Text

Final Written Examination with Answer Key (401-5 Format)
	ML100670463
Person / Time
Site:	Millstone
Issue date:	01/15/2010
From:	D'Antonio J; Operations Branch I
To:	; Dominion Nuclear Connecticut
	Hansell S
Shared Package
ML0924000035	List: ML100670463;
References
	50-336/10-301, ES-401, ES-401-8, TAC U01799, FOIA/PA-2011-0115 50-336/10-301
	Download: ML100670463 (76)
	v • d • e

{{#Wiki_filter:IANSWER KEYI ES-401 Site-Specific SRO Written Examination Form ES-401-8 Cover Sheet u.s. Nuclear Regulatory Commission Site-Specific SRO Written Examination Applicant Information Name: Atlswe(' ketl Date: !i1'/~'f / /(J I Facility/Unit: ~{H5t~nt;'U~:t ~ Region: I M II III IV Reactor Type: W...* CE IgJ BW []GE Start Time: Finish Time: Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination you must achieve a final grade of at least 80.00 percent overall, with 70.00 percent or better on the SRO-only items if given in conjunction with the RO exam; SRO-only exams given alone require a final grade of 80,00 percent to pass. You have 8 hours to complete the combined examination, and 3 hours if you are only taking the SRO portion, Applicant Certification All work done on this examination is my own, I have neither given nor received aid. AY1$Wt'V-g~t: Applicant s Signature Results R~~otal Examination Values _'_I a.,.'.J Points Applicant's Scores / a.5' Points Applicant's Grade ~I /~ Percent \\ANSWER KEyl ES-401. Page 31 of 33 IANSWER KEYI ES-401 Site-Specific SRO Written Examination Cover Sheet Form ES-401-8 u.s. Nuclear Regulatory Commission Site-Specific SRO Written Examination Applicant Information Name: Atlswe(' ketl Date: !i1'/~'f / /(J I Facility/Unit: ~{H5t~nt;'U~:t ~ Region: I M II III IV Reactor Type: W.. .* CE IgJ BW []GE Start Time: Finish Time: Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination you must achieve a final grade of at least 80.00 percent overall, with 70.00 percent or better on the SRO-only items if given in conjunction with the RO exam; SRO-only exams given alone require a final grade of 80,00 percent to pass. You have 8 hours to complete the combined examination, and 3 hours if you are only taking the SRO portion, Applicant Certification All work done on this examination is my own, I have neither given nor received aid. A Y1$Wt'V-g ~t: Applicant s Signature Results R~~otal Examination Values _'_I a.,.'.J Points Applicant's Scores / a. 5' Points Applicant's Grade ~I /~ Percent \\ANSWER KEyl ES-401. Page 31 of 33

SRO Exam Questions Only (No "Parents"Or "Odginals") Question #: 'f Question ID: 9000018 D RO ~ SRO D Student Handout? D Lower Order? l-SRO Ques. # Rev. 1 ~ Selected for Exam Origin: New D 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 568°F and stable - RCS pressure is 1210 psia and stable - Pressurizer level is 100%. - RWST level is 96% and slowly going down. - 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? D A Direct the Tectmical Support Center to develop a plan to restore RWST level. D B Direct the Balance of Plant Operator to align 24E to receive power from Unit 3. D C Direct the Reactor Operator to place the SI/CS Pump Miniflow switches in "OPERATE". [i'I D Direct the crew to commence a controlled cooldown and depressurization. Justification I o IS CORRECT; With ReS pressure stable at 1310 psi a 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 incorrel;t; JlJthough 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 bElcause step 8 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 required ONLY if the LOCA is outside of Containment. B is incorrect; Although the loss of 240 makes the "C" HPSI pump unavailable, the one available HPSI pump should be enough to mitigate the event, provided an RCS cooldown and depressurization is accomplished. Plausible the examinee may believe that starting a second HPSI pump is necessary to recover vessel level because the given conditions indicate that SI flow is presently inadequate (vessel level going down). C is incorrect; The SItCS Pump Miniflow switches are not placed in 'OPERATE" until RWST level is gO%. Plausible bElcause the examinee may feel that a Sump Recirc Actuation Signal is imminent; therefore, it would be appropriate to perform this step out of sequence. References I EOP 2532, "LOCA" and OP 2260, "EOP Users Guide" Comments and Question Modification History I NRC (comments on original question) - Distracter "8" does not relate to a Safety Function; Replace. Distracter "A" would be acceptable if an RWST level were added to the stem. RLC - In the stem: changed CET temp. from 578°F to 568°F, changed RCS pressure from 1310 psia to 1210 psia, and added "RWST level is 96% and slowly going down.". Also, changed choice "B" from realigning Condo Air Removal to aligning 24E to Unit 3. [12/30109) Bruce F. - D-3/C, No comment NRC KIA SystemtEtA System 009 Small Break LOCA 1---------------, Generic KIA Selected i NRC KIA Generic System 2.1 Conduct of Operations Page 1 of 75 Printed on 1/28/2010 at 12:02 I SRO Exam Questions Only (No "Parents"Or "Odginals") Question #: 'f Question ID: 9000018 D RO ~ SRO D Student Handout? D Lower Order? l-SRO Ques. # Rev. 1 ~ Selected for Exam Origin: New D 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 568°F and stable - RCS pressure is 1210 psia and stable - Pressurizer level is 100%. - RWST level is 96% and slowly going down. - 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? D A Direct the Tectmical Support Center to develop a plan to restore RWST level. D B Direct the Balance of Plant Operator to align 24E to receive power from Unit 3. D C Direct the Reactor Operator to place the SI/CS Pump Miniflow switches in "OPERATE". [i'I D Direct the crew to commence a controlled cooldown and depressurization. Justification I o IS CORRECT; With ReS pressure stable at 1310 psi a 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 incorrel;t; JlJthough 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 bElcause step 8 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 required ONLY if the LOCA is outside of Containment. B is incorrect; Although the loss of 240 makes the "C" HPSI pump unavailable, the one available HPSI pump should be enough to mitigate the event, provided an RCS cooldown and depressurization is accomplished. Plausible the examinee may believe that starting a second HPSI pump is necessary to recover vessel level because the given conditions indicate that SI flow is presently inadequate (vessel level going down). C is incorrect; The SItCS Pump Miniflow switches are not placed in 'OPERATE" until RWST level is gO%. Plausible bElcause the examinee may feel that a Sump Recirc Actuation Signal is imminent; therefore, it would be appropriate to perform this step out of sequence. References I EOP 2532, "LOCA" and OP 2260, "EOP Users Guide" Comments and Question Modification History I NRC (comments on original question) - Distracter "8" does not relate to a Safety Function; Replace. Distracter "A" would be acceptable if an RWST level were added to the stem. RLC - In the stem: changed CET temp. from 578°F to 568°F, changed RCS pressure from 1310 psia to 1210 psia, and added "RWST level is 96% and slowly going down.". Also, changed choice "B" from realigning Condo Air Removal to aligning 24E to Unit 3. [12/30109) Bruce F. - D-3/C, No comment NRC KIA SystemtEtA System 009 Small Break LOCA 1---------------, I Generic KIA Selected i NRC KIA Generic System 2.1 Conduct of Operations Page 1 of 75 Printed on 1/28/2010 at 12:02

Question #: 1 Ii'] SRO o Student Handout? o Lower Order? Question ID: 9000018 ORO J-SRO Ques. # Rev. 1 Ii'] Selected for Exam Origin: New o Past NRC Exam? Number 2.1.20 RO 4.6 SRO 4.6 CFRLink (CFR: 41.10/43.5/45.12) Ability to interpret and execute procedure steps. EOP 2532 Revision 029 Page 21 of95 l\\Ilillstone Unit 2 Loss of' Coolant Accident INSTR UCTIONS CONTINGENCY ACTIONS NOTE

1. Res cooldown should be initiated within one hour after the event to conserve condensate inventory and comply with the Long Term Cooling Analysis.

2. ReS cook/own rate greater than 40" Fihr should be maintained until the steam dump/bypass valves or atmospheric dump valves arc full open.

3. The starting point for the ReS cooldown should be the Tc or CET temperatures where ReS has stabilized.

4. Tcshould he used It.)r monitoring Res cooldown if in forced or natural circulation. CETs should be used for all other cases.

NOTE Technical Specilkation cooldown rates should he observed during the cooldown. The cooldown rates arc as follows: I. ReS Tc greater than 2200 F the cooldown rate is WOOF/hI'.

2. ReS Tc less than or equal to 2200 F th e cooldown ra te is 500 F/hr.

Perform Controlled Cool down

17JNITrATE a controlled cooldown using the steam dumps to establish shutdown cooling entry conditions.

17.1 INITIATE a controlled cooldown using the ADVs to establish shutdown cooling entt)' conditions. STOP THINK ACT REVIEW Page 2 of 75 Printed on 1/28/2010 at 12:02 Question #: 1 Question ID: 9000018 ORO Ii'] SRO o Student Handout? J-SRO Ques. # Rev. 1 Number 2.1.20 RO 4.6 SRO 4.6 Ability to interpret and execute procedure steps. l\\Ilillstone Unit 2 Loss of' Coolant Accident INSTR UCTIONS Ii'] Selected for Exam Origin: New CFRLink (CFR: 41.10/43.5/45.12) EOP 2532 Revision 029 Page 21 of95 CONTINGENCY ACTIONS NOTE

1.

Res cooldown should be initiated within one hour after the event to conserve condensate inventory and comply with the Long Term Cooling Analysis.

2.

ReS cook/own rate greater than 40" Fihr should be maintained until the steam dump/bypass valves or atmospheric dump valves arc full open.

3. The starting point for the ReS cooldown should be the Tc or CET temperatures where ReS has stabilized.

4.

T cshould he used It.)r monitoring Res cooldown if in forced or natural circulation. CETs should be used for all other cases. NOTE Technical Specilkation cooldown rates should he observed during the cooldown. The cooldown rates arc as follows: I. ReS Tc greater than 2200 F the cooldown rate is WOOF/hI'.

2.

ReS T c less than or equal to 2200 F th e cooldown ra te is 500 F/hr. Perform Controlled Cool down

17JNITrATE a controlled cooldown using the steam dumps to establish shutdown cooling entry conditions.

STOP THINK 17.1 INITIATE a controlled cooldown using the ADVs to establish shutdown cooling entt)' conditions. ACT REVIEW o Lower Order? o Past NRC Exam? Page 2 of 75 Printed on 1/28/2010 at 12:02

SRO Exam Questions Only (No "ParentsH.()r"Origillalstt) Question #: ., I Question ID: 9000018 ORO ~ SRO o Student Handout? o Lower Order? II-SRO Ques. # 1 Rev. 1 ~ Selected for Exam Origin: New o Past NRC Exam?

b. Procedural steps listed in alphanumeric order nre sequential steps and shall be addressed in that set1uence. Exceptions to this are as follows.

Asterisked steps, within the ORP or selected FRPs being implemented. may be brought forward to correct or preserve a Safety Function. Steps may be performed out of order after they have been accomplished once, if they are 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 piacekeeper when started. Since these steps will be performed repeatedly, the placekeeper is marked with a "cont," in the "Done" column. This signifies that the step is continuously performed and will not be completed during the performance of the EOP.

d. Bulleted lists arc 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 direct the performance of a task out of sequence, if the actions do not interfere with maintaining an existing Safety Function (e.g., transferring IXlwer supplies for "B" charging pump. preparing for power restoration).

1.9.2 Instructions and Contingent:), Actions

a. The EOPs are formatted with Instructions and Contingem,'v Actions. The instructions column presents the optimal~

method and sequence for accomplishing a specific task. 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 the Instructions column (left column).

OP 2260 I Level of Use] STOP THINK ACT REVIEW Rev. 009-03 InformatiOI2.,j 13 of 60 Page 3 of 75 Printed on 1/28/2010 at 12:02 Question #: II-SRO Ques. # 1 SRO Exam Questions Only (No "ParentsH.()r"Origillalstt) I Question ID: 9000018 ORO ~ SRO Rev. 1 ~ Selected for Exam o Student Handout? Origin: New o Lower Order? o Past NRC Exam?

b.

Procedural steps listed in alphanumeric order nre sequential steps and shall be addressed in that set1uence. Exceptions to this are as follows. Asterisked steps, within the ORP or selected FRPs being implemented. may be brought forward to correct or preserve a Safety Function. Steps may be performed out of order after they have been accomplished once, if they are 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 acekeeper when started. Since these steps will be performed repeatedly, the placekeeper is marked with a "cont," in the "Done" column. This signifies that the step is continuously performed and will not be completed during the performance of the EOP.

d.

Bulleted lists arc 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 direct the performance of a task out of sequence, if the actions do not interfere with maintaining an existing Safety Function (e.g., transferring IXlwer supplies for "B" charging pump. preparing for power restoration ). 1.9.2 Instructions and Contingent:), Actions

a.

The EOPs are formatted with Instructions and Contingem,'v Actions. The instructions column presents the optimal~ method and sequence for accomplishing a specific task. 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 the Instructions column (left column).

I Level of Use] InformatiOI2.,j STOP THINK ACT OP 2260 REVIEW Rev. 009-03 13 of 60 Page 3 of 75 Printed on 1/28/2010 at 12:02

Question #: 2 Student Handout? Lower Order? Question 10; 9000019 RO ~ SRO I-SRO Ques. Ii 2 Rev. 1 Selected for Exam Origin: New Past NRC Exam? The plant tripped from 100% power due to a Large Break LOCA. The crew successfully completed all actions of EOP 25~~5, Standard Post Trip Actions, and are presently performing EOP 2532, Loss of Coolant Accident. The following conditions exist approximately 2 hours 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 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 precipitation is beginning to occur causing intermittent blockage of flow channels. 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 Glogging 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 hours for Iodine scrubbing; however, core cooling (maintaining adequate SI flow to the core) takes precedence over Iodine scrubbing. A is incorrect; The Contclinment Spray System and CAR Coolers are designed to lower containment sump temperature enough to prevent cavitation of the !-IPSI Pumps. Plausible because the SFIAS 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 post event. It's NOT likely that Boron would be solidifying in the core at this time and blocking core flow. Plausible because accidE'nt analysis shows boron Precipitation will occur after a large break LOCA. The examinee may NOT remember the time frame for Boron Precipitation (8-10 hours after the LOCAl. Simultaneous Hot Leg and Cold Leg Injection is the appropriate action for Boron Precipitation. o is incorrect; Total Safe*ty 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 Questicln Modification History I NRC - (original question comment) Distracter "C" poor. Reword to "Boron precip. is beginning to occur causing intermittent blockage flow channels.... RLC Reworde,j Choice "cn per NRC comments. Bruce F. - D-4/C. No comment NRC KIA System/E/A System 011 Large Break LOCA Number EA2.10 RO 4.5 SRO 4.7 CFR Link (CFR 43.5/45.13) Ability to determine or interpret the following as they apply to a Large Break LOCA: Verification of adequate core cooling Page 4 of 75 Printed on 1/28/2010 at 12:02 Question #: 2 Question 10; 9000019 RO ~ SRO Student Handout? Lower Order? I-SRO Ques. Ii 2 Rev. 1 Selected for Exam Origin: New Past NRC Exam? The plant tripped from 100% power due to a Large Break LOCA. The crew successfully completed all actions of EOP 25~~5, Standard Post Trip Actions, and are presently performing EOP 2532, Loss of Coolant Accident. The following conditions exist approximately 2 hours 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 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 precipitation is beginning to occur causing intermittent blockage of flow channels. 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 Glogging 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 hours for Iodine scrubbing; however, core cooling (maintaining adequate SI flow to the core) takes precedence over Iodine scrubbing. A is incorrect; The Contclinment Spray System and CAR Coolers are designed to lower containment sump temperature enough to prevent cavitation of the !-IPSI Pumps. Plausible because the SFIAS 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 post-event. It's NOT likely that Boron would be solidifying in the core at this time and blocking core flow. Plausible because accidE'nt analysis shows boron Precipitation will occur after a large break LOCA. The examinee may NOT remember the time frame for Boron Precipitation (8-10 hours after the LOCAl. Simultaneous Hot Leg and Cold Leg Injection is the appropriate action for Boron Precipitation. o is incorrect; Total Safe*ty 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 Questicln Modification History I NRC - (original question comment) Distracter "C" poor. Reword to "Boron precip. is beginning to occur causing intermittent blockage flow channels.... RLC Reworde,j Choice "cn per NRC comments. Bruce F. - D-4/C. No comment NRC KIA System/E/A System 011 Large Break LOCA Number EA2.10 RO 4.5 SRO 4.7 CFR Link (CFR 43.5/45.13) Ability to determine or interpret the following as they apply to a Large Break LOCA: Verification of adequate core cooling Page 4 of 75 Printed on 1/28/2010 at 12:02

Question #: 2 Student Handout? Lower Order? Question ID: 9000019 0 RO ~ SRO II-SRO Ques. # 2 Rev. 1 ~ Selected for Exam Orlgin: New Past NRC Exam? Millstone Unit 2 Loss of Coolant Accident EOP 2532 Revision 029 Page 42 of95 INSTRUCTIONS CONTINGENCY ACTIONS NOTE Degradation in HPSI pump performance, post SRAS. may be indicative of debris fouling the LI~MT sump ~,crcen. Checking HPSI pump flow greater than 30 gpm ensures minimum now requircmcnL,<; are met [or pump protection when Res pressure is high and prohibiting 11m\\,. This presents differently than the sump blockage issue. "PSI Pump Post SI1:AS Performance Criteria

5IJ.JF SRAS has ac11lated, CHECK for adc(juate HPSI flow hy observing ALL of the following:

Flow greal:~r than or equal to 30 gpm for each operating pump Motor current stable Stable HPSI pump discharge presl"mre (continue) STOP THINK 50.1 IF unable to maintain HPSI 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 stahle Stahle HPSI pump discharge pressure 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 now:

a. IE CTMT pressure can be maintained less than 54 psig. AND at least ONE complete facility of CAR fans is operating. STOP CS pl1mps.

(conlin l1c) ACT REVIEW Page 5 of 75 Printed on 1/28/2010 at 12:02 Question #: 2 Question ID: 9000019 0 RO ~ SRO Student Handout? II-SRO Ques. # 2 Rev. 1 ~ Selected for Exam Orlgin: New Millstone Unit 2 Loss of Coolant Accident INSTRUCTIONS EOP 2532 Revision 029 Page 42 of 95 CONTINGENCY ACTIONS NOTE Degradation in HPSI pump performance, post SRAS. may be indicative of debris fouling the LI~MT sump ~,crcen. Checking HPSI pump flow greater than 30 gpm ensures minimum now requircmcnL,<; are met [or pump protection when Res pressure is high and prohibiting 11m\\,. This presents differently than the sump blockage issue. "PSI Pump Post SI1:AS Performance Criteria

5IJ.JF SRAS has ac11lated, CHECK for adc(juate HPSI flow hy observing ALL of the following:

Flow greal:~r than or equal to 30 gpm for each operating pump Motor current stable Stable HPSI pump discharge presl"mre (continue) STOP THINK 50.1 IF unable to maintain HPSI 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 stahle Stahle HPSI pump discharge pressure 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 now:

a. IE CTMT pressure can be maintained less than 54 psig. AND at least ONE complete facility of CAR fans is operating. STOP CS pl1mps.

(conlin l1c) ACT REVIEW Lower Order? Past NRC Exam? Page 5 of 75 Printed on 1/28/2010 at 12:02

Question #: .3 Question ID; 9000003 RO ~ SRO Student Handout? Lower Order? J-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 "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 F'ressure Controller, PIC-215, has malfunctioned. Using the Foxboro Controller, restore "A" RCP Bleedoff pressure and flow to the normal band. o The Rep Bleedoff Relief Valve has inadvertently opened. Evaluate "A" RCP Seal pressures to determine whether or not the "A" RCP may remain in operation. Justification 'I A is CORRECT; The RCP A UPPER SEAL PRES HI annunciator is indicative of a failure of the "AU RCP Middle or Lower Seal (or a combination of both). This resulted the a high Bleedoff flow through the "An 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 HCP 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 "A" RCP Middle Seal; however, the indications together would indicate a loss of Bleedoff flow through the "An RCP seals requiring the RCP to be stopped. If a Middle Seal had f<liled, 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 alarrrs 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 Controlier 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, "A" RCP Seal pressures are dependent on seal conditions and not necessarily the flow path of Bleedoff. Plausible because the ex,aminee 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 ARP 2590B-068 Comments and QUestion Modification History Bob K. - D-3/C Bill M. - D-3/C, 50150 Angelo - D-4/C; No comments. Bruce F. D-3/C, No comment NRC KIA System/E:JA System 015 Reactor Coolant Pump Malfunctions KIA NRC KIA Generic System 2.1 Conduct of Operations Number 2.1.7 RO 4.4 SRO 4.7 CFR Link (CFR: 41.5 I 43.5 145.12/45.13) Ability to evaluate plant performance and make operational judgements based on operating characteristics, reactor behavior, and instrument interpretation. Page 6 of75 Printed on 1/28/2010 at 12:02 Question #: .3 Question ID; 9000003 RO ~ SRO Student Handout? Lower Order? J-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 "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 F'ressure Controller, PIC-215, has malfunctioned. Using the Foxboro Controller, restore "A" RCP Bleedoff pressure and flow to the normal band. o The Rep Bleedoff Relief Valve has inadvertently opened. Evaluate "A" RCP Seal pressures to determine whether or not the "A" RCP may remain in operation. Justification 'I A is CORRECT; The RCP A UPPER SEAL PRES HI annunciator is indicative of a failure of the "AU RCP Middle or Lower Seal (or a combination of both). This resulted the a high Bleedoff flow through the "An 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 HCP 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 "A" RCP Middle Seal; however, the indications together would indicate a loss of Bleedoff flow through the "An RCP seals requiring the RCP to be stopped. If a Middle Seal had f<liled, 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 alarrrs 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 Controlier 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, "A" RCP Seal pressures are dependent on seal conditions and not necessarily the flow path of Bleedoff. Plausible because the ex,aminee 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 ARP 2590B-068 Comments and QUestion Modification History Bob K. - D-3/C Bill M. - D-3/C, 50150 Angelo - D-4/C; No comments. Bruce F. D-3/C, No comment NRC KIA System/E:JA System 015 Reactor Coolant Pump Malfunctions KIA NRC KIA Generic System 2.1 Conduct of Operations Number 2.1.7 RO 4.4 SRO 4.7 CFR Link (CFR: 41.5 I 43.5 145.12/45.13) Ability to evaluate plant performance and make operational judgements based on operating characteristics, reactor behavior, and instrument interpretation. Page 6 of75 Printed on 1/28/2010 at 12:02

SRO Exam Questions Only (No ftPar~nts"()r "Originals") Question #: 3 o Student Handout? o Lower Order? Question ID: 9000003 ORO ~ SRO II-SRO Ques. # 3 Rev. 0 ~ Selected for Exam Origin: New D Past NRC Exam? ApprOVlil Date 5/26/09 Effective Date 6/4109 Setpolnt: 2 gpm I 88-17 I RCPABLEED OFF FLOW HI AUTOMATIC FUNCTIONS

1.

None NOTE This alarm may be indicative of seal stage failure. One seal failure can cause high bleedoff flow alarm. Operation may continue with this alarm present providing seal bJeedoff temperature is within limits and seal (Jifferential pressures indicates that only one of the three lower seal stages I failed. A vapor stage failure will require a plant trip. CORRECTIVE: ACTIONS NOTE If seal flow reaches 10 g[Jm. 'W' RCP controlled bleedoff excess flow check valve closes to prevent blockage of bleetloff /low from other RCPs.

5.

Refer To the following guidance and DETERMINE if '~' Rep controlled bleedoff excess flow check valve has closed: High bleedoff flow alarm annunciates and clears follO\\ved by low bleedoff flow abrm which remains lit St:al pressures rise on all 3 seals (provided vapor seal is intact)

6.

IF "/'1\\' Rep controlled blcedoff excess flow check valve has closed, PERFORM the following.: 6.1 TRIP reactor and turbine. 6.2 STOP "/\\.' Rep. 6.3 Go To EOP 2525, "Standard Post Trip Actions" and PERFORM required ac tiolls. L________________ Page 7 of 75 Printed on 1/28/2010 at 12:02 SRO Exam Questions Only (No ftPar~nts"()r "Originals") Question #: 3 Question ID: 9000003 ORO ~ SRO o Student Handout? o Lower Order? II-SRO Ques. # 3 Rev. 0 ~ Selected for Exam Origin: New D Past NRC Exam? ApprOVlil Date 5/26/09 Effective Date 6/4109 Setpolnt: 2 gpm AUTOMATIC FUNCTIONS

1.

None NOTE I 88-17 I RCPABLEED-OFF FLOW HI This alarm may be indicative of seal stage failure. One seal failure can cause high bleedoff flow alarm. Operation may continue with this alarm present providing seal bJeedoff temperature is within limits and seal (Jifferential pressures indicates that only one of the three lower seal stages I failed. A vapor stage failure will require a plant trip. CORRECTIVE: ACTIONS NOTE If seal flow reaches 10 g[Jm. 'W' RCP controlled bleedoff excess flow check valve closes to prevent blockage of bleetloff /low from other RCPs.

5.

Refer To the following guidance and DETERMINE if '~' Rep controlled bleedoff excess flow check valve has closed: High bleedoff flow alarm annunciates and clears follO\\ved by low bleedoff flow abrm which remains lit St:al pressures rise on all 3 seals (provided vapor seal is intact)

6.

IF "/'1\\' Rep controlled blcedoff excess flow check valve has closed, PERFORM the following.: 6.1 6.2 6.3 TRIP reactor and turbine. STOP "/\\.' Rep. Go To EOP 2525, "Standard Post Trip Actions" and PERFORM required ac tiolls. L _______________ _ Page 7 of 75 Printed on 1/28/2010 at 12:02

SRO Exam Questions OnIY(No"Pirents" Or UOriginalsU) Question #: 4 Question ID: 9000004 D RO ~ SRO D Student Handout? D Lower Order? J-SRO Ques. # 4 Rev. 0 ~ Selected for Exam Origin: New D Past NRC Exam? The plant is operating at 100% power with the "s" Auxiliary Feedwater (AFW) Pump out of service for maintenance. Then the following 13vents 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? D 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 SOP 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 "s" Auxiliary Feedwater Regulating Valve, 2-FW-43S. This will prevent excessive auxiliary feedwater from overfilling the affected SG. D D Direct the SOP* 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 lo,~s 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. I n order to prevent overfeeding #2 SG, the "EI" 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 examinee 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; #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 I 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 Bruce F. - D-4/C, No comment NRC KIA SystemtEtA System 038 Steam Generator Tube Rupture (SGTR) I Generic KIA Selected] NRC KIA Generic System 2.1 Conduct of Operations Number 2.1.30 RO 4.4 SRO 4.0 CFR Link (CFR: 41.7/45.7) Ability to locate and operate components, including local controls. Page 8 of 75 Printed on 1/28/2010 at 12:02 SRO Exam Questions OnIY(No"Pirents" Or UOriginalsU) Question #: 4 Question ID: 9000004 D RO ~ SRO D Student Handout? D Lower Order? J-SRO Ques. # 4 Rev. 0 ~ Selected for Exam Origin: New D Past NRC Exam? The plant is operating at 100% power with the "s" Auxiliary Feedwater (AFW) Pump out of service for maintenance. Then the following 13vents 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? D 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 SOP 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 "s" Auxiliary Feedwater Regulating Valve, 2-FW-43S. This will prevent excessive auxiliary feedwater from overfilling the affected SG. D D Direct the SOP* 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 lo,~s 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. I n order to prevent overfeeding #2 SG, the "EI" 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 examinee 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; #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 I 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 Bruce F. - D-4/C, No comment NRC KIA SystemtEtA System 038 Steam Generator Tube Rupture (SGTR) I Generic KIA Selected] NRC KIA Generic System 2.1 Conduct of Operations Number 2.1.30 RO 4.4 SRO 4.0 CFR Link (CFR: 41.7/45.7) Ability to locate and operate components, including local controls. Page 8 of 75 Printed on 1/28/2010 at 12:02

i SRO Exam Questions Only (No "Parents" (}r "Originals") Question #: 4 Question ID; P-SRO Ques. # 4 Rev. [\\-fillstowe Unit 2 Loss of 120 VAC Vital 9000004 D RO ~ SRO D Student Handout? D Lower Order? 0 ~ Selected for Exam Origin; New D Past NRC Exam? AOP 2504D Rc\\isiun 003-07 Page 3 ur22 Instrument Panel '1j\\~20 [+] 1.0 PURPOSE 1.1 ()bj eeth*-e This procedure provides ilt.~tlUcti{)lt'; to be pClformed ulx)n the loss of 120 Volt AC Vilal Inslmmenl Panel VA-2O. 12 l>i.st'U s s ion The los, of VA-20 causes the Inss of many annunciators, indications, PPC inputs, intel"lock.~ and cOlllrol'>. Equipmefl I that affecls Unit operalion includes Ihe folklwing: FW-5ID. #2 SO FR\\'. fails as is on los>; of power, and closes when power is l'estmed FW41 B, #2 SO FRV bnm&<;, closes and control from C05 is lost "C" charging pump loses the plunger flush pump If channel "Y" is controlling pressurizer level, charging will go to maximum and letdown will go to minimum If pres.'<;urizer pressure control is sclected to channel "Y,"auto control of pressurizer sprays are 10M All pressurizer healers are deenergi:zed due 10 pressurizer level channel Y sending a pressurizer level low low heater culout "Il.:! restore pressurizer heaters, the healer select switch must be placed in the channel "X" poslition.

2 SO 'WfMOS DU\\fP MS 190B" control from {,-O5. C-l1 and C-10 is It'\\.~!.

Steam dump to condem;er pressure controller (PIC -4216) from C-05 is lost. Control from Foxboro computer is available.

2 Aux feed reg valve fails open and manual control from ('-05 is lost.

Page 9 of 75 Printed on 1/28/2010 at 12:02 SRO Exam Questions Only (No "Parents" (}r "Originals") Question #: 4 Question ID; 9000004 D RO ~ SRO D Student Handout? D Lower Order? P-SRO Ques. # 4 Rev. 0 ~ Selected for Exam Origin; New D Past NRC Exam? i [\\-fillstowe Unit 2 AOP 2504D Rc\\isiun 003-07 Loss of 120 VAC Vital Instrument Panel '1j\\~20 [+] Page 3 ur22 1.0 PURPOSE 1.1 () bj eeth*-e This procedure provides ilt.~tlUcti{)lt'; to be pClformed ulx)n the loss of 120 Volt AC Vilal Inslmmenl Panel VA-2O. 12 l>i.st'U s s ion The los, of VA-20 causes the Inss of many annunciators, indications, PPC inputs, intel"lock.~ and cOlllrol'>. Equ ipmefl I that affecls Unit operalion includes Ihe folklwing: FW-5ID. #2 SO FR\\'. fails as is on los>; of power, and closes when power is l'estmed FW41 B, #2 SO FRV bnm&<;, closes and control from C05 is lost "C" charging pump loses the plunger flush pump If channel "Y" is controlling pressurizer level, charging will go to maximum and letdown will go to minimum If pres.'<;urizer pressure control is sclected to channel "Y,"auto control of pressurizer sprays are 10M All pressurizer healers are deenergi:zed due 10 pressurizer level channel Y sending a pressurizer level low low heater culout "Il.:! restore pressurizer heaters, the healer select switch must be placed in the channel "X" poslition.

2 SO 'WfMOS DU\\fP MS 190B" control from {,-O5. C-l1 and C-10 is It'\\.~!.

Steam dump to condem;er pressure controller (PIC -4216) from C-05 is lost. Control from Foxboro computer is available.

2 Aux feed reg valve fails open and manual control from ('-05 is lost.

Page 9 of 75 Printed on 1/28/2010 at 12:02

SRO Exam Questions Only (No "Parents" Or "Originals") Question #: 4 Question ID; 9000004 D RO ~ SRO D Student Handout? D Lower Order? II-SRO Ques. # 4 Rev. 0 ~ Selected for Exam Origin: New D Past NRC Exam? r Sta Millstone Unit 2 ndard Post Trip Actions EOP 2525 Revision 023 PtlAC 16 of 26

6.

INSTRUCTIONS (o:mrinued)

c.

GIECK thal atleaSl one steam generator has BO'll) of the follo.:ving conditions met: 11;1ain feoowater Of' 'TWO auxiliary fredwater pumps are (~)erating to restore leve I 40 to 70%...

d. CHECK that RCS subcooling i,s greater than or equal to 30"F.

CONTINGENCY ACtIONS c.1 RESTORE levellO 40 lO 70S% in at least one steam generatOl' using ANY of the foll()\\ving: Main feemvater Motor-driven al.Lxiliary feemvater pump TDAF'W Pump. Refer 'Ill Appendix 6, uTDAFW Pump Normal Startup." TDAf'W Pump. Refer1h Appendix 7, "TDAF\\V Pump Abnormal Startup." d.l RESTORE steam generator level 40 to 70% by performing ONE of the follmving: FEED each unaffected steam generator greater than 300 gpm. FEED the least affected steam generator greater than 300 gpm. Page 10 of 75 Printed on 1/28/2010 at 12:02 SRO Exam Questions Only (No "Parents" Or "Originals") Question #: 4 Question ID; 9000004 D RO ~ SRO D Student Handout? D Lower Order? D Past NRC Exam? II-SRO Ques. # 4 Rev. 0 ~ Selected for Exam Origin: New r Millstone Unit 2 Standard Post Trip Actions EOP 2525 Revision 023

6.

INSTRUCTIONS (o:mrinued)

c.

GIECK thal atleaSl one steam generator has BO'll) of the follo.:ving conditions met: 11;1ain feoowater Of' 'TWO auxiliary fredwater pumps are (~)erating to restore leve I 40 to 70%...

d. CHECK that RCS subcooling i,s greater than or equal to 30"F.

Page 10 of 75 PtlAC 16 of 26 CONTINGENCY ACtIONS c.1 RESTORE levellO 40 lO 70S% in at least one steam generatOl' using ANY of the foll()\\ving: Main feemvater Motor-driven al.Lxiliary feemvater pump TDAF'W Pump. Refer 'Ill Appendix 6, uTDAFW Pump Normal Startup." TDAf'W Pump. Refer1h Appendix 7, "TDAF\\V Pump Abnormal Startup." d.l RESTORE steam generator level 40 to 70% by performing ONE of the follmving: FEED each unaffected steam generator greater than 300 gpm. FEED the least affected steam generator greater than 300 gpm. Printed on 1/28/2010 at 12:02

SR(lExam Questions Only (No "Parents"OJ.-.uOriginals") Question #: 4 Question ID: 9000004 0 RO ~ SRO o Student Handout? o Lower Order? II-SRO Ques. # Rev. 0 ~ Selected for Exam Origin: New o Past NRC Exam? Millstone Unit 2 Standard Post Trip Actions EOP 2525 Revision 023 Page 18 of26 '"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: Steam Plant Radiation Moniltors RM-5099, Steam Jet Air Ejector RM-4262, SG Blowdown RM-4299A and B, Main Steam Line 1 RM-4299C, Main Steam Line 2 (continu e) CONTINGENCY ACTIONS b.I IF feed is available to BOTH steam generators, THROITLE feed to the steam generator with the highest radiation readings to maintain level 40 to 45% bv performing ANY of the following;

1) OPERATE associated main feed reg bypass valve, FW-41A or FW-41B.

2) IF 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-43R (continue) Page 11 of 75 Printed on 1/28/2010 at 12:02 SR(lExam Questions Only (No "Parents"OJ.-.uOriginals") Question #: 4 Question ID: 9000004 0 RO ~ SRO o Student Handout? o Lower Order? o Past NRC Exam? II-SRO Ques. # Rev. 0 ~ Selected for Exam Origin: New 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: Steam Plant Radiation Moniltors RM-5099, Steam Jet Air Ejector RM-4262, SG Blowdown RM-4299A and B, Main Steam Line 1 RM -4299C, Main Steam Line 2 (con tinu e) Page 18 of26 CONTINGENCY ACTIONS b.I IF feed is available to BOTH steam generators, THROITLE feed to the steam generator with the highest radiation readings to maintain level 40 to 45% bv performing ANY of the following;

1) OPERATE associated main feed reg bypass valve, FW -41A or FW-41B.

2) IF 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-43R (continue) Page 11 of 75 Printed on 1/28/2010 at 12:02

SRO Exam Questions Only(Nq'"'Parents" Or "Originals") Question #: 4 Question ID: 9000004 0 RO Ii'J SRO o Student Handout? o Lower Ordelr? J-SRO Ques. # 4 Rev. 0 ~ Selected for Exam Origin: New o Past NRC Exam? EOP 2525, Standard Millstone Unit 2 Post Trip Actions Emergency Operating Procedure Technical Guide Page~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 ttle maximum expected normal containment pressure, high radiation inside or outside containment, or tile 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 Vliith 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 allows 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 Con1ingency 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. ~___"_____"___"___._____._______________._____________I Page 12 of 75 Printed on 1/28/2010 at 12:02 SRO Exam Questions Only(Nq'"'Parents" Or "Originals") Question #: 4 Question ID: 9000004 0 RO Ii'J SRO J-SRO Ques. # 4 Rev. 0 ~ Selected for Exam Millstone Unit 2 Emergency Operating Procedure Technical Guide o Student Handout? Origin: New o Lower Ordelr? o Past NRC Exam? EOP 2525, Standard Post Trip Actions Page~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 ttle maximum expected normal containment pressure, high radiation inside or outside containment, or tile 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 Vliith 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 allows 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 Con1ingency 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. ~ I Page 12 of 75 Printed on 1/28/2010 at 12:02

Question #: 5 Question ID: 9082581 D RO ~ SRO D Student Handout? ~ Lower Order? I-SRO Ques. # S Rev. 0 l~ Selected for Exam Origin: New D Past NRC Exam? The plant has experienced a loss of VA-1 0 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 action::; would be performed outside the Control Room and what is the reason for performing these actions in the listed order?

- - - - n ******************** III ******************************************

A 1. Place 2-SI-~106. 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-£)57. SOC Heat Exchanger Flow Control Valve. in manual and turn the handwheel in the clockwise direction as directed by the Control Room.

2-S1-306, SOC Total Flow Control Valve. must be opened first to provide minimum flow for the operating LPSI Pump.

1. Place 2-SI-£)57. SOC Heat Exchanger Flow Control Valve. in manual and turn the handwheel in the

~B clockwise direction as directed by the Control Room.

2. Place 2-SI<106. 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 ReS cooldown rate.

1. Place 2-SI-{i57. SOC Heat Exchanger Flow Control Valve. in manual and turn the handwheel in the DC counterclockwise direction as directed by the Control Room.

2. Place 2-SI-a06, SOC Total Flow Control Valve. in manual and turn the handwheel in the clockwise direction as dir'3cted by the Control Room.

2-SI-657. SOC Heat Exchanger Flow Control Valve, must be opened first to establish the desired ReS 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 2e*72, Loss of SDC. requires the PEO to first place 51-657 in local manual control to restore cooling to the RC5. 51-657 is a reverse operating valve (i.e., clockwise rotation is open. counterclockwise is close). On a loss of power (VA-10). 51-657 fails closed; therefore, 51-657 must be rotated in the clockwise direction to open it. 51-306 is also a reverse operated valve. 51-306 fails open on a loss of power (\\lA-10), and must be rotated in the counterclockwise direction to throttle it closed. 51-306 is operated last to allow more flow through the SDC Heat Exchangers, if necessary to provide additional cooling flow to the RC5. A is incorrect; The order of local valve operations is incorrect. Although 51-657 fails closed, 51-306 fails open; therefore, there is no need to establish minimum flow protection for the running LPSI Pump. Plausible because the eXlaminee 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 SDC 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. D 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 J AOP 2572 Loss of 5DC, 8ection 8.0 Comments and Question Modification History" I Bob K. - D-3/W (memory on reverse acting) Bill M. - D-:,/C, 50/50 Angelo D-3/C; No comments. Bruce F. - D-5/C, Is it important to remember that these valves are revers operating? Page 13 of75 Printed on 1/28/2010 at 12:02 Question #: 5 Question ID: 9082581 D RO ~ SRO D Student Handout? ~ Lower Order? I-SRO Ques. # S Rev. 0 l~ Selected for Exam Origin: New D Past NRC Exam? The plant has experienced a loss of VA-1 0 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 action::; would be performed outside the Control Room and what is the reason for performing these actions in the listed order?

- - - - n ******************** III ******************************************

A 1. Place 2-SI-~106. SOC Total Flow Control Valve. in manual and turn the handwheel in the counterclockwise direction as directed by the Control Room. ~B DC

2. Place 2-SI-£)57. SOC Heat Exchanger Flow Control Valve. in manual and turn the handwheel in the clockwise direction as directed by the Control Room.

2-S1-306, SOC Total Flow Control Valve. must be opened first to provide minimum flow for the operating LPSI Pump.

1. Place 2-SI-£)57. 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<106. 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 ReS cooldown rate.

1. Place 2-SI-{i57. 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-a06, SOC Total Flow Control Valve. in manual and turn the handwheel in the clockwise direction as dir'3cted by the Control Room.

2-SI-657. SOC Heat Exchanger Flow Control Valve, must be opened first to establish the desired ReS 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 2e*72, Loss of SDC. requires the PEO to first place 51-657 in local manual control to restore cooling to the RC5. 51-657 is a reverse operating valve (i.e., clockwise rotation is open. counterclockwise is close). On a loss of power (VA-10). 51-657 fails closed; therefore, 51-657 must be rotated in the clockwise direction to open it. 51-306 is also a reverse operated valve. 51-306 fails open on a loss of power (\\lA-10), and must be rotated in the counterclockwise direction to throttle it closed. 51-306 is operated last to allow more flow through the SDC Heat Exchangers, if necessary to provide additional cooling flow to the RC5. A is incorrect; The order of local valve operations is incorrect. Although 51-657 fails closed, 51-306 fails open; therefore, there is no need to establish minimum flow protection for the running LPSI Pump. Plausible because the eXlaminee 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 SDC 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. D 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 J AOP 2572 Loss of 5DC, 8ection 8.0 Comments and Question Modification History" I Bob K. - D-3/W (memory on reverse acting) Bill M. - D-:,/C, 50/50 Angelo D-3/C; No comments. Bruce F. - D-5/C, Is it important to remember that these valves are revers operating? Page 13 of75 Printed on 1/28/2010 at 12:02

Question #: 5 IQuestion ID; 9082581 RO ~SRO o Student Handout? ~ Lower Order? J-SRO Ques. # Rev. 0 ~ Selected for Exam Origin: New Past NRC Exam? NRC KIA System/EiA System 057 Loss of Vital AC Electrical Instrument Bus KIA Selected 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 75 Printed on 1/2812010 at 12:02 Question #: 5 IQuestion ID; 9082581 RO ~SRO o Student Handout? ~ Lower Order? J-SRO Ques. # Rev. 0 ~ Selected for Exam Origin: New Past NRC Exam? NRC KIA System/EiA System 057 Loss of Vital AC Electrical Instrument Bus KIA Selected 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 75 Printed on 1/2812010 at 12:02

Question #: IS 1 Question ID: 9082581 RO ~SRO ::J Student Handout? Lower Order? P-SRO Ques. # 5 Rev. 0 ~1 Selected for Exam Origin: New

J Past NRC Exam?

l\\:lillstone Unit 2 AOP 2572 Reyision 009-o.t Loss of Shutdown Cooling PURe 45 of70 8.0 Loss u(Puwer or Air to SI*657, SI*3060r Bolh INSTRUCTIONS CONTINGENCY ACTIONS NOTE 1.. Loss of power or air 10 SDC flow C'Ontrol valves has the fol\\ov.lng affect: SI.{j~i7 fails closed SI-306 faill. open to its limil slop (mid-position)

2. Lo,,;; of VA-lO fails both valves.

3. SI-J06 can be Ihronlcd from its maximum open position (ma'l:imum flow lim it stn~l position) when diverting ackhlional fklw through SHe heat exchangers is I'equircd.

4. Obtaininn reference posilionsof SDC 110w control valves maybe helpful during SOC rcstoralion.lf a loss of VA-to has occurred, the refercll\\.':C positions are onlv available as archive data in the ppe (PPC analog pOints 251657 and 1s13(6).

OBSERVE applicable controllers Of PPC analog !loinlS to obtain a i reference prn.il ion for SIX flow colUml valves: Output of FIC-306m archive PF'C data for 251306 Output of mC-3657 or archive PPC data fm' 251657 For Ihe failed '!alve(s), ADJUST the COnlroller output to match actual valve position. FlC-306 HIC-3657 Page 15 of 75 Printed on 1/28/2010 at 12:02 Question #: IS 1 Question ID: 9082581 RO ~SRO ::J Student Handout? P-SRO Ques. # 5 Rev. 0 ~1 Selected for Exam Origin: New i l\\:lillstone Unit 2 Loss of Shutdown Cooling AOP 2572 Reyision 009-o.t PURe 45 of70 8.0 Loss u(Puwer or Air to SI*657, SI*3060r Bolh INSTRUCTIONS CONTINGENCY ACTIONS NOTE 1.. Loss of power or air 10 SDC flow C'Ontrol valves has the fol\\ov.lng affect: SI.{j~i7 fails closed SI-306 faill. open to its limil slop (mid-position)

2.

Lo,,;; of VA-lO fails both valves.

3. SI-J06 can be Ihronlcd from its maximum open position (ma'l:imum flow lim it stn~l position) when diverting ackhlional fklw through SHe heat exchangers is I'equircd.

4. Obtaininn reference posilionsof SDC 110w control valves maybe helpful during SOC rcstoralion.lf a loss of VA-to has occurred, the refercll\\.':C positions are onlv available as archive data in the ppe (PPC analog pOints 251657 and 1s13(6).

OBSERVE applicable controllers Of PPC analog !loinlS to obtain a reference prn.il ion for SIX flow colUml valves: Output of FIC-306m archive PF'C data for 251306 Output of mC-3657 or archive PPC data fm' 251657 For Ihe failed '!alve(s), ADJUST the COnlroller output to match actual valve position. FlC-306 HIC-3657 Lower Order?

J Past NRC Exam?

Page 15 of 75 Printed on 1/28/2010 at 12:02

I 'SRO Exam Questions Only (No"Plirents" Or "Originals") Question #: 5 Question ID; 9082581 D RO ~ SRO D Student Handout? ~ Lower Order? II-SRO Ques. # 5 Rev. 0 ~ Selected for Exam Origin; New D Past NRC Exam? Millstone Unit 2 AOP 2572 Revision 009-04 Loss of Shutdlown Cooling I:.agc 49 or 70 JNSTRtJCTIONS CONTINGENCY ACTIONS CAUTION Care should be used when establishing SOC heat exchanger tlowdue to the tiltenrial for water in the SOC heat exchanger to be much cooler than RC remperature. Initiaring flow slr...vly allows temperatures to equalize. NOTE

l. SI-657 is !l reverse-operating valve; colmterdockwise rotation of rhe handwhed closes Ihe valve and d(){:kwi.w rotation of the handwheel npens the: valve.

1. WherI establishing ReS cooldtl""n rate, optimum temperature respoJt.<;.e is adlie\\'Cd by maintaining SI-657 ber.veen 35 and 6()%, open.

R.8 IF SI*{i57I lad a In,s of pl1wer or air, PERFOR.M the following: 8.. ~esillblishing coolclown, Refer to SP 26mB. "1hlllsient 'fempemtme, Prc&<;ure VcrifiGltion:' and PERFOru.1 the fol km-'ing:

MONITOR ReS conldown fa te ming T351 Y.
ENSURE system response is within cooldown limil~.

b. PERFORM the follo\\\\-'ing to take manual control of SI-657

(,-P;' ESF Room): 1 ) UNLOCK rhe mallual Itandwheel on valve.

2) CL05,E instrument air supply vaIve.and V ENT valve operator.

3} LOOSEN stem hex nut as required to allow stCIl! IIIovcment.

4) ROT,~fE SI*657 hRndwheel a5 directed bv tIlC Control Room.

(CI)ntinue) Page 16 of 75 Printed on 1/28/2010 at 12:02 'SRO Exam Questions Only (No"Plirents" Or "Originals") Question #: 5 Question ID; 9082581 D RO ~ SRO D Student Handout? ~ Lower Order? II-SRO Ques. # 5 Rev. 0 ~ Selected for Exam Origin; New D Past NRC Exam? I Millst AOP 2572 Revision 009-04 Loss of Sh one Unit 2 utdlown Cooling I:.agc 49 or 70 R.8 JNS TRtJCTIONS CONTINGENCY ACTIONS CAUTION Care sho uld be used when establishing SOC heat exchanger tlowdue to rial for water in the SOC heat exchanger to be much cooler than perature. Initiaring flow slr... vly allows temperatures to equalize. the til ten RC rem

l.

1.

NOTE SI-65 7 is !l reverse-operating valve; colmterdockwise rotation of rhe whed closes Ihe valve and d(){:kwi.w rotation of the handwheel s the: valve. hand npen Wher I establishing ReS cooldtl""n rate, optimum temperature respoJt.<;.e is \\'Cd by maintaining SI-657 ber.veen 35 and 6()%, open. adlie IF SI*{i57I lad a In,s of pl1wer or air, M the following: PERFOR. 8..

b.

~esillblishing coolclown, Refer to SP 26mB. "1hlllsient mtme, Prc&<;ure VcrifiGltion:' 'fempe and P ERFOru.1 the fol km-'ing: M ONITOR ReS conldown te ming T351 Y. fa E w NSURE system response is ithin cooldown limil~. PERF ORM the follo\\\\-'ing to take al control of SI-657 manu (,-P;' E SF Room): 1 )

2) 3}

4)

U NLOCK rhe mallual andwheel on valve. It C L05,E instrument air supply Ive.and va V ENT valve operator. L OOSEN stem hex nut as quired to allow stCIl! ovc men t. re III R OT,~fE SI*657 Rndwheel a5 directed bv lC Control Room. h tI (CI )ntinue) Page 16 of 75 Printed on 1/28/2010 at 12:02

Question #: Student Handout? ~ Lower Order? ,Question ID: 9082581 RO ~ SRO II-SRO Ques. # 5 Rev. o ~ Selected for Exam Origin: New Past NRC Exam? 1\\lillstoDII! Unit 2 AOP 2572 Rc\\ision 009-04 Loss of Shutdown Cooling Page 52 uf70 ,~_____________________________-i________________________________ INSTRUCTIONS CONTINGENCY ACnONS NOTE SI-30'l i~ a re\\/erse~{:tpcfating valve: cocmlc'rr:lockwi!;t;' mtation of the handwhcel closes lhe valve and clockwise mtatian of lite handwheel opens the valve. _8.9 As necessary, ESTABLISH local manual control of S{*306 as follows:

a.

ESit\\BLlSH communicalil)lts between ODe rators al the valve r;A:' ESF l~oom) and Control Room. b, CIDSE u\\!,!rumcnt air supply for S1-306, sec lolal flowennl!'ol.

c. OPEN pelcock on ins[rumenl air supply pre,sure regulator and VENT SI-306, SOC tObl1 fklW control, valve operator.

d.

UNLOCK and REMOVE chain from manllal handwlll!cl. c, ROTATE manual handwhecl cotllltercic>c:kl'd.,e and I ALIGN holes in OUlel' shaft with hole in inl1er shaft.

f.

INSERT tlile pin into shaft holes. g, ENSURE 51-306, SDC total flow emami, valve position indicator on tile manual actuator is at 1l!I'oUIed ("pen Pl)sitioll.

h. IE desis'ed to manually operate valve. POSITION S1-306 handwhecil as directed by the i

Control Room. _&.10 WIlEN recovery fmm manual operations is desired. PERFORM aetions specified by the 8M/US. Page 17 of75 Printed on 1/28/2010 at 12:02 Question #: ,Question ID: 9082581 RO ~ SRO Student Handout? II-SRO Ques. # 5 Rev. o 1\\lillstoDII! Unit 2 Loss of Shutdown Cooling ~ Selected for Exam Origin: AOP 2572 Rc\\ision 009-04 Page 52 uf70 New ,~ _____________________________ -i ______________________________ __ INSTRUCTIONS CONTINGENCY ACnONS NOTE SI-30'l i~ a re\\/erse~{:tpcfating valve: cocmlc'rr:lockwi!;t;' mtation of the handwhcel closes lhe valve and clockwise mtatian of lite handwheel opens the valve. _8.9 As necessary, ESTABLISH local manual control of S{*306 as follows:

a.

ESit\\BLlSH communicalil)lts between ODe rators al the valve r;A:' ESF l~oom) and Control Room. b, CIDSE u\\!,!rumcnt air supply for S1-306, sec lolal flowennl!'ol.

c. OPEN pelcock on ins[rumenl air supply pre,sure regulator and VENT SI-306, SOC tObl1 fklW control, valve operator.

d.

UNLOCK and REMOVE chain from manllal handwlll!cl. c, ROTATE manual handwhecl cotllltercic>c:kl'd.,e and I ALIGN holes in OUlel' shaft with hole in inl1er shaft.

f.

INSERT tlile pin into shaft holes. g, ENSURE 51-306, SDC total flow emami, valve position indicator on tile manual actuator is at 1l!I'oUIed ("pen Pl)sitioll.

h. IE desis'ed to manually operate valve. POSITION S1-306 handwhecil as directed by the i

Control Room. _&.10 WIlEN recovery fmm manual operations is desired. PERFORM aetions specified by the 8M/US. ~ Lower Order? Past NRC Exam? Page 17 of75 Printed on 1/28/2010 at 12:02

Question #: 6 l',;lueslion ID: 3100002 RO IY'J SRO Student Handout? IY'J Lower Order? I-SRO Ques. If 6 Rev. 2 IY'J 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 liD" Instrument Air Dryer After Filters. Assuming Instrument Air header pressure continues to lower, at what pressure in the Instrument Air System must the Unit Supelvisor (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 approximately 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. ['ll] 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. [J 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 I nstnument Air is acceptable. Plausible because 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 because 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 Bob K.* D-2/C (only need to know "80#") Changed pressure values on "A", "B" and "0" and reworded "An and "Bn slightly.* RLC Bill M. - D-2!C, K (90 psig is too high.) Changed values In dlstmctors "An and "B" to 85 psig vs 90 pslg. More plausible.* RJA Angelo - D-:!/C; No comments. Bruce F. - D*3/C, No comment NRC KIA System/ElA 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 Instnument Air: When to trip reactor if instrument air pressure is decreasing Page 18 of 75 Printed on 1/28/2010 at 12:02 Question #: 6 l',;lueslion ID: 3100002 RO IY'J SRO Student Handout? IY'J Lower Order? I-SRO Ques. If 6 Rev. 2 IY'J 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 liD" Instrument Air Dryer After Filters. Assuming Instrument Air header pressure continues to lower, at what pressure in the Instrument Air System must the Unit Supelvisor (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 approximately 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. ['ll] 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. [J 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 I nstnument Air is acceptable. Plausible because 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 because 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 Bob K.* D-2/C (only need to know "80#") Changed pressure values on "A", "B" and "0" and reworded "An and "Bn slightly.* RLC Bill M. - D-2!C, K (90 psig is too high.) Changed values In dlstmctors "An and "B" to 85 psig vs 90 pslg. More plausible.* RJA Angelo - D-:!/C; No comments. Bruce F. - D*3/C, No comment NRC KIA System/ElA 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 Instnument Air: When to trip reactor if instrument air pressure is decreasing Page 18 of 75 Printed on 1/28/2010 at 12:02

'.t 0*'. r Question #: 6 IQueStion ID; 3100002 0 RO ~ SRO Student Handout? ~ Lower Order? II-SRO Ques. # 6 Rev. 2 .". Selected for Exam Origin: Bank Past NRC Exam? Millstone Unit 2 AOP 2563 Revision 009-07 I~oss of Instrument Air Page 3 of32 1.0 PURPOSE 1..1 O~jccthte This procedure provides the operator with specific steps to be taken follmving a significant drop in instrument air pressure and actions to mitigate the effects of a reactor trip when the Instrument Air System is flot available. 1.1 Discussion This procedure is implemented when instmment air pressure is lowering below normal values. With a complete loss of instrument air, continued steady state plant operation is not possible. The loss of many important controls, such as in the feedwater system, could degrade ph,lnt conditions at the time of a reactor or turbine trip. Thcrdore. the reactor is tripped immediately when instmment air pressure lowers to the point where control of important system., is questionable. This I may he indicated by system response or instmmcnt air header pressure of less than ;~o psig. Should instrument air header pressure drop suddenly, as in the case of a main header rupture, the only initial means of decay and sensible heat removal is the main steam safeties. Subsequently, manual control of the atmospheric dump valves and use of the Auxiliary Fecdwater System can mitigate the pressure transient, thereby removing sensible and decay heat from the ReS in a controUed manner. Page 19 of 75 Printed on 1/28/2010 at 12:02 Question #: 6 IQueStion ID; 3100002 0 RO ~ SRO '.t 0*'. r Student Handout? ~ Lower Order? II-SRO Ques. # 6 Rev. 2 .". Selected for Exam Origin: Bank Past NRC Exam? I Millstone Unit 2 I~oss of Instrument Air 1.0 PURPOSE 1..1 O~jccthte AOP 2563 Revision 009-07 Page 3 of 32 This procedure provides the operator with specific steps to be taken follmving a significant drop in instrument air pressure and actions to mitigate the effects of a reactor trip when the Instrument Air System is flot available. 1.1 Discussion This procedure is implemented when instmment air pressure is lowering below normal values. With a complete loss of instrument air, continued steady state plant operation is not possible. The loss of many important controls, such as in the feedwater system, could degrade ph,lnt conditions at the time of a reactor or turbine trip. Thcrdore. the reactor is tripped immediately when instmment air pressure lowers to the point where control of important system., is questionable. This may he indicated by system response or instmmcnt air header pressure of less than ;~o psig. Should instrument air header pressure drop suddenly, as in the case of a main header rupture, the only initial means of decay and sensible heat removal is the main steam safeties. Subsequently, manual control of the atmospheric dump valves and use of the Auxiliary Fecdwater System can mitigate the pressure transient, thereby removing sensible and decay heat from the ReS in a controUed manner. Page 19 of 75 Printed on 1/28/2010 at 12:02

SRO Exam Questions Only (No "Pa.re~~sf'Oinprigipals~') Question #: 7 IQueStion ID: 9000020 ORO ~ SRO o Student Handout? o Lower Order? [-SRO Ques. # 7 Rev. 1 ~ Selected for Exam Origin: New o 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 tak.en to stabilize RCS temperature and the following conditions were observed:

Reactor power stable at - 99%.
Only the Upper Electrical Limit lights for all CEAs 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 aHer CEA #1 slipped, all required actions per AOP-2556, CEDS Malfunctions, have been completed, includin!~ 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? [J A Pulse counts must be reset to clear the Upper Core Stop and the CMI must be bypassed for CEA recovery. CEA #1 Pulse Count Indication and the CMI will be INOPERABLE while the CEA is being recovered. [J B CEA #1 Upper Electrical Limit must be defeated by lifting a lead and the CMI must be bypassed for CEA recovery. Reed Switch Indication for CEA #1 and the CIVIl will be II\\lOPERABLE while the CEA is being recovered. [J C Pulse counts must be reset to clear the Upper Core Stop and the CMI must be bypassed for CEA recovery. Onlv the CMI will be INOPERABLE while the CEA is being recovered. [~ D CEA #1 Upper Electrical Limit must be defeated by lifting a lead 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 tri!lgered based on the CEA #1 deviation from the other CEAs in Group 7, therefore it must be bypassed to recover the CEA. When tile 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 tile correct interlockslinhibits 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 Bruce Ferguson & Cliff Chapin - Change defeating the UEL to the actual action taken by the procedure. Modified choices "B" & "0" (correct answer) per reviewer comment. - RLC 12/22/09 NRC KIA System/EiA System 003 Dropped Control Rod Page 20 of 75 Printed on 1/28/2010 at 12:02 SRO Exam Questions Only (No "Pa.re~~sf'Oinprigipals~') Question #: 7 IQueStion ID: 9000020 [-SRO Ques. # 7 Rev. 1 ORO ~ SRO ~ Selected for Exam o Student Handout? Origin: New o Lower Order? o 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 tak.en to stabilize RCS temperature and the following conditions were observed:

Reactor power stable at - 99%.
Only the Upper Electrical Limit lights for all CEAs 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 aHer CEA #1 slipped, all required actions per AOP-2556, CEDS Malfunctions, have been completed, includin!~ 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? [J A Pulse counts must be reset to clear the Upper Core Stop and the CMI must be bypassed for CEA recovery. CEA #1 Pulse Count Indication and the CMI will be INOPERABLE while the CEA is being recovered. [J B CEA #1 Upper Electrical Limit must be defeated by lifting a lead and the CMI must be bypassed for CEA recovery. Reed Switch Indication for CEA #1 and the CIVIl will be II\\lOPERABLE while the CEA is being recovered. [J C Pulse counts must be reset to clear the Upper Core Stop and the CMI must be bypassed for CEA recovery. Onlv the CMI will be INOPERABLE while the CEA is being recovered. [~ D CEA #1 Upper Electrical Limit must be defeated by lifting a lead 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 tri!lgered based on the CEA #1 deviation from the other CEAs in Group 7, therefore it must be bypassed to recover the CEA. When tile 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 tile correct interlockslinhibits 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 Bruce Ferguson & Cliff Chapin - Change defeating the UEL to the actual action taken by the procedure. Modified choices "B" & "0" (correct answer) per reviewer comment. - RLC 12/22/09 NRC KIA System/EiA System 003 Dropped Control Rod Page 20 of 75 Printed on 1/28/2010 at 12:02

Question #: 7 IQuestion ID: 9000020 RO ~SRO

I Student Handout?

Lower Order? I-SRO Ques. # 7 Rev. 1 ~ Selected for Exam Origin: New Past NRC Exam? Number AA2.04 RO 3.4* SRO 3.6* CFR Link (CFR: 43.5/45.13) 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 1.0 PURPOSE AOP 2556 Revision 016 - lO c U This AOP contains [OP related material. U I. 1 Objectin~ This procedure provides instnlctions for the following malfunctions which could affect CEAs, CEDS, ACTM or CEA position indications: Multiple misaligned or untrippablc CEAs Misaligned CEA misaligned greatcr than 10 steps Inoperable CEA Position Indication System Inoperable CMl circuit Trippublc CEA U ntTi ppable CEA 1.2 Disclllssion following a CEA drop, operator action should be directed toward returning the plall1. to a slabk conditioll. At high power levels, if no action is taken following the CEA drop, reactor power will return to approximately the initial power level, but tH a reduced core average temperature (due to positive reactivity feedback from the negative moderator temperature coefficient). The following actions will minimize the affects of the CEA drop transient. Dropped CEAs result in reactor and turbine power mismatch. This mismatch is nulled by reducing turbine load to match reactor power (RCS temperatures not changing). Dropped CEAs al'lo result in undesirable neutron flux pattems. By correct operator respon~, the time span over which thesc patterns exist is minimized. It is desirable to record as much data as possible concerning abnormal flux patterns existing during and subsequent to rod drop. Proper usc of PPC, as outlined in this proccdure, produces data necessary for subsequent analysis r ___ ** _12y}~e~tJ'lr_E_n,gjEs.e!iE~L ____________________________ _

If during CEA motion, a core mimic light fails to clear or light and all other I indications are normal, it is acceptable to request I&C Department to temporarily I

lift kads for respective CEA upper c\\cctricallimit reed switch until it is aligned.

I This is not considereu a CEA position indication problem. I

~

Level OfUs~ ContinuOl~ STOP THINK ACT REVIEW Page 21 of 75 Printed on 1/28/2010 at 12:02 Question #: 7 IQuestion ID: 9000020 RO ~SRO

I Student Handout?

Lower Order? I-SRO Ques. # 7 Rev. 1 ~ Selected for Exam Origin: New Past NRC Exam? Number AA2.04 RO 3.4* SRO 3.6* CFR Link (CFR: 43.5/45.13) 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 1.0 PURPOSE AOP 2556 Revision 016 - lO c U This AOP contains [OP related material. U I. 1 Objectin~ This procedure provides instnlctions for the following malfunctions which could affect CEAs, CEDS, ACTM or CEA position indications: Multiple misaligned or untrippablc CEAs Misaligned CEA misaligned greatcr than 10 steps Inoperable CEA Position Indication System Inoperable CMl circuit Trippublc CEA U ntTi ppable CEA 1.2 Disclllssion following a CEA drop, operator action should be directed toward returning the plall1. to a slabk conditioll. At high power levels, if no action is taken following the CEA drop, reactor power will return to approximately the initial power level, but tH a reduced core average temperature (due to positive reactivity feedback from the negative moderator temperature coefficient). The following actions will minimize the affects of the CEA drop transient. Dropped CEAs result in reactor and turbine power mismatch. This mismatch is nulled by reducing turbine load to match reactor power (RCS temperatures not changing). Dropped CEAs al'lo result in undesirable neutron flux pattems. By correct operator respon~, the time span over which thesc patterns exist is minimized. It is desirable to record as much data as possible concerning abnormal flux patterns existing during and subsequent to rod drop. Proper usc of PPC, as outlined in this proccdure, produces data necessary for subsequent analysis r ___ ** _12y}~e~tJ'lr_E_n,gjEs.e!iE~L ____________________________ _

If during CEA motion, a core mimic light fails to clear or light and all other I indications are normal, it is acceptable to request I&C Department to temporarily I

lift kads for respective CEA upper c\\cctricallimit reed switch until it is aligned.

I This is not considereu a CEA position indication problem. I

~

Level OfUs~ ContinuOl~ STOP THINK ACT Page 21 of 75 REVIEW Printed on 1/28/2010 at 12:02

Question #: 7 IQueStion ID: 9000020 RO

.,. SRO Student Handout?

Lower Order? II-SRO Ques. # 7 Rev. 1 ~ Selected for Exam Origin: New Past NRC Exam? Millstone Unit 2 AOP 2556 Revision 016-10 CEA lVlalfunctions Page 13 of 55 INSTRUCTIONS CONTINGENCY ACTIONS 1£ necessary, PERFORM the following to modi[y dropped CEA position on PPC to match dropped CEA position indicated on "CEAPDS MONITOR:"

a. OBSERVE dropped CEAposition on "CEAPDS MONITOR."

b. Using "CEA Positions Menu" on

ppe, SELECT "CEA POSlTION EDITOR" I:Uld PERFORM directions as indicated on ppc.

Page 22 of 75 Printed on 1/28/2010 at 12:02 Question #: 7 IQueStion ID: 9000020 RO

.,. SRO Student Handout?

Lower Order? II-SRO Ques. # 7 Rev. 1 ~ Selected for Exam Origin: New Past NRC Exam? Millstone Unit 2 CEA lVlalfunctions AOP 2556 Revision 016-10 INSTRUCTIONS 1£ necessary, PERFORM the following to modi[y dropped CEA position on PPC to match dropped CEA position indicated on "CEAPDS MONITOR:"

a.

OBSERVE dropped CEA position on "CEAPDS MONITOR."

b.

Using "CEA Positions Menu" on

ppe, SELECT "CEA POSlTION EDITOR" I:Uld PERFORM directions as indicated on ppc.

Page 22 of 75 Page 13 of 55 CONTINGENCY ACTIONS Printed on 1/28/2010 at 12:02

Question #: 7 RO SRO Student Handout? o Lower Order? iQueStion ID; 9000020 [-SRO Ques. If 7 Rev. 1 ~ Selected for Exam Origin: New o Past NRC Exam? Millstone Unit 2 AOP 2556 Revision 016-10 CEA Malfunctions Page 14 of 55 INSTRUCTIONS 4.19 As desired, SELECr applicable CEA group " +/- 15 STEPS" or "FULL SCALE" on "CEAPDS MONITOR." _4.20 PRESS "'MANUAL INDIVIDUAL, MI" and CHECK light lit. _4.21 PRESS applicable group "[NHIBIT BYPASS" and CHECK the rollowing: Appropriate group red "INHIBIT BYPASS," lit

'TEA MOTION INHIBIT BYP" annunciator lit (BA-19, C-(4)

(depends on group selected) _4.22 PRESS applicable "GROUP SELECTION." _4.23 LOG entry into TIS LCD, 3.l.3.1, I ACTION B.1 (eMI bypassed). CONTINGENCY ACTIONS Page 23 of75 Printed on 1/28/2010 at 12:02 Question #: 7 iQueStion ID; 9000020 [-SRO Ques. If 7 Rev. 1 RO SRO ~ Selected for Exam Student Handout? Origin: New o Lower Order? o Past NRC Exam? Millstone Unit 2 CEA Malfunctions AOP 2556 Revision 016-10 INSTRUCTIONS 4.19 As desired, SELECr applicable CEA group " +/- 15 STEPS" or "FULL SCALE" on "CEAPDS MONITOR." _4.20 PRESS "'MANUAL INDIVIDUAL, MI" and CHECK light lit. _4.21 PRESS applicable group "[NHIBIT BYPASS" and CHECK the rollowing: Appropriate group red "INHIBIT BYPASS," lit 'TEA MOTION INHIBIT BYP" annunciator lit (BA-19, C-(4) (depends on group selected) _4.22 PRESS applicable "GROUP SELECTION." _4.23 LOG entry into TIS LCD, 3.l.3.1, I ACTION B.1 (eMI bypassed). Page 23 of75 Page 14 of 55 CONTINGENCY ACTIONS Printed on 1/28/2010 at 12:02

Question #: 8 Student Handout? IQueStion ID: 9000005 RO !" SRO ~ Lower Order? l-SRO Ques. # B Rev. 0 ~ Selected for Exam Otfgln: New D 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 the direction for controlling the affected SG level during the performance of EOP 2525, Standard Post Trip Actions, !md 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 rupturo 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. ~ 0 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 Stearn Plant Radiation Monitors are in alarm, EOP 2525 requires feed flow to maintained to the SG with the hi~lhest radiation reading to maintHin 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 IiTlits the gaseous release to the environment (The loss of off-site power results in a loss of condenser vacuum which requires 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 throu!Jh the broken tube(s). A is incorrect; Feedwater should NOT be secured to the affected SG until after the cooldown to 515°F, The addition of feedwater allows for Iodine scrubbinu and limits the radioactive release to the environment. Plausible because, up until a few years ago, feedwater ~ 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 70%. C is incorrect. While it is true that the addition of feedwater 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 2:525 requires feeding the unaffected (or least affected) SG at greater than 300 gpm. References 1 OP 2260, Unit 2 EOP User Guide EOP 2525, Standard post Trip Actions Comments and Questiol1 Modification History Bob K. - D-2/C (need only "40% - 45%" for answer) No changes made. Bill M. - D-2/C, K Angelo D-2/C; No comments. NRC - SG level range in Choice "S" justification does not match range in dlstracter. RLC Changed explanation in choice "B" from "40 to 75%" to "40 -70%", to match distracter. [11/13/09] Bruce F. D-2/C, No comment NRC KIA System/ElA System 060 Accidental Gaseous Radwaste Release NRC KIA Generic System 2.4 Emergency Procedures IPlan Number 2.4.18 RO 3.3 SRO 4.0 CFR Link (CFR: 41.10/43.1 145,13) Knowledge of the speCific bases for EOPs. Page 24 of75 Printed on 1/28/2010 at 12:02 Question #: 8 IQueStion ID: 9000005 l-SRO Ques. # B Rev. 0 RO !" SRO ~ Selected for Exam Student Handout? Otfgln: New ~ Lower Order? D 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 the direction for controlling the affected SG level during the performance of EOP 2525, Standard Post Trip Actions, !md 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 rupturo 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. ~ 0 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 Stearn Plant Radiation Monitors are in alarm, EOP 2525 requires feed flow to maintained to the SG with the hi~lhest radiation reading to maintHin 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 IiTlits the gaseous release to the environment (The loss of off-site power results in a loss of condenser vacuum which requires 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 throu!Jh the broken tube(s). A is incorrect; Feedwater should NOT be secured to the affected SG until after the cooldown to 515°F, The addition of feedwater allows for Iodine scrubbinu and limits the radioactive release to the environment. Plausible because, up until a few years ago, feedwater ~ 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 70%. C is incorrect. While it is true that the addition of feedwater 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 2:525 requires feeding the unaffected (or least affected) SG at greater than 300 gpm. References 1 OP 2260, Unit 2 EOP User Guide EOP 2525, Standard post Trip Actions Comments and Questiol1 Modification History Bob K. - D-2/C (need only "40% - 45%" for answer) No changes made. Bill M. - D-2/C, K Angelo D-2/C; No comments. NRC - SG level range in Choice "S" justification does not match range in dlstracter. RLC Changed explanation in choice "B" from "40 to 75%" to "40 -70%", to match distracter. [11/13/09] Bruce F. D-2/C, No comment NRC KIA System/ElA System 060 Accidental Gaseous Radwaste Release NRC KIA Generic System 2.4 Emergency Procedures IPlan Number 2.4.18 RO 3.3 SRO 4.0 CFR Link (CFR: 41.10/43.1 145,13) Knowledge of the speCific bases for EOPs. Page 24 of75 Printed on 1/28/2010 at 12:02

Question #: 8 IQuestion ID; 9000006 RO ~ SRO Student Handout? ~ Lower Order? I~11-;:S::Ro::::Q::ue::s::,1f::::=8;::::~====~R:e:;v::,.=====O===~~1s=e=l=ec=t=ed=fo=r=E=x=am===O~tU==n=:==N~e=w==~~*=P=8S=t NRC EXllm? I i 1 EOP 2525, "Standard Post Trip Actions," Implementation Guide (Shed 8 of 11)

b.

Due to current MSSY blowdowll setpoinls (- 880 psia), it is possible for MSSV(s) to be open post-trip when the steam dumps are not available AND have SO pressure within the normal control band. To assess if a MSSV is actually stuck open, it is recommended the CO adjust ADY automatic sctpoint( s) to the lower end of the control ban(.1. Additionally, temperature outside the Ilormal band should also be llsed to assess it stuck open MSSV.

c.

IfTe is less than 53()°F, the operator should determine iffcedwatcr flow is excessivl..~ to one or both SGs and adjust or isolate fcedwatcr flow as required. The operator should determine if the cause of the excessive flow is d lie to control system malfunction (e.g., FRY not closing on the trip or FRY bypass valves not at the rc()uired position following the trip), or due to leedwater flow to a SG blowing down from an ESDE. In the case of I11Cllfunctioning equipment. the operator should attempt to au.just feedwater flow manually, since a c{)mponent failure should not result in the loss of a SG for heal removal. The operator is required to isolate AFW to the affected SG within 30 minutes following the generation of MSI$ during an ESDE. For scenarios where isolation is not possible from the Control Room. allowance must be made for local operation of FW-43A(B) or FW-44. It has been validated that it will lake approximately 15 minutes to dose FW-43A(B) or FW-44 locally. Therefore. isolation of AFW to an affected steam gene;ator, from the control room, must be attempted within IS minutes of a MS1S.

d.

If Tc is less than 5300 F AND. the ESDE has been terminated. the operator is reqtlircd to operate the ADY or steam dumps to stabilize Te. Temperature should not be allowed to restore to the nnrmal band following an ESDE.

c.

If SG level is lowering and both MDAFW pumps are 110t operating, the operaHlr is required to start the TDAFP within to minute!; following a Loss of Normal Feedwater.

f.

If a SGTR has occurred. the operator is expected to throttle fced to the most affected SG as necessary to maintain level low in the band (40 to 45%). This will aid in maintaining SG pressure during the cooldown and aid in scrllbbing radioactive iodine. The top of the sa tube bundle is - 33%. If break t10w is restoring level to this band then feed flow is not necc'lsary [this may not be assessed until verification of Containment Iso iatio nl OP 2260 Level ofUs~ STOP THINK ACT REVIEW Rev. U09-03 Informatiol~ 2R of 60 Page 25 of75 Printed on 1/28/2010 at 12:02 Question #: 8 IQuestion ID; 9000006 RO ~ SRO Student Handout? ~ Lower Order? I~ 11-;:S::Ro::::Q::ue::s::, 1f::::=8;::::~====~R:e:;v::,.=====O===~~1 s=e=l=ec=t=ed=fo=r=E=x=am===O~tU==n=:==N~e=w==~~* =P=8S=t 1 NRC EXllm? I i EOP 2525, "Standard Post Trip Actions," Implementation Guide (Shed 8 of 11)

b.

Due to current MSSY blowdowll setpoinls (- 880 psia), it is possible for MSSV(s) to be open post-trip when the steam dumps are not available AND have SO pressure within the normal control band. To assess if a MSSV is actually stuck open, it is recommended the CO adjust ADY automatic sctpoint( s) to the lower end of the control ban(.1. Additionally, temperature outside the Ilormal band should also be llsed to assess it stuck open MSSV.

c.

IfTe is less than 53()°F, the operator should determine iffcedwatcr flow is excessivl..~ to one or both SGs and adjust or isolate fcedwatcr flow as required. The operator should determine if the cause of the excessive flow is d lie to control system malfunction (e.g., FRY not closing on the trip or FRY bypass valves not at the rc()uired position following the trip), or due to leedwater flow to a SG blowing down from an ESDE. In the case of I11Cllfunctioning equipment. the operator should attempt to au.just feedwater flow manually, since a c{)mponent failure should not result in the loss of a SG for heal removal. The operator is required to isolate AFW to the affected SG within 30 minutes following the generation of MSI$ during an ESDE. For scenarios where isolation is not possible from the Control Room. allowance must be made for local operation of FW-43A(B) or FW-44. It has been validated that it will lake approximately 15 minutes to dose FW-43A(B) or FW-44 locally. Therefore. isolation of AFW to an affected steam gene;ator, from the control room, must be attempted within IS minutes of a MS1S.

d.

If Tc is less than 5300 F AND. the ESDE has been terminated. the operator is reqtlircd to operate the ADY or steam dumps to stabilize Te. Temperature should not be allowed to restore to the nnrmal band following an ESDE.

c.

If SG level is lowering and both MDAFW pumps are 110t operating, the operaHlr is required to start the TDAFP within to minute!; following a Loss of Normal Feedwater.

f.

If a SGTR has occurred. the operator is expected to throttle fced to the most affected SG as necessary to maintain level low in the band (40 to 45%). This will aid in maintaining SG pressure during the cooldown and aid in scrllbbing radioactive iodine. The top of the sa tube bundle is - 33%. If break t10w is restoring level to this band then feed flow is not necc'lsary [this may not be assessed until verification of Containment Iso iatio nl Level ofUs~ Informatiol~ STOP THINK ACT Page 25 of75 OP 2260 REVIEW Rev. U09-03 2R of 60 Printed on 1/28/2010 at 12:02

Question #: 9 ~ Student Handout? Lower Order? I~uestion ID: 9000006 RO ~ SRO I-SRO Ques. f/. 9 Rev. 0 ~ Selected for Exam Origin: New 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 187°F 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/Gevels al'e 60% and stable - Containment is boing 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 Justification I C IS CORRECT; MP-26-EPI-FAP06-002, Equipment Failure, EA2, Uncontrolled temperature increase >10°F that results in RCS temperature >200°F. Due to the loss of SOC 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 240 would result In an Undervoltage actuation on Facility 2. Per RAC 14, Non-Emergency Station Events, an 8 hour report is required. (General Interest, Echo); however, the loss of SOC 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 classification must be reported. Plausible if examinee did not look at the higher Alert classification for Inability to Maintain Cold SID. o is incorrect; Per MP-26*EPI-FAP06-002, Equipment Failure, ES2(1.), No RCS Heat Removal via Steam Generators AND Once Through Cooling NOT effEtctive 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 steamin£1 AND Once Through Cooling would likely be effective if it were initiated. Plausible because the exzlmlnee may believe that no core cooling method is presently being utilized, therefore, core cooling is in serious jeopardy. . References I;P\\'tiVlGed 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. - 0*21Y, K (Add, "No other operator actions have been taken." to the bullet list)

Added "No other operatc)r actions have been taken." to the bullet list, per recommendation. RJA Angelo - D-3/C; No comments. Bruce F. D-3/C, No comment NRC KIA System/ElA ,...,,,...,,<:: KIA Selected NRC KIA Generic System System 074 2.4 Inadequate Core Cooling Emergency Procedures IPlan Page 26 of75 Printed on 1/28/2010 at 12:02 Question #: 9 I~uestion ID: 9000006 RO ~ SRO ~ Student Handout? Lower Order? I-SRO Ques. f/. 9 Rev. 0 ~ Selected for Exam Origin: New 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 187°F 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/Gevels al'e 60% and stable - Containment is boing 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 Justification I C IS CORRECT; MP-26-EPI-FAP06-002, Equipment Failure, EA2, Uncontrolled temperature increase >10°F that results in RCS temperature >200°F. Due to the loss of SOC 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 240 would result In an Undervoltage actuation on Facility 2. Per RAC 14, Non-Emergency Station Events, an 8 hour report is required. (General Interest, Echo); however, the loss of SOC 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 classification must be reported. Plausible if examinee did not look at the higher Alert classification for Inability to Maintain Cold SID. o is incorrect; Per MP-26*EPI-FAP06-002, Equipment Failure, ES2(1.), No RCS Heat Removal via Steam Generators AND Once Through Cooling NOT effEtctive 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 steamin£1 AND Once Through Cooling would likely be effective if it were initiated. Plausible because the exzlmlnee may believe that no core cooling method is presently being utilized, therefore, core cooling is in serious jeopardy. . References I ;P\\'tiVlGed 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. - 0*21Y, K (Add, "No other operator actions have been taken." to the bullet list)

Added "No other operatc)r actions have been taken." to the bullet list, per recommendation. RJA Angelo - D-3/C; No comments. Bruce F. D-3/C, No comment NRC KIA System/ElA System 074 Inadequate Core Cooling ,...,,,...,,<:: KIA Selected NRC KIA Generic System 2.4 Emergency Procedures IPlan Page 26 of75 Printed on 1/28/2010 at 12:02

Question #: 9 j:)uestion ID: 9000006 RO ~ SRO ['Ili Student Handout? o Lower Order? I-SRO Ques. It 9 Rev. 0 l"") Selected for Exam Origin: New Past NRC Exam? Number 2.4.:lD 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 NqC, or thH transmission system operator. Failure of Automallc Reactor rrip Ar!IIl ManuaJ Trip Was Successful I EA2 I IINABIUTY TO MAINTAIN GOLD SID I I Mode 5. 6

1. Uncontrolled RCS Temperature Increase '" 10°F That Resulls in RCS Temperature> 200' F

2. Inadvertent C'rucailty

~ toss OF ANNUNCIA1'ORSfffiANSIEN] I Mode 1. 2, 3, 4 I Loss of Most (75%) MeB AnnlJoclators '" 15 Minutes Al:!m EITHER of lhe Following: Significant Transient In Progress Loss of SPDS AIiQ ICC Instrumentation

1.

Loss of Shutdown Cooling", 15 Minutes AND Refuel Pool Waler level <: 35 FI.. 6 In.

2.

Uncontrolled FIGS Tempemturo Increase> 10'F

3. RCS Boron Cc.ncentration <: Minimum Required I EU2 I IREFUEL/SPENT FUEL POOL LEVELl Ir---M-Od-e-S-,-0----.

1, Uncontrolled Spert Fuel Pool Water level Decrease Causing loss of GaOling Suclion Flow

2.

Uncoctroiied fleluei Pool Vlater levei Decrease Requiring Containment Evacuation 8!'m All Spent Fuel Assemblies in Safe Storage locat ons IEU3 I L LOSS OF ANNUNCIATORS I I Mode 1. 2. 3. 4 I loss of Most (75%:, MeB Annunciators:> 15 Minutes AND SPDS QB ICC Instrums nlation Available I EU4 I G*OSS OF COMMUNICATIONS I I Mode ALL

1.

Loss of ALL Onsile Bectronic Communications Methods

2. Loss of,~u. Electronic Communications Methods With Government P,genc1es I EU5 I [SHUT-D-OW-N-L-C-O-E-XC-E-E-D-E-D~I I Mode 1. 2, 3. 4 I Unit NOT Brought fo Required Mode within Applicable LCO Action Statement Time Limits Page 27 of 75 Printed on 1/28/2010 at 12:02 Question #:

9 j:)uestion ID: 9000006 RO Ii*] SRO ['Ili Student Handout? o Lower Order? I-SRO Ques. It 9 Rev. 0 l"") Selected for Exam Origin: New Past NRC Exam? Number 2.4.:lD 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 NqC, or thH transmission system operator. Failure of Automallc Reactor rrip Ar!IIl M anuaJ Trip Was Successful I EA2 I IINABIUTY TO MAINTAIN GOLD SID I I Mode 5. 6

1.

Uncontrolled RCS Temperature Increase '" 10°F That Resulls in RCS Temperature> 200' F

2.

Inadvertent C'rucailty ~ toss OF ANNUNCIA1'ORSfffiANSIEN] I Mode 1. 2, 3, 4 I Loss of Most (75%) MeB AnnlJoclators '" 15 Minutes Al:!m EITHER of lhe Following: Significant Transient In Progress Loss of SPDS AIiQ ICC Instrumentation

1.

Loss of Shutdown Cooling", 15 Minutes AND Refuel Pool Waler level <: 35 FI.. 6 In.

2.

Uncontrolled FIGS Tempemturo Increase> 10'F

3.

RCS Boron Cc.ncentration <: Minimum Required I EU2 I I REFUEL/SPENT FUEL POOL LEVELl Ir---M-Od-e-S-, ---. 1, Uncontrolled Spert Fuel Pool Water level Decrease Causing loss of GaOling Suclion Flow

2.

Uncoctroiied fleluei Pool Vlater levei Decrease Requiring Containment Evacuation 8!'m All Spent Fuel Assemblies in Safe Storage locat ons I EU3 I L LOSS OF ANNUNCIATORS I I Mode 1. 2. 3. 4 I loss of Most (75%:, MeB Annunciators:> 15 Minutes AND SPDS QB ICC Instrums nlation Available I EU4 I G*OSS OF COMMUNICATIONS I I Mode ALL

1.

Loss of ALL Onsile Bectronic Communications Methods

2.

Loss of,~u. Electronic Communications Methods With Government P,genc1es I EU5 I [SHUT-D-OW-N-L-C-O-E-XC-E-E-D-E-D~I I Mode 1. 2, 3. 4 I Unit NOT Brought fo Required Mode within Applicable LCO Action Statement Time Limits Page 27 of 75 Printed on 1/28/2010 at 12:02

Question #: 10 IQueStion ID: 9000007 RO v: SRO Student Handout? i",,: Lower Order? I-SRO Ques. It 10 Rev. 1 ~ Selected for Exam Origin: New Past NRC EXilm? An RCS ChE!mistry 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> 1.0 micro-curies/gram DOSE EQUIVALENT 1-131, be IJ A in COLD SHUTDOWN within 36 hours after detection. Isotopic analysis of the primary coolant must be perforrr,ed once per hour when activity of the primary coolant> 1.0 micro-curies/gram DOSE EQUIVALENT !-131. The hourly sampling period allows time to obtain and analyze a sample. There is a low probability of a steam Hne bresik or S/G tube rupture in the next 36 hours and there is significant conservatism built into the RCS specific activity limit. Wittl the specific activity of the primary coolant> 1.0 micro-curies/gram DOSE EQUIVALENT 1-131, I~ B verify DOSE EQUIVALENT 1-131 ~ 60 micro-curies/gram once per 4 hours. Operation may continue for up to 48 hours while efforts are made to restore DOSE EQUIVALENT 1-131 to within the 1.0 micro curies/gram limit. The 4 hour sampling period allows time to obtain and analyze a sample. There is a low probability of a steam line breSik or S/G tube rupture in the next 48 hours and it is expected that normal coolant Iodine concentration would be restored within 48 hours. C With the specific activity of the primary coolant> 1.0 micro-curies/gram DOSE EQUIVALENT 1-131, be in HOT STANDBY within 6 hours from the time of detection and in COLD SHUTDOWN within 48 hours from the time of detection. The allowed completion times are reasonable 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 LOCI!\\. 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 hours or be in HOT STANDBY within the following 6 hours. It is expected tllat normal coolant Iodine concentration would be restored within 36 hours. 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 hours. b. With the specific activity of the primary coolant> 1.0 micl'o-curies/gram DOSE EQUIVALENT 1-131, but::: 60 micro-curies/gram, operation may continue for up to 48 hours while efforts are melde 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 ECIUIVALENT 1-131 greater than the LCO limit, samples at intervals of 4 hours must be taken to demonstr,ate that the specific activity is::: 60 micro-curies/gram. Four hours is required to obtain and analyze a sample. Sampling is continued every 4 hours to provide a trend. The DOSE EQUIVALENT 1-131 must be restored to within limit within 48 hours. The completion time of 48 hours is ac:ceptable 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 hours. Plausible because several other Tech Spec Action Statements require the plant to achieve COLD SHUTDOWN within 36 hours (Example: Containment Integrity, TSAS 3.6.1.1). C is incorrect; TSAS 3.4.8 does not require the plant to achieve HOT SHUTDOWN within 6 hours. Plausible because sever::,1 other Tech Spec Action Statements require the plant to achieve HOT SHUTDOWN within 6 hours and COLD SHUTDOWN within the following 36 hours. (Example: Specific Activity, TSAS 3.4.8c., DOSE EQUIVALENT 1-131 >60 micro curies/gram. ) D is incorrect; TSAS 3.4.8 does not require the plant to achieve HOT SHUTDOWN within 6 hours if RCS coolant specific activity cannot be lowered to ::: 1.0 micro-curies/gram DOSE EQUIVALENT 1-131 in 36 hours. Plausible if the examinee confuses the time requirements and the actual limit. References I Tech Spec 3.4.13, Specific Activity, and applicable Bases. Comments and Questicln Modification History I NRC - Remove the numbers (">0.1" and ">1.0") from all choices or make them all the same. RLC - Changed "> 0.1" in Choices "A" and "C" to "> 1.0". [11/30/091 Bruce F. - D-4/W, 50/50 (C and D) Page 28 of75 Printed on 1/28/2010 at 12:02 Question #: 10 IQueStion ID: 9000007 RO v: SRO Student Handout? i",,: Lower Order? I-SRO Ques. It 10 Rev. 1 ~ Selected for Exam Origin: New Past NRC EXilm? An RCS ChE!mistry 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? IJ A I~ B With the specific activity of the primary coolant> 1.0 micro-curies/gram DOSE EQUIVALENT 1-131, be in COLD SHUTDOWN within 36 hours after detection. Isotopic analysis of the primary coolant must be perforrr,ed once per hour when activity of the primary coolant> 1.0 micro-curies/gram DOSE EQUIVALENT !-131. The hourly sampling period allows time to obtain and analyze a sample. There is a low probability of a steam Hne bresik or S/G tube rupture in the next 36 hours and there is significant conservatism built into the RCS specific activity limit. Wittl 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 hours. Operation may continue for up to 48 hours while efforts are made to restore DOSE EQUIVALENT 1-131 to within the 1.0 micro-curies/gram limit. The 4 hour sampling period allows time to obtain and analyze a sample. There is a low probability of a steam line breSik or S/G tube rupture in the next 48 hours and it is expected that normal coolant Iodine concentration would be restored within 48 hours. C With the specific activity of the primary coolant> 1.0 micro-curies/gram DOSE EQUIVALENT 1-131, be in HOT STANDBY within 6 hours from the time of detection and in COLD SHUTDOWN within 48 hours from the time of detection. The allowed completion times are reasonable 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 LOCI!\\. 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 hours or be in HOT STANDBY within the following 6 hours. It is expected tllat normal coolant Iodine concentration would be restored within 36 hours. 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 hours. b. With the specific activity of the primary coolant> 1.0 micl'o-curies/gram DOSE EQUIVALENT 1-131, but::: 60 micro-curies/gram, operation may continue for up to 48 hours while efforts are mclde 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 ECIUIVALENT 1-131 greater than the LCO limit, samples at intervals of 4 hours must be taken to demonstr,ate that the specific activity is::: 60 micro-curies/gram. Four hours is required to obtain and analyze a sample. Sampling is continued every 4 hours to provide a trend. The DOSE EQUIVALENT 1-131 must be restored to within limit within 48 hours. The completion time of 48 hours is ac:ceptable 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 hours. Plausible because several other Tech Spec Action Statements require the plant to achieve COLD SHUTDOWN within 36 hours (Example: Conlainment Integrity, TSAS 3.6.1.1). C is incorrect; TSAS 3.4.8 does not require the plant to achieve HOT SHUTDOWN within 6 hours. Plausible because sever::,1 other Tech Spec Action Statements require the plant to achieve HOT SHUTDOWN within 6 hours and COLD SHUTDOWN within the following 36 hours. (Example: Specific Activity, TSAS 3.4.8c., DOSE EQUIVALENT 1-131 >60 micro-curies/gram. ) D is incorrect; TSAS 3.4.8 does not require the plant to achieve HOT SHUTDOWN within 6 hours if RCS coolant specific activity cannot be lowered to ::: 1.0 micro-curies/gram DOSE EQUIVALENT 1-131 in 36 hours. Plausible if the examinee confuses the time requirements and the actual limit. References I Tech Spec 3.4.13. Specific Activity, and applicable Bases. Comments and Questicln Modification History I NRC - Remove the numbers (">0.1" and ">1.0") from all choices or make them all the same. RLC - Changed "> 0.1" in Choices "A" and "C" to "> 1.0". [11/30/091 Bruce F. - D-4/W, 50/50 (C and D) Page 28 of75 Printed on 1/28/2010 at 12:02

Question #: 10 RO ~SRO Student Handout? ~ Lower Order? Question ID: 9000007 I-SRO Que*. # 10 Rev. 1 ~ Selected for Exam Origin: New Past NRC Exam? NRC KIA System/E/A System 076 High Reactor Coolant Activity Number AA2.02 RO 2.8 SRO 3.4 CFR Link (CFR: 43.5/45.13) Ability to dete'mine ancl 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 75 Printed on 1/28/2010 at 12:02 Question #: 10 I-SRO Que *. # 10 Question ID: 9000007 Rev. 1 RO ~SRO ~ Selected for Exam NRC KIA System/E/A System 076 High Reactor Coolant Activity Number AA2.02 RO 2.8 SRO 3.4 CFR Link (CFR: 43.5/45.13) Student Handout? ~ Lower Order? Origin: New Past NRC Exam? Ability to dete'mine ancl 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 75 Printed on 1/28/2010 at 12:02

Question #: 10 Question ID: 9000007 RO SRO Student Handout? ~ Lower Order? l-SRO Ques. # 10 Rev. 1 ~ Selected for Exam Origin: New D Past NRC Exam? October 27,2008 SPECIFlC.ACTIv1TY LIMITING CONDITION FOR OPERATION 3.4.8 The sp(~cific activity of the primary coolant shall be limited to:

a.

LO ~lCi/gram DOSE EQUNALENT 1-13 Land

b.

~ 1100 !-lCi/gram DOSE EQUIVALENT XE-133. APPLICABILITY~ MODES L 2, 3, 4. ACTION:

a.

With the specific activity of the primalY coolant> 1.0 ~lCilgralll DOSE EQUIVALENT 1-131, verify DOSE EQUIVALENT 1-13 1 ~ 60 11C i/gram once per 4 hours.

b.

With the specific activity ofthe prinuuy coolant> 1.0 ~lCi/gram DOSE EQUNALENT 1-131 but::;; 60 ~lCilgram, operation may continue for up to 48 hours while eff0l1s are made to restore DOSE EQUIVALENT 1-131 to within the 1.011Ci/gramlimit. Specification 3.0.4 is not applicable.

c.

With the specific activity ofthe primary coolant> 1.0 ~lCilgram DOSE EQUIVALENT 1-131 for more than 48 hours during one continuous time interval, or 60 JICilgram DOSE EQUIVALENT 1-131, be in HOT STANDBY within 6 hours and in COLD SHUTDOWN within 36 hours.

d.

With the specific activity of the primary coolant> 1100 ~lCi/gram DOSE EQUIVALENT XE-133, operatiolllllay continue for up to 48 hours while eff0l1s are made to restore DOSE EQUIVALENT XE-133 to within the 1100 JlCi/gram lillllt. Specificatioll 3.0.4 is not applicable.

e.

With the specific activity ofthe primary coolant> 1100 I1Ci/gram DOSE EQUIVALENT XE-133 for more than 48 hours during one continuous time interval, be in HOT STANDBY within 6 hours and in COLD SHUTDOWN within 36 hours. Page 30 of 75 Printed on 1/28/2010 at 12:02 Question #: 10 l-SRO Ques. # 10 Question ID: 9000007 Rev. 1 SPECIFlC.ACTIv1TY RO SRO ~ Selected for Exam LIMITING CONDITION FOR OPERATION Student Handout? ~ Lower Order? Origin: New D Past NRC Exam? October 27,2008 3.4.8 The sp(~cific activity of the primary coolant shall be limited to:

a.

LO ~lCi/gram DOSE EQUNALENT 1-13 Land

b.

~ 1100 !-lCi/gram DOSE EQUIVALENT XE-133. APPLICABILITY~ MODES L 2, 3, 4. ACTION:

a.

With the specific activity of the primalY coolant> 1.0 ~lCilgralll DOSE EQUIVALENT 1-131, verify DOSE EQUIVALENT 1-13 1 ~ 60 11 C i/gram once per 4 hours.

b.

With the specific activity of the prinuuy coolant> 1.0 ~lCi/gram DOSE EQUN ALENT 1-131 but::;; 60 ~l Cil gram, operation may continue for up to 48 hours while eff0l1s are made to restore DOSE EQUIVALENT 1-131 to within the 1.011Ci/gramlimit. Specification 3.0.4 is not applicable.

c.

With the specific activity ofthe primary coolant> 1.0 ~lCilgram DOSE EQUIVALENT 1-131 for more than 48 hours during one continuous time interval, or 60 JICilgram DOSE EQUIVALENT 1-131, be in HOT STANDBY within 6 hours and in COLD SHUTDOWN within 36 hours.

d.

With the specific activity of the primary coolant> 1100 ~lCi/gram DOSE EQUIVALENT XE-133, operatiolllllay continue for up to 48 hours while eff0l1s are made to restore DOSE EQUIVALENT XE-133 to within the 1100 JlCi/gram lillllt. Specificatioll 3.0.4 is not applicable.

e.

With the specific activity of the primary coolant> 1100 I1Ci/gram DOSE EQUIVALENT XE-133 for more than 48 hours during one continuous time interval, be in HOT STANDBY within 6 hours and in COLD SHUTDOWN within 36 hours. Page 30 of 75 Printed on 1/28/2010 at 12:02

Question #: 11 Student Handout? Lower Order? Ques60n ID: 9000008 RO ltl SRO I-SRO Ques. 1/ 11 Rev. 1 ~ Selected for Exam Ol1gin: New Past NRC E.xam? 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 "8" RCP operation. - "A" and "8" RCPs have just been started. - Shutdown Cooling has just been secured. - "C" and "0" RCP breakers have just been racked up. Then, the "8" RCP trips when the breaker's overcurrent relay actuated due to being jarred while movin~, staging (NOT an actual overcurrent condition). Which of the following actions is the correct response under the present conditions? ." A Immediately secure the "A" RCP, raise RCS pressure, then start "C" and "0" RCPs. B Immediately 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. Justification I A - CORRECT; The NPSH required pressure for "A" & "S" 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. S - WRONG; If NPSH requirements are not met, the RCP should be immediately secured. Plausible; Would be chmlen if avoiding the loss all RCS flow (violation of Tech. Specs.) is considered above possible RCP dama~le. C - WRONG; The NPSH required pressure for "A" & "S" RCPs is less than that for "C" & "0" RCPs (see OP-2201, Attach. 2 & 3). If the RCS pressure is at the minimum allowed for "A" & "S" RCP operation (initial conditions), it must be below the minimum required pressure for "C" & "0" Rep operation. Plausible; Would be chosen if "CO & "0" RCP NPSH required pressure is believed to be lower than "A" & "S" RCPs. D - 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 Hisi~;y I NRC - (original question comments) Reword stem question from "actions are required under" to "actions is the correct response undE!r" RLC - Reworded stem pElr NRC comments. [11/30109] Sruce F. 0-4/G, No comment NRC KIA System/E/A System 003 Reactor Coolant Pump System (RCPS) Generic KlA.~.E!I!C:~I.J NRC KIA Generic System 2.1 Conduct of Operations Number 2.1.23 RO 4.3 SRO 4.4 CFR Link (CFR: 41.10 143.5145.2/45.6) Ability to perform specifiG system and integrated plant procedures during all modes of plant operation. Page 31 of 75 Printed on 1/28/2010 at 12:02 Question #: 11 I-SRO Ques. 1/ 11 Ques60n ID: 9000008 Rev. 1 RO ltl SRO Student Handout? ~ Selected for Exam Ol1gin: New 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 "8" RCP operation. - "A" and "8" RCPs have just been started. - Shutdown Cooling has just been secured. - "C" and "0" RCP breakers have just been racked up. Lower Order? Past NRC E.xam? Then, the "8" RCP trips when the breaker's overcurrent relay actuated due to being jarred while movin~, staging (NOT an actual overcurrent condition). Which of the following actions is the correct response under the present conditions? ." A Immediately secure the "A" RCP, raise RCS pressure, then start "C" and "0" RCPs. B Immediately 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. Justification I A - CORRECT; The NPSH required pressure for "A" & "S" 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. S - WRONG; If NPSH requirements are not met, the RCP should be immediately secured. Plausible; Would be chmlen if avoiding the loss all RCS flow (violation of Tech. Specs.) is considered above possible RCP dama~le. C - WRONG; The NPSH required pressure for "A" & "S" RCPs is less than that for "C" & "0" RCPs (see OP-2201, Attach. 2 & 3). If the RCS pressure is at the minimum allowed for "A" & "S" RCP operation (initial conditions), it must be below the minimum required pressure for "C" & "0" Rep operation. Plausible; Would be chosen if "CO & "0" RCP NPSH required pressure is believed to be lower than "A" & "S" RCPs. D - 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 Hisi~;y I NRC - (original question comments) Reword stem question from "actions are required under" to "actions is the correct response undE!r" RLC - Reworded stem pElr NRC comments. [11/30109] Sruce F. 0-4/G, No comment NRC KIA System/E/A System 003 Reactor Coolant Pump System (RCPS) Generic KlA.~.E!I!C:~I.J NRC KIA Generic System 2.1 Conduct of Operations Number 2.1.23 RO 4.3 SRO 4.4 CFR Link (CFR: 41.10 143.5145.2/45.6) Ability to perform specifiG system and integrated plant procedures during all modes of plant operation. Page 31 of 75 Printed on 1/28/2010 at 12:02

Question #: 11 Lower Order? I-SRO Ques. # 11 ~ Selected for Exam Origin: New Past NRC Exam? Question ID: 9000008 RO SRO Student Handout? Rev. 1

1.

2.

V CAUTION V Thro 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 flow and any potential in *-surge. may result in a lowering of RCS pressure. Prompt operation of pn~ssurizer heaters (proportional and backup), should be anticipated.

1.

2.

NOTE When starting RCPs. it is desirable to maintain RCS pressure between RCP MNPSH and 265 psia, as indicated on Attachment 2, 3 or the applicable PPC NPSH display. £0 accommodate any pressure change on st~ut. If a temperature rise is expected upon RCP start, this must also be considered part of the RCS heatup. 4.4.12 After starting Reps, perform the following: SOC to RCS temperature, T351Y, must be lowered rapidly to compensate for the extra heat input from the RePs.

soc to RCS temperature, T351Y, is then raised to allow heatup to progress.

Monitor pressurizer pressure. 4.4.13 Refer To OP 230lC "Reactor Coolant Pump Operation" and PERFORM applicable actions to start selected RCPs (C-03). 4.4.14 WHEN RCPs are operating, ReferTb Attachments 1 and 2 or 3, or applicable PPC displays, and VERIFY RCS pressure to between MNPSH and 265 psia (C-03, PPC). 4.4.15 Using "SOC SYS HX FLOW CNTL, HIC-3657" (C-Ol) and "SOC SYS TOTAL FLOW, FIC-306", AOJUST SOC System 6.T to obtain value calculated in step 4.4.7 for concurrent SOC/RCP operation with O"F heatup rate. 4.4.16 As neccssmy, STABILIZE RCS temperature QR CONTINUE plant heatlip. 4.4.17 CLOSE manual disconnect switch, 89-S1652 (for SI-652, "SOC SYS suer CTMT IS0L," west wall of Control Room). 4.4.18 CLOSE "2-S1 -651, MANUAL DISCONNECf SWITCH, NS1651." (14'6" Al.lx Bldg, west wall. across from B51 enclosure) or 2201 STOP THINK ACT REVIEW Rev. 031-12 24 of 109 Page 32 of 75 Printed on 1/28/2010 at 12:02 Question #: 11 I-SRO Ques. # 11 Question ID: 9000008 RO SRO Rev. 1 ~ Selected for Exam V CAUTION V Student Handout? Origin: New

1.

Thro 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.

2.

Upon RCP start the initiation of bypass spray flow and any potential in *-surge. may result in a lowering of RCS pressure. Prompt operation of pn~ssurizer heaters (proportional and backup), should be anticipated. NOTE

1.

When starting RCPs. it is desirable to maintain RCS pressure between RCP MNPSH and 265 psia, as indicated on Attachment 2, 3 or the applicable PPC NPSH display. £0 accommodate any pressure change on st~ut.

2. If a temperature rise is expected upon RCP start, this must also be considered part of the RCS heat up.

4.4.12 After starting Reps, perform the following: SOC to RCS temperature, T351Y, must be lowered rapidly to compensate for the extra heat input from the RePs.

soc to RCS temperature, T351Y, is then raised to allow heatup to progress.

Monitor pressurizer pressure. 4.4.13 Refer To OP 230lC "Reactor Coolant Pump Operation" and PERFORM applicable actions to start selected RCPs (C-03). 4.4.14 WHEN RCPs are operating, ReferTb Attachments 1 and 2 or 3, or applicable PPC displays, and VERIFY RCS pressure to between MNPSH and 265 psia (C-03, PPC). 4.4.15 Using "SOC SYS HX FLOW CNTL, HIC-3657" (C-Ol) and "SOC SYS TOTAL FLOW, FIC-306", AOJUST SOC System 6.T to obtain value calculated in step 4.4.7 for concurrent SOC/RCP operation with O"F heatup rate. 4.4.16 As neccssmy, STABILIZE RCS temperature QR CONTINUE plant heat lip. 4.4.17 CLOSE manual disconnect switch, 89-S1652 (for SI-652, "SOC SYS suer CTMT IS0L," west wall of Control Room). 4.4.18 CLOSE "2-S1 -651, MANUAL DISCONNECf SWITCH, NS1651." (14'6" Al.lx Bldg, west wall. across from B51 enclosure) STOP THINK ACT REVIEW or 2201 Rev. 031-12 24 of 109 Lower Order? Past NRC Exam? Page 32 of 75 Printed on 1/28/2010 at 12:02

Question #: 11 Student Handout? Question ID; 9000008 RO ~ SRO Lower Order? II-SRO Ques. rI '11 Rev. 1 [;'] Selected for Exam Origin; New Past NRC Exam? Plant Heatup Conditional Actions (Sheet I or 5)

1. lE in MODE 3 AND one RCP loop is not operable, LOG in to Tech Spec Action slatemem 3.4.1.2.

2. IE at any time, RCP operation cannot continue, and it is necessary to restore SOc, PERFORM the following:

2.1 PERFORM one of the following: STOP affected RC:Ps. IE less than two RCPs will remain running. STOP all RCPs and LOG into the following. MODE 3. l~SAS 3.4.1.2 AcrION b MODE 4, TSAS 3.4~1.3 ACTION c Refer To applicable RCP NPSH Attachment, or PPC NPSH display. and CHECK NPSH for running RCPs is met. IF NPSH for running Reps is I/ot met, STOP RCPs. 2.2 IF applicable. Refer To one of the fbllowing and COOLDOWN plant to less than 300"F. EOP 2528, "Loss of Offsite Power/Loss of Forced Circulation" ICY OP 2207, "Plant Cooldown" 2.3 VERIFY all HPSI pump control switches are in "PULL TO LOCK"

3. IE at any time, during heatup, one RCP is lost (initially two RCPs operating AND SDC not i n service). PERFORM the following:

3.1 TRIP remaining RCP. 3.2 LOG into the following: MODE 3, "fSAS 3~4.1~2 ACI10N b MODE 4. TSAS 3.4.1.3 ACTION c 3.3 Refer To EOP 2528. "Loss of Om;;ite Power/Loss of Forced Circulation." 3.4 ADJUST ReS pressure to establish adequate NPSH for RCP operation as spedt1ed in Attachment 4. 3.5 IF 2 RCPs arc available tn be operated. Refer To OP 230 IC and START 2 RCPs. 3.5.1 LOG out of the following: MODE 3, TSAS 3#4. 'L2 ACTION b MODE 4, TSAS 3.4.1.3 ACTION c

Level,o! U;;e]

OP 2201 STOP THINK ACT REVIEW Rev. 031-12 Contlnuo~ 68 of 109 Page 33 of 75 Printed on 1/28/2010 at 12:02 Question #: 11 Question ID; 9000008 RO ~ SRO Student Handout? II-SRO Ques. rI '11 Rev. 1 [;'] Selected for Exam Origin; New Plant Heatup Conditional Actions (Sheet I or 5)

1. lE in MODE 3 AND one RCP loop is not operable, LOG in to Tech Spec Action slatemem 3.4.1.2.

2. IE at any time, RCP operation cannot continue, and it is necessary to restore SOc, PERFORM the following:

2.1 PERFORM one of the following: STOP affected RC:Ps. IE less than two RCPs will remain running. STOP all RCPs and LOG into the following. MODE 3. l~SAS 3.4.1.2 AcrION b MODE 4, TSAS 3.4~1.3 ACTION c Refer To applicable RCP NPSH Attachment, or PPC NPSH display. and CHECK NPSH for running RCPs is met. IF NPSH for running Reps is I/ot met, STOP RCPs. 2.2 IF applicable. Refer To one of the fbllowing and COOLDOWN plant to less than 300"F. 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. IE at any time, during heatup, one RCP is lost (initially two RCPs operating AND S DC not i n service). PERFO RM the following:

3.1 TRIP remaining RCP. 3.2 LOG into the following: MODE 3, "fSAS 3~4.1~2 ACI10N b MODE 4. TSAS 3.4.1.3 ACTION c 3.3 Refer To EOP 2528. "Loss of Om;;ite Power/Loss of Forced Circulation." 3.4 ADJUST ReS pressure to establish adequate NPSH for RCP operation as spedt1ed in Attachment 4. 3.5 IF 2 RCPs arc available tn be operated. Refer To OP 230 I C and START 2 RCPs. 3.5.1 LOG out of the following: MODE 3, TSAS 3#4. 'L2 ACTION b MODE 4, TSAS 3.4.1.3 ACTION c

Level,o! U;;e]

Contlnuo~ STOP THINK ACT REVIEW OP 2201 Rev. 031-12 68 of 109 Lower Order? Past NRC Exam? ICY Page 33 of 75 Printed on 1/28/2010 at 12:02

Question #: 12 Question ID: 9600016 -- RO ~ BRO Student Handout? ~ Lower Order? \\-SRO Ques. # 12 Rev. 0

jl Selected for Exam Origin

Mod Past NRC EKam? 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 PressLTizer Sprays in progress. " C02/3 annunciator in alarm; D-37, npZR PRESSURE SELECTED CHANNEL DEVIATION HI/LO" " Channe~1 ny" Pressurizer Level and Pressure controlling normally. Then, during the load reduction, Pressurizer Level Channel "X" falls to zero (0) and the following occur: " All control syst13ms 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 tile 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 ChannellX" from the leak rate calculation, to ensure valid trending of RCS Leak Rate by the PPC. D C Per the ARP for A-38; place the standby charging pumps in "Pull-To-Lock", to prevent the rate of the plant clownpower from accelerating above 50%/hr. D D Per AOP-257S, Rapid Downpower; shift pressurizer heater control to channel "Y" and restore the pressure controller setpoint, to prevent a plant trip on TM/LP. 'Ju;tificaticm I 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 "Fc)rcing PZR Sprays". B - WRONG: The high r<lte of power change exceeds the PPC program capabilities for calculating an accurate RCS leak rate. Plausible: The failed PZH 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 charging 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 setpoint, as necessary, to maintain pressure at the desired value. References I ARP-2590B-2150 Alarm C-38, PZR Level La-La 'Comments and Question Modification History I Bob K. - D-3/C (change to match given reactivity plan) Modified stem and cholice "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 downpower 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 downpower rate in dlstractor "c" to 50%/hr per recommendation. RJA Angelo - D-3/C; No comments. Bruce F. - D-3/C, No comment Page 34 of75 Printed on 1/28/2010 at 12:02 Question #: 12 \\-SRO Ques. # 12 Question ID: 9600016 -- RO ~ BRO Rev. 0

jl Selected for Exam Student Handout?

~ Lower Order? Origin: Mod Past NRC EKam? 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 PressLTizer Sprays in progress. " C02/3 annunciator in alarm; D-37, npZR PRESSURE SELECTED CHANNEL DEVIATION HI/LO" " Channe~1 ny" Pressurizer Level and Pressure controlling normally. Then, during the load reduction, Pressurizer Level Channel "X" falls to zero (0) and the following occur: " All control syst13ms 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 tile 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 ChannellX" from the leak rate calculation, to ensure valid trending of RCS Leak Rate by the PPC. D C Per the ARP for A-38; place the standby charging pumps in "Pull-To-Lock", to prevent the rate of the plant clownpower from accelerating above 50%/hr. D D Per AOP-257S, Rapid Downpower; shift pressurizer heater control to channel "Y" and restore the pressure controller setpoint, to prevent a plant trip on TM/LP. 'Ju;tificaticm I 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 "Fc)rcing PZR Sprays". B - WRONG: The high r<lte of power change exceeds the PPC program capabilities for calculating an accurate RCS leak rate. Plausible: The failed PZH 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 charging 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 setpoint, as necessary, to maintain pressure at the desired value. References I ARP-2590B-2150 Alarm C-38, PZR Level La-La 'Comments and Question Modification History I Bob K. - D-3/C (change to match given reactivity plan) Modified stem and cholice "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 downpower 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 dlstractor "c" to 50%/hr per recommendation. RJA Angelo - D-3/C; No comments. Bruce F. - D-3/C, No comment Page 34 of75 Printed on 1/28/2010 at 12:02

SR() Exam Questions Only (No "Parentsl'Ot:!'lQri;ipl.f$;f)'i'~

Question #: 12 Question ID; 9600016 RO ~SRO Student Handout? ~ Lower Order? I-SRO Ques. 1* 12 Rev. o ~ Selected for Exam Origin: Mod Past NRC Exam? NRC KIA System/E/A System 004 Chemical and Volume Control System Number A2.15 RO 3.5 SRO 3.7 CFRLink (CFR: 41.5/4315/45/3/45/5) Ability to (a) predict the impacts of the following malfunctions or operations on the CVCS; and (b) based on those predictions, USE! procedures to correct, control, or mitigate the consequences of those malfunctions or operations: High or low PZR level Page 35 of 75 Printed on 1/28/2010 at 12:02

SR() Exam Questions Only (No "Parentsl'Ot:!'lQri;ipl.f$;f)'i'~

Question #: 12 I-SRO Ques. 1* 12 Question ID; 9600016 RO ~SRO Student Handout? ~ Lower Order? Rev. o ~ Selected for Exam Origin: Mod Past NRC Exam? NRC KIA System/E/A System 004 Chemical and Volume Control System Number A2.15 RO 3.5 SRO 3.7 CFRLink (CFR: 41.5/4315/45/3/45/5) Ability to (a) predict the impacts of the following malfunctions or operations on the CVCS; and (b) based on those predictions, USE! procedures to correct, control, or mitigate the consequences of those malfunctions or operations: High or low PZR level Page 35 of 75 Printed on 1/28/2010 at 12:02

SRO Exam Questions Only (No UParents ff Or "Originals") Question #: 12 Question ID: 9600016 []RO ~ SRO D Student Handout? ~ Lower Order? J-SRO Ques. # 12 Rev. 0 ~ Selected for Exam Origin: Mod D Past NRC EKam? 02/16104 03/04/04 Approval Date Effective Date Setpol nt: 20% C-38 PRESSURIZER CHX LEVELLO LO AUTOMATIC: FUNCTIONS I. IF "SEL SW" is in x+Y' position, all heaters dc-energize. C.QBRECTIVE ACTIONS I. OBSERVE actual level on pressurizer level recorder, LR-IIO, pressurizer level controllers (C-OJ) and ppe.

2.

IF annunciator is flot valid, SHIff pressurizer level control to channel "'C' 2.1 SHIff pressurizer heater control "SEL SW" to channel Y.

3.

VERIFY the following: Available hackup charging pumps arc running (C-02). Letdown now is at minimum of 28 gpm on "LTDN FLO~ 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.

WHEN level rises to 4% below setpoint, VERIFY second back up charging pump stops.

7.

WHEN levc\\ ri;cs to 3% below setpoint, VERIFY first back lip charging pump stops.

8.

VERIFY level is restored to setpoinl.

9.

IF alarm was caused by channel X malfunctioning, SUBMIT cfiouble Report to I&C Department.

10.

Refer To Technical Specifications LeOs 3.3.3.5 and 3.3.3.R to determine ACTION Statem~~nt requirements. SUPPORTING INFORMATION I. Initiating Devices LC-llOXL

2.

CompUTer Points lJI0X'

3.

Possible Causes Controller malfunction ReS inventory loss

4.

Technical Specifications LeOs: 3.4.4,3.3.3.5 and 3.3.3.8

5.

Procedures OP 2304A, "Volume Control Portion of CVCS" AOP 256R, "Reactor Coolant System Leak"

6.

Control Room Drawings

!5203-32007, sh~ 57

7.

Annunciator Card Location: TBIO-JlZ ARP 2590B-215 Rev. {lOn Page I of 1 ---.--.------~---- --------------------------- Page 36 of 75 Printed on 1/28/2010 at 12:02 SRO Exam Questions Only (No UParents ff Or "Originals") Question #: 12 Question ID: 9600016 []RO ~ SRO D Student Handout? ~ Lower Order? J-SRO Ques. # 12 Rev. 0 ~ Selected for Exam Origin: Mod D Past NRC EKam? Approval Date 02/16104 Effective Date 03/04/04 Set pol nt: 20% C-38 PRESSURIZER CHX LEVELLO LO AUTOMATIC: FUNCTIONS I. IF "SEL SW" is in x + Y' position, all heaters dc-energize. C.QBRECTIVE ACTIONS I. OBSERVE actual level on pressurizer level recorder, LR-IIO, pressurizer level controllers (C-OJ) and ppe.

2.

IF annunciator is flot valid, SHIff pressurizer level control to channel "'C' 2.1 SHIff pressurizer heater control "SEL SW" to channel Y.

3.

VERIFY the following: Available hackup charging pumps arc running (C-02). Letdown now is at minimum of 28 gpm on "LTDN FLO~ 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.

WHEN level rises to 4% below setpoint, VERIFY second back up charging pump stops.

7.

WHEN levc\\ ri;cs to 3% below setpoint, VERIFY first back lip charging pump stops.

8.

VERIFY level is restored to setpoinl.

9.

IF alarm was caused by channel X malfunctioning, SUBMIT cfiouble Report to I&C Department.

10.

Refer To Technical Specifications LeOs 3.3.3.5 and 3.3.3.R to determine ACTION Statem~~nt requirements. SUPPORTING INFORMATION I. Initiating Devices LC-llOXL

2.

CompUTer Points lJI0X'

3.

Possible Causes Controller malfunction ReS inventory loss

4.

Technical Specifications LeOs: 3.4.4,3.3.3.5 and 3.3.3.8

5.

Procedures OP 2304A, "Volume Control Portion of CVCS" AOP 256R, "Reactor Coolant System Leak"

6.

Control Room Drawings

!5203-32007, sh~ 57

7.

Annunciator Card Location: TBIO-JlZ ARP 2590B-215 Rev. {lOn Page I of 1 ---.--.------~---- ---------------------------- Page 36 of 75 Printed on 1/28/2010 at 12:02

Question #: 12 Student Handout? ~ Lower Order? Question ID: 9600016 RO ~SRO J-SRO Ques. If 12 Rev. 0 ~ Selected for Exam Otfgin: Mod o Past NRC Enm? A-38

8.

IEindicated high or low level was L'(lUsed by controller or transmitter malfunction (other than Reactor Regulating System inputs), PERFORM the following: !'.I SI.UFT "LTDN FLOW CNTL, I lIC-l Hr' in "MAN" (C-02). K2 ADJUST "LTDN CNTL. HIC-tto" to stabilize Pressurizer kvcl and LI:!tdown flow (C-02). 8.3 IE desired, COMMENCE forcing Pressurizer sprays. 8.4 51 11FT Pressurizer level control to channel "Y" (C-03). 8.5 RESTORE Letdown to automatic as follows: 8.5.1 ADJUST hias to "0". using bhiLi< thumb\\,.,heel. 8.5.2 SHIFT "LTDN FLOW CNTL, HIC-IIO" to "AUTO:' 8.5.3 ADJUST bias to restore Prcssurizer levcl to setpoint. 8.5.4 SHIFf Pressurizer heater control "SEL SW" to channel "Y."

9.

As necessary, RESET the following Pressurizer heater breakers: "PROP HTR GROUP 1,. "PROP HTR GROUP 2" "BACKUP HTRS GROUP l" '"BACKUP HTRS GROUP 2" "BACKUP HTRS GROUP 3" "BACKUP HTRS GROUP 4"

10.

instrument malfunction is determined not to be the cause of low level, Refer lb the following, as applicable: AOP 2:')12, "Loss of All Charging" AOP 2568, "Reactor Coolant System Leak" AOP 2569, "Steam Generator Tube Leak" II. actual level was high or low, VERIFY level is restored to normal.

12.

alarm was caused by channel "X" malfunctioning, SUBMIT lfouble Report to Instrumentation & Control Department. ARP 2590B-213 Rev. 001-00 Page 37 of 75 Printed on 1/28/2010 at 12:02 Question #: 12 Question ID: 9600016 J-SRO Ques. If 12 Rev. 0 RO ~SRO ~ Selected for Exam Student Handout? Otfgin: Mod ~ Lower Order? o Past NRC Enm? A-38

8.

IE indicated high or low level was L'(lUsed by controller or transmitter malfunction (other than Reactor Regulating System inputs), PERFORM the following: !'.I SI.UFT "LTDN FLOW CNTL, I lIC-l Hr' in "MAN" (C-02). K2 ADJUST "LTDN CNTL. HIC-tto" to stabilize Pressurizer kvcl and LI:!tdown flow (C-02). 8.3 IE desired, COMMENCE forcing Pressurizer sprays. 8.4 51 11FT Pressurizer level control to channel "Y" (C-03). 8.5 RESTORE Letdown to automatic as follows: 8.5.1 ADJUST hias to "0". using bhiLi< thumb\\,.,heel. 8.5.2 SHIFT "LTDN FLOW CNTL, HIC-IIO" to "AUTO:' 8.5.3 ADJUST bias to restore Prcssurizer levcl to setpoint. 8.5.4 SHIFf Pressurizer heater control "SEL SW" to channel "Y."

9.

As necessary, RESET the following Pressurizer heater breakers: "PROP HTR GROUP 1,. "PROP HTR GROUP 2" "BACKUP HTRS GROUP l" '"BACKUP HTRS GROUP 2" "BACKUP HTRS GROUP 3" "BACKUP HTRS GROUP 4"

10.

instrument malfunction is determined not to be the cause of low level, Refer lb the following, as applicable: AOP 2:')12, "Loss of All Charging" AOP 2568, "Reactor Coolant System Leak" AOP 2569, "Steam Generator Tube Leak" II. actual level was high or low, VERIFY level is restored to normal.

12.

alarm was caused by channel "X" malfunctioning, SUBMIT lfouble Report to Instrumentation & Control Department. Page 37 of 75 ARP 2590B-213 Rev. 001-00 Printed on 1/28/2010 at 12:02

Question #: 13 RO ~ SRO Student Handout? ~ Lower Order? Question 10; 53730 J-SRO Ques. # 13 Rev. 4

.j-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 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. i,,1 C Trip the reactor and Go to EOP 2525, "Standard Post Trip Actions". D D Insert all CEAs per OP-2206, "Reactor Shutdown" and notify RE. Justification J C - CORRECT OP 2202, "Reactor Startup" Conditional Actions require 1l:iI;!.QiJ]g the reactor, and transilioning 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. o - 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). Referencesj 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-2IW, K (Did not remember the actual trip criteria value.) Angelo - D-3/C; No comments. Bruce F. - D-2/C. No comment NRC KiA System/l:/A System 012 Reactor Protection System --~--~] ~!,ner~ KIA Sel~c.:!~(j System 2.4 Emergency Procedures IPlan Number 2.4.4 R04.5 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 procedures. NRC KiA Generic Page 38 of 75 Printed on 1/2812010 at 12:02 Question #: 13 J-SRO Ques. # 13 Question 10; 53730 Rev. 4 RO ~ SRO

.j-Selected for Exam Student Handout?

Origin: Bank ~ Lower Order? 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 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. i,,1 C Trip the reactor and Go to EOP 2525, "Standard Post Trip Actions". D D Insert all CEAs per OP-2206, "Reactor Shutdown" and notify RE. Justification J C - CORRECT OP 2202, "Reactor Startup" Conditional Actions require 1l:iI;!.QiJ]g the reactor, and transilioning 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. o - 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). Referencesj 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-2IW, K (Did not remember the actual trip criteria value.) Angelo - D-3/C; No comments. Bruce F. - D-2/C. No comment NRC KiA System/l:/A System 012 Reactor Protection System --~--~] ~!,ner~ KIA Sel~c.:!~(j NRC KiA Generic System 2.4 Emergency Procedures IPlan Number 2.4.4 R04.5 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 procedures. Page 38 of 75 Printed on 1/2812010 at 12:02

Question #: 13 RO ~ SRO Student Handout? ~ Lower Order? Question ID; 53730 II-SRO Ques. # 13 Rev. 4 [~ Selected for Exam Origin: Bank Past NRC Exam? Reactor Startup Conditional Actions (Sheet 1 of 3)

l. IE at any time, the following conditions occur, PERFORM the specified action:

IF Tavg lowers to between 515 and 525 "F AND the reactor is critical, Refer to OP 2619A-OOl, "Controi Room Daily Smveinam:e," and RECORD RCS temperature once every hour. JF Tavg lowers to k~s than 5]5 OF AND the reactor is clitical, PERFORM the following: RAISE 1avg to greater than 515 OF within 15 minutes. Tavg is not greater than 515 OF within is minutes, PLACE plant in HOT STANDBY condition within the next 15 minutes. Refer lh T/S leO 3J.J,+5 and DETERMINE appUcabHity. an uncontroUed cooldown occurs (Tc less than son OF), PERFORM the following: TRIP reactor and INITIATE EOP 2525, "Standard Post Trip Actiof1.<;." STOP one of the 4 operuting RCPs (C-04). Refer Til AOP 2558, "Emergency Boration,'" and INITIATE emergent)' boratiofl. Refer III TIS LCO 3.4.9.1 and DETERMINE applicability.

2. IF at any time a sustained SUR of 1.0 dpm is achk:ved, TRIP reactor and Go'TI, EOP 2525, "Standard Post Trip Actions."

3. IF at any time during reactor startup, it appears that criticality is reached, or is predicted to be rcached, outside plus or minus 0.5% 8g ({}.9% Ag for initial startup after refueling) band of ECp, PERFORM the following:

3.1 INSERT all CEA rc!;;tulaling groups in sequence (C-04). 3.2 REQUEST Chemistry Department sample and determine RCS boron concentration. 3.3 INITIATE a CR for Reactivity Management tracking. 3.4 Refer lb OP 2208, "Reactivity Calculations" and, independent of CEA position, VERIFY adequate SHUTDOWN MARGIN using OP 2:208-013, "Shutdown Margin Determination." 3.5 NOTIFY Reactor Engineering. LevelOf~ Contim~ STOP THINK ACT REVIEW OP 2202 Rev. 021-06 36 of 56 Page 39 of 75 Printed on 1/28/2010 at 12:02 Question #: 13 II-SRO Ques. # 13 Question ID; 53730 RO ~ SRO [~ Selected for Exam Rev. 4 Student Handout? Origin: Bank Reactor Startup Conditional Actions (Sheet 1 of 3) ~ Lower Order? Past NRC Exam?

l. IE at any time, the following conditions occur, PERFORM the specified action:

IF Tavg lowers to between 515 and 525 "F AND the reactor is critical, Refer to OP 2619A-OOl, "Controi Room Daily Smveinam:e," and RECORD RCS temperature once every hour. JF Tavg lowers to k~s than 5]5 OF AND the reactor is clitical, PERFORM the following: RAISE 1avg to greater than 515 OF within 15 minutes. Tavg is not greater than 515 OF within is minutes, PLACE plant in HOT STANDBY condition within the next 15 minutes. Refer lh T/S leO 3J.J,+5 and DETERMINE appUcabHity. an uncontroUed cooldown occurs (Tc less than son OF), PERFORM the following: TRIP reactor and INITIATE EOP 2525, "Standard Post Trip Actiof1.<;." STOP one of the 4 operuting RCPs (C-04). Refer Til AOP 2558, "Emergency Boration,'" and INITIATE emergent)' boratiofl. Refer III TIS LCO 3.4.9.1 and DETERMINE applicability.

2. IF at any time a sustained SUR of 1.0 dpm is achk:ved, TRIP reactor and Go'TI, EOP 2525, "Standard Post Trip Actions."

3. IF at any time during reactor startup, it appears that criticality is reached, or is predicted to be rcached, outside plus or minus 0.5% 8g ({}.9% Ag for initial startup after refueling) band of ECp, PERFORM the following:

3.1 INSERT all CEA rc!;;tulaling groups in sequence (C-04). 3.2 REQUEST Chemistry Department sample and determine RCS boron concentration. 3.3 INITIATE a CR for Reactivity Management tracking. 3.4 Refer lb OP 2208, "Reactivity Calculations" and, independent of CEA position, VERIFY adequate SHUTDOWN MARGIN using OP 2:208-013, "Shutdown Margin Determination." 3.5 NOTIFY Reactor Engineering. LevelOf~ Contim~ STOP THINK ACT Page 39 of 75 REVIEW OP 2202 Rev. 021-06 36 of 56 Printed on 1/28/2010 at 12:02

Question #: 14 D Student Handout? ~ Lower Order? Question ID; 9000021 RO ~ SRO J-SRO Ques. # 14 Rev. 0 ~ Selected for Exam Origin: New Past NRC Exam? 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 (SDC) operation.
"B" LPSI pump in standby, aligned for SDC use.
"A" train of SpEint 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 trips on breaker overload. The Unit Supervisor (US) then directs the RO to recover SDC using the "B" Containment Spray (CS) pump. Which of tile following additional directions must the US give, while SDC flow is being supplied by a CS pump? ~ A All fuel movement in containment must remain secured. B SDC supplementing of SFP cooling must be secured. C Containment must remain evacuated of non-essential personnel. D Contarnment Closure must be fully set with all access doors closed. 'Justification J A - CORRECT; When SDC is being supplied by a CS pump, does not constitute an OPERABLE train of SDC. Therefore, ali fuel movement in CTMT must be secured. B - WRONG; A CS pum p 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 melY 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; GTMT 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 Clcsure must be set on initial loss of SDC flow, but once heat removal is regained, it may stop. Plausible; Examinee melY believe that with a CS pump supplying RHR needs (SDC not operable), CTMT closure must be maintained. References "! AOP-2572, Pg. 3, Discussion Section and Pages 28 &31 'Comments and Question Modification History I Bob K. - D-3/W (Did not remember SDC not Operable with CS Pump supplying). Corrected a typo in Distractor B. - RJA Bill M. - D-4/W, G (Did not realize that SDC was inoperable, which requires CORE ALTERATIONS to remain suspended.) Angelo - D-5/C: Difficult IJut fair. Bruce F. - D-4/C, 50/50 (A and B) NRC KIA System/E/A System 026 Containment Spray System (CSS) ~-=----il ~~_~~ Selected..; NRC KIA Generic System 2.4 Emergency Procedures IPlan Number 2.4.9 RO 3.8 SRO 4.2 CFRLink (CFR:41.10/43.5/45.13) Knowledge of low powel'/shutdown implications in accident (e.g., loss of coolant accident or loss of reSidual heat removal) mitigation strategies. Page 40 of 75 Printed on 1/28/2010 at 12:02 Question #: 14 Question ID; 9000021 RO ~ SRO J-SRO Ques. # 14 Rev. 0 ~ Selected for Exam 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 (SDC) operation.
"B" LPSI pump in standby, aligned for SDC use.
"A" train of SpEint Fuel Pool (SFP) cooling in service.

D Student Handout? ~ Lower Order? 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 SDC using the "B" Containment Spray (CS) pump. Which of tile following additional directions must the US give, while SDC flow is being supplied by a CS pump? ~ A All fuel movement in containment must remain secured. B SDC supplementing of SFP cooling must be secured. C Containment must remain evacuated of non-essential personnel. D Contarnment Closure must be fully set with all access doors closed. 'Justification J A - CORRECT; When SDC is being supplied by a CS pump, does not constitute an OPERABLE train of SDC. Therefore, ali fuel movement in CTMT must be secured. B - WRONG; A CS pum p 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 melY 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; GTMT 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 Clcsure must be set on initial loss of SDC flow, but once heat removal is regained, it may stop. Plausible; Examinee melY believe that with a CS pump supplying RHR needs (SDC not operable), CTMT closure must be maintained. References "! AOP-2572, Pg. 3, Discussion Section and Pages 28 & 31 'Comments and Question Modification History I Bob K. - D-3/W (Did not remember SDC not Operable with CS Pump supplying). Corrected a typo in Distractor B. - RJA Bill M. - D-4/W, G (Did not realize that SDC was inoperable, which requires CORE ALTERATIONS to remain suspended.) Angelo - D-5/C: Difficult IJut fair. Bruce F. - D-4/C, 50/50 (A and B) NRC KIA System/E/A System 026 Containment Spray System (CSS) ~-=----il ~~_~~ Selected..; NRC KIA Generic System 2.4 Emergency Procedures IPlan Number 2.4.9 RO 3.8 SRO 4.2 CFRLink (CFR:41.10/43.5/45.13) Knowledge of low powel'/shutdown implications in accident (e.g., loss of coolant accident or loss of reSidual heat removal) mitigation strategies. Page 40 of 75 Printed on 1/28/2010 at 12:02

Question #: 14 Question ID; 9000021 RO .~. SRO -- Student Handout? [Y'i Lower Order? J-SRO Ques. II 14 Rev. o Ii'] Selected for Exam Origin: New Millstone Unit 2 AOP 2572 Revision 009-04 of Shutdown Cooling Page 3 0(70 PURPOSE 1.1 Objective This procedure provides actions for recovering from a partial or tolalloss of shutdown cool ing. J.2 Ois(:ussion During SOC operation. there may not be flow past the loop RIDs. Core inlet and outlet temperatures are accurately measured during those conditions using SOC to RCS temperature, T351 Y, and RCS to SOC temperature, T351X, respectively. The average of these indicators provides a temperature that is equivalent to the average RCS lempcrature in the core. Containment Closure is established when all of the following conditions exist: The equipment door is closed 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 atmosphere is either: Closed by a manual or automatic isolation valve, blind flange, or equivalent, tlT Capable ofhcing closed under administrative control The usc of the CS pump for decay heat removal docs not meet the del1nition of an Operahle SOC train (LCO 3.9.R). Therefore no fuel movement is permitted when a CS pump is aligned to SOC per this procedure. 1.3 Aplfllicahility This procedure is applicable in MODEs 4, 5, 6 and Defueled. Use of the CS pumps is limited to MODE 6 and Defucled. Level of u$el STOP THINK ACT Continuo~ Page 41 of 75 Printed on 1/28/2010 at 12:02 Question #: 14 J-SRO Ques. II 14 Question ID; 9000021 RO .~. SRO -- Student Handout? [Y'i Lower Order? Rev. o Ii'] Selected for Exam Origin: New Millstone Unit 2 AOP 2572 Revision 009-04 of Shutdown Cooling Page 3 0(70 PURPOSE 1.1 Objective This procedure provides actions for recovering from a partial or tolalloss of shutdown cool ing. J.2 Ois(:ussion During SOC operation. there may not be flow past the loop RIDs. Core inlet and outlet temperatures are accurately measured during those conditions using SOC to RCS temperature, T351 Y, and RCS to SOC temperature, T351X, respectively. The average of these indicators provides a temperature that is equivalent to the average RCS lempcrature in the core. Containment Closure is established when all of the following conditions exist: The equipment door is closed and held in place by a minimum of four bolts. A minimum of one door in each airlock is closed. Each penetration providing direct access from the containment atmosphere to the outside atmosphere is either: Closed by a manual or automatic isolation valve, blind flange, or equivalent, tlT Capable ofhcing closed under administrative control The usc of the CS pump for decay heat removal docs not meet the del1nition of an Operahle SOC train (LCO 3.9.R). Therefore no fuel movement is permitted when a CS pump is aligned to SOC per this procedure. 1.3 Aplfllicahility This procedure is applicable in MODEs 4, 5, 6 and Defueled. Use of the CS pumps is limited to MODE 6 and Defucled. Level of u$el Continuo~ STOP THINK ACT Page 41 of 75 Printed on 1/28/2010 at 12:02

Question #: 14 ~SRO Student Handout? Lower Orda r? Question ID: 9000021 [=RO I-SRO Ques. # 14 Rev. 0 ~ Selected for Exam Origin: New Past NRC Exam? Millstone Unit 2 AOP 2572 Revision 009-04 Loss of Shutdown Cooling INSTRUC'TIONS 5.10 (conlinu*ed)

2) CHECK pressure at 2-RW-6fi, "SFPC/RW Purification Return Sample Stop," less than 30 psig.

Table JJ) Fuel Assemblies Flow splits with SFPC in service Off-Loaded Flow to RFP Flow to SFP 0-80 1700 81-170 1400 171-217 lIon

b. lEa pump is in service,

() 300 600 Page 28 of 70 CONTINGENCY ACTIONS

2) IF pressure at 2-RW -66 is greater than 30 psig, PERFORM the following::

a.

THROTILE 2-RW -15, "SDC to SFPC Stop" to obtain less than 30 psig at 2-RW-66.

b.

CONTACT Engineering for additional guidtUlCe on decay heat removaL Flow splits with SFPC flot in service Flow to RFP Flow to SFP 1250 450 750 950 400 130() PERFORM the followin gr:

1) THROITLE the following valves to ohtain Table 1.0 "

flow splits: 51 -615, "LPSI IN] VLVS" LOOP lA SI -625, "LPSIINJ VLVS" LOOP lB SI-635, "LPSI IN] VLVS" LOOP 2A SI -645, "LPS} IN] VLVS" LOOP 2B

2-RW-JS, "SDCto SPPC Stop"

( continue.) LevelofUi';'l STOP THINK Continuo~ ACT REVIEW Page 42 of 75 Printed on 1/28/2010 at 12:02 Question #: 14 Question ID: 9000021 [=RO ~SRO Student Handout? Lower Orda I-SRO Ques. # 14 Rev. 0 ~ Selected for Exam Origin: New Past NRC Ex Millstone Unit 2 AOP 2572 Revision 009-04 Loss of Shutdown Cooling INSTRUC'TIONS 5.10 (conlinu*ed)

2) CHECK pressure at 2-RW-6fi, "SFPC/RW Purification Return Sample Stop," less than 30 psig.

Table JJ) Fuel Assemblies Flow splits with SFPC in service Off-Loaded Flow to RFP Flow to SFP 0-80 1700 81-170 1400 171-217 lIon

b. lEa pump is in service, PERFORM the followin r: g

1) THROITLE the following valves to ohtain Table 1.0 "

flow splits: 51 -615, "LPSI IN] VLVS" LOOP lA SI -625, "LPSIINJ VLVS" LOOP lB SI-635, "LPSI IN] VLVS" LOOP 2A SI -645, "LPS} IN] VLVS" LOOP 2B 2-RW-JS, "SDCto SPPC Stop" ( continue.) () 300 600 LevelofUi';'l Continuo~ STOP THINK ACT Page 28 of 70 CONTINGENCY ACTIONS

2) IF pressure at 2-RW -66 is greater than 30 psig, PERFORM the following::

a.

THROTILE 2-RW -15, "SDC to SFPC Stop" to obtain less than 30 psig at 2-RW-66.

b.

CONTACT Engineering for additional guidtUlCe on decay heat removaL Flow splits with SFPC flot in service Flow to RFP Flow to SFP 1250 450 750 950 400 130() REVIEW r? am? Page 42 of 75 Printed on 1/28/2010 at 12:02

Question #: 14 Question ID: 9000021 RO ~ SRO D Student Handout? ~ Lower Order? I-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 31 of70 INSTRUC'TIONS CONTINGENCY ACTIONS NOTE 2-S1--306 has designed lcakhy that diverts now around the SDC heat exchangers and could challenge heat removal with CS pumps supplying SDC. _5_17 IF it CS pump is in scrvitc on SDC Hnd sufficient cooling cannot be ohtain(~d with 2-S1 -306 closed, CLOSE the applicable LPSI to SDC heat exchanger isolation valve: 2-S1-452, LPSI Pump Di5charge to "l'\\-' SOC Heat Exchanger 2 -,SI -453, I..iPSl Plln1p Discharge to "B" SOC Ileal Exchanger _5J8 REPEATstcps5.12 through 5.17 as needed to control RCS temperature. _5.19 WI*IEN ready to shift SOC from a CS pump to a LPSI pump. PERFORM Attachment R, "Realigning LPSI to Supply SOC and SFPC." _5.20 WHEf-t RCS pressure is stahle AND ReS temperature is less than 200°F"nd stable, STOP Containment Closure activitks. _5.21 Go lb Section 10.0. Level ofU~ STOP THINK ACT REVIEW Continuo~ Page 43 of 75 Printed on 1/28/2010 at 12:02 Question #: 14 I-SRO Ques. If 14 Question ID: 9000021 RO ~ SRO D Student Handout? ~ Lower Order? Rev. 0 ~ Selected for Exam Origin: New Past NRC Exam? Millstone Unit 2 Loss of Shutdown Cooling AOP 2572 Revision 009-04 Page 31 of70 INSTRUC'TIONS CONTINGENCY ACTIONS NOTE 2-S1--306 has designed lcakhy that diverts now around the SDC heat exchangers and could challenge heat removal with CS pumps supplying SDC. _5_ 17 IF it CS pump is in scrvitc on SDC Hnd sufficient cooling cannot be ohtain(~d with 2-S1 -306 closed, CLOSE the applicable LPSI to SDC heat exchanger isolation valve: 2-S1 -452, LPSI Pump Di5charge to "l'\\-' SOC Heat Exchanger 2 -,SI -453, I..iPSl Plln1p Discharge to "B" SOC Ileal Exchanger _5J8 REPEATstcps5.12 through 5.17 as needed to control RCS temperature. _5.19 WI*IEN ready to shift SOC from a CS pump to a LPSI pump. PERFORM Attachment R, "Realigning LPSI to Supply SOC and SFPC." _5.20 WHEf-t RCS pressure is stahle AND ReS temperature is less than 200°F"nd stable, STOP Containment Closure activitks. _5.21 Go lb Section 10.0. Level ofU~ Continuo~ STOP THINK ACT Page 43 of 75 REVIEW Printed on 1/28/2010 at 12:02

Question #: 15 o Student Handout? o Lower Order? Question ID: 9000010 RO ~SRO I-SRO Que~. 11 15 Rev. o ~. Selected for Exam Origin: New Past NRC Exam? The requimd survEliliance is being preformed on the "A" Service Water Strainer Flush Valve, 2-SW-90A. The data indicates the valve stroke time is slightly above the Maximum 'Normal Limit", but is within the Operational Readiness Results (aka; Acceptable Limits). Which of the following describes the condition of the "A" Service Water Strainer Flush Valve, 2-SW-90A and the required action? A The "A" ServiGe Water Strainer Flush Valve, 2-SW-90A is considered 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 Strainer Flush Valve, 2*SW-90A to OPERABLE. B The "A" Service Water Strainer Flush Valve, 2-SW-90A is considered inoperable. Obtain a different set of test equipment and immediately retest the "A" Service Water Strainer Flush Valve, 2-SW-90A, to verify that the previous data was accurate. C The "A" Service Water Strainer Flush Valve, 2-SW-90A is considered OPERABLE. Place the "A" Service Water Strainer in an Augmented Testing Program and test weekly to ensure the "A" Service Water Strainel" Flush Valve, 2-SW-90A, remains OPERABLE. ~ 0 The "P," Service Water Strainer Flush Valve, 2-SW-90A is considered OPERABLE. If the "An Service Water Strainel-Flush Valve, 2-SW-90A, falls outside the Normal Limits on an immediate retest, then the nAil Servic!~ Water Strainer Flush Valve, 2-SW-90A is inoperable. Justification I o IS CORRECT; Per SF' 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" stroke time limit has occurred. then the component is inoperable. A is incorrect; Even though the Service Water valve 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 due to a failure on the first test. Plausible because the examinee may consider the component inoperable from the first set of failed data. If the component is considered inoperable, then 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 el!aminee 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 valve to be considered OPERABLE. The Service Water valvH may be tested on a more freqlJent basis; however, the stoke time must be performed immediately after the initial failure to ensure it meets the "Normal" limit. Plausible because the el!aminee may not be aware of the need to immediately perform a second stoke time test. References I SP 2612A, "A" Service Water Pump Tests iComments and Question Modification History I Bob K. - D-3fW (Reword answer to say "immediate retest"). Reworded correct answer to include the term "immediate retest". RJA Bill M. - D-3fW. K (Didn't realize an immediate retest was required.) Angelo - D-4/C; No comments. Bruce F. - D-3/C. ChangE; "A" Service Water Pump to "A" Service Water Header. Changed "A" Service Water Pump to "A" Service Water Header where applicable. - RJA (12117/09) Cliff C. - Change stem to make surveillance be a regular "timed" testing of component, as opposed to a "post-repair retest", to ensure examinee does not conclude a new "base line" for data is being established. Modified stem per reviE,wer comments. - RLC 2/22/09 NRC KiA System/E/A System 076 Service Water System (SWS) I Generic KIA Selectee! 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 surveilianGe procedures. Page 44 of 75 Printed on 1/28/2010 at 12:02 Question #: 15 I-SRO Que~. 11 15 Question ID: 9000010 Rev. o RO ~SRO ~. Selected for Exam o Student Handout? o Lower Order? Origin: New Past NRC Exam? The requimd survEliliance is being preformed on the "A" Service Water Strainer Flush Valve, 2-SW-90A. The data indicates the valve stroke time is slightly above the Maximum 'Normal Limit", but is within the Operational Readiness Results (aka; Acceptable Limits). Which of the following describes the condition of the "A" Service Water Strainer Flush Valve, 2-SW-90A and the required action? A The "A" ServiGe Water Strainer Flush Valve, 2-SW-90A is considered 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 Strainer Flush Valve, 2*SW-90A to OPERABLE. B The "A" Service Water Strainer Flush Valve, 2-SW-90A is considered inoperable. Obtain a different set of test equipment and immediately retest the "A" Service Water Strainer Flush Valve, 2-SW-90A, to verify that the previous data was accurate. C The "A" Service Water Strainer Flush Valve, 2-SW-90A is considered OPERABLE. Place the "A" Service Water Strainer in an Augmented Testing Program and test weekly to ensure the "A" Service Water Strainel" Flush Valve, 2-SW-90A, remains OPERABLE. ~ 0 The "P," Service Water Strainer Flush Valve, 2-SW-90A is considered OPERABLE. If the "An Service Water Strainel-Flush Valve, 2-SW-90A, falls outside the Normal Limits on an immediate retest, then the nAil Servic!~ Water Strainer Flush Valve, 2-SW-90A is inoperable. Justification I o IS CORRECT; Per SF' 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" stroke time limit has occurred. then the component is inoperable. A is incorrect; Even though the Service Water valve 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 due to a failure on the first test. Plausible because the examinee may consider the component inoperable from the first set of failed data. If the component is considered inoperable, then 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 el!aminee 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 valve to be considered OPERABLE. The Service Water valvH may be tested on a more freqlJent basis; however, the stoke time must be performed immediately after the initial failure to ensure it meets the "Normal" limit. Plausible because the el!aminee may not be aware of the need to immediately perform a second stoke time test. References I SP 2612A, "A" Service Water Pump Tests iComments and Question Modification History I Bob K. - D-3fW (Reword answer to say "immediate retest"). Reworded correct answer to include the term "immediate retest". RJA Bill M. - D-3fW. K (Didn't realize an immediate retest was required.) Angelo - D-4/C; No comments. Bruce F. - D-3/C. ChangE; "A" Service Water Pump to "A" Service Water Header. Changed "A" Service Water Pump to "A" Service Water Header where applicable. - RJA (12117/09) Cliff C. - Change stem to make surveillance be a regular "timed" testing of component, as opposed to a "post-repair retest", to ensure examinee does not conclude a new "base line" for data is being established. Modified stem per reviE,wer comments. - RLC 2/22/09 NRC KiA System/E/A System 076 Service Water System (SWS) I Generic KIA Selectee! 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 surveilianGe procedures. Page 44 of 75 Printed on 1/28/2010 at 12:02

Question #: II-SRO Ques. # 15 15 Question ID; Rev. 9000010 RO ~SRO 0 ~ Selected for Exam Student Handout? Origin: New D Lower Order? D Past NRC Exam? 4.1.5 IF testing 2-SW-90A. ":4" SERVlCE WATER PUMP STRAINER FLUSH." PERFORM the following: NOTE Tv,lo Operators are required to time strainer Hush valves due to IOc'ltion of swi tches and quick operation of vaIves. level of Li;;l Continuo~

a.

Refer lh OP 2328A "Sodiwll Hypochlorite System." and VERIFY Sodium Hypochlorite Injection to 'W' Service Water I @ Pump is terminated.

h. LOG ENTRY into TSA."i 3.7.4.1 and TRMA."i 7. 1.2 LA.

IG)

c. IE at any time, valve does not stroke fully, Go To.

d. PLACE "1\\' Service Water Pump Strainer control switch in "HAND" (C-SSA).

10

c.

PRESS '~" Service Water Pump Strainer "START" button and MEASURE opell 'itroke time.

f.

RECORD 2-SW-90A open stroke time on SP 2612A-OO).

g.

PRESS ':4"' Service Water Pump Strainer "STOP" button and MEASURE close stroke time.

h. RECORD 2-SW-t)OAciose stroke time on SP 2612A-003.

1.

PLACE 'Pi. Service Water Pump Strainer control switch in ':AUTO" (C-58A). 1r4l1r;-.,3 1>..;/ \\.:V J. DOCUMENT2-SW-90A"Normal" limits Results on SP 2612A-003. STOP THINK REVIEW SP 2612A Rev. OW-OS 1-\\of36

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

I. IE Operational Readiness Results "llNSAT," Go To AUachment L

m. IF "Normal" limits Results "UNSAT," Go lh Attachment 2.

n. LOG EXIT from TSA.'i 3.7.4.1 and TRMAS 7.121.A.

I G) Page 45 of75 Printed on 1/28/2010 at 12:02 Question #: II-SRO Ques. # 15 15 Question ID; 9000010 RO ~SRO Student Handout? Rev. 0 ~ Selected for Exam Origin: New 4.1.5 IF testing 2-SW-90A. ":4" SERVlCE WATER PUMP STRAINER FLUSH." PERFORM the following: NOTE D Lower Order? D Past NRC Exam? Tv,lo Operators are required to time strainer Hush valves due to IOc'ltion of swi tches and quick opera tion of va Ive s.

a.

Refer lh OP 2328A "Sodiwll Hypochlorite System." and VERIFY Sodium Hypochlorite Injection to 'W' Service Water I @ Pump is terminated.

h.

LOG ENTRY into TSA."i 3.7.4.1 and TRMA."i 7. 1.2 LA. IG)

c. IE at any time, valve does not stroke fully, Go To.

d.

PLACE "1\\' Service Water Pump Strainer control switch in "HAND" (C-SSA). 10

c.

PRESS '~" Service Water Pump Strainer "START" button and MEASURE opell 'itroke time.

f.

RECORD 2-SW-90A open stroke time on SP 2612A-OO).

g.

PRESS ':4"' Service Water Pump Strainer "STOP" button and MEASURE close stroke time.

h.

RECORD 2-SW-t)OAciose stroke time on SP 2612A-003.

1.

PLACE 'Pi. Service Water Pump Strainer control switch in 1r4l1r;-.,3 ':AUTO" (C-58A). 1>..;/ \\.:V J. DOCUMENT2-SW-90A"Normal" limits Results on SP 2612A-003. level of Li;;l Continuo~ STOP THINK SP 2612A REVIEW Rev. OW-OS 1-\\of36

k.

DOCUMENT 2-SW-9UA Operational Readiness Results on SP 2612A-003. I. IE Operational Readiness Results "llNSAT," Go To AUachment L

m. IF "Normal" limits Results "UNSAT," Go lh Attachment 2.

n.

LOG EXIT from TSA.'i 3.7.4.1 and TRMAS 7.121.A. I G) Page 45 of75 Printed on 1/28/2010 at 12:02

i SRC) " Question #: 15 C Student Handout? Lower Order? Question ID; 9000010 RO ~ SRO II-SRO Ques. It 15 Rev. 0 i;'I/ Selected for Exam Origin: New Past NRC EKam? Actions for 1ST Data Outside hNormal" Limits (Sheet t of I) 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 I~ retest must he performed. L IF a second failure of "Normal",;troke time limit has (}(,"Curred, Go To Attachment L

2. VERIFY th.:! following meet test requirements:

Test prerequisites System conditions Procedure performance

3. REVrEW n:cortlcd data and DETERMINE if test equipment is providing accurate information.

4. Ti) retest componenL PERFORl\\1 the following:

4.1 OBTAIN new applicable Form data sheets (new cover sheet not required). 4.2 ENTER the following on applicable OPS Form cover sheet "Comments" section: "Retest of (specijj compolltmt) I'equired, additional dala sheets attached." 4.3 INDICATE on new OPS F\\.)rm data sheets thal data is from retest and ATIACH to original Form. 4.4 Go To applicable section of this procedure and PERFORM retest. Level OfU~ SP 2612A Continuo'~ STOP THINK ACT REVIEW Rev. 01O-0~ 350[36 Page 46 of 75 Printed on 1/28/2010 at 12:02 SRC) " gly <No :nPaifltltS!!!'D~~;)~~D Question #: 15 Question ID; 9000010 RO ~ SRO C Student Handout? II-SRO Ques. It 15 Rev. 0 i;'I/ Selected for Exam Origin: New i Actions for 1ST Data Outside '"Normal" Limits (Sheet t of I) 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 retest must he performed. L IF a second failure of "Normal",;troke time limit has (}(,"Curred, Go To Attachment L

2. VERIFY th.:! following meet test requirements:

Test prerequisites System conditions Procedure performance

3. REVrEW m:ortlcd data and DETERMINE if test equipment is providing accurate information.

4. Ti) retest componenL PERFORl\\1 the following:

4.1 OBTAIN new applicable Form data sheets (new cover sheet not required). 4.2 ENTER the following on applicable OPS Form cover sheet "Comments" section: "Retest of (specijj compolltmt) required, additional data sheets attached." 4.3 INDICATE on new OPS F\\.)rm data sheets thal data is from retest and ATIACH to original Form. 4.4 Go To applicable section of this procedure and PERFORM retest. Level OfU~ Continuo'~ STOP THINK ACT SP 2612A REVIEW Rev. 01O-0~ 350[36 Lower Order? Past NRC EKam? Page 46 of 75 Printed on 1/28/2010 at 12:02

I Question ID; 9000010 RO ~ SRO Student Handout? .:J Lower Order?

Question #:

15 II-SROQues.# 15 Rev. 0 Selected for Exam Origin; New Past NRC Exam? Actions for 1ST Data Outside "Acceptable" Limits (Sheet 1of 1)

1. CONSIDER component not OPERABLE and NOTIFY SM or US.

o

2..1F in MODE 1,2,3 or 4. LOG into TS 3.7.4.1, and TRMAS 7.1.21 A as reqllired.

1G)

3. SUBMIT CR and RECORD CR number in applicable Form cover sheet.

4. i\\lOTIFY Ih e following:

I~'T Coordi nalor System Engineer Level of Use SP 2612A STOP THINK ACT REVIEW Rev. 01O-m~ 34 of 36 Continuous Page 47 of75 Printed on 1/28/2010 at 12:02

Question #:

15 I Question ID; 9000010 RO ~ SRO Selected for Exam Student Handout? .:J Lower Order? II-SROQues.# 15 Rev. 0 Origin; New Actions for 1ST Data Outside "Acceptable" Limits (Sheet 1 of 1)

1. CONSIDER component not OPERABLE and NOTIFY SM or US.

2..1F in MODE 1,2,3 or 4. LOG into TS 3.7.4.1, and TRMAS 7.1.21 A as reqllired.

3. SUBMIT CR and RECORD CR number in applicable Form cover sheet.

4. i\\lOTIFY Ih e following:

I~'T Coordi nalor System Engineer Level of Use Continuous STOP THINK ACT SP 2612A REVIEW Rev. 01O-m~ 34 of 36 Past NRC Exam? 1o G) Page 47 of75 Printed on 1/28/2010 at 12:02

Question #: 16 D Student Handout? Lower Order? Question ID: 9000012 RO ~ SRO l-SRO Que~. if 16 Rev. 0 ~ Selected for Exam Origin: New Past NRC Exam? 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 the following actions must be taken and why? D 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. D B Immediately suspend CORE ALTERATIONS and restore the Radiation Monitor blower prior to resuming fuel movement, to ensure a potential fuel handling accident in Containment is NOT released to the environment. ." 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. D Restore the Containment Purge Valves to OPERABLE status within the next 30 minutes or immediately close the Pur~le Valves, to ensure Containment Closure is reestablished in case of a fuel handling accident in Containment. Justification I C IS CORRECT; TS 3.9.4 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 fUl~1 handlin!;} aCCident in Containment. A is incorrE~ct; GORE ALTERATIONS do NOT need to be suspended and Containment Closure is still available. Plausible if the examineE! 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 ffilJSt 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 repaired. Plausible if the examineE! 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 signal. 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. R.~ferences I Tech. Spec. 3.BA LCO; Containment Penetrations Comments and Question Modification History I Bob K. - D-3IC (Change "Designate" to "Ensure" for control room operator). Minor rewording of chc,ices "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-3IC; No comrnents. Bruce F. 3IC, No comment NRC KJA System/E/A I Gener~ KIA. SelecteciJ System 029 Containment Purge System (CPS) NRC KIA Generic System 2.1 Conduct of Operations Page 48 of 75 Printed on 1/28/2010 at 12:02 Question #: 16 l-SRO Que~. if 16 Question ID: 9000012 RO ~ SRO Rev. 0 ~ Selected for Exam 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. D Student Handout? Origin: New - Containment Atmosphere Radiation Monitor, RM-8123, is out of service for repairs. Lower Order? Past NRC Exam? 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? D 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. D B Immediately suspend CORE ALTERATIONS and restore the Radiation Monitor blower prior to resuming fuel movement, to ensure a potential fuel handling accident in Containment is NOT released to the environment. ." 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. D Restore the Containment Purge Valves to OPERABLE status within the next 30 minutes or immediately close the Pur~le Valves, to ensure Containment Closure is reestablished in case of a fuel handling accident in Containment. Justification I C IS CORRECT; TS 3.9.4 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 fUl~1 handlin!;} aCCident in Containment. A is incorrE~ct; GORE ALTERATIONS do NOT need to be suspended and Containment Closure is still available. Plausible if the examineE! 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 ffilJSt 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 repaired. Plausible if the examineE! 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 signal. 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. R.~ferences I Tech. Spec. 3.BA LCO; Containment Penetrations Comments and Question Modification History I Bob K. - D-3IC (Change "Designate" to "Ensure" for control room operator). Minor rewording of chc,ices "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-3IC; No comrnents. Bruce F. 3IC, No comment NRC KJA System/E/A System 029 Containment Purge System (CPS) I Gener~ KIA. SelecteciJ NRC KIA Generic System 2.1 Conduct of Operations Page 48 of 75 Printed on 1/28/2010 at 12:02

Question #: 16 RO;Jl SRO Student Handout? Lower Order? Question ID: 9000012 J-SRO Que&. ~ 16 Rev. o ~ Selected for Exam Origin: New Past NRC E;l(am? 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 REFUELING OPERATIONS CONTAIN1\\fENT PENETRATIONS LIMITING CONDHION FOR OPERATION 3.9.4 The containment penetrations shall be in the following status:

a.

The eqnipment door shall be either:

1.

dosed and held iu place by a minimum offour boIts, or

2.

open under administrative control* and capable ofbeing closed and held in place by a millinuull of four bolts,

b.

The persoll11el air lock shall be either: L closed by one persoll11el air lock door, or

2.

capable ofbeing closed by au OPERABLE personnel air lock door. under administrative control >1'. and

c.

Each penetration providing direct access from the coutainment atmosphere to the outside atmosphere shall be either:

1.

Closed by a manual or automatic isolation valve, blind flange, or equivalent, or 2, Be capable ofbeing closed lUlder administrative control >I< APPLICABILITY: DlU'illg movement of inadiated fuel assemblies within containment With the requirements ofthe above specification not satisfied, immediately suspend all operations iuvolving movement ofimtdiated fuel assemblies in the cOlltaimllellt. Administrative controls shall ensure that appropriate pel'somlel are aware that the equipment door, persollnel air lock door and/or other containment penetrations are open, and that a spedfic individual(s) is designated and available to close the equipment door. persolUlel air lock door and/or other contailllneut penetrations within 30 minutes if a fuel handling accident occurs. Any obstructions (e.g" cables 3ndlloses) that could prevent closure of the,equipment door, a personnel air lock door aud/or other containment penetration must be capable ofbeing quickly removed. - UNIT 2 3/4 9-4 Amendment No. 66, M. 98-, ~, 284 Page 49 of 75 Printed on 1/28/2010 at 12:02 Question #: 16 Question ID: 9000012 RO;Jl SRO Student Handout? J-SRO Que&. ~ 16 Rev. o ~ Selected for Exam Origin: New 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 REFUELING OPERATIONS CONTAIN1\\fENT PENETRATIONS LIMITING CONDHION FOR OPERATION 3.9.4 The containment penetrations shall be in the following status:

a.

The eqnipment door shall be either:

1.

dosed and held iu place by a minimum of four boIts, or

2.

open under administrative control* and capable of being closed and held in place by a millinuull of four bolts,

b.

The persoll11el air lock shall be either: L closed by one persoll11el air lock door, or

2.

capable of being closed by au OPERABLE personnel air lock door. under administrative control >1'. and

c.

Each penetration providing direct access from the coutainment atmosphere to the outside atmosphere shall be either:

1.

Closed by a manual or automatic isolation valve, blind flange, or equivalent, or 2, Be capable of being closed lUlder administrative control >I< APPLICABILITY: DlU'illg movement of inadiated fuel assemblies within containment With the requirements ofthe above specification not satisfied, immediately suspend all operations iuvolving movement of imtdiated fuel assemblies in the cOlltaimllellt. Administrative controls shall ensure that appropriate pel'somlel are aware that the equipment door, persollnel air lock door and/or other containment penetrations are open, and that a spedfic individual(s) is designated and available to close the equipment door. persolUlel air lock door and/or other contailllneut penetrations within 30 minutes if a fuel handling accident occurs. Any obstructions (e.g" cables 3ndlloses) that could prevent closure of the,equipment door, a personnel air lock door aud/or other containment penetration must be capable of being quickly removed. - UNIT 2 3/4 9-4 Amendment No. 66, M. 98-, ~, 284 Lower Order? Past NRC E;l(am? Page 49 of 75 Printed on 1/28/2010 at 12:02

Question #: 17 Student Handout? Lower Order? Question ID: 9019010 RO .~ SRO I-SRO Ques. 1/ 17 Rev. o ~. Selected for Exam Otfgin: Mod Past NRC Exam? The plant was at 1 00% power when CONVEX ordered Main Generator output be lowered from 900 MWe to 600 MWe in 15 minutes. AOP 2557, "Emer~lency Generation Reduction", was initiated and the following conditions now exist:

Group 7 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 eND", 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; 0-34).
RCS Tcold is 550 of and slowly rising (RPS).

Which one of the following actions should the US direct?

- - - I 11 I. I ***** I I ****** I *************** I"

......... I. I **************** I ************ 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 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. Justification. J B - CORRECT; Turbine load has been lowered ahead of the "A" Steam Dump Valve controller setpolnt, as indicated by TIC-4165 output being higher 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 already above the DNB Tech. Spec. limit, so the LCO must be entered. A - WRONG; Transferrirg 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 setpoint must be lowered to ensure the "A" steam dump stays aheac of the other 3 valves. C - WRONG; AOP 2557 requires RCS temperature to be maintained within 10°F of program or a plant trip is required. AOP 2557 maintains reactor power Gonstant, therefore, Tcold should be -545°F per Attachment 1. RPS indication (and C02/3 alarms) indicate Tcold is >/=549'F, which 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. 0- WRONG; Driving in CEAs will lower RCS temperature by lowering reactor power but, RCS temperature is "out-of-program" because generator load reduction 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. However, 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-31W (Did not read "2557" and thought Rx power was being reduced. Upon second thought. was able to dertermine the' correct answer. I Bill M. - D-31W. 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 "AU steam dump. Revised "B" to lower the setpoint on the "AU steam dump vs pick up load on the Turbine. - RJA Mike C. IW, Lower contl'Olier outputs in stem to put valves at 75% open. Done - RLC Angelo D-4/C; No comments. Bruce F.* D*2/C. No comment NRC KIA System/E/A System 045 Main Turbine Generator (MT/G) System 1(;;~~eri~KI AS~I~ct;;(jJ Page 50 of 75 Printed on 1/28/2010 at 12:02 Question #: 17 I-SRO Ques. 1/ 17 Question ID: 9019010 Rev. o RO .~ SRO ~. Selected for Exam Student Handout? Lower Order? Otfgin: Mod Past NRC Exam? The plant was at 1 00% power when CONVEX ordered Main Generator output be lowered from 900 MWe to 600 MWe in 15 minutes. AOP 2557, "Emer~lency Generation Reduction", was initiated and the following conditions now exist:

Group 7 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 eND", 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; 0-34).
RCS Tcold is 550 of and slowly rising (RPS).

Which one of the following actions should the US direct?

- - - I 11 I. I ***** I I ****** I *************** I"

......... I. I **************** I ************ 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 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. Justification. J B - CORRECT; Turbine load has been lowered ahead of the "A" Steam Dump Valve controller setpolnt, as indicated by TIC-4165 output being higher 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 already above the DNB Tech. Spec. limit, so the LCO must be entered. A - WRONG; Transferrirg 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 setpoint must be lowered to ensure the "A" steam dump stays aheac of the other 3 valves. C - WRONG; AOP 2557 requires RCS temperature to be maintained within 10°F of program or a plant trip is required. AOP 2557 maintains reactor power Gonstant, therefore, Tcold should be -545°F per Attachment 1. RPS indication (and C02/3 alarms) indicate Tcold is >/=549'F, which 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. 0- WRONG; Driving in CEAs will lower RCS temperature by lowering reactor power but, RCS temperature is "out-of-program" because generator load reduction 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. However, 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-31W (Did not read "2557" and thought Rx power was being reduced. Upon second thought. was able to dertermine the' correct answer. I Bill M. - D-31W. 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 "AU steam dump. Revised "B" to lower the setpoint on the "AU steam dump vs pick up load on the Turbine. - RJA Mike C. IW, Lower contl'Olier outputs in stem to put valves at 75% open. Done - RLC Angelo D-4/C; No comments. Bruce F.* D*2/C. No comment NRC KIA System/E/A System 045 Main Turbine Generator (MT/G) System 1(;;~~eri~KI AS~I~ct;;(jJ Page 50 of 75 Printed on 1/28/2010 at 12:02

SRO ExagtQuestionsOnly (No "Parents" Or "Originals") Question #: 17 Question ID: 9079010 []RO ~ SRO D Student Handout? D Lower Order? I-SRO Ques. # 17 Rev. 0 ~ Selected for Exam Origin: Mod D Past NRC Exam? System 2.4 Emergency Procedures IPlan 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. NRC KIA Generic Millstone Unit 2 Emergency Generation Reduction AOP 2557 Revision 006-06 Page 8 of 16 INSTRUL"fIONS CONTINGENCY ACrrONS CAUTION

1.

2.

PPC calorimetric is inaccurate due to SO level transients. The most accurate available indication of reactor power is d T power. Turbine exhaust hood temperature greater than or equal to 225°F requires manual turbine trip. ICD 3.9 WHEN transferring steam load from main turbine to steam dump and bypass valves. MONITOR the following: ,~T power Condellser backprcssure (ppe point P5127) Turbine exhaust hood temperature (lJR4500 "TURBINE TEMP & I~XPANSION," recorder poinL'> 8 and 9, ppe points T4319 and T4320) Condellsate header !low and pressure MVARs 3.10 WHEN reducing turbine load. MAINTAIN "1\\" steam dump bypass valve 20 to 1O{)9() open as follows: Using "STM DlJMPTAVO CNTL. TIC-4165". THROTTLE open "8," "e," and "D" steam dump bypass valves Level of Use I STOP THINK ACT REVIEW Continuous ~======::::::::'.__.~___________.___________.--J ______._________________________._______J Page 51 of 75 Printed on 1/28/2010 at 12:02 SRO ExagtQuestionsOnly (No "Parents" Or "Originals") Question #: 17 Question ID: 9079010 []RO ~ SRO D Student Handout? D Lower Order? I-SRO Ques. # 17 Rev. 0 ~ Selected for Exam Origin: Mod D Past NRC Exam? NRC KIA Generic System 2.4 Emergency Procedures IPlan 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 INSTRUL"fIONS CONTINGENCY ACrrONS CAUTION

1.

PPC calorimetric is inaccurate due to SO level transients. The most accurate available indication of reactor power is d T power.

2.

Turbine exhaust hood temperature greater than or equal to 225°F requires manual turbine trip. I CD 3.9 WHEN transferring steam load from main turbine to steam dump and bypass valves. MONITOR the following: ,~T power Condellser backprcssure (ppe point P5127) Turbine exhaust hood temperature (lJR4500 "TURBINE TEMP & I~XPANSION," recorder poinL'> 8 and 9, ppe points T4319 and T4320) Condellsate header !low and pressure MVARs 3.10 WHEN reducing turbine load. MAINTAIN "1\\" steam dump bypass valve 20 to 1O{)9() open as follows: Using "STM DlJMPTAVO CNTL. TIC-4165". THROTTLE open "8," "e," and "D" steam dump bypass valves STOP THINK ACT REVIEW Level of Use I Continuous ~======::::::::'. __ .~ ___________. ___________ .--J J Page 51 of 75 Printed on 1/28/2010 at 12:02

Question #: 17 Student Handout? Lower Order? I*SRO Ques. # 17 Rev. 0 ~ Selected for Exam Origin: Mod Past NRC Exam? Question ID: 9079010 D RO ~ SRO Millstone Unit 2 AOP 2557 Revision 006-06 Emergl!ncy Page 9 of 16 Reduction INSTRUCTIONS CONTINGENCY ACTIONS J Using one of the folluwing, ADJUST turbine to desired load: "LOAD LIMIT POT" '"LOAD SELECTOR, INCREASE" and "LOAD SELECTOR, DECREASE" buttons NOTE Receipt of annunciators DA-37, "I II COND D(T" AND DB-37, "HI COND DIS TEMP" is expected during this evolution (C-06/07). 3.12 IF annunciator DA-37 (C-06{1}7). "HI COND D/T' or DB-37 (C-06/07), "HI COND DIS TEMP,~' is received, Refer lb ARP 2590E, "Alarm Response for Control Room Panels, C..06/7." 3.13 WHEN desired load is achieved, STABILIZE tuit)inc load and RCS temperature. _3.14 ENSURE pressurizer level 35 to 709(. 3.14.1 IF the pressurizer level control system is not operating properly in automatically, RESTORE and MAlNTAJN pressurizer level 35 to 70% by pcrforming ANY of tbe following:

a.

OPERATE the pressurizer lcvel control sysrem.

b. Manually OPERATE charging and letdown.

Level of UBe I STOP THINK ACT REVIEW Continuous Page 52 of 75 Printed on 1/28/2010 at 12:02 Question #: 17 Question ID: 9079010 D RO ~ SRO Student Handout? I*SRO Ques. # 17 Rev. 0 ~ Selected for Exam Origin: Mod Millstone Unit 2 AOP 2557 Revision 006-06 Emergl!ncy Reduction INSTRUCTIONS J Using one of the folluwing, ADJUST turbine to desired load: "LOAD LIMIT POT" '"LOAD SELECTOR, INCREASE" and "LOAD SELECTOR, DECREASE" buttons Page 9 of 16 CONTINGENCY ACTIONS NOTE Receipt of annunciators DA-37, "I II COND D(T" AND DB-37, "HI COND DIS TEMP" is expected during this evolution (C-06/07). 3.12 IF annunciator DA-37 (C-06{1}7). "HI COND D/T' or DB-37 (C-06/07), "HI COND DIS TEMP,~' is received, Refer lb ARP 2590E, "Alarm Response for Control Room Panels, C.. 06/7." 3.13 WHEN desired load is achieved, STABILIZE tuit)inc load and RCS temperature. Lower Order? Past NRC Exam? _3.14 ENSURE pressurizer level 35 to 709(. 3.14.1 IF the pressurizer level control system is not operating properly in automatically, Level of UBe I Continuous STOP THINK ACT RESTORE and MAlNTAJN pressurizer level 35 to 70% by pcrforming ANY of tbe following:

a.

OPERATE the pressurizer lcvel control sysrem.

b.

Manually OPERATE charging and letdown. REVIEW Page 52 of 75 Printed on 1/28/2010 at 12:02

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? 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. " AI! other smokl3 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 ':he room has cleared. D The Fire Suppression system is warning that a discharge will occur after a timer countdown. Per TRM 3.3.3.i', "Fire Detection Instrumentation", the Zone 45 fire detection system is inoperable, establish a fire watch, when the room has cleared. Justification I 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 to pulse ClUICKL Y. 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 s~'stem 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 Prolection System", Pg 4, Discussion section.

2. ARP 25901. "Alarm Response for Fire Panel, C-26" (Zone 45). Pg 67-69 Comments 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. Bruce F. - D-4/C, 50/50 (A and B) NRC KIA System/E:/A System 086 Fire Protection System (FPS) Number A2.03 RO 2.7 SRO 2.9 CFR Link (CFR: 41.5 143.5/45.3/45.13) 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 th!~ FPS dU!~ to circuit failure or welding Page 53 of75 Printed on 1/2812010 at 12:02 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? 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. " AI! other smokl3 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 ':he room has cleared. D The Fire Suppression system is warning that a discharge will occur after a timer countdown. Per TRM 3.3.3.i', "Fire Detection Instrumentation", the Zone 45 fire detection system is inoperable, establish a fire watch, when the room has cleared. Justification I 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 to pulse ClUICKL Y. 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 s~'stem 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 Prolection System", Pg 4, Discussion section.

2. ARP 25901. "Alarm Response for Fire Panel, C-26" (Zone 45). Pg 67-69 Comments 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. Bruce F. - D-4/C, 50/50 (A and B) NRC KIA System/E:/A System 086 Fire Protection System (FPS) Number A2.03 RO 2.7 SRO 2.9 CFR Link (CFR: 41.5 143.5/45.3/45.13) 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 th!~ FPS dU!~ to circuit failure or welding Page 53 of 75 Printed on 1/2812010 at 12:02

"9'.'U.VJ..I,~ Only(N6"Pa~f~t6A~ijjr"Originalsn) Question #: 18 Question ID; 9000011 [J RO ~ SRO Student Handout? Lower Order? l-SRO Ques. # 18 Rev. o ~ Selected for Exam Origin; New o Past NRC Exam? Setpolnt: i ZONE 45 I Trouble - Detector string failure Fire WEST DC SWGR Detector string alarm ROOM FLP-6 Halon release from pressure switch ~MAnc FUNCTIONS L Activation of one smoke detector (photoelectric or ion) causes the following to occur: Al armet! t!clCctors red light illuminates. On FLP-6, alarmed zone annunciates. Locally strobe and horns pulses slowly. On FLP-6, detcctors location is shown on graphic annunciator. AJurm signal is sent to C-26 Zone 45~ Alarm lamp for graphk annunciator for cast 120 volt ~witchgear room illuminates.

2.

Activation of another smoke dctcdor of opposite type (ion or photoelectric), causes the following to occur: Alarmet! detectors red light illuminates. On FLP-6, alarmed zone annunciates. Locally strobe and horns pulses quickly. The following closes: 2-IIV-I3RVB, "Supply to 'B' Battery Room From DC Rooms" 2-HV-60IB, "Cable Vault to 'B' DC Room SWGR. Room Fire Damper" 60 second pre-discharge time delay begins. Upon expiration of time delay Halon System discharges and strobe and horn sounds s.teadily. Flashing lights operates. Upon expiration of time delay ha.lon system discharges into west 120 yolt switchgear room. ARP25901 LevelofUE~ STOP THINK ACT REVIEW Rev. 002-08 ReferencE:..j 67 of 104 Page 54 of 75 Printed on 1/28/2010 at 12:02 "9'.'U.VJ..I,~ Only(N6"Pa~f~t6A~ijjr" Originals n) Question #: 18 l-SRO Ques. # 18 Question ID; 9000011 [J RO ~ SRO Student Handout? Lower Order? Rev. o ~ Selected for Exam Origin; New o Past NRC Exam? i ZONE 45 I Setpolnt: Trouble - Detector string failure Fire - Detector string alarm Halon release from pressure switch ~MAnc FUNCTIONS WEST DC SWGR ROOM FLP-6 L Activation of one smoke detector (photoelectric or ion) causes the following to occur: Al armet! t!clCctors red light illuminates. On FLP-6, alarmed zone annunciates. Locally strobe and horns pulses slowly. On FLP-6, detcctors location is shown on graphic annunciator. AJurm signal is sent to C-26 Zone 45~ Alarm lamp for graphk annunciator for cast 120 volt ~witchgear room illuminates.

2.

Activation of another smoke dctcdor of opposite type (ion or photoelectric), causes the following to occur: Alarmet! detectors red light illuminates. On FLP-6, alarmed zone annunciates. Locally strobe and horns pulses quickly. The following closes: 2-IIV -I3RVB, "Supply to 'B' Battery Room From DC Rooms" 2-HV-60IB, "Cable Vault to 'B' DC Room SWGR. Room Fire Damper" 60 second pre-discharge time delay begins. Upon expiration of time delay Halon System discharges and strobe and horn sounds s.teadily. Flashing lights operates. Upon expiration of time delay ha.lon system discharges into west 120 yolt switchgear room. LevelofUE~ ReferencE:..j STOP THINK ACT Page 54 of 75 REVIEW ARP25901 Rev. 002-08 67 of 104 Printed on 1/28/2010 at 12:02

Question #: 18 Question 10; 9000011 RO ~ SRO Student Handout? D Lower Order? J-SRO Ques, # 18 Rev. 0 !!L Selected for Exam Odgin: New D Past NRC Exam? [ ZONE 45 8 8 WARNING When Halon Systems are actuated, the affected area i" neither oxygen deficient nor toxic; however, extended exposure to Halon may have harmful effects. ~--------------------------------------------------------------~ NOTE Activation ofthe manual pull station CHuses a halon release aftera five sccond time dc:lav. An abort S\\vitch can bc used to prevent a halon reka'lc until the affected panel can be reset Ifthe abort s\\\\1tch is turned back before the panel is reset.. the Halon discharges after it len second time delay. The reset is inside the respective FLP and a valvc lock kcy is rCtluired to get into the panel. The manua'! pull station activation overrides the abort. CORRECTIve ACTIONS

1.

Refer To Attachment 6. "FLP-6 and 6A Zone 45" and DETERMINE calise of alarm.

2.

IE fire ah:xm is valid, PERFORM the following: I@ 2.1 DETERMINE locatioll of fire. If alarm was t,'HUSC by actuation of Halon fire Suppression System ~ fire is verified, PERfORM the following: EVACUATE affected area. Refer lb AOP 2559, "Firc" and PERfORM applicable actions. 2.3 IF alarm is due to Halon discharge AND flO fire is presenL PERFORM the following: CAUTION When ventilating care must be taken not to discharge products of combustion into non-affected rooms. 2.:1..1 PERFORM actions to ventilate affected area. 2.:',.2 Refer To OP 2341A, "Fire Protection System;' and REMOVE appropriate Halon System from service. I ARP 25901 Level of Use STOP THINK ACT REVIEW Rev. 002-08 ReferenC43 68 of 104 Page 55 of 75 Printed on 1/28/2010 at 12:02 Question #: 18 Question 10; 9000011 RO ~ SRO Student Handout? D Lower Order? J-SRO Ques, # 18 Rev. 0 !!L Selected for Exam Odgin: New D Past NRC Exam? [ ZONE 45 8 WARNING 8 When Halon Systems are actuated, the affected area i" neither oxygen deficient nor toxic; however, extended exposure to Halon may have harmful effects. ~--------------------------------------------------------------~ NOTE Activation of the manual pull station CHuses a halon release after a five sccond time dc:lav. An abort S\\vitch can bc used to prevent a halon reka'lc until the affected panel can be reset If the abort s\\\\1tch is turned back before the panel is reset.. the Halon discharges after it len second time delay. The reset is inside the respective FLP and a valvc lock kcy is rCtluired to get into the panel. The manua'! pull station activation overrides the abort. CORRECTIve ACTIONS

1.

Refer To Attachment 6. "FLP-6 and 6A Zone 45" and DETERMINE calise of alarm.

2.

IE fire ah:xm is valid, PERFORM the following: 2.1 DETERMINE locatioll of fire. If alarm was t,'HUSC by actuation of Halon fire Suppression System ~ fire is verified, PERfORM the following: EVACUATE affected area. Refer lb AOP 2559, "Firc" and PERfORM applicable actions. 2.3 IF alarm is due to Halon discharge AND flO fire is presenL PERFORM the following: CAUTION When ventilating care must be taken not to discharge products of combustion into non-affected rooms. 2.:1..1 PERFORM actions to ventilate affected area. 2.:',.2 Refer To OP 2341A, "Fire Protection System;' and REMOVE appropriate Halon System from service. Level of Use I ReferenC43 STOP THINK ACT REVIEW ARP 25901 Rev. 002-08 68 of 104 I@ Page 55 of 75 Printed on 1/28/2010 at 12:02

Question #: 18 Question ID; 9000011 0 RO ~ SRO Student Handout? ',SRO Ques. # 18 Rev. o ~ Selected for Exam Origin: New Lower Order? ZONE 45 2.3.3 POST fire watch as necessary. 23.4 NOTIFY Fire Marshall. 2.3.5 SUBMIT Trouble Report to Maintenance Department.

3.

IF alarm is due to electrical malfunction, SUBMIT Trouble Report to Electrical Maintenance Department.

4.

For continued operation. CONSIDER supplemental room cooling.

5.

As applicable. Refer ']() Technical Requirements Manual, and DETERMINE system operahility. SUPPORTING INFORMATION

1.

Initiating Devin.'s FF'L-6 Detector string. FSD-49 3 ion detectors (smoke) 3 photoelectric detectors (smoke) PS-7696 IIS-7696 A & B (Manual Electric Release) Compute:r Points FLP-fi TE8436

3.

ll~chnical Requirements Manual, Section 11 subsection 1.0. [able A.3.1..4 E.3.1

4.

Procedures OJ> 2341 A. "Fire Protection System" AOP 2559. "Fire" AOP 2579F, "Fire Procedure for Hot Standby Appendix "R" Fire Area R -10" AOP 2579fF, "Fire Procedure for CooJdown and Cold Shuldowlt Appendix "R" fire Area R -10 and R-8" I ARP 25901 Level of Use STOP THINK ACT REVlEVJ Rev. 002-08 ReferenCE, 69 of 104 Page 56 of75 Printed on 1/28/2010,at 12:02 Question #: 18 ',SRO Ques. # 18 Question ID; 9000011 0 RO ~ SRO Rev. o ~ Selected for Exam Student Handout? Lower Order? Origin: New ZONE 45 2.3.3 POST fire watch as necessary. 23.4 NOTIFY Fire Marshall. 2.3.5 SUBMIT Trouble Report to Maintenance Department.

3.

IF alarm is due to electrical malfunction, SUBMIT Trouble Report to Electrical Maintenance Department.

4.

For continued operation. CONSIDER supplemental room cooling.

5.

As applicable. Refer ']() Technical Requirements Manual, and DETERMINE system operahility. SUPPORTING INFORMATION

1.

Initiating Devin.'s FF'L-6 Detector string. FSD-49 3 ion detectors (smoke) 3 photoelectric detectors (smoke) PS-7696 IIS-7696 A & B (Manual Electric Release) Compute:r Points FLP-fi TE8436

3.

ll~chnical Requirements Manual, Section 11 subsection 1.0. [able A.3.1.. 4 E.3.1

4.

Procedures OJ> 2341 A. "Fire Protection System" AOP 2559. "Fire" AOP 2579F, "Fire Procedure for Hot Standby Appendix "R" Fire Area R -10" AOP 2579fF, "Fire Procedure for CooJdown and Cold Shuldowlt Appendix "R" fire Area R -10 and R -8" Level of Use I ReferenCE, STOP THINK ACT Page 56 of75 REVlEVJ ARP 25901 Rev. 002-08 69 of 104 Printed on 1/28/2010,at 12:02

Question #: 18 RO ~SRO Student Handout? Lower Order? Question 10; 9000011 J-SRO Ques. # 18 Rev. 0 ~ Selected for Exam Otfgin: New Past NRC Exam? TECHNICAL REQUIREMENTS ~l4.3 INSTRUMENTATION 3l4*3,3 M9NII2B'tUi INiTBUMENTATION aL4.a.a,ZEIBE DEIECDQtj INSIBUMENIAIIQN LIMITING CCINDITION FOR OPERATION 3,33.7 As a minimum, the fire detection instrumentation for each fire detection zone in TRM Table 3.3-10 shall be OPERABLE. APPLICABIL.Irl;, Whenever equipment in that fire detection zone is required to be OPERABLE. ACIION: With the number of OPERABLE fire detection instrument(s) less than the minimum number of OF)ERABLE requirements of TRM Table 3.3*10:

a. Within 1 hour, 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 REQUIREMENT JABbE33-1Q fiRE PETECTION INSmUMSNTs Minimum Minimum Tolal No.

Channels Total No, Channels Q!.C~ ~ of Channels ~

4.

4.16 & 6.9 kV Switchgear Room (56'6") (40) 4 3 4,16 & 6.9 kV SWltcllgear Room (31'6") (18) 4 3 480 V We!t Switchgear Room {3S'S"} (18) 2 1 480 V East Switoh91!sr Room (36'6") (28) 2 1 East DC Equipment Room (43 Alarm) (FLP 5) 6 6 West DC Equipment Room (45 Alarm) (FLP 6) 6 6 Page 57 of 75 Printed on 1/28/2010 at 12:02 Question #: 18 J-SRO Ques. # 18 Question 10; 9000011 Rev. 0 RO ~SRO ~ Selected for Exam TECHNICAL REQUIREMENTS ~l4.3 INSTRUMENTATION 3l4*3,3 M9NII2B'tUi INiTBUMENTATION aL4.a.a,Z EIBE DEIECDQtj INSIBUMENIAIIQN LIMITING CCINDITION FOR OPERATION Student Handout? Lower Order? Otfgin: New Past NRC Exam? 3,33.7 As a minimum, the fire detection instrumentation for each fire detection zone in TRM Table 3.3-10 shall be OPERABLE. APPLICABIL.Irl;, Whenever equipment in that fire detection zone is required to be OPERABLE. ACIION: With the number of OPERABLE fire detection instrument(s) less than the minimum number of OF)ERABLE requirements of TRM Table 3.3*10:

a. Within 1 hour, 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 Switchgear Room (56'6") (40) 4,16 & 6.9 kV SWltcllgear Room (31'6") (18) 480 V We!t Switchgear Room {3S'S"} (18) 480 V East Switoh91!sr Room (36'6") (28) East DC Equipment Room (43 Alarm) (FLP 5) West DC Equipment Room (45 Alarm) (FLP 6) TECHNICAL REQUIREMENT JABbE33-1Q fiRE PETECTION INSmUMSNTs Minimum Tolal No. Channels Total No, Minimum Channels Q!.C~ ~ of Channels ~ Page 57 of 75 4 4 2 2 6 6 3 3 1 1 6 6 Printed on 1/28/2010 at 12:02

Question #: 19 RO ~SRO ~ Student Handout? Lower Order? Question ID: 9000013 I-SRO Ques. # 19 Rev. 1 ';,jl Selected for Exam Origin: New o 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 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 Down power. 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%, then transition to OP 2208, Attachment 5, Reactivity Thumbrules, to continue the power reduction to 55% to secure the Feed Pump. o 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 to 55% to secure the Feed Pump. ~ C Use Reactivity Plan RE-G-11 to reduce power to 55%, secure the Feed Pump, then raise power to the appropriate level using OP 2204, Load Changes, and a new Reactivity Plan. o Use Reactivity Plan RE-G-10 to reduce power to 55%, secure the Feed Pump, then transition to OP 2393, Core Power Distribution and Monitoring, to maintain ASI. justjfil::~ii~n I C IS CORRECT: Reactivity Plan RE-G-11 provides the Boration rate and CEA insertion required to down power to 55% within the one hour time constraint. OF> 2322, Main Feedwater, requires power to be at or less than 55% to remove the first Main Feed Pump from service. This reactivity 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 providl~s direction for use of a reactivity plan appropriate for the evolution being performed. Plausible because 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 thl~ pump is secured. D 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 remov*sd 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 mactivity plan will allow performance of the down power 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 include the effects of Xenon; therefore, the power reduction rate would actually be higher than indicated on RE-G-10. Refe~ences'l ProvIded RE-G-03, Rapid Down Power Reactivity Plans (Question Reference #1 of 3) AOP 2575, Rapid Downpower (NOT provided to examinees) Comments and Questil::m Modification History I NRC* (original question comments) Modify all choices to suggest use of the 55% Reactivity plan. RLC - Modified all four ciloices to allow for "55%" to be added to each, per NRC comments. [11/30109] Bruce F. - D-2/C, No comment NRC KIA System/l:/A System 2.1 Conduct of Operations [!l~~----- ~eneric KIA Selecte(! NRC KIA Generic System 2.1 Conduct of Operations Number 2:1.43 RO 4.1 SRO 4.3 CFR Link (CFR: 41.10 I 43.6 I 45.6) Ability to use procedures to determine the effects on reactivity of plant changes, such as reactor coolant system temperature, secondary plant, fuel depletion, etc. Page 58 of 75 Printed on 1/28/2010 at 12:02 Question #: 19 Question ID: 9000013 I-SRO Ques. # 19 Rev. 1 RO ~SRO ';,jl Selected for Exam ~ Student Handout? Origin: New Lower Order? o 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 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 Down power. 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%, then transition to OP 2208, Attachment 5, Reactivity Thumbrules, to continue the power reduction to 55% to secure the Feed Pump. o 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 to 55% to secure the Feed Pump. ~ C Use Reactivity Plan RE-G-11 to reduce power to 55%, secure the Feed Pump, then raise power to the appropriate level using OP 2204, Load Changes, and a new Reactivity Plan. o Use Reactivity Plan RE-G-10 to reduce power to 55%, secure the Feed Pump, then transition to OP 2393, Core Power Distribution and Monitoring, to maintain ASI. justjfil::~ii~n I C IS CORRECT: Reactivity Plan RE-G-11 provides the Boration rate and CEA insertion required to down power to 55% within the one hour time constraint. OF> 2322, Main Feedwater, requires power to be at or less than 55% to remove the first Main Feed Pump from service. This reactivity 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 providl~s direction for use of a reactivity plan appropriate for the evolution being performed. Plausible because 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 thl~ pump is secured. D 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 remov*sd 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 mactivity plan will allow performance of the down power 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 include the effects of Xenon; therefore, the power reduction rate would actually be higher than indicated on RE-G-10. Refe~ences'l ProvIded RE-G-03, Rapid Down Power Reactivity Plans (Question Reference #1 of 3) AOP 2575, Rapid Downpower (NOT provided to examinees) Comments and Questil::m Modification History I NRC* (original question comments) Modify all choices to suggest use of the 55% Reactivity plan. RLC - Modified all four ciloices to allow for "55%" to be added to each, per NRC comments. [11/30109] Bruce F. - D-2/C, No comment NRC KIA System/l:/A System 2.1 Conduct of Operations [!l~~----- ~eneric KIA Selecte(! NRC KIA Generic System 2.1 Conduct of Operations Number 2:1.43 RO 4.1 SRO 4.3 CFR Link (CFR: 41.10 I 43.6 I 45.6) Ability to use procedures to determine the effects on reactivity of plant changes, such as reactor coolant system temperature, secondary plant, fuel depletion, etc. Page 58 of 75 Printed on 1/28/2010 at 12:02

Question #: 19 I Question ID; 9000013 RO ~SRO ~ Student Handout? D LowerOrder? l-SRO Ques, # 19 Rev. 1 ,,: Selected for Exam Origin: New Past NRC Exam? RWST -1 pump RE-G-04 BA vol-880 gals 90% BA flow-44 gpm CEA pas - 180 steps rTIme'- 20 min. 0 Rate - 30%/hr RE-G-07 SA vol-2458 gals 70% BA flow - 44 gpm CEA pas - 180 steps (No CEAs) Time-56 min. e Rate - 32%/hr RE-G-10 I3A vol-3080 gals RWST -2 pump RE-G-05 BA vol-880 gals BA flow - 88 gpm CEA pas - 180 steps Time - 10 min. Rate - 60%/hr RE-G-08 BA vol-2640 gals BA flow - 88 gpm CEA pas - 180 steps Time - 30 min. Rate - 60%/hr RE-G-11 BA vol-3315 gals BAST - 1%/min RE-G-06 BA vol-204 gals BA flow - 20 gpm CEA pas - 180 steps Time - 10 min. Rate - 60%/hr RE-G-09 BA vol-606 gals BA flow - 20 gpm CEA pas - 180 steps Time - 30 min. Rate - 60%/hr RE-G-12 BA vol-720 gals SA flow - 44 gpm BA flow - 88 gpm BA flow - 18 gpm 55% CEA pas - 154 steps CEA pas - 154 steps C EA pas - 154 steps e Time - 70 min. Time - 38 min. Rate - 39%/hr Rate - 71 %/hr RE-G-13 RE-G-14 IBA vol-4400 gals BA vol-5120 gals ISA flow 44gpm BA flow - 88 gpm 15% CEApas 140 steps CEA pas - 140 steps Time-100 min.,Time-58 min. Rate - 51%/hr Rate - 88%/hr I Time - 40 min. Rate - 68%/hr RE-G-15 BA vol-1100 gals BA flow - 16 gpm CEA pas - 140 steps Time - 80 min. Rate - 64 %/hr Page 59 of75 Printed on 1/2812010 at 12:02 Question #: 19 I Question ID; 9000013 RO ~SRO ~ Student Handout? D LowerOrde r? l-SRO Ques, # 19 Rev. 1 ,,: Selected for Exam Origin: New Past NRC Ex am? RWST -1 pump RWST -2 pump BAST - 1%/min RE-G-04 RE-G-05 RE-G-06 90% BA vol-880 gals BA vol-880 gals BA vol-204 gals BA flow-44 gpm BA flow - 88 gpm BA flow - 20 gpm CEA pas - 180 steps CEA pas - 180 steps CEA pas - 180 steps 0 rTIme'- 20 min. Time - 10 min. Time - 10 min. Rate - 30%/hr Rate - 60%/hr Rate - 60%/hr RE-G-07 RE-G-08 RE-G-09 70% SA vol-2458 gals BA vol-2640 gals BA vol-606 gals (No CEAs) BA flow - 44 gpm BA flow - 88 gpm BA flow - 20 gpm CEA pas - 180 steps CEA pas - 180 steps CEA pas - 180 steps e Time-56 min. Time - 30 min. Time - 30 min. Rate - 32%/hr Rate - 60%/hr Rate - 60%/hr RE-G-10 RE-G-11 RE-G-12 I3A vol-3080 gals BA vol-3315 gals BA vol-720 gals 55% SA flow - 44 gpm BA flow - 88 gpm BA flow - 18 gpm CEA pas - 154 steps C EA pas - 154 steps CEA pas - 154 steps e Time - 70 min. Time - 38 min. Time - 40 min. Rate - 39%/hr Rate - 71 %/hr Rate - 68%/hr RE-G-13 RE-G-14 RE-G-15 IBA vol-4400 gals BA vol-5120 gals BA vol-1100 gals 15% ISA flow 44gpm BA flow - 88 gpm BA flow - 16 gpm CEApas 140 steps CEA pas - 140 steps CEA pas - 140 steps Time-100 min.,Time-58 min. Time - 80 min. Rate - 51%/hr I Rate - 88%/hr Rate - 64 %/hr Page 59 of75 Printed on 1/2812010 at 12:02

SRO Exam Questions.Only (No ".Parents" Or "Originals") Question #: 19 Question ID; 9000013 []RO ~ SRO ~ Student Handout? D Lower Order? iJ-SRO Ques. # 19 Rev. 1 ~ Selected for Exam Origin: New D Past NRC Exam? Millstone Unit 2 AOP 2575 Revision 004-01 Rapid Dowllpower Page 4of35 Reactivity plans are provided and should be llsed if the initial conditions specified in the plans (lOO~!c initial power, ARO. desired final power level) arc approximately as specified. These plans are an approximation of the required boration and CEA movement required to reach the desired power while controlling the ASI oscillation. The reactivity plan CEA positioning will maintain ASI v,'ithin COLR limits. OP-2393, "Core Power Monitoring Distribution and Control," provides direction for maintaining ASI control within specified bands during steady-state conditions. transient conditions, or at the direction of Reactor Engineering. The reactivity plans provide the above Reactor Engineering direction for AS) control. A") I control during the down power in ac;)rdance~with the reactivity plan is prefered. however, it shou Id not interfere with event mitigation. Once reactor power is stabilized, ASI should be maintained in acconJance with OP 2393 Of the reactivity plan. The rirst page of the reactivity plan provides the bora! ion 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. TIlis page should not be interpreted as procedural direction and deviation from this guidance is allowable to achieve the desired power level within the desired tim~. The third page provides a prediction of the relative ASI trend during the down power. TIle A'iI trend should not be llsed as an indication of a true absolute ASI 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 affects from ReS temperature changes vary significantly over core life. based upon Res boron concentration and resultant Moderator Temperature Coefficient (MTC) value. At 100% power, beginning of life (80L). after xenon equilibrium. the value of MTC is much less negative than at the end of life (EOL). "nlis means that at BOL a change of 1°F RCS temperature will calise approximately a 1/2 % change in power, whereas a 1°F ReS temperature chang.e at EOL will cause approximately a :!t;'{ change in power. Ll Applicability This procedure is applicable in Mode I at power levels greater than 20l}~, when an emergem..'Y power reduction is required. Level of USE~ ACT STOP THINK REVIEW Con1:jnuou~~ '--------------.------------------------------------------~ Page 60 of 75 Printed on 1/28/2010 at 12:02 SRO Exam Questions.Only (No ".Parents" Or "Originals") Question #: iJ-SRO Ques. # 19 Question ID; 9000013 []RO ~ SRO ~ Student Handout? 19 Rev. 1 ~ Selected for Exam Origin: New Millstone Unit 2 Rapid Dowllpower AOP 2575 Revision 004-01 Page 4of35 Reactivity plans are provided and should be llsed if the initial conditions specified in the plans (lOO~!c initial power, ARO. desired final power level) arc approximately as specified. These plans are an approximation of the required boration and CEA movement required to reach the desired power while controlling the ASI oscillation. The reactivity plan CEA positioning will maintain ASI v,'ithin COLR limits. OP-2393, "Core Power Monitoring Distribution and Control," provides direction for maintaining ASI control within specified bands during steady-state conditions. transient conditions, or at the direction of Reactor Engineering. The reactivity plans provide the above Reactor Engineering direction for AS) control. A") I control during the down power in ac;)rdance~ with the reactivity plan is prefered. however, it shou Id not interfere with event mitigation. Once reactor power is stabilized, ASI should be maintained in acconJance with OP 2393 Of the reactivity plan. The rirst page of the reactivity plan provides the bora! ion 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. TIlis page should not be interpreted as procedural direction and deviation from this guidance is allowable to achieve the desired power level within the desired tim~. The third page provides a prediction of the relative ASI trend during the down power. TIle A'iI trend should not be llsed as an indication of a true absolute ASI 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 affects from ReS temperature changes vary significantly over core life. based upon Res boron concentration and resultant Moderator Temperature Coefficient (MTC) value. At 100% power, beginning of life (80L). after xenon equilibrium. the value of MTC is much less negative than at the end of life (EOL). "nlis means that at BOL a change of 1°F RCS temperature will calise approximately a 1/2 % change in power, whereas a 1°F ReS temperature chang.e at EOL will cause approximately a :!t;'{ change in power. Ll Applicability This procedure is applicable in Mode I at power levels greater than 20l}~, when an emergem.. 'Y power reduction is required. Level of USE~ Con1:jnuou~~ STOP THINK ACT REVIEW D Lower Order? D Past NRC Exam? '--------------.------------------------------------------~ Page 60 of 75 Printed on 1/28/2010 at 12:02

Question #: 19 RO ~SRO ~ Student Handout? Question ID: 9000013 J-SRO Ques. # 19 Rev. 1 ~ Selected for Exam Otigin: New Millstone Unit 2 AOP 2575 Revision 004-01 Rapid Downpower Page 7 of35 Lower Order? INSTRUCTIONS CONTINGENCY ACTIONS CAUTION In the case of a dropped rod motion is nat used to initiate downpc-wer. 3.4 !E not down powering due to 11 dropped rod, INSERT Group 7 CEAs J0 +/-2 steps to initiate downpower. _3.5 Refer 'Ii> ppe or Reactor 3.5.1 IE reactor is !lot at the reactivity plan Engineering Curve and Data Book initial conditions. Refer To illld OBTAIN reactivity plan for the. and DETERMINE inital reactor power condition and desired rate of load reduction for time desired \\Gad reduction. in core life. 3.6 IF desired to borate from the RWST (prcfern:~d method) PERFORM the following;

a.

ENSURE at least one charging pump uperating.

h.

ENSURE CH-!.96, vcr makeup bypass, dused. ('. ENSURE CH-S04. RWST to charging suction. open.

d.

OPEN CH-192, d.l IF CH*192. RWST isolation, RWST isolat ion. can not be opened, Go'Iil step 3.8.

c.

CLOSE CH-511L VCT outlet isolation.

f.

CHECK charging flow f.1 START additional charging pumps at desired rate. as needed and balance charging and letdown. 3.7 Based on required rate of downpower, START additional charging pumps as necessary and balance charging and letdown.. LevelofUs~ STOP THINK ACT REVIEW Continuou~ Page 61 of 75 Printed on 1/28/2010 at 12:02 Question #: 19 Question ID: 9000013 RO ~SRO ~ Student Handout? J-SRO Ques. # 19 Rev. 1 ~ Selected for Exam Otigin: New Millstone Unit 2 Rapid Downpower INSTRUCTIONS AOP 2575 Revision 004-01 Page 7 of 35 CONTINGENCY ACTIONS CAUTION In the case of a dropped downpc-wer. rod motion is nat used to initiate 3.4 !E not down powering due to 11 dropped rod, INSERT Group 7 CEAs J 0 +/- 2 steps to initiate downpower. _3.5 Refer 'Ii> ppe or Reactor 3.5.1 Engineering Curve and Data Book illld OBTAIN reactivity plan for the inital reactor power condition and desired \\Gad reduction. 3.6 IF desired to borate from the RWST (prcfern:~d method) PERFORM the following;

a.

ENSURE at least one charging pump uperating.

h.

ENSURE CH-!.96, vcr makeup bypass, dused. ('. ENSURE CH-S04. RWST to charging suction. open.

d.

OPEN CH-192, RWST isolat ion.

c.

CLOSE CH-511L VCT outlet isolation.

f.

CHECK charging flow at desired rate. 3.7 Based on required rate of downpower, START additional charging pumps as necessary and balance charging and letdown.. LevelofUs~ Continuou~ STOP THINK IE reactor is !lot at the reactivity plan initial conditions. Refer To. and DETERMINE desired rate of load reduction for time in core life. ACT d.l IF CH*192. RWST isolation, can not be opened, Go'Iil step 3.8. f.1 START additional charging pumps as needed and balance charging and letdown. REVIEW Lower Order? Page 61 of 75 Printed on 1/28/2010 at 12:02

SRO Exam Questions Only (No "Parents" Or "Originals") Question #: 20 Question ID; 9000014 [J RO Ii'l SRO o Student Handout? Ii'l Lower Order? l-SRO Ques. If 20 Rev. 0 Ii'l Selected for Exam Origin: New o Past NRC E:Kam? Which of U,e following actions require authorization by the Refueling SRO? o A 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 OB from the guidE! pins. Oc In an emergency, insert a fuel assembly into the core and ungrapple it provided NO other fuel assemblies are adjacent. Ii'l D If an underload occurs prematurely, raise the fuel assembly, pull the mast detent pin, rotate slightly, and reinsert the assembly. Justification I o IS CORRECT; OP 22D9A, 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 mast to allow inserting a fuel assembly. A is incorrect..A. Cautior 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 incormct; While Attachment 4 allows the Refuel SRO to authorize several actions to free a fuel assembly (seen as an overload), manually manipulating the cable hoist is NOT one of them. The refuel machine may be moved in either horizontal plane to free an assembly. Plausible because Attacllment 4 allows the Refuel SRO to authorize movement of the refuel machine manually (hand crank) in either horizontal direction, just NOT in the vertical direction. C is incorrect; A fuel as~;embly 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. Refueling Operations Comments and Questitm Modification History I Bob K. - D-5/C (Operator involvement in refuel operations is very limited. Question is still acceptable). Bill M. - D-4fW, G (Didn't know due to limited involvement in fuel movement. Question is acceptable.) Angelo - D-5/W; Not an operator task. Bruce F. - D-2/W. 50/50 (C and D) NRC KIA S'vstem/E/A System 2.1 Conduct of Operations I Gener~=KlA_~lecte(O NRC KIA Generic System 2.1 Conduct of Operations Number 2.1.35 R02.2 SR03.9 CFRLink(CFR:41.10/43.7) Knowledge of the fuel-handling responsibilities of SROs. Page 62 of 75 Printed on 1/28/2010 at 12:02 SRO Exam Questions Only (No "Parents" Or "Originals") Question #: 20 l-SRO Ques. If 20 Question ID; 9000014 [J RO Ii'l SRO Rev. 0 Ii'l Selected for Exam o Student Handout? Origin: New Which of U,e following actions require authorization by the Refueling SRO? Ii'l Lower Order? o Past NRC E:Kam? o A When the grapple will NOT disengage the top of a fuel assembly, snap or twang the hoist cable to release the grapple. OB When an overload occurs, use the hand crank on the refuel machine hoist to free the fuel assembly from the guidE! pins. Oc In an emergency, insert a fuel assembly into the core and ungrapple it provided NO other fuel assemblies are adjacent. Ii'l D If an underload occurs prematurely, raise the fuel assembly, pull the mast detent pin, rotate slightly, and reinsert the assembly. Justification I o IS CORRECT; OP 22D9A, 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 mast to allow inserting a fuel assembly. A is incorrect..A. Cautior 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 incormct; While Attachment 4 allows the Refuel SRO to authorize several actions to free a fuel assembly (seen as an overload), manually manipulating the cable hoist is NOT one of them. The refuel machine may be moved in either horizontal plane to free an assembly. Plausible because Attacllment 4 allows the Refuel SRO to authorize movement of the refuel machine manually (hand crank) in either horizontal direction, just NOT in the vertical direction. C is incorrect; A fuel as~;embly 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. Refueling Operations Comments and Questitm Modification History I Bob K. - D-5/C (Operator involvement in refuel operations is very limited. Question is still acceptable). Bill M. - D-4fW, G (Didn't know due to limited involvement in fuel movement. Question is acceptable.) Angelo - D-5/W; Not an operator task. Bruce F. - D-2/W. 50/50 (C and D) NRC KIA S'vstem/E/A System 2.1 Conduct of Operations I Gener~=KlA_~lecte(O NRC KIA Generic System 2.1 Conduct of Operations Number 2.1.35 R02.2 SR03.9 CFRLink(CFR:41.10/43.7) Knowledge of the fuel-handling responsibilities of SROs. Page 62 of 75 Printed on 1/28/2010 at 12:02

SRO Exam Questions Only (No "~arellts"Or "Originals") Question #: 20 Question ID: 9000014 []RO ~ SRO o Student Handout? ~ Lower Order? II-SRO Ques. II 20 Rev. 0 ~ Selected for Exam Origin: New o Past NRC E:l(am? II Personnel Responsibilities During Refueling Operations (Shee( 5 of 6)

5. Refuel SRO Authority Stop CORE ALTERATIONS \\'<'hen deemed necessary.

Stop or defer any a<.1ivity around the refuel l100r which would jeopardize the safety of personnel or equipment. R.csponsibilities Be present on the Refuel Floor and responsible for maintaining OPS Procedures as required during the following CORE ALTERATIONS: Fuel Shuffle Movingfreplacing sources CEA shuffle in the reactor vessel Removing the upper guide structure from the reactor vessel I Uncoupling of CEA extension shafts Removal of CEA extension shafts from the lJGS Recoupling of CEA extension shafts Any other CORE ALTERATION as dctermined by Reactor Engineering As necessary, thc Refuel SRO following consultation with Reactor Engineering, authorizes that specified guidelincs Oil Attachment 4 bc 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-102, "Foreign Material Exdusi{}n" and DNAP-2000. "Dominion Work Management Process." Ensure proper radiological practices arc maintained around the refuel pool. Ensure safe load paths per MP 2712132, "Overhead Crane Operating Information;' and MP 2712131, "Control Heavy Load" are maintained. Ensure general safety of personnel and equipment. Ensure communications between refuel floor and the Control Room are maintained during CORE ALTERATIONS. Halt CORE ALTERATIONS if communications are lost. Ensure refueling equipment is operated in accordance with OPS-FH 215, "Refueling Machine Operation."

--L-ev-e-I-o-fu-'~

OP 2209A Referenc~ STOP THINK AGT REVIEW Rev. 026-06 46 of 64 Page 63 of75 Printed on 1/28/2010 at 12:02 Question #: 20 II-SRO Ques. II 20 II SRO Exam Questions Only (No "~arellts"Or "Originals") Question ID: 9000014 []RO ~ SRO o Student Handout? Rev. 0 ~ Selected for Exam Origin: New Personnel Responsibilities During Refueling Operations (Shee( 5 of 6) ~ Lower Order? o Past NRC E:l(am?

5. Refuel SRO Authority Stop CORE ALTERATIONS \\'<'hen deemed necessary.

Stop or defer any a<.1ivity around the refuel l100r which would jeopardize the safety of personnel or equipment. R.csponsibilities Be present on the Refuel Floor and responsible for maintaining OPS Procedures as required during the following CORE ALTERATIONS: Fuel Shuffle Movingfreplacing sources CEA shuffle in the reactor vessel Removing the upper guide structure from the reactor vessel Uncoupling of CEA extension shafts Removal of CEA extension shafts from the lJGS Recoupling of CEA extension shafts Any other CORE ALTERATION as dctermined by Reactor Engineering As necessary, thc Refuel SRO following consultation with Reactor Engineering, authorizes that specified guidelincs Oil Attachment 4 bc 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-102, "Foreign Material Exdusi{}n" and DNAP-2000. "Dominion Work Management Process." Ensure proper radiological practices arc maintained around the refuel pool. Ensure safe load paths per MP 2712132, "Overhead Crane Operating Information;' and MP 2712131, "Control Heavy Load" are maintained. Ensure general safety of personnel and equipment. Ensure communications between refuel floor and the Control Room are maintained during CORE ALTERATIONS. Halt CORE ALTERATIONS if communications are lost. Ensure refueling equipment is operated in accordance with OPS-FH 215, "Refueling Machine Operation."

--L-ev-e-I-o-f u-'~

OP 2209A Referenc~ STOP THINK AGT REVIEW Rev. 026-06 46 of 64 I Page 63 of75 Printed on 1/28/2010 at 12:02

Question #: 20 Student Handout? ~ Lower Order? Question ID: 9000014 RO ~SRO 20 Rev. o ~ Selected for Exam Origin: New Past NRC Eltam? Guidelines For Fuel Movement Operations (Sheet:! of 5) CALJTION Snapping or twanging ()f the hoist cable is prohibited. Hoist Cahle Manipulation Manipulation or pulling on the hoist cable is typically recommended only to assist in the follo~i!lg: Allnw engagement of the bottom nozzle with the (.,"orc support plate guide pins. Facilitate fuel movement when "hang-ups" or problems with grapple engagement or disengagement are encountered. Facilitate fuel movement when potential grit! interferences are encountered. Difficulty Inserting IE inserting a fuel assembly into the core, SFP storage rack, fuel elevator. or llpender can. an underload occurs. PERFORM the following: 1, RAISE the assembly until the underload is cleared. 2, CHECK alignment offuel a,,-;embly and fixture. 3, IF repositioning a fuel a5semblymuIluully over the core, ENSURE spreader is raised.

4. As necessary, REPOSITION fuel as..'icmbly and TRY rein.'>crting.

5, IF an underload is experienced again, PERFORM allY of the following: 5.1 PU LL the RFM mast detent pin out and TRY reinserting. 5.2 ROTATE the RFM mast 51 ightly in the clockwise or counterclockwise direction and TRY reinserting. 5.3 ENSURE fuel assembly is raised 4" from the core support plate to clear the guide pins and HAND CRANK the RFM up to 0.3" in any direction and TRY reinserting. 5.4 the above does not clear underload, manipulate the hoist cable to free the fuel assembly from potential grid interferences. OP 2209A STOP THINK ACT REVIEW Rev. 026-06 Page 64 of 75 Printed on 1/28/2010 at 12:02 Question #: 20 Question ID: 9000014 RO ~SRO Student Handout? ~ Lower Order? 20 Rev. o ~ Selected for Exam Origin: Guidelines For Fuel Movement Operations (Sheet:! of 5) CALJTION New Snapping or twanging ()f the hoist cable is prohibited. Hoist Cahle Manipulation Manipulation or pulling on the hoist cable is typically recommended only to assist in the follo~i!lg: Allnw engagement of the bottom nozzle with the (.,"orc support plate guide pins. Facilitate fuel movement when "hang-ups" or problems with grapple engagement or disengagement are encountered. Facilitate fuel movement when potential grit! interferences are encountered. Difficulty Inserting IE inserting a fuel assembly into the core, SFP storage rack, fuel elevator. or llpender can. an underload occurs. PERFORM the following: 1, RAISE the assembly until the underload is cleared. 2, CHECK alignment offuel a,,-;embly and fixture. 3, IF repositioning a fuel a5semblymuIluully over the core, ENSURE spreader is raised.

4. As necessary, REPOSITION fuel as..'icmbly and TRY rein.'>crting.

5, IF an underload is experienced again, PERFORM allY of the following: 5.1 PU LL the RFM mast detent pin out and TRY reinserting. 5.2 ROTATE the RFM mast 51 ightly in the clockwise or counterclockwise direction and TRY reinserting. 5.3 ENSURE fuel assembly is raised 4" from the core support plate to clear the guide pins and HAND CRANK the RFM up to 0.3" in any direction and TRY reinserting. 5.4 the above does not clear underload, manipulate the hoist cable to free the fuel assembly from potential grid interferences. STOP THINK ACT REVIEW OP 2209A Rev. 026-06 Past NRC Eltam? Page 64 of 75 Printed on 1/28/2010 at 12:02

Question #: 21 RO .~. SRO Student Handout? D Lower Order? Question ID: 9000024 r-SRO Ques. # 21 Rev. 1 ."'. Selected for Exam Odgin: New Past NRC Exam? The plant is operating at 1 00% power when ISO New England and CONVEX operators notify Millstone Station that a "Deg.raded 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? D 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 In anticipation of a plant trip, ensure "E" & "F" lACs are operating in the "Lead" and "Standby" modes.

JuStificatlon'!

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 running of t1ese components must be terminated. A - WRONG; The RSST breakers are not disabled until voltage drops below 88.9% of rated voltage. The EDG 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 breakl9rs must be disabled so they will NOT close on a possible plant trip. Also, prestaging the EDGs with a "slow start" would put minimum stress on the machines which are destined to carry all plant AC loads. However, this action is premature and over conservative 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. D 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' "" NRC - (original question comments) Add "In anticipation of a plant trip,.. to the beginning of Choice "0" and reword to make length comparable. RLC - Reworded Choice "0" per NRC comments. [11/30/09] Bruce F. D-3/C, No comment NRC KIA System/E/A System 2.2 Equipment Control NRC KIA Generic System 2.2 Equipment Control Number 2.2.17 RO 2.6 SRO 3.8 CFR Link (CFR: 41.10 I 43.5/45.13) Knowledge of the proce~;s for managing maintenance activities during power operations, such as risk assessments, work prioritization, and coordination with the transmission system operator. Page 65 of 75 Printed on 1/28/2010 at 12:02 Question #: 21 Question ID: 9000024 RO .~. SRO Student Handout? D Lower Order? r-SRO Ques. # 21 Rev. 1 ."'. Selected for Exam Odgin: New Past NRC Exam? The plant is operating at 1 00% power when ISO New England and CONVEX operators notify Millstone Station that a "Deg.raded 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? D 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 In anticipation of a plant trip, ensure "E" & "F" lACs are operating in the "Lead" and "Standby" modes.

JuStificatlon'!

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 running of t1ese components must be terminated. A - WRONG; The RSST breakers are not disabled until voltage drops below 88.9% of rated voltage. The EDG 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 breakl9rs must be disabled so they will NOT close on a possible plant trip. Also, prestaging the EDGs with a "slow start" would put minimum stress on the machines which are destined to carry all plant AC loads. However, this action is premature and over conservative 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. D 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' "" NRC - (original question comments) Add "In anticipation of a plant trip,.. to the beginning of Choice "0" and reword to make length comparable. RLC - Reworded Choice "0" per NRC comments. [11/30/09] Bruce F. D-3/C, No comment NRC KIA System/E/A System 2.2 Equipment Control NRC KIA Generic System 2.2 Equipment Control Number 2.2.17 RO 2.6 SRO 3.8 CFR Link (CFR: 41.10 I 43.5/45.13) Knowledge of the proce~;s for managing maintenance activities during power operations, such as risk assessments, work prioritization, and coordination with the transmission system operator. Page 65 of 75 Printed on 1/28/2010 at 12:02

Question It: 21 Student Handout? Lower Order? Question ID: 9000024 RO ~ SRO I-SRO Ques. # 21 Rev. 1 ~ Selected for Exam Origin: New Past NRC Exam? Mililstonc Unit 2 Degraded Voltage 3.0 Degraded Voltage INSTRUCtIONS 3.1 IF surveillunces of safety related pumps and motors arc in progress, TERMINATE surveillances during degraded voltage conditions. 3.2 REQlJ&~T the 8M refer to COP 200.8, "Response to ISO NE/CONEX Emergencies and Alerts." '" :U ClIEO!( actual degraded voltage condition exists by observation of Al'IIY or the following conditions: 4160 volt bus 24C 240 voltage less than 3,900 volts 480 volt bus 22E 22F voltage les:; than 440 volts AOP 2580 Revision 003-04 Page 6ofl2 CONTINGENCY ACTIONS Level of Use STOP THINK ACT REVIEW Continuous Page 66 of 75 Printed on 1/28/2010 at 12:02 Question ID: 9000024 RO ~ SRO Student Handout? Question It: 21 I-SRO Ques. # 21 Rev. 1 ~ Selected for Exam Origin: New Mililstonc Unit 2 Degraded Voltage 3.0 Degraded Voltage INSTRUCtIONS 3.1 IF surveillunces of safety related pumps and motors arc in progress, TERMINATE surveillances during degraded voltage conditions. 3.2 REQlJ&~T the 8M refer to COP 200.8, "Response to ISO NE/CONEX Emergencies and Alerts." '" :U ClIEO!( actual degraded voltage condition exists by observation of Al'IIY or the following conditions: 4160 volt bus 24C 240 voltage less than 3,900 volts 480 volt bus 22E 22F voltage les:; than 440 volts Level of Use Continuous STOP THINK AOP 2580 Revision 003-04 Page 6ofl2 CONTINGENCY ACTIONS ACT REVIEW Lower Order? Past NRC Exam? Page 66 of 75 Printed on 1/28/2010 at 12:02

SRO ExaJtt\\GHesflons Only (No "Parents" Or "OHghfaI~uJ. Question #: 22 Question ID: 53551 RO .Iof. SRO L Student Handout? ~ Lower Order? I-SRO Ques. 11 22 Rev. 2 ~ Selected for Exam Origin: Bank Past NRC Exam? A plant startup from a mid-cycle maintenance outage is in progress and the reactor has just been declared critical. Shortly after going critical, the SPO inadvertently overfeeds the S/Gs, and then regains control. This action results in average ReS temperature lowering to 519 of and then stabilizing. Which of the following operator actions should be taken, based on administrative requirements? ~ A Determine ReS Tave is within its limit once per hour until Tave reaches or exceeds 525 of. B Restore Tavg to within its limit within 15 minutes or be in HOT STANDBY within the next 15 minutes. o C Emergency Boration must be commenced immediately. D Immediately trip the Reactor and carry out EOP 2525. Justification I A - CORRECT; Per T.S. 3.1.1.5, when ever the reactor is critical, RCS Tavg must be >1= 515°F at all times. However, the T.S. Surveillance Requirements state that if Tavg drops below 525°F, then Tavg must be verified to be >1= 515°F (within its limit) at least once per hour. B - WRONG; This is the required action if Tavg dropped below 515°F. Plausible; Examinee may believe the limit for action is < 525*F as that is the reactor startup procedural limit (OP-2202). C - WRONG; This is the required action if Tavg dropped below 500*F. Plausible; Examinee may believe an uncontrolled cool down below the procedural required limit necessitates immediate action to ensure reactivity control. D - WRONG; This is the required action if the cooldown was not stopped (temperature control is not regained) before Tavg drops below 500"F. Plausible; Examinee may believe an uncontrolled cool down below the T.S. limit requires immediate action to ensure plant control. References I Tech Spec Surveillance 4.1.1.5 Comments and Question Modification History I NRC* (original question comments) Question is LOD 5, not required knowledge from memory. Replace question with one testing Tech. Spec. knowledge. RLC - Original question #220 replaced by Item #53551, rev. 2. Bruce F. - D-2/C, No comment NRC KlA System/E:/A System 2.2 Equipment Control I Gener~ KIA. Selecte(~ NRC KlA Generic System 2.2 Equipment Control Number 2.2:.38 RO 3.6 SRO 4.5 CFR Link (CFR: 41.7 141.10 143.1 145.13) Knowledge of conditions and limitations in the facility license. Page 67 of 75 Printed on 1/28/2010 at 12:02 SRO ExaJtt\\GHesflons Only (No "Parents" Or "OHghfaI~uJ. Question #: 22 Question ID: 53551 RO .Iof. SRO L Student Handout? ~ Lower Order? I-SRO Ques. 11 22 Rev. 2 ~ Selected for Exam Origin: Bank Past NRC Exam? A plant startup from a mid-cycle maintenance outage is in progress and the reactor has just been declared critical. Shortly after going critical, the SPO inadvertently overfeeds the S/Gs, and then regains control. This action results in average ReS temperature lowering to 519 of and then stabilizing. Which of the following operator actions should be taken, based on administrative requirements? ~ A Determine ReS Tave is within its limit once per hour until Tave reaches or exceeds 525 of. B Restore Tavg to within its limit within 15 minutes or be in HOT STANDBY within the next 15 minutes. o C Emergency Boration must be commenced immediately. D Immediately trip the Reactor and carry out EOP 2525. Justification I A - CORRECT; Per T.S. 3.1.1.5, when ever the reactor is critical, RCS Tavg must be >1= 515°F at all times. However, the T.S. Surveillance Requirements state that if Tavg drops below 525°F, then Tavg must be verified to be >1= 515°F (within its limit) at least once per hour. B - WRONG; This is the required action if Tavg dropped below 515°F. Plausible; Examinee may believe the limit for action is < 525*F as that is the reactor startup procedural limit (OP-2202). C - WRONG; This is the required action if Tavg dropped below 500*F. Plausible; Examinee may believe an uncontrolled cool down below the procedural required limit necessitates immediate action to ensure reactivity control. D - WRONG; This is the required action if the cooldown was not stopped (temperature control is not regained) before Tavg drops below 500"F. Plausible; Examinee may believe an uncontrolled cool down below the T.S. limit requires immediate action to ensure plant control. References I Tech Spec Surveillance 4.1.1.5 Comments and Question Modification History I NRC* (original question comments) Question is LOD 5, not required knowledge from memory. Replace question with one testing Tech. Spec. knowledge. RLC - Original question #220 replaced by Item #53551, rev. 2. Bruce F. - D-2/C, No comment NRC KlA System/E:/A System 2.2 Equipment Control I Gener~ KIA. Selecte(~ NRC KlA Generic System 2.2 Equipment Control Number 2.2:.38 RO 3.6 SRO 4.5 CFR Link (CFR: 41.7 1 41.10 143.1 145.13) Knowledge of conditions and limitations in the facility license. Page 67 of 75 Printed on 1/28/2010 at 12:02

Question #: 22 liRa ~ SRO Student Handout? ~l Lower Order? Question ID: 53551 I-SRO Ques..¥ 22 Rev. 2 ~ Selected for Exam Origin: Bank Past NRC Exam? September 25, 2003 MIl\\"1M1J:rvl TEMPERATURE FOR CRITICALITY LIMITING COl'fDITION FOR OPERATION 3.1.1.5 The Reactor Coolant System temperature (Tavg) shall be ;::: 515°F when the reactor is critical. MODES laud i'. With the Reactor Coolant System temperature (Tavg) < 515°F, restore Tavg to within its limit within 15 minutes or be in HOT STANDBY within the next 15 minutes. SURVEILLANCE REQUIREMENTS 4.1. 15 The Reactor Coolant System temperature (Tavg) shall be determined to be ;::: 515°E

3.

Within 15 minutes prior to making the reactor critical. and

h.

At least once per honr when the reactor is critical and the Reactor Coolant System temperature (Tavg) is < 525°E

With 1.0.

MILLSTONE - UNIT 2 3/41-7 AMENDMENT NO. -M. 280 Page 68 of75 Printed on 1/28/2010 at 12:02 Question #: 22 I-SRO Ques.. ¥ 22 Question ID: 53551 Rev. 2 liRa ~ SRO ~ Selected for Exam Student Handout? ~l Lower Order? Origin: Bank Past NRC Exam? September 25, 2003 MIl\\"1M1J:rvl TEMPERATURE FOR CRITICALITY LIMITING COl'fDITION FOR OPERATION 3.1.1.5 critical. The Reactor Coolant System temperature (Tavg) shall be ;::: 515°F when the reactor is MODES laud i'. With the Reactor Coolant System temperature (Tavg) < 515°F, restore Tavg to within its limit within 15 minutes or be in HOT STANDBY within the next 15 minutes. SURVEILLANCE REQUIREMENTS 4.1. 15 The Reactor Coolant System temperature (Tavg) shall be determined to be ;::: 515°E

3.

Within 15 minutes prior to making the reactor critical. and

h.

At least once per honr when the reactor is critical and the Reactor Coolant System temperature (Tavg) is < 525°E

With 1.0.

MILLSTONE - UNIT 2 3/41-7 AMENDMENT NO. -M. 280 Page 68 of75 Printed on 1/28/2010 at 12:02

Question #: 23 Question ID: 9000026 RO [ii] SRO Student Handout? [ii] Lower Order? [-SRO Ques. # 23 Rev. o [ii] Selected for Exam Origin: New Past NRC Exam? While preparing for an Aerated Waste Monitor Tank (AWMT) discharge, the PEO reported to the Shift Manager that he was unable to access PlOPS for the initial setup of the Aerated Waste System radiation monitor (RM-9116) because the PlOPS keypad appears to be failed. I&C reports it will take several hours to repair or replace the keypad. Which one of the following actions is required in order for the SM to approve the discharging of the AWMT? A A SECOND Chemist must FIRST approve the analysis results and flow requirements stated on the existing AWMT discharge permit. B The Chemist rnust be briefed to ensure the AWMT discharge is sampled 15 minutes AFTER the discharge has commenced. C The AWMT may be discharged if the Alert and Alarm limits presently set in PlOPS are lower than the limits the existing permit. D The PlOPS ke,ypad must FIRST be repaired to allow completion of the initial setup or another AWMT discharge permit is required. Justification.., D; CORRECT - SP-2617A requires the setup and limits programmed into the discharge radmonitor by PlOPS (Programmable Input/Output Processing System) be verified, up to and including a check of the I&C parameters for calibration and the performance of a "source check". None (If these can be done without accessing PlOPS. Therefore, the keypad must be repaired in order to perform the discharge under the existing permit, or a new permit must be generated following the procedure for an inoperable rad. monitor. A; WRONG - Tilis action does not meet the administrative requirements for discharging with an inoperable radmonitor. Plausible; The E~xaminee may recognize that the discharge may be performed under the existing conditions if the radmonitor is considered to be inoperable. However, the administrative requirements to discharge with an inoperable radmonitor state that a second sample must bEl drawn 81d analyzed by a second Chemist, not just reviewed by a second Chemist. B; WRONG - Tilis will help verify the radmonitor is operating and tracking the discharge radiation levels properly, but not that it is set up correctly to isolate it if limits are exceeded. Plausible; Examinee may believe this will ensure the limits stated in the discharge permit are met, regardless of the radmonitor setup. C; WRONG - This would meet the radiological requirements of the discharge if the radmonitor was known to be operable and in calibration. ThElse aspects can not be verified without acceSSing PlOPS. Plausible; Examinee may believe the setup of the AWMT radmonitor is irrelevant as long as the limits known to exist in PlOPS are more conservative than t10se on the discharge permit. References n4)t Comments and Question Modification History I New question generated, based on NRC and Exam Validator comments on previous questions. - RLC 12/23/09 NRC KIA System/E/A KIA System 2.3 Radiation Control System 2.3 Radiation Control Number 2.3.11 RO 2.7 SRO 3.2 CFR Link (CFR: 45.9/45.10) Ability to control radiation releases. NRC KIA Generic Page 69 of 75 Printed on 1/28/2010 ;3t 12:02 Question #: 23 Question ID: 9000026 [-SRO Ques. # 23 Rev. o RO [ii] SRO [ii] Selected for Exam Student Handout? Origin: New [ii] Lower Order? Past NRC Exam? While preparing for an Aerated Waste Monitor Tank (AWMT) discharge, the PEO reported to the Shift Manager that he was unable to access PlOPS for the initial setup of the Aerated Waste System radiation monitor (RM-9116) because the PlOPS keypad appears to be failed. I&C reports it will take several hours to repair or replace the keypad. Which one of the following actions is required in order for the SM to approve the discharging of the AWMT? A A SECOND Chemist must FIRST approve the analysis results and flow requirements stated on the existing AWMT discharge permit. B The Chemist rnust be briefed to ensure the AWMT discharge is sampled 15 minutes AFTER the discharge has commenced. C The AWMT may be discharged if the Alert and Alarm limits presently set in PlOPS are lower than the limits the existing permit. D The PlOPS ke,ypad must FIRST be repaired to allow completion of the initial setup or another AWMT discharge permit is required. Justification.., D; CORRECT - SP-2617A requires the setup and limits programmed into the discharge radmonitor by PlOPS (Programmable Input/Output Processing System) be verified, up to and including a check of the I&C parameters for calibration and the performance of a "source check". None (If these can be done without accessing PlOPS. Therefore, the keypad must be repaired in order to perform the discharge under the existing permit, or a new permit must be generated following the procedure for an inoperable rad. monitor. A; WRONG - Tilis action does not meet the administrative requirements for discharging with an inoperable radmonitor. Plausible; The E~xaminee may recognize that the discharge may be performed under the existing conditions if the radmonitor is considered to be inoperable. However, the administrative requirements to discharge with an inoperable radmonitor state that a second sample must bEl drawn 81d analyzed by a second Chemist, not just reviewed by a second Chemist. B; WRONG - Tilis will help verify the radmonitor is operating and tracking the discharge radiation levels properly, but not that it is set up correctly to isolate it if limits are exceeded. Plausible; Examinee may believe this will ensure the limits stated in the discharge permit are met, regardless of the radmonitor setup. C; WRONG - This would meet the radiological requirements of the discharge if the radmonitor was known to be operable and in calibration. ThElse aspects can not be verified without acceSSing PlOPS. Plausible; Examinee may believe the setup of the AWMT radmonitor is irrelevant as long as the limits known to exist in PlOPS are more conservative than t10se on the discharge permit. References n4)t Comments and Question Modification History I New question generated, based on NRC and Exam Validator comments on previous questions. - RLC 12/23/09 NRC KIA System/E/A System 2.3 Radiation Control KIA NRC KIA Generic System 2.3 Radiation Control Number 2.3.11 RO 2.7 SRO 3.2 CFR Link (CFR: 45.9/45.10) Ability to control radiation releases. Page 69 of 75 Printed on 1/28/2010 ;3t 12:02

Question #: 23 Question ID: 9000026 RO ~SRO Student Handout? Iil Lower Order? J-SRO Que,. If 23 Rev. o ~ Selected for Exam Otigin: New Past NRC E:lCam?

1. PURPOSE 1.1 Objective This procedure provides instructions to ensure during all MODES of operation all ret)uirements are met prior to and during all radioactive discharges from the following:

Aerated Liquid Radwaste System (A WMT), to satisfy REMODCM IV.C.l Surveillance Requirement, "Pdble ry.c-2, items l.b. and 3.b., "Radioactive Liquid Effluent Monitoring Instrumentation." Clean Liquid Radwaste System (CWMT..,,), to satisfy REMODCM IV.C.l Surveillance Requirement, Table IY.C-2, items La. and 3.3., "Radioactive Liquid Effluent Monitoring Instrumentation." 1.2 Discussion The objective of this procedure is satisfied by performing the following: SOURCE CHECK of applicable rauiHlion monitor prior to discharge: IF discharging CWMT, RE-9049 IF discharging AWMT. RE-9116 Checking flow recorders OPERABLE (tracking flow and indicating), prior discharge: IF di'icharging CWMT, "FR-9050, SYSTEM DISCHARGE FLOW RECORDER" IF discharging AWMT, "SYSTEM DISCHARGE FLOW. FR-9118" CHANNEL CHECK of applicable radiation monitor, 15 minutes after initiation of discharge If discharge is performed with radiation monitor not OPERABLE, ensuring 2 independent samples have been taken and analyzed Performing dual alld indepenuent verification of necessary steps for all discharges. CHANNEL CHECK, as used in this procedure, refers to the qualitative assessment of applicable radiation monitor behavior during operation, as compared to other similar indications of the same instrumentation. The Aerated and Clean Waste Rauiation Monitors have design nmges of 10 to 106 cpm [FSAR Table 7.5-5J. These radiation monitors and PlOPS ure not OPERABLE above l{)6 cpm. SP2617A LevelofLi;l STOP THINK ACT REVIEW Rev. 029-06 Continuc~ 20fSS Page 70 of 75 Printed on 1/28/2010 at 12:02 Question #: 23 Question ID: 9000026 RO ~SRO Student Handout? Iil Lower Order? J-SRO Que,. If 23 Rev. o ~ Selected for Exam Otigin: New

1. PURPOSE 1.1 Objective This procedure provides instructions to ensure during all MODES of operation all ret)uirements are met prior to and during all radioactive discharges from the following:

Aerated Liquid Radwaste System (A WMT), to satisfy REMODCM IV.C.l Surveillance Requirement, "Pdble ry.c-2, items l.b. and 3.b., "Radioactive Liquid Effluent Monitoring Instrumentation." Clean Liquid Radwaste System (CWMT..,,), to satisfy REMODCM IV.C.l Surveillance Requirement, Table IY.C-2, items La. and 3.3., "Radioactive Liquid Effluent Monitoring Instrumentation." 1.2 Discussion The objective of this procedure is satisfied by performing the following: SOURCE CHECK of applicable rauiHlion monitor prior to discharge: IF discharging CWMT, RE-9049 IF discharging AWMT. RE-9116 Checking flow recorders OPERABLE (tracking flow and indicating), prior discharge: IF di'icharging CWMT, "FR-9050, SYSTEM DISCHARGE FLOW RECORDER" IF discharging AWMT, "SYSTEM DISCHARGE FLOW. FR-9118" CHANNEL CHECK of applicable radiation monitor, 15 minutes after initiation of discharge If discharge is performed with radiation monitor not OPERABLE, ensuring 2 independent samples have been taken and analyzed Performing dual alld indepenuent verification of necessary steps for all discharges. CHANNEL CHECK, as used in this procedure, refers to the qualitative assessment of applicable radiation monitor behavior during operation, as compared to other similar indications of the same instrumentation. The Aerated and Clean Waste Rauiation Monitors have design nmges of 10 to 106 cpm [FSAR Table 7.5-5J. These radiation monitors and PlOPS ure not OPERABLE above l{)6 cpm. LevelofLi;l Continuc~ STOP THINK ACT REVIEW SP2617A Rev. 029-06 20fSS Past NRC E:lCam? Page 70 of 75 Printed on 1/28/2010 at 12:02

Question #: 24 Question ID: 9000016 [J RO .". SRO Student Handout? Lower Ord'er? I-SRO Ques. If 24 Rev. 2 ~ Selected for Exam Origin: New Past NRC EKam? The plant was operating normally at 100% power with ReS coolant activity at the maximum allowed limit due to known fuel defects. The crew manually tripped the plant due to rapidly lowering Pressurizer pressure and level. EOP 2525, Standard post Trip Actions, was successfully performed and the crew has entered EOP 2534, Steam Gem~rator Tube Rupture. Why does this pro::::edure require the crew to perform a cooldown on both Steam Generators until both hot leg temperatures are less than or equal to 515°F prior to isolating the affected Steam Generator? A The cooldown can be performed quicker with both SIG allowing earlier isolation of the affected S/G. B If the affected SIG is isolated prior to a cooldown, then increasing radiation levels in the affected SIG will result in a higher release later. e The initial cooldown will Significantly lower the affected SIG level resulting in more available volume for the coolant from the tube rupture. ~ D To limit the radiation release due to the potential for lifting the Main Steam Safeties on the affected S/G. Justification I D IS CORRECT; The Tochnical Basis for this step simply states that this is the reason for doing the cool down to 515"F before iSDlating theSG. A is incorrect; While the cool down can be performed more quickly on two steam Generators versus one Steam Generator, this is, NOT the reason for the cool down to 515°F. Plausible because this is a true statement; however, it answers the question, "Why use two Steam generators to cool down?" It does NOT answer the question, "Why cool down to 515"F?" B is incorrect; If the affected Steam Generator were isolated prior to the cooldown, radiation levels would actually be lower later. The higher Steam Generator pressure would limit the leakage of coolant from the RCS to the Steam Generator, resulting in lower activity levels in the affected Steam Generator. Plausible because the examinee could think that activity levels in the affected Steam Generator would continue to rise resulting in higher activity levels on H later release. C is incorrect; A potentililly true statement, but it is NOT the reason for cooling down to 515"F with both Steam Generators. Plausible because COOling down the affected Steam Generator could result in less inventory and allow more room for leakage from the RCS. References I EOP 2534. Steam Genel'alor Tube Rupture and associated Technical Guide Comments and Questi4>n Modification History I Potential additional distracter: "The initial cooldown of both Steam Generators ensures that both Steam Generators remain coupl!~d during the subsequent cooldown." Bruce F. - D-2/C, No comment NRC KIA System/E:/A System 2.3 Radiation Control NRC KIA Generic System 2.3 Radiation Control Number 2.3.14 RO 3.4 SRO 3.8 CFR Link (CFR: 41.12/43.4 /45.10) Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities. Page 71 of 75 Printed on 1/28/2010 at 12:02 Question #: 24 I-SRO Ques. If 24 Question ID: 9000016 [J RO .". SRO Rev. 2 ~ Selected for Exam Student Handout? Lower Ord'er? Origin: New Past NRC EKam? The plant was operating normally at 100% power with ReS coolant activity at the maximum allowed limit due to known fuel defects. The crew manually tripped the plant due to rapidly lowering Pressurizer pressure and level. EOP 2525, Standard post Trip Actions, was successfully performed and the crew has entered EOP 2534, Steam Gem~rator Tube Rupture. Why does this pro::::edure require the crew to perform a cooldown on both Steam Generators until both hot leg temperatures are less than or equal to 515°F prior to isolating the affected Steam Generator? A The cooldown can be performed quicker with both SIG allowing earlier isolation of the affected S/G. B If the affected SIG is isolated prior to a cooldown, then increasing radiation levels in the affected SIG will result in a higher release later. e The initial cooldown will Significantly lower the affected SIG level resulting in more available volume for the coolant from the tube rupture. ~ D To limit the radiation release due to the potential for lifting the Main Steam Safeties on the affected S/G. Justification I D IS CORRECT; The Tochnical Basis for this step simply states that this is the reason for doing the cool down to 515"F before iSDlating theSG. A is incorrect; While the cool down can be performed more quickly on two steam Generators versus one Steam Generator, this is, NOT the reason for the cool down to 515°F. Plausible because this is a true statement; however, it answers the question, "Why use two Steam generators to cool down?" It does NOT answer the question, "Why cool down to 515"F?" B is incorrect; If the affected Steam Generator were isolated prior to the cooldown, radiation levels would actually be lower later. The higher Steam Generator pressure would limit the leakage of coolant from the RCS to the Steam Generator, resulting in lower activity levels in the affected Steam Generator. Plausible because the examinee could think that activity levels in the affected Steam Generator would continue to rise resulting in higher activity levels on H later release. C is incorrect; A potentililly true statement, but it is NOT the reason for cooling down to 515"F with both Steam Generators. Plausible because COOling down the affected Steam Generator could result in less inventory and allow more room for leakage from the RCS. References I EOP 2534. Steam Genel'alor Tube Rupture and associated Technical Guide Comments and Questi4>n Modification History I Potential additional distracter: "The initial cooldown of both Steam Generators ensures that both Steam Generators remain coupl!~d during the subsequent cooldown." Bruce F. - D-2/C, No comment NRC KIA System/E:/A System 2.3 Radiation Control NRC KIA Generic System 2.3 Radiation Control Number 2.3.14 RO 3.4 SRO 3.8 CFR Link (CFR: 41.12/43.4 /45.10) Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities. Page 71 of 75 Printed on 1/28/2010 at 12:02

Question #: 24 Student Handout? Lower Order? Question ID; 9000016 RO ~SRO II-SRO Que *. # 24 Rev. 2 ~ Selected for Exam Odgln: New Past NRC Exam? EOfJ 2534, "Steam (;enerator Thbe Rupture," Implementation Guide (StH.;C! I of:1) Overview/Strategy I. The: primary Safety Functions challenged by a SGTR are RCS inventory control and Containment Isolation. Containment Isolation is re-established by the isolation of the affected SO and reducing RCS pressure below the SO relief valve setpoints.

2.

During a NC (."()oldown with an isolated SG, an inverted ~T (Tc greater than TH) will be observed due to a small amount of reverse heat transfer in the isolated SO. This has no effect on the NC flow in the intact SO.

3.

SGrR ~lmlplT\\l Cooldown RCS to < 515°FTH Identify, isolate, and confirm most affccted SG Depressurize RCS within 50 psi of isolated SO while maintaining: RCP NPSH limits At least 30°F subcooling Cooldown to SDC entry conditions Critical Tasks/Operator Credited Actions

1.

Thf: most affected SO is required to be isolated within 60 minutes from event initiation. ('I11e design basis assumes a SGfR occurs while operating at full power. Therefore, from time of the reactor trip).

2.

Thf: operator is assumed to lower RCS pressure to minimize the primary-secondary leak rate and prevent overfill of the affected SO.

3.

Tht~ operator cools the RCS using the intact SO and places SDC in service approximately 16 hours following the SGTR. OP 2260 Level of Use

STOP THINK ACT REVIEW Rev. 009-03 Informaticln 40 of 60 Page 72 of75 Printed on 1/28/2010 at 12:02 Question #:

24 Question ID; 9000016 RO ~SRO Student Handout? II-SRO Que *. # 24 Rev. 2 ~ Selected for Exam Odgln: New EOfJ 2534, "Steam (;enerator Thbe Rupture," Implementation Guide (StH.;C! I of:1) Overview/Strategy I. The: primary Safety Functions challenged by a SGTR are RCS inventory control and Containment Isolation. Containment Isolation is re-established by the isolation of the affected SO and reducing RCS pressure below the SO relief valve setpoints.

2.

During a NC (."()oldown with an isolated SG, an inverted ~T (Tc greater than TH) will be observed due to a small amount of reverse heat transfer in the isolated SO. This has no effect on the NC flow in the intact SO.

3.

SGrR ~lmlplT\\l Cooldown RCS to < 515°FTH Identify, isolate, and confirm most affccted SG Depressurize RCS within 50 psi of isolated SO while maintaining: RCP NPSH limits At least 30°F subcooling Cooldown to SDC entry conditions Critical Tasks/Operator Credited Actions

1.

Thf: most affected SO is required to be isolated within 60 minutes from event initiation. ('I11e design basis assumes a SGfR occurs while operating at full power. Therefore, from time of the reactor trip).

2.

Thf: operator is assumed to lower RCS pressure to minimize the primary-secondary leak rate and prevent overfill of the affected SO.

3.

Tht~ operator cools the RCS using the intact SO and places SDC in service approximately 16 hours following the SGTR. Level of Use

Informaticln STOP THINK ACT OP 2260 REVIEW Rev. 009-03 40 of 60 Lower Order?

Past NRC Exam? Page 72 of75 Printed on 1/28/2010 at 12:02

Question #: 25 Student Handout? Lower Order? Question ID; 9800061 RO ~SRO [-SRO Ques. 11 25 Rev. o ~ Selected for Exam Otfgin: 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 24D are de-energized due to a bus fault on 24D. - 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? f;;j' C-05 alarms indicating a Excess Steam Demand on the #1 SG, and C-08 alarm indicating VR-21 is de A energized. C-02/2, alarms indicating RCS temperatures are abnormally low and dropping, and both Boric Acid B Pumps are de-energized. C-05 alarms indicating both SG levels abnormally low, and only one Aux. Feedwater pump is feeding C just thl3 #2 SGt. C-01 alarms indicating CTMT Spray has actuated, and C-01 indicating only two CAR fans and one CS D pump are operating. Justification 1 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 em the "Au Main Steam header. VR-21 is deenergized based on the loss of 240. This will prevent the "S" Atmospheric Dump VaIVI~ (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). S - 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; Auxiliary 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. 0- 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. Plausible; Procedure givl~s guidance to repower and start all available ESAS components. However, this is not reguired to prevent exceeding a design limit. References I EOP 2525, Step 17 and AOP 2501, Pg. 10, Diagnostic Chart Comments and Question Modification History Sob K. D-4/C (OK - good question). Sill 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 tl) stem that all feedwater is secured to #1 SG. - RLC Bruce F. - D-3/C, Need to analyze each answer. NRC KIA System/EtA KiA Selected NRC KIA Generic System System 2.4 2.4 Emergency Procedure IPlan Emergency Procedures IPlan Page 73 of 75 Printed on 1/28/2010 at 12:02 Question #: 25 Question ID; 9800061 RO ~SRO Student Handout? Lower Order? [-SRO Ques. 11 25 Rev. o ~ Selected for Exam Otfgin: Mod Past NRC Exam? f;;j' 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 24D are de-energized due to a bus fault on 24D. - 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 B C D C-05 alarms indicating a Excess Steam Demand on the #1 SG, and C-08 alarm indicating VR-21 is de-energized. C-02/2, 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 thl3 #2 SGt. C-01 alarms indicating CTMT Spray has actuated, and C-01 indicating only two CAR fans and one CS pump are operating. Justification 1 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 em the "Au Main Steam header. VR-21 is deenergized based on the loss of 240. This will prevent the "S" Atmospheric Dump VaIVI~ (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). S - 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; Auxiliary 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. 0- 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. Plausible; Procedure givl~s guidance to repower and start all available ESAS components. However, this is not reguired to prevent exceeding a design limit. References I EOP 2525, Step 17 and AOP 2501, Pg. 10, Diagnostic Chart Comments and Question Modification History Sob K. D-4/C (OK - good question). Sill 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 tl) stem that all feedwater is secured to #1 SG. - RLC Bruce F. - D-3/C, Need to analyze each answer. NRC KIA System/EtA System 2.4 Emergency Procedure IPlan KiA Selected NRC KIA Generic System 2.4 Emergency Procedures IPlan Page 73 of 75 Printed on 1/28/2010 at 12:02

Question #: 25 Question ID: 9800061 RO li?l SRO Student Handout? 0 Lower Order? J-SRO Ques. Ii 25 Rev. 0 [ill Selected for Exam Origin: Mod Past NRC Exam? Number 2.4.45 RO 4.1 SRO 4.3 CFR Link (CFR: 41.10 143.5/45.3/45.12) Ability to prioritize and interpret the significance of each annunciator or alarm. Millstone Unit 2 Standard.Post Trip Actions INSTRUCTIONS Subsl,'quent Actions (continued)

17. IE steam generator prc&l:Iure is less than 572 psia the most affected steam generator has boiled dry, as indicated by CET temperature ril:ling, OPERATE the ADV for the least afft'Cted steam generator to stabilize CET temperature.

EOP 2525 Re\\'ision 023 Page 24 of26 CONTINGENCY ACTIONS Page 74 of 75 Printed on 1/28/2010 at 12:02 Question #: 25 Question ID: 9800061 RO li?l SRO Student Handout? 0 Lower Order? J-SRO Ques. Ii 25 Rev. 0 [ill Selected for Exam Origin: Mod Past NRC Exam? Number 2.4.45 RO 4.1 SRO 4.3 CFR Link (CFR: 41.10 143.5/45.3/45.12) Ability to prioritize and interpret the significance of each annunciator or alarm. Millstone Unit 2 EOP 2525 Re\\'ision 023 Standard.Post Trip Actions Page 24 of26 INSTRUCTIONS CONTINGENCY ACTIONS Subsl,'quent Actions (continued)

17. IE steam generator prc&l:Iure is less than 572 psia the most affected steam generator has boiled dry, as indicated by CET temperature ril:ling, OPERATE the ADV for the least afft'Cted steam generator to stabilize CET temperature.

Page 74 of 75 Printed on 1/28/2010 at 12:02

SR4)Exam QuestionsOnly(Non~~rentsn Or "Originals") Question #: 25 Question ID: 9800061 []RO ~SRO D Student Handout? D Lower Order? P-SRO Ques. # 25 Rev. 0 ~ Selected for Exam Origin: Mod D Past NRC Exam? Millstone Unit 2 AOP 2501 Re\\'ision 001-02 Diagnostic for Loss of Page 10 of 10 Electrical Power Lost Control Power (Sheet 1 of I ) Loss ofVR-ll Loss ofVR-21 Loss ofV..'\\-10 Loss ofVA-20 D1J:po~tJ[ Left side C-Ol & Right side C-Ol & "A" safety channels & "B" safety channels & lacCHin (dHIlU;U:N) CEAPDS Cort'~limic "A"RPS "B"RPS PA S}.,t~ln lost RRS Batt~ry BIL 1a.ts for 10 min. after \\"R-211o... C-02 Lddown I.ol.~. RRS battm BIU a"ailabl~ Chg Pp suction to (High TeIup.ensor failw'e) PZR..tpoint at prognm RWST

- RRS batteD' !!s21eted..

PZR Lvi control in Ch <<Y" PZR '<[point fail. to :rem - lZR I :rl tQDue] in Ch "X" - Maximum Cbarging flow Minimum Charging flow - Maximum Chlrging flow - Minimum Letdown flow ~laximum Lrtdown flow - Minimum Letdown flow C-03 Los. Alll~ZR Heakn Lo** All PZR Henters Lose All PZR Heaters Lo** A!LPZR Hrat~rs (Select iD "X" or "X+Y") (Select.n "Y" or "X~Y") (Select in "X" or "X+ Y") (Select in,.y" or "X+Y") PZR Backup Huten All four banks W1ll\\'ail.ble C--03 Bit: Charging Pumps ... RES battery drplrted ** Hle-IOOE dead CE" spray) HIe IOOF dead ("F".pray) Both.ta.1 automatic.lly Bit' Charging Pump' PZR Sp.*,,~' AUlo Control UD [Auto control wi HIC-IOOE (Dead control circuit) Botb.slll:p.""j PZR stpnt @ () ",.il.ble. [ManWlI 'Pny COD, ifP",,, control in ell. "X"] trol wi HIC-IOOF only] C-03 Loop I t,mp to RRS lost LGop 2 temp to RRS 10<' Loop I temp input to ICC, Loop 1 temp input to ICC, (RRS.ut" byp.....) (RRS auto b}'P.""') LTOP, RRS all loot LTOP, RRS aU I",t (RRS 1lU10 byp**~) (RRS aul<> bypa,...o) C-05 TIC 4165 Dead (To,'g cntrl) PI~-4223 Dead (#1 AJ)V) PIC-4216 Dead CA" CDV) PIC-4224 Dud (#2 ADV) C-05 Condf"nw-r St'fam Dumps Coud,pnw-r Stt*.m Dump'! All fo\\u: failed do.ed PIC-4216 de-flIugized (Open wi Quick Open only) [Contrl wi TIC-4165 @ COS or PlC-4~16 @FoxboroIA] RRS ba~ BiU nailable Quick Oprn OIl pmgnm

- RRS b.tteO' del1leted **

No QO to any dump volve C-05

1 AD" WIC-4223) avail.

RRS battm &'U available

1 ADV remote control lost
2 AD'- remole contr,,1 lost Manual.o;a.b: (Input pre"",e
2 AD\\' (PIC-4224) a"MI.

(C-05 & C-lI contrls dead) (C-05, -11. -10 contrl dead) frozen wbeo power lost) (Input ",*e..u.. frozen at ",,1 [Loc.I-Manual only} [Local-Manual only] ,,., ""ben powe,* "".. lo't). ... RRS hartel): deRleted....

2 AD\\' fail, dosed

[Control from CHiC 10] C-05 d A ~ SGFP Inse.1 dark, 000 '*B'" SGFP Insert dadc,. con

1 FRY Main Fail as i,
1 FRY :.\\lain Fail... io trois still 'Nork: (Indication on trob.till wad"

[Loc.I-ManWlI only} [Local-MaDu"lonly] PPC) (Indication 00 PPC) (Closes 00 power restore) (Closes OIl power..".tore) C--05 "AU SGFP :Uinimum Flow "B" SGFP lfinimtlm Flo..

1 FRY Bypass fails closed
1 FRV Bypa.. f~;1s closed Cntrt dead, \\'.I"e fail dosed Cntd dead, \\'a!"e fail closed (All control is lost)

(Ml contml i. lost) C-05 Blowdown isolates Blowdown isolates

1 Au" FRY rail. open
2 Aux FR" fails open (B..'n Rad Monitor dead)

(SJAE Rod Monitor dead) [Local-Manual only} [Local-Manual only) Level of use] STOP THINK ACT REVIEW Information Page 75 of 75 Printed on 1/28/2010 at 12:02 SR4) Exam QuestionsOnly(Non~~rentsn Or "Originals") Question #: 25 Question ID: 9800061 []RO ~SRO D Student Handout? D Lower Orde P-SRO Ques. # 25 Rev. 0 ~ Selected for Exam Origin: Mod D Past NRC Ex Millstone Unit 2 AOP 2501 Re\\'ision 001-02 Diagnostic for Loss of Page 10 of 10 Electrical Power Lost Control Power (Sheet 1 of I ) Loss ofVR-ll Loss ofVR-21 Loss ofV..'\\-10 Loss ofVA-20 D1J:po~tJ[ Left side C-Ol & Right side C-Ol & "A" safety channels & "B" safety channels & lacCHin (dHIlU;U:N) CEAPDS Cort'~limic "A"RPS "B"RPS PA S}.,t~ln lost RRS Batt~ry BIL 1a.ts for-10 min. after \\"R-211o... C-02 Lddown I.ol.~. RRS battm BIU a"ailabl~ Chg Pp suction to (High TeIup.ensor failw'e) PZR.. tpoint at prognm RWST

- RRS batteD' !!s21eted..

PZR Lvi control in Ch <<Y" PZR '<[point fail. to :rem - lZR I :rl tQDue] in Ch "X" - Maximum Cbarging flow Minimum Charging flow - Maximum Chlrging flow - Minimum Letdown flow ~laximum Lrtdown flow - Minimum Letdown flow C-03 Los. Alll~ZR Heakn Lo ** All PZR Henters Lose All PZR Heaters Lo ** A!LPZR Hrat~rs (Select iD "X" or "X+Y") (Select.n "Y" or "X ~ Y") (Select in "X" or "X + Y") (Select in,.y" or "X+Y") PZR Backup Huten All four banks W1ll\\'ail.ble C--03 Bit: Charging Pumps ... RES battery drplrted ** Hle-IOOE dead CE" spray) HIe IOOF dead ("F".pray) Both.ta.1 automatic.lly Bit' Charging Pump' PZR Sp.*,,~' AUlo Control UD- [Auto control wi HIC-IOOE (Dead control circuit) Botb.slll:p. ""j PZR stpnt @ () ",.il.ble. [ManWlI 'Pny COD, ifP",,, control in ell. "X"] trol wi HIC-IOOF only] C-03 Loop I t,mp to RRS lost LGop 2 temp to RRS 10<' Loop I temp input to ICC, Loop 1 temp input to ICC, (RRS.ut" byp..... ) (RRS auto b}'P.""') LTOP, RRS all loot LTOP, RRS aU I",t (RRS 1lU10 byp ** ~) (RRS aul<> bypa,... o) C-05 TIC 4165 Dead (To,'g cntrl) PI~-4223 Dead (#1 AJ)V) PIC-4216 Dead CA" CDV) PIC-4224 Dud (#2 ADV) C-05 Condf"nw-r St'fam Dumps Coud,pnw-r Stt*.m Dump'! All fo\\u: failed do.ed PIC-4216 de-flIugized (Open wi Quick Open only) [Contrl wi TIC-4165 @ COS or PlC-4~16 @FoxboroIA] RRS ba~ BiU nailable Quick Oprn OIl pmgnm

- RRS b.tteO' del1leted **

No QO to any dump volve C-05

1 AD" WIC-4223) avail.

RRS battm &'U available

1 ADV remote control lost
2 AD'- remole contr,,1 lost Manual.o;a.b: (Input pre"",e
2 AD\\' (PIC-4224) a"MI.

(C-05 & C-lI contrls dead) (C-05, -11. -10 contrl dead) frozen wbeo power lost) (Input ",*e.. u.. frozen at ",,1- [Loc.I-Manual only} [Local-Manual only] ,,., ""ben powe,* "".. lo't). ... RRS hartel): deRleted....

2 AD\\' fail, dosed

[Control from CHiC 10] C-05 d A ~ SGFP Inse.1 dark, 000- '*B'" SGFP Insert dadc,. con-

1 FRY Main - Fail as i,
1 FRY :.\\lain - Fail... io trois still 'Nork: (Indication on trob.till wad"

[Loc.I-ManWlI only} [Local-MaDu"lonly] PPC) (Indication 00 PPC) (Closes 00 power restore) (Closes OIl power.. ".tore) C--05 "AU SGFP :Uinimum Flow "B" SGFP lfinimtlm Flo..

1 FRY Bypass fails closed
1 FRV Bypa.. f~;1s closed Cntrt dead, \\'.I"e fail dosed Cntd dead, \\'a!"e fail closed (All control is lost)

(Ml contml i. lost) C-05 Blowdown isolates Blowdown isolates

1 Au" FRY rail. open
2 Aux FR" fails open (B..'n Rad Monitor dead)

(SJAE Rod Monitor dead) [Local-Manual only} [Local-Manual only) Level of use] Information STOP THINK ACT REVIEW r? am? Page 75 of 75 Printed on 1/28/2010 at 12:02}}

v t e Letter Sequence Other
TAC:U01799, (Open)
Initiation Acceptance... Administration Questions Answers Results Approval... Other: ML100610136, ML100670463, ML100670496, ML100670507, ML101160391
MONTHYEAR2010-01-15 [Table View]

ML100670463

Text

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