ML051400118

Final - RO & SRO Written (Folder 3)

ML051400118
Person / Time
Site:	Millstone
Issue date:	03/17/2005
From:	D'Antonio J Operations Branch I
To:	Wilson M Dominion Nuclear Connecticut
Conte R
References
Download: ML051400118 (106)

Text

1. I I 9 RO IM SRO Question ID: 5000001 Origin: New Memory Level The plant tripped due to the loss of the 'A' Feed Pump. The Standard Post Trip Actions of EOP 2525 have been completed. The Shift Technical Advisor has just satisfactorily completed the first Safety Function Status Check for EOP 2526 when the plant experienced another transient.

Based on the following information, identify the procedure that must be implemented upon exit from EOP 2526.

- Pressurizer pressure 2010 psia, decreasing slowiy

- Pressurizer Level 40%, decreasing slowly

- #I SG Pressure 880 psia, decreasing slowly

- #2 SG Pressure 900 psia, steady

- SG levels 50% in both SG

- RCS Subcooling 97 degrees F, increasing slowly

- Rad monitors indications normal, steady and not in alarm

- ADVS closed

- Tcold 531 degrees F, slowly decreasing

- Thot 533 degrees F, slowly decreasing

- RCPS operating

- Cntmt pressure 0.15 psig, steady A OP-2207, Plant Cooldown B EOP-2532, Loss of Coolant Accident C EOP-2536, Excess Steam Demand Event D OP-2272C, Plant Operation in MODE 3 prior to Reactor Startup Justification CHOICE (A) - NO WRONG: Indications reveal an uncontrolled cooldown in progress. Exiting the EOPs during an uncontrolled cooldown is not an appropriate action. "EOP 2526 Reactor Trip Recovery Technical Guide", states that EOPs are exited and operating procedures are entered when the desired condition of the plant is determined, and a plant procedure exists to establish and maintain the plant in the.desired conditm. While a cooldown is desired, the specific guidance for conducting the cooldown during an excess steam demand event is contained in the optimal recovery guideline.

VALID DISTRACTOR: an applicant may think use of EOP-2536 is preclude3 unless the threshold of steam generator pressure less than 800 psia is met in EOP-2541, Aopendiv ' , "Diagnostic Flowchart".

CHOICE (B)- NO WRONG: Indications are not consistent with a loss of coolant accident. Pressurizer pressure and level are decreasing.

However, containment pressure is steady and radiation monitor indications are normal. SG pressure and RCS temperatures are decreasing, indicative of excessike heat removal VALID DISTRACTOR: an applicant could misinterpret the cause of lowering pressurizer pressure and level as a loss of coolant accident.

CHOICE (C) - YES EOP-2541, Appendix 1, "Diagnostic Flowchart". dire& use of the "appropiate optimal recovery guideline" following a single event diagnosis. The existence of an inaccessible steam leak /s yivaii in the question stem. The appropriate optimal guideline for a steam leak is EOP-2536, " E x c c s Steam Deinann F.ent".

CHOICE (D) - NO WRONG: Indications reveal an uncontrolled coolduwn in progress. Exiting the EOPs during an uncontrolled cooldown IS not an appropriate action. "EOP 2526 Reactor Trip Recovery Technical Guide", states that EOPs are exited and operating procedures are entered when the desired csndi!ion of the plar;t is determined, and a plant procedure exists to establish and maintain the plant in the desired conditicn Current plmt conditions show that the plant cannot be maintained in Hot Standby because an uncontrolled cooldown is in progr6;s. The cooldown rate will increase as the decay heat generation rate drops.

VALID DISTRACTOR: an applicant may decide tin,-t, givrl: AI key ps:Xlfiat ?Estill within control bands specified in the EOP. that no further action is required other than tc, rcisintain the p l ; ~ ?in a shutdown condition References 1

J EOP 2526 Reactor Trip Recovery Technical C-di.,: Revisiu,i 1 5(P? 5 of 26) 2 EOP-2541, Appendix 1. "Diagnostic Flowchart", &vision 000 !:9!2,',.3t (Pg 1 of 1)

NRC KIA SysternlElA

System E02 Reactor Trip Recovery Number EK1 2 RO 3 0 SRO 3 4 CFR Link (CFR 41 8 I 4 1 10 45 3)

Knowledge of the operational implications of the following concepts as they apply to the (Reactor Trip Recovery)

Normal abnormal and emergency operating proceoures associated with (Reactor Trip Recovery)

NRC KIA Generic System Number RO SRO CFR Link

1. 2 I v'RO v SRO Question ID: 0054227 Origin: Bank Memory Level The reactor has just tripped from 100% power. As PPC) you are carrying out EOP-2525, Standard Post Trip Actions. You note the following conditions:

- Pressurizer pressure indicates 1800 psia and lowering

- Acoustic monitor indications are zero and steady

- Tavg indicates 535F and steady

- No rad monitor alarms are present

- Containment pressure is 0 psig and steady

- Pressurizer level is 33% and rising What is the appropriate action to take?

A Isolate letdown.

B Ensure MSI is actuated.

C Stop RCPs as necessary. d D Close PORV Block valves.

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Justification I CHOICE (A) - NO WRONG: EOP-2525. "Standard Post Trip Actions", (step 3 contingency) directs the operator to manually operate charging and letdown if pressurizer level is not being restored to the 3570% band. Level is at 33% and rising.

VALID DISTRACTOR: an applicant could decide to isolate letdown after incorrectly diagnosing the event as a pressurizer steam space leak.

CHOICE (6) - NO WRONG Indications are not consistent with an excess steam demand event. Pressurizer pressure is decreasing.

However. average coolant temperature and containment pressure are steady and radiation monitor indications are normal VALID DISTRACTOR: an applicant could misinterprEt the cause of lowering pressurizer pressure as an excess steam demand event.

CHOICE (C) - YES With the given indications, there is no LOCA (no containment radiation monitor indications and no containment pressure rise). no ESD (steady Tavg), and no SGTR (no secondary radiation monitor indications). Therefore the primary pressure decrease must be due to a stuck open spray valve. EOP-2525, "Standard Post Trip Actions", directs stopping RCPs as necessary if RCS pressure cannot be restored and maintained between 2225 and 2300 psia and any spray valve will not close "EOP-2525 Standard Post Trip Actions Technical Guide" states that differential pressure created by the RCPs provides the motive force for the pressurizer sprays and that securing the RCP will reduce the spray flow and the lowering of RCS pressure. Pressure control system lesson material (PLC-01-C) describes a stuck open spray valve event at MS2 on April 29, 1980. During this event operators stopped the 'A RCP (Loop 1A) and then the '6' RCP (Loop 1B). The uncontrolled depressurization was terminated afler the 2nd reactor coolant pump was stopped CHOICE (D)- NO WRONG. EOP 2525, "Standard Post Trip Actions". directs closing the PORV block valves if a PORV is open and pressure is less than 2250 psia. An open PORV el 1800 psia would be revealed by acoustic monitor indications.

However, acoustic monitor indications are given as zero and steady.

VALID DISTRACTOR: an applicant may incorrectly attribute the event to an open PORV References 1 EOP-2525. "Standard Post Trip Actions". Revision 20 (2122101) (Pg 8 9, 10 of 26) 2 EOP-2525 Standard Post Trip Actions Technical Guide, Revision 20 (Fg 11 of 38)

3. PLC-01-C, "Pressurizer Level 8 Pressure Control System" Lesson. Revision 3 (Pg 46, 47 of 61)

NRC KIA SysternlEIA System 008 Pressurizer (PZR) Vapor Spacz Accident (Relief Valve Stuck Open)

Number AA2.02 RO 3.9 SRO 4.1 CFR Link (CFR: 43.5/ 45 13)

Ability to determine and interpret the following as thcv apply to the Fiessurizer Vapor Space Accident PZR spray valve position indicators and acoustic monitors NRC KIA Generic System Number RO SRO CFR Link

1. 3 I r/ RO v SRO Question IP, 5000002 Origin: Bank r/ Memory Level A principle difference between a Large Break LOCA (LBLOCA) and a Small Break LOCA (SBLOCA) is:

A Only the LBLOCA clears the RCP loop seal B Only the LBLOCA requires heat removal from containment spray.

C Only the SBLOCA requires heat removal from the steam generators.

D Only the SBLOCA results in peak clad temperatures > 1500°F.

Justification I CHOICE (A) - NO WRONG. Plant-specific accident analysis shows both break categories can result in clearing the loop seal VALID DISTRACTOR: an applicant could reasonably assume that the smaller break, with a lower rate of mass loss, will not allow the loop seal to clear.

CHOICE (B) - NO WRONG. ECCS lesson material states that a failure ot the CS system during a LOCA or MSLB could challenge the containment building integrity.

VALID DISTRACTOR: an applicant could reasonably assume that heat removal via containment spray is only required for a large break LOCA CHOICE (C) - YES Plant-specific accident analysis shows both break categories can result in core uncovery. Further. "EOP 2532 Loss of Coolant Accident Technical Guide" (p1 of 18) states that for small breaks, heat removal via the flow out the break is not sufficient to provide cooling until at least the point where break uncovery occurs and, therefore. steam generator heat removal is required CHOICE (D) - NO WRONG Plant-specific accident analysis shows both break categories can result in peak cladding temperatures (PCT) in excess of 1500 degrees F The limiting LBLOCA results in PCT of 1814F and the limiting SBLOCA results in PCT of 20131F VALID DISTRACTOR: an applicant could reasonably assume that the larger break, with a very rapid blowdown and subsequent rapid reflood will not provide sufficient time with core uncovered to allow PCT to exceed 15OOF References 1 EOP-2532 Loss of Coolant Accident Technical Guide. Revision 2 I (Pg 1 of 18) 2 Millstone Unit 2 UFSAR Section 14.6. "Decreases in Reactor Coolant anventory", 2003 Revision (Pgs 14 6-14, -20. -

211

3 Source INPO Bank - Q# 22448 - Used at Diablo Canyon 1 , 1Oi112002 NRC K I A SysternlElA System 009 Small Break LOCA Number EK2 03 RO 3 0 SRO 3 3' CFR Link (CFR 41 7 / 45 7 )

Knowledge of the interrelations between the small nreak LOCA and the following SiGs NRC K I A Generic System Number RO SRO CFR Link

1. 4 I v RO *I SRO Question ID: 500r3003 Origin: New d Memory Level The reactor automatically tripped from full power. The :JS has just entered EOP-2525, "Standard Post Trip Actions". NO operator actions have been taker?.

Using the attached copy of the SPDS display, identify the event that has occurred A Feed Line Break B Loss of Coolant Accident C Steam Generator Tube Rupture on #2 SG D Excess Steam Demand inside Containment Justification I CHOICE (A) - NO WRONG Event is LBLOCA VALID DISTRACTOR Containment pressure is elevated CHOICE (B) - Y E S SPDS display copied off of simulator 1 minute after initiating fail open of #2 FRV followed by LBLOCA concurrent with trip from full power CHOICE (C) - NO WRONG Event is LBLOCA VALID DISTRACTOR SG #2 level higher than SG # 1 level CHOICE (D) - NO WRONG Event is LBLOCA VALID DISTRACTOR Thot, Tcold much lower than liormal post trip References 1

1 PPC-00 C "Plant Process Computer System Lesson Revision 1 I 12/22/03) (Pg 16 17 30 of 30) 2 PROVIDE APPLICANTS with copy of SPDS display for 3 inmute after initiating fail open of 82 FRV at full power followed by LBLOCA concurrent with the reactor trip N R C K/A SystemlEIA System 01 1 Large Break LOCA Number EA1 17 RO 3 5' SRO 4 1' CFR Link (CFR 41 7 I 4 5 5 / 45 6)

Ability to operate and monitor the following as they apply to a Large Breah LOCA, Safety parameter display system N R C KIA Generic System Number RO SRO CFR Link

1. S I v RO v SRO Questic:, ID: 500C004 Origin: New Memory Level Given:

Plant has been operating at 55% power with normal parameters for the last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Maintenance is being performed on the "A" Main Feedwater Pump. Mu!tiplh seal pressure alarms actuate. The following RCP seal pressure indications or. C-04R are observed and do NOT appear to be changing over time:

- vapor seal pressure:

RCP 1A = 70 psig and fluctuating +- 5 psic RCP 1B = 60 psig and fluctuating +- 5 psig RCP 2A = 70 psig and fluctuating +- 5 psi9 RCP 2 8 = 55 psig and fluctuating +- 5 psig

- upper seal pressure:

RCP 1A = 750 psig and fluctuating +- 30 psig RCP 1B = 1600 psig and fluctuating +- 50 psig RCP 2A = 1260 psig and fluctuating +- 40 psig RCP 28 = 780 psig and fluctuating +- 40 psig

- middle seal pressure:

RCP 1A = 1460 psig and fluctuating +- 30 psig RCP 1B = 2070 psig and fluctuating +- 80 psig RCP 2A = 1440 psig and fluctuating +- 120 psig RCP 2 8 = 1510 psig and fluctuating +- 30 psig Identify the correct diagnosis and proper response from the choices below.

A "8" RCP lower and middle seals have fail& andlor degraded. Trip reactor, then stop "B" RCP.

B "B" RCP lower and middle seals have failed and/or degraded. Start controlled plant shutdown v C "C" RCP lower and middle seals have failed and/or degraded. Trip reactor, then stop "C" RCP.

D "C" RCP lower and middle seals have failed and/or degraded. Start controlled plant shutdown Justification

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I Indications given result in the following pump seal diffsrentlal pressures "A" Pump (RCP 1A) lower - 790 niiddle - 710 upper - 680 "B" Pump (RCP 1B) lower - 180 middle - 470 upper - 1540 "C" Pump (RCP 2A) lower - 810 iniddle - 180 upper - 1190 "D"Pump (RCP 28) lower - 740 middle - 730 upper - 725 Seal is considered failed if differential pressure is less than 200 psi,: and RCS between 2200 and 2300 psia Seals are designed to operate with d/p less than 1500 psid indefinitely Witti :: sea! failed if either of the two intact seals starts to pump (pressure oscillations greater than +- 300 psid), start a conticlled piaiit shutdown If one seal stage IS failed anti d:p across either of the two intact seal stages is chmging at a slow rate (Izss than 10 psid every hour) then If any remaining seal stage dlp lowers to between 500 and 550 psid or rises to yeater than 1500 psid, start a controlled shutdown. With change at a faster rate, d/p setpotn! is hi$.er

CHOICE (A) - NO WRONG Trip action not required unless impending tdilure of all three sells VALID DISTRACTOR an applicant may determine 7 RCP Iowpr and mlddle seals failed and think correct action is to trip reactor CHOICE (6)-YES Lower seal has failed (<200 psid) Middle seal meets degradation criteria ( 4 0 0 psid) Procedure directs controlled 4iutdown CHOICE (C) - NO

\hlRONG Lower seal does not meet failure or degradation criteria Lower seal dlp is 810 psid Procedure directs continued operation VALID DISTRACTOR an applicant may determine C, RCP lower and middle seals failed and think correct action is to trip reactor CHOICE (D) - NO WRONG Lower seal does not meet failure or degrsdatior7 criteria I ower seal dlp is 810 psid Procedure directs continued operation VALID DISTRACTOR an applicant may determine C RCP lower arid middle seals failed References 1 OP-2301C, "Reactor Coolant Pump Operation", Revision 17 (1 1,6103) Section 4 13 "RCP Seal Failure Determination (Pg 34-38 of 45)

NRC KIA SystemlElA System 015/0 Reactor Coolant Pump (RCP, Malfunctions 17 Number AK2 10 RO 2 8' SRO 2 8 CFR Link (CFR 41 7 / 45 7)

Knowledge of the interrelations between the Reactor Coolant Pump Malfunctions (Loss of RC Flow) and the following RCP indicators and controls NRC KIA Generic System Number RO SRO CFR Link

1. 6 I 4 RO d SRO Question ID. 5000005 Origin: New Memory Level The plant is stable at 80% power with thi following conditiais:

- Letdown Flow Controller, HIC-110, is in MANUAL

- Charging and letdown flow are balanced Then, an RCS leak occurs, causing pressurizer level to lnwer at a rate of 2% every 10 minutes (2% I 10 rnin.).

The US instructs the PPO to stabilizes piessurizer level by adjusting the output of Letdown Flow Control HIC-1 IO.

Final conditions; HIC-110 has been adjusten, pressuirar level is now stable and there is NO makeup to the VCT.

Which one of the following describes the direction that the PPO adjusted the output of HIC-I 10 to stabilize pressurizer level, and at what rate will VCT level now lower.

A Lowered the output, VCT now dropping at 4% every 10 minutes. 4 B Lowered the output, VCT now dropping at 1% every 10 minutes.

C Raised the output, VCT now dropping at 6% every 10 minutes.

D Raised the output, VCT now dropping at 0.75% every 10 minutes.

Justification I Prcssurizer volume per % indicated level => 66 44 gals/% (at 2250 psid)

VCT volume per % indicated level => 34 gals/%

CHOICE (A) - YES Controller output must be lowered to reduce letdowri flowrate Rate o! VCT level decrease will be 1 954 (or approximately two) times the prior rate of pressurizer level decrease CHOICE (B) NO WRONG the actual rate of level decrease should be twice that of the pressurizer VALID DISTRACTOR applicant may think the rate of VCT level decrease will be 1/2 that of the pressurizer CHOICE (CI - NO WRONG the controller output must be lowered to reduce letdown flow and the VCT rate given is 3 times that of the pressurizer VALID DISTRACTOR applicant may think controller output must b e r-ised CHOICE (D) - NO WRONG the controller output must be lowered to reduce letdown flow and the VCT rate given is 1/3 times that of the pressurizer VALID DISTRACTOR applicant may think controller ocltput mus: t r raised References I 1 CVC-00-C "Chemical and Volume Control System" Lesson Revisior) 5/3(Pg 24 of 165) 2 OP-2304C "Make Up (Boration and Dilution) Poition of CVCS" rievision 21/9 (6/4/04) Section 4 6 Batch Makeup lo VCT" (Pg 23 of 78) 3 SP-2602A "Reactor Coolant Leakage". Revision 517 (8,311041 ACacbment 1, "RCS Pressure vs Pressurizer Volume" (Pg 16 of 20)

NRC KIA SystemIEIA System 022 Number RO SRO CFR Link NRC KIA Generic System 22 Equipment Control Number 222 RO 4 0 SRO 3 5 CFK Link (CFR 45 2)

Ability to manipulate the console controls as requirzr: to operd:e tht 'acilitd between shutdown and designated power levels

1. 7 d RO fl SRO Question ID: 5000006 Origin: New Memory Level The plant has begun a refueling outage an, IS cutrently in MODE 5. RCS is in reduced inventory and draining is in progress, with the following clinditions.

- 'A' LPSl Pump => runiiirig

- SDC Total Flow (F306) => 1601) gpm

- RCS to SDC Temp (T351X) => 97 degrees F

- Time after shutdown => 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br />

- No. 2 Hot Leg NR Lvl (L-122) => + 0.5 ilicht;s As the draining continues the following indications are observed:

- LPSl PUMP A SUCTION PRESSURE LO annunciator lit (C-01, A-8)

- Oscillating ' A LPSl Pump current Identify which of the following accounts for these indications.

A 'A' LPSl Pump cavitation due to vortexing in Loop 2 Hot Leg B A LPSl Pump is operating at shutoff head due to vortexing in Loop 2 Hot Leg C A' LPSl Pump cavitation due to fully open SI-306, SDC SYS TOTAL FLOW VALVE D 'A' LPSl Pump operating at shutoff head due to fully c!osed SI-306, SDC SYS TOTAL FLOW VALVE Justification CHOICE (A) - YES Vortexing causes air entrainment and leads to pump cavitatioii CHOICE ( 6 )- NO WRONG: Pump operating at shutoff head would indicate steady current and static head at pump suction VALID DISTRACTOR: an applicant may assume that vortexing will cause pump to operate at shutoff head conditions CHOICE (C) - NO

',VRONG: SI-306 failing closed would tend to reduc?:flow rate thereby reoucing chance of cavitation 51-306 failing open would have no effect on flowrate because the LPSl Loop Injection \'aIves were previously throttled to limit total flow to less than or equal to 1600 gpm. See Procedure OP-2310. (Pg 19-22 of 109)

VALID DISTRACTOR: an applicant may think that SI-306 failing open in this situation would result in a high flow condition.

CHOICE (D) - NO WRONG: In MODE 5, at 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> after shutdown, decay heat load requires flow through SDC HX via SI-657 System flow may be reduced by failure closed of SI-306. tiowever. flow would continue through the HX ensuring pump does not r u n at shutoff head.

VALID DISTRACTOR: an applicant may assume tna! :'losure uf Si-306 w 4 force pump to run at shutoff head conditions References 1 SDC-00-C. "Shutdown Cooling System" Lesson. Revision 314 (P3 47 of 79) 2 OP-231OC. "Shutdown Cooling System", Revisio-i 22 (918104). Secticn 4.4, "Reducing SDC Flow in Preparation for Reduced Inventor)"' (Pg 19-22 of 109) 3 OP-2301E, "Draining the RCS (IPTE), Revision 22 (11/4/03) (Pg 7 of 66:

4 AOP-2572. "Loss of Shutdown Cooling, Revision 9 i10/9103) (Pg 4 of 67)

NRC KIA SystemIEIA System 025 Loss of Residual Heat Removal C;ystein (RHRS)

Number AA207 R034 SRO 3 7 CFRLink (CFR 4 3 5 1 4 5 1 3 )

Ability to determine and interpret the following as they apply to tne Loss of Residual Heat Removal System Pump cavitation NRC KIA Generic System Number RO SRO CFR Link

1. 8 I d R 0 *I SRO Question ID: 5000007 Origin: Bank Memory Level The following conditions exist on Unit 2:

o The reactor is shutdown o Both trains of SDC in service o RCS temperature is 280°F o RCS pressure is 160 psia o RBCCW surge tank level is decreasing What leak location will produce these indications?

A Letdown Heat Exchanger B Thermal Barrier Heat Exchanger C Shutdown Cooling Heat Exchanger D Reactor Building Component Cooling Water Heat Exchanger

~

Justification CHOICE (A) - NO WRONG: Letdown pressure on letdown heat exchanger is g:eattlc than RBCCW pressure (160 psia vs 95 psig)

VALID DISTRACTOR. applicant may not understanu that letaown gressure on letdown heat exchanger RBCCW pressure CHOICE (B) - NO WRONG RCS pressure on thermal barrier heat exchangers is greater than RBCCW pressure on the same heat exchangers (160 psia vs 95 psig).

VALID DISTRACTOR: applicant may not understand that RCS pressure on thermal barrier > RBCCW pressure.

CHOICE (C) - NO WRONG: With both trains of shutdown cooling iii s-rvice. the SDC systeni pressure in the SDC HXs (-165 psig) exceeds that of RBCCW (-95 psig).

VALID DISTRACTOR. applicant may not understand that SUC pressure > RECCW pressure.

CHOICE (D) - YES Service water pressure on RBCCW heat exchanger is less than RBCCW pressure (-45 psig vs 95 psig) Maximum SW piimp delta-P of 65 psid in surveillance procedure ciata sheet (SP-2612A)

References 1 RBC-00-C. "Reactor Building Closed Cooling Water System" Lesson, Revision 6 (Pg 44, 45 of 73) 2 SP-2612A-003, "Surveillance Form". Revision 1 (416!04) (Pg ? of 7) 3 Source: INPO Bank - Q# 3392 - Used at Braidwuvrl 1, 91'1411998 NRC KIA SystemlElA System 026 Loss of Component Cooling Watzr (CCW)

Number AA1 05 RO 3.1 SRO 3 1 PFR Link (CFR 41 7 1 45.5 145.6)

Ability to operate and / or monitor the following as tnoy apply to the Loss of Component Cooling Water: The CCWS surge tank, including level control and level alarms, and radiation alarm NRC KIA Generic System Number RO SRO CFR Link

1. 9 I 4 RO 4 SRO Question ID: 0053364 Origin: Bank Memory Level Unit 2 is operating at 100% power, steady state. Pressurizer Pressure Transmitter PT-100X is selected for control and PT-1OOY is out-of-service for repairs.

Pressure on the selected transmitter. PT-1CnX, heyins risiig as indicated on Pressurizer Pressure Controller PIC-I OOX and the Plant Proces? Computer Poth pressurizer spray valves begin to open Pressure is decreasing on all Pressurizer Fressure Safety Channels Which one of the following actions, taken by themselves, would maintain pressure at approximately 2250 psia?

A Turn the pressurizer backup heater control switches tf) ON.

B Turn the pressurizer proportional heater control switches to ON.

C Place PIC-IOOX in MANUAL and lower its output as necessary. 4 D Place PIC-IOOX in MANUAL and raise its output as necessary J i i s t if i cati on I

CHOICE (A) - NO

'LVRONG Turning all backup heaters on manually will rlot maintain pressurp as the normal amount of spray valve flow

,ivailahle can override the output of all pressurizer hePters combined VALID DISTRACTOR applicant may assume that the additional heaters will stop the pressure reduction CHOICE (B) - NO WRONG Turning proportional heaters on manually i/i11 not maintairi pressure as the normal amount of spray valve flow ivnilable can override the output of all pressurizer heaters rombined Flso the proportional heaters cannot be placed in wrvice by simply closing their breaker hand switches Proportional heaters operate off of the pressure controller output siqnal

'IALID DISTRACTOR applicant may assume that thc ?oditiunal healers will stop the pressure reduction CHOICE (C) - YES With the pressure transmitter failedlfailing high the -ontroller must be set to manual and the output lowered far enough to cause the spray valves to close and the output of the proFortiona1 heaters to rise thereby restoring prcssure to normal CHOICE iD) NO

\VRONG Raising the controller output will open the spray valves more and cause pressure to drop faster

\. ALID DISTRACTOR applicant may think that raising controller output will increase heater output and reduce spray flow References 7 PLC-01-C "Pressurizer Level and Pressure Control System" ILesson Revision 3 (Pg 22 of 61)

NRC KIA SystemIEIA System 027 Number RO SRO CFR Link NRC KIA Generic System 2 1 Conduct of Operations Number 2 17 RO 3 7 SRO 4 1 CFR Link (CFR 43 5 / 45 12 / 45 131 4bility to evaluate plant performance and make cqerational Itidgin; nt nclsed on operating characteristics reactor tiehavior and instrument interpretation "

1. 10 1 d RO 4 SRO Question IU: 100U045 Origin: Bank 4 Memory Level A SGTR has occurred in #I SG concurrei! with a loss of off-site power. Initial cooldown on both RCS loops has been completed and #I SG has teen completely isolated.

What parameter and value would indicate that the RCS cooldown was too aggressive and that the loops had become uncoupled?

A #I loop Tc greater than or equal to 5" F lower than $2 loop Tc.

B #I loop Th greater than or equal to 10' F higher than #2 loop Th. 4 C #I loop delta-P greater than or equal to 5 psi lower than #2 loop delta-P D #I SG pressure greater than or equal !o 20 psi higher than #2 SG pressure Justification A isolated SG pressure remains elevated as part n! success strategy to minimize pri-to-sec leakage: C: natural circ rlclta-P is -1!2 # or less in loop #2, can't get 5# les-, E.once 21 SG is completely isolated there is no way for its Tc to he lower CHOICE ( A ) - NO

'LVRONG Once # 1 SG is completely isolated, #I loop Tc will iemain higher VALID DISTRACTOR. applicant may think that the ciifference in loop temperatures is indicative of uncoupling CHOICE (E) - YES Uiicoupling of the two loops is indicated by failure ot Th in the loop with tne isolated steam generator to track Th in the operating loop Hot leg temperatures differing by more than 10°F IS an indication that the isolated steam generator is limiting RCS cooldown and depressurization (2na note in 1 st note block. EOP-2534. Pg 24 of 64)

CHOICE (C) - NO

'>IYRONG. Natural circ delta-P is -112 # or less in loop #2,can't get 5# less VALID DISTRACTOR applicant may assume that the additional heaters will stop the pressure reduction CHOICE (D) - NO WRONG, Isolated SG pressure remains elevated as part ot success strategy to minimize pri-to-sec leakage VALID DISTRACTOR: applicant may assume that the differences in SG pressure indicative of uncoupling References _I 1 EOP-2534. "Steam Generator Tube Rupture", Revision 22 !3/22;02) (Pg 24 of 64)

NRC KIA SysternlElA System 038 Steam Generator Tube Rupture (SGTR)

Number EK1 03 RO 3 9 SRO 4 2 CFR Link (CFR 41 8 / 41 10 / 45 3)

Knowledge of the operational implications of the following Loncepts as they apply to the SGTR Natural circulation NRC KIA Generic System Number RO SRO CFR Link

1. 1. 11 I v RO v SRO Quer?ion !D: 307?64E Origin: Bank Memory Level The plant was operating normally at 100% iiower. when the following events occurred:

- Pressurizer Pressure, Level, and Reactor Coolant (RCS) Cold Leg Temperature (Tc) start dropping rapidly

- Reactor trips

- Main Steam Isolation (MSI) and Safety Injection Actuation Signal (SIAS) occur

- Reactor Coolant Pumps (RCP's) are secured

- Loop 2 Tc and Steam Generator (SIG)pressure are decreasing much faster than Loop 1 Tc and SIG pressure.

- Auxiliary Feedwater Actuation Signal (AFAS) has NOT actuated

- Containment pressure and temperature are increasing Which of the following actions must be taken on Panel C-05 in accordance with EOP-2536, "Excess Steam Demand Event" to mitigate this event?

A Place #2 SIG Auxiliary Feedwater Isolation Air Assisted Check Valve Switch to CLOSE.

B Shift # I and #2 Auxiliary Feedwater Regulating Valve Controllers to MANUAL and CLOSED.

C Place #I and #2 SG Auto Permissive OVERRlDElMANISTART RESET Switches to PULL- v TO-LOCK.

D Shift #2 SIG Auxiliary Feedwater Regulating Valve RESET/NORM/OVRD Switch momentarily to OVRD.

Justification I

CHOICE (A) - NO WRONG The air assisted check valves are designed to puvide ro,itainment isolation in the event of an accident inside rontainment These valves are are 6 inch swing cheLks that will preLent a reversal of flow Normal AFW flow will open the valves VALID DISTRACTOR an EOP Step (EOP-2536 .tep 9 L Pq 12 of 6.2) directs closing this valve in the event of a steam line break Applicant may think that closing tiits valve will ,,revent AFW from reaching the SG CHOICE (E) - NO WRONG An auto actuation signal will open the AFW feed regulating valves even in the manual loading stations are in AIANUAL and CLOSED VALID DISTRACTOR applicant may assume that tke valve will not automdtically open when in MANUAL CHOICE (C) - Y E S The AFW feed regulating valves will be closed unt4fAFAS is actuated Vacrng these switches in PULL TO LOCK prior to AFAS blocks the automatic initiation signal that ?peris the AFW feed regulating valves (AFW-00-C Pq 19 of 56)

CHOICE (D) - NO WRONG The RESET NORM OVRD switch will not prevent feeding the SG if Auto AFW trips after the RESET NORM OVRD was momentarily (spring return to normal) in OVRD VALID DISTRACTOR applicant may think that oncc overriden the valve hill not react to an auto actuation signal until this same switch is taken to RESET References 1 AFW-00-C "Auxiliarv Feedwater Svstem" Lesson Revision 5 (Po 19, 20 of 56) 2 EOP-2536. "Excess'Steam Demanb Event", Revision 20 (712710'i) (Pg 12 of 62)

NRC KIA SystemlElA System E05 Excessive Heat Transfer Number EA1 1 RO 3 9 SRO 't 2 CFH Link (CFR 41 7 / 45 5 / 45 6)

Ability to operate and / o r monltor the following as they dpply to the (Fxcessive Heat Transfer) Components and functions of control and safety systems including instriirnentation signsls tnterlocks failure modes and automatic and manual features NRC KIA Generic System Number RO SRO CFR Link

1. 12 I g RO g SRO Questinn ID: 500UC108 Origin: New Memory Level The plant has tripped from 100% power due to a loss of all feedwater. Upon completing Standard Post Trip Actions, the crew has noted that MFW. Condensata and AFW are UNAVAILABLE, and has transitioned to the Loss Of All Feedwater EOP Several minutes later, SlAS has been manually iniated and the US has decided to implement the Once-Through-Cooling (OTC) success path.

Which one of the following describes additional actions required for the succesful implementation of OTC, and the possible consequences if those actions are delayed?

A Open both ADVs 100% and open both PORVs. or core uncovery may occur due to g inadequate safety injection flow.

B Initiate an RCS cooldown NOT to exceed 80 "Fihi and Open PORVs, or the Reactor Vessel belt line may exceed design parameters C Open both PORVs and cooldown the RCS to minimize void formation and ensure single-phase NC flow, or core uncovery may result D Open both ADVs to cooldown the RCS at max rate then open PORVs at the 200 "F subcool line to prevent PTS of the RCS and reactor vessel Justification CHOICE (A) - Y E S I

ADVs and PORVs must both be opened to initiate or,ce-through cooling, or the limited PORV flow capacity will result in cventual core uncovery and fuel damage Per TG-254OD. i f the plant trips from power on low steam generator level following a loss of all feed, SG level could reach the ~nce-1t,:ough-c:)ollngaction point as early as 7 to 9 minutes after w e n t initiation The document also states that if oncc-through-cooling is not initiated before SGs are lost as a heat sink core uncovery and possible core damage could result CHOICE (E) - NO WRONG: Cooldown with limits is a possible contingency actions for a S/?'rR.if RCS or SG pressure IS holding up injection flow VALID DISTRACTOR: applicant may confuse this "legitimate contingency" for the required actions here CHOICE (C) - NO WRONG: Cooldown is initiated to maximize injection not to minimize vold formation. In an event where a LOOP has resulted in the loss of Force Flow, single-phase NC is most desirerable. But not in the event given in this question.

VALID DISTRACTOR applicant may confuse a desired condition wit!) -7 realistically achievable condition.

CHOICE (D) - NO WRONG, The concern is from an ESD event and a qossible contingency if the stated conditions cannot be controlled VALID DISTRACTOR. applicant may confuse this "!t:gltimate contingency" for the required actions here References 1 HPI-00-C. "High Pressure Safety Injection System" Lesson. Revision 6. (Pg 12 of 49) 2 CVC-00-C, "Chemical and Volume Control System" 1.~50'1. R e h i o n 8 (Pg 37 of 165) 3 EOP-2540D Functional Recovery of Heat Removal Technical Guide, Revision 18, (Pg 122 of 155)

NRC KIA SystemlElA System E06 Loss of Feedwater Number EA1 2 RO 3 4 SRO 4 0 CFRLink (CFR 41 7 / 4 5 5

• 4 5 6 )

Ability to operate and I or monitor the following as thfy apply to thr 'Loss of Feedwater) Operatinq behavioi characteristics of the facility NRC KIA Generic System Number RO SRO CFR Link

1. 13 I v RO v SRO Question ID: 0053332 Origin: Bank Memory Level The plant tripped from 100% power. A Station Blackou~has been diagnosed, and the appropriate EOP entered.

Which one of the following sets of conditions satisfy the requirements for stable natural circulation, two hours into the event?

A RCS Tcold => 445 degrees F and constant RCS Thot => 422 degrees F and going down CET => 443 degrees F and going down RCS Pressure => 1500 psia B RCS Tcold => 395 degrees F and constznt RCS Thot => 453 degrees F and going down C ET => 442 degrees F and going down RCS Pressure => 1600 psia C RCS Tcold => 458 degrees F and going down RCS Thot => 480 degrees F and going down CET => 481 degrees F and going down RCS Pressure => 700 psia D RCS Tcold => 458 degrees F and constant RCS Thot => 470 degrees F and constant CET => 469 degrees F and constant RCS Pressure => 930 psia Justification I CHOICE ( A ) - NO INRONG Choice (has Thot lower than Tcold Natural r irculatioii will not establish in reverse directicrn Also difference between Thot and CET is 11 degrees which is in excess of 10 degree lirrd

'IALID DISTRACTOR applicant may not recognize that Tcold must be lower than Thot CHOICE (B) - NO WRONG Loop delta-T is 58 degrees, which is in exr-ess of 55 degree criteria Also difference between Thot and CET is 1 1 degrees which is in excess of 10 degree limit VALID DISTRACTOR applicant may not know that maximum loop delta T IS < 55 degrees CHOICE (C) - NO lIS'RONG Subcooling IS 25 degrees which is belov. mlnimum subcooling criteria of 30 degrees

','ALlD DISTRACTOR applicant may not recognize that subcoolina is less than the required minimum CHOICE (D) - YES Tliot is greater than Tcold Loop delta-T is 12 degrees all temperatures 4's decreasing Subcoolinci IC 65 degree5 References 1

J EOP-2528"Loss of Offsite PoweriLoss of Forced Circulation" Revision 15 (2127101). (Pg 8 of 36) 2 EOP-2541 Appendix 2, "Figures". Revision 1 (3131104).Pg (1 of 7) 3 Steam Tables - SUPPLY TO APPLICANT FOR USE DURING EXAhl NRC KIA SystemlElA System 055 Loss of Offsite and Onsite Power (Station Blackou!)

Number EA202 RO 4 4 SRO 4 6 CFR Link (CFR 43 5 / 45 13)

Ability to determine or interpret the following as they apply to a Statlcr Blackout RCS core cooling throuqh natural irrulation cooling to SIG cooling NRC KIA Generic System Number RO SRO CFR Link

1. 14 1 v RO SRO Question ID: 50005W Origin: New Memory Level Power level is at 92% when the SPO reports that the FEEDWATER REGULATING VALVE 2 LOCKED alarm (8-8, C05) is actuated and all focr white indicating lights for #2 SG Feedwater Control are out.

Based on these indications, on direction of the US. the SPO will take the following action(s):

A Press RX TRIP TCBS pushbuttons and close #2 SG FRV Blocking Valve FW-42B to prevent ove rfiII.

B Press ' A and 'B' SGFP MAN pushbuttons and cori?rolpump speed manually to maintain level v in #2 SG.

C Press LIC-5269, #2 SG FRV Controller MAN pushbutton and maintain #2 SG level within the desired operating band.

D Press #2 SG FRV DOWNCOMER RESET pushbutton and control valve in manual to restore level to between 60 and 75%.

Justification I

Indications caused by a loss of Vital Instrument Bus VA-20. This bfrs supplies control power to #2 SG FRV Loss of power will cause normally open solenoid valves to close on the air supply lines to the Main and Bypass FRVs, which will fail in "as-is" position. Each FRV has four normally lit white control status lights. They indicate low instrument air header

~iressure.low control air pressure, high or low controller olitput or loss GI control power. All four lights will extinguish if power if Bus VA-20 is lost.

CHOICE (A) - NO WRONG: The ARP directs the operator to maintair power level constant Given the indications and the slow rate of power decrease. the operator will be able to controi SG level try ;arying SGFP speed.

VALID DISTRACTOR: applicant may assume that thc ioss of control wliile conducting a downpower will require a manual reactor trip. If the reactor is tripped, action to isolate feedwater would be appropriate since the FRV will not close The AOP for loss of the bus (2504D) contains a caution that warns operator the FRV will not close if a reactor trip occurs and states that since FRV fails 'as is'. a SG ledel transient may occur. This caution prepares operator for one possible outcome. but the following procedural guidance makes it apparent that actions are available to control SG level thereby avoiding the need for a reactor trip.

CHOICE (E) - YES ARP directs operator. if necessary, to place both SGFPs in manual ana to control level by pump speed This action is riecessary since without control power, the FRV cannnt be remotely positioned CHOICE (C) - NO WRONG With a loss of control power, the FRV canmt be controlled remotely in auto or in manual VALID DISTRACTOR: the alarm response. written for multiple possible causes for a FRV lock condition. does direct manual control of the FRV. The applicant must corrictly diagnose the cause of the problem based on given indications in order to determine the correct ARP actions CHOICE (D) - NO WRONG' Downcomer reset will not affect FRV control until control power is restored VALID DISTRACTOR: the ARP, written for multiple posslble ca!ises for a FRV lock condition, does direct the operator to press the pushbutton to restore manual control. The applicant must correctly diagnose the cause of the problem tiased on given indications in order to determine the correct ARP actions References i

1 ARP-2590D. Window B-8 (030). "FEEDWATER REGULATING VALVE 2 FAILURE", (2/12/04) 2 AOP 2504D. "Loss of 120 VAC Vital Instrument Panel VA-20" Revision 3 (6/24/04) (Pg 7 10.19 of 23) 3 OP-2385. "Feedwater Control System Operation" Revision 9 (2!7/02) (Pg 7 of 22 and Figure 51

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NRC KIA SystemIElA System 057 Number RO SRO CFR Link NRC KIA Generic System 24 Emergency Procedures /Plan Number 2 4 10 RO30 SRO 3 1 CFRLink (CFR 41 1 0 1 4 3 5 i 4 5 1 3 )

Knowledge of annunciator response procedures

1. 15 I 4 RO *, SRO Question ID: 00552C6 Origin: Bank g Memory Level A loss of 125 VDC Bus 201A causes a plart trip. Buses 25A and 24C fail to transfer to the RSST.

The 'A' D/G will A remain shutdown and CANNOT be started from the zontrol room B receive an emergency start and immediately [rip cn overspeed C come up to speed on the electrical govcmor and automatically load Bus 24C D come up to speed on the mechanical governor and have only limited protective features v available Justification CHOICE (A) - NO WRONG The diesel will start and run on the mechanical qovernor VALID DISTRACTOR Applicant may think diesel will remain shutdown CHOICE (B) - NO WRONG The diesel will start and run on the mechanical governor VALID DISTRACTOR The only available protective feature on a loss of DC control power is mechanical overspeed Applicant may think the overspeed trip will be challenqed h y the loss of DC control power CHOICE (C) - NO WRONG The diesel will start and run on the mechanical govirror and will not automatically load bus VALID DISTRACTOR Diesel is designed to auto start and auto load on a loss of power to Bus 24C Applicant may think diesel will function as designed CHOICE (D) - YES The diesel generator air start solenoid valves fail open on a loss of DC The diesel will start and run on thc mechanical governor with only the overspeed trip available all other trips need DC to operate The diesel output breaker will not close without DC control power, so the diesel can not provide power iu the bus Question requires applicant to understand effects of loss of DC on the EDG and on Bus 24C References 1 AOP 2505A. "Loss of Vital 125 VDC Bus 2 0 1 A Revision 1 12/12/03) (Fg 27 28 of 47) 2 LVD-00-C. "125 VDC/120 VAC" Lesson, Revisior 5 (Pg 52 of 7 7 )

NRC KIA SystemIEIA System 058 Loss of DC Power Number AK3 01 RO 3 4' SRO 3 7 CFR Link (CFR 41 5.41 10 / 45 6 / 45 1)

Knowledge of the reasons for the following respofises as they applv to the LOSS of DC Power Use of dc control power by DlGs NRC KIA Generic System Number RO SRO CFR Link

1. 16 1 RO SRO Question ID: 5000013 Origin: New Memory Level Plant is operating in MODE 3. 'C' Service Water Pump is w t of service for motor maintenance. '6' Service Water Pump is supplying the 'B' Service Water header. Bus 24E is aligned to Bus 24D. '6' RBCCW Heat Exchanger is aligned to provide minimum flow for the 'A' Service Water header. Long Island Sound water temperature is 37°F. EDG SW Bypass Valves SW-231A and SW-231 B are being maintained closed because of an issue with the adequacy of valve actuator spring pressure.

125 VDC Panel DVIO de-energizes due to a fault. You tiow have the following indications:

PPC Points:

F6433 'A' RBCCW HX SW Flow => 8885 gprn F6434 'B' RBCCW HX SW Flow => 795 gprn F6435 'C' RBCCW HX SW Flow => 8965 gpm Local EDG SW Flows:

FIC-6397 'A' EDG => 850 gpm FIC-6389 'B' EDG => 100 gpm Assuming NO operator actions related to Service Water, which of the following is correct concerning these indications?

A ' 6 ' EDG SW flow is LOWER than normal because SW-896, DG TCV is closed B 'A' EDG SW flow is HIGHER than norma! because nf 5 rupture downstream of 'A' EDG SW flow transmitter.

C 'A' RBCCW HX SW flow is LOWER than normal because of diversion of flow to the in-service TBCCW HX.

D 'C' RBCCW HX SW flow is HIGHER than normal because of a rupture downstream of 'C' 4 RBCCW HX SW flow transmitter.

Justification I

Pre-event plant configuration:

Both EDGs shutdown, SW flow = -150 gpm per EDG Fac 1 aligned to ' A RBCCW (400 gpm), 'A' TBCCW (1700 qprn) and 'E' TBCCW (2050 gprn for rntnimiim flow purposes

.inan vlv throttled)

Fac 2 aligned to 'C' RBCCW (400 gpm). '6' RBCCLL (1150 gprn for rnlnimurn flow purposes, man vlv SW-9B throttled).

and 'C' TBCCW (1 700 gpm)

Total SW flow = -3000 to 4000 gpm per header

'0'and 'C' SW pumps both powered from Facility 2 PS (24DU4E)

Expected plant response to the loss of DC Panel DVIO:

- MSlVs close, plant trips

- letdown isolates

~ Facility 1 ' A EDG starts, does not load

- TCBs 1 and 3 open

- Facility 1 RBCCW flow balance is disrupted due to numerous valves failing

- ' A RBCCW SW Outlet TCV, SW-8.IA (TV-6308) fails open

- EDG TCV. SW-89A (HV-6389) fails open EDG Bypass Valve, SW-231A (FY-6341) fails closed Bus 24C loses control power, fails to transfer to RSST. de-wiergizes Discussed system conditions with Dan Pantalone of blS2 on 11/19/04 Typical system flow conditions are expected operator knowledge. Plant SW conditions at power for winter operatlon (32-37°F Sound temp),

- currently operate with EDG Bypass Vlvs closed even when EDG is shutdown, based on issue with valve actuator spring force. when bypass was used, flow was 3000 gprn

- EDG SW flow with engine shutdown = 150 gpm. EDG flow with engine operating 1200 gprn

- RBCCW SW flow for in-service HX = 300 to 500 gpm. based on relatively small heat load at power and cold heat sink

- R0CCW SW flow for standby HX = 1000 to 130C gm, based on mairitatning adequate minimum system flow man vlv throttled

- TBCCW SW flow for in-service HX = 1500 to 200il gkm, bas& on relatively large TB heat load at power

- TBCCW SW flow for standby HX = 2000 gpm. based un rnalntainlr!g sa-quate minimum system flow, nian vlv throttled

- Total SW Header flow = 3000 to 4000 gprn CHOICE (A) - NO WRONG. Loss of DVI 0 causes a reactor trip. However. Factlity 2 equipment should not be challenged by loss of DVIO or the subsequent trip. Valve SW-89B was closed before the event a's,., w i l remain closed throughout the event. Flow may he slightly reduced because of diverted flow to the RBCCW HX !?ak. However, flow given is still approximately iiormal for the EDG in standby due to typical leakdge flow through the \*+Ibe,

VALID DISTRACTOR. Applicant may think flow is rediiced because TC?. is closed.

CHOICE (B) - NO

\?dRONG. The DC panel failure results in 'A' EDG TCV failing open UG Bypass is being maintained closed and fails

~-losr:d Flow will be higher than before the event bscause of open 1CV However, flow given is almost as expected dtiriiig EDG operation only somewhat reduced bemuse of div.:rted flow to open RBCCW HX TCV No leak is indicated

'JALID DISTRACTOR Applicant may think the given flow is I*igh?r thar normal due to a leak CHOICE (C) - NO

\IYRONG Winter mode is selected to rout HX discharge through 6 inch bypass line. However, the loss of DC Panel DV10 causes the outlet valve (TV-6308) in the 14 inch line to fail open Flow will be higher than prior to the event due to tho valve opening

'dALID DISTRACTOR. Applicant may think flow is lower than priu tc t h e event due to increased flow through the TBCCW heat exchanger TBCCW TCVs do not fail on loss of DV-10 CHOICE (D) - Y E S Flow is normally reduced during winter conditions. kll HX flow should be routed through the 6 inch outlet bypass line Indicated flow is significantly higher than would be expected wlth the Sound at 37°F.

References SWS-00-C, "Service Water System Lesson. Revision 5 (Pg 16. 49 of 59) 2 LVD-00-C. "125 VDCI120 V A C Lesson, Revisicn 5 (Pg 52. 53 of 77)

i AOP-2505A. "Loss of Vital 125 VDC Bus 2 0 1 A Revisim 1 (2112'03) (Pg 20, 22 of 47)

NRC KIA SysternlEIA System 062 Loss of Nuclear Service Water Number AA2 01 RO 2 9 SRO ': CFR Link (CFR 43 5 / 45 13)

Ability to determine and interpret the following as the; apply to thc Loss of Nuclear Service Water Location of a leak in the sws NRC KIA Generic System Number RO SRO CFR Link

1. 17 I 4 RO d SRO Question ID: 210COOO Origin: Bank Memory Level A Xenon free, End-Of-Life (EOL) reactor skrtup is ir: progres.

RCS temperature is being maintained on t h t 'A' Stean, Bypass valve. NO dilution is in progress.

Critical data is recorded and CEAs have been manually withdrawn 5 steps to raise reactor power to the POAH.

The SPO notes all 'High Power Trip Resets' on C04 are lit Which of the following actions is required for the given plant condition?

A Start an additional AFW pump to maintain SG levels B Insert CEAs to their 10E-4 position to null out the positive dpm startup rate C Operate Rx Trip pushbuttons due to CEAs stuck in continuous withdrawal. 4 D Depress RPS High Power Trip Resets and open MSlV bypasses to restore Tave to 532°F Justification CHOICE (A) - NO

'WRONG Reactor trip is required for an uncontrolled rod withdrawal Existing AFW alignment will provide sufficient flow post-trip to maintain SG levels VALID DISTRACTOR Applicant may mistakenly think that AFW flow should be increased to match steam demand CHOICE ( B ) - NO

'Jv'RONG Power is abnormally high, indicative of an large unLontrolled i.7ositive reactivity addition A rpactor trip is required VALID DISTRACTOR Applicant may 1nappropriate;y apply operating procedure guidance for actions upon reaching POAH CHOICE (C) - YES Hi-Power resets first liqht at -9% power much too hiqh for a 112 DPM SUR to achieve under the stated conditions All other sources of positlie reactivity are ruled out by s t k d conditions A reactor trip IS required From OP-2202 "If at

,Ir,ytime during this startup the condition of the reactor an? i + i respon4es dre not understood and controlled by the operators thc reactor must immediately be tripped :lid the actions et EOP-2525. Standard Post Trip Actions performed CHOICE (D)- NO

\YRONG Power is abnormally high indicative of an uncontrolled positive reactivity addition A reactor trip is required

'JALID DISTRACTOR Applicant may inappropriately +ply operating procedure guidance for normal power escalation References 1 OP-2202 "Reactor Startup IPTE" Revision 20 (5i5!041 (Pg 2 of 47)

NRC KIA SystemlElA System 001 Continuous Rod Withdrawal Number AAI 05 RO33 SRO 4 2 CFRLink ( C F R 4 1 7 i 4 5 5 1 4 5 6 )

Ability to operate and ior monitor the following as they applv to the Continuous Rod Withdrawal Reactor trip switches NRC KIA Generic System Number RO SRO CFR Link

Memory Level Unit 2 is operating at 100% power when tht. fol1owii:g alaims are received:

CEA MOTION INHIBIT CEA REG G R P S W P CEA REG G R P W P BKUP There are NO changes in indicated positiori to< any CEAs except for Shutdown Bank 'B' CEA #6, which indicates the following:

- red. amber and blue mimic lights out

- green mimic light lit

- PPC shows position at 180 steps

- CEAPDS shows position at 0 steps Identify the position of CEA #6 and the reason for the conflicting indications A CEA #6 is fully inserted. A CEA drop reed switch has failed. J B CEA #6 is fully withdrawn. A CEA drop reed switch has failed.

C CEA #6 is fully inserted. A CEA reed switch position transducer switch has failed.

D CEA #6 is fully withdrawn. A CEA reed switch position transducer switch has failed Justification I CHOICE (A) YES Each of the 61 CEAs has 4 stacks of reed switches One qrotip is a stack of 97 switches connected to the reed switch position tranducer This device generates a signal w n t to t i c CEAPDS Lo display CEA position on a CRT on C 04 A second group consists of a reed switch which actuaics .vhi the CEA IS fully inserted This switch activates the dropped f

CEA annunciator. lights that CEAs core mimic amb3r light and rezeros the PPC indication for that CEA at zero steps

\t third group consists of the switch responsible for the Upper Electrical Limit signal and turns on llic red light when CEA 5 > 180 steps The fourth group consists of the sv tch responsiblc for the Lower Electrical Limit signal and turns on the (lreen liqht when the CEA is < 1 step CEA #6 is fully inserteJ SII ,e the drop reed switch has failed there IS no amber drop light The PPC does not get a signal to reset position indication to 0 CHOICE (E) - NO WRONG CEA #6 is dropped VALID DISTRACTOR May think the CEA is fully witbdrawn based on lack of deviation alarms and PPC indication at 180 steps CHOICE(C) NO WRONG CEA reed switch position transducer hds not failed Position hidication on CEAPDS IS correct VALID DISTRACTOR May think CEA reed switch position transducer Frovides the PPC reset to zero signal CHOICE (D) - NO WRONG CEA is fully inserted VALID DISTRACTOR May think the CEA reed switch position transducer IS responsible for the CEA bottom light References 1 CED-01-C Rev 4 pages 30-31 NRC KIA SystemlElA System 003 Dropped Control Rod Number AA201 RO 3 7 SRO 3 9 CFR Link (CFR 43 5 / 45 13)

Ability to determine and interpret the following as they apply to the Dropped Control Rod Rod position indication to actual rod position NRC KIA Generic System Number RO SRO CFR Link

1. 1. 19 1 fl RO SRO Questic.n ID: 0054389 Origin: Bank d Memory Level The plant is in normal operation at 100% power. when a Flrs System Trouble annunciator is received on C06/7. Panel C-26 is checked and a PEO IS subsequently dispatched to the West DC Switchgear Room.

The PEO reports the following:

- Two Ion Chamber smoke detectors are in alarm

- The Halon strobe lights and horn are pulsating slowly Based on these conditions, what is the status of the West DC Switchgear Room Halon System?

A It is alarming as a warning of a poten!ial discharge if a photoelectric detector is activated. d B It IS presently discharging or completed dlschdrgiflg to the West DC Switchgear Room c It is warning that a discharge to the West DC Switchgear Room will occur after timer countdown.

D It is in an alarmed state, should have already discharged, but a system malfunction has occurred.

Justification Existing bank comments: OP 2341A (Rev 13): Discussion section 2nd ARP 25901. for Zone 45.$SSS(Copiedfrom Item No '4990' on 10/28/97 By RLC]

CHOICE (A) - YES The East and West DC switchgear rooms require tw3 zones (one photr,elP:!ric smoke detector and one ion smoke detector) to initiate a halon release. Activation of one s i i i o k ~dctcctcr, ion or photoelectric. will cause the strobe and horn to pulse slowly CHOICE (B) - NO WRONG: The East and West DC switchgear rooms also reqti're :,vo zcncs (one photoelectric smoke detector and one ion smoke detector) to initiate a halon release Activatiqa of one smoke detector, ion or photoelectric will came the strobe and horn to pulse slowly.

VALID DISTRACTOR: an applicant may think that tht pulsating horn announces a discharge in progress.

CHOICE (C) - NO WRONG: Activation of a second smoke detector of the opposite type. but in the same room, will cause the strobe and horns for the affected room to pulse QUICKLY The flashing lights will operate, and a 60 second pre-discharge time delay will begin Upon expiration of the time delay the dalon System will aischarge and the strobe and horn will sound steadily

?/ALID DISTRACTOR: an applicant may think that me SLOWLY pulsating horn and strobe light warn of a timer countdown to discharge halon.

CHOICE (D) - NO WRONG: Activation of a second smoke detector of the opposite type, but in the same room, will cause the strobe and horns for the affected room to pulse quickly. The flashing lights will operate, and a 60 second pre-discharge time delay will begin Upon expiration of the time delay the Halon System will discharye and the strobe and horn wIl sound steadily.

VALID DISTRACTOR: an applicant may think that the pulsdtiris h o c and strobe lights indicate that a s f s t e m malfunction has occurred.

References 1 OP 2341A. "Fire Protection System" Revision 15 (2'26/04) (Pg 3 of b I )

2 ARP 25901, "Alarm Response for Fire Panel C-2F' (Zone 45) Revisiol 02 (9/9/04) (Pg 67-70 97 of 106)

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NRC KIA SystemlElA System 067 Plant fire on site Number AA1 09 RO 3 0 SRO 3 3 C F H Link (CFR 41 7 / 45 5 / 45 6 )

Ability to operate and / or monitor the following a - they apply to the Plant Fire on Site Plant fire zone panel (incltiding detector location)

NRC KIA Generic System Number RO SRO CFR Link

1. 20 I d RO d SRO Quest,on ID. 5000012 Origin: New d Memory Level The plant IS operating at 55% power for maintenance on s MFW pump.

A fire breaks out in Control Room Panel C-01. Smoke is filling the control room. The Shift Manager orders a control room evacuation.

What action is required to be performed from the control room prior to evacuation if time permits?

A Trip the reactor at C-04. d B Open gravity feed valves at C-02 c Transfer in-house buses to the RSST at C-08 D Place both heater drain pumps in PULL-TO-LOCK at C-05.

Justification I

CHOICE (A) - YES Fire Procedure AOP-2579A directs a reactor trip from C-04 (Pg ti of 64)

CHOICE (B) - NO WRONG. because AOP-2579A does not direct this action from the control room The procedure directs lining up gravity feed from the BASTS locally (Step 21. Pg 15 of 64)

VALID DISTRACTOR: because the procedure directs opening these valves locally The applicant may think this is an action to be performed remotely prior to exiting the control room CHOICE (C) - NO WRONG because AOP-2579A does not direct this action from tlie control room. Per the procedure the operator will contact CONVEX and have them de-energize RSST 15G-22S (Step I S . Og 10 of 64)

VALID DISTRACTOR, because the non-fire shutdown fionl c d s i d r the control room procedure (AOP-2551) directs the transfer of in-house buses to the RSST prior to evacuation (Step 3 2 . Pg 5 of 21).

CHOICE (D1- NO WRONG: because AOP-2579A does not direct this action from the control room.

VALID DISTRACTOR. because the non-fire shutdown from outside the control room procedure (AOP-2551) directs the transfer of in-house buses to the RSST prior to evacuation (Step 3 2 Pg 5 of 21).

References 1 AOP-2579A "Fire Procedure for Hot Standby Appendix A Fire Area R-1", Revision 9 (9/15104) (Pq 6 10 15 of 64) 2 AOP-2551 "Shutdown from Outside the Control Room' Revision 9 (1/16/03) (Pg 5 6 of 21)

NRC KIA SysternlElA System 068 Control Room Evacuation Number AK2 02 RO 3 7 SRO 3 9 CFR Link (CFR 41 7 / 45 7)

Knowledge of the interrelations between the Control Room Evacuation and the following Reactor trip system NRC KIA Generic System Number RO SRO CFR Link

1. 21 I 4 RO SRO Question ID. Sr)iIOO:.I Origin: New Memory Level Unit 2 has been operating at 100% power steady state, for- several months.

Then, a technician working inside of Panel 83-03 accidently causes a momentary short circuit that trips

'A' and 'B' RCPs.

Assume all equipment responds as designed ai13 NO operator actions have yet been performed Which one of the following describes the expected conditions for the parameters listed below, once the 'A' & 'B' RCPs have completely stopped'!

Steam Pressure Steam Flow FRV Bypass Amount OPEN

~ _ _

A SG2 > SGI SG2 > SGI S G 2 open > S G I B SG2 > SGI SG2 > SGI S G 2 open = S G 1 4 C SG2 = SG1 SG2 < SGI S G 2 open < S G I D SG2 = SGI SG) = SGI S G 2 open = S G I Justification I

CHOICE (A) NO iVRONG Feed flow response for both steam generatois is the sari e The response is driven by the turbine trip signal riot SG level VALID DISTRACTOR because the applicant may thirin that feed flow will bz higher to maintain SG level with a higher

,teaming rate CHOICE (B) - YES The reactor protection system will initiate an RCP unwrspeed trip for 2 pumps at <830 rpm and an automatic reactor trip on low reactor coolant flow in Loops 1A and 1 B at a 2c$ flow setpoint Three minutes after the event the post-trip decay tirat is being removed via turbine bypass valves !however only Ldop 2 ill be steaming Flow in Loop 1 will be rwersed with Tcold pumped backward through Loops 1A and 16 The higher rate of heat transfer froin Loop 2 will rt,sult i r i HIGHER STEAM FLOW from SG2 Feed flow post-trip is deterinined by the design response of the feedwater control system The FRVs ramp closed in manual at programmed rate on turbine trip signal Bypass FRVs ramp open automatically to 40% over 3 minutes Althought S G levels may be different, both SGs will have the EQUAL FEED FLOW rate Steam PRESSURE will be HIGHER in SG2 because of the additional energy transfer Pressure in SG 1

\*iill be at saturation pressure for Tcold (900 psia at Tc=532 Fi CHOICE(CI NO 5VRONG SG2 pressure will be higher because of 'he Thot inlet wate' vs Tcold inlet water on SG1 VALID DISTRACTOR because the applicant may think pressures w l l ha equal due to the relatively low rate of heat transfer post trip CHOICE (D) - NO WRONG SG pressure will be higher because of the qreater eneryy t znsfei rate VALID DISTRACTOR because the applicant may thiiik pressures will be equal due to the relatively low rate of hpat transfer post-trip References 1 FWC-01-C "Feedwater Control System", Revision 2 (3:22/04) (Pg 22 23 of 46) 2 RPS-01-C "Reactor Protection System" Revision 6 (9/15/00)(Pg 2C of 8 0 )

3 hlain Steam Print 25203-26002-1 (swing check valves downstream gf MSIVs)

NRC KIA SystemIEIA System 003 Reactor Coolant Pump System IKCPSI Number K5 04 RO 3 2 SRO 3 5 C F R Link (CFR 41 5 / 45 7)

Knowledge of the operational implications of the 'ollowini- cmcepts a> tbev apply to the RCPS Effects of RCP shutdown on secondary parameters such as steam pressure steam flow and feed flow NRC KIA Generic System Number RO SRO CFt? Link

1. 22 I RO v SRO QueF'ron ID. 5000015 Origin: New d Memory Level The reactor is shutdown, maintaining Hot Standby conditions.

The following are indications for RCP I A :

- Motor Vibration ==> 0.001 inches peak to peak

- Seal Bleed-Off Flow ==> 0.95 gpm

- Upper Oil Reservoir Level ==> 82.5%

- Motor Stator Temperature ==> 265°F Given these conditions, identify the correct response reiated to conditions of RCP 1A.

A Trip pump because motor vibration exceeds limil B Trip pump not required because nothing exceeds limit C Trip pump because oil reservoir level exceeds limit D Trip pump because stator temperature exceeds limit 4 Justification I

CHOICE (A) - NO WRONG Motor vibration alarm setpoint is 0 002 13 0 005 inches peak to peak In this range an other groups are

[ ontacted for determination as to whether or not pump should be sbutdown VALID DISTRACTOR because the applicant may think that vihratil n is excessive and requiring a pump shutdown CHOICE (B) - NO WRONG Trip is required due to high stator temperature VALID DISTRACTOR because the applicant may all parameters within specification CHOICE (C) - NO WRONG Oil level is in expected range (75 to 85%) The high level aldrm actuates at 87 5%

VALID DISTRACTOR because the applicant m3y think oil level is excessive CHOICE (D) - Y E S RCP stator temperature is normally 160 to 180 degrees F 1?e high motor stator temperature alarm actuates at 260 ticqrees F Per the alarm response procedure opeia'ors are directed to trip the plant and then the piimp above 260 degrees F References 1 ARP-2590B-067 "RCP A VIBRATION HI ", Revision 0 2 ARP-2590B-066. "RCP A STR TEMP HI" Revision 0 3 ARP-25906-082, "RCP A UPR OIL RSVR LEVEL HI" Revision 0 4 CVC-00-C "Chemical and Volume Control System Lesson Revision 813 (Pg 14 of 165)

~~~

NRC KIA SystemIElA System 003 Reactor Coolant Pump System (RCPS)

Number A I 03 RO 2 6 SRO 2 6 CFR Link (CFR 41 5 I 4 5 5)

Ability to predict andlor monitor changes in paramerwi (to prevent exceeding design limits) associated wlth operating the RCPS controls including RCP motor stator winding temperaturFs NRC KIA Generic System Number RO SRO CFR Link

1. 1. 23 I d RO fl SRO Questan ,E. 0055022 Origin: Bank Memory Level You have just taken the shift as the P P n .>:ith tile folliwi-7 plant conditions:

- Plant is shutdown and in Mode 4

- RCS temperature is 290 degrees

- RCS pressure is 300 psia

- SDC warmup was initiated about ten miniites prior to yoiir taking the shift The Aux. Bldg. PEO calls up and reports tbat the "B" LPSl pump is making abnormal noises and the pump casing feels hot to the touch. He also reports that the "A" LPSl pump has normal running indications.

On C-01 the "6"LPSl pump motor current is lower than normal, and lower than the amps on the "A" LPSl pump.

The probable cause of this condition is that "B" LPSl Pump A has a failing pump bearing B is operating at runout conditions C has a seated discharge check valve D is experiencing a vortex in the SDC suction line.

Justification I

CHOICE (A) - NO WRONG A failling bearing would put load on the ouinp shaft causing amps to increase VALID DISTRACTOR because the applicant may think that the increased zasing temperature is due to bearinq failure CHOICE (B) - NO WRONG Runout conditions would be indicated by h gtier not lower m , p s VALID DISTRACTOR because the applicant the abnormal noise is caused by runout conditions CHOICE (C) - YES During SDC warmup. only one LPSl pump should be runn'iq to preveiil the performance imbalanres bctwsen the pumps from seating one of the pump's discharge check valves Thic causes that pump to run at shutcff head causing it tn overheat CHOICE(D) NO WRONG Suction line vortexing is a phenomenon asdciated with mid loop operations where air becomes entrained in the SDC suction Given the RCS IS 290 degrees F ana 300 psia conditions do not exist to allow SDC suctlon line vnrtexing VALID DISTRACTOR because the applicant may recognize that the low amps are one indication of loss of suction which can occur when air is entrained in the suction line References 1 SDC-00-C "Shutdown Cooling System" Reviaion 3 (1124103) (Pg .J of 79)

NRC KIA SystemIEIA System Number RO SRO CFR Link NRC KIA Generic System 21 Conduct of Operations Number 2 17 RO 3 SRO 4 4 CFRLink (CFR 4 3 5 1 4 5 1 2 , 45 13)

'Ability to evaluate plant performance and make owrations' IUOL riimls based on operating characteristics reactor behavior and instrument interpretation "

1. 24 I 9 RO !t SRO Question ID: 5000016 Origin: Bank Memory Level A LOCA Outside of Containment has occurred at the plant. In addition, during post trip EOP actions it was determined that reactor coolant radiation levels are significantly above normal. 15 minutes have elapsed since the reactor trip occurred. A General Emergency Classification has been made by the Shift Manager. The emergency response organization has NOT yet been staffed.

It has been determined that the LOCA can be isolated from the Mechanical Penetration area, however dose rates are very high. Radiation Protection Group surveys indicate that the general area dose rate is 50 REM per hour in the area of the Mechanical Penetration area.

Using Emergency Exposure Limits, what is the MAXIMUM stay time for an operator entering the area to isolate the leak with all dose extensions necessary for this condition granted?

A 6 minutes B 30 minutes 9 C 90 minutes D 120 minutes Justification I For "protection of large populations" the dose limit utilizing Emergency Exposure limits is 25 Rem. If the Dose Rate is 50 Remihr in the vicinity, the stay time would be 30 minutes. A, C, D distractors are equivalent to 5 Rem, 75 Rem. and 100 Rem - All plausible mer and Non Emer numbers CHOICE (A) - NO WRONG Dose would be 5 Rem Limit is 5 times annual non-emergency limit of 5 Rem TEDE. which is equal to 25 Rem VALID DISTRACTOR because 5 Rem, an established dose limit for non-emergency situations, is plausible CHOICE (B) -YES Dose would be 25 Rem which is the Emergency Exposure limit CHOICE (C) - NO WRONG Dose would be 75 Rem Limit is 5 times annual non-emergency limit of 5 Rem TEDE, which is equal to 25 Rem VALID DISTRACTOR because 75 Rem an established dose limit for life-threatening emergency situations ISplausible CHOICE (D) - NO WRONG Dose would be 100 Rem Limit is 5 times annual non-emergency limit of 5 Rem TEDE, which is equal to 25 Rem VALID DISTRACTOR because 100 Rem is plausible References J

1 10CFR20 "Standards for Protection Against Radiation" Subpart 20 1206, "Planned Special Exposures" 2 RPM 5 1 5, "Planned Special Exposures" 3 Source Indian Point 3 NRC Exam 1212003 NRC KIA SystemlElA System Number RO SRO CFR Link NRC KIA Generic System 23 Radiation Control Number 234 RO 2 5 SRO 3 i CFR Link (CFR 43 4 / 45 10)

Knowledge of radiation exposure limits and contamination Lontrol including permissible levels in excess of those authorized

1. 25 I *I RO *I SRO Question IC: 5000018 Origin: Bank *I Memory Level A steam generator tube rupture has occurred on SG2 EOP-2534, "Steam Generator Tube Rupture" has been implemented. Which of the following actions is performed in accordance with EOP-2534.

"Steam Generator Tube Rupture" to DIRECTLY limit the potential radiation release to the public?

A raise ruptured SG level above 40% using TDAFW pump B ensuring ruptured SG ADV setpoint at 920 psia and closed C tripping RCPs if pressurizer press less that 1714 psia and SlAS initiated D entering EOP-2536 (ESDE) for a SG pressure < 800 psia and subcooling going up Justification

~ ~

I CHOICE (A) - NO WRONG MSI blocked to facilitate controlled cooldown via preferred method (turbine bypass valves to condenser)

VALID DISTRACTOR because MSI is blocked by procedure CHOICE (B) YES The ADV is ensured to be in auto and its setpoint is raised to a value below the upper end of the band It is also ensured to be closed since steam pressure should be below this point This places the ADV in a condition to open prior lo pressure in the isolated steam generator reaching the MSSV lift setpoint, and minimizing the possibility that a MSSV w11 open and stick in an open position CHOICE (C) - NO WRONG RCP trip strategy based on worst-case LOCA concerns Continued operation of pumps is preferable during a SGTR event to allow for a prompt controlled RCS cooldown and depressurization VALID DISTRACTOR because RCP trip is directed by procedure under specified conditions CHOICE (D) - NO WRONG Functional Recovery Procedure is used to address multiple events from a symptom-based perspective EOP 2536 should not be entered and implemented from EOP-2534 with multiple events in progress VALID DETRACTOR because EOP-2534 does contain diagnosis confirmation steps and the functional recovery does address excess steam demand events References 1 EOP-2534 "Steam Generator Tube Rupture" Revision 22, (3122102) (Pg 10, 17 of 64) 2 TG-2534 Steam Generator Tube Rupture Revision 21 (Pg 14 22, 30 37 of 126) 3 Source Indian Point 3 NRC Exam 1212003 NRC KIA SysternlElA System Number RO SRO CFR Link NRC KIA Generic System 23 Radiation Control Number 2 3 11 RO 2.7 SRO 3 2 CFR Link (CFR' 45.9 145 10)

Ability to control radiation releases

1. 1. 26 I *I RO SRO Question ID: 500(~019 Origin: Bank v Memory Level Unit 2 is operating at 100% power. Which of the follov~ingUnit 2 activitieslevents requires direct notification of Unit 3 personnel?

A planned release of Waste Gas Decay Tank T - I 9A B entry into 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS action statement for a RWST b c - w sample result of 1685 ppm C manageable steam leak on body of HD-I03A, Feedwater Heater 1A Normal Dump Valve D small oily rag bin fire in turbine building extinguished within 10 minutes v Justification I

CHOICE (A) - NO WRONG Notification of Unit 3 not required for planned releases of waste gas decay tanks VALID DISTRACTOR A radioactive discharge from any unit is of general interest to the entire site The applicant may therefore think that notification of Unit 3 is required CHOICE (E) NO WRONG Notification of Unit 3 not required for entry into TS action statements VALID DISTRACTOR Tech Specs require a unit shutdown to COLD SHUTDOWN if RWST boron concentration remains out of spec for greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> A plant shutdown does require a plant announcement per MP-14-OPS GDL200 CHOICE (C) - NO WRONG Notification of Unit 3 not required for steam leaks on Unit 2 The normal dump valve IS isolable VALID DISTRACTOR Steam leak is a concern for personnel safety and continued operation C-OP-200 4 provides direction for addressing the event but does not require Unit 3 notification A steam leak requiring a unit shutdown would however be announced on plant page CHOICE (D) YES AOP 2559 'Fire" requires direct notification of Unit 3 for all fires References 1 MP-14-OPS-GDL200, "Conduct of Operations", Revision 8 (9/09/04) (Pg 27 of 42) 2 C-OP 200 4, "Response to Significant Plant Leaks" Revision 1 (11'26/96) 3 AOP-2559 "Fire", Revision 7 (3/24/04) (Pg 6 of 34) 4 Heater Drains Print 25203-26004 Sheet 3 of 3 (H-9) 5 TS 3 5 4 Refueling Water Storage Tank 6 Source Indian Point 3 NRC Exam, 12/2003 NRC KIA SystemlElA System Number RO SRO CFR Link NRC KIA Generic System 21 Conduct of Operations Number 2 1 14 RO 2 5 SRO 3 3 CFR Link (CFR 43 5 / 4 5 12)

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

1. 27 1 4 RO SRO Question ID: 5000025 Origin: New g Memory Level The plant is at full power. CEA partial movcmenl testing is in progress per SP-2620A when Reg Group 7 CEA #65 drops to 162 steps. In accordance with Technical Specification 3.1.3.1, reactor power is lowered to less than within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> primarily to reduce the effects on A 70°/0, long term power distributions from xenon redistribution B 85%, long term power distributions from xenon redistribution C 70%. available shutdown margin used in accident analyses D 85%. available shutdown margin used in accident analyses

~~~

Justification I

CHOICE (A) - YES Basis for Tech Spec 3 1 3 1 explains that power is lowered tc 70% to reduce the xenon redistribution effects on long term power distributions CHOICE (B) - NO WRONG The basis explains that for small misalignments ( ~ 2 steps) 0 there is a negligible effect on the available SHUTDOWN MARGIN VALID DISTRACTOR TS 3 10 2, "SPECIAL TEST EXCEPTIONS - GROUP HEIGHT AND INSERTION LIMITS" limits sower level to 85%

CHOICE (C) - NO WRONG The basis explains that for small misalignwents ( ~ 2steps) 0 there is a negligible effect on the available SHUTDOWN MARGIN VALID DISTRACTOR TS 3 1 3 1 limits power level to 70"k The basis states that the specifications of section 3 1 3 1 ensure that (1) acceptable power distribution limits are maintained (2) the minimum SHUTDOWN MARGIN is maintained and ( 3 ) the potential effects of a CEA ejeLtion accident are limited to acceptable levels CHOICE (D) - NO WRONG The basis explains that for small misalignments ( - 20 steps) there is a negligible effect on the available SHUTDOWN MARGIN VALID DISTRACTOR TS 3 10 2. "SPECIAL TEST EXCEPTIONS - GROUP HEIGHT AND INSERTION LIMITS" limits power level to 85%

References 1 1 T S 3 1 3 1 "MOVABLE CONTROL ASSEMBLIES CEA POSITION", Amendment 280 (Pg 314 1-20) 2 T S 3 1 3 1 Basis (Page B 3/4 1-3a B 314 1-4)

NRC KIA SystemIElA System 005 InoperablelStuck Control Rod Number AK3 05 RO 3 4 SRO 4 2 CFR Link (CFR 41 5.41 10 / 45 6 / 45 13)

Knowledge of the reasons for the following responses as they apply to the Inoperable / Stuck Control Rod Power limits uri rod misalignment NRC KIA Generic System Number RO SRCJ TFR Link

1. 28 I 4 RO rl SRO Question ID: 5000026 Origin: New Memory Level The plant is in MODE 6 with refueling operations in proylress.

One Wide Range Excore Nuclear Instrumelit has failed, repairs are in progress.

An I&C supervisor calls the control room aiid reports that, based on an audit of completed surveillances, it has been determined two of the remaining channels were improperly calibrated by an inexperienced technician and should be considered inoperable. The remaining channel was properly calibrated.

What impact does this have on fuel handliny activities and why?

A All fuel movement in containment and the spent fuel pool must be suspended due to inadequate remaining instrumentation for monitoring the state of the core.

B All fuel movement in to and out of the reactor core must be suspended due to inadequate 4 remaining instrumentation for monitoring the state of the core.

C Fuel movement may continue since the operability of the remaining channel is adequate for monitoring the state of the core.

D Fuel offload activities may proceed; fuel reload must be suspended due to inadequate remaining instrumentation for monitoring positive reactivity additions.

Justification I CHOICE (A) - NO WRONG CORE ALTERATIONS must be suspender: without two operable channels, activities in the spent fuel pool are not CORE ALTERATIONS VALID DISTRACTOR Plausible that all fuel handlinq wouln be stopped CHOICE (B) - YES CORE ALTERATIONS must be immediately suspended f m movement in the core IS a subset CHOICE (C) - NO WRONG Minimum channels operable requirement is TWO source I ange channels VALID DISTRACTOR Plausible that one channel sufficient for Lbr6 alterations CHOICE(D) NO LYRONG All CORE ALTERATIONS must be suspended not just those involving positive reactivity additions VALID DISTRACTOR Plausible that only concerned about addition of positive from new fuel loading Note This question on both the RO and SRO exams samples CFR 55 43(6) "Procedures and limitations involved in initial core loading alterations in core configuration control rod programming and determinabon of various internal and cxternal effects on core reactivity "

References T S 3 9 2 14 9 2 'REFUELING OPERATIONS - 'NSTRUMENTArION" Amendment 263 (Pg 314 9-2 B 314 9-1)

NRC KIA SystemlElA System 032 Loss of Source Range Nuclear instrumentation Number AK3 02 RO 3 7' SRO 4 1 CFR Link (CFR 41 5 41 10 1 45 6 1 45 13)

Knowledge of the reasons for the followlng responses as they apply to the Loss of Source Range Nuclear Instrumentation Guidance contained in EOP for loss of source-ranqe nuclear instrumentation NRC KIA Generic System Number RO SRO CFR Link

1. 1. 29 I RO d 8SRO Question ID: 3953696 Origin: Bank fl Memory Level Which one of following boration flowpaths would be a\:aiiable after a loss of 8-51?

A BAST gravity feed valves (508, 509) and a charging pump.

B BA pump, BA isolation (514) and a charging pump fl C BA pump, BA flow control valve (21L)Y), VCT outlet (501), and a charging pump.

D RWST isolation valve (192). RWST to chg pp suction (504) and a charging pump

~

Justification CHOICE (A) - NO WRONG CH-508 and CH-509 are powered from 6-51 VALID DISTRACTOR May think that gravity feed flowpath not affected by loss of 8-51 CHOICE (B) - YES No components in this flowpath are affected by a loss o f E-5' CHOICE (C) - NO WRONG CH-501 powered by 6-51 VALID DISTRACTOR May think that 501 unaffected by loss of 6-51 CHOICE ( D ) - NO WRONG CH-504 powered by 6-51 VALID DISTRACTOR May think that 504 unaffected by loss of 8-51 References i 1 ESA-01-C, "Engineered Safety Features Actuation System" Lesson Revision 3 (8/6/01)

NRC KIA SystemlElA System 004 Chemical and Volume Control >ystern Number A4 18 RO 4 3 SRO 4 1 CFR Link (CFR 4117 / 45 5 to 45 8 )

Ability to manually operate and/or monitor in the ccntrol rooni F.rtrgency borate valve NRC KIA Generic System Number RO SRO CFR Link

1. 30 I ~ R O~ S R O Questiori ID: 5000026: Origin: New d Memory Level The technical specification allowed outage time for one train of containment spray reflects the dual function of containment spray for .

A heat removal and iodine removal 4 B heat removal and sump pH control C hydrogen reduction and iodine removal D hydrogen reduction and sump pH control Justification CHOICE (A) - YES Per TS Basis The containment spray is more effective than the containment cooling system in reducing the tcniperature of superheated steam inside containment following a main steam line break In addition the containment spray system provides a mechanism for removing iodine from the containment atmosphere Therefore at least one train of containment spray is required to be OPERABLE when pressurizer pressure is >I750 psia and the allowed outage time for one train of containment spray reflects the dual function of containment spray for heat removal and iodine rernoval CHOICE (B) - NO WRONG' Sump pH control is provided by trisodium phosphate (TSP) dodecahydrate stored in dissolvirig baskets located in the containment basement. It functions to minimize the possibility of corrosion cracking of certain metal components during operation of the ECCS following a LOCA The TSP will get into solution during a LOCA even if mntainment spray is unavailable. Sump pH control is not a function of containment spray.

VALID DISTRACTOR: Control of pH is provided by TSP CHOICE (C) - NO LVRONG: Per lesson material: The introduction of highiy .:cidic borated water in a fine mist to the containment will rssult in the liberation of hydrogen gas in containment This is produced as a result of the metal-water reaction with

<tlurninumand zinc components. Corrosion of these components is minimal and therefore the brief exposure to containment spray will result in negligible loss of structural integrity of these components. The generation of hydrogen by this mechanism is minimized by controlling the iriventory of susceptible metals and by neutralizing the acidity of the water with Trisodium Phosphate.

VALID DISTRACTOR: Amount of hydrogen generation is minimized. but hydrogen concentration is not reduced. by sump pH control.

CHOICE (D) - NO WRONG Per lesson material: The introduction of highly acidic borated water in a fine mist lo the containment will result in the liberation of hydrogen gas in containment This is produced as a result of the metal-water reaction with aluminum and zinc components. Corrosion of these components is minimal and therefore the brief exposure to containment spray will result in negligible loss of structural integrity of these components. The generation of hydrogen by this mechanism is minimized by controlling the inventory of susceptible metals and by neutralizing the acidity of the water with Trisodium Phosphate.

VALID DISTRACTOR. Amount of hydrogen generation IS miriimizec'. but hydrogen concentration IS not reduced, by sump pH control.

References 1 TS 3:'4 6.2 1 Basis, "CONTAINMENT SYSTEMS - DEPEESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY AND COOLING SYSTEUS" Amendment 236 (Pg B 3/4 6 - 3 )

2 CSS-00-C. "Containment Spray System" Lesson. Revision 4 ( V 1 6 / C l ) , Section D.2 a (Pg 23 of 54)

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NRC KIA SystemlElA System 026 Containment Spray System (CS3I Number K406 RO 2 8 S R O 3 2' CFR Link (CFR 41 7)

Knowledge of CSS design feature(s) and/or interlork!s) which provide fol the following Iodine scavenging via the CSS NRC KIA Generic System Number RO SRO CFR Link

1. 31 I 4 RO 4 SRO Questilw ID: 0053dY4 Origin: Bank d Memory Level The plant is operating at 100% power with bus 24E aligned to bus 24D. The "B" RBCCW Heat Exchanger is not in service.

The "A" RBCCW Pump breaker trips and the first attempt to remotely reclose the breaker are not successful. A PEO is dispatched to determine why !he breaker cannot be closed remotely.

Per AOP-2564, "Loss of RBCCW", which c;f the following actions must be performed?

A Align and start the 'B' RBCCW pump to supply Facility 1 RBCCW Header. 4 B Immediately trip the reactor, then trip the affected RCPs due to the loss of RBCCW.

C Realign Bus 24E to Bus 24C and start the 'B' HBCCW Pump on Facility 1 RBCCW Header.

D Coordinate with PEO for a second attempt to reclose the motor breaker. If RCP seals exceed 250 degrees F, then trip reactor and affected RCPs.

Justification I

CHOICE (A) - Y E S The B RBCCW pump is available to supply Facility 1 within 5 minutes The guidance provided in AOP allows utilizing the 'B' pump to supply Facility 1 even though it is electrically aligned to Facility 2 CHOICE (B) - NO WRONG AOP directs compensatory actions A reactor trip would not be required unless unable to restore flow in a timely fashion VALID DISTRACTOR A sustained loss of one header will require a reactor trip CHOICE (C) - NO WRONG Insufficient time available to realign the h i s power source in accordance with procedurr Realignment of power source is not required by AOP VALID DISTRACTOR When performing routine rtalignments BO5 24E would be shifted to Bus 24C CHOICE (D) - NO vVRONG AOP specifically allows for only one attempt to restart Focusing all effort on restart of the affected pump coulcl result in a required reactor trip if unsuccessful Given that breaker has tripped and cannot be immediately rvclosed it is likely there may be a problem with the component VALID DISTRACTOR Applicant may think this is the appropriate action References I I

1 AOP-2564, "Loss of RBCCW", Revision 4 ( 1 2/6/02) (Pg 6. 10 of 46)

NRC KIA SystemlEIA System 008 Number RO SRO CFR Link NRC KIA Generic System 21 Conduct of Operations Number 2 1 23 RO 3 3 SRO 4 0 CFR Link (CFR 45 2 / 45 6)

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

1. 32 I d RO d SRO Questiclr? ID: 5000029 Origin: New Memory Level The plant is operating at full power with ali equipment functional, except for the 'B' HPSl Pump, which is 00s for maintenance.

Then, a large break LOCA occurs combined with a loss of Bus 24D (due to an electrical fault on 24D).

Which one of the choices correctly completes the following statement regarding the impact of the loss of ECCS pumps.

10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> after the event, a loss of the only available - adversely affect long term core cooling because the remaining .____

A HPSl pump would, LPSl pump does NOT have a system flowpath for boron precipitation control.

B HPSl pump would, LPSl pump could NOT be procedurally realigned for boron precipitation d control via hot leg injection.

C LPSl pump would NOT, HPSl pump is preferred for boron precipitation control D LPSl pump would NOT, HPSl pump could be procedurally realigned for boron precipitation control via hot leg injection.

Justification CHOICE (A) NO WRONG LPSl does have a system flowpath for boron p r x pitation control VALID DISTRACTOR Plausible that HPSl injection necessary for adequate injection flow CHOICE (B) YES 4 single HPSl pump will provide sufficient flow for lang term cooliwi \ LPSl pump could physically be aligned for hot 11-4injection but the EOPs do not provide prOCedUrdl gddance lor p 2rformingthis task CHOICE (C) - NO

'IZIRONG Loss of LPSl would have adverse affect because of inability to realign HPSl VALID DISTRACTOR HPSl does provide sufficient Lore cooling flou CHOICE (D) - NO WRONG HPSl could not be procedurally realigned for boron precipitation control because it IS needed for injection VALID DISTRACTOR HPSl could physically realigned but not iaw procedure References 1

i ECC 01-C "Emergency Core Cooling System" Revision 3 (6/28/01) (Frl 11 13 of 25) 2 OP 2541 Appendix 18 "Simultaneous Hot and Cold Leg Injection" NRC KIA SysternIElA System 006 Emergency Core Cooling System (ECCS)

Number K6 13 RO 2 8 SRO 3 ' CFF Link (CFR 41 7 / 45 7)

Krlowledqe of the effect of a loss or malfunction on the followi lq MII havf on the ECCS Pumps NRC KIA Generic System Number RO SRO CFR Link

1. 33 1 dR0 vSRO Questiun ID. 500003G Origin: Bank fl Memory Level The following Quench Tank parameters aiF: provided:

- Temperature is 150°F

- Level is 52%

- Pressure is 3 psig

- Oxygen concentration is 2%

What action is required to restore conditions to normal?

A Lower pressure to less than 1 psig.

B Lower level to less than 45%

C Lower 0 2 concentration to less than 156 D Lower temperature to less than 120°F V Justification I

CHOICE (A) - NO WRONG Pressure is normally maintained between 1 and 5 psig VALID DISTRACTOR Applicant may think pressure is too high CHOICE (B) NO uVRONG Level is maintained at 50% and must be maintained above '5%

VALID DISTRACTOR 45% is the low limit CHOICE (C) - NO LVRONG Concentration is maintained less than 4")" oxygen VALID DISTRACTOR Applicant may think concentration mu:! hc isducetl below 1%

CHOICE (D) - YES Ternoerature is maintained below 120°F References 1 OP-2301A. "PDT and Quench Tank Operation", Revision 10 (7:26/04) (Pg 4,6,8 16 of 37) 2 ARP-2590B-207, "QUENCH TANK TEMP HI" Revision 0 (3/4/04) 3 Source INPO Bank - Q# 19360 - Used at Kewaunee 1 12111l2000 NRC KIA SysternIElA System 007 Pressurizer Relief TanklQuench Tank System (PRTS)

Number A I 03 RO 2 6 SRO 2 7 CFR Link (CFR 41 5/45 5 )

Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating tho PRTS controls including Monitoring quench tank tenaerature NRC KIA Generic System Number RO SRO CFR Link

1. 34 I 4 RO 4 SRO Question ID: 5000031 Origin: New Memory Level Unit 2 was in the process of raising power 10 100% Given the following events and conditions:

Reactor power

- Q Power Channel 'A' = 72.0%

- Q Power Channel 'B'= 73.0%

- Q Power Channel 'C' = 72.0%

- Q Power Channel 'D' = 75.0%

Thermal power = 72.5%

Variable High Power Trip (VHPT) set points were last reset at Q=66.5%

Q Power Channel 'A' now FAILS HIGH.

Which one of the following statements is ccrrect concerning the effect of these conditions?

A The VHPT trips have actuated on channels "A" and "D" and the reactor will trip.

B The VHPT pretrip has actuated on only channel "A" and CEA withdrawal motion is inhibited.

C The VHPT pretrips have actuated on channels "A" and "D" and CEA withdrawal motion is fl inhibited.

D The VHPT pretrip has actuated on only channel "A" and CEA withdrawal motion is NOT inhibited.

Justification I

CHOICE (A) - NO WRONG AB. AC and AD logic ladders have tripped from tha signal in Channel " A but all other channels remain below tlie VHP trip setpoint. The reactor will not trip unless another channel exceeds 8 8%.

VALID DISTRACTOR: Applicant may think logic is met tor a reactcr trip CHOICE (5) - NO WRONG, Channels "A" and "D" have exceeded their pretrip setpoint 2 of 4 pretrips cause a CEA withdrawal inhibit and control rods will not move out.

VALID DISTRACTOR: If the applicant thinks that CH "A'is no longer in pretrip because it has already tripped then this could be a plausible answer.

CHOICE (C) - YES Channels "A" and "D" have exceeded their pretrip setpoint 2 of 4 pretrips cuase a CEA withdrawal inhibit and control rods will not move out.

CHOICE (D) - NO WRONG Channels " A and " D have exceeded their pretrip setpoint 2 o: 4 pretrips cause a CEA withdrawal inhibit and control rods will not move out VALID DISTRACTOR If the applicant does not recall the ','tlP trip setpcmt 7 r thinks that they are continually reset during a power ascension (as they are during a power decrease) and compares channel power to thermal power this distractor could be selected References 1 RPS-01-C. "Reactor Protection System" Lesson, Revision 6 (9:i5/73, (Pg 19 of 80 and Figures 7 20, and 331 NRC KIA SystemlElA System 012 Reactor Protection System Number K301 RO 3 9 SRO 4 0 CFR Link (CFR 41 7 / 45 6)

Knowledge of the effect that a loss or malfunctiov c f the RPS will have ?n :ne following CRDS NRC KIA Generic System Number RO SRO CFt' Link

1. 1. 35 I r, RO r, SRO Question ID. 00538eti Origin: Mod Memory Level Given the following conditions:

- 100% reactor power

- Inverter 2 has been isolated in preparation for repairs The DC input breaker on Inverter 6 is inadvertently opened while hanging the clearance on Inverter 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 'A' LPSI Pump will NOT be running.

B 'B' LPSl Pump will NOT be running r, C 'C' CAR Cooler Fan will be running in tast speed.

D 'D' CAR Cooler Fan will be running in slow speed Justification I CHOICE (A) - NO WRONG The LBLOCA will actuate SlAS Facility 1 ESAS equipment will operate as designed VALID DISTRACTOR 'A' LPSl Pump will be running CHOICE (B) - Y E S Opening DC input breaker on Inverter 6 with Inverter 2 out will deenergtze 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 which would normally result in an actuat,on '3' LPSl will remain stopped until manually started by the opcrator CHOICE ( C )- NO WRONG The LBLOCA will actuate SlAS Facility 1 ESAS equipment will operate as designed VALID DISTRACTOR 'A' CAR Fan will shift to slow speed CHOICE ( D ) - NO WRONG The LBLOCA will actuate SlAS Facility 2 ESAS equipment will not receive actuation signals VALID DISTRACTOR 'B' CAR Fan will remain I F fast speed References I 1 LVD-00-C, "125 VDC/120 VAC" Lesson, Revision 5 (Pg 9 of 81)

NRC KIA SystemlElA System 01 3 Engineered Safety Features Actuatior, qystem (ESFAS)

Number A204 RO36 SRO 4 2 CFRLink (CFR 41 5 / 4 3 5 : 4 5 3 / 4 5 1 3 )

Ability to (a) predict the impacts of the following malfunctions or operations on the ESFAS and (b) based Ability on those predictions use procedures to correct, control or nitigate the consequences of those malfunctions or operations Loss of instrument bus NRC KIA Generic System Number RO SRO CFR Link

1. 36 I d RO d SRO Question E: 5000C;33 Origin: New Memory Level The plant is operating at 75% power when Main Steam 'Line Pressure Transmitter PT-4224 for the #2 ADV fails high.

Which of the following describes the response of 'B' SG level to this instrument failure? Assume NO operator action.

A Level will NOT change, the feedwater control system will maintain level constant B Level will initially increase, then stabilize to mainkin a level equal to the level prior to the 4 failure.

C Level will initially decrease, then stabiliye to maintain a level equal to the level prior to the failure.

D Level will initially increase, then stabilize to maintain a level higher than the level prior to the failure. I Justification CHOICE (A) - NO WRONG: Level will swell when the ADV on #2 MSL opens. The feedwater control system will restore level to setpoint VALID DISTRACTOR. Applicant may not recognize the effect of the instrument failure on the ADV.

CHOICE ( 6 )- YES Lcwcl will swell when the ADV on #2 MSL opens. The feedwater control system will restore level to setpoint CHOICE (C) - NO WRONG Level will swell when the ADV on #2 MSI. opens. .'he re-dwater control system will restore level to setpoint

'VALID DISTRACTOR: Applicant may think that the predominant level effect will be shrink due to additional feedwater when FRV opens in response to steam-feed mismatch CHOICE (D) - NO WRONG. Level will swell when the ADV on #2 MSL cpens The feedwater control system will restore level to setpoint VALID DISTRACTOR: Applicant may think the controller will maintain level higher than setpoint.

References 1 MSS-00-C. "Main Steam System" Lesson, Revsion 6 (711 1/01) Page 16, 25 of 69) 2 Millstone Unit 2 FSAR, Revision 21, Section 7 4 7 (Pg 7 4-19)

NRC KIA SystemlEIA System 039 Main and Reheat Steam Systeni (MRSS)

Number K1 01 RO 3 1 SRO 3 2 CFR Link (CFR 41 2 to 41 9 / 45 7 to 45 8)

Knowledge of the physical connections and/or caux-effect relatioriships between the MRSS and the following svstems SIG N R C KIA Generic System Number RO SRO CFR Link

1. 37 I g RO g SRO Question ID: 0070216 Origin: Mod Memory Level The Main Steam Isolation Valves will autormticaily close in response to which one of the following sets of conditions?

A PT-I013A, SG 1 CHANNEL A PRESSURE = 585 psia PT-10138, SG 1 CHANNEL 6 PRESSURE = 570 psia B PT-I013A, SG 1 CHANNEL A PRESSURE = 564 psia PT-I023B, SG 2 CHANNEL B PRESSURE = 566 psia C PT-I023A, SG 2 CHANNEL A PRESSURE = 567 psia P i - I 0236, SG 2 CHANNEL B PRESSURE = 584 psia D PT-I013B, SG 1 CHANNEL B PRESSURE = 552 psia PT-I023B, SG 2 CHANNEL 6 PRESSURE = 555 psia Justification CHOICE (A) - NO WRONG Only CH B is < 572 psia.

VALID DISTRACTOR. Both channels are on same SCJ CHOICE (6) -YES Both pressures 4 7 2 psia. one is CH A the other CH B MSI is generated by 2/4 SG pressure <572 psia on any 2 channels provided they are not the same letter designatioi~ For example SG1 CH A and SG2 CH B (one A and one B)IS an acceptable combination. whereas S G l CH 4 and SG2 CH A (both A's) is not an acceptable combination CHOICE (C) - NO WRONG Only CH A IS 572 psia VALID DISTRACTOR Different channels on the same SG CHOICE (D) - NO WRONG Both transmitters have same designation (CH B)

VALID DISTRACTOR Both pressures are < 572 psia References I

1 SA-01-C "Engineered Safety Features Actuation System" Lesson Revision 3 (8/6/01) Table H and Figure 5 (Pg 52 53)

NRC KIA SystemlElA System 039 Main and Reheat Steam Systen, (VRSS)

Number A302 RO 3 1 SRO 3 5 CFR Link (CFR 41 5 / 45 5)

Ability to monitor automatic operation of the MRSS IncIuo r,q Isolatiori of the MRSS NRC KIA Generic System Number RO SRO CFR Link

1. 38 I d RO d SRO Question ID: 5000034 Origin: Mod Memory Level Given the following plant conditions:

- 100% power

- SG levels at setpoint

- Steam flow and feed flow matched

- SG2 Feed Flow Transmitter FT-5269A output fails high With NO operator actions, which of the fol!GWing dsscribes the expected plant response?

A SG level lowers, but stabilizes above tliz low level reactor trip.

8 SG level lowers to the low level reactor trip.

C SG level rises, but stabilizes below the high level turbine trip.

D SG level rises to the high level turbine trip.

Justification 1 CHOICE (A) - Y E S The relatively small change in the magnitude of the averaged feed flaw signal with this failure occurring at 100°/~power is not sufficient to drive level to the low level trip. The controller will act on the level deviation signal to restore level to setpoint.

CHOICE (B) - NO WRONG: Output from feed flow transmitters FT-5269A and FT-52693 i?nthe SG2 feed line are averaged for input to the three-element level control. Failing one transmifie( high drives Ihe average high However, at full power. the resulting change in the magnitude of the averaged signal is small. The control system will respond by throttling closed on the FRV. The level signal will act on the steam flow signal to moderafo !he response, restoring level to setpoint VALID DISTRACTOR- Applicant may think that feed flow/stearn flow misniatch will drive level to the low level trip setpoint CHOICE (C) - NO WRONG. SG will lower to the low level trip setpoint

'JALID DISTRACTOR: Applicant may think the higher indicated feed flow will cause SG level to rise but stabilize belon the high level trip based on input to the control system from the level signal CHOICE (D) - NO WRONG: SG will lower to the low level trip setpoint VALID DISTRACTOR. Applicant may think the higher indicated feed flow will cause SG level to rise to the turbine trip References 1 FWC-01-C, "Feedwater Control System", Revision 2 (3/22.'04) [Pg 7.8 of 46) 2 Source: INPO Bank - Q# 1942 - Used at Palisaues 1 B/lJUC)S-NRC K/A SystemlElA System 059 Main Feedwater (MFW) System Number K408 RO 2 5 SRO 2 7 CFR Link (CFR 41 7)

Knowledge of MFW design feature(s) and/or interlock(s) which provide for the following Feedwater regulatory valve operation (on basis of steam flow, feed flow mismatch)

NRC KIA Generic System Number RO SRC CFR Link

1. 1. 39 1 dR0 d SRO Ques:ion ID: 5000035 Origin: New Memory Level The plant is at 100% power. I&C is performing troubleshooting in the Facility 1 Auxiliary Feed Actuation Cabinet. A spurious Facility 1 AFAS is generated.

As a result of automatic actions associated witn this event, plant efficiency will and action will be taken to reduce A decrease, reactor power by inserting CEAs while maintaining turbine load constant B increase, reactor power by inserting CEAs while maintaining turbine load constant C decrease, turbine load by adjusting load limit to reduce eactor power *r D increase, turbine load by adjusting ioad limit to reduce reactor power Justification I

CHOICE (A) - NO WRONG, Inserting CEAs will insert negative reactivity but lowering RCS temperature will counter this effect Power will remain > l o o % until turbine load is reduced.

VALID DISTRACTOR. Insertion of CEAs does add negative reactivity and would reduce power in a reactor below the POAH CHOICE (B) - NO

\NRONG: Additional heat required to raise temperature of AFLY rntrring SGs, resulting in a decrease in plant efficiency VAILID DISTRACTOR. Main feedwater will be automaiisally throttletl to cxnpensate for the AFW flow. Applicant may think efficiency is improved by the reduction of main feedwater CHOICE (C) - YES Additional heat required to raise temperature of AFW entering SGs resulting in a decrease in plant efficiency OP-2204 "Load Changes", requires power to be maintained less than 100%

CHOICE (D) - N O WRONG Additional heat required to raise temperature of AFW entering SGs, resulting in a decrease in plant efficiency VALID DISTRACTOR. Main feedwater will be autoniatically throttled to compensate for the AFW flow. Applicant may think efficiency is improved by the reduction of mair! fcedwater References 1 OP-2204, "Load Changes", Revision 19 (6/29/04) (Pg 17 of 46)

~ _____

NRC KIA SystemIElA System 059 Main Feedwater (MFW) Syster.1 Number A2 01 RO 3 4' SRO 7 6' CFR Link (CFR 41 5 / 43 5 I 45 3 / 45 13)

Ability to (a) predict the impacts of the following malfunctions Z r qpcrations on the MFW and (b) based on those predictions use procedures to correct control or mitigdte the consr'quencss of those malfunctions or operations Feedwater actuation of AFW system NRC KIA Generic System Number RO S RO CFR Link

1. 40 1 ~ R Ov SRO Questiori ID. 50012036 Origin: Bank v Memory Level How is power supplied to 120 VAC lnstrutnent Bus VR21 when the LOAD CONNECTED TO NORMAL (amber) lamp is lit on Transfer Switch RS-2?

A 480 VAC from MCC B41A, rectified, and then inverted to 120 VAC B 125 VDC from battery, supplied to Bus 201 D. then to inverted to 120 VAC C 480 VAC from MCC 661, then through step-down transformer to 120 VAC v D 480 VAC from MCC B62, inverted to 120 VAC then isolating transformer to 120 VAC Justification

~~~ ~~

I CHOICE (A) - NO WRONG MCC B61 provides normal power to VR21 VALID DISTRACTOR MCC B41A provides alternate power to VR21 CHOICE (B) - NO WRONG MCC 661 provides normal power to VR21 VALID DISTRACTOR Bus 201 D provides power tc VA2C through INV 6 CHOICE (C) - YES MCC B61 provides normal power to VR21 CHOICE (D) - NO WRONG MCC B61 provides normal power to VR21 VALID DISTRACTOR MCC 862 provides emergency power to VR21 References 1 LVD-00-C "125 VDCI120 VAC" Revision 5 (Pa 10 33 of 77 and Fiaure 3) 2 In House Single Line Diagrams 25203-30001 and 25203-30024 3 Source INPO Bank - Q# 20751 - Used at Braidwood 1. 10/29/2001 NRC KIA SysternlElA System 062 A C Electrical Distribution Number K2 01 RO 3 3 SRO 3 4 CFR Link (CFR 41 7)

Knowledge of bus power supplies to the following Major system loads NRC KIA Generic System Number RO SRO CFR Link

1. 41 1 d RO d SRO Questiw ID: 500fi035 Origin: New d Memory Level Unit 2 tripped from 100% 90 minutes ago.

B u s 24C did NOT transfer to the RSST and the "A" Emergency Diesel Generator (EDG) tripped on DIESEL GEN 12U DIFFERENTIAL LOCKOUT. Electrical Maintenance has found and repaired a loose lead in the EDG trip circuitry which caused the trip.

O n e of the actions required to restart the "A" EDG is to press the Alarm Reset button on the local EDG skid panel.

Which one of the following describes the reason for pressing the Alarm Reset button at this time?

A closes a contact in the start circuitry which energizes the DC air start solenoid valves 8 resets the Shutdown Relay which allows the governor to admit fuel to the diesel d C energizes the Auto Start Relay which resets the mechanical overspeed trip solenoid D resets the Differential Lockout Relay which allows the DG output breaker to be closed Justification I A is incorrect The air start soleniods are NOT energized when the Alarm Reset button is pressed If that were the case the Diesel would start B is correct When the shutdown relay is energized the Diesel is automatically tripped and athe governor is dcctronically placed in a zero fuel position When the Alarm Reset button is pressed the shutdown relay is reset which illows the governor to admit fuel to the Diesel on dernsnd C is incorrect The Auto Start Relay does NOT rest the mwhanical overspeed trip D is incorrect The Differential Lockout Relay is NOT reset when the Alarm Reset button IS pressed References 1 EDG-00 C "Emergency Diesel Generator System' Lesson Revision 7 (8/27/02) (Pg 75 107 142 of 143) 2 OP 2346A 'Emergency Diesel Generators" Revision 25 (6115104) (Pg 17 of 99)

NRC KIA SystemlElA System 064 Emergency Diesel Generators (ED/G)

Number K403 R O 2 5 SRO 3 0 CFRLink (CFR 41 7 )

Knowledge of ED/G system design feature(s) and/or inter- lock(si which provide for the followinq Governor valve operation NRC KIA Generic System Number RO SRO CFR Link

1. 42 1 d RO d SRO Question ID: 5000032 Origin: Mod Memory Level Given the following conditions on Unit 2:

- SBLOCA resulted in a Manual SlAS

- A loss of offsite power occurred coinciderit with the Manual SlAS

- 41 60 Volt Bus 24E de-energized on 86-2 iockout

- 'A' and 'B' EDGs have energized their respective buses How many CAR fans will be operating?

A 1 c3 D 4

~~~~

Justification CHOICE (A) - NO WRONG CAR Fans are powered off of 480 Volt Buses BO5 and 306 These buses will re-energize from the EDGs VALID DISTRACTOR May assume Bus 24E fault w1Il keep Bus 24C or 24D from re-energizing May assume only 1 fan twill automatically restart on loss of offsite CHOICE (6) - N O WRONG CAR Fans are powered off of 480 Volt Buses BO5 and BO6 These buses will re-enerqize from the EDGs

\ / A I I D DISTRACTOR May assume Bus 24E fault tvill keep Bus 24C oi 24D from re-energizing hlay assume 2 fans will automatically restart on loss of offsite CHOICE (C) - NO WRONG CAR Fans are powered off of 480 Volt Buses BO5 znd BO6 These buses will re-energize from the EDGs VALID DISTRACTOR May assume Bus 24E fault will keep Bus 2417 or 24D from re-energizing May assume Bus 24E fault will keep Bus 24C or 24D from re-energizing CHOICE (D) - Y E S Buses BO5 and BO6 will reenergize from the EDGs CAR fans will restArt in slow speed on sequencer 1

References 1

CCS-00-C "Containment and Containment Systems' lesson Revision 8 (1 1/20/00) (Pg 33 of 83) 2 Source INPO Bank - Q# 23156 - Used at Salein 11!4,200:

~ _ _ _ _ _ ~ _ _ _ _ _ ~

NRC K / A SysternIEIA System 022 Containment Cooling System (CCS)

Number K201 RO 3 0 SRO 3 1 CFR Link (CFR 41 7)

Knowledge of power supplies to the following Containment cooling fans NRC KIA Generic System Number RO S RO CFR Link

1. I 43 I d RO d SRO Question ID: 0053401 Origin: Bank d Memory Level The plant is in Mode 4, on the RSST, wit!! bus 24E powered from bus 24D. when the "A" Service Water (SW) Pump breaker shorts internaliy causing a fault on Bus 24C and tripping the 24C-24G tie breaker.

If the appropriate equipment actuates on the Loss of No1mal Power to Bus 24C. which one of the following operator actions is required to prevent further equipment damage?

A Perform a normal shutdown of the "A' EDG B Start the "B" SW and RBCCW pumps on the Facility to which they are aligned C Align the "A" EDG to the Facility 2 SW header D Shutdown the "A" EDG using the Emergency Shutdown push buttons. d Justification I

Tlic " A EDG is running without any cooling water, it should be imrnrnediately tripped to prevent damaging the machine Thc fault on Bus 24C will prevent the A EDG breakpr from closing as well as no SW pump available to the facility CHOICE (A) - NO WRONG. EOP-2525 requires trip of the EDG.

VALID DISTRACTOR: Procedures and lesson materid stress that normal shutdown generally preferable because less stressful to engine.

CHOICE (B) - NO WRONG. The "3" pumps are aligned to Facility 2.

VALID DISTRACTOR: Start of a standby pump is a logical choice CHOICE (C) - NO WRONG No procedural guidance provided to allovi cross-tie of Facility 2 RBCCW with Facility 1 EDG VALID DISTRACTOR: Cross tie physically possible CHOICE (DI - YES EOP directs trip of running EDG on loss of service water References 1 EOP-2525. "Standard Post Trip Actions", Revision 20 (2/22101J (Pg 6 of 26) 2 EOP-2525 Standard Post Trip Actions Technicai Guide. Revision 2@(Pg 6 of 38)

NRC KIA SysternlElA System 076 Service Water System (SWS)

Number A2 02 RO 2 7 SRO 3 1 CFR Link (CFR: 41.5 / 43 5 / 4513 / 45/13)

Ability to (a) predict the impacts of the following malfunctions or operations on the SWS; and (b) based on those predictions. use procedures to correct, control, or mitigate the consequences of those malfunctions or operations-Service water header pressure NRC KIA Generic System Number RO SRO CFR Link

1. 44 I 4 RO 4 SRO Question ID: 5000038 Origin: New Memory Level The unit is operating at full power when il SGTR cscurs. Operators manually trip the plant and initiate SIAS. EOP-2525, "Standard Post-Trip Actions" are performed. While performing Step 2 of EOP-2534, "Steam Generator Tube Rupture", a report is received that a large rupture has been discovered on the Station Air header upstream of Containment Header Isolation 2-SA-42 in the 14 foot Aux Bldg General Area.

Assume the "D" Instrument Air Compressor F3D has beer1 installed and is in service suppling both Instrument Air and Station Air.

Under these conditions, what design feature will enable ttie Instrument Air header to remain pressurized?

A Opening of 2-SAS-6, Station Air Cross Tie to Unit 3 B Closing of 2-IA-642, Instrument Air to Station Air Excess Flow Check 4 C Auto opening of 2-SA-10.1, Cross Tie from Station Air to Instrument Air D Auto closing of 2-SA-23.1, Cross Tie from Station Ail to Containment Air

~~~ ~~~~

I Justification CHOICE (A) - NO WRONG: Given the location of the line rupture, the Unit 3 cross-tis woula supply the leak. The leak cannot be isolated from the Unit 3 air when cross-tied.

VALID DISTRACTOR: Step 32 of EOP-2534 directs alignment of Unit 3 to the Unit 2 Service Air System.

CHOICE (6)-YES Excess flow check valve will close to isolate the leak CHOICE ( C ) - NO WRONG: Valve 2-SA-IO.1 automatically opens when instrument air pressure drops below 85 psig to supply instrument air from station air. Valve 2-SA-I 1. I is interlocked to clos- :$,hen 2-SA- , C 1 is open to stop flow from the station air compressor to the station air system. However, a bypass line prevents this scheme from isolating the service air leak VALID DISTRACTOR: Applicant may think this automatic feature will correct the problem.

CHOICE (D) - NO WRONG Valve SA-23.1 is located within containment ano will not isolatu; the leak.

VALID DISTRACTOR: Valve SA-23.1 is on the service air to containnenr line and fails closed References 1

I ISA-00-C. "Station Air 8 Instrument Air Systems" Lesson. Revision 6 (Pg 12, 13 of 78) 2 OP-2332A-001, "Station Air System Valve Alignment", Revision 8 {Py 4 of 7) 3 Piping Diagram 25203-26009, "Instrument and Station Air System". Sneet 8 of 10, Revision 32 (9127101)

NRC KIA SystemlElA System 078 Instrument Air System (IAS)

Number K4.02 RO 3.2 SRO 3 5 CFR Link (CFR: 41.7)

Knowledge of IAS design feature(s) and/or interlock(s) which provide for the following: Cross-over to other air systems N R C KIA Generic System Number RO SRO CFR Link

1. 1. 45 1 g RO v SRO Question IO: 0072948 Origin: Bank d Memory Level The 'A' Instrument Air Compressor F3A proper!y :aflged for electrical troubleshooting. Electrical Maintenance has determined that the MANUAL i OFF / AUTO Control Switch requires replacement The yellow tag on the control switch must be A cleared prior to removal of the switch from the panel B lifted under a "temporary lift" until the new switch is installed C removed from the switch and attached beside the switch mounting location D maintained with the switch that is removed until transferred to the new switch Justification I CHOICE (A) NO WRONG Yellow tag is not required to be cleared prior to removing switch VALID DISTRACTOR Generally tags must be cleared before manipulating or working on boundary components CHOICE (B) - NO WRONG Yellow tag is not required to be cleared prior to removing switch VALID DISTRACTOR Plausible that tag would be temporary lifted to allow switch to be replaced since tag is for information purposes only CHOICE (C) - YES The tagging procedure states that "If tagged panel switch must be r m o v e d remove tag from switch and attach neat panel hole (yellow only) "

CHOICE (D) - NO WRONG Yellow tag is not required to be cleared priw to removing switch VALID DISTRACTOR Plausible that since tag is for information only it might be kept with the original switch until new switch installed as a way of maintaining control over the tag References 1 WC-2 "Tagging", Revision 6 (5/1/03) Attachment 4 " I ;jging Prqctices" (Pg 55 of 85)

NRC KIA SystemlElA System 078 Number RO SRO CFR Link NRC K/A Generic System 22 Equipment Control Number 2 2 13 RO 3 6 SRO 3 8 CFR Link (CFR 41 10 / 45 13)

Knowledge of tagging and clearance procedures

1. 1. 46 1 *I RO SRO Quesiion ID: 50000.10 Origin: New Memory Level Plant is operating in MODE 1 when operd!ors set. id::ations of a rapid rise in containment pressure coincident with lowering SG pressure. The ieactor is inanuaily tripped. The crew enters EOP-2525.

"Standard Post Trip Actions". While scanni!-ig the control boards, the SPO observes the following:

- ClAS ACTUATION SIG CH 1 TRIP alarn actuated

- ClAS ACTUATION SIG CH 2 TRIP alarrii actuated

- Containment Sump Pump P-33A stopped

- Containment Sump Pump P-33B running SSP-16 1 Containment Drain Sump Isolation Valve open SSP-16 2 Containment Drain Sump Isolation Valve dosed CH-505, RCP Bleedoff Isolation Valve closed CH-506. RCP Bleedoff Isolation Valve c,!osed These conditions indicate .__

A ESAS Block Relay 24VDC power has failed B SPO has overriden the ESAS signal tG SSP-16.1 C CTMT PRESS HI coincidence has not been met D An actuation module for ClAS has failed to actuate *I Justification I

CHOICE (A) - NO WRONG Loss of 24VDC block relay power would altect all Facility 2 CIAS components in same manner Would not have RCP Bleedoff Isolation 2-CH-506 closed with SSP-16 1 open.

VALID DISTRACTOR: 24VDC block power is associated with the Facility 2 ClAS actuation modules CHOICE (B) - NO W:RONG ESAS signal to SSP-16 1 cannot be overriden 1:ALID DISTRACTOR Many ESAS actuation signals can be overriden from switches on the main control boards CHOICE (C) - NO LVRONG ClAS ACTUATION SIG CH 2 TRIP alarr,) vould no! be in if coincidence not made up VALID DISTRACTOR Some of the indications p r o v i d d are consistent inth no ClAS CHOICE (D) - YES

,Abnormal ESF response caused by a failure of ESAS Actuation Module F V-607. The module actuates the following coinponents on a Facility 2 CIAS:

~ 2-RC-001, RC Hot Leg Sampling==Close 2-LRR-43 1 PDT Pump Discharge Valve==Close GR-11 I Waste Gas Surge Tank Inlet Valve==Close SSP-16 1 Containment Drain Sump Isolation Valve==Close

- P-33B Containment Drain Sump Pump==Stop 51-312 N2 to SI Tanks Shutoff Valve I

References 1 ESA 01-C "Engineered Safety Features Actuatiurl System' Lesson r i p w o n 3 (8/6/01) (Pg 44 of 73 and Tables 4

,>rid 5) 2 CCS 00-C 'Containment and Containment Systems LL;son Revision 8 (11/20/00) (Pg 15 of 83)

I ARP-2590A-138 "CIAS ACTUATION SIG CH T TRIP" Revision 0 NRC KIA SystemlElA System 103 Containment System Number A4 03 RO 2 7' SRO 7' CFH Link (CFR 41 7 / 45 5 to 45 8)

Ability to manually operate and/or monitor in the control rooni E3F slave ,clays NRC KIA Generic System Number RO SRO CFR Link

Question ID: 0053544 Origin: Bank Memory Level An estimated critical position ca.-ulation is being perforr?!ed to startup the reac.Jr 29 hours3.356481e-4 days <br />0.00806 hours <br />4.794974e-5 weeks <br />1.10345e-5 months <br /> after a plant trip from 100%. Boron concentration is 955 ppm.

Reference Data:

- Power = 100%

- Xenon = 2.41 Yo delta rho

- Samarium = 0.78% delta rho

- Tavg = 572F

- Burnup = 7,500 MWDlMTU

- Boron = 692 ppm The moderator temperature coefficient is _ _ and if moderator temperature is maintained during the startup at 2°F below the temperature assumed by the ECP, then the critical rod height will be than the calculated estimated critical position.

A negative, lower d B positive, lower C negative, higher D positive, higher Justification I CHOICE (A) -YES MTC is negative at 532°F when boron concentration IS below 1400 ppm CHOICE (B) - NO WRONG. MTC is negative at 532°F when boron c Jri:entra!ion is bciow 1400 ppm VALID DISTRACTOR: CEA height will be lower.

CHOICE (C) - NO WRONG: MTC is negative at 532°F when boron c ,ri?ntratic'n IS L.Aow '400 ppm VALID DISTRACTOR, MTC will be negative CHOICE (D) - NO WRONG- MTC is negative at 532°F when boron concentration IS below 1400 ppm VALID DISTRACTOR. If MTC was positive, then CE:? height would be higher.

References 1 OP-2208-003 "MODERATOR TEMPERATURE COEFFICIENT VERSUS BORON CONCENTRATION MOC CYCLE 16" Revision 044 (4/27/04)

NRC KIA SystemlElA System 001 Control Rod Drive System Number K5 26 RO 3 SRO 3 b CFR Link (CFR 41 5/45 7 )

Knowledge of the following operational implicatiuic AS tl'i. apply to tne CRDS Definition of moderator temperature Loefficient application to reactor control NRC KIA Generic System Number RO SRO C F H Link

1. 48 I 9 RO SRO Quecrrk- ID 5 ~ ~ 0 4 Origin: 1 New g Memory Level The plant IS operating at 100% power wher, a large break LOCA occurs If malfunctions prevent the use of either hydrogen recombiner, identify the approximate time that will elapse from the start of the event before hydrogen concentration will reach 3% by volume inside containment.

A greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 4 B 48 to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> C 24 to 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> D less than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Justification I

CHOICE (A) - YES HCS Lesson material and EOP-2532 Technical Guidr riotti st Ite the time at 12 to 16 days CHOICE (B) - NO WRONG Hydrogen concentration expected to reach 3'6 at 12 to 16 days

\/ALID DISTRACTOR Plausible that concentration would be approaching the limit within several davs CHOICE (C) - NO WRONG Hydrogen concentration expected to reach 3'6 at 12 to 16 days VALID DISTRACTOR Hydrogen monitor tech spec dction time I i r n i L is short term CHOICE (D) - NO WRONG Hydrogen concentration expected to reach 3"0at 12 lo 16 days VALID DISTRACTOR EOP-2532 directs simultaneous hot and cold ley injection at about 1/2 day into event 1

References 1

EOP-2532 Loss of Coolant Accident Technical Guide. Revision 1 1 (Pg 5 of 188) 2 HCS-00-C. "Hydrogen Control System" Lesson Revision 3 (6'27101) (Pg 44 of 48)

~ ~~~ ~

NRC KIA SysternIElA System 028 Hydrogen Recornbiner and Purae Control Syytem (hRPS)

Number K3 01 RO 3 3 SRO 4 0 CFR Link (CFR 41 7 / 45 6)

Knowledge of the effect that a loss or malfunction of the HRPS will have on the following Hydrogen concentration in 1 oritaininent NRC KIA Generic System Number RO SRO CFR Link

1. 49 I # RO 4 SRO Questto? ID: 5000U42 Origin: Mod Memory Level The following conditions exist on Unit 2:

- Reactor power is 8O%, steady state at EGC

- RCS boron concentration is 135 ppm

- All systems are in automatic control

'A' Main Steam Pressure Instpment PT-42 16 output urifts high causing 'A' Steam Dump Valve to Condenser, 2-MS-209, to stroke to approx 30Y0open Assuming NO immediate operator action, what IS the exnected response of the plant due to the steam dump valve failure AND what action can the operator take from the control room to stop the excess steam flow?

A Turbine load will decrease by approx. 3% AND isactor power will remain constant. The operator can stop dumping excess steam by placing the Bypass to Condenser Controller PIC-4216 to MAN and in the CLOSE position.

B Turbine load will decrease by less than 3% AND reactor power will increase by approx. 3%. 4 The operator can stop dumping exces.: steam by placing the Bypass to Condenser Controller PIC-4216 to MAN and in the CLOSE position.

C Turbine load will decrease by approx 3% AND reactor power will remain constant. The operator can stop dumping excess steam by taking the Quick Open Permissive Switch to OFF.

D Turbine load will decrease by less than 3°/0AND reactor power will increase by approx. 3%.

The operator can stop dumping excess steam by taking the Quick Open Permissive Switch to OFF.

Justification I

CHOICE (A) - NO

'VJRONG Reactor power will increase VALID DISTRACTOR Some steam flow will divert to the cor>densrr Turbine load will decrease slightly CHOICE (E) -YES Turbine load will decrease slightly due to lowered steam presstire Reactor power will increase because of greater steam demand The valve can be closed by taking rmtroller to manual and reducing output.

CHOICE (C) - NO WRONG. Reactor power will increase.

VALID DISTRACTOR Turbine load will decrease slightly CHOICE (D) - NO WRONG. Quick Open Permissive Switch will not close the valve VALID DISTRACTOR: Turbine load will decrease slightly and reactor power will increase. The quick open permissive tdocks quick open to the atmospherics.

References

_I 1 MSS-00-C, "Main Steam System" Lesson. Rev.;ion 6 (711 1/01) (Pg 34 35 of 74) 2 RRS-01-C. "Reactor Regulating System" Lessc.7, Revision 3 ( 7 X O I 1 3 ARP-2590D-024. "CONDENSER BYPASS VALVE NOT (:LOSL-D". Revision 0 (2/12/04)

Piping Diagram 25203-26002. "Main Steam Turbine. Sheet 4 a: 5. Xzvision 20 (9117101) (J-21 5 Source. INPO Bank - Q# 21444 - Used at Braidwoort 1. 7:17/2002 NRC KIA SystemlElA System 045 Main Turbine Generator (MTIG, system Number A2 08 RO 2 8 SRO 3  ?+ CFP Link (CFR 41 5/43 5/45 31'45 5 )

Ability to (a) predict the impacts of the following rmalfunctiolis or operation qn the MT1G system and (b) based on those predictions use procedures to correct. control or mitigate the consequences of those malfunctions or operations Steam dumps are not cycling properly at low load, p r stick cptm at highei load (isolate and use atmospheric reliefs when ricrpssaryi NRC KIA Generic System Number RO SdO iFH Link

1. 50 I v RO v SRO Questicjn ID: 5000039 Origin: New 4 Memory Level The plant is at 85% power. Heater Drain JWmp'A' has been removed from service for maintenance on the pump. Given the following tagout boundaries, identify the correct component operation sequence to prevent overpressurization cf piping

1. CLOSE 'A' Heater Drains Pump Suction Valve 2-HD-?A

2. CLOSE 'A' Heater Drains Pump Minimum Flow Recirc. Isolation 2-HD-45A

3. CLOSE 'A' Heater Drains Pump Discharge 2-HD-YA A 1-3-2 B 3-1-2 C 1-2-3 D 2-1-3 Justification I CHOICE (A) - NO WRONG Suction valve is closed before discharge valve VALID DISTRACTOR Procedure directs discharge before recirc CHOICE (B) - YES Discharge valve must always be closed before suction valve to prevent overpressurization of suction piping CHOICE (C) - NO WRONG Suction valve is closed before discharqe valve VALID DISTRACTOR Procedure directs drain aftei isolarrcis CHOICE (D) - NO WRONG Suction valve is closed before discharge valve VALID DISTRACTOR This sequence would isolate and depressurize piping References 1

1 OP-2320 "Feedwater Heater Drains and Vents". Revisiori 16 i 121231113), Section 4 5 (Pg 21 of 46)

NRC KIA SystemIElA System Number RO SRO CFR Link NRC KIA Generic System 2.2 Equipment Control Number 2 2 13 RO 3 6 SRO 3 8 CFR Link (CFR 41 1 0 1 4 5 13)

Knowledge of tagging and clearance procedures

1. 51 I RO v SRO Questic>,iID: 0054226 Origin: Mod Memory Level The plant was operating at 100% when a ieactor trip ocriirred. Given the following conditions and events:

- 2 charging pumps are operating.

- 3 CEAs failed to insert.

- Boric Acid Isolation, 2-CH-514, will NO7 ,pen.

- Gravity Feed Isolations, 2-CH-508 and 509. will NOT open Which one of the following statements coirectly describes the procedure and required actions to be taken?

A Continue EOP-2525, "Standard Post Trip Actions" to determine if any other problems exist.

Maintain Tavg at or above 500°F B Refer to AOP-2558, "Emergency Boration" and emergency borate for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> by opening Boric Acid Flow Control Valve, 2CH-210Y.

C Continue EOP-2541 "Appendix 3 - Emergency Boration" and emergency borate from the 4 RWST.

D Refer to EOP-2540A, "Functional Recovery of Reactivity Control" and emergency borate using success path RC-3 (Boration using SI).

Justification i

EOP 2525 contingency for Reactivity Control CHOICE (A) - NO If the candidate thinks that stuck 5 CEAs will not pose a problem as long as Tavg remains above 500 F this answer is plausible However EOP 2525 diiects the operator to emergency borate using EOP 2541 appendix 3 CHOICE (B) - NO If the candidate thinks that referiliij to the ACD 15 permissible under these cricumstances the flow path will provide boric acid flow to the RCS The thurnhule rcqiirernmt I the AOP is to borate 1 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> for P x h I

,itlditional CEA stuck beyond 1 In this case if the AOP was used the requirement to borate would bc 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> not 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> The applicant would select 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> if they used the thumbrule for 3 CEAs stuck out and neglected to recall that one of the CEAs is already considered stuck out in the saefty analysis Note that an entry condition for AOP 2558 is CHOICE [C] - YES Correct answer EOP 2525 directs the applicant to emergency borate using EOP 2541 appendix 3 ilipre is no time limit provided under emergency conditions This boration path uses the charqing pumps for injection CHOICE [D] - NO If the candidate thinks that a failuie of the CVCS prir,iary emergency boration flow path is justification for refering to EOP 2540A they could possibly select this answer This is the correct success path if RC-1 (CEA Insertion) and RC-2 (Boration using CVCS) are not ,Jailable This success path uses the SI pumps not the charging pumps Note that RC-2 directs the operator to borate usinq EOP 2541 Appendix 3 References 1 AOP-2558 pages 3-9 2 EOP-2525 page 3 3 EOP-2541 Appendix 3 pages 1-5 4 EOP-2540A pages 8-12 NRC KIA SysternlElA System 029 Number RO SRO CFR !.ink NRC KIA Generic System 2 1 Conduct of Operations Number 2 1 23 RO '1 9 SRO 4 0 CFR Link (CFR 45 2 / 45 6)

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

1. 52 1 d RO d SRO Questiclti ID: 5000056 Origin: New d Memory Level Unit 2 was operating at 100% power wher; a spurious reactor trip occurred. Given the following events and conditions:

- Pressurizer level rapidly dropped to 20%

- The operators reach step 3b of EOP-2525 (Standard Post Trip Actions) "Determine Status of RCS Inventory Control" What is the reason for verifying that RCS SUBCOOLING is greater than or equal to +30"F in step 3b?

A To ensure that pressurizer level indication xcurately represents total RCS inventory d B To ensure that RCS pressure is under automatic control and returning to the normal band C To determine if manual control of the pressurizer level control system is required D To determine if a LOCA is in progress and the contingency action is required

~

Justification CHOICE [A] - Y E S This is the correct reason for validating that subcooling IS > 30°F in this step CHOICE [E] - NO The PLCS does not control pressurizer pressure RCS pressure control is verified in step 4 of EOP 2525 CHOICEIC] - NO This distracter is the correct answer for step 3a not step 3b CHOICE [D]

This is a correct reason for monitoring RCS subcooling during other steps in the EOPs There is no contigency action for step 3b References I 1 EOP-2525 page 8 2 EOP Tech Guide 2525 pages 10-14 N R C K/A SysternIElA System 028 Pressurizer (PZR) Level Control Malfunction Number AK305 RO 3 7 SRO 4 1 CFR Link (CFR41 5 4 1 10 145 6 145 13)

Knowledge of the reasons for the following responsps as they apply to the Pressurizer Level Control Malfunctions Actions contained in EOP for PZR level malfunction NRC KIA Generic System Number RO SRO CFR Link

1. 53 I fl RO 9' SRO Questinr! ID: 5000957 Origin: New Memory Level Unit 2 was operating at 100% power wher a large break LOCA (design bases) occurred. Given the following events and conditions:

(/ 3l 1! j-

- The SDC HX "A" RBCCW outlet va ve ( > RB-13.1Aj jarrimed shut and will not open on SRAS

- The Unit Substation Transformer24B+-++ . I secondary windings open due to a fault and bus 22E is deenergized Z4cl-l ) t 3 -2 Which one of the following statements corfectly decribes the effect of this failure on the containment spray system if repairs CANNOT be made?

A The initial containment pressure spikd will exceed design pressure and long term heat removal will NOT be adequate B The initial containment pressure spike will exceed design pressure but long term heat removal will be adequate C The initial containment pressure spike will not exceed design pressure but long term heat removal will NOT be adequate D The initial containment pressure spike will not exceed design pressure and long term heat *I removal will be adequate Justification The loss of SDC to the A train of containment spray will prevent that train from removing heat during long term sump recirc operations The initial pressure spike will be mitigated because both trains of CSS will inject from the RWST during the initial pressurization of containment Long term cqoling requires operation of one complete train of ESF equipment which includes 2 CAR fans plus one CSS train Loss of the 22E emergency bus will cause CAR fans F14A and F14C to lose power CAR fans F14B and F14D will run from bus 22F CHOICE [A] - NO Both trains of containment spray will inject and maiiitaiii the initial pressurf spike below the design thrrshold for containment Long term heat removal requires only 1 train of contaiiimetit spray and 2 CAR fans to remove sufficient heat CHOICE [B] - NO Both trains of containment spray will inject and maintain the initial pressure spike below the design threshold for containment Long term heat removal will be adequate with 2 CAR fans and train B Containment spray CHOICE [C] - NO Long term heat removal from containment will be .i<<equatewith 2 CAR fans and Train B of CSS CHOICE [O]- Y E S References 1 ECC-01-C rev 3 page 19 I

2 ECCOI figure 6 (M2105-11-98) 3 CSS-00-C rev 4 chg 1 pages 6-8.37 4 FSAR Figure 25203-3001 Main Single Line Diagram 1-10-92 5 RBC-00-C rev 5 pages 24-25 NRC KIA SystemIEIA System 005 Residual Heat Removal System (RHRS:

Number K306 RO 3 1' SRO 3 2' CFR Link (CFR 41 7 I 4 5 6)

Knowledge of the effect that a loss or malfunction of the RHRS will have oil the following CSS NRC KIA Generic System Number RO SRO CFR Link

1. 1. 54 1 gRO d SRO Question ID: 3 Origin: Bank Memory Level The plant is at 15% power with a startup and power ascension in progress when the in-service pressurizer pressure controller is inadvertmtly shifted to manual. All other pressurizer control components are operating normally for 15% power.

Which one of the following describes the e :pected effcct on the Pressurizer Control System as a result of the pressure controller now being in manual? (Assume NO operator action.)

A Spray valves will begin to open if RCS pressure should begin to rise.

B Backup heaters can ONLY-be energized or deenergized manually C Low Pressurizer Pressure alarm will NOT actuate regardless of how low pressure drops D Proportional heater output will rise ONLY if pressurizer level rises to -4% above program level 4 Justification I A - wrong: With the controller in manual, spray valves will NOT operate B - wrong: BU htrs are controlled by bistables not the controller, therefore will operate at the bistable setpoints regardless of the controller condition.

C -wrong; alarm is controlled by bistables. not controller D - Correct; Prop htr controlled by Controller only, except level insuiye .;=

3.6% above setpoint References 1 P L C - O l C R 3 pages 9-12, 21-24, 27-28, 30 NRC KIA SystemlElA System 010 zer Pressure Control System (PZR PCSi Number A 3 02 RO 3 6 SRO 35 CFR Link (CFR: 41.7 145 51 Ability to monitor automatic operation of the PZR PCS, including PZR pressure NRC KIA Generic System Number RO SRO CFR Link

1. 55 1 4 RO 4 SRO Question ID: 5000W9 Origin: New Memory Level Unit 2 was operating at 100% power when a reactor trip occurred due to a loss of main feedwater Given the following events and conditions:

- A and B SIG levels dropped to 15%

- The Turbine driven AFW pump was started Then, Vital DC bus DVlO was deenergized due to a fault Which one of the following statement correctly describes the status of:

1. the turbine driven AFW pump, and

2. the AFW FRVs?

A 1. The turbine driven AFW pump will trip and CANNOT be restarted from the control room

2. FW-43A fails open, FW-43B is unaffected B 1. The turbine driven AFW pump CANNOT be controlled from the control room but can be controlled locally

2. FW-43A and FW-43B fail open C 1 . The turbine driven AFW pump can be controlled normally on C-05

2. FW-43A and FW-43B fail open D 1 . The turbine driven AFW pump can be controlled ormally on C-05 4

2. FW-43A fails open, FW-43B is unaffected Justification I

The TD AFW pump can be started if vital DC powei rroni DV20 ib sclected Lsing the key lock switches on C-05 FW-33A fails open on a loss of DC power from DVIO FW-43B will not lose power and will operate normally CHOICE [A] - NO WRONG This answer is partially correct The TD AFW p i m p will riot trip This choice is provided if the candidate does riot recognize that the TD AFW pump has an alternate vital DC supply that is selected on C-5 Part 2 of this answer is i orrect CHOICE [E] - NO WRONG The TD AFW pump can be started normally from C-05 FW-4.36 Is not affected by the loss of DVlO CHOICE [C] - NO WRONG This answer is also partially correct FW-436 is powered from DV20 CHOICE [D] - YES CORRECT This is the correct answer FW-436 is not powered from DV-IO References 1 AFW-00-C rev 5 chg 3 pages 14 and 19 1

2 EOP-2525 rev 20 page 16 J AOP-2506A rev 3 page 23 NRC KIA SystemIEIA System 061 Auxiliary I Emergency Feedwater ,AFW) Svstern Number K6 01 RO 2 5 SRO 2 8' CFR Link (CFR 41 7 / 45 7 )

Knowledge of the effect of a loss or malfunction of ?he following will have orl the AFW components Controllers and positioners NRC KIA Generic System Number RO SRO CFR Link

1. 56 I d RO v SRO Question ID: 0065167 Origin: Mod Memory Level Unit 2 was operating at 100% power whe!, an elcctriczl transient occurred. Given the following conditions and events in sequence:

- VA-20 was deenergized

- The plant tripped

- MSI actuated

- A SGTR occurred on the #I Steam Generator (SG)

- Upon reaching step 6 of EOP 2525 (SPTA) the SPO was directed to feed the #2 SG using Aux Feed Water (AFW)

Current level in both steam generators is 5b%

Which one of the following statements correctly describes:

1 . the required actions, and

2. the correct procedure to be used.

A 1 . Actions: Place both AFW "OVERRIDEIMANISTARTI RESET" hand switches in "Pull-To-Lock" - Close 2-FW Feed #2 SG with the turbine driven AFW pump only.

2. Implement EOP Appendix 6 (TDAFW Pump Normal Startup)

B 1. Actions: Manually initiate Facility Two AFW components - Close 2-FW Feed #2 SG with the turbine driven AFW pump only.

2. Implement EOP Appendix 7 (TDAFW Pump Abnormal Startup)

C 1 . Actions: Place Facility Two AFW "OVERRIDEIMANISTARTIRESET" hand switch in "Pull-To-Lock" - Ensure #I AFW Reg. Valve is closed - Locally control #2 Aux FRV.

2. Implement EOP Appendix 6 (TDAFW Pump Normal Startup)

D 1. Actions: Manually start both MDAFW,' pumps - Place both AFW "OVERRIDEIMANISTARTI 4 RESET" hand switches in "Pull-To-Lock - Ensure #I AFW Reg. Valve is closed - Locally control #2 Aux FRV.

. 2. Complete EOP 2525 step 6 without starting the TDAF'JV pump Justification I VA-20 powers the actuation logic for facility 2 AFAS a i d the actuation relays are energize-to-actuate Loss of VA-20 means that facility 2 AFW components will have to be manually operated. 1-he turbine driven AFW pump should not be used if a SGTR is in progress to prevent radiological contamination The correct answer is to NOT start the TD AFW pump and close 2-FW-43A (AFW FRV to the # I SlGi to prevent feeding the ruptured SIG. #2 SIG should be fed using hoth electric AFW pumps only.

Bank question 0065167 asked the applicants what the correct sequence would be if VA-10 was lost This question was modified from losing VA-IO to losing VA-20 In addition. the previous question appeared to assume that a loss of V A - I 0

vould fail open the FRV to the #1 S/G. This is not correct - 1055 of I 3 W 0 causes 2-FW-43A to fail open This modified question uses the previous bank question but corrects the earlier pioblems with that revision Variations of the original distracters are used in the event that appllicants memmzed the answer to the bank question CHOICE [A] - NO WRONG This was the previously correct answer to qdestion 0065167 in the MP-2 bank - which was written as a loss of VA-I0 instead of VA-20. It is not clear if this answer was ever truly correct. However, this answer IS provided as a valid distracter for applicants who may have memorized t k bank question 'Jsi.ig the turbine driven AFW pump to feed the

$2 SiG when a SGTR is occuring is not recomrnenurii when both eleitiw: driven AFW pumps are fully functional Selection of appendix 6 would be appropriate for ct;ir!irig the TDAFW !?liiiip and is consistent with the first part of the answer.

CHOICE [B] - NO WRONG Although this would result in feeding the #2 s!G. there Woiilci be no reason to manually initiate facility 2 AFW components if 2-FW-44 (AFW header cross-connect) was closed 'n addition. using the TD AF'a pump during a SGTR is not recommended. If the applicant thought that tk.2 loss of VA-20 W O ~ J ~prevent V a normal start of thc TDAFW pump, then use of appendix 7 would be correct CHOICE [C] - NO WRONG This distracter is incorrect because there IS no reason to plaLe the facility 2 hand switch in pull to lock and feeding the #2 SIG with the TDAFW pump would c a s e radiological probierns - i.e. a release to the environment. Part 1 was an original distracter from the rev 1 version of is queftion Use of appendix 6 would be appropriate if the TDAFW r!id not lose control power - which it does not with 1 :fJS? of L,?~?O CHOICE [D] - YES CORRECT The #1 AFW Reg valve (2-FW-43A) rema!ns fully furicllunal despite a loss of VA-20 This valve would fall open if DV10 was lost - which appears to be the p r e v w s correct an5wer :c the bank question. Facility 2 AFW

1 ornponents would have to be manually operateti hecwse their iicti iancii relay was deenergized whpn VA-20 lost power References I 1 AFW-00-C rev 5 chg 3 pages 33 and 35 2 EOP 2525 rev 20 page 16 3 AFW-00-C Figures 1 and 2 NRC KIA SystemlElA System 061 Auxiliary / Emergency Feedwater (AFW) System Number A205 RO 3.1' SRO 3 4 ' CFRLink (CFR: 4 1 . 5 / 4 3 . 5 / 4 5 3 / 4 5 . 1 3 )

Ability to (a) predict the impacts of the following malfuiictions or qxrations on the AFW: and (b) based on those predictions, use procedures to correct, control. or mitigate the consequences of those malfunctions or operations:

Automatic control malfunction NRC KIA Generic System Number RO SRO CFR Link

1. 57 I v RO d SRO Question ID: 5000060 Origin: New q Memory Level Which one of the following conditions will sdomatically terminate a liquid release from the AWMT (does NOT require manual operator action io stop the discnarge)?

A Reduction in CW flow below authorized dilution flow limit in discharge permit B Increase in AWMT sample pump discharge pressure above PlOPs setpoint limit C Reduction in AWMT sample pump discharge flow below PlOPs setpoint limit *I D Loss of power to the discharge flow recorder Justification

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I PlOPs monitors the discharge flow rate and will automatically terminate the discharge if the flow rate exceeds a high or low setpoint WRONG (A) - NO CW flow provides the dilution flow for the discharge 1 he reqdired a nout of dilution flow is stated on the discharge rfrrnit This flow is not monitored by PlOPs and requires operator x t i o n 13 terminate the discharge if it is exceeded NRONG [B] - NO Disharge pressure is monitored in PlOPs but only a low discharge pressure will cause an automatic isolation of the discharge path A high pressure does not automatically stop the discharge CORRECT (C) - YES Discharge flow rate is monitored by PlOPs and low flow will automatically terminate the discharge WRONG [D] - NO A loss of power to the flow recorder requires operator action to terminate the discharge There is no automatic action associated with a loss of power to the flow recorder 1

References ALR-04-C rev 3 page 10 I

2 RLD-04 C rev 1 pages 6 18 NRC KIA SystemlElA System 059 Accidental Liquid Radwaste Reie.tse Number AA206 RO 3 5 SRO 3 8 CFR Link (CFR 43 5 / 45 13)

Ability to determine and interpret the following as they apply to the Accidental Liquid Radwaste Release That the flow rate of the liquid being released is less than or equal to that specified on the release permit NRC KIA Generic System Number RO SRO CFR Link

1. 58 1 *I RO 9 SRO Question ID: 0054038 Origin: Bank Memory Level Unit 2 was operating at 100% power. Given the following events and conditions:

- SP-2601 D (Power Range Safety Channel an.1 Jelta-T Power Channel Calibration Surveillance Procedure) is in progress

- Reactor Protection System (RPS) Channel 'A' High Power, TM/LP and LPD trip units are bypassed

- RPS Channel 'C' linear power range drawer hiah voltage power supply fails Which one of the following statements corrzctly describe; the expected plant response based on the stated conditions with NO further operator sitions?

A The plant will trip if EITHER CH 'B' or 'D' processes a trip that receives an input from the *I linear power range Nls.

B The plant will trip immediately because the 2 out of 4 zoincidence logic has been satisfied C The plant will trip ONLY if BOTH CH 'B' and 'D' process trips that receive an input from the linear power range Nls.

D A partial trip will be processed because of :he opening of ONLY two reactor trip circuit breakers.

Justification I

CHOICE [A] - YES CORRECT Module RPS-01-C, section I1 D 7 c states that channel byptss keys operate relays that bypass the trip unit relay contacts Thus, Channel "A'in the above question does not contribute to 2/4 coincidence logic. OP 2380. Rev. 8.

Scction 4 4. Part B 3 d.2.c states that a high voltage power failure will cause a single channel trip (Channel " C " ) Thus, only one more channel trip ( " Bor "D") is necessary to meet 214 coincidencc logic.

CHOICE [B] - NO

'ViRONG With channel A bypassed, the logic ladder 15 in 2!3 channels A trip in channel C provides only 1 of 2 process inputs required.

CHOICE [C] - NO

'WRONG Channel A has input 1 trip signal. All that is required is on-?more trip signal, not two more trip signals CHOICE [D] - NO LVRONG A partial trip will not occur with channel A bypassed References 1 RPS-01-CR6 page 14 - 16 I

NRC K I A SysternlElA System 012 Reactor Protection System Number A3 02 RO 3 6 SRO 3 6 CFR Link (CFR 41 7 / 45 5)

Ability to monitor automatic operation of the RPS including Bistables NRC KIA Generic System Number RO SRO CFR Link

1. 1. 59 j d RO d SRO Questim ID: 50OUCl61 Origin: New Memory Level Unit 2 is shutdown in a refueling outage wi!h maintenance being conducted on DC Bus 201A. Given the following events and conditions:

- "125 VDC LOAD CENTER 201A TROUBLE" annunciator lit (C-08, A-21)

- 201A Battery voltage = 125 volts

- DC Bus 201A voltage = 133 volts

- Charger 201A/DC1 output voltage = 134 volts Which one of the following statements correctly describes the cause of this problem?

A The PEO has taken the DS-I open pole detection circuit to the FU-A position B Battery charger 201A/DC1 output has failed C A ground has developed on 125 VDC bus 201A D Battery disconnect D S - I has been opened d Justification I

CHOICE [A] - WRONG The DS 1 open pole detection circuit test swtich will ,101 cause this dnnunciator to actuate This test swtich is operated routinely by PEOs to verify battery voltage is equal to DC bus voltage This switch would be used to verifv the voltage ilifference between the battery and the battery bus CHOICE [B] WRONG Although annuciator C-08 A-21 would actuate the DC voltage output from the battery charger would not equal 134 VDC i f the charger had failed This could be selected if th 2 applicant thougli that the charger output would remain at 134 VDC i f the charqer failed sure to location of the tap for the voltage output CHOICE [C] -WRONG A ground on the DC bus would not actuate C 08 A 21 It wodd cause d different alarm to actuate This could be w k c t e d if the applicant did not understand the grtound detection circdit CHOICE [D] -CORRECT Opening battery disconnect DS-1 would cause this alarm if battery voltage was greater than 6 volts different between the battery and bus 201A Battery voltage is 8 volts lowrr than bus voltage Objective LVD-00-C R5 PEO Ob] 9 State the purpose and describe the operating characteristics including automatic functions and interlocks of the followinq major 125 VDC Electrical Distribution System components a s qiven in LVD-00 C (MB-000391 Ai Battery Charger B) Battery C ) Battery Disconnect D) Battery Fuses E l Ground Detector F ) 125 VDC Breakers G )Open Pole Detector H ) Kirk Key Interlocks References 1 LVD 00-C R5 C H I pages 13-14 2 h12104-09-98 87000443 Rev 1 (Figure 2) 3 ARP 2590F Rev 7-04 annuciator C-08 A-21 NRC KIA SysternlElA System 063 DC Electrical Distribution Systenr Number A3 01 RO 2 7 SRO 3 1 CFR Link (CFR 41 7 145 5)

Ability to monitor automatic operation of the DC elpctriLal system including Meters, annunciators dials recorders and indicating lights NRC KIA Generic System Number RO S RO CFR Link

1. 60 I 9 RO 9 SRO QuestiLrn ID: 0054057 Origin: Mod Memory Level Unit 2 is operating at 100% power when a plant trip occLrs concurrent with a loss of the RSST.

Given the following events and conditions.

- Instrument Air pressure drops slowly to 89 psig immediately following the trip and does not recover Which one of the following statements correctly describes.

1. The preferred method for restoring Unit 2 Instrument Air, and

2. The reason for selecting this method?

A 1. Cross-tie Unit 2 instrument air with ilnit 3 instrument air

2. To maintain air purity in the instrument air system and prevent introduction of contaminants B 1. Cross-tie Unit 2 station air with Unit 3 service air, then cross-tie Unit 2 instrument air with *I station air

2. This provides the simplest and fastest way to repressirrize instrument air C 1. Cross-tie Unit 2 instrument air with Unit 2 station air

2. This provides the simplest and fastest way to repressurize instrument air D 1. Cross-tie Unit 2 station air with Unit 3 service air. then cross-tie Unit 2 instrument air with station air

2. To maintain air purity in the instrument air system and prevent introduction of contaminants

~~~~

Justification I CHOICE [A] WRONG There is no direct physical connection between the Unit 2 and Unit 3 instrument air systems. In order to cross-tie with Unit 3, the operators must cross-tie Unit 2 station air wth Unit 3 s e r x e air - then corss-tie Unit 2 station air with Unit 2 instrument air. However, if there was a direct connection. the reason would be valid.

CHOICE [B] -CORRECT EOP 2525 step 19 requires cross tying Unit 2 to Unit 3 instrdment air systems and cross tying unit 2 station air to instrument air. This is the simplest and fastest method of repressurizing instrument air - but it introduces contaminants into the system because service air and station air do iiot have the same qir quality requirement as instrument air CHOICE [C] - WRONG Unit 2 station and instrument air systems are automatically cross-tied whenever instrument air pressure drops below 85 psig and can be manually cross-tied if needed but the Unit 2 station air compressor will not have power if the RSST fails - powered from bus 22C. The reason is valid IF the station air compressor had power.

CHOICE [D] -WRONG Partially correct - cross tying with unit 3 is the corred action to rake. but the station air system does not have the same level of air purity associated with instrument air The reason - "to maintaiii air purity. " is incorrect ISA-00-C lesson material contains a discussion of an occurrence under CR M2-97-2526 wnere operations cross-tied for approx 6 h o u r s Based on this occurrence, the operating procedure (OP-2332A) was modified to include a caution regarding minimizing cross-tie duration to prevent moisture buildup in the containment portion of the instrument air system.

References 1 EOP 2525 rev 20 step 19. Contingency Actions for loss of IA 2 EOP 2528. Step 9 Contingency Actions 3 ISA-00-C Rev 6 pages 11-16, 51

.I. FPS-04-C rev 3 chg 1 page 16 5 Instrument I Station Air Figure 1 M2 10/07/02 870~1262 6 EOP 2525 Technical Guide page 35 step 19 N R C KIA SystemlElA System 065 Loss of Instrument Air Number AK3 04 RO 3 1 ' SRO 3 2 CFR Link (CFR 41 5,41 10 I 4 5 6 145 13)

Knowledge of the reasons for the following responses as they apply to the LOSS of Instrument Air Cross-over to backup air supplies N R C KIA Generic System Number RO SRO CFP Link

1. 61 I fl RO *I SRO Question ID: 0054834 Origin: Mod fl Memory Level Unit 2 is'shutdown in MODE 6 moving spent fuel from the containment to the spent fuel pool Given the following events and conditions:

- Containment purge is in operation

- Spent Fuel Pool and Radwaste Ventilatio,~systems are aligned to exhaust to the Unit 2 stack.

- RM-81238 (Containment Gaseous Activitvj alarms due to high activity Which one of the following changes will occur automatically within the Containment Purge and/or the Main Exhaust systems in response to this condition?

A Fuel Handling Area Supply Fan (F-20) trips and Fuel Handing Area Isolation Valves 2-HV-165. 2-HV-170 and 2-HV-171 close B Purge supply fan (F-23) trips C Containment Purge Isolation Valves 2-AC-4. 2-AC-5, 2-AC-6 and 2-AC-7 close fl D Main Exhaust System Makeup Air Damper (2-AC-59) closes

~ ~~~

Justification CHOICE [A] -WRONG Main exhaust fans do not trip if RM-8123E alarms This occurs if an AEAS actuates in the spent fuel pool area CHOICE [B] -WRONG Piirge supply fan F-23 does not trip. If it did trip, it would stop or substantially reduce the flow of containment air through thc purge exhaust system.

CHOICE [C] - CORRECT Tlicsc purge exhaust valves will close This correct answer was modified from the original bank question CHOICE [D] -WRONG The hank question answer was that 2-AC-59 OPENFD - which is correct The old right answer was changed to 2-AC-59 CLOSES It is plausible to expect supply valves to close when a purge niust be stopped - but this valve is in the main exhaust system not the purge system.

References 1 RMS-00-C Rev 6 pages 30-33 2 RWV-00-C Rev 4 page 8. 20-22 NRC KIA SystemIEIA System 073 Process Radiation Monitoring Number K401 RO 4 0 SRO 4 3 CFR Link (CFR 41 7)

Knowledge of design feature(s) andlor interlocks which provide for the following Release termination when radiation cxceeds setpoint NRC KIA Generic System Number RO SRO CFR Link

1. 1. 62 I 4 RO fl SRO Ques?ron 13: 0053371 Origin: Mod 4 Memory Level The plant is operating at 100% power when you receive a !:igh radiation alarm on RM-8132A (Unit 2 Stack Particulate).

Given the following events and conditions:

- When checked on the PPC and RC-14, RM-8132A is found to be reading 7.5E04 cpm and stable

- An air sample by Health Physics confirms that this is a valid alarm

- HP recommends revising the RM-8132A rnodule setpoint(s) on RC-14 Using the provided attachment, which one of the following describes what must be done with the RM-8 132A module setpoints?

A Raise the alarm setpoint Do NOT change the alert or fail setpoints B Raise the alarm setpoint and the alert setpoint Do NOT change the fail setpoint C Raise the alarm, alert, and fail setpoints 4 D Do NOT change the alarm setpoints Justification RM-8132A is a process monitor listed in All 2 of OF2383C Rev 12-02 Raising the setpoints is allowed for all monitors in Att 2 CHOICE [A] -WRONG The &rrn setpoint must be raised by 2x not 1 5x The alert and fail setpoints must also be raised CHOICE [B] -WRONG The fail setpoint must also be raised to 1/5 of the new average reading CHOICE [C] -CORRECT Rhl 81 32A is a process monitor listed in Att 2 of OP2383C Rev 12-02 Raising the setpoints is allowed for all monitors i i i Att 2 The fail setpoint is raised to 1/5 of 7 75E04 The ?!?rt and alarm setpoints are raised to 1 5 and 2 0 x the new averaqe value CHOICE [D] -WRONG If RM-8132A was listed under All 1 of OF2383C this would be the correct answer This was previously the correct iiiiswer to the bank question which tested RM-81238 under similar circumstances This would also be a correct answer i f thP plant was undergoing a transient References I 1 OP2383C. "RADIATION MONITOR ALARM SETPOINT CONTROL", Revision 12-02 (08126104) (Fg 3 . 17-19, 29 of 32) 2 RMS-00-C Rev 6 chg 1 pages 79-80 7 Provide applicants with Attachments 1 and 2 of OP-2383C "RADIATION MONITOR ALARM SETPOINT CONTROL' (Pdyes 28 29 of 32)

_________ _____ ~ ~~ ~_____ ~

NRC KIA SystemlElA System 073 Process Radiation Monitoring (PRM) System Number A402 RO 3 7 SRO 3 7 CFR Link (CFR 41 7 145 5 to 45 8 )

Ability to manually operate andlor monitor in the r m t r o l loom Radia' I n iionitoring system control panel NRC KIA Generic System Number RO SRC CFF! Link

1. 63 I fl RO 4 SRO Question lu: 5000Cl13 Origin: New Memory Level Unit 2 had started up following a refueling outage. Given !he following events and conditions:

- The upper extension shaft on CEA-01 (rod in the center of the core) was not re-coupled following the reactor refueling

- The reactor is currently operating at 100% power Which one of the following statements correctly describes:

1. the CEA position indication for this condition. and

2. the predominant flux distribution concern (if any) if operation at 100% power continues?

A 1. The amber light will remain lit on the core mimic

2. Axial flux will be suppressed at the bottom of the ccre B 1. The amber light will remain lit on the core mimic

2. Radial flux peaking will occur in the center of the core C 1. CEA position indication will show that CEA-01 is fully withdrawn 4

2. Radial flux peaking will occur at the outer core regions D 1. CEA position indication will show that CEA-01 is fully withdrawn

2. There will be no abnormal flux distribution in the core Justification I

The upper extension shafl contains the magnet that the RPIS reed switches If the upper extension shafl is not coupled to the CEA there will be no direct indication of the problem because the upper extension shaft will actuate the reed switches as it is withdrawn CEA XX IS in the center of the core and the flux will be locally depressed by this rod remaining inserted there by causing a radial flux ddstribution problew CHOICE [A] - NO WRONG The amber rod bottom light will not remain lit If the applicant does not understand the difference between axial and radial flux suppression, the applicant may select this distracter because it is the only distracter that states axial flux is suppressed in the botlom of the core CHOlCE[B] NO WRONG - The amber rod bottom light will not rema'q lit CEA position indication will appear as if the rod IS fully withdrawn - but the rod remains fully inserted Radial flux suppression occurs in the center of the core causing radial flux peaking in the outer regions of the core CHOICE [C] - Y E S CORRECT- The amber rod bottom light will not be lit and there will be no direct indication that the CEA remains fully inserted The CEA will appear fully withdrawn on the core mimic PPC and CEAPS displays Radial flux will be suppressed in the center of the core - and will therefore peak in the outer $egionsof the core If the candidate does not rinderstand flux peaking and thinks (hat the reed switches are actuated bv the CEA and not the upper extension shaft the applicant could pick this distracter CHOICE [D] - NO WRONG - This choice is plausible if the applicant thinks that uncoupling the upper extension shaft does not prevent the control rod from being withdrawn References I 1 CED-01-C rlc04 pages 26-28, 53, 56 2 Figures 23 and 24 M2 04-02-98 86001050151 3 Core Mimic - Figure 23 N R C KIA SysternIEIA System 014 Rod Position Indication Systev. {RPIS)

Number A1.04 RO 3.5 SRO 3 8 CF!! Link (CFR: 41 5/45 5)

Ability to predict andlor monitor changes in parameters (to prevent exLeeding design limits) associated with operating the RPlS controls, including: Axial and radial power distribution N R C KIA Generic System Number RO SRO CFR Link

1. 1. 64 I [?I RO SRO Question ID: 5000065 Origin: New 4Memory Level Unit 2 is in MODE 5 preparing to commence a refueling outage. Given the following events and conditions:

- A containment purge is being lined up in accordance with OP-23148, "Containment and Enclosure Building Purge" using the containment cleanup mode

- Noble gas concentration inside containment exceeds the limits in OP-23148 Which one of the following statements correctly describes the purge path required?

A Containment purge flow will be directed through the EBFAS system through 2-AC-3 (E6 z PURGE SUPPLY DMPR)

B Containment purge flow will be directed through the EBFAS system through 2-AC-57 (CTMT [I PURGE EXH DMPR)

C Containment purge flow will be directed through the main exhaust system through 2-AC-3 (EB L; PURGE SUPPLY DMPR)

D Containment purge flow will be directed through the main exhaust system through 2-AC-57 0 (CTMT PURGE EXH DMPR)

Justification CHOICE [A] - CORRECT 1

If noble gas exceeds 1 25E2 Cilcc, then purging through the EBFAS system is required and 2-AC-57 must be red-tagged shut CHOICE [E] -WRONG Partially correct - containment purge flow IS directed through the EBFAS system but 2-AC-57 must be red tagged shut The path through the EBFAS system IS correct but 2-AC-57 IS not in this path CHOICE [C] -WRONG Partially correct - containment purge IS not directed through the main exhaust system but AC-3 IS used to control the release flow rate CHOICE [D] -WRONG Containment purge is not directed through the main exhaust system if noble gas concentration exceeds the limits in OP-23148 but if the main exhaust was used, AC-57 is the correct exhaust damper This is the normal purge release path for containment References 1 OP23148, 'Containment and Enclosure Building Purge", Revision 19-03 (9/17/04) (Pg 9,lO of 63) 2 RWV-00-C rev 4 page 11 3 RWV Figures 1, 2, 7

~~ ~ ~

NRC KIA SystemIEIA System 029 Containment Purge System (CPS)

Number A401 RO 2 5 SRO 2 5 CFR Link (CFR 41 7 I 4 5 5 to 45 8)

Ability to manually operate andlor monitor in the control room Containment purge flow rate NRC KIA Generic System Number RO SRO CFR Link

1. 65 1 d RO d SRO Question ID: 0071625 Origin: Bank d Memory Level Unit 2 was conducting refueling and a full core reload :s ii: progress when a design-basis earthquake occurs. Given the following events and following conditions:

- A spent fuel assembly had been lifted by the refuel machine and moved beyond the reactor vessel flange

- The upender was on the refuel pool side and empty

- The refuel machine operator noticed the level in the reactor cavity is lowering at a rate of about 5 inchedminute

- The location of the leak was reported to be from the reactor cavity pool seal

- AOP 2578 (Loss of Refuel Pool and Spent Fuel Pool Level) was entered

- The refueling SRO has directed the refueling machine operator to place the spent fuel assembly at the PRE-PROGRAMMED SAFE POINT and evacuate containment In accordance with AOP-2578, which of tlle following choices correctly describes necessary actions to comply with the directions of the refueling SRO?

A Transport to the designated safe point in the CORE and ungrapple the fuel assembly B Transport to the designated safe point in the NORTH SADDLE AREA of the refuel pool and ungrapple the fuel assembly.

C Transport to the designated safe point in the SFP REGION B STORAGE RACKS and do NOT ungrapple the fuel assembly.

D Transport to the designated safe point in the SOUTH SADDLE AREA of the refuel pool and vl do NOT ungrapple the fuel assembly Justification 1

CHOICE [A] -WRONG Returning the spent fuel assembly to the core is not :hc re.!l,ired act or1 pei AOP 2578 unless the assembly has not yet been moved beyond the reactor cavity flange In a r y case. the fuel asqembly should not be ungrappled This action was the correct answer to one version of the bank question.

CHOICE [B] -WRONG The north saddle area is not the quickest way to isolate the leak ant: restuw the bundle. The upper guide assembly is stored in the north saddle area. The safe point is in the south saddle crea. In any event, the fuel assembly should not be ungrappled.

CHOICE [C] - WRONG With the upender on the refuel pool side. the priority I S In close the refiAing canal gate. Transfering the fuel assembly to the spent fuel pool would only delay this action CHOICE [D] - CORRECT Ttiis is the correct action per AOP-2578 Tliis bank question has been used repeatedly with cc,.ect answers .?s .A (store in the SFP) and B (store in the core)

Storing the fuel assembly in the south saddle area is not une of the c0rre.t answers among the bank questions However. the programmed "safe point" for the refuel;liy machine IS the south saddle area This is where the fuel assembly would be stored if there was not enough time to put the assembly back in the core or move it to the spent fuel pool The purpose of this question is to determine if the applicants undersiand when it would be appropriate to store the fiiel assembly at the south saddle safe point Note: There is a precaution in AOP-2578 that statcs "3 5 A refuel cavity drain line failure in the south sar'dle w o u l ~rlr? i- inat wea completely. This eliminates this area and the transfer canal as a safe storage location."

The south saddle are has only 1-2 ft of water over the fuel bundles to protect and shield them if water drops to the level of the reactor cavity flange - this is not an appropriate storage location wI-~er,the location of the leak is from the south cavity drain This selection has never been a correct mswer for any prior ;/ersion of this bank question AOP-2578 tasks the refueling SRO with the followins priorities in rhe .NCI~ to a loss of refueling water level, NOTE The Refueling SRO is tasked with assessing water level loss rate time available to perform actions, and potential safe locations for rractor l w e i internal components not in the reactor. If worst case .r:ak iates are present, the following priorities are provided and sho;,'d be performeu 11:

parallel:

1 Ensuring Transfer Carriage is in the SFP side

2 Closing Transfer Tube Isolation. 2-RW-280 3 Lowering an irradiated fuel assembly in the refutiing inachiiie ,ntu the reactor vessel only if sufficient space is available to allow quick insertion of the fuel assembly [ e reload vs shuffle) 4 Lowering an assembly in the refueling machine into t4e south saddie 5 Lowering an assembly from the platform crane into a 3FP storage location References 1 AOP-2578, "Loss of Refuel Pool and Spent Fuel Pool Level", Revision 6 (05/23/03) Step 4 1 9 (Pg 8 9,lO of 15) 2 REF-04-C rev 3 pages 48-50 NRC KIA SystemlElA System 034 Fuel Handling Equipment System (FHES)

Number A1 02 RO 2 9 SRO 3 7 CFR Link (CFR 41 5 I 4 5 5)

Ability to predict andlor monitor changes in parameters (to prevent exce+edingdesign limits) associated with operating the Fuel Handling System controls includinq Water l w e l in the refueling canal NRC KIA Generic System Number RO SRO CFR Link

1. 66 1 tf RO 14SRO Question ID: 5000069 Origin: New Memory Level Unit 2 is conducting a reactor start up. Given the following events and conditions:

- Wide range (WR) logarithmic nuclear instrument (NI) channels C and D are out of service

- The reactor is not yet critical

- The ECP expected critical rod height is 100 steps on Regulating Group 6

- Regulating Group 4 is withdrawn to 60 steps

- WR NI Channel A failed low WRL NI Channel A c1.OE-I CPS WRL NI Channel B 6.2E2 CPS Which one of the following statements correctly describes the required action (if any) required to comply with TECHNICAL SPECIFICATIONS?

A Immediately trip the reactor.

B Insert all control rods and shutdown the reactor.

C Stop the startup until WRL NI Channel A has been repaired. NO other actions are required.

D Immediately ensure adequate shutdown margin. 9 Justification I

Tech Spec 3 3 1 requires 2 channels of WR Nl's to be operable in MODES 3,4 and 5 The reactor does not reach MODE 2 until group 4 CEAs are withdrawn to 92 steps Tech Spec LCO 3 3 1 1 applies under these conditions and requires immediate determination of adequate shutdown margin There are no other Tech Spec requirements that apply to this case CHOICE [A] - NO WRONG Tech specs require 2 channels of WRL Nl's to be oprable in MODE 3 Tnpptng the reactor is not required -

only determining adequate shutdown margin CHOICE [B] - NO WRONG Tech specs require 2 channels of WRL Nl's to be operable in mode 3 Immediately determining shutdown margin is required w t h the reactor in modes 3 CHOICE [C] - NO WRONG - Tech Spec 3.3.1.1 requires 2 WRL NI channels to be operable in mode 3. Stopping the startup may be a prudent action but it is not required by Tech Spec 3.3.1. l .

CHOICE [D] - YES CORRECT The reactor has not yet transitioned to mode 2 and 2 WRL NI channels are required in mode 3 With only 1 WRL NI channel operable, the tech spec action is to immediately determine shutdown margin References 1 OP-2202 rev 20-05 pages 4, 7-8,17. Att 1, A t l 4 1

2 NIS-01-C pages 9-16, 30-31 3 ARP C-04 AB-12 NRC KIA SystemIEIA System 033 Number RO SRO CFR Link NRC KIA Generic System 2.2 Equipment Control Number 2.2.22 RO 3.4 SRO 4.1 CFR Link (CFR: 43.2 I45.2)

Knowledge of limiting conditions for operations and safety limits

1. 67 1 d RO 9 SRO Questicrl ID: 5 0 o ~ n 7 0 Origin: New Memory Level Unit 2 is preparing to discharge the contenls of the CWMT :!lough the liquid radwaste system. Given the following conditions and events:

- The discharge tank contains 0.1 pCilml of Tritium as the primary isotope Which one of the following statements correctly describes:

1. The biological hazard presented by the contents of the tank

2. The administrative controls that ensure the activity iimits of 10CFR20 will NOT be exceeded?

A 1 . Tritium emits beta radiation and primarily presents an internal exposure hazard

2. The REMODCM limits the total smoun! -f Tritiui3t that can be released B 1 . Tritium emits gamma radiation and primarily presents a whole body exposure hazard

2. The NPDES limits the maximum volume of any radioactive discharge to 5000 gallons C 1 . Tritium emits beta radiation and primarily presents dn internal exposure hazard

2. The NPDES limits the maximum volume of any radioactive discharge to 5000 gallons D 1, Tritium emits gamma radiation and primarily presents a whole body exposure hazard

2. The REMODCM limits the total amount of Tritium that can be released Justification

_I CHOICE [A] - CORRECT Tritium decays by emission of a low energy beta particle which only presents an internal exposure problem to the human

!mdv TRM 3/4 11 limits the total activity of Tritium thnt can be released to ensure that the limits of 10CFR20 are not exceeded.

CHOICE [E] -WRONG Tritium emits a low energy beta particle - not a gam;lia N -.'LIES does not limit that total amount of activity that can be discharged - only limits other biological threats to tbc environment like chemicals and temperature of disharge water CHOICE [C] WRONG Partially correct - Tritium does emit a low energy bc'a prticlc whir+ preseqis an internal hazard NPDES does not limit the amount of activity that can be dicharged to ensure ll)CFRZ!l is riot err-eeded CHOICE [D] -WRONG P,irtially correct - Tritium does not emit gamma radiation and does not pre4ent a problem to whole body exposure However Tech Spec 3 11 1 is the guiding limlt to prevent exceeding the qGCFR20 limits References 1 RLD-04-C rev 1 pages 25-26 2 TRM 3/4 11 NRC KIA SystemlElA System 068 Liquid Radwaste System (LRS)

Number K5 04 RO 3 2 SRO 3 5 CFR Link (CFR 41 5 / 45 7)

Knowledge of the operatlonal implication of the follcwing 22 ,cepts as they apply to the Liquid Radwaste System Biological hazards of radiation and the resulting godl of ALARA NRC KIA Generic System Number RO SRO SF;!Lmk

1. 68 I 4 R o *ISRO Question ID: 5000071 Origin: New Memory Level Unit 2 was operating at 100% power in the process of conducting a gaseous waste discharge. Given the following events and conditions:

- The operators are releasing the contents of waste gas decay tank (WGDT) A in accordance with SP-2617B (Gaseous Waste Discharge)

- RM-9095 (Waste Gas Radiation Monitor) lost power 3 minutes after the release had started Which one of the following statements correctly describes:

1. The complete list of automatic actions that will occur, and

2. The actions required to restart the release7 A 1. 2-GR-37.1, 2-GR-37.2 (waste gas discharge isolation valves) AND 2-GR-8.1A (waste gas decay tank's outlet valve) will close.

2. The waste gas discharge may be restarted provided 2 independent samples of the WGDT have been taken and analyzed.

B 1. 2-GR-37.1, 2-GR-37.2 (waste gas discharge isolation valves) AND 2-GR-8.1A (waste gas decay tank's outlet valve) will close.

2. The waste gas discharge may not be restarted until RM-9095 has been restored to an operable status.

C 1. ONLY 2-GR-37.1 AND 2-GR-37.2 (waste gas discharge isolation valves) will close.

2. The waste gas discharge may be restarted provided 2 independent samples of the WGDT have been taken and analyzed.

D 1. ONLY 2-GR-37.1 AND 2-GR-37.2 (waste gas discharge isolation valves) will close.

2. The waste gas discharge may not be restarted until RM-9095 has been restored to an operable status.

Justification I CHOICE [A] CORRECT The waste gas discharge isolation valves and the isolation vale to the WGDT being released will automatically close RM-9095 is not required to be operable to conduct a waste gas release - but 2 independent sample of the WGDT contents must be taken and verified before the releas? can be restarted CHOICE [B] -WRONG RM-9095 is not required to be operable to conduct a waste gas release CHOICE [C] -WRONG 2-GR-8 1A also closes automatically in additon to 2-GR-37 1 8 2 Partially correct - the release may be restarted wlthout RM-9095 being operable provided the WGDT is independently sampisd by 2 people CHOICE [D] -WRONG 2-GR-8 1A also closes automatically in additon to 2-GR-37 1 8 2 RM-9095 is not required to be operable to conduct a waste gas release - but 2 independent sample of the WGDT contents must be taken and verified before the release can be restarted References 1 SP-26178 rev 11 page 3 2 RMS-00-C rev 6 pges 40-41 3 GRW-04-C Rev 513 pages 13-14. 17 NRC KIA SystemIEIA System 071 Waste Gas Disposal System (WGDS)

Number AZO5 RO25. SRO26 CFRLink (CFR 4 1 5 / 4 3 5 / 4 5 3 : 4 5 1 3 )

Ability to (a) predict the impacts of the following rnalfuroctionsor opLrations on the Waste Gas Disposal System and (b) based on those predictions. use procedures to correct control or rrtitio r!e the consequences of those malfunctions or operations Power failure to the ARM and PRM Sy'tems NRC KIA Generic System Number RO S RO CFR Link

1. 69 I d RO d SRO Question ID: 0056hlti Origin: Mod Memory Level Unit 2 was operating at 100% power. Giveil the following events and conditions.

- A reactor trip occurs at 0200 and the operators enter EOP-2525, "Standard Post Trip Actions"

- At 0210, several rad monitors are rising, including RM-4299 C (main steamline radiation monitor),

which slowly increases from 1.OE-2 to 1.5EO Rlhr Which one of the following statements corrLctly describes ine cause of this trend?

A A SGTR has occurred in the "B" SIC- a: 02 10 B A SGTR occurred on the "6" SIG prior to the trip 21 0200 C A large crud burst has occurred in the H C S at 0200 D Severe fuel damage has occurred at 021 0 4 Justification I CHOICE [A] - WRONG Thc main steamline radiation monitors respond pri narily to N i 6 gamma w i c h decays away after the reactor trip The N16 source term has decayed away 10 minutes aftL. the trip The applicart could pick this answer if tielshe did not rccognize that the main steam line rad monitors resp-rlo priirliarily t~ high energy gamma radiation RCS coolant would niiqrate into a ruptured SIG through the broken tube t d , onlv the gaseous activity would be transported down the main i t e m lines Typical RCS gaseous activites would nrt be siufficient LO cause the main steam line RAD monitor to alarm CHOICE [E] WRONG Applicants are trained to use the MSL RAD monitor. to di,.j lose ri S G TR so some may recall that the MSL monitors inspond to N16 if the SGTR occurs before the trip i t i e v wo~iltlbe iilLc)rrect for this condition because the NIG source Imn has decayed away after 10 minutes and the R ;i m s r o i i s actiwty ievels in the 'A" S/G are insufficent to cause RM-4299C to respond CHOICE [C] WRONG A large crud bust does not contain volatile radionucliL*-s ancl the acti:ltv from the crud burst will not migrate into the main strarn lines Particulate activity in the SIG will remain the S:G and not enter the main steam lines Sonie applicants inay not recognize that only the gaseous activity enter +liesteam linss CHOICE [D] - CORRECT R M 4299C will also respond to shine from Containmi qt (streaming throiq11steam line penetrations) during severe accidents with fuel damage RM-4299C IS used as arl nlt:rnate Cin+-inmenthigh range monitor because of this response References

-_I RMS-00-C rev 6 page 56-57 NRC KIA SystemlElA System 072 Area Radiation Monitoring (ARM' Svs, n Number K1 05 RO 2 8' SRO 2 9' CFR Link (CFR 41 2 to 41 9 ' 4 5 7 to 45 8 )

Knowledge of the physical connections and/or cause- effect relationships between the ARM system and the following systems MRSS NRC KIA Generic System Number RO SRO CFR Link

1. 70 I d RO SRO Question ID: 0054466 Origin: Mod d Memory Level The unit is operating at 100% power with all circulating water pumps in operation when the Main Condenser Waterbox 'D' Circ Water Inlet Valve, CW-11 E. red position indicating light on Panel C-06 IS noted out. Shortly thereafter, the following alarms are received:

- W E S T COND PIT SUMP LVL HI (C-06/7, BB-22)

- ClRC WATER PUMP A OVERLOADlTRlP (C-06/7. A-9)

- ClRC WATER PUMP B OVERLOAD/TRIP (C-06/7, B-9)

- ClRC WATER PUMP C OVERLOAD/TRIP (C-36/7. C-9)

- ClRC WATER PUMP D OVERLOAD/TRIP (C-06/7. D-9)

Which of the following explains the cause of the t r i p uf the circulating water pumps and the primary reason for this protection?

A switches actuated by level at 6 inches below the top of the west cond pit sump, to protect AFW pumps switches actuated by level at 10 inches in the cond pit area, to protect AFW pumps d C switches actuated by level at 6 inches below the top of the west cond pit sump, to protect main generator auxiliary systems D switches actuated by level at 10 inches in the cond pit area, to protect main generator auxiliary systems Justification ARP 2590E states any 2 of 4 level switches reaches 10" all CW pumps will trip CHOICE (A) - NO WRONG Level must reach 10 inches in the condenser pit area to t i i ~ali circulating water pumps.

VALID DISTRACTOR- Level at 6 inches below the top of the west cond pit sump actuates a level swttch to cause the sump level high alarm.

CHOICE (B) - Y E S 2 of 4 level switches in the condenser pit area actuated at >10 inches in pit will trip all circulating water pumps to prevent flooding from challenging the continued operability of !he AFW pumps CHOICE (C) - NO WRONG The primary reason is to protect AFW puinps. hot main gcn3:ator auxiliaries.

VALID DISTRACTOR. Level at 6 inches below the top of the west r.ond pit sump actuates a level switch to cause the sump level high alarm.

CHOICE (D) - NO WRONG. Level must reach 10 inches in the condenser pit area to trip all circulating water pumps VALID DISTRACTOR: Applicant may think that the purpoie is to protect main generator auxiliary systems since the AFW pumps are within protected rooms References 1

1 CWS-00-C. "Circulating Water and Water Box Priming Sysxern? RE;:won 9 (8/29/01). Section D 3 b. "Response to High Waer Level in Condenser Pit Area (Pg 29 of 45) 2 ARP-2590E-116, "WEST COND PIT SUMP LVL HI" NRC KIA SysternIElA System 075 Circulating Water System Number K307 RO 3 4' SRO 3 5' CFR Link (CFR 41 7 / 45 6 )

Knowledge of the effect that a loss or malfunctions of the ~irculalrnqwpter system will have on the following ESFAS NRC KIA Generic System Number RO SRO CFR Link

1. 71 1 4 RO lvl SRO Questior; 19: 5000067 Origin: New 4 Memory Level A transfer of a new fuel assembly is in progress from one location in the spent fuel pool to another using OP-23038,"SFP Fuel Handling Operations". The operator raises the hoist with the desired assembly grappled until upward motion is stopped by the upper limit switch interlock.

What must be done next?

A Release hoist raise switch, use the bridgeltrolley controls to move to destination.

B Stop all hoist and crane movement and notify Reactor Engineering immediately.

C Lower assembly into initial location and contact Reactor Engineering for resolution D Slowly lower hoist until load cell indicates 250 to 290 pounds, then continue move.

~ ~ ~~

Justification CHOICE (A) - NO WRONG: Procedure directs stopping all fuel movement.

VALID DISTRACTOR: Plausible that it is acceptable to have motion stopped by interlock CHOICE (B) -YES Danger of ungrappling and dropping fuel assembly. Must stop and notify immediately CHOICE (C) - NO WRONG: Procedure directs stopping all fuel movement.

VALID DISTRACTOR: Plausible that corrective action would be to lower into rack. This is correct action for fuel handling event.

CHOICE (D) - NO WRONG: Procedure directs stopping all fuel movement.

VALID DISTRACTOR: An applicant may think that use of the interlock affects the load cell. 250 to 290 pounds is the load identified by the procedure for a suspended assembly.

References

1. OP-23038, "SFP Fuel Handling Operations", Revision 1 (10118103) (Pg 1 0 , l l of 35)

2. REF-04-C, "Refueling Equipment" Lesson, Revision 3 (Pg 10 of 71)

NRC KIA SysternIEIA System Number RO SRO CFR Link NRC KIA Generic System 2.2 Equipment Control Number 2.2.28 RO 2.6 SRO 3.5 CFR Link (CFR: 43.7 / 45.13)

Knowledge of new and spent fuel movement procedures.

1. 72 1 IMRO 2 SRO Question ID: 5000066 Origin: New Memory Level Refueling is in progress. A new fuel assembly has just been lowered into core location A - I 1 (core map attached). You are the PPO and have noted the following before and after readings on the wide range logarithmic power channels:

BEFORE AFTER WR CH A 1.9E1 cps 2.OE1 cps WRCH B 1.8E1 cps 3.2E1 cps WR CH C 1.6E1 cps 1.9E1 cps WR CH D 1.OE1 cps 1.2E1 cps Based on these indications, which of the following is required?

A Suspend all core alterations and positive reactivity additions.

B Commence boration per AOP-2558, "Emergency Boration".

C Continue to monitor nuclear instruments, NO immediate action required. v Withdraw the fuel assembly and contact Reactor Engineering for guidance.

Justification CHOICE (A) - NO I

WRONG: Counts have not doubled. The only appreciable increase in counts is on CH B which is immediately adjacent to the location of the new assembly.

VALID DISTRACTOR: Per OP-2209A, if at any time, unanticipated count rate multiplication, (i e., doubling). is indicated. then suspend refuel operations.

CHOICE (B) - NO WRONG: Counts have not doubled. The only appreciable increase in counts is on CH B which is immediately adjacent to the location of the new assembly.

VALID DISTRACTOR: Per OP-2209A, if at any time, unanticipated count rate multiplication, (i.e., doubling). is indicated, then commence boration CHOICE (C) - YES Criteria for action is observation of an unanticipated count rate multiplication, (i,e., doubling). Counts have not doubled CHOICE (D) - NO WRONG: Counts have not doubled. The only appreciable increase in counts is on CH B which is immediately adjacent to the location of the new assembly.

VALID DISTRACTOR: Plausible that requirement is to remove the assembly to lower core reactivity while situation is evaluated.

References I

1. OP-2209A, "Refueling Operations", Revision 24 (1Z17103) (Pg 25 of 64)

2. Provide copy of NIS-O1-C Lesson Figure 2 to applicant.

NRC KIA SystemIElA System Number RO SRO CFR Link NRC KIA Generic System 2.2 Equipment Control Number 2.2.30 RO 3.5 SRO 3.3 CFR Link (CFR: 45.12)

"Knowledge of RO duties in the control room during fuel handling such as alarms from fuel handling area, communication with fuel storage facility. systems operated from the control room in support of fueling operations, and supporting instrumentation."

1. 73 I v RO d SRO Queslion ID: 5000064 Origin: New v Memory Level Which of the following conditions would Ni-)T require iintvediate entry into EOP-2525, "Standard Post-Trip Actions", if the condition were to occur inadvertently with the reactor operating at 100% power?

A Containment Isolation Actuation Signal on both Facilities 4 B Main Steam Isolation Signal on Facility 1 only C Overcurrent trip of normal feeder breaker to 6us 25B froin NSST D Loss of VA-20 with loss of HV to Linear Range CH '0' Justification CHOICE (A) - YES Spurious ClAS is addressed by AOP-2571 "Inad\ 2rtent Emerqency Core Cooling System Initiation' which provides direction for maintaining power operation while addressing the problems of inadvertent isolation CHOICE (B) - NO WRONG Each facility of ESAS can complete the safety function Both MSlVs will close resulting in either a manual or

<iutornaticreactor trip VALID DISTRACTOR Applicant may think that both Facilities must actuate to trip reactor CHOICE (C) - NO WRONG: An overcurrent trip on the bus normal feed will lock out the alternate feed from the RSST Fast transfer will not occur The bus will de-energize and its associated RCP breakers will trip. The reactor will trip on loss of low speed o f two RCPs.

VALID DISTRACTOR: Applicant may think that fast tranfer will occur, maintaining power to the RCPs CHOICE (D) - NO WRONG. Loss of VA-20 trips RPS CH 'B'. Loss of HV to a linear range NI causes associated channel trips through the PTTl With 2 RPS high power trips, a reactor trip will occur VALID DISTRACTOR- Applicant may think that loss of HV will not trip the variable high power on Channel D since the NI output will fail low.

References 1 AOP-2571, "Inadvertent Emergency Core Cooling System Initiation', Revision 4 (12/14/98) (Pg 4 of 15) 2 IHE-00-C. "In-House Electrical System" Lesson, Revision 3 (Pg 10 of 86) 0 Source INPO Bank - Q# 16 - Used at Braidwood 1 . 6/7/1999 NRC KIA SystemlElA System Number RO SRO CFR Link NRC KIA Generic System 24 Emergency Procedures /Plan Number 242 RO 3 9 SRO 4 1 CFR Link (CFR 41 7 I 4 5 7 / 45 81 Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions Note The issue of setpoints and automatic safety features is n d specificallv covered in the systems sections)

1. 74 I v RO v SRO Queslion ID: 5000063 Origin: New v Memory Level The unit is operating at power when multiple indicatiolis are received (radiation levels, feed system response) of a SGTR in #2 SG. The US orders a maliual reactor trip and SIAS. The crew enters EOP-2525, "Standard Post Trip Actions".

During initial performance of EOP-2525 Step 3 ("Determine Status of RCS Inventory Control")

numerous additional annunciators alarm. Operators observe a rapid depressurization of #I SG with a concurrent rapid rise in containment pressure.

Based on these conditions, the crew will immediately A return to Step 1 of EOP-2525 and commence the procedure again 4 B return to Step 1 of EOP-2525 but bring a step forward for manual ClAS C transition to an ORP based on Appendix 1, "Diagnostic Flowchart" D transition to an FRP based on Appendix 1, "Diagnostic Flowchart" Justification I

CHOICE (A) - YES OP-2260. Step 1 19 1 directs the action CHOICE (B) - NO WRONG: EOP-2525 does not contain any "CONTINUOUSLY APPLICABLE" steps.

VALID DISTRACTOR: When in ORP or FRP. selectee steps are identified as "CONTINUOUSLY APPLICABLE" CHOICE (C) - NO 1NRONG. OP-2260 directs performance of EOP-2525 Final steps then direct use of diagnostic flowchart VALID DISTRACTOR- For some events, transition to an ORP is tho cow=? of action that is concluded using the diagnostic flowchart CHOICE (0)- NO WRONG Appendix 1 is not used until completion of SPTAs VALID DISTRACTOR. Plausible that an FRP will be indicated to dedi with multiple events in progress References 1 OP-2260. "Unit 2 EOP User's Guide" Revision 8 (7/11/02) (Pg 16.18 of 32)

~~ ~ ~ ~~~

NRC KIA SystemIElA System Number RO SRO CFR Link NRC KIA Generic System 24 Emergency Procedures /Plan Number 2 4 12 RO 3 4 SRO 3 9 CFR Link (CFR 41 IO/45 12)

Knowledge of general operating crew responsibiliti6.s during ernergeric) I Ferations

1. 1. 75 I vRO d SRO Question ID: 0054207 Origin: Bank d Memory Level Which one of the following correctly describes the proper use of Event Specific EOPs?

A Steps NOT marked with a plus (+) sign may be performed in any sequence as determined by the USISM.

B Steps marked with an asterisk ( * ) sign may be performed in any sequence as determined by the USISM.

C Steps marked with an asterisk (*) may be pulled forward by the SM/US to prevent or correct a d loss of one or more safety functions.

D Actions which deviate from the written sequence of steps may be performed when appropriate guidance for that action is found in an AOP.

Justification A - WRONG; The "+" sign signifies a "hold" point Steps without the ' +" sign must still be performed in the proper sequence, unless the conditions of Choice "B" are met B - CORRECT. The step must be marked with an "*" in order to pull it forward, unless the SRO is invoking 50 54X C - WRONG- AOP actions can be used in an EOP but they are NOT to take the place of applicable steps in the EOP D - WRONG: These steps may be "pulled forward" if needed (per Choice "B"). but still must be performed in the written sequence under all other conditions.

References 1 EOP-2534. "Steam Generator Tube Rupture", Revision 9 (2127101) (Pg 3, 20, 21 of 37) 2 OP-2260. "Unit 2 EOP User's Guide", Revision 8 (7111/02) (Page 10.11 of 32)

NRC KIA SysternIElA System Number RO SRO CFR Link NRC KIA Generic System 24 Emergency Procedures /Plan Number 2 4 19 R027 SRO 3 7 CFRLink (CFR: 41 1014513)

"Knowledge of EOP layout, symbols, and icons "

1. 76 I RO fl SRO Question ID: 0055020 Origin: Mod Memory Level A reactor shutdown is in progress with the following conditions:

- Rx is at 0.8% power.

- Rods are being inserted in Manual Sequential mode

- Group 7 CEAs are at 30 steps.

- Group 6 CEAs are at 155 steps.

- A Group 6 CEA drops to the bottom of the core.

The PPO releases the CEA Control Switch, depresses the Regulating Group 6 INHIBIT BYPASS button, then places and holds the CEA Control Switch in INSERT.

Which one of the following describes the CEDS response and Tech Spec implications?

A ONLY Group 6 CEAs will insert, restore CEA Motion Inhibit to operable within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or ensure the Unit in MODE 3 within next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> B ONLY Group 7 CEAs will insert, restore CEA Motion Inhibit to operable within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or ensure the Unit in MODE 3 within next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> C BOTH Group 6 and Group 7 CEAs will insert, ensure Group 6 CEAS restored to within 10 s t e m of each other within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or ensure the Clnit in MODE 3 within next fi hours D NEITHER Group 6 nor Group 7 CEAs will insert, ensure Group 6 CEAS restored to within 10 fl steps of each other within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or ensure the Unit in MODE 3 within next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Justification SRO ONLY QUESTION - Samples 55 43(2) Facility operating !imitations in the technical specifications and their bases CHOICE (A) NO WRONG CMI bypass requires the operation of both the individual group Inhibit Bypass and the overall system CMI bypass buttons VALID DISTRACTOR Applicant may think that PPO i,tions bypassed CMI for Group 6 Applicant may think actions required for CMI operability CHOICE (B) NO WRONG CMI bypass requires the operation of both the individual qroiip Inhibit Bypass and the overall system CMI bypass buttons VALID DISTRACTOR Applicant may think only Group 6 affected by this CMI Applicant may think actions required for CMI operability CHOICE (C) - NO WRONG CMI bypass requires the operation of both the individual group Inhibit Bypass and the overall system CMI hypass buttons VALID DISTRACTOR Applicant may think CMI is bypassed for Group 6 and CMI is not active for Group 7 CHOICE (D) - YES CMI bypass requires the operation of both the indivioual goup Inhibit Bypass and the overall system CMI bypass buttons PPO actions did not bypass CMI for any group TS requires restoring CEA to within proper aliqnment within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or taking unit to MODE 3 within following 6 hnurs References 1 CED-01-C "Control Element Drive System" Lesson Revision 4 11/26/04) (Pg 28, 30 55. 56 57 of 67) 2 TS 3 1 3 1 "Movable Control Assemblies" - "CEA Position' NRC KIA SystemlElA System 003 Dropped Control Rod Number AA202 R027 SRO 2 8 CFRLink (CFR 4 3 5 i 4 5 1 3 )

Ability to determine and interpret the following as ihey apply to the h 6 p p s d Control Rod Signal inputs to rod control system NRC KIA Generic System Number RO SRO CFQ Link

1. 1. 77 I RO *I SRO Question ID: 5000017 Origin: New *I Memory Level The TS limiting condition for operation action requirements for a Safety Injection Tank (SIT) are less restrictive on time allowed to restore to OPERABLE status for boron concentration than for low level.

This is because the A tank volume requirements are based on minimizing the volume of nitrogen entering the SG tubes and preventing RCS heat removal.

B tank volume requirements are based on one tank emptying through the break and a passive failure of a second tank.

C boron requirements consider the average concentration in the total volume of three safety *I injection tanks D boron requirements assume sufficient shutdown margin due to void fraction during a large break LOCA Justification SRO ONLY QUESTION - Samples 55 43(2) Facility operating limitations in the technical specifications and their bases CHOICE (A) - NO WRONG Tank volume requirements are not based on nitrogen volume entering SG tubes VALID DISTRACTOR because nitrogen in SG tubes is a concern addressed during LOCA mitigation CHOICE (E) NO WRONG Tank volume requirements assume 3 tank5 remain available tc provide core cooling during the initial stages of a large break LOCA VALID DISTRACTOR because the applicant may remember that devgn assumptions are made regarding loss of one Icq of ECCS injection and on passive and active failures CHOICE (C) YES Boron concentration is a requirement to ensure abiliLj to maintain subcriticality during a LOCA However the effects of reduced concentration during core reflood are minor for two reasonc Core boiling tends to concentrate the boron in the inventory in the vessel and the boron concentration limit CI . based or1 :he average concentration in the total volume of three safety injection tanks CHOICE (D) - NO WRONG Boron concentration is a requirement to ensure ai~ilityto maintain subcriticality during a LOCA However the effects of reduced concentration during core reflooo are minor for two reasons Core boiling tends to concentrate the boron in the inventory in the vessel and the boron concentration limit is based on the average concentration in the total

\ oluine of three safety injection tanks VALID DISTRACTOR because the applicant may Assume that the bdsis for the boron requirements takes void fraction into account References

' Technical Specifications Section 8314 5 1 , "Safety Injection Tanks" Amendment 220 (Page 83/4-51)

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NRC KIA SystemlElA System Number RO SRO CFR Link NRC KIA Generic System 22 Equipment Control Number 2225 RO 2 5 SRO 3 7 CFR Link (CFR 43 2)

Knowledge of bases in technical specifications for limiting zmditior s for operations and safety limits

1. 1. 78 I RO *I SRO Question ID: 5000020 Origin: Mod Memory Level The plant is at 100% with Containment Plessure Chani?el "A" (PT-8113) failed high. The channel's associated bistable red "TRIP" light is lit at the Chsnnel A ESF sensor cabinet. The pressure bistable key bypass switch associated with this channel is in INHIBIT.

The crew receives the SlAS OR UV ACTUATION SIG CH 2 TRIP annunciator on C-01. The following are the pressures sensed by the Containment Pressure channels:

PT-8113, Containment Pressure CH A => 00s PT-8114, Containment Pressure CH B => 1.1 psig PT-8115, Containment Pressure CH C --> 1.Ipsig PT-8116, Containment Pressure CH D => 4.3 psig PT-8117, Containment Pressure LR => 1.2 psig PT-8238, Containment Wide Range Pressure => 1.O psiy PT-8239, Containment Wide Range Presssre => 1, I psig Containment radiation monitors are NOT in alarm and stable. Containment sump level is 46% and steady. NO ECCS equipment is actuated.

Which of the following statements is correct about the "SIAS OR UV ACTUATION SIG CH 2 TRIP" annunciator on C-01 and what procedural actions are necessary to address plant conditions?

A The alarm is valid, 1 of the associated Containment Pressure Channels are tripped, enter OP-231 4 8 , "Containment and Enclosure Building Purge".

B The alarm Is NOT valid, 1 of the associated Containment Pressure Channels are tripped, enter OP-2314B, "Containment and Enclosure Building Purge".

C The alarm is valid, 1 of the associated Containment Pressure Channels are tripped. enter EOP-2525, "Standard Post Trip Actions".

D The alarm is NOT valid, 1 of the assocnted containment Pressure Channels are tripped, enter EOP-2525, "Standard Post Trip Actions".

Justification SRO ONLY QUESTION - Samples 55 43(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations CHOICE (A) - NO WRONG Alarm actuated by 2 of 4 ESF channels (8 I13 thru 81 16) equal to or greater than 3 8 psig However PT-81 16 is the only ESF channel above the SlAS actuation setpoint The alarm is not valid VALID DISTRACTOR One channel is reading abovo 3 8 psig A rapid downpower is required because of h q h energy line break in containment CHOICE (6) -YES Alarm actuated by 2 of 4 ESF channels (81 13 thru 81 16)equal to or greater than 3.8 psig. However, PT-8116 IS the only ESF channel above the SlAS actuation setpoint. With less than 2 safety channels above setpoint. the alarm is not valid A rapid downpower is required because of high energy line break in containment.

CHOICE (C) - NO WRONG: A l a n actuated by 2 of 4 ESF channels (81 13 thru 81 16) equal to or greater than 3.8 psig. However, PT-81 16 is the only ESF channel above the SlAS actuation setpoint. With less than 2 safety channels above setpoint. the alarm is not valid.

VALID DISTRACTOR: ESF Pressure Transmittei a1 16 is reading abode 3.8 psig.

CHOICE (D) - NO WRONG: A high energy line break in containment with pressure at 1.O psig is reason to perform rapid downpower The plant is still at power. No reason to enter EOP-2525 VALID DISTRACTOR: Alarm actuated by 2 of 4 ESF .:hanneis (81 i 3 thru 81 16) equal to or greater than 3.8 psig PT-81 16 is the only ESF channel above the SlAS actuation setpoint The alarm is not valid.

References 1 ARP-2590A, "Alarm Response for Control Room Panel G O 1 ' Windqw 8-34 "SIAS OR UV ACTUATION SIG CH 2 TRIP (RED WINDOW)"

2 Containment Ventilation Print 25203 26028-1 Sheet 1 of 6 (J 4 1-12) 3 MP-16-CAP-SAPO1 "Condition Report Initiation ' Revision 1 (8:31/'14) 4 MP-16-MMM "Corrective Action' 5 OP-2384 "ESAS Operation", Revision 14 (1 1, I i103)(Pg 20 21 of 301 6 Source Indian Point 3 NRC Exam 1212003

NRC KIA SystemlElA System Number RO SRO CFP Link NRC KIA Generic System 24 Emergency Procedures/Plan Number 2446 R035 SRO 3 6 CFRLink (CFR 4 3 5 1 4 5 3 1 4 5 1 2 )

Ability to verify that the alarms are consistent with t h e plant conditions

1. 1. 79 1 RO 4 SRO Question ID. 5000021 Origin: Bank Memory Level The following plant conditions exist:

The plant is at 100% power. The '3' Emergency Diesel Generator (EDG) was declared INOPERABLE yesterday at 0600. At 0800 today, the Shift Manager discovers that the conditional 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> surveillance operability run on the 'A' EDG required by the 'B' EDG action statement has not been performed.

What action is required?

A The operability surveillance of A' EDG must. be performed successfully by 1000 today or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

B The operability surveillance of A' EDG must be performed successfully by 0900 today or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

C The operability surveillance of A' EDG must be performed successfully by 1200 today or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

D The operability surveillance of A' EDG mus! be performed successfully by 0800 tomorrow or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Justification SRO ONLY QUESTION - Samples 55 43(2). Facility operating limitations in the technical specifications and their bases CHOICE (A) - YES

'A' EDG becomes INOPERABLE at 0600 today if it is not run successfully Action statement (e 2) reads "Restore one of the inoperable diesel generators to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and " Two hours starts from time of declared inoperability (0830 + 2 hrs = 1000)

CHOICE (6) - NO WRONG 'A' EDG becomes INOPERABLE at 0600 today i f it is not run successfully Action statement (e 2) reads

'Restore one of the inoperable diesel generators to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and "

VALID DISTRACTOR TS 3 0 3 requires action taken withl,~1 hour 1" achieve HOT STANDBY within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> if TS and associated action requirement is not met CHOICE (C) - NO WRONG 'A' EDG becomes INOPERABLE at 0600 today if i t ir not run sLLcessfully Action statement (e 2 ) for two inoperable diesels reads "Restore one of the inoperable diesel generators to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and '

VALID DISTRACTOR Action e 2 does require HOT STANDBY with the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Applicant may think that 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> are available to perform required testing CHOICE (D) - NO WRONG 'A' EDG becomes INOPERABLE at 0600 today if it is not run successfully Action statement (e 2) reads

'Restore one of the inoperable diesel generators to OFERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and "

VALID DISTRACTOR TS 4 0 3 allows 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> belore applying required actions for a missed surveillance However this is a conditional surveillance in order to maintain complianc? with the action TS 4 0 3 does not apply to this situation References 1 Technical Specification 3/4 8 1 Pages 3/4 8 l a (Amendment 261) 3/4 8-2a (Amendment 2 7 7 )

2 Source INPO Bank Q# 24702 - Used at Seabrcok 05/30/2003 3 Provide TS 3 8 1 1 to applicants

~~~

NRC KIA SystemIEIA System Number NRC KIA Generic System 21 Conduct of Operations Number 2 1 10 RO 2 7 SRO 3 9 CFR Lmk (CFR 43 1 / 45 13)

Knowledge of conditions and limitations in the facility license

1. 80 I RO v SRO Question ID: 50C0022 Origin: Bank Memory Level The following conditions exist for a job performed on F sy-+em:

- The general area radiation levels are 10 mremihr

- The hot spot in the room is a pipe elbow that has radiatiosIlevels of 100 mrem/hr

- The job will be performed near the hot spot area Assuming the time to get to and from the job site is the same for each case and all shielding placement is done at 100 mrern/hr, which ONE (1) of the following results in the LEAST amount of personnel exposure?

A The job is performed by 2 operators for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> each c i the job at the hot spot B The job is performed by 2 operators for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> each on the job at the hot spot and a 3rd operator reading instructions in the general loom area for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

C The job is performed by 3 operators for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> each on the job at the hot spot and a 4th operator reading instructions in the general morn area for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

D 2 Health Physics technicians require 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> to install and remove 1 tenth thickness of lead shielding on the hot spot. The job is performed with the shielding in place by 2 operators for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> each.

Justification I SRO ONLY QUESTION - Samples 55 43(4) Radiation hazards that may a r m during normal and abnormal situations including maintenance activities and various contammation ponditions CHOICE (A) - NO WRONG Total dose for this plan equals 600 mrem The lowest dose of any choice provided is 310 mrem VALID DISTRACTOR This choice involves the fewest number of personnel CHOICE (B) - NO bVRONG Total dose for this plan equals 420 mrem The : ,west dose of sriy choice provided is 310 mrem VALID DISTRACTOR This choice requires less tinz lo complete the jnb than 2 other choices CHOICE (C) - YES This choice results in the lowest total dose of 310 m e m CHOICE iD) - NO vVRONG Total dose for this plan equals 360 mrem The lowest dose of any choice provided is 310 mrem VALID DISTRACTOR This choice installs shielding to reduce the dose t: workers References 1 Source Indian Point 3 NRC Exam, 12/2003 NRC KIA SystemlElA System Number RO SRO CFH Link NRC KIA Generic System 23 Radiation Control Number 2 3 10 RO 2 3 SRO 3 3 CFR Link (CFR 43 4 / 45 10)

Ability to perform procedures to reduce excessive ievels of rajiation and guard against personnel exposure

1. 81 I RO 4 SRO Question ID: 5009023 Origin: New 4 Memory Level Which of the following evolutions raises an immediate AL.AHA concern requiring notification of Health Physics Department? Consider the effects of the described action only.

A increasing CVCS letdown flow during normal power operations 4 B starting of the HPSl pumps by SlAS during a large break LOCA C increasing SFP cooling flow during spent fuel pool fuel moves D shifting from 'C' to 'A' Charging Pump running at 75% power Justification I SRO ONLY QUESTION - Samples 55 43(4) Radiation hazards that may arise during normal and abnormal situations including maintenance activities and various contamination conditioiis CHOICE (A) - YES Step 4 4 10 requires notification of Health Physics Department of any change in charging or letdown flow C H O I C E ( 6 ) - NO WRONG HPSl pumps are aligned for injection from the RWST and will not affect local dose rates until post SRAS EOP does not require HP notification at start of LOCA VALID DISTRACTOR A LOCA has the potential to raise local dose rates CHOICE (C) - NO WRONG Notification of HP is not required for increasing SFP cooling flow VALID DISTRACTOR Plausible that increasing SFP cooling flow might create ALARA concerns CHOICE (D) - NO WRONG Shifting charging pumps does not change charging flowrate and therefore does not present ALARA concerns Both of these pumps are located in the sdme general area VALID DISTRACTOR A caution states that HP should be notified for charqing charging flow conditions References 1 RPM-1 1 2, "Radiation Protection Program and AILARA Program' Rev,sion 3 (8/19/04) (Pg 5,8 9 10 11 16 of 33) 2 OP-2304E. "Charging Pumps", Revision 15 (03/09/04) S'ep 4 4 :O iPg 21 of 25)

NRC KIA SystemlElA System Number RO SRO CFR Link NRC K/A Generic System 23 Radiation Control Number 232 RO 2 5 SRO 2 9 CFR Link (CFR 41 12 / 43 4 45 9 / 4 5 10)

Knowledge of facility ALARA program

1. 1. 82 I RO *I SRO Questiog IC: 5000024 Origin: Bank Memory Level Unit 2 is operating near beginning of cycle dt a burnup of 50G MWDIMTU.

The following conditions exist AFTER a transient from 99% power:

- steam generator pressure is lower

- main generator megawatt output is lower

- indicated feedwater temperature is lower

- reactor coolant hot leg temperature is lower Which one of the following events caused this plant response and what is the applicable procedure for addressing the problem? Assume NO operator action.

A condenser backpressure rise (degraded vacuum). address with ARP-2590E (A-37). "COND VACUUM LO" B sensor input to throttle pressure limiter failed (0 psig), address with ARP-2590D (DA-22),

"10% TURBINE LOAD DECREASE" C feedwater heater extraction steam isolation valve closed (heater 1B), address with ARP-2590D (AA-18), "HEATER 1A LEVEL HI" D atmospheric dump stuck in intermediate position (50% open), address with ARP-2590D (B-6), *I "ATMOSPHERIC DUMP VALVE NOT CLOSED" Justification I

SRO ONLY QUESTION - Samples 55 43(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations CHOICE (A) - NO WRONG: Degraded vacuum with no movement of zontrol valves or cnntrol rods will result in no observable change in steam generator pressure. Steam flow will remain constant. Efficiency of the turbine will decrease Turbine will perform less work. The additional energy rejected to the condenser will be removea by circulating water systein Feedwater temperature will be unchanged and reactor power will be unLhangs4 VALID DISTRACTOR: Increasing backpressure will m u s e main geierator output to decrease.

CHOICE (B) - NO WRONG. Throttle pressure limiter is maintained in OFF during power operations to prevent undesirable load transients VALID DISTRACTOR: If on, the throttle pressure limiter would act to reduce turbine load.

CHOICE (C) - NO

'WRONG. Main turbine output will increase slightly with the isolation of an extraction line as extraction steam is redirected through subsequent turbine stages.

VALID DISTRACTOR. Loss of extraction will result i n lower feedwater temperature.

CHOICE ( D ) - YES Fully open ARV passes steam flow equivalent to approximately 7 5% reactor power. Steam flow will increase. Steam pressure will drop. With lower steam pressure, the main turbine output will drop. Feed flow will increase to maintain steam generator level. Increased feed flow with same extraction heating steam flow will result in lower feedwater temperature. The increased total steam flow will reduce average coolant tcmperature. The moderator temperature coefficient of reactivity will raise reactor power until equilib,;,m condi!ions are re-established. Reactor power and core delta-T will be higher, but Tave, Th and Tc will be lower References 1 MSS-00-C. "Main Steam System" Lesson, Revision 6 Section 19 b (Pg 27) 2 OP-2204, "Load Changes", Revision 19 (6/29/04) Attachment 6 "Temperature vs Power Program"(Pg 42 of 46) 3 Source INPO Bank - Q# 23848 - Used at Susquehanna 1 08/01/2002 NRC KIA SystemIElA System Number RO SRO CFR Link NRC KIA Generic System 24 Emergency Procedures /Plan Number 2 4 47 RO 3 4 SRO 3 7 CFR Link (CFR 41 10 43 5 / 45 12)

Ability to diagnose and recognize trends in an acciNiate and timely rnanriei utilizing the appropriate control room reference material

1. 83 I RO d SRO Questijll ID. 00715C9 Origin: Bank d Memory Level Given the following conditions exist after F reactor trip whik implementing EOP 2526, "Reactor Trip Recovery" :

- Pressurizer level 25% and lowering slowly

- RCS pressure 2230 psia and trending up slowly

- RCS Tavg 534°F and steady

- "A"SG level 41% and trending up slowly

- "B" SG level 55% and lowering slowly Identify the procedure which will be implemented next and the step that may be performed out of its given sequence within that procedure?

A EOP-2526, "Reactor Trip Recovery", manually adjust steam generator feed flows to control SG levels B EOP-2526, "Reactor Trip Recovery", manually adjust charging and letdown to control 4 pressurizer level C EOP-2536, "Excess Steam Demand Event", manually operate heaters and spray to control pressurizer pressure D EOP-2536, "Excess Steam Demand Event". manually operate steam dump/bypass valves to control RCS Tcold Justification SRO ONLY QUESTION - Samples 55 43(5) Assessment of facility conditions and selection of appropriate procedures during normal abnormal and emergency situations CHOICE (A) - NO WRONG SG levels are within band and do not require action to correct or preserve the safety function VALID DISTRACTOR EOP-2526 is the correct procedure CHOICE (13) YES Step 1 10 2 of the EOP User's Guide provides the following two conditions bhen EOP steps may be performed out of the order listed in the procedure 1) steps which are asterisked may be brought forward to correct or preserve a safety function and 2) steps may be performed out of order after they have been accomplished once if they are Continuously Applicable step as indicated by an asterisk Pressurizer level IS outside of the band given and trending in away from the hand CHOICE (C) - NO WRONG EOP-2536 is not the correct procedure VALID DISTRACTOR Manual control of pressurizer pressure is an asterisked step CHOICE (D) - NO WRONG RCS Tcold is within the band and does nc+reqdre action io correct or preserve the safety function VALID DISTRACTOR Manual control of Tcold via steam dupibypass dalves IS an asterisked step 1

References i OP-2260 "Unit 2 EOP Users Guide" Revislon F 17/11/02) SeL;iun 1 10 2 (Pg 11 of 32)

NRC KIA SystemlEIA System E02 Reactor Trip Recovery Number EA22 RO 3 0 SRO 4 0 CFR Link (CFR 43 5 I 4 5 13)

Ability to determine and interpret adherence to appronriate procedures rnd operation within the limitations in the facility s license and amendments as they apply to the Reactor Trip Recogery NRC KIA Generic System Number RO SRO CFR Link

1. 1. 84 I RO g SRO Questicn ID: 5000043 Origin: New Memory Level The unit is operating at 100% power, equilibrium conditions. RCS leakage has been steady (confirmed by manual calculation) over the last 2 days at 0.38 gpm, of which 0.01 gpm is known tube leakage on #1 SG.

The RO makes a 60 gallon water addition as measured by Flow Integrator F-210X to the RCS during the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> period of today's PPC leakrate calculation.

If Flow Integrator F-210X has inaccurately measured 5 gallons more than was actually injected, and the resulting leak rate change is attributed entirely to SG # I tube leakage, then the new SG # I tube leak rate based on PPC calculation will be _ _ .~

A 0.0308 gpm, which is less than the TS LCO for primary-to-secondary leakage. 9 B 0.218 gpm, which is less than the TS LGO for primary-to-secondary leakage.

C 0.0308 gprn, which exceeds the TS LCO for primary-to-secondary leakage and exceeds the pressure boundary leakage limit.

D 0.218 gpm, which exceeds the TS LCO for primary-to-secondary leakage and meets the pressure boundary leakage limit.

Justification SRO ONLY QUESTION - Samples 55.43(2) Facility operating limitations 111 the technical specifications and their bases CHOICE (A) - YES 5 !jallons less added to RCS than actually added. VJiIl be calculated a;, an increase of leakage of 5 gallons over 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> This equates to a calculated increase in leakage 01 J 0208 gpm. Added to existing SG tube leakage, new calculated pri-to-sec is 0.0308 gprn. TS limit is 0.035 gpm CHOICE(B) - NO WRONG. Measured leakrate is 0.0308 gpm VALID DISTRACTOR: Plausible value. off by factor of 10 CHOICE (C) - NO WRONG. Measured leakrate does not exceed the TS LCO VALID DISTRACTOR. Applicant may think this exceeds SG and pressure boundary limits CHOICE (D) - NO WRONG: Measured leakrate does not exceed the TS LCO VALID DISTRACTOR. Could determine leakage to be 0 218 gpm making this choice credible References I I

1 TS 3 4 6 2, "Reactor Coolant System Leakage" Amendment 228 (Pg ;/44-9) 2 SP-2602A "Reactor Coolant Leakage", Revision 3 (08/31/04) (Pg 3 of 20)

NRC KIA SysternlElA System 022 Number RO SRO CFR Link NRC KIA Generic System 22 Equipment Control Number 2 2 12 RO 3 0 SRO 3 4 CFR Link (CFR 41 10 / 45 13)

Knowledge of surveillance procedures

1. 85 I RO v SRO Question ID: 5000044 Origin: New g Memory Level In EOP-2528, "Loss of Offsite Power/Losx of Forced C~i&!lation", a note states that a cooldown rate of 30"F/hr should be observed when RCS Tcold IS below 230°F. This note is based on A preventing uncoupling of the core and loops B ensuring a margin of safety against brittle failure C preventing a cold water accident following RCP restart D ensuring no void formation due to upper head temperature Justification SRO ONLY QUESTION - Samples 55.43(2) Facility operating Iirnitatims in the technical specifications and their bases CHOICE (A) - NO WRONG: During natural circulation, the maximum cooldown rate at which the RCS loops can be maintained coupled is dependent on decay heat and RCS temperature , Due to this. the cooldown rate must be lowered as the cooldown progresses. A cooldown rate of 30 to 60 degrees an hour should be maintainable initially. and a rate of 10 to 25 degrees per hour should be sustainable until RCS temperature reaches 300°F While coupling is a concern. the higher limit of 30'Flhr is a technical specification requirement related to brittle fracture concerns.

VALID DISTRACTOR: Uncoupling is a concern during a natural circulation cooldown.

CHOICE(B) - YES Cooldown rate is maintained in accordance with Technical Specifications to ensure a margin of safety against brittle (non-ductile) failure CHOICE (C) - NO WRONG: Rate of cooldown does not affect RCP restart VALID DISTRACTOR: Plausible that stagnant legs can develop during na!ural circulation cooldown CHOICE (D) - NO WRONG. Cooldown limit is a technical specification brittle fracture concern.

VALID DISTRACTOR: Void formation is a concern during natural circulation.

References 1 EOP-2528, "Loss of Offsite PoweriLoss of Forced Circulation" Revision 15 (2f27101).(Page 21 of 36) 2 TS Basis 314.4.9,"RCS Pressure and Temperature Limits". Amendment 218 (Pg B 314 4-5. 4-6)

NRC KIA SystemlElA System 026 Number RO SRO CFR Link NRC KIA Generic System 2.4 Emergency Procedures /Plan Number 2.4.20 RO 3 3 SRO 4 0 CFR Link (CFR 41.10 I45 13)

"Knowledge of operational implications of EOP wa;,rii-,gs, cautinns. and notes "

1. 1. 86 I RO 4 SRO Questicn ID: 5000045 Origin: Mod 4 Memory Level The plant has been operating at 100% power. An automatic reactor trip has been initiated by RPS.

All CEA position indications show all rods are out of the core. The reactor trip breakers will not open at C-04 or locally and the CEDM feeder breakers will not open.

Operators implement the appropriate EOP. Identify the correct procedure entered and the procedurally identified criteria for verifying that the reactivity control acceptance criteria is met?

A EOP-2525, "Standard Post Trip Actions", emergency boration in progress 4 B EOP-2525, "Standard Post Trip Actions". power dropping and negative SUR C EOP-2540A, "Functional Recovery of Reactivity Control", emergency boration in progress D EOP-2540A, "Functional Recovery of Reactivity Control", power dropping and negative SUR Justification SRO ONLY QUESTION - Samples 55 43(5) Assessment of facility conditioris and selection of appropriate procedures during normal, abnormal and emergency situations CHOICE (A) - YES EOP-2525 provides contingency actions to trip the reactor from the control room by opening the MG set output breakers EOP-2540A is not entered until all SFSC evaluated Reactivity SF will be met before transition from EOP-2525 Cnteria are to take actions, already taken in stem to try to trip reactor Additionally if any rods stick out the condition will be addressed within EOP-2525 by emergency boraticrn CHOICE(B) - NO WRONG Reactivity SF in EOP-2525 does not require power dropping and negative SUR If these conditions are not met the contingency steps require actions to attempt to trip reactor Emergency boration is a required criterion for the stuck CEAs VALID DISTRACTOR Power dropping and negative SUR are rcqhired tu be checked to determine if continqency steps should be performed CHOICE(C) NO WRONG EOP-2525 will address the problem EOP 2540A will not neeo to be implemented VALID DISTRACTOR EOP-2540A addresses the loss of reactivity control safety function CHOICE (D) - NO WRONG Power decreasing and negative SUR are not the criteria for confirming reactor shutdown VALID DISTRACTOR If EOP-2540A were the procedure to be implemented, power dropping and negative SUR would be criteria for the SF References 1

I EOP-2525. "Standard Post Trip Actions" Revision 20 (2127/01)(Pg 3 of 26) 2 OP-2260, "Unit 2 EOP Users Guide" Revision 8 (7111/02')

3 Source INPO Bank - Q# 24679 - Used at Seabrook 1 5/30/2003 NRC KIA SystemlElA System 029 Anticipated Transient Without scram ( A T W S )

Number EA209 RO 4 4 SRO 4 5 CFR Link (CFR 43 5 / 45 13)

Ability to determine or interpret the following as they apply io a ATWC Occurrence of a main turbineheactor trip NRC K/A Generic System Number RO SRO CFR Link

1. 87 I RO d SRO Question IC: 0053649 Origin: Bank d Memory Level A reactor trip has occurred.

During the performance of EOP 2525, "Standard Post Trip Actions", the following conditions are reported:

PPO:

- SJAE RM is alarming

- Pressurizer level is 15% and lowering

- Pressurizer pressure is 1850 psia and lowering

- NO other apparent problems SPO:

- SIG Blowdown has isolated

- #1 S/G level is 9% and rising

- #2 S/G level is 34% and rising

- # I FRV Bypass - 60% open

- #2 FRV Bypass - 30% open

- Both SIG levels are rising at the same rate

- No other apparent problems Per EOP-2525, what actions must be taken with regard to steam generators?

A Secure feed to # I S/G.

B Secure feed to #2 S/G.

C Feed SGs to maintain levels of 10 to 80%.

D Feed S G s to maintain levels of 40 to 70%. d J tistification I

SRO ONLY QUESTION - Samples 55 43(5) Assessment of facility conditions and selection of appropriate procedures during normal abnormal and emergency situations CHOICE (A) - NO WRONG #1 SG is the unaffected SG Level will be Lontrolled betweeii 40 and 70%

VALID DISTRACTOR Plausible that #I unnecessary for heat removal CHOICE(B) - NO WRONG Step 7 contingency has the affected SG controlled between 40 3nd 70%

VALID DISTRACTOR Plausible that #2 would be isolated because of the SG tube leak CHOICE(C) NO LVRONG Both SGs will be maintained between 40 and 70%

VALID DISTRACTOR Step 6 has the operator ensure at leas! one SG is between 10 and 80% as a heat sink for the reactor coolant system CHOICE (D) - YES EOP 2525 step 7 b contingency actions for SJAE ractiation monitor unexplained activity requires that feed be throttled to maintain 40-70 % t o the SG with the highest radiatiori readings The highest radiation reading will be in the SG with the hiqher IevelAowest feed flow Step 6c restores level to 40 to 70%

References 1 EOP-2525 "Standard Post Trip Actions". Reviaun 20 (2127101)(Pg 16,18 of 26) 2 'EOP-2525. Standard Post Trip Actions Technicil Guide' Revision 2U (Pg 19 of 38)

NRC KIA SystemlElA System 038 Number RO SRO CFR Link NRC KIA Generic System 21 Conduct of Operations Number 2 123 RO 3 9 SRO 4 0 CFR Link (CFR 45 2 I 4 5 6)

Ability to perform specific system and integrated plaiit procedures cwrinq a1 modes of plant operation

1. 88 I RO v SRO Question ID: 0054362 Origin: Mod Memory Level While operating at 100% power, a plant trip occuis. While carrying out EOP-2525. Standard Post Trip Actions, the operators observe the following plant conaitions:

• All CEAs are inserted.

• All buses are energized.

Pressurizer Level is I O % , lowering.

• Pressurizer Pressure is 1700 psia, lowering

• Tavg is 505 OF, lowering.

• RCS subcooling is 100 "F,rising.

• Feeding both SGs with Main Feedwater

' #1 SG level 15% and dropping.

' #2 SG level 42% and rising.

• #I SG pressure 450psia and dropping.

• #2 SG pressure 650 psia and dropping.

' Containment pressure 1.5 psig, rising.

• SJAE rad monitor activity rising.

• #2 MSL rad monitor activity rising.

• NO rad monitors in alarm.

Which procedure will the operators implement next?

A EOP 2532, Loss of Coolant Accident B EOP 2534, S/G Tube Rupture C EOP 2536, Excess Steam Demand D EOP 2540, Functional Recovery Justification SRO ONLY QUESTION - Samples 55 43(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal and emergency situations CHOICE (A) - NO WRONG Multiple events (SGTR and ESDE) require antry into the functional recovery procedure VALID DISTRACTOR Pressurizer pressure is lower4ng pressurizer level is lowering CHOICE(B) - NO WRONG Multiple events (SGTR and ESDE) require entry iiito the tcinctional recovery procedure VALID DISTRACTOR Pressurizer pressure IS dropping no containme7t rad monitor alarms CHOICE (C) - NO WRONG Multiple events (SGTR and ESDE) requiie entry into t h t funcliunal recovery procedure VALID DISTRACTOR #I S/G level is dropping CHOICE (D) - YES Multiple events are in progress (SGTR and ESDE with failure of MSI), requiring entry into the functional recovery procedure References I 1 OP-2260, "Unit 2 SOP User's Guide" Revision 1 (7/11 / 3 2 )(Pg 9 10 Jf 32) 2 EOP-2541 Appendix 1 "Diagnostic Flowchart Revision 000 (10/2/q3) (Pg 1 of 1)

NRC KIA SystemIElA System A1 1 RCS Overcooling Number AA21 RO 2 9 SRO 3 J CFR Link (CFR 43 5 / 45 13)

Ability to determine and interpret the following as they ap ti!. to the {RCSOvercooling) Facility conditions and selection of appropriate procedures during abnormal and emergency ~ p e r a t i o ~ i s NRC KIA Generic System Number RO SRO CFR Link

1. 89 I RO d SRO Question ID: 5000G46 Origin: New Memory Level The plant is operating at 100% power when a large steam line rupture occurs on the upstream side of

1. I Main Turbine Stop Valve (SV-1). The reactor and turbine are manually tripped. While responding to the event. the operator attempts to close the MSIVs u;ing handswitches on C05.

Both MSlVs remain open. Which of the following choices is required to address this problem?

A Direct local manual closure of both MSlVs per EOP-2525, Standard Post-Trip Actions B Direct local manual closure of both MSlVs per EOP-2536, Excess Steam Demand Event.

C Place MSlV Bottle-Up Panel Isolation Switches in ISOL per EOP-2525. Standard Post-Trip d Actions.

D Place MSlV Bottle-Up Panel Isolation Switches in ISGL per EOP-2536, Excess Steam Demand Event.

Justification SRO ONLY QUESTION - Samples 55 43(5) Assessment of facility conditiolis and selection of appropriatr procedures during normal abnormal and emergency situations CHOICE (A) NO WRONG EOP-2525 contingency for MSlVs fail to close does not direct local manual closure VALID DISTRACTOR Plausible that local action may be required CHOICE (E) - NO WRONG EOP-2536 contingency for MSlVs fail to close does not direct local manual closure VALID DISTRACTOR Plausible that local action may be required CHOICE (C) - YES EOP 2525 Step 6 contingency directs use of bottle-up panel switches CHOICE(D) NO WRONG Operator would take action to operate tottle-up panel switches in EOP-2525 before entering EOP-2536 VALID DISTRACTOR Plausible that EOP-2536 would cover the necessary contingencies to address an MSlV fall to close References 1 EOP-2525. "Standard Post Trip Actions", Revision 20 (2/22/01) (Pg 13 of 26)

NRC KIA SystemlEIA System E05 Number RO SRO CFR Link NRC KIA Generic System 24 Emergency Procedures /Plan Number 2449 RO 4 0 SRO 40 CFR Link (CFR 41 10 / 43 2 ' 4 5 6)

Ability to perform without reference to procedures t h w e actions that require immediate operation of system components and controls

1. 90 RO d SRO Question ID: 5000047 Origin: New d Memory Level Refueling is in progress on Unit 2. During normal rounds, the Aux Building PEO reports that the red light on the SFP SW Area Radiation Monitor (RM-8139) local module is illuminated.

Which of the following is a possible reason for the reported indication?

A loss of power to the radiation monitor B local horn silence switch in the OFF position C actual high radiation condition in speni fuel pool area *I D Fuel Area Radn AEAS switch at ESF sensor cab in INHIBIT Justification J

SRO ONLY QUESTION - Samples 55 43(4) Radiation hazards that may arise during normal and abnormal situations including maintenance activities and various contaminatim conditions CHOICE (A) - NO WRONG Loss of power extinguishes the light VALID DISTRACTOR Plausible that function of red light is to indicate a loss of power CHOICE ( 6 ) NO WRONG Local horn silence keyswitch disables the audible aldtm Dut leaves the red light lit VALID DISTRACTOR Plausible that function of red light is to inform that audible is defeated CHOICE (C) - YES Local red light illuminates on sensed high radiation cvlidition at a readinq exceeding SOmRIhr CHOICE (D) - NO WRONG Keyswitch at ESF sensor cabinet functions to inhibit the trip and change logic from 2 14 to 2 out of 3 VALID DISTRACTOR Plausible that red light designed to provide local indication of defeated input to ESAS References 1

I RMS-00-C, "Radiation Monitoring System" Lessor iievision 6 (YlB'r12; (Pg 14.16 of 109)

NRC KIA SystemIElA System 061 Area Radiation Monitoring (ARM) System Alarms Number AA201 RO 3 5 SRO 3 7 CFR Link (CFR 43 5 1 45 131 Ability to determine and interpret the following as they apply to the Area Radiation Monitoring (ARM) System Alarms ARM panel displays NRC KIA Generic System Number RO SRO CFR Link

1. 91 I RO d SRO Question ID: 5003048 Origin: New Memory Level Plant is stable at 100% power with no activities in progress and no equipment out of service when the FIRE SYSTEM TROUBLE (AB-19, C-06/7) an(: 'he 'A' DIESEL ROOM (Zone 12, C-26) alarms are received. Approximately 1 minute later the following conditions are confirmed:

U3 Electric Fire Pump is stopped in standby U3 Diesel Fire Pump is stopped in standby U2 Electric Fire Pump is stopped in standby Jockey Fire Pump is running Panel C-26H ( ' A DG local fire panel) alarm light lit Panel C-26H one heat detector light lit Panel C-26H audible alarm horn actuated Other Panel C-26H indications are normal No other control room alarms have actuated The AB PEO has not entered the DG room This situation indicates a A failure of the heat detector, go to ARP-25901, "(2-26 Alarm Response" B fire in the diesel generator room, go to AOP-2559, "Fire" C melted fusible link in deluge system, go to RP-16. "Trodble Reporting" D loss of ventilation, go to OP-2315E, "Diesel Generator Ventilation System" Justification I

SRO ONLY QUESTION - Samples 55 43(5) Assessment ot facility conditions and selection of appropriate procedures during normal abnormal and emergency situations CHOICE (A) - Y E S Actual fire would melt fusible link, causing supervisorv air low pressure alarm CHOICE (B) - NO WRONG Fire would open deluge valve resulting in additional alarms VALID DISTRACTOR Plausible that detector has reported d valid condition CHOICE (C) - NO WRONG Melted fusible link would initiate spray flow actuating deluqe valve opening alarm VALID DISTRACTOR Plausible that alarm caused by melted link CHOICE (D) - NO WRONG Temperature Switch TS-8435 provides a Diesel Gen 12U Room Temp HiiLo alarm on C-08 at 110°F VALID DISTRACTOR Plausible that room temperature increase has cahsed the alarm References 1 FPS-O4-C, "Fire Protection System" Lesson Qevision 3 (Pg 28,29,52 of 82 NRC KIA SystemlElA System 067 Plant fire on site Number AA2 09 RO 2 4 SRO 2 7 CFR Link (CFR 43 5 / 45 13)

Ability to determine and interpret the following as they apply to thc Plapi Fire on Site That a failed fire alarm detector rxists NRC KIA Generic System Number RO SRO CFR Link

1. 92 I RO d SRO Question ID: 5000049 Origin: New d Memory Level 4 days ago the reactor was manually triplied from 1009'2 power because of a feed control malfunction. The problem has been corrected and a reactor startup is in progress. Current conditions are as follows:

-Tcold 532°F

- Pressurizer Pressure 2250 psia

- All RCPs in operation

- Reactor is critical

- Power is stabilized at 1E-3% while recording critical data

'B' RCP upper motor guide bearing temperature is observed to be 195°F and rising at 1 "F every 3 minutes. Which of the choices below is immediately required?

A 'B' RCP will remain running, implement OP-2202. "Reactor Startup IPTE" B 'B' RCP will have to be stopped, implement EOP-2525, "Standard Post Trip Operations".

C 'B' RCP will remain running, implement ARP-2590B-:24, "RCP B Upper Guide Temp Hi" D 'B' RCP will have to be stopped, implement TS 3.4.1, "Coolant Loops and Coolant Circulation" Justification 1

I SRO ONLY OUESTION - Samples 55 43(5) Assessment of facility conditions and selection of appropriate procedures during normal abnormal, and emergency situations CHOICE (A) - NO LVRONG ARP-25906 provides necessary guidance RCP must be stopped

'IALID DISTRACTOR May not know the temperature limit on RCP upper guide bearing CHOICE (B) - YES ARP directs trip of reactor/turbine stop of RCP and reference to EOP 2525 at >194"F guide bearing temperature CHOICE (C) - NO WRONG Bearing temperature is above 194°F 4RP requircas immeddte lrtp of reactorlturbine and RCP VALID DISTRACTOR May think bearing temperature is ~ e i o wmaximum allowed value ARP does provide necessary quidance on high temperature CHOICE (D) - NO WRONG TS 3 4 1 for MODE 3 only requires one R L F in operntion No TS impact to stopping RCP VALID DISTRACTOR May think TS LCO will require dction with less than 4 RCPs in operation References 1 ARP-2590B-124. "RCP B Upper Guide Temp Hi" Revision 0 (3/4/04)

NRC KIA SystemlElA System 003 Reactor Coolant Pump System (RCPS)

Number A2 03 RO 2 7 SRO 3 1 CFR Link (CFR 41 5 / 43 5/ 45 3 I4511 3)

Ability to (a) predict the impacts of the following mai'L8ctionsor operations on the RCPS, and (b) based on those predictions, use procedures to correct control or mitigate the consequences of those malfunctions or operations Problems associated with RCP motors, including faulty motors and current and winding and bearing temperature problems NRC KIA Generic System Number RO SRO CFR Link

1. 93 I RO d SRO Question ID: 5000050 Origin: New 4 Memory Level The unit is in a refueling outage. A modification was installed on the condenser air removal system to troubleshoot air in-leakage problems..

If this installation was performed as a "temporary modification", what time limitation is associated with the modification?

A Unless required more frequently by SORC, the modification shall be audited within 90 days after installation and at least once per calendar quarter thereafter.

B Unless required more frequently by SORC, the modification shall be audited within 90 days after installation and at least once per calendar year thereafter.

C Unless authorized by Station Director, the modification shall be removed prior to the end of the next refuel outage or a time not to exceed 18 months, whichever is shorter.

D Unless authorized by Station Director, the modification shall be removed prior to the end of *,

the next refuel outage or a time not to exceed 24 months, whichever is shorter.

Justification SRO ONLY QUESTION - Samples 55.43(3)Facility liceiisee procedures iuquired to obtain authority for design and operating changes in the facility.

CHOICE (A) - NO WRONG, Auditing is required monthly.

VALID DISTRACTOR: Package processing is required within 90 days CHOICE (B) - NO WRONG Auditing is required monthly and additional requirements are imposed every six months VALID DISTRACTOR: Completed index sheets are retained until all temporary modifications for a respective calendar year are restored.

CHOICE (C) - NO WRONG The modification removal requirement is to not exceea 24 rnmtnc.

VALID DISTRACTOR: " R for refueling is defined as every 18 months in Tech Specs.

CHOICE (D) - YES Reauirement is as written References 1 W C - I 0 "Temporary Modifications". Revision 5 (5126104)(Py45 of 521 NRC KIA SystemlEIA System 055 Number RO SRO CFR Link NRC KIA Generic System 22 Equipment Control Number 2220 RO 2 2 SRO 3 3 CFR Link (CFR 43 5 I 4 5 13)

Knowledge of the process for managing troubleshoot,ig dctivities

1. 94 I RO g SRO Qusa*lon ::1: 5000Gsl Origin: New Memory Level The unit is operating normally at 100% power when the following indications are noted:

- pressurizer pressure 2180 psia and decreasing

- pressurizer level 68% and increasing slowly

- containment atmosphere process radiation levels increasing pressurizer relief tailpipe temperatures a i 135°F and steady

- NO acoustic monitor alarms Given these indications and assuming NO operator action has been taken, identify the problem and the required action.

A reactor coolant liquid leak, enter EOP-2525, "Standard Post Trip Actions" B pressurizer steam space leak, enter ADP-2568. "Reactor Coolant System Leak" C leaking power operated relief valve, enter ARP-2590B-043, "PORV RC-402 Open" D secondary steam system leak in containment, evter AOP-2575, "Rapid Downpower" Justification I

SRO ONLY QUESTION - Samples 55 43(5) Assessment of facility conditioris and selection of appropriate procedures during normal, abnormal, and emergency situations CHOICE (A) - NO WRONG: Liquid leak cannot account for the increasing pressurizer level.

VALID DISTRACTOR, Given an uncontrollable RCS leak, the operator would have to trip the reactor and respond pel EOPs.

CHOICE (6) - Y E S Indications are consistent with a pressurizer steam sydce leak.

CHOICE (C) - NO WRONG. PORV leakage would result in tailpipe ternperatlira consistenr with an isenthalpic expansion across the PORV Slightly elevated tailpipe temperature is caused by steam leak in vicinity of temperature sensor VALID DISTRACTOR: PORV leakage would result in elevated tailpipe ienoerature CHOICE (D) - NO WRONG. Secondary steam leak should not result in elevated radiation levels VALID DISTRACTOR: A rapid downpower could reasonably be attempted References 1

I AOP-2568. "Reactor Coolant System Leak" Revision 7 (5/27/03)

NRC KIA SystemIEIA System 007 Number RO SRO CFR Link NRC KIA Generic System 21 Conduct of Operations Number 217 RO 3 ' SRO 4 4 CFRLink (CFR 4 3 5 1 4 5 1 2 / 4 5 13)

"Ability to evaluate plant performance and make oC.erational~ u d g m r ! ~ based

's on operatlng characteristics reactor behavior and instrument interpretation "

1. 95 I RO v SRO Question ID: 5000352 Origin: New fl Memory Level An automatic reactor trip and SIAS have x c u r r e d while operating at power 30 minutes into the event, the following conditions exist.

- Crew is performing EOP-2532 "Loss of Coolant Accident"

- CET temp 320°F

- pressurizer pressure 28 psia

- containment pressure 9 psig

- SIAS, CIAS, EBFAS, MSI and CSAS have actuated

- all equipment is functioning per design After several RWST LEVEL CH LO/LO alarms are received. the operator determines that SRAS has NOT actuated. Choose the NEXT correct action in response to the SRAS failure.

A Manually stop both Charging pumps.

B Manually close both Gravity Feed valves (CH-508/509)

C Manually stop both Low Pressure Safety Injection pumps. d D Manually close both RWST header outlet valves (CS-13 IA/B).

Justification I SRO ONLY QUESTION - Samples 55 43(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations CHOICE (A) - NO WRONG Charging pumps stopped after ensuring automatic actions of SRAS VALID DISTRACTOR Charging pumps are stopped after SRAS CHOICE (B) - NO WRONG Gravity feed valves are closed after ens~riiigatitomatic actions of SRAS VALID DISTRACTOR Gravity feed valves are closed after SRAS CHOICE (C) - YES Directed in Step 47 of EOP-2532 CHOICE (D) - NO WRONG RWST outlet valves closed after ensuring automatic actions 01 3RAS VALID DISTRACTOR RWST outlet valves closed after SRAS References 1 EOP-2532. "Loss of Coolant Accident", Revision 23 (3/31/04) (Pg 39, 40 of 95)

NRC KIA SysternIElA System 026 Containment Spray System (CSS)

Number A2.02 RO 4.2' SRO 4 4' CFR Link (CFR: 41 5 / 43 5 145.3 / 45 13)

Ability to (a) predict the impacts of the following malfunctions or operations on the CSS: and (b) based on those predictions. use procedures to correct, control.or mitigate the consequences of those malfunctions or operations:

Failure of automatic recirculation transfer NRC KIA Generic System Number RO SRO CFR Link

1. 96 1 RO *I SRO Question ID: 0054565 Origin: Bank *I Memory Level Following a plant trip, the Emergency Diesel Generators (EDG's) are supplying their respective Bus 24C and 24D due to a failure to transfer to the Reserve Station Service Transformer (RSST).

While still in EOP-2528, "Loss of Offsite PowerlLoss of Forced Circulation", the RSST is now available to supply Facility 1 electrical loads. NO electrical faults exist.

Based on these conditions, which of the following statements identifies the procedure and the CORRECT sequence of steps needed to restore Bus 24C to a normal post-trip alignment?

A Per EOP-2541, Appendix 23, "Restoring Electrical Power", reset ESAS UV signal, parallel *I RSST to 24C, open DIG breaker, close Btis 24A-24C tie breaker.

B Per AOP-2502C, "Loss of Vital 4.16 kV Bus 24C", reset ESAS UV signal, parallel RSST to 24C, open D/G breaker, close Bus 24A-24C tie breaker C Per EOP-2541, Appendix 23. "Restoring Electrical Power", reset ESAS UV signal. parallel RSST to 24C, close Bus 24A-24C tie breaker. open D/G breaker.

D Per AOP-2502C, "Loss of Vital 4.16 kV Bus 24C", reset ESAS UV signal, parallel RSST to 24C. close Bus 24A-24C tie breaker, open DIG breaker.

SRO ONLY QUESTION - Samples 55 43(5) Assessment of facility conditions and selection of appropriate procedures c'urinq normal abnormal and emergency situations CHOICE (A) -YES These actions are directed in the listed sequence in Appendix 23 CHOICE (3) NO WRONG AOP-2502C does not provide step sequence for re-energizing from normal source VALID DISTRACTOR AOP-2502C provides steps for energizing bus from the emergency diesel generator CHOICE (C) - NO WRONG Sequence is not correct Tie breaker is not closed until after DIG breaker is open VALID DISTRACTOR Plausible to think tie breaker should be closed before opening DIG breaker CHOICE (D) - NO WRONG AOP-2502C does not provide step sequence for re-energizing from normal source VALID DISTRACTOR Plausible to think that AOP would provide specific guidance for transferring bus back to normal source References 1 OP-2343. "4160 Volt Electrical System", Revision 20 (9/9/04) Section 4 20 "Restoring Bus 24C to Unit 2 RSST with Emergency Diesel Generator Supplying" (Pg 42 of 71)

NRC KIA SystemlElA System 062 Number RO SRO CFR Link NRC KIA Generic System 22 Equipment Control Number 222 RO 4 0 SRO J 5 CFR Link (CFR 45 2)

Ability to manipulate the console controls as required rn operate the faLility between shutdown and designated power levels

1. 1. 97 I RO d SRO Question ID: 5000053 Origin: New Memory Level Which of the following describes a condition required to be reported to the NRC under 10CFR50.72 and the correct time limit for reporting?

A 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> report due to deviation from the plant Technical Specifications authorized pursuant to d 10CFR50.54(x)

B 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> report due to a condition that could have prevented fulfillment of a safety function needed to mitigate consequences of an accident C 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> report due to failure to perform required surveillance test within technical specification allowable time limits D 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> report due to the nuclear power plant being in an unanalyzed condition that significantl) degrades plant safety Justification SRO ONLY QUESTION - Samples 55.43(1 ) Conditions and limitations in the facility license CHOICE (A) - Y E S 10CFR50 72(b)(l) requires a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> ENS notification if provisions of CFR50.54(x) invoked.

CHOICE (3) - NO WRONG: Fulfillment of a safety function is an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> notification under 10CFR50,72(b)(3)(v)

VALID DISTRACTOR: Plausible that safety function needed for acciaent mitigation would be a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> report CHOICE (C) - NO WRONG. Failure to perform a surveillance test within allowable time Iiniits is not 1, 4 or 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> reportable VALID DISTRACTOR: Plausible that missed surveillaice would be reportable as a violation of technical specification requirements CHOICE (D) - NO WRONG: 10CFR50,72(b)(3)requires an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> notificatior VALID DISTRACTOR: Plausible that an unanalyzed cond. '4Jn would I-equire a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> report References 1

J RAC-14. "Non-Emergency Station Events". Reiision 1 (3/24/04), Attachment 1 (Sheet 1, 2, 3 of 17)

NRC KIA SystemIElA System 076 Number RO SRO CFR Link NRC KIA Generic System 24 Emergency Procedures /Plan Number 2 4 30 RO 2 2 SRO 3 6 CFR Link (CFR, 43 5 / 4 5 1 1 )

Knowledge of which events related to system operations/status should be reported to outside agencies

1. 98 I RO V SRO Quesr,ofi ID: 500005ci Origin: Bank Memory Level On your shift, a monthly surveillance item is discovered overdue. Required due date was November 28th. Assume today is December 3rd and the performance of the Surveillance Test has begun. The previous surveillance tests for this componentkystem were Due and Completed as shown below.

Due Date: Completed Date:

- August 25 - August 28

- September 28 - October 1

- November 1 - October 28 Which ONE of the following statements describes the status of the componentkystern and the justification for that status?

A The componentlsystem is INOPERABLE because 3.25times the time interval for three consecutive tests has been exceeded.

B The componentlsystem is INOPERABLE because mcre than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> have elapsed from the due date plus the allowed extension.

C The componenvsystem is OPERABLE because the Technical Specifications allow time from previous tests to be carried forward.

D The componenvsystem is OPERABLE because the Technical Specifications allow a time 4 extension which has not been exceeded.

Justification I

SRO ONLY QUESTION - Samples 55 43(2) Facility cperating limitations in the technical specifications and their bases CHOICE (A) - NO

\/VRONG 25% extension still allows time to perform the L L !veillance VALID DISTRACTOR Plausible that consecutive t-st frequency is to be taken into account CHOICE (B) - NO WRONG Extension allows for 7 days from due date to perf-lrm lest Only 5 days have elapsed VALID DISTRACTOR TS 4 0 3 allows a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> exterwon from time aixovered missed surveillance CHOICE (C) NO WRONG TS 4 0 2 basis states "it is not intended that this provision be used repeatedly as a convenience to extend surveillance intervals beyond that specified for surveillances that are not performed during refueling outdges '

VALID DISTRACTOR Plausible to assume that testing frequency includes consideration of previous test history CHOICE (D) YES Per TS 4 0 2 each Surveillance Requirement shall be performed within the specified time interval with a maximum allowable extension not to exceed 25% of the surveillance time interval Per TS 4 0 3 if it IS discovered that a Surveillance was not performed within its specified surveillance interval then compliance with the requirement to declare the Limiting Condition for Operation not met may be delayed from the time of discovery up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified surveillance interval, whichever is greater References 1 Technical Specifications, Section 314 0 "Applicability" .*.,nendmsnt 27' (Pg 314 0-2) 2 Source INPO Bank - Q# 22908 - Used at DC Cook 1, 12/9/2002 NRC KIA SystemlEIA System 002 Number RO SRO CFR Link NRC KIA Generic System 22 Equipment Control Number 2 2 12 RO 3 0 SRO 3 4 CFR Link (CFR, 41 10 / 45 13)

Knowledge of surveillance procedures

1. 99 1 RO SRO Questiot ID 500005'7 Origin: New Memory Level The unit is at 100% power. SG LEVEL SETPOINT DEVIATION HVLO alarm is received (Panel C-05, D-I 6) The operator reports an apparent fdilure low of #I Steam Generator Pressure Transmitter, PT-4243 (steam flow pressure compensation channel)

What automatic plant response is expecled and what action is required?

A SG level has decreased, per OP-2385, "FEEDWATER CONTROL SYSTEM OPERATION",

take manual FRV control, select alternate steam pressure channel, return FRV to auto.

B SG level has increased, per OP-2385. "FEEDWATER CONTROL SYSTEM OPERATION",

take manual FRV control, select alternate steam pressure channel, return FRV to auto.

C SG level has decreased, per ARP-2590D-064, "SG LEVEL SETPOINT DEVIATION Hl/LO". d take manual FRV control, maintain level in manual until pressure transmitter repaired.

D SG level has increased, per ARP-2590D-064. "SG LE',,/EL SETPOINT DEVIATION HIILO".

take manual FRV control, maintain leve! in manual until pressure transmitter repaired.

Justification SRO ONLY QUESTION - Samples 55.43(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations CHOICE (A) - NO WRONG: Alternate channel unavailable due to pressure compensation failure.

VALID DISTRACTOR: A loss of one channel input typically requires selection of an alternate channel CHOICE (B) - NO WRONG. OP-2385 does not provide the guidance for response to a failed channel VALID DISTRACTOR Plausible that OP-2385 would provide the necrc;sary guidance CHOICE (C) - Y E S PT4243 provides pressure compensation to both control channel steam flow instruments on # I MSL. A failure low will reduce steam flow on both channels, resulting in the 3-elernsnl control system throttling closed FRVs Level will decrease. ARP-2590D-064 directs FRV control in manual. L'<oiiic have to maintain FRV in manual until pressure cornDensation restored.

CHOICE (D) - NO WRONG Level will decrease because the control svstem will percieve a reduction in steam flow VALID DISTRACTOR May think that low compensation pressure will result in higher indicated steam flow References 1 ARP-2590D-064. "SG LEVEL SETPOINT DE\,'IATION Hlil.0" Revision 000 (2112/04) 2 25203-26002. "MAIN STEAM FROM GENERATORS" SH 1 OF 5 Revision 57 (3/7/02) (K-8)

NRC KIA SystemIElA System 035 Steam Generator System (SIGS)

Number AZO3 RO 3 4 SRO 3 6 CFRLink (CFR 41 5 1 4 3 5 1 4 5 3 1 4 5 5 )

Ability to (a) predict the impacts of the following nialfuncti ;ns or operations on the GS, and (b) based on those predictions. use procedures to correct control or mitigate the consequences of those malfunctions or operations Pressurellevel transmitter failure NRC KIA Generic System Number RO SRO CFR Link

1. 100 1 RO 4 SRO Questiocl ID: 0053324 Origin: Mod Memory Level You are the Fuel Handling SRO. The refueling cavity is filled to 36' 8" and core alterations are in progress on Unit 2. The upper manways an SG #2 are off for inspection of the can decks. Both LPSl Pumps are running on SDC.

Which of the following activities CANNOT be authorized?

A remove pressurizer power operated relief valve B disassemble #2 main steam isolation valve 4 C shutdown 'A' LPSl and place in standby D lower refueling water storage tank level by 5%

Justification SRO ONLY QUESTION - Samples 55 43(7) Fuel hanciling facilities and procedures CHOICE (A) NO WRONG PORV removal does not affect refueling or containment integrity VALID DISTRACTOR Plausible that work could I ot be authorized during core alterations CHOICE (5) -YES With manways removed, disassembly of #2 MSlV allows a direct path to atmosphere Containment integrity is required when performing core alterations or moving irradiate? fuel in Containmant CHOICE (C) - NO WRONG Only 1 LPSl required to be in operation with full kiel pool VALID DISTRACTOR Plausible that not allowed to stop LPSl pump r'uring core alterations CHOICE (D)- NO WRONG Core alterations do not impose any limit >n RWST level \$henrefuel pool is filled VALID DISTRACTOR Plausible that limits are imposed on RWST ieve dur ng core alterations References 1

1 SP-2614B-002 "Containment Closure With SG Secondary Side Open to Containment" Revision 3 (4/20/01) (Pg 3 of 3) 2 Technical Specification 3 9 4, "Containment Penetrations" Amenament 745 (Pg 3/4 9-4)

NRC K/A SystemlEIA System Number RO SRO CFR Link NRC KIA Generic System 22 Equipment Control Number 2218 RO 2 3 SRO 3 6 CFR Link (CFR 43 5 / 4 5 13)

Knowledge of the process for managing maintenance activities dut ins shutdown operations