ML17122A082

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)

Text

ES-401 BWR Examination Outline Form ES-401-1 Facility: Cooper Nuclear Station Date of Exam: March 2017 Tier Group RO K/A Category Points SRO-Only Points K K K K K K A A A A G A2 G* Total 1 2 3 4 5 6 1 2 3 4

• Total

1. 1 3 3 4 4 3 3 20 4 3 7 Emergency &

Abnormal 2 2 1 1 N/A 1 1 N/A 1 7 1 2 3 Plant Evolutions Tier Totals 4 4 6 5 4 4 27 5 5 10 1 2 2 3 4 3 2 2 2 2 2 2 26 2 3 5 2.

Plant 2 2 1 1 1 1 1 1 1 1 1 1 12 0 2 1 3 Systems Tier Totals 4 3 4 5 4 3 3 3 3 3 3 38 4 4 8 1 2 3 4 1 2 3 4

3. Generic Knowledge and Abilities 10 7 Categories 3 3 2 2 2 2 1 2 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 Radiation Control K/A is allowed if the K/A is replaced by a K/A from another Tier 3 Category.)

2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions.

The final RO exam must total 75 points and the SRO-only exam must total 25 points.

3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted with justification; operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.

4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.

5. Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.

7. The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As.

8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in a category other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams.

9. For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.

G* Generic K/As Rev 2

ES-401 2 Form ES-401-1 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 K K A A G* K/A Topic(s) IR #

1 2 3 1 2 Knowledge of the reasons for the following responses 295001 Partial or Complete Loss of Forced as they apply to PARTIAL OR COMPLETE LOSS OF X 3.4 1 Core Flow Circulation / 1 & 4 FORCED CORE FLOW CIRCULATION: (CFR: 41.5)

AK3.01 Reactor water level response Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C.

295003 Partial or Complete Loss of AC / 6 X 4.2 2 POWER: (CFR: 41.7)

AA1.02 Emergency generators Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF D.C.

295004 Partial or Total Loss of DC Pwr / 6 X POWER: (CFR: 41.10) 3.2 3 AA2.01 Cause of partial or complete loss of D.C.

power 2.1.23 Ability to perform specific system and integrated 295005 Main Turbine Generator Trip / 3 X plant procedures during all modes of plant operation. 4.3 4 (CFR: 41.10)

Knowledge of the operational implications of the following concepts as they apply to SCRAM:

295006 SCRAM / 1 X 3.7 5 (CFR: 41.8 to 41.10)

AK1.03 Reactivity control Knowledge of the interrelations between CONTROL 295016 Control Room Abandonment / 7 X ROOM ABANDONMENT and the following: (CFR: 41.7) 4.4 6 AK2.01 Remote shutdown panel: Plant-Specific Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF 295018 Partial or Total Loss of CCW / 8 X 3.1 7 COMPONENT COOLING WATER: (CFR: 41.5)

AK3.07 Cross-connecting with backup systems Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF 295019 Partial or Total Loss of Inst. Air / 8 X 3.0 8 INSTRUMENT AIR: (CFR: 41.7)

AA1.03 Instrument air compressor power supplies Knowledge of the reasons for the following responses as they apply to LOSS OF SHUTDOWN COOLING:

295021 Loss of Shutdown Cooling / 4 X (CFR: 41.5) 3.6 9 AK3.05 Establishing alternate heat removal flow paths Ability to determine and/or interpret the following as they 295023 Refueling Acc / 8 X apply to REFUELING ACCIDENTS: (CFR: 41.10) 3.4 10 AA2.02 Fuel Pool Level 2.2.44 Ability to interpret control room indications to verify status and operation of a system, and understand 295024 High Drywell Pressure / 5 X 4.2 11 how operator actions and directives affect plant and system conditions. (CFR: 41.5)

Knowledge of the operational implications of the following concepts as they apply to HIGH REACTOR 295025 High Reactor Pressure / 3 X PRESSURE: (CFR: 41.8 to 41.10) 3.6 12 EK1.03 Safety/relief valve tailpipe temperature /

pressure relationships Knowledge of the interrelations between 295026 Suppression Pool High Water Temp. SUPPRESSION POOL HIGH WATER TEMPERATURE X 3.5 13

/5 and the following: (CFR: 41.7)

EK2.06 Suppression pool level 295027 High Containment Temperature / 5 NOT APPLICABLE Rev 2

ES-401 3 Form ES-401-1 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 K K A A G* K/A Topic(s) IR #

1 2 3 1 2 Ability to operate and/or monitor the following as they apply to HIGH DRYWELL TEMPERATURE:

295028 High Drywell Temperature / 5 X 3.8 14 (CFR: 41.7)

EA1.01 Drywell spray: Mark-I&II Ability to determine and/or interpret the following as they apply to LOW SUPPRESSION POOL WATER LEVEL:

295030 Low Suppression Pool Wtr Lvl / 5 X (CFR: 41.10) 3.5 15 EA2.04 Drywell/ suppression chamber differential pressure: Mark-I&II 2.4.1 Knowledge of EOP entry conditions and 295031 Reactor Low Water Level / 2 X 4.6 16 immediate action steps. (CFR: 41.10)

Knowledge of the operational implications of the following concepts as they apply to SCRAM 295037 SCRAM Condition Present and CONDITION PRESENT AND REACTOR POWER Reactor Power Above APRM Downscale or X 4.1 17 ABOVE APRM DOWNSCALE OR UNKNOWN:

Unknown / 1 (CFR: 41.8 to 41.10)

EK1.02 Reactor water level effects on reactor power Knowledge of the interrelations between HIGH OFF-295038 High Off-site Release Rate / 9 X SITE RELEASE RATE and the following: (CFR: 41.7) 3.6 18 EK2.02 Offgas system Knowledge of the reasons for the following responses as they apply to PLANT FIRE ON SITE:

600000 Plant Fire On Site / 8 X 2.8 19 AK3.04 Actions contained in the abnormal procedure for plant fire on site Ability to operate and/or monitor the following as they 700000 Generator Voltage and Electric Grid apply to GENERATOR VOLTAGE AND ELECTRIC X 3.9 20 Disturbances / 6 GRID DISTURBANCES: (CFR: 41.5 and 41.10)

AA1.05 Engineered safety features Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C.

295003 Partial or Complete Loss of AC / 6 X 3.5 76 POWER : (CFR: 43.5)

AA2.03 Battery status: Plant-Specific 2.2.40 Ability to apply Technical Specifications for a 295021 Loss of Shutdown Cooling / 4 X 4.7 77 system. (CFR:43.2 / 43.5)

Ability to determine and/or interpret the following as 295023 Refueling Acc / 8 X they apply to REFUELING ACCIDENTS: (CFR: 43.5) 4.1 78 AA2.04 Occurrence of fuel handling accident 2.4.18 Knowledge of the specific bases for EOPs.

295038 High Off-site Release Rate / 9 X 4.0 79 (CFR: 43.1)

Ability to determine and/or interpret the following as they apply to REACTOR LOW WATER LEVEL :

295031 Reactor Low Water Level / 2 X 4.8 80 (CFR: 43.5)

EA2.04 Adequate core cooling 2.4.30 Knowledge of events related to system operation

/ status that must be reported to internal organizations 600000 Plant Fire On Site / 8 or external agencies, such as the State, the NRC, or the 4.1 81 X transmission system operator.

(CFR: 43.5)

Ability to determine and/or interpret the following as they 700000 Generator Voltage and Electric Grid apply to GENERATOR VOLTAGE AND ELECTRIC Disturbances / 6 X GRID DISTURBANCES: (CFR:43.5) 3.8 82 AA2.05 Operational status of offsite circuit 3/ 3/ 20/

K/A Category Totals: 3 3 4 4 Group Point Total:

4 3 7 Rev 2

ES-401 4 Form ES-401-1 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 K K A A G K/A Topic(s) IR #

1 2 3 1 2

• Knowledge of the operational implications of the following concepts as they apply to LOSS OF MAIN CONDENSER 295002 Loss of Main Condenser Vac / 3 X 3.6 21 VACUUM : (CFR: 41.8 to 41.10)

AK1.03 Loss of heat sink 295007 High Reactor Pressure / 3 NOT SELECTED Knowledge of the interrelations between HIGH REACTOR 295008 High Reactor Water Level / 2 X WATER LEVEL and the following: (CFR: 41.7 / 45.8) 3.6 22 AK2.03 Reactor water level control 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 Knowledge of the reasons for the following responses 295015 Incomplete SCRAM / 1 X as they apply to INCOMPLETE SCRAM : (CFR: 41.5) 3.4 23 AK3.01 Bypassing rod insertion blocks 295017 High Off-site Release Rate / 9 NOT SELECTED Ability to operate and/or monitor the following as they apply to INADVERTENT CONTAINMENT ISOLATION :

295020 Inadvertent Cont. Isolation / 5 & 7 X 2.9 24 (CFR: 41.7)

AA1.03 Containment ventilation system: Plant-Specific Ability to determine and/or interpret the following as 295022 Loss of CRD Pumps / 1 X they apply to LOSS OF CRD PUMPS : (CFR: 41.10) 3.5 25 AA2.01 Accumulator pressure 295029 High Suppression Pool Wtr Lvl / 5 NOT SELECTED 295032 High Secondary Containment Area NOT SELECTED Temperature / 5 295033 High Secondary Containment Area NOT SELECTED Radiation Levels / 9 295034 Secondary Containment 2.4.31 Knowledge of annunciator alarms, indications, or X 4.2 26 Ventilation High Radiation / 9 response procedures. (CFR: 41.10) 295035 Secondary Containment High NOT SELECTED Differential Pressure / 5 Knowledge of the operational implications of the following concepts as they apply to SECONDARY 295036 Secondary Containment High X CONTAINMENT HIGH SUMP/AREA WATER LEVEL : 2.6 27 Sump/Area Water Level / 5 (CFR: 41.8 to 41.10)

EK1.02 Electrical ground/ circuit malfunction 500000 High CTMT Hydrogen Conc. / 5 NOT SELECTED 2.4.4 Ability to recognize abnormal indications for system 295014 Inadvertent Reactivity Addition / 1 X operating parameters that are entry-level conditions for 4.7 83 emergency and abnormal operating procedures.(CFR: 43.2)

Ability to determine and/or interpret the following as they apply to HIGH SUPPRESSION POOL WATER LEVEL :

295029 High Suppression Pool Wtr Lvl / 5 X 3.9 84 (CFR: 43.5)

EA2.01 Suppression pool water level 500000 High CTMT Hydrogen Conc. / 5 X 2.4.6 Knowledge of EOP mitigation strategies. (CFR: 43.5) 4.7 85 1 1 7/

K/A Category Point Totals: 2 1 1 1 / / Group Point Total:

3 1 2 Rev 2

ES-401 5 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 1 (RO / SRO)

System # / Name K K K K K K A A A A G K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4

• Knowledge of electrical power supplies to the 203000 RHR/LPCI: Injection X following: (CFR: 41.7) 2.7 28 Mode K2.03 Initiation logic Knowledge of the effect that a loss or malfunction of the SHUTDOWN COOLING SYSTEM (RHR 205000 Shutdown Cooling X SHUTDOWN COOLING MODE) will have on 3.3 29 following: (CFR: 41.7)

K3.01 Reactor pressure Knowledge of shutdown cooling system (RHR shutdown cooling mode) design feature(s) and/or 205000 Shutdown Cooling X 3.7 30 interlocks which provide for the following: (CFR: 41.7)

K4.02 High pressure isolation: Plant-Specific Knowledge of the operational implications of the following concepts as they apply to HIGH 206000 HPCI X PRESSURE COOLANT INJECTION SYSTEM : 3.3 31 (CFR: 41.5)

K5.05 Turbine speed control: BWR-2,3,4 Knowledge of the effect that a loss or malfunction of the following will have on the HIGH PRESSURE 206000 HPCI X 3.3 32 COOLANT INJECTION SYSTEM : (CFR: 41.7)

K6.02 D.C. power: BWR-2,3,4 207000 Isolation (Emergency)

NOT APPLICABLE Condenser Knowledge of LOW PRESSURE CORE SPRAY SYSTEM design feature(s) and/or interlocks which 209001 LPCS X 3.0 33 provide for the following: (CFR: 41.7)

K4.04 Line break detection 209002 HPCS NOT APPLICABLE Ability to predict and/or monitor changes in parameters associated with operating the STANDBY 211000 SLC X LIQUID CONTROL SYSTEM controls including: 3.7 34 (CFR: 41.5)

A1.08 RWCU system lineup Ability to (a) predict the impacts of the following on the REACTOR PROTECTION SYSTEM ; and (b) based on those predictions, use procedures to 212000 RPS X 4.0 35 correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5)

A2.16 Changing mode switch position Ability to monitor automatic operations of the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM 215003 IRM X 3.5 36 including: (CFR: 41.7)

A3.04 Control rod block status Ability to manually operate and/or monitor in the control room: (CFR: 41.7) 215004 Source Range Monitor X 3.4 37 A4.07 Verification of proper functioning /

operability 2.1.7 Ability to evaluate plant performance and make operational judgments based on operating 215005 APRM / LPRM X 4.4 38 characteristics, reactor behavior, and instrument interpretation. (CFR: 41.5)

Knowledge of the physical connections and/or cause/effect relationships between REACTOR CORE 217000 RCIC X ISOLATION COOLING SYSTEM (RCIC) and the 3.5 39 following: (CFR: 41.2 to 41.9)

K1.02 Nuclear boiler system Rev 2

ES-401 6 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 1 (RO / SRO)

System # / Name K K K K K K A A A A G K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4

• Knowledge of electrical power supplies to the 218000 ADS X following: (CFR: 41.7) 3.1 40 K2.01 ADS logic Knowledge of the effect that a loss or malfunction of the PRIMARY CONTAINMENT ISOLATION 223002 PCIS/Nuclear Steam X SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF 3.3 41 Supply Shutoff will have on following: (CFR: 41.7)

K3.20 Standby gas treatment system Knowledge of the operational implications of the following concepts as they apply to RELIEF/SAFETY 239002 SRVs X 3.7 42 VALVES: (CFR: 41.5)

K5.02 Safety function of SRV operation Knowledge of the effect that a loss or malfunction of 259002 Reactor Water Level the following will have on the REACTOR WATER X 3.3 43 Control LEVEL CONTROL SYSTEM: (CFR: 41.7)

K6.02 A.C. power Ability to predict and/or monitor changes in parameters associated with operating the REACTOR 259002 Reactor Water Level WATER LEVEL CONTROL SYSTEM controls X 2.9 44 Control including: (CFR: 41.5)

A1.05 FWRV/startup level control position: Plant-Specific Ability to (a) predict the impacts of the following on the STANDBY GAS TREATMENT SYSTEM ; and (b) based on those predictions, use procedures to 261000 SGTS X 3.2 45 correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5)

A2.11 High containment pressure Ability to monitor automatic operations of the A.C.

262001 AC Electrical X ELECTRICAL DISTRIBUTION including: (CFR: 41.7) 3.4 46 Distribution A3.04 Load sequencing Ability to manually operate and/or monitor in the 262001 AC Electrical control room: (CFR: 41.7)

X 3.3 47 Distribution A4.05 Voltage, current, power, and frequency on A.C. buses 2.1.27 Knowledge of system purpose and/or function.

262002 UPS (AC/DC) X 3.9 48 (CFR: 41.7)

Knowledge of the physical connections and/or cause/effect relationships between D.C.

263000 DC Electrical X ELECTRICAL DISTRIBUTION and the following: 3.2 49 Distribution (CFR: 41.2 to 41.9)

K1.02 Battery charger and battery Knowledge of the effect that a loss or malfunction of the EMERGENCY GENERATORS (DIESEL/JET) will 264000 EDGs X have on following: (CFR: 41.7 / 45.4) 4.1 50 K3.03 Major loads powered from electrical buses fed by the emergency generator(s)

Knowledge of EMERGENCY GENERATORS (DIESEL/JET) design feature(s) and/or interlocks 264000 EDGs X 3.8 51 which provide for the following: (CFR: 41.7)

K4.08 Automatic startup Knowledge of the operational implications of the following concepts as they apply to the 300000 Instrument Air X 2.5 52 INSTRUMENT AIR SYSTEM: (CFR: 41.5 / 45.3)

K5.01 Air compressors Rev 2

ES-401 7 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 1 (RO / SRO)

System # / Name K K K K K K A A A A G K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4

• Knowledge of CCWS design feature(s) and or 400000 Component Cooling X interlocks which provide for the following: (CFR: 41.7) 3.4 53 Water K4.01 Automatic start of standby pump Ability to (a) predict the impacts of the following on the RHR/LPCI: INJECTION MODE (PLANT SPECIFIC) ; and (b) based on those predictions, use 203000 RHR/LPCI: Injection X procedures to correct, control, or mitigate the 3.6 86 Mode consequences of those abnormal conditions or operations:

A2.04 A.C. failures 2.2.25 Knowledge of the bases in Tech Specs for 212000 RPS X 4.2 87 LCOs and Safety limits (CFR: 41.5 / 41.7 / 43.2)

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 215005 APRM / LPRM X 3.4 88 correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5)

A2.08 Faulty or erratic operation of detectors /

systems 2.4.20 Knowledge of the operational implications of 218000 ADS X EOP warnings, cautions, and notes. 4.3 89 (CFR: 41.10 / 43.5) 2.4.45 Ability to prioritize and interpret the 300000 Instrument Air X significance of each annunciator or alarm. 4.3 90 (CFR: 41.10 / 43.5) 2 2/ 26 K/A Category Point Totals: 2 2 3 4 3 2 2 2 2 / Group Point Total:

2 /5 3

Rev 2

ES-401 8 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 2 (RO / SRO)

System # / Name K K K K K K A A A A G K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4

• Knowledge of the physical connections and/or cause-effect relationships between CONTROL ROD DRIVE 201001 CRD Hydraulic X HYDRAULIC SYSTEM and the following: 3.4 54 (CFR: 41.2 to 41.9)

K1.07 Reactor protection system Knowledge of the effect that a loss or malfunction of the REACTOR MANUAL CONTROL SYSTEM will 201002 RMCS X 2.9 55 have on following: (CFR: 41.7)

K3.03 Ability to process rod block signals Knowledge of the operational implications of the 201003 Control Rod and Drive following concepts as they apply to CONTROL ROD X 3.1 56 Mechanism AND DRIVE MECHANISM : (CFR: 41.5)

K5.04 Rod sequence patterns 201004 RSCS NOT APPLICABLE 201005 RCIS NOT APPLICABLE 201006 RWM NOT SELECTED Knowledge of electrical power supplies to the 202001 Recirculation X following: (CFR: 41.7) 3.2 57 K2.02 MG sets: Plant-Specific Knowledge of RECIRCULATION FLOW CONTROL SYSTEM design feature(s) and/or interlocks which 202002 Recirculation Flow Control X provide for the following: (CFR: 41.7) 2.9 58 K4.07 Minimum and maximum pump speed setpoints 204000 RWCU NOT SELECTED 214000 RPIS NOT SELECTED 215001 Traversing In-Core Probe NOT SELECTED 215002 RBM NOT SELECTED Knowledge of the effect that a loss or malfunction of the following will have on the NUCLEAR BOILER 216000 Nuclear Boiler Inst. X 3.1 59 INSTRUMENTATION : (CFR: 41.7)

K6.01 A.C. electrical distribution 219000 RHR/LPCI: Torus/Pool NOT SELECTED Cooling Mode 223001 Primary CTMT and Aux. NOT SELECTED 226001 RHR/LPCI: CTMT Spray NOT SELECTED Mode 230000 RHR/LPCI: Torus/Pool NOT SELECTED Spray Mode Ability to predict and/or monitor changes in parameters associated with operating the FUEL POOL 233000 Fuel Pool Cooling/Cleanup X COOLING AND CLEAN-UP controls including: (CFR: 2.5 60 41.5)

A1.06 System flow Ability to monitor automatic operations of the FUEL 234000 Fuel Handling Equipment X HANDLING EQUIPMENT including: (CFR: 41.7) 3.1 61 A3.02 Interlock operation 239001 Main and Reheat Steam NOT SELECTED 239003 MSIV Leakage Control NOT SELECTED Rev 2

ES-401 9 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 2 (RO / SRO)

System # / Name K K K K K K A A A A G K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4

• Ability to (a) predict the impacts of the following on the REACTOR/TURBINE PRESSURE REGULATING SYSTEM ; and (b) based on those predictions, use 241000 Reactor/Turbine Pressure X procedures to correct, control, or mitigate the 3.7 62 Regulator consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6)

A2.07 Loss of condenser vacuum Ability to manually operate and/or monitor in the 245000 Main Turbine Gen. / Aux. X control room: (CFR: 41.7) 3.1 63 A4.02 Generator controls 256000 Reactor Condensate NOT SELECTED 259001 Reactor Feedwater NOT SELECTED 268000 Radwaste NOT SELECTED 271000 Offgas NOT SELECTED 272000 Radiation Monitoring NOT SELECTED 2.1.32 Ability to explain and apply system limits and 286000 Fire Protection X 3.8 64 precautions. (CFR: 41.10) 288000 Plant Ventilation NOT SELECTED 290001 Secondary CTMT NOT SELECTED 290003 Control Room HVAC NOT SELECTED Knowledge of the physical connections and/or cause-effect relationships between REACTOR VESSEL 290002 Reactor Vessel Internals X INTERNALS and the following: 3.4 65 (CFR: 41.2 to 41.9)

K1.15 Nuclear boiler instrumentation 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 223001 Primary CTMT and Aux. X 4.6 91 system integrity, containment conditions, radioactivity release control, etc.

(CFR: 41.7 / 43.5)

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, 239001 Main and Reheat Steam X control, or mitigate the consequences of those 3.9 92 abnormal conditions or operations:

(CFR: 41.5 / 45.6)

A2.10 Closure of one or more MSIV's at power Ability to (a) predict the impacts of the following on the REACTOR FEEDWATER SYSTEM ; and (b) based on those predictions, use procedures to correct, 259001 Reactor Feedwater X control, or mitigate the consequences of those 3.8 93 abnormal conditions or operations: (CFR: 41.5 / 45.6)

A2.07 Reactor water level control system malfunctions 1 1 12 K/A Category Point Totals: 2 1 1 1 1 1 1 / 1 1 / Group Point Total:

/3 2 1 Rev 2

Facility: Cooper Nuclear Station Date of Exam: March 2017 RO SRO-Only Category K/A # Topic IR # IR #

2.1.1 Knowledge of conduct of operations requirements. (CFR: 41.10) 3.8 66 Knowledge of procedures, guidelines, or limitations associated with 2.1.37 4.3 67 reactivity management. (CFR: 41.1)

Ability to identify and interpret diverse indications to validate the 2.1.45 4.3 68

1. response of another indication. (CFR: 41.7)

Conduct of Knowledge of individual licensed operator responsibilities related to Operations 2.1.4 shift staffing, such as medical requirements, no-solo operation, 3.8 94 maintenance of active license status, 10CFR55, etc. (CFR: 43.2)

Knowledge of procedures and limitations involved in core alterations.

2.1.36 4.1 95 (CFR: 43.6)

Subtotal 3 2 Knowledge of the process for making changes to procedures.

2.2.6 3.0 69 (CFR: 41.10)

Knowledge of the process for controlling equipment configuration or 2.2.14 3.9 70 status. (CFR: 41.10)

2. Knowledge of less than or equal to one hour Technical Specification 2.2.39 3.9 71 Equipment action statements for systems. (CFR: 41.7 / 41.10)

Control Knowledge of the process for managing troubleshooting activities.

2.2.20 3.8 96 (CFR: 43.5)

Knowledge of conditions and limitations in the facility license.

2.2.38 4.5 97 (CFR: 43.1)

Subtotal 3 2 Knowledge of radiation exposure limits under normal or emergency 2.3.4 3.2 72 conditions. (CFR: 41.12)

Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel

3. 2.3.12 3.2 73 handling responsibilities, access to locked high-radiation areas, Radiation aligning filters, etc. (CFR: 41.12)

Control Knowledge of radiation or contamination hazards that may arise 2.3.14 during normal, abnormal, or emergency conditions or activities. 3.8 98 (CFR: 43.4)

Subtotal 2 1 Knowledge of procedures relating to a security event (non-safeguards 2.4.28 3.2 74 information). (CFR: 41.10 / 43.5 / 45.13)

Knowledge of RO responsibilities in emergency plan implementation.

2.4.39 3.9 75 (CFR: 41.10)

4. Knowledge of EOP implementation hierarchy and coordination with Emergency other support procedures or guidelines such as, operating procedures, 2.4.16 4.4 99 Procedures / abnormal operating procedures, and severe accident management Plan guidelines. (CFR: 43.5)

Ability to take actions called for in the facility emergency plan, 2.4.38 including supporting or acting as emergency coordinator if required. 4.4 100 (CFR: 41.10 / 43.5 / 45.11)

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.

Rev 2

ES-401 Record of Rejected K/As Form ES-401-4 Randomly Tier / Group Selected K/A Reason for Rejection (Original)

(New)

RO T1/G1 Because CNS has a Mark I containment and NUREG 1123 states EPE 295027 295027 295027 High Containment Temperature is for Mark III containments only, Not Selected Not Applicable EPE 295027 was changed from NOT SELECTED to NOT APPLICABLE.

Page 1 point totals not affected by this change. (Rev 1)

SRO T1/G1 Because CNS has a Mark I containment and NUREG 1123 states EPE 295027 295038 295027 High Containment Temperature is for Mark III containments only, G2.4.18 G2.4.18 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 Because CNS does have a Low Pressure Core Spray system, System 239002 209001 209001 LPCS was changed from NOT APPLICABLE. Since System K4.04 K4.04 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 Because the Rod Sequence Control System is no longer used at CNS, 201004 201004 System 201004 RSCS was changed from NOT SELECTED to NOT Not Selected Not Applicable APPLICABLE. Page 1 point totals not affected by this change. (Rev. 1)

RO T2/G2 Because CNS does not have a Rod Control and Information System, 201005 201005 System 201005 RCIS was changed from NOT SELECTED to NOT Not Selected Not Applicable APPLICABLE. Page 1 point totals not affected by this change. (Rev 1)

RO T1/G1 Because a discriminatory, operationally valid RO question could not be 295026 295026 developed question, replaced 295026 EK2.03 with randomly selected EK2.03 EK2.06 EK2.06. Page 1 point totals not affected by this change. (Rev 2)

RO T2/G1 The only SLC flow indicator at CNS is a local float type meter on the SLC 211000 211000 Test Tank inlet piping. No flow indication is available for SLC injection to A1.06 A1.08 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 Single/Sequential turbine governor valve operation is no longer used at 245000 245000 CNS following high pressure turbine replacement during RE29. Because A4.07 A4.02 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: RO SRO Operating Test Number: 1 Administrative Topic Type Describe activity to be performed (see Note) Code*

A1, Perform Jet Pump Operability Check (RO)

Conduct of Operations R, D 2.1.25 (3.9/4.2)

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

A3, Determine Isolation Boundaries (RHR)

Equipment Control R, D 2.2.13 (4.1/4.3)

A4, Determine Workers Projected Total Dose Radiation Control R, N 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: RO SRO Operating Test Number: 1 Administrative Topic Type Describe activity to be performed (see Note) Code*

A5, Determine if Mode Change is Allowed Conduct of Operations R, D 2.1.20 (4.6/4.6), 2.2.35 (3.6/4.5), 2.2.40 (3.4/4.7)

A6, Reportable Occurrences to the NRC (#8)

Conduct of Operations R, N 2.1.18 (3.6/3.8), 2.1.20 (4.6/4.6), 2.4.30 (2.7/4.1)

A7, Review Jet Pump Operability and Recirc Pump Flow Checks Equipment Control R, M 2.2.12 (3.7/4.1), 2.2.42 (3.9/4.6)

A8, Authorize Stable Iodine Thyroid Blocking Radiation Control R, D 2.3.14 (3.4/3.8)

A9, Emergency Classification Emergency Plan R, D 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

ES-301 Control Room/In Plant Systems Outline Form ES-301-2 Facility: Cooper Nuclear Station Date of Examination: 03/06/2017 Exam Level: RO SRO-I SRO-U Operating Test No.: 1 Control Room Systems; * (8 for RO); (7 for SRO-I); 2 or 3 for SRO-U Safety System / JPM Title Type Code*

Function

a. JPM S1 - Secure SDG from Control Room L, N, S 6 264000 A4.04 (3.7/3.7)

b. JPM S2 - Defeat RPS Logic Trips During Failure to Scram (5.8.3) (Restoration) L, D, S 7 295037 EA1.01 (4.6/4.6)

c. JPM S3 - Conduct Alternate Pressure Control Using Reactor Feed Pumps L, D, S 3 259001 A4.02 (3.9/3.7)

d. JPM S4 - Level Recovery During Shutdown Conditions Using LPCI (Alternate Path) A, EN, L, N, S 2 203000 A4.05 (4.3/4.1)

e. JPM S5 - Perform 6.TG.303 Testing OPC Overspeed L, N, S 4 241000 A3.12 (2.9/2.9), 241000 A4.19 (3.5/3.4)

f. JPM S6 - Perform Standby Gas Treatment System Decay Heat Removal D, S 9 261000 A3.04 (3.0/3.1), 261000 A4.03 (3.0/3.0)

g. JPM S7 - Withdrawal of Control Rod From Position 00 (Alternate Path 2) A, L, N, S 1 201003 A2.01 (3.4/3.6)

h. JPM S8 - Verify Group 2 Isolation (Alt Path TIP Shear)

A, D, S 5 223002 A4.01 (3.6/3.5), 223002 A4.06 (3.6/3.7)

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)

A, N 8 286000 A4.05 (3.3/3.3), 286000 A1.05 (3.2/3.2)

j. JPM P2 - Place in Service/Remove From Service Hydrogen Cylinders/Tanks A, D 4 245000 A3.08 (2.5/2.6)

k. JPM P3 - Alternate Shutdown, Locally Operate SW-MO-89B for starting Torus Cooling D, E, L, R 5 219000 A1.08 (3.7 / 3.6), 295016 AK2.02 (4.0/4.1)

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 1 of 2 Rev 0

• Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 (5)

(C)ontrol room (D)irect from bank <9/<8/<4 (6)

(E)mergency or abnormal in-plant >1/>1/>1 (1)

(EN)gineered safety feature > 1 / > 1 / > 1 (control room system) (1)

(L)ow-Power / Shutdown >1/>1/>1 (7)

(N)ew or (M)odified from bank including 1(A) >2/>2/>1 (5)

(P)revious 2 exams < 3 / < 3 / < 2 (randomly selected) (0)

(R)CA >1/>1/>1 (1)

(S)imulator Page 2 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: RO SRO-I SRO-U Operating Test No.: 1 Control Room Systems:

• 8 for RO; 7 for SRO-I; 2 or 3 for SRO-U Safety System / JPM Title Type Code*

Function

a. JPM S1 - Secure SDG from Control Room L, N, S 6 264000 A4.04 (3.7/3.7)

b. JPM S2 - Defeat RPS Logic Trips During Failure to Scram (5.8.3) (Restoration) L, D, S 7 295037 EA1.01 (4.6/4.6) c.

d. JPM S4 - Level Recovery During Shutdown Conditions Using LPCI (Alternate Path) A, EN, L, N, S 2 203000 A4.05 (4.3/4.1)

e. JPM S5 - Perform 6.TG.303 Testing OPC Overspeed L, N, S 4 241000 A3.12 (2.9/2.9), 241000 A4.19 (3.5/3.4)

f. JPM S6 - Perform Standby Gas Treatment System Decay Heat Removal D, S 9 261000 A3.04 (3.0/3.1), 261000 A4.03 (3.0/3.0)

g. JPM S7 - Withdrawal of Control Rod From Position 00 (Alternate Path 2) A, L, N, S 1 201003 A2.01 (3.4/3.6)

h. JPM S8 - Verify Group 2 Isolation (Alt Path TIP Shear)

A, D, S 5 223002 A4.01 (3.6/3.5), 223002 A4.06 (3.6/3.7)

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)

A, N 8 286000 A4.05 (3.3/3.3), 286000 A1.05 (3.2/3.2)

j. JPM P2 - Place in Service/Remove From Service Hydrogen Cylinders/Tanks A, D 4 245000 A3.08 (2.5/2.6)

k. JPM P3 - Alternate Shutdown, Locally Operate SW-MO-89B for starting Torus Cooling D, E, L, R 5 219000 A1.08 (3.7 / 3.6), 295016 AK2.02 (4.0/4.1)

• 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 1 of 2 Rev 0

• Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 (5)

(C)ontrol room (D)irect from bank <9/<8/<4 (5)

(E)mergency or abnormal in-plant >1/>1/>1 (1)

(EN)gineered safety feature > 1 / > 1 / > 1 (control room system) (1)

(L)ow-Power / Shutdown >1/>1/>1 (6)

(N)ew or (M)odified from bank including 1(A) >2/>2/>1 (5)

(P)revious 2 exams < 3 / < 3 / < 2 (randomly selected) (0)

(R)CA >1/>1/>1 (1)

(S)imulator Page 2 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: RO SRO-I SRO-U Operating Test No.: 1 Control Room Systems:

• 8 for RO; 7 for SRO-I; 2 or 3 for SRO-U Safety System / JPM Title Type Code*

Function a.

b. JPM S2 - Defeat RPS Logic Trips During Failure to Scram (5.8.3) (Restoration) L, D, S 7 295037 EA1.01 (4.6/4.6) c.

d. JPM S4 - Level Recovery During Shutdown Conditions Using LPCI (Alternate Path) A, EN, L, N, S 2 203000 A4.05 (4.3/4.1) 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)

A, N 8 286000 A4.05 (3.3/3.3), 286000 A1.05 (3.2/3.2)

j. JPM P2 - Place in Service/Remove From Service Hydrogen Cylinders/Tanks A, D 4 245000 A3.08 (2.5/2.6)

k. JPM P3 - Alternate Shutdown, Locally Operate SW-MO-89B for starting Torus Cooling D, E, L, R 5 219000 A1.08 (3.7 / 3.6), 295016 AK2.02 (4.0/4.1)

• 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 1 of 2 Rev 0

• Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 (3)

(C)ontrol room (D)irect from bank <9/<8/<4 (3)

(E)mergency or abnormal in-plant >1/>1/>1 (1)

(EN)gineered safety feature > 1 / > 1 / > 1 (control room system) (1)

(L)ow-Power / Shutdown >1/>1/>1 (3)

(N)ew or (M)odified from bank including 1(A) >2/>2/>1 (2)

(P)revious 2 exams < 3 / < 3 / < 2 (randomly selected) (0)

(R)CA >1/>1/>1 (1)

(S)imulator Page 2 of 2 Rev 0

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

A E Scenarios P V 1 2 3 4 T M P E O I L N CREW CREW POSITION CREW POSITION CREW POSITION T N I T POSITION I

C A S A B S A B S A B S A B M A T L R T O R T O R T O R T O U N Y O C P O C P O C P O C P M(*)

T P E R I U RO RX 1 0 1 1 1 0 (R2,R4)

NOR 1 1 2 1 1 1 SRO-I I/C 2 3 5 4 4 2 MAJ 1 2 3 2 2 1 SRO-U TS NA NA NA 0 2 2 RO RX 0 0 0 1 1 0 (R1,R3)

NOR 0 1 1 1 1 1 SRO-I I/C 4 2 6 4 4 2 MAJ 1 2 3 2 2 1 SRO-U TS NA NA NA 0 2 2 RO RX 1 0 1 1 1 0 NOR 1 2 3 1 1 1 SRO-I I/C 4 4 8 4 4 2 SRO-U MAJ 1 2 3 2 2 1 (U1,U2)

TS 2 2 4 0 2 2 RO RX 0 0 0 0 1 1 0 NOR 0 0 2 2 1 1 1 SRO-I (I1,I4) I/C 4 4 4 12 4 4 2 MAJ 1 2 2 5 2 2 1 SRO-U 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 E Scenarios P V 1 2 3 4 T M P E O I L N CREW CREW POSITION CREW POSITION CREW POSITION T N I T POSITION I

C A S A B S A B S A B S A B M A T L R T O R T O R T O R T O U N Y O C P O C P O C P O C P M(*)

T P E R I U RO RX 1 0 0 1 1 1 0 NOR 1 1 1 3 1 1 1 SRO-I (I2,I5) I/C 4 3 2 9 4 4 2 MAJ 1 2 2 5 2 2 1 SRO-U TS 2 NA NA 2 0 2 2 RO RX 1 0 0 1 1 1 0 NOR 1 1 1 3 1 1 1 SRO-I(I3,I6) I/C 2 5 3 10 4 4 2 MAJ 1 2 2 5 2 2 1 SRO-U TS NA 2 NA 2 0 2 2 RO RX 0 1 1 1 1 0 NOR 1 0 1 1 1 1 SRO-I (I7)

I/C 5 2 7 4 4 2 SRO-U MAJ 2 1 3 2 2 1 TS 2 NA 2 0 2 2 RO RX 0 1 1 1 1 0 NOR 0 1 1 1 1 1 SRO-I (I8)

I/C 4 4 8 4 4 2 SRO-U 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 E Scenarios P V 1 2 3 4 T M P E O I L N CREW CREW POSITION CREW POSITION CREW POSITION T N I T POSITION I

C A S A B S A B S A B S A B M A T L R T O R T O R T O R T O U N Y O C P O C P O C P O C P M(*)

T P E R I U RO RX 0 1 1 1 1 0 NOR 1 1 2 1 1 1 SRO-I I/C 5 4 9 4 4 2 SRO-U(U3) MAJ 2 1 3 2 2 1 TS 2 2 4 0 2 2 RO (R5) RX 0 0 0 1 1 0 NOR 0 1 1 1 1 1 SRO-I I/C 4 3 7 4 4 2 SRO-U MAJ 2 1 3 2 2 1 TS NA NA NA 0 2 2 RO (R6) RX 0 1 1 1 1 0 NOR 1 0 1 1 1 1 SRO-I I/C 3 2 5 4 4 2 SRO-U MAJ 2 1 3 2 2 1 TS NA NA NA 0 2 2 RO RX 1 1 0 NOR 1 1 1 SRO-I I/C 4 4 2 SRO-U 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 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 Revision 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 1 Page 2 of 10 Event Malf. No. Event Type Event No. Description 1 N/A N (ATC) Shift CRD stabilizing valves 2 N/A R (ATC) Lower Reactor power by lowering RR pump speed (or) Reactor Building to Torus vacuum breaker fails open.

3 TS (CRS) zdipcswcs243av[2] CRS declares vacuum breaker inoperable.

C (BOP) Respond to reactor vessel flange seal leak alarm, enter 4 rr21 Procedure 4.6.3, and cycle the flange leak-off drain A (CREW) valves I (BOP, ATC) rp03a RPS EPA Breaker 1A1/1A2 trip, (half scram and half 5 A (CREW) PCIS group isolations) RMV-AO-10 fails to isolate.

(rf) rh32a CRS declares valve inoperable TS (CRS)

C (BOP, sw 11a REC Pump A trip. Start another REC pump. REC 6 ATC) sw11b Pump B trip. Manual scram due to loss of REC.

A(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 7 rd02a,b >3%, stop and prevent injection from all sources M (CREW)

(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.

(rf) rh29(A) First RHR loop to be put into suppression pool cooling 8 C (BOP)

(rf) rh30(A) 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)

Revision 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 1 Page 3 of 10 Quantitative Attributes Table ES-304-1 Attribute Actual Description Target Malfunctions after 1. RHR-MO-39A(B) fails to open.

1-2 1 EOP entry

1. RPV flange seal leak Abnormal Events 2-4 3 2. RPS EPA trip

3. Loss of multiple REC Pumps Major Transients 1-2 1 1. ATWS EOP entries requiring 1. EOP-6A 1-2 2 2. EOP-7A substantive action EOP contingencies requiring substantive 0-2 1 1. EOP- 7A action

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 EOP based Critical injects SLC before exceeding the Boron 2-3 4 Injection Initiation Temperature (BIIT) curve.

Tasks

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 1. Lower power using Reactor Recirculation pumps N/A 1 Manipulations

1. RPV Flange leak

2. RPS A EPA Breaker trip Instrument/ 3. Loss of REC pump A N/A 5 Component Failures 4. Loss of REC pump B

5. RHR-MO39A(B) valve fails to open

1. RPV Flange leak

2. RPS A EPA Breaker trip Total Malfunctions N/A 5 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

Revision 0

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

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

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Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 1 Page 6 of 10 Critical Tasks When control rods fail to scram, crew During failure to scram conditions with injects SLC and/or inserts control rods power >3%, stop and prevent injection from before exiting EOP-6A. (All control rods all sources (except boron, CRD, RCIC) as do not have to be fully inserted to satisfy necessary to lower RPV water level to this critical task; this only requires that below -60 CFZ (or LL, as applicable) and the crew is making progress to achieving control between -60 (or LL, as applicable) all rods in by fully inserting at least 5 to -183 CFZ prior to exiting EOP-7A.

control rods using RMCS.)

EVENT 6 6 Safety Failure to effect shutdown of the reactor Regarding lowering level below -60 CFZ, to significance when a RPS setting has been exceeded prevent or mitigate the consequences of any would unnecessarily extend the level of large irregular neutron flux oscillations induced degradation of the safety of the plant. This by neutronic/thermal-hydraulic instabilities, could further degrade into damage to the RPV water level is lowered sufficiently below principle fission product barriers if left the elevation of the feedwater sparger nozzles.

unmitigated. The crew is authorized and This places the feedwater spargers in the required by Conduct of Operations to take steam space providing effective heating of the mitigating actions when automatic safety relatively cold feedwater and eliminating the systems fail to perform their intended potential for high core inlet subcooling. For function. Action to shut down the reactor is conditions that are susceptible to oscillations, required when RPS and control rod drive the initiation and growth of oscillations is systems fail. 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 Revision 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 1 Page 7 of 10 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 Manual scram is initiated and numerous control rods indicate beyond position 00 and control rods indicate beyond position 00 and reactor power not downscale on panel 9-5 reactor power >3% on panel 9-5 indications indications. and SPDS and RPV level is >-60CFZ on SPDS.

Performance Operator manipulates keylocked switches for Operator manipulates Feedwater HMIs on indicator SLC B pump to START on panel 9-5. panel 9-5 or panel A as necessary to stop FW injection until RPV level goes below -60CFZ.

Operator selects individual control rods by depressing the respective pushbutton on the Operator manipulates HPCI controls on panel panel 9-5 matrix and inserts the rod by 9-3 to stop HPCI injection until RPV level is manipulating the emergency in switch on below -60CFZ.

panel 9-5.

Performance SLC B pump red light illuminated, SLC Feedwater flow indication on panel 9-5 indicate feedback discharge pressure rising, and SLC tank level zero.

lowering on panel 9-5.

HPCI flow indication on panel 9-3 indicates Operator selecting and inserting control rods zero and/or HPCI injection MOV indicates indicated by rod position decreasing to 00 for closed.

selected rod on panel 9-5.

Justification There is no time limit for effecting complete Applicability for this CT is during EOP-7A for the chosen reactor shutdown via boron injection or conditions where it is necessary to lower level performance control rod insertion. For the timeframe of to control power with Table 17 condition NOT limit this scenario, containment limits are not met (i.e. no high energy input into primary closely challenged and power oscillations are containment). There is no time limit for this not experienced. However, if the failure to lowering level, but it establishes margin to scram EOP were to be exited, other conditions where fuel damaging power procedures would not provide the guidance oscillations may theoretically occur. Before necessary to achieve reactor shutdown. exiting EOP-7A was chosen because Before exiting EOP-6A ensures guidance to other procedures would not provide the effect reactor shutdown is not removed. guidance necessary to establish margin for power oscillation mitigation. Before exiting Revision 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 1 Page 8 of 10 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 App. B, step RC/Q-6,RC/Q-7 App. B, Contingency #5 Group Appendix Revision 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 1 Page 9 of 10 Critical Tasks Inhibit ADS prior to automatic ADS valve When control rods fail to scram and energy opening during a failure to Scram. 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 With a Reactor Scram required, reactor not Failure to effect shutdown of the reactor when significance shut down, and conditions for ADS blowdown a RPS setting has been exceeded would are met, INHIBIT ADS to prevent an unnecessarily extend the level of degradation uncontrolled RPV depressurization and cold of the safety of the plant. This could further water injection from low pressure sources to degrade into damage to the principle fission prevent causing a significant power product barriers if left unmitigated. Action to excursion. 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.

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Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 1 Page 10 of 10 Performance Manipulation of ADS A and ADS B Inhibit Operator manipulates keylocked switches for indicator switches on panel 9-3 vertical section. SLC A(B) pump to START on panel 9-5.

Performance Inhibit switches click into the vertical, inhibit SLC A(B) pump red light illuminated, SLC feedback position on panel 9-3. discharge pressure rising, SLC tank level lowering on panel 9-5.

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

Justification The 105 second ADS timer allows sufficient If boron injection is initiated before suppression for the chosen time for the crew to recognize and override pool temperature reaches the BIIT, emergency performance automatic operation of the system. As long RPV depressurization may be precluded at limit as ADS is inhibited before ADS valves open, lower reactor power levels. At higher reactor reactor pressure will not be reduced to the power levels, however, the suppression pool shutoff heads of high volume, cold water heatup rate may become so high that the Hot systems. 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 App. B, step RC/Q-6 App. B, step RC/Q-6 Group Appendix Revision 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 2 Page 1 of 7 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 Revision 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 2 Page 2 of 7 Event Malf. No. Event Event No. Type Description N (BOP) 1 N/A Shift REC pumps TS (CRS)

OR: Auto function on CRD FCV controller fails requiring manual 2 I (ATC) zaicrdfc301 control to be used C (BOP) ms09e C (ATC) 3 OR: A Outboard MSIV 86A closes but leaks by zdipcissws4a (CREW)

TS (CRS)

C (ATC) 4 mc01 A Partial loss of condenser vacuum-scram (CREW) rp01 (a-d)

OR:

Electrical ATWS 5 zdirpssws1 M(CREW) When RPS fails to scram the reactor on a manual scram signal, within two minutes initiate the ARI System.

zdirpssws3a zdirpssws3b FW A line break inside primary containment.

fw18a RCIC spurious isolation M When RPV level lowers to -158 CFZ (TAF) and cannot be 6 rr20a (CREW) maintained above -183 CFZ (MSCWL) and insufficient high rc02 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.

ECCS system valves fail to auto open.

cs02a When operating injection systems cannot maintain RPV cs02b C (BOP) level and ECCS systems fail to automatically align for 8 injection, crew manually aligns ECCS systems for rh04a C (ATC) injection:

• For low pressure ECCS systems, prior to RPV rh04b 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)

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Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 2 Page 3 of 7 Quantitative Attributes Table ES-304-1 Attribute Actual Description Target Malfunctions after 1. Low Pressure ECCS injection valves fail to open.

EOP entry 1-2 2 2. RWCU fails to auto isolate

1. Outboard MSIV closure Abnormal Events 2-4 2 2. Partial loss of condenser vacuum

1. ATWS Major Transients 1-2 2 2. FW line break EOP entries requiring 1. EOP-1A, substantive action 1-2 2 2. EOP-3A EOP contingencies requiring substantive 0-2 1 1. EOP-2A action

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:

EOP based Critical 2-3

• For low pressure ECCS systems, prior to Tasks 3 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 1. None Manipulations N/A 1

1. CRD FCV controller failure

2. Outboard MSIV closure Instrument/ 3. Condenser in-leakage, loss of vacuum Component Failures N/A 5 4. RWCU fails to isolate

5. LP ECCS injection valves fail to open

1. CRD FCV controller failure

2. Outboard MSIV closure Total Malfunctions N/A 5 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.

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Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 2 Page 4 of 7 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.

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Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 2 Page 5 of 7 Critical Tasks When RPS fails to scram the reactor on a When RPV level lowers to -158 CFZ (TAF) manual scram signal, within two minutes and cannot be maintained above -183 CFZ initiate the ARI System. (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 RPS initiates a reactor scram when one or The MSCWL is the lowest RPV water level at significance more monitored parameters exceed their which the covered portion of the reactor core specified limits to preserve the integrity of the will generate sufficient steam to preclude any fuel cladding and the reactor coolant clad temperature in the uncovered portion of pressure boundary (RCPB) and minimize the the core from exceeding 1500F. When water energy that must be absorbed following a level decreases below MSCWL with injection, loss of coolant accident (LOCA). Failure to clad temperatures may exceed 1500F.

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.

Cueing Annunciators 9-5-2/A-1 (A-2) RX SCRAM Corrected Fuel Zone indication (SPDS) falls to CHANNEL A (B) in alarm with RPS -158 and lowering trend continues, and, remaining energized. 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 Operator depresses both manual scram Manipulation of any six SRV controls on panel indicator pushbuttons, or places the Reactor Mode 9-3:

Switch to SHUTDOWN on panel 9-5. SRV-71A SRV-71B SRV-71E SRV-71G SRV-71H SRV-71C SRV-71D SRV-71F Performance RPS Group lights de-energized on panel 9-5. Crew will observe SRV light indication go from feedback Control Rod full -in indication on panel 9-5. green to red, amber pressure switch lights Reactor power trend on nuclear illuminate, reactor pressure lowering on SPDS instrumentation on panel 9-5. and panel 9-3 and 9-5 meters and recorders, and SRV tailpipe temperatures rise on recorder MS-TR-166.

Justification Procedure 2.0.3, Conduct of Operations There is no time limit for effecting complete for the chosen requires upon recognition of a failure of The MSCWL (-183 CFZ) is the lowest RPV performance automatic action, the CRO shall manually water level at which the covered portion of the limit perform those actions necessary to fulfill the reactor core will generate sufficient steam to Revision 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 2 Page 6 of 7 safety function and report the completion of preclude any clad temperature in the the manual action to the CRS as soon as uncovered portion of the core from exceeding possible. Failure of RPS to automatically 1500F. Emergency depressurization is function would involve multiple sensor and allowed when level goes below TAF (-158 sensor relay failures. The complexity of an CFZ) and should be performed, if in the automatic RPS failure would necessarily judgment of the CRS, level cannot be require a short amount of time to diagnose maintained above -183 CFZ. Since it is and validate using control room indications. intended for the scenario supporting this CT to, Two minutes is a reasonable time for early in the event, clearly indicate no high operators to recognize a scram signal, verify pressure injection systems can be made the condition is valid, communicate available to reverse the lowering level trend, conditions to the crew, and insert a manual the crew will have time to communicate and scram, without unnecessarily extending the open 6 SRVs before -183 CFZ.

level of degradation to plant safety.

BWR Owners App. B, step RC-1 App. B, Contingency#1 Group Appendix Revision 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 2 Page 7 of 7 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 Failure to recognize the auto valve alignment significance 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 Manipulation of controls as required to open indicator 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 Red light illuminates and Green light feedback 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 Attempting to align high pressure ECCS for the chosen systems must be performed to determine performance their availability by the time TAF is reached in limit 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 App. B, Contingency 1, step C1-1 Group Appendix Revision 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 3 Page 1 of 7 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 Revision 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 3 Page 2 of 7 Event Malf. No. Event Event No. Type Description N

1 N/A Secure SGT after nitrogen makeup to the torus and drywell.

(BOP) 2 N/A R (ATC) Raise power to 95% with Reactor Recirculation OR C (BOP)

ZDIHVSWEFCI Running Battery Room Exhaust Fan trip, failure of standby fan 3 C[1]

to run, manual start of Essential Ventilation.

ZDIHVSWEFCI TS A[1] (CRS)

C (BOP-ATC)

SRV fails open.

4 ad06c A When a SRV fails open, close the SRV or prior to torus (CREW) bulk temperature reaching 110F, initiate a Reactor Scram TS (CRS)

C (ATC) 5 tu3i A Main Turbine Bearing #9 high vibration.

(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.

M When torus water level cannot be maintained above 9.6',

6 pc08 (CREW) 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.

OR zdimssws1a zdimssws1b C Only 2 SRVs open on emergency depressurization use of 7 zdimssws1a zdimssws4a (BOP) alternate ED systems.

zdimssws1d zdimssws1e (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal (TS) Tech Spec

• Critical Task (As defined in NUREG 1021 Appendix D)

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Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 3 Page 3 of 7 Quantitative Attributes Table ES-304-1 Attribute Actual Description Target Malfunctions after 1. All but 2 SRVs fail to open on emergency 1-2 1 depressurization EOP entry

1. SRV fails open.

Abnormal Events 2-4 2 2. MT Bearing #9 Hi Vibs Major Transients 1-2 1 1. Torus water leak EOP entries requiring 1. EOP-1A 1-2 2 2. EOP-3A substantive action EOP contingencies requiring substantive 0-2 1 1. EOP-2A action

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.

EOP based Critical 3. When torus water level cannot be maintained 2-3 4 above 9.6', scram the reactor prior to torus Tasks 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 1. Raise power 5% with Reactor Recirculation N/A 1 Manipulations

1. SRV fails open.

Instrument/ 2. Battery Room Exhaust fans fail.

N/A 4 3. MT Bearing #9 Hi Vibs Component Failures

4. 6 SRVs fail to open on emergency depressurization

1. SRV sticks open

2. Battery Room Exhaust fans fail.

Total Malfunctions N/A 4 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 Revision 0

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

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Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 3 Page 5 of 7 Critical Tasks When a SRV fails open, close the SRV or When torus water level cannot be prior to torus bulk temperature reaching maintained above 11', prevent HPCI 110F, initiate a Reactor Scram. operation prior to torus water level lowering below 11.0.

EVENT 4 5 Safety Closing the SRV or shutting down the reactor Operation of the HPCI System with its exhaust significance before 110°F in the Suppression Pool discharge device not submerged will directly ensures containment design limits due to pressurize the torus. HPCI operation is heat addition to the suppression pool will not therefore secured, as required, to preclude the be exceeded. 110°F in the Suppression Pool occurrence of this condition. The is both the Technical Specification limit and consequences of not doing so may extend to EOP-3A limit for effecting a reactor scram. failure of the primary containment from over-Tech Spec 3.6.2.1 requires that the Reactor pressurization, and thus, HPCI must be Scram be inserted at 110F. This secured irrespective of adequate core cooling requirement ensures that the unit will be shut concerns.

down at > 110F. The pool is designed to No comparable task regarding RCIC operation absorb decay heat and sensible heat but is provided because:

could be heated beyond design limits by the The exhaust flow rate of RCIC is no greater steam generated if the reactor is not shut than the steam generated by decay heat after down (TS Basis). Per PSTGs, the lowest reactor shutdown. The basis for determining temperature of the Boron Injection Initiation Primary Containment Pressure Limit assumes Temperature (BIIT) is specified as the action the operability of a containment vent capable level (110°F). A single value instead of a of removing decay heat 10 minutes after graph implements the BIIT in this step to reactor shutdown. Thus, any steam simplify the guideline. The BIIT specifies the discharged by RCIC into the torus airspace suppression pool temperature before which can be removed through the primary boron injection must be started. It is the containment vent and will not cause torus greater of: pressure to exceed PCPL even if the RCIC

• The highest suppression pool temperature exhaust is not submerged.

at which initiation of boron injection will Elevated torus pressure will cause the RCIC permit injection of the Hot Shutdown Boron turbine to trip much sooner than the HPCI Weight of boron before suppression pool turbine.

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.

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Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 3 Page 6 of 7 Cueing SRV open indications (solenoid lights, Lowering Torus water level, tailpipe pressure light, tailpipe temperature). approaching 11, as indicated on Step reduction in turbine generator load and SPDS and panel 9-3 indicators PC-steam flow. LRPR-1A and PC-LI-10.

Rising suppression pool temperatures on panel 9-3.

Performance Operator depresses both manual scram Crew stops and prevents HPCI by one of the indicator pushbuttons or places the Reactor Mode following on panel 9-3:

Switch to SHUTDOWN on panel 9-5, prior to

• Depressing and holding the HPCI trip exceeding 110°F in the Suppression Pool; pushbutton, and placing HPCI Aux Oil Pump or the operator closes the SRV IAW 2.4SRV control switch in PTL without exceeding 110°F in the Suppression

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

• Depressing HPCI MANUAL ISOLATION pushbutton, if initiation signal present Performance RPS Group lights de-energized on panel 9-5. HPCI speed lowers to zero on HPCI-SI-2792 feedback Reactor Power trend. on panel 9-3 Control Rod full-in indication. HPCI flow lowers to zero on HPCI-FIC-108 on SRV tailpipe pressure, steam flow, solenoid panel 9-3 lights, step increase in turbine generator load Steam supply isolation valve HPCI-MO-15 and steam flow and/or HPCI-MO-16 control switch green light Anytime when a SRV fails open and the illuminated and red light extinguished on panel actions addressed in Procedure 2.4SRV 9-3.

would be effective in closing the valve OR EOP-3A conditions when actions taken IAW 2.4SRV are ineffective or not attempted.

Justification 110ºF is both the EOP-3A step SP/T-2 limit If torus water level cannot be restored and for the chosen and the TS 3.6.2.1 limit for reactor shutdown maintained above 11 feet is the EOP-3A, step performance to limit heat addition to the suppression pool. SP/L-10 criteria for preventing HPCI operation limit Closing the failed open SRV would also to ensure HPCI exhaust does not directly terminate heat addition to the suppression impinge on the torus air space.

pool.

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

9.60.

EVENT 5 5 Revision 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 3 Page 7 of 7 Safety Energy in the RPV should be discharged The RPV is not permitted to remain at significance outside the primary containment, if possible, pressure if suppression of steam discharged and thereby reduce or limit the energy added from the RPV into the drywell cannot be to the suppression pool if emergency RPV assured. When the downcomer vent openings depressurization becomes necessary. Entry are not adequately submerged, any steam to the RPV Control guideline is therefore discharged from the RPV into the drywell may specified before reaching the elevation of the not condense in the suppression pool before downcomer openings so that the override torus pressure reaches unacceptable levels.

before Step RC/P-1 can be used to anticipate RPV depressurization is required at or before emergency RPV depressurization and rapidly the point at which this low water level condition depressurize the RPV, irrespective of the occurs. This reduces the amount of energy resulting cooldown rate. that may be discharged directly to the torus air Entering the EOP-1A at Step 1 assures that, space to as low as possible.

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.

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

Performance Operator depresses both manual Manipulation of any six SRV controls on panel indicator scram pushbuttons, or places the 9-3:

Reactor Mode Switch to SRV-71A SHUTDOWN on panel 9-5. SRV-71B SRV-71E SRV-71G SRV-71H SRV-71C SRV-71D SRV-71F Performance RPS Group lights de-energized on panel 9-5. Crew will observe SRV light indication go from feedback Control Rod full -in indication on panel 9-5. green to red, amber pressure switch lights Reactor power trend on nuclear illuminate, reactor pressure lowering on SPDS instrumentation on panel 9-5. and panel 9-3 and 9-5 meters and recorders, and SRV tailpipe temperatures rise on recorder MS-TR-166.

Justification Before torus water level drops to 9.6 is the Inability to maintain torus water level above for the chosen EOP-3A, step SP/L-11 criteria for 9.6 is the EOP-3A, step SP/L-12 criteria for performance transitioning to EOP-1A to shut down the transitioning to emergency depressurization.

limit reactor.

BWR Owners App. B, step SP/L-2.1 App. B, step SP/L-2.1 Group Appendix Revision 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 1 of 9 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 Revision 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 2 of 9 Event Malf. No. Event Type Event No. Description 1 N/A N(ATC) Shift RRMG lube oil pumps N (BOP) 2 rf rh14 Place RHR loop B in SPC, Min flow valve de-energizes open.

TS (CRS) 3 rd08b C (ATC) CRD Pump B trip.

C (BOP) 4 rr10a RR Pump A seal #1 leak and RR Pump A trip.

A (CREW)

C (BOP, ATC) rr04b 5 A (CREW) RR Pump A seal #2 leak, vent PC.

rr11a TS (CRS)

FW Line B break in PC-Scram M Initiate drywell sprays when torus pressure exceeds 10 6 fw18b psig, prior to drywell temperature reaching 280F and (CREW) prior to torus pressure exceeding the Pressure Suppression Pressure (PSP) curve.

C 7 hp01 HPCI fails to automatically start.

(BOP)

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 8 NBI various Main Steam Lines before drywell radiation reaches 150 M

R/hr or entering PC Flooding.

(CREW)

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)

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Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 3 of 9 Quantitative Attributes Table ES-304-1 Attribute Actual Description Target Malfunctions after 1. HPCI fails to automatically start.

1-2 1 EOP entry

1. RR pump trip.

Abnormal Events 2-4 2 2. RR seal leakage

1. FW line break inside PC Major Transients 1-2 2 2. Loss of all RPV level instruments EOP entries requiring 1. EOP-1A 1-2 2 2. EOP-3A substantive action EOP contingencies requiring substantive 0-2 1 1. EOP- 2B action

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 EOP based Critical reactor has been depressurized below the shutoff 2-3 4 Tasks 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.

1. Shift RRMG oil pumps Normal Events N/A 2 2. Place RHR Suppression Pool Cooling in service Reactivity N/A N/A 0 Manipulations

1. CRD Pump trip.

Instrument/ 2. RR pump A seal #1 with RR pump trip.

N/A 4 3. RR pump A seal #2 failure.

Component Failures

4. HPCI fails to automatically start

1. CRD Pump trip.

2. RR pump A seal #1 with RR pump trip.

Total Malfunctions N/A 4 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.

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

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Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 5 of 9 Critical Tasks When RPV level cannot be determined When RPV level cannot be determined and and torus level is above 6', open six SRVs the reactor has been depressurized below before drywell radiation reaches 150 R/hr the shutoff head of the respective pump(s),

or entering PC Flooding. 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 Depressurization of the RPV is necessary to Once the SRVs have been opened to significance perform the RPV flooding actions for the depressurize the RPV, injection systems are following reasons: aligned to flood the RPV and establish core The open SRVs establish a path cooling by submergence. The list of flooding from the RPV capable of rejecting methods includes all motor-driven systems energy in excess of decay heat to capable of injecting into the RPV. Any or all of ensure the RPV flooding actions are these systems may be used, as necessary, to successful. flood the RPV to the elevation of the main Reduced RPV pressure results in steam lines. Steam-driven systems are not increased injection flow rates, listed since, with SRVs open and the reactor reducing the total time required to shut down, the RPV will depressurize to below flood the RPV. the turbine stall pressures. Failing to raise Reduced RPV pressure reduces the RPV level to and observable point could water inventory loss through prevent recovery of RPV level above non-isolable leaks and breaks. MSCRWL, resulting in core damage.

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.

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Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 6 of 9 Failing to depressurize could prevent recovery of RPV level above MSCRWL, resulting in core damage Cueing Erratic or inconsistent indication on all RPV Erratic or inconsistent indication on all RPV level indications, and CRS declares RPV level indications, and CRS declare RPV level level cannot be determined. undetermined.

Six ADS valves have been manually opened.

Performance Manipulation of any six SRV controls on Crew establishes injection flow by indicator panel 9-3: manipulating controls as required to start the SRV-71A associated pumps and align system valves for SRV-71B injection using at least two pumps of the SRV-71E following systems:

SRV-71G Main condensate/booster pumps on panel A SRV-71H RHR/LPCI loop A and/or B on panel 9-3 SRV-71C Core spray A and/or B SRV-71D SRV-71F [Operator places affected ECCS pump(s) control switch(es) to START and valve control switches to OPEN (or CLOSE, if necessary)]

Performance Crew will observe SRV light indication go Indication that the RPV is flooded to the main feedback from green to red, amber pressure switch steam lines may include one or more of the lights illuminate, reactor pressure lowering on following indication on panels 9-3, 9-4, 9-5 or SPDS and panel 9-3 and 9-5 meters and field reports by the booth operator:

recorders, and SRV tailpipe temperatures

• Rising RPV pressure rise on recorder MS-TR-166.

• 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 Revision 0

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

• 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 Before 150R/hr in the drywell was chosen LOCA severity should result in a near linear for the chosen because this is an indicator of loss of RPV RPV level reduction that gives the crew an performance level and the shielding effect of the water, initial trend on all level instruments. Failing all limit indicating core exposure, yet it is much lower of the level instruments should occur within than the 2500R/hr trigger point during RPV about 30 seconds and should yield Flooding that indicates gross cladding failure inconsistent indications such that there is no is in progress. Before exiting to PC Flooding doubt level cannot be determined (e.g. LOCA was chosen because the design of the conditions with operation in the possible boiling scenario provides the crew with the means to region of the RPVST curve, minimal RPV restore and maintain adequate core cooling injection, level slowly lowering to -100 CFZ, IAW EOP-2B or 7B, and exiting to SAGs is then all level instruments fail upscale within 10 neither required nor authorized. 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 App. B, Contingency#4 App. B, Contingency #4.

Group Appendix Critical Tasks When RPV level cannot be determined Initiate drywell sprays when torus pressure and at least 1 SRV is manually opened, exceeds 10 psig, prior to drywell isolate MSIVs, MSL drains, HPCI steam temperature reaching 280F and prior to supply, and RCIC steam supply, prior to torus pressure exceeding the Pressure RPV water level rising to the bottom of the Suppression Pressure (PSP) curve.

main steam lines.

EVENT 7 5 Safety Steam lines connected to the RPV are Drywell sprays are initiated in two legs of EOP-significance isolated prior to initiating action to flood the 3A: Temperature and Pressure control.

RPV to preclude damage which may occur from cold water coming in contact with the hot metal, excessive loading of lines or Regarding drywell temperature, if operation of hangers not designed to accommodate the all available drywell cooling is unable to weight of water, and flooding of steam driven terminate increasing drywell temperature equipment (RCIC turbine, main turbine, etc.). before the structural design temperature limit Isolation is performed, however, only if the of 280ºF is reached, drywell sprays are status of SRVs assures the RPV will remain initiated to affect the required drywell depressurized during the flooding evolution.

For non-ATWS, only one SRV open is temperature reduction status of the DSIL and required to meet this condition. 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 Revision 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 8 of 9 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 Rising torus pressure indicated on SPDS and level indications, and CRS declares RPV panel 9-3 recorder PC-LRPR-1A.

level cannot be determined, and SRVs have been manually opened IAW EOP-2B or EOP-Cursor approaching unsafe boundary on PSP 7B for RPV depressurization.

graph display on SPDS.

Performance Crew places the following valve control Aligns torus spray on panel 9-3 using RHR indicator switches to CLOSE: loop A and/or B:

Inboard MSIVs on panel 9-3 MSL Drains on panel 9-4 places CONTMT COOLING 2/3 CORE HPCI steam supply on panel 9-3 VALVE CONTROL PERMISSIVE switch RCIC steam supply on panel 9-4 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 Indication for applicable isolation valves On panel 9-3, RHR pump/valve control switch feedback Green light illuminates and Red light light indication consistent with intended extinguishes. operation (Red - open/running, Green -

closed/stopped).

RHR flow rate rises on recorder RHR-FR-143 and indicator RHR-FI-133A(B.)

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Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 9 of 9 Torus/drywell pressure stabilizes/lowers on SPDS and panel 9-3 recorder PC-LRPR-1A.

Justification Equipment damage due to cold water cannot When torus pressure cannot be maintained for the chosen occur until water level reaches the main below PSP is the EOP-3A, step PC/P-4 criteria performance steam lines. requiring transition to emergency limit depressurization.

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

Group Appendix Revision 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 5 Page 1 of 8 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 Revision 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 5 Page 2 of 8 Event Malf. No. Event Event No. Type Description Withdraw control rods to establish 20-25% bypass valve 1 N/A R (ATC) position C (ATC)

A 2 rd10 (22-23) (CREW) Control rod 22-23 drifts out TS (CRS) 3 nm14a I (ATC) APRM A INOP failure (half scram)

C (BOP)

A 4 rc05 Spurious RCIC initiation (CREW)

TS (CRS)

I (BOP) hp06 5 TS HPCI steam line break and failure to isolate hp09 (CRS) ed05 M 6 Loss of off-site power ed06 (CREW)

DG 2 Fails to automatically start C

(BOP) When high pressure injection systems cannot maintain 7 dg06b RPV level and low pressure ECCS systems fail to A automatically start due to loss of AC power, crew (CREW) manually starts DG2 to energize LP ECCS systems prior to RPV water level falling below -158 CFZ (TAF).**

LOCA, ED on low RPV level When RPV level lowers to -158 CFZ (TAF) and cannot be rr20a M maintained above -183 CFZ (MSCWL) and insufficient 8 high pressure injection systems are available to restore (CREW) level, crew begins to Emergency Depressurize by opening the first of six SRVs before RPV level lowers below -183 CFZ.

Div 1 low pressure ECCS pumps fail to automatically start rh08a C When high pressure injection systems cannot maintain 9 rh08c RPV level and low pressure ECCS pumps fail to (BOP) automatically start, crew manually starts pumps to align cs06a 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.

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Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 5 Page 3 of 8 Quantitative Attributes Table ES-304-1 Attribute Actual Description Target

1. DG 2 fails to automatically start on LOOP.

Malfunctions after 1-2 2 2. RHR Loop A and Core Spray A pumps fail to auto EOP entry start post LOCA.

1. Control Rod drifts outward.

Abnormal Events 2-4 2 2. RCIC spurious initiation.

1. LOOP Major Transients 1-2 2

2. LOCA EOP entries requiring 1. EOP-1A 1-2 2 substantive action 2. EOP-3A EOP contingencies requiring substantive 0-2 1 1. EOP-2A action

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 EOP based Critical 2-3 3 systems fail to automatically start due to loss of Tasks 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 N/A 1 1. Withdraw control rods Manipulations

1. Control rod drifts out

2. APRM failure

3. Spurious RCIC initiation Instrument/

N/A 6 4. HPCI steam line break and failure to isolate.

Component Failures

5. Diesel Generator 2 fails to automatically start.

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

1. Control rod drifts out

2. APRM failure

3. Spurious RCIC initiation Total Malfunctions N/A 6 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.

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Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 5 Page 4 of 8 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 Revision 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 5 Page 5 of 8 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.

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Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 5 Page 6 of 8 Critical Tasks When high pressure injection systems When RPV level lowers to -158 CFZ (TAF) cannot maintain RPV level and low and cannot be maintained above -183 CFZ pressure ECCS systems fail to (MSCWL) and insufficient high pressure automatically start due to loss of AC injection systems are available to restore power, crew manually starts DG2 to level, crew begins to Emergency energize LP ECCS systems prior to RPV Depressurize by opening the first of six water level falling below -158 CFZ (TAF) SRVs before RPV level lowers below -183 CFZ.

EVENT 7 8 Safety Failure to recognize the auto start not The MSCWL is the lowest RPV water level at significance occurring and energizing of the safety bus, which the covered portion of the reactor core and failure to take manual action per will generate sufficient steam to preclude any Procedure 5.3EMPWR will result in clad temperature in the uncovered portion of unavailability of safety-related equipment the core from exceeding 1500F. When water necessary to provide adequate core cooling, level decreases below MSCWL with injection, otherwise resulting in core damage and a clad temperatures may exceed 1500F.

large offsite release.

Cueing Indication and/or annunciation that all ac Corrected Fuel Zone indication (SPDS) falls to emergency buses are de-energized -158 and lowering trend continues, and,

• Bus energized lamps extinguished before -158 CFZ is reached, initial conditions,

• Circuit breaker position field reports, and control room indications

• Bus voltage convey that adequate high pressure injection

• EDG status cannot be restored before level falls below -

Control room lighting dimmed 183 CFZ Performance Manipulation of controls as required to Manipulation of any six SRV controls on panel indicator energize Div 1(2) AC emergency bus from 9-3:

panel C: SRV-71A SRV-71H Operator places DIESEL GEN 1(2) BKR SRV-71B SRV-71C EG1(2) to CLOSE on panel C SRV-71E SRV-71D SRV-71G SRV-71F Performance Crew will observe light indication for Crew will observe SRV light indication go from feedback equipment powered by Division 1(2) AC green to red, amber pressure switch lights illuminate on panel 9-3 and bus voltage illuminate, reactor pressure lowering on SPDS

~4200V on panel C and panel 9-3 and 9-5 meters and recorders, and SRV tailpipe temperatures rise on recorder MS-TR-166.

Justification Attempting to start ECCS systems must be There is no time limit for effecting complete for the chosen performed to determine their availability by The MSCWL (-183 CFZ) is the lowest RPV performance the time TAF is reached in order to properly water level at which the covered portion of the limit implement EOP-1A decision steps regarding reactor core will generate sufficient steam to restoring and maintaining RPV level. 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 App. B, Contingency#1 App. B, Contingency#1 Group Appendix Revision 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 5 Page 7 of 11 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 Failure to recognize the auto start not significance 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 Manipulation of controls as required to start indicator 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 Crew will observe Red light illuminate and feedback Green light extinguish for the affected ECCS pump(s) on panel 9-3 Justification Attempting to start ECCS systems must be for the chosen performed to determine their availability by performance the time TAF is reached in order to properly limit implement EOP-1A decision steps regarding restoring and maintaining RPV level.

BWR Owners App. B, Contingency#1 Group Appendix Revision 0