ML20324A039

CN-2020-09-DRAFT Outlines

ML20324A039
Person / Time
Site:	Cooper
Issue date:	09/29/2020
From:	Greg Werner Operations Branch IV
To:	Nebraska Public Power District (NPPD)
References
Download: ML20324A039 (75)

Text

ES-401 BWR Examination Outline Form ES-401-1 Facility: Cooper Nuclear Station Date of Exam: September 29, 2020 Tier Group RO K/A Category Points SRO-Only Points K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G* Total A2 G* Total

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

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

3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 3 3 2 2 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outline sections (i.e., except for one category in Tier 3 of the SRO-only section, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 radiation control K/A is allowed if it is replaced by a K/A from another Tier 3 category.)

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

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

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

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

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

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

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

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

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

G* Generic K/As

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

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

Rev. 11

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

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

295001 (APE 1) Partial or Complete Loss of Ability to determine and/or interpret the Forced Core Flow Circulation / 1 & 4 following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE X FLOW CIRCULATION: (CFR: 41.10 / 43.5 3.3 39

/ 45.13)

AA2.03 Actual core flow 295003 (APE 3) Partial or Complete Loss of Knowledge of the operational implications AC Power / 6 of the following concepts as they apply to X PARTIAL OR COMPLETE LOSS OF A.C. 3.1 49 POWER: (CFR: 41.8 to 41.10)

AK1.02 Load shedding 295004 (APE 4) Partial or Total Loss of DC Knowledge of the reasons for the following Power / 6 responses as they apply to PARTIAL OR X COMPLETE LOSS OF D.C. POWER: 3.1 56 (CFR: 41.5 / 45.6)

AK3.03 Reactor SCRAM 295005 (APE 5) Main Turbine Generator Trip / Ability to determine and/or interpret the 3 following as they apply to MAIN TURBINE X GENERATOR TRIP: (CFR: 41.10 / 43.5 / 3.7 51 45.13)

AA2.04 Reactor pressure 295006 (APE 6) Scram / 1 Knowledge of the reasons for the following responses as they apply to SCRAM: (CFR:

X 41.5 / 45.6) 3.8 54 AK3.01 Reactor water level response 295016 (APE 16) Control Room Abandonment Knowledge of the interrelations between

/7 CONTROL ROOM ABANDONMENT and X the following: (CFR: 41.7 / 45.8) 2.9* 44 AK2.03 Control room HVAC 295018 (APE 18) Partial or Complete Loss of Knowledge of the operational implications CCW / 8 of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF X COMPONENT COOLING WATER: (CFR: 3.5 42 41.8 to 41.10)

AK1.01 Effects on component/system operations 295019 (APE 19) Partial or Complete Loss of Ability to determine and/or interpret the Instrument Air / 8 following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR:

X (CFR: 41.10 / 43.5 / 45.13) 3.6 40 AA2.02 Status of safety-related instrument air system loads 295021 (APE 21) Loss of Shutdown Cooling / Knowledge of the operational implications 4 of the following concepts as they apply to X LOSS OF SHUTDOWN COOLING: (CFR: 3.6 48 41.8 to 41.10)

AK1.04 Natural circulation 295023 (APE 23) Refueling Accidents / 8 Ability to operate and/or monitor the following as they apply to REFUELING X ACCIDENTS: (CFR: 41.7 / 45.6) 3.3 53 AA1.01 Secondary containment ventilation 295024 High Drywell Pressure / 5 2.4.35 Knowledge of local auxiliary operator tasks during an emergency and X 3.8 58 the resultant operational effects. (CFR:

41.10 / 43.5 / 45.13)

Rev. 11

ES-401 3 Form ES-401-1 295025 (EPE 2) High Reactor Pressure / 3 Knowledge of the interrelations between HIGH REACTOR PRESSURE and the X following: (CFR: 41.7 / 45.8) 3.9 47 EK2.09 Reactor power 295026 (EPE 3) Suppression Pool High Water 2.1.30 Ability to locate and operate Temperature / 5 X components, including local controls. 4.4 41 (CFR: 41.7 / 45.7) 295027 (EPE 4) High Containment Temperature (Mark III Containment Only) / 5 295028 (EPE 5) High Drywell Temperature Ability to operate and/or monitor the (Mark I and Mark II only) / 5 following as they apply to HIGH DRYWELL X TEMPERATURE: (CFR: 41.7 / 45.6) 3.9 50 EA1.03 Drywell cooling system 295030 (EPE 7) Low Suppression Pool Water 2.1.25 Ability to interpret reference Level / 5 X materials, such as graphs, curves, tables, 3.9 43 etc. (CFR: 41.10 / 43.5 / 45.12) 295031 (EPE 8) Reactor Low Water Level / 2 Knowledge of the interrelations between REACTOR LOW WATER LEVEL and the X following: (CFR: 41.7 / 45.8) 4.2 46 EK2.03 Low pressure core spray 295037 (EPE 14) Scram Condition Present Ability to operate and/or monitor the and Reactor Power Above APRM Downscale following as they apply to SCRAM or Unknown / 1 CONDITION PRESENT AND REACTOR X POWER ABOVE APRM DOWNSCALE OR 4.1* 52 UNKNOWN: (CFR: 41.7 / 45.6)

EA1.06 Neutron monitoring system 295038 (EPE 15) High Offsite Radioactivity Knowledge of the reasons for the following Release Rate / 9 responses as they apply to HIGH OFF-X SITE RELEASE RATE: (CFR: 41.5 / 45.6) 3.7 55 EK3.03 Control room ventilation isolation 600000 (APE 24) Plant Fire On Site / 8 Knowledge of the reasons for the following responses as they apply to PLANT FIRE X ON SITE: 2.8 57 AK3.04 Actions contained in the abnormal procedure for plant fire on site 700000 (APE 25) Generator Voltage and Ability to operate and/or monitor the Electric Grid Disturbances / 6 following as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID X DISTURBANCES: (CFR: 41.5 and 41.10 / 3.6 45 45.5, 45.7, and 45.8 )

AA1.01 Grid frequency and voltage K/A Category Totals: 3 3 4 4 3 3 Group Point Total: 20 Rev. 11

ES-401 4 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant EvolutionsTier 1/Group 2 (RO/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 Knowledge of the reasons for the following responses as they apply to HIGH X REACTOR WATER LEVEL: (CFR: 41.5 / 3.6* 61 45.6)

AK3.02 Reactor SCRAM 295009 (APE 9) Low Reactor Water Level / 2 Knowledge of the reasons for the following responses as they apply to LOW X REACTOR WATER LEVEL: (CFR: 41.5 / 3.2 65 45.6)

AK3.01 Recirculation pump run back 295010 (APE 10) High Drywell Pressure / 5 295011 (APE 11) High Containment Temperature (Mark III Containment only) / 5 295012 (APE 12) High Drywell Temperature /

5 295013 (APE 13) High Suppression Pool Temperature. / 5 295014 (APE 14) Inadvertent Reactivity Addition / 1 295015 (APE 15) Incomplete Scram / 1 Ability to determine and/or interpret the following as they apply to INCOMPLETE X SCRAM: (CFR: 41.10 / 43.5 / 45.13) 4.1* 64 AA2.02 Control rod position 295017 (APE 17) Abnormal Offsite Release 2.4.50 Ability to verify system alarm Rate / 9 setpoints and operate controls identified in X 4.2 60 the alarm response manual. (CFR: 41.10 /

43.5 / 45.3) 295020 (APE 20) Inadvertent Containment Isolation / 5 & 7 295022 (APE 22) Loss of Control Rod Drive Ability to operate and/or monitor the Pumps / 1 following as they apply to LOSS OF CRD X PUMPS: (CFR: 41.7 / 45.6) 3.1 62 AA1.01 CRD hydraulic system 295029 (EPE 6) High Suppression Pool Water Level / 5 295032 (EPE 9) High Secondary Containment Area Temperature / 5 295033 (EPE 10) High Secondary Knowledge of the interrelations between Containment Area Radiation Levels / 9 HIGH SECONDARY CONTAINMENT X AREA RADIATION LEVELS and the 3.7 63 following: (CFR: 41.7 / 45.8)

EK2.03 Secondary containment ventilation 295034 (EPE 11) Secondary Containment Knowledge of the operational implications Ventilation High Radiation / 9 of the following concepts as they apply to SECONDARY CONTAINMENT X VENTILATION HIGH RADIATION: (CFR: 4.1 59 41.8 to 41.10)

EK1.02 Radiation releases 295035 (EPE 12) Secondary Containment High Differential Pressure / 5 Rev. 11

ES-401 5 Form ES-401-1 295036 (EPE 13) Secondary Containment High Sump/Area Water Level / 5 500000 (EPE 16) High Containment Hydrogen Concentration / 5 K/A Category Point Totals: 1 1 2 1 1 1 Group Point Total: 7 Rev. 11

ES-401 6 Form ES-401-1 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) 2.4.18 Knowledge of the specific bases for X 3.3 17 RHR/LPCI: Injection Mode EOPs. (CFR: 41.10 / 43.1 / 45.13) 205000 (SF4 SCS) Shutdown Cooling Ability to (a) predict the impacts of the following on the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE); and (b) based on those X predictions, use procedures to correct, 3.3 2 control, or mitigate the consequences of those abnormal conditions or operations:

(CFR: 41.5 / 45.6)

A2.08 Loss of heat exchanger cooling 206000 (SF2, SF4 HPCIS) Knowledge of the operational implications High-Pressure Coolant Injection of the following concepts as they apply to X HIGH PRESSURE COOLANT 2.8 24 INJECTION SYSTEM: (CFR: 41.5 / 45.3)

K5.02 Turbine shaft sealing 207000 (SF4 IC) Isolation (Emergency) Condenser 209001 (SF2, SF4 LPCS) Ability to predict and/or monitor changes Low-Pressure Core Spray in parameters associated with operating the LOW PRESSURE CORE SPRAY X SYSTEM controls including: (CFR: 41.5 / 3.4 22 45.5)

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

High-Pressure Core Spray 211000 (SF1 SLCS) Standby Liquid Ability to manually operate and/or monitor Control in the control room: (CFR: 41.7 / 45.5 to X 45.8) 4.2* 7 A4.08 System initiation 212000 (SF7 RPS) Reactor Protection 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 X consequences of those abnormal 4.0 19 conditions or operations: (CFR: 41.5 /

45.6)

A2.11 Main steamline isolation valve closure 215003 (SF7 IRM) Knowledge of the effect that a loss or Intermediate-Range Monitor malfunction of the following will have on X the INTERMEDIATE RANGE MONITOR 3.0 3 (IRM) SYSTEM: (CFR: 41.7 / 45.7)

K6.04 Detectors 215004 (SF7 SRMS) Source-Range Ability to manually operate and/or monitor Monitor in the control room: (CFR: 41.7 / 45.5 to X 45.8) 3.9 8 A4.01 SRM count rate and period 215005 (SF7 PRMS) Average Power Ability to predict and/or monitor changes Range Monitor/Local Power Range in parameters associated with operating Monitor the AVERAGE POWER RANGE X MONITOR/LOCAL POWER RANGE 3.3 26 MONITOR SYSTEM controls including:

(CFR: 41.5 / 45.5)

A1.05 Lights and alarms Rev. 11

ES-401 7 Form ES-401-1 217000 (SF2, SF4 RCIC) Reactor Knowledge of the effect that a loss or Core Isolation Cooling malfunction of the REACTOR CORE X ISOLATION COOLING SYSTEM (RCIC) 3.5 11 will have on following: (CFR: 41.7 / 45.4)

K3.03 Decay heat removal 218000 (SF3 ADS) Automatic Knowledge of the operational implications Depressurization of the following concepts as they apply to X AUTOMATIC DEPRESSURIZATION 3.8 16 SYSTEM: (CFR: 41.5 / 45.3)

K5.01 ADS logic operation 223002 (SF5 PCIS) Primary Knowledge of the effect that a loss or Containment Isolation/Nuclear Steam malfunction of the PRIMARY Supply Shutoff CONTAINMENT ISOLATION X SYSTEM/NUCLEAR STEAM SUPPLY 2.8 20 SHUT-OFF will have on following: (CFR:

41.7 / 45.4)

K3.06 Turbine building radiation 239002 (SF3 SRV) Safety Relief Knowledge of electrical power supplies to Valves X the following: (CFR: 41.7) 2.8* 23 K2.01 SRV solenoids 259002 (SF2 RWLCS) Reactor Water Knowledge of the physical connections Level Control and/or cause-effect relationships between REACTOR WATER LEVEL CONTROL X SYSTEM and the following: (CFR: 41.2 to 3.2 5 41.9 / 45.7 to 45.8)

K1.15 Recirculation flow control system 261000 (SF9 SGTS) Standby Gas Ability to monitor automatic operations of Treatment the STANDBY GAS TREATMENT X SYSTEM including: (CFR: 41.7 / 45.7) 3.2 13 A3.02 Fan start 262001 (SF6 AC) AC Electrical Ability to monitor automatic operations of Distribution the A.C. ELECTRICAL DISTRIBUTION X including: (CFR: 41.7 / 45.7) 3.1 15 A3.01 Breaker tripping 262002 (SF6 UPS) Uninterruptable Knowledge of the effect that a loss or Power Supply (AC/DC) malfunction of the following will have on X the UNINTERRUPTABLE POWER 2.8 21 SUPPLY (A.C./D.C.): (CFR: 41.7 / 45.7)

K6.02 D.C. electrical power 263000 (SF6 DC) DC Electrical Knowledge of D.C. ELECTRICAL Distribution DISTRIBUTION design feature(s) and/or interlocks which provide for the following:

X (CFR: 41.7) 3.1 12 K4.02 Breaker interlocks, permissives, bypasses and cross ties 264000 (SF6 EGE) Emergency Knowledge of EMERGENCY Generators (Diesel/Jet) EDG GENERATORS (DIESEL/JET) design feature(s) and/or interlocks which provide X for the following: (CFR: 41.7) 4.0 4 K4.02 Emergency generator trips emergency/LOCA) 300000 (SF8 IA) Instrument Air Knowledge of (INSTRUMENT AIR SYSTEM) design feature(s) and or interlocks which provide for the following:

X (CFR: 41.7) 2.8 14 K4.03 Securing of IAS upon loss of cooling water Rev. 11

ES-401 8 Form ES-401-1 400000 (SF8 CCS) Component Knowledge of the physical connections Cooling Water and/or cause-effect relationships between X CCWS and the following: (CFR: 41.2 to 3.2 18 41.9 / 45.7 to 45.8)

K1.01 Service water system 510000 (SF4 SWS*) Service Water (Normal and Emergency) 209001 (SF2, SF4 LPCS) Knowledge of the effect that a loss or LowPressure Core Spray malfunction of the following will have on X the LOW PRESSURE CORE SPRAY 2.8 10 SYSTEM: (CFR: 41.7 / 45.7)

K6.04 D.C. power 215003 (SF7 IRM) Knowledge of the operational implications IntermediateRange Monitor of the following concepts as they apply to X INTERMEDIATE RANGE MONITOR 3.0 6 (IRM) SYSTEM: (CFR: 41.5 / 45.3)

K5.03 Changing detector position 217000 (SF2, SF4 RCIC) Reactor Knowledge of electrical power supplies to Core Isolation Cooling X the following: (CFR: 41.7) 2.7* 25 K2.03 RCIC flow controller 218000 (SF3 ADS) Automatic 2.2.12 Knowledge of surveillance X 3.7 9 Depressurization procedures. (CFR: 41.10 / 45.13) 263000 (SF6 DC) DC Electrical Knowledge of the effect that a loss or Distribution malfunction of the D.C. ELECTRICAL DISTRIBUTION will have on following:

X (CFR: 41.7 / 45.4) 3.4 1 K3.03 Systems with D.C. components (i.e.

valves, motors, solenoids, etc.)

K/A Category Point Totals: 2 2 3 3 3 3 2 2 2 2 2 Group Point Total: 26 Rev. 11

ES-401 9 Form ES-401-1 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 Knowledge of the effect that a loss or malfunction of the following will have on the CONTROL ROD DRIVE X HYDRAULIC System: (CFR: 41.7 / 2.8 29 45.7)

K6.06 Component cooling water systems 201002 (SF1 RMCS) Reactor Manual Control 201003 (SF1 CRDM) Control Rod and Drive Ability to monitor automatic Mechanism operations of the CONTROL ROD X AND DRIVE MECHANISM including: 3.7 33 (CFR: 41.7 / 45.7)

A3.01 Control rod position 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 Knowledge of electrical power X supplies to the following: (CFR: 41.7) 2.8 32 K2.03 APRM channels 216000 (SF7 NBI) Nuclear Boiler Knowledge of the operational Instrumentation implications of the following concepts as they apply to NUCLEAR X BOILER INSTRUMENTATION: 3.2 27 (CFR: 41.5 / 45.3)

K5.12 Effects on level indication due to rapid changes in void fraction 219000 (SF5 RHR SPC) RHR/LPCI: Ability to manually operate and/or Torus/Suppression Pool Cooling Mode monitor in the control room: (CFR:

X 41.7 / 45.5 to 45.8) 3.8* 35 A4.01 Pumps 223001 (SF5 PCS) Primary Containment and Auxiliaries 226001 (SF5 RHR CSS) RHR/LPCI:

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

X TORUS/SUPPRESSION POOL 3.6 38 SPRAY MODE and the following:

(CFR: 41.2 to 41.9 / 45.7 to 45.8)

K1.01 Suppression pool 233000 (SF9 FPCCU) Fuel Pool Cooling/Cleanup 234000 (SF8 FH) Fuel-Handling Equipment 239001 (SF3, SF4 MRSS) Main and Reheat Steam Rev. 11

ES-401 10 Form ES-401-1 239003 (SF9 MSVLCS) Main Steam Isolation Valve Leakage Control 241000 (SF3 RTPRS) Reactor/Turbine Knowledge of the effect that a loss Pressure Regulating or malfunction of the REACTOR/TURBINE PRESSURE X REGULATING SYSTEM will have on 2.8 37 following: (CFR: 41.7 / 45.4)

K3.23 Turbine trip testing 245000 (SF4 MTGEN) Main Turbine Generator/Auxiliary 256000 (SF2 CDS) Condensate 2.1.23 Ability to perform specific system and integrated plant X procedures during all modes of plant 4.3 30 operation. (CFR: 41.10 / 43.5 / 45.2 /

45.6) 259001 (SF2 FWS) Feedwater Ability to predict and/or monitor changes in parameters associated with operating the REACTOR X FEEDWATER SYSTEM controls 2.7 34 including: (CFR: 41.5 / 45.5)

A1.06 Feedwater heater level 268000 (SF9 RW) Radwaste 271000 (SF9 OG) Offgas 272000 (SF7, SF9 RMS) Radiation Monitoring 286000 (SF8 FPS) Fire Protection Knowledge of FIRE PROTECTION SYSTEM design feature(s) and/or interlocks which provide for the following: (CFR: 41.5 / 41.7 / 45.3 /

X 45.5 to 45.8) 3.6 36 K4.04 Personnel safety during halon and/or carbon dioxide system actuation 288000 (SF9 PVS) Plant Ventilation Knowledge of the effect that a loss or malfunction of the PLANT X VENTILATION SYSTEMS will have 2.9 28 on following: (CFR: 41.5 / 45.3)

K3.02 Reactor building temperature 290001 (SF5 SC) Secondary Containment 290003 (SF9 CRV) Control Room Ventilation 290002 (SF4 RVI) Reactor Vessel Internals Ability to (a) predict the impacts of the following on the REACTOR VESSEL INTERNALS; and (b) based on those predictions, use procedures to correct, control, or X mitigate the consequences of those 3.7 31 abnormal conditions or operations:

(CFR: 41.5 / 45.6)

A2.04 Excessive heatup/cooldown rate 51001 (SF8 CWS*) Circulating Water K/A Category Point Totals: 1 1 2 1 1 1 1 1 1 1 1 Group Point Total: 12 Rev. 11

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Cooper Nuclear Station Date of Exam:

Category K/A # Topic RO SRO-only IR # IR #

Knowledge of conduct of operations requirements. (CFR: 41.10 2.1.1 / 45.13) 3.8 69 Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, no-2.1.4 solo operation, maintenance of active license status, 3.3 70 10CFR55, etc. (CFR: 41.10 / 43.2)

Knowledge of industrial safety procedures (such as rotating 2.1.26 equipment, electrical, high temperature, high pressure, caustic, 3.4 67 chlorine, oxygen and hydrogen). (CFR: 41.10 / 45.12)

Subtotal 3 Ability to perform pre-startup procedures for the facility, 2.2.1 including operating those controls associated with plant equipment that could affect reactivity. (CFR: 41.5 / 41.10 / 43.5 4.5 73

/ 43.6 / 45.1)

2. Equipment Ability to track Technical Specification limiting conditions for 2.2.23 3.1 72 Control operations. (CFR: 41.10 / 43.2 / 45.13) 2.2.38 Knowledge of conditions and limitations in the facility license.

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

Subtotal 3 2.3.12 Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked 3.2 71 high-radiation areas, aligning filters, etc. (CFR: 41.12 / 45.9 /

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

Subtotal 2 2.4.5 Knowledge of the organization of the operating procedures network for normal, abnormal, and emergency evolutions. 3.7 75

4. Emergency (CFR: 41.10 / 43.5 / 45.13)

Procedures/Plan 2.4.27 Knowledge of fire in the plant procedures. (CFR: 41.10 / 43.5 3.4 74

/ 45.13)

Subtotal 2 Tier 3 Point Total 10 7 Rev. 11

ES-401 BWR Examination Outline Form ES-401-1 Facility: Cooper Nuclear Station Date of Exam: September 29, 2020 Tier Group RO K/A Category Points SRO-Only Points K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G* Total A2 G* Total

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

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

3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 2 2 1 2 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outline sections (i.e., except for one category in Tier 3 of the SRO-only section, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 radiation control K/A is allowed if it is replaced by a K/A from another Tier 3 category.)

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

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

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

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

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

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

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

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

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

G* Generic K/As

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

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

Rev. 11

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

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

295001 (APE 1) Partial or Complete Loss of Forced Core Flow Circulation / 1 & 4 295003 (APE 3) Partial or Complete Loss of AC Power / 6 295004 (APE 4) Partial or Total Loss of DC Ability to determine and/or interpret the Power / 6 following as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER:

X (CFR: 41.10 / 43.5 / 45.13) 3.6 77 AA2.01 Cause of partial or complete loss of D.C. power 295005 (APE 5) Main Turbine Generator Trip / 2.1.20 Ability to interpret and execute 3 X procedure steps. (CFR: 41.10 / 43.5 / 4.6 82 45.12) 295006 (APE 6) Scram / 1 2.1.7 Ability to evaluate plant performance and make operational judgments based on X operating characteristics, reactor behavior, 4.7 79 and instrument interpretation. (CFR: 41.5 /

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

/7 295018 (APE 18) Partial or Complete Loss of Ability to determine and/or interpret the CCW / 8 following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT X COOLING WATER: (CFR: 41.10 / 43.5 / 3.4 78 45.13)

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

4 295023 (APE 23) Refueling Accidents / 8 295024 High Drywell Pressure / 5 295025 (EPE 2) High Reactor Pressure / 3 295026 (EPE 3) Suppression Pool High Water 2.4.41 Knowledge of the emergency action Temperature / 5 X level thresholds and classifications. (CFR: 4.6 80 41.10 / 43.5 / 45.11) 295027 (EPE 4) High Containment Temperature (Mark III Containment Only) / 5 295028 (EPE 5) High Drywell Temperature 2.2.12 Knowledge of surveillance X procedures. (CFR: 41.10 / 45.13) 4.1 76 (Mark I and Mark II only) / 5 295030 (EPE 7) Low Suppression Pool Water Level / 5 295031 (EPE 8) Reactor Low Water Level / 2 Ability to determine and/or interpret the following as they apply to REACTOR LOW X WATER LEVEL: (CFR: 41.10 / 43.5 / 4.6 81 45.13)

EA2.01 Reactor water level 295037 (EPE 14) Scram Condition Present and Reactor Power Above APRM Downscale or Unknown / 1 295038 (EPE 15) High Offsite Radioactivity Release Rate / 9 600000 (APE 24) Plant Fire On Site / 8 700000 (APE 25) Generator Voltage and Electric Grid Disturbances / 6 K/A Category Totals: 3 4 Group Point Total: 7 Rev. 11

ES-401 3 Form ES-401-1 Rev. 11

ES-401 4 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant EvolutionsTier 1/Group 2 (RO/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 Ability to determine and/or interpret the Temperature. / 5 following as they apply to HIGH X SUPPRESSION POOL TEMPERATURE: 3.5 85 (CFR: 41.10 / 43.5 / 45.13)

AA2.02 Localized heating/stratification 295014 (APE 14) Inadvertent Reactivity Addition / 1 295015 (APE 15) Incomplete Scram / 1 295017 (APE 17) Abnormal Offsite Release Rate / 9 295020 (APE 20) Inadvertent Containment 2.4.6 Knowledge of EOP mitigation X strategies. (CFR: 41.10 / 43.5 / 45.13 4.7 83 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 Ability to determine and/or interpret the High Differential Pressure / 5 following as they apply to SECONDARY X CONTAINMENT HIGH DIFFERENTIAL 3.9 84 PRESSURE: (CFR: 41.8 to 41.10)

EA2.01 Secondary containment pressure 295036 (EPE 13) Secondary Containment High Sump/Area Water Level / 5 500000 (EPE 16) High Containment Hydrogen Concentration / 5 K/A Category Point Totals: 2 1 Group Point Total: 3 Rev. 11

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

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

203000 (SF2, SF4 RHR/LPCI)

RHR/LPCI: Injection Mode 205000 (SF4 SCS) Shutdown Cooling 2.4.4 Ability to recognize abnormal indications for system operating X parameters that are entry-level conditions 4.7 87 for emergency and abnormal operating procedures. (CFR: 41.10 / 43.2 / 45.6) 206000 (SF2, SF4 HPCIS)

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

Low-Pressure Core Spray 209002 (SF2, SF4 HPCS)

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

Intermediate-Range Monitor 215004 (SF7 SRMS) Source-Range 2.1.32 Ability to explain and apply system Monitor X limits and precautions. (CFR: 41.10 / 43.2 4.0 89

/ 45.12) 215005 (SF7 PRMS) Average Power Range Monitor/Local Power Range Monitor 217000 (SF2, SF4 RCIC) Reactor Core Isolation Cooling 218000 (SF3 ADS) Automatic Depressurization 223002 (SF5 PCIS) Primary Containment Isolation/Nuclear Steam Supply Shutoff 239002 (SF3 SRV) Safety Relief Valves 259002 (SF2 RWLCS) Reactor Water Level Control 261000 (SF9 SGTS) Standby Gas Treatment 262001 (SF6 AC) AC Electrical Ability to (a) predict the impacts of the Distribution following on the A.C. ELECTRICAL DISTRIBUTION; and (b) based on those predictions, use procedures to correct, X control, or mitigate the consequences of 4.2 86 those abnormal conditions or operations:

(CFR: 41.5 / 45.6)

A2.04 Types of loads that, if deenergized, would degrade or hinder plant operation 262002 (SF6 UPS) Uninterruptable Power Supply (AC/DC) 263000 (SF6 DC) DC Electrical Distribution 264000 (SF6 EGE) Emergency Generators (Diesel/Jet) EDG Rev. 11

ES-401 6 Form ES-401-1 300000 (SF8 IA) Instrument Air Ability to (a) predict the impacts of the following on the INSTRUMENT AIR SYSTEM and (b) based on those predictions, use procedures to correct, X control, or mitigate the consequences of 2.8 90 those abnormal operation: (CFR: 41.5 /

45.6)

A2.01 Air dryer and filter malfunctions 400000 (SF8 CCS) Component 2.4.2 Knowledge of system set points, Cooling Water interlocks and automatic actions X associated with EOP entry conditions. 4.6 88 (CFR: 41.7 / 45.7 / 45.8) 510000 (SF4 SWS*) Service Water (Normal and Emergency)

K/A Category Point Totals: 2 3 Group Point Total: 5 Rev. 11

ES-401 7 Form ES-401-1 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 2.4.6 Knowledge of EOP mitigation X strategies. (CFR: 41.10 / 43.5 / 45.13) 4.7 93 202002 (SF1 RSCTL) Recirculation Flow Control 204000 (SF2 RWCU) Reactor Water Cleanup Ability to (a) predict the impacts of the following on the REACTOR WATER CLEANUP SYSTEM; and (b) based on those predictions, use procedures to X correct, control, or mitigate the 2.8 91 consequences of those abnormal conditions or operations: (CFR: 41.5 /

45.6)

A2.10 Valve closures 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 Ability to (a) predict the impacts of the Spray Mode following on the RHR/LPCI:

CONTAINMENT SPRAY SYSTEM MODE; and (b) based on those X predictions, use procedures to correct, 3.0 92 control, or mitigate the consequences of those abnormal conditions or operations:

(CFR: 41.5 / 45.6)

A2.11 Motor operated valve failures 230000 (SF5 RHR SPS) RHR/LPCI:

Torus/Suppression Pool Spray Mode 233000 (SF9 FPCCU) Fuel Pool Cooling/Cleanup 234000 (SF8 FH) Fuel-Handling Equipment 239001 (SF3, SF4 MRSS) Main and Reheat Steam 239003 (SF9 MSVLCS) Main Steam Isolation Valve Leakage Control 241000 (SF3 RTPRS) Reactor/Turbine Pressure Regulating 245000 (SF4 MTGEN) Main Turbine Generator/Auxiliary 256000 (SF2 CDS) Condensate 259001 (SF2 FWS) Feedwater 268000 (SF9 RW) Radwaste 271000 (SF9 OG) Offgas 272000 (SF7, SF9 RMS) Radiation Monitoring 286000 (SF8 FPS) Fire Protection 288000 (SF9 PVS) Plant Ventilation Rev. 11

ES-401 8 Form ES-401-1 290001 (SF5 SC) Secondary Containment 290003 (SF9 CRV) Control Room Ventilation 290002 (SF4 RVI) Reactor Vessel Internals 51001 (SF8 CWS*) Circulating Water K/A Category Point Totals: 2 1 Group Point Total: 3 Rev. 11

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Cooper Nuclear Station Date of Exam:

Category K/A # Topic RO SRO-only IR # IR #

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

3.9 100 41.10 / 43.7)

1. Conduct of Operations 2.1.36 Knowledge of procedures and limitations involved in core 4.1 98 alterations. (CFR: 41.10 / 43.6 / 45.7)

Subtotal 2 2.2.21 Knowledge of pre- and post-maintenance operability 4.1 94 requirements. (CFR: 41.10 / 43.2)

2. Equipment Control 2.2.37 Ability to determine operability and/or availability of safety 4.6 96 related equipment. (CFR: 41.7 / 43.5 / 45.12)

Subtotal 2 2.3.6 Ability to approve release permits. (CFR: 41.13 / 43.4 / 45.10) 3.8 97

3. Radiation Control Subtotal 1 2.4.47 Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference 4.2 95

4. Emergency material. (CFR: 41.10 / 43.5 / 45.12)

Procedures/Plan 2.4.37 Knowledge of the lines of authority during implementation of 4.1 99 the emergency plan. (CFR: 41.10 / 45.13)

Subtotal 2 Tier 3 Point Total 7 Rev. 11

ES-401 Record of Rejected K/As Form ES-401-4 Tier / Randomly Reason for Rejection Group Selected K/A No rejected K/As on initial draft Rev. 11

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Cooper Nuclear Station Date of Examination: 9/21/2020 Examination Level: RO SRO Operating Test Number: CN-2020-9 Administrative Topic (see Note) Type Describe activity to be performed Code*

A1 Determine time to 200 F in the Spent Fuel Pool in accordance with 2.4FPC.

Conduct of Operations N/A K/A G2.1.25 (4.4) Ability to interpret reference materials, such as graphs, curves, tables, etc.

A2 Recognized a mis-positioned rod and the required actions Conduct of Operations R,N K/A G2.1.37 (4.3) Knowledge of procedures, guidelines, or limitations associated with reactivity management.

A3 Determine impact of pulling fuse Equipment Control K/A G2.2.15 (3.9) Ability to determine the R, N expected plant configuration using design and configuration control documentation, such as drawings, line-ups, tagouts, etc.

A4 Release Rate Determination (SJAE).

Radiation Control R,N K/A G2.3.11 (4.3) Ability to control radiation releases.

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

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

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

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

Rev 1

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Cooper Nuclear Station Date of Examination: 9/21/2020 Examination Level: RO SRO Operating Test Number: CN-2020-9 Administrative Topic (see Note) Type Describe activity to be performed Code*

A5 Determine instrumentation considered available while a fire is active in a location Conduct of Operations R, N in the reactor building K/A G2.4.3 (3.7) Ability to identify post-accident instrumentation.

A6 Review procedure 2.0.4 for Conduct of Operations discrepancies prior to shift turnover.

R, N K/A G2.1.3 (3.9) Knowledge of shift or short term relief turnover practices.

A7 Determine post maintenance testing for CREFS valve accumulator replacement Equipment Control R, N (HV-AOV-271)

K/A G 2.2.21 (4.1) Knowledge of pre-post-maintenance operability requirements .

A8 Determine additional requirements for drywell entry while tip maintenance is in Radiation Control progress.

K/A 2.3.12 (3.7) Knowledge of radiological R, N safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

A9 Determine Protective Action Recommendations Tabletop #11 (Shelter Emergency Plan R, D PAR).

K/A G2.4.44 (4.4) Knowledge of emergency plan protective action recommendations.

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).

Rev 1

ES-301 Administrative Topics Outline Form ES-301-1

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

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

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

Rev 1

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Cooper Nuclear Station Date of Examination: 9/21//2020 Exam Level: RO SRO-I SRO-U Operating Test Number: CN-2020-9 Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Code* Safety Function S1 Shifting CRD pumps, both trip and require a scram A, L, N, S 1 K/A 201001 A4.01 (3.8/3.9) Panel 9-5 S2 Quick restart of RFP, Thrust Bearing High pressure failure, Trip A, L, N, S 2 RFP K/A 259001 A3.01 (3.3/3.5) Panel A S3 Defeat Group 1 Low Level Isolation during an ATWS L,D,S 3 (Restoration)

K/A 239001 A1.07 (3.7/3.7)

S4 Perform emergency depressurization with Main Steam Drains L, N, S 4 K/A 239001 A4.02 (3.2/3.2) Panel 9-4 S5 Perform 5.2FUEL Att. 2 (Actions to minimize rad release) S, P. L 5 K/A 290001 A4.10 (3.4/3.3) Back Panels S6 start DG2 and just prior to connecting to bus high vibrations alarm A,N,S,EN 6 comes in and requires tripping.

K/A 264000 K4.01 (3.5/3.7) Panel C S7 Verify and reset a Group 2 Primary Containment Isolation A, L,D,S 7 K/A 215001 A2.07 (3.4/3.7) Back Panels S8. Align SW crosstie to REC IAW Emergency Procedure 5.2REC P, L, EN 8 Att. 6 K/A 400000 A4.01 (3.1/3.0) Back Panels page 1 of 2 Rev 1

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U P1 Transfer PMIS UPS static switch from inverter to alternate E, R, N 6 source K/A 262002 A4.01 (3.1/3.4)

P2 Startup RPS MG Set A (Alternate Path)

EN, D, A, 7 K/A 212000 A2.01 (3.7/3.9)

P3 Start FP-P-D using 2.2.30 section 10. The initial try in manual E, M 8 will fail due to bad battery, second attempt works, one value when warmed up will be out of spec requiring informing the CRS.

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

• All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions, all five SRO-U systems must serve different safety functions, and in-plant systems and functions may overlap those tested in the control room.

• Type Codes Criteria for R /SRO-I/SRO-U (Actual)

(A)lternate path 4-6/4-6 /2-3 (5/5/3)

(C)ontrol room (D)irect from bank 9/ 8/ 4 (5/5/3)

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

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

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

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

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

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

(S)imulator page 2 of 2 Rev 1

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Cooper Nuclear Station Date of Exam: 9/21/2020 Operating Test No.: 2020-09 A E Scenarios P V 4 3 2 1 T M P E O I L N CREW CREW CREW POSITION CREW POSITION T N I T POSITION POSITION I

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

T P E R I U RO RX 0 0 0 0 1 1 0 NOR 1 1 0 2 1 1 1 SRO-I (I1) I/C 7 5 3 15 4 4 2 SRO-U MAJ 1 1 2 4 2 2 1 TS 2 N/A N/A 2 0 2 2 RO RX 0 1 0 1 1 1 0 NOR 0 1 1 2 1 1 1 SRO-I (I2) I/C 4 7 4 15 4 4 2 SRO-U MAJ 1 1 2 4 2 2 1 TS N/A 2 N/A 2 0 2 2 RO RX 0 1 0 1 1 1 0 NOR 1 0 1 2 1 1 1 SRO-I (I3) I/C 3 4 6 13 4 4 2 SRO-U MAJ 1 1 2 4 2 2 1 TS N/A N/A 2 2 0 2 2 RO RX 1 1 0 NOR 1 1 1 SRO-I I/C 4 4 2 SRO-U MAJ 2 2 1 TS 0 2 2 Instructions:

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

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

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

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

Rev 0

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Cooper Nuclear Station Date of Exam: 9/21/2020 Operating Test No.: 2020-9 A E Scenarios P V 4 3 2 1 T M P E O I L N CREW CREW POSITION CREW POSITION CREW POSITION T N I T POSITION I

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

T P E R I U RO RX 0 0 0 0 1 1 0 NOR 1 1 0 2 1 1 1 SRO-I (I4) I/C 7 5 3 15 4 4 2 SRO-U MAJ 1 1 2 4 2 2 1 TS 2 N/A N/A 2 0 2 2 RO RX 0 1 0 1 1 1 0 NOR 0 1 1 2 1 1 1 SRO-I (I5) I/C 4 7 4 15 4 4 2 SRO-U MAJ 1 1 2 4 2 2 1 TS N/A 2 N/A 2 0 2 2 RO RX 0 1 0 1 1 1 0 NOR 1 0 1 2 1 1 1 SRO-I (I6 I/C 3 4 6 13 4 4 2 SRO-U MAJ 1 1 2 4 2 2 1 TS N/A N/A 2 2 0 2 2 RO RX 1 1 0 NOR 1 1 1 SRO-I I/C 4 4 2 SRO-U MAJ 2 2 1 TS 0 2 2 Instructions:

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

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

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

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

Rev 0

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Cooper Nuclear Station Date of Exam: 9/21/2020 Operating Test No.: 2020-9 A E Scenarios P V 4 3 2 1 T M P E O I L N CREW CREW POSITION CREW POSITION CREW POSITION T N I T POSITION I

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

T P E R I U RO RX 0 1 1 1 1 0 NOR 1 1 2 1 1 1 SRO-I I/C 7 7 14 4 4 2 SRO-U MAJ 1 1 2 2 2 1 (U1)

TS 2 2 4 0 2 2 RO RX 0 0 0 1 1 0 (R1) NOR 0 1 1 1 1 1 SRO-I I/C 4 5 9 4 4 2 SRO-U MAJ 1 1 2 2 2 1 TS N/A N/A 0 0 2 2 RO RX 0 1 1 1 1 0 (R2)

NOR 1 0 1 1 1 1 SRO-I I/C 3 4 7 4 4 2 SRO-U MAJ 1 1 2 2 2 1 TS N/A N/A 0 0 2 2 RO RX 1 1 0 NOR 1 1 1 SRO-I I/C 4 4 2 SRO-U MAJ 2 2 1 TS 0 2 2 Instructions:

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

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

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

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

Rev 0

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Cooper Nuclear Station Date of Exam: 9/21/2020 Operating Test No.: 2020-9 A E Scenarios P V 4 3 2 1 T M P E O I L N CREW CREW POSITION CREW POSITION CREW POSITION T N I T POSITION I

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

T P E R I U RO RX 1 1 0 1 1 0 NOR 1 0 1 1 1 1 SRO-I I/C 7 3 4 4 4 2 SRO-U MAJ 1 1 1 2 2 1 TS 1 N/A N/A 0 2 2 RO RX 1 1 0 NOR 1 1 1 SRO-I I/C 4 4 2 SRO-U MAJ 2 2 1 TS 0 2 2 RO RX 1 1 0 NOR 1 1 1 SRO-I I/C 4 4 2 SRO-U MAJ 2 2 1 TS 0 2 2 RO RX 1 1 0 NOR 1 1 1 SRO-I I/C 4 4 2 SRO-U MAJ 2 2 1 TS 0 2 2 Instructions:

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

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

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

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

Rev 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9/2020 Scenario 1 Page 1 of 12 Facility: Cooper Nuclear Station Scenario No.: 1 Op-Test No.: CN-2020-09 Examiners: ____________________________ Operators: _____________________________

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

1. Shift REC pumps from A in operation to B in operation

2. Respond to failure of REC Pump B

3. Raise power with Reactor Recirculation from 95% to 97%

4. Respond to Minor Earthquake and Gland seal exhauster tripping

5. Respond to Reactor Recirc Pump A impeller damage resulting in fuel failure

6. Respond to Major Earthquake causing RWCU Pump A leak in secondary containment

7. Respond to failure of Group 6 to automatically isolate/initiate

8. Respond to failure of group 3 to automatically isolate and loss of power to RWCU isolation valves

9. Respond to RFPT B governor valve failure.

10. Respond to HPCI controller failing low in auto Initial Conditions: Plant operating at 95% power near end of cycle.

Inoperable Equipment:

Turnover:

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

Planned activities for this shift are:

• Shift REC pumps

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

This is a new scenario.

Validation Time: XXX minutes Page 1 of 12 Rev 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9/2020 Scenario 1 Page 2 of 12 Event Malf. No. Event Type Event No. Description N (BOP,CRS) Shift REC pumps from A in operation to B in 1 N/A TS (CRS) operation (LCO 3.7.3)

Failure of REC Pump B (secure REC Pump 2 N/A C (BOP,CRS)

B) 3 N/A R (ATC,CRS) Raise power from 95% to 100%

hv02a (30)

C (BOP) Minor earthquake, Gland Seal Exhauster B 4 zdiarswgseb A (CREW) trip (start Gland Seal Exhauster A)

(STOP) rr50a (9)

C (ATC,CRS) Reactor Recirc Pump A impeller damage 5 cr01 (30) resulting in fuel failure (lower power IAW A (CREW) procedure 5.2FUEL) cr02 (40)

Major earthquake, RWCU Pump A leak into secondary containment CT#1 hv02b When a primary system is discharging 6 M (CREW) into the secondary containment through a cu01a break and cannot be isolated, the crew scrams the reactor prior to reaching Maximum Safe Operating radiation level in two areas.

Group 6 fails to automatically initiate/isolate rm02g (67.75) (initiate/isolate Group 6 manually) rm02h (67.5) CT#2 7 C (ATC,CRS) When Group 6 fails to automatically rm02p (68) initiate/isolate due to high radiation, manually initiate Group 6 isolation within rm02q (67.65) 90 after Reactor Building Exhaust Plenum radiation levels exceed 49 mR/hr.

Group 3 fails to isolate, RWCU-MO-15 and RWCU-MO-18 loss of power. (direct resetting rp12 breakers and manually close RWCU-MO-15 o/r zdipcissws15 and/or MO-18 from Panel 9-4) o/r zdipcissws16 CT#3 When a primary system is discharging 8 o/r zlopcissws15(1) M (CREW) into the secondary containment through a break and automatic isolation fails, the o/r zlopcissws15(2) crew takes manual action from the control o/r zlopcissws16(1) room which effects isolation of the system prior to reaching two Maximum Safe o/r zlopcissws16(2) Operating radiation values in two areas.

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Appendix D Scenario Outline Form ES-D-1 NRC CNS 9/2020 Scenario 1 Page 3 of 12 RFPT B governor valve fails closed (place 9 fw243 C (ATC)

RFP A in service) 10 hp04 (0) C (ATC) HPCI controller fails low in auto (manually control HPCI turbine speed)

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

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

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Appendix D Scenario Outline Form ES-D-1 NRC CNS 9/2020 Scenario 1 Page 4 of 12 Quantitative Attributes Table ES-301-4 Actual Attribute Description Target

1. Group 6 fails to automatically initiate/isolate

2. Group 3 fails to isolate, RWCU-MO-15 and RWCU-Malfunctions after MO-18 loss of power 1-2 4 EOP entry 3. RFPT B governor valve fails closed

4. HPCI controller fails low in auto

1. Minor earthquake Abnormal Events 2-4 2. Reactor Recirc Pump A impeller damage resulting in fuel failure Major Transients 1-2 2 1. RWCU leak into secondary containment

1. EOP-1A EOP entries requiring 2. EOP-3A 1-2 3 3. EOP-5A substantive action EOP contingencies requiring 1 per set 0 1. None substantive action

1. (CT#1) When a primary system is discharging into the secondary containment through a break and cannot be isolated, the crew scrams the reactor prior to reaching Maximum Safe Operating radiation level in two areas.

2. (CT #2) When Group 6 fails to automatically initiate/isolate due to high radiation, manually initiate Group 6 isolation/initiation within 90 after Pre-identified Reactor Building Exhaust Plenum radiation levels 2 3 Critical Tasks exceed 49 mR/hr.

3. (CT #3) When a primary system is discharging into the secondary containment through a break and automatic isolation fails, the crew takes manual action from the control room which effects isolation of the system prior to reaching two Maximum Safe Operating radiation values in two areas.

Normal Events N/A 1 1. Shift REC pumps Reactivity 1. Raise power from 95% to 100%

N/A 1 Manipulations Page 4 of 12 Rev 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9/2020 Scenario 1 Page 5 of 12

1. REC Pump B failure

2. Gland Seal Exhauster B trip

3. Fuel failure Instrument/ 4. Group 6 fails to automatically initiate/isolate Component N/A 7 Failures 5. Group 3 fails to isolate, RWCU-MO-15 and RWCU-MO-18 loss of power

6. RFPT B governor valve fails closed

7. HPCI controller fails low in auto

1. REC Pump B failure

2. Gland Seal Exhauster B trip

3. Fuel failure

4. Group 6 fails to automatically initiate/isolate Total N/A 7 5. Group 3 fails to isolate, RWCU-MO-15 and Malfunctions RWCU-MO-18 loss of power

6. RFPT B governor valve fails closed

7. HPCI controller fails low in auto Top 10 systems and operator actions important to risk that are tested:

• Primary Containment (Isolation), HPCI Page 5 of 12 Rev 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9/2020 Scenario 1 Page 6 of 12 SCENARIO

SUMMARY

The plant is operating at 95% power near the end of the operating cycle.

After the crew takes the watch, the BOP will shift REC pumps from A in operation to B in operation IAW Procedure 2.2.65.1. After the oncoming REC Pump B is started, the field report will indicate that the pump is making very loud noises. The crew will secure REC Pump B. The CRS will enter LCO 3.7.3 condition B.

After TS have been addressed for REC Pump B, the crew will raise power with Reactor Recirculation. When power reaches 97%, a minor earthquake will occur. This will cause the running gland seal exhauster to trip. The crew will take alarm card actions and start the standby gland exhauster. Four minutes after minor earthquake, high vibration will momentarily alarm Reactor Recirc Pump A due to slight impeller failure caused by the earthquake. The impeller failure will cause fuel damage, resulting in rising plant radiation levels.

A major earthquake will cause the breakers for isolation valves RWCU-MO-15 to trip. It will also cause a RWCU Pump A leak into secondary containment.

Secondary containment area radiation levels will continue to rise. The crew must insert a scram when RWCU Pump Room reaches the Maximum Safe Operating (MSO) radiation level before a second area reaches the MSO radiation level. (CT#1). One minute after the scram, the preferred RFP (B) will decrease to minimum speed requiring the crew to place RFP A in service.

Radiation level in Reactor Building Exhaust Plenum will rise to the HIGH-HIGH setpoint (10 mR/hr, actual), but a Group 6 isolation will not occur. The crew must take action to insert a manual Group 6 isolation/initiation before RB Exhaust Plenum radiation exceeds 49 mR/hr (TS allowable value) for 90 seconds (CT#2).

IAW with Procedure 5.2 fuel, the crew will close the MSIVs and shift level/pressure control to HPCI/RCIC/SRVs. HPCI controller will fail in auto but can be controlled in manual.

In order to avert emergency depressurization, the crew will direct the building operator to restore power to RWCU valves, then the crew will close RCIC-MO-15 and RCIC-MO18 to isolate the leak prior to a second area exceeding MSO radiation level (CT#3).

The exercise ends when RWCU is isolated and reactor water level and pressure are being controlled in the assigned bands.

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Appendix D Scenario Outline Form ES-D-1 NRC CNS 9/2020 Scenario 1 Page 7 of 12 CRITICAL TASK BASIS When a primary system is discharging into the secondary containment through a break and cannot be isolated, the Critical Task #1 crew scrams the reactor prior to reaching Maximum Safe Operating radiation level in two areas.

EOP-5A directs entering the RPV Control guideline and scramming the reactor when a maximum safe operating value is approached or exceeded. Failing to do so can result in an unnecessary offsite release and endanger plant personnel.

Safety Significance Entering the RPV Control guideline shuts down the reactor and allows controlled RPV pressure reduction to lower the driving head of the leak to reduce the effect on secondary containment, in order to avoid or anticipate emergency RPV depressurization.

Cues Indication of rising or Maximum Operating values in an area of a system which is connected to the RCS, combined with abnormal system parameters (e.g. such as levels, pressures, and flow rates).

Field reports of visible/audible leaks into secondary containment.

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

Indicators RPS Group lights de-energized on panel 9-5.

Performance Control Rod full -in indication on panel 9-5.

Feedback Reactor power trend on nuclear instrumentation on panel 9-5.

Applicability EOP-5A conditions where a system (primary or non-primary) is discharging into the secondary containment and manual isolation capability from the control room is possible. This includes manipulation of valve control switches and valve power supply control switches, as applicable. If the leaking system is required for adequate core cooling (or meets the other criteria in EOP-05A), this task is not applicable EOP-5A directs that this action be taken when a maximum safe operating value is approached or exceeded, and the source of a discharge cannot be isolated or determined. If Justification for the temperatures, radiation levels, or area water levels in any one chosen performance of the areas listed in the Secondary Containment Control limit guideline approach their maximum safe operating value, adequate core cooling, containment integrity, safety of personnel, or continued operability of equipment required to perform EPG actions can no longer be assured. A reactor Page 7 of 12 Rev 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9/2020 Scenario 1 Page 8 of 12 scram is initiated through entry of the RPV Control guideline to reduce the primary system discharge into secondary containment and in anticipation of possible emergency RPV depressurization. The scram is effected indirectly, through entry of the RPV Control guideline, rather than through an explicit direction in the Secondary Containment Control guideline, to ensure that RPV water level, RPV pressure, and reactor power are properly coordinated following the scram and to avoid potential conflicts with alternate rod insertion strategies in Step RC/Q if the RPV Control guideline is already in use. Failing to do so can significantly change the mitigation strategy as an unnecessary release will result and possibly endangering plant personnel.

Before reaching two Maximum Safe values in two areas for the same parameter was chosen because that is the next EOP-5A significant action threshold, when Emergency Depressurization is required. Shutting down the reactor and entering the RPV Control guideline before reaching this level will may avert the significant thermal transient on the RPV caused by Emergency Depressurization. If emergency depressurization is ultimately require, first shutting down the Justification reactor reduces the amount of energy introduced into secondary containment, thus mitigating the effects on secondary containment temperature, radiation, and water levels, as well as offsite release rates. Emergency depressurization without reactor shutdown creates an unnecessary challenge to the secondary containment barrier.

App. B, steps SC/T-4, SC/R-2, SC/L-2 BWR Owners Group Appendix The scenario must be able to drive secondary containment radiation to its Max Safe value in two plant areas. The CT Scenario Guide listed in the scenario should list which instruments/areas will Requirements exceed their MSO limit first and second.

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Appendix D Scenario Outline Form ES-D-1 NRC CNS 9/2020 Scenario 1 Page 9 of 12 When Group 6 fails to automatically initiate/isolate due to high radiation, manually initiate Group 6 Critical Task #2 isolation/initiation within 90 after Reactor Building Exhaust Plenum radiation levels exceed 49 mR/hr.

The isolation signals generated by the secondary containment isolation instrumentation are implicitly assumed in the safety analyses of USAR section V-3.0 and chapter XIV to initiate closure of valves and start the SGT System to limit offsite doses. High secondary containment exhaust radiation is an indication of possible gross failure of the fuel cladding.

Safety Significance When Reactor Building Exhaust Plenum Radiation-High is detected, secondary containment isolation and actuation of the SGT System are initiated to limit the release of fission products as assumed in the USAR safety analyses, Sections XIV-6.3 and XIV-6.4. With a fuel failure, releasing dose to the public through an unfiltered pathway increases risk to employees onsite and the public Cues Annunciators:

• 9-4-1/E-4 RX BLDG VENT HI-HI RAD

• 9-4-1/E-5 RX BLDG VENT HI RAD Group 6 lights lit on Group Isolation Status on Panel 9-5 Group 6 valves indicate open on Isolation Valve Status mimic on Panel 9-3 and at valve control switches SGT A and B remain idle Operator places Mode switch for RMP-RM-452A and 452B, RX BLDG VENT RAD MON CH A and CH B, to TRIP TEST.

Measurable OR Performance Indicators Operator places Mode switch for RMP-RM-452C and 452D, RX BLDG VENT RAD MON CH C and CH D, to TRIP TEST.

Group 6 lights OFF on Group Isolation Status on Panel 9-5 Performance Feedback Group 6 valves indicate closed SGT A and B start Applicability EOP-5A conditions, Reactor building isolation signal exceeding tech spec values and no group isolation occurs.

This CT is based upon RB Exhaust Plenum radiation exceeding the TS trip setpoint allowable value. Trip setpoints Justification for the are those predetermined values of output at which an action chosen performance should take place. The setpoints are compared to the actual limit process parameter (e.g., reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit)

Page 9 of 12 Rev 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9/2020 Scenario 1 Page 10 of 12 changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis or appropriate documents. The Allowable Values are derived from the analytic limits, corrected for calibration, process, and some of the instrument errors. The isolation of the reactor building within 90 seconds is IAW the USAR V-03.

Group 6 isolation is required to prevent an unfiltered/unmonitored release to the site and to the public.

This minimizes the consequences of the fuel failure to the public.

BWR Owners Group N/A Appendix The scenario must be able to drive the Reactor building Scenario Guide exhaust plenum to above 49 mR/hr.

Requirements Page 10 of 12 Rev 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9/2020 Scenario 1 Page 11 of 12 When a primary system is discharging into the secondary containment through a break and automatic isolation Critical Task #3 fails, the crew takes action which effects isolation of the system prior to reaching two Maximum Safe Operating radiation values in two areas.

EOP-5A directs entering the RPV Control guideline and isolating a leak prior to a maximum safe operating value is exceeded. Failing to do so can result in an unnecessary Safety Significance offsite release and endanger plant personnel. Entering the RPV Control guideline and following the actions, reduces the effect on secondary containment, in order to avoid emergency RPV depressurization..

Cues Indication of rising or Maximum Operating values in an area of a system which is connected to the RCS, combined with abnormal system parameters (e.g. such as levels, pressures, and flow rates).

Field reports of visible/audible leaks into secondary containment.

Measurable One of the 2 isolation valves is closed Performance Indicators Performance Green light on and red light off on either RWCU-MOV-15 or Feedback RWCU-MOV-18 Applicability EOP-5A conditions where a system (primary or non-primary) is discharging into the secondary containment and manual isolation capability from the control room is possible. This includes manipulation of valve control switches and valve power supply control switches, as applicable. If the leaking system is required for adequate core cooling (or meets the other criteria in EOP-05A), this task is not applicable EOP-5A directs that this action be taken prior to exceeding a maximum safe operating value. If temperatures, radiation levels, or area water levels in any one of the areas listed in the Secondary Containment Control guideline approach their Justification for the maximum safe operating value, adequate core cooling, chosen performance containment integrity, safety of personnel, or continued limit operability of equipment required to perform EPG actions can no longer be assured. Failure to isolate the leak can cause the mitigation strategy to fail, as an unnecessary release will result and possibly endangering plant personnel.

Before reaching two Maximum Safe values in two areas for the same parameter was chosen because that is the next Justification EOP-5A significant action threshold, when Emergency Depressurization is required. Isolating the leak and entering the RPV Control guideline before reaching this level may Page 11 of 12 Rev 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9/2020 Scenario 1 Page 12 of 12 avert the significant thermal transient on the RPV caused by Emergency Depressurization. Without the leak normal shutdown and depressurization can occur.

BWR Owners Group App. B, steps SC/T-3, SC/R-1, SC/L-1 Appendix The scenario must be able to drive secondary containment Scenario Guide radiation to its Max Safe value in two plant areas. The CT Requirements listed in the scenario should list which instruments/areas will exceed their MSO limit first and second.

Related Operating Experience CR-CNS-2018-07852 Gland Seal Exhauster A trip During ascension of reactor pressure from 150 to 475 psig, Gland Seal Condenser Exhauster A began showing erratic indications. Low Discharge Pressure and High Condenser level alarms began cycling. Operators were dispatched to adjust the blast gate, but prior the adjustment, Gland Seal Exhauster A began tripping and restarting.

Received a report from an electrician in the field that amps on the exhauster were reading well above nameplate. Gland Exhausters were shifted, to B running and A secured, and Gland Seal indications stabilized. E-shop reported that B Gland Exhauster amps were 17 amps, below nameplate. Blast gate for A Gland Exhauster was moved after the shift of exhausters to match the position of the blast gate for B Exhauster.

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

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

1. Perform Procedure 15.RF.101 valve test for RFPTs

2. Respond to APRM A and C failing high

3. Respond to spurious RCIC initiation

4. Respond to small Recirc loop B leak in the drywell

5. Respond to Recirc loop B line rupture in the drywell

6. Respond to loss of offsite power

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

8. Respond to HPCI auxiliary oil pump oil leak

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

Inoperable Equipment: Core Spray Pump A.

Turnover:

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

Planned activities for this shift are:

• Perform Procedure 15.RF.101

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

This is a new scenario.

Validation Time: XX minutes Rev 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-9 Scenario 2 Event Malf. No. Event Type Event No. Description 1 N/A N (BOP,CRS) Perform Procedure 15.RF.101 Spurious RCIC initiation CT#1 C (BOP,CRS) When a spurious RCIC initiation occurs that would cause power to rise above 100% rated 2 rd08a A (CREW) thermal power, crew lowers power to restore TS (CRS) and maintain power 100% rated thermal power by lowering Reactor Recirc flow before PMIS point NSSRP640 exceeds 2419 MWt.

nm09c @ 94 C (ATC,CRS) APRM C fails to 115%, APRM A fails to 106%

3 nm09e @ 85 TS (CRS) (bypass APRM C)

C (BOP,CRS) 4 rr20b @ 3 A (CREW) Small Recirc loop B leak in the drywell TS (CRS)

Recirc loop B line rupture in the drywell, loss of offsite power CT#2 When high pressure injection systems cannot maintain RPV level and low pressure ECCS systems fail to automatically start due to failure of DG1 to automatically energize the respective bus during loss of offsite AC power, crew manually starts DG1 to energize LP ECCS systems prior to RPV water level falling below -

183 CFZ (TAF) 5 rr20b @ 21 M (CREW)

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

Failure of DG2 output breaker to automatically 6 dg03b C (BOP,CRS) close (manually close breaker EG2)

HPCI auxiliary oil pump oil leak (place HCIC AOP 7 hp12 C (ATC,CRS) in PTL)

Injection valve RHR-MO-25A loss of power (align 8 rd08b C (ATC, BOP,CRS) RHR loop A pumps to containment spray or SPC)

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

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

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Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-9 Scenario 2 Quantitative Attributes Table ES-301-4 Actual Attribute Description Target

1. Failure of DG2 output breaker to automatically close Malfunctions after 2. HPCI auxiliary oil pump oil leak 1-2 3 EOP entry

3. RHR-MO-25A loss of power

1. Spurious RCIC initiation Abnormal Events 2-4 2 2. Small Recirc loop B leak in the drywell

1. Recirc loop B line rupture in the drywell Major Transients 1-2 2 2. Loss of offsite power EOP entries 1. EOP-1A requiring 1-2 2 2. EOP-3A substantive action EOP contingencies requiring 1 per set 1 1. EOP-2A Contingency #2 - Emergency Depressurization substantive action

1. (CT#1) When a spurious RCIC initiation occurs that would cause power to rise above 100% rated thermal power, crew lowers power to restore and maintain power 100%

rated thermal power by lowering Reactor Recirc flow before PMIS point NSSRP640 exceeds 2419 MWt.

2. (CT#2) When high pressure injection systems cannot maintain RPV level and low pressure ECCS systems fail to automatically start due to failure of DG1 to automatically energize the respective bus during loss of offsite AC power, crew manually starts DG1 to energize Pre-identified LP ECCS systems prior to RPV water level falling below -

2 3 Critical Tasks 183 CFZ (TAF).

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

Normal Events N/A 1 1. Perform Procedure 15.RF.101 Reactivity 1. none N/A 0 Manipulations

1. Spurious RCIC initiation

2. APRM C and A failure Instrument/ 3. Recirc loop B line rupture in the drywell Component N/A 6 4. Failure of DG2 output breaker to automatically close Failures

5. HPCI auxiliary oil pump oil leak

6. RHR-MO-25A loss of power Rev 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-9 Scenario 2

1. Spurious RCIC initiation

2. APRM C and A failure

3. Recirc loop B line rupture in the drywell Total Malfunctions N/A 6 4. Failure of DG2 output breaker to automatically close

5. HPCI auxiliary oil pump oil leak

6. RHR-MO-25A loss of power Top 10 systems and operator actions important to risk that are tested:

• Primary Containment, RCIC, Emergency AC Power (DG2), HPCI, RHR Rev 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-9 Scenario 2 SCENARIO

SUMMARY

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

After the crew takes the watch, the BOP will perform Procedure 15.RF.101, RFPT Stop Valve Test.

After procedure 15.RF.101 is completed, RCIC will spuriously initiate. The crew will enter Procedure 2.4CSCS and secure RCIC. RCIC injection must be secured or reactor power must be lowered to prevent exceeding the licensed power limit (CT#1). The CRS will enter TS 3.5.3 Condition A due to RCIC initiation capability being prevented.

After TS for RCIC have been addressed, APRMs A and C will fail. APRM A will fail to just below the upscale rod block, and APRM C will fail to just above the upscale rod block. The crew will respond IAW alarm cards and bypass APRM C. The CRS will enter TS 3.3.1.1 Condition A.

After TS for APRM A and C have been addressed, a small leak from Reactor Recirc loop B will develop. The crew will enter procedure 2.4PC and vent primary containment.

After primary containment venting has stabilized drywell pressure, Recirc loop B will rupture. Drywell pressure and temperature will rise, requiring a manual scram. EOPs 1A and 3A will be entered due to high drywell pressure. RPV water level will lower due to the leak. HPCI Auxiliary Oil Pump will develop and oil leak, HPCI AOP will be secured by the crew, preventing HPCI operation.

CRD and SLC will not be able to keep up with the leak. RPV water level will continue to fall to TAF, -158.

Shortly after the scram, a loss of offsite power will occur. DG2 output breaker wwill fail to automatically close, and the crew will close it manually from Panel C (CT#2) to energize Div 2 low pressure ECCS systems. Elevated drywell pressure will require operation of Torus/Drywell Spray using systems not required to maintain adequate core cooling. CRD flow will be maximized and SLC will be initiated for level control, but they will not be able to stabilize reactor water level. When level reaches TAF, as indicated on SPDS using Corrected Fuel Zone (CFZ) at -158, and with Core Spray Pump B and LPCI Pumps C and D lined up for injection, the crew will enter EOP-2A and conduct emergency depressurization due to RPV water level below TAF and cannot be restored and maintained above -183 CFZ (CT#3).

When reactor press lowers below the low pressure permissive for LPCI injection valves, RHR-MO-25A will lose power. The crew will align RHR loop A to containment spray or suppression pool cooling, since RHR Pumps A and C will Rev 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-9 Scenario 2 be operating on minimum flow. The crew will use Core Spray Pump B and RHR Pumps B and D to restore reactor water level.

The exercise ends when emergency depressurization is complete and reactor water level is being raised to the normal band, +3 to +54 inches.

Rev 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-9 Scenario 2 CRITICAL TASK BASIS When a spurious RCIC initiation occurs that would cause power to rise above 100% rated thermal power, crew lowers power to restore and maintain power 100% rated thermal Critical Task #1 power by lowering Reactor Recirc flow before PMIS point NSSRP640 exceeds 2419 MWt.

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

Cues Annunciators 9-4-1/A-1, RCIC Logic Initiated, RCIC turbine speed rising, RCIC-MO-21 opening, RCIC flow on RCIC-FI-91 rising, Reactor power rising indicated on IRM/APRM recorders NM-NR-46A-D, SPDS, PMIS.

Operator depresses RCIC Turbine Trip button and closes RCIC-MO-131 on panel 9-4, AND/OR Measurable Performance Indicators If reactor power rises above 100%, Operator Selects S on RR flow controllers RRFC-SIC-16A(B) on panel 9-4 and lowers RR pump flow (by turning speed demand counter-clockwise on one speed controller at a time) until power stabilizes below 100% on IRM/APRM recorders NM-NR-46A-D.

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

Applicability Any time a cold water injection from RCIC causes FW temperature to lower such that reactor power would exceed 100% with no operator intervention.

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

The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> point is the legal record for plant power level. PMIS point NSSRP640 (15 minute average) was chosen because it is Rev 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-9 Scenario 2 more limiting and provides the most rapid update to core thermal power available.

BWR Owners Group N/A Appendix Initial power level must be near 100% so injection from RCIC would Scenario Guide cause reactor power to rise above 100% with no operator Requirements intervention.

Rev 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-9 Scenario 2 When high pressure injection systems cannot maintain RPV level and low pressure ECCS systems fail to automatically start due to failure of DG2 to automatically energize 4160V Bus 1G Critical Task #2 during loss of offsite AC power, crew manually closes DG2 output breaker to energize LP ECCS systems prior to RPV water level falling below -183 CFZ (TAF)

Failure to recognize the failure of DG2 output breaker to automatically close and energize the safety bus and failure to take manual action per Procedure 5.3EMPWR will result in unavailability Safety Significance of safety-related equipment necessary to provide adequate core cooling, otherwise resulting in core damage and a large offsite release.

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

• Bus energized lamps extinguished

• Circuit breaker position

• Bus voltage

• EDG status

• Control room lighting dimmed Manipulation of controls as required to energize Div 2 AC Measurable Performance emergency bus from panel C:

Indicators Operator places DIESEL GEN 2 BKR EG2 to CLOSE on panel C Crew will observe light indication for equipment powered by Division Performance Feedback 2 AC illuminate on panel 9-3 and bus voltage ~4200V on panel C Applicability Loss of off-site power events when all sources of off-site power are lost and a diesel generator fails to energize its bus. This is only applicable if manual action from the Control Room would be effective in energizing the bus.

Attempting to energize ECCS systems must be performed to determine their availability by the time -183 is reached in order to properly implement EOP-1A decision steps regarding restoring and Justification for the maintaining RPV level. The MSCWL (-183 CFZ) is the lowest RPV chosen performance water level at which the covered portion of the reactor core will limit generate sufficient steam to preclude any clad temperature in the uncovered portion of the core from exceeding 1500°F.

BWR Owners Group App. B, Contingency#1 Appendix LOCA severity should result in a near linear RPV level reduction that causes level to fall to TAF over approximately 15-20 minutes Scenario Guide from the time the initial LOCA signal is received. (The LOCA Requirements malfunction severity may be ramped initially, but it should reach its final severity within approximately the first 3 minutes.)

Rev 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-9 Scenario 2 When RPV level lowers to -158 CFZ (TAF) and cannot be maintained above -183 CFZ (MSCWL) and it is apparent to the crew that insufficient high pressure injection systems will be available to restore level, crew Emergency Depressurizes by Critical Task #3 opening the first of 6 SRVs before RPV level lowers below -183 CFZ. (Momentary shrink below -183 CFZ due to automatic SRV operation in Low-Low Set mode does not constitute failure of this CT.)

The MSCWL is the lowest RPV water level at which the covered portion of the reactor core will generate sufficient steam to preclude any clad temperature in the uncovered portion of the core from Safety Significance exceeding 1500°F. When water level decreases below MSCWL with injection, clad temperatures may exceed 1500°F.

Cues Corrected Fuel Zone indication (SPDS) falls to -158 and lowering trend continues, and, before -158 CFZ is reached, initial conditions, field reports, and control room indications convey that adequate high pressure injection cannot be restored before level falls below -183 CFZ.

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

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

Applicability EOP-1A conditions with RPV pressure above the shutoff head of available low pressure injection systems or subsystems and any system injecting to the RPV (i.e. not in steam cooling).

The MSCWL (-183 CFZ) is the lowest RPV water level at which the covered portion of the reactor core will generate sufficient steam to preclude any clad temperature in the uncovered portion of the core from exceeding 1500°F. Emergency depressurization is allowed when level goes below TAF (-158 CFZ) and should be performed, if Justification for the in the judgment of the CRS, level cannot be maintained above -183 chosen performance CFZ. Since it is intended for the scenario supporting this CT to, limit early in the event, clearly indicate no high pressure injection systems can be made available to reverse the lowering level trend, the crew will have time to communicate and open 6 SRVs before -

183 CFZ.

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

Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-9 Scenario 2 LOCA severity should result in a near linear RPV level reduction that causes level to fall to TAF over approximately 15-20 minutes from the time the initial LOCA signal is received. It is very important to design the scenario such that the crew has information early during the LOCA event to determine high pressure injection systems cannot be recovered or optimized in order to stabilize level before -

183 CFZ is reached. The crew should know this within approximately 10 minutes from the start of the LOCA and by the Scenario Guide time level lowers to -100 CFZ to allow time to align/realign low Requirements pressure systems for injection before level reaches -158 CFZ, so that the only remaining action when TAF is reached will be to conduct emergency depressurization. (e.g As an initial condition, HPCI turbine is disassembled for maintenance. A field report for a RCIC valve malfunction states a valve has mechanical binding in the gearbox, cannot be manually opened and will take 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to repair.

Control room indications show a loss of offsite power and the dispatcher reports it cannot be restored for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. etc.)

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Facility: Cooper Nuclear Station Scenario No.: 3 Op-Test No.: CN-2020-09 Examiners: ____________________________ Operators: _____________________________

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

1. Withdraw control rods IAW the rod sequence to establish 17% bypass valve position

2. Respond to SRM D upscale

3. Place HPCI in standby

4. Respond to HPCI-MO-16 loss of power during stroke

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

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

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

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

9. Respond to failure RCIC isolation MOVs

10. Respond to 480V Bus 1A loss of power Initial Conditions: Plant operating at 5% power during startup, beginning of life Inoperable Equipment: none Turnover:

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

Planned activities for this shift are:

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

• Place HPCI in standby IAW Procedure 2.2.33.

• Continue startup IAW Procedure 2.1.1.

Scenario Notes:

This is a new scenario.

Validation Time: 60 minutes Rev 0

Event Malf. No. Event Type Event No. Description 1 N/A R (ATC,CRS) Raise reactor power by withdrawing control rods 2 nm02d @ 100 I (ATC,CRS) SRM D upscale 3 N/A N (BOP,CRS) Place HPCI in standby C (BOP,CRS) 4 r/f hp04 HPCI-MO-16 loss of power during stroke TS (CRS)

C (ATC,CRS) 5 rd04a @ 0 CRD-FCV-19A fails closed A (CREW)

C rp03c (ATC,BOP,CRS) 6 o/r zdisgtswefrf @ RPS B EPA breaker trip with SGT B fan trip off A (CREW)

TS (CRS)

RCIC steam line break in the Reactor Building CT#2 When a primary system is discharging into the secondary containment through a break and M cannot be isolated, the crew scrams the reactor 7 rc06 @ 7 (ATC,BOP,CRS) prior to reaching Maximum Safe Operating Temperatures in two areas. (For this scenario, RCIC-TE-77C Torus 890 ENE is the first area to exceed MSO and RHR-TE-99C Torus 885 NNW is the second area to exceed MSO.)

Failure of RCIC to automatically isolate on high area 8 rc07 C (BOP,CRS) temperature Failure of RCIC steam supply isolation valves to fully close CT#2 When a primary system is discharging into the secondary containment through an unisolable break, the crew Emergency Depressurizes by opening 6 SRVs when maximum safe operating r/f rc06a @ de-ener C temperatures are exceeded in two areas and 9 prior to exceeding Maximum Safe Operating r/f rc22a @ de-ener (ATC,BOP,CRS) temperature in a third area. (For this scenario, RCIC-TE-77C Torus 890 ENE and RHR-TE-99C Torus 885 NNW are the two areas that will exceed MSO first. RWCU-TE-117F Torus 896 W would be the third area to exceed MSO.)

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

10 ed09a C (BOP,CRS) 480V Bus 1A loss of power (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal (TS) Tech Spec

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

Rev 0

Quantitative Attributes Table ES-301-4 Actual Attribute Description Target

1. Failure of RCIC to automatically isolate on high area temperature Malfunctions after 2. Failure of RCIC steam supply isolation valves to fully 1-2 3 EOP entry close

3. 480V Bus 1A loss of power

1. CRD-FCV-19A fails closed Abnormal Events 2-4 2 2. RPS B EPA breaker trip with SGT B fan trip

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

1. (CT#1) When a primary system is discharging into the secondary containment through a break and cannot be isolated, the crew scrams the reactor prior to reaching Maximum Safe Operating Temperatures in two areas.

(For this scenario, RCIC-TE-77C Torus 890 ENE is the first area to exceed MSO and RHR-TE-99C Torus 885 NNW is the second area to exceed MSO.)

2. (CT#2) When a primary system is discharging into the Pre-identified secondary containment through an unisolable break, the 2 2 Critical Tasks crew Emergency Depressurizes by opening 6 SRVs when maximum safe operating temperatures are exceeded in two areas and prior to exceeding Maximum Safe Operating temperature in a third area. (For this scenario, RCIC-TE-77C Torus 890 ENE and RHR-TE-99C Torus 885 NNW are the two areas that will exceed MSO first.

RWCU-TE-117F Torus 896 W would be the third area to exceed MSO.) Anticipating Emergency Depressurization and fully opening Bypass valves also satisfies this CT.

Normal Events N/A 1 1. Place HPCI in standby Reactivity 1. Raise reactor power by withdrawing control rods N/A 1 Manipulations Rev 0

1. SRM D upscale

2. HPCI-MO-16 loss of power during stroke

3. CRD-FCV-19A fails closed Instrument/ 4. RPS B EPA breaker trip with SGT B fan trip Component N/A 7 5. Failure of RCIC to automatically isolate on high area Failures temperature

6. Failure of RCIC steam supply isolation valves to fully close

7. 480V Bus 1A loss of power

1. SRM D upscale

2. HPCI-MO-16 loss of power during stroke

3. CRD-FCV-19A fails closed

4. RPS B EPA breaker trip with SGT B fan trip Total Malfunctions N/A 7 5. Failure of RCIC to automatically isolate on high area temperature

6. Failure of RCIC steam supply isolation valves to fully close

7. 480V Bus 1A loss of power Top 10 systems and operator actions important to risk that are tested:

• HPCI, Primary Containment Isolation System, RCIC, RHRSW Rev 0

SCENARIO

SUMMARY

The plant is operating at 5% power during startup, beginning of life.

After the crew takes the watch, the ATC will withdraw control rods IAW the startup rod sequence to establish Bypass valves 17% open.

During the power ascension, SRM D will fail upscale. The crew will respond IAW the alarm card and bypass SRM D.

After Bypass valves are at least 17% open, the crew will align HPCI to standby IAW procedure 2.2.33. When HPCI-MO-16 is being opened, it will lose power.

The CRS will enter TS 3.6.1.3 Condition A and TS 3.5.1 Condition C. The crew will isolate HPCI steam supply.

After TS for HPCI-MO-16 have been addressed, CRD-FCV-19A will fail closed.

The crew will enter procedure 2.4CRD and shift CRD flow control valves.

After CRD flow control valves have been shifted, RPS B EPA breaker 1B2 will trip, resulting in a half scram, half PCIS Group 1, 2, 3, 7 isolations, and a full PCIS Group 6 isolation. RWCU will isolate and the pump will trip. The crew will transfer RPS B to its alternate supply per alarm card C-1/F-2, reset the Div 2 half scram IAW 2.1.5, and enter 2.1.22 to begin recovery from group isolations.

SGT B exhaust fan will trip after it automatically starts. The CRS will enter TS 3.6.4.3 Condition A for SGT B. The crew will not be given time to complete all of the actions of 2.1.22, such as restoring Reactor Building HVAC realignments, due to the timing of the scenario.

When response to the loss of RPS B is complete, a small RCIC steam line break will occur in the Reactor Building. Secondary containment area temperatures will rise, requiring entry into EOP-5A. If RCIC area temperatures reach 195°F, MSO limit and isolation setpoint, before the crew attempts manual isolation of RCIC steam supply valves, the automatic isolation will fail to occur.

When the crew attempts manual isolation of the RCIC steam supply valves, both valves will fail to fully close due to mechanical binding in their gear boxes.

EOP-1A entry and reactor shutdown will be required before two areas exceed the MSO limit (CT#1).

Following the scram, 480V Bus 1A will lose power. The crew will respond IAW procedure 5.3AC480 and will be required to start TEC Pump C.

When the crew has stabilized the plant following the scram, the RCIC steam leak will get worse, causing area temperatures in more than one area of the Reactor Building to approach the MSO limit. When two areas have reached the MSO limit, emergency depressurization will be required IAW EOP-5A, and Rev 0

the crew will enter EOP-2A to perform emergency depressurization (CT#2).

The crew may anticipate emergency depressurization and fully open Bypass valves, which also satisfies CT#2.

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

Rev 0

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

EOP-5A directs entering the RPV Control guideline and scramming the reactor when a maximum safe operating value is approached or exceeded. Failing to do so can result in an unnecessary offsite release and endanger plant personnel. Entering the RPV Control Safety Significance guideline shuts down the reactor and allows controlled RPV pressure reduction to lower the driving head of the leak to reduce the effect on secondary containment, in order to avoid or anticipate emergency RPV depressurization..

Cues Indication of rising or Maximum Operating values in an area of a system which is connected to the RCS, combined with abnormal system parameters (e.g. such as levels, pressures, and flow rates).

Field reports of visible/audible leaks into secondary containment.

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

RPS Group lights de-energized on panel 9-5.

Performance Feedback Control Rod full -in indication on panel 9-5.

Reactor power trend on nuclear instrumentation on panel 9-5.

Applicability EOP-5A conditions where a system (primary or non-primary) is discharging into the secondary containment and manual isolation capability from the control room is possible. This includes manipulation of valve control switches and valve power supply control switches, as applicable. If the leaking system is required for adequate core cooling (or meets the other criteria in EOP-05A), this task is not applicable.

EOP-5A directs that this action be taken when a maximum safe operating value is approached or exceeded and the source of a discharge cannot be isolated or determined. If temperatures, radiation levels, or area water levels in any one of the areas listed in the Secondary Containment Control guideline approach their maximum safe operating value, adequate core cooling, containment integrity, safety of personnel, or continued operability of equipment required to perform EPG actions can no longer be assured. A reactor scram is initiated through entry of the RPV Control guideline Justification for the to reduce the primary system discharge into secondary containment chosen performance limit and in anticipation of possible emergency RPV depressurization.

The scram is effected indirectly, through entry of the RPV Control guideline, rather than through an explicit direction in the Secondary Containment Control guideline, to ensure that RPV water level, RPV pressure, and reactor power are properly coordinated following the scram and to avoid potential conflicts with alternate rod insertion strategies in Step RC/Q if the RPV Control guideline is already in use. Failing to do so can significantly change the mitigation strategy as an unnecessary release will result and possibly endangering plant personnel.

Rev 0

Before reaching two Maximum Safe values in two areas for the same parameter was chosen because that is the next EOP-5A significant action threshold, when Emergency Depressurization is required. Shutting down the reactor and entering the RPV Control guideline before reaching this level may avert the significant thermal transient on the RPV caused by Emergency Depressurization. If BWR Owners Group emergency depressurization is ultimately required, first shutting Appendix down the reactor reduces the amount of energy introduced into secondary containment, thus mitigating the effects on secondary containment temperature, radiation, and water levels, as well as offsite release rates. Emergency depressurization without reactor shutdown creates an unnecessary challenge to the secondary containment barrier.

App. B, steps SC/T-4, SC/R-2, SC/L-2 Scenario Guide Requirements Rev 0

When a primary system is discharging into the secondary containment through an unisolable break, the crew Emergency Depressurizes by opening 6 SRVs when maximum safe operating temperatures are exceeded in two areas and prior to exceeding Maximum Safe Operating temperature in a third area. (For this scenario, RCIC-TE-77C Torus 890 ENE and Critical Task #2 RHR-TE-99C Torus 885 NNW are the two areas that will exceed MSO first. RWCU-TE-117F Torus 896 W would be the third area to exceed MSO.) Anticipating Emergency Depressurization and fully opening Bypass valves also satisfies this CT.

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

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

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

Manipulation of SRV controls on panel 9-3:

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

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

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

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

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

Rev 0

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

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

1. Place RHR Pump B in Suppression Pool Cooling

2. Respond to RHR-MO-34B loss of power/failure to open for SPC

3. Respond to CRD Pump A trip

4. Respond to Reactor Building Exhaust Ventilation failure

5. Respond to control rods 14-19, 18-19, and 34-35 drifting in

6. Respond to failure of the scram air header to vent (ATWS)

7. Respond to failure of ARI to initiate

8. Respond to failure of both SLC pumps

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

Inoperable Equipment: None.

Turnover:

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

Planned activities for this shift are:

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

Scenario Notes:

This is a new scenario.

Validation Time: XX minutes

Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-9 Scenario 4 Event Malf. No. Event Type Event No. Description 1 N/A N (BOP,CRS) Place RHR Pump B in Suppression Pool Cooling C (BOP,CRS) RHR-MO-34B loss of power/failure to open 2 r/f rh26a @ de-ener TS (CRS) (secure RHR Pump B)

C (ATC,CRS) 3 rd08a CRD Pump A trip (start CRD Pump B)

A (CREW) o/r zdihvswefr1a C (BOP,CRS)

@ off Reactor Building Exhaust Ventilation failure (start 4 A (CREW) o/r zdihvswefr1b SGT)

@ stndby TS (CRS)

Control rods 14-19, 18-19, and 34-35 drift in, manual scram required, failure of scram air header to vent CT#1 During failure to scram conditions when power rp01a is above 25% and the Main Turbine is on line rp01b and/or Bypass valves are available, prevent MSIV isolation by maintaining reactor water rp01c level above -104 inches or by installing MSIV 5 M (CREW) rp01d low level isolation bypass jumpers before wide range reactor water level lowers below -104 rd26 inches.

rd27 CT#2 Inhibit ADS prior to uncontrolled injection from high volume, low pressure systems and before exceeding the Tech Spec cooldown rate limit during a failure to Scram.

o/r zdirdsws53aa

@ normal Failure of ARI to initiate (manually trip Recirc 6 I (ATC,CRS) o/r zdirdsws53aa pumps)

@ normal sl01a Trip of both SLC pumps (place RPS test switches 7 C (ATC,CRS) to TRIP) sl01b CRD Pump B trip CT#3 When control rods fail to scram due to scram pilot air header failing to vent, CRD pumps are 8 rd08b C (ATC,CRS) not available, SLC is not available, and energy is discharging into primary containment, crew directs manually venting the scram pilot air header before exceeding Boron Injection Initiation Temperature Limit (BIIT) and before exiting EOP-6A.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-9 Scenario 4 RCIC flow controller fails low in AUTO (control 9 rc04 @ 0 C (BOP,CRS) RCIC speed in MANUAL)

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

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

Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-9 Scenario 4 Quantitative Attributes Table ES-301-4 Actual Attribute Description Target

1. Failure of ARI to initiate

2. Trip of both SLC pumps Malfunctions after 1-2 4 3. CRD Pump B trip EOP entry

4. RCIC flow controller fails low in AUTO

1. CRD Pump A trip Abnormal Events 2-4 2 2. Reactor Building Exhaust Ventilation failure

1. Control rods 14-19, 18-19, and 34-35 drift in, manual Major Transients 1-2 1 scram required, failure of scram air header to vent EOP entries 1. EOP-6A requiring 1-2 2 2. EOP-7A substantive action EOP contingencies requiring 1 per set 1 1. EOP-7A Contingency #5 - Level/Power Control substantive action

1. (CT#1) During failure to scram conditions when power is above 25% and the Main Turbine is on line and/or Bypass valves are available, prevent MSIV isolation by maintaining reactor water level above -104 inches or by installing MSIV low level isolation bypass jumpers before wide range reactor water level lowers below -104 inches.

2. (CT#2) Inhibit ADS prior to uncontrolled injection from Pre-identified high volume, low pressure systems and before exceeding 2 3 the Tech Spec cooldown rate limit during a failure to Critical Tasks Scram.

3. (CT#3) When control rods fail to scram due to scram pilot air header failing to vent, CRD pumps are not available, SLC is not available, and energy is discharging into primary containment, crew directs manually venting the scram pilot air header before exceeding Boron Injection Initiation Temperature Limit (BIIT) and before exiting EOP-6A.

Normal Events N/A 1 1. Place RHR Pump B in Suppression Pool Cooling Reactivity 1. none N/A 0 Manipulations

1. RHR-MO-34B loss of power/failure to open

2. CRD Pump A trip Instrument/ 3. Reactor Building Exhaust Ventilation failure Component N/A 6 4. Failure of ARI to initiate Failures

5. Trip of both SLC pumps

6. RCIC flow controller fails low in AUTO

Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-9 Scenario 4

1. RHR-MO-34B loss of power/failure to open

2. CRD Pump A trip

3. Reactor Building Exhaust Ventilation failure Total Malfunctions N/A 6 4. Failure of ARI to initiate

5. Trip of both SLC pumps

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

• Primary Containment, RHR, RHRSW

Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-9 Scenario 4 SCENARIO

SUMMARY

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

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

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

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

After response to loss of RB ventilation is complete, three control rods will simultaneously drift in. The crew will respond IAW the alarm card and insert a manual scram. RPS will de-energize, but the scram air header will fail to vent.

Manual initiation of ARI will fail. The crew will enter EOP-1A and transition to EOP-6A and EOP-7A. The crew will trip Recirc pumps. Power will be approximately 50%. The crew will install MSIV low level isolation bypass jumpers (CT#1) and inhibit ADS (CT#2) and lower reactor water level IAW EOP-7A to control power. When reactor water level is lowered and RCIC automatically initiates, its speed controller will fail low in automatic, and the crew will raise RCIC speed in MANUAL.

CRD Pump B will trip, preventing manual control rod insertion. Both SLC pumps will trip after they have been started, The crew must direct the building operator to manually vent the scram air header IAW procedure 5.8.3 (CT#3). If the crew has not manually vented the scram air header, the main turbine will spuriously trip 25 minutes following the scram, resulting in SRVs opening.

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

Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-9 Scenario 4 CRITICAL TASK BASIS During failure to scram conditions when power is above 30%

and the Main Turbine is on line and/or Bypass valves are available, prevent MSIV isolation by maintaining reactor water Critical Task #1 level above -104 inches or by installing MSIV low level isolation bypass jumpers before wide range reactor water level lowers below -104 inches.

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

Cues APRMs indicate >25% power following a scram.

Turbine Generator is on line.

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

• Terminals DD-1 to DD-2, BAY-1, PNL 9-15

• Terminals BB-1 to BB-2, BAY-3, PNL 9-15

• Terminals DD-1 to DD-2, BAY-1, PNL 9-17

• Terminals BB-1 to BB-2, BAY-3, PNL 9-17 OR Measurable Performance Indicators Crew maintains MSIVs open by maintaining RPV water level above

-104 indicated on Wide Range instruments:

• NBI-LR-1A

• NBI-LR-1B

• NBI-LI-85A

• NBI-LI-85B

• NBI-LI-85C

• SPDS Visual observation jumpers are installed in Panels 9-15 and 9-17.

Wide Range reactor water level >-104.

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

The actual setpoint for MSIV (Group 1) isolation on low reactor Justification for the water level is -104 (TS -113). The actual setpoint is modelled in chosen performance the simulator.

limit

Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-9 Scenario 4 BWR Owners Group App. B, step C5-2 Appendix Power level following trip of both Recirc pumps should be >30%.

Scenario Guide There should be not challenges to Feedwater or HPCI.

Requirements

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

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

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

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

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

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

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

Performance Feedback Receipt of ADS inhibited alarm panel 9-3-1/D-1.

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

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

Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-9 Scenario 4 BWR Owners Group App. B, step RC/Q-6 Appendix The scenario must be designed to make the crew lower RPV level per EOP-7A step FS/L-6 or FS/L-7 (i.e. ATWS with power >3%

Scenario Guide following trip of both recirc pumps, and Feedwater or HPCI Requirements maintaining level above -60 CFZ).

Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-9 Scenario 4 When control rods fail to scram due to scram pilot air header failing to vent, CRD pumps are not available, SLC is not available, and energy is discharging into primary containment, Critical Task #2 crew directs manually venting the scram pilot air header before exceeding Boron Injection Initiation Temperature Limit (BIIT) and before exiting EOP-6A.

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

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

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

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

The BIIT is a function of reactor power. If all control rods are inserted before suppression pool temperature reaches the BIIT, the HCTL will not be exceeded, so emergency RPV depressurization will be precluded.

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

RPS logics lights extinguished on Panel 9-5.

Absence of annunciator 9-5-2/F-5, Scram Valve Pilot Air Low Pressure.

Suppression Pool temperature rising on PMIS and panel indications.

Operator notifies (using phone/PA/etc) building operator to Measurable Performance manually vent the scram air header IAW Procedure 5.8.3.

Indicators Booth operator, as building operator notifies control room operator that the scram air header has been manually vented IAW Procedure 5.8.3.

Performance Feedback Receipt of annunciator 9-5-2/F-5, Scram Valve Pilot Air Low Pressure.

Indication of all control rods fully inserted on Panel 9-5 and/or PMIS.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 2020-9 Scenario 4 Indication of reactor power decaying to 0% on APRMs and/or SPDS.

Applicability ATWS with power >3% following trip of both recirc pumps per EOP-7A, no CRD pumps available, no SLC injection not available.

If all control rods are inserted before suppression pool temperature reaches the BIIT, the HCTL will not be exceeded, so emergency RPV depressurization will be precluded. BIIT is chosen for this scenario over HCTL because BIIT can be exceeded within the normal time frame of a scenario. Depending on conditions, HCTL Justification for the may not be exceeded within the normal timeframe of a scenario.

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

BWR Owners Group App. B, step RC/Q-7 Appendix Initial conditions, combined with the ATWS severity, should result in power >3% following trip of both recirc pumps per EOP-7A. A Scenario Guide condition must cause Suppression Pool temperature to rise if no Requirements crew action is taken. The scenario should be validated to exceed BIIT with no crew action.