ML071930241

Final Outlines (Folder 3)

ML071930241
Person / Time
Site:	Seabrook
Issue date:	06/22/2007
From:	St.Pierre G, Wright K Florida Power & Light Energy Seabrook
To:	Caruso J NRC Region 1
Sykes, Marvin D.
Shared Package
ML062050109	List: IR 05000443/2007301 ML070811054 ML071920480 ML071920500 ML071930191 ML071930235 ML071930238 ML071930241 ML071930246 ML071930252 ML071930259 ML072080317 ML072390373 ML072920097 ML072960127 SBK-L-07066, Draft - Outlines and Facility Ltr. Dtd 4/11/07 (Folder 2)
References
50-443/07-301, ES-401-2 50-443/07-301
Download: ML071930241 (37)

Text

, -

Tier

1.

Emergency &

Ab n o r m a I Plant Evolutions

2.

Plant Systems Seabrook Station 2007 NRC Written Exam Outline ES-401, Rev. 9 PWR Examination Outline Form ES-401-2 SRO-Only Points t q Y q - G T RO KIA Category Points Group K

K K

K K

K A

A A

A G

1 2

3 4

5 6

1 2

3 4

Total 1

2 2

4 3

4 3

18 4

2 6

NIA NIA 2

1 1

2 2

2 I

9 2

2 4

TierTotals 3

3 6

5 6

4 27 6

4 I O 1

4 1

4 4

1 2

3 2

3 2

2 28 2

3 5

2 2

1 1

1 1

1 1

1 0

0 1

10 1

2 3

TierTotals 6

2 5

5 2

3 4

3 3

2 3

38 3

5 8

Facility:

lo

3. Generic Knowledge and Abilities 1

2 3

4 4

2 2

2 Categories CCCM 1

ES-401, Rev. 9 Form ES-401-2 Seabrook Station 2007 NRC Exam Tier ?/Group1 RO ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions -Tier IIGtOup 1 (RO I SRO)

UAPE #I Name I Safety Function H

3 -

IR -

4.3 I

2.1.20 Ability to execute procedure steps.

(CFR: 41.10 I43.5 / 45.12) 000007 (BWlE0281E10; CUE02)

Reactor Trip - Stabilization - Recovery I1 000008 Pressurizer Vapor Space Accident I 3

4.1 3.3 2.6 responses as they apply to the Pressurizer Vapor EA2.15 Ability to determine or interpret the following as they apply to a small break LOCA.

RCS parameters 000009 Small Break LOCA I 3 00001 I Large Break LOCA I 3

00001 511 7 RCP Malfunctions I 4 3.5 I

RCP Seal Water Injection Subsystem I I I (CFR 41.7/45.5/45.6) 000022 Loss of Rx Coolant Makeup I 2 3.5 3.4 Actions contained in SOPS and EOPs for RCPs, loss of makeup, loss of charging, and abnormal (CFR 41.5/41.10/45.6/45.13)

(CFR: 41.10 I43.5 I45.13)

, charging.

, 2.4.11, Knowledge of abnormal condition procedures.

oooO25 Loss of RHR System I 4 000026 Loss of Component Cooling Water I 8 000027 Pressurizer Pressure Control System Malfunction I 3 2.8 I

ressure Control Malfunctions:

Expansion of liquids as temperature increases.

as they apply to a ATWS:

Charging Pumps.

(CFR 41.7/45.5/45.6) 000029 ATWS I1 3.4 1

000038 Steam Gen. Tube Rupture I 3

000040 (BWIEOS; CEEOS; WIE12)

Steam Line Rupture - Excessive Heat Transfer I 4

!.6 1

I I I I

Sensors and detectors (CFR 41.7/45.7) 000054 (CUE06) Loss of Main Feedwater I 4 000055 Station Blackout I 6 4.3 I I

I I onsitepower 1

r (CFR 41.5/41.10/45.6/45.13)

Seabrook Static 000057 Loss of Vital AC lnst. Bus I 6 I 000058 Loss of DC Power I 6 000062 Loss of Nuclear Svc Water I 4 000065 Loss of Instrument Air I 8 WIE04 LOCA Outside Containment I 3 WIEl 1 Loss of Emergency Coolant Recirc. I 4 BWIEW, WIE05 Inadequate Heat Transfer - Loss of Secondary Heat Sink I 4 KIA Category Totals:

following as the apply to Loss Of Vital AC Instrument responses as they apply to the Loss Of Nuclear Service The automatic actions (alignments) within the nuclear service water resulting from the actuation of When to commence plant shutdown if instrument Outside Containment:

ssociated with the LOCA operating within the limits of the facilities licensdamendments.

3 4 3

GroupPointTotal:

1.2 c

1.8 c

1.4 1.6 -

1.4 1.5 1.4 1

1 1

1 -

1 1

1 18/6 2

ES-401, Rev. 9 Form ES-401-2 Seabrook Station 2007 NRC Exam Tier IIGroupl SRO E M 1 PWR Examination Outline FOI~ ES401-2 Emergency and Abnormal Plant Evolutions -Tier IlGroup 1 (RO I SRO)

EIAPE # I Name I Safety Function 000007 (BWIEOP&ElO; CEIEO2)

Reactor Trip - Stabilization - Recovery I1 000008 Pressurizer Vapor Space Accident I 3 000009 Small Break LOCA I 3 00001 1 Large Break LOCA I 3

00001 511 7 RCP Malfunctions I 4 000022 Loss of Rx Coolant Makeup I 2 000025 Loss of RHR System I 4 000026 Loss of Component Cooling Water I 8

000027 Pressurizer Pressure Control System Malfunction I 3 000029 ATWS I 1

000038 Steam Gen. Tube Rupture I 3 000040 (BWIE05; CEIEOB; WIEl2)

Steam Une Rupture - Excessive Heat Transfer I 4 000054 (CEIE06) Loss of Main Feedwater I 4 000055 Station Blackout I 6

000056 Loss of Off-site Power I 6

000057 Loss of Vital AC lnst Bus I 6

000058 Loss of DC Power I 6

instrument interpretation.

(CFR. 43.5 I45.12 / 45.13) ns for emergency and 1

I 1

I I

I I I I 4 I 2 I Group PointTotal:

18/6 KIA Category Totals:

2

ES-401, Rev. 9 Form ES-401-2 Seabrook Station 2007 NRC Exam Tier 1IGroup 2 RO ES-401 PWR Examination Outline Form E-1-2 Ememencv and Abnormal Plant Evolutions - Tier 1IGrouD 2 [RO I SROI ElAPE # I Name I Safety Function 000001 Continuous Rod Withdrawal I1 000003 Dropped Control Rod I1 000005 InoperablelStuck Control Rod I I 000024 Emergency Boration I 1

000028 Pressurizer Level Malfunction I 2 000032 Loss of Source Range NI I 7

000033 Loss of Intermediate Range NI I 7

000036 fBWIAO81 Fuel Handlina Accident I 8 000037 Steam Generator Tube Leak I 3

000051 Loss of Condenser Vacuum I 4 000059 Accidental Liquid RadWaste Rel. I 9 000060 Accidental Gaseous Radwaste Rei. I 9 000061 ARM System Alarms I 7

~~

000067 Plant Fire On-site I 8 000068 (BWIAO8) Control Room Evac. I 8 000069 (WIE14) Loss of CTMT Integrity I 5

~

000074 (WIE06&E07) Inad. Core Cooling I 4 000076 High Reactor Coolant Activity I 9 WIEOI & E02 Rediagnosis & SI Termination I 3

WIEI 3 Steam Generator Over-pressure I 4 WIE15 Containment Flooding I 5 Dropped Control Rod:

Actions contained in the EOP for dropped Charging subsystem flow indicator and 1

Seabrook Station 2007 NRC Exam Tier 1IGroup 2 RO (Continued)

WlEl6 High Containment Radiation I 9 BWIAOI Plant Runback I 1

BWIAO2&AO3 LOSS Of NNI-X/Y I 7 BWlAW Turbine Trip I 4 BWIAOS Emergency Diesel Actuation I 6

~ _ _ _ _ _ _ _ _

BWIAO? Flooding I 8 BWIE03 Inadequate Subcooling Margin I 4 BWIEOI; WIEO3 LOCA Cooldown - Depress. I 4 BWIEOB; CEIA13; WIE09&E10 Natural Circ. I 4 BWIE13&E14 EOP Rules and Enclosures CEIAII; WIE08 RCS Overcooling - PTS I 4 CEIAI 6 Excess RCS Leakage I 2

CEIE09 Functional Recovery KIA Category Point Totals:

Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and utomatic and manual features.

Cooldown and Depressurization:

Normal, abnormal, and emergency operating procedures associated wth LOCA Cooldown and Depressunzation.

(CFR 41.5/41.10/45.6/45.13) implications of the following concepts as they apply to the Natural Circulation Cooldown:

Normal, abnormal, and emergency between the Pressurized Thermal Shock and Components and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and utomati and manual features.

I G~OUD Point Total:

3.1 3.4 3.3 3.4 1

I 2

ES-401, Rev. 9 Form ES-401-2 Seabrook Station 2007 NRC Exam Tier IIGroup 2 SRO ElAPE # I Name I Safety Function ES-401 K

H 1

2 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions -Tier IIGroup 2 (RO I SRO) 000001 Continuous Rod Withdrawal I I 000003 Dropped Control Rod I I I I 000005 InoperablelStuck Control Rod I1 000024 Emergency Boration I 1

000028 Pressurizer Level Malfunction I 2 000032 Loss of Source Range NI I 7

000033 Loss of Intermediate Range NI I 7

I I I I 000036 (BWIAOS) Fuel Handling Accident I 8 000037 Steam Generator Tube Leak I 3 000051 Loss of Condenser Vacuum I 4 000059 Accidental Liquid RadWaste Rel. I 9 000060 Accidental Gaseous Radwaste Rel. I 9 000061 ARM System Alarms I 7

000067 Plant Fire On-site I 8 I I 000068 (BWlAO6) Control Room Evac. I 8 I I 000069 (WIEl4) Loss of CTMT Integrity I 5

000074 (WIEO6&EO7) Inad. Core Cooling I 4 000076 High Reactor Coolant Activity I 9 I I WIEOI & E02 Rediagnosis & SI Termination I 3 WlEl3 Steam Generator Over-pressure I 4

WIEl5 Containment Flooding I 5 WIEl6 High Containment Radiation I 9 BWlAOl Plant Runback I 1

BWlAO2&AO3 LOSS of NNI-XN I 7

BWIAO4 Turbine Trip I 4 BWIAO5 Emergency Diesel Actuation I 6 BWIAO7 Flooding I 8 I I BWIE03 Inadequate Subcooling Margin I 4 BWIEOI; WIE03 LOCA Cooldown - Depress. I 4 BWIEOQ; CElAI3; WIEOQ&E10 Natural Circ. I 4 BWlEl3&E14 EOP Rules and Enclosures 1

I Seabrook Station 2007 NRC Exam Tier IIGrouD 2 SRO (Continued)

I KIA Category Point Totals:

I CUE09 Functional Recoverv I I 1 1 1 1 1 I

I I

I I I I

I : I G~OUP Point ~otel:

I 9/4 2

ES-401, Rev. 9 Form ES-401-2 Seabrook Station 2007 NRC Exam Tier 2IGroup I RO ES-401 System # I Name 003 Reactor Coolant Pump 004 Chemical and Volume Control 005 Residual Heat Removal 006 Emergency Core Cooling 007 Pressurizer ReliefIQuench Tank 008 Component Cooling Water 010 Pressurizer Pressure Control 012 Reactor Protection 01 3 Engineered Safety Features Actuation 022 Containment Cooling Plant S!

PWR Examination Outline F O I ~

ES-401-2 terns -Tier WGroup I (RO I SRO)

KIA Topic(s) 4 ips between the RCPS and have on the following:

Normal water supply for SiS relationships between the RPS and the following systems:

CFR 41.2 to 41.9145.7 to 45.8 Input channels and logic.

1 1

I 1

1 I

I I

1 1

1

025 Ice Condenser 026 Containment Spray 039 Main and Reheat Steam 059 Main Feedwater 061 AuxiliatylEmergency Feedwater 062 AC Electrical Distribution 063 DC Electrical Distribution 064 Emergency Diesel Generator 073 Process Radiation Monitoring 076 Service Water 078 Instrument Air 103 containment I

I G 2.4.2, Knowledge of system set Doints. interlocks and automatic operation of the MFW, including:

Programmed levels of the SIG.

of the following malfunctions or operations on the AFW, and based on those predictions, use procedures to correct, control, or mitigate the monitor changes in parameters associated with operating the ac nents using DC control relationships between the PRM system and the following systems:

Those systems served by PRM's.

I I

ESF Air Pressure automatic operation of the 5.9 2.8 2.9 3.4 3.4 3.5 3.9 3.6 3.6 3.1 3.9 1

7 1 -

1 -

1 1

2

Seabrook Station 2007 NRC Exam Tier 2IGroup 1 RO (Continued) 003 Reactor Coolant Pump 004 Chemical and Volume Control 005 Residual Heat Removal 010 Pressurizer Pressure Control 012 Reactor Protection 059 Main Feedwater 061 AuxlllaryIEmergency Feedwater KIA Category Point Totals:

following concepts as they apply to RCS heatup and cooldown effect or operations on the RPS, and based on those predictions, use procedures to correct, control, or mitigate the detectors and function generators.

Containment isolation valves affecting RCP operation.

41.7145.5 following concepts as they apply to RCS heatup and cooldown effect or operations on the RPS, and based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions Faulty or erratic operation of detectors and function generators.

41.5i43.5i45.3145.5 which provide for the following:

1 1

1 1

1 3

ES-401, Rev. 9 Form ES-401-2 Seabrook Station 2007 NRC Exam Tier 2IGroup 1 SRO ES-401 PWR Examination Outline Form ES-401-2 Plant S System # I Name I :I :I !

003ReactorCoolantPump I I I I

005 Residual Heat Removal 006 Emergency Core Cooling -1 007 Pressurizer RelIeflQuench Tank 008 Component Cooling Water 010 Pressurizer Pressure Control 01 2 Reador Protection 013 Engineered Safety Features Actuation 022 Containment Cooling 025 Ice Condenser 039 Main and Reheat Steam 059 Main Feedwater 7

061 AuxilierylEmergency Feedwater 062 AC Electrical Distribution 064 Emergency Diesel Generator 073 Process Radiation Monitoring items - Tier 2/Group I (RO I SRO)

Seabrook Station 2007 NRC Exam Tier 2IGroup 1 SRO (Continued) 2

ES-401, Rev. 9 Form ES-401-2 Seabrook Station 2007 NRC Exam Tier 21Group 2 RO ES-401 PWR Examination Outline Form ES-401-2 Plant Systems -Tier 2lGroup 2 (RO I SRO)

System # I Name 001 Control Rod Drive 002 Reador Coolant 011 Pressurizer Level Control 014 Rod Position Indication 015 Nuclear Instrumentation 016 Nonnuclear Instrumentation 01 7 ln-core Temperature Monitor 027 Containment Iodine Removal 028 Hydrogen Recombiner and Purge Control 029 Containment Purge 033 Spent Fuel Pool Cooling 034 Fuel Handling Equipment 035 Steam Generator 041 Steam Dump/Turbine Bypass Control 045 Main Turbine Generator 1

I I

1 -

1 I

I I

2

ES-401, Rev. 9 Form ES401-2 Seabrook Station 2007 NRC Exam Tier 2IGroup 2 SRO ES401 PWR Examination Outline Plant Systems -Tier WGroup 2 (RO I SRO)

F o

~

ES401-2 System # I Name K

K H

1 2

3 001 Control Rod Drive 002 Reador Coolant 01 1 Pressurizer Level Control 014 Rod Position Indication 015 Nuclear Instrumentation 016 Non-nuclear Instrumentation H

4 -

M 5 -

I I I I l 017 ln-core Temperature Monitor 027 Containment Iodine 028 Hydrogen Recombiner 035 Steam Generator 041 Steam Dumpfrurbine Bypass Control 045 Main Turbine Generator 055 Condenser Air Removal 056 Condensate 068 Liquid Radwaste 071 Waste Gas Disposal I I I I I 4

WA Topic(s)

I IR A2.03, Ability to predict the impacts of the following malfunctions or operations on the Spent Fuel Pool Cooling System and based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

loss of water level.

Abnormal spent fuel pool water level or and based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Dropped fuel element.

I I

I (CFR 43.4 I45.10) 1 1

I -

1

2

ES-401, Rev. 9 Generic Knowledge and Abilities Outline (Tier 3)

Form ES-401-3 Seabrook Station 2007 NRC Exam Tier 3 RO Facility:

Date of Exam:

Topic RO I

SRO-Only Category KIA #

IR 3.7 3.0 3.2 3.4 2.1.I G

(CFR: 41.10 145.13) 2.1.1 Knowledge of conduct of operations requirements.

1.

Conduct of Operations 2.1.I 1

G specification action statements for systems.

(CFR 43.2 I45.13) 2.1.11 Knowledge of less than one hour technical 2.1.28 2.1.28 Knowledge of the purpose and function of major system components and controls.

(CFR: 41.7) 2.1.33 2.1.33 Ability to recognize indications for system operating parameters which are entry-level conditions for technical specifications.

(CFR 43.2 143.3 I 45.3) 2.1.

2.1.

Subtotal

2.

Equipment Control 2.2.1 3 2.2.13 Knowledge of tagging and clearance procedures.

(CFR:

4 1.10 / 45.13) 3.6 3.4 2.2.22 Knowledge of limiting conditions for operations and safety limits.

(CFR: 43.2 / 45.2) 2.2.22 2.2.

2.2.

2.2.

2.2.

Subtotal 1

Seabrook Station 2007 NRC Exam Tier 3 RO (Continued) 2.3.1 Knowledge of 10 CFR: 20 and related facility radiation control requirements.

(CFR. 41.12 143.4.45.9 145.10) 2.3.4 Knowledge of radiation exposure limits and contamination control, including permissible levels in excess of those authorized.

(CFR. 43.4 / 45.10)

3.

Radiation Control 2.6 1

2.5 1

2.3.1

4.

Emergency Procedures I Plan 2.3.4 2.3.

2.3.

2.3.

2.3.

Subtotal 2-4-27 2.4.27 Knowledge of fire in the plant procedure.

(CFR 41.10 I43.5 / 45.13) 2.4.49 2.4.

2.4.49 Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.

(CFR: 41.10 / 43.2 145.6) 2.4.

2.4.

2.4.

Subtotal Tier 3 Point Total 2

ES401, Rev. 9 Generic Knowledge and Abilities Outline (Tier 3)

Form ES401-3 Seabrook Station 2007 NRC Exam Tier 3 SRO Faci I ity :

Date of Exam:

RO I

SRO-Only KIA #

Topic Category 1

I IR IR 2.1.4 2.1.4 Knowledge of shift staffing requirements.

(CFR: 41.10 / 43.2)

1.

Conduct of Operations 2.1.10 Knowledge of conditions and limitations in the facility license.

(CFR: 43.1 I45.13) 2.1.lo 2.1.

2.1.

2.1.

2.1.

Subtotal 2.2.21 2.2.21 Knowledge of pre-and post-maintenance operability requirements.

(CFR: 43.2)

2.

Equipment Control limiting conditions for operations and safety limits.

CFR 43.2 2.2.

2.2.

2.2.

2 Subtotal 2.3.4 2.3.4 Knowledge of radiation exposure limits and contamination control, including permissible levels in excess of those authorized.

(CFR: 43.4 / 45.10) 1

3.

Radiation Control 2.3.

2.3.

2.3.

2.3.

2.3.

Subtotal I

1

4.

Emergency Procedures I Plan Tier 3 Point Total Seabrook Station 2007 NRC Exam Tier 3 SRO (Continued) 2.4.21 2.4.38 2.4.

2.4.

2.4.

2.4.

2.4.21 Knowledge of the parameters and logic used to assess the status of safety functions including:

1. Reactivity control

2. Core cooling and heat removal

3. Reactor coolant system integrity

4. Containment conditions

5. Radioactivity release control.

(CFR. 43.5 / 45.12) 2.4.38 Ability to take actions called for in the facility emergency plan, including (if required) supporting or acting as emergency coordinator.

(CFR: 43.5 I45.11)

Subtotal 2

7 2

ES-401, Rev. 9 Record of Rejected KIAs Form ES-401-4 Tier I Group 1 I1 1 I1 2/1 2/1 2/2 Randomly Selected KIA 008 AKI.Ol E l l EA2.1 003 K I.02 012, Reactor Protection 2.1.27 Knowledge of system purpose and or function.

033, A2.03 Reason for Rejection Original proposed exam question was reviewed by Lead NRC Examiner and deemed to be fundamentals based.

The examiner recommended randomly selecting a different KIA from the same category (008). A new ramdomly selected KIA, 008 AK3.03 has been selected, and an associated new exam question has been submitted.

Original proposed exam question was reviewed by Lead NRC Examiner and deemed to be too similar to SIMJPMO9. The examiner recommended randomly selecting a different KIA from the same category (El I).

A new ramdomly selected KIA, E l l EA2.2 has been selected and an associated new exam question has been submitted.

~~

~

Original proposed question was reviewed by Lead NRC Examiner and deemed to be a KIA mismatch. Multiple attempts to replace the question under the same KIA resulted in unwanted duplication of material within the written exam. A new randomly selected KIA, K1.04 from the same category (003, RCPS) and K number (KI) has been selected and an associated new exam question has been submitted.

Original proposed KA was discussed with NRC Lead Examiner per teleconference. The Examiner suggested that the KA would be difficult to apply to an SRO level question. A new randomly selected KA, 061, Auxiliary Feedwater, 2.1.20 Ability to execute procedure steps.

(CFR: 41.I 0 I 43.5 I 45.1 2)

~~

Original question associated with 033, A2.03 deemed to simple by Lead NRC Examiner. Due to difficulty creating an appropriate SRO level question for the KA new KA 034, A2.01 was randomly sampled. A new question has been submitted.

ES-301, Rev. 9 Administrative Topics Outline Form ES-301-1 Seabrook Station 2007 NRC Exam Admin Topics-SRO Facility: Seabrook Examination Level: RO [1 SRO B Date of Examination: July 9,2007 Operating Test Number:

Administrative Topic (see Note)

Conduct of Operations

~~

Conduct of Operations Equipment Control

~~

Radiation Control Emergency Plan Type Code*

S or R,N Describe activity to be performed 2.1.6 Ability to supervise and assume a management role during plant transients and upset conditions.

Activity-Reporting Requirements For Onsite Event 2.1.I2 Ability to apply technical specifications for a system.

Activity-Technical Specifications and Allowed Outage Time.

2.2.12 Knowledge of surveillance procedures.

Activity-Verify RCS Steady State Leak Rate Determination.

2.3.10 Ability to perform procedures to reduce excessive levels of radiation and guard against personnel exposure.

Activity-VeMy a Liquid Effluent Waste Sample Request 2.4.40 Knowledge of the SROs responsibilities in emergency pian implementation.

Activity-Emergency Pian Classification and NotMcation 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 (I 3 for ROs; < 4 for SROs & RO retakes)

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

(P)revious 2 exams (< I; randomly selected)

ES-301, Rev. 9 Administrative Topics Outline Form ES-301-1 Seabrook Station 2007 NRC Exam Admin Topics-RO Facility: Seabroo k ExaminationLevel: RO SRO 0 Date of Examination: July 9,2007 Operating Test Number:

Administrative Topic (see Note)

Conduct of Operations

~~

Conduct of Operations Equipment Control Radiation Control Emergency Plan Type Code*

~

~

Describe activity to be performed 2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

Activity-Perform RCS Steady State Leakrate Calculations.

2.1.25 Ability to obtain and interpret station reference materials such as graphs, monographs, and tables which contain performance data.

Activity-Calculate Shutdown Margin in Mode 2 With Dropped Rod.

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

Activity-Spent Fuel Pool Blended Makeup Calculation.

2.3.10 Ability to perform procedures to reduce excessive levels of radlation and guard against personnel exposure.

Activity-Verify COP Exhaust RY Setpoints Prior to Gaseous Effluent Release.

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 (< 3 for ROs; 5 4 for SROs & RO retakes)

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

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

ES-301, Rev. 9 Control Roomlln-Plant Systems Outline Form ES-301-2 System I JPM Title Seabrook Station 2007 NRC Exam JPM-SROI Type Code*

Safety Function Facility: Seabrook ExamLevel: RO 0 SRO-I SRO-U 0

g. Emergency Core Cooling SystemlPerform SI Termination/

Reduction Date of Examination: July 9, 2007 Operating Test No.:

4%

E, M 3

i. AC Electrical DistributionKransfer Vital Instrument Bus

j. Component Cooling Water SystemlAlign Alternate Cooling To CCP Lube Oil Cooler L,D D,E,R

a. Steam Generator SystemlSteam Header Pressure PT-507 Fails Low 6

8 4

b. Emergency Diesel GeneratorslEmergency Trip of DG 1 B

c. Nuclear Instrumentation SystemlPower Range NI Failure

d. Containment Purge SystemlPlacing COP in Service

e. Reactor Coolant SystemlDepressurize RCS Using Aux Spray (E-3) 2 5

I I

f. Containment Spray System/ Transfer To Cold Leg Recirc (CBS-V-2 Fails)

h.

I

k. Chemical and Volume Control SystemlLocal Rapid Manual Boration 1

1

• Type Codes

~~

~

(A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (L)ow-Power I Shutdown (N)ew or (M)odified from bank including 1(A)

(P)revious 2 exams (S)imulator

( W A Criteria for RO I SRO-I I SRO-U 4-6 14-6 I 2-3 s 9 1 s 8 1 s 4 2 1 121 I 2 1 2 1 / 2 1 I 2 1 2 2 1 2 2 1 2 1 s 3 I s 3 I s 2 (randomly selected) 2 1 I 2 1 121 2

ES-301, Rev. 9 Control Roomlln-Plant Systems Outline Form ES-301-2 System I JPM Title Type Code*

a. Containment Spray System/ Transfer To Cold Leg Recirc A,S,E,N (CBS-V-2 Fails)

Seabrook Station 2007 NRC Exam JPM-SROU Safety Function 5

Facility: Seabrook Exam Level: RO 0 SRO-I 0 SRO-U

i. AC Electrical Distributionflransfer Vital Instrument Bus L,D

j. Component Cooling Water SystemlAlign Alternate Cooling To W, R CCP Lube Oil Cooler Date of Examination: JUIY 9,2007 Operating Test No.:

6 a

b. Emergency Diesel GeneratorsEmergency Trip of DG 1 B 6

2 I

I C. Reactor Coolant SystemlDepressurize RCS Using Aux Spray (3)

e.

1.

9.

h.

In-Plant Systems@ (3 for RO); (3 for SRO-I);

(3 or 2 for SRO-U)

k.

~

All control room (and in-plant) systems must be different and 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 (L)ow-Power I Shutdown (N)ew or (M)odified from bank including 1(A)

(P)revious 2 exams (RFA (S)imulator Criteria for RO I SRO-I I SRO-U 4-6 I 4 6 12-3 s 9 1 s a 1 s 4 21 I 2 1 />I 2 1 1 2 1 / 2 1 2 2 / 2 2 / 2 1 5 3 I s 3 I < 2 (randomly selected) 21121121 2

\\

ES-301, Rev. 9 Control Roomlln-Plant Systems Outline Form ES-301-2 System I JPM Title Seabrook Station 2007 NRC Exam JPM-RO Type Code*

Safety Function Facility: Seabrook ExamLevel: RO SRO-I 0 SRO-U 0

b. Emergency Diesel GeneratorslEmergency Trip of DG 1 B C. Nuclear Instrumentation SystemlPower Range NI Failure

d. Containment Purge SystemlPlacing COP in Service Date of Examination: July 9, 2007 Operating Test No.:

S,A,D 6

W, E 7

S,N 8

~~

~ _ _ _ _

a. Steam Generator SystemlSteam Header Pressure PT-507 Fails I E,N 2

I I

e. Reactor Coolant SystemlDepressurize RCS Using Aux Spray

f. Containment Spray System/ Transfer To Cold Leg Recirc (Loss I A,S,E,N I of Recirculation 5

g. Emergency Core Cooling SystemlPerform SI Termination/

Reduction 3

1 I

S9D I

h. Chemical and Volume Control SystemlShifting From CCP to 1 PDP I In-Plant Systems" (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

I i. AC Electrical Distributionrrransfer Vital Instrument Bus I

L,D I

6

j. Component Cooling Water SystemlAlign Alternate Cooling To CCP Lube Oil Cooler 8

k. Chemical and Volume Control SystemlLocal Rapid Manual Boration 1

All control room (and in-plant) systems must be different and serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

• Type Codes

~~

Criteria for RO I SRO-I I SRO-U 1

(A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (L)ow-Power I Shutdown (N)ew or (M)odified from bank including 1(A)

(P)revious 2 exams

( W A (S)imulator 4-6 14-6 I 2-3

< 9 / < 8 / < 4 2 1 / 2 1 / 2 1 2 1 I 2 1 I 2 1 2 2 / 2 2 / 2 1 s 3 I < 3 15 2 (randomly selected) 2 1 / 2 1 / 2 1 2

Appendix D, Rev. 9 Scenario Outline Form ES-D-1 11 Facility: Seabrook Station Scenario No.: 4 (D)

Op-Test No.:

Examiners:

Operators:

Initial Conditions: Plant is at 30% power. Power decrease due to D Atmospheric Steam Dump I/ leaking.

11 Turnover: Plant at 30% power. Continue power decrease at 5Yhr.

The D ASDV is leaking

!d.

Malf.

No.

P W P T 505 MfRCOl 6

MfSGOO 2D rvMSAV R50 Event Type*

N-ATC N-BOP N-SRO R-SRO R-ATC R-BOP I-BOP I-SRO I-ATC R-ATC R-SRO C-SRO C-ATC C-BOP C-SRO C-ATC C-BOP M-SRO M-ATC M-BOP M-SRO M-ATC R-SRO R-ATC R-BOP Event Description Power Decrease Tavgnref mismatch failure.

D RCP seal failure.

SG Tube Rupture @ 300 gpm.

D SG Safety Valve leaks past seat.

Appendix D, Rev. 9 Scenario Outline Form ES-D-I Facility: Seabrook Station Scenario No.: 2 (6)

OpTest No.:

Examiners:

Operators:

Initial Conditions: Middle of Life. 75% power after downpower at 20%lhr.

Turnover: Continue power increase Q 1 OYdhr. SUFP is aligned to Bus 4 for breaker testing of Bus 5 SUFP breaker.

Event I

2 3

4 5

6 7

Malf.

No.

CfRCLT 459 MfCSOl 6

~~

FWP32 B

MFRPS 001 RfMSVl 29 MfEDO3 8

Event Type*

R-SRO N-ATC N-BOP R-ATC R-BOP I-SRO I-ATC C-ATC C-SRO C-ATC C-SRO C-BOP M-SRO M-ATC R-SRO R-ATC R-BOP C-BOP C-SRO M-SRO M-ATC M-BOP Event DescriDtion Power Increase, Reactivity Change Controlling channel of PZR level fails low.

Charging pump 2A overcurrent trip.

'B' Main Feed Pump shaft shear.

Failure of reactor to trip.

Turbine Driven Emergency Feedwater Pump trip.

Loss of Offsite Power FR-H.1.

II *

(N)orrnal, (R)eactivity, (I)nstrurnent, (C)ornponent,

Atmendix D. Rev. 9 Scenario Outline Form ES-D-I Facility: Seabrook Station Scenario No.: 1 (A)

Op-Test No.:

ll Examiners:

Operators:

Initial Conditions: Plant is at 75% power. Middle of Life. Xenon is building in after a 20% per hour power decrease. Boron concentration is 11 71 ppm.

Turnover: Raise power to 100% Q 1 O%/hr.

Event 1

2 3

4 5

Malf.

No.

CtFWFK 530 L W L T 553 PtRCPT 455 MfRC04 8C MFR005 2

Event Type*

R-SRO N-SRO R-ATC R-BOP N-ATC N-RnD I-BOP C-BOP C-SRO I-SRO I-ATC I-SRO C-ATC C-SRO C-SRO C-ATC N-SRO R-SRO R-BOP R-ATC R-SRO R-ATC Event Description Power Increase, Reactivity Change C Fw Reg Valve Fails to 100% output and controlling channel of SG level fails low.

Controlling PZR pressure instrument fails high.

RCS Leak at 30 GPM to containment.

Shutdown portion of RCS leak abnormal 6

I I,,

Pressurizer manway failure.

M-SRO M-ATC M-BOP (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor