ML12111A190

Hope Creek - Draft - Outlines (Folder 2)

ML12111A190
Person / Time
Site:	Hope Creek
Issue date:	01/23/2012
From:	Fish T H Operations Branch I
To:	Public Service Enterprise Group
Jackson D E
Shared Package
ML113070699	List: IR 05000354/2012301 ML12061A285 ML12083A212 ML12083A223 ML12083A235 ML12083A242 ML12111A179 ML12111A188 ML12111A190 ML12111A200 ML12111A202 ML12172A046
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TAC U01845
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Text

ES-401 Written Examination Outline Form ES-401-1 II Facility:

Hope Creek Station Date of Exam: 03/05/2012 SRO-Only Points Tier Group K 1 K 2 K 3 A2 G* Total

1. 1 3 3 4 3 4 3 20 4 3 7 Emergency

& 1-------1-+--+---1 Abnormal Plant Evolutions 2 1 2 N/A 1 1 N/A 1 7 2 3 Tier Totals 4 4 6 4 5 4 27 5 5 10 2. Plant Systems r-___: __

Tier Totals 3 2 3 4 3 4 4 4 38 4 4 8 3. Generic Knowledge and Abilities Categories 1 2 2 3 4 2 3 3 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 (Le., 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. 7 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/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 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 KlAs 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) KlAs 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 KlAs. 8. On the following pages, enter the KIA 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 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 KlAs that are linked to 10 CFR 55.43.

ES-401 1 Form ES-401-1 Hope Creek Written Examination Emergency and Abnormal Plant Evolutions

-Tier 1 Group KJA Topic(s)EAPE # / Name Safety Function 295005 Main Turbine Generator Trip / 3 X AA2.07 -Ability to determine and/or interpret the tollowing as they apply to MAIN TURBINE GENERATOR TRIP: Reactor water level 3.6 76 295030 Low Suppression Pool Water Level / 5 X EA2.01 -Ability to determine and/or interpret the tollowing as they apply to LOW SUPPRESSION POOL WATER LEVEL: Suppression pool level 4.2 77 295006 SCRAM / 1 X AA2.02 -Ability to detemline and/or interpret the following as they apply to SCRAM: Control rod position 4.4 78 295021 Loss of Shutdown Cooling /4 X 2.4.34 -Emergency Procedures

/ Plan: Knowledge ofRO tasks performed outside the main control room during an emergency and the resultant operational effects. 4.1 79 295004 Partial or Total Loss of DC Pwr / 6 X 2.1.23 -Ability to pertorm specific system and integrated plant procedures during all modes of plant operation.

4.4 80 295024 High Drywell Pressure / 5 X 2.4.41 -Emergency Procedures i Plan: Knowledge of the emergency action level thresholds and classifications.

4.6 81 295031 Reactor Low Water Level / 2 X EA2.04 -Ability to determine and/or interpret the tollowing as they apply to REACTOR LOW WATER LEVEL: Adequate core cooling 4.8 82 600000 Plant Fire On-site / 8 X AKI.O I -Knowledge of the operation applications of the following concepts as they apply to Plant Fire On Site: Fire Classifications by type 2.5 39 295006 SCRAM / 1 X AKI.02 -Knowledge ofthe operational implications of the following concepts as they apply to SCRAM: Shutdown margin 3.4 40 295023 Refueling Acc Cooling Mode / 8 X AKl.03 -Knowledge of the operational implications of the following concepts as they apply to REFUELING ACCIDENTS:

Inadvertent criticality 3.7 41 295019 Partial or Total Loss of Inst. Air /8 X AK2.14 -Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR and the following:

Plant air systems 3.2 42 295004 Partial or Total Loss of DC Pwr /6 X AK2.02 -Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF D.C. POWER and the following:

Batteries 3.0 43 295031 Reactor Low Water Level /2 X EK2.14 -Knowledge of the interrelations between REACTOR LOW WATER LEVEL and the following:

Emergency generators 3.9 44 295003 Partial or Complete Loss of AC / 6 X AK3.01 -Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER: Manual and auto bus transfer 3.3 45 295026 Suppression Pool High Water Temp. / 5 X EK3.05 -Knowledge of the reasons for the following responses as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE:

Reactor SCRAM 3.9 46 295037 SCRAM Conditions Present and Reactor Power Above APRM Downscale or Unknown /1 X EK3.03 -Knowledge of the reasons for the following responses as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN : Lowering reactor water level 4.1 47 ES-401 2 Form Hope Creek Written Examination Emergency and Abnormal Plant Evolutions

-Tier 1 Group KIA Topic(s) Q# I EAPE # / Name Safety Function 295021 Loss of Shutdown Cooling / 4 X AA 1.04 -Ability to operate and/or monitor the following as they apply to LOSS OF SHUTDOWN COOLING: Alternate heat removal methods 3.7 48 295001 Partial or Complete Loss of Forced Core Flow Circulation / 1 &4 X AAl.07 -Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION:

Nuclear boiler instrumentation system 3.1 49 295016 Control Room Abandonment / 7 X AAI.02 -Ability to operate and/or monitor the following as they apply to CONTROL ROOM ABANDONMENT:

Reactor/turbine pressure regulating system 2.9 50 295024 High Drywell Pressure / 5 X EA2.0 I -Ability to determine and/or interpret the following as they apply to HIGH DRYWELL PRESSURE:

Drywell pressure 4.2 51 295038 High Off-site Release Rate / 9 X EA2.04 -Ability to determine and/or interpret the following as they apply to HIGH OFF-SITE RELEASE RATE: Source of off-site release 4.1 52 700000 Generator Voltage and Electric Grid Disturbances X AA2.02 -Ability to determine and/or interpret the following as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES:

Voltage outside the generator capability curve. 3.5 53 295005 Main Turbine Generator Trip / 3 X 2.1.31 -Ability to locate control room switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup. 4.6 54 295030 Low Suppression Pool Water Level / 5 X 2.2.12 -Equipment Control: Knowledge of surveillance procedures.

3.7 55 295028 High Drywell Temperature / 5 X 2.2.42 -Equipment Control: Ability to recognize system parameters that are entry-level conditions for Technical Specifications.

3.9 56 295018 Partial or Total Loss of CCW/8 X AA2.03 -Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER: Cause for partial or complete loss 3.2 57 295025 High Reactor Pressure /3 X EK3.02 -Knowledge of the reasons for the following responses as they apply to HIGH REACTOR PRESSURE:

Recirculation pump trip: Plant-Specific 3.9 58 KIA Category Totals: 3 3 4 3 4/4 3/3 Group Point Total: I 20/7 ES-401 3 Form Hope Creek Written Examination Emergency and Abnormal Plant Evolutions

-Tier 1 Group KIA Topic(s)EAPE # I Name Safety Function 295013 High Suppression Pool Temperature I 5 X AA2.01 High Suppression Pool Tempemture, Ability to detelmine and/or interpret the following as they apply to HIGH SUPPRESSION POOL TEMPERATURE:

Suppression Pool Temperature 4.0 83 295032 High Secondary Containment Area Temperature 15 X 2.1.20 -Conduct of Operations:

Ability to intclJlret and execute procedure steps. 4.6 84 295017 High Off-site Release Rate 19 X 2.4.30 -Emergency Procedures

/ Plan; Knowledge of events related to system operation,'

status that must be reported to internal organizations or external agencies, such as the state, the NRC, or the tmnsmission system opemtor. 4.1 85 295010 High Drywell Pressure 15 X AK1.03 -Knowledge of the opemtional implications of the following concepts as they apply to HIGH DRYWELL PRESSURE:

Temperature increases 3.2 59 295009 Low Reactor Water Level 12 X AK2.02 -Knowledge of the interrelations between LOW REACTOR WATER LEVEL and the following:

Reactor water level control 3.9 60 295034 Secondary Containment Ventilation High Radiation I 9 X EK3.03 -Knowledge ofthe reasons for the following responses as they apply to SECONDARY CONTAINMENT VENTILATION HIGH RADIATION:

Personnel evacuation 4.0 61 295035 Secondary Containment High Differential Pressure 15 X EAI.02 -Ability to operate and/or monitor the following as they apply to SECONDARY CONTAINMENT HIGH DIFFERENTIAL PRESSURE:

SBGT/FRVS 3.8 62 295036 Secondary Containment High SumplArea Water Levell 5 X EA2.02 -Ability to determine and/or intelJlret the following as they apply to SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL: Water level in the affected area 3.1 63 295008 High Reactor Water Level 12 X 2.1.27 -Conduct of Opemtions:

Knowledge of system pUlJlose and / or function.

3.9 64 295020 Inadvertent Cont. Isolation I 5 &7 X AK3.06 -Knowledge of the reasons for the following responses as they apply to INADVERTENT CONT AINMENT ISOLATION:

Suppression pool water level response 3.3 65 KIA Category Totals: 1 1 2 1 111 1/2 Group Point Total: I 7/3 ES-401 4 Form Hope Creek Written Examination Plant Systems -Tier 2 Group System # / Name K 1 K 2 K 3 K 4 K 5 K 6 A 1 A2 A 3 A 4 G Imp Q# 206000 HPCI X A2.04 -Ability to (a) predict the impacts of the tollowing on the HIGH PRESSURE COOLANT INJECTION SYSTEM; and l b) based on those predictions, use procedures to COITcct, control, or mitigate thc consequences of those abnonnal conditions or operations:

A.c. failures:

BWR-2,3,4 3.0 86 262001 AC X A2.04 -Ability to (a) predict the impacts of the tollowing on the A.C. ELECTRICAL DISTRIBUTION; and (b) based on those predictions, use procedures to cOlTect, control, or mitigate the consequences of those abnonnal conditions or operations:

Types of loads that, if de-energized, would hinder plant operdtion.

4.2 87 261000 SGTS X 2.2.44 -Equipment Control: Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives dIeet plant and system conditions.

4.4 88 215004 SRM X 2.2.38 Knowledge of conditions and limitations in the license 4.5 89 223002 PelS/Nuclear Steam Supply Shutoff X A2.09 -Ability to (a) predict the impacts of the tollowing on the PRIMARY CONTAINMENT ISOLA TION SYSTEM;NUCLEAR STEAM SUPPLY SHUT-OFF; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

System initiation 3.7 90 264000 EDGs X Kl.05 -Knowledge of the physical connectiollS ancll or cause-effect relationships between EMERGENCY GENERATORS (DIESEUJET) and the following:

Emergency generator fuel oil supply system 3.2 1 211000 SLC X Kl.Ol -Knowledge of the physical connectiollS ancllor cause-effect relationships between STANDBY LIQUID CONTROL SYSTEM and the following:

Core spray line break detection:

Plant-Specific 3.0 2 239002 SRVs X K2.01 -Knowledge of electrical power supplies to the following:

SRV solenoids 2.8 3 218000 ADS X K3.02 -Knowledge ofthe effect that a loss or malfunction of the AUTOMATIC DEPRESSURIZATION SYSTEM will have on the following:

Ability to rapidly depressurize the reactor 4.5 4 ES-401 5 Form Hope Creek Written Examination Plant Systems -Tier 2 Group System # I Name K K A 561 A2. Imp Q# 262001 AC Electrical Distribution X Ii: K3.01 -Knowledge of the effect that a loss or malfunction of the AC. ELECTRICAL DISTRIBUTION will have on following; Major System Loads 3.5 5 206000 HPCI X I';i!!",!'!ii":'

K3.01 Knowledge of the effect that a loss or malfunction of the HIGH PRESSURE COOLANT INJECTION SYSTEM will have on following; Reactor water level control: BWR-2,3,4 4.0 6 205000 Shutdown Cooling X ........ .... K4.03 Knowledge of SHUTDOWN COOLl:-lG SYSTEM (RHR SHUTOOWN COOLING MODE) design feature(s}

andior interlocks which provide for the following:

Low reactor water level: ific 3.8 7 262002 UPS (AC/DC) X Il K4.01 Knowledge of U:-IINTERRUPTABLE POWER SUPPLY (AC.lD.C.)

design feature(s) andior interlocks which provide for the following:

Transfer from preferred power to alternate power supplies 3.1 8 300000 Instrument Air X t* I**** ..

'. KS.Ol Knowledge of the operational implications of the following concepts as they apply to the INSTRUMENT AIR SYSTEM: Airc 2.5 9 263000 DC Electrical Distribution X .: KS.Ol Knowledge of the operational implications of the following concepts as they apply to D.C. ELECTRICAL DISTRIBUTION:

Hydrogen generation durin"" battery charging . 2.6 10 261000 SGTS X . '..:' I' ..... I",* K6.01 -Knowledge of the effect that a loss or malfunction of the following will have on the STANDBY GAS TREATMENT SYSTEM: A.C. electrical distribution 2.9 11 217000 RCIC X K6.04 -Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC): Condensate storage and transfer system 3.5 12 215005 APRM I LPRM X AI.07 -Ability to predict andior monitor changes in parameters associated with operating the AVERAGEPO\VERRANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM controls including; APRM (gain adjustment factor) 3.0 13 ES-401 6 Form Hope Creek Written Examination Plant Systems -Tier 2 Group Imp K K K K K K A A A A2 System # I Name G 1 2 3 4 5 6 1 4 3 AI,03 -Ability to predict and/or J monitor changes in parameters 209001 LPCS X " associated with operating the LOW 3.8 14 PRESSURE CORE SPRAY I'ii!'i":: " SYSTEM controls including:

Reactor water level ;" I,:"', A2.06 -Ability to (a) predict the impacts of the following on the 1 .' INTERMEDIATE RANGE 's,; MONITOR (IRM) SYSTEM; and 2150031RM X (b) based on those predictions, use 3.0 15 procedures to correct, control, or mitigate the consequences of those Il abnonnal conditions or operations:

i7 Faulty Range Switch A2,0 I -Ability to (a) predict the ,

impacts of the following on the SOURCE RANGE MONITOR 215004 Source Range (SRM) SYSTEM; and (b) based on 5< those predictions, use procedures to 2.7 16 Monitor ""'" conect, control, or mitigate the consequences of those abnonnal

"', conditions or operations:

Power supply degraded I,':,' A3.06 Ability to monitor 203000 RHRlLPCI:

Injection

' *.*. , automatic operations of the X RHRlLPCI:

INJECTION MODE 3.7 17 Mode ,: I**** ",' ** '**.* (PLANT SPECIFIC) including:

I ** ", ,.' Indicating lights and alanns I""'" A3,01 -Ability to monitor automatic operations of the CCWS 400000 Component Cooling X including:

Setpoints on instrument 3.0 18 Water Ii,' signal levels for nonnal operations, warnings, and trips that are applicable to the CCWS "

A4,02 -Ability to manually operate 223002 PCIS/Nuclear 3.9 19 Steam Supply Shutoff ,., X and/or morutor in the control room: ",*7: Manually initiate the system MOl 212000 RPS X I**** and/or monitor in the control room: 4.6 20 ProVIde manual SCRAM si2:Tlal(s)

I'h/ 2.1 30 -Conduct of Operations:

259002 Reactor Water .X' Ability to locate and operate 4.4 21 Level Control "; components, including local controls.

2.1.32 Ability to explain system 211000 SLC XF and apply system limits and 3.8 22 precautions.

2.4,11 -Emergency Procedures I 23 212000 RPS Plan: Knowledge of abnonnal 4.0 condition procedures.

X A4,04 Ability to manually operate 264000 Emergency and/or monitor in the Control Room: Manual start, loading, and 3.7 24 Generators (Diesel/Jet) stopping of emergency generator.

Plant Specific ES-401 7 Form ES-401-1 Hope Creek Written Examination Plant Systems -Tier 2 Group 239002 SRVs 217000 RCIC KJA Category Totals: 2 1 A3.0 I -Ability to monitor automatic operations of tbe RELIEFiSAFETY VALVES including:

SRV operation after ADS actuation K3.04 -Knowledge of the effect that a loss or malfunction of the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) will have on following:

Adequate core Group Point Total: 3.8 25 3.6 26 ES-401 8 Form Hope Creek Written Examination Plant Systems -Tier 2 Group System # / Name K 1 K 2 K 3 K 4 K 5 K 6 A 1 A2 A 3 A 4 G Imp. Q # 259001 Reactor Feedwater X A2.03 -Ability to (a) predict the impacts of the following on the REACTOR FEEDWATER SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Loss of condensate pump(s) 3.6 91 226001 RHR/LPCI:

CTMT Spray Mode X 2.2.42 -Equipment Control: Ability to recognize system parameters that are entry-level conditions for Technical Specifications.

4.6 92 223001 Primary CTMT and Aux. X 2.2.37 -Equipment Control: Ability to determine operability and / or availability of safety related equipment.

4.6 93 215001 Traversing In-core Probe X K1.08 -Knowledge of the physical connections and/or cause-effect relationships between TRAVERSING CORE PROBE and the following:

Reactor pressure vessel: (Not-BWR!l 2.5 27 286000 Fire Protection X K2.02 -Knowledge of electrical power supplies to the following:

Pumps 2.9 28 204000 RWCU X K3.02 -Knowledge of the effect that a loss or malfunction of the REACTOR WATER CLEANUP SYSTEM will have on following:

Reactor water level 3.1 29 202002 Recirculation Flow Control X K4.06 -Knowledge of RECIRCULATION FLOW CONTROL SYSTEM design feature(s) and/or interlocks which provide for the following:

Recirculation pump adequate NPSH: Plant-Specific 3.1 30 241000 ReactorlTurbine Pressure Regulator X K5.04 -Knowledge of the operational Implications of the following concepts as they apply to REACTORITURBINE PRESSURE REGULATING SYSTEM: Turbine inlet pressure vs. reactor pressure 3.3 31 201002 RMCS X K6.01 -Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR MANUAL CONTROL SYSTEM: Select matrix power 2.5 32 239001 Main and Reheat Steam X A 1.08 -Ability to predict and/or monitor changes in parameters associated with operating the MAIN AND REHEAT STEAM SYSTEM controls including:

Reactor pressure 3.8 33 ES-401 9 Form Hope Creek Written Examination Plant Systems -Tier 2 Group System # I Name K 1 K 2 K 3 K 4 K 5 K 6 A 1 A2 A 3 A 4 G Imp. Q # 288000 Plant Ventilation X A2.05 -Ability to (a) predict the impacts of the following on the PLANT VENTILATION SYSTEMS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Extreme outside weather conditions:

Plant-Specific 2.6 34 201001 CRD Hydraulic X A3.03 -Ability to monitor automatic operations of the CONTROL ROD DRIVE HYDRAULIC SYSTEM including:

System pressure 2.7 35 226001 RHR/LPCI:

CTMT Spray Mode X A4.15 -Ability to manually operate and/or monitor in the control room: Suppression chamber pressure:

Mark-I-II 3.6 36 259001 Reactor Feedwater X 2.4.2 -Emergency Procedures I Plan: Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions.

4.5 37 216000 Nuclear Boiler Inst. X A 1.04 -Ability to predict and/or monitor changes in parameters associated with operating the NUCLEAR BOILER INSTRUMENTATION controls including:

System venting 2.6 38 KIA Category Totals: 1 1 1 1 1 1 2 iii 1 1 1/2 Group Point Total: I 12/3 ES-401 10 Form Hope Creek Written Examination Generic Knowledge and Abilities Outline (Tier Facility:

Category 1. Conduct of Operations

2. Equipment Control 3. Radiation Control Hope Creek Station Date: 03/05/12 KIA # Topic Ability to use procedures to determine the 2.1.43 effects on reactivity of plant changes, such as RCS temperature, secondary plant, fuel depletion, etc. 2.1.35 Knowledge of the fuel-handling responsibilities of SRO's. Knowledge of industrial safety procedures (such as rotating equipment, electrical, high 2.1.26 temperature, high pressure, caustic, chlorine, oX)lgen and hydrogen).

Knowledge of how to conduct system 2.1.29 lineups, such as valves, breakers, switches, etc. Subtotal Knowledge of the process for managing maintenance activities during power 2.2.17 operations, such as risk assessments, work prioritization, coordination with the transmission system operator.

Knowledge of the process for making 2.2.6 changes to procedures.

Knowledge of less than or equal to one hour 2.2.39 Technical Specification action statements for systems. Subtotal Knowledge of radiation or containment 2.3.14 hazards that may arise during normal, abnormal, or emergency conditions or activities.

Knowledge of radiation monitoring systems, 2.3.15 such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. RO SRO-Only IR Q# IR Q# 4.3 94 3.9 98 3.4 66 4.1 67 2 2 3.8 95 3.0 68 3.9 69 ". 2 1 3.8 96 3.1 100 ES-401 11 Form Hope Creek Written Examination Generic Knowledge and Abilities Outline (Tier I I 4. Emergency Proced u res / Plan I I Tier 3 Point Total 2.3.11 Ability to control radiation releases.

3.8 70 Knowledge of Radiological Safety Principles pertaining to licensed operator duties, such as 2.3.12 containment entry requirements, fuel handling 3.2 71 responsibilities, access to locked high-radiation areas, aligning filters, etc. Ability to comply with radiation work permit 2.3.7 requirements during normal or abnormal 3.5 74 conditions.

Subtotal 3 ';%',!lii' 2 Knowledge of RO tasks performed outside 2.4.34 the main control room during an emergency 4.1 97 and the resultant operational effects. Ability to take actions called for in the facility 2.4.38 emergency plan, including supporting or 4.4 99 acting as emergency coordinator if required.

Knowledge of the bases for prioritizing safety 2.4.22 functions during abnormal/emergency 3.6 72 operations.

2.4.28 Knowledge of procedures relating to a 3.2 : 73 security event. 2.4.18 Knowledge of the specific bases for EOPs. 3.3 75 Subtotal 2 7 I I ES-401 12 Form Record of Rejected Tic.. I (:1roup 1 /1 1 / 1 1 /1 Randomly Selected KIA 295023/ AK1.02 295003/ AK.307 295016/ AA1.03 Reason for Rejection Question #41 -Knowledge of the operational implications of the following concepts as they apply to REFUELING ACCIDENTS:

Shutdown margin. Over-sampled topic, see question #40, almost identical subject matter. Randomly selected AK1.03 -Knowledge of the operational implications of the following concepts as they apply to

• RE::FUELING ACCIDENTS:

Inadvertent criticality.

Question #45, -Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER: Initiation of isolation condenser:

Plant-Specific Hope Creek does not have isolation condensers.

Randomly selected AK3.01 -Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER: Manual and auto bus transfer Question #50, -Ability to operate and/or monitor the following as they apply to CONTROL ROOM ABANDONMENT:

RPIS Control Rod Position Indication (RPIS) is not available outside of the Control Room. Randomly selected AA1.02 -Ability to operate and/or monitor the following as they apply to CONTROL ROOM ABANDONMENT:

Reactor/turbine pressure regulating system 1 /2 2950351 EA2.01 Question #83, -Ability to determine and/or interpret the following as they apply to SECONDARY CONTAINMENT HIGH DIFFERENTIAL PRESSURE:

Secondary containment pressure:

Plant-Specific.

Over-sampled topic, see question #62 -almost identical subject matter. Randomly selected 201 003/A2.1 0 Ability to (a) predict the impacts of the following on the CONTROL ROD At\ID DRIVE MECHANSIM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Excessive SCRAM time for a given drive mechanism 2/1 262001 I K3.04 Question #5 -Knowledge of the effect that a loss or malfunction of the A.C. ELECTRICAL DISTRIBUTION will have on following:

Uninterruptible power supply. Over-sampled topic, see question #8 -almost identical subject matter. Randomly selected K3.01 Knowledge of the effect that a loss or malfunction of the A.C. ELECTRICAL DISTRIBUTION will have on following:

Major system loads.

II 2/1 3 ES-401 13 Form Record of Rejected Question #4, -Knowledge of electrical power supplies to the following:

ADS logic. Over-sampled topic, see question #3 218000 / K2.01 almost identical subject matter. Randomly selected K3.02 -Ability to rapidly depressurize the reactor Question #24, -Ability to manually operate and/or monitor in the Control Room: TDRFP lockout reset. TDRFP Over-sampled topic, see questions

1. 21 and #89 very similar subject matter. 259002/ A4.09 . Randomly selected 264000/A4.04 Emergency Generators

! (Diesel/Jet)

Manual start, loading, and stopping of emergency 2/1

• generator.

Plant Specific 2/2 259001 / A2.08 2.2.23 1 / 1 295005/2.1.7 Question #91, -Ability to (a) predict the impacts of the following on the REACTOR FEEDWATER SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Loss of D.C. electrical power. Unable to write an SRO discriminating question for this topic. Randomly selected 259001/A2.03

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

Loss of condensate pump(s). Question #69, -Ability to track Technical Specification limiting conditions for operations.

Reactor Operators are not responsible for this task. Randomly selected 2.2.39 -Knowledge of less than or equal to one hour Technical Specification action statements for systems. Question #54 Conduct of Operations:

Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

Unable to write a discriminating question to ! adequately address all the attributes of the selected KIA. Randomly selected 2.1.31 -Ability to locate control room switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup. Question #89 -Emergency Procedures

/ Plan: Ability to verify that the alarms are consistent with the plant conditions.

Topic is 2 / 1 259002 I 2.4.46 sampled, see Q's #60 and #76 Randomly selected 215004 SRMs and 2.2.38 -Knowledge of conditions and limitations in the facility license. II E8-401 15 Form Record of Rejected Question # 80, -Emergency Procedures I Plan: Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures.

After review of question selected, question 1 11 2950041 2.4.4 #78 is very similar in nature and will be retained, however #80 will be reselected.

Randomly selected 295004/2.1.23

-Ability to perform specific system and integrated plant procedures during all modes of plant operation (Partial or Complete Loss of DC power) Question #22, -Equipment Control: Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

Determining the ! status of limiting conditions for operations (LCOs) is an SRO task 2/1 211000 / 2.2.36 at Hope Creek and is an unsuitable KIA for the RO section of the exam. I Randomly selected 211000/2.1.32

-Ability to explain system and II apply system limits and precautions. (SLC)

ES-401 14 Form Record of Rejected 1 12 201003/A2.10 Question #83, -Had randomly selected:

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

Excessive SCRAM time for a given drive mechanism.

This was chosen in error and was rejected due to being in the Tier 2 Group 2 NOT Tier 2 Group 1 as required by the outline. The original KIA, 2950351 EA2.01, Ability to determine and/or interpret the following as they apply to SECONDARY CONTAINMENT HIGH DIFFERENTIAL PRESSURE:

Secondary containment pressure:

Plant-Specific.

Was an over-sampled topic, see question #62 -almost identical subject matter. 2/1 215005/ A2.10 Randomly selected 295013/A2.01 High Suppression Pool ! Temperature, Ability to determine and/or interpret the following as they apply to HIGH SUPPRESSION POOL TEMPERATURE:

Suppression Pool Temperature Question #87, -A2.1 0 -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 correct, control, or mitigate the consequences of those abnormal conditions Changes in void concentration.

Due to a heavy concentration of Nuclear Instrumentation topics. Rejected this KIA and reselected an additional topic. 2/1 215003/2.2.44 Randomly selected 262001/A2.04

-Ability to (a) predict the impacts of the following on the AC. ELECTRICAL DISTRIBUTION; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Types of loads that, if de-energized, would hinder plant operation.

Question #88, -Equipment Control: Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives effect plant and system conditions.

Due to a heavy concentration of Nuclear Instrumentation topics. Reselected a different system and retained the original generic part of the KIA. This was done to maintain the balance of the outline. I Randomly selected 261000/2.2.44 Ability to interpret control room indications to verify the status and operation of a system, i and understand how operator actions and directives affect plant and system conditions. (SGTS)

HC ILT 2012 NRC Administrative Topics Outline Form ES-301-1 Facility:

Hope Creek Examination Level: Administrative Topic (See Note) Conduct of Operations Conduct of Operations Equipment Control Radiation Control Emergency Plan RO Type Code* S,D S,D,P S,M S,M Date of Examination:

3/5/2012 D SRO Operating Test Number: NRC 2012 Describe activity to be performed 2.1.31 ZZ024 Perform power distribution lineup. 2.1.18 ZZ016 Complete the Daily Logs (Complete Att 1A for 609,611, MSLRMS) (2009 NRC) 2.2.40 ZZ011 Re-start Reactor Recirc Pump lAW Attachment

2. 2.3.5 ZZ019 Calculate Noble Gas Release Rate. N/A NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.

• Type Codes & (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 ) {P)revious 2 exams (S;1; randomly selected)

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HC ILT 2012 NRC Administrative Topics Outline Form ES-301-1 Facility:

Creek Date of Examination:

3/5/2012 Examination Level: 0 RO SRO o perating Test Number: _:....::N:....::R:....;:C_2_0:....::1:....::2_

Administrative Topic

• Type Describe activity to be performed (See Note) I Code"

• Conduct of Operations

• R,N i 2.1.25 ZZ045 Perform On-Line Risk Controls Evaluation Conduct of Operations R,D,P 2.1.18 ZZ017 Review DL-26 (2009 NRC) 2.2.12 ZZ027 Review OP-IS.ZZ-0003 for Completeness Equipment Control R,M and Compliance with Acceptance Criteria.

Radiation Control R,D 2.3.6 ZZ003 Approve Containment Purge permit.

• 2.4.38 ECG003 Utilize the ECG to Classify an Event Emergency Plan R,M (Barrier Table General Emergency/PAR)

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

• Type Codes & (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 (P)revious 2 exams (::;1; randomly selected) (A)lternate Path I I Control Room/In-Plant Systems Outline Form ES-301-2 Facility:

Hope Creek Date of Examination:

Exam Level: RO [8] SRO-I 0 SRO-U 0 Operating Test No.: Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF) System / JPM Title a. AE004 Respond To Rising Drywell Pressure (KIA 223001 A2.01) BC015 Transfer Shutdown Cooling to the Standby Shutdown Cooling Loop (KIA 205000 A4.03) c. CG003 Respond to Main Condenser Low Vacuum (KIA 271000 A4.04) d. GS005 Vent To Control Containment Pressure With Suppression Pool Level Less Than 180 Inches (KIA 295024 EA1.19) e. BF011 Respond To An Uncoupled Control Rod (KIA 201003 A2.02) SB010 Respond To A Reactor Protection System Malfunction 212000 ED002 Respond To A Reactor Auxiliary Cooling Malfunction 295018 M2.02) (NRC

• h. AB001 Bypass MSIV Isolation Interlocks With MSIVs Closed/Open (KIA 239001 A2.03) In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U) EG003 Respond To A Safety Auxiliaries Cooling Water (KIA 400000 j. AB003 Respond To A Failed Open Safety Relief Valve (KIA 239002 A2.03) k. PK001 Respond To A Station Blackout (KIA 295003 M 1.04) Type Code* S,A, L, D S, A, L, N Safety Function 2 4 S, M S,A,D, E, L 9 5 S,A,D S,D,EN 1 7 S,A,D, P 8 S,D,E 3 D,R 8 D,E 3 D,E,R 6 @ All RO and SRO-I 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 (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power 1 Shutdown (N)ew or (M)odified from bank including 1 (A) * (P)revious 2 exams (R}CA (S)imulator Criteria for RO I SRO-II SRO-U 4-6 I 4-6 I 2-3 s9/s8/s4 ;::1/;::1/;::1 -I -I ;::1 (control room system) ;::1/::=1/::=1

2/;

:2/::::1 s 3/ s 31 s 2 (randomly selected)

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

Hope Creek Date of Examination:

3/5/2012 Exam Level: RO D SRO-I SRO-U D Operating Test No.: NRC2012 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF) System I JPM Title Type Code* Safety Function a. AE004 Respond To Rising Drywell Pressure (KIA 223001 A2.01) S,A, L, D 2 b. BC015 Transfer Shutdown Cooling to the Standby Shutdown Cooling Loop (KIA 205000 A4.03) S, A. L, N 4 c. CG003 Respond to Main Condenser Low Vacuum (KIA 271000 A4.04) IS, E, M 9 d. GS005 Vent To Control Containment Pressure With Suppression Pool Level Less Than 180 Inches (KIA 295024 EA 1.19) S, A. D, E, L 5 e. BF011 Respond To An Uncoupled Control Rod (KIA 201003 A2.02) S,A,D 1 f. SB010 Respond To A Reactor Protection System Malfunction (KIA 212000 A2.02) S,D,EN 7 g. ED002 Respond To A Reactor Auxiliary Cooling Malfunction (KIA 295018 M2.02) (NRC 2009) S,A,D,P 8 h. NA --I =

• In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U) i. EG003 Respond To A Safety Auxiliaries Cooling Water Malfunction (KIA 400000 A2.01) D,R I 8 j. AB003 Respond To A Failed Open Safety Relief Valve (KIA 239002 A2.03) D,E 3 k. PK001 Respond To A Station Blackout (KIA 295003 M1.04) D,E,R 6 @ All RO and SRO-I 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 1 SRO-I / SRO-U (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power

/ Shutdown (N)ew or (M)odified from bank including 1 (A) (P)revious 2 exams (R)CA (S)imulator 4-6 1 4-6 12-3 :59/:58/:54 2!1/2!1/2!1 -/ -1 2!1 (control room system) 2!1/2!1/2!1 2!2/2!21?:.1

5 3 / :5 3 / :5 2 (randomly selected) 2!1/?:.1/2!1 ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility:

Hope Creek Date of Examination:

3/5/2012 Exam Level: RO 0 SRO-I 0 SRO-U [R] Operating Test No.: NRC2012 i Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF) System 1 JPM Title Type Code* Safety Function ! a. AE004 Respond To Rising Drywell Pressure (KIA 223001 A2.01) S,A,L, ° 2 I b. BC015 Transfer Shutdown Cooling to the Standby Shutdown Cooling . Loop (KIA 205000 A4.03) S,A,L,N 4 c. NA --I d. NA --e. NA -i -f. SB010 Respond To A Reactor Protection System Malfunction (KIA 212000 A2.02) i S,D,EN 7 g. NA --h. NA --* In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U) i. NA --j. AB003 Respond To A Failed Open Safety Relief Valve (KIA 239002 D,E 3 A2.03) k. PK001 Respond To A Station Blackout (KIA 295003 M1.04) D,E,R 6 @ All RO and SRO-I 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 1 SRO-II SRO-U (A)lternate path 4-6/4-6 12-3 (C)ontrol room (D)irect from bank :;;9/:;;8/:;;4 (E)mergency or abnormal in-plant ;:;1/;:;1/;:;1 (EN)gineered safety feature -I -1 ;:;1 (control room system) (L)ow-Power I Shutdown ;:;1/;:;1/;::1 (N)ew or (M)odified from bank including 1 (A) ;:: 2/?:. 2/;:: 1 (P)revious 2 exams :s 3 / :s 3 / :s 2 (randomly selected)

! (R)CA ;:;1/?:.1/;::1

• (S)imulator il I I Appendix 0 Scenario Outline Form ES-O-1 Facility:

Hope Creek Scenario No.: J.... Op-Test No.: NRC2012 Examiners:

Operators: (SRO) (ATC) (BOP) Initial Conditions:

93% power. Turnover:

Raise reactor power to 98% per Load Dispatcher request. Event No. 1 Malf. No. N/A 2 MS09A 3 CD10A 4 PC07A ED16 5 RR31A1 6 7 PC07B EG12 DG08B DG02A DG02C DG02D HP01 HP06M RC02 RC05 Event R N I C C C C M Raise power to 98% with recirculation flow. PT-N076A MSL Pressure Fails Upscale (TS) "AI> CRD Pump Trip OBE Earthquake wi 10A403 Bus Fault & Lockout (TS) Small break LOCA / Manual Scram Aftershock wi LOP, Main Generator Lockout, "B" EDG Start Failure (recoverable), "A" &"0" EDG fail resulting in unrecoverable loss of 10A401 & 10A404 Buses I C (BOP) HPCI & RCIC auto start failure (RCIC recoverable) i I * (N}ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor APPENDIX 0, Page 38 of 39 Appendix 0 Scenario Outline Form ES-O-1 Facility:

Hope Creek Scenario No.:.£ Op-Test No.: NRC2012 Examiners:

Operators: (SRO) (ATe) (BOP) Initial Conditions:

84.5% power Turnover:

Power ascension in progress.

Raise power 84.5% to 90% using control rods. Place C RFPT in service. il Lve' ,v,alf. Event Type* Event No. No. Description 1 NA R (ATC) . Raise power 84.5% to 90% using control rods. N (BOP)

• Place C RFPT in service. N (SRO) I 2 CD032631 C (ATe) Stuck Control Rod. (TS SRO) C (SRO) 3 NM12B I (ATe) Flow Unit Fails Downscale w/half scram. (TS SRO) I (SRO) 4 TC07A C (ATC) A EHC Pump trip C (SRO) 5 TC16 C (All) Loss of EHC due to Filter Clogging wI Manual Scram i I 6 RP07 M (All) ATWS i 7 CU11A C (ATC) Failure of RWCU to auto isolate. CU11B 8 HP06E C (BOP) HPCI components failure to auto initiate HP14 il HP15 i HP16 i * (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor APPENDIX D, Page 38 of Appendix Scenario Outline Form ES-D-1 Facility:

Hope Creek Scenario No.: 4-LP Op-Test No.: NRC2012 Examiners: Initial Conditions:

3% power. Turnover:

Continue Reactor Startup using control Swap SSW pump alignment to remove D SSW Pump from service for Event Type* Event Malf. Description No. 1 NA Raise Reactor power with control rods. R (ATC) N (SRO) I 2 CD022603 Rod drifts out. (TS SRO) C (ATC) C (SRO) Swap Service Water Pumps 3 NA N (BOP) 4 CW05A C (BOP) Service Water Pump Malfunction (TS SRO) C (SRO) i 5 Single Reactor Recirc Pump Runaway (TS SRO) RR08B C (ATC) Recirc Pump Vibrations I I C(SRO) CR01 C (ALL) Fuel Failure With Scram 6 7 PC06 M (ALL) Torus Leak/Emergency Depressurization 8 RH03B RHR HX inlet valve F047B fails closed C (BOP) * (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor APPENDIX D, Page 38 of 39