ML15351A426

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2015-12 Draft Outlines
ML15351A426
Person / Time
Site: Cooper Entergy icon.png
Issue date: 12/08/2015
From: Vincent Gaddy
Operations Branch IV
To:
Nebraska Public Power District (NPPD)
References
50-298/15-012
Download: ML15351A426 (57)
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Text

ES-401 BWR Examination Outline Form ES-401-1 Facility: Cooper Nuclear Station Date of Exam: November 27,2015 Tier Group RO K/A Category Points SRO-Only Points K K K K K K A A A A G* A2 G* Total 1 2 3 4 5 6 1 2 3 4 Total

1. 1 3 4 4 3 3 3 20 7 Emergency & N/A N/A 2 2 0 2 0 2 1 7 3 Abnormal Plant Evolutions Tier Totals 5 4 6 3 5 4 27 10 1 2 3 2 3 2 3 2 2 2 3 2 26 5 2.

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

3. Generic Knowledge and Abilities Categories 1 2 3 4 10 1 2 3 4 7 3 2 2 3 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 Radiation Control K/A is allowed if the K/A 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 associated 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 shall be selected. 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 importance ratings (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

ES-401 2 Form ES-401-1 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 K K A A G* K/A Topic(s) IR #

1 2 3 1 2 2.2.36 Ability to analyze the effect of maintenance 295001 Partial or Complete Loss of Forced X activities, such as degraded power sources, on the 3.1 1 Core Flow Circulation / 1 & 4 status of limiting conditions for operations. (CFR:

41.10 / 43.2 / 45.13)

AK1. Knowledge of the operational implications of 295003 Partial or Complete Loss of AC / 6 X the following concepts as they apply to PARTIAL 2.7 2 OR COMPLETE LOSS OF A.C. POWER : (CFR:

41.8 to 41.10)

AK1.01 Effect of battery discharge rate on capacity AK2. Knowledge of the interrelations between 295004 Partial or Total Loss of DC Pwr / 6 X PARTIAL OR COMPLETE LOSS OF D.C. POWER 3.1 3 and the following: (CFR: 41.7 / 45.8)

AK2.01 Battery charger AK3. Knowledge of the reasons for the following 295005 Main Turbine Generator Trip / 3 X responses as they apply to MAIN TURBINE 3.8 4 GENERATOR TRIP: (CFR: 41.5 / 45.6)

AK3.07 Bypass valve operation AA1. Ability to operate and/or monitor the following 295006 SCRAM / 1 X as they apply to SCRAM : (CFR: 41.7 / 45.6) 4.2 5 AA1.01 RPS AA2. Ability to determine and/or interpret the 295016 Control Room Abandonment / 7 X following as they apply to CONTROL ROOM 3.3 6 ABANDONMENT : (CFR: 41.10 / 43.5 / 45.13)

AA2.06 Cooldown rate 2.1.23 Ability to perform specific system and 295018 Partial or Total Loss of CCW / 8 X integrated plant procedures during all modes of 4.3 7 plant operation. (CFR: 41.10 / 43.5 / 45.2 / 45.6)

AK2. Knowledge of the interrelations between 295019 Partial or Total Loss of Inst. Air / 8 X PARTIAL OR COMPLETE LOSS OF INSTRUMENT 2.8 8 AIR and the following: (CFR: 41.7 / 45.8)

AK2.04 Reactor water cleanup AK3. Knowledge of the reasons for the following 295021 Loss of Shutdown Cooling / 4 X responses as they apply to LOSS OF SHUTDOWN 3.3 9 COOLING : (CFR: 41.5 / 45.6)

AK3.04 Maximizing reactor water cleanup flow AA1. Ability to operate and/or monitor the following 295023 Refueling Acc / 8 X as they apply to REFUELING ACCIDENTS : (CFR: 2.9 10 41.7 / 45.6)

AA1.02 Fuel pool cooling and cleanup system EK1. Knowledge of the operational implications of 295024 High Drywell Pressure / 5 X the following concepts as they apply to HIGH 4.1 11 DRYWELL PRESSURE : (CFR: 41.8 to 41.10)

EK1.01 Drywell integrity: Plant-Specific EK2. Knowledge of the interrelations between 295025 High Reactor Pressure / 3 X HIGH REACTOR PRESSURE and the following: 3.9 12 (CFR: 41.7 / 45.8)

EK2.09 Reactor power EK3. Knowledge of the reasons for the following 295026 Suppression Pool High Water X responses as they apply to SUPPRESSION POOL 3.8 13 Temp. / 5 HIGH WATER TEMPERATURE: (CFR: 41.5 /

45.6)

EK3.01 Emergency/normal depressurization 295027 High Containment Temperature / 5 NA - MARK III

EA2. Ability to determine and/or interpret the 295028 High Drywell Temperature / 5 X following as they apply to HIGH DRYWELL 3.7 14 TEMPERATURE : (CFR: 41.10 / 43.5 / 45.13)

EA2.03 Reactor water level 2.4.20 Knowledge of the operational implications of 295030 Low Suppression Pool Wtr Lvl / 5 X EOP warnings, cautions, and notes. (CFR: 41.10 / 3.8 15 43.5 / 45.13)

EK1. Knowledge of the operational implications of 295031 Reactor Low Water Level / 2 X the following concepts as they apply to REACTOR 3.8 16 LOW WATER LEVEL : (CFR: 41.8 to 41.10)

EK1.02 Natural circulation: Plant-Specific EK2. Knowledge of the interrelations between 295037 SCRAM Condition Present and X SCRAM CONDITION PRESENT AND REACTOR 4.1 17 Reactor Power Above APRM Downscale or POWER ABOVE APRM DOWNSCALE OR Unknown / 1 UNKNOWN and the following: (CFR: 41.7 / 45.8)

EK2.03 ARI/RPT/ATWS: Plant-Specific EK3. Knowledge of the reasons for the following 295038 High Off-site Release Rate / 9 X responses as they apply to HIGH OFF-SITE 3.7 18 RELEASE RATE: (CFR: 41.5 / 45.6)

EK3.03 Control room ventilation isolation:

Plant-Specific.

AA1 Ability to operate and / or monitor the following 600000 Plant Fire On Site / 8 X as they apply to PLANT FIRE ON SITE: 2.6 19 AA1.08 Fire fighting equipment used on each class of fire AA2. Ability to determine and/or interpret the 700000 Generator Voltage and Electric Grid X following as they apply to GENERATOR VOLTAGE 3.6 20 Disturbances / 6 AND ELECTRIC GRID DISTURBANCES: (CFR:

41.5 and 43.5 / 45.5, 45.7, and 45.8)

AA2.04 VARs outside capability curve K/A Category Totals: 3 4 4 3 3 3 Group Point Total: 20

ES-401 3 Form ES-41-1 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 K K A A G* K/A Topic(s) IR #

1 2 3 1 2 295002 Loss of Main Condenser Vac / 3 AK3. Knowledge of the reasons for the following 295007 High Reactor Pressure / 3 X responses as they apply to HIGH REACTOR 4.0 21 PRESSURE : (CFR: 41.5 / 45.6)

AK3.04 Safety/relief valve operation: Plant-Specific 295008 High Reactor Water Level / 2 AA2. Ability to determine and/or interpret the following 295009 Low Reactor Water Level / 2 X as they apply to LOW REACTOR WATER LEVEL : 3.6 22 (CFR: 41.10 / 43.5 / 45.13)

AA2.02 Steam flow/feed flow mismatch 295010 High Drywell Pressure / 5 295011 High Containment Temp / 5 295012 High Drywell Temperature / 5 295013 High Suppression Pool Temp. / 5 295014 Inadvertent Reactivity Addition / 1 295015 Incomplete SCRAM / 1 295017 High Off-site Release Rate / 9 295020 Inadvertent Cont. Isolation / 5 & 7 AK1. Knowledge of the operational implications of the 295022 Loss of CRD Pumps / 1 X following concepts as they apply to LOSS OF CRD 3.3 23 PUMPS: (CFR: 41.8 to 41.10)

AK1.01 Reactor pressure vs. rod insertion capability EK1. Knowledge of the operational implications of the 295029 High Suppression Pool Wtr Lvl / 5 X following concepts as they apply to HIGH 3.4 24 SUPPRESSION POOL WATER LEVEL : (CFR: 41.8 to 41.10)

EK1.01 Containment integrity 295032 High Secondary Containment Area Temperature / 5 EK3. Knowledge of the reasons for the following 295033 High Secondary Containment Area X responses as they apply to HIGH SECONDARY 3.5 25 Radiation Levels / 9 CONTAINMENT AREA RADIATION LEVELS : (CFR:

41.5 / 45.6)

EK3.02 Reactor SCRAM 295034 Secondary Containment Ventilation High Radiation / 9 EA2. Ability to determine and/or interpret the following 295035 Secondary Containment High X as they apply to SECONDARY CONTAINMENT HIGH 2.8 26 Differential Pressure / 5 DIFFERENTIAL PRESSURE: (CFR: 41.8 to 41.10)

EA2.02 Off-site release rate: Plant-Specific 295036 Secondary Containment High Sump/Area Water Level / 5 2.1.28 Knowledge of the purpose and function of major 500000 High CTMT Hydrogen Conc. / 5 X system components and controls. (CFR: 41.7) 4.1 27 K/A Category Point Totals: 2 0 2 0 2 1 Group Point Total: 7

ES-401 4 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 1 (RO)

System # / Name K K K K K K A A A A G* K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 K3. Knowledge of the effect that a loss or 203000 RHR/LPCI: Injection X malfunction of the RHR/LPCI: INJECTION 4.3 28 Mode MODE (PLANT SPECIFIC) will have on following: (CFR: 41.7 / 45.4)

K3.01 Reactor water level K4. Knowledge of SHUTDOWN COOLING 205000 Shutdown Cooling X SYSTEM (RHR SHUTDOWN COOLING 3.8 29 MODE) design feature(s) and/or interlocks which provide for the following: (CFR: 41.7)

K4.03 Low reactor water level: Plant-Specific A1. Ability to predict and/or monitor changes in 206000 HPCI X parameters associated with operating the HIGH 3.5 30 PRESSURE COOLANT INJECTION SYSTEM controls including: (CFR: 41.5 / 45.5)

A1.03 Condensate storage tank level: BWR-2,3,4 2.4.49 Ability to perform without reference to procedures those actions that X require immediate operation of 4.6 31 system components and controls.

(CFR: 41.10 / 43.2 / 45.6) 207000 Isolation (Emergency) NA NA NA Condenser K6. Knowledge of the effect that a loss or 209001 LPCS X malfunction of the following will have on the 2.8 32 LOW PRESSURE CORE SPRAY SYSTEM :

(CFR: 41.7 / 45.7)

K6.05 ECCS room cooler(s) 209002 HPCS NA A2. Ability to (a) predict the impacts of the 211000 SLC X following on the STANDBY LIQUID CONTROL 4.1 33 SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6)

A2.08 Failure to SCRAM A3. Ability to monitor automatic operations of 212000 RPS X the REACTOR PROTECTION SYSTEM 4.2 34 including: (CFR: 41.7 / 45.7)

A3.06 Main turbine trip: Plant-Specific A4. Ability to manually operate and/or monitor 215003 IRM X in the control room: (CFR: 41.7 / 45.5 to 45.8) 2.9 35 A4.02 CRT display indications: Plant-Specific 3.3 A4.01 IRM recorder indication 2.1.20 Ability to interpret and execute procedure 215004 Source Range Monitor X steps. (CFR: 41.10 / 43.5 / 45.12) 4.6 36 A4. Ability to manually operate and/or monitor 215005 APRM / LPRM X in the control room: (CFR: 41.7 / 45.5 to 45.8) 3.6 37 A4.06 Verification of proper functioning/

operability

K1. Knowledge of the physical connections 217000 RCIC X and/or cause-effect relationships between 2.6 38 REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8)

K1.04 Main condenser K2. Knowledge of electrical power supplies to 218000 ADS X the following: (CFR: 41.7) 3.1 39 K2.01 ADS logic K3. Knowledge of the effect that a loss or 223002 PCIS/Nuclear Steam X malfunction of the PRIMARY CONTAINMENT 3.7 40 Supply Shutoff ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF will have on following:

(CFR: 41.7 / 45.4)

K3.07 Reactor pressure K6. Knowledge of the effect that a loss or malfunction of the following will have on the X PRIMARY CONTAINMENT ISOLATION 3.0 41 SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF : (CFR: 41.7 / 45.7)

K6.05 Containment instrumentation K4. Knowledge of RELIEF/SAFETY VALVES 239002 SRVs X design feature(s) and/or interlocks which provide 3.7 42 for the following: (CFR: 41.7)

K4.09 Manual opening of the SRV K5. Knowledge of the operational implications 259002 Reactor Water Level X of the following concepts as they apply to 3.1 43 Control REACTOR WATER LEVEL CONTROL SYSTEM : (CFR: 41.5 / 45.3)

K5.03 Water level measurement K6. Knowledge of the effect that a loss or 261000 SGTS X malfunction of the following will have on the 2.9 44 STANDBY GAS TREATMENT SYSTEM :

(CFR: 41.7 / 45.7)

K6.04 Process radiation monitoring K6. Knowledge of the effect that a loss or 262001 AC Electrical X malfunction of the following will have on the A.C. 3.6 45 Distribution ELECTRICAL DISTRIBUTION: (CFR: 41.7 /

45.7)

K6.02 Off-site power A1. Ability to predict and/or monitor changes in X parameters associated with operating the A.C. 3.1 46 ELECTRICAL DISTRIBUTION controls including: (CFR: 41.5 / 45.5)

A1.02 Effects of loads when energizing a bus 3.2 A1.05 Breaker lineups K3 Knowledge of the effect that a loss or 262002 UPS (AC/DC) X malfunction of the UNINTERRUPTABLE 2.6 47 POWER SUPPLY (A.C./D.C.) will have on following:(CFR: 41.7 / 45.4)

K3.07 Movement of control rods: Plant-Specific A2. Ability to (a) predict the impacts of the X following on the UNINTERRUPTABLE POWER 2.6 48 SUPPLY (A.C./D.C.) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6)

A2.01 Under voltage

K2. Knowledge of electrical power supplies to 263000 DC Electrical X the following: (CFR: 41.7) 3.1 49 Distribution K2.01 Major D.C. loads A3. Ability to monitor automatic operations of X the D.C. ELECTRICAL DISTRIBUTION 3.2 50 including: (CFR: 41.7 / 45.7)

A3.01 Meters, dials, recorders, alarms, and indicating lights A4. Ability to manually operate and/or monitor 264000 EDGs X in the control room: (CFR: 41.7 / 45.5 to 45.8) 3.7 51 A4.04 Manual start, loading, and stopping of emergency generator: Plant-Specific K1 Knowledge of the connections and / or 300000 Instrument Air X cause effect relationships between 3.1 52 INSTRUMENT AIR SYSTEM and the following:

(CFR: 41.2 to 41.9 / 45.7 to 45.8)

K1.05 Main Steam Isolation Valve air K2. Knowledge of electrical power supplies to 400000 Component Cooling X the following: (CFR: 41.7) 2.9 53 Water K2.01 CCW pumps K/A Category Point Totals: 2 3 3 2 2 3 2 2 2 3 2 Group Point Total: 26

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

System # / Name K K K K K K A A A A G* K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 201001 CRD Hydraulic A3. Ability to monitor automatic 201002 RMCS X operations of the REACTOR MANUAL 2.8 55 CONTROL SYSTEM including:

(CFR: 41.7 / 45.7)

A3.02 Rod movement sequence lights A2. Ability to (a) predict the impacts of 201003 Control Rod and Drive X the following on the CONTROL ROD 3.7 54 Mechanism AND DRIVE MECHANISM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

(CFR: 41.5 / 45.6)

A2.02 Uncoupled rod NA 201004 RSCS 201005 RCIS NA A4. Ability to manually operate and/or 201006 RWM X monitor in the control room: (CFR: 41.7 2.9 56

/ 45.5 to 45.8)

A4.02 Pushbutton indicating switches:

P-Spec(Not-BWR6) 2.4.30 Knowledge of events related to 202001 Recirculation X system operation/status that must bel 2.7 57 reported to internal organizations or external agencies, such as the State,l the NRC, or the transmission system operator. (CFR: 41.10 / 43.5 / 45.11) 202002 Recirculation Flow Control 204000 RWCU 214000 RPIS 215001 Traversing In-Core Probe 215002 RBM 216000 Nuclear Boiler Inst.

K1. Knowledge of the physical 219000 RHR/LPCI: Torus/Pool X connections and/or cause-effect 3.2 58 Cooling Mode relationships between RHR/LPCI:

TORUS/SUPPRESSION POOL COOLING MODE and the following:

(CFR: 41.2 to 41.9 / 45.7 to 45.8)

K1.06 Keep fill system 223001 Primary CTMT and Aux.

K1. Knowledge of the physical 226001 RHR/LPCI: CTMT Spray X connections and/or cause-effect 3.5 59 Mode relationships between RHR/LPCI:

CONTAINMENT SPRAY SYSTEM MODE and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8)

K1.03 LPCI/RHR pumps

230000 RHR/LPCI: Torus/Pool Spray Mode 233000 Fuel Pool Cooling/Cleanup 234000 Fuel Handling Equipment K2. Knowledge of electrical power 239001 Main and Reheat Steam X supplies to the following: (CFR: 41.7) 3.2 60 K2.01 Main steam isolation valve solenoids K3. Knowledge of the effect that a loss 239003 MSIV Leakage Control X or malfunction of the MSIV LEAKAGE 3.3 61 CONTROL SYSTEM will have on following: (CFR: 41.5 / 45.3)

K3.01 HRadiation release to the environment: BWR-4,5,6(P-Spec) 241000 K3.01 Knowledge of the effect that a loss or malfunction of the REACTOR/TURBINE PRESSURE REGULATING SYSTEM will have on following: Reactor power 241000 Reactor/Turbine Pressure X K3. Knowledge of the effect that a loss 4.1 61 Regulator or malfunction of the REACTOR/TURBINE PRESSURE REGULATING SYSTEM will have on following: (CFR: 41.7 / 45.4)

K3.01 Reactor power 245000 Main Turbine Gen. / Aux.

K4. Knowledge of REACTOR 256000 Reactor Condensate X CONDENSATE SYSTEM design 2.9 62 feature(s) and/or interlocks which provide for the following: (CFR: 41.7)

K4.11 Isolation of SJAE's on low flow: Plant-Specific K4.03 Condensate and/or booster 2.8 pump protection.

259001 Reactor Feedwater K5. Knowledge of the operational 268000 Radwaste X implications of the following concepts 2.7 63 as they apply to RADWASTE : (CFR:

41.5 / 45.3)

K5.01 Units of radiation, dose and dose rate 271000 Offgas K6 Knowledge of the effect that a loss 272000 Radiation Monitoring X or malfunction of the following will have 2.8 64 on the RADIATION MONITORING SYSTEM : (CFR: 41.7 / 45.7)

K6.03 A.C. power 286000 Fire Protection 288000 Plant Ventilation A1. Ability to predict and/or monitor 290001 Secondary CTMT X changes in parameters associated with 3.1 65 operating the SECONDARY CONTAINMENT controls including:

(CFR: 41.5 / 45.5)

A1.01 System lineups

290003 Control Room HVAC 290002 Reactor Vessel Internals 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) Form ES-401-3 Facility: Cooper Nuclear Station Date of Exam: November 27, 2015 Category K/A # Topic RO SRO-Only IR # IR #

2.1. 2.1.25 Ability to interpret reference materials, such as 3.9 66 graphs, curves, tables, etc. (CFR: 41.10 / 43.5 /

45.12) 2.1. 2.1.30 Ability to locate and operate components, including 4.4 67 local controls. (CFR: 41.7 / 45.7) 2.1. 2.1.31 Ability to locate control room switches, controls, and 4.6 68 1.

Conduct of indications, and to determine that they correctly Operations reflect the desired plant lineup. (CFR: 41.10 /

45.12) 2.1.

2.1.

2.1.

Subtotal 3 2.2. 2.2.12 Knowledge of surveillance procedures. (CFR: 41.10 3.7 69

/ 45.13) 2.2. 2.2.37 Ability to determine operability and/or availability of 3.6 70 safety related equipment. (CFR: 41.7 / 43.5 / 45.12) 2.

2.2.

Equipment Control 2.2.

2.2.

2.2.

Subtotal 2 2.3.7 Ability to comply with radiation work permit 2.3. 3.5 71 requirements during normal or abnormal conditions.

(CFR: 41.12 / 45.10) 2.3.13 Knowledge of radiological safety procedures pertaining 2.3. to licensed operator duties, such as response to 3.4 72 radiation monitor alarms, containment entry

3. requirements, fuel handling responsibilities, access to Radiation Control locked high-radiation areas, aligning filters, etc. (CFR:

41.12 / 43.4 / 45.9 / 45.10) 2.3.

2.3.

Subtotal 2 2.4. 2.4.18 Knowledge of the specific bases for EOPs. (CFR: 3.3 73 41.10 / 43.1 / 45.13) 2.4. 2.4.31 Knowledge of annunciator alarms, indications, or 4.2 74 response procedures. (CFR: 41.10 / 45.3) 2.4. 2.4.45 Ability to prioritize and interpret the significance of 4.1 75 4.

Emergency each annunciator or alarm. (CFR: 41.10 / 43.5 /

Procedures / Plan 45.3 / 45.12) 2.4.

2.4.

2.4.

Subtotal 3 Tier 3 Point Total 10

ES-401 Record of Rejected K/As Form ES-401-4 Tier / Group Randomly Selected K/A Reason for Rejection 1/1 295005 AK 3.07 Question #4 - Rejected K/A AK 3.08 Feedwater - HPCI actuation: BWR-2 due to being NA for Cooper. Replaced with randomly sampled K/A AK3.07 for same category 295005 3 All generic KAs were screened prior to random selection for Tier 3 as described in NUREG-1021, Section ES-401.D.1.b.

All All generic KAs The generic KAs in the initial NRC generated outline were selected in error. Instead of including the KAs listed in ES-401.D.1.b, they were excluded. The KAs initially selected in error were re-selected by the NRC in order to restore compliance with NUREG-1021. The replacement KAs were randomly selected.

1/1 295024, EK1.01 Question #11 - Rejected EK1.02 because it was specific to a Mark III containment.

CNS has a Mark 1 containment. Selected the remaining K/A in that category, EK1.01, Drywell Integrity, Plant-Specific.

1/1 295028, EA2.03 Question #14 - Rejected K/A 295028 EA2.02 because of oversampling; it was also selected on the SRO exam (Question #82). Replaced with randomly sampled K/A EA2.03 for same category 295028 1/1 295030, G2.4.20 Question #15 - rejected K/A G2.4.34, Knowledge of RO tasks performed outside the main control room duringl an emergency and the resultant operational effects, because could not find any explicitly RO tasks outside of the control room for low SP level. Replaced with randomly selected K/A G2.4.20 for the same category, 295030.

2/1 262002, K3.07 Question #47 - rejected K/A 4.01 for 262002, Knowledge of UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) design feature(s) and/or interlocks which provide for the following: Knowledge of UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.)

design feature(s) and/or interlocks which provide for the following: Transfer from preferred power to alternate power supplies, due to double jeopardy with question

  1. 48. K/A 4.01 was the only K/A in this category, so randomly selected K/A 3.07.

Question #55 - rejected K/A 201004 A3.04 Ability to monitor automatic operations of 2/2 201002, A3.02 the ROD SEQUENCE CONTROL SYSTEM (PLANT SPECIFIC) including: (CFR:

41.7 / 45.7) Annunciator and alarm signals: BWR-4,5 because RSCS is no longer used at CNS. Replaced with similar K/A for the RMCS, 201002, which is used at CNS. Randomly selected K/A A3.02 from subcategoryA3.

Question #35 - Rejected K/A A4.02, Ability to manually operate and/or monitor in the 2/1 215003, A4.01 control room: CRT display indications: Plant-Specific is not applicable to CNS.

Randomly selected from the same subcategory A4.01, IRM recorder indication.

Question #46 Rejected K/A A1.02 AC Electrical Distribution, Ability to predict and/or 2/1 262001 A1.05 monitor changes in parameters associated with operating the A.C. ELECTRICAL DISTRIBUTION controls including: Effects of loads when energizing a bus. Could not write a discriminatory question, randomly selected from the same subcategory A1.05 Breaker lineups.

Question #61 - Rejected K/A 239003 MSIV Leakage Control not applicable to CNS.

2/2 241000, K3.01 Randomly selected new system with same KA statement 241000 K3.01 Knowledge of the effect that a loss or malfunction of the REACTOR/TURBINE PRESSURE REGULATING SYSTEM will have on following: Reactor power Question #62 - Rejected K/A K4.11 - SJAE do not isolate on low flow at CNS.

2/2 256000, K4.03 Randomly selected from the same subcategory K4.03 Condensate and/or booster pump protection.

ES-401 BWR Examination Outline Form ES-401-1 Facility: Cooper Nuclear Station Date of Exam: November 27,2015 Tier Group RO K/A Category Points SRO-Only Points K K K K K K A A A A G* A2 G* Total 1 2 3 4 5 6 1 2 3 4 Total

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

Plant 2 12 0 2 1 3 Systems Tier Totals 38 5 3 8

3. Generic Knowledge and Abilities Categories 1 2 3 4 10 1 2 3 4 7 1 2 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 outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 Radiation Control K/A is allowed if the K/A 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 associated 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 shall be selected. 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 importance ratings (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

ES-401 2 Form ES-401-1 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 K K A A G* K/A Topic(s) IR #

1 2 3 1 2 295001 Partial or Complete Loss of Forced Core Flow Circulation / 1 & 4 295003 Partial or Complete Loss of AC / 6 295004 Partial or Total Loss of DC Pwr / 6 AA2. Ability to determine and/or interpret the 295005 Main Turbine Generator Trip / 3 X following as they apply to MAIN TURBINE 3.6 76 GENERATOR TRIP : (CFR: 41.10 / 43.5 / 45.13)

AA2.07 Reactor water level 2.2.37 Ability to determine operability and/or 295006 SCRAM / 1 X availability of safety related equipment. (CFR: 41.7 / 4.6 77 43.5 / 45.12) 295016 Control Room Abandonment / 7 AA2. Ability to determine and/or interpret the 295018 Partial or Total Loss of CCW / 8 X following as they apply to PARTIAL OR COMPLETE 3.5 78 LOSS OF COMPONENT COOLING WATER :

(CFR: 41.10 / 43.5 / 45.13)

AA2.03 Cause for partial or complete loss 2.1.7 Ability to evaluate plant performance and 295019 Partial or Total Loss of Inst. Air / 8 X make operational judgments based on operating 4.7 79 characteristics, reactor behavior, and instrument interpretation. (CFR: 41.5 / 43.5 / 45.12 / 45.13)

AA2. Ability to determine and/or interpret the 295021 Loss of Shutdown Cooling / 4 X following as they apply to LOSS OF SHUTDOWN 3.6 80 COOLING : (CFR: 41.10 / 43.5 / 45.13)

AA2.04 Reactor water temperature 2.4.41 Knowledge of the emergency action level 295023 Refueling Acc / 8 X thresholds and classifications. (CFR: 41.10 / 43.5 / 4.6 81 45.11) 295024 High Drywell Pressure / 5 295025 High Reactor Pressure / 3 295026 Suppression Pool High Water Temp. / 5 295027 High Containment Temperature / 5 EA2. Ability to determine and/or interpret the 295028 High Drywell Temperature / 5 X following as they apply to HIGH DRYWELL 3.9 82 TEMPERATURE : (CFR: 41.10 / 43.5 / 45.13)

EA2.02 Reactor pressure 295030 Low Suppression Pool Wtr Lvl / 5 295031 Reactor Low Water Level / 2 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown / 1 295038 High Off-site Release Rate / 9 600000 Plant Fire On Site / 8 700000 Generator Voltage and Electric Grid Disturbances / 6 K/A Category Totals: 4 3 Group Point Total: 7

ES-401 3 Form ES-401-1 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 K K A A G* K/A Topic(s) IR #

1 2 3 1 2 295002 Loss of Main Condenser Vac / 3 295007 High Reactor Pressure / 3 295008 High Reactor Water Level / 2 295009 Low Reactor Water Level / 2 295010 High Drywell Pressure / 5 295011 High Containment Temp / 5 295012 High Drywell Temperature / 5 295013 High Suppression Pool Temp. / 5 295014 Inadvertent Reactivity Addition / 1 2.2.25 Knowledge of the bases in Technical 295015 Incomplete SCRAM / 1 X Specifications for limiting conditions for 4.2 83 operations and safety limits. (CFR: 41.5 /

41.7 / 43.2)

AA2. Ability to determine and/or interpret the following 295017 High Off-site Release Rate / 9 X as they apply to HIGH OFF-SITE RELEASE RATE : 3.8 84 (CFR: 41.10 / 43.5 / 45.13)

AA2.05 Meteorological data 295020 Inadvertent Cont. Isolation / 5 & 7 295022 Loss of CRD Pumps / 1 2.4.6 Knowledge of EOP mitigation strategies.

295029 High Suppression Pool Wtr Lvl / 5 X (CFR: 41.10 / 43.5 / 45.13) 4.7 85 295032 High Secondary Containment Area Temperature / 5 295033 High Secondary Containment Area Radiation Levels / 9 295034 Secondary Containment Ventilation High Radiation / 9 295035 Secondary Containment High Differential Pressure / 5 295036 Secondary Containment High Sump/Area Water Level / 5 500000 High CTMT Hydrogen Conc. / 5 K/A Category Point Totals: 1 2 Group Point Total: 3

ES-401 4 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 1 (SRO)

System # / Name K K K K K K A A A A G* K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 203000 RHR/LPCI: Injection Mode 205000 Shutdown Cooling 206000 HPCI 207000 Isolation (Emergency)

Condenser 209001 LPCS 209002 HPCS NA A2. Ability to (a) predict the impacts of the 211000 SLC X following on the STANDBY LIQUID 3.4 86 CONTROL SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

(CFR: 41.5 / 45.6)

A2.03 A.C. power failures 212000 RPS 2.2.42 Ability to recognize system 215003 IRM X parameters that are entry-level 4.6 87 conditions for Technical Specifications. (CFR: 41.7 /

41.10 / 43.2 / 43.3 / 45.3) 215004 Source Range Monitor A2. Ability to (a) predict the impacts of the 215005 APRM / LPRM X following on the AVERAGE POWER RANGE 3.6 88 MONITOR/LOCAL POWER RANGE MONITOR SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

(CFR: 41.5 / 45.6)

A2.05 Loss of recirculation flow signal 217000 RCIC 218000 ADS 223002 PCIS/Nuclear Steam Supply Shutoff 239002 SRVs 2.2.12 Knowledge of surveillance 259002 Reactor Water Level X procedures. (CFR: 41.10 / 45.13) 4.1 89 Control 261000 SGTS 262001 AC Electrical Distribution 262002 UPS (AC/DC)

A2. Ability to (a) predict the impacts of the 263000 DC Electrical X following on the D.C. ELECTRICAL 3.2 90 Distribution DISTRIBUTION ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

(CFR: 41.5 / 45.6)

A2.01 Grounds 264000 EDGs 300000 Instrument Air 400000 Component Cooling Water K/A Category Point Totals: 3 2 Group Point Total: 5

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

System # / Name K K K K K K A A A A G* K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 201001 CRD Hydraulic 201002 RMCS 201003 Control Rod and Drive Mechanism 201004 RSCS 201005 RCIS 201006 RWM 202001 Recirculation 202002 Recirculation Flow Control 204000 RWCU 214000 RPIS 215001 Traversing In-Core Probe 215002 RBM A2. Ability to (a) predict the impacts of 216000 Nuclear Boiler Inst. X the following on the NUCLEAR BOILER 2.9 91 INSTRUMENTATION ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 /

45.6)

A2.12 Instrument isolation valve closures 219000 RHR/LPCI: Torus/Pool Cooling Mode 223001 Primary CTMT and Aux.

226001 RHR/LPCI: CTMT Spray Mode 230000 RHR/LPCI: Torus/Pool Spray Mode 233000 Fuel Pool Cooling/Cleanup 234000 Fuel Handling Equipment 239001 Main and Reheat Steam 239003 MSIV Leakage Control 241000 Reactor/Turbine Pressure Regulator 245000 Main Turbine Gen. / Aux.

256000 Reactor Condensate 259001 Reactor Feedwater 2.4.47 Ability to diagnose and 268000 Radwaste X recognize trends in an accurate and 4.2 92 timely manner utilizing the appropriate control room reference material. (CFR:

41.10 / 43.5 / 45.12)

271000 Offgas A2. Ability to (d) predict the impacts of 272000 Radiation Monitoring X the following on the RADIATION 4.1 93 MONITORING SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

(CFR: 41.5 / 45.6)

A2.10 Loss of coolant accident 286000 Fire Protection 288000 Plant Ventilation 290001 Secondary CTMT 290003 Control Room HVAC 290002 Reactor Vessel Internals K/A Category Point Totals: 2 1 Group Point Total: 3

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Cooper Nuclear Station Date of Exam: November 27, 2015 Category K/A # Topic RO SRO-Only IR # IR #

2.1.23 Ability to perform specific system and integrated plant 2.1.23 4.4 94 procedures during all modes of plant operation.

(CFR: 41.10 / 43.5 / 45.2 / 45.6) 2.1

1. 2.1.

Conduct of Operations 2.1.

2.1.

2.1.

Subtotal 1 2.2.12 Knowledge of surveillance procedures. (CFR: 41.10 2.2.12 4.1 95

/ 45.13) 2.2.25 Knowledge of the bases in Technical Specifications 2.2.25 for limiting conditions for operations and safety 4.2 96 limits. (CFR: 41.5 / 41.7 / 43.2)

2. 2.2.

Equipment Control 2.2.

2.2.

2.2.

Subtotal 2 2.3.4 Knowledge of radiation exposure limits under normal 2.3. or emergency conditions. (CFR: 41.12 / 43.4 / 45.10) 3.7 97 2.3.12 Knowledge of radiological safety principles pertaining 2.3. 3.7 98 to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to

3. locked high-radiation areas, aligning filters, etc. (CFR:

Radiation Control 41.12 / 45.9 / 45.10) 2.3.

2.3.

Subtotal 2 2.4.1 Knowledge of EOP entry conditions and immediate 2.4.1 4.8 99 action steps. (CFR: 41.10 / 43.5 / 45.13) 2.4.30 Knowledge of events related to system 2.4.30 4.1 100 operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system

4. operator. (CFR: 41.10 / 43.5 / 45.11)

Emergency 2.4 Procedures / Plan 2.4.

2.4.

2.4.

Subtotal 2 Tier 3 Point Total 7 7

ES-401 Record of Rejected K/As Form ES-401-4 Tier / Group Randomly Selected K/A Reason for Rejection 3 All generic KAs were screened prior to random selection for Tier 3 as described in NUREG-1021, Section ES-401.D.1.b.

All All generic KAs The generic KAs in the initial NRC generated outline were selected in error. Instead of including the KAs listed in ES-401.D.1.b, they were excluded. The KAs initially selected in error were re-selected by the NRC in order to restore compliance with NUREG-1021. The replacement KAs were randomly selected.

Question # 93 - Rejected K/A 272000 Radiation Monitoring, . Ability to (d) predict the 2/2 272000 A2.10 impacts of the following on the RADIATION MONITORING SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: A2.13 . Low reactor water level during refueling operations.

Rejected this K/A because it is too similar to K/A 295023 Refueling Acc, 2.4.41 Knowledge of the emergency action level thresholds and classifications, used in question #81. A low refueling pool level and elevated radiation levels were used in question #81. The original K/A would necessarily introduce testing of the same knowledge as does question #81, since pool level low enough to affect Rad Monitors would probably result in emergency classification.

Kept 272000 and randomly selected K/A A2.10 Loss of coolant accident by blindly drawing from a jumbled bag of 15 labeled pennies.

1/2 295029, 2.4.6 Question #85 - Rejected K/A 295029 High Suppression Pool Wtr Lvl , 2.4.20 Knowledge of the operational implications of EOP warnings, cautions, and notes.

Could not develop a discriminatory question at the SRO level for this K/A, since no specific EOP caution or note exists for high SP level. Kept K/A 295029 and randomly selected K/A from 2.4 Generics: 2.4.6 Knowledge of EOP mitigation strategies

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Cooper Nuclear Station Date of Examination: 11/30/2015 Examination Level: RO SRO Operating Test Number: CNS 14-01 Administrative Topic (see Note) Type Describe activity to be performed Code*

M Suppression Pool Temperature Calculation Conduct of Operations R (Version 2), SKL034-21-XX(131)

K/A: A295013 AA201 (3.8/4.0); 2.1.20 (4.6/4.6)

N/A Conduct of Operations M Determine Isolation Boundaries (Version 2),

Equipment Control R SKL034-50-XX(57)

K/As 2.2.41 (3.5/3.9)

P Radiation Protection Table Top, SKL034-30-63 D

Radiation Control R K/A 2.3.13 (3.4/3.8)

M Calculate Liquid Release Curie Content R

Emergency Plan (Version 2), SKL034-50-XX(75)

K/A 2.3.11 (3.8/4.3); 2.4.21 (4.0/4.6)

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

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

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

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

ES-301

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Cooper Nuclear Station Date of Examination: 11/30/2015 Examination Level: RO SRO Operating Test Number: CNS 14-01 Administrative Topic (see Note) Type Describe activity to be performed Code*

P Determine Shift Staffing Requirements for Mode D

Conduct of Operations Change, SKL034-20-114 R K/A 2.1.4 (2.9/3.9)

D Determine TS Actions for CRDM Removal, Conduct of Operations R SKL034-50-61 K/A 2.2.40 (3.4/4.7)

M Perform CRS Review of Jet Pump/Recirc Daily Equipment Control R Operability Checks (Version 2) SKL034 XX(113)

K/A 202001 K1.06 3.6/3.6; 2.2.42 (3.9/4.6)

N Authorize Emergency Exposure Radiation Control R SKL034-XX-XX K/A 2.3.4 (3.2/3.7)

N Emergency Classification SKL034-50-XX Emergency Plan S K/A 2.4.41 (2.9/4.6)

R 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)

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

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

ES-301

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Cooper Nuclear Station Date of Examination: 11/30/2015 Exam Level: RO SRO-I SRO-U Operating Test No.: NRC CNS 14-01 Control Room Systems:* 8 for RO; 7 for SRO-I; 2 or 3 for SRO-U System / JPM Title Type Code* Safety Function

a. SKL034-21-XX(45) Operate HPCI in pressure Control Mode M, A, EN, L,S 4 (Alternate Path) K/A 206000 K3.02 (3.8/3.8), A4.01 (3.8/3.7)
b. SKL034-20-XX(107), Establish Cooldown Rate with Bypass M, L, S 3 Valves K/A 241000 A4.06 (3.9/3.9)
c. SKL034-21-133, Respond To Uncoupled Control Rod (Alternate A, D, S 1 Path) K/A 201003 A2.02 (3.7/3.8)
d. SKL034-21-33 Transfer 4160 VAC Bus 1G From DG2 To 4160 D, EN, S 6 VAC Bus 1B K/A:262001 A4.04 (3.6/3.7), 264000 A4.05 (3.6/3.7)
e. Returning RHR Subsystem A to LPCI Standby from Suppression N, S, EN 2 Pool Cooling K/A 203000 A4.01 (4.3/4.1)
f. SKL034-21-61, Defeat Group 1 Low Level Isolation during an D, E, S 5 ATWS (Restoration). K/A 223002 (3.3/3.7)
g. SKL034-21-XX(44), Separation of REC Critical Loops K/A P, D, S, E 8 400000 A4.01 (3.1/3.0); 295018 AA1.03 (3.3/3.4), AK3.07 (3.1/3.2)
h. SKL034-21-XX(74) Withdraw SRMs during a Start-up (Alternate M, A, S 7 Path) K/A 215004.A1.01 (3.0/3.1), A2.03 (3.0/3.3), A4.04 (3.2/3.2)

In-Plant Systems* (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

i. SKL034-10-XX(96), Startup RPS MG Set A (Alternate Path) M, A 7 K/A 212000 A1.01 (2.8/2.9), A2.01 (3.7/3.9)
j. SKL034-10-95, Respond to NBPP Failure (Cntl Bldg Actions, P, A, D, E 6 Alternate Path) K/A 262002 K4.01(3.4/3.4)
k. SKL034-10-50 Fill the Skimmer Surge Tank D, R 9 K/A 233000.A2.02 3.1/3.3
  • All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all five SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
  • Type Codes Criteria for RO / SRO-I / SRO-U

A)lternate path 4-6 / 4-6 / 2-3 (C)ontrol room (D)irect from bank 9/8/4 (E)mergency or abnormal in-plant 1/1/1 (EN)gineered safety feature 1 / 1 / 1 (control room system)

(L)ow-Power / Shutdown 1/1/1 (N)ew or (M)odified from bank including 1(A) 2/2/1 (P)revious 2 exams 3 / 3 / 2 (randomly selected)

(R)CA 1/1/1 (S)imulator

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Cooper Nuclear Station Date of Examination: 11/30/2015 Exam Level: RO SRO-I SRO-U Operating Test No.: NRC CNS 14-01 Control Room Systems:* 8 for RO; 7 for SRO-I; 2 or 3 for SRO-U System / JPM Title Type Code* Safety Function

a. SKL034-21-XX(45) Operate HPCI in pressure Control Mode M, A, EN, L,S 4 (Alternate Path) K/A 206000 K3.02 (3.8/3.8), A4.01 (3.8/3.7)
b. SKL034-20-XX(107), Establish Cooldown Rate with Bypass M, L, S 3 Valves K/A 241000 A4.06 (3.9/3.9)
c. SKL034-21-133, Respond To Uncoupled Control Rod (Alternate A, D, S 1 Path) K/A 201003 A2.02 (3.7/3.8)
d. N/A
e. Returning RHR Subsystem A to LPCI Standby from Suppression N, S, EN 2 Pool Cooling K/A 203000 A4.01 (4.3/4.1)
f. SKL034-21-61, Defeat Group 1 Low Level Isolation during an D, E, S 5 ATWS (Restoration). K/A 223002 (3.3/3.7)
g. SKL034-21-XX(44), Separation of REC Critical Loops K/A P, D, S, E 8 400000 A4.01 (3.1/3.0); 295018 AA1.03 (3.3/3.4), AK3.07 (3.1/3.2)
h. SKL034-21-XX(74) Withdraw SRMs during a Start-up (Alternate M, A, S 7 Path) K/A 215004.A1.01 (3.0/3.1), A2.03 (3.0/3.3), A4.04 (3.2/3.2)

In-Plant Systems* (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

i. SKL034-10-XX(96), Startup RPS MG Set A (Alternate Path) M, A 7 K/A 212000 A1.01 (2.8/2.9), A2.01 (3.7/3.9)
j. SKL034-10-95, Respond to NBPP Failure (Cntl Bldg Actions, P, A, D, E 6 Alternate Path) K/A 262002 K4.01(3.4/3.4)
k. SKL034-10-50 Fill the Skimmer Surge Tank D, R 9 K/A 233000.A2.02 3.1/3.3
  • All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all five SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
  • Type Codes Criteria for RO / SRO-I / SRO-U

A)lternate path 4-6 / 4-6 / 2-3 (C)ontrol room (D)irect from bank 9/8/4 (E)mergency or abnormal in-plant 1/1/1 (EN)gineered safety feature 1 / 1 / 1 (control room system)

(L)ow-Power / Shutdown 1/1/1 (N)ew or (M)odified from bank including 1(A) 2/2/1 (P)revious 2 exams 3 / 3 / 2 (randomly selected)

(R)CA 1/1/1 (S)imulator

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Cooper Nuclear Station Date of Examination: 11/30/2015 Exam Level: RO SRO-I SRO-U Operating Test No.: NRC CNS 14-01 Control Room Systems:* 8 for RO; 7 for SRO-I; 2 or 3 for SRO-U System / JPM Title Type Code* Safety Function

a. SKL034-21-XX(45) Operate HPCI in pressure Control Mode M, A, EN, L,S 4 (Alternate Path) K/A 206000 K3.02 (3.8/3.8), A4.01 (3.8/3.7)
b. N/A
c. N/A
d. N/A
e. Returning RHR Subsystem A to LPCI Standby from Suppression N, S, EN 2 Pool Cooling K/A 203000 A4.01 (4.3/4.1)
f. SKL034-21-61, Defeat Group 1 Low Level Isolation during an D, E, S 5 ATWS (Restoration). K/A 223002 (3.3/3.7)
g. N/A
h. N/A In-Plant Systems* (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)
i. N/A
j. SKL034-10-95, Respond to NBPP Failure (Cntl Bldg Actions, P, A, D, E 6 Alternate Path) K/A 262002 K4.01(3.4/3.4)
k. SKL034-10-50 Fill the Skimmer Surge Tank D, R 9 K/A 233000.A2.02 3.1/3.3
  • All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all five SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
  • Type Codes Criteria for RO / SRO-I / SRO-U A)lternate path 4-6 / 4-6 / 2-3 (C)ontrol room (D)irect from bank 9/8/4 (E)mergency or abnormal in-plant 1/1/1 (EN)gineered safety feature 1 / 1 / 1 (control room system)

(L)ow-Power / Shutdown 1/1/1 (N)ew or (M)odified from bank including 1(A) 2/2/1 (P)revious 2 exams 3 / 3 / 2 (randomly selected)

(R)CA 1/1/1 (S)imulator

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Cooper Nuclear Station Date of Exam: 11/30/2015 Operating Test No.: NRC CNS 14-01 A E Scenarios P V 1 2 3 4 T M P E O I L N CREW CREW POSITION CREW POSITION CREW POSITION T N I T POSITION I

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

T P E R I U RO(R1) RX 1 0 0 1 1 1 0 NOR 0 1 1 2 1 1 1 SRO-I I/C 3 3 2 8 4 4 2 SRO-U MAJ 2 1 1 4 2 2 1 TS NA NA NA NA 0 2 2 RO(R2) RX 1 0 0 1 1 1 0 NOR 0 1 1 2 1 1 1 SRO-I I/C 3 3 2 8 4 4 2 SRO-U MAJ 2 1 1 4 2 2 1 TS NA NA NA NA 0 2 2 RO(R3) RX 0 1 1 1 1 0 NOR 0 0 0 1 1 1 SRO-I I/C 4 4 8 4 4 2 SRO-U MAJ 2 1 3 2 2 1 TS NA NA NA 0 2 2 RO RX 0 1 1 1 1 0 (R4)

NOR 0 0 0 1 1 1 SRO-I I/C 4 3 7 4 4 2 SRO-U MAJ 2 1 3 2 2 1 TS NA NA NA 0 2 2 Instructions:

1. Check the applicant level and enter the operating test number and Form ES-D-1 event numbers for each event type; TS are not applicable for RO applicants. ROs must serve in both the at-the-controls (ATC) and balance-of-plant (BOP) positions. Instant SROs (SRO-I) must serve in both the SRO and the ATC positions, including at least two instrument or component (I/C) malfunctions and one major transient, in the ATC position. If an SRO-I additionally serves in the BOP position, one I/C malfunction can be credited toward the two I/C malfunctions required for the ATC position.
2. Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.5.d) but must be significant per Section C.2.a of Appendix D. (*) Reactivity and normal evolutions may be replaced with additional instrument or component malfunctions on a one-for-one basis.
3. Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicants competence count toward the minimum requirements specified for the applicants license level in the right-hand columns.
4. For licensees that use the ATC operator primarily for monitoring plant parameters, the chief examiner may place SRO-I applicants in either the ATC or BOP position to best evaluate the SRO-I in manipulating plant controls.

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Cooper Nuclear Station Date of Exam: 11/30/2015 Operating Test No.: NRC CNS 14-01 A E Scenarios P V 1 2 3 4 T M P E O I L N CREW CREW POSITION CREW POSITION CREW POSITION T N I T POSITION I

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

T P E (I2) (R2) (R4) R I U RO RX 1 1 1 3 1 1 0 NOR 0 0 1 1 1 1 1 SRO-I (I1) I/C 6 3 5 14 4 4 2 MAJ 2 1 1 4 2 2 1 SRO-U TS 2 NA 2 4 0 2 2 RO RX 1 1 1 3 1 1 0 NOR 0 1 0 1 1 1 1 SRO-I(I2) I/C 6 4 4 14 4 4 2 MAJ 2 1 1 4 2 2 1 SRO-U TS 2 2 NA 4 0 2 2 RO RX 1 1 2 1 1 0 NOR 1 1 2 1 1 1 SRO-I I/C 4 5 9 4 4 2 SRO- MAJ 1 1 2 2 2 1 U(U1)

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

1. Check the applicant level and enter the operating test number and Form ES-D-1 event numbers for each event type; TS are not applicable for RO applicants. ROs must serve in both the at-the-controls (ATC) and balance-of-plant (BOP) positions. Instant SROs (SRO-I) must serve in both the SRO and the ATC positions, including at least two instrument or component (I/C) malfunctions and one major transient, in the ATC position. If an SRO-I additionally serves in the BOP position, one I/C malfunction can be credited toward the two I/C malfunctions required for the ATC position.
2. Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.5.d) but must be significant per Section C.2.a of Appendix D. (*) Reactivity and normal evolutions may be replaced with additional instrument or component malfunctions on a one-for-one basis.
3. Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicants competence count toward the minimum requirements specified for the applicants license level in the right-hand columns.
4. For licensees that use the ATC operator primarily for monitoring plant parameters, the chief examiner may place SRO-I applicants in either the ATC or BOP position to best evaluate the SRO-I in manipulating plant controls.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 1 Page 1 of 2 Facility: Cooper Nuclear Station Scenario No.: 1 Op-Test No.: CNS 14-01 Examiners: ____________________________ Operators: _____________________________

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

1. Raise reactor power during startup by withdrawing control rods.
2. Respond to erratic indication on IRM H requiring bypassing IRM H.
3. Respond to trip of TEC pump A.
4. Respond to trip of RPS MG Set B with trip of SGT A exhaust fan.
5. Respond to RCIC steam leak into Secondary Containment with inability to isolate the leak.
6. Take actions for a low power ATWS.
7. Conduct emergency depressurization due to degrading conditions in secondary containment.

Initial Conditions: Plant startup in progress at 4.5% power and rated pressure.

Inoperable Equipment: None Turnover:

The plant is at 4.5% power and rated pressure during startup from a mid-cycle outage. Planned activities for this shift are:

  • Continue power ascension IAW 2.1.1, Startup Procedure.

Primary Containment inerting is NOT in progress. There is no out of service equipment and integrated risk is GREEN. It is a division 1 work week.

Scenario Notes:

This is a new scenario.

Validation Time: 75 minutes

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 1 Page 2 of 2 Event Malf. No. Event Type Event No. Description Raise Reactor power by withdrawing control rods. (Startup 1 N/A R (ATC) control rod sequence; GOP 2.1.1, Startup Procedure )

I (ATC) Respond to IRM H erratic indication requiring bypass (ARP 9 2 nm05h TS (CRS) 1/E-8)

TEC Pump A trip requiring manual start of standby TEC pump 3 sw07a C (BOP)

(ARP M-2/A-5)

(rf) rp02 C (BOP,ATC)

RPS MG Set B trip with failure of SGT A Exhaust Fan (ARP C-4 (or) A (CREW) 1/F-2, ARP C-1/G-2, 2.1.5, 2.1.22) zdisgtswefre TS (CRS) rc06 C (BOP) rc07 RCIC steam leak in Secondary Ctmt with failure to automatically 5 A(CREW) isolate, failure of RCIC isolation valves MO-15 and MO-16 (rf) rc06a during manual isolation. (EOP-5A)

TS (CRS)

(rf) rc22a Manual scram due to Secondary Ctmt temperature, Hydraulic block ATWS < 3% power (EOP-5A, 1A, 6A, 6B, 7A)

C (ATC) When control rods fail to scram, crew injects SLC and/or 6 rd02a inserts control rods before exiting EOP-6A. (All control M (CREW) rods do not have to be fully inserted to satisfy this critical task; this only requires that the crew is making progress to achieving all rods in by fully inserting at least 5 control rods using RMCS.)

RCIC steam leak propagates during low power ATWS (EOP-5A, 1A, 6A, 6B, 7B) 7 M (CREW) When 2 areas exceed their Max Safe Temperature limit, crew Emergency Depressurizes by opening 6 SRVs before reaching Max Safe Temperature in a third area.

(or) During emergency depressurization, SRV-71G fails to open, 8 C(BOP) zdimssws1e(1) requiring opening a LLS SRV.

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

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

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 1 Page 3 of 2 Quantitative Attributes Table ES-304-1 Attribute Actual Description Target Malfunctions after 1. Hydraulic block ATWS EOP entry 1-2 2 2. SRV-71D failure to open

1. RPS B MG Set trip Abnormal Events 2-4 2 2. RCIC steam leak in Secondary Ctmt
1. ATWS Major Transients 1-2 2 2. RCIC steam line break requiring ED EOP entries requiring 1. EOP-1A,6A,7A substantive action 1-2 2 2. EOP-5A EOP contingencies
1. EOP-6A,7A requiring substantive 0-2 2 2. EOP-6B action
1. Inject SLC/insert control rods EOP based Critical 2. Emergency depressurize when 2 areas exceed Tasks 2-3 2 Max Safe temperature Normal Events N/A 0 N/A Reactivity Manipulations N/A 1 1. Withdraw control rods per the startup sequence
1. IRM H erratic
2. TEC pump A trip
3. RPS MG Set B trip Instrument/

Component Failures N/A 7 4. SGT A exhaust fan trip

5. RCIC isolation valve failure
6. SDV hyGraulic block
7. SRV-71G failure to open
1. IRM H erratic
2. TEC pump A trip
3. RPS MG Set B trip
4. SGT A exhaust fan trip Total Malfunctions N/A 8 5. Failure of RCIC auto isolation
6. RCIC isolation valve breaker trip
7. SDV hydraulic block
8. SRV-71G failure to open Top 10 systems and operator actions important to risk that are tested:

Reactor Protection System Primary Containment Isolation System Automatic Depressurization System(ADS)

Manually Initiate ADS

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 1 Page 4 of 2 SCENARIO

SUMMARY

Startup is in progress at 4.5% power and rated pressure at the beginning of the current fuel cycle when the crew takes the shift. No equipment is out of service.

After the crew takes the watch, the ATC operator will withdraw control rods IAW the control rod withdrawal sequence and the GOP.

While control rods are being withdrawn, IRM H indication will become erratic, simulating a degraded connection undervessel. Control rod blocks will be generated, requiring the ATC to respond IAW the alarm response procedure and bypass IRM H. The CRS will review TS and determine a potential LCO is required for IRM H relative to TS 3.3.1.1 and TRM 3.3.1.

When the Tech Spec assessment is complete, TEC pump A will trip. The BOP operator will respond and start a standby TEC pump, B or C, IAW the alarm response procedure. AOP entry will not be required since TEC pump C remains operating and the heat load on the TEC system is minimal with the Turbine Generator not yet on line.

After actions for TEC are complete, RPS MG Set B will trip, resulting in a half scram, half Group 1, 2, 3, 7 isolations, and a full Group 6 isolation.

RWCU will isolate and the pump will trip. Standby Gas Treatment A exhaust fan will trip. The crew will transfer RPS B to its alternate supply per the alarm response procedure, reset the Div 2 half scram IAW 2.1.5, and enter 2.1.22 to begin recovery from group isolations. The CRS will address TS 3.6.4.3 for SGT A exhaust fan failure. The crew will not be given time to complete all of the actions of 2.1.22, such as restoring Reactor Building HVAC realignments, due to the timing of the scenario.

After RPS B power has been restored and the lead examiner is ready to proceed, a small leak will develop from the RCIC steam supply in the NE Quad area. Secondary containment temperature in this area will rise, and if requested, field reports will inform the crew the leak is from RCIC. If the isolation setpoint is reached before the crew attempts to close RCIC steam supply valves, the automatic isolation will not occur. When the BOP attempts to manually isolate the steam supply, the MOVs will lose power while stroking. The crew will enter EOP-5A, EOP-1A, and scram due to an unisolable steam leak into secondary containment raising area temperature toward the Max Safe limit.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 1 Page 5 of 2 When the reactor is scrammed, a low power ATWS will occur due to hydraulic block of one scram discharge volume, and EOP-6A and 7A will be entered via EOP-1A. Power will be below 3%. The crew will inject SLC and/or install the necessary PTMs to bypass interlocks and drive control rods in individually via RMCS. Bypass valves will be available to control RPV pressure. Feedwater and HPCI will be available for RPV level control.

After the crew has stabilized conditions following the scram, the RCIC steam leak will degrade and cause temperature in the NE Quad and in the NW Quad to exceed the Max Safe limit, requiring emergency depressurization IAW EOP-5A and EOP-6B. When the handswitch for SRV-71G is placed to open, the SRV will remain closed, and the operator should open another SRV in its stead. Condensate will be available for RPV level control when MSCP pressure is reached.

The exercise ends when the reactor has been depressurized, RPV water level is being maintained -183 to +54, and SLC is injecting and/or control rods are being inserted via RMCS and/or scram.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 1 Page 6 of 2 Critical Tasks When control rods fail to scram, crew When 2 areas exceed their Max Safe injects SLC and/or inserts control rods Temperature limit, crew Emergency before exiting EOP-6A. (All control rods Depressurizes by opening 6 SRVs before do not have to be fully inserted to satisfy reaching Max Safe Temperature in a third this critical task; this only requires that area.

the crew is making progress to achieving all rods in by fully inserting at least 5 control rods using RMCS.)

EVENT 6 7 Safety Failure to effect shutdown of the reactor Should secondary containment parameters significance when a RPS setting has been exceeded, exceed their maximum safe operating values even at low power, would unnecessarily in more than one area, the RPV must be extend the level of degradation of the safety depressurized to preclude further degradation.

of the plant. This could further degrade into RPV depressurization places the primary damage to the principle fission product system in its lowest possible energy state, barriers if left unmitigated. The crew is rejects heat to the suppression pool in authorized and required by Conduct of preference to outside the containment, and Operations to take mitigating actions when reduces the driving head and flow of primary automatic safety systems fail to perform their systems that are unisolated and discharging intended function. Action to shut down the into the secondary containment.

reactor is required when RPS and control rod The criteria of "two or more areas" specified drive systems fail. identifies the rise in secondary containment parameters as a wide-spread problem which may pose a direct and immediate threat to secondary containment integrity, equipment located in the secondary containment, and continued safe operation of the plant.

Cueing Manual scram is initiated and numerous SPDS indication for NE and NW Quad areas control rods indicate beyond position 00 and both > 195°F reactor power not downscale on panel 9-5 indications.

Performance Operator manipulates keylocked switches for Manipulation of SRV controls on panel 9-3:

indicator SLC A and B pumps to START on panel 9-5. SRV-71A SRV-71B Operator selects individual control rods by SRV-71E depressing the respective pushbutton on the SRV-71G panel 9-5 matrix and inserts rods by SRV-71H manipulating the emergency in switch on SRV-71C panel 9-5. SRV-71D SRV-71F Performance SLC A and B pumps red lights illuminated on Crew will observe SRV light indication go from feedback panel 9-5. green to red, amber pressure switch lights illuminate, reactor pressure lowering on SPDS Operator selecting and inserting control rods and panel 9-3 and 9-5 meters and recorders, indicated by rod position decreasing to 00 for and SRV tailpipe temperatures rise on selected rod on panel 9-5. recorder MS-TR-166..

Justification There is no time limit for effecting complete Emergency Depressurization is required due to for the chosen reactor shutdown via boron injection or effects of a break spreading into and performance control rod insertion. For the timeframe of potentially affecting safety equipment and limit this low power scenario, containment limits operations in more than one area; however, are not challenged and power oscillations are emergency depressurization is not allowed not experienced. However, if the failure to until the second area exceeds its Max Safe scram EOP were to be exited, other limit. Before the Max Safe temperature limit is procedures would not provide the guidance exceeded in a third area gives reasonable time necessary to achieve reactor shutdown. for the crew to perform emergency Before exiting EOP-6A ensures guidance to depressurization before the leak hampers effect reactor shutdown is not removed.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 1 Page 7 of 2 equipment or operations in an even more widespread area.

BWR Owners App. B, step RC/Q-6,RC/Q-7 App. B, step SC/T-4.2 Group Appendix

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 2 Page 1 of 6 Facility: Cooper Nuclear Station Scenario No.: 2 Op-Test No.: CNS 14-01 Examiners: ____________________________ Operators: _____________________________

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

1. Place the third Condensate Pump and Condensate Booster Pump in service.
2. Raise reactor power by raising Reactor Recirc flow.
3. Respond to a spurious initiation of RCIC.
4. Respond to failure of Reactor Recirc Pump B seals.
5. Respond to Reactor Recirc B discharge line rupture with degraded ECCS.
6. Respond to Loss of Offsite power with failure of DG2 output breaker to automatically close.

Initial Conditions: Power ascension is in progress at 75% power following maintenance on Condensate Booster Pump C.

Inoperable Equipment: Core Spray Pump A Turnover:

The plant is at 75% power. Condensate Pump C and Condensate Booster Pump C are in standby following maintenance to repair an oil leak on Condensate Booster Pump C. Five Condensate F/Ds are in service. Planned activities for this shift are:

  • The BOP operator will place Condensate Pump C and Condensate Booster Pump C in service IAW 2.2.6, Condensate System

Core Spray Pump A is tagged out of service for motor PMs. The plant is 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> into 7 day LCO 3.5.1 Condition A for Core Spray A inoperable. It is a division 1 work week.

Scenario Notes:

This is a new scenario.

Validation Time: 90 minutes

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 2 Page 2 of 6 Event Malf. No. Event Type Event No. Description 1 N/A N (BOP) Place third CP and CBP in service (2.2.6, Condensate System)

Raise power to 80% using Reactor Recirc flow control (2.1.10, 2 N/A RX (ATC)

Station Power Changes)

I (BOP) 3 rc05 A (CREW) Spurious RCIC initiation (2.4CSCS)

TS (CRS)

C (ATC,BOP) rr10b Reactor Recirc Pump B seal #1 and #2 failure (ARP, 2.4RR, 4 A (CREW) rr11b 2.4PC) with inability to isolate.

TS (CRS)

Recirc B discharge line leak into the drywell resulting in scram and subsequent Loss of Offsite Power and inability to maintain rr12b C (ATC) RPV level, with trip of RHR A and C pumps upon initiation (EOP-1A, EOP-3A) 5 rh01a M(CREW)

When RPV level lowers to -158 CFZ (TAF) and cannot be rh01c maintained above -183 CFZ (MSCWL), crew Emergency Depressurizes by opening 6 SRVs before drywell radiation reaches 150 R/hr or entering PC Flooding.

Failure of Division 2 DG output breaker to automatically close (5.3EMPWR).

C (BOP) 6 dg03b When ECCS systems fail to automatically start due to loss of AC power, crew manually closes DG-2 output breaker to energize LP ECCS systems prior to RPV water level falling below -158 CFZ (TAF).

hp01 HPCI Oil Pump failure to auto start and subsequent trip after 7 C (ATC) hp11 manual 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)

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 2 Page 3 of 6 Quantitative Attributes Table ES-304-1 Attribute Actual Description Target Malfunctions after 1. DG-2 output breaker failure to auto close EOP entry 1-2 2 2. HPCI oil pump failure

1. Spurious RCIC initiation Abnormal Events 2-4 2 2. Recirc Pump B seal failure Major Transients 1-2 1 1. Recirc LOCA with Loss of Offsite Power EOP entries requiring 1. EOP-1A substantive action 1-2 2 2. EOP-3A EOP contingencies requiring substantive 0-2 1 1. EOP-2A action
1. Manually close DG-2 output breaker to power Div 2 EOP based Critical LP ECCS systems Tasks 2-3 2 2. Emergency depressurize when RPV level drops below -158 CFZ (TAF)

Normal Events N/A 1 1. Place third CP and CBP in service Reactivity Manipulations N/A 1 1. Raise power via Recirc flow

1. Spurious RCIC initiation
2. Recirc Pump B seal failure
3. Recirc pump B suction valve failure Instrument/ 4. Recirc A discharge line rupture Component Failures N/A 8 5. Loss of Offsite Power
6. DG-2 output breaker failure to auto close
7. RHR A and C pumps trip
8. HPCI oil pump failure
1. Spurious RCIC initiation
2. Recirc Pump B seal #1 failure
3. Recirc Pump B seal #2 failure
4. Recirc pump B suction valve failure
5. Recirc B discharge line rupture Total Malfunctions N/A 10 6. Loss of Offsite Power
7. DG-2 output breaker failure to auto close
8. RHR A pump trip
9. RHR C pump trip
10. HPCI oil pump failure Top 10 systems and operator actions important to risk that are tested:

Emergency AC power High Pressure Coolant Injection Residual Heat Removal Reactor Core Isolation Cooling Automatic Depressurization System(ADS)

Manually Initiate ADS Take Manual Action to Align ECCS

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 2 Page 4 of 6 SCENARIO

SUMMARY

The plant is operating at 75% power near the end of the current fuel cycle when the crew takes the shift. Core Spray Pump A is tagged out of service for motor PMs. Condensate Pump C and Condensate Booster Pump C are ready to be returned to service following maintenance on Condensate Booster Pump C.

After the crew takes the watch, the BOP operator will start Condensate Pump V and Condensate Booster Pump C IAW the SOP.

When Condensate Pump C and Condensate Booster Pump C are in operation, the ATC will raise power to 80% by raising the speed of the Reactor Recirc Pumps.

While power is being raised, a spurious RCIC initiation will occur due to a relay failure, and the crew will trip RCIC IAW 2.4CSCS, Inadvertent CSCS Initiation, and the ARP. This will render RCIC inoperable and require entry into TS 3.5.3 Condition A.

When the Tech Spec assessment is complete, the #1 seal on the RR Pump B fails. When actions for the #1 seal failure are complete, the #2 seal will fail on the same pump. The crew will enter 2.4RR, Reactor Recirculation Abnormal and secure Recirc Pump B. This will require implementation of controls for single recirc loop operation IAW TS 3.4.1. 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.

Drywell pressure and temperature rise requiring the crew to enter 2.4PC, Primary Containment Control, and vent the drywell.

After Recirc Pump B has been secured and drywell venting is in progress, a leak will develop on the Recirc B loop discharge line. Drywell pressure and temperature rise, resulting in a reactor scram. When the scram occurs, a loss of offsite power will occur, disabling Condensate and Feedwater.

High drywell pressure and low RPV water level are the entry conditions for EOPs 1A and 3A. RPV water level will lower due to shrink from the scram and because CRD and SLC will not be able to keep up with the leak, nor will RCIC if the crew manually restarts it. HPCI will fail to automatically initiate, and shortly after the crew starts it manually, its auxiliary lube oil pump will fail, resulting in loss of HPCI. RPV water level will continue to fall to TAF, -158.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 2 Page 5 of 6 Due to the unavailability of Core Spray Pump A and trip of RHR Pumps A and C, Division 1 low pressure ECCS will be unavailable. Division 2 DG will start on the LOOP, but its output breaker will fail to automatically close.

The crew will have to recognize this failure and manually close the output breaker from the control room to make Division 2 low pressure ECCS systems available before level goes below TAF.

Drywell pressure will be elevated, requiring operation of Torus/Drywell Spray using systems not required to maintain adequate core cooling. The crew may elect to reserve Division 2 RHR for LPCI mode in anticipation of going below TAF.

When level reaches TAF, as indicated on SPDS using Corrected Fuel Zone (CFZ) at -158, and with Division 2 core spray and LPCI lined up for injection, the crew will enter EOP-2B and conduct emergency depressurization due to RPV water level below TAF and cannot be restored and maintained above -

183 CFZ. The crew will restore level to +3 to +54 inches IAW EOP-1A using low pressure ECCS.

The exercise ends when the reactor has been depressurized, RPV water level is being restored +3 to +54 inches. The emergency classification will be made to satisfy and administrative JPM for SRO candidates after the scenario has ended.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 2 Page 6 of 6 Critical Tasks When RPV level lowers to -158 CFZ (TAF) When ECCS systems fail to automatically and cannot be maintained above -183 start due to loss of AC power, crew CFZ (MSCWL), crew Emergency manually closes DG-2 output breaker to Depressurizes by opening 6 SRVs before energize LP ECCS systems prior to RPV drywell radiation reaches 150 R/hr or water level falling below -158 CFZ (TAF) entering PC Flooding.

EVENT 5 6 Safety The MSCRWL is the lowest RPV water level Failure to recognize the auto start not significance at which the covered portion of the reactor occurring and energizing of the safety bus, core will generate sufficient steam to preclude and failure to take manual action per any clad temperature in the uncovered Procedure 5.3EMPWR will result in loss of portion of the core from exceeding 1500°F. safety-related equipment necessary to provide When water level decreases below MSCRWL adequate core cooling, otherwise resulting in with injection, clad temperatures may exceed core damage and a large offsite release.

1500°F.

Cueing Corrected Fuel Zone indication (SPDS) falls Indication and/or annunciation that all ac to -158 and lowering trend continues emergency buses are de-energized

  • Bus energized lamps extinguished
  • Circuit breaker position
  • Bus voltage
  • EDG status Control room lighting dimmed Performance Manipulation of SRV controls on panel 9-3: Manipulation of controls as required to indicator SRV-71A energize Div 2 AC emergency bus from panel SRV-71B C:

SRV-71E SRV-71G Operator places DIESEL GEN 2 BKR EG2 to SRV-71H CLOSE on panel C SRV-71C SRV-71D SRV-71F Performance Crew will observe SRV light indication go Crew will observe light indication for feedback from green to red, amber pressure switch equipment powered by Division 2 AC lights illuminate, reactor pressure lowering on illuminate on panel 9-3 and bus voltage SPDS and panel 9-3 and 9-5 meters and ~4200V on panel C recorders, and SRV tailpipe temperatures rise on recorder MS-TR-166..

Justification Before 150R/hr in the drywell was chosen This is so Emergency Depressurization will not for the chosen because this is an indicator of loss of RPV be delayed by the breaker failure to auto performance level and the shielding effect of the water, close. Per EOP-1A, TAF is the point at which limit indicating core exposure, yet it is much lower Emergency Depressurization may be entered than the 2500R/hr trigger point during RPV if level cannot be maintained above MSCWL, if Flooding that indicates gross cladding failure at least one Table 5 (high capacity) system is is in progress. MSCWL was not chosen lined up for injection. With the Div 2 because at CNS, cycling of one SRV could emergency bus energized, Div 2 ECCS result in a prompt swing in level from TAF to systems will auto align for injection.

below MSCWL, not allowing the crew time to properly execute the ED before MSCWL is indicated. Before exiting to PC Flooding was chosen because the design of the scenario provides the crew with the means to restore and maintain adequate core cooling IAW EOP-1A, and exiting to SAGs is not required or authorized.

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

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 3 Page 1 of 2 Facility: Cooper Nuclear Station Scenario No.: 3 Op-Test No.: CNS 14-01 Examiners: ____________________________ Operators: _____________________________

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

1. Shift CRD Pumps from B to A.
2. Reduce power by lowering Reactor Recirc flow.
3. Respond to Outboard MSIV 86A failing closed.
4. Respond to a control rod drifting outward.
5. Respond to a spurious Turbine trip and hydraulic block ATWS.
6. Respond to failure of Generator field breaker to open automatically.
7. Respond to CRD pump A trip.

Initial Conditions: Plant is at 100% power.

Inoperable Equipment: Core Spray Pump A Turnover:

The plant is at 100% power. CRD pumps require rotation to provide run time for CRD Pump A before scheduled maintenance on CRD Pump B next week. After CRD pumps are shifted, power will be reduced to remove Condensate Booster Pump C from service to repair an oil leak. Work Control is finalizing the repair plan and clearance order and verifying repair parts are on-hand. Also. Planned activities for this shift are:

  • Immediately following turnover, the BOP operator will shift CRD pumps IAW 2.2.8, Control Rod Drive Hydraulic System.
  • When notified by Work Control, remove Condensate Booster Pump C and Condensate Pump C from service IAW 2.2.6, Condensate System.

Core Spray Pump A is tagged out of service for motor PMs. The plant is 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> into 7 day LCO 3.5.1 Condition A for Core Spray A inoperable. It is a division 1 work week.

Scenario Notes:

This is a new scenario.

Validation Time: XX minutes

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 3 Page 2 of 2 Event Malf. No. Event Type Event No. Description 1 N/A N (BOP) Shift CRD pumps from B to A (2.2.8, CRD Hydraulic System)

Reduce power to 75% using Reactor Recirc flow control (2.1.10, 2 N/A RX (ATC) Station Power Changes)(actual power change will be ~2%

before next event)

C (ATC,BOP)

(or) Outboard MSIV 86A fails closed (2.4MSIV, Inadvertent MSIV 3 A (CREW) zdipcissws4a(1) Closure)

TS (CRS)

C (ATC) 4 Rd101827 A (CREW) Control Rod 18-35 drift outward (2.4CRD, CRD Trouble)

TS (CRS)

Spurious Turbine trip, hydraulic block ATWS >3% power with SLC A failure to initiate When control rods fail to scram, crew injects SLC and/or inserts control rods before exiting EOP-6A. (All control rods do not have to be fully inserted to satisfy this critical tc01 C (ATC) task; this only requires that the crew is making progress 5 rd02a M(CREW) to achieving all rods in by fully inserting at least 5 control rods using RMCS.)

rd02b Inhibit ADS before ADS valves open on an initiation signal.

During failure to scram conditions with power >3%, lower RPV water level to below -60 CFZ and control between -

60 to -183 CFZ prior to exiting EOP-7A.

Failure of Generator field breaker to open automatically (2.1.5 6 eg01 C (BOP)

Att 5) 7 rd08a C (ATC) CRD Pump A trip (ARP 5-2/A-6)

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

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

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 3 Page 3 of 2 Quantitative Attributes Table ES-304-1 Attribute Actual Description Target

1. Failure of Generator field breaker to open Malfunctions after EOP entry 1-2 2 automatically
2. CRD Pump A trip
1. Outboard MSIV 86A fails closed Abnormal Events 2-4 2 2. Control Rod 18-35 drift outward
1. Turbine trip, hydraulic block ATWS >3% power, Major Transients 1-2 1 SLC A fail to start EOP entries requiring substantive action 1-2 1 1. EOP-1A, 6A EOP contingencies requiring substantive 0-2 1 1. EOP-7A action
1. Inject SLC/insert control rods EOP based Critical 2. Inhibit ADS.

Tasks 2-3 3 3. Lower RPV water level to below -60 CFZ and control between -60 to -183 CFZ.

Normal Events N/A 1 1. Shift CRD pumps from B to A Reactivity Manipulations N/A 1 1. Reduce power via Recirc flow

1. Outboard MSIV 86A fails closed
2. Control Rod 18-35 drift out Instrument/

Component Failures N/A 5 3. Spurious Turbine trip/Hydraulic Block ATWS

4. Generator field breaker failure to open
5. CRD pump A trip
1. Outboard MSIV 86A fails closed
2. Control Rod 18-35 drift out
3. Spurious Turbine trip
4. Hydraulic Block of S SDV Total Malfunctions N/A 8 5. Hydraulic Block of N SDV
6. Generator field breaker failure to open
7. SLC A fail to start
8. CRD pump A trip Top 10 systems and operator actions important to risk that are tested:

Reactor Protection System Primary Containment Isolation System Nuclear Boiler Instrumentation

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 3 Page 4 of 2 SCENARIO

SUMMARY

The plant is operating at 100% power near the end of the current fuel cycle when the crew takes the shift. Core Spray Pump A is out of service to repair a jacket water leak.

When the crew takes the watch, the BOP operator will rotate CRD pumps, placing A in service and securing B IAW the SOP.

After CRD pumps have been shifted, the ATC operator will begin reducing power to 75% by lowering Recirc flow IAW the GOP.

During the power reduction, Outboard MSIV 86A will slowly fail closed.

Reactor pressure and flux will rise. Due to the increase in RPV pressure power will rise above the rod monitor block setpoint but will not exceed the RPS setpoint. The crew will enter 2.4MSIV and take the handswitch for the failed MSIV to CLOSE on panel 9-4 and lower power by lowering recirc flow to 40 mlbm/hr, followed by inserting control rods to below 70% power. MSIV 86A will not indicate fully closed. The CRS will enter TS 3.6.1.3 Condition A for the failed MSIV.

When control rod 18-25 is inserted during the power reduction, it will spuriously drift out after the RO inserts it. The ATC operator will reinsert the control rod using RMCS IAW 2.4CRD. The control rod will have to be fully inserted using its scram test and disarmed. The CRS will enter TS 3.1.3 Condition C for the control rod.

When the Tech Spec assessment is complete, a spurious trip of the main turbine will occur. The reactor will fail to shut down due to hydraulic blockages in both SDVs. Power will be approximately 25%. The crew will enter EOPs 6A and 7A via EOP-1A. The crew will inject SLC and/or install the necessary PTMs to bypass interlocks and drive control rods in individually via RMCS. Bypass valves will be available to control RPV pressure. Feedwater and HPCI will be available for RPV level control; however, injection from these systems will have to be stopped to lower RPV level below -60 CFZ. Level will then be maintained between -60 and

-183 CFZ.

Following the turbine trip, the generator field breaker will fail to automatically open when the generator trips on reverse power, and the BOP operator should trip it manually. SLC pump A will fail to start when attempted by the ATC. Also, CRD pump A will trip, requiring the ATC to start CRD pump B to accomplish control rod insertion.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 3 Page 5 of 2 The exercise ends when RPV water level is being controlled -183 to +54 and when SLC is injecting and/or control rods are being inserted via RMCS and/or scram.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 3 Page 6 of 2 Critical Tasks When control rods fail to scram, crew Inhibit ADS before ADS valves open on an injects SLC and/or inserts control rods initiation signal.

before exiting EOP-6A. (All control rods do not have to be fully inserted to satisfy this critical task; this only requires that the crew is making progress to achieving all rods in by fully inserting at least 5 control rods using RMCS.)

EVENT 5 5 Safety Failure to effect shutdown of the reactor when With a Reactor Scram required, reactor not significance a RPS setting has been exceeded would shut down, and conditions for ADS blowdown unnecessarily extend the level of degradation are met, INHIBIT ADS to prevent an of the safety of the plant. This could further uncontrolled RPV depressurization and cold degrade into damage to the principle fission water injection from low pressure sources to product barriers if left unmitigated. The crew prevent causing a significant power excursion.

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.

Cueing Manual scram is initiated and numerous ADS Timer initiated alarm on panel 9-3-1/A-1 control rods indicate beyond position 00 and reactor power not downscale on panel 9-5 indications.

Performance Operator manipulates keylocked switches for Manipulation of ADS A and ADS B Inhibit indicator SLC B pump to START on panel 9-5. switches on panel 9-3 vertical section.

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.

Performance SLC B pump red light illuminated, SLC Inhibit switches click into the vertical, inhibit feedback discharge pressure rising, and SLC tank level position on panel 9-3.

lowering on panel 9-5.

Receipt of ADS inhibited alarm panel 9-3-1/D-Operator selecting and inserting control rods 1.

indicated by rod position decreasing to 00 for selected rod on panel 9-5.

Justification There is no time limit for effecting complete The 105 second ADS timer allows sufficient for the chosen reactor shutdown via boron injection or time for the crew to recognize and override performance control rod insertion. For the timeframe of automatic operation of the system. As long as limit this scenario, containment limits are not ADS is inhibited before ADS valves open, closely challenged and power oscillations are reactor pressure will not be reduced to the not experienced. However, if the failure to shutoff heads of high volume, cold water scram EOP were to be exited, other systems.

procedures would not provide the guidance necessary to achieve reactor shutdown.

Before exiting EOP-6A ensures guidance to effect reactor shutdown is not removed.

BWR Owners App. B, step RC/Q-6,RC/Q-7 App. B, step RC/Q-6 Group Appendix

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 3 Page 7 of 2 Critical Tasks During failure to scram conditions with power >3%, lower RPV water level to below -60 CFZ and control between -60 to -183 CFZ prior to exiting EOP-7A.

EVENT 5 Safety To prevent or mitigate the consequences of significance any large irregular neutron flux oscillations induced by neutronic/thermal-hydraulic instabilities, RPV water level is lowered sufficiently below the elevation of the feedwater sparger nozzles. This places the feedwater spargers in the steam space providing effective heating of the relatively cold feedwater and eliminating the potential for high core inlet subcooling. For conditions that are susceptible to oscillations, the initiation and growth of oscillations is principally dependent upon the subcooling at the core inlet; the greater the subcooling, the more likely oscillations will commence and increase in magnitude.

24" below the lowest nozzle in the feedwater sparger has been selected as the upper bound of the RPV water level control band.

This water level is sufficiently low that steam heating of the injected water will be at least 65% to 75% effective (i.e., the temperature of the injected water will be increased to 65% to 75% of its equilibrium value in the steam environment). This water level is sufficiently high that the capability to bypass the low RPV water level MSIV isolation should be able to control RPV water level with feedwater pumps to preclude the isolation.

Maintaining RPV level above -183" (MSCRWL) assures adequate core cooling via steam cooling with injection. This 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. Affiliated with this task is the requirement to only use outside the shroud injection systems due to the potential for a large power excursion that may otherwise result.

Cueing Manual scram is initiated and numerous control rods indicate beyond position 00 and reactor power >3% on panel 9-5 indications and SPDS and RPV level is >-60CFZ on SPDS.

Performance Operator manipulates Feedwater HMIs on indicator panel 9-5 or panel A as necessary to stop FW injection until RPV level goes below -60CFZ.

Operator manipulates HPCI controls on panel 9-3 to stop HPCI injection until RPV level is below -60CFZ.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 3 Page 8 of 2 Performance Feedwater flow indication on panel 9-5 feedback indicate zero.

HPCI flow indication on panel 9-3 indicates zero and/or HPCI injection MOV indicates closed.

Justification Applicability for this CT is during EOP-7A for the chosen conditions where it is necessary to lower level performance to control power with Table 17 condition NOT limit met (i.e. no high energy input into primary containment). There is no time limit for this lowering level, but it establishes margin to conditions where fuel damaging power oscillations may theoretically occur. Before exiting EOP-7A was chosen because other procedures would not provide the guidance necessary to establish margin for power oscillation mitigation. Before exiting EOP-7A ensures guidance to effect this control is not removed.

NOTE This critical task must be evaluated carefully based on the level changes. If power is reduced significantly below 3%,

reactor water level may continue to rise above -60" with only CRD and SLC while driving rods this would not result in an UNSAT on this critical task unless the Operator failed to remove RCIC to help restore and maintain level in the band.

BWR Owners App. B, Contingency #5 Group Appendix

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 4(Alt) Page 1 of 6 Facility: Cooper Nuclear Station Scenario No.: 4(Alt) Op-Test No.: CNS 14-01 Examiners: ____________________________ Operators: _____________________________

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

1. Respond to APRM C signal failure high.
2. Respond to Main Bypass Valve B failing open.
3. Respond to DEH Pump B filter clogging.
4. Respond to Feedwater Line A break inside containment (LOCA).
5. Respond to HPCI failure.
6. Respond to failure of ECCS pump to automatically initiate.

Initial Conditions: The plant is operating at 100%.

Inoperable Equipment: Core Spray Pump A Turnover:

The plant is at 100% power. Planned activities for this shift are:

  • Maintain rated power..

Core Spray Pump A is tagged out of service for motor PMs. The plant is 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> into 7 day LCO 3.5.1 Condition A for Core Spray A inoperable. It is a division 1 work week.

Scenario Notes:

This is a new scenario.

Validation Time: XX minutes

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 4(Alt) Page 2 of 6 Event Malf. No. Event Type Event No. Description I (ATC) APRM C signal failure high (ARP panel 9-5-1/A-4,A-7,B-8; 9 1 nm09c TS (CRS) 2/A-1,B-1; 2.1.5)

C (ATC)

Bypass valve B fails open (ARP panel B-1/F-1, 2.4EXT-STM, 2 tc07b A (CREW) 2.4RXPWR)

TS (CRS)

C (BOP) 3 caep A (CREW) DEH Pump B filter clogging (ARP panel B-1/E-7) tc01 fw18a DEH pressure loss-Turbine trip / FW Line A break in DW / RPS rr20a C (ATC) failure to de-energize / ARI works 4 rp01a M(CREW) When RPV level lowers to -158 CFZ (TAF) and cannot be maintained above -183 CFZ (MSCWL), crew Emergency rp01b Depressurizes by opening 6 SRVs before drywell radiation reaches 150 R/hr or entering PC Flooding.

rp01c rp01d 5 (rf)hp06a4 C (BOP) HPCI injection MO-19 fails closed cs06a Failure of all Low Pressure ECCS to start cs06b When ECCS systems fail to automatically start due to rh08a C (BOP) loss of AC power, crew manually starts and aligns at least 6

rh08b two LP ECCS pumps prior to RPV water level falling below -158 CFZ (TAF).

rh08c rh08d (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal (TS) Tech Spec

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

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 4(Alt) Page 3 of 6 Quantitative Attributes Table ES-304-1 Attribute Actual Description Target Malfunctions after 1. HPCI injection valve failure EOP entry 1-2 2 2. LP ECCS failure to auto start

1. Bypass valve B fails open Abnormal Events 2-4 2 2. DEH Pump B filter clogging Major Transients 1-2 1 1. FW line A break in DW LOCA EOP entries requiring 1. EOP-1A substantive action 1-2 2 2. EOP-3A EOP contingencies requiring substantive 0-2 1 1. EOP-2A action
1. Manually start and align at least two LP ECCS EOP based Critical pumps Tasks 2-3 2 2. Emergency depressurize when RPV level drops below -158 CFZ (TAF)

Normal Events N/A 0 1. none Reactivity Manipulations N/A 0 1. none

1. APRM C signal failure high
2. Bypass valve B fails open
3. DEH Pump B filter clogging Instrument/

Component Failures N/A 7 4. Turbine trip/FW line A break in DW

5. RPS failure to de-energize
6. HPCI injection valve MO-19 failure
7. Failure of LP ECCS to auto start
1. APRM C signal failure high
2. Bypass valve B fails open
3. DEH Pump B filter clogging
4. DEH pressure loss/Turbine trip
5. FW line A break in DW Total Malfunctions N/A 9 6. Recirc LOCA to simulate FW check valve leakage
7. RPS failure to de-energize (4 total malfunctions)
8. HPCI injection valve MO-19 failure
9. Failure of LP ECCS to auto start (6 total malfunctions)

Top 10 systems and operator actions important to risk that are tested:

High Pressure Coolant Injection Residual Heat Removal Reactor Core Isolation Cooling Automatic Depressurization System (ADS)

Manually Initiate ADS Take Manual Action to Align ECCS

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 4(Alt) Page 4 of 6 SCENARIO

SUMMARY

The plant is operating at 100% power near the end of the current fuel cycle when the crew takes the shift. Core Spray Pump A is out of service to repair a jacket water leak.

After the crew takes the watch, APRM C will fail high, causing a control rod block and half scram. The ATC will respond IAW the ARPs, bypass APRM C, and reset the half scram IAW 2.1.5. The CRS will refer to TS 3.3.1.1 and T3.1.1 for initiating a Potential LCO for APRM C.

When the Tech Spec assessment is complete, Bypass valve B will fail open.

The BOP operator will respond IAW the ARP. Since the amount of extraction steam will be reduced, FW temperature will lower causing reactor power to rise (~1%). The crew will enter 2.4RXPWR and 2.4EXT-STM and decrease power by lowering Recirc flow IAW 2.1.10 to stabilize power below 100%. The CRS will enter TS 3.7.7 Condition A for the failed bypass valve.

When the Tech Spec assessment is complete, the discharge filter for DEH pump B will plug. DEH fluid pressure will lower. The BOP operator will respond IAW the filter dp high ARP, and the crew will rotate to DEH pump A.

After EHC pump A has been started and DEH pump B is secured, following a delay DEH pressure will fall, resulting in a Turbine trip. RPS will fail to de-energize, but ARI will function to insert all control rods. Upon the transient from the Turbine trip, Feedwater line A will rupture in the Drywell, and the A FW check valve disc will fail to seat, allowing RPV inventory loss back into the Drywell. Drywell pressure and temperature rise.

High drywell pressure and low RPV water level are the entry conditions for EOPs 1A and 3A. RPV water level will lower. CRD and SLC will not be able to keep up with the leak. RCIC will not help since injects into the FW line A, which is broken, so RCIC will not stabilize RPV level. HPCI injection valve MO-19 will fail to automatically open, and its breaker will trip when the BOP operator attempts to open it manually. RPV water level will continue to fall to TAF, -158.

Neither Division 1 nor Division 2 Low Pressure ECCS pumps will start on the LOCA signal. The crew will have to recognize this failure and manually start and align low pressure ECCS systems available before level goes below TAF.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 4(Alt) Page 5 of 6 Drywell pressure will be elevated, requiring operation of Torus/Drywell Spray using systems not required to maintain adequate core cooling. The crew may elect to reserve all RHRs for LPCI mode in anticipation of going below TAF.

When level reaches TAF, as indicated on SPDS using Corrected Fuel Zone (CFZ) at -158, and with Core Spray and/or LPCI lined up for injection, the crew will enter EOP-2B and conduct emergency depressurization due to RPV water level below TAF and cannot be restored and maintained above -183 CFZ. The crew will restore level to +3 to +54 inches IAW EOP-1A using low pressure ECCS.

The exercise ends when the reactor has been depressurized, RPV water level has been restored +3 to +54 inches. The emergency classification will be made to satisfy and administrative JPM for SRO candidates after the scenario has ended.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 4(Alt) Page 6 of 6 Critical Tasks When RPV level lowers to -158 CFZ (TAF) When ECCS systems fail to automatically and cannot be maintained above -183 start due to loss of AC power, crew CFZ (MSCWL), crew Emergency manually starts and aligns at least two LP Depressurizes by opening 6 SRVs before ECCS pumps prior to RPV water level drywell radiation reaches 150 R/hr or falling below -158 CFZ (TAF).

entering PC Flooding.

EVENT 5 6 Safety The MSCRWL is the lowest RPV water level Failure to recognize the auto start not significance at which the covered portion of the reactor occurring and failure to take manual action per core will generate sufficient steam to Conduct of Ops will result in loss of safety-preclude any clad temperature in the related equipment necessary to provide uncovered portion of the core from exceeding adequate core cooling, otherwise resulting in 1500°F. When water level decreases below core damage and a large offsite release.

MSCRWL with injection, clad temperatures may exceed 1500°F.

Cueing Corrected Fuel Zone indication (SPDS) falls Indication CS and LPCI pumps are not to -158 and lowering trend continues running:

  • On panel 9-3, Red lamps on pumps handswitches extinguished and green lamps on Performance Manipulation of SRV controls on panel 9-3: Manipulation of controls as required to start at indicator SRV-71A least 2 pumps on panel 9-3:

SRV-71B SRV-71E Operator places pump handswitch to START SRV-71G CS A, CS B, RHR A, RHR B, RHR C, RHR D SRV-71H SRV-71C Performance Crew will observe SRV light indication go Crew will observe red lamp illuminate and feedback from green to red and reactor pressure green lamp extinguish for respective CS and/or lowering on SPDS and panel 9-3 and 9-5 RHR pumps on panel 9-3 meters and recorders.

Justification Before 150R/hr in the drywell was chosen This is so Emergency Depressurization will not for the chosen because this is an indicator of loss of RPV be delayed by the LP ECCS system failure to performance level and the shielding effect of the water, auto start. Per EOP-1A, TAF is the point at limit indicating core exposure, yet t is much lower which Emergency Depressurization may be than the 2500R/hr trigger point during RPV entered if level cannot be maintained above Flooding that indicates gross cladding failure MSCWL, if at least one Table 5 (high capacity) is in progress. MSCWL was not chosen system is lined up for injection.

because at CNS, cycling of one SRV could result in a prompt swing in level from TAF to below MSCWL, not allowing the crew time to properly execute the ED before MSCWL is indicated. Before exiting to PC Flooding was chosen because the design of the scenario provides the crew with the means to restore and maintain adequate core cooling IAW EOP-1A, and exiting to SAGs is not required or authorized.

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

ES-401 BWR Examination Outline Form ES-401-1 Facility: Cooper Nuclear Station Date of Exam: November 27,2015 Tier Group RO K/A Category Points SRO-Only Points K K K K K K A A A A G* A2 G* Total 1 2 3 4 5 6 1 2 3 4 Total

1. 1 3 4 4 3 3 3 20 7 Emergency & N/A N/A 2 2 0 2 0 2 1 7 3 Abnormal Plant Evolutions Tier Totals 5 4 6 3 5 4 27 10 1 2 3 2 3 2 3 2 2 2 3 2 26 5 2.

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

3. Generic Knowledge and Abilities Categories 1 2 3 4 10 1 2 3 4 7 3 2 2 3 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 Radiation Control K/A is allowed if the K/A 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 associated 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 shall be selected. 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 importance ratings (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

ES-401 2 Form ES-401-1 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 K K A A G* K/A Topic(s) IR #

1 2 3 1 2 2.2.36 Ability to analyze the effect of maintenance 295001 Partial or Complete Loss of Forced X activities, such as degraded power sources, on the 3.1 1 Core Flow Circulation / 1 & 4 status of limiting conditions for operations. (CFR:

41.10 / 43.2 / 45.13)

AK1. Knowledge of the operational implications of 295003 Partial or Complete Loss of AC / 6 X the following concepts as they apply to PARTIAL 2.7 2 OR COMPLETE LOSS OF A.C. POWER : (CFR:

41.8 to 41.10)

AK1.01 Effect of battery discharge rate on capacity AK2. Knowledge of the interrelations between 295004 Partial or Total Loss of DC Pwr / 6 X PARTIAL OR COMPLETE LOSS OF D.C. POWER 3.1 3 and the following: (CFR: 41.7 / 45.8)

AK2.01 Battery charger AK3. Knowledge of the reasons for the following 295005 Main Turbine Generator Trip / 3 X responses as they apply to MAIN TURBINE 3.8 4 GENERATOR TRIP: (CFR: 41.5 / 45.6)

AK3.07 Bypass valve operation AA1. Ability to operate and/or monitor the following 295006 SCRAM / 1 X as they apply to SCRAM : (CFR: 41.7 / 45.6) 4.2 5 AA1.01 RPS AA2. Ability to determine and/or interpret the 295016 Control Room Abandonment / 7 X following as they apply to CONTROL ROOM 3.3 6 ABANDONMENT : (CFR: 41.10 / 43.5 / 45.13)

AA2.06 Cooldown rate 2.1.23 Ability to perform specific system and 295018 Partial or Total Loss of CCW / 8 X integrated plant procedures during all modes of 4.3 7 plant operation. (CFR: 41.10 / 43.5 / 45.2 / 45.6)

AK2. Knowledge of the interrelations between 295019 Partial or Total Loss of Inst. Air / 8 X PARTIAL OR COMPLETE LOSS OF INSTRUMENT 2.8 8 AIR and the following: (CFR: 41.7 / 45.8)

AK2.04 Reactor water cleanup AK3. Knowledge of the reasons for the following 295021 Loss of Shutdown Cooling / 4 X responses as they apply to LOSS OF SHUTDOWN 3.3 9 COOLING : (CFR: 41.5 / 45.6)

AK3.04 Maximizing reactor water cleanup flow AA1. Ability to operate and/or monitor the following 295023 Refueling Acc / 8 X as they apply to REFUELING ACCIDENTS : (CFR: 2.9 10 41.7 / 45.6)

AA1.02 Fuel pool cooling and cleanup system EK1. Knowledge of the operational implications of 295024 High Drywell Pressure / 5 X the following concepts as they apply to HIGH 4.1 11 DRYWELL PRESSURE : (CFR: 41.8 to 41.10)

EK1.01 Drywell integrity: Plant-Specific EK2. Knowledge of the interrelations between 295025 High Reactor Pressure / 3 X HIGH REACTOR PRESSURE and the following: 3.9 12 (CFR: 41.7 / 45.8)

EK2.09 Reactor power EK3. Knowledge of the reasons for the following 295026 Suppression Pool High Water X responses as they apply to SUPPRESSION POOL 3.8 13 Temp. / 5 HIGH WATER TEMPERATURE: (CFR: 41.5 /

45.6)

EK3.01 Emergency/normal depressurization 295027 High Containment Temperature / 5 NA - MARK III

EA2. Ability to determine and/or interpret the 295028 High Drywell Temperature / 5 X following as they apply to HIGH DRYWELL 3.7 14 TEMPERATURE : (CFR: 41.10 / 43.5 / 45.13)

EA2.03 Reactor water level 2.4.20 Knowledge of the operational implications of 295030 Low Suppression Pool Wtr Lvl / 5 X EOP warnings, cautions, and notes. (CFR: 41.10 / 3.8 15 43.5 / 45.13)

EK1. Knowledge of the operational implications of 295031 Reactor Low Water Level / 2 X the following concepts as they apply to REACTOR 3.8 16 LOW WATER LEVEL : (CFR: 41.8 to 41.10)

EK1.02 Natural circulation: Plant-Specific EK2. Knowledge of the interrelations between 295037 SCRAM Condition Present and X SCRAM CONDITION PRESENT AND REACTOR 4.1 17 Reactor Power Above APRM Downscale or POWER ABOVE APRM DOWNSCALE OR Unknown / 1 UNKNOWN and the following: (CFR: 41.7 / 45.8)

EK2.03 ARI/RPT/ATWS: Plant-Specific EK3. Knowledge of the reasons for the following 295038 High Off-site Release Rate / 9 X responses as they apply to HIGH OFF-SITE 3.7 18 RELEASE RATE: (CFR: 41.5 / 45.6)

EK3.03 Control room ventilation isolation:

Plant-Specific.

AA1 Ability to operate and / or monitor the following 600000 Plant Fire On Site / 8 X as they apply to PLANT FIRE ON SITE: 2.6 19 AA1.08 Fire fighting equipment used on each class of fire AA2. Ability to determine and/or interpret the 700000 Generator Voltage and Electric Grid X following as they apply to GENERATOR VOLTAGE 3.6 20 Disturbances / 6 AND ELECTRIC GRID DISTURBANCES: (CFR:

41.5 and 43.5 / 45.5, 45.7, and 45.8)

AA2.04 VARs outside capability curve K/A Category Totals: 3 4 4 3 3 3 Group Point Total: 20

ES-401 3 Form ES-41-1 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 K K A A G* K/A Topic(s) IR #

1 2 3 1 2 295002 Loss of Main Condenser Vac / 3 AK3. Knowledge of the reasons for the following 295007 High Reactor Pressure / 3 X responses as they apply to HIGH REACTOR 4.0 21 PRESSURE : (CFR: 41.5 / 45.6)

AK3.04 Safety/relief valve operation: Plant-Specific 295008 High Reactor Water Level / 2 AA2. Ability to determine and/or interpret the following 295009 Low Reactor Water Level / 2 X as they apply to LOW REACTOR WATER LEVEL : 3.6 22 (CFR: 41.10 / 43.5 / 45.13)

AA2.02 Steam flow/feed flow mismatch 295010 High Drywell Pressure / 5 295011 High Containment Temp / 5 295012 High Drywell Temperature / 5 295013 High Suppression Pool Temp. / 5 295014 Inadvertent Reactivity Addition / 1 295015 Incomplete SCRAM / 1 295017 High Off-site Release Rate / 9 295020 Inadvertent Cont. Isolation / 5 & 7 AK1. Knowledge of the operational implications of the 295022 Loss of CRD Pumps / 1 X following concepts as they apply to LOSS OF CRD 3.3 23 PUMPS: (CFR: 41.8 to 41.10)

AK1.01 Reactor pressure vs. rod insertion capability EK1. Knowledge of the operational implications of the 295029 High Suppression Pool Wtr Lvl / 5 X following concepts as they apply to HIGH 3.4 24 SUPPRESSION POOL WATER LEVEL : (CFR: 41.8 to 41.10)

EK1.01 Containment integrity 295032 High Secondary Containment Area Temperature / 5 EK3. Knowledge of the reasons for the following 295033 High Secondary Containment Area X responses as they apply to HIGH SECONDARY 3.5 25 Radiation Levels / 9 CONTAINMENT AREA RADIATION LEVELS : (CFR:

41.5 / 45.6)

EK3.02 Reactor SCRAM 295034 Secondary Containment Ventilation High Radiation / 9 EA2. Ability to determine and/or interpret the following 295035 Secondary Containment High X as they apply to SECONDARY CONTAINMENT HIGH 2.8 26 Differential Pressure / 5 DIFFERENTIAL PRESSURE: (CFR: 41.8 to 41.10)

EA2.02 Off-site release rate: Plant-Specific 295036 Secondary Containment High Sump/Area Water Level / 5 2.1.28 Knowledge of the purpose and function of major 500000 High CTMT Hydrogen Conc. / 5 X system components and controls. (CFR: 41.7) 4.1 27 K/A Category Point Totals: 2 0 2 0 2 1 Group Point Total: 7

ES-401 4 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 1 (RO)

System # / Name K K K K K K A A A A G* K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 K3. Knowledge of the effect that a loss or 203000 RHR/LPCI: Injection X malfunction of the RHR/LPCI: INJECTION 4.3 28 Mode MODE (PLANT SPECIFIC) will have on following: (CFR: 41.7 / 45.4)

K3.01 Reactor water level K4. Knowledge of SHUTDOWN COOLING 205000 Shutdown Cooling X SYSTEM (RHR SHUTDOWN COOLING 3.8 29 MODE) design feature(s) and/or interlocks which provide for the following: (CFR: 41.7)

K4.03 Low reactor water level: Plant-Specific A1. Ability to predict and/or monitor changes in 206000 HPCI X parameters associated with operating the HIGH 3.5 30 PRESSURE COOLANT INJECTION SYSTEM controls including: (CFR: 41.5 / 45.5)

A1.03 Condensate storage tank level: BWR-2,3,4 2.4.49 Ability to perform without reference to procedures those actions that X require immediate operation of 4.6 31 system components and controls.

(CFR: 41.10 / 43.2 / 45.6) 207000 Isolation (Emergency) NA NA NA Condenser K6. Knowledge of the effect that a loss or 209001 LPCS X malfunction of the following will have on the 2.8 32 LOW PRESSURE CORE SPRAY SYSTEM :

(CFR: 41.7 / 45.7)

K6.05 ECCS room cooler(s) 209002 HPCS NA A2. Ability to (a) predict the impacts of the 211000 SLC X following on the STANDBY LIQUID CONTROL 4.1 33 SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6)

A2.08 Failure to SCRAM A3. Ability to monitor automatic operations of 212000 RPS X the REACTOR PROTECTION SYSTEM 4.2 34 including: (CFR: 41.7 / 45.7)

A3.06 Main turbine trip: Plant-Specific A4. Ability to manually operate and/or monitor 215003 IRM X in the control room: (CFR: 41.7 / 45.5 to 45.8) 2.9 35 A4.02 CRT display indications: Plant-Specific 3.3 A4.01 IRM recorder indication 2.1.20 Ability to interpret and execute procedure 215004 Source Range Monitor X steps. (CFR: 41.10 / 43.5 / 45.12) 4.6 36 A4. Ability to manually operate and/or monitor 215005 APRM / LPRM X in the control room: (CFR: 41.7 / 45.5 to 45.8) 3.6 37 A4.06 Verification of proper functioning/

operability

K1. Knowledge of the physical connections 217000 RCIC X and/or cause-effect relationships between 2.6 38 REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8)

K1.04 Main condenser K2. Knowledge of electrical power supplies to 218000 ADS X the following: (CFR: 41.7) 3.1 39 K2.01 ADS logic K3. Knowledge of the effect that a loss or 223002 PCIS/Nuclear Steam X malfunction of the PRIMARY CONTAINMENT 3.7 40 Supply Shutoff ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF will have on following:

(CFR: 41.7 / 45.4)

K3.07 Reactor pressure K6. Knowledge of the effect that a loss or malfunction of the following will have on the X PRIMARY CONTAINMENT ISOLATION 3.0 41 SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF : (CFR: 41.7 / 45.7)

K6.05 Containment instrumentation K4. Knowledge of RELIEF/SAFETY VALVES 239002 SRVs X design feature(s) and/or interlocks which provide 3.7 42 for the following: (CFR: 41.7)

K4.09 Manual opening of the SRV K5. Knowledge of the operational implications 259002 Reactor Water Level X of the following concepts as they apply to 3.1 43 Control REACTOR WATER LEVEL CONTROL SYSTEM : (CFR: 41.5 / 45.3)

K5.03 Water level measurement K6. Knowledge of the effect that a loss or 261000 SGTS X malfunction of the following will have on the 2.9 44 STANDBY GAS TREATMENT SYSTEM :

(CFR: 41.7 / 45.7)

K6.04 Process radiation monitoring K6. Knowledge of the effect that a loss or 262001 AC Electrical X malfunction of the following will have on the A.C. 3.6 45 Distribution ELECTRICAL DISTRIBUTION: (CFR: 41.7 /

45.7)

K6.02 Off-site power A1. Ability to predict and/or monitor changes in X parameters associated with operating the A.C. 3.1 46 ELECTRICAL DISTRIBUTION controls including: (CFR: 41.5 / 45.5)

A1.02 Effects of loads when energizing a bus 3.2 A1.05 Breaker lineups K3 Knowledge of the effect that a loss or 262002 UPS (AC/DC) X malfunction of the UNINTERRUPTABLE 2.6 47 POWER SUPPLY (A.C./D.C.) will have on following:(CFR: 41.7 / 45.4)

K3.07 Movement of control rods: Plant-Specific A2. Ability to (a) predict the impacts of the X following on the UNINTERRUPTABLE POWER 2.6 48 SUPPLY (A.C./D.C.) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6)

A2.01 Under voltage

K2. Knowledge of electrical power supplies to 263000 DC Electrical X the following: (CFR: 41.7) 3.1 49 Distribution K2.01 Major D.C. loads A3. Ability to monitor automatic operations of X the D.C. ELECTRICAL DISTRIBUTION 3.2 50 including: (CFR: 41.7 / 45.7)

A3.01 Meters, dials, recorders, alarms, and indicating lights A4. Ability to manually operate and/or monitor 264000 EDGs X in the control room: (CFR: 41.7 / 45.5 to 45.8) 3.7 51 A4.04 Manual start, loading, and stopping of emergency generator: Plant-Specific K1 Knowledge of the connections and / or 300000 Instrument Air X cause effect relationships between 3.1 52 INSTRUMENT AIR SYSTEM and the following:

(CFR: 41.2 to 41.9 / 45.7 to 45.8)

K1.05 Main Steam Isolation Valve air K2. Knowledge of electrical power supplies to 400000 Component Cooling X the following: (CFR: 41.7) 2.9 53 Water K2.01 CCW pumps K/A Category Point Totals: 2 3 3 2 2 3 2 2 2 3 2 Group Point Total: 26

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

System # / Name K K K K K K A A A A G* K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 201001 CRD Hydraulic A3. Ability to monitor automatic 201002 RMCS X operations of the REACTOR MANUAL 2.8 55 CONTROL SYSTEM including:

(CFR: 41.7 / 45.7)

A3.02 Rod movement sequence lights A2. Ability to (a) predict the impacts of 201003 Control Rod and Drive X the following on the CONTROL ROD 3.7 54 Mechanism AND DRIVE MECHANISM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

(CFR: 41.5 / 45.6)

A2.02 Uncoupled rod NA 201004 RSCS 201005 RCIS NA A4. Ability to manually operate and/or 201006 RWM X monitor in the control room: (CFR: 41.7 2.9 56

/ 45.5 to 45.8)

A4.02 Pushbutton indicating switches:

P-Spec(Not-BWR6) 2.4.30 Knowledge of events related to 202001 Recirculation X system operation/status that must bel 2.7 57 reported to internal organizations or external agencies, such as the State,l the NRC, or the transmission system operator. (CFR: 41.10 / 43.5 / 45.11) 202002 Recirculation Flow Control 204000 RWCU 214000 RPIS 215001 Traversing In-Core Probe 215002 RBM 216000 Nuclear Boiler Inst.

K1. Knowledge of the physical 219000 RHR/LPCI: Torus/Pool X connections and/or cause-effect 3.2 58 Cooling Mode relationships between RHR/LPCI:

TORUS/SUPPRESSION POOL COOLING MODE and the following:

(CFR: 41.2 to 41.9 / 45.7 to 45.8)

K1.06 Keep fill system 223001 Primary CTMT and Aux.

K1. Knowledge of the physical 226001 RHR/LPCI: CTMT Spray X connections and/or cause-effect 3.5 59 Mode relationships between RHR/LPCI:

CONTAINMENT SPRAY SYSTEM MODE and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8)

K1.03 LPCI/RHR pumps

230000 RHR/LPCI: Torus/Pool Spray Mode 233000 Fuel Pool Cooling/Cleanup 234000 Fuel Handling Equipment K2. Knowledge of electrical power 239001 Main and Reheat Steam X supplies to the following: (CFR: 41.7) 3.2 60 K2.01 Main steam isolation valve solenoids K3. Knowledge of the effect that a loss 239003 MSIV Leakage Control X or malfunction of the MSIV LEAKAGE 3.3 61 CONTROL SYSTEM will have on following: (CFR: 41.5 / 45.3)

K3.01 HRadiation release to the environment: BWR-4,5,6(P-Spec) 241000 K3.01 Knowledge of the effect that a loss or malfunction of the REACTOR/TURBINE PRESSURE REGULATING SYSTEM will have on following: Reactor power 241000 Reactor/Turbine Pressure X K3. Knowledge of the effect that a loss 4.1 61 Regulator or malfunction of the REACTOR/TURBINE PRESSURE REGULATING SYSTEM will have on following: (CFR: 41.7 / 45.4)

K3.01 Reactor power 245000 Main Turbine Gen. / Aux.

K4. Knowledge of REACTOR 256000 Reactor Condensate X CONDENSATE SYSTEM design 2.9 62 feature(s) and/or interlocks which provide for the following: (CFR: 41.7)

K4.11 Isolation of SJAE's on low flow: Plant-Specific K4.03 Condensate and/or booster 2.8 pump protection.

259001 Reactor Feedwater K5. Knowledge of the operational 268000 Radwaste X implications of the following concepts 2.7 63 as they apply to RADWASTE : (CFR:

41.5 / 45.3)

K5.01 Units of radiation, dose and dose rate 271000 Offgas K6 Knowledge of the effect that a loss 272000 Radiation Monitoring X or malfunction of the following will have 2.8 64 on the RADIATION MONITORING SYSTEM : (CFR: 41.7 / 45.7)

K6.03 A.C. power 286000 Fire Protection 288000 Plant Ventilation A1. Ability to predict and/or monitor 290001 Secondary CTMT X changes in parameters associated with 3.1 65 operating the SECONDARY CONTAINMENT controls including:

(CFR: 41.5 / 45.5)

A1.01 System lineups

290003 Control Room HVAC 290002 Reactor Vessel Internals 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) Form ES-401-3 Facility: Cooper Nuclear Station Date of Exam: November 27, 2015 Category K/A # Topic RO SRO-Only IR # IR #

2.1. 2.1.25 Ability to interpret reference materials, such as 3.9 66 graphs, curves, tables, etc. (CFR: 41.10 / 43.5 /

45.12) 2.1. 2.1.30 Ability to locate and operate components, including 4.4 67 local controls. (CFR: 41.7 / 45.7) 2.1. 2.1.31 Ability to locate control room switches, controls, and 4.6 68 1.

Conduct of indications, and to determine that they correctly Operations reflect the desired plant lineup. (CFR: 41.10 /

45.12) 2.1.

2.1.

2.1.

Subtotal 3 2.2. 2.2.12 Knowledge of surveillance procedures. (CFR: 41.10 3.7 69

/ 45.13) 2.2. 2.2.37 Ability to determine operability and/or availability of 3.6 70 safety related equipment. (CFR: 41.7 / 43.5 / 45.12) 2.

2.2.

Equipment Control 2.2.

2.2.

2.2.

Subtotal 2 2.3.7 Ability to comply with radiation work permit 2.3. 3.5 71 requirements during normal or abnormal conditions.

(CFR: 41.12 / 45.10) 2.3.13 Knowledge of radiological safety procedures pertaining 2.3. to licensed operator duties, such as response to 3.4 72 radiation monitor alarms, containment entry

3. requirements, fuel handling responsibilities, access to Radiation Control locked high-radiation areas, aligning filters, etc. (CFR:

41.12 / 43.4 / 45.9 / 45.10) 2.3.

2.3.

Subtotal 2 2.4. 2.4.18 Knowledge of the specific bases for EOPs. (CFR: 3.3 73 41.10 / 43.1 / 45.13) 2.4. 2.4.31 Knowledge of annunciator alarms, indications, or 4.2 74 response procedures. (CFR: 41.10 / 45.3) 2.4. 2.4.45 Ability to prioritize and interpret the significance of 4.1 75 4.

Emergency each annunciator or alarm. (CFR: 41.10 / 43.5 /

Procedures / Plan 45.3 / 45.12) 2.4.

2.4.

2.4.

Subtotal 3 Tier 3 Point Total 10

ES-401 Record of Rejected K/As Form ES-401-4 Tier / Group Randomly Selected K/A Reason for Rejection 1/1 295005 AK 3.07 Question #4 - Rejected K/A AK 3.08 Feedwater - HPCI actuation: BWR-2 due to being NA for Cooper. Replaced with randomly sampled K/A AK3.07 for same category 295005 3 All generic KAs were screened prior to random selection for Tier 3 as described in NUREG-1021, Section ES-401.D.1.b.

All All generic KAs The generic KAs in the initial NRC generated outline were selected in error. Instead of including the KAs listed in ES-401.D.1.b, they were excluded. The KAs initially selected in error were re-selected by the NRC in order to restore compliance with NUREG-1021. The replacement KAs were randomly selected.

1/1 295024, EK1.01 Question #11 - Rejected EK1.02 because it was specific to a Mark III containment.

CNS has a Mark 1 containment. Selected the remaining K/A in that category, EK1.01, Drywell Integrity, Plant-Specific.

1/1 295028, EA2.03 Question #14 - Rejected K/A 295028 EA2.02 because of oversampling; it was also selected on the SRO exam (Question #82). Replaced with randomly sampled K/A EA2.03 for same category 295028 1/1 295030, G2.4.20 Question #15 - rejected K/A G2.4.34, Knowledge of RO tasks performed outside the main control room duringl an emergency and the resultant operational effects, because could not find any explicitly RO tasks outside of the control room for low SP level. Replaced with randomly selected K/A G2.4.20 for the same category, 295030.

2/1 262002, K3.07 Question #47 - rejected K/A 4.01 for 262002, Knowledge of UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) design feature(s) and/or interlocks which provide for the following: Knowledge of UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.)

design feature(s) and/or interlocks which provide for the following: Transfer from preferred power to alternate power supplies, due to double jeopardy with question

  1. 48. K/A 4.01 was the only K/A in this category, so randomly selected K/A 3.07.

Question #55 - rejected K/A 201004 A3.04 Ability to monitor automatic operations of 2/2 201002, A3.02 the ROD SEQUENCE CONTROL SYSTEM (PLANT SPECIFIC) including: (CFR:

41.7 / 45.7) Annunciator and alarm signals: BWR-4,5 because RSCS is no longer used at CNS. Replaced with similar K/A for the RMCS, 201002, which is used at CNS. Randomly selected K/A A3.02 from subcategoryA3.

Question #35 - Rejected K/A A4.02, Ability to manually operate and/or monitor in the 2/1 215003, A4.01 control room: CRT display indications: Plant-Specific is not applicable to CNS.

Randomly selected from the same subcategory A4.01, IRM recorder indication.

Question #46 Rejected K/A A1.02 AC Electrical Distribution, Ability to predict and/or 2/1 262001 A1.05 monitor changes in parameters associated with operating the A.C. ELECTRICAL DISTRIBUTION controls including: Effects of loads when energizing a bus. Could not write a discriminatory question, randomly selected from the same subcategory A1.05 Breaker lineups.

Question #61 - Rejected K/A 239003 MSIV Leakage Control not applicable to CNS.

2/2 241000, K3.01 Randomly selected new system with same KA statement 241000 K3.01 Knowledge of the effect that a loss or malfunction of the REACTOR/TURBINE PRESSURE REGULATING SYSTEM will have on following: Reactor power Question #62 - Rejected K/A K4.11 - SJAE do not isolate on low flow at CNS.

2/2 256000, K4.03 Randomly selected from the same subcategory K4.03 Condensate and/or booster pump protection.

ES-401 BWR Examination Outline Form ES-401-1 Facility: Cooper Nuclear Station Date of Exam: November 27,2015 Tier Group RO K/A Category Points SRO-Only Points K K K K K K A A A A G* A2 G* Total 1 2 3 4 5 6 1 2 3 4 Total

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

Plant 2 12 0 2 1 3 Systems Tier Totals 38 5 3 8

3. Generic Knowledge and Abilities Categories 1 2 3 4 10 1 2 3 4 7 1 2 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 outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 Radiation Control K/A is allowed if the K/A 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 associated 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 shall be selected. 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 importance ratings (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

ES-401 2 Form ES-401-1 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 K K A A G* K/A Topic(s) IR #

1 2 3 1 2 295001 Partial or Complete Loss of Forced Core Flow Circulation / 1 & 4 295003 Partial or Complete Loss of AC / 6 295004 Partial or Total Loss of DC Pwr / 6 AA2. Ability to determine and/or interpret the 295005 Main Turbine Generator Trip / 3 X following as they apply to MAIN TURBINE 3.6 76 GENERATOR TRIP : (CFR: 41.10 / 43.5 / 45.13)

AA2.07 Reactor water level 2.2.37 Ability to determine operability and/or 295006 SCRAM / 1 X availability of safety related equipment. (CFR: 41.7 / 4.6 77 43.5 / 45.12) 295016 Control Room Abandonment / 7 AA2. Ability to determine and/or interpret the 295018 Partial or Total Loss of CCW / 8 X following as they apply to PARTIAL OR COMPLETE 3.5 78 LOSS OF COMPONENT COOLING WATER :

(CFR: 41.10 / 43.5 / 45.13)

AA2.03 Cause for partial or complete loss 2.1.7 Ability to evaluate plant performance and 295019 Partial or Total Loss of Inst. Air / 8 X make operational judgments based on operating 4.7 79 characteristics, reactor behavior, and instrument interpretation. (CFR: 41.5 / 43.5 / 45.12 / 45.13)

AA2. Ability to determine and/or interpret the 295021 Loss of Shutdown Cooling / 4 X following as they apply to LOSS OF SHUTDOWN 3.6 80 COOLING : (CFR: 41.10 / 43.5 / 45.13)

AA2.04 Reactor water temperature 2.4.41 Knowledge of the emergency action level 295023 Refueling Acc / 8 X thresholds and classifications. (CFR: 41.10 / 43.5 / 4.6 81 45.11) 295024 High Drywell Pressure / 5 295025 High Reactor Pressure / 3 295026 Suppression Pool High Water Temp. / 5 295027 High Containment Temperature / 5 EA2. Ability to determine and/or interpret the 295028 High Drywell Temperature / 5 X following as they apply to HIGH DRYWELL 3.9 82 TEMPERATURE : (CFR: 41.10 / 43.5 / 45.13)

EA2.02 Reactor pressure 295030 Low Suppression Pool Wtr Lvl / 5 295031 Reactor Low Water Level / 2 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown / 1 295038 High Off-site Release Rate / 9 600000 Plant Fire On Site / 8 700000 Generator Voltage and Electric Grid Disturbances / 6 K/A Category Totals: 4 3 Group Point Total: 7

ES-401 3 Form ES-401-1 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 K K A A G* K/A Topic(s) IR #

1 2 3 1 2 295002 Loss of Main Condenser Vac / 3 295007 High Reactor Pressure / 3 295008 High Reactor Water Level / 2 295009 Low Reactor Water Level / 2 295010 High Drywell Pressure / 5 295011 High Containment Temp / 5 295012 High Drywell Temperature / 5 295013 High Suppression Pool Temp. / 5 295014 Inadvertent Reactivity Addition / 1 2.2.25 Knowledge of the bases in Technical 295015 Incomplete SCRAM / 1 X Specifications for limiting conditions for 4.2 83 operations and safety limits. (CFR: 41.5 /

41.7 / 43.2)

AA2. Ability to determine and/or interpret the following 295017 High Off-site Release Rate / 9 X as they apply to HIGH OFF-SITE RELEASE RATE : 3.8 84 (CFR: 41.10 / 43.5 / 45.13)

AA2.05 Meteorological data 295020 Inadvertent Cont. Isolation / 5 & 7 295022 Loss of CRD Pumps / 1 2.4.6 Knowledge of EOP mitigation strategies.

295029 High Suppression Pool Wtr Lvl / 5 X (CFR: 41.10 / 43.5 / 45.13) 4.7 85 295032 High Secondary Containment Area Temperature / 5 295033 High Secondary Containment Area Radiation Levels / 9 295034 Secondary Containment Ventilation High Radiation / 9 295035 Secondary Containment High Differential Pressure / 5 295036 Secondary Containment High Sump/Area Water Level / 5 500000 High CTMT Hydrogen Conc. / 5 K/A Category Point Totals: 1 2 Group Point Total: 3

ES-401 4 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 1 (SRO)

System # / Name K K K K K K A A A A G* K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 203000 RHR/LPCI: Injection Mode 205000 Shutdown Cooling 206000 HPCI 207000 Isolation (Emergency)

Condenser 209001 LPCS 209002 HPCS NA A2. Ability to (a) predict the impacts of the 211000 SLC X following on the STANDBY LIQUID 3.4 86 CONTROL SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

(CFR: 41.5 / 45.6)

A2.03 A.C. power failures 212000 RPS 2.2.42 Ability to recognize system 215003 IRM X parameters that are entry-level 4.6 87 conditions for Technical Specifications. (CFR: 41.7 /

41.10 / 43.2 / 43.3 / 45.3) 215004 Source Range Monitor A2. Ability to (a) predict the impacts of the 215005 APRM / LPRM X following on the AVERAGE POWER RANGE 3.6 88 MONITOR/LOCAL POWER RANGE MONITOR SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

(CFR: 41.5 / 45.6)

A2.05 Loss of recirculation flow signal 217000 RCIC 218000 ADS 223002 PCIS/Nuclear Steam Supply Shutoff 239002 SRVs 2.2.12 Knowledge of surveillance 259002 Reactor Water Level X procedures. (CFR: 41.10 / 45.13) 4.1 89 Control 261000 SGTS 262001 AC Electrical Distribution 262002 UPS (AC/DC)

A2. Ability to (a) predict the impacts of the 263000 DC Electrical X following on the D.C. ELECTRICAL 3.2 90 Distribution DISTRIBUTION ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

(CFR: 41.5 / 45.6)

A2.01 Grounds 264000 EDGs 300000 Instrument Air 400000 Component Cooling Water K/A Category Point Totals: 3 2 Group Point Total: 5

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

System # / Name K K K K K K A A A A G* K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 201001 CRD Hydraulic 201002 RMCS 201003 Control Rod and Drive Mechanism 201004 RSCS 201005 RCIS 201006 RWM 202001 Recirculation 202002 Recirculation Flow Control 204000 RWCU 214000 RPIS 215001 Traversing In-Core Probe 215002 RBM A2. Ability to (a) predict the impacts of 216000 Nuclear Boiler Inst. X the following on the NUCLEAR BOILER 2.9 91 INSTRUMENTATION ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 /

45.6)

A2.12 Instrument isolation valve closures 219000 RHR/LPCI: Torus/Pool Cooling Mode 223001 Primary CTMT and Aux.

226001 RHR/LPCI: CTMT Spray Mode 230000 RHR/LPCI: Torus/Pool Spray Mode 233000 Fuel Pool Cooling/Cleanup 234000 Fuel Handling Equipment 239001 Main and Reheat Steam 239003 MSIV Leakage Control 241000 Reactor/Turbine Pressure Regulator 245000 Main Turbine Gen. / Aux.

256000 Reactor Condensate 259001 Reactor Feedwater 2.4.47 Ability to diagnose and 268000 Radwaste X recognize trends in an accurate and 4.2 92 timely manner utilizing the appropriate control room reference material. (CFR:

41.10 / 43.5 / 45.12)

271000 Offgas A2. Ability to (d) predict the impacts of 272000 Radiation Monitoring X the following on the RADIATION 4.1 93 MONITORING SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

(CFR: 41.5 / 45.6)

A2.10 Loss of coolant accident 286000 Fire Protection 288000 Plant Ventilation 290001 Secondary CTMT 290003 Control Room HVAC 290002 Reactor Vessel Internals K/A Category Point Totals: 2 1 Group Point Total: 3

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Cooper Nuclear Station Date of Exam: November 27, 2015 Category K/A # Topic RO SRO-Only IR # IR #

2.1.23 Ability to perform specific system and integrated plant 2.1.23 4.4 94 procedures during all modes of plant operation.

(CFR: 41.10 / 43.5 / 45.2 / 45.6) 2.1

1. 2.1.

Conduct of Operations 2.1.

2.1.

2.1.

Subtotal 1 2.2.12 Knowledge of surveillance procedures. (CFR: 41.10 2.2.12 4.1 95

/ 45.13) 2.2.25 Knowledge of the bases in Technical Specifications 2.2.25 for limiting conditions for operations and safety 4.2 96 limits. (CFR: 41.5 / 41.7 / 43.2)

2. 2.2.

Equipment Control 2.2.

2.2.

2.2.

Subtotal 2 2.3.4 Knowledge of radiation exposure limits under normal 2.3. or emergency conditions. (CFR: 41.12 / 43.4 / 45.10) 3.7 97 2.3.12 Knowledge of radiological safety principles pertaining 2.3. 3.7 98 to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to

3. locked high-radiation areas, aligning filters, etc. (CFR:

Radiation Control 41.12 / 45.9 / 45.10) 2.3.

2.3.

Subtotal 2 2.4.1 Knowledge of EOP entry conditions and immediate 2.4.1 4.8 99 action steps. (CFR: 41.10 / 43.5 / 45.13) 2.4.30 Knowledge of events related to system 2.4.30 4.1 100 operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system

4. operator. (CFR: 41.10 / 43.5 / 45.11)

Emergency 2.4 Procedures / Plan 2.4.

2.4.

2.4.

Subtotal 2 Tier 3 Point Total 7 7

ES-401 Record of Rejected K/As Form ES-401-4 Tier / Group Randomly Selected K/A Reason for Rejection 3 All generic KAs were screened prior to random selection for Tier 3 as described in NUREG-1021, Section ES-401.D.1.b.

All All generic KAs The generic KAs in the initial NRC generated outline were selected in error. Instead of including the KAs listed in ES-401.D.1.b, they were excluded. The KAs initially selected in error were re-selected by the NRC in order to restore compliance with NUREG-1021. The replacement KAs were randomly selected.

Question # 93 - Rejected K/A 272000 Radiation Monitoring, . Ability to (d) predict the 2/2 272000 A2.10 impacts of the following on the RADIATION MONITORING SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: A2.13 . Low reactor water level during refueling operations.

Rejected this K/A because it is too similar to K/A 295023 Refueling Acc, 2.4.41 Knowledge of the emergency action level thresholds and classifications, used in question #81. A low refueling pool level and elevated radiation levels were used in question #81. The original K/A would necessarily introduce testing of the same knowledge as does question #81, since pool level low enough to affect Rad Monitors would probably result in emergency classification.

Kept 272000 and randomly selected K/A A2.10 Loss of coolant accident by blindly drawing from a jumbled bag of 15 labeled pennies.

1/2 295029, 2.4.6 Question #85 - Rejected K/A 295029 High Suppression Pool Wtr Lvl , 2.4.20 Knowledge of the operational implications of EOP warnings, cautions, and notes.

Could not develop a discriminatory question at the SRO level for this K/A, since no specific EOP caution or note exists for high SP level. Kept K/A 295029 and randomly selected K/A from 2.4 Generics: 2.4.6 Knowledge of EOP mitigation strategies

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Cooper Nuclear Station Date of Examination: 11/30/2015 Examination Level: RO SRO Operating Test Number: CNS 14-01 Administrative Topic (see Note) Type Describe activity to be performed Code*

M Suppression Pool Temperature Calculation Conduct of Operations R (Version 2), SKL034-21-XX(131)

K/A: A295013 AA201 (3.8/4.0); 2.1.20 (4.6/4.6)

N/A Conduct of Operations M Determine Isolation Boundaries (Version 2),

Equipment Control R SKL034-50-XX(57)

K/As 2.2.41 (3.5/3.9)

P Radiation Protection Table Top, SKL034-30-63 D

Radiation Control R K/A 2.3.13 (3.4/3.8)

M Calculate Liquid Release Curie Content R

Emergency Plan (Version 2), SKL034-50-XX(75)

K/A 2.3.11 (3.8/4.3); 2.4.21 (4.0/4.6)

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

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

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

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

ES-301

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Cooper Nuclear Station Date of Examination: 11/30/2015 Examination Level: RO SRO Operating Test Number: CNS 14-01 Administrative Topic (see Note) Type Describe activity to be performed Code*

P Determine Shift Staffing Requirements for Mode D

Conduct of Operations Change, SKL034-20-114 R K/A 2.1.4 (2.9/3.9)

D Determine TS Actions for CRDM Removal, Conduct of Operations R SKL034-50-61 K/A 2.2.40 (3.4/4.7)

M Perform CRS Review of Jet Pump/Recirc Daily Equipment Control R Operability Checks (Version 2) SKL034 XX(113)

K/A 202001 K1.06 3.6/3.6; 2.2.42 (3.9/4.6)

N Authorize Emergency Exposure Radiation Control R SKL034-XX-XX K/A 2.3.4 (3.2/3.7)

N Emergency Classification SKL034-50-XX Emergency Plan S K/A 2.4.41 (2.9/4.6)

R 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)

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

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

ES-301

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Cooper Nuclear Station Date of Examination: 11/30/2015 Exam Level: RO SRO-I SRO-U Operating Test No.: NRC CNS 14-01 Control Room Systems:* 8 for RO; 7 for SRO-I; 2 or 3 for SRO-U System / JPM Title Type Code* Safety Function

a. SKL034-21-XX(45) Operate HPCI in pressure Control Mode M, A, EN, L,S 4 (Alternate Path) K/A 206000 K3.02 (3.8/3.8), A4.01 (3.8/3.7)
b. SKL034-20-XX(107), Establish Cooldown Rate with Bypass M, L, S 3 Valves K/A 241000 A4.06 (3.9/3.9)
c. SKL034-21-133, Respond To Uncoupled Control Rod (Alternate A, D, S 1 Path) K/A 201003 A2.02 (3.7/3.8)
d. SKL034-21-33 Transfer 4160 VAC Bus 1G From DG2 To 4160 D, EN, S 6 VAC Bus 1B K/A:262001 A4.04 (3.6/3.7), 264000 A4.05 (3.6/3.7)
e. Returning RHR Subsystem A to LPCI Standby from Suppression N, S, EN 2 Pool Cooling K/A 203000 A4.01 (4.3/4.1)
f. SKL034-21-61, Defeat Group 1 Low Level Isolation during an D, E, S 5 ATWS (Restoration). K/A 223002 (3.3/3.7)
g. SKL034-21-XX(44), Separation of REC Critical Loops K/A P, D, S, E 8 400000 A4.01 (3.1/3.0); 295018 AA1.03 (3.3/3.4), AK3.07 (3.1/3.2)
h. SKL034-21-XX(74) Withdraw SRMs during a Start-up (Alternate M, A, S 7 Path) K/A 215004.A1.01 (3.0/3.1), A2.03 (3.0/3.3), A4.04 (3.2/3.2)

In-Plant Systems* (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

i. SKL034-10-XX(96), Startup RPS MG Set A (Alternate Path) M, A 7 K/A 212000 A1.01 (2.8/2.9), A2.01 (3.7/3.9)
j. SKL034-10-95, Respond to NBPP Failure (Cntl Bldg Actions, P, A, D, E 6 Alternate Path) K/A 262002 K4.01(3.4/3.4)
k. SKL034-10-50 Fill the Skimmer Surge Tank D, R 9 K/A 233000.A2.02 3.1/3.3
  • All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all five SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
  • Type Codes Criteria for RO / SRO-I / SRO-U

A)lternate path 4-6 / 4-6 / 2-3 (C)ontrol room (D)irect from bank 9/8/4 (E)mergency or abnormal in-plant 1/1/1 (EN)gineered safety feature 1 / 1 / 1 (control room system)

(L)ow-Power / Shutdown 1/1/1 (N)ew or (M)odified from bank including 1(A) 2/2/1 (P)revious 2 exams 3 / 3 / 2 (randomly selected)

(R)CA 1/1/1 (S)imulator

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Cooper Nuclear Station Date of Examination: 11/30/2015 Exam Level: RO SRO-I SRO-U Operating Test No.: NRC CNS 14-01 Control Room Systems:* 8 for RO; 7 for SRO-I; 2 or 3 for SRO-U System / JPM Title Type Code* Safety Function

a. SKL034-21-XX(45) Operate HPCI in pressure Control Mode M, A, EN, L,S 4 (Alternate Path) K/A 206000 K3.02 (3.8/3.8), A4.01 (3.8/3.7)
b. SKL034-20-XX(107), Establish Cooldown Rate with Bypass M, L, S 3 Valves K/A 241000 A4.06 (3.9/3.9)
c. SKL034-21-133, Respond To Uncoupled Control Rod (Alternate A, D, S 1 Path) K/A 201003 A2.02 (3.7/3.8)
d. N/A
e. Returning RHR Subsystem A to LPCI Standby from Suppression N, S, EN 2 Pool Cooling K/A 203000 A4.01 (4.3/4.1)
f. SKL034-21-61, Defeat Group 1 Low Level Isolation during an D, E, S 5 ATWS (Restoration). K/A 223002 (3.3/3.7)
g. SKL034-21-XX(44), Separation of REC Critical Loops K/A P, D, S, E 8 400000 A4.01 (3.1/3.0); 295018 AA1.03 (3.3/3.4), AK3.07 (3.1/3.2)
h. SKL034-21-XX(74) Withdraw SRMs during a Start-up (Alternate M, A, S 7 Path) K/A 215004.A1.01 (3.0/3.1), A2.03 (3.0/3.3), A4.04 (3.2/3.2)

In-Plant Systems* (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

i. SKL034-10-XX(96), Startup RPS MG Set A (Alternate Path) M, A 7 K/A 212000 A1.01 (2.8/2.9), A2.01 (3.7/3.9)
j. SKL034-10-95, Respond to NBPP Failure (Cntl Bldg Actions, P, A, D, E 6 Alternate Path) K/A 262002 K4.01(3.4/3.4)
k. SKL034-10-50 Fill the Skimmer Surge Tank D, R 9 K/A 233000.A2.02 3.1/3.3
  • All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all five SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
  • Type Codes Criteria for RO / SRO-I / SRO-U

A)lternate path 4-6 / 4-6 / 2-3 (C)ontrol room (D)irect from bank 9/8/4 (E)mergency or abnormal in-plant 1/1/1 (EN)gineered safety feature 1 / 1 / 1 (control room system)

(L)ow-Power / Shutdown 1/1/1 (N)ew or (M)odified from bank including 1(A) 2/2/1 (P)revious 2 exams 3 / 3 / 2 (randomly selected)

(R)CA 1/1/1 (S)imulator

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Cooper Nuclear Station Date of Examination: 11/30/2015 Exam Level: RO SRO-I SRO-U Operating Test No.: NRC CNS 14-01 Control Room Systems:* 8 for RO; 7 for SRO-I; 2 or 3 for SRO-U System / JPM Title Type Code* Safety Function

a. SKL034-21-XX(45) Operate HPCI in pressure Control Mode M, A, EN, L,S 4 (Alternate Path) K/A 206000 K3.02 (3.8/3.8), A4.01 (3.8/3.7)
b. N/A
c. N/A
d. N/A
e. Returning RHR Subsystem A to LPCI Standby from Suppression N, S, EN 2 Pool Cooling K/A 203000 A4.01 (4.3/4.1)
f. SKL034-21-61, Defeat Group 1 Low Level Isolation during an D, E, S 5 ATWS (Restoration). K/A 223002 (3.3/3.7)
g. N/A
h. N/A In-Plant Systems* (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)
i. N/A
j. SKL034-10-95, Respond to NBPP Failure (Cntl Bldg Actions, P, A, D, E 6 Alternate Path) K/A 262002 K4.01(3.4/3.4)
k. SKL034-10-50 Fill the Skimmer Surge Tank D, R 9 K/A 233000.A2.02 3.1/3.3
  • All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all five SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
  • Type Codes Criteria for RO / SRO-I / SRO-U A)lternate path 4-6 / 4-6 / 2-3 (C)ontrol room (D)irect from bank 9/8/4 (E)mergency or abnormal in-plant 1/1/1 (EN)gineered safety feature 1 / 1 / 1 (control room system)

(L)ow-Power / Shutdown 1/1/1 (N)ew or (M)odified from bank including 1(A) 2/2/1 (P)revious 2 exams 3 / 3 / 2 (randomly selected)

(R)CA 1/1/1 (S)imulator

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Cooper Nuclear Station Date of Exam: 11/30/2015 Operating Test No.: NRC CNS 14-01 A E Scenarios P V 1 2 3 4 T M P E O I L N CREW CREW POSITION CREW POSITION CREW POSITION T N I T POSITION I

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

T P E R I U RO(R1) RX 1 0 0 1 1 1 0 NOR 0 1 1 2 1 1 1 SRO-I I/C 3 3 2 8 4 4 2 SRO-U MAJ 2 1 1 4 2 2 1 TS NA NA NA NA 0 2 2 RO(R2) RX 1 0 0 1 1 1 0 NOR 0 1 1 2 1 1 1 SRO-I I/C 3 3 2 8 4 4 2 SRO-U MAJ 2 1 1 4 2 2 1 TS NA NA NA NA 0 2 2 RO(R3) RX 0 1 1 1 1 0 NOR 0 0 0 1 1 1 SRO-I I/C 4 4 8 4 4 2 SRO-U MAJ 2 1 3 2 2 1 TS NA NA NA 0 2 2 RO RX 0 1 1 1 1 0 (R4)

NOR 0 0 0 1 1 1 SRO-I I/C 4 3 7 4 4 2 SRO-U MAJ 2 1 3 2 2 1 TS NA NA NA 0 2 2 Instructions:

1. Check the applicant level and enter the operating test number and Form ES-D-1 event numbers for each event type; TS are not applicable for RO applicants. ROs must serve in both the at-the-controls (ATC) and balance-of-plant (BOP) positions. Instant SROs (SRO-I) must serve in both the SRO and the ATC positions, including at least two instrument or component (I/C) malfunctions and one major transient, in the ATC position. If an SRO-I additionally serves in the BOP position, one I/C malfunction can be credited toward the two I/C malfunctions required for the ATC position.
2. Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.5.d) but must be significant per Section C.2.a of Appendix D. (*) Reactivity and normal evolutions may be replaced with additional instrument or component malfunctions on a one-for-one basis.
3. Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicants competence count toward the minimum requirements specified for the applicants license level in the right-hand columns.
4. For licensees that use the ATC operator primarily for monitoring plant parameters, the chief examiner may place SRO-I applicants in either the ATC or BOP position to best evaluate the SRO-I in manipulating plant controls.

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Cooper Nuclear Station Date of Exam: 11/30/2015 Operating Test No.: NRC CNS 14-01 A E Scenarios P V 1 2 3 4 T M P E O I L N CREW CREW POSITION CREW POSITION CREW POSITION T N I T POSITION I

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

T P E (I2) (R2) (R4) R I U RO RX 1 1 1 3 1 1 0 NOR 0 0 1 1 1 1 1 SRO-I (I1) I/C 6 3 5 14 4 4 2 MAJ 2 1 1 4 2 2 1 SRO-U TS 2 NA 2 4 0 2 2 RO RX 1 1 1 3 1 1 0 NOR 0 1 0 1 1 1 1 SRO-I(I2) I/C 6 4 4 14 4 4 2 MAJ 2 1 1 4 2 2 1 SRO-U TS 2 2 NA 4 0 2 2 RO RX 1 1 2 1 1 0 NOR 1 1 2 1 1 1 SRO-I I/C 4 5 9 4 4 2 SRO- MAJ 1 1 2 2 2 1 U(U1)

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

1. Check the applicant level and enter the operating test number and Form ES-D-1 event numbers for each event type; TS are not applicable for RO applicants. ROs must serve in both the at-the-controls (ATC) and balance-of-plant (BOP) positions. Instant SROs (SRO-I) must serve in both the SRO and the ATC positions, including at least two instrument or component (I/C) malfunctions and one major transient, in the ATC position. If an SRO-I additionally serves in the BOP position, one I/C malfunction can be credited toward the two I/C malfunctions required for the ATC position.
2. Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.5.d) but must be significant per Section C.2.a of Appendix D. (*) Reactivity and normal evolutions may be replaced with additional instrument or component malfunctions on a one-for-one basis.
3. Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicants competence count toward the minimum requirements specified for the applicants license level in the right-hand columns.
4. For licensees that use the ATC operator primarily for monitoring plant parameters, the chief examiner may place SRO-I applicants in either the ATC or BOP position to best evaluate the SRO-I in manipulating plant controls.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 1 Page 1 of 2 Facility: Cooper Nuclear Station Scenario No.: 1 Op-Test No.: CNS 14-01 Examiners: ____________________________ Operators: _____________________________

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

1. Raise reactor power during startup by withdrawing control rods.
2. Respond to erratic indication on IRM H requiring bypassing IRM H.
3. Respond to trip of TEC pump A.
4. Respond to trip of RPS MG Set B with trip of SGT A exhaust fan.
5. Respond to RCIC steam leak into Secondary Containment with inability to isolate the leak.
6. Take actions for a low power ATWS.
7. Conduct emergency depressurization due to degrading conditions in secondary containment.

Initial Conditions: Plant startup in progress at 4.5% power and rated pressure.

Inoperable Equipment: None Turnover:

The plant is at 4.5% power and rated pressure during startup from a mid-cycle outage. Planned activities for this shift are:

  • Continue power ascension IAW 2.1.1, Startup Procedure.

Primary Containment inerting is NOT in progress. There is no out of service equipment and integrated risk is GREEN. It is a division 1 work week.

Scenario Notes:

This is a new scenario.

Validation Time: 75 minutes

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 1 Page 2 of 2 Event Malf. No. Event Type Event No. Description Raise Reactor power by withdrawing control rods. (Startup 1 N/A R (ATC) control rod sequence; GOP 2.1.1, Startup Procedure )

I (ATC) Respond to IRM H erratic indication requiring bypass (ARP 9 2 nm05h TS (CRS) 1/E-8)

TEC Pump A trip requiring manual start of standby TEC pump 3 sw07a C (BOP)

(ARP M-2/A-5)

(rf) rp02 C (BOP,ATC)

RPS MG Set B trip with failure of SGT A Exhaust Fan (ARP C-4 (or) A (CREW) 1/F-2, ARP C-1/G-2, 2.1.5, 2.1.22) zdisgtswefre TS (CRS) rc06 C (BOP) rc07 RCIC steam leak in Secondary Ctmt with failure to automatically 5 A(CREW) isolate, failure of RCIC isolation valves MO-15 and MO-16 (rf) rc06a during manual isolation. (EOP-5A)

TS (CRS)

(rf) rc22a Manual scram due to Secondary Ctmt temperature, Hydraulic block ATWS < 3% power (EOP-5A, 1A, 6A, 6B, 7A)

C (ATC) When control rods fail to scram, crew injects SLC and/or 6 rd02a inserts control rods before exiting EOP-6A. (All control M (CREW) rods do not have to be fully inserted to satisfy this critical task; this only requires that the crew is making progress to achieving all rods in by fully inserting at least 5 control rods using RMCS.)

RCIC steam leak propagates during low power ATWS (EOP-5A, 1A, 6A, 6B, 7B) 7 M (CREW) When 2 areas exceed their Max Safe Temperature limit, crew Emergency Depressurizes by opening 6 SRVs before reaching Max Safe Temperature in a third area.

(or) During emergency depressurization, SRV-71G fails to open, 8 C(BOP) zdimssws1e(1) requiring opening a LLS SRV.

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

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

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 1 Page 3 of 2 Quantitative Attributes Table ES-304-1 Attribute Actual Description Target Malfunctions after 1. Hydraulic block ATWS EOP entry 1-2 2 2. SRV-71D failure to open

1. RPS B MG Set trip Abnormal Events 2-4 2 2. RCIC steam leak in Secondary Ctmt
1. ATWS Major Transients 1-2 2 2. RCIC steam line break requiring ED EOP entries requiring 1. EOP-1A,6A,7A substantive action 1-2 2 2. EOP-5A EOP contingencies
1. EOP-6A,7A requiring substantive 0-2 2 2. EOP-6B action
1. Inject SLC/insert control rods EOP based Critical 2. Emergency depressurize when 2 areas exceed Tasks 2-3 2 Max Safe temperature Normal Events N/A 0 N/A Reactivity Manipulations N/A 1 1. Withdraw control rods per the startup sequence
1. IRM H erratic
2. TEC pump A trip
3. RPS MG Set B trip Instrument/

Component Failures N/A 7 4. SGT A exhaust fan trip

5. RCIC isolation valve failure
6. SDV hyGraulic block
7. SRV-71G failure to open
1. IRM H erratic
2. TEC pump A trip
3. RPS MG Set B trip
4. SGT A exhaust fan trip Total Malfunctions N/A 8 5. Failure of RCIC auto isolation
6. RCIC isolation valve breaker trip
7. SDV hydraulic block
8. SRV-71G failure to open Top 10 systems and operator actions important to risk that are tested:

Reactor Protection System Primary Containment Isolation System Automatic Depressurization System(ADS)

Manually Initiate ADS

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 1 Page 4 of 2 SCENARIO

SUMMARY

Startup is in progress at 4.5% power and rated pressure at the beginning of the current fuel cycle when the crew takes the shift. No equipment is out of service.

After the crew takes the watch, the ATC operator will withdraw control rods IAW the control rod withdrawal sequence and the GOP.

While control rods are being withdrawn, IRM H indication will become erratic, simulating a degraded connection undervessel. Control rod blocks will be generated, requiring the ATC to respond IAW the alarm response procedure and bypass IRM H. The CRS will review TS and determine a potential LCO is required for IRM H relative to TS 3.3.1.1 and TRM 3.3.1.

When the Tech Spec assessment is complete, TEC pump A will trip. The BOP operator will respond and start a standby TEC pump, B or C, IAW the alarm response procedure. AOP entry will not be required since TEC pump C remains operating and the heat load on the TEC system is minimal with the Turbine Generator not yet on line.

After actions for TEC are complete, RPS MG Set B will trip, resulting in a half scram, half Group 1, 2, 3, 7 isolations, and a full Group 6 isolation.

RWCU will isolate and the pump will trip. Standby Gas Treatment A exhaust fan will trip. The crew will transfer RPS B to its alternate supply per the alarm response procedure, reset the Div 2 half scram IAW 2.1.5, and enter 2.1.22 to begin recovery from group isolations. The CRS will address TS 3.6.4.3 for SGT A exhaust fan failure. The crew will not be given time to complete all of the actions of 2.1.22, such as restoring Reactor Building HVAC realignments, due to the timing of the scenario.

After RPS B power has been restored and the lead examiner is ready to proceed, a small leak will develop from the RCIC steam supply in the NE Quad area. Secondary containment temperature in this area will rise, and if requested, field reports will inform the crew the leak is from RCIC. If the isolation setpoint is reached before the crew attempts to close RCIC steam supply valves, the automatic isolation will not occur. When the BOP attempts to manually isolate the steam supply, the MOVs will lose power while stroking. The crew will enter EOP-5A, EOP-1A, and scram due to an unisolable steam leak into secondary containment raising area temperature toward the Max Safe limit.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 1 Page 5 of 2 When the reactor is scrammed, a low power ATWS will occur due to hydraulic block of one scram discharge volume, and EOP-6A and 7A will be entered via EOP-1A. Power will be below 3%. The crew will inject SLC and/or install the necessary PTMs to bypass interlocks and drive control rods in individually via RMCS. Bypass valves will be available to control RPV pressure. Feedwater and HPCI will be available for RPV level control.

After the crew has stabilized conditions following the scram, the RCIC steam leak will degrade and cause temperature in the NE Quad and in the NW Quad to exceed the Max Safe limit, requiring emergency depressurization IAW EOP-5A and EOP-6B. When the handswitch for SRV-71G is placed to open, the SRV will remain closed, and the operator should open another SRV in its stead. Condensate will be available for RPV level control when MSCP pressure is reached.

The exercise ends when the reactor has been depressurized, RPV water level is being maintained -183 to +54, and SLC is injecting and/or control rods are being inserted via RMCS and/or scram.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 1 Page 6 of 2 Critical Tasks When control rods fail to scram, crew When 2 areas exceed their Max Safe injects SLC and/or inserts control rods Temperature limit, crew Emergency before exiting EOP-6A. (All control rods Depressurizes by opening 6 SRVs before do not have to be fully inserted to satisfy reaching Max Safe Temperature in a third this critical task; this only requires that area.

the crew is making progress to achieving all rods in by fully inserting at least 5 control rods using RMCS.)

EVENT 6 7 Safety Failure to effect shutdown of the reactor Should secondary containment parameters significance when a RPS setting has been exceeded, exceed their maximum safe operating values even at low power, would unnecessarily in more than one area, the RPV must be extend the level of degradation of the safety depressurized to preclude further degradation.

of the plant. This could further degrade into RPV depressurization places the primary damage to the principle fission product system in its lowest possible energy state, barriers if left unmitigated. The crew is rejects heat to the suppression pool in authorized and required by Conduct of preference to outside the containment, and Operations to take mitigating actions when reduces the driving head and flow of primary automatic safety systems fail to perform their systems that are unisolated and discharging intended function. Action to shut down the into the secondary containment.

reactor is required when RPS and control rod The criteria of "two or more areas" specified drive systems fail. identifies the rise in secondary containment parameters as a wide-spread problem which may pose a direct and immediate threat to secondary containment integrity, equipment located in the secondary containment, and continued safe operation of the plant.

Cueing Manual scram is initiated and numerous SPDS indication for NE and NW Quad areas control rods indicate beyond position 00 and both > 195°F reactor power not downscale on panel 9-5 indications.

Performance Operator manipulates keylocked switches for Manipulation of SRV controls on panel 9-3:

indicator SLC A and B pumps to START on panel 9-5. SRV-71A SRV-71B Operator selects individual control rods by SRV-71E depressing the respective pushbutton on the SRV-71G panel 9-5 matrix and inserts rods by SRV-71H manipulating the emergency in switch on SRV-71C panel 9-5. SRV-71D SRV-71F Performance SLC A and B pumps red lights illuminated on Crew will observe SRV light indication go from feedback panel 9-5. green to red, amber pressure switch lights illuminate, reactor pressure lowering on SPDS Operator selecting and inserting control rods and panel 9-3 and 9-5 meters and recorders, indicated by rod position decreasing to 00 for and SRV tailpipe temperatures rise on selected rod on panel 9-5. recorder MS-TR-166..

Justification There is no time limit for effecting complete Emergency Depressurization is required due to for the chosen reactor shutdown via boron injection or effects of a break spreading into and performance control rod insertion. For the timeframe of potentially affecting safety equipment and limit this low power scenario, containment limits operations in more than one area; however, are not challenged and power oscillations are emergency depressurization is not allowed not experienced. However, if the failure to until the second area exceeds its Max Safe scram EOP were to be exited, other limit. Before the Max Safe temperature limit is procedures would not provide the guidance exceeded in a third area gives reasonable time necessary to achieve reactor shutdown. for the crew to perform emergency Before exiting EOP-6A ensures guidance to depressurization before the leak hampers effect reactor shutdown is not removed.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 1 Page 7 of 2 equipment or operations in an even more widespread area.

BWR Owners App. B, step RC/Q-6,RC/Q-7 App. B, step SC/T-4.2 Group Appendix

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 2 Page 1 of 6 Facility: Cooper Nuclear Station Scenario No.: 2 Op-Test No.: CNS 14-01 Examiners: ____________________________ Operators: _____________________________

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

1. Place the third Condensate Pump and Condensate Booster Pump in service.
2. Raise reactor power by raising Reactor Recirc flow.
3. Respond to a spurious initiation of RCIC.
4. Respond to failure of Reactor Recirc Pump B seals.
5. Respond to Reactor Recirc B discharge line rupture with degraded ECCS.
6. Respond to Loss of Offsite power with failure of DG2 output breaker to automatically close.

Initial Conditions: Power ascension is in progress at 75% power following maintenance on Condensate Booster Pump C.

Inoperable Equipment: Core Spray Pump A Turnover:

The plant is at 75% power. Condensate Pump C and Condensate Booster Pump C are in standby following maintenance to repair an oil leak on Condensate Booster Pump C. Five Condensate F/Ds are in service. Planned activities for this shift are:

  • The BOP operator will place Condensate Pump C and Condensate Booster Pump C in service IAW 2.2.6, Condensate System

Core Spray Pump A is tagged out of service for motor PMs. The plant is 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> into 7 day LCO 3.5.1 Condition A for Core Spray A inoperable. It is a division 1 work week.

Scenario Notes:

This is a new scenario.

Validation Time: 90 minutes

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 2 Page 2 of 6 Event Malf. No. Event Type Event No. Description 1 N/A N (BOP) Place third CP and CBP in service (2.2.6, Condensate System)

Raise power to 80% using Reactor Recirc flow control (2.1.10, 2 N/A RX (ATC)

Station Power Changes)

I (BOP) 3 rc05 A (CREW) Spurious RCIC initiation (2.4CSCS)

TS (CRS)

C (ATC,BOP) rr10b Reactor Recirc Pump B seal #1 and #2 failure (ARP, 2.4RR, 4 A (CREW) rr11b 2.4PC) with inability to isolate.

TS (CRS)

Recirc B discharge line leak into the drywell resulting in scram and subsequent Loss of Offsite Power and inability to maintain rr12b C (ATC) RPV level, with trip of RHR A and C pumps upon initiation (EOP-1A, EOP-3A) 5 rh01a M(CREW)

When RPV level lowers to -158 CFZ (TAF) and cannot be rh01c maintained above -183 CFZ (MSCWL), crew Emergency Depressurizes by opening 6 SRVs before drywell radiation reaches 150 R/hr or entering PC Flooding.

Failure of Division 2 DG output breaker to automatically close (5.3EMPWR).

C (BOP) 6 dg03b When ECCS systems fail to automatically start due to loss of AC power, crew manually closes DG-2 output breaker to energize LP ECCS systems prior to RPV water level falling below -158 CFZ (TAF).

hp01 HPCI Oil Pump failure to auto start and subsequent trip after 7 C (ATC) hp11 manual 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)

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 2 Page 3 of 6 Quantitative Attributes Table ES-304-1 Attribute Actual Description Target Malfunctions after 1. DG-2 output breaker failure to auto close EOP entry 1-2 2 2. HPCI oil pump failure

1. Spurious RCIC initiation Abnormal Events 2-4 2 2. Recirc Pump B seal failure Major Transients 1-2 1 1. Recirc LOCA with Loss of Offsite Power EOP entries requiring 1. EOP-1A substantive action 1-2 2 2. EOP-3A EOP contingencies requiring substantive 0-2 1 1. EOP-2A action
1. Manually close DG-2 output breaker to power Div 2 EOP based Critical LP ECCS systems Tasks 2-3 2 2. Emergency depressurize when RPV level drops below -158 CFZ (TAF)

Normal Events N/A 1 1. Place third CP and CBP in service Reactivity Manipulations N/A 1 1. Raise power via Recirc flow

1. Spurious RCIC initiation
2. Recirc Pump B seal failure
3. Recirc pump B suction valve failure Instrument/ 4. Recirc A discharge line rupture Component Failures N/A 8 5. Loss of Offsite Power
6. DG-2 output breaker failure to auto close
7. RHR A and C pumps trip
8. HPCI oil pump failure
1. Spurious RCIC initiation
2. Recirc Pump B seal #1 failure
3. Recirc Pump B seal #2 failure
4. Recirc pump B suction valve failure
5. Recirc B discharge line rupture Total Malfunctions N/A 10 6. Loss of Offsite Power
7. DG-2 output breaker failure to auto close
8. RHR A pump trip
9. RHR C pump trip
10. HPCI oil pump failure Top 10 systems and operator actions important to risk that are tested:

Emergency AC power High Pressure Coolant Injection Residual Heat Removal Reactor Core Isolation Cooling Automatic Depressurization System(ADS)

Manually Initiate ADS Take Manual Action to Align ECCS

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 2 Page 4 of 6 SCENARIO

SUMMARY

The plant is operating at 75% power near the end of the current fuel cycle when the crew takes the shift. Core Spray Pump A is tagged out of service for motor PMs. Condensate Pump C and Condensate Booster Pump C are ready to be returned to service following maintenance on Condensate Booster Pump C.

After the crew takes the watch, the BOP operator will start Condensate Pump V and Condensate Booster Pump C IAW the SOP.

When Condensate Pump C and Condensate Booster Pump C are in operation, the ATC will raise power to 80% by raising the speed of the Reactor Recirc Pumps.

While power is being raised, a spurious RCIC initiation will occur due to a relay failure, and the crew will trip RCIC IAW 2.4CSCS, Inadvertent CSCS Initiation, and the ARP. This will render RCIC inoperable and require entry into TS 3.5.3 Condition A.

When the Tech Spec assessment is complete, the #1 seal on the RR Pump B fails. When actions for the #1 seal failure are complete, the #2 seal will fail on the same pump. The crew will enter 2.4RR, Reactor Recirculation Abnormal and secure Recirc Pump B. This will require implementation of controls for single recirc loop operation IAW TS 3.4.1. 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.

Drywell pressure and temperature rise requiring the crew to enter 2.4PC, Primary Containment Control, and vent the drywell.

After Recirc Pump B has been secured and drywell venting is in progress, a leak will develop on the Recirc B loop discharge line. Drywell pressure and temperature rise, resulting in a reactor scram. When the scram occurs, a loss of offsite power will occur, disabling Condensate and Feedwater.

High drywell pressure and low RPV water level are the entry conditions for EOPs 1A and 3A. RPV water level will lower due to shrink from the scram and because CRD and SLC will not be able to keep up with the leak, nor will RCIC if the crew manually restarts it. HPCI will fail to automatically initiate, and shortly after the crew starts it manually, its auxiliary lube oil pump will fail, resulting in loss of HPCI. RPV water level will continue to fall to TAF, -158.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 2 Page 5 of 6 Due to the unavailability of Core Spray Pump A and trip of RHR Pumps A and C, Division 1 low pressure ECCS will be unavailable. Division 2 DG will start on the LOOP, but its output breaker will fail to automatically close.

The crew will have to recognize this failure and manually close the output breaker from the control room to make Division 2 low pressure ECCS systems available before level goes below TAF.

Drywell pressure will be elevated, requiring operation of Torus/Drywell Spray using systems not required to maintain adequate core cooling. The crew may elect to reserve Division 2 RHR for LPCI mode in anticipation of going below TAF.

When level reaches TAF, as indicated on SPDS using Corrected Fuel Zone (CFZ) at -158, and with Division 2 core spray and LPCI lined up for injection, the crew will enter EOP-2B and conduct emergency depressurization due to RPV water level below TAF and cannot be restored and maintained above -

183 CFZ. The crew will restore level to +3 to +54 inches IAW EOP-1A using low pressure ECCS.

The exercise ends when the reactor has been depressurized, RPV water level is being restored +3 to +54 inches. The emergency classification will be made to satisfy and administrative JPM for SRO candidates after the scenario has ended.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 2 Page 6 of 6 Critical Tasks When RPV level lowers to -158 CFZ (TAF) When ECCS systems fail to automatically and cannot be maintained above -183 start due to loss of AC power, crew CFZ (MSCWL), crew Emergency manually closes DG-2 output breaker to Depressurizes by opening 6 SRVs before energize LP ECCS systems prior to RPV drywell radiation reaches 150 R/hr or water level falling below -158 CFZ (TAF) entering PC Flooding.

EVENT 5 6 Safety The MSCRWL is the lowest RPV water level Failure to recognize the auto start not significance at which the covered portion of the reactor occurring and energizing of the safety bus, core will generate sufficient steam to preclude and failure to take manual action per any clad temperature in the uncovered Procedure 5.3EMPWR will result in loss of portion of the core from exceeding 1500°F. safety-related equipment necessary to provide When water level decreases below MSCRWL adequate core cooling, otherwise resulting in with injection, clad temperatures may exceed core damage and a large offsite release.

1500°F.

Cueing Corrected Fuel Zone indication (SPDS) falls Indication and/or annunciation that all ac to -158 and lowering trend continues emergency buses are de-energized

  • Bus energized lamps extinguished
  • Circuit breaker position
  • Bus voltage
  • EDG status Control room lighting dimmed Performance Manipulation of SRV controls on panel 9-3: Manipulation of controls as required to indicator SRV-71A energize Div 2 AC emergency bus from panel SRV-71B C:

SRV-71E SRV-71G Operator places DIESEL GEN 2 BKR EG2 to SRV-71H CLOSE on panel C SRV-71C SRV-71D SRV-71F Performance Crew will observe SRV light indication go Crew will observe light indication for feedback from green to red, amber pressure switch equipment powered by Division 2 AC lights illuminate, reactor pressure lowering on illuminate on panel 9-3 and bus voltage SPDS and panel 9-3 and 9-5 meters and ~4200V on panel C recorders, and SRV tailpipe temperatures rise on recorder MS-TR-166..

Justification Before 150R/hr in the drywell was chosen This is so Emergency Depressurization will not for the chosen because this is an indicator of loss of RPV be delayed by the breaker failure to auto performance level and the shielding effect of the water, close. Per EOP-1A, TAF is the point at which limit indicating core exposure, yet it is much lower Emergency Depressurization may be entered than the 2500R/hr trigger point during RPV if level cannot be maintained above MSCWL, if Flooding that indicates gross cladding failure at least one Table 5 (high capacity) system is is in progress. MSCWL was not chosen lined up for injection. With the Div 2 because at CNS, cycling of one SRV could emergency bus energized, Div 2 ECCS result in a prompt swing in level from TAF to systems will auto align for injection.

below MSCWL, not allowing the crew time to properly execute the ED before MSCWL is indicated. Before exiting to PC Flooding was chosen because the design of the scenario provides the crew with the means to restore and maintain adequate core cooling IAW EOP-1A, and exiting to SAGs is not required or authorized.

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

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 3 Page 1 of 2 Facility: Cooper Nuclear Station Scenario No.: 3 Op-Test No.: CNS 14-01 Examiners: ____________________________ Operators: _____________________________

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

1. Shift CRD Pumps from B to A.
2. Reduce power by lowering Reactor Recirc flow.
3. Respond to Outboard MSIV 86A failing closed.
4. Respond to a control rod drifting outward.
5. Respond to a spurious Turbine trip and hydraulic block ATWS.
6. Respond to failure of Generator field breaker to open automatically.
7. Respond to CRD pump A trip.

Initial Conditions: Plant is at 100% power.

Inoperable Equipment: Core Spray Pump A Turnover:

The plant is at 100% power. CRD pumps require rotation to provide run time for CRD Pump A before scheduled maintenance on CRD Pump B next week. After CRD pumps are shifted, power will be reduced to remove Condensate Booster Pump C from service to repair an oil leak. Work Control is finalizing the repair plan and clearance order and verifying repair parts are on-hand. Also. Planned activities for this shift are:

  • Immediately following turnover, the BOP operator will shift CRD pumps IAW 2.2.8, Control Rod Drive Hydraulic System.
  • When notified by Work Control, remove Condensate Booster Pump C and Condensate Pump C from service IAW 2.2.6, Condensate System.

Core Spray Pump A is tagged out of service for motor PMs. The plant is 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> into 7 day LCO 3.5.1 Condition A for Core Spray A inoperable. It is a division 1 work week.

Scenario Notes:

This is a new scenario.

Validation Time: XX minutes

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 3 Page 2 of 2 Event Malf. No. Event Type Event No. Description 1 N/A N (BOP) Shift CRD pumps from B to A (2.2.8, CRD Hydraulic System)

Reduce power to 75% using Reactor Recirc flow control (2.1.10, 2 N/A RX (ATC) Station Power Changes)(actual power change will be ~2%

before next event)

C (ATC,BOP)

(or) Outboard MSIV 86A fails closed (2.4MSIV, Inadvertent MSIV 3 A (CREW) zdipcissws4a(1) Closure)

TS (CRS)

C (ATC) 4 Rd101827 A (CREW) Control Rod 18-35 drift outward (2.4CRD, CRD Trouble)

TS (CRS)

Spurious Turbine trip, hydraulic block ATWS >3% power with SLC A failure to initiate When control rods fail to scram, crew injects SLC and/or inserts control rods before exiting EOP-6A. (All control rods do not have to be fully inserted to satisfy this critical tc01 C (ATC) task; this only requires that the crew is making progress 5 rd02a M(CREW) to achieving all rods in by fully inserting at least 5 control rods using RMCS.)

rd02b Inhibit ADS before ADS valves open on an initiation signal.

During failure to scram conditions with power >3%, lower RPV water level to below -60 CFZ and control between -

60 to -183 CFZ prior to exiting EOP-7A.

Failure of Generator field breaker to open automatically (2.1.5 6 eg01 C (BOP)

Att 5) 7 rd08a C (ATC) CRD Pump A trip (ARP 5-2/A-6)

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

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

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 3 Page 3 of 2 Quantitative Attributes Table ES-304-1 Attribute Actual Description Target

1. Failure of Generator field breaker to open Malfunctions after EOP entry 1-2 2 automatically
2. CRD Pump A trip
1. Outboard MSIV 86A fails closed Abnormal Events 2-4 2 2. Control Rod 18-35 drift outward
1. Turbine trip, hydraulic block ATWS >3% power, Major Transients 1-2 1 SLC A fail to start EOP entries requiring substantive action 1-2 1 1. EOP-1A, 6A EOP contingencies requiring substantive 0-2 1 1. EOP-7A action
1. Inject SLC/insert control rods EOP based Critical 2. Inhibit ADS.

Tasks 2-3 3 3. Lower RPV water level to below -60 CFZ and control between -60 to -183 CFZ.

Normal Events N/A 1 1. Shift CRD pumps from B to A Reactivity Manipulations N/A 1 1. Reduce power via Recirc flow

1. Outboard MSIV 86A fails closed
2. Control Rod 18-35 drift out Instrument/

Component Failures N/A 5 3. Spurious Turbine trip/Hydraulic Block ATWS

4. Generator field breaker failure to open
5. CRD pump A trip
1. Outboard MSIV 86A fails closed
2. Control Rod 18-35 drift out
3. Spurious Turbine trip
4. Hydraulic Block of S SDV Total Malfunctions N/A 8 5. Hydraulic Block of N SDV
6. Generator field breaker failure to open
7. SLC A fail to start
8. CRD pump A trip Top 10 systems and operator actions important to risk that are tested:

Reactor Protection System Primary Containment Isolation System Nuclear Boiler Instrumentation

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 3 Page 4 of 2 SCENARIO

SUMMARY

The plant is operating at 100% power near the end of the current fuel cycle when the crew takes the shift. Core Spray Pump A is out of service to repair a jacket water leak.

When the crew takes the watch, the BOP operator will rotate CRD pumps, placing A in service and securing B IAW the SOP.

After CRD pumps have been shifted, the ATC operator will begin reducing power to 75% by lowering Recirc flow IAW the GOP.

During the power reduction, Outboard MSIV 86A will slowly fail closed.

Reactor pressure and flux will rise. Due to the increase in RPV pressure power will rise above the rod monitor block setpoint but will not exceed the RPS setpoint. The crew will enter 2.4MSIV and take the handswitch for the failed MSIV to CLOSE on panel 9-4 and lower power by lowering recirc flow to 40 mlbm/hr, followed by inserting control rods to below 70% power. MSIV 86A will not indicate fully closed. The CRS will enter TS 3.6.1.3 Condition A for the failed MSIV.

When control rod 18-25 is inserted during the power reduction, it will spuriously drift out after the RO inserts it. The ATC operator will reinsert the control rod using RMCS IAW 2.4CRD. The control rod will have to be fully inserted using its scram test and disarmed. The CRS will enter TS 3.1.3 Condition C for the control rod.

When the Tech Spec assessment is complete, a spurious trip of the main turbine will occur. The reactor will fail to shut down due to hydraulic blockages in both SDVs. Power will be approximately 25%. The crew will enter EOPs 6A and 7A via EOP-1A. The crew will inject SLC and/or install the necessary PTMs to bypass interlocks and drive control rods in individually via RMCS. Bypass valves will be available to control RPV pressure. Feedwater and HPCI will be available for RPV level control; however, injection from these systems will have to be stopped to lower RPV level below -60 CFZ. Level will then be maintained between -60 and

-183 CFZ.

Following the turbine trip, the generator field breaker will fail to automatically open when the generator trips on reverse power, and the BOP operator should trip it manually. SLC pump A will fail to start when attempted by the ATC. Also, CRD pump A will trip, requiring the ATC to start CRD pump B to accomplish control rod insertion.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 3 Page 5 of 2 The exercise ends when RPV water level is being controlled -183 to +54 and when SLC is injecting and/or control rods are being inserted via RMCS and/or scram.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 3 Page 6 of 2 Critical Tasks When control rods fail to scram, crew Inhibit ADS before ADS valves open on an injects SLC and/or inserts control rods initiation signal.

before exiting EOP-6A. (All control rods do not have to be fully inserted to satisfy this critical task; this only requires that the crew is making progress to achieving all rods in by fully inserting at least 5 control rods using RMCS.)

EVENT 5 5 Safety Failure to effect shutdown of the reactor when With a Reactor Scram required, reactor not significance a RPS setting has been exceeded would shut down, and conditions for ADS blowdown unnecessarily extend the level of degradation are met, INHIBIT ADS to prevent an of the safety of the plant. This could further uncontrolled RPV depressurization and cold degrade into damage to the principle fission water injection from low pressure sources to product barriers if left unmitigated. The crew prevent causing a significant power excursion.

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.

Cueing Manual scram is initiated and numerous ADS Timer initiated alarm on panel 9-3-1/A-1 control rods indicate beyond position 00 and reactor power not downscale on panel 9-5 indications.

Performance Operator manipulates keylocked switches for Manipulation of ADS A and ADS B Inhibit indicator SLC B pump to START on panel 9-5. switches on panel 9-3 vertical section.

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.

Performance SLC B pump red light illuminated, SLC Inhibit switches click into the vertical, inhibit feedback discharge pressure rising, and SLC tank level position on panel 9-3.

lowering on panel 9-5.

Receipt of ADS inhibited alarm panel 9-3-1/D-Operator selecting and inserting control rods 1.

indicated by rod position decreasing to 00 for selected rod on panel 9-5.

Justification There is no time limit for effecting complete The 105 second ADS timer allows sufficient for the chosen reactor shutdown via boron injection or time for the crew to recognize and override performance control rod insertion. For the timeframe of automatic operation of the system. As long as limit this scenario, containment limits are not ADS is inhibited before ADS valves open, closely challenged and power oscillations are reactor pressure will not be reduced to the not experienced. However, if the failure to shutoff heads of high volume, cold water scram EOP were to be exited, other systems.

procedures would not provide the guidance necessary to achieve reactor shutdown.

Before exiting EOP-6A ensures guidance to effect reactor shutdown is not removed.

BWR Owners App. B, step RC/Q-6,RC/Q-7 App. B, step RC/Q-6 Group Appendix

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 3 Page 7 of 2 Critical Tasks During failure to scram conditions with power >3%, lower RPV water level to below -60 CFZ and control between -60 to -183 CFZ prior to exiting EOP-7A.

EVENT 5 Safety To prevent or mitigate the consequences of significance any large irregular neutron flux oscillations induced by neutronic/thermal-hydraulic instabilities, RPV water level is lowered sufficiently below the elevation of the feedwater sparger nozzles. This places the feedwater spargers in the steam space providing effective heating of the relatively cold feedwater and eliminating the potential for high core inlet subcooling. For conditions that are susceptible to oscillations, the initiation and growth of oscillations is principally dependent upon the subcooling at the core inlet; the greater the subcooling, the more likely oscillations will commence and increase in magnitude.

24" below the lowest nozzle in the feedwater sparger has been selected as the upper bound of the RPV water level control band.

This water level is sufficiently low that steam heating of the injected water will be at least 65% to 75% effective (i.e., the temperature of the injected water will be increased to 65% to 75% of its equilibrium value in the steam environment). This water level is sufficiently high that the capability to bypass the low RPV water level MSIV isolation should be able to control RPV water level with feedwater pumps to preclude the isolation.

Maintaining RPV level above -183" (MSCRWL) assures adequate core cooling via steam cooling with injection. This 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. Affiliated with this task is the requirement to only use outside the shroud injection systems due to the potential for a large power excursion that may otherwise result.

Cueing Manual scram is initiated and numerous control rods indicate beyond position 00 and reactor power >3% on panel 9-5 indications and SPDS and RPV level is >-60CFZ on SPDS.

Performance Operator manipulates Feedwater HMIs on indicator panel 9-5 or panel A as necessary to stop FW injection until RPV level goes below -60CFZ.

Operator manipulates HPCI controls on panel 9-3 to stop HPCI injection until RPV level is below -60CFZ.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 3 Page 8 of 2 Performance Feedwater flow indication on panel 9-5 feedback indicate zero.

HPCI flow indication on panel 9-3 indicates zero and/or HPCI injection MOV indicates closed.

Justification Applicability for this CT is during EOP-7A for the chosen conditions where it is necessary to lower level performance to control power with Table 17 condition NOT limit met (i.e. no high energy input into primary containment). There is no time limit for this lowering level, but it establishes margin to conditions where fuel damaging power oscillations may theoretically occur. Before exiting EOP-7A was chosen because other procedures would not provide the guidance necessary to establish margin for power oscillation mitigation. Before exiting EOP-7A ensures guidance to effect this control is not removed.

NOTE This critical task must be evaluated carefully based on the level changes. If power is reduced significantly below 3%,

reactor water level may continue to rise above -60" with only CRD and SLC while driving rods this would not result in an UNSAT on this critical task unless the Operator failed to remove RCIC to help restore and maintain level in the band.

BWR Owners App. B, Contingency #5 Group Appendix

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 4(Alt) Page 1 of 6 Facility: Cooper Nuclear Station Scenario No.: 4(Alt) Op-Test No.: CNS 14-01 Examiners: ____________________________ Operators: _____________________________

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

1. Respond to APRM C signal failure high.
2. Respond to Main Bypass Valve B failing open.
3. Respond to DEH Pump B filter clogging.
4. Respond to Feedwater Line A break inside containment (LOCA).
5. Respond to HPCI failure.
6. Respond to failure of ECCS pump to automatically initiate.

Initial Conditions: The plant is operating at 100%.

Inoperable Equipment: Core Spray Pump A Turnover:

The plant is at 100% power. Planned activities for this shift are:

  • Maintain rated power..

Core Spray Pump A is tagged out of service for motor PMs. The plant is 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> into 7 day LCO 3.5.1 Condition A for Core Spray A inoperable. It is a division 1 work week.

Scenario Notes:

This is a new scenario.

Validation Time: XX minutes

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 4(Alt) Page 2 of 6 Event Malf. No. Event Type Event No. Description I (ATC) APRM C signal failure high (ARP panel 9-5-1/A-4,A-7,B-8; 9 1 nm09c TS (CRS) 2/A-1,B-1; 2.1.5)

C (ATC)

Bypass valve B fails open (ARP panel B-1/F-1, 2.4EXT-STM, 2 tc07b A (CREW) 2.4RXPWR)

TS (CRS)

C (BOP) 3 caep A (CREW) DEH Pump B filter clogging (ARP panel B-1/E-7) tc01 fw18a DEH pressure loss-Turbine trip / FW Line A break in DW / RPS rr20a C (ATC) failure to de-energize / ARI works 4 rp01a M(CREW) When RPV level lowers to -158 CFZ (TAF) and cannot be maintained above -183 CFZ (MSCWL), crew Emergency rp01b Depressurizes by opening 6 SRVs before drywell radiation reaches 150 R/hr or entering PC Flooding.

rp01c rp01d 5 (rf)hp06a4 C (BOP) HPCI injection MO-19 fails closed cs06a Failure of all Low Pressure ECCS to start cs06b When ECCS systems fail to automatically start due to rh08a C (BOP) loss of AC power, crew manually starts and aligns at least 6

rh08b two LP ECCS pumps prior to RPV water level falling below -158 CFZ (TAF).

rh08c rh08d (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal (TS) Tech Spec

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

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 4(Alt) Page 3 of 6 Quantitative Attributes Table ES-304-1 Attribute Actual Description Target Malfunctions after 1. HPCI injection valve failure EOP entry 1-2 2 2. LP ECCS failure to auto start

1. Bypass valve B fails open Abnormal Events 2-4 2 2. DEH Pump B filter clogging Major Transients 1-2 1 1. FW line A break in DW LOCA EOP entries requiring 1. EOP-1A substantive action 1-2 2 2. EOP-3A EOP contingencies requiring substantive 0-2 1 1. EOP-2A action
1. Manually start and align at least two LP ECCS EOP based Critical pumps Tasks 2-3 2 2. Emergency depressurize when RPV level drops below -158 CFZ (TAF)

Normal Events N/A 0 1. none Reactivity Manipulations N/A 0 1. none

1. APRM C signal failure high
2. Bypass valve B fails open
3. DEH Pump B filter clogging Instrument/

Component Failures N/A 7 4. Turbine trip/FW line A break in DW

5. RPS failure to de-energize
6. HPCI injection valve MO-19 failure
7. Failure of LP ECCS to auto start
1. APRM C signal failure high
2. Bypass valve B fails open
3. DEH Pump B filter clogging
4. DEH pressure loss/Turbine trip
5. FW line A break in DW Total Malfunctions N/A 9 6. Recirc LOCA to simulate FW check valve leakage
7. RPS failure to de-energize (4 total malfunctions)
8. HPCI injection valve MO-19 failure
9. Failure of LP ECCS to auto start (6 total malfunctions)

Top 10 systems and operator actions important to risk that are tested:

High Pressure Coolant Injection Residual Heat Removal Reactor Core Isolation Cooling Automatic Depressurization System (ADS)

Manually Initiate ADS Take Manual Action to Align ECCS

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 4(Alt) Page 4 of 6 SCENARIO

SUMMARY

The plant is operating at 100% power near the end of the current fuel cycle when the crew takes the shift. Core Spray Pump A is out of service to repair a jacket water leak.

After the crew takes the watch, APRM C will fail high, causing a control rod block and half scram. The ATC will respond IAW the ARPs, bypass APRM C, and reset the half scram IAW 2.1.5. The CRS will refer to TS 3.3.1.1 and T3.1.1 for initiating a Potential LCO for APRM C.

When the Tech Spec assessment is complete, Bypass valve B will fail open.

The BOP operator will respond IAW the ARP. Since the amount of extraction steam will be reduced, FW temperature will lower causing reactor power to rise (~1%). The crew will enter 2.4RXPWR and 2.4EXT-STM and decrease power by lowering Recirc flow IAW 2.1.10 to stabilize power below 100%. The CRS will enter TS 3.7.7 Condition A for the failed bypass valve.

When the Tech Spec assessment is complete, the discharge filter for DEH pump B will plug. DEH fluid pressure will lower. The BOP operator will respond IAW the filter dp high ARP, and the crew will rotate to DEH pump A.

After EHC pump A has been started and DEH pump B is secured, following a delay DEH pressure will fall, resulting in a Turbine trip. RPS will fail to de-energize, but ARI will function to insert all control rods. Upon the transient from the Turbine trip, Feedwater line A will rupture in the Drywell, and the A FW check valve disc will fail to seat, allowing RPV inventory loss back into the Drywell. Drywell pressure and temperature rise.

High drywell pressure and low RPV water level are the entry conditions for EOPs 1A and 3A. RPV water level will lower. CRD and SLC will not be able to keep up with the leak. RCIC will not help since injects into the FW line A, which is broken, so RCIC will not stabilize RPV level. HPCI injection valve MO-19 will fail to automatically open, and its breaker will trip when the BOP operator attempts to open it manually. RPV water level will continue to fall to TAF, -158.

Neither Division 1 nor Division 2 Low Pressure ECCS pumps will start on the LOCA signal. The crew will have to recognize this failure and manually start and align low pressure ECCS systems available before level goes below TAF.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 4(Alt) Page 5 of 6 Drywell pressure will be elevated, requiring operation of Torus/Drywell Spray using systems not required to maintain adequate core cooling. The crew may elect to reserve all RHRs for LPCI mode in anticipation of going below TAF.

When level reaches TAF, as indicated on SPDS using Corrected Fuel Zone (CFZ) at -158, and with Core Spray and/or LPCI lined up for injection, the crew will enter EOP-2B and conduct emergency depressurization due to RPV water level below TAF and cannot be restored and maintained above -183 CFZ. The crew will restore level to +3 to +54 inches IAW EOP-1A using low pressure ECCS.

The exercise ends when the reactor has been depressurized, RPV water level has been restored +3 to +54 inches. The emergency classification will be made to satisfy and administrative JPM for SRO candidates after the scenario has ended.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 14-01 Scenario 4(Alt) Page 6 of 6 Critical Tasks When RPV level lowers to -158 CFZ (TAF) When ECCS systems fail to automatically and cannot be maintained above -183 start due to loss of AC power, crew CFZ (MSCWL), crew Emergency manually starts and aligns at least two LP Depressurizes by opening 6 SRVs before ECCS pumps prior to RPV water level drywell radiation reaches 150 R/hr or falling below -158 CFZ (TAF).

entering PC Flooding.

EVENT 5 6 Safety The MSCRWL is the lowest RPV water level Failure to recognize the auto start not significance at which the covered portion of the reactor occurring and failure to take manual action per core will generate sufficient steam to Conduct of Ops will result in loss of safety-preclude any clad temperature in the related equipment necessary to provide uncovered portion of the core from exceeding adequate core cooling, otherwise resulting in 1500°F. When water level decreases below core damage and a large offsite release.

MSCRWL with injection, clad temperatures may exceed 1500°F.

Cueing Corrected Fuel Zone indication (SPDS) falls Indication CS and LPCI pumps are not to -158 and lowering trend continues running:

  • On panel 9-3, Red lamps on pumps handswitches extinguished and green lamps on Performance Manipulation of SRV controls on panel 9-3: Manipulation of controls as required to start at indicator SRV-71A least 2 pumps on panel 9-3:

SRV-71B SRV-71E Operator places pump handswitch to START SRV-71G CS A, CS B, RHR A, RHR B, RHR C, RHR D SRV-71H SRV-71C Performance Crew will observe SRV light indication go Crew will observe red lamp illuminate and feedback from green to red and reactor pressure green lamp extinguish for respective CS and/or lowering on SPDS and panel 9-3 and 9-5 RHR pumps on panel 9-3 meters and recorders.

Justification Before 150R/hr in the drywell was chosen This is so Emergency Depressurization will not for the chosen because this is an indicator of loss of RPV be delayed by the LP ECCS system failure to performance level and the shielding effect of the water, auto start. Per EOP-1A, TAF is the point at limit indicating core exposure, yet t is much lower which Emergency Depressurization may be than the 2500R/hr trigger point during RPV entered if level cannot be maintained above Flooding that indicates gross cladding failure MSCWL, if at least one Table 5 (high capacity) is in progress. MSCWL was not chosen system is lined up for injection.

because at CNS, cycling of one SRV could result in a prompt swing in level from TAF to below MSCWL, not allowing the crew time to properly execute the ED before MSCWL is indicated. Before exiting to PC Flooding was chosen because the design of the scenario provides the crew with the means to restore and maintain adequate core cooling IAW EOP-1A, and exiting to SAGs is not required or authorized.

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

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