Text

2009-08-DRAFT Outlines
	ML093200394
Person / Time
Site:	Wolf Creek
Issue date:	11/16/2009
From:	Apger G; NRC Region 4
To:	; Wolf Creek
References
	50-482/09-301
	Download: ML093200394 (43)
	v • d • e

ES-401 Written Examination Outline Form ES-401-2 1

Facility:

Wolf Creek 2009 NRC Examination Date of Exam:

8/17/2009 RO K/A Category Points SRO-Only Points Tier Group K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G

Total A2 G*

Total 1

3 3

18 3

3 6

2 2

2 1

1 2

1 9

2 2

4

1.

Emergency Plant Evolutions Tier Totals 5

5 4

4 5

4 27 5

5 10 1

2 2

2 3

2 2

3 3

3 28 2

3 5

2 0

1 0

1 1

2 1

10 0

2 1

3

2.

Plant Systems Tier Totals 2

3 2

4 3

3 4

4 4

5 4

38 4

4 8

1 2

3 4

1 2

3 4

3. Generic Knowledge & Abilities Categories 2

3 2

3 10 1

2 2

2 7

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., except for one category in Tier 3 of the 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 that are not included on the outline should be added. Refer to section D.1.b of ES-401, for guidance regarding elimination of inappropriate 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 KAs having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

6.

Select SRO topics for Tiers 1 and 2 from the shaded systems and 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 or 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 (IR) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above. If fuel handling equipment is sampled in other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams.

9.

For Tier 3, select topics from Section 2 of the 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 10CFR55.43

ES-401 2

Form ES-401-2 Wolf Creek 2009 NRC Examination Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 EAPE # / Name Safety Function K1 K2 K3 A1 A2 G

K/A Topic(s)

Imp.

Q#

056 / Loss of Off-site Power / 6 X

AA2.32 - Ability to determine and interpret the following as they apply to the Loss of Offsite Power: Transient trend of coolant temperature toward no-load T-ave 4.3 76 011 / Large Break LOCA / 3 X

EA2.14 - Ability to determine or interpret the following as they apply to a Large Break LOCA: Actions to be taken if limits for PTS are violated 4.0 77 058 / Loss of DC Power / 6 X

AA2.02 - Ability to determine and interpret the following as they apply to the Loss of DC Power: 125V dc bus voltage, low/critical low, alarm 3.6 78 E05 / Inadequate Heat Transfer -

Loss of Secondary Heat Sink / 4 X

2.4.18 - Emergency Procedures / Plan:

Knowledge of the specific bases for EOPs.

4.0 79 009 / Small Break LOCA / 3 X

2.1.25 - Conduct of Operations: Ability to interpret reference materials, such as graphs, curves, tables, etc.

4.2 80 E12 / Steam Line Rupture -

Excessive Heat Transfer / 4 X

2.4.2 - Emergency Procedures / Plan:

Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions.

4.6 81 055 / Station Blackout / 6 X

EK1.02 - Knowledge of the operational implications of the following concepts as they apply to the Station Blackout : Natural circulation cooling 4.1 39 054 / Loss of Main Feedwater / 4 X

AK1.02 - Knowledge of the operational implications of the following concepts as they apply to Loss of Main Feedwater (MFW): Effects of feedwater introduction on dry S/G 3.6 40 011 / Large Break LOCA / 3 X

EK1.01 - Knowledge of the operational implications of the following concepts as they apply to the Large Break LOCA:

Natural circulation and cooling, including reflux boiling.

4.1 41 E11 / Loss of Emergency Coolant Recirculation / 4 X

EK2.1 - Knowledge of the interrelations between the (Loss of Emergency Coolant Recirculation) and the following:

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

3.6 42 E12 / Steam Line Rupture -

Excessive Heat Transfer / 4 X

EK2.1 - Knowledge of the interrelations between the (Uncontrolled Depressurization of all Steam Generators) and the following: Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

3.4 43 008 / Pressurizer Vapor Space Accident / 3 X

AK2.01 - Knowledge of the interrelations between the Pressurizer Vapor Space Accident and the following: Valves 2.7 44 007 / Reactor Trip - Stabilization -

Recovery / 1 X

EK3.01 - Knowledge of the reasons for the following as the apply to a reactor trip:

Actions contained in EOP for reactor trip 4.0 45 056 / Loss of Off-site Power / 6 X

AK3.01 - Knowledge of the reasons for the following responses as they apply to the Loss of Offsite Power: Order and time to initiation of power for the load sequencer 3.5 46

ES-401 3

Form ES-401-2 Wolf Creek 2009 NRC Examination Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 EAPE # / Name Safety Function K1 K2 K3 A1 A2 G

K/A Topic(s)

Imp.

Q#

E05 / Inadequate Heat Transfer -

Loss of Secondary Heat Sink / 4 X

EK3.4 - Knowledge of the reasons for the following responses as they apply to the (Loss of Secondary Heat Sink) RO or SRO function within the control room team as appropriate to the assigned position, in such a way that procedures are adhered to and the limitations in the facilities license and amendments are not violated.

3.7 47 E04 / LOCA Outside Containment /

3 X

EA1.2 - Ability to operate and / or monitor the following as they apply to the (LOCA Outside Containment) Operating behavior characteristics of the facility.

3.6 48 038 / Steam Generator Tube Rupture / 3 X

EA1.19 - Ability to operate and monitor the following as they apply to a SGTR: MFW System status indicator 3.4 49 025 / Loss of Residual Heat Removal System / 4 X

AA1.11 - Ability to operate and / or monitor the following as they apply to the Loss of Residual Heat Removal System: Reactor building sump level indicators 2.9 50 027 / Pressurizer Pressure Control System Malfunction / 3 X

AA2.13 - Ability to determine and interpret the following as they apply to the Pressurizer Pressure Control Malfunctions:

Seal return flow 2.8 51 057 / Loss of Vital AC Electrical Instrument Bus / 6 X

AA2.19 - Ability to determine and interpret the following as they apply to the Loss of Vital AC Instrument Bus: The plant automatic actions that will occur on the loss of a vital ac electrical instrument bus 4.0 52 077 / Generator Voltage and Electric Grid Disturbances X

AA2.07 - Ability to determine and interpret the following as they apply to Generator Voltage and Electric Grid Disturbances:

Operational status of engineered safety features 3.6 53 026 / Loss of Component Cooling Water / 8 X

2.1.23 - Conduct of Operations: Ability to perform specific system and integrated plant procedures during all modes of plant operation.

4.3 54 022 / Loss of Reactor Coolant Makeup / 2 X

2.2.40 - Equipment Control: Ability to apply technical specifications for a system.

3.4 55 029 / Anticipated Transient Without Scram (ATWS) / 1 X

2.1.30 - Conduct of Operations: Ability to locate and operate components, including local controls 4.4 56 K/A Category Totals:

3 3

3/3 3/3 Group Point Total:

18/6

ES-401 4

Form ES-401-2 Wolf Creek 2009 NRC Examination Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 2 EAPE # / Name Safety Function K1 K2 K3 A1 A2 G

K/A Topic(s)

Imp.

Q#

037 / Steam Generator Tube Leak /

3 X

AA2.05 - Ability to determine and interpret the following as they apply to the Steam Generator Tube Leak: Past history of leakage with current problem 3.3 82 E10 / Natural Circulation with Steam Void in Vessel with/without RVLIS / 4 X

EA2.1 - Ability to determine and interpret the following as they apply to the (Natural Circulation with Steam Void in Vessel with/without RVLIS) Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

3.9 83 E06 / Degraded Core Cooling / 4 X

2.1.28 - Conduct of Operations: Knowledge of the purpose and function of major system components and controls.

4.1 84 076 / High Reactor Coolant Activity /

9 X

2.4.4 - Emergency Procedures / Plan:

Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures.

4.7 85 061 / Area Radiation Monitoring (ARM) System Alarms / 7 X

AK1.01 - Knowledge of the operational implications of the following concepts as they apply to Area Radiation Monitoring (ARM) System Alarms: Detector limitations 2.5 57 036 / Fuel Handling Incidents / 8 X

AK2.01 - Knowledge of the interrelations between the Fuel Handling Incidents and the following: Fuel handling equipment 2.9 58 E14 / High Containment Pressure /

5 X

EK3.1 - Knowledge of the reasons for the following responses as they apply to the (High Containment Pressure) Facility operating characteristics during transient conditions, including coolant chemistry and the effects of temperature, pressure, and reactivity changes and operating limitations and reasons for these operating characteristics.

3.2 59 051 / Loss of Condenser Vacuum / 4 X

AA1.04 - Ability to operate and / or monitor the following as they apply to the Loss of Condenser Vacuum: Rod position 2.5 60 059 / Accidental Liquid RadWaste Release / 9 X

AA2.04 - Ability to determine and interpret the following as they apply to the Accidental Liquid Radwaste Release: The valve lineup for a release of radioactive liquid 3.2 61 003 / Dropped Control Rod / 1 X

2.2.39 - Equipment Control: Knowledge of less than one hour technical specification action statements for systems.

3.9 62 076 / High Reactor Coolant Activity /

9 X

AK2.01 - Knowledge of the interrelations between the High Reactor Coolant Activity and the following: Process radiation monitors 2.6 63 028 / Pressurizer Level Control Malfunction / 2 X

AA2.03 - Ability to determine and interpret the following as they apply to the Pressurizer Level Control Malfunctions:

Charging subsystem flow indicator and controller 2.8 64 005 / Inoperable/Stuck Control Rod /

1 X

AK1.04 - Knowledge of the operational implications of the following concepts as they apply to Inoperable / Stuck Control Rod: Definitions of axial imbalance, neutron error, power demand, actual power tracking mode, ICS tracking 3.0 65

ES-401 5

Form ES-401-2 Wolf Creek 2009 NRC Examination Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 2 EAPE # / Name Safety Function K1 K2 K3 A1 A2 G

K/A Topic(s)

Imp.

Q#

K/A Category Totals:

2 2

1 1

2/

2 1/

2 Group Point Total:

9/4

ES-401 6

Form ES-401-2 Wolf Creek 2009 NRC Examination Written Examination Outline Plant Systems - Tier 2 Group 1 System # / Name K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G

Imp.

Q#

007 Pressurizer Relief/Quench Tank X

A2.02 - Ability to (a) predict the impacts of the following malfunctions or operations on the P S; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Abnormal pressure in PRT 3.2 86 010 Pressurizer Pressure Control X

A2.03 - Ability to (a) predict the impacts of the following malfunctions or operations on the PZR PCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: PORV failures 4.2 87 022 Containment Cooling X

2.4.9 - Emergency Procedures / Plan:

Knowledge of low power / shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.

4.2 88 003 Reactor Coolant Pump X

2.4.30 - Emergency procedures/plan:

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

4.1 89 039 Main and Reheat Steam X

2.1.25 - Conduct of Operations: Ability to interpret reference materials, such as graphs, curves, tables, etc.

4.2 90 103 Containment X

K1.08 - Knowledge of the physical connections and/or cause-effect relationships between the containment system and the following systems: SIS, including action of safety injection reset 3.6 1

010 Pressurizer Pressure Control X

K1.08 - Knowledge of the physical connections and/or cause-effect relationships between the PZR PCS and the following systems: PZR LCS 3.2 2

003 Reactor Coolant Pump X

K2.01 - Knowledge of bus power supplies to the following: RCPS 3.1 3

006 Emergency Core Cooling X

K2.02 - Knowledge of bus power supplies to the following: Valve operators for accumulators 2.5 4

073 Process Radiation Monitoring X

K3.01 - Knowledge of the effect that a loss or malfunction of the PRM system will have on the following: Radioactive effluent releases 3.6 5

064 Emergency Diesel Generator X

K3.01 - Knowledge of the effect that a loss or malfunction of the ED/G system will have on the following: Systems controlled by automatic loader 3.8 6

076 Service Water X

K4.02 - Knowledge of SWS design feature(s) and/or interlock(s) which provide for the following: Automatic start features associated with SWS pump controls 2.9 7

ES-401 7

Form ES-401-2 Wolf Creek 2009 NRC Examination Written Examination Outline Plant Systems - Tier 2 Group 1 System # / Name K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G

Imp.

Q#

008 Component Cooling Water X

K4.01 - Knowledge of CCWS design feature(s) and/or interlock(s) which provide for the following: Automatic start of standby pump 3.1 8

005 Residual Heat Removal X

K5.09 - Knowledge of the operational implications of the following concepts as they apply the RHRS: Dilution and boration considerations 3.2 9

061 Auxillary/Emergency Feedwater X

K5.03 - Knowledge of the operational implications of the following concepts as the apply to the AFW: Pump head effects when control valve is shut 2.6 10 004 Chemical and Volume Control X

K6.13 - Knowledge of the effect of a loss or malfunction on the following CVCS components: Purpose and function of the boration/dilution batch controller 3.1 11 013 Engineered Safety Features Actuation X

K6.01 - Knowledge of the effect of a loss or malfunction on the following will have on the ESFAS: Sensors and detectors 2.7 12 012 Reactor Protection X

A1.01 - Ability to predict and/or monitor Changes in parameters (to prevent exceeding design limits) associated with operating the RPS controls including: Trip setpoint adjustment 2.9 13 063 DC Electrical Distribution X

A1.01 - Ability to predict and/or monitor changes in parameters associated with operating the dc electrical system controls including: Battery capacity as it is affected by discharge rate 2.5 14 059 Main Feedwater X

A2.01 - Ability to (a) predict the impacts of the following malfunctions or operations on the MFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Feedwater Actuation of AFW System 3.4 15 022 Containment Cooling X

A2.04 - Ability to (a) predict the impacts of the following malfunctions or operations on the CCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of service water 2.9 16 039 Main and Reheat Steam X

A3.02 - Ability to monitor automatic operation of the MRSS, including:

Isolation of the MRSS 3.1 17 026 Containment Spray X

A3.01 - Ability to monitor automatic operation of the CSS, including: Pump starts and correct MOV positioning 4.3 18 062 AC Electrical Distribution X

A4.01 - Ability to manually operate and/or monitor in the control room: All breakers (including available switchyard) 3.3 19 078 Instrument Air X

A4.01 - Ability to manually operate and/or monitor in the control room:

Pressure gauges 3.1 20

ES-401 8

Form ES-401-2 Wolf Creek 2009 NRC Examination Written Examination Outline Plant Systems - Tier 2 Group 1 System # / Name K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G

Imp.

Q#

007 Pressurizer Relief/Quench Tank X

2.4.50 - Emergency Procedures / Plan:

Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.

4.2 21 063 DC Electrical Distribution X

2.4.6 - Emergency Procedures / Plan:

Knowledge of EOP mitigation strategies.

3.7 22 078 Instrument Air X

A3.01 - Ability to monitor automatic operation of the IAS, including: Air pressure 3.1 23 076 Service Water X

A1.02 - Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the SWS controls including: Reactor and turbine building closed cooling water temperatures.

2.6 24 005 Residual Heat Removal X

K4.06 - Knowledge of RHRS design feature(s) and/or interlock(s) which provide or the following: Function of RHR pump miniflow recirculation 2.7 25 026 Containment Spray X

2.2.22 - Equipment Control: Knowledge of limiting conditions for operations and safety limits.

4.0 26 012 Reactor Protection X

A2.05 - Ability to (a) predict the impacts of the following malfunctions or operations on the RPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Faulty or erratic operation of detectors and function generators 3.1 27 004 Chemical and Volume Control X

A4.18 - Ability to manually operate and/or monitor in the control room:

Emergency borate valve 4.3 28 K/A Category Totals:

2 2

2 3

2 2

3 3

/

2 3

3 3

/

3 Group Point Total:

28/5

ES-401 9

Form ES-401-2 Wolf Creek 2009 NRC Examination Written Examination Outline Plant Systems - Tier 2 Group 2 System # / Name K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G

Imp.

Q#

075 Circulating Water X

A2.01 - Ability to (a) predict the impacts of the following malfunctions or operations on the circulating water system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Loss of intake structure 3.2 91 011 Pressurizer Level Control X

2.1.28 - Conduct of Operations:

Knowledge of the purpose and function of major system components and controls.

4.1 92 002 Reactor Coolant X

A2.03 - Ability to (a) predict the impacts of the following malfunctions or operations on the RCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of forced circulation 4.3 93 014 Rod Position Indication X

A4.01 - Ability to manually operate and/or monitor in the control room: Rod selection control 3.3 29 015 Nuclear Instrumentation X

K6.02 - Knowledge of the effect of a loss or malfunction on the following will have on the NIS:

Discriminator/compensation circuits 2.6 30 016 Non-nuclear Instrumentation X

2.4.6 - Emergency Procedures / Plan:

Knowledge of EOP mitigation strategies.

3.7 31 045 Main Turbine Generator X

K4.02 - Knowledge of MT/G system design feature(s) and/or inter-lock(s) which provide for the following:

Automatic shut of reheat stop valves as well as main control valves when tripping turbine 2.5 32 027 Containment Iodine Removal X

K2.01 - Knowledge of bus power supplies to the following: Fans 3.1 33 068 Liquid Radwaste X

K1.07 - Knowledge of the physical connections and/or cause effect relationships between the Liquid Radwaste System and the following systems: Sources of liquid wastes for LRS 2.7 34 028 Hydrogen Recombiner and Purge Control X

A1.02 - Ability to predict and/or monitor changes in parameter (to prevent exceeding design limits) associated with operating the HRPS controls including: Containment pressure 3.4 35 086 Fire Protection X

A4.01 - Ability to manually operate and/or monitor in the control room: Fire Water pumps 3.3 36 017 In-core Temperature Monitor X

A3.02 - Ability to monitor automatic operation of the ITM system including:

Measurement of in-core thermocouple temperatures at panel outside control room 3.4 37

ES-401 10 Form ES-401-2 Wolf Creek 2009 NRC Examination Written Examination Outline Plant Systems - Tier 2 Group 2 System # / Name K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G

Imp.

Q#

041 Steam Dump/Turbine Bypass Control X

A2.02 - Ability to (a) predict the impacts of the following malfunctions or operations on the SDS; and (b) based on those predictions or mitigate the consequences of those malfunctions or operations: Steam valve stuck open 3.6 38 K/A Category Totals:

1 1

0 1

1 1

/

2 1

/

1 Group Point Total:

10/3

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

Form ES-401-3 11 Facility:

Wolf Creek 2009 NRC Examination Date:

8/17/2009 RO SRO-Only Category K/A #

Topic IR Q#

IR Q#

2.1.36 Knowledge of procedures and limitations involved in core alterations.

4.1 94 2.1.25 Ability to interpret reference materials, such as graphs, curves, tables, etc.

3.9 66 2.1.13 Knowledge of facility requirements for controlling vital / controlled access.

2.5 67

1.

Conduct of Operations Subtotal 2

1 2.2.43 Knowledge of the process used to track inoperable alarms.

3.3 95 2.2.6 Knowledge of the process for making changes to procedures.

3.6 98 2.2.35 Ability to determine Technical Specification Mode of Operation.

3.6 68 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 69 2.2.7 Knowledge of the process for conducting special or infrequent tests.

2.9 75

2.

Equipment Control Subtotal 3

2 2.3.13 Knowledge of Radiological Safety Procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high radiation areas, aligning filters, etc.

3.8 96 2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions.

3.7 99 2.3.14 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

3.4 70 2.3.15 Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

2.9 71

3.

Radiation Control

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

Form ES-401-3 12 Subtotal 2

2 2.4.5 Knowledge of the organization of the operating procedures network for normal, abnormal, and emergency evolutions.

4.3 97 2.4.45 Ability to prioritize and interpret the significance of each annunciator or alarm.

4.3 100 2.4.25 Knowledge of fire protection procedures.

3.3 72 2.4.11 Knowledge of abnormal condition procedures.

4.0 73 2.4.34 Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects.

4.2 74

4.

Emergency Procedures /

Plan Subtotal 3

2 Tier 3 Point Total 10 7

ES-401 Record of Rejected K/As Form ES-401-4 13 Tier / Group Randomly Selected K/A Reason for Rejection 1 / 1 065 / 2.1.30 replaced by 029

/ 2.1.30 Excessive similar topic coverage for Instrument Air system 2 / 1 103 / K1.07 replaced by 103

/ K1.08 System does not exist at this facility 2 / 1 005 / K4.12 replaced by 005

/ K4.06 Piggyback mode not used at facility 2 / 2 014 / A4.03 replaced by 014

/ A4.01 Operationally insignificant and indication is not available in control room 2 / 2 072 / K1.03 replaced by 072

/ A4.01 No interface between ARMs and FHB Ventilation at this facility 2 / 2 072 / K1.02 replaced by 072

/ A4.01 This was a replacement for K1.03 but this is also not supported at facility.

2 / 2 072 / K4.02 replaced by 072

/ A4.01 This topic was a random reselection but topic not supported by facility design 2 / 1 059 / A2.04 replaced by 059

/ A2.01 Original topic was direct overlap with another test item on exam (Question 40) 2 / 2 068 K5.04 replaced by 068 K1.07 Original topic not operationally significant and no facility reference to support a test item at the license level required for examination 2 / 2 072 A4.01 replaced by 086 A4.01 Original topic operationally insignificant and function not performed by operations staff. No operations references support topic.

2 / 1 007 A2.06 replaced by 007 A2.02 Formation of a pressurizer bubble has no impact on the PRT, and the PRT is not used in any way for bubble formation at this facility.

2 / 1 003 G2.2.36 replaced by 003 G2.4.30 There is no LCO related to component related to degraded power sources, and the reference for degraded power sources is very limited, resulting in overlap with Question 53.

ES-401 Record of Rejected K/As Form ES-401-4 14

ES-301, Rev. 9 Administrative Topics Outline Form ES-301-1

{PRIVATE }Facility: ______Wolf Creek________ Date of Examination: __Aug 31 - Sept 4, 2009___

Examination Level: RO SRO X Operating Test Number: __________

Administrative Topic (see Note)

Type Code*

Describe activity to be performed Conduct of Operations (S.A.1.a)

R, M Complete a 1/M plot and determine the estimated critical position and required actions.

2.1.23 Ability to perform specific system and integrated plant procedures during all modes of operation. (CFR 41.10 / 43.5 /

45.2 / 45.6) (RO/SRO IR = 4.3 / 4.4)

Conduct of Operations (S.A.1.b)

N, R Review completed dilution requirement calculation for one hour after a power reduction.

2.1.37 Knowledge of procedures, guidelines or limitations associated with reactivity management. (CFR 41.1 / 43.6 /

45.6) (RO /SRO IR = 4.3 / 4.6)

Equipment Control (S.A.2)

R, N Using a completed surveillance (STS AL-101, MDAFW Pump A Inservice Pump Test), evaluate acceptance criteria.

2.2.12 Knowledge of surveillance procedures (CFR 41.10 /

45.13) (RO / SRO IR = 3.7 / 4.1)

Radiation Control (S.A.3)

N, R Given a Liquid Release Permit determine if it is ready to be authorized for release to the environment.

2.3.6 Ability to approve release permits. (CFR 41.13 / 43.4 /

45.10) (RO / SRO IR = 2.0 / 3.8)

Emergency Plan (S.A.4)

S, N In the simulator setting, perform the E-Plan classification.

2.4.41 Knowledge of the emergency action level thresholds and classifications. (CFR 41.10 / 43.5 / 45.11) (RO/SRO IR = 2.9 /

4.6)

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)

ES-301, Rev. 9 Administrative Topics Outline Form ES-301-1

{PRIVATE }Facility: ______Wolf Creek________ Date of Examination: __Aug 31 - Sept 4, 2009___

Examination Level: RO x SRO Operating Test Number: __________

Administrative Topic (see Note)

Type Code*

Describe activity to be performed Conduct of Operations (R.A.1.a)

R, M Complete 1/M plot and determine the estimated critical position.

2.1.23 Ability to perform specific system and integrated plant procedures during all modes of operation. (CFR 41.10 / 43.5 /

45.2 / 45.6) (RO/SRO IR = 4.3 / 4.4)

Conduct of Operations (R.A.1.b)

R, N Determine dilution requirements for one hour after a power reduction.

2.1.37 Knowledge of procedures, guidelines or limitations associated with reactivity management. (CFR 41.1 / 43.6 /

45.6) (RO /SRO IR = 4.3 / 4.6)

Equipment Control (R.A.2)

N, R Complete the calculation for surveillance STS AL-101, MDAFW Pump A Inservice Pump Test, evaluate/recommend operable/inoperable and required actions.

2.2.12 Knowledge of surveillance procedures (CFR 41.10 /

45.13) (RO / SRO IR = 3.7 / 4.1)

Radiation Control (R.A.3)

R, N Given radiological conditions evaluate the most efficient method to limit radiological exposure based on DAC hours.

2.3.12 Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc. (CFR 41.12 / 45.9 /

45.10) (RO/SRO IR = 3.2 / 3.7)

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)

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

Wolf Creek Date of Examination: Aug. 31 -

Sept. 4, 2009 Examination Level: RO SRO Operating Test Number:

Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

Bolded font = alternate path System / JPM Title Type Code*

Safety Function

a. 014 - Rod Position Indication System S1 -- Perform actions to retrieve a dropped or misaligned control rod.

(A2.04)

RO/SRO-I N, E, S 1

b. 013 - Engineered Safety Features Actuation System (ESFAS)

S2 -- Perform actions to ensure CSAS actuated correctly. (Auto CSAS failed, candidate manually initiates CSAS)

(A4.01)

PRA: ESFAS is a Top 10 Risk Significant System at Wolf Creek RO/SRO-I/SRO-U N, EN, S, A 2

c. 006 - Emergency Core Cooling System (ECCS)

S3 -- Perform action to ensure transfer of ECCS flowpath. (EJ HIS-8840 fails to open)

(A3.08)

PRA: BN & EJ are Top 10 Risk Significant System.

RO/SRO-I M, A, S, E 3

d. 076 - Service Water System (SWS)

S4 -- Perform actions to start Service Water pump or an Essential Service Water pump. (Service water pumps fail to start)

(A2.01)

PRA: Core Damage Frequency by Initiating Event (Loss of Service Water)

RO/SRO-I N, S, A 4S

e. 071 - Waste Gas Disposal System (WGDS)

S5 -- Perform actions to set process radiation monitor alarm setpoints for Vent Release Permit for Waste Gas Decay Tank.

(A4.25)

RO/SRO-I/SRO-U N, S 9

f. 016 - Non-Nuclear Instrumentation System S6 -- Perform actions to manually trip the Main Turbine. (Auto Turbine trip fails)

(K4.03)

RO/SRO-U N, A, S 7

g. 029 - Containment Purge System (CPS)

S7 -- Perform actions to ensure Containment Purge System isolation. (Auto CPIS fails, manual action required)

(A3.01)

RO/SRO-I N, A, S 8

h. 007 - Pressurizer Relief Tank (PRT)

S8 -- Perform actions to identify and isolate stuck open PORV.

(A2.01)

RO/SRO-I N, S 5

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

i. 063 - DC Electrical Distribution System P1 -- Perform actions of EMG C-0 Attachment C, DC Load Shed for NK11, NK12, NK13 and NK14.

(K2.01)

PRA: Risk significant system (NK); Core Damaging Frequency by Initiating Event (Loss of Offsite Power); CD Events - LERF (P = 2004 NRC exam)

RO/SRO-I/SRO-U D, P, L, E 6

j. 061 - Auxiliary / Emergency Feedwater (AFW) System P2 -- Perform actions to cooldown the TDAFWP piping to restore AFW to > 270,000 LBM per hour.

(A2.06)

PRA: Risk significant system (AL)

RO/SRO-I N, EN, L 4S

k. 006 - Emergency Core Cooling System (ECCS)

P3 -- Perform Attachment B of OFN BG-045, Gas Binding of CCPs or SI Pumps - vent the SI pump due to gas binding.

(A2.04)

SOER 97-1, Potential Loss of High Pressure Injection and Charging Capability from Gas Intrusion RO/SRO-I/SRO-U N, R, E 3

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 (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power / Shutdown (N)ew or (M)odified from bank including 1(A)

(P)revious 2 exams (R)CA (S)imulator 4-6 / 4-6 / 2-3 9 / 8 / 4 1 / 1 / 1

- / - / 1 (control room system) 1 / 1 / 1 2 / 2 / 1 3 / 3 / 2 (randomly selected) 1 / 1 / 1

Appendix D NUREG 1021 Revision 9 1

Appendix D Scenario Outline Form ES-D-1 Facility: __Wolf Creek_______ Scenario No.: ___1____ Op-Test No.: _______

Examiners: ____________________________ Operators: __________________________

Initial Conditions: The plant is at 100% power, Middle of Core life. The A ESW pump and the A EDG are tagged out of service.

Turnover: The A ESW pump is OOS for an upper motor bearing replacement, which causes the A EDG to be OOS. The National Weather Service has issued a Severe Thunderstorm Warning for Shawnee and Butler counties. Topeka System Dispatch (TSD) has notified WC of abnormal grid conditions based on weather conditions. Shift orders are to maneuver the plant generator output per TSD when conditions permit. TSD reports that under the current grid conditions that predicted grid voltage is 98.7% if WC were to come offline. OFN AF-025 Unit Limitations has been completed for the current plant conditions, and no further actions are required.

Event No.

Malf. No.

Event Type*

Event Description 1

t+1 mSE03A I - ATC, SRO Power range instrument SE NI-41B fails high. Rods step in. Manual Rod Control must be selected (or the reactor trips).

2 t+9 mBB21B fails to 2508 psig I - ATC, SRO Pressurizer Pressure Channel 456 fails high (PORV 456 cycles) 3 t+16 mSY03F R -- ATC N - BOP, SRO Loss of 345 kV Benton line. TSD will request Wolf Creek to lower load to 950 Mwe.

4 t+25 IRF pMA02 to 310 kV C -- All Grid voltage droops to 89.9% (310kV). Class 1E busses NB01/2 voltage will decrease to the point where the degraded voltage alarms come in.

mNE04B After a 94 sec time delay, the busses will shed and attempt to repower on the EDGs. Both EDGs are inoperable at this point.

5 t+28 mAL01 C - BOP, SRO The TDAFP fails to start in Auto, but can be started in manual. (CT Manually start the Turbine Driven Auxiliary Feed Pump)

Formatted

Appendix D NUREG 1021 Revision 9 2

6 t+28 mSY01 M -- All Loss All AC Power. The crew will enter EMG C-0, Loss of All AC Power, as neither safeguards bus is energized.

(CT Loss of All AC Power entry and response)

(CT Depressurize intact Steam Generator(s) at maximum rate) 7 t+35 Delete mNE04B M - All NB01/2 bus is restored (EDG B supplying NB02 bus)

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

Appendix D NUREG 1021 Revision 9 3

Appendix D Scenario Outline Form ES-D-1 Scenario Summary:

The crew assumes the watch at MOL core conditions. A Essential Service Water pumped is tagged out to replace a faulty motor bearing, necessitating the tagout of the A Emergency Diesel Generator.

The turnover will include weather conditions that result in an unstable grid. Westar energy has predetermined actions for unstable grid conditions, and notifying Wolf Creek (WC) of these conditions and predicting the effect of the loss of WCs generation capability are specified in Westars directives.

Power Range channel SE NI-41B fails high. The crew will refer to OFN SB-008, Instrument Malfunction, Attachment R:

Verify there is no turbine runback

Place Rod control in manual

Bypass the failed channel

Return Rod Control to Automatic Pressurizer Pressure channel BB PI-456 fails high. The crew will refer to OFN SB-008, Instrument Malfunction, Attachment K:

Identify failed channel

Take manual control of master pressure controller

Select out failed channel

Return system to auto When the 345 kV line is lost, TSD will request the crew to decrease generator output to 950 Mwe. The crew will refer to OFN AF-025, Unit Limitations. Actions taken by the crew will include:

Reduce Turbine load and reactor power per General Operating Procedures As the grid continues to degrade, the voltage will decrease to the point that Safeguards busses NB01/2 Degraded Voltage protection circuits are activated and function. The end result of the degraded voltage will be both NB busses deenergized with the balance of plant powered and the reactor not tripped. Shortly thereafter, the reactor will trip from a turbine trip that is caused by the unstable grid.

The Major event is a complete Loss of all AC power. The crew will respond using EMG C-0, Loss of All AC Power. Eventually EDG B will be available to power NB02.

Formatted: Bullets and Numbering Formatted: Bullets and Numbering Formatted Formatted: Bullets and Numbering

Appendix D NUREG 1021 Revision 9 4

The major actions of EMG C-0 are:

Perform Immediate Actions

Restore AC power

Maintain plant conditions for optimal recovery

Evaluate energized AC emergency bus

Select recovery guideline after AC power restoration Critical Tasks for this scenario are:

Manually start the Turbine Driven Auxiliary Feed Pump as this is the only method available to feed the steam generators during the rapid plant cooldown.

Loss of All AC Power entry and response.

Depressurize intact Steam Generator(s) at maximum rate to minimize RCS inventory loss.

Probabilistic Risk Analysis for scenario includes:

Core Damage Frequency By Event Tree:

Event Tree Core Damage Frequency (/yr)

Percent Contribution Station Blackout 6.46E-06 35.79%

Station blackout event is also very high on the Initiating event contributing to Core Damage Frequency.

Technical Specifications (TS) for this scenario:

Event 1: TS 3.3.1; Reactor Trip System Instrumentation, Table 3.3.1-1, FU 2a, 3a, 3b, 6, 18c, 18d, 18e Cond A: Enter the condition referenced in the table Cond D: Perform SR 3.2.4.2 (QPTR, 12 hrs) and place channel in trip condition (72 hr)

Cond E: One channel inoperable - place channel in trip condition (72 hrs)

Cond T & S: One hour to verify P-8, P-9 and P-10 interlocks are in their required state for plant condition TR 3.3.17, Reactivity Control and Power Distribution Alarms Cond A: Perform SR 3.2.3.1 (once per hr)

Cond D: Perform SR 3.2.4.1 or SR 3.2.4.2 as applicable Event 2: TS 3.3.1; Table 3.3.1-1, FU 6, 8a and 8b Cond A: Enter the condition referenced in the table Cond E: One channel inoperable - place channel in trip condition (72 hrs)

Cond M: One channel inoperable - place channel in trip condition (72 hr)

TS 3.3.2, ESFAS Instrumentation, Table 3.3.2-1, FU 1d, and 8b Cond A: Enter the condition referenced in the table Cond D: One channel inoperable - place channel in trip condition (72 hr)

Formatted: Bullets and Numbering Formatted: Bullets and Numbering

Appendix D NUREG 1021 Revision 9 5

Cond L: One hour to verify P-11 interlock in the required state for plant condition TS 3.3.6, Containment Purge Isolation Instrumentation - Refer to only, no Conditions to enter TS 3.3.7, Control Room Ventilation System Actuation Instrumentation - Refer to only, no Conditions to enter Event 4: TS 3.8.1, AC Sources - Operating Cond A: One offsite circuit inoperable, perform SR 3.8.1.1 for Operable offsite circuit (1 hr - STS NB-005)

Cond B: One DG inoperable, perform SR 3.8.1.1 for Operable offsite circuit (1 hr - STS NB-005)

Cond D: Two offsite circuits inoperable Cond E: One offsite and one DG inoperable Cond F: Two DG inoperable Cond I: Three or more required AC sources inoperable (TS 3.0.3 immediately)

TS 3.8.4, DC Sources - Operating Cond A: One DC electrical power subsystem inoperable (and TS 3.0.3)

TS 3.8.9, Distribution Systems - Operating Cond A: Enter applicable cond/req. actions LCO 3.7.8 (immediately)

Cond B: Restore AC electrical power distribution subsystem to operable (8 hr)

Cond F: Two trains inoperable (TS 3.0.3 - immediately)

Appendix D Scenario Outline Form ES-D-1 Appendix D NUREG 1021 Revision 9 1

Facility: __Wolf Creek________________ Scenario No.: ___2____ Op-Test No.: _____

Examiners: ____________________________ Operators: _____________________

Initial Conditions: The plant is restarting after a forced outage that lasted 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months . The Main Turbine is at 1800 RPM ready to be synchronized to the grid. The plant is at BOL conditions.

RCS boron concentration is @ 2033 ppm. Rod Control is in Manual with bank D at 180 steps.

GEN 00-003, Hot Standby to Minimum Load, is complete to step 6.41. SYS AC-120, Main Turbine Startup, is complete up to step 6.4, Synchronizing the Main Generator.

Turnover: Shift orders are to continue the plant startup. Complete SYS AC-120 to synchronize the Main Generator. Fuel conditioning limits are not in effect. The B Stator Cooling water pump is Out of Service for a rework condition following the outage. It is currently in PMT and is expected back in 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months .

Event No.

Malf. No.

Event Type*

Event Description 1

t + 1 N - All Synchronize the Main Generator to the grid and increase generator load to prevent reverse power.

2 t + 15 mBG01A R -- ATC I -- SRO Volume Control Tank Level control channel BG LI-112 fails low. The result is a 1/2 swapover from the VCT to the RWST.

3 t + 22 mAE15A 4

I - BOP, SRO Steam Generator Level Control channel AE LI-551 fails low.

4 t + 25 mAC06H C -- All The turbine trips on HI-HI bearing vibration.

5 t + 32 mAE14A C -- All A feedwater leak (8E6 lbm/hr) in Containment develops on the A Steam Generator feedline. (CT Isolate faulted Steam Generator) 6 t + 32 mSF17A mSF17B M -- All Failure of the RX to trip in Auto and Manual (CT Direct local Reactor Trip) 7 t + 32 mAL04A C - BOP, SRO Post trip malfunction Failure of the A MDAFP to auto start (CT Manually start A MDAFP)

Appendix D Scenario Outline Form ES-D-1 Appendix D NUREG 1021 Revision 9 2

8 t + 32 mSA27A B08, 09, 10 & 07 C - BOP, SRO Post trip malfunction Failure of the MSIVs to auto close (CT Manually close the MSIVs)

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

Appendix D Scenario Outline Form ES-D-1 Appendix D NUREG 1021 Revision 9 3

Scenario Summary:

The crew assumes the watch with the Main Turbine at 1800 RPM and ready to be synchronized to the grid. This is a normal evolution with no malfunctions. This event is complete when the generator is synced and some load is assumed by the generator to prevent reverse power. This is not a reactivity event at Wolf Creek because reactor power is increased prior to generator synchronization using the Steam Dump System.

This strategy allows a bumpless transfer.

At the discretion on the Lead Evaluator the malfunction for VCT level will be activated.

The result of the failure is that the VCT outlet valve that is controlled by the BG LI-112 channel will close and the RWST valve to the VCT will open. This event will require the SRO to transition to OFN SB-008, Instrument Malfunctions, and select Attachment U for guidance. Attachment U will direct the crew to isolate letdown flow, reduce charging to minimum and take local actions to close and deenergize the RWST to VCT valve.

Steam Generator Level channel AE LI-551 fails low. The crew will refer to OFN SB-008, Instrument Malfunctions, Attachment F. At low power the SGs are fed using the Main Feed Regulating Bypass valves. When using the bypass valves the control system uses actual level and nuclear power (programmed level) to develop a signal to the positioner. When level channel AE LI-551 fails low, the error signal will be large in the open valve direction. The operator will have to:

Take manual control of the Feed Reg Bypass valve

Match Steam flow and Feed flow

Select out the failed channel on RL006

Return the valve to automatic control When the Turbine Hi-Hi Vibration malfunction is activated, the Main Turbine will trip.

Since Reactor Power is less than 49% (P-9), the reactor will not trip. The crew will respond using OFN MA-001, Load Rejection or Turbine Trip.

At the discretion of the Lead Evaluator, the malfunction for a Steam Generator A feedwater leak inside Containment will be activated. This malfunction will result in:

A Steamline Pressure Safety Injection (SI), Reactor Trip and High Containment Pressure SI actuation. The crew will utilize EMG E-0, Reactor Trip or Safety Injection.

Major event: The reactor will fail to trip in Auto and Manual (Major event). The crew will transition to EMG FR-S1, Response to Nuclear Power Generation/ATWT, where local actions to trip the reactor will be successful. The crew will transition back to EMG E-0, Reactor trip or Safety Injection.

The A MDAFP will fail to start in Auto. It will start in manual.

The MSIVs will fail to close in Auto. They can be manually closed.

Appendix D Scenario Outline Form ES-D-1 Facility: __Wolf Creek________________ Scenario No.: ___3____

Op-Test No.: _______

Examiners: ____________________________ Operators:

Initial Conditions: ____100% power, EOL_________________________________________

Turnover: EDG B out of service for MTN PMs - expected back in six hours. TS 3.8.1 conditions A & B entered; STS NB-005, Breaker Alignment Verification has been completed

- due in seven hours.

MDAFW pump B tagged out/out of service for due to Emergent work (SM Concern).

Expected return is six hours. TS 3.7.5 condition B entered.

Event No.

Malf. No.

Event Type*

Event Description 1

t+2 mBB22 A

I --

ATC, SRO PZR level channel BB LI-459A failure high.

OFN SB-008, Instrument Malfunctions, Attachment J 2

t+8 bkrDPA D01A R -

ATC C -

BOP, SRO Condensate pump A trip OFN AF-025, Unit Limitations, Attachment A OFN MA-038, Rapid Plant Shutdown 3

t+25 ANN-E098 ANN-D099 ANN-C098 mSF15A mSF15B M-All Seismic event followed by a Reactor trip occurs.

EMG E-0, Reactor Trip or Safety Injection This event series sets up the scenario for the Major event EMG FR-H1, Response to Loss of Secondary Heat Sink.

mNB01 mNB02 (Post Reactor trip) NB01 & NB02 trip 4

t+25 mNE02 A

C --

ATC, SRO (Post reactor trip) EDG A autostart feature disabled - manual available (CT - start EDG A in order to energize NB01 bus)

Recall NB02 bus unavailable because EDG B out of service as part of Turnover item.

mAL02 mtrDPA L01A mBG13 A

TDAFW pump trip (broken linkage)

MDAFW pump A trip (shaft seizure)

CCP A trips due to overcurrent 5

t+25 M -

All Loss of all Auxiliary Feedwater EMG FR-H1, Response to Loss of Secondary Heat Sink (CT - Establish RCS bleed and feed before Steam Generators dry out)

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

Appendix D Scenario Outline Form ES-D-1 Scenario summary:

Plant is at 100% power, End of life (EOL). Emergency Diesel Generator (EDG) B is tagged out/out of service for MTN - Preventative Maintenance (PM). Expected return is six hours.

Technical Specifications (TS) 3.8.1 condition A & B entered. STS NB-005, Breaker Alignment Verification, complete (due in seven hours).

Motor Drive Auxiliary Feedwater Pump (MDAFWP) B is tagged out/out of service for due to Emergent work (SM Concern). Expected return is six hours. Technical Specification 3.7.5 condition B entered.

Pressurizer (PZR) level (controlling) channel BB LI-459A fails high. Crew responds using OFN SB-008, Instrument Malfunctions, Attachment J.

Attachment J: Select an alternate level channel as the controlling channel.

Crew stabilizes the plant.

Condensate pump A trips. The crew responds by entering OFN AF-025, Unit Limitations.

Attachment A will require a downpower evolution.

Downpower guidance per OFN AF-025: If one condensate pump is lost, reduce power as necessary to maintain Main Feed Pump (MFP) suction pressure greater than 340 psig on AEP0006 for A MFP and AEP0005 for B MFP.

Crew responds by performing a downpower.

Crew stabilizes the plant.

A seismic event occurs resulting in an inadvertent reactor trip. EMG E-0, Reactor Trip or Safety Injection is entered. (Major event)

Post trip, both NB01 and NB02 busses trip. Due to the Maintenance PMs (see Turnover item),

Emergency Diesel Generator B will not start and load onto NB02 bus. Emergency Diesel Generator A must be manually started by the Control Room and then it will load onto NB01 bus.

As the scenario progresses, the Turbine Driven Auxiliary Feedwater Pump (TDAFW pump) trips due to broken linkage and the Motor Driven Auxiliary Feedwater Pump A will trip due to shaft seizure and BOTH cannot be restarted. Motor Driven Auxiliary Feedwater Pump B cannot be started (see Turnover item). No Auxiliary Feedwater (AFW) to the Steam Generators is available. Note that the running CCP (CCP A) trips off due to overcurrent.

The crew enters EMG FR-H1, Response to Loss of Secondary Heat Sink (Major event).

Mitigation strategy: the crew uses the Foldout page of EMG FR-H1, Bleed and Feed.

EMG FR-H1, Response to Loss of Secondary Heat Sink, Major Actions:

Attempt restoration of RCS bleed and feed flow to the S/Gs

Initiation of RCS bleed and feed heat removal

Restore and verify secondary heat sink

Termination of RCS bleed and feed heat removal

Appendix D Scenario Outline Form ES-D-1 Facility: __Wolf Creek________________ Scenario No.: ___4____

Op-Test No.: _______

Examiners: ____________________________ Operators:

Initial Conditions: 68% power Turnover: Load reduction in progress per GEN 00-004, Power Operations, section 6.2, in order to remove A Main Feed pump from service due to high vibration.

Use SYS AE-320, Turbine Driven Main Feedwater Pump Shutdown.

Event No.

Malf.

No.

Event Type*

Event Description 1

t+1 R -

ATC N -

BOP, SRO Reduce power using GEN 00-004, Power Operations and remove A Main Feed Pump from service using SYS AE-320, Turbine Driven Main Feedwater Pump Shutdown (crew may reference / enter OFN AF-025, Unit Limitations, also) 2 t+15 mBB23 A

I -

ATC, SRO Pressurizer Spray valve (BB PCV-455B) fails open because PZR Spray Controller BB PK-455B fails high - manual control available OFN SB-008, Instrument Malfunctions, Attachment V 3

t+20 mAE15 C4 I --

BOP, SRO S/G C level AE LI-553 failure high OFN SB-008, Instrument Malfunctions, Attachment F ALR 00-110C, SG C Flow Mismatch ALR 00-110B, SG C Lev Dev 4

t+27 mBG13 C

C --

ATC, SRO Normal Charging Pump (NCP) trip; a Centrifugal Charging Pump must be started, letdown restored etc ALR 00-042A, Charging Line Flow HiLo (SYS BG-120, CVCS Startup or SYS BG-201, Shifting Charging Pumps -

either may be used to restore letdown)

ALR 00-042E, Charging Pump Trouble (Step 7 re-establishes letdown) 5 t+35 mBB02 B

M -- All 500 gpm Steam Generator Tube Rupture on S/G A OFN BB-07A, Steam Generator Tube Leakage (eventually EMG E-0, Reactor Trip or Safety Injection & EMG E-3, Steam Generator Tube Rupture)

5 EMG E-3, Steam Generator Tube Rupture actions:

CT - Isolate feed flow to the ruptured SG.

CT - Cooldown & Depressurize RCS to meet the following SI termination criteria before the end of the scenario.

RCS subcooling greater than 30°F

RCS pressure - stable or increasing

PZR level greater than 6%

6 t+43 mSA27 EM01 and mSA27 EM02 C -

ATC, SRO Post trip: BIT outlet valves (EM HIS-8801A and EM HIS-8801B) do not open. Manual open available CT - Open BIT outlet valves (EM HIS-8801A and EM HIS-8801B) before the end of the scenario or before needless red or orange path occurs.

EMG E-0, Reactor Trip or Safety Injection, Attachment F or allowed post Immediate Action completion (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

Appendix D Scenario Outline Form ES-D-1 Scenario summary Plant is at 68% power with a load reduction in progress to remove A Main Feed Pump (MFP) from service due to high vibration. GEN 00-004, Power Operations, section 6.2 in progress for load reduction. Crew will use SYS AE-320, Turbine Driven Main Feedwater Pump Shutdown, to remove A MFP from service.

Crew reduces power per GEN 00-004, Power Operations, section 6.2, Power Decrease. Crew may reference OFN AF-025, Unit Limitations, Attachment A (one main feedwater pump from service requires plant at 62% power). Crew removes A MFP from service using SYS AE-320, Turbine Driven Main Feedwater Pump Shutdown (see step 6.2.12 of GEN 00-004).

Crew stabilizes unit at lower power and removes pump from service.

Pressurizer (PZR) Spray Valve BB PCV-455B fails open due to PZR Spray Controller BB PK-455B failure (high). Manual control of BB PK-455B available.

Crew responds by performing OFN SB-008, Instrument Malfunctions, Attachment V.

Crew stabilizes the plant.

Steam Generator (S/G) C level channel AE LI-553 failure high. Crew responds by performing OFN SB-008, Instrument Malfunctions, Attachment F. An alternate level channel is selected for control.

ALR 00-110C, SG C Flow Mismatch or ALR 00-110B, SG C Lev Dev, may be used prior to entry of OFN SB-008, Instrument Malfunctions.

Crew stabilizes the plant.

Normal Charging Pump (NCP) trip. Crew responds using ALR guidance. A Centrifugal Charging Pump is started. Normal letdown must be restored (either using ALR guidance or SYS BG-120, CVCS Startup or SYS BG-201, Shifting Charging Pumps).

ALR 00-042A, Charging Line Flow HiLo

ALR 00-042E, Charging Pump Trouble

Crew stabilizes the plant.

Major event: A 500 gpm Steam Generator Tube Rupture (SGTR) on S/G A is diagnosed per OFN BB-07A, Steam Generator Tube Leakage.

SGTR requiring a manual Reactor trip and Safety Injection Signal actuation per Foldout page criteria of OFN BB-07A. Mitigation procedure will be EMG E-3, Steam Generator Tube Rupture.

1.

Crew responds using OFN BB-07A, Steam Generator Tube Leakage.

2.

Crew responds using EMG E-0, Reactor Trip or Safety Injection.

3.

Crew responds using EMG E-3, Steam Generator Tube Rupture.

EMG E-3, Steam Generator Tube Rupture, Major Actions:

a. Identify and isolate ruptured S/Gs

b. Cooldown and establish RCS subcooling margin

c. Depressurize RCS to restore inventory

d. Terminate SI to stop primary to secondary leakage

e. Prepare for cooldown to cold shutdown Scenario has three Critical Tasks (CT):

1. Open BIT outlet valves EM HIS-8801A & EM HIS-8801B

2. Crew isolates feed flow to ruptured steam generator.

3. Cooldown and Depressurize the RCS to meet Safety Injection termination criteria using EMG E-3, Steam Generator Tube Rupture

-- RCS subcooling greater than 30°F

-- RCS pressure stable or increasing

-- PZR level greater than 6%

Probabilistic Risk Analysis for this scenario includes:

Top 10 Human Action Failures by the Importance Measure Rankings F-V rank 1 Failure of RCS Cooldown & Depressurize - SGR event Core Damage Frequency by Initiating Event and by Event Tree and by Larger Early Release Frequency Consequence Steam Generator Tube Rupture Technical Specifications exercised:

Event 3 TS 3.3.1, Reactor Trip System Instrumentation, Table 3.3.1-1, FU 14 Cond A: Function / channel inoperable - enter condition per table (immediately)

Cond E: One channel inoperable - place channel in trip (72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months )

TS 3.3.2, Engineered Safety Features Actuation System Instrumentation, table 3.3.2-1, FU 5c and 6d Cond A: Function / channel inoperable - enter condition per table (immediately)

Cond D: One channel inoperable - place channel in trip (72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months )

Cond I: One channel inoperable - place channel in trip (72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months )

NOTE: Event 4 TR 3.1.9, Boron Injection System - Operating -- may be looked at

ES-301 Transient and Event Checklist Form ES-301-5 Facility:

Wolf Creek Date of Exam:

Aug 31 - Sept 4 Operating Test No.:

Scenarios 1

2 3

4 CREW POSITION CREW POSITION CREW POSITION CREW POSITION M

I N

I M

U M(*)

A P

P L

I C

A N

T E

V E

N T

T Y

P E

S R

O A

T C

B O

P S

R O

A T

C B

O P

S R

O A

T C

B O

P S

R O

A T

C B

O P

T O

T A

L R

I U

RX 0

2 1

1 1 0 NOR 3

1 2

1 1 1 I/C 45 45 4

4 4 2 MAJ 67 6

3 2

2 1 RO 1 SRO-I SRO-U TS 0

0 0

0 2 2 RX 0

2 1

1 1 0 NOR 3

1 2

1 1 1 I/C 45 45 4

4 4 2 MAJ 67 6

3 2

2 1 RO 2 SRO-I SRO-U TS 0

0 0

0 2 2 RX 0

0 2

1 1

1 0 NOR 3

1 0

2 1

1 1 I/C 45 345 78 14 9

4 4 2 MAJ 67 6

35 5

2 2 1 RO 3 SRO-I SRO-U TS 0

0 0

2 2 RX 3

0 1

1 1 0 NOR 0

1 1

1 1 1 I/C 124 2345 78 9

4 4 2 MAJ 67 6

3 2

2 1 RO SRO-I1 SRO-U TS 0

23 2

0 2 2

Instructions:

1.

Check the applicant level and enter the operating test number and Form ES-D-1 event numbers for each event type; TS are not applicable for RO applicants. ROs must serve in both the at-the-controls (ATC) and balance-of-plant (BOP) positions; Instant SROs must serve in both the SRO and the ATC positions, including at least two instrument or component (I/C) malfunctions and one major transient, in the ATC position.

If an Instant SRO additionally serves in the BOP position, one I/C malfunction can be credited toward the two I/C malfunctions required for the ATC position.

2.

Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.5.d) but must be significant per Section C.2.a of Appendix D. (*) Reactivity and normal evolutions may be replaced with additional instrument or component malfunctions on a 1-for-1 basis.

3.

Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicants competence count toward the minimum requirements specified for the applicants license level in the right-hand columns.

ES-301 Transient and Event Checklist Form ES-301-5 Facility:

Wolf Creek Date of Exam:

Aug 31 - Sept 4 Operating Test No.:

Scenarios 1

2 3

4 CREW POSITION CREW POSITION CREW POSITION CREW POSITION M

I N

I M

U M(*)

A P

P L

I C

A N

T E

V E

N T

T Y

P E

S R

O A

T C

B O

P S

R O

A T

C B

O P

S R

O A

T C

B O

P S

R O

A T

C B

O P

T O

T A

L R

I U

RX 3

0 1

1 1 0 NOR 0

1 1

1 1 1 I/C 124 2345 78 9

4 4 2 MAJ 67 6

3 2

2 1 RO SRO-I2 SRO-U TS 0

23 2

0 2 2 RX 0

2 0

1 1

1 0 NOR 3 1

0 2

1 1 1 I/C 1245 45 124 9

4 4 2 MAJ 67 6

35 5

2 2 1 RO SRO-I3 SRO-U TS 12 0

1 3

0 2 2 RX 3

0 0

1 1

1 0 NOR 0

1 0

1 1

1 1 I/C 124 2345 78 2

10 4

4 2 MAJ 67 6

35 5

2 2 1 RO SRO-I4 SRO-U TS 0

23 0

2 0

2 2 RX 0

0 0

1 1 0 NOR 3 1

2 1

1 1 I/C 1245 345 78 9

4 4 2 MAJ 67 6

3 2

2 1 RO SRO-I SRO-U1 TS 12 0

2 0

2 2

ES-301 Transient and Event Checklist Form ES-301-5 Facility:

Wolf Creek Date of Exam:

Aug 31 - Sept. 4 Operating Test No.:

Scenarios 1

2 3

4 Back up CREW POSITION CREW POSITION CREW POSITION CREW POSITION M

I N

I M

U M(*)

A P

P L

I C

A N

T E

V E

N T

T Y

P E

S R

O A

T C

B O

P S

R O

A T

C B

O P

S R

O A

T C

B O

P S

R O

A T

C B

O P

T O

T A

L R

I U

RX 0

0 0

1 1 0 NOR 3 1

2 1

1 1 I/C 1245 345 78 9

4 4 2 MAJ 67 6

3 2

2 1 RO SRO-I SRO-U2 TS 12 0

2 0

2 2 RX 1

1 0 NOR 1

1 1 I/C 4

4 2 MAJ 2

2 1 RO SRO-I SRO-U TS 0

2 2 RX 1

1 0 NOR 1

1 1 I/C 4

4 2 MAJ 2

2 1 RO SRO-I SRO-U TS 0

2 2 RX 1

1 1

1 0 NOR 1

1 1

1 1 1 I/C 2346 246 3

4 4

4 2 MAJ 5

5 5

1 2

2 1 RO SRO-I SRO-U TS 3

1 1

2 2

ES-301 Competencies Checklist Form ES-301-6 Facility:

Wolf Creek Date of Examination:

8-31to9/4 Operating Test No.

APPLICANTS RO1 SRO-I SRO-U RO2 SRO-I SRO-U RO3 SRO-I SRO-U RO SRO-I1 SRO-U SCENARIO SCENARIO SCENARIO SCENARIO Competencies 1

2 3

4 1

2 3

4 1

2 3

4 1

2 3

4 Interpret/Diagnose Events and Conditions 35 256 35 256 35 347 8

14 12 3

23 56 Comply With and Use Procedures (1) 356 7

26 356 7 26 35 67 134 6

14 5

12 36 7

12 34 56 Operate Control Boards (2) 356 126 356 126 35 6

134 567 8

14 5

12 36 Communicate and Interact 356 256 356 256 35 6 134 14 5 36 all Demonstrate Supervisory Ability (3) all Comply With and Use Tech. Specs. (3) 23 Notes:

(1)

Includes Technical Specification compliance for an RO.

(2)

Optional for an SRO-U.

(3)

Only applicable to SROs.

Instructions:

Check the applicants license type and enter one or more event numbers that will allow the examiners to evaluate every applicable competency for every applicant.

ES-301 Competencies Checklist Form ES-301-6 Facility:

Wolf Creek Date of Examination:

8-31to9/4 Operating Test No.

APPLICANTS RO SRO-I2 SRO-U RO SRO-I3 SRO-U RO SRO-I4 SRO-U RO SRO-I SRO-U1 SCENARIO SCENARIO SCENARIO SCENARIO Competencies 1

2 3

4 1

2 3

4 1

2 3

4 1

2 3

4 Interpret/Diagnose Events and Conditions 123 235 6

6 256 35 12 3

235 6

25 6 34 78 Comply With and Use Procedures (1) 123 67 123 456 all 26 all 12 36 7

123 456 23 5 all 13 46 Operate Control Boards (2) 123 6 126 12 36 25 13 45 67 8

Communicate and Interact 36 all all 256 all 36 all 23 5 all 13 4

Demonstrate Supervisory Ability (3) all all all all all Comply With and Use Tech. Specs. (3) 23 12 1

23 12 Notes:

(1)

Includes Technical Specification compliance for an RO.

(2)

Optional for an SRO-U.

(3)

Only applicable to SROs.

Instructions:

Check the applicants license type and enter one or more event numbers that will allow the examiners to evaluate every applicable competency for every applicant.

ES-301 Competencies Checklist Form ES-301-6 Facility:

Wolf Creek Date of Examination:

8-31to9/4 Operating Test No.

APPLICANTS RO SRO-I SRO-U2 RO SRO-I SRO-U RO SRO-I SRO-U RO SRO-I SRO-U SCENARIO SCENARIO SCENARIO SCENARIO Competencies 1

2 3

4 1

2 3

4 1

2 3

4 1

2 3

4 Interpret/Diagnose Events and Conditions 6

347 8

Comply With and Use Procedures (1) all 134 6

Operate Control Boards (2) 134 567 8

Communicate and Interact all 134 Demonstrate Supervisory Ability (3) all Comply With and Use Tech. Specs. (3) 12 Notes:

(1)

Includes Technical Specification compliance for an RO.

(2)

Optional for an SRO-U.

(3)

Only applicable to SROs.

Instructions:

Check the applicants license type and enter one or more event numbers that will allow the examiners to evaluate every applicable competency for every applicant.

ML093200394

Text

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