ML053330026

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CNS - 06-2005 - Initial Final Admin Outline
ML053330026
Person / Time
Site: Cooper Entergy icon.png
Issue date: 06/27/2005
From:
Nebraska Public Power District (NPPD)
To:
Office of Nuclear Reactor Regulation
References
ES-301, ES-301-1
Download: ML053330026 (36)
v • d • e


Text

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Cooper Nuclear Station Date of Examination: 6/6/2005 Examination Level (circle one): RO Operating Test Number:

Administrative Topic (*) Type Describe Activity to be Performed Code(s)

Conduct of Operations N Verify Valve Position (perform on > 1 mockup valve either at maintenance training building or staged in the training building; at least one valve verified open and close, one valve to be found not in the expected position, SKL034xxxx)

Conduct of Operations D S Perform RO Review of Daily Logs, SKL0345019 (collection of a variety of information from panels and the simulator PCIS using procedure 2.0.2, Operations Logs and Reports)

Equipment Control D Develop, Verify, and Implement Tagouts, SKL0345034 Radiation Control M S Perform Dose Assessment, #2, SKL0345037 (Perform a dose calculation using CNS Dose with data obtained from panel indication and PCIS),

NRC developed Radiation Control D Evaluate an RWP and identify the minimum radiological protection requirements.

Emergency Plan Not Tested NOTE: All item s (5 total) are required for SRO s. RO app licants require only 4 item s unless they are retaking only the administrative topics, when all 5 are required.

  • Type Co des & Criteria: (C) Control Room (D) Direct from Bank (#3 for RO s, # 4 for SROs and RO retakes)

(N) New or (M) Modified from Bank ($1 required)

(P) Previous 2 exams (#1, randomly selected)

(S) Simulator NUREG-1021, Revision 9

ES-301 Control Room / In-Plant Systems Outline Form ES-301-2 Facility: Cooper Nuclear Station Date of Examination: 6/62005 Exam Level (circle one): RO Operating Test Number:

Control Room Systems@ (8 for RO; 7 for SRO-Instant; 2 or 3 for SRO-Update)

System / JPM Title Type Codes (*) Safety Function

a. Recirc Flow Control System, SKL0342123, D S Reactivity Control Respond to Trip of Reactor Recirc Pump
b. Condensate System, SKL0342121, Perform D L S Reactor Inventory Feedwater Startup from 0 to 350 psig Control
c. ADS, SKL03420xx, Perform ADS Manual M A S Reactor Pressure Valve Actuation Surveillance (valve does not Control close when demanded)
d. RHR Shutdown Cooling Mode, SKL034xxxx, L N S Core Heat Removal Shutdown Cooling Cooldown Rate Adjustment
e. Primary Containment & Auxiliaries, M S Containment Integrity SKL0342025, Primary Containment Venting for PCPL
f. Reactor Equipment Cooling System, D S A Plant Service SKL0342144, Separation of REC Critical Loops Systems (REC pump trip)
g. APRM, SKL0342019, Perform APRM Gain D S A Instrumentation Adjustment for Single Loop Operations (potentiometer malfunction)
h. Plant Ventilation Systems, SKL0342075, D S Radiological Release Respond to Sustained Combustion in Offgas System In Plant Systems@ (3 for RO; 3 for SRO-Instant; 3 or 2 for SRO-Update)
i. Uninterruptible Power Supplies, SKL0341095, D A E Electrical Respond to No-Break Power Panel Failure
j. RPS, SKL034xxxx, 5.1ASD Failure to N A C Reactivity Control SCRAM; NRC developed
k. Reactor Core Isolation Cooling System, R N Reactor Inventory SKL034xxxx, Manual Start of the RCIC Turbine Control per 5.3ALT Strategy NUREG-1021, Revision 9

ES-301 Control Room / In-Plant Systems Outline Form ES-301-2

(@ ) All Control Room and In-Plant systems m ust be different and serve different safety functions; in-plant systems and functions may overlap those tested in the Control Room

  • Type Codes Criteria for RO / SRO-Instant / SRO-Upgrade (A) A lternate Pa th 4-6 / 4-6 / 2-3 (C) Control Room (D) Direct from Bank #9/ #8 / #4 (E) Emergency or Abnormal in plant $1 / $1 / $1 (L) Low Power $1 / $1 / $1 (N) New or (M) Modified from bank including 1(A) $2 / $2 / $2 (P) Previous 2 exams #3/ #3/ #2 (randomly selected)

(R) RC A Entry $1 / $1 / $1 (S) Simulator NUREG-1021, Revision 9

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Cooper Nuclear Station Date of Examination: 6/6/2005 Examination Level (circle one): SRO Operating Test Number:

Administrative Topic (*) Type Describe Activity to be Performed Code(s)

Conduct of Operations D S Reportability Determination per procedure 2.0.5 (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, given an event that should have resulted in ECCS discharge into the Reactor Coolant System as a result of a valid signal, SKL0343028, Reportable Occurrence to the NRC, #3)

Conduct of Operations N Shift Staffing Determination per procedure 2.0.3 section 10 (given a mode and a partial crew complement, determine what additional crew positions are required); NRC developed Equipment Control N Risk Assessment and Mock Safety Function Determination using procedure 0.26 per procedure 0.49, step 3.5 (evaluation of the schedule, including risk assessment, during periods outside normal office hours, for impact of emergent equipment problems including missed TS/TRM surveillances on scheduled activities & ensure mock safety function determination is performed to assess the impact of missed TS/TRM surveillances on safety-related equipment)

Radiation Control M Review and Approve Liquid Radioactive Waste Discharge per procedure 8.8.11, Attachment 1 (complete sections 1-3, provide information for SM to complete section 4, faulted - SM should not approve, SKL03450xx, Approve Radioactive Discharge Release Permit)

Emergency Plan M Protective Action Recommendation determination per procedure 5.7.20 and complete the appropriate section(s) of the offsite notification form (CNS Dose is not available; provide data to use Attachments 1 and 2, SKL03430xx, PAR Tabletop)

NOTE: All item s (5 total) are required for SRO s. RO app licants require only 4 item s unless they are retaking only the administrative topics, when all 5 are required.

NUREG-1021, Revision 9

ES-301 Administrative Topics Outline Form ES-301-1

  • Type Co des & Criteria: (C) Control Room (D) Direct from Bank (#3 for RO s, # 4 for SROs and RO retakes)

(N) New or (M) Modified from Bank ($1 required)

(P) Previous 2 exams (#1, randomly selected)

(S) Simulator NUREG-1021, Revision 9

ES-301 Control Room / In-Plant Systems Outline Form ES-301-2 Facility: Cooper Nuclear Station Date of Examination: 6/62005 Exam Level (circle one): SRO-U Operating Test Number:

Control Room Systems@ (8 for RO; 7 for SRO-Instant; 2 or 3 for SRO-Update)

System / JPM Title Type Codes (*) Safety Function

a. Recirc Flow Control System, SKL0342123, Not Tested Respond to Trip of Reactor Recirc Pump
b. Condensate System, SKL0342121, Perform Not Tested Feedwater Startup from 0 to 350 psig
c. ADS, SKL03420xx, Perform ADS Manual M A Reactor Pressure Valve Actuation Surveillance (valve does not Control close when demanded)
d. RHR Shutdown Cooling Mode, SKL034xxxx, L N Core Heat Removal Shutdown Cooling Cooldown Rate Adjustment
e. Primary Containment & Auxiliaries, Not Tested SKL0342025, Primary Containment Venting for PCPL
f. Reactor Equipment Cooling System, D S A Plant Service SKL0342144, Separation of REC Critical Loops Systems (REC pump trip)
g. APRM, SKL0342019, Perform APRM Gain Not Tested Adjustment for Single Loop Operations (potentiometer malfunction)
h. Plant Ventilation Systems, SKL0342075, Not Tested Respond to Sustained Combustion in Offgas System In Plant Systems@ (3 for RO; 3 for SRO-Instant; 3 or 2 for SRO-Update)
i. Uninterruptible Power Supplies, SKL0341095, D A E Electrical Respond to No-Break Power Panel Failure
j. RPS, SKL034xxxx, 5.1ASD Failure to SCRAM Not Tested
k. Reactor Core Isolation Cooling System, R N Reactor Inventory SKL034xxxx, Manual Start of the RCIC Turbine Control per 5.3ALT Strategy NUREG-1021, Revision 9

ES-301 Control Room / In-Plant Systems Outline Form ES-301-2

(@ ) All Control Room and In-Plant systems m ust be different and serve different safety functions; in-plant systems and functions may overlap those tested in the Control Room

  • Type Codes Criteria for RO / SRO-Instant / SRO-Upgrade (A) A lternate Pa th 4-6 / 4-6 / 2-3 (C) Control Room (D) Direct from Bank #9/ #8 / #4 (E) Emergency or Abnormal in plant $1 / $1 / $1 (L) Low Power $1 / $1 / $1 (N) New or (M) Modified from bank including 1(A) $2 / $2 / $2 (P) Previous 2 exams #3/ #3/ #2 (randomly selected)

(R) RC A Entry $1 / $1 / $1 (S) Simulator NUREG-1021, Revision 9

ES-301 Control Room / In-Plant Systems Outline Form ES-301-2 Facility: Cooper Nuclear Station Date of Examination: 6/62005 Exam Level (circle one): SRO-U Operating Test Number:

Control Room Systems@ (8 for RO; 7 for SRO-Instant; 2 or 3 for SRO-Update)

System / JPM Title Type Codes (*) Safety Function

a. Recirc Flow Control System, SKL0342123, Not Tested Respond to Trip of Reactor Recirc Pump
b. Condensate System, SKL0342121, Perform Not Tested Feedwater Startup from 0 to 350 psig
c. ADS, SKL03420xx, Perform ADS Manual M A Reactor Pressure Valve Actuation Surveillance (valve does not Control close when demanded)
d. RHR Shutdown Cooling Mode, SKL034xxxx, L N Core Heat Removal Shutdown Cooling Cooldown Rate Adjustment
e. Primary Containment & Auxiliaries, Not Tested SKL0342025, Primary Containment Venting for PCPL
f. Reactor Equipment Cooling System, D S A Plant Service SKL0342144, Separation of REC Critical Loops Systems (REC pump trip)
g. APRM, SKL0342019, Perform APRM Gain Not Tested Adjustment for Single Loop Operations (potentiometer malfunction)
h. Plant Ventilation Systems, SKL0342075, Not Tested Respond to Sustained Combustion in Offgas System In Plant Systems@ (3 for RO; 3 for SRO-Instant; 3 or 2 for SRO-Update)
i. Uninterruptible Power Supplies, SKL0341095, D A E Electrical Respond to No-Break Power Panel Failure
j. RPS, SKL034xxxx, 5.1ASD Failure to SCRAM Not Tested
k. Reactor Core Isolation Cooling System, R N Reactor Inventory SKL034xxxx, Manual Start of the RCIC Turbine Control per 5.3ALT Strategy NUREG-1021, Revision 9

ES-301 Control Room / In-Plant Systems Outline Form ES-301-2

(@ ) All Control Room and In-Plant systems m ust be different and serve different safety functions; in-plant systems and functions may overlap those tested in the Control Room

  • Type Codes Criteria for RO / SRO-Instant / SRO-Upgrade (A) A lternate Pa th 4-6 / 4-6 / 2-3 (C) Control Room (D) Direct from Bank #9/ #8 / #4 (E) Emergency or Abnormal in plant $1 / $1 / $1 (L) Low Power $1 / $1 / $1 (N) New or (M) Modified from bank including 1(A) $2 / $2 / $2 (P) Previous 2 exams #3/ #3/ #2 (randomly selected)

(R) RC A Entry $1 / $1 / $1 (S) Simulator NUREG-1021, Revision 9

ES-401 BWR Examination Outline Form ES-401-1 Facility: Cooper Nuclear Station Date of Exam: June 13, 2005 Tier Group RO K/A Catego ry Points SR O Only K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G* Total A2 G* Total

1. 1 2 3 3 4 3 5 20 4 3 7 N/A N/A 2 1 3 0 1 1 1 7 1 2 3 Tota ls 3 6 3 5 4 6 27 5 5 10
2. 1 4 2 2 2 2 2 4 2 4 1 1 26 2 3 5 2 1 0 1 2 0 2 1 1 1 1 2 12 1 2 3 Tota ls 5 2 3 4 2 4 5 3 5 2 3 38 3 5 8
3. Generic Knowledge and Abilities Categories 1 3 1 1 10 7 2 3 2 2 3 2 3 2 4 2 4 2 Note: 1. Ensure that at least 2 topics from every applicable KA category are sampled within each tier of the RO and SRO-only outlines (i.e. except for Category 1 in Tier 3 of the SRO-only outline, the Tier Totals in each KA category shall not be less than 2).
2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/- 1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points.
3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems that are not included on the outline should be added. Refer to ES-401, Attachment 2, for guidance regarding the elimination of inappropriate KA statements.
4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.
5. Absent a plant-specific priority only those KAs having an importance rating of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.
6. Select SRO topics for Tiers 1 and 2 from the shaded systems and KA categories.
7. (*) The generic (G) KAs in Tiers 1 and 2 shall be selected from Section 2 of KA Catalog but the topics must be relevant to the applicable evolution or system.
8. On the following pages enter the KA numbers, a brief description of each topic, the topics importance ratings for the applicable license level and the point totals for each system and category. Enter the group and tier totals for each category in the table above. Use duplicate pages for RO and SRO-only exams.
9. For Tier 3 select topics from Section 2 of the KA Catalog and enter the KA numbers, descriptions, IRs and point totals on Form ES 401-3. Limit SRO selections to KAs that are linked to 10 CFR 55.43.

ES-401 , NUREG 1021 Revision 9

ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1 / Group 1 (RO)

E/APE#/Name/Safety Function K K K A A G KA T opic(s) IR #

1 2 3 1 2 295001 Partial or Complete Loss of Forced 1 1 Relation between loss of forced core flow 3.3 AK2.04 Core Flow Circulation circulation and the reactor or turbine pressure regulating system Reasons for reduced loop operating 3.2 AK3.05 requirements as applies to loss of forced core flow circulation 2 9 50 0 3 P a rt ia l o r C o m p le te L os s o f A C 1 Operational implications of failsafe 2.6 AK1.05 component design applied to partial or complete loss of AC power 2 9 5 0 0 4 P a rti a l or Total Loss of DC Pwr 1 Determine or interpret the cause of partial 3.2 AA2.01 or complete loss of DC power 29 50 05 Ma in T urb ine Ge ne rato r Trip 1 Monitor or operate RPS as applies to main 3.6 AA1.02 turbine generator trip 2 9 50 0 6 S C R A M 1 Ability to prioritize and interpret the 3.3 2.4.45 significance of each annunciator and alarm 295016 Control Room Abandonment 1 Ability to determine or interpret reactor 4.1 AA2.01 power as applies to CR abandonment 2 9 50 1 8 P a rt ia l o r T o ta l L o ss of C C W 1 Reasons for reactor power reduction as 3.3 AK3.02 applied to partial or total loss of CCW 29 50 19 Pa rtial o r To tal Lo ss o f Inst A ir 1 Ability to evaluate plant performance and 3.7 2.1.07 make operational judgements based on operating characteristics, reactor behavior, and instrument interpretation 295021 Loss of Shutdown Cooling 1 Knowledge of the purpose and function of 3.9 2.1.28 major system components and controls.

295023 Refueling Accident 1 Monitor or operate radiation monitoring 3.4 AA1.04 equipment as applies to refueling accidents 295 024 High Dryw ell Pre ssu re 1 Operate or monitor RPS as applies to high 3.9 EA1.05 drywell pressure 295 025 High Re acto r Pres sure 1 Relationship with Safety Relief Valves 4.1 EK2.05 295026 Suppression Pool High W ater 1 Relationship between suppression pool 3.9 EK2.01 Te m pera ture cooling and high water temperature 295 027 High Co ntainm ent T em pera ture Not Applicable to Cooper 295 028 High Dryw ell Te m pera ture 1 Verify alarm setpoints and operate controls 3.3 2.4.50 as identified in the alarm response manual 295030 Low Suppression Pool Water Level 1 Monitor or operate RCIC as applies 3.4 EA1.02 Continued on the next page...

ES-401 , NUREG 1021 Revision 9

ES-401 BWR Examination Outline Form ES-401-1 Continued from the previous page...

Emergency and Abnormal Plant Evolutions - Tier 1 / Group 1 (RO)

E/APE#/Name/Safety Function K K K A A G KA T opic(s) IR #

1 2 3 1 2 295031 Reactor Low Water Level 1 Ability to located CR switches and 4.2 2.1.31 indications and determine they reflect the desired plant line up 295037 SCRAM C ondition Present and 1 Determine or interpret reactor water level 4.1 EA2.02 P o w e r A bo ve A PRM Downscale or Unknown as applied to SCRAM w/ATWS condition 295 038 High Offsite Re leas e R ate 1 Knowledge of the reasons for control room 3.7 EK3.03 ventilation isolation during conditions of high offsite release rate.

600 000 Plan t Fire O n S ite 1 Operational implications of fire fighting as 2.9 AK1.02 applies to plant fire on site KA Category Totals 2 3 3 4 3 5 Group Point Total: 20 ES-401 , NUREG 1021 Revision 9

ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1 / Group 2 (RO)

E/APE#/Name/Safety Function K K K A A G KA T opic(s) IR #

1 2 3 1 2 295002 Loss of Main Condenser Vacuum 295007 High Reactor Pressure 1 Determine or interpret reactor water level 3.7 AA2.03 as applies to high reactor pressure 295008 High Reactor Water Level 1 Monitor or operate HPCI as applies to high 3.1 AA1.04 reactor water level 295009 Low Reactor Water Level 1 Relations between reactor water level 3.9 AK2.02 control and low reactor water level 295010 High Drywell Pressure 295011 High Containment Temp. Not Applicable to Cooper 295012 High Drywell Temperature 295013 High Suppression Pool Temp. 1 Relation of suppression pool cooling 3.6 AK2.01 295014 Inadvertent Reactivity Addition 1 Ability to perform pre-startup procedures 3.7 2.2.01 for the facility including operating those controls associated with plant equipment that could affect reactivity 295015 Incomplete SCRAM 1 Knowledge of the operational implications 3.8 AK1.03 of the following concepts as they apply to incomplete scram: Reactivity affects 295017 High Offsite Release Rate Not Applicable to the RO Position 295020 Inadvertent Containment Isol 295022 Loss of CRD Pumps 295029 High Suppression Pool Level 295032 High Secondary Containment 1 Relation between CNMT area temperature 3.6 EK2.07 Area Temperature and leak detection system concepts 295033 High Secondary Containment Area Radiation Levels 295034 Secondary Containment Ventilation High Radiation 295035 Secondary Containment High Differential Pressure 295036 Secondary Containment High Sump / Area Water Level 500000 High CTMT Hydrogen Conc.

KA Category Point Total 1 3 0 1 1 1 Group Point T ota l: 7 ES-401 , NUREG 1021 Revision 9

ES-401 BWR Examination Outline Form ES-401-1 Plant Systems- Tier 2 / Group 1 (RO)

System # / Name K K K K K K A A A A G KA Topics IR #

1 2 3 4 5 6 1 2 3 4 203000 RHR LPCI: Injection Mode 1 Effect of malfunction on water lvl. 4.3 K3.01 205000 Shutdown Cooling 1 Design & Interlocks providing for 3.6 K4.05 reactor cool down rate 2 0 60 0 0 H P C I 1 Connections with Keep Fill systm 4.0 K1.09 207000 Isolation Condenser Not applicable to Cooper 2 0 90 0 1 L P C S 1 Predict/monitor System Lineup 3.3 A1.08 2 0 90 0 2 H P C S Not applicable to Cooper 211000 SLC 1 1 Relation with plant air systems 2.5 K1.03 Ability to apply Tech Specs 2.9 2.1.12 2 1 20 0 0 R P S 1 Knowledge of electrical power 3.2 K2.01 systems to RPS M/G sets 1 RPS Bus Voltage 2.8 A1.04 2 1 50 0 3 IR M 1 Operational implications of 3.0 K5.03 changing detector positions 215 004 So urce Ra nge Mon itors 1 Electrical power supplies to SRM 2.6 K2.01 channels or detectors 2 1 50 0 5 A R P M / L P R M 1 Monitor automatic operations of 3.3 A3.06 max. disagreement of flow comparator channels 21 70 00 RC IC 1 Predict/monitor Supp. Pool Level 3.3 A1.07 2 1 80 0 0 A D S 1 1 Effect of malfunction on ADS 3.8 K6.04 valve air supply Predict impact of small break LOCA on ADS and mitigate

[Deleted]

Ability to monitor automatic 4.2 A3.01 operations of the ADS including ADS valve operation 223002 PCIS / Nuclear Steam 1 Predict / Monitor changes assoc. 2.6 A1.04 Su pply S huto ff with individual relay status ES-401 , NUREG 1021 Revision 9

ES-401 BWR Examination Outline Form ES-401-1 239002 Safety Relief Valves 1 Connections with nuclear boiler 3.5 K1.03 instrumentation system Monitor SRV and acoustical mnt.

[Deleted]

1 Lights and Alarms 3.6 A3.08 259002 Reactor Water Level Control 1 Monitor auto operations and 3.2 A3.03 changes in main steam flow 26 10 00 Sta nd by G as Tre atm en t Sys 1 Moritor or operate fans from the 3.0 A4.03 Control Room 262001 AC Electrical Distribution 1 Physical connections with offsite 3.4 K1.03 power Plant Systems- Tier 2 / Group 1 (RO)

System # / Name K K K K K K A A A A G KA Topics IR #

1 2 3 4 5 6 1 2 3 4 2 6 20 0 2 U P S (A C - DC ) 1 Design & interlocks which 3.1 K4.01 provide for transfer from preferred to alternate power Predict/monitor changes assoc. 2.5 A1.02 with motor generator outputs

[Delete]

263000 DC Electrical Distribution 1 Effect of malfunction on systems 3.4 K3.03 with DC components 26 40 00 ED Gs 1 1 Operations implications of 3.4 K5.05 paralleling AC power sources Predict consequences of 2.9 A2.04 over/under-excited operation and mitigate 30 00 00 Instru m en t Air 1 Predict effect of air dryer and 2.9 A2.01 filter malfunctions and mitigate 4 0 00 0 0 C C W 1 Effect of loss or malfunction of 3.0 K6.05 pumps will have on CCW KA Category Point Totals: 4 2 2 2 2 2 4 3 3 1 1 Group Point Total: 26 ES-401 , NUREG 1021 Revision 9

ES-401 BWR Examination Outline Form ES-401-1 Plant Systems- Tier 2 / Group 2 (RO)

System # / Name K K K K K K A A A A G KA Topics IR #

1 2 3 4 5 6 1 2 3 4 201001 CRD Hydraulic 201002 RMCS 201003 Control Rod and Drive Mechanism 201004 RSCS Not Applicable to Cooper 201005 RCIS Not Applicable to Cooper 201006 RWM 202001 Recirculation 202002 Recirc Flow Control 204000 RWCU 214000 RPIS 1 Monitor or operate control rod 3.8 A4.02 position from the control room 215001 Traversing In-Core Probe 215002 Rod Block Monitor 216000 Nuclear Boiler Instrum.

219000 RHR LPCI: Torus / 1 Ability to predict and/or monitor 3.2 A1.04 Pool Cooling Mode changes in parameters ,

including suppression pool level.

223001 Primary CNMT & Aux. 1 Monitor automatic operation and 4.3 A3.05 Drywell pressure 226001 RHR LPCI: 1 Knowledge of the bases for 3.0 2.4.22 Containment Spray Mode prioritizing safety functions 23000 RHR LPCI: Torus / Pool Spray Mode 233000 Fuel Pool Cooling & 1 Knowledge of physical 2.9 K1.02 Cleanup connections and cause/effect with RHR 234000 Fuel Handling Equip.

239001 Main & Reheat Steam 1 Design & interlocks pertaining to 3.3 K4.09 equalization of MSIV pressure prior to opening 239003 MSIV Leakage Control Not Applicable to Cooper Continued on Next page...

ES-401 , NUREG 1021 Revision 9

ES-401 BWR Examination Outline Form ES-401-1 Continued from previous page...

Plant Systems- Tier 2 / Group 2 (RO)

System # / Name K K K K K K A A A A G KA Topics IR #

1 2 3 4 5 6 1 2 3 4 241000 Reactor / Turbine Pressure Regulator 245000 Main Turbine Generator

& Auxiliaries 256000 Reactor Condensate Effect of main steam system K6.10 loss or malfunction on condensate system [Delete]

1 Effect of main steam system 3.3 K6.06 loss or malfunction on the reactor feedwater system 259001 Reactor Feedwater 268000 Radwaste 271000 Offgas 1 Monitor auto operation of 2.9 A3.05 indicating lights and alarms 272000 Radiation Monitoring 1 Operator responsibilities during 3.0 2.1.2 all modes of operation 286000 Fire Protection 1 Effect of loss or malfunction on 3.2 K3.02 personnel protection 288000 Plant Ventilation 290001 Secondary Containment 290003 Control Room HVAC 1 Predict impact of initiation or 3.1 A2.01 reconfiguration and mitigate abnormal conditions 290002 Reactor Vessel 1 Design & interlocks provide for 3.3 K4.05 Internals natural circulation KA Category Point Totals 1 0 1 2 0 2 1 1 1 1 2 Group Point Total 12 ES-401 , NUREG 1021 Revision 9

ES-401 BWR Examination Outline Form ES-401-1 Average IR for Tier 1, Group 1 3.52 Tier Total 20 Average IR for Tier 1, Group 2 3.63 Tier Total 07 Average IR for Tier 2, Group 1 3.12 Tier Total 26 Average IR for Tier 2, Group 2 2.91 Tier Total 12 Average IR for Tier 3 3.06 Tier Total 10 Average IR for RO Exam 3.23 Exam Total 75 ES-401 , NUREG 1021 Revision 9

ES-401 BWR Examination Outline Form ES-401-3 Generic Knowledge and Abilities Outline (Tier 3)

Category KA # Topic RO SRO Only IR # IR #

1. 2..1.10 Knowledge of conditions and limitations in the facility 2.7 1 Conduct of license Operations 2..1.28 Knowledge of the purpose and function of major system 3.2 1 components and controls.

2..1.33 Ability to recognize indications for system operating 2.7 1 parameters which are entry conditions for technical specifications.

Subtotal 3

2. 2..2.22 Knowledge of limiting conditions for operations and safety 3.4 1 Equipment limits Control 2..2.30 Knowledge of RO duties in the CR during fuel handling 3.5 1 2..2.34 Knowledge of the process for determining the internal and 2.8 1 external effects on core reactivity Subtotal 3
3. Radiation 2..3.01 Knowledge of 10 CFR 20 and related facility radiation 2..6 1 Control control requirements 2..3.10 Ability to perform procedures to reduce excessive levels of 2..9 1 radiation and guard against personnel exposure Subtotal 2
4. Emergency 2..4.46 Ability to verify that alarms are consistent with plant 3.5 1 Procedures conditions and Plan 2..4.49 Ability to perform without reference to procedures those 3.3 1 actions which require immediate operations of system components and controls Subtotal 2 Tier 3 Point Total 10 ES-401 , NUREG 1021 Revision 9

ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1 / Group 1 (SRO)

E/APE#/Name/Safety Function A G KA T opic(s) IR #

2 295 001 Pa rtial or Co m plete L oss of Forc ed C ore 1 Knowledge of limiting conditions for operations and 4.1 2.2.22 Flow Circulation safety limits 2 9 50 0 3 P a rt ia l o r C o m p le te L os s o f A C 2 9 5 0 0 4 P a rti a l or Total Loss of DC Pwr 29 50 05 Ma in T urb ine Ge ne rato r Trip 1 Ability to determine or interpret feedwater 2.7 AA2.06 temperature as applied to a main turbine generator trip 2 9 50 0 6 S C R A M 295016 Control Room Abandonment 2 9 50 1 8 P a rt ia l o r T o ta l L o ss of C C W 29 50 19 Pa rtial o r To tal Lo ss o f Inst A ir 1 Knowledge of annunciators, alarms, and indications, 3.4 2.4.31 and use of the response instructions 295021 Loss of Shutdown Cooling 295023 Refueling Accident 295 024 High Dryw ell Pre ssu re Ability to determine or interpret suppression pool level as applied to high drywell pressure [Deleted]

295 025 High Re acto r Pres sure 1 Ability to determine or interpret suppression pool 4.1 EA2.03 temperature as applied to high reactor pressure 295 026 Su ppre ssion Po ol Hig h W ater T em pera ture 295 027 High Co ntainm ent T em pera ture 1 Ability to recognize indications for system operating 4.0 2.1.33 parameters which are entry-level conditions for technical specifications 295 028 High Dryw ell Te m pera ture 1 Ability to determine and/or interpret the following as 3.9 EA2.03 they apply to high drywell temperature: Reactor Water Level.

295030 Low Suppression Pool Water Level 1 Ability to determine or interpret drywell / 3.7 EA2.04 suppression pool differential pressure as applied to low suppression pool water level 295031 Reactor Low Water Level 295037 SCRAM C ondition Present and Power A b o ve A P RM D ownscale or Unknown 295 038 High Offsite Re leas e R ate 600 000 Plan t Fire O n S ite KA Category Totals 4 3 Group Point Total: 7 ES-401 , NUREG 1021 Revision 9

ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1 / Group 2 (SRO)

E/APE#/Name/Safety Function A G KA T opic(s) IR #

2 295002 Loss of Main Condenser Vacuum 295007 High Reactor Pressure 1 Knowledge of surveillance procedures 3.4 2.2.12 295008 High Reactor Water Level 295009 Low Reactor Water Level 295010 High Drywell Pressure 295011 High Containment Temp. Not Applicable to Cooper 295012 High Drywell Temperature 295013 High Suppression Pool Temp.

295014 Inadvertent Reactivity Addition 295015 Incomplete SCRAM 295017 High Offsite Release Rate 1 Knowledge of symptom-based EOP mitigation 4.0 2.4.6 strategies 295020 Inadvertent Containment Isol 295022 Loss of CRD Pumps 295029 High Suppression Pool Level 295032 High Secondary Containment Area Temperature 295033 High Secondary Containment Area Radiation Levels 295034 Secondary Containment Ventilation 1 Ability to determine or interpret ventilation radiation 4.2 EA2.01 High Radiation levels 295035 Secondary Containment High Differential Pressure 295036 Secondary Containment High Sump

/ Area Water Level 500000 High CTMT Hydrogen Conc.

KA Category Point Total 1 2 Group Point T ota l: 3 ES-401 , NUREG 1021 Revision 9

ES-401 BWR Examination Outline Form ES-401-1 Plant Systems- Tier 2 / Group 1 (SRO)

System # / Name A G KA Topics IR #

2 203000 RHR LPCI: Injection Mode 1 Knowledge of the process for managing maintenance 3.5 2.2.18 activities during shutdown operations 205000 Shutdown Cooling 2 0 60 0 0 H P C I 1 Ability to execute procedure steps 4.2 2.1.20 207000 Isolation Condenser Not Applicable to Cooper 2 0 90 0 1 L P C S 2 0 90 0 2 H P C S Not Applicable to Cooper 211000 SLC 2 1 20 0 0 R P S 2 1 50 0 3 IR M 1 Ability to determine Mode of Operation 3.3 2.1.22 215004 Source Range Monitor 2 1 50 0 5 A R P M / L P R M 21 70 00 RC IC 2 1 80 0 0 A D S 223 002 PC IS / Nu clea r Stea m Su pply S huto ff 23 90 02 SR Vs 259002 Reactor Water Level Control 261000 Standby Gas Treatment System 262001 AC Electrical Distribution 1 Ability to predict the impact of opening a disconnect 3.6 A2.08 under load and....correct, control or mitigate the consequences 2 6 20 0 2 U P S (A C - DC )

263000 DC Electrical Distribution 264 000 Em erge ncy D iese l Ge nera tors 1 Ability to predict the impacts of synchronization of the 3.6 A2.05 emergency generator with other electrical supplies, and....correct, control or mitigate the consequences 30 00 00 Instru m en t Air 4 0 00 0 0 C C W KA Category Point Totals: 2 3 Group Point Total: 5 ES-401 , NUREG 1021 Revision 9

ES-401 BWR Examination Outline Form ES-401-1 Plant Systems- Tier 2 / Group 2 (SRO)

System # / Name A G KA Topics IR #

2 201001 CRD Hydraulic 201002 RMCS 201003 Control Rod and Drive Mechanism Ability to perform specific system and integrated 4.0 2.1.23 plant procedures during different modes of plant operations [Deleted]

201004 RSCS Not Applicable to Cooper 201005 RCIS Not Applicable to Cooper 201006 RWM 202001 Recirculation 1 Ability to predict the impacts of recirculation scoop 3.4 A2.09 tube lockup on the recirculation system and on the basis of the prediction use procedures to correct, control, or mitigate the consequences of the abnormal conditions or operations.

202002 Recirc Flow Control 204000 RWCU 214000 RPIS 215001 Traversing In-Core Probe 215002 Rod Block Monitor 216000 Nuclear Boiler Instrumentation 219000 RHR LPCI: Torus / Pool Cooling Mode 223001 Primary CNMT & Aux.

226001 RHR LPCI: Containment Spray Mode 23000 RHR LPCI: Torus / Pool Spray Mode 233000 Fuel Pool Cooling & Cleanup 1 Ability to obtain and interpret station reference 3.1 2.1.25 materials...which contain performance data 234000 Fuel Handling Equip. 1 Ability to track limiting conditions for operations 3.8 2.2.23 239001 Main & Reheat Steam 239003 MSIV Leakage Control 241000 Reactor / Turbine Pressure Regulator 245000 Main Turbine Generator &

Auxiliaries ES-401 , NUREG 1021 Revision 9

ES-401 BWR Examination Outline Form ES-401-1 256000 Reactor Condensate

...continued on Next page...

...continued from Previous page...

Plant Systems- Tier 2 / Group 2 (RO / SRO)

System # / Name A G KA Topics IR #

2 259001 Reactor Feedwater 268000 Radwaste 271000 Offgas 272000 Radiation Monitoring 286000 Fire Protection 288000 Plant Ventilation 290001 Secondary Containment 290003 Control Room HVAC 290002 Reactor Vessel Internals Deleted KA Category Point Totals 1 2 Group Point Total 3 ES-401 , NUREG 1021 Revision 9

ES-401 BWR Examination Outline Form ES-401-1 Average IR for Tier 1, Group 1 3.69 Tier Total 7 Average IR for Tier 1, Group 2 3.87 Tier Total 3 Average IR for Tier 2, Group 1 3.64 Tier Total 5 Average IR for Tier 2, Group 2 3.76 Tier Total 3 Average IR for Tier 3 3.43 Tier Total 7 Average IR for SRO Exam 3.64 Exam Total 25 ES-401 , NUREG 1021 Revision 9

ES-401 BWR Examination Outline Form ES-401-3 Generic Knowledge and Abilities Outline (Tier 3)

Category KA # Topic RO SRO Only IR # IR #

1. 2.1.14 Knowledge of system status criteria which require 3.3 1 Conduct of notification of plant personnel Operations Subtotal 1
2. 2.2.14 Knowledge of the process for making configuration 3.0 1 Equipment changes Control 2.2.26 Knowledge of refueling administrative requirements 3.7 1 Subtotal 2
3. Radiation 2.3.3 Knowledge of SRO responsibilities for auxiliary 2.9 1 Control systems outside the control room (waste disposal and handling systems) 2.3.4 Knowledge of radiation exposure limits and 3.1 1 contamination control, including permissible levels in excess of those authorized Subtotal 2
4. 2.4.22 Knowledge of the bases for prioritizing safety 4.0 1 Emergency functions during abnormal or emergency operations Procedures and Plan 2.4.44 Knowledge of emergency plan protective action 4.0 1 recommendations Subtotal 2 Tier 3 Point Total 7 ES-401 , NUREG 1021 Revision 9

Facility: COOPER Scenario No.: 4 Op-Test No.:1 Examiners: _Paul Gage_______________ Operators: __________________________

_Steve Garchow___________ __________________________

_Kelly Clayton____________ __________________________

Plant Status: The plant is operating at 75% power near the end of the current fuel cycle when the crew takes the shift. The plant is in a normal configuration with the exception of Rx Feed Pump Lube Oil Pump A1 and Sump Pump D1 are out of service. Southeast Nebraska is in a severe thunderstorm warning, which includes intense electrical storm activity Turnover: The plant is operating at 75% power near the end of the current fuel cycle when the crew takes the shift. The plant is in a normal configuration with the exception of Rx Feed Pump Lube Oil Pump A1 and Sump Pump D1 are out of service. Southeast Nebraska is in a severe thunderstorm warning, which includes intense electrical storm activity.

Scenario: The plant is operating at 75% power near the end of the current fuel cycle when the crew takes the shift. The plant is in a normal configuration with the exception of Rx Feed Pump Lube Oil Pump A1 and Sump Pump D1 are out of service.

Southeast Nebraska is in a severe thunderstorm warning, which includes intense electrical storm activity. A surveillance testing the operability of main steam isolation valve 80A fails resulting in one MSL being isolated. A lightning strike on the grid results in the loss of the 69kV OPPD Nebraska City Line and requires performing a tech spec surveillance. When the Technical Specification assessment is complete, a MSL radiation monitor fails upscale. When the assessment of this is complete, APRM Channel B fails INOP. When the actions for the APRM are complete, a loss of off-site power, reactor scram, and a steam leak into primary containment occur. Later in the scenario EDG-2 fails. The scenario ends when RPV water level is being restored to the normal band, drywell pressure is being controlled and classifications have been made.

Event Malf. No. Event Event Description No. Type*

Trigger 1 N (SRO, Power reduction to < 70% using control rods and recirc flow 1.

RO)

Trigger 2 C, T Surveillance 6.MS.201 Section 5 (Failure of MSIV MOV-80A to open) 2.

(BOP,

Facility: COOPER Scenario No.: 2 Op-Test No.:1 Examiners: _Paul Gage_______________ Operators: __________________________

_Steve Garchow___________ __________________________

_Kelly Clayton____________ __________________________

Plant Status: The plant is operating at 60% power with instructions to continue the power ascension to 100%. The A1 Reactor Feedpump LO Pump and D1 Sump Pump are both tagged out due to a motor failure. The Sentinal Status is Green.

Turnover: The plant is operating at 60% power with instructions to continue the power ascension to 100%. The A1 Reactor Feedpump LO Pump and D1 Sump Pump are both tagged out due to motor failures. The Sentinal Status is Green. Reactor engineering is working on the computer program used to calculate AGAFs. Assume all AGAFs are in specification unless otherwise notified by the STA.

Scenario: The plant is operating at 60% power with instructions to continue the power ascension to 100%. The crew will raise power greater than 70% before the first event is called in. The A1 Feed Pump Lube Oil Pump and Sump Pump D1 are tagged out for maintenance.

Following the power ascension, an accumulator fault due to low N2 pressure is received. Following the Tech Spec assessments, an inadvertent initiation of HPCI occurs. The crew should respond per 2.4CSCS and the Technical Specifications.

After the Tech Spec assessment is complete, a tube rupture occurs in feedwater heater A5. This will require a diagnosis since the alarm clears. Eventually the crew will have to commence a plant shutdown. Once the shutdown is underway, an unisolable steam line leak will develop on the HPCI steam line. The automatic isolation for the steam supply valves will not function and the valves cannot be closed from the control room. The crew is expected to take action EOP-05, Secondary Containment Control and scram the plant before one area reaches a Maximum Safe Operating Temperature (MSOT).

Due to a hydraulic lock, many control rods will fail to insert. The crew should respond to the ATWS per EOP-06A, 7A and 5.8.3. Power level should be ~ 10%

after the Recirculation pump trip, so RPV water level will have to be lowered.

Control rods can be inserted via RMCS.

The secondary containment temperatures will continue to rise, resulting in MSOT in 2 areas. The crew is expected to take action iaw EOP-6B and perform an Emergency Depressurization. After the Emergency Depressurization is complete and RPV water level is being controlled, the control rods will insert the next time the scram is reset and scrammed again.

The scenario will terminate when the RPV has been depressurized, control rods have been inserted. and RPV water level has been restored to +15 to +40.

Facility: COOPER Scenario No.: 3 Op-Test No.:1 Examiners: _Paul Gage_______________ Operators: __________________________

_Steve Garchow___________ __________________________

_Kelly Clayton____________ __________________________

Plant Status: The plant is operating at 100% power near the end of the current fuel cycle when the crew takes the shift. The plant is in a normal configuration with Rx Feed Pump Lube Oil Pump A1 and Sump Pump D1 out of service. Southeast Nebraska is in a severe thunderstorm warning, which includes intense electrical storm activity Turnover: The plant is operating at 100% power near the end of the current fuel cycle when the crew takes the shift. The plant is in a normal configuration with Rx Feed Pump Lube Oil Pump A1 and Sump Pump D1 out of service. Southeast Nebraska is in a severe thunderstorm warning, which includes intense electrical storm activity.

Scenario: The plant is operating at 100% power near the end of the current fuel cycle when the crew takes the shift. The plant is in a normal configuration with Rx Feed Pump Lube Oil Pump A1 and Sump Pump D1 out of service. Southeast Nebraska is in a severe thunderstorm warning, which includes intense electrical storm activity.

When the Technical Specification assessment is complete, a bus ground results in a loss of MCC F. The loss of MCC F causes a trip of RFPT A due to low lube oil pressure, A1 pump tagged out, requiring a power reduction to maintain reactor water level and to reduce heat load to the capacity of the remaining pump. When conditions have stabilized, RFP B vibrations increase to the point that the pump must be tripped. The crew will scram the reactor and trip the last remaining RFP.

When RCIC initiates, a break develops on the "A" feedwater line inside the drywell. HPCI fails to automatically start, but may be manually started. RCIC will not inject due to the location of the leak. The feedwater line check valve leaks, and the leak continues. HPCI can maintain RPV water level for the selected leak size. Containment sprays will be required by the EOPs. Drywell sprays will be initiated. Drywell sprays will fail to isolate on low containment pressure. The operator must either maintain pressure by controlling spray flowrate or manally isolate drywell sprays when containment becomes negative before air is drawn into the primary containment. The scenario ends when RPV water level is being restored to the normal band, drywell pressure is being controlled and classifications have been made.

Event Malf. No. Event Event Description No. Type*

Initiated by N (RO) Power reduction of 100 MWe 1.

turnover Trigger 1 T, I Drywell Pressure Instrument Fails UPSC (PC-PS-12C) 2.

(RO, BOP, SRO)

Facility: COOPER Scenario No.: 3 Op-Test No.:1 Examiners: _Paul Gage_______________ Operators: __________________________

_Steve Garchow___________ __________________________

_Kelly Clayton____________ __________________________

Plant Status: The plant is operating at 100% power near the end of the current fuel cycle when the crew takes the shift. The plant is in a normal configuration with Rx Feed Pump Lube Oil Pump A1 and Sump Pump D1 out of service. Southeast Nebraska is in a severe thunderstorm warning, which includes intense electrical storm activity Turnover: The plant is operating at 100% power near the end of the current fuel cycle when the crew takes the shift. The plant is in a normal configuration with Rx Feed Pump Lube Oil Pump A1 and Sump Pump D1 out of service. Southeast Nebraska is in a severe thunderstorm warning, which includes intense electrical storm activity.

Scenario: The plant is operating at 100% power near the end of the current fuel cycle when the crew takes the shift. The plant is in a normal configuration with Rx Feed Pump Lube Oil Pump A1 and Sump Pump D1 out of service. Southeast Nebraska is in a severe thunderstorm warning, which includes intense electrical storm activity.

When the Technical Specification assessment is complete, a bus ground results in a loss of MCC F. The loss of MCC F causes a trip of RFPT A due to low lube oil pressure, A1 pump tagged out, requiring a power reduction to maintain reactor water level and to reduce heat load to the capacity of the remaining pump. When conditions have stabilized, RFP B vibrations increase to the point that the pump must be tripped. The crew will scram the reactor and trip the last remaining RFP.

When RCIC initiates, a break develops on the "A" feedwater line inside the drywell. HPCI fails to automatically start, but may be manually started. RCIC will not inject due to the location of the leak. The feedwater line check valve leaks, and the leak continues. HPCI can maintain RPV water level for the selected leak size. Containment sprays will be required by the EOPs. Drywell sprays will be initiated. Drywell sprays will fail to isolate on low containment pressure. The operator must either maintain pressure by controlling spray flowrate or manally isolate drywell sprays when containment becomes negative before air is drawn into the primary containment. The scenario ends when RPV water level is being restored to the normal band, drywell pressure is being controlled and classifications have been made.

Event Malf. No. Event Event Description No. Type*

Initiated by N (RO) Power reduction of 100 MWe 1.

turnover Trigger 1 T, I Drywell Pressure Instrument Fails UPSC (PC-PS-12C) 2.

(RO, BOP, SRO)

Facility: COOPER Scenario No.: 4 Op-Test No.:1 Examiners: _Paul Gage_______________ Operators: __________________________

_Steve Garchow___________ __________________________

_Kelly Clayton____________ __________________________

Plant Status: The plant is operating at 75% power near the end of the current fuel cycle when the crew takes the shift. The plant is in a normal configuration with the exception of Rx Feed Pump Lube Oil Pump A1 and Sump Pump D1 are out of service. Southeast Nebraska is in a severe thunderstorm warning, which includes intense electrical storm activity Turnover: The plant is operating at 75% power near the end of the current fuel cycle when the crew takes the shift. The plant is in a normal configuration with the exception of Rx Feed Pump Lube Oil Pump A1 and Sump Pump D1 are out of service. Southeast Nebraska is in a severe thunderstorm warning, which includes intense electrical storm activity.

Scenario: The plant is operating at 75% power near the end of the current fuel cycle when the crew takes the shift. The plant is in a normal configuration with the exception of Rx Feed Pump Lube Oil Pump A1 and Sump Pump D1 are out of service.

Southeast Nebraska is in a severe thunderstorm warning, which includes intense electrical storm activity. A surveillance testing the operability of main steam isolation valve 80A fails resulting in one MSL being isolated. A lightning strike on the grid results in the loss of the 69kV OPPD Nebraska City Line and requires performing a tech spec surveillance. When the Technical Specification assessment is complete, a MSL radiation monitor fails upscale. When the assessment of this is complete, APRM Channel B fails INOP. When the actions for the APRM are complete, a loss of off-site power, reactor scram, and a steam leak into primary containment occur. Later in the scenario EDG-2 fails. The scenario ends when RPV water level is being restored to the normal band, drywell pressure is being controlled and classifications have been made.

Event Malf. No. Event Event Description No. Type*

Trigger 1 N (SRO, Power reduction to < 70% using control rods and recirc flow 1.

RO)

Trigger 2 C, T Surveillance 6.MS.201 Section 5 (Failure of MSIV MOV-80A to open) 2.

(BOP,

ES-401 Record of Rejected KAs Form ES-401-4 For the RO Examination:

Tier Group Randomly Selected KA Reason for Rejection 1 1 295028 High Drywell Not applicable to the RO position; represents SRO-level Temperature, 2.1.6, Ability to knowledge supervise and assume a management role during plant transients and upset conditions 1 1 295031 Reactor Low Water Not applicable to the RO position; represents SRO-level Level, 2.4.41, Knowledge of the knowledge emergency action level thresholds and classifications 1 1 295031 Reactor Low Water IR for RO is less than 2.5 Level, 2.2.8, Knowledge of the process for determining if the proposed change (et.al.) Involves an unreviewed safety question 1 1 295019 Partial or Total Loss of IR for RO is less than 2.5 Instrument Air, 2.2.18, Knowledge of the process for managing maintenance activities during shutdown operations 1 1 2.4.02 Knowledge of system There are no EOP entry conditions associated with loss setpoints, interlocks, and of shutdown cooling, therefore another KA was selected automatic actions associated with at random.

EOP entry conditions, as applied to 295021, Loss of Shutdown Cooling 1 2 295014 Inadvertent Reactivity IR for RO is less than 2.5 Addition, 2.2.31, Knowledge of the effects of alterations on core configuration 1 2 295014 Inadvertent Reactivity Not applicable to the RO position; represents SRO-level Addition, 2.4.41, Knowledge of knowledge the emergency action level thresholds and classifications 1 2 295017 High Offsite Release Not applicable to the RO position (per M. Barton, Rate, AK1.02, Operational 1/19/05); replaced by EK3.03, Knowledge of the reasons Implications of high offsite for Control Room isolation as pertains to High Offsite release rate as affects protection Release Rate of the general public 2 1 262002 UPS, K2 No KAs in catalog for category K2 2 2 239003 MSIV Leakage Control Not Applicable to Cooper (per M. Barton, 1/19/05) ;

replaced by 233000 K1.02 ES-401 , NUREG 1021 Revision 9

ES-401 Record of Rejected KAs Form ES-401-4 2 2 272000 Radiation Monitoring No applicable reporting requirements System, 2.4.30, Knowledge of which events related to system operation/status which should be reported to outside agencies 2 2 256000, K6.10, Knowledge of the This KA was replace due to there being no relationship effect that a loss or malfunction of between main steam and the condensate system.

the main steam system will have Another KA was chosen from the same KA category.

on the condensate feedwater system.

2 2 219000 RHR LPCI: Suppression There is no automatic operation or repositioning of valves Pool Cooling Mode, A3.01, associated with RHR LPCI Suppression Pool Cooling Monitor automatic operation as Mode. Replaced by A1.04, Ability to predict and/or applied to valves monitor changes in parameters associated with RHR LPCI Suppression Pool Cooling Mode controls, including suppression pool level.

2 2 214000 Rod Position Indication Temperature indication is limited to PMIS (computer System, A4.03, Ability to system) with effectively no control over CR drive manually operate or monitor temperature. Rejected because no discriminating control rod drive temperature question can be written. Replaced by A4.02, Ability to from the control room manually operate or monitor control position from the control room.

3 0 2.1.26, Knowledge of non-nuclear IR for RO is less than 2.5 safety procedures 3 0 2.1.13, Knowledge of facility IR for RO is less than 2.5 requirements for controlling vital and controlled areas 3 0 2.2.19, Knowledge of IR for RO is less than 2.5 maintenance work order requirements 3 0 2.2.17, Knowledge of the process IR for RO is less than 2.5 for managing maintenance activities during power operations 3 0 2.2.4, Ability to explain variations Cooper is a single unit facility between units at a facility 3 0 2.215, Ability to identify and utilize IR for RO is less than 2.5 as-built design and configuration change documentation...

3 0 2.3.7, Knowledge of the process IR for RO is less than 2.5 for preparing a radiation work permit ES-401 , NUREG 1021 Revision 9

ES-401 Record of Rejected KAs Form ES-401-4 For the SRO Examination:

Tier Group Randomly Selected KA Reason for Rejection 2 1 Predict/monitor changes assoc. There are no MG sets associated with the UPS at with motor generator outputs Cooper. Another KA was randomly selected.

2 1 Monitor SRV and acoustical There are no SRV acoustical monitors at Cooper.

monitors Another KA was randomly selected from the same KA category.

2 1 Ability to predict the impacts of a A small break LOCA has no impact on the operation of small break LOCA on ADS, and the ADS. Another KA was randomly selected.

based on those predictions use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations.

1 1 Ability to determine or interpret Unable to construct an SRO only question for this KA.

suppression pool level as applied Another KA was randomly selected.

to high drywell pressure 2 2 Ability to perform specific system Unable to construct an SRO only question for this KA.

and integrated plant procedures Another KA was randomly selected.

during different modes of plant operations, as applied to 201003, Control Rod and Drive Mechanism.

2 1 Ability to execute procedure steps Unable to construct an SRO only question for this KA.

as applied to 206000, High Another KA was randomly selected.

Pressure Coolant Injection.

2 2 290002 Reactor Vessel Internals, Rejected because operators have very limited direct A2.05, Ability to predict the knowledge of reactor vessel internals and because they impacts of exceeding thermal are static components have little ability to operate the limits on reactor vessel internals components. Replaced by 202001 A2.09, Ability to and on the basis of the prediction predict the impacts of recirculation scoop tube lockup on use procedures to correct, the recirculation system and on the basis of the control, or mitigate the prediction use procedures to correct, control, or mitigate consequences of the abnormal the consequences of the abnormal conditions or conditions or operations. operations.

ES-401 , NUREG 1021 Revision 9

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