ML113330798
| ML113330798 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 08/03/2011 |
| From: | Public Service Enterprise Group |
| To: | Operations Branch I |
| JACKSON D RGN-I/DRS/OB/610-337-5306 | |
| Shared Package | |
| ML110030673 | List: |
| References | |
| TAC U0183 | |
| Download: ML113330798 (100) | |
Text
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In"<:lstionTopic A dropped rod while critical on Unit 2 has resulted in ReS Tavg lowering to 540.9"F, and the Rx remains critical. Which of the following choices contains ONLY reasons why Tavg must be raised to at least 541' for the Rx to remain critical lAW Salem Tech Specs? Raising Tavg to at least 541°F ensures ...
- 1. MTC is within its analyzed band.
- 2. protective instrumentation is within its normal operating range.
- 3. the P-12 interlock is above its allowable setpoint.
- 4. minimum shutdown margin is maintained.
- 5. that Heat Flux Hot Channel Factor does not exceed its limits in the lower region of the core.
- 6. the PZR is able to be operable with a bubble established.
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[~~~/~volut!§~ I,;;;D..;..ro;,.:;p;.:;;p,;;;e~d,;;;C,;;;o;,;.nt;;,,;ro;,;.I..;..R:.;;O,;;;d_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _---1 Tech-Spec limits for T-ave 55.41 (5, 10) The Tech Spec Minimum Temperature for Criticality limit of atleast 541°F (TSAS 3.1.1.4) Bases states that: 1. the moderator temperature coefficient is within its analyzed band, 2- the protective instrumentation is within its normal operating range, 3-the P-12 interlock is above it allowable setpoint, 4-the pressurizer is capable of being in an operable status with a steam bubble, and 5- the reactor pressure vessel is above its minimum RTNDT temperature. Of the 2 incorrect choices, #4 is plausible if the would tend to create higher power lower in' the core, but Fq-(Z) is a condition-which is kept in its allowable range by maintaining I acceptable power distribution limits as described in TSAS 3.2.2.
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~uestion Topic!
Given the following conditions:
- Unit 2 is operating at 100% power. - 21 SGFP trips, and a manual Rx trip is initiated based on lowering SG levels. - The Rx operator confirms the Rx trip during performance of the immediate actions of EOP* TRIP-1, Rx Trip or Safety Injection.
While performing the RCS Temperature Control section of EOP-TRIP-2. Rx Trip Response, the RO reports Tavg is -5510 and stable. Which of tf1l1nwlnr1 describes is -551° F? Reactor trip breaker position 55.41 (5,6,7) A manual Rx trip is initiated by the Rx Trip Handles. This sends a signal to Solid State Protection System Trains to open their respective reactor trip breakers. The P-4 signal is developed by the opening of each RTB. One of the functions of the
~~~~-=-'-'-- P-4 signal is to arm the steam dumps (Train A) and place them in the Plant Trip mode of operation (Train B), to control RCS Tavg at 547". Since the reactor trip was confirmed, at least one RTB opened. B is correct because the B P-4 signal not being present
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programmed Tavg of 547 at no load. A is incorrect because the load reject would arm the steam dumps, and Train~ B P-4 would place them in Plant Trip Mode and control at 547". C and D are incorrect because steam dumps would have to be in MS Pressure Control mode for the setpoin! on the control console to be the driving system Signal. Both C and D are plausible if candidate does not know normal mode of steam dump operation at ower, or function of Tav Control. L.Q.ri,jumber i' . J ITRP002E005 I ISTDUMPE008 I ISTDUMPE007 I
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'Question_i~~ IRO 3 Given the following conditions: - Unit 1 is operating at 100% power. - Spent Fuel Pool fuel moves are being performed lAW S1.0P-IO.ZZ-0010, Spent Fuel Pool Manipulations. - PZR spray demand lowers. - PZR level is rising very slowly. - The CRS enters S1.0P-AB.PZR-0001, Pressurizer Pressure Malfunction. - The crew identifies 1PR2, PZR Power Operated Relief Valve, is leaking.
Of the following choices, which is REQUIRED to be performed FIRST lAW Salem Tech Specs? Assume 1PR2 cannot be restored to an OPERABLE condition. Within one hour close 1PR7, PORV Block Valve, and maintain power to the valve. 1PR7, PORV Block Valve, and remove power from the valve.
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[SysterrifEV()lutionTitl~.1 IPressurizer Vapor Space Accident lJS!.. Statelll~~(:J ,...-_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _---; Knowled e of less than or e ual to one hour Technical Specification action statements for s stems. 55.41(10) A is incorrect because there is no requirement to stop moving fuel for this condition, nor is there a general caveat that says stop moving fuel for any Abnormal Procedure entry. B is incorrect but plausible, since it is the 3.0.3 action when there is no specific TSAS which applies, and fits with the other one hour requirements. C is correct. and D inconrect, because power is maintained to the block valve lAW TSAS 3.4.3 action a for excessive seat leakage, whereas power is removed from the block valve
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[questi;n Topicl IR04 - Outline Chanties r Given the following conditions:
- Unit 2 is operating at 100% power.
- 23 AFW pump is CIT.
- A 1,000 gpm LOCA occurs.
- The Rx is tripped and a Safety Injection initiated.
- NO AFW pumps automatically start.
Assuming that both MDAFW pumps could be manually started if demanded, which cf the following describes the MINIMUM required AFW pump status, and the reason wht that minimum is required lAW 2-EOP-LOCA-1, Loss of Reactor Coolant? [~ lONE MDAFW pump must be started to ensure a secondary heat sink is available. I TWO MDAFW pumps must be started to ensure a secondary heat sink is available. I I I.ONE MDAFW pump must be started to provide a static head of water to prevent primary to secondary leakage. I
@.J ITWO MDAFW pumps must be started to provide a static head of water to prevent primary to secondary leakage.
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ISmall Break LOCA Manual ESFAS initiation requirements 55.41(4,10) A is correct because one MDAFW pump will provide at least 22E4 Ibm/hr flow which is the minimum required to satisfy the Heat Sink CFST until at least one SG NR level is restored. Additionally, for a small to intermediate sized LOCA, this flow is required to ensure a secondary heat sink is available. The distracters all contain either 2 pumps or the incorrect reason, or both. The distracters which refer to the static head of water is ,the basis for a LBLOCA, , not a SBLOCA. The distinction
. must be made .that Iplausible because two-AFW pumps are required-during FRSM when >44E4Ibm/hr AFW flow is required.
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OHA D-22, 13 RCP BRG CLG WTR FLO LO, received 2 minutes ago. I Memory I Salem 1 & 2 9/26/20111 D 55.43(3)C is incorrect because the loss of charging pumps, and hence seal injecticn flow, is not a RCP trip criteria, since thermal barrier flow is still established. AS.CVC will not direct stopping the RCPs either. A is correct because #1 Seal Leakoff flow of greater than or equal to 6 gpm is RCP trip criteria based on Alt. 1 of AB.RCP. D is incorrect because the ARP for D-22 does not direct entry into AS.RCP unless accompanied by rising temp. B is incorrect because the setpoint for high seal outlet temp is o
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if RCP seal inlet temperature rises to 225 To prevent runout of a charging pump when a charging pump is started. The RCP seals and shafts may be damaged due to thermal shock when a charging pump is To prevent steam binding of the charging pumps due to RCP sealleakoff flashing to steam when a charging pump is started. The RCP thermal barrier heat exchangers may rupture due to thermal shock and water hammer when a charging pump is started. l~*am.L~i!u ISalem 1 &2 concepts as the apply to Loss of Reactor Coolant Makeup: Consequences of thermal shock to RCP seals EXP1~I'i~t~~ri()f'i 55.41(10.7) AS.CVC provides guidance in Att 3 and CAS 5.0 to isolate Rep seal inection if any seal inlet temp is greaterthan or
,Answers+/-;*;):i.'J equal to 225°F. This guidance is consistent with the guidance contained in LOPA-1 basis document. page 32. which states,
- - - - . "Isolating the RCP seal injection lines prepares the plant for recovery while protecting the RCPs from seal and shaft damage that may occur when a charging pump Is started as part of the recovery. With the RCP seal lines isolated. a centrifugal charging pump
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- but plausible since starting the first pump in a system has potential for runout after i..igh points have drained back into system. Cis incorrect but plausible since the high temperature in the sea! area on an extended loss of seal cooling may cause flashing in the seal area. and the gas could be transported back to evcs systern. but is not why the seals are isolated. 0 is incorrect but plausible because the CCW thermal barrier return is isolated for this reason during a LOPA.
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lQuestio~T()8~ IRO 7 I Given the following conditions:
- Unit 2 is in MODE 4 exiting a forced outage. - The reactor has been shutdown for 96 hours. - 23 RCP is in service. - 21 RHR loop is in service supplying shutdown cooling. - 22 RHR loop is aligned for EGGS. - RHR HX inlet temperature is 290°F and stable. - RCS pressure is 325 psig and stable. - ALL MS-10's are set for 200 psig in AUTO.
The NCO trips 21 RHR pump due to indications of cavitation. With NO operator action, determine how long it will take for MODE 3 will be reached? Assume only heat added to RGS is decay heat. References provided. [EJ 1< 15 minutes. I [EJ 115- 19 minutes. I 11M 120-25 minutes. I
@J I> 25 minutes.
I~~we;', fO' !Exl'!m Level! !~ §~!]J!iye~ey.~*CJ IApplication 1000025K101 EX, P:,',I,~n,',',a, [,Answers: ti" ' 55.41(10)The HUR will be determined using Attachment 5 of AB.RHR-1, page 2, with PZR level (Reps are running). The 96 hour mark is 4 days, and the HUR will be -3.8- 3.9°/minute. 60 degrees of temp change are needed to get from 290 to 350 (Mode 3) 60/3.8=15.8 minutes. 4°/minute would yield 15 minutes, but 4°1 minute is clearly above where the lines intersect on the graph. If the after core reload line is used this was a forced outage, not a refueling outage per stem), then the HUR would be 2.7/2.8 deg/min,
. _ ., 2 mjolltes If pace 3 HIIR is used then the H11e ,Mould 1 * *
- candidate uses the first page of attachment without checking to see if it is for the right conditions as stated in the stem, then they would use -0.31° which would give over 3 hours. Modified distracters to make all plausible. Provided Attachment 4 in addition to correct Attachment 5 to make question harder.
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Unit 2 is in MODE 3. NOP. NOT. returning from a refueling outage. 21 charging pump is in service.
- ALL CCW pumps trip due to a faulty electrical relay which was replaced on all CCW pumps during the outage.
Which of the following identifies an action that must be performed lAW S2.0P-AB.CC-0001, Component Cooling Abnormality, and the correct reason it must be Isolate letdown and swap charging pump suction to the RWST to prevent Place 23 charging pump in service since it is cooled by SW and both centrifugal charging pumps are cooled by CCW. Initiate Surge Tank makeup as level drops due to system pressure decay to ensure air is not introduced to CCW pump suction. IMemory t!Q>.:11000026G132 It1:~~_. [R~~§ ~B9Ya'ue~fIB ~~jl~
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to explain and a I all system limits and recautions. 55.41 (4.7.10) A is incorrect because while RCPs will be stopped, it is because them is no bearing cooling flow. The seal package is c...::;=~:=2l:oLJ still receiving seal injection flow from charging system. B is correct because in a total loss of CCW scenario where the loss of cooling to the letdown and seal water return heat exchangers (in addition to the loss of cooling to the Thermal Barrier Heat Exchanger) leads to a rising VCT temperature. which ultimately results in the loss of RCP seal injection, and failure of the Reactor PDP has CCW cooling. so no pump swap would be requireid in 'this case, but would be if the PDP (23) was in service. 0 is incorrect because surge tank M/U is initiated if a leak is present, not for a loss of all the pumps. The stem states the pumps tripped from an electrical fault, not because a system leak made them trip. I ~~~ I New IIQuestion Mv", .,.vau" Method: . 'I ~c:tLJestion~ource Comments 1 I .Comment I I I I
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- ~~:~~o_n T~i:J I_R~o~9 Given the following conditions:
Unit 1 is operating at 100% power.
- 13 Main Turbine Governor Valve fails shut.
The PZR Master Pressure Controller output fails as is at 2235 psig before any response to the Governor Valve closure occurs. Which of the following will occur naturally in the Pressurizer to help limit the magnitude of the resultin res sure transient on the prima system? An outsurge cools the PZR. This allows some steam to condense to water and limits the resulting pressure increase in the RCS. An insurge of hotter water heats the PZR. More liquid then flashes to steam helping to limit the resulting pressure drop in the RCS. An outs urge causes the steam space to expand in the PZR. This allows some liquid to flash t*:J steam and limits the resulting pressure drop in the RCS. An insurge of cooler water compresses the steam space In the PZR. Steam is condensed to water helping to limit the overall pressure increase in the RCS. IMemory Salem 1 & 2 [!<A: II 000027A 103 I~* * *~ [R(})f~ @i~~~9~
.:Si'stel11/E§~I.IJipn.l}itl6l IPressurizer Pressure Control Malfunction a pi to Pressurizer Pressure Contro! Malfunction: ~.EX. ,p.*.*. *. .I.~n.~tio.*I1 *.0.* * *.~. .,f. :.* 55.41( ) A is incorrect because the outsurge cause lower pressure in the PZR, which causes more flashing to restore pressure. B IAnswers:. . .... il is incorrect because an outsurge occurs, and if an insurge occurs it would be of colder water, not hotter. C is incorrect because an I
insurge would occur, but the action is correct. 0 is correct because the governor valve closing would cause an insurge based on the rising RCS pressure from the load reject. The insurge compresses the steam bubble, which cause condensing of the steam in the ~~rrt1)eF ~-c~.~". IABPZR1E001
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~~~..'..':'!'.'.'::J IRO 10 Given the following conditions:
- Unit 1 is operating at 100% power.
13 charging pump is in service.
- A manual Rx trip is initiated after 11 BF19 fails shut.
- The Rx does not trip, and can NOT be tripped from the control room.
Which of the following would be an expected control console indication for the Charging Pumps after the Rapid Boration Initiation steps have been completed in 1-EOP-FRSM-1, Response to Nuclear Power Generation? Assume Safety Injection is not actuated. 13 evcs Pum START, START, START. IAnswerl D E:=;;2C::~ lKA:l I000029A204
~~m'EV:oluti.&~'.Ti!fe.1 IAnticipated Transient Without Scram 55.41(10) D is correct because FRSM-1 has operator start 11 and 12 charging pumps. 13 charging pump is normally operating. All three pumps will indicate START. A is plausible if the candidate thinks the procedure starts ONLY one centrifugal charging pump, and stops the PDP. B is plausible if the candidate sees 13 PDP is in service and tllinks the procedure just needs at least one centrifugal charging pump in operation. C is plausible if the candidate thinks the procedure directs starting of either 11 AND 12
. !jres stoo0100 13 cbamino ollmo Stopping the operatiog . . .
I stoppedin other EOPs <imd ASs." - - _. .
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Given the following conditions:
- Salem Unit 2 is performing a Rx startup by control rods. - The reactor is critical at 3000 cps in the source range. - Control rods are withdrawn to raise power. - No additional operator action is taken and a reactor trip eventually occurs.
If a Source Range Nuclear Instrument Discriminator voltage was lost immediately after the rod pull was complete, would a reactor trip occur sooner or later than if the loss of voltage did not occur, and why? later. The recovery period of the instrument following the rod pull becomes longer. sooner. The recovery period of the instrument following the rod pull becomes shorter. sooner. The gamma level is added to the neutron level since it is not being discriminated out later. The gamma level is deducted from the neutron level since it is not being discriminated out. Iy to loss of Source Range Nuclear Instrumentation: 55.41(1,6) SR discriminator "screens out" gamma radiation at low power levels because it is not proportional to Rx power. If this voltage was lost, the SR NI would see all those extra gammas, and would indicate a higher power level. This would bring the SR NI to the trip level sooner than if the voltage had not been lost. 0 is incorrect because the gamma radiation is added, not deducted. A is incorrect because the trip would come sooner. The Pulse Amplifier has 3 functions: to amplify the signal output from the The 8F3 Proportional Counter used in the SR produces pulses proportional to the i3nergy of the ionization reaction in the chamber. A neutron will react with the 810 in the BF3 gas, to produce a large pulse. A gamma will react with the gas to produce a smaller pulse. The Discriminator portion of the Pulse Amplifier acts as a gate, to pass the large magnitude neutron pulses. and to reject the smaller amplitude gamma pulses. Because of its design, it will eliminate any pulses smaller than a neutron pulse, and so removes noise pulses. too.
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- Reactor power is 11 %.
- The low power trips have NOT been blocked.
Intermediate Range (IR) Channel I N35 previously failed and was removed from service lAW S1.0P-SO.RPS-0001, Nuclear Instrumentation Channel Trip / Restoration. Channel II N36 fails. The reactor will NOT trip, 1N36 indication will drop to zero. The reactor will NOT trip, but OHA F-25 SR FLUX HI will annunciate. Both Source Range channels will automatically energize, and the reactor will trip on high SR flux. I000033A201 1iAA2.01--! [~~Q]"~~@~IBlI~ §QG"()IJP:lU~2.~TI[Ji ISyst~ri11Ej,ol~L~ 1 Loss of Intermediate Range Nuclear Instrumentation
~m~ Abili to determine and inter ret the followin as the a pi to Loss of Intermediate Range Nuclear Instrumentation:
Ecuivalency between source-ran e, intermediate-ran e, and ower-ran e channel readings
~i~nati~l"1()fl 55.41(7) This is a valid KIA match because the candidate is required to understanc the relationship between where power is in the Answers:i<).,*;:*1 power range, the level at which automatic re-energization of the SR channels would occur from the IRNl's failing low and/or having 1** _ _ _ _ _ _ _ their bistables tripped because of the channel failure, and what the significance of the 11 % power in the power range has on NI automatic operation. This question requires knowledge of what levels in the 3 nuclear instruments would be present at 11 % power will read if they fail low and how that affects the SR instrumentation, and the P-10 permissive, which automatically blocks SR energization with Rx power >10% in the power range. Loss of Instrument power (high voltage DC 300-1500) to N36 will cause its indication to go low. The rx will NOT trip, because the automatic energization of P-lO permissive at 2/4 PRNI at 10% will already have occurred. (Stem 11 % power). This will prevent the SRNls from automatically reenergizing when the second IRNI channel lowers less than 7x10-11Amps. F-25 will not annunciate, and if it did it would trip the Rx.
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- Questhm Topic II RO 13 the following identifies why the ReS is depressurized during the response to a SOTR lAW Allow more rapid boron injection and refill the PZR.
Allow more rapid boron injection and minimize subcooling. Terminate primary-to-secondary leakage and refill the PZR. Terminate primary-la-secondary leakage and reduced secondary planl contamination.
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RO SkyScraper ISROSkVscraper '1 i RO Svste~IE~~I~i~n list:.' *.****SRO svsIJfuiEv~luik>~List '.1** Outline Chimqes I r - - * * * - - ._ .... - I Cluestion T?Pic:J RO 14 Which of the following identifies an automatic action and its correct initiating coincidence which will o,::;cur if a Main Steam piping rupture occurs at Steamline Isolation. 1/2 High Steam Flow on 2/4 SGs, with 2/4 RCS loops 2/3 detectors on 1/4 SG 100 psig lower than corresponding detectors on 2/~i remaining SGs. 2/3 detectors on 1/4 SG 100 psig lower than corresponding detectors on 1/3 remaining SGs. Knowied e of the interrelations between Steam Line Ru ture and the followin : 55.41(7) Main Steamline Isolation occurs as described in A. B is incorrect because the High steam Flow coincidence is 1/2, not 2/4. C is incorrect because there will be no Safety Injection on SG DIP because th'3 rupture is downstream of the MSIVs and ALL SGs will have lowering pressure. C COincidence is correct. D is incorrect for the same reason as C, and additionally has the wrong coincidence for the remaining SGs.
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IWhich of the following describes the reason Main Steam dum ps will be blocked from opening if Main Condenser vacuum degrades to 20" Hg? To prevent damaging the condenser on overpressure. To prevent further degradation of condenser vacuum. To prevent exceeding condenser design temperature. To prevent boiling circulating water which can damage condenser tubes. Loss of steam dum p ca abilit upon loss Of condenser vacuum 55.41(4) All of the distracters are things that can occur on a loss of vacuum, but am not the reason for having Control Grade Interlock C-9 blocking sleam dump operation. The condensers are not designed to operate at any pressure, and steam flow from L-_~,~.~,~---"-~ the steam dump system (as well as a Main Turbine trip) occur at 20"Hg decreasing, since at this level it is clear that there is something significantly wrong with the condenser and it must be removed from service. I rL;9:Number "',0 IABCONDE004 ~ateriaIR~q"lired fot @amination, I IOII""'~I9,?SOUl'ce: JIOther Facility I ~~n~odifica~ion.M~thod::JI Editorially Modified IQu~stion Sourc~EomrnentsJ ISeabrook May 2003 NRC Exam iel""'" ....... ' ' ' . ."'.'..",>. ...!::' " ...., I I I I I I----------------------------------~I
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Given the following conditions:
- Unit 2 is operating at 35% power with both SGFPs in service. - BOTH SGFPs trip simultaneously.
Which of the following describes the actions required in S2,OP-AB,CN-0001, Main Feedwater/CondElnsate S stem Abnormalit , and wh ? Trip the Rx and start all AFW pumps to maximize SG inventory. Trip the Main Turbine and start all AFW pumps to maximize SG inventory. Trip the Rx to ensure only decay heat and RCP heat are being added to the RCS. Trip the Main Turbine to ensure only decay heat and RCP heat are being added to the RCS. Reactor and/or turbine trl ,manual and automatic 55.41(10) In many cases, when a condition arises which calls for removal of steam flow, the Main Turbine would be tripped if Rx power is <49% (P-9). The Loss of Main Feed is NOT one of those cases, since SG inventory would quickly be depleted without the removal of steam demand from the Main Turbine. For a loss of fBed, the rx is tripped to ensure the only heat being added to the RCS is decay and RCP pump heat, which minimizes the amount of heat removal required from the SG's, and allows AFW system Inot possible nor reguired, since the 'procedu're knows they will already be running. fR~
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RO Sk'vscrap~rlsRO SkV~~~~p~~ L*', ROSvstemfE~olution Ust<1 'j'SROSvst~rrtJEvolution List IOutii~{Challaes [§~~tionTOr:i5J IRO 17 Given the following conditions: Unit 1 is operating at 100% power.
- All 3 Chillers have power, and are operating as expected for temperature conditions. - 12 chilled water pump is in service. - A loss of all off-site power occurs.
Which of the following identifies how the loss of off-site power will affect chiller and chilled water pump operation? Chillers will have their 460V breaker closed b their res ective SEC, and Chilled Water Pump s will be operatin to operate and I or monitor the foliowin HVAC chill water pum and unit 55.41(7) The SECs will shed then reclose the chiller 460 volt breakers on a MODE II (Blackout) SEC initiation following the EDG start. The SEC will close the Chiller 460V breakers and start both Chilled Water P'Jmps sequentially to satisfy the Chiller start permissive. This is unlike the ECAC interlock start, where the standby pump (11) starts only if the dedicated pump (12) fails to start.
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;.Cluestion Topic IRO 18 I
Given the following conditions:
- Unit 2 is operating at 100% power. - The 125 VDC control power normal supply breaker to the 2E 4KV Group Bus trips. - Prior to any action being taken in response to this breaker trip, a 200 gpm LOCA occurs.
i- Operators respond lAW the EOP network. Which of the followin9 describes how this 125 VDC breaker failure will affect the mitigation of the LOCA? IOperators will have to locally trip 22 RCP breaker at its cubicle to stop forced RCS flow when RCS pressure lowers to 1350 psig in EOP-TRIP-l 1 Rx Trip or Safety Injection, , I.operators will not be able to establish/maintain saturated PZR conditions if required in EOP-LOCA-2, Post LOCA Cooldown and Depressurization. I loperators will be required to adjust AFW flow to 22 SG lower than that supplied to the remaining 3 SGs in EOP-TRIP-1. I 9/26/20111~ [~ I.operators Will,~ot be able to restore normal charging and letdown in EOP-LOCA-2.
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[MStaternent'J _____ ~' _'_M'; Knowledge of EOP mitigation strategies. E?<P,I~r1'rtiOr1:6~ i,Answers:" 55.41(10,4) Control power for operation of 4KV busses is supplied from the 125VDC system. A normal and emergency supply are provided, and must be manually transferred when the emergency supply is required. The 125VDC control power supplies power to
',:'1" the TRI P coils of the 4KV breakers, and the breaker cannot trip without that power. However, the 4KV group buses are supplied from the APT (Main Generator output) wihen the unit is operating at 100% power. The same lack of control power will prevent the
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.. ,. nrc"l",.r fr,,;'" <:btirm P"",or "nn"t shut, and the bus \Nill become deenergized. This will cause 22 RCP to have no power, even though its 4KV breaker remains shut.
Loss of forced flow in a RC loop cause less heat transfer in that loop, requiring less AFW flow to supply less steam flow. Distracter A is incorrect because forced flow has already been lost in 22 RC loop AND ReS pressure will not lower to 1350 requiring all RCPs to be tripped. B is incorrect because PZR heater busses are powered from E AND G 460V buses, and 1 group of backup heaters remains available if required for PZR heatup to saturation. D is incorrect because letdown is not restored in LOCA-2, although normal charging is. This is Elausible if the correlation between PZR heaters/leveilletdown is confused. tL.iI {~},;;" Ret~renj:eTit'~' *,'~}~j..:nc,;l ~,,; FacilitYReferel1cet-,ll.lrnbe(;.. ,.* [Ref~re~!;eSeCti9r1~ ipage~o;jl ~ViSiO~ 12 Unit 125 VDC One Line 11223720 II II I 133 I I Rx Trip or Safety Injection II EOP-TRIP-1 -.JI II I 127 I I Post LOCA Cooldown and Depressurization II EOP-LOCA-2 II II I 1 25 I
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~~t!on T;'§J The operating crew has entered S2.0P-AB.RAD-0001 , Abnormal Radiation, due to RMS alarms on the plant vent. Operators have shifted both the Auxiliary Building and Fuel Handling Building Ventilation controls to HEPA PLUS CHARCOAL lAW their respective procedures.
Which statement describes the change, if any, in the release rate resulting from shifting the ventilation systems, if the problem is a stuck open relief valve on Waste Tank None - the WGDT relief valves discharge to the plant vent. The release rate is reduced by passing through the Auxiliary Building Ventilation System. The release rate is reduced by passing through the Fuel Handling Building Ventilation None - the WGDT relief valves discharge to the Auxiliary Building Ventilation exhaust fan suction plenum.
~;II 000060K202 rsystemiEV()luti()n;TitI~, \ Accidental Gaseous Radwaste Release lMStateli]eijtl,j r.;;;;.;.;;.;;..;=....;;;.;..;;.;.;;;..;;.;.;;;:.;..:.:::.;;:=~=c:.;.;;;~.:...:.:::=~::::.:....==-==..;...;.:="=;;.;;...:..;.:..;..:.:.;..:c.::...;:;;;.;.;;;....;;.;.;;..;.;;;;;.;.;;;.;.;;;;..:=..--------------!-
55,41 (13) The WGDT relief valves (205340 sheet 2, E-7) combine into a single header and go to (205322 sheet 1, G-3). which is the FHB Ventilation exhaust downstream of the exhaust fans. It then flows to the plant vent (205337 sheet 1, G-11 into the plant vent release pOint. A Release is detected by Plant Vent Monitors but does not pass through ventilation (8), (C). (D) Flowpath is not through filters nor
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[Question !~J:licJ IRO 20 Given the following conditions:
- Unit 2 is operating at 65% power.
- 21 Charging pump is in service.
#4 SW Bay has been isolated due to a bay leak lAW S2.0P-AB.SW-0003, Service Water Bay Leak.
- All #2 SW Bay pumps trip.
Which of the following describes an action that will be performed in S2.0P-AB.SW-0005, Loss of All Service Water, and wh ? Isolate CVCS letdown. This prevents resin damage in the CVCS demineralizers when letdown temperature exceeds 136*. Isolate CVCS letdown. This prevents heating of the fluid in the VCT to a temperature which will cause a loss of NPSH to the charging pumps. Swap charging to 23 charging pump if it is available. This reduces the heat load on the CCW system and extends the time available to a controlled shutdown of the to the RCPs, which allows sufficient time to install
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iKA:'II 000062A206 IsystEi'm/Evcijutipt1:1"\~fe"'1 ILoss of Nuclear Service Water 55.41 (10, 6,7) A and B are incorrect because while letdown is isolated, it is to m inirnize the heat input to the CCW system to prolong cooling to the Thermal Barriers and Rep seals. The CVCS demineralizers are automatically bypassed on high temperature at c..:.:..c::.c,-",,-~,",,-,,:...:..J 1360F. C is incorrect because a swap to the PDP is made since it is cooled by CCW, and an immediate action in AB.SW-005 is to trip the Rx, there will be no controlled shutdown. D is correct because the basis document states that the transfer to the Positive Imaintained by'the PDP until temporarl cooling hoses to th~ CCP's can be installed from the DM system. Lq':Number . . . \ IABSW04E004 I I I
RO skvscraper'l : SRO Skvs¢r~pei "'c"" r--,~"-, -,"' IRO 21 """""" " lQue~~on Topi£J I Given the following conditions: Unit 2 is in MODE 3, NOT and NOP. I- A total loss of all Control Air occurs. Which of the following describes the basis for transferring the charging pump suction to the RWST lAW S2.0P-AB.CA-0001, Loss of Control Air? I The RCS must be borated to CSD conditions, and RWST water is at a higher concentration than the VCT, so boron will be added to ReS at a faster rate. I [§] Ilf a centrifugal charging pump is running, its recirc line back to the VCT will short circuit any boron addition and raise the time required to achieve CSD boron concentration. , I [£J I CVCS letdown will be isolated, and VCT makeup is unavailable. This would cause a rapid depletion of VCT level and subsequent air induction to the CVCS charging header. I itive displacement charging pump is running, the minimum flow supplied at the low speed stop would not match the flow lost to the
,,'-'u am the seal retum relief valve lifting, and RCS inventory will be lost I 1!-n~~Eitl ~ [§~amteverll R IIVUl:lI!l<l..,el... eve! II Memory II:facmtY~11 Salem 1 & 2 11.E:xamDate:*11 9/26/20111
~IOOOO65K308 IAK3.08 IIRoxa,u~JI3.7LSR()V~'tie:TI 3.911~ec~ip~~ill~"R22rolJ~lIUls~q~rb~p:1 11 D [J ILoss of Instrument Air 55.41(10) ARCA basis document slates, "Prior to commencing a cooldown, the RCS must be borated to cold shutdown conditions. Withoul Control Air, there is no way to reduce RCS inventory. Thus, the amount 01' boron that can be added before the cooldown L-.>c~c_ ..~,_c~~'--' commences will depend on the available space in the PZR. Due to the slow rise in level, this may become limiting. Therefore, the charging pump suclion is transferred to the RWST early in the event. This ensures that any addition to the RCS is at RWST is incorrect because while the recirc line does gO' back to the VCT, it is not the b~s'is for the transfer. D is incorrect because seal Iretum is not isolated on loss of air. ~bb~~~~~~~~~~~~~~~i~~~~~~~~~~~~~~~~~~~~~IP<laeNOi.I!R~ ~~~~~==========~~~~~~======~~====:====~~===[~16==~ F=================~~============~F======~F=~~I==~ 1------------------______________~1------------------------~1---------------~ -----~I----~ IABCA01 E002 IlIJIab~rial Requiredfor Examination II II IQu~~~ionsoufce; *. ~ IPrevious 2 NRC Exams I ~stionIlJlOdificat\onMethod: iI Direct From Source IIUsed During Trainiri~~rogram 1 0 rQueStiO~~urce co~ I"J" ILOT RO NRC Exam August 2008 icv.. ", ",." ." .......
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~uestion Topic I RO 22 I
Given the following conditions: The Unit 2 operating crew is responding to an Inadequate Core Cooling condition lAW 2-EOP-FRCC-1, Response to Inadequate Core Cooling. The crew is unable to re-initiate ECCS flow.
- All SG WR levels are -5%.
- NO AFW pumps can be started.
- All core exit thermocouples indicate >1200F.
Prior to opening the PZR PORVs in an attempt to lower CET temperatures, how should RCPs be utilized? RCPs will NOT be started since possible rupture of hot SG tubes could occur. RCPs will NOT be started at this point since they are required later in the procedure. Start ONLY one RCP in any RCS loop. Continue operation of ONLY one RCP until core exit thermocouples are less than 1200 deg. Start one RCP in an available RCS loop. If core exit thermocouples remain> 1200F, start another RCP in an available loop. Continue until all available RCPs are running or CETs are <1200 deg. K"nI~ ~ ~illIl ~ IMemory ~l 1 & 2 I ~~ I [Eg1000074K201 1~~=:J~~~I2]~~§@3li§~I~§.§@~U~~U
!Inadequate Core Cooling and the followinc: I
'~xp'.an~tj,C),n,* .o.'1 5?,41 (1 0) With SG WR levels at -5%, the:e is no secondary heat sink. At step 13114 of FRCC-1., no SG level and no AFW f1?W II' Ari~wers:*i':fi: In directs operators to step 23. Step 25 caution states that RCP's are only to be started loops WIth SG NR levels l._ _ _ _ .~ __ --'~ there are none. No RCPs can be started. It is NOT because that they will be needed later in procedure.
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~. .- - . ' - ' - ~':::I status and SG WR levels 10 stem.
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~~uest~:::ropic i A small break LOCA has occurred on Unit 2. The actions of 2-EOP-LOCA-2. Post-LOCA Cool down and Depressurization are being performed, With ONE Charging Pump and ONE 51 Pump running. PZR level is stabilized at RCS pressure of 900 psi!]. With RCS subcooling at gOF. the operator turns on a set of backup heaters.
Which of the following describes the result of turning on the backup heaters? Break , ECCS Increased pressure will
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!Question TOpi:J IRO 24 Given the following conditions:
- Unit 2 is attempting to identify and isolate a LOCA outside containment.
2-EOP-LOCA-6, LOCA Outside Containment, has just been entered.
- The source of the water is backleakage from the 23 cold leg injection line.
Assuming that any valves required to be operated during lOCA-6 operate correctly, which of the following leak locations would NOT be isolated while using 2-EOP-LOCA-6? On the valve outlet flange on 21SJ49, RHR DISCH TO COLD LEGS. Between the 2RH20, RHR HX BYP VALVE and the 2RH26, HOT lEG ISOL VALVE. Between the 2RH2, RHR COMMON SUCT VALVE, and 22RH4, RHR PMP SUCT VALVE. IComprehension I rfa~i1ityj ISalem 1 & 2 IIEx~1(l9jlte:j [ 9/26/20111
~I 00WE04A102 I ~= C'illy~0'~~[J]~g[JI~ [ii~~D~~~lD l§yst~:gm!!eEJi!i~ j LOCA Outside Containment 1M StatemeJl!=1 ~~~~~~~~~~~~~~~~~~~~~~~~~~;~~~----------------------~
55.41(3,8) 2-EOP-LOCA-6 closes/checks closed the following valves: 2RH1 OR 2RH2, 21 and 22RH19s. 2RH26. 21 and 22SJ49s. Using drawing 205332-SIMP, it shows that any leak between the RH1/2 and the SJ49s will be isolated with the above
~~~..*J valves closed. The only location which wouldn't be affected by those valve being closed in the downstream/outlet side of the SJ49 valves. The stem statement of proper valve operation was inserted to preclude a candidate from assuming a leaking valve may not ------------__________~I----------------~I----------~~==~I--~ ~~:-.~
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;Question Topic I RO 25 Given the following conditions: - Unit 1 has experienced a LBLOCA. - 11 RHR pump is CIT electrically by it's 4KV breaker only. - RWST level is 14.8', and operators are performing 1-EOP-LOCA-3, Transfer to Cold Leg Recirculation.
Which of the following conditions will prevent transfer to CL recirc, and cause the crew to go to 1-EOP-LOCA-5, Loss of EmergencL Recirculation?
~ 112SJ44, Cant Sump Suct Valve, does NOT open when its Sump Auto Arm PB is depressed.
~ 11 RP4 Lockout Switch for 12RH4, RHR Pump Suction, is stuck in "Locked Out" position.
g] 112SJ44 open PB is depressed before the 12RH4 close PB is depressed. ISystemlEvolutioriTiUel I Loss of Emergency Coolant Recirculation I~~ IKAStatemElnt: I Knowledge of the interrelations between Loss of Emergency Coolant Recirculation and the following: I Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features. iEXP,la"ri,atio,.,n,.Of A is incorrect because Unit 1 SJ44 valves do not have auto swap function like unit 2 has. C is incorrect because the close signal iAnswers: ""'" will remain to the SJ44 even though it is interlocked to not open until the RH4 is shut, and there is no lockout to prevent it opening. B is incorrect because there is no lockout switch for RH4s on RP5. D is correct because if the 12RH4 cannot be closed, the 12SJ44 cannot be opened, and recire cannot be established, and procedure directs transition to LOCA-5. ~IT:::r=an:::s:::fe=r=to==C=O:::ld::::::::=;=;:::~:::::::i~C:::::::li~::::~=O=n======"'==:"fl=l~=~;E='
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RO Skvs~~ap~61*s~OSkVscraper RO******* ..*T* .. **<;**~*****'***** SvstemlEvolUtion : List SRQ. svste~JEvciIU~i~~l.istJ 1Outli ~eChanQes'" [9~e~!i~n TOPiciJ Given the following conditions:
- Unit 2 has experienced an event which has resulted in 24 SG pressure rising to 1115 psig .
- A MSLI has been performed.
How many SG Safety Valves will be open on 24 SG if pressure remains at 1115 psig, and all safety valves operate when ex ected? 55.41 (4,7) Each SG has 5 safeties, with lift setpoints of 1070, 11 DO, 1110, 1120, and 1125 psig.
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" ',.1 I I I I II I
Given the following conditions:
- Salem Unit 2 experienced a LOCA with failed fuel.
- Containment pressure peaked at 18 psig, and is currently 3 psig and slowly lowering, Containment radiation level is 1,1 E5 Rihr and rising very slowly.
- The TSC has not given the control room any direction on use of Adverse numbers while in the EOPs.
Which of the following describes how containment conditions will affect use of "Adverse" numbers while in the EOP's? radiation levels have the TSC has not provided any direction yet. they reset on lowering containment pressure and the radiation threshold has not been reached. [KAstatemerlt~ Abili to determine and inter ret the followin as the appl to Hi h Containment Radiation: Adherence to ap ropriate procedures and operation within the limitations in the facility's license and amendments. lEX. plar~ti . O**.n.. .* * *.**.*.Mj* 55.41 (9.12.10) Adverse containment numbers are used when containment pressure reaches 4 psig or containment radiation level reaches 1E5 Rlhr. The Adverse condition resets when containment pressure lowers less than 4 psig. but does NOT reset on high
- Answersz<*
="'~-' "-'-'-_.._. radiation. The conditions in stem are such that Adverse numbers have are in effect based on the containment radiation threshold being exceeded.
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~~~~.~~ I;!~~ar to NRC Exam questions from Sept 2001 Coo.k exam, Nov. 2002 Salem exam, Dec 2002 Beaver Valley I IComment . ' .......*..... ....... / .................
II I I I I I
Given the following conditions:
- Unit 2 is operating at 100% power.
- A failure in the automatic Rod Control circuit cause Bank D Control rods to step in at 72 steps per minute.
- Control rods insert 20 steps before the RO places rod control in manual and rod motion stops.
Which of the following describes the effect of the rod insertion with NO other operator action?
~ RCS temperature will lower to the RC Loops Tavg-Tref Deviation alarm setpolnt of 1.5"F.
fAns;;;~' E=J E~~L9JeIl ~ IApplication ~iiftYD ~.1 1 & 2 I ~a1if 1 9/26/20111 JKA: [~~~lsJtovi:iiui!:]'I@~§JI~ ~OGroup:IUlsR6,~roup:!U ISyste;:;:'jE;::~lutio~'Titiel IControl Rod Drive System IKA Statement] *Knowledge of the effect that a loss or malfunction of the Control Rod Drive System will have on the following: RCS IEXPlanation~ 55.41(5) A is incorrect because the RIL for 100% power is -170 steps on Bank D, so 20 steps of insertion from 230 steps (ARO) will iAnswers:"~::.::cJ not cause this alarm. B is incorrect because Main Generator load will lower due to lower steam pressure, but it will not remain lower
- - - - - since the Governor valves will automatically open to restore Turbine Steam line pressure to the value corresponding to 100% power.
C is correct because the rod insertion will cause a lowering of RCS pressure sufficient to full close the Spray valves, wi1ich are n",,,,,,,, I", _ gO/_ "non ';"0 t" rn . . "no ('\,,-,,,n "f P7Q'RII I h.,,,t,,,,,,, in m"nll,,1 nl\l n ie . ho,.."""" "hilo 10m no,,,+' "'" ,.11 Ilower, the setpoi'nt of 1SF is wrong. 1.5' low Tavg=Tref is wihen auto rod outward motion would occur. j............. .. . ....... " .. ReferenceTitle:;" ,,[1 ~tjiitY-f{~ferE!lJ*eNurnbef.::J ~efE![en~;;SectlQn " "IIPage No*1 ~1SjOO] ! ~Ic=o=nt=in;;;;;uo=u=s=RO=d=M=o=tiO;;;;;n===~=====:::::::=~11 S2.0P-AB.ROD-0003 II II 1121 I IOverhead Annunciator Window E II S2.0P-AR.ZZ-0005 It 1111 11 19 I I~c=o=ntr=o=1c=0=ns=0=le=C=C2=======~I~~1S::::2=.O=P=-A=R=.Z=Z=-O=0=12===~J I 1136 11 35 I [L.9.Number IABROD3E003 IRODSOOE019
Given the following conditions:
- Unit 2 was operating at 100% power when a 500 KV grid disturbance caused a Rx trip. - 2H 4KV Group bus deenergized upon the Rx trip. - 21 AFW tripped immediately upon starting. - SGSD flow is zero to all SGs.
All radiation monitors indicate as expected following a Rx trip. During performance of EOP-TRIP-2. Reactor Trip Response, the RO reports that 21 SG NR level is 8% higher than the other 3 steam generators. Loss of 23 MAC Panel has caused SGFPs to supply uneven feed flow to the SGs. 21 SG is steaming less than the other generators since its RCP is no longer running. The loss of 21 AFW pump has caused Pressure Override Protection circuit activation on 22 AFW pump. IComprehension I l~ac:,~i~~11 Salem 1 & 2 li<A St~ement:J Knowled e of the effect that a loss or malfunction of the Reactor Coolant Pump System will have on the following: S/G 55.41(5) The loss of 2H bus, which supplies power to 21 RCP, will cause the steaming rate in that SG to go down. With AFW flow the same to all SGs, the level in 21 SG will rise markedly, as it will not be steaming. A tube rupture would be indicated if there were no other reason for the level rise, as the diagnostic step asks if SG NR level is risin!~ uncontrollably. In this instance, it is not uncontrolled, since it is the natural reaction of the reduced steaming rate. SGSD flow was included in the stem to lend support to a different Group bus, and SGFP flow has been isolated by FWI at 554 0 Tavg following the trip anyway. The pressure override circuit only affects 2 SGs, in this case 21 and 22, and would be seen equally on each SG, and in the opposite direction since it acts Ito reduce flow to raise dischar e pressure.
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I i I I I
Given the following conditions:
- Unit 2 is operating at 100% power.
- 2CC190, RCP THERM BAR CC OUTLET V, fails shut.
Which one of the following describes the effect on RCP temperatures, if any, as a result of this failure? ALL RCP ... 1#1 sealleakoff temperatures will rise.
@J bearing temperatures will remain the same.
IReactor Coolant Pump System KAS"
~tatement: dJ IKnowi eJQe d af teoh f teeh ffect af a Iass or ma Ifunct'Ion on th e f0 II oWing WI'11 have on the Reactor Cooan 1 tP um~ si:stem :
Containment isolation valves affecting RCP operation I 11:~~I~~~t~ah;~~f ] line 55.41(3) The CCW line supplying the RCPs is a single line supplying both bearing cooling and Thermal Barrier cooling. Once the inside containment splits, the CCW from the Thermal Barriers has its own, separate return line, which is isolated by the 2CC190 Ar1?,weE'..!;-,~'::c;J. (inside containment) and 2CC131 (outside containment.) The Thermal Barrier CCW flow acts to cool reactor coolant flawing upwards through the thermal barrier upon a loss of seal injection flow. With normal seal injection, the loss of CCW to the thermal Ih~' . A ,U, Dt"O I i,;i",>!,::" t !P~'i!ti!~:;;;i',,::YI ii"i"; il'T:il(;nn~.""~,,,,,,."!,,,,,,. 1~~I"",,,!.~~'Secti6nJflinechanQeS 1 [QUe~iO-;TIPhll RO 35 IWhich of the following relief valves lifting would cause a rise in Pressurizer Relief Tank temperature or level? 55.41(3} The 3 distracters are components which used to be directed to the PRT but were re-routed to the containment trench
~~~1t..L~_.,J during a DCP.
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~==================~~============~~======~r~==~~159==~ ~================'~==========~:========~F==~~I==~ I-------------------------II-----------------~I----------~----~I----~ IpZRPRTE008 I !PZRPRTE003 I [ I iM,HeriaIJ~equired forgXamiQ~tlt)h;] I !I iOuestinnSource:iJ IFacility Exam Bank IIAU!!S~i?1l M"difi.cation,M~~hod:~il Significantly Modified! fUsed Jj:'tingTr~il1irigprog~~riJ D
~~==~~========~~~==~==~======~
Iguestionsol.l~':::?~mmerltsll Vision Q61390, inputs to PRT, changed correct answer and all distmeters. I I I, I I,----------------------------------~I
ROSkvScraoer. f ' SRO Skvscrap~~ I ,.RO Svstem/Evolution uit., I'SRosv~temlEvol~tion List 1..:HoJtH~e Ch~nries.
~':>ll'::'.'.'...'.. '::'~ IRO 36 Given the following conditions: - Unit 2 is operating at 65% power. - An intersystem leak develops between the CCW system and the Spent Fuel Pool Heat Exchanger. - CCW Console Alarm SURGE TANK LEVEL HI-LO alarms on 2CC1.
Which of the following describes the action required lAW S2.0P-ARZZ-0011 Control Console 2CC1? Drain the CCW surge tank from local drain valve to a 55 gallon drum to prevent overflowing tank to the in-service WHUT. I~~ Ib lle~roLe\ter! IR I [~ognitive""[~v~'J IApplication I 1f:(:jJit~ ISalem 1 & 2 tion
.. Modification
... .. Method IQuesti~n Source 90I'Qmentsii VISION Q50370
o""lstion Topic I IRO 37 I Given the following conditions:
- Unit 2 was operating at 75% power. - Both pressurizer PORV block valves (2PR6 and 2PR7) have been closed for several weeks due to seat leakage past the PORVs (2PR1 and 2PR2). - The PORVs are being maintained in MANUAL, with ALL Tech Spec required actions for the leaks complete. - Plant conditions develop which required a shutdown and depressurization. - RCS pressure is being maintained at 800 psig due to problems with isolating the SI accumulators. - The RCS cooldown continues to below 312°F.
Which of the following describes the effect if the operator were to arm the Pressurizer Overpressure Protection System (POPS) under these conditions? 12PR6 and 2PR7 would OPEN; 2PR1 and 2PR2 would OPEN [~ 12PR6 and 2PR7 would OPEN; 2PR1 and 2PR2 would remain CLOSED 12PR6 and 2PR7 would remain CLOSED; 2PR1 and 2PR2 would OPEN fd.l12PR6 and 2PR7 would remain CLOSED; 2PR1 and 2PR2 would remain CLOSED 55.41(7) With pressure above 375 and temperature < 312, arming POPS would cause the PR1 ,2,5,6 to open, regardless of the MANUAL selected for PR1 and 2.
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- - - - - - - - - - - - - I I - - - - - - - - - - - - ' I - - - - - - - - I I - - - - - l I - -.. . .
.Material Req~ired fCl~ Examination .II II IFacility Exam Bank liCluestion MOdific~~i~ethod;:-!I Direct From Source I[Used During Training Program i ~======~==~ IVision Q87628
RO SkyScraper I 'SR6'Sk'lscrap~~ I .,.Ro Slf~t~;nllf\lbtuti~n ListlFi sRe SystemiEvoiutlortlist. ., ()uiihle Cha'naes *1 iQu~!ion-Topicll RO 38 With Unit 2 operating normally at 100% power, what would be the effect on RCS pressure if the controlling PZR pressure channel instrument were to fail LOW, with no operator action?
~ T.
~ lower until the Rx trips on Low PZR pressure,
.4J rise until PZR spray valves open further to restore pressure to 2235 psig.
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IKA.Statemerit':l Knowled e of the effect that a loss or malfunction of the Pressurizer Pressure Control S stem will have on the foliowin RCS I~Wers:::"**"***.
,EXpl~n~ti~n!of 55.41(7) PZR pressure controller will see a low pressure condition from the failed instrument. This will cause spray valves to close fully, and all PZR heaters to energize. Actual RCS pressure will rise in response to the heaters being on, and will continue to rise until reaching the PZR PORV setpoint of 2335 psig. Only ONE PORV will open. PORV actuation coincidence is 2/2 channels>
2335. Since one of the channels is failed low, its PORV cannot open. B is incorrect because pressure would rise, but the trip is Iincorrect because the spray valves will shut on the channel failure, and be unable to respond to the rising pressure.
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IABPZR1E001 Material Required for Examination JI IJ I
!Question Source: IFacility Exam Bank IIQu~stion Modification.Method: . IEditorially Modified I [Used During Training prog~a~ D I 'Olll""tii-m Source Comments I IVision Q80493, replaced two distracters so that all choices did not include rising ReS pressure. Correct answer remains the same . I IComment , ".,
i I I I
SkvS~raper I .SROSkvs~r~p~r I . RO SvsteI11IEvolutio~List< I SRO Sv~t~mIEvolution List!Out:lineChanQeS I RO c'-----:-. I
- Question Topi~J RO 39 I
For each of the listed PZR heaters, identify their 460V bus power supply. 11 control group -- 11 B/U group NORMAL _ _ 11 B/U group EMERGENCY _ _ 12 B/U group NORMAL _ _ 12 B/U group EMERGENCY
~ IE, G, C, E, A.
I
~ I G, G, C, E, A.
I [ ] I E, E, A, G, C. I [ ] I G, E, A. G, C. I ,Answer II b 1 [E~am'Levelll R I [Cognitive Level II Memory I [Facil~ty: 11 Salem 1&2 I [EXamoateJ [ 9/26/20111 IKA: II 011000K202 IIK2.02 l~value:113.11IsRQvalue:113.21Isection:II~§i:>GrOllPJI 21~o.GrouP:11 21 :~~~3~1 0 I I Pressurizer Level Control System I~***,***._ ...,.----*~~***n._ ISystem/Evolution Title J IKA Statement:] Knowledge of bus power supplies to the following: I PZR heaters I
!Expl~lnation of 55.41 (3,7)The distracters all contain the possible busses, but in incorrect orders.
iAnswers: II r- .,' Reference Title ,/.
. I [; ,Facility Reference NUOlber:J IReferenye,SI~ction II Page No-:I [ReVisionl I No.1 Unit 1GP 480V bus PZR heaters.controlie 11203347-1 il II 11 13 1 I No 1 Unit Backup Group 11 11203348-1 JI il 11 9 I I No 1 & No 2 Units Group 12/22 backup groups [1247992-1 [I 11 3 I II IL.O.*Number IPZRP&LE005 I_ _ _ ---J IMaterial Required for Examination II II IQuestion Source: II Facility Exam Bank IlQuestion Modification Method: j Direct From Source Ilused ~ur~~g Training Progra~ 0
[QUestion Source Comments 1 [0801 C39 I I IComment I I I I I I I
RO sk~Sc:i~pe~. ..*. SRO Skyscraper I.. .Rcisv~t~';;iEv6rJtion li~t *.* ****SR()Svst~m(E,;()i~ti~n List** I~Otitlin~Ch~~Qes [Question TopiC] Which of the following identifies why a Safety Injection signal on Unit 2 is reset early in the EOPs after transition out of EOP-TRIP-1, Rx Trip or Safety Injection, following a LOCA? the Phase A signal must be reset to allow the ECCS pumps criteria have been and Phase A cannot be reset until it allows operators to regain control over plant equipment and prevents any equipment from automatically repositioning when RWST reaches 15.2', the Phase A signal must be reset to allow sampling of the SGs and RCS, and Phase A cannot be reset until the SI signal is reset. it allows operators to regain control over plant equipment and restore a sustained compressed air supply to containment. [*:11 012000G406 tKAState~1![J ,--________________________._ _____________--, Knowledge of EOP mitigation strategies, 55.41(7)The Phase A signal CAN be reset without resetting SI as it has a retentive memory circuit. See Note 10 on 221057, Bases docum ent states that the SI reset function is so that equipm ent can be aligned, and to restore a sustained, compressed air supply to allow control of air operated equipment in containment (e,g., charging and letdown valves, PZR PORV's, etc.) A is wrong both i because the reason and the Phase A reset logiC is incorrect. B is incorrect because it does not remove the standing "S" signal
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IReactor Protection Signal Safeguards actuation 11221057 II II IIF2=Z=::::::; I Loss of Reactor Coolant Bases Document 112-EOP-LOCA-1 II 1118 11 28 I II II II f IF===='
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- 9uesti~n Topic I During a power ascension from an initial power level of 4%, Permissive P-10 does not actuate when expected.
Which of the following describes the effect of this malfunction on the Reactor Protection System if the power ascension were to continue to 14% Rx High Steam Flow Safety Injection will remain blocked. A Loss of Off-Site power would NOT initially cause the RPS to trip the Rx. The low power Rx trips could NOT be blocked until P-13 energized during Main Turbine startup. t@]!a*e~ Knowled e of Reactor Protection S stem desi n feature s and or interlock s Automatic or manual enable/disable of RPS trips
. . .x.):l. '.* .a.*.**.na. . 't.i.*O 'E ~nswers:
r.'.* *. .. ':'.
. *.: . O.'.f.g.:*. 55.41(7) D is incorrect because the low power trips can NOT be blocked until P-10 is actuated. C is correct because the "at power" Rx trips (which are different from the "low power" trips) are not unblocked until P-? is actuated, wlhich gets its input from either P
---- 10(wlhich is not actuated) or P-13. (which is not actuated because the turbine is not online at 15% power). At 10% power the AFW pumps will have been secured, and wlhile the loss of off site power will cause a loss of the SGFPs, the SGs will not shrink to the 10 l¢ageNo:l [R~Yision.1
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[ciue~tion -T~pic] IRo42 I Given the following conditions:
- Unit 1 Rx power is 8.1 %.
1- Power is being raised slowly in preparation for rolling the Main Turbine.
- 11 SGFP is in service supplying FW to SGs.
- 12 SGFP is latched and at idle speed.
- All AFW pumps are aligned for normal standby operation.
11 SGFP trips. Which of the following describes the effect, if ant, this will have on the AFW pumps with NO operator action? IThe MDAFW pumps and the TDAFW pump will start when SG levels drop to the 10 10 level setpoint. j I.The MOAFW pumps will start when 11 SGFP trips. The TOAFW pump will start when SG levels lower following the Rx trip. I
~ IAll AFW pumps will remain in standby. Sufficient steam will be supplied through the 11-14MS18s, MS STOP BYP VALVES to supply 12 SGFP.
I II All AFW pumps will remain in standby. 12 SGFP will remain in service since at this power IEivel it is being supplied with steam from the Heating Steam System. 1
- Anslf.lerll a I [EJ(~lnLevelll R I [SRgl1my~Li!vel' IApplication I lF~9Httj{g [Salem 1 & 2 I 1.5~.a!'i~~,~;.: ! 9/26/20111
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~KA statement:'] Ability to predict and/or monitor changes in parameters associated with operating the Engineered Safety Features Actuation System controls includin :
S/G level 55.41 {4,7,8)0 is incorrect because the operating SGFP(s) will be placed on Main steam supply prior to exceeding 5% power (IOP*3, step 5.4.10), and 12 SGFP will not provide sufficient discharge pressure at 1100 rpm (idle speed), and the stem stales with no
=c..--,-~...._~-",,-,-, operator action so speed will not be raised. A is correct because the MDAFW pumps and TDAFW will start on 10 10 leve! in SGs as the source of FW is lost, and the SGs continue to steam. C is incorrect because the MS 185 could provide sufficient steam flow to
- 01 * *
- 1 MDAFW pumps is a trip of BOTH SGFPs..
1 - - - - - - - - - - - - - 1 1 - - - - - - - - - - - - ' - - - - - - 1 ~==.I-----' [h:£7NlIT~21 IAFW000E015- --I I I I I lMaterial R=quiredfor ~){aminatiori ] I II iQ~:~~1"l Sou~ce:** I IFacility Exam Bank ll9uestionMb~ification Method: ~ Editorially Modified IlU!ied Duri~grraining~rogr~TJ 0 IQuestion Source Comments I IVision 085462. Modified stem from MSLI to trip of operating SGFP with other SGFP latched. Correct answer
- ~- ...... remains the same. Enhanced distrac!ers. I IComment
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RO SkyScrape~ ISR.OSky~~r~perl ... .... [Q;:;esti;;:'-ToEi~11 RO 43 I Given the following conditions:
- Unit 2 is operating at 4% power during a startup. - A RCS leak causes PZR pressure and level to lower rapidly. - The operating crew initiates a Rx trip and Safety Injection. - When the 81 is initiated, a loss of off-site power occurs.
One minute after the loss of off-site power, which of the following describes a condition which indicates a failure of equipment to actuate as exeected, if no further operator action is taken?
~J I NO CCW pumps are running.
r [§J 124 SW pump running and 23 SW pump stopped. r
~ ICharging Systems SI Flowmeter reads 100 gpm.
I
@]J21 ABV Supply Fan running, 22 ABV Supply Fan NOT running.
I IAnsw~ Ic I [Exarn Level. I IR I [Cognitive Level. "11 Comprehension I [FaC;III§ij I Salem 1 & 2 I ~~.mDate:jl 9/26/2011 1 IKA: II 013000A401 11A4*01 ] ~I 4.51l sRO yalueil! 4.81Isection:U~ [!§OGroup:11 111sRO Group:'11 11 I~~tij D I I Engineered Safety Features Actuation System
--~--~-
ISystem/Evolution Title IKA StatementJ Abilf!y to manually operate and/or monitor in the control room: I ESFAS-initiated equipment which fails to actuate I I:~xplari~tic)~ of 55.41(7) CCW pumps are not started during MODE III (Accident plus Blackout). 24 SW pump is designated as the Lead SW pump Answers:.* ..*,'*1 on B 4KV bus, and only one pump per bus is started. 23 receives power from B bus also. 22 ABV Supply Fan is administratively locked out, and is what causes the B bus sequence not complete to occur on every SI initiation. The Charging systems SI flowmeter is BIT flow. 2 centrifugal charging pumps should have started, and even if RCS pressure were normal, which its not. the fin" thrn* ,,.,h tho Q I T r l h . ,;f;~~nth h;,.,hor r::ahor thoro ;c - nmhl"m ,,;th Iho fln,,,n<olh . " c:: 11? <on'; c: 11 ~ !=lIT n"ll"t ",I""c 1did not stroke full open, or less than 2 ch~rgi~g pumps started. TRIP-1 page 1, and TRIP-2 page 2 flowcharts, page 1, each Table I .*. *. Refer~nce Title' I Rx Trip or Safety Injection
=
A shows Accident Loads and Blackout loads respectivel~. I* .. Facility Reference Number=:] [Reference Section fI2-EOP-TRIP-1 II I[iage No.; IIRevisionl II F1 1127 I I Rx Trip Response ~ 12-EOP-TRIP-2 II II F1 1127 I I II JI II II I IL~O. Number ISECOOOE004 IMaterial Required for Examination II II IQuestion Source:
. .. jlNew IIQuestion Modification Method:~ :1 IlUsed During Training Program I D IQuestion Source comrnentsl I I r---
]Comment I
- I I I I I I I
Given the following conditions:
- Salem Unit 1 is operating at 100% power.
A small LOCA (300 gpm) occurs, and a Rx trip and Safety Injection are initiated.
- A loss of off-site power occurs when the Main Turbine trips.
NONE of the CFCUs start in Low Speed. How will the failure of the CFCUs affect containment instrumentation readings? CFCUs in service I022000K302 1EJ~9~@~~'~ ((~~D[§~g!~ iSy~t~rrij~~()fl.lH()i(tltt~J IContainment Cooling System IE~pian~ti9ri;of 55.41(5,7)The CFCU HEPA filter, which is in service following accident initiation, is designed to remove minute radioactive ~llswer~:J'~ particulate matter from the atmosphere so that offsite doses do not exceed the limits set by 10CFR100. A is incorrect because the lack of cooling function provided by the CFCU will cause the pressure to rise. C is incorrect because the Dew Point will rise due to higher temperature in containment. D is incorrect because the leak detection system will rise as condensation occurs on its cooling
*.ROskvscraperl ~.::S:R:O:S:kVt:S:C:ratPe:r:.~:*:R:6:S:vt:.*~:tet:mt:iE:**~:*6:1~:ti:rin:L:iS:t:**:JL.:':s:R:o:.*.:svt:*s:te:rril::E:v:o::lu:ti:()n:L::is:t~I~~:.o:***:~t:ii:n:e:Ct:**~:1:J1t:*~:e:s~====-'--'--'--'-======'71 i9uestion Topi"§] I_R;;...o:....-4;;...5_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _---l Given the following conditions:
- Unit 1 has experienced a LOCA.
- Control room operators have progressed to the point where the SEC has been reset.
- Containment pressure was 4.5 psig when the SECs were reset.
Which of the following describes the containment spray system response should a hi-hi containment pressure signal be generated at this point in the accident? would realign for spray, the CS pumps would be started by the SEC. would realign for spray, the CS pumps would have to be manually started. must be manually realigned for spray, the CS pumps would be started by the SEC. must be manually realigned for spray, the CS pumps would have to be manually started. IApplication I fF':@l1:t~~ll Salem 1 & 2 tSy$tem/E~c)hJti<?flTi~f~] IContainment Spray System IKAStatelllel'l!IJ Abili to monitor automatic operations of the Containment Spra Pump starts and correct MOV positionin I-EJ(.P.**.I* a..* *n. . . a~.\o.*. ~.*. *.()f.. . 155.41(7) The SEC controller operates the CS pumps at 2 di.fferent point in. the sequence U~TlL the SEC i~ reset. The SEC ONLY
,AIisw'ecrs; **. ,.... ..' controls the CS pumps, not the CS valves. The valves realign on the hi-hi cont pressure Signal whenever it IS received, but once the ~ . SEC is reset it will not start the CS pumps since the sequencer is no longer active.
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RO SkvScrllP~f I, S~OSkVsc{aper 112 keyswitches on BOTH safeguards trains. 2/2 keyswitches on BOTH safeguards trains. 1/2 keyswitches on EITHER safeguards train. 2/2 keyswitches on EITHER safeguards train. ~rsw:a Id I !E:xarnheYElI R i iC9s'h"lt!"e '-eyer IMemory I [FaB!!JE [Salem 1 & 2 ~I026000A401 1fA4.01--1[ROYa,u~:I@~~~@~I~~~DlsRdG:r911P:1 ~rrij~voll.ltionTitt~D IContainment Spray System !KASbt~q
~~~~~~~~~~~~~~~~~~~~------------------------------------------~
55.41(7) Salem has two safeguards trains, either of which performs its safety related function. Containment Spray requires 2/2
.*.""'.i keyswitches tumed simultaneously to the operate position to activate containment spray, due to the severe consequences which would occur if it were inadvertently actuated without it being required. Either train of safeguards will perform actuation. 2/2 keyswitches on EITHER train.
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~~~~~~~==~l~~=====~I=====~F==~~122=::::. ~========~I=======~~===='rF==~I~~ I--------------------------------~I-----------------------~I---------------~I-----~I----~ II
ROSkyScraper ISROSk'l!lc~ci~t ;>R6 S'l~tem;E~()I~ti6'nvd$t: ,I f SR(fS~$temlE~()I~tio~tist l_ottii~~Ch~n'~e~ [§ues_tiinTopic:] IRO 47 which one describes how the Spent Fuel Pool level will be lowered if required, lAW S1.0P-SO.SF-0001, Fill and Transfer of the Pumped with Spent Fuel Pool Cooling pump to RWST. Pumped with Refueling Water Purification pump to in service CVCS HUT. Drained via the Spent Fuel Pool Skimmer loop to the Drain Header to the in service evcs HUT. fKJilllo33000K105
~~mY'O'hitiri~~ I Spent Fuel Pool Cooling System s between Spent Fuel Pool Cooling System and the following: I 55.41(8,13,4) The RWST normal level is -41'. The puts the water level at 141'. The Spent Fuel Pool is -128'. There can be no gravity drain TO the RWST, although it will work coming FROM the RWST to the SFP. Transferring water from the SFP to the RWST is done by manipulating valves within the SFP Cooling system, and pumping the water with the Spent Fuel Pool pumps to the RWST. The other method is to drain it to the CVCS HUT. The Refueling Water Purification pump cannot take a suction on the
!performed by procedure, and it would d-rain to' the Fuel-Handling BidS ~ump, not the eves HUT.
~e~!i:,~::.* . .*'.' . .. ,,~~t'Elf~6geTiH~ ::~~i~l'" ."~~ ~iiiiYIij~f~fen<:~l1l1rn~eillIR~fer~h~fil s.~c~ [page No:1IReyisiori! 1Fill and Transfer of the Spent Fuel Pool [I S1.0P-SO.SF-0001 I[ II r 119 1 I Unit 1 Spent Fuel Cooling 11205223 II I: 11 26 I ITankCapacitYData IIS1.oP-TM.zz-0002 II II ila I iL::~Numb;';~ I SFPOOOE010 !MaterialRequired for Examination JI Ij IjUsed [)UringTraining 'Program 10
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* .RaSkYS(;r~p~r 1</SR~SkV~Cr~ller'l . "/RdS~~teIhlE~oi~ti6n Li~tl sRdS~sten1iEV()/iut{()ng!;tl.:.~utlirleCh~n~~s rQuestionTopiC] IRO 48 Unit 1 is performing a plant startup to full power.
From 0-20% power NR level will rise from 33% to 44%. From 20-100% power, NR level will rise to 48%. From 0-20% power NR level will rise from 33% to 44%. From 20-100% power, NR level will From 0-40% power NR level will rise from 33% to 44%. From 40-100% power, NR level will rise to 48%. From 0-40% power NR leve! will rise from 33% to 44%. From 40-100% power, NR level will remain at 44%. [i5A;iI035000A101 1~~~~yaw~0"~~~ei~[J]J~~I~~~DJ~OG~[]
~~ri17~§!i~~ ISteam Generator System ~iS~tementl~~~~~~~~~~~~~~~~~~~~~~~~~~~,~~~~~~~~~~~~~~~--+
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~u~stion TOPi,§J IR049 1 Given the following conditions: - Unit 2 is operating at 50% power. - 24MS167, Main Steam Isolation Valve Fast Close PB is depressed on the control console.
Assuming the reactor does NOT trip, which of the following describes the initial response of RCS Oelta-T and SG pressure in the AFFECTED loop? I
~ RCS Oelta-T rises md SG rises.
I IRCS Oelta-T rises and SG steam pressure lowers. I
@] I.RCS Delta-T lowers and SG steam pressure rises.
I iJ 1 RCS Oelta-T lowers and SG steam pressure lowers. I [n~\Ved ~ [§~amLexeD IR Ilqognitiyel..evel IComprehension I ~~cili!y:J ISalem 1 & 2 I iExa~D~i::ll 9/26/2011 1 iRQ.Y~lu~,;J~ !~RO}(<lIUeill2!l i§~'~~i9ri~ll~ [0 (3roupl D~(jGr()UP:,1 IkAn I039000A106 I~06
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~~l§yolllt'9rlI!~J w"% IMain and Reheat Steam System ikAstiltement? Ability to predict and/or monitor changes in parameters associated with operating the Main and Reheat Steam System controls indudin : Main steam ressure
§:JlI~r\~ti.oh9:fl 55.41(5) The unaffected SG's and therefore RCS loops provide the same amount of energy to the turbine - raising RCS OIT and AD!:,,#er;;;:':"j lowering SG steam pressure. In the affected loop, RCS DIT lowers to zero and SG steam pressure rises because heat removal is
. minimal. At least one of the conditions is incorrect in each distracter is wrong. '
~q:~fu~1':~ 1 MSTEAME015 1-__---' Ll'v'lat~rial' Rtl9uirectJor E~xamination II 'I 19~ti~s~q2i:] IFacility Exam Bank 1 [QuesiioI1Mq(!ificati()nM~th9q: j Direct From SIJurce llUsedDtiririgTr~ining Prograr!ij 0 ~U~ti~s~~~m~~~II_~j'~_OO_Q_OO_~_2________________________________~I
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RO skvscr~per'l. 'SRO Sk';;craper [aoosFi6n TOpiC] IRO 50 IWhich of the following describes why Main Steam Jines are drained prior to admitting steam into the headers?
~.-II.Ensures a vacuum pathway to the Main Condenser is available free of potential loop seals.
I I
~.J Removes any collected corrosion products or impurities to ensure Main Turbine blading is not impinged.
I IPreheats susceptible components such as steam traps prior to exposing them to full system 1: lperature. I IPrevents pressurized steam from forcing residual water in the piping to cause water hammer on downstream components. I , ~~ ~ [I:)(~~~evell ((:J ISogllitiliel..:~yeJi'i IMemory ! W~ ISalem 1 & 2 I~ I 9/26/20111 r¥ll039000K501 IL~~~~~~ITI"M~~1TI~I~§2'E~iU~~~U _ []
~yst~m/~Woluti~ill I Main and Reheat Steam System I ~.__~
Knowted e of the operational im Iications of the followin to the Main and Reheat Steam System: Definition and causes of steam/water hammer 55.41(4,10) The steam lines are designed to pass 99.25% quality steam at full power. Water which has accumulated in the piping during cooldown would be transported downstream, where it would impact inner piping walls and turbine blades if not removed. A is incorrect because the vacuum path would be from the condenser back through turbine control and stop valves which would be it shut 8 is incorrect because while remove anything in the condensed steam in the piping, it is not the reason why. C is incorrect
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~!J~stjon .~~<:: J IRQ 51 I Given the following conditions:
- Unit 2 is operating at 100% power.
- Main Steam Dumps are in MS Pressure Control* AUTO. set at 1005 psig.
. 2PT-505, Main Turbine Steamline Inlet Pressure Channel. fails LOW .
Which of the following identifies how the Main Steam Dum ps will respond to this failure prior to any Reactor Protection System response? Main Steam Dump valves will ...
! remain shut throughout the event I
ALL trip open and then modulate in the closed direction in response to lowering Tavg. I
~ ALL trip open and remain open until they automatically shut when Tavg lowers to 543 of.
I
@d initially remain shut, then modulate open as sleam pressure rises from the load reduction.
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- ISteam Dump System and Turbine Bypass Control IkASf' n", I~;;i Knowledge of the of the effect of a loss or malfunction on the following will have on the Steam Dump System and Turbine Bypass Control:
Controller and positioners, including ICS, S/G, CRDS I 'I,E~p'~~a~l~itorJ 55.41(5,6.7) When the Main Steam Dumps are placed in MS Pressure Control ModEl, that "arms" them, and they are sel up 10 Answers:')". respond to a deviation from its setpoint, which in the stem is stated as 1005 psig. Tl1at is. until Steam Pressure as sensed by a different steam pressure detector (PT -507. Steam Header Pressure, nol PT*505 Steamline Inlet Pressure) rises above 1005 psig, the steam dumps will remain closed. 100% power sleam pressure is - 800 psig, and it would rise as a load reduction occurred. wh.m PT_"n" f::.il" I", thi" ,..", "'" ::'l1tf'\!'T'!"tir mn . ,,..: ~~,',,~ Alhi,...h ill In\Al",r Reg "nn Icorrespondingly, secondary side steam pressure.
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~uestion Topic IR052
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I Which of the following identifies how a Main Turbine trip is confirmed in the EOP network lAW OP-AA-101-111-1003, Use of Procedures?
~ IOHA F-32. OEHC Tril I
I
~ Auto Stop Oil Pressure <45 psig.
I
~ IALL Main Turbine Stop Valves shut.
I [J IMain Turbine speed <1800 and lowering. I
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~J 1045000A406 11A4*o6 I~"a'ue:ll 2.81 [SRO,Value: 1I 2.71 i~!iption~lI~ ~§Gmup:11 2] [S~Q Grdue3] 1 21 ~~I~31 0
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~E:Ydlution'We*1 I Main Turbine Generator System 1[045 J IExplanationofl 55.41(4,10) A is incorrect because it is a result of the Turbine Trip, but is not used for trip confinnation. F-32 is plausible because iAnswersi >j one of the conditions which causes it to alarm is the correct answer(AIi main Turbine Stop Valves shut) but because other things (9
'" to be exact) also cause this F-32 to alann, it is not a TT confinnation. B is incorrect because while it is available to the operator on RP4 right next to the Turbine Stop Valve indication, it is not used to confirm the TT. It only show that's the oil pressure which has incorrect because speed less than 1800 rpm indicates the Main Generator is no longer connected to the grid, and speed lowering Iindicates no steam suppl for an unloaded turbine, it is not the definition of confirming the turbine Iri .
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[Question Topic: RO 53 Given the following conditions:
- Unit 2 is operating al 80% power, steady state.
- Power was reduced 2 days ago when 21 Condensate Pump tripped.
- 21 Condensate Pump remains DIS.
- The Condensate Polisher is in service with full flow.
Which of the following identifies the initial concern if 22 Condensate Pump were to trip, and the action which would be performed in response to that Lowering SGFP suction pressure. Open 21-23CN1 08 Polisher Bypass Valves. Lowering SG NR level. Initiate rapid load reduction at up to 5% I min to <49% power. Lowering SGFP suction pressure. Initiate rapid load reduction at up to 15% / min to <30% power. 9/26/20111 Ability to (a) predict the impacts of the following on the Condensate System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal 0 leration: Loss of condensate umps 55.41 (4.10)The limit per AB.CN for 2 cond pump and 3 HDP's in service is 85%pow."f. When the second pump trips as in the stem above, the power limitation is 30%. SGFP suction pressure will rapidly lower. SG N R levels will lower. but the SGFP will speed up
'-"---_... ~~~~lC-..J and the BF19s will open, which will tend to restore level but degrade SGFP suction pressure more. The polisher is bypassed. then the HP heater strings are bypassed in an effort to restore suction pressure. A load reduction will be performed. but it will be to o 0" * *
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RO SkvSc~ap~r' I ". SRO Sk~~~;~p~r . .~ R()Sv~t~~JE~61JtigriList "*'1 "'SRaSy~te~IE~()lutioll List :1 i6JtlineChanQ~s' *1 lQueliltio-~-To~ IRO 54 Given the following conditions:
- Unit 1 is performing a load reduction at 1% per minute due to lowering condenser vacuum.
- With turbine power at 75%, OHA G-7, ADFCS SWITCH TO MANUAL annunciates, and the PO reports 11 BF19, SG FEEDWATER CONT VALVE, has swapped to manual. - The load reduction continues to 70% power. where it is tenninated.
Which of the following describes how this failure will affect 11 SG NR water level, and how will it be corrected lAW S1.0P-AB.CN-0001, Main Feedwater/Condensate System Abnormality? lower than all other SGs. Manually adjust 11 BF1 9 in the OPEN direction to lower 11 SG NR greater than all other SGs. Manually adjust 11 BFi 9 in the CLOSED direction to lower 11 SG NR level. greater than all other SGs. Place SGFP Master Speed Controller in MANUAL and lower SGFP speed. Unaffected SG BF19s will respond and move in the open direction to re-establish equilibrium conditions. lower than all other SGs. Place SGFP Master Speed Controller in MANUAL and raise SGFP speed. Unaffected SG BFi9s will respond and move in the closed direction to re-establish equilibrium conditions. IApplication I Salem 1 & 2 Ability to (a) predict the impacts of the following on the Main Feedwater System and (b) based on those predictions, use procedures to correct. control. or mitl ate the consequences of those abnormal 0 eration: _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _"""'"" Failure of feedwater re ulatin valves 55.41 (5,7.10) OHA G-7 will swap the valve to manual, and control will still be available to the operator. AB.CN states that if a BF19 has swapped to manual, you are to establish control over the valve and stabilize SG NR level lAW SG Programmed Level (Att. 2). The valve failing as is at a higher power level that the other BF19s are currently controlling(75 vs 70 %j will cause 11 SG NR level to rise, since feed flow is > steam flow. Lowering SGFP speed would work as described. but it would not be done to pertubate the ~+/-.:c . . Refe[~!1c~.TI~m~ ((f.~~F!~fef~nce.Numb~~~ijtere~t:esE;Ctl<>riY]1 ~~l§J rR~ IMain Feedwater/Condensate system Abnormalit lis 1.oP-AB.CN-0001 II 1112 I 118 I I I II II II I I Ii II II II I ~NUl11ller . ell IADFWCSE012 I I I I I
- Mat~~ial Require~forExa'!ljnati()n II
[Ques=i~nsource: i j iQlJe~tio~M()diiicationMethOd:-:1 I~Jrln~TrainingProgrcn~j D I-N...;.e-w--------I1 I " [Question Source Comments I I i
ROSk';;Str~pe{ It SROSkVscr~p;r '1~()sJstemiE~()flltjbWLi~rH r,~ sROSv:ten;;Evdluli({I1'list' 11...~Jtil?~Ci;;n~E!~.1 [§~tiOr;T08§] Given the following conditions: Unit 2 tripped from 100% power.
- Neither MDAFW pump started or could be started.
- Total AFW flow Is 23E4 Ibm/hr.
- SG NR level on all SGs is 14% and rising slowly.
Which of the following describes the effect on 23 AFW pump when the PO lowers AFW flow to the SGs by throttling shut the 21-24AF11, AUX FEED - S/G LEVEL CONTROL VALVES?
~:II 061 000K503 !'ijystem/EvolPtiQh:;J;itie! j Auxiliary I Emergency Feedwater System 11 061 I M)3tl:iteroenill Knowledge of the operational implications of the following concepts as they apply to the Auxiliary / Emergency Feedwater System:
Pump head effects when control valve is shut IExpl~flat,@lof 55A1(5,4) 23 AFW pump Terry Turbine has its governor set to maintain a certain speed, not discharge pressure. As the AF11 i 'AI')~w:erS:!F>T.*m*: valves are throttled shut, the discharge pressure of the pump will rise, and remain at the new higher pressure as less work is
"-"":=~~~0L required of the turbine. A and B are incorrect because speed demand will remain constant, and discharge pressure will rise. C is incorrect because discharge pressure will rise.
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;.9.~~sticmTopiCJ IR056 Given the following conditions: - Unit 2 is operating at 100% power. - 2C Vital 4KV Bus is aligned to 24SPT (breaker 24CSD closed). - Power is lost to 2C Vital 125 VDC Bus. - Prior to restoring power to the 2C DC Bus, 24 SPT is deenergized.
Which of the following describes the status of 2C 4KV Vital Bus for these conditions? [J IL"""~'L"U from the 2C EDG. I IDeenergized with all in-feed breakers tripped. I IEnergized from 23SPT (breaker 23CSD closed). I fdll Deenergized with in-feed breaker 24CSD closed. I
- ij\hsvy¥rll d 115~~ltlte\j~I! IR I if::99I1ifiv~,t:~~~L~ I Comprehension I I~~~i.!ify:,i [Salem 1 &2 I 'l=y;~()ate:'11 9/26/20111
~:il 062000K103 1~93 1§9\Jalue:fI13.5I[S130~y~,~e;l14,01~tiql"l;1~~.Gro~EiJOlfRgiG~O)l~1 11 ~III 0 55.41(7) DC power is required to operate relays and contacts for the 4KV vital bus breakers. When DC power is lost, breakers will remain "as is". The EDG breaker can not close onto the bus even though it is deenergized because one of the interlocks to shut the L--"~~'-""--=c:::.= EDG output breaker is both infeed breakers open. The other (23) SPT cannot close its infeed breaker to the bus because it has an interlock which requires the other SPT's in feed breaker to be open.
IDCELECE013 l___ --l II 1()1I....~i~rlSqur~~;d IPrevious 2 NRC Exams IIQuestiO~~()~ific~t!ottMetho(j:'l Editorially Modified 1 iLJsed[)uri~g Training Progr,:ltllQ 0 jQuesti()risourcec()lnmentsll"J" ILOT RO NRC Exam - August 2008. Added "breaker" and "closed" to original distracter b, and swapped places I
-.-~ ...- - - . - - - - with distracler c due to its being longer now. Added words to lessen confusion and make it the same style as found IComrne l1 t .... ."
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,~'~~~~~~I~R~O~57____________________________________________________________- J Which of the following choices identifies an adverse effect of a ground on a 125VDC bus/battery, and the method in which operators perform ground isolation lAW S2.0P-SO.125-0004, 125VDC Ground Detection?
on the bus causes a higher level of current to flow in the system. Individually deenergize each load on the bus, then re-energize if that load is not the source of the on the battery associated with the bus causes a higher level of current to flow in the system. determine if the I/S is the cause of the on the bus causes voltage reading on the bus to become unreliable due to the excessive current flow. Transfer to the backup battery charger to determine if the lIS charger is the cause of the ground. on the battery associated with the bus causes voltage reading on the bus to become unreliable due to the excessive current flow. Individually deenergize each load on the bus, then re-energize if that load is not the source of the ground. jA;;wGr'j ~ ~~ ~ Memory ~1& 2 1 ~--:moate; I 9126/20111
~i11063000A201 1~~J~~@"~~Iill~§illJI~~~D~EB]DJ ID.C. Electrical Distribution Ability to (a) predict the impacts of the following on the D.C. Electrical Distribution and (b) based on those predictions, use rocedures to correct, control, or miti ate the consequences of those abnormal operation:
Grounds 55.41(7) The ground detection procedure has operators isolate individual loads. The ARP for low battery voltage has operators transfer to the standby battery charger if bus voltage is low, and battery current is pr<~sent. so these are plausible distracters. The
~~-""-_.,~~~"'..cJ bus voltage is higher than the battery voltage, so a ground on the battery would not cause bus current to rise.
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ROSkvScr~~~rl ,( SRQ Skvscra~~r I 8ue~tionlePicJ Given the following conditions:
- Salem Unit 1 is in MODE 3, NOT, NOP.
- A 500 KV grid disturbance causes a SEC MODE II actuation.
- During the electrical bus realignment, the 2A 4KV vital bus to 460VAC bus breaker trips.
Which of the following describes how this will affect 2A EDG operation, if NO corrective action is taken? 2A run until 22 Diesel Fuel Oil Storage Tank is empty. trip because its fuel oil supply pressure was lost when the EDG Fuel Oil Pumps lost power. trip because its lube oil supply pressure was lost when the EDG Lube Oil pumps lost power. run until its Fuel Oil Day Tank empties due to the loss of power to the Diesel Fuel Oil Transfer Pumps.
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[i<AS~ten1e~ Knowled(le of bus p_ower sup~ies to the following: I Fuel oil pumps J 55.41 (7,8) A is correct because there are 2 Diesel Fuel Oil Transfer pumps, powerec from A and B Vital 230V, each of which supplies fuel oil to all three EDGs. The loss of power to 2A DFO xfer pump, even if selected as the Lead Pump, will not affect EDG operation as the second pump has power from 2B bus, and will start on 10 level. ThE, DFOSTs have cross connect capability, but the cross connect is always shut. B is incorrect because the EDG Fuel Oil Pumps are shaft mounted, mechanically driven pumps. (,';" .h "h' +~. . *0, _ ,h "h' -h ;" ;n . ",h"n +h "'nr,,' ,h f. .. I=n~ starting. the Lube Oil pumps for operation are shaft driven" mechanical pumps. D is incorrect because there is stili power to one 1DFO xfer pump. which will fill all the EDG Day tanks based on level signals.
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- Unit 2 is operating at 100% power when OHA C-1 GAS ANL Y TRBL is received in the control room. - The NEO sent to investigate reports local alarm 8-3 OXYGEN HIGH/LOW on Waste Disposal Gas Analyzer PNL 110 is in alarm. - Local indication for in service Waste Gas Decay Tank (WGOT) 02 concentration is 4.1 %.
lAW Tech Specs, which choice describes what action is REQUIRED to be performed, and why? IPlace the Standby GOT in service and commence preparations to release the affected GOT. I IPlace the second Waste Gas Compressor in service to raise the total volume of gas in the WGDT in order to dilute the 02 concentration to l.below2%. I
! Reduce the oxygen concentration of the in service WGOT without delay to prevent potential releases of radioactive materials due to explosion of the GOT.
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@J IImmediately suspend all additions to the in service WGOT since the 02 concentration is above the 2% required to sustain combustion when I mixed with hydrogen.
~rl r=:J i~x<lIl1Llllvel*.1 m==i ~ltive'~eyelIll Memory I !F~f!lifY:1 ~1 &2 I ~~aJ1,9~te:\I1 9/26/20111 ~1071000K504 1~o4l~~@~~[I2J~~~~GWtJBJO"[S"'QGrQ~:[] IIl~ c I 55.41 (13) Tech Spec 3.11.2.5 requires the 02 concentration in the waste gas holdup system to be <2%. The action for 2-4% 02 is to reduce the 02 concentration to <2% in 48 hours. Distracter 0 is incorrect because while the immediate suspension of additions to the waste gas system is REQUIRED when >4% 02, the flammability percentage is wrong. release. C is the correct answer because the Tech Spec REQUIRES the reduction of 02 from 2-4% to less that 2% without delay. Also, the bases section for this tech specs describes that a potential explosion and release of radioactive materials from this explosion would not be lAW GDC 60, 10CFR50 Appx.A. Distracter A is in because TS ~~~~~~~~~~~@g~::~~~~~~~~~~~~~~~~~~~~~~~~~~=2==~~'~ ~==============='\=========~F====::::'F==~ 11=28=2====, I==================::::'I=============~F====:=~F=~~I~ I------------------------__________~ -------------------------~ ---------------~ ------~ I----~
I~~~~~n~~the following Area Radiation Monitors (ARM) will cause a ventilation system alignment change when it reaches its High Radiation Alarm I
~j 12R1A' Control room.
12R9, New Fuel Storage 12R32A, Fuel Handling Crane. . 12R52, Liquid PASS Room. I~~~~ EJ" ~tTl~ [K:J ~niti'lfl~vei'd IMemory I~ ISalem 1 & 2 I !~Il~TD~ I 9/26/201~J
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~- ' charcoal filters and starts both FHB Exhaust fans. C is incorrect but plausible because its auto function is to prevent Fuel Crane motio~ except in downward direction. Dis incorr:ct since it only has al~rm light outside the PASS room whi?h activate,S, but l'
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- 9ue~!fonToPicll RO 61 Given the following conditions:
Both Salem Units are operating at 100% power.
- 2R1 B-1, Unit 2 Intake Duct Rad Monitor Channel 1 loses power.
- BOTH Salem units Control Area Ventilation (CAV) systems remain in NORMAL Mode.
Which of the following identifies the status of the CAV systems. and how will the Control Rooms respond to the loss of power? are designed to remain in Normal Mode upon a loss of power to a single duct radiation monitor. No actions other than normal Corrective Action Program actions to troubleshoot and repair the power supply are required, with no speci:1c time limitation. are designed to remain in Normal Mode upon a loss of power to a single duct radiation monitor, The 2R1B-1 must be restored to operable within its Action or CAV must be in Accident Pressurized Mode on BOTH Units. should have swapped to AP Mode. Initiate AP Mode of operation by depressing initiating pushbutton on 2RP3 lAW S2.0P-SO.CAV-0001, Control Area Ventilation Operation. Initiate AP Mode of operation by aligning individual system components to their correct positions on 2RP3 Application Salem 1 & 2 55.41 (7) Any R18 channel losing power will automatically initiate Accident Pressurized Mode on 80TH Units. B is incorrect but plausible because the Tech Spec for R1 B says if one channel is inoperable, you have 14 days to restore before placing CAV in AP Mode. C is correct because manual initiation of AP Mode is accomplished by depressing the Accident pushbutton on RP3. 0 is incorrect because individual components are not aligned, still plausible if the candidate thinks the failure has affected the whole
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- Unit 1 is operating at 25% power.
- 1B EDG is running in parallel with station power on 1B 4KV Vital Bus.
- 13 and 16 SW pumps are in service, 11 SW pump is in AUTO.
1A 4KV Vital bus becomes deenergized due to a Bus Differential signal. 1 minute after the 1A 4KV Vital bus deene izes NO nne,""t,,," action which of the tr'IlI'~wir"" contains ALL the SW 115, 16. 1.qJ I076000K201 li K2 *01 1~]J§~~@f~MI~fiO~O!sROl3r~O
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1.~'~,1 1Service Water System 11 076 I ISAStat!.ment:l Knowledge of bus power supplies to the following: , Service water I I ,~ p.. ,.I,..,*.*.~r.,. :.t'.o.[I.**.'.*.O.*..f. I 55.41(7) A bus powers 15, and 16 SW pumps. I.E.,
*.. .**.. , .a. On a single bus UV as described in the stem, only that bus would load in blackout IAnsw:er!i':"~ loading. A bus is locked out on Bus Differential (deenergized), and the loss of 16 SW pump would cause header pressure to lower to where the auto pump (11) would start. Only one SW pump is aligned for AUTO which is the normal at power configuration for the SW pumps, one In auto, and the rest in manual. 12 SW pump would never start unless 11 pump did not on a SEC initiation, that is "h, ;t ;" nnt 1,,+0'-; ;n "n' nf tho -hni ,c 1 A <::11\, n, ,no" "'" ,1'-; ""I "bel cinC'o R h, ," n"","'" I"""" "A' ",," >Am;,.." 1<0 >Am" It ;cn" Iido,-i In any of the choices. There can be confusion about the running EDG and the loss of A vital bus' causing a MODE Ii (Blackout), which I would strip busses and load the primary SW pump on each bus.
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I Service Water Pump Operation II S1.0P-SO.SW-0001 II II IIF26=~ IUnit 1 4KV Vital Buses One line 1[203002 11 ) I IF I 3=4==.. I-------~I 11 ___----111 11_---' j~N4m6~~:n I SWBAYSE005 1_-'-------' iQue~ionS~Clurce:] IFacility Exa~m:::::B=a=nk:::::=:!ll:::g=u=e=st~)~on:::'=M=.?::d:::ifi:::IC:,::.a:,::t:::io::n=M:::*::et::h::o::U::,:=.* '=':::S=i:g=ni::::fic=a=n=tl=y=M=()=d=ifi=ed=:::::I.:ru~s:e::.d=D=u:r=rn=g=T::r:a=.in=i=ng=**::p::l".o=g=r=.a=~=.*~~!1 I l':~e~Sourcl:con:!l1entsi J-ROC61 Changed which pump is in auto (11 instead of 15) and that makes a different choice correct. I ic\,....... ~.. ... j *
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!Q':;:esti~n TOP5J IRO 63 Given the following conditions:
- All 3 Station Air Compressors have become unavailable.
- The NORMAL cooling water supply to the Unit 1 Emergency Control Air Compressor (ECAC) has been lost.
Describe the status of the Unit 1 ECAC. o eralion of the Unit 1 ECAC ... can continue since cooling water wiJI automatical1y swap to Demineralized Water through a check valve. must be discontinued until cooling water can be manually aligned through a spool piece from Service Water. must be discontinued until cooling water can be manually aligned through a spool piece from Demineralized Water. s between Instrument Air System and the following:
, 55.41 (4)The normal source of cooling water to the ECAC is the Chilled Water system. Upon a loss or unavailability of the Chilled Water system, SERVICE WATER can be supplied by MANUALLY installing a supply and a return spool piece. Demin water is plausible because it is as a backup cooling system for other systems (SI and charging pumps when normal cooling is lost.)
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- Salem Unit 1 is operating at 100% power.
OHA A-15 FIRE PUMP 1/2 RUN, and OHA A-23 FIRE PUMP 1/2 TRBL alarm in the control room.
- NO other fire system alarms are received.
- An NEO dispatched to the Fire Pump House reports fire main header pressure is 132 psig, and both fire pumps are operating.
- Fire Protection reports there are NO fire system actuations.
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.a .**.*..~*..*n**. *.o{11 55.41(7}Fire pumps will BOTH auto start upon a loss of power to their battery chargers. Losing the power from the ATS, supplied iE.X. PI.a.h.'..ti.:O
,Answers:,,,, . from #1 and #2 Misc yard panels will cause loss of both battery chargers. Distracter D is incorrect because a momentary drop in
~------ pressure will start the #1 pump, and the #2 pump has a time delay. Distracler A is incorrect in that the #1 pump will start as header pressure lowers, and the #2 pump will not start. Distracter B is wrong because a major pi ping rupture will cause header pressure to
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Given the following conditions:
- Unit 1 is operating at 100% power.
- Charging flow rose 1 gpm in 5 minutes and is now steady at 88 gpm.
- Computer trends show the increased RCS leakage started 10 minutes ago.
The CRS enters S 1.0P-AB.RC-0001, Reactor Coolant System Leak. Based on elevated 1R11A indications coincident with the rise in charging flow, the crew determines a small RCS leak in containment is occurring. Which of the following describes an action that will be performed lAW S1.0P-AB.RC-0001 based on the determination of the leak location and size, Place 2 CFCUs in slow speed and 2 CFCUs in high speed to prevent containment pressure from rising to the automatic Safety Injection setpoint. Place a centrifugal charging pump in service to ensure sufficient charging flow margin is available to maintain PZR level stable should the leak rate rise. Place 2 CFCUs in slow speed and 2 CFCUs in high speed to minimize the rise in containment humidity and prevent equipment damage from elevated moisture levels. Place a centrifugal charging pump in service to allow additional flow through the Mixed Bed Demineralizers to minimize the potential off-site release from containment. 1103000A101 I~~ [@~@!sR6V~iUe:clB~@lliJl~ ~~u~~llilD ISyst~l~tt~ l~c~o~n~ta~in~m~e~n~t~S~y~s~te~m~____________________________________________________________~ L -_ _ ~~ [KA St",fementJ Abili to redict and/or monitor chan es in arameters associated with 0 eratin the, Containment System controls including: Containment pressure, tem eralure, and humidit
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,c,.,.*. f.* 55.41 (9) AB.RC-1 Attachment 1, Continuous Action Summary, 4.0, states that at any time if Ie.ak is suspected to be in containment, Answers;', .:,.' then place 2 CFCUs in slow and 2 in fast. The Technical Bases Document, page 3, states that this is "to prevent an automatic -. ~- Safety Injection and minimize the potential for off-site releases when the leak is in Containment." B is incorrect because while it has the right action, the reason is wrong. Distracter C is incorrect but plausible, because a centrifugal charging pump is placed in incorrect because of the wrong reason for CFCU operation. D is incorrect as per B above: while additional flow through the demins I would reduce RCS activity levels, it is performed in AB.RC-2, Hi h RCS Activit. ~.*~~~~nR~~~~~~~0~~~~~~~~~~~~~~~~~~~~~~~~~~S~
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- I IOf the following. which would be considered a Core Alteration lAW S2.0P-IO.ZZ-0007. Cold Shutdowll to Refueling? I I When the first stud (during first pass of Reactor Head detensioning process) is detensioned.
I I When the RPV Head is lifted (1-2') to check for CRDM drive dis-engagement. I l~ I Insertion of a camera to the level of the RPV flange prior to fuel movement. I I Lifting of the first fuel bundle in the RPV. I
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55.41(10,13) D is correct because "CORE ALTERATION shall be the movement of any fuel. sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel." A is incorrect because it is when w;;cc.::..:::".::'::~'~'~ MODE 6, Refueling, is officially entered. B is incorrect because Core Alts cannot occur until the vessel head has been removed. C is !ncor~ect because "Refueling. activities that would constitu,te a CO~E ALTERATION do ~OT include the movement or components, except wihen these d-evices would result in the movement or manipulation of fuel. sources, or reactivity control Icomponents within the Reactor Pressure Vessel. [~~o:nJ l===========:::!~I==========~:=======4'r!====; 1:=16=~ ~~~~======:~~~~~~~~~~======+:==:.::; 1~~~~~~____________~1-~~~~~------~1----------~----~1-31---~ 1:=17==
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~es~onI6picll RO 67 i Given the following conditions:
- Unit 2 is operating at 66% power.
- Power was reduced when 22 SGFP tripped 2 days ago, and has remained at this level for 2 days.
- Operators are preparing to start a Main Turbine power ascension using dilution and automatic rod control to maintain Tavg and AFD on program,
- Tavg-Tref deviation is O°F.
lAW OP-AA-300, Reactivity Management, how should the power ascension be started? Assume this is a normal power ascension with all equi!2ment available.
~ IThe crew concurrently initiates the Main Turbine load ascension and a ReS dilution.
I I The crew initiates a ReS dilution. As soon as the dilution is in progress the Main Turbine load ascension is initiated. I I,The crew initiates a ReS dilution and waits until a ReS temp~rature rise is detected. Then the Main Turbine load ascension is initiated. I I The crew initiates the Main Turbine load ascension and waits until a ReS temperature lowering is detected. Then the ReS dilution is initiated. I
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~,"_,_.~._:.:::il Knowledge of procedures, guidelines, or limitations associated with reactivity management.
15X1?!~~,~tJO~ 55.41(5,10) Listed under the responsibilities of the Reactor Operator, page 8,4.6.5, "Typically, during planned load changes where I~~' ,,," .. i~"~,~*,l dilution or boralion is required, start with the dilution or boralion. The initial effects (ReS temperature change) of the reactivity change should be seen prior to initiating the load change." All the distracters have the correct actions in the wrong order. I t;\~,;ij!i;;fi:jC~i 1i!I~;,'i'i,'z.'ii: * , 1 t;j' . t',C3£'.~~~y~;,~.~."'*.~",** ~ un ",,:,1 [R~f'@";';;:;!;'< '" ''.'], IIJ~ag~ N6;eJ rs.evi~.M I RA:.ldivif\, m","",),j-""- len', Ilgp-AA-'lon II 18 114 I II II II I II I! I
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Given the following conditions:
- On your shift today, a component's normal monthly surveillance item is determined to have been scheduled incorrectly, and it has been 34 days since it was last performed, The component must be declared INOPERABLE at the time of discovery because the 24 hour tme limit allowed past the normal surveillance interval has been exceeded, The component remains OPERABLE because Technical Specifications allow a 25% time extension of the normal surveillance interval for surveillance which has not been exceeded.
The component remains OPERABLE ONLY for the next 24 hours after discovery, during v.hich a SAT surveillance must be performed to ensure OPERABILITY, otherwise the component must be declared INOPERABLE. The component must be declared INOPERABLE at the time of discovery ONLY if any redundant structure, system, or component (SSC) is also INOPERABLE for the system in which the affected component is required to be OPERABL.E.
~st~e~ent:ir- ______________________________________________________________________________________ ~
Knowled e of surveillance procedures. 55.41(10) Tech Spec 4.0.2 states that... "Each Surveillance Requirement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25 percent of the specified surveillance interval." Tech Spec 4.0.3 states that. ,," If it is discovered that a Surveillance was not performed within its specified frequency, then compliance with the requirement to declare the Limiting Condition for Operation not met may be delayed, from the time of discovery, up to 24 hours or S*urveillance. If the Surveillance is not p'erfonmed within the delay period, the Limitin\l Condition for Operation must immediately be declared not met and the applicable Actions must be entered." A is incorrect because the 25% "Grace Period" (which is 7.75 days for a 31 day monthly surveillance) has not been exceeded, C is incorrect (and different from Distracter A) because the 24 hour time limit is incorrect AND it says it would be applied from time of discovery, not added to the normal surveillance interval. D is incorrect because there is no requirement to check other sse of that system with regards to how it affects the component which has exceeded its monthly surveillance interval.
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RO skvsc~~perlisRQ'Skvscr~b~r L~~~~~?E~P;~ll RO 69. Given the following conditions on Unit 2:
- Reactor power is 75%.
- A RCS leak rate surveillance indicates the following:
- Total Corrected Volume leak rate is 4.0 gpm.
Leakage to PRT is 2.0 gpm.
- Leakage to the Reactor Coolant Drain Tank is 1.3 gpm.
- Total primary to secondary leakage is 0.125 gpm.
Which one. if any. of the following Technical Specification leakage limits has been exceeded? Knowied e of limitin conditions for operations and safe limits. EXP.* Ia.n.a*.*.*.t.*'.o .* .*.*.**.. . o. . ... f.*.~.* 55.41(5,10) Modified stem conditions to make a distracler (primary-to-secondary)correct, and the former correct (unidentified)
.*.* .*.*n Answers:*.* *.* answer incorrect. Salem TSAS 3.4.7.2 states the limits on RCS leakage to be 1 gpm for unidentified leakage, 10 gpm for identified
' . ,-..' leakage. and 150 gallons per day through anyone steam generator. The 0.125 gpm prHo-sec leakage is 180 gpd. PRT leakage is identified. so while it is >1 gpm, its limit is 10 gpm. RCDT leakage is unidentified, so since it is >1 gpm. its limit is exceeded.
ITECHSPE015 I I I I I [Material Reguired for Examil'lati()n iI II iQuestio~Source: ! IFacility Exam Bank Il9~estipl)NlOdificatjon Metho.d:il Significantly Modified IQuestionSciurc:e C.t:)/11:r§ri§ Salem 2002 RO NRC Exam (4 NRC Exam I ) I
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[QUIi;rtionT6p-iCll RO 70 Action? 26 SW pump is crr. ALL other SW pumps remain OPERABLE. performing a shutdown. 2N32 Source Range NI control power supply fuses blow. 21AF21 AUX FEED - SIG LEVEL CONTROL VALVE is discovered in the jacked shut position. 2PR3 PZR Code Safety Valve is declared INOPERABLE with 2PR4 and 2PR5. PZF~ Code Safety Valves OPERABLE. IApplication I fFaci'ifir;11 Salem 1 & 2 arameters which are ent -level conditions for technical s ecifications. 55.41(10) In MODE 4. only ONE PZR Code Safety is required to be OPERABLE. 3.4.2 In MODE 1,2 independent SW loops are required. An OPERABLE SW loop consists of one pump from each vital bus plus 2 pumps per bay. 3.7.4. SO.SW-OOOB. Att 2 page 1 of 36. In MODE 3.4.5 only ONE SR is required. 3.3.1.1 Table 3.3-1.6.b. ALL 3 AFW pumps and flow paths are required to be OPERABLE in MODES 1-3. 3.7.1,2 lill~l~+i;++:;::l.ii . . R~erer1~e+Tijle';:i"(i+'§ll1lf~~n~~ ~~~~~~~ ~~
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Given the following conditions: Units 1 and 2 are at 100% power.
- Unit 2 has experienced several fuel pin failures.
A leak must be repaired on a pipe in the Aux. Bldg. pipe tunnel. The general area dose rate in the location of the repair is 600 mrem/hr.
- In order to reach the location of the repair the worker must transit through a 6 Rem/hr high radiation area for 2 minutes and return via the same path. - The worker currently has an accumulated annual dose of 400 mrem.
Calculate the MAXIMUM allowable time that the worker can partiCipate in the repairs AT THE JOB SITE and NOT exceed the Current Annual TEDE
~S~~r- ____________________________________________________________________________ ~
Knowled e of radiation exposure limits under normal or emer enc conditions. 55.41(12) B is correct because: The Current Annual TEDE Administrative Dose Control Level is 2000 mrem for PSEG Nuclear LLC. Transient exposure is 400 mrem (6000mrem/hr x 4/60hr). (transit to and from the job). (Current) 400 mrem + (transit) 400 mrem :: 800 mrem. ADL of 2000 mrem 800 mrem = 1200 mrem allowable before maching the Annual TEDE Administrative Dose Control Level. 1200 mrem 1600 mrem/hr = 2 hours. A is incorrect, based on using limit of 1500 versus correct ADL (2000); C is IControl Level (2000) and NO transit dose. IRADCONE002 1_ _ _---1 IL Material RequIred for EXamination:':! I II lin,. . . ..~~urce::11 Facility Exam Bank Iq~e~~:ipl'Mod,ificatio~M~~J~ct From Source
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[Questio:nr;;p;~i IRO 72 Given the following conditions:
- Unit 1 is in Mode 2, with all Shutdown Banks withdrawn.
- Unit 2 is at 100% power
- All Unit 1 & 2 Circulating Pumps are in service 21,23 & 26 Service Water Pumps are running 21 cves Monitor Tank is being released via 22 CCW Hx to the Circulating Water System Choose the condition that would termination of the release.
~1194001G311 I~~~=I~~0"~~fJlID~l~ ~~]Dl~Fi~~r~]U
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Ability to control radiation releases. I.E.X. p.. p. ~lj.o.*.*.n.i'9,..f....1 header
' .8. ~sW&rs;.1..c 55.41(13) S2.0P-SO.WL-0001 states at step 5.5.9 that the released is to be terminated upon loss of dilution water flow. 22 SW discharges into the outlet of 1.1 A and 11 circulators. A Unit trip B ~ ~~es not co~stitute an "emergency", so the requirements of OP-SA-106-101-2001 Operating With an Emergency on the OpPosite Umt IS not applicable. Normal SW header pressure IS 105 125, the SW flow from the remaining 2 pumps in service will maintain flow to the CCI--lX. Trip of an operating CCW pump would I r"""lt in ..,,.., ""t, d-ort ,f tho ,t* ,~h n ,,,, n ..,,..,rl th" ,Y'\A, c"eto"" ro,." ",in" in coni,.."
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I RELEASE OF RADIOACTIVE LIQUID WASTE 1\ S2.0P-SO'wL-0001 II II 1124 I i II II II II I I Ii JI II II I
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LQuestion Topic IRO 73 I Given the following conditions:
- Unit 2 has experienced a steam line break inside containment. - Operators have entered 2-EOP-FRTS-1, Response to Imminent Pressurized Thermal Shock.
WhZ will the operators be instructed to terminate SI and start a RCP if possible? IThe soak required by FRTS-1 requires SI to be secured and a RCP to be running to provide ability to use spray to depressurize primary. I I.The soak required by FRTS-1 requires SI to be secured. A RCP should be started to equalize boron concentration throughout the primary to ensure proper shutdown margin as the ReS cools. J ISafety Injection flow is a significant contributor to any cold leg temperature decrease or overpressure condition and must be terminated. A Rep is started to minimize temperature gradient across S/G tube sheets. I ISafety Injection flow is a significant contributor to any cold leg temperature decrease or overpressure condition and must be terminated. A Rep is started to provide mixing of cold SI and warm reactor coolant water. I
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IKnowledge of the specific bases for EO?s. 55.41 (10) A is incorrect because the purpose for RCP operation is not the priority in FRTS-1, and. the soak is not the basis for terminating SI. B is incorrect because the soak is not the basis for terminating SI. 'C is incorrect because the 81 basis correct but the basis for Rep start is not correct. 0 is correct because Basis Document states for step 9 starting a Rep ... "in order to mix the cold incoming ECCS water and the warm reactor coolant water, and therefore decrease the likelihood of a PTS condition, an Rep
* " . . * * * * * **
- f1 Ihave contributed to the ReS cooldown or mat prevent a sUbseguent reduc1ion in ReS -pressure.~
IR~Yii?iqJil ~~~~~~~~~=~~~~~===~~====~~~=.:;125 , ~==============~~==========~:========~~==~I 1_ _ _ _ _ _ _ _ _ _ _ _- ' _ _ _ _ _ _ _ _ _ *...JI~ _ _ _*_ _....J 1_ _- - ' I I, ~erial~eqlJiredforE~~mlnafiC,!'111 ,.I
Given the following conditions: Unit 2 is operating at 100% power. OHA A-7, FIRE PROT FIRE, annunciates. Panel 2RP5 is checked and indicates the following:
- Zone 59 - Air and Water Deluge, Containment EI. 100 Panel 335 is lit.
- Zone 74 - Smoke and Fire Detector, Containment EI. 100 Panel 335 is lit.
Which of the follOwing describes how the Control Room crew should respond? Verify OHA A-15, FIRE PUMP 1/2 RUN is in alarm signifying a Diesel Fire Pump has auto started to supply Fire Protection water to associated valves in containment. Dispatch a NEO to open the associated deluge valves at their outside containment penetration location. Dispatch a NEO to open the 2FP147, FP CONTAINMENT IV.
@illiopen the 2FP147 from the control room.
iAn~V{~'(', Id I EXclrll"Lj[v;U IR 1 IMemory ISalem 1 & 2
~§thle~~n~~ ________________________________________________.____________________________ ~
Knowiedoe of "fire in the plant" procedures. i IE.,..'.x.,.*.**.p. . 1,.~j1.~.'. t.)#.***.n. .,...*".* *.p.*'.f.,.'.1 55.41(10) Overhead alarm A-7 states that if both zones 59 and 74 are received, then open 2FP147. This is controlled from the IIAnswers:**'. .* ""I control room on 2RP5. B is incorrect because the deluge valves are automatic and in containment. C is incorrect because the
. . FP147 is opened from the control room. A is incorrect because the Fire Pumps will not have auto started since the FP147 has to be opened, and is not an action contained in the ARP.
I No.1 & 2 Units Fire Protection 11205222-2 II I f::l ==~I 1 60 '~lo=v=e=rh=ea=d=A=n=nu=n=ci=at=o=rs=-=W=in=d=ow==A======~1~ls=2=.o=P=-A=R=.=ZZ~-=0=00=1=========;I~I============~'1 22 IFI5=1==~ 1_ _ _ _ _ _ _ _ _11 _ _ _ _ _ _1[ ____----111 11_----' l'::2:B~~.:=:=DJ 1 FIRPROE004 I I FIRPROE008 I 1---
~Mns~~:II~cil~~~a~m~B~a:n~k=~II~A~~~~~~~i~6~ri~~~~O~~~i~fi~~~1~fu~J~.~~~~~ili~O~9~j~.=]~rE:d:i:~:ri:a~~:M=~:i:fie:d==:I~~~.~~~~~*~D~~~~~ri~9~T=~~i~n~~~g~p~j~b~g~~~~~..~i~D~ !Question ~oG~ce C9mmehts]I - - ....~----..- - - -.. I.system. Modified chOices and added valve Identifiers.
Vision Q6980? modifi.ed stem to make it w~at w?uld control room crew do instead of what is the status of the FP I 1 C..... III1""".:, ...... ..>.. ' ' ', '.' '" , :". . ..":J"", t II J II 1 II J.
RO SkyScraper .'. 'sR6SkVs~rap~r 'I+R6s~st;fuIEJ&IJtidnList'?l ...*.. sif6'g~~t~irifEv~iJtl;nIi~t' Ij;lutlirle Chanrt~~1 [~~;.uo~~1 I=R=o=7=5======================::====::====::======================::============::========~1 An Alert has been declared at Salem. Which of the following identifies the PRIMARY method which the Primary Communicator will use to make notifications to the States of Delaware and New and how long from the Alert declaration do they have to make those notifications lAW Attachment 6, Primary Communicator of the I NETS phones within 30 minutes. I ESSX phones within 15 minutes. g:; I ESSX phones within 30 minutes.
~SW~(11 a I iExam~eveJ] IR I ~I194001G443 12;::.4=.~=3==*=m]==::::::====~~==:::::====::=========~.
I----------------------------------------------~~. . . --~ communications s stems and techniques. 55.41(10) Salem ECG, lists the communications systems in order of preference. The NETS (Nuclear Emergency Telecommunications System) is the primary closed circuit communication system for off-site notifications. The ESSX is also a closed circuit system, which is used as a backup for NETS. The notifications to the Sates must be made within 15 minutes of the declaration of an Emergency. [..". :~;, ~~fereo~~Jjtle:El ~f;~.J~cilit~~eT~rElric~Nul11!'eI'iX .~ !R:ef~r~ri§e'se(~HOIl *'1 i*f'~g~~od ~~ ISalem ECG - Attachment 6 II r 1~5=5=:::; I
, Emergency Preparedness Training Communicat II I
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I [9uestiofl TopiC] SRO 1 I Given the following conditions:
- Unit 2 is in MODE 4, performing aRCS heatup and pressurization lAW S2.0P-IO.zZ-0002, Cold Shutdown to Hot Standby. - RCS Tavg is 210°F. - RCS pressure is 310 psig. - 23 RCP is in service. - 21 Charging pump is in service. - 23 CCW pump is crr. - 2A and 2C 4KV Vital buses are powered from 24 SPT. - 2B 4KV vital bus is powered from 23 SPT. - 24 8PT loses power, and NEITHER 2A nor 2C 4KV vital buses transfer to alternate power due to faulty power available relay for 23 SPT.
1- 2A vital bus locks out on bus differential.
- 28 EDG fails to start.
Which of the following describes how the control room will respond? TheCRS will. .. BJ Ienter S2.0P-AB.RCP-0001 and trip 23 RCP 5 minutes after the receipt of OHA's 020-023 21 (22,23,24) RCP BRG CLG WTR FLO LO. I Ienter S2.0P-AB.RCP-0001, Reactor Coolant Pump Abnormality, and trip 23 RCP based on the loss of seal injection and thermal barrier flow Ito Rep;;. Ienter S2.0P-AB.CVC-0001, Loss of Charging, and check 22 charging pump has started to supply adequate seal injection to allow continued RCP operation. I [Eill.enter S2.0P-A8.CVC-0001, and take actions to start 22charging pump prior to 23 RCP seal package temperature rising to the point where seal injection flow is prohibited and 23 RCP trip is required. 1A.15~;;;r;1 [CJ IE~#IT!;'b~vel i ~ 19()gnf!iv~L.~\iel II Application lIE~fi}iti;;I [Salen1 &2 IIE~~r,rP~!~:)i! I 9/26/2011J
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55.43(5)The initial electrical lineup, combined with the conditions occurring in the stem, will result in only 2C 4KV vital bus having power. With the RCS temperature < 312', all charging pumps except one are CIT. 21 charging pump is powered from 2S 4KV vital
~~~~ bus. NO CCW pumps will be running since the only powered 4KV vital bus CCW pump is crr. With the only operating (or av~ila~te) charging pum~ tripped, the::re is no seal inj.e~tion flow to RCPs and no CCW flow to RCP ~hermal barn.ers. The CAS 5 minutes of the alarms annunciating, go to Attachment 2, Tripping RCPs, the loss of both seal injection AND CCW directs Rep trip without any time delay. C is incorrect because while AS.CVC will be entered on the trip of 21 charging pump, 22 charging pump will not have auto started due to its power supply being crr. 0 is incorrect because while the action will be taken to restore a char in to available status, the CAS action of AS.RCP will be the nori ,and seal acka e temperature will not stop that action.
IRevision ! F===================~~F===============~F=========~~==~~121==~ ~==~==============~~============~F=========;F==~FI9==~ --------------------------------~I----------------------~--------------~------~ I----~ fL..o, Number IABRCP1 E003 1------'
ISRO 2
~eSU~n I<.>PiC] 1 Given the fol/owing conditions:
- Unit 2 is in Mode 5 with 21 Residual Heat Removal (RHR) pump and HX in service for cooling.
- The RO reports that Pressurizer (PZR) level is slowly lowering unexpectedly.
- NO Overhead Annunciator alarms have been received.
- Refueling Water Storage Tank (RWST) level is stable.
- 21 Waste Hold Up Tank level is rising slowly.
Which of the foliowinq identifies the procedure which will be used and the action(s) taken in that procedure which will isolate this leak? iI.S2.0P-AB.RHR-0001. Loss of RHR. Close 2CV8, RHR Letdown. I
~ IS2.0P-AB.LOCA-1, Shutdown LOCA. Close 2eV8, RHR Letdown.
I IS2.0P-AB.RHR-0001, Loss of RHR. Place 22 RHR pump and HX in service and isolate 21 RHR loop. II JS2.0P-AB.LOCA-1, Shutdown LOCA. Place 22 RHR pump and HX in service and isolate 21 RHR loop.
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55.43(5) The 2 AB.LOCA distracters are incorrect because AB.LOCA is applicable in MODE 3 with Accumulators isolated or MODE
- 4. With MOOE 5 indicated in stem, it is the wrong procedure, although it has the correct action in distracter B. A is correct because the 2CVa will be isolated and it is the correct action for a leak which is causing the WHUT leve to rise. Distracter C is incorrect because it has the wrong action, with the right procedure. The action in the AB.RHR is to remove BOTH loops from service and I ABRHR1E005 I IABRHR1E004 I I I
~c:ll Required~()rE)(afuil1atl~n .*.. J I '/ ~~~~e£J IFacility Exam Bank 1[§uestionMO;lifiP'1!ti0rt Method: ]~ficantIY Modified 1[USed During Trainini;IProgram! 0 IQues~()n SOlirceComment~ Isalem 2003 SRO NRC Exam. (5 NRC Exams ago) Modified to Include which procedure to use.
IIComment ..... ...*.... .... ................ , '.' i I I I
~estionrOPiS ISRO 3 !
Given the following conditions:
- Unit 2 is operating at 21 % power.
- Turbine load is being raised at 10% per hour.
- Operators observe the following:
- OHA E-28 PZR HTR ON PRESS LO alarms.
- PZR level is 28% and lowering slowiy.
Control Rods begin stepping out in automatic at 8 spm.
- OHA C-38 CFCU LK DET HI alarms.
- Rx power is rising at 0.35% per minute.
1 Which of the following identifies how these conditions will be addressed?
~-:--lEnter S2.0P-AB-RC-0001, Reactor Coolant System Leak. If VCT level cannot be maintained 11 %, swap charging pump suction to the RWST and trip the Rx, confirm the Rx trip, initiate Safety Injection, GO TO EOP-TRIP-1, Reactor Trip Response
[~ , Enter S2.0P-AB.STM-0001, Excessive Steam Flow. Trip the Turbine, initiate AFW. lower power to <5% lAW S2.0P-IO.ZZ-0005, Minimum
<< Load to Hot Standby, close all MSIVs.
IEnter S2.0P-AB-RC-0001. If PZR level cannot be maintained stable or rising. trip the Rx and initiate Safety Injection, GO TO EOP-TRIP-1. IEnter S2.0P-AB.STM-0001. Trip the Reactor. confirm the Rx trip, initiate MSLI, initiate Safety Injection, GO TO EOP-TRIP-1. ! .. r::;----j I",. <.~ I,. . ,,,<, ,,; ., & ,.,. IF.~cili~y:'HSalem 1 & 2 1I;xamDate:.U 9/2612011 I-=---...J I-=---...J r - - - -----1 r ~I 000040A202 lAA2.02~ [5<l,,{,arue,: IB~()'lp'U~[J3 ~!E.JI~ ~2Gr9~O ~RQc;r~l[J] m
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ISteam Line Rupture Conditions requirin a reactor trip f.l IE.<<.<<~. .PI~h.Ylati~n.**. O.<*.< 55.43(5) D is correct because the indications in the stem are of a steam leak in containment. Rx power is rising at -21% per hour, A~swerS::;<.1 while the turbine load increase is at 10% per hour. AB.STM states that if Rx power is rising uncontrollably. take the actions as
-- described in D above. A is incorrect because only some of the indications present in stem would be present for aa RCS leak. and the action is incorrect also. B is incorrect because the actions described would be taken if the steam leak were outside containment
~~~~~~~======~~~~~~====~~~======~~~~110==~ I--------------------~----------~I-----------------------~ ---------------~!-----~ I----~ l§It:~~~J INew I~~~~.ifiCi~~~J e.uesti~n Source c:ommentS] I
Given the following conditions:
- 21 CVCS Monitor Tank is in recirc.
21 CVCS Monitor tank will be released via the Waste Discharge Cross Connection to Unit 1 Liquid Radwaste system lAW S2.0P-SOWL-0001, Release of Radioactive Liquid Waste from 21 CVCS Monitor Tank. Which of the following describes an action which would result in an unmonitored Radioactive Liquid Release if it was performed AFTER the Liquid Release was and action to the release? Unit 1 CRS authorizes a tagging request which results in the 1 Ri8, Liquid Waste Disposal Unit 2 CRS authorizes a tagging request which results in the 2R18, Liquid Waste Disposal Rae: monitor losing its power source. CRS authorizes a tagging request which results in the 2FR1064, Rad Waste Liquid Monitor Pumps Discharge flowmeter losing its CRS authorizes a tagging request which results in the 2FRi064, Rad Waste Liquid Monitor Discharge flowmeter losing its source. ISalem 1 &2
~I000059G236 1[~[~~~w~~[EJ~I~I~~~O~~D1 ~~t~wl~'VljIJtiOnI!tle] 'Accidental Liquid Radwaste Release ~~tate~~htDr- ____________________________~________~__~~__________~______~________________~
Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations. 55.43{4) The Salem ODCM is a supporting document to the Unit 1 and Unit 2 Technical Specifications. The previous LCOs that were contained in Radiological EffiuentTech Specs are now included in the ODCM. Using ODCM 3.3.3.8, lAW Salem Tech Specs 6.8.4.1.g.1, and Table 3.3-12 on page 19, the RiB and the FR1064 required to be operable. The R18 is interlocked with the WL51 so that if the R1810ses power, the WL51 will shut, preventing an unmonitored release, and no operator action would be required to between the flow recoder and the WL51, so the release would be considered unmonitored based on the fact that a required component was not available, and the release is ongoing. Loss of the flow recorder during the release requires operators to close the 2WL51 lAW step 5.5.9. The Unit 1 distraeters require knowledge of the flow path for a release which is directed through the cross connect line. l@~ ~====================~~==============~~========~r===~ ~126===~ ~~~~~~~~~~F=============~F======~~~'~24===~ 1~~~~~~____________-J1~~~_____________~1 ____________11 ____~1_18__~ IL.6.""un1Qer lJ .1 IWASLlQE009 I I I I I
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i, SRO Skyscraper' I rQue~tr(>n Topic1 SRO 5 I Given the following conditions:
- Unit 1 is operating at 100% power.
- 21, 24 and 26 SW pumps are in service.
- 21 and 22 SW header pressures are 108 psig.
- The following OHAs annunciate within 10 seconds of each other in this order:
- 8-13,21 SW HDR PRESS LO
- 8-14.22 SW HDR PRESS LO B-15. TURB AREA SW HDR PRESS LO.
- B-48, SW VL V RM FLOODED.
The standby SW pump starts automatically. and OHAs B-13, B-14, and B-15 clear. Which of the following describes the status of the SW s},:stem, and which erocedure will provide direct actions which will mitigate the event? I l~ A SW leak upstream of the 2ST901, TURB LO CLR ST RET V has occurred. S2.0P-AB.SW-0002, Loss of Service Water-Turbine Header. I I.A SW leak upstream of the 2ST901, TURB LO CLR ST RET V has occurred. S2.0P-SO.SW-C001, Loss of Service Water Header Pressure. I I£J ~A SW leak in the CFCU piping in the 78' Mechanical Penetration Area has occurred. S2.0P-AB.SW-0002, Loss of Service Water-Turbine Header. I
@j !A SW leak in the CFCU piping in the 78' Mechanical Penetration Area has occurred.
Pressure. S2.0P-SO.SW-0001. Loss of Service Water Header I
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[KAStatem~dB Ability to determine and interpret the following as they apply to Loss of Nuclear Service Water: The cause of possible SWS loss LE~&1~~.~ti.o.*.* *~.*. . o.J.*.. 155.43(5) The leak location could be in the TGA with t~e conditions in the stem e~cept for the SW valve room flooding. Knowledge
'Answers:':" i**,. of where the SW valve room and what piping is there IS needed to answer question. The 2ST901 would respond on a TGA leak, L .._ *-~-~-~ and depending on leak size eQuid cause a restoration of header pressures. If the leak is determied to be in the TGA, the AB.SW-2 procedure would be used. Although AB.SW-1 is entered for all the conditions in the stem, it does NOT provide direct actions (as
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- 1 Given the following conditions:
1- Unit 2 is in MODE 6 entering a Refueling Outage,
- No fuel assemblies have been removed from the Rx.
- Rx cavity level is 26' above the PRV flange.
- 21 RHR loop is in service in Shutdown Cooling.
- 22 RHR loop is O/S.
Which one of the following would [lI'event initiation of Rx defuelina on Unit 2? I [~ Loss of Control Air to containment. I I
~ Loss of Plant Page capability in containment.
I IRacking down the 22 RHR pump 4KV breaker. I [J IOnly one of the two SRNI's can provide audible indication in the Control Room, I IIAnswerl a I I ~,Jt:am Ie\f~] IS I [8:i~~it~vfl..e~eIII Memory 1 [§~Cility:] ISalem 1 &2 I ~~oate:11 9/26120111 l~:jl 0OOO65G136 1 @:1.36 ]l!3ov~lue:'iI3.oltSRo£i@~t~:.u~~I:frLJrRO~1 111$ROG.r6~1 1[ lIit ~ ISySt~p/EX9IutiqriTitl:~ I..;;L.;.oS;;,.;S..;;o;;,.;f...;.l...;.ns;;,.;t...:ru..;;m...:e..;;n..;;t..;;A..;;ir_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _- - l '-.~,_,..___' I~~@tem~htrlr- ____________________________________________________________________________~ KnowledQe of procedures and limitations involved in core alterations. I
~1<!~a!i~it9fl 55.43(7. 6) The requirement for SRNl's is BOTH operable and provideing VISUAL indication in the Control room, with ONE
- A11sWe(s:i0::.,<.c: providing AUDIBLE indication in the control room. The manipulator crane is air powered for gripping. so the loss of air to
- . -..- -..~ containment would preclude being able to perform core ails. Onl! ONE RHR loop is required to be in operation in MODE 6. 2 RHR loops are required to be OPERABLE when <23' level above the flange. ~~~rs~~reri;;~1:itY;'C~~]"'}1;2] P;:;:l:~"~~~r;s~:wn~~!0,.J ~~ferenc:~S~ctiM' ~llil ~,N.'~ ~~i F1R=ef=ue=lin::::9:::0::::p=er=at=ion=s=======:::;lls2,OP-10.ZZ-0107 /I [113 1117 FIR=e=ac=p=e=n=e=Ar=ea==&=c=on=t=co=n=tr=ol=A=ir======~IIF2=05=3=47=-=1,=3=============J~~I============~11 IIF42=,3=6~
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Given the following conditions:
- 100%
Unit 1 has experienced a large Main Steam line Break (MSLB) inside containment from power. 1- Safety Injection was initiated. I-
- MSLI has failed to shut ANY MS167 and they remain open.
ii AFP is crr.
- 12 and 13 AFP's tripped after starting. - RCS pressure is 700 psig. - ALL RCPs have been tripped. - Containment pressure is 18 psig and rising. - ALL Wide Range SG levels are 35% and dropping.
1- ALL SG pressures are 120 psig and dropping.
- RCS Tc's have dropped from 540°F to 230°F in 20 minutes.
Evaluatethe data and select the Rrocedure to be entered and action to be taken upon transition out of 1-EOP-TRIP-i. Reactor Trip Response.
~ 11-EOP-FRTS-i. Response to Imminent Pressurized Thermal Shock Conditions; shut all MSi0's md steam dUl ' valves.
I [gJ l.i-EOP-FRHS-1. Response to Loss of Seconday Heat Sink; initiate feed and bleed immediately. I 11-EOP-FRHS-1, initiate feed and bleed ONLY when 3/4 SG WR levels have dropped <32%. I
@J li-EOP-FRTS-1, reset Safeguards Actuation and restore normal charging and letdown.
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!kAs~i~meni8r- __________________________________________________________________________________~
Knowtedqe of EOP mitigation strateqies. I iE.x.r5lan~ti8nofj 55.43(5) B is correct because the conditions given in stem would transition to FRHS*1 due to a a RED path of no AFW flow and IAns'We~s:1J <9% NR level. The Bleed and Feed initiation criteria are when SIG WR levels are < 36%(adverse), NOT 32% as in distracter c.
'--.._-_... Distracters A&D are incorrect because it is a lower priority RED path, though it's action is correct for the procedure. ~f::~i;"IffR~~)-iti;c-~;~~ ~l~Fi3cilijYfi~f~~~llce"~ufll'ier.':~ ~f~~S;cii0L ~~ [Re,yTSjtiilj 1 Response to Loss of Seconday Heat Sink 111-EOP-FRHS-1 II r \=1 ==:;1124 IResponse to Imminent Pressurized Thermal Sh 111=1:::-E:::O:::P=-=F:::RT:::S:::-=1======~III=======;11 1125 1Critical Safety Function Status Trees j!_1-_E_O_P_-C_F_ST_-_1_ _ _ _ _--'II_____.___ i1 11 25 IL.Q.N~mber '. '"IJj IFRHSOOE005 I IFRHSOOE013 !
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l~Ue~tlCm}Opicj ISRO 8 I Given the following conditions:
- A Rx trip and SI were initiated based on a LOCA 1 hour ago. - Operators have transitioned out of EOP-TRIP-1, Reactor Trip or Safety Injection. - RCS pressure is 350 psig and lowering very slowly. - All RCPs are stopped. - RCS highest CET is 290°F and lowering very slowly.
Rx power is 300 cps and stable.
- RVLlS Full Range is 95% and stable. - 21-23 SG NR levels are 33% and stable, and 24 SG NR level is 0%. - Containment radiation levels are 500 mr/hr and stable. - Containment pressure is 5 psig and lowering very slowly. - Containment Sump level is 46% and stable.
Which of the following identifies the highest priority CFST for these conditions, and actions taken in that associated EOP? Reference provided. [] IYellow Path Core Cooling. Isolate ECCS accumulators, depressurize SGs to atmospheric I IPurple Path Containment Environment. Isolate Containment Pressure Relief flowpath (2VC5, :2VC6) and return to procedure in effect. I
~ IPuq;!le Path Containment Environment. Isolate fluid sources from outside containment which have corroborating indications of lower than I Iexpected levels.
Yellow Path Core Cooling. Check ECCS flow for current RCS conditions as expected or start pumps and align flow paths, and ensure PZR PORVs and RPV Head Vents are shut. ~~ ~ !S I IApplication ISalem._1...;&;...2_ _...l w..~C.0C..:c....,.-J _ _ _ _ _ _- ' [~IOOWE07A201 I LEA2.1 J~1rv~@~~~~~~[~£lJ~~~iDJl~Ro~r()lIrl:IDJ ISaturated Core Cooling as the appl to Saturated Core Cooling: conditions and selection of ap ro riate procedures durin abnormal and emer enc 55.43(5) Core Cooling is the highest priority based on: All RCPs off and no subcooling and CETs <700' with RVLlS level >39%. The actions in FRCC-3 check first if you need to be in a different procedure (SGTR-4 or AB.RHR-1). then has you reset safeguards, ,~~..~~C.'C,.~~~~ check ECCS flow vs RCS pressure and align valves if less than expected. Then it checks proper PZR PORV and RPV head vents shut. The CFST Tables for subcooling HAVE to be provided since they HAVE to be used to determine subcooling. Purple Path for
*
- 0 .
- correct. FRCE-1 actions-are partially cOrrect. in' that yo~ would isolate VC5 and 6 and return to procedure in effect, but the conditions for entering FRCE-1 were never met so you wouldn't be in procedure < 1.5 si cont pressure.
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[Questi~~iopiC J ISRO 9 I Given the following conditions:
- Unit 2 was operating at 15% power prior to synchronizing the Main Generator. - depressurizing A Main 8teamline rupture occurred that resulted in multiple Steam Generators in containment before 2 steam generators could be isolated from the 2 faulted 8Gs. - The 2 faulted 8Gs Tcs are reading 270"F and lowering. - The intact 2 SG Tes are 330°F and stable. - RCS pressure is 500 psig and slowly lowering, - Containment pressure is 16 psig and slowly lowering. - All SG NR levels are <9%. - Total AFW flow is 24E4 Ibm/hr. - Source Range Nis are NOT energized. - Intermediate Range SUR is 0.0 DPM.
With CFST's in effect, which of the followin~ identifies the procedure entry required, and actions whier will be performed in that procedure?
~ l2-EOP-FRTS-1, Response to Imminent Pressurized Thermal Shock Conditions. Maintain AFW flow >22E4 Ibm/hr until at least ONE intact SG NR level is >15%, stop all ECCS pumps except 21 or 22 charging pump. I
[EJ 1.2-EOP-FRSM-2, Response to Loss of Core Shutdown. Energize Source Range channels and verify SR SUR is 0 or negative. I 12-EOP-FRTS-1. Isolate any faulted SGs, depressurize RCS with ONE PORV to within 1OOo/hr Cooldown Curve. I J2-EOP-FRSM-2. Establish AFW flow >44E4Ibmihr, borate RCS untillR SUR is negative. I
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~StateJ1i~.~ Ability to determine and interpret the following as they apply to Pressurized Thermal Shock:
Facili~conditions and selection of apJ:1[opriate2focedures during abnormal and eml9rgency operations.
~lah#t,i~l'Iv~fl 55.43(5) A is correct be~use the stem conditions res~lt in a PURPLE path on FRT~~. Actions for maintaining AF.W flow (Step 3.5)
- Answers;)1 and ECCS pump reduction (Step 12) are correct. C IS Incorrect because depressuflzlng the RCS to restore oondltlons W1thn the
~._v_~~_ 1000F/hr curve is performed in FRTS-2 in response to a Yellow Priority Condition. FRTS-1 is entered from either RED or PURPLE conditions, and with SG NR levels <9% (which means you're less than 15% adverse) you are directed to maintain AFW flow> 22E4 Ih",,/h, ~{;th 'h SR NI" nl'1l . 10 CliO Ie ""., . ,.-l ho "'0 " Ih.",~ _n ') "';n"" " V::: I ('\\, ",+h
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- 2. The FRTS is a higher priority, and is a PURPLE path. B is incorrect because it is the wrong procedure with' the correct actions of that procedure. 0 is incorrect because it is the wrong procedure, and the actions are performed in FRSM-1.
1 2"~~~~iffi!E!~~ES] ~liiW:~~~~ill ~~~~u[~~i~~ ~YiSfdrl1 I Critical Safety Function Status Trees 112-EOP-CFST-1 II 11 11 25 I I Response to Imminent Pressurized Thermal Sh [\2-EOP-FRTS-1 " il 1125 I I Response to Loss of Core Shutdown 112-EOP-FRSM"2 JI II 11 20 1 [gi~r~m I FRTSOOE002 I I I I -.J
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[9uestion~ioPi~ SRO 10 I I Given the following:
- Unit 2 is in Mode 3 following a shutdown after a 200 day run. - All RCP's are in operation. - Main steam dumps are in MS PRESSURE CONTROL-AUTO @ 1005 psig.
24MS10 setpoints are 1015 psig.
- A transformer fault results in the total loss of off-site power. - 15 minutes after the transformer fault, with NO operator action, the following indications are present: - All RCS WR Thot's are 559°F and rising slowly. - All RCS WR Tcold's are 549°F and stable. - All SG pressures are 1015 psig and stable. - All SG NR levels are 39% and stable. - PZR level is 23% and rising slowly.
Which of the followinlij identifies the action that must be eerformed? [!] I.Lower 21-24MS10 setpoints to establish CET's stable or lowering lAW S2.0P-IO.ZZ-0008, Mai:ltaining Hot Standby. I
~ I.Lower 21-24MS10 setpoints to establish CET's stable or lowering lAW S2.0P*A8.RC-0004, Natural Circulation.
I "t.] ILower Main Steam Dume eressure seteoint to stabilize or reduce RCS Thots lAW S2,OP-AB.RC-0004. I Lower Main Steam Dump pressure setpoint to stabilize or reduce RCS Thots lAW S2.0P-IO.ZZ-OOOB.
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IKA: i loOWE09G411 I [~r=J[i@~g~~xT!f~~E!@~!~I~~~~O~~O 1~y"st~Ql!~\i~luti9nTTtlel INatural Circulation Operations
~St~em~~r- ________________________________________________________________________-.
Knowled e of abnormal condition procedures. IAE.X.R!~h~tiO~~ 55.43(5) The .Ios.s of off-site power will cause all ~CPs and Circulators. to st?p. Loss of the circulators will ~ause a loss of Steam nsWers: l;5,,1 Dumps Permissive and all steam dump valves Will shut. Steam dumping Will transfer to MS1 Os at 1015 pSlg from Steam Dumps at
-~~ 1005 psig. C and D are incorrect because steam dumps will have no effect. A is incorrect with right action but wrong procedure.
Ab.RC-4 is entered on a loss of forced circulation in MODES 3 and 4. The stem shows nat cir is NOT present, so the correct
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~~~i~j~.!CJ ISRO 11 Given the following conditions:
- Salem Unit 1 is operating at 30% power, performing a shutdown at 30% per hour.
The shutdown is being driven by Tech Spec 3.0.3 being entered when BOTH ECCS subsystems were declared inoperable.
- The unit is required to be in MODE 3 within the next 2 hours.
Which of the following describes the effect on the Rx if a loss of off-site power occurs, and how will operators perform the required cooldown? WILL NOT trip. Operators will perform a rapid cooldown lAW 1-EOP-TRIP-6, Natural WILL trip. Operators will perform a rapid cooldown lAW 1-EOP-TRIP-6, Natural Circulation Rapid Cool down with RVLlS. WILL NOT trip. Operators will perform a normal cooldown lAW 1-EOP-TRIP-4, Natural Circulation Cool down. Operators will perform a normal cooldown lAW 1-EOP-TRIP-4, Natural Circulation Cooldown. lCJ ~~£i~~ IApplication ~1& 2 IEIamD~i~ I 9126/2011 1
~ ~]I 002000A203 I~~~=]~@~Y1il~ rs;;cwll;,ll~ ~~~u~~~D1 !,,::,,~~~;',';;;:, 'Iltle':j IReactor Coolant System I i()02*~ i MStatement:] Ability to (a) predict the impacts of the following on the Reactor Coolant System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:
Loss of forced circulation
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. . . t 55.43(5) The Rx will trip on the loss of forced flow, based on any of the rCP trips associated with losing at least 2 RCPs when >P IAnii;wel's,:{.~~'fl10 and < P-8. In order to prevent violating Tech Specs, the unit must be cooled down to <350 OF (MODE 4) within 6 hours of being
. . in MODE 3. At the time MODE 3 is entered (on the Rx trip) Tavg is 547°F. 547-350=197° degrees of cooldown required in the next 6 hours, 197/6= 32.8°F/hr cooldown required. TRIP-4 maintains a maximum cooldown rate of 25 °F/hr, so the transition to TRIP-6
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~~;F{~fe~~ljceTltlrJr,*;..i: . .*. ~~'I ~f~~i!ityFi~fer~n~e N'~m9.~rmill ((~fer~ncif;SE;§fu;~~l ~geNo.l ~~11J INatural Circulation Rapid Cooldown with RVLlS "1-EOP-TRIP-6 1\ II 1 [22 1 INatural Circulation Cooldown 1/1-EOP-TR'P-4 il II 1122 I IRPS Primary Coolant System Trip Signals li221054 II II 11 10 1 IL.().. f:!umR~rfi* i&ill I RCSOOOE007 I
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IQuestion Topic! ISRO 12 I Given the following conditions:
- Salem Unit 1 has performed a Rx trip and SI based on rapidly lowering PZR pressure and level. - The crew has transitloned to 1-EOP-LOCA-1. Loss of Reactor Coolant.
Which of the following describes why the first several steps of LOCA-1 check for reasons OTHER than a LOCA for being in LOCA-1? LOCA-1 actions assume ... It loss of RCS inventory. If the ECCS injection flow is due to a Loss of Secondary Coolant, the event could be term inated by raising AFW flow. I
~ I a loss of RCS inventory. If the ECCS injection flow is due to a Loss of Secondary Coolant, the event could be terminated by isolating the faulted SG. I la SGTR is NOT causing the LOCA condition. If the ECCS injection flow is due to a SGTR, an uneccesary transition to LOCA-3, Transfer to Cold Leg Recirculation would be made. I I
~ a SGTR is NOT causing the LOCA condition. If the ECCS injection flow is due to a SGTR, the tube rupture must be addressed before retuming to LOCA-1 to terminate ECCS flow.
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[.!9';/1006000G422 '13*~4.22 *** n I~Y~lue:JI 3.61ISROVafufj 4.41 tseCtiqQ:'II~ [g.. §r<>~1 1'ls~qGrk'~1 11 .tel ~ rsystemlEv()lutionTi~IJU IEmergency Core Cooling System FKASt~eme~~~__~__~____~______~__~~~__~~~____~_________~~_______________________- . Knowledge of the bases for prioritizing safety functions during abnormal/emergency operations. I
~~pl.Matiqrl()fJ 55.43(5) The Major Action Categories for LOCA-1 are: 1. Check for Subsequent Failure, 2. Monitor Plant Equipment for Optimal
- Ar'!swe'rS:i>,,;J**! Mode of Operation, and 3. Determine optimal Method of Long-Term Plant Recovery. The first item checks that a faulted or L._._.~~.~~ ruptured SG is not the reason for ECCS injection, and either fixes it on the spot (faulted) or transitions to another more approriate procedure With a faulted SG being the cause of the ECCS flow, LOCA-1 has a "do loop" which will wait unril the SG has blown
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- in (V'. _1 . th t,,,nciti,,,,, t, (')('[L?' ""0 11 :.,. I" C::~TO i" performed to do actions which will terminate"the primary to secondary leakage, which is not performed in LOCA-1, Additionally, 1 there is no transition from SGTR-1 to LOCA-1.
L=os=s=of=R:::;:e=aC=to=rC=O::::O=lan=t======:::;1I:=1=-E=O=P-:::;:LO:::;:C:::;:A=-1=====:;iI.:========'.' ==:::}11l==2=5 =~ ISteram Generator Tube Rupture II II FI :=' j /1-EOP-SGTR-1 1124 i=1_L:::;:O_S~S~O=f_s~e=c_o~n==d_a=:ry~C~o=o_l=a_n~t~~~~~~~==-==-~~~I~I~1-~E_O==_P:::;:-~L_O::::S~C~-_1=_=-=_=_=_=_=_=_=_=--,:::;I*:=!.::_=~~~~~~~~~~~'I 111==2=0=~ ~~~D[CJ ILOCA01E005 ISGTR01E003 ILOSC01E002 IMaterial Requifed fqr Examinatiori . :j I l~estjo~so~rcri I_N_e_w_ _ _ _ _ _.....1[Qi!~stjOrlMOdificatiorl Method:] IfUsed DLlrjngTrainingPr?gra~ j 0 ~ues~~~~~ETc6mmentSJ I
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Given the following conditions: I
- Unit 2 is operating at 100% power.
- 2PR1 fails open and remains open, Which of the following identifies how this affects the PZR Master Pressure Controller (MPC) response, and what consequences, if any, are associated with the actions eerformed by the crew lAW S2,OP-AB.PZR-0001, Pressurizer Pressure Malfunction?
IMPC output will LOWER. The unit may continue to operate indefinitely after the initial litigative actions Ipleted. I
/.MPC output will RISE. The unit may continue to operate indefinitely after the initial mitigative actions are completed.
I [J IMPC output will LOWER. A unit shutdown will be required if 2PR1 cannot be restored to operable status, I IIMPC output will RISE, A unit shutdown will be required if 2PR1 cannot be restored to operable status. I IAf1$werll c I re;-arrfJ::~;"erll S I ICognitiye~~veIJ IApplication 1 ~;mtY:':11 Salem 1 & 2 1 [§x~m~etl!:11 9/26/2011/j c- -,./- .-.-,,,' ~---,",,- ~ ""'-r k w<<r,?$ 1t2,03 u [KAdlo10000A203 - .. --~~--
!§Jsterrt?5~luiiol\ Jlll~J IPressurizer Pressure Control System I i010 ~:Sfatehlellill Ability to (a) predict the impacts of the following on the Pressurizer Pressure Control System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:
PORV failures I
~p,an@q~*'*Ofl 55.43(2) This question is SRO level because of the Tech Spec knowledge required, and what actions TS directs for different PORV Ati!;we!s: . *.~._. malfunctions. Additionally, wihile the question doesn't specifically ask what procedure to use (too easy for AB.PZR), it does require knowledge of the actions IN that procedure. The MPC raises output wihen actual pressure rises, and lowers as actual pressure lowers. As pressure lowers due to the open PORV, the output will lower to turn on heaters and close spray valves. When the PORV
. 01 .~~ ,I, '~. In' 'h 01"'10\' 1\0 070 T, Q, '1 A e:; ~~tinn h ;f th~ 01"10\ ;~,' ,,-I Hh;n 7') h, -h"tdnlA/f'1 is required, A PORV isolation that DOESN'T require shutdown if not fixed is a leaking PORV, which is isolated by its Block Valve I with power maintained to the Block valve,
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!ABPZR1E002 I !PZRP&LE010 I I I
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Given the following conditions:
- Unit 2 is operating at 100% power.
- 23 charging pump is in service,
- A control circuit malfunction has resulted in the 2CV3, 45 gpm Letdown Orifice opening, and cannot be shut from the control room.
Which of the fall this failure will have on rm,,,,,l';nn of the , and how will rm,>r:.'tnr<:. .~~~r.r,,"1*J 23 charging pump speed will rise, Isolate letdown and place Excess Letdown in service lAW S2,OP-SO,CVC-0003, Excess Letdown Flow, 2CV55 will modulate in the open direction, Isolate letdown and place Excess Letdown in service lAW S2,OP-SO.CVC-0003, Excess Letdown Flow, 23 charging pump speed will rise, Transfer to a centrifugal charging pump lAW PZR LEVEL LO alarm response in S2,OP-AR.ZZ*0012, Control Console CC2. 2CV55 will modulate in the open direction, Transfer to a centrifugal charging pump lAW PZR LEVEL LO alarm response in S2,OP-AR.ZZ* 0012, Control Console CC2,
~!1011009A201 1~~MY~@~~~S!i~I~~~§§(JiEiU~~[Jj IPressurizer Level Control System IKAS!lilt'ernermJ i Ability to (a) predict the impacts of the following on the Pressurizer Level Control System and (b) based on those predictions, use Iprocedures to correct, control, or mitigate the consequences of those abnormal operation:
*Excessive letdown k~sl{llersz.:/,:*.'::
.*t; ,.~.I.?.'.a.,.*'.Q~t.:i.* o.'.,.n.*.*.* ***. o.'..',.*.f.* 55.43(5) Letdown orifii are Containment Isolation valves, and will require letdown isolation to isolate the CIV. For long term operation Excess Letdown must be placed in service. The 23 charging pump (PDP) is normally in service at power. It has a maximum flow of -96 gpm, Normally, letdown flow is 75 gpm with one 75 gpm orifice in service. With the 45 gpm orifice in service, letdown flow will be -110 gpm, The Master Flow Controller controls 23 charging pump speed when in service, and the CV55 when
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Given the following conditions: Unit 1 is operating at 100% power.
- 1PR5, PZR Safety Valve, falls open.
Which of the following identifies what will occur if NO operator action is taken until AFTER the Rx trips, and which procedure will be used upon the transition out of 1-EOP-LOCA-1 Loss of Reactor Coolant? The Main Turbine will runback on OT/DT, then the Rx will trip on OT/DT. 1-EOP-LOCA-2, Post LOCA Cooldown and Depressurization. The Main Turbine will run back on OT/DT, then the Rx will trip on OT/DT. 1-EOP-TRIP-3, The Rx will trip on Low PZR Pressure. 1-EOP-LOCA-2, Post LOCA Cooldown and Depressurization. The Rx will trip on Low PZR Pressure. 1-EOP-TRIP-3, Safety Injection Termination.
~'I 013000A203 I~~~~Y~B~Yllii!Q2i~~I~ ((~~D[SR(lG~U l~~tE!riil§vol~tiQnr~ IEngineered Safety Features Actuation System 11013~- "l LISA-Statement: I Ability to (a) predict the impacts of the following on the Engineered Safety Features Actuation System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:
Rapid depressurization rEx~atiOn~ 55.43(5) The PZR Safety failing open will cause a SBLOCA. Pressure will rapidly lower, and an OT/DT runback will initiate before l.AJ1siI.J~'fs:';£*:f!1 the OT/DT Rx trip. The Safety Injection will actuate, and pressure will stabilize - 985 psig. With no subcooling, the transition to
~.~"-'-'~--"""-J TRIP-3 cannot be made, and LOCA-2 will initiate a cooldown. The distracters either contain the wrong procedure, the wrong Rx trip actuation Signal, or both.
D':;;:' . ;:;.RefE!reijbe:ti~i~~;-:;~r~~~ F0~i=lI8fi~e~NU;r\t;~r"""7~ ~~S~ro;.;-r~ ~ l[~ IPost LOCA Cooldown and Depressurization !l1-EOP-LOCA-2 ...11 II 11 23 I I II , , :I II I I II II II fI I I ESFOOOE021 ~aierjal ~~qlJired f.or EXamination*.*.'.*. ! I ~~stion sourceIlI New lC......... ~ ...., ' I I I I
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Given the following conditions:
- Unit 2 is in MODE 6.
- Core off-load is in progress and is 2/3 complete.
- A fuel assembly is in the transfer cart at the Spent Fuel Pool.
- The mast tube of the manipulator crane is empty.
- Refueling cavity level begins to drop rapidly, approximately 6 inches per minute.
- Radiation Protection reports flooding in the lower elevations of containment.
Which of the following describes how the Fuel Transfer Tube Gate Valve should be operated lAW S2,OP-AB,FUEL-0002, Loss of Refueling Cavity or Spent Fuel Pool Level, and the consequence of that operation?
"START the transfer cart moving to containment, then shut the gate valve when the cart is clear. The fuel transfer cart will stop when the gate vaiveopelllimit is lost. I Ib~J START the transfer cart moving to containment, then shut the gate valve when the cart is clear. The fuel transfer cart will continue until it I
- reaches its normal travel stop on the containment side.
I
~ IMMEDIATELY shut the gate valve regardless of transfer cart position. If a fuel assembly is in the Spent Fuel Pool upender, transfer it to the Fuel Handling crane and place it into its designated SFP location. I
[J IIMMEDIATELY shut the gate valve regardless of transfer cart position. If a fuel assembly is in the Spent Fuel Pool upender. lower is to the horizontal position to reduce any radiation exposure due to dropping SFP level. I
!Ariswed a I IlI§!rn.Levefll s 1[Cognithie;t;evE!U IMemory 1 ~~c1J)tY:l [Salem 1 & 2 I~;nb~te:*ll 9/26/20111 ~I034000G435 I [4.35_'_J fROyallieTI[ 3.81 [$Rbya,(jeL~~ :secti?J.ruI~ (([groIJP:.11 21 [s~gOqr~ufuJl 21 Ita ~ ~~~![!it~~ 1Fuel Handling Equipment System ~Sfatemerit:l~__________________________________________~________.__~____~____________________~
Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects. 55.43(7) With a no fuel assembly in the mast tube and a fuel assembly in the transfer cart, the operator is directed to close the fuel pool gate valve at step 3.13 AFTER starting the cart to the containment side. The discussion section of S2,OP-AB.FUEL-0002 (2.4.C) states that the gate valve is not to be closed until the fuel handling cart is clear of the gate valve path. It also states that while the transfer cart may stop prior to reaching containment, it is acceptable because the water around the fuel with shield and
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ILoss of Refueling Cavity or Spent Fuel Pool Lev II S2.0P-AB.FUEL-0002 -11 II 11 10 1 iIF==========-1~~I======~I~1~IFI~ II 1/---------'11----11----'1 1--, [i+/-9~~~~2~)) IABFUE2E002 IQuestion~~l;'r~J [ Facility Exam Bank ll9ue~~i()n.tlllodificati()nMett1Od~IJ Direct From Source I [Questionsoor.cecol11mentsJ IVision Q71988
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. ROSk~SCra~eYrl ;'SROSkV~c~aDJ~tTRO a~~f!i9nToP~J I Given the following conditions:
- Salem Unit 2 is responding to a loss of all off-site power.
- 28 EDG is supplying 28 4KV vital bus.
2A and 2C EDGs have tripped.
- 2A and 2C SECs have been deenergized.
- A C02 actuation occurs in 2B EDG room.
Which of the following identifies the effect the C02 actuation will have on 2B EDG operation, and what actlon(s) will be performed? will automatically trip. Deenergize 2B SEC lAW 2-EOP-LOPA*i, Loss of All AC Power. will NOT automatically trip. Deenergize 2B SEC lAW S2.0P-SO.DGV-OOOi, Diesel {;p",p,r,>lI>r will automatically trip. Place the 2B Diesel Generator Supply Fans Emerg Bypass of C02 Shutdown Switch at 2RP5 in Emergency lAW 2 EOP-LOPA-i. will NOT automatically trip. Place the 2B Diesel Generator Supply Fans Emerg Bypass of C02 Shutdown Switch at 2RP5 to Emergency lAW S2.0P-SODGV-0001. mI 064000A222 I~E2~_J ~~~nfTI'~.~E.:J ;S~cfioh:<I~ fNiE~jO ~~~ IEmergency Diesel Generators There is no automatic EDG trip on C02 actuation, either in the EDG control room or the EDG area. The sEC does not control the EDG Area/control room supply fan. The SEC deenergization of the unaffected SECsfEDGs in stem is to make the EDG trip plausible if the candidate thinks that the SEC would prevent an EDG trip, since therE~ would be a standing Mode 11 SEC signal present with 2 vital buses deenergized. Bypassing the C02 shutdown of the supply fans allows them to restart since their control Ithere is an exhaust fan for the EDG FO Storage area.
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Given the following conditions: Unit 2 is operating at 100% power.
- 25 SW pump is INOPERABLE for scheduled maintenance.
- The appropriate TSAS Tracking entry for TSAS 3.7.4 has been made in OP-SA-1 08-115-1 001, Operability Assessment and Equipment Control Program.
23 SW pump strainer motor fails. Which for that action? Exit the TSAS Tracking statement for 25 SW pump, and enter TSAS 3.7.4 as an ACTIVE entry due to 1 SW loop being INOPERABLE. Exit the TSAS Tracking statement for 25 SW pump, and enter TSAS 3.7.4 as an ACTIVE entry due to 2 SW pumps being INOPERABLE. eking entry for 23 SW pump. Each 4KV vital bus requires only one OPERABLE SW pump to ensure full for 23 SW pump. IF ANY other SW pump were to be declared INOPERABLE, THEN entry OPERABLE SW loop would remain. IService Water System LKASt~emehl:I~~ __~__~__~~__~____~__~~__~~~__~____.____________________________-. Ability to determine operability and/or availability of safety related eouipment.
~p!araiitin()f] 55.43(2) Each SW pump must have its strainer operable for the SW pump to be operable. (SO.SW-005, page 97). With 2 SW ~!ii~rs:i~"<, pumps inoperable on different vital buses AND in different SW bays as described in stem, then BOTH SW !oops remain operable.
The requirement for 2 operable SW LOOPS are: One operable pump on A bus, One operable pump on B bus, One operable pump on C bus, and 2 operable pumps per bay. The requirement for ONE operable SW loop is 2 operable pumps powered from different
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operable SW loops. C is incorrect because the bases is wrong. If there were only one pump per vital bus, then there would only be one operable SW loop. A and B are wrong because entry into an active Tech Spec is not required for 2 SW pumps inop on different vital buses in different SW bays. t~"~~~e~I'I§,~1)~8:~~~~~~ful i~clf~fYJ~~fr~n~~NU~iJ ~~{se~~ ~~ ~~lc:It\]
/ Salem Tech Specs " 113.7.4 II 1/ I I Service Water System Operation II S2.0P-SO.SW-0005 II II 11 40 I I 'I 11 r I II I ILO;Num'ber,\> ~:..c_ ._,~:::~~_: __ .*...1____
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-~~~ I section to each chOice .
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Given the following conditions: A heatup is in progress on Unit 2 lAW S2.0P-IO.ZZ-0002, Cold Shutdown to Hot Standby.
- All RCPs in operation.
Seal injection flow is 6.0 gpm to each RCP During the heatup, VCT gas pressure is lowered from 35 psig to 15 psig. be addressed? Flow to the No.2 RCP seal will lower. Fill the RCP standpipe when the low level alarm annunciates lAW S2.0P-AR.ZZ-0011, Control Console 2CC1. A high RCP standpipe level alarm will actuate. Drain the RCP standpipe until the high level clears lAW S2.0P-AR.ZZ-0011. Control Console 2CC1. A high RCP standpipe level alarm will actuate. If required, raise seal injection flow to maintain at least 6 gpm to each RCP lAW S2.0P SO.CVC-OOOi. Flow to the No.2 RCP seal will lower. If required, raise seal injection flow to maintain at least 6 gpm to each RCP lAW S2.0P-SO.CVC-0001.
~11194001G123 j§~]l!N~@~~~~I~I§£~Ug2Ef~D ~Y;ten,'EVOll!tior£Jltj~ I I iGE!:lERIJ IKASt~e~e~~ ______________________________________________________________________________ ~
Ability to perform specific system and integrated plant procedures during all modes. of plant operation. [E:X,.,P,"',!,~;;, .Cltio"""**n"""".f'* ',:f,.,,'.,',',I* 55.43(5) Lowering the VCT pressure reduces the backpressure forcing flow to the #2 seal. With less backpressure. flow to the
~swer$:i;~B1d number 2 seal will lower (more of #1 sealleakoff is going to VCT) With less flow going to the #2 seal. number 2 sealleakoff will
'-'~'-'--. lower, and standpipe level cannot rise. The RCP standpipe will not lower to the alarm setpoint as long as there is any flow from the
#2 sealleakoff.
GX~C--=id':~efer~n¢~~i0T~ ~i;sW~ii~Qt¥NO~~f~ ~e(~r~ct;lEiec!iE~ :;~ ~~ ~~ ICharging, Letdown, and Seal Injection II S2.0P-SO.CVC-0001 II II II I il IIF====;11 II~=:::::; I II fI II 11---1 ~\.HJlb~r:>/J I RCPUMPE008 I I I [ I l~ateri<ll Required for Exai'ilillatioll II OIl":':7JISOllrc~;j IFacility Exam Bank ll9u~s~i()nModificall()nMeth6d:J Editorially Modified I illsedDuring "rai~ingl;)rogram *1 0 fQuestion sourceCorrln'lentsJ IViSion Q85461 L_,_ _ ,_~, .. ~._._.. _'~'_"'~ ... _ .
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Given the following conditions:
- Salem Unit 1 is operating at 100% power with equil ibrium core xenon-135.
- Power is lowered to 75% over a 15 minute period using the control rods.
- The NCO then adjusts control rod height as necessary to maintain average reactor coolant tem perature constant.
What will be the rod position and directional trend 30 hours after the power change? Reference rovided. Below the initial 75% position and inserting slowly. Above the initial 75% position and inserting slowly. Below the initial 75% position and withdrawing slowly. Above the initial 75% position and withdrawing slowly.
- KA=I!194001 G143 =r ~~~=J 1~9,Yal0e:lG~~~~L2~ ~~I~ ~2§~[3 ~JUR!?~D
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Ability to use procedures to determine the effects on reactivity of plant changes, such as reactor coolant system temperature, I secondary plant, fuel depletion, etc. I.E.. .~P.I.*. .a.*~a.tio. . . .n. .*. . Of.*. . 1 55.43(6) As Xenon builds in following the power reduction, control rods will have to be withdrawn. After the Xenon peak, Xenon will iAnswers;!:. "J be burning out, and control rods will be inserted to offset the less negative reactivity due to Xenon. Xenon will continue to burnout to I
-'~"~-' establish a new equilibrium level at 75% power, and the control rods will continue to be inserted past where they were inserted at 75%.
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§~e,stion To£~ 1 SRO 21 Given the following conditions:
Unit 1 is operating at 100% power. Operators have just satisfactorily completed SC.OP-PT.DG-0001 DIESEL GENERATOR MANUAL BARRING on 1B EDG.
- The only Tech Spec entered was 3.S.1.1.b for the 1 B EDG being declared INOPERABLE when the LOCKOUT SW was crr in the LOCKOUT position.
Which of the following identifies when the 18 EDG will be declared OPERABLE? CRS. When the LOCKOUT SW is released to the IN SERVICE position after successful completion of SC.OP-PT.DG-0001. When the NCO is directed by procedure to log the EDGs availability for operability testing in the Control Room Narrative Log. When the EDG has successfully met the acceleration, voltage and frequency Acceptance Criteria in the ST procedure used for the retest, and is synchronized to the Vital Bus. 1Memory I WaCil.l~( I Salem 1&2 11t:~~~Dat~:11 9/26/20111
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Ability to track Technical Specification limiting conditions for operations. i IE~t>lai1ationofl 55.43(2) The EDG beccmes inoperable when the Lockout Switch is placed in the lockout position. The EDG remains inoperable [!.nS'i\ferS:>>i during the performance of the barring. It cannot be declared operable until it has proved it meets the requirements for an operable
..._ _... EDG, since it has been affected by the barring operation. The requirements are successfully reaching rated speed, voltage and frequency requirements, AND being synchonized to the vital bus.
I________---lll II II 11=1====; fL':d.Number.. .:C071
~...,~~_:::;~:_~;;:~~.;Ed I EDGOOOE010 1 I I I I §~esti~~ource~2~me~ I"I" ILOT SRO CERT Exam 11/2006 IComment c.
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- I A Hope Creek Station employee has received 1900 mrem routine TEDE for the current calendar year, ALL at Hope Creek Station.
The employee is expected to receive an additional dose of 450 mrem on his current job assignment at SALEM. His lifetime exposure is 5500 mrem. lAW RP~AA-203, Exposure Control and Authorization, and prior to performing the job, written approval for increasing his dose limit to 3000 mrem TEDE for the calendar l'ear must be received from the work ~roup supervisor and the ... I RP .V.d. 'd\J!;l1 ONLY. I I Site Vb """,,,,Ye,,, ONLY. I
@J IRP Manager and Station Manager.
I II Station Manager and Site Vice President. 1 [Ans~~rJ Ia I ~mL.~v~111 S IIF()gr..~,~,,~~~el;11 Memory I [FaCility: Iisalem 1 & 2 J ,~J(#llll)ate :J 1 9/26/20111 iKA: 111 94001 G304 J~.3.4= IRo'\f~iue:11 3.21 f$~()V~lue~ 3.71~~~tiOP: IPWG 1((O.Gfo~p:: I 11Is~p~roup:ll 11 :i.~ ~ Syste I11/Ev()h:ltIPI1 Ti!!:J l__________________________________--'I[GEN@ Knowled e of radiation exposure limits under normal or emer enc IEXpla'nati6Hof 55.4:3(4) A is correct because the approval requirements are: Up to 3,000 mrem- RP Manager; up to 4,000 mrem- RP Manager and IAni;~er~:::,<!::~.. Station Manager, >4.000 Site Vice President. Distracters are all some form of combo of each of the positions.
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IRADCONE002
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Given the following conditions:
- Unit 2 is operating at 100% power.
The operating crew entered S2.0P-AB.RC-0002, High Activity in Reactor Coolant System, when RMS channEll 2R31, Letdown Line Monitor, went into WARNING. Which of the tnllflWlnn is the ant areas? Confinmation of 2R31 response. Radiation levels may have changed access requirements. 2R31 reads in CPM and therefore has no correlation to dose level changes. Detenmination of radiation type (gamma vs. neutron) will narrow possible sources of the higher radiation levels.
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iKAskte~e~r_------~------------------------------~~----~~w----------------------------~ Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities. I 1.~,~~J.~~~t@I.9f,1 55.43(4) B is correct as described in S2.0P-AB.RC-002 basis, so that prompt identification and subsequent notification of plant
!"' ,,,,~ ... !,,,~*"I personnel IS ensured.
A is incorrect because chemistry sampling confirms 2R31 readings, not survey results. C is incorrect because rising counts does indicate dose level changes. D is incorrect because the radiation type won't be determined, as neutrons won't be found. II_H=i=9h=A=C=tiV=i~=i=n=R=ea=m=or=C=o=ol=an~t=sy=s=re=m==~III-s=2.=O=P=-A=B=.R=C-=OO=0=2======~ill-==========~111-====~r,I_8==~ 1- [1- f 11- 11 ~I-=-=-=_=_=_=~_~_=-=-=-=-=-=-=-=~I~I========:;II II I~[~
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[Questio-; To;;j~ ISRo24 I Given the foliolNing conditions:
- Unit 2 has received a FIRE alarm for Zone 69, Fuel Handling Building (FHB).
- An operator in the area reports a fire in a bin of Protective Clothing in the FHS truck bay.
Which of the followin~1 identifies the correct 8rocedure for this event, and the actions reguired? I
~ I.S2.0P-AB.FIRE-0001, Control Room Fire Response, place Control Room Ventilation in FIRE OUTS IDE CONTROL AREA on both Unit 2 and c.. Unit 1, and secure ALL Unit 2 ONLY FHB supply and exhaust fans.
I I S2.0P-AB.FIRE-0001, Control Room Fire Response, place Control Room Ventilation in FIRE OUTSIDE l,;UN I KUL AREA and secure BOTH units FH B supply fans ONL Y. Unit 2 ONLY, I IS2.0P-AB.FIRE-0002, Fire Damage Mitigation, place Control Room Ventilation in FIRE OUTSIDE CONTROL AREA on both Unit 2 and Unit 1, and secure ALL Unit 2 ONLY FHB supply and exhaust fans. J
@]ls2.oP-AB.FIRE-0002, Fire Damage Mitigation, place Control Room Ventilation in FIRE OUTSIDE CONTROL AREA on Unit 2 ONLY, and secure BOTH units FHB supply fans ONLY. I
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[KAStatenient:! r-:------::--:-------------------------------------, Knowledge of "fire in the plant" procedures.
~ . ~.pl.ar~.fi.o.*.*. *.n.*.**'.***.o..:*..f.*. . . : 55.43(5) S2.0P-AB.FIRE-02 is the incorrect procedure because it is used for an uncontrolled fire in the Aux Bldg or 78 electrical Answers:.<j pen, not for a fire in the Fuel Handling Building. Steps 3.12 and 3.13 of S2.0P-AEI.FIRE-01 places BOTH Units in fire outside, and
..._ - - - 3.94 and 3.95 direct stopping FHB supply and Exhaust Fans. A is correct becausEI it is the correct procedure and action. The incorrect distracters are all a combination of either incorrect action, procedure, or both.
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[Question Topic I I
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SRO 25
- 1 Given the following conditions:
- Salem Unit 2 is experiencing an event which has resulted in an ALERT EAL being exceeded and recognized by the operating crew.
- After 5 minutes. the actions taken by the control room crew result in this ALERT EAL no longer being met.
1- Conditions for an UNUSUAL EVENT are present.
- No emergency declaration has been made yet.
Which of the followin~ identifies the actions required b:t the Emergenc~ Coordinator? I Declare an ALERT. terminate the ALERT. declare an UNUSUAL EVENT. I I.Deciare an ALERT. then reduce the Emergency Classification to an UNUSUAL EVENT. I I Declare an UNUSUAL EVENT, and make a non-emergency Afler-The-Fact 1 hour report to cocument the ALERT EAL. I I Declare an UNUSUAL EVENT, and ensure the ALERT EAL which was present for the 5 minutes is communicated to the NRC via the NRC Data Sheet. . I iAOswe-r] 1b liExam Le"'[Jl/ S II~ognl~I~~JlE!ver ;<1 1Memory 1~ 1Salem 1& 2 IlExa'",~(lte:11 9/26/20111 lKA: 11194001 G438
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_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _---1IIGENER!
~~yollltiO!lTitle. I ~St~emen!Jr- _____________________________________________________________________________ ~
Ability to take actions called for in the facilit ExPI~ri~tionRf' 55.43(5.1) Per the Salem ECG, Section I, page 8 of 10, Section IV, Event Classification Guide (ECG) Use, Subsection 0.2,
.AAswers: ~ ... "Short duration events that occur will be assessed and emergency classifications made, if appropriate, within about 15 minutes of ~~--.. control room indications or the receipt of the information, indicating that an EAL, has or had been exceeded. This classification is to be made eve~ if no EAL's are currently being exceeded, (Le. actions have bee~ taken to stabilize the \lant such that no EAL's EP.ZZ-040S, EMERGENCY TERMINATION REDUCTION - RECOVERY, d"escribes how to reduce the Emergency Classification if I the emer enc is not bein terminated. Since the UE is still bein met, the classification will be reduced, not terminated.
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