ML022890184
| ML022890184 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 05/14/2002 |
| From: | Vonsuskil J Exelon Generation Co, Exelon Nuclear |
| To: | Dyer J NRC/RGN-III |
| References | |
| 50-456/02301, 50-457/02301, BW020042 | |
| Download: ML022890184 (132) | |
Text
INITIAL SUBMITTAL OF THE WRITTEN EXAMINATION FOR THE BRAIDWOOD INITIAL EXAMINATION - JULY 2002
Exelon Generation Company, LLC www.exeloncorpxcomr Exek(nS.
A. o1--i;", Braidwood Station Nucl ear
-- 35100 South Rt 53, Suite 84 Braceville, IL 60407-9619 Tel. 815-458-2801 May 14, 2002 BW020042 James E. Dyer Regional Administrator - , - --. x 0'S. A~t Renfiotn dminis rator A . - i, 801 Warrenville Road - -
Lisle, IL 60532-4351 Braidwood Station, Units I and 2 Facility Operating License Nos. NPF-72 and NPF-77 l-NRC Docket Nos. 50-456 and 50-457
Subject:
Submittal of Integrated Initial License Training Examination Materials
., .. E f . .. , ..... . ;. ....
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Enclosed are the examination materials, which Braidwood Station is submitting in support of the Initial License Examination scheduled for the weeks of July 8, 2002 through July . - t I ; I J. , 4 :.
19, 2002, at the Braidwood Station.
This submittal includes the Senior Reactor Operator and Reactor Operator Written Examinations, Job Performance Measures, and Integrated Plant Operation Scenario Guides.
These examination materials have been developed in accordance with NUREG-1021, "Operator Licensing Examination Standards," Revision 8, Supplement 1. Please note that reference materials are attached to each individual examination question or item.
Some minor modifications have been made to the Integrated Examination Outline with regards to the operational scenarios in order to improve balance and content. These changes improve examination quality and are in compliance with NUREG-1021, Revision 8, Supplement 1.
Some modifications or adjustments to the examination material may be required due to procedural changes.
In accordance with NUREG 1021, Revision 8, Supplement 1, Section ES-201, "Initial Operator Licensing Examination Process," please ensure that these materials are withheld from public disclosure until after the examinations are complete.
May 14, 2001 U.S. Nuclear Regulatory Commission Page 2 Should you have any questions concerning this letter, please contact Amy Ferko, Regulatory Assurance Manager, at 815-417-2699. For questions concerning examination materials, please contact Mark Olson at 815-458-7856 or 815-458-7829.
Respectfully, Ja&sD.von Suskil Site ice Presidenf - -
Braidwood Station
Enclosures:
(Hand delivered to Mike Bielby, Chief Examiner, NRC Region l1l)
RO/SRO Composite Examination with references attached Control Room Systems and Facility Walk-Through Job'Performiance Measures with references attached Administrative Topic Job Performance Measures with refedences attached Integrated Plant Operation Scenario Guides Completed Checklists:
Operating Test Quality Checklist (Form ES-301-3)I Simulator Scenario Quality Checklist (Form ES-301-4)
Competencies Checklist (Form ES-301-6)
Written Exam Quality Checklist (Form ES-401-7)
Examination Security Agreements (Form ES-201-3)
Record of Rejected KAs (Form ES 401-10)
pei aContinuous Rod Withdrawal D6uring power operations, a continuous rod withdraw] accident has resulted in an ATWS situation on Unit 1.
Which of the following is REQUIRED to align the PREFERRED method of emergency boration for this event?
fI Open 1 CV8104, start the BA transfer pump, check emergency boration flow >30 gpm, verify charging flow
>30 gpm g lOpen 1CVI 12D or 1CV112E, close 1CV112B or 1CV112C, maximize charging flow, isolate letdown i] Open 1CV11 OA and 1CV11 OB, start the BA transfer pump, verify charging flow >30 gpm .
§ Open 1Sl8801A or 1S18801 B, locally throttle running CV pump discharge valve to match 1FI-917 and letdown flow I
1H Braidwood 7/19/02 0A104 REM A1 aE 3. E ro2 V M -. r P E i yuu 0 opurate ana I or monitor the following as they apply to Continuous Rod Withdrawal: ..I .. ',. '. t Operating switchfor emergency boration motor-operated valve operatinq switcIh -.
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E (A). is the preferred method (listed first) per I BwFR-S.1.( B and C) are backup methods listed in FR-S.1 if (A) is unsuccessful.
only an option listed in OA PRI-2 .
(D) is Functional Restoration -ATWS Procedure 1BwFR-S.1 - f Step 4 1 4I lAWOG I 'I I I i _ I W..Mi Ii-w MMM_ I MMMMAoMoIttca4 mm~e I IbrNg WKe i Mng FjI I II I
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M000003A1 06 . 1 1EP 1 r 1 r L2]1
$~steoil vo i iiiI Dropped Control Rod
_ Ablity to operate and I or monitor the following as they apply. to Dropped Control Rod:
2 RCS pressure and temperature a RCS temperature and pressure will decrease with power irimediately following a dropped control rod. (D) is Correct.
!ligned Rod F1BwOA ROD-3 Sympto1101 BwOA ROD-3 Lesson Plan 11-OA-XL-34 Iii j Ij2,5 I_.________________________
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i n Inoperable/Stuc Control Rod The following plant conditions exist on Unit 1
- PDMS is inoperable
- Control Bank D, Rod D-12 has become misaligned from the rest of the group by 10 steps
- Thermal power is 100% and stable
- QPTR associated with N41 has just been determined to be 1.10 If QPTR cannot be reduced to less than 1.10 over the next 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, thermal power will be limited to:
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. 'a A ~v pliatin Baid ood7/19/02 FO S~a~mnt: Kowledge of the operational implications of thefolwncnepsathyplyoIoerb/SukCtolRd Axial power imbalance.
. 7or 1 f Per 3.2.4 - reduce power greater than or equal to3%for each1% over 1.00 QPTR will be measured El thermal power reduced within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and -
each determination. (A) is incorrect - the TS is not applicable below 50% power. (C&D) are oMiainedRodIf1wARd389.
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An automatic reactor trip has occurred requiring entry into 1BwEP-0, Reactor Trip or Safety Injection. During performance of the first step, the operator cannot readily ascertain if the Reactor Trip and Bypass Breakers are open. All Rod Bottom lights are LIT and all Nuclear Instrumentation indicates neutron flux is rapidly decreasing with a -0.3 DPM startup rate.
W 10000072 IEK2 03 IVae [1Xf OI Fox Sf EPE R : [l21 1 & fS 2 tl Reactor Trip II 1007 t: Ittte Knowledge of the interrelations between Reactor Trip and the following: ,; A' . ' . i - .
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00 - PlJ - V -- - - .1 I Ewpa 1a r IPer 1BwEP-0 Step 1, actions are closed bulletted therefore reactor trip breakers must be verified open or the RN applied to ._.1 W
f rmanually m
.OUlU trip the reactor. After the manual trip attempt the operators may proceed to step YU .0.41XlJCIl~a~ ., ~.S. R~ . ~.j .1. JIVO1 aov..
2 (B incorrect). Transition to FR-S.1 is not
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the status of Reactor Trip Breakers while performing the RNO of step 1 (C incorrect). ..- . -. . . - :. .. I IRecoI. . .aii~e~ .~~i~ e:. r.c :Sca~ : .&Ri~~~ Lb..............:
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Reactor Trip or Safety Injection BwEP- ,Step 1 IOW 3 Fox--- E1= o FailtnEa Bank uew o ~.~ Editorially Modified I~dbi~i~~gPorm L s 1997 Bwd NRC Exam l5Cd1iKtyper~ moi _V YV& =' m2~ g TVZ fRevis'o I Per L
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&ues~~h Pressurizer (PZR) Vapor Space Accident (Relief Valve Stuck Open)
A large vapor space LOCA has occurred on Unit 1.
The operating crew has implemented the appropriate emergency procedures The STA is monitoring status trees. and is currently in 1BwEP-1, Loss of Reactor or Secondary Coolant.
The following indications are observed in the Main Control Room:
- Train 'A' CETCs indicate 720'F
- Train 'B' CETCs are de-energized.
- Thermocouple Map Display on CRT #2 indicates Average CETCs at 730'F.
- RVLIS indicates 15% in the plenum.
- RCS pressure is 350 psig.
Core cooling is (1) and will be ensured by performing (2)_.
(1) j2) g jjADEQUATE 1BwEP-1, Loss of Reactor or Secondary Coolant Nil SATURATED 1BwFR-C.3, Response to Saturated Core Cooling Ed DEGRADED 1BwFR-C.2, Response to Degraded Core Cooling J INADEQUATE 1BwFR-C.1, Response to Inadequate Core Cooling
~wer:~ j j j ~x a mC ~ ~ j C mpre ensio J ~~ y FB~raidwood 71 1 2 E008A1 3AA2.16 Y1U 3.8 S1~iu: jB 1d~h EP F 5 jJtS ~
Pres ui ze Vt10 08 A etermine and interpret the following as they apply to Pressurizer Vapor Space Accident:
RCS In-core thermocouple indicators; use of plant computer for interpretation
- I li~n~tio 'tf (C) Correct - given conditions present an ORANGE path on status trees. At >700'F the correct procedure is BwFR-C.2. (A&B) are incorrect as the ORANGE path overrides the normal EOP and a Yellow terminus. (D) Incorrect - >1200'F required for this I w Status Trees I~a~~~ti~e 4~W tiou7 II=
1 BwST-2
__=_____=__=
1BwST2 sttustree & Steam Tables 1WOG
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QLS36 o2001 Bwd NRC I SuBa~?,. ry j Faci.t Li0 ]
Unit 1 is in Mode 3 RCS pressure control was lost resulting in RCS pressure peaking at 2500 psig.
Both Pzr PORVs and 1 Pzr Safety valve opened, then closed.
Operators have subsequently stabilized RCS pressure at 2235 psig.
This event is (1) because _ (2)_
(1) (2)
Reorabl Th Pz P_ anaeisweecalne IS I n- knf.I. k. l.. ,. -
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'JnIIYy tLie IVr OdIVLy Valve was challenged 91 I Nnt RpnnrtAhl.- PC'Q ---
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000008G430 2e 3.6 E 2eEPif Sy~tmJS~o~uion ¶I~ Pressurizer Vapor Space Accident08 Knowledge of which events related to system operations/status should be reported to outside agencies.
P TS 5.6.4.- Monthly Operating Reports. Document all challenges to the Pzr PORVs or Safety Valves. (A) is correct.
(B) Incorrect i asthe PORVs were challengedandisalsorepoble(C&D) Incorrect - it is reportable . . . . .
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Folowig sm llbrek OCA sme reactor decay heat might be removed by "reflux flow,,.
of each Partially filled hot leg pipe.
J-:i et ,,Sd 2 bottom of each cold leg pipe.
Li qfuxidheat hescreiubLiqid issuseqeaedenty y he or oold nsie he te m g nerto tu esandreurnd o te ore vacu tei along onowtetpo on eahprilyfle men ot egapipe. ip atoso hefi n cnet thiebto of each cld le pipe XA'_g
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E Establishing high-head Safety Injection flow t fReducing RCS pressure by opening both Pzr PORVs l Rapidly depressurizing all Steam Generators to atmos AW r~ jtvtRi~ eelCr rhninF~~: FBradwod Exm7/19I22J 0001210 lEA2.10 1~~~e 14.1 19~~~Jj 41 EPE i 1: S 5 B
~S~~~it:Ability to determine and interpret the following as they apply to Large Break LOCA:
Verification of adequate core cooling Eant~Wt~fRenitatin Hgh ea SIis hemos efecivemetodto ecoerthe core and restore adequate core cooling"l1BwFRC0.1 ;*
background document. (B) is the correct response. ,. .
1~rebl a Ail ' lv~iu ! I '_______
NR _ ____________.__
L jpfi"Ker Background Documents -Inadequat cooling
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pueto pii, Large Break LOCA Which of the following explains why it is 1 fI To remove RCS decay heat via natural circulation coolant flow.
To prov tide a usecdary h for Post LOCA Cooldown and Depressurization An A Er a E aii L V I.B I or - one F acilit y Bra w o o~d E a7t : / 90 00O0011 <3~03 1 EK.0 4. v au :3j S Wcin P Gr~ j R iDp 9
~'~tt~fte jLarge Break LOCA Fo011 K~i~tm~it:Knowledgeof thereasons forthefollowing responses astheyapplytoLargeBreakLOCA:
Strigaxliary feed pumps and flow, ED/G, and service water pumps x~naib~founddocs - (A) Correct. Since SG's will eventually be depressurized, water level will prevent primary to secondary Per beackage )Incorrect - steam generators are not required as a heat sink for large break LOCAs (temps and pressures typically Loss o Reactoror Secondary Coolant BwP1F BakrudD cuments iEP-1 Se3j5 C _____
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Ne Qes~'f-ot.oModifi'cat'i'd'n'M"eAIi%-d- I[I NR Li
The2 and No. 3 RCP seals have failed and continued monitoring of RCP conditions is required No.
I Xs cM2 F00~001 7A12~2 1H-i_ B c!gith v Cor prehnin Bridood Exm : XM 71/2 1AAI.2-2 RO au: .f$O~u~j42 e(~<EE RGo~[ JS~r p~ ~
t opreact oolantPump Malfunctions (Loss of RC Flow)1 i,X lAbilitytooperateand/ ormonitorthefollowingastheyapplytoRecoColnPum Mancin (osfRCFw)
RCP seal failure/malfunction
'natid'f Indications are that the No. 1 seal has failed. The operator action summary of 1BwOA RCP-1 states to go to step 12 which states to trip the reactor and the RCP. Due to the high seal leakoff flow, continued monitoring is not the proper action to take. A controlled RCP Seal Failure j1 BwOA RCP-1 F OAS gRCSLP lIJAP-XL-01 8
RCP Seal Failure LP 1l1-OA-XL-27 if, `i 4 i 10 3l
__ _ _ 11 Coihi1yE C ~I 71 lb, 'eiw Cmbif' pef 7
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Control rods step out due to a red K jControl rods step in due to rising RCS temperature J 1j V-129 closes causing letdown relief valve to lift
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R;I306I 71 1r E 13 1024 1i
~ Ability to determine and interpret the following as they apply to Emergency Boration:
When boron dilution is taking place E-(C) Correct - CC 30 fails open on loss of air, cooling off letdown flow. At lower temperatures, mixed beds have a higher affinity for
^ boron. Less boron in the RCS causes power/RCS temperature to rise. Control rods will step in. (A) Incorrect
- letdown temperature l will not fail closed or cause letdown pressure to rise.
UnotoldDltolfBwOAPRI-12 I Symptoms/tp 4100 1.CVCSLP j1C XL0(5a9104 Mat ~ _____ ______ ______ _J____ ______
_ _ _ _J_
Qu~i Surc ~~sin~ns 2000 Bwd NRC 1991 Zion NRCEa
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I1 au:40 $yaue. ~t~n EE 38 1 1 1j3~i': j oain024 1Eegn Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures.
a(A) Correct - Per 1BwEP-0, step 1 RNO, enter 1BwFR S.1 which will step 4 direct the crew to emergency There are no entry symptoms for Pri-2, which does not direct a reactorat trip borate. (B) Incorrect, (C) Incorrect - No indications of an uncontrolled dilution
~AbnormalOperatingProceduresfrfh=jf I1 U.BwOA PRI-2,PR-1 2,ROD-1 JB,2 . _. Rvso 58,10, Emergency Operating Procedures i BwEP-0 j Step 1 3 100 Functional Restoration Procedures J I BwFR-S.1 1 Step 4 4 1A
[New SRO Assessmentof conditionsandselectionof appropriateprocedures...Per-
QuetioT~iG Emergency Bration Per the TRM, which of the following conditions meets the associated MINIMUM requirement for the Boric OPERABLE in Mode 3? Acid Storage System to be considered
~f Acontained borated water level of 35%
- A boron concentration of 6800 ppm A solution temperature of 69°F A.
s JA flowpath to the CV pump via 1CV11 OA,Boric Acid to Blender Vlv E~aii~LeeI Cogntfe~ev l Mem ry - 'Braidwood7/90 acNOW:
kA 0002l14 AK1.04 RV1ue f 2.8 SR au: eto:EE R1 3p EEPEm Ku SR3 53~ j7~
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KK~t~~egi~ Kowledge of the operational implications of the follwncoepsathyplyoEmrnyBrtin Low empeatue limits for boron concentration Expanakj C) orect-of T M equres40 leel,700 p mand 650F.(A&B)Inorc.()
noret-sr ila efr o pth nlu s lI f ReactivityControlSystems IITRM I 3.1.f 2 1
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What further indications will occur as a result of this failure?
91 The RCS Cooldown rate and CCW temperatures will both INCREASE
~jThe RCS Cooldown rate will INCREASE and CCW temperatures will DECREASE The RCS Cooldown rate and CCW temperatures will both DECREASE F00E0025K2 03 AK2.03 R l al;E Val: j e EPE Lsof Residual Heat Removal Sy e F025 1 Knowledge of the interrelations between Loss of Residual Heat Removal System and the following:
Service water or iaD Correct - with 1FT619 faiing low, more flow will be demanded from flow control valve 619, more flow will bypass the RH Heat A Exchanger,lessRCSflowthroughtheheatexchanger willdecreasetheRCS cooldownrate.Lessheatis transferredtoCCW and Normal Operating Procedures - RH Cooling l BwOP RH-6 F 14-15 26 Bwd Big NotesRHiRRCodw1 3
F- -- ___
Qdatiorla Sou* frce~me
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~1~$p~c Loss of Component Cooling Water(C )
The following conditons exist on Unit 1:
- A normal plant shutdown is in progress per 1BwGP 100-5, Plant Shutdown and Cooldown
- Train A of RH cooling was placed in service 5 minutes ago
- 3 minutes ago the following alarms were received:
Annunciator 1-7-E3, "RCP THERM BARR CC WTR TEMP HIGH" Annunciator 1-7-E5, "RCP BRNG CC WTR TEMP HIGH" Annunciator 1-2-C5, "CC HX OUTLET TEMP HIGH"
- The following readings exist on all running RCPS:
Motor bearing temperatures are 1650 F Lower radial bearings are 170'F Seal outlet temperatures are 1 350F Operator action in response to these conditions will be to (1) because (2 ).
.(1) id Immediately stop all running RCPs RCP bearing temperature limits have been exceeded due to a loss of cooling flow Reduce the RCS cooldown rate CCW heat exchanger temperatures have I exceeded design limits allowed for RCS cooiaown 1 Manuallyactuate Si, enter IBwEP-0 A loss of all Component Cooling Water has j occurred on Unit 1 E Start additional CCW pumps More flow is required through the CC Heat
. Exchanger to control CCW temperatures
-. ~i,~rb Fj~aL~vI s Con vEevI Copeeso ij Fahy BadodEJ*ae 7/19/02 F60026A2 04 F - R Ei AA204 V2.51 2 F EPEo j R J tgaLoss of Component Cooling Water FfD = _026 0 11
~t~rm~n:Ability to determine and interpret the following as they apply to Loss of Component Cooling Water:
Imtr>,an IThe normal values and upper limits for the temperatures of (A) Incorrect - RCP bearing temperatures are well within
<235 0 F. (B) Correct- per 1BwOA PRI-6, with CC suction temp and the components cooled by SWS limits. Motor bearings <1 950 F, Lower radial bearing <2250 F, Seal outlet discharge temps in alarm, heat exchanger outlet will be >120'F met, a total loss of CC is not occurring. (D) Incorrect - increasing CC flow through one heat exchanger will only serve to increase the RCS cooldown - contrary to actions required in OA-PRI-6.
Component Cooling Malfunction
~64 I 1BwOA PRI-6 M F7f 100 lAnnunciator ResponseProcedures I1-2-C5&D5, 1-2-E3&E5 j Cause Actions i van FTech Specs ijBasis 3.7.7 I 7-3 1 0 F6 SROof plant conditions and procedure use. TS temp limits -
Basis Knowledge l r 1l t- v .. ~ . _________ _ __ ___________J F Ai RR61
__}E _. _ I5
ME = ILoss of Component Cooling Water (CCW)
Unit 1 is operating at 100% reactor power, steady state conditions. All controlling systems are operating normally in automatic. Operators are performing steps in 1BwOA PRI-6, "Component Cooling Malfunction" due to a slowly lowering level in BOTH halves of the CC surge tank when the following sequence of annunciators is received:
2-E4, "CC SURGE TANK AUTO-M/U ON"
- 1-2-A5, "CC SURGE TANK LEVEL HIGH LOW"
- 1-2-A4, "CC PUMP TRIP"
- 1-7-A/B/C/D4, "RCP 1A/BIC/D THERM BARR CC WTR FLOW LOW" The NEXT procedure that must be entered by the operators is (1I) because _ (2)
(1) _(2).
W BwOP CC-5 Component Cooling Water Make-up Auto make-up to the surge tank has failed and must be restored ii BwOA RCP-2 Loss of Seal Cooling RCP seal failures are imminent -,,i due to the loss of thermal barrier cooling a 1IBwEP-0 Reactor Trip or Safety Injection The reactor must be manually tripped and all RCPs stopped immediately wEn loads must be stopped/prevented from starting I1. J1 S CEn23 [ A _ B E te: . 7/19/021 000026G404 2.4.4 V ue3 0 S 4.0.44' Va e.j9 4-M'3PELg1 M S31 W IFEP qGOi 3J p: 7j Loss of Component Cooling Water 026 Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and .
abnormal operating procedures. : . fo em:rencyan.
iffang of 1(C) Correct - symptoms are of decreasing surge tank level and loss of all running CC pumps. Operators are directed to enter PRI-6 for the CC malfunction, and EP-0 if the surge tank decreases to <13% to trip the reactor and stop all RCPs. (A) Incorrect - leakage making this a low priority. (B) Incorrect - the loss of thermal barrier cooling is not a concern as long as seal injection is maintained.
(D) Incorrect - ECCS / safe shutdown loads are cooled by SX R-f Wt&0 -~
Cing Malfunction 1BwOAPRI-6 AttachmentA 1 100 lAnnunciator Response IIBwAR 1-2-A4,A5,E4 l Operator Actions 1 E I
_= =___ ___ = =
QiW S.e ISRO Assessmentofplantcondition J uiJr Per T~ II 1 - i Facility Li
_ _ _ _ _ _ _ _ _ _ . j L
If the pressurizer master pressure controller were to fail in an "AS IS" condition during a large, rapid secondary load rejection, which of the following will occur naturally in the Pressurizer to help limit the magnitude of the resulting pressure transient on the primary system?
fcooler water compresses the steam space in the Pzr. Steam is condensed to water helping to limit the overall pressure increase An insurge of hotter water heats the Pzr. More liquid then flashes to steam helping to limit the resulting pressure drop in the RCS.
f An outsurge causes the steam space to expand in the Pzr. This allows some liquid to flash to steam and limits the resulting pressure drop in.. l the RCS.
An outsurge cools the Pzr. This allows some steam to condense to water and limits the resulting pressure increase in the RCS.
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. Ex1e,j B Application Braidwood i7/19102 0000271103 A1. AE 1 S6R MO R 2.9l S W EP 2G Pressurizer Pressure Control Malfunction 1027 MN = IKnowledgeoftheoperational implications of the following concepts as they apply to Pressurizer Pressure Control Malfunction:=
Latent heat of vaporization/condensation 1(A) Correct - load decrease causes an insurge into the Pzr as RCS heats up and expands. Insurge compresses the Pzr but raising pressure slightly above saturation, condensation occurs which tends to limit the pressure rise. (B) Incorrect - steam Iinto pzr. - II : -
E !-To limit the potential flooding of the spent Aswrc Ex I S E_ gn Braidwood 7/19/021l A000036A203 2.03 13.1*=
IN E Iu; 2 S WE. EE -
Fuel Handling Incidents 1036
~~ Ability to determine and interpret the following as they apply to Fuel Handling Incidents: .
Magnitude of potential radioactive release bo Per TRM 3.9.d . (Old TS 3.9.7) Crane travel with loads in excess of 2000 lbs is limited to ensure in the event the load is dropped, the activity release will be limited to that contained in a single fuel assembly and possible distortion of or fuel in the racks will not Refueling Operations ITRM j1 3.9.d 1j~
TS (old) Basis lTS 3/4 9-2 Basis 9-2 Al5
~tto loic. Ne _ ~ $i Molit~ ~to:Ue ~rn~rIig~it Quti_ ent *o0o1 I - - - - - - I
{SRO 0 jTS Basis I Pad.K i _._._...... I tiv sR I
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is
Rj Gripper will not open when closed unless it senses <500 lbs 9 Only one drive; bridge, trolley, or hoist, is operable at any one time c1Hoit mrtionwill - - A; - - we-gnnon tne nosIis
. ; 1.50 lbs
_..: ------- Zvil9gwvlllup UIR-0011 nlweight on thenoistis 1500 lbs .%. V; K jA hoist slow zone exists over the full range of lowering a fuel assembly into the core ns-e b ~ ~~4.RMmr ai~~ raidwood ExW E/g0 000036<302 AK3.02 e J : SE EPE R Dg 1S3 G
.,. 'M F36 I-_036 K Iatein d Knowledge of the reasons for the following responses as they apply to Fuel Handling Incidents:
Interlocks associated with fuel handling equipment
.;p an'ato_ a (A) incorrect - Gripper will not open when closed until it senses <1 200 lbs (B) Correct, per LP 1>2700lbs (D)Incorrect -thereare2 slow zones,at.÷/-1 0" topand bottom.
(C) Incorrect - weight restrcins None in the middle r .
Fuel Handling 1-FH-XL-01 i . - I 20-22 J 52 16 Qu!~~ir~ NwQuestionNjifV ca o e....... ... Tra"'ngPoin
~^ .
._____ __ =_ _ _ _ __ _ _ _
I EoLi 1 [X1
Qusinok Steam Generator (SIG) Tube Leak The following conditions existed on Unit 1:
- 100% reactor power
- Small Steam Generator Tube Leak (5 gpd) on 1A Steam Generator
- A shutdown has been ordered to repair the leak If the Main Turbine were to trip, what is the MAXIMUM power level that the turbine could trip from that would result in the least amount of direct radioactive release to the environment?
1 [40%
I
...;60% .. 7 ; ; *¢ " :;'
00031 A 2 EFAK3 09 1 a 3OIale EPE R IS enerator Tube Leak 037 e reasons for the folio responsesas they apply to Steam Generator Tube Leak: -
I Maximum load change capability of facility
-1LY
Question rTopiI Steam Generator Tube Rupture (SGTR)
A SGTR is in progress on Unit 1, and the control room operators are performing 1 BwEP-3, "Steam Generator Tube Rupture". The operators identify and isolate the ruptured Steam Generator, and they cooldown and depressurize the RCS. When conditions have been established that indicate Safety Injection flow is no longer required, the operators are directed to stop all but one CV pump and both SI pumps. 18 CV Pump and 1A SI Pump were successfully stopped.
When stopping the 1B SI Pump, the control switch indicated a GREEN (after trip) target, but positive indications of pump amps and discharge pressure went unnoticed by the operator. What effect will this have on continued operations if the status of the SI pump remains undetected?
The ruptured S/G will eventually fill with water, and the atmospheric relief valve will lift. . : ' *..
K jThe RCS will quickly repressurize and experience an overpressure transient..
Excessive cooldown of the RCS will occur, possibly causing a PTS concem in the RCS. . : :-:.
lj Damage to the SI pump will occur due to overheating from insufficient flow through the pump 2aal 6 i~L 0e B o'iie~~ opeeso riwo xmae~7/11 9/0022 Fj 00038Al24 JEA1.24 3.6;1 S3.41 Se EE 13 =2 ivl Steam Generator Tube Rupture 038 ta t:Ability to operate and / or monitor the following they apply to Steam Generator Tube Rupture:
Safety injec6On pump ammeter and indicators . .
Igt3YI *.tilhd (A) Correct-per the reference document, SI must be terminated when conditions are reached in order to prevent SG overfill. (B) :.
Incorrect-.with only 1 SIpump the repressurization would be slow, and only reach the shutoff head of ,the SI pump which is -1500. .
oen to the RWST..
1w Ino 1WOG background documents 1BwEP-3 J _
= E IEI Mr .Q.,.,..d M-0-A ,, Significantly Moifed LJd D~i-`g-'`rr-a-`-nn A 0_.__ ___
7 [- _ tLW j .ciii Li
10.20 mr/hr 00 0 8w 1 l EA2.1 1 L a a : 37 f S ~ i S~tufifl ISeamn Generator Tub-e Rupture
- ~ . b i E PE l Li 1 FII K~s f~no:
Ablityto
[§038 dterineandinterpret thefollowingasteapltoS amGnrorTbRuue:.
Local radiation reading on main steam lines TS Limit for SG Tube Leakage is 150 gpd through any 1 SG. (B) Correct - a .1 mr/hr increase over background estimated leak rate of 150 gpd per 1BwOS SG-1, Fig 2. (A) will yield an Incorrect - yields about 75 gpd est. (C) Incorrect - >than the minimum
'F__
kit LI at it.J M OM Steam Generator LeakageEstimation f BwOSSG-1 I F, Figure2 4 4 Tech Specs - Operational Leakage CO34.13-1 F&..1 IA98
=- _ -J E= E= DI 7M fG IaeiiBwOS SG-1i LeakageEstimation- Figure2 (page12)
_~e 0 tC I I 1 I Wctio fT r eer I I upJis
Q &estipp I Steam Line Rupture A reactor trip and safety injection has occurred. The operating crew has entered and performed all applicable steps of 1BwEP-0, "Reactor Trip and Safety Injection", up to and including step 30, "Check if ECCS Flow Should Be Terminated" The following conditions exist:
- RCS Tave is 4857F and decreasing
- RCS Pressure is 1300 psig and decreasing
- Containment pressure is 0.3 psig and stable
- Containment rad monitors are Green
- Steam Generator Parameters:
SG: _ A __B__ __C D Pressure 1000 psig stable 1000 psig stable 450 psig decreasing 1000 psig stable NR Level 30% increasing 28% increasing 0% (no trend) 30% increasing i MS Rad.: Green . Green Green . . Green Given the above conditions, which procedure transition should have been made in 1BwEP-0 while performing the diagnostic steps?
l /1 BwCA-1 .2 LOCA Outside Containment . .. . '
1BwEP-1 Loss of Reactor or Secondary Coolant 1jJ BwEP-2 Faulted Steam Generator Isolation 1BwCA-2.1 Uncontrolled Depress of all Steam Generators
~ J* g Cf g t iv ~ 17 A p i at o ~: Br ai~d FC) wood ~ x r t~ t ; .7 /1 9/0 2 000040G4Q4 2.4.4 a43:1 4O 4!0V It E IBIe Steam Line Rupture . . -
aVora oprtn poeus.. . . .
Ability to recognize abnormal indications for system operating parameters .. ....
which are entry-level conditions for emergency and
.... ~abnormal op erating procedures. - :-'
I Per BwEP-Odiagnostics(27-29)(C)Correct -CSGpressureisdecreasinginanuncontrolledmanner.Atthispointin M [ no controlled RCS cooldown in progress yet. (B) Incorrect - cnmt rad and pressure are normal post trip readings. (A&D) Incorrect-Reactor Trip or Safety Injection 1 BwEP-0 22,231o g
L = L==
! F= 1= E1
e=I Ion On NrwQs lfo-eoew
[oiretType ]nrfDiil'l~~*>
< ~e s9je Pee i1 X NRC LI is
QLoss of Main Feedwater (MEW)
The control room operators are responding to a RED condition on the heat sink conditions degrade to the point that RCS bleed-and-feed must be established. critical status tree. While they attempt to restore feed flow to a S/G, The reason RCS bleed and feed must be established QUICKLY is to prevent:
J)Inability to provide sufficient injection flow for core cooling due to high RCS pressure f High temperature and pressure failure of Steam Generator tubes An overpressurization challenge to the reactor vessel A rapid RCS overpressurization, followed by a rapid RCS depressurization due to RCP seal failures. .,1 NSw
.Fra Extv B cfgitvee iM -emory Facflt~:& Braidwood Ejm at:7/19/02 000054A1 04 MAA1 .04 I Va l 4.41 O j45 1S g . 1 S~tii/rnii Loss of Main Feedwater
-~ < .Or'W _un F05 4 121 F!i
~ Abilty to perateand 1or monitor the following as they apply to Loss of Main Feedwater; .
HPI, under total feedwater loss conditions I
xp' ai3atggs t(A) Correct - per HA background documents. Early bleed and feed allows maximum RCS pressure drop, greater St flow rates j~d Ar0v~ .,5:: :ensures effective heat removal. The further the transient is allowed to progress before and bleed and feed is initiated, the smaller the
=Funt1i o RestorationProcedures l l A3 p00 IF BackgroundDocuments i 1BwFR-H.1 le &Feed3,35
=~~ =.-
=E= == i ILLi
___i I ! e
_ee s ]_
F1acmit Li NRC LI _
Nis~EU J~an~I~ JC~ihtv ApplictoFaI*jBrdw dE* t 71/2 Sy v' jnil Station Blackout 1055 ]
M Ot RE AbilitytodetermineandinterpretthefollowingastheyapplytoStation Blackout:
Actions necessary to restore power (B) Correct - per 2BwCA-0.3, "It is preferred to prepare 4KV ESF Bus 241 for the Unit 1 corsstie to suport the motor driven AF Pump availability. (A) Incorrect- bus selection isok, but Unit l 2 must align Train B loads to support Unit loperation. (C&D)
Response to Opposite Unit Loss of All AC ll2BwCA-0.3 J step N r O a,= = IIstep______4___NOTE__________________
== 1 I _ _________
= jLRc, Li r~-Ad is
Qu79 o To I Loss of Offsite and Onsite Power (station Blackout)
While performing steps of 1BwCA-0.0, "Loss of All AC Power", which of the following steps if performed in the Main Control Room will NOT result in Ithe desired action(s) because only DC battery power is available?
JReset Containment Isolation Phase A j9 IClose CC from RCPs thermal barrier isol valve, 1CC685 lSync and Close BUS 241/141 reserve feed breaker, ACB 1414 A~sie Lve cR E~a o~iii~v Lve~ Aplcaion FaPi~4 Br-aidwood ~ a~te7/1/0 000055A204 EA2.04 IEl S 1e , 1 EE _ I 1 Sye~~uinie jStation l Blackout W slblle S = _055 055 1 KS~te~~rij:Ability todetermine and interprett~ -followingasthey applyto StationBlackout:
Instruments and controls operable with only dc battery power available ElDC power suppies 125VDC for both ESF divisions, including Rx trip switchgear, MCB ESF section, ESF switchgear control systems. MSIVs
.. will close, cnme isol phase Awill reset, ACB 1414 will close. (A,B,D) Incorrect. (C) Correct - 1CC685 is a motor jLossof AllACPower __ _ __ _ _ __ _ _ __ _ _ __ _ .- --- E__ r= _=j Bwd Big Notes - 125 VDC System DC-i 3 1 I r IQutestin on it IrA toueiemmms
_ iF Pee ib I ___ _ L L RL:J
TLoss of Offsite and Onsite power (Station Blackout)
§Tefolwin coniton exsInnt1
- - IOSS or dalH, power occurred 20 minutes ago
- The Emergency Director has classified the event in progress as a Site Emergency
- P11 QWM di IUNRCu inital notifications nave been made as required
- Maintenance now estimates 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> to restore AC power to either ESF Bus
- The Emergency Director has upgraded the classification to a General Emergency
- The time now is 01:15 The State of Illinois must be notified of this change in emergency plan classification NO LATER THAN:
9 01:30 Ej~20 1I I I r7]
Lm F, I -
n9-1 r, I
a xmLveSC~iti eoyFWliy riwo ~7119/02 x~
000055G430 2O.21 H 3eF3.1 6 Sc3 E S 1~ 111 jSaio 3 lcot055 =i Knowledge of which events related to system operations/status should be reported to outside agencies.
Per EP-AA-1 14 "Notifications" - offsite notifications must be made within 15 minutes of any classification level change. (A) is only EiT8fi.iRf Correct time frame.
Notifications 4 =P-AA-1 4.1.1 1 1 .1mj
==. . I
= If_-__________-_
I I
Quefstifri 906vil. ~ur 1- E~iictoi~et~
Ne I-- rigann '- ---
I l
.rga -ns I_- I SRO J Assessment of conditions 9r R -
NMv Ij1of
____-J II Nger [iJ
___ u _
2i-
ti TWO Ls The following conditions exit on Unit 1:
- A turbine trip / reactor trip has occurred concurrent with a loss of DC Bus 113
- The crew has completed the Immediate action steps of 1BwEP-0, "Reactor Trip or Safety Injection"
- Transition has been made to 1BwEP ES-0.1, "Reactor Trip Response"
- Concurrently, the SRO has entered 1BwOA ELEC-2, "Loss of DC Bus" Which of the following describes why an operator is dispatched in 1BwOA ELEC-2 to locally open the PMG breaker?
rA I
,I IHalf of the steam ULWI - "--1 dimn v h
fIiA I F-I 10 ea-i--fn fro I
IProtec---t low fron, on/ IvlnA -
' , .1UE t Half of the feedwater isolation valves have failed open WOW = ~~h6l onti~Lv~Mmr
_= B~raidwood FE7/119/02 J
0 K3AK3 02 71 S 4.2° S MO EjPE J Vz y1 o058 11t§,ta;,te~i l t Knowledge of the reasons for the following responses as they apply to Loss of DC Power:
1--- I, ,,.-o-11 --ori-n
-. tnr
-n .-
- ,1-,n PVVVV[ .1 .;n>
7'. _ _
~
- .- pii~,. (C) Correct - the main generator remains connected to the UAT. 4KV bus 143 and 6.9 bus 157 will have lost breaker control
. ., . 1and cannot ABT. They will remain energized, will all attendent loads, from the main generator as long as the power
- - - -. .- ..... ' ' , .. t PMG remains closed. 5
, 1. t "' I l direction. (B) Incorrect- steam dumps fail closed and are not affected by DC 113. (D) Incorrect - feedwater isolaton valves fail. . - -
closed and are controlled via the ESF DC Busses 111 and 112. .. ; ;
I FAbnormal OperationProcedures Ij BwOA ELEC-1 Athe A 3 100 g g BigBwd Notes I IDC-1 1 3 l tBwOA ELEC-1LessonPlan l 11-OA-CL-01 II 2 6
}Que@@fr^np3, i4r Qusio FclityExa Ban Method: 3QsooiitoModified f lli L 1[11 1~--=J---ji Peer i E7 I - _.__-_
11 II I ]
tLU~#
II IFi~r 3
{ L l JIR_
Qe 6 IAccidental Gaseous Radwaste Release Which of the following describes the actions associated with the Auxiliary Building Ventlelation System radiation alarm on ORT-AR055 (Train A Fuel Handling Incident). upon receipt of a Fuel Handling Building high OA Fuel Handling Charcoal absorber Charcoal absorber Charcoal Booster inlet damper Fan bypass damper Fan (OVA04CA) (OVA060Y) (OVA051Y)
Automatically Starts Opens Closes Started Manually Opens Closes f [Started Manually Closes Opens I Automatically Starts Closes Opens .-.- I ANOeR a EiiLe ont Level Memry FacElty: Braidood Exa7/1 9/0~2 F l 66 A60K 20 AK2o0 2 O lu 271 J3EPE 11 [ 21 6 tAccidental Gaseous Radwaste Release 1060 KA~~t~ii*:Knowledgie of the interrelations between Accidental Gaseous Radwaste,Releaseandtefloig IAuxiliary building ventilation system , [
Et~t~i~f Hi rad interlock from ORT-AR055 provides for auto start of the FHB Charcoal Booster Fan, auto opening of the charcoal absorber Aii~~ :~<.inlet_(and~outlet)_dampers,_and~auto closure of the~charcoal absorber bypass damper. (a) is the onlyvcorrect answer. 1
Qus onToi IAccidental Gaseous Radwaste Relas IWhich phase of the Large Break LOCA Accident Provides the basis for shifting of Auxiliary Building Ventilation to the Emergency Mode?
i 1jReflood Ej 1Recirculation IA302
__ 1=2il AccidentaleGaseous Radwaste Release i:
= -t .
Knolege f hereasons for the following responses as they applyvto Accidental Gaqseous Radwaste Release:
Isolation of the aventilation *~
- x Per TSBasis - 3.7.12 (D) correct - design basis is established by the large break LOCA. Assumes a passive ECCS failure outside I co nta inrnme nt, such as a SI pu mp seal1, wh ich o ccurs-d uring colId leg rec irc. (A-C) incorre ct - occuri ng inside of con tainmren t.
TS Basis jIB3.7.12 I Basis 37.12-3 0 11_
I 1 1_ ____J l J[}=
Mat fl I I ~~I t ,____
= =U1O_ , S I aIiiity ]
Q2uesttdi Area Radiation Monitoring (ARM) System Alarms Auxiliary Building General Area radiation monitors provide all of the following functions EXCEPT:
I
~ Trending of current and past radiological conditions 1 Local alarms for personnel protection Detecbon of unauthorized radioactive materials movement I I M I "OIL U1 MUA DUIIU II IY L.AIdfL;Udj toaster Tans 11 IMemory = Braidwood l
__1 = 12 1 : 27 I lliJ1 s~i~ii~odi~~,tii& Area Radiation Monitoring (ARM) System Alarms 06 i n ieW ~ i ~ : . k, > X : ~ 7 > i a 1 1 F ~ i S :
I uv\w euMUo systYrLI purpose ana or function. I Ex.!aiMiono{ SAux Building Charcoal Booster fans auto start upon receipt of a SI signal only; Not from Hi rad. (A-C) are, correct functions of the Ad2K2.w'. 1 general area radiation monitoring system. (D) is the incorrect (Correct,)answer.. s .
I- ---
Normal Operating Procedures IBwOP VA-5 E 2 10 .....
IRadiation MonitorsLP Jl11-AR-XL-o (49) IIl.A F1 2 1--_-------_______________ E ==_
IMtra I 4 iNew I 3f litnh I ~
I ~ ~ ~ ~ ~ ~ g
___________jW0 §
'3 W~,>>,jgi f:DII~tjU 1T : NOniR' Mi r I
ffim".M MINEM
------------------- ___j I_ ~i_ __ ___
I--,--.-
... _ ._ _ _ __ _ _ I
1 ----
C Li
The following conditions exist on Unit 1
- Power level is 100%, steady state
- 1A SX pump just tripped on overcurrent
- 1B SX pump could NOT be started
- Only 1 SX pump is available on Unit 2
- ra I Init 1 rouur crua orio i systems.
.o~tin systems
+;A
- What-actions are taik t redqu I ceth eh atloa oA K
Oo I ONE RCFC train on each unit is shutdown and isolated _
1~. 'I
~~~
All containment chillers on BOTH units are stopped and isolated I
U71
= I OA now LUWIK~ur-t. s on UNt unit is isolated I - ------- -
I I
I -7LI1 9/02 1000062A1.01 R W F-3.11 S e EPE E m Iij Xgi' - i Sysigvoh~on .tfe~Loss of Nuclear Service Water06 KA#.'(emnt:l~biity ooerate and /or monitor the following as they apply to Loss of Nuclear ServiceWater:
I Niiripnr ,.-n/w,- wnt., l-t1lUl ---
1-,
v -avcssIlsaul . .
I - 4 I(A) incorrect - isloation of 1 heat exchanger on each unit invokes LCO 3.0.3. (B) Correct- per 1BwOA PRI-8, attach B step 1bs. II RNO. (C) Incorrect - renders control of containment temperatures not possible (D) Incorrect - renders RCFCs incapable of I i l
_,,IurrjilgtCilpo LIC iJJiCiOLi L: ICLiit I I 116 Ilo .,1F,, iZ.
Q~s'o Su' I rn~fnt
[C~netTpeVg7; A 4 i~
I-.-fi,"'
Li
The following conditions exist on Unit 1:
- Reactor power is 100%, steady state with all systems in automatic control
- A secondary transient is preceeded by the following indications:
- Annunciator 1-21-El0, "125 VDC DIST PNL 111/113 VOLT LOW" alarm LIT
- MCB Indicator 1E1-DC001,"DC BUS 111 VOLTAGE" indicates 0 The IMMEDIATE action required to be taken by the operating crew is to:
1 Assume local emergency control of safe shutdown equipment
] Start-up/restore the 125 VDC ESF Bus Battery Charger . ' ' [
] ie the 125 VDC ESF Bus to Unit 2 ESF DC Power . .
j9 Verify Unit 1 reactor and turbine are tripped and ESF Busses are energized FdI s-L ApiainFclty riwo iii~ 7/19/02 Answ( LE Fj4Exani Ever'2 1000065G449 294e : E [ S[
e Loss of Instrument Air .1065 Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.
Ejp i f Loss of DC Bus 111 has occurred as evidenced by annunciator 1-21-E6 and DC, Bus voltage indicator. Loss of ESF DC results in, i .
loss of IA to the main feed regulating valves, which fail closed. Resulting closure of MFRVs results in or required an immediate i La :- :...-
lI I 101 I. t _..L loss of power. Train A equip will not be operated locally wo tripping protection if an operable train is available. (B) Incorrect -
indications exist that the batery charger has tripped off. (C) Incorrect - While it may be desireable to cross-tie DC Busses at some point, must first determine status of why 111 has tripped. Before that, the immediate concern is the failure of all FRVs on Unit 1, lowering SG level and imminent reactor trip.
IAbnormal Operating Proc - Loss of DC Bus I1 BwOA ELEC-1 Attach A 3 100 ii ____________________________________________________
II=__=__=_ ________
... __ _I _ _._
~ Nm1"ew IS ==_________~~~ During Trainin PrGog m I IQ~kguK I~n~e SRO~ ii1()Assmn fcniin eeto faporaepoeue Imediate actions)l E ,
KC __~~"! RC L J__
p Loss of Containment Integrity Which of the following transients is analyzed to result in the highest containment pressure AND greatest leakage out of containment?
l JS Desg bai LOCA i Design basis Steam Line Break inside containment 1 Ilnadvertant containment spray actuation Pressurizer vapor space LOCA AF~r Ea~~~ B- FMemory Fcly:Bra idwood ~7/1 9/02 000069KO1 [AK1 01 R W 2.61 e 3 tE f
'Loss of Containment Integrity 1069 KA~tterin~: nowledge of the operational implications of the following concepts as they apply to Loss of ContainmntIteriy Effect of pressure on leak rate exjahat6. Worst case LOCA generates larger mass and energy release than the worst case steam fine break. Inadvertant CS actuation would causepressure todecrease.evenif allRCPseals failed a DB LOCAis a largermassandenergy~release... .~
.nt 1-FR-XL-05 TechnicalSpecifications 1l3.6.4 I 2 j1 3 lf3 Basis B.34-'
Qi i ~ Fac ili y x a Ba~n k j Q et ~ i i a~ o eh d Direct From S u c f ui ' r i g r g r ORMV 1 M1f Question ir $m nt 2000 Bwd NRC PMI
%I M Faim Li N C__i
QInadequate Core Cooling Unit 1 reactor is shutdown with RCS pressure at 485 psig and decay inadequate core cooling situation, what pressure must be maintained heat being removed by the steam generators. In order to avoid approaching an in the steam generators to obtain a 50"F subcooling margin (Assume a negligible delta-T exists between the RCS and the in the RCS?
R f285 psig
- . )465 psig . :- c -i .: : :; ' l E jl665 psig .
785 psig A6Wer a ExmLv~B Cgi~Lvl Apication Faciftty,, FBr~aidwood Eabae 1190
- 0 007 8 };EK1.08 1 l~fe 2.5 j SRH~e 1J S[3~s EP RO'rp Li l 173jj Cooling074 KA~Sf~t~h~Knowledgeof theoperationalimplicationsof the followingcocpsa hyapyt ndqaeCr oln:*-.
Definition of subcooled liquid
- xp1~naW~f~n calculated value with the steam tables for a pressure of 485 psig. (500 psia has Tsat of 4670 F. 50F subcooled 417T is,300psa25si is 417TF Psat for
,Ina CoreCoolingLP i - u2 1
dequate l!-IT-XL-01 1 jM a Steam Tables l.I1_7 if W using steam tables to determine saturation at 485 psig and 50F lower - to 285 psig (A)
_ _ _ _ _ =1 II u Suer~sr
Cing Header Backpressure Control Valve.
1RH-607, RH Heat Exchanger Ou
. 1RH-619, RH Heat Exchanger Bypass Flow ConrlVve 455B, Pressurizer Spray Valve.
Lve~B ~ Cl~e plcto MAe Braidwoo 7/1 9/02 4n~ii:
KA~$a Kowledge of the interrelations between Inadequate Core Cooling and the following:
I Cntrlles ad ositioners p~a~i~t~ief. B. Correct. Decreasing demand on this controller will reduce flow from the RH pump to be injected into the core because it is on the 3 discharge of the pump and normally aligned 100% open. A. Incorrect. decreasing demand on 1CV-182 will not decrease charging asprayvalvewillnotdecreaseflowtothecorebecausethere wi eno srunningtoaffectRCSpressure.
System big notes ws CV-1,RH-i
,3 jOp asU~t ActionSummary Page Ill wEP0HY1 TipEC-sWhe Ek~iii~tIi~ None Q I QGS Wn6doh-(R NewWUs 1I I I Peer2Cmz.{
II Ad ii..ii o
c ontrol Rod Drive System The following conditions exist on Unit 1
- 90% power, steady state operating conditions
- All systems are operating normally in automatic
- Without warning, control rods begin to step
- Tave begins to increase above Tref which remains constant
- Pressurizer pressure is increasing
- Pressurizer level is increasing These symptoms are consistent with which of the following events?
E One control rod has ejected from the core
. A SG PORV has failed open A continuous rod withdrawl is occurring Il A pressurizer steam spaceleak has developed FBisAt v e Copehnio aclty raidwood 7/1 ~9/02J E347 00100000K3o02 K}T 3 1 K3.02 OVte J eto5SS19 3Oau:34J r~ F-9 I? R9rif 7j i Control Rod Drive System0 Sbta fn~egi Knowledge of the effect that a loss or malfunction of the Control Rod Drive System will have on the following:
., tiri§UP!,l l(A) incorrect - pressurizer pressure and level would decrease. (B).incorrect7 pressurizer pressure~and level would decrease asl Tave decreased (C) Correct - all symptoms of rod withdraw[ and Tave increase (D) incorrect - pressurizer pressure would decrease.
RodControl LP 1jl1-RD-XL-01 1I 1 2 1,20 E ===--= E=r ,
MA fQuestio
~ Facilty ExamBank Mddifc~tion efI~b& Significantly Mdfe .nn.rga r- ? _ - - I Q. s__ _, r _ _ _ _ __ _ _ __ _Ir Ue_ _iih1~
aJ
Que~riFToj:o 2 -trol Rod -DriveSys-tem ~
Unit 1 is at 100% reactor power, steady state A total loss of power has occurred in data cabinet B for the Digital Rod Position Indication (DRPI) System.
What affect does this have on DRPI?
i System accuracy shifts to +10, -4 steps I
h I PRnr :t Pmftn iI;kf., - ITr.r 4- -I.
t 1DRPI Urgent failure alarm annunciates LM F-J, I- - I 1UHLAIU1I I
I Anwr xntvlB ~ -----
CgiiER L~e eoy aiiy raidwood - 1 ---- 7-/19/021
./.02 nI 1 0 0 13 11 IK6 u e 1 Y 3 [
,ftonlt e Control Rod Drive System 001=
KA'~~t:rKnowledge of the effect of aloss or malfuinction on the following will have on the Control Rod Drive System:
~uet~o J~ Coolant System Top ~Reactor (RCS)
Which of the following parameters should be used to differentiate between the sized RCS LOCA inside containment? early stages of a moderately sized steam line break or a moderately IPressurizer level c E7I30 L!v CIiv g H'~
=002 Knowedg ofthe effect that a loss or malfunction of the Reactor Coolant System will have on the following:
Containment
~tioj
., E(C) correct-onlyRCSleakage will cause actual radiation levels to increase. (AB,D) direction regardless the transient or LOCA) incorrect - all three willchangein the same AT I[ntro to EP LP . l1-EP-XL-01 : acc ID chart
.Mae M7rRi.f6(4'ainE'tjz%.n .. tW,/fi'.
= .
iJ__________________________
iLi l_;~c1f2tj~i Li
_7______________7_7_
NR Li7
ED IK6.03
. I IReactor Coolant System02
.1SRO E F
tX S 1 X
=2 S 711j
=02 Knowedg oftheeffect of a loss or malfunction on the following will have on the Reactor Coolant System:
Reactor vessel level indication
_only 1 train of RVLIS is left available of the 2 total. Each train consists of 2 reactor head level indications and 6 reactor vessel plenum level indications (a) only correct answer.
K[LT Big Notes lCORE-2 RVLIS
= I_
=__=11
___ I C tTe* Gyilp In]it =eCo i/h~~NR >>a I
I; rlaciity jee Licg
_ q1
QuesI Reactor Coolant System (RS)I Durng a recent degassing operation of the RCS, Volume Control Tank pressure as level was raised. This caused Reactor Coolant Pump (VCT) level was increased to 70% without any concurrent adjustment (RCP) #1 seal leakoff flow to in VCT restore seal leakoff flows to normal. (1)_ , and will require _ (2)_ to I
Oecrease z Venting the VCT .-..
DDecrease Opening 1CV182 I00000A20.5 Sjs~ol~j~bi~~hfe 19.V~
A2.05 e FE5 Reactor Coolant Pump System03 N W J eS S e Ability to (a) predict the impacts of the following on the Reactor Coolant Pump procedures to correct, control, or mitigate the consequences of those abnormalSystem and (b) based on those predictions, use operation: : :
Effects of VCT pressure on RCP seal leakoif flows ; i .
I~la,~ I,1~ f (C) correct - increasing level causes pressure to increase, increasing backpressure flow. Venting the pressure from the VCT is the requiredaction on the RCP seals,Adecreasing #1 seal leakoff to take while degassing the RCS.(A&B) incorrect- backpressure.-
compensate'forincireasdVTpressure.
V-. SeaIl ekoffswillriotretu rRr t ,
ULno.rmal-parameters.
Af&Crn't6nZrra[Saramter pe Mechanical Degassing of the RCS j BwOP CV-14 l 6-14 I FK- _ -JE~- -
Matrrnal IQuein; Soarc C 26aiiei2t I ...
ei . ..
___ ___ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _L
QuetioTofc eactor Coolant Pum~p System (RCPS)_~
During normal operation. the Reactor Coolant Pump (RCP) motor windings are cooled by (1) and the RCP #1 seal is cooled by (2).(2 jAir CV.
CCw Cv :.
0000(0 IK4.04 1S S Vue: 99 Setg~ SY 1G~: [9JS Iu:fj nowl e CooantPumpSysem M esg emo ury 1 eco olntPm ytm*'003 Braidw F)andorineloty pooid e-fo xolloing Adequate cooling of RCP motor and seals Eiplanation of (B) Correct - air cools the motor windings and water (CV) cools the #1 seal (seal injection) , .- *,.
Aigfrma:' m i , ,,.le , M ,
A M .
RCP Lesson Plan 211,4 1,4 F131-RC-XL-oll
_ , =dlfoU=arn ic _ew$1 rl ueston 3 r_-~ -- i.R- -
-i _> §2 0Ji ~~ 7Rves~mp~
Pee s rv-ry Faif
Qu~tovir~fc Chemical and Volume Control System (CVCS)
The following plant conditions exist on Unit 1
- A small break LOCA has occurred
- Control Room indications suggest that the Pressurizer is solid
- RVLIS head and plenum indicate 100%
Which of the following describes the effect of changes in charging and letdown on the Reactor Coolant System Small mismatches between charging and letdown may cause large and sudden RCS pressure changes.
fLarge mismatches between charging and letdown may be required to induce small changes in RCS pressure. ,
ijJSmall mismatches between charging and letdown may cause large and sudden pressurizer level changes.. :
RCS pressure will respond exactly the way it responds to charging and letdown changes with a bubble present in the Pzr.
Fl4000A109 A1.09 .f j S e t E [3 I C m l d o e n l004 Ability to predict and/or monitor-changes in parameters associated with operating the Chemical and Volume Control System controls including ng:..
RCS pressure and temperature
~Iafi~K~t. (A) Correct - water is an incompressible fluid. In a solid condition, small variations in inventory (charging / letdown flows) will result me Em in large Pea variations in pressure. . - - ;i CVCSLP 1A-CV-XL-01 RCS Fill and Vent - BwOP RC-3 j F 161 14 Ma teril E. = I=
C&iF etpel -' l~i~ :07j Reiw I.,
'j"n- met l~~~FC++n'ye2r5ilty j
==2 -[_TC row____
[ G C
Chemical and Volume Control System (CVCS)
Two (2) minutes following a reactor trip with a loss of off-site power, which of the following motor operated valves will NOT have power available?
(Assume no operator actions) 1CV8104 "Emergency Boration Valve" 1CV1 12D "Charging Pump Suction from RWST Valve" 1J 1CV8105 "Charging Pump to Reactor Coolant Sys isolation Valve" 1CV8109 "Positive Displacement Pump Recirc Valve":
de RC o mpreh ensio n c y Bra o 7/19 /02 19ystemEviuo _il ChmcladVlmeCnrlSse 1004
'Knowledge of bus power supplies to the following MOVs EIna,;of~
p , i(D) correct- the only non-esf powered MOVin the list. 1CV8109 is powered from 133V1. (A,B,C) incorrect- these are ESFpowered
.,M !N i. _..... W MU 77GR&: ab] Iti:W,' 7 L CV ElectricalLineup BwOPCV-E1 If======_ F12 5
- ~F-- ===1=- == 0 _lit L1 N M -j r1 I
____i LI~er R Li
Thein foo Residual Heat Removal:
IThe following conditions exist on Unit 1:
- RCS temperature is 340F
- RCS pressure is 345 psig
- Pzr level is 54%
- 1A RH train is being aligned for shutdown cooling
- 1B RH train is aligned for injection Shortly after placing the 1A RH train in the shutdown cooling mode, the NSO notices RCS wide range pressure and Pzr level decreasing. A Field Operator on rounds informs the NSO that the 1 RH pump suction relief valve sounds like it is lifting.
Which of the following describes the possible cause of this event? (Evaluate each response separately 1SI8809B, RH TRN to RCS Cold Leg Injection valve, was inadvertently closed :
E 1CV131, Letdown Pressure Control valve, was left in AUTO when the 1A RH Pump was started 1RH8716SA, RH Disch. Header X-Tie valve, was left open prior to starting the 1A RH Pump -
f~J1 RH8701A, RCLoop Ato RHPump Suction valve,inadvertantly closed due to aninstrument failure Fo IE$wr
~~ MMMEMM~[f-J i
1gu Lvy .i~
uomprenension FcI~:Badod1 V i 9o 005000K3.01 K.1 aOV1: 3.9 Secioii SY 1 jJ Sl5i1: fI
~ ~&~1Tt
$ysirn/v Residual Heat Removal System 4
li i t~atent: nowledge, of the effect that a loss or malfunction of the Residual Heat Removal System will have on the following:
I RCS .
1 .
1 1 E (A) Incorrect - closure of the discharge will not increase suction pressure (B) Incorrect
- RCS is not solid (C) Correct - per ..
reference, closure of the disch X-tie valve is performed to prevent overpressurization U of the idle pump suction header. (D) Incorrect -
i PlacingtheRHsystemin S/D Cooling I BwOPRH-6 FF 12 22 6 --
E= = = F= a 1] - E 1=-=1 lity__Exam__ kI eFa itorially Modified- -- II C>,*,nt typ l~fl~
- _ 7 E REmgI ws '
___ _er] _
"a] NF1 d_
hesfollopin i Residual Heat Removal TIhe following conditions exist on Unit I n egeo heo aso id n plant": pre 7net m iatosown n[a~;~rl 311 Shecppytio: tr~
eitResiReidualHeatRemovloSytemSy. te Ptar osps d r n soi p an' pr s u e c Exfa~to ng d e to her elative in compressibility of w oJ(B or ec.L o s fRH cooling f ow willallowR O to ater heatup.1CV increase. willsensepressure 313- dropa sR p m pi trippedarid 'CIOs (C&D)incorrect RCSwillheat upasRHcoolingis lost Gure I0-f 01 I .5 _I 17
'CO w 1en l BMM-
[WNe 1Mdi st1ueoW ot i~ ~ TAm e
ai _ _ _ 'Fli I I~SupeiMiso_ _j
__________________HC(
QEmergency Core Cooling System (ECCS)
The following conditions exist on Unit 2
- A large break LOCA is in progress
- RWST level decrease requires the operators to transfer to
- 2RH8702A and B, "RC Loop 2C to RH Pump 2B Suction Cold Leg Recirculation
- 2SI8811 B, "Containment Sump 2B Isolation Valve", Isolation Valves", are both CLOSED is OPEN Which of the following actions MUST be performed to OPEN 2SI8804B, "2B RH Hx to CV/Sl Pump Suction Isolation Valve"?
S2SI8813 "SI Pump Common Miniflow Isolation Valve"
~ OPEN 2S18807B "SI/CV Pumps SucinHae rsteVle KOPEN 2CS009B "CS Pump 2B Sump Suction Valve" fj1CLOSE 2SI8812B "RH Pump 2B Scinfo WTIoainVle Ant~ ee e ~ o r ii ~ e e~ M m ~ I c ~~ ra d o d b t J7/ 9/0 2 1 X
- § ~ urisiif§'.
. R j j407 E r f J Sr : F2K4.17 1
Emrec Cor Coln Syte =. .
_
- 006 A~ta'eme'~ Kowledge of Emergency Core Cooling System design feature~s) and orinterlock(s) which provide for the followvingi:
Safety Injection valve interlocks o l(A)Correct - closing either (2S18920AND 2S18814) or closing 258813 satisfies the rest of the interlock to open the RH crosstie.:7 GPs -MainControlBoardValveInterlocks jg2wP1013-fr I= = =+! 7
'cf=
Qttii .-......-.............
- --.1....- .......-....... --...1.11 I.............
I..............
oi PrsuizrRIif Tank/Quench Tank System (PRTS) II The NSO has noted an increasing level in the Pressurizer Relief Tank (PRT).
Which one of the following RELIEF VALVES might be discharging to the PRT? I II f
.1. I-,. ' 4 ! " ' ', , jr 11, I fl, ? j - I ,I . r: ' :I : 1: J; .
.1. I :, ! - ;.','r. it": .ALr ,-; I
. I . . . " . I -- - 1. . -1
. j I
- 1. . - 4. - . , " ...... 6, 4 -. ; -i , , -.i:,.-. :., ., I I., I A..",I
. I
.i . i.
1CC9426A-D, RCP thermal barrier relief valves
[S18856A/B, 1~ RH Pump discharge relief valves" ... ~<;.K OReor FjEamp radoo 7/19/021 1007000A301 A310 a 27f .j coR
- op[j~ 3Q se~ouon*.jPressurizer ReifTnQunhakSytmQZ .:,,
j 100.77]
i MeMntp Ablt omntratmtcoeaions of the Pressurizer RA6iW Tanik/Quench Tank System including: :
rIIComponent which discharge to thPR ., - .. - .1 - ... > I (A) incorrect.-I CV81 18 reliev s to the VCT.P) Corrqct -:reli6veS to the PRT,(
C)Incorrect-relievestotheCnmtBidgFloorDrain--,-p.
umO (D) l6c6h-ect - r e 11' ieves to the HU
- %, I . . I I--- 11 I a " . -.. ---_, . ".. ." Ji..:, _ 1 7., ; -1 . 1,::` I
- FI i . .M ..lf I I FWAN , .., _
P,.&D M-64sheet5 . [II fij
[ ______ I . Ij
[f4 ~IF I 1SKO diF-1dr1-I iti tod . dtrilyMoiie Me t rr~rann rga 12001 Bwd NRCI I p 1so LI F acilit___
48
eComponent Cooling Water System (CCWS)
IA leak in which of the following components will result in an automatic closure of 1CCO17, "Component Cooling Surge Tank Vent Valve" J .1SealWater Heat Exchanger * ,. i '. .. .
Spent FdexPool E xchanger-, -.
I'Letdown Heat Exchanger ~ ~ '
fWaste Gas Compressor Heat Exchanger.,'.
tComponent Cooling Water System
'.I, 1.l1.0 l_-, . .-- e' . f C oi gW
, t Kowledge of the physo s e-efect relatonships between omponent Cooling Water System and the.
.. RMS ...
~x ~t~~
6~ 1C) orrct RS letdown is at a highe prsue than the CCW system :and will result in inleakage to CCW. (AB,D arei all at l peratirigpressure
,,ower than CCWa w resltin utekg ( w P 6'
., ZY . 1 - ;:,
qcI
Eio .i rediagnosis With Unit 1 operating at 100% power, the following events occurred:
- A reactor trip, coincident with a loss of Instrument Bus 114
- All systems responded as expected after the trip With NO operator actions, 5 minutes after the trip Steam Generator water levels will be HIGHER than normal post trip response due to a delay in ISOLATING AFW flow and the Rediagnosis procedure 1BwEP ES-0.0 shoulcldbe -
used.
HIGHER than normal post trip response due to a delay'In ISOLATIN AFW flo and -the Rediagnosis procedure 1BWEP ES-0.0 should NOT be used. . ..
~ LOWER than-normal post trip response due to DECREASED AW flOW;~nd the Rediagnosis procedure IIBwEP ES-0.shudbusd v,
~ILOWER than normal post trip response due to DECREASEDAWflow -andtheRediagnosis procedure 1BwEP ES-0.0 should NOT beused.
-w-t,
- 7/19102 OOWE0EK202 EK22 1R j 31_ . a. I E R aKnowledgeof theinterrelations between Rediagnosis an e owin:.-
Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations betwen he ropropraton f tiese systemnsto the o'perafi'on of the facility.-
A loss of inst bus 114 will cause B train of AFWflowcontrolval o rfow is sensed through themThis reduces total AFW flow to the SG's, reducing poststrip I e set ln of AFW vice 2seof red thnmltedu is t'those vavsfiepoe o lse .Icr e ve-re ~nosides not apply vi~ct eo rediadoi isii Lossof Instrument Bus I1 BwOA ELEC-2 Ta18ble jD iRediagnosis Ij1BwEPES---0 if 7E 1i--
IC ~pIicAP in IFclt riw
.LOCA W0 E G1 1.14 i 25 S a j 332.Fed E o L Outside Containment02 l FKnowleg o sytmsaucrtiaw chrequirethenotificationofpateron...*
~~ ~(C) Correct - the assembly of personnel is the next action to be performed (before the evacuation). This is accomnplished undrE-r j~ AA-113 ferncfor onsite personnel (A) Incorrect -assembly must j
-I. - -c-obe before evacuatioh
. . to give a full accounting
. agj.~of all onsite personnel (B)
- first
-, 5
____________4__
__ __ __ __ __ __ _2___ __ __ __rn Protetve MPersonne Actions Ine-Am113 f J!.. I . = 3 Til = :
51
Q s p iLOCA Outside Containment _ _ ,,.,, _ . .. .. .......................
While performing BwCA-1t.2, i "LOCA Outside Containment", under what condition would 1S8835, "Si Pumps to Cold Leg Isolation Valve', remain closed after being repositioned? ---
RCS pressure is increasing .. i. . '... . -
I' = _
F r'w v OOW II.jOIVj
.... !tA
= -
ng f~mn~rnti roe .ro 4---o.,er., .. .: ' I~, ~---
~j x~ e e ~jcA. iiL e Apliction I Aaciity Bridwod am~te>7/19/02
- I.r - ' I r Kt I E LOCA 2 Outside Containment 1 E .V: E04 .. I
, j l Knowledge of the reasons for thefollowing responses as they applyo LOCA Outside Cortainment:
Z ; ,
- ..... INormalaonormal and emergencyoperatng procedures associated with (LOCA Outside=Containment).
A)Correct 2 --------
as stated ihthe procedure 1BWCA-1 2 as the leak itssure increase. (BiCiD) are notp
__ 'L ' '
n
'1 U-tsgiven-inthe procedu
._ i.I Mt ,_
,, .. , n e LOCA Outside Containment (CA) -- 1BwCA-1.2 AHWOG
.. -I . i.. .A...., l .f i _ __
fm II 1 MM I
- I I ea uunIr a ni g om I
A .rAAgot A.M~,A ~ ~AA.l A. .A. . . ~ ~ _ .. Revews o~ pet1
.5z
9pc lPressurized Thermal Shock The following conditions exist on Unit 1
- Steam generator 1A tube rupture has occurred
- Crew is performing the initial RCS cooldown step of 1 BwEP-3, "Steam Generator Tube Rupture"
- All RCP's are OFF
, - Loop IA Tc indicates 180°F The STA has reported an ORANGE path on RCS Integrity.
The Unit Supervisor should (1) because (2)_ :..-
ReMain in 1BwEP-3 until the secor0d Cold injection Water is' cooling Loop IA
'CS -ep-r'ss-rzaonis cort 7.
co . -
~nneitl ransition to. 1BWFR-PJ.1, A -e~ecalneexists to theCF,- ':.S ..
Transition to 1BwFR-P.I as soon as Cooldown in 1BWEP-3 takes priority the initial cooldown is complete ov6ri 1w BtFR P1 P.
Remain in 1BwEP-3 until the appropriate An RCP will be started in 1 BwEP-3 O EA2.2 j3 3.51 0 ii 1 EP [ rou
,_- r i hr il 0h;1V 6- l termine FE and interpret the. following ey apply to ressurized Thermal Shock:
-- ., Ahrnet
------- prpriat procedures and opertonwithin te limitati'ons i the faiiysicnse and, amendmerits._ . 1.i (A) cor ect per 1wE~J Caution prior to'ste'p 6-rd No6f prior tostep 8ti1) IncorctOfancatoh'sttesN Tt~O F 5 7t1at thstmGC)lcr~t d - caution states to wait uni opeino tp2 Dliorc-cuinsywait until ste28rt SGTR - i1BwEP-3
'10, 36 low
_ _ _ I - E I_
,g-_j
Pressurized Thermal Shock Which of the following reflects the intent of the m Reduce RCS cooldown rate and decrease RCS pre colo nrtefn nres
....sRS coolEd C rssure
_ c,.
fincrease RCS cooldown rate and decrease RCS pressure Increase RCS cooddown rate and Increase RCS pressure
- ~A. ~ ~ 1K1 6~nif e~Vqv Memory Badod ~ mae
'.. di, 1EC' P16c 222E2L.. lEK3.2 0 k j3.61 13 r1-3 .1 [13R 1 g Prsurized Thermal Shock ,j8 l a l a e o p al Shock)=
p r stna ess(Cn (A)
Correctin r a i g t e c o d w a e i c e s s t et e m l s r s e . ( )icorrect, increasing RCS pressure increase th strsss orectreucs
() teralandvpessures sesses,per1BwFRP.1 .. * ',,
,,,I _ I Rlesponseto ImminentPTS 1lBwFR-P.1 ,
1I3______ A,__
Response toImminent PTS I Background Document P- 1 j ..1 . J
-- - ---J - -:-_
CJ
........ ........ ................. ~:-........
..I..... ......
Qiies~'t on ope [ Natural Circulation Operations ...... .... .............. ....
If operated within TS limits, all of the following preclude the hot fuel rod in the core from undergoing DNB during a loss of forced coolant flow accident EXCEPT:
.QPTR :
fiFNdH 1 _' .. Wri 1 1
A. j 'L[ j ontLie MeoyFctt.Braidwood
.. ~ 3 a 19e0 00WE09A202 EA . t a l 3.41 S- ali r 1.
Sy m~ou, Natural CirculationOperations ,..J..*I~K.
.. E09 fi e Ability to determine and interpret the following as they apply to Natural Circulaton Operations:>. *,. Z;*
Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments.
" RUM~i~i$ {ABD norc-perTS basis eacih'pro'tects against..DNBoin lo~ssoffore.flw
~321.pR324DB35 C or~
\_ 04 e
- je;r ! , !.-'.'. '. j's' it , , Ft,
- ~
w§l 0 - 0 0, imber1.:'g I . .. ,' ,, ,,,., '.I . .. . _ .~~~~F . 1
== j EII......T
==- F_=__________ =! i= _
TechSpech- Sassec:,.and3..Basi.,3..4k.2owledge~T i
Iaj I I .~c~~
Su2vs Li NR L 5S
p~ jNatural Circulation Operations Which of the following describes why it is important to run CRIJM faswe efrigantrlcrulatiion cooldown?
Provides the heat removal mechanism for the vessel head area . . : .
,!A i rsi -. ttur>crc circulation-fldw f~ i s tMi: throughs t e ..-RC v s h
_TeSe
. i __ __ _ __ _
fPreventserratic indicaion ofSRinstru fAids in natural circulation flow through the RCS .. - J, I... . i. .- ...
ti1
==
!Sr,;-
- tv n*Ver
^J ... Eeviie -j ,6g Mmr riwo
/90
.~. ~ %A ~ Knwlegeoftheinterrelations between Natural Circulato Operations andthefollowing:,i~~~.
Facility's heat removal systems, including primnary coolant, emergency coolant, the decay heat
~ removal systems and relations beteen te r)per operation of th'es'e systiems to-the' ojp'ration of Ife facility .. ,,
CRDM fanscool b cOdbynaturalcirculationflow, RxCavityventfansprovd ooing to,he SR Ni's.,,..,'-.
- t.
rg :':D
.. L Ps -Se ee.. i o L .
11vatll1ral-irculation~ ooldown - - i 1pBwEPES-0.2 . -: j Step22 RNO i 14 I .. 114 iWOG ii _ _
1CIRJ}
j I!
,"Facility Exam Bank~. ietFo ore Jd~rigTann fl r kI
=I MUM 1999 BwdNRC utIRNIEIM, M _W DietFo tL= piz I 1:ed
-'----'-s--
vrso NM 0 1 1lk] Li
. -1. ..-....11
. .. . .. ... . .... . 1- . . .....
et RoplG nU jNatural Circulation with Steam -.........-.....
1.11
..........- IL-1-
Void with/without RVLIS IThe following conditions exist on Unit 1 Ii -1BwEP ES-0.4, "Natural Circulation
- RCS Temperature is 450°F
- RCS Pressure is 800 psig Cooldown with Steam Void in Vessel (Without RVLIS) Unit 1" is in progress
- RVLIS is NOT available
- Charging and letdown flows are matched With RVLIS NOT available to monitor for void growth in the vessel, which of the following combined indications can be used to verify the presence of a void when letdown flow is increased > charging flow?
I I
RCS pressure will (1) and Pressurizer level wilr. (2)i. , . .-..........- , . _ -
I <-':..,i. . 'i.f :t ';;r , ,;-:i . '\tt .-
.. ,.I
',ncrease Increaseq- . . ....
. j n r se Decr ease 'A I6 4y.. ... . .
Decrea.selwih/ioltrhIg Decreas wiga ya pyo f te ~ * . -'_-
Operaing ehav or c aracteristics of the fa~cility. * .
~p~ j(D
~h~ton Core t -ra idl icrea-sing pres SUri level dUring th e ~C .S depre-ssurizati n is a sign that voids r for in inth6'pHrim ry;i
________ ____ system . Pressure dcreases, fluid flashes to~steam displacing pz-r level. (A&13) incorrect - ressure decl-ase s h eno s 13 'P~iqyenory i
.. IA-ICI~VC CYII t, COC ~..q C Vl7JOI7??II..
I kAU NatC ircC Cooldownw w/o RVLISI ll1BEPES-0.4 I lj7 1 1AWOG BackgroundD DocumentsI i 1BwEP ES-0.4I S tep8 3 °3 CI n X
=,_ _ _ _ __ I j j uesi onS' -- c: . w I eoni> se u r g riiir o~ i iI I I I I Su e v ~ i F]
Ir
~i~ 11
l i SteamGenerator Overpressure The following conditions exist on Unit 1:
- A spurious closure of all MSIVs occurred while operating at 100% power
- The reactor was manally tripped by the operators and immediate actions of 1BwEP-0 were performed
- Recovery operations are in progress utilizing 1BwEP
- The STA has identified a YELLOW path overpressure ES-0.1, "Reactor Trip Response" condition on IC SG with pressure at 1240 psig
- Checking 1PM04J, there is no steam flow indicated on the 1C steam generator
- All other steam generators and plant safety systems functioned as designed
- During subsequent repairs the unit has been holding in Mode 3 for the past 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> The condition of the IC steam generator is reportable to the NRC because: . :.
- fThe plant ehceededasaf lnit. .
.'".,i.'.'.;"..
JChallendes666c~re to saft vale *6 .'.-
loss ofiwfis'ion 6ioducth freirs ., I6ii e g JThe plant rs in a condition prohibited by tech specs
~a~ee plication F7/19/0Baidoo OWE13G430 2.4:36 2.2 g s A . 3 6. SP 'PE [ ' Ki 5yW~i;;Ev6 i n it jSteam Gentorverpressure i 'r t reportable i f primar;'syst.m power 'levl a ressure :ly. ( Safet valves a o ;'"'t P
_ _ _,_s
' o. '.of2 prdchtsi t mopra'reo e utsi a S ite En er nn;h* w I tsr Ir ,,l su tVItin, K, i r- Wf 's I. -=_
Functional Restoration proc SG Overpress 1BwFRH.2 H - ee=i FR=H.2 backgFRoun doumn FR-H ExelonReportabilityManual _ lS 1 2-4 - :
Steam Generator Overpressure The following sequence of events has occurred on Unit 1:
- Reactor has been manually tripped due to a secondary system malfunction
- 1BwEP-0 has been performed and a transition made to 1BwEP ES-0.1, "Reactor Trip Response'
- The STA has identified a YELLOW path on the Heat Sink Status Tree for steam generator pressure
- The crew has entered 1BwFR-H.2, "Response to Steam Generator Overpressure"
- The crew is preparing to dump steam from the affected steam generator
- The US reads a CAUTION that does not allow releasing steam from a SG with a narrow range level of greater than 93%
Why shouldn't the crew dump steam from the affected SG if NR level is >93%?
EJ j May causean uncontrolled radiation release since it islikely ti'atfthetsteam generator is ruptured .
May result in two'pha-i Kw andate'r hammer, ietrti'a' IIydangir pipe's, r IValve J....... . ,. '-.'i>. .7 jWill be ineffective in I w'erin ' g8 ~ter'slk r s ue i c he S y su co td .. ' V. : s...
g jWill cause a rapid pressure drop in the RCS, potetiallyiesuifin i=na safety injection C'a F -e _ MF M 7/19/021 OOWE 1 31<202V
_S
,:[FEecon- EPE Ri rti: V.1 I-Ate t Knowledge of th interrelationsbetween SteamGenerator Overpressure andthe following, Facility's heat removal systems, including primary coolant, emergency coolant between the proper operation of these systems to the6operation of the facility the ;.:
decay heat removal systems, and relations . .
{cInio (B)Coirrct - per FR-H s'eriesbbckgrdUnd ciumets.() ncrrctno indications, are present that wouldssetaST]a l occurred. (C) incorrect -at,1235#ajnidRCSTa've' osf ip, openingte PORV or SteamDumps willreleasIsteam (D)Incorrect -
, hoebumps ewnc:rc or~.. easeseao ) r66r ResponsetoSGOverpressure -
- - j1BwFR-H1
[Background Documents : - IIiFR-H. . .,,:t l-Wis*i___*_______
- pl jHigh Containment Pressure While performing actions of 1 BwFR-Z.1, "Response to High Containment Pressure',
what steps are taken to limit the peak pressure rise in Icontainment in the event on-e of the steam generators is faulted?
OOWE14K202 EK2.2:,,. RQVI3 _ 9 e 0' Kowldg of intbn High nContainment Pressure a tf oo vin:- t. ' E fKnoweg fteinerltosbetween High Contairimenit-reisstire and thp fn[lnwingri.
- Facility's heat removal systems, including primary coolant emergency coolant, the decay heat removal K' between thep)roper oPeration of these'system~stoithe operaon systems, and relations ,,
of the facility~~;.
.. 7 f~a 6~' (~) Corect -pr step 6 of FR-Z.1. (A).incorrect(-'RF af6enever run in fastseed in'adverse containment odtost rtc the fans :(C) incorrect:-AFW is only thro ed'to 45gprr if all stearm generators are faalted (D) incorrect - all stean enerators -
. ,,"ner . stfon y rted to,4 1 e ee dn rge.
,Responseto High CnmtPressure lI1BwFR-Z1 , , I lAWOG j 1- '=-IL . l I1 I
.... nto I _ _ _ __ _ _ _I CoeC~p cne~
anI~s~.- .. t.....
II f Sjrio L E=FacII it L NRC li
a IHigh Containment Pressure ,
1BwCA-1.1, "Loss of Emergency Coolant Recircula60on", is in progress when a RED path is identified for containment pressure. 1BwFR-Z. 1 "Response to High Containment Pressure", is entered immediately and containment isolation is verified. The operators then operate the containment spray system according to the directions found in 1BwCA-1.1, instead of 1BwFR-Z.1.
- Under th e the 1BwCA-1.1 pump operating criteria:
Ensure that the a um;heatreImval sst capacity iS used to reduce containment pressure..
/
~re oreresriciveensuriidigcontinuous contai m ntsray syst m o eain6rd
. ;Are lessrestrictive, permitting Wuc c ofi m n rsue edcontainment' sprayoperation to conserve R-WST water................,'.%.'.
- l E J Provide a more rapid means of verifying automatic actuation of the containment spray system.-,.-.-
4 1. w~
a ~ x J~ n ye L ve . A pi c t onF diif FBra idwoo -d 7 19 0 OOWE14K30
~i~t~nit7~
Ste HighContainment Pressur..
31 ~rV~u&IZ.71. e o'.fEPE ii~:J SOGu Jj
. 'Ž~~~"
1 JAZ teme Knowledge Nor _umal of the reasons for the followingresporises
., nrrd3P as they'apply toHighContainment Press;ure:..
t.' . l abnormalandemergency oteratingproed ressassociated C4.4........... _., _Q . .'*,
- w. (i I CninmentPressure)., ! . *;. '
xniV Coyrect - spra ope rtion requirementsiare relaxed to.allow conservation of RWST maiized but reduced '(B) incorrect - 1.1 criteria is le6ss restrictive, water nventory (A) increct i t I..1
~allowing n~o'CS 'pumps if allI RCFCs are a land ruanng
.: > :4 _ X. .... _
i,. _T.
e er~i~e. IMitl Fa~ii, eerriceui '.
,MWd M. r~.Ief"ceoo . . '. " agN
~.
I .
_ F Exa B Ie'SE.: __ _ _
Qi est i 6 _ E r M e _ _ : _iI ra ce gur e I J S pervsorJ j1 NMac~1y L
NNNNjN jContainment Flooding A large break LOCA has occurredonUi1.Teceiscretypromnstpin1B following conditions existed when P-"LsofRaorfScndyColt.Th the STA made his initial scan of the Status Trees:
- Pressurizer level was 0%
- Containment spray had automatically actuated.
Cnmt pressure was 12 psig and decreasing.
- Containment rad monitors 1 RT-ARO20 and 1RT-AR021 were in ALARM.
- Containment floor water level indicated 65 inches.
Which of the following procedures must be entered to address the above containment conditions?
~ 1BwFR-Z. 1 Response to HighCotimnPrsue Ji BwFR-Z.2 -Respos to Containment Flooding
- ~~~~~.,' .. ............... '
1BwFR-Z.X
.. Responsetoo High.Containment Radiation Level '..:,.gi+. .i. .
~ BF-..~.Rsos oLwPressurizer Level :i 4..
J 0mvLee' Aplicati on. j f l~ Braidwood Exmae, . 7/19/0 r0W~E15A2011 FEj_7 H EA2.fID&, R au 2.71 9JSet~
- jEPE R.Gi RGoq
, r -2 Flooding =C=ontainmnent
.E15 I
- Ability todeterminleandinterpret thefollowing asthey applytoContainmentFlooding:
I Fcilty onitions and selection of appropriate procedures during ..
abnormal and emergency operations.
I Sxplin' Vo~
. (A)lInco~rrectv- cnmt pressureis<0pi arid not requiredeob adnife .
dniido fo 6t nr yhhe M T.()Cret loigetyp s4Y 8 r- finches. This.s an ORANGE endpoint and the highest in the conditions present-.!(C) Incorrect
- Rad-monitors in:ALARM is a.
.Containment! inventor Status' Tres BwjTj
_______:~
ii a 0-e~
F_= --- Egl
_
i==jELi r---I AlSup'sowon t1(R L (o
'[ii no1 on~WHih tainment Radiation The following conditions exist on Unit 1:
- A small break RCS LOCA occurred 45 minutes ago
- The reactor was successfully tripped and Si actuated
- RCS pressure is 900 psig and increasing slowly
- RCS temperature is 500°F and decreasing slowly
- Pzr level is 25% and increasing slowly
- Containment pressure is 4 psig, decreasing slowly from a peak pressure of 22 psig
- Containment radiation levels are steady at 2.6E5 R/hr
- All S/Gs are intact with NR levels at 27% and increasing slowly
- The operating crew has performed all applicable steps of 1BwEP-0 and have transitioned to 1BwEP-1, "Loss of Reactor or Secondary Coolant" Which of the following sttements is true concerning the curndln oniions? .
-TotaluxFeed Fowr aybethrottled back to thntred0'eRCS toocldownffects, ; + jj ritti,tsra *, Wbe ~tped atariy time'deeh"ed ap rpria te- to congerve
- .: ~ ~
-,hi ,~b.otannei Spray.......................................
RWST-ivrty(-
I ~ ', ..- _ ~
t ., ;,: . ~ ."bIT :- *1g:
RCS subcooling is acceptable and would allow for SI termination if all other parameters are met g [Pzr level would require SI be immediately reinitiated if it had been previously terminated . . ,
~f ~ a~iL FB e e eI Application ailiBaiw od,;i A 1 9 00E 6A EA1. 3.11 SO a5FP [ 21
.I High Cotaihr -Rdiation 0 aei5;Ability to operate andI or monitor the followirtfa theyapply to High Contnmn Rdaon: ,
C-": omponents, and frtbns of controiad safetsy'ter s,;inudi instruentaion,signals,interIock failure odes, and uto nd maul . ,. . . .
if~~iktOd 'id FAdvers'e hontanmfc6ndiffions+eiit due t6-ffie'hi~gfi -a'd Ie-yl's'(>1 56rfhr)' -(A)-fn'o'rrect - requires pincorct - requiresrntmof SG lavels betwe'en 31'-5&/o1.`(B)
>2, hours.(C) riorrzct- 500 1 'requires950psig advers ni D orc Pzr.eveis requ. r-dd o LosO eco rScodr oln E _ 1BwEP-1 J ASses3,7 F K T i o~owo TI ~
re?,R --l'...- 7___
...-....1 ..................................-.................
I- - -I ..................
7r--- ...............................
- --- ....................... . ......-..... ....I...
I - . ...............................-. - . . ... ............................ ...........
-. 1111-M777 3F I -..- I __- 1. -
Which of the following Pressurizer level channels isNOT density compensated, making it read lower than actual level at normal operating temperature and pressure?
7.. .
mmI I-: II;.- ".i- 1% . .
,I -
.: Intel!..:
I I 7 I 1 .. ; -
t r .:! :: ;-
I., I .:!
11.1',.,
-- 11.11 , , 4, __ p -',-
'..` , I ,.'.: -,. j 'r .1` ,.,
,, ,. ; ,. 1:. , , ),. . " , .-
- ,:,. ,I.. . ,: ,' I .
, :,11 . .
, ,m I .
, i : T 'k ; I, ' '. I rLT-461:"" .;
1,4e1 .. rA' I Mill-- .:- I ., '.". 11 z" ; , ,
!analyzing-!" 6 1.
I J Ciiive~L~e>
~ns gJ xar't~ i~Meory Faili~ Baidwod ~Dte] 7/19/0~
011000K403
~ystF-T
. l e 1 OiIE 15YS j 2 6.0;1 ion.__2___2 j
I [ G . I!, j,
, ; t'.1.4 e Control System , 4,
... I e zM Knowledge~of Pressurizer Level Control System design feature(s) anAritroksthchpoiefrtefloing: : :- z ,
,. I .-tk I I -----
Density compensation of PZR level . ..
__, - i I - , I ;_ ._ ; I I - I I Sf,1
~5cp~naa L-46 sno :~ desil copenatedor aliratdtore'a~d a-ccurately at pressurizer temp~rat teof 4.(-A. f L ';
Ex~~~A;ifl....
680 F. (D)S.iscret :1'-
1 I PressurienP.. .,. Jll-RYrX C-01 >71K2 P I
I I IM MMI I FacilityI Exam -= BtF a:I u n lu__ . . ~I ----- __I I i _ 1 1MF I -_= I I!
MCD
jPressur'zer Level Control System (PZR LCS)
Instrumentation, LCO 3.3.3?
A..II
~1LT-459 alnd LT.
LT-460 and LT-462 .'-'i*."
~Stte~ n:Knowledge of the effect of a loss ormalfunctionon
- the following will have on the Pressunz;er Level Control System:,~
Function of PZR level gauges as postaccident monitors t
- r1' 3 r r nnelsoperable., L 4 ar usedfor Anstrument 2of tliee' inoperable will require entry into Conditon B foIAS333 s
( coret Coi W l011l ..
Of4entu y' ',i -.
I .
,. .Facili_
A- 0; ,n- - - -
p IloReactor Protection System Given the following plant conditions on Unit 1:
- The reactor was at full power with all systems in a normal, automatic lineup
- The reactor tripped on a LO-2 S/G narrow range level condition
- Reactor trip breaker B (RTB) did NOT open as expected I
I With NO operator action, the steam dumps will open on a signal from the (1) controller and will control Tave at (2)
I L (2)
-:- , a
.. _ ..... ~...- .. .....
I .. _ L .' . v i . . _
0 a
UL II Load reject 5A °
-v-F
. I I
I WIPlant ..m trinr rr 0
-v- - °F I
01200OK30 Ksn7-2 Kthe effect that alossormalfunctionof the ReactorProtectionSystem willhaveon thefollowing
-- , i i 4iflow01 ; -1,4'. - -Y . .; -; 4 ! .;, : .. - , !.7--"' .. 1-
. . I : , , .. . W RTA arms the steam dumps in the plant trip mode. (A&D) are incorrect as dumps will only be armed for the load _l Temperature will be controlled within a 3°F deadband - in this case from noload Tave. (C) is then correct.
rejection (>10%) l (B)Is incorrect I
. A. _ .
_ A_ . . .
OeaoBig Notes 7 j M4 F--J F-- __
II II ____ _ F - = =_
I M or~e O"f
~~to
.uv'C I
EAJ Facility Ex-am Bank I
-Iul n o alw toi jDirect From Sourc m'D7--_7.._--..
UF-u-n -'-
a jl
.I -. _ _--.
tt I
1l
'I' 0 YWY113, I,
I-
- ' WN Y, IL V
MY *-- 11 i III i 5
I I -- - -
I i
I -- fCfC~y D IL i
--- ii NR 60-
ieo I Reactor Protection System l
Which of the following reactor protection system tiserves as a BACK-UP to the Power Range Neutron Flux - High trip and is designed to ensure that the allowable heat generation rate (kw/ft) of the fuel is NOT exceeded?
~ OTdT
.. OPcI .
fPzr low pressure !, j . ! . .: '. ' . :
fbR MIA
. w- .
JRCS low flow E......_c
--,.rn.....Ev>1. , j Contv. vi . Meoy , 't Bradwoo..d . * *i~ *e i 7/19/0
. g! -
F5p20lK5O0 -i- - j K5.02 1----'@ i [
1012 .. J KAff e ~e ti Knowledge of the operational implications of the following coflcepts astheyapply tothe Reactr Protection System-., s . ,
FPowerTdensity-,
r TS response rps - r - -n-Tech Specs_ _ >fi. 1, I
- . .. =3 1 17 1j II- Fm
- 2 -Z eistloSo~ce e r== S
=--
le urrgiruig1rg E==
= 2I S§4'&~iso
,Fuf CI (ol
11............... .........
...... ...... . .... ......... _...... _. =n....... ............ ..
ow s l Engineered Safety Features Actuation System (ESFAS) _ = ....... ............
The following conditions exist on Unit 1:
- Safety Injection and all ESFAS equipment has actuated and is functioning as designed.
- The Emergency Director has declared an ALERT condition exists
- The crew is performing the actions of 1BwEP ES-1.2, "Post LOCA Cooldown and Depressurization"
- No CSF higher than YELLOW is in effect
- Annunciator 1-6-B7," RWST LEVEL LO-2" has just alarmed Which of the following actions should be taken?
IA Site Evacuation should be ordered AND the people directed to assemble at the New Training Building
.... ' 1,.:.
A plant announcement should be made wamingp psornne torestrict entry into the Aux Building due to potential highraito .. 4 The event should bereclassified asaiGeneral Emergency AND th'eNRC, Statea6n'd local governments notified immediately.:
16' Protective Action Recommendations (PARs) determined AND State and local governments notified within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> following evaluation l
0I300G1l14 72.1.14 2.51 S aS R of systemtstatusscriteria which ,. .
IKncwledgeofytmttsrtra crequire the notification of-plant personnel.
of>(A)Incrret- oradiologicalsafety hazzar warretng evacuation Is in progrest.
h (B) Correct - witchover to recirc may cause, levelsin the aux building. (C) Incorrect - No General Emergency conditions have been met, (D) Incorrect--; PARs are only Trnfr66-'e,dIJSwEP ES£.2i CUTO i;r gxAin,;
EmergClassifications& PARs IEP-AA-111I 1~'j'-~
FO-erth ci1=izt~ =
An.w j _n
.. ,~ e . ... _ _i ...
I2001 Prairie.. Island NRC
. . . ' , . . . . '.. . . e . . . _ iw Je j PerHI.
Sue'sr I]
l alty I
/,,la
--- -...."'. 1........-....'.'''''''-....'...:"................'
I.................
...I...-.........
Eepi I Engineered Safety Features Actuation System (ESFAS)
__ _.. ...I---..........._--1.1
..I........._..I---........
Unit 2 is presently at 90% power and shutting down due to an extended loss of Instrument Bus 214. All systems are in automatic when a loss of reactor coolant occurs. When pressurizer pressure reaches 1829 psig on 2/4 channels, which of the following will occur?
Automatic SI will occur. Train A ECCS equipment will 'automatically start. Train B ECCS equiprrlent must be manually started._ :
b jAutomaticSl 'will NOT occur.Train .,., ._ ~.. ~
Aand Train B ECCS equiprent willautomatica'lly startw.hen Si .is-manuallyactuated.
I
- -tom1-i- _ . - - - - - t - n - I - .Ab
-Jf = omatic:l. willdcur-0-an1Aan Train ECGueuuinmat wca Iluatec.-I-.., 'I :. ly-ac
~ utomatic SI will NOT occur. TraiiiA EGGS equipment must be manually started Traih B ECCS.-equipmenr cannot be started..
A f MJ;~ME FRfij pliato Badwo 7onivte1 01 2 $ }IK2
.Ol1300O yo t EngineeredSafetF a's yeaturesActu1ate0 R381 f _S@ I B :f3
- 1jC 13 A_to nt
- rKnowledge of bus power supplies to the olwn:
sfeguards equipment control ' ',
~. ,
~~ SI w~~ill automatically actuate, both trains. Train BES F'Loads however have lost the relay (energ ize to actuate) and ilntat tr
~~jasdesigned.They can'bemanually stgrted.:
,O~)OC L)II.UIC...;.101 f tVl CUt)
()Corrct: ()i Icorrect -SIlwillactuate,Train;Bwill CC .s> -.
notautostart' ()Train B willnot I (tq
ues~o op~ engineered Safety Features Actuation System (ESFAS) I Concerning the Engineered Safety Features Actuation System (ESFAS), there are (1) channels of narrow range steam generator level instrumentation on each steam generator which input to (2) independent safety trains of ESF.
(1) (2)
A &we.; ~j x ~m Cogniii a~Ve Me ory acil Braidwood a Ml e 1190 E013000K501 7FK5.01 7 2 JjB ei j fufto~ Engineered Safety Features Actuation, System *;
1 l Knowledge of the operational implications of the iowing concepts as they apply to the Engineered Safety Features-Actuation.. jl..
PeIritihofs fty train and ESF channel
.anato- There are2 independent trains of ESF (A&%j Each train w receivean i input from each one of4 level channalson each ,steam-gSF purposes. (TS 3.3.2) (C) is oHl e-re' 4c erne~co o.Rvson ~ r~ .:',',ciye ir i ,.. .... .......
... A, . ..... .......
-121%
..1. ........
II--.............
'I IRod Position Indication System (RPIS)
The following conditions exist on Unit 1:
- Reactor power is holding steady at 1x10e-8 amps during a normal reactor startup
- Individual and group position indicators show all control bank D rods at 120 steps withdrawn When the NSO begins to withdraw control rods to raise reactor power, the IR NIS indication suddenly drops by 1/3 decade and continues to decrease at a negative (-).25 DPM. There is no significant change in RCS Tave. The control bank D step counters now read 121 steps for both D1 and D2 groups. DRPI indicators for rods D-12, M-4, and H-8 indicate 0 steps. All other rod postion indicators (DRPI) are unchanged.
Which of the following has occurred based on these indications? ; - . -
The control bank step counters and associated DRPI indicators, along with the NIS indications are consistent with multiple droppedxrods..
- jThe individual rod position indicators appear to have failed, more than a single dropped rod would have resulted in a reactor trip~ e',ls,-~i.. [~ii.
J The control bank D group 2 step counter has failed, it should also read 0 steps if the rods in this group are fully inserted fEither the control bank D group step counter or3 DRPI indicators have failed, not enough information is provided to determine which
- j ogn ,y . Apiain I f .Braidwood A 7/19/02 0102 K0a 31 02SYS .e1u- S @ 2 J1G3
, vRod Position Indication System 1014 F lKnowledge of the physical connections and/or cause-effect relationships between .Rod Position IndicationSystemandthef;ollowing:
I I . .-.-. . . .... ...- ...
pi~na~o~ Indications provided areconsistent with multiple droppe rods. (A) is corre"ct. (B)incorrect
- thiereactor doesnttifo e ae I i -;,
________ii:_ .(C) incorrect - group counters are demand indicators only. (D) ncdrrect - given NIS response, DRP or group demand-countershave ONIIKO, leecicd geNoreevon
-I !
Abnormal Ops- Dropped or Misaligned Rod Symptoms; R- -i Ki-f 1BwOA ROD-3 _ 1 i 101.:
I~---E-..
II av-i -1y I -- C I~ir~-
i I ML)-t : m' i IIO DRPI Indicatons Lf2 j 1J...2 I- I II I lfl~
I -II11I m a w__ _ __ _ _ __ __I_
RM-71M"- I I QI4_&____ a M1 M - lt Ea ak-I~ioilyMoiid~~ F1 I I NM I &pRvisORy I acfl [
NtM Li
-if
plc jNuclear Instrumentation System Which of the following describes the effects of a short unintentional emergency in MANUAL. boration on the reactor at 75% power. Assume that control ro 1
Tave initially decreases caosingireactor~powerto decrease. Tave
.av thenderaeincreases toap to approximately mtl th the,i.ia nitial value.,.,,.
-,-:;' r'
- jTave initially idcreases causing reactor powerto decrease.Tavethenincreases t approximately theinitalvalue. ; ., '...,;'ii-'.'r
~ Reactor power initiallycdecreases causingTave to decrease. Reactoh-rpower then increases to approximately theinta I Reactor power initially decreasestcausingTaveto increase. Reactor power then.-increases to approximately the initial value.
I A~rj j a'nLevelf
- "~~~ FB ~j C atiLee Application 1.1.
acihtM adod ~ ae - - F'
-- - -- -r$.s 7/19/0 M 1015000A107 ~ f ~ ~ ~3.41 Scoi Fs-s' I j g .F777!.SR~o~p
-:~ij 113
,, Ability to predict and/or monitor chang ssociated with operating the Nuclear Instrumentation System controlst.,.
including: in e ter s s with operatin Changeron con ntra on j~i~i o'Bortio i isan dded poison, fewer neutrons available for absrption in the fuel; reactor power decreases.: The dera nrecr p.owerresui isn s adlowerdTave. As Tavepo eases.
(d reactivity is added decr causing reactor power to increase..(C.correct.respoonse.
, ,e r. = _ rd-e a 0-SNormal Ooeratin(a Procedir c .: .- r Inr ,e..
617
p lNuclear Instrumentation System The following conditions exist on Unit 2:
- A normal reactor startup is in progress
- Reactor power is steady at 1000 cps
- PR NI channel N-41 has failed low
- Operators have completed performing 2BwOA INST-1, "Nuclear Instrumentation Malfunction" for the failed PR channel Seconds later the control power fuses on PR channel N-43 both indicate blown Which of the following describes the next action to take:
Verify all rod at bottom lights LIT. .~.
1
¢ Manually reinsert all control bank rods - ,. f. ~ji:!' '..,'. .
Manually reinsert all control.'and shutdown banks'*, .:.< t ; :.4L.
. .i- ir > ;- , .. ..
j Complete the startup to <P-10 using only IR instruments ,- , ,V ----- -
ise ~' i'v~ %onivtil pliain Bra idwood
[~.. EAM,a g,l ~ 7/1 9/02 yse -vlui rte ji'i; Istrmntto System015 l. . I
. e: .Knowledge of the effect of a loss or malfunction x6n'the following will have on the Nuclear nstrumentation System: ';
Bistables and logic circuits .,.* .
- '.Ixp Jautomatic trpb~P4i jag active, even at <:indictae Ppoe.Sintuments are de-energized due to P-I pickup. (A):
is correct.
Candidates hay confuse 2 PRchannelIfailures wlth:P-1 0actuation and'blocking of SR:instrumients which make . f
d ... ig tes .JO UA,. . FLU
,,,'..~ riwo Big Noe _ F.'..>.~. ~*.>.............................................
- < II.J-rA ::
.LW'EdSS 3i -L I .. .
.1..
I I II . . ............ I-I "M I--- - I
' i Qi s -' New- -- -- sed bw-n ra nin 1-
.... I-0-......
2 IR F-=-=--j F-=--=--j I- _______j ME] M E=-=j I --- - 1 1:1 M 1-1 lMR~ [Z
Which ot I Non-Nuclear Instrumentation System I lBoard recorders. provi 1 LR-930 RWST Level . . . .,
Continmnt ressure.:
. 1PR,05t Ste m Generator Pres sure';" -t ;'$: ,'-G;,":': :<4a-'.'.:'1i<r es-7 n (ew: s'-
JfiFR-51 StamGenerator Steam Flow/Feed Flow-~.~~:
gj B~i~k~j~nie ~ Mmr i> riwo 7/19/021 O160LlA4c062 L A4202 u j271 e YS- I P :
K Non-Nuclear Instrumentation System :
azxa =
l016
. IW 06*.;-
- . Abiity tomanuallyoperateand/ormonitorintecontrol oqm
~i. *.
Recorders...
a ca at (, D hv n 1 PM 4J lfli..i vV,_.
1-J
j In-Core Temperature Monitor (=TM)
System The In-Core Temperature Monitoring System is utilized as part of the Power Distribution Monitoring System (PDMS). As such, the of Incore Thermocouples required to be OPERABLE is MINIMUM number (1_) with greater than or equal to (2) detector(s) per core quadrant.
7714 2 01 7000G222Z~ 22 ~ T~
u~j34 a ci sSj0ru 9 Rbru [~
In-Cre SstemT T mperture Monior 017 s .>.-. =,...I Knowldge.
of limiting c t fo .....
operations and safety limits.
AI23e7res 7
,r 4,. wih rete tanore2a per quadrant. (.C).Correct
. Req jTech $ uirements
- 2 Ii: 'PnalopRM . ii.tw ;t>;.tf7- ,,t Iwi6- t~i sor*Y.71t*
-xt;e
._ - .aii -ir
- 7 a 53 :NR Lieq =;
'w 7 _ _ . : 7 7 _ _ _ Z i i _ _ _ _ _ _ _ i ' _ _ _ i _ _ _ _ _ _ _ _ , 7 S _ _JI
..i.'7 .... o........New........
- ,I . .Used...........Trainingl .
.01
s In-Core Temperature Monitor (ITM) System __ . .
A Loss of Coolant Accident has occurred, core exit thermocouple (CETC) temperatures are reading 690.F and increasing.
Which of the following describes the expected response of the CETCs as the Reactor Coolant System and core exit temperatures continue to increase. (Assume no core cooling is present)
- The CETC's will indicate..
. Jlower than actual temperature above 700°F, and will stop indicating altogether as temperatures exceed 1200°F.
. .. ; l accuratelyupto 1800°F, and canbe used for endingpurposes up to23002F.
.,0 i -0' - r h t atu tpurbe0 a, ndca b . ..
actual temperature above 700F, r andthan cannot be ehigheupon foraccurate indication relied aov.'.
S accurately up to 1200°F, and will fail completely above 1800-F. . i 7 C'-
n r; mbj J C n vf Application Jj FcBr~aidwood - ,amaf:47/19/021 01700K102 K1.02
[ . 3.31 a 351 l:SYS [ 1. [
v.,iton In-Core Temperature Monitor System , l -X. ,:-.: . I 017
¢ta.t;e Knowledge of the physical connections and/or cause-effect relationships between - ore rature Mont Sytem and the following: .osisbtenI-oeTmeaueM o sma
.ngeis200-1800°F. Theyareexpected to indicate'up to 23000F with reducedaccuracy.(B)isd-rrect.i, J, !
usblerane Teyrc c. (B)isorc.vl.f':.. -_i1 I
12A and 2C RCFCs start immediately after the Si signal, 2B and 2D RCFCs start 20 seconds later J~ e j o~ieeF AplctinBraidwood ~ ~ - 71/2 IBM~~wa1 lA . E l 1 S 1 1 3 1 0 71022 S m Abilitytomanually operate and/or monitor in thecontrol room:
CCS fans xp 0nat4 0Fans already running in slow speed will remain running - none are in this example. All ii s.
high speed breakers trip (2A & 2C). After a 20 second, time delay, slow speed breakers close on non-running fans. (A) correct response. .
-- BwdBig Notes j otimn ooling VP-3 j~
l -=Facif=1tft -- Iso~ -
rci
-I-11
Containment Cooling System (CCS)
The following conditions exist on Unit 1:
- A small RCS LOCA has occurred
- The crew is performing 1BwOA PRI-1, "Excessive Primary Plant Leakage"
- Containment pressure is 2.9 psig and increasing slowly
- RCFC IC low speed breaker auto SI closure relay failed and was declared inoperable
- Annunciator 1-3-E5, "RCFC LOCAL CONT", is LIT.
- RCFC iC is in LOCAL control at the RSDP and running in HIGH SPEED Which of the following describes the available operation of the iC RCFC in the current configuration if containment pressure continues to rise to 4.0 psig?
g Tliihigh speed breaker will automhatically trip, the low speedbreaker CAIRN ONLY be closed from the RSDP 20 sedonds late s -. .
T hig braewl O9atoaial d h I~p~bek ANOT'1be 6osd affetthe~the' R8D h a,.. or1PMO6J.' ,
.. .. 6d _,au6 :.Zi 11 .
. The high speed breakerwill autdmatically trip, the'eow'speed breake AN BE'r iur lydcisedfroni'ilP J20s866mndsater. t b The high speed breaker will NOT autriatically trip, the low speed breaker MUST BE closed locally affdr the high speed breaker is tripped. l I
i _ D IC I: _=AM . .. 7/19a 0l2G44.31 1.j 3 3S 1 13 } J-SY f3
. Containment Cooing Sysfm 022 Knowledge of annunciators alarms and indications,and,useof theresponse instructions.
RFLCAL CO T in~diate thattthehigh'speed brdakerfor the'RC'Cis inilclcnto.Lwspe rakroeai~ s unaffected~by theLoc-al/,renote switch. (C) is'crr'c~t' (A) Incorrect'- l ispeed oration is n'ota'vailable at the RSDP. (B} 'K ; r1 II~i RaA;pC.A IIG-G -Is vX.- .. atC
, n
. ... M I --- In__ <
I I __
.. I.. I
T1S18809A RHR Cold Leg Injection Isolation Valve S1881.A A Sump solation Valve
-Containment c~1CS007A IACS PumpDischargelIsolation Valve I CS001A lA CSPump RWST Suction Isolation Valve mee , vee j'.g vemrBraidw-ood E~mae~7/190 N F5~2 200K10 J K1.01 FQ~~~14.21 R>4alu 213 SY-S feto: P F 21
~farri f~jContainment Spray System - ~tmentFKnowledge of the Physical connections and/or cause-effect relafionships between Containmet Spray
'S System and the following:
R 9lK ECCS M1IE"i1 09 Interlocks for openingl1CS009A are: lSl8811A Open,I1RH8701A Closed, CS toOpen.
(B)is only correct answer. (A.C,D) valvsI
-m'are not in interlock circuitry and are incorrect-CSLP 5ll-XL-CS-01 116IS.CS1-08-A IBwdBigNotes K i Containment Spray I CS-i n j ow Ineroksfr pnig S09ea e 1 S88 1 pe,1 H70R Cosd C o pn B)isolycrrc aser ( D eale e
~e~To Facilityrce;,~
ExamBankorially Modified e ran rga Lj
-70
Conta inmentSpray System (CCS)
The following conditions exist on Unit 1:
- A reactor trip and safety injection have occurred due to a large break RCS LOCA
- All ECCS equipment functioned normally upon receipt of the Si signal
- Five (5) minutes after the Si was received, Containment Spray (CS) actuation setpoint was exceeded
- Train B of CS started as designed
- Train A of CS did NOT auto start and could NOT be manually started from the Control Room Which of the following annunciators, if received JUST PRIOR to reaching the CS actuation setpoint, would result in the above status for CS?
f 1-21-A4 BUS 133X1133YFD BRKRTRIP 'i:.' . .. ;,i 1-22-A4: BUS 134Y134X FO: BRKR TRIP , - t, .:.:.;. .y.gi 11I I - - I..., 1. ". . I - I . " .. .- i; I !-i I
.. I ;- ;--*eIt ln: - . Ii !4 "
U I -- : -- - - - -, - _- I. "' '. k 1. . .1 1!:' :. n!
' . -'. n- -.. : I.:;,-! " '%
.`,): . : .' I
'; :;' ,, -1a !
, - . ,, : :: -,.! -'i '" 17" ' .' " . ... r . ' '
, ;1z ;'- , A'e4A'"---- -- . .I" . '-;'7, .":
.;_-4"
' .`ATI '_" I'.'e I%11.1" I
1
- M 1-22-A10 BUS 132XFDBRKRTRIP : -
I wii4 r_[ rp-M-0 rp-zl ____ - F-M-7MMUM r-_
I Joio2TmTK2o2 m l=_ M I rarrood .: 7/19/021 yste yol~ion. t ~< ;
Containment L~T~*
Spray System Z.J ~
~.~ " ~~' _ 9¶,,~JtI 2
% f2~
~~Knowledge F of bus power supplies to the following: IF , .*. ~
MOVs 4 .
.~ ~ Inte,,rlocks-,: to auto start 1CS019 must auto op'en. Pow.rsupply is 31X1. For CS to be-manually started, 1CS007A must'be;closed. '?
Powersupplyis 131X5. To start in recirc, 1S18811Arnust be open but we are not in recirc yet (C) is only correct answer. (D) ' -
CS P te c en. f.
IjC5lessonPlan Ii..i--
tl1-XL-C^S-01' ;'i l i l flu -' l21 G~R.C11' '1 -
1 -I:.--- - - - _
Bwd Big Notes ' " l j IJ Containment Spray . . .- I -
. .~
I I F. ____..l __.___......
_JC 'l I...... . IF__.
I Ir = ___JE-__JE=
I '
. _ -i 11 z _ -- ~ 11IIII~
i I- h~d14j~qiru rrExaminaltionT I
I ffil'f" OwtU WE
" -A N-w I ___ H ___ __J ffikM 11 II E=f_ I Facii NREL
Q M eHydrogen Recombiner and Purge Control System (HRPS)
Under normal plant conditions, opening the Post LOCA Hydrogen Monitoring outside suction and discharge isloation valves, 1PS228A/B -
1PS23OAIB, is accomplished in the Main Control Room from (1) , and indication of containment hydrogen concentration will normally be displayed on the (2) scale.
_(1) /9
- jJ 9o0 IjA4~ 03 [-- S 3 o!o 1 HydrogenRecombiner .n. and Purge Control System K 8"' 1028
'Ability to manually opeate-andor monitor in th control room:= .
Location and operation of hydrogen sampling and analysis of containment atmosphere including alarms and indications
.; . ,xpa~,laFt o1 (D) Correct - iControls and inrdications for 1PS228,,229,'and 23OA&Bvalves are all located on the.Containmen't lsolatidrPanei:
i,IPM1 1J.PS343 and PS344 are normally selected toindicate on the LO range (B&C) Incorrect - controls arenot locatedon.l r~~~~~ IVtJ , Ian.A ~c.4jj-,. IVIt
.. .ally ecct,etv Ajl tlOy iv i L%i
- I. ; *.s. .: ;iz Nnma n-ninPin:i~e'::
-':L,:-
muiilVlllyruurs~- Iljrn
. I::.: -: I l_.-:*
--. I-: I. .:--.
wvP-9 ____ : 1:.1::--
- i ............-.
I 1s.;;i2i8E2 I . I -- IB. I- I...........
_.__ I ............. 1_
l,:li-. __._1 f--- j---I'-I I~f -- I..... -i F--]--- ==
I z- " ~
U" II 4________-_____________________________________________I____ - I'
- I FN2 ""M -1 I nma " mm [:1 I
peerIII I . I ~ P rson I I 1 I -t : I SI
t rad monitor IRE-AROll is operable and continue the release prehension 02900iA204 1 J 2o 3 i ak I '
S iWyou0~. tL Containment Purge System sa ~ ~ ri.Li
- . j029 eSD ej tAbility to (a) predict the impacts of the following on the Containment Purge System and (b) based on those predictions, use .
procedures to correct, control, or mitigate the consequences of those abnormal operation:
.Health physics sampling of containment atmosphere -
R Isolating this pathway
.-, Run ,. ,.^ :,~~~~~~~~~~~~~ii; 1RE-PRO0l renders the noble gas activity monitor inoperable. RETS 2.2.la requires immediate suspension of purging via if this occurs. (A) Correct. (B) incorrect - this is an allowable option only for the iodine and particulates functions of I. W M
jContainment Purge System (CPS)
The following conditions exist on Unit 1:
- Operators are currently performing steps in Post LOCA Cooldown and Depressurization, 1BwEP ES-1.2
- Containment Mini-Flow Purge Exhaust and Post LOCA Purge Exhaust fans are aligned and running Which common mode failure will result in BOTH the Mini-Flow Purge Exhaust Fan AND the Post LOCA Purge Exhaust Fan tripping?
Closure of 1VQ005A, Mini-Flow Purge Exhaust Inside Cnmt Isolation.Valve
~. .~>
. j Closure of 1VQ005B, Mini-Flow Purge Exhaust Outside Cnr lation Valve ' .'-
. i Manual actuation of deluge inthe Post LOCA Purge Filter Unit. .. :.:' ri':: ", .r' ' .,' . .
EfHigh alarm on the Cnmt Purge rad monitor, 1RE-PRO01 . .. . .
n5,i ~ai I i[ 1joW e FvAp~plication I pill' Bra dwod .
7/1 9/
0 2900G449 2.4.49 a 4.01 d~ 5 D' c f I S I -;
io~t -C n Purge System029 Ability to: perform without reference to procedures those actions that require immediate ope!ation gf system components and
- p i b/ t(A)0Correct-VQ05A is interlocked withBOTHlthe. mini-flow purge exha.ustfan
- and-the post Incorrect - VQ005B interlocked with ONLY the rnini-flow purge efxhaust fan. (C incorrect-: this LOCA purge exhaust~fan. (B) c-s . ii action willl auto Cose:VQ00b which * - -'
r I.
BwlntPreVP-2 oe .. '-,,._ m.
- I115 _ F Ji,: ...
I==-== == === I ,g _
4012
L
[I ~W?106Sp OR j Spent Fuel Pool Cooling System (SFPCS)
If a leak develops on the discharge of the Spent Fuel Pool Heat Exchanger while the cooling loop is in operation, the Spent MAXIMUM of (1) before the FC Pump loses suction. Using BwOP FC-1 1, makeup water will be added back Fuel Pool will lose a the (2)__ to the Spent Fuel Pool via
()(2) 4Jfeet -Refueling'Water Storage Tank (RWST)
.~
IJ4 inches Volume ontrolTan'k (VCT) .. - ,~
,- 4 inches - Refuelihg Water Storage Tank (RWST) : i . ;.. ~i- ,
4 feet, .. Volume Control Tank (VCT)
.: , ... '.'.i;*
~nsJ ~' j ~Appicaton am ye ~ 't ~ Bradwoo E am~ae:~7/191021j FO33000A2~03 I A203 RO Iu F3.11 ~ ~ r3.5 FS Y r-Sj &riI.V2
.. ~s ... /Eigrol ,~o , ,pentFuelPol .Cooling System , ' '. ' ' 033
., ,,itto (a) predict the impacts of the flowing on the SpentFuel Pool Cooling System and (b) based on jthose predctons,-
- poedures to correct, control, or mitigate the cons66equenes of those abnormal operation:I. us sent e or loss of water level.
Eaa (A) Correct-the pump'gUbtion has stops.4 feetbelowinormal water level.
TheRWST-1sanitvailable source of makeup water ~B I rcorrect - the 4,inches relites to the anti-siphoh, hole orh~e F`P cooling discharge to thie pool The7h VCT also is not an viabem Spent FuelPoolLevel Adjustment Proc FC-. 1: f Spent FuelPoolCoolingLessonPlan figs,3 u- 56
,-I I .. _ _..I
JControls for the new fuel elevator will only travel in one direction
-'.thereis no upward motion available ;,;~$
- . Upward
. motion of the new fuel elevatorris stopped if surfaceradiation levels approach 100 mrhr <! :.'.f . i,,.*.-.i.;-
- JAn:upward motion interlock prevents lifting any loads greater than 12q0 bs with the new fuel elevator 2 s - , .a;-,,
A slack cable interiock prevents raising the much lighter spent fuel assembly via the new fuel elevator. .
W FE34000A302 I A3.02 Rf5:1 a 3 3FSY 32 ae~iiit~Abiity tomonitor automatic operations of the FuelHandling Equipment System including:.
T6Correct(C)- the upward motion interlock is to prevent raising spent fuel~out of the SFP, maintainingothe required depth of water ;
for shielding concerns&A):Incorrect- thewnew fuel elevatordoesbtravel upwardi (B)drncorrect -there is no rad interlock, l.,
lee 1t
~a~" --- Ld. i-m..
eskionSAUrc 7M OhrFclt1 i n FulHnln Pll-~~L0 d io o 'Sgificantly I . ._Modified /~iui9.i~~P 1 ._J _ ........ i £L... i [.J_.___.__.
r--ton ou& 01 Prii0sln R 01 S ffl O If1i~'
Oee:
0!;
Fuel Handling Equipment System (FHES) .. I Unit i is in Mode 6 and has commenced core off-load. The following conditions exist:
- 1A FHB Exhaust Filter Plenum is aligned and in service
- Containment mini-purge system is in service
- Fuel Handling Building Radiation Monitor, ORE-AR055, is OOS
- Fuel Handling Building Radiation Monitor, ORE-AR056, alarm circuitry has just failed. [MD is troubleshooting.
Which of the following describes the required ACTION, if any, to be taken in order to allow core off-load to continue?
. No ACTION is required, fuel movements may continue uninterrupted
~ Core off-load can:NOT beconducted until at least oneof-17&HB radonitors isrepaired Fuel movemeint may continiuefor up to 7.days while res~toring one (1) FHB rad monitor~to operable status~provided lBFH Plenum is aligned in the Emergency Operating Mode xhusilte v
- - - -l*'
Fuel movement may be conducted indefinitely provided an appropriate portable monitor is provided and the 1A FHB Exhaust Filter Plenum is aligned in the Emergency Operating Mode 220340001602 lK6.02 @2.61 O R@ 33 ' f ' ;
Fuel Handling Equipment System . ; ;*. ...
,03 tae Zn~Kowledge of theeffect of alossor malfunction onthe-folwn ilh onte FuelHandlingEquipment System: ~i IRadiation monitornngstysferns .
- i~ '- * . -
Wof ina!
p (A)incorrect - per TRM 3.3.0 with 2 channels inopm usf place 1 FHB Vent in emergency mode and'provide a fuel movemrents. (B) Incorrect- rnaymoefuel porta tle monitor or.stop with vent alignmentand protable mronitor:(C) Incorrect
-need to also place portable .
V. . C ; 1 2 5== ate aci e, G: Nu= Refe:== .. ...- _ m 7 iR4.
A turbine overspeed condition following a generator trip at power xrn ve Fsj - i~ ~ e ~ Cmrhnin at~ raidwood7//0 0O39000G2~25 1F2.2.25 a 2.51 SO3 71 ei s -ys I Of2p K3 G2U~
I Kowldgeof ases in technical specifications for limiting conditions for operations and safety limits.
iifp (A,B,C) Incorrect - these are not exceptions - they are basis statements for MVSIV operability. (D) Correct - Credit is not taken for MSIV operability to protect against turbine overspeed.
TehSe ai-MIsB 3.7.2 7 o l 81-
uestopi~ Main............. .................
Turbine .......
Generator ........ (MT/G) System The following conditions exist on Unit 1:
- Reactor power is 80%, steady state
- All systems are in automatic control
- One Main Steam Dump valve, 1MS004A, fails 100% open due to a valve positioner failure.
What is the expected response of the plant due to the steam dump valve failure AND what action can the operator take from the control room to stop the excess steam flow?
Turbine load Will decredse by approx. 3% AND reactor power will remain constant. he operator can stop dumping excess steamr either Bypass Interlock Switch to OFF/RESET. by taking:
Turbine load Will remain relatively constant.AND reactor poser will increaseby approx. 3%. The operator can stop taking the Steam Dump Mode Selector Switch to STEAM PRESSUR dumping excess steam by'.%
Turbine load will decrease byapprox. 3%/ AND-reactoripower will remnain -constant.The operator-can stop dumping, excess'steam the Steam Dump Mode Selector Switch to STEAM PRESSURE. by tkig
- ' ' ' '- -' >A' ' - ' i- Ad -
Turbine load will main relatively constant AND reactor power will; increase by approx. 3%. The operator can stop dumping taking either Bypass Interlock Switch to OFF/RESET.
gj-£ Ider H e pliainI1lf~ BadodE~ae 7/19/021 Tubn Geeato System8 ' :. l; E d .T' lE3' ID _ 1~045 -S..$
- -en' 'Ability
--AC to (a) predictthe procedures to correit impacts of thefollowingontheMainTurbineGeneratorSystem and(b)based control, or mitigate the consequences of those abnormal operation: D. onthose predictions,use
- jSemdumps are pot cycling properlyat lo~wload,or stickopen,it hig ieload(isolate n use atmosphericreliefs whenrleesry
...... . I , (A) Incorrect- turbine load will remain relatively constant with IMP IN (normal at 100%). Reactor
. j ., g ;.'.*><< increased steam flow. (B) Incorrect - selecting Steam Pressure Mode will not close the steam power will then increase'due to dumps if the-failure is in-the valve
j : increase
- ddue to increased stear low and either Steam Dump B dumpos (train A&13) , Byps tr swilli FFRESETwillose
.. , , '<:i<.-....c.il loeal t
~BdBgNotes IVIS-4 .Main Steam 6.i£.......i....
II= 3 E- --- J E
_=_1
== I 1
eo I< I-Main Turbine Generator (MT/G) System Once every 31 days, each of the 12 extraction steam nonreturn check valves are tested by observing freedom of movement of the weight arms on each valve. This testing is performed to ensure:
ISteam line breaks whicl-roccur outside the Auxiliary Building are positively isolated
~,.i-occuroodingdoesn in feedwater hedaters;limiting the'ab'lit to6restart fol oing a reactr trip . .,
. -,f ,-, . : .r ., _
, -Excessiveaversfe'ed of.theturbinegdoes notoccurfollowing a turbine.generator.trip>. rf j1 - i! X'.i 4':i;4 :
] Overpressurization of the main condenser does not occur if feedwater:heater levels increase too high
- --. Memor _ _ Braidwod 902 045002OG225 j.22 F.2.51 IM M 3.7j FScPYS rop 3 11 3R ys ioMainTurbine enerato .. .. System.'.' . S _ -. .
" Knowledge ofbasesin technicalspecifications for limiting conditions for operations andsafetylimits' : . -. -,'rjs-m Nonreiumn hkvaes1)r aroteturbir ovbrsl~ee~d r~Ofein ircuitry and thus protect thei turbine, frmoesed.
following a normal turbine trip, specifically fromnsteam-flashing in feedwater. heaters from reentering the MT.. (C) Is correct (A,B,D) vL...~i.Ls-corre
_ ,.',. ir= ~tacuO~ r pa I..........f _ iO QI tI US>~ _Jc~pv pcrutI. :'-.
-.'- . 11 - - - .. '.r. ,__ -__Aq
Main Feedwater (MFW) System The following conditions exist on Unit 1:
- Reactor is at 100% power, steady state
- All control systems are in automatic
- Instrument Air is lost to one feedwater regulating valve, 1FRV-51 0 If no action is taken in response to the FRV, which of the following describes the response of the plant AND followup action required by the MCB operator?
7 j'TURBINE TRIP ABOVE P-8" trips the reactor. All.Main-Feedwater Pumps AUTOMATICALLY trip.Operator must sihply VERIFY Feedwater AUTOMATICALLY . occurs. ......
. TURBINETRIP ABOVE P-8' trips close all Feedwater Isolation Valves.the reactor. All Main Feedwater ~.4~,'
Pumps must be.MANUALLY;tripped in EP-O. Operator must MANUALLY,:, .
!.S/GIALEVELLO-2"tripsthe 'reactor. Al_MainFeedwaterPumpsAUTOMATICALLYtrip Operatormust MANUALLY close allFedwater':~-ki.i. {0>s ..,,
Isolation Valves. - * . , , .
S"/G:lA LEVEL LO-2" trips the reactor. All Main Feedwater Pumps must be MANUALLY tripped in EP-O.
Operator must simply VERIFYs. t e. ,t !,f Feedwater Isolation AUTOMATICALLY occurs. L . '
nsr g j R~ e 1 R J CgiieL t Apiainc raidwood Exmbt 0559000A2'12 WA2.12 .'. a . 1.1*1 . ae f J sj l r1 S 17
- . s t -. Main Feedwater System. . .059
- .t 3 'a . Ability to (a) predict the impacts of the following .on the Main -Feedwater System and (b) based to correct, control, or mitigate the consequences of-those abnormal operation: on those predictions, use procedures
- . ;, i '; .,
Failure-of feedwater r.equlatinc'ivalves.
~ fio~ of(D) Correct'- FIRVs fail closed on loss'of air. SG level wjIl decrease to thelo-2 rx 4rip setpoint.
Lo-.2 level does not trip the;MFPs.~P-4:
initiates FW isolation: (A&B jincorrect - FRV fails cIosed levels decrease.(C)0!rcorrect Lo-2level does not trip the MFPs. P-4 '.
I t irI L IV ras uu .. , .. .' -V. _ -. _" > * . . L .. '
aillity umeK' ~
Annunciator response proc ' ' '.4IIBwAR 111-A8 ,
Bwd Big Notes - FW-1 -- l i
-I -
Ne ~et io-~o~id 4_ 1.............
_ n'~ n . ,.........
Ij
._.nT~e Cr .n _._ _ . . _.__ . ____4 . ... __ . _. =.' -:.,'
=:I= I GC 11 l[Clt o
... N D
MainFeedwater(MFW)System ty. _ . . . . . .. . .. .. =
At 50% power on both units, the Steam Generator programmed level for each unit is:
Unit 1 Unit 2 33.0%
2 ' 36.3% J :
",.'.: ::: - 0.%.
, 600% ..... 6 1 81.0% . .. ;.:! 80.8% .9i:'..-';6'l'..'
- -' t\.' -.-:s;r n TJ 6pte eMeory jac tBrdwd I nOae:7/19/0
- . . ?OA=l IA~A~ k*3.11 FNMI I -
- ~n Mai Fe dw te Syste -- .---
- ~ -: M~t'Ability to monitor automatic operations of
. . _ _ ._ . _. _ 1059 the Main Feedwater System incuding-Programmredl levels of the S/GI E~a~tj f-Porm areU 1a60/Uat 6 t 634fo h olrng o10 oe C) is onycortrepns . -
,.,.p;
- 9. -
essonPlan 0 L
ElI Maim~ri Mpqiini CoiiiiW7
~wso~tt s&
rvi or7]
.7Im q,
The following conditions exist on Unit 1:
I Pi
- Preparations are underway to perform a reactor startup per 1BwGP 100-2, "Reactor Startup"
- Steam Generator levels are being maintained utilizing tempering line flow via 1FW034A-D and 1FW035A-D
- When testing the reactor trip breakers per step 13 of 1BwGP 100-2, 1FW035D did not automatically close
- 1FW035D was manually closed by the NSO and the condition reported to the Unit Supervisor
- All other feedwater valves responded as designed Given the above failure, per Tech Specs the reactor startup will.
.BeALLOWED to cntinue with BQT FW035D and controlling tempering line flow. Sinceat least one vlve in the line I it'srequired position, theafetyfn f6tr is e ill not behalf-6 ileng. d B At tinue,however, IFW0350 must be arable ad closed wi power removed from its valve actuator. The Unit is en allowed to operate in enintely in i'di; -.. f. ,! -' ,oi NOT be allowed to continue because BOTH valves in this line must be OPERABLE to ensure positive isolation and prevent leakage inthe event of ani'acibideint. containment out- 7 NOT be allowed to continue. In addition to 1FW035D being inoperable, the Aux Relay function of feedwater isolation must inoperable which precludes any future Mode changes until be declared repaired. '
. 177 -1S a o, N 0.Cmrhnini Brai7wood021 Fo500`3 ~ 2.1.33 a3.~
_1l24 .
l~j.o 4.01 SS FS~i Orji by i R G -U~
1 Ed7 Main Feedwater=Std I'059 Ability to recognize indications for system operating parameters which are entry-level conditions for technical specificationsI Ipn (AlCorrc-n re o es hesafet'fUnctiori 1FW34D must be closed with power removed - single filure. (B) Co ret 3.6.3 does not require mode reducti-o'if the required action ~is co:mpleted within fhe time a llowed.
(C) Incorrect,-one valve may be
,cc ;n,i-,e- -,v. ic tuou I piVVc
-Vctc I--ICIIIUVUic.cv.I~~r CUiiat,uui i' uiie, U-,e,-, rIJ L I.L JVVCIi gU .JV Vl~it i relay functionhas
.theaux been affected..,
OAflI 20 ,-tUjVt.iui-oIyoa/ IIIIA.IICI.
not 7. _a , , ._=
1.A Webb,,,,
A dw WS.F I I I I
qt-
jAuxiliary/ Emergency Feedwater (AFW) System =
-The FIRST signal to automatically start both Aux Feedwater Pumps on each respective unit is received as steam generator level passes from operating level through (1) normal
% on Unit 1 and (2) % on Unit 2.
_ i 88.0 80.8 An, p .. ntv onFvLve eoy cbt riwo 7/19/02j 06110001i1i0 , K1.01 1in J4l l JJ JY .j yse v to *; Auxiliary/Emergency
. ffi l 23.07; Feedwater System 7- 0; i
-11;,." ~~~~~~~~~-? 13, ; ? ~K: ,,: :if!-l 1061, J et Knowledge of the physical connections and/or cause-effect relationships betweenA AxiliarI Emergency FeedwterSystemand theP pLO-2 SG Levels (2/4)will altomatically start both AFWpump on each respective unit. Lo-2 U-;
=18.0%, U-2=363% -; t0 q3
@Ue m lAuxiliary/EmergencyFeedwater(AFW)System The following conditions exist on Unit 1:
- A reactor trip / turbine trip has occurred
- The crew has transitioned out of 1BwEP-0 to 1BwEP ES-0.1, "Reactor Trip Response"
- Step 2 is being performed, "Maintain RCS Temperature Control"
- - All steam generator pressures are at 1050 psig and decreasing slowly
- All steam generator narrow range levels are <10%
- RCS temperature is 553°F and decreasing slowly Which of the following actions is required to control and minimize the cooldown of the RCS?
=
Maintain maxirum AFWflow until steam generator NR levels are
>25%, then decrease total AFW flow to 500 gpm .. - c -'2 *'->.
Decrease total AFW oIw,maintaining >500ogpm until SG NR~levels are >10%, then throttle as needed . to controlcooldown fImmdiaelydeceastotl
+ .
,d ecreass A.Wlflow, total
~ AW fotoaproximately 25 gpm per SG ':.
Id
. Stop the AFW pumps, if operating, and isolate them from the steam generators
. , <:..2 s i nsi~w~.v i b .0g vl Apicto ~~ ON j~anae 7/1Baidod
- 061000K:104;. K.04 - ei Syt y juiiayI mrency Feedwater System
. '.:s..j06 -------------
l Knowledge of the physical connections and/or cause-effect relationships between Auxiiary em following:
,iR _
!. . . ~ ',.: - xi,;.Emergency Feedwater System and the__
pr1-BwEP ES-0.1, step 2RNO. Maintain total AFWfl6w. >500 gpmuntiil NRI'evels-are >1a0/in at~leastone.SG The~n~no further restrictions are place bn AFW flow rates. (B) Correct (A) Incorrect-25%Helevel & 500 gpm are too high. (C)lncorrect.- cannot rTrip Resi & Basi hi1BtEPESO
_1 -. >J .~ 7s ep2RNO i 3.
100, FacilityExamBank 1 ~ orialyModified it's~ ~ .. agrurg[~
E==IJ I
F== =ac I I - - ..S......
icty sort-I lj 0L
lA.C. Electrical Distribution System A loss of all AC power has occurred on Unit 1. The crew is performing 1BwCA-O.O, "Loss of All AC Power", and is preparing to cross-tie to Unit 2 using a limited crosstie to ESF Bus 241. DG 2A is supplying the bus. You have been assigned to monitor ESF Bus amperage as loads are restored on Bus 141.
Whc. ftefloiglaswl K da h ags unn meae II~AMCR Chiller ..
. _I
~J1ACVPump .. .,
[
0600A0 T-rEA3.01 . ) e. Hi~ [T~J Am l~~A.C.Electrical RO *~[V 7 [ SOGop 7 Distributjon I06 ?
S :Ability to monitorautomaticoperationsofthe A.C. Electrical Distribution includihg:
L. .i Vital ac bus irnoeia~ge. :.i':<. .. .
(A norc hle raw .47 amps (B) Incorrect rC.C draws 14 ams(oseeLtrOsalodi d s amps(D)'Incorrect-lA CVdrawsl63 amps- . .) (C) Corrc AS i -et. .,
1:jA Condensate Pump 1A Heater Drain Pump Fl62000K2301 K2.01 m a j B3 1 R 1 jMS L1 Gte1i1 iE D 0 62
'State eai~ Knowledge of bus power supplies to the following: .
- j ~Major system loads* .
p an f~b ~ (A) Correct - powered from 156 (B) Incorrect - startup FWP from 159 (C) Incorrect - 1A CD/CB from 159 (D) Incorrect -lA HDP from 157 .
ACDistribution LP 1-AP-XL-01293 1
I 7EE=I =E Ii - m E_ i m =::31m
_ ._all _
j' ', TripsOPENdueto loss of power to the SHUNT coil.
[ .Trips OPEN due to loss of power to the UNDE c js NOT capable of tripping on a SHUNT trip is OTapale f tripping on a UNDERVOLTAGE trip
.er. 'a ever ~ .~igntt~ve ~ye<
F06 Application ~ riwo iae ~ 7190 O63000IK2 01 11( ;01 3 13 IR rO P.1 13~f~
em6 _s f DC lcrclDsrbto 6 1oul
$~a ern Knpowledge of bus power supplies to thefollowing: -. : J IMajor dc loads
!.:j A.nrrect because the sunt cl is normally de-energized.
B. & D.incorrect beause the undervoltage coil is supplied with 48v
~ ElectricalPrints 2E-430-RD6 SolidStateProtectionSstem jj P j RX 9,17
_ _ Fat Em B I'
ip R n 2 000BwdNRC. 1998Calloway NRCExam I M6Sup6&6iI]
El-
l Emergency Diesel GeneratED Syst . -
When synchronizing an Emergency Diesel Generator to an energized ESF bus, immediately after closing the generator output breaker, load the EDG to 500KW by going to (1) on the Diesel Generator (2)
(1) (2)
,1 Raise . Govemor Adjust Control ...... ,,,..;.. . . e:u...-.
JRaise ..... oltage Adjust Control .
Vl, ; , :.1E . ' cr'.
! fLower GovemorAdjustControl .! I . r ,., . .. '- .. "A4.'-
.... t :.-'
FLower Voltage Adjust Control l I
~ Le~'j a gn~i v eve, Appli ati onI ~ il ~ ~ > r aidwo od ,x i / 90 640OA 1 l A1. 08. ]4J E yG 1--- E: 2i
- iy^~~1=
064.-
lAbility to predict and/or with operating the Emergency Diesel Generatorscontrols inc uding: m c s. a naini~ng minimnum load on EDJYG (to prevent r verse power)
~plfj 6 f~ . B)iscorec.
_, A) Incorrect - lowering DG Speed~ will decrease load approaching the reverse power
,t trip setpoint (C&D) are: i incrret
-adjstng e vtta~cotrii~ wilt ndt affC~tbo~ loading J KA+:-.
==sl~ e~trtrtp
. . = j:= PD: 1 _,' : -- 13 23 - _ __ _ -j fNc
OMi [Liquid Radwaste System (LRS)
What TWO conditions will INDEPENDENTLY cause automatic closure of Liquid Radwaste Release Tank Discharge Key Locked Valve W5? I Low circulating water blowdown flow and high radiation sensed in the CWblowdown flow , ... - . . .
,, circulating water blowdown flow anid high radiation sensed in the release header:
. o ees edrfo n hg.-.
jHigh rees edrfo n ihradiation sensed in the release;header. O
.Hih release header flow and high radiation sensed in the Cblowdown flow
$ [bj kaiL j j tee Mmr aityBrdwoEamte7/19/02 03.81 [.4.04 R17-I 3.81 RO ae [? f RO
[3 r S[ t dw seSsemS~4 al/ iud R - Fo. 1068 t Aility to manuafly operate and/or monitor in the control room, Automatic isolation
-. =-
P26T1 (B)is only correct combination pr
[Liquidradwaste, rees or I d e 77T7 [wF8W 2 Iee _L33j UZJ dtralyMdfed l anigr~n Eln CZ. 1999 Bwd NRC C~I~1=p ~.RvesC qc
E }K5.03 I u !.1 W I Ii1 in E~ jSY I n [j <o t
I= FLiquid Radwaste Systemrn_____
1068 L t~n~Knwlege f teoperationalimplicationsof thefollowing
-concepts as they apply to theLiquidRadwasteSystem:-,
Units of radiaon, dose, and dose rate - -
~>~
E~pa js~mre in6m x 20mmn 50mremif donein 20minutes.Savingsof 75-50=25 mrem (B)Correct
- e @sif Iy ' c
_ vo Phbi- Am ~1o> PgNd Lfto%
Wohs as Disposal System (WGDS)
Which of the following REDUCES the possibility of an unintentional radioactive release to the atmosphere relief valve lifting?I from a Waste Gas Decay Tank (WGDT)
OGWO14, Waste Gas Discharge valve, will close automatically on detected ghradiaon in thedischarge header, isolating the relief path .'
WGDT relief valves discharge directly to the vent header so that flow is directed from the online tank directly to the standby tank- '..;, .' ; '
The waste gas Compressor discharge pressure is automatically limited to less th'n the WGDT. relief valve pressure setpoint:,:,
l WGDT inlet valve closes automatically on high pressure isolating theon-line WGDT and directing flow to the standby WGDT n ak+/-iane B ~ gIIev Application' BadodE~~atr Vaify /90 071 000K305 FiK3. 3.21 l .21 S F_ .
- iWaste Gas Disposal System 071 a n.Knowledgeof theeffect that~alossor malfunctionof the WseGsD poaSytmwlhav'eon the following:~~L-IARM and PRM systems R -.
(UCrec -tnks are automratically-switchead o"n high pressure. (A) incoret all relifdscag dwntemo W 14sth reie at B'In i otislte -dichre s C icrrect-isoltedB, correct-dsh esto the plantvent(Cicret iedischargeicagof onsraofW the compressor has no afc n 1
= = = Y*II=f, II ----------- - ------------. _
117 Q<< ston6c*
= a- 1999 Bwd NRC
=:zI S~u1d' R 9B _
10(
'pic Area Radiation Monitoring (ARM) System Radiation levels in the Fuel Handling Building INCREASED causing BOTH Fuel Handling Incident radiation monitors (AR055 and AR056) to simultaneously reach their actuation setpoints.
Which of the following would AUTOMATICALLY occur due to this condition?
~]jB Train FHB Charcoal Booster Fan starts, then A Train FHB Charcoal:1Booster Fan starts.. - ~~'z>
B Train:FHB Charcoal Booster Fan will start ONLY if A Train has'.failed to start. ~~: ~
i,,.'i.-.F'i,
- 1 ATrain FHB Charcoal Booster Fan starts, then B Train :Charcoal Booster Fanstart:: .,f .:~i . t '.'" i .
ATrain FHB Charcoal Booster Fan will start ONLY if B Train has failed to start. .
EE l0A301 1 a@ 3 ton' a ' RG [
te Area Radiation Monitoring System.
, ,<__ t_ _1072 i.
, Abil~~~tytomonitor automaticoperations of theAR sseicuiiie-;;>J.i ',) X . ttifi'i
___________Changes in ventilation alignment '
xjilan j.,.n6§ A . Incorrect. Damperkinterlocks.prevent both trains from starting. BiTrain gets a start signal first. When it starts, irs dampers position, an interlock preventing the start of ATrain.:B. Incorrect - it is the reverse of D, the correctanswer.
C. Incorrect B gets the:
NreNoteAux Bldg VA.
OL~LO~I ,O _i . , U. _J~ I ,1 ,- -,, --.
z.4.
FHB lnterinr-k- F T F7 F T 11 107-%L
f1VQ003C, Post LOCA purge exhaust fan is started iE.E...
072 00K401 1 M M1 1 u; 1[1 3 f v
-se it I Area Radiation Monitoring System
- 1 ~ M 072OO
$aEe e l Knowledge of ARM system design feature(s) and or interlock(s) which provide for the following:
Containment ventilation isolation . . .
~~paf~
o,~~j A)Incorrect-this fan is started via AR055&56 skids, notARllJ or 121 (B)
Correct-this is part of the cnmt isolation signal generated. (CD norc teercien auto actuation signal from any rad monitor.
RM-11 annunciator response BwAR 4-1AR01JI B 1 2 Bwd Big Notes - Cnmt Purge IIVP-2 J FI i . [
Ma_ e rial eqi e 1 E1 ~~i ,o J _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
Ij E _. Fat E 103
oCirculating Water System The following conditions exist on Unit 1:
- A loss of all Circulating Water Pumps has occurred due to excessive grass collection in the intake bay.
- A reactor trip / turbine trip was manually initiated by the operators.
- During performance of 1BwEP-0, a SGTR occurred on the 1B steam generator
- The crew transitioned to and performed actions contained in 1BwEP-3, "Steam Generator Tube Rupture"
- The RCS cooldown and depressurization steps to equalize RCS and ruptured SG pressure have been completed
- SI was terminated and the crew is now investigating the appropriate post-SGTR cooldown method to'.e
- While investigan coodown options, RCS subcooling was lost V i , * . . .
- Additiorail ECCS pumps have been started and aligned, but subcooling is not recbve ing, : . . - - ;.". ;t i Which of the following procedures must be used to continue the post SGTR'cooldown and recovery actions from this point?
~JBE-3 ,SGTR"_ must be continued until conditions exist fretbihnRHsudon V -
cooling!. ~~'
' ' 1BwEP
!; ES-3.1, "Post-SGTR Cooldown Using .Backl1must be used quickly to recover Pzr level and subcooling
- 4 . BwEP ES-3.3, "Post-SGTR Cooldown Using Stea Dumps" mustbe used as this is the preferred method of recovery
. E. 1BwCA-3.1, Rif Reacetor.Coolant, Su c oeLRe t _ppppWthLos o optioriavailable and must be implemented.f Subcooled Recovery Desired" is the onlyopnavibladmuteipeetd I /,!
sr fI~~ale.v jjogive N7075000G406 126 alu F e<3.11 2~~ ti*wr-ro; 4.0j -c~
6J etfionf 4 BridooS I~-8 f -xm~ =2e:
- &N9iiP ~- 19-_i-Q~p~7.;.f '-
!stf . ;t... t ~ Circulating W ter System , . '.' ,. .. ; lf - 5';
0@r: ,. - W-_symptoiibased EOPmitgatio-h-strafegi'e~:'
Knowl~~~dge 'h-2~'?) 1/2~~'y-~'s:. -WX
~ ~ ~ "~
~ t~ A nbtc-ser8oto Appropriate Post~SGTR C6oldown~ Metoi a transition.
-ASherestwithout r s ~no continuation frm heea ih' - arnsbr (B) incorrect -thodAis itiontoCA31wt the LAST step in Esub.ThereCi each procedure's OAS and step 2 in each post-SGTR cooldown procedure.
step 38 &OAS 45 100 step 2&OAS I l ETS-3.lstep 2 1F2~1 1 1iC j F Y
Foam.
Water *A'~~*'.,
am~t ~j e Cgni~i~ eve Memory Fcft Badwo mte . 71/0 0860K1K 03
~volt~oi~
e~ ~ Fire Protection System
- E l
[ ap 21se 08 I108L61 t m>Knowledge of the physical connections and/or cause-effect relationships between r ecton Sys te AFW System , ..... =
'jii~iaion ,I(D) Correct. it-s'the only FP system in the IlB AFW~ Room -
Ia'M I efrnc um ~re C* Wt~Oi Pa ego.
jBwig Notes . . .* i .l .. * ,
- 'IF2
____ ___ __ jpRQM . -
UI ij j Lz-
/0<
IFire Protection System (FPS)
In which ONE of the following areas is water NOT used as the MPT/UAT/S.AT transformers..,.
- Upper Cable SpreadingRoom . .
EHydrogen Seal Oil Units 2.' .. *. i 4.,.
4 '_ ..
Ventillation Charcoal Filters - - . . 'i'; ..-
,.': j fn vw A
I FB-J , ~j
_~-m i I e Le e ~ M m ryI ~ i ra idw ood FR O Mb Mi e 7/1 91 2 .
- 0 j IK5.03 3I F3.41 6 u ys em u866 F .
' j. e'i t.Knowledgeof theoperationalimpliCationsoftle following-conceptsasth IEffectofwatersprayonelectricalcomponents ... KE.,,,I, Eicpia it nt uised where damad may result from spra
- 2.: J.. ,K.. 0 _~'
onequipmeint. (B) is correct answer., -,..-~
.... ;.,. .. ;, ,. **~ *\ .
BwB Notes-:. Fir *:ij . .;
CY,#
o GGeneric FIn accordance with the Pre-Job
' What are the Critical Steps in this task? ;c ; .. -
are t E L Situations?
,.1: -:.. _ I a WhataDefensesarewrelyingson?
,v...
sinCharge of the evolution. ? jV. ' :' .
Fd*.i~i ef~j viv eoyFcf raldwood j xnb~7/19/021 194001.G3f1 2.1.1 b ae [ al jPWj 3 ef fG 3 G : [
IFGENERi7 Knowledge of conduct of operations requirements.
(AIBC) Incorrect - all 3 are included on theprejobbriefchecklist;The fourth is."What is the. Worst Thing that car go wrong" (D) is M loon E U- ceuno
- ootwa o t.,.
re-Job,Hehtened...Briefings ... HU-AA-1211 Attachment.:.: 1 I, ===== I I '= -. .
__f_ __~
l'I=
_ 101
In; IGeneric IIn accordance with BwAP 320-1, "Shift Staffing", the MINIMUM shift staffing requirement to comply witTehScswhBOHuisapo include: r F4i) 3 1E i.-!-.'.' ' ,,::*i ': ; ..........................
.4 s
- ? w3
= l G~~E N E R l - 0> - i Knowledge of shitsafnreuemt.
. . t (C) Correct - per TS 3.5.2, a RP Tech shall be onsite when fuel is in the reactor. 3 NSOs are required per 50.54 Shift Staff. g', :fj i
- Tech ;t;~~~
Specs - ::- :-:::- L -i -' -~ -: .
1522.
=F!=2-23 .- 9
_ __J
.D.
0w t 2001 BwdNRC E
F2SRO
_~~~ .iit .. _Li_ 7 ___ ee' HN=~
Sp Ls Ice
j _....Gene
........ r i.. ...........-=::; -....... ..:::::::::-
.Generic .............-.............
In accordance with OP-AA-101-1 10, "Reactivity management Controls", which of the following NON-LICENSED individuals can manipulate the controls of the reactor if under the direct supervision of the licensed Reactor Operator?
An individual enrolled ina approved training program ... ...
A System Engineer during surveillance testing ;:..
E [Any Non-Licensed Operator during sureillance testing ... . . .. .
O.[Any individual directed to operate controls by the Shift Manager ...
1940GF9 21 Anyj indviuad~_._irectued_ A- j-J n. e _ _ WL.M.
to GopeatecnrljyteSifMage!
- JK..................
Pacih
_7/19/02 ,..Braid.wood
.'.-'i f-:l*".rou',':'
~iT 2.'
_JGENERI l 4
- Abilitytodirect personnel activities in ontrol room -'
9 ((A) Correct per the reference, must "ensure traineesrri nnipulating reactivity controls are anddirectly superie bya licensedindividual"(B,C,D) arethen Incorrect..-, enrolled.in an approved training program
, * .,.... 1l r* ,,.. .j& ~ 0$~~
KeactvityManage mentControl ., 1 . IPAA~...A ill
. B. tt 0.
JReactivity Management Control LP - BIiG : NA iNA i iF2 5 - -- -- -
- 16:11u K604F Facility Exam Bank _ yModifi
}
QO"4iat~utr~iwi"W~s,12000 Bwd NRC 1996 Bwd NRC 0 W = i0 0MW J I u aIt LI 1 0Li I K 1E IEl
_f-q
E14 Overheating the valve motor F~EJ teve VjC~j~eOUei Application ~ c~~ raidwood 7/19/02 194d001G12 lF2.1.32 a1 aG I.t r j3101 Ev w- -8 - ... =~~ii fGENERI Abiit toexlan and apply all system limits and precautions.
p ana~o
- Per BwOP AF-5, starting duties for MOV-AF013(AH) is a max of 5 times w/l a one minute period. Prevents
= motor from excessive starting currents. (B) is only correct overheating the valve answer.
F1~6 II
Chloride AND Fluoride Specific activity is minimized Fluoride AND Oxygen Specific activity is minimized ONLY Chloride Structural integrity of the RCS nitvFs-r eVe FMemory Baiwo aDt,<7102
- 2194001G .1344 1 t R 9 1 1 P W 1 L 1 i 31J l[bilit to maintain primary and secondary plant chemistry within allowable limits.
l(p Ana onof(A) Incorrect - 02 has no limit in mode 5. (B) Incorrect -Fluoride is within allowable limits (<150 ppb) (C) Incor bo 02 a lw ... _ Fluoride are within limits (D) Correct - Chloride is out of limits-> 150 ppb. Also, TS basis is for RCS integrity, not RCS activity TRM - RCS Chemistry 1 Tech Requirements Manual 3.4.b 3.4.b-4 ReactorCoolantLP 11-RCXL-01 llA 35 1 TSBasis(old) ITS __34 _-5 A -92j J u'slno ~ 1 Fcility ExamBan Significantlyn I Mordifiedng U rora~ L SRO TS and Basis knowledge question I P O I I WcnEJ Li
_ _ - -- - --- - -- := Ut-
I..
.. ..... .......I.......... ......... ....... ..... ....... ......... ................. ................. ... ... ....................... ................ ............................. ......... . ......... ...... ............. ... .......
WE l Generic The following conditons exist on Unit 1:
- A reactor startup is in progress following an inadvertant plant trip
- The crew is performing steps of 1BwGP 100-2, "Reactor Startup"
- All control AND shutdown banks have been fully withdrawn
- The reactor is NOT critical Which of the following describes the required operator action?
Manually reinsert ALL Control and Shutdow Bank rods ife
.%a,, i IS. -l IManually reinsert ONLY the Control Bank rods ' . ls-:! F i
- Immediately open the Reactor Trip Breakers .
~ Rr AW evelr FR, F~g~te~e.Aplcto Braidwood j ~mar~7/190
~j 194'001G201 2.2.1 ~ ~ M7 ~ 2r 3 ysf m1EvoR~u io. i ec61 FPWG I Grt ifj O i 7111 iGENERIl Abilty to perform pre-startup procedures for the facility, d ncluding operating those controls associatedwith plant equipment that -s l could affect rea6ctivit'y."lt'nldn'peaigtoecnrosascae wt xp ar~ I t~ Per Attachment B, Contingency for not achieving criticality with all control rods fully withdrawn. Correct response is (A). (B)is action nm r for criicality belowLo-2 RIL. (C) Incorrect because"ALL rods must be fnserted:(D) incorrect - oniy applies to halted startups'durin'hg.
Pro edues . .BwGP 100-Ce Iii-
Feedwater pressure differential pressure program Control bank insertion limits vs. % Rated Thermal Power 1E9400.IG232.2.3 13.11 17 -
BGENERI; (multi-unit) Kriowledge of the design, procedural, and operational differences between units.
- (A) incorrect SDM both 1.3% (B) Incorrect U-&2 =3800 rect U-1"85-215psid, U-2-80-220psid (D) incorrect per.
PoerAsesionioGPGP~
isI
Ge n.............................................. .G Both Braidwood units undergo plant heatup and startups from cold shutdown conditons. Concerning the operation Isolation Valves, they are opened earlier in the startup of FW009A-D, Main Feedwater opening bypass isolabion flow control valves FW043A-Dprocess on (1) and must have startup purge logics satisifed before operating and FW046A-D on ___(2)-. by (1)(2 Unit 12 i Unit 2 : ; ; . : . - t- > i ; ? l i >.>t; 3. . ;- > i: ; ;; ;i t . X
.. Unit 2 . . Unit 1 ,; ,, , s tlor -. i:.i ! !.' j...ijj::'
f~¶ AnRe ~ ~ e,~ ntY ei Mmr i raidwood aM EJ 7/19/021 F1940~01G204 1 4a T2.-81,.
I~ ie[~ tuf E ~ ~ j *ru~ i (multi-unit) Ability to explain the" a nsin control board layouts, systems, instrumen ti a at a facilty.'...~nao~ndrcdrla.osetennt r opened on Unit .at-200F - in P 1001 On Unit2 opened at NOPNOT in GP 1004-O pur issixie ckts'() is only correct answer;ii_ _an Unit 2 still maint bas sb
,= 777, I "4
Que~tioi, opk IGeneric Greater than (1) feet of water must be maintained over the top of the reactor pressure vessel flange during movement of irradiated fuel assemblies within containment in order to (2)
,(1) (2) 23 E Have sufficient water depth available to remove 99% of the assumed -
10% iodine gap activity released from the rupture of an irradiated fuel . . .,, .. i, , , , . ...,
-KI 20 Provide sufficient wa volume to allow timforthe operator to nieihdi'ations 6"rec6 of ' 66Keff6an '-
. . . :T . 95. _.t KJ...._i _ i,_ _
I i.ii W j
B 123 * . . Maintain sufficient water volume as a heat sink for core cooling in the
' event' removal.
the operating RH loop fails to provide long term decay heat , t'. I
,;Wl t
-4 I . (
ir I --II I
- - -- .1 - . - I .. " n . ..
I, -J.
- 'O , Mhat the radiation I ',I U IL levels watret atthe adUVUoperatingtetop OT Me Ie Tuel assemolies elvto o ulhandling to ensure ; 4.
,equipment'rd beloZ4'mr/hr.ii.
- e sEx, ~ iie~ memory .. c.y riwo ~...711 9/021 194001 G225 2..2 l a , [
F1 G 0 [ - ,I
- s eD vol. (on itI .
Koldg ofbse'nteh dal spe'cil'ications fr lim~iting dniin o prbn nfaft iil antf(A)'Corre~ct .9e aid T (old) basis frMmin1imumcontaine, wate r dpt durngbemn of fue incnanmnA .
~efi tc itMFai t e~receurner e cecibtir~eh ue in i urn er Ild ITRM-. I 39.e-1 = -
IFH LP II I Ch -2FE _ -
ITS 3/4 9.10 (old) ,ITS Basis.(od) 'IBasis 9 A86: 7:
- IS _____
1 0 ~~n~ac~I~~I~I
- I afil0-~ FacilityExam Bank- Direct From Source ngtaigWg'ri r6Wtfi W&N .-- ~6W ae NRC OW 0 -E,:: EF I I W, iIER i
Generic efo i onditions exist on Unit 1:
- Reactor was tripped from 2% power during a normal coastdown for refueling
- 7/22/02 0900 Entered Mode 3, HOT STANDBY
- 7/22/02 1300 Entered Mode 4, HOT SHUTDOWN
- 7/23/02 0600 Entered Mode 5, COLD SHUTDOWN
- 7/23/02 2300 Entered Mode 6, REFUELING rthe earliest that fuel movement in the reactor vessel is allowed will be (1) to ensure that (2)
"17 7/251021100 Shod liatvAi fixss, , produci b .'] ' d , 7 7 7 7 .T T T.,T.T-'
r -- -- - -- - -_ _ .
I-. II .. .- .
. . I .I
- .E
- g. -
- I. - II ... .II .. I.
T7
,;-I ,I EL .21 2____ ____ 25/021100,;,.. ~ De~y hat removal abilit is ad~ut ua*x~ y,,~&;.rLK). ~~VK<. .,<- r?. . '.t, - . r I.-
., .1 1-7/qA/nq ENF linn I'll 0"-- V-1 -- _ .- ;. _;-
foo0 iive-u 1iIOn prouucts nave aecayed
_A.. . I 1. .-, - - I I 7I-I(V l1r,.
I W41 Decay neat removal ability is adequate
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._I .. :.:-i -.. I vGENERI Knowledge of eand spert flmOVemetprocedure.. s.....i ..............
- - II . I I I I - - ; . 7 , 11L -_- : -! - ! ; r , - , "I I TRM 3.9.a calls for 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />'subcritical before fuel movements can begin. (7/22/02 @ 0900 + 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> (3 days, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) =
7/26/02 @ 1300. TS Basis (old) defines the basis as ensuring the short lived fissiori Products have decayed off for radioactivity .*. 11I Hi ~
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7T7o&7:1 Other Faciliy 1~l ao Significantly odfej. .iihraigPrraLI Qu ,i neool&120 Prairie Island NRC 2000 Kewaunee NRC i I er II - --- __,I St~ s F]
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Unit is acompete eingrefuled fllowng ore ofloa in ccorance ith heuCrelLadinbPattrncspplid devitionfro orer o ththespeifie PWR Nucear omp nentTrasferLis (NC L),whil tr byNucler eFul iSevice.AAn NewFuelS nsprig fuel to or from the Spent Fuel Pool or the Vault, age requires the approval of _(1) _ AND _(2)_ before any further action is takenme.
Eniering Supervisor Station Nuclear Materials Custodian
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QaiidNuclear Engineer ~ Fuel HandigSprjo'~-,.
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~]Qualified Nuclear Engineer Station NuclearMaeilCutdn 22 I I eT H fiwE emory Braidwood xa 1901G31j2231M~4~ 2.2 dF~ 3J eiwPWG IG6p lo~p j- i
- *m ,. ~I GENERI Knowledge of procedures and limitations involved in initial core loading. . ~
p ntooerwA37-3,(A)Only correc epne BCD norc ..
l obntoso loal eiwr o actionsthat donot
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Generic The following conditions exist on Unit 1:
- Reactor power is 75%, steady state, equilibrium Xenon
- All controlling systems are operating in Automatic
- Turbine Impulse pressure transmitter PT-505 fails to its 50% value.
Control rods will respond by immediately stepping (1) at (2) steps per minute.
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194001G3 ii123 l2 R.ao .5 o PW ros"Giu [Ji I . . . 3 . - GENERI Knowledge ocotordpogamming. .. <. 333 ~ ..
(A) Correct - Tave program is 557'F-586Toradetao 9. A~t75%",tave is (.75)(29)+557-578.75oF Tref
(.5)(29)+(5S57)=571 .5F. A 7.2SIF mirnatch exists. ods will step in at 72 steps/min with anything at50% value is . ~
greater than a 5F mismatch 1t- - Iavc
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Today's date is November 12, 2002 (4th quarter). You have been assigned to authorize a rad worker to perform a routine task in the Auxiliary Building for which an estimated dose of 450 mrem will be received. Each has an exposure history this year as follows:
A. Age 46 Cumulated TEDE dose of 600 mrem Has a high lifetime exposure record B. Age 38 Cumulated TEDE dose of 48 mrem Has 2 quarters with an absent/no dose record C. Age 25....... *. .
Cumulated TEDE dose of 1260 nrem . . ;..: ' .
_Cumulated SDE dose of6'Rem to the.left hand .* ;i- ' .- -p- ^;, ;
D Age.i Cumulated TEDE dose of 80 rnrem - :- :.- ,
Cumdlated SDE dose of i15 mrem tothe fofea'rm upper '-i'.
Which of the ab6ve operators can be assigned'the task without' -e'ding any of Exelon's radiation exposure limits or submitting'approvil for exposure limit extensions?
Worker A 1
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~~.I. ,. . .-_X Knowledge of 10 CFR: 20 and related facility radiation control requirements. -,
(A) Incorrect - High lifetimeexposurerecordlimitsthisworkertoanannualdoseof1000mrem. (B) Incorrect - alloweddoses decreased by 1250 mrem for EACH absent/no dose record on file. (C) total dose received would remain below the 2000 mrem liiM..tI. I i11.44lIO gt10M A.CUC 44 \.4 MJ 14M011 li U i14iiMti_ U.17 MIVMMM11 R .. rei tl
.Fa...y ...... .e :.::e::. :::::o =. re..e N b.. Ref eto~ . ~u~ I Exposure Control and Authorization i RP-AA-203 42-5 INGET Student Study Guide Rad Potection jJJ~
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ue~onf {k ~Generic Whic he flloing of s anSROresponsibility?
flfPlacing the placard "Gas Decay Tank ReleaseIn Progress" on OPM02J priorto comnnoncing a release ;is
[ Performing second verification of the lineup to transfer a blowdown tank to the condensate storage tank: *...::.., 5 - . _ _
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rn o e pendg e v ef i ctO f toe li n t l ace ar e e se t k on eh oh n ystem . - - . .
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- owledge of SRO responsibilities for auxiliary systems that are outside the contrl room(g.watdisposaoadhadln A. Correct per reference. (B) Incorrect- second verifier is notirequired o be an SRO r . ,i release rate D) orect - isnotrequired tobe done yanSR (C) Incorrect - Rad Protection determines Ij ~ I Ito
-i U0-a-,G 'eneicWINI ,O, yMe ~ l iOm I A Site Area Emergency has been declared at Braidwood Station due to a LOCA outside containment. The LOCA is direct pathway to the environment exists, and limited makeup to the RWST is available. An operator has volunteered into the Auxiliary Building, a locally isolate the leak. This action would significantly reduce offsite dose and has all required approvals from the TSC.
to enter the Aux Building to If The operator has a lifetime exposeure of 3200 mrem TEDE and an exposure for the current year of 230 mrem.
What is the maximum exposure this operator may receive while performing actions to isolate this leak?
~]50 Rem TEDE Meoniy Freveood- jii~ 1~ ~ 7/19/021 4
3-.3 15 Ml E P
=4 A o 2_3.4 F GENERIl Knowledge of radiation exposure limits and contamination control, including permissible levels in excess of those authorizd xp a o Per EP-AA-1 13, Personnel Protective Actions - 25 Rem TEDE is the emergency exposure limit. IT shall be voluntary and limited once in a lifetime. (C) is correct to? >
. Personnel Protec veActons l EP-AA-113 413 ExposureControl and Authorization I RP-A-23 l j J 1 J
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pE~eitGeneric
- F Unit 1 is in Mode 4. Containment Purge is in Progress using the Mini-purge Supply and Exhaust Fans. While the purge is in progress, I RE-PROO1, Containment Purge Effluent Rad monitor, exceeds the ALERT setpoint.
Which of the following must be performed ?
jMANUALLY stop the containment purge in progress - ~ .*9 L ALVERIFY containment purge AUTOMATICALLY s S {VERIFY Post LOCA Purge filter unit AUTOMATICALLY aligns. - ji ,, ;,
MANUALLY alignPost LOCA Purge fIte unit
- 194 001G3` 239 _ 1' I 3J' Se4F PWG3 MO I lGENERI Knowledge of the process for performing a containment purge.
ra o oNA. Correct. B. Incorrect. The AR011/12 auto isolates-the purge path, not 1RE-PROO1:
ost urgefilterunit. D.Incorrect.Procedure referencedirectsstoppingpurge C. Incorrect. There is no auto alignment of.4i'l theoca
-r er (vicemanuallyaligningtheflrunt ..
rnurnit)q~ R e filter-efreie
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.11 I......... I Given the following plant conditions:
- Unit 1 is at 100% power
- Unit 2 is at 100% power
- OPR09J "CC HX Outlet Unit 0 Radiation Monitor" is in HIGH alarm
- A confirmed High Alarm has been determined by Chemistry
- The 0 CC HX has been subsequently isolated The crew should now verify:
E Only 1C0017 is closed and enterthe LCOfor Unit i cw ~ - I ....... ,,
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Only 2CC017 is closed and enter the LCO for Unit 2 CCW .-.. :',
sj w*.j*:1 . '.,: " i j Both 10C017 and 2CC017 are ctosed and enter the LCO'forboth units for CCW . ,,:* r ...-.-. :..i,'.
[ Both 1CCO17 and 2CC017 are closed and donotne6d't entera LCO for either unit *.-4l' ..- +1 . 'i,/.j o'u~;
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'iBo~gu .{ .i. lGENERIl Ab.ilityto control radiation releases. . .*....j......
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- Bothivent valves receive a closure signal from thecommon CC heat exchanger rad monitor. Must enter LCO for both units I
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~d~&i~ure M f 1,2000 Bwd NRC Exam
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Which of the following conditions would NOT require immediate entry into 1BwEP-0, "Reactor Trip or Safety Injection", if the condition were to occur I inadvertently with the reactor operatino at 100% power?
I Safety Injection actuation o Train n 'A' '..',.; . ....:. .l.
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Loss of *Instrument t Bus 112 With PR Instrument N-44 failed
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Radioactive steam does NOT contaminate the main steamlines;> I ns*R Jev -. og[eLv' Comprehensio 1t~ Braidwood t.-7/1/02j 1 1194001G406 2.4.6 .. 0 ir . 1-2 4.O. . . -
yse tI lGENERIl UKnowledge symptom based EOP mitigation strategies. L"' . b ' i 9f P-3bacgro~jn doumets (B Corec. r_ Onts rupitured~SG'depressurization during upcoming-RCS cooldown steps. (A), .
Incorrect-the ruputed SG will not be used for cooldown .unless it is the only intct SG.. (C) Incorrect - No in E-3 ig with
-ore...oi s heolyitat G..( E3mitigatin I I It.iVa 1.L I I U : 39 V3.. ;IV U ,U y
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I ., 33 10 I1 BwEP-3 [Step 4 . 11.. d. ZCl,
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The following conditions exist on Unit 1:
- Bus 141 is DE-ENERGIZED
- Bus 142 is DE-ENERGIZED
- RCS pressure is 2220 psig and decreasing slowly
- Pzr level is 31% and decreasing slowly
- Preparations are being made to cool the RCS to 350°F in order to minimize further RCS inventory loss Operators are performing steps in (1) and are CAUTIONED NOT to decrease RCS Hot Leg temperatures below 350°F to prevent (2)
(1) (:2)
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'Ell BWE -0 Re ctor Trip....or SI ; PrsurizedThermal Sh dit.o.
EdI1 BwEP-0,Reactoir Trip or Si Accumulator Nitrogen injection i s : : ; .. -
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hot leg temps should not be decreased foless fthan 35r (B) Incorect - E 0 . :.3
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.1steps are 3 in progrossj.(D)lnco(~rredt-Y. OO
.!pIJ~3 II a ,3430 AI~..3I PTS is nota t66n r(n brlhqr~kcrnifiind dfrrti~monfXc 3~J)IPUSJIV
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Loss of All AC LP il1-CA-XL-01 27 6 Loss of All AC Power BwCA-0.0 B1 _ s 3 F39 F1CE Emergency Response Guidelines ;ERGCA-0.0 I 1 1118 i 11C 1 L-_aa aI.. qu..... o rji . a . n ________________________________________________..___I____________-------------__-
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[Generic An emergency call has been received by the Assist NSO in the control room, reporting a large fire in the Turbine Building.
Which of the following is NOT one of the Shift Managers responsibilities while completing the checklist for a Fire/Hazmat Spill Response?
ai Verify the Fire Brigade has been notified/dispatched , . -.. .: '
0 [Notificati of Rad Protection to dispatch personnelto the t e,. .7`7.;' 5 ;
JAssessment of the fire/scearioad classification for thelEinegen'cy Plan~-
g Announcementofthe fire over the planitPA system' and sounding of the plantfire alarr:m eMemory Braidwood 7/19/021 194001 G427 2 27;4 I PWGG Mr l3 !-t y trl lGENERI IKnowledgeoffireintheplantprocedure . . , . *>. . .! ' .
l~niM~o~('A-C) are all part of the checklist for the S.Manager topeform :(D) Correct - this jis the assist operators reisponsibility per~the 'i A, i pP116 0Fire,/HazmatSpillResponse, endixA&D _4,.:l; I~ E~ E .. I->
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I es n I j Generic . . . - - ----- ---- .. I A large Break LOCA concurrent with a loss of containment integrity has been INITIALLY classified as a General Emergency.
The offsite state authorities will be notified of this event on the (1) phone, and the NRC will be notified on the (2) phone.
(1) - (2) _
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.. 926 ~~ . 9 ~to PG L - G~&rp EjS3 ru Tj the emergencv Dlan.
-:. -- ---- 11.1 . .. 1. ---- L 4),Dedicated lines in the MCR for NARs is GREEN and the NRC is RED. ' s:....' 5<;" .-- ii' I r e WN 4
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Notificaaions Q~eti~ou I ew eMc i o ti&Isdurn rlPora< j NR~Li IZ8
-194001G432~ ;aq 31 R)VVeG _P24.2f Or I S- r p: Li o IGdENERI ~4 I Knowledge of operator response to loss of all annunciators. II I t (A Corec - erBwd EALs: - MU6 and 1BwOS AN.lA AAR A.3 (BCD) Incorrect7-the event does warrant EP classification anhd xp!nahn 7 ~does requircotinuous monitoring .2 ~ ~ - ; I I 1...,
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Loss of Annunciators IlwSNl .~~Iz~
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o ~~New 1L~~n ~ ~ AUe uii SRO jKnowledge of EALs and Actions in the ~AAR (selecting correct procedure) FieeiL I I $u~El