ML19099A116

2019-02 Draft Outlines

ML19099A116
Person / Time
Site:	Columbia
Issue date:	04/09/2019
From:	Greg Werner Operations Branch IV
To:	Energy Northwest
References
Download: ML19099A116 (68)

Text

ES-401 BWR Examination Outline Form ES-401-1 Rev. 11 Facility: Columbia Generating Station Date of Exam:

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 3

3 N/A 4

4 N/A 3

20 7

2 1

1 1

1 2

1 7

3 Tier Totals 4

4 4

5 6

4 27 10

2.

Plant Systems 1

2 2

2 2

3 3

3 3

2 2

2 26 5

2 1

1 1

2 1

1 1

1 1

1 1

12 3

Tier Totals 3

3 3

4 4

4 4

4 3

3 3

38 8

3. Generic Knowledge and Abilities Categories 1

2 3

4 10 1

2 3

4 7

3 3

3 3

Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outline sections (i.e., except for one category in Tier 3 of the SRO-only section, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 radiation control K/A is allowed if it is replaced by a K/A from another Tier 3 category.)

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

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

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

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

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

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

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

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

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

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

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

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 operate and/or monitor the following as they apply to Partial or Complete Loss of Forced Core Flow Circulation: (CFR: 41.7 / 45.6)

AA1.06 Neutron monitoring system 3.3 1

295003 (APE 3) Partial or Complete Loss of AC Power / 6 X

Knowledge of the interrelations between Partial or Complete Loss of A.C. Power and the following: (CFR: 41.7 / 45.8)

AK2.01 Station batteries 3.2 2

295004 (APE 4) Partial or Total Loss of DC Power / 6 X

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

(CFR: 41.10 / 45.3) 4.2 3

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)

AA2.06 Feedwater temperature 2.6 4

295006 (APE 6) Scram / 1 X

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

41.5 / 45.6)

AK3.05 Direct turbine generator trip 3.8 5

295016 (APE 16) Control Room Abandonment

/ 7 X

Ability to operate and/or monitor the following as they apply to Control Room Abandonment: (CFR: 41.7 / 45.6)

AA1.02 Reactor/turbine pressure regulating system 2.9*

6 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)

AA2.05 System pressure 2.9 7

295019 (APE 19) Partial or Complete Loss of Instrument Air / 8 X

Knowledge of the interrelations between Partial or Complete Loss of Instrument Air and the following: (CFR: 41.7 / 45.8)

AK2.19 RHR/LPCI 2.7 8

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)

AK1.01 Decay heat 3.6 9

295023 (APE 23) Refueling Accidents / 8 X

Knowledge of the reasons for the following responses as they apply to Refueling Accidents: (CFR: 41.5 / 45.6)

AK3.04 Non-coincident SCRAM function 3.0 10 295024 High Drywell Pressure / 5 X

Knowledge of the reasons for the following responses as they apply to High Drywell Pressure: (CFR: 41.5 / 45.6)

EK3.05 RPV flooding 3.5 11 295025 (EPE 2) High Reactor Pressure / 3 X

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

41.10 / 43.5 / 45.11) 2.9 12 295026 (EPE 3) Suppression Pool High Water Temperature / 5 X

Knowledge of the operational implications of the following concepts as they apply to Suppression Pool High Water Temperature: (CFR: 41.8 to 41.10)

EK1.02 Steam condensation 3.5 13 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

Knowledge of the interrelations between High Drywell Temperature and the following: (CFR: 41.7 / 45.8)

EK2.02 Components internal to the drywell 3.2 14 295030 (EPE 7) Low Suppression Pool Water Level / 5 X

Ability to operate and/or monitor the following as they apply to Low Suppression Pool Water Level: (CFR: 41.7 / 45.6)

EA1.02 RCIC 3.4 15

ES-401 3

Form ES-401-1 Rev. 11 295031 (EPE 8) Reactor Low Water Level / 2 X

Ability to operate and/or monitor the following as they apply to Reactor Low Water Level: (CFR: 41.7 / 45.6)

EA1.04 High pressure core spray 4.3*

16 295037 (EPE 14) Scram Condition Present and Reactor Power Above APRM Downscale or Unknown / 1 X

2.2.22 Knowledge of limiting conditions for operations and safety limits. (CFR: 41.5 /

43.2 / 45.2) 4.0 17 295038 (EPE 15) High Offsite Radioactivity Release Rate / 9 X

Knowledge of the operational implications of the following concepts as they apply to High Off-Site Release Rate: (CFR: 41.8 to 41.10)

EK1.03 Meteorological effects on off-site release 2.8 18 600000 (APE 24) Plant Fire On Site / 8 X

Ability to determine and interpret the following as they apply to Plant Fire On Site:

AA2.11 Time limit for use of respirators 2.9 19 700000 (APE 25) Generator Voltage and Electric Grid Disturbances / 6 X

Ability to determine and/or interpret the following as they apply to Generator Voltage and Electric Grid Disturbances:

(CFR: 41.5 and 43.5 / 45.5, 45.7, and 45.8)

AA2.06 Generator frequency limitations 3.4 20 K/A Category Totals:

3 3

3 4

4 3

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 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 X

Ability to determine and/or interpret the following as they apply to High Drywell Temperature: (CFR: 41.10 / 43.5 / 45.13)

AA2.02 Drywell pressure 3.9 21 295013 (APE 13) High Suppression Pool Temperature. / 5 295014 (APE 14) Inadvertent Reactivity Addition / 1 295015 (APE 15) Incomplete Scram / 1 X

Knowledge of the interrelations between Incomplete SCRAM and the following:

(CFR: 41.7 / 45.8)

AK2.11 Instrument air 3.5 22 295017 (APE 17) Abnormal Offsite Release Rate / 9 X

2.4.20 Knowledge of the operational implications of EOP warnings, cautions, and notes. (CFR: 41.10 / 43.5 / 45.13) 3.8 23 295020 (APE 20) Inadvertent Containment Isolation / 5 & 7 295022 (APE 22) Loss of Control Rod Drive Pumps / 1 X

Ability to determine and/or interpret the following as they apply to Loss of CRD Pumps: (CFR: 41.10 / 43.5 / 45.13)

AA2.01 Accumulator pressure 3.5 24 295029 (EPE 6) High Suppression Pool Water Level / 5 X

Knowledge of the reasons for the following responses as they apply to High Suppression Pool Water Level: (CFR: 41.5

/ 45.6)

EK3.03 Reactor SCRAM 3.4 25 295032 (EPE 9) High Secondary Containment Area Temperature / 5 X

Knowledge of the operational implications of the following concepts as they apply to High Secondary Containment Area Temperature: (CFR: 41.8 to 41.10)

EK1.04 Impact of operating environment on components 3.1 26 295033 (EPE 10) High Secondary Containment Area Radiation Levels / 9 295034 (EPE 11) Secondary Containment Ventilation High Radiation / 9 295035 (EPE 12) Secondary Containment High Differential Pressure / 5 295036 (EPE 13) Secondary Containment High Sump/Area Water Level / 5 500000 (EPE 16) High Containment Hydrogen Concentration / 5 X

Ability to operate and monitor the following as they apply to High Containment Hydrogen Control: (CFR: 41.7 / 45.6)

EA1.03 Containment atmosphere control system 3.4 27 K/A Category Point Totals:

1 1

1 1

2 1

Group Point Total:

7

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 X

Knowledge of the operational implications of the following concepts as they apply to RHR/LPCI: Injection Mode: (CFR: 41.5 /

45.3)

K5.01 Testable check valve operation 2.7*

28 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)

A2.05 System isolation 3.5 29 206000 (SF2, SF4 HPCIS) High Pressure Coolant Injection 207000 (SF4 IC) Isolation (Emergency) Condenser 209001 (SF2, SF4 LPCS) Low Pressure Core Spray X

Ability to (a) predict the impacts of the following on the Low Pressure Core Spray 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)

A2.08 Valve openings 3.1 30 209002 (SF2, SF4 HPCS) High Pressure Core Spray X

Knowledge of the effect that a loss or malfunction of the High Pressure Core Spray System (HPCS) will have on following: (CFR: 41.7 / 45.4)

K3.03 Adequate core cooling 3.9 31 211000 (SF1 SLCS) Standby Liquid Control X

Knowledge of the operational implications of the following concepts as they apply to Standby Liquid Control System: (CFR:

41.5 / 45.3)

K5.01 Effects of the moderator temperature coefficient of reactivity on the boron 2.7 32 212000 (SF7 RPS) Reactor Protection X Knowledge of the physical connections and/or cause-effect relationships between Reactor Protection System and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8)

K1.01 Nuclear instrumentation 3.7 33 215003 (SF7 IRM) Intermediate Range Monitor X 2.2.12 Knowledge of surveillance procedures. (CFR: 41.10 / 45.13) 3.7 34 215004 (SF7 SRMS) SourceRange Monitor X

Ability to predict and/or monitor changes in parameters associated with operating the Source Range Monitor (SRM) System controls including: (CFR: 41.5 / 45.5)

A1.06 Lights and alarms 3.1 35 215005 (SF7 PRMS) Average Power Range Monitor/Local Power Range Monitor X

Knowledge of the effect that a loss or malfunction of the Average Power Range Monitor/Local Power Range Monitor System will have on following: (CFR: 41.7

/ 45.4)

K3.02 Reactor recirculation system 3.5 36 217000 (SF2, SF4 RCIC) Reactor Core Isolation Cooling X

Knowledge of the physical connections and/or cause-effect relationships between Reactor Core Isolation Cooling System (RCIC) and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8)

K1.02 Nuclear boiler system 3.5 37

ES-401 6

Form ES-401-1 Rev. 11 218000 (SF3 ADS) Automatic Depressurization X

Knowledge of the effect that a loss or malfunction of the following will have on the Automatic Depressurization System:

(CFR: 41.7 / 45.7)

K6.04 Air supply to ADS valves 3.6 38 223002 (SF5 PCIS) Primary Containment Isolation/Nuclear Steam Supply Shutoff X

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

A4.01 Valve closures 3.6 39 239002 (SF3 SRV) Safety Relief Valves X

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

A4.08 Plant air system pressure 3.2 40 259002 (SF2 RWLCS) Reactor Water Level Control 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 41 261000 (SF9 SGTS) Standby Gas Treatment X

Ability to monitor automatic operations of the Standby Gas Treatment System including: (CFR: 41.7 / 45.7)

A3.03 Valve operation 3.0 42 262001 (SF6 AC) AC Electrical Distribution X

Knowledge of the effect that a loss or malfunction of the following will have on the A.C. Electrical Distribution: (CFR: 41.7

/ 45.7)

K6.01 D.C. power 3.1 43 262002 (SF6 UPS) Uninterruptable Power Supply (AC/DC)

X Knowledge of Uninterruptable Power Supply (A.C./D.C.) design feature(s) and/or interlocks which provide for the following: (CFR: 41.7)

K4.01 Transfer from preferred power to alternate power supplies 3.1 44 263000 (SF6 DC) DC Electrical Distribution X

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

K2.01 Major D.C. loads 3.1 45 264000 (SF6 EGE) Emergency Generators (Diesel/Jet) EDG X

Ability to monitor automatic operations of the Emergency Generators (Diesel/Jet) including: (CFR: 41.7 / 45.7)

A3.02 Minimum time for load pick up 3.1 46 300000 (SF8 IA) Instrument Air X

Knowledge of the effect that a loss or malfunction of the following will have on the Instrument Air System: (CFR: 41.7 /

45.7)

K6.07 Valves 2.5 47 400000 (SF8 CCS) Component Cooling Water X

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

K2.01 CCW pumps 2.9 48 510000 (SF4 SWS*) Service Water (Normal and Emergency) 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)

A1.05 Torus/suppression pool water level 3.5 49 215003 (SF7 IRM) Intermediate Range Monitor X

Ability to (a) predict the impacts of the following on the Intermediate Range Monitor (IRM) 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)

A2.01 Power supply degraded 2.8 50 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 51

ES-401 7

Form ES-401-1 Rev. 11 262001 (SF6 AC) AC Electrical Distribution X

Knowledge of A.C. Electrical Distribution design feature(s) and/or interlocks which provide for the following: (CFR: 41.7)

K4.06 Redundant power sources to vital buses 3.6 52 264000 (SF6 EGE) Emergency Generators (Diesel/Jet) EDG X

Ability to predict and/or monitor changes in parameters associated with operating the Emergency Generators (Diesel/Jet) controls including: (CFR: 41.5 / 45.5)

A1.04 Crank case temperature and pressure 2.6 53 K/A Category Point Totals:

2 2

2 2

3 3

3 3

2 2

2 Group Point Total:

26

ES-401 8

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 X

Knowledge of Reactor Manual Control System design feature(s) and/or interlocks which provide for the following: (CFR: 41.7)

K4.02 Control rod blocks 3.5 54 201003 (SF1 CRDM) Control Rod and Drive Mechanism 201004 (SF7 RSCS) Rod Sequence Control X

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

41.7 / 45.5 to 45.8)

A4.01 System bypass switches 3.4 55 201005 (SF1, SF7 RCIS) Rod Control and Information 201006 (SF7 RWMS) Rod Worth Minimizer 202001 (SF1, SF4 RS) Recirculation X

2.2.42 Ability to recognize system parameters that are entry-level conditions for Technical Specifications. (CFR: 41.7 / 41.10 /

43.2 / 43.3 / 45.3) 3.9 56 202002 (SF1 RSCTL) Recirculation Flow Control 204000 (SF2 RWCU) Reactor Water Cleanup 214000 (SF7 RPIS) Rod Position Information 215001 (SF7 TIP) Traversing InCore Probe X

Knowledge of TRAVERSING IN-CORE PROBE design feature(s) and/or interlocks which provide for the following: (CFR: 41.7)

K4.01 Primary containment isolation 3.4 57 215002 (SF7 RBMS) Rod Block Monitor 216000 (SF7 NBI) Nuclear Boiler Instrumentation 219000 (SF5 RHR SPC) RHR/LPCI:

Torus/Suppression Pool Cooling Mode X

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

Torus/Suppression Pool Cooling Mode and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8)

K1.04 LPCI/RHR pumps 3.9 58 223001 (SF5 PCS) Primary Containment and Auxiliaries 226001 (SF5 RHR CSS) RHR/LPCI:

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

Torus/Suppression Pool Spray Mode 233000 (SF9 FPCCU) Fuel Pool Cooling/Cleanup X

Knowledge of the effect that a loss or malfunction of the Fuel Pool Cooling and Clean-Up will have on following: (CFR: 41.7 /45.6)

K3.06 Area radiation levels 2.9 59 234000 (SF8 FH) FuelHandling Equipment 239001 (SF3, SF4 MRSS) Main and Reheat Steam 239003 (SF9 MSVLCS) Main Steam Isolation Valve Leakage Control

ES-401 9

Form ES-401-1 Rev. 11 241000 (SF3 RTPRS) Reactor/Turbine Pressure Regulating X

Knowledge of the operational Implications of the following concepts as they apply to Reactor/Turbine Pressure Regulating System: (CFR: 41.5 /

45.3)

K5.04 Turbine inlet pressure vs.

reactor pressure 3.3 60 245000 (SF4 MTGEN) Main Turbine Generator/Auxiliary X

Ability to predict and/or monitor changes in parameters associated with operating the Main Turbine Generator and Auxiliary Systems controls including: (CFR: 41.5 / 45.5)

A1.05 Reactor pressure 3.5 61 256000 (SF2 CDS) Condensate X

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

K2.01 System pumps 2.7*

62 259001 (SF2 FWS) Feedwater 268000 (SF9 RW) Radwaste X

Ability to (a) predict the impacts of the following on the Radwaste; 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)

A2.01 System rupture 2.9 63 271000 (SF9 OG) Offgas 272000 (SF7, SF9 RMS) Radiation Monitoring X

Ability to monitor automatic operations of the Radiation Monitoring System including: (CFR:

41.7 / 45.7)

A3.03 Liquid radwaste isolation indications 3.1 64 286000 (SF8 FPS) Fire Protection 288000 (SF9 PVS) Plant Ventilation 290001 (SF5 SC) Secondary Containment X

Knowledge of the effect that a loss or malfunction of the following will have on the Secondary Containment: (CFR: 41.7 / 45.7)

K6.05 Auxiliary building ventilation 2.9 65 290003 (SF9 CRV) Control Room Ventilation 290002 (SF4 RVI) Reactor Vessel Internals 51001 (SF8 CWS*) Circulating Water K/A Category Point Totals:

1 1

1 2

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: Columbia Generating Station Date of Exam:

Category K/A #

Topic RO SRO-only IR IR

1. Conduct of Operations 2.1.20 Ability to interpret and execute procedure steps. (CFR:

41.10 / 43.5 / 45.12) 4.6 66 2.1.36 Knowledge of procedures and limitations involved in core alterations. (CFR: 41.10 / 43.6 / 45.7) 3.0 67 2.1.37 Knowledge of procedures, guidelines, or limitations associated with reactivity management. (CFR: 41.1 /

43.6 / 45.6) 4.3 68 Subtotal 3

2. Equipment Control 2.2.2 Ability to manipulate the console controls as required to operate the facility between shutdown and designated power levels. (CFR: 41.6 / 41.7 / 45.2) 4.6 69 2.2.21 Knowledge of pre-and post-maintenance operability requirements. (CFR: 41.10 / 43.2) 2.9 70 2.2.43 2.2.43 nowledge of the process used to track inoperable alarms. (CFR: 41.10 / 43.5 / 45.13) 3.0 71 Subtotal 3

3. Radiation Control 2.3.5 Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

(CFR: 41.11 / 41.12 / 43.4 / 45.9) 2.9 72 2.3.11 Ability to control radiation releases. (CFR: 41.11 / 43.4 /

45.10) 3.8 73 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 74 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 75 Subtotal 2

Tier 3 Point Total 10

ES-401 BWR Examination Outline Form ES-401-1 Rev. 11 Facility: Columbia Generating Station Date of Exam:

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 4

3 7

2 7

1 2

3 Tier Totals 27 5

5 10

2.

Plant Systems 1

26 3

2 5

2 12 1

1 1

3 Tier Totals 38 5

3 8

3. Generic Knowledge and Abilities Categories 1

2 3

4 10 1

2 3

4 7

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 X

Ability to determine and/or interpret the following as they apply to Partial Or Complete Loss of A.C. Power: (CFR: 41.10

/ 43.5 / 45.13)

AA2.02 Reactor power / pressure / and level 4.3*

76 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)

AA2.02 Extent of partial or complete loss of D.C. power 3.9 77 295005 (APE 5) Main Turbine Generator Trip /

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

/ 7 X

Ability to determine and/or interpret the following as they apply to Control Room Abandonment: (CFR: 41.10 / 43.5 / 45.13)

AA2.04 Suppression pool temperature 4.1 78 295018 (APE 18) Partial or Complete Loss of CCW / 8 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 X

2.4.45 Ability to prioritize and interpret the significance of each annunciator or alarm.

(CFR: 41.10 / 43.5 / 45.3 / 45.12) 4.3 79 295026 (EPE 3) Suppression Pool High Water Temperature / 5 X

Ability to determine and/or interpret the following as they apply to Suppression Pool High Water Temperature: (CFR:

41.10 / 43.5 / 45.13)

EA2.03 Reactor pressure 4.0 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

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

/ 45.12) 4.0 81 295030 (EPE 7) Low Suppression Pool Water Level / 5 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 82 295031 (EPE 8) Reactor Low Water Level / 2 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

ES-401 3

Form ES-401-1 Rev. 11 700000 (APE 25) Generator Voltage and Electric Grid Disturbances / 6 K/A Category Totals:

0 0

0 0

4 3

Group Point Total:

7

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

2.2.37 Ability to determine operability and/or availability of safety related equipment. (CFR: 41.7 / 43.5 / 45.12) 4.6 83 295014 (APE 14) Inadvertent Reactivity Addition / 1 X

Ability to determine and/or interpret the following as they apply to Inadvertent Reactivity Addition: (CFR: 41.10 / 43.5 /

45.13)

AA2.01 Reactor power 4.2*

84 295015 (APE 15) Incomplete Scram / 1 295017 (APE 17) Abnormal Offsite Release Rate / 9 295020 (APE 20) Inadvertent Containment Isolation / 5 & 7 295022 (APE 22) Loss of Control Rod Drive Pumps / 1 295029 (EPE 6) High Suppression Pool Water Level / 5 295032 (EPE 9) High Secondary Containment Area Temperature / 5 295033 (EPE 10) High Secondary Containment Area Radiation Levels / 9 295034 (EPE 11) Secondary Containment Ventilation High Radiation / 9 295035 (EPE 12) Secondary Containment High Differential Pressure / 5 295036 (EPE 13) Secondary Containment High Sump/Area Water Level / 5 X

2.2.25 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

(CFR: 41.5 / 41.7 / 43.2) 4.2 85 500000 (EPE 16) High Containment Hydrogen Concentration / 5 K/A Category Point Totals:

0 0

0 0

1 2

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 X

2.4.8 Knowledge of how abnormal operating procedures are used in conjunction with EOPs. (CFR: 41.10 /

43.5 / 45.13) 4.5 86 205000 (SF4 SCS) Shutdown Cooling 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 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)

A2.10 Reactor/turbine pressure control system low pressure 3.8 87 215003 (SF7 IRM)

Intermediate-Range Monitor 215004 (SF7 SRMS) Source-Range Monitor 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 X

Ability to (a) predict the impacts of the following on the Relief/Safety Valves; 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)

A2.01 Stuck open vacuum breakers 3.3 88 259002 (SF2 RWLCS) Reactor Water Level Control X

Ability to (a) predict the impacts of the following on the Reactor Water Level Control 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)

A2.08 Receipt of an ECCS initiation signal: FWCI 4.5*

89 261000 (SF9 SGTS) Standby Gas Treatment

ES-401 6

Form ES-401-1 Rev. 11 262001 (SF6 AC) AC Electrical Distribution 262002 (SF6 UPS) Uninterruptable Power Supply (AC/DC) 263000 (SF6 DC) DC Electrical Distribution 264000 (SF6 EGE) Emergency Generators (Diesel/Jet) EDG 300000 (SF8 IA) Instrument Air X

2.4.1 Knowledge of EOP entry conditions and immediate action steps. (CFR:

41.10 / 43.5 / 45.13) 4.8 90 400000 (SF8 CCS) Component Cooling Water 510000 (SF4 SWS*) Service Water (Normal and Emergency)

K/A Category Point Totals:

0 0

0 0

0 0

0 3

0 0

2 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 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 216000 (SF7 NBI) Nuclear Boiler Instrumentation 219000 (SF5 RHR SPC) RHR/LPCI:

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

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

Torus/Suppression Pool Spray Mode 233000 (SF9 FPCCU) Fuel Pool Cooling/Cleanup 234000 (SF8 FH) Fuel-Handling Equipment X

Ability to monitor automatic operations of the Fuel Handling Equipment including: (CFR: 41.7 /

45.7)

A3.02 Interlock operation 3.7 91 239001 (SF3, SF4 MRSS) Main and Reheat Steam 239003 (SF9 MSVLCS) Main Steam Isolation Valve Leakage Control 241000 (SF3 RTPRS) Reactor/Turbine Pressure Regulating X

Ability to (a) predict the impacts of the following on the Reactor/Turbine Pressure Regulating System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

(CFR: 41.5 / 45.6)

A2.14 Loss of main turbine PMG 2.7 92 245000 (SF4 MTGEN) Main Turbine Generator/Auxiliary 256000 (SF2 CDS) Condensate 259001 (SF2 FWS) Feedwater 268000 (SF9 RW) Radwaste 271000 (SF9 OG) Offgas

ES-401 8

Form ES-401-1 Rev. 11 272000 (SF7, SF9 RMS) Radiation Monitoring 286000 (SF8 FPS) Fire Protection 288000 (SF9 PVS) Plant Ventilation 290001 (SF5 SC) Secondary Containment 290003 (SF9 CRV) Control Room Ventilation 290002 (SF4 RVI) Reactor Vessel Internals X

2.2.38 Knowledge of conditions and limitations in the facility license.

(CFR: 41.7 / 41.10 / 43.1 / 45.13) 4.5 93 51001 (SF8 CWS*) Circulating Water K/A Category Point Totals:

0 0

0 0

0 0

0 1

0 1

1 Group Point Total:

3

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

Form ES-401-3 Rev. 11 Facility:

Date of Exam:

Category K/A #

Topic RO SRO-only IR IR

1. Conduct of Operations 2.1.6 Ability to manage the control room crew during plant transients. (CFR: 41.10 / 43.5 / 45.12 / 45.13) 4.8 94 2.1.25 Ability to interpret reference materials, such as graphs, curves, tables, etc. (CFR: 41.10 / 43.5 / 45.12) 4.2 95 Subtotal 2

2. Equipment Control 2.2.15 Ability to determine the expected plant configuration using design and configuration control documentation, such as drawings, line-ups, tag-outs, etc. (CFR: 41.10 /

43.3 / 45.13) 4.3 96 2.2.36 Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations. (CFR: 41.10 / 43.2 /

45.13) 4.2 97 Subtotal 2

3. Radiation Control 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) 3.1 98 Subtotal 1

4. Emergency Procedures/Plan 2.4.22 Knowledge of the bases for prioritizing safety functions during abnormal/emergency operations. (CFR: 41.7 /

41.10 / 43.5 / 45.12) 4.4 99 2.4.23 Knowledge of the bases for prioritizing emergency procedure implementation during emergency operations. (CFR: 41.10 / 43.5 / 45.13) 4.4 100 Subtotal 2

Tier 3 Point Total 10 7

ES-401 Record of Rejected K/As (Rev 1 - 9/19/18)

Form ES-401-4 Page 1 of 5 RO Examination Tier /

Group Randomly Selected K/A Reason for Rejection 1/1 295006.AK3.05 (RO Question 5)

Original K/A: Knowledge of the reasons for the following responses as they apply to SCRAM: Direct turbine generator trip. (CFR: 41.5 /

45.6) IR: 3.8 Reason for Rejection: There is no direct turbine generator trip on a scram at CGS.

Recommended replacement K/A: 295006.AK3.06 - Knowledge of the reasons for the following responses as they apply to SCRAM:

Recirculation pump speed reduction. (CFR: 41.5 / 45.6) IR: 3.2 Replacement K/A randomly selected per guidance in NUREG 1021.

1/1 295019.AK2.19 (RO Question 8)

Original K/A: Knowledge of the interrelations between Partial or Complete Loss of Instrument Air and the following: RHR/LPCI.

(CFR: 41.7 / 45.8) IR: 2.7 Reason for Rejection: No interrelations between instrument air and RHR/LPCI operation.

Recommended replacement K/A: 295019.AK2.05 - Knowledge of the interrelations between Partial or Complete Loss of Instrument Air and the following: Main steam system. (CFR: 41.7 / 45.8) IR: 3.4 Replacement K/A randomly selected per guidance in NUREG 1021.

1/1 295024.EK3.05 (RO Question 11)

Original K/A: Knowledge of the reasons for the following responses as they apply to High Drywell Pressure: RPV flooding (CFR: 41.5 /

45.6) IR: 3.5 RPV flooding is not a response to High Drywell Pressure at CGS.

Recommended replacement K/A: 295024.EK3.08 - Knowledge of the reasons for the following responses as they apply to High Drywell Pressure: Containment spray. (CFR: 41.5 / 45.6) IR: 3.7 Replacement K/A randomly selected per guidance in NUREG 1021.

1/1 295025.2.4.41 (RO Question 12)

Original K/A: Knowledge of the emergency action level thresholds and classifications. (CFR: 41.10/43.5/45.11) IR: 2.9 Reason for Rejection: K/A requires knowledge of EAL thresholds and classifications. This is not an RO function at CGS.

Recommended replacement K/A: 295025.2.4.2 - Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions. (CFR: 41.7 / 45.7 / 45.8) IR: 4.5 Replacement K/A randomly selected per guidance in NUREG 1021.

ES-401 Record of Rejected K/As (Rev 1 - 9/19/18)

Form ES-401-4 Page 2 of 5 1/1 295038.EK1.03 (RO Question 18)

Original K/A: Knowledge of the operational implications of the following concepts as they apply to High Off-Site Release Rate:

Meteorological effects on off-site release. (CFR: 41.8 to 41.10) IR:

2.8 Reason for Rejection: Meteorological effects of offsite release not within the RO area of responsibility.

Recommended replacement K/A: 295038.EK1.02 - Knowledge of the operational implications of the following concepts as they apply to High Off-Site Release Rate: Protection of the general public.

(CFR: 41.8 to 41.10) IR: 4.2 Replacement K/A randomly selected per guidance in NUREG 1021.

1/2 295017.2.4.20 (RO Question 23)

Original K/A: Knowledge of the operational implications of EOP warnings, cautions, and notes. (CFR: 41.10/43.5/45.13) IR: 3.8 Reason for Rejection: There are no warnings, cautions or notes on PPM 5.4.1, Radioactivity Release Control.

Recommended Replacement K/A: 295017.2.4.18 - Knowledge of the specific bases for EOPs. (CFR: 41.10/43.1/45.13) IR: 3.3 Replacement K/A randomly selected per guidance in NUREG 1021.

1/2 500000.EA1.03 (RO Question 27)

Original K/A: Ability to operate and monitor the following as they apply to High Containment Hydrogen Control: Containment atmosphere control system (CFR: 41.7 / 45.6) IR: 3.4 Reason for Rejection: The Containment Atmosphere Control system is Retired in Place at CGS.

Recommended Replacement K/A: 500000.EA1.04 - Ability to operate and monitor the following as they apply to High Containment Hydrogen Control: Drywell recirculating fans. (CFR:

41.7 / 45.6) IR: 2.9 Replacement K/A randomly selected per guidance in NUREG 1021.

2/1 203000.K5.01 (RO Question 28)

Original K/A: Knowledge of the operational implications of the following concepts as they apply to RHR/LPCI: Injection Mode:

Testable check valve operation: (CFR: 41.5 / 45.3) IR: 2.7 Reason for Rejection: RHR/LPCI testable check valves are Retired in Place at CGS.

Recommended Replacement K/A: 203000.K5.02 - Knowledge of the operational implications of the following concepts as they apply to RHR/LPCI: Injection Mode: Core Cooling Methods: (CFR: 41.5 /

45.3) IR: 3.5 Replacement K/A randomly selected per guidance in NUREG 1021.

ES-401 Record of Rejected K/As (Rev 1 - 9/19/18)

Form ES-401-4 Page 3 of 5 2/1 215005.K3.02 (RO Question 36)

Original K/A: Knowledge of the effect that a loss or malfunction of the Average Power Range Monitor/Local Power Range Monitor System will have on following: Reactor recirculation system: (CFR:

41.7 / 45.4) IR: 3.5 Reason for Rejection: A failed APRM does not affect RRC system operation.

Recommended Replacement K/A: 215005.K3.01 - Knowledge of the effect that a loss or malfunction of the Average Power Range Monitor/Local Power Range Monitor System will have on following:

RPS (CFR: 41.7 / 45.4) IR: 4.0 Replacement K/A randomly selected per guidance in NUREG 1021.

2/1 239002.A4.08 (RO Question 40)

Original K/A: Ability to manually operate and/or monitor in the control room: Plant air system pressure: (CFR: 41.7 / 45.5 to 45.8)

IR: 3.2 Reason for Rejection: Original K/A samples knowledge that is similar to required knowledge for Question RO-38 K/A.

Recommended Replacement K/A: 239002.A4.06 - Ability to manually operate and/or monitor in the control room: Reactor water level (CFR: 41.7 / 45.5 to 45.8) IR: 3.9 Replacement K/A randomly selected per guidance in NUREG 1021.

2/2 201004.A4.01 (RO Question 55)

Original K/A: Ability to manually operate and/or monitor in the control room: System bypass switches: (CFR: 41.7 / 45.5 to 45.8)

IR: 3.4 Reason for Rejection: Rod Sequence Control System (RSCS) no longer in operation at CGS.

Recommended Replacement K/A: 201006.A4.06 - Ability to manually operate and/or monitor in the control room: Selected rod position indication (CFR: 41.7 / 45.5 to 45.8) IR: 3.2 New system and K/A randomly selected in accordance with NUREG 1021 and supplemental guidance provided from chief examiner.

ES-401 Record of Rejected K/As (Rev 1 - 9/19/18)

Form ES-401-4 Page 4 of 5 2/2 268000.A2.01 (RO Question 63)

Original K/A: Ability to (a) predict the impacts of the following on the RADWASTE ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: System rupture. (CFR: 41.5 / 45.6) IR: 2.9 Reason for Rejection: Radwaste system operation is performed by NLOs outside the control room. There are no procedures that licensed operators would use in the control room to combat a radwaste system rupture.

Recommended Replacement K/A: 268000.A4.01 - Ability to manually operate and/or monitor in the control room: Sump Integrators (CFR: 41.7 / 45.5 to 45.8) IR: 3.4 All other K/As in the original K/A group have importance ratings <

2.5. New K/A group and specific K/A randomly selected in accordance with NUREG 1021 and supplemental guidance provided from chief examiner.

SRO Examination Tier /

Group Randomly Selected K/A Reason for Rejection 1/2 295036.2.2.25 (SRO Question 85)

Original K/A: Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits. (CFR: 41.5 /

41.7 / 43.2) IR: 4.2 Reason for Rejection: There are no Technical Specifications /

Licensee Controlled Specifications covering Sump / Area Water Levels Recommended Replacement K/A: 2.2.44 - Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions. (CFR: 41.5 / 43.5 / 45.12) IR: 4.4 New K/A randomly selected in accordance with NUREG 1021 and supplemental guidance provided from chief examiner.

ES-401 Record of Rejected K/As (Rev 1 - 9/19/18)

Form ES-401-4 Page 5 of 5 Tier /

Group Randomly Selected K/A Reason for Rejection 2/1 259002.A2.08 (SRO Question 89)

Original K/A: Ability to (a) predict the impacts of the following on the Reactor Water Level Control System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Receipt of an ECCS initiation signal: FWCI (CFR: 41.5 / 45.6) IR: 4.5*

Reason for Rejection: CGS does not have FWCI.

Recommended Replacement K/A: 259002.A2.03 - Ability to (a) predict the impacts of the following on the Reactor Water Level Control System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of reactor level input.

(CFR: 41.5 / 45.6) IR: 3.7 Replacement K/A randomly selected per guidance in NUREG 1021.

2/2 241000.A2.14 (SRO Question 92)

Original K/A: Ability to (a) predict the impacts of the following on the Reactor/Turbine Pressure Regulating System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of main turbine PMG (CFR: 41.5 / 45.6) IR: 2.7 Reason for Rejection: Unable to write an SRO-level question.

Recommended Replacement K/A: 241000.A2.20 - Ability to (a) predict the impacts of the following on the Reactor/Turbine Pressure Regulating System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Low reactor/turbine pressure regulating system oil level (CFR: 41.5 / 45.6) IR: 2.6 Replacement K/A randomly selected per guidance in NUREG 1021.

3/4 2.4.23 (SRO Question 100)

Original K/A: Knowledge of the bases for prioritizing emergency procedure implementation during emergency operations. (CFR:

41.10 / 43.5 / 45.13) IR: 4.4 Reason for Rejection: Since BWRs use symptom-based EOPs, there is no process for prioritizing procedure implementation; it is based on symptoms/plant conditions.

Recommended Replacement K/A: 2.4.9 - Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.

(CFR: 41.10 / 43.5 / 45.13) IR: 4.2 Replacement K/A randomly selected per guidance in NUREG 1021.

ES-301 Administrative Topics Outline (Rev 0 - 9/11/18)

Form ES-301-1 Facility:

Columbia Generating Station Date of Examination:

2/25/19 Examination Level: RO SRO Operating Test Number:

1 Administrative Topic (see Note)

Type Code*

Describe activity to be performed A-1 Conduct of Operations K/A: 2.1.20 (4.6 / 4.6)

(M)(P)(R)

ALTERNATE DETERMINATION OF DRYWELL IDENTIFIED LEAK RATE

==

Description:==

Given plant data, determine the Calculated Identified Drywell Leak Rate per SOP-EDR-OPS (Equipment Drain System Operation).

A-2 Conduct of Operations K/A: 2.1.25 (3.9 / 4.2)

(D)(R)

RRC-P-1B DELTA-T CAVITATION ALARM VERIFICATION (Time Critical)

==

Description:==

Validate in-coming cavitation alarm (using associated annunciator response procedure) by calculating RRC-P-1B Delta-T between pump suction temperature and temperature associated with RPV steam dome pressure (as read from graph). Determine that a power reduction is necessary.

A-3 Equipment Control K/A: 2.2.41 (3.5 / 3.9)

(D)(R)

EXPLAIN RHR-P-2C FAILURE TO START USING ELECTRICAL WIRING DIAGRAM (EWD)

==

Description:==

For plant conditions given, explain using EWD why RHR-P-2C did not start when given a manual initiation signal.

A-4 Radiation Control K/A: 2.3.4 (3.2 / 3.7)

(M)(R)

DETERMINATION OF STAY TIME IN A HIGH RADIATION AREA

==

Description:==

Determine personal maximum stay time for job performed in a high radiation area such that it will not exceed the allowable Annual Administrative Dose Hold Point value.

OE: SER 32-86 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 (4)

(D)irect from bank ( 3 for ROs) (2)

(N)ew or (M)odified from bank ( 1) (2)

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

ES-301 Administrative Topics Outline (Rev 0 - 9/11/18)

Form ES-301-1 Facility:

Columbia Generating Station Date of Examination:

2/25/19 Examination Level: RO SRO Operating Test Number:

1 Administrative Topic (see Note)

Type Code*

Describe activity to be performed A-5 Conduct of Operations K/A: 2.1.5 (2.9* / 3.9)

(D)(R)

VERIFY ON-COMING CREW QUALIFICATIONS

==

Description:==

Given information for the on-coming crew, verify their watch qualifications per station procedure. Expound on any qualification issues found.

A-6 Conduct of Operations K/A: 2.1.25 (3.9 / 4.2)

(D)(P)(R)

DETERMINE IF VOLUNTARY ENTRY INTO AIA IS ALLOWABLE

==

Description:==

Given plant conditions, determine if voluntary entry into the Area of Increased Awareness (AIA) is allowed per PPM 3.2.1 (Normal Plant Shutdown).

A-7 Equipment Control K/A: 2.2.40 (3.4 / 4.7)

(D)(R)

DETERMINE IF REACTOR MODE CHANGE IS ALLOWED

==

Description:==

Given plant conditions, determine if a reactor mode change from Mode 4 to Mode 2 is allowed. Provide justification.

A-8 Radiation Control K/A: 2.3.6 (2.0 / 3.8)

(M)(R)

APPROVAL OF CW BLOWDOWN

==

Description:==

Evaluate a request from Chemistry to commence Circulating Water (CW) Blowdown to the Columbia River while ensuring adherence to the National Pollutant Discharge Elimination System (NPDES) Permit.

A-9 Emergency Plan K/A: 2.4.41 (2.9 / 4.6)

(D)(R)

COMPLETE CNF FOR GENERAL EMERGENCY

==

Description:==

Complete Classification Notification Form (CNF) for a General Emergency based upon completed Dose Assessment. (Time Critical)

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 (5)

(D)irect from bank ( 4 for SROs) (4)

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

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

ES-301 Control Room/In-Plant Systems Outline (Rev 0 - 9/11/18)

Form ES-301-2 Facility:

Columbia Generating Station Date of Examination:

2/25/19 Exam Level: RO SRO-I SRO-U Operating Test Number:

1 Control Room Systems:* 8 for RO System/JPM Title Type Code*

Safety Function S-1: START B CRD PUMP - CONTROLLER FAILS LOW UPON SHIFTING TO AUTO (A)(N)(S) 1

==

Description:==

When Control Rod Drive (CRD) Pump B is started (following trip of the A pump), the CRD flow controller fails low when shifted to Auto. Operator action required to manually raise output signal to re-establish CRD flow.

K/A: 201001 A4.03 (2.9 / 2.8)

S-2: TRANSFER SM-1 FROM TR-N TO TR-S & TRANSFER SM-7 FROM SM-1 TO TR-B (A)(N)(S) 6

==

Description:==

4160 VAC Bus SM-1 is manually transferred from Normal Transformer (TR-N1) to the Startup Transformer (TR-S).

During the transfer, the TR-N1 feed breaker to SM-1 fails to automatically trip and must be tripped manually. In addition, 4160 VAC Bus SM-7 (Division 1 ESF Bus) is manually transferred to the Backup Transformer (TR-B).

K/A: 262001 A4.04 (3.6 / 3.7)

S-3: EQUALIZE AND OPEN OUTBOARD MSIVs (D)(L)(S) 3

==

Description:==

Outboard Main Steam Isolation Valves (MSIVs) are opened after validating acceptable MSIV differential pressure and resetting the MSIV isolation logic.

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

S-4: PERFORM REACTOR FEED PUMP A RESTART (D)(L)(S) 2

==

Description:==

Following trip of both Reactor Feed Pumps on high RPV water level, a restart of Reactor Feed Pump A is performed using SOP-RFT-RESTART-QC.

K/A: 259001 A4.02 (3.9 / 3.7)

ES-301 Control Room/In-Plant Systems Outline (Rev 0 - 9/11/18)

Form ES-301-2 S-5: MANUALLY INITIATE CONTAINMENT ISOLATION FOR TIP CHANNEL #5 - MANUAL SQUIB FIRING REQUIRED (A)(D)(L)(S) 7

==

Description:==

Upon discovering that a Traverse In-Core Probe (TIP) valve (TIP-V-5) did not isolate upon a containment isolation signal, action is taken to manually close TIP-V-5 to include manually firing associated squib valve.

K/A: 215001 A2.07 (3.4 / 3.7)

S-6: ALIGN SERVICE WATER TO FUEL POOL HEAT EXCHANGERS (D)(L)(P)(S) 9

==

Description:==

Following a total loss of normal cooling, action is taken to align Plant Service Water (TSW) to both Fuel Pool heat exchangers per SOP-FPC-OPS.

K/A: 233000 A2.08 (2.9 / 3.1)

S-7: START SGT SUBSYSTEM A - FAILURE REQUIRES COMPONENT LEVEL START (A)(EN)(N)(S) 5

==

Description:==

Upon failing to start Standby Gas Treatment Train A using Quick Card, a manual start of SGT Train A will be required at the component level using either the Train A Lead fan or Lag fan per SOP-SGT-START.

K/A: 290001 A4.01 (3.3 / 3.4)

S-8: TRANSFER SDC FROM RHR LOOP B TO RHR LOOP A -

SW-P-1A FAILS TO AUTO START AND SW-V-2A FAILS TO AUTO OPEN (A)(L)(N)(S) 4

==

Description:==

With Residual Heat Removal (RHR) Loop B already secured from shutdown cooling (SDC), shift the SDC lineup from the B SDC loop to the A SDC loop per SOP-RHR-SDC. Service Water Pump 1A (SW-P-1A) fails to automatically start. When manually started, Service Water pump discharge valve (SW-V-2A) fails to automatically open requiring manual operation.

K/A: 205000 A4.09 (3.1 / 3.1)

ES-301 Control Room/In-Plant Systems Outline (Rev 0 - 9/11/18)

Form ES-301-2 In-Plant Systems:* 3 for RO P-1: PERFORM ATTACHMENT 7.4 OF ABN-CR-EVAC ON A CONTROL ROOM EVACUATION (TIME CRITICAL)

(D)(E)(L)(R) 1

==

Description:==

Following a Control Room evacuation due to a fire, perform ABN-CR-EVAC, Attachment 7.4, which includes pulling fuses to ensure 4160 VAC Bus SM-8 (Division 2 ESF Bus) is tripped and opening Reactor Protection System (RPS) breakers to ensure a complete scram occurs. Time Critical - 10 minutes.

295016 AA1.01 (3.8 / 3.9)

P-2: PERFORM WETWELL VENTING LOCALLY USING HARDENED CONTAINMENT VENT (HCV) SYSTEM (E)(L)(N)(R) 5

==

Description:==

Using PPM 5.5.14 (Emergency Wetwell Venting), vent the Wetwell from the HCV Remote Operating Station (ROS) through local manipulation of HCV valves.

K/A: 223001 A2.07 (4.2* / 4.3*)

P-3: BYPASS REACTOR CORE ISOLATION COOLING (RCIC)

TRIPS AND ISOLATIONS (C)(D)(E)(L)

(P)(R) 2

==

Description:==

During Station Blackout (SBO), bypass the RCIC high area temperature isolation from the Control Room and then proceed to the field to defeat the RCIC 25 psig high exhaust pressure trip by isolating local RCIC pressure switches.

295003 AA1.03 (4.4* / 4.4*)

All RO control room (and in-plant) systems must be different and serve different safety functions. In-plant systems and functions may overlap those tested in the control room.

• Type Codes Criteria for RO /SRO-I/SRO-U (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power/Shutdown (N)ew or (M)odified from bank including 1(A)

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

ES-301 Control Room/In-Plant Systems Outline (Rev 0 - 9/11/18)

Form ES-301-2 Facility:

Columbia Generating Station Date of Examination:

2/25/19 Exam Level: RO SRO-I SRO-U Operating Test Number:

1 Control Room Systems:* 7 for SRO-I System/JPM Title Type Code*

Safety Function S-1: START B CRD PUMP - CONTROLLER FAILS LOW UPON SHIFTING TO AUTO (A)(N)(S) 1

==

Description:==

When Control Rod Drive (CRD) Pump B is started (following trip of the A pump), the CRD flow controller fails low when shifted to Auto. Operator action required to manually raise output signal to re-establish CRD flow.

K/A: 201001 A4.03 (2.9 / 2.8)

S-2: TRANSFER SM-1 FROM TR-N TO TR-S & TRANSFER SM-7 FROM SM-1 TO TR-B (A)(N)(S) 6

==

Description:==

4160 VAC Bus SM-1 is manually transferred from Normal Transformer (TR-N1) to the Startup Transformer (TR-S).

During the transfer, the TR-N1 feed breaker to SM-1 fails to automatically trip and must be tripped manually. In addition, 4160 VAC Bus SM-7 (Division 1 ESF Bus) is manually transferred to the Backup Transformer (TR-B).

K/A: 262001 A4.04 (3.6 / 3.7)

S-3: EQUALIZE AND OPEN OUTBOARD MSIVs (D)(L)(S) 3

==

Description:==

Outboard Main Steam Isolation Valves (MSIVs) are opened after validating acceptable MSIV differential pressure and resetting the MSIV isolation logic.

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

S-5: MANUALLY INITIATE CONTAINMENT ISOLATION FOR TIP CHANNEL #5 - MANUAL SQUIB FIRING REQUIRED (A)(D)(L)(S) 7

==

Description:==

Upon discovering that a Traverse In-Core Probe (TIP) valve (TIP-V-5) did not isolate upon a containment isolation signal, action is taken to manually close TIP-V-5 to include manually firing associated squib valve.

K/A: 215001 A2.07 (3.4 / 3.7)

ES-301 Control Room/In-Plant Systems Outline (Rev 0 - 9/11/18)

Form ES-301-2 S-6: ALIGN SERVICE WATER TO FUEL POOL HEAT EXCHANGERS (D)(L)(P)(S) 9

==

Description:==

Following a total loss of normal cooling, action is taken to align Plant Service Water (TSW) to both Fuel Pool heat exchangers per SOP-FPC-OPS.

K/A: 233000 A2.08 (2.9 / 3.1)

S-7: START SGT SUBSYSTEM A - FAILURE REQUIRES COMPONENT LEVEL START (A)(EN)(N)(S) 5

==

Description:==

Upon failing to start Standby Gas Treatment Train A using Quick Card, a manual start of SGT Train A will be required at the component level using either the Train A Lead fan or Lag fan per SOP-SGT-START.

K/A: 290001 A4.01 (3.3 / 3.4)

S-8: TRANSFER SDC FROM RHR LOOP B TO RHR LOOP A -

SW-P-1A FAILS TO AUTO START AND SW-V-2A FAILS TO AUTO OPEN (A)(L)(N)(S) 4

==

Description:==

With Residual Heat Removal (RHR) Loop B already secured from shutdown cooling (SDC), shift the SDC lineup from the B SDC loop to the A SDC loop per SOP-RHR-SDC. Service Water Pump 1A (SW-P-1A) fails to automatically start. When manually started, Service Water pump discharge valve (SW-V-2A) fails to automatically open requiring manual operation.

K/A: 205000 A4.09 (3.1 / 3.1)

In-Plant Systems:* 3 for SRO-I P-1: PERFORM ATTACHMENT 7.4 OF ABN-CR-EVAC ON A CONTROL ROOM EVACUATION (TIME CRITICAL)

(D)(E)(L)(R) 1

==

Description:==

Following a Control Room evacuation due to a fire, perform ABN-CR-EVAC, Attachment 7.4, which includes pulling fuses to ensure 4160 VAC Bus SM-8 (Division 2 ESF Bus) is tripped and opening Reactor Protection System (RPS) breakers to ensure a complete scram occurs. Time Critical - 10 minutes.

295016 AA1.01 (3.8 / 3.9)

ES-301 Control Room/In-Plant Systems Outline (Rev 0 - 9/11/18)

Form ES-301-2 P-2: PERFORM WETWELL VENTING LOCALLY USING HARDENED CONTAINMENT VENT (HCV) SYSTEM (E)(L)(N)(R) 5

==

Description:==

Using PPM 5.5.14 (Emergency Wetwell Venting), vent the Wetwell from the HCV Remote Operating Station (ROS) through local manipulation of HCV valves.

K/A: 223001 A2.07 (4.2* / 4.3*)

P-3: BYPASS REACTOR CORE ISOLATION COOLING (RCIC)

TRIPS AND ISOLATIONS (C)(D)(E)(L)

(P)(R) 2

==

Description:==

During Station Blackout (SBO), bypass the RCIC high area temperature isolation from the Control Room and then proceed to the field to defeat the RCIC 25 psig high exhaust pressure trip by isolating local RCIC pressure switches.

295003 AA1.03 (4.4* / 4.4*)

All SRO-I control room (and in-plant) systems must be different and serve different safety functions.

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

• Type Codes Criteria for RO /SRO-I/SRO-U (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power/Shutdown (N)ew or (M)odified from bank including 1(A)

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

I1 (Rev 0 - 9/19/18)

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Columbia Generating Station Date of Exam: 2/25/2019 Operating Test No.: 1 A

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2 3

4 T

O T

A L

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CREW POSITION CREW POSITION CREW POSITION CREW POSITION S

R O

A T

C B

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S R

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B O

P S

R O

A T

C B

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S R

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B O

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RO SRO-I SRO-U RX 1

1 1

1 0

NOR 1

1 1

1 1

I/C 2

4,5 3,4,8 6

4 4

2 MAJ 5,8 7

6 4

2 2

1 TS 2,4 2

0 2

2 RO SRO-I SRO-U RX NOR I/C MAJ TS RO SRO-I SRO-U RX NOR I/C MAJ TS Instructions:

1.

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

2.

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

3.

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

4.

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

I1

I2 (Rev 0 - 9/19/18)

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Columbia Generating Station Date of Exam: 2/25/2019 Operating Test No.: 1 A

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2 3

4 T

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A L

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CREW POSITION CREW POSITION CREW POSITION CREW POSITION S

R O

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C B

O P

S R

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T C

B O

P S

R O

A T

C B

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S R

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B O

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RO SRO-I SRO-U RX 0

1 1

0 NOR 1

2 2

1 1

1 I/C 4,7 3,6,8 2,4,5, 8

9 4

4 2

MAJ 5,8 7

6 4

2 2

1 TS 2,3 2

0 2

2 RO SRO-I SRO-U RX NOR I/C MAJ TS RO SRO-I SRO-U RX NOR I/C MAJ TS Instructions:

1.

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

2.

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

3.

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

4.

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

I2

I3 (Rev 0 - 9/19/18)

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Columbia Generating Station Date of Exam: 2/25/2019 Operating Test No.: 1 A

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2 3

4 T

O T

A L

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CREW POSITION CREW POSITION CREW POSITION CREW POSITION S

R O

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C B

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T C

B O

P S

R O

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C B

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RO SRO-I SRO-U RX 0

1 1

0 NOR 0

1 1

1 I/C 2,3,6 3,4 2,5 7

4 4

2 MAJ 5,8 7

6 4

2 2

1 TS 3,4 2

0 2

2 RO SRO-I SRO-U RX NOR I/C MAJ TS RO SRO-I SRO-U RX NOR I/C MAJ TS Instructions:

1.

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

2.

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

3.

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

4.

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

I3

I4 (Rev 0 - 9/19/18)

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Columbia Generating Station Date of Exam: 2/25/2019 Operating Test No.: 1 A

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2 3

4 T

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CREW POSITION CREW POSITION CREW POSITION CREW POSITION S

R O

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B O

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R O

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C B

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RO SRO-I SRO-U RX 1

1 1

1 0

NOR 0

1 1

1 I/C 2

4,5 2,4,5, 8

7 4

4 2

MAJ 5,8 7

6 4

2 2

1 TS 2,4 2,3 4

0 2

2 RO SRO-I SRO-U RX NOR I/C MAJ TS RO SRO-I SRO-U RX NOR I/C MAJ TS Instructions:

1.

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

2.

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

3.

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

4.

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

I4

I5 (Rev 0 - 9/19/18)

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Columbia Generating Station Date of Exam: 2/25/2019 Operating Test No.: 1 A

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2 3

4 T

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CREW POSITION CREW POSITION CREW POSITION CREW POSITION S

R O

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RO SRO-I SRO-U RX 0

1 1

0 NOR 1

1 2

1 1

1 I/C 4,7 3,4 3,4,8 7

4 4

2 MAJ 5,8 7

6 4

2 2

1 TS 3,4 2

0 2

2 RO SRO-I SRO-U RX NOR I/C MAJ TS RO SRO-I SRO-U RX NOR I/C MAJ TS Instructions:

1.

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

2.

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

3.

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

4.

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

I5

R1 (Rev 0 - 9/19/18)

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Columbia Generating Station Date of Exam: 2/25/2019 Operating Test No.: 1 A

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2 3

4 T

O T

A L

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CREW POSITION CREW POSITION CREW POSITION CREW POSITION S

R O

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R O

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C B

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RO SRO-I SRO-U RX 1

1 0

NOR 1

1 1

I/C 4

4 2

MAJ 2

2 1

TS 0

2 2

RO SRO-I SRO-U RX 0

1 1

0 NOR 2

1 1

1 1

I/C 2,3,6 3,6,8 2,5 8

4 4

2 MAJ 5,8 7

6 4

2 2

1 TS 0

0 2

2 RO SRO-I SRO-U RX 1

1 0

NOR 1

1 1

I/C 4

4 2

MAJ 2

2 1

TS 0

2 2

Instructions:

1.

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

2.

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

3.

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

4.

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

R1

R2 (Rev 0 - 9/19/18)

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Columbia Generating Station Date of Exam: 2/25/2019 Operating Test No.: 1 A

P P

L I

C A

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E V

E N

T T

Y P

E Scenarios 1

2 3

4 T

O T

A L

M I

N I

M U

M(*)

CREW POSITION CREW POSITION CREW POSITION CREW POSITION S

R O

A T

C B

O P

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O A

T C

B O

P S

R O

A T

C B

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S R

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RO SRO-I SRO-U RX 1

1 0

NOR 1

1 1

I/C 4

4 2

MAJ 2

2 1

TS 0

2 2

RO SRO-I SRO-U RX 0

1 1

0 NOR 1

2 2

1 1

1 I/C 4,7 3,6,8 5

4 4

2 MAJ 5,8 7

3 2

2 1

TS 0

0 2

2 RO SRO-I SRO-U RX 1

1 0

NOR 1

1 1

I/C 4

4 2

MAJ 2

2 1

TS 0

2 2

Instructions:

1.

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

2.

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

3.

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

4.

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

R2

R3 (Rev 0 - 9/19/18)

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Columbia Generating Station Date of Exam: 2/25/2019 Operating Test No.: 1 A

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2 3

4 T

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CREW POSITION CREW POSITION CREW POSITION CREW POSITION S

R O

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RO SRO-I SRO-U RX 1

1 0

NOR 1

1 1

I/C 4

4 2

MAJ 2

2 1

TS 0

2 2

RO SRO-I SRO-U RX 1

1 1

1 0

NOR 0

1 1

1 I/C 2,3,6 4,5 5

4 4

2 MAJ 5,8 7

3 2

2 1

TS 0

0 2

2 RO SRO-I SRO-U RX 1

1 0

NOR 1

1 1

I/C 4

4 2

MAJ 2

2 1

TS 0

2 2

Instructions:

1.

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

2.

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

3.

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

4.

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

R3

R4 (Rev 0 - 9/19/18)

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Columbia Generating Station Date of Exam: 2/25/2019 Operating Test No.: 1 A

P P

L I

C A

N T

E V

E N

T T

Y P

E Scenarios 1

2 3

4 T

O T

A L

M I

N I

M U

M(*)

CREW POSITION CREW POSITION CREW POSITION CREW POSITION S

R O

A T

C B

O P

S R

O A

T C

B O

P S

R O

A T

C B

O P

S R

O A

T C

B O

P R

I U

RO SRO-I SRO-U RX 1

1 0

NOR 1

1 1

I/C 4

4 2

MAJ 2

2 1

TS 0

2 2

RO SRO-I SRO-U RX 0

1 1

0 NOR 1

2 2

1 1

1 I/C 4,7 3,6,8 5

4 4

2 MAJ 5,8 7

3 2

2 1

TS 0

0 2

2 RO SRO-I SRO-U RX 1

1 0

NOR 1

1 1

I/C 4

4 2

MAJ 2

2 1

TS 0

2 2

Instructions:

1.

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

2.

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

3.

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

4.

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

R4

R5 (Rev 0 - 9/19/18)

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Columbia Generating Station Date of Exam: 2/25/2019 Operating Test No.: 1 A

P P

L I

C A

N T

E V

E N

T T

Y P

E Scenarios 1

2 3

4 T

O T

A L

M I

N I

M U

M(*)

CREW POSITION CREW POSITION CREW POSITION CREW POSITION S

R O

A T

C B

O P

S R

O A

T C

B O

P S

R O

A T

C B

O P

S R

O A

T C

B O

P R

I U

RO SRO-I SRO-U RX 1

1 0

NOR 1

1 1

I/C 4

4 2

MAJ 2

2 1

TS 0

2 2

RO SRO-I SRO-U RX 1

1 1

1 0

NOR 0

1 1

1 I/C 2,3,6 4,5 5

4 4

2 MAJ 5,8 7

3 2

2 1

TS 0

0 2

2 RO SRO-I SRO-U RX 1

1 0

NOR 1

1 1

I/C 4

4 2

MAJ 2

2 1

TS 0

2 2

Instructions:

1.

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

2.

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

3.

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

4.

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

R5

Appendix D NRC SCENARIO OUTLINE (SC-1) (Rev 0 - 9/13/2018) FORM ES-D-1 Columbia Generating Station CGS 2019 NRC Exam Scenario 1 Outline Page 1 of 6 Facility:

Columbia Generating Station Scenario No.:

1 Op Test No.:

1 Examiners:

Operators:

Initial Conditions:

Reactor power is 100%. RHR-P-2B tagged out for replacement of broken shaft coupling, discovered during last nights surveillance run. Expected to be restored within the next 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. LCOs 3.5.1 A.1, 3.6.1.5 A.1, 3.6.2.3 A.1 and RFO 1.6.1.5 A.1 have been entered. RWCU-P-1B was shutdown 30 minutes ago per SOP-RWCU-SHUTDOWN, section 5.1. Both RWCU Filter Demineralizers have been removed from service and are being backwashed.

Turnover:

Immediately following shift turnover, CRO1 is to perform a Quick Restart of RWCU using RWCU-P-1B per SOP-RWCU-START, section 5.3. Following step 5.3.6, adjust RWCU pump suction flow (as read on RWCU-FI-609) to ~400 gpm using RWCU-V-44. OPS 2 has been briefed on the evolution and is on station standing by to assist.

Critical Tasks:

CT-1 Re-energize SM-7 by closing EITHER breaker B-7 OR the DG-1 Output breaker within 10 minutes after the loss of SM-7 voltage.

CT-2 Initiate Emergency Depressurization (ED) by opening seven (7) Safety Relief Valves (ADS Preferred) within 10 minutes of exceeding the Pressure Suppression Pressure (PSP) limit. CT considered met if any combination of 7 Safety Relief Valves are opened.

NOTE: An unintentional or unnecessary RPS or ESF actuation may result in the creation of a post-scenario Critical Task, if that actuation results in a significant plant degradation or significantly alters a mitigation strategy.

Event No.

Trigger Event Type*

Event Description 1

N (ATC)

Perform RWCU (Reactor Water Cleanup) System Quick Start using RWCU-P-1B per SOP-RWCU-START, section 5.3.

2 TRG-2 C (BOP,SRO)

SGT (Standby Gas Treatment) Train A high temperature occurs due to strip heaters failing to de-energize. (Tech Spec)

TS (SRO) 3 TRG-3 C (BOP)

HD-LIC-3B (Low Pressure Heater 3B Level Indicating Controller) failure in Auto results in low level in Low Pressure heater 3B.

4 TRG-4 I (ATC Ground on RFT-MOP/1A (Reactor Feedwater Turbine 1A Main Oil Pump) with previously faulted AOP (Auxiliary Oil Pump) causes a trip of RFW-P-1A (Reactor Feedwater P -1A). RRC-P-1A (Reactor Recirculation Pump 1A) fails to runback to 30Hz. Manual runback only successful to 45Hz. (Tech Spec)

TS (SRO) 5 TRG-5 M (ALL)

TR-M1 (Main Power Transformer #1) trip causes Load Reject and Automatic Scram. All rods insert. Breaker failures result in the loss of 4160 VAC Bus SM-1, 480 VAC Bus SL-11 and Division 1 Safety Bus SM-7. (See Event 6) 6 I (BOP)

Backup Transformer breaker B-7 malfunction occurs with DG-1 output breaker failing to automatically close on SM-7 under voltage. Breaker B7 or DG-1 Output breaker is closed to manually to re-energize SM-7. (CT-1) 7 I (ATC)

RRC (Reactor Recirculation) pump breakers (3B & 4B) fail to open following Main Turbine trip requiring manual action to secure RRC-P-1A (RRC Pump 1A).

8 TRG-8 M (ALL)

Inboard MSIV closure causes steam leak upstream of MSL 'C' flow restrictor which causes high Drywell pressure and temperature. DG-2 fails to start on high DW pressure and cannot be started manually.

Appendix D NRC SCENARIO OUTLINE (SC-1) (Rev 0 - 9/13/2018) FORM ES-D-1 Columbia Generating Station CGS 2019 NRC Exam Scenario 1 Outline Page 2 of 6 9

Using RHR-P-2A (Residual Heat Removal Pump 2A), initiate Wetwell spray when Wetwell pressure reaches 2 psig. Spray Drywell when Wetwell pressure exceeds 12 psig and prior to exceeding PSP. Pressure reduction successful.

10 Two minutes after Drywell sprays are put into operation, an OBE causes Drywell floor failure, an overcurrent trip of SW-P-1B followed by a trip of RHR-P-2A due to shaft seizure. Emergency Depressurization required on inability to stay below PSP. (CT-2)

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications Target Quantitative Attributes Actual Description Malfunctions after EOP entry (1-2) 3 RRC-P-1B Fails to Trip; Steam Leak in Drywell; Loss of Drywell Spray Capability Abnormal events (2-4) 3 SGT A High Temperature; HD-LIC-3B Controller Failure; RFW-P-1A Trip with Failed RRC Runback Major transients (1-2) 2 Main Generator Load Reject; Steam Leak in Drywell EOPs entered/requiring substantive actions (1-2) 2 PPM 5.1.1 (RPV Control); PPM 5.2.1 (Primary Containment Control)

Entry into a contingency EOP with substantive actions ( 1 per scenario set) 1 PPM 5.1.3 (Emergency RPV Depressurization)

Pre-identified Critical tasks ( 2) 2 See Critical Task Sheets.

Appendix D NRC SCENARIO OUTLINE (SC-1) (Rev 0 - 9/13/2018) FORM ES-D-1 Columbia Generating Station CGS 2019 NRC Exam Scenario 1 Outline Page 3 of 6 SCENARIO 1

SUMMARY

Event 1 Immediately following shift turnover, CRO1 performs a Quick Restart of RWCU (Reactor Water Cleanup) using RWCU-P-1B per SOP-RWCU-START, section 5.3. Following step 5.3.6 (and as directed by the turnover), CRO1 adjusts RWCU pump suction flow (as read on RWCU-FI-609) to ~400 gpm using RWCU-V-44 (RWCU Demineralizer Bypass).

Event 2 (TRG-2) Annunciators P851-S1 1-3 (SGT DIV 1 BOARD K1 Trouble) and P827-K1 2-4 (CHARCOAL FLTR A-1 OUTLET TEMP HI) come in due to high temperature condition inside Standby Gas Treatment (SGT) Train A caused by local strip heaters failing to de-energize. SGT-TI-6A (Carbon Filter 1 temperature indication) shows 265°F (normally ~107°F). ARM-RIS-8 (SGT Filter Area Radiation Monitor) also shows a small rise. CRS enters ABN-SGT-TEMP/RAD which directs starting the unaffected (B) SGT Train per SOP-SGT-START and recirculating the affected (A) SGT Train per ABN-SGT-TEMP/RAD, section 4.2. Following SGT system start, temperature and area radiation will slowly lower.

With SGT Train A Inoperable, the CRS refers to Technical Specifications and determines the following LCO applies:

LCO 3.6.4.3 A.1 - Restoring SGT Train A to Operable status within 7 days Event 3 (TRG-3) Annunciator P840-A2 9-3 (LP HEATER 3B LOW LEVEL) comes in when the HD-LIC-3B (Low Pressure Heater 3B Level Indicating Controller) output fails high in Auto resulting in low level in Low Pressure heater 3B. As directed by the Annunciator Response Procedure (ARP), the heater level controller on H13-P835 Board Y must be placed in Manual and output reduced to slowly restore heater water level to its normal level.

Event 4 (TRG-4) Annunciator P800-C3 7-2 (BUS 11 GROUND) comes in as a result of a ground on RFT-MOP/1A (Reactor Feedwater Turbine 1A Main Oil Pump) which causes the MOP to trip. RFT A control oil pressure (normally ~120 psig as read on pressure indicator RFT-PI-2/1A) starts lowering. RFT-P-AOP/1A (Reactor Feedwater Turbine 1A Auxiliary Oil Pump) fails to auto start as control oil pressure lowers to 70 psig and cannot be manually started. RFW-P-1A trips when control oil pressure reaches 50 psig (as indicated by annunciator P840-A1 1-1 (TURB A TRIP)).

Following the feed pump trip, and when RPV water level lowers to L4 (31.5), RRC (Reactor Recirculation) Pump 1B starts running back (from ~55 Hz) to 30 Hz, as designed but RRC Pump 1A fails to runback. When the RRC Pump 1A flow controller is used to manually runback the pump, it only lowers to 45 Hz. The CRS enters ABN-CORE and ABN-POWER to assess core operating conditions and RRC pump operating restraints.

With RRC loop flow mismatch not within limits, the CRS refers to Technical Specifications and determines the following LCO applies:

LCO 3.4.1 A.1 - Declare the recirculation loop with lower flow to be not in operation within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Event 5 (TRG-5) TR-M1 (Main Power Transformer #1) trip causes Load Reject and automatic scram. All rods insert. Startup transformer breaker (S-1) fails to close (auto or manually) causing a loss of 4160 VAC Bus SM-1 and 480 VAC Bus SL-11. Division 1 Safety Bus (SM-7) also de-energizes due to failure of breaker B-7 (Backup Transformer feed breaker for SM-7) and the Division 1 EDG (DG-1) output breaker feed to SM-7 to automatically close. This results in DG-1 running without cooling water. (See Event 6)

CRS enters PPM 5.1.1 (RPV Control) on low reactor water level and ABN-ELEC-SM1/SM7 due to loss of busses.

PPM 3.3.1 (Reactor Scram) is also entered. The crew will not be able to shift the feedwater lineup to the startup flow control valves until power is restored to SL-11 (which powers key MOVs). The crew restores power to SM-7 per ABN-ELEC-SM1/SM7 (see Event 6) and SL-11 per SOP-ELEC-480V-OPS-QC. Meanwhile, RPV injection is accomplished with RCIC (Reactor Core Isolation Cooling) and/or HPCS (High Pressure Core Spray), as needed.

Appendix D NRC SCENARIO OUTLINE (SC-1) (Rev 0 - 9/13/2018) FORM ES-D-1 Columbia Generating Station CGS 2019 NRC Exam Scenario 1 Outline Page 4 of 6 Event 6 TR-B breaker B-7 and DG-1 Output breaker (DG1-7) fail to automatically close on SM-7 under voltage resulting in DG-1 running without the required SW (Service Water) cooling. The crew re-energizes SM-7 by closing EITHER breaker B-7 OR the DG-1 Output breaker within 10 minutes after the loss of SM-7 voltage. (CT-1) This action restores SW cooling to DG-1 and allows SM-7 to be used for Containment sprays. (See Event 9)

Event 7 When the Main Turbine trips following the scram, RRC (Reactor Recirculation) pump breakers (3B & 4B) fail to open allowing RRC-P-1B (RRC Pump 1A) to continue to run. The STOP PB will not secure the pump. Opening breaker RRB, 3B or 4B will secure the pump. (This is particularly important when RCC (Reactor Closed Cooling Water) cooling is lost to the pump on high DW pressure signal). (See Event 8)

Event 8 (TRG-8) Inboard MSIV closure causes steam leak upstream of MSL 'C' flow restrictor which causes high Drywell pressure and temperature. DG-2 fails to start on high DW pressure (due to an incomplete start sequence) and cannot be started manually. Division 2 Safety Bus (SM-8) remains energized from TR-S (Startup Transformer).

The CRS re-enters PPM 5.1.1 (RPV Control) on high Drywell pressure (1.68 psig) and enters PPM 5.2.1 (Primary Containment Control) on high Drywell pressure and high Drywell temperature (135°F).

Event 9 The CRS sets a key parameter for Wetwell pressure reaching 2 psig. When reached, RHR-P-2A (Residual Heat Removal Pump 2A) is used to spray the Wetwell per SOP-RHR-SPRAY-WW-QC. The CRS also sets a key parameter for Wetwell pressure exceeding 12 psig. When exceeded (and prior to exceeding Pressure Suppression Pressure (PSP) - PPM 5.2.1, Fig. F), RHR-P-2A is used to spray the Drywell per SOP-RHR-SPRAY-DW-QC. Drywell pressure reduction is successful.

Event 10 Two minutes after Drywell sprays are put into operation, an OBE (Operating Basis Earthquake) causes Drywell floor failure, an overcurrent trip of SW-P-1B followed by a trip of RHR-P-2A due to shaft seizure. With sprays no longer available (RHR pump 'A', RHR pump 'B', & SW pump 'B' are not available for Containment sprays), the CRS directs an Emergency Depressurization per PPM 5.1.3 (RPV Depressurization) once the Pressure Suppression Pressure (PSP) limit is exceeded per PPM 5.2.1, Fig. F. (CT-2)

TERMINATION CRITERIA: The scenario will be terminated when Drywell sprays have been initiated, an Emergency Depressurization has been performed, and RPV level is being controlled in the prescribed band OR as directed by the Lead Examiner.

Appendix D NRC SCENARIO OUTLINE (SC-1) (Rev 0 - 9/13/2018) FORM ES-D-1 Columbia Generating Station CGS 2019 NRC Exam Scenario 1 Outline Page 5 of 6 CT-1 Critical Task Statement:

Re-energize SM-7 by closing EITHER breaker B-7 OR the DG-1 Output breaker within 10 minutes after the loss of SM-7 voltage.

Safety Significance:

Loss of power to SM-7 removes redundant source of power needed to prevent a significant challenge to Safety Functions required to maintain a safe shutdown condition. With RHR-P-2B initially out of service, it is essential to restore power to SM-7 to allow use of RHR-P-2A to mitigate the event through spraying the Drywell (which reduces the challenge to Containment). DG-1 can run for 10.1 minutes without SW cooling before reaching its temperature limit of 220°F. Failure to provide SW cooling within this time frame could result in failure of the DG further impacting Safe Shutdown margin. (Ref: OI-69 (TCOA-1))

Initiating Cue:

• Annunciators showing SM-7 Loss of Voltage

• Bus SM-7 voltage as indicated on H13-P800 shows 0 VAC

• Control Board Breaker Status Measurable Performance Standard:

The operator will manually close breaker B7 OR breaker DG1-7 to re-energize Bus SM-7.

Performance Feedback:

Bus SM-7 voltage as indicated on H13-P800 shows approximately 4160 VAC.

Appendix D NRC SCENARIO OUTLINE (SC-1) (Rev 0 - 9/13/2018) FORM ES-D-1 Columbia Generating Station CGS 2019 NRC Exam Scenario 1 Outline Page 6 of 6 CT-2 Critical Task Statement:

Initiate Emergency Depressurization (ED) by opening seven (7) Safety Relief Valves (ADS Preferred) within 10 minutes of exceeding the Pressure Suppression Pressure (PSP) limit. CT considered met if any combination of 7 Safety Relief Valves are opened.

Safety Significance:

If Wetwell or drywell sprays could not be started or if their operation was not effective in reducing primary containment pressure, the RPV is depressurized to minimize further release of energy from the RPV to the primary containment. The PSP is a function of suppression pool water level. It is used in the EOPs to ensure that pressure suppression capability sufficient to accommodate emergency RPV depressurization is maintained while the RPV is at pressure. (Ref: PPM 5.0.10)

Initiating Cue:

Containment pressure is approaching the PSP limit as indicated in PPM 5.2.1 (Primary Containment Control), Fig. F.

Measurable Performance Standard:

7 Safety Relief Valves are manually opened.

Performance Feedback:

The valve light indications for each of the 7 Safety Relief Valves will change from Green lit to Red lit when control switch is taken to Open. Reactor pressure is lowering as indicated on MS-LR/PR-623A, MS-LR/PR-623B, or MS-PI-9.

Appendix D NRC SCENARIO OUTLINE SC-2 (Rev 0 - 9/13/2018) FORM ES-D-1 Columbia Generating Station CGS 2019 NRC Exam Scenario 2 Outline Page 1 of 7 Facility:

Columbia Generating Station Scenario No.:

2 Op Test No.:

1 Examiners:

Operators:

Initial Conditions:

Reactor power is 92%. Power ascension to 100% using RRC Flow will continue after assuming the shift. OSP-HPCS/IST-Q701, HPCS System Operability Test, is in-progress.

Turnover:

Continue with the performance of OSP-HPCS/IST-Q701, starting at Step 7.3. Initialed steps have already been performed. LCO 3.5.1 B.1 & B.2 was entered just before your shift to support the surveillance. The two year VPI and channel calibration are NOT due. Non-intrusive testing of HPCS-V-16 & 24 is not required. Placing RHR in SP Cooling is not required. The pre-job brief has been completed, and Equipment Operators are on station to support completion of the surveillance. HP has been informed of surveillance performance. The power increase and surveillance are to be performed concurrently.

Critical Tasks:

CT-1 Trip Division 1 Diesel Generator (DG-1) within 10 minutes of Division 1 Safety Bus (SM-7) lockout.

CT-2 Initiate Emergency Depressurization (ED) by opening seven (7) Safety Relief Valves (ADS preferred) after RPV water level reaches TAF (-161 inches) and within 10 minutes of level dropping below TAF. CT considered met if any combination of 7 Safety Relief Valves are opened.

CT-3 After ED, and within 10 minutes of RPV pressure lowering to 200 psig, restore and maintain RPV water level above TAF (-161 inches) using low pressure systems.

NOTE: An unintentional or unnecessary RPS or ESF actuation may result in the creation of a post-scenario Critical Task, if that actuation results in a significant plant degradation or significantly alters a mitigation strategy.

Event No.

Trigger Event Type*

Event Description 1

R (ATC)

Raise reactor power to 100% using RRC (Reactor Recirculation) Flow.

2 N (BOP)

Perform the HPCS (High Pressure Core Spray) System Operability Test, OSP-HPCS/IST-Q701. Performed concurrently with raising power.

3 C (BOP,SRO)

The HPCS min flow, HPCS-V-12, fuses clear while closing. (Tech Spec)

TS (SRO) 4 TRG-4 C (ATC,SRO)

High vibrations occur on RRC-P-1A resulting in lower seal failure which requires securing pump. Subsequent failure of upper seal requires isolating RRC loop 'A'.

(Tech Spec)

TS (SRO) 5 TRG-5 C (ATC)

RRC-P-1B trips requiring a manual scram due to no forced RRC flow.

6 I (BOP)

Main Generator fails to Auto Trip requiring a manual trip.

(CGS LER 2016-004-01) 7 TRG-7 M (ALL)

OBE (Earthquake) causes RRC LOCA, broken RCIC (Reactor Core Isolation Cooling) shaft, and lockout of the Startup Transformer. DG-3 trips upon start and is unavailable (no HPCS (High Pressure Core Spray)).

Appendix D NRC SCENARIO OUTLINE SC-2 (Rev 0 - 9/13/2018) FORM ES-D-1 Columbia Generating Station CGS 2019 NRC Exam Scenario 2 Outline Page 2 of 7 8

C (BOP)

When Drywell pressure reaches 1.68 psig, RHR-P-2A will automatically start with an overcurrent condition. The output breaker fails to trip which causes a lockout on SM-7. RHR-P-2B will not start and cannot be started. DG-1 must be emergency tripped due to loss of SW cooling. (CT-1) (TCOA to trip DG-1 within 10 minutes - REF: OI-69 TCOA-1) 9 Initiate Wetwell Spray (using PPM 5.5.2 (RHR/SW Cross Tie Lineup)) when Wetwell pressure reaches 2 psig. Spray Drywell when Wetwell pressure exceeds 12 psig and prior to exceeding PSP. Pressure reduction successful.

10 Initiate Emergency Depressurization (ED) on low RPV level (CT-2) and restore RPV level to above TAF using low pressure systems. (CT-3)

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications Target Quantitative Attributes Actual Description Malfunctions after EOP entry (1-2) 2 Loss of High Pressure Injection; Loss of Low Pressure Injection (except LPCS & SW B)

Abnormal events (2-4) 4 HPCS Min Flow Fuses Clear; RRC-P-1A High Vibrations/Seal Failure; RRC-P-1B Trip; Main Generator fails to Auto Trip Major transients (1-2) 1 OBE (Earthquake) with Loss of High Pressure Feed EOPs entered/requiring substantive actions (1-2) 2 PPM 5.1.1 (RPV Control); PPM 5.2.1 (Primary Containment Control)

Entry into a contingency EOP with substantive actions ( 1 per scenario set) 1 PPM 5.1.3 (Emergency RPV Depressurization)

Pre-identified Critical tasks ( 2) 3 See Critical Task Sheets.

Appendix D NRC SCENARIO OUTLINE SC-2 (Rev 0 - 9/13/2018) FORM ES-D-1 Columbia Generating Station CGS 2019 NRC Exam Scenario 2 Outline Page 3 of 7 SCENARIO 2

SUMMARY

Event 1 CRO1 raises reactor power to 100% with RRC (Reactor Recirculation) flow at the rate of 1% per minute or 1 Hz per minute.

Event 2 CRO2 performs the HPCS System Operability Test, OSP-HPCS/IST-Q701 starting with step 7.3. The surveillance is performed concurrently with the power rise.

Event 3 During performance of OSP-HPCS/IST-Q701 (step 7.3.16), while raising HPCS system flow beyond 3000 gpm, the fuses clear for HPCS-V-12 (HPCS min flow valve). Annunciator P601.A1 6-8 (High Pressure Core Spray Out of Service) comes in. Investigation reveals BISI (Bypass and Inoperable Status Indication) for HPCS System shows a MOV Network Power Loss/Overload and that the position indications lights for HPCS-V-12 have gone out.

With power lost to HPCS-V-12 (which is a Primary Containment Isolation Valve), the CRS refers to Technical Specifications and determines the following LCOs apply:

LCO 3.6.1.3 C.1 - Isolate the affected penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, or blind flange.

LCO 3.6.1.3 C.2 - Verify the affected penetration flow path is isolated once per 31 days.

Event 4 (TRG-4) Annunciator P602.A6 2-4 (Recirc A System VIB High) comes in due to high vibrations on RRC Pump 1A.

OPS2 will report RRC Pump 1A pump casing (X and Y) vibrations at 4.0 PK-G (above alarm set point), RRC Pump 1A pump casing (Z) vibrations at 4.5 PK-G (above alarm set point) and all other pump vibrations slightly higher but below their alarm set points. CRS directs a RRC flow reduction per PPM 3.2.6 (Power Maneuvering) with both RRC pumps (to reduce Pump 1A speed) in effort to reduce RRC Pump 1A vibrations. Vibrations slightly lower after speed reduction but remain above their alarm setpoints and continue to slowly rise thereafter.

Several minutes after the vibrations began, the RRC Pump 1A lower seal will start degrading as evidenced by rising RRC Pump 1A cavity pressure on RRC pressure instrument 602A. CRS enters ABN-RCC-SEAL. RRC Pump 1A upper seal pressure eventually exceeds 925 psig requiring tripping of the pump and entering ABN-RRC-LOSS. The CRS will also enter ABN-CORE to assess core operating conditions as it pertains to OPRM enabled, Reactor Recirculation flow, and Rod Line. Eventually both seals fail for RRC-P-1A (as evidenced by the RRC Pump 1A upper and lower seal pressures failing low) requiring isolation of the A RRC Loop per ABN-RRC-SEAL.

With only one RRC pump operating and the requirements of LCO 3.4.1 not met (LCO 3.2.1, LCO 3.2.2, and LCO 3.3.1.1 (Function 2b), the CRS refers to Technical Specifications and determines the following LCO applies:

LCO 3.4.1 B.1 - Satisfy the requirements of the LCO within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Event 5 (TRG-5) RRC-P-1B trips requiring a manual scram due to no forced RRC flow as directed per ABN-RRC-LOSS Immediate Operator Actions. The CRS enters PPM 5.1.1 (RPV Control) on low RPV level. With no RRC pumps running, actions will be taken to limit thermal stratification of the RPV lower head region to include reducing CRD (Control Rod Drive) flow to the reactor and controlling RPV pressure in batches to induce water mixing.

Appendix D NRC SCENARIO OUTLINE SC-2 (Rev 0 - 9/13/2018) FORM ES-D-1 Columbia Generating Station CGS 2019 NRC Exam Scenario 2 Outline Page 4 of 7 Event 6 Following insertion of a manual scram, the Main Turbine trips but the Main Generator fails to automatically trip. The Main Generator is manually tripped by depressing either the Unit Emergency Trip pushbutton OR the Unit Overall Trip pushbutton at H13-P800 as directed in PPM 3.3.1 (Reactor Scram).

Recent Columbia OE illustrates a case when manually tripping the Main Generator (when required) failed to occur.

(CGS LER 2016-004-01)

Event 7 (TRG-7) OBE (Earthquake) causes RRC LOCA, broken RCIC (Reactor Core Isolation Cooling) shaft, and lockout of the Startup Transformer. DG-3 trips upon start and is unavailable (no HPCS (High Pressure Core Spray)).

LOCA develops causing a large loss of inventory and a rise in Drywell pressure. The CRS re-enters PPM 5.1.1 (RPV Control) and PPM 5.2.1 (Primary Containment Control) on high Drywell pressure (1.68 psig). Without the Startup Transformer following a unit trip, the normal source of feedwater injection is not available (feed and condensate).

RCIC (Reactor Core Isolation Cooling) and HPCS (High Pressure Core Spray) are also not available. The remaining high pressure (but low capacity) sources such as SLC (Standby Liquid Control) and CRD (Control Rod Drive) are insufficient to maintain RPV water inventory. RPV level starts trending down.

Event 8 When Drywell pressure reaches 1.68 psig, RHR-P-2A automatically starts with an overcurrent condition. The output breaker fails to trip which causes a lockout on SM-7. RHR-P-2B will not start and cannot be started. This leaves the crew with only RHR Pump 2C (which can be used for post-ED injection) and SW (Service Water) Pump 1B (which can be used for Wetwell and Drywell sprays and post-ED injection).

With the lockout of SM-7, DG-1 (which previously started on high Drywell pressure) is running without cooling water (SW (Service Water) Pump 1A de-energized when SM-7 locked out). DG-1 must be emergency tripped due to loss of SW cooling. (CT-1) (TCOA to trip DG-1 within 10 minutes - REF: OI-69 TCOA-1)

Event 9 As directed by PPM 5.2.1 (Primary Containment Control), the crew Initiates Wetwell Spray (using PPM 5.5.2 (RHR/SW Cross Tie Lineup and SW pump 1B)) when Wetwell pressure reaches 2 psig. Crew will spray Drywell (with SW pump 1B) when Wetwell pressure exceeds 12 psig and prior to exceeding PSP. Drywell pressure will start trending down once Drywell sprays are placed in operation.

Event 10 Crew Initiates Emergency Depressurization (ED) on low RPV level (CT-2) and restores RPV level to above TAF using low pressure systems. (CT-3)

TERMINATION CRITERIA: The scenario will be terminated when Drywell sprays have been initiated, an Emergency Depressurization has been performed, and RPV level is being controlled above TAF (- 161 inches) OR as directed by the Lead Examiner.

Appendix D NRC SCENARIO OUTLINE SC-2 (Rev 0 - 9/13/2018) FORM ES-D-1 Columbia Generating Station CGS 2019 NRC Exam Scenario 2 Outline Page 5 of 7 CT-1 Critical Task Statement:

Trip Division 1 Diesel Generator (DG-1) within 10 minutes of Division 1 Safety Bus (SM-7) lockout.

Safety Significance:

Service Water provides cooling for Diesel Generator Diesel Cooling Water System (DCW). The DCW high temperature trip of the Diesel Generator is bypassed on a LOCA signal. The reservoir tank and the DCW volume provide adequate cooling for 10 minutes without Service Water (ME-02-94-42). If Service Water cannot be started within 10 minutes of the start of the Diesel Generator, the Diesel Generator must be tripped to avoid potential damage caused by over-heating. No Service Water available with SM-7 lockout present. (Ref: OI-69: TCOA-1)

Initiating Cue:

DG-1 is running with a valid start signal and SW-P-1A is not running as indicated by SW-PI-32A, Pump Discharge Pressure, SW-FI-9A, Loop A Flow, and SW-P-1A Motor Current indications all reading zero.

Measurable Performance Standard:

DIESEL GEN 1 EMERGENCY TRIP pushbutton on H13-P800 (Board C) used to trip DG.

Performance Feedback:

DG-1 Bkr 7DG1 Trip annunciator in. DG-1 Control Switch status indicator changes from red (running) to green (not running).

Appendix D NRC SCENARIO OUTLINE SC-2 (Rev 0 - 9/13/2018) FORM ES-D-1 Columbia Generating Station CGS 2019 NRC Exam Scenario 2 Outline Page 6 of 7 CT-2 Critical Task Statement:

Initiate Emergency Depressurization (ED) by opening seven (7) Safety Relief Valves (ADS preferred) after RPV water level reaches TAF (-161 inches) and within 10 minutes of level dropping below TAF. CT considered met if any combination of 7 Safety Relief Valves are opened.

Safety Significance:

Preclude core damage by establishing conditions that allow low pressure ECCS systems to restore water level above TAF (Safety Limit). (Ref: CGS Technical Specifications - 2.1.1.3)

Initiating Cue:

Procedural direction in PPM 5.1.1 (RPV Control) when RPV level drops to -161 inches (TAF) as indicated as indicated on MS-LR-615.

Measurable Performance Standard:

7 Safety Relief Valves are manually opened.

Performance Feedback:

The valve light indications for each of the 7 Safety Relief Valves will change from Green lit to Red lit when control switch is taken to Open. Reactor pressure is lowering as indicated on MS-LR/PR-623A, MS-LR/PR-623B, or MS-PI-9.

Appendix D NRC SCENARIO OUTLINE SC-2 (Rev 0 - 9/13/2018) FORM ES-D-1 Columbia Generating Station CGS 2019 NRC Exam Scenario 2 Outline Page 7 of 7 CT-3 Critical Task Statement:

After ED, and within 10 minutes of RPV pressure lowering to 200 psig, restore and maintain RPV water level above TAF (-161 inches) using low pressure systems.

Safety Significance:

Precludes core damage by establishing conditions that allow low pressure injection systems to restore water level above TAF (Safety Limit). (Ref: CGS Technical Specifications - 2.1.1.3)

Initiating Cue:

Procedural direction in PPM 5.1.1 (RPV Control) which directs maximizing injection into RPV with all available sources.

Measurable Performance Standard:

All available low pressure injection sources are aligned to restore RPV level.

Performance Feedback:

Indication of applicable injection flow. RPV level rises to greater than TAF.

Appendix D NRC SCENARIO OUTLINE SC-3 (Rev 0 - 9/13/2018) FORM ES-D-1 Columbia Generating Station CGS 2019 NRC Exam Scenario 3 Outline Page 1 of 7 Facility:

Columbia Generating Station Scenario No.:

3 Op Test No.:

1 Examiners:

Operators:

Initial Conditions:

Reactor power at 80% due to economic dispatch. Maintenance is to be performed on RCC-P-1A during your shift.

Turnover:

Swap RCC pumps to RCC-P-1B running and RCC-P-1A in standby. Pre-Job brief has been completed and OPS2 is standing by near the RCC pumps.

Critical Tasks:

CT-1 Stop and prevent injection into the RPV, with the exception of SLC, RCIC, and CRD, to establish an LL of -65 inches in an ATWS with GT 5% power within 20 minutes of entry into PPM 5.1.2.

CT-2 With reactor scram required and the reactor not shutdown, commence inserting control rods per PPM 5.5.11 (Attachment 6.1 - Tab B) within 20 minutes of entry into PPM 5.1.2.

CT-3 Initiate Emergency Depressurization (ED) by opening the first of seven (7) Safety Relief Valves (ADS preferred) before Wetwell level lowers to 19 feet, 2 inches. CT considered met if any combination of 7 Safety Relief Valves are opened within 10 minutes of Wetwell level lowering below 19 feet, 2inches.

NOTE: An unintentional or unnecessary RPS or ESF actuation may result in the creation of a post-scenario Critical Task, if that actuation results in a significant plant degradation or significantly alters a mitigation strategy.

Event No.

Trigger Event Type*

Event Description 1

N (BOP)

Swap RCC (Reactor Closed Cooling) Pumps.

2 TRG-2 C (ATC/SRO)

Rod Block Monitor (RBM) NUMAC Critical Failure results in Rod Out Block requiring Manual Bypass. (Tech Spec)

TS (SRO) 3 TRG-3 I (BOP)

LD-TE-4A (RCIC (Reactor Cooling Isolation Cooling) Equipment Room temperature element) fails high causing a RCIC isolation and RCIC turbine trip.

RCIC-V-8 does not auto close but can be closed manually. (Tech Spec)

TS (SRO) 4 TRG-4 C (BOP,SRO)

OBE causes Wetwell wall rupture common to RHR 'A' & RHR 'B' Pump Rooms.

Wetwell level starts lowering requiring emergency makeup per PPM 5.5.23.

HPCS will subsequently trip following Wetwell level rise.

5 C (ATC,SRO)

Manual scram required before Wetwell level reaches 19 ft 2 in. May scram earlier based on plant damage caused by earthquake.

6 M (ALL)**

High Power Hydraulic ATWS (-80" to -140" Band): Terminate and prevent injection into the RPV with the exception of SLC, RCIC, and CRD, to establish a LL. (CT-1) 7 SLC-P-1A experiences severe flow reduction and SLC-P-1B experiences shaft break when SLC injection required.

8 C (BOP,SRO)

Perform PPM 5.5.11 to insert control rods. (CT-2) All control rods insert on first Scram-Reset-Scram (S/R/S) attempt. Manual rod insertion successful until S/R/S performed. CRS exits PPM 5.1.2 and returns to PPM 5.1.1 once all rods are in.

9 Perform Emergency Depressurization when Wetwell level cannot be maintained above 19 ft 2 in. (CT-3)

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications

• • Event forms portion of significant CGS PSA Accident Sequence (TTC044)(Ref: PSA-1-SM-0001)

Appendix D NRC SCENARIO OUTLINE SC-3 (Rev 0 - 9/13/2018) FORM ES-D-1 Columbia Generating Station CGS 2019 NRC Exam Scenario 3 Outline Page 2 of 7 Target Quantitative Attributes Actual Description Malfunctions after EOP entry (1-2) 3 HPCS Pump Trip; Hydraulic ATWS; Degradation of SLC Flow Abnormal events (2-4) 3 RBM NUMAC Critical Failure; LD-TE-4A Fails High; OBE Major transients (1-2) 1 Hydraulic ATWS EOPs entered/requiring substantive actions (1-2) 3 PPM 5.1.1 (RPV Control); PPM 5.2.1 (Primary Containment Control); PPM 5.3.1 (Secondary Containment Control)

Entry into a contingency EOP with substantive actions ( 1 per scenario set) 2 PPM 5.1.2 (RPV Control - ATWS); PPM 5.1.3 (Emergency RPV Depressurization)

Pre-identified Critical tasks ( 2) 3 See Critical Task Sheets.

Appendix D NRC SCENARIO OUTLINE SC-3 (Rev 0 - 9/13/2018) FORM ES-D-1 Columbia Generating Station CGS 2019 NRC Exam Scenario 3 Outline Page 3 of 7 SCENARIO 3

SUMMARY

Event 1 RCC (Reactor Closed Cooling) pumps are swapped to RCC-P-1B running and RCC-P-1A in standby per SOP-RCC-OPS to support maintenance.

Event 2 (TRG-2) Annunciator P603.A8 3-5 (RBM UPSL OR INOP) comes in due to NUMAC failure of the B RBM (Rod Block Monitor). A physical withdraw block is generated preventing outward rod movement. The ARP (Annunciator Response Procedure) directs checking RBM status which shows it as the cause for the rod block due to RBM Downscale and RBM Inoperative trips.

The CRS directs bypassing the faulty RBM at H13-P603. Once bypassed, the physical Withdraw Block clears again allowing outward rod motion.

With RBM B bypassed (and made Inoperable), the CRS refers to Technical Specifications and determines the following LCO applies:

LCO 3.3.2.1 A.1 - Restore RBM channel to Operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Event 3 (TRG-3) Annunciator P601.A12 5-2 (LEAK DET RCIC EQUIP AREA TEMP HIGH) comes due to LD-TE-4A (RCIC (Reactor Cooling Isolation Cooling) Equipment Room temperature element) failing high. Temperature as read on LD-TRS-608 (RCIC Equipment Area Temperature Recorder) shows 400°F (failed high).

The resultant RCIC isolation signal generates a RCIC isolation and RCIC turbine trip. RCIC-V-8 does not auto close but can be closed manually.

With one channel of RCIC Function 3.e Inoperable (see LCO 3.3.6.1 Table 3.3.6.1-1), the CRS refers to Technical Specifications and determines the following LCO applies:

LCO 3.3.6.1 A.1 - Place channel in trip within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

With RCIC-V-8 (RCIC Turbine Steam Supply Isolation) failing to automatically isolate and therefore becoming Inoperable, the CRS refers to Technical Specifications and determines the following LCOs apply:

LCO 3.6.1.3 A.1 - Isolate the affected penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

LCO 3.6.1.3 A.2 - Verify the affected penetration flow path is isolated once per 31 days.

With RCIC system made Inoperable, the CRS refers to Technical Specifications and determines the following LCOs apply:

LCO 3.5.3 A.1 - Verify by administrative means High Pressure Core Spray System is Operable (Immediately).

LCO 3.5.3 A.2 - Restore RCIC System to Operable status within 14 days.

Event 4 (TRG-4) Annunciators P851.S1 5-1 (OPERATING BASIS EARTHQUAKE) followed by P601.A3 2-7 (Leak Detection Reactor Building Floor Sump R1 Leakage High) and P601.A2 3-2 (Leak Detection Reactor Building Floor Sump R2 Leakage High) come in due to the earthquake causing flooding into the RHR (Residual Heat Removal) Pump 2A and 2B pump rooms from the Wetwell. As the R1 and R2 sumps fill up, annunciators P602.A13 1-1 (Reactor Building Floor Sump R2 Level Hi Hi) and P602.A13 2-1 (Reactor Building Floor Sump R1 Level Hi Hi) also come in.

The CRS enters ABN-EARTHQUAKE, ABN-FLOODING and PPM 5.2.1 (Primary Containment Control). Wetwell emergency makeup per PPM 5.5.23 using HPCS (High Pressure Core Spray) commences. One minute after HPCS

Appendix D NRC SCENARIO OUTLINE SC-3 (Rev 0 - 9/13/2018) FORM ES-D-1 Columbia Generating Station CGS 2019 NRC Exam Scenario 3 Outline Page 4 of 7 is injecting into the Wetwell for makeup, HPCS trips effectively securing makeup allowing Wetwell level to start lowering again. As RHR Pump A and B rooms continue to be flooded, annunciators P601.A4 5-3 (RHR A PUMP ROOM WATER LEVEL HIGH) and P601.A2 2-7 (RHR B PUMP ROOM WATER LEVEL HIGH) come in requiring the CRS to enter PPM 5.3.1 (Secondary Containment Control). Due to the rupture being unisolable with a continuing downward Wetwell level trend, the CRS directs a manual scram (see Event 5). Note: CRS may also direct a manual scram based on degrading plant conditions per ABN-EARTHQUAKE.

Event 5 CRS directs manual scram before Wetwell level reaches 19 ft 2 in. May scram earlier based on plant damage caused by earthquake. Upon placing the reactor Mode switch in Shutdown, no control rods insert. Use of the manual scram pushbuttons and Alternate Rod Insertion (ARI) also fail to shut down the reactor. SLC (Standby Liquid Control) is initiated but to little effect (see Event 7).

Event 6 High Power Hydraulic ATWS - CRS enters PPM 5.1.1 (RPV Control) and transitions to PPM 5.1.2 (RPV Control -

ATWS).

CRS directs the following major additional actions:

Inhibit ADS (Automatic Depressurization System)

Take manual control of HPCS Perform PPM 5.5.6 (Bypassing the MSIV Isolation Interlocks on High Tunnel Temperature and Low RPV Level)

Perform PPM 5.5.1 (Overriding ECCS Valve Logic to Allow Throttling RPV Injection)

Terminate and prevent injection into the RPV with the exception of SLC, RCIC, and CRD, to establish a LL

(-80" to -140" Band) (CT-1)

Event forms portion of significant CGS PSA Accident Sequence (TTC044) (Ref: PSA-1-SM-0001)

Event 7 SLC-P-1A experiences severe flow reduction and SLC-P-1B experiences shaft break when SLC injection required.

Leaves control rod insertion as primary means to shutdown reactor under all conditions. (See Event 8)

Event 8 CRS directs performance of PPM 5.5.11 (Alternate Control Rod Insertions) to insert control rods based on a hydraulic ATWS (CT-2). Attachment 6.1 - Tab B is used to insert control rods manually or through one or more Scram-Reset-Scram attempts.

All control rods insert after the first Scram-Reset-Scram is performed. CRS exits PPM 5.1.2 and returns to PPM 5.1.1 once all rods are in.

Event 9 As Wetwell level continues to lower during ATWS actions, a decision will have to be made by the CRS as to whether an Emergency Depressurization (ED) is to be performed per PPM 5.1.5 (Emergency RPV Depressurization - ATWS) if rods have yet to be fully inserted or per PPM 5.1.3 (Emergency RPV Depressurization) after all control rods have been inserted. It is expected that the crew will be able to insert all rods before an ED is required since all control rods fully insert on the first Scram-Rest-Scram attempt.

In either case, ED is initiated by opening the first of seven (7) Safety Relief Valves (ADS preferred) before Wetwell level lowers to 19 feet, 2 inches. CT considered met if any combination of 7 Safety Relief Valves are opened within 10 minutes of Wetwell level lowering below 19 feet, 2inches. (CT-3)

TERMINATION CRITERIA: The scenario will be terminated when all control rods have been fully inserted, an Emergency Depressurization has been performed, and RPV level is being controlled in the prescribed band OR as directed by the Lead Examiner.

Appendix D NRC SCENARIO OUTLINE SC-3 (Rev 0 - 9/13/2018) FORM ES-D-1 Columbia Generating Station CGS 2019 NRC Exam Scenario 3 Outline Page 5 of 7 CT-1 Critical Task Statement:

Stop and prevent injection into the RPV, with the exception of SLC, RCIC, and CRD, to establish an LL of -65 inches in an ATWS with GT 5% power within 20 minutes of entry into PPM 5.1.2.

Safety Significance:

If reactor power is above 5% or unknown, direction will be given to lower RPV level below the elevation of the feedwater spargers to prevent or mitigate the consequences of any irregular neutron flux oscillations induced by neutronic/thermal-hydraulic instabilities. With the spargers in the steam space, heating of the feedwater by contact with the steam reduces core inlet subcooling. The initiation and growth of oscillations is principally dependent upon the core inlet subcooling; the greater the subcooling, the more likely oscillations will commence and increase in magnitude. -65 in. (twenty-four inches below the feedwater sparger) has been selected as the upper bound of the RPV water level control band. This water level is sufficiently low that steam heating of the injected water will be at least 75% effective.

Initiating Cue:

Reactor scram required and reactor not shutdown as indicated by APRM Downscale lights not lit and Reactor Power indicating >5% on APRMs or cannot be determined.

Measurable Performance Standard:

Stop and prevent injection with the exception of SLC, RCIC, and CRD in accordance with OI-15, within 20 minutes of entry into PPM 5.1.2.

Performance Feedback:

RPV level is lowering as indicated on RPV NR & WR Level instrumentation and reactor power is lowering as indicated on APRMs.

Appendix D NRC SCENARIO OUTLINE SC-3 (Rev 0 - 9/13/2018) FORM ES-D-1 Columbia Generating Station CGS 2019 NRC Exam Scenario 3 Outline Page 6 of 7 CT-2 Critical Task Statement:

With reactor scram required and the reactor not shutdown, commence inserting control rods per PPM 5.5.11 (Attachment 6.1 - Tab B) within 20 minutes of entry into PPM 5.1.2.

Safety Significance:

If reactor power is elevated (above the APRM downscale trip setpoint) or cannot be determined, the core may be susceptible to large, irregular neutron flux oscillations.

Analyses of neutronic / thermal-hydraulic instabilities during failure-to-scram conditions have been performed. Instabilities are manifested by oscillations in reactor power which, if the reactor cannot be shutdown, may increase in magnitude. If the oscillations remain small or moderately sized, they tend to repeat on approximately a two second period.

Under certain circumstances, however, the oscillations may continue to grow and become sufficiently large and irregular to cause localized fuel damage. The initiation and growth of these oscillations is principally dependent upon the subcooling at the core inlet: the greater the subcooling, the more likely oscillations will commence and increase in magnitude. Reactor shutdown on control rod insertion alone is the preferred way to ensure reactor stays shutdown under all conditions.

Initiating Cue:

Reactor scram required and all control rods are not full in with Reactor Power indicating

>5% on APRMs or cannot be determined.

Measurable Performance Standard:

Control rods are being inserted per PPM 5.5.11 (Attachment 6.1 - Tab B).

Performance Feedback:

Reactor Power as indicated on the APRMs / IRMs / SRMs is decreasing as control rods are being inserted.

Appendix D NRC SCENARIO OUTLINE SC-3 (Rev 0 - 9/13/2018) FORM ES-D-1 Columbia Generating Station CGS 2019 NRC Exam Scenario 3 Outline Page 7 of 7 CT-3 Critical Task Statement:

Initiate Emergency Depressurization (ED) by opening the first of seven (7) Safety Relief Valves (ADS preferred) before Wetwell level lowers to 19 feet, 2 inches. CT considered met if any combination of 7 Safety Relief Valves are opened within 10 minutes of Wetwell level lowering below 19 feet, 2inches.

Safety Significance:

Wetwell level must be maintained above the elevation of the downcomer vent openings to ensure that steam discharged from the drywell into the Wetwell following a primary system break will be adequately condensed. (Results of the Bodega Bay Mark I containment tests indicate 95% steam condensation may be achieved from a vertical downcomer vent that discharges at a level six inches above the suppression pool surface.) If Wetwell level cannot be maintained above the specified minimum value, steam may not be adequately condensed and primary containment pressure could exceed allowable limits. Since the RPV may not be kept at pressure when pressure suppression capability is unavailable, Emergency RPV Depressurization is required.

Initiating Cue:

Suppression Pool water level is approaching 19 ft 2 inches as indicated on CMS-LR-3 or CMS-LR-4.

Measurable Performance Standard:

7 Safety Relief Valves are manually opened.

Performance Feedback:

The valve light indications for each of the 7 Safety Relief Valves will change from Green lit to Red lit when control switch is taken to Open. Reactor pressure is lowering as indicated on MS-LR/PR-623A, MS-LR/PR-623B, or MS-PI-9.

Appendix D NRC SCENARIO OUTLINE SC-4 (Spare) (Rev 0 - 9/13/2018) FORM ES-D-1 Columbia Generating Station CGS 2019 NRC Exam Scenario 4 Outline Page 1 of 6 Facility:

Columbia Generating Station Scenario No.:

4 Op Test No.:

1 Examiners:

Operators:

Initial Conditions:

Reactor startup is in progress with power at 11%. Plant recently entered Mode 1. RPV pressure is

~960 psig with DEH in Auto with Turbine Bypass Valves ~28% open. TDRFP 'A' is in Auto on the Master Level Controller (MLC) with TDRFP B rolling at ~800 RPM. TSW-P-1B is out of service for the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to allow work on discharge check valve TSW-V-52B.

Turnover:

After turnover, withdraw control rods as required to raise reactor power to ~15%. Next in-sequence rod move is per RWM Page 36, Step 1 (Rod 02-35). Continuous rod withdrawal is permitted.

Critical Tasks:

CT-1 When a primary system is discharging into the secondary containment, manually scram the reactor before any area exceeds its Maximum Safe Operating Temperature (MSOT).

CT-2 When a primary system is discharging into the secondary containment, isolate the discharge before area temperatures exceed Maximum Safe Operating Temperature (MSOT) in more than one area.

NOTE: An unintentional or unnecessary RPS or ESF actuation may result in the creation of a post-scenario Critical Task, if that actuation results in a significant plant degradation or significantly alters a mitigation strategy.

Event No.

Trigger Event Type*

Event Description 1

N (ATC)

Withdraw control rods as required to raise reactor power to ~15%.

2 C (ATC/SRO)

Second (2nd) control rod withdrawn to notch 48 fails coupling check requiring full insertion of rod. (Tech Spec)

TS (SRO) 3 TRG-3 I (BOP,SRO)

RCIC-LS-15B (RCIC Suction Switchover CST Level Low) Failure - Auto switchover fails to occur (RCIC-V-31 opens but RCIC-V-10 fails to close).

(Tech Spec).

TS (SRO) 4 TRG-4 C (ATC,SRO)

APRM 'D' Fails Upscale requiring Manual Bypass.

5 TRG-5 C (BOP)

DEH-P-1B Fails. DEH-P-1A (the standby pump) does not auto start but can be manually started.

6 TRG-6 C (SRO,ATC)

Steam leak develops upstream of RCIC-V-8 (RCIC Turbine Steam Isolation) in TIP Mezzanine area in the Reactor Building. RCIC-V-63 (RCIC Steam Supply Inboard Isolation) fails to completely close requiring manual scram before TIP Mezzanine area exceeds its Maximum Safe Operating Temperature (MSOT).

(CT-1) 7 M (ALL)

OBE (Earthquake) results in piping rupture in Main Steam Tunnel. All inboard MSIVs fail to automatically close on high tunnel temperature. Manual closure required to isolate the rupture before exceeding 2nd MSOT. (CT-2)

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications

Appendix D NRC SCENARIO OUTLINE SC-4 (Spare) (Rev 0 - 9/13/2018) FORM ES-D-1 Columbia Generating Station CGS 2019 NRC Exam Scenario 4 Outline Page 2 of 6 Target Quantitative Attributes Actual Description Malfunctions after EOP entry (1-2) 2 RCIC-V-63 Fails Intermediate; Inboard MSIVs Fail to Auto Close Abnormal events (2-4) 4 Rod Uncoupled; RCIC-LS-15B Failure; APRM D Fails Upscale; DEH-P-1B Shaft Shear Major transients (1-2) 1 Steam Leak in Main Steam Tunnel EOPs entered/requiring substantive actions (1-2) 2 PPM 5.1.1 (RPV Control); PPM 5.3.1 (Secondary Containment Control)

Entry into a contingency EOP with substantive actions ( 1 per scenario set) 0

[NONE]

Pre-identified Critical tasks ( 2) 2 See Critical Task Sheets.

Appendix D NRC SCENARIO OUTLINE SC-4 (Spare) (Rev 0 - 9/13/2018) FORM ES-D-1 Columbia Generating Station CGS 2019 NRC Exam Scenario 4 Outline Page 3 of 6 SCENARIO 4

SUMMARY

Event 1 CR01 is tasked with withdrawing control rods as required to raise reactor power to ~15%. Rod movement is performed per SOP-CR-MOVEMENT. The first rod (02-35) will be withdrawn from notch 12 to full out (notch 48) followed by a coupling check, which passes. The second rod to be withdrawn (34-59) is also withdrawn from notch 12 to full out (notch 48) followed by a coupling check. This time the coupling check fails. (See Event 2).

Event 2 When the second (2nd) control rod (34-59) is withdrawn to notch 48 and a coupling check performed, annunciator P603.A7 1-8 (ROD OVERTRAVEL) will come in. In addition, the rod position indicator goes from showing notch 48 to a blank position (beyond notch 48). This is evidence of an uncoupled control rod. Annunciator Response Procedure (ARP) directs fully inserting affected control rod and disarming its HCU (Hydraulic Control Unit).

With control rod 34-59 Inoperable, the CRS refers to Technical Specifications and determines the following LCOs apply:

LCO 3.1.3 C.1 - Fully insert Inoperable control rod within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.

LCO 3.1.3 C.2 - Disarm the associated CRD within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Event 3 (TRG-3) A failure of level switch RCIC-LS-15B (RCIC Suction Switchover CST Level Low) results in an automatic shift of the RCIC (Reactor Core Isolation Cooling) suction from the Condensate Storage Tanks to the Wetwell.

Condensate Storage Tank (CST) levels are normal. During the automatic transfer, RCIC-V-31 (Wetwell suction) opens but RCIC-V-10 (Condensate Storage Tank suction) fails to automatically close. Operator action required to close RCIC-V-10 manually.

With one channel of RCIC Function 3 Inoperable (see LCO 3.3.5.2 Table 3.3.5.2-1), the CRS refers to Technical Specifications and determines the following LCOs applies:

LCO 3.3.5.2 A.1 - Enter the Condition referenced in Table 3.3.5.2-1 for the channel (Immediately).

LCO 3.3.5.2 D.2.1 - Place channel in trip within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> - OR -

LCO 3.3.5.2 D.2.2 - Align RCIC pump suction to the suppression pool within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Event 4 (TRG-4) Annunciators P603.A7 2-7 (ROD OUT BLOCK), P603.A8 2-6 (APRM UPSCALE), and P603-A8 1-6 (APRM UPSCL OR INOP TRIP) come in as a result of an APRM (Average Power Range Monitor) upscale failure.

Investigation of the ODAs (Operator Display Assemblies) at P603 or the APRM Chassis at P608 reveals APRM D upscale failure.

The CRS will direct APRM D channel bypassed and refer to LCO 3.3.1.1 (RPS Instrumentation). With 3 (three) required APRM channels remaining Operable, no further Tech Spec action is required (other than tracking the Inoperable APRM).

Event 5 (TRG-5) Annunciator P820-B1 6-5 (DEH PUMP DISCHARGE PRESS LOW) comes in as a result of a shaft shear on the running DEH (Digital Electro-Hydraulic) pump (DEH-P-1B). DEH discharge header pressure as read on the EHC MAIN DISPLAY starts trending down. As pressure lowers below 1800 psig, the standby EHC pump (DEH-P-1A) fails to automatically start.

Appendix D NRC SCENARIO OUTLINE SC-4 (Spare) (Rev 0 - 9/13/2018) FORM ES-D-1 Columbia Generating Station CGS 2019 NRC Exam Scenario 4 Outline Page 4 of 6 CRO may directly start the standby pump once EHC pressure lowers below 1800 psig by noting that the standby pump should have automatically started but didnt. The Annunciator Response Procedure (ARP) also directs starting the standby pump. DEH discharge header will be restored once the standby pump has been started.

Event 6 Note: The below annunciators are the main annunciators for the event. The ES-D-2 forms will describe the initiating annunciators in addition to those below.

(TRG-6) Steam leak develops upstream of RCIC-V-8 (RCIC Turbine Steam Isolation) in TIP Mezzanine area in the Reactor Building. Annunciators P601.A2 2-2 (LEAK DET RWCU ROOMS TEMP HI-HI), P601.A3 2-5 (LEAK DET RWCU ROOMS TEMP HI-HI), P601.A2 1-1 (LEAK DET RCIC PIPE ROUTING AREA TEMP HI-HI), and P601.A3 1-8 (LEAK DET RCIC PIPE ROUTING AREA TEMP HI-HI) come in when the TIP Mezzanine area temperature reaches 160°F.

The annunciators are accompanied by generated isolation signals which cause the RWCU (Reactor Water Cleanup) system to isolate by closing pump suctions (RWCU-V-1 and RWCU-V-4) and then tripping the RWCU pumps.

The same generated isolation signals attempt to isolate the RCIC (Reactor Core Isolation Cooling) system by closing RCIC-V-8 (RCIC Turbine Steam Supply Isolation) which does close and RCIC-V-63 (RCIC Steam Supply Line Inboard Isolation Valve) which does not close (valve goes intermediate and stops). Any manual attempt to close RCIC-V-63 is unsuccessful.

CRS enters ABN-HELB (High Energy Line Break) and PPM 5.3.1 (Secondary Containment Control).

Manual scram is required before the TIP Mezzanine area temperature reaches the Maximum Safe Operating Temperature (MSOT) of 212°F. (CT-1)

Event 7 Note: The below annunciators are the main annunciators for the event. The ES-D-2 forms will describe the initiating annunciators in addition to those below.

(TRG-7) Five (5) minutes after the manual scram, annunciators P851.S1 5-1 (OPERATING BASIS EARTHQUAKE) followed by P601.A2 3-1 (LEAK DET MSL TUNNEL TEMP HI-HI) and P601.A3 1-7 (LEAK DET MSL TUNNEL TEMP HI-HI) occur as a result of an earthquake causing a steam leak in the Main Steam Tunnel.

The annunciators are accompanied by generated isolation signals which attempt to isolate the Main Steam system by closing all four (4) Inboard MSIVs (MS-V-22A through D) and all four (4) Outboard MSIVs (MS-V-28A through D). All of the Outboard MSIVs automatically close but none of the Inboard MSIVs automatically close.

As a minimum, MS-V-22D has to be manually closed to isolate the leak. Once isolated, Main Steam Tunnel temperatures start trending down thereby preventing a second Maximum Safe Operating Temperature (MSOT).

(CT-2)

TERMINATION CRITERIA: The scenario will be terminated when a manual scram has been inserted and the main steam leak in the Main Steam Tunnel has been isolated OR as directed by the Lead Examiner.

Appendix D NRC SCENARIO OUTLINE SC-4 (Spare) (Rev 0 - 9/13/2018) FORM ES-D-1 Columbia Generating Station CGS 2019 NRC Exam Scenario 4 Outline Page 5 of 6 CT-1 Critical Task Statement:

When a primary system is discharging into the secondary containment, manually scram the reactor before any area exceeds its Maximum Safe Operating Temperature (MSOT).

Safety Significance:

If a primary system is discharging into the secondary containment, one of three conditions must exist:

A primary system break cannot be isolated because the system is required for damage control or must be operated in PPM 5.1.1 through PPM 5.2.1.

No isolation valves exist upstream of a primary system break, or if isolation valves do exist, they cannot be closed because of some mechanical / electrical /

pneumatic failure.

The source of the discharge cannot be determined.

If any temperature in any one of the areas listed in Table 23 approaches its MSOT, adequate core cooling, containment integrity, safety of personnel, or continued operability of equipment required to perform EOP flowchart actions can no longer be assured. PPM 5.1.1, RPV Control, must be entered to make certain the reactor is scrammed. Scramming the reactor reduces decay heat levels and the energy that the RPV may be discharging to the secondary containment.

Initiating Cue:

Primary system is discharging into the secondary containment and any area is approaching its MSOT.

Measurable Performance Standard:

Manual reactor scram initiated by placing the mode switch to Shutdown.

Performance Feedback:

Reactor shutdown as indicated by APRMs downscale.

Appendix D NRC SCENARIO OUTLINE SC-4 (Spare) (Rev 0 - 9/13/2018) FORM ES-D-1 Columbia Generating Station CGS 2019 NRC Exam Scenario 4 Outline Page 6 of 6 CT-2 Critical Task Statement:

When a primary system is discharging into the secondary containment, isolate the discharge before area temperatures exceed Maximum Safe Operating Temperature (MSOT) in more than one area.

Safety Significance:

Should secondary containment temperature continue to increase and exceed its MSOT in more than one area, the RPV must be Emergency Depressurized. To prevent an unwanted Emergency RPV depressurization, any discharge from the primary containment to secondary containment must be isolated before exceeding a second MSOT. Isolating the high energy system mitigates a direct and immediate threat to secondary containment integrity, protects equipment located in secondary containment, and protects personnel on and off the site.

Initiating Cue:

Primary system is discharging into the secondary containment and a second area is approaching its MSOT.

Measurable Performance Standard:

Leak from primary to secondary containment is isolated through operator action.

Performance Feedback:

Primary system stops discharging into secondary containment and associated secondary containment temperatures start to lower.