ML12353A170
| ML12353A170 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 09/14/2012 |
| From: | Operations Branch I |
| To: | Public Service Enterprise Group |
| Jackson D | |
| References | |
| TAC U01859 | |
| Download: ML12353A170 (127) | |
Text
the following conditions:
Unit 1 is operating at 100% power.
1E 460 Volt bus is deenergized following a trip of its feed breaker.
Tagging is in progress to allow troubleshooting of 1E 460 Volt bus.
I
,The 1G 460 Volt bus feed breaker is opened instead of the 1E 460 Volt bus feed breaker, deenergizing the 1G 460 Volt bus.
~~iCh of the followif!~!)qibe~_how this will affect Control Rods?
.~=:.-.-
~ I.AL.L. control rods w.. ill drop into th.e. core d.ue to the loss of the o.nlYoperating Rod Drive Motor-
~nerator (RDM(3) set.
.... ___________----!.l
[I into the core due to the loss of one of the two RDMG sets.
into the core since both RDMG sets remains
@] ~~c_o_n_tr()l_ro_d~_\\,Nill drop into t~e core since one RDfy1G set Eemain!)=r=u=n~ni;;:::n::g=.:::;:::::=.=======~=".
IAnswerlli-,lsxarnLevell ~u JICognitive leVell ~~ti~IJ~IFacility;lIsalem 1&2 JIExam Date: I
.. *.**1~3/201~1
~ [gm.E:'.r:gen~y and Abnormal Plant Evolutions IRO.Group I*
ISRO Group I lOOOoo:fK20s~~
fRe.x;rd Numbe~
.rn,'"<r""n Control Rod and the tnll,'"'u/in...*
~5.41.b(6} Unit one Rod Drive Motor Generator sets-are powered from 2 (1 E Bkr 15X and 1F Bkr 15X) of L--___--'Ithe 4 non vital 460V buses (1E, 1F, 1G, 1H). Normally, both RDMG sets are operating in parallell for redundancy in supplying power through the Reactor Trip Breakers to the control rods. With 1E 460 Volt bus deenergized from the initial conditions in the stem, removing power from the 1G 460 Volt bus will Icause the only operating RDMG set to lose power. The loss of power to the control rods will cause the
- stationary gripper coils to un-grip, and ALL control rods will drop into the core. B is incorrect because no RDMG sets will be running, and even if one remained running no rods would drop. C is incorrect becaus I,both RDMG sets do not have power. 0 is incorrect because no RDMG sets are running. The mOdificat~.oni to this question was required since there is no longer a negative flux rate Reactor trip, and the question IoriginCl"y asked whether the reactor would tripor not. Modified to al'ik about what happens to control r()ds.
Reference Title*.
I Unit 1 1E 460V One Line Describe the function of the following components and how their normal and abnormal operation affects the Rod Control and Position Indication Systems:
Rod Cluster Control Assembly (RCCAl Control Rod Drive Mechanism (CRDM)
Rod Drive MG Sets Reactor Trip and Trip Bypass breakers Reactor Control Unit Power Cabinets Logic Cabinet components:
Pulser Master Cycler Slave Cyclers Bank Overlap Unit
- h.
DC Hold Cabinet
- i.
Rod Position Indicator (RPI) Coils
- j.
Signal Conditioning Modules
- k.
PulseutouAnalog (P to Al Converters 14,
Rod Bottom Bistables Rod Insertion Limit Comparator Counters RODSOOE007 L ______**
NCT State the power supplies to the following Rod Control and Position Indication systems components:
Rod Drive MG Sets Reactor Trip and Trip Bypass breakers Power Cabinets Logic Cabinet DC Hold Cabinet IMaterial Required for Examination Question Source:
IQuestion Modification Method:
Editorially Modified Friday,~eptember 14, 2012 12:0!:11Pt0:]
i
of the following identifies the MINIMUM required AFW flow following a Rx trip, and the that amount?
EI Ibm/hr. Ensures SG tubes are covered to prevent SG pressure drop and Ibm/hr. Ensures SG tubes are covered to prevent SG pressure drop and excessive
~ ~:I~:~es sufficient flow for deca~heat removalplus allo::es for nor::=J
@J Ibm/hr. Provides sufficient flow for decay heat removal plus allowances for normal IAnswer I IExam Level I ICognitive Level I
!Facility: I Salem.1 & 2 IIexam Date: 1 m
1213/20121 EJ ~Jgency and!\\bnormaIPlantEvolutions_! IRQ Group I[=:1] ISRQ Group I[]
OOOOOZK106 IReactor Trip.
]
=======~============~
ill] Knowledge of theoperation<:il implications ofthe following cOl"lceptsas they appiYto Reactor.~"""""--.*
IEK1.o61 ~elationship of emergency fE'!E'!dwater flow to§IG and decay ~E'!at removal follc)l,>\\Iing reactor triem=-:J I 3.7
'mm~_~mmm, 55.41.b(4) At step 3 of TRIP-2, Reactor Trip Response, operators are asked if total AFW flow is >22E4 I~bm/hr. The bases document states that this amount of flow is.,."the minimum safeguards AFW flow requirement for heat removal plus allowances for normal channel accuracy (typically one AFW pump capac. ity at.. de.Sig.n, p.ressure..,)". T,he 44E41bm/hr ChO.ices are.Plausible becaus.e...i,t is the minimum required AFW flow in FRSM-1. The SG pressure drop bases is found in SGTR series procedures to prevent a rapid pressure drop in a ruptured sG which would reinitiate primary to secondary 1000....'<:;
Reference Title
[Reactor Trip Response Describe the basis for each step. caution, note, and continuous action summary item in 2-EOP-TRIP-2 1Material Required for Examination IQ,:,estiOQ.$Qurce: " 'I IQuestion Modificetion Method:
m IQuestion Source. CClmments:
___~, ~__________~,
I C_
14,
[GiVen the following conditions:
Unit 2 is operating at 100% power when a catastrophic failure of RCS loop 21 cold leg piping occurs.
i-RCS pressure rapidly dropped to 35 psig.
22 RHR pump failed to start, and remains stopped.
Initial RWST level was 41.1 feet.
IOf the following, W.. hiCh is CLOSEST to the the time available from the failure until the swap to Cold
~eg recirc will be re_q~~ed?
~
~~===~.
~
=---~.~------------~.~--......~------------------------------------- -.----------...---------~
@J r24 min-utes.
--~---.-........
@]
minutes.
!Answer ! ~] !Exam Level! Li !Cognitlve Level !1.Application-~ !FaCllity: ![~alem 1 & 2
--!Exam. Date: !~--=-~~l
[Emergerlcy and j\\bnorl11al Plant Evolu!ions
[000011K202
---~---.
-"--'--'--____.......1 l=La=-r:,:,g-",e-=B=r=~=~k,~.....=L..=O=C:-,-A-=====::::::::========::::::::::::::::::--:-___.~_:::::::
...-=-=-=.::::::::====:::::::===::::::::==~.IReCOrd Num~_er_1 '---.......___-----"
nO~"'le(lae of the interrelations between Break LOCA and the Tnll,r'l\\AlI.nn*
rc:.,55.41.b(8) Question stem describes LBLOCAwith power. With the RCS at35 psig, all ECCS pumps will J' 1..-___--1 be injecting at their maximum rate. The flow rates used are: Charging pumps 2x560= 1120gpm(page 23); SI pumps 2x675= 1350gpm (page 26); RHR 1x4500= 4500 (page 34); and Containment Spray pum
. flow of 2x2600=5200. (page 17)
So, 1120+ 1350+4500+5200= 12,170 gpm total. With the initial RWST I of 41.1' equating to 370,000 gallons, and 15.2' level of 150,000, you need to pump in 220,000 That's 18.08 minutes. Distracter A is if the failed RHR pump is included. Distracter D is the time would take to in the entire RWST volume. Distracter C is the time if CS pump flow is not included.
Reference Title* '.
For the analyzed transients/accidents:
A Inadvertent depressurization of the RCS - Stuck open safety (Condition" - Faults of Moderate Frequency)
B.
Small Break LOCA 4" diameter cold leg break (Condition'" Infrequent Fault)
C.
Large Break LOCA Double-ended break of an RCS cold leg (Condition IV 0 Limiting Fault)
- 1. Describe the analysis assumptions
- 2. Describe the protective features that mitigate the event
- 3. Describe the expected plant response
- 4. State whether the indicates fuel describe the fuel failure mechanism
!Material Required for Examination Tank Capacity Data. Page 28 of 34, RWST Tank I=.!Q!""ue_st~ion.;.... *"=s.;..()u....r_ce.;..l~~=::=~. ';::.====-=-======== IQUestion MOdifi~tion Method: **lLEdit~~f\\.1()(:li~=~=-===;
Vision Q48704. Modified correct answer from 18 to 19 minutes to reduce probablility of candidate not choosing 1..-________--'___---1 correct answer because it was LESS than calculated time. Also modified stem to say what is the CLOSEST Spr.tpIY,l'Ipr 14, 2012 12:07:11
following conditions:
Unit 2 is operating at 30% power, steady state.
OHA 0-29,22 RCP BKR OPEN/FLO LO is received.
All 22 loop RC flows are 85% and dropping.
The red START bezel for 22 RCP is illuminated.
i-The reactor has NOT tripped.
IYVhiCh of the follow~~91~lifi~s what has_occurred?
................. ~=====~====~
~
~~..................
............ --------------------~..................
............ --------------------------~
a leak.
IFacility: l~m1 ~2
_jlexam Date: 1m ___ 12/3/2012[
- ~[I~RO~G~r~ou~p~1;[= 1[lsRO Group I[JJ Malfunctions and the 55.41.b(3,7) With a Rep shaft shear, there is no event that would cause the Rep breaker to open. For a.;..;;;.;.;;;";';";~_---Ithis reason, that is why the START bezel will still be illuminated, even though loop flows are all dropping.
Distracter is incorrect because between 10%(P-10) and 36%(P-8), 1/4 ReS loop 10 flow will NOT cause a
!Rx trip, the coincidence is 2/4. Distracter is incorrect because there are 3 low pressure flow taps, and 1 icommon high pressure flow tap. Distracter is incorrect because a seized Rep shaft would cause its Isupply breaker to trip on overcurrent. The indication in the stem is that the breaker is closed. Answer is
- correct because a sheared shaft would cause that loop flow to drop, even while the bezel indication showed the breaker is still closed.
Reference Title*
Window D Learning Objectives RCPUMPE008 LOR Identify and describe the Control Room controls, indications, and alarms associated with the Reactor Coolant Pump, including:
The Control Room location of Reactor Coolant Pump control bezels and indications. (Licensed Operator & STA only)
The function of each Reactor Coolant Pump Control Room control and indication. (Licensed Operator & STA only)
The effect each Reactor Coolant Pump control has upon Reactor Coolant Pump components and operation. (Licensed Operator & STA only)
The plant conditions or permissives required for Reactor Coolant Pump Control Room controls to perform their intended function. (Licensed Operator & STA only)
The set~~nts associated with the~(i)<lc:t()r Coolant Pump control room alarms. (Licensed ()~!3.r<ltor & STA only) the setpoints, coincidence, blocks and permissives for automatic actuations associated with the L.Il.t" ""'U Operator &STA only)
IMaterial Required for Examination Question Source:
Exam Bank IQuestion Modification Method:
'--_________--1lviSiOn9~?312Iast NRC Exam usage4exams ago (12f2006)
IFrid8Y:September 14, 2012 12:07:11 PM
conditions:
Unit 1 is operating at 100% power.
13 Charging pump is in service.
Normal letdown is in service.
The 1A 4KV to 460V bus feeder breaker opens, deenergizing the 1A 460/230V bus.
i\\Nith N~operator action, vyhich of th~f()nowing identifief)§I<:;~g~ce, if§ll'ly,~()'UI1is event?
~
level wi! remain stable.
~
level wiIL~~ risi/"lg at :J% per minute.
minute.
IExam Levell
.~lc~O~gn~it~iv~e~.L~eve~1~I~~;~~=~= !Facility: 1~1&2_IExam Date: I 1213/2012]
~ ~rgen~yand At>normal PlantE;:yolutions IIROGroup I 0 ISROGroup 1:3
~OOQg~1Q3 Q22
] it.:OsSof 8~actor Coolant Makeup
- -IIRecordNumber I L___~
nnUlQ"nA of the operational implications of the following concepts as they apply to Loss of Reactor t'h!"rninn flow and PZR level
.b(7) Normal power operation has the Positive Diplacment charging pump in service (13), which is from 1A 460 volt bus. The two centrifugal charging pumps, (11 and 12) are powered from Band 4KV buses respectively. The thumbrule for PZR level at NOT is 75 gallons per % of level.(Page 15 of iLesson Plan) When the 1A 460 volt bus is deenergized, the breaker for the 13 charging pump does NOT Itrip, it does NOT have a UV trip. Therefore, the interlock for automatically closing the 3 letdown oriface isolation valves is not satisfied (all 3 charging pump breakers open.) Letdown reamains in service at 75 gpm, which is normal at power letdown flow. This will cause PZR level to lower at 1 % per minute. PZR level would remain stable if it is thought that a charging pump remains in operation because of not knowing correct power supplies. The VCT rule of thumb is 20 gallons per % level. With lewtdown flow Istill entering the VCT at 75 gpm and no chargin gpump taking suction and pumping from VCT, VCT level will be RISING at -4% per minute, not lowering. The 1 % VCT dstracter is if there is confusion about will be at 1 %
minute.
Reference Title Loss of Charging and PRT Lesson Plan IMaterial Required for EXjlmination IQuestion Source: I IQuestion Modification MethOd:
IQuestion Source Comments: II m
_ _ _ _ _ _ _mmmm._ _
~----
14,201212:07:11 PM
of the following is an automatic response for Component Cooling Water system Ito a Safety Injection signal on a LOCA, and why?
Assume ~ontainr'!lent pressure peaks at 5 El '2CC215 and2CC113, Excess Letdown Heat Exchanger CCW isolation valves, receive a lclo~e signal to ensure t~non-ess~r-l!ial containment flowp(;lth is isolated.
g 9
~.121 and 22C......1...
.......I....iOrl.V.*.81.Ves~receive an open.*.*.*****Si... nalto e.n
... rf3that 10n...
-I
......C.....6,R.*.*.HR HX CCW is.O...a..t
.. su term cooling of theBCS is in service wher1Jhe swap to Cold Leg Recirc is reqyired.
@J 5 and 2CC113, Excess Letdown Heat Exchanger CCW isolation valves, receive a to ensure ALL CCW and return from the containment is isolated.
22CC16, RHR HX CCW isolation valves, receive an open signal to ensure that RHR do not overheat if the RCS remains above the shutoff heat of the RHR IAnswer I~J IExam Levell [i3:=J ICognitive Level IMemqrv I I Facility: I~~& 2 IIExam Date: I 12/3/2012
~J:gency and Abnormal Plant EV91utions IIRQ Group ICi ISRQ Group I
!QQQ()1QK302
_____J Loss of Co.r:nPonimt Cooling\\.ryater I [Record Number '---__---1 IAKi *iKnoWiedgeof the reasons for the foliowir"lg responses as they appl¥io LOS!; of Component C()()Iir1gWater]
iAK3.021 [he automati~~~@li.gnments) vyit~in the CC~resulting fr9r11 the actuation of theE5~f\\~ill ~
.b(7,8) The 51 signal sends a close signal to the CC215 and CC113, Excess Letdown HX isolation as they are Containment Phase A isolation valves. The purpose of closing Phase A isolation
,valves is to ensure all non-essential containment penetrations are isolated on a 51. B is incorrect becaus CC16s do not receive an open signal until the ARM PB is depressed and the RW5T level is 15.2' and manual alignment is required to place ECC5 in CLR. C is incorrect because ALL CCW supply and return are not isolated on a 51 signal, the RCP CCW is still being supplied until a Phase B signal at 15 psig in cont. 0 is incorrect because the CC16s do not receive an open signal until the ARM PB is depressed and the RW5T level is 15.2', and the RHR pumps are cooled by either flow through the pump from RW5T (LBLOCA)()r recirc flow (5BLOCA ul'1til pp is SID).
Reference Title Component CoolingLe~~on Plan Learning Obiectives CCWOOOE004 Describe the function of the following components and how their normal and abnormal operation affects the Component Water System:
(:nnnnm1.. nt Cooling Water Surge Tank r:"nnn",'"nl Cooling Water Pumps Con100flAni Cooling Water Heat Exchangers
'"nl;,tin,nlC":nntirnl Valves RCP Thermal Barrier Discharge Valve
& 118, RCP Cooling Water Inlet Valves
& 187, RCP Bearing Cooling Outlet Valves
& 113, Excess Letdown Heat Exchanger CCW Inlet & Outlet Valves RHR Heat Exchanger Outlet Isolation Valves
& 18, CCW Pump Suction Cross-connect Valves Component Cooling Water Pump Outlet
&31, Component Cooling Heat Exchanger Outlet to Auxiliary Header (Non-safety Related Header Isolation Valves)
, Letdown Temperature Control Valve Surge Tank Vent Valve
, RCP Thermal Barrier Discharge Flow Control Valve Radiation Monitors IMaterial Required for Examination Friday. September 14, 201212:07:11 PM ml
Exam Bank IQuestion Modification Method: IrIC-o-n-ce-p-t-U-s-e---d-****--_***_
'--------------l IVision Q42744, m~e 2 and~and added why to stem.
Salem Unit 1 is operating at 100% power when the PZR Master Pressure Controller demand fails
,high.
'How will this failure affect PZR pressure control components in AUTO?
PZR B/U heaters will be PZR Spray Valves will be __ BOTH PZR PORV's will I
be
~...-
E]
shut.
~
shut.
shut.
§] L...... _____~___ ___________________________________________________________________.....
~
~
~~~~~~~~~C~~=::;-~IFacility: I~m 1 & 2m __llexam Date: Lm---12~
ISRO Group I 1000027A101 I
IP!lSsuri:ZlL Pressure Colltrol Malfullction
-=, ~dNumber*1 nn'~r"'lt~ and / or monitor the tn!ln'AI,n to Pressurizer Pressure Control and PORVs
.b(7) Master Pressure controller scale runs from 0-100% demand. With MPC demand failing high, output of the MPC calls for maximum spray, and no backup heaters. The PZR PORVs are controlled of the MPC demand from actual PZR pressure channels 1-4. Since actual PZR pressure i PORVs have no demand Reference Title Pressurizer Pressure and Level Control Power Relief Valves IQuestion Modification Method:
Changed from MPC failing low to failing high, which changes correct answer to one of the 201212:07:11 I
Page 9 of 90 J
I IGiVen the following conditions:
i-Unit 1 experienced an ATWT in MODE 1 where both Train "A" and "B" Reactor Trip Breakers (RTBs) failed to open.
The reactor was tripped when the pressurizer heater buses were deenergized.
The RTBs remain shut.
An momentary inadvertent safety injection signal was generated and has cleared.
Which of the following describes the impact of depressing the Train A and Train B RESET SI IPUshbuttons on 2CC1??
-rhe SI signal will... ~~~~~~~~~~~~~~~~~~~~~~~~~~~~
~ Ireset, and Auto SI will be blocked.
~ Ireset, and Auto SI will NOT be blocked.
I I
ElINOT reset, and Auto SI will be blocked.
@] INOT reset, and Auto SI will NOT be blocked.
IAnswer I~J IExam Levelll&::=:J ICognitive Level IIApplication
~ ~ergency and Abnormal Plant Evolutions IIRO Gro 1029
] IAnticipated Transient Without Scram IIFacility: IISalem 1& 2 up II1J ISRO Group I OJ I
. n
_ _J IIExam Date: I 12/3/20121 IQQmg~I5~Q6 I I Record Number 11"---__...J8 1
~ IKnowledge of the interrelations between Anticipated Transient Without Scram and the following:
I IAK2.061l§:reakers, rela'y~, and disconnects 112.9*113.1 *1
==============================================~==~~
Explanation of 55.41.b(7) The SI signal can be reset as shown on 221057 grid F-2 AND box and downstream LATCH
- 1. Manually pushing the reset pb, (
A_ns_w_e_r__..... :RESET. This shows the SI can be reset if 2 conditions are present.
MANUAL SI RESET AND BLOCK), and the 1-2 minute TD has timed out after the SI signal was generated. 2. The LATCH-RESET button is reset. Right next to that AND box is another AND box, whose purpose is to block a second SI after the Rx has been tripped. Since the Rx has not tripped, there is no output from this AND box, and a NO signal will be input into the NOR box to the right of it. The 0
,signal into this box will produce a 1 signal out of this box, which is one of 2 inputs to the AND box to its right. This AND box needs 2 signals to produce an output (Auto SI). The second input into this AND box is a safety injection from any of the 4 auto SI signals above it. Hi Steamline Flow with 10 steamline pressure or 10-10 Tavg, High Steamline Differential pressure, PZR low pressure, or Containment hi Reference Title IRPS Safeguards Actuation Signals I
========================~
I Learning Objectives RXPROTE027 LOR Given a Reactor protection System Failure, predict the effect of the Reactor protection System failure on the following:
(Licensed Operator and STA Only) a)
Control Rod Drive System b)
Main Turbine/Generator c)
Engineering Safeguards System d)
Reactor Fuel e)
Containment IMaterial Required for Examination II'---__________________________~
I Friday, September 14, 201212:07:11 PM I
I Page 10 of 90. I
IQuestion Source: I !Facility I:;xam Bank IIQuestion Modification Method:
IISignificantly Modified IQuestion Source Comments:
I Vision 044139. Modifed inadvertent SI signal from remaining present to being clear This changes correct naswer to one of the distracters.
Friday, September 14, 2012 12:07:11 PM I
the following conditions:
Unit 2 will be performing a controlled shutdown and cooldown from 100% power due to a 5 gpm tube leak on 22 SG lAW S2.0P-AB.SG-0001, Steam Generator Tube Leak.
After completing the Immediate Actions of EOP-TRIP-1, Reactor Trip or Safety Injection, following the Rx trip, the RO reports that control rod 202 is stuck in the fully withdrawn position.
Which of the following identifies the action, if any, the crew will perform in response to the stuck lrod?.~
E1ll!'i!@te a r~ld~!>gration for 35 minutes during~~rf<:>iinarlce of EOP-TRIP-2.
~
a rapid boration for 35 minutes in S2.0P-AB.SG-0001 after exiting the TRIP series actions are required for a single stuck rod because SOM for the cooldown to
",,,t*,,,....,. are required for a single stuck rod until the Auto SI Block is performed during ReS to 1900 are used in with EOPs.
- 0) Step 3.26.H in AB.SG states to trip the turbine, then trip the Rx at 20% power during the The 5 gpm size of the leak will allow for a controlled shutdown, and will also allow the crew to Itr!l,,,,,,itinn to TRIP-2. There are no SGTR diagnostic steps in TRIP-2 that would cause a transition to
. AB.SG would be re-entered at step 3.27 following exit of TRIP-2, and 3.28 directs rapid borati01 stuck rod for 35 minutes. The rapid boration will be initiated before any depressurization starts so the distracter is incorrect.
None Determine the expected alarms and indications Describe the analysis assumptions Describe the protective features that mitigate the event.
Describe the IMaterial Required for Examination Ques~ion Source:
.__ ~-IQuestion Modification Method:
I iEdiiori-an-y-M-Odified=*****====i 1...-__.;.;....;._'--___...... ~~5_0_37___________________________
[J'age 12 of 90
following conditions:
Unit 2 is operating at 100% power, MOL.
i_ The Condensate Polisher is in service -full flow.
21 SGFP trips.
NO operator action is taken in response to the SGFP trip, and the Rx does NOT trip.
~hich of the foll()yying is an UNEXF'~gI~[) alarm if it is locked in 2 minut~~c:1fte~~§£!> trips?
~
EHC SYS TRBL.
~
Alarm RC PRESS DEVIATION HI.
Alarm RC LOOPS TAVG-TREF DEVIATION.
IFacility: I!Salem 1 &2 IExam Date: L 12/3/20121 mmm
--==="--I....=::::;===-=~",:;:=:,:=:;--;= =:-1;:::::S==RO;:G:;r=ou=p:::;I-'==;-
'cQ00054G446 10 1
1
:~=======================~=
mm.~
that the alarms are consistent with the
.b(5)B is incorrect because the condensate polisher trouble alarm will be in due to the CN 108s 1...-___--' open on a SGFP trip) AND the CN109 being open at the same time (polisher in service). D is incorrect because RC loops Tavg-Tref deviation will be expected as rods are driving in due to the turbine runback to 65%. C is correct because the RC pressure deviation would not be expected, since the setpoint (+75 ipsig deviation)equates to when the spray valves are full open. The spray valves should be shut after the insurge due to the load rejection and then the large amount of inward rod motion. A is incorrect because be in alarm since it receives input from the EHC Control and Status computer, which will have a of Feed Runback alarm in..
Reference Title
[~O~05~l'J~~~W_09__________
=,.:.:_:_=_:......:_:..:_-:--........................-:-c::-====_==-----'==_
Learning Objectives the operation of the following system as applied to S2.0P-AB.CN-0001:
Full Flow Demineralizer Operation Heater string bypass valve operation (CN-4
- 7) c)
Feedwater control valves (BF-19 & BF-40).
d)
Main Feed Pump trips e)
Main Feed Pump speed control i)
Include effects on discharge pressure ADFWCS analyzed transients/accidents:
Loss of Feedwater Excessive Heat Removal Due to Feedwater System Malfunctions Determine the expected alarms and indications.
Describe the analysis assumptions.
Describe the protective features that mitigate the event.
Describe the Question Source:.
IQuestion Modification*Method:
","".",,",,,, Modified Vision Q113267. Removed procedure transition part of question to make RO level (alarm not expected)vs. SRO L..-_~_______"'" !Ievel (unexpected alarm AND~fltprocedure addres~~s unexpected alarm.)
14,201212:07:11 PM
Friday, September 14, 2012 12:07:11 PM I
the following condition:
Unit 2 was operating at 100% power when a total loss of all AC power occurred.
15 minutes after the power loss, operators have locally started 2B EDG.
of the following is an action that is REQUIRED to have been performed PRIOR to energizing and non-essential DC loads to extend the time the Vital Instrument Inverters can power their loads.
and reset SI to prevent the auto start of a centrifugal charging pump and shock to the RCP seals.
EJ iDe.. en.ergizeALL SECs and depress stop PBs for SEC actuated components to prevent o-",~~loadi!!9t~~e 2B 4KV vital bus.
@J IStartthe Station BlackouiCornpressor to provide air for operation of 21-24AF11, AUX FEED IS/G LEVEL CONTROL VLVS, t<?prevent over fee~if'lg the SGs when 2? AFW pp starts.
IAnswer I
'Exam Levell fFCJ 'Cognitive Level, MemoJY~~:=:--l'Facility: 1:?~I11~~~
IIExam Date: I
~1:u3/20121
~if}~~J~I'l1~rgenC:~Clnd Abnormal Plant Evolution~ IRO Group I ISRO Group I r60QQ§5A263~
1055_:~=] [StationBI.9~()LJt to Station Blackout:
.b(10) The Continuous Action Step for energizing a denenergized vital bus with an EDG comes the step to deenergize all SEC's. The Bases Document states on page 15 that the reason to the SECs and depress the Stop PB for all SEC controlled safety related loads is to prevent th bus from overloading. It additionally states that a further reason is to prevent charging pump automatic start and possible thermal shock to the RCP seals. SI is initiated at Step 21 NOT to prevent a charging ipump from running, but rather to prevent the SI actuated valve realignment that will occur if an SI signal is sensed after power is restored. Non essential DC loads are shed at Step 35 to extend the batteries power capability. The SBO is started as part of Blackout Coping Actions in Attachment 2 Part A of AB.LOOP-1. All the distracters are actions which will be taken during an extended loss of all AC power,
,but the correct answer is the only one that is required to be performed AND has the correct reason for doing it prior to power restoration. D will be performed, but it is NOT the correct reason, and is required 60 minutes of Blackout. A and B will be performed, but are not required to be performed prior to restoration.
Reference Title flOSS of All AC Power
.... **cc===___ ******
IMaterialRequired for Examination.
Question Source:
.....'H'V'OI.S 2 NRC Exams IQuestion Modification Method: IDirect From S~urce
'--________--' 108-01 NRc~oexam (May 2010) Vision Q133~~8_____.
~
_______.........____~i Friday. September 14, 2012 12:oi:11PM
IGiVen the following conditions:
1-Unit 2 is in MODE 4.
RCS pressure is 290 psig.
RHR HX inlet temperature is 270 0 F.
21 RHR pump is in service in shutdown cooling.
22 RHR loop is aligned for ECCS.
A loss of all off-site power occurs.
.Which of the following identifies why S2.0P-AB.LOOP-0001, Loss of Off-Site Power directs loperators to initiate S2.0P-AB.RHR-0001, Loss of RHR?
E1IThe 2A SEC trips 21 RHR pump and does not restart it when 2A EDG connects to 2A vital bus. I
@J ~~~r:~;s~!;~i~ea~lit~~nbn~~~S~CW pumps, and they d~~~r~~:h_~_~_t~eEDGS connect to EJ IS2.0P-AB.LOOP-0001 does not know what the initial plant conditions are, and always directs initiation of S2.0P-AB.RHR-0001 regardless of whether or not RHR is in operation.
@] The 22RH18, RHR HX Flow Control Valve, fails shut, and the 2RH20 RHR HX Bypass Flow Control Valve fails open. Action is contained in S2.0P-AB.RHR-0001 to re-establish positive control of RHR HX flow.
IAnswer I@:] IExam Level I ~ ICognitive Level l1Q9mprehension I[Facility: IISalem 1 & 2 IIExam Date: I 12~3/201~
ITier: IIEmergency and Abnormal Plant Evolutions IIRO Group IOJ ISRO Group InJ
[000056K302
~-----] [Loss of Off-Site Power I I Record Number I [
-121 I6iSLllKnowledge of the reasons for the following responses as they apply to Loss of Off-Site Power:
I i6k3.021 ~ctions contained in EOP for loss of offsite power I[4AI L4.7I Explanation of 155.41.b(7,8) The loss of off-site power with NO Safety Injection signal is a SEC MODE II actuation, Answer Blackout. All 3 EDG's will start, the SEC will strip all loads of it's vital bus, shut the EDG output breaker jand sequnce on BLACKOUT loads. The rHR pumps are NOT blackout loads and will not be started.
AB.LOOP-1 asks, at step 3.8, if a RHR pump was running in SDC mode. If the answer is yes, it directs initiation of AB.RHR since the LOOP will result as described above. The CCW pumps WILL be started by their respective SEC's. AB.RHR is NOT always directed, only if a RHR pump was running in SDC mode.
The 22RH18 fails as is, and is not the reason for initiating AB.RHR.
Reference Title ILoss of Off-site Power I
I __ ____________________________________________
Learning*Objectives ABLOP1 E002 I Describe, in general terms, the actions taken in S2.0P-AB.LOOP-0001 (q)and the bases for the actions.
IMaterial Required for Examination II Question Source:
INew IIQuestion Modification Method:
II I Friday, September 14, 2012 12:07:11 PM I
Page 16 of 90 I
the following conditions:
Unit 1 is in MODE 2 performing a startup by control rods lAW S1.0P-IO.zZ-0003, Minimum Load to Hot Standby.
Rx power is stable at 4%.
Vital Instrument Bus 1 D inverter output breaker trips and deenergizes 1 D 115VAC.
Vital Instrument Bus.
,One minute after the loss of 1 D VIB, which of the following contains the indication(s) that will be illuminated on Reactor Status Panel 1 RP4, with NO action?
I!J lamp.
~
~------
---.-~ ~--------------
- ........................... --------------------------------------~
E] @~~ Over Power Rod_~e Manual-Byp~~~for CH IV lamp.
@]
Power and Low Power.
- ==~'--:. 'L=======-===:"...I=c=Og=n=it=ive=L=e=ve=I~I_=[A=..R=P~liCanQD.
II Facility: Iisalem 1 8..~.. ___ IIExam Date: 1= 1213/20121
_n IRO Group I [JJ ISRO Group I
[00057A203
~~==:~~=====================
to Loss of Vital AC
.b(7) A is incorrect because the reactor has no trip demand from the loss of
"'v"'U"' the CH IV indication is associated with "G" 4KV RCP group bus. 4 group buses are H,E,F,G fro 1, 12, 13, and 14 RCPs. C is incorrect because the over power block bypass must be nmanually D is correct because the CH IV for BOTH the High power hi flux and low power high flux will be Reference Title ILoss of 1 D 115 Vital In~tument bus Learning Objectives the operation of the following as applied to S1/S2.QP-AB.115-0001(Q), S1/S2.0P-AB.1 003(q)and S1/S2.0P-AB.115-0004(q) 115 V Vital Bus distribution EQ~ue=s7.ti=on~&ru~.=-~_e:~~~~~~~~~==============..bl~~u~es~ti~on~M~O~d~rr~i~~t~io~n~M~~~h~OO~:~~================
14,201212:07:11 PM Page 17 oHIO J
,Given the following conditions:
Salem Unit 2 is operating at 100% power.
All Station Air Compressors trip and none can be restarted.
i-The Unit 2 ECAC does not start, and cannot be started.
The Unit 1 ECAC starts and trips after 5 minutes.
Which of the following identifies an action taken on Salem Unit 2 lAW S2.0P-AB.CA-0001, Loss of g~rltroJt-~-, ~d_vv_~y~_
~ IAn o.pe.r.ator is-dispatch.ed to locally Sh.ut the. 2DR6' AFWST MjI.TvaIVe~This.. iS. to p. revent o.ve.r 1
~o\\Ali.l~g !Il_e_~~'{\\'§..I~~_cC!~i~g~~piILof hydrogen peroxide to the storm d~airl_sY's~~t"Il_"________J
~ The Rx is tripped when EITHER Control Air header lowers to <80 psig. This is to prevent an automatic trip on 10-10 SG NR level when the BF19s associated with that header start to drift shut.
~ All Radwaste releases in progress are terminated. This ensures that during a gradual depressurization of the Control Air system a release is not in progress when the dilution medium flowrate may be changing.
@J An operator is dispatched to manually control 23 AFW pump speed which is running at the high speed stop. This ensures the pump does not become steam bound due to the 21-24AF11 valves failing shut with only limited recirc flow provided.
IAnswer I EJ IExam Level I ~llcognitive Level IIMemory IIFacility: IISaiem 1 & 2 I I Exam Date: I 12/3/20121 ITier: I ~rgency and Abnormal Plant Evolutions IIRO Group I OJ ISRO Group I OJ 1000065G314
[065
] ILoss of Instrument Air I I Record Number II
[LJ Radiation Control I
23:14] [Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or I [ 3.4] [3.8]
~mergency conditions or activities.
IExplanation of I55.41.b(1 0) A is incorrect because while the 2DR6 will be operated locally, the concern is the overflow of IAnswer water with hydrazine in it, not ammonium hydroxide. C is correct because the bases document says that on page 8 of 12. B is incorrect because BOTH CA header pressures have to be below 80 psig before the rx is directed to be tripped, but the reason is correct. D is incorrect because the action is correct, but the reason is wrong. The AF11 s fail open, and pump runout is a concern with the speed failed at the high speed stop and higher steam supply pressure present. (page 9 of 12)
Reference Title ILoss of Control Air I
Learning Objectives
! ABCA01 E002 Describe, in general terms, the actions taken in S2.0P-AB.CA-0001 (q)and the bases for the actions in accordance with the Technical Bases Document.
:=1==
.~.
~IM=a=~~ri=al=R~e=qu=ir=e=d~fu~r~Ex=a~m=in~a=ti=on~=====I~I==============~~~~~~~~~~==~~~~~~~~~=I
'Question Source:
I I Question Modification Method:
I,Editorially Modified
.J Vision Q41379 modified from the release valves are shut because...to what do you have to do (terminate any L...-________---I release in progress) and why. The why was the original questions 4 choices. Also modified per technical review comment that B could also be correct Friday, September 14, 201212:07:11 PM I
the following conditions:
1"\\0.."",1",.,r<:> are performing actions in 2-EOP-FRCC-1, Response to Inadequate Core Cooling.
1-With no other RCPs in service, 23 RCP has been started lAW direction in FRCC-1 and Rx core temperature is lowering.
23 RCP was the only RCP able to be started.
Which of the foll()w~,!g id~nt!fies w~y 23 RCP would be stopped lAW FRCC-_1___?c_:~===":"c, __~:::===.:.
E] IRVLlS level hasrisento >57010 which shows that the fuel is covered and injection flow is lpresent.
-~-~----
[]
SG NR levels have lowered <9% which indicates insufficient heat transfer will be RCS
@J r3 RCP #1~seaID/P has lowered to less than 250 psid which is less than the minimumm~l require(j to prevent mechanical damage to RCP.
IAnswer'1
~=~::::::::=-==~~:;:::=::;-;:::"IIFaCility: 1~1'l11 &. 2 ~_IIExam Date: 11273120131 IIRO Group I ISRO Group I =:1l 1000074K3<M=:J of the reasons for the Tnlln'Allnn RCPs 55.41.b( 1 O)The step to isolate the ECCS accumulators is 28.1, just prior to stopping RCPs at step 29.
I L..-~_--'---I The ECCS accumulators are isolated after intermittent RHR flow has been verified, since this means they I have discharged based on RHR discharge pressure capacity. However, SI or Charging system flow is n01 checked until AFTER the running RCP is stopped in step 29 when RCS Thots (at least 2) are <350"F.
!There Is no concem for Rep damage based on seal DIP. As discussed at Step 23, normal conditions for I
',RCP oper.ation are desired, but not reqUired... The only t,hi..ng that will p,re.. vent RCP sta,rt.. is no SG NR,..Ievel
<<9%) based on potential creep failure of the high temperature SG tubes, but LOSS of SGNR level does Inot require stopping the RCPS. RVLlS level must be >57% at step 31 and be combined with at least 2 RCS Thots <350 and all RCP's already stopped to exit ~RCC-1 to LO~A-1.
Reference Title
=.......
~nadequate Core Cooling Describe the EOP mitigation strategy for the following:
A.
Response to Inadequate Core Cooling.
B.
Response to Degraded Core Cooling.
C.
to Saturated Core Conditions IMaterial Required for Examination IQuestion Source:. I.1N_e
..~_~
..._ __.....J IQuestlOr:l Modification Method:
....w.~*___
_____~-_-_-_~....~
IQuestion Source Comments: II L-___________________________________________________________________
I I
Friday, September 14, 201212:07:11 PM Page 20 of 9~J
the Unit 2 Rx operating at 100% power, which of the following radiation monitors would be the RST to indication that a nuclear fuel rod had a substantial leak?
Containment Iodine.
Containment 130' elevation.
0, Plant Vent Noble Gas Release Rate.
'Answer I~J IExam Level I
'Cognitive Level I Comprehension
['Facility: 1~~lem 1&~~2_~['Exam Date: I 12f~/2~
~ ~i9~Yi!llCL~bnormal PlantEvolutions
'IRO Group' [] 'SRO Group I[]
[()QQQl6A1QL
~76
]
Reactor Coolant 81"1"".'.,.
~
.--~--.
......~--
..-~
..........~.
i55.41 (11) With the plant operating at 100% power, letdown will be in service. Failed fuel would 1...-___---1 iimmediately release fission products into the RCS. The letdown line would transport these radionuclides and be detected quickly by the 2R31. The 2R 12B would only see the failed fuel if there were a way for it to get out of the RCS, and then it would have to expand and travel throughout containment. The 2R2
- would respond the same way as the 212B but even slower, since its an area monitor and area radiation Ileveis would take a long time to rise. The 2R41 D would see increased radiation levels if the letdown fluid l;~~:~~;et outsidet:e letdown line or v~~ or charging line, ~~d would be after th:~R31 saw the
.~
Reference Title LeamingObjectives
[ ABRC02EO~DeSCribethe operation;fthe~f=-olc-Io~wi:-ng-s-y~st-ems'~s applied-t-oS-2-.0-P--*AB.RC~0602:'~-*---"-'_....
- a) 2R31 Letdown Line Failed Fuel Monitor b)
C'{~S Demineralizer Operations Source I Page 21 of 90
- Given the following conditions:
Unit 2 is operating at 100% power when the operators receive several alarms related to the 500KV grid.
The Electric System Operator calls Unit 2 and directs them to perform a rapid load reduction to 875 MW due to grid instability issues.
jWhich of the following describes how the load reduction will be performed lAW S2.0P-AB.GRID 0001, Abnormal Grid?
I iAt the EHC Console the PO will depress...
§J Ithe GO pushputton, and ensure the runback automatically stops at -66% turbine power.
~ ISMD #2 RUNBACK and GO PBs, then depress HOLD when Main Generator load lowers <
875 MW.
~ ISMD #2 RUI\\JBACK and GO PBs, and ensure the load reduction stops automatically at -66%
turbine power.
@] IEITH. ER the GO pushbutton OR SMD #2 RUN BACK and GO PBs, then depress HOLD when Main Generator load lowers < 875 MW.
IAnswer I L IExam Levell ~ ICognitive Level IIMemory IIFacility: IISaiem 1 & 2 I I Exam Date: I 12/3/20121
~ ~ergency and Abnormal Plant Evolutions IIROGroup 1[3 ISROGroup 1[3
- 0000llA102 lOll
] IGeneratorSoltage and Electric Grid Disturbances I I Record Number II 171 IAA1. J ~AbjlitYto-operate and/or monitor the following as they applyto-G-enerato-r=Voltageand Electric Grid Disturbances:
r-IAA-1.-02-=;III~bine / gen~~ator controls._________________ _
I[Mil 3.71 Explanation ofI55.41.b(1 0,7) AS.GRID directs the load reduction directed by the ESO due to grid instability be performed
....An_sw_e_r__..... lAW AU 4, which says to push SMD #2 if the required end point is >765MW AND <942 MW. It says to Ilpress HOLD when the MW value is less than or equal to that directed by the ESO. While depressing the GO PS would work, the procedure says to do it a certain way to ensure consistency amongst the crews (Note on AU 4), so it would be wrong. The MT is normally set up to do a 15% per minute run back to 66%
turbine load (-810 Mwe), so while it would get load about where it is supposed to, the procedure doesn't allow you to do it that way. Candidate needs to know how to initiate load reduction, and how it is directed to be stopped.
Reference Title IAbnormal Grid I
1 Learning Objectives ABGRIDE003 Describe, in general terms, the actions taken in S2.0P-AB.GRID-0001 (Q) and the bases for the actions.
~==~==
~IM=a=m=ri=al=R~e=qu=ir=e=d~fu~r~E=xa=m=in~a=ti=on==-~Ic==~~============~~~~~~~~~~~~-=-=--=--=-=-=*~=-------=-----_--J~
Question Source:
IFacili~E~amBank IIQuestion Modification Method: IEditoriallyModifted I
Question Source Comments:
Vision Q120134. Modified one distracter from SMD 2 or SMD 3 to SMD 2 or GO PB since SMD 3 only appeared in one choice, and GO only appeared in one choice.
I Friday, September 14, 2012 12:07:11 PM I
I Page 22 of 9iJ
the following:
i Unit 1 experienced a Rx trip and Safety Injection from full power due to a RCS leak.
i-The control room crew is currently performing 1-EOP-TRIP-3, Safety Injection Termination.
11 Charging pump is in service and 12 Charging pump has been secured.
Charging pump flow through the BIT has been isolated and 1CV68 and 1CV69, Charging Discharge Valves, have been opened.
The RO fully opens 1 CV55, Charging Flow Control Valve, and reports current PZR level is 45% and lowering slowly.
i iWhich of !bE?following describe~the action the contr<?lroom crew should take lAW 1-EO'=:TRIP-3?
§J flow the BIT and close 1CV68 and 1CV69.
~ ~nitia~~~!llrl]eCtiOnand retu~hto 1 EOP-TRIP-1, ~~Trip or Safety Injection, Step 1.
§J to establish PZR level stable or and continue
@] Allow-~10 minutes to allow for system conditions to stabilzewhilemonitoringPZ~RleveT--J IAnswer I ~llExam Levell [!C] ICognitive Level I ~~_
I!Facility: I~~~.. IIExam Date: I 12/3f201~1 Tier:
llE:_mergency~~~-n-o-rm-aIPlant Evo[~tio-ns--IIRO Group I ISRO Group I OOWE02K201~1 E02
.'===
..~.=~"'::.===:==,=============================================:...':R=e=C=Or=d=N=u=m=be=r=cl~=;_.............-,
nrnnnr'Qnt., and functions of control and safety systems, including instrumentation, ntQrln,r-I<e> failure and automatic and manual features.
5.41 (7,8) After stopping one of two centrifugal charging pumps at step 4, charging flow is re-directed L...:...:.=";';"':~_---I. om BIT to normal charging line. If this flowpath cannot maintain stable or rising PZR level, the operator will re-establish BIT flow and go to LOCA-2, Post LOCA Cooldown and depressurization since control of RCS inventory is greater than the capacity of normal charging. The basis document specifically says not ito re-start the idled CVCS pump, because that would restore subcooling/PZR level and you would end up Iback at the same step if you went to LOCA-1, then back to TRIP-3. C is incorrect because TRIP-3 is not
!continued. 12 CVCS pump MIGHT be started lAW CAS to start ECCS pumps as necssary since stem is non-specific about actual PZR level, but continuing in TRIP-3 is not true. Step 7 has operators restablish charging flow through the BIT and go to LOCA-2. There is no cAS action to initiate SI and go back to iTRIP-1.. There is no prision (nor reason) to allow plant conditions to stabilize, maximum charging flow
- should act on PZR level in not minutes to level.
Reference Title cSafety Injection Termination IQuestion Modification Method:
L---~___~[
14,201212:07:11 PM
the following conditions:
Unit 2 was operating at 100% power when the RCS developed a SBLOCA.
45 minutes after the trip, 2B 4KV vital bus locked out on bus differential.
Containment pressure is 2.4 psig and lowering very slowly.
Operators are now performing actions in 2-EOP-LOCA-2, Post LOCA Cooldown and Depressurization.
of the following contains an alarm, which if received during performance of LOCA-2, would requir~ th~_~~~~ed r~~ponse?
I!J rpZR Low Level alarm at 17%. Start ECg~~~ asnec~s~
~
Lo Level console alarm at 15.2 feet. Transfer RCS to Cold EJ 21 SG Program Deviation Setpoint Actual console alarm at 28% NR level. Open 21AF21 SG Control Valve to raise level in 21 SG.
IAnswerl ~] IExam Level] R!ICognitive Level I !QQmprehension II Facility: IISaiem 1 & 2::] IExam Date: I
.......... 12/3/20121
~ ~rgency8ildl'_bnormal Plant Evolutions~ IRO Group I ISROGroup I lQciY;Vf;03K19_:t..J of the operational implications of the following concepts as they apply to LOCA \\"'O()loc)wn 1nf"<:Itl"lr<> and conditions indicating signals, and remedial actions associated with the
\\"'O()loc)wn and
.b(1 0,11) RWST 10 level at 15.2' indicates the need to transfer RCS cooling to cold leg recirculation,
'--~__.....I as identified by the CAS action in LOCA-2. The SG program deviation setpoint is +/-5%, so the alarm at 28% would be valid. However, 22 AFW pump has no power, so opening the 21AF21 would have no
,effect. The R53s are N2 monitors in the Main Steam Lines, and after the Rx is shutdown do not provide indication of any use. The PZR level at which ECCS pumps are started is 11 %, (19% adverse). With containment pressure at 2.4 psig. normal values would be used. The PZR low level alarm comes in at I.5% belo~ prog~am, which would ~~ -22% with the low Tavg expected during a SBLOCA.
Reference Title Learning Obiective$
the indications that are monitored to ensure proper system/component operation for each step in POST AND DEPRESSURIZATION.
I Material Required for Examination Question Source:
I~uestion Modification Method:
Vision Q127166. Removed procedure from LOCA outside containment, removed response from opening AF21 I...-__________.....J jvalve a~part of distracter.
Friday, September 14, 201212:07:11 PM
the following conditions:
Unit 2 is attempting to identify and isolate a 400 gpm LOCA into the RHR system which occurred while operating at 75% power.
2-EOP-LOCA-6, LOCA Outside Containment, was entered from 2-EOP-TRIP-1, Rx Trip or Safety Injection.
The source of the water is back leakage from the 23 cold leg injection line.
A large leak in the RHR system is located on the piping between 21 and 22RH19s, RHR HX DISCH X-CONN VALVES.
of the following components, if it failed to respond when directed by LOCA-6, would prevent the RCS leak outside containment?
'----""~- -
~ [~~~J~~B!:iBPI§gtl TO COLD LEGS._
E;]
@]
IIFacility: IISaiem 1 &2 II Exam Date: I 12/3/2~1_~]
- -=====~~==~:::=:::=::::::::::;=.==-==~~~:::;-~"
ISRO Group I lQQwE04Ai9iJ
- 0) The leakage from 23 Cold leg flows back through the 21 SJ49, then through the 21 RH 19 connect to reach the leak. Leak isolataion is attempted first by ensuring closed the RCS-RHR isolation valves RH1 and RH2. Then BOTH the RH19s are shut. Since the leakage from the has to flow through the 21 RH 19 to reach the leak, and it is a normally open valve. Its failure to reposition when directed would prevent leak isolation. The SJ69 is closed after the leak is isolated as above. The 22SJ49 is on the opposite RHR train. Candidate also has to know which RHR train feeds which cold legs when not cross connected.. 21 RHR feeds 21 and 23 cold legs, while 22 RHR train 22 and 24 Reference Title lQf~~ Qu~ide Containment Learning Obiectlives IMaterial Required for Examination Question.Source:
14, Page 25 of 90
- IFRHS-1, Response to Loss of Secondary Heat Sink, Step 3 asks, "Is RCS pressure greater than ANY intact or ruptured SG pressure".
yvh!~h ofthef9l1ow!ng~~ll1er1!s is correct if the opera!or~~s N9~....__.
[!J iIMMEDIATELY-goto Step 23, Bleed and Feed Initiation, since there is no decayheat removal.
occurring through the SGs.
.....-J
~ ~;~eP:~::~~~~c::;=tt~~:s~r:::~~ni~~t9~~!~k~~~~:~_re_.__~
El to Procedure in effect. The RCS has experienced a LOCA large enough that a heat sink is NOT because heat is removed break flow.
@] IIMMEDIATELYtrip all RCPs to prevent further loss of reactor coolant through the LOCA, since I ia LOOP later in the event could cause a more severe loss of reactor coolant or two-phase IRCSHow.
IAnswer I [J tExamLevel1 fif.=J ICognitive Level Il~ry.__JIFacllity: 1~.&_2~_.IIExam Date: I.
1213/2012]
~ ~~.§l.fl.C!A~'!Ial Plant E",olutions _] IRQ Group I ISRQ Group I lQQ\\liLE:Q'§~
it,=05_
~§.§..of$~~QIl<:laryJ;~$inkn.~_
_= l~ecord Number 1'---_... --'-'
of the reasons for the Heat Sink:
abnormal and emergency operating procedures associated with (Loss of Secondary Heat 55.41.b( 1 0. 14) The reason for checking RCS pressure> intact or ruptured SG pressure is to check if 1...--'-.:__--1 there is a need to be worried about a secondary heat sink. If RCS pressure is below SG pressures, then a
.LOCA of sufficient size is present, and break flow will be removing decay heat, along with ECCS
- injection. Oistracter A is incorrect because the criteria for going to bleed and feed is SG WR level.
10istracter B is incorrect because a secondary heat sink could actually be established, and could reduce IRCS. temperatu.re by dumping steam f.rom the SG.s... 0
..... is incorre... t it is n.ot...a
....R 1
istrac.ter....o
...c
...... CAS of F....HS but is reason for tripping RCPs in TRIP-1 or LOCA-1.
Reference Title ILoss of Secondary Heat Sink.
I==Q=u=es=ti==on...s~o=u=rce-==:..L:;=;;;;;;;;;;;=,-;=c====::.====::::======JbQ~ue=s=ti=on=M........"od...lfi...c=at...io=n=M..."eth""""o.,...d.,...:.......L'============
!Vision Q127090 I...---'-.:-'-.:~~~---I.--......
=..........,
14,201212:07:11 l~ge 26 of 90 i
the following conditions:
Unit 2 has experienced a steam line break inside containment.
Operators have entered FRTS-1, Response to Imminent Pressurized Thermal Shock.
IWhY will the oper~9rs be in~tructed to terminate SI and start RCP(s) if possible?
~ The soak required by FRTS-1 requires SI to be secured. RCPs should be started to boron concentration throughout the primary to ensure proper cools.
Injection flow is a significant contributor to any cold leg temperature decrease or condition and must be terminated. RCPs are started to minimize temperature across S/G tube sheets.
@J Safety Injection-flow -is a significant contributortoany cold leg tEmlperature decrease or overpressure condition and must be terminated. RCPs are started to and warm reactor coolant water.
IAnswer I ~-!Exam Levell
!Cognitive Level !MemOry I!Facility: ! ~ll'lm1_&_2__d,-Ex_a_m_D_a_te_:71===1=2/=3/=2=01c~:~:::J Emer9E;lncy and_Abnormal Plant Evolutions IIRO Group I ISRO Group I
~-'-'=--" ""!Urn,,,. abnormal and emergency operating procedures associated with (Pressurized Thermal
.b.(10,8) incorrect - purpose for RCPs is not priority in FRTS-1, soak is not basis for SI. B - incorrect C - incorrect SI basis RCP basis not accurate. O-correct Reference Title Imminent Pressurized Thermal Shock Conditions Learning Objectives caution. and note in 2-EOP-FRTS-1
- 2. and EOP-CFST-1. Figure 4 & 4A Originallyon Seabrook 2003 NRC exam.
September 14, 2012 12:07:11 PM
the following conditions:
Operators are performing a natural circulation rapid cooldown on Unit 1 lAW 1-EOP-TRIP-5, Natural Circulation Rapid Cooldown Without RVLlS.
NO RCPs are running or can be started.
The control room crew has completed the initial RCS cooldown I depressurization to 500°F I 1600 psig.
The current time is 1300.
i Of the following, which one identifies the EARLIEST time RCS Thot temperatures could be reduced
[below 450°?
I Assum(3 the cooldownwill start at 1309 and instantaneously ~e at th~ maximum rate allowed.J EJ
- 6.
~
l... _________________________________________________________.....____________..~
EJ
~
ICognitive Level IIApplicaiion-j IFacility: 1~1& 2
...J IExam Date: I 1213/2012!
m and Abnormal Plant IRO Group I ISRO Group I
,00VIJJ~lQ,*~_1Q~.-J
[Natural Circulation with Steam Void in Vessel with/without RVLlS:=J ~~()rd Numberl to operate and / or monitor the following as they apply to Natural Circulation with Steam Void in with/without RVLlS:
behavior characteristics of the
.55.41.b(10) RCS temp is at 500 degrees per the stem. The next temp reduction will be to 450° L..-~~_........I step 9, and the cooldown is directed to be performed at <100°F per hour. This means 30 minutes of cooldown is required. The 1401 distracter is if the 500/hr rate of step 7 (initial cooldown to 500°F) is used. The 1316 distracter is if the 2000/hr PZR cooldown limit perTS 3.4.10.2.b is used. The 1501 is for both of the and if the 100°F hour rate is used for an entire Reference Title
!Natural Circulation Rapi~ Cooldown Without RVLlS m
Learning Obi*ecti'ves indications that are monitored IMaterial Required for Examination Question Source:
____.. IQuestion. Modification Method:
Q116968. Modified to include procedure name. Added that the C/D to 500°F has been performed.
1 minute to each choice since the stem asks when can be below 450, and 30 minutes of cooldown to 450.
I
'Given the following conditions for Unit 1:
1-A reactor trip and SI occurred at 0700 due to a 1500 gpm RCS LOCA.
RHR system problems have resulted in a loss of recirculation capability.
Current time is 1300 hours0.015 days <br />0.361 hours <br />0.00215 weeks <br />4.9465e-4 months <br />.
Conditions present when transitioning to 1-EOP-LOCA-5, Loss of Emergency Recirculation, from 1 EOP-LOCA-1, Loss of Reactor Coolant, due to the loss of recirc capability are:
RCS subcooling is 10°F.
All RCPs are secured 11 and 12 Charging Pumps are running BIT flow - 350 gpm RVLlS full range 95%
11 SI Pump flow - 250 gpm 12 SI Pump flow-250 gpm Containment pressure 4.1 psig Which of the following identifies the ECCS pumps that should be run following determination of Minimum SI Flow for Decay Heat Removal?
Assume:
--~--~------------------------------------------------------~
Equal flow from each Charging Pump, and each pump will supply half the original total flow if the other charging pump is secured.
Each SI pump flow remains constant if the other SI pump is secured.
RCS subcooling remains between 10°F - 45°F for the duration of this question.
ElIOI\\JE charging pump and BOTH SI pumps.
I
[] lONE Charging pump and ONE SI pump.
I ElIONE. Charging pump only.
I
@] 1QBE. SI pump o_n--,,--Iy_._____
~
IAnswer I [J IExam Level I IiL:-=:J ICognitive Level IIApplication IIFacility: IISaiem 1& 2 I I Exam Date: I 12/3/20121
~ [Er!18rgency and Abn()rmal Plant Evolutions I IRQ Group It 1J ISRQ Group I nl 100WE11 A202 I E11
] [LOssoTEmergency Coolant Recirculation J[ReCOrCi"N-umber] 1 241
~iAbility to determine and interpret the following as they apply to Loss of Emergency Coolant Recirculation:
I
- EA2.2 IlAdherence to appropriate procedures and operation within the limitations in the facility's license and I 3.411 4.21
~mendments.================================================----==
Explanation of 155.41.b(1 0,8) A 1,500 gpm RCS LOCA will deplete the RWST in 2.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. The transfer to CL Recirc will Answer
!already have been performed. During step 14 of LOCA-5, charging pumps will be reduced to ONE centrifugal, and SI pumps will be reduced to ONE. Starting at Step 19 of LOCA-5, with RCP's secured with <50 degrees subcooling, will use Figure A to determine th ECCS flow required vs. time after trip. 6 Ihours equals 360 minutes, which is -225 gpm, but definitely LESS THAN 250 gpm. With the stem stating that charging pump flows remain the same, a single charging pump will be insufficient to supply the required flow. A single SI pump, however, supplying 250 gpm will supply sufficient flow.
Reference Title Friday, September 14, 2012 12:07:11 PM:::J
Recirculation IMaterial Required for Examination 1-EOP-LOCA-5 flowchart 1 & 2 Question Source:
- ~-IQuestion Modification Method:
'--_-'--________..... IVision Q42181, used 3 NRC exams ago (Class 07-01, Aug 2008)..~___~~ ~~__~~ ~_~
14,201212:07:11 PM
the following conditions:
Unit 1 has experienced a MSLB at the Main Turbine inlet steam piping.
All attempts at Main Steam line Isolation have failed.
Operators have transitioned out of 1-EOP-TRIP-1, Reactor Trip or Safety Injection.
RCS cooldown rate is 120o/hr.
RCS pressure is 1300 psig and dropping.
Charging system SI flowmeter indicates 290 gpm.
The RCS cooldown is NOT being controlled.
Which choice identifies an action that must be performed lAW 1-EOP-LOSC-2, Multiple Steam Gener~r Depress~LJ~ization, and why?
__.....~
~
RCP's to minimize heat to the RCS.
~
AFW to minimize cooldown while still k~(9pingthe SG tu~eswet_._______~._.~_
EJ BOTH RHR pumps to prevent damage to RHR pumps from continued operation above shutoff head.
operators to close all BF19's, BF40's, and BF22's to re-establish a secondary pressure SG.
IAnswer I tiL] IExam Level I ICognitive Level I IAlmlic:§tion
!Tier:.!Emergency and Abl'1()rmal PlaniEvolutions
! ~IR~O~G=ro=u=p:;-1=:::;-;:::::::;:;:~=:;-.::::=;--.1 r----=~====
~=--___ Uncontrolled Depressuri?=C:ltion of all Steam Generators nmiVlel:me of the operational implications of the following concepts as they apply to of all Steam Generators:
capa~ity, and function of emergency
~~~~'ifr;:;:;.....
~..~.I.....~..I...........
........................-~-..~.~.-
..='=..I
!55.41.b(1 0,4 )Once out of TRIP-1, no actions other than attempting to close MSLI valve are taken in
'--___..... LOSC-1 prior to going to LOSC-2. Maintaining> 1 E4 Ibm/hr to each S/G keeps tubes from drying out, among other things. Do not trip RCP's because pressure is dropping due to cooldown, and the reason is wrong. Doesn't matter if it's uncontrolled or not. Distracter A is incorrect because we don't close BF22's,
.but the rea!:l<?l!i!:l right. Don't stop RHR pumps becausepr(:1ssure is still dropping, reason is right.
Reference Title
.MultipleSteam Generatori?E:lpressurization Learning.Objectives IMaterial Required for Examination
====~=.:...=.:................::....=:====-=~..----:-----c...~.~.. IQuestlon Modification Method:
Used on "H" RO NRC Exam (2004, 5 NRC exams ago.)
14,2012
the following conditions:
1 has experienced a LBLOCA.
The crew is responding lAW the EOP network.
3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> after the transfer to Cold Leg Recirc has been accomplished, the STA reports a Purple Path exists for containment Environment due to containment sump level being 80%.
~hi~h of ~~~lJ'\\Iil!g~~would assist in validating that_c:lhigh containment sumel~Ic:l~tuc:llly exists?
~ ~]'.. contains 200,OQ9 gallons.
~
Tank levels are both 85%.
in service with SW header each.
.JIExam Date: I lQQWE15G1451
'[~-ec-o-rd-N-u-m-be-r'"l 12/3t?~
to and diverse indications to validate the of another indication.
- .,55.41b.C7,9)FRCE_2 checks for possible sources of the excessi\\/esump level in containment: They are:..~
'--___---' CFCU SW flow, FP to containment isolation valve position, CCW Surge Tank level, Demin Water Storage Tank Level, Primary Water Storage Tank Level.
PWST of 200,000 gallons is not enough to have raised containment sump level that much. Tank Capacity completely full is -240,000 gallons.
SW header pressure of 108 psig is within the normal operating range of 105-115 and would have to be a Ilot lower with 3 SW pumps in service to inject enough water into containment.
IIEach CFCU SW flow is normally -1600 gpm, and indicates a 3,000 gpm leak into containment is possible. Fire protection Storage tank level change would be inadequate to raise sump level that much.
Reference Title
!Response toljigh Con~aJl1ment ~LJmp Level Data Learning Objectives the indications that are monitored to ensure proper system/component operation for each step in 2-EOP-FRCE-1 thru 3 and EOP-CFST-1, 5
IQuestion Modification Method:
14,201212:07:11
I of the following identifies the radiation monitor(s) that must be sensing high radiation for either channel of the Subcooling Margin Monitor to automatically shift to the
.ADVERSE Mode?
.._..._-_....__...===
EJ ~HE.F3~9B~~4:~QQ!1.t~l!!rl1~t!::!l9~(3~e~ __
--~
---.~....
~
... ~ A"!~~4~_Con!~!'l~rll!iigll~ge.
Containment 130' OR R7 In-Core Seal Table.
R2, Containment 130', AND R7 In-Core Seal Table.
~-iIExam Date: I 12/3/2612]
~I;::SR~O~G;::rou==p~I;::=:;--
lQQWE16A4QU as Containment Radiation:
=--==-.J llrihc,ron,('", to appropriate procedures and operation within the limitations in the facility's license and j55.41.b( 11) Either of the Containment high Range monitors reaching 1 E5 Rlhr will automatically place L..-...-.;__--liSMM in ADVERSE Mode. The other area monitors in containment listed do not into the SMM.
Reference Title.
Learning ObiectivEIS Identify and describe the Control Room controls, indications, and alarms associated with the Radiation including:
The Control Room location of Radiation Monitoring System control bezels and indications. (Licensed Operator & STA The function of each Radiation Monitoring System Control Room control and indication. (Licensed Operator & STA only)
The effect each Radiation Monitoring System control has upon Radiation Monitoring System components and operation.
(Licensed Operator & STA only)
The conditions or permissives required for Radiation Monitoring System Control Room controls to perform their to ensure proper system/component operation for each step in
__ -.-1 Question Source:
IQuestionModification Meth()d:.. I Vision Q73944 replaced area monitors outside containmentwith 2 others inSide.co.n!ainmen.!.. Removed W.indoW]
I...-.:...-__________..J dressing.
the following conditions:
Unit 2 is operating at 70% power and stable after a load reduction was completed 10 minutes ago.
Rod Control is in MANUAL control.
The highest actual Tave-Tref deviation is 4.0°F.
Which choice identifies the rod speed that would be present initially if the Rod Control Selector Switch w~~y~~(j~ in AUTO?
EJi24
~~~-----~~~--------~-------
-~--------
~--
~--.......-------~~
IAnswer I ~~=i IExam Levell rs=J ICognitive Level I ~pplicCl~_
IFacility: 1~2~IExam Date:. 1 n~~~1~1
!Tier:-]
IROGroup I ISROGroup I 19Q1000A101 speed is determined in AUTO by Auct High Tavg vs. Tref (PT-505, Turbine Pressure). It doesn't matter which channel is connected to Terr recorder on 2RP3. The AUTO rod
'speed program is 8 spm from 1.5-3.0°F deviation. From 3.0-5.0 it ramps up linearly from 8 spm to 72 spm. This correlates to 16 spm per 112°F temp change. 8 spm (@3.0) + 32 spm (from 3.0 to 4.0)= 40 spm.
The 56 spm distracter is if a linear ramp from 1.5 - 5.0 degrees was used.
The 24 spm distracter is if the Terr connected to the chart was used.
The 48 spm distracter is normal manual rod control speed.
The Power mismatch circuit in rod control will have cycled through over 5 time constants and its effect on rod speed will be zero ten minutes after the load reduction has been LeamingObjectives IMaterial Required for Examination Question Source:
Exam Bank IQuestion Modification Method:
'--____--'-___~ !viSion Q8806_2_**_______ _ _______________
14,201212:07:11 PM
I the following conditions:
1-Unit 1 is operating at 30% power returning from a mid-cycle outage.
1-Rod Control is in Manual.
1-14 RCP trips.
Which of the following describes how 14 RC Loop Tavg will be affected 5 minutes later with NO
.operator action when compared to pre-event Tavg in 14 RC Loop?
14:~BC Lot?PJ""avg~~~:_~.
E]
because backflow from the unaffected
....f!-(j-~e.:gual Thot.~1 will cause T c to rTs-e-a
~.~~--
~~.
[]
because backflow from the unaffected will cause Tc to rise and Thot.
~ 'LOwer beca~_~_l~~b.~elt tr(;lI!~l!I.s!ccur in 14 SG due to the loss of forced f10vv il!~t:lp,
@]
because less heat transfer will occur in 14 SG due to the loss of forced flow in Icognitive Level I [Comprehen§ion IIFacility:I~.1 &2.*=~IExam Date: 1.~.~El
- ~~=~;======
- "'-'.
- .~=--:'-':=======:'..==~I~RO~G~r:::ou::":'pI[JJ ISROGroup ILf
~QQIS§QL~
I ~cord Number
- L.....___ '
~~====~~~.=~=~~~--=.=~~-=-~=-=~
of the operational implications of the following concepts as they apply to the Reactor Coolant With Rx power <36%, the reactor will not trip upon a loss of a single RCP. 14 RCP will cost and flow will reverse in that loop. A differential pressure will exist across a loop in which a RCP is running due to the head of other RCPs applied to the cold leg side of the vessel. This high pressure (cold leg side) will induce flow in an idle loop in the reverse direction. This means reactor coolant from Icold leg will flow back through the idle RCP and SG. This will lower Tavg in the idle loop. Th and Tc in
!affected loop will reverse, and the original Tc will be greater than Th. The 2 distracters with the Th/Tc eqLJaling are incorrect because Tc will rise >Th in that Reference Title
§eactor Coolant System Lesson Plan I Materia I Required for Examination I=.!Q~u_es~ti~on"",=,so_u_rce_
.. -::~=~=,...!::!c======...!
..::.::.::===-=-=====' IQuestion MOdifi~=thOd:
I Friday, September 14, 201212:07:11 PM
the following conditions:
Unit 2 is operating at 100% power.
21 Charging pump is CIT 23 Charging pump trips.
1-22 Charging pump cannot be immediately started.
I IWhich e>!the followi£l9 describes t~e impact of!~e loss of S~<:ll Injection FI()W?m~
l!J level will lower due to the lower sealleakoff flow, and auto makeup to the VCT will
[]
Ilf seal injection cannot be restored within 5 minutes, operators~WiIl trip the Rx lAW CAS actionl in S~.OP-AB.RCf=l-0001, Reac!or Coolant ~ump Abnormi3lity.
......_________--!J
~ flow from the RCS past the Thermal Barrier heat exchanger will maintain RCP seal
!temp~rature and allow plant operation to con!inue while at!~mpting to re~tore charging flow.
@] ~3 Charging pump (Unit 1) will be started and supply Unit 2 charging header lAW S2.0P AB.CVC-0001, Loss of Charging. This will require a Unit 1 shutdown to be initiated due to 13 9harging pumesu~tion alignm~f"lt to the Unit 1 RWST.
of the of the effect of a loss or malfunction on the following will have on the Reactor
.b(7) A is incorrect because when all 3 charging pump breakers are open, letdown oriface isolation automatically shut, so letdown flow is zero. Charging pumps are using no VCT capacity. Seal will still be going to VCT, so VCT level will be rising. B is incorrect because as AB.RCP directs Rx trip if BOTH seal injection and Thermal Barrier flows are lost, not just one or the other. C is correct Ibecause it describes the flowpath of RCP seal cooling flow when normal seal injection flow has been lost..
o is incorrect because AB.CVC-1 direct lining up Unit 1 PDP (13). it would require a Unit 2 shutdown, not!
a Unit 1 shutdown based on higher borated water supplied from Unit 1 RWST to Unit 2 RCS. Trainee Imay recognize that TS 3.0.3 is present when no charging pumps are available, but correct answer is worded to allow time to to restore flow.
Reference Trtle
.. m ________
.--......... ~-
...1 RE!actor Coolant Learning
.--'---_~_.m.m_.....*
i RCPUMPE008 LOR Identify and describe the Control Room controls, indications, and alarms associated with the Reactor Coolant Pump, including:
The Control Room location of Reactor Coolant Pump control bezels and indications. (Licensed Operator & STA only)
The function of each Reactor Coolant Pump Control Room control and indication. (Licensed Operator & STA only)
The effect each Reactor Coolant Pump control has upon Reactor Coolant Pump components and operation. (Licensed Operator & STA only)
The plant conditions or permissives required for Reactor Coolant Pump Control Room controls to perform their intended function. (Licensed Operator & STA only)
The setpoints associated with the Reactor Coolant Pump control room alarms. (Licensed Operator & STA only)
_m_
_m..
.~_
IMaterial Required for Examination Page 36 of 90
Question Source:
Exam IQuestion Modification Method: I[Concept Used
'--_-'-________--' ~=onceptOf ReS flowing up shaft to supply seals. Added what is the effect part toillatch KIA.
14,2012
, Letdown Demin Bypass Valve, will automatically reposition to bypass the eves Mixed Bed Demineralizers at t~(Sl~~tdown HX Outlet temperature of....
EI to of the resin beads.
§]
of the resin beads.
E1
@]
~~=:;'~~;==;I~==~lc;o=g=n;iti~v=e;Le=w~1~I=M=e=m=o=~========~I;I;Fa=C;il;iw=:~lils=a=~le=m==1&;;2====~II;EX=a=m=D;a=w=:~L****.*.*.****************===1;2!;3/;2;01~21 IRO Group I [JJ ISRO Group I OJ
~()()4()QQ!<416 J nOliVle(:lOe of Chemical and Volume Control System design feature(s) and or interlock(s) which
.b(5) The 2CV21 will reposition to divert flow from the CVCS demineralizers at 136°F. All the n""r"""7""rC! are in series, with the Mixed bed normally in service. The Cation Bed is placed in service Lithium control at power, and the deborating bed is placed in service at EOL from RCS boron concentration control. The stem states "Mixed Bed Oemin". As temperature rises, resins affinity for boron lowers and boron is released in system, not removed more.
Learning Objectives LOR NCT Describe the function of the following components and how their normal and abnormal operation affects the and Volume Control System:
Letdown/Charging Letdown Isolailon Valves, CV2, CV277 Regenerative Heat Exchanger Letdown Orifices Letdown Orifice Isolation Valves, CV3, CV4, CV5 Letdown ReleifValve, CV6 Letdown Line Containment Isolation Valve, CV7 RHR Flow Control Valve, CV8 Letdown Heat Exchanger Low Pressure Letdown Control Valve, CV18 Temperature Control Valve, CV21 Demineralizers (Mixed Bed, Cation, and Deborating Inlet Valve to Deborating Demin, CV27 Reactor Coolant Filter Diversion Valve, CV35 CVCS Holdup Tanks Volume Control Tank VCT Isolation Valves, CV40, CV41 Chemical Mixing Tank Charging Pumps (Centrifugal and PD)
Miniflow Recirc. Valves, CV139, CV140 Seal pressure Control Valve, CV71 Chg. Line Containmentlsol. Valves, CV68, CV69 Charging to Loop 3 Valve, CV77, Loop 4 Valve, CV79 PZR Auxiliary Spray Valve, CV75 CCP Flow Control Valve, CV55
- b. RCP Seal Water Seal Water Injection Filters Seal Bypass Flow Valve, CV114 Seal Water Retum Isolation Valve, CV104 Seal Water Return Relief Valve, CV115 Seal Return Cont. lsol. Valves, CV116, CV284 Seal Return Filter
Seal Water Heat Exchanger
- c. Excess letdown Excess Letdown Isolation Valves, CV278, CV131 Excess Letdown Heat Exchanger Excess letdown Flow Colrol Valve, CV132 Excess Letdown Diversion Valve, CV134
- d. Makeup Primary Water Storage Tank Primary Water Makeup Pumps Boric Acid Batch Tank Boric Acid Tanks Boric Acid Transfer Pumps Boric Acid Filter Boric Acid Blender i Primary Water Flow Control Valve, CV179 l
Boric Acid Flow Control Valve, CV172 Charging Pump Suction Valve, CV185 VCT Makeup Isolation Valve, CV181 Rapid Borate Stop Valve, CV175
...--- -------.__....... :---====-=:--:--=:==c.=-c-===
Question Source:
IQuestion Modification Me~l====u=se=d=======
Vision Q39286 Used concept of letdown flow divert at 136"F, made 2 and 2 and also added why it diverts.
~------------------~
14,201212:07:11
Given the following conditions:
Unit 2 is in MODE 4.
RCS Cooldown is in progress.
21 RHR Pump and Heat Exchanger are in service to provide shutdown cooling.
22 RHR loop is aligned for ECCS.
CRS directs the RCS cooldown rate be REDUCED.
101 the 1",lowing, which describes how RHR system flow will be adjusted to lower the coldown rate?
~ IThrottle closed on 21 RH18, RHR Heat Exchanger Flow Control valve, while throttling clo.se.d o.n j 2RH20, RHR Heat Exchanger Bypass valve to maintain total RHR flow constant.
[]
IThrottle open on 21 RH18, RHR Heat Exchanger Flow Control valve, while throttling closed on 2RH20, RHR Heat Exchanger Bypass valve to lower total RHR flow.
E1IThro.ttle open on 21RH18, while throttling open on 2RH20 to raise RHR Heat Exchanger bypass flow.
@] IThro.ttle closed on 21 RH18, while throttling open on 2RH20 to maintain total RHR flow constant.
IAnswer I[] IExam Level] ~llcognitive Level IIApplication IIFacility: IISaiem 1 & 2 I I Exam Date: I 12/3/201]
~ ~t Systems IIRO Group I 0 ISRO Group I [JJ 005000A402 1
'005
] IResidual Heat Removal System I I Record Number i 32 1
~Ip'~ility t~_Il1~I1~.ally operate and/or monitor in the control room:
.1
'A4.02 IlEeat exchanger bypass flow control I~ W]
Explanation of i55.41.b(5,8)Throttling closed on the RHR HX outlet valve while throttling the bypass valve open will pass Answer iless water through the RHR heat exchanger, therefore reducing the cooldown rate while maintianing stable total RHR system flow. A is incorrect because throttling closed both the RH 18 and RH20 will not Ir:naintain flow constant. B. & C are incorrect because it. WOUld. raise..the cooldown rate by passing more lflow through RHR HX.
Reference Title
~nitiatingR H R I
===========================================================~
IRHR Simplified Drawing I
I
=====================~I Learning Objectives i RHROOOE004 LOR NCT Describe the function of the following components and how their normal and abnormal operation affects the Residual Heat Removal System:
a)
RHR Pumps b)
Refueling Water Storage Tank c)
Heat Exchangers d)
Motor Operated Valves i)
RH1 and RH2, Inlet Isolation Valves ii) RH4, Pump Suction Isolation Valves iii) SJ44, Containment Sump Isolation Valves iv) SJ69, RWST to RHR Suction v)
RH29, Miniflow Recirc. Valves vi) RH19, Loop Isolation Valves vii) SJ45, RHR to SI or Charging/SI Pump Suction viii) SJ113, CCP-SIP Suction Cross-Connect Valves ix) SJ49 Outlet Isolation Valve x)
RH26, RHR Hot Leg Isolation xi) CS36, Spray Recirculation from RHR Valve e)
Air-Operated Valves i)
RH18, RHR HX Outlet Valves ii) RH20, RHR HX Bypass Valve f)
Other System Valves Friday, September 14, 201212:07:11 PM I
[~_ag~_~~of90 _
i)
RH3, RCS to RHR Inlet Relief Valve ii) RH25, RHR to RCS Hot Leg Relief Valve iii) SJ48, RHR to RCS Cold Leg Relief Valves iv) RH12, RHR HX Bypass Valve v)
RH17, RHR to CVCS Letdown vi) RH21, RHR to RWST Containment Sump Anti-Vortex Baffle Orifices IMaterlal Required for Examination Modified I=:Q~u=e=st=:=lo=n:::::=:=s=ou.r...ce==::;:::L:==~=i-;==:==:================:== IQuestlon Modification Method:
]
...-~---
lFriday,Se~14, 201212:07:11PM!
the following conditions:
Unit 2 is in MODE 5.
BOTH loops of RHR are in service for Shutdown Cooling.
RCS temperature is 190°F and stable.
Each loop is supplying 1800 gpm flow.
2RH20 is 10% open.
Conditions to transition to MODE 4 are NOT met.
1 The air line supplying the 21 RH18, RHR HX Outlet FCV breaks, and air is lost to 21 RH18.
. Which of~~~followi~g describes the initial effect thi~ airline failure will have?
1 E]
will be no effect on the RHR
~
will have to be throttled in the direction to
@J 8 will have to be throttled in the closed direction to aRCS cooldown.
~------------------------------------------............ ----~----------------
~
@] 122~R~H1 B~iII have to be throttled opened to ensure RCS temperature is mail!tained <200°F.--]
.. -=:J ~~;'dN~ __......... __-'-'-'
nOIlVle(me of Residual Heat Removal System design feature(s} and or interlock(s) which provide for the of RHR heat flow pp
'~55.. 1.b(7.8} A is c
.. ause the RH18. valve.s are fai.1 a Lo.si.ng the air su. Iy will not affect. I
...4
.. o..rrect bec
.. s-is valve.
L..,...;.......;......_---I the stable system conditions as sdescribed in the stem. B is incorrect but plausible if it is thought the 21 RH18 fails open. C is incorrect but plausible if it is thought the 2'1 RH18 fails open. 0 is incorrect but plc:llJsible if it is tholJ9ht the 21 RH 18 fails shut.
j Reference Title I
~s oiControl Air LOR NCT Describe the function of the following components and how their normal and abnormal operation affects the Heat Removal System:
a)
RHR Pumps b)
Refueling Water Storage Tank c)
Heat Exchangers d)
Motor Operated Valves i)
RH1 and RH2, Inlet Isolation Valves Ii) RH4, Pump Suction Isolation Valves iii) SJ44, Containment Sump Isolation Valves iv) SJ69, RWST to RHR Suction v)
RH29, Miniflow Recirc. Valves vi) RH19, Loop Isolation Valves vii) SJ45, RHR to 81 or Charging/SI Pump Suction viii) SJ113, CCP-8IP Suction Cross-Connect Valves ix) SJ49 Outlet Isolation Valve x)
RH26, RHR Hot Leg Isolation xi) CS36, Spray Recirculation from RHR Valve e)
Air-Operated Valves i)
RH18, RHR HX Outlet Valves Ii) RH20, RHR HX Bypass Valve f)
Other System Valves i)
RH3, RCS to RHR Inlet Relief Valve ii) RH25, RHR to RCS Hot Leg Relief Valve iii) 8J48, RHR to RCS Cold Leg Relief Valves Ifn(iay~septe~-be-r-14-,2-0 1212:(jT:~
.---:::-.......__::--:--:-:c---,
Iv) RH12. RHR HX Bypass Valve v)
RH17. RHR to CVCS Letdown vi) RH21. RHR to RWST Containment Sump Anti-Vortex Baffle Orifices IMaterial Required for Examination F.Q:=u=e=s7.ti:=o=n=:s:=o=u=rc=e=:=:=!-==~=,--r=============.=.....=m.m=m,.IQUestion Modification Method:
14.2012 12:07:11 PM
~~~~m~~~~~___
Unit 1 is operating at 100% power when a total loss of ALL AC power occurs.
iWhiCh of the following identifies a consequence if ALL AC power remains deenergized for at least
~e day?_
-'--_--'----=:-~.=-=--.=-_==-'--========~
I!JLossoiECCS pumped injection capability coupled with RCP seal leakage will result in core uncovery.
~ IContainment degradation due to the sustained pressure above 15 psig after RCDr reliefs lift
~nd remain opel1'~
........~.....................
EJ in containment as RCS inventory is released which will complicate recovery when AC nn"AI<:.r is restored.
@JLoss of makeup capability to the RWST will result in lowering level below Tech Spec required for accident re~~very.
IAnswer IliJ lexam Levell ~ ICognitive Level IlI,n0''''''' I IFacility: IISaiem 1 & 2
~~~=~~~==~"==~========~==~IR~O~G=ro=up~I~1 11ISROGrooplc=]
=-"--__..-1IEmergency Core Cooling $ystem Knowledge of the effect that a-loss-or-malfunctiOn ofihe Emergency Core Cooling-System will have on
[following:
K~:QL..
pressure is expected to rise to - 3 psig and 40°F as the RCS drains through the RCP I..._____----'.seals. Time to core uncovery as shown on Figure 40 in lesson plan, best case, is <20 hours. RCS inventory will be released to containment, however. the containment is designed for a LBLOCA in which the mass in the RCS is released to the containment and long term recovery is not affected. RWST will not be since it has no flow nor motive force.
Leaming Objectives the rRsrlOm'R of the reactor coolant seals to a IQuestionModification Method:
I 1...-________---1 [6CCOOk2002 NR?~~~Ill' modified to Salem conditions and replaced poor distracters.
SAn,lF!rTlt"lF!r 14, 2012 12:07:11 PM
Given the following condition:
- Unit 1 has intiated a Safety Injection in response to a LBLOCA.
Choose the set of valves which would prevent some portion of ECCS injection flow from occurring if Ithe~NOT rep~on upon !b~§1 Signal.
~.__
~.__.......
El 2 AND 1SJ13, BIT Outlet.
~ 1-1 ECCS Accumulator Outlet.
EJ ANI:)---"?~~ RHR Discharg~~~~~eg.
@J 1SJ44 AND 1 Containment Isolation.
1
';;:;::::::::~=~======='--'==='-=lc=o=g=n=iti=ve=L=e=ve=I=I-=l6=p=Q=ii~ca~t::;:io~n=.:::::::;"mIIFacility: I [salenlI& 2-- Iexam Date: I 12/~!201~
IRO Group I ISRO Group IQQ6J!QQ.K.~iO
==::~~=======================~==~~~====~
1006~-=J[Eii1efgen~y Core Cooling $ystem
=="'I~ecordNumber I n"\\llfi",,1nA of the of the effect of a loss or malfunction on the following will have on the Emergency Core
.b(8) A is correct because BIT inlet (SJ4/5) and outlet valves (SJ12/13) are normally shut and L..:..:.::"';';":'~_....I*r'~ran/a an open signal from SSPS on a SI. B is incorrect because SJ54 valves are opened and
.rl"":>n""rni:7,,,rl at 1,000 psig during a plant startup. Stem states LBLOCA so plausible since accumulators will inject during LBLOCA. C is incorrect because SJ49 valves are normally open at power, and do not during a LOCA, but would expect to have ECCS injection flow from RHR pumps during LBLOCA.
is incorrect because SJ44 valves are opened in LOCA-3 during transfer to CL recirc, and do not nrr-nllnl'"
ECCS flow.
Reference Title
! ECCSOOE016 Given a Emergency Core Cooling System failure, predict the effect of the Emergency Core Cooling following: (License Operator and STA only)
Reactor Coolant System Containment Nuclear Fuel IMaterial Required for Examination Question Source:
IQuestion Modification' MethC)d;
'---------_.....1 '--__~...... __ ____
.~
..~~ge4~5~o~f9~O----'J
[Friday, Septembf:)r 14. 201212:07:11~
==-P~
Which of the following describes Pressurizer Relief Tank (PRT) response when a bubble is being drawn in the PZR after a vacuum refill of the Reactor Coolant System lAW S2.0P-SO.RCS-0002, Vacuum Refill of the RCS?
PRT....
~ Ipressure will rise SlOW.Iy as operators vent air and non-condensibles by opening the Pressurizer PORVs.
~ Ilevel will rise rapidly as the Pressurizer PORVs cycle automatically in response to the solid PZR expanding.
§] Ipress.ure will rise rapidly as the Pressurizer PORVs cycle automatically in response to the solid PZR expanding
~
@J Ilevel. will rise slowly as operators maintain Pressurizer pressure during RCP bumps by opening I the Pressurizer PORVs.
IAnswer I~J IExam Levell rFL::J ICognitive Level IIMemo_r)' ______
I !Facility: I [Salem 1& 2 I I Exam Date: I 12/3/20@
n
~ ~It Systems IIRO Group I [JJ ISRO Group I [JJ 1007000K502
[Q07____] leiessurizer Relief Tank/Quench Tank System
.((Record Number II 36 1
!jS§=:J IKnowledge of the operational implications of the following concepts as they apply to the Pressurizer Relief ITank/QuenchTan~k~S~y-s~te-m-:----~~~~========~~============================~~
U K5.02 ] ~ethod offorrningUa steam bubble in the PZR IeM [M Explanation of 55.41.b(3) A is correct because operators will perform a 10-15 minute vent of the PZR while drawing a 1
Answer
- bubble (Step 5.3.28), with PZR level 40-60% (Step 5.3.5). There will be minimal liquid carryover, but
!venting will slowly raise PRT pressure. Band C are incorrect because the PORVs are controlled in Imanual. 0 is incorrect because the RCP bumps are performed prior to a vacuum being used in the RCS, I,and PORVs are in auto during bumps, but will be opened after the rCP is secured for venting. (Step Reference Title IVacuum Refill of the RCS I
===========================================:
I Learning Objectives PZRPRTE012 NeT Discuss the procedural requirements associated with the Pressurizer and Pressurizer Relief Tank, including an explanation of major precaution and limitations in the Pressurizer and Pressurizer Relief Tank procedures I
I IMaterial Required for Examination II Question Source:
IIQuestion Modification Method:
II L..---____Il_
I Friday, September 14, 201212:07:11 PM I
one of the following describes the normal and loss-of-air positions of the Component Water Surge Tank Vent Valve 2CC-149?
and fails
~
L..~___~~~.______________________.~______~.______________________________________~
@J
~~=~~~~~~====~~~====~~~~==~~~~~~~~~
- ===:;-:--=~.======~:::==_'"lc=o=g=n=iti=ve==Le=v=el=_'II:.:.M:::.~~~-. __JjFacility:Il~~~....
J!exam Date: I 12/3/201~
IRO Group I ISRO Group I
[008000K4Q.!LJ
~~==~~~=====================~~~
cQomponenlCooling WCiter Systemm~Record~umber *i of Component Cooling Water System design feature(s) and or interlock(s) which provide for the associated valves and controls 2CC149 is a normally open vent valve, and fails shut on loss of air (and loss of control Reference Title Learning Obiecti!veS CCWOOOE006
- NCT Outline the interlocks associated with the following Component Cooling Water System components:
-.------I CC-149, Surge Tank Vent Valve CC-131, RCP Thermal Barrier Discharge Flow Control Valve RHR Heat Outlet Isolation Valves IMaterial Required for EXamination Question Source:
Bank ___...... __
... ___............ ____ IQuestionM~dificationM~od:
I
........~d
'-'-_________.... [ViSion Q39_3()2 editorially modified to remove window dressing and common com~~ents foundin.all.choices.....
14,
I
- Following a loss of offsite power, 2A 4KV Vital Bus fails to reenergize.
IWhiCh of the following describes the PZR heater group{s) which are available, or will be made la"<:1il<:1~,_to maint<:1~,!~Z~JJressure vvhil~rE?sJ>()~9 lAW TR~J~s EOPs?
~
21
[]
~-.. ~~
and control heat~I_gr~o_u-,-p_s_.____
~~~i~~_~'}~~~~$$.~~~~~~~~~~~;;==~.=IF~ac~iI~i~~:~I~;a=le=m~1~&=2~~I~I~_a_m_D_a_m_:71=====1=V=3=12=Ol=.21 ISRO Group I[JJ
.b(7) Control Group heaters are powered from 2G non vital bus, and does not have an emergency supply. 21 Backup Heater Group is normally powered from 2G non vital bus, but has an emergency power supply from the 2C vital bus. 22 Backup Heater Group is normally powered from 2E non vita~bus, but has an emergency power supply from the 2A vital bus.
480V Pressurizer Heater Bus One-Line 4BOV Pressurizer Heater Bus One-Line Learning Objectives power supply to the Pressurizer Pressure and level Control components:
Heaters Heaters Valves F.:Q:=u=es=::t=;i=io=n=::::s:==ou=r_ce=:~~~=~I'::;::-===:==::=:===C=:===::::~ JQuestion MOd~~~~ethOd:
14,
the following conditions:
Unit 2 is operating at 100% power.
1-A power reduction from 100% to 20% Rx power will be performed at 1 % per minute lAW S2.0P
!AB.LOAD-0001.
Prior to initiating the down power, the PZR Master Flow Controller is placed in manual and is NOT adjusted during the down power.
I IWhich of the following is CLOSEST to what actual PZR level will be when the down power is c;omelete~_§md RCS Tavg_~)<<(lctly ()~~g~am?
El ~.=_____
...__.__.,,'=_=-==-:::c_._...
[]
~
EJ [£O{o.
@J L-________ ________
______________________________________________________________~
IAnswer I!b ] !Exam Level!
!Cognitive Level Il!iomp@IJ~~.llFacllity:ISalem 1 ~.._2_.... I I Exam Date: I __ m~.c'12
~~==~==='~~~~=====~~
~f:]
IROGroup I[~ !SROGrouP IOJ
.011000K6.Q+/--J
~!O-"1-"-1_____J.Pressurizer Level Control System I Record Number 1,---------"
iK6..~ IKnowled-ge-oiihe o-f-t-h-e-effectof a loss or mal-fu-nc-tion onth-e-fo-liowing-w-i-lihave on the Pressuri-ze-r-L-evel JI"~*rl'lTlnn of PZR level controllers
.b(7). Program PZR level is clipped at 59%. As the downpower occurs, there will be an outsurge I the PZR as the RCS contracts due to lowering Tavg.
At 20% power, ReS Tavg exactly on nrr",r*::o..... is 551.6.°, (AB.ROD-3 Attachment 1) which would give a program level of 28.3%. (AB.ROD-3 ). The PZR mass does not change from the down power. A is incorrect but plausible since it i is the no load PZR prograsm level. 0 is incorrect but plausible if the candidate thinks that with charging inl level remains at its current level. C is for Reference Title 1~-O-I1-!i-I1.l!...-O-usRo-d--M-o-tion Leaming Objectives.---------'_.
1_-
Identify and describe the Control Room controls, indications, and alarms associated with the Pressurizer Pressure and Level Control System, including: (Licensed Operator & STA only)
The Control Room location of Pressurizer Pressure and Level Control System control bezels and indications.
The function of each Pressurizer Pressure and Level Control System Control Room control and indication.
The effect each Pressurizer Pressure and Level Control System control has upon Pressurizer Pressure and Level Control System components and operation.
The plant conditions or permissives required for Pressurizer Pressure and Level Control System Control Room controls to their intended function.
IMaterial Required for Examination Question SourCt):
IQuestion Modification Method:
S"r.t"l'YlhAr 14, 2012
!With Unit2isai100% pO\\~er,Containmer'liPressure Channell (one) indication became erratic and the channel was removed from service lAW S2.0P-SO.RPS-0005, Placing Containment Pressure Channel in Tripped Condition.
Ipr~<:lit:;t the plant response if Contc:lir'lrnentPressure Channel IV (four) subsequently failsbigh.
I!J ~~.~~~~--~-
.......................... --~~-~~~~~~.......................... -~.---~~~~~............. -~~~-~-.....
~
actuation on 2/3 channels EJ Injection, Containment Spray, Main Steamline Isolation and Phase B Isolation all actuate.
Steamline Isolation and Phase B Isolation. Containment Spray valves rer:losltlon do not start.
ICognitive Level I!ApplicatiolJm IIROGroup I of the Reactor Protection '\\Jc~tolrn 55.41.b(7) Cont press Channel I only feeds Cont Hi-Hi (Spray act) it does not feed the Cont Hi L...-.;.......;;____---I circuits. Containment Spray system bistables are energized to actuate, so when the failed removed from service, its Spray actuation bistable is NOT tripped, it is removed from inputting to coincidence to prevent one of the remaining channels from actuating cant spray if it fails. This SI ciruitry still 213 on channels I, II, and III, and the containment spray actuation goes to 213 of the Iron"\\"'C,.,inn channels.
Reference Title a Containment Pressure Channel in the Condition Learning Objectives LOR State the setpoints. coincidence, blocks and permissives for all Reactor Trips and Safety Injections actuations (Licensed Operator and STA Only)
IQuestion Modification Method:
a LOCA, choose the ONLY one of the following which automatically occurs at 15 psig in containment.
IAssume there is a 3 minute ramp in c0rl!~inment~~s~lIr~J~om 4 psigto 15 EJ A Isolation.
~
Isolation.
~
~--..---------------------
............------------------------~
!Cognitive Level 1 nno,.,..."""
!Facility: !rsa1em1& 2 IIExam Date:!
mm1?~
~:=;-~=~~====::..::==-======:-:::.-=~~:;;;::::~-;:: =;-;!;:;SR~O~G;:rou==p~II=='I;-I--'
I013000~4Q:?u__ J
---'1 Record Number j l.................... _---.!
and/or monitor in the control room:
.b(7) All the distracters occur at 4 psig in containment pressure (81). Only the Containment pressure signal. The stem states there is a 3 minute time between the not a race as mightpccur during a LBLOC!\\.
Reference Title Learning Objectives ESFOOOE021 State the for automatic actuations associated with the Features
!Material Required for Examination.
- 1 Question Source:
Friday, September 14,2012 12:07:12 PM
the following, which one describes the purpose of the Enginneered Safety Features lAW Salem fuel c~(3_rl!perature to 2500°F.
.~.~.--
-~----
[] [li..imiting fission product dispersal to minimize popula~ti-o-n-e-x-p-o~sure for an..
..ac.Cident.a.1 re..le.a5el.
l~~yond the containment.
~...... ___
__ ~__. __~J
~ ~..SU.re.....t.he des
.. re. ' inconjun
...... with re
.. 1a.**** nd protec.t.iO.n systems w.ill
.. f9. n of the co
.. c.tio.n
.. a.ctor contro pre-,,_ent release of fission prpducts beyoJ1~the fuel qladding.
~_____ ~
@;] [EnSUre retentionoffission productsTn the Reactor Coolant System (RCS), and for operation-al'l and ~~!eleases beyond the RCS, retention of fissiol1 products ~y the contail1ment.,
IAnswer*1 !d! IExam Level I~] ICognitive Level II Memory.....
. I Facility: I[Salem 1 & 2~-_ *I Exam Date: r 12/3/20121
~ ~!§)'stems IRO Group I ISRO Group I 1013000GilLJ 013
] [Engineered Safety Features Actuation System
== lRecord Numberj and/or function.
55.41.b(3,9) A is incorrect because it is one of the ECCS Acceptance Criteria, not a pupose of ESF. B is l
!;"';;';;;';';';~_--I incorrect because ESF is designed to keep fission products in containment. C is incorrect because it is a precursor to having ESF components. That is, core design is meant to keep fission products in the fuel, whereas ESF is designed to keep them in the RCS, and upon leakage of the RCS, to keep them in
!containment. 0 is correct per Salem UFSAR, Section 6, page 6.1-1 which states:"The engineered safety*
features are the provisions in the station which embody methods 2 and 3 above..... " Methods 2 and 3 are the 2 parts of tbe correct ans'v\\fers.
i Reference Title
~~AR
.. __...._.......===
=
Learning Objectives State the purpose of the Engineered Safely Features. Include in this the following:
Residual thermal energy Barriers to fissi()n produc;t!elease
~IM~ate==,ri=a'=R~e=qu=jr=e=d~~~r=Eu==m=in=a=tio=n======I,=~=.~-~~-~=..======~-=-~.~==~=-~~~~~~~=11==~-~====~~~====
I'!!Q~u_es~tio!",n_s~.o_u_rce~:~=~;;.....,..~~~
_____ !QUeSfiOnModlflcation Method:
IGiVen the following conditions:
i_
Unit 2 is operating at 80% power performing a Tech Spec required load reduction at 1 % per minute.
Rx power must be below 50% in the next 35 minutes.
Boration and automatic rod control are maintaing RCS Tavg. 1.0-2.0°F above program.
- Power Range Nuclear Instrument Channel IV, 'I N44, fails HIGH.
I IWhiCh of the following identifies:
- 1. The effect this failure will have.
- 2. The action which should be performed.
~ IControl rods begin stepping OUT at 72 spm. Stop the load reduction until ROD STOP BYPASS is placed in BYPASS.
~ IControl rods begin stepping IN at 72 spm. Stop the load reduction until ROD STOP BYPASS is placed in BYPASS.
~
ElIControl rods begin stepping IN at 72 spm. Place control rods in manual.
@J ~rol rods begin stepping OUT at 72 spm. Place control rods in manual.
IAnswer I [J IExam Levell ~ ICognitive Level IIApplication
~IFacility:l~ 1~~___ JIExam Date: I 12/3/2012]
ITier:
IIEla_nt_~y~le~s________ _
] IRO Group I 0 ISRO Group I 0 1()15000A202 J
[01K-----] [Nuclear-Instrumentation* System IIRecord Number] I 431 IA2.~ Ability to (a) predict the impacts of the following on the Nuclear Instrumentation System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:
1A2.02J ~ulty or erratic:. operation of detectors or compensating components 1[3.1113.5*1 Explanation of :55.41.b(6,1 0) With a Tech Spec required power reduction underway,-and-having-slirn margin to achieve-i-t, Answer the load reduction will continue. As in both AB.ROO-3 (Continuous Rod Motion) and AB.NIS-1 Nuclear
.Instrumentation Malfunction, operators are directed to place control rods in manual (steps 3.1 and 3.1.1 respectively) and adjust rods in manual to control Tavg (steps 3.5 and 3.1.2 respectively.) AB.NIS has istep for terminating load reduction if in progress as a OR step with placing rods in manual. The ROD STOP is for outward rod motion only, and does not affect inward rod motion.
Reference Title IContinuous Rod Motion
====================================================~
INuciear Instrumentation System Malfunction I
============================:
Learning Objectives
~NIS1E003 a) determine the appropriate abnormal procedure in accordance with this lesson plan.
b) describe the plant response to actions taken in the abnormal procedure in accordance with this lesson plan.
,-----________,'==C)==de~cribe the final plant condition that is established by the abnormal procedure in accordance with this lesson plan.
1_____ __ "----_________________________________------'
L IMaterial Required for Examination II Question Source:
INew __
IIQuestion Modification Method:
II I
L...----------'L
_ ________..1 I Friday, September 14,2012 12:07:12 PM I
the following conditions:
LOGA is in progress.
Operators have transitioned out of EOP-TRIP-1, Reactor Trip or Safety Injection.
of the following indicates a superheat condition exists in the core, and what GFST is GETs> 1200°F. RED ISRO Group I[1]
017__j JD;Q(.)"riI~l'l1j:)~r<:lture Monitor Systemm____...*~
nOlivle(lae of the operational implications of the following concepts as they apply to the In-Core Monitor
,55.41.b.{5, 1 O)With CETs indicating> 1200°F.... "this temperature indicates that most liquid inventory
~=:"'-_....J already been removed from the RCS and that core decay heat is superheating steam in the core."(page
- 16)
This is a FRCC RED path entry to FRCC-1 Response to Inadequate Core Cooling. With CETs 1>700°F... "superheat at the core exit is indicated. An ICC condition will exist if in the next block RVLlS indicates <3.5 feet collapsed level in the core" 39% RVLlS=3.5'. This means distracter C is incorrect because RVLlS is 51 %. D is incorrect because it is a RED path entry to FRCC-1, not a PURPLE path
,entry into FRCC::?: 8 is inc()rr~ct becau~e it is the IiVrong (PURr:>LE path) pri()rity.
Reference Title
[§iticaISaf~~Functiol1~§tatus Trees Basis Document Learning Objectives
. IMaterial Required for Examination "
Question Source:"
IQuestlon Modification Method:
[Friday, September 14, 2012 12:07:12 PM
Given the following conditions:
Unit 2 has experienced a LBLOCA coincident with a loss of off site power.
2C 4KV vital bus locked out on bus differential.
2B SEC did not actuate.
IASSUming one train of ECCS equipment is operating, which of the following identifies the FIRST laction which will restore the minimum complement of equipment to assure containment integrity is
!maintained lAW Salem FSAR?
EI 2C SEC.
[g E1 START PB for 21 CFCU.
START PB for 22 AND 24 CFCUs.
Ability to predict and/or monitor changes in parameters associated with operating the S\\J'~tolm controls 5541.b.(8)FSAR Section 6 and 15 both state that the minimum complement of Containment Spray L......;......____---I,Pump/CFCUs required to ensure containment integrity along with a train of ECCS in operation is 1 CS
!pump and 3 CFCUs. With the conditions in the stem, only 21 CS pump will be running on A bus, C bus
!will be deenergized because a bus differential signal locks out all power to the bus. Additionally, the Ipower supplies to the CFCUs are A,B,C,B,C for 21-25 CFCUs, so only 21 CFCU will be in operation. A is incorrect because the SEC can't start any loads until the bus has power. B is incorrect because it is already running. C is correct because that will restore the 3 CFCUs needed. D is incorrect because 21
- CS pump will already be r~r1r1ing..
Reference Title the design bases of the Containment the design bases and values for the following parameters associated with the
,",V<"PrY'" (Licensed Operator & STA only)
Negative Pressure Positive Pressure IMaterial Required for Examination Question Source:
IQuestion Modification Method:
'I'vision Q113269. M.O
..... answers (A and D) had
.....dified Distracter D from 22 to 21 CS pumps since 2 of the L-________---I equipment that would not be available due to loss of 2C bus.
.................... _____~ ~_.....~
14,2012 12:07:12 PM
the following conditions:
1-
.A loss of reactor coolant has occurred which results in containment pressure rapidly rising to 1 pSlg.
walking down the control boards 25 minutes later to prepare for a crew brief, which of the II"\\\\o.,,,nl'1 locked in Overhead alarms would be EXPECTED for these conditions?
DET VOLT TRBL.
] IExam Date: 1-- 1273720121 1026000G431 J n01NIE~dare of annunciator or i5t.55.41.b(8)Spray eductor flow is nominally 75 gpm.Normal level 75%. Administratively maxievel 900/01 L..-___---I 1(3900 gal as shown in SC.CH-AO.CS-0415) Alarm occurs at 67%. Normal level 3,400 gallons. Alarm at I
- 3,050 gallons. 0-43 will occur - 5 minutes into event. Transition out ofTRIP 15 minutes. E-5 is not !
expected to be in alarm 25 minutes after the trip because it is indication of the SR instruments being i.*
energized when they should not be with respect to turbine power above or below 15%. Source range
!automatically energize between 15-18 minutes following a trip. C-29 is not expected because it indicates I*
'CFCU SW valve alignment problem with the CFCU running. It is plausible because the CFCU Airflow Trouble alrm WILL be in alarm, as it occurs when the damper alignment is not correct for running in HIGH i speed, and the CFCU will be running in LO speed following an accident. B-7 would not be in alarm as SW
,to the TC?A is automatically i~olated by the SECs:
Reference Title
.._J 1..8i11HIIf.Y* Objectives I c~
Identify and describe the Control Room controls, indications, and alarms associated with the Containment Spray System, including:
The Control Room location of Containment Spray System control bezels and indications. (Licensed Operator & STA only)
The function of each Containment Spray System Control Room control and indication. (Licensed Operator & STA only)
The effect each Containment Spray System control has upon Containment Spray System components and operation.
(Licensed Operator & STA only)
The plant conditions or permissives required for Containment Spray System Control Room controls to perform their function.
Operator & STA only)
The associated with the Containment control roam alarms.
1Material Required for Examination QUestion Source:
IQuestion Modification Method:
'----------11 14, L£'age 5aof 90 I
~.~"...--
the following conditions:
Salem Unit 1 was operating at 100% power when a LOCA occurred.
A manual reactor trip and manual SI were initiated.
When the Main Generator output breakers opened, a loss of off-site power occurred.
1 A vital bus locked out on bus differential.
I
~~~~~~~;t~~~~II~Wi~9_id~~tifies which Contai~ment Iodine Removal Units (IRUs) can be started if El 1 IRU ONLY.
~
I IL ""...I to the following:
55.41.b(9) Containment IRUs are powered from G and E non-vital 460VAC. With the loss of off-site a;...;;.;;';;";';";~_--I iPower, none of the non vital busses are energized. The distracters are based on the operator knowing
!that the I~ading of a vital bus in Mode IV doesn't have any bearing on IRU operation.
Reference Title.
[1,E1 Aux Building 460-230V One line 460-230V One line I CONTMTE004 State the power supply to the components, including voltage level and 1 E/Non 1 E.
Containment Fan Cooling Units, including breaker alignment for Fast and Slow speed.
Containment Iodine Removal Fans (Licensed Operator & STA only)
Control Rod Drive Ventilation Fans (Licensed Operator & STA only)
Reactor Nozzle Support Ventilation Fans (Licensed Operator & STA only)
Reactor Shield Ventilation Fans (Licensed Operator & STA only)
Hydrogen Reco_lllbillers (Licensed Operator & STA only)
IMaterial Required for Examination
[Friday, September 14, 2012 12:07:12 PM
IWhiCh of the following describes how uncovering of the fuel in the Spent Fuel Pool (SFP) is prevented if a leak were to develop on the in-service Spent Fuel Pool Cooling Pump?
EJ IAutorl"l~~()l~~~u~f Cooling,pump0r] I()w level ill t~e SFF" I
~ ~utorl"l_aJi~mJ3keup~() the SFP com~in~~~l!bt~~lo le_veLal§r_~!()~_rt~~e Control Room.
I E] ILocating the SFP Cooling pump suction line close to the surface of the pool, and an anti-siphon hole in the return line to the SFP.
@J ~;~nK~;;~f;_~~~~!~~ ~~:t~ ~~~~El~~~~~~o;~~op~_~~_~~~~e of the pool and an anti-IAnswer IEJ IExam Level] ~ ICognitive Level IIMemory IIFacility: IISalem 1 & 2 I I Exam Date: I 12/3/20@
ITier: I~t Systems IIRO Group I [] ISRO Group I []
lO_33QQOI<-403__J
[033
] ISpent Fuel Pool Cooling System I [Record Number 11 48 1
.K4~~rK~()""'ledge ofSpent Fuel Po()1 Coo~ng Systemdesign feature(s) and or interlock(s) which provide for the
- following:
~--'=i IK4.03 I [~nti-siphon devices 1[2][2]
IExplanation of I55.41.b(4) There are neither auto SFP pump trips nor auto M/U to the SFP. The suction line is located 4 IAnswer ft below the surface to minimize level lost if a leak were to develop below the level of the pool in the SFP Cooling system, while the return to the SFP is located 6' above the fuel. The return line has a 1/2' hole drilled in it to prevent it from siphoning the level back to the SFP Cooling pumps on a leak.
Reference Title ISpent Fuel Pool Cooling I
=======================i I
Learning Objectives SFPOOOE013 Given a Spent Fuel Pool Cooling System failure, predict the effect of the Spent Fuel Pool Cooling System failure on the following: (License Operator and STA only)
Spent Fuel Ventilation Radiation Monitoring System Spent Fuel Temperature IMaterial Required for Examination II Question Source:
__ -c=_IJQUesti~~~~~~~~~~ IIConcept Used
~""",!,~~_~,L.;;;=~=,--,;=:;;=;=========
Vision Q60120 regarding when in service pump would lose suction
~----------------~
I Friday, September 14, 201212:07:12 PM I
I. Page 58 of 90 _ I
a steam leak, the CRS directs the PO to FAST close 11 MS167 from the 1 CC3 bezel. The depresses the NORMAL close PB on 1 CC2 instead, and the 11 MS167 starts closing IWhich choice describes what will happen if the operator then pushes the FAST close PB for
[11MS167?
~
vent valves11 MS169 and 111V1S171 immediately open, allowing hydraulic pressure to iclo~E:lvalve agaii"\\st only a~l'l'"l0spheric pressure.
~
MSIV hydraulic pump immediately stops, depressurizing the hydraulic header, and allows steamp,"-~§su~to ~Iose the__ v___a_l_v_e_._______ _
E] iThehydraulic sequence will continue until the Valve Fully Closed (33CVO) contact is closed.
IAII o!hergperation of the valve i~ locked out until the hydraulic pump is deenergized.
@J valve immediately opens, equalizing hydraulic pressure on both sides of the piston, and vent valves 11 MS169 and 11 MS171 open to allow main cto~rn the valve.
ICognltive Level I [.A.pplicatic.n IFacility: 11~~1~2------=IIExam Date: I 12..':3/2012
- .;
- .
- =;--~
ISRO Group I IQ39000G12S-1
=,"--__.-1 IMain and Reheat Steam System I [Record NUlllber-1 i2LIcOnduct Of Operations
~============================================~==
- 2.1.28 I[Knowledge of the purpose and function of major system components and controls.
IW 14.1l Explanation of
.b(4) Using Logic drawings 239916 and 239917: The Emergency Trip signal is generated from the Answer CLOSE PB, and acts just the same as a Safeguards Train MSLI or High Stm Line Flow SI signal.
- SV-1 closes (was open to direct hydraulic pressure to bottom of hydraulic piston.) SV-3 opens, equalizing ihydraulic pressure on both sides of the hydraulic piston, and allows the hydraulic fluid to act as buffer to Iprevent 11MS167 from slamming closed. The solenoids for 11MS169 and 11MS171 open, venting air,
- and the valve open to allow MS pressure on the bottom of the lower operating piston to drive the disc up iagainst atmospheric p~~ssure. The hydraulic pump does immediately stop rlJ~ning.
Reference Title the power supply to the Main Isolation Valve QHydraulic Pumps Isolation Valve Vent Valves IMaterial Required for Examination
~Q=ue=s='ti=on=s~o=u=ru=::'~~~~~*****~******~****~-~-********~--~========~IQ~u~e~st~io~n~M~.od~"~i~~tiO~n~M~e~th~O~d:~..~J~ID~i~re=ct~F=ro=m=s=ou=r=re~*~~~l
[Vision Q80671
~~------------~
following conditions:
A Unit 2 plant startup is in progress.
Reactor power is stable at 18%.
The Main Generator is rolling unloaded at 1800 rpm.
Main Steam Dumps are controlling in AUTO in MS Pressure control.
MS Dump Pressure setpoint is lowered 5 psig.
,With no other ope~ator action, ~everal minutes latf3~_~_~_otice~____... __ _
I!J--~........--.
~
is> 18%.
El [COrltrol rods ~~\\,e stepped in.._
@J out.
IExam Levell rs= ICognitive Level I ~pJic;ation.__ ] IFacUity: IIS!lIE":lrl1~ 2 mm~IIExam Date: I
--12/3/2§1]
OJ IQ39000K508 ]
~~====~~==~~===============..~==~J~~====~_~=-__ fRee-arC! Number *'--____---'
==========~~'=~==~~~~~~~~~~==-
of the operational implications of the following concepts as they apply to the Main and Reheat of steam removal on reactivity
. b(5}With steam dumps open in MS pressure control auto, lowering the setpoint will cause sterna L;..;;.;.;;..;.....;___"-----l dumps to open to reduce steam header pressure to setpoint pressure This will cause a higher steam flow and lower temperature, which causes higher Rx power. Control rods are not placed in auto until >P-2, which is '15% Turbine power, which is not online yet, so rods will be in manual and no operator action istated in stem.
Reference TItle I
m J
IMaterial Required for Examination From Source
- .==,====C::::::,C.==.IQu~~~o_n_M_o_d~~tion Method:
modified from rod pull to lowering MS dump setpoint.
Friday, SE":lptember 14, 2012 12:07:~~ii'i]
the following conditions:
Unit 1 is operating at 100% power when a Main Turbine trip causes a reactor trip.
The Main Steam Dumps do NOT arm.
of the following describes the effect this failure to arm will have on the Reactor Coolant over the next 5 minutes.
will exceed below 2335 IExam Level I LJ ICognitive Level IiA~)plicaJ~ ___IIFaCility: I 1
IExam Date: L IRO Group IQ ISRO Group I[]
155.41.b(4,5,) With a reactor trip, core heat production will lower rapidly. The Steam Generator L--___-IIAtmosphereic Reliefs, MS10s, will open to establish ReS temp -551-552 psig. The ReS pressure will Inot rise enough to open the PORV's, much less the PZR Safeties, which are meant to relieve a loss of
'load with the Rx still at power. PZR Sprays will open rapidly and fully id required to prevent PORV operation. The Tavg distracters will not occur as the Safeties will not open, and would reset well before lowering pressure <1005p~ig (no load temp for 54rF)
Reference Title eZR and PRT Lesson Plan Leli!ming.Objectives
[JS~T~~~IIJ~O~R~S~tate the setpoints, coincidence, blocks and permissives for automatic actuations associated with the Steam Dump ystem. (Licensed Operator & STA only)
IMaterial Required for Examination Question Source:
___.............____.lQuestlon Modification Method:
Page 61 of 90
Of the following, choose the choice which contains ONLY actions that automatically occur on a Unit 2 Main Turbine trip from 100% power, with NO operator action.
I.
Running EHC pumps trip ill.
500KV breakers 1-9 and 9-10 open.
III.
Emergency Bearing Oil pumps start.
!IV. 4KV Vital buses swap power supplies.
Iv. 4KV Group buses swap power supplies
~I. Main Generator Exciter Field Br_e~~E:lr_gpel"l~'_
EI 01, ~I'-------
~
E111, IV, VI.
E1~vl.
@J (([JY~V.
,-;;:::=:::;::='::;
IAnswer Irc-IIExam Level IlL] ICognitive Level IlMifDi~__IIFacility: Iisalem 1&2 IIExam Date: I 12/3/20121
~ ~t Systems IIRO Group I [2] ISRO Group I Q045000A311
- Q45______ ] IMain Turbine Generator System I I Record Number.11 521
==========================:====~====~~~
~'Ability to monitor automatic operations of the lVIain Turbine Generator System including:
~
I6i 11 I l§enerator trip 1[2.6*1*[2.9*1 IExplanation of li55.41.b( 4) Running EHC pumps do not auto stop, but plausible because F-32 DEHC trip occurs on IAnswer iturbine trip. 1-9 and 9-10 are the Unit 2 Main Generator output breakers, and they open automatically on every turbine trip. Emergency bearing oil pumps do not start but plausible because aux bearing oil pump will start. 4 KV group buses are powered from APT when Main Generator is operating, an automatically swap to Station Power Transformers powered from off site power upon when the output breakers open.
4KV vital bus swap does not occur as vital buses are powered from off site source. Exciter Field breaker trips upon a Main Turbine trip Reference Title IOverhead Annunciators Windows F,G,H I
====================================================~
IGenerator Voltage regulator Exciter Fi_e:=ld:=B=r=e=a=k=e=r=========
_. __._____________________.__ J
[
_n - - - - - - - - - - - - - - - - -1 Learning Objectives I MNTURBE006 LOR NCT Outline the interlocks associated with the following Main Turbine System components: (Licensed Operator & Non licensed Operator only)
Tuming Gear Turbine Drain Valves LgXCTR2EO~ Describe the interlocks associated with the following Unit 2 Main Generator Exciter and 25 Kv Systems components:
A.
Exciter Field Breaker B.
Key interlocks associated with the Power Rectifiers C.
Main Transformer cooling fans D.
Auxiliary Transformer cooling fans L
~IM~a=te=r~ia=I~Re~q=u=ir=e=d~w~r~E=xa=m=i=n=at=io=n====~I=I================~~~==~~~==~~~~~==============~==~
IQuestion Source: I [New -
llQuestion Modification Method: Ii I
IQuestion Source Comments:
II I
L-________________________________________________________________~
I Friday, September 14, 201212:07:12 PM I
I Page 62 of 9i]
the following conditions:
Unit 1 is operating at 45% steady state power.
All Heater Drain Pumps are O/S.
All Steam Flows and Feed Flows are 40% and stable Ni's indicate 45% on each channel and stable.
RCS TavglTref deviation is O.O°F.
- I'W.hich of t.h.e..f...OlIowing is CLOSEST to. the p..og.r... m..med value of SG Feed Delta-P lAW
...r..
.. a SO.CN-OOQ~, Steam Generator Feed Pump Operation?
[!J 150.-E~~d._.. ___
Ell60 ps_id_.__...__~...___...___... ___...____..___...__._...___...___...__....._....__...._
~~E~i~:
@] 50
!Exam Level! [R--.~] !Cognitive Levell iA2Qlica!i91"L
__I!Facility: 1~1&_~=:--IIExam Date: 1...__-
1~/~/20~?1 ISRO Group I
~QQ.Q&LQz::::J
~:;==::;..:;iF.;;;;_!-l~
for ICS
...E:!~<! PUI1'lE!>peed,J.n<;I':Jdin9_n()!"..mal control s~e~
- 55.41 :1;(4) SG Feed DIP (delta between feed---'---pr--'.e=s=su"*r*-e-a--'-nd=S=G**** pressure) is controlledbyadjusting SGFP
'--____... speed, and is programmed based on total % Steam Flow. Actual SGFP speed in rpm is only a result, not a controlled parameter. The KIA intent is met by asking how the feed pressure DIP is controlled, which itself controls SGFP speed. 50 is the minimum DIP from 0-15%. 150 psig is the 100% DIP. 60 psid is if l
~.9ida_t~_us~dli.~ear _scaLE:! frol1'l 0~-1 00% steam fI()'.'\\f:___
ReferenceTitle Learning Obiiecltives LOR Identify and describe the Control Room controls, indications, and alarms associated with the Condensate and Feedwater System, including:
The Control Room location of Condensate and Feedwater System control bezels and indications. (Licensed Operator & STA only)
The function of each Condensate and Feedwater System Control Room control and indication. (Licensed Operator & STA only)
The effect each Condensate and Feedwater System control has upon Condensate and Feedwater System components and operation. (Licensed Operator & STA only)
The plant conditions or permissives required for Condensate and Feedwater System Control Room controls to perform th~ir.
intended function. (Licensed Operator & STA only)
L.2t'e setpoinlsassociatedwith the COfldensate and Feedwater System control room alarms. (Licensed Operator & STA only)
IMaterlal Required for Examination Question Source:
IQuestlon Modification Method:
I
Which of the following identifies how over-cooling of the ReS is prevented on an uncomplicated manual Rx trip from 1 00% p<:>"Y~r? _
ao,rh.,'<:>t':,r Isolation.
[ts:LJ !Knowledge of the physical connections and/or cause=effeCt relationships between Main Feedwater System I
!~nc:lthefollowing:
.___J
-K1-.o-5-"-'i1 [B'cs 1!3.1*1 55.41.b(7)Feedwater interlock actuates when 3/4 RCS Tavgs <554°F and at least one
'---___---' associated bypass breaker open. This shuts the BF19's and BF40 Feed Reg Valves. Feedwater Isolatlcln loccurs when 2/3 SG NR levels on 1/4 SG's reaches 67% OR on a SI signal, and shuts BF19s, 40s, 13s,
'and trips the SGFPs. On an uncomplicated Rx trip this will not occur. P-10 is 3/4 PRNls <10% power, blocks the low power Rx trips. P-12 is 3/4 RCS Tavgs <543°F, and will shut the Steam Dump valves.
an uncomplicated trip, steam dumps will modulate to control Tavg at 54r, and temp will not reach Reference Title Reactor Tripl3:~~ponse Learning Objectives
************r----~~~--~---c.--.-.......----
CN&FDWE006 LOR NCT Outline the interlocks associated with the following Condensate and Feedwater System components:
Condensate Polishing System Inlet and Bypass Valves Steam Generator Feedwater Pump Warmup and Suction Isolation Valves Conden~ate Pump Discharge and Condensate Polisher~etum Chemical Injection Valves IMaterial Required for Examination Question Source:
Friday,
[
Page 64 of 90 _ I
.Given the following conditions:
1 1-Salem Unit 2 is operating at 100% power.
21 AFW pump is CIT for pump oil bubbler repair.
A 400 gpm tube rupture occurs on 23 SG.
The Rx is tripped and a SI initiated successfully.
2A 4KV vital bus locks out on Bus Differential.
'WhiCh of the following describes how 23 AFW pump should be utilized lAW 2-EOP-SGTR-1, I
,StE:)(;Im Generator Tu~e Ruptu~E:)?
El 23 AFW pump speed to minimum and trip 23 AFW pump regardless of MDAFW pump**
to terminate the unmonitored radioactive release from its steam exhaust.
~ ~~ ~~AJ:,p~u~t~= ~~~~~i~~~~ :~~t~3~~~Do~t8rtUnti~~45... 1 EJ running 23 AFW pump since 22 AFW pump will only be supplying feed to a single
@] ~inue rUTll"ling 23 AFW P':!I'nP because it i~the only sourcE:) of feedflowto!~e SGs.
mm
]
___IIExam Date: 1_
12/3/201~
ISROGroup I
[6100QI<103 of the physical connections and/or cause-effect relationships between Auxiliary /
'-.;""t""rn and the Tnlln'Allinn' steam !':.\\/!':.:IAm
.b{4,1 0,13) 23 Turbine Driven AFW pump is supplied from 21 and 23 Main steam lines. Each has a tap upstream of the MSIV. Each line has its own isolation valve (21 MS45 and 23MS45) its own check valve before the 2 lines combine into one line which feeds the TDAFW pump. During
!a SGTR, the TDAFW pump remians in service ONLY if it is the SOLE source of feed flow to the SG's. In the stem, 22 AFW pump will be running, since it is powered off of 2B 4KV vital bus, and it has received auto start signal on SG level. A is incorrect but plausible because there IS an unmonitored rlease from I;AFW pump since it exhausts directly to atmosphere, but is only secured if there is another source of feed. B is correct per Steps 4.4,4.5, and 4.7 of SGTR-1, which state to lower speed and trip if it is not thei sale source of feed, and to not restart until 23MS45 has been shut. C is incorrect because a single SG ein9...fed is sufficient ~.or...heat Sin.k status a
.. il..I.
... run.
.. ct..
.. nd.. 23 AFW pump w not continue to.be D is incorre because 22 AFW pump is powered from 2B 4KV vital bus and it remains energized. 21 AFW pump is powt:lred from 2A 4KV \\lilal bus which wasl()st.
~
J Reference Title 1
~(;lm Generator Tube Rupture 2 Main Steam Learning Objectives A. Determine a discrete path through the EOP.
S.
Detennine an transition out of the EOP IMaterial ~equired for Examination I'!Q_ue_s~ti~on..s!."our~==:::=I-!=~=,-;=~cc::::::-:=...::::cc::::::-:cc::::::-:cc::::::-:=~-'-'====' IQuestion Modification Method:
I Friday, September 14,201212:07:12 PM Page 65 of 90
Friday, September 14. 201212:07:12 PM I
l Given the following conditions:
-~
i-Unit 2 has tripped from 100% power from a faulty SSPS relay on the unaffected train during SSPS testing.
No Safety Injection has occurred or is required.
,Which oft~e following describes the ~ffect if 21 AFW pumpfails to start?
~ !Operator action to throttle the 21=24AF11, S/G LEVEL CONTROL VAL VES,will be directed to prevent overfeeding the SGs, since 23 AFW pump speed will NOT be lowered to minimum
~peed unless BOTHAFW pump~are runnil"!.9 in EOP-II3IP-2, Rx Trip Resp(?l'1s_e_.__
~ IOperator action to throttle the 21-24AF11, S/G LEVEL CONTROL VALVES, willbe-d-ir-e-ct-e-d-to--'
iprevent overfeeding the SGs, since 23 AFW pump will NOT be secured unless BOTH AFW pump~are running in EOP=IRIF>:1 Rx Trip or Safety lr1j~ction.
........... _....... ______~
~
of the RCS will occur during the first 5 minutes following the trip and a MSLI will to l'irnit the excessive cooldown.
nO\\lvleejae of the effect that a loss or malfunction of the Auxiliary I Emergency Feedwater System will 21 MDAFW pump supplies AFW flow to 23 and 24 SG. 23 TDAFW pp supplies all 4 SGs.
a Rx trip, operators will transition to TRI P-2 after the Immediate Actions of TRIP-1 are I'\\Qr'tf'lrrl'\\QI'1 and a SI is not required. After stopping the SGFPs in TRIP-2, step 3, 23 AFW pp speed is to minimum or 22E4 Ibm/hr. Since there would be no flow to 23 and 24 SGs if speed was Ilf'I\\A/Qr~,1'1 to minimum, operators will throttle the AF11s to balance flow to each of the SGs and maintain and pressures approximate. C and D are incorrect because overfeeding will NOT occur since
",,.,,::>r<>lrnrc are directed to lower AFW flow. B is incorrect because operators will have transitioned to TRIP Reference Title
[l3:aactor Trip Response.
Learning Objectives'
---:-=-:-c.,-::-::-=-cc:---:-:..........--........-'--;-------c_____--:
NCT Given plant conditions. relate the Auxiliary Feedwater System with the following:
Steam Generators Main Feedwater System Main Steam System Reactor Coolant System Condensate System Demineralized Water System Service Water System Fresh Water Fire Protection System Room Coolers Rl'If,,,.,,,,.rrl,, t-n. u:nmp,nt Controllers IMaterial Required for Examination.
[~.ri(j<3Y. September 14.201212:g~:1?~
Question Source:
ion Modification Method:
IDirect From Source..
IVision 0134732. Used on Salem 07-01 NRC RO exam.(052, 3 exams ago)
'------------~
~-~.......-- - - - - - - - -
-~'
14,
IWhich of the following describes how control room instrumentation will be affected if the 1 C I
- Instrument Bus (VIB) Inverter were to experience a latched transfer?
Instrl:l~rltati~rlJ::>5)wered from 1 C VIB...
I!J the transfer since it occurs in milli-seconds.
[]
INOPERABLE until the VIB inverter is restored to its n~f11alpower El ffi~~~§EI~JOw during the trarisier (1-2seconds)~li~':Arn to fulifunction~1 status.
@]
10 VIB will momentarily be lost during the transfer since their inverters are powered from same 230 VAC source.
_IIExam Date: I._ 52/3/?~~~
P620QQAIQLi r[R-ec-o-r-cd-N-u-m-be-r-'-'I
-=-=-=.~r;"====
....=============...=======
...==:=-====..':'..==-=-='~=-=
....======....:::====~....
A 1.-.JAbility to predict and/or monitor changes in parameters associated with operating the A.C. Electrical Pistributicm contr:~s_!nclu_cjing:__ __
~W:L] ((ffect o~~!nentati()n-an~ols of switc;hing power suppTie~
J [2.5 l?,§j
'155.41.b(7) B is incorrect because the VIB, and the instrumentation powered from it, remain OPERABLE L...;.;.,;,.;;;.;:....-_---I as long as the inverter is powering the Vital Bus. (P&L 3.5) The transfer of a VIB inverter takes 2/3 of 1 cycle, which is 11.1 milli seconds, which will not give enough time for the lights to respond on the L~nstrume~!ation. D i~ing()rrect be~l:\\use ir1dicl:\\tion w()n'~~e lost,ancj1D V~B i~power~d!r()rT"I....1.1? bus.
Learning Ob.ieCitivel$
the power supply to the following 115VAC components:
A. B. C. and D Vital Bus Inverters Essential Control Power System Emergency Lighting Power System Unit 2 RMS Power System SPDS Power System Control Rod Control and Indication Power Systems Security Power System Telecommunications Pov.rer Sy~tem Question Source:
Source Spr,t"lY,hAr 14, 2012
the one component below which does NOT receive power from the Unit 1 Vital 125 Instrument Bus Inverter.
H 4K'{§roup Bus controleower.
~________
@J !Supervisc:>ry Control and Data Acquisition (SCADA) System.
ICognitive Level '[Memory
.IFacility: I Salem 1 & 2
- 'Exam Date: I 12/3/20121
.~======='-==='-'========='---':':::=~:;::;==~ ::;-1;:;;:S;:RO;;:;Gr=ou=p:::;I-;:=~-
I063000K201 IRecordNurii§Q to the t.....II.....'A1'nn
.b(4) The SCADA system is powered from the Circ Water 125 VDC system, not the vital 125VDC All the distracters are powered from the 125 VDC vital buses. 11 Essentail controls inverter rt:>rlt:>\\1"", power from 1 C 125 VDC bus. 10 Vital Instrument Bus Inverter recieves power from 1 B 125 bus. The 1 H 4KV Group Bus recieves its control power from 1A 125 VDC bus (reg) and 25 VDC One Line I-IO,I"Tr,lr"" S\\I'tAme:: Lesson Plan Learning Objectives and describe the local controls and indications associated with the DC Electrical System, including:
The location of DC Electrical System local controls and indications. (Licensed Operator & Non-licensed Operator only)
The function of DC Electrical System local controls and indications. (Licensed Operator & Non-licensed Operator only)
The plant and conditions or permissives required for DC Electrical System local controls to perform their intended function.
(Licensed Operator & Non-licensed Operator only)
The setpoints associated with the DC Electrical Systems local alarmS.
IMaterial Required for Examination Question Source:
IQuestion Modification Method:
l,==========~
--1 L--.;..____-! 1_ __
[ Friday, September 14,?012 12:07:12 PM I
Page 70 of 90
of the following describes the Design Basis for the capacity of the EDG Diesel Fuel Oil Storage Tanks (DFOSTs) following a LOCA coincident with a LOOP?
El TWO.
'---<--'--~<<<<<
[] EACH; THREE.
TWO ED volume of BOTH; THREE.
I!Facility: !ISalem 1 &2 IIExam Date: !
12/3/20121
~~~:==~~====~==~========~==~~==~*~I~SR~O~G~ro=u=p~!==~~
~~~lQ3
ILIR~ec~o=rd==N=um=b~e=r~I~==--~
~======.===<<< -~----::---=-=:=---::-----::--=::::====-==.
nOllvle(:1ae of the physical connections and/or cause-effect relationships between Emergency Diesel
';Ar'Ar,::nnr<:: and the TnllnlAl,nn Salem FSAR states....The combined volume of both 30,000 gallon fuel oil storage tanks L...;;.;.;;";';";~_--' lI"'n,nt""nC! sufficient fuel oil at the Technical Specification minimum volume to supply two diesel generators, at the most limiting accident mitigation profile for LOCA with loss of offsite power, for 4-5 days." <<<<<<<<<<<<<_~..~.
Learning Objectives Describe the function of the following components and how Diesel Generators:
Lubricating Oil System Jacket Cooling Water System Fuel Oil System Starting Air System Turbo-charger Turbo Boost Air System Exciter Q Regulator Iv..n1nr/~'ru>.>rl Control the design bases of the Emergency Diesel Generators.
& STAonly)
!Material Required for Examination Question Source:
IQuestionModification Method.:
Page 71 of 90
Ilf a leak were to develop on 21 RHR pump room cooler, which of the following tanks would show
'rising level because of the leak?
ElIAuxil~ary Buildil1_g_~l.JrrlpTarlk.
u_
]
[] l~r1___se'"Y~~_\\lVa_ste Hold Up Tank.
~ ~n-se~~_eCVCS Hold Up Tan~.
____1
@J ~dry, Hot S_hower, and Chemical Drain Tank.
IAnswer I ~] lexam Level I ~IICognitive Level IIApplication I I Facility: Iisalem 1& 2 I I exam Date: I 12/3/20121
~ ~t Systems IIRO Group I [2] ISRO Group I [2]
- 068000K107 1
1068
] ILiquid Radwaste System 11Record Number II 60 IiS1:=J !Knowledge of the physical connections and/or cause-effect relationships between Liquid Radwaste System lan_d the following:...........
r-IK-1-.0-7--".I[§:ources of liquid'--w-'-a-"'.s=te=s=f=or-=L'-"R=S-=========================;Ir=lli=='2.7 [}]
IExplanationof I55.41.b(13) RHR pump rooms each have a sump for recieveing draains and leakage in the room. Each IAnswer RHR pump room sump is pumped to the inservice Waste Hold up Tank. A is incorrect because the Aux Building sump tank collects floor drains from locations above it, and it is located on 64'.
C is incorrect because the CVCS HUT system recieves influent which can be processed and recovered as CVCS quaility water, not floor drains. D is incorrect because Laundry and Hot Shower drain collection points are provided for the contaminated laundry, showers, and sink utilized for protective clothing and personnel decontamination.
Reference Title IFloor Drains - Contaminated I
====================================================~
IWaste Disposal Liquid I
I
_________________I Learning Objectives
[WASLlQE004 NCT Describe the function of the following components and how their normal and abnormal operation affects the Radioactive Liquid Waste System:
Waste Evaporator Feed Pump Waste Monitor Holdup Tank Pump Auxiliary Building Sump Tank Pumps Chemical Drain Tank and Laundry and Hot Shower Tank Pumps Reactor Coolant Drain Tank Pumps Containment Sump Pumps Reactor Sump Pump RHR Sump Pumps FHB Sump Pump I
I I
r..___**
'---T
===:=:;-;::==============================-~J IMaterial Required for Examination II Question Source:
IIQuestion Modification Method:
II L...--____-----! L______
I Friday, September 14, 2012 12:07:12 PM I
I Page 72 of 90 I
the following conditions:
Salem Unit 1 is operating at 100% power, with no equipment out of service.
Salem Unit 2 is in MODE 5, during a return from a refueling outage.
Unit 2 is performing a normal Liquid Release from 21 eves Monitor Tank via 22 SW header to Unit 1 ew lAW S2.0P-SO.WL-0001, Release of Radioactive Liquid Waste from 21 CVCS monitor Tank.
The Radwaste Overboard Discharge Flow Recorder 2FR-1064 is OPERABLE.
The 2R18 Liquid Waste Disposal and the 2R41D Plant Vent Release Rate monitors are OPERABLE.
After commencing the release, the field operator is recording the Initial Release Data.
When recording the 2R18 reading, it reads 10 E6 cps.
1-The 2R18 red alarm light is lit on the 104 panel.
The 2WL51 Liquid Release Stop Valve indicates OPEN in the control room.
of the following describes what these indications mean, and how the operating crew should 2R18 reading is an expected short duration spike for a release of a monitor tank filled with activities liquid. The 2WL51 closure time delay must time out before the 2WL51 be expected to shut. Continue with the liquid release and ensure 2R18 reading returns to normal.
@]Withthe 2FR-1064 and 2R41D OPERABLE, block the 2R18 input to the 2WL51 on 2RP1 prior
~!~~time delay!iming out and continuE?!bE? liquicJr~leC1~e_.~~~~~~_
ElI.The2WL51 should have immediately a.uto.matically closed o.n. the 2R18 high radiaiio.*.n. alarm. J
!The f'.J_<29J!lJb~_~9JJ29m should sh.uJ?~L51.
@]
should have immediately automatically closed on t EO should shut 2WL51 he 2R18 1213/201~
55.41.b(13) 6.82E5 cps is the ALARM setpoint which will automatically shut the 2WL51 L.....;.;..;"";""__.....I 0028). The current reading is above the setpoint at which the 2WL51 should have automatically shut, the NCO should shut the valve remotely. The red alarm light for high radiation indicates the 2WL51 have shut, memorization of alarm setpoint is not necessary. Additionally, S2.0P-SO.WL-0001 for releasing the tank, Step 5.5.9, says if the 2R18 alarms, then the NEO is to inform the NCO to shut the 12WL51. Th..e
... re is no provision for closing the valve locally. There is no time delay for 2WL51 closure, nor is there.§l!1.y provision for releasing a ho~t<:ll'l~ or is there an expected spike.
Radioactive Liquid Waste CVCS monitor Tank.
Process Radiation Monitor LiOiumno Objectives Outline the interlocks associated with the following Radiation Monitoring System components:
61
R1 S, Control Room Inlet Duct Monitor R5, FHS 0 SFP Area Radiation Monitor R7, In-core Seal Table Area Radiation Monitor R9, FHS 0 New Fuel Storage Area Radiation Monitor R10A, Personnel Hatch 0 Containment Elev 100,lE Area Monitor RiDS, Personnel Hatch 0 Containment Elev 130,lE Area Monitor R11A, R12A, R12S, Containment Particulate, Noble Gas, and Iodine Monitor R13A, S, C D & E CFCU Service Water Monitors R17A and S, Component Cooling Liquid Monitor R18, liquid Waste Disposal R19A, S, C, & D, Steam Generator Slowdown liquid Monitors R32A, Fuel Handling Crane Area Radiation Monitor R36, Evaporator and Feed Pre heaters Condensate Monitor R41 D, Plant Vent Radiation Monitor PASS Room Area Radiation Monitor IMaterial Required for Examination Question Source: ~~~~f1kJ IQuestion Modification Method: I Modified
'--_..;..________--'.... I~~s:~g~~~:.6 mOdifi~d to provide~ore info in stem about ~perabieeQuipment and made answers more]
14,2012
[GiVen the following conditions:
1-Unit 2 is operating at 100% power steady state.
A field operator reports a SW leak in 2C EDG room, just upstream of 23SW39, 2C DIESEL CLG iSWVLV.
1-The RO reports Service Water pressure on both 21 and 22 headers has lowered from 112 to 101 psig and continues to lower.
IWhiCh of the following describes the expected SW system response, and how the operating crew iwill respond lAW S2.0P-AB.SW-0001, Loss of Service Water Header Pressure?
Assume SW header pressure can be restored.
IThe~~st~(j!>y SWpUrTI~~t~~Vllhen §",,~(j~~J)ressure lowers to...
I!J 195.0PSi9. Lock out 2C EDG and declare it INOPERABLE, shut21SW21 AND 22SW21, ~ n DIESEL CLG SW INLET VALVES, to isolate the leak.
~1.199.5.P.**.*.Si9. Lock out 2C EDG an
.. d. d. ecl~re it INO.PERABLE, shut 21SW21 AND 22SW.2.1:_..
.~~EL CLG SW INLET VALVE~S,~1()lsolate the leak.
~
..... __..1 EJ psig. Lock out 2C EDG and declare it INOPERABLE, isolate the leak by shutting 1SW37 AND 2C DIESEL CLG SW INLET VALVES.
Lock out 2C EDG and declare it INOPERABLE, isolate the leak by shutting AND 2C DIESEL CLG SW INLET VALVES.
~==~.::::~~========:="...=======-===~;::;==~=lFacility:l~ 1 & 2 ~~-IIExam Date: I 1213/20121 IRQ Group I ~ ISRQ Group I lQZ~02J i076
~cord Number* 1 [_~
A2~-!Ability to (a) predict the impacts of the following on the Service Water System and (b) based on those Ipredic!i2!1~,use procedures to c()~~ect, control,or mitiga!~the consequences o!!bose abnormal opera!i9n: *
- 8..2.~ ~ervice water headerp~~~su!e JW 155:41.b(7~ 10)Lock out the EDG(s) that will be affected and isolate the leak. Step 3.11 has operators
'-----'-__-I isolate the leak. EDG must be locked out to prevent starting with no SW available. The only way to isolat the leak is to isolate both supplies from both SW headers by closing both SW37's. Cannot isolate both SW21 s per AU 4, Steps 4.0 Band C because it would render ALL EDGs inoperable. OHA for SW Iheader pressure low state?8uto start for standy f?.'!'1 pumps is 95.5 psi9:
Reference Title ILoss of Service Water Header Pressure Window B the actions in I:>1'1',Clrrl",nN>
IMaterial Required for Examination Questic;m Source:
IQuestionModlfication Method; L-_______----' l~~:o~~~~~~8(~ahd~n~n~~ :~~a~~a~,~o do to isolate leak (4 choices) to what pressure auto pump will start and *l 1
1 Friday, September 14, 2012 12:07:12 PM
[Friday, September 14, 201212:07:12 PM
the following identifies normal Control Air header pressure, and the pressure at which Air will start?
ICognitive Level I ~n<'r'rI,nnJ IFacility: I ~I~~ 1 &2
! I Exam Date: I 1213/2012
~~~=~~====~~====='~======~==~~==~[J=1J~I~SR~O~G=ro=uP~I~~-
J"\\n,or<,flrmc of the Instrument Air..... \\"~TQITI 55.41.b(7) AB.CA Basis Document states... "When supplied from SA through the dryers, CA pressure
'--____..... iruns approximately 5 psig below SA pressure. CA pressure cycles along with the SA cycle. Thus, CA ipressures normally run between 95 and 105 psig.The Emergency Control Air Compressor (ECAC),
IShould1:lLJto start if CA pressure drops to 85 psig.
Reference Title Air
~~.::=:==
Operation _____
~~~s of CorltroTAir
--..--~~_____ _
~~---
Learning Objectives CONAlREOOa--r..Id~ntify anddes~ribe the Control Room controls, indications, and alarms associated with the Control Air System, including:
The Control Room location of Control Air System control bezels and indications. (Licensed Operator & STA only)
The function of each Control Air System Control Room control and indication. (Licensed Operator & STA only)
The effect each Control Air System control has upon Control Air System components and operation. (Licensed Operator &STA
- only)
, The plant conditions or penmissives required for Control Air System Control Room controls to perfonm their intended function.
(Licensed
&STA only)
The associated with the Control Air control room alarms.
IQuestion Modification Method; n Q42094 changed to normal pressure and auto start pressure, from a what starts the ECAC.,
Friday, September 14, 2012 12:07:12 PM Page 7_7_o__f__-----.J I
Given the following conditions:
Unit 2 is operating at 100% power when the following occurs:
OHA A-7, FIRE PROT FIRE.
2RP5 Fire Protection Panel, Zone 148 - Work Control Ops Ready Room lamp illuminates, as does the FIRE lamp for that row of alarms.
The audible coded fire alarm is broadcast over the plant PA system.
For these conditions, which of the following identifies:
. What these indications mean.
2RP5 Fire Protection Panel is reset when the condition has cleared.
active fire suppression system (water/C02/Halon) has activated. Reset from the Control active fire suppression system (water/C02/Halon) has activated. Reset from has activated. Reset from the Control Room.
IAnswer I [J IExam Levell IExam Date: I 12/312012
~:::::;-:~=;--=========="':::":=======~~~:::::;-;='=""';~~=:; f==~_*i
[Q86000A402 l~~cord NumooiJ nn,,,r:;OIl'~ and/or monitor in the control room:
i55.41.b(4) 1 )Early warning detectors provide the Control Room with early indication of fire, but do cause a suppression system to actuate. Early warning Smoke Detectors and Fire Detectors the plant are arranged to alarm to the Control Room by zone.a) If a detector on a zone actuates, the zone I indicating light on Panel RP5 will illuminate along with the appropriate group "FIRE" light, and the coded fire alarm assigned to the zone will be broadcast over the station PA system.
Once a zone actuates, the alarm remains illuminated until the zone is manually reset from the fire protection panels in the Relay Room.
Reference Title
!Fire Protection Systerrl Less()11 F1lan Leaming'Objectives' FIRPROEoOfTldentityand describe the Control' Roomcontrols, indications, andafarms'associated with the Fire Protection System, including:
i The Control Room location of Fire Protection System control bezels and indications. (Licensed Operator &STA only)
The function of each Fire Protection System Control Room control and indication, (Licensed Operator &STA only)
The effect each Fire Protection System control has upon Fire Protection System components and operation, (Licensed Operator & STA only)
The plant conditions or permissives required for Fire Protection System Control Room controls to perform their intended function.
Operator & STA only)
"",lrv,int" associated with the Fire Protection control room alarms.
IMaterial Required for Examination Question SOUf96:
IQuestion Modification Method:
I =.=========;
m Friday, September 14, 2012 12:07:12 PM Page 78 of 90
IGiVen the following conditions:
i-Unit 1 is operating at 100% power.
Control room operators are preparing to perform a Containment Pressure Relief lAW S1.0P-SO.CBV-0002, CONTAINMENT PRESSURE-VACUUM RELIEF SYSTEM OPERATION.
Containment radiation levels are NORMAL for 100% power operation with no failed fuel.
lAtter opening the 1VC5 and 1VC6 to initiate the pressure relief, which choice describes how the
'respective radiation monitors indication will respond?
i1R12A - Containment Gas Effluent
'I R41 B - Plant Vent Noble Gas Intermediate Range
'I R41 0 - Plant Vent Noble Gas Release Rate
[!J 11 R12A constant; 1 R41 B constant; 1 R41 0 rises.
~ 11 R12A rises; 1 R41 B constant; 1 R41 0 constant.
@J 11R12A constant; 1R41B rises; 1R41D constant.
@] §I2A rises; 1 R41 B rises; 1 R41 0 rises.
IAnswer I@:] IExam Level I B::=:] ICognitive Level IIApplication JIFacility: IISalem 1& 2 I I Exam Date: I 12/3/20121
~ ~t Systems IIRO Group I [JJ ISRO Group I [JJ
'103000A409 103
] IContainment System IIRecord Number i I 651 M::::J t.~ility_to m?nLJallyoperate and/or monitor in the control room:
J M.09] If"ontainment vacuum system I[Ijj @El Explanation of 1 55.41.b( 11) 1 R 12A is sampling containment atmosphere, so it will NOT rise when the pressure relief is Answer started. 1R41 B is an intermediate range monitor that normally does not have sample flow through it. It's sample flow will start when the 10 range 1 R41 A monitor nears its high end of monitoring range. It's
,indication will not change during a pressure relief with NORMAL containment radiation levels. The R41 0 Iprovides the gaseous effluent release rate (uCi/sec) by combining (product of) the on-range R41A throug R41C with plant vent flow (cc/sec). It will rise when the pressure relief is initiated, and also provides automatic termination of release on hi gaseous effluent.
Reference Title IContainment Pressure-Vacuum Relief System Operation I
Learning Objectives
~--~~----------------
CONTMTE007 Identify and describe the Control Room controls, indications, and alarms associated with the Containment and Containment Support Systems, including:
The Control Room location of Containment and Containment Support Systems control bezels and indications. (Licensed Operator &STA only)
The function of each Containment and Containment Support Systems Control Room control and indication. (Licensed Operator
& STAonly)
The effect each Containment and Containment Support Systems control has upon Containment and Containment Support Systems components and operation_ (Licensed Operator & STA only)
The plant conditions or permissives required for Containment and Containment Support Systems Control Room controls to perform their intended function_ (Licensed Operator & STA only)
The setpoints associated with the Containment and Containment Support Systems control room alarms. (Licensed Operator &
STA.c>.'2'y.) ________
,------~---
IMaterial Required for Examination II
~----------~~==============~------------~
I Friday, SElptember 14, 2012 12:07:12 PM I Page 79 of 90 I
IQuestion Source:
Exam Bank IQuestion Modification Method:
IQuestion Source Comments:
Ilvision Q75012
the following conditions:
After taking the NRC Initial License Exam in May, you are assigned to an operating crew.
The NRC issues you your Reactor Operator License on June 25th.
The LORT Annual Requalification Exam (Segment 3) starts on June 26th.
lAW TQ-AA-106, Licensed Operator Requal Training Program, which of the following describes if you are/are not required to participate in the Licensed Operator Requalification Training Program for Segrl1ent ~?
~ ~~RE req--u---ir-~e-d----to tCike the p.,nnual R~qual exam ANt:? Segment 3 trairJ_in-=g_.___________~
[]
to take the Annual exam nor attend are NOT required to take the Annual Requal Exam but must still attend Segment 3 ARE required to take the Annual Requal exam but do not have to attend Segment 3 IAnswer I IExam Level I~~] ICognitive Level I[MemQIY~~_~ IIFacility: I~~rl1I~~2
-.~.~ IExam Date: 1~~213/20131 ITler:J ~11~ricKn()~~(:ige ar'l(rAbilities~--________
ISRO Group 11940018104
-,==--'--:'~I===,-
[rfeCOrdNuill§] i 66
!GENERIC_~
.b(10) "Trainees who obtain a license in a given year may be exempted from taking the Annual
..,_c..*.,....,t'1 Test that is scheduled to be administered during the first requalification training cycle in which operator participates. However, operators who complete one or more training cycles before the scheduled Annual Operating Test shall I~~ke the test to ensure that they do not exceed the allowed testing intervals." In addition, while the
~~~~~~;,n~;~~~~~~e~~hgema;n~u3a!r~~~~~~~ived, thereqUirem_nt t~_atten~ requal isnotwaived, and the Reference Title
_0.
Lk:ensed Operator Requal Training Program IMaterial Required for Examination
.....~
Question Source~
IQuestion Modification Method:
1...-____-'-__---' '----__----------
-_----=--=--- --------~I
[
Friday,September~4, 201212:07:12 PIVl'
Thermal Barrier is to...,...r.1rOl"t the Rep...
due to excessive heat
~==:-;::~.........~======='L:'~~=========~==,~~~...;:...:-=-=;_....... _,IFaciiity: 1~~1& 2 IRQ Group 1=-_] ISRQ Group 11_ 11 and function of and controls.
. b.(3) The RCP thermal barrier prevents hot reactor coolant from flowing up the shaft to the radial L...-';""""__.......J bearing and RCP seal package when normal seal injection flow is lost. The normal flow of cool seal injection water past the raidial barrier into the RCS normally performs that function. In the event of a loss lof seal injection flow, reactor coolant would flow up across the tubes of the heat exchanger(cooled by Iccw) afl.!!irl!o the Pljmp to provide the coolant and lubricant for the radial bearing and seals Reference Title Lesson Plan Describe the function of the following components and how their nonmal Pump:
Turning Vane Diffuser Diffuser Adapter Thenmal Barrier and Heat Exchanger Pump Radial Bearing Controlled Leakage Seal Assembly Lower Motor Radial Bearing Upper Motor Radial Bearing Flywheel Anti-Reverse Rotation Device Oil Lift Pump Motor Heaters IMaterial Required for Examination*
Question* Source:
IQuestionModification Method:
'---'---'---_____---1 IVision Q43989 rr1ade pschometrically balanced_. _______
14,2012
of the following conditions will REQUIRE the suspension of fuel movement in the Unit 2
. reports BOTH 100' elevation containment airl~C::~~§l!~()pen.
~Contai,"!ment Radiation Monitor 2R12Afaiis causin9~()~rn~nt Venti~riJ~!9tion~!9nal. ]
@] I.BOT.*.H..... u.nit 2 Control Room Ventilation intake duct radiation monitors 2R 1B-1 and 1 R,iB.**.*-.2. a.r....e...J declared inoperable when intake damper CAA40 shuts, and the PO depresses Fire Outside Control Room.
ICognitive Level I~~ation llExam Date: I 1213iio121 Abilities
~~==~::;-;;;:~==::;-::==;--...J
- 1940o_1G14_0-'_,
IRecord Number IL..___.......:.l administrative 55.41.b(11) D is correct because operation in the recirculation mode requires suspension of fuel
'--___-' movement. (SO.CAV P&L 3.6.3 When aligned to FIRE OUTSIDE CONTROL AREA (Recirculation iMode},Core Alterations and movement of irradiated fuel is NOT permitted (TIS Bases 3/4.7.6)}.
Cis lincorrect because Containment Radiation monitors are not required to be operable for Mode 6 or Fuel Movement or Core Alts per Tech Specs. B is incorrect because the airlock doors are only required to be CAPABLE of being shut, and can be open (S2.0P-ST.CAN-0007, page 8) A is incorrect because the COLR limit for boron concentration is 2139 Reference Title Closure Learning Objectives Discuss the procedural requirements associated with the Refueling System, including an explanation of major precaution and lirTlitations in the Refueling System procedures. (Licensed Operator & Non-licensed Operator only)
Question Source:
IQuestion ModificatiooMethod:
I ~~om SOll~
I Friday, September 14, 2012 12:07:12 PM I 1
the following conditions:
Unit 2 is operating at 100% power.
Excess Letdown is in service due to a problem with the control circuit for 2CV35, VCT 3 WAY INLET V.
I&C troubleshooting is in progress on 2CV35.
The RO reports that 2CV35 has just swapped to the Flow to HUT position.
The RO also reports that during the pre-job brief for the 2CV35 troubleshooting, it was stated that 2CV35 actual position would NOT be affected during the troubleshooting.
~~fthe following identifies the FIRST action the crew ~hould tak~?
[!J S2.0P-AB.CVC-0001, Loss of Charging to address the unanticipated CVCS system the WCC to initiate a tagout for 2CV35 since the troubleshooting needs to have the deactivated.
El !Contact the I&C Supervisor and stop work on 2CV35 based on being outside of Procedures, iPar§l"Tl~ters ()~J:l"9ge!3!3~~()0f>_§11~W HU-AA-101~rilJ!l!an Error Prevenji()~__.
@J the RO place 2CV35 in the Auto position to maintain status control since the Component Normal and Off Normal does not reflect the valves current
.1 ISRO Group I[))
",~r,,'l\\ltlnn activities.
55.41.b(10) Excess letdown does not flow through 2CV35, so its movement will not affect RCS letdown,
~;;...;.;.;;.;;......;._-,!and AB.CVC-1 will not address the valve movement. While a tagout may need to be initiated, it would not!
happen without stopping the job in progress and finding out in more detail from I&C how the valve could be affected and what the blocking points needed to be. OOPS requires stopping the job because they system realignment was stated to NOT going to occur during troublshooting. Status control does not movement of components just to align with the Off Normal, the Off Normal position would be in SAP.
IQuestion Modification Method:
iThe Unit 2 control room reciev9s a call from theRad Waste Operator, who states that an iSOlate-dl IGas Decay Tank in hold-up has lowered in pressure from 90 psig to 40 psig over the last 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and continues to lower slowly.
I ii:i~~~~~hde~~~O~i~~n~~~~ti~~oL~;o~~~;~mation that tank p~:ure has IO~:~d (vs instrum~:J EJ the 2R41 D trend reClQ~1"l2RP1. An unapp~~d relea~~r11_ClY be~~gress.
~ ~Iay the 2R416tre6~Qi!lJiir1?BI::!:~~~~~~~~~~~rlP~e_ss_._
E1 IAnswer Iii] IExam Levell
.~lc~O~g~ni~tiv~e~L~e~v~el~I,-~~~~~~=:::;-;~IFaCillty: l[saiem1&~-=, IExam Date: I 12/3/201?J
~ ~ericKn2~E?~ge and Abilities
--~ IRO Group I 0 OJ 1~QJ~244. J IGENERI~
i~ecord N~miiJ=e=rL*::::::::::~_....J to interpret control room indications to verify the status and operation of a system, and how actions and directives affect and conditions.
- 1) R41 (plant vent) monitors in control room have trend function which can display historical release is in progress its being monitored, but is unapproved. Area Monitors do not have trend but some of them are trended on P-250 and control room.
I Reference Title lF3adiation Monitoring System Oper(ijtion LeamingObjectives I RMSOOOE007 _1 Identify and describe the Control Room controls, indications, and alarms associated with the Radiation Monitoring System, including:
The Control Room location of Radiation Monitoring System control bezels and indications. (Licensed Operator &STA only)
The function of each Radiation Monitoring System Control Room control and indication. (Licensed Operator & STA only)
The effect each Radiation Monitoring System control has upon Radiation Monitoring System components and operation.
(Licensed Operator & STA only)
The plant conditions or permissives required for Radiation Monitoring System Control Room controls to perform their intended function.
_._. ___.. =
..._.-::---~-===:.=-::---~==.=-::---~--==:.=::==~-c==.====--==..=-::---~
IMaterial Required for Examination QuestionSoul'e$:
IQuestion Modification Method:
14,2012
WhiCh of the following describes when rising radiation levels on 2R19A, STM GEN BLOWDOWN j IRAD MONITOR, will automatically close the 21GB4, SG BID OUTLET ISOL VALVE, and why?
2R19A in will close the 21 GB4 EJ to minimize the spread of contamination from a Steam Generator Tube Rupture on 21 Steam Generator to "'O'"'I"\\I'"1.rI
~Alarm.' to minimize the spread of contamination from a Steam Generator TI.Jbe Rupture~
I(SGTR) on?1 Steam Generat9r to secondary sy~tems.
mm~
mm
~
to prevent backfeeding contamination from 21 Steam Generator to any other the unaffected Steam Generators blowdown lines.
@J Alarm, to prevent backfeeding contamination from 21 Steam Generator to any other Steam Generator the unaffected Steam Generators blowdown lines.
ICognitive Level 11\\1It>mn'....'
IFacility: I.Salef!11 &? __IIEXam Date: I 12/3/20121
~~~~=i~~~~~,~~A\\:"b""ili:",tie~s~===:::::::: IRQ Group I ISRQ Group I OJ
!194DO-.lG~JD
___-_____======::::-:-:-::----:--~I. ~C~~d Number 1'--__--"
no'Vvle(jae of radiation or contamination hazards that may arise during normal, abnormal, or conditions or activities.
55.41.b(11) B is correct because the GB4 is automatically shut on the Hi Rad Alarm, whereas the 21 t......-.;.......;__--I24GB10s, 21-24185s, and 2GB50 shut on Hi Rad Warning. Isolating the blowdown path from the S/G to the condenser will prevent the spread of contamination, and also will prevent any type of release from the main condenser to atmosphere. Each S/G has its own blowdown line, so backfeeding contamination is
[l.Cltpossible through the~l2.wdown lines Reference Title IRadiatio... n Monitoring System Operation J
Describe, in general terms, the actions taken in and the bases for the actions in accordance with the Technical Bases Document IMaterial Required for Examination Question Source:
Significantly~odified modified to make a 2 and 2 vs just a why do we isolate blowdown 1
Which of the following Area Radiation Monitors is checked in 2-EOP-LOCA-1, Loss of Reactor 1~()~~VY~~~ci~.!!'i~!~gJf(3J:.()~.A.~ci~lJ!I!t~coBt~i!!I'll~~~~c::~i~g~ __....__.~~~.
§J Electrical Penetration.
Fuel Han-.91Ll'!g Building-Spent Fuel.
.b(11) LOCA-1, step 16, checks for radiation outside containment by looking at 2R4, charging pump 2R41 D plant vent process, 2R34, 1 R3 Radio Chem lab area, 1 R6A Sampling room, 1 R20B countin Reference Title flOSs of Reactor Coolant 11..\\.1:::__
[§2.0P-SO--:-RM=0001 Learning Objectives*',
Identify and describe the local controls, indications, and alarms associated with the Radiation Monitoring The location of Radiation Monitoring System local controls and indications. (Licensed Operator & Non-licensed The function of Radiation Monitoring System local controls and indications. (Licensed Operator & Non-licensed nn<,.~t,~r The plant conditions or permissives for Radiation Monitoring System local controls to perform their
& Non-licensed,,~~*.~.
Accident, and describe how the actions of 2-EOP-LOCA-1 Identify possible radioactivity release paths for a LOCA OUTSIDE CONTAINMENT IMaterial Required for Examination Question Source:
Iquestion Modification Method:
Modified iviSion Q125697, replaced distracter with 2R1 OA (used to assist in determining LOCA is ocurring IN cont, not
'----'---------...1.outside.
.... ___....___ j
~--...~-.----
- Friday, September 14, 201212:07:12 PM Page 87 of 90
1 IGiVen the following:
The unit has been tripped and Safety Injection initiated due to a LOCA.
1-The STA observes a PURPLE path displayed by SPDS for the CORE COOLING Status Tree, Iwith no other RED or PURPLE paths on SPDS.
, The SM M is blinking dashes on both channels.
SPDS is displaying question marks for all CET's.
1-Plant Computer CET indication shows all CET's between 50-60°F.
Local CET Display is unavailable.
'Which of the following identifies how RCS saturation temperature will be determined lAW
.CFST-1, Critical Safety Function Status Trees?
lyyideFiarlg~RCS Thot and...
I!J lBCE) J:>re~sureJ~J::103~1-405~be':J~c!_in{)<:>~I1Gtio!'l~h Steam Tables.
[] ~:pressure (~I_-~rE!~~'.VJIU~~~~d in conj':Jnction with CFST~ubcoolin~!~_J E]
pressure channels (PI-455A, 456, 457 or 474A) will be used in conjunction with Steam
@][p.ZR press...u. r.e channels (PI-455A, 456, 457 or 474A) will be used in conjunction with.CFST r-::---~-..ubcooling Tables.
. -;::::::==~-=::::;
IExam LevelllR IICognitive Level IIMemo~
m~-=IExam Date: 1..12t~!20121
.~ ;;;IRo';:;:Gr=ou=p:::;-1~=;-;:;:;~===;~==;--
[i~4001G400 rRec()rd Number : ~__...........J to instrumentation.
155.41.b(7.10) EQP-CFST Basis Document, page 2 of 18, "The SMM should be used to determine RCS
'--___.....J subcooling. If the SMM is inoperable. then calculate and log RCS subcooling on Table D. The value of T Sat is obtained by using Table A for Normal Containment or Table B for Adverse Containment." Table D
~otnotemstate~:.. "*RS;S t~rnper<!ture~Use fET'!;)(WR_Thotl3TD's if Ci:;T's a~e notavailable)."_
Reference Title Function Status Trees Learning Obiectilves IMaterial Required for Examination Question Source:
E<:c~~El<<lm Ba~ ___......__
.. ~ IQu~tion Modification Method:
Modified
!Vision Q81104. U.sed slem and. expanded an.swer reqUired. f.rom generic "what to do n.ext" to howto.determine
'--------'------' il3CS saturation temp.
14,2012 Page 88 of 90
I I
During performance of Emergency Operating Procedures some Cautions apply Iremainder of the EOP in progress.
Which of the following describes how these Continuous Cautions are identified?
~
box to the affected.
~
box pri()i to steEJ ofthe procedlJre.
..~-.-..~..~
E] ll2~le-borderedb()xprior to theSte:ps affected.
@]
bordered box er!e>r to steE 1 otthe procedure. ~__..~.~
and icons.
155:41.b(10)4. "A Continuous Caution applies from the point at which it is encountered through the
'---___---I 'remainder of the EOP. The Continuous Caution symbol has double borders. The use of Continuous Cautions should be minimized,"
Reference Title Learning Objectives
.....--..--,-c~---.......--~-~--..........----~--.......~------'--.......------~-......:.-----,
TRPO~~EOO~
Describe the EOP usage rules associated with the following in accordance with SC.OP-AP.ZZ-0102(Q):
A._Immediate Actions B._Continuous Action Summaries C._Communications D. Log Keeping I
E.':::Application of Notes and Cautions F. Transitions G:='~ct..verse c::ontainment IQuestion Modification Method.:
I ~~i~n=Y=M=()=difi=le=d=====
'--_-..;._______....1 ~Q81103RePI~c;ed imPlaUSibledlstract=r___
........ _______....... ____~
14, 2012
which one of the following Unit 2 procedures are the Emergency Diesel Generator FIRE BYPASS switches directed to be in BYPASS?
Gontrol Room Evacuation,
__ FIR~~QQQJ,Gontrql Room Fire.~espon~_,____
@JIS..
...P-AS:CR-OOO.2, Gon
.. 1 R
...oomEvac.U8tion Due to Fir.e in the Control Room, Relay Roo
....m:l 2.,.O
..tro 460j230V Switc~gear ~o().rn.....Er 4KV Switchgear Room..
IAnswer I [] !ExamLevel!
!Cognitive Level I,\\/I<,rn*.."nt
!Exam Date: I 12/3/2012:
~==~~;'~!=;;*~=K;n=o=w;le=cI~g*e=a=nd=
......='='A'::Qi~Ii!~Le=..§=_====.... ::::=~;;;:;===;-;:: =:;-;:;:;:;~==:::;-r===,~.-J 194001 G427 J
!Record Number I rn<:l,rn<>nr'\\f Procedures I Plan n('\\\\JI/IAl,nA of "fire in the
.b(10) AB.CR-2, Attachment 4, Reactor Operator, pages 15, 19, and 22 place the keylock switches bypass. The 3 distracters do not contain steps to perform this action. The bypass switches remove the control from the EDG control, and are only operated when the control room has been evacuated due SEC be aberrant.
Reference Title Control Room Evacuaiion Due to Fire in the Control Room, RelayRoom, 460/230V Switchgear Room, or 4KV Switchgear Room.
Learning Objectives the appropriate abnormal procedure.
Describe the plant response to actions taken in the abnormal procedure.
Describe the final condition that is established the abnormal Question Source:
IQuestion Modification Method:
~~~~~~~~===============
14,
U.s. Nuclear Regulatory,Commission
"'~, ------- """"----~-"~--------~--------Site-Speeifi,e ----------------------'
Written Examination Applicant Information Name:
I Date: 12/3/2012 1 & 2 Level: SRO Reactor Type: W Start Time:
Finish Time:
Instructions
-,~---....-----------....
Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination you must acl1ieve a final grade of at least 80.00 percent overall, with 70.00 percent or better on the SRO-only items if given in conjunction with the RO exam; SRO-only exams given alone require a final grade of 80.00 percent to pass. You have 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to complete the combined examination, and 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> if you are only taking the SRO portion.
Applicant Certification --
Ail work done on this examination is my own. I have neither given nor received aid, Applicant's Signature
--...-~.
-=-----
Results r------'
RO/SRO-OnlylTotal Examination Values I
I Points
-.-~-.-
--...~---.---~
Applicant's Score I
I Points
- ___________.M_ _ _
~--~-.
Applicant's Grade I
I Percent
Senior Reactor Operator Answer Key 1. b e
- 3. b
- 4. d
- 5. d
- 6. a
- 7. a
- 8. e
- 9. b
- 10. b 11. a
- 12. a
- 13. b
- 14. d
- 15. b
- 16. d
- 17. a
- 18. e
- 19. a
- 20. b 21. d
- 22. e
- 23. a
- 24. e
- 25. a Page 1
Given the following conditions:
Unit 1 has performed a rapid downpower lAW S1.0P-AB.LOAO-0001, Rapid Load
(
Reduction, due to severe Circulating Water System grassing.
3 Circulators remain in service, one on each "A" waterbox.
With Rx power at 8%, the RO notices that over borating has caused RCS Tavg to drift below the Minimum Temperature for Criticality, and begins continuously withdrawing control rods to raise Tavg.
As control rods continue to be withdrawn, the CRS notices Rx power has risen to 22%
and continues to rise.
Which of the following identifies the proper response to this condition, and why?
pirect the 139 to...
___.__..___.__._~.__.~_o__.. _.. ____......______. ___. __._'__ 0'
[triP the Rx. Multiple RPS functions have failed-to trip-ihe-Rxa'nd-adequate 6NBR'cannot be ass~<!:.
__~........_... __.______.......___.....______
II the Rx. A control int~r!()gk has failed to E~~~~nt ap.r:)!oa~li~i~'Rx'tripsetP-()~~-..~]
limme~i9tel}'_~!()p rod motion. If RCS Ta"g. is >541°F,adjust..Stear1'!g:t:Jr1'!p~~g!().~er Rx powei.J
-~-
-.-~
>535°F.
functions have not takef!.2@~~_._._.....
~~r~========~~~=~====~======~~*=
This was the precursor to the Salem April 1994 event, in which continuous rod withdrawl was trying to raise tavg afet overborating and rod insertion. In the stem, Rx rods are still at 22% power, when Control Grade interlock C-1 should have acted at 20% power to movement. A is incorrect because it is not a multiple failure, either P-10 (10% Rx power) to reinstate the low power trips, or the C-1 failed to stop rod motion. C and D are incorrect in that Rx is rapidly approaching an automatic Rx trip setpoint, stopping rod motion and attempting to recove the is non-conservative and incorrect.
Identify and describe the Control Room controls, indications, and alarms associated with the Reactor Protection System, including: (Licensed Operator and STA Only) a)
The Control Room location of Reactor Protection System control bezels and indications b)
The function of each Reactor Protection System Control Room control and indication c)
The effect each Reactor Protection System control has upon Reactor Protection System components and operation d)
The plant conditions or permissives required for Reactor Protection System Control Room controls to perform their intended function The associated with the Reactor Protection control room alarms
.._....~....__..
______ ~..._. :c':*:.C*: :'::c:::",-C::=="".:.c:.:::-::""::.c::::::::::=. ccc::::::c: '.........____............
L *.~__...:.._u.....s...e....d............._
[)age1_-~f~~*~_1
October 11. 2012 3:32:57 PM
.-.-.-~....---......--..~--.'----
the following conditions:
Unit 1 was operating at 100% power, EOL.
(
A 70 gpm RCS leak in containment was identified.
Operators tripped the Rx and initiated a Safety Injection lAW S1.0P-AB.RC-0001, Reactor Coolant System Leak.
While performing EOP-TRIP-1, Rx Trip or Safety Injection, Step 23, PZR PORV STATUS, the RO reports that PZR PORV 2PR1 indicates open.
The RO reports 2PR1 will not manually shut.
The RO reports that 2PR6 PZR PORV Block Valve will not shut.
i IWhich ofthE3}()I!()~Jng identifies how the CRS should oroceE~a IIIContinue-in TRIP-1. The combirled' size ofthe-PZR PORV and RCS leaks will allow a
~!an§iJti()rlto TI3IP:;.~,..§.~fety 1_r]ec~i()r1_ Terr\\'l!r1ati(?l'l. ____.........
iii in TRIP-1. A transition to LOCA-1, Loss of Reactor Coolant will be made based on containment radiation IFITran$iti'on"to'LoCA~{A transition to TRIP-3 will be made after PZR PORV status is checked again in LOCA-1.
L..........._... _~._
11 to LOCA-1. 81 Termination criteria cannot be with either P0I3\\1not..f\\J11Y...~~ut.
=-'--'-'-""'-"'...,,==,::-=to:.:.. i::n:t,e,:::}:pJ:E3t~D.q.E3'S~~g~J)~8~~~~!~_~e_2_~:_=__..-
....-~::-=,_,,_
[I§J
......::-...::....::::::.::::-=.=:...::.... -
5.43.(5) A 100 gpm leak would require SI as stated in the stem, but charging pump flow would be able maintain both PZR level and RCS pressure. If a PORV is open in TRIP-1 and it or its block valve cannot be shut, a transition is made to LOCA-l because the EOP doesn't know how big the PORV leak is. Once in LOCA~1, there are steps redundant to those performed in TRIP-1 to attempt to close the PORV or its block valve, then it continues{there is no transition to another procedure based on PORV/block valve status.) Immediately after the PORV Status step in LOCA-1, SI Flow Reduction criteria are checked. All conditions should be met which are: Subcooling >QoF, AFW flow/Adequate SG NR level, RCS pressure stable or rising, PZR level >11 %. With RCS pressure given in stem of 2235 psig and stable, this would indicate the PORV is only slightly open vs full open. A is incorrect because while a transition WILL be made to TRIP-3, it is not made in TRIP-1, it is made in LOCA-1. B is incorrect because while a transition to LOCA-1 might be made based on rising containment radiation levels, the transition must be made now. 0 is incorrect because the small size of the RCS leak combined with the small~i?~()1.!b~.~9~Y()J:>_e!li.!!g_(§ls.~~0'.IV.!1 normal RCS SI termination criteria can be met.
...~~.-...._.-._._._--
~
(
11, 20123:32:57 PM Page 4 of 35
the following conditions:
- Unit 1 was tripped from 100% power when a SBLOCA occurred.
(
!- While performing EOP-TRIP-1, 11 SG was identified as being faulted.
1_ All actions have been completed to isolate 11 SG.
After transitioning to EOP-LOCA-1, the crew is performing the Faulted SG Evaluation steps.
11 SG pressure is 740 psig and lowering.
1-12-14 SG pressures are all 960 psig and lowering very slowly.
1Y'{~ic~~!h~.. fo~?""i~£J<:!~ntifi~~.D_?W t.h_~C RS_~.h9~lcj,~~_spg~cjL.~~""hy.?______________
III r..-..-.---.--;'.;-.-----,--... -.---.-.---.------.---.... -..-.. ------...........-.--......---.-.....-------...---.. --..***-------*----------*..-----1
'..'" IT~;~J~~~fs~r;;~~O~~;~;~~~;~~;:.~~~:~~~~_~I;~;~;J~~~~j~~...~h~~~~~~~~~Y2.~~~;;~~~~~__._1 1\\ rco-ntinuein-LOCA~1sincetheECCS-injectiOniscoollng the RCS and causing the unisolated I~:~~~;~~~~~ to lower. Going to LOSC-1 would only perform steps which have already been II in LOCA-1 since the ECCS injection is cooling the RCS and causing the unisolated pressures to lower. The additional subcooling provided will allow an earlier transition to d
SI Flow Reduction II fTransltiO-n-dire-ctlyto-EOP::LO-SC~2because' all SGs are now faulted. Steps to if 81 Itermination can be performed will be adversely affected due to the lowering RCS pressure from the SG an_d.___c___a_.._u___s_e._..._u... _n__e_.c~~§a._ry_p_r()c_e_dIJ.r:..~.E~()!m~!1_~~~._..... ___......____._... ___.J
_-~.-
....' [~:J'.'_Is_.]
to interpret control room indications to verify the status and how actions and directives affect and The procedure transition to LOSC-1 is made before LQCA-1 when in TRIP-1. The stem syas been completed to isolated the faulted SG. These actions would have to be performed in LOSC--1. The transition out of LOSC-1 is SGTR-1 if there is a rupture or to LOCA-1. The first step in LOCA-1 is to check for faulted SG's that have not been isolated. C is incorrect because RCS pressure will still be lowering (not stable or rising) based on the faulted SG. A is incorrect because the other SGs are not faulted, they are reacting to the cool ECCS water being pumped into the RCS, but the reason is pICl~sible. 0 is incorrect because there is no di~ecttransi~ion to L()SC-2, you first have to enter LOSC-1.
(
Thursday. October 11, 2012 3:32:57 PM
the following conditions:
/
Unit 2 is in MODE 5 after refueling the Rx.
The Rx was shutdown 20 days ago.
21 RHR loop providing shutdown cooling, and 22 RHR loop aligned for EGGS.
21 RHR HX inlet temperature is 105"F and stable.
RGS pressure is 205 psig.
21 RHR pump begins cavitating due to a valve being mispositioned during a tagging release.
The GRS enters S2.0P-AB.RHR-0001, Loss of RHR, and stops 21 RHR pump.
During performance of S2.0P-AB.RHR-0001, the crew has time for normal restoration and local venting of the RHR system.
1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after the initial loss of RHR, 22 RHR pump is started.
1-The RO reports RHR flow is oscillating between 1,500-3,000 gprn.
1-The highest GET temperature is 184°F.
IWhiCh of the describes how the GRS shl:)LlI~ pr()~~~_~?~
II and initiate Attachmerlt 7, Hot Leg Inj)~tion.
55.43(5) The conditions given in the stem indicate aRCS heatup rate of -1.3 deg per minute. With the initial temp of 105, this will add 78 degrees and when the RHR pump is started RCS temp will be 183°.
RHR flo'w is required to be stable between 1800-3000 after the venting and starting of the RHR pump to allow exiting the procedure with all other RHR system parameters normal.(page 12) With 1,500 gpm flow swings, this will be answered NO. Step 3.30 states to stop any running RHR pump. Step 3.32 (page 18)states to initiate an alternate method of Decay Heat Removal. With CET temps <200°F, the preferred method is Attachment 8 Cold Leg injection. Attachment 7 Hot leg injection is not preferred due to RCS
,!)eJ.I1~JI'l~a,<::t and <200°F, With the P~c)~E:ldl'.r~~~t~n~g~tB...~t92all RHR pumps, startil19g1 is l10t directed.
Q80328 concept used, expanded to make SRO level. Changed 2 distrac\\ers to start idle RHR pump.
October 11, 2012 3:32:57 PM
Thursday~
11,20123:32:57 PM I
IGiven the following conditions:
1-Unit 2 is operating at 100% power.
1-The RO reports that the PZR Cold Cal level channel (LI-462) is indicating 0%.
i I iWhich of the following describes how the CRS should respond?
,,~~,--
,,-.""~----
--~.. -~.--.--.--------
§J jEnterTSAS 3.3.1.1 Reactor Trip-System Instrumentation. This Tech Specisbased on being -,
jable to provide the overall reliability, redundancy, and diversity assumed available in the facility idesign for the protection and mitigation of accident and transient conditions. Place a single
- pje~~ of~d translucent tape across LI-462.
@;]IE.nterTSAS3.i.3..7. ACcidentMon.. itoring-'instrumen.tatio.n. -T.'his T.*e..cll* spec. is. base..d....on enSUring.**]
lsufficient information is available on selected plant parameters to monitor and assess these "a.riables followi~.9an accid~nt Place al"l!~fO sticker onLI~.?~ _
l
@] INo Tech Spec entry will be made since TSASi3.1.1 andTSAS 3.3.3.7 are applicable to this '-1
- instrument ONLY in MODES 4-6 and during movement of irradiated fuel. Place a single piece.
Lof red transluce':!t!~pe across LI-462.
I
@]lrNoTe.chSpece'ntryWill*'bemadesirlCeTSAS3.3.1.1 an'd TSAS 3.3.3.7 are never applicable 1 to this instrument. Place an INFO sticker on LI-462.
~--
/ArislNel' I\\dl It;xam H~yell ~s' :!:t;Q9niti~~1~vell'ME:moryn.. ',~='i'F'l(:flity:lIsalem 1 & 21\\exa,Pf!te; L ~. 12/312012j iEmerg~t1.c:Y and Abnormal PI§lnt Evolution~
IRbGfoupll.21IsR,Q;\\Grbupll_~21 lc)Qo_028G;24.3.. 1
[028
.. llpre.s§.urizer LeyeIGon,troLMalfynction. -
- ]
51
- 2.2-:l~gUlpment Control 12.2.4.3,1 iKnowleq~.e ofthe process' used trackin()perable.Cl.larms...
i 3.0113,,:3J 155.43(2)(5) The PZR cold 'calcha-nnel is not included in the Rx Trip Instrumentation or theAcCldent.
,--,,---,----,----,-.-;..JjMonitoring Tech Specs. It is used when RCS temperature is <200°F as directed in S2.0P-IO.zZ-0006, iHot Standby to Cold Shutdown. A is incorrect because TS 3.3.1.1 is not entered, although the Bases is Icorre,ct.. Also, the red translucent ta~~Is used to identitx_an inop<;~~~le alarm.
,Control Room Instrumentation and Alarms j
Hot Standby to Cold Shutdown Given a situation dealing with Pressurizer Pressure and Level Control System operability, examine the situation and apply the appropriate Technical Specification action. (License Operator and STA only)
NCT State the Technical Specification associated with the component, parameters and operation of the Pressurizer Pressure
- and Level Control System including the Limiting Condition for Operation(s) (LCO) and the applicability of the LCO(s) (Non*
l.li(;ensed Operator)
I
[
IMateria. Reqtilre(t for/,:x~mfhation'
,II I
I I
Thursday, October 11, 2012 3:32:57 PM Page 9 of 35
.Given the following conditions:
Unit 1 is operating at 100% power.
(
The RO reports the following:
PZR level slowly lowering.
Charging flow slowly rising.
PZR pressure 2235 psig and stable.
Condenser radiation monitor R15 in WARNING and rising.
I I
~s~~~ht~!~~o:~~~ennS~~~dt~h~~ocedure which will provide the most effective mitigating rs-2.6Fi-=-AB.SG.. -0001, Steam G.e.n.*s*raior-fub.e. Leak... Actions Providsdin-.proc.edure*WiII reduce l!b~sp~~~.~LoL~<?I1!?_~inat19~.._. ~__._.
._~_
[I rS2:0P=AB~RC-:0001, Reactor-CoolantSystem Leak. This procedure checks in a prioritized l.r!!§I!::~neifo!.~IL~~~rces.ofBg§I.~<:!kage*__oo________
~ IS2~P=AB.SG=OOO~Steam-Generator-TubeLeak. Actions providedinp-rocedure will ensure that th9.! anY_(3.ffected SG tube will....n
....._o_t__'-.. 0....
_..0_._._____._00_.
o J
IIIIS2.0-P~ABRC-0001, Reactor Coolant System Leak. This procedure checks first to ensure the I~~~tl~;k;;f~i; ~~~~f~I;_~\\~e~~i~ i~~r~~~~vc~.:~~::_~~~~_~~ty a~_~provides direction for Ii] ~lli 1I;911M1~lliQ[)IiC~~=_*_~.=_]_1i~~'e~ 1 &~~.~.-]...I[~_
l~ll1~rgency'and_6bnSJ!!11§I_.F:!§i'LE:-VOllJ~QQ~_=:J 1li1IM c*:--:--_**-*
[§!~?ln G~nerator~~b_e.Lji~L
,c:::::.....===
nfQrrm::.t the following as to Steam Generator Tube Leak:
to detect leaks C is incorrect because while the actions to shutdown the reactor and cooldown/depressurize will likelihood that the tube will not rupture, it does not ensure that it will not. A is correct because the procedure checks that PZR level can be maintained stable or rising, actions are taken to Iminimize the spread of contamination by isolating the SG to the effect that it can be isolated. AB.RC is Inot the most effective procedure because it will require additional time to get through the procedure If~~~~~~;~~~i~~;'~J~~~~nt.o AB.SG, lengthening the time before actions are taken in AB.SG to minimize Q39610. Originally had 4 procedure choices with no why part.
Thursday, October 11, 2012 3:32:57 PM
the following conditions:
Unit 1 is operating at 100% power.
(
12 SG NR Channell has been removed from service while undergoing a Channel Calibration lAW S 1.IC-CC.RCP-0045, 1 L T -529 #12 Steam Generator Level Protection Channell.
12 SG NR Channel IV fails high.
Control rods begin stepping in at 72 spm.
b_~~Ef~~~___. describes how the CRS should EOP-TRIP-1 Reactor Trip or Safety Injection, and attempt to trip the Rx from the control room. If unsuccessful, enter FRSM-1 Response to Nuclear Power Generation and dispatch an
~~~t~)~.~~~~!Iy~r-ip_!~.B~_~____._....__.___._.._.__...... ___........ _..._..__.. ______________.__.
IIlf [Enter EOP-TRIP-1 and attempt to trip the Rx and Main-Turbine from the control room. If unsuccessful, dispatch an operator to locally trip the Main Turbine, then enter FRSM-1 and
!cJl~~~~_~I'!~~r~~~_~l2._I~g~)!~!':ip~~_____.__... _
FRSM-1 directly, and ensure that the Main Turbine is AMSAC will start ONLY MDAFW when SG NR level is < 5% for> 25 seconds.
to Loss of Main Feedwater:
signal is generated by 2/3 NR level channels on 12 SG being >67%, The P-14 signal trips the Main Turbine (which is indicated by 72 spm rod insertion) trips the Main Feed pumps and shuts the BF19s and 40s and 13s. (FW Isolation signal) The reactor should have tripped on the Main Turbine trip, but has not as evidenced by control rods stepping in on the Main Turbine load reject vs being on the bottom. The FRSM distracters are both incorrect because FRSM-1 is not entered directly, even though the actions of D are correct with the correct setpoints. C has incorrect action and setpoint. B is incorrect because the l'II1~i!l1uI.~i!1~.b~~_~I~~~cly_!J:iP pe~___
[Given the following conditions:
1 1-Unit 2 is at 40% power performing a shutdown.
(
1-4 SW Bay is isolated due to a leak on the 25SW3, 25 SW Pump Discharge Isolation Valve.
1-Operators are performing the shutdown to comply with TSAS 3.7.4 because difficulties
, arose during the leak repair of the 25SW3.
2A EDG is supplying 2A 4KV vital bus for a scheduled surveillance.
21 and 23 SW pumps are in service.
1 of the following describes how the control room crew will respond if 2A EDG output breaker on Bus Differential, and 23 SW pump trips 1 minute later on over current when supplying SW at runout conditions?
~=-~c:=::::::.
fllTri*p.-t.he.-Mai.n-.T.*.U.r..bin.e t.o. reduce Rx powe
.....r.* and heat input to the ReS, and enterS2.0p=---J
~§~II3~=9.Qq~1,_ I~rpln.~_TIip_~-=9_._..__..._....._~..___~_..._~__.. _.... ______..__
II S2.0P-AB.SW-0001, Loss of Service Water Header Pressure. Trip the Main Turbine, reduc~_~x~()we~'::~_~~oJll~.r:.c:!.~I.t()""e.I§tce AF'J'!. in service.
tIllEnter S2:0-P-AB.SW~OO()5~Loss**ofAli-SerViceWater. Trip the Rx, confirm the trip, and stop lB.Q~lS~t().lil"T!i!!b.~__hea~inp~!._!~Jh~~gq'.0L~y~!~~t"!l_,_?1l2p~eserve ReP. ~~~I.P~E~~g~lS:.. _..
11 ~~~~=s~n~ne~ ~~t:~~~~~~A~e1~~~~~t~::~i':~~;n:~~~ :~~;~~s~~;~~~:~~~or
,yvate~_pume.s op~!atill_g~_.
1~.-8il~'
~Ii~-:I-.--_-,*.
_.-=~~=1_~t?~oI2j
.=*W~'
-~.J""'.diI\\j=,;1
[9~QQ§?GAQ§~-~]
155.43(5) A is incorrect because even if the actions are (some ofthose) performed in AB.SW-5, the next procedure entry to AB.TRB is incorrect since the Rx is tripped in AB.SW-5. B is incorrect because
!AB.SW-1 doesn't perform those actionS. C is correct. 0 is incorrect because AB.SW-5 should be
~rltI:E~<:l~ef9E~~xi!irl.9_t~~JRI p se
...~r...ie...s._..... ". ___.. ~_
Given a set of initial plant conditions:
A.
Determine the appropriate abnormal procedure.
B.
Describe the plant response to actions taken in the abnormal procedure C.D~.sc!i~~t~e_fil1l>!eIClnt<::.Slfldl~on.!.hat i.s!3sta~is~ed b.¥tI1_e.abn(}~01al pr(}c;edure Modified
r-----.---....
.~--
IGiven the following conditions:
I 1-Unit 2 has experienced a MSLB at the Mixing Bottle.
All attempts at MSLI have failed, and 21-24MS167s remain open.
Operators have just completed SI termination steps in EOP-LOSC-2, Multiple Steam Generator Depressurization, and PZR level is being maintained stable.
i-AFW flow to each SG is 1.0E4 Ibm/hr.
1-The RO reports rising pressure in 22 SG.
['I(\\'hich of the follo\\IVing describes!!.ow the CRS shoul~proceed. and why?
§J ~Transition to EOP-LOSC-1 and stop -RCPs if RCS pressureis<1350 psig. since Reps cannot L~e st()pped in LOSC-2._.. _ __..._......._____
[§J ~Transition to EOP-LOSC-1, Loss of Secondary Coolant, since one SG is n-ow available for lsubsequent recovery §lctions.
[5ZIIRe-main in E6p~LOSC-2 until positive-control can be established over'the cooidowrl after the Irel"l1ainin_g~tE:;am Generators h~v~}u}Iy':iepressurized,t.h~~ tr§l~sJ!i.on to EOP-L9SC~_1 ~_
~ IR-emain in EOP-LOSC-2 since retur-ningto EOP-LOS-C-1 will require atransition to EOP=LOCA~I l~~~;~scu~~~~~tion and ~~"~_ompllcate re_cOVery_afterthe_r~:aining SG~h~VefUny_...
_ I
...",--~.,.,
IEx<im Date: I 1~J~~2~1~1 lQOWE-12A2011 IE12--- _l!Jricontrolli;?9 Depressurizati9D. ()j _aU StearnG~n.era_t9rs
'-"'--q-'--~' -.-',
9i
[gt\\=?,J.!A..biiitYtodetermine andinte. rpret tile -following as theyapply to Uncontrolled Depr
..es.surization of all Steam ****.1*
iGenerators:
TJ 1~;~~;~~nOs~ditiOnS ~~~ selection of appropria~~_~rocedures dU~~9 abnorm~~ an_d emer=e~cy' 3.211 4~oi 1
f55.43(5) LOSC-2-CAS states that upon a pressure -rise in any SG-except when performing SI termination
,in Steps 8-20, GO TO EOP-LOSC-1. The stem states that it is after Step 20. LOSC-1 Basis Document,
!page 7, states that.."Any cooldown operations that are performed as subsequent recovery actions will require at least one nonfaulted SG_"
1,~arni~90J>lectjV~~..
c......:.;..;;_.*..".... _..........* ::
A Determine a discrete path through the EOP.
S.
Determine an appropriate transition out ofthe EOP iFacility Exam Bank IIQi::'~s;ti6n!V\\odificationl\\lletqod:'-
.- Ilconcept Used IVisit:m Q57956 conceptused, and added the "why" to question,
~~---~~~---~~
r Thursday. October 11.2012 332:57 PM-]
Page 13 of 35
I"
-_.......-~...
IGiven the following conditions:
~-
I i-Salem Unit 2 was performing a Rx shutdown due to indications of failed fuel I
after chemistry reported reactor coolant activity to be 500 uCilgm dose equivalent 1-131.
1-During the shutdown, the RO reports lowering PZR pressure and level.
1-With 21 charging pump in service, PZR level continues to rapidly lower, and the RO I reports containment pressure is also rising.
The RO trips the Rx and initiates a Safety Injection, and all equipment responds as expected for the SI.
II-The SM declares a Site Area Emergency.
I-RCS pressure continues to lower, and 30 minutes later is 35 psig.
I
[Which of the following identifies a condition which would require notification of the NRC, and the I<:orr~t time f~r that notification?.
....~..
§J [fhe.\\iVinddTrec!ion shifts from 0° t01ffoo. 15 minutes.
~ (gontainl11e-rrtri3dlationlevel exc~~.d~_~QOO R /h*r. 60 minutes.
I
@J [g.ont~inment slJ~.!~IJ,~! indicated on 2CC1 has remai~ed stableai46%. 15 minutes.
I It] ~he control roofrlmust be evacuated and operators cannot access theAuxiliary Building due'.1 Ito hi~~h radiation. 60 minutes..._.~
tA!lswer lib Hexarn Leve" ICQ9ritiveLever Ilt.Q2ILC?!i9D
. J[FaCility:. I[~~!ef11~!*~.2-I!ExaritOa~e: I 1273/2012]
'Emergency. an.d)\\bnorl}1al Plantl::~olutjons I,ROGroup !1:- ~I 1$~Q.~~i:)~p;1 21
[OOVYEI6G430i
,-"~"""",,,..,1;0'-'.***,*.*,*,**", l.
10J
[~i~.1 !l-lighJ~ontairlment Radiation I!2.4 I'LE:!!'ler!:lencyerocedures I Plan..
'.'.*CC.IIICC.=*..
!2.4:.3<fl [Knowledge of events related to system operation/status that must be reported to internal I'
I?rganizations or external agen_<?i~s~such as_ the State,t!1_e NRC, or the transmission system op~r~toL, 155.43(1,2) The 15 minute times in the question are from the 15 minute notifications required during
,emergenCies to the states. The NRC is not required to be notified for 60 minutes. The second knowledge part of the question is what would cause a notification to be required, i.e. a more sever E plan classification is made. The radiation levels> 2,000 R1hr adds 2 points from the containment barrier. The stem states that the SM declared a SAE, which would have been under FB4.L and RB2.L each of which is 5 points. The 2 additional points would put the unit in a GE. The wind shift while in a would not require a notification because no PAR would have been made for the SAE. The containment sump level is NOT expected, with containment pressure at 35 psig, the entire contents of the RCS are on the floor and level would have risen, as is seen for LBLOCAs. The 15 minute time for this condition is wrong, however. The CR evac and inability to establish control of the plant in 15 minutes is a SAE, and the plant is already in a SAE.
lSalem ECG 1
- tM
- ::.a=te~ria~j=*.~::::eq~U=lr=ed=f:;'"or:-:*E--:.x=amin=at=io=n~-::",:"-J"liSRO 10 Salem ECG' IGl~estionsqurce:.IINew I. Thursday. October 11. 2012 3:32:57 PM r Page 14 of 35 -1
Thursday, October 11,2012 3:32:57 PM... J Page 15 of 35
the following condition:
Unit 1 is in MODE 5 during a plant startup.
11 RH R loop is in service.
12 RHR loop is aligned for ECCS.
13 RCP is in service.
11 charging pump is in service.
RCS Tavg is 175°.
RCS pressure is 310 psig.
PZR level is 60%.
When placing the second RCP in service, RCS pressure momentarily rises to 390 psig.
l\\~\\L~lCh of~he follo~_~!l9E3~g~i~s_!~~RHR and how the CRS shouI9J:I~g<2E3E3d?~__
~ ~he -1R'H3,-R~HRSAF RLF VLV TO CONTAINMENT SUMP opens. Enter S2.0P-AB.PZR 10001, PZR Pressure Malfunction, and ensure that any PZR PORV that opened in response to the RCS E.re~~!:,r~ ha~ shut. _ _
IlIThe TRH2, RHR-'COrVIMON -S'UCr-MOV-auto-m-atically shuts. Enter S2.0P-AB.PZR-0001 and 1~!1s~~I"lY_~?B.EQB""!~~~g2~'2~s!"if'l~_s29.!1!E3_~C?!~~..R<2§_E~~_~S.~_~_a._s.sJ:ll:Jt_____... ____.__,.
II n-he1 RH3 opens. Enter S2.0P-AB.LOCA-0001, Shutdown LOCA, and isolate letdown to
~~~~§i,'2\\J~~lgs.~~ ___..
1RH2 automatically shuts. Enter S2.0P-AB.LOCA-0001 and isolate letdown to minimize loss.
the impacts of the on the Residua! Heat Removal System and (b) based nrofi'l"'t.nnc use procedures to correct, control, or mitigate the consequences of those abnormal durin!;J cold shLit9g""'II____c,_c_..:_:::.,,'=::::.:::::::.-::::::-===:::"::==C====-:=.=::c==::=:::::::::-::::=:':::,:-::':::::::::::':::',
With RHR in service both the PZR PORVs and the 1 RH3 will open at their 375 psig setpoints.
opening will not be apprent to the control room, but the PORV opening will. AB.PZR,, will ensure the PORV has shut The 1RH2 has an OPENING interlock that requires RCS I[:pressure to be <375 psig, then a keyswitch opens the valve. There is no automatic closure associated with this valve on high pressure. AB.LOCA is used in MODE 3 and MODE 4 with the accumulators isolated, and with the unit in MODE 5 as described in the would not be entered,
""..,---~-~-.---..------.-
LOR NCT Describe the function of the following components and how their normal and abnormal operation affects the Residual Heat Removal System:
a)
RHR Pumps b)
Refueling Water Storage Tank c)
Heat Exchangers d)
Motor Operated Valves i) RH1 and RH2, Inlet Isolation Valves Ii) RH4, Pump Suction Isolation Valves Thursday, October 11, 2012 3:32:57 PM
iii) SJ44. Containment Sump Isolation Valves Iv) SJ69, RWST to RHR Suction v)
RH29, Miniflow Reclrc_ Valves vi) RH19, Loop Isolation Valves vii) SJ45, RHR to SI or Charging/SI Pump Suction viii} SJ113, CCP-SIP Suction Cross-Connect Valves Ix) SJ49 Outlet Isolation Valve x)
RH26, RHR Hot Leg Isolation xi) CS36, Spray Recirculation from RHR Valve e)
Air-Operated Valves i)
RH18, RHR HX Outlet Valves ii) RH20, RHR HX Bypass Valve f)
Other System Valves i)
RH3, RCS to RHR Inlet Relief Valve ii) RH25, RHR to RCS Hot Leg Relief Valve iii) SJ48, RHR to RCS Cold Leg Relief Valves iv) RH12, RHR HX Bypass Valve v)
RH17. RHR to CVCS Letdown vi) RH21, RHR to RWST Containment Sump Anti-Vortex Baffle Orifices October 11, 2012 3:32:57 PM
Given the following conditions:
1-Unit 2 is operating at 9% power, performing a power ascension prior to rolling the Main Turbine.
OHA F-17 IR FLUX HI is received in the control room.
The Bistable for 2N36 on 2RP4 is lit.
The reactor remains at power.
Which choice identifies what has happened, and what action(s) isfare required to be performed?
1.."..."...*".__.- ".~...*....."... ~........__._.. _._**__.....~.____...... _.**..*_*..___._.......*._
!ll~%~~~~~;;~~~~:;~_;:j~~J~t~h~~~;~~;~7.~~~:I~oierify the turbine is tripped,
~ IThisaTa,:m"lsexpected at -10% power. BLOCK bot-hintermedlate-rl:lng'echannels by
!dep~~~il'1~L!he_I?LQg~INTE::RMEDIATE RANGE A and B PBs lAW OHA F-17 ARP.
IA failu'reof-theTRhigh"flux tri*p"blOckhasoccurred~Lowe"rRx power to less than 5% and Idepress BLOCK INTERMEDIATE RANGE B pushbutton to block Tn3in B lAW S2.0P-IO.ZZ 10004 POWER OPERATION.
r-.. ------. -~.---.-
IGiven the following conditions:
1-Unit 2 is at 100% power.
"LOSS OF TRIPPING CAPABILITY" Alarm is received for "A" Reactor Trip Breaker (RTB).
IWhiCh of the following describes the effect on RTB "A" from this condition and how should the CRS Iproceed?
~_t?§l9tor lj"]p Bre~~~r A.:..::.. _______
rfr;\\1
-'--:-1 r--~~--**------*-------~--~-***********-*----*-*-----*.-
~ Iwill NOT trip from an Automatic Safety Injection signal since the UV coil is unavailable. Restore I lL:JY..E~~biU!"~L~ithi!!_4B~oLl~s or ~_t?)r-l.tjot Stand within the next six hours.
II !will trip from ALL Reactor TriP-initiation signals EXCEPT a Manual~:;a-fetyTr1jection since the
~;aun~~~o~!~~J~~_~n~_~_:i=~~.~~~~re_Shunt ~~ capability within 4~_hours or be in Hot Ii tWill NOT trip from an-Automatic Safety-IilTe-ctlo-n-signalsince the UV coil is unavailable. Place
- Reactor Trip Bypass Breaker A in service and open Reactor Trip Breaker A within one hour or in Hot within the next six hours.
RTB Shunt Coils are energized to trip. The RTBs also have a deenergize to actuate UV Both of these open the RTB. The Loss of Tripping Capability alarm indicates that 125VDC power been lost to the Shunt Coil. As shown on drawing 221051, the only Reactor Trip which goes SOLELY the Shunt trip coil is a Manual Safety Injection. A is incorrect because an auto SI will trip the Rx, but action is correct. C is incorrect because of A above and the action is incorrect. B is correct because above and TSAS 3.3.1.1, Table 3.3-1, Functional Unit 21. page 3/4 3-4. in Modes 1 and 2, directs
- 14. Action 14 states that with either the UV or shunt trip unavailable. restore it to Operable within hours or be in HSB within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
. ----~---.~.~
Bank
===========:===~===~
E~iO~-Q77831.Ad(iedTeCh Spec part to make SRO level
[~age1-9-~f.~5.*~]
I
- l. Thursda~*October 11, 20123. 32:57 -~~.-'
Page 20 of 35
Given the following conditions:
I Unit 1 is performing a Reactor startup lAW S1.0P-IO.zZ-0003 Hot Standby to Minimum Load.
!1-All Shutdown Bank control rods have been fully withdrawn.
1-Control Bank A is fully withdrawn.
1-As Control Bank B is withdraws past 20 steps, the RO reports OHA E-48, ROD I
BOTTOM has just alarmed and remains locked in.
1-No other alarms are received.
1
[yvhich of !~s>II~irlg describes how the~~ystem i~ ~~perating, and how the 9R§ sh0l-ll~pro~ee5~?
[!J iThe-Rod Bottom Bistabie~~causes OHA E=-48 to a'iannas eachcontrolbank'is withdrawn past 120 steps and is expected. The CRS should direct the reset of the Non-Urgent Failure to reset l!!1e.CiI~Em,th~continue the startLJP'
~ jThe Rod Bottom Bistable causes OHA E-48to alarm as each control bank iswithdrawn past
~ 'I
- ~~~;~~~~~~s1~b~~~~~s;r~~~:~r~,0~~~d~~~ii~hu~ ~h~ ~:;~;~ss the STARTUP J
~ IThe Rod Bottom Bistabie Clearecf when Control Bank A was withdrawn past 20 steps and OHA IE-48 is unexpected at this time. The CRS should enter S1.0P-AB.ROD-0002 Dropped Rod l~,rI~~I~l~~ct ~he.c>peQ!111tof the ReactoE~Breakers !o~~rminate the Rx~tCirtuJ?_._...
@J [The Rod Bottom Bistable cleared when Control Bank Awas withdrawn past 20 steps and OHA
!E-48 is unexpected at this time. The CRS should place the startup on hold and initiate S1.0P IAB.ROD-0002, Dropped Rod, or S1.0P-AB.ROD-0004, Rod Position Indication Failure, to idetermine what malfunction has occurred.
I.
IIEx~mi~e¥ef! ~_.l ~C99nitip~i;WelllQ2rJ1Rr~h~nsioD J,~~ilitY;lrs~al~m~1 &2_
'1IEX'!rnr'patepLu-12/312012]
IPI~'lt§ystems IR6.Gro~tH. 21 !SRq;r;;rouR,!
i01B060G2:37., J u
'llN9n~Nuc~Erln~ktLiDeQt~tj9_n Sy~t~I11.~~___~_, ____..1 l?:2*I[E_~Jpment Control
.,,.,1
- 2.2..~I~ i [Abili~to determine o.fl~rabilityandior* av~f~bility of safety related equ(e::me-I'It~~
.......J[3.6i [*i'I3=j The Rod Bottom alarrn CLEARS when CB A is withdrawn past 20steps This is because aRod Bottom '!
J..-.-.C....l."--___'-I Bistable Bypass for each of the other three control banks B,C,D bypass the alarm for their respective I'
bank when all rods in that group are below 35 steps. This means that the alarm was CLEAR when it I
alarmed. it did not reflash. Since no other alarms occurred, the CRS should place the startup on hold and' enter AB. ROD-2 (which will direct entry into AB. ROD-4) or enter AB. ROD-4 directly to investigate the I
failure. There is a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> window for having to terminate the startup (IOP-3, step 5.2.19) so opening the Itrip bre§lkers isnot required..
~
iRod Control anc:f Position Indicating Systems Lesson Plan iDropped R()d
[~od Position Indication Failure NCT Describe the function of the following components and how their normal and abnormal operation affects the Rod Control and Position Indication Systems:
Rod Cluster Control Assembly (RCCAl Control Rod Drive Mechanism (CRDM)
Rod Drive MG Sets Reactor Trip and Trip Bypass breakers Reactor Control Unit Power Cabinets Thursday, October 11, 2012 3:32:57 PM Page 21 of 35 J
~
Logic Cabinet components:
Pulser Master Cycler Slave Cyclers Bank Overlap Unit
- h.
DC Hold Cabinet Rod Position Indicator (RPI) Coils
- j.
Signal Conditioning Modules
- k.
PulseOtoQAnalog (P to A) Converters I.
Rod Bottom Bistables
- m.
Rod Insertion Limit Comparator Counters
__ :=-==::==:cc=,=:':
---.---.------~-..-.~-
.- 1 Used rv.-.i5l0nQ60249 expanded from rod bottom bistable question to whether alarm is expected and what CRSshould -I ldo.
..-~-,,-.--"--.-- -
.. ~."'-~-"----'--' -_..... --
Unit 2 is operating at 100% power.
Technicians are performing a sensor calibration of Containment Pressure Channel IV lAW S2.IC-SC.RCP-0066, 2PT-948A CONTAINMENT PRESSURE PROTECTION CHANNEL IV.
All bistables and test switches are in their proper alignment for the calibration.
1 While I&C Technicians are performing the calibration, the control room recieves OHA C-16, PHASE B CNTMT ISOL ACT.
Which of the following identifies what has happened, and how the CRS should respond to this
. ~'~'-=':':.=-".'=:"-=:'.:::~:.:=:::':':"':....::"=:.~.::':::'=':':
II Phase B isolation valves-mwill~shutONLY. Th'e isolation will not be able to be reset untill&C returns their bistables and test switches to normal. Trip the Rx, stop all RCPs, and GO TO EOP-TRIP-1.
iii IPhase B isolation valves will shut and ContaTnment~Sp~rayvalves will open. Attempt to reset and open the Phase B isolation valves, If unable to reset Phase B, GO TO S2.0P-AB.RCP 0001 RCP Abn()rl1'1§llity:~_..... __._______......__._...__._.
..... ~~._.
r _____..................mm..___...................m..____._..____~
__._m..._...___.
B isolation valves will shut, Containment Spray valves will open, and both Containment Spray pumps will start. Attempt to reset and open the Phase B isolation valves. If unable to
~~~~ Phase
.._._.e..._ct__
the Rx and GO TO EOP-TRIP-1...R..._a..or_,---___.........~--".__..__ ____.
Ia [ThiS is an expected alarm during performance of the test because the Channel IV Containment ISpray bistables were previoously tripped, and the Phase B Isolation signal is not expected to j§~~.;;;;~~~;g~;~~;~~~;~g~~~!~'~ff:~:~:h_~~~._~~~~~ti~:_~~I~~s and Containme ___t____..._ _
..._n
_1...1~caiiOn--_=*.*,.~mT&2mmmm=-J
!lI!!:.!II'!. a.,ll'!'"..
~_ [JJ _11111 nnl*r;::I1'A controls identified in the alarm rA<::nnr'!::A manual.
i~==:-====:====-==-===~==~=.:='
55.43(5) This alarm is not expected to occur during the sensor cal, as it requires 2/4 containment ijpreSSure channels to see 15 psig. The ARP says if cont pressure is <15 psig, attempt to reset and Phase B isoaltion valves, and if unsuccessful go to AB.RCP based on the loss of CCW to the RCPs.
.~§y~<3ly~~.~i!I!~Eo~ti~n! !J~!tlle_ Q..s..~.r11p~__~illl'l~()ts!art.
lQ~erhead Annunciators Window C Identify and describe the Control Room controls, indications, and alarms associated with the Containment Spray System, including:
The Control Room location of Containment Spray System control bezels and indications. (Licensed Operator &STA only)
The function of each Containment Spray System Control Room control and indication. (Licensed Operator 8. STA only)
The effect each Containment Spray System control has upon Containment Spray System components and operation.
(Licensed Operator &STA only)
The plant conditions or permissives required for Containment Spray System Control Room controls to perform their intended function.
Operator &STA only)
The associated with the Co~~.!.n.'!lent ~~¥.~Y~~.'!lc.?ntrol room alarms.
the following conditions:
i-Fuel handling is in progress in the Unit 2 Spent Fuel Pool when a fuel assembly in the
(
Spent Fuel Handling Tool is dropped.
l Gas bubbles are observed in the vicinity of the dropped fuel assembly.
1, 2R5, Fuel Handling Building radiation monitor, goes into alarm.
IWhiCh of the following describes the effect of this event, and contains actions that will be performed lAW S2.0P-AB.FUEL-0001 Fuel Incident?
II Auxiliary Building ventilation automatically swaps to place the Charcoal Filter in serviceto-"--J IPE~_'I's:I!! a relea§~to th e _~Jlvi~.~'!':!:Iel!t.§.llsur~J:b.~_f!:i~~.l::Ipply Fa.ll,-!~J"!:l nl!in9~__......_._____..____
_[A
.. LL Fuel Hand.l.i.n.g crane. m
.. o.tiO.n... except downward movement is locked out to prevent raising
?_9amaged fuel_§l:~~~!!lPJx..t::I1~ure the FHB Truck Roll Door is closed.
[ContainmentVent.ilation Isolation activates to ensure any containment pressure/vacuum relief-li_~ service is terminated. Ensure the Fuel Transfer Cart is NOT at the Spent Fuel Pool and 19l9.~~.!.~.E?.S3_~~.._.___..____......_____.___.._..____.._........___....
. J 1
Handling Building ventilation automatically swaps to place the Charcoal Filter in service a release to the environment. Ensure the FHB Watertight Door remains closed JlJIIiIIL ___ !~3~13l l0340QQ~t'-=]
~=~~'~~:=:========~======~====:=~~~=~.==::===~~!
Ability to (a) predict the impacts of the following on the Fuel Handling Equipment System and (b) based those use procedures to correct, control, or mitigate the consequences of those abnormal The 2R5 in alarm swaps the FHB exhaust ventilation to the Charcoal Filter and starts Fans. The normal configuration for FHB ventilation is the single Supply Fan running, and BOTH Exhaust Fans running. A is incorrect because ABV does not auto swap to charcoal filter upon a 2R5 high alarm, nor does the Supply fan get an auto start signal. B is incorrect because while the FH crane only locks out as described with the 2R32A rad monitor on the crane itself goes into alarm, not the 2R5 area monitor, and the action is correct. C is incorrect but plausible because CVI does not actuate, but FHV oes discb_~Q~.t().l~~EI§t1.!__~~n.t~.,..be a~io~ll£:)i~ correct.
Q109317. Added why and action to be performed in AB.Fuel-2.
["GiVen the following conditions-:
1-Unit 2 is in MODE 3.
Welding in the Turbine Building has caused an actual deluge actuation to occur in the Turbine Building.
1-The control room recieves the following alarms:
OHAs A-7 FIRE PROT FIRE iI OHA A-15 FIRE PUMP 1/2 RUN Coded Fire alarm 2-2-1 TURBINE GEN AREA -88' ELEV Which of the following identifies how the Fire Protection system has responded, and how should ItheQ~~9ceed?...
~ IONLY one dieseTfire pump ~as started: Enter ~2.0.P-AB.FI.RE-0001, Control Room Fire If{es.ponse. p_lace BOTH U~lt 1an~ Unit 2 CAV In Fire ()utslde Cgn!ro.lA.r~~. _
I§J IONL Y one diesel fire pump-has started. Enter S2.0P-AB.FP-=-0001,-Fire Protection System l~r~J~~~~~~y~~~:0~~~;~~~~~e required due to the cur:~nt ~a:~ilitYOf the Fire
§J I.B.O.T.H. diesel firepu-mpS-have. started. Enter S2.0P-AB.F
.. IRE-OO.O.-1, c.o.-ntrol R.o.om Fire II Resp()nse. Place BO~H_ Uri_it 1 and Unit 2 CAV in Fire Outsi_<le 9~n.!r()1 Are_a.__ _
~ fSOTH diese(fire pumps have started. Enter S2.0P-AB.FP-OOO-1,-FireProtection System iMalfunction. A Unit shutdown will be required due to the current capability of the Fire iprotections.ystem bei,:~ ~~grade<!:
___ ~_..__
IA!1Jwer Iia **1 iEx~l1ILevelll§.-=*.~llcogijitt'J'e"~eieLllApplj.c;ation
_.JIF~Cillty:l!Salem.1m~?___
1~~:Dat&zl 12/3/2012 Iplant$ystems.
-. --.~=~~-_
~-":"'....:...J...... 2l ~~J~!,qtApll 21 I08Ei09Qj>,2Q~ J r
- 086 lFiC~ protection§.l'st~.m__
171
- 8.2.:.. IIAbility to (a) predictihe impacts of the following on the Fire Protection System and (b) based on those ler~<:iiction~use procedures to.co~r~.c:t,_control, or mJtig§lte the co~~.eguel1_c~sof those abnormal operation:
[j\\g,gi -, llnadvertent actuation of ~h~EF'S~ue to circuit failure;'or weldfQg~~'
[2:~
.E~p,~natl?nof 155.43(5) A deluge valve opening as stated in stem will cause FP system header pressure to lower to the-I "niwer";*** IPoint that #1 Fire pump will start at 85 psig, and will restore header pressure. Each Fire Pump is rated to isupply all fire protection needs. The second Fire Pump will NOT start, as its auto start pressure is set at 75 psig. When the deluge occurs, the CRS will not know it is inadvertent. The CRS will respond to the auto start of the pump lAW ARP for A-7 FIRE PROT FIRE and A-15 based on the deluge valve opening.
This directs implementation of AB.FIRE, which checks if the CR is affected, then directs placing CR ventilation in fire outside the CR AB.FP is NOT entered, because there is no indication of a malfunction, but indication of a valid deluge valve actuation. The shutdown distracter is plausible because it is the action required in ARFP if both normal and backup fire protection systems are unavailable.
[Control R00t11 Fire Response. =___ _
_J
[fire Protection Systern Less~nPlan I
[Fire Protection System Malfunction.
LearnlngObje'ctives
,~,,,;:,.~ "<<.~;.>~j;~£~l~"*l:lL,-,,~,
Describe the function and operating characteristics for the following Fire Protection System components:
Fire Barrier Components:
Fire Doors Fire Dampers Penetration Seals Fire Proofing Marinite Wails Thursday, October 11, 2012 3:32:57 PM Page 26 of 35
Energy Shields Protective Wraps and Coatings b,
Fire Detection Devices:
Ionization detector Thermal detector Smoke and Fire detectors
- c.
Fire Protection Subsystems:
Water Supply System Preaction Deluge System Wet-Pipe Sprinkler System Foam System Carbon Dioxide System Halon System d,
Fire Header Pressure Switches
]
Ii I
llQuestiOIl ModifiClltionlVlethod:
Thursday, October 11, 2012 3:32:57 PM Page 27 of35
---~---
- Given the following conditions:
Unit 2 has experienced a LOCA.
21 RHR pump has been CfT for the last 2 days.
Containment pressure is 14 psig and rising slowly.
22 RHR pump trips while after performing Safeguards Reset actions in EOP-LOCA-1, Loss of Reactor Coolant.
The CRS transitions to LOCA-5, Loss of Emergency Recirculation.
i-The STA reports a valid PURPLE path on Containment Environment with containment I
pressure at 15 psig and rising slowly, and no higher PURPLE or any RED paths present.
.Which of the following describes how the CRS should use Containment Spray pumps?
lThe_~RS sh~ld start!~t~p Contail']ll1~nt.§.pray.Eumps in.....
~ IEOP-FRCE-1, Response to Excessive Containment Pressure, because Containment Spray 1 lE~p~lJ',fi~_~~_auto start_with the.?_ECs reset
~~.EOP-FRC.. E-1*t)ecausea1Ie
.... ne Containm. entspra.y pump is.*. r.e.quin~d.t.o be running
'1
.. ast. o llJ',fheneve.r containment P!~~s_~~~ !§>~'1~.psig..
I
[£;] [EOP-LOCA~~ to establ.ish minim~~~~9uJr-.~cLg~S flowjn-o~der to-c_onserve RyY_~T inverl!o_rY..
@J,EOP-LOCA-5 since FRPs-are not in-effect when LOCA-5-is In effect-~-
L ~
~_ ~
fA.nswer.1
. -J IExam Le\\(ell ICogllitiV~l~v~~ I [M*~rl1Q.ry.
tfl!!cU,ity~1 ~ale.rn 1 & 2 IEXam Date:!'.
12/3/20121
[Plant SJ's~errl~.
.1 IROGrQuP! [JI I~RO'qfOl!pl r. _1:
[1 Q;3_00_0f31~~..
[10*3.
IfContainment§Y§1ern 18i l2-4~J[Emergen'cyPro~edu!es I Plan L4.4.22J [Knowled'ge *ofthe bases for eri()ritiz_~ng. sClfe!yJul1ctionsduri~£ abn()r,!!~l/em~r~enc¥ operations.
55.43(5) The transition to FRCE-1 is required upon a valid PURPLE path (15 psig containment). A is incorrect because FRCE-1 specifically asks if LOCA-5 is in effect, and if so, direct CS pumps to be operated lAW LOCA-S. B is incorrect because of the above reason, and additionally, using the table found in LOCA-5 as the bases, there are conditions with cont press>15 psig that NO CS pumps will be directed to be started. C is correct. D is incorrect because FRPs are in effect after the transition out of TRIP~1; and there is nodirection tos_uspE;nd FRPs_either in LQg~~1 or LOCA-?
[Loss of Emergency Recirculation IResponse to Excessive Containment Pressure Discuss the procedural requirements associated with the Containment and Containment Support Systems. including an explanation of major precaution and limitations in the Containment and. Containment Supeort Systems procedures LOCA05E006 Describe the basis for each step, caution, note, and Continuous Action Summary item in LOSS OF EMERGENCY RECIRCULATION caution, and note in 2-EOP-FRCE-1 thru 3 and EOP-CFST-1, Figure 5
!MateriatJ~eqUired for ExaminatiQn' ".
I IGt!lestlon Source.;lIPrevious 2 NRC Exams.
Ilaliestioll M(jdifieatiOlfMeth,g!1; UEd~~OriaIlY Modified IQliesti~n$ol.lrceco~ments:* Ilvision~77740m~dl~ied to ask 'PJhat pr~cedure O~~01~~C Exam
_ J
[Given the following conditions:
i Unit 2 is in MODE 3 preparing for a startup.
j-21 RDMG set motor AND generator breakers are closed, and BOTH Reactor Trip I Breakers A and B are shut for rod control testing.
IWhich of the following identifies how many Reactor Coolant loops are required to be in operation
!IAW Salem Tech Spec 3.4.1.2.c, Reactor Coolant System, Hot Standby, and correctly reflects its
[Bases?
~ ~~~~9i~:~:ause single fail~r~_~onsidera~~_ns require aIIIOO~s~~_ oper~tion when r~d_.c:ntrol is... j
~ r6ne~ because it"is sufficient to provide positive pressure control of the ReS with a bubble '
l~st~blish~Q in th~F'_Z~
§JFour, because potential energy additions from tile secondarysystem'require 4 loops to absorb-I lt~~~energ'y to pr~~ent exceedin_g th.~.!~rnits of Ap!?~.!lQ~x_9 to_1qCFRP~r!.50., __.
@J lOne, because-it provides adequate flow to ensure -rnixing, preve-nt stratffication, and produce Igradual reactivity changes during boron concentration reductions in the Reactor Coolant I
l?!.'.~tem.
IAnswerl ~'- 1 'EX~fl')leveq Is ICQgrt!jlv~teve~ I [rv1~1T10_ry_ ---
lifatHitv:[~ate-m 1 & 2 ----]liSi$ari}"Dafe: [
12/3/2~_121
!Generic Knowledge and Abilitiis
__ j IRQ Groqp Il ISROGrSll.tpl r' 11 1194001G132 "1
lGEJi~~RIC "!
19]
!2.1.. ' ![Conduct Of Operations
]2j,3i=-I-rAbility't~-~X-Plain and ap~y~alls-ysterll_IT~jts andpr~c_autions..
55.43(2) All of the choices have their reasons pulled from the Bases section of Tech Spec for RCP L:;.;..;~~-,,-,,-~ 1operation. The conditions in the stem indicate that Rod Control is energized. With rod control energized, i4 RCPs must be in operation. As per the Bases on page B314 4-1, it is for single failure criteria. B is
'incorrect because 4 loops are required. C is incorrect because the secondary system heat concerns are
'Ifor starting a RCP with RCS cold legs <312c F. 0 is incorrect because it is one RCP, but has the correct bases for when only one Rep is required.
Tech Specs_
[Salem Tech Sp~cs Bases I
LOR Given a situation dealing with Reactor Coolant Pump operability, examine the situation and apply the appropriate Technical Specification action. (License Operator and STA only)
NCT State the Technical Specification associated with the component. parameters and operation of the Reactor Coolant Pump including the Limiting. Condition for Operation(s) (LCO) and the applicabiiity of the LCO(s) (Non-licensed Operator)
Il\\Ilaterial~!!IqUiredfor~arninpti~rr' I IQue~ti,(i~~ource;.IIFacility Exam Bank
- QU$Stli>I1,Moplficati ISjgnificantlY~Odifjed IQ~esti()n*{30l!rCee~n,q:lE:mts: LI r0'27905Question originally asked why all RCPs had to be running with energized rod control. Modified so that Icandidate had to determine rod control is energized, then decide how many RCPs had to be running, in addition
[to bases.
Thursday, October 11,2012 3:32:57 PM Page 29 of 35
Given the following conditions:
Reactor Engineering (RE) contacts U2 Control Room at 0630 and reports that a shipment of new
(
fuel was scheduled to arrive at 6 a.m.
RE reports the fuel has not arrived and the shipping company is unable to provide a prompt determination of where the shipment is.
'[WhiCh of the following identifies the most restrictive reporting requirement for this event?
~
is require~d..to be ~m~de by
.~ ~ ~~~~~
II [~:.§~~Jjg~E_~E?~'?rt~_Q?Qg.~~__._._~~........
I] IA One Hour 0730.-~*~*-*~
III
~
077417 changed from 4 different reports 1,4,24 and UE, to 1 or 4, and added the required report by time.
Thursday, October 11, 2012 3:32:57 PM
10f the following, which identifies who will approve the installation of a Temporary Configuration iChange lAW CC-AA-112-1 001, Attachment 5, Temporary Configuration Change Package llnstallation?
L~"m
..._~______
~ 1?y.~I~~I\\t1a nas~!(s)foraffecte~syster1'1{s).~~'-- "
___.""~ __
[EJ l~ite Er1.9ineeringg!r~G!or.
[g IOp~~~ti~r:!~_ ~anage.r.
r.:i-' "r'~'~'~~--. -"
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6 Shift M~r1Clg~rJ~RS..
IAnswer**1 IlExaffi;levet! i§-" IC09~itiv~Leve!llryiemory HFacitlty:lrS~lem 1 & 2 _
IIEx~mpate:l 12i3/2012~
[GenericKnowledg~~and Abilities I[RO /3roupll.
ISROGr!)ltp II 11
- 194Q.OJ G211.J 21!
[~EtiE..RIQ ]
--,-..I
[2.2.I[EguipmE:nt CO,l1trol".......
~_
[2.21J 'J rKno~e~ge of the pro'cE:s,sfor controlling-t~l}1porary desig~_chang~s:._ "
~
55.43(3) CC-AA-112 and associated T&RM delineate who approves Temporary Configuration changes.
'--'-_-'--.....;;;...~!A is incorrect because the System Manager (8M) performs a review of the TCCP. The 8M is also responsible for reviewing the limited duration requirements for temporary changes installed per Maintenance Rule (a)(4). SM is responsible for assisting the RE with post installation testing. 8M is also responsible for ensuring that TCCP Extended Installation Justification is approved. B is incorrect because, the Site Engineering Director (SED) only has overall responsibility for the Temporary
~<:l~figuration Change Pr()9!Cim. C is incorrect because th,e ()p~_~<:l~ager is not required to<:lp'p~()ye~.,..CCs..
[Tei'Tlporary Configurat~on.Change Imple~~ntation-T&RM..
r.
[
Thursday, October 11,20123:32:57 PM]
Page 31 of 35
IDuriflgnormal operations in MODE 1, -it-is-discovered that 2A ED(fmonthly-s-urveillance was-not Ipetiormed within its 31 day required periodicity.
The required surveillance was last performed 33 days ago.
jWhich of the following identifies the status of 2A EDG lAW Tech Specs, and why?
I 12A EDG
______~
is...
L....
_.~
_____ ~_.,,__
~ NOPER~_'=l::__bec.ause it has exceeded its 31_~§i~urVeill~!1~~-.!equirement.==-_=__.___ _
__ J
[] NOPERABLE since the 24-hour delay til11~~st!Q~_31 ~ayr~9uir~rllef'lLh~~ been.. E:;~<?ie~esJ:
~ IO-PER'ABLE because the normal surveill~nce inte'rvalPlus 25o~'eXtensionhas not been iexce~ded.
@] !OPERABLE-because the surveillance canbeperformed-within-the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> delay time which I
starts L.!E,<?f'l_~is<?9_vi:ryof the mi~s~~~':.l~eillance.__,
_I iAn$W~r>llg nIIExa~Lev,e_!1 '8 IIC09nitiye,Levef IIApplic:ationH~aciHty:llSalem 1& 2
!Exam Dat!l{1 -_ _
1213/2012!
IGene!!cKnowledge and Abilities.
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22 r2.2 Ii ~quipment Control li.7:~i IIj\\bility_to determine operability and/oravailability of safetyr~lated_eql::lipmeni,'-'
-55.43(2) Tech Spec 4.0.2 states.., "Each Surveillance Requirement shall be performed within the I...____""""",,,,,,,,,-.;,;;u specified surveillance interval with a maximum allowable extension not to exceed 25 percent of the specified surveillance interval." Since the 25% of 31 days has not been exceeded, the EDG remains
[OPERABLE, since its surveillance is not required to be performed until 31+7.75 days, A is incorrect Ibecause of the 25% time. B is incorrect because the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> delay time is not applicable until after the 25% extension expires, and since it is a 31 day frequency could be allowed to gq to 31 days (per Tech Ispec 4.0.3) 0 is incorrect because the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> delay time is N/A first because the allowable time would
~e longer,and second because it is not ap~icable yet.
ISalem Tech~pe~s.._
I
[
Describe the term Surveillance Requirement as it applies to the Technical Specifications TECHSPE014 I - Describe the gene~ral requlrements associated with Specifications 4,0.1 through 4.0,5 relating to implementation of the Technical Specification Surveillance Requirements am Bank IIQuestl()!}'Modlficatio~M£!thO~:,' I Concept Used
'-'--'--______._'-' lVision 60~76 made intoa spe~ificoperability question,
[. Thursday, October 1 f 2012 i32:5i'PM i
32 of 35
iGiven the following condition:
I 1-Unit 2 was manually tripped to enter a refueling outage at 20:00:00 on January 21 st.
Ilf all other requirements are met, which of the following is the EARLIEST time that movement of Jl:lel in!he,Rx vessel could oC<::_lJ~If\\\\I\\I Te~h Se~cs?
§;] [Q~-OOon Jclnuary25th.-
~
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- b. t9QQQc:m ~~,l"l~~'Y1.~th: __
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.._. J[Cognitive Le)lel IIAQl;)li.gl!i-l::>n.--_.lIFa~lit¥:1 iSalem 1 & 2___J IExam,Date: 1__
-12i3/20121 IGenE,rjc_Knowledg!3_~niL~~Uties.
I~OGroup 11-11 ISROGiovpl [ ]
[19~Q()1G3~13
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rR;~,";dji;;;:;;i~;:::":,
[2.3.. -!iRcldiation Control 12,3J 3~J [KnOWledge of radiological safety procedures pertaining to licensed operator duties, such as 13.81
!response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, i
_access tolocked higl1-radiation areas, alignit'!£t filters,_ etc, i
~,.,,--...,.
155.43(7,2) TSAS 3.9.3.a states that for refueling outages between Oct. 15-May 15th, the reactor shall be
'--~~='---I !subcritical for 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />. 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> from 2000 on January 21 is 0400 on January 25th. Per tech Specs IBases. the minimum requirement for Rx subcriticality prior to movement of irradiated fuel assemblies in the Reactor Pressure Vessel ensures sufficient decay time has elapsed to allow the radioactive decay of the short lived fission products. The 80 hour9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> decay time (LAR S08-01) is consistent with the assumptions Iused in !he fuel handling~c~i~E!n~ analy§i§.
Discuss the procedural reqUirements associated with the Refueling System, including an explanation of major precaution and limitationsi.~ the Refuel!l!.g S)iste.rn procedures. 3Licensed Op~_ra!or & Non-licerlsed Operator only)
IMatEfriaL~~qUIi:ed.f9r,~.?'aminati-Oh II ility Exam Bank QuestionSoutI:8ComPlen~;
IVision 084937
..w.
Thursday, October 11, 2012 3:32:57 PM. I Page 33 of 35
I An-expiosion~a~nd fire-at the RAP tank area has resulted in a possible large spill o(radioactive water tin the area. An Alert has been declared and all required facilities are activated and staffed. The Fire IDepartment has determined that off-site assistance from the local fire department is needed.
!IAW S2.0P-AB.FIRE-0001, Control Room Fire Response, which choice identifies who must iC!uthoriz~,~~glJe~!l!~£L~ff-siteJire departmenLa~~istance?_
BJ l§_~~uritY~QlJtLS~eelVisor.
~ [f\\J~~T.~a*r Fire. Prote~!ior'!,§Llpervisor..
~ fS-M I Emergency. Duty Officer(ED6f g] [Radiological Ass_essmen~Coordinator (F3~9)_~--'
IAnsv.;er I IIExaw level I [8, llGogriftivEtkeyerllM~rnoJY
_!!Facility: I[saiem 1 &.3._
[Q.~~ljg Kn()wlf::cJ.ge and Abilities
____J IRa Gr<iup 111 Isac)qnmp '-1' 1Exam !pate:l.
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1
-i41
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[24 iiEmerg'ency Procedures I Plan
.. - "1 lL!A2§'] TKnow.ledge of fa.G.ility protection requirements, including fire. brigade and portable fire 'fi9.hting 113.1]
lequipment usag~...
r=------:-------;-:-,-,'S5.43(7)55.43(5) CAS An. 1,Fire Dept. Support,'Caution prior to step 3.0 in AB.FIRE-1 states, "In-the event of a radiological emergency, the Nuclear Fire Protection Supervisor should obtain permission from the EDO/SM prior to calling for off-site assistance." A is plausible because security is required to be notified whenever off-site assistance is requested (CAS 2.0) B is plausible because they will be leading the fire brigade and will be the person to request the off-site assistance through the EDO. D is plausible be~ause during 5:l0 Emergencythe RAC Is associated with the radiologi<::.§llaspect of the emergel1"cy:
Identify and describe the local controls, indications, and alarms associated with the Fire Protection System, including:
The location of Fire Protection System local controls and indications (Licensed Operator & Non-licensed Operator only)
The function of Fire Protection System local controls and indications. (Licensed Operator & Non-licensed Operator only)
The pla'lt and conditions or permlssives required Fire Protection System local controls to perform their intended function.
(Licensed Operator only)
The setp()intsassociated with the Fire Protection System local alarms..: (~i<;ensed Operator only)
[Material Requ!r~d f91' ~~n1lnatlon,,
tQuestl().ri~ooree:, IIPrevious 2 NRC Exams IIQuestionMoqi{~ea~i.oIlMet!'lod:.IID~ect From Sou~ce IQues;~c.)Il~ourcet:omments~ I~al~11'l 08-01 SRO NRC exa (5/17/2010)" "
rT1 Thursday, October 11, 2012 3:32:57 PM
_Page 34 of 35 I
IGiVen the following conditions:~
i-22 cves Monitor Tank was released to the Delaware River earlier this shift.
The release was secured 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> ago during the performance of S2.0P-SO.WL-0002, Radioactive Release from 22 evcs Monitor Tank.
During a sample review after the release was secured, the Chemistry Department I recognized that a Radioactive Liquid Release to the Delaware River was performed with an iisotopic I concentration which exceeded the ECG EAL RU1.3 Unusual Event threshold of 2X the I
ODCM for >60 minutes.
!\\Nh~ch of the following <:!escribes how this should be addressed?
EJ Ilnitiate9~~.Q=-~merg~ncY5m~_.b()ur r~port fELthls Af!er-th_e~~?lct eve-nT~
~ INotifyihe NJDEP withinone hour of the report by Chemistry Departmentthat the release rate _I'.1 le}<ceeded the ODeM.
~
§ iDeciare an Unusual Event-based on exceeding the-EAL at-the Time of theevent, then-.----* -l tte~min~te the UE because the EAL t~~eshold is Q~Loflg_er bel!Jg_~~~~eded._ _
~ iDeclare an Unusual Event based on e-.x.ce-e.d ing
.. the EAL at the time of the eve.n.t.' then retract. -j'
. lthe UE because the EAL ~hreshold~~~ N(:n ~~<:;ee~eq when the declaratio.~was made.
ICognitive Lever IMemorY. _
.. 1!l=acility:.I l.salem 1 & 2
-JIE~mDate:I'_
12/3/20121
,GeneiiGKnowledg~§.l}9f'.bilitLes IIR(),Gt9up! i11IsRQ"9~upl L l194001G42~_1
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[~.4. I~~rnergency Procedures i Plan 1
- -------, roo. -""
!2A,2~J iKnolfl/!e(jge of the emergen<:;), plan r
55.43.(5) Salem ECG, Introduction and Usage, Section 8.6, Conditions Discovered After-the-Fact,
~......_-' describes an after-the-fact event as an event that exceeded an EAL threshold and was not recognized at the time of occurrence but is identified greater than one hour after the conditions has occurred and the condition no longer exists. The stem identifies that the liquid release was terminated 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> ago, which terminates the release exceeding the ODeM. After the Fact events that occur will be assessed and evaluated to ensure that no EAL current applies. An emergency declaration is NOT required and a non Iemergency One-Hour Report should be initiated. The NJ DEP notification is required for spills, not
~ischargesnorm.a!lypE:lrformed.
lSalem ECG_
I
[
IMateria. Requhied fo.fExamincttl()f},.' *tt 1!.
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- =---=-=~ IObjective from VISION is EPTRAiNINGE1: Given EP related iss~es and topics, analyze the issue, classify
- I
""-"--"-'-_-----'--""-'~'-'-""_--'U evet:l!:; and communicate to the State.s._IAW app!:.o\\led statiCl!l2rocedures. Not!~s~ed in dropdown menu.
l Page 35 of 35 'j