ML20324A042

CN-2020-09-FINAL Outlines

ML20324A042
Person / Time
Site:	Cooper
Issue date:	09/29/2020
From:	Greg Werner Operations Branch IV
To:	Nebraska Public Power District (NPPD)
References
Download: ML20324A042 (72)

Text

ES-401 BWR Examination Outline Form ES-401-1 Rev. 11 Facility: Cooper Nuclear Station Date of Exam: September 29, 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.

Emergency and Abnormal Plant Evolutions 1

3 4

4 N/A 4

3 N/A 2

20 7

2 1

1 2

1 1

1 7

3 Tier Totals 4

5 6

5 4

3 27 10

2.

Plant Systems 1

2 2

3 3

3 3

2 2

2 2

2 26 5

2 1

1 2

1 1

1 1

1 1

1 1

12 3

Tier Totals 3

3 5

4 4

4 3

3 3

3 3

38 8

3. Generic Knowledge and Abilities Categories 1

2 3

4 10 1

2 3

4 7

3 3

2 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.

ES-401 2

Form ES-401-1 Rev. 11 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant EvolutionsTier 1/Group 1 (RO/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 Forced Core Flow Circulation / 1 & 4 X

Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: (CFR: 41.10 / 43.5

/ 45.13)

AA2.03 Actual core flow 3.3 39 295003 (APE 3) Partial or Complete Loss of AC Power / 6 X

Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF A.C.

POWER: (CFR: 41.8 to 41.10)

AK1.02 Load shedding 3.1 49 295004 (APE 4) Partial or Total Loss of DC Power / 6 X

Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER:

(CFR: 41.5 / 45.6)

AK3.03 Reactor SCRAM 3.1 56 295005 (APE 5) Main Turbine Generator Trip /

3 X

Ability to determine and/or interpret the following as they apply to MAIN TURBINE GENERATOR TRIP: (CFR: 41.10 / 43.5 /

45.13)

AA2.04 Reactor pressure 3.7 51 295006 (APE 6) Scram / 1 X

Knowledge of the reasons for the following responses as they apply to SCRAM: (CFR:

41.5 / 45.6)

AK3.01 Reactor water level response 3.8 54 295016 (APE 16) Control Room Abandonment

/ 7 X

Knowledge of the interrelations between CONTROL ROOM ABANDONMENT and the following: (CFR: 41.7 / 45.8)

AK2.01 Remote Shutdown Panel 4.4*

44 295018 (APE 18) Partial or Complete Loss of CCW / 8 X

Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER: (CFR:

41.8 to 41.10)

AK1.01 Effects on component/system operations 3.5 42 295019 (APE 19) Partial or Complete Loss of Instrument Air / 8 X

Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR:

(CFR: 41.10 / 43.5 / 45.13)

AA2.02 Status of safety-related instrument air system loads 3.6 40 295021 (APE 21) Loss of Shutdown Cooling /

4 X

Knowledge of the operational implications of the following concepts as they apply to LOSS OF SHUTDOWN COOLING: (CFR:

41.8 to 41.10)

AK1.04 Natural circulation 3.6 48 295023 (APE 23) Refueling Accidents / 8 X

Ability to operate and/or monitor the following as they apply to REFUELING ACCIDENTS: (CFR: 41.7 / 45.6)

AA1.01 Secondary containment ventilation 3.3 53 295024 High Drywell Pressure / 5 X

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

(CFR: 41.10 / 45.3) 4.2 58

ES-401 3

Form ES-401-1 Rev. 11 295025 (EPE 2) High Reactor Pressure / 3 X

Knowledge of the interrelations between HIGH REACTOR PRESSURE and the following: (CFR: 41.7 / 45.8)

EK2.09 Reactor power 3.9 47 295026 (EPE 3) Suppression Pool High Water Temperature / 5 X

Knowledge of the interrelations between SUPPRESSION POOL HIGH WATER TEMPERATURE and the following: (CFR:

41.7 / 45.8)

EK2.06 Suppression pool level 3.5 41 295027 (EPE 4) High Containment Temperature (Mark III Containment Only) / 5 295028 (EPE 5) High Drywell Temperature (Mark I and Mark II only) / 5 X

Ability to operate and/or monitor the following as they apply to HIGH DRYWELL TEMPERATURE: (CFR: 41.7 / 45.6)

EA1.03 Drywell cooling system 3.9 50 295030 (EPE 7) Low Suppression Pool Water Level / 5 X

2.1.25 Ability to interpret reference materials, such as graphs, curves, tables, etc. (CFR: 41.10 / 43.5 / 45.12) 3.9 43 295031 (EPE 8) Reactor Low Water Level / 2 X

Knowledge of the interrelations between REACTOR LOW WATER LEVEL and the following: (CFR: 41.7 / 45.8)

EK2.03 Low pressure core spray 4.2 46 295037 (EPE 14) Scram Condition Present and Reactor Power Above APRM Downscale or Unknown / 1 X

Ability to operate and/or monitor the following as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: (CFR: 41.7 / 45.6)

EA1.06 Neutron monitoring system 4.1*

52 295038 (EPE 15) High Offsite Radioactivity Release Rate / 9 X

Knowledge of the reasons for the following responses as they apply to HIGH OFF-SITE RELEASE RATE: (CFR: 41.5 / 45.6)

EK3.03 Control room ventilation isolation 3.7 55 600000 (APE 24) Plant Fire On Site / 8 X

Knowledge of the reasons for the following responses as they apply to PLANT FIRE ON SITE:

AK3.04 Actions contained in the abnormal procedure for plant fire on site 2.8 57 700000 (APE 25) Generator Voltage and Electric Grid Disturbances / 6 X

Ability to operate and/or monitor the following as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES: (CFR: 41.5 and 41.10 /

45.5, 45.7, and 45.8 )

AA1.01 Grid frequency and voltage 3.6 45 K/A Category Totals:

3 4

4 4

3 2

Group Point Total:

20

ES-401 4

Form ES-401-1 Rev. 11 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant EvolutionsTier 1/Group 2 (RO/SRO)

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 295008 (APE 8) High Reactor Water Level / 2 X

Knowledge of the reasons for the following responses as they apply to HIGH REACTOR WATER LEVEL: (CFR: 41.5 /

45.6)

AK3.02 Reactor SCRAM 3.6*

61 295009 (APE 9) Low Reactor Water Level / 2 X

Knowledge of the reasons for the following responses as they apply to LOW REACTOR WATER LEVEL: (CFR: 41.5 /

45.6)

AK3.01 Recirculation pump run back 3.2 65 295010 (APE 10) High Drywell Pressure / 5 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 following as they apply to INCOMPLETE SCRAM: (CFR: 41.10 / 43.5 / 45.13)

AA2.02 Control rod position 4.1*

64 295017 (APE 17) Abnormal Offsite Release Rate / 9 X

2.4.50 Ability to verify system alarm setpoints and operate controls identified in the alarm response manual. (CFR: 41.10 /

43.5 / 45.3) 4.2 60 295020 (APE 20) Inadvertent Containment Isolation / 5 & 7 295022 (APE 22) Loss of Control Rod Drive Pumps / 1 X

Ability to operate and/or monitor the following as they apply to LOSS OF CRD PUMPS: (CFR: 41.7 / 45.6)

AA1.01 CRD hydraulic system 3.1 62 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 X

Knowledge of the interrelations between HIGH SECONDARY CONTAINMENT AREA RADIATION LEVELS and the following: (CFR: 41.7 / 45.8)

EK2.03 Secondary containment ventilation 3.7 63 295034 (EPE 11) Secondary Containment Ventilation High Radiation / 9 X

Knowledge of the operational implications of the following concepts as they apply to SECONDARY CONTAINMENT VENTILATION HIGH RADIATION: (CFR:

41.8 to 41.10)

EK1.02 Radiation releases 4.1 59 295035 (EPE 12) Secondary Containment High Differential Pressure / 5

ES-401 5

Form ES-401-1 Rev. 11 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

2 1

1 1

Group Point Total:

7

ES-401 6

Form ES-401-1 Rev. 11 ES-401 BWR Examination Outline Form ES-401-1 Plant SystemsTier 2/Group 1 (RO/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 X 2.4.18 Knowledge of the specific bases for EOPs. (CFR: 41.10 / 43.1 / 45.13) 3.3 17 205000 (SF4 SCS) Shutdown Cooling X

Ability to (a) predict the impacts of the following on the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

(CFR: 41.5 / 45.6)

A2.08 Loss of heat exchanger cooling 3.3 2

206000 (SF2, SF4 HPCIS)

High-Pressure Coolant Injection X

Knowledge of the operational implications of the following concepts as they apply to HIGH PRESSURE COOLANT INJECTION SYSTEM: (CFR: 41.5 / 45.3)

K5.02 Turbine shaft sealing 2.8 24 207000 (SF4 IC) Isolation (Emergency) Condenser 209001 (SF2, SF4 LPCS)

Low-Pressure Core Spray X

Ability to predict and/or monitor changes in parameters associated with operating the LOW PRESSURE CORE SPRAY SYSTEM controls including: (CFR: 41.5 /

45.5)

A1.01 Core spray flow 3.4 22 209002 (SF2, SF4 HPCS)

High-Pressure Core Spray 211000 (SF1 SLCS) Standby Liquid Control X

Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)

A4.08 System initiation 4.2*

7 212000 (SF7 RPS) Reactor Protection X

Ability to (a) predict the impacts of the following on the REACTOR PROTECTION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 /

45.6)

A2.11 Main steamline isolation valve closure 4.0 19 215003 (SF7 IRM)

Intermediate-Range Monitor X

Knowledge of the effect that a loss or malfunction of the following will have on the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM: (CFR: 41.7 / 45.7)

K6.04 Detectors 3.0 3

215004 (SF7 SRMS) Source-Range Monitor X

Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8)

A4.01 SRM count rate and period 3.9 8

215005 (SF7 PRMS) Average Power Range Monitor/Local Power Range Monitor X

Ability to predict and/or monitor changes in parameters associated with operating the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM controls including:

(CFR: 41.5 / 45.5)

A1.05 Lights and alarms 3.3 26

ES-401 7

Form ES-401-1 Rev. 11 217000 (SF2, SF4 RCIC) Reactor Core Isolation Cooling X

Knowledge of the effect that a loss or malfunction of the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) will have on following: (CFR: 41.7 / 45.4)

K3.03 Decay heat removal 3.5 11 218000 (SF3 ADS) Automatic Depressurization X

Knowledge of the operational implications of the following concepts as they apply to AUTOMATIC DEPRESSURIZATION SYSTEM: (CFR: 41.5 / 45.3)

K5.01 ADS logic operation 3.8 16 223002 (SF5 PCIS) Primary Containment Isolation/Nuclear Steam Supply Shutoff X

Knowledge of the effect that a loss or malfunction of the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF will have on following: (CFR:

41.7 / 45.4)

K3.07 Reactor pressure 3.7 20 239002 (SF3 SRV) Safety Relief Valves X

Knowledge of electrical power supplies to the following: (CFR: 41.7)

K2.01 SRV solenoids 2.8*

23 259002 (SF2 RWLCS) Reactor Water Level Control X

Knowledge of the physical connections and/or cause-effect relationships between REACTOR WATER LEVEL CONTROL SYSTEM and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8)

K1.15 Recirculation flow control system 3.2 5

261000 (SF9 SGTS) Standby Gas Treatment X

Ability to monitor automatic operations of the STANDBY GAS TREATMENT SYSTEM including: (CFR: 41.7 / 45.7)

A3.02 Fan start 3.2 13 262001 (SF6 AC) AC Electrical Distribution X

Ability to monitor automatic operations of the A.C. ELECTRICAL DISTRIBUTION including: (CFR: 41.7 / 45.7)

A3.01 Breaker tripping 3.1 15 262002 (SF6 UPS) Uninterruptable Power Supply (AC/DC)

X Knowledge of the effect that a loss or malfunction of the following will have on the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.): (CFR: 41.7 / 45.7)

K6.02 D.C. electrical power 2.8 21 263000 (SF6 DC) DC Electrical Distribution X

Knowledge of D.C. ELECTRICAL DISTRIBUTION design feature(s) and/or interlocks which provide for the following:

(CFR: 41.7)

K4.02 Breaker interlocks, permissives, bypasses and cross ties 3.1 12 264000 (SF6 EGE) Emergency Generators (Diesel/Jet) EDG X

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

K4.02 Emergency generator trips emergency/LOCA) 4.0 4

300000 (SF8 IA) Instrument Air X

Knowledge of (INSTRUMENT AIR SYSTEM) design feature(s) and or interlocks which provide for the following:

(CFR: 41.7)

K4.03 Securing of IAS upon loss of cooling water 2.8 14

ES-401 8

Form ES-401-1 Rev. 11 400000 (SF8 CCS) Component Cooling Water X

Knowledge of the physical connections and/or cause-effect relationships between CCWS and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8)

K1.01 Service water system 3.2 18 510000 (SF4 SWS*) Service Water (Normal and Emergency) 209001 (SF2, SF4 LPCS)

LowPressure Core Spray X

Knowledge of the effect that a loss or malfunction of the following will have on the LOW PRESSURE CORE SPRAY SYSTEM: (CFR: 41.7 / 45.7)

K6.04 D.C. power 2.8 10 215003 (SF7 IRM)

IntermediateRange Monitor X

Knowledge of the operational implications of the following concepts as they apply to INTERMEDIATE RANGE MONITOR (IRM) SYSTEM: (CFR: 41.5 / 45.3)

K5.03 Changing detector position 3.0 6

217000 (SF2, SF4 RCIC) Reactor Core Isolation Cooling X

Knowledge of electrical power supplies to the following: (CFR: 41.7)

K2.03 RCIC flow controller 2.7*

25 218000 (SF3 ADS) Automatic Depressurization X 2.2.12 Knowledge of surveillance procedures. (CFR: 41.10 / 45.13) 3.7 9

263000 (SF6 DC) DC Electrical Distribution X

Knowledge of the effect that a loss or malfunction of the D.C. ELECTRICAL DISTRIBUTION will have on following:

(CFR: 41.7 / 45.4)

K3.03 Systems with D.C. components (i.e.

valves, motors, solenoids, etc.)

3.4 1

K/A Category Point Totals:

2 2

3 3

3 3

2 2

2 2

2 Group Point Total:

26

ES-401 9

Form ES-401-1 Rev. 11 ES-401 BWR Examination Outline Form ES-401-1 Plant SystemsTier 2/Group 2 (RO/SRO)

System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*

K/A Topic(s)

IR 201001 (SF1 CRDH) CRD Hydraulic X

Knowledge of the effect that a loss or malfunction of the following will have on the CONTROL ROD DRIVE HYDRAULIC System: (CFR: 41.7 /

45.7)

K6.06 Component cooling water systems 2.8 29 201002 (SF1 RMCS) Reactor Manual Control 201003 (SF1 CRDM) Control Rod and Drive Mechanism X

Ability to monitor automatic operations of the CONTROL ROD AND DRIVE MECHANISM including:

(CFR: 41.7 / 45.7)

A3.01 Control rod position 3.7 33 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 215002 (SF7 RBMS) Rod Block Monitor X

Knowledge of electrical power supplies to the following: (CFR: 41.7)

K2.03 APRM channels 2.8 32 216000 (SF7 NBI) Nuclear Boiler Instrumentation X

Knowledge of the operational implications of the following concepts as they apply to NUCLEAR BOILER INSTRUMENTATION:

(CFR: 41.5 / 45.3)

K5.12 Effects on level indication due to rapid changes in void fraction 3.2 27 219000 (SF5 RHR SPC) RHR/LPCI:

Torus/Suppression Pool Cooling Mode X

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

41.7 / 45.5 to 45.8)

A4.01 Pumps 3.8*

35 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 X

Knowledge of the physical connections and/or cause-effect relationships between RHR/LPCI:

TORUS/SUPPRESSION POOL SPRAY MODE and the following:

(CFR: 41.2 to 41.9 / 45.7 to 45.8)

K1.01 Suppression pool 3.6 38 233000 (SF9 FPCCU) Fuel Pool Cooling/Cleanup 234000 (SF8 FH) Fuel-Handling Equipment 239001 (SF3, SF4 MRSS) Main and Reheat Steam

ES-401 10 Form ES-401-1 Rev. 11 239003 (SF9 MSVLCS) Main Steam Isolation Valve Leakage Control 241000 (SF3 RTPRS) Reactor/Turbine Pressure Regulating X

Knowledge of the effect that a loss or malfunction of the REACTOR/TURBINE PRESSURE REGULATING SYSTEM will have on following: (CFR: 41.7 / 45.4)

K3.01 Reactor power 4.1 37 245000 (SF4 MTGEN) Main Turbine Generator/Auxiliary 256000 (SF2 CDS) Condensate X

2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation. (CFR: 41.10 / 43.5 / 45.2 /

45.6) 4.3 30 259001 (SF2 FWS) Feedwater X

Ability to predict and/or monitor changes in parameters associated with operating the REACTOR FEEDWATER SYSTEM controls including: (CFR: 41.5 / 45.5)

A1.06 Feedwater heater level 2.7 34 268000 (SF9 RW) Radwaste 271000 (SF9 OG) Offgas 272000 (SF7, SF9 RMS) Radiation Monitoring 286000 (SF8 FPS) Fire Protection X

Knowledge of FIRE PROTECTION SYSTEM design feature(s) and/or interlocks which provide for the following: (CFR: 41.5 / 41.7 / 45.3 /

45.5 to 45.8)

K4.04 Personnel safety during halon and/or carbon dioxide system actuation 3.6 36 288000 (SF9 PVS) Plant Ventilation X

Knowledge of the effect that a loss or malfunction of the PLANT VENTILATION SYSTEMS will have on following: (CFR: 41.5 / 45.3)

K3.02 Reactor building temperature 2.9 28 290001 (SF5 SC) Secondary Containment 290003 (SF9 CRV) Control Room Ventilation 290002 (SF4 RVI) Reactor Vessel Internals X

Ability to (a) predict the impacts of the following on the REACTOR VESSEL INTERNALS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

(CFR: 41.5 / 45.6)

A2.04 Excessive heatup/cooldown rate 3.7 31 51001 (SF8 CWS*) Circulating Water K/A Category Point Totals:

1 1

2 1

1 1

1 1

1 1

1 Group Point Total:

12

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

Form ES-401-3 Rev. 11 Facility: Cooper Nuclear Station Date of Exam:

Category K/A #

Topic RO SRO-only IR IR 2.1.1 Knowledge of conduct of operations requirements. (CFR: 41.10

/ 45.13) 3.8 69 2.1.4 Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, no-solo operation, maintenance of active license status, 10CFR55, etc. (CFR: 41.10 / 43.2) 3.3 70 2.1.26 Knowledge of industrial safety procedures (such as rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and hydrogen). (CFR: 41.10 / 45.12) 3.4 67 Subtotal 3

2. Equipment Control 2.2.1 Ability to perform pre-startup procedures for the facility, including operating those controls associated with plant equipment that could affect reactivity. (CFR: 41.5 / 41.10 / 43.5

/ 43.6 / 45.1) 4.5 73 2.2.39 Knowledge of less than or equal to one hour Technical Specification action statements for systems. (CFR: 41.7 /

41.10 / 43.2 / 45.13) 3.9 72 2.2.38 Knowledge of conditions and limitations in the facility license.

(CFR: 41.7 / 41.10 / 43.1 / 45.13) 3.6 68 Subtotal 3

3. Radiation Control 2.3.12 Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc. (CFR: 41.12 / 45.9 /

45.10) 3.2 71 2.3.15 Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. (CFR: 41.12 / 43.4 / 45.9) 2.9 66 Subtotal 2

4. Emergency Procedures/Plan 2.4.5 Knowledge of the organization of the operating procedures network for normal, abnormal, and emergency evolutions.

(CFR: 41.10 / 43.5 / 45.13) 3.7 75 2.4.27 Knowledge of fire in the plant procedures. (CFR: 41.10 / 43.5

/ 45.13) 3.4 74 Subtotal 2

Tier 3 Point Total 10 7

ES-401 BWR Examination Outline Form ES-401-1 Rev. 11 Facility: Cooper Nuclear Station Date of Exam: September 29, 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.

Emergency and Abnormal Plant Evolutions 1

N/A N/A 20 3

4 7

2 7

2 1

3 Tier Totals 27 5

5 10

2.

Plant Systems 1

26 2

3 5

2 12 2

1 3

Tier Totals 38 4

4 8

3. Generic Knowledge and Abilities Categories 1

2 3

4 10 1

2 3

4 7

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.

ES-401 2

Form ES-401-1 Rev. 11 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant EvolutionsTier 1/Group 1 (RO/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 Forced Core Flow Circulation / 1 & 4 295003 (APE 3) Partial or Complete Loss of AC Power / 6 295004 (APE 4) Partial or Total Loss of DC Power / 6 X

Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER:

(CFR: 41.10 / 43.5 / 45.13)

AA2.01 Cause of partial or complete loss of D.C. power 3.6 77 295005 (APE 5) Main Turbine Generator Trip /

3 X

2.1.20 Ability to interpret and execute procedure steps. (CFR: 41.10 / 43.5 /

45.12) 4.6 82 295006 (APE 6) Scram / 1 X

2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. (CFR: 41.5 /

43.5 / 45.12 / 45.13) 4.7 79 295016 (APE 16) Control Room Abandonment

/ 7 295018 (APE 18) Partial or Complete Loss of CCW / 8 X

Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER: (CFR: 41.10 / 43.5 /

45.13)

AA2.01 Component temperatures 3.4 78 295019 (APE 19) Partial or Complete Loss of Instrument Air / 8 295021 (APE 21) Loss of Shutdown Cooling /

4 295023 (APE 23) Refueling Accidents / 8 295024 High Drywell Pressure / 5 295025 (EPE 2) High Reactor Pressure / 3 295026 (EPE 3) Suppression Pool High Water Temperature / 5 X

2.4.41 Knowledge of the emergency action level thresholds and classifications. (CFR:

41.10 / 43.5 / 45.11) 4.6 80 295027 (EPE 4) High Containment Temperature (Mark III Containment Only) / 5 295028 (EPE 5) High Drywell Temperature (Mark I and Mark II only) / 5 X 2.2.12 Knowledge of surveillance procedures. (CFR: 41.10 / 45.13) 4.1 76 295030 (EPE 7) Low Suppression Pool Water Level / 5 295031 (EPE 8) Reactor Low Water Level / 2 X

Ability to determine and/or interpret the following as they apply to REACTOR LOW WATER LEVEL: (CFR: 41.10 / 43.5 /

45.13)

EA2.01 Reactor water level 4.6 81 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 Electric Grid Disturbances / 6 K/A Category Totals:

3 4

Group Point Total:

7

ES-401 3

Form ES-401-1 Rev. 11

ES-401 4

Form ES-401-1 Rev. 11 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant EvolutionsTier 1/Group 2 (RO/SRO)

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 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 /

5 295013 (APE 13) High Suppression Pool Temperature. / 5 X

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

(CFR: 41.10 / 43.5 / 45.13)

AA2.02 Localized heating/stratification 3.5 85 295014 (APE 14) Inadvertent Reactivity Addition / 1 295015 (APE 15) Incomplete Scram / 1 295017 (APE 17) Abnormal Offsite Release Rate / 9 295020 (APE 20) Inadvertent Containment Isolation / 5 & 7 X 2.4.6 Knowledge of EOP mitigation strategies. (CFR: 41.10 / 43.5 / 45.13 4.7 83 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 X

Ability to determine and/or interpret the following as they apply to SECONDARY CONTAINMENT HIGH DIFFERENTIAL PRESSURE: (CFR: 41.8 to 41.10)

EA2.01 Secondary containment pressure 3.9 84 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:

3

ES-401 5

Form ES-401-1 Rev. 11 ES-401 BWR Examination Outline Form ES-401-1 Plant SystemsTier 2/Group 1 (RO/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 X

2.4.4 Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures. (CFR: 41.10 / 43.2 / 45.6) 4.7 87 206000 (SF2, SF4 HPCIS)

High-Pressure Coolant Injection 207000 (SF4 IC) Isolation (Emergency) Condenser 209001 (SF2, SF4 LPCS)

Low-Pressure Core Spray 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 X

2.1.32 Ability to explain and apply system limits and precautions. (CFR: 41.10 / 43.2

/ 45.12) 4.0 89 215005 (SF7 PRMS) Average Power Range Monitor/Local Power Range 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 Level Control 261000 (SF9 SGTS) Standby Gas Treatment 262001 (SF6 AC) AC Electrical Distribution X

Ability to (a) predict the impacts of the following on the A.C. ELECTRICAL DISTRIBUTION; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

(CFR: 41.5 / 45.6)

A2.04 Types of loads that, if deenergized, would degrade or hinder plant operation 4.2 86 262002 (SF6 UPS) Uninterruptable Power Supply (AC/DC) 263000 (SF6 DC) DC Electrical Distribution 264000 (SF6 EGE) Emergency Generators (Diesel/Jet) EDG

ES-401 6

Form ES-401-1 Rev. 11 300000 (SF8 IA) Instrument Air X

Ability to (a) predict the impacts of the following on the INSTRUMENT AIR SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: (CFR: 41.5 /

45.6)

A2.01 Air dryer and filter malfunctions 2.8 90 400000 (SF8 CCS) Component Cooling Water X

2.4.2 Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions.

(CFR: 41.7 / 45.7 / 45.8) 4.6 88 510000 (SF4 SWS*) Service Water (Normal and Emergency)

K/A Category Point Totals:

2 3 Group Point Total:

5

ES-401 7

Form ES-401-1 Rev. 11 ES-401 BWR Examination Outline Form ES-401-1 Plant SystemsTier 2/Group 2 (RO/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 X

2.4.6 Knowledge of EOP mitigation strategies. (CFR: 41.10 / 43.5 / 45.13) 4.7 93 202002 (SF1 RSCTL) Recirculation Flow Control 204000 (SF2 RWCU) Reactor Water Cleanup X

Ability to (a) predict the impacts of the following on the REACTOR WATER CLEANUP SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 /

45.6)

A2.10 Valve closures 2.8 91 214000 (SF7 RPIS) Rod Position Information 215001 (SF7 TIP) Traversing In-Core Probe 215002 (SF7 RBMS) Rod Block Monitor 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 X

Ability to (a) predict the impacts of the following on the RHR/LPCI:

CONTAINMENT SPRAY SYSTEM MODE; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

(CFR: 41.5 / 45.6)

A2.11 Motor operated valve failures 3.0 92 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 259001 (SF2 FWS) Feedwater 268000 (SF9 RW) Radwaste 271000 (SF9 OG) Offgas 272000 (SF7, SF9 RMS) Radiation Monitoring 286000 (SF8 FPS) Fire Protection 288000 (SF9 PVS) Plant Ventilation

ES-401 8

Form ES-401-1 Rev. 11 290001 (SF5 SC) Secondary Containment 290003 (SF9 CRV) Control Room Ventilation 290002 (SF4 RVI) Reactor Vessel Internals 51001 (SF8 CWS*) Circulating Water K/A Category Point Totals:

2 1 Group Point Total:

3

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

Form ES-401-3 Rev. 11 Facility: Cooper Nuclear Station Date of Exam:

Category K/A #

Topic RO SRO-only IR IR

1. Conduct of Operations 2.1.35 Knowledge of the fuel-handling responsibilities of SROs. (CFR:

41.10 / 43.7) 3.9 100 2.1.36 Knowledge of procedures and limitations involved in core alterations. (CFR: 41.10 / 43.6 / 45.7) 4.1 98 Subtotal 2

2. Equipment Control 2.2.21 Knowledge of pre-and post-maintenance operability requirements. (CFR: 41.10 / 43.2) 4.1 94 2.2.37 Ability to determine operability and/or availability of safety related equipment. (CFR: 41.7 / 43.5 / 45.12) 4.6 96 Subtotal 2

3. Radiation Control 2.3.6 Ability to approve release permits. (CFR: 41.13 / 43.4 / 45.10) 3.8 97 Subtotal 1

4. Emergency Procedures/Plan 2.4.47 Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material. (CFR: 41.10 / 43.5 / 45.12) 4.2 95 2.4.37 Knowledge of the lines of authority during implementation of the emergency plan. (CFR: 41.10 / 45.13) 4.1 99 Subtotal 2

Tier 3 Point Total 7

ES-401 Record of Rejected K/As Form ES-401-4 Rev. 11 Tier /

Group Randomly Selected K/A Reason for Rejection 2 / 1 223002 / K3.06 (Q20)

Could not develop a psychometrically sound question at the RO level for the subject K/A. Resampled to 223002 / K3.07 2 / 2 241000 / K3.23 (Q37)

Could not develop a psychometrically sound question at the RO level for the subject K/A. Resampled to 241000 / K3.01 1 / 1 295026 / 2.1.30 (Q41)

Not any discriminating local controls associated with high suppression pool temperature. Resampled to 295026 /

AK2.06 1 / 1 295016 / AK2.03 (Q44)

Could not develop a psychometrically sound question at the RO level for the subject K/A. Resampled to 295016 / AK2.01 1 / 1 295024 / 2.4.35 (Q58)

Could not develop a psychometrically sound question at the RO level for the subject K/A. Resampled to 295024 / 2.4.31 3

2.2.23 (Q72)

Could not develop a psychometrically sound question at the RO level for the subject K/A. Resampled to 2.2.39 3

2.4.28 (Q95)

Could not develop a security level question without getting into safeguards information. Resampled to 2.4.47 3

2.1.43 (Q98)

Could not develop a tier 3 question on using procedures for effects of reactivity on plant changes. Everything would be system specific. Resampled to 2.1.36

ES-301 Administrative Topics Outline Form ES-301-1 Rev 1 Facility:

Cooper Nuclear Station Date of Examination:

9/21/2020 Examination Level: RO SRO Operating Test Number:

CN-2020-9 Administrative Topic (see Note)

Type Code*

Describe activity to be performed Conduct of Operations R, M A-1 Determine time to 200 F in the Spent Fuel Pool in accordance with 2.4FPC.

K/A G2.1.25 (3.9)

Conduct of Operations R, M A-2 Determine Actions for a Mispositioned Control Rod K/A G2.1.37 (4.3)

Equipment Control R, M A-3 Determine impact of pulling fuse in RWCU control circuit K/A G2.2.15 (3.9)

Radiation Control R, M A-4 Perform ERP Effluent Release Rate Determination Using SJAE Pathway K/A G2.3.11 (3.8)

Emergency Plan N/A 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)

ES-301 Administrative Topics Outline Form ES-301-1 Rev 1 Facility:

Cooper Nuclear Station Date of Examination:

9/21/2020 Examination Level: RO SRO Operating Test Number:

CN-2020-9 Administrative Topic (see Note)

Type Code*

Describe activity to be performed Conduct of Operations R, N A5 Determine impact of fire in Reactor Building on Reactor Water Level Instrumentation K/A G2.1.20 (4.6)

Conduct of Operations R, D A6 Determine Actions for Chemistry Out of Limits K/A G2.1.34 (3.5)

Equipment Control R, M A-7 Determine Post-Maintenance Testing Requirements for HV-AOV-271AV K/A G2.2.21 (4.1)

Radiation Control R, N A-8 Determine Drywell Entry Conditions to Perform Work in the Drywell on TIP Drives K/A G2.3.13 (3.8)

Emergency Plan R, D A9 Determine Protective Action Recommendations Tabletop #11 K/A G2.4.44 (4.4)

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)

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 page 1 of 2 Rev 6 Facility:

Cooper Nuclear Station Date of Examination:

9/21//2020 Exam Level: RO SRO-I SRO-U Operating Test Number:

CN-2020-9 Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Code*

Safety Function S-1 Shifting CRD pumps, both trip and require a scram K/A 201001 A4.01 (3.1/3.1) Panel 9-5 RO, SRO-I, SRO-U N, A, S 1

S-2 Quick restart of RFP, Thrust Bearing High pressure failure, Trip RFP K/A 259001 A4.02 (3.9/3.7) Panel A RO, SRO-I, SRO-U M, A, L, S 2

S-3 Defeat Group 1 Low Level Isolation during an ATWS (Restoration)

K/A 239001 K4.01 (3.8/3.8) Back Panels RO, SRO-I, SRO-U D, L, EN, S 3

S-4 Perform emergency depressurization with Main Steam Line Drains K/A 239001 A4.02 (3.2/3.2) Panel 9-4 RO, SRO-I N, L, S 4

S-5 Respond to Fuel Element Failure - Align Sump Pumps IAW 5.2FUEL K/A 295036 EA1.01 (3.2/3.3), EK3.04 (3.1/3.4); 290001 A4.10 (3.4/3.3) Back Panels RO, SRO-I D, P, L, S 5

S-6 Start DG1 and just prior to connecting to bus high vibrations alarm comes in and requires tripping.

K/A 264000 K4.01 (3.5/3.7), A4.04 (3.7/3.7) Panel C RO, SRO-I M, A, EN, S 6

S-7 Verify and reset a Group 2 Primary Containment Isolation K/A 215001 A2.07 (3.4/3.7) Back Panels RO D, A, L, EN, S 7

S-8. Separation (Split) of REC Critical Loops IAW Emergency Procedure 5.2REC Att. 5 K/A 400000 4.01 (3.1/3.0) Back Panels RO, SRO-I M, L, EN, S 8

Paired: 1/3, 2/5, 6/8 Not paired: 4, 7

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 page 2 of 2 Rev 6 In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U P-1 Transfer PMIS UPS static switch from inverter to alternate source K/A 262002 K4.01 (3.1/3.4)

RO, SRO-I, SRO-U N, E, R 6

P-2 Startup RPS MG Set A, voltage cannot be adjusted, secure RPS MG Set A K/A 212000 A2.01 (3.7/3.9)

RO, SRO-I, SRO-U D, P, A 7

P-3 Start Diesel Driven Fire Pump using 2.2.30 section 10. Attempt in first manual mode fails, attempt in second manual mode is successful.

K/A 286000 A2.08 (3.2/3.3)

RO, SRO-I M

8 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 (Actual)

(A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power/Shutdown (N)ew or (M)odified from bank including 1(A)

(P)revious 2 exams (R)CA (S)imulator 4-6/4-6 /2-3 (4/5/3) 9/ 8/ 4 (3/4/2) 1/ 1/ 1 (1/1/1) 1/ 1/ 1 (control room system) (3/4/1) 1/ 1/ 1 (5/5/2) 2/ 2/ 1 (7/6/3) 3/ 3/ 2 (randomly selected) (2/2/1) 1/ 1/ 1 (1/1/1)

Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-09 Scenario 1 Page 1 of 51 Page 1 of 51 Rev 3 Facility: Cooper Nuclear Station Scenario No.: 1 Op-Test No.: CN-2020-09 Examiners: ____________________________ Operators:

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

1. Raise power with Reactor Recirculation from 95% to 100%

2. Respond to RRMG A speed control failure high

3. Respond to REC Pump A trip

4. Respond to Minor Earthquake causing Gland Steam Exhauster B trip, Drywell Hydrogen/Oxygen Analyzer failure

5. Respond to RFPT A governor valve failure high

6. Respond to Feedwater Line B break inside containment

7. Respond to failure of control rods 22-23, 22-27, 26-23, 26-27, 30-31 to scram

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

9. Respond to intermittent failure of SRM/IRM drive-in switch

10. Respond to failure of RCIC to automatically start Initial Conditions: Plant operating at 95% power near end of cycle.

Inoperable Equipment: Core Spray Pump A Turnover:

Plant operating at operating at 95% power near end of cycle.

Planned activities for this shift are:

• Raise power from 95% to 100% power Scenario Notes:

This is a new scenario.

Validation Time: 60 minutes

Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-09 Scenario 1 Page 2 of 51 Page 2 of 51 Rev 3 Event No.

Malf. No.

Event Type Event Description 1

N/A R (ATC,CRS)

Raise power from 95% to 100%

2 rr17a I (ATC,CRS)

RRMG A speed control failure high 3

Sw11a C (BOP,CRS)

TS (CRS)

REC Pump A trip (LCO 3.7.3) 4 hv02a (30) zdiarswgseb (STOP)

C (BOP,CRS)

TS (CRS)

A (CREW)

Minor earthquake, Gland Steam Exhauster B trip (start Gland Steam Exhauster A), Drywell Hydrogen/Oxygen Analyzer failure (TLCO 3.3.3) 5 fw24a @ 30 C

(ATC,BOP,CRS)

A (CREW)

RFPT A governor valve failure high (trip RFPT A, trip RRMG A) 6 fw18b @ 20 M (CREW)

Feedwater line B break inside containment with check valve leakage (LOCA, LOFW) 7 rd122223 rd122227 rd122623 rd122627 rd153031 C (ATC,CRS)

Failure of control rods 22-23, 22-27, 26-23, 26-27, 30-31 to scram (attempt to insert rods using RMCS)

CT#1 Inhibit ADS prior to uncontrolled injection from high volume, low pressure systems and before exceeding the Tech Spec cooldown rate limit during a failure to Scram.

CT#2 When Emergency Depressurization is required during a failure to scram condition, stop and prevent all injection into the RPV except for RCIC, Boron Injection, and CRD prior to opening an SRV for emergency depressurization. (Failure to fully stop and prevent a subsystem IAW the hard card before an SRV is opened does not constitute failure of this CT if the subsystem is subsequently aligned such that it does not inject during depressurization.)

8 rd01c C (ATC,CRS)

Failure of Scram Discharge Volume vent and drain valves to automatically close 9

o/r zdinmswz1a7 C (ATC,CRS)

Intermittent failure of SRM/IRM drive-in switch (reinsert SRMs/IRMs) 10 rc01 C (BOP,CRS)

Failure of RCIC to automatically start (attempt to manually align RCIC)

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

Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-09 Scenario 2 Page 1 of 40 Page 1 of 40 Rev 3 Facility: Cooper Nuclear Station Scenario No.: 2 Op-Test No.: CN-2020-09 Examiners: ____________________________ Operators:

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

1. Perform Procedure 15.TG.304, Main Turbine DEH High Pressure Fluid System Functional Test

2. Respond to spurious RCIC initiation

3. Respond to Main Steam Line Radiation Monitor D failure downscale

4. Respond to APRM C failure high

5. Respond to small Recirc Loop B leak in the Drywell

6. Respond to Recirc Loop B line rupture in the Drywell

7. Respond to loss of offsite power

8. Respond to failure of DG2 output breaker to automatically close

9. Respond to HPCI auxiliary oil pump oil leak

10. Respond to injection valve RHR-MO-25A loss of power Initial Conditions: Plant operating at 100% power near the end of the operating cycle.

Inoperable Equipment: Core Spray Pump A.

Turnover:

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

Planned activities for this shift are:

• Perform Procedure 15.TG.304

• Continue preventive maintenance on Core Spray Pump A Scenario Notes:

This is a new scenario.

Validation Time: 60 minutes

Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-09 Scenario 2 Page 2 of 40 Page 2 of 40 Rev 3 Event No.

Malf. No.

Event Type Event Description 1

N/A N (BOP,CRS)

Perform Procedure 15.TG.304 2

rc05 I (ATC,BOP,CRS)

A (CREW)

TS (CRS)

Spurious RCIC initiation (TS 3.5.3)

CT#1 When a spurious RCIC initiation occurs that would cause power to rise above 100% rated thermal power, crew lowers power to restore and maintain power 100% rated thermal power by lowering Reactor Recirc flow before PMIS point NSSRP641 exceeds 2419 MWt.

3 rm02d @ 0 TS (CRS)

Main Steam Line Radiation Monitor D failure downscale (TS 3.3.6.1) 4 nm09c @ 92 C (ATC,CRS)

APRM C failure high (115%)

5 rr20b @ 3 C (ATC,BOP,CRS)

A (CREW)

Small Recirc Loop B leak in the Drywell 6,7 rr20b @ 21 ed05 ed06 ed07 M (CREW)

Recirc Loop B line rupture in the Drywell, loss of offsite power CT#2 When high pressure injection systems cannot maintain RPV level and low pressure ECCS systems fail to automatically start due to failure of DG2 to automatically energize 4160V Bus 1G during loss of offsite AC power, crew manually closes DG2 output breaker to energize LP ECCS systems prior to RPV water level falling below -183 CFZ (TAF)

CT#3 When RPV level lowers to -158 CFZ (TAF) and cannot be maintained above -183 CFZ (MSCWL) and it is apparent to the crew that insufficient high pressure injection systems will be available to restore level, crew Emergency Depressurizes by opening the first of 6 SRVs before RPV level lowers below -183 CFZ. (Momentary shrink below -183 CFZ due to automatic SRV operation in Low-Low Set mode does not constitute failure of this CT.)

Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-09 Scenario 2 Page 3 of 40 Page 3 of 40 Rev 3 8

dg03b C (BOP,CRS)

Failure of DG2 output breaker to automatically close (manually close breaker EG2) 9 hp12 C (ATC,CRS)

HPCI auxiliary oil pump oil leak (place HPIC AOP in PTL) 10 r/f rh17a C (ATC,BOP,CRS)

Injection valve RHR-MO-25A loss of power (align RHR loop A pumps to containment spray or SPC)

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

Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-09 Scenario 3 Page 1 of 38 Rev 2 Facility: Cooper Nuclear Station Scenario No.: 3 Op-Test No.: CN-2020-09 Examiners: ____________________________ Operators:

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

1. Withdraw control rods IAW the control rod sequence package to establish 18-20% bypass valve position

2. Respond to SRM D failure

3. Respond to HPCI-MO-15 loss of power during stroke while placing HPCI in standby

4. Respond to CRD-FCV-19A failing closed

5. Respond to RPS B EPA breaker trip with SGT B fan trip

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

7. Respond to failure of RCIC to automatically isolate on high area temperature

8. Respond to failure RCIC isolation MOVs

9. Respond to 480V Bus 1A loss of power Initial Conditions: Plant operating at 5% power during startup at the beginning of the operating cycle.

Inoperable Equipment: none Turnover:

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

Planned activities for this shift are:

Withdraw control rods IAW the control rod sequence package to establish 18-20% bypass valve position.

Place HPCI in standby IAW Procedure 2.2.33.

Continue startup IAW Procedure 2.1.1.

Scenario Notes:

This is a new scenario.

Validation Time: 60 minutes

Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-09 Scenario 3 Page 2 of 38 Rev 2 Event No.

Malf. No.

Event Type Event Description 1

N/A R (ATC,CRS)

Raise reactor power by withdrawing control rods 2

nm02d @ 100 I (ATC,CRS)

SRM D failure 3

r/f hp04 C (BOP,CRS)

TS (CRS)

Place HPCI in standby, HPCI-MO-15 loss of power during stroke (TS 3.6.1.3/3.5.1) 4 rd04a @ 0 C (ATC,CRS)

A (CREW)

CRD-FCV-19A fails closed 5

rp03c o/r zdisgtswefrf @

off C

(ATC,BOP,CRS)

A (CREW)

TS (CRS)

RPS B EPA breaker trip with SGT B fan trip (TS 3.6.4.3) 6 rc06 @ 7 M

(ATC,BOP,CRS)

RCIC steam line break in the Reactor Building CT#1 When a primary system is discharging into the secondary containment through a break and cannot be isolated, the crew scrams the reactor prior to reaching Maximum Safe Operating Temperatures in two areas. (For this scenario, RCIC-TE-77C Torus 890 ENE is the first area to exceed MSO and RHR-TE-99C Torus 885 NNW is the second area to exceed MSO.)

7 rc07 C (BOP,CRS)

Failure of RCIC to automatically isolate on high area temperature 8

r/f rc06a @ de-ener r/f rc22a @ de-ener C

(ATC,BOP,CRS)

Failure of RCIC steam supply isolation valves to fully close 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 Maximum Safe Operating temperatures are exceeded in two areas and prior to exceeding Maximum Safe Operating temperature in a third area. (For this scenario, RCIC-TE-77C Torus 890 ENE and RHR-TE-99C Torus 885 NNW are the two areas that will exceed MSO first. RWCU-TE-117F Torus 896 W would be the third area to exceed MSO.)

Anticipating Emergency Depressurization and fully opening Bypass valves also satisfies this CT.

9 ed09a C (BOP,CRS) 480V Bus 1A loss of power (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal (TS) Tech Spec Critical Task (As defined in NUREG 1021 Appendix D)

Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-9 Scenario 4 Page 1 of 40 Page 1 of 40 Rev 2 Facility: Cooper Nuclear Station Scenario No.: 4 Op-Test No.: CN-2020-09 Examiners: ____________________________ Operators:

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

1. Place RHR Pump B in Suppression Pool Cooling, RHR-MO-34B failure to open

2. Respond to CRD Pump A trip

3. Respond to Reactor Building Exhaust Ventilation failure

4. Respond to Control Rods 30-23, 38-15, and 26-07 drifting in and failure of the scram air header to vent (ATWS)

5. Respond to failure of ARI to initiate

6. Respond to failure of both SLC pumps

7. Respond to CRD Pump B trip

8. Respond to RCIC speed control failure low in AUTO Initial Conditions: Plant operating at 100% power near the middle of the operating cycle.

Inoperable Equipment: None.

Turnover:

Plant operating at operating at 100% power near the middle of the operating cycle.

Planned activities for this shift are:

• Place RHR Pump B in Suppression Pool Cooling IAW Procedure 2.2.69.3.

Scenario Notes:

This is a new scenario.

Validation Time: 60 minutes

Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-9 Scenario 4 Page 2 of 40 Page 2 of 40 Rev 2 Event No.

Malf. No.

Event Type Event Description 1

r/f rh26a @ de-ener C (BOP,CRS)

TS (CRS)

Place RHR Pump B in Suppression Pool Cooling, RHR-MO-34B failure to open (TS 3.6.2.3) 2 rd08a C (ATC,CRS)

A (CREW)

CRD Pump A trip 3

o/r zdihvswefr1a

@ off o/r zdihvswefr1b

@ stndby C (BOP,CRS)

A (CREW)

TS (CRS)

Reactor Building Exhaust Ventilation failure (TS 3.6.4.1) 4 rd03j rd03i rd03h rp01a rp01b rp01c rp01d rd26 rd27 M (CREW)

Control Rods 30-23, 38-15, and 26-07 drift in, manual scram required, failure of scram air header to vent CT#1 During failure to scram conditions when power is above 30% and the Main Turbine is on line and/or bypass valves are available, prevent MSIV isolation by maintaining reactor water level above -104 inches or by installing MSIV low level isolation bypass jumpers before wide range reactor water level lowers below -

104 inches.

CT#2 Inhibit ADS prior to uncontrolled injection from high volume, low pressure systems and before exceeding the Tech Spec cooldown rate limit during a failure to scram.

5 o/r zdirdsws53aa

@ normal o/r zdirdsws53ba

@ normal I (ATC,CRS)

Failure of ARI to initiate 6

sl01a sl01b C (ATC,CRS)

Trip of both SLC pumps 7

rd08b C (ATC,CRS)

CRD Pump B trip 8

rc04 @ 0 C (BOP,CRS)

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

Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-9 Scenario 4 Page 3 of 40 Page 3 of 40 Rev 2 Quantitative Attributes Table Attribute ES-301-4 Target Actual Description Malfunctions after EOP entry 1-2 4

1. Failure of ARI to initiate

2. Trip of both SLC pumps

3. CRD Pump B trip

4. RCIC flow controller fails low in AUTO Abnormal Events 2-4 2

1. CRD Pump A trip

2. Reactor Building Exhaust Ventilation failure Major Transients 1-2 1

1. Control Rods 30-23, 38-15, and 26-07 drift in, manual scram required, failure of scram air header to vent EOP entries requiring substantive action 1-2 2

1. EOP-6A

2. EOP-7A EOP contingencies requiring substantive action 1 per set 1

1. EOP-7A Contingency #5 - Level/Power Control Pre-identified Critical Tasks 2

3

1. (CT#1) During failure to scram conditions when power is above 30% and the Main Turbine is on line and/or bypass valves are available, prevent MSIV isolation by maintaining reactor water level above -

104 inches or by installing MSIV low level isolation bypass jumpers before wide range reactor water level lowers below -104 inches.

2. (CT#2) Inhibit ADS prior to uncontrolled injection from high volume, low pressure systems and before exceeding the Tech Spec cooldown rate limit during a failure to scram.

Normal Events N/A 0

1. None Reactivity Manipulations N/A 0

1. None Instrument/

Component Failures N/A 7

1. RHR-MO-34B failure to open

2. CRD Pump A trip

3. Reactor Building Exhaust Ventilation failure

4. Failure of ARI to initiate

5. Trip of both SLC pumps

6. CRD Pump B trip

7. RCIC flow controller fails low in AUTO

Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-9 Scenario 4 Page 4 of 40 Page 4 of 40 Rev 2 Total Malfunctions N/A 7

1. RHR-MO-34B failure to open

2. CRD Pump A trip

3. Reactor Building Exhaust Ventilation failure

4. Failure of ARI to initiate

5. Trip of both SLC pumps

6. CRD Pump B trip

7. RCIC flow controller fails low in AUTO Top 10 systems and operator actions important to risk that are tested:

Primary Containment, RHR, RHRSW

Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-9 Scenario 4 Page 5 of 40 Page 5 of 40 Rev 2 SCENARIO

SUMMARY

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

After the crew takes the watch, the BOP will place RHR Pump B in Suppression Pool Cooling IAW Procedure 2.2.69.3. When the operator attempts to open RHR-MO-34B, it will fail to open due to a faulty control circuit. RHR Pump B, operating on minimum flow, must be secured within 15 minutes. The CRS will enter TS 3.6.2.3 Condition A.

After TS for RHR-MO-34B have been addressed, CRD Pump A will trip. The crew will respond IAW the alarm card and place CRD Pump B in service.

After CRD Pump B has been placed into service, Reactor Building Exhaust Fan 1A trips and Fan 1B will not start. Secondary Containment pressure will rise, requiring entry into EOP-5A and TS 3.6.4.1 Condition A. The crew must enter 2.4HVAC to restore Reactor Building negative pressure with SGT.

After response to loss of Reactor Building ventilation is complete, three control rods will simultaneously drift in. The crew will respond IAW the alarm card and insert a manual scram. RPS will de-energize, but the scram air header will fail to vent. Manual initiation of ARI will fail. The crew will enter EOP-1A and transition to EOP-6A and EOP-7A. The crew will trip Recirc pumps. Power will be approximately 50%. The crew will install MSIV low level isolation bypass jumpers (CT#1) and inhibit ADS (CT#2) and lower reactor water level IAW EOP-7A to control power. When reactor water level is lowered and RCIC automatically initiates, its speed controller will fail low in automatic, and the crew will raise RCIC speed in MANUAL.

CRD Pump B will trip, preventing manual control rod insertion. Both SLC pumps will trip after they have been started. The crew must direct the building operator to manually vent the scram air header IAW procedure 5.8.3. If the crew has not manually vented the scram air header, the main turbine will spuriously trip 30 minutes following the scram, resulting in SRVs opening.

After the scram air header has been vented and all control rods have fully inserted, the crew will transition to EOP-1A and stabilize plant conditions. The exercise ends when all control rods are inserted and reactor water level is being restored to the normal band, +3 to +54 inches.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-9 Scenario 4 Page 6 of 40 Page 6 of 40 Rev 2 CRITICAL TASK BASIS Critical Task #1 During failure to scram conditions when power is above 30% and the Main Turbine is on line and/or bypass valves are available, prevent MSIV isolation by maintaining reactor water level above -104 inches or by installing MSIV low level isolation bypass jumpers before wide range reactor water level lowers below -104 inches.

Safety Significance If the MSLs isolate with power >30%, SRV operation is required to control RPV pressure. If the reactor is still generating appreciable power, the Heat Capacity Temperature Limit could then be reached in a relatively short time. Exceeding HCTL would challenge primary containment integrity, especially during ATWS conditions, and would require emergency depressurization. Step C5-2 therefore defeats certain main steam line isolations to prevent closure of the MSIVs and permit continued use of the main condenser as a heat sink, even if RPV water level must be lowered below the low level isolation setpoint.

Cues APRMs indicate >30% power following a scram.

Turbine Generator is on line.

EOP-7A directs lowering reactor water level Measurable Performance Indicators Operator installs jumpers IAW procedure 5.8.20:

Terminals DD-1 to DD-2, BAY-1, PNL 9-15 Terminals BB-1 to BB-2, BAY-3, PNL 9-15 Terminals DD-1 to DD-2, BAY-1, PNL 9-17 Terminals BB-1 to BB-2, BAY-3, PNL 9-17 OR Crew maintains MSIVs open by maintaining RPV water level above -104 indicated on Wide Range instruments:

NBI-LR-1A NBI-LR-1B NBI-LI-85A NBI-LI-85B NBI-LI-85C SPDS Performance Feedback Visual observation jumpers are installed in Panels 9-15 and 9-

17.

Wide Range reactor water level >-104.

MSIVs indicate open on MSIV control switches and Isolation Valve Status mimic on Panel 9-3 Applicability Failure to scram conditions, power >30%.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-9 Scenario 4 Page 7 of 40 Page 7 of 40 Rev 2 Justification for the chosen performance limit The actual setpoint for MSIV (Group 1) isolation on low reactor water level is -104 (TS -113). The actual setpoint is modelled in the simulator.

BWR Owners Group Appendix App. B, Step C5-2 Scenario Guide Requirements Power level following trip of both Recirc pumps should be

>30%. There should be not challenges to Feedwater or HPCI.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-9 Scenario 4 Page 8 of 40 Page 8 of 40 Rev 2 Critical Task #2 Inhibit ADS prior to uncontrolled injection from high volume, low pressure systems and before exceeding the Tech Spec cooldown rate limit during a failure to scram.

Safety Significance In order to effect a reduction in reactor power, actions in Contingency #5 may deliberately lower RPV water level to a level below the automatic initiation setpoint of ADS. Actuation of this system imposes a severe thermal transient on the RPV and complicates the efforts to maintain RPV water level within the ranges specified in Contingency #5. Further, rapid and uncontrolled injection of large amounts of relatively cold, unborated water from low pressure injection systems may occur as RPV pressure decreases to and below the shutoff heads of these pumps. Such an occurrence would quickly dilute in-core boron concentration and reduce reactor coolant temperature. When the reactor is not shutdown, or when the shutdown margin is small, sufficient positive reactivity might be added in this way to cause a reactor power excursion large enough to severely damage the core. Therefore, ADS initiation is purposely prevented as the first action of the level/power control procedure. When required, explicit direction to depressurize the RPV is provided in the PSTG, thereby negating any requirement to maintain the automatic initiation capability of ADS.

Cues ADS Timer Actuated alarm 9-3-1/A-1.

Wide Range and Fuel Zone/CFZ RPV level indications approaching or exceeding Level 1 (-113).

ADS valve control switch red and amber indicating lights on Panel 9-3 ON.

Measurable Performance Indicators Manipulation of ADS A and ADS B Inhibit switches on Panel 9-3 vertical section.

Performance Feedback Inhibit switches click into the vertical, inhibit position on Panel 9-3 prior to breaking the tech spec required cooldown rate.

ADS Timer actuated alarm 9-3-1/A-1.

Applicability ATWS with power >3% following trip of both recirc pumps per EOP-7A.

Justification for the chosen performance limit Inhibiting ADS before injection from high volume, cold water systems occurs ensures a related power excursion will not be experienced that could challenge to the fuel barrier. Inhibiting ADS before the Tech Spec cooldown limit is exceeded ensures the RPV fission product barrier is not challenged by a significant thermal transient.

BWR Owners Group Appendix App. B, Step RC/Q-6

Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-9 Scenario 4 Page 9 of 40 Page 9 of 40 Rev 2 Scenario Guide Requirements The scenario must be designed to make the crew lower RPV level per EOP-7A Step FS/L-6 or FS/L-7 (i.e. ATWS with power >3% following trip of both recirc pumps, and Feedwater or HPCI maintaining level above -60 CFZ).

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 2020-9 Scenario 4 Page 10 of 40 Page 10 of 40 Rev 2 SIMULATOR SET-UP A. Materials Required - none B. Initialize the simulator in IC19 or password protected IC-354, 100% (MOC)

1. Ensure this schedule file is in the schedule file directory:

Schedule File Name - 2020-9sc4.sch

2. Click on Open in the Schedule window and Open Schedule File 2020-9sc4.sch (in the Schedule Directory)

3. In Schedule window, click on the Stopped red block. The red block will change to a green arrow and indicate the schedule file is active (Running).

4. Take the simulator out of freeze.

5. Click the Summary tab in the Director window. Verify the schedule files are loaded and opened per Section C below. (Note: Any actions in the schedule file without a value in the @Time column will not load into the director until the event trigger goes active.)

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 2020-9 Scenario 4 Page 11 of 40 Page 11 of 40 Rev 2 C. File loaded verification:

(Continued on next page.)

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 2020-9 Scenario 4 Page 12 of 40 Page 12 of 40 Rev 2 (Continued on next page.)

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 2020-9 Scenario 4 Page 13 of 40 Page 13 of 40 Rev 2

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 2020-9 Scenario 4 Page 14 of 40 Page 14 of 40 Rev 2

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 2020-9 Scenario 4 Page 15 of 40 Page 15 of 40 Rev 2 D. Panel Setup

1. Ensure PMIS IDTs are blank.

2. Ensure RR Controllers are selected to P.

3. Ensure CRD Pump A is in service and CRD Pump B is in standby.

4. Ensure CRD Drive D/P at 265 psid.

5. Ensure CRD Cooling Water flow at 50 gpm.

6. Ensure RHR SWBP B is in service per Procedure 2.2.70 (steps 6.2.1 thru 6.2.11 complete) in preparation for placing RHR Pump B in Suppression Pool Cooling mode.

7. Ensure all four SW pumps are in service.

8. Ensure SSST Y voltage to RRMGs is set to TAP 2.

9. SSST X Voltage card should read TAP POS 3, MAX 4469 and MIN 4361.

10. Provide markup to crew of Procedures 2.2.69.3 (steps 7.1 thru 7.20 complete) and 2.2.70 (steps 6.2.1 thru 6.2.11 complete) for placing RHR Pump B in Suppression Pool Cooling

11. Ensure IC19, Rod Sequence Notebook under the RO desk.

12. Ensure LPCI B is shown INOP on Safety System Status section of Panel 9-5.

13. Ensure pink flags are removed from all annunciator windows.

14. Ensure Preferred tag is on SGT B.

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 2020-9 Scenario 4 Page 16 of 40 Page 16 of 40 Rev 2 Op-Test No.: CNS 2020-4 Scenario No.:

4 Event No.:

1 Event

Description:

Place RHR Pump B in Suppression Pool Cooling, RHR-MO-34B failure to open Time Position Applicants Action or Behavior CRS Directs BOP to place RHR Pump B in Suppression Pool Cooling per Procedure 2.2.69.3, Section 7, beginning at Step 7.22.

Booth Operator ROLE PLAY:

If asked, as Rx Bldg NLO report RHR Pump B pre-start checks are complete, SAT.

If directed to check breaker for RHR-MO-34B, (fused disconnect 9B on MCC-Y), wait 3 minutes, then report the disconnect is ON and you see nothing abnormal.

If directed to check the position of RHR-MO-34B locally, wait 3 minutes, then report RHR-MO-34B is fully closed.

BOP (Procedure 2.2.69.3 [RHR Suppression Pool Cooling and Containment Spray])

7. PLACING RHR SUBSYSTEM B SUPPRESSION COOLING MODE IN SERVICE 7.22 IF manual override required for 2/3 Core Valve Control Permissive, THEN PERFORM following: (Step is N/A) 7.23 IF manual containment cooling valve control required, THEN PLACE CONTMT COOLING VLV CONTROL PERMISSIVE switch to MANUAL. (Step is N/A) 7.24 OPEN RHR-MO-39B, SUPPR POOL COOLING/TORUS SPRAY VLV.

7.25 IF following exist: (Step is N/A)

RPV less than or equal to 300 psig.

Injection not desired.

THEN CLOSE RHR-MO-27B, OUTBD INJECTION VLV.

NOTE 1 - RHR-MO-16B, LOOP A MIN FLOW BYP VLV, remains open when RHR Subsystem B less than or equal to 2107 gpm.

NOTE 2 - Reactor Building high ambient temperature may cause Suppression Pool temperature to rise before lowering.

CAUTION - RHR pump damage may occur from minimum flow operation greater than 15 minutes.

7.26 START RHR Pump B or D. (Starts RHR Pump B) 7.27 THROTTLE OPEN RHR-MO-34B, SUPPR POOL COOLING INBD THROTTLE VLV, to obtain rated cooling flow or as directed by CRS.

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 2020-9 Scenario 4 Page 17 of 40 Page 17 of 40 Rev 2 Op-Test No.: CNS 2020-4 Scenario No.:

4 Event No.:

1 Event

Description:

Place RHR Pump B in Suppression Pool Cooling, RHR-MO-34B failure to open Time Position Applicants Action or Behavior BOP Recognizes RHR-MO-34B fails to reposition when switch was placed to open, informs CRS.

BOP May send operator to check breaker for RHR-MO-34B, (fused disconnect 9B on MCC-Y).

BOP Advocates securing RHR Pump B due to operation on minimum flow limited to

<15 minutes.

CRS Directs BOP to secure RHR Pump B per Procedure 2.2.69.3.

BOP Secures RHR Pump B IAW Procedure 2.2.69.3, Section 9:

9.1 IF CM-38, PCV-266 BYPASS (R-958-SW), closed in Step 7.29, THEN OPEN CM-38. (Step is N/A) 9.2 IF RHR Pumps B and D running, THEN PERFORM following: (Step is N/A) 9.3 PERFORM following concurrently:

CLOSE RHR-MO-34B, SUPPR POOL COOLING INBD THROTTLE.

(Cannot perform as written due to RHR-MO-34B failure; therefore, closed RHR-MO-39B)

WHEN RHR Subsystem B flow zero, THEN STOP RHR Pump B or D.

(Stops RHR Pump B) 9.4 ENSURE RHR PUMP B switch in auto after stop (green flagged).

9.5 ENSURE RHR PUMP D switch in auto after stop (green flagged). (Step is N/A) 9.6 ENSURE RHR-MO-16B, LOOP B MIN FLOW BYP VLV, open.

CRS Directs BOP to secure RHRSWBP B IAW Procedure 2.2.70.

BOP Secures RHR SWBP B IAW Procedure 2.2.70, RHR Service Water Booster Pump System, Section 8:

8.2 REMOVE SWBP B from service as follows:

8.2.1 ENSURE SWBP Subsystem at ~ 4000 gpm for about 15 minutes prior to securing SWBP B.

8.2.2 ADJUST SW-SW-MO89B, HX-B SW DISCH VLV 89B, to

~ 2800 gpm.

8.2.3 PLACE SWBP B switch to STOP or PULL-TO-LOCK.

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 2020-9 Scenario 4 Page 18 of 40 Page 18 of 40 Rev 2 Op-Test No.: CNS 2020-4 Scenario No.:

4 Event No.:

1 Event

Description:

Place RHR Pump B in Suppression Pool Cooling, RHR-MO-34B failure to open Time Position Applicants Action or Behavior 8.2.4 WHEN SWBP B breaker opens, THEN VERIFY SW-MO-89B, HX-B SW DISCH VLV (Panel 9-3),

closes.

8.2.5 ENSURE SWBP B switch in NORM AFTER STOP (green flagged).

CRS Determines entry into TS 3.6.2.3 Condition A is required for RHR B Suppression Pool Cooling subsystem inoperable.

TS 3.6.2.3 CONDITION REQUIRED ACTION COMPLETION TIME A. One RHR suppression pool cooling subsystem inoperable.

A.1 Restore RHR suppression pool cooling subsystem to OPERABLE status.

7 days END OF EVENT Notes Booth Operator Proceed to next event at the direction of the Lead Examiner.

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 2020-9 Scenario 4 Page 19 of 40 Page 19 of 40 Rev 2 Op-Test No.: CNS 2020-4 Scenario No.:

4 Event No.:

2 Event

Description:

CRD Pump A trip Time Position Applicants Action or Behavior Booth Operator When directed by Lead Examiner, trigger Event 2, rd08a CRD Pump A trip.

ATC Responds to the following annunciator:

9-5-2/A-6, CRD Pump A Breaker Trip Reports CRD Pump A trip.

Performs alarm card 9-5-2/A-6 actions:

1.1 Restore CRD as follows:

1.1.1 Place CRD-FC-301 in MAN.

1.1.2 Adjust CRD-FC-301 to minimum.

1.1.3 WHEN in service FLOW CONTROL VLV AO 19A or 19B indicates closed, THEN start CRD Pump B.

1.1.3.1 IF pump fails to start, THEN go to Step 1.2.

1.1.4 Adjust CRD-FC-301 to obtain flow of 50 gpm.

1.1.5 Balance CRD-FC-301.

1.1.6 Place CRD-FC-301 to BAL.

Reports CRD Pump B has been placed into service CRS Directs ATC to start CRD Pump B IAW alarm card.

Booth Operator ROLE PLAY:

If sent to check CRD Pump A, wait 3 minutes, then report the motor is hotter to the touch than normal, but you see nothing else abnormal.

If sent to check CRD Pump A breaker (EE-CB-480F(CRDP-A)), wait 3 minutes, then report the ground overcurrent relay is tripped for CRD Pump A breaker.

BOP Reports the following annunciator came in and cleared:

C-2/G-3, 480V Bus 1F Ground Sends building operator to investigate CRD Pump A.

Sends operator to check breaker for CRD Pump A, Breaker EE-CB-480F(CRDP-A).

END OF EVENT

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 2020-9 Scenario 4 Page 20 of 40 Page 20 of 40 Rev 2 Op-Test No.: CNS 2020-4 Scenario No.:

4 Event No.:

2 Event

Description:

CRD Pump A trip Time Position Applicants Action or Behavior Notes Booth Operator Proceed to next event when directed by the Lead Examiner.

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 2020-9 Scenario 4 Page 21 of 40 Page 21 of 40 Rev 2 Op-Test No.: CNS 2020-4 Scenario No.:

4 Event No.:

3 Event

Description:

Reactor Building Exhaust Ventilation failure Time Position Applicants Action or Behavior Booth Operator When directed by the Lead Examiner, trigger EVENT 3, (fails RB Exhaust Fans 1A and 1B using their control switches)

BOP Responds to annunciators:

• R-2/E-4, Reactor Bldg Exhaust Fan Failure

• R-2/C-4, Reactor Bldg HV-R-1A Fan Failure

• R-2/D-4, Reactor Bldg Booster Fan Failure Recognizes and reports Reactor Building exhaust and supply fans are not running.

Attempts to start Reactor Building exhaust fans.

Reports Reactor Building exhaust fans cannot be started.

Updates crew that high reactor building pressure is EOP-5A entry condition if D/P above 0.

CRS Enters 2.4HVAC.

If Reactor Building pressure rises to 0 wg, enters EOP-5A. (not expected due to scenario timing)

Directs BOP to perform 2.4HVAC subsequent actions.

Performs 2.4HVAC Attachment 1 actions to start SGT:

(Steps 1 and 2 are N/A)

3. IF Reactor Building supply and exhaust fans cannot be started, THEN perform following:

3.1 At VBD-K, start SGT System per Procedure 2.2.73 to maintain Reactor Building pressure negative.

Performs 2.2.73 section 6:

6.1 At VBD-R, place SGT-DPIC-546, RX BLDG/SGT DP, in AUTO 6.2 ADJUST SGT-DPIC-546 setpoint to less than or equal to -0.25" wg.

Note to Examiners: The operator is likely to go to step 6.3.2 because SGT B is labeled Preferred, and step 6.3.1 will be N/A.

NOTE 1 - Both SGT Subsystems may be placed in service for EOPs.

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 2020-9 Scenario 4 Page 22 of 40 Page 22 of 40 Rev 2 Op-Test No.: CNS 2020-4 Scenario No.:

4 Event No.:

3 Event

Description:

Reactor Building Exhaust Ventilation failure Time Position Applicants Action or Behavior NOTE 2 - Preferred fan does not set a requirement but is meant to help equalize charcoal run time. Either fan may be run at any time based upon circumstance.

CAUTION - Secondary Containment pressure transient and TS LCO 3.6.4.1 entry may occur from starting SGT.

6.3 PERFORM one of following (VBK-K):

6.3.1 PLACE EF-R-1E, SGT A EXHAUST FAN, in service.

6.3.1.1 PLACE EF-R-1E in RUN.

6.3.1.2 VERIFY following:

SGT-AO-249, SGT A INLET, opens.

SGT-AO-251, SGT A DISCHARGE, opens.

6.3.2 PLACE EF-R-1E, SGT B EXHAUST FAN, in service.

6.3.1.1 PLACE EF-R-1FE in RUN.

6.3.1.2 VERIFY following:

SGT-AO-250, SGT B INLET, opens.

SGT-AO-252, SGT B DISCHARGE, opens.

6.4 REMOVE Reactor Building H&V System from service per Procedure 2.2.47.

6.5 ADJUST SGT-DPIC-546, RX BLDG/SGT DP, as necessary, to obtain greater than or equal to 800 scfm on SGT-FI-545, SGT DISCHARGE HEADER FLOW, to maintain electric heater energized.

Sends operator to investigate RB HVAC failure.

Booth Operator Role Play: If sent to investigate RB Exhaust Fan failure, wait 5 minutes, then report you can find nothing unusual.

CRS Determines entry into TS 3.6.4.1 Condition A is required for the period of time secondary containment pressure above -0.25. (TS SR 3.6.4.1.1)

TS 3.6.4.1 CONDITION REQUIRED ACTION COMPLETION TIME A. Secondary containment inoperable in MODE 1, 2, or 3.

A.1 Restore secondary containment to OPERABLE status.

4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> END OF EVENT

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4 Event No.:

3 Event

Description:

Reactor Building Exhaust Ventilation failure Time Position Applicants Action or Behavior Notes Booth Operator Proceed to next event when directed by the Lead Examiner.

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4 Event No.:

4, 5 Event

Description:

Control rods 30-23, 38-15, and 26-07 drift in, manual scram required, failure of scram air header to vent, failure of ARI to initiate Time Position Applicants Action or Behavior Booth Operator When directed by Lead Examiner, trigger EVENT 4, rd03j, rd03i, rd03h - Control rods 30-23, 38-15, and 26-07 drift in Proceed to next event when ATC reports ATWS conditions.

ATC Recognizes and reports alarm:

9-5-1/C-4, Rod Drift Reports 2.4CRD entry condition.

Performs alarm card actions:

2.1 Check full core display for rod drift light(s) to determine which rod(s) is drifting. (Determines and reports drift lights on for control rods 30-23, 38-15, and 26-07.)

2.2 IF more than one rod is drifting, THEN SCRAM and concurrently enter Procedure 2.1.5.

CRS Enters 2.4CRD [CRD Trouble].

Directs ATC to insert manual scram.

ATC Performs 2.1.5 Mitigating Task Scram Actions:

1.1 Press both RX SCRAM buttons.

1.2 Place REACTOR MODE switch to REFUEL.

1.3 IF reactor power > 3%, THEN perform following:

1.3.1 Place REACTOR MODE switch to SHUTDOWN.

1.3.2 Initiate ARI.

Recognizes and reports ARI failure due to failure to ARM.

ATC Perform Attachment 2 Reactor Power Control of 2.1.5 as follows:

1 REACTOR POWER CONTROL 1.1 Ensure REACTOR MODE switch is in SHUTDOWN.

1.2 Verify all SDV vent and drain valves are closed.

1.3 Ensure operating RR pumps have run back to 22% speed 1.4 Verify all control rods are fully inserted.

Recognizes and reports not all rods fully inserted.

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4 Event No.:

4, 5 Event

Description:

Control rods 30-23, 38-15, and 26-07 drift in, manual scram required, failure of scram air header to vent, failure of ARI to initiate Time Position Applicants Action or Behavior ATC Recognizes and reports scram pilot air header failed to vent due to absence of annunciators:

9-5-2/F-5, Scram Valve Pilot Air Low Pressure 9-5-1/G-4, North SDV Not Drained 9-5-1/G-5, South SDV Not Drained END OF EVENT Notes Booth Operator Proceed to next event when ATC reports ATWS conditions.

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4 Event No.:

6, 7, 8 Event

Description:

Trip of both SLC pumps, CRD Pump B trip, RCIC speed controller failure low in AUTO Time Position Applicants Action or Behavior Booth Operator Both SLC pumps will trip 10 seconds after they are started.

CRD Pump trips 30 seconds after the manual scram.

CRS Enter EOP 1A [RPV Control] and transition to EOP 6A [RPV Pressure/Power Control (Failure-to-Scram)] and EOP 7A [RPV Level Control (Failure-to-Scram)]:

CRS (EOP-6A)

Directs BOP to:

Stabilize pressure <1050 psig using Table 12 systems as necessary

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4 Event No.:

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Description:

Trip of both SLC pumps, CRD Pump B trip, RCIC speed controller failure low in AUTO Time Position Applicants Action or Behavior CRS (EOP-6A)

Directs ATC to runback RR pumps to minimum speed, then trip RR pumps:

ATC Reports RR pumps have runback to minimum speed.

Trips RR pumps by placing switches DRIVE MOTOR BKR 1CN and DRIVE MOTOR BKR 1DS to STOP.

Reports reactor power ~48%

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4 Event No.:

6, 7, 8 Event

Description:

Trip of both SLC pumps, CRD Pump B trip, RCIC speed controller failure low in AUTO Time Position Applicants Action or Behavior CRS (EOP-6A)

Directs ATC to Reset ARI and insert rods IAW EOP-6A Step FS/Q-17.

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4 Event No.:

6, 7, 8 Event

Description:

Trip of both SLC pumps, CRD Pump B trip, RCIC speed controller failure low in AUTO Time Position Applicants Action or Behavior CRS Directs ATC to inject SLC before BIIT curve is exceeded IAW EOP-6A:

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4 Event No.:

6, 7, 8 Event

Description:

Trip of both SLC pumps, CRD Pump B trip, RCIC speed controller failure low in AUTO Time Position Applicants Action or Behavior CRITICAL TASK #1: During failure to scram conditions when power is above 30% and the Main Turbine is on line and/or bypass valves are available, prevent MSIV isolation by maintaining reactor water level above -104 inches or by installing MSIV low level isolation bypass jumpers before wide range reactor water level lowers below -104 inches.

CRS Directs BOP to defeat MSIV low water level isolation IAW Procedure 5.8.20

[EOP Plant Temporary Modifications]

Bypasses MSIV low water level isolation IAW Procedure 5.8.20:

3.

BYPASSING LOW RPV WATER LEVEL MSIV ISOLATION 4.1 IF any MSIV closed, THEN place its respective control switch to CLOSE (PNL 9-3).

4.1.1 MS-AO-AO80A.

4.1.2 MS-AO-AO80B.

4.1.3 MS-AO-AO80C.

4.1.4 MS-AO-AO80D.

4.1.5 MS-AO-AO86A.

4.1.6 MS-AO-AO86B.

4.1.7 MS-AO-AO86C.

4.1.8 MS-AO-AO86D 4.2 Ensure following valves closed (PNL 9-4):

4.2.1 RR-AO-740, OUTBD ISOL VLV.

4.2.2 RR-AO-741, IND ISOL VLV.

4.3 To bypass Group 1 Low RPV Water Level Isolation interlock to maintain any MSIVs open or to open MSIVs to re-establish main condenser as a heat sink, perform following; N/A if not required by plant conditions:

4.3.1 Install MSIV LOW WATER LEVEL ISOL BYPASS between Terminals DD-1 and DD-2 (BAY-1, PNL 9-15).

4.3.2 Install MSIV LOW WATER LEVEL ISOL BYPASS between Terminals BB-1 and BB-2 (BAY-3, PNL 9-15).

4.3.3 Install MSIV LOW WATER LEVEL ISOL BYPASS between Terminals DD-1 and DD-2 (BAY-1, PNL 9-17).

4.3.4 Install MSIV LOW WATER LEVEL ISOL BYPASS between Terminals BB-1 and BB-2 (BAY-3, PNL 9-17).

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4 Event No.:

6, 7, 8 Event

Description:

Trip of both SLC pumps, CRD Pump B trip, RCIC speed controller failure low in AUTO Time Position Applicants Action or Behavior CRITICAL TASK #2: Inhibit ADS prior to uncontrolled injection from high volume, low pressure systems and before exceeding the Tech Spec cooldown rate limit during a failure to Scram.

CRS Directs BOP to inhibit ADS IAW EOP-7A:

BOP Inhibit ADS when directed by CRS:

At Panel 9-3, places ADS A and ADS B INHIBIT switches to INHIB.

CRS Directs BOP to stop and prevent injection to lower RPV level below -60 inches.

Performs Stop and Prevent IAW EOP 5.8 [Emergency Operating Procedures (EOPs)] Att. 4 HARD CARD:

1. STOP INJECTION 1.1 Stop HPCI by performing one of following:

1.1.1 IF HPCI is not running, THEN place AUXILIARY OIL PUMP switch to PULL-TO-LOCK.

1.1.2 IF HPCI is running, THEN perform one of following:

EOP 7A RPV Level EOP 7A RPV Level

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4 Event No.:

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Description:

Trip of both SLC pumps, CRD Pump B trip, RCIC speed controller failure low in AUTO Time Position Applicants Action or Behavior BOP BOP 1.1.2.1 Place HPCI controller to MANUAL and lower on controller to maintain > 100 psig below low-end of RPV pressure band, and

a. Maintain turbine speed > 2050 rpm.

1.1.2.2 Trip HPCI turbine by performing following:

a. Ensure AUXILIARY OIL PUMP control switch in START.

b. Press and hold TURBINE TRIP button.

c. After turbine stops, THEN place AUXILIARY OIL PUMP switch to PULL-TO-LOCK.

d. Release TURBINE TRIP button.

1.2 Stop feedwater by performing following:

1.2.1 IF ENABLE INJECTION button is yellow, THEN on STARTUP VALVE screen, press ENABLE INJECTION button and press "YES" to confirm cancel.

1.2.2 At a RVLC/RFPT HMI, select STARTUP VALVE screen, press STOP AND PREVENT button, and confirm "YES" in pop-up box.

1.2.3 WHEN RF-MO-29 and RF-MO-30 are closed, THEN on STARTUP VALVE screen, press ENABLE INJECTION button and confirm "YES" in pop-up box.

1.2.4 IF RF-MO-29 or RF-MO-30 are open and cannot be closed from Control Room, THEN perform following:

1.2.4.1 Trip both RFP's.

1.2.4.2 Trip all operating condensate booster pumps.

CAUTION - If Core Spray and RHR pumps are placed in PULL-TO-LOCK before system flow is reduced to minimum, draining of system may occur.

1.3 Place both core spray pumps in PULL-TO-LOCK.

1.4 Stop RHR by ensuring one of following:

1.4.1 Both RHR Systems secured with pumps in PULL-TO-LOCK.

1.4.2 RHR outboard injection valves automatic open signal bypassed per Procedure 5.8.20 (PTMs 97 through 100) with injection valves closed.

1.4.3 IF RPV pressure is maintained 500 psig, THEN operate RHR aligned to suppression pool cooling and/or containment spray per Procedure 2.2.69.3.

2. PREVENT INJECTION

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4 Event No.:

6, 7, 8 Event

Description:

Trip of both SLC pumps, CRD Pump B trip, RCIC speed controller failure low in AUTO Time Position Applicants Action or Behavior BOP 2.1 Prevent both RHR Subsystems by performing one of following in each loop:

2.1.1 Both RHR Systems secured with pumps in PULL-TO-LOCK.

2.1.2 IF RPV pressure is maintained 500 psig, THEN operate RHR aligned to suppression pool cooling and/or containment spray per Procedure 2.2.69.3; and 2.1.2.1 Bypass RHR outboard injection valves automatic open signal per Procedure 5.8.20 (PTMs 97 through 100) with injection valves closed.

2.1.3 RHR outboard injection valves automatic open signal bypassed per Procedure 5.8.20 (PTMs 97 through 100) with injection valves closed.

2.2 Prevent feedwater by performing following:

2.2.1 Ensure RF-MO-29 is closed.

2.2.2 Ensure RF-MO-30 is closed.

2.2.3 Trip condensate and condensate booster pump(s), as required.

2.3 Prevent CS by performing following:

2.3.1 Ensure CS-MO-12A is closed.

2.3.2 Ensure CS Pump A control switch in PULL-TO-LOCK.

2.3.3 Ensure CS-MO-12B is closed.

2.3.4 Ensure CS Pump B control switch in PULL-TO-LOCK.

2.4 Prevent HPCI by performing following:

2.4.1 IF HPCI is not running, THEN ensure AUXILIARY OIL PUMP switch is in PULL-TO-LOCK.

2.4.2 IF HPCI is running, THEN trip HPCI turbine by performing following:

2.4.2.1 Press and hold TURBINE TRIP button.

2.4.2.2 WHEN turbine is at zero rpm, THEN place AUXILIARY OIL PUMP switch to PULL-TO-LOCK.

2.4.2.3 Release TURBINE TRIP button.

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4 Event No.:

6, 7, 8 Event

Description:

Trip of both SLC pumps, CRD Pump B trip, RCIC speed controller failure low in AUTO Time Position Applicants Action or Behavior Reports injection systems are stopped and prevented.

Booth Operator ROLE PLAY:

When directed by BOP to install EOP PTMs97-100, wait 3 minutes then put in the overrides for the PTMs. Report back to BOP when PTMs installed.

BOP Direct NLO to install EOP PTMs 97 through 100 for RHR injection valve control.

BOP Recognizes and reports RCIC flow controller failed low when reactor water level lowers below -42 Wide Range.

Places RCIC-FIC-91 in MANUAL on Panel 9-4 and raises output to establish injection flow.

BOP When RPV level lowers to less than -60 inches, maintains RPV level between

-60 inches and -183 inches using EOP 5.8.13 systems (Feedwater, HPCI and/or RCIC, CRD and SLC).

ATC When directed by the CRS inject SLC IAW Procedure 2.2.74, Standby Liquid Control System, Section 6:

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4 Event No.:

6, 7, 8 Event

Description:

Trip of both SLC pumps, CRD Pump B trip, RCIC speed controller failure low in AUTO Time Position Applicants Action or Behavior

• Place both keys in SLC PUMP A and SLC PUMP B keylock switches on Panel 9-5 and place switches to START.

• Check both SLC pumps start.

• Check white SQUIB VALVE READY DS-3A (1106A) and SQUIB VALVE READY DS-3B (1106B) lights turn off (Panel 9-5).

• Check pressure on SLC-PI-65, PUMP PRESSURE (Panel 9-5), is greater than reactor pressure.

• Check Annunciator 9-5-2/G-7, LOSS OF CONT TO SQUIB VLVS, alarms.

• Ensure RWCU-MO-15, INBD ISOL VLV (Panel 9-4), is closed.

• Ensure RWCU-MO-18, OUTBD ISOL VLV (Panel 9-4), is closed.

• Ensure both RWCU pumps are off (Panel 9-4).

• Ensure RWCU-MO-74, DEMIN SUCTION BYPASS VLV (Panel 9-4), is throttled open.

Provides CRS initial SLC tank level indicated on SLC-LI-66 TANK LEVEL on Panel 9-5 (~80%).

Recognizes and reports SLC Pumps A and B tripped.

CRS Directs ATC to perform Alternate Boron Injection using RCIC IAW Procedure 5.8.8.

ATC Directs building operator to perform local alignments for Alternate Boron Injection using RCIC IAW Procedure 5.8.8 Section 7.

ATC Performs 5.8.3 [Alternate Rod Insertion Methods] Attachment 1 flowchart:

Recognizes scram valves are NOT open by absence of scram valve Places RPS TEST A1, A2, B1, and B2 TRIP LOGIC switches on Panels 9-15 and 9-17 to TRIP, then returns switches to NORMAL

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4 Event No.:

6, 7, 8 Event

Description:

Trip of both SLC pumps, CRD Pump B trip, RCIC speed controller failure low in AUTO Time Position Applicants Action or Behavior Booth Operator ROLE PLAY:

When directed to manually vent the scram air header, wait 5 minutes, then delete malfunction rd26 (Backup Scram Valve failure), and report the scram air header is being vented.

ATC (Procedure 5.8.3, Attachment 1)

Sends building operator to manually vent the scram pilot air header IAW procedure 5.8.3, Section 8.

ATC (Procedure 5.8.3, Attachment 1)

Recognizes and reports CRD Pump B tripped; therefore, no CRD pump available to drive rods using RMCS.

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4 Event No.:

6, 7, 8 Event

Description:

Trip of both SLC pumps, CRD Pump B trip, RCIC speed controller failure low in AUTO Time Position Applicants Action or Behavior ATC Recognizes and reports all control rods have fully inserted due to scram air header venting and reactor power is lowering to 0%.

Directs building operator to restore the scram air header IAW Procedure 5.8.3, Section 8.

CRS When all rods have been inserted to at least position 02, exits EOP-6A and EOP-7A and re-enters EOP-1A, directs EOP-1A Step RC/L-3 Directs BOP to slowly raise RPV level to new band of +3 to +54.

BOP Secures HPCI and RCIC.

Slowly raises reactor water level by throttling Feedwater Startup FCV using HMI.

NOTE to Examiners: The main turbine is expected to trip on reverse power due to the crew manually venting the scram air header. However, if the crew has not manually vented the scram air header, the main turbine will spuriously trip 30 minutes after the crew initially attempts to scram the reactor.

BOP Recognizes and reports annunciator:

B-1/A-1, Turbine Trip Verifies turbine trip IAW 2.2.77 [Main Turbine], Att. 2 hard card:

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4 Event No.:

6, 7, 8 Event

Description:

Trip of both SLC pumps, CRD Pump B trip, RCIC speed controller failure low in AUTO Time Position Applicants Action or Behavior 1.1 Simultaneously DEPRESS TURB TRIP 1 and TURB TRIP 2 buttons:

(N/A if turbine automatically tripped; BOP will check following) 1.2 VERIFY following valves closed (Panel B, DEH Main display):

Both stop valves.

All governor valves.

All reheat stop valves.

All interceptor valves.

1.3 ENSURE following breakers open (Panel C):

1.3.1 345KV BUS, PCB-3310.

1.3.2 345KV BUS, PCB-3312.

1.3.3 GEN EXCITER FIELD BKR.

CRS If Suppression Pool Temperature exceeds 95°F, enters EOP-3A [Primary Containment Control].

Directs BOP to place RHR A in Suppression Pool Cooling mode.

BOP Places RHR A in Suppression Pool Cooling IAW Procedure 2.2.69.3 [RHR Suppression Pool Cooling and Containment Spray], Attachment 1 [RHR System Operations Hard Card]:

1. SUPPRESSION POOL COOLING 1.1 PLACE RHR SW System in service.

1.1.1 START SWBP(s).

1.1.2 ADJUST SW-MO-89A to maintain flow between 2500 and 4000 gpm.

1.2 If required, with CRS permission, PLACE CONTMT COOLING 2/3 CORE VALVE CONTROL PERMISSIVE switch to MANUAL OVERRD. (Step is N/A) 1.3 If required, PLACE CONTMT COOLING VLV CONTROL PERMISSIVE switch to MANUAL.

1.4 ENSURE RHR-MO-39A open.

1.5 IF reactor pressure less than or equal to 300 psig and injection not desired, THEN CLOSE RHR-MO-27A, OUTBD INJECTION VLV. (Step is N/A) 1.6 ENSURE RHR PUMP running.

CAUTION - RHR pump damage may occur from minimum flow operation greater than 15 minutes.

1.7 THROTTLE RHR-MO-34A to obtain desired cooling flow.

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4 Event No.:

6, 7, 8 Event

Description:

Trip of both SLC pumps, CRD Pump B trip, RCIC speed controller failure low in AUTO Time Position Applicants Action or Behavior 1.8 THROTTLE RHR-MO-38A to split flow for torus sprays to maintain desired containment pressure, if required. (Step is N/A) 1.9 THROTTLE RHR-MO-66A to obtain desired cooling rate.

1.10 IF directed by EOP 3A, THEN OBTAIN maximum cooling.

1.11 IF PCIS Group 6 lights lit on Panel 9-5, THEN ENSURE one of following open:

• REC-MO-711.

• REC-MO-714.

ATC Maintains level +3 to + 54 using feedwater, as directed.

NOTE to Examiners: Scenario objectives have been met when all rods have inserted and level is being raised to the normal band of +3 inches to +54 inches, OR at Lead Examiners discretion.

END OF SCENARIO Notes

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 2020-9 Scenario 2 Page 40 of 40 Page 40 of 40 Rev 2 INITIAL CONDITIONS A.

Plant Status:

1.

100% power, steady state, near the middle of the operating cycle.

B.

Tech. Spec. Limitations in effect:

1.

Day 1 of 7-day LCO per 3.5.1 A.1 for LPCI mode of RHR subsystem B INOPERABLE in preparation for placing RHR Pump B in Suppression Pool Cooling.

C.

Significant problems/abnormalities:

1.

none D.

Risk

1.

PRA Risk is Green

2.

Aggregate Risk is 0 E.

Evolutions/maintenance for the on-coming shift:

1.

BOP to place RHR Pump B in Suppression Pool Cooling per Procedure 2.2.69.3, RHR Suppression Pool Cooling and Containment Spray, Section 7, beginning at Step 7.22.

(RHR SWBP B has already been placed in service to support Suppression Pool Cooling per Procedure 2.2.70.)

2.

Continue full power operation.