ML14080A029

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Initial Exam 2013-302 Draft Administrative Documents
ML14080A029
Person / Time
Site: Sequoyah  Tennessee Valley Authority icon.png
Issue date: 03/21/2014
From:
NRC/RGN-II
To:
Tennessee Valley Authority
Shared Package
ML14080A058 List:
References
50-327/13-302, 50-328/13-302
Download: ML14080A029 (27)
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ES-301 Administrative Topics Outline Form ES-301-1 Facility: Sequoyah Nuclear Station 1 & 2 Date of Examination:12/2/2013 Exam Level: RO SRO Operating Test No: 2013-302 Administrative Topic (see Type Describe activity to be performed Note) Code*

Calculate the Boric Acid Controller Setting Conduct of Operations R, M 2.1.37 (4.3) Knowledge of procedures, guidelines, or limitations associated with reactivity management.

Determine Maintenance of Active License Status.

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

Perform a Defense-In-Depth Assessment.

Equipment Control R, N 2.2.18 (2.6) Knowledge of the process for managing maintenance activities during shutdown operations.

Radiation Control Not examined Perform RO Actions During Aircraft Probable Threat AOP-T.01 App D.

Emergency Procedures/Plan R, N 2.4.39 (3.9) Knowledge of RO responsibilities in emergency plan implementation.

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; 4 for SROs & RO retakes)

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

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

A.1.a Given plant data with Unit 1 in MODE 5, the examinee will calculate the Boric Acid Controller Setting of 71.2 using 0-SO-62-7 Boron Concentration Control Appendix M.

A.1.b Given a situation with five different Senior Reactor Operators that are not currently on shift, the Examinee will assess the work schedules for five different operators and using OPDP 10 License Status Maintenance, Reactivation and Proficiency for Non-Licensed Positions determines Operator #2, #3 and #4 have maintained an active license status and that Operator

  1. 1 and #5 have not accumulated sufficient time required to maintain an active license status.

A.2 Given the following conditions:

Unit 1 is in MODE 5 following a Refueling Outage.

S/G Manways are installed.

RCS Boron concentration is sufficient for Shutdown Margin requirements.

The RCS is in a Partial Drain Condition.

The Switchyard is protected.

The 1A Charging pump is out of service for maintenance.

The 1B 6.9 kv Shutdown Board was de-energized and locked out while performing work.

The Examinee will perform a Defense-In-Depth Assessment using 1-PI-OPS-000-020.2, OPERATOR AT THE CONTROLS DUTY STATION CHECKLISTS-MODES 5, 6 AND DE-FUELED and determines a RED condition exists on Decay Heat removal and an ORANGE condition exists on Power Availability and that Decay Heat Removal is the most significant challenged safety function.

A.3 Not examined.

A.4 During an Aircraft PROBABLE Threat event the examinee will perform AOP-T.01, SECURITY EVENTS, Appendix D Aircraft PROBABLE Threat Notifications. The Examinee will notify the SM to classify an ALERT based on EAL 4.6, determine that inadequate time for rapid evacuation exists, and ensures REP Responders are dispatched using the REP Paging system.

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Sequoyah Nuclear Station 1 & 2 Date of Examination:12/2/2013 Exam Level: RO SRO Operating Test No: 2013-302 Administrative Topic (see Type Describe activity to be performed Note) Code*

Determine Actions Required Following a Reactivity Management Event When at Power.

Conduct of Operations R, M 2.1.37 (4.6) Knowledge of procedures, guidelines, or limitations associated with reactivity management.

Determine Maintenance of Active License Status.

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

Perform a Defense-In-Depth Assessment.

Equipment Control R, N 2.2.18 (3.9) Knowledge of the process for managing maintenance activities during shutdown operations.

Approve a Waste Gas Decay Tank Release for Maintenance and Radiation Monitor RM-118 Radiation Control R, M INOPERABLE.

2.3.6 (3.8) Ability to approve release permits.

Classify The Event Using The EPIP-1 and Complete a State Notification Form.

Emergency Procedures/Plan R, M 2.4.41 (4.6) Knowledge of the emergency action level thresholds and classifications.

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; 4 for SROs & RO retakes)

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

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

A.1.a Given a sequence of events while acting as the Unit Supervisor when a CVCS Purification mixed bed ion exchanger was placed in service with insufficient boron loading, the examinee will determine the following:

The severity of the event as a Minor Reactivity Management Event using NPG-SPP-10.4 Reactivity Management Program.

That site operations management and duty plant manager are the required internal notifications using NPG-SPP-3.5 Regulatory Reporting Requirements.

This task is based on a Sequoyah internal operating event.

A.1.b Given a situation with five different Senior Reactor Operators that are not currently on shift, the Examinee will assess the work schedules for five different operators and using OPDP 10 License Status Maintenance, Reactivation and Proficiency for Non-Licensed Positions determines Operator #2, #3, #4 and #5 have maintained an active license status and that Operator #1 and #5 have not accumulated sufficient time required to maintain an active license status.

A.2 Given the following conditions:

Unit 1 is in MODE 5 following a Refueling Outage.

S/G Manways are installed.

RCS Boron concentration is sufficient for Shutdown Margin requirements.

The RCS is in a Partial Drain Condition.

The Switchyard is protected.

The 1A Charging pump is out of service for maintenance.

The 1B 6.9 kv Shutdown Board was de-energized and locked out while performing work.

The Examinee will perform a Defense-In-Depth Assessment using 1-PI-OPS-000-020.2, OPERATOR AT THE CONTROLS DUTY STATION CHECKLISTS-MODES 5, 6 AND DE-FUELED and determines a RED condition exists on Decay Heat removal and an ORANGE condition exists on Power Availability and that Decay Heat Removal is the most significant challenged safety function.

A.3 Given a situation while acting as the Unit Supervisor when a Waste Gas Decay Tank B release is planned with 0-RE-90-118 inoperable, the Examinee determines the following requirements are necessary to accommodate the Waste gas release and intrusive maintenance:

A Purge required prior to maintenance.

ABGTS Train B will be used for the release.

Two independent samples and analyses of the Waste Gas Decay are performed.

Two independent calculations of the Waste Gas Decay Tank release rate are performed.

Two independent of the verifications discharge valve lineup are performed.

The requirements listed are necessary to demonstrate the appropriate administrative controls that are in place to preclude the possibility of an inadvertent release radioactive in excess of limits to the public.

A.4 Acting as the Site Emergency Director during a MODE 5 LOCA, the Examinee classifies the event as a SITE AREA EMERGENCY based on EAL 6.1 and the Examinee completes a TVA Initial Notification for Site Area Emergency form with no errors on items noted with an *.

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Sequoyah Nuclear Station 1 & 2 Date of Examination: 12/2/2013 Exam Level: RO SRO-I SRO-U Operating Test No: 2013-302 Control Room Systems (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System / JPM Title Type Code* Safety Function

a. Initiate Emergency Boration with Multiple Control Rods Not Fully M, A, S 1 Inserted. EPE 024 EA 1.06 (3.2/3.1)
b. Depressurize an Unisolable Cold Leg Accumulator. 006 A4.02 (4.0/3.8) M, A, EN, L, 2

S

c. Respond to a Shutdown LOCA with a Failure of Containment Isolation N, A, EN, L, S 3 and Containment Ventilation Isolation. EPE 011 EA 2.04 (3.8/4.0)
d. Establish Once Through Cooling by Initiating RCS Bleed and Feed. N, A, EN, L, 4P EPE E05 EA 2.2 (3.7/4.3) EN, S
e. Synchronize the Main Generator to the Grid. 045 A4.02 (2.7/2.6) N, A, S 4S
f. Perform Equipment Checks Following ESF Actuation. 103 A4.01 M, EN, L, S 5 (4.5/4.8)
g. Respond to a Main Control Room High Radiation Alarm with a Failure D, S 7 of Control Room Isolation. APE 061 AA1.01 (3.6/3.6)
h. Perform CR Actions for Fire in the Auxiliary Building. APE 067 AA2.17 N, L, S 8 (3.5/4.3)

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

i. Respond to Decreasing RCS Pressure from the Aux CR. APE 068 D, E 8 AA1.12 (4.4/4.4)
j. Align Upper Containment Radiation Monitor to Lower Containment. D, R 2 002 A3.01 (3.7/3.9)
k. Cycle the Unit 2 Main Generator PCB 062A4.04 (2.6/2.7) N, L 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 / SRO-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 (C)ontrol room (D)irect from bank 9/8/4 (E)mergency or abnormal in-plant 1/1/1 (EN)gineered safety feature - / - / 1 (control room system)

(L)ow-Power / Shutdown 1/1/1 (N)ew or (M)odified from bank including 1(A) 2/2/1 (P)revious 2 exams 3 / 3 / 2 (randomly selected)

(R)CA 1/1/1 (S)imulator

a. The examinee will assume the shift in MODE 3 following a Reactor Trip with multiple Control Rods that are not fully inserted. The examinee will be directed to initiate emergency boration from the A BAT using EA-68-4 Emergency Boration. The examinee will not be able to establish emergency boration using the CHARGING FLOW CONTROL valve and will implement the alternate path to establish emergency boration through the Charging Pump ECCS Discharge (CCPIT) valves.
b. The examinee will assume the shift during an event with E-1, LOSS OF REACTOR OR SECONDARY COOLANT in progress. The examinee will isolate the Loop 1, 2 and 3 Cold Leg Accumulators (CLA) but discovers the Loop 4 CLA discharge valve will not close. The Examinee will implement the alternate path and vent the Loop 4 CLA using EA-63-1, VENTING UNISOLATED COLD LEG ACCUMULATOR.
c. The examinee will assume the shift in MODE 5 and will respond to plant conditions. The Examinee will diagnose a loss of RCS Inventory and manually stop Reactor Coolant and Residual Heat Removal Pumps and align injection from the Charging Pumps. The Examinee will implement the alternate path to manually close Containment Radiation Monitor and Containment Purge isolation valves using AOP-R.02 SHUTDOWN LOCA.
d. The examinee will assume the shift in MODE 3 with an event in progress. The Examinee will restore a Heat Sink using FR-H.1 LOSS OF SECONDARY HEAT SINK. The Examinee will implement the alternate path to manually start the Train A Safety Injection Pump and vent the RCS using the Reactor head vents which establishes RCS Bleed and Feed.
e. The examinee will assume the shift in MODE 1 and will start and synchronize the Main Generator the grid using 0-GO-4, POWER ASCENSION FROM LESS THAN 5% REACTOR POWER TO 30%

REACTOR POWER. The examinee will respond to a failure of the voltage regulator and implement the alternate path to to trip the main turbine.

f. The examinee will assume the shift in MODE 3 following an inadvertent Reactor Trip and Safety Injection. The examinee will perform EA-0-1 EQUIPMENT CHECKS FOLLOWING ESF ACTUATION and will start at least one train of Emergency Gas Treatment and Auxiliary Building Gas Treatment fans.
g. The examinee will assume the shift in MODE 1 and will respond to plant conditions. The examinee will diagnose a Main Control Room high radiation condition and manually place one train of Emergency Control Room Ventilation in service and isolate the normal Control Room Ventilation using 0-SO-30-2 Control Room Isolation.
h. The examinee will assume the shift in MODE 3 with an uncontrolled fire in the Unit 1 Aux Building Penetration Room. The examinee will perform AOP-N.08 APPENDIX R FIRE SAFE SHUTDOWN to perform time critical actions which will stop Charging pumps, and manually align Charging pump suction and discharge to the ECCS using AOP-N.08, APPENDIX R FIRE SAFE SHUTDOWN.
i. While responding to lowering RCS pressure during a Control Room abandonment situation, the examinee will take local control of the Pressurizer PORVs at the Auxiliary Control Panel and shut the PORVs.
j. The examinee will assume the shift in MODE 1 when the Lower Containment radiation monitor becomes unavailable due to a fault. The examinee will start a sample pump and align Upper Containment radiation monitor 2-RM-90-112 to sample lower containment using 2-SO-90-2 Gaseous Process Radiation Monitoring System.
k. The examinee will assume the shift in MODE 5 with maintenance complete on the Unit 2 main Generator PCB. The examinee cycle the Unit 2 Main Generator circuit breaker with generator using 2-SO-57-1, GENERATOR CIRCUIT BREAKER. This component is a recent plant modification.

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Sequoyah Nuclear Station 1 & 2 Date of Examination: 12/2/2013 Exam Level: RO SRO-I SRO-U Operating Test No: 2013-302 Control Room Systems (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System / JPM Title Type Code* Safety Function

a. Initiate Emergency Boration with Multiple Control Rods Not Fully M, A, S 1 Inserted. EPE 024 EA 1.06 (3.2/3.1)
b. Depressurize an Unisolable Cold Leg Accumulator. 006 A4.02 (4.0/3.8) M, A, EN, L, 2

S

c. Respond to a Shutdown LOCA with a Failure of Containment Isolation N, A, EN, L, S 3 and Containment Ventilation Isolation. EPE 011 EA 2.04 (3.8/4.0)
d. Establish Once Through Cooling by Initiating RCS Bleed and Feed. N, A, EN, L, 4P EPE E05 EA 2.2 (3.7/4.3) EN, S
e. Synchronize the Main Generator to the Grid. 045 A4.02 (2.7/2.6) N, A, S 4S
f. Perform Equipment Checks Following ESF Actuation with a Failure of M, EN, L, S 5 ESF Slave Relays. 103 A4.01 (4.5/4.8)
g. Not examined N/A N/A
h. Perform CR Actions for Fire in the Auxiliary Building with a Failure of N, L, S 8 CCPIT Valves to Close. APE 067 AA2.17 (3.5/4.3)

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

i. Respond to Decreasing RCS Pressure from the Aux CR. APE 068 D, E 8 AA1.12 (4.4/4.4)
j. Align Upper Containment Radiation Monitor to Lower Containment. D, R 2 002 A3.01 (3.7/3.9)
k. Cycle the Unit 2 Main Generator PCB 062A4.04 (2.6/2.7) N, L 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 / SRO-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 (C)ontrol room (D)irect from bank 9/8/4 (E)mergency or abnormal in-plant 1/1/1 (EN)gineered safety feature - / - / 1 (control room system)

(L)ow-Power / Shutdown 1/1/1 (N)ew or (M)odified from bank including 1(A) 2/2/1 (P)revious 2 exams 3 / 3 / 2 (randomly selected)

(R)CA 1/1/1 (S)imulator

a. The examinee will assume the shift in MODE 3 following a Reactor Trip with multiple Control Rods that are not fully inserted. The examinee will be directed to initiate emergency boration from the A BAT using EA-68-4 Emergency Boration. The examinee will not be able to establish emergency boration using the CHARGING FLOW CONTROL valve and will implement the alternate path to establish emergency boration through the Charging Pump ECCS Discharge (CCPIT) valves.
b. The examinee will assume the shift during an event with E-1, LOSS OF REACTOR OR SECONDARY COOLANT in progress. The examinee will isolate the Loop 1, 2 and 3 Cold Leg Accumulators (CLA) but discovers the Loop 4 CLA discharge valve will not close. The Examinee will implement the alternate path and vent the Loop 4 CLA using EA-63-1, VENTING UNISOLATED COLD LEG ACCUMULATOR.
c. The examinee will assume the shift in MODE 5 and will respond to plant conditions. The Examinee will diagnose a loss of RCS Inventory and manually stop Reactor Coolant and Residual Heat Removal Pumps and align injection from the Charging Pumps. The Examinee will implement the alternate path to manually close Containment Radiation Monitor and Containment Purge isolation valves using AOP-R.02 SHUTDOWN LOCA.
d. The examinee will assume the shift in MODE 3 with an event in progress. The Examinee will restore a Heat Sink using FR-H.1 LOSS OF SECONDARY HEAT SINK. The Examinee will implement the alternate path to manually start the Train A Safety Injection Pump and vent the RCS using the Reactor head vents which establishes RCS Bleed and Feed.
e. The examinee will assume the shift in MODE 1 and will start and synchronize the Main Generator the grid using 0-GO-4, POWER ASCENSION FROM LESS THAN 5% REACTOR POWER TO 30%

REACTOR POWER. The examinee will respond to a failure of the voltage regulator and implement the alternate path to to trip the main turbine.

f. The examinee will assume the shift in MODE 3 following an inadvertent Reactor Trip and Safety Injection. The examinee will perform EA-0-1 EQUIPMENT CHECKS FOLLOWING ESF ACTUATION and will implement the alternate path to start at least one train of Emergency Gas Treatment and Auxiliary Building Gas Treatment fans.
g. Not Examined
h. The examinee will assume the shift in MODE 3 with an uncontrolled fire in the Unit 1 Aux Building Penetration Room. The examinee will perform AOP-N.08 APPENDIX R FIRE SAFE SHUTDOWN to perform time critical actions which will stop Charging pumps, and manually align Charging pump suction and discharge to the ECCS.
i. While responding to lowering RCS pressure during a Control Room abandonment situation, the examinee will take local control of the Pressurizer PORVs at the Auxiliary Control Panel and shut the PORVs.
j. The examinee will assume the shift in MODE 1 when the Lower Containment radiation monitor becomes unavailable due to a fault. The examinee will start a sample pump and align Upper Containment radiation monitor 2-RM-90-112 to sample lower containment using 2-SO-90-2 Gaseous Process Radiation Monitoring System.
k. The examinee will assume the shift in MODE 5 with maintenance complete on the Unit 2 main Generator PCB. The examinee cycle the Unit 2 Main Generator circuit breaker with generator using 2-SO-57-1, GENERATOR CIRCUIT BREAKER. This component is a recent plant modification.

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Sequoyah Nuclear Station 1 & 2 Date of Examination: 12/2/2013 Exam Level: RO SRO-I SRO-U Operating Test No: 2013-302 Control Room Systems (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System / JPM Title Type Code* Safety Function

a. Initiate Emergency Boration with Multiple Control Rods Not Fully M, A, S 1 Inserted. EPE 024 EA 1.06 (3.2/3.1)
b. Not Examined N/A N/A
c. Respond to a Shutdown LOCA with a Failure of Containment Isolation N, A, EN, L, S 3 and Containment Ventilation Isolation. EPE 011 EA 2.04 (3.8/4.0)
d. Not Examined N/A N/A
e. Synchronize the Main Generator to the Grid. 045 A4.02 (2.7/2.6) N, A, S 4S
f. Not Examined N/A N/A
g. Not Examined N/A N/A
h. Not Examined N/A N/A In-Plant Systems (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)
i. Respond to Decreasing RCS Pressure from the Aux CR. APE 068 D, E 8 AA1.12 (4.4/4.4)
j. Align Upper Containment Radiation Monitor to Lower Containment. D, R 2 002 A3.01 (3.7/3.9)
k. Not Examined N/A N/A

@ 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 / SRO-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 (C)ontrol room (D)irect from bank 9/8/4 (E)mergency or abnormal in-plant 1/1/1 (EN)gineered safety feature - / - / 1 (control room system)

(L)ow-Power / Shutdown 1/1/1 (N)ew or (M)odified from bank including 1(A) 2/2/1 (P)revious 2 exams 3 / 3 / 2 (randomly selected)

(R)CA 1/1/1 (S)imulator

a. The examinee will assume the shift in MODE 3 following a Reactor Trip with multiple Control Rods that are not fully inserted. The examinee will be directed to initiate emergency boration from the A BAT using EA-68-4 Emergency Boration. The examinee will not be able to establish emergency boration using the CHARGING FLOW CONTROL valve and will implement the alternate path to establish emergency boration through the Charging Pump ECCS Discharge (CCPIT) valves.
b. Not Examined
c. The examinee will assume the shift in MODE 5 and will respond to plant conditions. The Examinee will diagnose a loss of RCS Inventory and manually stop Reactor Coolant and Residual Heat Removal Pumps and align injection from the Charging Pumps. The Examinee will implement the alternate path to manually close Containment Radiation Monitor and Containment Purge isolation valves using AOP-R.02 SHUTDOWN LOCA.
d. Not Examined
e. The examinee will assume the shift in MODE 1 and will start and synchronize the Main Generator the grid using 0-GO-4, POWER ASCENSION FROM LESS THAN 5% REACTOR POWER TO 30%

REACTOR POWER. The examinee will respond to a failure of the voltage regulator and implement the alternate path to to trip the main turbine.

f. Not Examined
g. The examinee will assume the shift in MODE 1 and will respond to plant conditions. The examinee will diagnose a Main Control Room high radiation condition and manually place one train of Emergency Control Room Ventilation in service and isolate the normal Control Room Ventilation using 0-SO-30-2 Control Room Isolation.
h. Not Examined
i. While responding to lowering RCS pressure during a Control Room abandonment situation, the examinee will take local control of the Pressurizer PORVs at the Auxiliary Control Panel and shut the PORVs.
j. The examinee will assume the shift in MODE 1 when the Lower Containment radiation monitor becomes unavailable due to a fault. The examinee will start a sample pump and align Upper Containment radiation monitor 2-RM-90-112 to sample lower containment using 2-SO-90-2 Gaseous Process Radiation Monitoring System.
k. Not Examined.

ES-401 PWR Examination Outline Form ES-401-2 Facility: SEQUOYAH Date of Exam: DECEMBER 2013 ROKCaoyPnts SRO-OnIy_Points Tier Group K K K K K KAAAA G A2 G* Total 1 2 3 4 5 6 1 2 3 4 Total

1. 1 iiL 1L I 18 3 3 6 Emergency &

Abnormal 2 I N/A j_ N/A 9 2 2 4 Plant Evolutions TierTotals 4 5 5 4 5 4 27 5 5 10 1 3 3 3 3 2 2 2 3 3 2 2 28 2 3 5 2.

Plant 2 L+/-J.. L+/-+/- 10 0 1 2 3 Systems Tier Totals 3 4 4 4 3 3 3 4 4 3 3 38 3 5 8

3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 2 3 3 2 1 2 2 2 Note: 1 . Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outlines (i.e., exceptfor one category in Tier 3 ofthe SRO-only outline, the Tier Totals in each K/A 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 K/A statements.
4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.
5. Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected.

Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.

7** The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution o(system. Refer to Section D.1 .b of ES-401 for the applicable K/As.

8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in other than Category A2 or G* the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams.
9. For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 1 0 CFR 55.43.

ES-401, REV 9 T1G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO 007EK2.02 - -

Reactor Trip Stabilization Recovery 2.6 2.8 Breakers, relays and disconnects j j J J

/1 009EA2.09 Small Break LOCA 1 3 2.8 3.3 j Low-pressure SWS activity monitor 01 5AK1 .01 RCP Malfunctions/4 4.4 4.6 J j j Natural circulation in a nuclear reactor power plant 025AK3.02 Loss ofRi-IR System I 4 3.33.7 j j J Isolation of RHR low-pressure piping prior to pressure increase above specified level O26AA1 .04 Loss of Component CoolingWater/ 8 2.7 2.8 j CRDM high-temperature alarm system 027AK1 Pressurizer Pressure Control System 2.8 3.1 j j j j Expansion ofqtnds as temperature increases Malfunction I 3 038EG2.4.34 Steam Gen. Tube Rupture/3 4.2 4.1 j Knowledge of RO tasks performed outside the main J

control room during an emergency and the resultant operational effects 055EA2.03 Station Blackout /6 3.9 4 LI LI LI LI LI LI LI LI [] [ Acons necessary to restore power 056AG2.1.20 Loss of Off-site Power /6 4.6 4.6 LI LI LI LI LI LI LI LI LI LI AbHity to execute procedure steps.

O57AA1 .04 Loss of Vital AC Inst. Bus /6 3.5 3.6 RWST and VCT valves LI LI LI LI LI LI [] LI LI LI LI 058AK3.Ol Loss of DC Power/6 3.4 3.7 Use of dc control power by D/Gs Page 1 of 2 4/17/2013 8:12AM

ES-401, REV 9 T1G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME/SAFETYFUNCTION: IR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO 062AA2.06 Loss of Nuclear Svc Water I 4 2.8 3.1 J J J j The length of time after the loss of SWS flow to a component before that component may be damaged 065AK3.04 Loss of Instrument Air 1 8 3 3.2 j Cross-over to backup air supplies

[J 077AG2.2.42 Generator Voltage and Electric Grid 3.9 4.6 j Ability to recognize system parameters that are entry-Disturbances I 6 level conditions for Technical Specifications WEO4EK2.2 LOCA Outside Containment I 3 3.8 4.0 j j J Facilitys heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems and relations between the proper operation of these systems to the operation of the facility.

WEO5EK1 .3 Inadequate Heat Transfer Loss of 3.9 4.1 El El El El El El El El El Annunciators and conditions indicating signals, and Secondary Heat Sink I 4 remedial actions associated with the (Loss of Secondary Heat Sink).

WE1 1 EA1 .3 Loss of Emergency Coolant Recirc./4 3.7 4.2 El El LI El Desired operathig results during abnormal and LI LI El []

emergency situations.

WE12EK2.1 -

Steam Line Rupture Excessive Heat 3A31 EEElElElElElElElEl Components and functions of control and safety systems, Transfer I 4 including instrumentation, signals, interlocks, failure modes and automatic and manual features.

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ES-401, REV 9 T1G2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO OO1AA1 .04 Continuous Rod Withdrawal I 1 3.8 3.6 j j j Operating switch for emergency boration motor-operated valve operating switch 036AK3.02 Fuel Handling Accident I 8 2.9 3.6 j j Interlocks associated with fuel handung equipment 037AK1 .02 Steam Generator Tube Leak I 3 3.5 3.9 j j j Leak rate vs. pressure drop J j j 051AA2.02 Loss of Condenser Vacuum I 4 3.9 4.1 j j Conditions requiring reactor and/or turbine trip J

068AA2.05 Control Room Evac. / 8 4.2 4.3 j j j Availability of heat sink 076AK2.OiHighReactor Coolant Activity / 9 6 3 j j Process radiation monitors weO2EG2.2.44 SI Termination /3 4.2 4.4 j j j Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions WEO3EK2.2 LOCA Cooldown - Depress. /4 3.7 4.0 j Facilitys heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems and relations between the proper operation of these systems to the operation of the facility.

WEO8EK3.3 RCSOvercooling - PTS /4 3.7 3.8 El El El El Manipulation of controls required to obtain desired operating results during abnormal and emergency situations.

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ES-401, REV 9 T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO 003K614 Reactor Coolant Pump 2.6 2.9 j j Starting requirements J j 004K2.07 Chemical and Volume Control 2.7 3.2 Heat tracing 004K3.06 Chemical and Volume Control 3.4 3.6 RCS temperature and pressure j J j j 005K5.03 Residual Heat Remov 2.9 3.1 Reacvity effects ofRHRfW water 006A4.O3Emergency Core Cooling 3.5 3.5 j j Transfer from boron storage tank to boron injection tank J J 007K5.02 Pressurizer Relief/Quench Tank 3.1 3.4 Method of forming a steam bubble hi the PZR j

008K3.03 Component Coolhig Water 4.1 4.2 RCP 01 0K1 .08 Pressurizer Pressure Control a23.5 -

LI LI LI LI LI LI LI LI 012A2.05 Reactor Protection 3.1 3.2 Faulty or erratic operation of detectors and function LI LI LI LI LI LI LI LI LI LI generators Ol3KaOl Engineered Safety Features Actuation 3.6 3.8 ESFAS/safeguards equipment control LI LI LI LI LI LI LI LI LI LI 022A3.0l Containment Cooling 4.1 Initia tion of safeguards mode of operation LI LI LI LI LI LI LI LI LI LI Page 1 of 3 4/17/2013 8:12 AM

ES-401, REV 9 T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME/SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO 022K3.01 Containment Cooling 2.9 3.2 El D D D El El El D El El Containment equipment subject to damage by high or low tem perature, hum idity and pressure 025A3.02 Ice Condenser 3.4 DDDDDDDDDD Isolationvalves 026G2.2.36 Containment Spray 3.1 4.2 LI LI El LI D LI LI El LI LI Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions of operations 039A4.04 Main and Reheat Steam 3.8 3.9 El D El LI El LI El El LI i D Emergency feedwater pump turbines 059A2.1 1 Main Feedwater 3.0 3.3 LI LI El El El D D i LI D El Failure of feedwater control system 059K4.16 Main Feedwater 3.1 LI D D Automatic trips for MEW pumps LI D LI LI El D D 061K1.07 Auxiliary/Emergency Feedwater 3.6 Emergency water source El El D El El LI El LI LI D 061K6.01 Auxiliary/Emergency Feedwater 2.5 --:8 Controllers and positioners El D D El D D El D LI 062A1 .01 AC Electncal Distribution 3.4 3.8 SignificanceofD/coadHmits D D LI LI LI El D LI El D 063A1 .01 DC Electrical Disffibutkrn 2.5 3.3 Battery capacity as it is affected by discharge rate D D El El LI D 1 D LI El LI 063A2.01 DC Electrical Distribution 2.5 3.2 DDDDDDDDDD Grounds Page 2 of 3 4/17/2013 8:12AM

ES-401, REV 9 T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO 064G2.4.4 Emergency Diesel Generator 4.5 4.7 EJ j j Ability to recognize abnormal indications for system Ej Ej J j operating parameters which are entry-level conditions for emergency and abnormal operating procedures.

073K4.02 Process Racation Monitoring 3.3 3.9 Letdown isolation on high-RCS activity El El El ] El El El El El El El O76A32 Service Water 3.7 3.7 El El El El El El El El Emergency heat loads El El 076K2X)4 ServftDe Water 2.52 Reactor building closed cooling water El El El El El El El El El El 078K1.Ol Instrument Air 2.8 2.7 Sensorair El El El El El El El El El El 103K4.04 Contnment 25 3.2 Personnel access hatch and emergency access hatch El El El El El El El El El El Page 3of 3 4/17/2013 8:12AM

ES-401, REV 9 T2G2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO I I I I I I I I I I 001K4.23 Control Rod Drive a4 3.8 J J j Rodmotion inhibit PressurizerLevel Control 313.1 OperatknofPZRlevel controllers 015A1 .01 Nuclear Instrumentabon 3.5 3.8 j j j NIS calibration by heat balence j

028K5.04 Hydrogen Recombiner and Purge 2.6 3.2 j j The selective removal of hydrogen j L LI LI El Control 041 K2.02 Steam Dump/TurbineBypass Control 2.8 2.8 CS inverter breakers J J 045K3.Ol Main TurLne Generator 2.9 3.2 El j Remainder of the plant 055A3.03 Condenser Aw Removal 2.5 2.7 Automat diversion of CARS exhaust LI El El El LI El El El El El 056A2.04 Condensate 2.6 2.8 Loss of condensate pumps El El El El El El El El El El 072A4.03 Area Radiation Momtoring 3.1 3.1 Check source for operability demonstration El El El El El El El El El El 075G2.4.l Circulating Water 4.6 4.8 Knowledge of EOP entry conditions and immediate acbon El El El El El El El El El El steps.

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ES-401, REV 9 T3 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO G2.l .32 Conduct of operations 3.8 4.0 j j Ability to explain and apply all system limits and J j precautions.

G2.l .45 Conduct of operations 4.3 4.3 j j j j Ability to identify and interpret diverse indications to validate the response of another indication G2.2.17 EquipmentControl 638 j Knowledge ofthe process for manangmaD1tenance J

activities during power operations.

G2l 8 EquipmentControl 2.6 3.8 j j j Knowledge of the process for managing maitenance activities during shutdown operations.

G2.2.38 Equipment Control 3.6 4.5 Knowledge of conditions and limitations in the facility J j j j license.

G2.3.13 Radiation Control 3.4 3.8 j j j j j Knowledge of radiological safety procedures pertaining to licensed operator duties G2.3.14 Radtion Control 43.8 Knowledge of radiation or contamination hazards that El El El may arise during normal, abnormal, or emergency conditions or activities G2.3.4 Radiation Control 3.23.7 Knowledge of radiation exposure limits under normal and El El El El El El El El El El emergency conditions G2.4.20 Emergency Procedures/Plans 3.8 4.3 Knowledge of operationaHmplications of EOP warmng El El El El El El El El El El cautions and notes.

G2.4.25 Emergency Procedures/Plans 3.3 3.7 Knowledge of fire protection procedures.

El El El El El El El El El El Page 1 of 1 4/17/2013 8:12 AM

ES-401, REV 9 SRO T1G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO 008AG2.l .7 Pressurizer Vapor Space Accident / 3 4.4 4.7 j J j j j Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior and instrument interpretation.

011 EA2.07 Large Break LOCA / 3 3.2 3.4 j That equipment necessary for functioning of critical pump j j j j EJ J water seals is operable 025AA2.01 Loss of RI-IR System /4 2.7 2.9 Proper amperage of running LPI/decay heat removal/RHR pump(s) 056AG2.2.40 Loss of Off-sue Power /6 34 4.7 El El LI El ] Ability to apply technical specifications for a system.

065AA2.06 Loss of Instrument Air /8 3.6 4.2 When to trip reactor if instrument air pressure is de LI El El LI LI El El El El El creasing weO4EG2.4K LOCA Outside Containment /3 3.3 4.0 Knowledge of the specific bases for EOPs.

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ES-401, REV 9 SRO T1G2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR Kl K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO 069AG2.l .23 Loss of CTMT Integrity /5 4.3 4.4 El El El El LI El El El El Ability to perform specific system and integrated plant procedures during all modes of plant operation.

076AG2.4.46 High Reactor Coolant Activity / 9 4.2 4.2 j Ability to verify that the alarms are consistent with the J j j J plant conditions.

WEO3EA2.1 -

LOCA Cooldown Depress. /4 3.4 4.2 El El El El El El El J [] [] [ FaciBty conditions and selection of appropriate procedures during abnormal and emergency operations.

WEO7EA2.2 Saturated Core Cooling Core Cooling 3.3 3.9 El El El El El El El Adherence to appropriate procedures and operation

/4 LI LI El within the limitations in the facilitys license and amendments.

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ES-401, REV 9 SRO T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME/SAFETY FUNCTION: IR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO 005A2.01 Residual Heat Removal 2.7 2.9 El El LI LI Failure modes for pressure, flow, pump motor amps, El LI El motor temperature and tank level instrumentation 008G2.4.50 Corn ponent Cooling Water 4.2 4.0 Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.

026G2.2.25 Containment Spray 3.2 4.2 Knowledge of the bases in Technical Specifications for j j limiting conditions for operations and safety limits.

062A207 AC Ectric Distribution 3.03.4 Consequences of opening a disconnect under load El El LI LI El El LI E] [ El 063G2.4.31 DC Electrical Distribution 4.2 4.1 Knowledge of annunciators alarms, indications or response procedures Page 1 of 1 4/17/2013 8:12AM

ES-401, REV 9 SRO T2G2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME/SAFETY FUNCTION: IR Kl K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO 002G2.2.37 Reactor Coolant 3.6 4.6 EJ D LI LI LI LI LI LI LI [} [] Ability to determine operability and/or availability of safety related equipment 028G2.1.30 Hydrogen Recombiner and Purge 4.4 4.0 LI LI LI LI LI LI LI LI LI LI Ability to locate and operate components, including local Control controls.

079A2.O1 Station Air 2.9 3.2 Cross-connection with lAS LI LI LI LI LI LI LI EiLI LI LI Pagelof 1 4/17/2013 8:12AM

ES-401, REV 9 SRO T3 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR Kl K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO G2.1 .42 Conduct of operations 2.5 3.4 Knowledge of new and spent fuel movement procedures G2.2.15 Equipment Control 3.9 Ability to determine the expected plant configuraon using design and configuration control documentaion G2.2.20 Equipment Control 2.6 3.8 Knowledge of the process for managing troubleshooting activities.

G24 Radiation Control 3.2 3.7 Knowledge of radiation exposure limits under normal and emergency conditions G2.3.7 Radiation Control 3.5 3.6 Ability to comply with radiation work permit requirements during normal or abnormal conditions G2.4.35 Emergency Procedures/Plans 3.8 4.0 Knowledge of local auxiliary operator tasks during emergency and the resultant operational effects G2.4.38 Emergency Procedures/Plans 2.4 Ability to take actions called for in the facility emergency plan, including supporting or acting as emergency coordinator.

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