Latest revision as of 18:37, 4 February 2020

Text

2017-03 Final Outlines
	ML17122A086
Person / Time
Site:	Cooper
Issue date:	03/16/2017
From:	Vincent Gaddy; Operations Branch IV
To:	; Nebraska Public Power District (NPPD)
References
	Download: ML17122A086 (42)
	v • d • e

ES-401 BWR Examination Outline Form ES-401-1 Facility: Cooper Nuclear Station Date of Exam: March 2017 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 3 4 4 3 3 20 4 3 7 Emergency &

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

Plant 2 2 1 1 1 1 1 1 1 1 1 1 12 0 2 1 3 Systems Tier Totals 4 3 4 5 4 3 3 3 3 3 3 38 4 4 8 1 2 3 4 1 2 3 4

3. Generic Knowledge and Abilities 10 7 Categories 3 3 2 2 2 2 1 2 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only 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 Rev 4

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 / SRO)

E/APE # / Name / Safety Function K K K A A G* K/A Topic(s) IR #

1 2 3 1 2 Knowledge of the reasons for the following responses 295001 Partial or Complete Loss of Forced as they apply to PARTIAL OR COMPLETE LOSS OF X 3.4 1 Core Flow Circulation / 1 & 4 FORCED CORE FLOW CIRCULATION: (CFR: 41.5)

AK3.01 Reactor water level response Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C.

295003 Partial or Complete Loss of AC / 6 X 4.2 2 POWER: (CFR: 41.7)

AA1.02 Emergency generators Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF D.C.

295004 Partial or Total Loss of DC Pwr / 6 X POWER: (CFR: 41.10) 3.2 3 AA2.01 Cause of partial or complete loss of D.C.

power 2.1.23 Ability to perform specific system and integrated 295005 Main Turbine Generator Trip / 3 X plant procedures during all modes of plant operation. 4.3 4 (CFR: 41.10)

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

295006 SCRAM / 1 X 3.7 5 (CFR: 41.8 to 41.10)

AK1.03 Reactivity control Knowledge of the interrelations between CONTROL 295016 Control Room Abandonment / 7 X ROOM ABANDONMENT and the following: (CFR: 41.7) 4.4 6 AK2.01 Remote shutdown panel: Plant-Specific Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF 295018 Partial or Total Loss of CCW / 8 X 3.1 7 COMPONENT COOLING WATER: (CFR: 41.5)

AK3.07 Cross-connecting with backup systems Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF 295019 Partial or Total Loss of Inst. Air / 8 X 3.0 8 INSTRUMENT AIR: (CFR: 41.7)

AA1.03 Instrument air compressor power supplies Knowledge of the reasons for the following responses as they apply to LOSS OF SHUTDOWN COOLING:

295021 Loss of Shutdown Cooling / 4 X (CFR: 41.5) 3.6 9 AK3.05 Establishing alternate heat removal flow paths Ability to determine and/or interpret the following as they 295023 Refueling Acc / 8 X apply to REFUELING ACCIDENTS: (CFR: 41.10) 3.4 10 AA2.02 Fuel Pool Level 2.2.44 Ability to interpret control room indications to verify status and operation of a system, and understand 295024 High Drywell Pressure / 5 X 4.2 11 how operator actions and directives affect plant and system conditions. (CFR: 41.5)

Knowledge of the operational implications of the following concepts as they apply to HIGH REACTOR 295025 High Reactor Pressure / 3 X PRESSURE: (CFR: 41.8 to 41.10) 3.6 12 EK1.03 Safety/relief valve tailpipe temperature /

pressure relationships Knowledge of the interrelations between 295026 Suppression Pool High Water Temp. SUPPRESSION POOL HIGH WATER TEMPERATURE X 3.5 13

/5 and the following: (CFR: 41.7)

EK2.06 Suppression pool level 295027 High Containment Temperature / 5 NOT APPLICABLE Rev 4

ES-401 3 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (RO / SRO)

E/APE # / Name / Safety Function K K K A A G* K/A Topic(s) IR #

1 2 3 1 2 Ability to operate and/or monitor the following as they apply to HIGH DRYWELL TEMPERATURE:

295028 High Drywell Temperature / 5 X 3.8 14 (CFR: 41.7)

EA1.01 Drywell spray: Mark-I&II Ability to determine and/or interpret the following as they apply to LOW SUPPRESSION POOL WATER LEVEL:

295030 Low Suppression Pool Wtr Lvl / 5 X (CFR: 41.10) 3.5 15 EA2.04 Drywell/ suppression chamber differential pressure: Mark-I&II 2.4.1 Knowledge of EOP entry conditions and 295031 Reactor Low Water Level / 2 X 4.6 16 immediate action steps. (CFR: 41.10)

Knowledge of the operational implications of the following concepts as they apply to SCRAM 295037 SCRAM Condition Present and CONDITION PRESENT AND REACTOR POWER Reactor Power Above APRM Downscale or X 4.1 17 ABOVE APRM DOWNSCALE OR UNKNOWN:

Unknown / 1 (CFR: 41.8 to 41.10)

EK1.02 Reactor water level effects on reactor power Knowledge of the interrelations between HIGH OFF-295038 High Off-site Release Rate / 9 X SITE RELEASE RATE and the following: (CFR: 41.7) 3.6 18 EK2.02 Offgas system Knowledge of the reasons for the following responses as they apply to PLANT FIRE ON SITE:

600000 Plant Fire On Site / 8 X 2.8 19 AK3.04 Actions contained in the abnormal procedure for plant fire on site Ability to operate and/or monitor the following as they 700000 Generator Voltage and Electric Grid apply to GENERATOR VOLTAGE AND ELECTRIC X 3.9 20 Disturbances / 6 GRID DISTURBANCES: (CFR: 41.5 and 41.10)

AA1.05 Engineered safety features Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C.

295003 Partial or Complete Loss of AC / 6 X 3.5 76 POWER : (CFR: 43.5)

AA2.03 Battery status: Plant-Specific 2.2.40 Ability to apply Technical Specifications for a 295021 Loss of Shutdown Cooling / 4 X 4.7 77 system. (CFR:43.2 / 43.5)

Ability to determine and/or interpret the following as 295023 Refueling Acc / 8 X they apply to REFUELING ACCIDENTS: (CFR: 43.5) 4.1 78 AA2.04 Occurrence of fuel handling accident 2.4.18 Knowledge of the specific bases for EOPs.

295038 High Off-site Release Rate / 9 X 4.0 79 (CFR: 43.1)

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

295031 Reactor Low Water Level / 2 X 4.8 80 (CFR: 43.5)

EA2.04 Adequate core cooling 2.4.30 Knowledge of events related to system operation

/ status that must be reported to internal organizations 600000 Plant Fire On Site / 8 or external agencies, such as the State, the NRC, or the 4.1 81 X transmission system operator.

(CFR: 43.5)

Ability to determine and/or interpret the following as they 700000 Generator Voltage and Electric Grid apply to GENERATOR VOLTAGE AND ELECTRIC Disturbances / 6 X GRID DISTURBANCES: (CFR:43.5) 3.8 82 AA2.05 Operational status of offsite circuit 3/ 3/ 20/

K/A Category Totals: 3 3 4 4 Group Point Total:

4 3 7 Rev 4

ES-401 4 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (RO / SRO)

E/APE # / Name / Safety Function K K K A A G K/A Topic(s) IR #

1 2 3 1 2

Knowledge of the operational implications of the following concepts as they apply to LOSS OF MAIN CONDENSER 295002 Loss of Main Condenser Vac / 3 X 3.6 21 VACUUM : (CFR: 41.8 to 41.10)

AK1.03 Loss of heat sink 295007 High Reactor Pressure / 3 NOT SELECTED Knowledge of the interrelations between HIGH REACTOR 295008 High Reactor Water Level / 2 X WATER LEVEL and the following: (CFR: 41.7 / 45.8) 3.6 22 AK2.03 Reactor water level control 295009 Low Reactor Water Level / 2 NOT SELECTED 295010 High Drywell Pressure / 5 NOT SELECTED 295011 High Containment Temp / 5 NOT SELECTED 295012 High Drywell Temperature / 5 NOT SELECTED 295013 High Suppression Pool Temp. / 5 NOT SELECTED 295014 Inadvertent Reactivity Addition / 1 NOT SELECTED Knowledge of the reasons for the following responses 295015 Incomplete SCRAM / 1 X as they apply to INCOMPLETE SCRAM : (CFR: 41.5) 3.4 23 AK3.01 Bypassing rod insertion blocks 295017 High Off-site Release Rate / 9 NOT SELECTED Ability to operate and/or monitor the following as they apply to INADVERTENT CONTAINMENT ISOLATION :

295020 Inadvertent Cont. Isolation / 5 & 7 X 2.9 24 (CFR: 41.7)

AA1.03 Containment ventilation system: Plant-Specific Ability to determine and/or interpret the following as 295022 Loss of CRD Pumps / 1 X they apply to LOSS OF CRD PUMPS : (CFR: 41.10) 3.5 25 AA2.01 Accumulator pressure 295029 High Suppression Pool Wtr Lvl / 5 NOT SELECTED 295032 High Secondary Containment Area NOT SELECTED Temperature / 5 295033 High Secondary Containment Area NOT SELECTED Radiation Levels / 9 295034 Secondary Containment 2.4.31 Knowledge of annunciator alarms, indications, or X 4.2 28 Ventilation High Radiation / 9 response procedures. (CFR: 41.10) 295035 Secondary Containment High NOT SELECTED Differential Pressure / 5 Knowledge of the operational implications of the following concepts as they apply to SECONDARY 295036 Secondary Containment High X CONTAINMENT HIGH SUMP/AREA WATER LEVEL : 2.6 27 Sump/Area Water Level / 5 (CFR: 41.8 to 41.10)

EK1.02 Electrical ground/ circuit malfunction 500000 High CTMT Hydrogen Conc. / 5 NOT SELECTED 2.4.4 Ability to recognize abnormal indications for system 295014 Inadvertent Reactivity Addition / 1 X operating parameters that are entry-level conditions for 4.7 83 emergency and abnormal operating procedures.(CFR: 43.2)

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

295029 High Suppression Pool Wtr Lvl / 5 X 3.9 84 (CFR: 43.5)

EA2.01 Suppression pool water level 500000 High CTMT Hydrogen Conc. / 5 X 2.4.6 Knowledge of EOP mitigation strategies. (CFR: 43.5) 4.7 85 1 1 7/

K/A Category Point Totals: 2 1 1 1 / / Group Point Total:

3 1 2 Rev 4

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

Knowledge of electrical power supplies to the 203000 RHR/LPCI: Injection X following: (CFR: 41.7) 2.7 26 Mode K2.03 Initiation logic Knowledge of the effect that a loss or malfunction of the SHUTDOWN COOLING SYSTEM (RHR 205000 Shutdown Cooling X SHUTDOWN COOLING MODE) will have on 3.3 29 following: (CFR: 41.7)

K3.01 Reactor pressure Knowledge of shutdown cooling system (RHR shutdown cooling mode) design feature(s) and/or 205000 Shutdown Cooling X 3.7 30 interlocks which provide for the following: (CFR: 41.7)

K4.02 High pressure isolation: Plant-Specific Knowledge of the operational implications of the following concepts as they apply to HIGH 206000 HPCI X PRESSURE COOLANT INJECTION SYSTEM : 3.3 31 (CFR: 41.5)

K5.05 Turbine speed control: BWR-2,3,4 Knowledge of the effect that a loss or malfunction of the following will have on the HIGH PRESSURE 206000 HPCI X 3.3 32 COOLANT INJECTION SYSTEM : (CFR: 41.7)

K6.02 D.C. power: BWR-2,3,4 207000 Isolation (Emergency)

NOT APPLICABLE Condenser Knowledge of LOW PRESSURE CORE SPRAY SYSTEM design feature(s) and/or interlocks which 209001 LPCS X 3.0 33 provide for the following: (CFR: 41.7)

K4.04 Line break detection 209002 HPCS NOT APPLICABLE Ability to predict and/or monitor changes in parameters associated with operating the STANDBY 211000 SLC X LIQUID CONTROL SYSTEM controls including: 3.7 34 (CFR: 41.5)

A1.08 RWCU system lineup Ability to (a) predict the impacts of the following on the REACTOR PROTECTION SYSTEM ; and (b) based on those predictions, use procedures to 212000 RPS X 4.0 35 correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5)

A2.16 Changing mode switch position Ability to monitor automatic operations of the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM 215003 IRM X 3.5 36 including: (CFR: 41.7)

A3.04 Control rod block status Ability to manually operate and/or monitor in the control room: (CFR: 41.7) 215004 Source Range Monitor X 3.4 37 A4.07 Verification of proper functioning /

operability 2.1.7 Ability to evaluate plant performance and make operational judgments based on operating 215005 APRM / LPRM X 4.4 38 characteristics, reactor behavior, and instrument interpretation. (CFR: 41.5)

Knowledge of the physical connections and/or cause/effect relationships between REACTOR CORE 217000 RCIC X ISOLATION COOLING SYSTEM (RCIC) and the 3.5 39 following: (CFR: 41.2 to 41.9)

K1.02 Nuclear boiler system Rev 4

ES-401 6 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 1 (RO / 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

Knowledge of electrical power supplies to the 218000 ADS X following: (CFR: 41.7) 3.1 40 K2.01 ADS logic Knowledge of the effect that a loss or malfunction of the PRIMARY CONTAINMENT ISOLATION 223002 PCIS/Nuclear Steam X SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF 3.3 41 Supply Shutoff will have on following: (CFR: 41.7)

K3.20 Standby gas treatment system Knowledge of the operational implications of the following concepts as they apply to RELIEF/SAFETY 239002 SRVs X 3.7 42 VALVES: (CFR: 41.5)

K5.02 Safety function of SRV operation Knowledge of the effect that a loss or malfunction of 259002 Reactor Water Level the following will have on the REACTOR WATER X 3.3 43 Control LEVEL CONTROL SYSTEM: (CFR: 41.7)

K6.02 A.C. power Ability to predict and/or monitor changes in parameters associated with operating the REACTOR 259002 Reactor Water Level WATER LEVEL CONTROL SYSTEM controls X 2.9 44 Control including: (CFR: 41.5)

A1.05 FWRV/startup level control position: Plant-Specific Ability to (a) predict the impacts of the following on the STANDBY GAS TREATMENT SYSTEM ; and (b) based on those predictions, use procedures to 261000 SGTS X 3.2 45 correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5)

A2.11 High containment pressure Ability to monitor automatic operations of the A.C.

262001 AC Electrical X ELECTRICAL DISTRIBUTION including: (CFR: 41.7) 3.4 46 Distribution A3.04 Load sequencing Ability to manually operate and/or monitor in the 262001 AC Electrical control room: (CFR: 41.7)

X 3.3 47 Distribution A4.05 Voltage, current, power, and frequency on A.C. buses 2.1.27 Knowledge of system purpose and/or function.

262002 UPS (AC/DC) X 3.9 48 (CFR: 41.7)

Knowledge of the physical connections and/or cause/effect relationships between D.C.

263000 DC Electrical X ELECTRICAL DISTRIBUTION and the following: 3.2 49 Distribution (CFR: 41.2 to 41.9)

K1.02 Battery charger and battery Knowledge of the effect that a loss or malfunction of the EMERGENCY GENERATORS (DIESEL/JET) will 264000 EDGs X have on following: (CFR: 41.7 / 45.4) 4.1 50 K3.03 Major loads powered from electrical buses fed by the emergency generator(s)

Knowledge of EMERGENCY GENERATORS (DIESEL/JET) design feature(s) and/or interlocks 264000 EDGs X 3.8 51 which provide for the following: (CFR: 41.7)

K4.08 Automatic startup Knowledge of the operational implications of the following concepts as they apply to the 300000 Instrument Air X 2.5 52 INSTRUMENT AIR SYSTEM: (CFR: 41.5 / 45.3)

K5.01 Air compressors Rev 4

ES-401 7 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 1 (RO / 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

Knowledge of CCWS design feature(s) and or 400000 Component Cooling X interlocks which provide for the following: (CFR: 41.7) 3.4 53 Water K4.01 Automatic start of standby pump Ability to (a) predict the impacts of the following on the RHR/LPCI: INJECTION MODE (PLANT SPECIFIC) ; and (b) based on those predictions, use 203000 RHR/LPCI: Injection X procedures to correct, control, or mitigate the 3.6 86 Mode consequences of those abnormal conditions or operations:

A2.04 A.C. failures 2.2.25 Knowledge of the bases in Tech Specs for 212000 RPS X 4.2 87 LCOs and Safety limits (CFR: 41.5 / 41.7 / 43.2)

Ability to (a) predict the impacts of the following on the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM ; and (b) based on those predictions, use procedures to 215005 APRM / LPRM X 3.4 88 correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5)

A2.08 Faulty or erratic operation of detectors /

systems 2.4.20 Knowledge of the operational implications of 218000 ADS X EOP warnings, cautions, and notes. 4.3 89 (CFR: 41.10 / 43.5) 2.4.45 Ability to prioritize and interpret the 300000 Instrument Air X significance of each annunciator or alarm. 4.3 90 (CFR: 41.10 / 43.5) 2 2/ 26 K/A Category Point Totals: 2 2 3 4 3 2 2 2 2 / Group Point Total:

2 /5 3

Rev 4

ES-401 8 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 2 (RO / 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

Knowledge of the physical connections and/or cause-effect relationships between CONTROL ROD DRIVE 201001 CRD Hydraulic X HYDRAULIC SYSTEM and the following: 3.4 54 (CFR: 41.2 to 41.9)

K1.07 Reactor protection system Knowledge of the effect that a loss or malfunction of the REACTOR MANUAL CONTROL SYSTEM will 201002 RMCS X 2.9 59 have on following: (CFR: 41.7)

K3.03 Ability to process rod block signals Knowledge of the operational implications of the 201003 Control Rod and Drive following concepts as they apply to CONTROL ROD X 3.1 56 Mechanism AND DRIVE MECHANISM : (CFR: 41.5)

K5.04 Rod sequence patterns 201004 RSCS NOT APPLICABLE 201005 RCIS NOT APPLICABLE 201006 RWM NOT SELECTED Knowledge of electrical power supplies to the 202001 Recirculation X following: (CFR: 41.7) 3.2 57 K2.02 MG sets: Plant-Specific Knowledge of RECIRCULATION FLOW CONTROL SYSTEM design feature(s) and/or interlocks which 202002 Recirculation Flow Control X provide for the following: (CFR: 41.7) 2.9 58 K4.07 Minimum and maximum pump speed setpoints 204000 RWCU NOT SELECTED 214000 RPIS NOT SELECTED 215001 Traversing In-Core Probe NOT SELECTED 215002 RBM NOT SELECTED Knowledge of the effect that a loss or malfunction of the following will have on the NUCLEAR BOILER 216000 Nuclear Boiler Inst. X 3.1 55 INSTRUMENTATION : (CFR: 41.7)

K6.01 A.C. electrical distribution 219000 RHR/LPCI: Torus/Pool NOT SELECTED Cooling Mode 223001 Primary CTMT and Aux. NOT SELECTED 226001 RHR/LPCI: CTMT Spray NOT SELECTED Mode 230000 RHR/LPCI: Torus/Pool NOT SELECTED Spray Mode Ability to predict and/or monitor changes in parameters associated with operating the FUEL POOL 233000 Fuel Pool Cooling/Cleanup X COOLING AND CLEAN-UP controls including: (CFR: 2.5 60 41.5)

A1.06 System flow Ability to monitor automatic operations of the FUEL 234000 Fuel Handling Equipment X HANDLING EQUIPMENT including: (CFR: 41.7) 3.1 61 A3.02 Interlock operation 239001 Main and Reheat Steam NOT SELECTED 239003 MSIV Leakage Control NOT SELECTED Rev 4

ES-401 9 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 2 (RO / 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

Ability to (a) predict the impacts of the following on the REACTOR/TURBINE PRESSURE REGULATING SYSTEM ; and (b) based on those predictions, use 241000 Reactor/Turbine Pressure X procedures to correct, control, or mitigate the 3.7 62 Regulator consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6)

A2.07 Loss of condenser vacuum Ability to manually operate and/or monitor in the 245000 Main Turbine Gen. / Aux. X control room: (CFR: 41.7) 3.1 63 A4.02 Generator controls 256000 Reactor Condensate NOT SELECTED 259001 Reactor Feedwater NOT SELECTED 268000 Radwaste NOT SELECTED 271000 Offgas NOT SELECTED 272000 Radiation Monitoring NOT SELECTED 2.1.28 Knowledge of the purpose and function of major 286000 Fire Protection X 4.1 64 system components and controls. (CFR: 41.7) 288000 Plant Ventilation NOT SELECTED 290001 Secondary CTMT NOT SELECTED 290003 Control Room HVAC NOT SELECTED Knowledge of the physical connections and/or cause-effect relationships between REACTOR VESSEL 290002 Reactor Vessel Internals X INTERNALS and the following: 3.4 65 (CFR: 41.2 to 41.9)

K1.15 Nuclear boiler instrumentation 2.4.21 Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant 223001 Primary CTMT and Aux. X 4.6 91 system integrity, containment conditions, radioactivity release control, etc.

(CFR: 41.7 / 43.5)

Ability to (a) predict the impacts of the following on the MAIN AND REHEAT STEAM SYSTEM ; and (b) based on those predictions, use procedures to correct, 239001 Main and Reheat Steam X control, or mitigate the consequences of those 3.9 92 abnormal conditions or operations:

(CFR: 41.5 / 45.6)

A2.10 Closure of one or more MSIV's at power Ability to (a) predict the impacts of the following on the REACTOR FEEDWATER SYSTEM ; and (b) based on those predictions, use procedures to correct, 259001 Reactor Feedwater X control, or mitigate the consequences of those 3.8 93 abnormal conditions or operations: (CFR: 41.5 / 45.6)

A2.07 Reactor water level control system malfunctions 1 1 12 K/A Category Point Totals: 2 1 1 1 1 1 1 / 1 1 / Group Point Total:

/3 2 1 Rev 4

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Cooper Nuclear Station Date of Exam: March 2017 RO SRO-Only Category K/A # Topic IR # IR #

2.1.1 Knowledge of conduct of operations requirements. (CFR: 41.10) 3.8 66 Knowledge of procedures, guidelines, or limitations associated with 2.1.37 4.3 67 reactivity management. (CFR: 41.1)

Ability to identify and interpret diverse indications to validate the 2.1.45 4.3 68

1. response of another indication. (CFR: 41.7)

Conduct of Knowledge of individual licensed operator responsibilities related to Operations 2.1.4 shift staffing, such as medical requirements, no-solo operation, 3.8 94 maintenance of active license status, 10CFR55, etc. (CFR: 43.2)

Knowledge of procedures and limitations involved in core alterations.

2.1.36 4.1 95 (CFR: 43.6)

Subtotal 3 2 Knowledge of the process for making changes to procedures.

2.2.6 3.0 69 (CFR: 41.10)

Knowledge of the process for controlling equipment configuration or 2.2.14 3.9 70 status. (CFR: 41.10)

2. Knowledge of less than or equal to one hour Technical Specification 2.2.39 3.9 71 Equipment action statements for systems. (CFR: 41.7 / 41.10)

Control Knowledge of the process for managing troubleshooting activities.

2.2.20 3.8 96 (CFR: 43.5)

Knowledge of conditions and limitations in the facility license.

2.2.38 4.5 97 (CFR: 43.1)

Subtotal 3 2 Knowledge of radiation exposure limits under normal or emergency 2.3.4 3.2 72 conditions. (CFR: 41.12)

Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel

3. 2.3.12 3.2 73 handling responsibilities, access to locked high-radiation areas, Radiation aligning filters, etc. (CFR: 41.12)

Control Knowledge of radiation or contamination hazards that may arise 2.3.14 during normal, abnormal, or emergency conditions or activities. 3.8 98 (CFR: 43.4)

Subtotal 2 1 Knowledge of procedures relating to a security event (non-safeguards 2.4.28 3.2 74 information). (CFR: 41.10 / 43.5 / 45.13)

Knowledge of RO responsibilities in emergency plan implementation.

2.4.39 3.9 75 (CFR: 41.10)

4. Knowledge of EOP implementation hierarchy and coordination with Emergency other support procedures or guidelines such as, operating procedures, 2.4.16 4.4 99 Procedures / abnormal operating procedures, and severe accident management Plan guidelines. (CFR: 43.5)

Ability to take actions called for in the facility emergency plan, 2.4.38 including supporting or acting as emergency coordinator if required. 4.4 100 (CFR: 41.10 / 43.5 / 45.11)

Subtotal 2 2 Tier 3 Point Total 10 7 Rev 4

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Revision statement:

Rev 2 added question numbers and added importance rating for RO K/A 286000 2.1.32 and corrected importance ratings for SRO K/As 203000 A2.04, 215005 A2.08, 239001 A2.10, and 259001 A2.07. Also, replaced 1/1 K/A 295026 EK2.03 with EK2.06. Correct T1/G2 totals on category totals for RO K1 and K3 on ES-401-1 page 1.

Rev 3 corrected RO T1/G2 Tier Totals for K1 and K3 on ES-401-1 page 1.

Added header for form ES-401-3.

Swapped question 26 and 28 question number assignments to prevent having four answers the same in a row on the written exam.

Swapped question 55 and 59 question number assignments to prevent having four answers the same in a row on the written exam.

Rev 4 replaced K/A 286000 G2.1.32 for RO T2/G2 question 64 with K/A 286000 G2.1.28.

Rev 4

Rev 04 ES-401 Record of Rejected K/As Form ES-401-4 Randomly Tier / Group Selected K/A Reason for Rejection (Original)

(New)

RO T1/G1 Because CNS has a Mark I containment and NUREG 1123 states EPE 295027 295027 295027 High Containment Temperature is for Mark III containments only, Not Selected Not Applicable EPE 295027 was changed from NOT SELECTED to NOT APPLICABLE.

Page 1 point totals not affected by this change. (Rev 1)

SRO T1/G1 Because CNS has a Mark I containment and NUREG 1123 states EPE 295027 295038 295027 High Containment Temperature is for Mark III containments only, G2.4.18 G2.4.18 EPE 295027 was replaced with randomly selected EPE 295038 High Off-site Release Rate. KA G2.4.18 was not changed. Page 1 point totals not affected by this change. (Rev 1)

RO T2/G1 Because CNS does have a Low Pressure Core Spray system, System 239002 209001 209001 LPCS was changed from NOT APPLICABLE. Since System K4.04 K4.04 239002 SRVs was one of the systems sampled twice, 239002 SRVs K4.04 was replaced with 209001 LPCS K4.04 so that LPCS is sampled at least once. Page 1 point totals not affected by this change. (Rev 1)

RO T2/G2 Because the Rod Sequence Control System is no longer used at CNS, 201004 201004 System 201004 RSCS was changed from NOT SELECTED to NOT Not Selected Not Applicable APPLICABLE. Page 1 point totals not affected by this change. (Rev. 1)

RO T2/G2 Because CNS does not have a Rod Control and Information System, 201005 201005 System 201005 RCIS was changed from NOT SELECTED to NOT Not Selected Not Applicable APPLICABLE. Page 1 point totals not affected by this change. (Rev 1)

RO T1/G1 Because a discriminatory, operationally valid RO question could not be 295026 295026 developed, replaced 295026 EK2.03 with randomly selected EK2.06.

EK2.03 EK2.06 Page 1 point totals not affected by this change. (Rev 2)

RO T2/G1 The only SLC flow indicator at CNS is a local float type meter on the SLC 211000 211000 Test Tank inlet piping. No flow indication is available for SLC injection to A1.06 A1.08 the RPV. A discriminatory, operationally valid question could not be developed. Replaced A1.06 with randomly selected A1.08 under the same K/A category. Page 1 point totals not affected by this change.

(Rev 2)

RO T2/G2 Single/Sequential turbine governor valve operation is no longer used at 245000 245000 CNS following high pressure turbine replacement during RE29. Because A4.07 A4.02 of this and the design of the DEH system, a discriminatory question could not be developed. Replaced A4.07 with randomly selected A4.02 under the same K/A category. Page 1 point totals not affected by this change.

(Rev 2)

RO T2/G2 Because a discriminatory, operationally valid RO question could not be 286000 286000 developed for fire protection system limits and cautions, replaced RO G2.1.32 G2.1.28 T2/G2 question 64 K/A 286000 G2.1.32 with randomly selected K/A 286000 G2.1.28. Page 1 point totals not affected by this change.(Rev 4)

Revision statement: (Rev 04)

Replaced RO T2/G2 question 64 K/A 286000 G2.1.32 with K/A 286000 G2.1.28.

Rev 04

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Cooper Nuclear Station Date of Examination: 3/06/2017 Examination Level: RO SRO Operating Test Number: 1 Administrative Topic Type Describe activity to be performed (see Note) Code*

A1, Perform Jet Pump Operability Check (RO)

Conduct of Operations R, D 2.1.25 (3.9/4.2)

A2, Perform SLC Operability Checks Conduct of Operations R, N 2.1.20 (4.6/4.6)

A3, Determine Isolation Boundaries (RHR)

Equipment Control R, D 2.2.13 (4.1/4.3)

A4, Determine Workers Projected Total Dose Radiation Control R, N 2.3.14 (3.4/3.8)

Emergency Plan NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, then all 5 are required.

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

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

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

ES-301, Page 22 of 27 Rev. 0

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Cooper Nuclear Station Date of Examination: 3/06/2017 Examination Level: RO SRO Operating Test Number: 1 Administrative Topic Type Describe activity to be performed (see Note) Code*

A5, Determine if Mode Change is Allowed Conduct of Operations R, D 2.1.20 (4.6/4.6), 2.2.35 (3.6/4.5), 2.2.40 (3.4/4.7)

A6, Reportable Occurrences to the NRC (#8)

Conduct of Operations R, N 2.1.18 (3.6/3.8), 2.1.20 (4.6/4.6), 2.4.30 (2.7/4.1)

A7, Review Jet Pump Operability and Recirc Pump Flow Checks Equipment Control R, M 2.2.12 (3.7/4.1), 2.2.42 (3.9/4.6)

A8, Authorize Stable Iodine Thyroid Blocking Radiation Control R, D 2.3.14 (3.4/3.8)

A9, Emergency Classification Emergency Plan R, D 2.4.41 (2.9/4.6)

NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, then all 5 are required.

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

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

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

ES-301, Page 22 of 27 Rev. 0

Rev 1 ES-301 Control Room/In Plant Systems Outline Form ES-301-2 Facility: Cooper Nuclear Station Date of Examination: 03/06/2017 Exam Level: RO SRO-I SRO-U Operating Test No.: 1 Control Room Systems; * (8 for RO); (7 for SRO-I); 2 or 3 for SRO-U Safety System / JPM Title Type Code*

Function S1 - Secure SDG from Control Room L, N, S 6 264000 A4.04 (3.7/3.7)

S2 - Start Torus Cooling from ASD Room L, D, S 5 219000 A1.02 (3.5/3.5)

S3 - Conduct Alt Pressure Control Using Reactor Feed Pumps L, D, S 3 259001 A4.02 (3.9/3.7)

S4 - Level Recovery During Shutdown Conditions Using LPCI A, EN, L, N, S 2 203000 A4.05 (4.3/4.1)

S5 - Perform 6.TG.303 Testing OPC Overspeed L, N, S 4 241000 A3.12 (2.9/2.9), 241000 A4.19 (3.5/3.4)

S6 - Align REC System IAW 5.3EMPWR L, N, S 8 400000 A4.01 (3.1/3.0)

S7 - Withdraw Control Rod From Position 00 A, L, N, S 1 201003 A2.01 (3.4/3.6)

S8 - Verify Group 2 Isolation (TIP Shear)

A, D, S 7 223002 A4.01 (3.6/3.5), 223002 A4.06 (3.6/3.7)

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

P1 - Locally Secure Fire Pump C A, N 8 286000 A4.05 (3.3/3.3), 286000 A1.05 (3.2/3.2)

P2 - Locally Align RHRSW Crosstie for RPV Injection D, E, L 4 2.1.29 (4.1/4.0); 295031 EA1.08 (3.8/3.9)

P3 - (ASD) Locally Operate SW-MO-89B for Torus Cooling D, E, L, R 5 219000 A1.08 (3.7 / 3.6), 295016 AK2.02 (4.0/4.1)

All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all five SRO-U systems must serve different safety functions; 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 A 4-6 / 4-6 / 2-3 (4)

(C)ontrol room C -----

(D)irect from bank D <9 / <8 / <4 (5)

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

(EN)gineered safety feature EN >1 / >1 / > 1 (control room sys) (1)

(L)ow-Power / Shutdown L >1 / >1 / >1 (9)

(N)ew or (M)odified from bank including 1(A) N-M >2 / >2 / >1 (6)

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

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

(S)imulator S -----

Rev 1

Rev 1 ES-301 Control Room/In Plant Systems Outline Form ES-301-2 Facility: Cooper Nuclear Station Date of Examination: 03/06/2017 Exam Level: RO SRO-I SRO-U Operating Test No.: 1 Control Room Systems; * (8 for RO); (7 for SRO-I); 2 or 3 for SRO-U Safety System / JPM Title Type Code*

Function S1 - Secure SDG from Control Room L, N, S 6 264000 A4.04 (3.7/3.7)

S2 - Start Torus Cooling from ASD Room L, D, S 5 219000 A1.02 (3.5/3.5)

S4 - Level Recovery During Shutdown Conditions Using LPCI A, EN, L, N, S 2 203000 A4.05 (4.3/4.1)

S5 - Perform 6.TG.303 Testing OPC Overspeed L, N, S 4 241000 A3.12 (2.9/2.9), 241000 A4.19 (3.5/3.4)

S6 - Align REC System IAW 5.3EMPWR L, N, S 8 400000 A4.01 (3.1/3.0)

S7 - Withdraw Control Rod From Position 00 A, L, N, S 1 201003 A2.01 (3.4/3.6)

S8 - Verify Group 2 Isolation (TIP Shear)

A, D, S 7 223002 A4.01 (3.6/3.5), 223002 A4.06 (3.6/3.7)

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

P1 - Locally Secure Fire Pump C A, N 8 286000 A4.05 (3.3/3.3), 286000 A1.05 (3.2/3.2)

P2 - Locally Align RHRSW Crosstie for RPV Injection D, E, L 4 2.1.29 (4.1/4.0); 295031 EA1.08 (3.8/3.9)

P3 - (ASD) Locally Operate SW-MO-89B for Torus Cooling D, E, L, R 5 219000 A1.08 (3.7 / 3.6), 295016 AK2.02 (4.0/4.1)

All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all five SRO-U systems must serve different safety functions; 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 A 4-6 / 4-6 / 2-3 (4)

(C)ontrol room C -----

(D)irect from bank D <9 / <8 / <4 (4)

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

(EN)gineered safety feature EN >1 / >1 / > 1 (control room sys) (1)

(L)ow-Power / Shutdown L >1 / >1 / >1 (8)

(N)ew or (M)odified from bank including 1(A) N-M >2 / >2 / >1 (6)

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

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

(S)imulator S -----

Rev 1

Rev 3 ES-301 Control Room/In Plant Systems Outline Form ES-301-2 Facility: Cooper Nuclear Station Date of Examination: 03/06/2017 Exam Level: RO SRO-I SRO-U Operating Test No.: 1 Control Room Systems; * (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U)

Safety System / JPM Title Type Code*

Function S1 - Secure SDG from Control Room L, N, S 6 264000 A4.04 (3.7/3.7)

S4 - Level Recovery During Shutdown Conditions Using LPCI A, EN, L, N, S 2 203000 A4.05 (4.3/4.1)

S7 - Withdraw Control Rod From Position 00 A, L, N, S 1 201003 A2.01 (3.4/3.6)

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

P1 - Locally Secure Fire Pump C A, N 8 286000 A4.05 (3.3/3.3), 286000 A1.05 (3.2/3.2)

P3 - (ASD) Locally Operate SW-MO-89B for Torus Cooling D, E, L, R 5 219000 A1.08 (3.7 / 3.6), 295016 AK2.02 (4.0/4.1)

All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all five SRO-U systems must serve different safety functions; 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 A 4-6 / 4-6 / 2-3 (3)

(C)ontrol room C -----

(D)irect from bank D <9 / <8 / <4 (1)

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

(EN)gineered safety feature EN >1 / >1 / > 1 (control room sys) (1)

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

(N)ew or (M)odified from bank including 1(A) N-M >2 / >2 / >1 (4)

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

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

(S)imulator S -----

Rev 3

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 1 Page 1 of 44 Facility: Cooper Nuclear Station Scenario No.: 1 Op-Test No.: CNS 15-01 Examiners: ____________________________ Operators: _____________________________

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

1. Shift CRD Stabilizing valves.

2. Lower reactor power using RR pumps.

3. Respond to Reactor Bldg to Torus Vacuum Breaker PC-AO-243 failing open.

4. Respond to RPV flange leakage.

5. Respond to trip of RPS A EPAs with failure of RMV-AO-10 to close.

6. Respond to loss of multiple REC pumps.

7. ATWS Level Power control

8. Respond to RHR SPC valve failing to open.

Initial Conditions: Plant operating at 100% power.

Inoperable Equipment: HPCI inoperable. Auxiliary Oil pump motor replacement.

TS LCO 3.5.1, Condition C Turnover:

The plant is at 100% power.

Planned activities for this shift are:

Shift CRD Stabilizing valves per Procedure 2.2.8 (Rev. 95)

Lower power to 95% with RR Pumps per Procedure 2.1.10 (Rev. 113)

Electrical Maintenance working on replacing HPCI AOP motor Scenario Notes:

This is a new scenario.

Validation Time: 75 minutes Rev. 2

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 1 Page 2 of 44 Event Malf. No. Event Type Event No. Description 1 N/A N (ATC,CRS) Shift CRD stabilizing valves 2 N/A R (ATC, CRS) Lower Reactor power by lowering RR pump speed.

(or) Reactor Building to Torus vacuum breaker fails open.

3 TS (CRS) zdipcswcs243av[2] CRS declares vacuum breaker inoperable.

C (BOP,CRS) Respond to reactor vessel flange seal leak alarm, 4 rr21 enter Procedure 4.6.3, and cycle the flange leak-off A (CREW) drain valves.

I rp03a (BOP,ATC,CRS) RPS EPA Breaker 1A1/1A2 trip, (half scram and half 5 PCIS group isolations) RMV-AO-10 fails to isolate.

(rf) rh32a A (CREW) CRS declares valve inoperable.

TS (CRS) sw 11a C (BOP, ATC) REC Pump A trip. Start another REC pump. REC 6

sw11b A(CREW) Pump B trip. Manual scram due to loss of REC.

Hydraulic block ATWS > 3% power (EOP-1A, 3A, 6A, 6B, 7A)

(CT-1) 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 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.)

(CT-2) Inhibit ADS prior to automatic ADS valve opening during a failure to Scram.

7 rd02a,b (CT-3) During failure to scram conditions with M (CREW) power >3%, stop and prevent injection from all sources (except boron, CRD, RCIC) as necessary to lower RPV water level to below -

60 CFZ (or LL, as applicable) and control between -60 (or LL, as applicable) to -183 CFZ prior to exiting EOP-7A.

(CT-4) When control rods fail to scram and energy is discharging to the primary containment, crew injects SLC before exceeding the Boron Injection Initiation Temperature (BIIT) curve.

(rf) rh29(A) First RHR loop to be put into suppression pool 8 C (BOP,CRS)

(rf) rh30(A) cooling has RHR-MO-39A(B) fail to open.

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

Critical Task (As defined in NUREG 1021 Appendix D)

CREW notation for Abnormal (A) and Major (M) events denotes ATC, BOP, and CRS are credited.

Rev. 2

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 1 Page 3 of 44 Quantitative Attributes Table ES-304-1 Attribute Actual Description Target Malfunctions after EOP entry 1-2 1 RHR-MO-39A(B) fails to open.

RPV flange seal leak Abnormal Events 2-4 3 RPS EPA trip Loss of multiple REC Pumps Major Transients 1-2 1 ATWS EOP-3A EOP entries requiring substantive action 1-2 2 EOP-6A EOP contingencies requiring substantive 0-2 1 EOP- 7A action (CT-1) 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 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.)

(CT-2) Inhibit ADS prior to automatic ADS valve opening during a failure to Scram.

(CT-3) During failure to scram conditions with EOP based Critical power >3%, stop and prevent injection from all 2-3 4 Tasks sources (except boron, CRD, RCIC) as necessary to lower RPV water level to below -

60 CFZ (or LL, as applicable) and control between -60 (or LL, as applicable) to -183 CFZ prior to exiting EOP-7A.

(CT-4) When control rods fail to scram and energy is discharging to the primary containment crew injects SLC before exceeding the Boron Injection Initiation Temperature (BIIT) curve.

Normal Events N/A 1 Shift CRD Stabilizing valves.

Reactivity Manipulations N/A 1 Lower power using Reactor Recirculation pumps RPV Flange leak Reactor Bldg to Torus vacuum breaker fails open Instrument/ RPS A EPA Breaker trip N/A 6 Component Failures Loss of REC pump A Loss of REC pump B RHR-MO39A(B) valve fails to open RPV Flange leak Reactor Bldg to Torus vacuum breaker fails open RPS a EPA Breaker trip Total Malfunctions N/A 6 Loss of REC pump A Loss of REC pump B RHR-MO39A(B) valve fails to open Top 10 systems and operator actions important to risk that are tested:

Rev. 2

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 1 Page 4 of 44 Reactor Protection System (Event 5)

Residual Heat Removal System in Suppression Pool Cooling Mode (Event 8)

SCENARIO

SUMMARY

The plant is operating at 100% power. HPCI Auxiliary Oil Pump motor replacement is taking place.

Event 1 After the crew takes the watch, the ATC shifts the CRD Stabilizing valves per Procedure 2.2.8. (Event 1)

Event 2 After shifting stabilizing valves, the ATC lowers power ~5% per Load Dispatcher schedule.

NOTE: Events 3 and 4 are triggered simultaneously due to Event 4 taking ~ 10 minutes to manifest itself.

Event 3 (Triggered by Lead Examiner)

After lowering power the Reactor to Torus vacuum breaker PC-AO-243 fails open. The CRS enters LCO 3.6.1.7, Condition A and declares the vacuum breaker inoperable.

Event 4 (Triggered by Lead Examiner)

The RPV inner seal develops a small leak requiring the BOP cycle the leak-off isolation valves from the control room to clear the alarm.

Event 5 (Triggered by Lead Examiner)

After actions for RPV flange leakage are complete, RPS EPAs on Division I trip causing a half reactor scram and half Group 1, 2, 7, and full Group 3, 6 isolations.

RMV-AO-10 fails to isolate on the loss of RPS. The CRS enters LCO 3.6.1.3, Condition A for RMV-AO-10 failing to isolate and determines a potential LCO for TS 3.3.8.2 Condition A is required for the EPA breaker. LCO 3.3.8.2 entry is not required, since the EPA is no longer supplying RPS.

Event 6 (Triggered by Lead Examiner)

After RPS A power has been restored from the alternate supply and RRMG cooling restored, and the half scram is reset, REC Pump A trips requiring the BOP to start the standby pump per alarm procedures. Shortly after the standby pump is started, REC Pump B trips requiring entry into Emergency Procedure 5.2REC. The ATC will insert a reactor scram. The CRS will not have time to enter Technical Specifications for the REC pumps.

Rev. 2

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 1 Page 5 of 44 Event 7 (No Trigger required)

When the reactor is scrammed, a low power ATWS occurs due to hydraulic block of both scram discharge volumes, and EOP-6A and 7A are entered via EOP-1A.

Reactor power is above 3%. The crew injects SLC and/or installs the necessary PTMs to bypass interlocks and insert control rods individually via RMCS (CT-1, CT-4).

ADS is manually inhibited to prevent automatic operation (CT-2). Stop and Prevent is required because reactor power is above 3%. RPV level is intentionally lowered below -60 inches wide range in order to lower core inlet subcooling and lower reactor power (CT-3). Only 1 Main Turbine Bypass valve is available to control RPV pressure. SRVs have to be used to supplement pressure control. Feedwater injection is available for RPV level control.

Event 8 (Automatically Triggered when opening the first MO39A(B) is attempted)

After the crew has stabilized conditions following the scram, the selected RHR suppression pool cooling loop cannot be placed into service because RHR-MO39A (B) fails to open. The BOP transfers to the other division of RHR and places it into suppression pool cooling.

After the scram has been reset twice, the control rods are allowed to be fully inserted with the next scram. The CRD transitions from ATWS to non-ATWS flowcharts, SLC injection is halted and RPV level restoration is directed.

The exercise ends when control rods are inserted, and RPV water level is being maintained between -183 inches and +54 inches.

Rev. 2

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 1 Page 6 of 44 Critical Tasks (CT-1) When control rods fail to scram, (CT-2) Inhibit ADS prior to automatic ADS crew injects SLC and/or inserts control valve opening during a failure to Scram.

rods 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 7 7 Safety Failure to effect shutdown of the reactor With a Reactor Scram required, reactor not significance when a RPS setting has been exceeded shut down, and conditions for ADS blowdown would unnecessarily extend the level of are met, INHIBIT ADS to prevent an degradation of the safety of the plant. This uncontrolled RPV depressurization and cold could further degrade into damage to the water injection from low pressure sources to principle fission product barriers if left prevent causing a significant power excursion.

unmitigated. The crew is authorized and required by Conduct of Operations to take mitigating actions when automatic safety systems fail to perform their intended function. Action to shut down the reactor is required when RPS and control rod drive systems fail.

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 Rev. 2

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 1 Page 7 of 44 Critical Tasks (CT-3) During failure to scram conditions (CT-4) When control rods fail to scram and with power >3%, stop and prevent energy is discharging to the primary injection from all sources (except boron, containment, crew injects SLC before CRD, RCIC) as necessary to lower RPV exceeding the Boron Injection Initiation water level to below -60 CFZ (or LL, as Temperature (BIIT) curve.

applicable) and control between -60 (or LL, as applicable) to -183 CFZ prior to exiting EOP-7A.

EVENT 7 7 Safety Regarding lowering level below -60 CFZ, to Failure to effect shutdown of the reactor when significance prevent or mitigate the consequences of any a RPS setting has been exceeded would large irregular neutron flux oscillations unnecessarily extend the level of degradation induced by neutronic/thermal-hydraulic of the safety of the plant. This could further instabilities, RPV water level is lowered degrade into damage to the principle fission sufficiently below the elevation of the product barriers if left unmitigated. Action to feedwater sparger nozzles. This places the shut down the reactor is required when RPS feedwater spargers in the steam space and control rod drive systems fail.

providing effective heating of the relatively cold feedwater and eliminating the potential The Boron Injection Initiation Temperature for high core inlet subcooling. For conditions (BIIT) is the greater of:

that are susceptible to oscillations, the initiation and growth of oscillations is

The highest suppression pool temperature at principally dependent upon the subcooling at which initiation of boron injection will permit the core inlet; the greater the subcooling, the injection of the Hot Shutdown Boron Weight of more likely oscillations will commence and boron before suppression pool temperature increase in magnitude. exceeds the Heat Capacity Temperature Limit.

24" below the lowest nozzle in the feedwater

The suppression pool temperature at which a sparger has been selected as the upper reactor scram is required by plant Technical bound of the RPV water level control band. Specifications.

This water level is sufficiently low that steam heating of the injected water will be at least The BIIT is a function of reactor power. If 65% to 75% effective (i.e., the temperature of boron injection is initiated before suppression the injected water will be increased to 65% to pool temperature reaches the BIIT, emergency 75% of its equilibrium value in the steam RPV depressurization may be precluded at environment). This water level is sufficiently lower reactor power levels. At higher reactor high that the capability to bypass the low power levels, however, the suppression pool RPV water level MSIV isolation should be heatup rate may become so high that the Hot able to control RPV water level with Shutdown Boron Weight of boron cannot be feedwater pumps to preclude the isolation. injected before suppression pool temperature reaches the Heat Capacity Temperature Limit Regarding lowering level below LL, the even if boron injection is initiated early in the combination of high reactor power (above the event. Since failure-to-scram conditions may APRM downscale trip), high suppression pool present severe plant safety consequences, the temperature (above the Boron Injection requirement to initiate boron injection is Initiation Temperature), and an open SRV or independent of any anticipated success of high drywell pressure (above the scram control rod insertion. When attempts to insert setpoint) are symptomatic of heat being control rods satisfactorily achieve reactor rejected to the suppression pool at a rate in shutdown, the requirement for boron injection excess of that which can be removed by the no longer exists. (Control rod insertion is Suppression Pool Cooling System. Unless directed under Step RC/Q-7 concurrently with mitigated, these conditions ultimately result in Step RC/Q-6.)

loss of NPSH for ECCS pumps taking suction on the suppression pool, containment over-pressurization, and (ultimately) loss of primary containment integrity, which in turn could lead to a loss of adequate core cooling and uncontrolled release of radioactivity to the environment. The conditions listed, combined with the inability to shut down the Rev. 2

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 1 Page 8 of 44 reactor through control rod insertion, dictate a requirement to promptly reduce reactor power since, as long as these conditions exist, suppression pool heatup will continue.

If torus water temperature was allowed to exceed the HCTL prior to commencing the lowering of level, a RPV depressurization would be required. Failure to completely stop RPV injection flow (with the exception of CRD and SLC) prolongs the elevated reactor power condition; thus, depositing more energy than necessary into the suppression pool.

Cueing Manual scram is initiated and numerous Manual scram is initiated and numerous control rods indicate beyond position 00 and control rods indicate beyond position 00 and reactor power >3% on panel 9-5 indications reactor power not downscale on panel 9-5 and SPDS and RPV level is >-60CFZ on indications.

SPDS.

Suppression Pool temperature rising on panel 9-3 indication.

Performance Operator manipulates Feedwater HMIs on Operator manipulates keylocked switches for indicator panel 9-5 or panel A as necessary to stop SLC A(B) pump to START on panel 9-5.

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.

Performance Feedwater flow indication on panel 9-5 SLC A(B) pump red light illuminated, SLC feedback indicate zero. discharge pressure rising, SLC tank level lowering on panel 9-5.

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 If boron injection is initiated before suppression pool for the chosen conditions where it is necessary to lower level to temperature reaches the BIIT, emergency RPV performance control power with Table 17 condition NOT met depressurization may be precluded at lower reactor (i.e. no high energy input into primary power levels. At higher reactor power levels, limit containment). There is no time limit for this however, the suppression pool heatup rate may lowering level, but it establishes margin to become so high that the Hot Shutdown Boron conditions where fuel damaging power oscillations Weight of boron cannot be injected before may theoretically occur. Before exiting EOP-7A suppression pool temperature reaches the Heat was chosen because Capacity Temperature Limit even if boron injection is other procedures would not provide the guidance initiated early in the event. Since failure-to-scram necessary to establish margin for power oscillation conditions may present severe plant safety mitigation. Before exiting EOP-7A ensures consequences, the requirement to initiate boron guidance to effect this control is not removed. injection is independent of any anticipated success of control rod insertion.

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.

BWR Owners App. B, Contingency #5 App. B, step RC/Q-6 Group Appendix Rev. 2

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 2 Page 1 of 37 Facility: Cooper Nuclear Station Scenario No.: 2 Op-Test No.: CNS 15-01 Examiners: ____________________________ Operators: _____________________________

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

1. Shift REC Pumps.

2. CRD FCV auto function fails requiring manual control.

3. One outboard MSIV fails closed.

4. Partial loss of main condenser vacuum requiring manual scram.

5. Electric ATWS

6. FW line break inside PC with loss of RCIC.

7. RWCU fails to auto isolate

8. Emergency Depressurize on low RPV level.

9. Low pressure injection valves fail to automatically open.

Initial Conditions: Plant operating at 100% power.

Inoperable Equipment: HPCI inoperable. Auxiliary Oil pump motor replacement.

TS LCO 3.5.1, Condition C Turnover:

The plant is at 100% power.

Planned activities for this shift are:

Maintain present power level.

Electrical Maintenance working on installing HPCI AOP motor.

Scenario Notes:

This is a new scenario.

Validation Time: 75 minutes Rev. 2

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 2 Page 2 of 37 Event Malf. No. Event Type Event No. Description N (BOP, CRS) 1 N/A Shift REC pumps TS (CRS)

OR: I (ATC,CRS) Auto function on CRD FCV fails causing manual control 2

zaicrdfc301 A (CREW) to be used C

ms09e (ATC,BOP,CRS) 3 OR: Outboard MSIV 86A closes but leaks by A (CREW) zdipcissws4a TS (CRS)

C (ATC,CRS) 4 mc01 Partial loss of condenser vacuum-scram A (CREW) rp01 (a-d)

OR: Electrical ATWS 5 zdirpssws1 M(CREW) (CT-1 When RPS fails to scram the reactor on a manual zdirpssws3a scram signal, within two minutes initiate the ARI System.

zdirpssws3b FW A line break inside primary containment.

RCIC spurious isolation fw18a (CT-2) When RPV level lowers to -158 CFZ (TAF) and cannot be maintained above -183 CFZ (MSCWL) and 6 rr20a M (CREW) insufficient high pressure injection systems are available to restore level, crew begins to Emergency Depressurize rc02 by opening the first of six SRVs before RPV level lowers below -183 CFZ.) (Momentary shrink below -183 due to automatic SRV closure does not constitute failure of this critical task).

7 rp12 C (BOP,CRS) RWCU fails to automatically isolate.

ECCS system valves fail to auto open.

cs02a C (CT-3 )When operating injection systems cannot maintain cs02b RPV level and ECCS systems fail to automatically align for 8 (ATC,BOP,CRS) rh04a injection, crew manually aligns ECCS systems for injection:

rh04b

For low pressure ECCS systems, prior to RPV pressure lowering below 200 psig.

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

Critical Task (As defined in NUREG 1021 Appendix D)

CREW notation for Abnormal (A) and Major (M) events denotes ATC, BOP, and CRS are credited.

Rev. 2

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 2 Page 3 of 37 Quantitative Attributes Table ES-304-1 Attribute Actual Description Target Malfunctions after Low Pressure ECCS injection valves fail to open.

1-2 2 EOP entry RWCU fails to auto isolate Outboard MSIV closure Abnormal Events 2-4 2 Partial loss of condenser vacuum ATWS Major Transients 1-2 2 FW line break EOP entries requiring EOP-1A 1-2 2 substantive action EOP-3A EOP contingencies requiring substantive 0-2 1 EOP-2A action (CT-1) When RPS fails to scram the reactor on a manual scram signal, within two minutes initiate the ARI System.

(CT-2) When operating injection systems cannot maintain RPV level and ECCS systems fail to automatically align for injection, crew manually aligns ECCS systems for injection: For low pressure ECCS systems, prior to RPV pressure EOP based Critical 2-3 lowering below 200 psig.

3 Tasks (CT-3) When RPV level lowers to -158 CFZ (TAF) and cannot be maintained above -183 CFZ (MSCWL) and insufficient high pressure injection systems are available to restore level, crew begins to Emergency Depressurize by opening the first of six SRVs before RPV level lowers below -183 CFZ.) (Momentary shrink below -183 due to automatic SRV closure does not constitute failure of this critical task).

Normal Events N/A 1 Shift REC Pumps Reactivity Manipulations N/A 1 1 CRD FCV controller failure Outboard MSIV closure Instrument/

Component Failures N/A 5 Condenser in-leakage loss of vacuum RWCU fails to isolate.

LP ECCS injection valves fail to open CRD FCV controller failure Outboard MSIV closure Total Malfunctions N/A 5 Condenser in-leakage rise RWCU fails to isolate.

LP ECCS injection valves fail to open Top 10 systems and operator actions important to risk that are tested:

Reactor Protection System (Event 5)

ADS/SRV (Event 6)

Residual Heat Removal System in LPCI Mode (Event 8)

Operator fails to depressurize with SRVs (Event 6)

Operator fails to initiate ADS and initiate ECCS early (Event 6)

Rev. 2

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 2 Page 4 of 37 SCENARIO

SUMMARY

The plant is operating at 100% power at the end of the operating cycle. HPCI Auxiliary Oil Pump motor is removed and a replacement is being installed.

Event 1 After the crew takes the watch, the BOP operator shifts REC pumps by starting B and securing A. The CRS is required to declare REC Div I subsystem inoperable per LCO 3.7.3, Condition B, Event 2 (Triggered by Lead Examiner)

After TS are addressed for the REC pump shift, the CRD FCV automatic setpoint fails downscale requiring the ATC to take manual control and return CRD cooling water flow and pressure to normal.

Event 3 (Triggered by Lead Examiner)

After the CRD system flows are returned to normal in manual, outboard MSIV 86A partially closes. The crew enters Abnormal Procedure 2.4MSIV and the RO rapidly lowers reactor power to <70%. The BOP places the effected MSIV control switch to CLOSE to prevent reopening. The CRS enters LCO 3.6.1.3, Condition A and declares the PCIV inoperable.

Event 4 (Triggered by Lead Examiner)

After TS are addressed for the partially closed MSIV, condenser in-leakage rises requiring reactor power to be lowered to maintain vacuum > 23 inches mercury.

Condenser vacuum continues to lower requiring the reactor scram.

Event 5 (No Trigger required)

On the manual reactor scram, the crew recognizes the ATWS is an electric block ATWS. Manual ARI initiation successfully inserts the control rods (CT-1).

Event 6 (Automatically triggered when RFP Discharge Valve automatically closes,

~2 minutes after ARI is initiated)

After the control rods are inserted, Feedwater A line break inside the PC commences and the CRS enters EOP 3A. The torus and drywell are sprayed to control containment pressure and temperature. RPV water level continues to drop. RCIC will be unavailable due to a spurious isolation signal.

Event 7 (No Trigger required)

RWCU fails to isolate on low RPV level. Manual isolation from the control room is required.

Rev. 2

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 2 Page 5 of 37 Event 8 (No Trigger required)

RPV level lowers to TAF requiring the crew to emergency depressurize (CT-2). As RPV level and pressure lower, RHR injection valves fail to open and cannot be opened. The CS injection valves fail to open and can be opened from the control room (CT-3).

The exercise ends when emergency depressurization is complete and RPV level restoration is being controlled.

Rev. 2

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 2 Page 6 of 37 Critical Tasks (CT-1) When RPS fails to scram the (CT-2) When RPV level lowers to -158 CFZ reactor on a manual scram signal, within (TAF) and cannot be maintained above -

two minutes initiate the ARI System. 183 CFZ (MSCWL) and insufficient high pressure injection systems are available to restore level, crew begins to Emergency Depressurize by opening the first of six SRVs before RPV level lowers below -183 CFZ. (Momentary shrink below -183 due to automatic SRV closure does not constitute failure of this critical task).

EVENT 5 6 Safety RPS initiates a reactor scram when one or The MSCWL is the lowest RPV water level at significance more monitored parameters exceed their which the covered portion of the reactor core specified limits to preserve the integrity of the will generate sufficient steam to preclude any fuel cladding and the reactor coolant clad temperature in the uncovered portion of pressure boundary (RCPB) and minimize the the core from exceeding 1500F. When water energy that must be absorbed following a level decreases below MSCWL with injection, loss of coolant accident (LOCA). Failure to clad temperatures may exceed 1500F.

effect shutdown of the reactor when a RPS setting has been exceeded, even at low power, would unnecessarily extend the level of degradation of the safety of the plant. This could further degrade into damage to the principle fission product barriers if left unmitigated. The crew is authorized and required by Conduct of Operations to take mitigating actions when automatic safety systems fail to perform their intended function. Action to shut down the reactor is required when RPS and control rod drive systems fail.

Cueing Annunciators 9-5-2/A-1 (A-2) RX SCRAM Corrected Fuel Zone indication (SPDS) falls to CHANNEL A (B) in alarm with RPS -158 and lowering trend continues, and, remaining energized. before -158 CFZ is reached, initial conditions, field reports, and control room indications convey that adequate high pressure injection cannot be restored before level falls below -

183 CFZ Performance Operator depresses both manual scram Manipulation of any six SRV controls on panel indicator pushbuttons, or places the Reactor Mode 9-3:

Switch to SHUTDOWN on panel 9-5. SRV-71A SRV-71B SRV-71E SRV-71G SRV-71H SRV-71C SRV-71D SRV-71F Performance RPS Group lights de-energized on panel 9-5. Crew will observe SRV light indication go from feedback Control Rod full -in indication on panel 9-5. green to red, amber pressure switch lights Reactor power trend on nuclear illuminate, reactor pressure lowering on SPDS instrumentation on panel 9-5. and panel 9-3 and 9-5 meters and recorders, and SRV tailpipe temperatures rise on recorder MS-TR-166.

Justification Procedure 2.0.3, Conduct of Operations The MSCWL (-183 CFZ) is the lowest RPV for the chosen requires upon recognition of a failure of water level at which the covered portion of the performance automatic action, the CRO shall manually reactor core will generate sufficient steam to limit perform those actions necessary to fulfill the preclude any clad temperature in the Rev. 2

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 2 Page 7 of 37 safety function and report the completion of uncovered portion of the core from exceeding the manual action to the CRS as soon as 1500F. Emergency depressurization is possible. Failure of RPS to automatically allowed when level goes below TAF (-158 function would involve multiple sensor and CFZ) and should be performed, if in the sensor relay failures. The complexity of an judgment of the CRS, level cannot be automatic RPS failure would necessarily maintained above -183 CFZ. Since it is require a short amount of time to diagnose intended for the scenario supporting this CT to, and validate using control room indications. early in the event, clearly indicate no high Two minutes is a reasonable time for pressure injection systems can be made operators to recognize a scram signal, verify available to reverse the lowering level trend, the condition is valid, communicate the crew will have time to communicate and conditions to the crew, and insert a manual open 6 SRVs before -183 CFZ.

scram, without unnecessarily extending the level of degradation to plant safety.

BWR Owners App. B, step RC-1 App. B, Contingency#1 Group Appendix Rev. 2

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 2 Page 8 of 37 Critical Tasks (CT-3) When operating injection systems cannot maintain RPV level and ECCS systems fail to automatically align for injection, crew manually aligns ECCS systems for injection:

For low pressure ECCS systems, prior to RPV pressure lowering below 200 psig.

EVENT 8 Safety Failure to recognize the auto valve alignment significance not occurring, and failure to take manual action per Conduct of Ops will result in unavailability of safety-related equipment necessary to provide adequate core cooling, otherwise resulting in core damage and a large offsite release.

Cueing Indication ECCS valves are not opening with initiation conditions present:

Green light on and Red lamp extinguished at respective injection handswitch on panel 9-3 or 9-4.

Indication of Drywell Pressure 1.83 psig Indication of RPV water level -113 RPV pressure below injection valve open permissive setpoint Performance Manipulation of controls as required to open indicator the affected ECCS injection valve(s) or pump turbine controls from panel 9-3 or 9-4:

Operator places affected ECCS injection valve(s) control switch(es) to OPEN on panel 9-3 or 9-4.

Performance Red light illuminates and Green light feedback extinguishes for the affected ECCS injection valve(s), as applicable, on panel 9-3 or 9-4.

RCIC or HPCI turbine speed and flow rate rises, as applicable, on panel 9-3 or 9-4.

Justification Attempting to align high pressure ECCS for the chosen systems must be performed to determine performance their availability by the time TAF is reached in limit order to properly implement EOP-1A decision steps regarding restoring and maintaining RPV level. Attempting to align low pressure ECCS systems can only be done one RPV pressure falls below the injection valve RPV pressure permissive and will only be effective once RPV pressure falls below the shutoff head of the respective ECCS pump. The reduction in RPV pressure will normally be via Emergency Depressurization, which is a separate critical task bounded by a minimum RPV level.

BWR Owners App. B, Contingency 1, step C1-1 Group Appendix SIMULATOR SET-UP Rev. 2

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 1 of 47 Facility: Cooper Nuclear Station Scenario No.: 4 Op-Test No.: CNS 15-01 Examiners: ____________________________ Operators: _____________________________

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

1. Shift RRMG oil pumps

2. Place RHR SP cooling in service

3. Respond to CRD pump trip

4. Respond to a RR Pump A #1 seal failure and subsequent pump trip

5. Respond to a RR Pump A #2 seal failure. Vent PC

6. Respond to a FW line break inside PC

7. Respond to failure of HPCI to automatically start

8. Respond to loss of RPV level indication, flood the RPV to the MSLs Initial Conditions: Plant is operating at 100% power Inoperable Equipment: None Turnover:

The plant is operating at 100% power.

Planned activities for this shift are:

Shift RRMG oil pumps Place RHR SPC in service Maintain present power level Scenario Notes:

This is a new scenario.

Validation Time: 75 minutes Rev. 2

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 2 of 47 Event Malf. No. Event Type Event No. Description 1 N/A N (ATC,CRS) Shift RRMG lube oil pumps N (BOP,CRS) 2 rf rh14 Place RHR loop B in SPC, Min flow valve de-energizes open.

TS (CRS) 3 rd08b C (ATC,CRS) CRD Pump B trip.

C (BOP) 4 rr10a A (CREW) RR Pump A Seal #1 leak and RR Pump A trip.

TS (CRS)

C rr04b 5 (BOP,ATC,CRS) RR Pump A Seal #2 leak, vent PC.

rr11a A (CREW)

FW Line B break in PC-Scram M (CT-1) Initiate drywell sprays when torus pressure 6 fw18b exceeds 10 psig, prior to drywell temperature reaching (CREW) 280F and prior to torus pressure exceeding the Pressure Suppression Pressure (PSP) curve.

7 hp01 C (BOP,CRS) HPCI fails to automatically start.

Loss of RPV level instruments, RPV flooding.

(CT-2) When RPV level cannot be determined and torus level is above 6', open six SRVs before drywell radiation reaches 150 R/hr or entering PC Flooding.

(CT-3) When RPV level cannot be determined and the reactor has been depressurized below the shutoff head M of the respective pump(s), inject into the RPV to flood 8 NBI various to the Main Steam Lines before drywell radiation (CREW) reaches 150 R/hr or entering PC Flooding.

(CT-4) When RPV level cannot be determined and at least 1 SRV is manually opened, isolate MSIVs, MSL drains, HPCI steam supply, and RCIC steam supply, prior to RPV water level rising to the bottom of the main steam lines.

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

Critical Task (As defined in NUREG 1021 Appendix D)

CREW notation for Abnormal (A) and Major (M) events denotes ATC, BOP, and CRS are credited.

Rev. 2

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 3 of 47 Quantitative Attributes Table ES-304-1 Attribute Actual Description Target Malfunctions after 1-2 1 HPCI fails to automatically start.

EOP entry RR pump trip.

Abnormal Events 2-4 2 RR seal leakage FW line break inside PC Major Transients 1-2 2 Loss of all RPV level instruments EOP-1A EOP entries requiring substantive action 1-2 2 EOP-3A EOP contingencies requiring substantive 0-2 1 EOP- 2B action (CT-1) Initiate drywell sprays when torus pressure exceeds 10 psig, prior to drywell temperature reaching 280F and prior to torus pressure exceeding the Pressure Suppression Pressure (PSP) curve.

(CT-2) When RPV level cannot be determined and torus level is above 6', open six SRVs before drywell radiation reaches 150 R/hr or entering PC Flooding.

EOP based Critical (CT-3) When RPV level cannot be determined and the 2-3 4 reactor has been depressurized below the shutoff head Tasks of the respective pump(s), inject into the RPV to flood to the Main Steam Lines before drywell radiation reaches 150 R/hr or entering PC Flooding.

(CT-4) When RPV level cannot be determined and at least 1 SRV is manually opened, isolate MSIVs, MSL drains, HPCI steam supply, and RCIC steam supply, prior to RPV water level rising to the bottom of the main steam lines.

Shift RRMG oil pumps Normal Events N/A 2 Place RHR Suppression Pool Cooling in service Reactivity Manipulations N/A 0 N/A CRD Pump trip.

Instrument/ RR Pump A Seal #1 failure with RR pump trip.

N/A 4 Component Failures RR Pump A Seal #2 failure.

HPCI fails to automatically start CRD Pump trip.

RR Pump A Seal #1 failure with RR pump trip.

Total Malfunctions N/A 4 RR Pump A Seal #2 failure.

HPCI fails to automatically start Top 10 systems and operator actions important to risk that are tested:

Nuclear Boiler Instrumentation (Event 8)

Residual Heat Removal in Containment Spray Mode (Event 6)

HPCI (Event 7)

ADS/SRV Operator fails to depressurize with SRVs Operator fails to initiate ADS and initiate ECCS early Rev. 2

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 4 of 47 SCENARIO

SUMMARY

The plant is operating at 100% power.

Event 1 After the crew takes the watch, the ATC shifts RRMG oil pumps B1 and B3 per procedure 2.2.68.1. The oil pump shift is in preparation for tagging out the oil pump later in the shift.

Event 2 The BOP then places RHR in Suppression Pool Cooling in preparation a HPCI run the next shift. As the system's minimum flow valve starts to close it de-energizes in an intermediate position. The CRS declares the LPCI subsystem inoperable per LCO 3.5.1, Condition A. The valve is declared inoperable per LCO 3.6.2.3, Condition A.

Event 3 (Triggered by Lead Examiner)

After Technical Specifications are addressed for LPCI inoperable, the operating CRD pump trips requiring the ATC to start the standby pump.

Event 4 (Triggered by Lead Examiner)

After the CRD pump trip is addressed, RR Pump A develops a #1 seal failure. The crews responds to rising seal temperatures and lowers RR pump speed.

Subsequently the RR pump trips, placing plant operation near the buffer region of the power to flow map. The CRS enters TS LCO 3.4.1.

Event 5 (Triggered by Lead Examiner)

After the RR pump trip is addressed, the pump's #2 seal develops a leak requiring the pump to be isolated and the PC to be vented with Standby Gas Treatment.

Event 6 (Triggered by Lead Examiner)

After the #2 seal failure is addressed, FW line B develops a leak inside PC. The reactor scrams on high drywell pressure. The crew initiates Torus and Drywell Sprays (CT-1).

Event 7 (No Trigger required)

HPCI fails to automatically start on high drywell pressure and must be started manually.

Event 8 (Triggered by Lead Examiner)

All RPV level instrumentation is lost and the crew emergency depressurizes (CT-2).

Steam lines are isolated (CT-3) and the crew uses injection systems to flood the RPV to the bottom of the Main Steam Lines (CT-4).

Rev. 2

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 5 of 47 The exercise ends when emergency depressurization is complete and RPV level is maintained at the bottom of the MSLs.

Rev. 2

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 6 of 47 Critical Tasks (CT-1) Initiate drywell sprays when torus (CT-2) When RPV level cannot be pressure exceeds 10 psig, prior to drywell determined and torus level is above 6',

temperature reaching 280F and prior to open six SRVs before drywell radiation torus pressure exceeding the Pressure reaches 150 R/hr or entering PC Flooding.

Suppression Pressure (PSP) curve.

EVENT 6 8 Safety Drywell sprays are initiated in two legs of Depressurization of the RPV is necessary to significance EOP-3A: Temperature and Pressure control. perform the RPV flooding actions for the following reasons:

The open SRVs establish a path from Regarding drywell temperature, if operation the RPV capable of rejecting energy of all available drywell cooling is unable to in excess of decay heat to ensure the terminate increasing drywell temperature RPV flooding actions are successful.

before the structural design temperature limit Reduced RPV pressure results in increased injection flow rates, initiated to affect the required drywell reducing the total time required to flood the RPV.

temperature reduction status of the DSIL and Reduced RPV pressure reduces the adequate core cooling permitting. Spray water inventory loss through operation effects a drywell pressure and non-isolable leaks and breaks.

temperature reduction through the combined Dynamic loading on the SRVs and effects of evaporative cooling and convective downstream piping is minimized as cooling. RPV water level reaches and is discharged through these valves.

Regarding drywell pressure, operation of RPV depressurization can be most easily and drywell sprays reduces primary containment rapidly accomplished by opening SRVs. The pressure by condensing any steam that may ADS valves are used first since they are the be present and by absorbing heat from the most reliable, considering component containment atmosphere through the qualifications, pneumatic supply systems, combined effects of evaporative and initiation circuitry, and control power. In convective cooling. Drywell sprays are addition, the relative locations of the ADS initiated when torus pressure exceeds the valve discharges provide uniform distribution of Torus Spray Initiation Pressure (10# torus the heat load around the suppression pool.

pressure) to preclude chugging the cyclic The direction to open all ADS valves requires condensation of steam at the downcomer manual action, even if the valves are already openings of the drywell vents. When a steam open on high pressure. Automatic valve bubble collapses at the exit of the operation in the relief or safety mode does not downcomers, the rush of water drawn into accomplish the objective of this step, even if the downcomers to fill the void induces low-low set logic has actuated. RPV flooding stresses at the junction of the downcomers conditions are defined based on steam flow and the vent header in Mark I containments through the SRVs. Direct manual control must and at the junction of the downcomers. be established to ensure that the valves Repeated application of such stresses could remain open as RPV pressure decreases.

cause fatigue failure of these joints; thereby, SRVs may be opened only if suppression pool creating a direct path between the drywell water level is above the elevation of the top of and torus. When drywell sprays are initiated, the discharge devices. If the SRVs were the resulting pressure reduction opens the opened with the discharge devices exposed, vacuum breakers, drawing non-condensable steam would pass directly into the suppression from the torus back into the drywell. This chamber airspace, bypassing the suppression condition defines the Torus Spray Initiation pool. The resulting pressure increase could Pressure. As the drywell atmosphere is exceed the maximum pressure capability of purged to the torus and replaced by steam, the primary containment.

torus pressure increases. The SCSIP is the Failing to depressurize could prevent recovery lowest torus pressure which can occur when of RPV level above MSCRWL, resulting in core 95% of the non-condensable in the drywell damage have been transferred to the torus. Since the failure mode is based on fatigue failure, a precise time limit or pressure cannot be provided. Therefore, prompt initiation of drywell sprays is required based on existing Rev. 2

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 7 of 47 EOP priorities.

Cueing Rising torus pressure indicated on SPDS and Erratic or inconsistent indication on all RPV panel 9-3 recorder PC-LRPR-1A. level indications, and CRS declares RPV level cannot be determined.

Cursor approaching unsafe boundary on PSP graph display on SPDS.

Performance Aligns torus spray on panel 9-3 using RHR Manipulation of any six SRV controls on panel indicator loop A and/or B: 9-3:

SRV-71A places CONTMT COOLING 2/3 CORE SRV-71B VALVE CONTROL PERMISSIVE SRV-71E switch to MANUAL OVERRD SRV-71G SRV-71H opens RHR-MO-39B, if closed SRV-71C SRV-71D closes close RHR-MO-27B, OUTBD SRV-71F INJECTION VLV, if necessary starts RHR PUMP(s), if not running For drywell spray, opens RHR-MO-31B Performance On panel 9-3, RHR pump/valve control Crew will observe SRV light indication go from feedback switch light indication consistent with green to red, amber pressure switch lights intended operation (Red - open/running, illuminate, reactor pressure lowering on SPDS Green - closed/stopped). and panel 9-3 and 9-5 meters and recorders, RHR flow rate rises on recorder RHR-FR-143 and SRV tailpipe temperatures rise on and indicator RHR-FI-133A(B.) recorder MS-TR-166.

Torus/drywell pressure stabilizes/lowers on SPDS and panel 9-3 recorder PC-LRPR-1A.

Justification When torus pressure cannot be maintained Before 150R/hr in the drywell was chosen for the chosen below PSP is the EOP-3A, step PC/P-4 because this is an indicator of loss of RPV performance criteria requiring transition to emergency level and the shielding effect of the water, limit depressurization. indicating core exposure, yet it is much lower than the 2500R/hr trigger point during RPV Flooding that indicates gross cladding failure is in progress. 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-2B or 7B, and exiting to SAGs is neither required nor authorized.

BWR Owners App. B, step PC/P-1. App. B, Contingency#4 Group Appendix Rev. 2

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 8 of 47 Critical Tasks (CT-3) When RPV level cannot be (CT-4) When RPV level cannot be determined and the reactor has been determined and at least 1 SRV is manually depressurized below the shutoff head of opened, isolate MSIVs, MSL drains, HPCI the respective pump(s), inject into the steam supply, and RCIC steam supply, RPV to flood to the Main Steam Lines prior to RPV water level rising to the bottom before drywell radiation reaches 150 R/hr of the main steam lines.

or entering PC Flooding.

EVENT 8 8 Safety Once the SRVs have been opened to Steam lines connected to the RPV are isolated significance depressurize the RPV, injection systems are prior to initiating action to flood the RPV to aligned to flood the RPV and establish core preclude damage which may occur from cold cooling by submergence. The list of flooding water coming in contact with the hot metal, methods includes all motor-driven systems excessive loading of lines or hangers not capable of injecting into the RPV. Any or all designed to accommodate the weight of water, of these systems may be used, as and flooding of steam driven equipment (RCIC necessary, to flood the RPV to the elevation turbine, main turbine, etc.). Isolation is of the main steam lines. Steam-driven performed, however, only if the status of SRVs systems are not listed since, with SRVs open assures the RPV will remain depressurized and the reactor shut down, the RPV will during the flooding evolution. For non-ATWS, depressurize to below the turbine stall only one SRV open is required to meet this pressures. Failing to raise RPV level to and condition.

observable point could prevent recovery of RPV level above MSCRWL, resulting in core damage.

Cueing Erratic or inconsistent indication on all RPV Erratic or inconsistent indication on all RPV level indications, and CRS declare RPV level level indications, and CRS declares RPV level undetermined. cannot be determined, and SRVs have been Six ADS valves have been manually manually opened IAW EOP-2B or EOP-7B for opened. RPV depressurization.

Performance Crew establishes injection flow by Crew places the following valve control indicator manipulating controls as required to start the switches to CLOSE:

associated pumps and align system valves Inboard MSIVs on panel 9-3 for injection using at least two pumps of the MSL Drains on panel 9-4 following systems: HPCI steam supply on panel 9-3 Main condensate/booster pumps on panel A RCIC steam supply on panel 9-4 RHR/LPCI loop A and/or B on panel 9-3 Core spray A and/or B

[Operator places affected ECCS pump(s) control switch(es) to START and valve control switches to OPEN (or CLOSE, if necessary)]

Performance Indication that the RPV is flooded to the main Indication for applicable isolation valves Green feedback steam lines may include one or more of the light illuminates and Red light extinguishes.

following indication on panels 9-3, 9-4, 9-5 or field reports by the booth operator:

Rising RPV pressure

If a main steam line is not isolated, field report of two-phase flow conditions audible in the vicinity of the steam tunnel, main steam equalizing header, or main turbine stop and bypass valves

Actuation of HPCI, RCIC or main steam line high flow logic

Field report of water leakage from HPCI or Rev. 2

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 9 of 47 RCIC turbine shaft seals

HPCI/RCIC STM DRAIN POT LEVEL HI alarms

SRVs re-open and stay open at RPV pressures below 50 psig above torus pressure

If injection sources are aligned with torus suction, torus water level:

- decreases as RPV and steam lines are flooded

- stabilizes when steam lines are full

Local torus water temperatures near open SRVs Justification LOCA severity should result in a near linear Equipment damage due to cold water cannot for the chosen RPV level reduction that gives the crew an occur until water level reaches the main steam performance initial trend on all level instruments. Failing lines.

limit all of the level instruments should occur within about 30 seconds and should yield inconsistent indications such that there is no doubt level cannot be determined (e.g. LOCA conditions with operation in the possible boiling region of the RPVST curve, minimal RPV injection, level slowly lowering to -100 CFZ, then all level instruments fail upscale within 10 seconds, simulating all reference legs flashing).

BWR Owners App. B, Contingency #4. App. B, Contingency#4, step C4-2.2 Group Appendix Rev. 2

ML17122A086: Difference between revisions

Latest revision as of 18:37, 4 February 2020

Contents

Text

SUMMARY

SUMMARY

SUMMARY

Navigation menu

ML17122A086: Difference between revisions

Latest revision as of 18:37, 4 February 2020

Text

SUMMARY

SUMMARY

SUMMARY

Navigation menu

Search