ML18285A125

2018-09 Final Outlines

ML18285A125
Person / Time
Site:	Cooper
Issue date:	10/02/2018
From:	Vincent Gaddy Operations Branch IV
To:	Nebraska Public Power District (NPPD)
References
50-298/18-09 50-298/OL-18
Download: ML18285A125 (71)

Text

ES-401 BWR Examination Outline Form ES-401-1 Rev 2 Facility: Cooper Nuclear Station Date of Exam: September 2018 Tier Group RO K/A Category Points SRO-Only Points K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G*

Total A2 G*

Total

1.

Emergency &

Abnormal Plant Evolutions 1

3 3

4 N/A 3

4 N/A 3

20 2

1 1

1 2

1 1

7 Tier Totals 4

4 5

5 5

4 27

2.

Plant Systems 1

2 3

3 3

3 2

2 2

2 2

2 26 2

2 1

1 1

1 1

1 1

1 1

1 12 Tier Totals 4

4 4

4 4

3 3

3 3

3 3

38

3. Generic Knowledge and Abilities Categories 1

2 3

4 10 1

2 3

4 3

3 2

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

2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points, and the SRO-only exam must total 25 points.

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

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

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

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

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

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

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

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

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

ES-401 2

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

E/APE # / Name / Safety Function K

1 K

2 K

3 A

1 A2 G*

K/A Topic(s)

IR 295001 (APE 1) Partial or Complete Loss of Forced Core Flow Circulation / 1 & 4 X

Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION :

AK1.03 Thermal Limits 3.6 1

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

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

POWER :

AK3.06 Containment isolation 3.7 2

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

Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF DC POWER:

AA1.02 systems necessary to assure safe plant shutdown 3.8 3

295005 (APE 5) Main Turbine Generator Trip /

3 X

Knowledge of the operational implications of the following concepts as they apply to MAIN TURBINE GENERATOR TRIP:

AK1.03 Pressure effects on reactor level 3.5 4

295006 (APE 6) Scram / 1 X

2.1.32 Ability to explain and apply system limits and precautions 3.8 5

295016 (APE 16) Control Room Abandonment /

7 X

Ability to determine and/or interpret the following as they apply to CONTROL ROOM ABANDONMENT :

AA2.05 Drywell pressure 3.8 6

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

Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER :

AA1.02 System loads 3.3 7

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

Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR and the following:

AK2.18 ADS 3.5 8

295021 (APE 21) Loss of Shutdown Cooling / 4 X

Ability to determine and/or interpret the following as they apply to LOSS OF SHUTDOWN COOLING :

AA2.02 RHR/shutdown cooling system flow 3.4 9

295023 (APE 23) Refueling Accidents / 8 X

2.4.35 Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects.

3.8 10 295024 High Drywell Pressure / 5 X

Knowledge of the interrelations between HIGH DRYWELL PRESSURE and the following:

EK2.11 Drywell spray (RHR) logic Mark I & II 4.2 11 295025 (EPE 2) High Reactor Pressure / 3 X

Ability to determine and/or interpret the following as they apply to HIGH REACTOR PRESSURE:

EA2.04 Suppression pool level 3.9 12 295026 (EPE 3) Suppression Pool High Water Temperature / 5 X

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

EK2.01 Suppression pool cooling 3.9 13 295027 (EPE 4) High Containment Temperature (Mark III Containment Only) / 5 N/A for CNS (Mark III only) 295028 (EPE 5) High Drywell Temperature (Mark I and Mark II only) / 5 X

Knowledge of the operational implications of the following concepts as they apply to HIGH DRYWELL TEMPERATURE:

EK1.02 Equipment environmental qualification 2.9 14 295030 (EPE 7) Low Suppression Pool Water Level / 5 X

Knowledge of the reasons for the following responses as they apply to LOW SUPPRESSION POOL WATER LEVEL; EK3.07 NPSH considerations for ECCS pumps 3.5 15

ES-401 3

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

E/APE # / Name / Safety Function K

1 K

2 K

3 A

1 A2 G*

K/A Topic(s)

IR 295031 (EPE 8) Reactor Low Water Level / 2 X

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

EA2.04 Adequate core cooling 4.6*

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

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

EA1.06 Neutron monitoring system 4.1 17 295038 (EPE 15) High Offsite Radioactivity Release Rate / 9 X

2.1.19 Ability to use plant computers to evaluate system or component status 3.9 18 600000 (APE 24) Plant Fire On Site / 8 X

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

AK3.04 Actions contained in the abnormal procedure for plant fire onsite 2.8 19 700000 (APE 25) Generator Voltage and Electric Grid Disturbances / 6 X

Knowledge of the reasons for the following responses as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES:

AK3.02 Actions contained in abnormal operating procedure for voltage and grid disturbances 3.6 20 K/A Category Totals:

3 3

4 3

4 3

Group Point Total:

20

ES-401 4

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

E/APE # / Name / Safety Function K

1 K

2 K

3 A

1 A

2 G*

K/A Topic(s)

IR 295002 (APE 2) Loss of Main Condenser Vacuum / 3 NOT SAMPLED 295007 (APE 7) High Reactor Pressure / 3 NOT SAMPLED 295008 (APE 8) High Reactor Water Level /

2 NOT SAMPLED 295009 (APE 9) Low Reactor Water Level /

2 NOT SAMPLED 295010 (APE 10) High Drywell Pressure / 5 X

Knowledge of the interrelations between HIGH DRYWELL PRESSURE and the following:

AK2.02 Drywell/Suppression chamber Differential Pressure 3.3 21 295011 (APE 11) High Containment Temperature (Mark III Containment only) / 5 N/A for CNS (Mark III only) 295012 (APE 12) High Drywell Temperature

/ 5 X

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

AA1.02 drywell cooling system 3.8 22 295013 (APE 13) High Suppression Pool Temperature. / 5 NOT SAMPLED 295014 (APE 14) Inadvertent Reactivity Addition / 1 X

Ability to operate and/or monitor the following as they apply to INADVERTENT REACTIVITY ADDITION:

AA1.06 Reactor/turbine pressure regulating system 3.3 26 295015 (APE 15) Incomplete Scram / 1 X

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

AK3.01 Bypassing rod insertion blocks 3.4 24 295017 (APE 17) Abnormal Offsite Release Rate / 9 NOT SAMPLED 295020 (APE 20) Inadvertent Containment Isolation / 5 & 7 X

Ability to determine and/or interpret the following as they apply to INADVERTENT CONTAINMENT ISOLATION:

AA2.04 Reactor pressure 3.9 25 295022 (APE 22) Loss of Control Rod Drive Pumps / 1 NOT SAMPLED 295029 (EPE 6) High Suppression Pool Water Level / 5 X

2.4.4 Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.

4.5 23 295032 (EPE 9) High Secondary Containment Area Temperature / 5 NOT SAMPLED 295033 (EPE 10) High Secondary Containment Area Radiation Levels / 9 X

Knowledge of the operational implications of the following concepts as they apply to HIGH SECONDARY CONTAINMENT AREA RADIATION LEVELS:

EK1.02 Personnel protection 3.9 27 295034 (EPE 11) Secondary Containment Ventilation High Radiation / 9 NOT SAMPLED 295035 (EPE 12) Secondary Containment High Differential Pressure / 5 NOT SAMPLED 295036 (EPE 13) Secondary Containment High Sump/Area Water Level / 5 NOT SAMPLED 500000 (EPE 16) High Containment Hydrogen Concentration / 5 NOT SAMPLED K/A Category Point Totals:

1 1

1 2

1 1

Group Point Total:

7

ES-401 5

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

System # / Name K

1 K

2 K

3 K

4 K

5 K

6 A

1 A2 A

3 A

4 G*

K/A Topic(s)

IR 203000 (SF2, SF4 RHR/LPCI)

RHR/LPCI: Injection Mode X

Knowledge of RHR/LPCI: INJECTION MODE (PLANT SPECIFIC) design feature(s) and/or interlocks which provide for the following:

K4.07 Emergency generator load sequencing 3.7 28 205000 (SF4 SCS) Shutdown Cooling X

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

A4.01 SDC/RHR pumps 3.7 29 206000 (SF2, SF4 HPCIS)

High-Pressure Coolant Injection X

Ability to a) predict the impacts of the following on the HIGH PRESSURE COOLANT INJECTION SYSTEM and b) based on those predictions, use procedures to correct, control, or mitigate the consequences on those abnormal conditions or operations:

A2.12 Loss of room cooling 3.4 30 207000 (SF4 IC) Isolation (Emergency) Condenser N/A for CNS 209001 (SF2, SF4 LPCS)

Low-Pressure Core Spray X

X Knowledge of the operational implications of the following concepts as they apply to LOW PRESSURE CORE SPRAY SYSTEM :

K5.01 Indications of pump cavitation Ability to monitor automatic operations of the LOW PRESSURE CORE SPRAY SYSTEM including:

A3.06 Lights and alarms 2.6 3.6 31 32 209002 (SF2, SF4 HPCS)

High-Pressure Core Spray N/A for CNS 211000 (SF1 SLCS) Standby Liquid Control X

X Knowledge of the physical connections and/or cause-effect relationships between STANDBY LIQUID CONTROL SYSTEM and the following:

K1.06 Reactor Vessel Ability to predict and/or monitor changes in parameters associated with operating the STANDBY LIQUID CONTROL SYSTEM controls including:

A1.04 Valve operations 3.7 3.6 33 34 212000 (SF7 RPS) Reactor Protection X

X Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR PROTECTION SYSTEM :

K6.04 D.C. electrical distribution 2.4.31 Knowledge of annunciator alarms, indications, or response procedures.

2.8 4.2 35 36 215003 (SF7 IRM)

Intermediate-Range Monitor X

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

K5.03 Changing detector position 3.0 37 215004 (SF7 SRMS)

Source-Range Monitor X

Knowledge of SOURCE RANGE MONITOR (SRM)

SYSTEM design feature(s) and/or interlocks which provide for the following:

K4.02 Reactor SCRAM signals 3.4 38

ES-401 6

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

System # / Name K

1 K

2 K

3 K

4 K

5 K

6 A

1 A2 A

3 A

4 G*

K/A Topic(s)

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

Knowledge of the effect that a loss or malfunction of the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM will have on following:

K3.03 Reactor manual control system 3.3 39 217000 (SF2, SF4 RCIC) Reactor Core Isolation Cooling X

Knowledge of electrical power supplies to the following:

K2.02 RCIC initiation signals (logic) 2.8*

40 218000 (SF3 ADS) Automatic Depressurization X

Knowledge of the physical connections and/or cause-effect relationships between AUTOMATIC DEPRESSURIZATION SYSTEM and the following:

K1.05 Remote shutdown system: Plant-Specific 3.9 41 223002 (SF5 PCIS) Primary Containment Isolation/Nuclear Steam Supply Shutoff X

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

K3.05 Drainage sump levels Ability to manually operate and/or monitor in the control room:

A4.03 Reset system isolations 2.7 3.6 42 43 239002 (SF3 SRV) Safety Relief Valves X

Knowledge of electrical power supplies to the following:

K2.01 SRV solenoids 2.8*

44 259002 (SF2 RWLCS) Reactor Water Level Control X

Knowledge of REACTOR WATER LEVEL CONTROL SYSTEM design feature(s) and/or interlocks which provide for the following:

K4.12 Manual and automatic control of the system 3.5 45 261000 (SF9 SGTS) Standby Gas Treatment X

Knowledge of the effect that a loss or malfunction of the following will have on the STANDBY GAS TREATMENT SYSTEM :

K6.04 Process radiation monitoring 2.9 46 262001 (SF6 AC) AC Electrical Distribution X

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

K3.01 Major system loads 3.5 47 262002 (SF6 UPS)

Uninterruptable Power Supply (AC/DC)

X 2.1.20 Ability to interpret and execute procedure steps.

4.6 48 263000 (SF6 DC) DC Electrical Distribution X

Ability to monitor automatic operations of the D.C.

ELECTRICAL DISTRIBUTION including:

A3.01 Meters, dials, recorders, alarms, and indicating lights 3.2 49

ES-401 7

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

System # / Name K

1 K

2 K

3 K

4 K

5 K

6 A

1 A2 A

3 A

4 G*

K/A Topic(s)

IR 264000 (SF6 EGE) Emergency Generators (Diesel/Jet) EDG X

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

K5.06 Load sequencing Ability to predict and/or monitor changes in parameters associated with operating the EMERGENCY GENERATORS (DIESEL/JET) controls including:

A1.01 Lube oil temperature 3.4 3.0 50 51 300000 (SF8 IA) Instrument Air X

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

A2.01 Air dryer and filter malfunctions 2.9 52 400000 (SF8 CCS) Component Cooling Water X

Knowledge of electrical power supplies to the following:

K2.02 CCW valves 2.9 53 K/A Category Point Totals:

2 3

3 3

3 2

2 2

2 2

2 Group Point Total:

26

ES-401 8

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

System # / Name K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G*

K/A Topic(s)

IR 201001 (SF1 CRDH) CRD Hydraulic NOT SAMPLED 201002 (SF1 RMCS) Reactor Manual Control X

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

A2.03 Select block 2.9 54 201003 (SF1 CRDM) Control Rod and Drive Mechanism NOT SAMPLED 201004 (SF7 RSCS) Rod Sequence Control Not Applicable 201005 (SF1, SF7 RCIS) Rod Control and Information Not Applicable 201006 (SF7 RWMS) Rod Worth Minimizer NOT SAMPLED 202001 (SF1, SF4 RS) Recirculation X

Knowledge of the physical connections and/or cause-effect relationships between RECIRCULATION SYSTEM and the following:

K1.19 Feedwater flow 3.2 55 202002 (SF1 RSCTL) Recirculation Flow Control NOT SAMPLED 204000 (SF2 RWCU) Reactor Water Cleanup X

Knowledge of the effect that a loss or malfunction of the REACTOR WATER CLEANUP SYSTEM will have on following:

K3.02 reactor water level 3.1 56 214000 (SF7 RPIS) Rod Position Information NOT SAMPLED 215001 (SF7 TIP) Traversing In-Core Probe NOT SAMPLED 215002 (SF7 RBMS) Rod Block Monitor NOT SAMPLED 216000 (SF7 NBI) Nuclear Boiler Instrumentation NOT SAMPLED 219000 (SF5 RHR SPC) RHR/LPCI:

Torus/Suppression Pool Cooling Mode X

Knowledge of the operational implications of the following concepts as they apply to RHR/LPCI:

TORUS / SUPPRESSION POOL COOLING MODE:

K5.04 Heat exchanger operation 2.9 57 223001 (SF5 PCS) Primary Containment and Auxiliaries X

Knowledge of electrical power supplies to the following:

K2.09 Drywell cooling fans 2.7 58 226001 (SF5 RHR CSS) RHR/LPCI:

Containment Spray Mode X

Ability to predict and/or monitor changes in parameters associated with operating the RHR/LPCI: CONTAINMENT SPRAY SYSTEM MODE controls including:

A1.10 Emergency generator loading 3.0 59 230000 (SF5 RHR SPS) RHR/LPCI:

Torus/Suppression Pool Spray Mode NOT SAMPLED 233000 (SF9 FPCCU) Fuel Pool Cooling/Cleanup NOT SAMPLED

ES-401 9

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

System # / Name K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G*

K/A Topic(s)

IR 234000 (SF8 FH) Fuel-Handling Equipment X

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

A4.02 Control rod drive system 3.4 60 239001 (SF3, SF4 MRSS) Main and Reheat Steam NOT SAMPLED 239003 (SF9 MSVLCS) Main Steam Isolation Valve Leakage Control NOT SAMPLED 241000 (SF3 RTPRS) Reactor/Turbine Pressure Regulating X

Ability to monitor automatic operations of the REACTOR/TURBINE PRESSURE REGULATING SYSTEM including:

A3.08 Steam bypass valve operation 3.8 61 245000 (SF4 MTGEN) Main Turbine Generator/Auxiliary X

Knowledge of the effect that a loss or malfunction of the following will have on the MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS:

K6.06 Electrical distribution 3.0 62 256000 (SF2 CDS) Condensate NOT SAMPLED 259001 (SF2 FWS) Feedwater NOT SAMPLED 268000 (SF9 RW) Radwaste NOT SAMPLED 271000 (SF9 OG) Offgas X

Knowledge of the physical connections and/or cause-effect relationships between OFFGAS SYSTEM and the following:

K1.06 Main steam system 2.8 63 272000 (SF7, SF9 RMS) Radiation Monitoring NOT SAMPLED 286000 (SF8 FPS) Fire Protection X

2.4.50 Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.

4.2 64 288000 (SF9 PVS) Plant Ventilation NOT SAMPLED 290001 (SF5 SC) Secondary Containment NOT SAMPLED 290003 (SF9 CRV) Control Room Ventilation X

Knowledge of CONTROL ROOM HVAC design feature(s) and/or interlocks which provide for the following:

K4.01 System initiations/reconfiguration: Plant-Specific 3.1 65 290002 (SF4 RVI) Reactor Vessel Internals NOT SAMPLED K/A Category Point Totals:

2 1

1 1

1 1

1 1

1 1

1 Group Point Total:

12

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

Form ES-401-3 Rev 2 Facility: Cooper Nuclear Station Date of Exam: September 2018 Category K/A #

Topic RO SRO-Only IR IR

1.

Conduct of Operations 2.1.3 Knowledge of shift or short term relief turnover practices 3.7 66 2.1.5 Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.

2.9*

67 2.1.34 Knowledge of primary and secondary plant chemistry limits 2.7 68 Subtotal 3

2.

Equipment Control 2.2.6 Knowledge of the process for making changes to procedures 3.0 69 2.2.14 Knowledge of the process for controlling equipment configuration or status.

3.9 75 2.2.22 Knowledge of limiting conditions for operations and safety limits 4.0 71 Subtotal 3

3.

Radiation Control 2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions 3.2 72 2.3.11 Ability to control radiation releases 3.8 73 Subtotal 2

4.

Emergency Procedures /

Plan 2.4.3 Ability to identify post-accident instrumentation 3.7 74 2.4.20 Knowledge of the operational implications of EOP warnings, cautions, and notes 3.8 70 Subtotal 2

Tier 3 Point Total 10

Rev 2 Revision statement:

Rev 1 Corrected IR values for questions 3, 4, 48 Corrected K/As for question 3 from AA1.01 to AA1.02 and question 22 from AA1.01 to AA1.02, Corrected question 56 K/A description from that of K6 (Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR WATER CLEANUP SYSTEM) to that of K3 (Knowledge of the effect that a loss or malfunction of the REACTOR WATER CLEANUP SYSTEM will have on following:)

Replaced original K/A for question 6 295016 Ability to determine and/or interpret the following as they apply to CONTROL ROOM ABANDONMENT (AA2.01: Reactor power) with (AA2.05 Drywell pressure). Recorded on ES-401-4, Record of Rejected K/As (Rev 1).

Corrected wording for Q#27 EK1.02 of 295033 (was from EK2)

Updated all column 1 data for Tier 1 and 2 to match 1021 Rev 11 forms.

Rev 2 swapped question numbers for #23 and #26 to eliminate four Answer Bs in a row swapped question numbers for #70 and #75 to eliminate four Answer Ds in a row

ES-401 BWR Examination Outline Form ES-401-1 Rev 1 Facility: Cooper Nuclear Station Date of Exam: September 2018 Tier Group RO K/A Category Points SRO-Only Points K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G*

Total A2 G*

Total

1.

Emergency &

Abnormal Plant Evolutions 1

3 4

7 2

2 1

3 Tier Totals 5

5 10

2.

Plant Systems 1

3 2

5 2

0 2

1 3

Tier Totals 5

3 8

3. Generic Knowledge and Abilities Categories 1

2 3

4 1

2 3

4 7

2 2

1 2

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

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

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

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

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

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

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

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

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

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

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

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

ES-401 2

Form ES-401-1 Rev 1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (SRO)

E/APE # / Name / Safety Function K

1 K

2 K

3 A

1 A

2 G*

K/A Topic(s)

IR 295001 (APE 1) Partial or Complete Loss of Forced Core Flow Circulation / 1 & 4 NOT SAMPLED 295003 (APE 3) Partial or Complete Loss of AC Power / 6 NOT SAMPLED 295004 (APE 4) Partial or Total Loss of DC Power

/ 6 X

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

AA2.04 System lineups 3.3 76 295005 (APE 5) Main Turbine Generator Trip / 3 NOT SAMPLED 295006 (APE 6) Scram / 1 NOT SAMPLED 295016 (APE 16) Control Room Abandonment / 7 X

Ability to determine and/or interpret the following as they apply to CONTROL ROOM ABANDONMENT:

AA2.06 Cooldown rate 3.5 77 295018 (APE 18) Partial or Complete Loss of CCW

/ 8 NOT SAMPLED 295019 (APE 19) Partial or Complete Loss of Instrument Air / 8 NOT SAMPLED 295021 (APE 21) Loss of Shutdown Cooling / 4 X

G2.2.25 Knowledge of the bases in Technical Specifications for limiting conditions of operations and safety limits 4.2 78 295023 (APE 23) Refueling Accidents / 8 X

G2.4.18 Knowledge of the specific bases for EOPs.

4.0 79 295024 High Drywell Pressure / 5 NOT SAMPLED 295025 (EPE 2) High Reactor Pressure / 3 NOT SAMPLED 295026 (EPE 3) Suppression Pool High Water Temperature / 5 NOT SAMPLED 295027 (EPE 4) High Containment Temperature (Mark III Containment Only) / 5 NOT SAMPLED 295028 (EPE 5) High Drywell Temperature (Mark I and Mark II only) / 5 NOT SAMPLED 295030 (EPE 7) Low Suppression Pool Water Level / 5 NOT SAMPLED 295031 (EPE 8) Reactor Low Water Level / 2 NOT SAMPLED 295037 (EPE 14) Scram Condition Present and Reactor Power Above APRM Downscale or Unknown / 1 X

G2.4.30 Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator 4.1 80 295038 (EPE 15) High Offsite Radioactivity Release Rate / 9 X

Ability to determine and/or interpret the following as they apply to HIGH OFF-SITE RELEASE RATE:

EA2.03 Radiation levels 4.3 81 600000 (APE 24) Plant Fire On Site / 8 NOT SAMPLED

ES-401 3

Form ES-401-1 Rev 1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (SRO)

E/APE # / Name / Safety Function K

1 K

2 K

3 A

1 A

2 G*

K/A Topic(s)

IR 700000 (APE 25) Generator Voltage and Electric Grid Disturbances / 6 X

G2.2.40 Ability to apply Technical Specifications for a system 4.7 82 K/A Category Totals:

0 0

0 0

3 4

Group Point Total:

7

ES-401 4

Form ES-401-1 Rev 1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (SRO)

E/APE # / Name / Safety Function K

1 K

2 K

3 A

1 A

2 G*

K/A Topic(s)

IR 295002 (APE 2) Loss of Main Condenser Vacuum / 3 NOT SAMPLED 295007 (APE 7) High Reactor Pressure / 3 NOT SAMPLED 295008 (APE 8) High Reactor Water Level / 2 NOT SAMPLED 295009 (APE 9) Low Reactor Water Level / 2 X

Ability to determine and/or interpret the following as they apply to LOW REACTOR WATER LEVEL AA2.01 Reactor water level 4.2 83 295010 (APE 10) High Drywell Pressure / 5 NOT SAMPLED 295011 (APE 11) High Containment Temperature (Mark III Containment only) / 5 N/A for CNS (Mark III only) 295012 (APE 12) High Drywell Temperature /

5 NOT SAMPLED 295013 (APE 13) High Suppression Pool Temperature. / 5 NOT SAMPLED 295014 (APE 14) Inadvertent Reactivity Addition / 1 NOT SAMPLED 295015 (APE 15) Incomplete Scram / 1 NOT SAMPLED 295017 (APE 17) Abnormal Offsite Release Rate / 9 NOT SAMPLED 295020 (APE 20) Inadvertent Containment Isolation / 5 & 7 NOT SAMPLED 295022 (APE 22) Loss of Control Rod Drive Pumps / 1 NOT SAMPLED 295029 (EPE 6) High Suppression Pool Water Level / 5 NOT SAMPLED 295032 (EPE 9) High Secondary Containment Area Temperature / 5 X

Ability to determine and/or interpret the following as they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURE:

EA2.03 Cause of high area temperature 4.0 84 295033 (EPE 10) High Secondary Containment Area Radiation Levels / 9 NOT SAMPLED 295034 (EPE 11) Secondary Containment Ventilation High Radiation / 9 NOT SAMPLED 295035 (EPE 12) Secondary Containment High Differential Pressure / 5 NOT SAMPLED 295036 (EPE 13) Secondary Containment High Sump/Area Water Level / 5 NOT SAMPLED 500000 (EPE 16) High Containment Hydrogen Concentration / 5 X

G2.4.41 Knowledge of the emergency action level thresholds and classifications.

4.6 85 K/A Category Point Totals:

0 0

0 0

2 1

Group Point Total:

3

ES-401 5

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

System # / Name K

1 K

2 K

3 K

4 K

5 K

6 A

1 A2 A

3 A

4 G*

K/A Topic(s)

IR 203000 (SF2, SF4 RHR/LPCI)

RHR/LPCI: Injection Mode NOT SAMPLED 205000 (SF4 SCS) Shutdown Cooling NOT SAMPLED 206000 (SF2, SF4 HPCIS)

High-Pressure Coolant Injection X

G2.2.25 knowledge of the bases in Technical Specifications for limiting conditions of operations and safety limits 4.2 86 207000 (SF4 IC) Isolation (Emergency) Condenser N/A for CNS 209001 (SF2, SF4 LPCS)

Low-Pressure Core Spray NOT SAMPLED 209002 (SF2, SF4 HPCS)

High-Pressure Core Spray N/A for CNS 211000 (SF1 SLCS) Standby Liquid Control NOT SAMPLED 212000 (SF7 RPS) Reactor Protection NOT SAMPLED 215003 (SF7 IRM)

Intermediate-Range Monitor NOT SAMPLED 215004 (SF7 SRMS)

Source-Range Monitor NOT SAMPLED 215005 (SF7 PRMS) Average Power Range Monitor/Local Power Range Monitor NOT SAMPLED 217000 (SF2, SF4 RCIC) Reactor Core Isolation Cooling X

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

A2.19 High suppression pool temperature 3.6 87 218000 (SF3 ADS) Automatic Depressurization NOT SAMPLED 223002 (SF5 PCIS) Primary Containment Isolation/Nuclear Steam Supply Shutoff NOT SAMPLED 239002 (SF3 SRV) Safety Relief Valves X

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

A2.02 Leaky SRV 3.2 88 259002 (SF2 RWLCS) Reactor Water Level Control NOT SAMPLED 261000 (SF9 SGTS) Standby Gas Treatment X

G2.2.12 Knowledge of Surveillance procedures 4.1 89 262001 (SF6 AC) AC Electrical Distribution NOT SAMPLED

ES-401 6

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

System # / Name K

1 K

2 K

3 K

4 K

5 K

6 A

1 A2 A

3 A

4 G*

K/A Topic(s)

IR 262002 (SF6 UPS)

Uninterruptable Power Supply (AC/DC)

NOT SAMPLED 263000 (SF6 DC) DC Electrical Distribution NOT SAMPLED 264000 (SF6 EGE) Emergency Generators (Diesel/Jet) EDG X

Ability to (a) predict the impacts of the following on the EMERGENCY GENERATORS (DIESEL/JET) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

A2.07 Loss of offsite power during full load testing 3.7 90 300000 (SF8 IA) Instrument Air NOT SAMPLED 400000 (SF8 CCS) Component Cooling Water NOT SAMPLED K/A Category Point Totals:

0 0

0 0

0 0

0 3

0 0

2 Group Point Total:

5

ES-401 7

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

System # / Name K

1 K

2 K

3 K

4 K

5 K

6 A

1 A2 A 3

A 4

G*

K/A Topic(s)

IR 201001 (SF1 CRDH) CRD Hydraulic NOT SAMPLED 201002 (SF1 RMCS) Reactor Manual Control NOT SAMPLED 201003 (SF1 CRDM) Control Rod and Drive Mechanism NOT SAMPLED 201004 (SF7 RSCS) Rod Sequence Control N/A for CNS 201005 (SF1, SF7 RCIS) Rod Control and Information NOT SAMPLED 201006 (SF7 RWMS) Rod Worth Minimizer NOT SAMPLED 202001 (SF1, SF4 RS) Recirculation NOT SAMPLED 202002 (SF1 RSCTL) Recirculation Flow Control X

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

A2.04 Recirc pump speed mismatch between loops (plant specific) 3.2 91 204000 (SF2 RWCU) Reactor Water Cleanup NOT SAMPLED 214000 (SF7 RPIS) Rod Position Information NOT SAMPLED 215001 (SF7 TIP) Traversing In-Core Probe NOT SAMPLED 215002 (SF7 RBMS) Rod Block Monitor NOT SAMPLED 216000 (SF7 NBI) Nuclear Boiler Instrumentation NOT SAMPLED 219000 (SF5 RHR SPC) RHR/LPCI:

Torus/Suppression Pool Cooling Mode NOT SAMPLED 223001 (SF5 PCS) Primary Containment and Auxiliaries NOT SAMPLED 226001 (SF5 RHR CSS) RHR/LPCI:

Containment Spray Mode NOT SAMPLED 230000 (SF5 RHR SPS) RHR/LPCI:

Torus/Suppression Pool Spray Mode NOT SAMPLED 233000 (SF9 FPCCU) Fuel Pool Cooling/Cleanup NOT SAMPLED 234000 (SF8 FH) Fuel-Handling Equipment X

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

A2.01 Interlock failure 3.7 92 239001 (SF3, SF4 MRSS) Main and Reheat Steam NOT SAMPLED 239003 (SF9 MSVLCS) Main Steam Isolation Valve Leakage Control NOT SAMPLED 241000 (SF3 RTPRS) Reactor/Turbine Pressure Regulating NOT SAMPLED 245000 (SF4 MTGEN) Main Turbine Generator/Auxiliary NOT SAMPLED 256000 (SF2 CDS) Condensate X

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

4.2 93 259001 (SF2 FWS) Feedwater NOT SAMPLED 268000 (SF9 RW) Radwaste NOT SAMPLED 271000 (SF9 OG) Offgas NOT SAMPLED 272000 (SF7, SF9 RMS) Radiation Monitoring NOT SAMPLED 286000 (SF8 FPS) Fire Protection NOT SAMPLED

ES-401 8

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

System # / Name K

1 K

2 K

3 K

4 K

5 K

6 A

1 A2 A 3

A 4

G*

K/A Topic(s)

IR 288000 (SF9 PVS) Plant Ventilation NOT SAMPLED 290001 (SF5 SC) Secondary Containment NOT SAMPLED 290003 (SF9 CRV) Control Room Ventilation NOT SAMPLED 290002 (SF4 RVI) Reactor Vessel Internals NOT SAMPLED K/A Category Point Totals:

0 0

0 0

0 0

0 2

0 0

1 Group Point Total:

3

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

Form ES-401-3 Rev 1 Facility: Cooper Nuclear Station Date of Exam: September 2018 Category K/A #

Topic RO SRO-Only IR IR

1.

Conduct of Operations 2.1.8 Ability to coordinate personnel activities outside the control room 4.1 94 2.1.29 Knowledge of how to conduct system lineups, such as valves, breakers, switches, etc 4.0 95 Subtotal 2

2.

Equipment Control 2.2.5 Knowledge of the process for making design or operating changes to the facility 3.2 96 2.2.17 Knowledge of the process for managing maintenance activities during power operations, such as risk assessments, work prioritization, and coordination with the transmission system operator 3.8 97 Subtotal 2

3.

Radiation Control 2.3.6 Ability to approve release permits 3.8 98 Subtotal 1

4.

Emergency Procedures / Plan 2.4.16 Knowledge of EOP implementation hierarchy and coordination with other support procedures or guidelines such as operating procedures, abnormal operating procedures, and severe accident management guidelines.

4.4 99 2.4.42 Knowledge of emergency response facilities.

3.8 100 Subtotal 2

Tier 3 Point Total 7

Rev 1 Revision Statement:

Rev 1 Corrected IR values for questions 76 and 81 Corrected K/A number for question 81 from E2.03 to EA2.03 Question 92 replaced K/A 234000 A3.01 Ability to monitor automatic operations of the FUEL HANDLING EQUIPMENT including: Crane/Refuel bridge movement (plant specific) with K/A 234000 A2.01 Ability to (a) predict the impacts of the following on the FUEL HANDLING EQUIPMENT; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Interlock failure. Updated tier totals on pages 7 and 1. Recorded on ES-401-4, Record of Rejected K/As, (Rev 1).

Updated all column 1 data for Tier 1 and 2 to match 1021 Rev 11 forms.

Added NOT SAMPLED, as appropriate, to K/A Topic(s) column for Tiers 1 and 2 to be consistent with the provided ES-401-1 RO outline.

ES-401 Record of Rejected K/As Form ES-401-4 Rev 1 Tier / Group Randomly Selected K/A Reason for Rejection RO T1/G1 295016 AA2.01 295016 AA2.05 Because CNS procedures do not contain instructions for observing or determining reactor power during control room abandonment, 295016 AA2.01 (reactor power) was replaced with 295016 AA2.05 (Drywell pressure). Page 1 point totals not affected by this change. (Rev 1)

SRO T2/G2 234000 A3.01 234000 A2.01 Could not develop question at SRO-only level for 234000 A3.01, so replaced with 234000 A2.01 and updated tier totals on pages 7 and 1. (Rev 1)

page 1 of 1 rev 1 ES-301 Administrative Topics Outline Form ES-301-1 Facility: Cooper Nuclear Station Date of Examination:

9/24/2018 Examination Level:

Operating Test Number:

RO SRO CNS 9/2018 Administrative Topic (see Note)

Type Code*

Describe activity to be performed Conduct of Operations R, D A1, Determine Diesel Fuel Oil Availability IAW 5.3EMPWR (SKL034-50-79)

K/A G2.1.23 (4.3/4.4)

Conduct of Operations R, D A2, Perform DW Unidentified Leak Rate Checks IAW daily surveillance 6.LOG.601 (SKL034-50-74)

K/A G2.1.18 (3.6/3.8)

Equipment Control R, N A3, Determine impact of pulling fuse (interpret RPS electrical drawing)

K/A G2.2.15 (3.9/4.3)

Radiation Control N/A Emergency Plan R, M A4, Calculate Liquid Release Curie Content from South CST IAW Emergency Plan Implementing Procedure 5.7.16 (new Version

3)

K/A G2.4.39 (3.9/3.8)

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 & Criteria:

(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (< 3 for ROs; < 4 for SROs & RO retakes) (2)

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

(P)revious 2 exams (< 1; randomly selected) (0)

page 1 of 1 rev 1 ES-301 Administrative Topics Outline Form ES-301-1 Facility: Cooper Nuclear Station Date of Examination:

9/24/2018 Examination Level:

Operating Test Number:

RO SRO CNS 9/2018 Administrative Topic (see Note)

Type Code*

Describe activity to be performed Conduct of Operations R, N A5, Review IST stroke time surveillance for RRMG ventilation damper (plant OE)

K/A G2.1.20 (4.6)

Conduct of Operations R, N A6, Determine required actions for plant chemistry not within limits K/A G2.1.34 (3.5)

Equipment Control R, D A7, Determine Post-Maintenance Testing Requirements IAW 7.0.5 (SKL034-50-18)

K/A G2.2.21 (4.1)

Radiation Control R, M A8, Authorize Emergency Exposure IAW 5.7.12 (Modified version of 12/2015 NRC ILT JPM A8)

K/A G2.3.4 (3.7)

Emergency Plan R, D A9, Determine Protective Action Recommendation IAW 5.7.20 (Tabletop 2)

(SKL034-30-11)

K/A G2.4.44 (4.4)

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

• Type Codes & Criteria:

(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (< 3 for ROs; < 4 for SROs & RO retakes) (2)

(N)ew or (M)odified from bank (> 1) (3)

(P)revious 2 exams (< 1; randomly selected) (0)

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 page 1 of 2 Rev 3 Facility:

Cooper Nuclear Station Date of Examination:

9/24/2018 Exam Level: RO SRO-I SRO-U Operating Test Number:

CNS 9/2018 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. Reset RR scoop tube lockout IAW SOP 2.2.68.1 sect 16 (Alternate Path - RR pump speed increases when lockout is reset, requiring locking out again)

K/A 202002 A1.01 (3.2/3.2)

N, A 1

S2. Restart RWCU following Group 3 isolation IAW SOP 2.2.66 sect 6

K/A 204000 A4.01 (3.1/3.0)

N 2

S3. Lowering DEH pressure setpoint IAW SOP 2.2.77.1 hard card (Alternate Path - During pressure reduction, DEH Throttle Pressure signals fail Bypass valves to open, requiring manual control of Bypass valves.) (modified SKL034-20-107)

K/A 241000 A4.06 (3.9/3.9)

M, L, A 3

S4. Place HPCI in STBY from isolated condition IAW SOP 2.2.33 sect 6 (Alternate Path - HPCI steam leak in Reactor Bldg when HPCI steam supply line reaches 100 psig, must isolate HPCI manually)

K/A 206000 A4.04 (3.7/3.7)

N, A, EN 4

S5. Verify Group 6 Isolation IAW GOP 2.1.22 (SGT Train A did not automatically initiate, drywell vent inlet line failed to isolate, RB Exh Fan did not stop and its discharge valves did not close, requires manual initiation of SGT, manual securing RB Exh Fan, and isolation of valves) (SKL034-20-127)

K/A 223002 A4.01 (3.6/3.5)

D, L 5

S6. Transferring Bus 1F from Emergency Transformer to Bus 1A IAW SOP 2.2.18.1 sect 10 K/A 262001 A4.01 (3.4/3.7)

N, EN 6

S7. Withdraw IRMs during a Start-Up IAW SOP 2.1.1 and Instrumentation Operations Procedure 4.1.2 (Alternate Path -

IRM B sticks and requires being freed,) (SKL034-20-131)

K/A 215003 A1.01 (3.4/3.3)

D, L, A 7

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 page 2 of 2 Rev 3 S8. Align SW crosstie to REC IAW Emergency Procedure 5.2REC Att. 6 K/A 400000 A4.01 (3.1/3.0)

N, L 8

In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U P1. Fill and Vent DG 1(2) Fuel Oil Booster Pump piping IAW SOP 2.2.12 sect 6 (SKL034-11-13)

K/A 264000 K1.05 (3.2/3.3)

D 6

P2. Local Component Alignments to Shift Core Spray A(B) Suction from CST to Torus IAW SOP 2.2.9 sect 12.1.3 (SKL034-11-14)

K/A 209001 K1.02 (3.4/3.4)

D, R, L 2

P3. ASD Control Building Actions, De-energize RPS IAW Emergency Procedure 5.1ASD Att 2 (SKL034-11-11) 295016 AA1.01 (3.8/3.9)

D, E 7

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

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

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

(2) 1/ 1/ 1 (5) 2/ 2/ 1 (6) 3/ 3/ 2 (randomly selected)

(0) 1/ 1/ 1 (1)

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 page 1 of 2 Rev 3 Facility:

Cooper Nuclear Station Date of Examination:

9/24/2018 Exam Level: RO SRO-I SRO-U Operating Test Number:

CNS 9/2018 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. Reset RR scoop tube lockout IAW SOP 2.2.68.1 sect 16 (Alternate Path - RR pump speed increases when lockout is reset, requiring locking out again)

K/A 202002 A1.01 (3.2/3.2)

N, A 1

S2. N/A N/A N/A S3. Lowering DEH pressure setpoint IAW SOP 2.2.77.1 hard card (Alternate Path - During pressure reduction, DEH Throttle Pressure signals fail Bypass valves to open, requiring manual control of Bypass valves.) (modified SKL034-20-107)

K/A 241000 A4.06 (3.9/3.9)

M, L, A 3

S4. Place HPCI in STBY from isolated condition IAW SOP 2.2.33 sect 6 (Alternate Path - HPCI steam leak in Reactor Bldg when HPCI steam supply line reaches 100 psig, must isolate HPCI manually)

K/A 206000 A4.04 (3.7/3.7)

N, A, EN 4

S5. Verify Group 6 Isolation IAW GOP 2.1.22 (SGT Train A did not automatically initiate, drywell vent inlet line failed to isolate, RB Exh Fan did not stop and its discharge valves did not close, requires manual initiation of SGT, manual securing RB Exh Fan, and isolation of valves) (SKL034-20-127)

K/A 223002 A4.01 (3.6/3.5)

D, L 5

S6. Transferring Bus 1F from Emergency Transformer to Bus 1A IAW SOP 2.2.18.1 sect 10 K/A 262001 A4.01 (3.4/3.7)

N, EN 6

S7. Withdraw IRMs during a Start-Up IAW SOP 2.1.1 and Instrumentation Operations Procedure 4.1.2 (Alternate Path -

IRM B sticks and requires being freed,) (SKL034-20-131)

K/A 215003 A1.01 (3.4/3.3)

D, L, A 7

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 page 2 of 2 Rev 3 S8. Align SW crosstie to REC IAW Emergency Procedure 5.2REC Att. 6 K/A 400000 A4.01 (3.1/3.0)

N, L 8

In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U P1. Fill and Vent DG 1(2) Fuel Oil Booster Pump piping IAW SOP 2.2.12 sect 6 (SKL034-11-13)

K/A 264000 K1.05 (3.2/3.3)

D 6

P2. Local Component Alignments to Shift Core Spray A(B) Suction from CST to Torus IAW SOP 2.2.9 sect 12.1.3 (SKL034-11-14)

K/A 209001 K1.02 (3.4/3.4)

D, R, L 2

P3. ASD Control Building Actions, De-energize RPS IAW Emergency Procedure 5.1ASD Att 2 (SKL034-11-11) 295016 AA1.01 (3.8/3.9)

D, E 7

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

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

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

(2) 1/ 1/ 1 (5) 2/ 2/ 1 (5) 3/ 3/ 2 (randomly selected)

(0) 1/ 1/ 1 (1)

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

Cooper Nuclear Station Date of Examination:

9/24/2018 Exam Level: RO SRO-I SRO-U Operating Test Number:

CNS 9/2018 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. Reset RR scoop tube lockout IAW SOP 2.2.68.1 sect 16 (Alternate Path - RR pump speed increases when lockout is reset, requiring locking out again)

K/A 202002 A1.01 (3.2/3.2)

N, A 1

S2. N/A N/A N/A S3. N/A N/A N/A S4. Place HPCI in STBY from isolated condition IAW SOP 2.2.33 sect 6 (Alternate Path - HPCI steam leak in Reactor Bldg when HPCI steam supply line reaches 100 psig, must isolate HPCI manually)

K/A 206000 A4.04 (3.7/3.7)

N, A, EN 4

S5. N/A N/A N/A S6. Transferring Bus 1F from Emergency Transformer to Bus 1A IAW SOP 2.2.18.1 sect 10 K/A 262001 A4.01 (3.4/3.7)

N, EN 6

S7. N/A N/A N/A S8. N/A N/A N/A In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U P1. N/A N/A N/A P2. Local Component Alignments to Shift Core Spray A(B) Suction from CST to Torus IAW SOP 2.2.9 sect 12.1.3 (SKL034-11-14)

K/A 209001 K1.02 (3.4/3.4)

D, R, L 2

P3. ASD Control Building Actions, De-energize RPS IAW Emergency Procedure 5.1ASD Att 2 (SKL034-11-11) 295016 AA1.01 (3.8/3.9)

D, E 7

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 page 2 of 2 rev 1 All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions, all five SRO-U systems must serve different safety functions, and in-plant systems and functions may overlap those tested in the control room.

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

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

(2) 1/ 1/ 1 (1) 2/ 2/ 1 (3) 3/ 3/ 2 (randomly selected)

(0) 1/ 1/ 1 (1)

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 1 Page 1 of 51 Rev. 3 Facility: Cooper Nuclear Station Scenario No.: 1 Op-Test No.: CNS 9/2018 Examiners: ____________________________ Operators:

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

1. Restore LT-59D to service in RVLCS using HMI

2. Raise reactor power by withdrawing control rods

3. Respond to IRM E upscale during startup

4. Respond to single rod drifting in

5. Respond to loss of Battery Room Exhaust Fans

6. Respond to MSL A leak in the steam tunnel, manual scram, ATWS Level/Power control

7. Respond to RHRSW valve MO-89A(B) loss of power

8. SLC Pump A failure to start Initial Conditions: Plant operating at 5% power during startup Inoperable Equipment: none Turnover:

The plant is at 5% power.

Planned activities for this shift are:

Restore LT-59D to service in RVLCS per Procedure 4.4.1 section 9.

Withdraw control rods IAW Procedure 10.13 and the rod sequence to establish 20-25% bypass valve position.

Continue startup IAW Procedures 2.1.1.

Scenario Notes:

This is a new scenario.

Validation Time: 90 minutes

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 1 Page 2 of 51 Rev. 3 Event No.

Malf. No.

Event Type Event Description 1

N/A N

(BOP,CRS)

Restore LT-59D to service in RVLCS using HMI 2

N/A R

(ATC,CRS)

Raise reactor power by withdrawing control rods 3

nm05e I

(ATC,CRS)

IRM E fails upscale 4

rd03h C

(ATC,CRS)

A (CREW)

TS (CRS)

Control rod 26-07 drifts in 5

(Overrides)

ZDIHVSWEFCIC[1]

ZDIHVSWEFCIA[1]

C (BOP,CRS)

A (CREW)

TS (CRS)

Loss of both Battery Room Exhaust Fans 6

ms03a rp04 rd02a rd02b M (CREW)

MSL A leak in the steam tunnel, failure of Group 1 to automatically isolate (manually scram and Group 1 isolation required)

Hydraulic block ATWS > 3% power, SLC Pump A fail to start (EOP-1A, 5A, 6A, 7A)

CT#1 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 Maximum Safe values in two areas for the same parameter.

(Steam tunnel area will reach MSO temperature first. To meet this CT, crew must isolate MSLs before SW Quad area or SE Quad area reach MSO temperature.)

CT#2 When control rods fail to scram and energy is discharging to the primary containment (e.g.

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

7 sw07a C

(BOP,CRS)

RHRSW valve MO-89A(B) loss of power 8

(Override) zdislcsws1a C

(ATC, CRS)

SLC Pump A failure to start (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal (TS) Tech Spec Critical Task (As defined in NUREG 1021 Appendix D), If an operator or the crew significantly deviates from or fails to follow procedures that effect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the post-scenario review.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 1 Page 3 of 51 Rev. 3 Quantitative Attributes Table Attribute ES-304-1 Target Actual Description Malfunctions after EOP entry 1-2 3

1. Failure of Group 1 to automatically isolate

2. SW-MO-89A(B) loss of power

3. SLC Pump A failure to start Abnormal Events 2-4 2

1. Single control rod drift in

2. Loss of Battery Room Exhaust Fans Major Transients 1-2 1

1. ATWS, power >3%

EOP entries requiring substantive action 1-2 2

1. EOP-5A

2. EOP-6A/7A EOP contingencies requiring substantive action 1/set 1

1. EOP-6A/EOP-7A, Contingency #5 - level/power control EOP based Critical Tasks 2-3 2

1. (CT#1) 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 Maximum Safe values in two areas for the same parameter. (Steam tunnel area will reach MSO temperature first. To meet this CT, crew must isolate MSLs before SW Quad area or SE Quad area reach MSO temperature.)

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

Normal Events N/A 1

1. Restore LT-59D to service in RVLCS using HMI Reactivity Manipulations N/A 1

1. Raise reactor power by withdrawing control rods Instrument/

Component Failures N/A 6

1. IRM E fails upscale

2. Control rod 26-07 drifts in

3. Loss of Battery Room Exhaust Fans

4. Group 1 failure to auto isolate

5. SW-MO-89A(B) loss of power

6. SLC Pump A failure to start Total Malfunctions N/A 6

1.

IRM E fails upscale

2.

Control rod 26-07 drifts in

3.

Loss of Battery Room Exhaust Fans

4.

Group 1 failure to auto isolate

5.

SW-MO-89A(B) loss of power

6.

SLC Pump A failure to start Top 10 systems and operator actions important to risk that are tested:

Reactor Protection System, Primary Containment Isolation System, RHRSW

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 1 Page 4 of 51 Rev. 3 SCENARIO

SUMMARY

The plant is operating at 5% power during startup.

After the crew takes the watch, the BOP restores reactor level transmitter LT-59D to service in RVLCS using an HMI per Procedure 2.1.1.

After LT-59D has been returned to service, the ATC will withdraw control rods IAW the startup rod sequence to establish Bypass valves 20-25% open.

During the power ascension, IRM E will fail upscale, resulting in a RPS A half scram and a control rod withdrawal block. The crew will respond IAW the alarm card, bypass IRM E, and reset the half scram IAW procedure 2.1.5. The CRS will determine a potential LCO for IRM E is required.

When response to IRM E failure is complete, control rod 26-07 will drift inward.

The crew will respond IAW the alarm card and procedure 2.4CRD. The ATC will fully insert control rod 26-07. The CRS will enter TS 3.1.6 Condition A and TS 3.1.3 Condition C for control rod 26-07.

After actions for control rod 26-07 drift are complete, Battery Room Exhaust Fan 1C will trip. The BOP will respond IAW the alarm card. The BOP will attempt to start standby fan 1A. Battery Room Exhaust Fan 1A will fail to start. The BOP will then start Essential Control Building Ventilation per Procedure 2.2.38. The CRS will enter TRM LCO T3.8.1 Conditions A and B.

Condition B is exited when Essential Control Building Ventilation has been placed into operation.

When response to loss of Battery Room Exhaust Fans is complete, a leak will develop on MSL A in the steam tunnel. The CRS will enter EOP-5A on high MSL tunnel temperature. Unless the crew manually isolates the leak first, steam tunnel area temperature will rise to the Maximum Safe Operating level.

A Group 1 isolation will fail to automatically occur. The BOP must close Group 1 valves before temperature in a second area rises to the MSO level (CT#1).

When the reactor is scrammed, an ATWS occurs due to hydraulic block of both scram discharge volumes. EOP-6A and 7A are entered via EOP-1A. Reactor power is above 3%. The crew injects SLC and installs the necessary PTMs to bypass interlocks and insert control rods individually via RMCS. With Group 1 valves closed and SRV operating to control reactor pressure, either SLC must be initiated or all controls rods inserted to at least position 02 before Suppression Pool temperature exceeds the Boron Injection Initiation Temperature (BIIT) curve (CT#2). SLC Pump A will fail to start.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 1 Page 5 of 51 Rev. 3 Stop and Prevent is required because reactor power is above 3%. RPV level is intentionally lowered below -60 inches wide range in order to lower core inlet subcooling and lower reactor power. ADS is inhibited when EOP-7A is entered to avert uncontrolled depressurization due to lowering level. HPCI and RCIC are available for RPV level control.

Suppression Pool Cooling will be required due to SRV operation. SW-MO-89A(B), RHR heat exchanger Service Water discharge valve, associated with the first loop of SPC attempted to be placed into service will lose power, requiring the operator to transition to the other RHR loop for SPC.

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

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

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 1 Page 6 of 51 Rev. 3 CRITICAL TASK BASIS Critical Task #1 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 values in two areas for the same parameter. (Steam tunnel area will reach MSO temperature first. To meet this CT, crew must isolate MSLs before SW Quad area or SE Quad area reach MSO temperature.)

Safety Significance EOP-5A directs isolating primary system leaks into secondary containment when a maximum normal operating value is exceeded.

Failing to do so can result in an unnecessary offsite release and endanger plant personnel. Isolating the leak terminates the RCS discharge into secondary containment.

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 Indicators Crew places the control switches for the following isolation valves to CLOSE:

VLV AO 80A, INBOARD STEAM ISOLATION VLV AO 86A, OUTBOARD STEAM ISOLATION VLV AO 80B, INBOARD STEAM ISOLATION VLV AO 86B, OUTBOARD STEAM ISOLATION VLV AO 80C, INBOARD STEAM ISOLATION VLV AO 86C, OUTBOARD STEAM ISOLATION VLV AO 80D, INBOARD STEAM ISOLATION VLV AO 86D, OUTBOARD STEAM ISOLATION MS-MO-74, INBD ISOL VLV MS-MO-77, OUTBD ISOL VLV Performance Feedback Indication for applicable isolation valves Green light illuminates and Red light extinguishes.

Secondary Containment parameter(s) eventually stabilizes and lowers.

RPV and/or associated system parameters indicate leak has been isolated.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 1 Page 7 of 51 Rev. 3 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.

Technical Bases EOP-5A directs that this action be taken when a maximum normal operating value is exceeded. Failing to do so can significantly change the mitigation strategy as an unnecessary release will result and possibly endangering plant personnel.

Justification for the chosen performance limit 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. Isolating the leak before reaching this level will avert the significant thermal transient on the RPV caused by Emergency Depressurization.

BWR Owners Group Appendix App. B, steps SC/T-3, SC/R-1, SC/L-1 Scenario Guide Requirements The scenario must be able to drive at least one secondary containment parameter to its Max Safe value in two plant areas if the crew does not take action to isolate the leak. The crew scramming and reducing RPV pressure to reduce the driving head of the leak should not alone prevent reaching the Max Safe value for a parameter in two plant areas.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 1 Page 8 of 51 Rev. 3 Category EOPS Critical Task #2 When control rods fail to scram and energy is discharging to the primary containment (e.g. SRVs, LOCA), crew injects SLC or inserts all control rods to at least position 02 before exceeding the Boron Injection Initiation Temperature (BIIT) curve.

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

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

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

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

The BIIT is a function of reactor power. If boron injection is initiated before suppression pool temperature reaches the BIIT, emergency RPV depressurization may be precluded at lower reactor power levels. At higher reactor power levels, however, the suppression pool heatup rate may become so high that the Hot Shutdown Boron Weight of boron cannot be injected before suppression pool temperature reaches the Heat Capacity Temperature Limit even if boron injection is initiated early in the event. Since failure-to-scram conditions may present severe plant safety consequences, the requirement to initiate boron injection is independent of any anticipated success of control rod insertion. When attempts to insert control rods satisfactorily achieve reactor shutdown, the requirement for boron injection no longer exists. (Control rod insertion is directed under Step RC/Q-7 concurrently with Step RC/Q-6.)

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

Suppression Pool temperature rising on panel 9-3 indication.

Measurable Performance Indicators Operator manipulates keylocked switch for SLC B pump to START on panel 9-5. (SLC Pump A fails to start in this scenario.)

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

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 1 Page 9 of 51 Rev. 3 Performance Feedback SLC B pump red light illuminated, SLC discharge pressure rising, SLC tank level lowering by 26% on panel 9-5 (from ~80% to below ~54%

for this scenario).

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

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

Justification for the chosen performance limit If boron injection is initiated or all control rods are inserted to position 02 before suppression pool temperature reaches the BIIT, emergency RPV depressurization may be precluded at lower reactor power levels. At higher reactor power levels, however, the suppression pool heatup rate may become so high that the Hot Shutdown Boron Weight of boron cannot be injected before suppression pool temperature reaches the Heat Capacity Temperature Limit even if boron injection is initiated early in the event. Since failure-to-scram conditions may present severe plant safety consequences, the requirement to initiate boron injection is independent of any anticipated success of control rod insertion.

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

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

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 2 Page 1 of 50 Rev. 4 Facility: Cooper Nuclear Station Scenario No.: 2 Op-Test No.: CNS 9/2018 Examiners: ____________________________ Operators:

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

1. Shift DEH pumps from Pump B to Pump A

2. Respond to DEH Pump A degraded, failure to maintain pressure

3. Respond to RHR Pump B trip

4. Respond to Turbine Bypass Valve A failing partially open

5. Respond to Turbine Control Valves 1, 3, 4 drift closed, failure of automatic scram/ARI

6. Respond to Control Rods 34-11 and 42-39 failure to scram

7. Respond to CRD Pump B trip

8. Respond to Earthquake, resulting in Torus leak to <9.6 feet Initial Conditions: Plant operating at 100% power near middle of cycle with RHR B operating in SPC Inoperable Equipment: LPCI Loop B due to RHR B in SPC Turnover:

The plant is at 100% power.

RHR B is in Suppression Pool Cooling.

Planned activities for this shift are:

Shift DEH pumps from Pump B to Pump A.

Scenario Notes:

This is a new scenario.

Validation Time: 80 minutes

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 2 Page 2 of 50 Rev. 4 Event No.

Malf. No.

Event Type Event Description 1

N/A N

(BOP,CRS)

Shift DEH pumps from Pump B to Pump A 2

(pump variable) tckehpp2=1410 C

(BOP,CRS)

Oncoming DEH Pump A degraded, failure to maintain pressure 3

rh01b C

(BOP,CRS)

A (CREW)

TS (CRS)

RHR Pump B trip 4

tc07a C (ATC, CRS)

A (CREW)

TS (CRS)

Turbine Bypass Valve A fails partially open

• CT#1 When a loss of feedwater heating 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.

5 tc09a tc09c tc09d rd27 (overrides) zdirpsscrmja1 zdirpsscrmja2 zdirpsscrmjb1 zdirpsscrmjb2 M

(CREW)

Turbine Control Valves 1, 3, 4 drift closed, failure of automatic scram and ARI (manual scram/ARI functional)

• CT#2 When RPS fails to automatically scram the reactor on a valid signal and energy is discharging into primary containment, initiate RPS scram circuits via scram push buttons or Reactor Mode Switch and/or initiate the ARI System, as necessary, to depressurize the scram air header before average Suppression Pool Temperature exceeds 110°F.

6 rd153411 rd154239 C

(ATC,CRS)

Control Rods 34-11 and 42-39 fail to scram 7

rp12 C

(ATC,CRS)

CRD Pump B trip 8

sw07a C

(BOP,CRS)

Earthquake, Torus leak to <9.6 feet

• CT#3 When torus water level cannot be maintained above 9.6', crew Emergency Depressurizes by opening 6 SRVs prior to torus water level falling to 9.60. (Anticipating Emergency Depressurization and fully opening Bypass Valves also satisfies this critical task.)

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal (TS) Tech Spec Critical Task (As defined in NUREG 1021 Appendix D), If an operator or the crew significantly deviates from or fails to follow procedures that effect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the post-scenario review.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 2 Page 3 of 50 Rev. 4 Quantitative Attributes Table Attribute ES-304-1 Target Actual Description Malfunctions after EOP entry 1-2 3

1. Control Rods 34-11 and 42-39 fail to scram

2. CRD Pump B trip

3. Torus leak Abnormal Events 2-4 2

1. RHR Pump B trip

2. Turbine Bypass Valve A fails partially open Major Transients 1-2 1

1. Turbine Control Valves 1, 3, 4 drift closed, failure of automatic scram and ARI EOP entries requiring substantive action 1-2 2

1. EOP-1A

2. EOP-3A EOP contingencies requiring substantive action 1/set 1

1. EOP-6A/EOP-7A, Contingency #5 - level/power control EOP based Critical Tasks 2-3 3

1. (CT#1) When a loss of feedwater heating 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 RPS fails to automatically scram the reactor on a valid signal and energy is discharging into primary containment, initiate RPS scram circuits via scram push buttons or Reactor Mode Switch and/or initiate the ARI System, as necessary, to depressurize the scram air header before average Suppression Pool Temperature exceeds 110°F.

3. (CT#3) When torus water level cannot be maintained above 9.6', crew Emergency Depressurizes by opening 6 SRVs prior to torus water level falling to 9.60.

(Anticipating Emergency Depressurization and fully opening Bypass Valves also satisfies this critical task.)

Normal Events N/A 1

1. Shift DEH pumps from Pump B to Pump A Reactivity Manipulations N/A 1

1. Raise power to 100% using Reactor Recirc Instrument/

Component Failures N/A 6

1. Oncoming DEH Pump A degraded, failure to maintain pressure

2. RHR Pump B trip

3. Turbine Bypass Valve A fails partially open

4. Control Rods 34-11 and 42-39 fail to scram

5. CRD Pump B trip

6. Torus leak to ~10.5 feet

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 2 Page 4 of 50 Rev. 4 Total Malfunctions N/A 6

1. Oncoming DEH Pump A degraded, failure to maintain pressure

2. RHR Pump B trip

3. Turbine Bypass Valve A fails partially open

4. Control Rods 34-11 and 42-39 fail to scram

5. CRD Pump B trip

6. Torus leak to ~10.5 feet Top 10 systems and operator actions important to risk that are tested:

Reactor Protection System, RHR, Primary Containment (overpressure protection), HPCI

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 2 Page 5 of 50 Rev. 4 SCENARIO

SUMMARY

The plant is operating at 100% power in mid-cycle.

The BOP will shift DEH pumps from B to A in operation. After DEH Pump A has been started, when pump B is secured, DEH pressure will lower due to DEH Pump A degraded. The BOP will respond IAW the alarm card and restart DEH Pump B.

After DEH Pump B has been restarted, RHR Pump B trips. The crew responds IAW the alarm card. Suppression Pool temperature does not require continued operation of SPC. Since RHR Loop B partially drains when RHR Pump B trips, since SP return MOVs are open, the crew will realign valves for LPCI standby and initiate action to fill and vent the loop. The CRS will declare LPCI B, Containment Spray B, and SPC B inoperable IAW TS 3.6.1.9 Condition A and TS 3.6.2.3 Condition A.

When response to RHR Pump B trip is complete, Turbine Bypass Valve A will fail partially open. The BOP will respond IAW the alarm card. Since the amount of extraction steam will be reduced, FW temperature will lower causing reactor power to rise (~1%). The crew will enter 2.4EX-STM and decrease power by lowering Recirc flow IAW 2.1.10 to stabilize power below 100% (CT#1). The CRS will enter TS 3.7.7 Condition A for the failed bypass valve.

After actions for Bypass Valve A are complete, three Turbine governor valves drift closed. RPV pressure will rise, but neither RPS nor ARI will automatically initiate on high RPV pressure. SRVs will open due to high RPV pressure.

The crew must insert a manual scram before average Suppression Pool temperature exceeds 110°F (CT#2). The effect of the Turbine trip on DEH following the scram results in freeing the failed Bypass valve.

Two control rods will fail to scram due to failure of their scram valves to open when the scram air header depressurizes. Although reactor will be trending downscale, the crew will enter EOP-6A and EOP-7A via EOP-1A, since more than one control rod remains withdrawn. The two control rods will be inserted manually by the ATC. CRD Pump B trips before the control rods are inserted, requiring the ATC to start the standby CRD pump to achieve rod insertion.

After the crew has stabilized RPV level and pressure and re-entered EOP-1A following control rod insertion, an earthquake will occur, resulting in a leak from the Torus. The crew will respond IAW procedure 5.1QUAKE and enter EOP-3A on low Torus water level. Before Torus level falls below 11.0 feet, the crew will prevent HPCI operation IAW the hard card. When the CRS determines

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 2 Page 6 of 50 Rev. 4 Torus water level cannot be maintained above 9.6 feet, the crew will emergency depressurize before Torus level falls to 9.6 feet (CT#3).

The exercise ends when emergency depressurization has been performed and RPV level and pressure are being controlled in the specified bands.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 2 Page 7 of 50 Rev. 4 CRITICAL TASK BASIS Critical Task #1 When a loss of feedwater heating 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.

Safety Significance 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 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 B-1/F-1, DEH Trouble and 9-5-2/F-4, RVLC System Logic Initiated, Reactor power rising indicated on IRM/APRM recorders NM-NR-46A-D, SPDS, PMIS, Feedwater temperature lowering on FW Temperature PMIS point NSSRP617 and/or PMIS FW Heating Display.

Measurable Performance Indicators 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.

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

Applicability Any time a reduction in extraction steam causes FW temperature to lower such that reactor power would exceed 100% with no operator intervention.

Justification for the chosen performance limit 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 are not considered intentional. Step 11.1 states It is prohibited to intentionally operate greater than the applicable licensed power limit as determined in Section 10. If core thermal power 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> average calculation exceeds applicable licensed power limit, action shall be taken to ensure subsequent hourly average remains less than or equal to applicable limit. Step 11.4 directs monitoring and maintaining PMIS Point NSSRP643 (running 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> average) and PMIS Point NSSRP645 (running 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> average) below 2419 MWt. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> point is the legal record for plant power level. PMIS point NSSRP640 (15 minute average) was chosen because it is more limiting and provides the most rapid update to core thermal power available.

BWR Owners Group Appendix N/A

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 2 Page 8 of 50 Rev. 4 Scenario Guide Requirements Initial power level must be near 100% and the severity of the loss of feedwater heating must cause reactor power to rise above 100% with no operator intervention.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 2 Page 9 of 50 Rev. 4 Critical Task #2 When RPS fails to automatically scram the reactor on a valid signal and energy is discharging into primary containment, initiate RPS scram circuits via scram push buttons or Reactor Mode Switch and/or initiate the ARI System, as necessary, to depressurize the scram air header before average Suppression Pool Temperature exceeds 110°F.

Safety Significance RPS initiates a reactor scram when one or more monitored parameters exceed their specified limits to preserve the integrity of the fuel cladding and the reactor coolant pressure boundary (RCPB) and minimize the energy that must be absorbed following a loss of coolant accident (LOCA). Failure to effect shutdown of the reactor when a RPS setting has been exceeded, even at low power, would unnecessarily extend the level of degradation of the safety of the plant. This could further degrade into damage to the principle fission product barriers if left unmitigated. The crew is authorized and required by Conduct of Operations to take mitigating actions when automatic safety systems fail to perform their intended function. Action to shut down the reactor is required when RPS and control rod drive systems fail. The postulated DBA against which the primary containment performance is evaluated is the entire spectrum of postulated pipe breaks within the primary containment. Inputs to the safety analyses include initial suppression pool water volume and suppression pool temperature. An initial pool temperature of 95°F is assumed for the safety analyses. Reactor shutdown at a pool temperature of 110°F and vessel depressurization at a pool temperature of 120°F are assumed for the safety analyses. The pool is designed to absorb decay heat and sensible heat but could be heated beyond design limits by the steam generated if the reactor is not shut down.

Cues Annunciators 9-5-2/A-1 (A-2) RX SCRAM CHANNEL A (B) in alarm with RPS remaining energized. Annunciator 9-3-1/A-2 RELIEF VALVE OPEN. Torus water temperature rising on SPDS, Suppression Pool Temperature recorders PC-TR-24 and PC-TR-25 on VBD-J, and annunciators J-1/A-1, SUPPR POOL DIV 1 WATER HIGH TEMP and J-1/A-2, SUPPR POOL DIV 2 WATER HIGH TEMP Measurable Performance Indicators Operator depresses both manual scram pushbuttons, or places the Reactor Mode Switch to SHUTDOWN, or arms and depresses both ARI initiation pushbuttons on panel 9-5.

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

ARI MAN INIT red light lit on panel 9-5.

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

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

Applicability Any time a parameter exceeds a scram setting and RPS fails to trip and energy is discharging into primary containment via SRVs or a primary system leak.

This task is only critical if a manual scram or ARI actuation would be successful in fully inserting control rods.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 2 Page 10 of 50 Rev. 4 Justification for the chosen performance limit Procedure 2.0.3, Conduct of Operations requires upon recognition of a failure of automatic action, the RO shall manually perform those actions necessary to fulfill the safety function and report the completion of the manual action to the CRS as soon as possible. Failure of RPS to automatically function would involve multiple sensor and sensor relay failures. The complexity of an automatic RPS failure would necessarily require a short amount of time to diagnose and validate using control room indications.

BWR Owners Group Appendix App. B, step RC-1 Scenario Guide Requirements All automatic RPS and ARI actuation must be defeated.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 2 Page 11 of 50 Rev. 4 Category EOPS Critical Task #3 When torus water level cannot be maintained above 9.6', crew Emergency Depressurizes by opening 6 SRVs prior to torus water level falling to 9.60. (Anticipating Emergency Depressurization and fully opening Bypass Valves also satisfies this critical task.)

Safety Significance The RPV is not permitted to remain at pressure if suppression of steam discharged from the RPV into the drywell cannot be assured.

When the downcomer vent openings are not adequately submerged, any steam discharged from the RPV into the drywell may not condense in the suppression pool before torus pressure reaches unacceptable levels. RPV depressurization is required at or before the point at which this low water level condition occurs. This reduces the amount of energy that may be discharged directly to the torus air space to as low as possible.

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

Measurable Performance Indicators Manipulation of any six SRV controls to OPEN on panel 9-3:

SRV-71A SRV-71B SRV-71E SRV-71G SRV-71H SRV-71C SRV-71D SRV-71F OR if anticipating ED, Indication on DEH HMI all 3 Bypass Valves have been fully opened manually using DEH HMI controls.

Performance Feedback Crew will observe SRV light indication go from green to red, amber pressure switch lights illuminate, (OR if opening Bypass Valves in anticipation of ED, Bypass Valve positions on DEH HMI indicate 100%) reactor pressure lowering on SPDS and panel 9-3 and 9-5 meters and recorders, and SRV tailpipe temperatures rise on recorder MS-TR-166.

Applicability EOP-3A conditions with torus water level lowering.

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

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

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 2 Page 12 of 50 Rev. 4 Scenario Guide Requirements The scenario should be designed such that it will be apparent to the crew within 10 minutes that they will be unable to stop the drop in suppression pool level (e.g. torus leak) This includes any necessary reports from the field. The initial report from the field should include the elevation of the leak in the torus (e.g. for this CT, at least below

9) and should state the leak is unisolable. The rate of fall in suppression pool level should give the crew at least 20 minutes from the start of the major event before reaching 9.6 feet. This is because preventing HPCI operation and scramming will also be required before 9.6 feet, normally.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 3 Page 1 of 47 Rev. 4 Facility: Cooper Nuclear Station Scenario No.: 3 Op-Test No.: CNS 9/2018 Examiners: ____________________________ Operators:

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

1. Shift TEC pumps from Pump B to Pump C

2. Respond to CRD flow controller fail high

3. Respond to ESST loss of power

4. Respond to SRV-71A failing open

5. Respond to Reactor Recirc discharge line break with degraded low pressure ECCS

6. Respond to failure of 4160V Buses 1A and 1B to transfer to SSST, with failure of DG1 to auto start and RHR Pumps C and D trip

7. Respond to failure of HPCI to auto start

8. Respond to failure of Core Spray A injection valve to automatically open Initial Conditions: Plant operating at 100% power near end of cycle Inoperable Equipment: Core Spray Pump B Turnover:

The plant is at 100% power.

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

TEC Pumps A and B are in service.

Planned activities for this shift are:

Shift TEC pumps from Pump B to Pump C.

Scenario Notes:

This is a new scenario.

Validation Time: 70 minutes

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 3 Page 2 of 47 Rev. 4 Event No.

Malf. No.

Event Type Event Description 1

N/A N (BOP,CRS)

Shift TEC pumps from Pump B to Pump C 2

(override) zaicrdfc301[2]

I (ATC,CRS)

CRD flow controller fail high 3

ed06 C (BOP,CRS)

A (CREW)

TS (CRS)

ESST loss of power 4

ad06c C (ATC,BOP,CRS)

A (CREW)

TS (CRS)

SRV-71A fails open

• CT#1 When a SRV fails open, prior to torus bulk temperature reaching 110°F, close the SRV.

5 rr20a rh01b rh01d M

(CREW)

Reactor Recirc A line break with degraded ECCS

• 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.)

6 ed03a ed03a dg01a C (BOP,CRS) 4160V Buses 1A and 1B fail to transfer to SSST, with failure of DG1 to auto start

• 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

-158 CFZ (TAF) 7 hp01 C (BOP,CRS)

HPCI fails to auto start 8

cs02a C (ATC,CRS)

Core Spray A injection valve fails to automatically open (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal (TS) Tech Spec Critical Task (As defined in NUREG 1021 Appendix D), If an operator or the crew significantly deviates from or fails to follow procedures that effect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the post-scenario review.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 3 Page 3 of 47 Rev. 4 Quantitative Attributes Table Attribute ES-304-1 Target Actual Description Malfunctions after EOP entry 1-2 2

1. 4160V Buses 1A and 1B failure to transfer to SSST

2. Failure of DG1 to auto start Abnormal Events 2-4 2

1. ESST loss of power

2. SRV-71A fails open Major Transients 1-2 2

1. Reactor Recirc A line break with degraded ECCS

2. Loss of Feedwater EOP entries requiring substantive action 1-2 2

1. EOP-1A

2. EOP-3A EOP contingencies requiring substantive action 1/set 2

1. EOP-1A, Contingency #1 - Alternate Level Control

2. EOP-2A, Contingency #2 -Emergency Depressurization EOP based Critical Tasks 2-3 3

1. (CT#1) When a SRV fails open, prior to torus bulk temperature reaching 110°F, close the SRV.

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 LP ECCS systems prior to RPV water level falling below -

158 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. Shift TEC pumps from Pump B to Pump C Reactivity Manipulations N/A 0

1. none Instrument/

Component Failures N/A 7

1. CRD flow controller fail high

2. ESST loss of power

3. SRV-71A fails open

4. 4160V Buses 1A and 1B failure to transfer to SSST

5. Failure of DG1 to auto start

6. HPCI failure to auto start

7. Core Spray A injection valve fails to automatically open

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 3 Page 4 of 47 Rev. 4 Total Malfunctions N/A 6

1.

CRD flow controller fail high

2.

ESST loss of power

3.

SRV-71A fails open

4. 4160V Buses 1A and 1B failure to transfer to SSST

5.

Failure of DG1 to auto start

6.

HPCI failure to auto start

7.

Core Spray A injection valve fails to automatically open Top 10 systems and operator actions important to risk that are tested:

Emergency AC Power, ADS/SRV, HPCI Operator fails to depressurize following high pressure injection failure, Operator fails to manually initiate ECCS, Operator fails to maximize CRD

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 3 Page 5 of 47 Rev. 4 SCENARIO

SUMMARY

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

After the crew takes the watch, the BOP will shift TEC pumps from B to C in operation IAW procedure 2.2.76.

After TEC pumps have been shifted, the CRD flow controller output will fail high.

The crew will enter procedure 2.4CRD. The ATC will take manual control of CRD flow controller IAW 2.4CRD Att. 5 and restore CRD parameters to normal.

When response to CRD flow controller failure is complete, the ESST will lose power. The BOP will respond IAW the alarm cards, and the crew enters procedure 5.3GRID. The BOP removes the ESST from service IAW procedure 2.2.17. The CRS will enter TS 3.8.1 Condition A for the failed offsite AC source.

After actions for the ESST are complete, SRV-71A fails open. The crew enters abnormal procedure 2.4SRV, lowers power to below 90% with Reactor Recirc, and inhibits ADS (CT#1). The valve closes. The CRS declares the valve inoperable per LCO 3.5.1. Conditions E and F and declares ADS trips systems inoperable per TS 3.3.5.1 Conditions F and G.

TS 3.3.5.1 may be exited when fuses for SRV 71A are removed and ADS Inhibit switches are returned to normal.

After actions for SRV-71A are complete, a leak develops on Reactor Recirc loop A. Drywell pressure quickly rises to the scram and ECCS initiation setpoint. RHR Pumps C and D trip upon initiation. HPCI fails to automatically start. If the crew manually starts HPCI, it will trip shortly after it is started and will be unrecoverable. The crew will enter EOP-1A and EOP-3A on high drywell pressure.

When the Main Generator trips following the scram, 4160V Buses 1A and 1B fail to transfer from the NSST to the SSST, resulting in loss of Condensate/Feedwater. DG1 fails to automatically start, so 4160V emergency Bus 1F is de-energized, resulting in loss of numerous control panel indicators and Div 1 powered low pressure ECCS. The crew must manually start DG1 from panel C (CT#2).

RCIC is available for high pressure injection but will not overcome the leak rate. The CRS will exercise the Alternate Level Control leg of EOP-1A. The crew will maximize CRD and inject SLC for level control. RPV level will slowly fall below TAF (-158 on Corrected Fuel Zone, CFZ). When RPV level goes below TAF, the crew must begin emergency depressurization IAW EOP-2A

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 3 Page 6 of 47 Rev. 4 using SRVs before level falls below the Minimum Steam Cooling Reactor Water Level, -183 CFZ (CT#3).

LPCI Pumps A and B and Core Spray A are available to restore level.

However, CS A injection MOV 12A will not automatically open when RPV pressure falls below the pressure permissive, but the crew can open it from panel 9-3.

The exercise ends when emergency depressurization is complete and level is above TAF and being restored to the normal band.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 3 Page 7 of 47 Rev. 4 CRITICAL TASK BASIS Critical Task #1 When a SRV fails open, prior to torus bulk temperature reaching 110°F, close the SRV.

Safety Significance Closing the SRV or shutting down the reactor before 110°F in the Suppression Pool ensures containment design limits due to heat addition to the suppression pool will not be exceeded. 110°F in the Suppression Pool is both the Technical Specification limit and EOP-3A limit for effecting a reactor scram. Tech Spec 3.6.2.1 requires that the Reactor Scram be inserted at 110°F. This requirement ensures that the unit will be shut down at > 110°F. The pool is designed to absorb decay heat and sensible heat but could be heated beyond design limits by the steam generated if the reactor is not shut down (TS Basis). Per PSTGs, the lowest temperature of the Boron Injection Initiation Temperature (BIIT) is specified as the action level (110°F). A single value instead of a graph implements the BIIT in this step to simplify the guideline. The BIIT specifies the suppression pool temperature before which boron injection must be started. It is the greater of:

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

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

The BIIT is a function of reactor power. It is utilized to establish a requirement for boron injection following a failure-to-scram. If boron injection is initiated before suppression pool temperature reaches the BIIT, emergency RPV depressurization may be precluded at lower reactor power levels. At higher reactor power levels, however, the suppression pool heatup rate may become so high that the HSBW cannot be injected before suppression pool temperature reaches the Heat Capacity Temperature Limit even if boron injection is initiated early in the event. Entering the RPV Control guideline at Step RC-1 ensures that, if possible, the reactor is scrammed before boron injection is required and in anticipation of possible RPV depressurization in Step SP/T-3.

Closing the SRV before 110°F precludes the need for manual RPS actuation.

Cues SRV open indications (solenoid lights, tailpipe pressure light, tailpipe temperature).

Step reduction in turbine generator load and steam flow.

Rising suppression pool temperatures on panel 9-3 and SPDS.

Measurable Performance Indicators Operator closes the SRV IAW 2.4SRV by placing SRV control switch to CLOSE or ADS Inhibit switches to INHIBIT on panel 9-3 before exceeding 110°F in the Suppression Pool.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 3 Page 8 of 47 Rev. 4 Performance Feedback SRV tailpipe pressure lowering (SRV amber light off), MSL steam flow rising, SRV solenoid lights (green on, red off), step increase in turbine generator load and steam flow Applicability Anytime when a SRV fails open and the actions addressed in Procedure 2.4SRV would be effective in closing the valve OR EOP-3A conditions when actions taken IAW 2.4SRV are ineffective or not attempted.

Justification for the chosen performance limit 110°F is both the EOP-3A step SP/T-2 limit and the TS 3.6.2.1 limit for reactor shutdown to limit heat addition to the suppression pool.

Closing the failed open SRV would also terminate heat addition to the suppression pool.

BWR Owners Group Appendix App. B, step SP/T-2 Scenario Guide Requirements Initial suppression pool temperature must be low enough, combined with the SRV open severity, to allow the crew reasonable time to perform applicable steps of 2.4SRV. To ensure the CT is based solely on crew action, the SRV should be made to close when the crew places its control switch back to auto or inhibits ADS on panel 9-3.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 3 Page 9 of 47 Rev. 4 Critical Task #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 and/or closes DG-1 output breaker to energize LP ECCS systems prior to RPV water level falling below -158 CFZ (TAF)

Safety Significance Failure to recognize the auto start not occurring and energizing of the safety bus, and failure to take manual action per Procedure 5.3EMPWR will result in unavailability 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 Measurable Performance Indicators Manipulation of controls as required to energize Div 1 AC emergency bus from panel C:

Operator places DIESEL GEN 1 BKR EG1 to CLOSE on panel C Performance Feedback Crew will observe light indication for equipment powered by Division 1 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 auto start or energize its bus.

This is only applicable if manual action from the Control Room would be effective in energizing the bus.

Justification for the chosen performance limit Attempting to start ECCS systems must be performed to determine their availability by the time TAF is reached in order to properly implement EOP-1A decision steps regarding restoring and maintaining RPV level.

BWR Owners Group Appendix App. B, Contingency#1 Scenario Guide Requirements 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. (The LOCA malfunction severity may be ramped initially, but it should reach its final severity within approximately the first 3 minutes.)

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 3 Page 10 of 47 Rev. 4 Critical Task #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.)

Safety Significance 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 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.

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

SRV-71B SRV-71E SRV-71G SRV-71H SRV-71C SRV-71D SRV-71F Performance Feedback Crew will observe SRV light indication go from green to red, amber pressure switch lights illuminate, reactor pressure lowering on SPDS 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).

Justification for the chosen performance limit 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 in the judgment of the CRS, level cannot be maintained above -183 CFZ. Since it is intended for the scenario supporting this CT to, early in the event, clearly indicate no high pressure injection systems can be made available to reverse the lowering level trend, the crew will have time to communicate and open 6 SRVs before -

183 CFZ.

BWR Owners Group Appendix App. B, Contingency#1

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 3 Page 11 of 47 Rev. 4 Scenario Guide Requirements 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 time level lowers to -100 CFZ to allow time to align/realign low 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.)

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 4 Page 1 of 40 Rev. 5 Facility: Cooper Nuclear Station Scenario No.: 4 Op-Test No.: CNS 9/2018 Examiners: ____________________________ Operators:

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

1. Secure RHRSWBP B

2. Raise power to 77% using Reactor Recirc

3. Respond to APRM F INOP signal

4. Respond to Service Water Pump D trip

5. Respond to Recirc Pump B oil leak/high vibration requiring manual trip, Exclusion region entry

6. Respond to binding of Recirc Pump A and Thermal Hydraulic Instability

7. Respond to failure of 480V Bus 1B

8. Respond to RCIC steam leak in Reactor Building with failure of Group 5 to automatically isolate Initial Conditions: Plant operating at 75% power near end of cycle Inoperable Equipment: none Turnover:

The plant is at 75% power during power ascension following a control rod pattern adjustment.

RHRSWBP B is operating following SPC operations.

Planned activities for this shift are:

Secure RHRSWBP B.

Raise power to 77% using Reactor Recirc.

Scenario Notes:

This is a new scenario.

Validation Time: 60 minutes

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 4 Page 2 of 40 Rev. 5 Event No.

Malf. No.

Event Type Event Description 1

N/A N (BOP,CRS)

Secure RHRSWBP B 2

N/A R (ATC,CRS)

Raise power to 77% using Reactor Recirc 3

nm14f I

(ATC,CRS)

APRM F INOP 4

sw01d C (BOP,CRS)

A (CREW)

TS (CRS)

Service Water Pump D trip 5

ad06c C (ATC,BOP,CRS)

A (CREW)

TS (CRS)

Recirc Pump B oil leak/high vibration 6

rr20a rh01b rh01d M (CREW)

Recirc Pump A binding resulting in trip, Thermal Hydraulic Instability, manual scram required

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

7 ad03a ad03b ad06d ad06f C (BOP, ATC, CRS) 480V Bus 1B loss of power 8

ed03a dg01a C (BOP,CRS)

RCIC steam leak in Reactor Building, failure to auto isolate

• CT#2 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 Maximum Safe values in two areas for the same parameter. (RCIC-TE-77C [RCIC Area 859 NE]

will reach MSO temperature first. To meet this CT, crew must close RCIC-MO-15 or RCIC-MO-16 before RHR-TE-99C [Torus Area 885 NNW]

reaches MSO temperature, 195°F.)

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal (TS) Tech Spec Critical Task (As defined in NUREG 1021 Appendix D), If an operator or the crew significantly deviates from or fails to follow procedures that effect the maintenance of basic safety functions, those actions may form the basis of a CT identified in the post-scenario review.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 4 Page 3 of 40 Rev. 5 Quantitative Attributes Table Attribute ES-304-1 Target Actual Description Malfunctions after EOP entry 1-2 2

1. 480V Bus 1B loss of power

2. RCIC steam leak in Rx Bldg, failure to auto isolate Abnormal Events 2-4 2

1.

Service Water Pump D trip

2. Recirc Pump B oil leak/high vibration Major Transients 1-2 1

1. Recirc Pump A binding resulting in trip, Thermal Hydraulic Instability, manual scram required EOP entries requiring substantive action 1-2 2

1. EOP-1A

2. EOP-5A EOP contingencies requiring substantive action 1/set 0

1. None EOP based Critical Tasks 2-3 2

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

2. (CT#2) 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 Maximum Safe values in two areas for the same parameter. (RCIC-TE-77C [RCIC Area 859 NE] will reach MSO temperature first. To meet this CT, crew must close RCIC-MO-15 or RCIC-MO-16 before RHR-TE-99C [Torus Area 885 NNW] reaches MSO temperature, 195°F.)

Normal Events N/A 1

1. Secure RHRSWBP B Reactivity Manipulations N/A 1

1. Raise power to 77% using Reactor Recirc Instrument/

Component Failures N/A 5

1. APRM F INOP

2. Service Water Pump D trip

3. Recirc Pump B oil leak/high vibration

4. 480V Bus 1B loss of power

5. RCIC steam leak in Rx Bldg, failure to auto isolate Total Malfunctions N/A 5

1. APRM F INOP

2. Service Water Pump D trip

3. Recirc Pump B oil leak/high vibration

4. 480V Bus 1B loss of power

5. RCIC steam leak in Rx Bldg, failure to auto isolate Top 10 systems and operator actions important to risk that are tested:

RHRSW, Service Water, Primary Containment Isolation System, RCIC

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 4 Page 4 of 40 Rev. 5 SCENARIO

SUMMARY

The plant is operating at 75% power during power ascension following a control rod pattern adjustment near end of cycle.

After the crew takes the watch, the BOP will secure RHRSWBP B per procedure 2.2.70.

After RHRSWBP B has been secured, the ATC raises power to 77% by raising Recirc pump speed IAW procedure 2.1.10.

After power has been raised, APRM F fails INOP, resulting in a control rod withdrawal block and RPS B half scram. The crew responds IAW the alarm card and bypasses APRM F and resets RPS B IAW procedure 2.1.5. The CRS will determine a potential LCO is required for APRM F.

When response for APRM F INOP is complete, Service Water Pump D will trip.

The BOP will respond IAW alarm cards and start Service Water Pump B. The CRS will declare Service Water Pump D inoperable per TS 3.7.2 Condition A.

DG2 will be declared inoperable per TS 3.8.1 pending the crew placing Service Water Pump B mode switch in standby.

After actions for SW Pump D are complete, RR Pump B will experience high vibration and high motor bearing temperature due to an oil leak. The crew will respond IAW alarm cards and secure RR Pump B. When Recirc Pump B is secured, operation will be in the Stability Exclusion Region of the Power-Flow Map. The crew will insert control rods to exit the region. The crew will begin performing steps for single loop operation IAW procedure 2.2.68.1. The CRS will enter TS 3.4.1 Condition A pending exit of the Stability Exclusion Region and Condition B pending implementation of setpoint changes required for single loop operation.

After the Stability Exclusion region has been exited, RR Pump A binding will occur, resulting in drive motor breaker trip. The ATC will respond by inserting a manual scram IAW procedure 2.4RR due to no RR Pumps operating and power above 1%. Thermal Hydraulic Instability will slowly occur. A manual scram must be inserted before 25% peak to peak flux oscillations occur or APRM high flux trips are received (CT#1). The crew will enter EOP-1A on low reactor water level due to shrink from the scram.

480V Bus 1B will lose power following the scram, requiring entry into procedure 5.3AC480. The BOP will start TEC Pump B due to the resulting loss of power to TEC Pump C. RFP B discharge valve will lose power and will

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 4 Page 5 of 40 Rev. 5 not fully close. The ATC must control RFPs manually using the HMI to maintain level +3 to +54 until the RFP B discharge valve is manually closed.

After the crew has stabilized RPV level and pressure and begun scram recovery, a leak will develop on the RCIC steam supply in the Reactor Building. The CRS will enter EOP-5A on high Reactor Building area temperature. RCIC-TE-77C [RCIC Area 859 NE] temperature will rise to the Maximum Safe Operating level. A Group 5 isolation will fail to automatically occur. The BOP must close Group 5 valves before temperature in another area rises to the MSO level (CT#2).

The exercise ends when RCIC steam supply has been isolated and RPV pressure and level are being controlled in the prescribed bands.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 4 Page 6 of 40 Rev. 5 CRITICAL TASK BASIS Critical Task #1 Manually scram the reactor when both recirculation pumps trip, prior to exceeding 25% peak to peak neutron flux oscillations or APRM auto scram setpoint due to neutron flux oscillations.

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

Cues SRM period alarms.

Oscillating power indications on neutron monitoring instrument.

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

Performance Feedback On panel 9-5:

RPS status lights de-energized.

Reactor power level trend.

Control rods inserted.

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

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

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

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 4 Page 7 of 40 Rev. 5 Scenario Guide Requirements Initial power level following the recirc pump trips, with RPV level in the normal band, should be in the Stability Exclusion Region.

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

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 4 Page 8 of 40 Rev. 5 Critical Task #2 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 Maximum Safe values in two areas for the same parameter. (RCIC-TE-77C [RCIC Area 859 NE] will reach MSO temperature first. To meet this CT, crew must close RCIC-MO-15 or RCIC-MO-16 before RHR-TE-99C [Torus Area 885 NNW] reaches MSO temperature, 195°F.)

Safety Significance EOP-5A directs isolating primary system leaks into secondary containment when a maximum normal operating value is exceeded.

Failing to do so can result in an unnecessary offsite release and endanger plant personnel. Isolating the leak terminates the RCS discharge into secondary containment.

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 Indicators Crew places the control switch for RCIC-MO-15 and/or RCIC-MO-16 to CLOSE.

Performance Feedback Indication for RCIC-MO-15 and/or RCIC-MO-16, as applicable, Green light illuminates and Red light extinguishes.

Secondary Containment parameter(s) eventually stabilizes and lowers.

RPV and/or associated system parameters indicate leak has been isolated.

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.

Technical Bases EOP-5A directs that this action be taken when a maximum normal operating value is exceeded. Failing to do so can significantly change the mitigation strategy as an unnecessary release will result and possibly endangering plant personnel.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 9-2018 Scenario 4 Page 9 of 40 Rev. 5 Justification for the chosen performance limit 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. Isolating the leak before reaching this level will avert the significant thermal transient on the RPV caused by Emergency Depressurization.

BWR Owners Group Appendix App. B, steps SC/T-3, SC/R-1, SC/L-1 Scenario Guide Requirements The scenario must be able to drive at least one secondary containment parameter to its Max Safe value in two plant areas if the crew does not take action to isolate the leak. The crew scramming and reducing RPV pressure to reduce the driving head of the leak should not alone prevent reaching the Max Safe value for a parameter in two plant areas.