Text

Final - RO & SRO Written Examination with Answer Key (401-5 Format) (Folder 3)
	ML091070550
Person / Time
Site:	Millstone
Issue date:	02/19/2009
From:	Jordan A; Dominion Nuclear Connecticut
To:	Hansell S; Operations Branch I
	Hansell S
Shared Package
ML082600232	List: IR 05000336/2009301; ML082681139; ML090210018; ML090540381; ML090690320; ML090690400; ML090690487; ML090700775; ML090700829; ML091070550; ML091120006; ML091120062; ML091120075; ML091120121; ML091270250; ML091330638; ML091350479;
References
	08-0606F, 50-336/09-301, TAC U01634 50-336/09-301
	Download: ML091070550 (109)
	v • d • e

Dominion Nuclear Connecticut, Inc.

Mill'tonc I'ower Station Dominion Kopc Fcrry Road W<lrerford. CT 06185 FEB 1 9 2009 Mr. Samuel L. Hansell, Jr., Chief Serial No. 08-0606F Operations Branch, Region I MPS Lic/TC RO U.S. Nuclear Regulatory Commission Docket No. 50-336 475 Allendale Road License No. DPR-65 King of Prussia, PA 19406 DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 2 SENIOR REACTOR OPERATOR AND REACTOR OPERATOR FINAL WRITTEN EXAMINATIONS As agreed to by the NRC lead examiner and Dominion Nuclear Connecticut, Inc. examination developers, enclosed are the final written examinations and supporting documentation for the senior reactor operator and reactor operator. , "Written Examinations and Supporting Documentation" is being furnished in accordance with 10 CFR 55.40(b)(3) by an authorized representative of the facility.

Consistent with guidance contained in NUREG-1021, Examination Standard 201, Attachment 1, n

"Examination Security and Integrity Considerations , the written examinations and supportirrg documentation contained in Enclosure 1 should be ,withheld from public disclosure until after the examination has been completed. No redacted ver:sions are being supplied.

Should you have any questions regarding this submittal, please contact Mr. Jeff T. Spence at (860) 437-2540.

Sincerely, n Site ~ i cPresident i - Millstone

Serial No. 08-0606F Senior Reactor Operator and Reactor Operator Final Written Examinations Page 2 of 2

Enclosures:

1 Commitments made in this letter: None.

cc: (WIOenclosure)

U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406-1415, Ms. Carleen Sanders Project Manager, Mail Stop 08B3 U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station U.S. Nuclear Regulatory Comrrrission ATTN: Document Control Desk Washington, DC 20555

Serial No.08-0606F Docket No. 50-336 Enclosure 1 Final Written Examinations and Supportinq Documentation Millstone Power Station Unit 2 Dominion Nuclear Connecticut, Inc. (DNC)

Grade:

I I 86) I All work done on this examination is m A ..(rve peither given, nor received aid4

RO and SRO Exam Questions (No "Parents" Or "Originals1')

RO/ I-SRO Ques #: 1 Question ID: 10001 10 RO C]SRO 0 Student Handout? 0Lower Order?

Rev. 2 Selected for Exam Origin: Bank aPast NRC Exam?

A reactor trip occurred from 100% power. On the plant trip, the BOP notes that both 6900 volt AC buses are deenergized due to a failure to transfer to the RSST. All other electrical buses are energized from their normal source.

Which of the following describes the effect of the loss of power on the secondary system and the appropriate action to be taken?

A Condenser vacuum will be lost due to the loss of cooling to the steam jet air ejector. Per EOP 2525, Standard Post Trip Actions, the BOP must close the MSlVs and open 2-AR-17, Condenser Vacuum Breaker.

nB The loss of cooling water to the Gland Exhaust Condenser will cause a loss of condenser vacuum.

Per Appendix 4, the BOP must manually control steam seal pressure using MS-182A, Gland Seal supply bypass.

0C Excessive water hammer in the feedwater heaters will begin due to the loss of condensate flow. Once Standard Post Trip Actions are completed, the BOP must close all extraction steam supply valves on C0617.

D The loss of Main Feed supply to the SGs will cause hotwell reject to overflow the condensate surge tank. Once Standard Post Trip Actions are completed, the BOP must instruct lonics to secure makeup flow.

I

.Justificationp A - Correct; A loss of 6900 Volt AC buses results in a loss of condensate pumps. EOP 2525, step 11 states that if offs~tepower is lost or no condensate pumps are operating, then the condenser is NOT available; close both MSlVs and open 2-AR-17, condenser vacuum breaker.

B -Wrong, the turbine seals will still be maintained by gland seal steam until the MSlVs are closed. The gland seal supply valves will automatically throttle to maintain the appropriate steam seal pressure.

C - Wrong; On a loss of condensate pumps, the subsequent actions of EOP 2525 require the BOP to close the MSIVs, which will isolate steam flow to the feedwater heaters as much as is practical. Although this will NOT eliminate all hydraulic stress (lots of noise) seen by the feedwater heaters due to the sudden loss of condensate and feed flow, this is all that can be done at this time.

D - Wrong; Because condensate pump discharge pressure provides flow to the CST through the "reject" valve, the loss of 6900 volt AC buses causes a loss of condensate pumps; therefore the CST will NOT fill up.

References 1 LP CAR-00-C, R-3, C-5, Condenser Air Removal, (three linked locations)

EOP-2525, R-23, St. 10 (Inst & Cant)

Comments and Question Modification History Per NRC comments. the followina chanoes were made 02/18/09

1) Corrected justificstions for choices "2 & "B".

2) Removed "During the performance of EO-2525, Standard Post Trip Actions" from the stem.

3) Reworded choice "B" to reference actions taken per App. 4, controlling of gland seal manually

4) Replaced choice "C" required action (limit AFW flow based on high hotwell level).

5) Replaced choice " D required action (close the reject valve to the surge tank).

NRC WA System/E/A System E02 Reactor Trip Recovery Number EA1.l RO 3.7 SRO 3.7 CFR Link (CFR: 41.7 145.5 145.6)

Ability to operate andlor monitor components and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes and automatic and manual features as they apply to the Reactor Trip Recovery.

Page 1 of 268 Printed on 211912009 at 16:49

RO and SRO Exam Questions (No "Parents" Or "Originals")

ROl I-SRO Ques #: 2 Question ID: 8000066 KO E] SRO n Student Handout? [I Lower Order?

Rev. 0 @ Selected for Exam Origin: New aPast NRC Exam?

The plant had tripped from 100% power due an inadvertent MSI on Facility 1.

On the trip, one Pressurizer Safety valve opened slightly and remains partially open. All other plant equipment responded as designed. The crew completed EOP 2525 and has transitioned to the applicable event specific EOP.

The BOP has commenced a plant cooldown by opening " A Steam Dumprrurbine Bypass Valve (SD/TBV) 40%, using controller PIC-4216 on C-05 in manual mode.

However, the " A steam dump valve operator is failed such that the valve only opens 20%.

Which one of the following indications must the BOP use to verify the " A SDITBV is open 20%?

A The SDKBV "red" light on C-05 cB Controller PIC-4216 on C-05

@ C Foxboro IA SD/TBV system screen D The SDrrBV annunciator on C-05 ZstifLtirlp C - CORRECT; The Condenser Steam Dump Valves operator has remote feedback as to actual valve position, which can be read on (or controlled from) the Foxboro IA SDITBV control screen (or the equivalent PPC screen).

A -WRONG; A limit switch operated by valve stem movement energizes the "red" light on C05 when the valve is about 15% open.

B - WRONG, The controller on C-05 is DEMAND indication, NOT actual valve position.

D - WRONG; The SDlTBV annunciator on C-05 will give indication when the valves is about 15% open.

References I Loss-Of-Control-Power Operator Aid, R-I, C-0 LP ESA-01-C, Enaineered Safetv Features Actuation Svstem. Pa. 19 One-Line ~ i a g r a m oSteam f ~ u k ~ ~ l ~ u rBypass b i n e control system, CL242 Comments and Question Modification History I Rephrased the stem of the question to shorten it and to make it less confusing. 11111108 Revised to focus question on "control room" position indication capabilities. 0111312009 Fixed "typo" in stem, per NRC comments. 02/18/09 NRC KIA System/E/A System 008 Pressurizer (PZR) Vapor Space Accident (Relief Valve Stuck Open)

Number AA2.17 RO 2.5 SRO 2.7* CFR Link (CFR: 43.51 45.13)

Ability to determine and interpret the following as they apply to the Pressurizer Vapor Space Accident: Steam dump valve controller (position)

Page 4 of 268 Printed on 2/19/2009 at 16:49

RO and SRO Exam Questions (No "Parents" Or "Originals")

ROfl-IRO Ques#: 3 1 Question ID: 8000045 a RO C]SRO rJ Student Handout? Lower Order?

Rev. 0 @] Selected for Exam Origin: New Past NRC Exam?

The plant was at 20% power, downpowering for a refueling outage, when the following occurred:

- The plant was tripped due to a rupture of the charging header somewhere near the RCS loop.

- Two CEAs stuck out (fully withdrawn) on the trip.

- The Main Steam Header ruptured in the Aux. Feed Pump Room, on the steam supply to the Turbine Driven Aux. Feed Pump.

Present conditions 15 minutes post trip are:

- The crew has entered EOP-2540 for a Small Break LOCA and a Steam Line Break.

- Reactor power is stable at 5 E- 05% -

- RCS pressure dropped to, and is stable at 1500 psia. -

- RCS Tavg is stable at 460°F -

- The charging header has been isolated from the control room.

- Ruptured steam supply to the TDAFP has been isolated.

- SIAS, CIAS, EBFAS and MSI have been verified as fully actuated.

- All other plant equipment and conditions are as expected for this event.

The Reactivity Control Safety Function is

................................................................... m . . , . . . . . . . . . . .

A NOT met based on two CEAs fully withdrawn and no boron injection.

L B NOT met based on little or no boron injection or Xenon production.

C presently met based on the existing reactor power level and trend.

D presently met based on SIAS boron injection and stable RCS Tavg.

Justification )

C - Correct, EOP-2540 Reactivity Control Safety Function requires a reactor power level at or below 1 X 10 -04% and stable to meet "Condition 2" of the acceptance criteria A - Wrong; This is the requirement of EOP-2525 for more than one CEA not fully inserted on the trip. However, EOP-2540 looks for a shutdown reactor based on actual power level B -Wrong; There would have been little or no boron injection and not enough time for any appreciable Xenon production. However, these criteria do not apply for the Reactivity Control Safety Function They would come into play during accident recovery much later.

D -Wrong; If either accident had lowered RCS pressure 250 psi, this condition would help ensure the reactor remained shutdown However, there is no SI flow due to RCS pressure and the charging header rupture.

References I EOP-2540A, R-10, St. 3, Condition 2 (Inst)

EOP-2541, R-2, App. 2, Figure 3 - Pre-SRAS Minimum Required SI Flow Comments and Question Modification History Revised stem to be a 'fill in the blank.' Revised answers to allow each of them to complete the sentence. 11111/08 NRC KIA System/E/A System 009 Small Break LOCA Number EA2.32 RO 3.2' SRO 3.6* CFR Link (CFR 43.5 145.13)

Ability to determine or interpret the following as they apply to a small break LOCA: SDM Page 6 of 268 Printed on 2/19/2009 at 16:49

RO and SRO Exam Questions (No "Parents" Or "Originals")

Question ID: 8000002 &I KO SRO student Handout? [7Lower Order?

Rev. 0 m Selected for Exam Origin: New Ti Past NRC Exam?

The plant has tripped due to a Large-Break LOCA and the crew has been successfully mitigating the event using the applicable EOP.

The following plant conditions exist six hours into h e event:

- SIAS, CIAS, CSAS, MSI and SRAS have all been verified as completely actuated.

- CTMT pressure = 2.3 psig and lowering slowly.

- CTMT temperature = 220°F and lowering slowli.

- Reactor Vessel Level = 19% and stable.

- All other plant equipment is functioning as designed.

Then, at that time, a state wide blackout causes a loss of the RSST and the following conditions now exist:

- Facility 1 components are unavailable due to an electrical fault on bus 24C and/or 24E (24E is presently aligned to 24C).

- "B" EDG starts and its output breaker closes but the Fac. 2 Sequencer has failed at Sequence '0' and does NOT re-start any components.

Which of the following lists the pumps that are procedurally required to be placed in service and why?

A Service Water, RBCCW and HPSl pumps for core cooling.

Service Water and RBCCW pumps and CAR Fans for C'TMT cooling.

B Service Water, RBCCW and LPSl pumps for core cooling.

Service Water and RBCCW pumps and CAR Fans for CTMT cooling.

nc Service Water, RBCCW and HPSl pumps for core cooling.

Service Water, RBCCW, and CTMT Spray purnps for CTMT cooling.

0D Service Water, RBCCW and LPSl pumps for RCS cooling.

Service Water, RBCCW, and CTMT Spray purnps for CTMT cooling.

A - Correct; SW is the heat sink to RBCCW. RBCCW is the heat sink to the RCS because vessel level is too low to use the SGs as a heat sink. HPSl is required for flow through the core because LPSl cannot be used durina sum^ Recirc. Present CTMT Dressure and time dictate that CAR ians be used for CTMTcooling, not CTMT Spray.

B - Wrong; In sump recirc (SRAS) the LPSl pumps are not used for RCS or core cooling.

C - Wrong; With CTMT pressure less than 7 psig, CTMT spray would be secured by procedure.

D - Wrong; With CTMT pressure less than 7 psig, CTMT spray would be secured by procedure. Also, in sump recirc (SRAS) the LPSl pumps are not used for RCS or core cooling.

R e f e r e 4 EOP-2532, St 5, 13, 22, 23 and 60 Reworded all four choices to reduce number of words and improve readability.

Reworded the stem from "pumps that MUST be operating" to "pumps that are procedurally required to be placed in service" to eliminate confusion on what exactly is required.

Reworded choices " A - "D" to symetrically align (R-R, W-R, R-W. W-W), per NRC comments. 02/18/09 Reworded justifications for "C" & " D to match changes made on 32/18/09, per NRC comments 02/20/09 NRC KIA System/E/A System 011 Large Break LOCA Number EK3.03 RO 4.1 SRO 4.3 CFR Link ICFR 41.5 141.10 / 45.6145.13)

Knowledge of the reasons for the following responses as the apply to the Large Break LOCA: Starting auxiliary feed pumps and flow, EDIG, and service water pumps P a g e 9 of 268 Printed on 2/23/2009 at 10:30

RO and SRO Exam Questions (No "Parents" Or "Originals")

Question ID: 8054462 B RO SRO C1 student Handout? U Lower Order?

Rev. 1 Selected for Exam Origin: Mod Past NRC Exam?

The plant IS operating in MODE 3, Normal Operatirig Temperature and Pressure, when a B RCP BLEEDOFF FLOW HI annunciator alarms. Several seconds later, the annunc~atorclears and IS immediately followed by a B RCP BLEEDOFF FLCW LO annunciator.

Which one of the following describe the consequerrces of this sequence of alarms?

0 . 1..1..........11............IIIIII..mImm.n.......m..,,,.......m..,,,,.,,....,m,,

A The change in "B" RCP bleedoff flow will raise the differential pressure across the Upper and Middle seals.

B The "B" RCP must be secured due to the loss of seal flow and the potential for failure of the RCP seal.

C "B" RCP bleedoff flow must be diverted to the PDT to prevent exceeding temperature and pressure limits.

LJ D "B" RCP seal bleedoff flow will automatically divert to the PDT, resulting in a decrease in letdown flow.

B - Correct; If bleedoff flow to an RCP seal is blocked, it would cause damage to the seals. The applicable plant procedures require the RCP to be immediately secured, even if it involves a required plant trip (operation in MODE 1 or 2).

A - Wrong; The purpose of the Excess Flow Check valve is to stop the excessive flow from the affected RCP. When the bleedofl flow stops, pressure across the affected RCP's seal will equalize with primary pressure and differential pressure across each seal stage will go to zero.

C - Wrong; This is only true if the normal bleedoff flow path out o i CTMT is isolated (by something like a CIAS). This would occur if the check valve did NOT seat and the bleedoff flow exceeded the flow capacity of the system relief valve.

D - Wrong; A possible scenario if when the bleedoff check valve closed, flow was diverted to the PDT by way of the bleedoff line relief valve. However, although this relief valve may possibly lift on the initial high bleedoff flow "spike", it would be isolated from the affected RCP when the high flow check valve closes.

References ARP for Control Board alarm CA-21, "RCP 'B' Bleedoff Flow Lo". ARP-2590B-101, R-0, C - I , St 4 & 5 Comments and Question Modification History P

Added 'All' to the beginning of distractor C to provide clarification and to make the distractor more plausible. (Cannot align any

~ndividualRCP bleedoff to the PDT) 11111/08 Added MODE 3 to the stem to ensure the correct answer is not a partial answer (trip reactor and turbine) 1/13/09 Changed "undesirable effect" in stem to "consequences of'. 1/13/09 Distractors are plausible if the examinee is not familiar with how the bleedoff system operates. 1/13/09 Changed stem to provide indications vs. the consequences of the indications. This requires examinee to diagnose what happened.

This also makes the distractors more plausible if the examinee does not know how the system functions.

Also changed distractor A from 'isolating bleedoff flow from the other three RCPs' to 'raising bleedoff flow from the other three RCPs'.

NRC comment incorporated on 2/18/09 02/20/09; Reworded choices "A" & "D" per NRC comments 02/23/09; Fixed justification for Choice "DMbased on modificatiorls made on 02/20/09 NRC KIA System/E/A System 015 Reactor Coolant Pump Malfunctions Number AK2.07 RO 2.9 SRO 2.9 CFR Link (CFR 41.7 145.

Knowledge of the interrelations between the Reactor Coolant Pump Malfunctions (Loss of RC Flow) and the following: RCP seals Page 14 of 268 Printed on 2/23/2009 at 10:31

RO and SRO Exam Questions (No "Parents" Or "Originals")

ROI I-SRO Ques #: 6 Question ID: 8000004 [3;] RO SRO n Student Handout? Lower Order?

Rev. 2 Id Selected for Exam Origin: Mod Past NRC Exam?

A plant downpower was performed for Control Valve Testing.

Then a rupture of a charging pump discharge dampener requires charging and letdown to be secured.

The plant is being maintained constant at 90% power while corrective actions are being performed on the charging system.

Pressurizer level has peaked at 70%, but is now dropping at the steady rate of 5% per hour.

How long before pressurizer level lowers to the point where administrative requirements will require a plant trip?

m , , I . I . . , . . . " . , , . . . . . , . . . . . . , . . . m . . . , , m m . . , , , m m m . .

1_7 A 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months B 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months C 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months D 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months

.Justification 1 B - Correct; AOP-2512, Loss Of All Charging, requires the plant be tripped if pressurizer level drops to 10% below programmed Pressurizer level setpoint. At 90% power, programmed Pressurizer level has NOT started ramping down yet and is still 65%.

Therefore, a trip is required when level drops to 55%. At this rate, it will take; 70% - 55% = 15% drop required; 15% x 15 min/l% =

75 min.

A - Wrong; This is the time it would take level to drop below programmed level setpoint.

C - Wrong; This assumes a trip is required at 35%. This used to be the lower Tech. Spec. limit, but has been removed on a recent license change (now there is only an upper limit). When it was applicable, this would require action and eventual plant shutdown and/or trip.

D -Wrong; This assumes a trip is required when pressurizer level reaches 20%, which causes all heaters to automatically de-energize. There is an administrative requirement to shutdown upon loss of the Proportional Heaters, but NOT a plant trip.

References 1

[NOT provided to Examinees] AOP-2512, R-I, C-2. Note preced~ngSt. 3 1 and St. 3 1 Comments and Question Modification History Minor modification to stem wording of existing PZR level being at 70%, per NRC comments 02117/09 NRC WA SystemlElA System 022 Loss of Reactor Coolant Makeup Number AA2.04 RO 2.9 SRO 3.8 CFR Link (CFR: 43.5 145.13)

Ability to determine and interpret the following as they apply to the Loss of Reactor Coolant Pump Makeup: How long PZR level can be maintained within limits Page 16 of 268 Printed o n 211912009 at 16:49

RO and SRO Exam Questions (No "Parents" Or "Originals")

Question ID: 8054131 @ HO SRO Student Handout? n Lower Order?

Rev. 0 Selected for Exam Origin: Mod uPast NRC Exam?

The plant is on Shutdown Cooling using the " A LPSl pump. The "B" LPSl Pump is aligned for service on SDC with its handswitch in the NORMAL-AFTER-STOP position.

A Loss of Normal Power (LNP) surveillance is scheduled to be performed on Facility 2. The LNP surveillance involves opening the RSST supply breaker to bus 24D, in order to test the Emergency Diesel Generator (EDG) and the ESAS Sequencer's ability to restore specific vital loads.

Immediately after the LNP test is initiated, the " A LPSl pump breaker trips on overload due to a seized impeller.

Which one of the following statements is correct regarding the status of the "B" LPSl Pump breaker after opening the RSST supply breaker to bus 24D?

3A Because of the anti-pumping circuit, the undervoltage signal on ESAS must be reset in order to start the "B" LPSl pump.

B Because the "B" LPSl pump handswitch is in the "Neutral" position, it will automatically start on the applicable sequencer step.

nC Because the "B" LPSl pump is aligned for SDC use, it will automatically start on the applicable sequencer step.

D Because there is NO SIAS present on ESAS, the "B" LPSl pump must be manually started once 24D is reenergized.

Justification I D - Correct, There is no automatic restart of the LPSl pumps on an LNP, but they do get a load shed signal A - Wrong; Although the LPSl pump has an "anti-pump" circuit, it is only armed if an automatic start signal is present and the pump is then shutdown by a signal OTHER THAN a load shed. Therefore, it is NOT interlocked off under these conditions.

B - Wrong; The LPSl Pump hand switch position does NOT affect the LNP start signal. However, this does require the operator place the handswitch to the stop position before attempting to restart the pump.

C -Wrong; When the plant enters the shutdown mode, plant procedure steps jumper out some of the control interlocks that would automatically trip the LPSl pumps on certain switchyard (grid) events. The only "automatic" restart for a LPSl pump is from a SIAS.

These signals are not triggered by the LNP surveillance, nor do the control jumpers have any impact on the LNP load shed of the LPSl pumps.

LP SDC-00-C, R-4, Pg. 12; Causes of SDC pump trip and effect of LNP on running pump.

Comments and Question Modification History Reworded the stem to change the order of the LNP test and the loss of the " A LPSI Pump. The Facility 2 LNP test would NOT continue if the " A LPSl Pump were lost immediately before the test. 11111/08 Per CC; Changed "restartlrestarted" to "startlstarted". 01/06/09

1) Clarified status of "B" LPDl Pump in stem. NRC comment. 2/18/09

2) Reworded question to ensure answers match. NRC comment. 2/18/09

3) Reworded Choice B to match end of Choice C. NRC comment. 2/18/09 NRC KIA System/E/A System 025 Loss of Residual Heat Removal System (RHRS) 1 Generic KIA Selected 1 NRC WA Generic System 2.2 Equipment Control Number 2.2.12 RO 3.7 SRO 4.1 CFR Link (CFR: 41.10 145.13)

Knowledge of surveillance procedures.

Page 1 8 o f 268 Printed o n 2/19/2009 at 16:49

RO and SRO Exam Questions (No "Parents" Or "Originals")

Question ID: 8000065 KO SRO Student Handout? Lower Order?

Rev. 0 Selected for Exam Origin: New nPast NRC Exam?

.. Which of the following events would require the Reactor Operator to immediately manually trip the reactor?

I . . . . . . . . . . . . . . . . . . . . . . . . . . . . , . , . . . . . . . . . . . . . . , , . . . . . . , , . . . . . . , , . . . , , , , , , . . , , , , ,

nA At 100% power, during CEA testing, a Group 7 CEA slips to 172 steps followed by another Group 7 CEA, which slips to 144 steps.

B While operating at 100% power, Bus 24A is 10:jt on a fault and Bus 24C is energized by its associated EDG.

C At 50% power, shortly after the loss of the "A" RBCCW Pump, " A and "C" RCP Upper, Lower, and Thrust Bearings read 198°F.

D While maintaining power at 1 to 3% for warming the main steam lines, RCS temperature (Tavg) lowers to 524°F.

~ u s t i f i d C is correct. If any RCP bearing temperature exceeds 194'F, the RCP must be tripped; therefore, a manual Reactor trip is requi~ed.

The RO would be responsible for observing RCP temperatures during a loss of RBCCW event and is also responsible for tripping the reactor and affected pumps when bearing temperatures exceed the alarm setpoint.

"A" is incorrect. If one CEA is misaligned by >20 steps and anotiler CEA IS misaligned by greater than 10 steps, then the Reactor must be tripped immediately. In this case one CEA is misaligned by only 8 steps and the other is misaligned by >20 steps. This will require a power reduction, but will NOT require an immediate trip.

"B" is incorrect A loss of Bus 24A will result in the loss of two Cin: Water Pumps, which will likely require a downpower, but NOT an immediate trip. If two Circ Water Pumps are lost in ONE condenser, then the plant must be immediately tripped. In this case, the two Circ Water Pumps are in different condensers.

"DMis incorrect If RCS temperature is lower by 10°F during a sta~tup,an immediate plant trip is required. At 1-3% power, Tavg is approximately 530 -532". Even at its highest point, TCS temperature only lowered by 8"F, which does NOT require a trip.

Additionally, Tech Specs requires RCS temperature to be logged every 15 minutes if Tavg falls below 525'F while critical. Again, a trip is NOT required.

References 1 OP 2202, Reactor Startup Replaced or~g~nal KIA and quest~on Orig~nalKIA determ~nedto oe SRO level knowledge 1/13/09 No load Tavg is 530°F and full load Tavg is 569"F, making 3 % power Tavg approximately 531 2 "F. Changed temperature in Chioce D to 524°F per NRC comment. 2/18/09 Added power levels to "A" and "C" per NRC comments 02120109 NRC KIA System/E/A System 026 Loss of Component Cooling Water (CCW)

G e n e r l c X t Selected

~ 1 NRC KIA Generic System 2.1 Conduct of Operations Number 2.1.2 RO 4.1 SRO 4.4 CFR Link (CFR: 41.10 145.13)

Knowledge of operator responsibilities during all modes of plant operation.

Page 2 0 o f 268 Printed o n 212012009 at 14:39

RO and SRO Exam Questions (No "Parents" Or l'Originals'l)

Question ID: 8000067 131 HO C]SRO [IStudent Handout? Lower Order?

Rev. 0 Selected for Exam Origin: New Past NRC Exam?

The following initial plant conditions exist:

- 100% steady-state

- Channel "Y" Pressurizer Level and Pressure Control set up as the controlling channels Then, VR-21 deenergizes due to a problem with its static switch.

Which of the following describes the effect on the applicable components, assuming NO operator actions have been taken?

0A Channel "Y" pressurizer pressure input would fail low, causing pressure control to slowly raise actual pressurizer pressure.

B Channel "Y" pressurizer level input would fail low, causing pressurizer level control to slowly raise actual pressurizer level.

C All pressurizer heaters are deenergized, RCS pressure would lower to 2200 psia causing the backup heaters to reenergize.

[31 D All pressurizer heaters are deenergized and spray valve bypass flow would cause RCS pressure to continue to lower.

Justification 1 D - Correct; The Pressurizer Heater Selector switch is normally in the "Both" position, which means a loss of VR-11 OR VR-21 will cause all PZR heaters to deenerqize due to the failure of the heater low level cutout circuit. The recovery of the heaters requires the operators to de-select the failedlde-energized circuit (select Ch. "X" only) and reclose both Proportional heater breakers.

A - Wrong; PZR pressure input is powered by VA-20 (VIAC), NOT VR-21, a non-vital power supply.

B -Wrong; Although the Ch. "Y" PZR level control is powered from non-vital sources, the level transmitter (input) is powered by a VIAC.

C -Wrong; With VR-21 deenergized, the backup heaters are unavailable, regardless of operator or system actions. This is because the loss of VR-21 causes the High Pressurizer Pressure heater trip to fail in the "triggered" mode, which prevents the Backup heaters from being re-energized by operator OR control system action.

References I AOP-25048, R-3, C-11, St. 3.2 Loss-Of-Control-PowerOperator Aid, R-1, C-0 Comments and Question Modification History Rewrote Q# 90 per NRC comments, such that 2 choices do NOT de-energize PZR htrs 01/13/2009 NRC KJA System/E/A System 027 Pressurizer Pressure Control System (PZR PCS) Malfunction Number AA1.O1 RO 4.0 SRO 3.9 CFR Link (CFR 41.7 145.5 145.6)

Ability to operate and / or monitor the following as they apply to the Pressurizer Pressure Control Malfunctions: PZR heaters, sprays, and PORVs Page 22 of 268 Printed o n 211912009 at 16:49

RO and SRO Exam Questions (No "Parents" Or "Originals")

ROI I-SRO Ques #: 10 Question ID: 8000009 &I RO El SRO Student Handout? Lower Order?

Rev. 0 @] Selected for Exam Origin: New [I Past NRC Exam?

The plant is at 100% power, steady state, when a grid disturbance causes the main turbine to trip. Before the RO can trip the reactor, helshe notices that all CEAs are inserting. The BOP reports that all electrical buses are energized.

Which one of the following, by itself, would indicate that the ATWS Mitigation Circuit (i.e.; Diverse SCRAM System) has actuated?

.......................... m . m m , m m m m m . . m m , . . . ~ , . . . , , . . . , m . . . , . . . . , . . . m , m m m . , m m . ~ , m .

nA All four Turbine Trip Undervoltage Relays are tripped.

Both MG Set output contactors are open.

0c Both AFW Auto Start annunciators are in alarm.

D Both IMG Set 480 VAC supply breakers are tripped.

Justification 1 6 - Correct; The DSS actuation trips both MG set output contactors as an additional way to shutdown the reactor, separate from RPS.

A - Wrong; The Turbine Trip Undervoltage Relays are tripped whenever the CEDS busses are deenergized, regardless of the cause.

Although the same 10CFR50, App 65 law that requires an independent circuit to trip the reactor (DSS), also requires the trip of the main turbine on a reactor trip, this circuit is NOT actuated by the DSS.

C -Wrong; The AFAS alarm could be in due to the level shrink on a plant trip driven by a turbine trip. The load reject would cause a spike in SG pressure and result in a higher than expected shrink in SG level.

D - Wrong; The MG Set 480 VAC breakers do NOT get a trip sigqal from the Diverse Scram System.

References 1 LP CED-01-C, R-4, Pg. 7, 9, & 36 (excerpts)

Comments and Question Modification History Changed distractor 'D' from: Indication the PORVs opened and closed, to: One control channel NI is reading 50%. This is to ensure Distractor 'D' is absolutely incorrect, but plausible due the input to the Automatic Auxiliary Fedwater Actuation vs. the Diverse Scram System.

Changed Choice D from 'One control channel NI is reading >20%,' to 'Both MG Set 480 VAC supply breakers are tripped'. Based on validiator feedback.

Changed Choice " A from "All eight TCBs are open." to "All four Turbine Trip Undervoltage Relays are tripped." to remove any connection to coponents directly actuated by RPS. 01/13/09

2) Included in stem, ' all electrical buses are energized'. NRC comment. 2/18/09

3) Modified Chioce C slightly to more closely match the wording of the annunciator. NRC comment. 2/18/09 NRC KIA System/E/A System 029 Anticipated Transient Without Scram (ATWS)

Number EK2.06 RO 2.9* SRO 3.1' CFR Link (CFR 41.7 145.7)

Knowledge of the interrelations between the and the following an ATWS: Breakers, relays, and disconnects Page 24 of 2 6 8 Printed on 2/19/2009 at 16:49

RO and SRO Exam Questions (No "Parents1'Or "Originals1')

Question ID: 8054019 KO C SRO 0 student Handout? Lower Order?

Rev. 0 Selected for Exam Origin: Mod Past NRC Exam?

The following plant conditions exist:

- A SGTR occurred with all systems operating normally on the plant trip.

- EOP 2525 has been completed and EOP 2534 is being properly implemented.

- The RCS cooldown to 515°F has just been completed.

How would SIG pressures respond if the cooldowr to 515°F had been accomplished with a loss of the RSST at the time of the trip?

..................... rn m , , , . . . . . . . . . , , . . . . . , , m , . . . . . . , . , , . . . . . . . , , , , , . . . . . . , , , , . . . .

A SG pressure will be lower in both of the SGs because of the larger RCS Delta-T.

B SG pressure will be lower in the ruptured SG because it will be depressurized by the RCS leakage.

C SG pressure will be the same because the cooldown is always to the same RCS temperature.

D SG pressure will be higher in the ruptured SG because of the larger RCS Delta-T Justification I A - Correct; The large Delta-T results in a lower TC SG pressure most closely tracks TC.

B -Wrong; RCS leakage into the SG through a ruptured tube could act like "spray flow" into the pressurizer, causing SG pressure to drop more than expected. However, this effect is NOT possible I~the SG level is maintained >/= 40% per the EOP-2534.

C - Wrong, It is RCS Tcold that decides SG pressure. The cooldown is to a Thot of 515°F. On a loss of the RSST, RCPs will be lost Tcold will be lower in natural circulation than with RCPs running D -Wrong; The two loops should remain coupled as both SGs are being used in the cooldown Therefore, both loop Tcold temperatures should be about the same. Any differences in SG pressures would be based solely on the throttled position of the individual ADVs during the cooldown References I PT E34-01-S, Pg 12 Comments and Question Modification History Reworded Distractor 'B' from " the ruptured SG because the SGs are no lonqer hnked by the Main Steam header and the ruptured SG will be ..." to "...the ruptured SG because, after the MSlVs are closed, the ;uplured SG will be ..." This is to provide clarity to the distractor and still maintain plausibility. 11111/08 In the stem, changed "to less than 151°F" to "515°F". 1/13/09 Removed words after "RCS Delta-T" from distractors A and D. 1/13/09 Removed 'after the MSlVs are closed" from distractor C. 1/13/09 Reworded stem to remove reference to 'differences' in S/G pressures. This allows chioce C to remain a valid distractor. NRC comment 2/18/09 NRC K/A System/E/A System 038 Steam Generator Tube Rupture (SGTR)

Number EK1.03 RO 3.9 SRO 4.2 CFR Link (CFR 41.81 41.10145.3)

Knowledge of the operational implications of the following concepts as they apply to the SGTR: Natural circulation Page 26 of 268 Printed on 211 912009 at 16:49

RO and SRO Exam Questions (No "Parents1'Or "Originals")

ROI I-SRO Quei #. 12 1 Question ID: 8073999 @ RO SRO Student Handout? Lower Order?

Rev. 2 p]Selected for Exam Origin: Mod [I Past NRC Exam?

A startup is in progress with the plant at 8000 MWaIMTU, in MODE 1 at 6% power.

Then the plant tripped due to an Excess Steam Demand Event in the Enclosure Building.

Fifteen minutes after the trip, the crew has completed EOP 2525 and is transitioning to the event specific EOP.

The following plant conditions now exist:

Core Exit Thermocouple (CET) temperature is 420°F and stable.

Pressurizer pressure is 1380 psia and slowly rising.

Pressurizer level is at 10% and slowly rising.

All equipment responded as expected and the appropriate operator actions were performed.

Based on the existing plant conditions, which one of the following Technical Specification LCOs are NOT met?

aA The RCS cooldown is greater than the maximum allowed.

Reactor Coolant Temperature is less than the minimum allowed nC Moderator Temperature Coefficient is more positive than allowed.

D The Pressurizer is below the minimum water level allowed.

Justification I A - correct; The RCS cooldown has exceeded the Tech Spec limit of 100"Flhr; therefore LC0 3.4.9.1a applies. The cooldown rate is NOT an issue if it was caused by a LOCA or the mitigating action for a SG Tube Rupture. However, it does apply for an ESD event.

The cooldown rate limit applies whenever RCS temperature is greater than 220°F.

6 - wrong; The minimum RCS temperature is 51 5"F, but the spec only applies in MODE 1.

C -wrong; Even though the RCS has experienced a significant cooldown and a significant amount of Boric Acid has been added, the MTC will NOT be positive at a Middle of Life condition (8000 MWDIMTU).

D -wrong; The Pressurizer level Tech Spec LC0 is applicable in this MODE; however, there is no lower limit on Pressurizer level.

(The previous revision of this Tech Spec LC0 had a lower limit of 35% Pressurizer level.)

References I Millstone Unit 2 Technical Specification 3.4.9.1. (NOT provided.)

Comments and Question Modification History Replaced Distractor 'C'. Original was not plausible. 11111/08 Rewrote question to ensure it was at the RO level. 1/14/09 Changed 'below' in choice b to 'less than'. NRC comment. 2/18/09 NRC System/E/A System E05 Excess Steam Demand

' ~ ~ e n e rKIA i c Selected /

NRC KIA Generic System 2.2 Equipment Control Number 2.2.22 RO 4.0 SRO 4.7 CFR Link ICFR: 41.5 143.2 145.2)

Knowledge of limiting conditions for operations and safety limits.

Page 28 of 268 Printed on 211912009 at 16:49

RO and SRO Exam Questions (No "Parents" Or "Originals")

ROI 1-SRO Ques #: 13 Question ID: 8071926 HO SRO Student Handout? [7Lower Order?

Rev.

0 Selected for Exam Origin:

Mod nPast NRC Exam'?

Following a trip from 100% power due to loss of all feedwater, the following plant conditions exist:

Buses 25A and 25B are deenergized due to a failure to automatically fast transfer.

Bus 24E is aligned to Bus 24C.

Bus 24C is deenergized; the associated DIG wil NOT start (PEO dispatched).

"B" Aux Feedwater pump breaker tripped on fault. (PEQ dispatched)

The Terry Turbine tripped on overspeed and will NOT reset. (PEO dispatched)

The Condensate System is NOT in operation.

2 SIG level is 130 inches and lowering.

I SIG level is at 100 inches and lowering.

Trending indicates #I SG level will be at 70 inches within the next 10 minutes.

All other conditions are as expected.

Early implementation of Once-Through -Coolirlg m m m m m m m m m . . . . .

A will IVOT be necessary at this time because Feedwater may be restored prior to reaching 70 inches in either SIG.

B should be initiated now because the Condenser Steam Dumps are NOT available for heat removal.

nC will NOT be necessary at this time because both Atmospheric Dump Valves are available for heat removal from the SIGs.

D should be initiated now because only one train of HPSl is available for heat removal with the PORVs.

Justification I D is correct; Note prior to step 5 of EOP 2537 states:

Once through cooling should be initiated prior to SG wide range level reaching 70 inches if any of the following exists:

1. Main or Auxiliary Feedwater is NOT expected to be restored.

2. Less than two trains of HPSI, PORVs, or ADVs are available.

3. NO Charging Pumps are available.

Additionally, OP 2260 EOP User's Guide states that OTC should be initiated at 100" to ensure it is complete by the time SIG level reaches 70".

A is incorrect; Although it is a possibility that feedwater may be restored prior to reaching 70 inches in either SIG, with only one HPSI available, Once-through-Cooling must be initiated early to ensure adequate heat removal.

B is incorrect; Although the Condenser Steam Dumps are NOT available due to the loss of Condensate (MSIVs are closed), this is NOT a criteria for early initiation of Once-Through-Cooling.

C is incorrect; Although both ADVs are available for heat removal at this time, the loss of feed to the SIG will result in a loss of heat removal from the SlGs when inventory is depleted. Once Through Cooling must be initiated early to ensure adequate heat removal with only one HPSI Pump injecting.

References 1 EOP 2537, Loss of All Feedwater, note prior to Step 5 Comments and Question Modification History Changed #2 SG level from 235 inches to 150 ~nchesand # I SG level from 150 inches to 110 inches. This is to ensure the examinee realizes that 'early initiation' should begin now and not wait to see if components can be restored before 'early initiation' is attempted.

11111108 Lowered SG levels to 130" and loo", per NRC comments, to more effectively fit the knowledge requirements of an RO.

0111312009 The following objectives signifiy that an RO is required to know the conditions, actions, and bases for the initiation of once through cooling:

MB-5960, LOlT Given a set of plant conditions concerning a loss of all feedwater, determine if criteria for the following are met as specified in EOP 2537, Loss of All Feedwater:

B. initiation of once through cooling MB-05961. LOlT Describe the condition dependent actions and their bases for the following as specified in EOP 2537, Loss of All Feedwater:

A. initiation of once through cooling B. fully implememting once through cooling via PORVs.

NRC K/A System/E/A System E06 Loss of Feedwater Number EA2.2 RO 3.0 SRO 4.2 CFR Link (CFR: 43.5 145.13)

Ability to determine and interpret adherence to appropriate procedures and operation within the limitations in the facility's license and amendments as they apply to the Loss of Feedwater.

Page 30 of 268 Printed on 2/19/2009 at 16:49

RO and SRO Exam Questions (No "Parents" Or "Originals")

Question ID: 8080324 @ KO SRO n Student Handout? Lower order?

Rev. 1 Selected for Exam Origin: Mod UPast NRC Exam?

To conserve the vital batteries, EOP 2530, Station Blackout, requires specific loads be deenergized if a battery charger cannot be restored within 60 minutes.

Which of the following are required to be deenergized for this reason?

A Distribution Panels D-I I , 12, 21 and 22 cB Inverters 5 and 6 c lnverters 1,2, 3, and 4 D ESAS Actuation Cabinets 5 and 6 ust ti fie at ion 1 D is correct; Without vital AC, ESAS is very limited on what it can actuate. However, the cabinets are still deenergized at the last possible minute.

A is incorrect; Only specific non-vital DC loads are secured, based on key indications and components that, although NOT safety rated, are important for plant control.

B is incorrect; lnverters 5 and 6 will probably be secured to save the turbine battery, but are NOT secured by this procedure or at this time.

C is incorrect; lnverters 1 and 2 are NOT secured, due to the need to keep VA-10 and 20 energized, even though these VlACs are backed up by inverters powered by the turbine battery.

References I EOP 2530, Step 12, Comments and Question Modification History Rewrote question, per NRC comments, to evaluate knowledge of specific loads that are required to be deenergized. 01/13/2009

1) Changed choices A and D to noun name followed by number. NRC comment 2118109

2) Removed the phrase 'specific loads' from stem. NRC comment 2/18/09 NRC K/A System/E/A System 055 Loss of Offsite and Onsite Power (Station Blackout)

Number EA1.05 RO 3.3 SRO 3.6 CFR Link (CFR 41.7 145.5 145.6)

Ability to operate and monitor the following as they apply to a Station Blackout: Battery, when approaching fully discharged Page 33 of 268 Printed on 211912009 at 16:49

RO and SRO Exam Questions (No "Parents" Or "Originals")

RO/ I-SRO Ques #: 15 Question ID: 8000003 @ RO SRO n Student Handout? @ Lower Order?

Rev. 0 Selected for Exam Origin: New Past NRC Exam?

Why are the loads that start on ONLY a Loss of Normal Power (LNP) different than the loads that start on a Safety Injection Actuation Signal (SIAS) with a concurrent LNP?

A The load limit for the Diesel Generators is lower for an LNP ONLY event, than it is for an LNP concurrent with a SIAS, due to SlAS realignment of Service Water.

B Starting the Enclosure Building Filtration System (EBFS) fans during an LNP WITHOUT a concurrent SlAS will result in overheating of the EBFS charcoal filters.

C The additional components that start on an LNP concurrent with a SlAS are required to mitigate the consequences of a LOCA or ESD event inside CTMT.

D Components experience higher flows and temperatures in an accident situation, therefore, the starting sequence for specific components is changed for a SIAS.

Justification 1 C is correct. The components started on Sequence 3 on an LNP with a concurrent SlAS are. LPSl Pump, Containment Spray Pump, and Enclosure Buildina Filtration Fan. These comoonents are NOT needed for accident mitiaation durina an LNP ONLY situation.

A is incorrect. ~ l t h o u g h the Service Water system is realigned auring a SlAS to eliminate the heat from non-safety related components, the EDG load LIMIT for an LNP ONLY and an LNP with a concurrent SlAS is the same.

B is incorrect. Starting the Enclosure Building Filter Fan on an LNP ONLY would NOT result in overheating of the EBFS charcoal filters. It is simply unnecessary to start the EBFS fans for an LNP.

D is incorrect. While most components experience higher flows and temperature during an accident event, the starting sequence is maintained for both conditions. In Sequence 3, additional compenents are started for a SIAS, but the sequence remains the same.

References I LP ESA-01-C, Pg 31 LP ESA-01-C, EDG Load Sequence - No SlAS LP ESA-01-C, EDG Load Sequence - With SlAS Comments and Question Modification History Modified choice "6" and "D" per NRC comments. 01/14/2009

1) Reworded the question stem to eliminate the phrase 'NO components start on Sequence 3'. NRC comment. 2/18/09

2) Corrected justification for Choice B to match rewording. NRC comment. 2/18/09 NRC WA System/E/A System 056 Loss of Offsit<?Power Number AK3.01 RO 3.5 SRO 3.9 CFR Link (CFR 41.5,41.10 145.6 145.13)

Knowledge of the reasons for the following responses as they apply to the Loss of Offsite Power: Order and time to initiation of power for the load sequencer Page 36 of 268 Printed on 211912009 at 16:49

RO and SRO Exam Questions (No "ParentsttOr "Originalstt)

RO/ I-SRO Ques #. 16 Question ID: 8000005 @] 1 x0 1 SRO g,Student Handout? n Lower Order?

Rev. 0 m Selected for Exam Origin: New uPast NRC Exam?

'The plant is operating at 100% power with normal conditions when all Channel "C" safety related indications are suddenly deenergized. All other Control Room indications are energized.

Assuming all equipment responded as designed, based on these indications, what action must the Balance of Plant (BOP) operator take with regard to control of Steam Generator level?

............................ m m . m m . . m . m m m m . . . ~ . . . . . . . . . . . . . . . . . . . . . . . . m . m m . m . m . . m . .

A Place ONLY the " A Main Feed Pump speed control in MANUAL and raise speed.

m Verify "A" Main Feed Pump speed control is in AUTO and operating properly.

C Place BOTH Main Feed Pump speed controls in MANUAL and adjust speeds.

0D Start any two Auxiliary Feed Pumps, after the plant is manually tripped.

Justification 1 3 is correct. Although VA-30 is considered a "Vital" control power supply, it is the backup control power supply to the "A" MFP speed control. VR-11 is the 'main' power supply and it will continue to supply power to the feed pump speed control, which results in NO change to feed pump speed. The "A" MFP Trouble alarm will annunciate when VA-30 is lost, which requires the operator to monitor feed pump speed and SIG level.

A is incorrect. This would be true if VR-11 was not maintaining power to the control system. However, because the control circuitry is NOT impacted by a loss of VA-30, there is no need to place the A" MFP speed control in MANUAL.

C is incorrect. This is the required SGFP action for a failed stearn flow transmitter, which is NOT powered by VA-30, but VR-11/21.

Therefore, there is no need to place both feed pump controls in MANUAL. The loss of VA-30 will NOT impact either Feed Pump.

D is incorrect. If VR-11 were also lost, the " A MFP would trip on loss of control power. However, the feed pumps are NOT affected, therefore; the plant, will NOT be tripped.

References 1 LP MFW-01-C, Pg 37, Table of Power Supplies ARP-2590D-001, SGFP A Trouble alarm Comments and Question Modification History Fixed typo in Distractor 'C' Changed 'Main Fed Pump' to 'Main Feed Pump' 11111/08 Modified question and distractors based on NRC comments. 1/13/09

1) Fixed typo in stem. NRC comment. 2/18/09

2) Eliminated reference to any alarms that may lead examinee to think VR-11 is lost. NRC comment. 2/18/09 NRC KIA System/E/A System 057 Loss of Vital AC Electrical Instrument Bus Number AA1.03 RO 3.6* SRO 3.6 CFR Link (CFR 41.7 145.5 145.6)

Ability to operate and I or monitor the following as they apply to the Loss of Vital AC Instrument Bus: Feedwater pump speed to control pressure and level in S/G Page 39 of 268 Printed on 211 912009 a t 16:50

RO and SRO Exam Questions (No "Parents" Or "Originals")

RO/ 1-SRO Ques #: 17 Question ID: 8000007 @] RO SRO Student Handout? n Lower Order?

Rev. 0 Selected for Exam Origin: New [I Past NRC Exam?

The plant is operating at 100% power in a normal configuration. The AC supply breaker for the "B" Battery Charger suddenly opens. Several minutes later, the "125 VDC Battery 201 B Undervoltage" annunciator alarms.

Which of the following correctly describes the impact of this condition over the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months if the Charger is NOT restored?

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . m ,

A The associated instrumentation will become increasingly inaccurate.

1B Facility 2 AC breakers can NO longer be operated remotely.

C Battery voltage is degraded, but is still adequate for all DC loads.

D Letdown Isolation Valve, CH-515,will automatically isolate.

Justification I C is correct. The undewoltage alarm is a annunciated to alert the operator to a degrading condition. The alarm is NOT meant to provide a warning of malfunctioningequipment. Therefore, DC loads are NOT impacted when this alarm annunciates. The Station Batteries are analyzed to provide DC voltage to required loads for at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following the loss of its associated charger.

A is incorrect. The output from the inverters is NOT impacted by the lower DC voltage; therefore, connected instrumentation is NOT affected.

B is incorrect. Control power is NOT lost to AC breakers until the battery is depleted. The battery is designed to last for at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months which is sufficient to maintain adequate voltage to all AC breakers for the 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months stated.

D is incorrect. Letdown Isolation Valve, CH-515, will NOT close. DC power will be maintained.

References I LP LVD-01-C,Pg 12, Design Basis of Batteries APR-2590F, 1 2 5 ~ a~t c ~ e 201 r ~~B Undervoltage alarm Comments and Question Modification History Slight change to stem Changed 'Ten minutes...' to 'Several minutes...'. Ten minutes may be considered too soon for the alarm, which may cause the examinee to think the problem is much more severe.

Changed '...at an indicated 126 volts.' to '...due to the lowering votlage.' The actual alarm setpopint may be confusing to the examinee if helshe believes that the alarm setpoint is different. (NOT required to be memorized.)

In Distractor 'D', changed 'overvcurrent' to 'undervoltage'. Makes the distractor more plausible. Examinee is more inclined to believe components trip on undewoltage than overcurrent. 11111108 NRC KIA System/E/A System 058 Loss of DC Power Number ,4142.02 RO 3.3* SRO 3.6 CFR Link (CFR: 43.5 145.13)

Ability to determine and interpret the following as they apply to the Loss of DC Power: 125V dc bus voltage, lowlcritical low, alarm Page 42 of 268 Printed o n 2/19/2009 at 1 6 5 0

RO and SRO Exam Questions (No "Parents" Or Originals")

Roi I-sRo Ques #: 1 8 1 Question ID: 79976 @ KO 1SRO 1Student Handout? @ Lower Order?

Rev. 2 a Selected for Exam Origin: Bank Past NRC Exam?

The plant is in normal operation at 100% power. The "LOAD LIMIT LIMITING" light is energized on the C0617 EHC insert.

Which of the following describes the response of the turbine controls to changes in grid frequency.

A The turbine will respond only to a significantly !ow grid frequency; the turbine speed control unit demand will cause the control valves to open.

L B The turbine will NOT respond to grid frequency changes in this mode; load limit prevents the turbine speed control unit demand from opening or clasing the control valves.

C The turbine continuously responds to grid frequency changes in this mode, the control valves open and close in response to speed control unit demand to maintain 60 Hz.

D The turbine will respond to a significantly high grid frequency, the turbine speed control unit demand will cause the control valves to close.

Justification 1 D is correct. The load reference signal and the speed error signal are combined in such a way that even with a "valves wide open" reference signal (+5VDC), the control valves (CV) will close proportionately as speed increases from 100% (60 Hz) to 105% (63 Hz).

A is incorrect. The control valves will NOT open due to a change in load; however, the control valves are able to automatically close proportionally to a load change.

B is incorrect. The control valves WILL respond to an increase in grid frequency by closing proportionally.

C is incorrect. The control valves will NOT open due to a change in load; however, the control valves are able to automatically close proportionally to a load change.

References 1 LP MTC-01-C, Pg 40 & 42, Main Turbine. Controls on "Load Liwit" (NOComments or Question Modification History at this t i r n e l 7 NRC KIA System/E/A System 077 Generator Voltage and Electric Grid Disturbances Number AK2.07 RO 3.6 SRO 3.7 CFR Link (CFR: 41.4, 41.5, 41.7,41.10 145.8)

Knowledge of the interrelations between Generator Voltage and Electric Grid Disturbances and the following: TurbineIgenerator control.

Page 45 o f 268 Printed o n 211912009 at 1650

RO and SRO Exam Questions (No "Parents1'Or "Originals")

mi I-SRO Ques #: 19 Question ID: 2100001 [3] R 0 SRO [IStudent Handout? [aLower Order?

Rev. 1 Selected for Exam Origin: Bank Past NRC Exam?

The plant is operating at 100% power and the monthly CEA operability surveillance is in progress. The Reactor Operator (RO) has just finished inserting CEA #I (Group 7) six steps from the fully withdrawn position, when it suddenly slips to the 126 step position.

Which one of the following combinations of CEAPDS and PPC position indications matches those that would be displayed on C04 under these conditions?

....................... rn m , , . . . . . . . . . . , , . . , , . , , , . . . . . . . . . . . . . . . . , . . . , , m .,,..,,...,,

A CEAPDS indicates 126 steps Computer indicates 126 steps 4 B CEAPDS indicates 126 steps Computer indicates 174 steps

~1 C CEAPDS indicates 174 steps Computer indicates 126 steps D CEAPDS indicates 174 steps Computer indicates 174 steps Justification I B - Correct; CEAPDS will display the slipped CEA position because it monitors the reed switches for the individual CEA. However, the PPC will only display a change in CEA position if the CEDM was actually "pulsed" to move the CEA.

A - Wrong; The PPC will not sense the CEA has moved because the CEA slipped to 126 steps and the CEDM was not pulsed to move it.

C -Wrong; This choice has the two indications reversed, if the concept is partially understood.

D -Wrong; This choice assumes both indications need to see the CEDM pulsed to move the CEA References I AOP-2556, Pg. 4, Discussion on CEA Pulse Counting possible error.

LP CED-01-C, Pg 26, explanation of Pulse and Reed CEA position display.

/ N O Comments or Question Modification History at this time. 1 NRC KIA System/E/A System 003 Dropped Control Rod Number AK2.03 RO 3.1* SRO 3.2" CFR Link (CFR 41.7 145.7)

Knowledge of the interrelations between the Dropped Control Rod and the following: Metroscope Page 47 of 268 Printed o n 2/19/2009 at 16:50

RO and SRO Exam Questions (No "Parents" Or "Originals1')

Question ID: 8600020 RO L]SRO n Student Handout? n Lower Order?

Rev. 0 @ Selected for Exam Origin: Mod aPast NRC Exam?

The plant is at 100% power, steady state, all equi~mentfunctioning normally.

Then, the high voltage power supply to the Channel " A Wide Range Nuclear Instruments fails, such that the channel is now reading eight (8) decades lower.

Which of the following describes the change in plantlcomponent conditions due to this power supply failure?

[7 A The Zero Power Mode Bypass will arm on Channel "A".

U B The Power Trip Test Interlock (PTTI) will initiate on Channel " A .

C The PDlL alarm and interlock on CEAPDS wilt be bypassed.

D The Level 1 and Level 2 Bistables will reset or1 Channel " A .

Justification 1 D - CORRECT: The Level 112 bistables will RESET on this channel when the signal drops below 1 X 10-4% The channel would now be reading 1 X 10-6% power.

A -WRONG: When a Wide Range NI channel drops below 1 X 10-4%, Level 112 bistables will ALLOW arming of the Zero Power Mode Bypass, but a bypass key on the channel (not normally in place) must be turned to the "bypass" position. This key would only be in place if testing were being done on the channel.

B - WRONG, The PTTl interlock would indeed be armed for this channel, IF the failed detector power supply were on a Linear Channel.

C - WRONG: The PDlL bypass would activate on this failure, but ALL four channels must activate for the applicable interlocks to be affected.

References 1 NIS-01-C, Rev 3, Ch. 2, Pg 15 of 57, Pg 31 of 57 ARP 2590C-092, Rev 000

~ N OComments or Question Modification History at this- t i m e . 1 NRC WA System/E/A System 032 Loss of Source Range Nuclear Instrumentation Number AA2.01 RO 2.6 SRO 2.9* CFR Link (CFR: 43.5 145.13)

Ability to determine and interpret the following as they apply to the Loss of Source Range Nuclear Instrumentation: Normallabnormal power supply operation Page 50 of 268 Printed on 2/19/2009 at 1650

RO and SRO Exam Questions (No "Parents" Or "Originals")

ROI I-SRO Ques #: 21 Question ID: 8055940 [31 RO 1 1SRO n Student Handout? n Lower Order?

Rev. 0 @] Selected for Exam Origin: Mod [I Past NRC Exam?

An Operator is at the Aerated Waste Panel, about to start a radwaste discharge. All applicable administrative requirements have been properly completed and verified up to this point. The Operator starts the sample pump and the "AERATED WASTE EFFLUENT RADIATION HI" annunciator clears.

Additionally, the "RM9116 LOSS OF FLOW" annu~ciatorclears and then immediately re-alarms.

Then, the Operator turns the control switch for the first discharge valve directly to OPEN but the red light does NOT light. The Operator then turns the hand switch for each discharge valve in the CLOSE direction first, then to OPEN. The red lights on both valves energize and, a couple seconds later, the green lights on both valves go out.

The Operator continues the discharge with the following observations:

- Flow indication on the discharge flow recorder is about half what was seen on the last discharge.

- Aerated Waste Monitor Tank level is lowering at a rate expected for the lower discharge flow.

- The Aerated Rad. Waste Discharge filter delta-P is much higher than that seen on the last discharge.

- All recorded parameters appear to be within acceptable ranges and tracking normally.

Which of the following describes the impact of these conditions?

aA The abnormally low discharge flow will cause the discharge to isolate on low sample flow.

The abnormally low discharge flow will result in a more conservative radiation monitor reading.

C The discharge will isolate ONLY on a loss of control power to the discharge valves.

D The discharge will isolate on a Rad Monitor failure, but will NOT isolate on a Rad Monitor High alarm.

Justification I C - Correct; Starting the sample pump often clears the low sample flow alarm but then triggers the high sample flow condition as sample flow stabilizes. Although this sample flow fluctuation w o ~ l dnormally isolate the discharge, over-riding a high sample flow alarm also prevents a low sample flow condition from closing the valves. The sample flow and radiation alarms on PlOPS must be acknowledged and cleared BEFORE the discharge valves are opened, or opening them means they have been "over-ridden" open.

Based on the conditions stated, the discharge valves HAVE BEEN over-ridden open, and will NOT close for ANY alarm condition triggered by the Rad. Monitor.

A - Wrong; Because the discharge valves were overridden, they will NOT close. The Rad. Monitor sample flow is a separate slip stream driven by a sample pump. The sample flow would NOT be effected by the discharge flow rate.

B - Wrong; Because the discharge valves were overridden, they will NOT close due to a high radiation monitor reading; therefore, the Rad Monitor is inoperable. The low discharge flow may result in a greater sampling of the rad. waste as it passes by the sample pump suction. However, this potential "over-sampling" could only result in an artificially high radiation reading, which is more conservative.

D - Wrong; The discharge valve over-ride is designed to allow for a rad. waste discharge with the rad. monitor de-energized.

Therefore, a radiation monitor failure of any kind will NOT secure the discharge.

References I RMS-00-C, Radiation Monitoring, Rev. 7, Ch. 2 ARP 2593A, Rev 1, Ch. 3 ALR-04-C, Aerated Liquid Radwaste, Rev. 3, Ch. 1 Comments and Question Modification History Reworded the stem to eliminate the second operator Question now requires the examinee to determine whether or not the rad monitor is operable or not. 11111/08

1) Fixed typo in question. NRC comment. 2118109

2) Placed a period at the end of Choice d. NRC comment. 2/18/09

3) Reworded question to ensure it encompasses all answers. NRC comment. 2/18/09 Modified Choice A. Was originally correct. Validation. 2/18/09 NRC WA System/E/A System 059 Accidental Liquid Radwaste Release Number AK3.03 RO 3.0 SRO 3.7 CFR Link (CFR 41.5,41.10 145.6145.13)

Knowledge of the reasons for the following responses as they apply to the Accidental Liquid Radwaste Release: Declaration that a radioactive-liquid monitor is inoperable Page 53 of 268 Printed on 211912009 at 16:50

RO and SRO Exam Questions (No "Parents1'Or "Originals1')

ROI I-SRO Ques #: 22 Question ID: 8000055 HO SRO Student Handout? Lower Order?

Rev. 1 @ Selected for Exam Origin: New Past NRC Exam?

A control room evacuation is required due to toxic gas. All required immediate actions of AOP 2551, Shutdown from Outside the Control Room, have been performed and the operating crew has gathered at C-21, Hot Shutdown Panel.

Which of the following describes the component location and method of operation, as required by procedure, for aligning and controlling auxiliary feed flow to the Steam Generators?

U A AUX Feed Water Control Transfer switches on C-21 are in "LOCAL" Aux Feed OverridelManlStarVResethandswitches on C-21 are in "Pull-To-Lock" Aux Feed Pump Emergency OverrideIReset switches on C-05 are in "NORM" Aux Feed Reg Valve controllers on C-21 are in "M" Operate Aux Feed pumps and Aux Feed Reg Valves as desired from C-21 B AUX Feed Water Control Transfer switches or: C-21 are in "REMOTE" Aux Feed OverridelManlStartlReset handswitches on C-21 are in "Pull-To-Lock" Aux Feed Pump Emergency OverridelReset switches on C-21 are in "OVRD" Aux Feed Reg Valve controllers on C-21 are in "M" Operate Aux Feed pumps and Aux Feed Reg Valves as desired from C-21 C Aux Feed Water Control Transfer switches or, C-21 are in "REMOTE" Aux Feed OverridelManlStartlReset handswitches on C-05 are in "START" Aux Feed Pump Emergency OverrideIReset switches on C-21 are in "OVRD" Aux Feed Reg Valve controllers on C-21 are in " A Operate Aux Feed pumps and Aux Feed Reg Valves locally as desired D Aux Feed Water Control Transfer switches on C-21 are in "LOCAL" Aux Feed Override/Man/Start/Resethandswitches on C-21 are in "Start" Aux Feed Pump Emergency OverrideIReset switches on C-21 are in "OVRD" Aux Feed Reg Valve controllers on C-21 are in " A Operate Aux Feed pumps and Aux Feed Reg Valves locally as desired Justification 1 B is correct. AOP 2551 requires tripping the plant and both main feed pumps. In order to now feed the SGs, all switches on C-21 to be aligned such that the entire Auxiliary Feed System can be operated from C-21.

A is incorrect. If the Aux Feed Water Control Transfer switches on C-21 are in "LOCAL", NO Aux Fed components can be operated from C-21 C is incorrect. If the Aux Feed Reg Valve controllers on C-21 are in "A", then feed flow cannot be manually controlled. It doesn't matter what position the Aux Feed OverridelManlStarVResethandswitches on C-05 are in as long as the Aux Feed OverridelManlStaNReset handswitches on C-21 are in "Pull-To-Lock" (not listed). Additionally, the Aux Feed Pumps and Aux Feed Reg Valves would be operated from C-21, but MAY be operated locally.

D is incorrect. If the Aux Feed Reg Valve controllers on C-21 are in "A", then feed flow cannot be manually controlled. Additionally, If the Aux Feed Water Control Transfer switches on C-21 are in "LOCAL", NO Aux Feed components can be operated from C-21 References I AOP-2551, Pg. 16, Step 1 . 6 ~

Comments and Question Modification History Modified question to match WA NRC comment. 1/14/09 NRC WA System/E/A System 068 Control Room Evacuat~on Number AK3.01 RO 3.9 SRO 4.2 CFR Link (CFR 41.5,41.10 145.6 145.13)

Knowledge of the reasons for the following responses as they apply to the Control Room Evacuation: System response to reactor trip Page 58 of 268 Printed on 211912009 at 1650

RO and SRO Exam Questions (No "Parents" Or "Originals")

m i I-SRO Ques #. 23 Question ID: 8000061 RO n SRO u Student Handout? Lower Order?

Rev. 0 @ Selected for Exam Origin: New UPast NRC Exam?

The plant has just started up from a refueling outage and is stable at 30% power on a secondary chemistry hold.

Then, DC bus 201B de-energizes due to a bus fault, resulting in the following conditions:

- Both MSlVs close

- The "B" main Steam header ruptures in containment

- 248 and 24D are de-energized (along with all lower voltage busses powered by them)

- Facility One SIAS, CIAS, EBFAS, MSI and CSAS have all fully actuated

- All other plant systems and components that have power are functioning as designed.

The crew is evaluating numerous alarms and indications caused by the power loss and subsequent ESD.

Which of the following alarm indications will require contingency actions be taken to prevent exceeding a design limit?

......................... .............................................~...........

nA C05 alarms indicating an ESD on #2 SG and C08 alarm indicating VR-21 is de-energized.

C0213 alarms indicating RCS Th and Tc are abnormally low and both Boric Acid Pumps are de-energized.

C C04 alarms indicating Facility One Aux. Feedwater has actuated and C08 alarm indicating loss of DV-20.

rg D C01 alarms indicating CTMT Spray has actuated and C01 indicating only two CAR fans and one CS pump are operating.

Justification I C - CORRECT; All alarms and indications mentioned in the four choices are expected for the given event, a loss of DC bus 2018 and subsequent ESD on the "0" Main Steam header. However, Choice "C"information indicates ~uxiliary Feedwater will be feed the affected steam generator. The Design Basis ESD in CTMT states that ALL feed to the affected steam generator must be secured within 30 minutes to meet the design criteria for CTMT Integrity. In this criteria, only one facility of ESAS equipment is assumed to be functioning and available.

A - WRONG; VR-21 is deenergized, based on the given event. However, this would prevent the "B" Atmospheric Dump Valve (ADV) from being operated from the control room. If the other steam header was ruptured, this would be the correct choice, as it would require immediate action to get an operator to C21 (Remote Shutdown Panel) to control RCS temperature when the affected SG boils dry (thus preventing PTS)

B - WRONG; This gives indication of an excessive cooldown of the RCS with a potential problem with boric acid injection. However, the other facility of power is available to allow automatic alignment of a boric acid source to the remaining charging pump, which is sufficient (although not optimum) to meet "reactivity control". Procedure steps will ensure additional boron injection is aligned, but this is above the required amount.

D -WRONG; One facility of CTMT Cooling and Pressure Control is certainly NOT optimum during and ESD, but it is designed to be sufficient to maintain CTMT Integrity, provided all feed is secured to the affected SG in the required time frame.

References 1 OP-2260, Pg. 43; ESD Mitigation Requirements and Critical Tasks Comments and Question Modification History Per exam validation, modified Choice "B" to de-energize both "Boric Acid Pumps" instead of both "Gravity Feed Valves". 02/17/09 NRC KIA System/E/A System 069 Loss of Containment Integrity 1 Generic WA ~ e d NRC KIA Generic System Emergency Procedures /Plan Number 2.4.45 RO 4.1 SRO 4.3 CFR Link (CFR: 43.5 145.3 145.12)

Ability to prioritize and interpret the significance of each annunciator or alarm.

Page 61 of 268 Printed on 211 912009 at 16:50

RO and SRO Exam Questions (No "Parents" Or "Originals")

Question ID: 8000056 M HO SRO 0 student Handout? n Lower Order?

Rev. 0 E Selected for Exam Origin: New Jr Past NRC Exam?

The plant was at 100% power when a leak developed on the charging header connection to RCS loop one.

The crew attempted to isolate CVCS from the leak, but charging header isolation to loop one, CH-519, would NOT close. The leak was subsequently isolated from CVCS by closing Charging Header Isolation, CH-429.

The leak subsequently degraded to a rupture, resulting in a Small-Break LOCA.

The following conditions now exist:

- RCS pressure is 1400 psia and lowering slowly.

- The crew has transitioned to EOP-2532, Loss Of Coolant Accident.

- The US has determined that there is NO injection flow.

- All other plant equipment is operating as designed for the present plant conditions.

Which of the following alignments would raise injection flow?

n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . m . . . , . . m ~ , , . . . . . , , . . . . , , . . . . , , . . . ~ , , . m m . . , , m m m . , , . .

A Align the charging pumps discharge to the " A HPSl injection line.

B Align the " A HPSl pump to inject through the auxiliary spray line.

C Align the charging pumps to inject through the auxiliary spray line.

aD Align the charging pumps to discharge to the RCS loop two header.

I tificatin A - CORRECT; SB-LOCA, has the potential to cause RCS inventory to be lost faster than RCS pressure. With RCS pressure > 1250 psia (HPSI shutoff head), the charging pumps are the only pumps capable of injecting into the RCS. Based on the leak location and the failure of CH-519 to close, all three normal charging header injection paths are lost. Therefore, the alternate injection path, through "A" HPSl injection line, must be used.

B - WRONG; This path could be used if RCS pressure were lower and the leak were in a different location.

C - WRONG; This path would lower RCS pressure and allow HPSl to inject, but the leak location and CH-519 failure prevents it.

D - WRONG; The RCS loop injection lines are separate, and protected by check valves and isolation valves, but the leak location and failure of CH-519 eliminate this option.

References I CVCS One-Line Diagram OEP-2541, Pre-SRAS SI Flow Curve F o C o m m e n t s or Question ~ o d i f i c a t G History at this t i m e - 1 NRC KIA SystemlEIA System 074 lnadequate Core Cooling Number EAI.09 RO 3.7 SRO 3.8 CFR Link [CFR 41.7 145.5 145.6)

Ability to operate and monitor the following as they apply to a lnadequate Core Cooling: CVCS Page 63 of 268 Printed on 211912009 at 16:50

RO and SRO Exam Questions (No "Parents" Or "Originals")

ROl I-SRO Ques #: 25 Question ID: 8000012 @ RO SRO Student Handout? @ Lower Order?

Rev. 0 k Selected for Exam Origin: New 0Past NRC Exam?

The plant is at 100% power, steady state, when a normally scheduled primary sample shows high fission product activity in the RCS.

Which of the following describes a procedurally required action and why that action must be accomplished?

rn m . , . . . . , . . . . . . . . , . . . , . . . . . . . . . . . . . , . . . . . . . . . . . . * . . . . . . . rn m A Place a second purification demineralizer in parallel with the in-service purification demineralizer.

The second demineralizer will double the rate of activity removal from the RCS.

B Start an additional Charging Pump and adjust the Letdown Flow Controller bias to raise Letdown flow.

The additional flow will provide a higher rate of removal of activity from the RCS.

nC Divert Letdown flow to the Clean Waste System and perform a continuous blended makeup to the VCT.

Diverting Letdown will allow the activity to decay; a blended makeup will dilute the activity in the RCS.

D Replace the 0.5 micron Letdown pre-filter and post-filter with the fine mesh, 0.1 micron filters.

The finer mesh of the 0.1 micron filter will filter out smaller particles and reduce RCS activity faster.

Justification I B - Correct; Starting an additional Charging Pump and raising Letdown is the action required by AOP 2511, High RCS Activity A -Wrong; Placing an additional demineralizer in parallel with the first demineralizer will NOT double the removal rate of activity from the RCS. Additionally, the two spare demineralizers are used for other purposes; the first one is for removing Lithium form the RCS; the second one is for removing Boron form the RCS at end of life.

C - Wrong; Diverting Letdown to Clean Waste and performing a blended makeup to the VCT will dilute the activity in the RCS; however, 1) This is NOT procedurally directed, 2) Filling the Clean Waste System would require discharging additional activity to the environment needlessly, 3) The expense of processing the water would be financially prohibitive.

D -Wrong; Although this method will remove activity from the RCS more quickly, the finer mesh filters clog very quickly which will require them to be changed more frequently, which will result in needless addition exposure and an additional expense for disposal of the filters.

References 1 AOP-2511, Step 3.3 Comments and Question Modification History Modified question based on NRC comments. 1/14/09 I

Corrected minor "tyos" in choice " A & "DM.Also, reworded Choice "B", second sentence, per NRC comments. 02/16/09 NRC WA System/E/A System 076 High Reactor Coolant Activity Number AK3.05 RO 2.9 SRO 3.6 CFR Link (CFR 41.5,41.10 145.6 145.13)

Knowledge of the reasons for the following responses as they apply to the High Reactor Coolant Activity: Corrective actions as a result of high fission-product radioactivity level in the RCS Page 66 of 268 Printed on 211 912009 at 1650

RO and SRO Exam Questions (No "Parents" Or "Originals")

ROI I-SRO Ques #: 26 Question ID: 8680010 RO SRO U Student Handout? 5 Lower Order?

Rev. 0 Selected for Exam Origin: Mod Past NRC Exam?

The plant was manually tripped from 100% power due to a rupture of the " A Main Steam Header in the containment.

On the trip, VA-20 was lost due to a fault on the bus.

All other plant equipment is operating normally (except for ALL loads on VA-20, which are still deenergized).

The crew has transitioned to EOP-2536, Excess Steam Demand Event, and has carried out all applicable steps.

Where must the #2 ADV be operated from in order to stabilize RCS temperature?

m m , . . . . , . . . m , , . . . . , . . . . , , . . . , , . . . . , , . . . , , . . . . , . . . , 1 1 1 1 ,

A The ADV controller on C21.

B Local-Manual at the ADV.

[7 C The ADV controller on C05.

D The ADV controller on C10.

- - - -... - -.. -.. 1 Justification B - Correct; VA-20 powers the entire #2 ADV control circuit outside of the control room. With a loss of VA-20, the #2 ADV can NOT be o~eratedremotelv from ANY location. The valve can be o~eratedlocallv due to the location of the steam r u ~ t u r e(CTMT).

A - wrong; The loss of VR-21 requires the #2 ADV be operaied from C-21; however, the loss of VA-20 also de-energizes the C21 part of the ADV control circuit.

C - Wrong; The #1 ADV control power has been modified to allow operation in manual only, from the C05 controller upon a loss of VR-

11. However, the #2 ADV C05 controller is powered from VA-20 and would be de-energized.

D - Wrong; The C10 Fire Shutdown panel is designed for used when the control room must be evacuated due to an during an Appendix " R type fire. Although it is very protected due to its function, the loss of VA-20 will still prevent the operation of the #2 ADV from here.

References I AOP-2504D (Loss of VA-20). Pg. 3, Discussion of #2 ADV loss of control Comments and Question Modification History Disagree with NRC suggestion to modify Parent question and use ~tinstead of Modified version. Suggested modifications to Parent would cause all choices to be wrong. If suggestions incorporated to Modified version, it would become very LOD. 01/14/2009 NRC KIA System/E/A System A1 1 RCS Overcooling Number AK1.l RO 3.1 SRO 3.3 CFR Link I,CFR: 41.8 141.10 145.3)

Knowledge of the operational implications of components, capacity and function of emergency systems as they apply to the RCS Overcooling.

Page 68 of 268 Printed on 2/19/2009 at 16:50

RO and SRO Exam Questions (No "Parents" Or "Originals")

ROI I-SRO Ques #: 27 Question ID: 876163 @ RO a SRO Student Handout? [J Lower Order?

Rev. 1 @I Selected for Exam Origin: Mod [IPast NRC Exam?

The plant was tripped due to an Excess Steam Demand event inside the turbine building. During the performance of EOP 2525, the following additional conditions were noted:

Three (3) CEAs stuck fully withdrawn.

Reactor power is 5 X 10-2% and stable.

The normal charging flow path is NOT available due to a pipe rupture between CH-429, Charging Header Isolation Valve, and the containment wall.

Pressurizer pressure = 1750 psia and rising.

Pressurizer level = 26% and rising.

RCS Tavg = 490 O F and rising.

Which of the following subsequent procedures is required to be used for guidance to meet the highest priority safety function?

m m m m , . . . . . . . . . . . . . . . . . . . . . ~ . . . . . . . . . . . . . . . . . m . m m . . m m , m m . . .

@ A AOP-2558; Emergency Boration B EOP 2541, Appendix 3; Emergency Boration C AOP-2512; Loss Of All Charging D EOP-2540A; Functional Recovery of Reactivity Justification 1 D - Correct; EOP 2540A has guidance for use of the alternate charging flow path through the safety injection header. Upon completing EOP-2525, the crew must transition to an event driven EOP for subsequentguidance a mitigation strategy. The stated conditions require the use of the alternate charging header, which is addressed in EOP-2540A.

A -Wrong; AOP-2558 does provide guidance for boric acid injection using the alternate charging flow path through the SI header.

However, procedure usage guidelines require the completion of E'OP-2525 before an AOP is referenced for guidance in event mitigation.

6 - Wrong; EOP-2541, Standard Appendices, provides guidance for the boric acid injection in EOP space. However, the normal charging header must be available.

C - Wrong; Although "all charging" has been lost, AOP-2512 is NOT written to recover charging flow within an EOP event. The assumption with the AOP is that the loss of charging flow is the event of concern.

References I EOP-2541, App. 1, Diagnostic Flow Chart for No Reactivity Control.

EOP-2540A, Pg. 6, St. 2, Charging Pump alignment to alternate path INO Comments or Question Modification History at this time. 1 NRC K/A System/E/A System E09 Functional Recovery Number EA2.1 RO 3.2 SRO 4.4 CFR Link ICFR: 43.5 145.13)

Ability to determine and interpret facility conditions and selection of appropriate procedures during abnormal and emergency operations as they apply to the Functional Recovery.

Page 70 of 268 Printed on 211912009 at 1650

RO and SRO Exam Questions (No "Parentsf'Or "Originals")

RO/ I-SRO Ques #. 28 Question ID: 54426 rn Fa.0 U SRO LJ Student Handout? @ Lower Order?

Rev. 1 rn Selected for Exam Origin: Bank C]Past NRC Exam?

A plant heatup is in progress and the RO has been directed to start the first two RCPs.

Which of the following describes the INTERLOCKS that must be satisfied in order to start an RCP?

................................................................................ rn .

A Minimum lift pump oil pressure and minimum RBCCW flow.

B Minimum lift pump oil pressure and minimum seal bleedoff flow.

C Minimum Pressurizer pressure and minimum RBCCW flow.

D Minimum RBCCW pressure and minimum lift pump oil flow.

Justification I A - Correct; RCP lift pump must be running with oil pressure to tbe bearings and the RCP Cooling Water Flow Low alarm must be out (orooer RBCCW flow) to make-uo aux. contacts in the RCP brea~erand allow it to be closed from C-03 B - WRONG; ~bnormalseal bleedoff flow will cause alarms that require the RCP to be secured, but these alarms will NOT prevent it from being started.

B -WRONG; It is against procedures to start an RCP with RCS pressure below NPSH. However, there is NO interlock to prevent this from occurring.

D -WRONG; Starting the RCP under these conditions could result in serious damage, but they are NOT the specific parameters the starting circuit monitors.

References 1 LP RCS-00-C, R8C1, Pg 41 w o m m e n t s or Question Modification History at this time. 1 NRC KIA System/E/A System 003 Reactor Coolant Pump System (RCPS)

Number A4.03 RO 2.8 SRO 2.5 CFR Link (CFR: 41.7 145.5 to 45.8)

Ability to manually operate andlor monitor in the control room: RCP lube oil and lifl pump motor controls Page 73 of 268 Printed on 211912009 at 16:50

RO and SRO Exam Questions (No "Parents" Or "Originals")

Question ID: 78836 R0 SRO [I Student Handout? u Lower Order?

Rev. 0 Selected for Exam Origin: Bank Past NRC Exam?

The plant is at 100% power, normal operation, when the Letdown Backpressure Controller, PIC-201, transmitter fails to 200 psig.

Which one of the following is the system response to this instrument failure?

1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 . . 1 1 A Indicated letdown flow will remain constant.

Pressurizer level will remain constant.

B Indicated letdown flow will go to 0 gpm.

Pressurizer level will slowly rise.

C Indicated letdown flow will remain constant.

Pressurizer level will slowly rise.

mD Indicated letdown flow will go to 0 gpm.

Pressurizer level will remain constant.

1

~

Justification D is correct; The failure of the backpressure controller transmitter will cause the backpressure control valve to close causing the Letdown line relief valve, 2-CH-345, to lift. Because the relief valve is located upstream of the letdown flow indicator, indicated letdown flow will go to zero. The Letdown relief will still pass the flow allowed by the Letdown flow control valve and Charging flow is NOT affected; therefore, Pressurize level will NOT be affected.

A is incorrect. If the Letdown Backpressure Controller, PIC-201, transmitter fails to a value below the desired pressure of 300 psig, then the backpressure control valve will close and cause actual backpressure to exceed the lift setting of the Letdown Line Relief Valve; however, indicated pressure will read whatever pressure the transmitter has failed to. Because the Letdown flow indicator is downstream of the relief valve, letdown flow will indicate 0 gpm even though actual letdown flow through the relief valve is hasn't changed.

B is incorrect. If the Letdown Backpressure Controller, PIC-201, transmitter fails to a value below the desired pressure of 300 psig, then the backpressure control valve will close and cause actual backpressure to exceed the lift setting of the Letdown Line Relief Valve; however, indicated pressure will read whatever pressure the transmitter has failed to. Because letdown and Charging flows are still matched, Pressurizer level will NOT change.

C is incorrect. If the Letdown pressure transmitter fails to 200 psig, then letdown backpressure will read 200 psig; however, letdown flow will NOT go to 140 gpm regardless of how the backpressure transmitter fails. The Letdown flow control valve will maintain approximately 40 gpm.

References I OP-2304A, Pg. 4, Precaution 3 5 cvc-00-C, Pg Comments and Question Modification History Modified all four choices per NRC comments. 02/19/09 NRC IOA System/E/A System 004 Chemical and Volume Control System Number K4.11 RO 3.1 SRO 3.6 CFR Link (CFR: 41.7)

Knowledge of CVCS design feature(s) andlor interlock@)which provide for the following: Temperaturelpressure control in letdown line: prevent boiling, lifting reliefs, hydraulic shock, piping damage, and burst Page 75 of 268 Printed on 211912009 at 1 6 5 0

RO and SRO Exam Questions (No "Parents" Or "Originals")

Ro/ I-sRo Ques #: 30 1 Question ID: 8200015 a RO 0 SRO Student Handout? Lower Order?

Rev. 0 @ Selected for Exam Origin: Mod Past NRC Exam?

Initial Conditions:

- The plant has tripped due to a state wide blackout (the grid is lost).

- three (3) CEAs are stuck fully withdrawn.

- 24D is de-energized due to a bus fault.

- 24E is aligned to 24D.

- "B" Charging pump is aligned to Facility 2.

- The RO has successfully initiated emergency boration using the " A charging pump.

- ALL other plant equipment has responded as designed.

If the " A charging pump discharge relief valve were to stick full open at this time, which of the following describes an action needed to allow concentrated boric acid to be injected into the RCS and the reason for that action?

rn m m m a . m . m , m , n m . . . . . . . . . . . . . . . . . . . . . . . . . . . . .... m . . m u m m . .

A Align charging pump flow to Aux. Spray to reduce the charging pump's discharge back pressure.

B The "B" charging pump must be started on Facility 1; " A & "C" charging pumps are unavailable.

0C " A charging pump must be aligned to the alternate charging path; Facility 2 pumps are unavailable.

D HPSl must be used for boron injection; the discharge relief capacity exceeds three charging pumps.

Justification I B - Correct; The loss of 24D de-energizes the "C" charging pump and the "B" charging - - .pump . due to its initial power supply alignment Each charging pump has its own discharge relief valve;which when lifting is designed t relieve the entire capacity of the respective pump. As has been seen on MP2, when this valve fails open, 100% of the flow from the applicable charging pump is diverted to Clean Liquid Radioactive Waste. Procedural guidance states that if the "B" charging pump is not available solely because of its present power supply alignment, its must be shifted to power from the other facility and started.

A -Wrong; Reducing RCS pressure is the eventual requirement if there are NO charging pumps available. In that instance, Choice "D" may be correct.

C -Wrong; This would be the choice if the "B" charging pump were NOT available for reasons OTHER than loss of power to the facility it was presently aligned to.

D - Wrong; The LETDOWN line relief valve has the capacity to relieve all three charging pumps. However, each charging pump has its own discharge relief valve that only relieves that applicable charging pumps discharge.

References 1 EOP-2541, App 3, Pg 3, Emergency Boration wl "B" CCP.

Comments and Question Modification History Replaced Choice " A [Isolate RCP bleedoff flow, the discharge relief ensures pump minimum flow of approximately 4 gpm.] (too easy) 11111/08 NRC KIA System/E/A System 004 Chemical and Volume Control System Number A2.14 RO 3.8' SRO 3.9 CFR Link (CFR: 41.51 4315 14513 14515)

Ability to (a) predict the impacts of the following malfunctions or operations on the CVCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Emergency boration Page 78 of 268 Printed on 211 912009 at 16:50

RO and SRO Exam Questions (No "Parents" Or "Originals")

ROI I-sRO Ques #: 31 I Question ID: 8055096 4 RO SRO Student Handout? a Lower Order?

Rev. 2 Selected for Exam Origin: Mod Past NRC Exam?

The plant is in Mode 4, preparing to enter a refueling outage. Cooldown and depressurizing of the RCS is on hold for equipment testing before Mode 5 entry.

The following conditions now exist:

- RCS pressure is stable at 150 psia.

- RCS Tavg is stable at 225 O F

- Pressurizer (PZR) level is 40%, as seen on the Cold Cal. Indication L-103.

- Channel " X and "Y" PZR level indicate 46% on both controllers.

- ALL PZR Backup Heaters are in "PULL-TO-LOCK.

Which one of the following control actions is required to maintain PZR level and pressure stable?

................................................................ m . m . m m . m m m . m m " m . . .

A Level controller in AUTO with LocalIRemote switch in LOCAL and setpoint set to 46%.

Pressure controller in MANUAL with output adjusted as necessary.

B Level controller in MANUAL with output adjusted as necessary.

Proportional Heater Breakers opened and closed as necessary.

C Foxboro IA Level Setpoint MANUALLY set to 46%.

Proportional Heater Breakers opened and closed as necessary.

D Level controller in AUTO set to 40% with Setpoint Override switch (C-02) in OVERRIDE.

Pressure controller in MANUAL with output adjusted as necessary.

I

- - - - p p p p p - - - -

~ustificatio.

B is correct. With RCS temperature in the range for SDC operation, the Reactor Regulating System (RRS) would calculate a PZR setpoint of 40%. However, because the PZR level control channels are calibrated for NOTINOP, they would indicate a level of -46%.

46%-40%=6% mismatch in level The PZR Level Control System will see this mismatch as a "level insurge" and respond accordingly.

With level >/= 3.6% above setpoint, the response will cause all Proportional Heaters to come on at maximum output, regardless of the Pressure Controller's output, unless they are manually secured by opening their individual breakers.

A is incorrect. This would bypass the RRS level setpoint (of 40%) and allow automatic control of PZR level (although NOT normally done). However, because the "insurge" relay that causes the proportional heaters to be at maximum output does not receive any input from the level controllers, the pressure controllers are still unable to control Proportional Heaters.

C is incorrect The RRS is a subsystem of the Foxboro IA computer system. The Foxboro IA generates the PZR level setpoint signal that is used by the level control circuit, using RCS Tc and Th inputs. Although five other setpoints generated by the Foxboro IA can be manually overridden, the only manual control for PZR level setpoint generation is which RCS loop is used for temperature input.

D is incorrect. Placing the PZR Setpoint Override Switch in the "OVERRIDE" position will override the insurge signal that prevents the backup charging pumps from running if manually started. However, it does NOT override the insurge signal that drives the proportional heaters to maximum output, therefore the pressure controllers are still unable to control Proportional Heaters.

References I OP-2204, Attachment. 4, PZR Level Control Program Comments and Question Modification History Based on technical complexities of the control circuit, some of the NRC suggestions for quesiton rev~sionwould weaken LOD Mod~f~ed stem slightly and completely revised all four choices to incorporate some of NRC suggestions and improve LOD 01/16/2009

[See original question and modify based on comments there. 01/22/09]

NRC WA System/E/A System 005 Residual Heat Removal System (RHRS)

Number A4.03 RO 2.8' SRO 2.7* CFR Link (CFR: 41.7 145.5 to 45.8)

Ability to manually operate andlor monitor in the control room: RHR temperature, PZR heaters and flow, and nitrogen Page 8 0 of 268 Printed o n 2/19/2009 at 16:50

RO and SRO Exam Questions (No "Parents" Or "Originals")

ROI I-SRO Ques #: 32 I Question ID: 8000015 a RO SRO n Student Handout? @ Lower Order?

Rev. 0 Selected for Exam Origin: New U Past NRC Exam?

The plant tripped from 100% power. On the trip, tt-e supply breaker to Bus B-51 tripped and CANNOT be reset (B-51 and VR-11 remain de-energized).

While the crew was performing EOP 2525, Standard Post Trip Actions, a Large Break LOCA developed.

All other systems and components operated as expected. The crew subsequently entered the appropriate procedure.

To ensure maximum obtainable Safety lnjection flow is established, a PEO must be dispatched to open

...................................................................... m a . . m a . . . rn . rn A ONLY LPSl lnjection Valves, SI-615 and Sl-625 locally.

B ONLY LPSl lnjection Valves, 9-615 and 3-635 locally.

C all four Facility 1 HPSl lnjection Valves locally.

[7 D Facility 1 HPSl Header Stop, Sl-656, locally.

Justification I A is correct. Bus 8-51 supplies power to the Facility 1 LPSl Valves. SI-615 and Sl-625. Unlike the HPSl lnjection valves which are maintained open, the valves will automatically open on SlAS Due to the loss of power, the only way to open the valves is in the local, manual mode. All other LPSl components should function as designed to ensure maximum flow is obtained.

B is incorrect. LPSl lnjection Valve, Sl-635, is powered from Facility 2 and is already open. (Examinee may think all odd numbered valves are power from Facility 1, as is the normal convention.) Failing to open Sl-625 locally prevent maximum SI flow.

C is incorrect. The Facility 1 HPSl lnjection Valves they are dee~iergizedwith the loss of 851. Although they get a SlAS open signal, the Facility 1 HPSl lnjection Valves are maintained open all the time.

D is incorrect. Although the HPSl Header Stop is deenergized, il is maintained open all the time above MODE 5.

References I LP ECC-01-C (ECCS) , Pg. 12, LPSl lnjection Valve discussion Comments and Question Modification History Changed all distractors and correct answer per NRC comments 1115/09 NRC KIA System/E/A System 006 Emergency Core Cooling System (ECCS)

Number K2.04 RO 3.6 SRO 3.8 CFR Link (CFR: 41.7)

Knowledge of bus power supplies to the following: ESFAS-operated valves Page 82 of 268 Printed on 2/19/2009 at 16:50

RO and SRO Exam Questions (No "Parents" Or "Originals")

mi I-SRO Ques #: 33 Question ID: 8054464 RO n SRO Student Handout? Lower Order?

Rev. 0 Selected for Exam Origin: Mod 0Past NRC Exam?

The plant is at 100% power, steady state, with all equipment operating as designed.

Then, an RCS Safety Valve begins leaking by, causing a slow rise in Quench Tank parameters.

Which of the following is procedurally directed to control Quench Tank temperature?

....................... m m m m . . . . . . . . . . . . m . . . . . . . . . . . . . . . . . . . . . . . . . . . . . m m . m , . . m m m . * .

U A Place the Quench Tank on recirculation to maintain temperature less than the maximum of 150°F.

B Place the Quench Tank on recirculation to maintain temperature less than the maximum of 120°F.

C Feed the Quench Tank from PMW and drain to the Primary Drain Tank as necessary to maintain temperature less than the maximum of 150°F.

0D Feed the Quench Tank from PMW and drain to the Primary Drain Tank as necessary to maintain temperature less than the maximum of 120°F.

Justification I B - Correct; The Recirculation alignment is the same as the coolirig - alignment and OP 2301A requires tank temperature to maintained less than 120°F.

A -Wrong; 150°F is NOT the maximum allowed temperature.

C -Wrong; Although it may be effective, feed and bleed is NOT the procedurally required method of cooling the Quench Tank.

Additionally, 150°F is NOT the maximum allowed temperature.

D - Wrong; Although it may be effective, feed and bleed is NOT the procedurally required method of cooling the Quench Tank.

References I OP-2301A (PDT & Quench Tank Operation), Pg. 4, Cautionllnstruction of valve control when cooling.

Comments and Question Modification History 1 Fixed typo in stem. Changed "...slowly rise ..." to "...slow rise ..." 11111/08 Fixed typo in stem and choices C and D. NRC comment. 2/19/00 NRC WA System/E/A System 007 Pressurizer Relief TankIQuench Tank System (PRTS)

Number A1.03 RO 2.6 SRO 2.7 CFR Link (CFR: 41.5145.5)

Ability to predict andlor monitor changes in parameters (to prevent exceeding design limits) associated with operating the PRTS controls including: Monitoring quench tank temperature Page 84 of 268 Printed on 211912009 at 16:50

RO and SRO Exam Questions (No "Parents" Or "Originals")

RO/ I-SRO Ques #: 34 I Question ID: 79028 a RO 0 SRO Student Handout? Lower Order?

Rev. 3 Selected for Exam Origin: Bank Past NRC Exam?

With the plant operating at 100% power with Bus 24E aligned to Bus 24C, the following alarms are received within a 5 minute period of time:

- RBCCW HDR B PRESS LO (C-0617)

- RBCCW HDR B FLOW HI (C-0617)

- RBCCW SURGE TK LEVEL HIILO (C-0617)

- AUX BLDG SUMP LEVEL HI (C-0617)

- PMW HEADER LOW PRESSURE (C-0213)

- Various low flow annunciators for components supplied by "B" RBCCW header NO other annunciators are in alarm.

Which of the following conditions will cause these indications and what actions, per AOP 2564, Loss of RBCCW, will be required to mitigate the consequences of this event?

m m . . . . . , . . . . . . .

A The RBCCW supply piping has ruptured at the inlet to the "C" RBCCW Heat Exchanger.

Align the "B" RBCCW Pump and Heat Exchanger to supply Facility 2 and place them in service; Isolate "C"RBCCW Heat Exchanger.

The RBCCW piping that supplies the Letdown Heat Exchanger, Sample Coolers, and the Degasifier has ruptured.

Align the "B" RBCCW Pump and Heat Exchanger to supply Facility 2 and place them in service; Isolate "C" RBCCW Heat Exchanger.

C The RBCCW supply piping has ruptured at the inlet to the "C" RBCCW Heat Exchanger.

Place the "C" RBCCW Pump in Pull-To-Lock, secure RBCCW Surge Tank Make Up, 2-RB-215, trip the reactor, secure the "B" and "DMRCPs.

D The RBCCW piping that supplies the Letdown Heat Exchanger, Sample Coolers, and the Degasifier has ruptured.

Place the "C" RBCCW Pump in Pull-To-Lock, secure RBCCW Surge Tank Make Up, 2-RB-215, trip the reactor, secure the "B" and "D" RCPs.

,Justification (

D is correct The RBCCW low pressure alarm is indicative of a header rupture downstream of the RBCCW Pump. The RBCCW high flow would narrow down the rupture to downstream of the flow instrument which is downstream of the RBCCW heat exchanger. The various low flow annunciators for components supplied by the "B" RBCCW header would further narrow down the location to a component supplied by the "0" RBCCW header.

A is incorrect A rupture of the "B" RBCCW header at the inlet to the RBCCW heat exchanger would NOT result in a hlgh flow alarm.

Additionally, an RBCCW sump high level annunciator would also alarm.

B is incorrect. The RBCCW header flow instrument (and annunciator) are located downstream of the RBCCW heat exchanger discharge isolation valve, therefore, a low flow annunciator would alarm for this condition Additionally, an RBCCW sump high level annunciator would also alarm C is incorrect. A rupture in the suction piping to the RBCCW Pump would NOT result in an RBCCW header high flow alarm References I ARP for RBCCW Hi Flow alarm Comments and Question Modification History Discussed the need to add Letdown Hlgh Temperature alarm to the list of annunciataors. Determined that it was NOT appropriate.

Annunciator will NOT alarm for this condition. 11111/08 NRC WA System/E/A System 008 Component Cooling Water System (CCWS)

Number A2.07 RO 2.5* SRO 2.8' CFR Link (CFR: 41.5 143.5 145.3 145.13)

Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Consequences of high or low CCW flow rate and temperature; the flow rate at which the CCW standby pump will start Page 86 of 268 Printed on 211 912009 at 16:50

RO and SRO Exam Questions (No "Parents" Or "Originalsw)

Question ID: 8066598 4 RO SRO n Student Handout? Lower Order?

Rev. 0 k Selected for Exam Origin: Mod Past NRC E:xam?

The plant was operating at 100% power with all equipment in a normal alignment. Bus 24C is supplying Bus 24E. The "C" RBCCW Pump suddenly trips on fault. The Balance of Plant (BOP) operator starts the "B" RBCCW Pump to supply Facility 2 and places the 'C' RBCCW Pump handswitch in Pull-To-Lock. NO other operator action is taken.

Then, a circuit failure in the Facility 2 ESAS Actuation Cabinet causes an inadvertent actuation of SIAS, CIAS, EBFAS and UV ALL plant systems and components function as designed (including those driven by the Facility 2 inadvertent actuation).

Which of the following correctly describes the condition of the RBCCW Pumps and Heat Exchanger TCVs based on the inadvertent ESAS actuation?

~1 A The 'A' and 'B' RBCCW Pumps are running.

'A' heat exchanger TCV throttling on temperature.

'B' and 'C' heat exchanger TCVs full open.

B The 'A' and 'B' RBCCW Pumps are running.

'A' and 'B' heat exchanger TCVs throttling on temperature.

'C' heat exchanger TCVs full open.

C Only the 'A' RBCCW Pump is running

'A' heat exchanger TCV throttling on temperature.

'B' and 'C' heat exchanger TCVs full open.

D Only the 'A' RBCCW Pump is running

'A' and 'B' heat exchanger TCVs throttling on temperature.

'C' heat exchanger TCVs full open.

i~ustifiCation I C is correct. The '8' RBCCW Pump SIASILNP Block handswitch at the breaker was lefl in BLOCK which will NOT allow the 'B" RBCCW Pump to start on SIAS. A Fac. 2 SIAS will cause the 'B' (swing) and 'C' RBCCW heat exchanger TCVs to get a full open signal.

A is incorrect. The 'B' RBCCW Pump will NOT start on SIAS due to the SIASILNP Block handswitch being lefl in the BLOCK position.

B is incorrect. The 'B' RBCCW Pump will NOT start on SIAS due to the SIASILNP Block handswitch being lefl in the BLOCK position.

Additionally, the 'B' RBCW heat exchanger is the swing heat exchanger and its TCV will get an open signal on a SIAS from either facility of ESAS.

D is incorrect. While it is true that only the 'A' RBCCW Pump will start on the SIAS, the '6' RBCW heat exchanger is the swing heat exchanger and its TCV will get an open signal on a SIAS from either facility of ESAS.

References I AOP-2564, Pg. 15 + 16 ARP-2590-097, "B" RBCCW Pump SlASlLNP Manually Blocked Comments and Question Modification History Modified stem to state an LNP occurred on the trip from the LOCA Reworded question per NRC comments. 02/18/09 NRC K/A System/E/A System 008 Component Cooling Water System (CCWS)

Number A3.08 RO 3.6* SRO 3.7* CFR Link (CFR: 41.7 I45.5)

Ability to monitor automatic operation of the CCWS, including: Automatic actions associated with the CCWS that occur as a result of a safety injection signal Page 88 of 268 Printed on 2/19/2009 at 1650

RO and SRO Exam Questions (No "Parents" Or "Originals")

ROI I-SRO Ques #: 36 Question ID: 1000021 RO SRO Student Handout? Lower Order?

Rev. 2 Selected for Exam Origin: Bank Past NRC Exam?

The plant is at 100% power, steady state, with 480 V buses 22A and 22B cross-tied due to 22A's 41601480 VAC step-down transformer being~taggedout.

Chemistry has determined that the RCS and pressurizer boron concentration differ by more than 50 ppm.

The crew must commence forcing pressurizer sprays for boron equalization at this time.

How must the pressurizer heaters be aligned to force sprays?

A Groups '3' and '4' backup heaters ON Groups 'I'and '2' backup heaters in 'Pull-To-Lock'.

aB Groups 'I'and '2' backup heaters ON Groups '3'and '4' backup heaters in 'Pull-To-Lock' C Groups 'It, '3' and '4' backup heaters ON Group '2' backup heaters in 'Pull-To-Lock'.

D Groups '2', '3' and '4' backup heaters ON Group 'I'backup heaters in 'Pull-To-Lock' Justification 1 A - CORRECT; OP 2204, Section 4.2 , has guidance for energizing heaters as required and adjusting the in service pressure controller to maintain pressure. cross-tying480 V buses requiresthat both associated backupheater groups be tagged out due to the potential to overload 228 supply transformer; therefore, only group 3 and Group 4 heaters are available for forcing sprays.

B - WRONG; Group 1 and Group 2 heaters are NOT available. With Bus 228 supplying Bus 22A, the heaters are placed in Pull-To-Lock and tagged to prevent overloading Bus 228 transformer.

C -WRONG; Placing Group 2 heaters in Pull-To-Lock will help to prevent overloading Bus 228 by keeping loads on 22B at a minimum, but the procedure for cross tying 480 Volt buses does NOT allow energizing backup heaters on Bus 22A. The limit is imposed too protect Bus 228 transformer, not the crosstie breaker.

D - WRONG; This assumes that the limiting factor is the cross-tie breaker (22A - 228) and that only the Group 1 heaters must be placed in Pull-To-Lock. Therefore, only the far side heater group 1 (Bus 22A) may NOT be used, but Group 2 (Bus 228) may be used normally.

References I OP-2344A, Step 4.13, Cross-tie 480 load center 22A with 228 Comments and Question Modification History Deselected KIA due low discriminatory value of "from memory" knowledge of power supply to all control board indications.

Question from bank modified for new KIA 01/16/2009

[Modify Rev 1 based on Cliff C. 01/22/09]

NRC comments:

I ) and 2) Fixed justification for all choices. 2/19/09

3) selected 'Lower order. 2118/09

4) and 6) Reworded stem 2/19/09

5) Fixed choice C? 2119/09 NRC K/A System/E/A System 010 Pressurizer Pressure Control System (PZR PCS)

Number K2.01 RO 3.0 SRO 3.4 CFR Link (CFR: 41.7)

Knowledge of bus power supplies to the following: PZR heaters Page !33 of 268 Printed o n 211 912009 at 1 6 5 0

RO and SRO Exam Questions (No "Parents" Or "Originals")

RO/ I-SRO Ques #: 37 I Question ID: 1000022 RO rJ SRO 1Student Handout? Lower Order?

Rev. 0 Selected for Exam Origin: Bank C7 Past NRC Exam?

I&C is performing a function test on RPS channel 'A' and has bypassed the power related trip on the channel. Vital instrument bus VA-30 is suddenly lost.

Based on the above, what is the resulting condition of the RPS?

. . . . . . . . . . . . . . . . . . . . . . . . . I . . . , , . . . , I . . . . . , . . . . . . . . . . , . . . . . . . . . . . , . . . . . , . . " . . . .

A The loss of Channel 'C' causes four TCBs to open, but the reactor has NOT tripped.

B With Channel 'A' bypassed and Channel 'C' de-energized, ALL TCBs remain closed.

aC Coincident trip signals are processed from Channels 'A' and 'C' resulting in a reactor trip.

D With the RPS in a 1 out of 3 configuration, the loss of Channel 'C' results in a reactor trip.

Justification 1 A is correct. The K3 relay (powered from VA-30) is deenergized resulting In two TCBs opening (TCB 3 and 7). Additionally, due to the loss of another matrix relay (powered from VA-30) a contact opeps in thicircuitry for the K4 r i a y causing TCBS 4 and 8 to open. The end result is that half of the TCBs are open, but the remaining TC,Bs are still providing power to the CEDMs; therefore, the reactor did NOT trip.

B is incorrect. Channel 'A' is bypassed and Channel 'C' is de-energized; however, the loss of Channel 'C' will result in 4 TCBs opening C is incorrect. Selected trip units on Channel 'A' are in bypass and will NOT process a trip signal for those units. The reactor will NOT trip from the trip signal processed from the loss of Channel 'C' because a signal from 2 channels is needed to cause a reactor trip.

D is incorrect. Placing Channel 'A' in bypass will NOT place the RPS in a 1 out of 3 configuration; therefore, the reactor will NOT trip when Channel 'C' is lost.

References I LP RPS-01-C, Fig. 8, RPS Drawer Power Supplies Comments and Question Modification History Reworded the stem such that I&C is perbrming a functional test on RPS Channel " A instead of the daily RPS surveillance. The daily surveillance no longer requieres the operator to bypass any RPS trips when performing the surveillance;~therefore,Distractor '0' was-implausible. 11111/08 No NRC comments 01/16/2009 NRC KIA System/E/A System 012 Reactor Protection System Number K1.O1 RO 3.4 SRO 3.7 CFR Link (CFR: 41.2 to 41.9 145.7 to 45.8)

Knowledge of the physical connections andlor cause effect relationships between the RPS and the following systems: 120V vital/instrument power system Page 95 of 268 Printed on 211912009 at 16:50

ROI I-SRO Ques #: 38 Question ID: 8053886 a RO SRO n Student Handout? ,@ Lower Order?

Rev. 0 @ Selected for Exam Origin: Mod u Past NRC Exam?

Given the following conditions:

- 100% reactor power

- lnverter 2 has been deenergized in preparation for emergent repairs The DC input breaker on lnverter 6 is inadvertently opened while hanging the clearance on lnverter 2 .

If a large break LOCA were to occur inside Containment with the plant in this configuration, which of the

....................... ................~.....................................~....

following would be an expected condition two minutes after the event? Assume NO operator action.

A 'B' LPSl Pump has automatically started.

B 'A' LPSl Pump has automatically started.

C 'C' CAR Cooler Fan is running in fast speed.

D 'D' CAR Cooler Fan is running in slow speed.

Justification I B is correct. Facility 1 ESAS equipment will be unaffected by the loss of Power to VA-20; therefore, 'A' LPSl Pump will automatically start A is incorrect Opening the DC input breaker on lnverter 6 with lnverter 2 out will deenergize Vital AC Bus VA20, which will deenergize Facility 2 ESAS Actuation Cabinet. All Facility 2 ESAS associated equipment will be prevented from responding to conditions whlch would normally result in an actuation. 'B' LPSl will remain stopped until manually started by the operator C is incorrect The 'C"' CAR Fan is a Facility 1 Component Facility 1 ESAS equipment will operate as designed. 'C' CAR Fan will shift to slow speed on the SIAS.

D is incorrect. Facility 2 ESAS equipment will not receive an actbation s~gnalof any kind. 'D' CAR Fan will remain in fast speed.

References I 120 VAC One-Line Diagram ESA-OIC, ESAS Lesson Text Comments and Question Modification History Changed annotation to "Low" (from Memory) 01/16/2009

- - - - p p p p p p NRC KIA System/E/A System 013 Engineered Safety Features Actuation System (ESFAS)

Number A3.02 RO 4.1 SRO 4.2 CFR Link (CFR: 41.7 145.5)

Ability to monitor automatic operation of the ESFAS including: Operation of actuated equipment Page 97 of 268 Printed on 211 912009 at 16:50

RO and SRO Exam Questions (No "Parents" Or "Originals")

RO/ 1 - sQues~ #. 39 Question ID: 8000068 m RO SRO a Student Handout? Lower Order?

Rev. 0 m Selected for Exam Origin: New Past NRC Exam?

The plant has experienced a Loss of Coolant Accident and the following conditions exist:

- Sump Recirculation Actuation has occurred.

- Both LPSl Pumps are running and their respective breakers cannot be opened from either the Control Room or locally.

- All other SRAS actuated components have been properly actuated.

Which one of the following statements describe the appropriate action to take?

A Override and close three LPSl lnjection Valves to include LPSl lnjection Valve, 2-Sl-635.

B Override and open SI Pump Minimum Flow Recirculation Valves, 2-Sl-659 and 2-3-660.

c If any LPSl Pump is exhibiting signs of cavitation, then override and stop one HPSl Pump.

D Place the LPSl system in a Boron Precipitation Control configuration.

Justification 1 A - Correct; Per EOP 2532, contingency step 48 b 1, if the LPSl Pumps cannot be secured, then close Sl-635 and two other LPSl lnjection Valves One LPSl lnjection valves must remain open B - Wrong, Although reopening the minimum flow isolation valves will provlde the LPSl Pumps with minimum flow and prevent pump overheating, this would also result in a release to the atmosphere through the RWST vent C - Wrong, Stopping a HPSl Pump IS the appropriate action for I;! Pump cavitation due to a sump clogging condition, not for the failure of the LPSl Pump breakers to open on SRAS.

D -Wrong, Although a LPSl Pump may be used for Boron Precipitation Control and this would provide a flow path to prevent over heating, ~twould be much too early at thrs point. Additionally, only one LPSl Pumps may be used for Boron Precipitation.

EOP 2532, Contingency Step 48.b.l Comments and Question Modification History Changed question to address NRC comments. Boarderline low LOD. WA not tight. Choices B, C, and D not plausible New question requires operator to monitor LPSl pumps due to failure to turn off. AISO-requiresmanual operation in the Control Room due to a failure of ESFAS-initiated equipment to actuate. WA match Distractors are all plausible because each of them will provide some level of LPSl Pump protection, but they are either inappropriate at this time or they are for the wrong reason 2/2/09 NRC comments.

I)and 2) Fixed Typos. 2119/09

3) Reworded justifiction for Choice D to explain why it is plausible 2/19/09

4) 'Immediately' deleted from chioice D 2/19/09

5) Changed SI pump to LPSl pump 2/19/09

6) Changed to lower order 2/19/09 NRC KIA System/E/A System 013 Engineered Safety Features Actuation System (ESFAS)

Number A4.01 RO 4.5 SRO 4.8 CFR Link (CFR: 41.7 145.5 to 45.8)

Ability to manually operate andlor monitor in the control room: ESFAS-initiated equipment which fails to actuate Page 100 of 268 Printed on 211912009 at 16:50

RO and SRO Exam Questions (No "Parents" Or "Originals")

Question ID: 8000046 k]RO 1- SRO U Student Handout? [ 1 Lower Order?

Rev. 0 [31 Selected for Exam Origin: New C]Past NRC Exam?

Unit 2 was operating at 100% power, with "B" RBCCW Heat Exchanger out of service for hydrolazing, when the Instrument Air line that supplies the RBCCW temperature control valves (TCVs) ruptured.

Instrument Air Supply Stop to RBCCW TCVs, 2-lA-255, was closed to isolate the ruptured air line. " A and "C" RBCCW Heat Exchanger Service Water Outlet Valves, 2-SW-9A and 2-SW-9C, were manually throttled to normal flow rates to maintain normal te~nperatc~res on RBCCW cooled components.

Then, the plant trips due to a Large Break LOCA in Containment.

ALL plant equipment functions as designed.

Which one of the following statements describes how the present Service Water (SW) System alignment will affect this event?

1 1 . . . . . . . . . . . . . . . . . ~ ~ . ~ ~ ~ ~ ~ . . . ~ ~ . ~ ~ ~ ~ ~ ~ ~ . . . . . . ~ . ~ ~ ~ . ~ . . . . . . . . . ~ . ~ . ~ . . ~ ~ ~ ~ .

14A Upon SRAS initiation, the RBCCW System heat exchangers will NOT provide adequate cooling for CTMT and will cause CTMT temperature and pressure to rise.

B The initial containment pressure spike will approach design pressure due to the limited cooling from the CAR fans, but long term heat removal will t)e adequate.

nC Flow through the SDC heat exchangers shell side will NOT be adequate by the time SRAS occurs and core cooling will be challenged.

LA D The higher RBCCW system temperatures frorrr the lower Service Water flow will cause the Spent Fuel Pool temperature to rise.

A - CORRECT: The given conditions will result in significantly diminished post-accident SW flow to RBCCW which will result in a significantly higher RBCCW temperature due to the higher heat loading from the CAR Coolers (LOCA in Containment). RBCCW flow to the CAR Coolers remains the same, but significantly more heai is added to the RBCCW System as a result of the LOCA. The restricted Service Water flow to the RBCCW Heat Exchangers will result in significantly higher RBCCW temperatures to all supplied components. This is especially true after SRAS with the SDC Heat Exchangers now in service.

B - WRONG: CTMT Spray will prevent CTMT pressure from exceeding the design pressure. Additionally, adequate long term cooling will not be available due to the loss of cooling to RBCCW needed for the post-SRAS environment.

C - Wrong; When SRAS occurs, RBCCW flow through the SDC Heat Exchangers will NOT change. The increase in RBCCW temperature due to the low SW flow will result in significantly less heat removal from Containment. However, the increase in RBCCW temperature will NOT cause a change in RBCCW flow due to the lack of temperature control valves in components presently being cooled by the system.

D - WRONG: This would occur if RBCCW were left aligned to the SFP cooling system under the stated conditions. However, SlAS actuation isolates RBCCW to the SFP cooling heat exchangers. SFP temperatures will rise, but only due to the loss of cooling (NOT due to higher temp. in RBCCW) which is restored 4 to 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months after the LOCA.

Millstone Unit 2 Technical Soecifications Bases. Paae 8314 6-3 RCB-00-C, Pages 9 and 42 SWS-00-C, Page 33 In the stem, replaced 'three' with 'inservice'; Was: Service Water flow to the three RBCCW heat exchangers... Now: Service Water flow to the in-service RBCCW heat exchangers ... The 'spare' heal exchanger TCV will either be full open if on minimum flow or isolated if not on minimum flow. 11/11/08 Changed distractor B from 'CAR Fans unavailable . . ' to 'limited cooling from the CAR Fans.. ' Added which valves are throtteld in stem. The small TCVs are open due to the loss of air; however, the flow through them is limited to approximately 2200 gpm as opposed to 9,000 gpm through the large TCVs which are throttled to a normal flow of approximatley 1500 gprn (normal flow depends on time of year; 300-1500 gpm). Added IA valve name and number to stem. 2/2/09 NRC comments:

1) Corrected typo in stem. 2/19/09

2) Deleted referece to the state of the Service Water System after the LOCA, in the stem. 2/19/09

3) Replaced Choice D. 02/19/09 Modified choice "D" to eliminate the SFP temperature "would exceed the design limit." per NRC comments. 02/20/09 NRC K/A System/E/A System 022 Containment Cooling System (CCS)

Number K1.O1 RO 3.5 SRO 3.7 CFR Link (CFR: 41 2 to 41.9 145.7 to 45 8)

Page 102 of 2 6 8 Printed o n 212012009 at 14:40

RO and SRO Exam Questions (No "Parents" Or "Originals")

Question ID: 8000016 3 RO SRO a Student Handout? Lower Order?

Rev. 0 @ Selected for Exam Origin: New Past NRC Exam?

The plant was at 100% power when it automatic all!^ tripped due to a Small Break Loss of Coolant Accident approximately 4.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months ago.

- Immediately prior to the trip, bus 22F was deenergized due to a fault.

- Containment pressure peaked at -12 psig, but is now reading 6.8 psig and slowly dropping.

- Containment temperature is slowly lowering with pressure.

- The Containment High Range Radiation Monitors are NOT in alarm.

The RO has just overridden and secured all actuated Containment Spray components at C-01.

The RO notes the following annunciators are still in alarm:

- All four channels of CTMT PRES HI HI

- Both facilities of CSAS ACTUATION SIG TRIP Subsequently, the break size increases, causing Containment pressure to rise rapidly above 10 psig.

Which of the following describes the subsequent condition andlor actions associated with the Containment Spray system, based on the rise in containment pressure?

A Both facilities of Containment Spray will automatically actuate and supply >I300 gpm flow to each Containment Spray header.

B Facility 1 Containment Spray Pump must be started and the associated valve opened by pushing the actuation button on C-01.

p~C Facility 1 Containment Spray Pump must be manually started and the associated valve opened from the C-01 vertical section.

D Both facilities of Containment Spray must be manually started and the associated valves opened from the C-01 vertical section.

Justification I C - Correct; EOP-2532, LOCA, dictates that CTMT spray should be secured once CTMT pressure drops below 7 psig. The system is secured by overriding the pump off, overriding the spray valve closed. Then, the actuating signal on ESAS is "reset" so auto actuat~on is again available, if needed However, based on the steps accomplished, ESAS has NOT yet been reset. Therefore, spray flow must be manually restarted.

A -Wrong; Although the RO has secured CS flow, the actions taken so far have NOT completed the applicable procedure steps in that ESAS was NOT reset. The CS actuation signal on CTMT pressure, unlike the MSI actuation signal on SG pressure, does NOT automatically "re-arm" when pressure drops below setpoint and then rises back up. Therefore, the ESAS CS signal is still active, and can NOT "automatically" restart CS on rising pressure.

B - Wrong; Pushing the CS actuation buttons on C01 is the "expected" method to "re-activate" the CS signal. However, because the ESAS CS signal is triggered only by high CTMT pressure with a coincident SlAS (unlike SlAS which is triggered by CTMT pressure or RCS pressure), once a component is overridden the C-01 buttons will NOT re-trigger actuation.

D -Wrong; Starting the "B" Containment Spray Pump is NOT appropriate because 2-CS-4.1B, " B Containment Spray Valve, will NOT open due to the loss of power (22F). Even though the Spray valve was NOT overridden, the loss of Bus 22F will prevent it from opening.

~eferekes I EOP 2532 Tech Guide, Page 109 CSS-00-C, Containment Spray. Page 14 Comments and Question Modification History Reworded stem and choices baed on NRC comments. 02/02/09 Modified stem to remove statement that ESAS was not reset and added alarm indication noted by RO, per NRC comments. 02/18/09 NRC comments:

1) Deleted statement. "The fault cannot be repaired." 2/19/09

2) Removed, 'Five minutes later ...' and replaced with 'Subsequently...'. 2/19/09

3) Provided indication that CSAS had NOT been reset vs. just stating CSAS had NOT been reset. 2/19/09 NRC KIA System/E/A System 022 Containment Cooling System (CCS)

Number A1.02 RO 3.6 SRO 3.8 CFR Link (CFR: 41.5 145.5)

Ability to predict andlor monitor changes in parameters (to prevent exceeding design limits) associated with operating the CCS controls including: Containment pressure Page 108 of 268 Printed on 211912009 at 16:50

RO and SRO Exam Questions (No "Parents" Or "Originals")

RO/ I-SRO Ques #: 42 Question ID: 8000069 [31 KO I ]

SRO U Student Handout? Lower Order?

Rev. 0 @ Selected for Exam Origin: New D Past NRC Exam?

The plant has tripped from 100% power due to an intersystem LOCA into the RBCCW System. The RBCCW Containment Isolation Valves were succc!ssfully closed within 10 minutes after entry into Loss of Coolant Accident, EOP 2532.

The following conditions existed approximately 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months after the trip:

SRAS actuated.

Containment pressure is 5 psig and slowly lowering.

RCS pressure is 360 psia and slowly lowering.

RCS subcooling is 0°F.

Vessel level is at 100%.

Containment Sump level indicates -18 feet.

Both Containment Spray Pumps are running normally.

" A and "C" HPSl Pump current and flow are fluctuating.

Per EOP 2532, Loss of Coolant Accident, which of the following actions must be taken?

m , , m m m I I . m m I , . . . . , . . . . , . . . . , . . . , , . . . , , . m m m , m m m . , , m m m , .

A Secure ONLY one Containment Spray Pump.

nB Secure one of the HPSl Pumps.

C Throttle HPSl injection flow.

14D Secure both Containment Spray Pumps.

Justification I D is correct. Sump suction strainer clogging will cause a lower suction pressure in all the running SI pumps. The lower suction pressure will cause the HPSl Pumps to cavitate. EOP-2532 directs the CS pumps be secured (ifnot needed) to limit the competition for sump suction flow.

A is incorrect. Securing only one Containment Spray Pump will provide some relief for the HPSl Pumps; however, the Containment Sump level is NOT abnormally low and is NOT the reason for the HPSl Pumps to cavitate.

B is incorrect. Securing a HPSl Pump would significantly reduce the flow to the vessel. With the given conditions, the Containment Spray Pumps are NOT required to be running; therefore, securing a HPSl Pump would NOT be appropriate.

C is incorrect. Procedurally, HPSl flow is lowered as needed AFTER the Containment Spray pumps are secured. Additionally, the hot water in Containment is NOT flashing in the HPSl Pumps suctions.

References 1 EOP-2532, St. 50, Indications of CTMT Sump Clogging Comments and Question Modification History Selected new KIA. Unable to develop a question based on old KIA. 2/3/09 Deleted second sentence of each choice, based on NRC comments. 02/18/09 NRC comments:

1) Removed the word 'initial' from stem and added 'EOP 2532'. 2/19/09

2) Second part of all choices (reason) removed. 2/19/09

3) Added 'ONLY' to Choice A. 2/19/09 NRC KIA System/E/A System 026 Containment Spray System (CSS)

I Generic KIA Selected 1 NRC KIA Generic System 2.1 Conduct of Operations Number 2.1.23 RO 4.3 SRO 4.4 CFR Link (CFR: 45.2 145.6)

Ability to perform specific system and integrated plant procedures during all modes of plant operation.

Page 112 of 268 Printed on 211912009 at 16:50

RO and SRO Exam Questions (No "Parents" Or "Originals")

Question ID: 8078895 a Fro A:[ SRO U Student Handout? Lower Order?

Rev. 2 Selected for Exam Origin: Mod nPast NRC Exam?

The plant is at 100% power when the extraction steam supply valve to the 2A Feed Water Heater closes.

Which one of the following describes the plant contlition approximately 5 minutes after the valve closes., as compared to steady state conditions prior to the va ve clos~ng?Assume NO operator action.

... 11...................*..............1......111.................................

A RCS Tcold on RPS is lower.

B Delta-T power on RPS is lower.

aC Generator MWe output is lower.

3D 1 Main Condensate Flow is lower A - Correct; The loss of steam supply to the feedwater heater will result in lower feedwater temperatures entering the SG. This will result in a drop in RCS Tcold.

B - Wrong; The lowering of Tcold combined with an unchanged seam demand will result in a RlSE in delta-T power.

C - Wrong; The loss of steam flow to the extraction results in a gain of steam flow directly to the main turbine. This extra steam will cause generator electrical output to RISE.

D - Wrong; The extra steam going through the main turbine equaies to extra water in the main condenser and a RlSE in condensate flow.

LP FHD-01-C. Pg. 22, Potential impact on power when removing a Feedwater Heater.

OP-2318, Precaution and Notes on potential impact on power when removing isolating steam to a Feedwater Heater Comments and Question Modification History Based on NRC comments, reworded question to include an approximate time and changed "will" to "is" in each choice 02/02/09 Minor wording changes to stem question statement, per NRC conirnents. 02/19/09 Minor typo, changed "Assuming" to "Assume" in stem per NRC comments. 02120109 NRC KIA System/E/A System 039 Main and Rehaat Steam System (MRSS)

Number K3.05 RO 3.6 SRO 3.7 CFR Link (CFR. 41.7 145.6)

Knowledge of the effect that a loss or malfunction of the MRSS will have on the following: RCS Page 114 of 268 Printed on 212012009 at 14:40

RO and SRO Exam Questions (No "Parents" Or "Originals")

ROl 1-SRO Ques #: 44 Question ID: 8000050 RO SRO n Student Handout? @ Lower Order?

Rev. 1 Selected for Exam Origin: Mod Past NRC Exam?

You have been directed to establish "local-manual" control of the #1 Main Feedwater Regulating Valve (MFRV).

Communications have been established with the Control Room and the valve restraining device has been removed.

Which of the following additional actions must be performed to take "local-manual"control of the # I MFRV?

......................... m , , . . , , . . , , , . , , , . , , , . . . . . . . . . . . , , . . , , . . . , , . . . , . . . , , . m , , . .

A 1. Isolate all control signals to the # I MFRV at the Bottleup Panel.

2. Rotate the manual handwheel until it contacts the valve operator.

3. Open the # I MFRV equalizing stop to put the manual handwheel in local control.

C] B 1. Rotate the manual handwheel on the # I MFRV until it makes contact with the valve operator.

2. lsolate air to the # I MFRV air operator.

3. Vent the remaining air out of the valve operator to put the manual handwheel in local control.

C 1. Rotate the large handwheel on the #I MFRV until it makes contact with the valve

2. Rotate the small handwheel to lock the large handwheel and put it in control of the valve.

3. Open the # I MFRV equalizing stop.

nD 1. Rotate the large handwheel on the # I MFRV until it makes contact with the valve

2. Isolate air to the #I MFRV air operator.

3. Rotate the small handwheel to equalize air across the operator to put the large handwheel in control of the valve.

Justification I C - CORRECT; Per OP 2385, the sequence is as follows:

1. TURN larae handwheel clockwise to its limit of travel.

2. TURN small handwheel counterclockwise until tight, locking large handwheel in "engaged" position.

3. OPEN feed regulating equalizing stop.

4. OPERATE the valve per Control Room instructions.

A -WRONG; That is a combination of the methods to take local control of an Atmospheric Dump Valve (ADV) and the MFRV.

B -WRONG; That is a combination of the methods to take local control of an ADV, an Aux. Feed Reg. Valve (AFRV) and the MFRV.

D -WRONG; That is a combination of the methods to take local control of an AFRV and the MFRV.

References I 0P-2385, Pg. 21 - 22, Step 4 7.4, Local Operation of No. 1 FRV Comments and Question Modification History Reselected a bank question and modified it based on NRC comments. 02/03/09 Minor corrections to choice "Dm,per NRC comments. 02/19/09

- Corrected typo; #2 MFRV changed to #1 MFRV

- Changed item 2. from "lsolate all backup air ..."to "lsolate air . . "

NRC WA System/E/A System 059 Main Feedwater (MFW) System 1 Generic KIA Selected 1 NRC KIA Generic System 2.1 Conduct of Operations Number 2.1.30 RO 4.4 SRO 4.0 CFR Link (CFR: 41.7 145.7)

Ability to locate and operate components, including local controls.

Page 117 of 268 Printed on 211912009 at 16:50

RO and SRO Exam Questions (No "Parents" Or "Originals")

Question ID: 8000070 a RO n SRO C]Student Handout? Lower Order?

Rev. 0 Selected for Exam Origin: New Past NRC E ~ a m ?

The plant has just come out of a refueling outage a i d is presently operating at 1.5 % power, making preparations to warm the Main Steam lines.

Chemistry reports that although the nitrogen blanket on the Condensate Storage Tank (CST) is in place, the oxygen concentration is still out of specification, high. CST level is being maintained constant at 97% with 80 gpm blowdown per Steam Generator.

What action must be taken for this condition?

..... 1..........................~.........,...,,...,,....,.,..,,,,......,.,.

u A Maximize CST make-up flow and drain to the hotwell.

a B Transfer CST to the hotwell and hotwell to the Surge Tank.

C Place the CST nitrogen sparger in service.

1.3 D Place the CST recirc pumps in service.

Justification 1 C - Correct; Per OP 23198, a Nltrogen blanket is maintained on the CST to limit the transfer of Oxygen to the Steam Generators via the Auxiliary Feed System. If required, the CST sparger bubbles Nitrogen into the CST to lower the Oxygen concentration.

A - Wrong; A feed and bleed of the CST would lower Oxygen levels; however, this would create a large volume of waste water that would cause an environmental concern when discharging.

B - Wrong; Transferring the CST to the Hotwell and then to the Surge Tank will help to eliminate the Oxygen in the CST, especially if the CST is refilled from the water purification vendor; however, this creates an environmental problem with the waste water that would be created.

D - Wrong; The CST recirc pumps will help de-aerate the water by raising ii temperature, but this is not designed to lower the oxygen level.

~ e f e r e d OP-2319B, D~scuss~on of CST nltrogen blanket and sparger Comments and Question Modification History Per NRC comments, reworded questlon to eliminate unnecessary complexity beyond the KIA 02119109

- Reworded stem questlon statement to ellmlnate 'why' actlon must be taken

- Reduced the word~ngof each cho~ceby removlng the 'reason' 'or each actlon Modified stem to include notation of existing nitrogen blanket, per NRC comments. 02/20/09 NRC KIA System/E/A System 061 Auxiliary IEmergency Feedwater (AFW) System Number K1.05 RO 2.6' SRO 2.8' CFR Link (CFR: 41.2 to 41.9 145.7 to 45.8)

Knowledge of the physical connections andlor cause-effect relationships between the AFW and the following systems: Condensate system Page 120 of 268 Printed on 212012009 at 14:40

Question ID: 54554 KO L SRO Student Handout? Lower Order?

Rev.

p 1

p p

&; Selected for Exam Origin:

Bank p p nPast NRC Exam?

p A Loss of Normal Power has occurred, and Bus 24C is being powered from the " A Emergency Diesel Generator.

The Reserve Station Services Transformer (RSST) is now available for use.

When should the undervoltage relays on the Engineered Safeguards Actuation System be reset, to allow transferring Bus 24C to the RSST?

................................ m . , , , . , . , . , , , . . , , , . . , , . . . . . . . , . . . , , . m m , , m m , , . m m , . .

A lmmediately AFTER stripping all loads (except 22E) off bus 24C.

L] B Immediately PRIOR to the RSST being energized from the grid.

0C Immediately AFTER paralleling the RSST with the Diesel Generator.

D immediately PRIOR to paralleling the RSST with the Diesel Generator.

Justification 1 D - Correct; With 24C being powered from the EDG due to an LNP, the ESAS Undervoltage signal would still be present This signal prevents closing ~nany other source of power to the bus, other than the EDG. Therefore, the signal must be reset before the RSST breaker can be closed A -Wrong; This action is taken only if 24C IS deenergized and is about to be repowered.

B - Wrong, This is NOT allowed as it would prevent the sequencer from slowly loading emergency equipment on the EDG if a subsequent accident resulted in a SlAS C - Wrong; This is NOT possible due to the UV interlock from ESAS with the EDG powering the buss.

References I EOP-2541, App 23, Attachment 23-H, Transferring 24C form ED(; to RSST Comments or Question Modification History at this time. 1 NRC K/A System/E/A System 062 A.C. Electrical Distribution Number K1.02 RO 4.1 SRO 4.4 CFR Link (CFR: 41.2 to 41.9)

Knowledge of the physical connections andlor cause- effect relationships between the ac distribution sys- tem and the following systems: EDIG Page 122 of 268 Printed on 211912009 at 1650

RO and SRO Exam Questions (No "Parents" Or "Originals")

Question ID: 73066 2 RO n SRO n Student Handout? [I Lower Order?

Rev. 1 Selected for Exam Origin: Bank Past NRC Exam?

The following conditions exist:

- Mode 2

- Reactor Startup in progress

- All switchyard breakers are CLOSED

- The Main Generator links are installed Then, a feed control problem causes a reactor trip on low steam generator level.

Based on EOP 2525, Standard Post Trip Actions, which one of the following describes the required actions to ensure the switchyard and transformer yard are properly configured?

.................................. m . , . . " , , . . . . . . . . , . . . , , . . . , . . . . , . . . m . . m m m m m . " . . m .

A Verify the Motor Operated Disconnect (15G-2x1-4) OPEN and the Generator Output Breakers (1%-

8T-2 & 15G-9T-2) CLOSED.

Verify all buses energized by the RSST.

B Verify the Motor Operated Disconnect (15G-2X 1-4) CLOSED and the Generator Output Breakers (1 5G-8T-2 & 15G-9T-2) OPEN.

Verify all buses energized by the RSST.

nC Verify the Motor Operated Disconnect (l5G-2x1-4) CLOSED and the Generator Output Breakers (15G-8T-2 & 15G-9T-2) CLOSED.

Verify all buses energized by the IVSST.

D Verify the Motor Operated Disconnect (15G-2x1-4) OPEN and the Generator Output Breakers (15G-8T-2 & 15G-9T-2) CLOSED.

Verify all buses energized by the NSST.

~usticati0.l A - Correct; With the ring bus closed, the 8T & 9T are closed and should remain that way. The 15G-2x1-4 must, therefore, be open because the Main Generator is off line B - Wrong; The 8T & 9T are closed in this conf~guration(Mode 2) and do NOT get a trip signal because the Main Generator never got a trip signal C -Wrong; Unit 2 does NOT have a Main Generator output breaker, therefore, the 15G-2x1-4 must be open if the ring bus is closed.

D - Wrong; With the Main Generator links installed, the unit can NOT be backfeeding through the NSST.

References I Training Diagram for 345 KV Comments and Question Modification History Added "Based on EOP 2525, Standard Post Trip Actions" to provide a basis for the required actions.

Discussed adding a failure of the Main Steam Stop valves to close which, under normal power conditions, would prevent the 15G-8T-2 and 15G-9T-2, Generator Output Breakers from opening. In this case the 15G-8T-2 and 15G-9T-2, Generator Output Breakers should NOT open. Therefore it was deemed unnecessary to add the malfunction.

Added noun names to the breakers and disconnect. 11111/08 Slightly rewording of stem and complete rewording of four choices per NRC comments. 02/03/09 NRC K/A System/E/A System 062 A.C. Electrical Distribution Number A4.01 RO 3.3 SRO 3.1 CFR Link (CFR: 41.7 145.5 1 to 45.8)

Ability to manually operate and/or monitor in the control room: All breakers (including available switchyard)

Page 124 of 268 Printed on 211912009 at 16:50

RO and SRO Exam Questions (No "Parents" Or "Originals")

ROI I-SRO Ques #: 48 Question ID: 8073622 m RO c' SRO [IStudent Handout? Lower Order?

Rev. 0 Selected for Exam Origin: Mod 'J Past NRC Exam?

The plant has just tripped from 100% power and the following conditions now exist:

- All breaker indicating lights for Bus 24C are deenergized.

- Breaker indicating lights for TCBs # I and #5 are deenergized.

- The appropriate breaker indicating lights for all other buses are lit.

- Buses 24A and 24C are deenergized.

All other plant equipment is functioning as designed, based on the given plant conditions.

Which of the following describes other system or component responses to the loss of electrical power, WITHOUT operator actions?

..................... rn m . . . . . . . . . . . . . . . . , . . . , , . . . . . . . , , . . . . . . . , , . , , , . m , , , . , , , , , , , , .

A At least one Condensate pump and two RCPs were lost.

[7 B "A" EDG is running with ONLY the emergency trips available.

17 C #I Atmospheric Dump Valve controller on C05 is deenergized.

D The " A and "C" RCPs are running with NO cooling water flow.

I

~uzficiiti, D - Correct; This is the indication for the loss of DV-10 The loss of DC (control power) will also cause a loss of 24A & 24C on the trip, because the RSST-24C breaker and the "A" DIG output breaker cannot close. With no facility 1 power there is no facility 1 RBCCW, so the two RCPs are running without cooling water and should be immediately tripped manually.

A -Wrong, The Condensate Pumps and RCPs are powered from Buses 25A & B, which still have power. This would be true if the DC bus 201A was lost, but that would cause the breaker lights on 24A to also deenergize.

B -Wrong; The "A" EDG will start on a loss of DC (DV-10) with only "overspeed" protection. The "emergency"trips are NOT available This would be the expected condition on a normal loss of offsite power start.

C -Wrong, The # I ADV controller is powered by VA-10, NOT DV-10 Vital control power bus VA-10 is still being powered from 201A.

References 1 AOP-2506A, Loss of DV-10 Load List Comments and Question Modification History Modifled choice "D" (correct answer) slightly to clar~fy"negative" wordlng per NRC comment. 02/03/09 NRC KIA System/E/A System 063 DC Electrical Distribution System Number A4.01 RO 2.8* SRO 3.1 CFR Link (CFR: 41.7 145.5 to 45.8)

Ability to manually operate andlor monitor in the control room: Major breakers and control power fuses Page 127 of 268 Printed on 211912009 at 1 6 5 0

RO and SRO Exam Questions (No "Parents" Or "Originals")

Question ID: 8056971 @j RO SRO [I student Handout? Lower Order?

Rev. 1 @ Selected for Exam Origin: Mod Past NRC Exam?

Following an automatic start of both Diesel Generators (DGs) due to a Loss of Normal Power (LNP), a "DIESEL GENERATOR 13U TROUBLE" alarm is received.

A PEO, dispatched to investigate the alarm, reports that the "JACKET COOLING TEMP HIGH" alarm is active and the Service Water (SW) flow meter to the 'B' DG is reading abnormally LOW.

Which of the following would cause this condition?

............................. m , . m rn rn rn m . rn rn rn rn . . , a m rn rn rn rn 0A The Diesel Generator Jacket Cooler has a tube rupture.

B The Diesel Generator Air Cooler SW Inlet is pl~gged.

C The Diesel Generator SW Bypass Valve is full open.

D The Diesel Generator SW inlet crosstie is open.

Justification 1 B - Correct, The SW supply to the DG flows through the Air, Lube Oil and Jacket Coolers in series. Therefore, a blockage in any one of them would result in the stated conditions.

A - Wrong, The SW strainer delta-P is normally about 4 psld, therefore, a low reading is NOT indicative of low cooling water flow.

C -Wrong, Although this would result in a slightly lower SW flow, the SW flow instrument is upstream of the bypass valve. Therefore, this is NOT indicative of the low SW flow seen on the meter.

D - Wrong; Cross-tying the two SW supply headers puts the two supplies in parallel to the single DG supply line. This would NOT lower the supply pressure an appreciable amount and; therefore. NOT lower DG SW flow References I ARP-25918-009, "B" EDG Jacket Coolant Temp. High alarm response.

Comments and Question Modification History Modified question based on NRC comments and suggested changes. 02/03/09 Fixed typo in Choice A. NRC comment. 2119109 NRC KIA System/E/A System 064 Emergency Diesel Generators (EDIG)

Number K1.02 RO 3.1 SRO 3.6* CFR Link (CFR: 41.2 to 41.9 145.7 to 45.8)

Knowledge of the physical connections andlor cause- effect relationships between the EDIG system and the following systems: DIG cooling water system Page 131 of 268 Printed on 211 912009 at 16:50

RO and SRO Exam Questions (No "Parents" Or "Originals")

Question ID: 8000072 RO U SRO L] Student Handout? Lower Order?

Rev. 0 Selected for Exam Origin: New Past NRC Exam?

While operating at 100% power, all conditions normal, a VENT STACK RADMONITOR HIIFAIL annunciator alarms on C-0617. The Kaman Radiation Monitor, HM 8168, is intermittently reading a maximum value of 1E+5 microcurieslcc. NO other Process or Area Radiation Monitor shows any upward trend.

Which of the following describes the consequences of this condition and the required response?

A Gaseous Radwaste discharges CANNOT be performed on either unit.

Perform the actions of OPS Form 2617A-007, Augmented Sampling Requirements.

B The Main Exhaust Fans are tripped to prevent an unrnonitored release.

Override and start Main Exhaust Fans as required for normal ventilation.

M C The Millstone Stack Gaseous Radiation Monitor, RM-8132B, is inoperable.

Place the Stack Gas Effluent Purge Valve, RV-8132, in the 'DISABLE' position.

D The Radwaste Ventilation discharge paths to the Millstone stack are automatically isolated.

Purge RM-8168 and restore isolated discharge paths to a normal alignment.

C is correct. Intermittent voltage spikes in the detector power supply are causing intermittent alarms on the Kaman is at 2E-2 microcuries/cc. If the Kaman reaches the alarm setpoint, it will automatically cause the Millstone Stack Gaseous Radiation Monitor, RM-81328, to purge. RM-81328 is inoperable while in purge; therefore, ARP-2590E-061 requires the 'Stack Gas Effluent Purge Valve, RV-8132, to be placed in the 'DISABLE' position on a failure of RM 8168.

A is incorrect. Although RM-8168 and RM-81328 are gaseous monitors for the Millstone stack, neither are required to perform Radwaste discharges from either unit. Augmented sampling requirements are implemented only if RM-8132B cannot be restored to operable.

B is incorrect. Main Exhaust Fans are automatically tripped on a ClAS to prevent an unplanned release, not on a high radiation signal D is incorrect. Discharge paths to the Millstone stack are NOT automatically isolated on a high radiation signal; therefore, a failure of RM-8168 will NOT result in the isolation of any discharge paths. 3ther Radiation Monitors in alarm will cause various discharge paths to isolate.

Q-50, ARP-2590E-061, Vent Stack HI Rad Replaced question. Original question overlapped SRO question $83. NRC comment. 2/19/09 Per NRC comments; 02/19/09

- Modified Choice "8" & "D" per NRC comments.

- Clarified cause of intermittent alarms in justification.

Per NRC comments; 02/23/09

- Fixed spelling error in Choice "D", Ventilation init~allyspelt with 2 "L's"

- Modified first two words in Choice "D"; "rad. waste" changed to "Radwaste" and first letter in "Ventilation" now capitalized.

NRC K/A System/E/A System 073 Process Radiation Monitoring (PRM) System Number A2.01 RO 2.5 SRO 2.9" CFR Link :CFR: 41 5 143.5 145.3 145.13)

Ability to (a) predict the impacts of the following malfunctions or operations on the PRM system; and (b) based on those predictions, use procedures to cor- rect, control, or mitigate the consequences of those malfunctions or operations: Erratic or failed power supply P a g e 133 o f 268 Printed o n 2/23/2009 at 10:32

RO and SRO Exam Questio~ls(No "Parents" Or "Originals")

RO/ I-SRO Ques #: 51 Question ID: 1000069 [17J Rf3 SRO u Student Handout? ti Lower Order?

Rev. 1 [17J Selected for Exam Origin: Bank Past NRC Exam?

'The plant was tripped and EOP 2536, ESDE, entered after EOP 2525 due to a large Main Steam Line Break on #2 SG inside containment.

The following conditions existed at the time:

- #2 SG has just blown dry but has not yet been isolated.

- RCS temperature and pressure have been stabil zed with Th subcooled margin at 94" F.

- Containment pressure is 21 psig and lowering.

- Containment temperature is 260 O F and lowering.

- There are NO indications of any fuel clad failures.

Suddenly, pressurizer level and sub-cooled margin start lowering.

RCS temperatures are stable.

The STA reports that helshe suspects a SGTR has occurred in #2 SG.

Which of the following radiation monitor indications would change if the only additional casualty was a tube rupture on #2 SG?

m . . . . . . . . . . . . . . . . . . . . . . . . . m . m m . m m . m , , , , , , m . . . . . . . . . , . . . , . . . . . . . . . . . , m , , , , , , , , . . . . m A Containment Atmospheric Radiation Monitors 0B Steam Jet Air Ejector Radiation Monitor C Facility 2 Main Steam Line Radiation Monitor D Refueling Floor Area Radiation Monitor

~ustificati& 1 D - CORRECT: With low RCS activity and the ruptured SG already faulted this RM and the personnel access hatch area RM are the only RMs capable of alarming. The STA will note this problem when performing the Safety Function Status Check for EOP-2536.

A -WRONG: Containment Atmospheric Rad. Monitors sarr~plingpath was isolated on the ClAS triggered by high containment pressure.

B -WRONG: Ordinarily one of the first indications of a SGTR. However, the MSlVs closed on the MSI from high containment pressure, therefore, the SJAE do NOT have a steam supply.

C - WRONG: The location of RM and 30 mrlhr alarm setpoint would require significant clad failure for alarm to come in (design function).

References I EOP-2536, Safety Function Status Check Main Steam System Diagram Comments and Question Modification History No NRC Comments NRC KIA System/E/A System 073 Process Radiation Monitoring (PRM) System G e n e r m ~ e c t e di NRC KIA Generic System 2.1 Conduct of Operations Number 2.1.31 RO 4.6 SRO 4.3 CFR Link (CFR: 45.12)

"Ability to locate control room switches, controls and indications and to determine that they are correctly reflecting the desired plant lineup."

Page 135 of 268 Printed o n 211 912009 at 16:51

RO and SRO Exam Questions (No "Parents" Or "Originals")

Question ID: 8071544 RO SRO '

IStudent Handout? n Lower Order?

Rev. 0 Selected for Exam Origin: Mod Past NRC Exam?

The plant is operating in MODE 1 at 100% power with the following conditions:

- Long Island Sound injection temperature is 73°F'

- " A and "C" Service Water Pumps are supplying Facility 1 and 2, respectively.

- Bus 24E is aligned to bus 24C.

Then, the 'A' Service Water Pump trips on overload. The BOP attempted to start the "B" Service Water Pump on Facility 1, but the breaker would NOT close. Subsequently, the " A RBCCW Header high temperature alarm annunciates. Within a few minutes, the BOP informs the US that the " A RBCCW heat exchanger outlet temperature is reading 121°F and rising.

Which of the following describes the minimum procedurally required actions for these conditions?

A Restore the Facility 1 Service Water Header within one hour or commence a plant shutdown.

i]B Open both Service Water header cross tie valves and commence a normal plant shutdown.

C Enter the Tech Spec for degraded Facility 'I RBCCW and commence a Rapid Downpower.

D Secure the Facility 1 RBCCW Pump and manually trip the reactor and the " A & "C" RCPs.

,Justification 1 D - CORRECT, The design temperature of the RBCCW system is 120°F In accordance with AOP 2565 (Loss Of Service Water),

Section 10, ~fRBCCW heat exchanger outlet temperature approaches 120°F (or higher) and restoration is NOT imminent, the associated RBCCW pump must be tripped Also, AOP-2564, (Loss of RBCCW) gives guidance on RBCCW Heat Exchanger outlet temperature of >I20 "F, which requires a plant trip A - WRONG; AOP-2565 and AOP-2564 glve guidance for logg~nginto varlous TSAS due to the loss of a Service Water and RBCCW headers. This could infer Tech. Spec. 3.0 3 applies due to the many systems and components affected by this malfunction.

B -WRONG, Procedural guidance exists for this action, but it is administratively prohibited in this mode of operation.

C -WRONG; When RBCCW header temperature exceeds 120 OF, the header is inoperable. Ordinarily, an inoperable coollng system header would requlre logging into the applicable TSAS and performing a plant shutdown. However, at this temperature the RBCCW header is NOT considered just degraded, but LOST, and the appropr~ateactions must be taken due to the vulnerability of the affected heat loads References I AOP-2565, Loss of Service Water, St. 10 3 AOP-2564, Loss of RBCCW, St 3 3 1 (Contingency)

Comments and Question Modification History Modified question choices based on NRC comments. 02/03/09 NRC KIA System/E/A System 076 Service Water System (SWS)

Number K1.09 RO 3.0' SRO 3.1' CFR Link (CFR: 41.2 to 41.9 145.7 to 45.8)

Knowledge of the physical connections and/or cause-effect relationships between the SWS and the following systems: Reactor building closed cooling water Page 138 of 268 Printed on 211 912009 at 1 6 5 1

RO and SRO Exam Questions (No Parents" Or "Originals")

RO/I-SRO Ques #: 53 I Question ID: 8053533 RO SRO 1 Student Handout? 1 1Lower Order?

Rev. 4 Selected for Exam Origin: Mod Past NRC Exam?

'The plant was operating at loo%, all components are operating normally with the following Instrument Air alignment:

- The "D" Instrument Air (IA) compressor is aligned to operate in lead.

- The "En and "F" Air Compressors are in Standby.

Then, the reactor automatically tripped due to a state wide blackout and loss of the grid. On the trip, Bus 24C deenergizes due to a bus fault.

All other plant equipment responded normally to the existing conditions.

The crew has just entered EOP 2525, Standard Post Trip Actions.

What is the present status of the Instrument Air System?

..................... m m m m . . . . . . . . . . . . . . . . . . . . ~ . . . . . . . . . . . . . . . . . . . . , . m m . . . . m m m m . . a .

A Only Unit 3 is available to supply Station Air, which must then be cross-tied to the IA header.

B The "F" IA compressor is running or will automatically start on low IA header pressure.

C The "F" IA compressor is available, but will NOT run until manually started locally.

U D Only the "DMIA compressor is running, with "F" IA Compressor available for backup.

Justification I C - CORRECT; "F" IA compressor is the only compressor that has power. However, the "F" IA Compressor (IAC) will NOT auto-start even after the bus IS reenergized by the EDG Both "EM& "F" lACs must be given a local start signal to reset them back in "auto" mode after a loss of power.

A -WRONG; This was correct before the new "vital" lACs ("En& "F") were recently installed. The "F" IAC has power available, but requires local operator action to supply the IA header.

B - WRONG, The "F" IAC requires local operator action to restore it to "auto start" mode.

D -WRONG; "D" IA Compressor is powered from 22C (Non-Vital480 VAC), which is deenergized because of the 24C bus fault References I EOP-2525, St. 18, Subsequent Action for IA restoration.

OP-23288, Discussion on IAC Auto Start Requirements Comments and Question Modification History Changed the stem to state that both "Emand "F" Air Compressors are in standby. Previous wording had one compressor available and one in standby. New wording more accurately reflects actual conditions.

Changed the wording in Answer C from ' ..reset locally' to '. .started locally'. More acturately reflects the action taken. 11/11/08 NRC KIA System/E/A System 078 Instrument Air System (IAS) r ~ i n e r i ~l~silectedl c -

NRC KIA Generic System 2.2 Equipment Control Number 2.2.44 RO 4.2 SRO 4.4 CFR Link (CFR: 41.5 / 43.5 / 45.12)

Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

Page 141 of 268 Printed on 211912009 at 1651

RO and SRO Exam Questions (No ttParentsllOr "Originals")

ROI I-SRO Ques #: 54 Question ID: 54821 RO SRO Student Handout? U Lower Order?

Rev. 4 Selected for Exam Origin: Bank u Past NRC Exam?

Core alterations are in progress on Unit 2. The #I Steam Generator (SIG) upper manways have been removed to perform an inspection of the moisture separators. #2 SIG is intact.

It was subsequently discovered that a Main Steam Safety Valve had been removed from the " A Main Steam Header for maintenance, instead of removirrg a safety from the intended "B" Main Steam Header.

Which of the following is a required action in response to this discovery?

....................................... I , . . . . . . . . . . . . . . . . . . . . . . . m , m . m m , . . . m , . r m , , .

@ A Suspend the movement of any irradiated fuel assemblies in CTMT.

fl B Verify plant ventilation is maintaining a negative pressure in CTMT uC Verify reactor vessel level is maintained 4" below RCS centerline.

D Ensure a contingency plan exists to close the " A MSSV opening.

Justification I A - CORRECT, T S 3.9.4 - With the manways removed, a flow path now exists from the CTMT atmosphere, into the SIG through the manwavs, out the SlGs down the Main Steam line (~enetratinaCTMT) and out the hole in the Main Steam l~ne left bv the removed safety val;e. This would allow a direct path for radibactivity riea:;e tdthe atmosphere if fuel damage were to occur in CTMT B -WRONG; OK if "Core alterations" were NOT in progress.

C - WRONG; This assumes the LOWER SIG manways (RCS loo0 area) were removed.

D -WRONG; Although FME is a concern, this assumes concern is for foreign object damage only.

References I Main Steam Diagram showing path through CTMT wall.

Tech. Spec Action Required for violation of CTMT Integrity.

Tech Spec Definition of "Core Alteration" and CTMT Integrity TS.

Comments and Question Modification History I Added the sentence, "#2SIG is intact." To provide information as to status of #2 SIG. For clarity n ~ and added that it was a subsequent discovery that a safety had been removed Deleted reference to the the Auxiliary ~ u i l d i PEO from " A Main Steam header instead of the "6"Main Steam Header. The discovery would not likely be from the Aux Building PEO Modified four choices per suggestions in NRC Comments. 02/03/09 NRC WA System/E/A System 103 Containment System Number K3.01 RO 3.3' SRO 3.7* CFR Link (CFR: 41 7 145.6)

Knowledge of the effect that a loss or malfunction of the containment system will have on the following: Loss of containment integrity under shutdown conditions Page 143 o f 268 Printed o n 2/19/2009 at 16:51

RO and SRO Exam Q Question ID: 8000020 rn R.3 SRO Student Handout? Lower Order?

Rev. 0 Selected for Exam Origin: New Past NRC Exam?

The plant is at 100% power, steady state, with the following additional conditions:

- RCS leakage has risen over the last couple days to approximately 1.5 gpm.

- HP and Operations are performing an Emergency Containment Entry to investigate the source of leakage.

- All other plant conditions and systems are normal.

The team has entered containment (CTMT) and observes that 2-CH-442, the Letdown Header isolation to the RCS, has a small body-to-bonnet leak.

While exiting CTMT, the door interlock mechanism fails, causing both doors to jam open about 50%.

Various ventilation radiation monitors are now beginning to slowly rise.

Which one of the following describes a consequence of the broken CTMT air lock and an applicable mitigating strategp m . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .....................................,,,,........,.

A Radiation from CTMT that is entering the Enclosure Building is a "ground release". Main Exhaust must be aligned to Millstone Stack.

B CTMT radiation is entering the Enclosure Building. Initiate an Enclosure Building Purge and evaluate securing Main Exhaust fans.

C CTMT leakage is now much greater. Ensure all Enclosure Building doors are closed and start any available Main Exhaust fans.

D The CTMT Barrier is lost and the RCS Barrier is degraded. A plant trip is required, followed by an immediate cooldown to Mode 5.

Justification I B - Correct; With the rise in leakage from CTMT, an Enclosure Building (EB) Filtration Actuation Signal will start the EB Fans, filter the radiation released from CTMT with HEPA filters and realign EB ventilation to the millstone stack to eliminate the "ground release" effect.

A - Wrong; main exhaust can NOT be aligned to the millstone stack. It discharges only to the MP2 stack, which is considered a "ground release" C - Wrong; CMTM Integrity is violated as long as the air lock doors are not closed and sealed. Integrity can NOT be re-established regardless of actions taken to secure the EB.

D - Wrong; The barriers are lost, but a plant trip is NOT required and not necessarily conservative. A trip from 100% power will put a lot of stress on plant components, including the leaking valve, and should not be done when a Rapid Downpower is an option.

References 1 OP-2314G, Enclosure Building Filtration System Discussion EBFAS D~agram Comments and Question Modification History Much d~scussionon whether or NOT this is RO knowledge Based on WA importance - yes Modified four choices based on suggestions in NRC comments. 32/03/09 Per NRC comment; corrected typo in choice "B"; 'Initiating' changed to 'Initiate'. 02/19/09 NRC WA System/E/A System 103 Containment System Number A2.05 RO 2.9 SRO 3.9 CFR Link (CFR: 41.5 / 43.5 / 45.3 / 45.13)

Ability to (a) predict the impacts of the following malfunctions or operations on the containment system-and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations Emergency containment entry Page 147 of 268 Printed on 2/19/2009 at 1 6 5 1

RO and SRO Exam Questions (No "Parents" Or "Originals")

ROi I-SRO Ques #: 56 Question ID: 8080579 RO n SRO 3 Student Handout? I (Order?

Lower Rev. 0 Selected for Exam Origin: Mod Past NRC Exam?

The reactor is at Middle of Life (MOL), and has tripped frorri 100% power equilibrium conditions. The cause of the trip has been determined and corrected with NO other maintenance issues.

Critical rod position has been calculated for a reactor startup eight (8) hours after the trip and the RCS boron concentration has been adjusted per the ECF).

Which of the following conditions would cause the actual critical rod position to be lower than the predicted critical rod position?

.................................... m . . , , , . . . ~ , . , , . . . . , , , . . . . . , . . . . . , , , . . m m m , , m . . .

A Startup is delayed for an additional six (6) hours beyond the time used for the original ECP calculations.

B When Borating to the RCS, CVCS was accidentally aligned to the " A BAST, instead of the RWST.

C While performing the reactor startup, the #I Steam Generator steam flow transmitter begins to fail high D Beginning of Life curves were used by mistake in performing the ECP calculations used for the startup.

Justification I A - Correct; Delaying the startup an additional 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months w ~ lput l the startup well after Xenon has peaked and the concentration has lowered below whatlevels would have existed 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months affer the trip. he lower Xenon concentration amounts to a positive reactivity addition, resulting in a lower CEA height for criticality.

B - Wrong; The BAST have a higher concentration of boron than the RWST, which results in an increase in boron in the RCS and a negative reactivity addition, or a higher CEA height for criticality.

C -Wrong; At greater than 15% power, this would cause the Main FRVs to open and over-feed the Steam Generator, causing a drop in RCS temperature and a positive reactivity addition. But at the power level that a reactor startup is performed, steam flow has NO affect on the amount of feed water going to the SG.

D - Wrong; BOL curves would have assumed a much greater excess reactivity present in the core. This would result in the ECP requiring a higher RCS boron concentration.

References I Reactor Engineering, Curve and Data Book, MOC Life Post-Trip Xenon Decay Comments and Question Modification History If peak Xenon is assumed to occur 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months post trip, then a delay of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (Answer 'A') will result in the same Xenon concentration as the original ECP. If the startup is delayed 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , Xenon is less than the ECP.

Changed distractor "B" to state that boration was from the BAST instead of the RWST. Originally was reversed, making the distractor a correct answer. 11111108 NRC comment of typo in choice "D" corrected in initial exam review. 02103109 NRC KIA System/E/A System 001 Control Rod Drive System Number K5.13 RO 3.7 SRO 4.0 CFR Link (CFR: 41.5145.7)

Knowledge of the following operational implications as they apply to the CRDS: Effects of past power history on xenon concentration and samarium concentration Page 150 of 268 Printed on 2/19/2009 at 16:51

RO and SRO Exam Questions (No "Parents" Or "Originals")

Ro/I-sRoQues# 57 1 Question ID: 81726 RO SRO n Student Handout? Lower Order?

Rev. 0 Selected for Exam Origin: Bank 0Past NRC Exam?

The plant is at 98% power with Group 7 CEAs at 170 steps withdrawn.

The Plant Process Computer has just been lost (totally shutdown).

Which of the following describes when Manual Group rod motion will automatically stop, if Group 7 rods are now withdrawn in Manual Group mode?

[7 A When the FIRST rod in the group reaches the lJpper Electrical Limit.

When ALL the rods in the group reach their Upper Electrical Limit.

C When the LAST rod in the group reaches the Upper Core Stop.

D When the FIRST rod in the group reaches the lJpper Core Stop.

Justification I B - CORRECT; UCS is driven by the PPC, therefore, the CEAs will only stop when they reach the UEL.

A -WRONG; The UEL does NOT input into the CEA Group controllers in the CEDS Logic Cabinets. It inputs directly into the Individual CEA controllers for the purpose of stopping ALL withdraw commands before the CEA reaches the mechanical limit of the CEDM.

C -WRONG, This is when group withdrawal would normally be stopped by the Upper Core Stop. However, the UCS is driven by the PPC and is, therefore, unavailable.

D - WRONG, This is a common misconception of when group motion would be stopped by the Upper Core Stop, based on an UCS setpoint of 177 PULSES (RPI from the PPC) and the a UEL setpoint of 180 steps (RPI from reeds). When rods are being withdrawn and a slight misalignment within the group exists, ofien the difference between the two readings makes it appear that rod motion does not stop until at least one CEA has reached its UEL.

References I OP-2302A, Pg. 48, Attachment 5 Comments and Question Modification History Reworded question to improve legibility.

Per NRC comments, removed unnecessary acronyms in stem and the four choices. 02/19/09 NRC KIA System/E/A System 014 Rod Position Indication System (RPIS)

Number K4.01, RO 2.5* SRO 2.7* CFR Link (CFR: 41 5 / 45.7)

Knowledge of RPIS design feature(s) and/or interlock(s) which provide for the following: Upper electrical limit Page 152 o f 268 Printed o n 211912009 at 16:51

RO and SRO Exam Questions (No "Parents" Or "Originals")

ROI I-SRO Ques #. 58 Question ID: 8064354 RO SRO U Student Handout? LJ Lower Order?

Rev. 0 Selected for Exam Origin: Mod Past NRC Exam?

A plant down power is in progress with present power level at -12% and dropping slowly.

Three of the four RPS Linear Power Range bistables have been reset (LEDs have gone out). However, the Channel "D" Linear Power Range bistable will NOT reset (the LED remains lit and is not blinking).

I&C investigation reveals the RPS Channel "DMlevel 1 bistable is failed in the "armed" state, but all other components of Channel "D" are operating normally and are expected to continue functioning as designed.

Which one of the following describes the effect of the RPS Level 1 Bistable's existing status?

.......................... m m . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . rn rn rn m . m . m .

A The PDlL CEA Motion Inhibit will remain enabled.

B The Zero Power Mode Bypass CANNOT be bypassed on Channel "DM.

c A Turbine trip will NOT result in a Channel "D" Reactor trip.

D The Local Power Density trip on Channel "D" is still armed.

Justification (

D - Correct; Level 1 Bistables will "reset" below 15% NI power as sensed by the Linear channels to bypass the turbine trip and LDP trip for that channel of RPS. Therefore, they are still armed for this channel.

A - Wrong; The CEAPDS PDlL CMI is bypassed at 10E-4% by the Wide Range Level 2 bistable, not the linear range channels. If ONE wide range channel bistable fails to reset, then the CEAPDS PDlL CMI will remain enable (cannot be bypassed).

B -Wrong; The Zero Power Mode Bypass is enabled by the Wide Range Level 1 bistable, not linear range channels. If the Level 1 bistable failed on Channel "DMWide Range, then Channel "D" ZPMB could not be bypassed.

C - Wrong; RPS channel "D" will process the turbine trip and trigger. However, because of the 214 logic, the reactor will NOT trip. The level 1 bistable (Turbine triplReactor trip) is bypassed at 15% linear range power lowering.

References I OP-2205, R-14, Pg 14 8 15 Comments and Question Modification History Changed Distracter " A from " signal to the 15G-8T-2 and 9T-2 is blocked ' to '...signal from the "D" Turbine Stop valve is blocked.'

This is a more plausible distracter. 11111108 Changed distracters A, B, and C to address NRC comments. 314109 NRC WA System/E/A System 015 Nuclear Instrumentation System Number K6.04 RO 3.1 SRO 3.2 CFR Link (CFR: 41.7 145.7)

Knowledge of the effect of a loss or malfunction on the following will have on the NIS: Bistables and logic circuits Page 154 of 268 Printed on 211 912009 at 16:51

RO and SRO Exam Questions (No "Parents" Or llOriginalsll)

Question ID: 8600105 [dRO SRO a student Handout? Lower Order?

Rev. 0 Selected for Exam Origin: Mod Past NRC Exam?

The plant is in normal operation at 100% power with all systems and components aligned normally and functioning as designed.

Then, the Loop 1 Thot input to the Reactor Regulating System suddenly fails low.

Which of the following actions must be verified or taken to ensure Pressurizer level is maintained on program for the actual plant conditions?

C] A Verify the Plant Process Computer has "bypassed" the failed Thot input and calculated outputs are at pre-failure levels.

B Verify the Foxboro IA program has "bypassed" the failed Thot input and calculated outputs are at pre-failure levels.

C To ensure proper operation in the event of a trip, select "Local-Setpoint"on the selected pressurizer level controller.

D To ensure proper operation in the event of a trip, transfer PZR level control to Channel 'X' on the Foxboro IA controller screen.

Justification I B - Correct, the Foxboro IA will automatically de-select an input that 1s fa~ledout-of-range and use only the other loops Thot for the calculation of pressurizer level program setpoint and steam d u m ~ valve auto demand setpoint. However, it should be verified that this occurs per the design.

A - Wrong, Although the Foxboro IA is controlled from a terminal and screen used for interface with the PPC, the program is running on a totally different computer system Also, the Foxboro IA sends data to the PPC for use and display, NOT the other way around.

C -Wrong; This will PREVENT the RRSlFoxboro IA from controll ng pressurizer level as designed in the event of a plant trip D -Wrong; This action may be warranted if the Foxboro IA Tavg signal were to fail low with the failed input. However, when a loop temperature fails low, the Foxboro 1A automatically bypasses the bad input, therefore, Tavg and PZR level control will NOT be affected.

References I RRS-01-C, Reactor Regulating System, Rev. 3, Ch 3, Page 18 Comments s and Question Modification History Changed the failure on Reactor Reg from 1200°F (high) to a low failure. If the temperature fails high and the Foxboro IA does not bypa s the failed input, then nothing happens; if the the temperature fails low and the Foxboro IA does not bypass the failed input, then operator intervention is required. More challenging.

Changed Distractor 'D' to transfer PZR level control to channel 'X' instead of steam dump controls. Makes the distractor more plausible.

Addressed NRC comments. 314109 NRC KIA System/E/A System 016 Non-Nuclear Instrumentation System (NNIS)

Number A4.01 R 0 2.9' S R 0 2.8* CFR Link (CFR: 41 7 145.5 to 45.8)

Ability to manually operate andlor monitor in the control room: NNI channel select controls Page 157 of 268 Printed on 211912009 at 1 6 5 1

I-sRo Ques #: 60 I Question ID: 8100018 [jRO SRO Student Handout? u Lower Order?

Rev. 1 Selected for Exam Origin: Mod Past NRC Exam?

The following conditions exist:

- The plant has tripped due to an unisolable Small Break Loss of Coolant Accident.

- Bus 24C was lost on the trip and cannot be restored. (Bus 24E is aligned to Bus 24C)

- "C" HPSl Pump tripped on overload.

- The RVLMS indicates 0% vessel level [Note: The 7% Level Indication (point #8) has been jumpered out on both channels due to instrument failure.]

- The crew has just completed the Diagnostic Flow Chart.

Which of the following sets of data would provide definite indication that the core is actually uncovered, likely resulting in core damage, and what action must be taken to mitigate the effects of this condition?

rn rn m . . a m m . rn rn ....

A Pressurizer pressure: 600 psia; Maximum HJTC temperature: 520°F; CET Max: 487°F; CET High:

485°F Reduce RCS pressure in accordance with EOP 2532, Loss of Coolant, to obtain LPSl flow.

U B Pressurizer pressure: 600 psia; IMaximum HJTC temperature: 520°F; CET Max: 487°F; CET High:

485°F Energize Bus 24E and start "B" HPSl Pump in accordance with EOP 2532, Loss of Coolant.

p~ C Pressurizer pressure: 700 psia; Maximum HJTC temperature: 540" F; CET Max: 530°F; CET High:

528°F Energize Bus 24E and start "B" HPSl Pump in accordance with EOP 2532, Loss of Coolant.

D Pressurizer pressure: 700 psia; Maximum HJTC temperature: 540" F; CET Max: 530°F; CET High:

528°F Reduce RCS pressure in accordance with EOP 2532, Loss of Coolant, to obtain LPSl flow.

Justification I C - CORRECT; The assumption at Millstone 2 is that the highest CET (CET max) is likely to be somewhat inaccurate; therefore, the next highest CET (CET High) is used when determining subcooling. With 700 psia and CET High at 530°F (the lowest indicated temperature), conditions indicate superheat at the top of the core. Superheated conditions at the top of the core are indicative of core uncovery. The only way to cover the core is to increase Safety Injection flow (or stop the leak). This is accomplished by energizing Bus 24E and starting "B" HPSl Pump. Additional actions will require a plant cooldown to depressurize the RCS and enable LPSl flow.

A -WRONG; At an RCS pressure of 600 psia and CET High at 485, the RCS is at saturation conditions. The core is NOT uncovered with saturation or subcooled conditions. Additionally, if RCS pressure were reduced at this point without the "B" HPSl Pump, it would increase the bulk boiling in the core and cause core uncovery sooner.

B -WRONG; At an RCS pressure of 600 psia and CET High at 485, the RCS is at saturation conditions. The core is NOT uncovered with saturation or subcooled conditions.

D - WRONG; If RCS pressure were reduced at this point, LPSl flow would be established at approximately 360 psia. Without any Safety Injection flow, initially lowering RCS pressure would increase the bulk boiling in the core and cause core uncovery sooner.

References I Pr&d@d EOP-2541 PressureiTemperature Requirements [NOT provided to Examinees]

[Provide Steam Tables during exam]

Comments and Question Modification History Added 'on both channels' to stem to clarify that both channels have the lower probe jumper out. 11111108 Fixed distactors and justification to address NRC comments. 214109 NRC KIA System/E/A System 017 In-Core Temperature Monitor System (ITM)

Number A2.02 RO 3.6 SRO 4.1 CFR Link (CFR: 41.5 143.5 145.3 145.5)

Ability to (a) predict the impacts of the following malfunctions or operations on the ITM system; and (b) based on those predictions, use procedures to cor- rect, control or mitigate the consequences of those malfunctions or operations: Core damage Page 159 of 268 Printed on 2/19/2009 at 1651

RO and SRO Exam Questions (No "Parents" Or "Originals")

Question ID: 8000021 RO SRO 7 Student Handout? Lower order?

Rev. 1 Selected for Exam Origin: New Past NRC Exam?

'The plant tripped due to a Large Break LOCA several hours ago and the following conditions now exist:

- EOP-2532 in progress

- All plant systems and components functioning as designed for the LB-LOCA

- Containment pressure = 2.5 psig and lowering slowly

- Containment Spray has been secured per procedure, based on improving containment conditions

- Containment Hydrogen purge in progress

- All other plant systems and components functioring as designed Then, the break in the RCS gets larger, causing Containment pressure to start rising.

Which of the following describes how the Containment Hydrogen Purge duct work will be protected from an over-pressure condition?

rn m . rn p

J A Hydrogen Purge Dampers must be manually closed from their C-01 control switches.

B Hydrogen Purge Dampers will automatically close on rising radiation levels in CTMT C Hydrogen Purge Dampers will close by manual or automatic re-actuation of a CSAS.

nD Hydrogen Purge Dampers will close by manual or automatic re-actuation of a CIAS.

Justification I A - Correct; The isolation valves must be overridden open and, therefore, must be closed by re-operation of their control switches B -Wrong; This implies that the a subsequent high CTMT radiation signal will still close the-dampers, even though a CIAS signal was overridden to open them High CTMT radiation also closes these dampers, and this condition would be present. However, this is NOT correct because high radiation must already exist in CTMT for the excessive generation of hydrogen to take place (severe fuel damage) Therefore, the Hi Rad signal must have already been overridden to initially open the dampers C -Wrong; Although CSAS is "reset" when spray is secured, and it will automatically reactuate on rising CTMT pressure, this signal will NOT automatically close the purge isolation valves D - Wrong; If CIAS where to reactuate on rising CTMT pressure it would automatically close the purge isolation valves. However, the given conditions do NOT state that the CIAS signal has been reset and, unlike the direction given to secure and reset CSAS, EOP-2532 does NOT direct CIAS be reset on improving CTMT conditions. Although a SlASlClAS signal is required for ESAS to generate a CSAS, the SIAS/CIAS s~gnaldoes NOT have to be reset In order to reset the CSAS signal Unlike CIAS, which requires SlAS be reset on ESAS in order to reset a CIAS signal.

References I LP CSS-01-C, CTMT Purge Isolation Valves wl CIAS.

Comments and Question Modification History Modified question per NRC comments; 02/19/09

- Replaced specific ESAS signals that actuated in stem with "All plant systems and components functioning as designed .."

- Changed accident from LOCA to LB-LOCA to improve plausibility of hydrogen generation.

- Added clairification to chioce "D" justification to improve understanding of plausibility.

NRC KIA System/E/A System 028 Hydrogen Recombiner and Purge Control System (HRPS)

Number K1.O1 RO 2.5* SRO 2.5 CFR Link (CFR: 41.2 to 41.9 145.7 to 45.8)

Knowledge of the physical connections and/or cause-effect relationships between the HRPS and the following systems:

Containment annulus ventilation system (including pressure limits)

Page 161 of 268 Printed on 211912009 at 16:51

RO and SRO Exam Questions (No "Parents" Or '*Originals")

I-SRO Ques #. 62 Question ID: 8000022 RO n SRO a Student Handout? Lower Order?

Rev. 0 Selected for Exam Origin: New Past NRC Exam?

During refueling operations, the KO notices that the audible count rate from the Nuclear Instruments (Nls) appears to have suddenly started to rise.

The US directs the RO to validate the audible count rate change, before initiating any mitigating actions.

Which of the following would verify the audible count rate change heard in the control room is a valid indication?

. . . . . . . . . . . . . . . . . . . . . . . . . . a .......................................................

0A Monitor NI Safety Channel indication on RPS or C04.

U B Request audible count rate indication from the Refuel Machine.

aC Compare wide range channel trends on the recorder on C04.

D Monitor NI Control Channel indication on RPS or C04.

Justification I C - Correct; Any of the wide range channels can be selected to plot on the 604 recorder that is dedicated for those instruments. This would give a quick indication of whether the count rates are actually rising.

A - Wrong; The Safety Channel Nls are in the "power" range and were NOT designed for indication of count rate changes at this level.

B -Wrong; The audible count rate indication heard on the refuel machine is from the same wide range channel heard in the control room. Checking this indication would only verify the audible contra1 circuits both work, NOT that the instrument feeding them is reading correctly.

D -Wrong; The Control Channel Nls are also in the "power" range and were NOT designed for indication of count rate changes at this level.

References 1 LP Wide Range NI Diagram Excerpt from OP-2202,-RXSIU directing use of various Wide Range NI indications Comments and Question Modification History Modified question statement slightly to clearly state that the task 1s to VALIDATE the instrument indicating a rising count rate.

Choices " A 8 "D" are verifing that the student, without the use of any reference material, recognises thatthe vast majority of nuclear power indications in the control room do NOT read in the source range. Based on this reasonning, it is believed that choices " A 8 "D" ARE plausible. Also, as stated in the Justification field, choice "B" is NOT correct because it is monitoring the same instrument that is being called into question. The NRC comment stated that choice "B" could be arguably correct, therefore it is considered plausible.

02105109 NRC KIA System/E/A System 034 Fuel Handling Equipment System (FHES)

Number A4.02 RO 3.5 SRO 3.9 CFR Link (CFR: 41.7 145.5 to 45.8)

Ability to manually operate andlor monitor in the control room: Neutron levels Page 163 of 268 Printed on 211912009 at 16:51

RO and SRO Exam Questions (No "Parents" Or "Originals")

ROl I-SRO Ques #. 63 Question ID: 8000023 M RRD 0 SRO 0 Student Handout? @jLower Order?

Rev. 0 Selected for Exam Origin: New Past NRC Exam?

Which of the following plant scenarios would cause the greatest change in the numerical value of the Moderator Temperature Coefficient (MTC)?

................................................................ m . . . . . . . . . . . . . . . . .

17 A Raising plant power from IxlOE-4% to 1% during the initial reactor startup directly after a refueling outage.

B Raising output from 600 MWe to 900 MWe, recovering from an Emergency Generation Reduction.

C Raising RCS average temperature, from 564°F to the 100% program value, during End-of-Cycle coastdown.

w D Returning plant power to 100% after the recovery of a dropped CEA, during Middle-Of-Life conditions.

I-

~urification D - Correct; Reactor power must be lowered to less than 70%, and turbine load with it, to comply with Tech Specs for a dropped CEA.

Boron injection must be used (due to the dropped CEA) and at MOL conditions, this will raise RCS boron concentrat~onenough to effect the value of MTC (-54 ppm). Once the CEA is recovered, the added boron must be diluted out to return power to 100% Hence the 'relationship between MTC and Boron concentration as T/G load is increased.' The increase in power from 70% to 100% will require a reduct~onof 54 ppm Boron (Excluding the effects of Xenon) from the RCS. Using RE-E-03, MTC vs. Boron Concentration, cycle 19 MOC, this change in Boron will cause MTC to change from -0.015%delta k/k/"F. to -0.017%delta k/kl"F (a change of 0.002

%delta k/k/"F).

A - Wrong; The amount of dilut~onrequired to raise power from 1E-4% to 1% is NOT significant enough to change MTC very much A four decade change in power at this level w~llhave NO impact on MTC.

B -Wrong, Reactor power and RCS temperature are NOT changed, by procedure, in an Emergency Generation Reduction event C -Wrong, RCS Boron concentration is NOT altered during this e ~ e n t RCS temperature is raised to program by lowering turbine load. The 5 "F change in RCS temperature will cause MTC to rise slightly, but the change is insignificant. The affect on MTC from the required boron concentration reduction to achieve a 30% power change is much more significant than the affect seen by a 5°F change in Tavg.

References 1 AOP-2556, Power Reduction Requirement Comments and Question Modification History Changed Distractor 'A' changed 0-30% power to lxlOE-4% power to 1%. To show a wider range of power and still be incorrect.

I Ill1108 Changed distractors slightly per NRC comments. Added more information to justification to clarify why the correct answer is correct and the distractors are plausible. but wrong. 2/5/09 NRC WA System/E/A System 045 Main Turbine Generator (MTIG) System Number K5.17 RO 2.5' SRO 2.7' CFR Link (CFR: 41.5 145.7)

Knowledge of the operational implications of the following concepts as the apply to the MT/B System: Relationship between moderator temperature coefficient and boron concentration in RCS as TIG load increases Page 166 o f 268 Printed o n 211912009 at 16:51

RO and SRO Exam Questions (No "Parents" Or Originals")

KO/I-SRO Ques a: 64 I Question ID: 8056807 [dRO a SRO 7 Student Handout? Lower Order?

Rev. 0 Selected for Exam Origin: Mod Past NRC Exam?

Fuel is being moved in the Spent Fuel Pool (SFP) area during a refueling outage, when a SFP area radiation monitor fails high.

What effect, if any, would this radiation monitor failure have on the SFP ventilation system and can fuel handling operations in the SFP continue?

0A SFP ventilation will shift to AEAS mode, Fuel Handling may continue.

B SFP ventilation alignment will NOT be affected, Fuel Handling may continue.

C SFP ventilation can NOT shift to AEAS mode, Fuel Handling must stop.

D SFP ventilation will shift to EBFAS mode, Fuel Handling must stop.

Justification I B - Correct; AEAS requires the triggering or failure of 2 area rad. monitors to actuate an AEAS (214 logic).

A - Wrong; It is CTMT ventilation (Refuel Pool area) that requires only 1 rad monitor to actuate (114 logicj, NOT SFP ventilation.

C - Wrong; Even with a rad monitor failed, any 2 of the other 3 can triggered and realign the SFP ventilation system to AEAS.

D -Wrong; An EBFAS actuating would block an AEAS and require fuel movement be stopped. However, only a manual actuation or SlAS can trigger an EBFAS.

References I LP ESA-01-C, ESAS Text Explanation of AEAS and actuation log c Comments and Question Modification History Modified per NRC comments as follows: 02/04/09

- Added "if any" to stem.

- Deleted "however" and "therefore" from all choices.

- In choice "D", changed the word "has" to "will".

NRC WA System/E/A System 072 Area Radiation Monitoring (ARM) System Number K3.02 RO 3.1 SRO 3.5 CFR Link (CFR: 41.7 145.6)

Knowledge of the effect that a loss or malfunction of the ARM system will have on the following: Fuel handling operations Page 168 of 268 Printed on 211912009 at 1 6 5 1

RO and SRO Exam Questions (No "Parents" Or "Originals")

Question ID: 8680012 RO n SRO n Student Handout? Lower Order?

Rev. 0 Selected for Exam Origin: Mod Past NRC Exam?

The plant is in "end-of-cycle" coastdown and workers have just begun erecting scaffolding in the "B" Emergency Diesel Generator (EDG) room. None of the plant systems have been tagged out yet.

Then, one of the scaffold bars hits a heat detector above the EDG, which causes the heat detector to fail in the "actuated" mode.

Which of the following describes the effect of this inadvertent actuation of this heat detector on plant systems or components?

......................... . . . . . . . . . . . . . . b m . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . m . . . . e . . . .

A The sprinkler system has NOT actuated because both a heat and a smoke detector actuation are required.

B The 13U DG room deluge has been actuated, ~ uthe t nozzles are NOT spraying down the EDG or the room.

c The 13U Diesel Generator Deluge Supervisory Air System alarm activated on the Fire Panel in the control room.

D The sprinkler system has NOT actuated because a second heat or smoke detector actuation is required.

Justification I B - Correct; Unlike the DC switchgear rooms, a single heat detector actuation will trigger the fire suppression system. However, the sprinkler system in the EDG rooms is a "dry" system. A fusible link in each nozzle must melt before that nozzle can spray down the room A - Wrong; This is how the Vital DC switchgear room fire suppression system works, NOT the EDG fire system. Although there are similarities, the two systems are different in actuation requiremerts.

C -Wrong; This would occur if the scaffolding broke a sprinkler head nozzle (the fus~blelink). The purpose of the supervisory air system is to detect the loss of a nozzle seal.

D -Wrong; Other suppression systems require a trigger on more than one detector to avoid an inadvertent actuation if just such an event occurs. However, because of the fusible links in each noale, the EDG system does NOT utilize that failure prevention method References I LP FPS-04-C, Pages 26,28 & 29 Describe EDG Fire Deluge System Comments and Question Modification History Modified Choice "B" per NRC comments; changed "the EDG and the room" to "the EDG or the room" ('and' changed to 'or') 02/04/09 NRC K/A System/E/A System 086 Fire Protection System (FPS)

Number K6.04 RO 2.6 SRO 2.9 CFR Link (CFR: 41.7 145.7)

Knowledge of the effect of a loss or malfunction on the Fire Protection System following will have on the: Fire, smoke, and heat detectors Page 170 of 268 Printed on 211 912009 at 16:51

RO and SRO Exam Questions (No "Parents" Or "Originals")

Rot I-SRO Ques #: 66 Question ID: 75144 @ Rt3 SRO Student Handout? @ Lower Order?

Rev. 3 Selected for Exam Origin: Bank aPast NRC Exam?

The Letdown Valve Controller on Hot Shutdown Panel C-21 is shifted to MANUAL and given a maximurn output signal.

Which one of the following describes the response rhat will be observed on the control room letdown flow controller on C02?

A The controller output indicator will remain as-is, and letdown flow will not change from the value maintained by the C02 controller.

B The controller output indicator will remain as-is, and letdown flow will go to the maximum allowed by a full open flow control valve.

c The controller output indicator will track to 100%, and letdown flow will go to the maximum allowed by a full open flow control valve.

D The controller output indicator will remain as-is, and letdown flow will go to the maximum flow allowed by the letdown limiter.

Justification _(

B - Correct; The C-21 controller is downstream in the control c~rcuitfrom the C02 controllers and, therefore, have no effect on the C02 controller operation or output. However, the letdown valves will ful!y open, based on the output from the C21 controller because the C21 controller output does NOT go through the Letdown Limiter circuit.

A -Wrong; This is true of the controllers on C10 because those controllers first have to be put in the circuit by the "isolate" switch.

C -Wrong; The flow control valve operation is correct but the indication stated is based on that seen for Foxboro controllers that utilize the Foxboro IA system.

D - Wrong; Controller indication is correct but the letdown limiter is "upstream" of this controllers output and, therefore, has NO effect on the controllers output to the valves.

References I Letdown Flow Control Valve Signal Flow Path Diagram Comments and Question Modification History Added "control roomn to question stem.

NRC KIA System/E/A System 2.1 Conduct of Operations F e n e r i c KIA Selected 1 NRC WA Generic System 2.1 Conduct of Operations Number 2.1.28 RO 4.1 SRO 4.1 CFR Link (CFR: 41.7)

Knowledge of the purpose and function of major system components and controls Page 172 of 268 Printed on 211912009 at 16:51

RO and SRO Exam Questions (No "Parents" Or "Originals")

RO/ I-SRO Ques # 67 Question ID: 8054123 M

- Rt3 SRO Student Handout? Lower Order?

Rev. 1 Selected for Exam Origin: Mod Past NRC Exam?

The plant is in Mode 4, with a plant heatup in progress and the " A & "B" RCPs in service. The control room noted that " A RCP temperatures are slightly above normal and directs two operators to perform a verification of the RBCCW system valve line-up in containment.

While performing the valve line-up verification, the operators observe that the valve controlling RBCCW flow to the " A RCP appears to be opened about half as much as the valve for the "C" RCP. Per the valve lineup, the RBCCW valves to both " A and "C" RCPs should be opened 1+1/2 turns.

Which one of the following actions must be taken by the operators discovering the mispositioned RBCCW valve for the " A RCP?

.......................... rn m , . m m . m . m . . , . .............................. .. a . rn rn rn m . . m .

A The first operator repositions the valve per the valve lineup while the second operator observes the repositioning.

B The first operator repositions the valve per the valve lineup then the second operator performs an independent verification.

aC The two operators stop and contact the control room for authorization before repositioning the valve per the lineup.

D The first operator slowly opens the valve while the second operator leaves to verify RCP conditions are returning to normal.

Justification I

1 C - CORRECT; Per administrative requirements, "If a discrepancy is discovered (the performers of a valve lineup) immediately notify first line supervision for resolution" before proceeding.

A -WRONG; This is the correct method to reposition a throttled valve in the RBCCW system, however, this is NOT allowed without first getting permission from the on-watch SRO.

B - WRONG; NOT allowed even with permission because now only ONE check has been made of the valve's throttled position.

D -WRONG; This would be allowed if the operators had already been given instructions to re-align any mispositioned valves and the system were NOT vital and required an independent verification of all valve positions.

References I OP-AA-5000, Independent Verification Requirements Comments and Question Modification History Rewrote the question and choices using suggestions from the NRC comments. 02/05/09 NRC System/E/A System 2.1 Conduct of Operations 1 Generic WA Selected 1 NRC WA Generic System 2.1 Conduct of Operations Number 2.1.29 RO 4.1 SRO 4.0 CFR Link (CFR: 41 . I 0 145.1 145.12)

Knowledge of how to conduct system lineups, such as valves, breakers, switches, etc.

Page 174 of 268 Printed on 2/19/2009 at 16:51

RO and SRO Exam Questions (No 'lParentsVOr llOriginalsl')

I-sRoQues #: 68 1 Question ID: 8074001 RO SRO 7 Student Handout? Lower Order?

Rev. 1 Selected for Exam Origin: Mod Past NRC Exam?

Which of the following would be considered a CONSERVATIVE decision?

m m m . m . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . m . . m m . . m m .

A Initiation of Once-Through-Cooling at a S/G level > 100" with a total loss of S/G feed flow and bus 24C deenergized.

B Manual initiation of a SRAS if the RWST lowers to 20% level while injecting into the RCS for a large-break LOCA.

C Performing a watch relief to avoid exceeding a 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months shift with a reactor startup approaching its 5th doubling of counts.

D Initiation of Emergency Boration if startup rate exceeds 1.0 dpm while withdrawing CEAs during a reactor startup.

Justification I A - CORRECT; Although the EOPs state OTC must be initiated by the time SIG level reaches 70", guidance exists for early initiation of OTC at 100" level in any SIG, if less than optimal equipment IS available B -WRONG; Early initiation of SRAS is non-conserve because of the loss of inventory to the CTMT sump.

C -WRONG; The reactor startup procedure states that shifl turnover should NOT occur with the reactor approaching criticality due to the reactivity control concern (even in light of the new NRC overtime rules, so long as Operator-At-The-Controls FFD is NOT a concern).

D -WRONG; Although initiation of Emergency Boration would quickly shutdown the reactor at this power level, the reactor startup procedure requires a "reactor trip" if SUR exceeds 1.0 dprn.

OP-2260, Pg. 12, St. 1.8.2 Comments and Question Modification History Rewrote question (now ModifiedlNew) to test for a "Conservative Decision", so that the question is not solic~tinga "negative" action among "positive" act~ons Also, replaced choice "D", thus addressing the two NRC comments. 02/05/09 NRC KIA System/E/A System 2.1 Conduct of Operations 1 Generic KIA Selected 1 NRC KIA Generic System 2.1 Conduct of Operations Number 2.1.39 RO 3.6 SRO 4.3 CFR Link (CFR: 41.10 143.5 145.12)

Knowledge of conservative decision making practices.

Page 176 of 268 Printed on 211 912009 at 16:51

RO and SRO Exam Questions (No "Parents" Or "Originals")

ROil-SRO QuesX: 69 ( Question ID: 64620 RO SRo n Student Handout? Lower Order?

Rev. 4 Selected for Exam Origin: Bank Past NRC Exam?

A "blue dot" has been placed on the annunciator window; (3-0617,D-17, "RADWASTE AREA SUPPLY AIR TROUBLE".

What is the significance of the blue dot?

. . . . . . . . . . . . . . . . . . . . . . . . . . . , , . . . . I ................................................

A The alarm circuit is out of calibration and can not be solely utilized as an indication that mitigating actions are required.

0B The monitored equipment associated with the annunciator is inoperable, but an alarm may indicate a status change.

C The associated alarm logic card has been removed, preventing the annunciator from triggering by the input device.

D The annunciator can be triggered by multiple inputs and one or more of these inputs are known to be out of service.

Justification I C - Correct; Per OP-2387A; A blue dot is placed on an annunciator window to indicate that it is a hanging or nuisance alarm that was taken out of service. An annunciator is considered out of service when the associated alarm (logic) card is removed.

A - Wrong; A trouble Report would be filled for this situation, containing details of the problem. An orange label (with magnetic backing), containing the TR number and date, would be placed or the control board as close to the alarm as possible.

B - Wrong; This implies the equipment is out of service due to a "blue" tag, which would allow technicians to test the component and possibly cause a periodic alarm. However, an item out of service due to a tagout requires a RED dot, regardless of the tag color.

D -Wrong; This alarm condition would require an ORANGE dot on the window.

References )

OP-2387A, Definitions Comments and Question Modification History Per NRC comments, modified stem from " a blue dot on one of the annunciator windows." to state a specific annunciator window Per reviewer, changed "any" in correct answer to "the", to better align with meaning of the blue dot. 02/17/09 NRC WA System/E/A System 2.2 Equipment Control 7

1 Generic KIA S e l e c t e d NRC WA Generic System 2.2 Equipment Control Number 2.2.14 RO 3.9 SRO 4.3 CFR Link (CFR: 41.10 143.3 145.13)

Knowledge of the process for controlling equipment configuration or status.

Page 178 of 268 Printed on 211912009 at 16:51

RO and SRO Exam Questions (No "Parents" Or "Originals1')

Question ID: 80785 RO n SRO C]Student Handout? ,Q Lower Order7 Rev. 2 Selected for Exam Origin: Bank Past NRC Exam?

Duriqg a refueling outage, the " A TBCCW Pump breaker, B0107, is properly red tagged for troubleshooting due to an electrical problem. Electrical Maintenance has determined that the breaker has to be removed and bench tested in the maintenance shop. The intent is to install the original breaker when repairs are complete.

..... ............................,..................................,.....,...~...

The tag on the breaker A must be replaced by an operating permit tag before removal of the breaker from the cubicle.

must remain attached to the breaker until all related work is completed.

oc must be replaced by a yellow tag before removal of the breaker from the cubicle.

D must be removed from the breaker and attached to the cubicle door.

Justification I D is correct. OP-MP-200.1001, Equipment Clearance Process, Attachment 1. Section 3.1 describes the process for removing an MCC bucket from its cubicle when the breaker has been red tagged, as follows: Remove tag, Remove MCC bucket from cubicle, Rehang the tag on the door.

A is incorrect. Operating Permit tags are used to allow a qualified person to operate a device or introduce energy into a system or device. Operating Permit tags are NOT used to remove a component from its normal location; therefore, this would be an inappropriate use of the Operating Permit tag.

B is incorrect. Even though the breaker will be physically removed from the cubicle and the pump cannot be started, the tagging procedure requires that the tag be removed from the breaker and placed on the cubicle door when a breaker is being removed frorn its cubicle.

C is incorrect. A yellow tag is used to inform personnel of off-normal conditions, special instructions, or precautions, but does NOT preclude the operation of a component. The procedure does NOT allow the use of a yellow tag for this condition.

,References I OP-MP-200-1001, Equipment Clearance Process, Attachment 1. Section 3 1 Comments and Question Modification History Incorporated NRC comments Modified distractors "A" and " Cto be more plausible. 2/5/09 NRC WA System/E/A System 2.2 Equipment Control I Generic KIA Selected 1 NRC WA Generic System 2.2 Equipment Control Number 2.2.13 RO 4.1 SRO 4.3 CFR Link (CFR: 41.10 145.13)

Knowledge of the tagging and clearance procedures.

Page 180 of 268 Printed on 211912009 at 16:51

RO and SRO Exam Questions (No "Parents" Or "Originals")

ROI I-SRO Ques #: 71 Question ID: 80625 RO SRO Student Handout? u Lower Order?

Rev. 3 Selected for Exam Origin: Bank C Past NRC Exam?

The plant was operating at 100% power, steady state, with all CEAs fully withdrawn.

Then, CEA #I in Reg. Group 7 drops to the bottom of the core (0 steps withdrawn)

The crew subsequently performs all required actions, per the applicable AOP, and is awaiting the I&C "go-ahead" to begin recovery of the dropped CEA.

It has now been one hour and 50 minutes since the CEA dropped, and I&C has just informed the control room that CEA recovery steps may begin.

Which of the following is the required action?

...................... m m . . . . . . . . . . . . . . . . . . . , . . . . . . . . . , , . . . . . . . . . . . . . m . . . ~ . . m m . m . . m A Immediately commence a plant shutdown to MODE 3 by Boration only.

B Withdraw the dropped CEA to at least 170 steps within the next 10 minutes.

0C Within the next 10 minutes, initiate steps to recover the dropped CEA.

D Trip the plant and maintain the reactor shut down for a minimum of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

Justification I A - Correct; Per the requirements of TSAS 3.1.3.1 action A.l and AOP 2556 step 4.28.k: IF the misaligned CEA is not realigned to within 10 steps of all other CEAs in its' group within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , PLACE the plant HOT STANDBY condition within the next 6-hours.

The remaining 10 minutes is NOT enough time for recovery of the CEA.

B -Wrong: Recovering a CEA this quickly is in violation of the AOP guidance, as it has a strong possibility of damaging the fuel.

Based on the AOP recovery guidelines, it is mathematically impossible to recover the dropped CEA within the Tech. Spec. time limit.

C -Wrong; Initiating the recovery of the CEA within two hours does NOT meet the requirements of TSAS 3.1.3.1 action A.l and AOP 2556 step 4.28.k. The CEA must be within 10 steps if its group within two hours.

D -Wrong; A plant trip is NOT required and would be considered an overly aggressive plant shutdown and a non-conservative action Tech. Spec. 3.1.3.1 CEA Position, Action " A ; Misaligned by >20 Steps.

AOP-2556; Step 4.28, Misaligned CEA Recovery.

Comments and Question Modification History Eliminated the references per NRC comment Changed Distractor "C".Reactor Operators do NOT apply Tech Specs.

NRC WA System/E/A System 2.2 Equipment Control 1 Generic KIA Selected 1 NRC WA Generic System 2.4 Equipment Control Number 2.2.40 RO 3.4 SRO 4.7 CFR Link (CFR: 41.10 143.2 143.5 145.3)

Ability to apply Technical Specifications for a system.

Page 182 of 268 Printed on 211912009 at 16:51

RO and SRO Exam Questions (No "Parents" Or "Originals")

RO/ I-SRO Ques #: 72 Question ID: 8500016 Re3 SRO Student Handout? Lower Order?

Rev. 0 Selected for Exam Origin: Mod Past NRC Exam?

A LOCA Outside of Containment has occurred at the plant. In addition, excessive fuel damage has resulted in radiation levels significantly above normal in various equipment locations. A General Emergency Classification has been declared and the Station Emergency Response Organization is fully staffed.

It was determined that the LOCA could be isolated by manual valve manipulation in an area where the dose rates are approximately 60 REM per hour. An operator has entered the area and isolated the leak, but appears to have suffered a stroke and now needs assistance to leave the high radiation area.

ALL dose extensions necessary for this situation have been granted per the Emergency Exposure Limits guidelines.

Which of the following exposure requirements are still applicable for the volunteer who enters the high radiation area to assist the injured operator?

rn A Any male or female can volunteer; stay time is limited to 6 minutes.

B Only males over 50 can volunteer; stay time limit is 25 minutes.

nC Only males any age can volunteer; stay time is up to that individual.

D Any male or female can volunteer; stay time is up to that individual.

Justification I D - Correct; For "life saving situations" the dose limit per Emergency Exposure Limits is strictly up to the individual who volunteers to give assistance. In this instance, procedures do NOT have different requirements for males or females as the person must be a volunteer.

A - Wrong; This equates to a dose of 5 rem, which is the normal limit for non-emergency scenarios.

B - Wrong; This equates to a dose of 25 Rem, which is the Emergency Exposure limit for non-volunteers performing accident mitigation. The "male over 50" is a company guideline when solic~tingvolunteers for high exposure missions, but it is NOT a requirement.

C -Wrong; This is the correct dose for life-threatening emergency situations. However, although excluding females is plausible, it is NOT an administrative requirement.

References I MP-26-EPI-FAPO9, Radiation Exposure Controls F0-or Question Modification History at this time. 7 NRC K/A System/E/A System 2.3 Radiation Control L ~ e n e r i KIA

-- c ~electe4 NRC KIA Generic System 2.3 Radiation Control Number 2.3.4 RO 3.2 SRO 3.7 CFR Link (CFR: 41 . I 2 143.4 145.10)

Knowledge of radiation exposure limits under normal or emergency conditions.

Page 186 of 268 Printed o n 211912009 at 1651

RO and SRO Exam Questions (No "Parents" Or "Originals")

Question ID: 1000109 B RO SRO student Handout? @ Lower Order?

Rev. 1 Selected for Exam Origin: Bank Past NRC Exam?

Radiography is being performed in the Auxiliary Building and has caused an area radiation monitor to alarm.

The US has directed the applicable module on RC-14 be placed in ALARM DEFEAT until the operation is complete.

Which of the following describes why the radiation monitor's switch is placed in the ALARM DEFEAT position?

............................................................................. I . . . .

A TO silence the radiation monitor's horn on the local module.

B TO reset any automatic action caused by the radiation monitor.

C To clear the radiation monitor's red andlor amber lights on RC-14.

D To allow other radiation monitor alarms to annunciate on C-0617.

Justification I D - Correct; Placing the applicable ALARM DEFEAT switch in the ALARM DEFEAT position will allow other area radiation monitor alarms to be annunciatedon C-0617. The red 'HIGH' and amber "'FAIL' lights will be'lit on the applicable rad monitor on RC-14. The local horn will need to be bypassed with a key on the local module.

A - Wrong; The ALARM DEFEAT switch will NOT silence the local horn.

B -Wrong; The ALARM DEFEAT switch will NOT reset any automatic action caused by the rad monitor. In fact, the ALARM DEFEAT switch will result in a rad monitor failure which will prevent resettir~gany automatic function.

C - Wrong; The ALARM DEFEAT switch will NOT clear the red and amber lights on the RC-14 module. In fact, the ALARM DEFEAT switch will cause the red and amber lights to be lit.

References I ARP-2590E-128, Defeat Area Rad Monitor Alarm N O Comments or Question Modification History at this time:-1 NRC KIA System/E/A System 2.3 Radiation Cortrol 1 Generic KIA ~ e l e c t e q NRC KIA Generic System 2.3 Radiation Control Number 2.3.15 RO 2.9 SRO 3.1 CFR Link (CFR: 41.12 143.4 145.9)

Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

Page 188 of 268 Printed on 211912009 at 1651

Question #: 74 I QuestionID: 8310054 RO SRO Student Handout? Lower Order?

Rev. 0 Selected for Exam Origin: Mod 0Past NRC Exam?

A plant cooldown is in progress using OP 2207.

The protected facility is Facility 1.

RCS temperature is 270" F and pressure is 375 psia vvith the 'A' & 'B' RCPs running.

The Shutdown Cooling System is in recirc for the warmup/pressurization leak check using the 'B' LPSl pump.

'The 'A' HPSl pump is available with its handswitch in Pull-To-Lock

'B' and 'C' HPSl pump breakers are racked out.

Then, a seismic event occurs resulting in the following plant conditions:

- Pressurizer level and pressure are dropping rapidly.

- Containment pressure is <psi I and stable.

- RBCCW surge tank is rising rapidly.

- 'B' RCP has tripped and 'A' RCP has been manually secured.

...Which of the following describes actions required for this event?

m . . . . . . . . . , 1 . . . . . . . , m . , , . . . . ~ . . , . . . , . . , , , , . . . , , , , , , , . , , , . . , , . . . , , . , , , m , , , a . , , . .

A Align 'A' HPSl pump and start it, then secure 'B' LPSl pump, realign LPSI pumps for Safety Injection, then manually start both LPSl pumps.

B Ensure the 'A' HPSl and 'A' LPSl Pumps have automatically started and their associated injection valves automatically have opened.

C Take 'A' HPSl Pump handswitch to OFF, secure SDC and realign LPSl pump suction, and observe automatic start of Facility 1 SI pumps.

D Manually actuate SlAS via C01 push button and verify both Facility 1 SI Pumps automatically start and all safety injection valves realign.

1-t i i c a t o n A - Correct; SlAS is blocked and w~llNOT respond on lowering pressurizer pressure. Accordingly, the 'A' HPSl is required to be in PTL due to lowered RCS temperature and LPSl alignment prevents injection until realigned, these actions are directed by OP 2207 in response to a LOCA.

B - Wrong; This would occur if Facility 1 HPSl and LPSl were maintained fully operable and SlAS was NOT blocked or the LOCA was raising CTMT pressure.

C - Wrong; This would occur if LPSl injection valves were powered and SlAS was NOT blocked or the LOCA was raising CTMT pressure.

D - Wrong; This would occur if LPSl did NOT needs to be manually realigned and the HPSl pump "Pull-To-Lock" feature allowed for automatic start on an ESAS actuation (similar to the Charging Pumps).

References 1 0P-2207, Attachment 9, Step 8, Actions for a LOCA while shutdown k0Comments o r Question Modification History at this time. 1 7

NRC KIA SystemlElA System 2.4 Emergency Procedure /Plan L ~ e n e r i cKIA ~ e l e c- tedA NRC KIA Generic System 2.4 Emergency Procedures /Plan Number 2.4.9 RO 3.8 SRO 4.2 CFR Link (CFR: 41.10 143.5 145.13)

Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.

Page 220 of 295 Printed on 1111412008 at 12:0!3

RO and SRO Exam Questions (No "Parents" Or "Originals")

RO/I-SROQucs#: 75 I Question ID: 8000051 rJl RO U SRO 1 Student Handout? k Lower Order?

Rev. 0 [31 Selected for Exam Origin: New Past NRC Exam?

An RO qualified operator, attached to Crew "D", is cn site, temporarily assigned to the Procedure Group.

Then, Crew "EMoperators, who presently have the watch in the control room, declare an emergency event requiring activation of the Station Emergency Response Organization (SERO). The Unit 2 Shift Manager (SM) makes an announcement over the site paging system, for all SERO members to report to their designated emergency response facility.

Which of the following describes the location where the Crew "D" RO must now report, based on the SM's site page?

. . . . . . . . . . . . . . . . . . . . . . . . . . . . m m m , , , , , m m . I . , , , , , . . . . . . , , , . . . . . . . , , , . . . . . . . . , . . . . m , , ,

A The Unit 2 Control Room.

B Remain in Work Control.

C Building 475 Cafeteria.

D Technical Support Center Justification I C - CORRECT; All "OFF-Shift" Operations personnel are required to report to the Operational Support Center Assembly Area, wh~chis the cafeteria in building 475. he^ will be dispersed from there to various plant locations, as they are needed.

A -WRONG; The RO temporarily assigned to the Procedure Group is NOT considered an "On-Shift" operator, which means he should NOT go to the Control Room. Only operators that actually have the "Shift" are supposed to report to the Control Room during an event when SERO is activated.

B - WRONG; This would only be correct if the announcement stated personnel should seek immediate shelter, such as during a security event or fast moving natural phenomenon.

D - WRONG; Some support personnel with SERO positions, temporarily assigned to the Procedure Group, would report to this location. But Not an RO qualified operator.

References I SERO CBT, Slide 20, Operator SERO Reporting Requirement N O Comments or Question Modification History at this time. 2' NRC KIA System/E/A System 2.4 Emergency Procedure /Plan 1 Generic WAS-NRC KIA Generic System 2.4 Emergency Procedures /Plan Number 2.4.29 RO 3.1 SRO 4.4 CFR Link (CFR: 41.10 143.5 145.11)

Knowledge of the emergency plan.

Page 192 o f 268 Printed o n 211912009 at 16:51

RO and SRO Exam Questions (No "Parents" Or "Originals")

Ques #: 76 1 Question ID: 8000025 IRO SRO Student Handout? Lower Order?

SRO-U Ques. # 1 Rev. 0 m Selected for Exam Origin: New r]Past NRC Exam?

The plant was tripped from 100% power due to a c rculating water header rupture in the East condenser pit which resulted in the condenser pit water level reaching 10 inches. All equipment responds as expected.

The Standard Post Trip Actions of EOP 2525 have been completed and the crew has just entered EOP 2526, Reactor trip Recovery.

While performing EOP 2526, Reactor Trip Recovery, the RSST is suddenly lost.

The following plant conditions exist 15 minutes later:

- Pressurizer pressure is 2060 psia and going up slowly

- Pressurizer Level is 38% and rising slowly

- # I SG pressure is 900 psia and stable

- #2 SG pressure is 950 psia and slowly rising

- # I SG level is 40% and steady

- #2 SG level is 55% and slowly rising

- 'A' DG is supplying Bus 24C

- 'B' DG was emergency tripped because its output breaker would NOT close

- Tcold is 530°F and slowly rising

- Thot is 543°F and rising

- Containment Sump level is 58% and shows a 1% rise every 10 minutes.

- Facility 1 Containment Atmospheric radiation monitors are flagged with an "Unexpected Rise" on the PPC, values are NOT changing.

All other equipment and parameters are as expected.

Which of the following actions must now be directed by the US?

m , I m m , I . m , m . , , , . m , , . . , , . . . , , . . , , * . , , ~ . . , . . . m , , . . , . . . . , m m .

CA Repeat EOP 2525, Standard Post Trip Actions, then transition to EOP 2528, Loss of Offsite PowerlLoss of Forced Circulation.

@ B Re-perform Appendix 1, Diagnostic Flow Chart, then transition to EOP 2528, Loss of Offsite PowerILoss of Forced Circulation.

C Complete EOP 2526, Reactor Trip Recovery, and refer to the AOP(s) that are determined to be appropriate for the existing conditions.

D Re-diagnose conditions, continue with EOP 2526, Reactor Trip Recovery, and refer to AOP 2568, Reactor Coolant System Leak.

Justification 1 B is correct. OP 2260, Rev 009-02, step 1 . 9 . 4 ~ states, "If, during the performance of an ORP, a major change in plant conditions occurs, the US should return to the beginning of the procedure in use and commence the procedure again at the Confirm Diagnosis section, which, if appropriate, will lead to the review of Appendix 1 Diagnostic Flowchart." Therefore, the US is required to return to step 1 and review the Diagnostic Flowchart to determine that the appropriate procedure for the given conditions is EOP 2528.

A is incorrect. If a major change occurs in plant conditions during the performance of EOP 2525, then the US should return to the beginning and commence the procedure again. In this case, EOP 2525 is complete and EOP 2526 is in progress; therefore, it would be inappropriate to repeat EOP 2525, even though the crew will ultimately end up in EOP 2528.

C is incorrect. OP 2260 allows the US to perform AOPs while performing an optimal EOP. AOP 2568 would address the RCS leak and AOP 2583 would address the loss of offsite power in MODE 3, however, the loss of offsite power occurred during the performance of EOP. Which was entered due to a reactor trip. The loss of offsite power EOP is applicable.

D is incorrect. OP 2260 allows the US to perform AOPs while performing an optimal EOP In this case, EOP 2526 would NOT be appropriate due to the loss of the RSST, however, the RCS leak MAY be pursued (NOT required at this time) in conjunction with EOP 2528, Loss of Offsite PowerlLoss of Forced Circulation.

References I OP 2260, Rev 009-02. step 1 . 9 . 4 ~

Comments and Question Modification History

[Too easy] (Add parameters, 12/01/08)

Added indications of an RCS leak and a loss of a vital 4160 volt bus. Reworded distractors to be more plausible for the change in conditions. (Fixed RJA 12/02/08)

In Distractor D, added hyphen to Re-diagnose. 2/9/09

.I)-.- 111- m. .-L--~-IL P..~(..-

Page 194 of 268 Printed on 2/19/2009 at 16:51

RO and SRO Exam Questions (No "Parents" Or "Originals")

Question ID: 8000025 1RO SRO Student Handout? @ Lower Order?

SRO-U Ques. # 1 Rev. 0 @ Selected for Exam Origin: New uPast NRC Exam?

NKL r v aysrernltlw

~ =yaralll 02 Reactor Trip iecovery Number EA2.1 RO 2.7 SRO 3.7 CFR Link (CFR: 43.5 145.13)

Ability to determine and interpret facility conditions and selectiorl of appropriate procedures during abnormal and emergency operations as they apply to the Reactor Trip Recovery.

Page 195 of 268 Printed on 211912009 at 16:51

Question ID: 8000036 I] R() SRO Student Handout? Lower Order?

SRO-U ~ues\# 2 Rev. 2 @ Selected for Exam Origin: New Past NRC E ~ r n ?

\

was about to place Shutdown Cooling (SDC) in Coolant Pumps (RCPs) in service when Pressurizer Pressure Narrow Range the plant?

SDC in service.

B C

The Facility 1 Safety Override and close SDC Suction Isolation 4

Tank (SIT) Outlet Valves will open causing the ITSto inject.

Valves and restore Pressurizer level to t desired setpoint.

3-652, will NOT open until the high Direct I&C to bypass nnect Pressurizer Pressure Narrow D

SpecificationLCO.

Justification (

k The "B" Power Operated Reli Valve (PORV) will open and RCS Immediately place the associat PORV Block Valve to 'Close' a

\ /

P ssure will rapidly lower.

log into the LTOP Technical established by initiating SDC.

NOT align with 51-652 (Facility 2).

References I Plant Cooldown, OP 2207 ARP 25908-209

?\

erous varied comments f m validation and still meeting the KIA, the question Replaced question. Too 9 ny conflicting opinions on fixes. (NRC and Validator) 1/23/09

\

nt to close PORV Block valve. (Does NOT automatically close below NPSH before the PORV Block Valve can fully close. Va y paced the LTOP switch in high and immediately after, placed w NPSH. In this condition RCS pressure is only -25 psia abov NRC K / P / S y s t e m / ~ / ~system 008 Pressurizer (PZR) Vapor Space Accident (Relief Valve ~ t u c $ ~ e n )

NR P KlAGeneric g m b e r 2.2.44 System RO 4.2 2.2 SRO 4.4 Equipment Control CFR Link (CFR: 41.5 143.5 145.12) \

.hbility to interpret control room indications to verify the status and operation of a system, and understand how operator directives affect plant and system conditions.

Page 197 of 268 Printed on 211

RO and SRO Exam Questions (No "Parents" Or "Originals")

I-SRO Q U ~ S#: 78 1 Question ID: 8000027 ;] RO SRO [IStudent Handout? Lower Order?

SRO-U Ques. # 3 Rev. 0 Selected for Exam Origin: New Past NRC Exam?

The plant is in Natural Circulation after a reactor trip from 100% due to a small break LOCA with a concurrent loss of the RSST. Standard Post Trip Actions, EOP 2525, has been successfully performed You have directed the RO to check for single phase natural circulation in accordance with Loss of Coolant Accident, EOP 2532.

Which of the following sets of stable parameters is acceptable (administratively allowed) for Natural Circulation after the cooldown and depressurization has been initiated?

(Assume the cooldown limitations of the Technical Specifications and the Technical Requirements Manual are maintained.)

rn a , m . . . . . . . , , . . . . . . . . , , , . . . . . . . . . . . . . . . . . . , , , . . m m m m rn m , , . . m m m rn A Pressurizer pressure is 812 psia Pressurizer level is 25%

Highest CET is 520°F Tc is 485°F Th is 515°F B Pressurizer pressure is I100 psia Pressurizer level is 12%

Highest CET is 520°F Tc is 455°F Th is 518°F C Pressurizer pressure is 610 psia Pressurizer level is 27%

Highest GET is 455°F Tc is 423°F Th is 443°F D Pressurizer pressure is 931 psia Pressurizer level is 8%

Highest CET is 505°F Tc is 465°F Th is 500°F Justification

- ..-.... - -. . 1 D is correct. Loop delta-T is 35°F (maximum is 55"); Th and Tc are constant (stated in question stem); CET subcooling is 31°F (Minimum operating limit of the PIT curve for CETs is 30°F); Difference between Th and CET temperature is 5°F (maximum is 10°F)'

Pressurizer level ofgreater than 20% is NOT a requirement for natural circulation during a LOCA ( ~ i n i m u mof , 20% for all other events). 8% correlates to the minimum PZR level displayed by the PPC, regardless of how low PZR level then drops.

A is incorrect. A CET temperature of 512°F is the saturation temperature for 812 psia. Although Tc is greater than 30°F subcooled, the requirement for ensuring adequate natural circulation for a LOCA requires CET temperature to be at least 30°F subcooled.

B is incorrect. Although a CET temperature is 36°F subcooled for 1100 psia (meets the requirement of greater than 3OoF), the loop delta-T of 63°F does not meet the requirement of less than 55°F for natural circulation.

C is incorrect. Although a CET temperature of 455°F is approximately 33°F subcooled, Th and CET temperature are greater than 1 O0F,which does not meet the requirements for natural circulation References I EOP-2532, St. 39, Verification of Single Phase NC Flow Comments and Question Modification History Although all reviewers missed this question, it was determined to be valid based on later discussions. This may indicate a knowledge deficiency issue and will be discussed with supervision after the exam is administered 11/14/08

[redo "A" just ] (fixed 12/01/08)

NRC WA System/E/A System 009 Small Break LOCA Number EA2.37 RO 4.2 SRO 4.5 CFR Link (CFR 43.5 145.13)

Ability to determine or interpret the following as they apply to a small break LOCA: Existence of adequate natural circulation Page 119 of 268 Printed on 2/19/2009 at 1651

RO and SRO Exam Questions (No "Parents" Or "Originals")

Question ID: 54119 [IRO [31 SRO C]Student Handout? Lower Order?

SRO-U Ques. # 4 Rev. 4 Selected for Exam Origin: Bank Past NRC Exam?

The following conditions exist:

- A Large Break LOCA occurred approximately 9 clours ago.

- SRAS was initiated approximately 7.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months ago.

- Chemistry Department reports the current RCS sample indicates 1500 ppm boron concentration.

- An RCS sample taken 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ago indicated a boron concentration of 1945 ppm.

- ONLY Facility 1 power is available.

- All other conditions are as expected for this event.

Given these conditions, which one of the following is the procedurally preferred operator action that must be

.. ..................................,........~...........,........,,......,.......

directed by the US?

A Stop the 'A' HPSl Pump, align the 'A' HPSl pump to the auxiliary spray line, and start the 'A' HPSl Pump.

Ensure 'A' HPSl pump is running, align the 'A' LPSl pump to the SDC suction line, and start "A" LPSl Pump.

C Make-up to the BASTS,restart all of the available charging pumps, and inject into the 'A' HPSl header.

D Align the 'A' LPSl pump to inject to the RCS through the auxiliary spray line, maintain HPSl injection as-is.

Justification 1 A - CORRECT; Boron precipitation is expected. With Facility 2 NOT available, a Facility 1 HPSl pump is utilized to establish flow through the PZR spray noule (via Aux. spray), through the Pressurizer into the #2 hot leg, and back through the core such that water in the core is flushed out the cold leg break. Flow via this path is required to prevent the boric acid concentration in the fuel region from reaching the level at which crystallization would occur. (EOP 2541, App. 18-B)

B - WRONG; If ONLY Facility 1 is available, then the "A" LPSl Purnp cannot be used to inject to the SDC suction line.

C -WRONG; This would be the actions if additional boron injection were required and the normal path was not available.

D -WRONG; With ONLY one facility available, the "A" LPSl Pump cannot he used to inject into the RCS through the auxiliary spray line.

References (

EOP-2532, St. 56, Hot Leg Injection.

EOP-2541, App. 18B, Hot Leg Injection wlonly Fac. 1 Available Comments and Question Modification History

[Remove CET temp. from stem and reword dist "B"] (fixed 12/01/08)

Modified distractors 'B' and 'D' based on validation comments. 1/16/09 Changed justification to include, "with Facil;ity 2 NOT available." Choice A is correct. In stem, "ONLY Facility 1 is available." 2/9/09 NRC KIA System/E/A System 01 1 Large Break LOCA Number EA2.11 RO 3.9 SRO 4.3 CFR Link (CFR 43.5 145.13)

Ability to determine or interpret the following as they apply to a Large Break LOCA: Conditions for throttling or stopping HPI Page 202 of 268 Printed o n 211912009 at 1 6 5 1

v t e Letter Sequence Other
TAC:U01634, (Withdrawn, Closed)
Initiation Acceptance... Administration Questions Answers Results Withdrawal Other: ML090700775, ML090700829, ML091070550, ML091120006, ML091120062, ML091120075, ML091270250, ML091330638, ML091350479
MONTHYEAR2009-02-02 [Table View]

ML091070550

Text

Enclosures:

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