ML20241A241

1-CN-2020-04-Draft Outlines 1-2-3

ML20241A241
Person / Time
Site:	Cooper
Issue date:	08/04/2020
From:	Greg Werner Operations Branch IV
To:	Nebraska Public Power District (NPPD)
References
Download: ML20241A241 (64)

Text

ES-401 1 Form ES-401-1 Facility: Cooper Nuclear Station Date of Exam: 04/2020 Tier Group RO K/A Category Points SRO-Only Points K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G* Total A2 G* Total

1. 1 1 4 4 4 4 3 20 7 Emergency and N/A N/A 2 1 1 1 2 1 1 7 3 Abnormal Plant Evolutions Tier Totals 2 5 5 6 5 4 27 10 1 2 2 2 3 2 2 2 3 3 2 3 26 5 2.

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

3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 2 3 2 3 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 outline sections (i.e., except for one category in Tier 3 of the SRO-only section, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 radiation control K/A is allowed if it 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 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. 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 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

• These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan. They are not required to be included when using earlier revisions of the K/A catalog.

• • These systems/evolutions may be eliminated from the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan.

Rev 2

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

E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G* K/A Topic(s) IR #

295001 (APE 1) Partial or Complete Loss of Knowledge of the reasons for the following 2.8 38 Forced Core Flow Circulation / 1 & 4 responses as they apply to PARTIAL OR X COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION:

AK3.03 Idle loop flow 295003 (APE 3) Partial or Complete Loss of Ability to determine and/or interpret the 3.9 28 AC Power / 6 following as they apply to PARTIAL OR X COMPLETE LOSS OF A.C. POWER:

AA2.05 Whether a partial or complete loss of A.C. power has occurred 295004 (APE 4) Partial or Total Loss of DC X 2.2.22 Knowledge of limiting conditions for 4.0 52 Power / 6 operations and safety limits.

295005 (APE 5) Main Turbine Generator Trip / Ability to operate and/or monitor the 3.1 58 3 following as they apply to MAIN TURBINE X

GENERATOR TRIP:

AA1.01 Recirculation system: Plant-Specific 295006 (APE 6) Scram / 1 Ability to operate and/or monitor the 3.7 48 following as they apply to SCRAM:

X AA1.03 Reactor/turbine pressure regulating system 295016 (APE 16) Control Room Abandonment Knowledge of the interrelations between 4.4 45

/7 CONTROL ROOM ABANDONMENT and X the following:

AK2.01 Remote shutdown panel: Plant-Specific 295018 (APE 18) Partial or Complete Loss of Knowledge of the reasons for the following 3.3 46 CCW / 8 responses as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT X

COOLING WATER:

AK3.06 Increasing cooling water flow to heat exchangers 295019 (APE 19) Partial or Complete Loss of Ability to determine and/or interpret the 3.5 60 Instrument Air / 8 following as they apply to PARTIAL OR X

COMPLETE LOSS OF INSTRUMENT AIR:

AA2.01 Instrument air system pressure 295021 (APE 21) Loss of Shutdown Cooling / Ability to determine and/or interpret the 3.2 29 4 following as they apply to LOSS OF X

SHUTDOWN COOLING:

AA2.06 Reactor pressure 295023 (APE 23) Refueling Accidents / 8 Ability to determine and/or interpret the 3.6 42 following as they apply to REFUELING X

ACCIDENTS:

AA2.01 Area radiation levels 295024 High Drywell Pressure / 5 Knowledge of the reasons for the following 4.0 32 responses as they apply to HIGH X

DRYWELL PRESSURE:

EK3.06 Reactor SCRAM Rev 2

ES-401 3 Form ES-401-1 295025 (EPE 2) High Reactor Pressure / 3 Knowledge of the interrelations between 3.9 51 HIGH REACTOR PRESSURE and the X

following:

EK2.09 Reactor power 295026 (EPE 3) Suppression Pool High Water Knowledge of the operational implications of 3.5 49 Temperature / 5 the following concepts as they apply to X SUPPRESSION POOL HIGH WATER TEMPERATURE:

EK1.02 Steam condensation 295027 (EPE 4) High Containment N/A for CNS Temperature (Mark III Containment Only) / 5 295028 (EPE 5) High Drywell Temperature Knowledge of the interrelations between 3.6 41 (Mark I and Mark II only) / 5 HIGH DRYWELL TEMPERATURE and the X

following:

EK2.04 Drywell ventilation 295030 (EPE 7) Low Suppression Pool Water X 2.2.40 Ability to apply Technical 3.4 44 Level / 5 Specifications for a system.

295031 (EPE 8) Reactor Low Water Level / 2 X 2.4.3 Ability to identify post-accident 3.7 57 instrumentation.

295037 (EPE 14) Scram Condition Present Ability to operate and/or monitor the 3.9 36 and Reactor Power Above APRM Downscale following as they apply to SCRAM or Unknown / 1 CONDITION PRESENT AND REACTOR X

POWER ABOVE APRM DOWNSCALE OR UNKNOWN:

EA1.07 RMCS: Plant-Specific 295038 (EPE 15) High Offsite Radioactivity Knowledge of the reasons for the following 3.7 34 Release Rate / 9 responses as they apply to HIGH OFF-SITE X RELEASE RATE:

EK3.03 Control room ventilation isolation:

Plant-Specific 600000 (APE 24) Plant Fire On Site / 8 Ability to operate and / or monitor the 2.6 64 following as they apply to PLANT FIRE ON X SITE:

AA1.08 Fire fighting equipment used on each class of fire 700000 (APE 25) Generator Voltage and Knowledge of the interrelations between 3.1 39 Electric Grid Disturbances / 6 GENERATOR VOLTAGE AND ELECTRIC X

GRID DISTURBANCES and the following:

AK2.02 Breakers, relays K/A Category Totals: 1 4 4 4 4 3 Group Point Total: 20/7 Rev 2

ES-401 4 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant EvolutionsTier 1/Group 2 (RO)

E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G* K/A Topic(s) IR #

295002 (APE 2) Loss of Main Condenser Vacuum / 3 295007 (APE 7) High Reactor Pressure / 3 Ability to operate and/or monitor the 3.7 33 following as they apply to HIGH REACTOR X PRESSURE:

AA1.05 Reactor/turbine pressure regulating system 295008 (APE 8) High Reactor Water Level / 2 295009 (APE 9) Low Reactor Water Level / 2 295010 (APE 10) High Drywell Pressure / 5 295011 (APE 11) High Containment Temperature (Mark III Containment only) / 5 295012 (APE 12) High Drywell Temperature / Ability to determine and/or interpret the 2.8 40 5 following as they apply to HIGH DRYWELL X

TEMPERATURE:

AA2.03 Drywell humidity: Plant-Specific 295013 (APE 13) High Suppression Pool X 2.4.18 Knowledge of the specific bases for 3.3 55 Temperature. / 5 EOPs.

295014 (APE 14) Inadvertent Reactivity Addition / 1 295015 (APE 15) Incomplete Scram / 1 295017 (APE 17) Abnormal Offsite Release Knowledge of the operational implications 3.3 35 Rate / 9 of the following concepts as they apply to HIGH OFF-SITE RELEASE RATE:

AK1.03 Meteorological effects on off-site release X X Knowledge of the interrelations between HIGH OFF-SITE RELEASE RATE and the following:

AK2.10 Process radiation monitoring system 295020 (APE 20) Inadvertent Containment Isolation / 5 & 7 295022 (APE 22) Loss of Control Rod Drive Knowledge of the interrelations between 3.6 62 Pumps / 1 LOSS OF CRD PUMPS and the following:

AK2.03 Accumulator pressures X X Knowledge of the operational implications of the following concepts as they apply to LOSS OF CRD PUMPS:

AK1.02 Reactivity control 295029 (EPE 6) High Suppression Pool Water Level / 5 Rev 2

ES-401 5 Form ES-401-1 295032 (EPE 9) High Secondary Containment Knowledge of the reasons for the following 3.8 53 Area Temperature / 5 responses as they apply to HIGH X SECONDARY CONTAINMENT AREA TEMPERATURE:

EK3.03 Isolating affected systems 295033 (EPE 10) High Secondary Containment Area Radiation Levels / 9 295034 (EPE 11) Secondary Containment Ventilation High Radiation / 9 295035 (EPE 12) Secondary Containment Ability to operate and/or monitor the 3.8 56 High Differential Pressure / 5 following as they apply to SECONDARY X CONTAINMENT HIGH DIFFERENTIAL PRESSURE:

EA1.02 SBGT/FRVS 295036 (EPE 13) Secondary Containment High Sump/Area Water Level / 5 500000 (EPE 16) High Containment Hydrogen Concentration / 5 K/A Category Point Totals: 1 1 1 2 1 1 Group Point Total: 7/3 Rev 2

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

System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G* K/A Topic(s) IR #

203000 (SF2, SF4 RHR/LPCI) 4.2 26 Ability to predict and/or monitor changes in RHR/LPCI: Injection Mode parameters associated with operating the X RHR/LPCI: INJECTION MODE (PLANT SPECIFIC) controls including:

A1.01 Reactor water level 205000 (SF4 SCS) Shutdown Cooling 2.6 19 Knowledge of the effect that a loss or malfunction of the SHUTDOWN COOLING X SYSTEM (RHR SHUTDOWN COOLING MODE) will have on following:

K3.05 Fuel pool cooling assist: Plant-Specific 206000 (SF2, SF4 HPCIS) 4.2 10 2.4.35 Knowledge of local auxiliary operator High-Pressure Coolant Injection tasks during an emergency and the resultant X operational effects.

2.4.31 Knowledge of annunciator alarms, indications, or response procedures.

207000 (SF4 IC) Isolation N/A for CNS (Emergency) Condenser 209001 (SF2, SF4 LPCS) 2.8 25 Knowledge of the operational implications of Low-Pressure Core Spray the following concepts as they apply to LOW X

PRESSURE CORE SPRAY SYSTEM:

K5.04 Heat removal (transfer) mechanisms 209002 (SF2, SF4 HPCS)

N/A for CNS High-Pressure Core Spray 211000 (SF1 SLCS) Standby Liquid 3.2 3 Ability to (a) predict the impacts of the Control following on the STANDBY LIQUID CONTROL SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

X X A2.03 A.C. power failures 4.0 17 Ability to monitor automatic operations of the STANDBY LIQUID CONTROL SYSTEM including:

A3.06 RWCU system isolation: Plant-Specific 212000 (SF7 RPS) Reactor Protection 4.0 20 Ability to (a) predict the impacts of the following on the REACTOR PROTECTION SYSTEM; and (b) based on those predictions, X use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

A2.11 Main steamline isolation valve closure Rev 2

ES-401 7 Form ES-401-1 215003 (SF7 IRM) 3.0 13 Knowledge of the operational implications of Intermediate-Range Monitor the following concepts as they apply to X INTERMEDIATE RANGE MONITOR (IRM)

SYSTEM:

K5.03 Changing detector position 215004 (SF7 SRMS) Source-Range 3.7 11 Knowledge of the effect that a loss or Monitor malfunction of the SOURCE RANGE X MONITOR (SRM) SYSTEM will have on following:

K3.04 Reactor power and indication 215005 (SF7 PRMS) Average Power 3.7 24 Knowledge of AVERAGE POWER RANGE Range Monitor/Local Power Range MONITOR/LOCAL POWER RANGE Monitor X MONITOR SYSTEM design feature(s) and/or interlocks which provide for the following:

K4.07 Flow biased trip setpoints 217000 (SF2, SF4 RCIC) Reactor 3.4 21 Knowledge of the effect that a loss or Core Isolation Cooling malfunction of the following will have on the X REACTOR CORE ISOLATION COOLING SYSTEM (RCIC):

K6.01 Electrical power 218000 (SF3 ADS) Automatic 4.0 2 Knowledge of the physical connections and/or Depressurization cause-effect relationships between AUTOMATIC DEPRESSURIZATION X

SYSTEM and the following:

K1.02 Low pressure core spray: Plant-Specific 223002 (SF5 PCIS) Primary 2.9 14 Ability to (a) predict the impacts of the Containment Isolation/Nuclear Steam following on the PRIMARY CONTAINMENT Supply Shutoff ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

X X A2.02 D.C. electrical distribution failures Ability to monitor automatic operations of the 2.5 6 PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF including:

A3.03 SPDS/ERIS/CRIDS/GDS: Plant-Specific 239002 (SF3 SRV) Safety Relief 3.7 4 Ability to predict and/or monitor changes in Valves parameters associated with operating the RELIEF/SAFETY VALVES controls including:

X X A1.06 Reactor power 2.1.32 Ability to explain and apply system 3.8 16 limits and precautions.

Rev 2

ES-401 8 Form ES-401-1 259002 (SF2 RWLCS) Reactor Water 3.5 15 Ability to manually operate and/or monitor in Level Control the control room:

X A4.11 High level lockout reset controls: Plant-Specific 261000 (SF9 SGTS) Standby Gas 3.2 7 Knowledge of the physical connections and/or Treatment cause-effect relationships between X STANDBY GAS TREATMENT SYSTEM and the following:

K1.02 Drywell 262001 (SF6 AC) AC Electrical 3.3 8 Knowledge of electrical power supplies to the Distribution X following:

K2.01 Off-site sources of power 262002 (SF6 UPS) Uninterruptable 3.1 5 Knowledge of UNINTERRUPTABLE POWER Power Supply (AC/DC)

SUPPLY (A.C./D.C.) design feature(s) and/or interlocks which provide for the following:

K4.01 Transfer from preferred power to alternate power supplies X X 2.7 12 Knowledge of the effect that a loss or malfunction of the following will have on the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.):

K6.03 Static inverter 263000 (SF6 DC) DC Electrical 3.3 23 Ability to manually operate and/or monitor in Distribution the control room:

X A4.01 Major breakers and control power fuses: Plant-Specific 264000 (SF6 EGE) Emergency 3.1 22 Ability to monitor automatic operations of the Generators (Diesel/Jet) EDG EMERGENCY GENERATORS (DIESEL/JET) including:

X X A3.06 Cooling water system operation 4.2 18 2.4.46 Ability to verify that the alarms are consistent with the plant conditions.

300000 (SF8 IA) Instrument Air 3.0 1 Knowledge of (INSTRUMENT AIR SYSTEM) design feature(s) and or interlocks which X

provide for the following:

K4.02 Cross-over to other air systems 400000 (SF8 CCS) Component 2.9 9 Knowledge of electrical power supplies to the Cooling Water X following:

K2.02 CCW valves K/A Category Point Totals: 2 2 2 3 2 2 2 3 3 2 3 Group Point Total: 26/5 Rev 2

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

System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G* K/A Topic(s) IR #

201001 (SF1 CRDH) CRD Hydraulic Knowledge of CONTROL ROD 2.9 65 DRIVE HYDRAULIC SYSTEM X design feature(s) and/or interlocks which provide for the following:

K4.12 Controlling CRD system flow 201002 (SF1 RMCS) Reactor Manual Control 201003 (SF1 CRDM) Control Rod and Drive Knowledge of the effect that a loss 3.0 54 Mechanism or malfunction of the following will X have on the CONTROL ROD AND DRIVE MECHANISM:

K6.02 Reactor pressure 201004 (SF7 RSCS) Rod Sequence Control 201005 (SF1, SF7 RCIS) Rod Control and Information 201006 (SF7 RWMS) Rod Worth Minimizer 202001 (SF1, SF4 RS) Recirculation 202002 (SF1 RSCTL) Recirculation Flow Control 204000 (SF2 RWCU) Reactor Water Cleanup 2.4.4 Ability to recognize abnormal 4.5 37 indications for system operating X parameters that are entry-level conditions for emergency and abnormal operating procedures.

214000 (SF7 RPIS) Rod Position Information 215001 (SF7 TIP) Traversing In-Core Probe Ability to (a) predict the impacts of 2.7 61 the following on the TRAVERSING IN-CORE PROBE; and (b) based on those predictions, use procedures to X correct, control, or mitigate the consequences of those abnormal conditions or operations:

A2.01 Low reactor water level: Mark-I&II (Not-BWR1) 215002 (SF7 RBMS) Rod Block Monitor 216000 (SF7 NBI) Nuclear Boiler A2.09, Ability to (a) predict the 3.1 61 Instrumentation impacts of the following on the NUCLEAR BOILER INSTRUMENTATION ; and (b)

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

Jet pump flow: Design-Specific 219000 (SF5 RHR SPC) RHR/LPCI: Knowledge of the operational 2.9 43 Torus/Suppression Pool Cooling Mode implications of the following concepts as they apply to X

RHR/LPCI: TORUS/SUPPRESSION POOL COOLING MODE:

K5.04 Heat exchanger operation 223001 (SF5 PCS) Primary Containment and Auxiliaries Rev 2

ES-401 10 Form ES-401-1 226001 (SF5 RHR CSS) RHR/LPCI:

Containment Spray Mode 230000 (SF5 RHR SPS) RHR/LPCI: Knowledge of the physical 3.0 50 Torus/Suppression Pool Spray Mode connections and/or cause-effect relationships between RHR/LPCI:

X TORUS/SUPPRESSION POOL SPRAY MODE and the following:

K1.06 Keep fill system 233000 (SF9 FPCCU) Fuel Pool Cooling/Cleanup 234000 (SF8 FH) Fuel-Handling Equipment 239001 (SF3, SF4 MRSS) Main and Reheat Knowledge of the effect that a loss 2.8 59 Steam or malfunction of the MAIN AND X REHEAT STEAM SYSTEM will have on following:

K3.05 Condenser air removal 239003 (SF9 MSVLCS) Main Steam Isolation Valve Leakage Control 241000 (SF3 RTPRS) Reactor/Turbine Pressure Regulating 245000 (SF4 MTGEN) Main Turbine Generator/Auxiliary 256000 (SF2 CDS) Condensate 259001 (SF2 FWS) Feedwater Ability to predict and/or monitor 2.7 31 changes in parameters associated with operating the REACTOR X

FEEDWATER SYSTEM controls including:

A1.06 Feedwater heater level 268000 (SF9 RW) Radwaste 271000 (SF9 OG) Offgas Ability to manually operate and/or 2.9 63 X monitor in the control room:

A4.02 System flows 272000 (SF7, SF9 RMS) Radiation Monitoring 286000 (SF8 FPS) Fire Protection Knowledge of electrical power 2.9 47 X supplies to the following:

K2.02 Pumps 288000 (SF9 PVS) Plant Ventilation Knowledge of the effect that a loss 2.8 27 or malfunction of the PLANT VENTILATION SYSTEMS will have X

on following:

K3.01 Secondary containment temperature: Plant-Specific 290001 (SF5 SC) Secondary Containment 290003 (SF9 CRV) Control Room Ventilation Ability to monitor automatic 3.3 30 operations of the CONTROL ROOM X

HVAC including:

A3.01 Initiation/reconfiguration 290002 (SF4 RVI) Reactor Vessel Internals K/A Category Point Totals: 1 1 2 1 1 1 1 1 1 1 1 Group Point Total: 12/3 Rev 2

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Cooper Nuclear Station Date of Exam: 04/2020 Category K/A # Topic RO SRO-only IR # IR #

2.1.20 Ability to interpret and execute procedure steps. 4.6 74

1. Conduct of 2.1.37 Knowledge of procedures, guidelines, or limitations 4.3 67 Operations associated with reactivity management.

Subtotal 2 2.2.7 Knowledge of the process for conducting special or 3.9 72 infrequent tests.

2.2.14 2.2.14 Knowledge of the process for controlling

2. Equipment equipment configuration or status.

Control 2.2.40 Ability to apply technical specifications for a system. 3.4 75 2.2.43 Knowledge of the process used to track inoperable 3.0 73 alarms.

Subtotal 3 2.3.5 Ability to use radiation monitoring systems, such as 2.9 71 fixed radiation monitors and alarms, portable survey

3. Radiation instruments, personnel monitoring equipment, etc.

Control 2.3.4 Knowledge of radiation exposure limits under normal or 3.2 69 emergency conditions.

Subtotal 2 2.4.47 Ability to diagnose and recognize trends in an accurate 4.2 66 and timely manner utilizing the appropriate control room reference material.

4. Emergency 2.4.49 Ability to perform without reference to procedures those 4.6 68 Procedures/Plan actions that require immediate operation of system components and controls.

2.4.3 Ability to identify post-accident instrumentation. 3.7 70 Subtotal 3 Tier 3 Point Total 10 7 Rev 2

ES-401 12 Form ES-401-1 Facility: Cooper Nuclear Station Date of Exam: 04/2020 Tier Group RO K/A Category Points SRO-Only Points K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G* Total A2 G* Total

1. 1 20 4 3 7 Emergency and N/A N/A 2 7 2 1 3 Abnormal Plant Evolutions Tier Totals 27 6 4 10 1 26 2 3 5 2.

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

3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 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 outline sections (i.e., except for one category in Tier 3 of the SRO-only section, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 radiation control K/A is allowed if it 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 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. 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 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

• These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan. They are not required to be included when using earlier revisions of the K/A catalog.

• • These systems/evolutions may be eliminated from the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan.

Rev 2

ES-401 13 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant EvolutionsTier 1/Group 1 (SRO)

E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G* K/A Topic(s) IR #

295001 (APE 1) Partial or Complete Loss of Ability to determine and/or interpret the 3.1 82 Forced Core Flow Circulation / 1 & 4 following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE X

FLOW CIRCULATION:

AA2.04 Individual jet pump flows: Not-BWR-1&2 295003 (APE 3) Partial or Complete Loss of AC Power / 6 295004 (APE 4) Partial or Total Loss of DC Power / 6 295005 (APE 5) Main Turbine Generator Trip /

3 295006 (APE 6) Scram / 1 295016 (APE 16) Control Room Abandonment

/7 295018 (APE 18) Partial or Complete Loss of 2.4.35 Knowledge of local auxiliary operator 4.0 90 CCW / 8 X tasks during an emergency and the resultant operational effects.

295019 (APE 19) Partial or Complete Loss of Ability to determine and/or interpret the 3.7 91 Instrument Air / 8 following as they apply to PARTIAL OR X COMPLETE LOSS OF INSTRUMENT AIR:

AA2.02 Status of safety-related instrument air system loads (see AK2.1 - AK2.19) 295021 (APE 21) Loss of Shutdown Cooling /

4 295023 (APE 23) Refueling Accidents / 8 295024 High Drywell Pressure / 5 2.4.4 Ability to recognize abnormal 4.7 81 indications for system operating parameters X that are entry-level conditions for emergency and abnormal operating procedures.

295025 (EPE 2) High Reactor Pressure / 3 Ability to determine and/or interpret the 3.8 77 following as they apply to HIGH REACTOR X

PRESSURE:

EA2.06 Reactor water level 295026 (EPE 3) Suppression Pool High Water Temperature / 5 295027 (EPE 4) High Containment NA for CNS Temperature (Mark III Containment Only) / 5 295028 (EPE 5) High Drywell Temperature X 2.2.12 Knowledge of surveillance 4.1 83 (Mark I and Mark II only) / 5 procedures.

295030 (EPE 7) Low Suppression Pool Water Level / 5 295031 (EPE 8) Reactor Low Water Level / 2 Rev 2

ES-401 14 Form ES-401-1 295037 (EPE 14) Scram Condition Present and Reactor Power Above APRM Downscale or Unknown / 1 295038 (EPE 15) High Offsite Radioactivity Release Rate / 9 600000 (APE 24) Plant Fire On Site / 8 700000 (APE 25) Generator Voltage and Ability to determine and/or interpret the 4.0 88 Electric Grid Disturbances / 6 following as l they apply to GENERATOR VOLTAGE AND ELECTRIC GRID X

DISTURBANCES:

AA2.07 Operational status of engineered safety features K/A Category Totals: 4 3 Group Point Total: 20/7 Rev 2

ES-401 15 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant EvolutionsTier 1/Group 2 (SRO)

E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G* K/A Topic(s) IR #

295002 (APE 2) Loss of Main Condenser X 2.2.42 Ability to recognize system 4.6 87 Vacuum / 3 parameters that are entry-level conditions for Technical Specifications.

295007 (APE 7) High Reactor Pressure / 3 295008 (APE 8) High Reactor Water Level / 2 295009 (APE 9) Low Reactor Water Level / 2 295010 (APE 10) High Drywell Pressure / 5 X Ability to determine and/or interpret the 3.9 89 following as they apply to HIGH DRYWELL PRESSURE:

AA2.02 Drywell pressure 295011 (APE 11) High Containment Temperature (Mark III Containment only) / 5 295012 (APE 12) High Drywell Temperature /

5 295013 (APE 13) High Suppression Pool Temperature. / 5 295014 (APE 14) Inadvertent Reactivity Addition / 1 295015 (APE 15) Incomplete Scram / 1 X Ability to determine and/or interpret the 4.2 85 following as they apply to INCOMPLETE SCRAM:

AA2.02 Control rod position 295017 (APE 17) Abnormal Offsite Release Rate / 9 295020 (APE 20) Inadvertent Containment Isolation / 5 & 7 295022 (APE 22) Loss of Control Rod Drive Pumps / 1 295029 (EPE 6) High Suppression Pool Water Level / 5 295032 (EPE 9) High Secondary Containment Area Temperature / 5 295033 (EPE 10) High Secondary Containment Area Radiation Levels / 9 295034 (EPE 11) Secondary Containment Ventilation High Radiation / 9 295035 (EPE 12) Secondary Containment High Differential Pressure / 5 295036 (EPE 13) Secondary Containment High Sump/Area Water Level / 5 500000 (EPE 16) High Containment Hydrogen Concentration / 5 K/A Category Point Totals: 2 1 Group Point Total: 7/3 Rev 2

ES-401 16 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant SystemsTier 2/Group 1 (SRO)

System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G* K/A Topic(s) IR #

203000 (SF2, SF4 RHR/LPCI)

RHR/LPCI: Injection Mode 205000 (SF4 SCS) Shutdown Cooling 206000 (SF2, SF4 HPCIS)

High-Pressure Coolant Injection 207000 (SF4 IC) Isolation (Emergency) Condenser 209001 (SF2, SF4 LPCS) 3.4 84 Ability to (a) predict the impacts of the Low-Pressure Core Spray following on the LOW PRESSURE CORE SPRAY SYSTEM; and (b) based on those X predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

A2.01 Pump trips 209002 (SF2, SF4 HPCS)

High-Pressure Core Spray 211000 (SF1 SLCS) Standby Liquid Control 212000 (SF7 RPS) Reactor Protection 215003 (SF7 IRM)

Intermediate-Range Monitor 215004 (SF7 SRMS) Source-Range Monitor 215005 (SF7 PRMS) Average Power 4.5 78 2.2.38 Knowledge of conditions and Range Monitor/Local Power Range X limitations in the facility license.

Monitor 217000 (SF2, SF4 RCIC) Reactor Core Isolation Cooling 218000 (SF3 ADS) Automatic Depressurization 223002 (SF5 PCIS) Primary Containment Isolation/Nuclear Steam Supply Shutoff 239002 (SF3 SRV) Safety Relief Valves 259002 (SF2 RWLCS) Reactor Water 4.1 76 2.4.9 Knowledge of low power/shutdown Level Control implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.

X 2.4.30 Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.

261000 (SF9 SGTS) Standby Gas Treatment Rev 2

ES-401 17 Form ES-401-1 262001 (SF6 AC) AC Electrical Distribution 262002 (SF6 UPS) Uninterruptable Power Supply (AC/DC) 263000 (SF6 DC) DC Electrical Distribution 264000 (SF6 EGE) Emergency Generators (Diesel/Jet) EDG 300000 (SF8 IA) Instrument Air 4.7 86 X 2.2.40 Ability to apply Technical Specifications for a system.

400000 (SF8 CCS) Component 3.4 92 Ability to (a) predict the impacts of the Cooling Water following on the CCWS and (b) based on those predictions, use procedures to correct, X

control, or mitigate the consequences of those abnormal operation:

A2.01 Loss of CCW pump K/A Category Point Totals: 2 3 Group Point Total: 26/5 Rev 2

ES-401 18 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant SystemsTier 2/Group 2 (SRO)

System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G* K/A Topic(s) IR #

201001 (SF1 CRDH) CRD Hydraulic 201002 (SF1 RMCS) Reactor Manual Control 201003 (SF1 CRDM) Control Rod and Drive Mechanism 201004 (SF7 RSCS) Rod Sequence Control 201005 (SF1, SF7 RCIS) Rod Control and Information 201006 (SF7 RWMS) Rod Worth Minimizer 202001 (SF1, SF4 RS) Recirculation 202002 (SF1 RSCTL) Recirculation Flow Control 204000 (SF2 RWCU) Reactor Water Cleanup 214000 (SF7 RPIS) Rod Position Information 215001 (SF7 TIP) Traversing In-Core Probe X 2.4.4 Ability to recognize abnormal 4.7 80 indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.

215002 (SF7 RBMS) Rod Block Monitor X 2.2.22 Knowledge of limiting 4.7 79 conditions for operations and safety limits.

216000 (SF7 NBI) Nuclear Boiler Instrumentation 219000 (SF5 RHR SPC) RHR/LPCI:

Torus/Suppression Pool Cooling Mode 223001 (SF5 PCS) Primary Containment and Auxiliaries 226001 (SF5 RHR CSS) RHR/LPCI:

Containment Spray Mode 230000 (SF5 RHR SPS) RHR/LPCI:

Torus/Suppression Pool Spray Mode 233000 (SF9 FPCCU) Fuel Pool Cooling/Cleanup 234000 (SF8 FH) Fuel-Handling Equipment 239001 (SF3, SF4 MRSS) Main and Reheat Steam 239003 (SF9 MSVLCS) Main Steam Isolation Valve Leakage Control 241000 (SF3 RTPRS) Reactor/Turbine Pressure Regulating 245000 (SF4 MTGEN) Main Turbine Generator/Auxiliary 256000 (SF2 CDS) Condensate Rev 2

ES-401 19 Form ES-401-1 259001 (SF2 FWS) Feedwater X Ability to (a) predict the impacts of 3.3 93 the following on the REACTOR FEEDWATER SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

A2.02 Feedwater heater isolation 268000 (SF9 RW) Radwaste 271000 (SF9 OG) Offgas 272000 (SF7, SF9 RMS) Radiation Monitoring 286000 (SF8 FPS) Fire Protection 288000 (SF9 PVS) Plant Ventilation 290001 (SF5 SC) Secondary Containment 290003 (SF9 CRV) Control Room Ventilation 290002 (SF4 RVI) Reactor Vessel Internals K/A Category Point Totals: 1 2 Group Point Total: 12/3 Rev 2

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Cooper Nuclear Station Date of Exam: 04/2020 Category K/A # Topic RO SRO-only IR # IR #

2.1.4 Knowledge of individual licensed operator 3.8 98 responsibilities related to shift staffing, such as medical

1. Conduct of requirements, no-solo operation, maintenance of Operations active license status, 10CFR55, etc.

2.1.35 Knowledge of fuel handling responsibilities of SROs. 3.9 100 Subtotal 2 2.2.

2.2.38 Knowledge of conditions and limitations in the facility 4.5 95

2. Equipment license.

Control 2.2.42 Ability to recognize system parameters that are entry- 4.6 94 level conditions for Technical Specifications.

Subtotal 2

3. Radiation 2.3.11 Ability to control radiation releases. 4.3 97 Control Subtotal 1 2.4.8 Knowledge of how abnormal operating procedures are 4.5 96 used in conjunction with EOPs.

4. Emergency 2.4.19 Knowledge of EOP layout, symbols, and icons. 4.5 99 Procedures/Plan 2.4.40 2.4.40 Knowledge of SRO responsibilities in emergency plan implementation.

Subtotal 2 Tier 3 Point Total 10 7 Rev 2

ES-401 Record of Rejected K/As Form ES-401-4 Tier / Group Randomly Reason for Rejection Selected K/A RO T1/G1 295037 295037 EA1.07 (Q#36) CNS does not have RC&IS. Replaced 295037 EA1.08 EA1.08 (RC&IS) with 295037 EA1.07 (RMCS). Page 1 point totals not affected by this change. (Rev 1)

RO T2/G1 212000 212000 A2.11 (Q#20) RPS does not receive input from MSL high radiation at CNS. Replaced 212000 A2.17 (MSL high radiation) with A2.17 212000 A2.11 (MSIV closure). Page 1 point totals not affected by this change. (Rev 1)

RO T2/G1 239002 A1.06 (Q#4) CNS does not have acoustical monitors. Replaced 239002 239002 A1.02 (acoustical monitor noise) with 239002 A1.06 A1.02 (reactor power). Page 1 point totals not affected by this change. (Rev 1) 206000 2.4.31 RO T2/G1 (Q#10) There are no local auxiliary operator tasks associated 206000 HPCI during emergencies except for reading REC flow locally, which is not operationally significant.

G2.4.35

[NRC] Replaced with 2.4.31 Knowledge of annunciator alarms, indications, or response procedures.

295017 AK2.10 RO T1/G2 (Q#35) Could not develop discriminatory question at the RO 295017 level regarding meteorological effects on off-site release.

ROs are not involved in dose assessment or PARs at CNS.

AK1.03

[NRC] Replaced with K/A AK2.10: Knowledge of the interrelations between HIGH OFF-SITE RELEASE RATE and the following: AK2.10 Process radiation monitoring system.

NOTE: No other AK1 K/As were acceptable for RO testing for this E/APE, so AK2 K/A was chosen. Because this dropped the Tier 1 K1 K/As to less than 2, another T1/G2 System with a K2 K/A was changed to K1 for Q#62.

RO T1/G2 295022 AK1.02 (Q#62) [NRC] System 295022 AK2.03 Loss of Control Rod Drive Pumps:

295022 Replaced with K1 K/A AK1.02, Knowledge of the operational AK2.03 implications of the following concepts as they apply to LOSS OF CRD PUMPS: Reactivity control This is to make up for loss of 1 K1 K/A per Q#35 question change, above.

Rev 2

ES-401 Record of Rejected K/As Form ES-401-4 RO T2/G2 216000 A2.09 (Q#61) TIP question overlaps SRO Q#80 and has low IR 295017 (2.7). Recommend replacing with A2 from K/A 216000 NBI (same safety function)

AK1.03

[NRC] Agree, replaced system with 216000 A2.09, Ability to 215001 (a) predict the impacts of the following on the NUCLEAR A2.01 BOILER INSTRUMENTATION ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Jet pump flow: Design-Specific RO T3 G2.2.14 (Q#72) Could not develop discriminatory question at the RO level regarding Infrequently Performed Tests and Evolutions.

ROs have no detailed responsibilities regarding IPTEs at G2.2.7 CNS.

[NRC] Replaced with 2.2.14 Knowledge of the process for controlling equipment configuration or status.

SRO T2/G1 259002 (Q#76) Could not develop question at SRO level for K/A 2.4.9 259002 G2.4.30 related to RVLCS due to K/A is for low power/shutdown accident mitigation strategies.

G2.4.9

[NRC] Replaced with G2.4.30 Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.

SRO T2/G1 N/A (Q#78) There are no License Conditions [10 CFR 55.43(b)(1) 215005 topics] related to APRMs.

G2.2.38 [NRC] Technical Specifications are part of the facility license, test on Tech Spec aspects of APRMs to satisfy conditions and limitations in the facility license to satisfy the original K/A.

SRO T3 G2.4.40 (Q#99) Could not develop question at SRO level for knowledge of EOP layout, symbols, and icons.

G2.4.19 [NRC] Replace with 2.4.40 Knowledge of SRO responsibilities in emergency plan implementation.

Rev 2

ES-401 Record of Rejected K/As Form ES-401-4 Revision statement:

Rev 1

• For Q#36, replaced original K/A 295037 EA1.08 related to RC&IS with 295037 EA1.07 related to RMCS, since CNS does not have RC&IS, and documented on ES-401-4. Tier totals were not affected by this change.

• For Q#3, corrected K/A 211000 from A3.03 to A2.03, since A2 category was indicated by the first X in A2 column. Tier totals were not affected by this change.

• For Q#20, replaced original K/A 212000 A2.17 related to MSL high radiation with 212000 A2.11 related to MSIV closure, since MSL radiation is not an input to RPS at CNS, and documented on ES-401-4. Tier totals were not affected by this change.

• For Q#4, replaced original K/A 239002 A1.02 related to acoustical monitoring noise with A1.06 related to reactor power, since CNS does not have acoustical monitors, and documented on ES-401-4. Tier totals were not affected by this change.

• For Q#23, corrected K/A 263000 from A3.01 to A4.01, since A4 category was indicated by X in A4 column. Tier totals were not affected by this change.

• For Q#63, corrected K/A 271000 from A3.02 to A4.02, since A4 category was indicated by X in A4 column. Tier totals were not affected by this change.

• For question 83, corrected description and IR of listed K/A 2.2.12.

Rev 2 The following K/As were requested to be replaced by CNS, and NRC selected the replacement K/As (This is the status of the written exam outlines at the time of draft operating test submittal):

• For Q#10, replaced 206000 G2.4.35 with 206000 G2.4.31.

• For Q#35, replaced 295017 AK1.03 with 295017 AK2.10. No other AK1 K/As were acceptable for RO testing for this E/APE, so AK2 K/A was chosen. Because this dropped the Tier 1 K1 K/As to less than 2, another T1/G2 System with a K2 K/A was changed to K1 (ref. Q#62).

• For Q#62, replaced 295022 AK2.03 with 295022 AK1.02 to make up for loss of one K1 K/A per Q#35 question change, above.

• For Q#61, replaced 215001 (TIPS) A2.01 with 216000 (NBI) A2.09.

• For Q#72, replaced G2.2.7 with G2.2.14.

• For Q#76, replaced 259002 G2.4.9 with 259002 G2.4.30.

• For Q#99, replaced G2.4.19 with G2.4.40.

Rev 2

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Cooper Nuclear Station Date of Examination: 4/27/2020 Examination Level: RO SRO Operating Test Number: CN-2020-4 Administrative Topic (see Note) Type Describe activity to be performed Code*

A1, modified SKL034-21 Drywell Average Temperature Calculation Conduct of Operations R, N (Normal Method) IAW 5.8.10 K/A G2.1.23 (4.3)

A2, new - Determine mechanical and electrical isolation boundaries for seal Equipment Control R, N replacement on REC Pump C K/A G2.2.13 (4.1)

From Previous 2 exams randomly selected 3/2017 ILT A4 Radiation Control R, P A3. Determine Workers Projected Total Dose (using survey map) 2.3.14 (3.4)

A4, new- calculate DW release rate IAW Commented [WG1]: Is this really an RO Emergency Plan Implementing function?

Emergency Plan R, D Procedure 5.7.16 K/A G2.4.39 (3.9)

NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).

• Type Codes and Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank ( 3 for ROs; 4 for SROs and RO retakes)

(N)ew or (M)odified from bank ( 1)

(P)revious 2 exams ( 1, randomly selected)

Rev 0

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Cooper Nuclear Station Date of Examination: 4/27/2020 Examination Level: RO SRO Operating Test Number: CN-2020-4 Administrative Topic (see Note) Type Describe activity to be performed Code*

SA5, From Previous 2 exams randomly selected 9/2018 NRC SA5 Conduct of Operations R, P Review IST stroke time surveillance for RRMG ventilation dampers (plant OE)

K/A G2.1.20 (4.6)

SA6, new - Determine reportability for Control Room Emergency Filtration System BF-C-1A Emergency Booster Fan non-functional Conduct of Operations in Mode 1. (Group 6 occurs due to R, N inadvertent short (invalid ESF actuation) while I&C working in panel, CREF fan motor fails) (Procedure 2.0.5)

K/A 2.1.18 (3.8), 2.1.20 (4.6)

SA7, SKL0345061 (modified to rod 10-11) -

Determine TS Actions for Removal of a Equipment Control R, M Single CRD during Refueling K/A 2.2.40 (4.7)

SA8, new - Determine emergency dose limit, estimated exposure, and whether KI Radiation Control authorization is required for operator to R, N manually close RCIC-MO-15 Commented [WG1]: Very similar to combination of JPMs from 2018 exam (JPM SA8 - Determine K/A G2.3.4 (3.7) Emergency Exposure Requirements - RCIC room) and 2017 exam (JPM A8 - Authorize Stable Iodine Thyroid Blocking). Not sure this is SA9, SKL034-30-64 (Emergency Classification) sufficiently different, but a combination from the previous 2 exams. Thoughts?

Emergency Plan R, D K/A G2.4.41 (4.6)

NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).

• Type Codes and Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank ( 3 for ROs; 4 for SROs and RO retakes)

(N)ew or (M)odified from bank ( 1)

(P)revious 2 exams ( 1, randomly selected)

Rev 0

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Cooper Nuclear Station Date of Examination: 4/27/2020 Exam Level: RO SRO-I SRO-U Operating Test Number: CN-2020-4 Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Code* Safety Function S1. Respond to two rods drifting during 6.CRD.301 (Alt Path) N, A 1 K/A 201003 A2.03 (3.4/3.7)

S2. Restart RWCU following Group 3 isolation IAW SOP 2.2.66 sect P 2 Commented [WG1]: Shouldnt this also be coded 6 (9/2018 ILT NRC JPM S2 randomly selected) as D?

9/2018 NRC K/A 204000 A4.01 (3.1/3.0) ILT S2 S3. SKL034-20-29R0, Open MSIVs with RPV pressurized (Alt D, A, EN, L 3 Path)

K/A 239001 A4.01 (4.2/4.0)

S4. SKL034-20-137, Operate RCIC in pressure control (Alt Path) D, A, L 4 K/A 217000 A4.01 (3.7/3.7)

S5. Perform 5.2FUEL Att. 2 (Actions to minimize rad release) N, L 5 KA 295036 EK3.04 (3.1/3.4)

S6. Energize 4160V critical Bus 1G from Emergency Station Service D, EN 6 XFMR (ESST) (SKL034-20-43 Rev 0)

K/A 262001 A4.01 (3.4/3.7)

S7. Restore Level Transmitter NBI-LT-59D input to RVLC to service N 7 IAW IOP 4.4.1 section 9 K/A 216000 K1.12 (3.6/3.7)

S8. Start SGT B for HPCI operation for level control post-scram. N, A, EN, L 8 2.2.73 Section 8. (Alt Path)

KA 261000 A4.07 (3.1/3.2) page 1 of 2 Rev 0

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U P1. new - Alt Path - Locally secure unloaded DG N, A, E 6 KA 264000 K4.07 (3.3/3.4)

P2. SKL034-10-65R9, Conduct alternate rod insertion by D, R, L, E 1 manually venting scram air header K/A 295037 EA1.05 (3.9/4.0) EK2.05 (4.0/4.1)

P3. SKL-L-0341019- Place the A Air Dryer in Service D 8 K/A 295019 AA1.02 (3.3/3.1)

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

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

(C)ontrol room -----

(D)irect from bank 9/ 8/ 4 (6/6/3)

(E)mergency or abnormal in-plant 1/ 1/ 1 (2/2/2)

(EN)gineered safety feature 1/ 1/ 1 (control room system) (3/3/1)

(L)ow-Power/Shutdown 1/ 1/ 1 (5/5/4)

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

(P)revious 2 exams 3/ 3/ 2 (randomly selected) (1/0/0)

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

(S)imulator ------

1.9.1 Plant System Organization by Safety Function

1. Reactivity Control

2. Reactor Water Inventory Control

3. Reactor Pressure Control

4. Heat Removal From the Core

5. Containment Integrity

6. Electrical

7. Instrumentation

8. Plant Service Systems

9. Radioactivity Release.

page 2 of 2 Rev 0

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Cooper Nuclear Station Date of Exam: 4/27/2020 Operating Test No.: 2020-4 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 RX 0 1 1 1 1 0 NOR 1 0 1 1 1 1 U1 I/C 7 6 13 4 4 2 MAJ 2 1 3 2 2 1 TS 2 2 4 0 2 2 RX 0 1 0 1 1 1 0 NOR 1 1 0 2 1 1 1 I1 I/C 3 7 3 13 4 4 2 MAJ 2 1 1 4 2 2 1 TS N/A 2 N/A 2 0 2 2 RX 0 1 1 1 1 0 NOR 0 1 1 1 1 1 R1 I/C 6 5 11 4 4 2 MAJ 2 1 3 2 2 1 TS N/A N/A 0 0 2 2 RX 0 1 1 1 1 0 NOR 0 0 0 1 1 1 R2 I/C 4 4 8 4 4 2 MAJ 1 1 2 2 2 1 TS N/A N/A 0 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 I/C 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 new reactor facility 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.

Rev 0

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Cooper Nuclear Station Date of Exam: 4/27/2020 Operating Test No.: 2020-4 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 RX 0 0 1 1 0 NOR 1 1 1 1 1 U2 I/C 7 7 4 4 2 MAJ 2 2 2 2 1 TS 2 2 0 2 2 RX 1 1 1 1 0 NOR 1 1 1 1 1 U3 I/C 7 7 4 4 2 MAJ 1 1 2 2 1 TS 2 2 0 2 2 RX 0 0 0 1 1 0 NOR 1 0 1 1 1 1 R3 I/C 3 4 7 4 4 2 MAJ 2 1 3 2 2 1 TS N/A N/A 0 0 2 2 RX 0 1 1 1 1 0 NOR 0 1 1 1 1 1 R4 I/C 6 5 11 4 4 2 MAJ 2 1 3 2 2 1 TS N/A N/A 0 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 I/C 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 new reactor facility 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.

Rev 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 4/2020 Scenario 1 Page 1 of 12 Facility: Cooper Nuclear Station Scenario No.: 1 Op-Test No.: CN-2020-04 Examiners: ____________________________ Operators: _____________________________

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

1. Shift CRD Pump B to CRD Pump A in operation

2. Respond to narrow range level transmitter NBI-LT-52C failure

3. Respond to Critical power panel CPP-1 loss of power

4. Respond to Condensate Pump A trip, Reactor Recirc Pump A runs back too far, trips

5. Respond to Reactor Recirc Pump B trip with Thermal Hydraulic Instability

6. Respond to a Main Steam Line D leak in the drywell

7. Respond to DW Spray A permissive switch failure

8. Respond to Torus to Drywell Vacuum Breaker failure with failure of RHR Loop B containment sprays Initial Conditions: Plant operating at 100% power near end of cycle.

Inoperable Equipment: Core Spray Pump A is tagged out of service for motor PMs.

Turnover:

Plant operating at operating at 100% power near end of cycle.

Planned activities for this shift are:

• Continue steady state power operation.

• Continue maintenance on CS Pump A.

Scenario Notes:

This is a new scenario.

Validation Time: XX minutes Page 1 of 12 Rev 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 4/2020 Scenario 1 Page 2 of 12 Event Malf. No. Event Type Event No. Description 1 N/A N (ATC,CRS) Shift CRD Pumps (B to A)

I (ATC,CRS) 2 rr27c Narrow range level transmitter NBI-LT-52C failure TS (CRS)

C (BOP,CRS) 3 ed11d Critical power panel CPP-1 loss of power A (CREW)

C (ATC,BOP,CRS) fw14b Condensate Pump A trip, RR Pump A runs back 4 A (CREW) rr17a too far, Reactor Recirc Pump A trip TS (CRS)

Reactor Recirc Pump B shaft binding with Thermal Hydraulic Instability CT#1 Manually scram the reactor when both rr03b 5 M (CREW) recirculation pumps trip, prior to exceeding cr04b 25% peak to peak neutron flux oscillations or APRM auto scram setpoint due to neutron flux oscillations.

6 ms01d M (CREW) Main Steam Line D leak in the drywell (Override) 7 C (BOP,CRS) DW Spray A permissive switch failure zdirhrsws17a 8 pco2a C (BOP,CRS) Torus to Drywell Vacuum Breaker failure 9 rh01b C (BOP,CRS) RHR Pump B trip RHR Pump D trip CT#2 C When torus pressure cannot be maintained 10 rh01d (ATC,BOP,CRS) below the Pressure Suppression Pressure (PSP), the crew completes Emergency Depressurization prior to exceeding PCPL-A.

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

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

Page 2 of 12 Rev 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 4/2020 Scenario 1 Page 3 of 12 Quantitative Attributes Table ES-301-4 Actual Attribute Description Target

1. DW Spray A permissive switch failure

2. Torus to Drywell Vacuum Breaker failure Malfunctions after 1-2 4 3. RHR Pump B trip EOP entry

4. RHR Pump D trip

1. Critical power panel CPP-1 loss of power Abnormal Events 2-4 2 2. Condensate Pump A trip, RR Pump A runs back too far, Reactor Recirc Pump A trip

1. Reactor Recirc Pump B shaft binding with Thermal Major Transients 1-2 2 Hydraulic Instability

2. Main Steam Line D leak in the drywell EOP entries 1. EOP-1A requiring 1-2 2 2. EOP-3A substantive action EOP contingencies requiring 1 per set 1 1. EOP-2A Contingency #2 - Emergency Depressurization substantive action

1. (CT#1) Manually scram the reactor when both recirculation pumps trip, prior to exceeding 25% peak to peak neutron flux oscillations or APRM auto scram setpoint due to neutron flux oscillations.

Pre-identified 2 2 Critical Tasks 2. (CT#2) When torus pressure cannot be maintained below the Pressure Suppression Pressure (PSP), the crew completes Emergency Depressurization prior to exceeding PCPL-A.

Normal Events N/A 1 1. Shift CRD pumps Reactivity 1. none N/A 0 Manipulations

1. Narrow range level transmitter NBI-LT-52C failure

2. Critical power panel CPP-1 loss of power

3. Condensate Pump A trip, RR Pump A runs back too far, Reactor Recirc Pump A trip Instrument/

Component N/A 7 4. DW Spray A permissive switch failure Failures 5. Torus to Drywell Vacuum Breaker failure

6. RHR Pump B trip

7. RHR Pump D trip Page 3 of 12 Rev 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 4/2020 Scenario 1 Page 4 of 12

1. Narrow range level transmitter NBI-LT-52C failure

2. Critical power panel CPP-1 loss of power

3. Condensate Pump A trip, RR Pump A runs back too far, Reactor Recirc Pump A trip Total Malfunctions N/A 7 4. DW Spray A permissive switch failure

5. Torus to Drywell Vacuum Breaker failure

6. RHR Pump B trip

7. RHR Pump D trip Top 10 systems and operator actions important to risk that are tested:

• Primary Containment, RHR, RHRSW Page 4 of 12 Rev 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 4/2020 Scenario 1 Page 5 of 12 SCENARIO

SUMMARY

The plant is operating at 100% power near the end of the operating cycle.

After the crew takes the watch, the ATC will shift CRD pumps from CRD Pump A in operation to CRD Pump B in operation IAW Procedure 2.2.8.

After CRD pumps have been shifted, narrow range level transmitter NBI-LT-52C fails high, resulting in a half Main Turbine/RFPT trip. The crew will respond IAW the alarm card and Procedure 4.4.1 to diagnose the failure using the RVLCS HMI. The CRS will enter TS 3.3.2.2 Condition A.

After TS for the failed level transmitter have been addressed, 120V AC power panel CPP-1 will lose power. The crew will respond IAW Procedure 5.3AC120.

The BOP will start a turbine building exhaust fan to restore TB HVAC.

After response to loss of CPP-1 is complete, Main Condensate Pump 1B will trip. The crew will respond to CP B trip and enter 2.4MC-RF to address the tripped MC Pump B. At this power level the Reactor Recirc pumps should run back towards 45% speed and stop running back once Condensate pump discharge header pressure restores, which should occur approximately 50%

Reactor Recirc pump speed. However, Reactor Recirc Pump A will continue to runback for several seconds. If the crew identifies RR Pump A continues to run back, they will lockout the RR A scoop tube. This will require entry into 2.4RR.

After a delay RR Pump A will trip. The CRS will enter TS 3.4.1 Condition B due to single loop operation.

After the TS for RR Pump A failure have been addressed, RR Pump B will trip.

The ATC will respond by inserting a manual scram IAW procedure 2.4RR due to no RR Pumps operating and power above 1%. If a manual scram is delayed, thermal hydraulic instability will occur, also requiring a manual scram per 2.4RR (CT#1).

When the reactor is scrammed, a leak develops on MSL D in the drywell.

Drywell pressure and temperature rise. The crew will enter EOP-1A and EOP-3A due to high drywell pressure. Torus Spray will be required. Only RHR Loop B is available for containment spray due to failure of the spray valve control permissive for RHR Loop A. A Torus to Drywell Vacuum Breaker will fail open, requiring use of Drywell Spray.

Drywell Spray will fail a short time after it has been started due to trip of RHR Pump B. If RHR Pump D is started for Drywell Spray, it will also trip. Drywell pressure will rise above the Pressure Suppression Pressure (PSP) limit.

Emergency Depressurization will be required (CT#2).

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Appendix D Scenario Outline Form ES-D-1 NRC CNS 4/2020 Scenario 1 Page 6 of 12 The exercise ends when the reactor has been depressurized, RPV water level is being controlled +3 to +54 inches.

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Appendix D Scenario Outline Form ES-D-1 NRC CNS 4/2020 Scenario 1 Page 7 of 12 CRITICAL TASK BASIS Manually scram the reactor when both recirculation pumps trip, prior to exceeding 25% peak to peak neutron flux oscillations Critical Task #1 or APRM auto scram setpoint due to neutron flux oscillations.

Analyses of neutronic/thermal-hydraulic instabilities during failure-to-scram conditions have been performed. Instabilities are manifested by oscillations in reactor power, which, if the reactor cannot be shut down, may increase in magnitude. If the oscillations remain small or moderately sized, they tend to repeat on approximately a two second period. Under certain circumstances, however, the oscillations may continue to grow and become sufficiently large and Safety Significance irregular to cause localized fuel damage. Analytical results indicate that the fuel clad may experience boiling transition during THI but that it subsequently rewets and is adequately cooled even for oscillations that resemble reactivity excursion events. For an occasional large pulse, however, rewetting of some of the highest-powered locations within the highest-powered fuel bundles may not occur; the clad could then continue to heat up over several oscillation cycles.

Cues SRM period alarms.

Oscillating power indications on neutron monitoring instrument.

Operator depresses both manual scram pushbuttons, or places the Measurable Performance Reactor Mode Switch to SHUTDOWN on panel 9-5.

Indicators On panel 9-5:

RPS status lights de-energized.

Performance Feedback Reactor power level trend.

Control rods inserted.

Applicability Operation in the Stability Exclusion Region with no Recirc pump in operation.

The threshold of 25% peak-to-peak neutron flux oscillations has been chosen to ensure an attempt to shut down the reactor using RPS for the same conditions that require boron injection during failure to scram events. Analysis has shown that APRM peak-to-peak amplitudes reach 25% of rated thermal power well before any Justification for the individual LPRM signal reaches an amplitude for which fuel damage chosen performance might be possible. Since analysis has shown THI to exhibit a limit negative decay ratio, power level would be expected to rise to the automatic high flux scram setpoint. Requiring manual reactor shutdown before an automatic scram on high neutron flux is reasonable to allow a short time to recognize and diagnose THI, communicate plant conditions, and effect the manual scram.

BWR Owners Group App. B, step RC/Q-6 Appendix Page 7 of 12 Rev 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 4/2020 Scenario 1 Page 8 of 12 Initial power level following the recirc pump trips, with RPV level in the normal band, should be in the Stability Exclusion Region.

Thermal Hydraulic Instability malfunction CR04A must be inserted at Scenario Guide a ramped severity that will eventually cause an automatic high flux Requirements scram, but at slow enough ramp rate to allow the crew to recognize and diagnose THI and 25% oscillations, then communicate and insert a manual scram before the high flux trip setpoint is reached.

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Appendix D Scenario Outline Form ES-D-1 NRC CNS 4/2020 Scenario 1 Page 9 of 12 When torus pressure cannot be maintained below the Pressure Suppression Pressure (PSP), the crew completes Emergency Critical Task #2 Depressurization prior to exceeding PCPL-A.

If suppression pool and/or drywell sprays cannot be initiated or are ineffective in reversing the increasing trend of primary containment pressure, as evidenced by not being able to maintain torus pressure below the Pressure Suppression Pressure, the RPV is depressurized to minimize further release of energy from the RPV to the primary containment. This action serves to terminate or reduce as much as possible any continued primary containment pressure increase.

The Pressure Suppression Pressure (PSP) is the lesser of:

• The highest torus pressure which can occur without steam in the torus air space.

• The highest torus pressure at which initiation of RPV depressurization will not result in exceeding Primary Containment Pressure Limit before RPV pressure drops to the Minimum RPV Flooding Pressure.

Safety Significance

• The highest torus pressure which can be maintained without exceeding the suppression pool boundary design load if SRVs are opened.

The Primary Containment Pressure Limit (PCPL) is the lesser of:

• The pressure capability of the primary containment.

• The maximum primary containment pressure at which vent valves sized to reject all decay heat from the containment can be opened and closed.

• The maximum primary containment pressure at which SRVs can be opened and will remain open.

The PCPL is a function of primary containment water level.

Exceeding the limit may challenge primary containment vent valve operability, SRV operability, or the structural integrity of the primary containment.

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

Cursor approaching unsafe boundary on PSP graph display on SPDS.

Manipulation of any six SRV controls on panel 9-3:

SRV-71A Measurable Performance SRV-71B Indicators SRV-71E SRV-71G SRV-71H Page 9 of 12 Rev 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 4/2020 Scenario 1 Page 10 of 12 SRV-71C SRV-71D SRV-71F Crew will observe SRV light indication go from green to red, amber pressure switch lights illuminate, reactor pressure lowering on Performance Feedback SPDS and panel 9-3 and 9-5 meters and recorders, and SRV tailpipe temperatures rise on recorder MS-TR-166.

Applicability EOP-3A conditions with energy being added to primary containment (LOCA) and torus/drywell pressure rising.

When torus pressure cannot be maintained below PSP is the EOP-3A, step PC/P-4 criteria requiring transition to emergency depressurization. Failure to ED early enough to avert exceeding Justification for the PCPL-A could result in failure of the containment barrier. .

chosen performance limit Depressuring to the Decay Heat Removal Pressure (50 psig) before PCPL-A is exceeded ensures energy discharged to the primary containment will be within the capacity of the containment vent.

BWR Owners Group App. B, step PC/P-2.

Appendix A leak with failure of torus/drywell sprays and failure of a Torus to Drywell vacuum Breaker is normally required to effect exceeding Scenario Guide PSP. It is important to design the scenario such that the crew has Requirements information early during the LOCA event to determine the rising DW pressure cannot be mitigated and that PSP will be exceeded.

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Appendix D Scenario Outline Form ES-D-1 NRC CNS 4/2020 Scenario 1 Page 11 of 12 Related Operating Experience SEN 57 Malfunction of Rosemount pressure and differential pressure transmitters Various failures involving loss of sensing fluid or electronics failure INPO ICES 307495 Vital 120V Instrument AC Uninterruptable Power Supply Inverter Output Breaker Trip On 2/23/13 Class 1E 120V Instrument (Inst) AC uninterruptable power supply (UPS) inverter IY23 output breaker 52-23B2 tripped at approximately 04:27 and de-energized all of the Class 1E 120V Inst AC PY23 panel loads. The IY23 output breaker tripping rendered the inverter and its respective instrument AC distribution panel PY23 inoperable. Visual inspection, analysis of plant process computer data did not find evidence of a fault. The output breaker was replaced and the PY23 panel was re-energized. The Apparent Cause Analysis did not find a cause for spurious breaker trip and the presumptive cause was due to aging of the electronic trip unit.

(CNS) CR 2006-09515 ANN -ANN -(C-4/F-6), CRIT INST & CONT PNL CPP LOSS OF VOLT Received Received Annunciator ANN -ANN -(C-4/F-6), CRIT INST & CONT PNL CPP LOSS OF VOLT, in and reset. Checked MCC-TX feeder breaker. All conditions normal. No other indications of loss of panel CPP.

INPO ICES 407759 Condensate Pump Breaker Lockout and Pump Trip On 03/05/2017 steam generator water level deviation annunciators alarmed and the main feedwater reg valves were noted to be opening. Subsequently the condensate pump was discovered to be tripped. This resulted in the unit down powering to 90 percent until repairs to the condensate motor could be made. The condensate motor failed due to an electrical short caused by loose stator bolts in the motor. This event is Consequential because of the unplanned downpower.

SER 23-93 Delayed Manual Scram Following a Core Flow Reduction Event And Entry into a Region of Core Instability (Perry) On July 9, 1993, the reactor was operating at 100 percent power when both reactor recirculation pumps unexpectedly downshifted to low speed, placing the reactor in a region of the operating domain susceptible to core thermal-hydraulic instability. The Page 11 of 12 Rev 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 4/2020 Scenario 1 Page 12 of 12 operators manually scrammed the reactor several minutes after entering this region of the power-to- flow map where an immediate manual scram is procedurally required.

(CNS) CR 2007-03555 Stability Exclusion Region of Power-to-Flow Map entered At 0030 5-19-2007 Abnormal Procedure 2.4RR was entered was entered due to entry into the stability exclusion region. The stability exclusion region was entered when the "A" RRMG was secured for planned maintenance. LCO 3.4.1 Condition A was entered and recirc flow was raised to exit the region as required by 2.4RR and LCO 3.4.1.

SEN 136 Mispositioned control rod recovery without management approval.

(CNS) On January 7, 1996, during a load line reduction following a reactor recirculation pump trip, the wrong control rod group was inserted when both the reactor operator and second verifier selected the wrong page of the rod movement sheet. After insertion of the first group of rods, and noting the error, the operator and verifier decided to begin recovery from the error without notifying management of the mispositioning or the plan to recover.

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Appendix D Scenario Outline Form ES-D-1 NRC CNS 4/2020 Scenario 2 Page 1 of 11 Facility: Cooper Nuclear Station Scenario No.: 2 Op-Test No.: CN-2020-04 Examiners: ____________________________ Operators: _____________________________

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

1. Transfer Reactor Mode Switch to RUN

2. Withdraw control rods IAW the rod sequence to establish 20-25% bypass valve position

3. Respond to RMCS withdraw timer malfunction

4. Respond to control rod double notch during withdrawal

5. Respond to spurious HPCI initiation

6. Respond to HPCI steam line break in the Reactor Building requiring scram

7. Respond to failure of Scram Discharge Volume vent and drain valves to automatically close

8. Respond to Reactor Mode Switch failing in RUN position

9. Respond to failure of HPCI to automatically isolate on high area temperature

10. Respond to failure HPCI isolation MOVs Initial Conditions: Plant operating at 5% power during startup Inoperable Equipment: none Turnover:

The plant is at 5% power at beginning of cycle.

Planned activities for this shift are:

• Transfer to RUN Mode per Procedure 2.1.1.

• Withdraw control rods IAW the rod sequence to establish 20-25% bypass valve position.

• Continue startup IAW Procedure 2.1.1.

Scenario Notes:

This is a new scenario.

Validation Time: XX minutes Page 1 of 11 Rev 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 4/2020 Scenario 2 Page 2 of 11 Event Malf. No. Event Type Event No. Description 1 N/A N (ATC,CRS) Transfer Reactor Mode Switch to RUN 2 N/A R (ATC,CRS) Raise reactor power by withdrawing control rods I (ATC,CRS) 3 rd19 RMCS withdraw timer malfunction A (CREW)

C (ATC,CRS) 4 rd235027 A (CREW) Control rod double notches on withdraw TS (CRS)

Spurious HPCI initiation with failure of HPCI trip on C (BOP,CRS) high water level hp05 5 A (CREW) CT#1 rr28b@58 When HPCI spuriously initiates at low reactor TS (CRS) power, stop and prevent HPCI injection before RFPT high water level trip.

HPCI steam line break in the Reactor Building 6 hp06 M (CREW) requiring scram rd01a rd01b Failure of Scram Discharge Volume vent and drain 7 C (ATC,CRS) rd01c valves to automatically close rd01d (override) C 8 Reactor Mode Switch fails in RUN position zdirpssws1 (ATC,BOP,CRS)

Failure of HPCI to automatically isolate on high 9 hp09 C (BOP,CRS) area temperature Failure of HPCI steam supply isolation valves to fully close CT#2 When a primary system is discharging into the secondary containment through an unisolable (Overrides) break, the crew Emergency Depressurizes by zdihpcisws1 opening 6 SRVs when maximum safe operating zlohpcisws1(2) 10 C values are exceeded in two areas for the same (ATC,BOP,CRS) parameter and prior to exceeding maximum (Remote) safe operating value in a third area for that hp04a parameter. (For this scenario, RHR-TE-99G Torus 900 SW and RHR-TE-99C Torus 885 NNW are the two areas that will exceed MSO first. RWCU-TE-117F Torus 896 SE would be the third area to exceed MSO.)

(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 4/2020 Scenario 2 Page 3 of 11 Quantitative Attributes Table ES-301-4 Actual Attribute Description Target

1. Reactor Mode Switch fails in RUN position

2. Failure of Scram Discharge Volume vent and drain valves to automatically close Malfunctions after 3. Failure of HPCI to automatically isolate on high area 1-2 4 EOP entry temperature

4. Failure of HPCI steam supply isolation valves to fully close

1. RMCS withdraw timer malfunction Abnormal Events 2-4 2 2. Spurious HPCI initiation with failure of HPCI trip on high water level

1. HPCI steam line break in the Reactor Building requiring Major Transients 1-2 1 scram EOP entries 1. EOP-5A requiring 1-2 2 2. EOP-1A substantive action EOP contingencies requiring 1 per set 1 1. EOP-2A Contingency #2 - Emergency Depressurization substantive action

1. (CT#1) When HPCI spuriously initiates, stop and prevent HPCI injection before RFPT high water level trip.

2. (CT#2) When a primary system is discharging into the secondary containment through an unisolable break, the crew Emergency Depressurizes by opening 6 SRVs when Pre-identified maximum safe operating values are exceeded in two 2 2 Critical Tasks areas for the same parameter and prior to exceeding maximum safe operating value in a third area for that parameter. (For this scenario, RHR-TE-99G Torus 900 SW and RHR-TE-99C Torus 885 NNW are the two areas that will exceed MSO first. RWCU-TE-117F Torus 896 SE would be the third area to exceed MSO.)

Normal Events N/A 1 1. Transfer Reactor Mode Switch to RUN Reactivity 1. Raise reactor power by withdrawing control rods N/A 1 Manipulations

1. RMCS withdraw timer malfunction

2. Spurious HPCI initiation with failure of HPCI trip on high water level

3. Reactor Mode Switch fails in RUN position Instrument/ 4. Failure of Scram Discharge Volume vent and drain Component N/A 6 valves to automatically close Failures

5. Failure of HPCI to automatically isolate on high area temperature

6. Failure of HPCI steam supply isolation valves to fully close Page 3 of 11 Rev 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 4/2020 Scenario 2 Page 4 of 11

1. RMCS withdraw timer malfunction

2. Spurious HPCI initiation with failure of HPCI trip on high water level

3. Reactor Mode Switch fails in RUN position

4. Failure of Scram Discharge Volume vent and drain Total Malfunctions N/A 6 valves to automatically close

5. Failure of HPCI to automatically isolate on high area temperature

6. Failure of HPCI steam supply isolation valves to fully close Top 10 systems and operator actions important to risk that are tested:

• HPCI, Primary Containment Isolation System, RPS Page 4 of 11 Rev 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 4/2020 Scenario 2 Page 5 of 11 SCENARIO

SUMMARY

The plant is operating at 5% power during startup.

After the crew takes the watch, the crew transfers the Reactor Mode Switch to RUN per Procedure 2.1.1.

After the Reactor Mode Switch has been transferred to RUN, the ATC will withdraw control rods IAW the startup rod sequence to establish Bypass valves 20-25% open.

During the power ascension, a RMCS withdraw timer malfunction will occur, resulting in a control rod select block and requiring entry into 2.4CRD. The ATC will reset the timer malfunction IAW 2.4CRD Att. 8.

After the RMCS withdraw timer malfunction has been cleared, the crew will resume control rod withdrawal. Control rod 50-27 (4th rod withdrawn) will double notch past the required banked position, resulting in a control rod withdrawal block. The crew will respond IAW the alarm card. The CRS will determine a LCO 3.1.6 Condition A entry is required due to rod position not in compliance with BPWS.

When response to the control rod double notch is complete, HPCI will spuriously initiate. The crew will respond IAW procedure 2.4CSCS and stop and prevent HPCI to avert a RFPT trip due to reactor water level reaching the RFPT high water level trip setpoint (CT#1). The CRS will enter TS 3.5.1 Condition C for HPCI inoperable due to it being prevented from injecting After actions for HPCI spurious initiation are complete, a small HPCI steam line break will occur in the Reactor Building. Secondary containment area temperatures and radiation levels will rise, requiring entry into EOP-5A. If HPCI area temperatures reach 195°F, MSO limit and isolation setpoint, before the crew attempts manual isolation of HPCI steam supply valves, the automatic isolation will fail to occur. When the crew attempts manual isolation of the HPCI steam supply valves, both valves will fail to fully close due to mechanical binding in their gear boxes. EOP-1A entry and reactor shutdown will be required.

When the reactor is manually scrammed, Scram Discharge Volume vent and drain valves will fail to automatically close. The ATC will isolate the valves using the panel 9-5 control switch. The Reactor Mode Switch will fail in RUN position. A Group 1 isolation will occur when reactor pressure lowers to 835 psig due to low decay heat and the HPCI steam leak, requiring transition to RCIC and CRD for level control.

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Appendix D Scenario Outline Form ES-D-1 NRC CNS 4/2020 Scenario 2 Page 6 of 11 The HPCI steam leak will get worse, causing area temperatures in more than one area of the Reactor Building to approach the MSO limit. When two areas have reached the MSO limit, emergency depressurization will be required IAW EOP-1A, and the crew will enter EOP-2A to perform emergency depressurization (CT#2).

The exercise ends when the reactor has been depressurized and RPV water level is being restored to between +3 and +54 inches.

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Appendix D Scenario Outline Form ES-D-1 NRC CNS 4/2020 Scenario 2 Page 7 of 11 CRITICAL TASK BASIS When HPCI spuriously initiates at low reactor power, stop and Critical Task #1 prevent HPCI injection before RFPT high water level trip.

An avoidable RFPT trip presents an unnecessary challenge to plant safety. Loss of FW transients are a significant contributor to core Safety Significance damage frequency (ref. CNS PRA Rev. 6).

Cues On panel 9-3:

Annunciator 9-3-2/A-1, HPCI Logic Actuated HPCI Stm to Turb Vlv MO-14 red light ON (opening)

HPCI Injection Valve MO-19 red light ON (opening)

Flow rising on Flow Controller HPCI-FC-108 Reactor water level rising on various control room indicators Manipulation of HPCI controls on panel 9-3:

HPCI Turbine Trip push button depressed Measurable Performance HPCI-MO-14 control switch to CLOSE Indicators HPCI Auxiliary Oil Pump control switch to PTL As HPCI is secured, on panel 9-3 crew will observe Annunciator HPCI Turbine Trip, HPCI-MO-14 Green light ON and Red light OFF, HPCI AOP Green light ON and Red light OFF, flow lowering on Performance Feedback HPCI-FC-108, HPCI Turbine Speed lowering on HPCI-SI-2792, and reactor water level stop rising on various control room indicators Applicability Low power conditions, where HPCI rated flow 4250 gpm is higher than reactor steam flow, resulting in reactor water level rising.

CNS Procedure 2.4CSCS immediate operator action requires securing HPCI upon a spurious initiation. Procedure 2.4RXLVL Justification for the immediate operator action requires a manual scram if reactor water chosen performance limit level cannot be maintained below +50 inches on narrow range instruments. RFPTs trip on high reactor water level, +53.5 inches.

BWR Owners Group N/A Appendix Power level must be low enough for HPCI at rated flow to cause reactor water level to rise above 53.5 inches with feedwater flow reduced to zero. To reach the RFPT trip setpoint, wide range level instrument NBI-LIS-101B should be set slightly below the HPCI high Scenario Guide water level trip setpoint using malfunction rr28b, to the defeat the Requirements HPCI trip, which occurs at a lower actual water level than the narrow range RFPT high water level trip. The scenario should be designed to provide at least ~1 minute before level reaches +53.5 inches.

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Appendix D Scenario Outline Form ES-D-1 NRC CNS 4/2020 Scenario 2 Page 8 of 11 When a primary system is discharging into the secondary containment through an unisolable break, the crew Emergency Depressurizes by opening 6 SRVs when maximum safe operating values are exceeded in two areas for the same parameter and prior to exceeding maximum safe operating Critical Task #2 value in a third area for that parameter. (For this scenario, RHR-TE-99G Torus 900 SW and RHR-TE-99C Torus 885 NNW are the two areas that will exceed MSO first. RWCU-TE-117F Torus 896 SE would be the third area to exceed MSO.)

Should secondary containment parameters exceed their maximum safe operating values in more than one area, the RPV must be depressurized to preclude further degradation. RPV depressurization places the primary system in its lowest possible energy state, rejects heat to the suppression pool in preference to outside the containment, and reduces the driving head and flow of primary systems that are unisolated and discharging into the Safety Significance secondary containment.

The criteria of "two or more areas" specified identifies the rise in secondary containment parameters as a wide-spread problem which may pose a direct and immediate threat to secondary containment integrity, equipment located in the secondary containment, and continued safe operation of the plant.

Cues SPDS indication for secondary containment parameters indicate area radiation, area temperature, or area water level has exceeded its maximum safe operating value in two areas.

Manipulation of SRV controls on panel 9-3:

SRV-71A SRV-71B SRV-71E Measurable Performance SRV-71G Indicators SRV-71H SRV-71C SRV-71D SRV-71F Crew will observe SRV light indication go from green to red, amber pressure switch lights illuminate, reactor pressure lowering on Performance Feedback SPDS and panel 9-3 and 9-5 meters and recorders, and SRV tailpipe temperatures rise on recorder MS-TR-166.

Applicability EOP-5A conditions, RCS leaks into secondary containment with the RPV pressurized.

Emergency Depressurization is required due to effects of a break spreading into and potentially affecting safety equipment and Justification for the operations in more than one area; however, emergency chosen performance limit depressurization is not allowed until the second area exceeds its Max Safe limit. Before the Max Safe limit is exceeded in a third area gives reasonable time for the crew to perform emergency Page 8 of 11 Rev 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 4/2020 Scenario 2 Page 9 of 11 depressurization before the leak hampers equipment or operations in an even more widespread area.

BWR Owners Group App. B, steps SC/T-4.2, SC/r-2.2, SC/L-2.2.

Appendix The scenario must be able to drive the selected parameter to its Max Safe value in three plant areas. If temperature is chosen, a failure to scram event, where RPV pressure is not allowed to be lowered, is well suited. Also, ensure the leak severity itself, or subsequent cold water injection, does not deplete RPV pressure Scenario Guide (driving head) so low that Max Safe in a third area cannot be Requirements reached. The crew should be driven to ED, versus just reducing pressure, to provide a consistent, measurable performance indicator. The CT listed in the scenario should list which instruments/areas will exceed their MSO limit first, second, and third.

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Appendix D Scenario Outline Form ES-D-1 NRC CNS 4/2020 Scenario 2 Page 10 of 11 Related Operating Experience (CNS) CR 2006-06868 RMCS timer failure Entered Abnormal Procedure 2.4 CRD due to control rod 18-43 will not notch out from position 46 to 48. When attempting to withdraw control rod by continuous or notch withdrawal the settle light illuminates immediately and no drive water flow is observed, the drive in and drive out lights do not illuminate. Performed Attachment 6 for stuck control rod, all attempts to move control rod are unsuccessful. Based on indications, fault appears to with the timer. Method of discovery: SP 6.CRD.301. Requirement not met: RMCS should operate HCUs to position control rods as required.

(CNS) CR 2011-01763 RMCS timer test failed during surveillance test During performance of the Withdraw Timer Malfunction Test at step 5.2.2 of 6.CRD.301 the Timer Test failed to function at all to block rod withdrawal. White light did not turn on

, Red light did not turn on and the selected Control Rod remained selected on the Full Core Display.

INPO ICES 306797 Control Rod double notching (CNS) On June 8, 2013, Control Rod 18-35 double notched during notch withdrawal from position 02 to 06. The control rod was being withdrawn from position 00 to 18 using notch withdrawal. All other notches were as expected. CRD Drive Water differential pressure was verified to be at 265 psig. The control rod withdrawal did not appear to be too fast. Control rod should not double notch during notch withdrawal.

After double notch occurrence, stop rod withdrawal. notified Reactivity Manager, Control Room Supervisor, and Shift Manager. Continued rod withdrawal after being evaluated by Reactor Engineering and Operation (CNS) CR 2014-01430 Control rod 18-35 has exhibited multiple double notches SEN 136 Mispositioned control rod recovery without management approval (CNS) On January 7, 1996, during a load line reduction following a reactor recirculation pump trip, the wrong control rod group was inserted when both the reactor operator and second verifier selected the wrong page of the rod movement sheet. After insertion of the first group of rods, and noting the error, the operator and verifier decided to begin recovery from the error without notifying management of the mispositioning or the plan to recover.

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Appendix D Scenario Outline Form ES-D-1 NRC CNS 4/2020 Scenario 2 Page 11 of 11 OE26672 Mispositioned control rods during reactor startup (River Bend) On March 8, 2008, during plant startup at approximately 25 percent the rod line was being adjusted in accordance with the Reactivity Control Plan (RCP) in preparation for reactor recirculation pump up-shift. During manipulations, 6 control rods were withdrawn from position 16 to 24 instead of 16 to 20 as indicated on the control rod pull sheet. The root cause evaluation determined that the primary factor in the event was failure to appropriately utilize human performance tools and inadequate oversight.

SER 40-84 Inadvertent initiation of ESF systems due to spurious operation of trip units Various design inadequacies, component failures, and personnel errors associated with dc buses caused voltage variations resulting in tripping of inverters. Following the dc voltage transients, the inverters automatically reset. This caused analog trip units, which are powered by the inverters, to produce short duration actuation signals. The signals were caused by the trip unit's load relay becoming energized before the loop current reestablished itself above the trip setpoint. These momentary signals have inadvertently initiated emergency diesel generators, emergency core cooling systems (ECCS) and other safety- related systems.

INPO ICES 466852 Unusual Event due to failure of Reactor Mode Switch to fully actuate (Peach Bottom) During GP-3 shutdown of Unit 3 on 10/21/19, when the Reactor Operator moved the Unit 3 Reactor Mode Switch to shutdown, it failed to generate a full reactor scram.

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Appendix D Scenario Outline Form ES-D-1 NRC CNS 4/2020 Scenario 3 Page 1 of 12 Facility: Cooper Nuclear Station Scenario No.: 3 Op-Test No.: CN-2020-04 Examiners: ____________________________ Operators: _____________________________

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

1. Raise reactor power using Reactor Recirculation flow

2. Respond to a control rod drifting out

3. Respond to Service Water Pump A trip

4. Respond to Circulating Water pump inlet screen clogging and partial loss of CW pumps, resulting in degraded condenser vacuum

5. Respond to complete loss of Circulating Water pumps, loss of condenser vacuum

6. Respond to hydraulic block ATWS

7. Respond to failure of Reactor Recirc Pump B to trip on ATWS/RPT

8. Respond to SLC B relief valve failing open

9. Respond to trip of RHRSWB pump operating for Suppression Pool Cooling Initial Conditions: Plant operating at 97% power following a sequence exchange near middle of cycle.

Inoperable Equipment: Core Spray Pump A is tagged out of service for motor PMs.

Turnover:

Plant operating at 97% power following a sequence exchange near middle of cycle.

Core Spray Pump A is tagged out of service for motor PMs.

Planned activities for this shift are:

• Raise power to 100% IAW Procedure 2.1.10. There are NO preconditioning holds.

• Continue maintenance on CS Pump A.

Scenario Notes:

This is a new scenario.

Validation Time: XX minutes Page 1 of 12 Rev 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 4/2020 Scenario 3 Page 2 of 12 Event Malf. No. Event Type Event No. Description Raise reactor power using Reactor Recirculation 1 N/A R (ATC,CRS) flow Control rod 18-19 drifts out CT#1 C When an inadvertent reactivity addition occurs (ATC,BOP,CRS) (in this case, rod drift out) that would cause 2 rd101819 power to rise above 100% rated thermal power, A (CREW) crew lowers power to restore and maintain TS (CRS) power 100% rated thermal power by inserting the control rod and/or lowering Reactor Recirc flow before PMIS point NSSRP641 (30 minute power average) exceeds 2419 MWt.

C (BOP,CRS) 3 sw01c Service Water Pump C trip TS (CRS) mc06a mc06b mc06c C (ATC,CRS) Circulating Water pump inlet screen clogging and 4 partial loss of CW pumps, resulting in degraded mc06d A (CREW) condenser vacuum mc05a mc05b Complete loss of Circulating Water pumps, loss of condenser vacuum, hydraulic block ATWS mc05a CT#2 When control rods fail to scram and energy is mc05b discharging to the primary containment (e.g.

5 M (CREW) rd02a SRVs, LOCA), crew injects SLC or inserts all control rods to at least position 02 before rd02b exceeding the Boron Injection Initiation Temperature (BIIT) curve.

Failure of Reactor Recirc Pump B to trip on 6 rr24b C (ATC,CRS) ATWS/RPT 7 sl03b C (ATC,CRS) SLC B relief valve fails open 8 sw04a(b, c, or d) C (BOP,CRS) First RHRSWB pump started for Suppression Pool Cooling trips (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 4/2020 Scenario 3 Page 3 of 12 Quantitative Attributes Table ES-301-4 Actual Attribute Description Target

1. Failure of Reactor Recirc Pump B to trip on ATWS/RPT Malfunctions after 2. SLC B relief valve fails open 1-2 3 EOP entry 3. First RHRSWB pump started for Suppression Pool Cooling trips

1. Control rod 18-19 drifts out

2. Circulating Water pump inlet screen clogging and partial Abnormal Events 2-4 2 loss of CW pumps, resulting in degraded condenser vacuum

1. Complete loss of Circulating Water pumps, loss of Major Transients 1-2 1 condenser vacuum, hydraulic block ATWS EOP entries 1. EOP-3A requiring 1-2 2 2. EOP-6A substantive action EOP contingencies requiring 1 per set 1 1. EOP-7A Contingency #5 - Level/Power Control substantive action

1. (CT#1) When an inadvertent reactivity addition occurs (in this case, rod drift out) that would cause power to rise above 100% rated thermal power, crew lowers power to restore and maintain power 100% rated thermal power by inserting the control rod and/or lowering Reactor Recirc flow before PMIS point NSSRP641 (30 minute Pre-identified power average) exceeds 2419 MWt.

2 2 Critical Tasks

2. (CT#2) When control rods fail to scram and energy is discharging to the primary containment (e.g. SRVs, LOCA), crew injects SLC or inserts all control rods to at least position 02 before exceeding the Boron Injection Initiation Temperature (BIIT) curve.

Normal Events N/A 1 1. none Reactivity 1. Raise reactor power using Reactor Recirculation flow N/A 1 Manipulations

1. Control rod 18-19 drifts out

2. Service Water Pump C trip

3. Circulating Water pump inlet screen clogging with CWP A and B trip Instrument/

Component N/A 6 4. Failure of Reactor Recirc Pump B to trip on Failures ATWS/RPT

5. SLC B relief valve fails open

6. First RHRSWB pump started for Suppression Pool Cooling trips Page 3 of 12 Rev 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 4/2020 Scenario 3 Page 4 of 12

1. Control rod 18-19 drifts out

2. Service Water Pump C trip

3. Circulating Water pump inlet screen clogging with CWP A and B trip Total Malfunctions N/A 6 4. Failure of Reactor Recirc Pump B to trip on ATWS/RPT

5. SLC B relief valve fails open

6. First RHRSWB pump started for Suppression Pool Cooling trips Top 10 systems and operator actions important to risk that are tested:

• RPS, RHR, RHRSW Page 4 of 12 Rev 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 4/2020 Scenario 3 Page 5 of 12 SCENARIO

SUMMARY

The plant is operating at 97% power during power ascension.

After the crew takes the watch, the crew raises power to 100% using Reactor Recirculation flow control.

After the crew has raised power, control rod 18-19 drifts out. The crew responds IAW Procedure 2.4CRD and inserts rod 18-19, but the rod does not latch. The crew scrams rod 18-19 IAW Procedure 2.4 and it then latches. The crew must insert rod 18-19 and/or lower Reactor Recirc flow to prevent the 30-minute power average from exceeding the License Condition limit (CT#1). The CRS enters TS 3.1.3 Condition C for rod 18-19.

After the TS for rod 18-19 has been addressed, Service Water Pump C will trip.

The BOP will start SW Pump B IAW the alarm card. The CRS will enter TS 3.7.2 Condition A for SW Pump C.

After the TS for SW Pump C has been addressed, Circulating Water travelling screens will begin to block due to river debris. After a delay, CW Pump A will trip, followed later by CW Pump B. Condenser vacuum will lower. The crew will respond IAW Procedure 2.4VAC and lower power to stabilize condenser vacuum.

After power has been lowered and vacuum stabilized, CW Pumps C and D will trip due to travelling screen failure. The crew will insert a manual scram IAW Procedure 2.4VAC when vacuum cannot be maintained above 23 Hg.

When the reactor is scrammed, control rods will fail to insert due to blockages in both scram discharge volumes. Reactor power will be approximately 30%.

EOP-6A and 7A are entered via EOP-1A. The crew injects SLC and installs the necessary PTMs to bypass interlocks and insert control rods individually via RMCS. MSIVs will close on low condenser vacuum. With MSIVs closed and SRV operating to control reactor pressure, either SLC must be initiated or all controls rods inserted to at least position 02 before Suppression Pool temperature exceeds the Boron Injection Initiation Temperature (BIIT) curve (CT#2). SLC B relief valve will fail open when SLC Pump B is started.

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. ADS is inhibited when EOP-7A is entered to avert uncontrolled depressurization due to lowering level. HPCI and RCIC are available for RPV level control.

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Appendix D Scenario Outline Form ES-D-1 NRC CNS 4/2020 Scenario 3 Page 6 of 12 Suppression Pool Cooling will be required due to SRV operation. The RHRSWB pump associated with the first loop of SPC attempted to be placed into service will trip, requiring the operator to start the other SWB pump in that RHR loop or transition to the other RHR loop for SPC.

Once several control rods have been inserted, the ATC begins alternately resetting RPS, driving rods individually while allowing the SDV to drain, and reinserting manual scrams. When all control rods have been inserted to at least position 02, the CRS transitions from ATWS to non-ATWS flowcharts, SLC pumps are stopped and RPV level restoration is directed.

The exercise ends when control rods are inserted or Hot Shutdown Boron weight has been injected, and the CRS has reset the level band to +3 to +54.

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Appendix D Scenario Outline Form ES-D-1 NRC CNS 4/2020 Scenario 3 Page 7 of 12 CRITICAL TASK BASIS When an inadvertent reactivity addition occurs (in this case, rod drift out) that would cause power to rise above 100% rated thermal power, crew lowers power to restore and maintain power 100%

Critical Task #1 rated thermal power by inserting the control rod and/or lowering Reactor Recirc flow before PMIS point NSSRP641 (30 minute power average) exceeds 2419 MWt.

License Condition C.1 states 2419 MWt is the Maximum Power Level authorized. The licensed 100% power limit is a basis for assumptions in the plants safety analysis. Sustained operation above 2419 MWt may Safety Significance place unit operation outside of the plant design basis. NUREG 1021 App. D, section D states a CT must be essential to safety, and lists actions to for which operation or correct performance prevents violation of a facility license condition as one example of a CT.

Cues Annunciators 9-5-1/C-4, Rod Drift, Red drift LED lit for rod 18-19 on full core display on panel 9-5, When selected, control rod 18-19 position rising on panel 9-5, Reactor power rising indicated on IRM/APRM recorders NM-NR-46A-D, SPDS, PMIS, Operator selects rod 18-19 and places Emergency Notch Override switch to EMER ROD IN on panel 9-5, and/or Operator Selects S on Measurable RR flow controllers RRFC-SIC-16A(B) on panel 9-4 and lowers RR Performance pump flow (by turning speed demand counter-clockwise on one speed Indicators controller at a time) until power stabilizes below 100% on IRM/APRM recorders NM-NR-46A-D.

Reactor power stabilizes below 100% on IRM/APRM recorders NM-NR-Performance 46A-D and on PMIS points NSSRP640, NSSRP641, NSSRP642, Feedback NSSRP643, and NSSRP645.

Applicability Any time a rod drift out would cause reactor power to exceed 100% with no operator intervention.

License Condition C.1 lists 2419 MWt as the Maximum Power Level for CNS. Procedure 2.1.10, Station Power Changes, sections 10 and 11 describes the methodology for adherence to this limit. A note at step 11.1 states Minor power fluctuations due to automatic control system response, random processes such as bi-stable flow, and flow meter measurement uncertainties are inherent to BWR operating characteristics. Small, short-term fluctuations in power that are not Justification for the under the direct control of a Licensed Reactor Operator are not chosen performance considered intentional. Step 11.1 states It is prohibited to intentionally limit operate greater than the applicable licensed power limit as determined in Section 10. If core thermal power 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> average calculation exceeds applicable licensed power limit, action shall be taken to ensure subsequent hourly average remains less than or equal to applicable limit. Step 11.4 directs monitoring and maintaining PMIS Point NSSRP643 (running 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> average) and PMIS Point NSSRP645 (running 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> average) below 2419 MWt. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> point is the legal record for plant power level. PMIS point NSSRP641 (30 minute Page 7 of 12 Rev 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 4/2020 Scenario 3 Page 8 of 12 average) was chosen because it is more limiting and provides timely update to maintain adequate margin to the core thermal power limit.

BWR Owners Group N/A Appendix Scenario Guide Initial power level must be near 100% and the rod drift out must cause Requirements reactor power to rise above 100% with no operator intervention.

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Appendix D Scenario Outline Form ES-D-1 NRC CNS 4/2020 Scenario 3 Page 9 of 12 Category EOPS When control rods fail to scram and energy is discharging to the primary containment (e.g. SRVs, LOCA), crew injects SLC or inserts all control rods to at least position 02 before Critical Task #2 exceeding the Boron Injection Initiation Temperature (BIIT) curve.

Failure to effect shutdown of the reactor when a RPS setting has been exceeded would unnecessarily extend the level of degradation of the safety of the plant. This could further degrade into damage to the principle fission product barriers if left unmitigated. Action to shut down the reactor is required when RPS and control rod drive systems fail.

The Boron Injection Initiation Temperature (BIIT) is the greater of:

• The highest suppression pool temperature at which initiation of boron injection will permit injection of the Hot Shutdown Boron Weight of boron before suppression pool temperature exceeds the Heat Capacity Temperature Limit.

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

Safety Significance 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.)

Cues 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.

Operator manipulates keylocked switch for SLC A pump to START on panel 9-5. (SLC Pump B relief valve fails open in this scenario.)

Measurable Performance Indicators Operator selects individual control rods by depressing the respective pushbutton on the panel 9-5 matrix and inserts the rod by manipulating the emergency in switch on panel 9-5.

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Appendix D Scenario Outline Form ES-D-1 NRC CNS 4/2020 Scenario 3 Page 10 of 12 SLC A pump red light illuminated, SLC discharge pressure rising, SLC tank level lowering by 26% on panel 9-5 (from ~80% to below

~54% for this scenario).

Performance Feedback Operator selecting and inserting control rods indicated by rod position decreasing to 00 for selected rod on panel 9-5.

Applicability ATWS with power >3% following trip of both recirc pumps per EOP-7A, energy being discharged to Primary Containment causing Torus water temperature to rise.

If boron injection is initiated or all control rods are inserted to position 02 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 Justification for the Temperature Limit even if boron injection is initiated early in the chosen performance event. Since failure-to-scram conditions may present severe plant limit safety consequences, the requirement to initiate boron injection is independent of any anticipated success of control rod insertion.

If the failure to scram EOP were to be exited, other procedures would not provide the guidance for control rod insertion necessary to achieve reactor shutdown. Before exiting EOP-6A ensures guidance to effect reactor shutdown is not removed.

BWR Owners Group App. B, step RC/Q-6 Appendix Initial conditions, combined with the ATWS severity, should result in power >3% following trip of both recirc pumps per EOP-7A.

Scenario Guide Suppression Pool temperature must be rising due to unstoppable Requirements condition such as loss of the main condenser or LOCA. The scenario should be validated to exceed BIIT; therefore, ability to achieve control rod insertion may need to be hampered or delayed.

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Appendix D Scenario Outline Form ES-D-1 NRC CNS 4/2020 Scenario 3 Page 11 of 12 Related Operating Experience INPO ICES 150203 Single Control Rod Drift Out During Reactor Startup Single Control Rod Drift Out During Reactor Startup On December 19, 1994, while withdrawing control rods for unit startup, Control Rod 50-35 drifted from position 06 to 48 (full out). This Control Rod was selected and being withdrawn from position 04 to 06 when the event occurred. Operations personnel recognizing the rod drift condition attempted to stop the control rod by applying an insert signal.

INPO ICES 205668 Control Rod Drift Due to Transponder Card Failure (OE16861)

(Nine Mile Point 1)Control Rod Drift Due to Transponder Card Failure During power ascension from Forced Outage, indications were received that a control rod had drifted.

At the time of the annunciation no control rods were being withdrawn. A failed transponder card was determined to be the cause of event. Previous recurring transponder card failures generally caused a lockup of the Reactor Manual Control System. This particular transponder card failure resulted in a control rod drift.

INPO ICES 230540 Service Water Pump Trip (OE26446)

(Indian Point 2) On 1/27/2008, Service Water Pump tripped on long delay overload due to failure of the 'B' phase termination at 21SWP Motor. 21 Service Water Pump was on the essential SW header at the time thus IP2 entered a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> LCO. Service Water Pump (SWP) tripped on long delay overload due to failure of the 'B' phase termination at 21SWP Motor.

INPO ICES 443918 Service Water Pump Trip (Indian Point 2) On 9/3/2018, Service Water Pump (SWP) breaker tripped open while running, which caused Operations personnel to place the pump in trip-pullout. The 22 SWP was declared inoperable, which resulted in a Maintenance Rule Functional Failure as well as a Mitigating System Performance Index (MSPI) failure of the 22 SWP. The cause for this event was a turn to turn short of the motor. Based on the damage to the motor winding, no definitive apparent cause could be found.

(CNS) CR 2012-02665 Traveling screen trips Received D-1/D-2 traveling screen HI and HI-HI D/P alarms. Started CWP B and D. All traveling screens are in HAND and FAST, current screen coverage is 50%-75%

consisting of grass and leaves due to recent heavy rains in the area.

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Appendix D Scenario Outline Form ES-D-1 NRC CNS 4/2020 Scenario 3 Page 12 of 12 INPO ICES 306387 Trip of Circulating Water Traveling Screen Causes Power Reduction of <10%

(CNS) On dayshift ~ 1701 received A1 Traveling water screen High differential pressure (DP), ~3 minutes later HI-HI DP and A1 Screen Trip alarms were received.

Reactor power was reduced, and Circulating Water Pump was secured. Consequences, downpower of ~9.5% and Reactivity Management Event Level 4. Cause was due to high debris in the river due to weather conditions and a restricted flowpath to 'A' Circulating Water Pump.

SEN 134 Failure of control rods to fully insert (Wolf Creek) After a manual reactor scram on January 30, 1996, five control rods failed to fully insert. One control rod indicated 18 steps out, two control rods indicated 12 steps out, and two control rods indicated six steps out.

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