ML112030746

Draft - Written Exam (Folder 2)

ML112030746
Person / Time
Site:	Millstone
Issue date:	05/13/2011
From:	NRC Region 1
To:	Dominion Resources
JACKSON D RGN-I/DRS/OB/610-337-5306
Shared Package
ML110030662	List: IR 05000423/2011302 ML111860296 ML112030663 ML112030719 ML112030734 ML112030746 ML112030794 ML112030812 ML112030862 ML112031001 ML112490411
References
TAC U01832, 50-423/11-302 50-423/11-302
Download: ML112030746 (104)

Text

Examination Outline Cross-reference: Level RO _S~R~O~____

Question # 1 Tier # 1 Reactor Trip: Group #

Knowledge of the reasons for the actions in the EOP KIA # EPE.007 .EK3.0 1 Importance Rating 4.0 _4,,;,,:.. :. . 6______

Proposed Question:

The plant is initially at 100% power with RCS boron at 800 ppm, when the following sequence of events occurs:

1. The reactor trips.

2. Seven rods do not fully insert on the trip.

3. The crew enters ES-0.1, Reactor Trip Response.

4. ES-0.1 directs the crew to borate until RCS Boron concentration is increased by 1300 ppm.

5. The crew commences immediate boration per AOP 3566, Immediate Boration.

Why is the crew required to raise RCS boron concentration by 1300 ppm?

a) This ensures adequate SHUTDOWN MARGIN for any number of stuck rods, including all rods stuck out.

b) This is the assumed maximum boration achievable for the case where the RWST is the boration source.

c) This ensures the amount shutdown will be at least 1.3% following the outside design basis event of more than one stuck rod.

d) This ensures adequate SHUTDOWN MARGIN by borating 200 ppm for each stuck rod, minus 100 ppm for one assumed stuck rod.

Proposed Answer: A Explanation (Optional): If two or more control rods are not fully inserted (or ifDRPI is lost), ES-O.l will direct immediate boration due to shutdown margin concems. The RNO directs the crew to immediate borate by 200 ppm for each rod not fully inserted ("D" plausible), up to either a maximum of 1300 ppm ("D" wrong, since 200 ppm x 7 rods = 1400 ppm), or an RCS boron concentration of 2600 ppm is reached. The 200 ppm per rod limit, OR the 1300 ppm total boration limit ensures SHUTDOWN MARGIN is met for any number of stuck rods including the all rods out configuration plus a 10% uncertainty ("A" correct). The 2600 ppm limit on RCS boron concentration ensures SHUTDOWN MARGIN is met for the design basis stuck rod condition, ensuring the amount shutdown to be at least 1.3% following the outside design basis event of more than one stuck rod ("C" wrong, but plausible), while ensuring the required boration can be accomplished when relying on the RWST as the boration source ("B" wrong, but plausible).

Technical Reference(s): ES-O.l step (Rev. (24), step 6 (Attach if not previously provided) ES-O.l step deviation document (Rev. 24), step 6 (including version/revision number)

Proposed references to be provided to applicants during examination: None Leaming MC-05512 DISCUSS the basis of major procedure steps &/or sequence of steps in (As available)

Objective: EOP 35 ES-0.1.

Question Source: Bank # 70391 Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.1,41.5 and 41.10 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 2 Tier # 1 Pressurizer Vapor Space Accident: Group #

Ability to operate and/or monitor control of Pzr level KIA # APE.008.AAI.06 Importance Rating 3.6 3.6 Proposed Question:

The following sequence of events has occurred:

1. The "A" Pzr PORV spuriously opens, resulting in a safety injection.

2. When attempting to close the "A" PORV Block Valve, its breaker trips.

3. After exiting E-O, the crew is able to reenergize and close the "A" PORV block valve.

4. The crew terminates SI and restores letdown.

5. The crew enters FR-I.l, Response to High Pressurizer Level.

What actions will the crew take to regain control of pressurizer level?

a) Tum on Pzr Heaters to generate steam in the Pressurizer, and control charging and letdown as necessary to maintain RCS pressure steady while Pzr level comes back into the desired band.

b) Tum on Pzr Heaters to generate steam in the Pressurizer, and commence venting the reactor vessel head to collapse any vessel head voids, lowering Pzr level back into the desired band.

c) Tum off all Pzr Heaters to minimize heat input to the RCS, and commence venting the reactor vessel head to collapse any vessel head voids, lowering Pzr level back into the desired band.

d) Tum off all Pzr Heaters to minimize heat input to the RCS, and decrease charging flow to lower Pressurizer level to the desired band.

Proposed Answer: _A_ _

Explanation (Optional): At the onset of the vapor space break, RCS pressure rapidly dropped to saturation for the vessel head and hot legs. This causes two phase flow up the surge line into the Pzr, filling the pressurizer, with Pzr level remaining at 100% as long as the vapor space break is active. After the crew isolates the break and terminates 81, FR-I.I will direct the crew to tum on Pzr heaters ("C" and "D" wrong) to add heat to the Pzr water, restoring saturation conditions, and generating steam in the Pzr. "C" and "D" are plausible, since all heaters are turned off in ES-I.2 during the cooldownldepressurization to minimize heat input. "B" and "C" plausible, since FR-I.3 vents the vessel head to remove a bubble in the head. "D" is also plausible since the normal means to lower Pzr level when it is on scale is to decrease charging. "A" is correct, and "B" wrong, since the crew is directed to control charging and letdown as necessary to maintain pressure stable while Pzr level lowers while a steam bubble forms. If there is no void in the head, pressure will increase as the steam bubble forms, requiring charging to be decreased or letdown to be increased. If there is a void in the head, Pzr level will drop as the void collapses due to the pressure increase, requiring charging to be increased or letdown to be decreased.

Technical Reference(s): FR-I.l (Rev. 008-00), steps 9-12 (Attach if not previously provided) WOG Bkgd Doc for FR-I.l (Rev. 2), steps 7-10 (including versionlrevision number)

Proposed references to be provided to applicants during examination: _N.....o_n_e_ _ _ _ _ _ _ _ __

Learning (As available)

Objective: MC-04914 Outline the unique characteristics ofa Pressurizer Vapor Space LOCA.

Question Source: New Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 41.3, 41.1 0, and 41.14 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 3 Tier #

Small Break LOCA: Group #

Ability to operate/monitor CVCS KIA # EPE.009.EAl.04 Importance Rating 3.5 Proposed Question:

With the plant initially at 100% power, the following sequence of events occurs:

1. A small-break LOCA occurs.

2. The crew enters AOP 3555, Reactor Coolant Leak.

3. The crew starts a second Charging Pump.

4. The crew actuates Safety Injection and enters E-O, Reactor Trip or Safety injection.

5. RCS pressure gradually decreases to 1950 psia, and stabilizes.

6. The RO has just been directed to verify proper emergeney ECCS valve alignment and is currently monitoring the Charging System.

What is the expected positions of the VCT outlet valves (3CHS*LCVI12B and C), and the Charging Cold Leg Injection Valves (3SIH*MV880IA and B)?

3CHS*LCVl12B/C 3SIH*MV8801A1B a) OPEN OPEN b) CLOSED CLOSED c) CLOSED OPEN d) OPEN CLOSED Proposed Answer: B Explanation (Optional): When SI actuates, 3CHS*LCVI12D opens, aligning the RWST to the Charging Pump Suetion, then 3CHS*LCVII2B will close ("A" and "D" wrong) after 112D is open with SI actuated (or VCT level is less than 4%) to isolate the VCT from the Charging Pump. The cold leg injection valves will only open if RCS pressure drops less than P-19 setpoint of 1900 psia ("C" wrong, "B" correct).

Technical Reference(s): P&IDs 104D (No. 29) and I13A (No. 32)

(Attach if not previously provided) LSKs 1-6F (No.9), 26-2.2C (No.8), and 27-2A (No. 11)

(including version/revision number) £-0 Step Dev Doc (Rev. 25) for step 16 Proposed references to be provided to applicants during examination: None Learning MC-04203 For the below listed plant events, partial or complete, describe the effects on (As available)

Objective: the Chemical and Volume Control System and its interrelated systems ... Safety Injection Actuation ...

Question Source: Bank # 68372 Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 41.7 Comments:

Examination Outline Cross-reference: Level RO _S~R_O~____

Question # 4 Tier #

Large Break LOCA: Knowledge of the interrelations Group #

between pumps and a large break LOCA KIA # EPE.Oll.EK2.02 Importance Rating 2.6 _2~.7 Proposed Question:

A large, cold leg break LOCA occurs, and the RCS rapidly depressurizes to CTMT pressure.

What provides the actual motive force that drives ECCS water up into the active fuel region during the reflood phase of this accident?

a) SI Accumulator head.

b) Residual Heat Removal Pump head.

c) Charging Pump head.

d) Vessel Downeomer water elevation head.

Proposed Answer: D Explanation (Optional): Chapter 15 defines the reflood time as follows: The reflood phase of the transient is defined as the time period lasting from the end-of-refill until the reactor vessel has been filled with water to the extent that the core temperature rise has been terminated. From the later stage of blowdown and then the beginning-of-reflood, the safety injection accumulator tanks rapidly discharge borated cooling water into the RCS, contributing to the filling of the reactor vessel downcomer. The downcomer water elevation head provides the driving force required for the reflooding of the reactor core. The low head and high head safety injection pumps supply water to the RCS cold legs. Injection into the broken loop is lost out the break.

Injection into the other loops enters the vessel downcomer via the eold legs, and spills out the broken loop's cold leg ("A", "B", and "C" wrong), while maintaining the downcomer full, up to the bottom of the cold legs.

The water level in the downcomer provides a driving head due to elevation difference between its level and the level in the active fuel region of the core (liD" correct). "A" is plausible, since accumulators provide water for refill. "B" and "C" are plausible, since these pumps are injecting during a large break LOCA, and provide motive force on smaller breaks.

Technical Reference(s): Westinghouse MITCORE Text, page 16-49 (Attach ifnot previously provided) Westinghouse MlTCORE Text, Figure 36 (including version/revision number)

Proposed references to be provided to applicants during examination: ~N...;o...;n...;e.,..-_ _ _ _ _ _ _ __

Learning MC-04912 For a Large Break LOCA, EXPLAIN Core Cooling during the 4 major stages of (As Objective: the Event, DESCRIBE the symptoms of the Event, & OUTLINE the automatic Protective & available)

Operator Credited Actions required to mitigate the consequences of the Event.

Question Source: Bank # 70091 Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 41.2,41.3, and 41.8 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 5 Tier # 1 RCP Malfunctions: Ability to determine/interpret the Group #

cause of RCP failure KIA # APE.015/017.AA2.0l Importance Rating 3.5 Proposed Question:

With the plant at 100% power, the following sequence of events oceurs:

1. RCP" A" number I sealleakoff flow decreases from 2.76 gpm to 1.22 gpm.

2. The "RCP A No.2 Seal LeakoffHi" annunciator is received on MB4.

What is the eause of the above indications?

a) Loss ofRCP "A" seal injection.

b) Failure ofRCP "A" #1 seal.

c) Failure ofRCP "A" #2 seal.

d) Failure ofRCP "A" #3 seal.

Proposed Answer:

-C- -

Explanation (Optional): "A" is wrong since isolation of seal injection results in RCS flow past the thermal barrier, maintaining seal flows. "A" is plausible, since sealleakoff flow has decreased. "B" is wrong, since

1. I seal failure would result in HIGH # I sealleakoff flow. "B" is plausible, since # 1 Sealleakoff flow has changed, and would be correct if#l sealleakoffhad also gone up. "C" is correct, since if#2 seal fails, flow through #2 will increase, and # 1 sealleakoff will decrease as more flow goes through #2 seal. "D" is wrong since failure of #3 seal will not affect operation of either the # I or #2 seal. Its purpose is to limit gas release to CTMT. "D" is plausible, since a portion of#3 sealleakoffjoins #2 sealleakoff. This would be correct if

1. 1 sealleakoffhad not changed.

Technical Reference(s): OP3353.MB4B, 1-2 (Rev. 004-11)

(Attach if not previously provided) P&ID 103A (No. 24)

(including version/revision number)

Proposed referenees to be provided to applicants during examination: _N~o-:-:n...;.e~_ _ _ _ _ _ _ __

Learning MC-05434 Explain the effects of, and describe the required actions for the following (As available)

Objective: RCP seal failures: A. #1 Seal Failure B. #2 Seal Failure C. #3 Seal Failure Question Source: Bank # 75458 Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 10CFR55.41.3 and 41.5 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 6 Tier #

Loss of Rcactor Coolant Makeup: Group #

Ability to determine/interpret Charging Pump problems KIA # APE.022.AA2.02 Importance Rating 3.2

-3.7 Proposed Question:

The plant is at 100% power, and initial conditions are as follows:

• The "C" Charging (CHS) Pump is running, aligned to the "A" Train.

• The "A" Charging Pump Cooling (CCE) Pump is running.

The following sequence of events occurs:

1. A fire breaks out in the Cable Spreading Room.

2. The crew enters EOP 3509, Fire Emergency.

3. Over the next several minutes, several spurious actuations occur.

4. The "C" CHS Pump trips, and its MB3 Green and Amber lights illuminate.

5. The crew enters EOP 3506, Loss ofAll Charging.

6. The crew holds a brief, and determines the following abnormal conditions exist:

• The "A" CCE Pump has TRIPPED.

• The "B" "RHR TO CHG" Valve (3CHS*MV8468B) has spuriously CLOSED.

• A PEO reports the "C" CHS Pump Breaker white "Auxiliary Circuit" light is OFF.

7. All other equipment operates as designed.

What was the cause of the "C" CHS Pump trip?

a) The pump overheated due to the loss of the "A" CCE Pump.

b) The pump cavitated due to an isolated suction flowpath.

c) The pump breaker tripped due to a blown UC fuse on its 4KV breaker.

d) The pump breaker tripped due to a blown UT fuse on its 4KV breaker.

Proposed Answer: B Explanation (Optional): This question is based on Millstone 3 OE, where a CHS Pump was started and damaged with an RHR TO CHG Valve closed. "B" is correct, since the CHS Pump suction from the VCT is sent to the "B" Train, and must pass through both cross-connect valves 3CHS*MV8468A and B (failed closed) to reach the "A" train suction. "A" is wrong, since on a loss of running CCE pump ("A" plausible),

the standby pump will start on low pressure, and the CCE system cross-connect valves are open. "C" is wrong, since a blown UC fuse will de-energize the closing coil, but not cause the breaker to trip. "C" is plausible, since this fuse has blown, as indicated by the green and amber lights remaining lit, with the white aux circuit light no longer receiving power. "D" is wrong, since the UT fuse provides power to the MB indicating lights, which are still indicating. "D" is plausible, since a blown UT fuse will de-energize the trip coil, which would result in a CHS pump trip.

Technical Reference(s): _P_&_ID_l_0_4_A-,CN,--0_._52-')'--_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

(Attach if not previously provided) _P_&_ID_I_O_5_A-,(.....N_o_._2-'3)'--_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

(including version/revision number) _E_S_K_5_A---,>(_N_o_._13~)'__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

Proposed references to be provided to applicants during examination: -=-N.;.:o~n:.:.e_ _ _ _ _ _ _ _ __

Learning MC-04202 Describe the operation of the Chemical and Volume Control System (As available)

Objective: under normal, abnormal, and emergency operating conditions.

Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 41.3,41.7, and 41.8 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 7 Tier #

Loss of RHR Systcm: Group #

Knowledge of the operational implications of a KJA# APE.025.AKl.Ol loss ofRHRS during a1l modes of operation Importance Rating 3.9 _4';';'.:;..3_ __

Proposed Question:

The plant is being cooled down in preparation for refueling after a 400 day run, and current conditions are as follows:

• RCS temperature is 175°F

• RCS pressure is 305 PSIA

• The Pressurizer is solid.

• Charging flow control is in manual

• RHR is in the cooldown mode, with the "A" Train in service.

A tube in the "A" RHR heat exchanger suddenly fails, causing a 50-gpm tube leak.

Assuming no opcrator action is taken, what will be the result of the tube leak?

a) RPCCW surge tank level will decrease until the RPCCW Surge Tank Makeup Valve opens. This will maintain surge tank level.

b) RPCCW surge tank level will decrease until the "A" RPCCW pump trips. This will result in a loss of shutdown cooling.

c) Actual RHR Pump flow will increase and Pressurizer pressure will decrease. RCS temperature will begin to increase.

d) Actual RHR Pump flow will increase and Pressurizer pressure will decrease. RCS temperature will begin to decrease.

Proposed Answer: _C-"--__

Explanation (Optional): RHR pressure is higher than CCP pressure; so 50 gpm is flowing from the RHR system into the RPCCW system ("A" and "B" wrong). "A" and "B" are plausible, since there is a leak in the RPCCW System. With Charging flow control valve 3CHS"'FCV121 in manual, 50 gpm is being lost from the RCS, and pressure will decrease in the pressurizer. RHR pump flow (actual) will increase, since pressure will drop at its discharge with the tube leak. Flow transmitter 3RHS*FT618 will see less flow and send a signal to RHS*FCV618 to throttle open to maintain 4000 gpm total flow. Now a greater percentage of the RHR flow retuming to the RCS is bypassing the RHS heat exchanger, so RCS temperature begins to increase

("C" correct, liD" wrong). "D" is plausible, since total flow through the RHR pump has increased.

Technical Reference(s): OP 3208 (Rev. 021**06), steps 4.3.10 and 4.3.11 (Attach ifnot previously provided) OP 3310A (Rev. 017-03), 4.5 (including version/revision number) P&ID 112A (No. 049)

Proposed references to be provided to applicants during examination:

Learning MC-05459 Given a failure, partial or complete, of the residual heat removal system (As available)

Objective: detennine the effects on the system and on interrelated systems.

Question Source: Bank # 78921 Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 41.8 and 41.10 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 8 Tier #

Loss of Component Cooling Water: Ability to perform Group #

actions without reference to procedures requiring KJA# APE.026.GEN .2.4.49 immediate operation of system components and controls Importance Rating 4.6 _4_.4_ _ __

Proposed Question:

With the plant at 100% power, thc following sequence of events occurs:

I. The "A" RPCCW Pump trips.

2. The crew enters AOP 3561, Loss o/Reactor Plant Component Cooling Water.

3. Prior to opening the procedure, the US directs the RO to isolate charging and letdown in order to stabilize the plant.

What actions will the RO take to complete this task?

a) Simultaneously CLOSE Charging Header Flow Control Valve 3CHS*FCVI21 and the in-service Letdown Orifice Isolation Valve 3CHS* AV8149B or C. Then CLOSE Charging Isolation Valve 3CHS*MV8106.

b) Simultaneously CLOSE Charging Header Flow Control Valve 3CHS*FCV121 and the in-service Letdown Orifice Isolation Valve 3CHS*AV8149B or C. Then CLOSE Letdown Isolation Valves 3RCS*LCV459 and 460.

c) CLOSE the in-service Letdown Orifice Isolation Valve3CHS* A V8149B or C. Then CLOSE Charging Header Flow Control Valve 3CHS*FCVI21. Then CLOSE Charging Isolation Valve 3CHS*MV8106.

d) CLOSE the in-service Letdown Orifice Isolation Valve3CHS* AV8149B or C. Then CLOSE Charging Header Flow Control Valve 3CHS*FCV121. Then CLOSE Letdown Isolation Valves 3RCS*LCV459 and 460.

Proposed Answer:

-A- -

Explanation (Optional): Loss ofletdown cooling will cause the VCT to heatup, rapidly reaching foldout page RCP Trip Criterion, since VCT water is sent to the RCP seals. "A" is correct, since the RO will simultaneously CLOSE Charging Header Flow Control Valve 3CHS*FCV121 and the in-service Letdown Orifice Isolation Valve3CHS *AV814 7B or C, to minimize thermal stresses due to loss of Regen Heat Exchanger heating/cooling ("C" and "D" wrong). "C" and "D" are plausible, since this would remove the heat source for letdown prior to removing the cooling source to the Regen Heat Exchanger. Then Charging Isolation Valve 3CHS*MV8106 is CLOSED in order to isolate the charging path (FCV121 is a fail-open valve) ("B" wrong). "B" is plausible, since the Letdown Isolation Valves are in the letdown path, which is being isolated.

Technical Reference(s): AOP 3561 (Rev. 011-02), Attachment A, step 1 (Attach if not previously provided) AOP 3561 basis doc (Rev. 010), for Attachment A, step 1 (including version/revision number) Ops Department Skill of the Trade Policy (09/12/07)

Proposed references to be provided to applicants during examination: -=-N:..:o.:.;n:;:.e_ _ _ _ _ _ _ _ __

Learning (As available)

Objective: MC-03933 Describe the major action categories contained within AOP 3561.

Question Source: New Question History:

Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 41.41.3,41.4, and41.10 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 9 Tier #

ATWS: Knowledge oflocal auxiliary operator tasks and Group #

the resultant operational effects K/ A # EPE.029.GEN.2.4.35 Importance Rating 3.8 _4_._0_ __

Proposed Question:

4160Y busses 34C and 34A deenergize, and the following sequence of events occurs.

1. The crew enters FR-S.l Response to Nuclear Power GenerationlATWS.

2. The US directs the RO to immediate borate the RCS per FR-S.I, steps 4-6.

3. The RO reports that Charging Flow Control Valve 3CHS*FCY121 has spuriously closed, and will not open.

4. The RO aligns the safety grade boration path by opening Charging Header Flow Control Valve 3CHS*HCY190B.

5. A PEO locally aligns RCP seal supply through "B" Charging Pump Seal Supply Bypass Valve 3CHS*Y270.

6. The RO opens Cold Leg Injection Valve 3SIH*MY8801B.

7. The RO closes Charging Header Isolation Valve 3CHS*MY8438B to provide a throttlable horation flowpath.

Why does FR-S.l require a PEO to locally align the RCP seal supply through 3CHS*Y270 while aligning Charging through 3CHS*HCYI90B?

a) This ensures adequate seal injection and immediate boration flow, since CHS*HCY190B is undersized to supply both paths.

b) This prevents completely isolating the RCP Seal supply path from the "Bft CHS pump.

c) This protects the "B" CHS Pump from damage by ensuring net positive suction head (NPSH) is maintained to the pump.

d) This protects against RCP Seal damage by preventing the Seal Injection Filter from being bypassed.

Proposed Answer: --=...:A:....,..,-...,.

Explanation (Optional): All charging flow will be aligned through HCY190B to prevent a problem with excess RCS inventory ifFCY121 had failed open. "A" is correct since a separate seal supply path through 3CHS*Y270 is aligned while charging through 3CHS*HCY190B since it is undersized to supply both the charging and seal injection paths "B" is wrong since flow will not be isolated to the seals when MY8438B is closed, but plausible, since if the "B" train had lost power, all seal injection flow through the normal path would be isolated. "C" is wrong, since flow capability will be reduced when charging through HCY190B, but plausible, since this is the basis for limiting charging flow while using the gravity borate path. "D" is wrong, since the seal injection filter is not bypassed in this alignment, but plausible, since an alternate charging and seal injection path are aligned.

Technical Reference(s): -=.,;FR:.:-.. ::S:..:. .I:.. 1.:{R.:.;e:..;,v.:. . .O;,. ;I:.,:9,L),'-=S:,:te.::.&p:.:s.. ,;4.. ,;-6==--_ _ _ _ _ _ _ _ _ _ _ _ _ _ __

(Attach ifnot previously provided) FR-S.l (Rev. 019) Step Dev Document, steps 4-6 (including version/revision number) AOP 3566 (Rev. 10) Basis Document, steps 1-3 P&ID 104A (No. 52)

Proposed references to be provided to applicants during examination: ."..::..N:..:o:.::::n:.::.e.,-_ _ _ _ _ _ _ __

Learning MC-04626 Discuss the basis of major procedure steps and/or sequence of steps in (As available)

Objective: EOP 35 FR-S.I Question Source: Bank # 80923 Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 41.5,41.8, and 41.1 0 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 10 Tier #

Steam Gen. Tube Rupture: Group #

Ability to determine operability and/or availability of KJA# EPE.038.GEN.2.2.37 safety related equipment Importance Rating 3.6 _4.,;..;..,;. ;6_ __

Proposed Question:

The following sequence of events occurs:

1. A Steam Generator Tube Rupture occurs on the "B" SG.

2. Offsite power is lost on the trip.

3. The crew enters E-3, Steam Generator Tube Rupture.

4. The crew prepares to depressurize the RCS to minimize break flow and refill the pressurizer.

By what method is the crew required to depressurize the RCS, and why?

a) Depressurize the RCS using the Normal Pzr Spray Valves, to avoid adverse containment conditions.

b) Depressurize the RCS using the Normal Pzr Spray Valves, to minimize the loss of reactor coolant.

c) Depressurize the RCS using the Auxiliary Spray Valve, to provide a controlled depressurization rate.

d) Depressurize the RCS using a Pzr PORV, to minimize thermal stresses on the Pzr spray nozzle.

Proposed Answer:

-D- -

Explanation (Optional): On a loss of offsite power, the RCPs have lost power, so there is no motive force for normal spray ("A" and "B" wrong). Instrument air has also been lost, but will be restored. "An and "B" are plausible, since normal Pzr spray is preferred if it is available, for the reasons given in "A", "B", and "D".

Using a PORV can lead to adverse Ctmt conditions, and loses RCS inventory. "D" is correct, and "C" wrong, since at this point in E-3, letdown has not yet been restored to provide aux spray pre-heating, so thermal shock of the Aux Spray nozzle will occur if spray is initiated. A Pzr PORV is used instead of aux spray to minimize thermal stresses on the spray nozzle. "C" is plausible since aux spray is preferred in ES-3.1, since by that time, letdown should be in service, and using Aux Spray does provide better control ofpressure and thanaPORV.

Technical Reference(s): E-3 (Rev. 022), step 16 (Attach if not previously provided) WOG Bkgd (Rev. 2) for E-3, step 16 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning MC-07441 Given a set of plant conditions, determine the required actions to be (As available)

Objective: taken per E-3.

Question Source: Bank # 75598 Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 41.7 and 41.10 Comments:

RO Examination Outline Cross-reference: Level

-SRO -----

Question # II Tier #

Loss of Main Feedwater: Group #

Knowledge of the operational implications of the effects KIA # APE.054.AKI.02 of feed introduction on a dry SG Importance Rating 3.6 _4..;;.:.=-2_ __

Proposed Question:

With the plant initially at 100% power and the TDAFW pump tagged out, the following sequence of events occurs:

I. The "A" Feed Reg Valve fails closed.

2. The reactor trips on Lo-Lo level in the "A" SG.

3. Both MDAFW pumps auto-start.

4. The BOP reports AFW flow can NOT be established to the "A" SG, since MDAFW flow control valve to the "A" SG (3FWA *HIC31Al) has failed closed.

5. The crew completes ES-O.l, Reactor Trip Response and transitions to FR-H.5, Response to Steam Generator Low Level.

Maintenance reports that the MDAFW flow control valve to the "A" SG has been repaired, and current conditions are as follows:

• RCS Tave: 557°F

• Total AFW flow: 600 gpm

• "A" SG Wide Range level: 3%

In accordance with the WOG EOP Basis Document, what is the most significant operational concern if the crew reestablishes AFW flow to the "A" SG?

a) This will extend the mass and energy release to the atmosphere.

b) This will result in thermal or mechanical shocks to the SG tubes that could result in a tube rupture.

c) This will reinitiate the cooldown, which could result in a Shutdown Margin concern.

d) This will reinitiate the cool down, which could result in a pressurized thermal shock (PTS) event.

Proposed Answer:

-B- -

Explanation (Optional): Feeding a hot (>550°F), dry (WR level <12%) will create significant thermal stresses on SG components. Reestablishing feed flow to a dry SG will result in thermal/mechanical shock to SG tubes, increasing the risk of either a tube leak or a tube rupture ("B" correct, "A", "e", and "D" wrong).

Therefore, FR-H.5 directs the crew to request the ADTS to evaluate refilling the affected SG as part oflong term recovery actions, and transitions out ofFR-H.5, skipping the step to restore AFW flow "A" is plausible since this is a basis for isolating feedwater to a faulted SG. "C" and "D" are plausible since increasing feed water flow increases the cooldown rate, and cooling down adds thermal stress to the RCS and also adds positive reactivity.

Technical Reference(s): WOG Bkgd Doc for FR-H.5 (Rev. 2), step 4 (Attach if not previously provided) WOG Bkgd Doc for FR-H.l (Rev. 2), step 1, second Caution (including version/revision number)

Proposed references to be provided to applicants during examination: None

~--~------------------

Learning MC-05979 Discuss the basis of major procedure steps and/or sequence of steps in (As available)

Objective: EOP FR-H.5.

Question Source: Bank # 80862 Question History: Millstone 3 2007 NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.8 and 41.10 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 12 Tier #

Station Blackout: Group #

Knowledge of the reasons for the actions contained in KIA # EPE.055.EK3.02 the EOP for station blackout Importance Rating 4.3 _4~.6~_ __

Proposed Question:

The plant has experienced a loss of all AC power, and the operators are carrying out the actions ofECA-O.O, Loss ofAll AC Power.

An operator is dispatched to locally close RCP Seal Water Return Containment Isolation Valve 3CHS*MV8100.

Why is the crew directed to close this valve?

a) Prevent thermal shock ofRCP seals.

b) Prevent the formation of steam in the RPCCW System.

c) Prevent over-pressurizing the PRT, potentially releasing radioactivity in containment.

d) Prevent overflowing the VCT, potentially releasing radioactivity in the Aux Building.

Proposed Answer: -:;..D_ _

Explanation (Optional): Isolating at the RCP seal return line prevents seal leakage from filling the volume control tank (via VCT relief valve) with potential for radioactive release within the auxiliary building (liD" correct). Such a release, without auxiliary building ventilation available, could limit personnel access for local operations. "An is wrong, but plausible since this is the reason for isolating seal injection during a loss of all AC. "B" is wrong, but plausible since this is the reason for isolating RPCCW to CTMT during a loss of all AC. "c" is wrong, but plausible since the PRT also receives radioactive vents, and has a rupture disk that relieves to CTMT.

Technical Reference(s): WOG Bkgd Doc for ECA-O.O (Rev. 2), step 8 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: ~N.:..:o:..:;n:.;;.e_ _ _ _ _ _ _ _ __

Learning ~C-?3852 Discuss the basis of major procedure steps and/or sequence of steps (As available)

Objective: WIthin EOP 35 ECA-O.O.

Question Source: Bank # 67592 Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.5 and 41.10 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 13 Tier #

Loss of Vital AC Elec. lnst. Bus: Group #

Ability to operate andlor monitor manual inverter KJA# APE.057.AAl.OI swapping during loss of vital AC Bus Importance Rating 3.7 3.7 Proposed Question:

With the plant initially at 100% power, the following sequence of events occurs:

1. VIAC 1 deenergizes.

2. The crew enters AOP 3564, Loss ofOne Protective System Channel.

3. The crew is successful at restoring power to VIAC 1 from MCC 32-2R via the static switch.

4. A PEO is dispatched to transfer VIAC-l to Invcrter 1 in accordance with OP 3345B, i20 Volt Vital instrument AC.

After VIAC 1 has been transferred to Inverter 1, what will be the position ofthe Bypass Line to UPS Breaker and the Bypass Switch?

BYQass Line to UPS BYQass Switch Breaker a) "ON" NORMAL OPERATION" b) "OFF" "BYPASS TO LOAD" c) "ON If "BYPASS TO LOAD" d) "OFF" "NORMAL OPERATION" Proposed Answer: A Explanation (Optional): "A" is correct, since the Inverter 1 "BYPASS LINE TO UPS" breaker will be placed in "ON" to allow synchronization ("B" and "D" wrong). Then the Inverter 1 Bypass Switch will be placed in "NORMAL OPERATION" to allow the inverter output to supply VIAC 1 through the bypass switch ("C" wrong). Then the 3VBA*INV-I "INVERTER TO LOAD" pushbutton will be depressed, to select the inverter, rather than the alternate source, to supply the static switch. "B", "C", and "D" are plausible, since these are the actual positions on the breaker and switch.

Technical Reference(s): OP 3345B (Rev. 11), section 4.5 (Attach if not previously provided) EE-IBA (No. 29)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning MC-05009 Describe the operation of 120 VAC Distribution System Controls and (As available)

Objective: Interlocks: A. Static Transfer Switch Operation. B. Bypass Line Regulator. C.

Manual Bypass Switch. D. Inverter Indication and Control.

Question Source: New Question History:

Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 41.8 and 41.10 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 14 Tier #

Loss of DC Power: Group #

Knowledge of the operational implications of battery KIA # APE.058.AK1.01 charger equipment/instrumentation Importance Rating 2.8 3.1 Proposed Question:

With the plant at 100% power and Battery Charger 3 supplying DC Bus 3 (301A-2), the following sequence of events occurs:

1. DC Bus 3 (301 A-2) loses power.

2. The crew enters AOP 3563, Loss ofDC Bus.

3. A PEO is dispatched to DC Bus 3.

4. The PEO reports Battery 3 output breaker is open.

5. The PEO reports Battery Charger 3 output breaker is open.

The crew considers energizing DC Bus 3 from swing Batte:ry Charger 7.

Which ofthe following components could be damaged if the Battery Charger is placed in service on the deenergized DC Bus?

a) The Battery Bank b) The 120V Inverter c) The Charger Rectifier Stack d) The DC Bus Loads Proposed Answer: --=C_ _

Explanation (Optional): HC" is correct, and "A", "B", and "D" wrong, since placing a battery charger on a deenergized bus will draw excessive current on the charger, potentially damaging the rectifier stack. "A",

"B", and "D" are plausible, since these components are all normally tied to the DC Bus.

Technical Reference(s): OP3345C (Rev. 016-04), Precaution 3.6 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None

~~--------------------

Learning MC-05014 Describe the major administrative or procedural precautions and (As available)

Objective: limitations associated with the 125 VDC Distribution System, including the basis for each, identified within OP 3345C.

Question Source: Bank # 68020 Question History: Millstone 3 2004 NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 41.8 and 41.10 Comments:

Examination Outline Cross-reference: Level RO _S~R~O~____

Question # 15 Tier #

Loss of Nuclear Service Water: Group #

Ability to determine/interpret valve lineup to restart KJA# APE.062.AA2.03 SWS with portion of system bypassed Importance Rating 2.6 ...;2;;.;..9::...-_____

Proposed Question:

Initial Conditions:

• The plant is in MODE 5.

• An "A" Electrical Train Outage is in progress.

• The "B" Service Water Pump is running.

The following sequence of events occurs:

1. The RPCCW HX SW FLOW HI/LO annunciator is received on MB 1C.

2. A PEO is dispatched, and reports a SWP pipe break just downstream of the HB" Train Service Water to RPCCW Supply Valve 3SWP*MOY50B.

3. The HB" Service Water Pump trips.

4. The RO isolates the break by closing 3SWP*MOY50B.

5. The crew is preparing to start the "D" SWP Pump.

What combination of Service Water System loads should be aligned with 3SWP*MOV50B isolated to provide SWP minimum flow requirements, and prevent exceeding maximum SWP flow limits?

a) The "c" RPCCW Heat Exchanger.

b) One (1) TPCCW Heat Exchanger.

c) The "C" RPCCW Heat Exchanger and one (1) TPCCW Heat Exchanger.

d) Two (2) TPCCW Heat Exchangers.

Proposed Answer:

-D- -

Explanation (Optional): The minimum flow requirement for one SWP train is two TPCCW HX's ("D" correct and "B" wrong), or one RPCCW HX. The maximum flow allowed is one RPCCW HX and one TPCCW HX. "An and "C" are wrong because closing 3SWP*MOY50B isolates the "B" AND the "C" RPCCW HX. "A" and "c" are plausible, since these would provide adequate flow, if available. "B" is plausible, since TPCCW provides a flowpath for SWP with 3SWP*MOV50B closed, and there is a concern with excessive flow as well as minimum flow.

Technical Reference(s): OP 3326 (Rev. 023-06), Precautions 3.9 and 3.10 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: -.::..N;.;;o;;;.:n:.:.e_ _ _ _ _ _ _ _ __

Learning MC-05716 Describe the major administrative or procedural precautions and limitations (As Objective: placed on the operation of the Service Water System, and the basis for each. available)

Question Source: New Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.10 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 16 Tier#

Loss ofInstrument Air: Group #

Ability to operate andlor monitor remote manual loaders KIA # APE.065.AAl.OI Importance Rating 2.7 2.5 Proposed Question:

With the plant initially at 100% power, the following sequence of events occurs:

1. The "An Instrument Air Compressor trips.

2. The crew enters AOP 3562, Loss ofInstrument Air.

3. A PEO is dispatched to the instrument air compressors.

In accordance with AOP 3562, what action will the PEO be direeted to take?

a) The PEO will place the "B" instrument air compressor CONTROL SWITCH to "CS".

b) The PEO will place the "B" instrument air compressor CONTROL SWITCH to "AUTO".

c) The PEO will place the "B" instrument air compressor LOAD TRANSFER SWITCH in "I".

d) The PEO will place the "B" instrument air compressor LOAD TRANSFER SWITCH in "2".

Proposed Answer:

-A- -

Explanation (Optional): Control switehes for the lAS and SAS compressors have three positions. "A" is correct, and "B", "C", and "0" wrong, since AOP 3562 places the switch in CS to ensure the running compressor is continuously loaded. AUTO allows the compressor to automatically start at a set pressure ("B" plausible). AUTOMATIC operation is used for a compressor in a standby mode. "e" and "0" are plausible, since Position 1 and Position 2 are positions used to admit air to "unloaders" that prevent one half or the other half of the inlet valves to cylinder halves from closing.

Technical Reference(s): AOP 3562 (Rev. 007-01), step 2.a.RNO (Attach if not previously provided) LSK 12-1F (No. 10)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning MC-05323 Describe the operation ofthe plant air systems under the following ... (As available)

Objective: Low instrument air pressure ...

Question Source: New Question History:

Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 41.4 and 41.10 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 17 Tier #

Loss of Secondary Heat Sink: Group #

Knowledge of ops implication of components, capacity, KIA # EPE.W/E05.EK1.1 function of emergency systems Importance Rating 3.8 4.1 Proposed Question:

A loss of secondary heat sink has occurred.

Bleed and Feed has been initiated using only one PORV.

Compared to using both PORVs, how effective is Bleed and Feed cooling?

a) Bleed and Feed cooling is LESS effective, since less RCS depressurization allows less subcooled S1 flow.

b) Bleed and Feed cooling effectiveness is NOT AFFECTED, since the RCS depressurizes to saturation in either case.

c) Bleed and Feed cooling effectiveness is NOT AFFECTED, since the RCS pressurizes to the PORV setpoint in either case.

d} Bleed and Feed cooling is MORE effective, since less RCS mass is lost through a single PORV.

Proposed Answer: _A_ _

Explanation (Optional): "A" is correct, and "B", "C", and "D" wrong, since one open PZR PORV will lower RCS pressure at a slower rate than with two PORVs, during which time the RCS will continue to heat up. The PORVs will lower pressure only to saturation in the RCS, so pressure will not lower as far with the one PORV. So less injection flow will occur, and may not be sufficient to maintain adequate RCS feed flow. "B" and "c" are plausible, since the feed source is the same with one or two PORVs, and the basis information relates to feed and bleed versus bleed and feed cooling. "D" is plausible, since mass is lost through PORVs, and this is a basis for using aux spray rather than PORV s to depressurize in the EOPs.

Technical Reference(s): FR-H.I (Rev. 020-01), steps 13 and 14 (Attach if not previously provided) WOG Bkgd Doc for FR-H.1 (Rev. 2), step 16 (including version/revision number)

Proposed references to be provided to applicants during examination: ..:;,.N;..:o;,:;n;,:;e_ _ _ _ _ _ _ _ __

Learning MC-04941 APPRAISE each Operator-initiated recovery technique in its ability to (As available)

Objective: restore the Heat Sink Critical Safety Function.

Question Source: Bank # 70058 Question History: Millstone 2 2000 NRC Exam Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 41.5,41.7 and 41.8 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 18 Tier #

Generator Voltage and Electric Grid Disturbances: Group #

Knowledge of the operational implications of definition KJ A # APE.077.AKl.OI of volts, watts, amps, V ARs, power factor Importance Rating 3.3 3.5 Proposed Question:

Initial Main Generator/345KV Switchyard Conditions:

• Real Load: 1280 MWe

• Reactive Load: 350 MVARS out

• Switchyard Voltage: 345KV

• Frequency: 60.0 Hz The grid becomes unstable, and the BOP reports the following parameters:

Switchyard Voltage has dropped to 330 KV.

Frequency has remained at 60.0 Hz.

Assuming the reactor does NOT trip, how will Main Generator Amps respond to this event; and which limit (MWe or MV AR) will most likely be exceeded?

Generator Amps Limit most likely to be exceeded a) Increase MVARs b) Decrease MVARs c) Increase MWe d) Decrease MWe Proposed Answer: A Explanation (Optional): Since Frequency has not changed, turbine speed remains constant, since it is tied to the grid. The turbine control valves will remain steady, maintaining real load constant (HC" and "0" wrong).

A generator's MVAR load increases when generator terminal voltage increases above grid voltage. This can be caused either by raising excitation voltage, or by decreasing grid voltage, which has happened in the transient described in the stem ofthe question. "A" is correct, and "B" wrong, since raising MVARs increases generator amps. "B", "C", and "0" are plausible, since a transient is in progress.

Technical Reference(s): General Physics Motors and Generators Text (Rev. 2), Figure 5-44 (Attach ifnot previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: -.:-N;.;;o..;;;n~e_ _ _ _ _ _ _ _ __

Learning MC-0331? Given a failure of the 345KV Distribution System, or a portion ofthe (As available)

Objective: system, determine the effects on the System and on the interrelated systems ...

Question Source: New Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 41.4 and 41.5 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 19 Tier #

Inoperable/Stuck Control Rod Group # 2 Knowledge of operational implication of Flux tilt KJA# APE.005.AKl.02 Importance Rating 3.1 3.9 Proposed Question:

Reactor power is 80%, with a load decrease in progress per OP 3204, At Power Operation, when the following sequence of events occurs:

1. An NIS UPPER DET FLUX DEVIATIONIAUTO DEFEAT annunciator is received on MB4.

2. The RO reports one Control Bank D rod has not been inserting with the rest of the bank, and is misaligned by 14 steps.

3. The crew enters AOP 3552, Malfunction o/the Rod Drive System.

4. The US directs the RO to check Quadrant Power Tilt Ratio on the Plant Process Computer.

Does LCO 3.2.4 "Quadrant Power Tilt Ratio" apply at this power level; and based on flux tilt concerns, how long does the crew have to realign the rod before AOP 3552 will place additional restrictions on the recovery of the rod?

a) LCO 3.2.4 does NOT apply, since power is below 85%. The crew has ONE hour to realign the rod.

b) LCO 3.2.4 does NOT apply, since power is below 85%. The crew has FOUR hours to realign the rod.

c) LCO 3.2.4 DOES apply, since power is above 50%. The crew has ONE hour to realign the rod.

d) LCO 3.2.4 DOES apply, since power is above 50%. The crew has FOUR hours to realign the rod.

Proposed Answer: --=.C_ _

Explanation (Optional): The QPTR LCO does not apply below 50% (HA" and HB" wrong). "A" and "B" are plausible, since 85% is the NIS Hi Flux setpoint that will be selected rod is misaligned fore greater than an hour. "C" is correct, and "D" wrong, since AOP 3552 has the crew reduce power to less than 75% prior to rod recovery if the rod is misaligned for greater than an hour. This is based on industry OE (Arkansas Nuclear One, 1983), where fuel was damaged recovered due to flux peaking when a rod was realigned that had been misaligned for an extended period of time. "D" is plausible, since four hours is the ACTION time to reduce the NIS Hi Flux setpoint with an inoperable rod.

Technical Reference(s): LCO 3.2.4 (Amendment 60), Quadrant Power Tilt Ratio, and its Applicability (Attach if not previously provided) AOP 3552 (Rev. 010-01), Attachment A, step 4 (including version/revision number)

Proposed references to be provided to applicants during examination: None

--~~~----------------

Learning MC-04889 DESCRIBE the major parameter changes associated with Reactivity & (As available)

Objective: Power Distribution Anomalies Question Source: New Question History:

Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 41.6 and 43.2 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 20 Tier # 1 Emergency Boration: Group # 2 2 Knowledge of the interrelations between emergency KJA# APE.024.AK2.04 boration and pumps Importance Rating 2.6 2.5 Proposed Question:

With the plant at 100% power, the following sequence of events occurs:

1. The crew enters AOP 3566, Immediate Boration.

2. The crew aligns for gravity boration.

What is the maximum allowed net charging flow while aligned for gravity boration?

a) 33 gpm b) 53 gpm c) 75 gpm d) 100 gpm Proposed Answer: ......;;;C_ _

Explanation (Optional): With no Boric Acid Transfer Pump running, the crew must limit net charging flow to the RCS to LESS THAN 75 gpm (charging + seal injection - RCP seal return) ("C" correct, "A", "B", and "D" wrong). "A" is plausible, since this is the minimum required boration flow that ensures adequate boration is occurring. "B" is plausible, since this is the approximate minimum flow if seal injection is included. "D" is plausible, since this is the flow required if the suction source is the RWST.

Technical Reference(s): AOP 3566 (Rev. 010-01), step l.b (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: .....;;..N-'o-'n-'e_ _ _ _ _ _ _ _ __

Learning MC-03961 Describe the major action categories within AOP 3566, Immediate (As available)

Objective: Boration.

Question Source: Bank # 65130 Question History:

Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 41. 8 and 41.1 0 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 21 Tier #

Loss ofIntermediate Range NI: Group # 2 2 Knowledge of operational implications of EOP K/A# APE.033.GEN.2.4.20 warnings, cautions, and notes Importance Rating 3.8 4.3 Proposed Question:

Initial Conditions:

1. A plant startup is in progress per OP 3203, Plant Startup.

2. Reactor power is 8%.

The IR HI FLUX ROO STOP annunciator is received on MB4C.

What failure may have caused this annunciator to come in, and what operational implications exist?

a) P-IO has cleared sooner than it should have. The crew will not be able to block the Intermediate Range Hi Flux Rod Stop when directed in OP 3203.

b) P-IO has cleared sooner than it should have. The crew will not be able to block the Intermediate Range Hi Flux Reactor Trip when directed in OP 3203.

c) An Intermediate Range NIS Channel is slowly failing high. The reactor will trip if the failed channel reaches the current equivalent of 25% power.

d) An Intermediate Range NIS Channel is slowly failing high. The crew will need to block the Intermediate Range Hi Flux Rod Stop before they can proceed with raising reactor power.

Proposed Answer: _C_ _

Explanation (Optional): In the note box in Attachment E of AOP 3571, Instrument Failure Response, this annunciator is listed as a symptom of an IR Channel Failure. This annunciator is not received when P-l 0 clears ("A" and "B" wrong). "C" is correct, and "0" wrong, since below P-l 0, the IR Rod Stop and IR Hi Flux Trip cannot be blocked. "0" is plausible, since thc crew would normally block the Rod Stop when P-I 0 clears. "A" and "B" are plausible, since P-lO clears at 10% power, and it feeds into the block circuits for the IR Hi Flux Rod Stop, IR Hi Flux Trip, and the PR Hi Flux Lo Setpoint Reactor Trip.

Technical Reference(s): AOP 3571 (Rev. 009-07), Attachment E, Note Box, top of page 1 of3 (Attach if not previously provided) _O_P_3_35_3_.M_B_4_C-,(~R_ev_._0_0_6-_0_7),-,-,_6_-7_ _ _ _ _ _ _ _ _ _ _ _ _ __

(including version/revision number) Functional Owg 3 (No. G) and 4 (No. G)

Proposed references to be provided to applicants during examination: --=..N;..;;o;..;;n;.:.e_ _ _ _ _ _ _ _ __

Learning MC-05225 Describe the operation of the Nuclear Instrumentation System Control (As available)

Objective: and Interlocks ... Reactor Trip Signals ... Protection Signals ...

Question Source: New Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 41.7 and 41.10 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 22 Ticr# 1 I Fuel Handling Accident: Group # 2 2 Ability to operate andlor monitor ARM system KJA# APE.036.AA 1.02 Importance Rating 3.1 3.5 Proposed Question:

The unit is in MODE 0, with fuel assemblies being moved to different locations in the fuel pool, when the following sequence of events occurs:

1. A RADIATION HI annunciator is received on MB2.

2. The RO reports 3HVR-REI7-1 (Fuel Building Exhaust) is in ALARM.

3. The RO reports Area Radiation Monitor trends as follows:

• 3RMS-RE08-\ (Spent Fuel Pool Bridge Hoist Area) shows slightly increasing radiation levels.

• 3RMS-RE36-J (Fuel Pool Area) shows slightly increasing radiation levels.

What is the most likely cause of the increasing radiation levels?

a) Spent Fuel Pool level is decreasing due to a Refueling Cavity Seal leak.

b) Spent Fuel Pool level is decreasing due to a Fuel Pool Cooling System leak.

c) Fuel Assemblies are being moved closer to the radiation monitors.

d) One of the Fuel Assemblies has been damaged during fuel movement.

Proposed Answer: ~D_ _

Explanation (Optional): "A" and "B" are wrong, since a loss oflevel would show no increase seen on the ventilation monitor, since it is in the Auxiliary Building, and there is no release of radioactive material. "A" and "B" are plausible, since loss of refueling cavity/fuel pool level will reduce area shielding, causing an increase in area monitor radiation indication. "C" is wrong, since Area Monitors RE08 and RE36 are not next to each other, and they are both increasing. "C" is plausible, since moving fuel in the reactor vessel affects NIS detectors. "D" is correct, since both ARMs and the PRM are showing increases.

Technical Reference(s): AOP 3573 (Rev. 018-01), Att A, pg 9; Att B, pg 2 and 5 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None

~~~------------------

Learning MC-00165 Describe the function and location of the following Radiation Monitors (As available)

Objective: (including Local and Remote Indicating Control Units ... RMS-RE-36/08 ...

Question Source: Modified Bank # 75625 Original Question Attached Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 41. 7 and 41.11 Comments:

Original Bank Question #75625 on following page.

Original Bank Question #75625 The unit is in MODE 0, and current conditions are as follows:

• All steps in OP3210B Refueling Operations, Section 4.1 "Core Off-Load" are complete.

• Radiation Monitor Trends are as follows:

• 3RMS-RE08-J (Spent Fuel Pool Bridge Hoist Area) shows increasing radiation levels.

• 3RMS-RE36-1 (Fuel Pool Area) shows increasing radiation levels.

• 3HVR-RE17-1 (Fuel Building Exhaust) shows no change in radiation levels.

What is the most likely cause ofthc increasing radiation levels on RE08 and RE36?

a) Spent Fuel Pool level is decreasing due to a Refueling Cavity Seal leak b) Spent Fuel Pool level is decreasing due to a Fuel Pool Cooling System leak c) One of the Fuel Assemblies is leaking gases from a cladding penetration d) One of the Fuel Assemblies was damaged during offload Correct Answer: B

Examination Outline Cross-reference: Level RO SRO Question # 23 Tier #

Loss of CTMT Integrity: Group # 2 2 Ability to determine/interpret adherence to appropriate KIA # EPE.W/EI4.EA2.2 procedures Importance Rating 3.3 _3_._8_ __

Proposed Question:

An earthquake occurs, resulting in the following sequence of events:

1. Safety Injection actuates due to four faulted SGs inside Containment.

2. The BOP operator throttles total AFW flow to 550 gpm.

3. The crew enters FR-Z.I, Response to High Containment Pressure.

4. The crew reaches FR-Z.I, step 10, "Check If Auxiliary Feedwater Flow Should Continue to All SGs".

What action is the crew required to take with AFW flow?

a) Stop AFW flow to all Steam Generators, to minimize the mass/energy release to CTMT.

b) Throttle AFW flow down to 100 gpm to each SG to minimize the mass/energy release to CTMT.

c) Control AFW flow to stabilize RCS temperature while maintaining minimum heat sink requirements.

d) Maintain current AFW flow rate to maintain minimum heat sink requirements.

Proposed Answer: _B_ _

Explanation (Optional): The caution prior to step 10 states "If all SGs are faulted, at least 100 gpm feed flow should be maintained to each SG ("A" wrong)." "A" is plausible since, per step 10 operators are required to isolate AFW flow to a faulted SG (but not with all four SGs being faulted). "B" is correct, and "c" and "D" wrong, since, per the caution, and the basis document, AFW is required to be throttled to 100 gpm to minimize the energy input to CTMT. "C" is plausible, since in other places in the EOP network, AFW flow is controlled to maintain hot leg temperature stable if hot legs were heating up. liD" is plausible, since normally, AFW flow is desired to be maintained at greater than 530 gpm due to heat sink concerns.

Technical Reference(s): FR-Z.l (Rev. 016-02), step 10, and Cautions prior to step 10 (Attach if not previously provided) WOG Bkgd Document (Rev. 2), for FR-Z.l, step 6, and Caution prior to step 6 (including version/revision number)

Proposed references to be provided to applicants during examination: -:..N;,.:o:.:.;n;.:.e_-:-_ _ _ _ _ _ __

Learning MC-07464 Given a set of plant conditions, properly apply the notes and cautions of (As available)

Objective: FR-Z.I.

New Question Source:

Question History:

Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 41.10 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 24 Tier #

Steam Generator Over-pressure: Group # 2 2 Knowledge of the reasons for manipulation of controls KJA# EPE.W/EI3.EK3.3 to obtain desired results Importance Rating 3.2 -,3;;..;..4~_ __

Proposed Question:

A turbine runback occurs, resulting in the following sequence of events:

1. The reactor trips.

2. The crew enters FR-H.2, Response to Steam Generator Overpressure.

3. FR-H.2 directs the crew to check RCS hot leg WR temperature 2: 530°F.

4. Based on RCS hot leg WR temperature being 2: 530°F, FR-H.2 directs the crew to conduct a cool down by dumping steam from the unaffected SGs.

Why does FR-H.2 direct the crew to cool down the RCS from the unaffected SGs?

a) To reduce affected SG pressure, since excessive heat transfer from the primary may be the cause of the overpressure condition.

b) To reduce thermal stresses on the U-tubes of the affected SG, since excessive pressure stress already exists on the tubes.

c) To prevent a rapid depressurization of the RCS, since the affected SG may not be at saturation conditions.

d) To prevent a radiation release, since the cause of the overpressure condition may be a SG Tube Rupture.

Proposed Answer: _A::..c::.-__

Explanation (Optional): "A" is correct, and "B", HC", and "0" wrong, since excessive heat transfer from the primary may be the cause of the affected SO overpressurization. Therefore, a check on RCS hot leg temperatures is made. JfRCS hot leg temperatures are greater than or equal to 530°F (Which is the saturation temperature of the lowest steamline safety valve setpoint, including allowances for channel accuracy), a cooldown is initiated by dumping steam from the unaffected SO(s) to aid in reducing the temperature and pressure in the affected SO(s). "B" is plausible, since an overpressure condition exists. "C" is plausible, since this is a basis related to high pressurizer level in FR-1.2, Response to High Pressurizer Level, and the SO may have a high pressure due to overfilling (but this would have led to an earlier transition to FR-H.3, Response to Steam Generator High Level. "0" is plausible, since this is a basis for cooling down the RCS if tube leakage is excessive.

Technical Reference(s): FR-H.2 (Rev. 009), step 7 (Attach if not previously provided) WOO Bkgd Document (Rev. 2), for FR-H.2, step 7 (including version/revision number)

Proposed references to be provided to applicants during examination: . . .; . N.;.:o. ;: n;.:.e_ _ _ _ _ _ _ _ __

Learning MC-05976 Discuss the basis of major procedure steps and/or sequence of steps in (As available)

Objective: EOP FR-H.2 Question Source: New Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 41.10 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 25 Tier #

High Containment Radiation: Group # 2 2 Knowledge of annunciators, indications, or response KJA# EPE.WIE 16.GEN.2.4.31 procedures Importance Rating 4.2 Proposed Question:

With the plant initially at 100% power, the following sequence of events occurs:

1. The main turbine trips, causing a reactor trip.

2. The crew enters ES-O.l, Reactor Trip Response.

3. SIS actuates due to a fault on the "A" Steam Generator inside Containment.

4. The crew returns to E-O, Reactor Trip or Safety Injection.

5. Containment Hi Range Radiation Monitors 3RMS-RE04A and 05A start rapidly increasing.

6. Fuel Drop Radiation Monitors 3RMS-RE41 and RE42 show no change in radiation levels.

7. The Containment Status Tree changes color, indicating FR-Z.3, Response to High Containment Radiation Levels.

What action, ifany, is the crew required to take based on the high radiation indications on 3RMS-RE04A and RE05A?

a) The crew IS required to transition to FR-Z.3, since FR-Z.3 is a Red Path procedure.

b) The crew IS required to transition to FR-Z.3, since FR-Z.3 is an Orange Path procedure.

c) The crew is NOT required to transition to FR-Z.3. The input to the status tree is 3RMS-RE41, which has not changed. The status tree is invalid.

d) The crew is NOT required to transition to FR-Z.3. The 3RMS-RE04N05A readings are NOT valid for about 2 to 5 minutes due to TIC effects.

Proposed Answer:

-D- -

Explanation (Optional): "D" is correct, since the effects of Therrnal Induced Current are significant but of relatively short duration. Generally, the effects ofthe TIC will dissipate within 2-5 minutes. Until the TIC effects have dissipated the RE04NRE05A readings should not be considered valid. "A" and "B" are wrong, since FR-Z.3 us a YELLOW path procedure, which is considered a supplementary set of actions provided to address individual parameters in an off-normal state. The Operator is not required to implement a YELLOW path procedure if it is considered inappropriate based on available time or current plant status. "C" is wrong, since the rad monitors that input to the status tree are the high range monitors. "A" and "B" are plausible, since indicated increasing radiation has caused a status tree (:olor change, and red and orange paths require transition. "C" is plausible, since a failed input to a status tree allows the crew to manually check the tree, and not make a transition if the status tree indication is found to be false.

Technical Reference(s): EP-MP-26-EPA-REF03 (Rev. 015), page 50 of 141 (Attach if not previously provided) OP 3272 (Rev. 008-11), Attachment 4, Sheets 4 and 5 of7 (including version/revision number) EOP35 F-05 (Rev. 004), CTMT Status Tree Proposed references to be provided to applicants during examination: None Learning (As available)

Objective: MC-05961 Identify plant conditions that require entry into EOP35 FR-Z.3.

Question Source: New Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 41.10 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 26 Tier # 1 LOCA Cooldown -Depress: Group # 2 2 Ability to operate and/or monitor for desired operating KJA# EPE. WIE03 .EA 1.3 results Importance Rating 3.7 4.1 Proposed Question:

A small break LOCA has occurred, and the operators are cooling down the plant per E8-1.2 Post-LOG'A Cooldown and Depressurization.

What RCP configuration is desired, and why?

a) One RCP running for effective heat transfer and pressure control while minimizing heat input.

b) One RCP running for ECC8 mixing considerations while minimizing inventory loss.

c) All RCPs stopped to minimize heat input into the Reactor Coolant System.

d) All RCPs stopped to prevent possible core uncovery should a loss of offsite power occur.

Proposed Answer: A Explanation (Optional): ES-l.2 background states that forced flow is the preferred mode of operation ("C" and "D" wrong) to allow for normal RCS cooldown and provide pressurizer spray ("A" correct and "B" wrong). All but one are stopped to minimize heat input to the RCS. "B" is plausible, since ECC8 mixing is a basis for starting an RCP in FR-P.1. "C" is plausible, since this the basis for stopping RCPs in FR-H.l "D" is plausible, since this is a basis for RCP trip criteria, which do not apply once a controlled cool down is commenced.

Technical Reference(s): E8-1.2 (Rev. 017-01), step 12 (Attach if not previously provided) WOG Bkgd Document (Rev. 2), for ES-1.2, step 12 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning (As available)

Objective: MC-05529 Describe the major action categories within EOP 35 E8-1.2.

Question Source: Bank # 70250 Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 41.10 Comments:

Examination Outline Cross-reference: Level RO SRO 1

~l _ _ __

Question # 27 Tier#

RCS Overcooling PTS: Group # 2 2 Ability to determine/interpret selection of appropriate KIA # EPE.W/E08.EA2.1 procedures Importance Rating 3.4 _4..;..:.=-2_ __

Proposed Question:

INITIAL CONDITIONS:

• The crew has been performing a natural circulation cooldown per ES-O.2, Natural Circulation Coo/down ..

• Tcold has been steady at 335°F for the past hour.

THE FOLLOWING SEQUENCE OF EVENTS OCCURS:

1. A large steamline break occurs in Containment.

2. The crew actuates safety injection and returns to E-O, step 1.

3. The STA continues manually monitoring the status trees, recording Teold every 5 minutes.

RCS cold leg temperature trending indicates the following:

TIME Tcold 1300 335°F 1305 274°F 1310 253°F 1315 232°F 1320 215°F At what time was the crew required to transition to FR-P.l, Response to Imminent Pressurized Thermal Shock Condition?

a) 1305 b) 1310 c) 1315 d) 1320 Proposed Answer:  :---:-C-:--_

Explanation (Optional): The status tree is green until temperature has cooled down by 100°F in the past 60 minutes (" A" and "B" wrong). With the 1OO°F cooldown, the status tree will tum ORANGE below 256°F

("C" correct, liD" wrong). "A" is plausible, since temperature is below where the status tree would tum YELLOW, if the 100°F cooldown had occurred. "B" is plausible, since temperature is below the transition temperature of 256°F. "D" is plausible, since the status tree will have turned from ORANGE to RED.

Technical Reference(s): 35 EOP F.04 (Rev. 006), Integrity Status Tree (Attach ifnot previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning (As available)

Objective: MC-04551 Identify plant conditions that require entry into EOP 35 FR-P.1.

Question Source: Bank # 75642 Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 41.10 Comments:

Examination Outline Cross-reference: Level SRO Question # 28 Tier # 2 2 Reactor Coolant Pump: Group #

Predict and/or monitor changes in RCP Standpipe level KJA# 003.A1.l0 associated with operating Controls Importance Rating 2.5 2.7 Proposed Question:

With the plant at 100% power, conditions are as follows:

I. The Primary Grade Water CTMT Isolation Valves are open.

2. The RO opens RCP A Standpipe Fill Valve (3PGS-LCVI81) at MB4.

How does RCP A standpipe level respond if no further operator action is taken, and how can the RO monitor standpipe level?

a) Level will increase until the standpipe high level setpoint is reached. The RO can monitor standpipe level on a meter on MB3.

b) Level will increase until the standpipe high level setpoint is reached. The RO can indirectly monitor standpipe level via a standpipe HiILo level annunciator on MB4.

c) Level will increase until the standpipe overflows to the CDTT. The RO can monitor standpipe level on a meter on MB3.

d) Level will increase until the standpipe overflows to the CDTT. The RO ean indirectly monitor standpipe level via a standpipe HilLo level annunciator on MB4, Proposed Answer: ~B_ _

Explanation (Optional): The standpipe fill valve will automatically close when the standpipe high level setpoint is reached (HC" and "D" wrong), "B" is correct, and "A" wrong, since the only way to monitor standpipe level in the eontrol room is the high/low level alarms. "A" is plausible, since numerous RCP Seal injectionlleakoff indications are available on MB3. "c" and "D" are plausible, since if the standpipe overflows, it overflows to the CDTT.

Technical Reference(s): OP 3353.MB4B (Rev. 004-11), 2-2B (Attach if not previously provided) LSKs 26-2.6B (No. 008) and 35-1 C (No. 08)

(including version/revision number) P&IDs 103A (No. 024) and 119A (No. 30)

Proposed references to be provided to applicants during examination: None Learning (As available)

Objective: MC-05427 Describe the following RCP fluid flowpaths.,. Seal flow ...

Question Source: New Question History:

Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 41.3 and41A Comments:

Examination Outline Cross-reference: Level RO SRO Question # 29 Tier # 2 2 Chemical and Volume Control: Group #

Predict impact/mitigate isolation of letdown/makeup KJA# 004.A2.07 Importance Rating 3.4 3.7 Proposed Question:

With the plant initially at 100% power, the following sequence of events occurs:

1. The LETDOWN RELIEF VV TEMP HI annunciator is received on MB3A.

2. The US enters the associated Annunciator Response Procedure.

3. The RO reports VCT level is 44% and decreasing.

To where is letdown currently routed; and what action will the ARP direct the crew take to mitigate the event?

a) Letdown flow is being directed to the CDTT. The RO will commence a manual makeup to the VCT.

b) Letdown flow is being directed to the CDTT. The RO will place 3CHS*PKI31 in MANUAL and attempt to restore letdown header pressure to normal.

c) Letdown flow is being directed to the PRT. The RO will commence a manual makeup to the VCT.

d) Letdown flow is being directed to the PRT. The RO will place 3CHS*PK131 in MANUAL and attempt to restore letdown header pressure to normal.

Proposed Answer: -:::;..D_ _

Explanation (Optional): Letdown Relief Valve 3CHS*RV8117 (600 psig lift setpoint) is lifting, directing letdown flow to the PRT ("A" and "B" wrong). "A" and "B" are plausible, since the Containment Drains Transfer Tank (CDTT) receives numerous drains inside CTMT. The ARP directs the crew to take manual control of 3CHS*PK131 to open 3CHS*PCV131 to restore letdown flow and lower letdown header pressure

("D" correct). "C" is wrong, since, even though letdown is no longer flowing to the VCT, and VCT level is dropping (HC" plausible), auto makeup will still function at 41 %.

Technical Reference(s): _O_P_3_3_53_._M_B....3_A....;{.....R_e_v._O_O_2_-0_8.(..!.),_4_-6_ _ _ _ _ _ _ _ _ _ _ _ _ __

(Attach if not previously provided) -=-P.::.:&=lD::;....:1...;.O;;;;.2F~(N;..;..::.:o.:....:1:..:6L)_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

(including version/revision number) --"-P.::..&:.::lD==-.;1:....:0:....:4.::..A=-(N~0:..:...:;;5-=2L)_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

Proposed references to be provided to applicants during examination:

Learning MC-04202 Describe the operation ofthe Chemical and Volume Control System -------------

None (As available)

Objective: under normal, abnormal, and emergency operating conditions.

Question Source: New Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 41.3,41.5 and 41.1 0 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 30 Tier # 2 2 Residual Heat Removal: Group # 1 Knowledge of power supplies to RCS pressure KJA# OOS.K2.03 boundary MOVs Importance Rating 2.7 2.8 Proposed Question:

Which "A" RHR Pump Suction Valve is powered from MCC 32-3U?

a) 3RHS*MV870IA, "A" RHR Pump Suction Inside Ctmt Isolation from the RCS.

b) 3RHS*MV870IB, "A" RHR Pump Suction Outside Ctmt Isolation from the RCS.

c) 3RHS*MV870IC, "A" RHR Pump Suction Inside Ctmt Isolation from the RCS.

d) 3SIL*MV8812A, "A" RHR Pump Suction Isolation from the RWST.

Proposed Answer: -=.B_ _

Explanation (Optional): The "A" Residual Heat Removal pump suction line is isolated from the RCS by three normally-dosed motor-operated valves (3RHS*MV870IA, B, and C) in series ("A" and "C" plausible). The two normally closed isolation valves inside Containment (3RHS*MV870IA and C) receive power from the same Class IE source as the RHS pump in that train, while the valve outside containment (3RHS*MV870lB) is powered by the opposite train. This arrangement ensures that single failure requirements for RHS accessibility and isolation are met. The suction ofthe "A" RHR pump is also connected to the RWST through 3SIL*MV8812A, located in the "A" RHR Pump room ("D" plausible).

The power supplies to the Valves are as follows:

3RHS*MV8701A: 32-2R ("A" wrong) 3RHS*MV8701B: 32-3U (HB" correct) 3RHS*MV8701C: 32-2R (HC" wrong) 3SIL*MV8812A: 32-4T ("0" wrong)

Technical Reference(s): OP 331OA-005 (Rev. OOS-Ol), page 3 of7 (Attach ifnot previously provided) FSAR (Rev. 21.3), Section 5.4.7.1, page 5.4-26 (including version/revision number)

Proposed references to be provided to applicants during examination: --:-N_o_n_e_ _ _ _ _ _ _ _ __

Learning MC-OS45S Deseribe the operation ofthe following Residual Heat Removal system (As Objective: equipment controls and interlocks: A. Loop Suction Valves B. RWST Suction Valves ... available)

Question Source: New Question History:

Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 41.3 and 41.8 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 31 Tier# 2 2 Emergency Core Cooling: Group # 1 Ability to perform without reference immediate actions KIA # 006.GEN.2.4.49 Importance Rating 4.6 4.4 Proposed Question:

With the plant initially at 100% power, the following sequence of events occurs:

1. An inadvertent Train "B" Safety Injection Signal is received.

2. Train "A" Safety Injection Signal is NOT received.

3. The crew enters E-O Reactor Trip or Safety Injection.

Prior to operator action, what is the status of suction to the Charging Pumps; and what immediate action will E-O direct the operators to take to specifically address the current ECCS alignment?

a) Charging Pump suction is aligned to both the RWST and the VCT. The crew will NOT actuate SIS, to minimize the mass added to the RCS during this inadvertent SIS.

b) Charging Pump suction is aligned to both the RWST and the VCT. The crew WILL actuate SIS to establish a known ECCS system alignment prior to proceeding in the EOP network.

c) Charging Pump suction is aligned to the RWST, and isolated from the VCT. The crew will NOT actuate SIS, to minimize the mass added to the RCS during this inadvertent SIS.

d) Charging Pump suction is aligned to the RWST, and isolated from the VCT. The crew WILL actuate SIS to establish a known ECCS system alignment prior to proceeding in the EOP network.

Proposed Answer: D Explanation (Optional): On a Safety Injection, the Charging Pump Suction Valves from the RWST (3CHS*LCVl12D and E) open (Train-specific). Since these valves are in parallel, suction is aligned.

Charging Pump suction from the VCT isolates (3CHS*LCVl12B and C) isolate (Train-specific). Since these valves are in series, suction from the VCT has been isolated (itA" and "B" wrong). "A" and "B" are plausible, since pump suction is desirable, and only a single train ofSI has actuated. "D" is correct, and "c" wrong, since if only a single train of SIS has actuated E-O directs the crew to actuate the second train of SIS. "C" is plausible, since the SIS is inadvertent, and a single train of SIS will add less mass to the RCS. This action was taken at Salem in response to the eel grass event, and the second train of SIS actuated later in the event.

Technical Reference(s): ~E....;-O....;(>.:.R;;;;.e..;.v..;..0;; ;.2; ; .6:. t.),'-'s;. ;.te; .1p:. . 4.:;..:;.;;;.b.;.;;;.RN~O=--_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

(Attach if not previously provided) WOG Bkgd Doc (Rev. 2), for E-O step 4 (including version/revision number) -.::...P=&=ID=-..:1:.:0....;4.::D:....(;:.N..:.:o::.:..=29::...),~_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

Proposed references to be provided to applicants during examination: None

~~~--------------

Learning MC-06289 Given a failure (partial or complete) of the Emergeney Core Cooling (As available)

Objective: System, determine the effects on the system and on interrelated systems.

Question Source: Bank # 85249 Question History: Millstone 3 2009 NRC Exam Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 41.7 and 41.10 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 32 Tier #

Pressurizer Relief/Quench Tank: Group #

Physical connections/cause-effect between PRT and KJA# 007.Kl.Ol CTMT System Importance Rating 2.9 3.1 Proposed Question:

PLANT CONDITIONS:

• The plant is in MODE 3.

• The "A" PORV (3RCS*PCV455A) is leaking by and will not reseat.

• Block valve 3RCS*MV8000A will not close.

Assume Safety Injection does not actuate, and no further operator action is taken.

To where will the excess water entering the PRT eventually be routed?

a) When the PRT Rupture Disk fails, the water will be released to the Containment atmosphere. It will then collect in the Unidentified Leakage Sump, and from there, be pumped to the Identified Leakage Sump, which is then pumped to Rad Waste.

b) When the PRT Rupture Disk fails, the water will be released to the Containment atmosphere. It will then collect in the Identified Leakage Sump, and from there, be pumped to the Unidentified Leakage Sump, which is then pumped to Boron Recovery.

c) When the PRT overfills, the water will automatically relieve to the Primary Drains Transfer Tank (PDTT). When the PDTT overfills, it will relieve to Rad Waste.

d) When the PRT overfills, the water will automatically relieve to the Containment Drains Transfer Tank (CDTT). When the CDTT overfills, it will relieve to Boron Recovery.

Proposed Answer:

-A- -

Explanation (Optional): "A" is correct since the PRT is protected from overpressure by a rupture disk, which will rupture and release its contents to the containment atmosphere as it flashes to steam ("C" and "D" wrong.

As the steam condenses, it will collect on the CTMT walls and floor, draining into the unidentified leakage sump ("B" wrong, but plausible). From here, the water is pumped to the identified leakage sump, which is pumped to rad waste. HC" and "D" are plausible, since the CDTT and PDTT receive reactor plant gaseous drains from numerous sources, and their discharge can be routed to either rad waste or boron recovery.

Technical Reference(s): P&lD 102F (No. 16)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: -"-N.. .;o.. .;n.. .;e_ _ _ _ _ _ _ _ __

Learning MC-05349 Describe the Pressurizer Relief Tank System operation, or operations (As available)

Objective: required, under the following normal, abnormal, or emergency operating conditions or procedures... Pressurizer Safety Valve OR Power Operated Relief Valve discharge ...

Question Source: New Question History:

Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 41.7 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 33 Tier # 2 2 Pressurizer Relief/Quench TanJe Group #

Design feature andJor interlock that provides for KJA# 007.K4.01 PRT cooling Importance Rating 2.6 2.9 Proposed Question:

Initial Conditions:

• The plant is at 100% power.

• The Primary Grade Water (PGS) CTMT Isolation Valves are closed.

The following sequence of events occurs:

1. A discharge to the PRT occurs.

2. PRT temperature is high.

3. The US directs the RO to cool the PRT by filling it from PGS using OP 3301A, Pressurizer ReliefTank and Reactor Vessel Flange LeakoffOperations.

In accordance with OP 330lA, what actions are required by the RO to commence cooling the PRT with PGS water?

a) The RO must manually open the PRT Fill Valve at MB4, and PGS flow will initiate, since the PGS Containment Isolation Valves will have automatically opened.

b) The RO must manually open the PGS Containment Isolation Valves at MBl, and manually open the PRT Fill Valve at MB4, initiating PGS flow.

c) The RO must manually open the PGS Containment Isolation Valves at MBl, and PGS flow will initiate, since the PRT Fill Valve will have automatically opened.

d) The RO is required to monitor for proper valve operation, since both the PGS Containment Isolation Valves and the PRT Fill Valve will have automatically opened, initiating PGS flow.

Proposed Answer: B Explanation (Optional): "B" is correct, and "A", "C", and "D" wrong, since all of these valves are manually opened and closed. "A", "C", and "0" are plausible, since the CIVs have an auto-close feature, and the PRT Vent Vale has an auto-close feature. Also, the PRT drain valve is interlocked with PRT level.

Technical Reference(s): OP 330lA (Rev. 008-05), Section 4.2 (Attach if not previously provided) -:;..P.:.;&;.;;;;ID;;.. .:1..;:1.:.;9A;;.;;.. J>(N.:.;. ;:o.:.;.3=-.0:..<)_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

(including version/revision number) ESKs 7Z (No. 10) and 7EZ (No. 06)

Proposed references to be provided to applicants during examination: -:;..N;..;o.:.;n.:.;e_ _-:--:-_ _ _ _ __

Learning MC-05349 Describe the Pressurizer Relief Tank System operation, or operations required, (As Objective: under the following normal, abnormal, or emergency operating conditions or available) procedures... Restoring from a high Pressurizer Relief Tank Temperature condition ...

Question Source: Bank # 80878 Question History: Millstone 3 2007 NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 41.7 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 34 Ticr# 2 2 Component Cooling Water: Group #

Predict impact/mitigate the shutting of the letdown K/A# 00S.A2.0S cooler isolation valves Importance Rating 2.5 2.7 Proposed Question:

With the plant at 100% power and the Degassifier out of service, the following sequence of events occurs:

1. The "LETDOWN HX OUT TEMP HI" annunciator is received on MB3A.

2. The RO reports RPCCW to Letdown Heat Exchanger Valve 3CCP-TCVl72 has failed CLOSED.

What will be the initial effect on the plant; and for this event, if the crew is slow to respond, what parameter will be exceeded that will require the RO to trip the plant?

a) Demineralizer Divert Valve 3CHS*TCVI29 will automatically bypass the CHS demineralizers. The RO will trip the plant when YCT temperature exceeds 135°F.

b) Demineralizer Divert Valve 3CHS*TCYI29 will automatically bypass the CHS demineralizers. The RO will trip the plant when RCP bearing oil temperatures exceed 195°F.

c) Flashing will occur at the discharge of the in-service Letdown Orifice. The RO will trip the plant when VCT temperature exceeds 135°F.

d) Flashing will occur at the discharge of the in-service u~tdown Orifice. The RO will trip the plant when RCP bearing oil temperatures exceed 195°F.

Proposed Answer: A Explanation (Optional): Loss ofRPCCW to the Letdown Heat Exchanger causes Letdown Temperature to increase, and when temperature increases to 134°F, letdown flow will bypass the demineralizers. The hotter water in the letdown stream will heatup the VCT water, which is supplied to the RCP seals. The crew will receive a VCT high temperature annunciator, which will direct the crew to AOP 3561. AOP 3561 requires a plant trip ifVCT temperature reaches 135°F ("A" correct, and "B" wrong). "B" is plausible, since RPCCW provides cooling to the RCP bearings, CHS provides RCP seal cooling, and 195°F bearing temperature is a trip criterion on the AOP 3561 Foldout Page. "c" and "D" are wrong, since the letdown orifice is upstream of the letdown heat exchanger, and the location vulnerable to flashing in the letdown stream on loss of cooling to the letdown heat exchanger is downstream of the letdown heat exchanger, where temperature will be higher than normal, and pressure is at its lowest. "c" and "D" are plausible, since cooling has been lost to the letdown stream, and a pressure reduction occurs through the letdown orifice.

Technical Reference(s): _O_P_3_3_53_.M_B_3_A_C,..,.R_e_v._0_0_2_-0_S"""),_5_-5_ _ _ _ _ _ _ _ _ _ _ _ _ __

(Attach if not previously provided) OP 3353.MB3A (Rev. 002-0S), 5-10 (including version/revision number) AOP 3561 (Rev. 011-02), Foldout Page P&ID 104A (No. 52)

Proposed references to be provided to applicants during examination: _N....o....n_e_ _ _ _ _ _ _ _ __

Learning MC-07542 Given a set of plant conditions, properly apply the notes, cautions, and (As available)

Objective: foldout page items of AOP 3561.

Question Source: New Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 41.3,41.4,41.7, and 41.10 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 35 Tier# 2 2 Component Cooling Water: Group #

Ability to monitor automatic operation due to SIS KiA # 008.A3.0S Importance Rating 3.6 3.7 Proposed Question:

With the plant initially at 100% power, the following sequence of events occurs:

1. Safety Injection actuates.

2. The RO is monitoring the automatic RPCCW System response to the event.

What RPCCW valve positions will the RO observe?

CCP Non-Safety Hdr Valves CCP/CDS X-Tie Valves a) Open Open b) Close Close c) Open Close d) Close Open Proposed Answer: D Explanation (Optional): SIS actuates CIA, and SIS/Containment Isolation Phase A automatically operates the following CCP valves as follows:

The RPCCW Non-Safety Header valves will CLOSE ("A" and "c" wrong).

The CCP to CDS cross-connect valves will OPEN, supplying loads normally cooled by Chilled Water ("B" wrong, and "D" correct). "A", "B", and "C" are plausible, since both ofthese valves receive signals on SIS/CIA, and CCP isolates to CTMT on a CDA signal.

Technical Reference(s): P&ID 121 A (No. 32)

(Attach if not previously provided) P&ID 121 B (No. 20)

(including version/revision number)

Proposed references to be provided to applicants during examination: -:.N.:..:o:..::n:.,:e_ _ _ _ _ _ _ _ __

Learning MC-04154 Describe the operation of the Reactor Plant Component Cooling System (As available)

Objective: under the following normal, abnormal, or emergency conditions ... Sequence Safeguards Signal actuation ...

Question Source: New Question History:

Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 41.4 and 41.7 Connnents:

Examination Outline Cross-reference: Level RO SRO Question # 36 Tier # 2 2 Pressurizer Pressure Control: Group # 1 1 Physical connections/cause-effect between PZR PCS KJA# OlO.Kl andPZRLCS Importance Rating 3.2 3.5 Proposed Question:

With the plant initially at 100% power, the following sequence of events occurs:

1. A turbine ronback initiates.

2. During the ronback, both PZR spray valves start to throttle open.

3. As the runback continues, the PZR backup heaters energize, even though spray valves are still open.

Why have the backup heaters energized?

A. The PZR level controller is responding to a greater than 5% outsurge from the downpower, to restore the PZR liquid to saturation conditions.

B. The PZR level controller is responding to a greater than 5% insurge from the downpower, to restore the PZR liquid to saturation conditions.

C. The PZR pressure controller is responding to the PZR pressurizer pressure rise via a rate/lag compensated circuit, to prevent pressure oscillations as pressure is restored to 2250 psia.

D. The PZR pressure controller is responding to the PZR pressurizer pressure via a rate/lag compensated circuit, to prevent pressure oscillations as pressure is restored to 2250 psia.

Proposed Answer: B Explanation (Optional): The downpower will cause RCS temperature to increase due to a decrease in heat removal. This will cause RCS water to expand, resulting in a insurge to the pressurizer, so both PZR pressure and level will increase. The increase in pressure causes spray valves to open, and when pressurizer level increases by 5%, the heaters will energize ("B" is correct, nAn is wrong). The reason for this is that the temperature of the insurging water is not as hot as the pressurizer water, and if an outsurge follows with the pressurizer water at less than saturation temperature, RCS pressure could rapidly drop. "A" is plausible, since pressurizer level is changing due to a heat imbalance between the primary and secondary plant. "C" and "0" are wrong since backup heaters cycle around 2225 to 2233 psia, and spray valves cycle around 2275 to 2325 psia, and shouldn't both be on together based on pressure. "c" and "0" are plausible, since pressurizer heaters are normally controlled by an input from pressurizer pressure.

Technical Reference(s): Functional Sheet 11 (No. H)

(Attach ifnot previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: ~N_o,-n....:.e.,...-_ _ _ _ _ _ _ __

Leaming MC-05341 Describe the operation of the Pressurizer Pressure and Level Control (As available)

Objective: System under Normal, Abnormal, and Emergency Operating conditions.

Question Source: Bank #68619 Question History: Millstone 3 2009 NRC Exam Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.5 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 37 Tier # 2 2 Reactor Protection: Group #

Ability to recognize entry conditions for EOPs/ AOPs KJA# O12.GEN.2.4.4 Importance Rating 4.5 4.7 Proposed Question:

The plant is initially at 100% power with all selectable controllers on the Main Boards selected to Channell.

The following initial sequence of events occurs:

1. Pressurizer pressure instrument 3RCS*PI458 (Channel 4) fails HIGH.

2. The crew enters AOP 3571, Instrument Failure Response.

3. The crew completes all actions of AOP 3571, including tripping all bistables.

RCS Loop A Tave instrument 3RCS*TI412 (Channel 1) fails LOW due to its Tcold narrow range temperature instrument failing LOW.

Is the crew required to enter E-O, Reactor Trip or Safety Injection due to an automatic reactor trip? If so, which reactor trip signal was received?

a) The crew is not required to enter E-O. The reactor remains at power.

b) The crew will enter E-O, based on an OPaT reactor trip.

c) The crew will enter E-O, based on an OTaT reactor trip.

d) The crew will enter E-O, based on a high Pzr pressure reactor trip.

Proposed Answer:

-C- -

Explanation (Optional): The original pressure transmitter failure's corrective action required the crew to trip the OTDT bistable for loop "D", along with other bistables. When the Tcold channel fails low, DT fails high, bringing in an OTDT and OPDT bistable (liB" plausible) for channell, meeting the 2 of 4 coincidence for an OTDT trip ("C" correct, "A" wrong). "B" is wrong, since OPDT bistable was not tripped on a failed pressure instrument. liD" is wrong, since the high pressure trip reqwres 2 of 4 bistables, and only one bistable is tripped. "A" is plausible, since two different types of instruments failed. "D" is plausible, since the original failure was a pressure channel.

Technical Reference(s): AOP 3571 (Rev. 009-07), Attachment A, page 6 of6 (Attach ifnot previously provided) AOP 3571 (Rev. 009-07), Attachment B, page 6 of6 (including version/revision number) --=-P:::.;.&=ID:::.......:l:....:0=2:::.;.A:::.;.(N~0:..:.-=2=9.L)_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

Proposed references to be provided to applicants during examination: None Learning MC-05493 Describe the operation of the following RPS controls and (As available)

Objective: interlocks... Reactor Trip Signals ...

Question Source: New Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 41.7 and 41.10 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 38 Tier # 2 Engineered Safety Features Actuation: Group #

Knowledge of the effect of a loss or malfunction of KJA#

detectors on ESF AS Importance Rating 2.7 3.1 Proposed Question:

INITIAL CONDITIONS:

The plant is operating at 100% power Containment Pressure Channel III (PT-935) has failed high The appropriate bistables have been tripped/bypassed.

I & C are about to begin troubleshooting the failed channel Which safeguards signal(s) would be generated if the I&C technician inadvertently de-energized the control and instrument power for the Channel II Containment Pressure instrument?

a) Only an SI signal will be generated.

b) Only SI and M81 signals will be generated.

c) Only 81 and COA signals will be generated.

d) 81, MS1, and COA signals will be generated.

Proposed Answer: ~B_ _

Explanation (Optional): The SI and MSI bistables are de energize to actuate. One channel is already tripped, when the second channel is de energized, the logic is complete and the signals will be actuated ("A" and "C" wrong). The COA signal is energize to actuate and therefore will not be affected with the second channel loss of power (liB" correct, "C" and "0" wrong). "A", "C", and "0" are plausible, since each of these ESFAS signals receive input from CTMT pressure.

Technical Reference(s): Functional sheet 8 ,No. J)

(Attach ifnot previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: --,-N~o~n;.;;.e_ _-:-_ _ _ _ _ __

Learning MC-05498 Given a failure (partial or complete) of the RP8, determine the effects on the (As available)

Objective: system and on inter-related systems: A. Power Failure. B. Instrumentation Failure.

Question Source: Bank # 69327 Question History: Millstone 3 2004 NRC Exam Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 41. 7 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 39 Tier # 2 2 Containment Cooling: Group #

Predict and/or monitor changes in Containment humidity KJA# 022.Al.03 due to operating controls Importance Rating 3.1 3.4 Proposed Question:

The crew has just started a second CAR fan.

How will the crew's action affect Containment humidity, and where can this be monitored in the control room?

a) Humidity will increase. This can be monitored on the dewpoint meter on MB2.

b) Humidity will increase. This can be monitored on the dewpoint meter at VP 1.

c) Humidity will decrease. This can be monitored on the dewpoint meter on MB2.

d) Humidity will decrease. This can be monitored on the dewpoint meter at VPl.

Proposed Answer: --:;;;C_ _

Explanation (Optional): Each CAR fan draws air across the cooling coil assembly and discharges the air to a common duct which distributes it through secondary ducts to different levels of the containment. This will increase CTMT cooling, decreasing CTMT temperature. Also, the cold (below the dewpoint) cooling coil will condense water vapor from the CTMT air as it passes over the coils, decreasing CTMT humidity ("A" and "B" wrong). Dewpoint meter 3LMS-ME22C indicates on MB2 ("C" correct, "D" wrong). "A" and "B" are plausible, since increasing cooling affects humidity. "D" is plausible, since VPl contains Ctmt Ventilation System Controls.

Technical Reference(s): OP 3313B (Rev. 007-02), section 1.2 (Attach if not previously provided) --"-P&.:. ;.;:.;ID;;.. .;;.I.:;..54..;.;;A..;:.. .>;(N..;.;;o.:;.. _2..;.;;6."-2_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

(including version/revision number) www.newworldencyclopedia.org/entry/Air conditioning Proposed references to be provided to applicants during examination: None Learning MC-04248 Describe the operation of the following ... Containment Leakage (As available)

Objective: Monitoring System... Indications... Containment humidity monitors.

Question Source: New Question History:

Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 41.4 and 41.9 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 40 Tier # 2 2 Containment Spray: Group # 1 1 Knowledge of the effect of a loss or malfunction of CSS KJA# 026.K3.01 on the Ctmt Cooling System Importance Rating 3.9 4.1 Proposed Question:

With the plant initially at 100% power, a LOCA occurs, resulting in the following sequence of events:

1. Safety Injection actuates.

2. CDA actuates.

3. Both Quench Spray Pumps fail to start.

4. All other CDA-related signals and equipment operate as designed.

5. The crew is at E-O, Reactor Trip or Safety Injection, step 4.

What is the status of Containment Cooling Systems?

a) ALL previously running Containment Recirculation (CAR) Fans and CRDM Fans remain RUNNING.

b) ALL previously running CAR Fans are TRIPPED. ALL previously running CRDM Fans remain RUNNING.

c) ALL Safety Related Containment Recirculation (CAR) Fans and CRDM Fans remain RUNNING. ALL Non-Safety related CAR and CRDM Fans are TRIPPED.

d) ALL Safety Related CAR and CRDM Fans are TRIPPED. ALL previously running Non-Safety Related CAR Fans and the CRDM Fans remain RUNNING.

Proposed Answer: -:=.D_ _

Explanation (Optional): A HI-3 signal is generated at Containment Pressure (2/4) >23 psia which generates a CDA signal. The QSS Pumps should have started on the CDA signal, but have not. The CDA signal will trip the safety related CTMT Recirc and CRDM fans whether or not QSS Pumps start ("B" and "C" wrong), since they have lost cooling. There is no ESF actuation signal which trips the Non-safety related fans ("D" correct). "A" is wrong, since OP 3313B requires that 2 Containment Recirculation Fans be operating and that both safety related fans not be nonnally operated together to prevent overheating the pressurizer cubicle. OP 3313C requires that 2 CRDM fans be operated when required to have them in service, but has no restriction on which fans are operated together. "A", "B", and "c" are plausible, since a combination of safety and non safety fans are initially running, with some remaining running and some tripping.

Technical Reference(s): OP 3313B (Rev. 007-02), section 4.1 Cautions and Notes (Attach if not previously provided) --"-O.....

P...........

33......

13C . . . .............

(Rev-'

. . . . . . . -.-'-00""'.6. -0.:..:3:...<)"-,

.. s""'te.:. .\.p. . 4. .:..:......

1.;.:.1_ _ _ _ _ _ _ _ _ _ _ _ _ __

(including version/revision number) _P& __.. .I.. D_l 53A.......>...(N_o_._2_8_L) _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

Proposed references to be provided to applicants during examination: None

~~-----------

Learning MC-04259 Describe the operation of the following Containment Ventilation System (As available)

Objective: controls and interlocks ... Containment Air Recirculation System... Containment Rod Drive Mechanism Cooling System ...

Question Source: Bank # 72259 Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 4l.7 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 41 Tier # 2 2 Containment Spray: Group #

Predict and/or monitor changes in Ctmt Spray Pump KJA# 026.Al.06 cooling associated with operating controls Importance Rating 2.7 3.0 Proposed Question:

A LOCA has occurred, and initial conditions are as follows:

• All four Containment Recirculation (RSS) Pumps are running.

• CDA has been reset.

• The crew has just entered ES-l.3, Transfer to Cold Leg Recirculation.

• The crew has not yet taken any actions in ES-l.3.

The RO inadvertently depresses the CLOSE pushbutton for "A" RSS Pump Discharge Valve 3RSS*MY20A.

What will be the effect on cooling for the "A" RSS Pump?

a) "A" RSS pump cooling will be lost.

b) "A" RSS pump cooling will be maintained, since 3RSS*MV20A will not stroke closed.

c) "A" RSS pump cooling will be maintained, since recirculation valve 3RSS*MY38A will automatically open.

d) "A" RSS pump cooling will only be maintained ifthe 3RSS*MV38A "Override" pushbutton has been depressed on MB2.

Proposed Answer: _C.:::..-__

Explanation (Optional): The discharge MOY will stroke closed, since CDA has been reset ("B" wrong). "B" is plausible, since the valve would not stroke closed if the CDA signal were not reset. "C" is correct, and "A" wrong, but plausible, since the recirc valve auto-opens to maintain cooling flow to the pump only if all of the following are met: The discharge valve is not open, flow is low (it is, since there is no other flowpath available with the discharge valve closed), and the pump has been running (it has). "D" is wrong, but plausible, since the Block-Auto pushbutton is used to allow opening the recirc valve manually with the discharge valve open.

Technical Reference(s): LSK 27-11C (No. LO)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: ~N;..;;o;..;;n;..;;e_ _ _ _ _ _ _ _ __

Learning MC-04171 Describe the operation of the following containment depressurization (As available)

Objective: system components controls and interlocks ... Containment Recirculation Pump Recirculation Valves ...

Question Source: New Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 41.7,41.9, and 41.14 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 42 Tier # 2 Main and Reheat Steam: Group #

Ability to manually operate andlor monitor steam dump KJ A # 039.A4.07 valves Importance Rating 2.8 Proposed Question:

Initial Conditions:

• The plant is at 100% power.

• Turbine impulse pressure transmitter 3MSS-PT506 has failed.

• Steam dumps have been placed in the "Steam Pressure" Mode.

The following sequence of events occurs:

I. The plant trips.

2. RCS Tave drops to 550°F, and the stcam dump valves close.

3. The crew desires to commence a plant cool down.

What initial action is required to cooldown the plant with the steam dump valves?

a) Take the Mode Selector Switch to RESET.

b) Take the Bypass Interlock Selector Switches to OFF/RESET.

c) Take the Bypass Interlock Selector Switches to BYP INTLK.

d) Take the steam pressure controller to MANUAL and depress the INCREASE pushbutton.

Proposed Answer:

-C- - -

Explanation (Optional): To bypass low-low Tav interlock both interlock bypass switches must be taken to bypass ("C" correct). Taking the mode switch to reset resets the load rejection C-7 arming memory (" A" wrong, but plausible). Taking the bypass interlock select switches to off/reset position will block steam dump operation (liB" wrong, but plausible). Shifting steam pressure controller to manual allows adjusting steam dump valve position if they are not blocked ("D" wrong, but plausible).

Technical Reference(s): Functional Dwg 10 (No. J)

(Attach ifnot previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None

---

~~----------------

Learning MC-05630 Describe the operation ofthe following steam dump system controls and (As available)

Objective: interlocks: a. Steam dump mode selector switch b. Steam dump interlock selector switch c. Steam dump steam pressure controller d. C-9 control interlock e. P-4 permissive interlock f. C-7 control interlock g. P-12 permissive interlock Question Source: Bank # 73214 Question History:

Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 41.4,41.5 and 41.7 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 43 Tier # 2 2 Main Feedwater: Group #

Physical connections/cause-effect between MFW and KIA # 059.K1.04 Steam Generator Water Level Control Importance Rating 3.4 3.4 Proposed Question:

With the plant initially at 100% power, Main Feedwater Header Pressure starts to decrease due to a significant feed header leak upstream of the feed flow venturis.

Prior to operator action, how will Feed Pump speeds and the Feed Regulating Valve positions initially respond to the decreasing pressure?

a) Feed Pump speed will INCREASE. SGWLC will throttle the Feed Regulating Valves in the OPEN direction.

b) Feed Pump speed will INCREASE. SGWLC will throttle the Feed Regulating Valves in the CLOSED direction.

c) Feed Pump speed will DECREASE. SGWLC will throttle the Feed Regulating Valves in the OPEN direction.

d) Feed Pump speed will DECREASE. SGWLC will throttle the Feed Regulating Valves in the CLOSED direction.

Proposed Answer:

-A- -

Explanation (Optional): As pressure drops, less feedwater enters the SGs. This causes a feed flow-steam flow mismatch, causing Feed Reg Valves to throttle open ("B" and "D" wrong). This further lowers feed pressure.

As the pressure sensed by 3FWS-PT508 decreases, the sensed DP between steam pressure and feed pressure decreases. The Master Speed controller increases pump speed to restore normal DP across thc feed regulating valves ("A" correct, "C" wrong. Actual feed pressure will increase as the feed pumps speed up, increasing actual feed flow. After a lag, which allows for the effects of shrink and swell to dissipated, the level error circuit will respond to the lower SG level (due to the initial decrease in feed flow) and also send a signal to open the feed regulating valves. "B", "C", and "D" are plausible, since the feed header pressure inputs to feed pump speed and SGWLC, and a transient is in progress.

Technical Reference(s): Functional Drawing 13 (No. 1)

(Attach if not previously providcd)

(including version/revision number)

Proposed references to be provided to applicants during examination: -.:..N.;.;:o~n:.:e_ _ _ _ _ _ _ _ __

Leaming MC-04664 Given the following failures (partial or complete) ofthe Main Feedwater & (As available)

Objective: Steam Generator Water Level Control Systems, DETERMINE the effects on the system

& on interrelated systems ... Main Feed System Pressure Transmitter (PT508) ...

Question Source: New Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 41.4,41.5, and 41.7 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 44 Tier # 2 2 Auxiliary! Emergency Feedwater: Group # I Knowledge of design features or interlocks which KIA # 061.K4.08 provide for AFW recirculation Importance Rating 2.7 -=.2.:.:..9_ _ __

Proposed Question:

The crew is shifting all AFW pump suctions from the DWST to the CST in accordance with GA-4, "Transfer AFW Pump Suction and Fill DWST".

When the lineup is complete, to where are the AFW Pump recirculation flowpaths aligned?

a) The recife paths arc isolated while suction is aligned to the CST.

b) The reeirc paths are directed to the CST.

c) The recirc paths are directed to the DWST.

d) The recirc paths are directed to the suctions of the AFW pumps.

Proposed Answer: C Explanation (Optional): "C" is correct, and "A", "S", and "D" wrong, since the AFW pump recirculation path is always aligned to the DWST via manual valves. This path is not changed when realigning suction to the CST. "A" is plausible, since the CST is the alternate, non-safety suction source. "B" is plausible, since the suction has been aligned to the CST. "D" is plausible, since numerous pumps, such as RHR pumps, have their recirc paths aligned back to their suctions.

Technical Reference{s): EOP 3503 (Rev. 015), Note prior to step 18 (Attach if not previously provided) P&ID 130B (No. 43)

(including version/revision number)

Proposed references to be provided to applicants during examination: _N...;o..,..n_e_ _ _ _ _ _ _ _ __

Learning MC-04638 DESCRIBE the following operations of the Auxiliary Feedwater System (As Objective: under the following normal, abnormal, & emergency conditions ... Shifting Auxiliary available)

Feedwater Pump (Motor & Turbine Driven) suction between the DWST & the CST ...

Question Source: Bank # 64300 Question History:

Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 55.41.7 and 41.8 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 45 Tier # 2 2 AC Electrical Distribution: Group # I Ability to monitor automatic operation, including safety- KIA # 062.A3.05 related indicators and controls Importance Rating 3.5 3.6 Proposed Question:

With the plant operating at 100%, a "brown-out" condition on the grid results in MP3 bus voltages being supplied at 80% for 5 minutes.

How will the 6.9KV and 4.16KV busses be affected?

a) A116.9KV buses to remain energized at the low voltage. The norma14.16KV buses remain energized at the low voltage, while the emergency 4.l6KV buses are powered from the RSSTs.

b) AIl6.9KV buses to remain energized at the low voltage. The norma14.16KV buses remain energized at the low voltage, while the emergency 4.16KV buses arc powered from the Emergency Diesels.

c) All 6.9KV buses de-energize. The normal4.l6KV buses dc-energize, while the emergency 4.16KV buses are powered from the RSSTs.

d) All 6.9KV buses de-energize. The normal4.16KV buses de-energize, while the emergency 4.16KV buses are powered from the Emergency Diesels.

Proposed Answer: B Explanation (Optional): The UV signal for the 6.9 KV buses is voltage less than 70% for 1.0 seconds, at which point thc load breakers from the affected bus trip. Since voltage is at 80%, they will not de energize

("C" and UD" wrong, but plausible). Undervoltage to lockout NSST for the 4KV buses is also 70% so the NSST Supply breakers will not open. However, due to brownout conditions (<90% voltage) without SIS or CDA present, after 4.5 minutes the 4160 V bus tie breakers will trip, the EDGs will start. If voltage on the emergency busses degrades to <30%, and RSST voltage is ~~97%, the RSST supply breakers will attempt to close; but since RSST voltage is less than 97%, fast and slow transfer will not occur ("A" wrong, but plausible). With the RSST breaker failing to close, emergency bus voltage will drop to less than 30%, and after 5 more seconds, the EDG breakers will close to reenergize the bus (UB" correct).

Technical Reference(s): OP 3353.MB8A (Rev. 002-10), 2-2,2-9,3-12 and 5-12 (Attach ifnot previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None

~~--------------------

Learning MC-05023 Describe the 4 kV ~i~tribution System operation under normal, (As available)

Objective: abnormal, and emergency condItIOns ... Loss ofNSSA...

Question Source: Bank # 68084 Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 41.7 and 41.8 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 46 Tier # 2 2 AC Electrical Distribution: Group #

Knowledge of abnormal condition procedures KIA # 062.GEN.2.4.11 Importance Rating 4.0 4.2 Proposed Question:

While at 100% power, the following sequence of events occurs:

1. Emergency bus 34D receives a Bus Differential and deenergizes.

2. The crew enters AOP 3577, Loss a/Normal and Offsjte Power To A 4.16KV Emergency Bus.

3. AOP 3577 directs the crew to shift the seal return flowpath to the top of the VCT.

What does shifting seal return to the top of the VCT prevent?

a) An unmonitored heatup of the charging pump suction water, which is supplied to the RCP seals.

b) An uncontrolled pressure increase in the VCT due to loss of cooling to the Letdown Heat Exchanger.

c) A heatup in the VCT, resulting in cavitation when a charging pump is subsequently started.

d) A heatup in the VCT, resulting in decreased cooling to the CVCS Regenerative Heat Exchanger.

Proposed Answer: -::..:.A~_

Explanation (Optional): "A" is correct, since seal return is normally directed to the suction ofthe charging pumps, and with the loss of bus 34D, cooling is lost to the seal return HX. This results in hotter water flowing to the charging pump suction and to the RCP seals, with the inability to monitor actual seal injection temperature, which is normally monitored via VCT temperature. Directing return to the top of the VCT allows for monitoring seal injection temperature using 3CHS-TII16 at MB3. HB" is wrong, but plausible, since a loss of cooling to the letdown HX occurs with a loss of bus 34C. "C" and "D" are plausible, since switching the heated seal return to the top of the VCT will cause VCT to heatup faster, and VCT water is supplied to the Charging Pump suction and the Regenerative Heat Exchanger.

Technical Referenee(s): AOP 3577 (Rev. 001-03), step 8 (Attach ifnot previously provided) AOP 3577 Basis Document (Rev. 001), step 8 basis (including version/revision number)

Proposed references to be provided to applicants during examination: None

--~~------------------

Learning MC-07396 Discuss the basis of major precautions, procedure steps, and/or step sequence (As Objective: in AOP-3577, Loss Of Normal And Offsite Power To A 4.16KV Emergency Bus. available)

Question Source: Bank # 80558 Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 41.3, 41.5, and 41.1 0 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 47 Tier # 2 2 DC Electrical Distribution: Group #

Physical connections/cause-effect between DC System KIA # 063.Kl.03 and battery charger and battery Importance Rating 2.9 3.5 Proposed Question:

With the plant initially at 100% power with all electrical systems aligned normally, Battery Charger 1 output breaker inadvertently trips open.

What voltage will the BOP operator observe on DC Bus I?

a) 0 volts b) 125 volts c) 133 volts d) 140 volts Proposed Answer:

-B- -

Explanation (Optional): "B" is correct, and "A" is wrong, since the DC bus will still be energized by the battery, which puts out 125 volts. "C" is wrong, but plausible, since this is the voltage put out by the battery charger, which the bus normally indicates. liD" is wrong, but plausible, since this is the voltage put out by the rectifier from the 480 vac bus to the inverter. This voltage is prevented from flowing back to the battery bus by a reverse biased diode.

Technical Reference(s): EE IBA (No. 29)

(Attach ifnot previously provided) OP 3345C (Rev. 016-04), Sections 1.1,2.1.2, and 2.1.3 (including version/revision number)

Proposed references to be provided to applicants during examination: -.::..N;.;;,o~n:.:.e_ _ _ _ _ _ _ _ __

Learning MC 03309 Given a failure of the 125 VDC distribution system or a portion of the (As available)

Objective: system, determine the effects on the system and on interrelated systems a. Loss of de bus effect on control systems b. Loss of dc bus effect on VIAC Question Source: New Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 41.8 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 48 Tier # 2 2 DC Electrical Distribution: Group # 1 1 Knowledge of design features or interlocks which KIA # 063.K4.02 provide for breaker interlocks and cross-ties Importance Rating 2.9 3.2 Proposed Question:

What is the purpose of the Kirk Key Interlock associated with Battery Charger 8?

a) To prevent tying an "A" Train 480V MCC with a "B" Train MCC through the Swing Charger 8 AC input breakers.

b) To prevent tying two "B" Train 480V MCC's together through the Normal Charger 2 and Swing Charger 8 AC input breakers.

c) To prevent tying an "A" Train DC Bus with a "B" Train DC Bus through the Swing Charger 8 output breakers.

d) To prevent tying two "B" Train DC Busses together through the Swing Charger 8 output breakers.

Proposed Answer: --,,=-D_ _

Explanation (Optional): Kirk key interlocks are provided for swing battery chargers 7,8,9 to prevent cross connecting the following 125 VDC buses:

Charger 7: Bus 1 (301A-l) and Bus 3 (30IA-2)

Charger 8: Bus 2 (3018-1) and Bus 4 (3018-2)

Charger 9: Bus 5 (301 C-l) and Bus 6 (3010-1)

The Kirk key allows only one swing charger output breaker to be closed at a time ("A" and "B" wrong),

preventing it from electrically cross-tying two "B" Train DC Busses "c" wrong, "D" correct). "A" and "B" are plausible, since all chargers have AC input breakers. "C" is plausible, since the Kirk key interlocks prevent tying DC busses together.

Technical Reference(s): OP 3345C (Rev. 016-04), Section 1.2 (Attach if not previously provided) EE-! BA (No. 29)

(including version/revision number)

Proposed references to be provided to applicants during examination: None

~~~------------------

Learning MC-03306 Describe the operation of 125 VDC distribution system controls and (As available)

Objective: interlocks a. Standby charger Kirk Key interlocks...

Question Source: Bank # 68199 Question History:

Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 41.7 and 41.8 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 49 Tier # 2 2 Emergency Diesel Generator: Group # 1 Physical connections/cause-effect between EDG and AC KIA # 064.KLOI Importance Rating 4.1 4.4 Proposed Question:

With the plant initially at 100% power, the following sequence of events occurs:

1. VIAC-l de-energizes.

2. On the resulting transient, the following events occur:

• The reactor trips,

• Safety Injection actuates,

• Offsite power is lost.

3. The crew enters E-O, Reactor Trip or Safety Injection.

Without operator action, how will the "A" Emergency Diesel Generator (EDG) respond?

a) The "A" EDG will NOT automatically start. Bus 34C will remain de-energized.

b) The "A" EDG starts, but Bus 34C will NOT strip, and the EDG output breaker will NOT automatically close. Bus 34C will remain de-energized.

c) The "A" EDG starts, but Bus 34C will NOT strip. The EDG output breaker will automatically close onto a loaded Bus 34C.

d) The "A" EDG starts, Bus 34C strips, and the EDG output breaker automatically closes onto Bus 34C.

But Bus 34C loads will NOT automatically sequence onto the bus.

Proposed Answer: _B_ _

Explanation (Optional): Loss ofVIAC-J or 2 deenergizes the associated EDG sequencer. The loss ofa sequencer with SIS actuating results in the following:

The associated diesel will not start (except LOP) (OIA" wrong, since LOP occurred, but plausible, since the EDG would not start on SIS). With a de-energized sequencer, bus stripping will not occur ("C" plausible),

and the associated train's load sequencing will not occur ("0" plausible). Since the sequencer normally sends a "Bus stripped" auto-dose permissive signal to the EDG output breaker, the EDG output breaker will not auto-close ("B" correct, "c" and "D" wrong).

Technical Reference(s): AOP 3564 (Rev. 009-03), Caution prior to step 1 (Attach ifnot previously provided) LSK-24-9.4A (No. 12)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning -------------------------

MC-04417 Given a failure (partial or complete) of a emergency diesel load (As available)

Objective: sequencer, determine the effects on the system and on interrelated systems.

Question Source: Bank # 69225 Question History:

Question Cognitive Level: Comprehension or Analysis J0 CFR Part 55 Content: 41.7 and 41.8 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 50 Tier # 2 2 Emergency Diesel Generator: Ability to manually Group #

operate and/or monitor need for/consequences of manual KJA# 064.A4.l0 load shedding on safeguards bus Importance Rating 3.3 3.4 Proposed Question:

Current Conditions:

• The reactor has tripped due to a Loss of All AC Power.

• The crew is performing the actions ofECA-O.O, Loss ofAll AC Power.

• Initial actions to restore AC power have been unsuccessful.

What actions arc required to be taken by the crew with 4KV Bus 34C breakers that supply power to pumps?

a) The control switches are verified to be in Auto-After-Stop (except for the Charging Pump), to allow the sequencer to respond to the LOP when power is restored. The Charging Pump is placed in Pull-To-Lock to prevent thermal-shocking the RCP seals when power is restored.

b) The control switches are verified to be in Auto-After-Stop (except for a Service Water Pump), to ensure RCS makeup is promptly restored when power is restored. The Service Water Pump is placed in Pull To-Lock to prevent overloading the EDG when power is restored.

c) The control switches are placed in Pull-to-Lock (except for the Charging Pump), to prevent a potential overload ofthe power source when power is restored. The Charging Pump is left in Auto-After-Stop to provide core cooling when power is restored.

d) The control switches are placed in Pull-to-Lock (except for a Service Water Pump) to prevent a potential overload ofthe power source when power is restored. The Service Water Pump is left in Auto-After Stop to provide EDG cooling when power is restored.

Proposed Answer: -:;.D_ _

Explanation (Optional): Equipment is placed in Pull-To-Lock ("A" and "B" wrong) to permit the operator to evaluate the status of the restored bus and sequence loads onto the bus consistent with plant conditions. liD" is correct, and "C" wrong, since one service water pump per train is left in AUTO to provide EDG cooling.

"A" and "B" are plausible, since normally, the preferred switch position is Auto-After-Stop, to ensure prompt loading on an LOP, and preventing thermal shock to the seals is a reason for placing the CHS pump in Pull To-Lock. "C" is plausible, since restoring makeup to the core is a major concern in ECA-O.O.

Technical Reference(s): ECA-O.O (Rev. 022-02), step 8 (Attach if not previously provided) WOG Bkgd Doc (Rev. 2), for ECA-O.O, step 8 (including version/revision number)

Proposed references to be provided to applicants during examination: -..:..N~o~n~e_ _ _ _ _ _ _ _ __

Learning (As available)

Objective: MC-03851 Describe the major action categories within EOP 35 ECA-O.O.

Question Source: Bank # 78405 Question History:

Question Cognitive Level: Memory or Fundamental Knowledge IO CFR Part 55 Content: 41.10 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 51 Tier # 2 2 Process Radiation Monitoring: Group # --"-1_ _ _ __

Predict and/or monitor changes in radiation levels KJ A # 073.A 1.01

~~-------------

associated with operating controls Importance Rating 3.2 3.5 Proposed Question:

With the plant at 100% power and all equipment operating normally, the following sequence of events occurs:

L The RO commences adjusting the Process Radiation Monitor 3HVR-REI2 (Degassifier area) ALARM setpoint at the RMS Console.

2. The RO adjusts the setpoint too low, bringing in its HI ALARM.

Which area radiation monitor will show an increasing trend due to changing area radiation levels caused by the 3HVR-RE12 ALARM setpoint adjustment?

a) 3RMS-RE07 (Aux 66', Calibration room area).

b) 3RMS-RE16 (Aux 43', VCT and boric acid tank area).

c) 3RMS-RE13 CAux 24', Heat exchanger area).

d) 3RMS-REIO CAux 04', Resin discharge pipe chase area).

Proposed Answer: B Explanation (Optional): "B" is correct, since the automatic action resulting from the alarm on 3HVR-REI2 causes 3CHS* AOV71 to divert letdown flow to the VCT (away from the degassifier). Radioactive gasses are no longer stripped out ofthe letdown stream, allowing radioactive gasses to accumulate in the VCT gas space, increasing area radiation. "A", "C", and ltD" are plausible, since these area monitors are also in the Auxiliary Building.

Technical Reference(s): AOP 3573 (Rev. 018-01), Att. A, pg 8 (Attach ifnot previously provided) AOP 3573 (Rev. 018-01), Att. B, pg 3 (including version/revision number)

Proposed references to be provided to applicants during examination: -:;..N.;.;;o:.;;;n:.;;;e_ _ _ _ _ _ _ _ __

Learning MC-05472 Given a failure of the Radiation Monitoring System (partial or (As available)

Objective: complete), determine the effects on the system and on inter-related systems.

Question Source: Bank # 75621 Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 41.5,41.11 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 52 Tier # 2 2 Service Water: Group #

Knowledge of power supplies for SWP ESF-actuated KiA # 076.K2.08 MOVs Importance Rating 3.1 3.3 Proposed Question:

A CDA actuates, and the RO reports that Service Water to RSS Heat Exchanger Inlet Isolation Valve 3SWP"'MOV54A did not stroke open.

The crew dispatches a PEO to check the power supply to the valve.

To which MCC is the PEO dispatched?

a) MCC 32-lT b) MCC 32-3T c) MCC 32-4T d) MCC 32-5T Proposed Answer: --",C_ _

Explanation (Optional): "D" is correct, since the power supply to 3SWP*MOV54A is MCC 32-4T. "A",

"B", and "C" are plausible, since each of these are MCCs powered from load center 32T. "A" is wrong, since 32-lT supplies loads in the Emergency Diesel Enclosure. "B" is wrong, since 32-3T supplies loads in the Turbine Building. "D" is wrong, since 32-5T supplies loads in the Intake Structure.

Technical Reference(s): OP 3326-023 (Rev. 007-04), page 3 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning MC-05714 Describe the operation of the following Service Water System (As available)

Objective: components controls and interlocks ...

Question Source: New Question History:

Question Cognitive Level: Memory or Fundamental Knowledge 10 CPR Part 55 Content: 41.8 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 53 Tier # 2 2 Service Water: Group # I I Knowledge of design features or interlocks which KIA # 076.K4.06 provide for Service Water train separation Importance Rating 2.8 ___

3._2_ _ __

Proposed Question:

How is Service Water System train separation maintained at the point in the system where both trains supply the RPCCW Heat Exchangers?

a) The two SWP trains are physically independent, with no interconnecting piping.

b) The interconnections between the two SWP trains automatically isolate on a Safety Injection signal.

c) The interconnections between the two SWP trains have check valves that maintain separation between the two trains.

d) The interconnections between the two SWP trains use normally-closed manual isolation valves to maintain separation.

Proposed Answer: _D_ _

Explanation (Optional): The two trains are for the most part physically separate ("A" plausible). The interconnecting portions include RPCCW and AFW, in which the interconnecting piping is kept separate via closed manual isolation valves CD" correct; "A", "B", and "c" wrong); TPCCW, which isolates on an LOP signal ("B" plausible), and Post Accident Sampling, which is kept separate by check valves ("C" plausible).

Technical Reference(s): P&ID 133B (No. 74)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: -'-N...;;o_n_e_ _ _ _ _ _ _ _ __

Learning MC-O 8718 Describe the operation of the Service Water System under the following normal, (As Objective: abnormal, and emergency conditions ... Loss of Offsite Power ... Safety Injection ... available)

Question Source: New Question History:

Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 41.8 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 54 Tier # 2 Instrument Air: Group #

Knowledge of design features or interlocks which KIA # 078.K4.01 provide for manual/auto tmnsfer of control Importance Rating 2.7 2.9 Proposed Question:

With the plant initially at 100% power, the following sequence of events occurs:

1. The plant trips due to a loss of offsite power.

2. The crew enters ES-O.1, Reactor Trip Response.

3. The crew resets LOP at MB2 per ES-O.l direction.

4. The crew manually starts an Instrument Air Compressor at MB 1.

Which Instrument Air Compressor did the RO manually start, and would the Air Compressor have started if the crew failed to reset LOP at MB2?

a) The RO started the "A" lAS compressor. The compressor would NOT have started if the LOP signal had not been reset at MB2.

b) The RO started the "A" lAS compressor. The compressor WOULD have started even if the signal had not been reset at MB2.

c) The RO started the "B" lAS compressor. The compressor would NOT have started if the LOP signal had not been reset at MB2.

d) The RO started the "B" lAS compressor. The compressor WOULD have started even if the LOP signal had not been reset at MB2.

Proposed Answer: D Explanation (Optional): On an LOP, the EDGs will re-energize the emergency busses. The "A" instrument air (lAS) compressor will not have power available, since it is powered from non-emergency bus 32P ("A" and "B" wrong). The "B" lAS compressor has power ("B" plausible), but its breaker tripped on the LOP signal. ES-O.I directs the crew to close the "B" lAS compressor breaker at MB 1 to manually restore lAS header pressure. Resetting LOP is not necessary, since the MB2 LOP reset allows manually stopping loads.

The manual start block (HC" plausible) clears automatically 40 seconds after the EDG energizes the bus, and this time has passed well before reaching the step in ES-O.l ("D" correct, "C" wrong).

Technieal Reference(s): ES-O.1 (Rev. 024), steps 3.d and 3.h.l (Attach if not previously provided) OP 3332A-004 (Rev. 004-03), page 2 (including version/revision number) LSK 12-lE (No.7); 24.9.4.A (No. 12); 24.904.B (No. 12); 24.904.P (No. 10)

Proposed references to be provided to applicants during examination: -::..N;.,:o.:,;n;.:,e_ _ _ _ _ _ _ _ __

Learning MC-05323 Describe the operation of plant air systems under the following normal, (As available)

Objective: abnormal, and emergency operating conditions ... Loss of offsite power (LOP) ...

Question Source: New Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 4104,41.7, and 41.l0 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 55 Tier # 2 2 Containment: Group #

Ability to manually operate andlor monitor ESF slave KIA # 103.A4.03 relays Importance Rating 2.7 2.7 Proposed Question:

Current Conditions:

• The crew is performing SP 3646A.8, Slave Relay Testing - Train A.

• The crew is perfonning a pre-job brief before performing step 4.1, "Containment Isolation Phase A S804 - Relay K624 - Continuity Check."

• The RO informs the crew that Relay K624 feeds CIA Valve 3CHS*MV8112.

Will 3CHS*MV8112 stroke during this test; and if an actual CIA signal is received during this test, will Slave Relay K624 respond to the actual CIA signal?

a) 3CHS*MV8112 will NOT stroke during this test. The Slave Relay WILL respond to an actual CIA.

b) 3CHS*MV8112 will NOT stroke during this test. The Slave Relay will NOT respond to an actual CIA.

e) 3CHS*MV8112 WILL stroke during this test. The Slave Relay WILL respond to an actual CIA.

d) 3CHS*MV8112 WILL stroke during this test. The Slave Relay will NOT respond to an actual CIA.

Proposed Answer:

-B- -

Explanation (Optional): The test checks continuity of the slave relay, but does not actually operate the associated valve ("C" and "D" wrong). "C" and "D" are plausible, since Tech Spec Acceptance Criteria for a slave relay test requires the energization of each slave relay and verification of operability of each relay.

Also, "Go" testing of slave relays actually operates components. "B" is correct, and "A" wrong, since the slave relay is blocked from responding to actual signals during continuity testing. "AU is plausible, since during some testing, such as Sequencer Test 1 testing, the equipment gets automatically removed from the Test Mode to respond upon receipt of an actual signal.

Technical Reference(s): SP 3646A.8 (Rev. 023-05), Note prior to step 4.1.3 (Attach if not previously provided) SP 3646A.8-001 (Rev. 014), page 2 of2 (including version/revision number)

Proposed references to be provided to applicants during examination: -=..N.:. .;o;.; ;n;.; ;e_ _ _ _ _ _ _ _ __

Learning MC-05497 Describe the operation of the RPS under the following normal, abnormal, (As available)

Objective: and emergency conditions ... Slave Relay Testing 1. Go Testing 2. No Go Testing ...

Question Source: New Question History:

Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 41.7 and 41.9 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 56 Tier # 2 2 Pressurizer Level Control: Group # 2 2 Knowledge of the effect on PLC of a loss or malfunction KIA # 0l1.K6.04 of pressurizer level controllers Importance Rating 3.1 3.1 Proposed Question:

With the plant initially at 100% power, the following sequence of events occurs:

1. The crew commences a downpower using OP 3204, At Power Operations.

2. The pressurizer level controller 3CHS-LK459 level controller setpoint sticks at its original value.

3. The PZR Level Controller is left in AUTOMATIC.

How will Charging Flow Control Valve 3CHS*FCY121 respond throughout the course of the downpower?

a) 3CHS*FCY121 will throttle closed and letdown will isolate on low pressurizer level.

b) 3CHS*FCY121 will remain at the same position and maintain PZR level at 64%.

c) 3CHS*FCY121 will throttle open and maintain PZR level at 64%.

d) 3CHS*FCY121 will open fully and the reactor will trip on high pressurizer level.

Proposed Answer: C Explanation (Optional): As Thot decreases on the downpower, the RCS water becomes more dense and PZR level will tend to decrease ("B" wrong). "B" is plausible, since Pzr level setpoint is being maintained constant. As power is reduced, the charging flow control valve will throttle open (" A" wrong) to maintain 64% level ("C" correct and "D" wrong). "A" and "D" are plausible, since the pressurizer level controller has malfunctioned with a transient in progress.

Technical Reference(s): Functional sheet 11 (No. H)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None

--~~--~-------------

Learning MC-05342 Given a failure, partial or complete, ofthe Pressurizer Pressure and Level (As available)

Objective: Control System, determine the effects on the system and on interrelated systems.

Question Source: Bank # 73553 Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 41.8,41.10 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 57 Tier # 2 2 Nuclear Instrumentation: Group # 2 2 Ability to interpret control room indications and KIA # 015.GEN.2.2.44 understand how operator actions affect system and plant Importance Rating 4.2 4.4 Proposed Question:

With the plant at 100% power, the following sequenee of events occurs:

1. NIS Channel N41 fails high.

2. The operators take prompt action to place rod control in MANUAL.

3. The crew enters AOP 3571, Instrument Failure Response.

4. The crew is preparing to operate the switches on the NIS drawers per AOP 3571.

Which of the switch manipulations directed by AOP 357 I is/are physically required in order to restore proper manual and automatic operation of rod control?

a) The "Comparator Channel Defeat" switeh must be taken to Channel "N41" only.

b) The "Power Mismatch Bypass Switch" must be taken to Channel "N41" only.

c) Both the "Comparator Channel Defeat" switch and the "Rod Stop Bypass" switch must be taken to Channel "N41".

d) Both the "Power Mismatch Bypass" switch and the "Rod Stop Bypass" switch must be taken to Channel "N41".

Proposed Answer: --=.D_ _

Explanation (Optional): The "Power Mismatch Bypass" switch removes the faulty channel from the auctioneering input to rod control. The "Comparator Channel Defeat" switch compares all 4 NIS channels, and provides an annunciator if the highest reading channel differs from the lowest channel by >2% ("A" and "C" wrong). Since C-2 is a 114 coincidence, and deenergizes to actuate, the "Rod Stop Bypass" switch must also be operated to allow manual rod withdrawal CD" correct, "B" wrong). "A" and "C" are plausible, since this switch is also operated on a failed NIS channel. "A" and "B" are plausible since many coincidences are 2/4, in which case the "Rod Stop Bypass" switch would not have to be operated.

Technical Reference(s): AOP 3571 (Rev. 009-07), Att. D, pg 2 of7 (Attach if not previously provided) Functional sheets 4 (No. G) and 9 (No. H)

(including version/revision number) --=O-=-P-=3:..::3~5.::...3:.:...M;.:;;;B::.4.:..C;::....;:(R:.:e.:..v:.:..;.:;;;0~0.::...6-;.:;;;0.:..7L?),;.:;;;3-=-3=--_ _ _ _ _ _ _ _ _ _ _ _ _ __

Proposed references to be provided to applicants during examination: ~N:.:.o.=n:.:.e_ _ _ _ _ _ _ _ __

Learning MC-05225 Describe the operation of the Nuclear Instrumentation System Control and (As Objective: Interlocks ... Control Interlocks ... Power Mismatch Bypass ... Rod Stop Bypass ... available)

Question Source: Bank # 75600 Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 55.41.7 and 41.10 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 58 Tier # 2 2 Non-nuclear Instrumentation: Group # 2 2 Knowledge ofthe effect of a loss or malfunction ofNNI KJA# 0l6.K3.06 will have on AFW Importance Rating 3.5 3.7 Proposed Question:

Reactor Power is 2%, and the following initial conditions exist:

• The crew is placing the "A" TDMFP in service, and removing AFW from service.

• All SG Narrow Range Levels are 50%.

The following sequence of events occurs:

L "B" SG Narrow Range Level Channel 3FWS*LT527 fails to 0%.

2. The BOP operator takes the control switch for the "B" MDAFW pump to STOP.

How will the "B" MDAFW Pump respond?

a) The pump stops, and remains off when the BOP operator lets go of the switch.

b) The pump stops, and restarts as soon as the BOP operator lets go of the switch.

c) The pump remains running with the Amber breaker disagreement light LIT.

d) The pump remains running, with the Amber breaker disagreement light OFF.

Proposed Answer: .....;;...A,--_

Explanation (Optional): The failed "B" SG NR level channel is below the low-low level MDAFW Pp auto start setpoint (1/4 SGs). But, since the coincidence for AFW Pump Auto Start is 2/4 channels, and only one instrument has failed low, the pump will stop (HA" correct, "B", "C", and "D" wrong). "B", "C", and "D" are plausible, since on an actual low-low level on 1 /4 SGs, the pump cannot be stopped by the operator at the Main Board until Lo-Lo level is reset.

Technical Reference(s): Functional Sheets 7 (No. M) and 13 (No. J)

(Attach ifnot previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: . . .:. N.:. :;o; :;n:.; .e_ _ _ _ _ _ _ _ __

Learning MC-04635 DESCRIBE the operation of the following Auxiliary Feedwater System (As available)

Objective: component controls & interlocks: A. Motor Driven Auxiliary Feedwater Pumps B.

Turbine Driven Auxiliary Feedwater Pump ...

Question Source: Modified Bank Question #69635 Original Question Attached Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 41.7 Comments: Original Question #69635 on next page

Original Question #69635 The reactor has tripped from 100% power due to a fault on the "A" Steam Generator, and Current Conditions are as follows:

• SG NR levels:

• "A" SG: 0%

• "B" SG: 25%

• "c" SG: 31%

• "D" SG: 25%

• SIS has been reset.

• No other operator actions have been taken.

The BOP operator takes the control switch for the "B" MDAFW pump to STOP.

How will the "B" MDAFW Pump respond?

a) The pump stops, with the Amber breaker disagreement light LIT.

b) The pump stops, with the Amber breaker disagreement light OFF.

e) The pump remains running with the Amber breaker disagreement light LIT.

d) The pump remains running, with the Amber breaker disagreement light OFF.

Correet Answer = "D"

Examination Outline Cross-reference: Level RO SRO Question # 59 Tier # 2 2 Containment Iodine Removal: Group # 2 2 Knowledge of operational implications of the purpose of KIA # 027.K5.01 charcoal filters Importance Rating 3.1 3.4 Proposed Question:

The reactor has tripped, and the following complications exist:

• An RCS leak into CTMT has significantly changed the following parameters in the CTMT atmosphere:

• CTMT humidity has increased.

• CTMT iodine levels have increased.

• CTMT particulate levels have increased.

The crew enters FR-Z.3, Response to High Containment Radiation Level.

Per FR-Z.3 guidance, the ADTS is considering the use of the Containment Air Filtration (CAF) System to lower radiation levels in CTMT.

What is the greatest impact the elevated CTMT humidity level has on the effectiveness of the CAF System?

a) The HEPA filters will be less effective at reducing atmospheric particulate levels.

b) The HEP A filters will be less effective at reducing atmospheric iodine levels.

c) The charcoal filters will be less effective at reducing atmospheric particulate levels.

d) The charcoal filters will be less effective at reducing atmospheric iodine levels.

Proposed Answer:

-D- -

Explanation (Optional): The CAF system is most effective at removing iodine in the charcoal filters ("C" wrong) and particulates in the HEP A filters ("B" wrong). "D" is correct, and"A" wrong, since increasing humidity significantly reduces the effectiveness of the charcoal filters, but not HEPA filters (until 95%

humidity conditions are reached). "A", "B", and "c" are plausible, since the CAF System has both HEPA and charcoal filters, and removes both particulates and iodine.

Technical Reference(s): -"-FR;;..;;..;;-Z;;;';';:..3..lo.(R;;.;;e;:..v...;..-,,0..;;.O;:..5)~,;:..st:..:.e.l:.p-,,I_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

(Attach if not previously provided) ___ FS.;..A_R ______ 9A...;.'7___._l.2..:.(Rev...;..___

___ ___ 2...;.2.. .;.3... ) _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

(including version/revision number) www.novent.homestead.comlfileslcarbon.htm www.filt-air.comlResourcesiArticlesihepa/hepa filters.aspx Proposed references to be provided to applicants during examination: -"-N,...:o,:;.n;.;.e-:---:-_ _ _ _ _ _ __

Learning MC-04261 Describe the major administrative or procedural precautions and limitations (As available)

Objective: placed on the operation ofthe Containment Ventilation System, and the basis for each.

Question Source: New Question History:

Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 41.10 and 41.12 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 60 Tier# 2 Containment Purge: Group # 2 2 Knowledge of design features or interlocks which KJA# 029.K4.03 provide for automatic purge isolation Importance Rating 3.2 3.5 Proposed Question:

The plant is in MODE 5, with the CTMT Purge System (HVU) in operation.

Fuel Drop monitor 3RMS*RE42 goes into HIGH alarm.

What is the immediate effect ofthe alarm on the running CTMT Purge Exhaust Fan (3HVR-FN4A or 4B),

and on the purge supply valves (3HVU*CTV32A and 33A) and exhaust valves (3HVU*CTV32B and 33B)?

a) The CTMT Purge Exhaust Fan continues to run. One supply valve and one exhaust valve close.

b) The CTMT Purge Exhaust Fan trips. One supply valve and one exhaust valve close.

c) The CTMT Purge Exhaust Fan continues to run. All four supply and exhaust valves close.

d) The CTMT Purge Exhaust Fan trips. All four supply and exhaust valves close.

Proposed Answer: _A_ _

Explanation (Optional): High Radiation isolates the CTMT Purge Supply and Exhaust paths if EITHER fuel drop monitor goes into alarm. One monitor isolates the supply and return inside containment valves, and the other monitor isolates the outside containment valves ("C" and "D" wrong). The fans do not receive an auto trip signal ("A" correct, "B" wrong). "C" and "D" are plausible, since each of these isolates CTMT. "B" is plausible, since the fan is running without a suction path.

Technical Reference(s): AOP 3573 (Rev. 018-01), Attachment B, page 6 of6.

(Attach ifnot previously provided) P&IDs 148A (No. 40) and 153A (No. 28)

(including version/revision number)

Proposed references to be provided to applicants during examination: -:-N....;o-"n....;e_ _ _ _ _ _ _ _ __

Learning MC-04259 Describe the operation of the following Containment Ventilation System (As available)

Objective: controls and interlocks ... Containment Purge Air System ...

Question Source: Bank # 73071 Question History:

Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 41.7 and 41.11 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 61 Tier # 2 2 Steam Generator: Oroup# 2 2 Knowledge of the effect of a loss or malfunction of SOS KiA # 035.K3.01 on the RCS Importance Rating 4.4 4.6 Proposed Question:

A SO Tube Rupture occurs on the "C" Steam Generator, resulting in the following sequence of events:

1. The crew enters E-3, Steam Generator Tube Rupture.

2. The crew completes the steps that isolate the ruptured SO.

3. The crew was NOT able to close the "C" and "D" MSIVs.

What adverse effect would there be on the RCS if the crew continued on in E-3 with the MSIVs open?

a) After the RCS cooldown is completed, an excessive dilution of the RCS would occur.

b) While the RCS is being depressurized, a loss ofRCS subcooling would occur.

c) After the RCS depressurization is completed, the RCS would repressurize.

d) While the RCS is being cooled down, RCS conditions will reach RCP trip criteria.

Proposed Answer: -=.B_ _

Explanation (Optional): "B" is correct since the ruptured SG needs to be isolated from the intact SOs before decreasing the intact steam generator pressures, since this minimizes radiological releases and ensures RCS subcooling when primary to seeondary leakage is terminated in subsequent steps. "A" is wrong, since isolating the ruptured SO allows its pressure to remain higher than the pressures of the SOs being steamed to cooldown the RCS. The goal is to depressurize the RCS to the point where backflow will occur. "A" is plausible, since this is a basis for choosing ES-3.2 or 3.3 over ES-3.l if there is a concern about backflow.

"C" is wrong, since SI has not yet been terminated, so the RCS will re-pressurize. "C" is plausible, since this is the basis for terminating SI. "D" is wrong because during the cooldown, RCP trip criteria do not apply, but plausible, since RCS pressure will drop below RCP trip criterion pressure during the depressurization.

Technical Reference(s): WOO Bkgd Doc (Rev. 2) for E-3 Step 5 Caution (Attaeh if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: -=.. N. .;;o.;::n;.;;.e_ _ _ _ _ _ _ _ __

Leaming MC-04372 Discuss the basis of major procedure steps and/or sequence of steps in (As available)

Objective: EOP 35 E-3 Question Source: Bank # 65981 Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 41.5 and41.10 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 62 Tier # 2 2 Area Radiation Monitoring: Group # 2 2 Knowledge of operational implications of radiation KJA# 072.K5.01 theory, including sources, type, effects Importance Rating 3.0

-2.7 Proposed Question:

Which type of radiation is the Millstone 3 Area Radiation Monitors designed to detect?

a) Alpha radiation b) Beta radiation c) Gamma radiation d) Neutron radiation Proposed Answer: C Explanation (Optional): "c" is correct, and "A", "B", and "D" wrong, since the purpose of area radiation monitors is to monitor gamma radiation levels in various areas of the plant which are subject to changing radiological conditions. "A", "B", and "D" are plausible, since these are all types of ionizing radiation of concern at nuclear plants.

Technical Reference(s): Millstone 3 Radiation Monitor Manual (No Revision Number), page 83 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning MC-00165 Describe the funetion and location ofthe following Radiation (As available)

Objective: Monitors...

Question Source: New Question History:

Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 41.11 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 63 Tier # 2 2 Station Air: Group # 2 2 Predict impact/mitigate cross-connection with lAS KIA # 079.A2.01 Importance Rating 2.9 3.2 Proposed Question:

With the plant initially at 100% power, the following sequence of events occurs:

I. The RO reports that Instrument Air (lAS) header pressure is at 88 psig and decreasing.

2. The crew enters AOP 3562, Loss ofInstrument Air.

3. A PEO reports an air leak on an lAS branch line in the turbine building.

4. When lAS pressure reaches 85 psig, the following automatic actions occur:

• Service Air (SAS) to lAS Cross-Connect Valve 3IAS-AOV14 automatically opens.

• SAS Header Supply Valve 3SAS-AOV33 automatically closes.

5. lAS header pressure stabilizes at 85 psig.

6. The US directs the PEO to isolate the leak by closing the nearest isolation valve.

7. The PEO successfully isolates the leak.

How will Service Air header pressure respond now that the leak is isolated; and by what method can the PEO locally realign SAS to supply the Service Air header, if needed?

a) SAS header pressure will continue to decrease to zero, unless local manual actions are taken. The PEO can manually realign to supply SAS via switches at the local lAS annunciator panel.

b) SAS header pressure will continue to decrease to zero, unless local manual actions are taken. The PEO can manually realign to supply SAS by isolating air to 3IAS-AOVI4 and 3SAS-AOV33.

c) SAS header pressure will recover, since 3IAS-AOVI4 and 3SAS-AOV33 automatically realign when lAS pressure reaches 103 psig. If the valves fail to automatically reposition, the PEO can manually realign to supply SAS via switches at the local lAS annunciator panel.

d) SAS header pressure will recover, since 3IAS-AOV14 and 3SAS-AOV33 automatically realign when lAS pressure reaches 103 psig. If the valves fail to automatically reposition, the PEO can manually realign to supply SAS by isolating air to 3IAS-AOVI4 and 3SAS-AOV33.

Proposed Answer: C Explanation (Optional): When lAS header pressure decreases to 85 psig, SAS realigns to supply lAS.

Additionally, when lAS header pressure increases to 103 psig, the AOV's automatically realign to their normal positions, and SAS pressure will be maintained ("A" and "B" wrong). "A" and "B" are plausible, since maintaining lAS pressure is a higher priority than SAS pressure. "C" is correct and "D" wrong, since the capability exists to locally align the valves via switches at the lAS annunciator panel. "D" is plausible, since the cross connect valves are AOVs, with normal positions being to supply the SAS header.

Technical Reference(s): LSK-12-1C (No.6), 12-2C (No.8)

(Attach if not previously provided) OP 3353.1S (Rev. 001), 1-1 (including version/revision number)

Proposed references to be provided to applicants during examination: None

--~~------------------

Learning MC-05323 Describe operation of plant air systems under the following normal, (As available)

Objective: abnormal, and emergency operating conditions ... Low instrument air pressure ...

Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 41.7 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 64 Tier # 2 2 Fire Protection: Group # 2 2 Knowledge of the operational implications of water KJA# 086.K5.03 spray on electrical components Importance Rating -=...3.:.:1_ _ __ 3.4 Proposed Question:

A fire breaks out in the "A" Train Switchgear room, and current conditions are as follows:

• The Shift Manager and fire brigade leader discuss the option of discharging a fire hose in the area.

• It is decided that if a fire hose is used, the fire brigade will use a spray pattern rather than a solid stream, to minimize the potential for electrocuting the fire brigade members.

If water were to be discharged in this area, what other immediate operational concern would exist?

a) This could make the area uninhabitable, complicating a shutdown outside the control room, which would be required if the fire causes spurious equipment actuations.

b) This could cause the loss of individual loads or the loss of an entire 4KV Bus, due to the creation of electrical short circuits.

c) This could increase the corrosion rate to the electrical components in the area.

d) This could incrcase the likelihood of are-flash ifthe fire is a "Class A" fire.

Proposed Answer: _B_ _

Explanation (Optional): "B" is correct, since water is a good conductor of electricity, potentially creating paths from energized electrical bus bars directly to ground. "A" is wrong, but plausible, since this is a concern if C02 were to be discharged into the area. "C" is wrong, but plausible, sinee this is a bigger concern with chemical extinguishing agents, and is a long-term, rather than immediate concern. "D" is wrong, but plausible, since this is a concern with Halon or C02, which do not remove the heat as effectively as water.

Teclmical Reference(s): "Common Fire Extinguishing Agents" (US Department of Labor Website)

(Attach if not previously provided) (www .osha. govI dpcl outreach traininglhtmlfiles/extagent.html)

(including version/revision number)

Proposed references to be provided to applicants during examination: _N-'o_n...;.e_ _ _ _ _ _ _ _ __

Learning MC-04620 Describe the major administrative or procedural precautions and (As available)

Objective: limitations placed on the operation of the Water Fire Protection (FPW) system, including the basis for each.

Question Source: New Question History:

Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 41.8 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 65 Tier# 2 2 Site Specific AMSAC: Oroup # 2 2 Knowledge of AMSAC design features or interlocks KIA # Site Specific.AMSAC.K4 Importance Rating Site Site Proposed Question:

With the plant initially at 65% power, the following sequence of events occurs:

T=O: Turbine Impulse Pressure Transmitter 3MSS-PT505 fails to ZERO.

T+3.0 minutes: A loss of main feedwater event occurs.

T+3.4 minutes: "A" SO level reaches the Lo-Lo level reactor trip setpoint T+3.4 minutes: The reactor does NOT trip.

T+3.5 minutes: Indicated SO NR levels stabilize at the following levels:

• SO "A": 12%

• SO "B": 16%

• SO "C": 20%

• SO "D": 13%

Assuming no operator action is taken, and SO NR level indications remain at their current levels, how will AMSAC respond to this event?

a) AMSAC will trip the reactor and start the AFW pumps in about 25 seconds.

b) AM SAC will trip the main turbine and start the AFW pumps in about 25 seconds.

c) AMSAC will NOT actuate because SO levels are above the AMSAC actuation setpoint.

d) AMSAC will NOT actuate because it will disarm before the actuation timer expires.

Proposed Answer: B Explanation (Optional): AMSAC will stay armed for 260 seconds after PT505 failed low and AMSAC actuation is delayed for only 25 seconds after levels go below 16.6% in 3 of 4 SOs ("C" wrong). "B" is correct, and "A" wrong, since AMSAC starts AFW and trips the turbine. "A" is plausible, since AMSAC is designed to mitigate a loss of feed ATWS event. "C" is plausible, since one SO level is above the RPS Lo-Lo level reactor trip setpoint, and another level is not significantly below the RPS SO Lo-Lo level setpoint. "D" is wrong because the timer for actuation is 25 seconds (SO level) which when added to the 210 seconds that have transpired, is still shorter than the 260 second disarming time. "D" is plausible, since Turbine Impulse Pressure failed below the AMSAC disarming pressure, and there are time delays associated with both dis arming and actuation.

Technical Reference(s): OP3350 (Rev. 006-04), Note 1 prior to step 4.2.1 (Attach if not previously provided) OP3350 (Rev. 006-04), Attachment 3 (including version/revision number)

Proposed references to be provided to applicants during examination: -::-N.:,;o:..::n:.:;e_ _ _ _ _ _ _ _ __

Learning MC-04086 Describe operation of AMSAC circuitry including the following: (As available)

Objective: a. Inputs b. Input/output logic c. Outputs Question Source: Bank # 72277 Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 41.7 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 66 Tier # 3 3 Ability to coordinate personnel activities outside the Group # 1 control room KIA # GEN.2.1.8 Importance Rating 3.4 4.1 Proposed Question:

The crew is preparing to place the tube side of an isolated HP Feedwater Heater in service per OP 3321, Main Feedwater, and initial conditions are as follows:

• The BOP operator has been designated as the activity coordinator, and has the master copy of the "Continuous Use" procedure in-hand in the control room.

• The BOP operator will be directing a PEO to perform several procedure steps in the Turbine Building.

The BOP operator directs the PEO to locally perform OP 3321, steps 4.18.2.d through 4.18.2.i.

In accordance with AD-AA-l 02, Procedure Use and Adherence, is the PEO required to have a copy ofthe procedure in-hand locally; and how will the BOP/PEO document the performance of the PEO's actions in the master copy ofthe procedure?

a) The PEO is NOT required to have a copy of the procedure in hand. The BOP CAN place-keep in the master procedure copy based on the PEO's report that a step has been performed.

b) The PEO is NOT required to have a copy of the procedure in hand. The BOP CANNOT place-keep in the master procedure copy for the PEO's actions.

c) The PEO IS required to have a copy of the procedure in hand. The BOP CAN place-keep in the master procedure copy based on the PEO's report that a step has been performed.

d) The PEO IS required to have a copy ofthe procedure in hand. The BOP CANNOT place-keep in the master procedure copy for the PEO's actions.

Proposed Answer: C Explanation (Optional): All personnel who are performing actions in a Continuous Use procedure and are not able to view the master copy shall have a copy of the procedure in hand ("A" and "B" wrong). "A" and "B" are plausible, since the control room routinely directs PEO actions in the plant. With multiple procedure copies in use, the assigned coordinator place-keeps in the master copy as acknowledgements from the other individuals are received that steps have been performed ("C" correct, "D" wrong). "D" is plausible, since, with one procedure performer, the performer is required to initial the procedure as the steps are performed.

Technical Reference(s): AD-AA-102 (Rev. 4), Section 3.8 (Attach ifnot previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning MC-06799 Outline the methods for updating the Master Copy of a document when (As available)

Objective: multiple users per performing tasks Question Source: New Question History:

Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 41.10 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 67 Tier # 3 3 Ability to locate control room switches, controls, Group #

and indications, and to determine they correctly KJA# GEN.2.1.31 reflect the desired plant lineup Importance Rating 4.6 4.3 Proposed Question:

After a mid-cycle reactor trip, a plant startup is in progress per OP 3203, Plant Startup, and current conditions are as follows:

• The plant is stable at 12% power.

• The oncoming BOP operator observes the following switch/indicator positions on MB5:

MB5 SwitchlIndication Position

• FW PUMPS P4 TRIP BYPASS Switch: NORMAL

• 3MSS-N07, Steam Dump "MODE SELl! Switch: STMPRESS

• Atm Relief Bypass 3MSS*MOV74A Lockout Switch (MB5R): LOCKOUT

• Feed Isolation Valve 3FWS*MOV35A Position Indication: GREEN Which switch position should the BOP operator report as "NOT expected" for current plant conditions?

a) The FW PUMPS P4 TRIP BYPASS Switch should be in BYPASS.

b) The Steam Dump "MODE SELl! Switch should be in T AVE Mode.

c) The Atm Relief Bypass Valve Lockout Switch should be in NORMAL.

d) The Feed Isolation Valve position indicator should indicate RED.

Proposed Answer: A Explanation (Optional): With the plant at 12% power, the "FW PUMPS P4 TRIP BYPASS" selector switch should be in BYPASS ("A" correct), since it is not placed in "NORMAL" until power is above 25% power.

This is a two position "NORMALIBYP ASS" selector switch, located on MB5, and is used to enable or bypass the Reactor Trip signal which trips the MFW Pumps. The Steam Dump "MODE SELl! Switch should be in the Steam Pressure Mode, since it is not placed in Tave Mode until power is above 15% ("B" is wrong, but plausible). The Atm Relief Bypass Valve Cutout Switch should be in LOCKOUT, since these BTP 9.5-1 Fire Safety cut out switches are normally in bypassed to prevent spurious operation in the event of a "hot short". These switches are switched to "Operate" prior to operating the valves, but the condenser steam dumps are in operation ("C" wrong, but plausible). The Feed Isolation Valve position indicator should indicate GREEN, since these valves are bypassed by the Feed Reg Bypass Valves, and are utilized to isolate the Main Feedwater Regulating Valves while feeding with the bypass valves. They are not opened until the crew shifts to the Main Feed Reg Valves at 25% power ("D" wrong, but plausible).

Technical Reference(s): OP 3203 (Rev. 019-11), steps 4.3.56,57, and 64 (Attach if not previously provided) OP 3204 (Rev. 017-13), step 4.1.14

---

~~~------~--~------------------------------------

(including version/revision number) __ G_A_-_2_6-"(R

__ ev_._0_0_l.l.-'),c_s_te..p_6 Proposed references to be provided to applicants during examination: -=..N.:. .:o:. : :n:. :e___________________

Learning Objective: MC-03384 Describe the major action categories contained within OP 3203. (As available)

Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 41.4,41.7, and 41.10 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 68 Tier # 3 3 Knowledge of the proeess for making changes to Group # 2 2 procedures KJA# GEN.2.2.6 Importance Rating 3.0 3.6 Proposed Question:

Current Conditions:

• The RO is commencing a surveillance procedure for the "A" SIH pump.

• The RO recognizes the flowrate listed in the procedure has not been updated based on an impeller modification made during the previous outage.

What action, if any, is the RO allowed to take to complete the surveillance?

a) Proceed with the surveillance. If the obtained flowrate ends up being within that specified in the procedure, sign off the surveillance as completed satisfactorily. Then initiate a procedure change.

b) Notify the Unit Supervisor. Obtain Engineering concurrence, and then with US permission, make a pen and-ink change to the procedure, Then proceed with the surveillance.

c) Notify the Unit Supervisor. Obtain concurrence from a second SRO, and then with US permission, make a pen-and-ink change to the procedure. Then proceed with the surveillance.

d) The surveillance is required to be stopped. Ensure the plant is in a safe condition, notify the Unit Supervisor, and initiate a procedure change.

Proposed Answer: -'=.D_ _

Explanation (Optional): When there is a procedure discrepancy, stop the work, ensure the plant is in a safe condition, inform supervision, and initiate a procedure change ("D" correct, "A", "B", and "c" wrong),

Continue the procedure when corrections have been made. "A" is plausible, since it requires flow to be within spec before completing the surveillance. "B" and "e" are plausible, since these both get supervision involved prior to changing the procedure.

Technical Reference(s): AD-AA-102 (Rev. 4), section 3.6.2 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning MC-05167 Describe actions to be taken if a procedure yields inadequate or (As available)

Objective: unexpected results Question Source: INPO Exam Bank Question History: 2000 Byron NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 41.10 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 69 Tier # 3 3 Knowledge of the process for controlling Group # 2 2 equipment configuration or status KJA# GEN.2.2.14 Importance Rating 3.9 4.3 Proposed Question:

An equipment deficiency necessitates a change to the normal configuration ofthe Radioactive Liquid Waste System.

What condition would allow the crcw to track this as an Alternate Plant Configuration?

a) The new alignment is NOT within the design ofthe Radioactive Liquid Waste System.

b) The new alignment is NOT addressed by OP 3335D, Radioactive Liquid Waste System.

c) The new alignment IS projected to be a long-term alignment.

d) The new alignment IS a complex alignment.

Proposed Answer:

-B- -

Explanation (Optional): An Alternate Plant Configuration is an alignment that is not considered to be the normal configuration ofa system, but is consistent with system design ("A" wrong). It is required when an equipment deficiency necessitates a change to the normal configuration and the change is not specifically addressed by plant procedures ("B" correct). They are intended as temporary, and should be returned to normal as soon as practicable ("C" wrong). They are also intended to be relatively simple configurations. If the configuration is complex, it should be addressed using routine operations such as an approved procedure

("D" wrong). "A", "C", and "D" arc plausible, since they are all involved in allowing an Alternate Plant Configuration.

Technical Reference(s): OP-AA-1500 (Rev. 5), Note prior to step 3.2.1 (Attach if not previously provided) OP-AA-1500 (Rev. 5), Note prior to step 3.3.5 (including version/revision number)

Proposed references to be provided to applicants during examination:

Learning MC-05079 Discuss the general documentation requirements for equipment (As available)

Objective: manipulation.

Question Source: New Question History:

Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 41.10 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 70 Tier # 3 Ability to use radiation monitoring systems, such as Group # 3 3 fixed radiation monitors and alarms, portable survey KJA# GEN.2.3.5 instruments, personnel monitoring equipment, etc. Importance Rating 2.9

-2.9- - -

Proposed Question:

An operator is in the Turbine Driven Auxiliary Feedwater Pump Room, and is preparing to perform a manual frisk prior to exiting the ESF building.

How far away from the surface being checked is the operator required to hold the probe; and what is the minimum count rate increase above background that the operator would be considered contaminated?

Probe distance Minimum countrate increase a) lh inch I 00 counts per second b) lh inch 200 counts per second c) 1 inch 100 counts per second d) I inch 200 counts per second Proposed Answer: A Explanation (Optional): Manual Frisking - Verify that the hand-held frisker is on the "xl" scale, and the background is less than 200 cpm. Pick up the probe and pass it slowly over your body, holding the probe one-half inch away from the surface being checked ("C" and "D" wrong). The probe must be moved very slowly; one to two inches per second. While frisking, watch the needle on the meter face and listen for the clicks. If you observe an increase of 100 counts per minute above background, you are considered contaminated and must contact HP ("A" correct, "B" wrong). "C" and "D" are plausible, since I to 2 inches per second is the speed that you are required to move the probe. HB" is plausible, since 200 counts per second is the maximum background level above which you are not allowed to frisk.

Technical Reference(s): Radiation Protection Manual 5.2.2 (Rev. 015), page 21 of 40 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning MC-05128 State the precautions which must be followed while using an RM-14 to (As available)

Objective: perform a frisk.

Question Source: New Question History:

Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 41.12 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 71 Tier# 3 3 Knowledge of radiological safety principles pertaining Group # 3 3 to licensed operator duties, such as containment entry, KJA# 12 fuel handling, access to locked high-radiation areas, etc. Importance Rating 3.2 3.7 Proposed Question:

The plant is in MODE 5, and preparations are being made for the initial Containment entry.

An operator will be required to enter the MIDS area inside Containment.

What requirements exist for entry into this area in addition to the normal requirements for entering a locked high radiation area?

a) The Incore System Drives must be tagged out.

b) An access control guard must be stationed at the locked entrance.

c) The access point locking mechanism must prevent personnel entry and exit.

d) A flashing light must be at the entry point to warn of an open gate or door.

Proposed Answer:

-A- -

Explanation (Optional): "A" is correct, since the Incore Drives must be tagged out. "B" is wrong, but plausible, since this is a requirement for both a locked high radiation and the MIDS area when work requires the lock to be defeated. "C" is wrong, since the lock must not prevent egress, but plausible, since this would be a requirement above that required for a locked high radiation area, and the exiting individual is required to wait at the exit until HP peer-checks the exit door locked. "D" is wrong, but plausible, since this is required for a locked high radiation area.

Technical Reference(s): RP-AA-201 (Rev. 5), Precaution 4.1.1 (Attach ifnot previously provided) RP-AA-201 (Rev. 5), steps 5.4.5,5.6.2, and 5.7.l.f.

(including version/revision number) RP-AA-201 (Rev. 5), Attachment 4, step 2 Proposed references to be provided to applicants during examination: -:;..N.;.;o:.;;.n:..;.e_ _ _ _ _ _ _ _ __

Learning MC-05135 List the requirements which must be met for entry into the Unit 3 (As available)

Objective: "MIDS" Very High Radiation Area Question Source: New Question History:

Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 41.12 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 72 Tier # 3 3 Ability to recognize entry conditions for EOPs/AOPs Group # 4 4 KJA# GEN.2.4.4 Importance Rating 3.7 4.7 Proposed Question:

A reactor startup is in progress in accordance with OP 3202, Reactor Startup.

Which plant condition meets a termination criterion that requires the crew to trip the reactor and enter E-O, Reactor Trip or Safety Injection?

a) Expected time of criticality is greater than I hour from the time estimated in the ECe.

b) The 11M plot indicates the reactor may go critical below the RIL.

c) A control rod does not move offthe core bottom along with its banle d) Sustained startup rate reaches 1.0 decades per minute in the Intermediate Range.

Proposed Answer: --:;..D_ _

Explanation (Optional): "D" is correct, and "A", "B", and "c" wrong, since the two termination criteria that require a reactor trip are: Sustained SUR of 1.0 dpm, and an uncontrolled cooldown that results in Tc being less than 530°F. "A" is plausible, since expected critical time not within I hour of time estimated in the ECC requires the crew to recalculate the ECe. "B" is plausible, since if calculations indicate reactor may go critical below the RIL, the crew needs to perform AOP 3566, Immediate Boration, and fully insert all Control Banks. "c" is plausible, since, when the Rod Banks are moved to I steps off the bottom, the crew checks ° Rod Position Indication, and if any control rods do not move off core bottom, the crew is directed to to recover the control rod.

Technical Referenee(s): OP 3202 (Rev. 021-01), Section 3.14 (Attach ifnot previously provided) OP 3202 (Rev. 021-0 I ), Step 4.24 (including version/revision number) --,,-O.;;..P...;:3.;;..2-,-0=.2",,(R:. ;:e.; . v.; . .,.;:.0.=2.=1--=0..::.1L!.),-=S..::.te:.. p;.. .4-=..;;..25'--_ _ _ _ _ _ _ _ _ _ _ _ _ __

Proposed references to be provided to applicants during examination: None Learning Objective: MC-04335 Identify plant conditions that require entry into EOP 35 E-O. (As available)

Question Source: New Question History:

Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 41.10 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 73 Tier # 3 3 Knowledge of EOP layout, symbols, and icons Group # 4 4 KJA# GEN.2.4.19 Importance Rating 3.4 4.1 Proposed Question:

The crew has entered the EOP network, and is currently performing a procedurally directed sub-step that is preceded by a "bullet" ( .).

In accordance with OP 3272, EOP User's Guide, what does the bullet signify?

a) The sub-step is part of an immediate action step.

b) The sub-steps may be performed in any order.

c) The sub-steps may require a procedure transition.

d) The sub-step is part of a continuous action step.

Proposed Answer:

-B- -

Explanation (Optional): "B" is eorrect, and "C" and "D" wrong, since a bullet signifies the steps may be performed in any order. "A" is wrong, but plausible, since immediate action steps are marked with an asterisk (*). "C" and "D" are plausible, since transition steps and continuous action steps are part of the EOP network.

Technical Reference(s): OP 3272 (Rev. 008-1 ] ), Attachment 2, sheet 2 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None

--~~~------------

Learning MC-04446 Describe the use and applicability of "notes" and "cautions" contained (As available)

Objective: within the emergency operating procedures network including transitions to another EOP procedure or step in the same EOP procedure.

Question Source: New Question History:

Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 41.10 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 74 Tier # 3 3 Knowledge of operational implications ofEOP Group # 4 4 warnings, cautions, and notes. KJA# GEN.2.4.20 Importance Rating 3.8 4.3 Proposed Question:

A large break LOCA occurs, and the crew addresses EOP Notes and Cautions as follows as they progress through the EOP network:

1. The crew enters E-O, Reactor Trip or Safety Injection.

2. The US reads the Note prior to E-O, step 1 silently, rather than out loud to the crew.

3. The crew transitions to E-l, Loss ofReactor or Secondary Coolant.

4. The US paraphrases the E-l Caution prior to step 1 about maintaining RCP seal injection, rather than reading it verbatim.

5. The US chooses to read the Caution prior to E-1, step 5 about paR V cycling silently, rather than out loud, since it does not apply to the event in progress.

6. After completing E-l, step 7, the crew transitions to ES-1.3, Transfer to Cold Leg Recirculation.

7. The crew transitions back to the "Procedure and Step in Effect," which is E-l, step 8 "Check if RHR Pumps Should 8e Stopped", without reading the Caution preceding step 8.

Which action violated the standards of OP 3272, EOP Users' Guide?

a) The US was required to read the Note prior to E-O, step lout loud.

b) The US was required to read the Caution about seal injection verbatim.

c) The US was required to read the Caution about PORV cycling out loud.

d) The US was required to read the Caution prior to E- 1, step 8.

Proposed Answer:

-D- -

Explanation (Optional): "A" is wrong, since the US is not required to read notes and cautions prior to immediate action steps out loud. "A" is plausible, since notes and cautions are normally read out loud. "8" is wrong, since the US is allowed to paraphrase notes and cautions. "8" is plausible, since EOP steps are normally read verbatim. "C" is wrong, since notes and cautions are required to be read out loud when they do not apply to the event in progress. "C" is plausible, since notes and cautions are normally read out loud. "D" is correct, since the US is required to read notes and cautions prior to the step at which they enter an EOP.

Technical Reference(s): OP 3272 (Rev. 008-11), Attachment 2, sheets 4 and 5 (Attach if not previously provided) E-O (Rev. 026), Cautions prior to steps 1 and 24 (including version/revision number) E-l (Rev. 024), Cautions prior to steps 1, 5, and 8 Proposed references to be provided to applicants during examination: None Learning MC-04446 Describe the use and applicability of "Notes" and "Cautions" contained (As available)

Objective: within the emergency operating procedures network including transitions to another EOP procedure or step in the same EOP procedure.

Question Source: New Question History:

Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 41.10 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 75 Tier # 3 3 Ability to diagnose and recognize trends utilizing Group # 4 4 appropriate control room reference material KJA# GEN.2.4.47 Importance Rating 4.2 4.2 Proposed Question:

A plant transient occurs with rod control in manual, and major parameters stabilize relative to their initial values as follows:

• Reactor Power (based on calorimetric): Higher than prior to the transient.

• Tave: Returned to its original value.

• MWe: Lower than prior to the transient.

What event has occurred?

a) A dilution of the RCS has occurred.

b) A Main Turbine ronback event has occurred.

c) Extraction steam has isolated to a Feed Heater.

d) A Main Steamline break has occurred.

Proposed Answer: --:;;;.C_ _

Explanation (Optional): "C" is correct, since MWe changing opposite of reactor power with Tave constant is indicative of a change in plant efficiency. Due to loss of efficiency, Tc will go down and T h will go up, but Tave will be constant. "A" is wrong, since with a dilution, MWe would be increasing. "B" is wrong, since with a turbine ronback, Reactor Power would be decreasing. "0" is wrong, since with a steam break, Tave would be decreasing. "A" is plausible, since reaetor power is increasing. "B" is plausible, since MWe are decreasing. "D" is plausible, since Reactor power is increasing, and MWe are decreasing.

Technical Reference(s): C06407C (Rev. 0) Training Event Diagnostic Flowchart (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: -.:. . N:.;:o.;;;n;;:..e_ _ _ _ _ _ _ _ __

Learning MC-05349 For given plant conditions, qualitatively state the effect of any RCS, (As available)

Objective: secondary plant or reactivity induced transient in any number of the 4 plant loops on the following parameters (RCP trip, turbine trip, dropped rod, etc.): reactor power, rod position, RCS loop average temperatures (affected and non-affected loops), RCS loop delta-t (affected and non-affected loops), steam pressure (affected and non-affected loops), pressurizer pressure, and pressurizcr level.

Question Source: Bank # 65340 Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 41.1,41.14 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 76 Tier#

RCP Malfunctions: Group #

Ability to detennineiinterpret calculation of flow in loop KiA # APE.OI5/17.AA2.07 with stopped RCP Importance Rating 2.9 Proposed Question:

With the plant at 30% power, the following sequence of events occurs:

1. The crew enters AOP 3554, RCP Trip, or Stopping a RCP at Power.

2. The crew removcs thc "c" RCP from service.

3. The plant remains on line during the initial shrink in the affected SG.

Based on the RCS flow response in the "C" loop, what action will the US direct with feed flow to the "C" SG; and what ACTION, if any, is the US required to direct with RCS Loop "C" temperature instrumentation that fecds Reactor Protection?

a) Feed flow to the "C" SG will need to be increased to maintain SG level, since "C" RCS loop flow is now due to natural circulation, which increases "C" SG temperature to approximately RCS hot. The temperature bistables are NOT required to be tripped, since the instruments are still operable.

b) Feed flow to the "c" SG will need to be reduced to prevent overfilling the SG, since RCS loop flow has reversed, lowering "C" SG temperature to approximately RCS T co1d ' The temperature bistables are NOT required to be tripped, since the instruments are still operable.

c) Feed flow to the "C" SG will need to be increased to maintain SG level, since "c" RCS loop flow is now due to natural circulation, which increases "C" SG temperature to approximately RCS T hot. The temperature bistables ARE required to be tripped using AOP 3571, Instrument Failure Response.

d) Feed flow to the "C" SG will need to be reduced to prevent overfilling the SG, since RCS loop flow has reversed, lowering "C" SG temperature to approximately RCS Tco1d ' The temperature bistables ARE required to be tripped using AOP 3571, Instrument Failure Re~ponse.

Proposed Answer: _D~__

Explanation (Optional): When the "C" RCP is stopped, flow in the affected loop will coast down and stop.

Flow will then reverse, since the running RCPs in the other 3 loops will maintain a DP across the core, with the affected loop cold leg pressure at RCP discharge pressure. The idle loop temperature will become very close to Tcold ofthe steaming loops. The affected steam generator will have a significant reduction in its steaming due loss of energy input from the RCS. So, the US will direct the BOP to stop feeding the "c" SG (HA" and "C" wrong). "A" and "C" are plausible, since if all four RCPs trip, the hot legs will heat up, and natural circulation flow will commence. As for the temperature instrumentation in the affected loop, although it still functions ("A" and "B" plausible), it is only monitoring Tcold temperature, so it is no longer providing a useful input into the DT trips. So the US will direct the affected RCS loop temperature instrument bistables to be tripped using AOP 3571, Instrument Failure Response ("B" wrong, "D" correct).

Technical Reference(s): -.:;:..FS;:.;AR:..;;;;;.;. .l. (R:.,;;c.;;..v;...,;. ,; ; 2. ; ,.1.;.; ;.312.},. : ;.F. :;ig2.:u: . re:. . .: . ;15;.,;. 3; ,. -. ;. 9_ _ _ _ _ _ _ _ _ _ _ _ _ _ __

(Attach if not previously provided) AOP 3554 (Rev. 008-01), steps 5.e, 7, and 8.a (including version/revision number) AOP 3571 (Rev. 009-07), Attachment A, step 6, and Table Proposed references to be provided to applicants during examination: -.:. .N:. . :o.; .:n:.;:.e_ _ _ _ _ _ _ _ __

Learning MC-03349 For given plant conditions, qualitatively state the effect of... RCP trip ... on the (As Objective: following parameters: reactor power, rod position, RCS loop average temperatures (affected available) and non-affected loops), RCS loop delta-t (affected and non-affected loops), steam pressure (affected and non-affected loops), pressurizer pressure, and pressurizer level.

Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 41.5 and 43.5 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 77 Tier #

Loss of Reactor Coolant Makeup: Group #

Knowledge ofEOP mitigation strategies KIA # APE.022.GEN.2.4.6 Importance Rating 4.7 Proposed Question:

With the plant at 100% power, the following sequence of events occurs:

1. RO notices the "A" CHS Pump amps are oscillating.

2. Before the operators take action, the "A" CHS Pump trips.

3. The crew enters EOP 3506, Loss ofAll Charging.

4. The RO isolates letdown.

5. The STA reports RCP #1 seal inlet temperatures for all four RCPs indicate 140°F.

What action is required to be taken by the crew?

a) Trip the reactor and enter E-O, Reactor Trip or Safety injection, since seal inlet temperatures are elevated.

b) Trip the reactor and enter E-O, Reactor Trip or Safety injection, since seal injection flow has been lost.

c) Vent the "B" CHS Pump gravity feed boration line, verify a suction path aligned, isolate the RCP seal injection path, and then start the liB" CHS Pump.

d) Vent the liB" CHS Pump gravity feed boration line, verify a suction path aligned, check RCP seal cooling, and then start the liB" CHS Pump.

Proposed Answer: D Explanation (Optional): liD" is correct, since with signs of cavitation, the charging pump suction must be vented prior to starting a CHS Pump. n A" and liB" are wrong, but plausible, since a reactor trip is not required unless RCP inlet tcmperatures are elevated, or Pzr level is lost. "C" is wrong, but plausible, since the crew is only required to isolate seal injection if all seal cooling was lost (thermal barrier cooling is still present).

Technical Reference(s): EOP 3506 (009-02), steps 1-7 (Attach if not previously provided) EOP 3506 (009-02), Foldout Page (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning MC-07517 (SRO, 8T A) Given a set of plant conditions, determine the required (As available)

Objective: actions to be taken per EOP 3506.

Question Source: Bank # 76155 Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 43.5 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 78 Tier #

Steam Gen. Tube Rupture: Group #

Ability to determine/interpret plant conditions from KJA# EPE.038.EA2.07 survey of control room indications Importance Rating 4.8 Proposed Question:

A Steam Generator tube ruptures on the "D" Steam Generator, resulting in the following sequence of events:

1. The crew enters E-3, Steam Generator Tube Rupture.

2. RCS cooldown and depressurization are completed.

The crew is checking to determine ifECCS flow should be terminated, and current conditions are as follows:

• RCS subcooling is 35°F

• RCS pressure: is slowly decreasing

• Pressurizer level is 20% and slowly decreasing

• "D" SG pressure: is 1100 psig and stable

• SG NR levels are as follows

• "A" SG: 5% and increasing.

• "B" SG: 4% and increasing

• "c" SG: 6% and increasing

• "D" SG: 55% and decreasing What action is the crew required to perform?

a) Stop ECCS pumps. Remain in E-3 and establish normal charging flow.

b) Do not stop ECCS pumps. Go to ECA-3.1 SGTR with Loss ofReactor Coolant - Subcooled Recovery Desired.

c) Do not stop ECCS pumps. Transition to FR-H.I Response to Loss ofSecondary Heat Sink.

d) Do not stop ECCS pumps. Remain in E-3 and initiate RCS cooldown.

Proposed Answer:

-B- -

Explanation (Optional): 1. Checks subcooling > 32°F it is, 35°F

2. Verify secondary heat sink satisfied by one SG> 8% NR level.

3. RCS pressure stable or increasing.

It is decreasing, therefore apply the RNO. Do not stop ECCS pump. Go to ECA 3.1 ("B" correct).

4. PZR level> 16%. Pzr level is 20%.

"A" is plausible since this action would be taken if SI termination criteria were met. "C" is plausible since this action would be taken if heat sink criteria were not met, and intact SG levels are below 8%. "D" is plausible since this action would be performed ifPZR level was low.

Technical Reference(s): E-3 (Rev. 022), Step 18 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: -"-N-.:o.;;;n.;;;c_ _ _ _ _ _ _ _ __

Learning MC-04373 Discuss conditions which require transition to other procedures from (As available)

Objective: EOP 35 E-3.

Question Source: Bank #65978 Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 43.5 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 79 Tier#

Loss of DC Power: Group #

Knowledge of how AOPs are used in conjunction with KIA # APE.058.GEN.2.4.8 EOPs Importance Rating 4.5 Proposed Question:

With the plant initially at 30% power, the following sequence of events occurs:

1. Battery Bus 1 deenergizes.

2. The crew enters AOP 3563, Loss ofDC Bus.

3. The reactor automatically trips.

4. The US desires to continue implementing AOP 3563, s,ince it has guidance on the impact of the loss of DC bus that will assist in stabilizing the plant after the trip.

The SM is considering how to direct the US to implement AOP 3563, E-O, Reactor Trip or Safety Injection, and ES-O.l, Reactor Trip Response to mitigate the event.

Which procedure implementation strategy complies with OP 3272, EOP User's Guide in this situation?

a) Prior to entering E-O, the US directs the RO to perform the first two steps of AOP 3563, Attachment 1 "Loss of DC Bus I" which establishes RCS temperature control, and ensures proper control of AFW.

The crew then enters E-O. After completing E-O, step 4, the crew transitions to ES-O.l, and completes the remaining steps of AOP 3563 in parallel with ES-O.l.

b) The US hands AOP 3563 to the extra SRO, who directs the BOP to perform the steps of AOP 3563, Attachment 1 "Loss of DC Bus I" in parallel with the US, who enters E-O and directs the RO to perform the steps ofE-O. After completing E-O, step 4, the crew transitions to ES-O.l, with the extra SRO and BOP completing all of AOP 3563, while the US and RO perform ES-O.l.

e) The crew enters E-O. After completing E-O, step 4, the crew transitions to AOP 3563, and the US directs the RO and BOP to perform the steps of AOP 3563, Attachment 1 "Loss of DC Bus 1" to deal with the impact of the loss of DC bus 1. After the crew completes all of AOP 3563, the crew transitions to ES 0.1.

d) The crew enters E-O. After completing E-O, step 4, the crew transitions to ES-O. l, with the US implementing AOP 3563, Attachment 1 "Loss of DC Bus 1" in parallel with ES-O.l. Only the steps of AOP 3563 that are necessary to ensure success ofES-O.l are performed, without completing all of AOP 3563.

Proposed Answer: ,....=.D-,-_

Explanation (Optional): It is acceptable to perform the actions of an AOP in parallel with an ERG derived EOP provided the actions ofthe ERG-derived procedure receives priority ("C" wrong) and the actions of the AOP are not initiated before completing all immediate actions of the ERG derived procedure ("A" and liB" wrong). It is not necessary to perform all steps in the parallel procedure. Only those steps necessary to ensure success of the ERG derived procedure need to be performed (liD" correct).

Technical Reference( s): ---=O-"'-P-=3:. .: 2:. :. 7.: .2. . >.(R::. : . : . ev-=.-=O.. :.O. :. 8--=-1:. .: 1L.2),--"S:. . :e. :.ct=-io: . : n.: . . : . :1. . :. .7_ _ _ _ _ _ _ _ _ _ _ _ _ __

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: --=-N;.. :o;.:.:n;.;:.e_ _ _ _ _ _ _ _ __

Learning MC-04455 Describe the usage of abnormal operating procedures while in the (As available)

Objective: emergency operating procedure network.

Question Source: Bank # 78929 Question History: Millstone 3 2004 NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 43.5 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 80 Tier #

Loss of Emergency Coolant Recirc: Group #

Ability to determine/interpret facility conditions and KJA# EPE.W/Ell.EA2.1 selection of appropriate procedure Importance Rating 4.2 Proposed Question:

A LOCA outside Containment occurs, resulting in the following sequence of events:

I. The crew enters E-O, Reactor Trip or Safety Injection.

2. With RCS pressure at 1650 psia, the RO reports "A" RHR Pump flow indicates 300 gpm.

3. The crew transitions to ECA-I.2, LOCA outside Containment.

4. Immediately after closing Cold Leg Injection Valve SIL *MV8809A, RHR flow drops to 0 gpm.

Has the leak been isolated from the RCS; and to which procedure is the crew required to transition?

a) The leak has NOT been isolated from the RCS. The crew will transition to E-J, Loss ofReactor or Secondary Coolant.

b) The leak has NOT been isolated from the RCS. The crew will transition to ECA-l.1, Loss ofEmergency Coolant Recirculation.

c) The leak has been isolated from the RCS. The crew will transition to E-l, Loss ofReactor or Secondary Coolant.

d) The leak has been isolated from the RCS. The crew will transition to ECA-l.l, Loss ofEmergency Coolant Recirculation.

Proposed Answer: B Explanation (Optional): Since RHR flow initially existed with RCS above RHR pump shutoff head, the leak was into the RHR System. Since RHR flow dropped to zero with the discharge valve closed, the leak is on the RCS side of the isolation valve, so the leak is still active ("C" and "D" wrong). If ECA-1.2 is not successful at isolating a LOCA outside CTMT, the crew will transition to ECA-l.l since no water is entering the CTMT sump to support CTMT recirc ("B" correct, "A" wrong). "A" is plausible, since the crew would transition to E-l if the leak was isolated, and a LOCA is still in progress. "C" and "D" are plausible, since RHR flow dropped to zero when the valve was closed.

Technical Reference(s): ECA-l.l (Rev. 016-02), Entry Conditions (Attach ifnot previously provided) ECA-1.2 (Rev. 008), step 5 (including version/revision number)

Proposed references to be provided to applicants during examination: None Learning (As available)

Objective: MC-03870 Identify plant conditions that require entry into EOP 35 ECA-l.l.

Question Source: New Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFRPart 55 Content: 43.5 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 81 Tier #

Inadequate Heat Transfer Loss of Secondary Heat Sink Oroup #

Ability to diagnose and recognize trends utilizing KIA # EPE.W/EOS.OEN.2.4.47 appropriate control room reference material Importance Rating 4.2 Proposed Question:

With the plant initially at 100% power, the following sequence of events occurs:

0250 The plant trips due to a loss of all Main Feedwater Pumps 0324 The crew enters FR-H.l, Response to Loss ofSecondary Heat Sink.

0330 All SO WR levels indicate 50% and decreasing.

0337 RCS Tcold reaches 561°F and the SO Atmospheric Relief Valves open.

0355 Current conditions are as follows:

• All SO WR levels indicate 32% and stable.

• RCS pressure is cycling on the PORVs.

• CETCs indicate 639°F and increasing.

• RCS Loop AT is decreasing.

What required action is the highest priority for the crew at this point?

a) Attempt to restore feed from AFW pumps.

b) Attempt to restore feed from the Main Feed pumps.

e} Attempt to restore feed from the Main Condensate pumps.

d) Immediately initiate bleed and feed of the RCS.

Proposed Answer: D Explanation (Optional): This event has an element from Diablo Canyon OE: (NRC-IN 2002-10) Feb 9, 2002, where SG level stabilized at 7.5% NR, above the 7.2% trip & AFW aetuation setpoint, even though actual level was still decreasing, due to DP across the SG moisture separator. With all WR levels stable above 29%

WR, the 2nd B&F criterion needs to be applied, which is CETCs increasing with RCS pressure at 2350 psia due to a loss of heat sink. "D" is correct, and "A", "B", and "C" wrong, since RCS temperature is increasing with secondary relief valves open and subcooling decreasing, showing heat sink is not adequate. "An, "B",

and "C" are plausible since these are actions in FR-H.l to restore heat sink ifimmediate bleed and feed is not yet required, and SG levels are all above the 29% B&F actuation setpoint. Also, there are several ways to be at 2350 psia with CETCs increasing that are not due to loss of heat sink, such as if SI actuates due to a SG fault, with the RCS heating up after the SG blows dry. "A", "B", and "c" are plausible, since these are heat sink restoration strategies in FR-H.I, wide range SG levels are above bleed and feed criterion, and there are several conditions where CETCs could be increasing with RCS pressure at 2350 psia (such as SIS with a blown down faulted SG) other than due to a loss of heat sink.

Technical Reference(s): FR-H.l (Rev. 020-0 I), Caution prior to step 3 (Attach if not previously provided) WOO Bkgd Doc (Rev. 2), for FR-H.l, section 2.2.4.5 (including version/revision number) WOG Exec Vol. (Rev. 2), Generic Issue: Natural Circulation, Section 2.1 Proposed references to be provided to applicants during examination: --'-N.:. .;o..;;:n..:.e_ _ _ _ _ _ _ _ __

Learning MC-07461 Given a set of plant conditions, determine the required actions to be (As available)

Objective: taken per FR-H.l.

Question Source: Bank # 78739 Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 43.5 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 82 Tier # I High Reactor Coolant Activity: Group # 2 Ability to determine/interpret corrective actions for high KIA # APE.076.AA2.02 fission product activity Importance Rating 3.4 Proposed Question:

With the plant initially at 100% power, the following sequence of events occurs:

1. The crew performs a rapid downpower to 80%.

2. The US directs the Chemistry Department to sample the RCS for Iodine.

3. Five hours after the downpower was commenced, Chemistry reports the following:

• DOSE EQUIVALENT 1-131 is 32 uO per gram.

• DOSE EQUIVALENT Xe-133 is 82 uCi per gram.

Using the attached copy of LCO 3.4.8, ""Specific Activity, what ACTION is the crew required to take, and what is the Tech Spec basis for the limits on RCS Activity?

a) Power operation may continue up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> while efforts are made to restore RCS Activity to within limits. The basis is to minimize the dose consequences of a Steam Line Break or a Steam Generator Tube Rupture.

b) Power operation may continue up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> while efforts are made to restore RCS Activity to within limits. The basis is to minimize the dose consequences of a LOCA inside or outside Containment.

c) Be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The basis is to minimize the dose consequences of a Steam Line Break or a Steam Generator Tube Rupture.

d) Be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The basis is to minimize the dose consequences of a LOCA inside or outside Containment.

Proposed Answer: A Explanation (Optional): LCO 3.4.8 has three ACTIONS for 1-131 limits, based on activity level and the amount of time the limit is exceeded. It also has two ACTIONS for Xe-133 limits. With 1-131 between I and 60 *Ci per gram for less than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, power operation may continue up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> while efforts are made to restore RCS Activity to within limits. "C" and "D" are wrong, since the ACTION for Xe-I33 above 81.2 *Ci per gram is also 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. "AI! is correct, and "B" wrong, since the basis for limiting RCS activity is to minimize the dose consequences of a Steam Line Break or a Steam Generator Tube Rupture. "B" is plausible, since radiation would be released on a LOCA as well. "C" and "D" are plausible, since the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> ACTION can be triggered by either 1-131 or Xe-I33, and by exceeding 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

Technical Reference(s): Tech Spec LCO 3.4.8 (Amendment 246)

(Attach if not previously provided) Tech Spec Basis for LCO 3.4.88 (LBDCR No. 08-MP3-013)

(including version/revision number)

Proposed references to be provided to applicants during examination: Tech Spec LCO 3.4.8 Learning MC-0753 1Given a set of plant conditions, determine the required actions to be taken (As available)

Objective: per AOP 3553.

Question Source: New Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 43.2 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 83 Tier #

Steam Generator Over-pressure: Group # 2 Ability to apply Technical Specifications for a system KfA# W/EJ 3.GEN.2.2AO Importance Rating 4.7 Proposed Question:

With the plant at middle oflife conditions, a grid instability event occurs, resulting in the following sequence of events:

1. The crew observes that the two lowest-set SG safety valves do not appear to be opening at the required setpoints for each of the Steam Generators.

2. The reactor trips.

3. The crew enters FR-H.2, Response to Steam Generator Overpressure.

4. The plant is stabilized at no-load conditions.

5. Plant management desires to start up the plant as soon as practical.

During the event review, it is determined that the SG Safety Valves lifted at the following SG pressures:

MSS Valve "A" SG "B" SG "C" SG "D" SG RV22A-D 1200 psig 1200 psig 1225 psig 1225 psig RV23A-D 1210psig 1220 psig 1230 psig 1235 psig RV24A-D 1205 psig 1210 psig 1210 psig 1205 psig RV25A-D 1215 psig 1215 psig 1215 psig 1215 psig RV26A-D 1230 psig 1230 psig 1225 psig 1225 psig Using the attached copy of Tech Spec LCO 3.7.1.1, what is the maximum power level allowed by the ACTION of LCO 3.7.1.1, assuming the NIS Hi Flux setpoints are lowered, if required, to the required setpoint?

a) 0%

b) 25.5%

c) 42.8%

d) 60.1%.

Proposed Answcr: C Explanation (Optional): The Tech Spec required lift settings per Table 3.7-3 is within 1% following safety valve testing (not the case at Middle of Life), and within 3% for normal conditions. So the required lift setpoint for the lowest set safety valves (RV22A-D) is 1185 psig x 1.03 1220.55 psig, meaning RV22C and D are not operable. The required lift setpoint for the second-lowest set safety valves (RV23A-D) is 1195 psig x 1.03 == 1230.85 psig, meaning RV22D is not operable. So, the "D" SG has two inoperable safety valves, requiring the crew to enter ACTION b. This requires power to be reduced (limited) to the limit of Table 3.7-1; and with two inoperable safety valves, the "D" SG still has 3 operable safety valves, requiring power to be limited to 42.8% ("C" correct, "A", "B", and "D" wrong). "A" is plausible, since ACTION d requires the plant to be shutdown if 4 or more safety valves are inoperable on one or more SGs, and a total of 8 safety valves lifted more than 1% outside of their lift setpoint. Also, surveillance requirement 4.0.3 prevents increasing modes while relying on an action statement that requires a plant shutdown, but a shutdown is not required for current conditions. "B" and "D" are plausible, since these setpoints are also on table 3.7-1, depending on how many safety valves are operable.

Technical Reference(s): Tech Spec LCO 3.7.1.1 (Amendment 242)

(Attach ifnot previously provided) Tech Spec Table 3.7-1 (Amendment 242)

(including version/revision number) Tech Spec Table 3.7-3 (Amendment 106)

Proposed references to be provided to applicants during examination: Tech Spec LCO 3.7.1.1 and Tables Learning MC-05007 Given a plant condition or equipment malfunction, use provided (As available)

Objective: reference material to ... evaluate technical specification applicability and determine required action requirements ...

Question Source: New Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 43.2 and 43.5 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 84 Tier#

Site Specific Instrument Failure Response: Group # 2 Ability to perform specific system and integrated KiA # Site Specific. AOP 3571.GEN.2.1.23 plant procedures during all modes of operation Importance Rating 4.4 Proposed Question:

With the plant at 100% power, the following sequence of events occurs:

1. NIS power range channel N-41 fails low.

2. The crew enters AOP 3571, Instrument Failure Response.

3. The US directs the RO to remove channel N-41 from input to the AFD and QPTR monitor alarms for computer program 3R5.

4. The RO completes this step 45 minutes after the instrument initially failed.

Now that the RO's actions are completed, what additional action(s), if any, is/are the US required to direct, and why?

a) The US is required to implement AFD monitoring since the AFD Monitor Alarm is considered INOPERABLE. The US is also required to direct preparations to be made for incore flux mapping, since the QPTR Monitor Alarm is considered INOPERABLE.

b) The US is required to implement AFD monitoring since the AFD Monitor Alarm is considered INOPERABLE. The US is NOT required to direct preparations to be made for incore flux mapping, since the QPTR Monitor Alarm is considered OPERABLE.

c) The US is NOT required to implement AFD monitoring since the AFD Monitor Alarm is considered OPERABLE. The US is required to direct preparations to be made for incore flux mapping, since the QPTR Monitor Alarm is considered INOPERABLE.

d) The US is NOT required to implement AFD monitoring since the AFD Monitor Alarm is considered OPERABLE. The US is NOT required to direct preparations to be made for incore flux mapping, since the QPTR Monitor Alarm is considered OPERABLE.

Proposed Answer: C Explanation (Optional): The AOP actions have removed the affected input from the AFD program, which restores the AFD Monitor Alarm to OPERABLE, so AFD monitoring is not required ("A" and "B" wrong).

The AOP actions have also removed the affected input from the QPTR program, but this does not restore OPERABILITY to the QPTR Monitor Alarm ("C" correct and "D" wrong). "A" and "B" are plausible, since the AFD alarm would be considered INOPERABLE if the RO had taken longer than one hour to complete the AOP 3571 step. "B" and "D" are plausible, since the RO has taken action to remove the failed channel from the QPTR alarm program.

Technical Reference(s): AOP 3571 (Rev. 009-07), Attachment D, steps 6 and 7, and associated Notes (Attach if not previously provided) Tech Spec Surveillance Req. 4.2.1.1.1.b and 4.2.4.l.b (Amendment 60)

(including version/revision number)

Proposed references to be provided to applicants during examination: None

~~~------------------

Learning MC-07563 (SRO, STA) Given a set of plant conditions, determine the required (As available)

Objective: actions to be taken per AOP 3571.

Question Source: Bank # 80911 Question History: Millstone 3 2007 NRC Exam Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 43.2 and 43.5 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 85 Tier #

Site Specific - Loss of all AC Power Recovery with the Group # 2 SBO Diesel, ECA-O.3: Ability to determine/interpret KiA # Site Specific. ECA-0.3.EA2 adherence to appropriate procedure. Importance Rating Site Specific Proposed Question:

The following sequence of events occurs:

1. The plant trips due to a loss of all AC power.

2. The crew enters ECA-O.O, Loss ofAll AC Power.

3. The crew restores power to Bus 34C from the SBO diesel.

4. The crew transitions out of ECA-O.O.

5. The crew starts the "A" Charging Pump.

Current Conditions are as follows:

• Pressurizer level: 10%

• RCS pressure: 1550 psia

• Core Exit Thermocouples: 564°F

• CTMT temperature: 125°F What action is the US required to direct?

a) Open Charging Flow Control Valve 3CHS*FCV 121 to increase PZR level above 16%.

b) Open one charging pump cold leg injection valve and increase PZR level above 16%.

c) Actuate Safety Injection, and remain in ECA-O.3 Loss ofAll AC Power Recovery with the SEa Diesel.

d) Actuate Safety Injection, and transition to ECA-0.2 Loss ofAll AC Power Recovery with S1 required.

Proposed Answer: B Explanation (Optional): The erew has just started a Charging Pump, so the crew has just completed step 6.

With the SBO diesel as the only source of power, significant loading limitations exist, so the crew will not transition to another EOP, since other EOPs assume at least one emergency bus is available ("D" wrong).

The crew will not actuate SIS ("C" wrong), since SI is directed to be reset to allow manual loading of equipment (and avoid overloading the SBO diesel) per the caution prior to step 1 ofECA-O.3. liB" is correct, and "A" wrong, since the cold leg injection valve will supply the maximum amount of water from one charging pump with low Pzr level. "A" is plausible, since this would raise PZR level, the Charging pump is currently supplying water through FCV 121, and this action is directed in other procedures, such as AOP 3555 RCS leak. "C" and liD" are plausible, since actuating SI would raise PZR level, PZR level is below the SI reinitiation setpoint on the foldout page of several EOPs. Also, ECA-O.2 would be the correct choice if either offsite power or an EDG were supplying power.

Technical Referencc(s): ECA-O.3 (Rev. 13), Caution prior to step 1, steps 6 and 7 (Attach ifnot previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None

~~~------------------

Learning MC-074 11 (SRO, STA) Given a set of plant conditions, determine the required (As available)

Objective: actions to be taken per ECA-O.3.

Question Source: Bank # 67595 Question History: Millstone 3 2009 NRC Exam Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 43.5 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 86 Tier # 2 Chemical and Volume Control: Group #

Predict impact/mitigate depressurizing RCS while hot KfA# 004.A2.28 Proposed Question: Importance Rating 4.3 The following initial sequence of events occurs:

I. A Steam Generator Tube Rupture occurs on the "C" SO.

2. The crew enters E-3, Steam Generator Tube Rupture.

3. The crew prepares to depressurize the RCS to minimize break flow and refill the pressurizer.

4. The crew is NOT able to commence depressurizing the plant by any of the methods directed in E-3.

5. The crew transitions to the appropriate Emergency Contingency Action procedure, with current conditions as follows:

• RCS subcooling is 80°F.

• "C" SG NR level is 70%.

• "C" SG pressure is 1085 psig.

The crew commences depressurizing the RCS by stopping one Charging Pump, both SIH pumps, and realigning charging through the nonnal charging flowpath.

To what pressure will the RCS initially depressurize ifno operator action is taken; and when is the crew required to take action to terminate the depressurization?

a) The RCS will depressurize to and stabilize at ruptured SG pressure. The crew will terminate the depressurization if subcooling decreases to 32°F, or if Plenum level drops to less than 19%.

b) The RCS will depressurize to and stabilize at ruptured SG pressure. The crew will terminate the depressurization when the RCS reaches saturation conditions, or ifPzr level exceeds 73%.

c) The RCS will depressurize to and stabilize at CETC saturation pressure. The crew will terminate the depressurization if subcooling decreases to 32°F, or if Plenum level drops to less than 19%.

d) The RCS will depressurize to and stabilize at CETC saturation pressure. The crew will terminate the depressurization when the RCS reaches saturation conditions, or ifPzr level exceeds 73%.

Proposed Answer: ,....:..:A=--_

Explanation (Optional): With no pressure control, the crew is required to transition to ECA-3.3. This will stop ECCS and realign charging to cause depressurization and terminate break flow. With no SI flow, break flow will rapidly empty the Pzr, and partially drain the hot leg. When steam hits the subcooled hot leg water, it will condense, continuing the depressurization, until the RCS reaches SG pressure ("C" and "D" wrong).

"C" and "D" are plausible, since during a LOCA, the RCS would continue to depressurize, and for loss of heat sink events, pressure would hold up at the point where the hot legs are saturated. "A" is correct, and "B" wrong, since termination criteria are subcooling ::: 32°F, or plenum level < 19%. "B" is plausible since this is termination criteria from ECA-3.2, which the crew would enter based on high SG level, and it is elevated.

Technical Reference(s): _E;:;;.-. .::3.. .:(.:;.R:.:.ev..:. ;...::2:.:.2;.:2),...::s.:..:.teLP-=I...::6...::a~-d::..-_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

(Attach if not previously provided) -"-E.. ;;.C.;;..A.;;..-3;.. ..;.;;;;.2-"(R:..:. :. .ev.;;....;;..O.;;..1_4.-...;;O...;;.I~},...;;s.;;..te""p_I...;;8_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

(including version/revision number) _E_C_A_-.;;..3.;;...3....:{....R_e_v_.O_1_3_-_0~1}"'-,_st_e.... p_s_7_-1_0_ _ _ _ _ _ _ _ _ _ _ _ _ _ __

WOG Bkgd Doc (Rev. 2) for ECA-3.3, page 31, and pages 24-28 Proposed references to be provided to applicants during examination: --:;..N.;.;o;.;;n;..:.e_ _ _ _ _ _ _ _ __

Learning MC-07441 Given a set of plant conditions, determine the required actions to be (As available)

Objective: taken per E-3.

Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 43.5 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 87 Tier # 2 Residual Heat Removal: Group #

Predict impact/mitigate RHR Pump malfunction KIA # 005.A2.03 Importance Rating 3.1 Proposed Question:

INITIAL CONDITIONS:

• Plant is in MODE 5, solid plant operations, on the "A" Train ofRHR

• "B" Electrical Distribution Train outage is in progress, and cannot be immediately restored

• RCS temperature is 150°F

• RCS pressure is 150 psi a

• All RCS loops are full, swept, and vented.

• All Steam Generators are in Wet-Layup

• The "A" Charging Pump is running

• No RCPs are running The:"A" RHR Pump shaft shears, and its flow drops to zero.

What action will the crew take to remove decay heat?

a) Open both PORVs, and fill the RCS using one Charging Pump from the RWST.

b) Open both PORVs, and fill the RCS using one SI Pump from the RWST.

c) Throttle open the charging line flow control val ve to raise RCS pressure to >170 psi a, and open the Steam Generator Atmospheric Relief Valves.

d) Start one Reactor Coolant Pump, check proper differential pressure across its #1 seal, and open the Steam Generator Atmospheric Relief Valves.

Proposed Answer: ---=-C_ _

Explanation (Optional): "C" is correct, since the RCS is already full and steam generators are available. The procedure has conditions established for natural circulation, RCS pressure is increased to ensure subcooled natural circulation cooling, and the steam generators are used to dump steam. "A" and "B" are wrong, but plausible, since bleed and feed is only used if natural circulation cooling is unsuccessful. The charging pump is the preferred feed source, and the SI Pump is the backup source of feed. "D" is wrong, but plausible, since forced cooling is only used if a RCP is already running.

Technical Reference(s): EOP 3505 (Rev. 010-03), AttachmentB, steps 8-11 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: -=..N.:. .;o;..:;n:.;;e_ _ _ _ _ _ _ _ __

Leaming MC-07513 Given a set of plant conditions, determine the required actions to be (As available)

Objective: taken per EOP 3505.

Question Source: Bank # 76224 Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 43.5 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 88 Tier # 2 Engineered Safety Features Actuation: Knowledge ofless Group #

than or equal to one hour Tech Spec action statements KJ A # o13.GEN.2.2.39 Proposed Question: Importance Rating 4.5 A refueling outage has just been completed, and Initial Conditions are as follows:

• Two Pzr Pressure Transmitters (3RCS*PT455 and PT456) were replaced during the outage.

• The plant is in MODE 3.

• The crew is perfonning a plant heatup per OP 3201, Plant Heatup.

The following sequence of events occurs:

1. Pzr pressure increases above 2000 psia as indicated on Pzr Pressure Transmitters 3RCS*PT457 and 458.

2. The RO reports 3RCS*PT455 and PT456 still indicate 1950 psia.

3. 3RCS*PT455 and PT456 are both declared INOPERABLE.

Based on the failed instruments and current permissive status, what ACTION, if any, is the crew required to take?

a) The plant startup may proceed provided the associated Bistables are tripped within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b) The plant startup may proceed provided the associated Bistables are tripped within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

c) Suspend the plant startup, enter LCO 3.0.3, and within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> check the status ofthe P-ll permissive annunciator window.

d) Suspend the plant startup, and within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> check the status of the P-ll permissive annunciator window.

Entry into LCO 3.0.3 is NOT required.

Proposed Answer: --:;;C_ _

Explanation (Optional): With an ESFAS Instrumentation Channel or Interlock Channel found to be inoperable, declare the channel inoperable and apply the applicable ACTION statement requirements of Table 3.3-3 until the channel is restored to OPERABLE status.

Per Functional Unit 9: Engineering Safety Features Actuation System Interlocks for Pressurizer Pressure, P 11, ACTION 21 applies, with the total channels 3, and minimum channels 2. With less than the Minimum Number of Channels OPERABLE, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> determine by observation ofthe associated permissive annunciator window(s) that the interlock is in its required state for the existing plant condition ("A" and "B" wrong), and since P-ll is now in the wrong state (2 of3 channels less than 2000 psia), apply Specification 3.0.3 ("C" correct, "D" wrong). The crew also needs to enter LCO 3.0.3 due to Low Pzr Pressure SI less than minimum channels per Functional Unit Ld, ACTION 20. "A" and "B" are plausible, since numerous RPS and ESF bistables require tripping within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and others within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. And this action would be correct, if one channel was low (less than total channels, but not less than minimum). "D" is plausible, since this would be correct for P-ll if the permissive were in the required state.

Technical Reference(s): Tech Spec LCO 3.3.2, ACTION b (Amendment 159)

(Attach ifnot previously provided) Tech Spec Table 3.3-3 (Various Amendments starting with 70)

(including version/revision number) _F_u-"-n_ct_io_n_a_I....D....r....aWl_*n_.g"--6.,.,,(N_o_.....G-"-)_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

Proposed references to be provided to applicants during examination: -:.,N,..;o;;:;n:.:,e_ _ _ _ _ _ _ _ __

Learning MC-05776 Given a plant condition or equipment malfunction, use provided (As available)

Objective: reference material to: a. Determine entry conditions to applicable plant procedures

b. Evaluate technical specification applicability and determine required actions Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 43.2 Comments:

Examination Outline Cross-reference: Level RO SRO Question # S9 Tier # 2 Main Feedwater: Group #

Ability to recognize entry conditions for EOPs/AOPs KJA# 059.GEN.2.4.4 Importance Rating 4.7 Proposed Question:

With the plant at 100% power, a transient occurs, and initial conditions are as follows:

• Reactor power is 100% and stable.

• RCS T ave is 58S0F and slowly increasing.

• RCS pressure is 2260 psia and slowly increasing.

• "A" SG NR Level is 35% and decreasing.

• Containment pressure is 16 psia and increasing.

CTMT HI-I pressure is reached, and SIS actuates.

For this event, which operator credited action is the crew required to complete to mitigate this event; and to which procedure will the crew transition after completing E-O?

a) The crew is required to check PORV Block valves OPEN within 70 minutes from initiation of safety inj ection. They will transition from E-O to E-I, Loss ofReactor or Secondary Coolant.

b) The crew is required to check PORV Block valves OPEN within 70 minutes from initiation of safety injection. They will transition from E-O to E-2, Faulted Steam Generator Isolation.

c) The crew is required to isolate auxiliary feedwater to the affected steam generator within 30 minutes from event initiation. They will transition from E-O to E-I, Loss ofReactor or Secondary Coolant.

d) The crew is required to isolate auxiliary feedwater to the affected steam generator within 30 minutes from event initiation. They will transition from E-O to E-2, Faulted Steam Generator Isolation.

Proposed Answer: _D_ _

Explanation (Optional): A feedline break inside CTMT is in progress (CTMT pressure is increasing, since energy is being released to CTMT; and SG level is decreasing since feed is not reaching the SG, and SG inventory is being lost out ofthe break). An operator-credited action for a feedline break is to isolate auxiliary feedwater to the affected steam generator within 30 minutes from time ofbreak ("A" and "B" wrong). After the feedline uncovers, the SG will depressurize, and the crew will be required to enter E-2 (HC" wrong and "D" correct). "A" and "B" are plausible, since E-I would be entered for a LOCA, and energy is being released to CTMT, and normal indication for a faulted SG on a stearnline break is RCS cooldown. "c" is plausible, since this is an operator credited action for an inadvertent SI.

Technical Reference(s): FSAR (Rev. 21.3), Chapter 15.2.S , and Figures 15.2.10 and 15.2.1S (Attach ifnot previously provided) COP 200.IS (Rev. 000-01), Attachment 3, Sheet 5 of20 (including version/revision number) ~E__

-O::.. . ;:(R;,;;.e:.. ;v;,;;..;,;;.0.::26..:.,)'-'.,;,;;.st.:,:eLP..:;;2:..:6______________________

Proposed references to be provided to applicants during examination: None

---

~----------------

Learning MC-04887 DESCRIBE the major parameter changes associated with decreased heat (As available)

Objective: removal by the Secondary System.

Question Source: New Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 43.1 and 43.5 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 90 Tier # 2 Instrument Air: Group #

Predict impact/mitigate air dryer and filter malfunctions KJA# 078.A2.01 Proposed Question: Importance Rating 2.9 A plant heatup is in progress per OP 3201, Plant Heatup, and initial conditions are as follows:

• RCS temperature is 190°F and stable.

• RCS pressure is 350 psia and stable.

• The PZR is solid.

The following sequence of events occurs:

1. Instrument Air pressure starts rapidly decreasing.

2. The crew enters AOP 3562, Loss ofInstrument Air.

3. A PEO who is dispatched to investigate reports the following:

• The INST AIR A AFT CLR OUT TEMP HI annunciator is lit.

• Instrument Air Filter Differential Pressure is 6 psid.

In accordance with AOP 3562, Attachment A, "Loss of Instrument Air Local Actions," what action will the US direct the PEO to take; and iflAS pressure continues to decrease, what will be the effect on RCS pressure?

a) The US will direct the PEO to place the emergency instrument air dryer in service. IflAS pressure continues to decrease, RCS pressure will increase to a maximum of about 455 psia.

b) The US will direct the PEO to place the emergency instrument air dryer in service. If lAS pressure continues to decrease, RCS pressure will increase to a maximum of about 2350 psia.

c) The US will direct the PEO to swap the in-service instrument air filters. IfIAS pressure continues to decrease, RCS pressure will increase to a maximum of about 455 psia.

d) The US will direct the PEO to swap the in-service instrument air filters. If lAS pressure continues to decrease, RCS pressure will increase to a maximum of about 2350 psia.

Proposed Answer: C Explanation (Optional): Question is considered SRO, since it requires detailed assessment of plant conditions, including Tech Spec required status of COPPS in lower MODEs. AOP 3562, Attachment A requires the crew to swap filters ifDP is above 4 psid, but does not require swapping air dryers unless specific annunciators are lit, and they are not ("A and "B wrong). "A" and "B" are plausible, since an lAS annunciator is lit. IfIAS pressure is lost, RHR letdown valve 3CHS*PCV 131 will fail closed, causing RCS pressure to increase. Since COPPS is required by Technical Specifications, and will remain ARMED until 230°F, and since the PORVs are not air operated valves, COPPS will mitigate the pressure rise at 455 psia

("C" correct, "D" wrong). "D" is plausible, since if COPPS were blocked, the PORVs would lift at 2350 psia.

Technical Reference(s): AOP 3562 (Rev. 007-01), Attachment A (Attach ifnot previously provided) OP 3201 (Rev. 021-05), steps 4.3.37, and 4.4.3 (including version/revision number) Tech Spec 3.4.9.3 (Amendment 197)

P&ID I04A (No. 52)

Proposed references to be provided to applicants during examination: None Learning MC-05324 Given a failure, partial or complete, of plant air systems, determine (As available)

Objective: effects on the systems and interrelated systems Question Source: New Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 43.2 and 43.5 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 91 Tier #

In-core Temperature Monitor: Group # 2 Predict impact/mitigate thermocouple open/short circuits KJA# 017.A2.01 Proposed Question: Importance Rating 3.5 Initial Conditions:

• The plant is in MODE 2, with a Reactor Startup in progress.

• Train "A" ICC/RVLMS indications are normal.

• Train "B" ICC/RVLMS indications are normal, except for the 100% head level sensor, which is not functioning due to a failed heater.

The following sequence of events occurs:

I. An open circuit occurs in the Train "B" 63% Head RVLMS unheated thermocouple.

2. Plant Process Computer indication for the Train "B" 63% Head RVLMS detector turns BLUE and shows an "X" Quality Tag.

3. I&C determines that the head 63% thermocouple is not repairable with the plant on line.

Using the attached copy ofLCO 3.3.3.6, what ACTION, if any, is required?

a) No ACTION is required. Both Trains ofRVLMS are still OPERABLE.

b) No ACTION prohibiting the reactor startup is required. Either restore the "B" Train ofRVLMS to OPERABLE within 7 days or submit a Special Report to the NRC within 30 days.

c) Halt the startup, and restore the "B" Train of RVLMS to OPERABLE within 7 days, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

d) Halt the startup until an alternate method of measuring reactor vessel inventory has been implemented.

Submit a Special Report to the NRC within 30 days, and restore the "B" Train ofRVLMS to OPERABLE status at the next scheduled refueling.

Proposed Answer:

-B- -

Explanation (Optional): Train B is INOPERABLE since, per Table 3.3-10, a probe is considered OPERABLE if half or more ofthe head and half or more of the plenum detectors are OPERABLE; and both head detectors are inoperable. "A" is plausible, since train "A" is fully functional, and less than half of Train "B" detectors are inoperable, but they are both in the head. "B" is correct, since, per T.S 3.3.3.6 "Accident Monitoring" ACTION e. is applicable for one channel failure. ACTION g. allows transition to an OPERATIONAL MODE while in the ACTIONS of3.3.3.6. "C" and "D" are wrong, since only one channel has failed, and the other channel is still OPERABLE. "C" is plausible, since this is the action for less than total channels for Accident Monitoring except RVLMS and CTMT Hi Range Monitor. "D" is plausible, since this is related to the action for RVLMS less than minimum channels.

Technical Reference(s): _L...;.C_O_3;..;..;;..3.;.;;.3. .;. 6_C....A_m_e.;;..n...;.d;...m_e...;.n_t_22_4....)_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

(Attach ifnot previously provided) Table 3.3.10 (Amendments 46 and 229)

(including version/revision number)

Proposed references to be provided to applicants during examination: LCO 3.3.3.6, including Table Learning MC-04834 Given a plant condition or equipment malfunction, use provided (As available)

Objective: reference material to: a. Determine entry conditions to applicable plant procedures.

B. Evaluate technical specification applicability and determine required actions ...

Question Source: Bank # 84463 Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 43.2 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 92 Tier # 2 Steam Generator: Group # 2 Predict impact/mitigate impact on SGS of faulted or KIA # 035.A2.01 ruptured SGs Importance Rating 4.6 Proposed Question:

Initial Conditions:

1. The crew is in E-3, Steam Generator Tube Rupture due to a tube rupture on the "B" SG.

2. The crew is preparing to commence the cool down of the RCS.

A safety valve lifts on the ruptured "B" SG, and sticks fully open.

How will "B" SG Wide Range level respond, and what procedure will the crew initially enter to mitigate this event?

a) Ruptured SG Wide Range level will decrease. The crew will transition to E-2, Faulted Steam Generator Isolation.

b) Ruptured SG Wide Range level will increase. The crew will transition to E-2, Faulted Steam Generator Isolation.

c) Ruptured SG Wide Range level will decrease. The crew will transition to ECA-3.1, SGTR with Loss of Reactor Coolant Subcooled Recovery Desired.

d) Ruptured SG Wide Range level will increase. The crew will transition to ECA-3.l, SGTR with Loss of Reactor Coolant - Subcooled Recovery Desired.

Proposed Answer: -=..;A~_

Explanation (Optional): Ruptured SG WR level and pressure will decrease due to loss of mass out the failed open safety valve ("B" and "D" wrong). "B" and "D" are plausible, since NR level would experience swell as the SG depressurizes, and RCS break flow rate will increase as the ruptured SG depressurizes. However, a design basis tube rupture is about 300 gpm, and a safety valve can pass about 5% steam flow, which is over 500 gpm. "A" is correct, and "C" wrong, since the crew is required to transition to E-2 to isolate the faulted SG prior to continuing with mitigation per E-3 series procedures. "C" is plausible, since the crew is at a step that will send them to ECA-3.1 if SG pressure drops to 530 gpm, and they will ultimately transition to ECA 3.1 to mitigate the faulted, ruptured SG.

Technical Reference(s): FSAR, Figure 15.1.17 (Rev. 21.3)

(Attach if not previously provided) SGS035C (Training Lesson Plan) (Rev. 1, Ch. 2), pages 16, 17, 22, and 23 (including version/revision number) E-3 (Rev. 022;L)2,.';;.;st=ep.r:.. =52..,::::.an::;:d=-=.F.=0:..;ld:..;o=u:,:.t. cp::::.ag<<.e=--_ _ _ _ _ _ _ _ _ _ _ _ __

Proposed references to be provided to applicants during examination: ....;...N_o_n..;.e_ _ _ _ _ _ _ _ __

Learning MC-04373 Discuss conditions which require transition to other procedures from (As available)

Objective: EOP 35 E-3 Question Source: Bank # 65975 Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 43.5 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 93 Tier # 2 Site Specific AMSAC: Group # 2 Ability to diagnose and recognize trends utilizing KiA # Site Specific AMSAC.GEN.2.4.47 appropriate control room reference material Importance Rating 4.2 Proposed Question:

Initial Conditions:

• Reactor power is 38% with a load increase to 100% in progress per OP 3204, At Power Operations.

• AMSAC-related annunciators indicate as follows:

• AMSAC NOT ARMED (MB4C): LIT

• AMSAC TROUBLE/BYPASS (MB4C): NOT LIT

• AMSAC TRIP (MB7B): NOT LIT The following sequence of events occurs:

1. The AMSAC TROUBLEIBYP ASS annunciator comes in.

2. I&C investigates and reports that an AMSAC software error exists which will prevent AMSAC from functioning.

What Tech Spec/TRM ACTION, if any, is required?

a) The power increase to 100% may continue, but efforts shall be made to return AMSAC to OPERABLE as soon as possible.

b) Operation at power may continue, but power must be restricted to less than 40% of full turbine load.

c) No ACTION is required, since no ACTION applies below 40% power.

d) AMSAC must be repaired within one hour, or the plant must be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Proposed Answer: -..::;..B_ _

Explanation (Optional): "B" is correct, and "A", "C", and "D" wrong, since TRM 7.2.1 ACTION 2 requires power to be restricted to less than 40% of full turbine load if AMSAC is not OPERABLE when below 40%

power. "A" is plausible since this is the ACTION required if power is above 40%. "C" is plausible since AMSAC does not ARM until 40% power. "D" is plausible since these are ACTIONs from LCO 3.0.3.

Technical Reference(s): TRM 7.2.1 (LBDCR 07-MP3-018)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during exalnination: .,. .::-N,.:o.:;n;.;. e_ _ _ _ _ _ _ _ __

Learning MC-04091 Given a plant condition or equipment malfunction, use provided (As available)

Objective: reference material to: a. Determine entry conditions to applicable plant procedures

b. Evaluate technical requirements applicability and determine required actions Question Source: Bank # 80132 Question History:

Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 41.7,41.8 and 43.2 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 94 Tier # 3 Ability to interpret and execute procedure steps Group #

KJA# GEN.2.1.20 Importance Rating 4.6 Proposed Question:

A plant cooldown is in progress in accordance with OP 3208, Plant Cooldown, and initial conditions are as follows:

• Both trains of RHR are in service in the Cooldown Mode.

• Pressurizer level is stable at 55%.

• RCS cold leg temperatures are 250°F and decreasing.

• RCS pressure is 350 psia and stable.

• PZR temperature and surge line temperature are both stable at 430°F.

• All Pressurizer heaters are energized.

The RO reports that pressurizer surge line temperature has started decreasing, indieating 420°F.

What adverse plant condition exists, and what action is the US required to direet?

a) Spmy flow has initiated with AT across the Pressurizer spray nozzle in excess ofTRM limits. The US will direct the extra senior licensed operator to notify Engineering and initiate a CR.

b) Spray flow has initiated with AT across the Pressurizer spray nozzle in excess ofTRM limits. The US will direct the RO to deenergize pressurizer heaters to restore AT to within limits within 30 minutes.

c) A pressurizer insurge is in progress, with AT between the RCS and Pzr applying thermal stress to the PZR Surge Line. The US will direct the RO to adjust Charging Flow Control Valve 3CHS*FCV121 to decrease charging flow.

d) A pressurizer insurge is in progress, with AT between the RCS and Pzr applying thermal stress to the PZR Surge Line. The US will direct the RO to adjust Letdown Pressure Controller 3RCS-PK131 to decrease letdown flow.

Proposed Answer: --=.C_ _

Explanation (Optional): In this situation, the pressurizer level control system is being used to maintain PZR level constant with spray flow adding water to the PZR at a rate greater than the net charging mte to the RCS as the RCS contracts during the cooldown. This establishes a continuous PZR outsurge, preventing a PZR insurge and the associated thermal transient. If net charging flow increases above the 35 gpm spray flow, an insurge occurs, as evidenced by the surge line temperature drop. The US must either increase letdown flow

("D" wrong, but plausible) or decrease charging flow ("C" correct). There is a 182°F temperature difference between the RCS and the PZR, which is within the 200°F spray nozzle administrative limit. "A" lists actions required if the 200°F limit is exceeded, and "B" lists the actions related to the TRM 320°F limit. (" A" and "B" wrong, but plausible).

Technical Reference(s): TRM 3A.9.2.C (LBDCR 07-MP3-0l8)

(Attach if not previously provided) ---:;.O.;;.P...:3...;;:2:...:.0:...:.8~(R...:e.;;.v"";"....;.O_2.;;.1-...:0....;.6L!.),....;.s.....:te~p_4.;;..3;....'.;;.33:..-.._ _ _ _ _ _ _ _ _ _ _ _ _ __

(including version/revision number) OP 3208 basis document (Rev. 20-20), steps 4.3.31 Proposed references to be provided to applicants during examination: .-:;,.N;,. ;o;,;;;n:..:.e_ _ _ _ _ _ _ _ __

Learning MC-07S03 Given a set of plant conditions, dettlrmine the required actions to be (As available)

Objective: taken per OP 3208.

Question Source: Bank # 78785 Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 43.5 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 95 Tier # 3 Knowledge of the refueling process Group #

KJA# GEN.2.1.41 Importance Rating 3.7 Proposed Question:

The plant has just been shutdown for a refueling outage.

Which of the following evolutions will require a Refueling SRO to be present on the Refueling Floor?

a) Removing the Upper Guide Structure from the reactor vessel b) Installing or removing the Fuel Transfer Tube blind flange c) Start filling the reactor vessel from the RWST using the RHR system d) Initial Reactor Vessel Stud detensioning Proposed Answer: -=-..:A,--_

Explanation (Optional): "A" is correct since OP 32 lOA, step 4.1.15 and 4.1.16 require a refueling SRO for this evolution. "B" is wrong, blind flange removal occurs at step 4.1.7 ofOP 32 lOA, Refueling Preparations, before step 4.1.15 which stations the Refueling SRO, and is NOT identified as a CORE ALT. "c" is wrong, filling the reactor vessel from the RWST occurs at step 4.3.13 ofOP 32 lOA, Refueling Preparations, does not require stationing the Refueling SRO, and is NOT identified as a CORE ALT. "D" is wrong, since detensioning occurs at step 4.1.6 ofOP 3210A, Refueling Preparations, before step 4.1.15 which stations the Refueling SRO and is NOT identified as a CORE ALT. HB", "C", and "D" are plausible, since each ofthese are actions taken in OP 32 lOA as part of the refueling process.

Technical Reference(s): OP 3210A (Rev. 013-07), Steps 4.1.7,4.1.15,4.1.16, and 4.3.13 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning MC-04544 Describe the following: (As available)

Objective: A. Core Alterations and what specifically marks the start of Core Alterations B. Who has authority to direct and/or approve all core eomponent movements Question Source: Bank # 78793 Question History: Millstone 3 2002 NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 43.6 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 96 Tier # 3 Knowledge of conditions and limitations in the facility Group #

license KIA # GEN.2.2.38 Importance Rating 4.5 Proposed Question:

A small, short-term transient occurs at the end of the computer-defined 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shift, and the following reports are made:

• Current calorimetric 4-minute average power indication is at 3660 MWth based on a steam flow calculation.

• The 12-hour shiftly average power comes in at 3650.1 MWth.

In accordance with OP 3204, At Power Operation, what are the minimum actions (in addition to submitting a CR. and maintaining the shiftly average power less than or equal to 3650 MWth) required to be performed by the crew?

a) Monitor 4-Minute Average Power, and ensure it returns to less than or equal to 3650 MWth within 20 minutes, and notify Reactor Engineering. This is NOT a reportable event.

b) Monitor 4-Minute Average Power, and ensure it returns to less than or equal to 3650 MWth within 20 minutes, and notify Reactor Engineering, requesting they determine Reportability.

c) Reduce power to less than or equal to 3650 MWth, and notify Reactor Engineering. This is NOT a reportable event.

d) Reduce power to Jess than or equal to 3650 MWth, and notify Reactor Engineering, requesting they determine Reportability.

Proposed Answer: --",-0_ _

Explanation (Optional): The erew is directed to maintain the 4 minute average less than or equal to 3650 MWth. "A" and "B" are wrong, since allowance is made for brief, statistical fluctuations of up to 0.2%

power (3657 MWth), but this has been exceeded. IF the "4 minute average" THERMAL POWER (CVQRPA) exceeds 100.2% (3,668 MWth), the crew is required to reduce power to less than 3650 MWth.

"A" and "B" are plausible, since if power exceeds 3650 MWth by less than 0.2%, the crew is generally allowed 15 to 20 minutes to analyze the transient prior to having to reduce power. "D" is correct, and "C" wrong, since if the 12-hour average exceeds 3650 MWth, Reactor Engineering must be requested to determine Reportability. "An is plausible, since power has not exceeded 102% by either 4-minute or shiftly calculation, which is the 4-minute average power trigger to determine reportability.

Technical Reference(s): Facility License (Amendment 242), section 2.C.(l)

(Attach if not previously provided) OP-3204 (Rev. 017-13), sections 1.2 and 4.3.1 (including version/revision number)

Proposed references to be provided to applieants during examination: None

~~~~----------------

Learning MC-05227 Describe the major administrative or procedural precautions and limitations (As Objective: placed on the operation of the Nuclear Instrumentation System and the basis for each. available)

Question Source: Modified Bank # 69062 Original Question Attached.

Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 43.1 and 43.5 Comments:

Original Bank Question #69062 is attached on the next page.

Original Bank Question #69062 A transient occurs at the end of the computer-defined 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shift, and the following reports are made:

• Current calorimetric 4-minute average power indication is at 3705 MWth based on a steam flow calculation.

• The 12-hour shiftly average power comes in at 3650.0 MWth.

In accordance with OP 3204, At Power Operation, what actions (in addition to submitting a CR, and maintaining the shiftly average power less than or equal to 3650 MWth) are required to be performed by the crew?

a) Immediately reduce power to less than or equal to 3650 MWth, and notify Reactor Engineering.

b) Promptly power to less than or equal to 3650 MWth within 15 minutes, and request Reactor Engineering to determine Reportability.

c) Immediately reduce power to less than or equal to 3650 MWth, and notify the NRC within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

d) Promptly reduce power to less than or equal to 3650 MWth, and notify the NRC within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Answer: B

Examination Outline Cross-reference: Level RO SRO Question # 97 Tier # 3 Knowledge ofless than or equal to one hour Tech Spec Group # 2 action statements for systems KJA# GEN.2.2.39 Importance Rating 4.5 Proposed Question:

The crew is preparing to take critical data with the reactor at 1 x 10-8 amps.

The STA reports that Tave has dropped to 550°F.

What Technical Specification ACTION, if any, is required?

a) Restore Tavg within 15 minutes OR be in HOT STANDBY within the next 15 minutes.

b) Restore Tavg within 15 minutes AND be in HOT STANDBY within the next 15 minutes.

c) Restore Tavg within hour AND be in HOT STANDBY within one hour.

d) No ACTION is required, since the plant is NOT in MODE 1. Restore Tavg prior to entering MODE l.

Proposed Answer:

-A- -

Explanation (Optional): With a Reactor Coolant System operating loop temperature (Tavg) less than 551°F, restore Tavg to within its limit within 15 minutes or be in HOT STANDBY within the next 15 minutes.

Therefore, with the plant in MODE 2 the ACTION is appropriate ("A" correct, "B", "C", and "D" wrong).

"B" is plausible, since it is close to the required action. "c" is plausible, since this is the action required if a core safety limit were violated. "D" is plausible, since in MODE 2, and the action would not apply ifK..,rr were less than one.

Technical Reference(s): Tech Spec LCO 3.1.1.4 (Amendment 60)

(Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: None Learning -------------------------

MC-00016 Given a plant condition or equipment malfunction, use provided reference ma (As Objective: terial to... Evaluate Technical Specifications applicability and determine required actions ... available)

Question Source: Bank # 68646 Question History:

Question Cognitive Level: Memory or Fundamental Knowledge 10 CFR Part 55 Content: 43.2 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 98 Tier # 3 Ability to control radiation releases Group # 3 KIA # GEN.2.3.11 Importance Rating 4.3 Proposed Question:

With the plant at 100% power and a SLCRS Fan running, the following sequence of events occurs:

1. The "SLCRS Filter Fan Discharge to Millstone Stack" radiation monitor 3 HVR *RE 19B goes into alann.

2. The crew enters AOP 3573, Radiation Monitor Alarm Response.

What two actions are required to be directed by the US to check for potential sources of the radioactive release?

a) Check the Condenser Air Ejector Discharge Radiation Monitor 3ARC-RE21 trend, and check the CTMT Purge Exhaust Fans running.

b) Check the RPCCW Radiation Monitor 3CCP-RE31 trend, and check the CTMT Purge Exhaust Fans running.

c) Cheek the Condenser Air Ejector Discharge Radiation Monitor 3ARC-RE21 trend, and check the CTMT Vacuum Pumps running.

d) Check the RPCCW Radiation Monitor 3CCP-RE31 trend, and check the CTMT Vacuum Pumps running.

Proposed Answer: --:;;C_ _

Explanation (Optiona\): "C" is correct since both thc Condenser Air Ejectors and the CTMT Vacuum Pumps discharge to Gaseous Waste, which goes to the Millstone Stack. "A" and "B" are wrong, since the Containment Purge System exhausts to the Turbine Bldg stack, but plausible since it draws on CTMT, as does the CTMT vacuum pumps. "B" and "D" are wrong, since RPCCW is monitored by CCP-RE31, which is not alarming. "B" and "D" are plausible, since RPCCW overflows to the Aux Bldg, which is drawn on by SLCRS.

Technical Reference(s): AOP 3573 (Rev. 018-01), Attachment A, page 10 (Attach if not previously provided)

(including version/revision number)

Proposed references to be provided to applicants during examination: -:; ,.N.:. ,:o.:. .:n.:. .:e_ _ _ _ _ _ _ _ __

Learning MC-07567 Given a set of plant conditions, determine the required actions to be (As available)

Objective: taken in accordance with AOP 3573 Radiation Monitor Alann Response Question Source: Bank # 78805 Question History: Millstone 3 2004 NRC Exam Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 43.4 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 99 Ticr# 3 Knowledge ofEOP mitigation strategies Group #

KIA # GEN.2.4.6 Importance Rating 4.7 Proposed Question:

Current Plant Conditions:

• Core Exit Thermocouples read 1250°F.

• The crew is progressing through FR-C.l, Response to Inadequate Core Cooling.

• The crew has been unsuccessful at restoring High Head Safety Injection.

• The crew has been unsuccessful at depressurizing the secondary plant.

Which is the next major action that will be attempted, and how/why will the step be implemented?

a) One Rep is started at a time to maximize the time core cooling is provided, temporarily cooling the core with ReS loop crossover leg water.

b) All RCPs are started at once to provide symmetrical cooling of the core via heat transfer to the Steam Generators.

c) Both Pzr PORVs and both Reactor Head Vent Valves are opened to reduce ReS pressure low enough to allow low pressure safety injection from the SIL Accumulators.

d) Both Pzr PORVs are opened to reduce RCS pressure low enough to allow low pressure safety injection from the RHR Pumps.

Proposed Answer: ......;;...;.A,-_

Explanation (Optional): In FR-C.l, the third major action is to start all RCPs ("C" and "D" wrong). The background document states that starting an RCP (with adequate heat sink) will force two phase flow through the core, temporarily keeping it cool with crossover leg water. The steps are designed to start one RCP at a time, to extend the time the core is kept cool while the plant staff works at restoring either high head injection or the ability to depressurize the secondary plant ("A" correct, and "B" wrong). "B" is plausible, since the next major action is to start Reps, and symmetrical cooling is a basis for starting an RCP while on natural circulation. "C" and "D" are plausible, since the next major action strategy after starting RCPs is depressurizing the ReS by opening primary plant relief valves, and the basis is to get low head injection from accumulators and/or RHR pumps.

Technical Reference(s): FR-C.l (Rev. 017), steps 17 through 22 (Attach if not previously provided) WOG Bkgd Doc (Rev. 2) for FR-C.1, step 18 (including version/revision number)

Proposed references to be provided to applicants during examination: -'-N...;o-:"n...;e_ _ _ _ _ _ _ _ __

Learning MC-07457 Given a set of plant conditions, detenmne the required actions to be (As available)

Objective: taken per FR-C.l.

Question Source: Bank # 72428 Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 43.5 Comments:

Examination Outline Cross-reference: Level RO SRO Question # 100 Tier # 3 Knowledge of the bases for prioritizing emergency Group #

procedure implementation during emergency operations KIA # GEN.2.4.23 Importance Rating 4.4 Proposed Question:

A small break LOCA occurs, resulting in the following sequence of events:

1. Core Exit Thermocouples increase to 720°F.

2. The crew transitions to the appropriate Functional Response Procedure.

3. The crew commences depressurizing all intact SGs to 190 psig.

During the SG depressurization, the INTEGRITY critical safety function status tree turns RED.

Is the US required to direct a transition to FR-P.l, Response to Imminent Pressurized Thermal Shock? Why or why not?

a) Yes, since FR-P.l will be addressing an "extreme" challenge to a fission product barrier, while the current procedure is only addressing a "severe" challenge to a barrier.

b) Yes, since FR-P.1 will take actions that more directly protect a higher priority fission product barrier than that being addressed by the current procedure.

c) No, since the current procedure is addressing an "extreme" challenge to a fission product barrier, while FR-P.l would only be addressing a "severe" challenge to a barrier.

d) No, since the current procedure is taking actions that more directly protect a higher priority fission product barrier than that being addressed by FR-P .1.

Proposed Answer:

-D- -

Explanation (Optional): Status tree monitoring is based on RED being the highest priority "extreme" challenge to a fission product barrier, while an orange path (current core cooling tree color, above 718°F) is considered a lower priority "severe" challenge. Also the trees are prioritized based on which barrier they protect, with the priority being the clad first, then the RCS, and finally CTMT. For this event, it is expected that during the SG depressurization, the integrity tree may turn RED, since accumulators will inject as RCS pressure lowers, and cold accumulator water will flow past the RCS Tcold instruments, and into the vessel.

"An and nB" are wrong, since for this event, the operators are required to remain in C.2 ("C" plausible), to restore adequate core cooling, since the core cooling status tree directly protects the Clad, which is a higher priority barrier than the RCS ("D" correct, "c" wrong). "A" and "B" are plausible, since normally (when actions of one FR do not conflict with another FR), the RED path ("extreme challenge) on integrity would be a higher priority than the Orange path ("severe" challenge) fi)r core cooling.

Technical Reference(s): WOG Exec Vol. (Rev. 2) Writers Guide, Figure 12 (Attach ifnot previously provided) WOG Exec. Vol. (Rev. 2) Description, pages 23-26, and Figure 2 (including version/revision number) FR-C.2 (Rev. 013), Caution prior to step 11 Proposed references to be provided to applicants during examination: -:;..N.;.,;o:.;:n:.;:e_ _ _ _ _ _ _ _ __

Learning MC-04531 Diseuss conditions which require transition to other procedures from (As available)

Objective: EOP 35 FR-C.2.

Question Source: New Question History:

Question Cognitive Level: Comprehension or Analysis 10 CFR Part 55 Content: 43.5 Comments: