Text

2017-03 Proposed Outlines
	ML17122A082
Person / Time
Site:	Cooper
Issue date:	03/16/2017
From:	Vincent Gaddy; Operations Branch IV
To:	; Nebraska Public Power District (NPPD)
References
	Download: ML17122A082 (63)
	v • d • e

ES-401 BWR Examination Outline Form ES-401-1 Rev 2 Facility: Cooper Nuclear Station Date of Exam: March 2017 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 7

1 2

3 Tier Totals 4

4 6

5 4

4 27 5

5 10

2.

Plant Systems 1

2 2

3 4

3 2

2 2

2 26 2

3 5

2 2

1 1

12 0

2 1

3 Tier Totals 4

3 4

5 4

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

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 2 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)

AK3.01 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)

AA1.02 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)

AK1.03 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)

AK2.01 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)

AK3.07 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)

AA1.03 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)

AA2.02 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)

EK1.03 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 2 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)

EA1.01 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)

EA2.04 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)

EK1.02 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)

EK2.02 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)

AA1.05 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)

AA2.03 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)

AA2.04 Occurrence 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)

AA2.05 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 2 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 26 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 2 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 28 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)

A1.08 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)

A3.04 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)

K1.02 Nuclear boiler system 3.5 39

ES-401 6

Form ES-401-1 Rev 2 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)

K5.02 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)

A1.05 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)

A2.11 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)

A4.05 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)

K1.02 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)

K3.03 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 2 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

/

3 Group Point Total:

26

/5

ES-401 8

Form ES-401-1 Rev 2 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 55 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 Rod 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 59 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 Interlock 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 2 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

2.1.32 Ability to explain and apply system limits and precautions. (CFR: 41.10) 3.8 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 power 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

/

2 1

1 1

/

1 Group Point Total:

12

/3

Rev 2 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 2.1.37 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

Rev 2 Revision 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.

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

Randomly Selected K/A (New)

Reason for Rejection RO T1/G1 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 T2/G1 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 T2/G2 201004 Not Selected 201004 Not Applicable Because the Rod Sequence Control System is no longer used 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 201005 Not Selected 201005 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 question, 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 A4.08 with randomly selected A4.02 under the same K/A category. Page 1 point totals not affected by this change.

(Rev 2)

Revision statement:

Replaced RO T1/G1 K/A 295026 EK2.03 with EK2.06.

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

3/06/2017 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 Workers 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:

3/06/2017 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 Checks 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

Page 1 of 2 Rev 0 ES-301 Control Room/In Plant Systems Outline Form ES-301-2 Facility: Cooper Nuclear Station Date of Examination:

03/06/2017 Exam Level:

Operating Test No.:

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

a. JPM S1 - Secure SDG from Control Room 264000 A4.04 (3.7/3.7)

L, N, S 6

b. JPM S2 - Defeat RPS Logic Trips During Failure to Scram (5.8.3) (Restoration) 295037 EA1.01 (4.6/4.6)

L, D, S 7

c. JPM S3 - Conduct Alternate Pressure Control Using Reactor Feed Pumps 259001 A4.02 (3.9/3.7)

L, D, S 3

d. JPM S4 - Level Recovery During Shutdown Conditions Using LPCI (Alternate Path) 203000 A4.05 (4.3/4.1)

A, EN, L, N, S 2

e. JPM 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

f.

JPM S6 - Perform Standby Gas Treatment System Decay Heat Removal 261000 A3.04 (3.0/3.1), 261000 A4.03 (3.0/3.0)

D, S 9

g. JPM S7 - Withdrawal of Control Rod From Position 00 (Alternate Path 2) 201003 A2.01 (3.4/3.6)

A, L, N, S 1

h. JPM S8 - Verify Group 2 Isolation (Alt Path TIP Shear) 223002 A4.01 (3.6/3.5), 223002 A4.06 (3.6/3.7)

A, D, S 5

In-Plant Systems * (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

i.

JPM P1 - Securing Fire Pump C Locally (Alternate Path) 286000 A4.05 (3.3/3.3), 286000 A1.05 (3.2/3.2)

A, N 8

j.

JPM P2 - Place in Service/Remove From Service Hydrogen Cylinders/Tanks 245000 A3.08 (2.5/2.6)

A, D 4

k. JPM P3 - Alternate Shutdown, Locally Operate SW-MO-89B for starting 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.

Page 2 of 2 Rev 0

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 4-6 / 4-6 / 2-3 (5)

< 9 / < 8 / < 4 (6)

> 1 / > 1 / > 1 (1)

> 1 / > 1 / > 1 (control room system) (1)

> 1 / > 1 / > 1 (7)

> 2 / > 2 / > 1 (5)

< 3 / < 3 / < 2 (randomly selected) (0)

> 1 / > 1 / > 1 (1)

Page 1 of 2 Rev 0 ES-301 Control Room/In Plant Systems Outline Form ES-301-2 Facility: Cooper Nuclear Station Date of Examination:

03/06/2017 Exam Level:

Operating Test No.:

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

a. JPM S1 - Secure SDG from Control Room 264000 A4.04 (3.7/3.7)

L, N, S 6

b. JPM S2 - Defeat RPS Logic Trips During Failure to Scram (5.8.3) (Restoration) 295037 EA1.01 (4.6/4.6)

L, D, S 7

c.

d. JPM S4 - Level Recovery During Shutdown Conditions Using LPCI (Alternate Path) 203000 A4.05 (4.3/4.1)

A, EN, L, N, S 2

e. JPM 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

f.

JPM S6 - Perform Standby Gas Treatment System Decay Heat Removal 261000 A3.04 (3.0/3.1), 261000 A4.03 (3.0/3.0)

D, S 9

g. JPM S7 - Withdrawal of Control Rod From Position 00 (Alternate Path 2) 201003 A2.01 (3.4/3.6)

A, L, N, S 1

h. JPM S8 - Verify Group 2 Isolation (Alt Path TIP Shear) 223002 A4.01 (3.6/3.5), 223002 A4.06 (3.6/3.7)

A, D, S 5

In-Plant Systems * (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

i.

JPM P1 - Securing Fire Pump C Locally (Alternate Path) 286000 A4.05 (3.3/3.3), 286000 A1.05 (3.2/3.2)

A, N 8

j.

JPM P2 - Place in Service/Remove From Service Hydrogen Cylinders/Tanks 245000 A3.08 (2.5/2.6)

A, D 4

k. JPM P3 - Alternate Shutdown, Locally Operate SW-MO-89B for starting 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.

Page 2 of 2 Rev 0

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 4-6 / 4-6 / 2-3 (5)

< 9 / < 8 / < 4 (5)

> 1 / > 1 / > 1 (1)

> 1 / > 1 / > 1 (control room system) (1)

> 1 / > 1 / > 1 (6)

> 2 / > 2 / > 1 (5)

< 3 / < 3 / < 2 (randomly selected) (0)

> 1 / > 1 / > 1 (1)

Page 1 of 2 Rev 0 ES-301 Control Room/In Plant Systems Outline Form ES-301-2 Facility: Cooper Nuclear Station Date of Examination:

03/06/2017 Exam Level:

Operating Test No.:

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

a.

b. JPM S2 - Defeat RPS Logic Trips During Failure to Scram (5.8.3) (Restoration) 295037 EA1.01 (4.6/4.6)

L, D, S 7

c.

d. JPM S4 - Level Recovery During Shutdown Conditions Using LPCI (Alternate Path) 203000 A4.05 (4.3/4.1)

A, EN, L, N, S 2

e.

f.

g.

h.

In-Plant Systems * (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

i.

JPM P1 - Securing Fire Pump C Locally (Alternate Path) 286000 A4.05 (3.3/3.3), 286000 A1.05 (3.2/3.2)

A, N 8

j.

JPM P2 - Place in Service/Remove From Service Hydrogen Cylinders/Tanks 245000 A3.08 (2.5/2.6)

A, D 4

k. JPM P3 - Alternate Shutdown, Locally Operate SW-MO-89B for starting 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.

Page 2 of 2 Rev 0

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 4-6 / 4-6 / 2-3 (3)

< 9 / < 8 / < 4 (3)

> 1 / > 1 / > 1 (1)

> 1 / > 1 / > 1 (control room system) (1)

> 1 / > 1 / > 1 (3)

> 2 / > 2 / > 1 (2)

< 3 / < 3 / < 2 (randomly selected) (0)

> 1 / > 1 / > 1 (1)

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Cooper Nuclear Station Date of Exam: 03/06/2017 Operating Test No.:

A P

P L

I C

A N

T E

V E

N T

T Y

P E

Scenarios 1

2 3

4 T

O T

A L

M I

N I

M U

M(*)

CREW POSITION CREW POSITION CREW POSITION CREW POSITION S

R O

A T

C B

O P

S R

O A

T C

B O

P S

R O

A T

C B

O P

S R

O A

T C

B O

P R

I U

RO (R2,R4)

SRO-I SRO-U RX 1

0 1

1 1 0 NOR 1

1 2

1 1 1 I/C 2

3 5

4 4 2 MAJ 1

2 3

2 2 1 TS NA NA NA 0

2 2 RO (R1,R3)

SRO-I SRO-U RX 0

0 0

1 1 0 NOR 0

1 1

1 1 1 I/C 4

2 6

4 4 2 MAJ 1

2 3

2 2 1 TS NA NA NA 0

2 2 RO SRO-I SRO-U (U1,U2)

RX 1

0 1

1 1 0 NOR 1

2 3

1 1 1 I/C 4

4 8

4 4 2 MAJ 1

2 3

2 2 1 TS 2

2 4

0 2 2 RO SRO-I (I1,I4)

SRO-U RX 0

0 0

0 1

1 0 NOR 0

0 2

2 1

1 1 I/C 4

4 4

12 4

4 2 MAJ 1

2 2

5 2

2 1 TS NA NA 2

2 0

2 2 Instructions:

1.

Check the applicant level and enter the operating test number and Form ES-D-1 event numbers for each event type; TS are not applicable for RO applicants. ROs must serve in both the at-the-controls (ATC) and balance-of-plant (BOP) positions. Instant SROs (SRO-I) must serve in both the SRO and the ATC positions, including at least two instrument or component (I/C) malfunctions and one major transient, in the ATC position. If an SRO-I additionally serves in the BOP position, one I/C malfunction can be credited toward the two I/C malfunctions required for the ATC position.

2.

Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.5.d) but must be significant per Section C.2.a of Appendix D. (*) Reactivity and normal evolutions may be replaced with additional instrument or component malfunctions on a one-for-one basis.

3.

Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicants competence count toward the minimum requirements specified for the applicants license level in the right-hand columns.

4.

For licensees that use the ATC operator primarily for monitoring plant parameters, the chief examiner may place SRO-I applicants in either the ATC or BOP position to best evaluate the SRO-I in manipulating plant controls.

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Cooper Nuclear Station Date of Exam: 03/06/2017 Operating Test No.:

A P

P L

I C

A N

T E

V E

N T

T Y

P E

Scenarios 1

2 3

4 T

O T

A L

M I

N I

M U

M(*)

CREW POSITION CREW POSITION CREW POSITION CREW POSITION S

R O

A T

C B

O P

S R

O A

T C

B O

P S

R O

A T

C B

O P

S R

O A

T C

B O

P R

I U

RO SRO-I (I2,I5)

SRO-U RX 1

0 0

1 1

1 0 NOR 1

1 1

3 1

1 1 I/C 4

3 2

9 4

4 2 MAJ 1

2 2

5 2

2 1 TS 2

NA NA 2

0 2 2 RO SRO-I(I3,I6)

SRO-U RX 1

0 0

1 1

1 0 NOR 1

1 1

3 1

1 1 I/C 2

5 3

10 4

4 2 MAJ 1

2 2

5 2

2 1 TS NA 2

NA 2

0 2 2 RO SRO-I (I7)

SRO-U RX 0

1 1

1 1 0 NOR 1

0 1

1 1 1 I/C 5

2 7

4 4 2 MAJ 2

1 3

2 2 1 TS 2

NA 2

0 2 2 RO SRO-I (I8)

SRO-U RX 0

1 1

1 1 0 NOR 0

1 1

1 1 1 I/C 4

4 8

4 4 2 MAJ 2

1 3

2 2 1 TS NA 2

2 0

2 2 Instructions:

1.

Check the applicant level and enter the operating test number and Form ES-D-1 event numbers for each event type; TS are not applicable for RO applicants. ROs must serve in both the at-the-controls (ATC) and balance-of-plant (BOP) positions. Instant SROs (SRO-I) must serve in both the SRO and the ATC positions, including at least two instrument or component (I/C) malfunctions and one major transient, in the ATC position. If an SRO-I additionally serves in the BOP position, one I/C malfunction can be credited toward the two I/C malfunctions required for the ATC position.

2.

Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.5.d) but must be significant per Section C.2.a of Appendix D. (*) Reactivity and normal evolutions may be replaced with additional instrument or component malfunctions on a one-for-one basis.

3.

Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicants competence count toward the minimum requirements specified for the applicants license level in the right-hand columns.

4.

For licensees that use the ATC operator primarily for monitoring plant parameters, the chief examiner may place SRO-I applicants in either the ATC or BOP position to best evaluate the SRO-I in manipulating plant controls.

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Cooper Nuclear Station Date of Exam: 03/06/2017 Operating Test No.:

A P

P L

I C

A N

T E

V E

N T

T Y

P E

Scenarios 1

2 3

4 T

O T

A L

M I

N I

M U

M(*)

CREW POSITION CREW POSITION CREW POSITION CREW POSITION S

R O

A T

C B

O P

S R

O A

T C

B O

P S

R O

A T

C B

O P

S R

O A

T C

B O

P R

I U

RO SRO-I SRO-U(U3)

RX 0

1 1

1 1 0 NOR 1

1 2

1 1 1 I/C 5

4 9

4 4 2 MAJ 2

1 3

2 2 1 TS 2

2 4

0 2 2 RO (R5)

SRO-I SRO-U RX 0

0 0

1 1 0 NOR 0

1 1

1 1 1 I/C 4

3 7

4 4 2 MAJ 2

1 3

2 2 1 TS NA NA NA 0

2 2 RO (R6)

SRO-I SRO-U RX 0

1 1

1 1 0 NOR 1

0 1

1 1 1 I/C 3

2 5

4 4 2 MAJ 2

1 3

2 2 1 TS NA NA NA 0

2 2 RO SRO-I SRO-U RX 1

1 0 NOR 1

1 1 I/C 4

4 2 MAJ 2

2 1 TS 0

2 2 Instructions:

1.

Check the applicant level and enter the operating test number and Form ES-D-1 event numbers for each event type; TS are not applicable for RO applicants. ROs must serve in both the at-the-controls (ATC) and balance-of-plant (BOP) positions. Instant SROs (SRO-I) must serve in both the SRO and the ATC positions, including at least two instrument or component (I/C) malfunctions and one major transient, in the ATC position. If an SRO-I additionally serves in the BOP position, one I/C malfunction can be credited toward the two I/C malfunctions required for the ATC position.

2.

Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.5.d) but must be significant per Section C.2.a of Appendix D. (*) Reactivity and normal evolutions may be replaced with additional instrument or component malfunctions on a one-for-one basis.

3.

Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicants competence count toward the minimum requirements specified for the applicants license level in the right-hand columns.

4.

For licensees that use the ATC operator primarily for monitoring plant parameters, the chief examiner may place SRO-I applicants in either the ATC or BOP position to best evaluate the SRO-I in manipulating plant controls.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 1 Page 1 of 10

Revision 0

Facility: Cooper Nuclear Station Scenario No.: 1 Op-Test No.: CNS 15-01 Examiners: ____________________________ Operators:

Objectives: To evaluate the candidates ability to operate the facility in response to the following evolutions:

1. Shift CRD Stabilizing valves

2. Lower reactor power using RR pumps

3. Respond to Reactor to Torus Vacuum Breaker PC-AO 243 failing open

4. Respond to RPV flange leakage

5. Respond to trip of RPS 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 Turnover:

The plant is at 100% power.

Planned activities for this shift are:

Shift CRD Stabilizing valves per Procedure 2.2.8.

Lower power to 95% with RR Pumps per Procedure 2.1.10.

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 10

Revision 0

Event No.

Malf. No.

Event Type Event Description 1

N/A N (ATC)

Shift CRD stabilizing valves 2

N/A R (ATC)

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)

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)

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)

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.)

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 -60 CFZ (or LL, as applicable) and control between -60 (or LL, as applicable) to -183 CFZ prior to exiting EOP-7A.

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.

Inhibit ADS prior to automatic ADS valve opening during a failure to Scram.

8 (rf) rh29(A)

(rf) rh30(A)

C (BOP)

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)

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 1 Page 3 of 10

Revision 0

Quantitative Attributes Table Attribute ES-304-1 Target Actual Description Malfunctions after EOP entry 1-2 1

1.

RHR-MO-39A(B) fails to open.

Abnormal Events 2-4 3

1.

RPV flange seal leak

2.

RPS EPA trip 3.

Loss of multiple REC Pumps Major Transients 1-2 1

1.

ATWS EOP entries requiring substantive action 1-2 2

1.

EOP-6A

2.

EOP-7A EOP contingencies requiring substantive action 0-2 1

1.

EOP-7A EOP based Critical Tasks 2-3 4

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.)

2.

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.

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 -

60 CFZ (or LL, as applicable) and control between -60 (or LL, as applicable) to -183 CFZ prior to exiting EOP-7A.

4.

Inhibit ADS prior to automatic ADS valve opening during a failure to Scram.

Normal Events N/A 1

1.

Shift CRD Stabilizing valves.

Reactivity Manipulations N/A 1

1.

Lower power using Reactor Recirculation pumps Instrument/

Component Failures N/A 5

1.

RPV Flange leak

2.

RPS A EPA Breaker trip

3.

Loss of REC pump A

4.

Loss of REC pump B 5.

RHR-MO39A(B) valve fails to open Total Malfunctions N/A 5

1.

RPV Flange leak

2.

RPS A EPA Breaker trip

3.

Loss of REC pump A

4.

Loss of REC pump B 5.

RHR-MO39A(B) valve fails to open Top 10 systems and operator actions important to risk that are tested:

Reactor Protection System Residual Heat Removal System in Suppression Pool Cooling Mode SCENARIO

SUMMARY

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 1 Page 4 of 10

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The plant is operating at 100% power. HPCI Auxiliary Oil Pump motor replacement is taking place.

After the crew takes the watch, the ATC shifts the CRD Stabilizing valves per Procedure 2.2.8.

After shifting stabilizing valves, the ATC lowers power ~5% per Load Dispatcher schedule.

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.

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.

After actions for RPV flange leakage are complete, RPS EPAs on Division I trip causing a half reactor scram and half Group 1, 2, 3, 7, and full Group 6 isolations.

PC-AO-10 fails to isolate on the loss of RPS. The CRS determines a potential LCO 3.3.8.2 Condition A for the EPA breaker and enters LCO 3.6.1.3, Condition A for PC-AO-10 failing to isolate.

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. A reactor scram is required. The CRS will not have time to enter Technical Specifications for the REC pumps.

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. 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. 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.

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.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 1 Page 5 of 10

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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 10

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Critical Tasks 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.)

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 -60 CFZ (or LL, as applicable) and control between -60 (or LL, as applicable) to -183 CFZ prior to exiting EOP-7A.

EVENT 6

6 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.

Regarding lowering level below -60 CFZ, to 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.

Regarding lowering level below LL, the 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

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 1 Page 7 of 10

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with the inability to shut down the 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.

Maintaining RPV level above -183" (MSCWL) assures adequate core cooling via steam cooling with injection. This 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. Affiliated with this task is the requirement to only use outside the shroud injection systems due to the potential for a large power excursion that may otherwise result.

Cueing Manual scram is initiated and numerous control rods indicate beyond position 00 and reactor power not downscale on panel 9-5 indications.

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 >-60CFZ on SPDS.

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.

Operator manipulates Feedwater HMIs on panel 9-5 or panel A as necessary to stop FW injection until RPV level goes below -60CFZ.

Operator manipulates HPCI controls on panel 9-3 to stop HPCI injection until RPV level is below -60CFZ.

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.

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.

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.

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

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 1 Page 8 of 10

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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.

BWR Owners Group Appendix App. B, step RC/Q-6,RC/Q-7 App. B, Contingency #5

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 1 Page 9 of 10

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Critical Tasks Inhibit ADS prior to automatic ADS valve opening during a failure to Scram.

When control rods fail to scram and energy is discharging to the primary containment (e.g. SRVs, LOCA), crew injects SLC before exceeding the Boron Injection Initiation Temperature (BIIT) curve.

EVENT 6

6 Safety significance 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.

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:

The highest suppression pool temperature at 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.

The suppression pool temperature at which a 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.)

Cueing ADS Timer initiated alarm on panel 9-3-1/A-1 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.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 1 Page 10 of 10

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Performance indicator Manipulation of ADS A and ADS B Inhibit switches on panel 9-3 vertical section.

Operator manipulates keylocked switches for SLC A(B) pump to START on panel 9-5.

Performance feedback Inhibit switches click into the vertical, inhibit position on panel 9-3.

Receipt of ADS inhibited alarm panel 9 1/D-1.

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 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.

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, step RC/Q-6 App. B, step RC/Q-6

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 2 Page 1 of 7

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Facility: Cooper Nuclear Station Scenario No.: 2 Op-Test No.: CNS 15-01 Examiners: ____________________________ Operators:

Objectives: To evaluate the candidates ability to operate the facility in response to the following 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 injections valves fail to automatically open Initial Conditions: Plant operating at 100% power Inoperable Equipment: HPCI inoperable. Auxiliary Oil pump motor replacement.

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 7

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Event No.

Malf. No.

Event Type Event Description 1

N/A N (BOP)

TS (CRS)

Shift REC pumps 2

OR:

zaicrdfc301 I (ATC)

Auto function on CRD FCV controller fails requiring manual control to be used 3

ms09e OR:

zdipcissws4a C (BOP)

C (ATC)

A (CREW)

TS (CRS)

Outboard MSIV 86A closes but leaks by 4

mc01 C (ATC)

A (CREW)

Partial loss of condenser vacuum-scram 5

rp01 (a-d)

OR:

zdirpssws1 zdirpssws3a zdirpssws3b M(CREW)

Electrical ATWS 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 When RPV level lowers to -158 CFZ (TAF) and cannot be maintained above -183 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 -183 CFZ.)

7 rp12 C (BOP)

RWCU fails to automatically isolate.

8 cs02a cs02b rh04a rh04b C (BOP)

C (ATC)

ECCS system valves fail to auto open.

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)

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 2 Page 3 of 7

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Quantitative Attributes Table Attribute ES-304-1 Target Actual Description Malfunctions after EOP entry 1-2 2

1.

Low Pressure ECCS injection valves fail to open.

2.

RWCU fails to auto isolate Abnormal Events 2-4 2

1.

Outboard MSIV closure 2.

Partial loss of condenser vacuum Major Transients 1-2 2

1.

ATWS 2.

FW line break EOP entries requiring substantive action 1-2 2

1.

EOP-1A,

2.

EOP-3A EOP contingencies requiring substantive action 0-2 1

1.

EOP-2A EOP based Critical Tasks 2-3 3

1.

When RPS fails to scram the reactor on a manual scram signal, within two minutes initiate the ARI System.

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.

3.

When RPV level lowers to -158 CFZ (TAF) and cannot be maintained above -183 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 -183 CFZ.)

Normal Events N/A 1

1.

Shift REC Pumps Reactivity Manipulations N/A 1

1.

None Instrument/

Component Failures N/A 5

1.

CRD FCV controller failure

2.

Outboard MSIV closure

3.

Condenser in-leakage, loss of vacuum

4.

RWCU fails to isolate 5.

LP ECCS injection valves fail to open Total Malfunctions N/A 5

1.

CRD FCV controller failure

2.

Outboard MSIV closure

3.

Condenser in-leakage, loss of vacuum

4.

RWCU fails to isolate 5.

LP ECCS injection valves fail to open Top 10 systems and operator actions important to risk that are tested:

Reactor Protection System ADS/SRV Residual Heat Removal System in LPCI Mode 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 2 Page 4 of 7

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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.

After the crew takes the watch, the BOP operator shifts REC pumps by starting B and securing A. The CRS is required to either declare REC Div I subsystem inoperable per LCO 3.7.3, Condition B, or DG1 inoperable per LCO 3.8.1, Condition B, depending upon how the REC NORMAL/STANDBY selector switches are manipulated.

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.

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.

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.

On the manual reactor scram, the crew recognizes the ATWS is an electric block ATWS. Manual ARI initiation successfully inserts the control rods.

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 and RWCU fails to isolate on low RPV level. Manual isolation from the control room is required.

RPV level lowers to TAF requiring the crew to emergency depressurize. 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.

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 5 of 7

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Critical Tasks When RPS fails to scram the reactor on a manual scram signal, within two minutes initiate the ARI System.

When RPV level lowers to -158 CFZ (TAF) and cannot be maintained above -183 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 -183 CFZ.

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 Annunciators 9-5-2/A-1 (A-2) RX SCRAM CHANNEL A (B) in alarm with RPS remaining energized.

Corrected Fuel Zone indication (SPDS) falls to

-158 and lowering trend continues, and, before -158 CFZ is reached, initial conditions, field reports, and control room indications convey that adequate high pressure injection cannot be restored before level falls below -

183 CFZ.

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 There is no time limit for effecting complete The MSCWL (-183 CFZ) is the lowest RPV water level at which the covered portion of the reactor core will generate sufficient steam to

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 2 Page 6 of 7

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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.

preclude any clad temperature in the uncovered portion of the core from exceeding 1500F. Emergency depressurization is allowed when level goes below TAF (-158 CFZ) and should be performed, if in the judgment of the CRS, level cannot be maintained above -183 CFZ. Since it is 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 -183 CFZ.

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 7 of 7

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Critical Tasks 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.

Indication of Drywell Pressure 1.83 psig Indication of RPV water level -113 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

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 3 Page 1 of 7

Revision 0 Facility: Cooper Nuclear Station Scenario No.: 3 Op-Test No.: CNS 15-01 Examiners: ____________________________ Operators:

Objectives: To evaluate the candidates ability to operate the facility in response to the following evolutions:

1. Secure Standby Gas Treatment after nitrogen makeup to the drywell and torus.

2. Raise reactor power with Reactor Recirculation to 95%.

3. Respond to Battery Room Exhaust Fans failing.

4. Respond to a stuck open SRV.

5. Respond to TG bearing 9 high vibration.

6. Respond to a leak in the torus.

7. Respond to failure of 6 SRVs to open when emergency depressurizing, use of alternate ED systems.

Initial Conditions: Plant operating at 100%power.

Inoperable Equipment: None Turnover:

The plant is at 90% power.

Planned activities for this shift are:

Secure SGT A from nitrogen purge operation.

Raise power to 95% with Reactor Recirculation.

Scenario Notes:

This is a new scenario.

Validation Time: 75 minutes

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 3 Page 2 of 7

Revision 0 Event No.

Malf. No.

Event Type Event Description 1

N/A N

(BOP)

Secure SGT after nitrogen makeup to the torus and drywell.

2 N/A R (ATC)

Raise power to 95% with Reactor Recirculation 3

OR ZDIHVSWEFCI C[1]

ZDIHVSWEFCI A[1]

C (BOP)

TS (CRS)

Running Battery Room Exhaust Fan trip, failure of standby fan to run, manual start of Essential Ventilation.

4 ad06c C

(BOP-ATC)

A (CREW)

TS (CRS)

SRV fails open.

When a SRV fails open, close the SRV or prior to torus bulk temperature reaching 110F, initiate a Reactor Scram 5

tu3i C (ATC)

A (CREW)

Main Turbine Bearing #9 high vibration.

6 pc08 M

(CREW)

Torus water leak-Emergency Depressurization When torus water level cannot be maintained above 11',

prevent HPCI operation prior to torus water level lowering below 11.0.

When torus water level cannot be maintained above 9.6',

scram the reactor prior to torus water level falling below 9.6.

When torus water level cannot be maintained above 9.6',

crew Emergency Depressurizes by opening 2 SRVs prior to torus water level falling below 9.6.

7 OR zdimssws1a zdimssws1b zdimssws1a zdimssws4a zdimssws1d zdimssws1e C

(BOP)

Only 2 SRVs open on emergency depressurization use of alternate ED systems.

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal (TS) Tech Spec Critical Task (As defined in NUREG 1021 Appendix D)

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 3 Page 3 of 7

Revision 0 Quantitative Attributes Table Attribute ES-304-1 Target Actual Description Malfunctions after EOP entry 1-2 1

1.

All but 2 SRVs fail to open on emergency depressurization Abnormal Events 2-4 2

1.

SRV fails open.

2.

MT Bearing #9 Hi Vibs Major Transients 1-2 1

1.

Torus water leak EOP entries requiring substantive action 1-2 2

1.

EOP-1A

2.

EOP-3A EOP contingencies requiring substantive action 0-2 1

1.

EOP-2A EOP based Critical Tasks 2-3 4

1.

When a SRV fails open, close the SRV or prior to torus bulk temperature reaching 110F, initiate a Reactor Scram

2.

When torus water level cannot be maintained above 11', prevent HPCI operation prior to torus water level lowering below 11.0.

3.

When torus water level cannot be maintained above 9.6', scram the reactor prior to torus water level falling below 9.6 prior to torus water level lowering below 9.6.

4.

When torus water level cannot be maintained above 9.6', crew Emergency Depressurizes by opening 2 SRVs prior to torus water level falling below 9.6.

Normal Events N/A 1

1.

Secure SGT after nitrogen makeup evolution Reactivity Manipulations N/A 1

1.

Raise power 5% with Reactor Recirculation Instrument/

Component Failures N/A 4

1.

SRV fails open.

2.

Battery Room Exhaust fans fail.

3.

MT Bearing #9 Hi Vibs 4.

6 SRVs fail to open on emergency depressurization Total Malfunctions N/A 4

1.

SRV sticks open

2.

Battery Room Exhaust fans fail.

3.

MT Bearing #9 Hi Vibs 4.

6 SRVs fail to open on emergency depressurization Top 10 systems and operator actions important to risk that are tested:

ADS/SRV Operator fails to depressurize with SRVs

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 3 Page 4 of 7

Revision 0 SCENARIO

SUMMARY

The plant is starting up at 90% power.

After the crew takes the watch, Standby Gas Treatment fan is secured following nitrogen makeup evolution.

After SGT is secured, reactor power is raised 5% using Reactor Recirculation pumps per Procedure 2.1.10.

The running Battery Room Exhaust Fan trips and the standby fan cannot be started.

The Essential Control Building Ventilation system is required to be placed into service.

The CRS determines TLCO 3.8.1 is not met and declares both Battery Room Exhaust Fans inoperable.

After the TRM is addressed for the Battery Room Exhaust fans, SRV C sticks open.

The crew enters the abnormal procedure 2.4SRV, lowers power to below 90% with Reactor Recirculation, and inhibits ADS. The valve closes. The CRS declares the valve inoperable per LCO 3.5.1. Condition E.

The Main Turbine bearing #9 develops high vibrations requiring the crew to lower reactor power with Reactor Recirculation flow and control rod insertion. The power drop lowers bearing vibrations.

After the Main Turbine #9 bearing high vibration is addressed, the torus develops a water leak on the bottom beyond makeup capability. The reactor is scrammed and emergency depressurization performed once torus level nears the bottom of the downcomers. Only 2 SRVs can be open, and alternate ED systems are used.

The exercise ends when emergency depressurization is complete and RPV level recovery is under control.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 3 Page 5 of 7

Revision 0 Critical Tasks When a SRV fails open, close the SRV or prior to torus bulk temperature reaching 110F, initiate a Reactor Scram.

When torus water level cannot be maintained above 11', prevent HPCI operation prior to torus water level lowering below 11.0.

EVENT 4

5 Safety significance Closing the SRV or shutting down the reactor before 110°F in the Suppression Pool ensures containment design limits due to heat addition to the suppression pool will not be exceeded. 110°F in the Suppression Pool is both the Technical Specification limit and EOP-3A limit for effecting a reactor scram.

Tech Spec 3.6.2.1 requires that the Reactor Scram be inserted at 110F. This requirement ensures that the unit will be shut down at > 110F. The pool is designed to absorb decay heat and sensible heat but could be heated beyond design limits by the steam generated if the reactor is not shut down (TS Basis). Per PSTGs, the lowest temperature of the Boron Injection Initiation Temperature (BIIT) is specified as the action level (110°F). A single value instead of a graph implements the BIIT in this step to simplify the guideline. The BIIT specifies the suppression pool temperature before which boron injection must be started. It is the greater of:

The highest suppression pool temperature at 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.

The suppression pool temperature at which a reactor scram is required by plant Technical Specifications.

The BIIT is a function of reactor power. It is utilized to establish a requirement for boron injection following a failure-to-scram. 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 HSBW cannot be injected before suppression pool temperature reaches the Heat Capacity Temperature Limit even if boron injection is initiated early in the event. Entering the RPV Control guideline at Step RC-1 ensures that, if possible, the reactor is scrammed before boron injection is required and in anticipation of possible RPV depressurization in Step SP/T-3.

Operation of the HPCI System with its exhaust discharge device not submerged will directly pressurize the torus. HPCI operation is therefore secured, as required, to preclude the occurrence of this condition. The consequences of not doing so may extend to failure of the primary containment from over-pressurization, and thus, HPCI must be secured irrespective of adequate core cooling concerns.

No comparable task regarding RCIC operation is provided because:

The exhaust flow rate of RCIC is no greater than the steam generated by decay heat after reactor shutdown. The basis for determining Primary Containment Pressure Limit assumes the operability of a containment vent capable of removing decay heat 10 minutes after reactor shutdown. Thus, any steam discharged by RCIC into the torus airspace can be removed through the primary containment vent and will not cause torus pressure to exceed PCPL even if the RCIC exhaust is not submerged.

Elevated torus pressure will cause the RCIC turbine to trip much sooner than the HPCI turbine.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 3 Page 6 of 7

Revision 0 Cueing SRV open indications (solenoid lights, tailpipe pressure light, tailpipe temperature).

Step reduction in turbine generator load and steam flow.

Rising suppression pool temperatures on panel 9-3.

Lowering Torus water level, approaching 11, as indicated on SPDS and panel 9-3 indicators PC-LRPR-1A and PC-LI-10.

Performance indicator Operator depresses both manual scram pushbuttons or places the Reactor Mode Switch to SHUTDOWN on panel 9-5, prior to exceeding 110°F in the Suppression Pool; or the operator closes the SRV IAW 2.4SRV without exceeding 110°F in the Suppression Pool.

Crew stops and prevents HPCI by one of the following on panel 9-3:

Depressing and holding the HPCI trip pushbutton, and placing HPCI Aux Oil Pump control switch in PTL

Placing HPCI-MO-16, STM SUPP OUTBD ISOL VLV control switch in CLOSE

Depressing HPCI MANUAL ISOLATION pushbutton, if initiation signal present Performance feedback RPS Group lights de-energized on panel 9-5.

Reactor Power trend.

Control Rod full-in indication.

SRV tailpipe pressure, steam flow, solenoid lights, step increase in turbine generator load and steam flow Anytime when a SRV fails open and the actions addressed in Procedure 2.4SRV would be effective in closing the valve OR EOP-3A conditions when actions taken IAW 2.4SRV are ineffective or not attempted.

HPCI speed lowers to zero on HPCI-SI-2792 on panel 9-3 HPCI flow lowers to zero on HPCI-FIC-108 on panel 9-3 Steam supply isolation valve HPCI-MO-15 and/or HPCI-MO-16 control switch green light illuminated and red light extinguished on panel 9-3.

Justification for the chosen performance limit 110ºF is both the EOP-3A step SP/T-2 limit and the TS 3.6.2.1 limit for reactor shutdown to limit heat addition to the suppression pool.

Closing the failed open SRV would also terminate heat addition to the suppression pool.

If torus water level cannot be restored and maintained above 11 feet is the EOP-3A, step SP/L-10 criteria for preventing HPCI operation to ensure HPCI exhaust does not directly impinge on the torus air space.

BWR Owners Group Appendix App. B, step SP/T-2 App. B, step SP/L-2.2 Critical Tasks When torus water level cannot be maintained above 9.6', scram the reactor prior to torus water level falling below 9.6 prior to torus water level lowering below 9.60.

When torus water level cannot be maintained above 9.6', crew Emergency Depressurizes by opening 2 SRVs prior to torus water level falling below 9.6.

EVENT 5

5

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 3 Page 7 of 7

Revision 0 Safety significance Energy in the RPV should be discharged outside the primary containment, if possible, and thereby reduce or limit the energy added to the suppression pool if emergency RPV depressurization becomes necessary. Entry to the RPV Control guideline is therefore specified before reaching the elevation of the downcomer openings so that the override before Step RC/P-1 can be used to anticipate emergency RPV depressurization and rapidly depressurize the RPV, irrespective of the resulting cooldown rate.

Entering the EOP-1A at Step 1 assures that, if possible, the reactor is scrammed and shutdown is assured by control rod insertion before RPV depressurization is initiated.

Entry into the EOP-1A must be explicitly stated because conditions requiring entry into the EOP-3A do not necessarily require entry into EOP-1A. Therefore, a scram may not yet have been initiated. Directing that EOP-1A be entered, rather than explicitly stating here "Initiate a Reactor Scram", coordinates actions currently being executed if the EOP-1A has already been entered. In addition, entry to EOP-1A must be made because it is through EOP-1A that EOP-2A "Emergency RPV Depressurization", is performed.

The RPV is not permitted to remain at pressure if suppression of steam discharged from the RPV into the drywell cannot be assured. When the downcomer vent openings are not adequately submerged, any steam discharged from the RPV into the drywell may not condense in the suppression pool before torus pressure reaches unacceptable levels.

RPV depressurization is required at or before the point at which this low water level condition occurs. This reduces the amount of energy that may be discharged directly to the torus air space to as low as possible.

Cueing Lowering Torus water level, approaching 9.6, as indicated on SPDS and panel 9-3 indicators PC-LRPR-1A and PC-LI-10.

Lowering Torus water level, approaching 9.6, as indicated on SPDS and panel 9-3 indicators PC-LRPR-1A and PC-LI-10.

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 Before torus water level drops to 9.6 is the EOP-3A, step SP/L-11 criteria for transitioning to EOP-1A to shut down the reactor.

Inability to maintain torus water level above 9.6 is the EOP-3A, step SP/L-12 criteria for transitioning to emergency depressurization.

BWR Owners Group Appendix App. B, step SP/L-2.1 App. B, step SP/L-2.1

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 1 of 9

Revision 0 Facility: Cooper Nuclear Station Scenario No.: 4 Op-Test No.: CNS 15-01 Examiners: ____________________________ Operators:

Objectives: To evaluate the candidates ability to operate the facility in response to the following 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 #1 seal failure.

5. Respond to a RR pump #2 seal failure and pump trip. 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 operating at 100%power.

Inoperable Equipment: None Turnover:

The plant is 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 9

Revision 0 Event No.

Malf. No.

Event Type Event Description 1

N/A N(ATC)

Shift RRMG lube oil pumps 2

rf rh14 N (BOP)

TS (CRS)

Place RHR loop B in SPC, Min flow valve de-energizes open.

3 rd08b C (ATC)

CRD Pump B trip.

4 rr10a C (BOP)

A (CREW)

RR Pump A seal #1 leak and RR Pump A trip.

5 rr04b rr11a C (BOP, ATC)

A (CREW)

TS (CRS)

RR Pump A seal #2 leak, vent PC.

6 fw18b M

(CREW)

FW Line B break in PC-Scram 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)

HPCI fails to automatically start.

8 NBI various M

(CREW)

Loss of RPV level instruments, RPV flooding.

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.

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.

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)

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 3 of 9

Revision 0 Quantitative Attributes Table Attribute ES-304-1 Target Actual Description Malfunctions after EOP entry 1-2 1

1.

HPCI fails to automatically start.

Abnormal Events 2-4 2

1.

RR pump trip.

2.

RR seal leakage Major Transients 1-2 2

1.

FW line break inside PC 2.

Loss of all RPV level instruments EOP entries requiring substantive action 1-2 2

1.

EOP-1A

2.

EOP-3A EOP contingencies requiring substantive action 0-2 1

1.

EOP-2B EOP based Critical Tasks 2-3 4

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.

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.

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.

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

1.

Shift RRMG oil pumps 2.

Place RHR Suppression Pool Cooling in service Reactivity Manipulations N/A 0

N/A Instrument/

Component Failures N/A 4

1.

CRD Pump trip.

2.

RR pump A seal #1 with RR pump trip.

3.

RR pump A seal #2 failure.

4.

HPCI fails to automatically start Total Malfunctions N/A 4

1.

CRD Pump trip.

2.

RR pump A seal #1 with RR pump trip.

3.

RR pump A seal #2 failure.

4.

HPCI fails to automatically start Top 10 systems and operator actions important to risk that are tested:

Nuclear Boiler Instrumentation Residual Heat Removal in Containment Spray Mode HPCI 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 9

Revision 0 SCENARIO

SUMMARY

The plant is at 100% power.

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.

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.

After Technical Specifications are addressed for LPCI inoperable, the operating CRD pump trips requiring the ATC to start the standby pump.

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.

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.

After the #2 seal failure is addressed, FW line B develops a leak inside PC. The reactor scrams on high drywell pressure.

HPCI fails to automatically start and must be started manually.

All RPV level instrumentation is lost and the crew emergency depressurizes and floods the RPV to the bottom of the steam lines.

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 5 of 9

Revision 0 Critical Tasks 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.

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.

EVENT 7

7 Safety significance 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.

The direction to open all ADS valves requires 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.

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.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 6 of 9

Revision 0 Failing to depressurize could prevent recovery of RPV level above MSCRWL, resulting in core damage Cueing Erratic or inconsistent indication on all RPV level indications, and CRS declares RPV level cannot be determined.

Erratic or inconsistent indication on all RPV level indications, and CRS declare RPV level undetermined.

Six ADS valves have been manually opened.

Performance indicator 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 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)]

Performance feedback 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.

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:

Rising RPV pressure

If a main steam line is not isolated, field 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

Actuation of HPCI, RCIC or main steam line high flow logic

Field report of water leakage from HPCI or RCIC turbine shaft seals

HPCI/RCIC STM DRAIN POT LEVEL HI alarms

SRVs re-open and stay open at RPV pressures below 50 psig above torus pressure

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 7 of 9

Revision 0

If injection sources are aligned with torus suction, torus water level:

- decreases as RPV and steam lines are flooded

- stabilizes when steam lines are full

Local torus water temperatures near open SRVs Justification for the chosen performance limit 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.

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 -100 CFZ, then all level instruments fail upscale within 10 seconds, simulating all reference legs flashing). The scenario should also be validated to provide clear, consistent indication when the RPV has been flooded to the MSLs.

BWR Owners Group Appendix App. B, Contingency#4 App. B, Contingency #4.

Critical Tasks 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.

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.

EVENT 7

5 Safety significance 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.

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 of 280ºF is reached, drywell sprays are 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

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 8 of 9

Revision 0 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 EOP priorities.

Cueing 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.

Rising torus pressure indicated on SPDS and panel 9-3 recorder PC-LRPR-1A.

Cursor approaching unsafe boundary on PSP graph display on SPDS.

Performance indicator 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 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 Performance feedback Indication for applicable isolation valves Green light illuminates and Red light extinguishes.

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.)

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 9 of 9

Revision 0 Torus/drywell pressure stabilizes/lowers on SPDS and panel 9-3 recorder PC-LRPR-1A.

Justification for the chosen performance limit Equipment damage due to cold water cannot occur until water level reaches the main steam lines.

When torus pressure cannot be maintained below PSP is the EOP-3A, step PC/P-4 criteria requiring transition to emergency depressurization.

BWR Owners Group Appendix App. B, Contingency#4, step C4-2.2 App. B, step PC/P-1.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 5 Page 1 of 8 Revision 0 Facility: Cooper Nuclear Station Scenario No.: 5 Op-Test No.: CNS 15-01 Examiners: ____________________________ Operators:

Objectives: To evaluate the candidates ability to operate the facility in response to the following evolutions:

1. Withdraw control rods to establish 20-25% bypass valve position.

2. Respond to a control rod drifting out.

3. Respond to APRM A INOP failure.

4. Respond to a spurious RCIC initiation.

5. Respond to a HPCI steam line break and failure to isolate.

6. Respond to a LOOP.

7. Respond to a LOCA.

8. Respond to failure of DG 2 to automatically start.

9. Respond to RHR Loop A and Core Spray A pumps failure to automatically start.

10. Emergency Depressurize on low RPV level.

Initial Conditions: Plant operating at 5% power.

Inoperable Equipment: None Turnover:

The plant is at 5% power.

Planned activities for this shift are:

Continue reactor startup.

Scenario Notes:

This is a new scenario.

Validation Time: 75 minutes

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 5 Page 2 of 8 Revision 0 Event No.

Malf. No.

Event Type Event Description 1

N/A R (ATC)

Withdraw control rods to establish 20-25% bypass valve position 2

rd10 (22-23)

C (ATC)

A (CREW)

TS (CRS)

Control rod 22-23 drifts out 3

nm14a I (ATC)

APRM A INOP failure (half scram) 4 rc05 C

(BOP)

A (CREW)

TS (CRS)

Spurious RCIC initiation 5

hp06 hp09 I (BOP)

TS (CRS)

HPCI steam line break and failure to isolate 6

ed05 ed06 M

(CREW)

Loss of off-site power 7

dg06b C

(BOP)

A (CREW)

DG 2 Fails to automatically start When high pressure injection systems cannot maintain RPV level and low pressure ECCS systems fail to automatically start due to loss of AC power, crew manually starts DG2 to energize LP ECCS systems prior to RPV water level falling below -158 CFZ (TAF).**

8 rr20a M

(CREW)

LOCA, ED on low RPV level When RPV level lowers to -158 CFZ (TAF) and cannot be maintained above -183 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 -183 CFZ.

9 rh08a rh08c cs06a C

(BOP)

Div 1 low pressure ECCS pumps fail to automatically start When high pressure injection systems cannot maintain RPV level and low pressure ECCS pumps fail to automatically start, crew manually starts pumps to align LP ECCS systems for injection prior to RPV water level falling below -158 CFZ (TAF).**

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal (TS) Tech Spec Critical Task (As defined in NUREG 1021 Appendix D)

- Performing either Event 7 or Event 9 CT allows the RPV to be flooded. One or the other is required.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 5 Page 3 of 8 Revision 0 Quantitative Attributes Table Attribute ES-304-1 Target Actual Description Malfunctions after EOP entry 1-2 2

1.

DG 2 fails to automatically start on LOOP.

2.

RHR Loop A and Core Spray A pumps fail to auto start post LOCA.

Abnormal Events 2-4 2

1.

Control Rod drifts outward.

2.

RCIC spurious initiation.

Major Transients 1-2 2

1.

LOOP 2.

LOCA EOP entries requiring substantive action 1-2 2

1.

EOP-1A 2.

EOP-3A EOP contingencies requiring substantive action 0-2 1

1.

EOP-2A EOP based Critical Tasks 2-3 3

1.

When RPV level lowers to -158 CFZ (TAF) and cannot be maintained above -183 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 -183 CFZ.)

2.

When high pressure injection systems cannot maintain RPV level and low pressure ECCS systems fail to automatically start due to loss of AC power, crew manually starts DG2 to energize LP ECCS systems prior to RPV water level falling below -158 CFZ (TAF)

3.

When high pressure injection systems cannot maintain RPV level and low pressure ECCS pumps fail to automatically start, crew manually starts pumps to align LP ECCS systems for injection prior to RPV water level falling below -

158 CFZ (TAF).

Normal Events N/A 0

N/A Reactivity Manipulations N/A 1

1.

Withdraw control rods Instrument/

Component Failures N/A 6

1.

Control rod drifts out

2.

APRM failure

3.

Spurious RCIC initiation

4.

HPCI steam line break and failure to isolate.

5.

Diesel Generator 2 fails to automatically start.

6.

RHR Loop A and Core Spray A pumps fail to automatically start.

Total Malfunctions N/A 6

1.

Control rod drifts out

2.

APRM failure

3.

Spurious RCIC initiation

4.

HPCI steam line break and failure to isolate.

5.

Diesel Generator 2 fails to automatically start.

6.

RHR Loop A and Core Spray A pumps fail to automatically start.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 5 Page 4 of 8 Revision 0 Top 10 systems and operator actions important to risk that are tested:

Emergency AC power/DGs HPCI ADS/SRV Residual Heat Removal in LPCI injection MODE.

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 5 Page 5 of 8 Revision 0 SCENARIO

SUMMARY

The plant is starting up at approximately 5% power and the Reactor MODE switch is in RUN.

After the crew takes the watch, the ATC withdraws control rods to raise power to establish 20-25% bypass valve position While withdrawing control rods, control rod 22-23 begins drifting out. The control rod is fully inserted per abnormal procedure 2.4CRD. The CRS declares the control rod inoperable per TS LCO 3.1.3, Condition C.

After the TS call is made for the inoperable control rod, APRM A INOP/TRIP occurs. The crew determines it is an instrument failure and bypasses the APRM. The CRS determines it is a potential LCO and no TS LCO entry is required.

After the APRM failure is addressed, RCIC spuriously initiates. The crew enters 2.4CSCS and the BOP trips RCIC. RCIC-MO-131 control switch is overridden OPEN so RCIC cannot be reset and used for injection after the LOCA. The CRS enters TS LCO 3.5.3, Condition A and declares the RCIC inoperable.

Once TS are addressed for RCIC, a steam line break in HPCI occurs and HPCI fails to isolate. The BOP manually isolates HPCI from the main control room. The CRS enters TS LCO 3.5.1 Condition C and determines RCIC not operable. The CRS then enters TS LCO 3.5.1 Condition G.

After TS are addressed for HPCI, a LOOP and LOCA occurs. The reactor scrams and only Diesel Generator 1 connects to its bus. RHR A and C pumps and Core Spray A pump fail to automatically start and must be manually started. Diesel Generator 2 fails to start and must be manually started so it can automatically load onto its respective bus.

The CRS enters EOP 1A to control RPV parameters and EOP 3A to control PC parameters. The torus and drywell are sprayed to control containment pressure and temperature.

RPV level lowers to TAF requiring the crew to emergency depressurize.

The exercise ends when emergency depressurization is complete and RPV level is being restored.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 5 Page 6 of 8 Revision 0 Critical Tasks When high pressure injection systems cannot maintain RPV level and low pressure ECCS systems fail to automatically start due to loss of AC power, crew manually starts DG2 to energize LP ECCS systems prior to RPV water level falling below -158 CFZ (TAF)

When RPV level lowers to -158 CFZ (TAF) and cannot be maintained above -183 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 -183 CFZ.

EVENT 7

8 Safety significance Failure to recognize the auto start not occurring and energizing of the safety bus, and failure to take manual action per Procedure 5.3EMPWR will result in unavailability of safety-related equipment necessary to provide adequate core cooling, otherwise resulting in core damage and a large offsite release.

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 Indication and/or annunciation that all ac emergency buses are de-energized

Bus energized lamps extinguished

Circuit breaker position

Bus voltage

EDG status Control room lighting dimmed Corrected Fuel Zone indication (SPDS) falls to

-158 and lowering trend continues, and, before -158 CFZ is reached, initial conditions, field reports, and control room indications convey that adequate high pressure injection cannot be restored before level falls below -

183 CFZ Performance indicator Manipulation of controls as required to energize Div 1(2) AC emergency bus from panel C:

Operator places DIESEL GEN 1(2) BKR EG1(2) to CLOSE on panel C Manipulation of any six SRV controls on panel 9-3:

SRV-71A SRV-71H SRV-71B SRV-71C SRV-71E SRV-71D SRV-71G SRV-71F Performance feedback Crew will observe light indication for equipment powered by Division 1(2) AC illuminate on panel 9-3 and bus voltage

~4200V on panel C 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 Attempting to start 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.

There is no time limit for effecting complete The MSCWL (-183 CFZ) 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. Emergency depressurization is allowed when level goes below TAF (-158 CFZ) and should be performed, if in the judgment of the CRS, level cannot be maintained above -183 CFZ. Since it is 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 -183 CFZ.

BWR Owners Group Appendix App. B, Contingency#1 App. B, Contingency#1

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 5 Page 7 of 11 Revision 0 Critical Tasks When high pressure injection systems cannot maintain RPV level and low pressure ECCS pumps fail to automatically start, crew manually starts pumps to align LP ECCS systems for injection prior to RPV water level falling below -158 CFZ (TAF).

EVENT 9

Safety significance Failure to recognize the auto start 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 pumps are not running with initiation conditions present:

Green light on and Red lamp extinguished at respective pump handswitch on panel 9-3 Indication of Drywell Pressure 1.83 psig Indication of RPV water level -113 Performance indicator Manipulation of controls as required to start the affected ECCS pump(s) from panel 9-3:

Operator places affected ECCS pump(s) control switch(es) to START on panel 9-3 Performance feedback Crew will observe Red light illuminate and Green light extinguish for the affected ECCS pump(s) on panel 9-3 Justification for the chosen performance limit Attempting to start 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.

BWR Owners Group Appendix App. B, Contingency#1

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