ML20210N813
| ML20210N813 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 05/02/1997 |
| From: | Biermann F UNION ELECTRIC CO. |
| To: | Bundy H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| Shared Package | |
| ML20210N765 | List: |
| References | |
| RTR-NUREG-1021 NUDOCS 9708260128 | |
| Download: ML20210N813 (117) | |
Text
l 1
Fran Diermann -Operating Supervisor Union Electric Company Callaway Plant P.O. Box 620 Fulton,MO. 65251 May 2,1997 Iloward Dundy-Lead Examiner US NRC Region IV 611 Rynn Plaza Drive, Suite 400 Arlingten, TX. 76011 8064
Dear Howard:
This letter accompanies the examination outline for the retake examination at the Callaway Plant. Tids outline is for the retake of the written section for the Reactor Operator license exam which is scheduled to be administered on June 27,1997.
This outline was prepared based on the guidelines in NUREG 1021, " Operator Licensing Examination Standards for Power Reactors," Interim Revision 8. During the preparation of this outline, the Callaway Plant IPE and specific plant operating events were reviewed for topic selection.
The Exandnation Outline Quality Checklist, ES 2012, has been reviewed and is included with the submitted outline. This review indicated that while no specific K/A topic has been repeated from the RO exam administered on February 24, there were 11 of 100 K/A topics that could be considered as related.
Thank you for your time on this examination review. I am looking forward to hearing from you with any feedback you have on : Ids outline. Please contact me at (573) 676-8404 or Bruce Moody (573) 676 8194 with any conunents.
Sin y,
F Biermann Attachments (1) Outline QA Checklist ES-201-2 (2) Exam Outline ES-401
)
9708260128 970820
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1 4
ES-401 Callaway Plant Re'ake RO Examination Outline i
Facility:
Callaway RO Retake Date of 6/27/97 Exam Level: RO Exam:
K/A Category Points Tier Group Point K
K K
K K
K A
A A
A G
Total 1
2 3
4 5
6 1
2 3
4 1.
1 2
2 4 N @? E 2
6 m..
W('
Wa 16 q
4.h N t
" [.N A
kD,':h (;I A[ M:'
k,
-k
/' v
.^
C j
Emergency &
2 5
3 5
gg; ' pig TJ 2
2 g i}n Na 17
+,n e
n w,
Abnormal 3
2 s' O-1 l;g &..
W*
3
~
~
f{@!h g
gp 5
8 Plant 5
5 10 We 36 Tier
)
Evolutions pjf g
- g Totals ps 49 A
i 1
5 1
1 4
2 1
1 3
1 4
Wa 23 2.
2 3
1 0
3 1
1 1
4 2
4 W*
20 Plant 3
1 0
1 2
0 0
2 1
1 0
8 Systems 9
2 2
9 3
2 4
8 4
8 Na 51 Tier Totals
- 3. Generic Knowledge and Cat 1 Cat 2 Cat 3 Cat 4 Abilities 4
3 1
5 13 Note:
Attempt to distribute topics among all K/A categories; select.
at least one topic from every K/A category within each tier.
Actual point totals must match those specified in the table.
Select topics from many systems; avoid selecting more than two or three K/A topics from a given system unless they relate to plant-specific priorities.
Systems / evolutions within each group are identified on the associated outline.
The shaded areas are not applicable to the category / tier.
NUREG-1021 Interim Rev. 8, January 1997
~
ES-401 Callaway RO June 1997 Re-Examination Outline Form ES-401-4 Emergency and Abnormal Plant Evolutions - Tier 1/ Group 1 K
K K
A A
G E/ APE # / Name / Safety Function WA Topi@
g pm A203 Required Actions for Stuck Rod 3.5/4.4 1
000005 Inoperable / Stuck Control Rod / I X
000015/17 RCP Malfunctions / IV X
A123 Response to High RCP Vibration 3.1/3.2 1
W/E09&E10 Natural Cire. / IV X
K2.2 Subcooling and Cooldown Rate on Natural Cirr 3.6/3.9 1
000024 Emergency Boration / I X
A201 Verification of Emergency Boration Flow 3.6/3.9 1
000026 Loss of Component Cooling Water / Vill X
K303 Immediate Actions for Loss of CCW Pump 4.0/4.2 1
000027 Pressurizer Pressure Control System Malfunction /Ill 000040 (W/E12) Steam Une X
A2.2 Restoration from an uncontrotted depressuization 3.4/3.9 1
1 of all Steam Generators Rupture - Excessive Heat Transfer / IV W/E08 RCS Overcooling - PTS / IV X
A1.3 RCS Post Soak C/D Limits following PTS 3.6/4.0 1
000051 Loss of Condenser Vacuum / IV X
K301 Loss of Steam Dumps with Loss of Vacuum 2.8/3.1 1
000055 Station Blackout / VI X
K302 ECA0.0 Actions for ESW Pump (IPE) 4.3/4.6 1
000057 Loss of Vital AC Elec. Inst. Bus / VI X
A219 Immediate Actions for Less of NN02 (IPE-Flood) j RO/4.3 1
000062 Loss of Nuclear Service Water / IV 000067 Plant Fire On-site / IX X
K101 Classification / Suppress on of Fire (IPEE) 2.9/3.9 1
000068 Control Room Evac. / Vill 000069 (W/E14) Loss of CTMT Integrity / V X
E14K2.1 Manual Actions for High Containment 3.4/3.7 1
Pressure 000074 (W/E06&E07) Inad. Core Cooling / IV X
E07K1.2 Use of Saturated Core Coofing Procedure 3.1/3.6 1
X E06K3.1 Stopping RCPs in inadequate Core Cooling 3.4/3.8 1
X 074A201 Determine conditions meeting Subcooling 4.6/4.9 1
000076 High Reactor Coolant Activity / IX X
A202 High RCS Activity Samp!hg Requirements 2.8/3.4 1
i K/A Category Totats:
2 2
4 2
6 Group Point Total:
16 Target 16
ES-401 Callaway RO June 1997 Re-Examination Outline Form ES-40i-4 Emergency and Abnormal Plant Evolutions -Tier 1/ Group 2 K
A A
G E/ APE # / Name / Safety Function K/A Topic (s)
Imp.
Points 2
000001 Continuous Rod Withdrawal / I X
K206 Actions on Continuous Rod Wdhdravd 3.0/3.1 1
000003 Dropped Control Rod / I 000007 Reactor Trip - Stabilization - Recovery / I X
K105 Time for Source Range Energization afterTrip 3.3/3.8 1
000008 Pressurizer Vapor Space Accident / Ill X
AK303 E-0 Response to failed Open Spray Valve 4.1/4.6 1
000009 Small Break LOCA / Ill X
A101 Diagnosis of PRI vs SEC Leak from parameters 4.4/4.3 1
000011 Large Break LOCA /111 X
K101 Heat Removal on Primary LOCA 4.1/4A 1
'NE04 LOCA Outside Containment /111 X
EK1.2 Precaution during valve strokes in ECA-12 3.5/42 1
W/E03 LOCA Cooldown/ Depress. / IV X
K2.1 Block Low Steamune Pressure SI 3.6/4.0 1
X EK1.2 Reason for attemating ECCS trains 3.6/4.1 1
WE11 Loss of Emergency Coolant Recirc. / IV X
K1.1 Makeup Sources in ECA-1.2 3.5/4.2 1
WE02 Si Termination /11!
X EA22 Procedure Guidance fo!!owing Si termination 3.5/4.0 1
000022 Loss of Reacter Coolant Makeup /11 X
K302 Loss of Charging Flow due to valve failure 3.5/3.8 1
000025 Loss of RHR System / IV X
A102 Air entrainment during RHR operation 3.8/3.9 1
000029 Anticipated Transient w/o Scram / I X
K306 Failure of Turbine to Trip 42/4.3 1
000032 Loss of Source Range NI / Vil 000033 Loss of Intermediate Range NI / Vil 000037 Steam Generator Tube Leak /111 X
K310 Response to high rad on SG Tube Leak 3.3/3.7 1
000038 Steam Generator Tube Rupture /111 X
K301 Preferred order of depressurizing Primary 4.1/4.3 1
w/SGTR 000054 Loss of Main F.eedwater/ IV X
A203 AFW Pump Start Signals 4.1/42 1
WEOS Inadequate Heat Transfer-Loss of X
K2.2 Transition from FR-H.1 to E-1 3.9/42 1
Secondary Heat Sink / IV 000058 Loss of DC Power / VI 000059 Accidental Liquid RadWaste Rel. / IX 000060 Accidental Gaseous Radwaste Rel. / IX 000061 ARM System Alarms / VII I
W/E16 High Containment Radiation / IX K/A Category Totals:
5 3
5 2
2 Group Point Total: 17 Target 17 NUREG-1021 Interim Rev. 8. January 1997
ES-401 Callaway RO June 1997 Re-Examination Outline Form ES-401-4 Emergency and Abnormal Plant Evolutions - Tier 1/ Group 3 I(
^
^
E/ APE # / Name / Safety Function K/A Topic (s)
Imp.
Points 3
000028 Pressurizer Level Malfunction / II 000036 Fuel Handling Accident / Vill 000056 Loss of Off-site Power / VI 000065 Loss of Instrument Air / Vill X
K308 E-0 Actions forloss of Air to EFHV43/44 3.7/3.9 1
W/E13 Steam Generator Over-pressure / IV X
E13EA2.1 Response to SG Overpressure 2.9/3.4 1
W/E15 Containment Flooding / V X
E15EK3.3 Emergency isolation of ECCS Accumulator 2.9/2.9 1
l K/A Category Totals:
2 1
n/a Group Point Total: 3 Target 3 NUREG-1021 Interim Rev. 8. January 1997
ES-401 Ca!!away RO June 1997 Re-Examination Outline Form ES-401-4
~
Piant Systems -Tier 2 / Group 1 K
K K
K K ^ ^ ^ ^
E/ APE # / Name / Safety Function K/A Topic (s)
Imp.
Points p
3 g
5 6
2 3
4 001 Control Rod Drive X
K565 Power Mismetch Effect on Rod Control 32/3.6 1
X A102 Plant Response to Pimp Failure 3.1/3.$
1 X
K504 Rod Insertion Limit w/P-A Cony.
4.3/4.7 1
Malfunct 003 Reactor Coolant Pump X
A404 Monitor #1 Seal DP on RCP 3.1/3.0 1
X K112 RCS Leak Thermal Banier Isotation 3.0/3.3 1
004 Chemical and Volume Control X
K107 Plant response to Flux Doubling 2.6/2.9 1
X K106 Response to VCT Level Channel 3.1/3.1 1
Failure X
A227 RWST Operability in Mode 6 3.5/4.2 1
l 013 Engineered Safety Features Actuation X
K113 Fuel Bldg Ventilation on a SIS 2.8/3.1 1
[
X K201 Downpower cross-trip biccks 3.6/3.8 2.7/3.1 1
X A402 Reset of ESFAS Channels 4.3/4.4 1
j 015 Nuclear instrumentation X
K302 Response to Power Range N1 Failure 3.3/3.5 1
X K604 Source Range NI Failure 3.1/32 1
X A201 Idling current to IR Nuclear Inst 3.5/3.9 1
017 In-core Temperature Monitor X
K403 IncoreThermocouple Accuracy 3.1/3.3 1
022 Containment Cooling X
A301 Flow to Containment Coolers on SI 4.1/4.3 1
056 Condensate X
K103 FeedwaterTemp Response to 2.6/2.6 1
Feedwater HeaterIsolation 059 Main Feedwater X
A403 FRV/ Feed AP on Powerincrease 2.9/2.9 1
X A212 MFP trips from FRV failure 3.1/3.4 1
061 Auxiliary / Emergency Feedwater X
K401 AFAS/LSP Actuation Requirements 3.9/42 1
X K404 Flow Control Valve Operations 3.1/3.4 1
068 Liquid Radwaste X
K401 Auto isolation of Radwaste Discharge 3.4/4.1 1
072 Area Radiation IVonitoring X
A403 Area Rad MON Source Check 3.1/3.1 1
K/A Category Totals:
5 1
1 4
2 1
1 3
1 4
0 Group Point Total: 23 Target 23 l
NUREG-1021 Interim Rev. 8. January 1997
ES-401 Callaway RO June 1997 Re-Examination Outline Form ES-401-4 Plant Systems - Tier 2 / Group 2 K
K K ^ ^ ^ ^
E/ APE # / Name / Safety Function K/A Tcpic(s)
Imp Points 3
5 6
2 3
4 002 Reactor Coolant X
A303 Master Pzr Press Controller Setting 4,4/4 _
j l
006 Emergency Core Cooling X
A402 ECCS Response to Valve Failure 4.0/3.8 1
010 Pressurizer Pressure Control X
K611 Response to Pzr Press Channel Failure 2.7/3.1 1
011 Pressurizer Level Control X
A210 PZR Level Channel Failure 3.4/3.6 1
012 Reactor Protection X
A406 Operation of Rx trip breakers 4.3/4.3 1
014 Rod Position Indication X
A402 Selecting Auto Rod Speed 3.4/3.2 1
l 016 Non-nuclear Instrumentation X
K112 Requirements for AMSAC Actuation 3.5/3.5 1
i 026 Containment Spray X
A301 Plant Response to CSAS 4.3/4.5 1
029 Containment Purge X
A201 Shutdown Purge Operations 2.9/3.6 1
033 Spent Fuel Pool Cooling X
A101 Normal Makeup to SFP 2.7/3.3 1
035 Steam G,enerator X
K403 initiation of BSPSIS 2.6/2.8 1
039 Main and Reheat Steam X
K508 Plant response to S/G PORV failure 3.6/3.6 1
055 Condenser Air Removal 062 AC Electrical Distribution X
K104 Offsite AC PowerSources 3.7/4.2 1
X K201 Power Supplies to Major t.oads 3.3/3.4 1
063 DC Electrical Distribution X
A401 Monitor NK Iineup in the Control Room 2.8/3.1 1
064 Emergency Diesel Generator X
A212 HVAC Operability for DGs 2.8/3.1 1
X K402 D/G Output BreakerTrips 3.9/4.2 1
073 Process Radiation Monitoring X
K101 Response to CCW Rad Mon Alarm 3.6/3.9 1
075 Circulating Water X
K401 Cooling Twr Bypass Valve Operation 2.5/2.8 1
079 Station Air 086 Fire Protection X
A201 Manual Shutdown of Firepumps 2.9/3.1 1
K/A Category Totals:
3 1
0 3
1 1
1 4
2 4
0 Group Point Total: 20 Target 20 i
NUREG-1021 Interim Rev. 8. January 1997
l l
ES-401 Callaway RO June 1997 Re-Examination Outline Form ES-401-4 Plant Systems -Tier 2 / Group 3 K
K K
K ^
^
E/ APE # / Name / Safety Function
^k^3 K/A Topic (s) imp. -
Points p
3 g
6 4
005 Residual Heat Removal X
K407 RHR Valve Interlocks 3.2/3.5 1
[
X A103 Flashing in RHR HX 2.5/2.6 1
l l
007 Pressurizer Relief / Quench Tank X
A205 Relief paths to the PRT 3.2/3.6 1
I 008 Component Cooling Water X
K104 Source of CCW Leakage 3.3/3.3 1
027 Containment lodine Removal 028 Hydrogen Recombiner and Purge Control i
l 034 Fuel Handling Equipment 041 Steam Dump / Turbine Bypass Control X
A101 Plant Response to Tavg Failure 2.9/2.9 1
l 045 Main Turbine Generator X
K411 Inputs to Turbine Runbeck 3.3/3.6 1
076 Service Water i
078 Instrument Air X
K302 Valve fait position on loss of air 3.4/3.6 1
l 103 Containment X
A301 Rad Mon response to C!SA 3.9/4.2 1
i l
F r
i t
K/A Category Totals:
1 0
1 2
0 0
2 1
1 0
0 Group Point To'.al: 8 Target 8 NUREG-1021 Interim Rev. 8, January 1997 l
d ES-401 PWR RO Examination Outline Facility:
Callaway Date of 6/27/97 Exam Level: RO Exam:
Category K/A #
Topic Imp.
Points 2.1.17 Verbal Communication Policy 3.5/3.6 1
2.1.1 License candidates requirements in the CR 3.7/3.8 1
Conduct of 2.1.32 Precautions / Limitations for Radwaste Supply 3.4/3.8 1
Operations 2.1.11 Tech Spec-Min Tomp for Criticality 3.0/3.8 1
Total 4
l 2.2.13 Protective Tagging Partial Clearance 3.6/3.8 1
2 2.2.22 Tech Spec Safety Limits 3.4/4.1 1
i Equipment 2.2.32 Fuel Handiling Accident immediate Actions 3.5/3.3 1
Control Total 3
2.3.2 Radiological Posting Requirements 2.5/2,9 1
Radiation Control Total 1
2.4.11 Control Room Evacuation 3.4/3.6 1
2.4.20 RCP Trip Criteria 3.3/4.0 1
Emergency 2.4.1 Reactor Trip Requirements 4.3/4.6 1
Procedures 2.4.22 Critical Safety Function Implementation 3.0/4.0 1
and Plan 2.4.11 Plant Computer Outage 3.4/3.0 1
Total 5
Tier 1 Target Point Total RO 13 NUREG-1021 Interim Rev. 8. January 1997
)
CALLAWAY PLANT IPE/PRA Referenced to Callaway Plant RO Exam of 2/?4/97 Individual l'lant Evaluation RO Written Major Event (% CDF)
Question #
TOl'IC IntomalFlooding (30.5%)
37 Loss of DC coritrol power to major NB breaker (Effect of Flooding) 92 Loss of Vital AC Instrument Bus NN02 (Effect of Flooding) 97 Loss of NK01 battery charger (Effect of Flooding)
Station Blackout (30.3%)
2 Makeup to Spent Fuel Pool during Loss of All AC 39 Loss of All AC response per ECA-0.0 63 Depressurization of Ali S/Gs while in ECA-0.0 LOCAs(19.0%)
10 Large Break RCS LOCA - response in ES-1.3 12 Hydrogen Control following RCS LOCA 20 Maintaining RCP seal Cooling - (prevention of seal induced LOCA) 56 SI Termination Criteris while in E-1 88 Adverse containment while in E-1.
33 Loss of all circualting water / loss of vacuum Transients (17.7%)
42 St. sam break outside containment 73 Pressurizer Steam Space Leak 74 Loss of NB01 Electrical Bus 84 Actions for isolating steam leak in E-2 90 Loss of offsite power - natural cire cooldown 31 Inventory Control while in E-3 SGTR (1.4%)
95 Steam Generator isolation in E-3 19 Basis for Turbine Trip in Response to ATWS ATWS (0.7%)
54 Operation of the AMSAC System PRA Risk Sianificant Systems
(% Contributing)
KJ/NE (30%)
41 LSELS actuation while perfoming Diesel Testing 50 Standby Diesel Generator operation following Emergency Start ESW (23%)
53 Use of ESW as backup supply to AFW AFW (12%)
61 TD Auxiliary Feodwater Pump setpoints for speed control RHR (7%)
75 RCS Pressure Control while on RHR 77 Loss of RHR while in Reduced Inventory CCW (4%)
8 Loss of CCW to RCPs while in EOPs 45 CCW response to loss of NB uus supply transformer 76 CCW toads during ECCS actuation
.-..~.
.QBJEF EXAMINER OUTLINE COMMENTS CALLAWAY RO RETAKE 6/27/97 BO WRITTEN Tier 1, Group 1, K/A 000024A201. The importance factors should be 3.8/4.1 vice 3.6/3.9. Resolution: Licensee will correct.
Tier 1, Group 2, K/A 000001K206 The topic description does match the K/A selection which refers to interactions with the Tavg/ Tref deviation meter.
Resolution: Licensee will change topic description to match selected K/A, Tier 1, Group 2, K/A W/E11K11. Importance factor is 3.7/4,0 vice 3.5/t,2.
Resolution: Licensee will correct, Tier 2, Group 1, K/A 013K201, importance factor is 3.6/3.8 vice 2.7/3.1, Resolution: Licensee will correct.
'ier 2, Group 3, K/A 007A205, Topic does not appear to relate to exceeding pressure limits for PRT, Resolution: Question is intended to relate to leak paths which cause PRT pressure limits to be exceeded. Licensee will reword topic to
- clarify, Tier 2, Group 3 K/A 045K411. Topic appears to relate to K/A 045K412 vice 045K411. Resolution: Licensee will correct K/A reference.
Tier 3, Radiation Control. Topic appears to relate rnore closely to K/A 2.3.1 than
2.3.2. Resolution
Licensee will change reference to K/A 2.3.1.
Tier 3, Target Point Total RO, Tier 1 is incorrectly referenced. Resolution: Licensee will correct reference.
1
{
LALLAWAY PLANT EXAMINATION COVER SHEET TRAINING DEPARTh1ENT COURSE TITLE:
RO INITIAL LICENSE EXAMINATION DATE:
l NAME (Prin:):
SCORING:
SIGNATUP.E:
Points Possibic; 100 Points Missed:
..... ~. _.
Grade.
DIRECrlONS: lil ACK OUT CORRECT ANSWCRS E LB_ll C l fE l^llBllCI E si l ^ l E I C l l D_J E l B I LCJ Fol i
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52 lAllBI M IDI L^J I B I E l o l 2
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lAlIhl E lDl 29 lAllBllCl E 54 E lBllCllDl 79 %lBl C E 4
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$ E lBllCllDl 30 ElBllCllDl lAl E h lDl l A l fiil E LoJ 55 80 l
l 6 I Al [ B l l C l E 31 l AllBIE IDI so l a.J I B i l C l E si Fill B I LCJ E ElBllCllDl 32 lAllBllCl E 57 E lBllCllDl 82 lAllBllClE 7
EB ICilDI 33 l^l E ICllDI 58 l^11BI M IDI E lBilCilDI S
83 9 I AllBIE lDI L^J I B I E l o l 59 l ^ l E I C l LDJ l^l E ICllol 34 84
[A} E lCllDl 35 E lBllCl[Dj lAl E lCllDl 85 l A l E l C l l D l 10 60 l
11 lAl @jfCl E IAl l B l E l D l 61 l AllBllCl E E lBl W lDj 36 86 IAllBl E [M 12 l AllBllCl E 37 62 l AllBllCl E
[AjlBllClE S7 l A l_LBJ_EE_ 38.Eg I C l_l D l
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l 14 ElBllCllDl E IBllCl[M l AllBllCl E E lBllCllDl 39 64 89 15 E lBllCllDl lA] E lC)lDl 65 E lBl[CjlDl IIl E lCllDl 40 90 lAlIBl E lDl 41 l AllBl E lDl l Al E lCllDl l AllBllCl E 16 66 91 E lBllCllDl 42 l AllBllCl E l Al E lCllDl l AllBlE lDl j
17 67 92 18 lAl E ICllDI 43 E IBllCllDI cs l A I E I C l l D I lAjlBilCI E 93 19 lAllBllCl E ElBllCllDl 69 ElBllCllDl l AllBllCl E 44 94 E lBllCllDl l AllBl E lDl lAllBllCl E 95 l AllBl E lDl j
20 45 70 l
21 l AllBl E lDl l AllBl E lDl 71 l Al E lCllDl lAllBl E lDl 46 96 l Al E lCllDl 47 l AllDllCl E 72 E l B l l C l l.D l E lBllCllDl 22 97 23 l Al E lCllDl l Al E lCllDl 73 lAllBl E lDl 98 lAllBllCl E 4S
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E lBllCllDI E lBllCllDI IAllBI E lDI 99 lAllBI E lDI 24 49 74 f
25 ElBllCllDl l AllBl E lDl 75 l AllBllCl E l Al E lCllDl 50 100
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i 3
QUESTION #1 A Safety injection has occurred due to a Steam Generator Tube Rupture. The crew has just completed verifying that both NB01 and NB02 are energized per Step 3 of E-0, Reactor Trip or Safety injection. Which one of the following describes the HVAC flowpath for the Fuel Building
- at this time?-
1 A.
Fuel Building supply and normal axhaust stops; emergency exhaust dampers align to the
_ Aux. Buildingc_ z -.
B.
Fuel Building supply and normal exhaust stops; emergency exhaust dampers align to l
l Fuel and Aux Building; L
- C.
Fuel Buildi.ng supply keeps running or starts; normal and emergency Fuel Building exhaust isolates.
l D.
Fuel Building supply and exhaust keeps running or starts; emergency exhaust dampers l
align to the Fuel _ Building.
ANSWER:
A. Fuel Building supply and normal exhaust stops; emergency exhaust dampers align to the Aux Building.
K/A #: 103K113 2,8/3.1 KA DESCRIPTION: FB Ventilation on SIS :
OBJECljyE #:
011039.O_D
REFERENCES:
T61.0110.6 LP 39, Page 40 AUTHOR: RBM SOURCE: BANK Modified Y-L DISTRACTER EXPLANATION:
Response B is incorrect because the emergency exhaust does not align to the Fuel Bldg Response C is incorrect because the Fuel Bldg supply does not keep running. Response D is incorrect because the Fuel Bldg supply and exhaust does not keep running and the dampers do not align to the Fuel Bldg.
RO Outline #32
= n.
QUESTION #2:
The plant is at 100% power, alisystems lined up for normal operation with the NCP operating with 120 gprp letdown flow. A complete loss of offsite power occurs without a safety injection signal. All plant systems respond as designed except that 2 rods in the *C" Control Bank are at
- 200 steps withdrawn.
- Which ONE of the below describes the actions that will be performed during the initial response
. to the above conditions?
A.
- Manually start either CCP, open suctions from RWST to CCP since immediate boration flow from Doric Acid Transfer Pumps are not available.
B.'
Ensure NCP is running, open BGHV8104, Emergency Boration to CCP suction, and start either Boric Acid Transfer pump.
C.
. Ensure either CCP is running, open suctions from RWST to CCP since immediate boration flow from Boric Acid Transfer Pumps are not available.
D.
Ensure either CCP is running, open BGHV8104, Emergency Boration to CCP suction, and start either Boric Acid Transfer pump.
ANSWER:
D. Ensure either CCP is running, open BGHV8104, Emergency Boration to CCP suction, and start either Boric Acid Transfer pump.
K/A #: 005A203 3.5/4.4 KA DESCRIPTION: Required Actions for Stuck Rod OBJECTIVE #:
003B610A --
REFERENCES - OTO-ZZ-00003.
AUTHOR: FXB SOURCE: NEW - HO DISTRACTER EXPLANATION:
A. -
CCP Starts automatically.
B.
NCP has loss of power C.
Boric Acid system is NOT shed on Loss of Offsite Power RO Outline #52 I
m~
- - ~ ~
..=.
)
i QUESTION #3 Plant conditions are as follows:
. Safety injection occurred on low steam line pressure
. ECA-2.1 Uncontrolled Depressurization of All Steam Generators
. ECA-2.1, Step 2, Control Auxiliary Feed Flow to Minimize RCS Cooldown, has AFW flow at 15,000 lbm/hr to each S/G.
Level
- Pressure A S/G 20% WR 355 psig stable B S/G 15% WR 300 psig decreasing C S/G 15% WR 290 psig decreasing D S/G 25% WR 355 psig increasing Which ONE of the following actions should be taken ?
A.
Continue with performance of ECA-2.1, Uncontrolled Depressurizaton of All Steam Generators B.
Transition to E-2, Faulted Steam Generator Isolation C.
Transition to FR-H.1, Respons to Loss of Secondary Heat Sink D.
Transition to E-0, Reactor Trip or Safety injection ANSWER:
B. Transition to E-2, Faulted Steam Generator Isolation K/A #: E12EA2.2 3.4/3.9 KA DESCRIPTION: Restoration from Uncontrolled Depressurization of All S/G's OBJECTIVE #:
003D160D
REFERENCES:
ECA 2.1 Page 1 AUTHOR: PJM SOURCE: NEW - HO DISTRACTER EXPLANATION:
A. Plant conditions have changed requiring a transition.
B. Correct. Fold out page re "es transition to E-2 on any S/G pressure increasing.
C. Since AFW flowis under ator control, the transition criteria to FR-H.1 is not met.
D. No signals have been re lived that would require transition to E-O from this point.
RO Outline #66
s, QUESTION #4
-i You are the on_-duty Reactor Operator; In accordance with plant policy, which une.of the
- following non licensed individuals may you allow to start the 'A' Safety injection (SI) Pump from Panel RLO17 in the Control Room?
A. -
- Any system engineer authorized by the Shift Supervisor who is performing Si system surveillances.
~
. Any assistant eqlipment operatEr' performing OJT on the SI system who is being B,
~
monitored by the Control Room Supervisor, C,
Any individual who is in a license training program under my direct observation.
t
' D; Any electrical maintenance supervisor troubleshooting why the SI pump vibration readings are abnormal.
ANSWER:
r C Any individual who is in a license training program under my direct observation.
K/A #: 2.1,1 3.7/3.8 KA DESCRIPTION: License Candidate Requirements in Main CR OBJECTIVE #:
003A040H
REFERENCES:
T61,003A.6 LP 4 ODP ZZ 00010 AUTHOR: RBM-SOURCErNEWvL-DISTRACTER EXPLANA'ilON:
-The only non-licensed ind viduals authorized to operate main control board controls are personnel enrolled in a license training program.- Individuals listed in responses A, B and D would have an interest in starting the pump but could not be authorized to start it from the main control board.
RO Outline #95 l
..-. ~
-., a
1-i QUESTION #5 The plant is in MODE 6 with core reload in progress. The intermediate range nuclear l
instruments read as follow:
N35 N36 1E-11 amps On the low stop Which ONE of the following conditions is the status of the Intermediate Range Nuclear.
instruments?
A.
N35 is operable. The idling current is sei at 1E-11 amps.
B.
N36 is operable. There is insufficient fuel to create a power level.
C.
N35 and N36 are operable due to being within their channel check.
D.
N35 and N36 are inoperable due to being outside their channel check.
ANSWER:
A. N35 is operable. The idling current is set at 1E-11 amps.
K/A #: 015A201 3.5/3.9 KA DESCRIPTION: Idling Current to IR NI OBJECTIVE #:
0110280B
REFERENCES:
T61.0110.6 LP 28 Page 15 AUTHOR: PJM
, ~
SOURCEn4EW-L DISTRACTER EXPLANATION:
The design of the intermediate range nuclear instruments has an applied idling current adjusted to 1E 11 amps to keep the meters from being pegged low when power levelis below the scale.
B.
The IR should not rest on the low peg due idling current.
C.
Both should read the same due to idling current.
D.
Only one is inoperable due to idling current.
RO Outline #51
i m.
- QUESTION #6 Th' plant is operating at 100% power with all equipment in its normal lineup. A spurious SI e
. occurs during some l&C testing. All equipment functions as designed.
1Which ONE of the'following procedures will the CRS transition to upon completion of the Emergency Procedure guidance?
A.
OTG-ZZ-00001, PlantHeatup. Cold Shutdown to Hot Standby B.-
OTG-ZZ-0001 A, Shutdown Bank Withdrawal C.
OTG-ZZ-00002, Reactor Startup D.
OTG-ZZ-00005, Plant Shutdown 20% Power to Hot Standby ANSWER:
D. OTG-ZZ-00005, Plant Shutdown 20% Power to Hot Standby K/A #: E02EA2.2 3.5/4.0 KA DESCRIPTION: Procedural Guidance Following St Termination OBJECTIVE #:
003D090G
REFERENCES:
. ES 1,1 AUTHOR: PJM SOURCE: NEW - L DISTRACTER EXPLANATION:
~ U porra -R eactor-Tripand13afet3rinjectiorrE:0willtemnteredrES:tTwillte transitioned to from there.~-
~
Upon completion of ES-1.1, the guidance is to perform OTG-ZZ-00005 or 8.
A. Incorrect B.. Incorrect C. Incorrect D. Correct RO Outline #87
~
i
)
QUESTION #7 The following plant conditions exist at 0400:
. T avg 557'F
= PZR Press 2235 psig
. Source Range Channel 31 indicates 19 cps
. Source Range Channel 32 indicates 20 cps The following plant conditions exist at 0409:
. Tavg 556*F
. PZR Press 2240 psig
. Source Range Channel 31 indicates 39 cps
. Source Range Channel 32 indicates 46 cps Which ONE of the following includes a plant response to the above conditions?
A.
The "B" CCP suction valve from the RWST, BNLCV112E, opens.
B.
The Containment Evacuation Alarm sounds.
C.
The High Flux at Shutdown Alarm sounds.
D.
The reactor trip breakers open on high source range counts.
ANSWER:
_. A..The "B" CCP suction valve from the RWST, BNLCV112E, opens.
K/A #: 004K107 2.6/2.9 KA DESCRIPTION: Plant Response to Flux Doubling OBJECTIVE #:
0110280A
REFERENCES:
T61.0110.6 LP 28 AUTHOR: RBM SOURCE: BANK Modified Y - HO DISTRACTER EXPLANATION:
All 4 responses are associated with actions associated with the NIS source range instruments.
Responses B and C have a setpoint of 5 times background and D has a setpoint of 10E5 cps.
RO Outline #3
QUESTION #8 Which ONE of the following describes the Flow Control Valves for the Motor Driven Au Feedwater Pumps?
A.
The valves are motor operated and throttle close to limit runout flow under all seconda side pressure conditions.
. B.. _The valves are motor operated and throttle close automatically.at high flow rates to limit _
. containment pressure increase caused by a steam line rupture in containment.
C.
The valves are air operated and throttle close to limit runout flow under all secondary sid pressure conditions.
D, The valves are air operated and throttle close automatically at high flow rates to limit containment pressure increase caused by a steam line rupture in containment.
ANSWER:
A. The vc!ves are motor operated and throttle close to limit runout flow under all secondary side pressure conditions.
K/A #: 061K404 3.1/3.4 KA DESCRIPTION: Flow Control Valve Operation OBJECTIVE #:
0110250C
REFERENCES:
T61.0110.6 LP 25, Page 12 AUTHOR: PJM SOURCE: BANK Modified N - L DISTRACTER EXPLANATION:
A. The motor operated valves throttle automatically to limit pump runout and cavitation.
B. The orifice are the ones that limit containment pressure increase.
C. The turbine driven aux feed pump has air operated valves.
D. The turbine driven aux feed pump has air operated valves and the orifice are the ones that limit containment pressure increase.
RO Outline #44
L QUESTION #9 i.
- Fuel handling operations are in progres~s to support refueling. Both Fuel Building Rsdiation Monitors, GG RE-27 and GG_ RE 28 alarm, resulting in a Fuel Building Ventilation Isolation Signal (FBVIS). Which of the following would be the correct immediate actions to be taken in response to this event?
A; Store any fuel assembly in transfer in a safe location.
i
__ _ Ensure at least one personnel airlock. door is closed. _ _
l. _._. Ensure.the Fuel Building roll-up. door is closed._.. ~. _.
B.
Evaluate the need to evacuate the Fuel Building.
Ensure at least one personnel airlock door is closed.
Ensure the Fuel Building roll-up door is closed, i
C.
Evaluate the need to evacuate the Fuel Building.
i Store any fuel assembly in transfer in a safe location.
Ensure at least one personnel airlock door is closed.
D.
Evaluate the need to evacuate the Fuel Building.
- Store any fuel assembly in transfer in a safe location.
Ensuro tho Fuo! Building rolleup door is closed.
ANSWER:
C. Evaluate the need to evacuate the Fuel Building.
Store any fuel assembly in transfer in a safe location.
Ensure at least one personnel airlock door is closed.
K/A #d 2.32__3.5/3.3 KA DESCRIPTION: Immediate Actions /OTO-KE-00001 OBJECTIVE #:
003E0501
REFERENCES:
T61.003E.6 LP 5 OTO-KE-00001 AUTHOR: RBM SOURCE: NEW - HO DISTRACTER EXPLANATION:
These are all actions which would be taken in response to a fuel handling accident. All are immediate actions to be taken LAW OTO-KE-00001 except for the need to ensure the Fuel Building roll-up door is closed. That requirement is only a subsequent action.
- RO Outline #89
QUESTION #10 The crew is responding to a plant transient and are currently in procedure ECA 1.2, ?LOCA
'Outside Containment *.-
Why should operators wait some amount of time during each valve repositioning per this procedure?
A, i Prevents overcurrent trips on valve mot.or breakers.
B.-
Allows system pressure to respond to repositioning.
' C.
Ensures valve strokes completely prior to attempiing movement.
D.
- To allow check on indications of leak in auxiliary building.
ANSWER:
B. Allows system pressure to respond to repositioning.
K/A #: E04EK1.2 3.7/4.0 KA DESCRIPTION: Precaution During Valve Strokes in ECA-1.2 OBJECTIVE #:
003D140D
REFERENCES:
T61.003D.6 LP 14 AUTHOR: FXB SOURCE: BANK - Modified N - L
-OISTRACTER-EXPLANATION:
A.
Breakers overcurrent trips are jumpered C.-
Procedure states to observe pressure after valve is closed D. -
No remote indication required, but note on page 2 has personnel searching RO Outline #86.
l
a
- s-
~~4.~..
~
[
i
- QUESTION #11--
RCS activity is 50 microcuries per gram dose equivalent 1 131.
Which ONE of the following is the Chemistry sampling requirements per OTO-BB-00005?
A.
Normat 3 day sample requirements are necessary.
B.
_Once. per day until activity decreases fo.r. 3, consecutive days.
,---..: a
~
j C.
.Once per hour until activity decreases less than Tech Spec limit.
i
' D.-
' Once per two hours until activity decreases on 3 consecutive ssmples.
[
ANSWER:
L
.e D. Once per two hours until activity decreases on 3 consecutive samples.
.i.
K/A #: 076AA2.02 2.8/3.4 KA DESCRIPTION:.-High RCS A'ctivity Sampling Requirements-OBJECTIVE #:
003B180A
REFERENCES:
OTO-BB 00005 Page 2 Tech Spec 3,4.8 i
AUTHOR: PJM SOURCE: NEW - L DISTRACTER EXPLANATION:
A. RCS activity is above the Tech Spec limit. Increased sampling is required,
[
B._ Time requirements are_ wrong.
C. Time requirements are wrong.
p D. Con ect.
. RO Outline #65 1-Yt'.
I 1
4
-t
9 e
QUESTION #12-Eh Which one of the following instructional communications between the Reactor Operator and the i
i Primary Operator would be in accordance with the'Callaway Plant Policy on verbal
- communications?
L
-'A Open the CCW heat exchanger A tube side drain, EGV0021.
Bc iOpen the CCW heat exchanger alpha tube side drain EGV0021
.s_
i
~ _ _ _ _. _ _ _. _..
C.
- Primary Operator, open the CCW heat exchanger A tube side drain, EGV0021.
D.
Primary Operator, open the CCW heat exchanger alpha tube side drain, EGV0021.
ANSWER;
- D. Primary Operator, open the CCW heat exchanger alpha tube side drain, EGV0021.
K/A #: 2.1.17 3.5/3.6 KA DESCRIPTION: - Verbal Communication Policy.
REFERENCES:
UEND-Comm-01 AUTHOR: RBM SOURCE: NEW - L DISTRACTER EXPLANATION:
Communications must be directed and the phonetic alphabet used for train related equipment, Response A is not directed and has no phonetics. Response B is not directed. Response C does not use phonetics.- Only Response D is directed and uses phonetics.
RO Outline #400
~ _ _ _ _ _ _
._ _.._. __ _ _ _.~. _ _ _ _ _... -
J 4-
\\
- OUESTION #13 -
'\\
i LThe plant is operating at 100% Power; An inadvertent Safety injection occurs when an l&C i
- technician is performing a surveillance. WHICH ONE of the following _will open and remain open until recovery from the SI is commenced?
1 A.
RCP DISCH PZR SPRY LINE ISO PCV, BBPCV04558 and BBPCV0455C.
B.
L CVCS REGEN HX TO PZR AUX SPRAY ISO HV, BGHV8145.
..~ - -,.
C.
CVCS REGEN HX TO LOOP 4 COLD LEG ISO Hh, BbHV8147.
< D.
CVCS LTDN ORIFICE B OUT ISO HV, BGHV81498.
ANSWER:
I d
K/A #: 078K302: 3.4/3.6 KA DESCRIPTION: Valve fait position on loss of air 1
OBJECTIVE #: - 0110110M
REFERENCES:
OTO-KA-00001, Att 5 Page 1 I
AUTHOR: PJM SOURCE: NEW-HO i
DISTRACTER EXPLANATION:
L A1 Fail close on loss of air.
j; B.: Fail close pn loss of air.
F_w C,. ASIS-willcause4CIS ArWhen the.CIS-A-occurs the-InstrurecMir-to-Containment-Isolation-valve e
(
KAPV0029, receives a close signal. This results in a loss of instrument air to containment. The fail r
position of BGHV8147 is open. The fait position of all the other valves is closed, D. Fail close on loss of air.~
[
_ RO Outline #1 e -
s QUESTION #14 Which ONE of the following describes the containment atmosphere radiation monitors GT RE 31 and GT RE 327 A.
They sample containment via the hydrogen control symm and are isolated from containment by a glS A actuation.
B.
They cample upstream'of the containment isolation valves for the hydrogen control system and are NOT isolated by a CIS A actuation.
C.
They sample between the containment isolation valves on the mini purge exhaust line and initiate a CPIS on high high activity.
D.
They sample from the containment purge exhaust line outside containment and initiate a CPIS on high high activity,
- ANSWER:
A.- They sample containment via the hydrogen control system and are isolated from l
containment by a CIS A actuadon.
K/A #: 103A301 3.9/4.2 I
KA DESCRIPTION: Rad Monitor Response to CISA OBJECTIVE #:
0110400K
REFERENCES:
M22GS01 AUTHOR: PJM SOURCE: BANK Modified N L DISTRACTER EXPLANATION:
A. GT RE 31 and GT RE-32 are physically connected to draw containment air from the hydrogen control system. The isolation valves in these lines receives a signal to close on a CIS A. There are no isolation signals generated from these two radiation monitors.
B. They sample through the iso valves and are isolated.
C. They do not sample from the purge system.
D. They do not sample from the purge system.
RO Outline #31
4 QUESTION #15 A loss of condenser vacuum is occurring due to unknown reasons, and power has been reduced from 100% to 75% over the last 5 minutes.
The following conditions exist:
e Auct High Tavg 593'F e Reactor / Turbine Power 75% / 775 Mwe
. LP 'A' Condenser Pressure 5.8 Hga e LP 'B' Condenser Pressure 6.2 Hga 4 LP 'C' Condenser Pressure
_6.5 Hga Which ONE of the below describes the expected operation of the condenser steam dumps with these conditions:
A.-
Less than 12 steam dumps are available and all available dumps are FULLY OPEN.-
B.
ALL 12 condenser steam dumps are available and all are FULLY OPEN.
C.
Less than 12 steam dumps are available and all available dumps are PARTIALLY OPEN.
D.
Less than 12 steam dumps are available and all are CLOSED, ANSWER:
A. Less than 12 steam dumps are availrble and all available dumps are FULLY OPEN.
K/A #: 051AK3,01 2.8/3.1 i
KA DESCRIPTION: Loss of Steam Dumps with Loss of Vacuum OBJECTIVE #:
00380 BOA 01B
REFERENCES:
OTO-AD-00001 7250D64 Sheet 10 AUTHOR: FXB SOURCE: NEW HO DISTRACTER EXPLANATION:
B. Condenser B, C greater than 6 in Hga.
C. Dumps avai!able are fully open.
D. Dumps available are not closed.
RO Outline #71 UKI
4 l
OUESTION #16 l
The plant is at 8% power preparing to synchronize the main generator to the grid when the running main feed pump trips. As the Balance of Pital operator you observe the following steam generator narrow range levels:
A Indicates 16%
B indicates 15%
C Indicates 14%
D Indicates 17%
Which one of the following correctly describes the status of the Auxillary Feedwater (AFW)
System?
A.
No AFW pumps are running I
B.
Only the turbine driven AFW pump is running 4
- C.
Only the motor driven AFW pumps are running D.
All the AFW pumps are running ANSWER:
i C. Only the motor driven AFW pumt:s are running K/A #: 054AA2.03 4.1/41 KA DESCRIPTION: AFW PP Start Signals OBJECTIVE #: - 0110250H, -
=
REFERENCES:
T61.0110.6 LP 25 OTO SA-00001 AUTHOR: RBM SOURCE: NEW - HO DISTRACTER EXPLANATION:
All answers are plausible, depending on whether a MDAFAS or TDAFAS signal has been actuated. With the given conditions, only a MDAFAS signal would be generated, therefore C is the correct answer.
RO Outline #84
QUESTION #17 The followin0 conditions exist:
1
- PRZR Relief Tank Level Hi/Lo - ALARMING on HIGH LEVEL
. PRZR Relief Tank Pressure
-- ALARMING on HIGH PRESSURE Which ONE of the below combinations contains possible sources that should be monilored for leakage.into the PRT7 A.
RHR Pump Suction Reliefs (EJ8708A/B), RCP Seal Leakoff Relief (BG8121), and CVCS Letdown Relief (BG8117).
B.
ECCS Accumulator Reliefs (8855A-D), RHR Pump Suction Relief (tiJ8708A/B), and CVCS Letdown Relief (BG8117).
C.
RCP Seal Leakoff Rollef(BG8121), CVCS Letdown Relief (BG8117) and RHR Discharge Reliefs (EJ885GA/B).
D.
Safety injection Pump Suction Reliefs (EM8858A), RHR Pump Suction Reliefs (EJ8708A/B), and RCP Seal Leakoff Rollef (BG8121).
ANSWER:
A. RHR Pump Suction Reliefs (EJ8708A/B), RCP Seal Leakoff Relief (BG8121), and CVCS Letdown Relief (BG8117).
K/A #: 007A205 - 3.2/3.0
.KA DESCRIPTION: Impact of Pressure-t-on PRT----- ---
OBJECTIVE #:
01100901
REFERENCES:
M22BB02 AUTHOR: FXB SOURCE: NEW HO DISTRACTER EXPLANATION:
B.
ECCS Reliefs go to atmosphere C-RHR Discharge Reliefs go to RHUT D-Safety injection Suction Relief to RHUT RO Outline #38
QUESTION #18 Per ODP ZZ-00310 WPA Tagging, which one of the following situations would require that a new tagout be generated?
A.
Tags need to be removed to perform rotation checks following maintenance.
B.
There is a need for a sixth partial on an existing tagout.
=.......-.
C.-.. Monthly audit identifies missing tags on an existing tagout.
D.
Additional tags are needed after maintenance has boon started on a component.
ANSWER:
D. There is a need for a sixth partial on an existing tagout.
K/A #: 2.2.13 3.0/3.8 KA DESCRIPTION: WPA Partial Clearing OBJECTIVE #:
003A330B
REFERENCES:
T61.003A.6 LP 33 ODP ZZ 00310 AUTHOR: RBM SOURCE: NEW L DISTRACTER EXPLANATION:
A would use the temporary lift option, not a new tagout.
C would result in new tags being generated, not a new tagout.
D would result in new tags being added to the existing tagout, not a new tagout.
RO Outline #99
1 l
QUESTION #19 The electric fire pump has started due to a rupture in the main fire header going into the Radwaste Building. When can the pump be secured and returned to standby?
A.
Anytime from Panel KC008 in the main control room.
B.
Anytime from the local control panel.
ZZ6. I Noril PanefRC508'inlhe AEnio'n't'rol' room aiie'r~syster5'p5siurehas~bden restored.
~ ~
D.
From the locs cNdd panel after system pressure has been restored.
ANSWER:
D. From the local control panel after system pressure has been restored.
K/A #: -086A201 2.9/3.1 KA DESCRIPTION: Fire Protection System. Manual S/D of Firepumps OBJECTIVE #:
0110350 0
REFERENCES:
T61.0110.6 LP 35, Page 78 AUTHOR: RBM SOURCE: BANK Modified Y L DISTRACTER EXPLANATION:
Responses A and C are incorrect because the fire system pumps cannot be secured from the main control room anytime, in order for the system to be restored to normal (standby), system pressure must l
be restored to normal, thus Response D is the only correct choice.
RO. Outline.#36__.___ _,_
l
QUESTION #20 Callaway Plaat is in Mode 4. "B" RHR is in service. A plant cooldown is in progress. The Reactor Operator is directed to stop the cooldown. EGHV102, *B" CCW to *B" RHR heat exchanger is CLOSED.
Which ONE of the following events occur?
A.
B.
- B" ESW flashes in the *B' CCW heat exchanger causing water hammer in the *B" ESW.
C.
D.
Service Watei 9N a in the 'B' CCW heat exchanger causing water hammer in the Service Water g ^ an. -
ANSWER:
A. "B" CCW flashes in the *B" RHR heat exchanger causing the *B" CCW surge tank level to increase.
K'A #: 005A103 2.5/2.6 KA DESCRIPTION: Flashing in RHR HX OBJECTIVE #:
0110100E
REFERENCES:
' SOS 86 0054 AUTHOR: PJM SOURCE: BANK Modified Y - HO DISTRACTER EXPT.ANATION:
A. When CCW flow was stopped in the plant to the RHR heat exchanger, CCW flashed to steam in the RHR heat exchanger upon heating up.
B. ESW cools CCW. ESWis at a higher pressure so should not flash.
C. O RHR is pressurized so should not flash.
D. Service water cools CCW. Service water is at a higher pressure so should not flash.
RO Outline #27
l a
4 OUESTION #21 Plant conditions are as follows:
- Mode i 100% Reactor Power e
e All required Tech Spec and FSAR equipment is operable.
. An inadvertent CSAS occurs.
--Which ONE of the following sets of valves will OPEN due to the CSAS7 -
A.
RWST Isolation Valves, DNHV3 and BNHV4.
B.
Containment Sump Isolation Valves ENHV1 and ENHV7, i
i
~ ~ C.
Containment Spray Pump Discharge Valves ENHV6 and ENHV12.
D.
Containment Spray Add Eductor Outlet Valves ENV5 and ENV11.
ANSWER:
C. Containment Spray Pump Discharge Valves ENHV6 and ENHV12.
K/A #: 026A301 4.3/4.5 VA DESCRIPTION: Plant Response to CSAS OBJECTIVE #:
0110180H
REFERENCES:
M22BN01 M22EN01.
g p
SOURCE: BANK Modified N HO DISTRACTER EXPLANATION:
When a CSAS occurs, BNHV3, BNHV4, ENHV6, and ENHV12 receive open signals.
A. Bl4HV3 and BNHV4 are nonnally open valves. They will not reposition.
B. ENHV1 and ENHV7 are operator opened only.
C. ENHV6 and ENHV12 are normally closed and will open due to the CSAS.
D. ENVS and ENV11 are normally locked open valves.
- RO Outline #17.
7
l 1
t QUESTION #22 The plant is operating at 100% power. The controlling PZR pressure transmitter, PT-455, fails high. As a result of this failure, HOW will the following PZR components initially respond?
(Select ONE of the following)
PZR PORVs SPRAY VALVES PZR B/U HEATERS
_.._Am_ Bot!Lw.ill OP_EN____Will CLOSE.
_.W.ill_Deenergizo...__
B.
One will OPEN -
Will OPEN Will Deenergizo C.
One will OPEN Will CLOSE Will. Energize D.
Neither will OPEN Will OPEN Will Doenergize i
ANSWER:
i
-l B. One will OPEN Will OPEN Will Deenergize K/A #: 010K601 2.7/3.1 KA DESCRIPTION: Response to Pzr Press Channel Failure OBJECTIVE #:
0110300K i
REFERENCES:
T61.0110.6 LP 30 Pages 22 24 8756D37 Sheet 26 AUTHOR: PJM
.SOURCErNEWrHO-
- ~ - - - ~~-
DISTRACTER EXPLANATION:
A. Under different plant conditions, this could occur.
B. The controlling PZR press channelis the upper selected channel on the PZR press control switch.
This channel will also feed PORV 455A PORV 456 is controlled from the lower selected channel. When PT455 fails high, PORV 455A responds by opening while PORV 456 is not effected.
3 The spray valves and the PZR backup heaters are controlled from the master PZR pressure controller.
The pressure input to the master is from the upper channel on the PZR press control switch. When PT455 fails high, the spray valves open to reduce pressure while the backup heaters turn off to reduce -
heat input and lower pressure.
C&D Under different plant conditions this could occur.
RO Outline #9
?
1 QUESTION #23 FR H.1, Response to Loss of Secondary Heat Sink has just been entered. Plant conditions are as follows:
. - Total available AFW flow 200,000 lbm/hr
._ A,B,0 SIG levels 2% NR
- C S/G level-20% WR
. A,B,0 S/G pressures 950 psig
. O SIG pressure - completely depressurized-
~ ~~ ~ ~ ~~ Containment temperatureT190'F ~ ^
~~ ~
~
~
~
. Containment activity - SE 2 microcuries per gram
. RCS pressure - 900 psig
. PZR level-off scale low
. RCP's - secured Which ONE of the following actions should be taken in accordance with FR-H.1 ?
I A.
Immediately initiate steps 10 through 16, RCS bleed and feed i
i B.
-Transition to E-1, Loss of Reactor or Secondary Coolant C.
Transition to E 2, Faulted Steam Generator Isolation D.
Return to procedure and step in effect when S/G level > 4%
ANSWER:
B. Transition to E-1, Loss of Reactor or Secondary Coolant K/A #: E05E'K2.2 3.9/4.2 KA DESCRIPTION: Transition from FR H.1 to E 1 OBJECTIVE #:
003D200D
REFERENCES:
FR H.1 Page 2 AUTHOR: PJM SOURCE: BANK - Modified N - HO DISTRACTER EXPLANATION:
A. These are part of the procedure but conditions are not requiring this:
B. With S/G pressure less than RCS pressure, Heat Sink is not required. Transition to E 1 is ne to deal with RCS problems.
C, Conditions are met for Faulted Steam Generator but that should have been already dealt with or w be dealt with when the plant conditions are right.
D.' This is a transition when feedwater is restored, but with the containment temperature of 190'F, adverse containment conditions exist changing the requirements to 35%
RO Outline #78
QUESTION #24 Which ONE of the following electricallineups is a normallineup for the Switchyard?
A.
ESF XFMR XNB01 energized by Safeguards XFMR *B" via ring bus breaker 52-3.
B.
MTGY-CAL-7 transmission line energized by the main step-up XFMR's via switchyard j
breaker V53.
l C.
- CAL-BLAN 1 transmission line energized by 345 KV SWYD Bus *A* via switchyard breaker V41.
D.
200 series site load XFMR's energized by Safeguards XFMR *A* via ring bus breaker 52 4.
ANSWER:-
- - A.- ESF XFMR XNB01 energized by Safeguards XFMR *B" via ring bus breaker 52-3.
K/A #: 062K104 3.7/4.2 KA DESCRIPTION: Offsite AC Power Sources OBJECTIVE #:
0110010B i
REFERENCES:
8618 X 94080 i
AUTHOR: PJM SOURCE: BANK Modified Y L DISTRACTER EXPLANATION:
A. The enly physical lineup which can be made in the switchyard is supplying the ESF transformer with ring bus breaker 52 3 from the B Safeguards transformer.
B. MTGY-CAL-7 goes through V85.
C. CAL-BLAN-1 goes through V43 and V41 to get to A bus.
D. 200 series from XFMR A through 52 2.
RO Outline #15 m
4 QUESTION #25 l
A reactor coolant system leak develops in the 'C' RCP thermal barrier. The CCW system is desi ned for this failure by isolating the thermal barriers on HIGH CCW _
0 i
I A.
Flow B.
Radiation
__._.C._
Pressure.
D.
Tank Level ANSWER:
- A;- Flow K/A #: 003K112 3.0/3.3 l
KA DESCRIPTION: RCS Leak Thermal Barrier Isolation OBJECTIVE #:
003B160A
REFERENCES:
T61.0110.6 LP 9 AUTHOR: PJM SOURCE: NEW L DISTRACTER EXPLANATION:
B.
Alarm and Isolates makeup and Surge Tank Vent C.
Designed for high pressure D.
High levelisolates makeup RO.OutlineJtSO hui
n QUESTION #26 The plant has experienced a loss of offsite power All equipment functioned as designed. Five minutes later, NB0111, *A" D/G output breaker, trips open. Which ONE of the following caused NB0111 to open?
A.
High lube oil temperature B._ _O_yercu_rrent _,_ _,,_. _ _, __ _, __, _.... _. _ _. _..,, __ _, _
C.
LowJacket water pressure D.
- ANSWER: -
1
- D. Ovsispeed K/A #: 064K402 3.9/4.2 KA DESCRIPTION: D/G Output Breaker Trips OBJECTIVE #: - 011003OH
REFERENCES:
T61.0110.6 LP 3, Pages 59 & 65 AUTHOR: PJM SOURCE: BANK Modified Y - HO.
DISTRACTER EXPLANATION:
A. High lube oil temperature does not cause an Engine Shutdown Relay trip.
B._Overcurrent. trips.thet/G outputbreaker oni normal. start but istlocked onan emergency-start.-
C. LowJacket water pressure does not cause an Engine Shutdown Relay trip.
D. Overspeed causes and Engine Shutdown Relay actuation. This trips the D/G breaker open on any type of start.
RO Outline #16
4 4
-QUESTION #27
- NCP is running,120 gpm letdown e BGFT121, CVCS CHG HDR TO REGEN HX FLOW XMTR, fails, e As a result, BGFCV124 closes, e The NCP handswitch red light is lit
. BGHV8109, NCP Recirculation Valve, is open.
The following annunciators are received:
. CHARGING LINE FLOW LOW e SEAL INJECTION TO RCP FLOW LOW e NCP FLOW LOW Which ONE of the following actions should be taken immediately?
A.
Take manual control of BGFCV124 and open it.
B.
Start a CCP and secure t(ie NCP, C, _
Open the sealinjection target rocks D,
Isolate allletdown ANSWER:
A. Take manual control of BGFCV124 and open it.
K/A #: 022AK3,02 3.5/3,8 KA DESCRIPTION: Valve Closure in Charging Line OBJECTIVE #:
003B22OA
REFERENCES:
OTO BG-00002 Page 2 AUTHOR: PJM SOURCE:- NEW HO 1
'DiSTRACTER EXPLANATION:
A, Correct Since the NCP is not a problem, then the next immediate action required by OTO BG-00002 is to open any valves that have closed.
B This would be the correct action if the NCP had failed, C This would work on a CCP but the NCP has no physical connection to BGHV8353, Sealinjection, D. Isolating letdown is a subsequent action if charging cannot be restored expeditiously.
RO Outline #85
t OUESTION #28 I
Which ONE of the below NK01 bus electricallineups meets the requirements for continuous ooeration while in MODE 17 NORM CHARGER ALT CHARGER DATTERY NV21 NK25 NK11
.. _ A.
Connected Disconnected
_ _ _ Disc _onnected C.
Disconnected Connected from PG bus Connected D.
Disconnected Disconnected Connected ANSWER:
B.
Disconnected Connected from NG bus Connected K/A #: 063A401 2.8/3.1 KA DESCRIPTION: Monitor NK Lineup in the Control Room OBJECTIVE #:
0110060G
REFERENCES:
OTN.NK 00001 AUTHOR: FXB SOURCE: NEW - L DISTRACTER' EXPLANATION:
A.
Battery Disconnected C.
NK25 powered from PG bus D.
No charger connected RO Outline #40
l QUESTION #29 l
Compare the output of the roactor control unit, with rod control in auto, for the two conditions l
I listed below;
- A 2% stop chango in turbine load at 90%.
. A 2% stop change in turbino load at 40%.
Which.Ot4E of_the following describes the comparison of the. signal output and tho.teason_for.. _
__.the differenco?..
A.
Larger at 90% due to the responso of the Variable gain unit.
B.
Larger at 90% due to the responso of the Non Linear gain unit.
C.
Smaller at 90% due to the responso of the Non Linear gain unit.
D.
Smaller at 90% due to the responso of the Vorlable gain unit.
ANSWER:
D. Smaller at 90% due to the response of the Variable gain unit.
K/A #: 001K565 3.2/3.6 KA DESCRIPTION: Power Mismatch Effect on Rod Control OBJECTIVE #:
0110260B
REFERENCES:
T01.0110.0 LP 26, Page 17
' AUTHOR: PJM
~ SOURCE:~ NEWTHO
~ ~ - ~ -
~~~~
DISTRACTER EXPLANATION:
A. The variatde gain unit inserts a high gain at low power levels and a low galn at power levels above 50%.
B&C. The nonlinear gain unit causes larger load changes to have a larger effect.
D. Correct RO Outline #7
UUESTION #30 The plant is operating at 100% power. l&C has dolormined the *A* train ESFAS has a bad power supply that has to be replaced. The *A* train ESFAS will be downpoworod using OTS SA-00001.
Which ONE of tho tollowing actions provent an inadvertent actuation from tho *B" ESFAS train?
_ _. A Tho cross trip _ block switch is placed in the BLOCK position.
~
~
B.
The assoelated *B" bistables are placed in the BYPASS position.
C.
The dual voltage power supply,15 vdc/ 48 vde, is placed in the OFF position.
D.
The 48 vde power supply is placed in the OFF position.
u ANSWER:
A. The cross trip block switch is placed in the BLOCK position.
K/A #: 013K201 3.6/3.8 KA DESCRIPTION: Downpower Cross Trip Block OBJECTIVE #:
0110520C
~
REFERENCES:
J.104 00176 AUTHOR: PJM l-SOURCE: BANK Modified N - HO DISTRACTER EXPLANATION:
A. The cross trips are blocked with the cross trip block switch.
B. Bypassing a bistable makes it inoperable. With the other train downpowered, there would be no operable bistables.
C&D. Turning off a power supply will trip its associated bistables.
RO Outline #33
l l
QUESTION #31 E-0,
- Reactor Trip /Sofety injection", defines RCP trip critoria. When stopping all RCP's in a LOCA condition, E-0 directs that steam dumps be operated in tho " steam pressuro* modo.
Which one of the following is also req'Jired to be verified when stopping all RCP's por E-07 A.
VSIV's open.
Z1TTBotE~56Pi,ln 5peidlio'A~ ~
~
^
~~
~
C.
Either CCP or Si in operation.
D.
AFW flow >300 KIbm/br.
ANSWER:-
C. Elther CCP or SIin operation.
K/A #: 2.4.20 3.3/4.0 KA DESCRIPTION: RCP Trip Criteria OBJECTIVE #:
003D020A
REFERENCES:
E 0 Foldout Page AUTHOR: FXB SOURCE: NEW L DISTRACTER EXPLANATION:
A, P.equired for steam dump operation.
8,. Either CCP, -
D. Required for secondary heat sink.
RO Outline #92
QUESTION #32 Which one of the following posted areas must be deposted beforo plant personnel may bo permitted to enter?
l I
A.
CHRA, Caution High Radiation Area l
B.
DHRA, Dangor High Radiation Area i
TZC<.
DREA, Danger High Radiation Area Radiological Exclusion Area D.
VHRA, Very High Radiation Area ANSWER:
D, VHRA,- Very High Radiation Area -
K/A #: 2,3,2 2.5/2.9 KA DESCRIPTION: Radiological Posting Requirements OBJECTIVE #;
003A310D 003A310F
~
AUTHOR: RBM SOURCE: NEW L DISTRACTER EXPLANATION:
All of the areas listed have special entry requirements The entry requirement for a VHRA is that it must.
be deposted before entry can be authorized.
RO Outline #96 h4.
i QUESTION #33 Which ONE of the following statements correctly describes the purpose of the interlocks associated with the RHR suction isolation valves (EJHV8701 A/B and BBPV8702A/B) while placing RHR in the shutdown cooling rnode?
A.
Prevent draining the RWST to the containment sump.
~ B.,__Provent overpressurization of the RHR system.
l C.
Prevents exceeding RHR system design temperature.
D.
Prevent contamination of the RWST, ANSWER:
B. Prevent overpressurization of the RHR system.
K/A #: 005K407 3.2/3.5 KA DESCRIPTION: RHR Valve Interlocks OBJECTIVE #;
01100708
REFERENCES:
T61.0110.6 LP 7, Pages 10 and 11 M22EJ01 AUTHOR: PJM SOURCE: NEW 1.
DISTRACTER EXPLANATION:
A. BNHVBB12A/B and EJHV8811 A/B are interlocked to prevent draining the RWST to the containment
..recirc sump._._. _._ _
B. The interlocks associated with BBPV8702A/B and EJHV8701A/B are to prevent overpressurization of the RHR and depressurization of the RCS.
C. RCS Temperature is high but not a design problem.
D. EJHV8BD4 and EMHV14/13 are interlocked to prevent contamination of the RWST.
RO Outline #11
i QUESTION L. 4
\\
' : Plant don'ditions are as follows:
- At 0700, a loss of all offsite power occurred
- At 0710, a Sl occurred due to stuck open S/G Safety
. All equipment operated as designed e Restoration of instr' ment air to Containment is in progress.
. EFHV43 and EFHV44, ESW to 'A' and 'B' air compressors are closed.
Which ONE of the following has caused EFHV43 and EFHV44 to close?
A, Blackout load shed i
B.
LOCA load shed C.
Loss ofinstrument air
_ D.
High DIP ~
ANSWER:
C. Loss of instrument alt K/A #; 065AK3.08 3.7/3,9 KA DESCRIPTION: E-0 Actions for Loss of Air to EFHV43/44 OBJECTIVE #;
003D04EE
REFERENCES:
~E-0 Page 31 AUTHOR: PJM SOURCE: NEW L OlSTRACTER EXPl.ANATION:
A. A Blackout Icad shed occurs due to loss of offsite power. EFHV43/44 are not shed.
B, A 1.OCA load shed occurs due to Sl. EFHV43/44 are not shed.
C, Correct. When all the instrument air compressors stop, instrument air pressure decreases until EFHV43/44 fall close.
D. EFHV43/44 do close on high D/P but no such conditions exist such as a pipe break.
RO Outline #58
QUESTION #35 A precaution and limitation in OTN.EG 00001, Component Cooling Water System, informs the operator that EGHV0069A/B (EG HS 60) and EGHV0070A/B (EG HS 70), CCW Supply / Return to Radwaste, must be opened simultaneously. The reason for this requirement is to:
A.
Satisfy the system high flow interlock.
B.
Satisfy the system low flow interlock.
C.
Minimize potential of system water hammers.
D.
Ensure proper flow is maintained to containment system loads.
ANSWER:
A. Satisfy the system high flow interlock.
K/A #: 2.1.32 3.4/3.8 KA DESCRIPTION: Precautions and Limitations for Radwaste Supply OBJECTIVE #:
003A000C
REFERENCES:
T61.003A.6 LP 9 OTN EG 00001 4
AUTHOR: RBM SOURCE: - NEW L DISTRACTER EXPLANATION:
These valves have an auto closure signal on high differential pressure. This signal results from system-high flow versus system low flow. Valve cycling can result in system wr'er hammers and improper flow balances, however this is not the reason for the precaution askedgtit in this qug ion.
RO Outline #94
OUESTION #36 During a surveillance test with the plant at 90% power, an instrument technician removes the instrument power fuses from N42 prior to bypassing channel N42 inputs. Assuming no operator actions the control rods will
_ (Select ONE of the following.)
A.
drive IN at 72 spm then drive OUT as temperature DECREASES.
l
.___ B._... drive OUT.at 72 spm then drive IN as temperature INCREASES. ___
C.
be blocked from moving OUT in AUTO and MANUAL.
D.
be blocked from moving OUT in AUTO only.
ANSWER:
C. be blocked from moving OUT in AUTO and MANUAL.
K/A #: 015K302 3.3/3.5 KA DESCRIPTION: Response to Power Range NI Failure OBJECTIVE #:
0110280E
REFERENCES:
T61.0110.6 LP 28, Pages 20 & 37 OTO SE 00003 7250D64 Sheet 9 AUTHOR: PJM SOURCE: NEW HO DISTRACTER EXPLANATION:
.Ac-if the Niinput-to rod controlis increased rapidly, this would appear that turbine load decreased so-rods would go in.
B, if the NIinput to Rod Control decreased rapidly, this would appear that the turbine load increased so rods would go out.
C. Loss of instrument power occurs when the I&C technician pulls the instrument power fuses. When instrument power is lost, C 2 is activated. Since this is a 1/4 coincidence, this results in blocking outward rod motion in auto and manual.
D. Some blocks are auto only blocks.
RO Outline #8
l OUESTION #37 The following conditions exist:
. %" and *B' CCW Pumps are running.
c i
i e The Service Loop is being supplied by *B' CCW Train.
. 'A' Component Cooling Water Radiation Monitor EGRE0009 has exceeded the Hi Hi ALARM setpoint.
-Which ONE of the following automatic actions occurin addition to receiving an audible ALARM - -
on the RM 117 A.
EGRV0009, CCW SRG TK A VENT CTRL VLV remains OPEN, EGRV0010, CCW SRG TK B VENT CTRL VLV remains OPEN.
~ Bc ~ ~ EGRV0009,-CCW SRG TK A VENT CTRL VLV remains OPEN,~ EGRV0010, CCW SRG
~
TK B VENT CTRL VLV CLOSES.
C.
EGRV0009, CCW SRG TK A VENT CTRL VLV CLOSES, EGRV0010, CCW SRG TK B VENT CTRL VLV remains OPEN.
D.
EGRV0009, CCW SRG TK A VENT CTRL VLV CLOSES, EGRV0010, CCW SRG TK B VENT CTRL VLV CLOSES, ANSWER:
C, EGRV0009, CCW SRG TK A VENT CTRL VLV CLOSES, EGRV0010, CCW SRG TK B VENT CTRL VLV remains OPEN, K/A #: 073K101 3.6/3.9 KA DESCRIPTION: Response to CCW Rad Mon Alarm OBJECTIVE #:
0110100E
REFERENCES:
T61,0110.0 - LP10, Page 13 M22EG01 and M22EG02 AUTHOR: PJM SOURCE:' NEW L DISTRACTER EXPLANATION:
A HiHi radiation does affect the surge tank vent valves.
B. The B vent valve is not affected, the A is.
C. The vent valve and makeup valve for a CCW train close on Hi Hi radiation on its associated train radiation monitor, The Service Loop has no effect on which vent and makeup valve isolate, D. HiHi radiation on one train does not affect the other.
RO Outline #4
OUESTION #38 i
Which ONE of the following will result in a shift of Auxiliary Foodwater (AF) System suction from the Condensate Storage Tank to the Essential Service Water System ?
A.
AFW pump suction pressure of 20 psia coincident with a loss of offsite power.
9 AFW pump suction pressure of 22 psla coincident with Lo Lo levelin ALL steam
_ generators.
' Ci AFW pump suction pressure of 20 psia coincident with Pressuriher pressure of 1850 psig.
I D.
AFW pump suction pressure of 22 psia coincident with a phase B isolation.
ANSWER:
^
A. AFW pump suction pressure of 20 psla coincident with a loss of offsite power.
K/A #: 061K401 3.9/4.2 KA DESCRIPTION: AFAS/LSP Actuation Requirements OBJECTIVE #:
0110250H/Ol
REFERENCES:
T61.0110.6 LP 25 Pages 15,16 and 19 AUTHOR: PJM SOURCE: NEW HO DISTRACTER EXPLANATION:
A. For an automatic swapover from the CST to the ESW, the suction pressure must be less than 21.71 psia with a AFAS present.
lhe. conditions.lbat. generate _anAEAlare;
- S/G lo lo level Undervoltage on NB01 or NB02 AMSAC Trip of both main feed pumps Manual B. AFW suction pressure not low enough for LSP.
C. No signals actuated to cause AFAS.
D. AFW suction pressure not low enough for LSP.
RO Outline #12
t QUESTION #39-Plant conditions are as follows:
100% Rx Power e
588.3'F Tavg Loop 1 588.1'F Tavg Loop 2 e 588.4'F Tavg Loop 3 588,2*ETavg Loop 4_._ _
.. __ __. 5 8 5. 4 'F_Tr e f._._-- -.
Which ONE of the following would occur if the Rod Control Selector Switch were placed in AUTO 7 Rod speed Rod movement direction A.
8 spm IN B.'
48 spm IN i
C.
8spm OUT D,
48 spm OUT ANSWER:
A.
8 spm IN K/A #: 014A402 3.4/3.2 KADESCRIP. TION:_ Selecting Auto RodSpeed OBJECTIVE #:
0110260C
REFERENCES:
T61.0110.6 LP 26, Page 18 AUTHOR: PJM -
SOURCE: NEW HO DISTRACTER EXPLANATION:
Tref - Tavg = - 3*F-3'F on the chart shows an inward rod movement at 8 spm.
RO Outline #42 -
i
L L
QUESTION #40 Plant conditions are as follows:
l e A LOCA outside of containment has occurred, i
e No fluid is collecting in the containment recirc sump.
i e The RWST is empty.
e _ All ECCS pumps have been stopped.
. Which ONE of the following can be used to provide water to the ECCS pumps per.ECA-1.-11 -- -
A.
Condensate Storage Tank
_ B.
Reactor Makeup Water torage Tank
- C; ~ - Spent Fuel Pool-
+ " - - -
~ ~
D.
Waste Collection Tank ANSWER:
B. Reactor Makeup Water Storage Tank K/A #: E11EK1.1 3.7/4.0 KA DESCRIPTION: Makeup Sources of Water on Loss of Recire Sump OBJECTIVE #:
003D13OK
REFERENCES:
ECA 1.1 Page 16 AUTHOR: PJM 700RCt:: NEW HO
~
~
~~
DISTRACTER EXPLANATION:
A.
No physical connection possible.
B. -
RMWST is lined up with a BAT to provide normal makeup to either the VCT or the RWST.
C.
. No physical connection possible.
D.
No physical connection possible.
RO Outline #72
OUESTION #41
_ Given the following conditions:
l
. The plant was operating normally at 100% power
. SIG A narrow range level rapidly decreases to 2%
- S/Gs B, C, and D narrow range levels are at 50%
. The reactor and turbine do NOT trip
. Auxillary Feedwater pumps do NOT start
. An Anticipated Transient Without Trip [ATWS) condition is announced"
~T Control ~ rods ar6 manually inserted' ~ ~ ~ ~
~~
. Power Range Instrumentation is decreasing at 10% per minuto due to rod insertion
}
Which ONE of the following is the expected response of the ATWS Mitigating System (AMSAC)?
A.
AMSAC will automatically trip the reactor which then causes a turbine trip.
B.
AMSAC will trip the turbine and automatically start all AFW pumps.
C.
AMSAC will not actuate since the required S/G low level logic has not been satisfied.
D.
AMSAC is blocked frorn actuation since power level will be less than 30% power before the AMSAC time delay expires.
ANSWER:
C. AMSAC will not actuate since the required S/G low level logic has not been satisfied.
K/A #: 016K112 S.0/3.5 KA DESCR1P.T10N;_ Requirements for AMSAC Actuation OBJECTIVE #:
0110650B
REFERENCES:
T61.0110.6 LP54, Page 1 OTA RL RK083A AUTHOR: PJM SOURCE: NEW HO DISTRACTER EXPl.ANATION:
A. AMSAC trips the turbine not the reactor.
B. AMSAC does trip the turbine and give an AFAS but the required conditions are not met.
C. The following conditions are necessary for AMSAC actuation. Pimp >33% with a 2/2 coiicidence.
This will stay armed for 360 seconds after Pimp decreases below the setpoint. This signal will not go away with the power reduction in progress. Thc S/G level necessary is <5% in 3/4 S/G's. So this requirement is not met, if AMSAC conditions are met, then after 25 seconds, AMSAC trips the turbine and actuates AFAS.
D. AMSAC stays armed for 360 sec following power decrease so the actuation is not blocked.
RO Outline #5
QUESTION #42 Which ONE of the following describes the functional relationship between the Feed Hdr/ Steam
- Hdr AP, FW Reg Valve Position, and the FW Rog Bypass Valve Position, while increasing power from 10 - 20%7 FRV BYPASS Feed Hdr / Steam Hdr AP FW REG VALVE VALVC
!.Z CMainiain~e[c'isTant at 150 psid._ Throttled.OPEN._ Throttled OPEN_- _._ __. __._ _
l o
I B..
Slowly lowered to 90 psid Throttled OPEN
. Throttled CLOSED C.
Maintained constant at 150 psid Remain CLOSED Throttled OPEN
' - - D.
Slowly lowered to 90 psid -
-Throttled OPEN
-Throttled OPEN -
ANSWER:
i D.- Slowly lowered to 90 psid Throttled OPEN Throttled OPEN K/A #: 059A403 - 2.9/2.9 KA DESCRIPTION: FRV/ Feed AP on Powerincrease l
i OBJECTIVE #:
011023OF
REFERENCES:
OTN AE 00001 i
AUTHOR: FXB 100RCE"NEW"Ndifie~d N/ATHO OlSTRACTER EXPLANATION:
Differential pressure is slowly lowered and the bypass FRVs are left open until 30% load.
RO Outline #22
QUESTION #43-The followhg conditions exist:-
. Reactor Power is 60%
. 'A' RCP #1 Seal AP is 170 psid and decreasing 1 psid per minute
. 'A' RCP #1 Seal Leak-off flow is 3 gpm
. 'A' RCP Frame Vibration is 3 mils and increasing 1 mil /hr Which:one of the.following is the proper operator action per.OTO BB-00002, Reactor Coolant Pump Off-Normal?
A..
Reduce power to <4fW end trip the affected RCP.
l L
B.
' Trip the reactor and tu+ine, then trip the affected RCP.
\\
' C.
Trip the affected RCP and be in Mode 3 within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
D.
Continue to monitor vibration on 'A' RCP.
ANSWER:
A. Reduce power to <48% and trip the affected RCP, K/A #: 015/17AA1.23 3.1/3.2 KA DESCRIPTION: Respond to High RCP Vibration OBJECTIVE #:
00381508
REFERENCES:
OTO-BB-00002 SOS 96-1775 AUTHOR: -RBM SOURCE: BANK Modified Y - HO DISTRACTER EXPLANATION:
B would be true if vibration was at 5 mils.
C is plausible in that you wouid have 6 hrs to shut down the plant after securing an RCP except in this case the reactor would trip if the pump was secured due to P-8.
D is plausible if vibrations were steauy.
- RO Outline #56
t
.t
. QUESTION #44 -
The plant is responding to a safety injection and the crew is currently in E-1, Loss of Primary or-
- Secondary Coolant, with the following plant conditions:
. RCS Pressure; 1700 psig - Stable
. Core Exit TC 560*F Stable
. ECCS systems All Actuated as Required 2._CTMI.& essure __.10 psig Stable.
. CTMT Temperature _195?F_ Slowly. Decreasing. -_._ _
.: CTMT Radiation SE2 R/HR
- S/G Aux Feedflow.
100 Kibm/hr'to each S/G
. S/G Pressures 1100 psig - Stable
. RCPs All off
~~ ~ With~these^ conditions the core decay heat is being removed primF :?y byf
~ '"
^^
A.
- Heat transfer from the RCS to the S/G's due to natural circulation.
B.
Heat transfer from the RCS to the SIG's due to reflux boiling.
. C.'
Injection of water from the RWST and removal of steam / water out from the break.
D,.
Injection of water from the RWST and removal of steam / water out the pressurizer PORV.
ANSWER:
A. Heat transfer from the RCS to the S/G's due to natural circulation.
K/A-#:--011K101-t1/4:4 KA DESCRIPTION: : Core Cooling on RCS LOCA OBJECTIVE #:
003D080X
REFERENCES:
T61.003D.6 LP 8 AUTHOR: FXB SOURCE: NEW-HO DISTRACTER EXPLANATION:
B. Occurs on Mid Size LOCAs (up to 1ft' Breaks)in which case RCS pressure would be slightly above S/G Safety Pressure.
C. _Large Break LOCA.
D.- Used for bleed and feed.
RO Outline #79
- QUESTION #45 Which one of the following would violate the Technical Specification Safety Limit concerning Reactor Coolant System pressure? -
A.-
Mode.1_ - 2550 psig B.
Mode 2_- 2350 psig
~
,C _._ Mode E s756~psl6Z l.Zil TE.ZZZ Z_Zill_.,Z li~ ~~TZill_. _.._
~
D.
_ Mode 4 - 2450 psig ANSWER:
- C, Mode 3 - 2750 psig K/A #: 2.2.22 3.4/4.1
)
VA DESCRIPTION: - Tech Spec Safety Limits OBJECTIVE #:
003A3608 and 003A360C REF ERENCES: T61.00'3A.6 LP 36 Tech Specs AUTHOR: RBM SOURCE:. NEW - HO DISTRACTER EXPLANATION:
-The plant mode does not make any difference conceming the safety limit. Any pressure that exceeds
.2735_psig xiolate s_the_s a fe ty_ limit-2550 is above the design pressure of 2485.
2350 is above the PORV setpoint of 2335.
2450 is above the reactor trip setpoint of 2385.
. RO Outline #98
QUESTION #46 NE01 ('A' Emergency Diesel) Room Supply Fan is being placed in the Pull-to-Lock position.
Which ONE of the following is correct for the operability of NE01 while the Room Supply Fan is under the control of the Reactor Operator?
Placing NE01 Room Supply Fan in Pull-to-Lock will have NO effect on the operability of A.
NE01.
___ZBCNE0iTshoGd bs decl5 Fed irspifhbie.~when.ttiblReactoTOper5for piadesifs suppif fan in _ 'lTT Pull-to-Lock.
C.
NE01 should be declared inoperable if outside air temperature exceeds 65 F.
i D.
NE01 should be declared inoperable if NE01 room temperature increases to 100 F.
ANSWER:
C. NE01 should be declared inoperable if outside air temperature exceeds 65'F.
l K/A #: 064A212 2.8/3.1 l
KA DESCRIPTION: HVAC Operability for DGs OBJECTIVE #:
0110600H
REFERENCES:
T61.0110.6 LP 60 AUTHOR: RBM SOURCE: BANK Modified N - HO
-DlSTRACTER EXPLANATION:
A DG may be considered operable with it's associated supply fan inoperable as long as the outside air temperature is less than or equal to 65'F. Response A is not true -it could have an effect. B is not correct if outside air temperature is less than 65. Room temperature does not effect operability unless it reaches 119' Response C is the only correct response for this question.
RO Outline #2
3 QUESTION #47 1
A truck in the switchyard backs into the "B" Safeguards transformer causing a loss of the entire
- switchyard. "A" and "B" D/G's fail to start.
- Which ONE of the following power supplies would be unavailable?
A.
_"A" MFP emergency oil pump power B _ RLO1-RLO6. board power ___ _ _...____. _..__ ~ _.. Z_ _.l_Z_
C.
PA01 breaker control power 1
D.
Main Turbine turning gear oil pump ANSWER:
D. Main Turbine turning gear oil pump.
K/A #: 062K201 3.3/3.4 KA DESCRIPTION: Power Supplies to Major Loads OBJECTIVE #:
011001 01
REFERENCES:
ECl AUTHOR: PJM SOURCE: BANK Modified N - HO DISTRACTER EXPLANATION:
.A
'A* MFP-emergency oil pumpJs powered 4 rom.thePJ01. battery.
B.'._All MCB power is NK power.
C. All major bus PA,PB,NB control power is DC.
_D. Main Turbine turning gear oil pump is powered from PG130.
RO Outline #20
QUESTION #48 The Containment Fan Coolers are running in fast speed when an SIS occurs. Which one of the
- following describes the expected response of the Fan Coolers?
A.
They continue to run unless containment spray is initiated.
B.
They trip and are restarted in slow speed by the LOCA sequencer.
Z_lCllThey_shif'tTo~siow.spe~eTh}feis160isec. delay.ZZZTZil_T.Z.Til'Z_1 ~ _-_i l
D.
ThoyJip and are manually restarted in slow speed.
ANSWER:
B. They trip and are restarted in slow speed by the LOCA sequencer.
~
K/A #: 022A301 4.1/4.3 WA DESCRIPTION: Flow to Containment Coolers on an SI u.-
l OBJECTIVE #:
0110400D
~
l AUTHOR: FXB SOURCE: NEW - L DISTRACTER EXPLANATION:
A. Only SIS snifts cont, spray to slow.
C. Coolers stopped and started @ 35 sec.
D. Not shed and auto start present.
RO Outline #00
QUESTION #49
-j The plant conditions are as follows:
)
. 'A' S/G Tube Rupture
. Manual Safety Injection
. All equipment functioned as designed
.- 'A' S/G atmospheric is OOS
. 'A' S/G pressure 1240 psig.
e 'A' S/G level > NR level indication -
1 L
- 'B', 'C', 'D' S/G levels < NR level indication L
. RCS pressure 1240 psig
. PZR level 40%
Which ONE of the following should be implemented when the STA makes a pass through the CSF's ?
- A.
FR-H.2 Response to Steam Generator Overpressure B.
FR-H.3 Response to Steam Generator High Level C.
FR-H 4 Response to Loss of Normal Steam Release Capabilities D.
FR-H.S Response to Steam Generator Low Level ANSWER:
A. FR-H.2 Response to Steam Generator Overpressure K/A #: E13EA2.1 2,9/3,4 KA DESCRIPTION: Determine Response to S/G Overpressure OBJECTIVE #:
003D260U
REFERENCES:
FR-H.2 Page 1 AUTHOR: PJM SOURCE: NEW - L DISTRACTER EXPLANATION:
A.
Correct. S/G overpressure is highest priority on list due to concern for structuralintegrity of the Steam Generator.
B.
. A' S/G levelis high but pressure is higher priority.
C.
A' S/G pressure meets loss of normal steam release capabilities but overpressure is higher priority.
D.-
'B', 'C', 'D' SIG's levels are all low but overpressure is higher priority.
RO Outline #55
QUESTION #50 Assuming 100% power and 75 gpm letdown. Which ONE of the following failures will'esult in a loss of makeup capability AND cause a reduction in Reactor Power if tne Reactor Makeup
.. Control system is in AUTO and NO operator action is taken?
A.
Loss of instrument air to BGLCV112A B.
BGLT0112 fails high_
C.
BGLT0149 fails high D.
BGLT0185 fails high ANSWER:
C. BGLT0149 fails high K/A #: 004K106 3.1/3.1 KA DESCRIPTION: Response to VCT Level Channel Failure OBJECTIVE #:
0110110F
REFERENCES:
T61.0110.6 LP 11, Pages 19 and 21 OTO-BG-00004 8756D37 Sheet 28 AUTHOR: PJM SOURCE: NEW-HO DISTRACTER EXPLANATION:
AB_GLGV112Maits_to_theXCT_p sition.on_ajossatair.lhis.wilthave_no.effeet.ortthe plant o
B. BGLT0112 failing does not effect VCT makeup system.
C.- BGLT149 failing high will result in BGLCV112A going to the full divert to the RHUT. This will re lowering VCT level. Since 149 also controls makeup, no auto makeup will occur with the channel fa high. When VCT level gets to 5%, an auto swapover to the RWST will occur due to BGLT0112 and 0185. This will result in lowering reactor power due to the boron being added at a higher concentration from the VCT.
D. BGLT0185 failing does not effect VCT makeup system.
. RO Outline #6
4
- QUESTION #51 Plant conditions are as follows:
. Mode 1 100% Reactor Power i Tavg 588.4*F
. NCP is running with BGFCV124 in Auto
. Letdown flow is 125 gpm
. PZR level controller, BBLK459 is in Auto
. BBLT459 is the selected channel BBLT459 fails HIGH. NO operator action is taken.
Which ONE of the following occurs in the plant due to BBLT459 failing?
A.
The Reactor willimmediately trip on PZR high level.
- B.
The Reactor will, after a period of time, trip on high PZR level.
C.
The Reactor will, after a period of time, trip on low PZR level.
D.
The Reactor does not trip due to PZR level.
ANSWER:
B. The Reactor will, after a period of time, trip on high PZR level.
K/A #: 011A210. 3.4/3.6 KA DESCRIPTION: PZR Level Channel Failure _
l OBJECTIVE #:
0110300L -
REFERENCES:
8756D37 Sheet 11 and 27 AUTHOR: PJM SOURCE: BANK Modified Y - HO DISTRACTER EXPLANATION:
A. Rx Trip is 2/3 coincidence.
B. When BBLT459 fails high, the NCP flow control valve will close down to minimum flow, about 40 gpm, to reduce PZR level while letdown is still 125 gpm. When PZR level gets to 17%, BBLT460 will isolate latdown with BGLCV460. With no letdown while still charging, the PZR level will increase until a high level Rx Trip occurs at 92% 2/3 coincidence.
C. No Rx trips on low level.
D. Rx trips on high level.
RO Outline #29
QUESTION #52_-
Which one of the following actions occur _on a high rad alarm (red) on the SG sample monitor, SJ RE 02?
A.
BM-FV-54, SG Blowdown Surge Tank Outlet Valve closes B.
_ BM-HV-65 thru 68, SG Sample isolation Valves close
~
JZ3M:H9AlbruAJG b76566tainment isolationyalves close_ZZ lE_il.
1.ZZZ
~
~
~
D.
PBM01 A, SG Blowdown Discharge Pump trips ANSWER:
- C. BM-HV-1 thru 4 SG B/D Containment isolation Valves close K/A #: 037AK3.10 3.3/3.7 KA DESCRIPTION: Auto Actions to High Rad Levels from SG Tube Leak OBJECTIVE #:
003B140A
REFERENCES:
OTO-BB-00001 AUTHOR: RBM.
SOURCE: BANK Modified N - L DISTRACTER EXPLANATION:
BM-FV-54 and PBM01 A receive input from rad monitors BMRE25 or 52. BM-HV-65 thru 68 are closed by a SGBSIS signal.
RO Outline #59
.- QUESTION #53-E-3, " Steam Generator Tube Rupture", performs RCS depressurization to refill the pressurizer.
Which ONE of the following shows the preferred order for depressurization.
A.
Normal Pressurizer Spray, Auxiliary Spray, Pressurizer PORV I
j l
B.
3orm_al Pressurizer Spray,_ Pressurizer P_ORV, Auxiliary Spray l
~ _ _ _ _ _.. _.. -. _ _. _ -
_-. ~ _
- ~ _
C.
Pressurizer PORV, Auxiliary Spray, Normal Pressurizer Spray j-D.
Pressurizer PORV, Normal Pressurizer Spray, Auxiliary Spray ANSWER:
B. Normal Pressurizer Spray, Pressurizer PORV, AuxiFary Spray
' i K/A #: 038K301 4.1/4.3-KA DESCRIPTION: Preferred Order for Depressurization in SGTR OBJECTIVE #:
003D170V01A
REFERENCES:
T61,003D.6 LP 17 AUTHOR: FXB SOURCE: BANK - Modified N - HO DISTRACTER EXPLANATION:
All Distracters contain methods used by E 3 for depressurizing.
. RO Outline #80 N
- QUESTION #54 Using the attached graph, select the temperature combination at which the cooling tower bypass valves should be open.
Basin Temperature Ambient Dry Bulb Temperature A.
40*F 25'F
.._ _ B_ _ _ _ 50?F_
. _.. _. _.. _. 2 5' F.. _. _ _ _ _.
C.
70*F 10*F I
l D.
75'F 10*F ANSWER:
A.
40'F 25'F K/A #: 075K401 2.5/2.8 KA DESCRIPTION: CT Bypass Valve Operatica OBJECTIVE #:
0110040E
REFERENCES:
T61.0110.6 LP 4. Pg. 66 OTN-DA-00001 AUTHOR: RBM SOURCE: NEW - HO DISTRAGTER-EXPl:ANATiON:
OTN-DA-00001, Circulating Water System, requires the operator to ensure the CT bypass va open if temperature is below the bold line on Attachment 1. Plotting the four options given results in a temperature combination requiring the bypass valves to be open.
RO Outline #39
OTN.DA 00001 Itev. 9 Cooling Tower Freeze Protection Curve
- ~
~ ~ ~ -
+90 -
.j.
\\
I on Valves {0 pen i
+70 -.
3
.i valves _......
Basin Valves Cicsed
- opened / closed l
Temp.
+60 -
j (deg. F) i
+50 -
.Eypass.. Operation.
+40 -
+3 0 '-
i i
i i
i i
i i
i i
i i
30
-20
-10
+0
+10
+20
+30i\\e i
+40
+32 Ambient Dry Bulb Temp. (deg. F)
Select one of t}'e following to maintain backpressure at approximately 2.1 HgA:
I.
3 pump operation and throttle the lower C.W. passes II.
3 pump operation and FREEZE-PROTECT or
)
FREEZE-PROTECT and throttled III.
2 pump operation and normal throttle Th - t
.P*
QUESTION #55 A high high shelllevel occurs in the 1 A heater. Which ONE of the following is the response of Main Feedwater due to the high high level?
A.
Main Feedwater temperature goes up.
B.
Main Feedwater temperature goes down.
-C.-
Main Feedwater flow goes.up,_-
D.
Main Feedwater flow goes down.
ANSWER:
i B. Main Feedwater temperature goes down K/A #: 056K103 2.6/2.6 KA DESCRIPTION: Feedwater Temperature Response to LP Htr Isolation l
l OBJECTIVE #:
011022OG
REFERENCES:
OTS-AF-00003 Page 2 AUTHOR: PJM SOURCE: BANK Modified Y HO DISTRACTER EXPLANATION:
A. Temperature goes down.
B. When a high high level occurs in the 1 A heater, the A LP heater string willisolate. Less heat transfer area,less heating of the main feedwater, temperature goes down.
C.._ Flow doesn'Lchange D. Flow doesn't change.
RO Outline #19 l
QUESTION #56 L Callaway is in MODE 1. All systems are li_ned up in their normal configuration and operating properly. A component cooling water system leak has occurred.
The following conditions are noted by the Reactor Operator:
( The CCW surge tank level is 53% and DECREASING
. VCT levelis 65% and INCREASING Which ONE of the following leakage sources would result in these conditions?--
r L
A.
CCW heat exchanger B.
f.etdown heat exchanger C.
RHR heat exchanger D.
. Seal water heat exchan'ger ANSWER:-
D.. Seal water heat exchanger K/A #: 008K104 3.3/3.3-KA DESCRIPTION: Source of CCW Leakage OBJECTIVE #:
0110100H
REFERENCES:
-T61.0110.6 LP 10 Pages 21-23 AUTHOR: PJM SOURCE: BANK - Modified Y - HO DISTRACTER EXPLANATION:
A.- A leak in the CCW heat exchanger would result in a leak out of the CCW system due to service water being at a lower pressura than CCW. VCT level would be^ steady.
B A leak in the letdown heat exchanger would result in a leak into the CCW system.
C. A leak in the RHR heat exchanger would result in a leak out of the CCW system due to RHR pressure being lower than CCW pressure. VCT level would be steady.
D. A leak in the seal water heat exchanger would result in a leak out of the CCW system due to the seal water retum pressure being less than CCW pressure. The water leaking out of the CCW system would be leaking into the CVCS system causing an increase in VCT level.
RO Outline #10
0UESTION #57-Main Control Board annunciator window 62C, AREA RAD MON FAIL, has alarmed. The BOP-determines it was caused by SDRE0035, New Fuel Storage Area Rad monitor. The CRS directs you to source check the monitor for a response.
Which ONE of the following will you use to perform a source check?
A.
The green button B.
The yellow button C.-
The red button D.
The white button l
ANSWER:
A. The green button K/A #: 072A403. 3.1/3.1 KA DESCRIPTION: Area Rad Mon Source Check OBJECTIVE #:
0110360F
REFERENCES:
T61.0110.6 LP 36 Page 20 AUTHOR: PJM SOURCE: NEW-L DISTRACTER EXPLANATION:
A. if-the green normallight pushbutton is pushedca source check-isinitiated.
B._ This is an alarm acknowledge button.
C. -This is an alarm acknowledge button.
D. This provides source check on the RM11 process rad monitors but not the area rad monitors.
RO Outline #46 h
a i.
L QUESTION #58 i
The plant is being cooled down per ES-0.2, Natural Circulation Cooldown, due to a sustained
!=
loss of all offsite power following a tornado.
The cooldown rate will be limited to less than
'F/hr while the RCS subcooling will be maintained greater than
'F more subcooled than instrument error.
_ _ _ A... _50,.50___
B.
100,100 C..
50,100 D.
100,50 w
ANSWER:
C. 50,'100 K/A #: E09EK2.2 3,6/3.9 KA DESCRIPTION: Subcooling and C/D Rate on Natural Cire OBJECTIVE #:
003D070R & OS
REFERENCES:
ES-0.2 AUTHOR: FXB SOURCE: NEW - L DISTRACTER EXPLANATION:
-A;8;D-AllCRDM fans are-NOT available-due tolosstf offsite powerand;therooldownTateis1imited to-less than 50'F/hr.
RO Outline #69
QUESTION #59 i
The Control Room is being evacuated due to a fire in Panel KC008 You are the Reactor Operator and have been directed by the Shift Supervisor to perform your immediate actions.
Which of the following locations will you initially go to when you leave the Control Room?
A.
Front standard on the main turbine.
B,_.. South (Train B) ESF Switchgear Room..(CB-2000) _. ___. __.
4 horth HVAC R$om ( B 2bd7)
C t
D.
Auxiliary Shutdown Panel I
ANSWER:
B. South (Train B) ESF Switchgear Room (CB-2000)
K/A #: 2.4.11 3.4/3.6 KA DESCRIPTION: Control Room Evacuation OBJECTIVE #:
003B590C
REFERENCES:
T61.0038.6 LP 59 OTO ZZ-00001 4
AUTHOR: RBM SOURCE: NEV9 - 1.
DISTRACTER EXPLANATION:
Each location is a correct location where someone from the CR would respond to during a CR evacuation.as.follows:
A.
Balance of Plant Operator B.
Reactor Operator C.
Primary Equipment Operator D.
Shift Supervisor RO Outline #93
QUESTION #60 The following plant conditions exist:
- Reactor power 23%
- Gen MW 208 l
. CB D roos 130 steps
- Rod Control Automatic
___.______._MER_ Speed Control _ Automatic
. -Main FRV-
- Automatic If turbine impulse pressure controlling channel, PT-505, fails HIGH, the rods in control bank D will (Select ONE of the following.)
A.
{
~
be Inserted into the core (tripped) by a turbine trip signal l
B.
move in the outward direction C.
move in the inward direction E,.
not move ANSWER:
B. move ;n the outward direction K/A #: 001A102 3.1/3A N,T
'CTION: Plant Response to Pimp Failure 08s 11102608 REFEF.
.'61.0110.6 LP 26 Page 16 OTO-AC-00003 AUTHOR: PJM SOURCE: NEW HO DISTRACTER EXPLANATION:
A. There are failures that result in turbine trip but not this one.
B. Two signals will call for outward rod movement.
N1 power being less than turbine power will generate an outward rod demand to catch reactor power to turbine power. Tref will be greater than Tavg. This will generate an outward rod demand to match Tavg to Tref.
C. Wrong direction.
D. Some failures do not result in rod movement.
RO Outline #13
- QUESTION #61
. Plant conditions are as follows:
. Mode 1
. - 100% Reactor Power
. All Automatic controls in their normal condition
. BBTY411, Loop 1 Tavg, fails at 588.5'F
.- A Reactor Trip occurs.
NO Operator Actions are taken. Which ONE of the following occurs in the plant?
A.
. Steam dumps will be open.The RCS will cool down until main steam pressure reaches 615 psig. Then the MSIV's will CLOSE and the RCS will heat up again.
B.
Steam dumps will be open. The RCS will cool down until main steam pressure reaches -
1092 psig. Then the steam dumps will throttle to maintain this pressure.
~ ~ C.
Steam dumps will be open. The RCS will cool down until Tavg reaches 557'F. Then the steam dumps will throttle to maintain this temperature.
7 D.
Steam dumps will be open. The RCS will cool down until Tavg reaches 550*F. Then the steam dumps will cycia closed and open to maintain this temperature.
ANSWER:
D. Steam dumps will be open. The RCS will cool down until Tavg reaches 550*F. Then the steam dumps will cycle closed and open to maintain this temperature.
K/A #: 041A101 2.9/2.9 KA DESCRIPTION: Plant Response to Tavg Failure OBJECTIVE #:
0110200K
REFERENCES:
7250D64 Sheet 10 AUTHOR: PJM SOURCE: BANK Modified N - HO DISTRACTER EXPLANATION:
A. Steam pressure will not reach 615 psig because of P12.
B.' Steam dumps on steam pressure.
C.'. If the steam dump worked properly, this would be true.
D. With no operator action, the input into the steam dump controller from Tavg will be 588.5'F because highest auctioneered Tavg is used.
This will result in Steam Dumps staying open trying to lower Tavg to reference temperature of 557"F. P12 closes the steam dumps at 550 F to prevent excessive cooldown.
RO Outline #18
QUESTION #62
. A turbine runback can be caused by high stator cooling water temperature or low stator cooling water flow.
A.
inlet; inlet B.
inlet: outlet
_._C. - - outlet;- outlet ------_-__ _ -
D.
outlet; inlet ANSWER:
D. outlet; inlet K/A #: 045K412 3.3/3.6 KA DESCRIPTION: Inputs to Turb Runback i
)
OBJECTIVE #:
0110330E
REFERENCES:
T61.0110.6 LP 33, Page 13 -
AUTHOR: RBM SOURCE: BANK Modified N - L DISTRACTER EXPLANATION:
A turbine runback is caused by high stator CW outlet temperature and low stator CW inlet flow.
RO Outline #34 m
. QUESTION #63 -
L All reactor coolant pumps secured with the following conditions:
.. RCS WR Pressure (BBPl405) 400
. RCS NR Pressure (BBPl406) 350
. Charging Header Pressure (BGPl120) 550
. VCT Pressure (BGPl115)-
50 psig
--Using the attached graphc Which ONE of the below #1 seal leak-off flowrates is the MAXIMUM flow that would still allow starting reactor coolant pumps?
A.;
- 1.0 gpm
. B.
1.5 gpm C.
2.0 gpm.
D.
2.5 gpm
. ANSWER:
C. 2.0 gpm K/A #: 003A404 - 3.1/3.0 KA DESCRIPTION: Monitor RCP #1 Seal D/P
.OBJECTIVEJd0.3A200A3 __._ _.
REFERENCES:
.OTN BB-00003, Att.1 AUTHOR: FXB SOURCE: NEW-HO DISTRACTER EXPLANATION:
A. Max flow for 200 psid B. Max flow for 400 psid '
C. 550-50=500 psid=:2.0 gpm D. Max flow for 700 psid RO Outline #48 i
OTh' llll.0000.1 Rev. 10 I
ricu n 1 NO.1 SEAL NORMAL OPERATING RANGE 6
I l'
i i
q. _..
/
/
. /
/
_ _ NO.1
/
SEAu y
LEAK 3 RATE
~
(GrW-NORMAL OPERATING RANGE
/
7 9
i
.w' i
I I
.2
,i 0
1 0
200 400 600 800 1000 1200 1400 1600 1800 2000 220b 2400 NO,L. SEAL-O!FFERENTIAL. PRESSURE 4 PSI)--
2500 Use the fouod:y; indicators ifleu than 400 PS!D 2250 RCP $al Differential Press. Indicator -
A BB PI - 151A B
BB P1 152A
~
C' BB PI 151A D'
BB Pl.15M Otherwise use BG P1 - 120A BG Pl.'115 p,n,snrp l
i v e a es,s o m
-e.
}
QUESTION #64
..The following conditions exist:
. Safetyinjection Actuated
. Procedure in Effect E-1 Loss of Reactor or Secondary Coolant
. RCS Core Exit Tc-580'F
. RCS_ Cold Leg Temp _
560*F
. CTMT Pressuro 9 psig
. CTMTTemperature 190*F
. CTMT Radiation 1E3 R/HR Using the provided attachment 2 from E-1, determine which ONE of the below combinations of RCS pressure and temperature would satisfy subcooling requirements.
^
- RCS WR Pressure Core Exit TC A.
1800 psig 590'F 1
B.-
1500 psig 570'F F
- C.
1200 psig 550'F D.
.1000 psig 500'F.
ANSWER:
D.
_1000 psig 500'F K/A #: 074EA201 4.6/4.9 l
KA DESCRIPTION: Determine Subcooling Requirements OBJECTIVE #:
003D070S -
REFERENCES:
E-1, Att. 2 AUTHOR: FXB f
SOURCE:' NEW-HO DISTRACTER EXPLANATION:
A.
Must use page 2 of attachment-below dotted line for adverse B.C - Below dotted line for adverse containment RO Outline #62 i
f s
f
l Coouo : LOSS or F. tac:oa ca.SocosoAar l A::acnmen e ocee. no.
l nev.
,)
2n2 t-1 RCS SUDCOOLING CURVES RCS StiBC00UNC Cl;R\\TS 5 b4m Up a u. *a<.hoa -1.'<e: - t %.:
.u.:. va:. i.d A. W.J.EN ?A SSUAE iS GRE ATEh !..AN !?00 #5iG.
O 1
A 1
c ACCEPTBLE REC 10N IS TO TBE LEFT OE THE APPLICBLE CURYE 5 3 5 0 0 -- J. _....._._.u- - - - - ' _ _. _ _. _
l.
j
-L 3.. p. -. _.- - _ _ 7. :- - - - ~ ~ - -
~ ~ _ _ --
__3
!/
I i 2
/ i [a i
i i
A l
= _ -. - -
,o
/l i
, N
- ji i
j 3
3 0 0 0 l
/
1
' 0
/
'/
l i,
/
l l2
/i/
i I
- N
/
i i
.i l G 2500
/
f-
- /
[i i
j,,
l[
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-['.
l l
i i;
,/
/
i i
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/
i is
. U 2000
,/
i j_j _,_ __
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/
a i
. /;
l l
i I I l*
.i
.. ~ /
i
/
i l
i lS1500,li iP l
y i
G 500 550 600 650 700 s
I RCS TEMPEAtTURE IN DEG. F, I.
i (i;iERM0 COUPLE DATA - NORMAL CTHI ONLY - USE SOLIO CURN}E THERM 0 COUPLE DATA ADVERSE CTHI - USE DASHED CURVE l
,if IHERMDCOUPLES ARE UNAVAllABLE KR RIO DATA CAN BE USED FOR ALL CTMT CONDII10AS - USE DASHED CURVE)
Page 1
of 2
Proced. !;o.
LOSS OT REACTOR OR.SEColiDARY Attachment Rev.
)
E-1 COOLMIT 2
3B2 ECS SUDCooLING CURVES RCS SUllC00lJNG CURES tr.sse.
NUit:
UHAiW0k 1 niiH H 5 nt'
Uat IrugA. WHEN PgESSURE Ij5 ggt HANUt rW:THAN 1700 S G.
DA L
A:
ACCEPTABLE REC 10NIS TO THE LEFT OF Tile APPLICABLE CURL'E 2000' g. __ j
,' _ l.. 'l _,i _
.t _'_'_j_g'. i,
--4--I
._ l ll I i l II I
l ffI Ii 5
ll ll l
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l il a /,'
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ll l l 1!!il!_j.il 1i lii!l!l
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l il
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l I. I I !
lN ll 1
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_ g.y; ll l
M'X l
l II ll l
P
?
!..M "
I o
l
,,il
,,,,,l l
200 300 400 500 600 700 RCS TEMPERATURE IN DEG. F.
(TlftRMDCDUPLE DATA - NORMAL CTMT ONLY -
THERMOCOUPLE DATA - ADVERSE CTHT - USE DASHED CURVE
)
,lFTHERMDCOUPLESAREUNAVAILABLF XR R10 DA1A CAN EE USED FOR ALL CTHI CONDIT10'h5 - USE DASHED CURVE)
Page 2
of 2
QUESTION #65 Plant conditions are as follows:
. A' Circ Pump lockout
- Setback in progress
'. MCB annunciator 81D, Rod Bank Lo lit
. MCB annunciator 81C, Rod Bank Lolo lit -
l
. Which ONE of the following indications will meet the requirements for current plant conditions ?
.A.
- 40 gpm flow indicated on BGFl183A, EMERG BORATE TO CCP A & B HDR FLOW IND A.
B.
40 gpm flow indicated on BGFY1108, BA TO BA BLENDING TEE FLOW BATCH COUNTER.
C.
BGHISSA, A BORIC ACID XFER PMP PBG02A HAND IND SW, red light lit.
D.
BGHV8104, EMERG BORATE TO CCP A & B HDR ISO HV, red light lit ANSWER:
A. 40 gpm flow indicated on BGFl183A, EMERG BORATE TO CCP A & B HDR FLOW IND A.
K/A #: 024AA2.01 3.8/4.1-KA DESCRIPTION: Verification of Emergency Borate Flow OBJECTIVE #:. 0038610A
REFERENCES:
' OTO-ZZ-00003 Page 2
~
AUTHOR:. PJM SOURCE: NEW-HO DISTRACTER EXPLANATION:
A. Correct. If flow on that meter is bss than 30 gpm, the operator has addition actions to perform to get boration started.
B. This counter will not indicate during emergency borate flow.
C. The pump running by itself is not proof of emergency boration.
D. The valve being open by itself is not proof of emergency boration.
RO Outline #74
OUESTION #66 Which one of the following conditions would requito action to be taken within one hour to avoid violating the plant ToJinical Specifications?
A.
The plant is at 40% power with /FD outsido its specified targot band.
B.
The plant is at 2% power with Tav0 at 550'.
~
~
C._ - The plant li'ai'40% power vihen 520 voli AC bus NN01 losos power.
~ ~l T ~
D.
The plant is at 2% power when SR channel N31 fails.
ANSWER
- B. The plant is at 2% power with Tavg at 550'.
K/A #: 2.1.11 3.0/3.8 KA DESCRISTION: TS 3.1.1.4 Minimum Temp for Criticality OBJECTIVE #:
003A0300
REFERENCES:
~
AUTHOR: RBM SOURCE: NEW. HO DISTRACTER EXPLANATION:
l A would be a 15 min response if above 50% power.
B :s a 15 min response.
C requires action within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
D would require immediate action if below P 0.
RO Outline #88
QUESTION #67 A reactor startup is in progress with IR power at 3x10-11 amps. The sourco range Hi h Flux 0
Trip has NOT boon blocked. Which OWE of the following describes the Reactor Protection Systorn responso if a CONTROL POWER fuse blows on N 31 with the sourco rango Lovel Trip Bypass Switch in the positions indicatod?
Loyol Trip Bypass: NORMAL Lovel Trip Bypass: BYPASS
. __. A _
. _ - _ _ no trip
-. no trip B.
reactor trip t.) trip C.
no trip reacior trip D.
It: actor trip reactor trip ANSWER:
l B.
reactor trip no trip K/A #: 015K604 3.1/3.2 KA DESCRIPTION: Source Range Ni Failure OBJECTIVE #:
0110280D
REFERENCES:
T61.0110.6 LP 28 Pages 9 & 10 7250D64 Sheet 3 AUTHOR: PJM SOURCE: NEW HO DISTRACTER EXPLANATION:
A. Trip signal processed when in normal.
B. When a Control Power fuse blows on the Source Range Instrument, the trip bistable turns on to the tripped condition. The Level Trip Bypass switch, when in the normal position will allow a trip signal to process through to the reactor protection system. The Level Trip Bypass switch in the BYPASS position, blocks the Reactor Trip signal from the reactor protection system.
C. Trip signal processed when in normal.
D. No trip when in bypass.
RO Outline #14
e QUESTION #68 The plant is at 100% power with the following conditions:
e *C" CCW pump OOS e 'A' CCW Train in Service, *B" CCW Train in Standby
- e NCP in Service with 120 gpm Letdown Flow The "A" CCW pump trips due to unknown reasons.
Which ONE of the below include:: reylmd immediate actions?
A.-
Restart the "A" CCW pump, if pump falls to restart, then start either *B" or *D" CCW pump and transfer the service loop to the *B" CCW Train.
- B.
Start either "B" or *D" CCW pump then transfer the service loop to the "B" CCW Train.
C.
Trip the Reactor Coolant Pumps and the Reactor, enter E 0, Reactor Trip / Safety
- Injection, then transfer the service loop to the "B" CCW Train.
D.. -Verify "B" CCW pump starts automatically then transfer the service loop to the "B" CCW Train.
. ANSWER:
B. Start either *B" or "D" CCW pump then transfer the service loop to the "B" CCW Train.
NIA #: _026K303
_.4.0/4.2 KA DESCRIPTION: Loss of CCW Pump Operator Actions.
OBJECTIVE #:
003B290B
REFERENCES:
OTO-EG 00001 AUTHOR: FXB SOURCE: NEW HO DISTRACTER EXPLANATION:
A Procedure OTO EG 00001 specifically states otherwise.
C 10 min or high RCP temp limit would be E-0 criteria.
D.
No automatic start for opposite train CCW.
RO Outline #60
I 4
OljESTlON #68 Th plant is at 100% power with the following conditions:
. *C" CCW pump OOS e *A" CCW Train in Service, "B" CCW Train in Standby
. NCP in Service with 120 gpm Letdown Flow The "A" CCW pump trips due to unknown reasons.
Which ONE of the below includes required immediate actions?
A.
Restart the "A" CCW pump, if pump falls to restart, then start either "B" or *D" CCW pump and transfer the service loop to the "B" CCW Train.
B.
Start either *B" or *D" CCW pump then transfer the service loop to the *B" CCW Traln.
1
-C.
Trip the Reactor Coolant Pumps and the Reactor, enter E 0, Reactor Trip / Safety Injection, then transfer the service loop to the "B" CCW Train.
D.
Verify "B" CCW pump starts automatically then transfer the service loop to the *B" CCW Train.
ANSWER:
B. Start either *B" or "D" CCW pump then transfer the service loop to the "B" CCW Train.
K/A #; 026K303 4.0/4.2.
KA. DESCRIPTION:. Loss of CCW Pump Operator Actions OBJECTIVE #:
003B2908-
REFERENCES:
OTO EG-00001 AUTHOR: FXB SOURCE: NEW - HO DISTRACTER EXPLANAllON:
A Procedure OTO EG-00001 specifically states otherwise.
C 10 min or high RCP temp limit would be E 0 criteria.
D.
No automatic start for opposite train CCW.
RO Outline #60
1 QUESTION #G9 A steam line has ruptured inside containment resulting in a reactor trip and safety injection.
E-0, Reactor Trip and Safety injection has been entered and the operating crew is in the progress of transitioning to E 1, Loss of Reactor or Secondary Coolant. While monitoring the CSF status trees, you determine that an ORANGE path exists for SUBCRITICALITY.
Which one of the following actions should be performed by the crew:
i__.1 ? Continue current pass through the status trees,if no RED pith fs enc'ountered then
~
Implement FR S.1.
B.
Complete the actions of E-1, then implement FR S.1.
C.
Immediately implement FR-S.1, then continue current pass through the status trees.
D.
Implement FR S.1 at the discretion of the Shift Supervisor.
ANSWER:
A. Continue current pass through the status trees, if no RED path is encountered then implement FR S.1.
3 K/A #: 2.4.22 3.0/4.0 KA DESCRIPTION: CSF Implementation Requirements OBJECTIVE #:
003D010B
REFERENCES:
~SOURCEr-BANK Modified NvHO-
~
DISTRACTER EXPLANATION:
All options are plausible in that they are actions that could be considered by the crew as to when to implement FR S.1. According to the EOP rules of usage, however, response A is the correct action.
RO Outline #90
(
QUESTION #70 The plant is operating at full power when a Loop 1 Wide Range Tc RTD fails HIGH. This event will have which one of the following offects on indicated subcooling on the 'B' Train Core Subcooling Monitor?
A.
No offect sinco only core exit thermocouples provide input.
B.- No offect sinco the Subcooling Monitor is bypassed with the reactor trip breakers closed.
CL Small decrease sir ce the Subcooling Monitor usos the average of the Wide Rango RTDs and the core exit thermocouples.
]
D.
Largo decrease sinco the Subcooling Monitor usos the highest reading Wide Rango RTD or core exit thermocouple.
ANSWER:
D. Largo decrease since the Subcooling Monitor uses the highest reading Wido Range RTD or core oxit thermocouple.
K/A #: 017K401 3.4/3.7 KA DESCRIPTION: Incore Thermocouple inputs OBJECTIVE #:
0110300C
REFERENCES:
8756D37 Sheet 6 AUTHOR: FXB SOURCE: NEW - L
~ iSTRACTER EXPLANATION:'
~ ~ ~
~ ~ ~ ~ ~~
~ ~ ' ~ ~
~~
D The candidate must know that loop 1 Tc feeds train B only and that the temperature input is auctioneered high of Th, Tc, and thermocouples.
RO Outline #21
QUESTION #71 The plant is operating at 50% steady stato power, with Tavg and Trof matched. Rods are withdrawn in manual continuously for 20 steps.
Which ONE of the following combinations includes indications that would be received in the control room?
Trof/Touct _
Rod Control Temp
_.. _._ Annunciator.
Error Indicatiort _
_ Turbine Load
~_
A.
HI Positive Decreasing B,
LO Positive Increasing C.
HI Negative increasing D.
LO Negative Decreasing ANSWER:
B.
LO Positive Increasing i
K/A #; 001AK2.06 3.0/3.1 KA DESCRIPTION: Tave/Trof deviation motor OBJECTIVE #:
0038530A
REFERENCES:
OTO SF 00002 Page 2
~
' AUTHOR: FXB
~
~
SOURCE: NEW-HO DISTRACTER EXPLANATION:
A. Annunciator Hi and Turbine Load Incorrect C. Hi Annunciator, and Error Indication incorrect D. Error Indication, and Turbine Load incorrect RO Outlino #75
QUESTION #72 A LOCA has occurred. ES1.2, Post LOCA Cooldown and Depressurization is in progress. Just prior to initiating the RCS cooldown, the low steamline pressure Sl is blocked.
Which ONE of the following is the reason for blocking the low steamline SI ?
A.
To prevent the MSIVs from auto closure on low steamline St.
,.il.Zjo dMion thibiealtilin's isolation from high rate signal.
i.Z T i J r
- C, To prevent any Si equipment from actuating.
D.
To allow manual operation of si equipment.
ANSWER:
A.- To prevent the MSIVs from auto closure on low steamline SI.
K/A #: E03EK2.1 3.0/4.0 KA DESCRIPTION: Block Low Steamline Press Sl Prior to Cooldown OBJECTIVE #:
003D100E
REFERENCES:
-T61.003D.6 LP 10 AUTHOR: PJM SOURCE: NEW LO DISTRACTER EXPLANATION:
A. Correct. It is preferable to use steam dumps for a cooldown. The MSIVs need to be open to use
.them-
=
B, When low steamline si is blocked, the high rate signal is unblocked. This is not why the SI signalls blocked.
C. SI has already occurred at this point.
D. Resetting Si not blocking SI would allow equipment with the SI signal affecting them to be operated.
RO Outline #83 4-
QUESTION #73 Which one of the following actions must be performed if the plant computer is down for 15 minutes 7 A.
Perform a heat balanco.
B.
Perform control room computer down logs.
C.-Perform AFD monitor alarm inoperable surveillance. -
D.
Verify locally that area room temperatures are within specifications.
l ANSWER:
C. Perform AFD monitor alarm inoperable surveillanco.
K/A #: 2.4.11 3.4/3.6 KA DESCRIPTION: Plant Computer Outage OBJECTIVE #:
003B4608
REFERENCES:
T61.0038.6 LP 40 OTO-RJ 00001 AUTHOR: RBM SOURCE: BANK Modified N - L DISTRACTER EXPLANATION:
All responses are actions that are taken in response to the computer being down but are required after different-amounts of time.-A-and B are not required until the computer has been down for-4 hours and D-is not required until the computer has been down for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
RO Outline #97
l QUESTION #74 Callaway Plant is in Mode 3 cooling down for a Refueling outage. The Reactor Operator has been directed to decrease RCS pressure to 1950 psig. Which ONE of the following would the Reactor Operator have to set the Pressurizer Pressure Master Control, BBPK455A, to maintain the RCS at 1950 psig in Auto? Narrow Range PZR pressure range is from 1700 to 2500 psig.
i A.
1.77 turns I-2.55 turns _
C.
3.13 turns D.
4.41 turns l
ANSWER:
C. 3.13 turns K/A #: 002A303 4.4/4.6 KA DESCRIPTION: Master Pzr Press Controller Setting OBJECTIVE #:
0110090C
REFERENCES:
v0A RL 00004 AUTHOR: PJM SOURCE: BANK Modified N - HO DISTRACTER EXPLANATION:
1950
..-1700 -
250 80 psig per turn
= 3.13 RO Outline #23
QUESTION #75 l
The plant is operating at 100% power. The instrument air line to the "A" S/G FRV ruptures.
The plant trips on S/G low level.
Which ONE of the following caused both main feed pumps to trip?
A.
MFP discharge pressure high high
..-~ BT' ' Trip of all condensate pumps -
~
~
~ --
C.
FWIS ANSWER:
D. FWIS K/A #: 059A212 3.1/3.4 KA DESCRIPTION: MFP Trips from FRV Failure OBJECTIVE #:
0110230D
REFERENCES:
T61.0110.6 LP 23, Pages 26,27, and 29 AUTHOR: PJM SOURCE: BANK Modified N HO DISTRACTER EXPLANATION:
A. This is a trip but FWIS will actuate first.
B~Thrs~1sTtilp bWel5EdsFssts pumps are not'165t7
~ ~ ~
~ ~ ~ ~ ~ - -
C. This is a trip but no conditions exist to overspeed the turbine, i
D. When the S/G low low reactor trip occurs, a FWIS is generated from the S/G low low level also. This will trip both MFP's.
RO Outline #43
QUESTION #76 During post LOCA cooldown and depressurization CCP's and Si pumps are stopped on alternato ECCS trains when possible.
Which one of the following describes the reason for alternating the ECCS trains?
A.
Increase the probability of RCS injection flow if one ECCS train becomes inoperable.
~
. EqUaiize flows on ECCS trains to maintain pump. suction head.'
~
B.
C.
Minimize primary leakage while ensuring adequate subcooling.
D.
Equalize loading on the NB busses, since it is assumed that the d,esels are supplying power.
ANSWER:
A. Increase the probability of RCS injection flow if one ECCS train becomes inoperable.
I K/A #: E03EK1.2 3.6/4.1 KA DESCRIPTION: Reason for Alternating ECCS Trains OBJECTIVE #:
003D100J RF"ERENCES: TG1.003D.6 LP 10 I
AUTHOR: FXB SOURCE: BANK Modified N - L DISTRACTER EXPLANATION:
Responses B,-Crand D are statements that are true and would be desirable in system design and-operational considerations. 'A'is the basis for securing ECCS pumps on alternate trains.
RO Outline #68 1
t QUESTION #77 The plant experienced a Si on low PZR press. RCS pressure is 1750 psig and decreasing. E-0, Reactor Trip or Safety injection, Step 21 is in progress. PZR spray valve BBPCV4558 is partially open and will not close in manual.
Which ONE of the following is the action to perform per E-07
_ A.
. Trip the B RCP B.
Trip the D_RCP C.
Trip the A and D RCP's D.
Trip the B and D RCP's ANSWER:
C. Trip the A and D RCP's l
K/A #: 008AK3.03 4.1/4.6 KA DESCRIPTION: E 0 Response to Spray Valve Open OBJECTIVE #:
0110090B
REFERENCES:
- E 0 Page 21 AUTHOR: PJM SOURCE: NEW-HO DISTRACTER EXPLANATION:
.A Two RCP's are required by procedurec-B. Two RCP's are required by procedure although stopping just D does stop all spray flow.
C. Correct. This will stop spray flow through this valve.
D. This pump combination is for the BBPCV455C spray valve.
RO Outline #82
QUESTION #78 The plant is in MODE 6 with core reload in progress.
Which ONE of the below sets of conditions will allow fuel movement to continue with both Boric Acid Siorago Tanks (BAT) drained?
- Level Boron -
Temperature A.
96,500 gal 2585 35'F B.
77,500 gal 2750 55'F C.
59,350 gal 2320 75'F l
D.
38,300 gal 2100 95'F ANSWER:
B, 77,500 gal 2750 55'F K/A #: 004A227 3.5/4.2 KA DESCRIPTION: RWST Operability in Mode 6 OBJECTIVE #:
0110130E02A REFERENCED. FSAR 16.1.2.5 AUTHOR:- FXB-SOURCE: BANK Modified N - HO
~
DISTRACTER EXPLANATION:
RWST Requirements MODE 5 or 6 55,416 gal 2350 ppm Boron 37 'F Min Temperature RO Outline #25
i QUESTION #79 '
Plant conditions are as follows:
- Rod H-8 in CB D was misaligned low.
. Rod H 8 was withdrawn 15 steps to align it with the other rods in CB D.
-. The P/A converter AUTO / MAN switch is broke in the Auto position.
-.. Which ONE of the following conditions occurs because of the P/A converter being in AUTO
-when rod H-8 was recovered? --
A. -
Rod Control Non Urgent Failure alarm when CB D rods initially moved in.
B.
Rod Control Urgent Failure alarm when CB D initially moved in.
C.
Rod Bank Lo alarm will be received with CB D actually above setpoint.
l D.
Rod Bank Lo alarm will be received with CB D actually below setpoint.
ANSWER:
D. Rod Bank Lo alarm will be received with CB D actually below setpoint K/A #: 001K504 4.3/4.7 KA DESCRIPTION: Rod insertion Limit /PA Converter Malfunction OBJECTIVE #:
0110260 0
REFERENCES:
T61.0110.6 LP 26, Pages 20,27,29,30 and 31 AUTHOR: PJM SOURCE: NEW HO DISTRACTER EXPLANATION:
A'. Some rod problems cause a Non Urgent etarm but not this problem.
B. Some rod problems cause an Urgent alarm but not this problem.
C. Wrong Direction D.- With the PA converter still in Auto when the CB D rod is moved out, the PA converter will count the rods 15 steps further out than they actually are. When this is input into the rod insertion limit computer, which generates the ROD BANK LO and LOLO alarms, the rods will be 15 steps below where they need to be when an alarm occurs.
RO Outline #49 -
\\
QUESTION #80 During the performance of ECA 3.2, SGTR with Loss of Reactor Coolcnt Saturated Recovery Desired, the current pass through the CSF's indicate a yellow path on Core Cooling requiring consideration of performing FR C 3, Response to Saturated Core Cooling.
Which ONE of the following actions should be performed 7 A.
Proceed immediately to FR-C.3.
~
B.
Pcrform FR C.3 in parallel with ECA-3.2.
C.
Finish performing ECA-3.2.
D.
Proceed immediately to ES-0.0, Rediagnosis.
1 ANSWER:
l C. Finish performing ECA 3.2.
)
l K/A#: E07EK1.2 3.1/3.6 KA DESCRIPTION: Saturated Core Cooling Procedure Usage OBJECTIVE #:
003D250U
REFERENCES:
FR C.3 Page 2 AUTHOR: PJM-SOURCE: NEW HO DIST RACTER EXPLANATION:----------
A. FR C.3 should not be performed during ECA 3.2 B. FR C.3 should not be performed during ECA 3.2 C. Correct D. Rediagnosis is not necessary.
RO Outline #G3 I
/
QUESTION #81 The following conditions exist:
- SPENT FUEL POOL Hl/LO Level Annunciatoris received.
. The plant is at 100% power.
- Makeup to the SFP is required due to evaporation.
. SFP boron concentration is 2450 ppm.
~
~
-- Which 65E of the' below sources would be used for makeup? S A.
Blended Flow via CVCS 8.
Refueling Water Storage Tank C.
- Recycle Holdup Tanks D.
Reactor Makeup Water ANSWER:
D. Reactor Makeup Water K/A #: 033A101 2.7/3,3 KA DESCRIPTION: Normal Makeup to SFP OBJECTIVE #:
011024 0)
REFERENCES:
T61.0110.6 Lesson 24 4
OTN EC 00001, Sec. 5,9 l
'AUTHORF FXB '
~
~~
~~
~
~
~
~~
SOURCE: NEW Modified N/A L DISTRACTER EXPLANATION:
Makeup via RMW due to evaporation. Choices A,B.C are plausible since they are used for other conditions of makeup other than evaporations.
RO Outline #24
-k
QUESTION #82 Radwasto is discharging DMT *A*. HBFV866, LRW DISCH FCV, automatically CLOSED.
Which ONE of the following could have caused HBFV866 to CLOSE automatically.
A.
Servico Water Radiation Monitor Hi Hi B.
Liquid Radwaste Discharge Flow High C.
Liquid Radwaste Discharge Flow Low D.
Dilution Water Flow Low ANSWER:
D. Dilution Water Flow Low K/A #: 068K401 3.4/4.1 KA DESCRIPTION: Auto Isolation of Radwaste Discharge OBJECTIVE #;
0110160N
REFERENCES:
M22HB05 AUTHOR: PJM SOURCE: BANK Modified N L DISTRACTER EXPLANATION:
A. Service Water Radiation Monitor does not input to HBFV866.
B. Liquid Radwaste Flow does not input to HBFV866.
C. Same as B D."The conditions'that will auto close HBFV866 are:
~ ~ ~
~ ~ ~~ ^ "~~~~
^
Loss of Air Loss of Voltage HBRE18 Hi Hi Low Dilution Water Flow RO Outline #28
t
~ QUESTION #83-Plant coriditions are as follows:
- RCS large break LOCA occurred e Safety injection on low pressurizer pressure e All equipment function as designed e Cold Leg Recirculation in Progress e RCS Temperature is 325'F
-.-- + RCS Pressure is 35 psig
- E 1 Step 15 isolation of accumulators is in progress Which ONE of the following actions is the preferred order of isolating the 'B' accumulator?
A.
Close the isolation valve, open one of the vent valves, open the other vent valve
. B.
Open one of the vont valves,- open the other vent valve, close the isolation valve l
l C.
Close the isolation valve, open the vent valve l
t i
D.
Open the vent valve, close the isolation valve 4
ANSWER:
A. Close the isolation valve, open one of the vent valves, open the other vent valve K/A #: E15EK3.3 2.9/2.9 KA DESCRIPTION: Emergency isolation of St Accumulators OBJECTIVE #:
01101908
REFERENCES:
E 1. Step 15 RNO AUTHOR: PJM SOURCE: NEW HO DISTRACTER EXPLANATION:
Plant design has the motor operated accumulator isolation valves for 'B' and 'C' below the water levelin an accident that fills containment, it is assumed the isolation valves do not work for that reason, To meet single failure criteria; a second vent valve was added to the 'B' and 'C' accumulators.
A. Correct B,' Order wrong by procedure.
C. Order ior 'A' and 'D' accumulators.
D. Order wrong by procedure.
RO Outline #53
OUESTION #84 The following plant conditions exist:
. ANNUN 88C, HI Ctml Press SI Rx Trip, LIT
. Reactor Power 97 %
...
- Gen MW 1200
.__..LCRG19i.._
. Open. ___
- LCPG20 Open
. Control Rods Being manually inserted by the RO e Turbine Stop Valves Open
{
As the Balance of Plant operator, which one of the following would be the first action you would take?:
A.
Manually runback the main turbine.
B.
' Manually trip the main turbine, t
C.
Manually open switchyard breakers V53 and V55.
D.
Fast close the MSIVs.
ANSWER:
B. Manually trip the main turbine.
-. - - ---K/A #H020EK3.06-4;2/4:3-KA DESCRIPTION: FR S.1, Failure of Turbine to Trip OBJECTIVE #:
003D2008
REFERENCES:
FR S.1-AUTHOR: FXB SOURCE: NEW-HO DISTRACTER EXPLANATION:
All options are plausible. They are all actions found in the RNO response column of Step 2 in FR.S.1 for ensuring the turbine has tripped The order they would be performed in FR.S.1 is B A-D-C.
RO Outline #81
QUESTION #85 A large steam line break occurs inside containment. A Safety injection occurs on Containment Pressure. Containment pressure is 30 psig when step 14 of E-0, ' Check if CTMT Spray is Required', is performed.
Which ONE of the following is the reason for stopping all four RCP's ?
A.
They are an unnecessary addition of heat to Containment.
B.
All RCP cooling water flow is automatically isolated.
C.
Air is too dense for the motor cooler fans to keep the motor cool.
l D.
Containment structural failure is imminent.
ANSWER:
B. All RCP cooling water flow is automatically isolated.
K/A #: E14EK2.1 3.4/3.7 KA DESCRIPTION: Manual Actions on High Containment Pressure OBJECTIVE #:
003D040W
REFERENCES:
T61.003D.6 LP 4 Page 42 E-0 Pages 13 and 14 AUTHOR: PJM SOURCE: NEW L
.DlSTRACTER EXPLANATION-At 27 psig, a CSAS and CISB will occur.
A. Although they are adding heat, this is incorrect.
B. Correct. When the CISB is received, all CCW to the RCP's will automatically isolate.
C. The Containment Coolers are running in slow speed because of the air density.
D. Containment pressure will cause containment structural failure but design pressure is 60 psig.
RO Outline #61 9
1 I
OUESTION #86 Which ONE of the following will initiate a Blowdown and Sample Process Isolation Signal (BSPIS)?
A.
Blowdown surge tank level high.
I B.
Undervoltage on either NB bus.
. $_ S C,
.High ter$perature at non regen heat exchanger. outlet.
~
D.
Blowdown domineralizer high differential pressure.
' ANSWER:
A. Blowdown surge tank level high.
l K/A #: 035K403 2.6/2.8 l
KA DESCRIPTION: Initiation of BSPSIS OBJECTIVE #:
0110120D
REFERENCES:
OTO-SA 00001 T61.0110.6 LP12, Pages 910 AUTHOR: FXB SOURCE: BANK Modified N L DISTRACTER EXPLANATION:
B -Initiater a-SGBSIS--
C Closes BMHV1 4
~
D Causes Alarm Only RO Outline #26
i QUESTION #87 An inadvertent SIS occurs. All systems actuato as designed.
l Which ONE of the following can be roset by itself with no other signals being resot?
A.
CPIS B.
CRVIS I
C.
SGB5lS -
'l
~
D.
CISA ANSWER:
D. CISA K/A #: 013A402 4.3/4.4 l
KA DESCRIPTION: Roset of ESFAS Channels OBJECTIVE #:
01105208
REFERENCES:
T61.0110.6 LP 52, Pa0es 8,16, and 23 AUTHOR: PJM SOURCE: NEW L DISTRACTER EXPLANATION:
A. CPIS requires CIS A reset.
B. CRVIS requires CIS-A reset.
C. SGBSIS requires SIS reset.
.RO. Outline #47
QUESTION #88 A slowly increasing RCS level during mid loop operations could be a symptom of l
A.
air entrainment into the RHR system.
B.
nozzle dam leakby into the RCS system.
high differential pressure between the Containment and Fuel Building.
C.
D.
pressurizer nitrogen overpressure too high.
ANSWER:
A. air entralnment into the RHR system.
K/A #: 025AA1.02 3.8/3.9 KA DESCRIPTION: Air Entrainment RHR OperaUori OBJECTIVE #:
003E030E
REFERENCES:
LP 3 AUTHOR: RBM SOURCE: NEW - HO DISTRACTER EXPLANATICN.
Responses B and C would cause level changes, but tne result would be a decrease in vessellevel.
Response D would be true except that the pressurizer is vented to atmosphere durin0 mid loop pps.
- RO Outlinn #70
QUESTION #89 The plant is in Mode 1 at 25% reactor power. Which one of the following would be cause to enter E-0, Reactor Trip or Safety injection?
A.
'A' Steam Generator level is at 10% on all channels and the reactor has not tripped.
B.
The main turbine stop valves have closed and the reactor has not tripped.
Z~_T15.S Pressurizer' level channel 459 is at 98% and the reactor has not tripped.
~
~
~
. D.-
'C' RCP breaker has just tripped open and the reactor has not tripped.
ANSWER:
A.- 'A' Steam Generator level is at 10% on all channels and the reactor has not tripped.
K/A #: 2.4.1 4.3/4.6 KA DESCRIPTION: Reactor Trip Requirements 25%
OBJECTIVE #:
003D040A
REFERENCES:
T61.003D.6 LP 4 AUTHOR: RBM SOURCE: BANK Modified N L DISTRACTER EXPT.ANATION:
Response A should have resulted in a reactor trip. Response B would only result in a reactor trip if.
above 50% power.- Pressurizerlevel trip requires 2 of 3 channels above 72% The trip of one RCP only trips the reactorif above 48% power.
RO. Outline.#91__
pr.
I QUESTION #90 The plant is in Mode 6 with refueling in progress.
Which ONE of the following is the ventilation lineup for Containment?
A.
Mini purge supply and exhaust in service.
l D.
Shutdown purge supply and exhaust in service.
C.
Mini purgo exhaust in service with supply through the equipment hatch.
D.
Shutdown purge exhaust in service with supply through the equipment hatch.
l ANSWER:
B. Shutdown purgo supply and exhaust in service.
K/A #: 029A201 2.9/3.0 KA DESCRIPTION: Shutdown Purge Operation OBJECTIVE #:
0110400 0
REFERENCES:
Tech Specs AUTHOR: PJM SOURCE: NEW L DISTRACTER EXPLANATION:
A. S/D purge is placed in service during Mode 5. Mini purge is removed from service.
B. Tech Spec requirement is to have the equipment hatch closed during fuel movement.
Shutdown purge is placed in service during Mode & Mini purge is removed from service.-- ---- - --
C. Same as A.
D, This lineup is used in Mode S. With refueling in progress, the equipment hatch is required to be closed.
RO Outline #45 4
QUESTION #91 l
Which ONE of the below sets of valvo positions would maintain their associated systems l
OPERABLE to perform their intended function for MODE 1 operation?
EMHV8814A DNV8717 EJHV8840 St Pump A RHR to RHR Train A/B l
Recirc iso RWST Hot Leg Recire
- A.
CLOSED CLOSED CLOSED B.
CLOSED OPEN OPEN C,
OPEN OPEN OPEN D.
OPEN CLOSED CLOSED ANSWER:
D.
OPEN CLOSED CLOSED f
K/A #: 006A402 4.0/3.8 KA DESCRIPTION: ECCS Response to Valve Failure OBJECTIVE #:
0110170G
REFERENCES:
T61.0110,6 LP 17, Page 11 WC ECCS LER Event AUTHOR: FXB SOURCE: NEW HO
~
DISTRACTER EXPLANATION:
None needed See reference RO Outline #41
4 QUESTION #92 The plant is responding to a loss of offsite and onsite power.
During implementation of ECA-0.0, Loss of All A.C., the AC supply breakers to the ESW pumps are:
A.
Placed in Pull-to-Lock on the MCB so they are under positive operator control.
7E.' B.-
Mar $ually closed locally since there is no closing power to the MCB handswitch.-
- - - --- - - 1
~
i C.
Placed in Normal After-Stop on the MCB for start when the DG is started locally.
l D. -
Verified closed locally prior to starting DG or reenergizing the NB bus.
ANSWER:
C. Placed in Normal-After-Stop on the MCB for start when the DG is started locally.
K/A #: 055K302 4.3/4.6 i
KA DESCRIPTION: Actions in ECA 0.0 for ESW Pump OBJECTIVE #:
003D22OL l
REFERENCES:
ECA 0.0
- 1. P. E.
AUTHOR: FXB SOURCE: NEW L DISTRACTER EXPLANATION:
i A.-
- All other NB loads are placed in PTL-B.
Closing Power is still available from NK batteries D. =
~ Closed automatically by shutdown sequencer RO Outline #54
QUESTION #93 -
1.
A plant startup is in progress. Power level is 1E-7 amps. The Reactor is tripped when PA02 is denergized by Relay Test.
Which ONE of the following is the approximate length of time before the Source Range N1's automatically energize ?
A.--
2 minutes B.
3 minutes L
C, 9 minutes D.
11 minutes l
l l
At4SWER:
Dc-11 minutes K/A #: 007EK1.05 3.3/3.8 KA DESCRIPTION: How Long for Scurce Ranges to Energize on Rx Trip l OBJECTIVE #:
003D060E l
AUTHOR: PJM SOURCE: NEW-LO DISTRACTER EXPLANATION:
SUR following a Reactor trip is -1/3 DPM.
Source Range NI's automatically energize at 6E-11.
1E 7 to BE-11 is 3.4 Decades 3.4 Decades at.33 DPM = 10.3 minutes RO Outline #76 L
\\
LQUESTION_#94 With the plant in MODE 3 the MCB handswitch for the Reactor Trip Breakers is rotated to the f
CLOSE position.
-I i
Which ONE of the below conditions would prevent the reactor trip breakers from closing?
'A.
Pressurizer Level at 95%
~
ZBd 1RCS Pies ~sure it3800'psig J '
'~
C, Pressurizer Level at 17%
Di RCS Pressure at 2400 psig -
ANSWER:
D. RCS Pressure at 2400 psig K/A #: 012A406 4.3/4.3 KA DESCRIPTION: Operation of Rx Trip Breakers OBJECTIVE #:
0110270D REFERENCEG: OTO-SA 00001 Table 111 E 0 Attachment 1 AUTHOR: FXB SOURCE: NEW Modified N/A - HO DISTRACTER EXPLANATION:
~
.A-
- Level above trip _setpoint but trip.not-enabled -
B--
. Pressure below trip setpoint but trip not enabled C-Letdown isolation Setpoint Only - No trip RO Outline #35 4'
QUESTION #95 The r$ actor is exactly critical at 10E-8 amps when the "A" S/G PORV fails open, Assuming NO operator action and NO reactor trip, which ONE of the following best _ describes the values of Tavg and nuclear power for the resulting new steady state conditions?
A.
Final Tavg is greater than initial Tavg. Final reactor power is above the POAH.
B.
Final Tavg is greater than initial Tavg. Final reactor power is at the POAH.
C.
Final Tavg is less than initial Tavg. Final reactor power is above the POAH, D.
Final Tavg is less than initial Tavg. Final reactor power is at the POAH.
ANSWER:
C. Final Tavg is less than initial Tavg. Final reactor power is above the POAH, K/A #: 039K508 3.3/3.6 KA DESCRIPTION: Plant Response to S/G PORV Failure OBJECTIVE #:
0110200F
REFERENCES:
SER 9-97,2. on Event -
AUTHOR: PJM SOURCE: NEW - HO DISTRACTER EXPLANATION:
When the S/G PORV opens, the steam flow from the S/G will cool the RCS. This will add positive reactivity, The positive reactivity will cause reactor power to increase. Each S/G PORV provides 2 3% of 100% reactor power steam flow. This will result in reactor power stabilizing at 2-3%. No reactivity was cdded to offset the power defect so Tavg stays lower than initial to compensate keeping the reactor critical.
RO Outline #37 -
.s
i QUESTION #96 The plant is in MODE 2 commencing warmup of the main turbine.
Which ONE of the following could ba a direct result of a loss of Vital AC Instrument Bus NN02.
A.
Charging Pump Suction Swaps to the RWST.
B.
Source Range Hi Flux Reactor Trip.
C.
Intermediate Range High Flux Reactor Trip.
D.
Idle Component Cooling Water Pump Start.
ANSWER:
C. Intermediate Range High Flux Reactor Trip.
K/A #: 057AA219 4.0/4.3 KA DESCRIPTION: Auto Actions on Loss of NN02 l
OBJECTIVE #:
003B450A
REFERENCES:
OTO-NN 00001 IPE Flooding AUTHOR: FXB SOURCE: BANK Modified Y L DISTRACTER EXPLANATION:
A.
NN01 or NN04 B._. Source Range High_ Flux Trip Blocked._
C.
NN01 or NN04 RO Outline #57 lin
QUESTION #97 Which ONE of the following parameters can be used to distinguish between a primary and secondary leak?
>4 A
RCS Tavg -
-Bc Pressurizer Level
_.- _J.C.... Containment Pressure _..
~
D, Pressurizer Pressure ANSWER:
A. RCS Tavg K/A #: 009EA1.01 4.4/4.3 KA DESCRIPTION: Response of Plant Parameters to LOCA OBJECTIVE #:
003D030F
REFERENCES:
T61.003D.6 LP 3 AUTHOR: FXB SOURCE: NEW - LO f-DISTRACTER EXPLANATION:
B.
Level decreases with both C.
CTMT pressure may increase with both -
'F-D.
Pressurizer Pressure may decrease with both '
.80 Outi.ine #Z7 -
-QUESTION #98
- ~ FR-Ci1 in in progress.
- S/G's have been depressurized to 120 psig.
. All RCP's are stopped.
Which ONE of the below is the reason for stopping the RCP's?
A.
Reduce the heat input to the RCS.
_m.._
B.
Normal PZR spray flow is unnecessary.
C.
Prevent cavitation from N2 from the accumulators.
D.
Prevent damage to Number 1 seal.
ANSWER:
D. Prevent damage to Number 1 seal.
-K/A #: E06EK3.1 3.4/3.8 KA DESCRIPTION: Stopping RCP's in FR-C.1 Inadequate Core Cooling OBJECTIVE #:
003D2501
REFERENCES:
T61.003D.6 LP 25 Page 47 AUTHOR: PJM SOURCE: NEW - L DISTRACTER EXPLANATION:
~
A RCP's at1his~oointTnay have~beerrproviding some' forced flow cooling of the core 7
B. RCS pressure control should not be necessary with having just cooled down the plant which would lower pressure.
C. The accumulators should have been isolated in the previous step.
D. Correct. The next step will depressurize the RCS resulting in loss of seal D/P.
RO Outline #64
QUESTION #99 FR P,1,' Response to imminent Pressurized Thermal Shock Condition, is in progress due to a steam leak which has been isolated.
Which ONE of the following conditions is. acceptable using Attachment 7 for RCS Post-Soak Cooldown Limitations during recovery from the PTS condition 7 l
_ _ _ A.
RCS cold legsy 200*F. RCS wide range pressure = 0 psig.
B.
RCS cold legs =_250*F. RCS wide range pressure = 300 psig l
C.
RCS cold legs = 300*F. RCS wide range pressure = 400 psig.
~ D.'
RCS cold legs = 400*F. RCS wide range pressure = 300 psig.-
,1 l
ANSWER:
1 C. RCS cold legs = 300*F. RCS wide range pressure = 400 psig.
K/A #: E08EA1.3 - 3.6/4.0 KA DESCRIPTION: RCS Post Soak C/D Limits Following PTS E
OBJECTIVE #:
003D280E
REFERENCES:
FR P.1 Att. 7 AUTHOR: PJM-SOURCE: BANK - Modified Y - HO DISTRACTER EXPLANATION:-
- ~ -
~ ~--- -~ ~ ~
A.
Outside the allowed band.
B.
Outside the allowed band.
C.
Correct.
D.
Outside the allcwed band.
- RO Outline #73 X
w
Proced. 14 0.
RESPotiSE TO X MMll4EliT Attac.went
- Hey, TR-P.I PRESSURIZED THERF.AL SHOCK 7
COtiDITIOt1 1D)
RCS POST-SOAK C00LDOVil LIMITATIotts CURVE RCS POST-bME C001D0TN INM10NS MEMBLt UC10NISMTTl!K W CITHIS t,
~~
3000 j[/ k
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t S 1000 0 l
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Page 1
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- 4 -
QUESTION #100 A fire is reported in PG24, Pressurizer Variable Heater, Load Center.
This is an example of a--
fire and would be extinguished
. A.
Class B, by pre-action sprinklers.
l
- f,l'_~,Bc -Class C, by halon.
i C.
Class B,- by halon.
-l J
D.
Class C, by pre-action sprinklers.
JANSWER:
f
- {
B. ~ Class C, by halon.
K/A #: 067K101 2.9/3.9 KA DESCRIPTION: Fire in PG24, Type and Suppression l
OBJECTIVE #:
Orientation - Industrial Safety IPEEE - Fire Onsite
REFERENCES:
Industrial Safety AUTHOR: FXB SOURCE: NEW L DISTRACTER EXPLANATION:
A.
Wrong class, wrong protection
-. ~ - -
C.-
Wr6ti'g cinH-
~-
~ ~ ~ ~ ~ ' ~ ~ ~ ~ ~ ~ ~ - -
~~~
D.
Wrong protection RO Outline #67 l
-M