ML20205A441
| ML20205A441 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 12/17/1998 |
| From: | Boyd J NEBRASKA PUBLIC POWER DISTRICT |
| To: | Pellet J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| Shared Package | |
| ML20205A435 | List: |
| References | |
| NTD980383, NUDOCS 9903300381 | |
| Download: ML20205A441 (150) | |
Text
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NTD980383 December 17,1998 John Pellet U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive Suite 400 -
Arlington, Texas 76011
Dear Mr. Pellet:
Subject:
Transmittal ofInitial Licensed Operator Examination Outline and Draft License Applications Enclosed is the initial examination outline for RO written examination to be given the week of February 8,1999. I am the designated NPPD representative / reviewer for this examination.
Mr. Harley McDaniel, Elden Plettner, Terry Borgan, Phil Ballard and Ed Bowles (EXCEL l
Corporation) and I have signed an ODG 210 " Attachment A" pre-examination security I agreement, which I have in file. These examination materials shall be withheld from public disclosure until after the examinations are completed.
Also enclosed are preliminary copies of the license applications for the candidates. l Respectfully, Jeffrey W. Boyd SRO/RO Instructor ,
4 Cooper Nuclear Station pc: S. Blake w/o attach H. McDaniel w/o attach N. Robinson w/o attach NTD file w/o attach i
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I ES-401 BWR RO Examination Outline Form ES-401-2 Facility: Cooper Date of Exam: February 1999 Exam Level: RO 7
K/A Category Points
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Tier Group Point K K K K K K A A A A C Tel 1 2 3 4 5 6 1 2 3 4
- 1. 1 3 2 2 h d 2 2 hh 2 13 Emergency & 2 3 3 3 h 4 3 hM 3 19 p
3 0 1 1 hhh 0 1 hh 1 4 Evolutions Tier 6 6 6 6 6 6 36 1 3 3 2 3 1 2 3 3 3 3 2 28
- 2. 2 2 2 1 3 1 2 1 2 2 1 2 19 3 1 0 0 0 2 0 0 0 0 1 0 4 Systems Tier 6 4 5 5 4 4 4 5 5 5 4 51 Totals _
- 3. Generic Knowledge and Abilities Cat 1 Cat 2 Cat 3 Cat 4 4 3 3 3 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 16 of 39 Interim Rev. 8, January 1997
ES-401 BWri RO Examination Outline Form ES-401-2 Emergency and Abnormat Plant Evolutions - Tier 1/ Group 1 E/ APE # / Name / Safety Function K1 K2 K3 A1 A2 G K/A Topic (s) Imp. Points 295005 Main Turbine Generator Trip / tit 01 Reason for reactor scram 3.8 1 295006 SCRAM / f 05 Ability to monitor / operate neutron monitoring 4.2 1 295007 High Reactor Pressure / Ill 04 Ability to monitor / operate SRVs 3.9 1 295009 Low Reactor Water Level / It 05 Implications of low RPV level and natural circulation 3.3 1 295010 High Drywell Pressure / V 01 Relationship between high drywell pressure and Supp. pool level 3.2 1 295014 Inadvertent Reactivity Addition / I 03 Determine the cause of reactivity addition 4.0 1 295015 Incomplete SCRAM / l 04 Relationship between incomplete scram and RPS 4.0 1 295024 High Drywett Pressure / V x 2.4.1, (nowledge of EOP entry conditions 4.3 1 295025 High Reactor Pressure /111 02 Reason for recire pump trip on high pressure 3.9 1 295031 Reactor Low Water Levet /11 x 2.4.48, interpret and verify status 3.5 1 295031 Reactor Low Water Level /11 01 Knowledge of adequate core cooling 4.6 1 295037 SCRAM Condition Present and Power 07 Knowledge of shutdown margin 3.4 1 Above APRM Downscale or Unknown / I 500000 High Containment Hydrogen Cene. / V 03 Determine combustibility limits for the drywell 3.3 1 K/A Category Totals: 3 2 2 2 2 2 Group Point Total: 13 NUREG 1021 17 of 39 Internc Rev. 8, January 1997
ES-401 BWR RO Examination Outline Form ES-4012 Emergency and Abnormal Plant Evolutions - Tier 1/ Group 2 E/ APE # / Name / Safety Function K1 K2 K3 A1 A2 G K/A Topicts) Imp. Points 295001 Partial or Complete Loss of Forced Core 01 Interrelationship with the recirc system 3.6 1 Flow Circutation / I & IV 295002 Loss of Main Condenser Vacuum / Ill 08 Interrelationship cire water system 3.1 1 295003 Partial or Complete Loss of AC Pwr / VI 01 Monitor / operate AC electrical distribution system 3.7 1 295004 Partial or Complete Loss of DC Pwr / VI O2 Knowledge of redundant supplies 3.2 1 295008 High Reactor Water Level /11 02 Interpret / determine steam flow / feed flow mismatch 3.4 1 295012 High Drywelt Temperature / V O2 Monitor / operate drywett cooling system 3.8 1 295013 High Suppression Pcol Temp. / V . x 2.4.4. Recognize EOP entry conditions 3.9 1 295016 Control Room Abandonment / Vil 03 Reason for disabling control room controls 3.5 1 295017 High Off-site Release Rate / IX 04 Interpret / determine source of release 3.6 1 295018 Partial or Complete Loss of CCW / Vlif 01 Effects on component / system operation 3.5 1 295019 Partial or Complete Loss of Inst. Air / x 2.4.11. Knowledge of AOPs 3.4 1 Vill 295020 Inadvertent Cont. Isotation / V & Vil 01 Monitor / operate PCIS/NSSSS 3.6 1 295072 Loss of CRD Pumps /1 01 Reactor pressure vs. Rod insertion capability 3.3 1 295026 High Suppression Pool Water Temp. / V 05 Reason for reactor scram 3.9 1 295028 High Drywell Temperature / V x 2.4.20, Knowledge of warnings, conditions and notes 3.3 1 295029 High Suppression Pool Water Level / V 01 Peason for emergency depressurization 3.5 1 295030 Low Suppression Pool Water Leve! / V 04 Interrelationship with RHR/LPCI 3.7 1 295033 High Secondary Containment Area 01 Interpret / determine area radiation levels 3.8 1 Radiation Levels / IX 295034 Secondary Containment Ventilation 03 Monitor / operate secondary containment ventilstion 4.0 1 High Radiation / IX K/A Category Pc' Totats: 3 3 3 4 3 3 Group Point Totaf: 19 NUREG 1021 18 Of 39 Interim Rev. 8, January 1997
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ES-401 BWR RO Examination Outline Form ES-401-2 Plant Systems - Tier 2/ Group 1 System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G K/A Topic (s) Imp. Points 201001 CRD Hydraulic 04 Design features, backup scram valves 3.6 1 201001 CRD Hydraulic 03 Effect of malfunction on CRD mechanisms 3.1 1 201002 RMCS O2 Predict impact and respond to rod drift 3.2 1 201002 RMCS O2 Predict / monitor control rod position 3.4 1 202002 Recirculation Flow Control 03 Monitor automatic operation of the scoop 3.1 1 tube 203000 RHR/LPCI:In_iection Mode 01 Pump power supplies 3.5 1 206000 HPCI 12 Turbine trip control 4.0 1 206000 HPCI 01 Electrical power supplies to system valves 3.2 1 209001 LPCS 05 Relationship with ADS 3.7 1 211000 SLC 08 Design features / interlocks with SBLC 4.2 1 control switch 212000 RPS 01 Knowledge of electrical power supplies 3.2 1 212000 RPS O2 Knowtedge of specific logic arrangements 3.3 1 215003 IRM O2 Predict consequences of inoperative IRM 3.5 1 215001 SRM x 2.1.7, Ability to make operational 3.7 1 determination 215005 APRM / LPRM 07 Knowledge of flow biased setpoints 3.7 1 216000 Nuclear Boiler instrumentation 13 Relationship with feedwater system 3.4 1 217000 RCIC 04 Effect of a loss of the CST 3.5 1 218000 ADS O2 Ability to monitor logic 4.2 1 223001 Primary CTMT and Auxiliaries 09 Monitor supp pool temperature 3.5 1 223002 PCIS/ Nuclear Steam Supply Shutoff n 02 Monitor auto operation including valve 3.5 1 closurcs 239002 SRVs 04 Effects of a loss of DC 3.0 1 241000 Reactor / Turbine Pressure Reoulator 06 Effects of a malfunction of the BPVs 4.1 1 NUREG 1021 20 of 39 Interim Rev. 8, January 1997
t 259001 Reactor Feedwater 10 Monitor automatic actions during a pump 3.4 1 trip 259002 Reactor Water Level Control 02 Predict / monitor changes in feed flow 3.6 1
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259002 Reactor Water Level Control 05 Relationship between feedwater and 3.6 ' 1 feedwater control 261000 SGTS O2 Ability to monitor suction vafves 3.1 1-264000 EDGs 10 Predict the consequences of a t.OCA 3.9 1 264000 EDGs , x 2.1.32. Explain / apply system limits and 3.4 1 precautions ;
K/A Category Point Totals: 3 3 2 3 1 2 3 3 3 3 2 Group Point Total: 28 l
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NUREG 1021 21 of 39 Interim Rev. 8, January 1997 ;
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ES-401 BWR RO Examination Outline Form ES-401-2 l Plant Systems - Tier 2/ Group 2 i System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G K/A Topic (s) Imp. Points [
i 201003 Control Rod and Drive Mechanism Of Ability to predict reactor power 3.7 1 201006 RWM 03 Predict the impact of a rod drift 3.0 1
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202001 Recirculation 16 Interlocks for a recire runback 3.3 ' 1 ;
204000 R.WCU 01 Connections with the RPV 3.1 1 !
205000 Shutdown Cooling 08 Effects of a foss of RHRSW 3.5 1 f l
214000 RPIS 03 Ability to monitor proper operation 3.5 1
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215002 RBM x. 2.1.28, Kncwledge of purpose / function of 3.2 1 components or controts 219000 RHR/LPCf: Torus / Pool Cooling Mode 02 Abitity to operate valve lineup 3.7 1 .
226001 RHR/LPCI: CTMT Spray Mode 01 Malfunction of containment spray 3.6 1 f i
239001 Main and Reheat Steam 01 Power supplies to MS!V solenoids 3.2 1
!nterlocks for turbine control 3.1 i 245000 Main Turbine Gen. and Auxiliaries 09 1 262001 AC Electrical Distribution 01 Effects of a loss of AC 3.5 1 1
262002 UPS (AC/DC) 02 Effects of a loss of DC power 2.8 1 f f !
4 263000 DC Electrical Distribution x 2.4.11, Knowledge of AOPs 3.4 1 [
! 271000 Offgas 04 _
impact of high radiation 3.7 1 f
286000 Fire Protection 05 Effect on EDG operation 3.0 1 2
290001 Secondary CTMT 01 Ability to monitor containment isolation 3.9 1 i
- I 3OOON) Instrument Air 02 Effects of a loss of air on vatves 3.3 1 i
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400000 Component Cooling Water 02 Relationship with loads 3.2 1 j L
i Ct/A Category Point Totals: 2 1 3 2 1 2 1 2 2 1 2 Group Point Total: 19 L
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ES -401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-5 Facility: Cooper Date of Exam: February 1999 Exam Level: RO Categorv K/A # Topic Imp. Points 2.1.1 Knowledge of conduct cf operations requirements. 3.7 1 2.1.20 Ability to execute procedural steps. 4.3 1 Operations
_ _2.1.29 Knowledge of how to conduct and verify valve lineups. 3.4 1 2.1.32 ~.bility ,to explain and apply system limits and 3.4 1 precautions.
Total 4 2.2.2 Ability to manipulate controls. 4.0 1 2.2.13 Knowledge of tagging and clearance procedures. 3.6 1 Equipment 2.2.22 Knowledge of LCOs and safety limits.
Control 3.4 1 Total 3 2.3.1 Knowledge of 10CFR20 and related documents 2.6 1 2.3.4 Knowkdge of exposure limits and contamination control. 2.5 1 Radiation 2.3.10 Ability to reduce rad levels and guard against personnel 2.9 1 Control exposure.
Tot 3 2.4.4 Ability to recognize EOP and AOP entry conditions. 4.0 1 1 4.11 Knowledge of AOPs. 3.4 1 j Emer ncy Proce ures and 2.4.13 Knowledge of crew roles and responsibilities during EOP 3.3 1 i Plan use. l Total 3
_ Total 13 NUREG 1021 37 of 39 Interim Rev 8, January 1997
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J ES-401 BWR RO Examination Outline Form ES-401-2 1
l Facility: Cooper Date of Exam: February 1999 Exam level: RO'
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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 3 2 2 //h, 2 2 // /, 2 13 Emergency & 2 3 3 3 // 4 3 3 19 Abnormal Plant Evolutions 3 1 1 h / 1 h') 1 4 Tier 6 6 6 6 6 6 36
, Totals , , ,
1 3 3 2 3 1 2 3 3 3 3 2 28
! 2- 2 2 1 3 2 1 2 1 2 2 1 2 19 2 ""
3 1 0 0 0 2 0 0- 0 0 1 0 4 Systems Tier 6 4 5 5 4 4 4 5 5 5 4 51 Totsis
- 3. Generic Knowledge and Abilities Cat 1 Cat 2 Cat 3 Cat 4 4 3 3 3 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 l
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Systems / evolutions within each group are identified on the associated outline.
The shaded areas are not applicable to the category / tier.
i i
NUREG 1021 16 of 39 Interim Rev. 8, January 1997
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295031 Reactor Low Water Level /11 01 Knomi of adequatecorecooling 4.6 1 t R$AC LbW$A EVYL 4$No 41 295037 SCRAM Condinon Present and Power 07 Know of shutdown m 3.4 1 Above APRM Downscale or Unknown /I Knowl gfggrgggigg gfggconcepgasthey F NO E APRM DOWNSCALE OR UNKNOWN:
Shutdown margin (CFR: 41.8 to 41.10) 500000 High Containment Hydrogen Conc./ V 03 Determine combus 3.3 1 limits for the drfen hlhA hTYI ENT YhFk NNO CNR :
Combustible limits for drvwell (CFR: 41.10 / 43.5 / 45.f3)
K/A Category Totals: 3 2 2 2 2 2 Group Point Total: 13 NUREG-1021 18 of 39 Interim Rev. 8, January 1997
ES-401 BWR RO Examination Ouoine Form ES-401-2 Emergency and Abnormal Plant Evolubons - Tier 1/Groe 2 i
E/ APE # / Name / Safety Function K1 K2 K3 A1 A2 G K/A Topic (s) imp. Points 295001 Partial or Complete Loss of Forced Core 01 Interreta ' with the recirc system 3.6 1 Flow Circulation /I & IV Knowledoe of interrelaDons between PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION and the fo!!owing:
Recirculation system (CFR: 41.7 / 45.8) 295002 Loss of Main Condenser Vacuum / til 08 Interrelationship cire water system 3.1 1 Knowle. doe of tne interrelations between LOSS OF MAIN CONDENSER VACUUM and the fol!owing:
g y f45 K 295003 Partial or Complete Loss of AC Pwr / VI 01 Monitor! operate AC electrical distribunon system 3.7 1
$$OYA$LEYLOYOFN.$PYE A.C. electncal distnbution system (CFR: 41.7 / 45.6) 295004 Partial or Complete Loss of DC Pwr / VI 02 Knowledge of redundant s@pties 3.2 1 Knowledoe of the operationalimolications of the followino concepts as they a_pply to PARTIAL Oli COMPLETE LOSS OF D.C. POWER:
Hedundant D.C. power supplies: Plant-Specific (CFR:41.8 to 41.10) 295008 High I?eacter Water Level /11 02 Int et/ determine steam flow / feed flow mismatch 3.4 1 g gefgyR g[nterpret the fotlowing as they apply to HIGH Steam flow /feedflow mismatch (CFR: 41.10 / 43.5 / 45.13) 295012 High Drywell Temperature / V 02 Monitor / operate drywell cooling system . 3.8 i Ability to operate and/or monitor the following as they apply to HIGH DRYWELL TEMPERATURE:
CYR k 295013 High Suppression Pool Temp. / V x 2.4.4, Recognize EOP entry condtions 3.9 1 Ability to recognize abnornial indications for system operati,no parameters whacli are entry-level condiDons for emergency and abnormaIoperating procedures.
295016 Control Room Abandonment / Vil 03 Reason for disabling control room controis 3.5 1 L C IIAfkA D N ElIT CN:41.Y4 6 295017 High Off-site Release Rate / IX 04 Int et/ determine source of release 3.6 1 Abili to determine and/or interpret the following as they apply to HIGH ,
OFF ITE RELEASE RATE: >
tSource of off-site release fCFR:41.10/43.5/45.13)
NUREG-1021 19 Of 39 Interim Rev. 8, January 1997
295018 Partial or Cornpiete Loss of CCW / Vill 01 Effects on cv., -
io anon 3.5 1 CNS-Respond to Loss of REC Pumps Knowl e of t al.Eona..
era l heations of the follow cone gg ARTIAL R COMPL E LOSS OF COMPON NTNGCgas they ,
ICA'*4Tig"G'Y** P""" ""
295019 Partial or Complete Loss of inst. Air / Vill x 2.4.11. Knowledge of AOPs 3.4 1
- Knowledge cf abnormal condition procedures.
tM Inadvertent Cont. Isolation / V & Vil 01 Monitor / operate PCIS/NSSSS 3.6 1 NERE COYAN$ T SONIOl:'
PCIS/NSSSS <
(CFR:41.7/45 6) 295022 Loss of CRD Pumps /I 01 Reactor essure vs. Rod insernon capabilit - 3.3 1 Knowl gggtg im hcations oythe following concepts as they ka tor pressure vs. rod insertion capability (CFR: 41.8 to 41.10) 295026 High Suppression Pool Water Temp. / V 05 Reason for reactor scram 3.9 1 P S PNL$GNATE b P N TURE$
Reactor SCRAM (CFR:41.5/45.6) 295028 High Drywell Temperature / V x 2.420. Knowledge of warnings, conditions end notes 3.3 1 Knowledge of operanonal imphcations of EOP warnings / cautions / and notes.
295029 High Suppression Pool Water Levet / V 01 3.5 1 ;
Reason for emergency yessurization YtMPRESSIEPOO A N '
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295030 Low Suppressbn Pool Water Level / V 04 Interrelationship with RHR/LPCI 3.7 1 Knowledge of Ine interrelations between LOW SUPPRESSION POOL WATER LEVEL and the following:
RHR/LPCI '
(CFR: 41.7 / 45.8) 295033 High Secondary Containment Area 01 Int et/ determine area radation levels 3.8 1 SE'bO DN N ON N EA R D TENL EN:
Area radiation levels (CFR: 41.10 / 43.5 / 45.13) 295034 Secondary Containment Ventilation High 03 Monitor / operate secondary w a.;-- 4 ventilation 4.0 1 b DNY ONentNN V N LYT ventitation H A ION:
ggta -
K/A Category Point Totals: 3 3 3 4 3 3 Group Point Total: 19 r
k NUREG-1021 20 of 39 Interim Rev. 8, Jarmary 1997
ES-401 BWRRO tion Outline Form ES-401-2 Emergency and Atmrmal Evolunons -Tier 1/Grote 3 E/ APE # / Name / Safety Funcuon K1 K2 K3 A1 A2 G K/A Topic (s) Imp. Points 295021 Loss of Shutdown Cooling /IV x 2.4.9. goflowy g AHR) mhtion strategies.
g 3.3 1 or loss 295023 Refueling Accidents / Vill 04 Interpret / determine the occurrence of an accident 3.4 1 n
fEkktkeECggenterpret the following as they apply to fC NO/43 5 )
295032 Hg Containment Area 03 3.8 1 Reason for isola 5ng affected gems ONRYbN NM i ARM N YRNTU iC1*e"fi 5 f?S'i>$v$t***
ry Containment High Sump / Area 01 (nterrelations cf wgerr i b een SfCbN CONTAINMENT HIGH P/ AREA WATER LEVEL and the following Secondary containment equipment and floor drain system (CFR:41.7 / 45.8) i K/A Category Point Totals: 1 1 1 1 Group Point Total: 4 NUREG-1021 21 of 39 Interim Rev. 8, January 1997
ES-401 BWR RO Examhmtion Outline Form ES-401-2 Plant Systems - Tier 2, Group 1 System # / Name K K K K K K A A A A G K/A Topic (s) Imp. Points 1 2 3 4 5 8 1 2 3 4 201001 CRD Hydraulic 04 gfeeg,ggrgges 3.6 1 HYDR IC SYSTEM desion feattre(s and/or interlocks which provTde for the )
following:
Scrammina control rods with inoperative SCRAM solenoid vatves (back-up SCRAM CYR 1.7) 201001 CRD Hydraulic 03 Effect of malfunction on CRD ir-d.= a 3.1 1 m u f NO TNOl k$) RIVE HYDRAULIC SNTEM will have mwing:
Cgoigg echanisms 201002 RMCS 02 Predict impact and respond to rod drift 3.2 1 fI REA T$ WAf CONTk SYSTEM : and (b) based on those on 01, rn a[the oN NIsof those od f akarm (CFR: 41.5/45.6) 201002 RMCS 02 Predict / monitor control red position 3.4 1 Ability to predict and/or monstor changes in EA hl controls includi :
A.C TN$YTM C 4 SNS SY tM Recirculanon Flow Control 03 Monitor automa5c gation of the tube 3.1 1 EbChlA IN FLOkChNYFWL ,
SYSTEM includig:
FY .b b) 203000 RHR/LPCt:Injecuon Mode 01 Pump power stcpiles 3.5 1 Knowl. edge of electncal power supplies to the following:
CN: 41.7) 206000 HPCI 12 Turbine trip control 4.0 1 CNS-Respond to HPCI Trips and Isoladons Ability to manualty operate and/or monitor in the control room:
C 4 5 5 to 45 )
L NUREG-1021 22 of 39 Interim Rev. 8, January 1997
tyIrmate HPCI edg 7e elec pN t@pliestYlhe System alves: BWR-2,3,4 (CFR: 41.7) 209001 LPCS 05 RelatonsNp with ADS, 3.7 1 YR follow kR [TE and the C 412 4. . o .
SLC 08 211gEMWe U E W AMS features / Interlocks with SBLC control 42 1 Enowledge of STANDBYlaQUID CONTROL SYSTEM desian feature (s) and/or anterlocks which providelor the following:
System initiation upon operation of SBLC -
Control switch (CFR: 41.7) 212000 RPS 01 of electrical power st@ plies 32 1 Knowl of electncal power st@plNr3 to the R rhor-generator sets (CFR:41.7) f oper onaEirnp of EANT N R N C lhN S :
( : 1 / 4Y3 215003 IRM 02 Predict consequences of inoperative IRM 3.5 1 fI i NkDNE RANGE MONINA (IRM) SYSTEM : and (b) based on those edictions, use procedures to correct. ,
Yeor or oper N
( R If5 5 i
, 215004 SRM X 2 to make operational 3.7 1 Abihty to evaluate plant performance and i make operationell ments based on l Nirber terhetath 215005 APRM / LPRM 07
{nowl _R P RANGE MONIT AL POWER RANGE MONIT TEM design feature (s and/or t interlocks which provide For the follow)ing:
gR bi e tnp setpoints -
NUREG-1021 23 of 39 Interim Rev. 8, January 1997
__ __ .-- _ _ _ - _ _ _ . - _ _ _ - _ _ _ - - _ - - - _ - - _ _ _ _ - - _ _ _ _ - _ . - _ - - -__N
216000 Nuclear Boiler instrumentauori 13 Relationship with feedwater system 3.4 1 Know e of the physical connecnons and/or
$UNE$R bL R$WRUMENATION and the following:
Feedwater system (CFR:412 to 41.9 / 45.7 to 45.8) 217000 RCIC 04 Effect of a less of the CST 3.5 1 K9g of the effect that a loss or kEACTOR C EINOL N RCIC) :
LW l gY (CFR: 41.f/ 4$.W 218000 ADS 02 Abmty to monitor logic 42 1 NUREG-1560 inhibit ADS Ability to manuaHy operate and/or monitor in the control room:
fCFR: . 4 to 45.8) 223001 Primary CTMT and Auxmaries 09 3.5 gggtur 1 SWE AND blMARECktY AUXILIAR E controlsincluding:
Par *ri.57PS%'*""*ta 223002 PCIS/ Nuclear Steam Supply Shutoff 02 Monitor auto operation including valve 3.5 1 AY ffT$ NN NI SYSTEM / NUCLEAR STEAM SUPPLY SHUT-Of OFF including:
Valve closures (CFR:41.7/45.7) 239002 SRVs 04 Effects of aloss of DC 3.0 1 Know e of the effect that a loss or REUEF/SA E :
'ChN15/
241000 Reactor / Turbine Pressure Regulator 06 Effects of a malfunc6cn of the BPVs 4.1 1 Knowledge of the effect that a loss or ENEtEE UL t will have on following:
hk41 5.4) 259001 Reactor Feedwater 10 Monitor automatic actions during
$YMO717E'[DVfAY[R S[ST'EM including:
g tr 3.4 1 (CN:W7 / 45.7) 1.
NUREG-1021 24 of 39 Interim Rev. 8, January 1997
- - _ __ =-_
1 259002 Reactor Water Level Control 02 Prm& i/nr0Ntor chan 3.6 1 Abilg edtet and,ges or morntorin feedes inflow kACTORNAT SYSTEM controts including:
$ N T OL Reactor feedwater f'av (CFR: 41.5/45.5) 259002 Reactor Water Level Contro 05 R 'een feedwater and 3.6 1
$*IC$$'5inEENRL5 SYSTEM and the following:
Reactor feedwater s 41.9 / 45.7 to 45.8) ystem (CFR: 412 to 261000 SGTS 02 Abill to monitor sucson valves 3.1 1 Abili to manuaHy operate and/or monitor in the c of room:
Suction vatves (CFR: 41.7 / 43.5 to 45.8) 264000 EDGs 10 Predct the c of a LOCA 3.9 1 CNS-Respond to Loss of AC Power - Use of Ability.to (a) the imDacts of the Emergency AC Power followir!g EMERGENCY GENEHA (DIESEUJET): and (b) based on those predictions, use procecures to
[t LOCA e $c ft orope$r N.
(CFR:41.5/45.6) 264000 EDGs x 2.1.32 Explain / apply system limits and 3.4 1 Ab ty to explain and apply system limits and precautrons.
K/A Category Point Totals: 3 3 2 3 1 2 3 3 3 3 2 Group Point Total: 28 i
NUREG-1021 25 of 39 Interim Rev. 8 January 1997 i
ES-401 - BWR RO Examinanon Ouunne Form ES-401-2 P: ant Systems -Tw 2Karoo p 2 System # / Name .K K K K K K A A A A G K/A Topic (s) Imp. Points 1 2 3 4 5 6 1 2 3 4 201003 Control Rod ard Drtve Mechanism 01 Ability to predict reactor power 3.7 1 Abd to preoict and/or monitor es in bSTkOEWA DR VN! C SM controfs including:
C'FN: 4W/55.5) 201006 RWM 03 Predict the impact of a rod drift 3.0 1 ,
f i NdINNhElN IZER SYbEl(RWH)(PLAN SPECIFIC (b) based on those predicDons, use ) ; and procedures to correct, control, or mitigate the consequences of those abnormal condiDons o I h (Not-BWR6)(CFR: 41.5/
45.6) 202001 Recirculation 16 Interiocks for a recire runback 33 1 Knowledge of RECIRCULATION System ,
provide for the fo)llowing: design feature (s and/or interlocks which Recirculation pump downshift / runback: Plant-fN 41.7) 204000 RWCU 01 Connections with the RPV 3.1 1 Knowl e of the hyscalconnections and/or RE$bfOh1[INER CL UNSTEM and e7sel (CFR: 412 to 41.9 / 45.7 to 45.8) 205000 Shutdown Cooling 08 Effects of a loss of RHRSW 3.5 1 Knowledge of the effect that a foss or malfunction of the follow ill have on the D MOD PCFR"*iT% int "*"t-Sa# ,
210300 RPIS 03 Ability to monitor proper operation 3.5 1 dD POSlTYlINFSM$TTdf TN includig: functioning / operability gfgigfgr 215002 RBM x 2.1.28. edge on of 32 1 I Know edge of the purpose and function of maior system comconents and controls.
NUREG-1021 26 of 39 Interim Rev. 8 January 1997
=
l Mode ~ 02 valv 3.7 hlkgRHR/LPCI:T g i Suppression ochng the control room:
C :D/ 45.5 to 45.8) 220CD1 RHR/LPCI: CTMT Spray Mode 01 3.6 1 Malfuncson of containment g m of theNHR/L%I:
C fNMENT SPRAY SYSTEM MODE will C"oaan" ret"a@ews.,r-on cnenur YCEII.7 / 45.4) 239001 Main and Reheet Steam 01 Power suppNes to MSIV solenoids -
3.2 1 Knowledge of electricsl power supphes to the a m isolation valve solenoids 245000 Main Turbine Gen. and Auxiliaries 09 interiocks for turbine 3.1 1 AN7A L Y YS e()
gor interlocks which provide the 41 tor e a Loss of AC of the gtglg DISTRIBUTION will have on following:
CNNN4 262002 UPS (AC/DC) 02 Eff ts foes ofgg 2.8 1 0 NNRRU ABL PO NS I PYY I e h (CFR: 41.7 / 4her 88= oc ""'ac=' "'"'"" a "
,6*A%a""!* d mhon p,ocedu,es: ' '
e 271000 Offgas 04 3.7 1 g of p hFGAS STEM : a nd (b) followigon ect, con , or 'the c es of thoseabnormalco bons or oper E4ESN5 2ee rire ermanon 05 3.0 i p, =t g g g g g ,,,, ,,,,,,,,,, ,
I"RoTECEN kT
?CT: Es'*/E'm NUREG-1021 27 of 39 Interim Rev. 8, January 1997
290001 Secondary CTMT 01 3.9 1 Abig to monitor contakwnent isolation YD dONT EY udi keco)dary containment isolation (CF R):1.7 /
n 45.7) 300000Instrurnent Air 02 Effects of aloss of aironvalves 3.3 1 Ifulc E f the N TNNE kR SYSTEM)
Systems naving will have on thevalves pneumate following:
ari d (N*41.7 / 45.6) 400000 Component Cooling Water 02 Relations with loads 32 1 the ohysical connections and /
or ca ect relationships between CCWS and the following:
Loads cooled by CCWS
_ (CFR: 412 to 41.9 / 45.7 to 45.8) 1 K/A Category Point Totats: 2 1 3 2 1 2 1 2 2 1 2 Grote Point Total: 19 s
NUREG-1021 28 of 39 Interim Rev. 8, January 1997
2 i
ES-401 BWR RO Examinmuon Ouulne Form ES-401-2 Plant Sysoms -Tier 2A3nn p 3 i
System #/Name .K K K K K K A A A A G K/ Atopic (s) Imp. Points '
1 2 3 4 5 6 1 2 3 4 [
215001 Traversing in-core Probe 05 M 1 ~
K c
d W mh & '
ect rela between ;
TRAVERSING IN OR PROBE and the i following: '
Primary containment isolation system: (Not- -
BWR1)41.2 to 41.9 / 45.7 to 45.8)
(CFR:
234000 Fuel Handhng Equipment 02 3.1 '
g impurasans of refusung 1 Knowledge of the operagonalimphcanons of y
AN hNkN l&O'M"a*3fa'"*"""'"'"" i 268000 Radweste 01 Monitor stsnp integrators 3.4 1 !
Abihty to manually operate and/or morwtor in r the control room:
- /YS to 45.8) [
L 288000 Plant Venglation 02 Operamonalimpucagons of DP coreal 3.2 1 i Knowledge of the operabonalimplicahons of f
~
P T Ek STN SY -
Differential pressure control (CFR: 41.7 / j 45.4) i i
I t
[
f t
K/A Category Point Totats: 1 2 1 Grote Point Total: 4 i e
i 1 !
i l NUREG-1021 29 of 39 Interim Rev. 8 January 1997
______.______________i
ES - 401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-5 Facility: Cooper Date of Exam: February 1999 Exam level: RO Category K/A # Topic Imp. Points 2.1.1 Knowledge of conduct of operations requirements. 3.7 1 Knowledge of conduct of operations requirements.
(CFR: 41.10/45.13)
OCP t ns 2.1.20 Abilit 4.3 1 Ability {o execute procedure stepsto execute procedural steps.
(CFR: 41.10/43.5/45.12) 2.1.29 Knowledge of how to conduct and verify valve lineups. 3.4 1 Knowledge of how to conduct and venty valve lineups.
(CFR: 41.10/45.1/45.12) 2.1.32 Ability to explain and ply system litnits and precautions. 3.4 1 Abihty to explain and app system hmits and precautions.
(CFR: 41.10/43.2/45.1 ) j Tomi 4 2.2.2 Ability to manipulate controls. 4.0 1 Abihty to manipulate the console controls as required to operate l the f acihty between shutdown and designated power levels. i 1
(CFR: 452) 1 Equipment 2.2.13 Knowledge of taggin oced 3.6 1 I Control fcN'N8 yf' 83'hy clearance 3
agandproc clearancegures. ures.
2.2.22 Knowledge of LCOs and safety limits. 3.4 1 Knowledge of hmiting conditions for operations and safety hmits.
(CFR: 432/452)
Total 3 2.3.1 Knowledge of 10CFR20 and related documents. 2.6 1 Knowledge of 10 CFR: 20 and related facihty radiation cor.'.rol requirements.
(CFR: 41.12/43.4.45.9/45.10)
Radiation 2.3.4 Knowledge of exposure limits and contamination control. 2.5 1 Knowledge of radiaDon e osure hmits and contamination control / !
Control including permissible leve s in excess of those authonzed. l l
(CFR: 43.4/45.10) I 2.3.10 Ability to reduce rad levels and guard against personnel 2.9 1 exposure.
Abihty to perform procedures to reduce excessive levels of radiation and guard against personnel exposure.
(CFR: 43.4/45.10)
Total 3 2.4.4 Ability to recognize EOP and AOP entry conditions. 4.0 1 CNS- Abihty to recoonize abnormalindication= for system operating Stauan Bl** U"' parameters which are eng"-level
conditions for emergency and Emergency Procedures and
" *8@/*$"]/$'6) hFn: 4 Plan 2.4.11 Knowledge of AOPs. 3.4 1 Knowledge of abnormal condition procedures.
(CFR: 41.10/43.5/45.13) 2.4.13 Knowledge of crew roles and responsibilities during EOP 3.3 1 use.
Knowledge of crew roles and responsibikties during EOP flowchart use.
(CFR: 41.10 / 45.12)
Total 3 Total 13 NUREG 1021 37 of 39 Interim Rev. 8. January 1997
-n .,
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- :and A BWR RO Eneminado..JuSne Plant Evolueens *. Tier e 1/Grou%'1%+d W' ES-401 s o - < - -
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. . . , . m.. 4 2' #x K/A,Topicte
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- ~~~ ~s.4~ ~"**g' E/ APE # 1 Name / Safety Function
- K14 *- K2 e "K37 ^A1- *A2' LG: #,
F ~ bima,t? INx._W 295005 Main Turbine Generator Trip / Ifl 01 _
Reason for reactor scram 3.8 1 395006 SCRAM / I 05 ' Ability to monitor / operate neutron moneoring 4.2
- 1 295007 High Reactor Pressure / I!! 04 Ability to monitor / operate SRVs 3.9 1 295009 Low Reactor Water Level / II 05 Implications of low RPV level and natursi circulation - 3.3 1 295010 High Drywell Pret..ure / V 01 Relationship between high drywell pressure and Supp. pool level 3.2 1 A95014 Inadvertent Reactivity Addition / I 03 Determine the cause of reactnnty addition 4.0 1 295015 incomplete SCRAM / I 04 Relationship between incomplete scram and RPS 4.0 1 295024 High Drywell Pressure / V x 2.4.1, Knowledge of EOP entry conditions '
4.3 1~
295025 High Reactor Pressure / Ill 02 Reason for recire purnp trip on high pressure 3.9 1 295031 Reactor low Water Level /11 x 2.4.48, Interpret and verify status 3.5 1 R95031 Reactor Low Water Level /11 01 Knowledge of adequate core cooling 4.6 1 295037 SCRAM Condition Present and Power 07 Knowledge of shutdown rnargin 3.4 1-Above APRM Downscale or Unknown / I 500000 Hit Containment Hvdrogen Conc. / V 03 Determine combustibility limits for the drywell 3.3 1' I t
I 1 {
KIA Category Totals: 3 2 2 2 2 2 Group Point Total: 13 :
, )
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- _ _ _ - . _ ~ - - - - - - . . _ _ _ - - _ _ - - - - - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - . _ _ _ _ _ _ - _ _ - - _ .
4 ,
. I w: ,. _ - ~ -
ES-401 s
.. J BWR RO Examinadon Oudine . . E . - > Form ES-401-2 !
- Emor pncy and Abnormal Plant Evolutions - Tier 1/ Group 2 x
~
0 E/ APE # / Nemo'/ Sofety' Function K1' =K2: K3i A11 A2' -G 7 K/A TopleN)'
- lenp." 2 Point [
t i 295001 Partial or Complete Loss of Forced Core 01 Interrelationship with the recirc system 3.6 1 Flow Circulation / i & IV i !
295002 Loss of Main Condenser Vacuum /111 08 Interrelationship cire water system 3.1 1 R95003 Partial or Complete Loss of AC Pwr / VI _ 01 Monitor / operate AC electrical distribution system 3.7 1 295004 Partial or Complete Loss of DC Pwr / VI 02 Knowledge of redundant supplies 3.2 1 f
, 295008 High Reactor Water Level /11 02 Interpret / determine steam flow / feed flow mismatch 3.4 1 l 4
295412 High Drywell Ternperature / V 02 Monitor / operate drywell cooling system 3.8 1 295013 High Suppression Pool Smp. / V x 2.4.4, Recognize EOP entry conditions 4.0 1 295016 Control Room Abando ent / Vil 03 Reason for disabling control room controls 3.5 1 t
295017 High Off-site Release Rate / IX 04 Interpret / determine source of release 3.6 1 l 295018 Partial or Complete loss of CCW / Vill 01 Effects on ecmponent/ system operation 3.5 1 295019 Partial or Complete Loss of inst. Air / Vill x 2.4.11, Knowledge of AOPs 3.4 1 :
i !
295020 Inadvertent Cont. Isolation / V & Vil 01 Monitor / operate PCIS/NSSSS 3.6 1
^
295022 Loss of CRD Pumps / I 01 Reactor pressure vs. Rod insertion capability 3.3 1
~
- 295026 High Suppression Pool Water Temp. / V L
~ ~
01$ Reason for normeWemergency depres' s urizadon5(Ses K/A .; 3.s 2 1c
~ ~
changes)i 295028 High Drywell Temperature / V x 2.4.20, Knowledge of warnings, conditions and notes 3.3 1
.. . . .. s .
. ... . .4 296029 High Suppression Pool Water Level / V i. . . . . .013 Reason for resetor scram '(See KIA changes) s 3.5: h1 ~
i 295030 Low Suppression Pool Water Level / V 04 interrelationship with RHR/LPCI 3.7 1 !
295033 High Secondary Containment Area 01 Interpret / determine area radiation levels 3.8 1 Radiation Levels / IX 6 i
295034 Secondary Containment Ventilation High 03 Monitor / operate secondary containment ventilation 4.0 1
- Radiation / IX i t
K/A Category Point Totals: 3 3 3 4 3 3 Group Point Total: 19 I 4 ,. !
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. ES-401 BWR RO ExamkatioriOutline Form ES-401-2
' Plant Systems - Tier 2/ Group 1
- System # / Name - LK1 K2- K3- K4 K5 K6' A1 A2 ' A3 - ~A4 G : K/A Topic (s)-- : Imp.' ' Points 201001 CRD Hydraulic 04 Design features, backup scram valves 3.6 1 201001 CRD Hydraulic 03 Effect of malfunction on CRD mechanisms 3.1 1 201002 RMCS O2 Predict impact and respond to rod drift 3.2 1 201002 RMCS 02 Pred.ct! monitor control rod position 3.4 1 202002 Recirculation Flow Control 03 Monitor automatic operation of the scoop 3.1 1 tube 203000 RHR/LPCf: Injection Mode 01 Pump power supplies 3.5
- 1 206000 HPCI _.
12 Turbine trip cortrot 4.0 1 206000 HPCI 01 Electrical power supplies to system valves 3.2
changes) -
211000 SLC 08 Design features / interlocks with SBLC 4.2
- 1 control switch 212000 RPS 01 Knowledge of electrical power supplies 3.2 1 212000 RPS O2 Knowledge of specific logic arrangements 3.3 1 5 215003 IRM 02 Predict consequences of inoperative tRM 3.5 1 215004 SRM x 2.1.7, Ability to make operational 3.7 1 1 determination 215005 APRM / LPRM 07 Knowledge of flow biased setpoints 3.7 1 216000 Nuclear Boiler Instrumentation 13 Relationship with feedwater system 3.4 1 217000 RCIC 04 Effect of a loss of the CST 3.5 1 218000 ADS O2 Ability to monitor logic 4.2
- 1 223001 Primary CTMT and Auxiliaries 09 Monitor supp pool temperature 3.5 1 223002 PCIS/ Nuclear Steam Supply Shutoff 02 Monitor auto operation including valve 3.5 1 closures l
239002 SRVs 04 Effects of a loss of DC 3.0 1 !
241000 Reactor / Turbine Pressure Reautator 06 Effects of a malfunction on the BPVs 4.1
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ES-401 BWR RO Examination Outtine Form ES-401-2 Piant Systems - Tier 2/ Group 2 l
System # / Name' 'Ki K2 K3 K4 K5 K6 A1 A2 A3 A4 G' / K/A Topic (s) ~ Imp. Points i 201003 Control Rod and Drive Mechanism 01 Ability to predict reactor power 3.7 1 201006 RWM 03 Predict the impact of a rod drift 3.0 1 202001 Recirculation _
16 Interlocks for a recire runbeck 3.3 1 2O_4000 SWCU 01 Connections with the RPV 3.1 1 205000 Shutdewn Cooling 08 Effects of a loss of RHRSW 3.5 1 214000 RPIS 03 Ability to monitor proper operation 3.5 1 215002 RBM x 2.1.28, Knowledge of purpose / function of 3.2 1 components or controls 219000 RHR/t PCI: Torus / Pool Cooling Mode 02 Ability to operate valve lineup 3.7
- 1 226001 RNR/t.PCI: CTMT Spray Mode 01 Malfunction of containment spray 3.6 1 239001 Main and Reheat Steam Oi Power supplies to MStV solenoids 3.2
- 1 245000 Main Turbine Gen. and Auxiliaries - 09 Interlocks for turbine control (See KIA 3.1 1 changes) 262001 AC Electrical Distribution 01 Effects of a loss of AC 3.5 1 262002 UPS (AC/DC) 02 Effects of a loss of DC power 2.8 1 263000 DC Electrical Distribution x 2.4.11, Knowledge of AOPs 3.4 1 271000 Offgas 04 Impact of high radiation 3.7 1 286000 Fire Protection 05 Effect on EDO operation 3.0 1 290001 Secondary CTMT _
01 Ability to monitor containment isoletion 3.9 1 300000 instrument Air 02 Effects of a loss of air on valves 3.3 1 400000 Component Cooling Water 02 Relationship with loads 3.2 1 5
K/A Category Point Totals: 2 1 3 2 l 2 1 2 2 1 2 Group Point Total: 19 7
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Facility: Cooper Date of Exam: February 12,1999 Exam Level: RO Category - K/A # Topic - Imp. Points l l
2.1.1 Knowledge of conduct of operations requirements. 3.7 1 2.1.20 Ability to execute procedural steps. 4.3 1 l
Conduct of Operations 2.1.29 Knowledge of how to conduct and verify valve lineups. 3.4 '
1 2.1.32 Ability to explain and apply system limits and precautions. 3.4 1 Total 4 2.2.2 Ability to manipulate controls. 4.0 1 2.2.13 Knowledge of tagging and clearance procedures. 3.6 1 Equipment 2.2.22 Knowledge of LCOs and safety limits. 3.4 1 Control Total 3 2.3.I Knowiedge of 10CFR20 an.1 related documents 2.6 1 2.3.4 Vnerwiedge of exposure limits and contamination control. 2.5 1 l
l Radiation 2.3.10 Ability to reduce rad levels and guard against personnel 2.9 l 1
Control exposure. i l
Total 3 2.4.4 Ability to recognize EOP and AOP entry conditions. 4.0 1 2.4.11 Knowledge of AOPs. 3.4 1 Emergency Procedures and 2.4.13 Knowledge of crew roles and responsibilities during EOP 3.3 1 Plan use.
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Total 3 Total 13 l l
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- L Topic ' . K/A # L hap? Reason i Original 295026 High Suppression Pool Water Terr perature / V K3.05 3.9 This K/A duplicates one from the original NRC written examination. After review, it has been determined that the K/A is so specific that duplication of ;
Reason for reactor soem the originc! NRC exam question would occur. Therefore,the K/A ve changed to E/ APE # 295026. K3.01, the renson for normal / emergency depressurtration on high t New 295026 High Suppression Pool Water Temperature / v K3.01 3.8 suppressim pM temperature R is beGewed that this change wRI not effect the overaB j sample plan and wiB provide sufficient difference from the previous question. ,
Reason for normat/ emergency depressurization ,
i Original 295029 High Suppression Pool Water Levet / V K3.01 3.5 ouestion wrRten was determened to be very similar to a question used on the previous NRC exam. Because the KfA is so specific, R is believed that any question written to Reason for emergency depressuriration address this K/A would stiH dupBeste the previouty used question. Therefore, the K/A i we changed to E/ APE # 295029 K3.03, the reason for a reactor scram on high
-New 295029 High Suppression Pool Water Level / V K3.03 3.4 suppressen pmi imi. It is benewed that this change wiR not effect the overed sample plan and win provide sufficient difference from the previous question. l Reason for reactor scram a
Z Original 209001 LPCS Kl.05 3.7 Because of a conflict with a previous examinatio'n another sample plan K/A i change is required. Specificatty K/A 209001, K1.05, Low Pressure Core !
Relationship with ADS Spray System physical connections and/or cause and effect relationship with [
ADS. This K/A was used on the last exam. Because of it's narrow topic dueloping a different from the previous exam is not possible. Therefore to New 209001 LPCS Kl.13 2.8 maintain the current sample plan balance and prevent repeating a question replace K/A 209001 K1.05 with 209001, K1.13. Low Pressure Core Spray f Relationship with leek detection. System physical connections and/or cause and effect relationship with leak detection.
- I
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l Original 245000 Main Turbine Gen. and Auxiliaries K4.09 3.1 After examining the sample plan. R was determined that Tier 2. Group 2, K/A 245000, .r K4.09, be changed to K/A 245000. K4.06, to prevent a double jeopardy problem with Interlocks for turbine contro Tier 2, Group 1, K/A 241000 K3.06. The K/A will st!N provide for a queshon on the i main generator protective relaying, as R relates to plant safety.
New 245000 Main Turbine Gen. and Auxiliaries K4.06 2.7 Interloch for generster protection 10
ES-401 - BWR RO Examination Outline Fomi ES-401-2 l
. A:\SPLNOTE.R4 l-Facility: Cooper Date of Exam: 1999 Exam 12 vel: RO 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
{
- l. l' 3 2 2 h //2 2 h 2 13 I f Emergency & .2 ' 3 3 3 bh 4 3 [/ 3 19
[/ h hh 3 0 1 1 0 1 1 4 Evolutions e. .
Tier 6 6 6 6 6 6 36 Totals 1 3 3 2 3 1 2 3 3 3 3 2 28 1
- 2. 2 2 2 2 1 3 1 1 2 2 1 2 19 Plant 3 1 0 0 0 2 0 0 0 0 1 0 4 Systems Tier 6 4 5 5 4 4 4 5 5 5 4 51 Totals
- 3. Generic Knowledge and Abilities Cat 1 Cat 2 Cat 3 Cat 4 4 3 3 3 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 topic S m many systems; avoid selecting more than two or three K/A topics fiu a gi 'en system unless they relate to plant-specific priorities.
Systems /escudons within each group are identified on the associated outline.
The shaded areas are not applicable to the category / tier.
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- 295005 Main Turtnne Generator Trip / lil 01 Reason for reactor scram 3.8 1 Knowledoe of the reasons for the following responses as they apply to MAIN Tl/RBINE GENERATOR TRIP:
Reactor SCRAM (CFR: 41.5 / 45.6) 295006 SCRAM / I 05 Abili to monitor /ooerste neutron monitoring 4.2
- 1 Abili to operate ardor morutor the following as they apply to SCR : Neutron morutoring system (CFR: 41.7 / 45.6) 295007 High Reactor Pressure /111 04 At!I to monitor /ooerste SRVs 3.9 1 Abi to operate and/or morntor the following as they apply to HIGH REA TOR PRESSURE:
Safety / relief valve operation- Plant-Specific (CFR:41.7/45.6) 295009 Low Reactor Water Level /11 05 Implications of low RPV level and natural circulation 3.3 1 Knowledg9 of the oDerationalimDlications of the followr'ng. Concepts as they apply to LOW REACTOR WATER LEVEL Natural circulation (CFR: 41.8 to 41.10) 295010 High Drywell Pressure / V 01 Relationship between high dryweft presgure and Suco 3.2 1 Knowledge of the interrelations between NGH DRYWEL. pool Ipvel and L PRESSURE the following:
(C 41.7 5 )
295014 Inadvertent Reactivity Addition / I 03 Determine the cause of reactivity addition 4.0 1 Abel!fy to determine and/or enterpret the following as they apply to INADVERTENT REACTIVITY ADDITION:
Cause of reactivity addition (CFR:41.10/43.5/45.13) 295015 incomplete SCRAM / I 04 Relationship between incomplete scram anQ RPS 4.0 1 Knowledge of the interrelations between INCOMPLETE SCRAM and the following:
RPS (CFR: 41.7 / 45.8) 295024 High Crywell Pressure / V x 2.4.1, Knowledge of EOP entry conditiona 4.3 1 Knowfedge of EUP entry conditions and imrnediate action steps.
2950. 3 Mgh Reactor Pressure / lit 02 Reason for recirc pump trip on hiah pressure 3.9 1 Knowledge of the reasons for the fo'llowing responses as they apply to HIGH REACTOR PRESSURE:
Recirculation o tnp: Plant-Specific (CFR:41.5/45.
295031 Reactor Low Water Level / Il x 2.4.48. Interpret and verify status 3.5 1 Abstify to interpret control room indications to verify the status and operation of system / and understand how operator action s and directrves affect plant and system conditions.
295031 Reactor Low Water Level /11 01 Knowledge of adequate core cooling '
4.6 1 Knowledoe of the operational imoheations of the following concepts as they a to REACTOR LOW WATER LEVEL:
te core coolin (C : 41.8 to 41.10)g 3
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N0 IM 295037 SCRAM Condition Present and Power 07 Knowl of shutdown me 'n 3.4 1 Above APRM Downscale or Unknown / I Knowl e of the operationali tions of the followinQ concepts as they to AM CONDITION P ENT AND REACTOR POWER A E PRM DOWNSCALE OR UNKNOWN: Shu'down marain (CFR: 41.8 to 41.10) 500000 High Containtnent Hydrogen Conc. / V 03 Determine combustibility limits for the drywell 3.3 1 Ability to determine and / or interpret the followino as thev I' to HIGH PRIMARY CONTAINMENT HYUROGEN CONCENTRATI N: Combustible limits for drvweil (CFR: 41.10 / 43.5 / 45.f3) K/A Category Totals: 3 2 2 2 2 2 Group Point Total: 13 4
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- Im' p.b ' Points 295001 Partiet or Complete Loss of Forced Core 01 Interrelationship with the recire system 3.6 1 Flow Circulation !l & IV Knowledoe of tt e interretations between PARTIAL OR COMPLETE LOSS '
OF FORCED C' ORE FLOW CIRCULATION and the following: Recirculation system (CFR:41.7/45.8) 295002 Loss of Main Condenser Vacuum / III 08 Interrelationship circ water system 3.1 1 Knowledoe of the interrelanons between LOSS OF MA!N CONDENSER VACUUM and the following: Condenser circulatrng water system
, (CFR:41.7/45.8) 295' ;tial or Complete Loss of AC Pwr / VI 01 Monitor / operate AC electricef distribution system 3.7 1 Abihty to operate and/or monrtor the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER:
A.C. electrical distnbution system (CFR: 41.7 / 45.6) 295004 Partiel or Complete Loss of DC Pwr / VI 02 Knowted of redundant supplies 3.2 1 Knowl e of the operational em.pheations of the followinq concepts as they apply to ARTIAL OR COMPLETE LOSS OF D.C. POWER: Hedundant D.C. power supplies: Plant-Specific (CFR:41.8 to 41.10) 295008 High Reactor Water Level /11 02 Interpret / determine steem flow / feed flow mismatch 3.4 1 Ability to determine and/or interpret the following as they apply to H!GH REACTOR WATER LEVEL: Steam flow /teedflow mismatch (CFR:41.10/43.5/45.13) 295012 High Drywell Temperature / V 02 DRbELNEMPERATURE:Mopitorlopereg f ggi cool ngy flSing as they ap PCW'W/ R"ef*'"" 295013 High Suppression Pool Temp. / V x 2.4.4, Recognize EOP entry conditions 4.0 1 Ability to recograze abnormal indications for system operatina parameters whicli are entry-level condibons for emergency and abnormaToperating procedures. 295016 Control Room Abandonment / Vit 03 Reason for disabling control room controts 3.5 1 Knowle. doe of the reasons for the fo8 towing responses as they apply to CONTROL ROOM ABANDONMENT:
. Disabling control room controls (CFR: 41.5 / 45.6) 295017 High Off site Release Rate / IX 04 inte t/ determine source of release 3.6 1 Abit to determine and/or interpret the fcFowing as they apply to HIGH OF ITE RELEASE RATE:
tSource of off-site release (CFR: 41.10 / 43.5 / 45.13) 295018 Partial or Complete Loss of CCW / Vill 01 Effects on component / system o ation 3.5 1 Knowledaeof theoperat ons of the followina concepts as they aooty to PARTIAL OR COMPL E LOSS OF COMPONENT COOLING WATER: Effects on evnIt.10)m (CFR: 41.8 to 4/ system operations 5
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-Me _ ' imp. - Polms i 295019 Partief or Complete Loss of inst. Air / x 2.4.11. Knowledge of AOPs 3.4 1 Viti Knowledge of abnormal condition procedures.
295020 inadvertent Cont. Isolation / V & Vil 01 Monitor /ooerste PCIS/NSSSS 3.6 1 Abit!?v to ope. rate ansor morwtor the followi as they app!y to INADVERTENT CONTAINMENT ISOLATIOi : PCIS/NSSSS (CFR:41.7/45.6) 295022 Loss of CRO Pumps / t 01 Reactor pressure vs. Rod insertion capability 3.3 1 Knowledoe of tne coerational imphcations of the following concepts as they I to EOSS OF CRD PUMPS: e tor pressure vs. rod insertion capability (CFR:41.8 to 41.10) 295026 High Suppression Pool Water Temp. / V 01 Renao for normat/ emergency depressurization. 3.8 1 OS Knowledge of the reasons for the following responses as they apply to 9:9 SUPPRESSION POOL HIGH WATER TEMPERATURE: Normal /emerger cy depressurization [CFR: 4h5 / 45.6)
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This K/A duplicates one from the original NRC written examination. After review, it has been determined that the K/A is so specific that duplication of thai, original NRC rxam guestion would occur. Therefore. the K/A we changed to E/ APE # 295026, K3.01 the reason for normat/6T.e.goacy depressurization on high suppression pool temperature. It is belteved that this change will not effect the overalt sample >lan and will provide sutticient difference from the previous question. 295028 High Drywell Temperature / V x 2.4.20, Knowledge of warrwngs, conditions and notes 3.3 1 Knowledge of operational imphcations of EOP wamings / cautions / and notes. 295029 High Suppression Pool Water Level / V 03 Reason for reactor scram
- 3.4 1 04- Knowledge of the reasons for the followi responses as they apply to 0-6 HIGH SUPPRESSION POOL WATER L L-Reactor scram (CFR:41 5 /45.6) __
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,am Question written was determined to be v simitar to a question used on the orevious NRC exam. Because the K/A is so gecific, it is believed that any question wrttten to address this K/A would still duplicate the touty used Question. Therefore. fhe K/A we chanced to E/ APE # 295029 K103, me reason for a reactor s6 ram on high suppression pool level. It is believed the this change will not effect the overall samole plan a6d will provide sufSc6 erd derference frorn the previous question 295030 Low Suppression Pool Water Leve! / V 04 interrelationship with RHR/LPCI 3.7 1 KrxrMedoe of the interrelations between LOW SUPPRESSION POOL WAT ER LEVEL and the following:
RHR/LPCI (CFR: 41.7 / 45.8) 295033 High Secondary Containment Area 01 Inte et/ determine area radiation levels 3.8 1 Radiation Levels / IX Abih to determine and/or interpret the followina as they appl SEC NDARY CONTAINMENT AREA RADIATION LEVELS:y to HIGH Area radiation levels (CFR: 41.10 / 43.5 / 45.13) 6
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MYu Imp;' PeInts 295034 Secondary Containment Ventilation High 03 Monitor / operate secondary containment ventitorion 4.0 1 Radiation / IX Ability to operate and/or monitor the following as they to SECONDARY CONTAINMENT VENTILATION HIGH IATION: Secondary containment ventilation ' (CFR:41.7/45.6) , K/A Category Point Totsis: 3 3 3 4 3 3 Group Point Totet: 19 1 I b i P 6 6 i 7 t
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295021 Loss of Shutdown Cooling / IV x 2.4. Knowledge of low wer/ shutdown implications 3.3 1 Know e of low power / imphcations irl accident (e.g. LOCA or loss o RNR) mibgation strategies. 295023 Refueling Accidents / Vit! 04 Interpret / determine the occurrence of en accident 3.4 1 Abthty to determine and/or interpret the following as they apply to REFUELING ACCIDENTS: toccurrence of fuel handling accident (CFR: 41.10 / 43.5 / 45.13) t 295032 High Seconday Containment Area 03 Reason for isolating effected systems' 3.8 1 Temperature / V Knowledle of the reasons for the followina responses as they apply to HIGH StUONDARY CONTAINMENT AREA TEMPERATURE: Isolating affected systems (CFR:41.5/45.6) 295038 Secondary Containment High 01 Interrefetionship with sourpment and floor drain system 3.1 1 Sump / Ares Water Levet / V Knowledoe of the interret# pons between SECCNDARY CONTAINMENT HIGH SUMP / AREA WATER LEVEL and the following Secondary containment equipment and floor drain system (CFR: 41.7 / 45.8) f l K/A Category Point Totsis: 1 1 1 1 Group Point Total: 4 8
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following: Scrammina control rods with inoperative SCRAM so'lenoid valves (back-up SCRAM valves)1.7) (CFR 4 201001 CRD Hydraufic 03 Effect of malfunction on CRO mechanisms 3.1 1 Knowledge of the effect that a loss or malfunction of the CONTROL ROD DRIVE HYDRAULIC SYSTEM will have on following: Control rod drive mechanisms (CFR: 41.7/45.4) 201002 RMCS 02 Predict impact and respond to rod drift 3.2 1 Atxtity to (a) edsci the imDacts of the follows on REACTOR MANUAL CON L SY TEM : and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal concitions or operations: Rod drift alarm (CFR: 41.5/45.6) 201002 RMCS 02 Predict / monitor control rod position 3.4 1 Ability to predict and/or monitor es in parameters associated with opera the REACTOR MANUAL CONTHOL SY TEM controls including: osition Control (CFR: 41.5 rod p/ 45.5) 202002 Recirculation Flow Control 03 Monitor automatic operation of the scoop 3.1 1 tube CU i L T SYSTEM includi : Scoop tube operaaon: BWR-2,3,4 (CFR: 41.7/45.7) 203000 RHR/LPCI: Injection Mode 01 Pump power supplies 3.5* 1 Knowledge of electncal power supplies to the following:
- 41.7) 206000 HPCI 12 Turbine trip control 4.0 1 Ability to manuall the control room:y operate and/or monitor in Turtune tno controls: BWR-2,3.4 (CFR: 41.7 / 45.5 to 45.8) 206000 HPCI 01 Electrical power supplies to system valves 3.2* 1 Knowledge of electncal power supplies to the following:
System vafves: BWR-2,3,4 ' (CFR: 41.7) 9
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- 1 control switch Knowledge of STANDBY LtOU!D CONTROL SYSTEM design feature (s) and/or interlocks which provide for the following:
System initiation upon operation of SBLC control switch (CFR: 41.7) 212000 RPS 01 Knowledge of electrical power supplies 3.2 1 Knowledge of electncal power suppwes to the NErnhgenerator sets (CFR:41.7) 212000 RPS O2 Knowledge of specific logic arrangements 3.3 1 Knowledge of the operational implications of the followino conceots as they to REACTOR PROTECTION SYS : ic arran Specific (CFR: 41.5 log / 45.3) gements 2150031RM 02 Predict consequences of ino rative IRM 3.5 1 Abihty to (a) oredict the ' of the followinc on the INTERM DIATE RANGE MONITOR (IRM) SYSTEM ; and (b) based on those predictions. use procedures to correct. control, or mitigate the consequences of those i abnormal conditions or operations; 1RM inop condition (CFR:41.5/45.6) 215004 SRM x 2.1.7, Ability to make operational 3.7 1 determination Abdity to evaluate fant performance and make operational ments based on operating charact / reactor behavior / and instrument interDretation. 10
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WPoints . _ 215005 APRM / LPRM 07 K 3.7 of flow biaseggints 1 A ER RANGh MbNIT M NIT R SY TEM desian feature (s) and/or interfocks which provide f6r the followtng: Flow biased tnp setpoints (CFR:41.7) 216000 Nuclear Boiler instrumentation 13 Relationship with fesdwater system 3.4 1 Knowledge of the connections and/or cause- enect rela s between NUCLEAR BOILER IN UMENTATION and the following: Feedwater system (CFR:41.2 to 41.9 / 45.7 to 45.8) 217000 RCIC 04 Effect of a loss of the CST 3.5 1 Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR CORE ISOLAHON COOLING SYSTEM (RCIC): Condensate storaae and transfer system (CFR: 41.7 / 45.7J 218000 ADS O2 Ability to monitor logic 4.2
- 1 Abi!ity to manuall the control room:y operate and/or rnonitor in ADS logic initiation (CFR: 41.7 / 45.5 to 45.8) 223001 Primary CTMT and Auxiliaries 09 Monitor supp pool temperature 3.5 1 Ability to predict arxfor monitor cha es in Darameters associated with opera the PRIMARY CONTA!NMENT SYST E AND AUXillARIES controls including:
CN: 1.5 ) 223002 PCIS/ Nuclear Steam Supply Shutoff 02 Monitor auto operation including valve 3.5 1 closures Ability to monitor automatic operations of the PRIMARY CONTAINMENT ISOLATION SYSTEM / NUCLEAR STEAM SUPPLY SHUT-OFF including: Valve ctosures . (CFR:41.7/45.7) 239002 SRVs 04 Effects of a loss of DC 3.0 1 Knowledge of the effect that a loss or malfunction of the followina will have on the REllEF/ SAFETY VALVES': ( N15/ 241000 Reactor / Turbine Pressure Regulator 06 Effects of a malfunction of the BPVs 4.1
- 1 Knowledge of the effect that a loss or malfunction of the REACTOR / TURBINE PRESSURE REGULATING SYSTEM will have on following:
Bypass vatves (CFR: 41.7/45.4) 11
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- 4 7/45.7) 259002 Reactor Water Level Control 02 Predict / monitor changes in feed flow 3.6 1 Abihty to predict and/or monitor che es in carameters associated with operan the REACTOR WATER LEVEL CONTR SYSTEM controls including:
Reactor feedwater flow (CFR: 41.5/45.5) 259002 Reactor Water Level Contro! 05 Relationship between feedwater and 3.6 1 feedwater control Knowledge of the physical connections arw1/or cause-effect relationships between REACTOR WATER LEVEL CONTROL SYSTEM and the following: Reactor feedwater s 412 to 4t.9 / 45.7 to 45.8) ystem (CFR: 261000 SGTS O2 Abili to monitor suction valves 3.1 1 Abi to manuany operate and/or monitor in the c of room:. Succon vaives (CFR: 41.7 / 45.5 to 45.8)- 264000 EDGs 10 Predict the consequences of a LOCA 3.9 1 Abihty to(a) theim acts of the following on EMERGE Y GENERATOR (DIESEUJ : and (b) based on those predictions, use procecures to correct, control, or mr ate the consecuences of thoseabnormalc atrons or operations: LOCA (CFR:41.5/45.6) 264000 EDGs x 2.1.32. Explain / apply system limits and 3.4 1 precautions Ability to explain and apply system limits and precautions. K/A Category Point Totels: 3 3 2 3 1 2 3 3 3 3 2 Group Point Total: 28 1 12
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201003 Control Rod and Drive Mechanism 01 Ability to predict reactor power 3.7 1 Ability to predict aral/or monitor che in Darameters associated with opera the CONTROL ROD AND DRIVE MEC ISM controls including: Heactor power (CFR: 41.5/45.5) 201006 RWM 03 Predict the impact of a rod drift 3.0 1 Abihty to (a) predict the imoacts of the following on the ROD WORTH MINIMlZER SYSTEM (RWH)(PLANT SPECIFIC); and (b) based on those pred6ctons, use procedures to correct, control, or mitigate the consequences of those abnormal condiDons or operations: Rod drift P-Spec (Not-BWR6)(CFR:41.5/ 45.6) 202001 Recirculation 16 Intertocks for a recire runback 3.3 1 Knowledge of RECIRCULATION System provide for the fo)llowing: design feature (s and/or interlocks which Recirculation pump downshift / runback- Plant-( : 41.7) 204000 RWCU 01 Connections with the RPV 3.1 1 Knowledge of the physical connections arxt/or cause-etrect retationshios between REACTOR WATER CLEANUP SYSTEM and the following: Reactor vessel (CFR: 412 to 41.9 / 45.7 to 45.8) 205000 Shutdown Cooling 08 Effects of a loss of RHRSW 3.5 1 Knowledge of the effect that a loss or malfunction of the followina win have on the SHUTDOWN CQQLING SYSTEM (RHR SHUTDOWN RHR service water: COOLING MODE)fic Plant-Foeci (CFR: 413 /45.7) 214000 RPIS 03 Abili to monitor proper operation 3.5 1 Abi monitor automatic operations of the RO SITION INFORMATION SYSTEM including: Venficauon of pro _per functioning / operability (CFR:41.7/453) 215002 RBM x 2.1.28. Knowledge of purpose / function of 3.2 1 components or controls Knowledge of the purpose and function of maior system components and controls. 219000 RHR/LPCI: Torus / Pool Cooling Mode 02 Ability to operate valve lineup 3.7* 1 Abihty to manually operate and/or monitor in the control room: Valve hneup ' (CFR:41.7 / 45.5 to 45.8) 13
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the main be pr generator relevino, as it relates plant 4 262001 AC Electricot Distribution 01 Effects of a loss of AC 3.5 1 DISTRIBUTION will have on following5 of the E T k Mes'or system loads (CFR:41.7/45.4) 262002 UPS (AC/DC) 02 . Effects of a loss of DC nower 2.8 1 Knowledge of the effect that a loss or malfuncnon of the foHowing wm have on the UNINTERRUPTABLE POWER SUPPLY s ' (A.C/D.C.): D.C. electncel power (CFR: 41.7 / 45.7) 263000 DC Electrical Distribution x 2.4.11, Knowledge of AOPs 3.4 1 Knowledge of abnormal condition procedures. 1-14
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followina on the OFFGAS SYSTEll: and based oh those predictions, use procedure s to(b) c ect, control, or miDoate the consecuences e abnormal con &tions or operabons: system hsch radiaDon (C : 4f.5 / 45.61 286000 Fire Protection 05 Effect on EDG operation 3.0 1 Knowledge of the operational implications of the followino concepts as they apply to FIRE PROTECTION SYSTEM: Diesel operations (CFR: 41.5 / 45.3) 290001 Secondary CTMT 01 Abiti o monitor containment isolation 3.9 1 Abili to monitor automaticjperations d the E NDARY CONTAINMLNT (CFR:4 r) containmentisolabon including:1.7/
- 45. .
300000 instrument Air 02 Effects of a loss of air on vahres 3.3 1 Knowledge of the effect that a loss or malfunction of the (INSTRUMENT AIR gYSTEM wil.1 vinghave on the valves fo3 towing: an d ggs pneumatic (CFR: 41.7 / 45.6) 400000 Component Cooling Water 02 Relationship with loads 3.2 1 Knowledge of the physical connections and / or causeffect relanonships between CCWS and the following: LoaJs cooled by CCWS (CFR:41.2 to 41.9 / 45.7 to 45.8) K/A Category Point Totals: 2 1 3 2 1 2 1 2 2 1 2 Group Point Total: 19 15 l
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g fa ik ;k( , , wenwwww w we ac ws ww m wmmamwesen wtar 215001 Traversing in-core Probe 05 Relationship with PCIS 3.3 1 Knowfedgeof the alsw &Ls ard'or cause- enect reta between TRAVERSING IN OR PROBE and the following: Prima containment isolation system: (Not-BWR1 (CFR: 1.2 to 41.9 / 45.7 to 45.8) 234000 Fuel Handling Equipment 02 Operationalimplications of refueling 3.1 1 interlocks Knowledge of the operationalimplications of the follows concepts as they apply to FUEL HANDLIN QUIPMENT: uel handlina ment interlocks FR:41.5/45 268000 Radweste 01 Monitor sump integrators 3.4 1 Ability to manuall the control room:y operate and/or monitor in S intearators ( F : 41.7 / 45.5 to 45.8) 288000 Plant Ventilation 02 Operationalimplications of DP control 3.2 1 Knowledge of the oparationalimplications of the following_ concepts as they anoty to PLANT VEN' 1ILATION SYSTEMS: Differential pressure control (CFR:41.7/ 45.4) K/A Category Point Totals: 1 2 1 Group Point Total: 4 s 16
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Facility: Cooper . Date of Exam: 1999
- Exam Leved: RO I
Category K/A # Topic Imp. Polats 2.1.1 Knowledge of conduct of operations requirements. 3.7 1 Knowledge of conduct of operahons requirements. (CFR: 4f.10/45.13) 2.1.20 a 3 Abilitfo Abihty execute procedure steps.to execute procedural4.3 steps. 1 (CFH: 41.10/43.5/45.12) 2.1.29 Knowledge of how to conduct and verify valve lineups. 3.4 i Knowledoe of how to conduct and verity valve lineups. (CFR: 4f.10 / 45.1/ 45.12) 2.132 Ability to explain and apply system limits and precautions. 3.4 1 Abihty to explain and apply system limits and precautions. (CFR: 41.10 / 43.2 / 45.12) Total 4 2.2.2 Ability to manipulate controls. 4.0 1 Abihty to marupulat; the console controls as required to operate the f acility betwoen shutdown and designated power levels. (CFR: 45.2) Equipment Control 2.2.13 3.6 Knowled *otof tagginband clearancebrocedures. es. 1 np a clearance proe 2.2.22 Knowledge of LCOs and safety limits. 3.4 1 Knowl e of hmibng cond100ns for operations and safety hmits. (CFR: .2 / 45.2) Total 3 2.3.1 Knowledge of 10CFR20 and related documents. 2.6 1 Knowledge of 10 CFR: 20 and related facility radiation control (C : 1 43.4.45.9/45.10) Radiation Control 2.3.4 Knowled 2.5 1 Knowledgfe of er.fon esure oi radsa limits ure hmits andand contamination contaminaDon control /control. ggingr,gle levpin excess of those authonzed. 2.3.10 Ability to reduce rad levels and guard against personnel 2.9 I A ty p'erform procedures to reduce excessive levels of radiation and guard against personnel exposure. (CFR: 43.4/45.10) Total 3 2.4.4 Ability to recognize EOP and AOP entry conditions. 4.0 1 Abikty lo recoonize abnormalindicabons for system operatng parameters which are en evel conditions for emergency and Emergency "CF :
- 1. 3 56)
Procedures and Plan 2.4.11 Knowledge of AOPs. Knowledge of abnormal condition procedures. 3.4 i (CFR: 41.10/43.5/45.13) 2.4.1.1 Knowledge of crew roles and responsibilities during EOP 3.3 I use. Knowledge of crew roles and responsibikties during EOP flowchart NR: 41.10/45.12) Total 3 Total 13 17
- .. . _ . - .- - _ _ . . _ - = - . -__. --
l RO Written Examination Question No.: 62 K/A: 295005 K3.01 Importance: 3.8 Tier: 1 Group: 1 Cognitive Level: 1 Exam Bank No.: new
Reference:
Tech. Spec Bases, Sect. B3.3.1.1 Objective: COR001-14-02, 2.c, 4.b Which one of the following is the bases for a reactor scram on a main turbine trip when reactor power is at 35%7 l a. Protects the reactor from the effects of a loss of heat sink,
- b. Anticipates a reactor power rise due to the colder feedwater.
- c. Ensure the bottom of the RPV steam dryer separator skirt is NOT uncovered.
- d. Provide a backup scram to the RPV pressure and APRM high reactor scrams.
Answer: a.
- b. No, this is a concern below 30% power
- c. Prevents exceeding the MCPR safety limit
- d. These scrams backup the turbine trip scram 4
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l RO Written Examination Question No.: 66 K/A: 295006 A1.05 Importance: 4.2* Tier: 1 Group: 1 Cognitive Level: 1 Exam Bank No.: new
Reference:
2.1.5, Section 8.8 Objective: COR002-12-02,3.a Following a scram from full power when all control rods fully insert, which one of the following meets the requirements of General Operating Procedure 2.1.5, " Emergency Shutdown from Power?"
- a. Fully insert SRM de actors and verify lowering SRM readings.
Fully insert IRM detectors and verify IRM Range 6 to Range 7 overlap.
- b. Fully insert SRM detectors and verify lowering SRM readings.
Fully insert IRM detectors, range IRMs on scale and verify lowering IRM readings. 5
- c. Insert SRM detectors to maintain 10' to 10 cps and verify lowering SRM readings.
Fully insert IRM detectors and verify IRM Range 6 to Range 7 overlap. 5
- d. Insert SRM detectors to maintain 10' to 10 cps and verify lowering SRM readings.
Fully insert IRM detectors, range IRMs on scale and verify lowering IRM readings. Answer: b. Insert both SRMs and IRMs, then check power lowering. Range 6 to 7 overlap 'is done 5 on a startup and is not required per 2.1.5. SRM detectors are withdrawn maintain 10 to 10 cps during a startup. l l l i 2 l l
RO Written Examination Question No.: 30 K/A: 295007 A1.04 Importance: 3.9 Tier: 1 Group: 1 Cognitive Level: 2 Exam Bank No.: new
Reference:
2.4.2.3.3, Sect 2.1, COR002-16-02 Objective: COR002-16-02,5.c The plant is operating at full power when all MSIVs close. All control rod, fully insert into the reactor. Reactor pressure rises to 1093 psig. Assume that the Safety Relief Valves (SRVs) and the Safety Valves (SVs) function at their design set point (* 0.0 psig).
~
Which one of the following describes the response of the nuclear pressure relief system?
- a. Two SRVs will open and then close. NO SRVs cycle. NO SVs open.
- b. Five SRVs and one SV will open and then close. Two SRVs will cycle.
- c. Two SRVs and two SVs will open and then close. Two or three SRVs cycle.
- d. Five SRVs will open and then close. One or two SRVs cycle. NO SVs open.
Answer: d. 5 will open based on SRV setpoints, then pressure will be controlled by one or two of the low-low set SRVs
- a. 5 SRVs will open
- b. No SV opens.
- c. 5 SRVs will open. No SV opens.
3
RO Written Examination Question No.: 15 K/A: 295009 K1.05 Importance: 3.3 Tier: 1 Group: 1 Cognitive Level: 1 Exam Bank No.: new __
Reference:
2.4.2.4.1, Attachment 4 Objective: COR002-22-02, obj. Sh COR002-23-02, obj. 9d The reactor has been shutdown for 18 hours and is currently in Cold Shutdown (MODE 4). A cooldown is in progress with reactor coolant temperature at 162 *F. RHR Loop "A" is in Shutdown Cooling with both reactor recirculation pumps tripped. A Group 2 isolation signal trips the RHR system and RHR CANNOT be restaned. Which one of the following describes where RPV water level is required to be maintained for the current conditions and why?
- a. At least +48 inches on the narrow range RPV water level instmments to promote natural circulation.
- b. At 0.0 inches on the wide range RPV water level instruments to support alternate heat removal using RWCU.
- c. Flooded (solid) on the shutdown range RPV water level instnnnents to support alternate heat removal using the SRVs.
- d. Between +27.5 inches and +42.5 inches on the narrow range RPV water level instruments to miniudze thermal stratification in the reactor pressure vessel.
Answer: a.
- b. Water level is not high enough to support this method of heat removal
- c. Not an appros ed method ofheat removal
- d. Circulation is needed to minimize thermal stratification 4
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. . . - . . . . -. . ._ . . - - _ . _ = - . . RO Written Examination Question No.: 45 K/A: 295010 K2.01 Importance: 12 Tier: 1 Group: 1 Cognitive Level: 2 Exam Bank No.: new
Reference:
EOP-3 A, INT 008-06-13, Graph Objective: INT 008-06-18,1. 2. 10 Which one of the following is assured by emergency depressurizing the reactor if torus pressure CANNOT be maintained in the SAFE region of the Pressure Suppression Pressure Limit (Graph 10, Pressure Suppression Pressure)? To prevent exceeding the ...
- a. Drywell Spray Initiation Limit (DWSIL).
- b. Heat Capacity Temperature Limit (HCTL).
- c. Primary Containment Pressure Limit (PCPL).
- d. Safety Relief Valve Tailpipe Level Limit (SRVTPLL).
Answer: c.
- a. PCPL is not related to DWSIL.
- b. PCPL is not related to HCTL.
- d. PCPL is not related to SRVTPLL, the SRVTPLL is 16', not 16' 6" (right side of PSPL).
Attachments: All the EOP Graphs, FULL SIZE (EOP 5.8 Attachment 2) and ruler. 5
i RO Written Examination Question No.: 21 K/A: 295014 A2.03 Importance: 4.0 Tier: 1 Group: 1 Cognitive Level: 2 Exam Bank No.: new l
Reference:
2.4.1.7, Section 6.2 Objective: COR002-22-02,6b,6d,6h j While operating steady state the following indications are observed: Reactor power lowers Narrow Range reactor water level rises Indicated core plate d/p lowers Indicated core flow rises "A" and "B" recirculation loop flows rise Which one of the following failures caused the above conditions?
- d. One (1) of the Jet pumps has failed.
i
- b. A shroud support access hole cover has failed.
- c. One (1) recirculation pump's speed has raised to maximum.
- d. Flow through a control cell (four fuel bundles) has been blocked.
Answer: b.
- a. Loop flows will only rise in one loop and reactor water level change would not be discernible.
- c. Would not provide these indications
- d. This would lower core flow 6
1 RO Written Examination Question No.: 65 i K/A: 295015 K2.04 Importance: 4.0 Tier: 1 Group: 1 Cognitive Level: 2 Exam Bank No.: new
Reference:
2.3.2.28, COR002-21-02, Figure 3. Objective: COR002-21-02, 5.a While operating at full power a power excursion to 125% occurs cnd the following annunciators are reo-ived-
- 9-5-2/A-3, REACTOR SCRAM CHANNEL B - 9-5-2/B-1, NEUTRON MONITORING TRIP j l
NO control rods moved. At the 9-5 vertical panel, you observe the followmg: i 1 White CRD Scram Solenoid Group lights for RPS Trip System "A" are lit. White CRD Scram Solenoid Group lights for RPS Trip System "B" are off. j NO operator actions have been taken in response to the conditions stated above. ! If the SA-KISA and the 5A-K15C relays will NOT change state, which one of the following operator actions will cause ALL control rods to fully insert? l l
- a. Depressing the "A" manual scram pushbutton. j
- b. Placing the Reactor Mode Switch to SHUTDOWN.
- c. Resetting RPS and then inserting a manual reactor scram.
- d. Placing "A" and "C" RPS trip channel test switches to TRIP.
1 Answer: d. i
- a. K15A and K15C must both actuate to insert all control rods
- b. KISA and K15C must both actuate to insert all control rods
- c. K15A and K15C must both actuate to insert all control rods Attachments: RPS Trip System A figure (COR002-21-02, Figure 3) l I
l
l . I RO Written Examination Question No.: 33 K/A: 295024 Generic 2.4.1 Importance: 4.3 Tier: 1 Group: 1 Cognitive Level: 1 Exam Bank No.: new
Reference:
EOP-3 A, INT 008-06-02 Objective: INT 008-06-13, I Which one of the following describes the EOP(s) required to be entered if drywell pressure rises to 2.0 psig?
- a. EOP-1 A, "RPV Control" only.
- b. EOP-3 A, " Primary Containment Control" only.
- c. EOP-1 A, "RPV Control," and EOP-3 A, " Primary Containment Control" only.
- d. EOP-1 A, "RPV Control," EOP-3A, " Primary Containment Control, and EOP-5A,
" Secondary Containment Control" Answer: c.
Entry into EOP-1 A and EOP-3 A is required. EOP-5A has no entry condition met. Note: Any EOPs supplied for exam need to have entry conditions blanked out. l 8
l 1 l l l RO Written Examination l Question No.: 22 K/A: 295025 K3.02 Importance: 3.9 Tier: 1 Group: 1 Cognitive Level: 1 Exam Bank No.: new i
Reference:
2.3.2.28,9-5 2/C-8 Objective: COR002-33-02,8c ! While the plant is operating at full power a Turbine trip causes RPV pressure to peak at 1095 psig. Which one of the following describes the effect on the Recirculation Pumps, including why?
- a. Both field breakers trip to insert negative reactivity. ,
- b. Both drive motor breakers trip to prevent over-pressurizing the pump discharge piping.
- c. Both drive motor breakers trip because of a turbine trip " lockout" of the normal transformer.
- d. Both pumps run back to 45% speed as feedwater flow lowers to ensure they have adequate NPSH.
Answer: a.
- b. Basis is Not to protect piping
- c. One drive motor breaker trips if powered from the NSST
- d. Drive motor breakers trip 9
RO Written Examination Question No.: 46 K/A: 295031 Generic 2.4.48 Importance: 3.5 Tier: 1 Group: 1 Cognitive Level: 2 Exam Bank No.: new
Reference:
EOP-1 A, RPV Control Objective: INT 008-06-18,2 During conduct of the EOPs, the following parameters exist: Reactor pressure 20 psig Drywell pressure 8 psig Drywell temperature 300*F Torus temperature 105 F Rx Building temperature 150 F If actual reactor water level is at the top of active fuel (TAF) and NO instrument mu boiling is observed, which one of the following describes the RPV level instrumentation that can L,' used to confirm reactor water level?
- a. RPV level CANNOT be determined.
- b. Fuel Zone level instruments can be used.
- c. Wide Range level instruments can be used.
- d. Narrow Re.nge level instruments can be used.
Answer: b. NEW Caution 1. Although in the unsafe region of Graph 1, instrument can be used as long as no boiling is observed.
- a. Although in the unsafe region of Graph 1, instmment can be used as long as no boiling is observed.
- b. below minimum indicated level
- c. below minimum indicated level Attachments: All the EOP Graphs and iust Caution 1 10
RO Written Examination Question No.: 47 K/A: 295031 K1.01 Importance: 4.6 Tier: 1 Group: 1 Cognitive Level: 3 Exam Bank No.: new
Reference:
INT 008-06-02, EOPs-1 A, 6A Objective: INT 008-06-02, 8 Which one of the following conditions assures adequate core cooling? Note: All RPV levels are as INDICATED on the Fuel Zone instruments.
- a. All control rods are fully inserted, Reactor Pressure 128 psig, RPV level -40 inches, NO SRVs open, the only available injection is ECCS pressure maintenance.
- b. All control rods are fully inserted, Reactor Pressure 200 psig, RPV level -50 inches, NO SRVs open, the only available injection is one (1) Core Spray pump.
- c. ATWS with reactor power at 5%, Reactor Pressure 60 psig, RPV level-20 inches, Three (3) SRVs open, the only available injection is one (1) RHR pump.
- d. ATWS with reactor power at 14%, Reactor Pressure 385 psig, RPV level-50 inches, One (1) SRV open, the only available injection is (1) Condensate pump.
Answer: c. Level is above -30 inches for adequate steam cooling and 3 SRVs are open with 1 Minimum Alternate RPV Flooding Pressure met !
- a. -40 inches is too low for adequate steam cooling RC/L-16
- b. -50 inches corrected is below minimum steam cooling level
- d. -50 inches corrected is below minimum steam cooling level but above old minimum steam cooling level Attachments: All the EOP Graphs, EOP 1A, 6A & 7A with entry conditions," Exit" override, and all Cautions except Caution 1 blanked out.
4 4 11
_. - . - =- - - -- - .- . . . . RO Written Examination Question No.: 42 K/A: 295037 K1.07 Importance: 3.4 Tier: 1 Group: 1 Cognitive Level: 2 _ Exam Bank No.: new
Reference:
2.4.1.1.1, Tech Specs. Objective: COR002-04-02,2 Immediately following a reactor scram, it is determined that seven (7) control rods located randomly throughout the core are stuck between positions 06 and 34. Conditions are as follows: Reactor pressure 920 psig Reactor water level +25 inches (stable on the narrow range) Drywell pressure 1.0 psig Drywell temperature 130 F Torus temperature 85 F The seven control rods will NOT respond to RMCS 1 In accordance with EOP-6A and EOP-7A, " Failure to Scram," which one of the following describes the condition allowing exit from EOP-6A and EOP-7A7 1
- a. ALL operable APRMs indicate downscale.
- b. Hot shutdown boron weight injected into the reactor core.
- c. Cold shutdown boron weight injected into the reactor core.
- d. ALL control rods except 26-27 are fully inserted into the reactor core.
Answer: d. The only condition allowing exit of EOP 6A & 7A is the reactor will remain shutdown under all conditions without boron 12
l i RO Written Examination Question No.: 37 K/A: 500000 A2.03 Importance: 3.3 Tier: 1 Group: 1 l Cognitive Level: 2 Exam Bank No.: new
Reference:
EOP-3A Objective: INT 008-06-13,4 COR002-03-02,14e A LOCA has occurred and the following conditions exist: l Drywell H2 concentration is 7% l Torus H2 concentration is 4% Drywell 02 concentration is 4% Torus 02 concentration is 6% In accordance with the EOPs, which one of the following describes the Primary Containment i H2/02 combustible limit status and required actions? I The Primary Containment H2/02 concentration is ... ; i
- a. below the combustible limit. A Reactor scram and emergency depressurization is required.
- b. below the combustible limit. A Reactor scram and emergency depressurization is NOT required.
- c. above the combustible limit. A Reactor scram and emergency depressurization is required.
l
- d. above the combustible limit. A Reactor scram and emergency depressurization is I NOT required.
Answer: c. l l The limits,6%, H2 and 5%,02 in either torus or drywell are the limits for the primary l containment. Combustible limit exceeded requires a reactor scram and emergency l depressurization. I Attachments: EOP-1A & EOP-3A with Entry Conditions blanked out and all Cautions except Caution 1 blanked out. 13 l
. . - - .. -- .. ~ .
RO Written Examination Question No.: 20 K/A: 295001 K2.01 Importance: 3.6 Tier: 1. Group: 2 Cognitive Level: 2 Exam Bank No.: new
Reference:
2.4.2.2.1, 2.1.22 Objective: COR002-22-02, 4.h, 5.j, 6.c, 6.j, ! 7.k,10.h l i A reactor startup in MODE 1 is in progress with reactor power at 15%. A spurious group 6 isolation occurs and all equipment operates as designed. When the isolation is reset, it is observed i that the Reactor Recirculation Motor Generator (MG) set ventilation system CANNOT be re-started. Reactor Recirculation MG set internal air temperatures are as follows: Reactor Recirculation MG set "A" motor air temperature is 276*F. ' Reactor Recirculation MG set "B" generator internal air temperature is 271 *F. 1 Which one of the following IMMEDIATE operator actions is required for the above plant conditions?
- a. Manually scram the reactor,
- b. Stop power changes in progress.
- c. Press the Recirculation MG set "A" Scoop Tube Lockout push button.
l d. Reduce RRMG speed as necessary to lower RRMG set temperatures below 210*F. 1 Answer: a.
- b. Immediate operator action for 2.4.1.7, "Unexpirdned Decrease In Reactor Power."
- c. Action from 2.4.2.2.2, " Reactor Recirculation Flow Control System Failure."
l d. Action from 2.3.2.26 alarm card 9-4-3/C-4. Not applicable due to RRMG sets being tripped. l 14
RO Written Examination Question No.: 92 K/A: 295002 K2.08 Importance: 3.1 Tier: 1 Group: 2 Cognitive Level: 2 Exam Bank No.: new
Reference:
2.4.9.3.5, Objective: COR001-02-02, 3.b, 4.c, 7.b While operating at 96% power a backwash sequence is initiated on the "l Al" condenser. As the backwash sequence starts the following annunciators are received:
- A-4/E-1, CONDENSER A/B BACKWASH TROUBLE B-1/B-3, TG LOW VACUUM PRE-TRIP Main condenser vacuum is slowly degrading. Which one of the following is the cause of degrading main condenser vacuum?
The backwash sequence initiated and ...
- a. the "l Al" condenser water box inlet valve did NOT open.
b the "l A2" condenser water box inlet valve did NOT open.
- c. the "l Al" condenser water box outlet valve did NOT close.
- d. the "l A2" condenser water box outlet valve did NOT open.
A _swer: c. With the outlet valve open circ water vill bypass the l Al condenser.
- a. This normally occurs during a backwast
- b. Tids valve does not reposition from open.
- d. This normally occurs during a backwash.
1 15
RO Written Examination Question No.: 74 K/A: 295003 ' A1.01 Importance: 3.7 Tier: I Group: 2 Cognitive Level: 2 Exam Bank No.: new
Reference:
2.3.13, COR001-01-02 Objective: COR001-01-02, 6.b, 7.a.13.c The reactor is operating at 100% power when the Auto-Transformer becomes de-energized. Which one of the following will occur? Power will be lost to ...
- a. one (1) of the Reactor Recirculation pumps, requiring single loop operation.
- b. the intake structure equipment, requiring a shutdown in accordance with GOP 2.1.5.
- c. the 12.5 KV system, requiring the system to be restored from the Cornfield substation.
- d. one (1) Condensate and one (1) Condensate Booster pump, resulting in a low RPV water level reactor scram.
Answer: c.
- a. The startup transformer will be supplied by the 161KV Aubum line
- b. The intake structure is not effected.
- d. The normal transformer is NOT effected 16 i
i l
.-- - - - - . .- ..- ~- .- -. - _ _ _ _ _ _ -
l l I 1 l RO Written Examination Question No.: 78 , I K/A: 295004 K1.02 Importance: 3.2 1 Tier: 1 Group: 2 _ Cognitive Level: 1 Exam Bank No.: new
Reference:
2.2.25A, pp. 9,10,12 Objective: 2.2.24A, p. 2 COR002-07-02, obj: 6j, 6h, 8c, 9a,1Sb COR002-07-02, Figures 1 & 2 The plant is operating at 75% power when a fault causes a complete loss of 125 VDC bus "B." Which one of the following describes equipment that has been de-energized and has the ability to be manually transferred to its' alternate electrical power source? _ l
- a. Main Turbine Emergency Oil pump and the Air Side Seal Oil Backup Pump.
- b. The ASD Panel and HPCI-MO-16, " Steam Supply Outboard Isolation valve."
l
- c. MS-MO-77, " Outboard Isolation valve" and RWCU-MO-18, " Outboard Isolation valve." l
- d. RCIC-MO-41, " Torus Pump Suction valve" and RCIC-MO-131, " Steam Supply to Turbine valve."
Answer: c.
- a. These are transferrable, but are powered by 250 VDC and are not de-energized. Cannot be transferred during power operations due to SBO calculations.
- b. The ASD Panel and HPCI-MO-16 do not have the ability to be transferred to alternate.
- d. RCIC is normally powered from Division 1 125 VDC.
i i 17
RO Written Examination Question No.: 88 K/A: 295008 A2.02 Importance: 3.4 Tier: 1 Group: 2 Cognitive Level: 3- Exam Bank No.: new
Reference:
2.4.5.1, Section 2.2.2 Objective: COR002-32-02, 8d,9c 1 Given the following conditions: I Reactor poweris steady at 50% i Reactor Vessel Level Control System is in 3-element control l Reactor level detector channel "B" is selected The Channel "B" feedwater flow signal fails to ZERO. Which one of the following describes the result (s) and why? Actual Reactor level will ...
- a. lower, then return to the original level due to the level error signal overriding the steam flow / feed flow error signal.
- b. lower, and stabilize at a lower level due to a mismatch between the level error signal and the total feed flow signal.
- c. rise, and stabilize at a higher level due to a mismatch between the total steam flow signal and the total feed flow signal.
- d. rise until the main turbine and feedpumps trip due to a mismatch between the total steam flow signal and the total feed flow signal.
Answer: c.
- a. Level rises
- b. Level rises
- d. Level rises but should not reach the high level setpoint at this steam flow.
/
Attachments: Provide Figure 2 from COR002-32-02. 18
RO Written Examination Question No.: 6 K/A: 295012 A1.02 Importance: 3.8 Tier: 1 Group: 2 Cognitive Level: 2 Exam Bank No.: new
Reference:
EOP-3A, 5.8.9 Objective: COR002-03-02,13.d,13.e,15.a.1, 17.a A small break LOCA has occurred with the following conditions: Reactor waterlevel +45" indicated on the narrow range instruments Reactor pressure 560 psig Drywell pressure 3.1 psig Drywell temperature 195"F Primary containment water level 14 feet Which one of the following actions is required to operate ALL available drywell cooling?
- a. The drywell cooling FCUs CANNOT be re-started.
- b. Start all the FCUs by placing their control switches in OVERRIDE.
- c. Start all the FCUs by placing their control switches in OVERRIDE and then RUN.
- d. Installjumpers to bypass the high drywell pressure signal and place all the FCUs control switches in RUN.
Answer: b.
- a. FCUs can be started
- c. If the control switches are placed in RUN the FCUs will trip
- d. Nojumpers are required 19
RO Written Examination Question No.: 39 K/A: 295013 Generic 2.4.4 Importance: 4.0 Tier: 1 Group: 2 Ccgnitive Level: 2 Exam Bank No.: new
Reference:
EOP-3A, EOP-5A Objective: INT 008-06-13,1; INT 008-06-17, I The plant is operating at 100% power with HPCI testing in progress. A loss of cooling for the HPCI Room has occurred. The following conditions exist: Annunciator 9-3-1/E-10, AREA HIGH TEMP alanns Ronan Annunciator (1522), HPCI ROOM (E 878') AREA TEMP HIGH, is displayed Temperature Module HPCI-TS-105D indicates 180*F on the Plant Area Temperature Monitor Panel Average Suppression Pool temperature is 98 F Which one of the following describes the EOP(s) required to be entered?
- a. EOP-5A, " Secondary Containment Control" only.
- b. EOP-1 A, "RPV Control," and EOP-3 A, " Primary Containment Control" only.
- c. EOP-3 A, " Primary Containment Control," and EOP-5 A, " Secondary Containment ]
Centrol" only. '
- d. EOP-I A, "RPV Control," EOP-3 A, " Primary Containment Contrci," and EOP-5A,
" Secondary Containment Control".
Answer: c. EOP-1 A has no entry condition met. I 20
RO Written Examination Question No.: 25 K/A: 295016 K3.03 Importance: 3.5 Tier: 1 Group: 2 Cognitive Level: 1 Exam Bank No.: new 1
Reference:
5.4.3.2 Objective: COR002-34-02,4 b i Emergency Procedure 5.4.3.2, " Post-Fire Shutdown to Mode 4 Outside Control Room," requires the ASD panel isolation switches be placed in the ISOLATE Position. l l Which one of the following describes the reason for this action? l I
- a. To ensure automatic operation ofECCS remains available.
1
- b. To isolate wire runs to meet divisional physical separation criteria. l
- c. To prevent overloading the associated DG during a design basis LOCA.
I
- d. To disconnect control room control circuits to prevent spurious operation of the I associated equipment. !
Answer: d.
- a. HPCI automatic features (except overspeed) are disabled when operated from the ASD panel.
- b. The ISOLATE position does not change the physical routing or location of equipment.
- c. These valves draw the same power from the ASD panel as from the control room.
l I i i 21
RO Written Exam: nation Question No.: 84 K/A: 295017 A2.04 Importance: 3.6 Tier: 1 Group: 2 Cognitive Level: 2 Exam Bank No.: new
Reference:
2.3.2.24, 2.4.7.1, 2.4.1.2 Section . Objective: COR001-16-02, 5, 7.b,10.h 6.1 While operating at full power the following alarms are received: i 9-4-1/C-4, OFFGAS TIMER INITIATED 9-4-1/C-5, OFFGAS HIGH RAD Which one of the following caused these alarms?
- a. HPCI steam leak in the HPCI room.
- b. Fuel element leak in the spent fuel pool.
- c. High moisture in the AOG charcoal beds.
- d. Blockage of flow to at least one (1) fuel assembly. I Answer: d.
1
- a. No effect on off-gas. Direct effect on reactor building and ERP radiation level.
- b. This would effect reactor building vent rad monitor
- c. This will ERP radiation monitor to alarm, but not offgas.
22
-. - . . . . . - _- . - . . _ _ . .. . _ --- ~
l J RO Written Examination Question No.: 68 l K/A: 295018 K1.01 Importance: 3.5 Tier: 1 Group: 2 i Cognitive Level; 2 Exam Bank No.: new
Reference:
5.2.4 Objective
COR002-19-02,2.b,4.a,4.c 5.a, 5.c, 6.b, 6.d, The unit is operating at 85% power with REC pumps "A", "B" and "C" operating. REC pump control switches are positioned as follows:
"A" REC pump STANDBY "B" REC pump NORMAL "C" REC pump STANDBY "D" REC pump NORMAL An operator mistakenly de-energizes MCC-K and ten (10) seconds later re-energizes MCC-K.
Twenty (20) seconds after MCC-K is re-energized, which one of the following will restore REC cooling with three (3) REC pumps in operation? Manually start ...
- a. two (2) REC pumps only ("A," "B" and/or "D").
- b. "A" or "B" REC pumps and verify REC pump "D" automatically starts.
- c. two (2) REC pumps only ("A," "B" and/or "D") and then open the non-critical header supply, drywell supply isolation, HX outlet, and augmented radwaste supply.
- d. "A" or "B" REC pumps, verify REC pump "D" automatically starts, and then open the non-critical header supply, drywell supply isolation, HX outlet, and augmented radwaste supply. 1 Answer; a.
No pumps auto start. If two of "A,""B" and or "D" pumps are started within 40 seconds of the pump trips, no header isolation valves close. ' 23
)
i l l RO Written Examination Question No.: 81 l K/A: 295019 Generic 2.4.I1 Importance: 3.4 Tier: 1 Group: 2 l Cognitive Level: l' Exam Bank No.: new '
Reference:
5.2.8 Objective
SKLO10-01-02, A.4, B.3 l l 1 While the plant is operating at power, a failure causes Service Air pressure to lower. ' If Service Air pressure continues to lower below 90 psig, which one of the following requires a manual reactor scram per Emergency Procedure 5.2.8, " Loss ofInstrument Air?" i
- a. Only one (1) control rod starts to insert or instrument air pressure lowers to 84 psig.
- b. More than one (1) control rod starts to insert or instrument air pressure lowers to 76 psig.
- c. Less than two (2) compressors running and Service Air Supply Header automatically isolated.
- d. The in-service instrument air dryer becomes clogged and Drywell Pneumatic Header Low Pressure alarm is received.
Answer; b. 1 Procedure requires a scram when more than one rod drifts in < 77 psig instmment air pressure. l i I I i v 24
RO Written Examination Question No.: 28 K/A: 295020 A1.01 Importance: 3.6 Tier: 1 Group: 2 Cognitive Level: 1 Exam Bank No.: new
Reference:
2.1.22 Objective: COR002-11-02,8.b A false high drywell pressure signal has caused an automatic initiation of HPCI. An operator then depresses the Manual Isolate pushbutton instead of the Turbine Trio oushbutton on the 9-3 panel when attempting to secure HPCI. Which of the following will occur?
- a. itPCI Inboard Steam Isolation valve, HPCI-MO-15 closes, the ECST Suction valve HPCI-MO-17 receives a close signal and the HPCI turbine trips.
- b. HPCI Outboard Steam Isolation valve, HPCI-MO-16 closes, the Suppression Pool l suction valve HPCI-MO-58, receives a close signal and the HPCI turbine trips.
- c. Both HPCI Inboard and Outboard Steam Isolation valves, HPCI-MO-15 and HPCI-MO-16, close and both HPCI Suction valves, HPCI-MO-17 and HPCI-MO-58 i receive a close signal and the HPCI turbine trips.
J
- d. Both HPCI Inboard and Outboard Steam Isolation valves, HPCI-MO-15 and HPCI- -
MO-16, close and both HPCI Suction valves, HPCI-MO-17 and HPCI-MO-58 receives a close signal, HPCI turbine coasts down but does NOT trip. Answer: b.
- a. This is logic A which is not tripped by the manual pushbutton.
- c. The manual pushbutton only trips logic B
- d. The manual pushbutton only trips logic B and the turbine trips on an isolation signal.
l 25
l 1 RO Written Examination Question No.: 49 K/A: 295022 K1.01 Importance: 3.3 1 Tier: 1 Group: 2 l Cognitive Level: 2 Exam Bank No.: new l
Reference:
2.4.1.1.4, Section 6.1 Objective: COR002-04-02, 8.a,10.b,13.g i During a plant startup RPV pressure is 850 psig when a loss of suction causes the "A" CRD pump to trip. The "B" CRD pump is out-of-service for maintenance. 1 After one (1) minute, which one of the following statements is correct? I l Rod motion control with RMCS is ... 1
)
- a. unaffected.
( Scram times will exceed technical specification limits.
- b. unaffected.
Scram times will be within technical specification limits. ) l
- c. unavailable. 1 Scram times will exceed technical specification limits. l
- d. unavailable. I Scram times will be within technical specification limits.
Answer: d.
)
- a. Normal rod motion is lost. Scram times are OK as long as accumulators are charged.
- b. Normal rod motion is lost !
- c. Scram times are OK as long as accumulators are charged.
l 1 26
l l RO Written Examination Question No.: 34 K/A: 295026 K3.01 Importance: 3.8 Tier: 1 Group: 2 Cogstive Level: 1 Exam Bank No.: new l
Reference:
EOP-3 A, INT 008-06-13 Objective: INT 008-06-13,4 INT 008-06-18 l Which one of the following describes why an emergency depressurization is required per EOP-3A, " Primary Containment Control," when suppression pool temperature cannot be maintained below the Heat Capacity Temperature Limit (HCTL)? Ensures initiation of RPV depressurization will NOT result in exceeding ...
- a. Net Positive Suction Head (NPSH) limits for low pressure ECCS when they are required for adequate core cooling.
- b. Boron Injection Initiation Temperature (BIIT) when Hot Shutdown Boron Weight has NOT been injected into the reactor.
- c. Suppression Chamber Spray Initiation Pressure (SCSIP) while sufficient energy remains in the RPV to exceed containment limits.
- d. Primary Containment Pressure Limit (PCPL) while the rate of energy transfer to the suppression pool exceeds containment vent capacity.
Answer: d.
- a. Not related to NPSH for pumps.
- b. HSBW is not the reason for emergency depressurization
- c. SCSIP is used to preclude cyclic condensation of steam (chugging) at the downcomer
. openings of the drywell vents.
l i 27
RO Written Examination Question No.: 41 K/A: 295028 Generic i;. ' 20 , I portance: 3.3 Tier: 1 Group: 2 Cognitive Level: 1 Exam Bank No.: new
Reference:
EOP-7A Objective: INT 008-06-10,3 While performing EOP-7A, "RPV Level / Failure to Scram," with power below 3%, which one of the following CAUTIONS applies as reactor water level is controlled?
- a. Lowering RPV water level to -42 inches will result in an ADS initiation if ADS is NOTinhibited.
- b. Prior to lowering RPV water level to -110, ensure low pressure ECCS injection is stopped and prevented.
- c. Lowering RPV water level to -110 inches will result in an MSIV isolation and loss of the main condenser as a heat sink.
- d. Prior to lowering RPV water level to -42 inches, block injection from all low pressure ECCS systems NOT required for RPV level control.
Answer: c. a,b,d These cautions do not exist in EOP-7A Attachments: EOP-7A with all CAUTIONS except Caution I blanked out (Verify CAUTION 6 blanked out as it is in the flowchart steps). r 28
RO Written Examination Question No.: 35 K/A: 295029 K3.03 Importance: 3.4 Tier: 1 Group: 2 Cognitive Level: 2 Exam Bank No.: new
Reference:
EOP-3 A, INT 008-06-13, II.I.4 Objective: INT 008-06-13, 4 Which one of the following describes why a reactor scram is required per EOP 3A, " Primary Containment Control," when suppression pool water level CANNOT be maintained below sixteen (16) feet?
- a. To prevent covering the Torus to Drywell vacuum breakers.
- b. Ensures sufficient suppression chamber free air space if a LOCA occurs.
- c. To ensure the reactor is shutdown by control rod insertion prior to performing an emergency depressurization.
- d. A subsequent LOCA blowdown above this suppression pool level will exceed the suppression pool downcomer design differential pressures.
Answer: c. EOP set-point basis.
- a. Basis for 16.5'
- b. Basis for the HCLL. -
- d. Not the reason for the scram. EOP terms that sound technical.
29
RO Written Examination Question No.: 44 K/A: 295030 K2.04 Importance: 3.7 Tier: 1 Group: 2 Cognitive Level: 3 Exam Bank No.: new
Reference:
EOP-1A, EOP/ SAG Graphs Objective: INT 008-06-18, l., 2. A LOCA has occurred with the following conditions: All control rods are fully insened RHR pumps "A," "B" and "D" are unavailable HPCI, and RCIC are unavailable RHR "C" operating in LPCI mode is being used to maintain RPV water level Actual RPV water level is twenty (20) inches above the top of active fuel (TAF) and rising ten (10) inches per minute RHR flow rate 8500 gpm (maximum available) Toms pressure 4 psig Suppression Pool temperature 185 F Suppression Poollevel 7 feet Which one of the following describes the use of RHR as an injection system?
- a. Reduce RHR flow to 4000 gpm.
- b. Reduce RHR flow to 5500 gpm.
- c. Reduce RHR flow to 7000 gpm.
- d. Maintain RHR flow at 8500 gpm.
Answer: c. With 4 psig pressure and 3 feet of water above the suctions there is 5.29 psig overpressure i requiring that flow be reduced to no more than 7000 gpm for NPSH concerns. Flow is also in the UNSAFE region of the Vortex limit curve, but this curve is less limiting than the NPSH curve in this case. 10 inches per minute RPV rise means that an excess flow of 1500 to 2000 gpm is available and flow could be reduced to 6500 and still maintain RPV water level above TAF. Attachments: EOP and FUL'L SIZE SAG graphs must be attached to the exam. Calculators and ruler. i 30 l
RO Written Examination Question No.: 40 K/A: 295033 A2.01 Importance: 3.8 Tier: 1 Group: 2 Cognitive Level: 2 Exam Bank No.: new
Reference:
EOP-5A Objective: INT 008-06-17,3 Which one of the following describes the conditions in secondary containment that require a reactor shutdown if a primary system is NOT discharging into the Secondary Containment?
- a. RMA-RA-7, " Neutron Monitor Sys Drive Mech Area," and RMA-RA-10, "HPCI Pump Room," alarm and both indicate upscale high.
- b. RMA-RA-27, " Torus HPV Area (Southwest)," and RMA-RA-4, "RWCU Precoat Area," alarm and both indicate upscale high.
- c. RMA-RA-3, "New Fuel Area," and RMA-RA-5, "RWCU Sludge and Decant Pump Area," alarm and both indicate upscale high.
- d. RMA-RA-8, "CRD Hydraulic Equip Area (South)," and RMA-RA-9, "CRD Hydraulic Equip Area (North)," alarm and both indicate upscale high.
Answer: b. Two areas must exceed max safe, b. is the only one that meets this. RMA-RA-3 and RMA-RA-7 are not on the table. Attachments: EOP 5A with entry conditions blanked out. EOP Tables 9,10 & 11 31
RO Written Examination Question No.: 4 K/A: 295034 A1.03 Importance: 4.0 Tier: 1 Group: 2 Cognitive Level: 1 Exam Bank No.: new
Reference:
2.4.8.4.1, Objective: COR001-08-02,11.b,19.c I 2.1.22, Section 8.6 In accordance with General Operating Procedure 2.1.22, " Recovering From A Group Isolation," which one of the following methods is used to ensure that the Secondary Containment has ISOLATED?
- a. Verify NO flow is indicated on the reactor building exhaust flow recorder.
- b. Verify ALL supply fan discharge valves and ALL exhaust fan discharge valves are closed using the valve position indicators.
- c. Verify ALL reactor building supply and ALL reactor building e.haust fans trip and SGT starts by observing fan status light indicators.
- d. Verify ALL supply fan discharge dampers are closed locally and verify ALL exhast fans discharge dampers are closed by observing control room indications.
Answer; b.
- a. Will not verify an isolation l
- b. These are the fan isolations not the building isolations
- c. Checking fan status does not verify damper isolation has occurred.
l 32
M RO Written Examination Question No.: 16 K/A: 295021 Generic 2.4.9 Importance: 3.3 , Tier: 1 Group: 3 Cognitive Level: 2 Exam Bank No.: new
Reference:
AP 2.4.2.4.1, Attachment 5 Objective: COR002-23-02,7.a The reactor is in Cold Shutdown with RPV metal temperatures at 150 F The reactor has been shut down for 33 hours with the recirculation pumps secured. RPV water level is just below full scale on the Narrow Range indicators. The RPV head vents are OPEN. A loss of shutdown cooling occurs. Assume NO operator actions are taken. Which one of the following is the approximate time (in hours) the reactor will remain in Cold Shutdown if shutdown cooling CANNOT be restored?
- a. I hour
- b. 1.4 hours
- c. 2.4 hours
- d. 3.3 hours Answer: a.
- b. hours after si .tdown with reactor water level at RPV flange curve
- c. days after shutdown with reactor water level at high level trip curve
- d. days after shutdown with reactor water level at RPV flange curve Attachments: 2.4.2.4.1, Attachment 5 (all of the attachments) 33
RO Written Examination Question No.: 59 K/A: 295023 A2.04 Importance: 3.4 Tier: 1 Group: 3 Cognitive Level: 1 Exam Bank No.: new
Reference:
EP 10.25, section 4.1.2.5 Objective: SKLO10-01-02, A4 Refueling activities are in progress with a new fuel bundle being lowered into reactor core location 21-40. Which one of the following requires the Control Room Monitor to direct fuel loading be immediately terminated per 10.25, " Refueling - Core Unload, Reload, and Shuflie?" ;
- a. A full scram is received due to Scram Discharge Volume high level.
- b. SRM "A" and SRM "B" c unt rates rise by a factor of ten (10) to 300 eps.
- c. Shutdown Cooling is lost with less than 24 hours estimated for " time to boil."
l
- d. Fuel Pool Cooling is lost with less than 24 hours estimated for " time to boil."
I Answer; b. Note below step 4.1.2.4 states "SRM count rates normally do not exceed 100 cps." l a, c, d - None of these conditions require fuel loading be terminated per 10.25.
Attachment:
Provide Fig 4 of COR002-30-02 I 34 i
f RO Written Examination Question No.: 38 1 K/A: 295032 K3.03 Importance: 3.8 Tier: 1 Group: 3 Cognitive Level: 1 Exam Bank No.: new
Reference:
EOP-5A, INT 008-06-17 Objective: INT 008-06-17,4 Which one of the following describes the EOP-5A, " Secondary Containment Control," basis for isolating a system discharging into the secondary containment?
- a. To minimize RPV inventorylosses.
- b. To backup PCIS automatic functions.
- c. To maintain ALL areas of the Reactor Building accessible to personnel.
- d. To terminate rising temperatures, radiation levels, and water levels in s' . ondary containment.
Answer: d.
- a. This is covered by other EOPs
- b. PCIS automatic actions may not have been required
- c. Secondary Containment Control does not maintain habitability for all areas. The Max Safe values are based on equipment operability and personnel access necessary for EOP actions.
1 l l \ : 1 1 35 P
RO Written Examination Question No.: 11 K/A: 295036 K2.01 Importance: 3.1 Tier.1 Group: 3 Cognitive Level: 2 Exam Bank No.: new
Reference:
2.2.27, 2.3.2.20, S-1/A-1, B&R- Objective: COR002-03-02,4,12.k 2038 COR001-11-02, 5 A 300 gpm leak on the "A" RHR heat exchanger has resulted in flow into the NW quadrant sump "l A." As the level rises the following annunciator is received: - S-1/A-1, REACTOR BLDG A SUMP HI HI LEVEL Which one of the following describes how this will effect water level in secondary contaimnent?
- a. The sump level will rise and the sump will overflow until the heat exchanger is isolated.
- b. The discharge valve from the sump will isolate, level in the NW quad will rise at a faster rate.
l
- c. The isolation valve on the inlet into the sump will close, directing water from the heat exchanger into the torus area.
- d. The second sump pump will start. Provided flow to the sump remains constant, sump level will be maintained at the current value.
Answer: c. l
- a. Sump level should lower with the inlets isolated
- b. The discharge from the sump is NOT isolated, sump level shouid lower.
- d. The lines into the sump are isolated so level should eventually lower as the water is pumped from the sump and water is directed into the toms area.
1 I 36 . l
RO Written Examination Question No.: 64 < K/A: 201001 K4.04 Importarice: 3.6 Tier: 2 Group: 1 Cognitive Level: 1 Exam Bank No.. new
Reference:
COR002-04-02 Objective: COR002-04-02,4.d, Which one of the following describes the operation of the Backup Scram Valves?
- a. Both valves must be energized to depressurize the scram air header.
- b. Both valves must be de-energized to depressurize the scram air header.
- c. Only one (1) valve must be energized to depressurize the scram air header.
- d. Only one (1) valve must be de-energized to depressurize the scram air header.
Answer: c.
- a. Only one valve is required by design.
- b. Only one valve is required by design.
- d. These are DC valves, energized to actuate.
l 37
RO Written Examination Questioa No.: 54 K/A: 201001 K3.03 Importance: 3.1 Tier: 2 Group: 1 Cognitive Level: 2 Exam Bank No.: new
Reference:
2.4.1.1.1, COR002-04-02, Fig. 5 Objective: COR002-04-02,12.c The plant is operating at 50% power with the following CRD system indications: Drive water differential pressure 265 psid Drive flow 0.0 gpm Charging Header pressure 1450 psig CRD system flow 50 gpm While attempting to insert control rod 18-19, drive water flow is observed to be 0.0 gpm. When attempting to withdraw control rod 18-19, drive water flow is observed to be 2.0 gpm. The control rod does NOT move. Which one of the following describes the cause of the above indications? Directional Control Valve ...
- a. 122 is stuck open.
- b. 123 is stuck open.
- c. 122 is stuck closed.
- d. 123 is stuck closed.
Answer: d.
- a. 122 stuck open would provide continuous withdrawal flow
- b. 123 stuck open would provide continuous insert flow
- c. 122 stuck closed would prohibit withdrawal flow but allow insert flow.
Attachment:
Provide Fig 5 of reference 38
l RO Written Examination Question No.: 50 K/A: 201002 A2.02 Importance: 3.2 Tier: 2 Group: 1 Cognitive Level: 1 Exam Bank No.: new l
Reference:
2.4.1.1.3, Section 4.4 Objective: COR002-20-02, 4.c, 7.b 2.3.2.27, 9-5-1/C-4 Reactor Power is 50%. With NO control rod selected, a partially withdrawn control rod is slowly moving into the core. Which one of the following is required in accordance with procedure 2.4.1.1.3, " Failure of Drive ; to Latch?" Select the specific control rod and ...
- a. fully insert the specific control rod with the EMERGENCY IN switch.
- b. insert a scram on the specific control rod for a minimum of five (5) seconds.
- c. raise CRD cooling water pressure and monitor the specific control rod movement.
- d. attempt to stop the specific rod move:,ent by placing the ROD MOVEMENT CONTROL switch in WITHDRAW.
Answer: a.
- b. This action would be taken if the control was drifting out after fully inserted.
- c. Would cause the control rod to drift in faster ifd/p was the problem and not covered by CNS procedures.
- d. Unacceptable yet, similar in concept to the control rod drifting out actions. l l
1 i 39 l l
RO Written Examination Question No.: 51 K/A: 201002 A1.02 Importance: 3.4 i Tier: 2 Group: 1 Cognitive Level: 2 Exam Bank No.: new i
Reference:
COR002-20-02 Objective: COR002-20-02, 4.a, 7.a IOP 4.3, Sect - 12.3.2
~ With the plant at 65% power, a control rod is single notch withdrawn from notch 24 to notch 26.
While the control rod is being withdrawn, a malfunction of the RMCS timer causes a continuous withdrawal signal to be sent to the selected control rod. i Assume NO additional operator actions. l l Which one of the following describes the final position of the control rod?
- a. Notch 00.
b.' Notch 28. i
- c. Notch 30.
I
- d. Notch 48.
l Answer: b. Timer malfunction deselects the rod after 2 seconds (which is % second longer than the normal timer), causing the control rod to be de-selected
- a. From the given power level a Rx scram will not occur.
- c. An extra % second will not result in a two (2) notch change. l
- d. Rod will be de-selected after 2 seconds of motion.
40
i RO Written Examination Question No.: 19 K/A: 202002 A3.03 Importance: 3.1 Tier: 2 Group: 1 Cognitive Level: 2 Exam Bank No.: new
Reference:
COR002-22-02, 2.2.68, Sections . Objective: COR002-22-02, 6.I, 6j,10c,10j, 4.1.4,8.2.8 10n The "A" Reactor Recirculation Pump is being started. The JOG BYPASS switch is the JOG-IN position. After the MG SET switch has been placed to START, the following events occur in sequence: RRMG"A" drive motor breaker closes. 15 seconds later, RRMG "A" drive motor breaker trips and locks out The field breaker does NOT close during the start attempt. Given the above panel indications why did the Reactor Recirculation pump fail to start?
- a. RRMG set room ventilation not in operation.
- b. Recirculation pump discharge valve failed to open.
- c. Recirculation pump suction valve is NOT fully open.
- d. Scoop tube positioner failed to ramp to the startup position.
Answer: d. RRMG set trips and lockout on incomplete sequence
- a. This is not a recirculation pump trip, it is a breaker close permissive.
- b. Discharge valve shall be partially open within one (1) minute and full open within two (2) minutes from time drive motor breaker is closed to prevent pump from tripping.
- c. This would prevent the drive motor breaker from closing 1
l l l 41 ; i
. . . ~ .. . __ _
l l RO Written Examination Question No.: 70 K/A:203000 K2.01 Importance: 3.5* Tier: 2 Group: 1 Cognitive Level: 2 Exam Bank No.: new
Reference:
2.2.68.1, COR002-23-02 Objective: COR002-23-02, 2.a, 3.f, 8.a, 8.c The following sequence of events has occurred: A LOCA has occurred resulting in LPCI injection 20 seconds later, all offsite power is lost DG1 will NOT start DG2 starts and loads as designed Under these conditions, which one of the following describes the status of the RHR pumps thirty (30) seconds after the loss of power and why?
- a. "A" and "B" pumps are available but NOT operating because the pump stop signal has sealed in.
- b. "A" and "B" pumps are operating because the breaker anti-pump circuitry was sealed in when the associated 4160 volt bus was re-energized.
- c. "C" and "D" pumps are available but NOT operating because the breaker anti-pump circuitry has sealed in.
- d. "C" and "D" pumps are operating because the breaker anti-pump circuitry was reset when the associated 4160 volt bus was de-energized.
Answer: d.
- a. Not available as they're powered by DG 1. Stop signal has NOT been energized (switch not taken to OFF)
- b. Not operating as they're powered by DG 1 I
- c. Anti pump will NOT prevent the pumps from starting 42
I RO Written Examination Question No.: 27 K/A: 206000 A4.12 Importance: 4.0 i l Tier: 2 Group: 1 ! Cognitive Level: 2 Exam Bank No.: new
Reference:
2.2.33, Sect. 4.2.6 . Objective: COR002-11-02, 5.b, 8.a, 8.c,10.f HPCI initiated due to low RPV water level due to a loss of Reactor Feedwater Pumps. HPCI subsequently tripped on RPV high water level. The following conditions are observed: l l RPV water level is +20 inches and slowly lowering 1 Drywell pressure is 1.0 psig and slowly rising
]
Which one of the following describes how HPCI will respond as water level continues to lower 7 I If the operator depresses the ...
- a. INITIATION SIGNAL RESET pushbutton on panel 9-3, IIPCI will start and inject into the RPV regardless of RPV water level and Drywell pressure.
- b. HI REACTOR WATER LEVEL TRIP RESET pushbutton, HPCI will start and inject into the RPV regardless of RPV water level and Drywell pressure.
- c. INITIATION SIGNAL RESET pushbutton, HPCI will start and inject into the RPV i if Drywell pressure rises to 1.84 psig regardless of RPV water level.
- d. HI REACTOR WATER LEVEL TRIP RESET pushbutton, HPCI will restart only ,
after water level lowers to -42 inches or Drywell pressure rises to above 1.84 psig. Answer: b.
- a. Does not reset the high level trip, removes the "open" signal to HPCI-MO-14. l'
- c. Will not reset the high level trip.
- d. These conditions do not require low water level or high drywell pressure to stan HPCI. l l
i I l l l 43 i i
RO Written Examination Question No.: 26 K/A: 206000 K2.01 Importance: 3.2* Tier: 2 Group: 1 Cognitive Level: 2 Exa'n Bank No.: new R eference: 2.2.33, Sect 2.2.5 Objective: COR002-11-02, 5.g, 5.h, 6.a, 6.c, 10.b With NO AUTOMATIC HPCI initiation signal present, which one of the following describes the effect that a loss of ALL AC power has on HPCI system operation and why? HPCI can be started for RPV...
- a. injection only. The Pressure Control Mode is unavailable because HPCI-MO-21, Test Bypass to ECST valve, CANNOT be opened due to loss of power.
- b. injection only. The Pressure Control Mode is unavailable because HPCI-MO-24, ECST Test Line Shutoff valve, CANNOT be opened due to loss of power,
- c. level control and/or Pressure Control. The HPCI-MO-15, Steam Supply Inboard Isolation valve, will NOT re-position due to loss of power.
- d. level control or pressure control. The interlock between the HPCI-MO-19, Injection valve, and the HPCI-MO-24, Outboard ECST Test Line Shutoff valve, will NOT ftmetion due to loss ofpower.
Answer: c. a, b, d The only AC powered valve in the HPCI system is HPCI-MO-15. T 1 44 5
. . - - - . . = . . -_. - - - . .
RO Written Examination Question No.: 12 K/A: 209001 Kl.13 Importance: 2.8 Tier: 2 , Group: 1 Cognitive Level: 1 Exam Bank No.: new
Reference:
2.3.2.23, COR002-06-02 Objective: COR002-06-02, 3.d, 5.d, 6.b, 9.c During operation at full power the following annunciator is received: 9-3-3/A-5, CORE SPRAY B BREAK DETECTION NO other annuaciators alarm. A station operator is sent to the d/p indicating switch and reports that the d/p is oscillating at around 4.0 psid. Which one of the following caused this indication? 1
- a. Water is leaking by the core spray injection check valve.
- b. A break has occurred in the core spray sparger or core spray piping to the sparger inside the core shroud.
- c. The core spray piping is broken between its' Outboard Injection valve and its' Inboard Injection valve. l 1
- d. There is a break in the core spray line within the reactor vessel between the core shroud and RPV penetration.
Answer: d.
- a. This would also have to be leaking by the injection valves, this would cause an annunciator for high pressure valve leak, 9-3-3/C-5.
- b. This would not cause a change in reading or the alarm.
- c. A break in this location is isolated from the break detection instrumentation by the closed Inboard Injection valve.
1 l 45
RO Written Examination Question No.: 9 K/A: 211000 K4.08 Importance: 4.2* Tier: 2 Group: 1 Cognitive Level: 1 Exam Bank No.: new
Reference:
COR002-29-02 Objective: COR002-29-02, 5.g, The keylock switch for Standby Liquid Control (SLC) Pump "A" is turned to the START position. Aside from starting the "A" SLC pump, what else will this switch movement initiate?
- a. Only the "A" squib valve fires, only RWCU-MO-15 isolates.
- b. Both "A" and "B" squib valve fires, only RWCU-MO-18 isolates.
- c. Only the "A" sqtiib valve fires, both RWCU-MO-15 and RWCU-MO-18 isolates.
- d. Both "A" and "B" squib valve fire, both RWCU-MO-15 and RWCU-MO-18 isolate.
Answer: a.
- b. Only the "A" squib fires, RWCU-MO-15 closes, RWCU-MO-18 does not close.
- c. Only the RWCU-MO-15 isolates.
- d. Only the "A" squib fires, Only the RWCU-MO-15 isolates.
l l I 46
, RO Written Examination Question No.: 60 K/A: 212000 K2.01 Importance: 3.2 Tier 2 Group: 1 Cognitive Level: 1 Exam Bark No.: new
Reference:
2.4.6.6, COR002-21-02 Objective: COR002-21-02, 9.a,11.a The reactor is operating at 100% reactor power when the feeder breaker to MCC-L trips. Which one of the following describes the response of the Reactor Protection System (RPS) to this feeder breaker trip? Pcwer is lost to the ...
- a. "A" RPS MG set. After a small time delay, power is lost to the "A" RPS logic, and a
- % scram is received.
- b. "B" RPS MG set. After a small time delay, power is lost to the "B" RPS logic, and a l
% scram is received
- c. "A" RPS MG set. After a small time delay, the "A" RPS bus automatically shifts to l the alternate power supply. NO % scram is received. !
- l i d. "B" RPS MG set. After a small time delay, the "B" RPS bus automatically shifts to !
the alternate power supply. NO % scram is received. i 1 4 Answer: a. )
- b. MCC-L powers the Div 1 "A" RPS bus.
] c. The RPS system does not automatically transfer power supplies.
- d. The RPS system does not automatically transfer power supplies.
i I i l I l k 47 I
l l RO Written Examination Question No.: 63 K/A: 212000 K5.02 Importance: 3.3 Tier: 2 Group: 1 Cognitive Level: 1 Exam Bank No.: new
Reference:
COR002-21-02, Sect. II. D. 4 & Objective: COR002-21-02, 4.m, 5.a,10.c Figure 7 The reactor is operating at 25% reactor power. Both the "A" Inboard MSIV and the "D" Outboard MSIV close. All others MSIVs remain fully open. Which one of the following describes the design erTect on the RPS logic? a A full scram is received.
- b. A % scram on RPS Trip System "A" is received.
- c. A % scram on RPS Trip System "B" is received.
- d. Neither a % scram, nor a full scram will be received.
Answer: d. a,b,c. A combination ofMSIV valve closure in the "A" and "D" or "B" and "C" steam lines will not result in RPS % or full scram. l l 48
RO Written Examination Question No.: 67 K/A: 215003 A2.02 Importance: 3.5 Tier: 2 Group: 1 Cognitive Level: 1 Exam Bank No.: new
Reference:
4.1.2, Sect 4.2 Objective: COR002-12-02, 3.b, 3.d, 5.a, 5.b, 6.b, 6.c, 7.e A normal plant startup is in progress with the reactor mode switch in STARTUP. The following Intermediate Range Monitor system conditions exist: IRM Channel"A"is failed downscale IRM Channel"A" is bypassed IRM Channel "A" Mode Switch is in STANDBY All the IRM range switches, including IRM channel "A", are on Range 2.
- Which one of the following describes the automatic action (s) that occur when IRM "A" is taken out ofbypass?
- a. % scram only.
- b. Control rod block only.
- c. IRM downscale alarm only.
- d. Control rod block and % scram.
Answer: d. It's Mode switch is out of OPERATE
- a. Yes, but it also generates a rod block
- b. Normally a downscale is a rod block, but in this case it's been disabled with the Mode Switch. INOP generates a % scram
- c. A rod block and % scram are received.
i 49
RO Written Examination Question No.: 53 K/A: 215004 Generic 2.1.7 Importance: 3.7 Tier: 2 Group: 1 Cognitive Level:'1 Exam Bank No.: new
Reference:
4.1.1, 2.4.1.6, COR002-30-02 Objective: COR002-30-02, 3.b, 5.a, 5.b , During a reactor startup with the reactor close to criticality, control rod 18-19 is withdrawn from position 08 to 12. During movement of the Control Rod Drive Mechanism (CRDM), ALL of the SRM count rate meters remain at 4 x 10' cps. Which one of the following is the cause of this indication?
- a. The SRM detectors have been withdrawn too far out of the core.
1
- b. Source neutrons contribution is NOT measurable at this power level.
- c. This control rod is located too far from any SRM for this movement to be detected.
- d. This control rod is uncoupled from it's control rod drive and is stuck somewhere in the core.
4 Answer: d. J This would not prevent an indicated flux change from occurring. 4 x 104 cps is well vithin a. the required value for detection of changes. l . b Source neutrons are the major contributor at this power level..
- c. This control rod is right next to the SRM. Any rod movement near criticality weidd be
! detected by at least one SRM detector.
1 1 l l l l l l l l l 50
RO Written Examination Question No.: 23 K/A: 215005 K4.07 Importance: 3.7 Tier: 2 Group: 1 Cognitive Level: 2 Exam Bank No.: new
Reference:
COR002-01-02 Objective: COR002-01-02, obj. 8a, 8e, 9c 4.1.3, Section 4.3 2.3.2.27, 9-5-1/F-4 The reactor is operating at 100% rated thermal power. The following instrument readings are observed: APRM A 102 % - APRM B 99 % APRM C 100 % - APRM D 99% APRM E 100 % - APRM F 99 % Flow Unit A 98 % - Flow Unit B 87 % Which one of the following describes the efvect due to the above conditions if all trips and alarms occur at their design setpoint(s)? A control rod withdrawal block and ...
- a. an APRM Upscale alarm.
- b. a Flow Reference OffNormal alarm.
- c. an APRM INOP alarm and a RPS Trip System B trip (% scram).
- d. a Flow Reference Off Normal alarm and a RPS Trip System A trip (% scram).
Answer; b. If Flow Unit A if greater than flow unit B by more than 10%, comparator A trip unit trips causing a rod block, a FLOW REF OFF NORMAL annunciator, a COMPAR white indicating light or. panel 9-5, a COMPAR amber light on the respective flow unit and on panel 9-14. Scram setpoint is .58w + 61 = 111.46% Rod Block setpoint is .58w + 49.5 = 99.96% 51
l l l RO Written Examination Question No.: 90 1 K/A: 216000 Kl.13 Importance: 3.4 ' Tier: 2 Group: I f Cognitive Level: 3 Exam Bank No.: new
Reference:
COR002-15-02 Objective: COR002-15-02, obj. 2f, 6e, 4a, 5a i 4.6.1, Section 4.1 The plant is operating at 15% reactor power. The Main Turbine has just been synchronized with I the grid. 1 The Feedwater Level Control System is in automatic - three element control. NBI-LT-52B is l inoperable due a failure of the level transmitter. The level transmitter failure caused the control l panel indication to indicate +60 inches. The "A" level instrument (transmitter NBI-LT-52A) is selected for control. Prior to removing , the NBI-LT-52B level transmitter from service for maintenance, the eg alizing valve for NBI-LT- l 52A is fully opened by I&C. 4 Assume NO operator actions are taken. l l Which one of the following describes the effect of these failures?
- a. The RFPs and the Main Turbine will trip.
- b. Only a low reactor water level alarm is received.
- c. Only a high reactor water level alarm is received.
- d. Only a % scram is received on RPS trip system "A" Answer: a.
- b. A full scram is received on lowlevel.
- c. A high level trip occurs.
- d. A full scram is received on lowlevel.
Attachment:
Provide a drawing of the level instrumentation for the narrow range instruments (B&R 2026 Sh.1) 52
1 RO Written Examination Question No.: 31 K/A: 217000 K6.04 Importance: 3.5 Tier: 2 Group: 1 Cognitive Level: 2 Exam Bank No.: new
Reference:
COR002-18-02 . Objective: COR002-18-02,10c,1lb 2.2.67, Section 4.2.1.12 A reactor scram due to MSIV closure while at 100% power occurred. RPV water level reached a minimum of -20 inches indicated on the wide range RPV water level instrument. RCIC has been placed in a test lineup, recirculating water to the ECSTs. The following conditions exist: RPV water level +20 inches indicated on wide range RPV level instrument (stable) - Suppression Pool water level 0.0 inches indicated on the narrow range RPV level instrument ECST A LEVEL 1.8 feet above the bottom of the tank ECST B LEVEL 1.7 feet above the bottom of the tank Which one of the following describes the effect on the RCIC valve alignment?
- a. ECST suction valve MO-18 remains open and the suppression pool suction valve MO-41 remains closed. Test Bypass to ECST valve MO-30 and ECST Test Line Shutoffvalve MO-33 remain open.
- b. Suppression pool suction valve MO-41 fully opens, and then ECST suction valve MO-18 closes. Test Bypass to ECST valve MO-30 and ECST Test Line Shutoff valve MO-33 close.
- c. ECS i suction valve MO-18 remains open and the suppression pool suction valve MO-41 remains closed. Test Bypass to ECST valve MO-30 and ECST Test Line Shutoffvalve MO-33 close.
- d. Suppression pool suction valve MO-41 fully opens, and then ECST suction valve MO 18 closes. Test Bypass to ECST valve MO-30 and ECST Test Line Shutoff valve MO-33 remain open ;
Answer: b. A low level in either ECST causes the ECST suction valve MO-18 to automatically close when the open limit switch of the suppression pool suction valve MO-41 is actuated. The MO-30 and MO-33 auto close when the MO-41 opens. l l 53 j i i
l I RO Written Examination Question No.: 29 K/A: 218000 A4.02 Importance: 4.2* Tier: 2 Group: 1 Cognitive Level: 2 Exam Bank No.: new
Reference:
COR002-16-02 Objective: COR002-16-02, obj. Sb, 6a 2.4.4.1, Section 4.1.3 The following conditions have been present for 2 minutes: RPV water level indicates -148 inches on the wide range RPV level instrument Reactor pressure is 300 psig Drywell pressure is 22 psig Assume ALL equipment operates as designed. Which one of the following describes the current status of the ADS valves, and the actions necessary to close or maintain them closed? The ADS valves are ...
- a. open.
The ADS A INHIBIT and the ADS B INHIBIT switches must be placed in INHIBIT.
- b. closed.
The ADS A INHIBIT and the ADS B INHIBIT switches must be placed in INHIBIT.
- c. closed.
The ADS LOGIC A TIMER and the ADS LOGIC B TIMER pushbuttons must be depressed at least every 90 seconds.
- d. open.
The ADS A INHIBIT and the ADS B INHIBIT switches must be placed in INHIBIT and then the ADS LOGIC A TaiER and ADS LOGIC B TIMER pushbuttons must be depressed. Answer: a.
- b. ADS valve logic is satisfied and the valves are open l
- c. ADS valve logic is satisfied and the valves are open ;
- d. Depressing the reset push buttons is not required l
l l 54 1
RO Written Examination Question No.: 80 K/A: 223001 A1.09 Importance: 3.5 Tier: 2 Group: 1 Cognitive Level: 2 Exam Bank No.: nr.w
Reference:
2.6.1, sect 6.1.1 Objective: SKL012-42-03,02j 2.2.25, sect 2.2.6 Given the following conditions: ALL 4160 volt busses are de-energized VBD-H Manual Transfer switch is in ALTERNATE ALL Division I DC power sources are unavailable Which one of the following describes the indicators available to be used as an information source to take action regarding Suppression Pool Temperature without reliance on other indications?
- a. PMIS/SPDS only.
- b. Alternate Shutdown Panel instruments only.
- c. One (1) of the Suppression Chamber Water Temperature recorders only.
- d. Both PMIS/SPDS and one (1) of the Suppression Chamber Water Temperature recorders.
Answer: b.
- a. PMIS cannot be used as a sole source.
- c. NBPP is not available as DIV I DC is de-energized and no AC power is available.
- d. NBPP is not available as DIV I DC is de-energized and no AC power is available to the temperature recorder, PMIS cannot I ; used as a sole source.
55
'b
/
RO Written Examination Question No.: 7 K/A: 223002 A3.02 Importance: 3.5 Tier: 2 Group: 1 Cognitive Level: 2 Exam Bank No.: new
Reference:
COR002-03-02 . Objective: COR002-03-02, Sa, 6a, 6b, 6d, 6f, 2.1.22, Section 8.0 6j 2.2.33, Section 4.1 4.9, Section 2.0 Given the following parameters: Drywell pressure is 12.7 psig RPV water level lowered to -50 inches on the wide range RPV level instrument Main Condenser vacuum has degraded to 14"Hg Reactor pressure is 85 psig Which one of the following describes equipment that isolates under these conditions?
- a. MSIVs and Main steam line drains, HPCI steam supply line, RWCU isolation valves.
- b. RWCU isolation valves, Drywell floor and equipment drains, HPCI steam supply line.
- c. RCIC steam supply line, Recirculation loop sample valves, MSIVs and Main steam ;
line drains. ;
- d. Recirculation loop sample lines, RCIC steam supply line, Drywell Floor and .
Equipment Drain valves ! Answer; b. i
- a. Group 1 isolation is not met.
l
- c. Group 1 and 5 isolatkus are not met.
- d. Group 5 isolation is not met.
56
i RO Written Examination Question No.: 77 K/A: 239002 K6.04 Importance: 3.0 Tier: 2 Group: 1 Cognitive Level: 1 Exam Bank No.: new
Reference:
COR002-16-02 Objective: COR002-16-02, obj. 8f, 2b 2.4.2.3.1, Section 4.8 During normal operation at 100% power,125 VDC panel "A" is lost. Which one of the following desemibes the effect on the Low-Low Set SRVs? , All Low-Low Set SRVs ...
- a. remain powered from their normal power supply,
- b. automatically transfer to their alternate power supply.
- c. are de-energized with NO alternate power supply available.
- d. must be manually transferred to their alternate power supply.
Answer: b. Both LLS logic channels are normally powered from 125 VDC panel AA2, with an alternate supply from 125 VDC panel BB2. On a loss of power (panel AA2), both channels will automatically transfer to the alternate supply. 4 l 4
)
I l 1 1 57
RO Written Examination Question No.: 93 K/A: 241000 K3.06 Importance: 4.1* Tier: 2 Group: 1 Cognitive Level: 2 Exam Bank No.: new
Reference:
2.4.5.2.3 Objective: COR002-09-02, 6.h, 7.b, 7.e, During a plant startup, the turbine hasjust been synchronized per SOP 2.2.14 "22 KV Electrical System." At this point, the digital controller fails and prevents generator load from automatically ramping up, this causes a turbine generator trip on reverse power. Which one of the following statements describes how reactor pressure control will respond? The pressure setpoint ... '
- a. remains in automatic. The bypass valves remain in their pre-tripped position until opened by the operator.
- b. transfers to manual at the existing setpoint. The bypass valves automatically maintain reactor pressure at that setpoint.
- c. remains in automatic. The bypass valves transfer to manual and must be manually positioned to control reactor pressure.
- d. tiansfers to manual at the existing setpoint. The bypass valves remain in their pre-tripped position until opened by the operator.
Answer: b.
- a. DEH shifts to manual and closes the control valves. BPVs will open to control pressure at the preset pressure.
- c. Pressure control shifts to manual and the bypasses respond to control pressure because the pressure transducer has NOT failed.
- d. BPVs will open to control pressure at the preset pressure.
58
1 1 RO Written Examination Question No.: 89 K/A: 259001 A3.10 Importance: 3.4 Tier: 2 Group: 1 Cognitive Level: 1 Exam Bank No.: new
Reference:
COR002-02-02 Objectiv~e: COR002-02-02, obj. 6g 2.4.9.4.3 step 5.1.1 During operation at 75% power a breaker electrical fault causes a trip of one (1) Condensate Booster Pump. RFP suction pressure lowers to 225 psig for 13 seconds. Which one of the following describes the response of the feedwater system?
- a. Both RFPs trip.
- b. Both RFPs continue to operate.
- c. The "A" RFP will continue to operate. The "B" RFP will trip.
- d. The "B" RFP will continue to operate. The "A" RFP will trip.
Answer: d. When RFP suction lowers to 260 psig, the "A" RFP trips afler a 10-second time delay and the "B" RFP trips after a 15-second time delay. I l l l 59 , I
\
l
RO Written Examination Question No.: 86 K/A: 259002 A1.02 Importance: 3.6 Tier: 2 Group: 1 Cognitive Level: 2 Exam Bank No.: new
Reference:
COR0C?.-32-02 Objective: COR002-32-02, obj. 5b, 7a, 7b 2.4.5.1, Section 4.4, 6.3 The plant is operating at power with the following reactor vessel level control alignment: RFC-LC-83, MASTER LEVEL CONTROLLER in balance RFC-MA-84A, FW CONTROLLER STATION A in balance RFC-MA-84B, FW CONTROLLER STATION B in balance Feedwater flow is approximately 9.6x10'lbm/hr. Steam flow is approximately 9.6x106 lbm/hr. RPV water level is +35 inches. The Master Controller OUTPUT slowly fails downscale. RPV water level lowers to +27 inches when the operator places the "A" and "B" RFP controllers to MANUAL. Assume NO additional action is taken by the operator. Which one of the following describes the response of Feedwater Flow and RPV water level? Feedwater flow will ... 6
- a. rise to 9.6x10 lbm/hr. Level will rise to +42 inches.
6
- b. rise to 9.6x10 lbm/hr. Level will remain at +27 inches.
6
- c. rise above 9.6x10 lbm/hr. Level will rise to +42 inches. l l
- d. remain below 9.6x10' lbm/hr. Level will continue to lower.
1 Answer: b.
- a. Level will not rise. )
6
- c. Feed flow will not rise above 9.6x10 lbm/hr. Level will not nse. ,
6
- d. Feed flow rises to 9.6x10 lbm/hr. Level does not lower j 1
I 1 l 60 I i l h
RO Written Examination Question No.: 85 K/A: 259002 Kl.05 Importance: 3.6 Tier: 2 Group: 1 Cognitive Level: 1 Exam Bank No.: new
Reference:
2.2.28.1, Sect. 8.2.11, Objective: COR002-32-02,9.c 2.3.2.28, 9-5-2/G-4 2.3.2.1, A-1/F-6 6 A Startup is in progress with feedwater flow at lx10 lbm/hr. "A" Feedpump control is in automatic on the Master Level Controller (RFC-LC-83). The Startup Master Controller (RFC-LC-130) is in manual. Which one of the following would cause the feedwater pump to go into " Track and Hold?"
- a. Startup Master Controller slowly fails upscale.
- b. Startup Master Controller slowly fails downscale.
- c. Selected RPV water level instrument slowly fails upscale.
- d. Selected RPV water level instrument slowly fails downscale.
Answer: d. The selected instrument failure will cause a < 6 ma. output to be sensed by the track , and hold circuit.
- a. The startup level controller output is not sensed by the track and hold circuit.
- b. The startup level controller output is not sensed by the track and hold circuit.
- c. The selected level transmitter output must drop < 6 ma. to initiate the track and hold circuit.
An upscale failure would result in a 50 ma. output. 61
t RO Written amination Question No.: 2 K/A: 2v1000 A4.02 Importance: 3.1 Tier: 2 Group: 1 l Cognitive Level: 1 Exam Bank No.: new
Reference:
2.4.8.4.1,2.2.73, Sect. 3.3, & 5. Objective: COR002-28-02, 5.a, 9.b 1 While operating at full power, a problem with the Reactor Building HVAC system has resulted in ! loss ofReactor Building Ventilation. Which one of the following is required to aid in improving the Reactor Building dhTerential ) pressure using the "A" Standby Gas Treatment (SGT)? Start the "A" SGT fan, ...
- a. verify SGT-AO-249, SGT A INLET, and SGT-AO-251, SGT A DISCHARGE automatically open only.
- b. manually open SGT-AO-249, SGT A INLET, and SGT-AO-251, SGT A l DISCHARGE and then open AD-R-1B, PRIMARY CONTAINMENT '
ISOLATION. l
- c. manually open SGT-AO-249, SGT A INLET and SGT AO-251, SGT A DISCHARGE, and then open HPCI-AO-275, HPCI GLAND EXHAUST 1 DISCHARGE TO SGT.
- d. verify SGT-AO-249, SGT A INLET, and SGT-AO-251, SGT A DISCHARGE l
automatically opens and then open AD-R-1C, STANDBY GAS TREATMENT ROOM SUPPLY. l Answer; a. I
- b. SGT-AO-249 & 251 should automatically open, AD-R-1B, PRIMARY CONTAINMENT l ISOLATION would take a suction on a dead leg of pipe.
l
- c. SGT-AO-249 should automatically HPCI-AO-275, HPCI GLAND EXHAUST DISCHARGE TO SGT would take a suction on a dead leg of pipe.
- d. AD-R-IC is failed open 1
l I l l l i l 62 1
RO Written Examination Question No.: 71 K/A: 264000 A2.10 Importance: 3.9 Tier: 2 Group: 1 Cognitive Level: 2 Exam Bank No.: new
Reference:
COR002-08-02 Objective: COR002-08-02,13c,9b l DG2 has been started and loaded to 3850 KW for the monthly surveillance when a reactor scram due to high drywell pressure occurs. Two (2) minutes following the LOCA, ALL offsite sources are lost. Which one of the following describes the effect the above conditions will have on DG2 and 4160 l Bus 1G7 I I
- a. DG2 output breaker will NOT trip. DG2 will remain connected to Bus 1G. ;
l
- b. DG2 engine will trip and is NOT available until the Diesel Generator over current lockout is reset.
1
- c. DG2 output breaker will trip when the LOCA signal is received. DG2 remams unloaded when offsite power is lost.
- d. DG2 output breaker will trip when the LOCA signal is received. DG2 will re-connect l to Bus 1G when offsite power is lost.
Answer: d. The DG output breaker receives a trip signal opening the breaker when the LOCA signal occurs. The DG would then run unloaded. The DG will pick up 4160 Bus 1G when it is de-energized (LOOP). 63 j l
- ~. . . - - .- -- . .- .
RO Written Examination Question No.: 72 K/A: 254000 Generic 2.1.32 Importance: 3.4 Tier: 2 Group; 1 ) 1 Cognitive Level: 1 Exam Bank No.: new
Reference:
COR002-08-02 , Objective: COR002-08-02,9g 2.2.20, Section 5.0 COR002-34-02, 2b The following Diesel Generator Isolation Switches are positioned to ISOLATE: l IS/DG-1 A and IS/DG-1B IS/EGI and IS/EGI-CT Which one of the following describes the effect on DGl? l DG1 must be started ...
- a. locally. DG1 output breaker must be manually closed from panel"C" locally D31 output breaker must be manually closed from the local panel.
- c. from the control room. DGl output breaker must be manually closed from panel"C".
1
- d. from the control room. DGl output breaker must be manually closed from the local i panel.
l Answer: b. a,c,d All remote and automatic start fea..ses of DG1 are disabled. Panel"C" indications and controls are disabled. I 1 l l l j l l l 64 i
RO Written Examination Question No.: 55 K/A: 201003 A1.01 Importance: 3.7 Tier: 2 Group: 2 Cognitive Level: 1 Exam Bank No.: new
Reference:
COR002-05-02 Objective: COR002-05-02, obj. llb COR002-04-02 COR002-04-02, obj.12c The unit operating at 100% power near the end of cycle with all control rods fully withdrawn. The scram inlet valve (CRD-AOV-126) for control rod 30-31 opens. Which one of the following describes the response of the plant over the next five (5) minutes? Reactor power will ...
- a. be downscale on APRMs.
- b. remain at 100% reactor power.
- c. rise, but the unit will continue to operate at power.
- d. lower, but the plant will continue to operate at power.
Answer: d.
- a. A reactor scram will not occur. The SDV level will not change.
- b. A single control scram will reduce reactor power.
- c. A single control scram will reduce reactor power.
l I 65
. -. - - - .. -. . _ .- .. =- - .-_-_ .
RO Written Examination Question No.: 57 K/A: 201006 A2.03 Importance: 3.0 Tier: 2 Group: 2 Cognitive Level: 1 Exam Bank No.: new
Reference:
COR002-26-02 Objective: COR002-26-02, obj. 8a,9 2.3.2.27 2.4.1.1.3 The plant is operating at 9% reactor power. All control rods in the current rod group are at their insert limit of 36. One of the control rods in the current group drins in from position 36 to position 00. Which one of the following describes the effect on the Rod Worth Minimizer (RWM) if the drining control rod is selected? RWM will identify the control rod as ...
- a. an Insert Error. A control rod block will NOT be enforced.
- b. a Withdrawal Error. A control rod block will NOT be enforced.
- c. an Insert Error and a Select Error. A control rod block will be enforced.
- d. an Insert Error and a Withdrawal Error. A control rod block will be enforced.
Answer: a. A Select Error occurs when a non-error rod is selected. The drining rod is an error rod. The rod will not be a Withdrawal Error at position 00. A rod past its' insert limit is an insert error. No rod block occurs for a single insert error. 66
. . - . _ .- _ _ - - . __ . . - = _ . .- ..
J l j RO Written Examination Question No.: 18 ) K/A: 202001 K4.16 Importance: 3.3 Tier: 2 Group: 2 I Cognitive Level: 2 Exam Bank No.: new
Reference:
2.2.68, COR002-22-02 Objective: COR002-22-02, 5.d, 6.I,10.1,10.m During a plant startup the "A" recirculation pump trips causing the following conditions: Reactor poweris 39% l Steam flow is 40% l Feedwater flowis 41%
"B" recirculation pump is operating "A" feedwater pump is operating.
l Both recirculation MG sets M/A transfer stations are in MANUAL set at 57 % demand. 1 l What is the expected scoop tube speed position setpoint on the "A" recirculation pump 4 minutes I after the recirculation pump trip? (Assume operator actions for the tripped recirculation pump have been completed.) I l
- a. 0% 1
- b. 22%
l 1
- c. 45 %
- d. 57 %
Answer: b.
- a. c. d. Pump speed is limited by the dual limiter to 22% speed because the discharge valve is closed on the tripped ("A") pump.
I 1 67 l l
i RO Written Examination Question No.; 17 K/A: 204000 Kl.01 Importance: 3.1 Tier: 2 Group: 2 Cognitive Level: 2 Exam Bank No.: new
Reference:
COR001-20-02 Objective: COR001-20-02, obj. 4k, 7f, 7h The unit is in MODE 2 with a startup in progress. Reactor pressure is being maintained at 300 psig using the main turbine bypass valves. The "A" reactor recirculation pump trips and then a Group 3 isolation signal is received. Assume NO operator action is taken. ; Which one of the following describes the consequence on the plant?
- a. Reactor water level will rise outside the allowed band.
- h. Reactor water level will lower and a reactor scram will be received. j
- c. A prerequisite for the "A" reactor recirculation pump start CANNOT be determined. I l
- d. RWCU non-regenerative heat exchanger outlet temperature will rise damaging the l demineralizer resin. )
I Answer: c. l
- a. Reactor water level will rise but will be within the required band (a shutdown is not required based on reactor water level)
- b. Reactor water level will rise.
- d. Temperature willlower.
I l 68
I . \ RO Written Examination Que: tion No.: 14 K/A: 205000 K6.08 Importance: 3.5 Tier: 2 Group: 2 ; Cognitive Level: 1 Exam Bank No.: new l
Reference:
COR002-27-02 Objective: COR002-27-02, obj. 6,4e 2.3.2.21, 9-3-1/A-3 SKL010-01-02, A.4, B1 2.4.2.4.1, Attachment 3, Section 1.1 2.2.70, Section 2.2.4 4. The plant is shutdown with the following conditions "C" RHR Pump operating in Shutdown Cooling Mode "A" RHR EWBP is operating ) RPV Level is in the prescribed RPV level band for shutdown cooling operation , l The following annunciator is received:
- 9-3-1/A-3, RHR SWBP A TRIP Which one of the following actions is required to be taken?
- a. Manually start the "C" RHR SWBP.
- b. Throttle closed RHR-MO-66A, HX Bypass valve.
- c. Verify that the "C" RHR SWBP automatically starts.
l
- d. Throttle open SW-MO-89A, HX-A SW Outlet valve.
t Answer: a.
- b. Valve should be opened.
- c. No automatic start.
- d. During MODE 4 or 5 operations, a normal SW flow of 4000 gpm can be supplied to HX A or B at service water pressure with SWBPs A and C windmilling. This requires lifled leads.
l l l 69
RO Written Examination Question ~No.: 52 K/A: 214000 A3.03 Importance: 3.5 Tier: 2 Group: 2 Cognitive Level: 2 Exam Bank No.: new
Reference:
COR002-20-02 Objective: COR002-20-02, obj. I1,13b A non-selected control rod at position 36 becomes uncoupled. The CRDM will be fully withdrawn. While fully withdrawing the control rod to position 48, which one of the fol!owing describes when the uncoupled control rod can be identified using RPIS?
- a. As soon as the rod is selected.
- b. When the CRDM is moved from position 36.
- c. When the CRDM coupling check is perfonned.
- d. When the RMCS timer times out at position 48.
Answer: c. An uncoupled control rod cannot be detected by RPIS until it is withdrawn to the overtravel position. This is done during the coupling check. l l l l 1 1 i
)
70 , I l 1
RO Written Examination Question No.: 56 K/A: 215002 Generic 2.1.28 Importance: 3.2 Tier: 2 Group: 2 Cognitive Level: 2 Exam Bank No.: new
Reference:
COR002-24-02 Objective: COR002-24-02, obj.1,4a 4.1.5, Section 2.0 A plant startup is in progress with reactor power at 32%. When withdrawing a control rod 22-23 from position 08 to position 16, the Reactor Operator mistakenly continues to withdraw control rod 22-23 using Rod Out Notch Override control when a control rod block is received. Which one of the following is the reason for this control rod block?
- a. Prevent exceeding the MCPR safety limit.
- b. Prevent exceeding a peak fuel enthalpy of 280 cal /gm.
4
- c. Preserve the integrity of the reactor coolant pressure boundary.
- d. Minimize the energy that must be absorbed following a LOCA.
Answer: a. The rod block is a result of the RBM. The design of the RBM is to prevent exceeding the MCPR safety limit. l i I i i I 71
RO Written Examination Question No.: 24 K/A: 219000 A4.02 Importance: 3.7* Tier: 2 Group: 2 Cognitive Level: 2 Exam Bank No.: new
Reference:
COR002-23-02 Objective: COR002-23-02, Obj. 3p, 5b 2.2.69.3, Section 4.0 The following conditions are present following a LOCA: i Drywell pressure 12 psig and slowly rising RPVlevel(Fuel Zone instrument) + 5 inches and slowly rising Reactor pressure 20 psig RHR Loop A and B Secured CS Loops A and B Injecting following automatic initiation 1 Three (3) minutes have elapsed since annunciator RX LOW PRESS 291-436 PSIG i alarmed. The Control Room Operator is directed to place RHR Loop B into Suppression Pool Cooling. For the current plant conditions, which of the following operator action (s) must be performed to manually close the RHR-MO-27B, Outboard Injection valve?
- a. Wait two (2) additional minutes.
- b. Depress the Loop B Initiation Logic Reset pushbutton.
- c. Place the containment cooling valve control permissive switch is in MANUAL.
- d. Place the contLinment cooling permissive switch in MANUAL and the containment cooling 2/3 core permissive switch in MANUAL OVERRIDE.
l Answer: a. 5 minute timer must time out. I
- b. Cannot be reset with the current conditions. 1
- c. No input into MO-27B logic. l
- d. No input into MO-27B logic. !
l l 72 l
RO Written Examination Question No.: 36 I'JA: 226001 K3.01 Importance: 3.6 Tier: 2 Group: 2 Cognitive Level: 1 Exam Bank No.: new
Reference:
INT 008-06-13 Objective: INT 008-06-13, obj. 3 A LOCA is in progress and drywell sprays have been initiated. Which one of the following will result if drywell sprays are NOT terminated and drywell pressure lowers below 0.0 psig,7
- a. Chugging at the outlet of the downcomer.
- b. Partial de-inerting of the Primary Containment.
1
- c. Mechanical failure (collapse) of the Torus downcomer ring header.
- d. Mechanical failure of the Reactor Building to Torus vacuum breakers.
1 Answer: b. l
- a. Phenomenon associated with initiation of DW sprays.
- c. Phenomen associated with evaporative cooling due to spraying while in the unsafe region of the DW spray initiation limit curve.
- d. Event is within the design of the vacuum breakers.
1 73
_ _ _ . . - - - . . . . - . = _ . _ .-. ._ . RO Written Examination Question No.: 61 K/A: 239001 K2.01 Importance: 3.2* Tier: 2 Group: 2 Cognitive Level: 2 Exam Bank No.: new
Reference:
COR002-14-02 Objective: COR002-14-02, obj. 2a, 3h, 7j 2.1.22, Section 8.1 , The plant is operating at 100% power with the RPS "A" Power Transfer Switch and the RPS "B" Power Transfer Switch in the MG-SET position. The 480V Station Service Transformer IG is de-energized. Which one of the following describes the position of the MSIVs following the power loss?
- a. Inboard and outboard MSIVs CLOSE.
- b. Inboard and outboard MSIVs remain OPEN.
l
- c. Inboard MSIVs CLOSE. Outboard MSIVs remain OPEN.
- d. Outboard MSIVs CLOSE. Inboard MSIVs remain OPEN.
Answer: b. i a,c,d The AC and DC solenoid power that remains maintains the MSIVs open. l l l l i i 74
RO Written Examination Question No.: 94 K/A: 245000 K4.06 Importance: 2.7 Tier: 2 Group: 2 Cognitive Level: 2 Exam Bank No.: new
Reference:
2.2.14, Section 4.4,4.6,4.7 Objective: COR001-13-02, Obj. 6c, 6d, 7a . 2.4.9.2.2 The plant is operating at 100% reactor power when the following conditions are observed: Annunciator C-3/G-1, MAIN GEN VOLTAGE REG TROUBLE, alarms Ronan Annunciator (4022), MAIN GEN VOLT REG FORCING ALARM, is displayed Generator reactive load has risen by 300 MVARS out and continues to rise Assume NO operator action is taken. . Which one of the following describes the design response of the Main Generator to the above conditions? Field excitation current will ... )
- a. raise to correct the problem. If the problem continues, the voltage regulator will l remain in AUTOMATIC. Aner a time delay, the Main Generator will trip. !
- b. lower to correct the problem. If the problem continues, the voltage regulator will 4
remain in AUTOMATIC. After a time delay, the Main Generator will trip.
- c. raise to correct the problem. If the problem continues, aner a time delay, the voltage regulator will trip (shin to MANUAL). If the problem continues with the regulator in MANUAL for an additional time delay, the Main Generator will trip.
- d. Iower to correct the problem. If the problem continues, aner a time delay, the voltage regulator will trip (shift to MANUAL). If the problem continues with the regulator in MANUAL for an additional time delay, the Main Generator will trip.
l Answer: d. Forcing Alarm Overexcitation (OXP-2) protects the Main Generator filed windings from excessive temperature due to prolonged overexcitation. OXP-2 lowers the field excitation current. If the field excitation current is not reduced to a safe value within a specified time, then the automatic voltage regulator is tripped, and a second timer starts. If the over-excited condition still exists when the second timer times out, then the Main Generator trips. d 75
RO Written Examination Question No.: 69 K/A: 262001 K3.01 Importance: 3.5 Tier: 2 Group: 2 Cognitive Level: 2 Exam Bank No.: new
Reference:
COR002-27-02 Objective: COR002-27-02, obj. Sc, 3g, 4c 2.2.71, Section 4.0 l 5.2.5, Section 2.7 The unit is operating at 100% reactor power. SW pump alignment is as follows: SW pumps "A," "B" and "C" are operating Mode Selector switches for the "A" and "B" SW pumps are in STANDBY Mode Selector switches for the "C" and "D" SW pumps are in AUTO A loss of offsite power occurs. Both DGs start and energize busses IF and IG. Assume NO operator actions are taken. i Which one of the following describes the Service Water pumps that will be operating by design following this event?
- a. A and B
- b. A and C
. c. B and D
- d. C and D Answer: a.
4 Only the SW pumps selected to standby start 13 seconds after buses IF and IG are energized
- from an emergency power source.
4 4
, 76
RO Written Examination Question No.: 73 K/A: 262002 K6.02 Importance: 2.8 Tier: 2 Group: 2 Cognitive Level: 1 Exam Bank No.: new
Reference:
COR002-07-02 Objective: COR002-07-02, obj. 8q 2.4.6.7, Section 2.1 The 250 VDC supply to the No Break Power Panel (NBPP) Inverter is lost. Which one of the following describes how the NBPP is powered after this event?
- a. The NBPP will automatically transfer to MCC-R.
- b. The inverter will automatically transfer to the alternate 250 VDC Bus.
- c. The NBPP will NOT automatically transfer but can be manually transferred to MCC-L.
- d. The inverter will NOT automatically transfer but can be manually transferred to ;
MCC-LX. Answer; a. l l NRC exam #94 evaluated location of a transfer switch. )
- b. Does not transfer to DIV II DC.
- c. Automatically transfer and cannot be manually powered from MCC-L.
- d. Automatically transfer and cannot be manually powered from MCC-LX.
1 77
i RO Written Examination Question No.: 79 K/A: 263000 Generic 2.4.11 Importance: 3.4 Tier: 2 Group: 2 Cognitive Level: 2 Exam Bank No.: new 1
Reference:
COR002-07-02 Objective: COR002-07-02, obj. 6f 2.4.6.10, Section 4.8 SKL010-01-02, A.4, B.1, B.3 The unit is operating at 60% power when a loss of 125 VDC Panel bbl occurs. i In accordance with Abnormal Procedure 2.4.6.10, "125 VDC System Failures," which one of the i following is required? l
- a. Operate the "B" CRD pump.
- b. Transfer DG2 control power to its alternate source.
I
- c. Transfer the "B" recirculation pump to the startup transformer. l
- d. Entry into single loop operations on the "A" recirculation pump.
i Answer: d.
- a. The "A" CRD pump is operated.
- b. Cannot be performed with current plant design.
- c. The "A" recirculation pump needs to be transferred to the startup transformer.
I j l 78
RO Written Examination Question No.: 83 K/A: 271000 A2.04 Importance: 3.7 Tier: 2 Group: 2 Cognitive Level: 1 Exam Bank No.: new
Reference:
CO.R001-16-02 Objective: COR001-16-02, obj. 8g,10b During a reactor heatup and pressurization at 300 psig. The OFF-GAS HIGH RAD and the OFFGAS TIMER INITIATED annunciators alarm. If the OFFGAS TIMER INITIATED alarm is sustained for 20 minutes, which one of the following automatic actions will occur?
- a. Mechanical Vacuum Pumps trip.
- b. AOG-AO-902 "AOG Return valve" closes. I
- c. OG-AO-254 "Offgas System Isolation valve" opens.
- d. AR-AO-12 "30 Minute Holdup Pipe Drain valve" opens.
1 Answer: b. I I
- a. MSL high-high radiation
- c. OG-AO-254 closes
- d. AR-AO-12 closes l
l 79
E RO Written Examination Question No.: 76 K/A: 286000 K5.05 Importance: 3.0 Tier: 2 Group: 2 Cognitive Level: 2 Exam Bank No.: new
Reference:
COR002-08-02, COR001-05-02 Objective: COR002-08-02, obj. 6d 2.2.2, Section 2.0 COR001-05-02, obj. 5f Both Emergency Diesel Generators are running following a start on a LOCA signal. Which one of the following signal (s) will actuate one (1) of the Emergency Diesel Generator CO2 Fire Suppression System, including the effect on the associated DG room ventilation?
- a. Actuation of the manual release station at the exit to the Turbine Building will initiate CO2 immediately and trip the DG room ventilation.
- b. One (1) thermal detector in the DG room sensing high temperature will initiate CO2 immediately and DG room ventilation continues to operate.
- c. One (1) thermal detector in the DG fuel oil day tank room sensing high temperature will initiate CO2 after a time delay and trip the DG room ventilation.
- d. Two (2) of the four (4) DG area smoke detectors activated in a DG room will initiate CO2 aller a time delay and DG room ventilation continues to operate.
Answer: d.
- a. DG room ventilation will not trip because a LOCA signal is present
- b. There are no thermal detectors in the DG room.
- c. DG room ventilation will not trip because a LOCA signal is present 80
RO Written Examination Question No.: 5 K/A: 290001 A3.01 Importance: 3.9 Tier: 2 Group: 2 Cognitive Level: 2 Exam Bank No.: new
Reference:
COR001-08-02 Objective:COR001-08-02, obj. Ilb 2.1.22, Section 8.5 The unit is at 100% power. Irradiated fuelis being arranged in the fuel pool to support receipt of new fuel when annunciator 9-4-1/E-4, RX BLDG VENT HI-HI RAD is received. RONAN CRTs display the following: (1763) RX BLDG VENT MONITOR A HI-HI RAD (1764) RX BLDG VENT MONITOR B HI-HI RAD (1780) RX BLDG VENT MONITOR D HI-HI RAD Which one of the following describes the effect on the Secondary Containment and why? The Reactor Building HVAC supply and exhaust fans ...
- a. trip and the system isolates because a Group 6 isolation is actuated.
- b. trip and the system isolates because a Group 2 isolation is actuated.
- c. continue to operate and the system does NOT isolate because the DIV I logic has NOT tripped.
- d. continue to operate and the system does NOT isolate because the DIV II logic has NOT tripped.
Answer: a.
- b. Group 2 isolation is high drywell pressure and low reactor water level c, d Both DIV I and DIV II are above the setpoint for the group 6 isolation.
A group 6 isolation is actuated and the fans trip and the system isolates. 81
n- - - , - - - -e. neae J.. an - A - ,S - sa ,n~m->-= M + RO Written Examination Question No.: 82 K/A: 300000 K3.02 Importance: 3.3 Tier: 2 Group: 2 Cognitive Level: 1 Exam Bank No.: new
Reference:
5.2.8 Objective
COR002-02-02, obj. 5d,8a COR002-02-02, p. 45 COR002-32-02, obj. 8a , COR002-32-02, p. 31 ' The unit is operating at 100% reactor power. The instrument air header completely ruptures. Assume NO operator action is taken. Which one of the following describes the effect on the Condensate & Feedwater system valves?
- a. MC-AOV-FCV17 " System Minimum Flow" valve fails open.
- b. MC-AOV-FCV11 A/B "RFP Minimum Flow" valves fail closed.
- c. MC-AOV-B1 " Condensate Demin System Bypass " valvt -ails closee.
- d. RF-AOV-FCVI1 AA/BB "RFPA/B "Startup Flow Control" valvei fail open.
Answer: a.
- b. valves fail open
- c. valve fails open
- d. valves fail closed.
l l l t 4 l 82 i I i
)' l I RO Written Examination Question No.: 13 K/A: 400000 Kl.02 Importance: 3.2 Tier: 2 Group: 2 Cognitive Level: 2 Exam Bank No.: new
Reference:
COR002-19-02 Objective: COR002-19-02, obj. 6a, 6i, 6j 2.3.2.16, M-1/A-4, Section 3 l Which one of the following will cause a high water level in the Reactor Equipment Cooling (REC) Surge Tank?
- a. RHR pump seal coolerleak.
J
- b. Tube leak in the REC heat exchanger.
1
- c. REC heat exchanger outlet header pressure at 65 psig. ;
I
- d. Tube leak in the RWCU non-regenerative heat exchanger.
Answer: d.
. Possible leakage sources are the RWCU NRHX, RWCU and Reactor Recirculation seal water coolers and Fuel Pool Cooling heat exchangers 83
l RO Written Examination Question No.: 8 K/A: 215001 Kl.05 Importance: :.3 Tier: 2 Group: 3 Cognitive Level: 1 Exam Bank No.: new
Reference:
COR002-31-02 Objective: COR002-31-02, obj. 9e Which one of the foli', whig describes the design response of a TIP detector that is in the reactor core when a Group 2 cnd a Group 6 isolation signal is received?
- a. Group 2 Isolation will cause the TIP to withdraw. Group 6 closes the ball valve.
- b. Group 6 Isolation will cause the TIP to withdraw. Group 6 closes the ball valve.
- c. Group 2 Isolation will cause the TIP to withdraw. Group 2 closes the ball valve.
- d. Group 2 QR Group 6 Isolation will cause the TIP to withdraw and close the ball valve.
Answer: c. a,b,d The group 2 isolation will cause a group 6 isolation, however, a group 6 isolation has no effect on TIPS. l 84
k RO Written Examination Question No.: 58 K/A: 234000 K5.02 Importance: 3.1 i Tier: 2 Group: 3 Cognitive Level: 2 Exam Bank No.: new
Reference:
COR001-21-02 Objective: COR001-21-02, obj. Sb 2.2.31, Attachment 3 Conditions during a core OFFLOAD are as follow: Mode Switch in the REFUEL position ALL rods are fully inserted into the reactor core A fuel assembly has been released in the fuel pool and the Main Hoist is raised to the Normal-Up position. The next step requires that a fuel assembly be removed from the reactor core and be transferred to a spent fuel pool location. Which one of the following describes when the ROD BLOCK refueling interlock is ACTUATED as the next step is being performed? When the refueling bridge is moved over the reactor core...
- a. with the Main Hoist in the Normal-Up position.
- b. and the Main Hoist is lowered from the Normal-Up position.
- c. the Main Hoist is lowered, and the fuel assembly is grappled.
- d. the Main Hoist is lowered, the fuel assembly is grappled, and the fuel assembly is raised.
Answer: b. 1
- a. Main hoist must be loaded with fuel _o.r NOT Normal-Up with the refueling bridge over the .
reactor core.
- c. The Rod Block is received when the Main Hoist is lowered. >
- d. The Rod Block is received when the Main Hoist is lowered.
i l 85
l l l 1 RO Writ +en Examination Question No.: 10 K/A: 268000 A4.01 Importance: 3.4 Tier: 2 Group: 3 Cognitive Level: 2 Exam Bank No.: new
Reference:
COR001-11-02 Objective: COR001-11-02, obj. 2d, 5 With the unit operating at 80% reactor power, the follovnng annunciators are received: 1 9-4-2/D-1, DRYWELL EQUIP. SUMP G HIGH LE'2L 9-4-2/C-1, DRYWELL EQUIP. SUMP G HI-HI LEVEL 9-4-2/B-1, DRYWELL EQUIP. SUMP G HIGH FILL-UP RATE ! 9-4-2/A-1, DRYWELL EQUIP. SUMP G HIGH TEMP l The last drywell equipment sump pump to operate was pump 1-G-1. ! i Which one of the following describes the response of the Drywell Equipment Drain system?
- a. Only the 1-G-1 pump is operating and the Sump "G" totalizer value is changmg.
Water is NOT recirculated through the heat exchanger. ) l
- b. Only the 1-G-2 pump is operating and water is recirculated through the heat l exchanger. The Sump "G" totalizer value is NOT changing.
)
- c. Both the 1-G-1 and the 1-G-2 pumps are operating and water is recirculated through the heat exchanger. The Sump "G" totalirer value is NOT changing. i
- d. Both the 1-G-1 and the 1-G-2 pumps are operating and the Sump "G" totalizer value is changing. Water is NOT recirculated through the heat exchanger.
Answer; c. l l
- a. Both pumps start when the hi-hi level is received. The recirculation valve opens and the l
discharge to Radwaste closes on high temperature.
- b. Both pumps start when the hi-hi level is reached. I
- d. The recirculation valve opens and the discharge to Radwaste closes on high temperature. !
86
RO Written Examination Question No.: 1 K/A: 288000, K5.02 Importance: 3.2 Tier: 2 Group: 3 Cognitive Level: 1 Exam Bank No.: new
Reference:
COR001-08-02 Objective: COR001-08-02, obj.13b Which one of the following describes how the Reactor Building Ventilation System maintains the required 0.25 inches of negative water pressure in the Reactor Building during nonnal operation of the system?
- a. At least one (1) more exhaust fan than supply fan is operated.
- b. A d/p controller regulates the operating supply fans vortex damper position.
- c. A d/p controller regulates the operating exhaust fans vortex damoer position.
- d. The capacity of the exhaust fans is greater than the capacity of the supply fans.
Answer: c.
- a. capacity is not used to maintain d/p
- b. d/p controller on suction dampers maintain flow through the filters
- d. not in accordance with system design 87
I RO Written Examination Question No.: 3 K/A: Generic 2.1.1 Importance: 3.7 Tier: 3 Group: N/A l 1 Cognitive Level: 1 Exam Bank No.: new l
Reference:
SKL008-01-02, Watchstanding Objective: SKL008-01-02, obj.12,13 Principles (RO) SKLO10-01-02, A.4 0.1, Section 3.5,3.6 While placing the "A" SGT train in service to support HPCI surveillance testing, the Control Room Operator recognizes that the procedure step identifies the control switch to open SGT FLOW /RX BLDG DP CONT, as SGT-DPCV-546B, rather than SGT-DPCV-546A. Which one of the following is required for this identified condition? ) 1
- a. Complete the start of the "A" SGT train, then make a pen and ink correction to the l
procedure. Notify the procedure owner aficr placing SGT in service.
- b. Stop action, make , pen and ink correction to the procedure, then proceed with starting the "A" SCf train. Notify the procedure owner after placing SGT in service.
- c. Discontinue actions to start the "A" SGT train and place SGT back in standby. The procedure shall be revised using the procedure change process prior to placing SGT l m service.
l
- d. Discontinue actions to start the "A" SGT train and leave all SGT components in their current position / condition. The procedure shall be revised using the procedure change process prior to placing SGT in service.
Answer: b. Work should continue for obvious typographical errors, spelling errors, title changes, l procedure number changes.
- a. Corrective action for the procedure is required prior to proceeding.
- c. Procedure change prior to implementation is only required if the procedure is adversely affected. , j
- d. SGT would be aligned to standby if procedure was terminated. Procedure change prior to implementation is only required if the procedure is adversely affected.
I I 88 I
RO Written Examination Question No.: 96 K/A: Generic 2.1.20 linportance: 4.3 Tier: 3 Group: N/A Cognitive Level: 1 Exam Bank Nr : new
Reference:
OI-7, Attachment E Objective: fLO10-01-02, A.4 2.3.1, section 8.1 A Surveillance Procedure is being performed on the ADS system. The expected annunciators have been " flagged" with translucent colored tape per Operations Instruction #7. The CRS has been informed of all expected alarms. The operator has referred to the alarm card for all expected alarms. Which one of the following describes the required actions when one of the " flagged" alarms is received as expected at the appropriate time? The operator shall acknowledge the alarm and ...
- a. is NOT required to report the alarm to the CRS. The operator does NOT have to refer to the associated alarm card.
] b. is NOT required to report the alarm ta the CRS. The associated alarm card shall be referred to and performed. i c. the annunciator shall be reponed to the CRS. The operator does NOT have to refer to the associated alarm card.
- d. the annunciator shall be reported to the CRS. The associated alarm card shall be referred to and performed.
Answer: a.
- b. The Alarm Procedure does not have to be referred to under these conditions per 0l#7 and 2.3.1.
- c. No report is required to the CRS per OI#7.
- d. No report is required to the CRS per OI#7. The Alarm Procedure does not have to be referred to under these conditions per OI#7 and 2.3.1.
89
RO Written Examination Question No.: 99 K/A: Generic 2.1.29 Importance: 3.4 Tier: 3 . Group: N/A Cognitive Level: 1 Exam Bank No.: new
Reference:
SKL008-01-02, Watchstanding Objective: SKLOO8-01-02, obj.10 Principles (RO) SKLO10-01-02, A.4 0.31, Section 8.2 Note In accordance with Administrative Procedure 0.31, " Equipment Status Control," which one of the following set of conditions permit the concurrent verification for a procedure step to be waived?
- a. The valve requires the use of a ladder so that it is accessible.
- b. The valve location makes egress difficult should the valve malfunction.
- c. The valve is required to be locked and is locked in position by the performer.
- d. The verification will result in a radiation exposure of 12 mrem to the verifier.
Answer: d.
- a. Not a permitted waiver for procedure steps
- b. Not a permitted waiver for procedure steps
- c. Not a permitted waiver for procedure steps 90
I RO Written Examination Question No.: 32 l l K/A: Generic 2.1.32 Importance: 3.4 ! l Tier: 3 Group: N/A Cognitive Level: 1 Exam Bank No.: new
Reference:
Objective: COR002-18-02, 05.i 2.2.67, Section 5.19 & 5.20 SKL012-42-18,05 SIL 548 In accordance with 2.2.67, " Reactor Core Isolation Cooling System," which one of the following states the reason for securing the RCIC Gland Seal Vacuum Pump whenever RCIC is NOT available for injection? j
- a. To reduce contamination levels in the North East Quad.
- b. To eliminate addition of oxygen to the Suppression Chamber.
- c. To reduce post-LOCA DC loads within the assumptions of the load calculations.
- d. To eliminate unnecessary wear on the Gland Seal Vacuum Pump mechanical seals.
Answer: b.
- a. Would not cause a reduction in contamination levels; potentially could raise contamination levels is secured shortly after RCIC operation.
- c. This is not the reason for placing the pump in PTL.
- d. The precaution has nothing to do with pump seals. The pump seals are not mechanical, they are packing type.
91
RO Written Examination Question No.: 43 K/A: Generic 2.2.2 Importance: 4.0 Tier: 3 Group: N/A , Cognitive Level: 2 Exam Bank No.: new
Reference:
01-7 Objective: SKL010-10-01, A3 During an ATWS, the Reactor Operator is directed to perform alternate control rod insertion. The Reactor Operator will be performing the actions to insert control rods by resetting RPS and inserting a manual reactor scram. Assume the CRS has NOT suspended any peer check requirements. Which one of the following describes the peer checking requirements to perform this task? Peer checking is required for ...
- a. all steps of the task.
- b. all steps except for panel 9-5 actions only.
1
- c. all steps except forjumper installation only.
- d. the jumper installation, and is waived for all other steps.
l ; i Answer: c. Jumper installation is waived in accordance with 01-7 as it is a back panel action. i l Peer check will be performed by operators in the Control Room for front panel manipulations prior to manipulating controls. This verification will be consistently performed during steady state I manipulations and whenever reasonably possible during abnormal and transient conditions. Inunediate operator actions shall not be delayed to wait for peer check. Peer check can be suspended for specific tasks during transients by the CRS as he deems reasonable and necessary. I l 92
RO Written Examination Question No.: 100 fl K/A: Generic 2.2.13 Importance: 3.6 Tier: 3 Group: N/A ' Cognitive Level: 1 Exam Bank No.: new i
Reference:
0.9, Section 8.4 Objective:SKL010-01-02, A.4 ^ Which one of the following describes when a CAUTION TAG shall be posted on a control switch located in the Control Room? . a. To identify that non-operations personnel can operate the control switch.
- b. To prevent the operation of a component so that maintenance can be performed on the component.
, c. To provide protection to personnel or equipment when a component is undergoing a design modification. $ d. To provide instructions regarding the safe operation of a component as a result of an ! abnormal condition. Answer: d. ,
- a. a blue test tag would be used for this purpose
- b. a danger tag would be used for this purpose i c. caution tags are not used for personnel protection.
a 93
RO Written Examination Question No.: 95 K/A: Generic 2.2.22 Importance: 3.4 Tier: 3 Group: N/A Cognitive Level: 2 Exam Bank No.: new
Reference:
Technical Specifications 3.5.1 Objective: INT 007-05-06,1 & 3 The unit is operating at 100% power when the following Technical Specification conditions are discovered: February 1,1999 at 1200 the "A" RHR pump is declared inoperable. February 3,1999 at 1200 the "C" RHR pump is declared inoperable. February 6,1999 at 0600 the "A" RHR pump is restored to OPERABLE status. February 6,1999 at 0800 the HPCI system is declared inoperable. Apply any extensions that are permitted by Technical Specifications. Assume NO other equipment will be restored to OPERABLE status. Which one of the following describes the time and date when the unit shall be in MODE 37
- a. February 6,1999 at 2100.
- b. February 8,1999 at 2400.
- c. February 9,1999 at 2000.
- d. February 9,1999 at 2400.
Answer: c. When HPCI is declared inoperable, entry into Condition D is required. After 72 hours, entry into Condition G is required. The unit shall be in MODE 3 within the next 12 hours.
- a. Assumes entry into Condition H and LCO 3.0.3 which requires MODE 3 within 13 hours of HPCI becoming inoperable.
- b. Assumes entry into Condition B following the 7 day allowed outage time for the first inoperable pump. Incorrect because a 24 hour completion time extension is permitted.
- d. Assumes entry into Condition B following the 7 day allowed outage time for the first inoperable pump plus an extension of 24 hours for the second pump. This time is greater than that for Condition D and Condition G.
Attachment:
Provide Technical Specification 1.0,3.0,3.5 and 3.6. Do not provide the l Bases. ' 94
RO Written Examination Question No.: 97 K/A: Generic 2.3.1 Importance: 2.6 Tier: 3 Group: N/A Cognitive Level: 1 Exam Bank No.: new
Reference:
9.RADOP.3 Objective: SKL010-01-02, A4 & A5 ITS 5.7.1 & 5.7.2 When comparing 9.RADOP.3," Area Posting and Access Control," requirements for a High Radiation Area to the requirements for a locked High Radiation Area, which one of the following ONLY applies to the locked High Radiation Area?
- a. The Control Room shall be contacted.
- b. A special work permit, SWP, shall be read and understood.
- c. An Administrative Technical Specification LCO shall be entered.
- d. An alarming dose rate meter that continuously integra'es the dose rate shall be in the J possession of the operator.
Answer: a.
- b. A RWP is required for entry into an unlocked and a locked High Radiation Area.
- c. High Radiation Area and Locked High Radiation Area requirements are de:; cubed in f Technical Specification 5.7, however, entry into a Radiation Area or locked High Radiation j area does not constitute violation of Technical Specifications unless the specific requirements are not met. I
- d. This is an optional requirement for entry into both areas.
l 1 95 ;
i RO Written Examination Question No.: 98 K/A: Generic 2.3.4 Importance: 2.5 Tier: 3 Group: N/A Cognitive Level: 2 Exam Bank No.: new
Reference:
GEN 001-01-03 Objective: GEN 001-01-03, Limiting Radiation Dose Obj. D, E 9.ALARA.1, Section 7.2.1 SKL010-01-02, A.4 A station operator has an accumulated TEDE of 1.6 rem for the year as permitted by a previous , extension. Because of dose projections during the assigned outage work, the individual is ! expected to receive an accumulated TEDE of 2.4 rem. In accordance with 9.ALARA.1, " Personnel Dorimetry and Occupational Radiation Exposure Program," which one of the following describes the authorization required for the worker to receive the expected dose?
- a. Plant Manager
- b. Outage Manager
- c. Radiological Manager
- d. Site Vice President - Nuclear Answer: c.
a,b,d Authorizations are required by the Radiological Manager above 2000 mrem. Site V.P. is required above 3000 mrem. 96 l l
RO Written Examination Question No.: 48 K/A: Generic 2.3.10 Importance: 2.9 Tier: 3 Group: N/A Cognitive Level: 1 Exam Bank No.: new
Reference:
Objective:SKL010-01-02, A.4 5.8.3, Section 6.1 INT 008-06-02, 7 Given: Emergency Director declared a Site Area Emergency five (5) minutes ago. TSC is NOT operational. All Area Radiation Monitors (ARMS) on the Reactor Building 903' elevation are alarming and indicate off-scale high. Drywell radiation monitor RMA-RM-40A and RMA-RM-40B indicate 2 x 10 3rem / hour. A station operator is directed to perform manual draining of the SDV. Which one of the following describes the requirements to perform the directed EOP actions outside the Control Room?
- a. The operator shall be accompanied by a Radiological Protection Technician.
- b. The operator may NOT enter Secondary Containment until the TSC is operational.
- c. The operator shall carry a survey instrument capable of monitoring radiation dose rates. ,
- d. The operator is only required to follow standard Radiological Protection practices and procedures.
Answer: a.
- b. This is true if the Drywell radiation monitor RMA-RM-40A or RMA-RM-40B indicate
>10' rem / hour.
- c. This is not an option when the ARM is off-scale high.
- d. This is tme ifNO ARMS are in alarm.
97
RO Written Examination Question No.: 75 l K/A: Generic 2.4.4 Importance: 4.0 Tier: 3 Group: N/A Cognitive Level: 1 Exam Bank No.: new
Reference:
COR001-01-02 Objective: SKL010-01-02, A.4, B.1 2.2.15, Section 2.2 A reactor scram has occurred. Which one of the following describes conditions needed to enter , Emergency Procedure 5.2.5.1, " Loss of All Site AC Power - Station Blackout"? )
- a. Only a loss of the STARTUP transformer.
- b. A loss of the EMERGENCY transformer and the STARTUP transformer.
- c. A loss of the EMERGENCY transformer and a failure of the DGs to start and load.
- d. A loss of the EMERGENCY transformer, STARTUP transformer, and a failure of the DGs to start and load.
Answer: d. I
- a. DG power is still available. 69 KV is still available.
- b. DG poweris still available.
- c. Offsite circuit is still available 4
98
r RO Written Examination Question No.: 91 K/A: Generic 2.4.11 Importance: 3.4 Tier: 3 Group: N/A Cognitive Level: 1 Exam Bank No.: new
Reference:
COR001-14-02 Objective: SKL010-01-02, B.3 2.4.9.3.5, Section 3.0 The unit is operating at 100% reactor power with three (3) Circulating Water Pumps operating. The "A" Circulating Water Pump trips and the TG LOW VACUUM PRE-TRIP annunciator is l received. Recirculation flow is reduced to slow the rate ofloss of vacuum but condenser vacuum continues to slowly degrade. In accordance with Abnormal Procedure 2.4.9.3.5, " Loss of Condenser Vacuum," what IMMEDIATE action is required?
- a. Trip the Main Turbine only.
- b. Start a mechanical vacuum pump,
- c. Start a third circulating water pump.
- d. Manually scram the reactor, then trip the Main Turbine Answer: c.
- a. not an immediate action for the condition provided
- b. not an immediate action for loss of condenser vacuum
- d. not an immediate action for loss of condenser vacuum 99
.4
RO Written Examination Question No.: 87 K/A: Generic 2.4.13 importance: 3.3 Tier: 3 Group: N/A Cognitive Level: 1 Exam Bank No.: new
Reference:
INT 008-06-02, Section II.F.1 Objective: INT 008-06-02, Obj. 9 5.8, section 1.12 While performing Abnormal Procedures 2.4.9.4.1, "RFP Turbine Control Failure," an entry condition into the Emergency Operating Procedures (EOPs) is met. Which one of the following describes the Abnormal Procedure and Emergency Operating Procedure (EOP) use for this condition? Enter all applicable EOPs and execute ...
- a. all flow paths concurrently for the EOPs entered. The Abnormal Procedures are exited when the EOPs are entered.
- b. the flow path for the most degraded plant parameter first. The Abnormal Procedures are exited when the EOPs are entered.
- c. all flow paths concurrently for the EOPs entered. Execute the remaining steps of the Abnormal Procedures when the plant is stable.
- d. the flow path for the most degraded plant parameter first. Execute the other flow paths and the remaining steps of the Abnormal Procedures when the plant is stable.
Answer: c. l When EOPs are entered, all paths are pursued simult.neously. Abnormal procedures are not exited just because EOPs are entered. ! ( I 100}}