ML14261A309

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ML14261A309
Person / Time
Site: FitzPatrick Constellation icon.png
Issue date: 07/23/2014
From:
FitzPatrick
To: Todd Fish
Operations Branch I
Shared Package
ML14101A125 List:
References
TAC U01887
Download: ML14261A309 (30)
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ES-401 Written Examination Outline Form ES-401-1 Facility: James A. Fitzpatrick Date of Exam : September 2014 RO KIA Category Points SRO-Only Points Tier Group K K K K K K A A A A G Total A2 G* Total 1 2 3 4 5 6 1 2 3 4

  • 1 3 4 3 3 3 4 20 3 4 7 1.

Emergency

& 2 1 2 1 1 1 1 7 1 2 3 Plant Evolutions Tier Totals 4 6 4 4 4 5 27 4 6 10 1 3 2 2 3 2 3 2 2 2 3 2 26 2 3 5 2.

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

3. Generic Knowledge & Abilities 10 7 Categories 3 2 2 3 1 2 2 2 Note: 1. Ensure that at least two topics from every applicable KIA 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 KIA category shall not be less than two).
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 and justified; operationally important, site-specific systems that are not included on the outline should be added. Refer to section D.1.b of ES-401 , for guidance regarding elimination of inappropriate KIA 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 KAs 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 KIA categories.

7.* The generic (G) KIAs in Tiers 1 and 2 shall be selected from Section 2 of the KIA 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 KIA's

8. On the following pages, enter the KIA numbers, a brief description of each topic, the topics' importance ratings (IR) 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 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 KIA Catalog, and enter the KIA numbers, descriptions, IRs, and point totals(#) on Form ES-401-3. Limit SRO selections to KIAs that are linked to 10CFR55.43

ES-401 2 Form ES-401-1 JAF 14-2 NRC Exam Written Examination Outline Emergency and Abnormal Plant Evolutions -Tier 1 Group 1 EAPE # I Name Safety Function KIA Topic(s)

EA2.03- Ability to determine and/or 295030 Low Suppression Pool interpret the following as they apply to X LOW SUPPRESSION POOL WATER 3.9 76 Water Level / 5 LEVEL: Reactor pressure AA2.02- Ability to determine and/or interpret the following as they apply to 295019 Partial or Complete PARTIAL OR COMPLETE LOSS OF X 3.7 77 Loss of Instrument Air I 8 INSTRUMENT AIR: Status of safety-related instrument air system loads (see AK2.1 - AK2.19)

EA2.05- Ability to determine and/or interpret the following as they apply to 295028 High Drywell X HIGH DRYWELL TEMPERATURE: 3.8 78 Temperature /5 Torus/suppression chamber pressure:

Plant-Specific 2.4.4 - Emergency Procedures I Plan:

Ability to recognize abnormal indications 295038 High Off-site Release X for system operating parameters that 4.7 79 Rate /9 are entry-level conditions for emergency and abnormal operating procedures.

2.2.37 -Ability to determine operability 700000 Generator Voltage and X and/or availability of safety related 4.6 80 Electric Grid Disturbances eQuipment.

2.4.11 - Emergency Procedures I Plan:

295016 Control Room Abandonment I 7 X Knowledge of abnormal condition 4.2 81 procedures.

2.4.41 - Emergency Procedures I Plan:

295023 Refueling Accidents I 8 X Knowledge of the emergency action 4.6 82 level thresholds and classifications.

AK1 .01 - Knowledge of the operational 295001 Partial or Complete implications of the following concepts as Loss of Forced Core Flow X they apply to PARTIAL OR COMPLETE 3.5 39 Circulation I 1 & 4 LOSS OF FORCED CORE FLOW CIRCULATION: Natural circulation AK1 .03- Knowledge of the operational implications of the following concepts as 295023 Refueling Accidents I 8 X 3.7 40 they apply to REFUELING ACCIDENTS: Inadvertent criticality AK1.02- Knowledge of the operation applications of the following concepts 600000 Plant Fire On-site I 8 X 2.9 41 as they apply to Plant Fire On Site: Fire Fighting AK2 .07- Knowledge of the interrelations 295006 SCRAM I 1 X between SCRAM and the following : 4.0 42 Reactor pressure control EK2.07- Knowledge of the interrelations between LOW SUPPRESSION POOL 295030 Low Suppression Pool X WATER LEVEL and the following : 3.5 43 Water Level / 5 Downcomer/horizontal vent submergence AK2 .01 - Knowledge of the interrelations 295004 Partial or Complete between PARTIAL OR COMPLETE X 3.1 44 Loss of DC Power I 6 LOSS OF D.C. POWER and the following : Battery charger AK3.03- Knowledge of the reasons for 295016 Control Room the following responses as they apply to X 3.5 45 Abandonment I 7 CONTROL ROOM ABANDONMENT:

Disabling control room controls 295025 High Reactor Pressure I EK3.03- Knowledge of the reasons for X 3.8 46 3 the following responses as they apply to

ES-401 2 Form ES-401-1 JAF 14-2 NRC Exam Written Examination Outline Emergency and Abnormal Plant Evolutions -Tier 1 Group 1 EAPE # I Name Safety Function KIA Topic(s)

HIGH REACTOR PRESSURE: HPCI operation : Plant-Specific EK3.05- Knowledge of the reasons for 295031 Reactor Low Water the following responses as they apply to X REACTOR LOW WATER LEVEL:

4.2 47 Level/2 Emergency depressurization AA 1.02- Ability to operate and/or 295003 Partial or Complete monitor the following as they apply to X 4.2 48 Loss of AC Power I 6 PARTIAL OR COMPLETE LOSS OF A. C. POWER: Emergency generators AA 1.02- Ability to operate and/or monitor the following as they apply to 295019 Partial or Complete X PARTIAL OR COMPLETE LOSS OF 3.3 49 Loss of Instrument Air /8 INSTRUMENT AIR: Instrument air system valves: Plant-Specific EA 1.07- Ability to operate and/or 295038 High Off-site Release monitor the following as they apply to X 3.6 50 Rate /9 HIGH OFF-SITE RELEASE RATE:

Control room ventilation: Plant-Specific EA2.02- Ability to determine and/or 295024 High Drywell Pressure I interpret the following as they apply to X 3.9 51 5 HIGH DRYWELL PRESSURE: Drywell temperature AA2.04- Ability to determine and/or interpret the following as they apply to 295018 Partial or Complete X PARTIAL OR COMPLETE LOSS OF 2.9 52 Loss of CCW I 8 COMPONENT COOLING WATER:

System flow AA2 .07- Ability to determine and/or 295021 Loss of Shutdown interpret the following as they apply to Cooling /4 X LOSS OF SHUTDOWN COOLING:

2.9 53 Reactor recirculation flow 2.4.30- Emergency Procedures I Plan :

Knowledge of events related to system 700000 Generator Voltage and operation I status that must be reported Electric Grid Disturbances X to internal organizations or external 2.7 54 agencies, such as the state, the NRC, or the transmission system operator.

2.1.20 - Conduct of Operations: Ability 295005 Main Turbine Generator X to interpret and execute procedure 4.6 55 Trip /3 steps.

2.4.18- Emergency Procedures I Plan:

295026 Suppression Pool High Water Temperature I 5 X Knowledge of the specific bases for 3.3 56 EOPs.

EK2.03- Knowledge of the interrelations 295028 High Drywell between HIGH DRYWELL Temperature /5 X TEMPERATURE and the following:

3.6 57 Reactor water level indication 2.4.4 - Emergency Procedures I Plan:

295037 SCRAM Conditions Ability to recognize abnormal indications Present and Reactor Power X for system operating parameters that 4.5 58 Above APRM Downscale or are entry-level conditions for emergency Unknown /1 and abnormal operating procedures.

KIA Category Totals: 3 4 3 3 3/3 4/4 Group Point Total:

I 20/7

ES-401 3 Form ES-401-1 JAF 14-2 NRC Exam Written Examination Outline Emergency and Abnormal Plant Evolutions -Tier 1 Group 2 EAPE # I Name Safety Function KIA Topic(s)

AA2.02- Ability to determine and/or interpret the following as they apply to 295015 Incomplete SCRAM /1 X 4.2 83 INCOMPLETE SCRAM: Control rod position 2.4.47- Emergency Procedures I Plan:

295036 Secondary Ability to diagnose and recognize trends Containment High Sump/Area X in an accurate and timely manner 4 .2 84 Water Level/ 5 utilizing the appropriate control room reference material.

2.4.31 -Emergency Procedures I Plan :

295009 Low Reactor Water X Knowledge of annunciator alarms, 4.1 85 Level/2 indications, or response procedures.

EK1 .01 - Knowledge of the operational implications of the following concepts as 295029 High Suppression Pool X they apply to HIGH SUPPRESSION 3.4 59 Water Level/ 5 POOL WATER LEVEL: Containment integrity AK2 .05- Knowledge of the interrelations 295002 Loss of Main between LOSS OF MAIN CONDENSER X 2.7 60 Condenser Vacuum I 3 VACUUM and the following : Feedwater system AK3.01 - Knowledge of the reasons for the following responses as they apply to 295009 Low Reactor Water X LOW REACTOR WATER LEVEL: 3.2 61 Level/2 Recirculation pump run back: Plant-Specific AA 1.01 -Ability to operate and/or 295010 High Drywell Pressure I monitor the following as they apply to X 3.4 62 5 HIGH DRYWELL PRESSURE: Drywell ventilation/cooling AA2.03- Ability to determine and/or 295020 Inadvertent interpret the following as they apply to X 3.7 63 Containment Isolation I 5 & 7 INADVERTENT CONTAINMENT ISOLATION: Reactor power 2.1.31 -Conduct of Operations: Ability to locate control room switches, controls, 500000 High Containment Hydrogen Concentration I 5 X and indications, and to determine that 4.6 64 they correctly reflect the desired plant lineup.

AK2.09- Knowledge of the interrelations between HIGH REACTOR WATER 295008 High Reactor Water X LEVEL and the following: Reactor water 3.1 65 Level/2 cleanup system (ability to drain): Plant-Specific KIA Category Totals: 1 2 1 1 1/1 1/2 Group Point Total: 7/3

ES-401 4 Form ES-401-1 JAF 14-2 NRC Exam Written Examination Outline Plant Systems - Tier 2 Group 1 K K K K K K A A A Imp.

System #I Name A2 G Q#

1 2 3 4 5 6 1 3 4 A2.01 -Ability to (a) predict the impacts of the following on the REACTOR WATER LEVEL CONTROL SYSTEM; and (b) 259002 Reactor Water Level based on those predictions, use X 3.4 86 Control System procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of any number of main steam flow inputs A2.11 -Ability to (a) predict the impacts of the following on the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM 223002 PC IS/Nuclear Steam SUPPLY SHUT-OFF; and (b)

X based on those predictions, use 3.9 87 Supply Shutoff procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Standby liquid initiation 2.2.38 - Equipment Control:

262001 AC Electrical X Knowledge of conditions and 4.5 88 Distribution limitations in the facility license.

2.1.25 - Conduct of Operations:

Ability to interpret reference 2150031RM X 4.2 89 materials, such as graphs, curves, tables, etc.

2.1 .7 - Conduct of Operations:

Ability to evaluate plant performance and make 215005 APRM I LPRM X operational judgments based on 4.7 90 operating characteristics, reactor behavior, and instrument interpretation.

K1 .01 -Knowledge of the physical connections and/or cause-effect relationships 203000 RHRILPCI : Injection between RHRILPCI :

X INJECTION MODE (PLANT 2.8 1 Mode SPECIFIC) and the following :

Condensate storage and transfer system: Plant-Specific K1.04- Knowledge of the physical connections and I or cause-effect relationships 400000 Component Cooling X between CCWS and the 2.9 2 Water following: Reactor coolant system, in order to determine source (s)

K2.02- Knowledge of electrical 209001 LPCS X power supplies to the following: 2.5 3 Valve power K2.01 -Knowledge of electrical 212000 RPS X power supplies to the following : 3.2 4 RPS motor-generator sets

ES-401 4 Form ES-401-1 JAF 14-2 NRC Exam Written Examination Outline Plant Systems - Tier 2 Group 1 K K K K K K A A A System #I Name A2 G Imp. Q#

1 2 3 4 5 6 1 3 4 K3.11 - Knowledge of the effect that a loss or malfunction of the UNINTERRUPTABLE POWER 262002 UPS (AC/DC) X 2.8 5 SUPPLY (A.C./D.C.) will have on following : MSIVs: Plant-Specific K3.02 - Knowledge of the effect that a loss or malfunction of the SHUTDOWN COOLING 205000 Shutdown Cooling X SYSTEM (RHR SHUTDOWN 3.2 6 COOLING MODE) will have on following: Reactor water level:

Plant-Soecific K4.06- Knowledge of PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF design 223002 PCIS/Nuclear Steam X feature(s) and/or interlocks 3.4 7 Supply Shutoff which provide for the following:

Once initiated, system reset requires deliberate operator action K4.04- Knowledge of SOURCE RANGE MONITOR (SRM) 215004 Source Range SYSTEM design feature(s)

X 2.8 8 Monitor and/or interlocks which provide for the following : Changing detector position K5.03 -Knowledge of the operational implications of the 259002 Reactor Water Level following concepts as they apply X 3.1 9 Control to REACTOR WATER LEVEL CONTROL SYSTEM: Water level measurement K5.01 -Knowledge of the operational implications of the following concepts as they apply 218000 ADS X 3.8 10 to AUTOMATIC DEPRESSURIZATION SYSTEM: ADS loaic ooeration K6.05- Knowledge of the effect that a loss or malfunction of the following will have on the 261000 SGTS X 3.1 11 STANDBY GAS TREATMENT SYSTEM: Reactor protection svstem: Plant-Soecific K6.03- Knowledge of the effect that a loss or malfunction of the following will have on the 217000 RCIC X 3.5 12 REACTOR CORE ISOLATION COOLING SYSTEM (RCIC):

Suppression pool water suoolv A 1.01 -Ability to predict and/or monitor changes in parameters associated with operating the 264000 EDGs X 3.0 13 EMERGENCY GENERATORS (DIESEUJET) controls including :

Lube oil temoerature

ES-401 4 Form ES-401-1 JAF 14-2 NRC Exam Written Examination Outline Plant Systems -Tier 2 Group 1 K K K K K K A A A System # I Name A2 G Imp. Q#

1 2 3 4 5 6 1 3 4 A 1.06 -Ability to predict and/or monitor changes in parameters associated with operating the 215003 IRM X INTERMEDIATE RANGE 3.3 14 MONITOR (IRM) SYSTEM controls including : Lights and alarms A2.01 -Ability to (a) predict the impacts of the following on the D.C. ELECTRICAL DISTRIBUTION ; and (b) based 263000 DC Electrical 15 X on those predictions, use 2.8 Distribution procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Grounds A2.07 -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 215005 APRM I LPRM X 3.2 16 predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions:

Recirculation flow channels flow mismatch A3.02 -Ability to monitor automatic operations of the 206000 HPCI X HIGH PRESSURE COOLANT 3.8 17 INJECTION SYSTEM including :

System Flow: BWR-2,3,4 A3.04- Ability to monitor automatic operations of the 239002 SRVs X RELIEF/SAFETY VALVES 3.6 18 including : Acoustical monitor noise: Plant-Specific A4.07- Ability to manually 211000 SLC X operate and/or monitor in the 3.6 19 control room : Lights and alarms A4.01 -Ability to manually 300000 Instrument Air X operate and/or monitor in the 2.6 20 control room : Pressure gauges 2.4.21 - Emergency Procedures I Plan: Knowledge of the parameters and logic used to assess the status of safety 262001 AC Electrical functions, such as reactivity X 4.0 21 Distribution control, core cooling and heat removal , reactor coolant system integrity, containment conditions, radioactivity release control, etc.

2.4.2 - Emergency Procedures I Plan : Knowledge of system set 217000 RCIC X points, interlocks and automatic 4.5 22 actions associated with EOP entry conditions.

ES-401 4 Form ES-401-1 JAF 14-2 NRC Exam Written Examination Outline Plant Systems - Tier 2 Group 1 K K K K K K A A A System # I Name A2 G Imp. Q#

1 2 3 4 5 6 1 3 4 K1.01 - Knowledge of the physical connections and/or cause-effect relationships 263000 DC Electrical X between D.C. ELECTRICAL 3.3 23 Distribution DISTRIBUTION and the following : A.C. electrical distribution K4.08- Knowledge of STANDBY LIQUID CONTROL SYSTEM design feature(s) 211000SLC X and/or interlocks which provide 4.2 24 for the following: System initiation upon operation of SBLC control switch A4.09- Ability to manually operate and/or monitor in the 206000 HPCI X 3.8 25 control room: Suppression pool level: BWR-2,3.4 K6.01 - Knowledge of the effect that a loss or malfunction of the 259002 Reactor Water Level following will have on the Control X REACTOR WATER LEVEL 3.2 26 CONTROL SYSTEM: Plant air systems KIA Category Totals: 3 2 2 3 2 3 2 2/2 2 3 2/3 Group Point Total:

I 26/5

ES-401 5 Form ES-401-1 JAF 14-2 NRC Exam Written Examination Outline Plant Systems - Tier 2 Group 2 K K K K K K A A A System # I Name A2 G Imp. Q#

1 2 3 4 5 6 1 3 4 A2 .09- Ability to (a) predict the impacts of the following on the RECIRCULATION FLOW CONTROL SYSTEM; and (b) 202002 Recirculation Flow based on those predictions, use X 3.3 91 Control procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Recirculation flow mismatch: Plant-Specific 2.4.50 - Emergency Procedures 245000 Main Turbine I Plan: Ability to verify system Generator and Auxiliary X alarm setpoints and operate 4.0 92 Systems controls identified in the alarm response manual.

A2 .01 -Ability to (a) predict the impacts of the following on the PLANT VENTILATION

, SYSTEMS; and (b) based on 288000 Plant Ventilation those predictions, use X procedures to correct, control, 3.4 93 Systems or mitigate the consequences of those abnormal conditions or operations: High drywell pressure: Plant-Specific K1 .14- Knowledge of the physical connections and/or cause-effect relationships 233000 Fuel Pool X between FUEL POOL 2.5 27 Cooling/Cleanup COOLING AND CLEAN-UP and the following : Reactor building ventilation K1 .20 - Knowledge of the physical connections and/or cause-effect relationships 259001 Reactor Feedwater X between REACTOR 3.1 28 FEEDWATER SYSTEM and the following : Main steam system:

TDRFPs-Onlv K3.02- Knowledge of the effect that a loss or malfunction of the 239001 Main and Reheat X MAIN AND REHEAT STEAM 3.1 29 Steam SYSTEM will have on following :

Condenser K4.03 -Knowledge of RECIRCULATION FLOW CONTROL SYSTEM design 202002 Recirculation Flow X feature(s) and/or interlocks 3.0 30 Control which provide for the following :

Signal failure detection : Plant-Soecific K5.03 - Knowledge of the operational Implications of the 241000 Reactor/Turbine following concepts as they apply Pressure Regulating X to REACTOR/TURBINE 3.5 31 System PRESSURE REGULATING SYSTEM: Reactor power vs.

reactor oressure

ES-401 5 Form ES-401-1 JAF 14-2 NRC Exam Written Examination Outline Plant Systems - Tier 2 Group 2 K K K K K K A A A System #I Name A2 G Imp. Q#

1 2 3 4 5 6 1 3 4 K6.01 - Knowledge of the effect that a loss or malfunction of the following will have on the FIRE 286000 Fire Protection X 3.1 32 PROTECTION SYSTEM: A. C.

electrical distribution : Plant-Specific A 1.01 -Ability to predict and/or monitor changes in parameters associated with operating the 272000 Radiation RADIATION MONITORING X 3.2 33 Monitoring SYSTEM controls including:

Lights, alarms, and indications associated with normal operations A2.01 -Ability to (a) predict the impacts of the following on the FUEL HANDLING EQUIPMENT; and (b) based on 234000 Fuel Handling X those predictions, use 3.3 34 Equipment procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Interlock failure A3.01 -Ability to monitor automatic operations of the 288000 Plant Ventilation X PLANT VENTILATION 3.8 35 Systems SYSTEMS including :

Isolation/initiation signals A4.09 -Ability to manually operate and/or monitor in the 271000 Offgas X 3.3 36 control room : Offgas system controls/components 2.4.47- Emergency Procedures I Plan: Ability to diagnose and 290003 Control Room recognize trends in an accurate HVAC X 4.2 37 and timely manner utilizing the appropriate control room reference material.

A2.01 -Ability to (a) predict the impacts of the following on the TRAVERSING IN-CORE PROBE; and (b) based on those 215001 Traversing In-core predictions, use procedures to X 2.7 38 Probe correct, control, or mitigate the consequences of those abnormal conditions or operations: Low reactor water level: Mark-I&II(Not-BWR1)

KIA Category Totals: 2 0 1 1 1 1 1 2/2 1 1 1/1 Group Point Total:

I 12/3

Facility: James A. Fitzpatrick Date: September 2014 RO SRO-Only Category KIA# Topic IR Q# IR Q#

Knowledge of primary and secondary plant 2.1 .34 3.5 94 chemistry limits.

Ability to explain and apply all system limits 2.1 .32 3.8 66

1. and precautions.

Conduct Knowledge of industrial safety procedures of Operations (such as rotating equipment, electrical, high 2.1 .26 3.4 67 temperature, high pressure, caustic, chlorine, oxygen and hydrogen).

Ability to locate and operate components, 2.1.30 4.4 75 including local controls.

Subtotal 3 1 Knowledge of the process for managing maintenance activities during shutdown 2.2.18 3.9 95 operations, such as risk assessments, work prioritizaion, etc.

Ability to apply technical specifications for a 2.2.40 4.7 99 system.

2.

Equipment Ability to perform pre-startup procedures for Control the facility, including operating those controls 2.2.1 4.5 68 associated with plant equipment that could affect reactivity.

Knowledge of the process used to track 2.2 .43 3.0 69 inoperable alarms.

Subtotal 2 2 Knowledge of radiation or containment hazards that may arise during normal, 2.3.14 3.8 96 abnormal, or emergency conditions or activities.

Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms,

3. 2.3.5 2.9 100 portable survey instruments, personnel Radiation monitoring equipment, etc.

Control Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, 2.3.15 2.9 70 portable survey instruments, personnel monitoring equipment, etc.

Knowledge of Radiological Safety Principles pertaining to licensed operator duties, such 2.3.12 as containment entry requirements, fuel 3.2 71 handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

Subtotal 2 2 Knowledge of local auxiliary operator tasks 2.4.35 during an emergency and the resultant 4.0 97 operational effects.

Knowledge of events related to system operation I status that must be reported to 2.4.30 internal organizations or external agencies, 4.1 98 such as the state, the NRC, or the transmission system operator.

4.

Emergency Knowledge of the parameters and logic used Procedures I to assess the status of safety functions , such Plan as reactivity control, core cooling and heat 2.4.21 4.0 72 removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

2.4.27 Knowledge of "fire in the plant" procedures. 3.4 73 Knowledge of procedures relating to a 2.4.28 3.2 74 security event (non-safeguards information).

Subtotal 3 2 Tier 3 Point Total 10 7

ES-401 Record of Rejected K/As Form ES-401-4 Tier I Randomly Selected KIA Reason for Rejection Group The following topics I K/As were excluded from the systematic and random sampling process:

295027 High Containment This topic applies to plants with Mark Ill Temperature containments only. The facility has a Mark I 1I 1 containment.

295011 High Containment This topic applies to plants with Mark Ill Temperature containments only. The facility has a Mark I 1I 2 containment.

207000 Isolation (Emergency) This system is not installed at the facility.

2I 1 Condenser 209002 HPCS This system is not installed at the facility.

2I1 201004 RSCS This system is no longer installed at the facility.

2/2 201005 RCIS This system is not installed at the facility.

2/2 2.2.3 Knowledge of the design, This KIA applies to multi-unit facilities only.

procedural, and operational G differences between units.

2.2.4 Ability to explain the This KIA applies to multi-unit facilities only.

variations in control board/control room layouts, G systems, instrumentation, and procedural actions between units at a facility.

The following KJAs were reiected following the systematic and random sampling process:

Question 91 The facility is a BWR-4 and does not have Recirculation flow control valves (FCVs).

202002 Recirculation Flow Control Randomly re-selected KIA 202002 ..

Recirculation Flow Control A2.01 - Ab1hty to (a)

A2 .08- Ability to (a) predict the predict the impacts of the following on the impacts of the following on the RECIRCULATION FLOW CONTROL SYSTEM; RECIRCULATION FLOW and (b) based on those predictions, use 2/2 CONTROL SYSTEM; and (b) procedures to correct, control, or mitigate the based on those predictions, use consequences of those abnormal conditions or procedures to correct, control, operations: Recirculation pump trip or mitigate the consequences of those abnormal conditions or operations: FCV lockup: BWR-5,6 Question 28 The facility has steam-driven Feedwater pumps, not motor-driven Feedwater pumps.

259001 Reactor Feedwater Randomly re-selected KIA 259001 Reactor K2.01 -Knowledge of electrical Feedwater K1.20- Knowledge of the physical power supplies to the following:

2/2 connections and/or cause-effect relationships Reactor feedwater pump(s): between REACTOR FEEDWATER SYSTEM Motor-Driven-Only and the following: Main steam system:

TDRFPs-Only Question 31 The facility does not have a unitized actuator as part of the Reactor/Turbine Pressure 241000 Reactor/Turbine Regulating System.

Pressure Regulating System Randomly re-selected KIA 241000 K5.07- Knowledge of the Reactor/Turbine Pressure Regulating System operational Implications of the K5.03- Knowledge of the operational 2/2 following concepts as they Implications of the following concepts as they apply to REACTOR/TURBINE apply to REACTOR/TURBINE PRESSURE PRESSURE REGULATING REGULATING SYSTEM: Reactor power vs.

SYSTEM: Unitized actuator reactor pressure operation: Fermi-Only

Question 17 Unable to be develop a question for the randomly selected KIA without overlapping with 206000 HPCI Question 12.

A3.08- Ability to monitor Randomly reselected KIA 206000 HPCI A3.02-automatic operations of the Ability to monitor automatic operations of the 2/1 HIGH PRESSURE COOLANT HIGH PRESSURE COOLANT INJECTION INJECTION SYSTEM including: SYSTEM including: System Flow: BWR-2,3,4 Condensate storage tank level :

BWR-2,3,4 Question 21 Unable to develop a question at the correct license level for the randomly selected generic 262001 AC Electrical KIA related to the given system.

Distribution Randomly reselected KIA 262001 AC Electrical 2.4.41 - Emergency Procedures Distribution 2.4.21 -Emergency Procedures I I Plan: Knowledge of the Plan: Knowledge of the parameters and logic 2/1 emergency action level used to assess the status of safety functions, thresholds and classifications. such as reactivity control , core cooling and heat removal , reactor coolant system integrity, containment conditions, radioactivity release control , etc.

Question 23 Unable to develop a question without overlap with question on last NRC exam . Resampling 263000 DC Electrical to limit overlap.

Distribution Randomly reselected KIA 263000 DC Electrical K5.01 - Knowledge of the Distribution K1 .01 -Knowledge of the physical operational implications of the connections and/or cause-effect relationships 2/1 following concepts as they between D.C. ELECTRICAL DISTRIBUTION apply to D.C. ELECTRICAL and the following : A. C. electrical distribution DISTRIBUTION: Hydrogen generation during battery charging

Question 30 !he facility does not have a design feature or Interlock specifically related to break of a scoop 202002 Recirculation Flow tube.

Control Randomly re-selected KIA 202002 K4.01 -Knowledge of Recirculation Flow Control K4.03- Knowledge RECIRCULATION FLOW of RECIRCULATION FLOW CONTROL 2/2 CONTROL SYSTEM design SYSTEM design feature(s) and/or interlocks feature(s) and/or interlocks which .provide for the following : Signal failure which provide for the following:

detection: Plant-Specific Scoop tube break: Plant-Specific Question 58 The randomly selected generic KIA is oversampled on this exam (Questions 37 58 295037 SCRAM Conditions 84, 89). ' '

Present and Reactor Power Above APRM Downscale or Randomly re-selected KIA 295037 SCRAM Unknown Conditions Present and Reactor Power Above APRM .Downscale or Unknown 2.4.4 - Ability to 1 /1 2.4.47- Emergency Procedures recogmze abnormal indications for system I Plan: Ability to diagnose and operating parameters that are entry-level recognize trends in an accurate conditions for emergency and abnormal and timely manner utilizing the operating procedures.

appropriate control room reference material.

Question 61 The facility does not have a specific Feedwater pump run-out flow control feature.

295009 Low Reactor Water Level Randomly re-selected KIA 295009 Low Reactor Water Level AK3.01 -Knowledge of the AK3.02- Knowledge of the reasons for the following responses as they reasons for the following 1 /2 apply to LOW REACTOR WATER LEVEL*

responses as they apply to Recirculation pump run back: Plant-Specific.

LOW REACTOR WATER LEVEL: Reactor feed pump runout flow control : Plant-Specific

Question 81 The randomly selected generic KIA does not support developing an operationally relevant 295016 Control Room question for the randomly selected abnormal Abandonment plant evolution.

2.2.42 - Equipment Control: Randomly re-selected KIA 295016 Control 1 11 Ability to recognize system Room Abandonment 2.4.11 - Emergency parameters that are entry-level Procedures I Plan: Knowledge of abnormal conditions for Technical condition procedures.

Specifications.

Question 89 The randomly selected generic KIA is oversampled on this exam.

215003 IRM Randomly re-selected KIA 215003 IRM 2.1 .25-2.4.47- Emergency Procedures Emergency Procedures I Plan: Ability to I Plan : Ability to diagnose and interpret reference materials, such as graphs, 211 recognize trends in an accurate curves, tables, etc.

and timely manner utilizing the appropriate control room reference material.

Question 94 The randomly selected generic KIA does not support developing a written question at the 2.1.45 -Ability to identify and appropriate license level and is tested interpret diverse indications to 3 extensively during the operating portion of the validate the response of another examination.

indicator.

Randomly re-selected KIA 2.1.34 - Knowledge of primary and secondary plant chemistry limits.

Question 91 Unable to develop an appropriate SRO-Ievel question for the randomly selected KIA 202002 Recirculation Flow Control Randomly re-selected KIA 202002 Recirculation Flow Control A2.09- Ability to (a)

A2.01 -Ability to (a) predict the predict the impacts of the following on the impacts of the following on the RECIRCULATION FLOW CONTROL SYSTEM ;

RECIRCULATION FLOW 212 and (b) based on those predictions, use CONTROL SYSTEM; and (b) procedures to correct, control, or mitigate the based on those predictions, use consequences of those abnormal conditions or procedures to correct, control, operations: Recirculation flow mismatch: Plant-or mitigate the consequences of Specific.

those abnormal conditions or operations: Recirculation pump trip

Question 80 The randomly selected generic KIA does not support developing an operationally relevant 700000 Generator Voltage and question without testing minutia for the Electric Grid Disturbances randomly selected abnormal plant evolution.

2.4.45 -Ability to prioritize and 1 /1 Randomly re-selected KIA 700000 Generator interpret the significance of Voltage and Electric Grid Disturbances 2.2.37-each annunciator or alarm.

Ability to determine operability and/or availability of safety related equipment.

ES-301 Administrative Topics Outline Form ES-301-1 Facility: James A. Fitz12atrick Date of Examination: Se12tember 2014 Examination Level: RO Operating Test Number: 14-2 Administrative Topic Type Describe activity to be performed (see Note) Code*

P, D,S Perform RHR Lineup Verification Conduct of Operations 14-1 NRC KIA2.1.29 (4.1), ST-2AN Re-activation of RO License Conduct of Operations M,R KIA 2.1.4 (3.3), ODS0-30 Perform ST-23C, Jet Pump Operability- Two Loop Equipment Control M,R KIA2.2.12 (3.7), ST-23C Determine Worker Exposure for Emergent Work Radiation Control D,R KIA 2.3.4 (3.2), EN-RP-201 Emergency Procedures/Plan NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.

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

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

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

ES-301 Administrative Topics Outline Form ES-301-1 Facility: James A. Fitz12atrick Date of Examination : Se12tember 2014 Examination Level: SRO Operating Test Number: 14-2 Administrative Topic Type Describe activity to be performed (see Note) Code*

P, D, R Verify Fuel Movement Sheets Conduct of Operations 12-1 NRC KIA 2.1 .35 (3 .9), OSP-66.001 Re-activation of SRO License Conduct of Operations D,R KIA 2.1.4 (3.8), ODS0-30 Review ST-23C, Jet Pump Operability- Two Loop Equipment Control M, R KIA2.2.12 (4.1), ST-23C Determine Visitor RCA Access Requirements Radiation Control D, R KIA 2.3.4 (3.7), EN-RP-201 , EN-RP-202 Classify General Emergency, Determine PARs, and Emergency Procedures/Plan M,R Perform Event Notification KIA 2.4.44 (4.4), IAP-1 , IAP-2 NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required .

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

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

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

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: James A. Fitzpatrick Date of Examination: September 2014 Exam Level : RO/SRO Operating Test No.: 14-2 Control Room Systems@ (8 for RO); (7 for SR0-1); (2 or 3 for SRO-U, including 1 ESF)

System I JPM Title Type Code* Safety Function

a. Restore CRD to Normal Alignment Following ATWS, CRD Controller Fails in Automatic P, D,A,S 1

2014-1 NRC KIA 201001 A2 .07 (3.2/3.1 }, EP-3, OP-25

b. Restore ESW After Injection Into RBCLC P, D, S 8

KIA 400000 A4.01 (3.1/3.0}, OP-21, OP-40 2012-2 NRC

c. Start First Feedwater Pump N, L,S 2 KIA 259001 A4.02 (3.9/3.7} , OP-2A
d. Start EDGs A and C Due to Hurricane Warning N, EN , S 6 KIA 264000 A4.04 (3.7/3.7) , OP-22
e. Restore Shutdown Cooling After Isolation, RHR Pump Trips M,A, L, S 4 KIA 205000 A4.01 (3.7/3.7) , AOP-30
f. Standby Gas Treatment Initiation Verification with Initiation Failure and Low Reactor Building D/P D, A, EN,S 9 KIA 261000 A4.06 (3.3/3.6} , OP-20
g. Verify and Reset Group 2 & RWCU Isolation, Multiple Valves Open D, A, EN , S 5 KIA 223002 A4.03 (3.6/3.5} , AOP-15
h. (RO Only) Plant Shutdown From Outside the Control Room -

ATC Actions D, S 7 KIA 295016 AA1 .01 (3 .8/3.9), AOP-43 In-Plant Systems@ (3 for RO) ; (3 for SR0-1); (3 or 2 for SRO-U)

i. Vent Torus to Lower Primary Containment Pressure D, A, E 5 KIA 295010 AA1 .05 (3.1/3.4) , EP-6
j. Lineup SLC Test Tank for Injection D, R, E 2 KIA 295031 EA1 .08 (3 .8/3.9) , EP-8
k. Electrically Disarm a CRD HCU D, R 1 KIA 201003 A2.02 (3.7/3.8) , OP-25

@ All RO and SR0-1 control room (and in-plant) systems must be different and serve different safety functions; all 5 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 I SR0-11 SRO-U (A)Iternate path 4-6 I 4-6 I 2-3 (C)ontrol room (D)irect from bank S91S81S4 (E)mergency or abnormal in-plant ~11~11~1 (EN)gineered safety feature - I - I ~1 (control room system)

(L)ow-Power I Shutdown ~11~11~1 (N)ew or (M)odified from bank including 1(A) ~21~21~1 (P)revious 2 exams s 3 Is 3 Is 2 (randomly selected)

(R)CA ~11~11~1 (S)imulator

Appendix D Scenario Outline Form ES-D-1 Facility: James A. Fitzpatrick Scenario No.: NRC-1 Op-Test No.: 14-2 Examiners: Operators:

Initial Conditions: The plant is operating at approximately 90% power. RCIC is out of service for maintenance. Torus cooling is in service on RHR loop A.

Turnover: Secure Torus cooling per OP-13B sections F.1 and F.7.

Event Malf. Event Event No. No. Type* Description N- Shutdown Torus Cooling 1 N/A BOP, SRO OP-13B RHRSW Pump A Trip 2 SW04:A C-SRO ARP 09-3-1-25, Technical Specifications 1-ATC, APRM B Fails Upscale 3 NM14:B SRO OP-16, ARP 09-5-1-4 I-BOP, HPCI Inadvertent Initiation 4 HP05 SRO AOP-77, AOP-32 , Technical Specifications C-BOP FW05:A Feedwater Pump A High Vibration and Pump Trip 5 R-ATC ,

FW01 :A SRO ARP 09-6-4-11, AOP-41 RR15:A Loss of Coolant Accident 6 M-All MS02:A EOP-2, EOP-4 FW01 :B Feedwater Pump B Trip, HPCI Trip 7 C-All HP02 EOP-2 ED43:A Loss of Offsite Power 8 C-All ED43:B AOP-72

  • (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor

Facility: James A. Fitzpatrick Scenario No. : NRC-1 Op-Test No. : 14-2

1. Total malfunctions (5-8) 7 Events 2, 3, 4, 5, 6, 7, 8
2. Malfunctions after EOP entry (1-2) 2 Events 7 & 8
3. Abnormal events (2-4) 3 Events 3, 4, 5
4. Major transients (1-2) 1 Event 6
5. EOPs entered/requiring substantive actions (1-2) 2 EOP-2, EOP-4
6. EOP contingencies requiring substantive actions (0-2) 2 EOP-2 Alt Level Leg, EOP-2 Emergency Depress Leg
7. Critical tasks (2-3) 2 CRITICAL TASK DESCRIPTIONS:

CT-1: Given a coolant leak inside the Containment, the crew will spray the Drywell, in accordance with EOP-4.

CT-2: Given a coolant leak, a loss of high pressure injection systems, and the inability to restore and maintain Reactor water level above the Top of Active Fuel (TAF), the crew will initiate actions for an Emergency RPV Depressurization before Reactor water level lowers below -19", in accordance with EOP-2.

Appendix D Scenario Outline Form ES-D-1 Facility: James A. Fitzpatrick Scenario No.: NRC-2 Op-Test No.: 14-2 Examiners: Operators:

Initial Conditions: The plant is operating at approximately 100% power. Air Compressor B is out of service for maintenance. Core Spray pump A is ready for full flow testing .

Turnover: Perform Core Spray full flow testing per ST-3PA, starting at step 8.7.6.

Event Malt. Event Event No. No. Type* Description N- Perform Core Spray Full Flow Test 1 N/A BOP, SRO ST-3PA C- Core Spray Pump Overload 2 Override BOP, SRO ARP 09-3-1-31, Technical Specifications I-BOP, Power Control Requests Minimum Generator Reactive Load 3 Report SRO OP-11A C- Control Rod Drive Pump Trip, One Control Rod Drive Accumulator RD06 4 ATC, Low Pressure RD09 SRO AOP-69, ARP-09-5-1-43, Technical Specifications R- Main Turbine Bearing High Vibration 5 TU04 BOP, SRO AOP-66 R-ATC Loss of Main Condenser Vacuum 6 MC01 C- BOP, SRO AOP-31 , AOP-1 RP01AB Failure of RPS and ARI to Actuate 7 RP01BB M-All EOP-2, EOP-3 RP09 C- SLC Squib Valves Fail to Fire, Recirculation Pumps Fail to SL02 8 ATC, Automatically Trip RR13 SRO EOP-3 EG01 C- Main Generator Trip, Two Turbine Bypass Valves Fail Closed 9 BOP, TC04 SRO EOP-3

  • (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor

Facility: James A. Fitzpatrick Scenario No.: NRC-2 Op-Test No. : 14-2

1. Total malfunctions (5-8) 8 Events 2, 3, 4, 5, 6, 7, 8, 9
2. Malfunctions after EOP entry (1-2) 2 Events 8 & 9
3. Abnormal events (2-4) 4 Events 3, 4, 5, 6
4. Major transients (1-2) 1 Event?
5. EOPs entered/requiring substantive actions (1-2) 1 EOP-2
6. EOP contingencies requiring substantive actions (0-2) 1 EOP-3
7. Critical tasks (2-3) 2 CRITICAL TASK DESCRIPTIONS:

CT-1: Given a failure to scram with Reactor power above 2.5%, the crew will lower Reactor power by one or more of the following methods, in accordance with EOP-3:

  • Terminating and preventing all RPV injection except SLC, RCIC, and CRD

Appendix D Scenario Outline Form ES-D-1 Facility: James A. Fitzpatrick Scenario No. : NRC-3 Op-Test No. : 14-2 Examiners: Operators:

Initial Conditions: The plant is operating at approximately 100% power. Air Compressor B is out of service for maintenance.

Turnover: Swap RBCLC pumps per OP-40 section G.1. Then lower Reactor power to 95% using Recirculation flow.

Event Malf. Event Event No. No. Type* Description N- Swap RBCLC Pumps 1 N/A BOP, SRO OP-40 R- Lower Reactor Power to 95% Using Recirculation Flow 2 N/A ATC, SRO OP-27 1-ATC, Recirc Flow Unit Failure 3 RR23:B SRO OP-16, Technical Specifications C- Electrical Fault on Bus L-15 4 ED21 :A BOP, SRO AOP-18A, Technical Specifications C- SRV C Inadvertently Opens 5 AD06 :C BOP, SRO AOP-36 AD06:G SRV G Fails Open, SRV Tailpipe Break Into Torus Airspace 6 AD08 :G M-All AOP-36 , AOP-1 , EOP-2 , EOP-4 MS16:G 1-ATC, Main Turbine Fails to Trip 7 TC11 SRO AOP-1 RH01 RHR Pump Trip, RHR Pump Suction Valve Fails Closed, RHRSW C-8 Remotes BOP, Crossties Fail Closed SRO EOP-4, EOP-2 Overrides

  • (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor

Facility: James A. Fitzpatrick Scenario No.: NRC-3 Op-Test No.: 14-2

1. Total malfunctions (5-8) 6 Events 3, 4, 5, 6, 7, 8
2. Malfunctions after EOP entry (1-2) 2 Events 7 & 8
3. Abnormal events (2-4) 2 Events 4 & 5
4. Major transients (1-2) 1 EventS
5. EOPs entered/requiring substantive actions (1-2) 2 EOP-2, EOP-4
6. EOP contingencies requiring substantive actions (0-2) 1 EOP-2 Emergency Depress Leg
7. Critical tasks (2-3) 2 CRITICAL TASK DESCRIPTIONS:

CT-1: Given a stuck open SRV, the crew will scram the Reactor before Torus water temperature exceeds the Boron Injection Initiation Temperature, in accordance with AOP-36 and EOP-4.

CT-2: Given the inability to maintain Primary Containment conditions inside the Pressure Suppression Pressure, the crew will perform an Emergency RPV Depressurization, in accordance with EOP-4.

Appendix D Scenario Outline Form ES-D-1 Facility: James A. Fitzpatrick Scenario No.: NRC-4 Op-Test No.: 14-2 Examiners: Operators:

Initial Conditions: The plant is operating at approximately 65% power. Circulating Water pump C is ready to be returned to service following maintenance.

Turnover: Start Circulating Water pump C per OP-4 section 0 .17. Then, raise Reactor power using Recirculation flow per the provided reactivity instruction .

Event Malf. Event Event No. No. Type* Description N- Start Circulating Water Pump C 1 N/A BOP, SRO OP-4 R- Raise Reactor Power with Recirculation Flow 2 N/A ATC, SRO OP-27 1-ATC, Recirculation Pump B Runs High 3 Override SRO OP-27, Technical Specifications C- Loss of Steam Packing Exhauster A 4 Override BOP, SRO ARP 09-7-3-43, OP-240 C- Loss of RBCLC Flow to Recirculation Pump B 5 SW01:B BOP, SRO OP-27, AOP-8, Technical Specifications Main Steam Leak into Turbine Building 6 MS05 M-All AOP-40, AOP-1, EOP-2 1-ATC, All MSIVs Fail to Automatically Close 7 RP12 SRO AOP-40 FW03/04 Feedwater Pump Controllers Fail Low, HPCI Flow Controller RC02 I-BOP, Failure in Automatic, RCIC Fails to Automatically Initiate, SLC Fails 8 To Initiate, CRO Pumps Trip RD06 SRO Override EOP-2 , OP-19 I-BOP, 9 HP02 HPCI Trips SRO RC03 RCIC Trips, ADS Fails to Automatically Initiate 10 I-AII A001 EOP-2

  • (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor

Facility: James A. Fitzpatrick Scenario No.: NRC-4 Op-Test No. : 14-2

1. Total malfunctions (5-8) 8 Evenm3,4,5,6, 7,8,9,10
2. Malfunctions after EOP entry (1-2) 4 Evenm 7, 8, 9, 10
3. Abnormal events (2-4) 3 Evenm 3, 4, 5
4. Major transients (1-2) 1 EventS
5. EOPs entered/requiring substantive actions (1-2) 1 EOP-2
6. EOP contingencies requiring substantive actions (0-2) 0
7. Critical tasks (2-3) 3 CRITICAL TASK DESCRIPTIONS:

CT-1 : Given a loss of all RBCLC flow to an operating Recirculation pump, prevent a prolonged Recirculation seal loss of coolant accident by one or more of the following, in accordance with OP-27:

  • Securing the Recirculation pump
  • Isolating the Recirculation pump CT-2: Given a Main Steam leak into the Turbine Building and a failure of the MSIVs to automatically close, isolate the Main Steam lines, in accordance with AOP-40.

CT-3: Given the Reactor above the injection pressure of adequate injection sources, lower Reactor pressure to allow adequate Reactor injection, in accordance with EOP-2.

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