ML17122A086

2017-03 Final Outlines

ML17122A086
Person / Time
Site:	Cooper
Issue date:	03/16/2017
From:	Vincent Gaddy Operations Branch IV
To:	Nebraska Public Power District (NPPD)
References
Download: ML17122A086 (42)

Text

ES-401 BWR Examination Outline Form ES-401-1 Rev 4 Facility: Cooper Nuclear Station Date of Exam:

March 201 7 Tier Group RO K/A Category Points SRO-Only Points K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4 G* Total A2 G* Total 1. Emergency & Abnormal Plant Evolutions 1 3 3 4 N/A 4 3 N/A 3 20 4 3 7 2 2 1 1 1 1 1 7 1 2 3 Tier Totals 5 4 5 5 4 4 27 5 5 10 2. Plant Systems 1 2 2 3 4 3 2 2 2 2 2 2 26 2 3 5 2 2 1 1 1 1 1 1 1 1 1 1 12 0 2 1 3 Tier Totals 4 3 4 5 4 3 3 3 3 3 3 38 4 4 8 3. Generic Knowledge and Abilities Categories 1 2 3 4 10 1 2 3 4 7 3 3 2 2 2 2 1 2 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO

-only outline, the "Tier Totals" in each K/A category shall not be less than two). (One Tier 3 Radiation Control K/A is allowed if the K/A is replaced by a K/A from another Tier 3 Category.)

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 with justification; operationally important, site

-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES

-401 for guidance regarding the elimination of inappropriate K/A 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 K/As having an importance rating (IR) 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 K/A categories.

7. The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES

-401 for the applicable K/As.

8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics' importance ratings (IRs) 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; if fuel handling equipment is sampled in a category other than Category A2 or G* on the SRO

-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO

-only exams.

9. For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES

-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.

G* Generic K/As

ES-401 2 Form ES-401-1 Rev 4 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions

- Tier 1/Group 1 (RO / SRO)

E/APE # / Name / Safety Function K 1 K 2 K 3 A 1 A 2 G* K/A Topic(s)

IR # 295001 Partial or Complete Loss of Forced Core Flow Circulation / 1 & 4 X Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION

(CFR: 41.5

) AK3.0 1 Reactor water level response 3.4 1 295003 Partial or Complete Loss of AC / 6 X Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER: (CFR: 41.7)

AA1.0 2 Emergency generators 4.2 2 295004 Partial or Total Loss of DC Pwr / 6 X Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF D.C.

POWER: (CFR: 41.10)

AA2.01 Cause of partial or complete loss of D.C. power 3.2 3 295005 Main Turbine Generator Trip / 3 X 2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation.

(CFR: 41.10) 4.3 4 295006 SCRAM / 1 X Knowledge of the operational implications of the following concepts as they apply to SCRAM:

(CFR: 41.8 to 41.10)

AK1.0 3 Reactivity control 3.7 5 295016 Control Room Abandonment / 7 X Knowledge of the interrelations between CONTROL ROOM ABANDONMENT and the following:

(CFR: 41.7)

AK2.0 1 Remote shutdown panel: Plant

-Specific 4.4 6 295018 Partial or Total Loss of CCW / 8 X Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER

(CFR: 41.5)

AK3.0 7 Cross-connecting with backup systems 3.1 7 295019 Partial or Total Loss of Inst. Air / 8 X Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR

(CFR: 41.7)

AA1.0 3 Instrument air compressor power supplies 3.0 8 295021 Loss of Shutdown Cooling / 4 X Knowledge of the reasons for the following responses as they apply to LOSS OF SHUTDOWN COOLING

(CFR: 41.5)

AK3.05 Establishing alternate heat removal flow paths 3.6 9 295023 Refueling Acc / 8 X Ability to determine and/or interpret the following as they apply to REFUELING ACCIDENTS

(CFR: 41.10)

AA2.0 2 Fuel Pool Level 3.4 10 295024 High Drywell Pressure / 5 X 2.2.44 Ability to interpret control room indications to verify status and operation of a system, and understand how operator actions and directives affect plant and system conditions. (CFR: 41.5) 4.2 11 295025 High Reactor Pressure / 3 X Knowledge of the operational implications of the following concepts as they apply to HIGH REACTOR PRESSURE: (CFR: 41.8 to 41.10)

EK1.0 3 Safety/relief valve tailpipe temperature / pressure relationships 3.6 12 295026 Suppression Pool High Water Temp. / 5 X Knowledge of the interrelations between SUPPRESSION POOL HIGH WATER TEMPERATURE and the following:

(CFR: 41.7)

EK2.06 Suppression pool level 3.5 13 295027 High Containment Temperature / 5 NOT APPLICABLE

ES-401 3 Form ES-401-1 Rev 4 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions

- Tier 1/Group 1 (RO / SRO)

E/APE # / Name / Safety Function K 1 K 2 K 3 A 1 A 2 G* K/A Topic(s)

IR # 295028 High Drywell Temperature / 5 X Ability to operate and/or monitor the following as they apply to HIGH DRYWELL TEMPERATURE

(CFR: 41.7)

EA1.0 1 Drywell spray: Mark

-I&II 3.8 14 295030 Low Suppression Pool Wtr Lvl / 5 X Ability to determine and/or interpret the following as they apply to LOW SUPPRESSION POOL WATER LEVEL

(CFR: 41.10)

EA2.0 4 Drywell/ suppression chamber differential pressure: Mark

-I&II 3.5 15 295031 Reactor Low Water Level / 2 X 2.4.1 Knowledge of EOP entry conditions and immediate action steps. (CFR: 41.10) 4.6 16 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown / 1 X Knowledge of the operational implications of the following concepts as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN

(CFR: 41.8 to 41.10)

EK1.0 2 Reactor water level effects on reactor power 4.1 17 295038 High Off

-site Release Rate / 9 X Knowledge of the interrelations between HIGH OFF-SITE RELEASE RATE and the following:

(CFR: 41.7)

EK2.0 2 Offgas system 3.6 18 600000 Plant Fire On Site / 8 X Knowledge of the reasons for the following responses as they apply to PLANT FIRE ON SITE

AK3.04 Actions contained in the abnormal procedure for plant fire on site 2.8 19 700000 Generator Voltage and Electric Grid Disturbances / 6 X Ability to operate and/or monitor the following as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES

(CFR: 41.5 and 41.10)

AA1.0 5 Engineered safety features 3.9 20 295003 Partial or Complete Loss of AC / 6 X Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C.

POWER : (CFR: 43.5)

AA2.0 3 Battery status: Plant

-Specific 3.5 76 295021 Loss of Shutdown Cooling / 4 X 2.2.40 Ability to apply Technical Specifications for a system. (CFR: 43.2 / 43.5) 4.7 77 295023 Refueling Acc / 8 X Ability to determine and/or interpret the following as they apply to REFUELING ACCIDENTS: (CFR: 43.5)

AA2.0 4 fOccurrence of fuel handling accident 4.1 78 295038 High Off

-site Release Rate / 9 X 2.4.18 Knowledge of the specific bases for EOPs.

(CFR: 43.1) 4.0 79 295031 Reactor Low Water Level / 2 X Ability to determine and/or interpret the following as they apply to REACTOR LOW WATER LEVEL :

(CFR: 43.5)

EA2.04 Adequate core cooling 4.8 80 600000 Plant Fire On Site / 8 X 2.4.30 Knowledge of events related to system operation / status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.

(CFR: 43.5) 4.1 81 700000 Generator Voltage and Electric Grid Disturbances / 6 X Ability to determine and/or interpret the following as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES:

(CFR:43.5) A A2.0 5 Operational status of offsite circuit 3.8 82 K/A Category Totals:

3 3 4 4 3/4 3/3 Group Point Total: 20/7 ES-401 4 Form ES-401-1 Rev 4 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions

- Tier 1/Group 2 (RO / SRO)

E/APE # / Name / Safety Function K 1 K 2 K 3 A 1 A 2 G* K/A Topic(s)

IR # 295002 Loss of Main Condenser Vac / 3 X Knowledge of the operational implications of the following concepts as they apply to LOSS OF MAIN CONDENSER VACUUM : (CFR: 41.8 to 41.10)

AK1.03 Loss of heat sink 3.6 21 295007 High Reactor Pressure / 3 NOT SELECTED 295008 High Reactor Water Level / 2 X Knowledge of the interrelations between HIGH REACTOR WATER LEVEL and the following: (CFR: 41.7 / 45.8)

AK2.03 Reactor water level control 3.6 22 295009 Low Reactor Water Level / 2 NOT SELECTED 295010 High Drywell Pressure / 5 NOT SELECTED 295011 High Containment Temp / 5 NOT SELECTED 295012 High Drywell Temperature / 5 NOT SELECTED 295013 High Suppression Pool Temp. / 5 NOT SELECTED 295014 Inadvertent Reactivity Addition / 1 NOT SELECTED 295015 Incomplete SCRAM / 1 X Knowledge of the reasons for the following responses as they apply to INCOMPLETE SCRAM : (CFR: 41.5)

AK3.01 Bypassing rod insertion blocks 3.4 23 295017 High Off

-site Release Rate / 9 NOT SELECTED 295020 Inadvertent Cont. Isolation / 5 & 7 X Ability to operate and/or monitor the following as they apply to INADVERTENT CONTAINMENT ISOLATION : (CFR: 41.7)

AA1.03 Containment ventilation system: Plant

-Specific 2.9 24 295022 Loss of CRD Pumps / 1 X Ability to determine and/or interpret the following as they apply to LOSS OF CRD PUMPS : (CFR: 41.10)

AA2.01 Accumulator pressure 3.5 25 295029 High Suppression Pool Wtr Lvl / 5 NOT SELECTED 295032 High Secondary Containment Area Temperature / 5 NOT SELECTED 295033 High Secondary Containment Area Radiation Levels / 9 NOT SELECTED 295034 Secondary Containment Ventilation High Radiation / 9 X 2.4.31 Knowledge of annunciator alarms, indications, or response procedures. (CFR: 41.10) 4.2 28 295035 Secondary Containment High Differential Pressure / 5 NOT SELECTED 295036 Secondary Containment High Sump/Area Water Level / 5 X Knowledge of the operational implications of the following concepts as they apply to SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL :

(CFR: 41.8 to 41.10)

EK1.02 Electrical ground/ circuit malfunction 2.6 27 500000 High CTMT Hydrogen Conc. / 5 NOT SELECTED 295014 Inadvertent Reactivity Addition / 1 X 2.4.4 Ability to recognize abnormal indications for system operating parameters that are entry

-level conditions for emergency and abnormal operating procedures.(CFR: 43.2) 4.7 83 295029 High Suppression Pool Wtr Lvl / 5 X Ability to determine and/or interpret the following as they apply to HIGH SUPPRESSION POOL WATER LEVEL : (CFR: 43.5)

EA2.01 Suppression pool water level 3.9 84 500000 High CTMT Hydrogen Conc. / 5 X 2.4.6 Knowledge of EOP mitigation strategies.

(CFR: 43.5) 4.7 85 K/A Category Point Totals:

2 1 1 1 1/1 1/2 Group Point Total:

7/3 ES-401 5 Form ES-401-1 Rev 4 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems

- Tier 2/Group 1 (RO / SRO)

System # / Name K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4 G* K/A Topic(s)

IR # 203000 RHR/LPCI: Injection Mode X Knowledge of electrical power supplies to the following:

(CFR: 41.7)

K2.03 Initiation logic 2.7 26 205000 Shutdown Cooling X Knowledge of the effect that a loss or malfunction of the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) will have on following: (CFR: 41.7)

K3.01 Reactor pressure 3.3 29 205000 Shutdown Cooling X Knowledge of shutdown cooling system (RHR shutdown cooling mode) design feature(s) and/or interlocks which provide for the following:

(CFR: 41.7) K4.02 High pressure isolation: Plant

-Specific 3.7 30 206000 HPCI X Knowledge of the operational implications of the following concepts as they apply to HIGH PRESSURE COOLANT INJECTION SYSTEM :

(CFR: 41.5)

K5.05 Turbine speed control: BWR

-2,3,4 3.3 31 206000 HPCI X Knowledge of the effect that a loss or malfunction of the following will have on the HIGH PRESSURE COOLANT INJECTION SYSTEM : (CFR: 41.7)

K6.02 D.C. power: BWR

-2,3,4 3.3 32 207000 Isolation (Emergency)

Condenser NOT APPLICABLE 209001 LPCS X Knowledge of LOW PRESSURE CORE SPRAY SYSTEM design feature(s) and/or interlocks which provide for the following: (CFR: 41.7)

K4.04 Line break detection 3.0 33 209002 HPCS NOT APPLICABLE 211000 SLC X Ability to predict and/or monitor changes in parameters associated with operating the STANDBY LIQUID CONTROL SYSTEM controls including:

(CFR: 41.5)

A1.0 8 RWCU system lineup 3.7 34 212000 RPS X Ability to (a) predict the impacts of the following on the REACTOR PROTECTION SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

(CFR: 41.5)

A2.16 Changing mode switch position 4.0 35 215003 IRM X Ability to monitor automatic operations of the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM including:

(CFR: 41.7)

A3.0 4 Control rod block status 3.5 36 215004 Source Range Monitor X Ability to manually operate and/or monitor in the control room:

(CFR: 41.7)

A4.07 Verification of proper functioning

/ operability 3.4 37 215005 APRM / LPRM X 2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. (CFR: 41.5

) 4.4 38 217000 RCIC X Knowledge of the physical connections and/or cause/effect relationships between REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) and the following:

(CFR: 41.2 to 41.9)

K1.0 2 Nuclear boiler system 3.5 39 ES-401 6 Form ES-401-1 Rev 4 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems

- Tier 2/Group 1 (RO / SRO)

System # / Name K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4 G* K/A Topic(s)

IR # 218000 ADS X Knowledge of electrical power supplies to the following:

(CFR: 41.7)

K2.01 ADS logic 3.1 40 223002 PCIS/Nuclear Steam Supply Shutoff X Knowledge of the effect that a loss or malfunction of the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT

-OFF will have on following:

(CFR: 41.7)

K3.20 Standby gas treatment system 3.3 41 239002 SRVs X Knowledge of the operational implications of the following concepts as they apply to RELIEF/SAFETY VALVES: (CFR: 41.5)

K5.0 2 Safety function of SRV operation 3.7 42 259002 Reactor Water Level Control X Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR WATER LEVEL CONTROL SYSTEM: (CFR: 41.7)

K6.02 A.C. power 3.3 43 259002 Reactor Water Level Control X Ability to predict and/or monitor changes in parameters associated with operating the REACTOR WATER LEVEL CONTROL SYSTEM controls including:

(CFR: 41.5)

A1.0 5 FWRV/startup level control position: Plant

-Specific 2.9 44 261000 SGTS X Ability to (a) predict the impacts of the following on the STANDBY GAS TREATMENT SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

(CFR: 41.5

) A2.1 1 High containment pressure 3.2 45 262001 AC Electrical Distribution X Ability to monitor automatic operations of the A.C.

ELECTRICAL DISTRIBUTION including:

(CFR: 41.7)

A3.04 Load sequencing 3.4 46 262001 AC Electrical Distribution X Ability to manually operate and/or monitor in the control room:

(CFR: 41.7)

A4.0 5 Voltage, current, power, and frequency on A.C. buses 3.3 47 262002 UPS (AC/DC) X 2.1.27 Knowledge of system purpose and/or function. (CFR: 41.7) 3.9 48 263000 DC Electrical Distribution X Knowledge of the physical connections and/or cause/effect relationships between D.C. ELECTRICAL DISTRIBUTION and the following:

(CFR: 41.2 to 41.9)

K1.0 2 Battery charger and battery 3.2 49 264000 EDGs X Knowledge of the effect that a loss or malfunction of the EMERGENCY GENERATORS (DIESEL/JET) will have on following:

(CFR: 41.7 / 45.4)

K3.0 3 Major loads powered from electrical buses fed by the emergency generator(s) 4.1 50 264000 EDGs X Knowledge of EMERGENCY GENERATORS (DIESEL/JET) design feature(s) and/or interlocks which provide for the following: (CFR: 41.7)

K4.08 Automatic startup 3.8 51 300000 Instrument Air X Knowledge of the operational implications of the following concepts as they apply to the INSTRUMENT AIR SYSTEM:

(CFR: 41.5 / 45.3)

K5.01 Air compressors 2.5 52 ES-401 7 Form ES-401-1 Rev 4 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems

- Tier 2/Group 1 (RO / SRO)

System # / Name K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4 G* K/A Topic(s)

IR # 400000 Component Cooling Water X Knowledge of CCWS design feature(s) and or interlocks which provide for the following:

(CFR: 41.7)

K4.01 Automatic start of standby pump 3.4 53 203000 RHR/LPCI: Injection Mode X Ability to (a) predict the impacts of the following on the RHR/LPCI: INJECTION MODE (PLANT SPECIFIC) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

A2.04 A.C. failures 3.6 86 212000 RPS X 2.2.25 Knowledge of the bases in Tech Specs for LCOs and Safety limits (CFR: 41.5 / 41.7 / 43.2) 4.2 87 215005 APRM / LPRM X Ability to (a) predict the impacts of the following on the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

(CFR: 41.5)

A2.08 Faulty or erratic operation of detectors / systems 3.4 88 218000 ADS X 2.4.20 Knowledge of the operational implications of EOP warnings, cautions, and notes.

(CFR: 41.10 / 43.5) 4.3 89 300000 Instrument Air X 2.4.45 Ability to prioritize and interpret the significance of each annunciator or alarm.

(CFR: 41.10 / 43.5) 4.3 90 K/A Category Point Totals:

2 2 3 4 3 2 2 2/2 2 2 2/3 Group Point Total:

26/5 ES-401 8 Form ES-401-1 Rev 4 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems

- Tier 2/Group 2 (RO / SRO)

System # / Name K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4 G* K/A Topic(s)

IR # 201001 CRD Hydraulic X Knowledge of the physical connections and/or cause

-effect relationships between CONTROL ROD DRIVE HYDRAULIC SYSTEM and the following:

(CFR: 41.2 to 41.9

) K1.07 Reactor protection system 3.4 54 201002 RMCS X Knowledge of the effect that a loss or malfunction of the REACTOR MANUAL CONTROL SYSTEM will have on following:

(CFR: 41.7

) K3.03 Ability to process rod block signals 2.9 5 9 201003 Control Rod and Drive Mechanism X Knowledge of the operational implications of the following concepts as they apply to CONTROL ROD AND DRIVE MECHANISM :

(CFR: 41.5

) K5.04 fRod sequence patterns 3.1 56 201004 RSCS NOT APPLICABLE 201005 RCIS NOT APPLICABLE 201006 RWM NOT SELECTED 202001 Recirculation X Knowledge of electrical power supplies to the following:

(CFR: 41.7)

K2.02 MG sets: Plant

-Specific 3.2 57 202002 Recirculation Flow Control X Knowledge of RECIRCULATION FLOW CONTROL SYSTEM design feature(s) and/or interlocks which provide for the following:

(CFR: 41.7)

K4.07 Minimum and maximum pump speed setpoints 2.9 58 204000 RWCU NOT SELECTED 214000 RPIS NOT SELECTED 215001 Traversing In

-Core Probe NOT SELECTED 215002 RBM NOT SELECTED 216000 Nuclear Boiler Inst.

X Knowledge of the effect that a loss or malfunction of the following will have on the NUCLEAR BOILER INSTRUMENTATION :

(CFR: 41.7

) K6.01 A.C. electrical distribution 3.1 5 5 219000 RHR/LPCI: Torus/Pool Cooling Mode NOT SELECTED 223001 Primary CTMT and Aux.

NOT SELECTED 226001 RHR/LPCI: CTMT Spray Mode NOT SELECTED 230000 RHR/LPCI: Torus/Pool Spray Mode NOT SELECTED 233000 Fuel Pool Cooling/Cleanup X Ability to predict and/or monitor changes in parameters associated with operating the FUEL POOL COOLING AND CLEAN

-UP controls including:

(CFR: 41.5) A1.06 System flow 2.5 60 234000 Fuel Handling Equipment X Ability to monitor automatic operations of the FUEL HANDLING EQUIPMENT including:

(CFR: 41.7

) A3.02 fInterlock operation 3.1 61 239001 Main and Reheat Steam NOT SELECTED 239003 MSIV Leakage Control NOT SELECTED

ES-401 9 Form ES-401-1 Rev 4 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems

- Tier 2/Group 2 (RO / SRO)

System # / Name K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4 G* K/A Topic(s)

IR # 241000 Reactor/Turbine Pressure Regulator X Ability to (a) predict the impacts of the following on the REACTOR/TURBINE PRESSURE REGULATING SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6)

A2.07 Loss of condenser vacuum 3.7 62 245000 Main Turbine Gen. / Aux.

X Ability to manually operate and/or monitor in the control room: (CFR: 41.7)

A4.02 Generator controls 3.1 63 256000 Reactor Condensate NOT SELECTED 259001 Reactor Feedwater NOT SELECTED 268000 Radwaste NOT SELECTED 271000 Offgas NOT SELECTED 272000 Radiation Monitoring NOT SELECTED 286000 Fire Protection X 2.1.28 Knowledge of the purpose and function of major system components and controls.

(CFR: 41.7) 4.1 64 288000 Plant Ventilation NOT SELECTED 290001 Secondary CTMT NOT SELECTED 290003 Control Room HVAC NOT SELECTED 290002 Reactor Vessel Internals X Knowledge of the physical connections and/or cause

- effect relationships between REACTOR VESSEL INTERNALS and the following:

(CFR: 41.2 to 41.9)

K1.15 Nuclear boiler instrumentation 3.4 65 223001 Primary CTMT and Aux.

X 2.4.21 Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

(CFR: 41.7 / 43.5

) 4.6 91 239001 Main and Reheat Steam X Ability to (a) predict the impacts of the following on the MAIN AND REHEAT STEAM SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

(CFR: 41.5 / 45.6)

A2.10 Closure of one or more MSIV's at powe r 3.9 92 259001 Reactor Feedwater X Ability to (a) predict the impacts of the following on the REACTOR FEEDWATER SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

(CFR: 41.5 / 45.6)

A2.07 Reactor water level control system malfunctions 3.8 93 K/A Category Point Totals:

2 1 1 1 1 1 1 1/2 1 1 1/1 Group Point Total:

12/3 ES-401 Generic Knowledge and Abilities Outline (Tier 3)

Form ES-401-3 Rev 4 Facility: Cooper Nuclear Station Date of Exam:

March 2017 Category K/A # Topic RO SRO-Only IR # IR # 1. Conduct of Operations

2.1.1 Knowledge

of conduct of operations requirements.

(CFR: 41.10) 3.8 66 2.1.3 7 Knowledge of procedures, guidelines, or limitations associated with reactivity management.

(CFR: 41.1) 4.3 67 2.1.45 Ability to identify and interpret diverse indications to validate the response of another indication. (CFR: 41.7) 4.3 68 2.1.4 Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, "no

-solo" operation, maintenance of active license status, 10CFR55, etc.

(CFR: 43.2) 3.8 94 2.1.36 Knowledge of procedures and limitations involved in core alterations. (CFR: 43.6) 4.1 95 Subtotal 3 2 2. Equipment Control 2.2.6 Knowledge of the process for making changes to procedures.

(CFR: 41.10) 3.0 69 2.2.14 Knowledge of the process for controlling equipment configuration or status. (CFR: 41.10) 3.9 70 2.2.39 Knowledge of less than or equal to one hour Technical Specification action statements for systems. (CFR: 41.7 / 41.10) 3.9 71 2.2.20 Knowledge of the process for managing troubleshooting activities.

(CFR: 43.5) 3.8 96 2.2.38 Knowledge of conditions and limitations in the facility license.

(CFR: 43.1) 4.5 97 Subtotal 3 2 3. Radiation Control 2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions. (CFR: 41.12) 3.2 72 2.3.12 Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high

-radiation areas, aligning filters, etc.

(CFR: 41.12

) 3.2 73 2.3.14 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

(CFR: 43.4) 3.8 98 Subtotal 2 1 4. Emergency Procedures / Plan 2.4.28 Knowledge of procedures relating to a security event (non

-safeguards information). (CFR: 41.10 / 43.5 / 45.13) 3.2 74 2.4.39 Knowledge of RO responsibilities in emergency plan implementation.

(CFR: 41.10

) 3.9 75 2.4.16 Knowledge of EOP implementation hierarchy and coordination with other support procedures or guidelines such as, operating procedures, abnormal operating procedures, and severe accident management guidelines.

(CFR: 43.5) 4.4 99 2.4.38 Ability to take actions called for in the facility emergency plan, including supporting or acting as emergency coordinator if required. (CFR: 41.10 / 43.5 / 45.11) 4.4 100 Subtotal 2 2 Tier 3 Point Total 10 7 ES-401 Generic Knowledge and Abilities Outline (Tier 3)

Form ES-401-3 Rev 4 Rev ision statement:

Rev 2 added question numbers and added importance rating for RO K/A 286000 2.1.32 and corrected importance ratings for SRO K/As 203000 A2.04, 215005 A2.08, 239001 A2.10, and 259001 A2.07

. Also, replaced 1/1 K/A 295026 EK2.03 with EK2.06.

Correct T1/G2 totals on category totals for RO K1 and K3 on ES-401-1 page 1.

Rev 3 corrected RO T1/G2 Tier Totals for K1 and K3 on ES

-401-1 page 1. Added header for form ES

-401-3. Swapped question 26 and 28 question number assignments to prevent having four answers the same in a row on the written exam.

Swapped question 55 and 59 question number assignments to prevent having four answers the same in a row on the written exam.

Rev 4 replaced K/A 286000 G2.1.32 for RO T2/G2 question 64 with K/A 286000 G2.1.28.

Rev 04 Rev 04 ES-401 Record of Rejected K/As Form ES-401-4 Tier / Group (Original)

Randomly Selected K/A (New) Reason for Rejection RO T 1/G 1 295027 Not Selected 295027 Not Applicable Because CNS has a Mark I containment and NUREG 1123 states EPE 295027 High Containment Temperature is for Mark III containments only

, EPE 295027 was changed from NOT SELECTED to NOT APPLICABLE

. Page 1 point totals not affected by this change. (Rev 1) SRO T1/G1 295027 G2.4.18 295038 G2.4.18 Because CNS has a Mark I containment and NUREG 1123 states EPE 295027 High Containment Temperature is for Mark III containments only, EPE 295027 was replaced with randomly selected EPE 295038 High Off

-site Release Rate. KA G2.4.18 was not changed.

Page 1 point totals not affected by this change

. (Rev 1) RO T 2/G 1 239002 K4.04 209001 K4.04 Because CNS does have a Low Pressure Core Spray system, System 209001 LPCS was changed from NOT APPLICABLE. Since System 239002 SRVs was one of the systems sampled twice

, 239002 SRVs K4.04 was replaced with 209001 LPCS K4.04 so that LPCS is sampled at least once.

Page 1 point totals not affected by this change. (Rev 1) RO T 2/G 2 201004 Not Selected 201004 Not Applicable Because the Rod Sequence Control System is no longer use d at CNS , System 201004 RSCS was changed from NOT SELECTED to NOT APPLICABLE. Page 1 point totals not affected by this change.

(Rev. 1) RO T2/G2 20100 5 Not Selected 20100 5 Not Applicable Because CNS does not have a Rod Control and Information System , System 201005 RCIS was changed from NOT SELECTED to NOT APPLICABLE.

Page 1 point totals not affected by this change. (Rev 1) RO T1/G1 295026 EK2.03 295026 EK2.06 Because a discriminatory, operationally valid RO question could not be developed, replaced 295026 EK2.03 with randomly selected EK2.06. Page 1 point totals not affected by this change. (Rev 2) RO T2/G1 211000 A1.06 211000 A1.08 The only SLC flow indicator at CNS is a local float type meter on the SLC Test Tank inlet piping. No flow indication is available for SLC injection to the RPV. A discriminatory, operationally valid question could not be developed. Replaced A1.06 with randomly selected A1.08 under the same K/A category.

Page 1 point totals not affected by this change. (Rev 2) RO T2/G2 245000 A4.07 245000 A4.02 Single/Sequential turbine governor valve operation is no longer used at CNS following high pressure turbine replacement during RE29. Because of this and the design of the DEH system, a discriminatory question could not be developed. Replaced A4.0 7 with randomly selected A4.02 under the same K/A category.

Page 1 point totals not affected by this change. (Rev 2) RO T2/G2 286000 G2.1.32 286000 G2.1.28 Because a discriminatory, operationally valid RO question could not be developed for fire protection system limits and cautions, replaced RO T2/G2 question 64 K/A 286000 G2.1.32 with randomly selected K/A 286000 G2.1.28. Page 1 point totals not affected by this change

.(Rev 4) Revision statement

(Rev 0 4) Replaced RO T2/G2 question 64 K/A 286000 G2.1.32 with K/A 286000 G2.1.28.

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Cooper Nuclear Station Date of Examination:

3/06/201 7 Examination Level:

Operating Test Number:

RO SRO 1 Administrative Topic (see Note)

Type Code* Describe activity to be performed Conduct of Operations R, D A1, Perform Jet Pump Operability Check (RO) 2.1.25 (3.9/4.2)

Conduct of Operations R, N A2, Perform SLC Operability Checks 2.1.20 (4.6/4.6)

Equipment Control R, D A3, Determine Isolation Boundaries (RHR) 2.2.13 (4.1/4.3)

Radiation Control R, N A4 , Determine Worker's Projected Total Dose 2.3.14 (3.4/3.8)

Emergency Plan NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, then all 5 are required.

• Type Codes & Criteria:

(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (< 3 for ROs;

< 4 for SROs & RO retakes) (2) (N)ew or (M)odified from bank (> 1) (2) (P)revious 2 exams (< 1; randomly selected) (0) ES-301, Page 22 of 27 Rev. 0 ES-301 Administrative Topics Outline Form ES-301-1 Facility: Cooper Nuclear Station Date of Examination:

3/06/201 7 Examination Level:

Operating Test Number:

RO SRO 1 Administrative Topic (see Note)

Type Code* Describe activity to be performed Conduct of Operations R, D A5 , Determine if Mode Change is Allowed 2.1.20 (4.6/4.6), 2.2.35 (3.6/4.5), 2.2.40 (3.4/4.7)

Conduct of Operations R, N A6 , Reportable Occurrences to the NRC (#8)

2.1.18 (3.6/3.8), 2.1.20 (4.6/4.6), 2.4.30 (2.7/4.1) Equipment Control R, M A7 , Review Jet Pump Operability and Recirc Pump Flow Check s 2.2.12 (3.7/4.1), 2.2.42 (3.9/4.6)

Radiation Control R, D A8 , Authorize Stable Iodine Thyroid Blocking

2.3.14 (3.4/3.8) Emergency Plan R, D A9 , Emergency Classification

2.4.41 (2.9/4.6) NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, then all 5 are required.

• Type Codes & Criteria:

(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (< 3 for ROs;

< 4 for SROs & RO retakes) (3) (N)ew or (M)odified from bank (> 1) (2) (P)revious 2 exams (< 1; randomly selected) (0) ES-301, Page 22 of 27 Rev. 0 Rev 1 Rev 1 ES-301 Control Room/In Plant Systems Outline Form ES-301-2 Facility: Cooper Nuclear Station Date of Examination:

03/06/201 7 Exam Level:

Operating Test N o.: RO SRO-I SRO-U 1 Control Room Systems

* (8 for RO); (7 for SRO

-I); 2 or 3 for SRO

-U System / JPM Title Type Code* Safety Function S1 - Secure SDG from Control Room 264000 A4.04 (3.7/3.7) L, N, S 6 S 2 - Start Torus Cooling from ASD Room 219000 A1.02 (3.5/3.5) L, D, S 5 S3 - Conduct Alt Pressure Control Using Reactor Feed Pumps 259001 A4.02 (3.9/3.7) L, D, S 3 S4 - Level Recover y During Shutdown Conditions Using LPCI 203000 A4.05 (4.3/4.1) A, EN, L, N, S 2 S5 - Perform 6.TG.303 Testing OPC Overspeed 241000 A3.12 (2.9/2.9), 241000 A4.19 (3.5/3.4)

L, N, S 4 S6 - Align REC System IAW 5.3EMPWR 400000 A4.01 (3.1/3.0)

L, N , S 8 S7 - Withdraw Control Rod From Position 00 201003 A2.01 (3.4/3.6) A, L, N, S 1 S8 - Verify Group 2 Isolation (TIP Shear) 223002 A4.01 (3.6/3.5), 223002 A4.06 (3.6/3.7)

A, D, S 7 In-Plant Systems

• (3 for RO); (3 for SRO

-I); (3 or 2 for SRO

-U) P1 - Locally Secur e Fire Pump C 286000 A4.05 (3.3/3.3), 286000 A1.05 (3.2/3.2)

A, N 8 P2 - Locally Align RHRSW Crosstie for RPV Injection 2.1.29 (4.1/4.0); 295031 EA1.08 (3.8/3.9)

D, E, L 4 P3 - (ASD) Locally Operate SW

-MO-89B for Torus Cooling 219000 A1.08 (3.7 / 3.6), 295016 AK2.02 (4.0/4.1)

D, E, L, R 5

• All RO and SRO-I control room (and in

-plant) systems must be different and serve different safety functions; all five SRO-U systems must serve different safety functions; in

-plant systems and functions may overlap those tested in the control room.

• Type Codes Criteria for RO

/ SRO-I / SRO-U (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in

-plant (EN)gineered safety feature (L)ow-Power / Shutdown (N)ew or (M)odified from bank including 1(A) (P)revious 2 exams (R)CA (S)imulator A 4-6 / 4-6 / 2-3 (4) C ----- D < 9 / < 8 / < 4 (5) E > 1 / > 1 / > 1 (2) EN > 1 / > 1 / > 1 (control room sys)

(1) L > 1 / > 1 / > 1 (9) N-M > 2 / > 2 / > 1 (6) P < 3 / < 3 / < 2 (randomly selected)

(0) R > 1 / > 1 / > 1 (1) S -----

Rev 1 Rev 1 ES-301 Control Room/In Plant Systems Outline Form ES-301-2 Facility: Cooper Nuclear Station Date of Examination:

03/06/201 7 Exam Level:

Operating Test N o.: RO SRO-I SRO-U 1 Control Room Systems

* (8 for RO); (7 for SRO-I); 2 or 3 for SRO

-U System / JPM Title Type Code* Safety Function S1 - Secure SDG from Control Room 264000 A4.04 (3.7/3.7) L, N, S 6 S 2 - Start Torus Cooling from ASD Room 219000 A1.02 (3.5/3.5) L, D, S 5 S4 - Level Recover y During Shutdown Conditions Using LPCI 203000 A4.05 (4.3/4.1) A, EN, L, N, S 2 S5 - Perform 6.TG.303 Testing OPC Overspeed 241000 A3.12 (2.9/2.9), 241000 A4.19 (3.5/3.4)

L, N, S 4 S6 - Align REC System IAW 5.3EMPWR 400000 A4.01 (3.1/3.0)

L, N , S 8 S7 - Withdraw Control Rod From Position 00 201003 A2.01 (3.4/3.6) A, L, N, S 1 S8 - Verify Group 2 Isolation (TIP Shear) 223002 A4.01 (3.6/3.5), 223002 A4.06 (3.6/3.7)

A, D, S 7 In-Plant Systems

• (3 for RO); (3 for SRO-I); (3 or 2 for SRO

-U) P1 - Locally Secure Fire Pump C 286000 A4.05 (3.3/3.3), 286000 A1.05 (3.2/3.2)

A, N 8 P2 - Locally Align RHRSW Crosstie for RPV Injection 2.1.29 (4.1/4.0); 295031 EA1.08 (3.8/3.9)

D, E, L 4 P3 - (ASD) Locally Operate SW

-MO-89B for Torus Cooling 219000 A1.08 (3.7 / 3.6), 295016 AK2.02 (4.0/4.1)

D, E, L, R 5

• All RO and SRO-I control room (and in

-plant) systems must be different and serve different safety functions; all five SRO-U systems must serve different safety functions; in

-plant systems and functions may overlap those tested in the control room.

• Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in

-plant (EN)gineered safety feature (L)ow-Power / Shutdown (N)ew or (M)odified from bank including 1(A) (P)revious 2 exams (R)CA (S)imulator A 4-6 / 4-6 / 2-3 (4) C ----- D < 9 / < 8 / < 4 (4) E > 1 / > 1 / > 1 (2) EN > 1 / > 1 / > 1 (control room sys)

(1) L > 1 / > 1 / > 1 (8) N-M > 2 / > 2 / > 1 (6) P < 3 / < 3 / < 2 (randomly selected)

(0) R > 1 / > 1 / > 1 (1) S -----

Rev 3 Rev 3 ES-301 Control Room/In Plant Systems Outline Form ES-301-2 Facility: Cooper Nuclear Station Date of Examination:

03/06/201 7 Exam Level:

Operating Test N o.: RO SRO-I SRO-U 1 Control Room Systems

* (8 for RO); (7 for SRO

-I); (2 or 3 for SRO

-U) System / JPM Title Type Code* Safety Function S1 - Secure SDG from Control Room 264000 A4.04 (3.7/3.7) L, N, S 6 S4 - Level Recover y During Shutdown Conditions Using LPCI 203000 A4.05 (4.3/4.1) A, EN, L, N, S 2 S7 - Withdraw Control Rod From Position 00 201003 A2.01 (3.4/3.6)

A, L, N, S 1 In-Plant Systems

• (3 for RO); (3 for SRO

-I); (3 or 2 for SRO

-U) P1 - Locally Secure Fire Pump C 286000 A4.05 (3.3/3.3), 286000 A1.05 (3.2/3.2)

A, N 8 P3 - (ASD) Locally Operate SW

-MO-89B for Torus Cooling 219000 A1.08 (3.7 / 3.6), 295016 AK2.02 (4.0/4.1)

D, E, L, R 5

• All RO and SRO

-I control room (and in

-plant) systems must be different and serve different safety functions; all five SRO-U systems must serve different safety functions; in

-plant systems and functions may overlap those tested in the control room.

• Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in

-plant (EN)gineered safety feature (L)ow-Power / Shutdown (N)ew or (M)odified from bank including 1(A) (P)revious 2 exams (R)CA (S)imulator A 4-6 / 4-6 / 2-3 (3) C ----- D < 9 / < 8 / < 4 (1) E > 1 / > 1 / > 1 (1) EN > 1 / > 1 / > 1 (control room sys)

(1) L > 1 / > 1 / > 1 (4) N-M > 2 / > 2 / > 1 (4) P < 3 / < 3 / < 2 (randomly selected)

(0) R > 1 / > 1 / > 1 (1) S -----

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 1 Page 1 of 44 Rev. 2 Facility: Cooper Nuclear Station Scenario No.: 1 Op-Test No.: CNS 15-01 Examiners: ____________________________ Operators: _____________________________ ____________________________ _____________________________ ____________________________ _____________________________ Objectives: evolutions: 1. Shift CRD Stabilizing valves. 2. Lower reactor power using RR pumps. 3. Respond to Reactor Bldg to Torus Vacuum Breaker PC-AO-243 failing open. 4. Respond to RPV flange leakage. 5. Respond to trip of RPS A EPAs with failure of RMV-AO-10 to close. 6. Respond to loss of multiple REC pumps. 7. ATWS Level Power control 8. Respond to RHR SPC valve failing to open. Initial Conditions: Plant operating at 100% power. Inoperable Equipment: HPCI inoperable. Auxiliary Oil pump motor replacement. TS LCO 3.5.1, Condition C Turnover: The plant is at 100% power. Planned activities for this shift are: Shift CRD Stabilizing valves per Procedure 2.2.8 (Rev. 95) Lower power to 95% with RR Pumps per Procedure 2.1.10 (Rev. 113) Electrical Maintenance working on replacing HPCI AOP motor Scenario Notes: This is a new scenario. Validation Time: 75 minutes Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 1 Page 2 of 44 Rev. 2 Event No. Malf. No. Event Type Event Description 1 N/A N (ATC,CRS) Shift CRD stabilizing valves 2 N/A R (ATC, CRS) Lower Reactor power by lowering RR pump speed. 3 (or) zdipcswcs243av[2] TS (CRS) Reactor Building to Torus vacuum breaker fails open. CRS declares vacuum breaker inoperable. 4 rr21 C (BOP,CRS) A (CREW) Respond to reactor vessel flange seal leak alarm, enter Procedure 4.6.3, and cycle the flange leak-off drain valves. 5 rp03a (rf) rh32a I (BOP,ATC,CRS) A (CREW) TS (CRS) RPS EPA Breaker 1A1/1A2 trip, (half scram and half PCIS group isolations) RMV-AO-10 fails to isolate. CRS declares valve inoperable. 6 sw 11a sw11b C (BOP, ATC) A(CREW) REC Pump A trip. Start another REC pump. REC Pump B trip. Manual scram due to loss of REC. 7 rd02a,b M (CREW) Hydraulic block ATWS > 3% power (EOP-1A, 3A, 6A, 6B, 7A) (CT-1) When control rods fail to scram, crew injects SLC and/or inserts control rods before exiting EOP-6A. (All control rods do not have to be fully inserted to satisfy this critical task; this only requires that the crew is making progress to achieving all rods in by fully inserting at least 5 control rods using RMCS.) (CT-2) Inhibit ADS prior to automatic ADS valve opening during a failure to Scram. (CT-3) During failure to scram conditions with power >3%, stop and prevent injection from all sources (except boron, CRD, RCIC) as necessary to lower RPV water level to below -between --prior to exiting EOP-7A. (CT-4) When control rods fail to scram and energy is discharging to the primary containment, crew injects SLC before exceeding the Boron Injection Initiation Temperature (BIIT) curve. 8 (rf) rh29(A) (rf) rh30(A) C (BOP,CRS) First RHR loop to be put into suppression pool cooling has RHR-MO-39A(B) fail to open. (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal (TS) Tech Spec

• Critical Task (As defined in NUREG 1021 Appendix D) CREW notation for Abnormal (A) and Major (M) events denotes ATC, BOP, and CRS are credited.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 1 Page 3 of 44 Rev. 2 Quantitative Attributes Table Attribute ES-304-1 Target Actual Description Malfunctions after EOP entry 1-2 1 RHR-MO-39A(B) fails to open. Abnormal Events 2-4 3 RPV flange seal leak RPS EPA trip Loss of multiple REC Pumps Major Transients 1-2 1 ATWS EOP entries requiring substantive action 1-2 2 EOP-3A EOP-6A EOP contingencies requiring substantive action 0-2 1 EOP- 7A EOP based Critical Tasks 2-3 4 (CT-1) When control rods fail to scram, crew injects SLC and/or inserts control rods before exiting EOP-6A. (All control rods do not have to be fully inserted to satisfy this critical task; this only requires that the crew is making progress to achieving all rods in by fully inserting at least 5 control rods using RMCS.) (CT-2) Inhibit ADS prior to automatic ADS valve opening during a failure to Scram. (CT-3) During failure to scram conditions with power >3%, stop and prevent injection from all sources (except boron, CRD, RCIC) as necessary to lower RPV water level to below -trol between --prior to exiting EOP-7A. (CT-4) When control rods fail to scram and energy is discharging to the primary containment crew injects SLC before exceeding the Boron Injection Initiation Temperature (BIIT) curve. Normal Events N/A 1 Shift CRD Stabilizing valves. Reactivity Manipulations N/A 1 Lower power using Reactor Recirculation pumps Instrument/ Component Failures N/A 6 RPV Flange leak Reactor Bldg to Torus vacuum breaker fails open RPS A EPA Breaker trip Loss of REC pump A Loss of REC pump B RHR-MO39A(B) valve fails to open Total Malfunctions N/A 6 RPV Flange leak Reactor Bldg to Torus vacuum breaker fails open RPS a EPA Breaker trip Loss of REC pump A Loss of REC pump B RHR-MO39A(B) valve fails to open Top 10 systems and operator actions important to risk that are tested:

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 1 Page 4 of 44 Rev. 2 Reactor Protection System (Event 5) Residual Heat Removal System in Suppression Pool Cooling Mode (Event 8) SCENARIO

SUMMARY

The plant is operating at 100% power. HPCI Auxiliary Oil Pump motor replacement is taking place. Event 1 After the crew takes the watch, the ATC shifts the CRD Stabilizing valves per Procedure 2.2.8. (Event 1) Event 2 After shifting stabilizing valves, the ATC lowers power ~5% per Load Dispatcher schedule. NOTE: Events 3 and 4 are triggered simultaneously due to Event 4 taking ~ 10 minutes to manifest itself. Event 3 (Triggered by Lead Examiner) After lowering power the Reactor to Torus vacuum breaker PC-AO-243 fails open. The CRS enters LCO 3.6.1.7, Condition A and declares the vacuum breaker inoperable. Event 4 (Triggered by Lead Examiner) The RPV inner seal develops a small leak requiring the BOP cycle the leak-off isolation valves from the control room to clear the alarm. Event 5 (Triggered by Lead Examiner) After actions for RPV flange leakage are complete, RPS EPAs on Division I trip causing a half reactor scram and half Group 1, 2, 7, and full Group 3, 6 isolations. RMV-AO-10 fails to isolate on the loss of RPS. The CRS enters LCO 3.6.1.3, Condition A for RMV-AO-10 failing to isolate and determines a potential LCO for TS 3.3.8.2 Condition A is required for the EPA breaker. LCO 3.3.8.2 entry is not required, since the EPA is no longer supplying RPS. Event 6 (Triggered by Lead Examiner) After RPS A power has been restored from the alternate supply and RRMG cooling restored, and the half scram is reset, REC Pump A trips requiring the BOP to start the standby pump per alarm procedures. Shortly after the standby pump is started, REC Pump B trips requiring entry into Emergency Procedure 5.2REC. The ATC will insert a reactor scram. The CRS will not have time to enter Technical Specifications for the REC pumps.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 1 Page 5 of 44 Rev. 2 Event 7 (No Trigger required) When the reactor is scrammed, a low power ATWS occurs due to hydraulic block of both scram discharge volumes, and EOP-6A and 7A are entered via EOP-1A. Reactor power is above 3%. The crew injects SLC and/or installs the necessary PTMs to bypass interlocks and insert control rods individually via RMCS (CT-1, CT-4). ADS is manually inhibited to prevent automatic operation (CT-2). Stop and Prevent is required because reactor power is above 3%. RPV level is intentionally lowered below -60 inches wide range in order to lower core inlet subcooling and lower reactor power (CT-3). Only 1 Main Turbine Bypass valve is available to control RPV pressure. SRVs have to be used to supplement pressure control. Feedwater injection is available for RPV level control. Event 8 (Automatically Triggered when opening the first MO39A(B) is attempted) After the crew has stabilized conditions following the scram, the selected RHR suppression pool cooling loop cannot be placed into service because RHR-MO39A (B) fails to open. The BOP transfers to the other division of RHR and places it into suppression pool cooling. After the scram has been reset twice, the control rods are allowed to be fully inserted with the next scram. The CRD transitions from ATWS to non-ATWS flowcharts, SLC injection is halted and RPV level restoration is directed. The exercise ends when control rods are inserted, and RPV water level is being maintained between -183 inches and +54 inches.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 1 Page 6 of 44 Rev. 2 Critical Tasks (CT-1) When control rods fail to scram, crew injects SLC and/or inserts control rods before exiting EOP-6A. (All control rods do not have to be fully inserted to satisfy this critical task; this only requires that the crew is making progress to achieving all rods in by fully inserting at least 5 control rods using RMCS.) (CT-2) Inhibit ADS prior to automatic ADS valve opening during a failure to Scram. EVENT 7 7 Safety significance Failure to effect shutdown of the reactor when a RPS setting has been exceeded would unnecessarily extend the level of degradation of the safety of the plant. This could further degrade into damage to the principle fission product barriers if left unmitigated. The crew is authorized and required by Conduct of Operations to take mitigating actions when automatic safety systems fail to perform their intended function. Action to shut down the reactor is required when RPS and control rod drive systems fail. With a Reactor Scram required, reactor not shut down, and conditions for ADS blowdown are met, INHIBIT ADS to prevent an uncontrolled RPV depressurization and cold water injection from low pressure sources to prevent causing a significant power excursion. Cueing Manual scram is initiated and numerous control rods indicate beyond position 00 and reactor power not downscale on panel 9-5 indications. ADS Timer initiated alarm on panel 9-3-1/A-1 Performance indicator Operator manipulates keylocked switches for SLC B pump to START on panel 9-5. Operator selects individual control rods by depressing the respective pushbutton on the panel 9-5 matrix and inserts the rod by manipulating the emergency in switch on panel 9-5. Manipulation of ADS A and ADS B Inhibit switches on panel 9-3 vertical section. Performance feedback SLC B pump red light illuminated, SLC discharge pressure rising, and SLC tank level lowering on panel 9-5. Operator selecting and inserting control rods indicated by rod position decreasing to 00 for selected rod on panel 9-5. Inhibit switches click into the vertical, inhibit position on panel 9-3. Receipt of ADS inhibited alarm panel 9-3-1/D-1. Justification for the chosen performance limit There is no time limit for effecting complete reactor shutdown via boron injection or control rod insertion. For the timeframe of this scenario, containment limits are not closely challenged and power oscillations are not experienced. However, if the failure to scram EOP were to be exited, other procedures would not provide the guidance necessary to achieve reactor shutdown. Before exiting EOP-6A ensures guidance to effect reactor shutdown is not removed. The 105 second ADS timer allows sufficient time for the crew to recognize and override automatic operation of the system. As long as ADS is inhibited before ADS valves open, reactor pressure will not be reduced to the shutoff heads of high volume, cold water systems. BWR Owners Group Appendix App. B, step RC/Q-6,RC/Q-7 App. B, step RC/Q-6 Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 1 Page 7 of 44 Rev. 2 Critical Tasks (CT-3) During failure to scram conditions with power >3%, stop and prevent injection from all sources (except boron, CRD, RCIC) as necessary to lower RPV water level to below -applicable) and control between -LL, as applicable) to -exiting EOP-7A. (CT-4) When control rods fail to scram and energy is discharging to the primary containment, crew injects SLC before exceeding the Boron Injection Initiation Temperature (BIIT) curve. EVENT 7 7 Safety significance Regarding lowering level below -prevent or mitigate the consequences of any large irregular neutron flux oscillations induced by neutronic/thermal-hydraulic instabilities, RPV water level is lowered sufficiently below the elevation of the feedwater sparger nozzles. This places the feedwater spargers in the steam space providing effective heating of the relatively cold feedwater and eliminating the potential for high core inlet subcooling. For conditions that are susceptible to oscillations, the initiation and growth of oscillations is principally dependent upon the subcooling at the core inlet; the greater the subcooling, the more likely oscillations will commence and increase in magnitude. 24" below the lowest nozzle in the feedwater sparger has been selected as the upper bound of the RPV water level control band. This water level is sufficiently low that steam heating of the injected water will be at least 65% to 75% effective (i.e., the temperature of the injected water will be increased to 65% to 75% of its equilibrium value in the steam environment). This water level is sufficiently high that the capability to bypass the low RPV water level MSIV isolation should be able to control RPV water level with feedwater pumps to preclude the isolation. combination of high reactor power (above the APRM downscale trip), high suppression pool temperature (above the Boron Injection Initiation Temperature), and an open SRV or high drywell pressure (above the scram setpoint) are symptomatic of heat being rejected to the suppression pool at a rate in excess of that which can be removed by the Suppression Pool Cooling System. Unless mitigated, these conditions ultimately result in loss of NPSH for ECCS pumps taking suction on the suppression pool, containment over-pressurization, and (ultimately) loss of primary containment integrity, which in turn could lead to a loss of adequate core cooling and uncontrolled release of radioactivity to the environment. The conditions listed, combined with the inability to shut down the Failure to effect shutdown of the reactor when a RPS setting has been exceeded would unnecessarily extend the level of degradation of the safety of the plant. This could further degrade into damage to the principle fission product barriers if left unmitigated. Action to shut down the reactor is required when RPS and control rod drive systems fail. The Boron Injection Initiation Temperature (BIIT) is the greater of: which initiation of boron injection will permit injection of the Hot Shutdown Boron Weight of boron before suppression pool temperature exceeds the Heat Capacity Temperature Limit. reactor scram is required by plant Technical Specifications. The BIIT is a function of reactor power. If boron injection is initiated before suppression pool temperature reaches the BIIT, emergency RPV depressurization may be precluded at lower reactor power levels. At higher reactor power levels, however, the suppression pool heatup rate may become so high that the Hot Shutdown Boron Weight of boron cannot be injected before suppression pool temperature reaches the Heat Capacity Temperature Limit even if boron injection is initiated early in the event. Since failure-to-scram conditions may present severe plant safety consequences, the requirement to initiate boron injection is independent of any anticipated success of control rod insertion. When attempts to insert control rods satisfactorily achieve reactor shutdown, the requirement for boron injection no longer exists. (Control rod insertion is directed under Step RC/Q-7 concurrently with Step RC/Q-6.)

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 1 Page 8 of 44 Rev. 2 reactor through control rod insertion, dictate a requirement to promptly reduce reactor power since, as long as these conditions exist, suppression pool heatup will continue. If torus water temperature was allowed to exceed the HCTL prior to commencing the lowering of level, a RPV depressurization would be required. Failure to completely stop RPV injection flow (with the exception of CRD and SLC) prolongs the elevated reactor power condition; thus, depositing more energy than necessary into the suppression pool. Cueing Manual scram is initiated and numerous control rods indicate beyond position 00 and reactor power >3% on panel 9-5 indications and SPDS and RPV level is >-SPDS. Manual scram is initiated and numerous control rods indicate beyond position 00 and reactor power not downscale on panel 9-5 indications. Suppression Pool temperature rising on panel 9-3 indication. Performance indicator Operator manipulates Feedwater HMIs on panel 9-5 or panel A as necessary to stop FW injection until RPV level goes below - Operator manipulates HPCI controls on panel 9-3 to stop HPCI injection until RPV level is below - Operator manipulates keylocked switches for SLC A(B) pump to START on panel 9-5. Performance feedback Feedwater flow indication on panel 9-5 indicate zero. HPCI flow indication on panel 9-3 indicates zero and/or HPCI injection MOV indicates closed. SLC A(B) pump red light illuminated, SLC discharge pressure rising, SLC tank level lowering on panel 9-5. Justification for the chosen performance limit Applicability for this CT is during EOP-7A conditions where it is necessary to lower level to control power with Table 17 condition NOT met (i.e. no high energy input into primary containment). There is no time limit for this lowering level, but it establishes margin to conditions where fuel damaging power oscillations may theoretically occur. Before exiting EOP-7A was chosen because other procedures would not provide the guidance necessary to establish margin for power oscillation mitigation. Before exiting EOP-7A ensures guidance to effect this control is not removed. NOTE This critical task must be evaluated carefully based on the level changes. If power is reduced significantly below 3%, reactor water level may continue to rise above -60" with only CRD and SLC while driving rods this would not result in an UNSAT on this critical task. If boron injection is initiated before suppression pool temperature reaches the BIIT, emergency RPV depressurization may be precluded at lower reactor power levels. At higher reactor power levels, however, the suppression pool heatup rate may become so high that the Hot Shutdown Boron Weight of boron cannot be injected before suppression pool temperature reaches the Heat Capacity Temperature Limit even if boron injection is initiated early in the event. Since failure-to-scram conditions may present severe plant safety consequences, the requirement to initiate boron injection is independent of any anticipated success of control rod insertion. BWR Owners Group Appendix App. B, Contingency #5 App. B, step RC/Q-6 Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 2 Page 1 of 37 Rev. 2 Facility: Cooper Nuclear Station Scenario No.: 2 Op-Test No.: CNS 15-01 Examiners: ____________________________ Operators: _____________________________ ____________________________ _____________________________ ____________________________ _____________________________ Objectives: evolutions: 1. Shift REC Pumps. 2. CRD FCV auto function fails requiring manual control. 3. One outboard MSIV fails closed. 4. Partial loss of main condenser vacuum requiring manual scram. 5. Electric ATWS 6. FW line break inside PC with loss of RCIC. 7. RWCU fails to auto isolate 8. Emergency Depressurize on low RPV level. 9. Low pressure injection valves fail to automatically open. Initial Conditions: Plant operating at 100% power. Inoperable Equipment: HPCI inoperable. Auxiliary Oil pump motor replacement. TS LCO 3.5.1, Condition C Turnover: The plant is at 100% power. Planned activities for this shift are: Maintain present power level. Electrical Maintenance working on installing HPCI AOP motor. Scenario Notes: This is a new scenario. Validation Time: 75 minutes Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 2 Page 2 of 37 Rev. 2 Event No. Malf. No. Event Type Event Description 1 N/A N (BOP, CRS) TS (CRS) Shift REC pumps 2 OR: zaicrdfc301 I (ATC,CRS) A (CREW) Auto function on CRD FCV fails causing manual control to be used 3 ms09e OR: zdipcissws4a C (ATC,BOP,CRS) A (CREW) TS (CRS) Outboard MSIV 86A closes but leaks by 4 mc01 C (ATC,CRS) A (CREW) Partial loss of condenser vacuum-scram 5 rp01 (a-d) OR: zdirpssws1 zdirpssws3a zdirpssws3b M(CREW) Electrical ATWS (CT-1 When RPS fails to scram the reactor on a manual scram signal, within two minutes initiate the ARI System. 6 fw18a rr20a rc02 M (CREW) FW A line break inside primary containment. RCIC spurious isolation (CT-2) When RPV level lowers to -cannot be maintained above -insufficient high pressure injection systems are available to restore level, crew begins to Emergency Depressurize by opening the first of six SRVs before RPV level lowers below - (Momentary shrink below -automatic SRV closure does not constitute failure of this critical task). 7 rp12 C (BOP,CRS) RWCU fails to automatically isolate. 8 cs02a cs02b rh04a rh04b C (ATC,BOP,CRS) ECCS system valves fail to auto open. (CT-3 )When operating injection systems cannot maintain RPV level and ECCS systems fail to automatically align for injection, crew manually aligns ECCS systems for injection: For low pressure ECCS systems, prior to RPV pressure lowering below 200 psig. (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal (TS) Tech Spec

• Critical Task (As defined in NUREG 1021 Appendix D) CREW notation for Abnormal (A) and Major (M) events denotes ATC, BOP, and CRS are credited.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 2 Page 3 of 37 Rev. 2 Quantitative Attributes Table Attribute ES-304-1 Target Actual Description Malfunctions after EOP entry 1-2 2 Low Pressure ECCS injection valves fail to open. RWCU fails to auto isolate Abnormal Events 2-4 2 Outboard MSIV closure Partial loss of condenser vacuum Major Transients 1-2 2 ATWS FW line break EOP entries requiring substantive action 1-2 2 EOP-1A EOP-3A EOP contingencies requiring substantive action 0-2 1 EOP-2A EOP based Critical Tasks 2-3 3 (CT-1) When RPS fails to scram the reactor on a manual scram signal, within two minutes initiate the ARI System. (CT-2) When operating injection systems cannot maintain RPV level and ECCS systems fail to automatically align for injection, crew manually aligns ECCS systems for injection: For low pressure ECCS systems, prior to RPV pressure lowering below 200 psig. (CT-3) When RPV level lowers to -(TAF) and cannot be maintained above -CFZ (MSCWL) and insufficient high pressure injection systems are available to restore level, crew begins to Emergency Depressurize by opening the first of six SRVs before RPV level lowers below - (Momentary shrink below -not constitute failure of this critical task). Normal Events N/A 1 Shift REC Pumps Reactivity Manipulations N/A 1 1 Instrument/ Component Failures N/A 5 CRD FCV controller failure Outboard MSIV closure Condenser in-leakage loss of vacuum RWCU fails to isolate. LP ECCS injection valves fail to open Total Malfunctions N/A 5 CRD FCV controller failure Outboard MSIV closure Condenser in-leakage rise RWCU fails to isolate. LP ECCS injection valves fail to open Top 10 systems and operator actions important to risk that are tested: Reactor Protection System (Event 5) ADS/SRV (Event 6) Residual Heat Removal System in LPCI Mode (Event 8) Operator fails to depressurize with SRVs (Event 6) Operator fails to initiate ADS and initiate ECCS early (Event 6)

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 2 Page 4 of 37 Rev. 2 SCENARIO

SUMMARY

The plant is operating at 100% power at the end of the operating cycle. HPCI Auxiliary Oil Pump motor is removed and a replacement is being installed. Event 1 After the crew takes the watch, the BOP operator shifts REC pumps by starting B and securing A. The CRS is required to declare REC Div I subsystem inoperable per LCO 3.7.3, Condition B, Event 2 (Triggered by Lead Examiner) After TS are addressed for the REC pump shift, the CRD FCV automatic setpoint fails downscale requiring the ATC to take manual control and return CRD cooling water flow and pressure to normal. Event 3 (Triggered by Lead Examiner) After the CRD system flows are returned to normal in manual, outboard MSIV 86A partially closes. The crew enters Abnormal Procedure 2.4MSIV and the RO rapidly lowers reactor power to <70%. The BOP places the effected MSIV control switch to CLOSE to prevent reopening. The CRS enters LCO 3.6.1.3, Condition A and declares the PCIV inoperable. Event 4 (Triggered by Lead Examiner) After TS are addressed for the partially closed MSIV, condenser in-leakage rises requiring reactor power to be lowered to maintain vacuum > 23 inches mercury. Condenser vacuum continues to lower requiring the reactor scram. Event 5 (No Trigger required) On the manual reactor scram, the crew recognizes the ATWS is an electric block ATWS. Manual ARI initiation successfully inserts the control rods (CT-1). Event 6 (Automatically triggered when RFP Discharge Valve automatically closes, ~2 minutes after ARI is initiated) After the control rods are inserted, Feedwater A line break inside the PC commences and the CRS enters EOP 3A. The torus and drywell are sprayed to control containment pressure and temperature. RPV water level continues to drop. RCIC will be unavailable due to a spurious isolation signal. Event 7 (No Trigger required) RWCU fails to isolate on low RPV level. Manual isolation from the control room is required.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 2 Page 5 of 37 Rev. 2 Event 8 (No Trigger required) RPV level lowers to TAF requiring the crew to emergency depressurize (CT-2). As RPV level and pressure lower, RHR injection valves fail to open and cannot be opened. The CS injection valves fail to open and can be opened from the control room (CT-3). The exercise ends when emergency depressurization is complete and RPV level restoration is being controlled.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 2 Page 6 of 37 Rev. 2 Critical Tasks (CT-1) When RPS fails to scram the reactor on a manual scram signal, within two minutes initiate the ARI System. (CT-2) When RPV level lowers to -(TAF) and cannot be maintained above -pressure injection systems are available to restore level, crew begins to Emergency Depressurize by opening the first of six SRVs before RPV level lowers below -CFZ. (Momentary shrink below -automatic SRV closure does not constitute failure of this critical task). EVENT 5 6 Safety significance RPS initiates a reactor scram when one or more monitored parameters exceed their specified limits to preserve the integrity of the fuel cladding and the reactor coolant pressure boundary (RCPB) and minimize the energy that must be absorbed following a loss of coolant accident (LOCA). Failure to effect shutdown of the reactor when a RPS setting has been exceeded, even at low power, would unnecessarily extend the level of degradation of the safety of the plant. This could further degrade into damage to the principle fission product barriers if left unmitigated. The crew is authorized and required by Conduct of Operations to take mitigating actions when automatic safety systems fail to perform their intended function. Action to shut down the reactor is required when RPS and control rod drive systems fail. The MSCWL is the lowest RPV water level at which the covered portion of the reactor core will generate sufficient steam to preclude any clad temperature in the uncovered portion of the core from exceeding 1500F. When water level decreases below MSCWL with injection, clad temperatures may exceed 1500F. Cueing 2/A-1 (A-2) RX SCRAM CHANNEL A (B) in alarm with RPS remaining energized. Corrected Fuel Zone indication (SPDS) falls to -before -field reports, and control room indications convey that adequate high pressure injection cannot be restored before level falls below - Performance indicator Operator depresses both manual scram pushbuttons, or places the Reactor Mode Switch to SHUTDOWN on panel 9-5. Manipulation of any six SRV controls on panel 9-3: SRV-71A SRV-71B SRV-71E SRV-71G SRV-71H SRV-71C SRV-71D SRV-71F Performance feedback RPS Group lights de-energized on panel 9-5. Control Rod full in indication on panel 9-5. Reactor power trend on nuclear instrumentation on panel 9-5. Crew will observe SRV light indication go from green to red, amber pressure switch lights illuminate, reactor pressure lowering on SPDS and panel 9-3 and 9-5 meters and recorders, and SRV tailpipe temperatures rise on recorder MS-TR-166. Justification for the chosen performance limit Procedure 2.0.3, Conduct of Operations requires upon recognition of a failure of automatic action, the CRO shall manually perform those actions necessary to fulfill the The MSCWL (-water level at which the covered portion of the reactor core will generate sufficient steam to preclude any clad temperature in the Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 2 Page 7 of 37 Rev. 2 safety function and report the completion of the manual action to the CRS as soon as possible. Failure of RPS to automatically function would involve multiple sensor and sensor relay failures. The complexity of an automatic RPS failure would necessarily require a short amount of time to diagnose and validate using control room indications. Two minutes is a reasonable time for operators to recognize a scram signal, verify the condition is valid, communicate conditions to the crew, and insert a manual scram, without unnecessarily extending the level of degradation to plant safety. uncovered portion of the core from exceeding 1500F. Emergency depressurization is allowed when level goes below TAF (-CFZ) and should be performed, if in the judgment of the CRS, level cannot be maintained above -intended for the scenario supporting this CT to, early in the event, clearly indicate no high pressure injection systems can be made available to reverse the lowering level trend, the crew will have time to communicate and open 6 SRVs before - BWR Owners Group Appendix App. B, step RC-1 App. B, Contingency#1 Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 2 Page 8 of 37 Rev. 2 Critical Tasks (CT-3) When operating injection systems cannot maintain RPV level and ECCS systems fail to automatically align for injection, crew manually aligns ECCS systems for injection: For low pressure ECCS systems, prior to RPV pressure lowering below 200 psig. EVENT 8 Safety significance Failure to recognize the auto valve alignment not occurring, and failure to take manual action per Conduct of Ops will result in unavailability of safety-related equipment necessary to provide adequate core cooling, otherwise resulting in core damage and a large offsite release. Cueing Indication ECCS valves are not opening with initiation conditions present: Green light on and Red lamp extinguished at respective injection handswitch on panel 9-3 or 9-4. - RPV pressure below injection valve open permissive setpoint Performance indicator Manipulation of controls as required to open the affected ECCS injection valve(s) or pump turbine controls from panel 9-3 or 9-4: Operator places affected ECCS injection valve(s) control switch(es) to OPEN on panel 9-3 or 9-4. Performance feedback Red light illuminates and Green light extinguishes for the affected ECCS injection valve(s), as applicable, on panel 9-3 or 9-4. RCIC or HPCI turbine speed and flow rate rises, as applicable, on panel 9-3 or 9-4. Justification for the chosen performance limit Attempting to align high pressure ECCS systems must be performed to determine their availability by the time TAF is reached in order to properly implement EOP-1A decision steps regarding restoring and maintaining RPV level. Attempting to align low pressure ECCS systems can only be done one RPV pressure falls below the injection valve RPV pressure permissive and will only be effective once RPV pressure falls below the shutoff head of the respective ECCS pump. The reduction in RPV pressure will normally be via Emergency Depressurization, which is a separate critical task bounded by a minimum RPV level. BWR Owners Group Appendix App. B, Contingency 1, step C1-1 SIMULATOR SET-UP Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 1 of 47 Rev. 2 Facility: Cooper Nuclear Station Scenario No.: 4 Op-Test No.: CNS 15-01 Examiners: ____________________________ Operators: _____________________________ ____________________________ _____________________________ ____________________________ _____________________________ Objectives: evolutions: 1. Shift RRMG oil pumps 2. Place RHR SP cooling in service 3. Respond to CRD pump trip 4. Respond to a RR Pump A #1 seal failure and subsequent pump trip 5. Respond to a RR Pump A #2 seal failure. Vent PC 6. Respond to a FW line break inside PC 7. Respond to failure of HPCI to automatically start 8. Respond to loss of RPV level indication, flood the RPV to the MSLs Initial Conditions: Plant is operating at 100% power Inoperable Equipment: None Turnover: The plant is operating at 100% power. Planned activities for this shift are: Shift RRMG oil pumps Place RHR SPC in service Maintain present power level Scenario Notes: This is a new scenario. Validation Time: 75 minutes Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 2 of 47 Rev. 2 Event No. Malf. No. Event Type Event Description 1 N/A N (ATC,CRS) Shift RRMG lube oil pumps 2 rf rh14 N (BOP,CRS) TS (CRS) Place RHR loop B in SPC, Min flow valve de-energizes open. 3 rd08b C (ATC,CRS) CRD Pump B trip. 4 rr10a C (BOP) A (CREW) TS (CRS) RR Pump A Seal #1 leak and RR Pump A trip. 5 rr04b rr11a C (BOP,ATC,CRS) A (CREW) RR Pump A Seal #2 leak, vent PC. 6 fw18b M (CREW) FW Line B break in PC-Scram (CT-1) Initiate drywell sprays when torus pressure exceeds 10 psig, prior to drywell temperature reaching 280F and prior to torus pressure exceeding the Pressure Suppression Pressure (PSP) curve. 7 hp01 C (BOP,CRS) HPCI fails to automatically start. 8 NBI various M (CREW) Loss of RPV level instruments, RPV flooding. (CT-2) When RPV level cannot be determined and torus level is above 6', open six SRVs before drywell radiation reaches 150 R/hr or entering PC Flooding. (CT-3) When RPV level cannot be determined and the reactor has been depressurized below the shutoff head of the respective pump(s), inject into the RPV to flood to the Main Steam Lines before drywell radiation reaches 150 R/hr or entering PC Flooding. (CT-4) When RPV level cannot be determined and at least 1 SRV is manually opened, isolate MSIVs, MSL drains, HPCI steam supply, and RCIC steam supply, prior to RPV water level rising to the bottom of the main steam lines. (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal (TS) Tech Spec

• Critical Task (As defined in NUREG 1021 Appendix D) CREW notation for Abnormal (A) and Major (M) events denotes ATC, BOP, and CRS are credited.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 3 of 47 Rev. 2 Quantitative Attributes Table Attribute ES-304-1 Target Actual Description Malfunctions after EOP entry 1-2 1 HPCI fails to automatically start. Abnormal Events 2-4 2 RR pump trip. RR seal leakage Major Transients 1-2 2 FW line break inside PC Loss of all RPV level instruments EOP entries requiring substantive action 1-2 2 EOP-1A EOP-3A EOP contingencies requiring substantive action 0-2 1 EOP- 2B EOP based Critical Tasks 2-3 4 (CT-1) Initiate drywell sprays when torus pressure exceeds 10 psig, prior to drywell temperature reaching 280F and prior to torus pressure exceeding the Pressure Suppression Pressure (PSP) curve. (CT-2) When RPV level cannot be determined and torus level is above 6', open six SRVs before drywell radiation reaches 150 R/hr or entering PC Flooding. (CT-3) When RPV level cannot be determined and the reactor has been depressurized below the shutoff head of the respective pump(s), inject into the RPV to flood to the Main Steam Lines before drywell radiation reaches 150 R/hr or entering PC Flooding. (CT-4) When RPV level cannot be determined and at least 1 SRV is manually opened, isolate MSIVs, MSL drains, HPCI steam supply, and RCIC steam supply, prior to RPV water level rising to the bottom of the main steam lines. Normal Events N/A 2 Shift RRMG oil pumps Place RHR Suppression Pool Cooling in service Reactivity Manipulations N/A 0 N/A Instrument/ Component Failures N/A 4 CRD Pump trip. RR Pump A Seal #1 failure with RR pump trip. RR Pump A Seal #2 failure. HPCI fails to automatically start Total Malfunctions N/A 4 CRD Pump trip. RR Pump A Seal #1 failure with RR pump trip. RR Pump A Seal #2 failure. HPCI fails to automatically start Top 10 systems and operator actions important to risk that are tested: Nuclear Boiler Instrumentation (Event 8) Residual Heat Removal in Containment Spray Mode (Event 6) HPCI (Event 7) ADS/SRV Operator fails to depressurize with SRVs Operator fails to initiate ADS and initiate ECCS early Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 4 of 47 Rev. 2 SCENARIO

SUMMARY

The plant is operating at 100% power. Event 1 After the crew takes the watch, the ATC shifts RRMG oil pumps B1 and B3 per procedure 2.2.68.1. The oil pump shift is in preparation for tagging out the oil pump later in the shift. Event 2 The BOP then places RHR in Suppression Pool Cooling in preparation a HPCI run the next shift. As the system's minimum flow valve starts to close it de-energizes in an intermediate position. The CRS declares the LPCI subsystem inoperable per LCO 3.5.1, Condition A. The valve is declared inoperable per LCO 3.6.2.3, Condition A. Event 3 (Triggered by Lead Examiner) After Technical Specifications are addressed for LPCI inoperable, the operating CRD pump trips requiring the ATC to start the standby pump. Event 4 (Triggered by Lead Examiner) After the CRD pump trip is addressed, RR Pump A develops a #1 seal failure. The crews responds to rising seal temperatures and lowers RR pump speed. Subsequently the RR pump trips, placing plant operation near the buffer region of the power to flow map. The CRS enters TS LCO 3.4.1. Event 5 (Triggered by Lead Examiner) After the RR pump trip is addressed, the pump's #2 seal develops a leak requiring the pump to be isolated and the PC to be vented with Standby Gas Treatment. Event 6 (Triggered by Lead Examiner) After the #2 seal failure is addressed, FW line B develops a leak inside PC. The reactor scrams on high drywell pressure. The crew initiates Torus and Drywell Sprays (CT-1). Event 7 (No Trigger required) HPCI fails to automatically start on high drywell pressure and must be started manually. Event 8 (Triggered by Lead Examiner) All RPV level instrumentation is lost and the crew emergency depressurizes (CT-2). Steam lines are isolated (CT-3) and the crew uses injection systems to flood the RPV to the bottom of the Main Steam Lines (CT-4).

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 5 of 47 Rev. 2 The exercise ends when emergency depressurization is complete and RPV level is maintained at the bottom of the MSLs.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 6 of 47 Rev. 2 Critical Tasks (CT-1) Initiate drywell sprays when torus pressure exceeds 10 psig, prior to drywell temperature reaching 280F and prior to torus pressure exceeding the Pressure Suppression Pressure (PSP) curve. (CT-2) When RPV level cannot be determined and torus level is above 6', open six SRVs before drywell radiation reaches 150 R/hr or entering PC Flooding. EVENT 6 8 Safety significance Drywell sprays are initiated in two legs of EOP-3A: Temperature and Pressure control. Regarding drywell temperature, if operation of all available drywell cooling is unable to terminate increasing drywell temperature before the structural design temperature limit initiated to affect the required drywell temperature reduction status of the DSIL and adequate core cooling permitting. Spray operation effects a drywell pressure and temperature reduction through the combined effects of evaporative cooling and convective cooling. Regarding drywell pressure, operation of drywell sprays reduces primary containment pressure by condensing any steam that may be present and by absorbing heat from the containment atmosphere through the combined effects of evaporative and convective cooling. Drywell sprays are initiated when torus pressure exceeds the Torus Spray Initiation Pressure (10# torus pressure) to preclude chugging the cyclic condensation of steam at the downcomer openings of the drywell vents. When a steam bubble collapses at the exit of the downcomers, the rush of water drawn into the downcomers to fill the void induces stresses at the junction of the downcomers and the vent header in Mark I containments and at the junction of the downcomers. Repeated application of such stresses could cause fatigue failure of these joints; thereby, creating a direct path between the drywell and torus. When drywell sprays are initiated, the resulting pressure reduction opens the vacuum breakers, drawing non-condensable from the torus back into the drywell. This condition defines the Torus Spray Initiation Pressure. As the drywell atmosphere is purged to the torus and replaced by steam, torus pressure increases. The SCSIP is the lowest torus pressure which can occur when 95% of the non-condensable in the drywell have been transferred to the torus. Since the failure mode is based on fatigue failure, a precise time limit or pressure cannot be provided. Therefore, prompt initiation of drywell sprays is required based on existing Depressurization of the RPV is necessary to perform the RPV flooding actions for the following reasons: The open SRVs establish a path from the RPV capable of rejecting energy in excess of decay heat to ensure the RPV flooding actions are successful. Reduced RPV pressure results in increased injection flow rates, reducing the total time required to flood the RPV. Reduced RPV pressure reduces the water inventory loss through non-isolable leaks and breaks. Dynamic loading on the SRVs and downstream piping is minimized as RPV water level reaches and is discharged through these valves. RPV depressurization can be most easily and rapidly accomplished by opening SRVs. The ADS valves are used first since they are the most reliable, considering component qualifications, pneumatic supply systems, initiation circuitry, and control power. In addition, the relative locations of the ADS valve discharges provide uniform distribution of the heat load around the suppression pool. manual action, even if the valves are already open on high pressure. Automatic valve operation in the relief or safety mode does not accomplish the objective of this step, even if low-low set logic has actuated. RPV flooding conditions are defined based on steam flow through the SRVs. Direct manual control must be established to ensure that the valves remain open as RPV pressure decreases. SRVs may be opened only if suppression pool water level is above the elevation of the top of the discharge devices. If the SRVs were opened with the discharge devices exposed, steam would pass directly into the suppression chamber airspace, bypassing the suppression pool. The resulting pressure increase could exceed the maximum pressure capability of the primary containment. Failing to depressurize could prevent recovery of RPV level above MSCRWL, resulting in core damage Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 7 of 47 Rev. 2 EOP priorities. Cueing Rising torus pressure indicated on SPDS and panel 9-3 recorder PC-LRPR-1A. Cursor approaching unsafe boundary on PSP graph display on SPDS. Erratic or inconsistent indication on all RPV level indications, and CRS declares RPV level cannot be determined. Performance indicator Aligns torus spray on panel 9-3 using RHR loop A and/or B: places CONTMT COOLING 2/3 CORE VALVE CONTROL PERMISSIVE switch to MANUAL OVERRD opens RHR-MO-39B, if closed closes close RHR-MO-27B, OUTBD INJECTION VLV, if necessary starts RHR PUMP(s), if not running For drywell spray, opens RHR-MO-31B Manipulation of any six SRV controls on panel 9-3: SRV-71A SRV-71B SRV-71E SRV-71G SRV-71H SRV-71C SRV-71D SRV-71F Performance feedback On panel 9-3, RHR pump/valve control switch light indication consistent with intended operation (Red open/running, Green closed/stopped). RHR flow rate rises on recorder RHR-FR-143 and indicator RHR-FI-133A(B.) Torus/drywell pressure stabilizes/lowers on SPDS and panel 9-3 recorder PC-LRPR-1A. Crew will observe SRV light indication go from green to red, amber pressure switch lights illuminate, reactor pressure lowering on SPDS and panel 9-3 and 9-5 meters and recorders, and SRV tailpipe temperatures rise on recorder MS-TR-166. Justification for the chosen performance limit When torus pressure cannot be maintained below PSP is the EOP-3A, step PC/P-4 criteria requiring transition to emergency depressurization. Before 150R/hr in the drywell was chosen because this is an indicator of loss of RPV level and the shielding effect of the water, indicating core exposure, yet it is much lower than the 2500R/hr trigger point during RPV Flooding that indicates gross cladding failure is in progress. Before exiting to PC Flooding was chosen because the design of the scenario provides the crew with the means to restore and maintain adequate core cooling IAW EOP-2B or 7B, and exiting to SAGs is neither required nor authorized. BWR Owners Group Appendix App. B, step PC/P-1. App. B, Contingency#4 Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 8 of 47 Rev. 2 Critical Tasks (CT-3) When RPV level cannot be determined and the reactor has been depressurized below the shutoff head of the respective pump(s), inject into the RPV to flood to the Main Steam Lines before drywell radiation reaches 150 R/hr or entering PC Flooding. (CT-4) When RPV level cannot be determined and at least 1 SRV is manually opened, isolate MSIVs, MSL drains, HPCI steam supply, and RCIC steam supply, prior to RPV water level rising to the bottom of the main steam lines. EVENT 8 8 Safety significance Once the SRVs have been opened to depressurize the RPV, injection systems are aligned to flood the RPV and establish core cooling by submergence. The list of flooding methods includes all motor-driven systems capable of injecting into the RPV. Any or all of these systems may be used, as necessary, to flood the RPV to the elevation of the main steam lines. Steam-driven systems are not listed since, with SRVs open and the reactor shut down, the RPV will depressurize to below the turbine stall pressures. Failing to raise RPV level to and observable point could prevent recovery of RPV level above MSCRWL, resulting in core damage. Steam lines connected to the RPV are isolated prior to initiating action to flood the RPV to preclude damage which may occur from cold water coming in contact with the hot metal, excessive loading of lines or hangers not designed to accommodate the weight of water, and flooding of steam driven equipment (RCIC turbine, main turbine, etc.). Isolation is performed, however, only if the status of SRVs assures the RPV will remain depressurized during the flooding evolution. For non-ATWS, only one SRV open is required to meet this condition. Cueing Erratic or inconsistent indication on all RPV level indications, and CRS declare RPV level undetermined. Six ADS valves have been manually opened. Erratic or inconsistent indication on all RPV level indications, and CRS declares RPV level cannot be determined, and SRVs have been manually opened IAW EOP-2B or EOP-7B for RPV depressurization. Performance indicator Crew establishes injection flow by manipulating controls as required to start the associated pumps and align system valves for injection using at least two pumps of the following systems: Main condensate/booster pumps on panel A RHR/LPCI loop A and/or B on panel 9-3 Core spray A and/or B [Operator places affected ECCS pump(s) control switch(es) to START and valve control switches to OPEN (or CLOSE, if necessary)] Crew places the following valve control switches to CLOSE: Inboard MSIVs on panel 9-3 MSL Drains on panel 9-4 HPCI steam supply on panel 9-3 RCIC steam supply on panel 9-4 Performance feedback Indication that the RPV is flooded to the main steam lines may include one or more of the following indication on panels 9-3, 9-4, 9-5 or field reports by the booth operator: report of two-phase flow conditions audible in the vicinity of the steam tunnel, main steam equalizing header, or main turbine stop and bypass valves PCI, RCIC or main steam line high flow logic Indication for applicable isolation valves Green light illuminates and Red light extinguishes.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 9 of 47 Rev. 2 RCIC turbine shaft seals alarms -open and stay open at RPV pressures below 50 psig above torus pressure sources are aligned with torus suction, torus water level: - decreases as RPV and steam lines are flooded - stabilizes when steam lines are full SRVs Justification for the chosen performance limit LOCA severity should result in a near linear RPV level reduction that gives the crew an initial trend on all level instruments. Failing all of the level instruments should occur within about 30 seconds and should yield inconsistent indications such that there is no doubt level cannot be determined (e.g. LOCA conditions with operation in the possible boiling region of the RPVST curve, minimal RPV injection, level slowly lowering to -CFZ, then all level instruments fail upscale within 10 seconds, simulating all reference legs flashing). Equipment damage due to cold water cannot occur until water level reaches the main steam lines. BWR Owners Group Appendix App. B, Contingency #4. App. B, Contingency#4, step C4-2.2