ML13158A172

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Draft - Outlines (Folder 2)
ML13158A172
Person / Time
Site: Nine Mile Point Constellation icon.png
Issue date: 03/22/2013
From:
Nine Mile Point
To: Todd Fish
Operations Branch I
Jackson D
Shared Package
ML12356A123 List:
References
TAC U01861
Download: ML13158A172 (27)
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Text

ES-401 BWR Examination Outline Form ES-401*1 II Nine Mile Point Unit 1 Date of Exam: May 2013 RO KIA Category Points SRO-Only Points Tier Group K K K K K K A A A A G A2 G*

Total

1.

Emergency &

Abnormal Plant 1

2 1 2 3 r;r;=r 4 2

0 2

4 5

N/A 6

1 4

1 2

3 1

3 4 N/A 4

1 Total 20 7

3 2

4 1

7 3

Evolutions Tier Totals 4

3 6

5 4

5 27 5

5 10 1

3 2

3 4

3 0

2 3

3 1

2 3

Plant 2

0 0

0 2

1 2

2 3

Systems

!l~

Tier Totals 3

2 3

61~

3 4

38 3

5 8

2.

00:

2

3. Generic Knowledge and Abilities 1

2 3

4 10 1

2 3

4 7

Categories 3

3 2

2 II 2 1

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 (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.

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 2

Form ES-401*1 ES-401 BWR Examination Outline Form ES-401-1 E/APE # I Name I Safety Function 295026 Suppression Pool High Water Temperature 15 295030 Low Suppression Pool Water Levell 5

295003 Partial or Complete Loss of AC Power/6 295005 Main Turbine Generator Trip I 3 295023 Refueling Accidents / 8 295016 Control Room Abandonment 17 600000 Plant Fire On Site / 8 295001 Partial or Complete Loss of Forced Core Flow Circulation /1 & 4 295003 Partial or Complete Loss of AC Power 16 295004 Partial or Complete Loss of DC Power I 6 295005 Main Turbine Generator Trip / 3 295006 SCRAM /1 295016 Control Room Abandonment 17 295018 Partial or Complete Loss of CCW I 8 295019 Partial or Complete Loss of Instrument Air 18 295021 Loss of Shutdown Cooling 14 K

K K ~8G 1

2 3

X X

X X

X X

X X

X X

X X

X X

X X

/SRO)

KIA Topic(s)

! IR 2.1.25 Ability to interpret reference materials, such 4.2 76

, etc.

2.1.23 Ability to perform specific system and 4.4 77 integrated plant procedures during all modes of plant operation.

AA2.04 Ability to determine and/or interpret the 3.7 78 following as they apply to PARTIAL OR COMPLETE LOSS OF AC. POWER: System lineups AA2.08 Ability to determine and/or interpret the 3.9 79 following as they apply to MAIN TURBINE GENERATOR TRIP: Reactor power AA2.03 Ability to determine and/or interpret the 80 following as they apply to REFUELING ACCIDENTS: Airborne contamination levels 2.1.7 - Ability to evaluate plant performance and 4.7 81 make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

2.4.11 - Knowledge of abnormal condition 4.2 82 procedures.

AK3.03 Knowledge of the reasons for the following 2.8 1

responses as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: Idle loop flow AA1.01 - Ability to operate and/or monitor the 3.7 2

following as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER: A.C. electrical distribution system AK2.02 - Knowledge of the interrelations between 3.0 3

PARTIAL OR COMPLETE LOSS OF D.C. POWER and the foliowinQ: Batteries AK2.08 - Knowledge of the interrelations between 3.2 4

MAIN TURBINE GENERA TOR TRIP and the following: A.C. electrical distribution 2.4.6 - Knowledge of EOP mitigation strategies.

AA1.08 - Ability to operate andlor monitor the 4.0 6

following as they apply to CONTROL ROOM ABANDONMENT: Reactor pressure AA2.02 - Ability to determine and/or interpret the 3.1 7

following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER:

Cooling water temperature AA1.01 - Ability to operate andlor monitor the 35 8

following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR: Backup air supply AA2.06 - Ability to determine andlor interpret the 3.2 9

following as they apply to LOSS OF SHUTDOWN COOLING: Reactor pressure

295023 Refueling Accidents 18 295024 High Drywell Pressure I 5 295025 High Reactor Pressure 13 295026 Suppression Pool High Water Temperature I 5 295028 High Drywell Temperature /5 295030 Low Suppression Pool Water Levell 5

295031 Reactor Low Water Levell 2 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown I 1 295038 High Off-site Release Rate 19 600000 Plant Fire On Site 18 700000 Generator Voltage and Electric Grid Disturbances I 6 KIA Category Totals:

X X

X X

X X

X X

X X

r:E 2

3 4

4 3 4

I

/

3 4

AK2.02 - Knowledge of the interrelations between 2.9 10 REFUELING ACCIDENTS and the following: Fuel pool cooling and cleanup system EA2.02 - Ability to determine andlor interpret the 3.9 11 following as they apply to HIGH DRYWELL PRESSURE: Drywell temperature 2.1.27 - Knowledge of system purpose and/or 3.9 12 function.

EA1.03 - Ability to operate andlor monitor the 3.9 13 following as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Temperature monitoring EK1.02 - Knowledge of the operational implications 2.9 14 of the following concepts as they apply to HIGH DRYWELL TEMPERATURE: Equipment environmental qualification 2.4.1 - Knowledge of EOP entry conditions and 4.6 15 immediate action steps.

EK1.01 - Knowledge of the operational implications 4.6 16 of the following concepts as they apply to REACTOR LOW WATER LEVEL: Adequate core cooling EK3.01 - Knowledge of the reasons for the following 4.1 17 responses as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: Recirculation pump trip/runback: Plant-Specific EK3.03 - Knowledge of the reasons for the following 3.7 18 responses as they apply to HIGH OFF-SITE RELEASE RATE: Control room ventilation isolation:

Plant-Specific AK3.04 - Knowledge of the reasons for the following 2.8 19 responses as they apply to PLANT FIRE ON SITE:

Actions contained in the abnormal procedure for plant fire on site 2.4.45 - Ability to prioritize and interpret the

'u significance of each annunciator or alarm.

Group Point Total:

20/7

ES-401 3

Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 1">1.

2 (RO/SRO)

E/APE # I Name / Safety Function 295015 Incomplete SCRAM /1 K

1 K

2 K

3 A

1 A

2 G

X KIA Topic(s) 2.4.8 - Knowledge of how abnormal operating procedures are used in conjunction with EOPs. §;

295036 Secondary Containment High Sump/Area Water Levell 5 X

EA2.01 Ability to determine and/or interpret the following as they apply to SECONDARY CONTAINMENT HIGH SUMPfAREAWATER LEVEL:

IlfJonents within the affected area 3.2 84 295007 High Reactor Pressure I 3 X

AA2.03 Ability to determine and/or interpret the following as they apply to HIGH REACTOR PRESSURE: Reactor water level 3.7 85 295010 High Drywell Pressure / 5 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 21 295012 High Drywell Temperature 15 X

AK1.01 Knowledge of the operational implications of the following concepts as they apply to HIGH DRYWELL TEMPERATURE: Pressureltemperature relationship 3.3 22 295013 High Suppression Pool Temperature I 5 X

AK3.02 Knowledge of the reasons for the following responses as they apply to HIGH SUPPRESSION POOL TEMPERATURE: Limiting heat additions 3.6 23 295014 Inadvertent Reactivity Addition /1 X

AK3.02 Knowledge of the reasons for the following responses as they apply to INADVERTENT REACTIVITY ADDITION: Control rod blocks 3.7 24 295020 Inadvertent Containment Isolation /

5&7 X

AK1.02 Knowledge of the operational implications of the following concepts as they apply to INADVERTENT CONTAINMENT ISOLATION:

Power/reactivity control 3.5 25 295033 High Secondary Containment Area Radiation Levels / 9 X

EA1.02 - Ability to operate and/or monitor the following as they apply to HIGH SECONDARY CONTAINMENT AREA RADIATION LEVELS: Process radiation monitoring system 3.7 26 500000 High Containment Hydrogen Concentration I 5 X

EA2.01 Ability to determine and / or interpret the following as they apply to HIGH PRIMARY CONTAINMENT HYDROGEN CONCENTRATIONS:

Hydr stem availability 27 KIA Category Point Totals:

2 0

2 1

Group Point Total:

713

ES*401 4

Form ES-401*1 I

-~~.v BWR Examination Outline F

PI.",m21Gro" 1 IRO I SRO)

I System # / Name K

K K

K A

G KIA Topic(s)

IR 1

2 3 456 4

207000 Isolation (Emergency)

X 2.4~ 18 - Knowledge of the specific bases for 4.0 86 Condenser EOPs.

215005 APRM I LPRM X 2.4.47 - Ability to diagnose and recognize 4.2 87 trends in an accurate and timely manner utilizing the appropriate control room reference material.

206000 HPCI X

A2.04 - Ability to (a) predict the impacts of the 3~0 88 following on the HIGH PRESSURE COOLANT INJECTION SYSTEM; and (b) based on those predictions, use procedures to correct, control. or mitigate the consequences of those abnormal conditions or operations: A.C. failures: BWR-2,3,4 212000 RPS X

A2.03 - Ability to (a) predict the impacts of the 3.5 89 following on the REACTOR PROTECTION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Surveillance testing 211000 SLC X 22.38 - Knowledge of conditions and 4.5 90 i

limitations in the facility license.

n 205000 Shutdown Cooling X

A1.01 - Ability to predict and/or monitor 3.3 28 changes in parameters associated with operating the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) controls including: Heat exchanger cooling flow 205000 Shutdown Cooling X

K4.01 - Knowledge of SHUTDOWN 3.4 29 COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) design feature(s) and/or interlocks which provide for the following:

High temperature isolation: Plant-Specific 206000 HPCI X 2.4.46 - Ability to verify that the alarms are 4.2 30 consistent with the plant conditions.

207000 Isolation (Emergency)

X A 1.05 - Ability to predict and/or monitor 4.0 31 Condenser changes in parameters associated with operating the ISOLATION (EMERGENCY)

CONDENSER controls including: Reactor pressure: BWR-2,3 209001 LPCS X

K2.01 - Knowledge of electrical power 3.0 32 supplies to the following: Pump power 211000 SLC X

A2.02 - Ability to (a) predict the impacts of the 3~6 33 following on the STANDBY LIQUID CONTROL SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Failure of explosive valve to fire 212000 RPS X

K5.02 - Knowledge of the operational 3.3 34 implications of the following concepts as they apply to REACTOR PROTECTION SYSTEM:

Specific logic a

2150031RM 215004 Source Range Monitor 215004 Source Range Monitor 215005 APRM / LPRM 218000 ADS 223002 PCIS/Nuclear Steam Supply Shutoff 223002 PCIS/Nuclear Steam Supply Shutoff 239002 SRVs 259002 Reactor Water Level Control 261000 SGTS 262001 AC Electrical Distribution 262002 UPS (ACIDC) 262002 UPS (AC/DC)

Ixl X

X X

X X

X X

X X

X X

X A3,02 - Ability to monitor automatic 3,3 35 operations of the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM including:

Annunciator and alarm signals K1,02 - Knowledge of the physical 3.4 36 connections and/or cause-effect relationships between SOURCE RANGE MONITOR (SRM)

SYSTEM and the following: Reactor manual control A3,01 - Ability to monitor automatic 3,2 37 operations of the SOURCE RANGE MONITOR (SRM) SYSTEM including: Meters and recorders K3,01 - Knowledge of the effect that a loss or 4,0 38 malfunction of the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM will have on following:

RPS K5.01 - Knowledge of the operational 3.8 39 implications of the following concepts as they apply to AUTOMATIC DEPRESSURIZATION SYSTEM: ADS logic operation K4.03 - Knowledge of PRIMARY 3.5 40 CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF design feature(s) and/or interlocks which provide for the following:

Manual initiation capabilitv: Plant-Specific K1.14 - Knowledge of the physical 2.8 41 connections and/or cause-effect relationships between PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF and the following:

Containment drainage system K2.01 - Knowledge of electrical power 2.8 42 supplies to the following: SRV solenoids K5.01 - Knowledge of the operational 3.1 43 implications of the following concepts as they apply to REACTOR WATER LEVEL CONTROL SYSTEM:

GEMAC/Foxboro/Bailey controller operation:

Plant-Specific K3.02 - Knowledge of the effect that a loss or 3,6 44 malfunction of the STANDBY GAS TREATMENT SYSTEM will have on following: Off-site release rate A4.01 - Ability to manually operate and/or 3.4 45 monitor in the control room: All breakers and disconnects (including available switch yard):

Plant-Specific K3.14 - Knowledge of the effect that a loss or 2.8 46 malfunction of the UNINTERRUPTABLE POWER SUPPLY (AC.lD.C.) will have on following: Rx power: Plant-Specific K4.01 Knowledge of UNINTERRUPTABLE 3,1 47 POWER SUPPLY (AC.lD,C,) design feature(s) and/or interlocks which provide for the following: Transfer from preferred power to alternate po lies

263000 DC Electrical Distribution X

A2.01 - Ability to (a) predict the impacts of the following on the D.C. ELECTRICAL DISTRIBUTION; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Grounds 2.8 48 263000 DC Electrical Distribution X 2.2.36 - Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

3.1 49 264000 EDGs X

A2.09 - Ability to (a) predict the impacts of the following on the EMERGENCY GENERATORS (DIESEUJET); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Loss of AC. power 3.7 50 264000 EDGs X

K1.06 - Knowledge of the physical connections and/or cause-effect relationships between EMERGENCY GENERATORS 3.2 51 (DIESEUJET) and the following: Starting system 300000 Instrument Air X

K4.02 - Knowledge of (INSTRUMENT AIR SYSTEM) design feature(s) and or interlocks which provide for the following: Cross-over to other air systems 3.0 52 400000 Component Cooling Water X

A3.01 - Ability to monitor automatic operations of the CCWS including: Setpoints on instrument signal levels for normal operations, warnings, and trips that are applicable to the CCWS 3.0 53 KIA Category Point Totals:

3 2

3 4

W 3

0 2

3 I

2 3

1 2 I

3 Group Point Total:

2615

ES-401 5

Form ES-401-1 BWR Examination Outline r::=

Plant Systems Tier 2/Group 2 (RO I SRO)

System # f Name K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G

KIA Topic(s)

I 202002 Recirculation Flow Control X 2.2.22 Knowledge of limiting conditions for operations and safety limits.

201001 CRD Hydraulic X

A2.14 - Ability to (a) predict the impacts of the following on the CONTROL ROD DRIVE HYDRAULIC SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Low drive header pressure 2.8 92 233000 Fuel Pool Cooling/Cleanup X 2.4.31 - Knowledge of annunciator alarms, indications, or response procedures.

4.1 93 201001 CRD Hydraulic X

A4.06 - Ability to manually operate and/or monitor in the control room: SDV isolation valve test switch 2.8 54 201002 RMCS X 2.1.20 - Ability to interpret and execute procedure steps.

4.6 55 201006 RWM X

K4.02 - Knowledge of ROD WORTH MINIMIZER SYSTEM (RWM) (PLANT SPECIFIC) design feature(s) and/or interlocks which provide for the following:

Withdraw blocks/errors: P-Spec (Not BWR6) 3.5 56 202001 Recirculation X

A 1.04 - Ability to predict and/or monitor changes in parameters associated with operating the RECIRCULATION SYSTEM controls including: Reactor water level 3.3 57 216000 Nuclear Boiler Instrumentation X

K6.01 Knowledge of the effect that a loss or malfunction of the following will have on the NUCLEAR BOILER 3.1 58 INSTRUMENTATION: A.C. electrical distribution 219000 RHRlLPCI: ToruS/Pool Cooling Mode I !

X A4.02 Ability to manually operate and/or monitor in the control room:

Valve lineup 3.7 59 226001 RHRlLPCI: Containment Spray Mode X 2.4.20 - Knowledge of the operational implications of EOP warnings, cautions, and notes.

3.8 60 239001 Main and Reheat Steam 271000 Offgas Ix X

I A2.11 - 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, control, or mitigate the consequences of those abnormal conditions or operations:

Steam line break K6.11 - Knowledge of the effect that a loss or malfunction of the following will have on the OFF GAS SYSTEM:

4.1 3.2 61 62 Condenser vacuum 288000 Plant Ventilation X

K4.03 - Knowledge of PLANT VENTILATION SYSTEMS design feature(s) and/or interlocks which provide for the following: Automatic starting and st fans 2.8 63

290001 Secondary Containment X

K6.08 - Knowledge of the effect that a loss or malfunction of the following will have on the SECONDARY CONTAINMENT: Plant air systems 2.7 64 290003 Control Room HVAC X

K5.03 - Knowledge of the operational implications of the following concepts as they apply to CONTROL ROOM HVAC:

re control 2.6 65 KIA Category Point Totals:

0 0

0 2

1 3

1 1

I 1

0 2 2 I

2 Group Point Total:

12 f 3

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

Form ES-401-3 Date of Exam:

Category KIA #

Topic RO SRO-Only IR IR 2.1.6 Ability to manage the control room crew during plant transients.

4.8 94

1.

Conduct 2.1.20 Ability to i,Ill:"II.JII:a and execute procedure steps.

4.6 95 of Operations 2.1.32 Ability to explain and apply system limits and precautions.

3.8 66 2.1.14 Knowledge of criteria or conditions that require plant-wide announcements, such as pump starts, reactor trips, mode 3.1 67 vi IQII~"'<>, etc.

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

4.4 68 Subtotal 3 ~

2 2.2.6 Knowledge of the process for making changes to procedures.

3.6 96

2.

Equipment Control 2.2.2 Ability to manipulate the console controls as required to operate the facility between shutdown and designated power levels.

4.6 69 2.2.20 Knowledge of the process for managing troubleshooting activities.

2.6 70 2.2.13 Kr of tagging and clearance procedures.

4.1 71 Subtotal 3

1 2.3.5 Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey 2.9 97 instruments, tJ",I<>VIlIIt:a monitoring equipment, etc.

3.

Radiation Control 2.3.4 Knowledge of radiation exposure limits under normal or

" ""~"'"<-y conditions.

3.7 98 2.3.7 Ability to comply with radiation work permit requirements during normal or abnormal conditions.

3.5 72 2.3.11 Ability to control radiation releases.

3.8 73 Subtotal 2 -

2 2.4.28 Knowledge of procedures relating to a security event (flv, -<>dfe::yuards information).

4.1 99

4.

Emergency Procedures I 2.4.44 Knowledge of emergency plan protective action recommendations.

4.4 100 Plan 2.4.14 Kn

.of general guidelines for EOP usage.

3.81 74 2.4.30 Knowledge of events related to system operation/status that must be reported to internal organizations or external 2.7 75 agencies, such as the State, the NRC, or the U QI '<>1 '"<><>'V' system vtJ"" CllU' SUbtotal 2.2 Tier 3 Point Total 10

ES-401 Record of Rejected KlAs Form ES-401-4 1 /1 1 /2 2/1 2/1 2/1 2/2 2/2 2/2 2/2 2/2 G

Randomly Selected KIA 295027 High Containment Temperature 295011 High Containment Temp 203000 RHR/LPCI: Injection Mode 209002 HPCS 217000 RCIC 201004 RSCS 201005 RCIS 215002 RBM 230000 RHRlLPCI: Torus/Pool Spray Mode 239003 MSIV Leakage Control 2.2.3 Knowledge of the deSign, procedural, and operational differences between units.

Reason for Rejection This topic applies to plants with Mark III containments only. The facility has a Mark I containment.

This topic applies to plants with Mark III containments only. The facility has a Mark I containment.

This system is not installed at the facility.

This system is not installed at the facility.

This system is not installed at the facility.

This system is not installed at the facility.

This system is not installed at the facility.

This system is not installed at the facility.

This system is not installed at the facility.

This system is not installed at the facility.

This KIA applies to multi-unit facilities only.

2.2.4 Ability to explain the variations in control board/control room layouts, systems, instrumentation, and G

procedural actions between units at a facility.

Question 29 205000 K4.02 - Knowledge of SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN 2/1 COOLING MODE) design feature(s) and/or interlocks which provide for the following:

High pressure isolation Question 51 264000 K 1.02 - Knowledge of the physical connections and/or cause-effect relationships between EMERGENCY GENERATORS (DIESEUJET) 2/1 and the following: D.C.

electrical distribution Question 82 600000 Plant Fire On Site 2.1.19 - Ability to use plant 1 /1 computers to evaluate system or component status.

This KIA applies to multi-unit facilities only.

The facility does not have a high pressure isolation for Shutdown Cooling.

Randomly re-selected KIA 205000 K4.01 Knowledge of SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) design feature(s) and/or interlocks which provide for the following: High temperature isolation: Plant-Specific The KIA overlaps with Question 49 (263000 DC Electrical Distribution - 2.2.36 - Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.) and DC Electrical Distribution is heavily sampled on the written exam.

Randomly re-selected KIA 264000 K1.06 Knowledge of the physical connections and/or cause-effect relationships between EMERGENCY GENERATORS (DIESEUJET) and the following: Starting system An operationally relevant question at the appropriate license level could not be written due to lack of relationship between Plant Fire On Site and use of the Plant Process Computer.

Randomly re-selected Generic KIA 2.4.11 Knowledge of abnormal condition procedures.

Question 79 295005 Main Turbine Generator Trip AA2.08 - Ability to determine 1 /1 and/or interpret the following as they apply to MAIN TURBINE GENERATOR TRIP: Electrical distribution status An operationally relevant question at the appropriate license level could not be written for the randomly sampled KIA.

Randomly re-sampled KIA 295005 AA2.05 Ability to determine and/or interpret the following as they apply to MAIN TURBINE GENERATOR TRIP: Reactor power.

Administrative Topics Outline Facility: Nine Mile Point Unit 1 Examination Level: RO Date of Examination: May 2013 Operating Test Number: LC1 11-01 Administrative Topic (see Note)

Type Code'"

Describe activity to be performed Conduct of Operations P,D,S 2010 NRC Perform Reactor Water Level Instrument Checks N1-ST-DO, KIA 2.1.18 (3.6)

Conduct of Operations N,R Perform Heat-up Rate Determination N1-0P-43A Attachment 18, KIA 2.1.7 (4.4)

Equipment Control M,R Explain RPS Manual Scram Circuit Using Prints C-19859-C, KIA 2.2.41 (3.5)

Conduct Alert Emergency Announcement and Evacuation Emergency Procedures/Plan N,S EPIP-EPP-18, KIA 2.4.43 (3.2)

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 (:'5 3 for ROs; :'54 for SROs & RO retakes)

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

(P)revious 2 exams (:'5 1; randomly selected)

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Nine Mile Point Unit 1 Examination Level: SRO Administrative Topic (see Note)

Conduct of Operations Conduct of Operations Equipment Control Radiation Control Emergency Procedures/Plan Date of Examination: May 2013 Operating Test Number: LC1 11-01 Type Code*

Describe activity to be performed M,R Review Reactor Water Level Instrument Checks and Determine Technical Specification Impact N1-ST-DO, KIA 2.1.7 (4.7)

D,R Determine Reportability Requirements for Loss of Offsite Power with EDG Failure CNG-NL-1.01-1004, NUREG 1022, KIA 2.1.18 (3.8)

M,R Explain RPS Manual Scram Circuit Using Prints and Determine Technical Specification Requirements for RPS Manual Scram Pushbutton C-19859-C, KIA 2.2.41 (3.9)

N,R Determine Actions for Inoperable Stack Radiation Monitor ODCM, KIA 2.3.11 (4.3)

N,S Post-Scenario Emergency Event Classification EPIP-EPP-18, KIA 2.4.29 (4.4)

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 Control Room/In-Plant Systems Outline Fonn ES-301-2 Facility: Nine Mile Point Unit 1 Date of Examination: May 2013 Exam Level: RO/SRO-IISRO-U Operating Test No.: LC1 11-01 Control Room Systems@ (B for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System 1JPM Title Type Code" Safety Function

a. Rapid RWCU System Restoration for Level Control P,D, L,S 2

KIA 204000 A4.06 (3.0/2.9), N1-EOP-HC 2010 NRC

b. Cool RBEVS Charcoal Filter P, D, L, EN, S 9

KIA 261000 A4.03 (3.0/3.0), N1-0P-10 2009 NRC

c. Re-Open MSIVs With Reactor Pressurized and Lower Pressure N,L,S 3

KIA 239001 A4.01 (4.2/4.0), N1-EOP-HC

d. Transfer Torus Water to WeT D,EN,S 5

KIA 295029 EA1.03 (2.9/3.0), N1-EOP-1

e. Control Room Evacuation. Manual Scram Fails, ARI Works, One Main Steam Line Fails to Isolate, Powerboard 11 Fails to M,A,EN.S 7

Fast Transfer KIA 295016 AA 1.01 (3.B/3.9), N1-S0P-21.2

f. Restore Emergency Condenser To Service D,A,EN,S 4

KIA 207000 A4.05 (3.5/3.7), N1-0P-13

g. EDG 103 SID - PB 103 Return to Normal Power D,A,S 6

KIA 264000 A4.04 (3.7/3.7), N1-0P-45

h. Reset Reactor Scram (RO Only)

M, L, S 1

KIA 295006 AA1.07 (4.1/4.1), N1-S0P-1 I In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

i. Initiate Emergency Condenser Locally M,A,L,E,R 4

KIA 207000 A2.08 (3.8/3.8), N1-0P-13

j. Respond to RBCLC Makeup Tank Level Alarm D,A,E,R 8

KIA 295018 AA2.04 (2.9/2.9), N1-ARP-H1

k. Lineup to Flood the Reactor Vessel Using the Diesel Fire Pump D,E,R 2

KIA 295031 EA1.0B (3.B/3.9)

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 I Shutdown (N)ew or (M)odified from bank including 1 (A)

(P)revious 2 exams

{R)CA (S)imulator Pairings:

AandB CandD E then F G alone H, possibly with RO admin JPM Criteria for RO 1SRO-II SRO-U 4-6 / 4-61 2-3

$9/$8/$4 2:1/2:1/2:1

- / - / 2:1 (control room system) 2:1/2:1/2:1 2:2/'ii!2/'ii!1

$ 3/ $ 3/ $ 2 (randomly selected)

'ii!1/2:1/2:1

Appendix D Scenario Outline Form ES-D-1 Facility: Nine Mile Point Unit 1 Scenario No.: NRC-1 Op-Test No.: LC1 11-01 Examiners:

Operators:

Initial Conditions: The plant is operating at approximately 95% power. Containment Spray pump 112 is out of service for maintenance. Steam Packing Exhauster 12 is caution tagged due to high vibrations.

Turnover: Secure Reactor Recirculation pump 12. Maintain Reactor power near the initial level while securing the pump.

Event Event

~

Type*

Description 1

N/A N

BOP, SRO Secure Reactor Recirculation Pump 12 While Maintaining Reactor Power R-ATC N 1-0P-1, Technical Specifications 2

FW02A Override C

BOP, SRO Feedwater Booster Pump 11 Trips with Failure of Feedwater Booster Pump 13 to Auto-start N1-S0P-1S.1, Technical Specifications Powerboard 11 Electrical Fault 3

ED04 C-AII N1-S0P-30.1, N1-S0P-1.3, N1-S0P-1.1, Technical Specifications 4

EC01 M-AII Steam Leak in Primary Containment N1-S0P-1, N1-EOP-2, N1-EOP-4, N1-EOP-8 5

PC10A C-AII Torus to Drywell Vacuum Breaker Inadvertently Opens PC10C N1-EOP-4 FW28A S

FW28B FWOS C

BOP, SRO HPCI Fails to Auto-Initiate, Feedwater Pump 13 Disengages, and Core Spray Valves Fail to Auto-Open N1-EOP-2 CSO?

?

Overrides C

ATC, SRO Partial Primary Containment Isolation Failure N1-S0P-40.2 8

CT01A C-AII Containment Spray Pump 111 Trips N1-EOP-4 (N)ormal (R )eactivity, (I )nstrument, (C)omponent (M)ajor

Facility: Nine Mile Point Unit 1 Scenario No.: NRC*1 Op-Test No.: LC111-01

1. Total malfunctions (5-8) 7 Events 2, 3,4, 5, 6, 7, 8
2. Malfunctions after EOP entry (1-2) 4 Events 5, 6, 7, 8
3. Abnormal events (2-4) 2 Events 2 &3
4. Major transients (1-2) 1 Event 4
5. EOPs entered/requiring substantive actions (1-2) 2 N1*EOP*2, N1-EOP-4
6. EOP contingencies requiring substantive actions (0-2) 1 N1*EOP*8
7. Critical tasks (2-3) 2 CRITICAL TASK DESCRIPTIONS:

CT*1 - Given a LOCA in the Drywell and a failure of HPCI to initiate, inject with preferred and alternate injection systems to restore and maintain RPV water level above -84 inches, in accordance with N1-EOP

2.

CT*2 - Given a LOCA in the Drywell and degraded Contai nment Spray capability, execute N1-EOP-8, RPV Blowdown, when it is determined Torus pressure cannot be maintained below the Pressure Suppression Pressure limit, in accordance with N1 EOP-4.

Appendix D Scenario Outline Form ES-D-1 Facility: Nine Mile Point Unit 1 Scenario No.: NRC-2 Op-Test No.: LC1 11-01 Examiners:

Operators:

Initial Conditions: The plant is operating at approximately 55% power. Containment Spray pump 112 is out of service for maintenance. Steam Packing Exhauster 12 is caution tagged due to high vibrations.

Circulating Water pump 11 was returned to service earlier in the shift following waterbox cleaning.

Turnover: Raise Reactor power with Recirculation flow.

Event Malf.

Event Event No.

No.

Type*

Description R-Raise Reactor Power With Recirculation Flow 1

N/A

ATC, SRO N1-0P-43B, N1-0P-1 2

NM36A I-SRO Recirculation Flow Unit Fails Upscale ARP F2-2-6, Technical Specifications 3

Report C

ATC, SRO Power Control Requests Emergency MVAR Support OP-32 4

EC03B I-BOP.

SRO Emergency Condenser 12 Inadvertent Initiation ARP K1-1-5. N1-0P-13, Technical Specifications 5

CW17 C-AII Intake Grassing N1-S0P-1B.1. N1-S0P-1.1 6

FW14C I-BOP, SRO Feedwater Level Control Fails As-is in Automatic N1-S0P-16.1 7

EC02 M-AII Emergency Condenser 11 Steam Leak in Reactor Building ARP K1-1-1, K1-4-3, N1-S0P-1, N1-EOP-2, N1-EOP-5 C

Mode Switch Fails to Scram B

Overrides

ATC, SRO N1-S0P-1 EC07A 9

ECOBA C-AII Emergency Condenser 11 Fails to Isolate N1-EOP-5. N1-EOP-B ECOBB 10 Override I-BOP, SRO Bypass Opening Jack Fails to Open Turbine Bypass Valves N1-EOP-2 (N)ormal, (R)eactivity,

{l}nstrument, (C)omponent, (M)aior

I Facility: Nine Mile Point Unit 1 Scenario No.: NRC-2 Op-Test No.: LC111-01

1. Total malfunctions (5-8) 9
Events 2, 3, 4,5,6,7,8,9,10
2. Malfunctions after EOP entry (1-2) 3 Events 8, 9,10
3. Abnormal events (2-4) 3 Events 4, 5, 6
4. Major transients (1-2) 1 Event 7
5. EOPs entered/requiring substantive actions (1-2) 2 N1-EOP-2, N1-EOP-5
6. EOP contingencies requiring sUbstantive actions (0-2) 1 N1-EOP-8
7. Critical tasks (2-3) 3 CRITICAL TASK DESCRIPTIONS:

CT Given the plant at power with lowering intake water level, remove a Circulating Water pump from service in order to preserve use of the lake as a heat sink, in accordance with N1-S0P-18.1.

CT Given an un-isolable Emergency Condenser leak outside primary containment and one general area temperature above the maximum safe limit, insert a manual reactor scram, in accordance with N1*EOP-5.

CT*3 - Given an un-isolable Emergency Condenser leak outside primary containment and two general area temperatures above the maximum safe limit, execute N1-EOP-8, RPV Blowdown, in accordance with N1-EOP*5.

Appendix 0 Scenario Outline Form ES-O-1 Facility: Nine Mile Point Unit 1 Scenario No.: NRC-3 Op-Test No.: LC111-01 Examiners:

Operators:

Initial Conditions: The plant is operating at approximately 100% power. Containment Spray pump 112 is out of service for maintenance. Steam Packing Exhauster 12 is caution tagged due to high vibrations.

Turnover: Maintain operation at rated power.

Event Malf.

Event Event No.

No.

Type*

Description C

1 RP01A

BOP, SRO RPS MG Set 131 Trip ARP F1-3-7, N1-0P-48, N1-S0P-16.1, N1-S0P-1.1 R-ATC Reactor Pressure Instrument Fails Downscale 2

RP16B I-SRO ARP F4-4-2, Technical Specifications C

CRD Flow Control Valve Fails Closed 3

RD36A

BOP, N1-S0P-5.1, Technical Specifications SRO MPR Setpoint Drifts Low 4

Override I-All N1-S0P-31.2, N1-S0P-1 Failure to Scram 5

RD33 M-AII N1-EOP-2, N1-EOP-3 LP01A Liquid Poison Pumps Trip 6

C-AII LP01B N1-EOP-3, N1-EOP-3.2 C

CRD Pump 12 Trips 7

RD35B

ATe, N1-EOP-3.1, N1-S0P-5.1 SRO (N)ormal.

(R)eactivitv.

(I)nstrument.

(C)omponent.

(M)ajor

Facility: Nine Mile Point Unit 1 Scenario No.: NRC-3 Op-Test No.: LC111-01

1. Total malfunctions (5-8) 7 Events 1, 2, 3, 4, 5, 6, 7
2. Malfunctions after EOP entry (1-2) 2 Events 6 & 7
3. Abnormal events (2-4) 3 Events 1, 3, 4
4. Major transients (1-2) 1 Event 5
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 of the Reactor to scram with power above 6% or unknown and RPV water level above -41 inches, lower power to below 6% by:

  • Terminating and preventing all injection except boron and CRD, in accordance with N1-EOP-3.

CT Given a failure of the Reactor to scram, insert control rods, in accordance with N1*EOP*3.

Appendix 0 Scenario Outline Form ES-D-1 Facility: Nine Mile Point Unit 1 Scenario No.: NRC-4 Op-Test No.: LC1 11-01 Examiners:

Operators:

Initial Conditions: The plant is operating at approximately 100% power. Containment Spray pump 112 is out of service for maintenance. Steam Packing Exhauster 12 is caution tagged due to high vibrations.

Turnover: Remove Line 4 from service.

Event Malf.

Event No.

No.

Description

J N-Remove Line 4 from Service 1

NfA BOP, N 1-0P-33A, Technical Specifications SRO C

BOP, ERV Inadvertently Opens 2

AD05 SRO N1-S0P-1.4, N1-S0P-1.1, Technical Specifications R

ATC C-Powerboard 16A Electrical Fault 3

ED12A BOP, ARP L4-3-6, N1-EOP-4 SRO CW04A All RBCLC Pumps Trip 4

CW04B C-AII N1-S0P-11.1, N1-S0P-1, N1-EOP-2 CW04C FW03A Feedwater Pumps Trip 5

FW03B C-AII N1-EOP-2 FW06 CU01 Coolant Leak Inside Primary Containment 6

M All EC01 N1-EOP-2, N1-EOP-4 Fuel Zone Level Instrument Flashing 7

RR87 I-All N1-EOP-2, N1-EOP-7 (N}ormal, (R)eactivity, (l)nstrument (C)omponent (M)aior

Facility: Nine Mile Point Unit 1 Scenario No.: NRC-4 Op-Test No.: LC111-01

1. Total malfunctions (5-8) 6 Events 2, 3, 4, 5, 6, 7
2. Malfunctions after EOP entry (1-2) 2 Events 5 & 7
3. Abnormal events (2-4) 3 Events 2, 3, 4
4. Major transients (1-2) 1 Event 6
5. EOPs entered/requiring substantive actions (1-2) 2 N1-EOP-2, N1-EOP-4
6. EOP contingencies requiring substantive actions (0-2) 1 N1-EOP-7
7. Critical tasks (2-3) 3 CRITICAL TASK DESCRIPTIONS:

CT Given an inadvertently open ERV at power, close the ERV or insert a manual scram prior to Torus temperature exceeding 110°F, in accordance with N1-S0P-1.4.

CT Given a LOCA in the Drywell, initiate Containment Sprays prior to exceeding the Pressure Suppression Pressure limit, in accordance with N1 EOP-4.

CT Given the plant with RPV water level unknown, execute N1-EOP-7, RPV Flooding, in accordance with N1-EOP-2.

Appendix D Scenario Outline Form ES-D-1 Facility: Nine Mile Point Unit 1 Scenario No.: NRC-5 Op-Test No.: LC1 11-01 Examiners:

Operators:

Initial Conditions: The plant is operating at approximately 5% power during a startup. Steam Packing Exhauster 12 is caution tagged due to high vibrations.

Turnover: Raise Reactor power by withdrawing control rods. Place the Mode Switch in RUN and withdraw the IRMs.

Event Malf.

Event Event No.

No.

Type*

Description R-Raise Reactor Power with Control Rods 1

N/A

ATC, SRO N1-0P-5, N1-0P-43A N

Place Mode Switch in RUN and Withdraw IRMs 2

N/A

BOP, SRO N1-0P-43A Loss of Line 1 3

ED02A C

SRO ARP A8-1-1, Technical Specifications Loss of Condenser Vacuum 4

MC01 C-AII N1-S0P-25.1, N1-S0P-1 Loss of Une 4 5

ED01A C

All N1-S0P-33A.1, Technical Specifications 6

DG04A C

BOP, SRO Emergency Diesel Generator 102 Fails to Automatically Start N1-S0P-33A.1 Loss of Coolant Accident 7

RR29 M-AII N1-EOP-2, N1-EOP-4, N1-EOP-8 8

CS06 C

All Core Spray Fails to Auto-Start N1-EOP-2 9

EC04A C

ATC, Emergency Condensers Fail to Operate from Control Room EC04B SRO N1-EOP-2, N1-EOP-8 (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor

Facility: Nine Mile Point Unit 1 Scenario No.: NRC-5 Op-Test No.: LC111-01 1 Total malfunctions (5-8) 7 Events 3, 4, 5, 6, 7, 8, 9

2. Malfunctions after EOP entry (1-2) 2 Events 8 & 9
3. Abnormal events (2-4) 2 Events 4 &5
4. Major transients (1-2) 1 Event 7
5. EOPs entered/requiring substantive actions (1-2) 2 N1-EOP-2, N1-EOP-4
6. EOP contingencies requiring substantive actions (0-2) 2 N1-EOp-2 Alternate Level Control, N1-EOP-8
7. Critical tasks (2-3) 3 CRITICAL TASK DESCRIPTIONS:

CT Given a LOCA in the Drywell, initiate Containment Sprays prior to exceeding the Pressure Suppression Pressure limit, in accordance with N1 EOP-4.

CT Given Reactor water level below the top of active fuel and Reactor pressure above the capacity of available low pressure systems, perform an emergency depressurization of the Reactor, in accordance with N1-EOP-2 and N1-EOP-8.

CT Given a LOCA in the Drywell and Reactor water level below the top of active fuel, inject with Preferred and Alternate Injection Systems to restore and maintain Reactor water level above *84 inches, in accordance with N1-EOP-2.

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