ML13248A594

WC-2013-07 Final Outlines

ML13248A594
Person / Time
Site:	Wolf Creek
Issue date:	07/26/2013
From:	David Strickland Operations Branch IV
To:	Wolf Creek
laura hurley
References
50-482/13-007
Download: ML13248A594 (44)

Text

ES-401 PWR Examination Outline Form ES-401-2 Facility: Wolf Creek Date of Exam: July 2013 RO K/A Category Points SRO-Only Points Tier Group K K K K K K A A A A G A2 G* Total 1 2 3 4 5 6 1 2 3 4

• Total

1. 1 3 3 3 3 3 3 18 3 3 6 Emergency &

Abnormal 2 2 2 2 N/A 1 1 N/A 1 9 2 2 4 Plant Evolutions Tier Totals 5 5 5 4 4 4 27 5 5 10 1 2 2 2 3 3 3 3 3 3 2 2 28 2 3 5 2.

Plant 2 1 1 1 1 1 1 1 1 1 1 10 2 1 3 Systems Tier Totals 3 3 3 4 4 4 4 4 3 3 3 38 4 4 8

3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 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/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 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 (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams.

9. For Tier 3, select topics from Section 2 of the 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 10 CFR 55.43.

ES-401 2 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (RO / SRO)

E/APE # / Name / Safety Function K K K A A G K/A Topic(s) IR #

1 2 3 1 2 EA1 Ability to operate and monitor the 000007 (BW/E02&E10; CE/E02) Reactor X following as they apply to a reactor trip: CVCS 3.2 1 Trip - Stabilization - Recovery / 1 (CFR 41.7 / 45.5 / 45.6)

AA2. Ability to determine and interpret the 000008 Pressurizer Vapor Space X following as they apply to the Pressurizer 3.9 2 Accident / 3 Vapor Space Accident: PORV isolation (block) valve switches and indicators (CFR: 43.5 / 45.13) 2.1.20 Ability to interpret and execute 000009 Small Break LOCA / 3 X procedure steps. 4.6 3 (CFR: 41.10 / 43.5 / 45.12)

EK1 Knowledge of the operational implications 000011 Large Break LOCA / 3 x of the following concepts as 4.1 4 they apply to the Large Break LOCA :

Natural circulation and cooling, including reflux boiling (CFR 41.8 / 41.10 / 45.3)

AK2. Knowledge of the interrelations between 000015/17 RCP Malfunctions / 4 x the Reactor Coolant Pump 2.9 5 Malfunctions (Loss of RC Flow) and the following: RCP seals (CFR 41.7 / 45.7)

AK3. Knowledge of the reasons for the 000022 Loss of Rx Coolant Makeup / 2 X following responses as they apply to 3.5 6 the Loss of Reactor Coolant Makeup: Actions contained in SOPs and EOPs for RCPs, loss of makeup, loss of charging, and abnormal charging (CFR 41.5, 41.10 / 45.6 / 45.13)

AA1. Ability to operate and / or monitor the 000025 Loss of RHR System / 4 X following as they apply to the Loss of Residual 2.8 7 Heat Removal System: RHR heat exchangers (CFR 41.7 / 45.5 / 45.6)

AA2. Ability to determine and interpret the 000026 Loss of Component Cooling X following as they apply to the Loss of 2.5 8 Water / 8 Component Cooling Water: The normal values and upper limits for the temperatures of the components cooled by CCW (CFR: 43.5 / 45.13) 2.4.4 Ability to recognize abnormal indications 000027 Pressurizer Pressure Control X for system operating parameters that are entry- 4.5 9 System Malfunction / 3 level conditions for emergency and abnormal operating procedures.

(CFR: 41.10 / 43.2 / 45.6) l EK1 Knowledge of the operational implications 000029 ATWS / 1 X of the following concepts as 2.6 10 they apply to the ATWS: Definition of reactivity (CFR 41.8 / 41.10 / 45.3)

EK3 Knowledge of the reasons for the 000038 Steam Gen. Tube Rupture / 3 X following responses as the apply to the 4.4 11 SGTR: Prevention of secondary PORV cycling (CFR 41.5 / 41.10 / 45.6 / 45.13)

AK2. Knowledge of the interrelations between 000040 (BW/E05; CE/E05; W/E12) X the Steam Line Rupture and the 2.6* 12 Steam Line Rupture - Excessive Heat following: (CFR 41.7 / 45.7)

Transfer / 4 AA1. Ability to operate and / or monitor the 000054 (CE/E06) Loss of Main X following as they apply to the Loss of Main 4.4 13 Feedwater / 4 Feedwater (MFW): Manual startup of electric and steam-driven AFW (CFR 41.7 / 45.5 / 45.6)

AA2. Ability to determine and interpret the 000055 Station Blackout / 6 X following as they apply to the Pressurizer 3.5 14 Pressure Control Malfunctions: Conditions requiring plant shutdown (CFR: 43.5 / 45.13) 2.4.3 Ability to identify post-accident 000056 Loss of Off-site Power / 6 X instrumentation. 3.7 15 (CFR: 41.6 / 45.4)

AK3. Knowledge of the reasons for the 000057 Loss of Vital AC Inst. Bus / 6 X following responses as they apply to 4.1 16 the Loss of Vital AC Instrument Bus: Actions contained in EOP for loss of vital ac electrical instrument bus (CFR 41.5,41.10 / 45.6 / 45.13)

AK1. Knowledge of the operational implications 000058 Loss of DC Power / 6 X of the following concepts as 2.8 17 they apply to Loss of DC Power: Battery charger equipment and instrumentation (CFR 41.8 / 41.10 / 45.3) 000062 Loss of Nuclear Svc Water / 4 000065 Loss of Instrument Air / 8 EK2. Knowledge of the interrelations between W/E04 LOCA Outside Containment / 3 X the (LOCA Outside Containment) and the 3.8 18 following: Facility*s 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. (CFR:

41.7 / 45.7)

W/E11 Loss of Emergency Coolant Recirc. / 4 BW/E04; W/E05 Inadequate Heat Transfer - Loss of Secondary Heat Sink / 4 000077 Generator Voltage and Electric Grid Disturbances / 6 K/A Category Totals: 3 3 3 3 3 3 Group Point Total: 18/6

ES-401 3 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (RO / SRO)

E/APE # / Name / Safety Function K K K A A G K/A Topic(s) IR #

1 2 3 1 2 AK1. Knowledge of the operational 000001 Continuous Rod Withdrawal / 1 X 2.8 19 implications of the following concepts as they apply to Continuous Rod Withdrawal: Definitions of core quadrant power tilt (CFR 41.8 / 41.10 / 45.3) 000003 Dropped Control Rod / 1 000005 Inoperable/Stuck Control Rod / 1 AK2. Knowledge of the 000024 Emergency Boration / 1 X 2.7 20 interrelations between Emergency Boration and the following: Valves (CFR 41.7 / 45.7) 000028 Pressurizer Level Malfunction / 2 000032 Loss of Source Range NI / 7 AK3. Knowledge of the reasons for the 000033 Loss of Intermediate Range NI / 7 X 3.6 21 following responses as they apply to the Loss of Intermediate Range Nuclear Instrumentation: Guidance contained in EOP for loss of intermediate range instrumentation (CFR 41.5,41.10 / 45.6 /

45.13) 000036 (BW/A08) Fuel Handling Accident / 8 000037 Steam Generator Tube Leak / 3 AA1. Ability to operate and / or monitor 000051 Loss of Condenser Vacuum / 4 X 2.5* 22 the following as they apply to the Loss of Condenser Vacuum: Rod position (CFR 41.7 / 45.5 / 45.6) 000059 Accidental Liquid RadWaste Rel. / 9 000060 Accidental Gaseous Radwaste Rel. / 9 000061 ARM System Alarms / 7 000067 Plant Fire On-site / 8 000068 (BW/A06) Control Room Evac. / 8 000069 (W/E14) Loss of CTMT Integrity / 5 000074 (W/E06&E07) Inad. Core Cooling / 4 000076 High Reactor Coolant Activity / 9 W/EO1 & E02 Rediagnosis & SI Termination / 3 EA2. Ability to determine and interpret W/E13 Steam Generator Over-pressure / 4 X 3.0 23 the following as they apply to the (Steam Generator Overpressure)

Adherence to appropriate procedures and operation within the limitations in the facility*s license and amendments.

(CFR: 43.5 / 45.13) 2.4.31 Knowledge of annunciator W/E15 Containment Flooding / 5 X 4.2 24 alarms, indications, or response procedures.

(CFR: 41.10 / 45.3)

W/E16 High Containment Radiation / 9 BW/A01 Plant Runback / 1

BW/A02&A03 Loss of NNI-X/Y / 7 BW/A04 Turbine Trip / 4 BW/A05 Emergency Diesel Actuation / 6 BW/A07 Flooding / 8 BW/E03 Inadequate Subcooling Margin / 4 EK1. Knowledge of the operational BW/E08; W/E03 LOCA Cooldown - Depress. / 4 X 3.5 25 implications of the following concepts as they apply to the (LOCA Cooldown and Depressurization) Annunciators and conditions indicating signals, and remedial actions associated with the (LOCA Cooldown and Depressurization).

(CFR: 41.8 / 41.10 / 45.3)

EK2. Knowledge of the interrelations BW/E09; CE/A13; W/E09&E10 Natural Circ. / 4 X 3.6 26 between the (Natural Circulation Operations) and the following: Facility*s 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.

(CFR: 41.7 / 45.7)

BW/E13&E14 EOP Rules and Enclosures EK3. Knowledge of the reasons for the CE/A11; W/E08 RCS Overcooling - PTS / 4 X 3.6 27 following responses as they apply to the (Pressurized Thermal Shock) Normal, abnormal and emergency operating procedures associated with (Pressurized Thermal Shock).

(CFR: 41.5 / 41.10, 45.6, 45.13)

CE/A16 Excess RCS Leakage / 2 CE/E09 Functional Recovery K/A Category Point Totals: 2 2 2 1 1 1 Group Point Total: 9/4

ES-401 4 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 1 (RO / SRO)

System # / Name K K K K K K A A A A G K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 K4 Knowledge of RCPS design 003 Reactor Coolant Pump X 3.2 28 feature(s) and/or interlock(s) which provide for the following:

Minimizing RCS leakage (mechanical seals) (CFR: 41.7)

K5 Knowledge of the 004 Chemical and Volume X 3.8 29 Control operational implications of the following concepts as they apply to the CVCS: Relationship between temperature and pressure in CVCS components during solid plant operation (CFR: 41.5/45.7)

K6 Knowledge of the effect of a 005 Residual Heat Removal X 2.5 30 loss or malfunction on the following will have on the RHRS: RHR heat (CFR: 41.7 / 45.7)

A1 Ability to predict and/or 006 Emergency Core Cooling X 4.2 31 monitor changes in parameters (to prevent exceeding design limits) associated with operating the ECCS controls including:

ECCS flow rate (CFR: 41.5 / 45.5)

A2 Ability to (a) predict the 007 Pressurizer Relief/Quench X 3.6 32 Tank 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:

Overpressurization of the PZR(CFR: 41.5 / 43.5 / 45.3 /

45.13)

A3 Ability to monitor automatic 008 Component Cooling Water X 2.9 33 operation of the CCWS, including: A3.04 Requirements on and for the CCWS for different conditions of the power plant (CFR: 41.7 / 45.5)

A4 Ability to manually operate 010 Pressurizer Pressure Control X 3.6 34 and/or monitor in the control room: PZR heaters (CFR: 41.7 /

45.5 to 45.8)

2.4.45 Ability to prioritize and 012 Reactor Protection X interpret the significance of each 4.1 35 annunciator or alarm.

(CFR: 41.10 / 43.5 / 45.3 / 45.12)

K1 Knowledge of the physical 013 Engineered Safety Features X connections and/or cause effect 3.4 36 Actuation relationships between the ESFAS and the following systems: MFW System (CFR: 41.2 to 41.9 / 45.7 to 45.8)

K2 Knowledge of power supplies 022 Containment Cooling X 3.0* 37 to the following: Containment cooling fans (CFR: 41.7)

K3 Knowledge of the effect that 028 Hydrogen Recombiner and X 3.3 38 Purge System a loss or malfunction of the HRPS will have on the following: Hydrogen concentration in containment (CFR: 41.7 / 45.6)

K4 Knowledge of CSS design 026 Containment Spray X 2.8 39 feature(s) and/or interlock(s) which provide for the following:

Prevention of material from clogging nozzles during recirculation (CFR: 41.7)

K5 Knowledge of the operational 039 Main and Reheat Steam X implications of the following 3.6 40 concepts as they apply to the MRSS:

Effect of steam removal on reactivity (CFR: 441.5 / 45.7)

A1 Ability to predict and/or monitor 059 Main Feedwater X changes in parameters (to prevent 2.7* 41 exceeding design limits) associated with operating the MFW controls including: Power level restrictions for operation of MFW pumps and valves (CFR: 41.5 / 45.5)

K6 Knowledge of the effect of a loss 061 Auxiliary/Emergency X or malfunction of the following will 2.6 42 Feedwater have on the AFW components:

Pumps (CFR: 41.7 / 45.7)

A2 Ability to (a) predict the impacts 062 AC Electrical Distribution X of the following malfunctions or 2.5 43 operations on the ac distribution system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Degraded system voltages (CFR: 41.5 / 43.5 / 45.3 / 45.13)

A3 Ability to monitor automatic 063 DC Electrical Distribution X operation of the DC electrical 2.7 44 system, including: Meters, annunciators, dials, recorders, and indicating lights (CFR: 41.7 / 45.5)

A4 Ability to manually operate 073 Process Radiation X and/or monitor in the control room: 3.9 45 Monitoring Effluent release (CFR: 41.7 / 45.5 to 45.8) 2.2.22 Knowledge of limiting 064 Emergency Diesel Generator X conditions for operations and safety 4.0 46 limits. (CFR: 41.5 / 43.2 / 45.2)

K1 Knowledge of the physical 076 Service Water X 3.5* 47 connections and/or cause- effect relationships between the SWS and the following systems: RHR system (CFR: 41.2 to 41.9 / 45.7 to 45.8)

K2 Knowledge of bus power 078 Instrument Air X 2.7 48 supplies to the following:

Instrument air compressor (CFR: 41.7)

K3 Knowledge of the effect that 103 Containment X 3.8 49 a loss or malfunction of the containment system will have on the following: Loss of containment integrity under normal operations (CFR: 41.7 / 45.6)

K4 Knowledge of RHRS design 005 Residual Heat Removal X 3.2 50 feature(s) and/or interlock(s) which provide or the following:

System protection logics, including high-pressure interlock, reset controls, and valve interlocks (CFR: 41.7)

K5 Knowledge of the 039 Main and Reheat Steam X 2.9 51 operational implications of the following concepts as they apply to the MRSS: Definition and causes of steam/water hammer (CFR: 441.5 / 45.7)

A2 Ability to (a) predict the 059 Main Feedwater X 3.0* 52 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: Failure of feedwater control system (CFR: 41.5 / 43.5 / 45.3 / 45.13)

K6 Knowledge of the effect of a 061 Auxiliary/Emergency X 2.5 53 Feedwater loss or malfunction of the following will have on the AFW components: Controllers and positioners (CFR: 41.7 / 45.7)

A1 Ability to predict and/or 064 Emergency Diesel Generator X 2.8 54 monitor changes in parameters (to prevent exceeding design limits) associated with operating the ED/G system controls including: Crankcase temperature and pressure (CFR: 41.5 / 45.5)

A3 Ability to monitor automatic 103 Containment X 3.9 55 operation of the containment system, including: Containment isolation (CFR: 41.7 / 45.5)

K/A Category Point Totals: 2 2 2 3 3 3 3 3 3 2 2 Group Point Total: 28/5

ES-401 5 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 2 (RO / SRO)

System # / Name K K K K K K A A A A G K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 A4 Ability to manually operate 001 Control Rod Drive X 3.5 56 and/or monitor in the control room: Determination of SDM (CFR: 41.7/45.5 to 45.8) 002 Reactor Coolant X 2.4.18 Knowledge of the 3.3 57 specific bases for EOPs.

(CFR: 41.10 / 43.1 / 45.13)

K1 Knowledge of the physical 011 Pressurizer Level Control X 3.8 58 connections and/or cause-effect relationships between the PZR LCS and the following systems: RPS (CFR: 41.2 to 41.9 / 45.7 to 45.8)

K3 Knowledge of the effect that a 014 Rod Position Indication X 2.5 59 loss or malfunction of the RPIS will have on the following: Plant computer (CFR: 41.7 / 45.6)

K2 Knowledge of bus power 015 Nuclear Instrumentation X 3.3 60 supplies to the following: NIS channels, components, and interconnections (CFR: 41.7) 016 Non-nuclear Instrumentation K4 Knowledge of ITM system 017 In-core Temperature Monitor X 3.4 61 design feature(s) and/or interlock(s) which provide for the following: Input to subcooling monitors CFR: 41.7)

K5 Knowledge of the operational 027 Containment Iodine Removal X 3.1* 62 implications of the following concepts as they apply to the CIRS: Purpose of charcoal filters (CFR: 41.7 / 45.7) 028 Hydrogen Recombiner and Purge Control 029 Containment Purge 033 Spent Fuel Pool Cooling 034 Fuel Handling Equipment 035 Steam Generator 041 Steam Dump/Turbine Bypass Control

045 Main Turbine Generator 055 Condenser Air Removal 056 Condensate K6 Knowledge of the effect of a 068 Liquid Radwaste X 2.5 63 loss or malfunction on the following will have on the Liquid Radwaste System : Radiation monitors (CFR: 41.7 / 45.7)

A1 Ability to predict and/or 071 Waste Gas Disposal X 2.5 64 monitor changes in parameters(to prevent exceeding design limits) associated with Waste Gas Disposal System operating the controls including: Ventilation system (CFR: 41.5 / 45.5)

A2 Ability to (a) predict the 072 Area Radiation Monitoring X 2.8 65 impacts of the following malfunctions or operations on the ARM system- and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Detector failure (CFR: 41.5 / 43.5 / 43.3 / 45.13) 075 Circulating Water 079 Station Air 086 Fire Protection K/A Category Point Totals: 1 1 1 1 1 1 1 1 1 1 Group Point Total: 10/3

ES-401 2 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (RO / SRO)

E/APE # / Name / Safety Function K K K A A G K/A Topic(s) IR #

1 2 3 1 2 000007 (BW/E02&E10; CE/E02) Reactor Trip - Stabilization - Recovery / 1 000008 Pressurizer Vapor Space X 2.4.11 Knowledge of abnormal 4.2 76 Accident / 3 condition procedures.

(CFR: 41.10 / 43.5 / 45.13) 000009 Small Break LOCA / 3 X EA2 Ability to determine or interpret the 3.6 77 following as they apply to a small break LOCA: Charging pump flow indication (CFR 43.5 / 45.13) 000011 Large Break LOCA / 3 000015/17 RCP Malfunctions / 4 000022 Loss of Rx Coolant Makeup / 2 000025 Loss of RHR System / 4 000026 Loss of Component Cooling Water / 8 000027 Pressurizer Pressure Control X AA2. Ability to determine and interpret the 4.0 78 System Malfunction / 3 following as they apply to the Pressurizer Pressure Control Malfunctions: Actions to be taken if PZR pressure instrument fails high (CFR: 43.5 / 45.13) 000029 ATWS / 1 X 2.4.8 Knowledge of how abnormal 4.5 79 operating procedures are used in conjunction with EOPs.

(CFR: 41.10 / 43.5 / 45.13) 000038 Steam Gen. Tube Rupture / 3 000040 (BW/E05; CE/E05; W/E12)

Steam Line Rupture - Excessive Heat Transfer / 4 000054 (CE/E06) Loss of Main Feedwater / 4 000055 Station Blackout / 6 000056 Loss of Off-site Power / 6 X AA2. Ability to determine and interpret the 3.6 80 following as they apply to the Loss of Offsite Power: Operational status of CCW pump (CFR: 43.5 / 45.13) 000057 Loss of Vital AC Inst. Bus / 6 X 2.2.37 Ability to determine 4.6 81 operability and/or availability of safety related equipment.

(CFR: 41.7 / 43.5 / 45.12) 000058 Loss of DC Power / 6

000062 Loss of Nuclear Svc Water / 4 000065 Loss of Instrument Air / 8 W/E04 LOCA Outside Containment / 3 W/E11 Loss of Emergency Coolant Recirc. / 4 BW/E04; W/E05 Inadequate Heat Transfer - Loss of Secondary Heat Sink / 4 000077 Generator Voltage and Electric Grid Disturbances / 6 K/A Category Totals: 3 3 Group Point Total: 18/

6

ES-401 3 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (RO / SRO)

E/APE # / Name / Safety Function K K K A A G K/A Topic(s) IR #

1 2 3 1 2 000001 Continuous Rod Withdrawal / 1 000003 Dropped Control Rod / 1 000005 Inoperable/Stuck Control Rod / 1 X AA2. Ability to determine and 4.4 82 interpret the following as they apply to the Inoperable / Stuck Control Rod: Required actions if more than one rod is stuck or inoperable (CFR: 43.5 / 45.13) 000024 Emergency Boration / 1 000028 Pressurizer Level Malfunction / 2 000032 Loss of Source Range NI / 7 000033 Loss of Intermediate Range NI / 7 000036 (BW/A08) Fuel Handling Accident / 8 X 2.4.47 Ability to diagnose and 4.2 83 recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.

(CFR: 41.10 / 43.5 / 45.12) 000037 Steam Generator Tube Leak / 3 000051 Loss of Condenser Vacuum / 4 000059 Accidental Liquid RadWaste Rel. / 9 000060 Accidental Gaseous Radwaste Rel. / 9 000061 ARM System Alarms / 7 000067 Plant Fire On-site / 8 000068 (BW/A06) Control Room Evac. / 8 000069 (W/E14) Loss of CTMT Integrity / 5 X AA2. Ability to determine and 4.4 84 interpret the following as they apply to the Loss of Containment Integrity: Verification of automatic and manual means of restoring integrity (CFR: 43.5 / 45.13) 000074 (W/E06&E07) Inad. Core Cooling / 4 000076 High Reactor Coolant Activity / 9 W/EO1 & E02 Rediagnosis & SI Termination / 3 W/E13 Steam Generator Over-pressure / 4 W/E15 Containment Flooding / 5 W/E16 High Containment Radiation / 9 BW/A01 Plant Runback / 1

BW/A02&A03 Loss of NNI-X/Y / 7 BW/A04 Turbine Trip / 4 BW/A05 Emergency Diesel Actuation / 6 BW/A07 Flooding / 8 BW/E03 Inadequate Subcooling Margin / 4 X 2.4.30 Knowledge of events 4.1 85 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.

(CFR: 41.10 / 43.5 / 45.11) l BW/E08; W/E03 LOCA Cooldown - Depress. / 4 BW/E09; CE/A13; W/E09&E10 Natural Circ. / 4 BW/E13&E14 EOP Rules and Enclosures CE/A11; W/E08 RCS Overcooling - PTS / 4 CE/A16 Excess RCS Leakage / 2 CE/E09 Functional Recovery K/A Category Point Totals: 2 2 Group Point Total: 9/4

ES-401 4 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 1 (RO / SRO)

System # / Name K K K K K K A A A A G K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 003 Reactor Coolant Pump X 2.1.31 Ability to locate 4.3 86 control room switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup.

(CFR: 41.10 / 45.12) 004 Chemical and Volume Control 005 Residual Heat Removal A2 Ability to (a) predict the 006 Emergency Core Cooling X 4.2 87 impacts of the following malfunctions or operations on the ECCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Inadvertent SIS actuation (CFR: 41.5 / 45.5) 007 Pressurizer Relief/Quench Tank 008 Component Cooling Water 010 Pressurizer Pressure Control 012 Reactor Protection 013 Engineered Safety Features X 2.4.49 Ability to perform 4.4 88 Actuation without reference to procedures those actions that require immediate operation of system components and controls.

(CFR: 41.10 / 43.2 / 45.6) 022 Containment Cooling 025 Ice Condenser

A2 Ability to (a) predict the 026 Containment Spray X 3.7 89 impacts of the following malfunctions or operations on the CSS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Safe securing of containment spray when it can be done)

(CFR: 41.5 / 43.5 / 45.3 / 45.13) 039 Main and Reheat Steam 059 Main Feedwater 061 Auxiliary/Emergency Feedwater 062 AC Electrical Distribution 063 DC Electrical Distribution X 2.1.27 Knowledge of system 4.0 90 purpose and/or function.

(CFR: 41.7) l 064 Emergency Diesel Generator 073 Process Radiation Monitoring 076 Service Water 078 Instrument Air 103 Containment K/A Category Point Totals: 2 3 Group Point Total: 28/5

ES-401 5 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 2 (RO / SRO)

System # / Name K K K K K K A A A A G K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 001 Control Rod Drive 002 Reactor Coolant 011 Pressurizer Level Control 014 Rod Position Indication A2 Ability to (a) predict the 015 Nuclear Instrumentation X 3.8 91 impacts of the following malfunctions or operations on the NIS; and (b based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Core void formation (CFR: 41.5 / 43.5 / 45.3 / 45.5) 016 Non-nuclear Instrumentation 017 In-core Temperature Monitor 027 Containment Iodine Removal 028 Hydrogen Recombiner and Purge Control 029 Containment Purge 033 Spent Fuel Pool Cooling 034 Fuel Handling Equipment 035 Steam Generator X 2.4.9 Knowledge of low 4.2 92 power/shutdown implications in accident (e.g.,

loss of coolant accident or loss of residual heat removal) mitigation strategies.

(CFR: 41.10 / 43.5 / 45.13) l 041 Steam Dump/Turbine Bypass Control

A2 Ability to (a) predict the 045 Main Turbine Generator x 2.9 93 impacts of the following malfunctions or operation on the MT/G system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Malfunction of electrohydraulic control (CFR: 41.5 / 43.5 / 45.3 / 45.5) 055 Condenser Air Removal 056 Condensate 068 Liquid Radwaste 071 Waste Gas Disposal 072 Area Radiation Monitoring 075 Circulating Water 079 Station Air 086 Fire Protection K/A Category Point Totals: 2 1 Group Point Total: 10/3

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Wolf Creek Date of Exam: July 2013 Category K/A # Topic RO SRO-Only IR # IR #

2.1. 2.1.23 Ability to perform specific system and 4.3 66 integrated plant procedures during all

1. modes of plant operation.

Conduct of Operations (CFR: 41.10 / 43.5 / 45.2 / 45.6) 2.1. 2.1.25 Ability to interpret reference 3.9 67 materials, such as graphs, curves, tables, etc.

(CFR: 41.10 / 43.5 / 45.12) 2.1. 2.1.32 Ability to explain and apply system 3.8 68 limits and precautions.

(CFR: 41.10 / 43.2 / 45.12) 2.1. 2.1.15 Knowledge of administrative 3.4 94 requirements for temporary management directives, such as standing orders, night orders, Operations memos, etc.

(CFR: 41.10 / 45.12) 2.1. 2.1.35 Knowledge of the fuel-handling 3.9 95 responsibilities of SROs.

l(CFR: 41.10 / 43.7) 2.1.

Subtotal 3 2 2.2. 2.2.12 Knowledge of surveillance procedures. 3.7 69 (CFR: 41.10 / 45.13) l

2. 2.2. 2.2.35 Ability to determine Technical 3.6 70 Equipment Specification Mode of Operation.

Control (CFR: 41.7 / 41.10 / 43.2 / 45.13) 2.2. 2.2.5 Knowledge of the process for making 3.2 96 design or operating changes to the facility.

(CFR: 41.10 / 43.3 / 45.13) 2.2. 2.2.21 Knowledge of pre- and post- 4.1 97 maintenance operability requirements.

l(CFR: 41.10 / 43.2) 2.2.

2.2.

Subtotal 2 2

2.3. 2.3.12 Knowledge of radiological safety 3.2 71 principles pertaining to licensed operator l 3.

duties, such as containment entry Radiation requirements, fuel handling responsibilities, l Control access to locked high-radiation areas, aligning filters, etc.

(CFR: 41.12 / 45.9 / 45.10) 2.3. 2.3.14 Knowledge of radiation or 3.4 72 contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

(CFR: 41.12 / 43.4 / 45.10) 2.3. 2.3.6 Ability to approve release permits. 3.8 98 (CFR: 41.13 / 43.4 / 45.10) 2.3.

2.3.

2.3.

2 1 2.4. 2.4.2 Knowledge of system set points, 4.5 73 interlocks and automatic actions associated

4. with EOP entry conditions.

Emergency Procedures / (CFR: 41.7 / 45.7 / 45.8)

Plan 2.4. 2.4.20 Knowledge of the operational 3.8 74 implications of EOP warnings, cautions, and notes. (CFR: 41.10 / 43.5 / 45.13) 2.4. 2.4.34 Knowledge of RO tasks performed 4.2 75 outside the main control room during an l emergency and the resultant operational effects.

(CFR: 41.10 / 43.5 / 45.13) 2.4. 2.4.4 Ability to recognize abnormal 4.7 99 indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.

(CFR: 41.10 / 43.2 / 45.6) 2.4. 2.4.23 Knowledge of the bases for 4.4 100 prioritizing emergency procedure implementation during emergency operations. (CFR: 41.10 / 43.5 / 45.13) 2.4.

Subtotal 3 2 Tier 3 Point Total 10 7

ES-401 Record of Rejected K/As Form ES-401-4 Tier / Randomly Reason for Rejection Group Selected K/A T1 G1 025 AA1.23 Overlap/Conflict with #30 KA 005 K6.03. Unable to select another >2.5 KA in 005 K6 area. Selected KA in 025:

025AA1.01 Bottom line: 025 AA1.23 replaced with 025 AA1.01 RCS/RHR cooldown rate (3.6/3.7)

T1 G1 055 AA2.06 AA2.06 not applicable. It is for 027 not 055. Recommend replace with 055 EA2.02.

Bottom Line: 055 AA2.06 is replaced with 055 EA2.02 (Ability to determine or interpret the following as they apply to a station blackout: RCS core cooling through natural circulation cooling to SG cooling.) (CFR 43.5/45.13) (4.4/4.6)

T2 G1 028 K3.01 028 K3.01 is not a T2 G1 - it is a T2 G2. Recommend replace with 003 RCPs K3.02 (CFR 41.7/45.6) (3.5/3.8)

Bottom line: Replaced 028 K3.01 with 003 K3.02 (3.7/4.0)

T2 G2 072 A2.02 Radiation monitors are oversampled. Recommend replacement with 028 A2.02.

Bottom line: Replaced 072 A2.02 with 028 A2.02 (3.5/3.9).

See next page - had to replace 028 A2.02, too.

FINAL

T2 G2 028 A2.02 TS Amendment 157: The revised 10 CFR 50.44 no longer defines a design-basis LOCA hydrogen release, and eliminates the requirements for hydrogen control systems to mitigate such a release. The installation of hydrogen recombiners and or vent and purge systems required by 10 CFR 50.44(b)(3) was intended to address the limited quantity and rate of hydrogen generation that was postulated from a design-basis LOCA. The Commission has found that this hydrogen release is not risk-significant because the design-basis LOCA hydrogen release does not contribute to the conditional probability of a large release up to approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the onset of core damage. In addition, these systems were ineffective at mitigating hydrogen releases from risk-significant beyond design-basis accidents. Therefore, the Commission eliminated the hydrogen release associated with a design-basis LOCA from 10 CFR 50.44 and the associated requirements that necessitated the need for the hydrogen recombiners and the backup hydrogen vent and purge systems. As a result, the NRC staff finds that the requirements related to hydrogen recombiners no longer meet any of the four criteria in 10 CFR 50.36(c)(2)(ii) for retention in the TSs and the existing TS requirements may, therefore be eliminated for all plants.

Hence at Wolf Creek: all references to the hydrogen recombiners have been eliminated from the EMGs. There is no need to test them and the SYS was voided, too.

Bottom line: Replaced 028 A2.02 with 056 A2.04 T2 G1 003 2.1.31 Unable to write an SRO-level question. Selected another (SRO) system - kept the K/A Bottom line: Replaced 003 2.1.31 with 004 2.1.31 T3 (SRO) C of O; 2.1.15 2.1.15 has no 55.43 connection - See NOTE 9 of ES-401-1.

Bottom line: Replaced 2.1.15 with 2.1.34.

FINAL

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Wolf Creek Date of Examination: July 2013 Examination Level: RO SRO Operating Test Number:

Administrative Topic Type Describe activity to be performed (see Note) Code*

Using EMG ES-04, Natural Circulation Cooldown, step R.A.1 N, R 10b: Verify Cold Shutdown Boron Concentration by Sampling: Determine RCS boron concentration on a Conduct of Operations total mass basis, using Attachment A, DETERMINATION OF RCS BORON CONCENTRATION BASED ON TOTAL MASS 2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation.

(CFR 41.10/43.5/45.2/45.6) (4.3/4.4)

Manually determine Quadrant Power Tilt Ratio (QPTR)

R.A.2 D, R using STS RE-012, QPTR Determination.

Conduct of Operations 2.1.20 Ability to interpret and execute procedure steps.

(CFR 41.10/43.5/45.12) (4.6/4.6)

Complete STS BG-005A, BORIC ACID TRANSFER R.A.3 N, R SYSTEM INSERVICE PUMP A TEST, Attachment A, Data Sheet.

Equipment Control 2.2.12 Knowledge of surveillance procedures. (CFR 41.10/45.13) (3.7/4.1)

Using a Radiation Work Permit (RWP) and previously R.A.4 N, R received dose, calculate the amount of time an Operator has to complete hanging tags on a tagout.

Radiation Control 2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions. (CFR 41.12/43.4/45.10) (3.2/3.7)

Emergency Procedures/Plan Not used in 2013 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)

FINAL 1

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Wolf Creek Date of Examination: July 2013 Examination Level: RO SRO Operating Test Number:

Administrative Topic Type Describe activity to be performed (see Note) Code*

Review/Approve reactivity calculation for an up power of S.A.1 N, R 10%.

Conduct of Operations 2.1.37 Knowledge of procedures, guidelines, or limitations associated with reactivity management. (CFR 41.1/43.6/45.6) (4.3/4.6)

Review/Approve manual calculation of RTP (STS SE-S.A.2 D, R 002, Manual Calculation of Reactor Thermal Power)

Conduct of Operations 2.1.20 Ability to interpret and execute procedure steps.

(CFR 41.10/43.5/45.12) (4.6/4.6)

Review Quadrant Power Tilt Ratio and applicable S.A.3 D, R Technical Specifications.

Equipment Control 2.2.12 Knowledge of surveillance procedures (CFR 41.10/45.13) (3.7/4.1) 2.2.42 Ability to recognize system parameters that are entry level conditions for Technical Specifications.

(CFR 41.7/41.10/43.2/43.3/45.3) (3.9/4.6)

Review/Approve/Evaluate a Containment Purge Permit S.A.4 N, R (CPP) for correctness prior to restart.

Radiation Control 2.3.6 Ability to approve release permits. (CFR 41.13/43.4/45.10) (2.0/3.8) 2.3.11 Ability to control radiation releases. (CFR 41.11/43.4/45.10) (3.8/4.3)

In the simulator setting, perform Emergency Plan S.A.5 D, S classification within fifteen minutes, and accurately and correctly complete an Emergency Notification form (EPF Emergency Procedures/Plan 06-007-01).

Time Critical JPM (only the classify).

2.4.41 Knowledge of the emergency action level thresholds and classifications. (CFR 41.10/43.5/45.11)

(2.9/4.6) 2.4.44 Knowledge of emergency plan protective action recommendations. (CFR 41.10/41.12/43.5/45.11)

(2.4/4.4)

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

FINAL 1

• 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)

FINAL 2

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Wolf Creek Date of Examination: July 2013 Exam Level: RO SRO-I SRO-U Operating Test Number: 1 RO-only JPM in bold Control Room Systems (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

Safety System / JPM Title Type Code*

Function

a. (S1) (003.A2.11) Dropped Control Rod During Rod Parking N,S 1

b. (S2) (004.A3.03) Letdown HX High Temperature Divert N,A,S 2

c. (S3) (006.A3.01) Isolate Accumulators following a LOCA D,A,S 3

d. (S4) (003.A4.06) Start a Reactor Coolant Pump D,S,L 4P (Note: RO only)

e. (S5) (045.A4.02) Synchronize Main Generator to the Grid M,S,L 4S

f. (S6) (027.A4.03) Start Containment Atmosphere Control Fan N,A,S 5

g. (S7) (073.A4.02) Place Unit Vent Monitor in Accident Mode of Operation N,S 7

h. (S8) (008.A2.01) Transfer CCW System Service Loop D,A,S 8 In-Plant Systems (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

i. (P1) (076.AK3.06) Place Cation Bed Demin in Service for High RCS Activity D,R 1

j. (P2) (E09.EA1.3) Natural Circulation -Depressurize Inactive SG N,R,A,E 4S

k. (P3) (057.AK3.01) Align 120VAC Vital Bus to SOLA Transformer D,A,E 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 ES-301, Page 23 of 27 Rev 2 Rev 2: rev follows initial submittal

Appendix D Scenario Outline Form ES-D-1 Facility: ____Wolf Creek_________ Scenario No.: ___1___ Op-Test No.: _______

Examiners: ____________________________ Operators: _____________________________

Initial Conditions: 100%, Middle of Life Turnover: Motor Driven AFW pump A is tagged out for maintenance activities. Technical Specification (TS) 3.7.5 Condition B.1 (restore AFW train to OPERABLE in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) was entered.

Expected return is 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

Event Malf. Event Event No. No. Type* Description 1 mBB01F I RCS temperature, BB TI-421 (T-cold), fails high.

SRO ATC TS 3.3.1, REACTOR TRIP SYSTEM INSTRUMENTATION, Condition A (Immediately entered due to failure) and from Table 3.3.1-1, Fu 6 and 7, Condition E (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to trip bistables).

OFN SB-008, INSTRUMENT MALFUNCTIONS, Attachment L.

2 mAB01A I Steam Generator A steam pressure, AB PI-514A, fails low.

1 SRO BOP TS 3.3.2, ENGINEERED SAFETY FEATURES ACTUATION INSTRUMENTATION, Condition A (Immediately due to failure) and from Table 3.3.2-1, Fu 1.e and 4.e, Condition D (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to trip bistables).

ALR 00-108B, SG A LEV DEV or ALR 00-108C, SG A FLOW MISMATCH and/or OFN SB-008, INSTRUMENT MALFUNCTIONS, Attachment C.

1 FINAL NRC 1 9

3 mBB21B I Pressurizer pressure instrument, BB PI-456, fails low.

SRO ATC TS 3.3.1, REACTOR TRIP SYSTEM INSTRUMENTATION, Condition A (Immediately entered due to failure) and from Table 3.3.1-1, Fu 6 and 8, Conditions E and M are entered (both are 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to trip bistables).

TS 3.3.2, ENGINEERED SAFETY FEATURES ACTUATION INSTRUMENTATION, Condition A (Immediately due to failure) and from Table 3.3.2-1, 1.d, 3.a.3, 5.d, 6.e and 8.b, Conditions D (1.d, 3.a.3, 5.d, 6.e: 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to place channel in bypass) and L (one hour to verify P-11 interlock in correct state) are entered.

OFN SB-008, INSTRUMENT MALFUNCTIONS, Attachment K.

4 mAE08D C Main Feed Regulating Valve D fails closed; manual control available using controller AE FK-540.

SRO BOP ALR 00-111C, SG D FLOW MISMATCH or ALR 00-111B, SG D LEV DEV.

5 mSG01 M Seismic event with an inadvertent Reactor trip and Safety Injection (SI) signal and a Loss of all Auxiliary Feedwater. (Critical Task mSF15A SRO (CT) - Establish feedwater flow into at least one SG before mSA01B ATC RCS bleed and feed is initiated and before SGs dry out.)

mAL02 BOP bkrDPAL EMG E-0, REACTOR TRIP OR SAFETY INJECTION, EMG FR-01B H1, RESPONSE TO LOSS OF SECONDARY HEAT SINK and SYS AP-122, NON-SAFETY AUX FEED PUMP OPERATION.

6 mAC02 C Preloaded and post trip: Main turbine fails to trip (auto), manual trip C available. BOP depressed both MAIN TURBINE MASTER TRIP SRO A and B pushbuttons: AC HS-002A and AC HS-002B. (CT -

mAC02B Manual Main Turbine trip)

BOP Immediate Action step 2RNO EMG E-0, REACTOR TRIP OR SAFETY INJECTION.

7 mSA27 C Preloaded and post trip: Containment Fan Coolers A and C are GN03B not running in SLOW speed.

SRO mSA27 GN05B ATC EMG E-0, REACTOR TRIP OR SAFETY INJECTION, Attachment F.

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor 2

FINAL NRC 1 9

SCENARIO

SUMMARY

Turnover and Initial Conditions: Unit is at 100% power, Middle of Life. Motor Driven AFW pump A is tagged out for maintenance activities. Technical Specification (TS) 3.7.5 Condition B.1 (restore AFW train to OPERABLE in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) was entered. Expected return is 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

Event 1: Reactor Coolant System (RCS) temperature T-cold instrument BB TI-421 fails high. Control rods step inward. The crew identifies and diagnoses the temperature instrument failure and enters OFN SB-008, INSTRUMENT MALFUNCTIONS. Attachment L, Narrow Range RTD Malfunction, is used to identify and mitigate the instrument failure. Memory Action steps are performed by the BOP (verify no load rejection in progress) and ATC (take rods to manual using SE HS-9). Technical Specifications are identified by the SRO. TS 3.3.1, REACTOR TRIP SYSTEM INSTRUMENTATION, Table 3.3.1-1, Fu 6 and 7 are identified and Conditions A and E are entered.

Event 2: Steam pressure channel for Steam Generator A, AB PI-514A, fails low. The crew identifies and diagnoses the steam pressure channel failure and enters Alarm Response procedure ALR 00-108B, SG A LEV DEV or ALR 00-108C, SG A FLOW MISMATCH and/or OFN SB-008, INSTRUMENT MALFUNCTIONS. OFN SB-008, INSTRUMENT MALFUNCTIONS, Attachment C, Steam Pressure Channel Malfunction, is used to identify and mitigate the instrument failure. Memory action steps are performed by the BOP (A Main Feed Regulating Valve placed in manual and Steam Generator level controlled manually). Technical Specifications are identified by the SRO. TS 3.3.2, ENGINEERED SAFETY FEATURES ACTUATION INSTRUMENTATION, Table 3.3.2-1, Fu 1.e and 4.e are identified and Conditions A and D are entered.

Event 3: Pressurizer (PZR) pressure instrument, BB PI-456, fails low. The crew identifies and diagnoses PZR pressure instrument failure and enters OFN SB-008, INSTRUMENT MALFUNCTIONS. Attachment K, PZR Pressure Malfunction, is used to identify and mitigate the instrument failure. Technical Specifications are identified by the SRO. TS 3.3.1, REACTOR TRIP SYSTEM INSTRUMENTATION, Table 3.3.1-1, Fu 6 and 8 are identified and Conditions A, E and M are entered. TS 3.3.2, ENGINEERED SAFETY FEATURES ACTUATION INSTRUMENTATION, Table 3.3.2-1, Fu 1.d, 3.a.3, 5.d, 6.e and 8.b are identified and Conditions A, D and L are entered.

Event 4: Steam Generator D Main Feed Regulating Valve (MFRV) closes in automatic. The crew identifies and diagnoses the MFRV failure and enters Alarm Response procedure ALR 00-111C, SG D FLOW MISMATCH or ALR 00-111B, SG D LEV DEV, to mitigate the MFRV failure.

Event 5: The Major event is accompanied by a seismic alarm. An Inadvertent Reactor trip and Safety Injection Signal occurs followed by a Loss of all Auxiliary Feedwater. The crew diagnoses the seismic event and Reactor Trip and Safety Injection actuation. The crew enters EMG E-0, REACTOR TRIP OR SAFETY INJECTION.

During the performance of EMG E-0, REACTOR TRIP OR SAFETY INJECTION, the crew diagnoses the Loss of all Auxiliary Feedwater (AFW). At step 8 RNO, the crew ensures the BIT valves are open and transitions to Functional Recovery procedure EMG FR-H1, RESPONSE TO LOSS OF SECONDARY HEAT SINK.

Success path for the scenario is accomplished at step 8 of EMG FR-H1, RESPONSE TO LOSS OF SECONDARY HEAT SINK, when AFW flow from the Non-Safety Related Auxiliary Feedwater Pump is established to the Steam Generators. SYS AP-122, NON-SAFETY AUX FEED PUMP OPERATION, is performed.

Critical Task (CT) Establish feedwater flow into at least one SG before RCS bleed and feed is initiated and before SGs dry out.

3 FINAL NRC 1 9

Event 6: Post trip, the BOP determines the Main Turbine failed to trip. The BOP depresses both MAIN TURBINE MASTER TRIP A and B pushbuttons (AC HS-002A and AC HS-002B) during the performance of Immediate Actions step 2 RNO of EMG E-0, REACTOR TRIP OR SAFETY INJECTION.

Critical Task - Manual Main Turbine trip is performed.

Event 7: Post trip, the ATC/BOP determines that Containment Fan Coolers A and C are not running in SLOW speed. EMG E-0, REACTOR TRIP OR SAFETY INJECTION, Attachment F, step F8 RNO directs starting the fans in SLOW speed.

SCENARIO TERMINATION:

Successful mitigation of the scenario requires the crew restore secondary heat sink by performance of EMG FR-H1, RESPONSE TO LOSS OF SECONDARY HEAT SINK, using the Non-Safety Related Auxiliary Feedwater Pump per procedure SYS AP-122, NON-SAFETY AUX FEED PUMP OPERATION.

CRITICAL TASKS (CT)

Event 5: Establish feedwater flow into at least one SG before RCS bleed and feed is initiated and before SGs dry out (RCS bleed and feed is initiated when 3 of 4 SGs indicate 12% wide range level. SG dryout is indicated by at least 3 SGs with wide range level less than 9%). Restore AFW to the Steam Generators using Non-Safety Related Aux Feed (NSAFW) Pump per procedure SYS AP-122, NON-SAFETY AUX FEED PUMP OPERATION, entered from EMG FR-H1, RESPONSE TO LOSS OF SECONDARY HEAT SINK.

Event 6: Manual Main Turbine trip is performed. Manual Main Turbine trip before IR SUR becomes positive and before any RCS cold leg temperature decreases by more than 100°F in a 1-hour period and reaches the T1 limit (240°F) and prior to transition out of EMG E-0. Due to the new design/controls, both MAIN TURBINE MASTER TRIP A and B pushbuttons (AC HS-002A and AC HS-002B) are manipulated.

TECHNICAL SPECIFICATIONS:

Event 1: Reactor Coolant System (RCS) temperature T-cold instrument BB TI-421 fails high.

TS 3.3.1, REACTOR TRIP SYSTEM INSTRUMENTATION, Condition A (Immediately entered due to failure) and from Table 3.3.1-1, Fu 6 and 7, Condition E (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to trip bistables).

Event 2: Steam pressure channel for Steam Generator A, AB PI-514A, fails low.

TS 3.3.2, ENGINEERED SAFETY FEATURES ACTUATION INSTRUMENTATION, Condition A (Immediately due to failure) and from Table 3.3.2-1, Fu 1.e and 4.e, Condition D (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to trip bistables).

Event 3: Pressurizer (PZR) pressure instrument, BB PI-456, fails low.

TS 3.3.1, REACTOR TRIP SYSTEM INSTRUMENTATION, Condition A (Immediately entered due to failure) and from Table 3.3.1-1, Fu 6 and 8, Conditions E and M are entered (both are 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to trip bistables).

TS 3.3.2, ENGINEERED SAFETY FEATURES ACTUATION INSTRUMENTATION, Condition A (Immediately due to failure) and from Table 3.3.2-1, 1.d, 3.a.3, 5.d, 6.e and 8.b, Conditions D (1.d, 3.a.3, 5.d, 6.e: 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to place channel in bypass) and L (one hour to verify P-11 interlock in correct state) are entered.

4 FINAL NRC 1 9

PRA/PSA: On March 31, 2013, NE 13-0022 provided the Notice of Probabilistic Risk Assessment (PRA)

Model Revision 6.

Scenario PRA application Description Scenario 1 Top Operator Action Failure to Enter EMG FR-H1 Note: Crew does enter EMG FR-H1 and the success path is to feed the S/Gs using the NSAFW pump.

Scenario 2 Core Damage Frequency Switchyard centered LOOP (CDF) by Initiating Event Note: This event is complicated when the only Large Early Release available EDG experiences a fuel failure and the Frequency (LERF) by crew enters EMG C-0.

Initiating Event Scenario 3 Core Damage Frequency Large steamline break outside CTMT (CDF) by Initiating Event 5

FINAL NRC 1 9

Appendix D Scenario Outline Form ES-D-1 Facility: ____Wolf Creek_________ Scenario No.: ___3___ Op-Test No.: _______

Examiners: ____________________________ Operators: _____________________________

Initial Conditions: ~2% power - startup in progress. Beginning of Life.

Turnover: Crew across the hall is being briefed to continue power escalation. Your crew tasked to maintain current plant conditions stable steady state. GEN 00-003, HOT STANDBY TO MINIMUM LOAD, in progress at step 6.39. Main Turbine is not synced to the grid. Pre-heating in service.

Event Malf. Event Event No. No. Type* Description 1 mBB21C I Pressurizer (PZR) pressure channel, BB PI-457, fails high.

SRO Technical Specification (TS) 3.3.1, REACTOR TRIP SYSTEM ATC INSTRUMENTATION, Condition A (Immediately entered due to failure) and from Table 3.3.1-1, Fu 6 and 8, Condition E (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to trip bistables) are identified.

TS 3.3.2, ENGINEERED SAFETY FEATURES ACTUATION INSTRUMENTATION, Condition A (Immediately entered due to failure) and from Table 3.3.2-1, Fu 1.d, 3.a.3, 5.d, 6.e, and 8.b, Condition D (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to trip bistables) and Condition L (one hour to verify interlock P-11 in correct state) are identified.

OFN SB-008, INSTRUMENT MALFUNCTIONS, Attachment K.

2 mAE15D I Steam Generator D level channel, AE LI-549 (controlling 3 channel), fails low.

SRO BOP TS 3.3.1, REACTOR TRIP SYSTEM INSTRUMENTATION, Condition A (Immediately entered due to failure) and from Table 3.3.1-1, Fu 14 Condition E (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to trip bistables) is identified.

TS 3.3.2, ENGINEERED SAFETY FEATURES ACTUATION INSTRUMENTATION, Condition A (Immediately entered due to failure) and from Table 3.3.2-1, Fu 5.c and 6.d are identified.

Conditions I and D (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to trip bistables) respectively.

ALR 00-111B, SG D LEV DEV or ALR 00-111A, SG D LEV HILO and/or OFN SB-008, INSTRUMENT MALFUNCTIONS, Attachment F.

1 FINAL NRC 3 12

3 bkrPB00 C Normal Charging Pump (NCP) trip.

301 SRO ATC ALR 00-042E, CHARGING PMP TROUBLE 4 mAB07B C Steam Generator B Atmospheric Relief Valve (ARV) fails open, manual closure available.

SRO BOP TS 3.7.4, ATMOSPHERIC RELIEF VALVES (ARVs), Condition A (7 days to restore to OPERABLE status).

AP 15C-003, PROCEDURE USERS GUIDE ABNORMAL OPERATIONS, step 6.1.7, or OFN AB-041, STEAM LINE OR FEEDLINE LEAK.

Per AP 15C-003 step 6.1.7, the Operator should take manual control when components are not performing correctly.

5 mAB04B M B Steam Line break outside Containment. (Critical Task (CT) -

Isolate Auxiliary Feedwater (AFW) to the Faulted Steam SRO Generator)

ATC BOP OFN AB-041, STEAM LINE OR FEEDLINE LEAK, EMG E-0, REACTOR TRIP OR SAFETY INJECTION, EMG E-2, FAULTED STEAM GENERATOR ISOLATION.

Time Critical Action (TCA): Isolate Auxiliary Feedwater to a faulted Steam Generator following a Steam Line Break event within twenty minutes (AI 21-016, OPERATOR TIMED CRITICAL ACTION VALIDATION, Attachment A, Time Critical Action List.)

6 mNB01 C Preloaded and post trip: Emergency Bus NB01 trips, Emergency Diesel Generator (EDG) A starts and loads. (CT - Manually mEF05A SRO start Essential Service Water pump A)

ATC Essential Service Water (ESW) A autostart failure, manual start available.

AP 15C-003, PROCEDURE USERS GUIDE ABNORMAL OPERATIONS, step 6.1.7 or EMG E-0, REACTOR TRIP OR SAFETY INJECTION, Attachment F.

Per AP 15C-003 step 6.1.7, the Operator should take manual control when components are not performing correctly.

2 FINAL NRC 3 12

7 bkrDPE C Preloaded and post trip: Component Cooling Water (CCW) trip of G01B B pump. CCW D autostart defeated, manual start available.

SRO (CT - Manually start at least one CCW pump in the train with mEG14 required ECCS equipment operating)

D ATC AP 15C-003, PROCEDURE USERS GUIDE ABNORMAL OPERATIONS, step 6.1.7 or EMG E-0, REACTOR TRIP OR SAFETY INJECTION, Attachment F.

Per AP 15C-003 step 6.1.7, the Operator should take manual control when components are not performing correctly.

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor 3

FINAL NRC 3 12

SCENARIO

SUMMARY

Turnover and Initial Conditions: ~2% power - startup in progress. Beginning of Life. Crew across the hall is being briefed to continue power escalation. Your crew tasked to maintain current plant conditions stable steady state. GEN 00-003, HOT STANDBY TO MINIMUM LOAD, in progress at step 6.39. Main Turbine is not synced to the grid. Pre-heating in service.

Event 1: Pressurizer (PZR) pressure channel, BB PI-457, fails high. The crew identifies and diagnoses the failure and enters OFN SB-008, INSTRUMENT MALFUNCTIONS. OFN SB-008, INSTRUMENT MALFUNCTIONS, Attachment K, PZR Pressure Malfunction, is used to identify and mitigate the instrument failure. Memory Action steps are performed by the ATC (identify failed channel, select manual on PZR Pressure Master Controller, control pressure and select out the failed channel). Technical Specifications are identified by the SRO. TS 3.3.1, REACTOR TRIP SYSTEM INSTRUMENTATION, Condition A (Immediately entered due to failure) and from Table 3.3.1-1, Fu 6 and 8, Condition E (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to trip bistables) are identified. TS 3.3.2, ENGINEERED SAFETY FEATURES ACTUATION INSTRUMENTATION, Condition A (Immediately entered due to failure) and from Table 3.3.2-1, Fu 1.d, 3.a.3, 5.d, 6.e, and 8.b, Condition D (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to trip bistables) and Condition L (one hour to verify interlock P-11 in correct state) are identified.

Event 2: Steam Generator D controlling level channel, AE LI-549, fails low. The crew identifies and diagnoses the level channel failure and enters Alarm Response procedure ALR 00-111B, SG D LEV DEV or ALR 00-111A, SG D LEV HILO and/or OFN SB-008, INSTRUMENT MALFUNCTIONS. OFN SB-008, INSTRUMENT MALFUNCTIONS, Attachment F, S/G Level Channel Malfunction, is used to identify and mitigate the instrument failure. Memory Action steps are performed by the BOP (identify the failed instrument, place D Feed Regulating Bypass Valve in manual and control Steam Generator level manually). Technical Specifications are identified by the SRO. TS 3.3.1, REACTOR TRIP SYSTEM INSTRUMENTATION, Condition A (Immediately entered due to failure) and from Table 3.3.1-1, Fu 14 Condition E (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to trip bistables) is identified. TS 3.3.2, ENGINEERED SAFETY FEATURES ACTUATION INSTRUMENTATION, Condition A (Immediately entered due to failure) and from Table 3.3.2-1, Fu 5.c and 6.d are identified. Conditions I and D (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to trip bistables) are entered respectively.

Event 3: Normal Charging Pump (NCP) trip. The crew identifies and diagnoses the NCP trip and enters ALR 00-042E, CHARGING PMP TROUBLE, to mitigate the component failure. A Memory Action is performed by the ATC (isolate letdown by closing any open Letdown Orifice Isolation valves). A Centrifugal Charging pump is started and letdown re-established per actions of ALR 00-042E.

Event 4: Steam Generator B Atmospheric Relief Valve (ARV) fails open, manual closure available. The crew identifies and diagnoses the failure. The BOP closes the open ARV using AB-PIC-2A, SG B STEAM DUMP TO ATMS CTRL, per procedure AP 15C-003, step 6.1.7 or OFN AB-041, STEAM LINE OR FEEDLINE LEAK, step 5. Technical Specifications are identified by the SRO. TS 3.7.4, ATMOSPHERIC RELIEF VALVES (ARVs), Condition A (7 days to restore to OPERABLE status).

AP 15C-003, PROCEDURE USERS GUIDE ABNORMAL OPERATIONS, step 6.1.7, allows the Operator to take manual control of components not performing their intended function.

Event 5: Major event: A 1.2 E+6 lb/hr B Steam Line break outside Containment occurs. The crew identifies the Steam Line break outside Containment and may enter OFN AB-041, STEAM LINE OR FEEDLINE BREAK, to mitigate the consequences; however, a Reactor trip and Safety Injection are required and performed. The crew enters EMG E-0, REACTOR TRIP OR SAFETY INJECTION. The Main Steam Isolation Valves are closed, and Steam Generator B is identified as the faulted Steam Generator. Auxiliary Feedwater is isolated to the faulted Steam Generator per EMG E-0 REACTOR TRIP 4

FINAL NRC 3 12

OR SAFETY INJECTIONs Foldout page criteria #3, Faulted S/G Isolation Criteria. The crew transitions to EMG E-2, FAULTED STEAM GENERATOR ISOLATION and based on plant conditions transitions to EMG ES-03, SI TERMINATION or EMG E-1, LOSS OF REACTOR OR SECONDARY COOLANT.

Critical Task (CT) Isolate Auxiliary Feedwater (AFW) to the Faulted Steam Generator before completion of EMG E-2.

Event 6: Post trip, Emergency Bus NB01 trips, Emergency Diesel Generator (EDG) A starts and loads.

Essential Service Water (ESW) A autostart failure, manual start available. The ATC diagnoses ESW A must be started in order to supply cooling water to EDG A and the NB01 loads. ESW A is started using handswitch EF HIS-55A per AP 15C-003, PROCEDURE USERS GUIDE ABNORMAL OPERATIONS, step 6.1.7 or EMG E-0, REACTOR TRIP OR SAFETY INJECTION, Attachment F, Automatic Signal Verification, step F7 RNO.

Critical Task to manually start Essential Service Water pump A is performed. Manually start at least the minimum required number of ESW pumps in an operating safeguards train before required Diesel Generator(s) trip or before the completion of Attachment F of EMG E-0).

Event 7: Post trip: Component Cooling Water (CCW) B pump trips. CCW D autostart is defeated, however manual start available using handswitch EG HIS-24. The ATC diagnoses the lack of running Component Cooling Water pumps. CCW D pump must be started in order to supply cooling water to safeguard components e.g. Centrifugal Charging Pump oil coolers, Safety Injection pump oil coolers etc.

AP 15C-003, PROCEDURE USERS GUIDE ABNORMAL OPERATIONS, step 6.1.7 or EMG E-0, REACTOR TRIP OR SAFETY INJECTION, Attachment F, Automatic Signal Verification, step F6 RNO.

Critical Task to manually start at least one CCW pump in the train with required ECCS equipment operating before completion of Attachment F of EMG E-0.

SCENARIO TERMINATION Successful mitigation of the scenario requires the faulted Steam Generator is isolated and based on plant conditions, transition to EMG ES-03, SI TERMINATION or EMG E-1, LOSS OF REACTOR OR SECONDARY COOLANT.

CRITICAL TASKS (CT):

Event 5: Isolate Auxiliary Feedwater (AFW) to the Faulted Steam Generator before completion of EMG E-2 is performed. Auxiliary Feedwater is isolated to the Faulted Steam Generator per EMG E-0 REACTOR TRIP OR SAFETY INJECTIONs Foldout page criteria #3, Faulted S/G Isolation Criteria.

When the crew transitions to EMG E-2, FAULTED STEAM GENERATOR ISOLATION, actions will be performed to ensure the faulted Steam Generator is isolated. Foldout #3 from EMG E-0: Main Steam Isolation valves are closed (not critical). Critical: To isolate AFW: the BOP closes AL HK-9A, SG B MD AFP AFW REG VLV CTRL and AL HK-10A, SG B TD AFP AFW REG VLV CTRL.

Event 6: Manually start at least the minimum required number of ESW pumps in an operating safeguards train before required Diesel Generator(s) trip, e.g. EDG A or before the completion of Attachment F of EMG E-0. ESW PUMP A handswitch EF HIS-55A is manipulated to RUN position, starting ESW A pump before the EDG A trips.

Event 7: Manually start at least one CCW pump in the train with required ECCS equipment operating before completion of Attachment F of EMG E-0. Bravo train CCW pump D is started. Manipulate CCW PUMP D handswitch EG HIS-24 to RUN position, starting CCW D pump, providing cooling water to ECCS loads.

5 FINAL NRC 3 12

TECHNICAL SPECIFICATIONS:

Event 1: Pressurizer (PZR) pressure channel, BB PI-457, fails high.

TS 3.3.1, REACTOR TRIP SYSTEM INSTRUMENTATION, Condition A (Immediately entered due to failure) and from Table 3.3.1-1, Fu 6 and 8, Condition E (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to trip bistables) are identified.

TS 3.3.2, ENGINEERED SAFETY FEATURES ACTUATION INSTRUMENTATION, Condition A (Immediately entered due to failure) and from Table 3.3.2-1, Fu 1.d, 3.a.3, 5.d, 6.e, and 8.b, Condition D (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to trip bistables) and Condition L (one hour to verify interlock P-11 in correct state) are identified.

Event 2: Steam Generator D controlling level channel, AE LI-549, fails low. TS 3.3.1, REACTOR TRIP SYSTEM INSTRUMENTATION, Condition A (Immediately entered due to failure) and from Table 3.3.1-1, Fu 14 Condition E (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to trip bistables) is identified. TS 3.3.2, ENGINEERED SAFETY FEATURES ACTUATION INSTRUMENTATION, Condition A (Immediately entered due to failure) and from Table 3.3.2-1, Fu 5.c and 6.d are identified. Conditions I and D (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to trip bistables) are entered respectively.

Event 4: Steam Generator B Atmospheric Relief Valve (ARV) fails open, manual closure available. TS 3.7.4, ATMOSPHERIC RELIEF VALVES (ARVs), Condition A (7 days to restore to OPERABLE status).

PRA/PSA: On March 31, 2013, NE 13-0022 provided the Notice of Probabilistic Risk Assessment (PRA)

Model Revision 6.

Scenario PRA application Description Scenario 1 Top Operator Action Failure to Enter EMG FR-H1 Note: Crew does enter EMG FR-H1 and the success path is to feed the S/Gs using the NSAFW pump.

Scenario 2 Core Damage Frequency Switchyard centered LOOP (CDF) by Initiating Event Note: This event is complicated when the only Large Early Release available EDG experiences a fuel failure and the Frequency (LERF) by crew enters EMG C-0.

Initiating Event Scenario 3 Core Damage Frequency Large steamline break outside CTMT (CDF) by Initiating Event TIME CRITICAL/TIME SENSITIVE ACTIONS:

Per AI 21-016, OPERATOR TIME CRITICAL ACTIONS VALIDATION, form AIF 21-016-02, Time Verification Form, will be used to capture the completion time and routed to Operations Support and Safety Analysis for review.

Time Critical Action (TCA): Isolate Auxiliary Feedwater to a faulted Steam Generator following a Steam Line Break event within twenty minutes (AI 21-016, OPERATOR TIMED CRITICAL ACTION VALIDATION, Attachment A, Time Critical Action List.)

6 FINAL NRC 3 12

Appendix D Scenario Outline Form ES-D-1 Facility: ____Wolf Creek_________ Scenario No.: ___4___ Op-Test No.: _______

Examiners: ____________________________ Operators: _____________________________

Initial Conditions: 100%, Beginning of Life.

Turnover: Motor Driven Auxiliary Feedwater Pump (MDAFW) A tagged out for preventative maintenance activities. Technical Specification (TS) 3.7.5 Condition B.1 (restore AFW train to OPERABLE in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) was entered. Expected return is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Event Malf. Event Event No. No. Type* Description 1 mBB22A I Pressurizer (PZR) level channel, BB PI-459, fails low.

SRO ATC Technical Specification (TS) 3.3.1, REACTOR TRIP SYSTEM INSTRUMENTATION, Condition A (Immediately entered due to failure) and from Table 3.3.1-1, Fu 9, Condition M (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to trip bistables) is identified.

OFN SB-008, INSTRUMENT MALFUNCTIONS, Attachment J.

2 mAE12C I Steam Generator B feed flow controlling channel, AE FT-520, fails high.

SRO BOP ALR 00-109C, SG B FLOW MISMATCH, ALR 00-109B, SG B LEV DEV and/or OFN SB-008, INSTRUMENT MALFUNCTIONS, Attachment E.

3 bkrWS0 C Service Water Pump A trip.

1PA SRO ATC Technical Requirement Manual (TRM) 3.7.8, SERVICE WATER SYSTEM, Condition A (60 days to restore to FUNCTIONAL status)

ALR 00-009B, SERV WTR PMP TRIP or ALR 00-008B, SERV WTR PRESS HI LO.

1 FINAL NRC 4 9

4 mAB01C I Steam Generator C controlling pressure channel, AB PI-535A, 2 fails high.

SRO BOP TS 3.3.2, ENGINEERED SAFETY FEATURES INSTRUMENTATION, Condition A (Immediately entered due to failure) and from Table 3.3.2-1, Fu 1.e and 4.e, Condition D (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to trip bistables respectively) are identified.

ALR 110C, SG C FLOW MISMATCH, ALR 00-110B, SG C LEV DEV and/or OFN SB-008, INSTRUMENT MALFUNCTIONS, Attachment C.

5 R Reactivity event: Shift Manager declares Motor Driven Auxiliary Feedwater Pump B INOPERABLE but AVAILABLE.

SRO ATC TS 3.7.5, Condition C, (Two AFW trains inoperable), Required BOP Action C.1 (Be in MODE 3 within six hours).

Crew utilizes pre-shift 10% downpower brief or OFN MA-038, RAPID PLANT SHUTDOWN.

6 mBB06C M 600 gpm Cold Leg break, Loop C - Loss Of Coolant Accident (LOCA).

SRO ATC OFN BB-007, RCS LEAKAGE HIGH; EMG E-0, REACTOR TRIP BOP OR SAFETY INJECTION; EMG E-1, LOSS OF REACTOR OR SECONDARY COOLANT; then based on plant conditions transitions to EMG ES-11, POST LOCA COOLDOWN AND DEPRESSURIZATION.

7 mAL01 C Preloaded and post trip: Turbine Driven Auxiliary Feedwater Pump (TDAFP) autostart failure, manual start available. MDAFW B rAL11 SRO AFW discharge to Steam Generators A and D throttled.

rAL09 BOP (Critical Task (CT) - Establish 270, 000 lbm/hr Auxiliary Feedwater flow before completion of Attachment F of EMG E-0.)

AP 15C-003, PROCEDURE USERS GUIDE ABNORMAL OPERATIONS, step 6.1.7 or EMG E-0, REACTOR TRIP OR SAFETY INJECTION, step 8, RNO b. or Attachment F.

Per AP 15C-003 step 6.1.7, the Operator should take manual control when components are not performing correctly.

2 FINAL NRC 4 9

8 mSA18B C Preloaded and post trip: Train Bravo CPIS and CISA autostart failure, manual actuation available; however, CTMT ATMS mSA23B SRO MONITOR SPLY CTMT ISO VLV, GS HIS-36 and CTMT ATMS mSA27 ATC MONITOR RETURN CTMT ISO VLV, GS HIS-34, remain open, GS16 manual closure available. (CT - Close containment isolation valves such that at least one valve is closed on each critical mSA27 phase-A penetration before completion of Attachment F of GS17 EMG E-0.)

EMG E-0, REACTOR TRIP OR SAFETY INJECTION, Attachment F.

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor 3

FINAL NRC 4 9

SCENARIO

SUMMARY

Turnover and Initial Conditions: Unit is at 100%. Beginning of Life. Motor Driven Auxiliary Feedwater Pump (MDAFW) A tagged out for preventative maintenance activities. Technical Specification (TS) 3.7.5 Condition B.1 (restore AFW train to OPERABLE in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) was entered. Expected return is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Event 1: Pressurizer (PZR) level channel, BB PI-459, fails low. The crew identifies and diagnoses the failure and enters OFN SB-008, INSTRUMENT MALFUNCTIONS. OFN SB-008, INSTRUMENT MALFUNCTIONS, Attachment J, PZR Level Channel Malfunction, is used to identify and mitigate the instrument failure. Technical Specifications are identified by the SRO. TS 3.3.1, REACTOR TRIP SYSTEM INSTRUMENTATION, Condition A (Immediately entered due to failure) and from Table 3.3.1-1, Fu 9, Condition M (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to trip bistables) is identified.

Event 2: Steam Generator B feed flow controlling channel, AE FT-520, fails high. The crew identifies and diagnoses the failure and enters either ALR 00-109C, SG B FLOW MISMATCH, ALR 00-109B, SG B LEV DEV, and/or OFN SB-008, INSTRUMENT MALFUNCTIONS. OFN SB-008, INSTRUMENT MALFUNCTIONS, Attachment E, Feedwater Flow Channel Malfunction is used to identify and mitigate the instrument failure. Memory Action steps are performed by the BOP (identify the failed instrument, place Main Feed Regulating Valve, AE FK-520, in manual and control Steam Generator level).

Event 3: Service Water Pump A trip. The crew identifies and diagnoses Service Water Pump A trip and enters either ALR 00-009B, SERV WTR PMP TRIP, or ALR 00-008B, SERV WTR PRESS HI LO, to mitigate the component failure. A standby Service Water Pump is started to establish discharge pressure greater than 85 psig. The SRO identifies Technical Requirement (TR) 3.7.8, SERVICE WATER SYSTEM, Condition A (60 days to restore to FUNCTIONAL status).

Event 4: Steam Generator C controlling pressure channel, AB PI-535A, fails high. The crew identifies and diagnoses the failure and enters ALR 110C, SG C FLOW MISMATCH, ALR 00-110B, SG C LEV DEV and/or OFN SB-008, INSTRUMENT MALFUNCTIONS. OFN SB-008, INSTRUMENT MALFUNCTIONS Attachment C, SG Pressure Channel Malfunction is used to identify and mitigate the instrument failure. Memory actions are performed by the BOP (identify the failure, place C Main Feed Regulating Valve, AE FK-530, in manual, and control Steam Generator level). Technical Specifications are identified by the SRO. TS 3.3.2, ENGINEERED SAFETY FEATURES INSTRUMENTATION, Condition A (Immediately entered due to failure) and from Table 3.3.2-1, Fu 1.e and 4.e, Condition D (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to trip bistables respectively) are identified.

Event 5: Reactivity event: The Shift Manager (cue) informs the Control Room Supervisor that Motor Driven Auxiliary Feedwater Pump B has been declared INOPERABLE but AVAILABLE. The SRO determines per Technical Specification 3.7.5, Condition C, (Two AFW trains inoperable), Action C.1 (Be in MODE 3 within six hours), that a downpower must be initiated. If the pre-shift brief for a 10% downpower is not begun, the Shift Manager cues that the crew downpower using OFN MA-038, RAPID PLANT SHUTDOWN.

Event 6: Major event: 600 gpm Cold Leg break, Loop C - Loss Of Coolant Accident (LOCA). Once the downpower is initiated, a 600 gpm LOCA occurs. The crew diagnoses the LOCA per OFN BB-007, RCS LEAKAGE HIGH, and determines that a Reactor Trip and Safety Injection must be actuated. The crew enters EMG E-0, REACTOR TRIP OR SAFETY INJECTION. The crew will transition to EMG E-1, LOSS OF REACTOR OR SECONDARY COOLANT and then based on plant conditions, transition to EMG ES-11, POST LOCA COOLDOWN AND DEPRESSURIZATION or EMG ES-03, SI TERMINATION.

4 FINAL NRC 4 9

Event 7: Preloaded and post trip: Turbine Driven Auxiliary Feedwater Pump (TDAFP) autostart failure, manual start available. MDAFW B AFW discharge to Steam Generators A and D are throttled. The BOP diagnoses the TDAFW pump did not autostart and that MDAFW B discharge to Steam Generators A and D is low. AFW total flow must be greater than 270,000 lbm/hr until narrow range level in at least one Steam Generator is greater than 6%. TDAFW pump must be started manually from the Control Room.

Critical Task: Establish 270, 000 lbm/hr Auxiliary Feedwater flow before completion of Attachment F of EMG E-0.

AP 15C-003, PROCEDURE USERS GUIDE ABNORMAL OPERATIONS, step 6.1.7 or EMG E-0, REACTOR TRIP OR SAFETY INJECTION, step 8 RNO b (start the pumps and throttle AFW) and/or Attachment F, Automatic Signal Verification, step F4 RNO b (starts TDAFW pump).

Event 8: Preloaded and post trip: Train Bravo CPIS and CISA autostart failure occurs; however, manual actuation available using SA HS-15 and SB HS-48 respectively; additionally, upon manual actuation, CTMT ATMS MONITOR SPLY CTMT ISO VLV, GS HV-36 and CTMT ATMS MONITOR RETURN CTMT ISO VLV, GS HV-34, remain open, manual closure available using GS HIS-36 and GS HIS-34.

Per EMG E-0, REACTOR TRIP OR SAFETY INJECTION, Attachment F, Automatic Signal Verification, step F3 RNOa, the ATC actuates CISA for Train Bravo using SB HS-48 and at step F9 RNOa actuates CPIS, SA HS-15 for Train Bravo and closes GS HV-36 and GS HV-34, isolating Containment.

Critical Task: Close containment isolation valves such that at least one valve is closed on each critical phase-A penetration before completion of Attachment F of EMG E-0.

SCENARIO TERMINATION:

Successful mitigation of the scenario requires the crew identify and mitigate the LOCA per EMG E-1, LOSS OF REACTOR OR SECONDARY COOLANT and then based on plant conditions, transition to EMG ES-11, POST LOCA COOLDOWN AND DEPRESSURIZATION or EMG ES-03, SI TERMINATION.

CRITICAL TASKS (CT):

Event 7: Establish 270, 000 lbm/hr Auxiliary Feedwater (AFW) flow before completion of Attachment F of EMG E-0. NOTE: AFW total flow must be greater than 270,000 lbm/hr until narrow range level in at least one Steam Generator is greater than 6%. TDAFW pump must be started manually from the Control Room.

Event 8: Close containment isolation valves such that at least one valve is closed on each critical phase-A penetration before completion of Attachment F of EMG E-0. Close CTMT ATMS MONITOR SPLY CTMT ISO VLV, GS HV-36 and CTMT ATMS MONITOR RETURN CTMT ISO VLV, GS HV-34, isolating Containment.

5 FINAL NRC 4 9

TECHNICAL SPECIFICATIONS:

Event 1: Pressurizer (PZR) level channel, BB PI-459, fails low. TS 3.3.1, REACTOR TRIP SYSTEM INSTRUMENTATION, Condition A (Immediately entered due to failure) and from Table 3.3.1-1, Fu 9, Condition M (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to trip bistables) is identified.

Event 3: Service Water Pump trip. TR 3.7.8, SERVICE WATER SYSTEM, Condition A (60 days to restore to FUNCTIONAL status) is identified.

Event 4: Steam Generator C pressure channel, AB PI-535A, fails high. TS 3.3.2, ENGINEERED SAFETY FEATURES INSTRUMENTATION, Condition A (Immediately entered due to failure) and from Table 3.3.2-1, Fu 1.e and 4.e, Condition D (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to trip bistables respectively) are identified.

Event 5: The SRO determines per Technical Specification 3.7.5, Condition C, (Two AFW trains inoperable), Action C.1 (Be in MODE 3 within six hours), that a downpower must be initiated.

PRA/PSA: On March 31, 2013, NE 13-0022 provided the Notice of Probabilistic Risk Assessment (PRA)

Model Revision 6.

While the official Top Ten risk significant systems have not been officially determined, by analyzing the Core Damage Frequency (CDF) by Initiating Event and Large Early Release Frequency (LERF) by Initiating Event tables, the following systems are very important:

Service Water see Scenario 4 6

FINAL NRC 4 9