es -401 pwr examination outline (ro) form es-401-2 · the generic (g) k/as in tiers 1 and 2 shall...

Rev 1

ES-401 PWR Examination Outline (RO) Form ES-401-2

Facility: Callaway Date of Exam: December 2014

Tier

Group

RO K/A Category Points SRO-Only Points

K1

K2

K3

K4

K5

K6

A1

A2

A3

A4

G *

Total

A2 G* Total

1. Emergency &

Abnormal Plant

Evolutions

1 4 2 2

N/A

3 4

N/A

3 18 6

2 1 2 0 2 1 3 9 4

Tier Totals 5 4 2 5 5 6 27 10

2.

Plant Systems

1 3 3 2 3 2 2 3 2 3 3 2 28 5

2 1 0 1 2 1 1 1 1 1 1 0 10 3

Tier Totals 4 3 3 5 3 3 4 3 4 4 2 38 8

3. Generic Knowledge and Abilities Categories

1 2 3 4 10 1 2 3 4 7

3 3 2 2

Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outlines (i.e., 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.

Rev 1

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)

E/APE # / Name / Safety Function K1

K2

K3

A1

A2

G K/A Topic(s) IR #

000007 (BW/E02&E10; CE/E02) Reactor Trip - Stabilization - Recovery / 1

X EA1.04 Ability to operate and monitor the following as they apply to a reactor trip: RCP operation and flow rates. (CFR 41.7 / 45.5 / 45.6)

3.6 1

000008 Pressurizer Vapor Space Accident / 3

000009 Small Break LOCA / 3 X EK3.15 Knowledge of the reasons for the following responses as the apply to the small break LOCA: Closing of RCP thermal barrier outlet valves.

(CFR 41.5 / 41.10 / 45.6 / 45.13)

3.2 2

000011 Large Break LOCA / 3 X EK2.02 Knowledge of the interrelations between the Large Break Loca and the following: Pumps. (CFR 41.7 / 45.7)

2.6 3

000015/17 RCP Malfunctions / 4

000022 Loss of Rx Coolant Makeup / 2 X AA2.04 Ability to determine and interpret the following as they apply to the Loss of Reactor Coolant Makeup: How long PZR level can be maintained within limits.

(CFR 43.5/ 45.13)

2.9 4

000025 Loss of RHR System / 4 X AK1.01 Knowledge of the operational implication of the following concepts as they apply to A loss of Residual Heat Removal System: Loss of RHRS during all modes of operation.

(CFR 41.8 / 41.10 / 45.3)

3.9 5

000026 Loss of Component Cooling Water / 8

X AA1.01 Ability to operate and / or monitor the following as they apply to the Loss of Component Cooling Water: CCW temperature indications.

(CFR 41.7 / 45.5 / 45.6)

3.1 6

000027 Pressurizer Pressure Control System Malfunction / 3

X AA2.03 Ability to determine and interpret the following as they apply to the Pressurizer Pressure Control Malfunctions: Effects of RCS pressure changes on key components in plant.

(CFR: 43.5 / 45.13)

3.3 7

000029 ATWS / 1 X EK1.01 Knowledge of the operational implications of the following concepts as they apply to the ATWS: Reactor nucleonics and thermo-hydraulics behavior.

(CFR 41.8 / 41.10 / 45.3)

2.8 8

Rev 1

000038 Steam Gen. Tube Rupture / 3 X EA1.19 Ability to operate and monitor the following as they apply to a SGTR: MFW System status indicator.

(CFR 41.7 / 45.5 / 45.6)

3.4 9

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 X EK1.01 Knowledge of the operational implications of the following concepts as they apply to the Station Blackout : Effect of battery discharge rates on capacity. (CFR 41.8 / 41.10 / 45.3)

3.3 10

000056 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 (CFR 41.5,41.10 / 45.6 / 45.13)

3.5 11

000057 Loss of Vital AC Inst. Bus / 6 X AA2.03 Ability to determine and interpret the following as they apply to the Loss of Vital AC Instrument Bus: RPS panel alarm annunciators and trip indicators .

(CFR: 43.5 / 45.13)

3.7 12

000058 Loss of DC Power / 6 X 2.2.37 Ability to determine operability and/or availability of safety related equipment. (CFR: 41.7 / 43.5 / 45.12)

3.6

13

000062 Loss of Nuclear Svc Water / 4 X AA2.02 Ability to determine and interpret the following as they apply to the Loss of Nuclear Service Water: The cause of possible SWS loss.

(CFR: 43.5 / 45.13)

2.9 14

000065 Loss of Instrument Air / 8

W/E04 LOCA Outside Containment / 3 X EK1.2 Knowledge of the operational implications of the following concepts as they apply to the (LOCA Outside Containment): Normal, abnormal and emergency operating procedures associated with (LOCA Outside Containment).

(CFR: 41.8 / 41.10, 45.3)

3.5 15

W/E11 Loss of Emergency Coolant Recirc. / 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.

(CFR: 41.7 / 45.7)

3.6 16

Rev 1

BW/E04; W/E05 Inadequate Heat Transfer - Loss of Secondary Heat Sink / 4

X 2.4.21 Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc. (CFR: 41.7 / 43.5 / 45.12)

4.0 17

000077 Generator Voltage and Electric Grid Disturbances / 6

X 2.1.27 Knowledge of system purpose and/or function. (CFR: 41.7)

3.9 18

K/A Category Totals: 4 2 2 3 4 3 Group Point Total: 18

Rev 1

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)


K2

K3

A1

A2

G K/A Topic(s) IR #

000001 Continuous Rod Withdrawal / 1

000003 Dropped Control Rod / 1 X AK1.02 Knowledge of the operatinal implicaitons of the following concepts as they apply to Dropped Control Rod: Effects of turbine-reactor power mismatch on rod control.

3.1 26

000005 Inoperable/Stuck Control Rod / 1

000024 Emergency Boration / 1 X AK2.01 Knowledge of the interrelations between Emergency Boration and the following: Valves .

(CFR 41.7 / 45.7)

2.7 19

000028 Pressurizer Level Malfunction / 2 X 2.4.46 Ability to verify that the alarms are consistent with the plant conditions. (CFR: 41.10 / 43.5 / 45.3 / 45.12)

4.2 20

000032 Loss of Source Range NI / 7 X AA2.02 Ability to determine and interpret the following as they apply to the Loss of Source Range Nuclear Instrumentation: Expected change in source range count rate when rods are moved. (CFR: 43.5 / 45.13)

3.6 21

000033 Loss of Intermediate Range NI / 7

000036 (BW/A08) Fuel Handling Accident / 8

000037 Steam Generator Tube Leak / 3 X AA1.04 Ability to operate and / or monitor the following as they apply to the Steam Generator Tube Leak: Condensate air ejector exhaust radiation monitor and failure indicator . (CFR 41.7 / 45.5 / 45.6)

3.6 22

000051 Loss of Condenser Vacuum / 4 X 2.1.20 Ability to interpret and execute procedure steps. (CFR: 41.10 / 43.5 / 45.12)

4.6 23

000059 Accidental Liquid RadWaste Rel. / 9

000060 Accidental Gaseous Radwaste Rel. / 9 X 2.2.38 Knowledge of conditions and limitations in the facility license. (CFR: 41.7 / 41.10 / 43.1 / 45.13)

3.6 24

000061 ARM System Alarms / 7

000067 Plant Fire On-site / 8

000068 (BW/A06) Control Room Evac. / 8

Rev 1

000069 (W/E14) Loss of CTMT Integrity / 5 X EA1.3 Ability to operate and / or monitor the following as they apply to the (High Containment Pressure): Desired operating results during abnormal and emergency situations. (CFR: 41.7 / 45.5 / 45.6)

3.3 25

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

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 X EK2.2 Knowledge of the interrelations between the (Pressurized Thermal Shock) 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)

3.6 27

CE/A16 Excess RCS Leakage / 2

CE/E09 Functional Recovery

K/A Category Point Totals: 1 2 0 2 1 3 Group Point Total: 9

Rev 1

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 K1

K2

K3

K4

K5

K6

A1

A2

A3

A4

G K/A Topic(s) IR #

003 Reactor Coolant Pump X A1.09 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RCPS controls including: Seal flow and D/P

(CFR: 41.5 / 45.5)

2.8 28

003 Reactor Coolant Pump X 2.1.28 Knowledge of the purpose and function of major system components and controls. (CFR: 41.7)

4.1 29

004 Chemical and Volume Control

X A2.25 Ability to (a) predict the impacts of the following malfunctions or operations on the CVCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Uncontrolled boration or dilution.

(CFR: 41.5/ 43.5 / 45.3 / 45.5)

3.8 30

005 Residual Heat Removal X A1.01 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RHRS controls including: Heatup/cooldown rates (CFR: 41.5 / 45.5)

3.5 31

006 Emergency Core Cooling X K5.06 Knowledge of the operational implications of the following concepts as they apply to ECCS: Relationship between ECCS flow and RCS pressure. (CFR: 41.5 / 45.7)

3.5 32

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 the PRT.

(CFR: 41.5 / 43.5 / 45.3 / 45.13)

2.6 33

Rev 1

008 Component Cooling Water X A4.01 Ability to manually operate and/or monitor in the control room: CCW indications and controls.

(CFR: 41.7 / 45.5)

3.3 34

008 Component Cooling Water X K1.05 Knowledge of the physical connections and/or cause-effect relationships between the CCWS and the following systems: Sources of makeup water. (CFR: 41.2 to 41.9 / 45.7 to 45.9)

3.0 35

010 Pressurizer Pressure Control X K2.03 Knowledge of bus power supplies to the following: Indicator for PORV position

(CFR: 41.7)

2.8 36

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 . (CFR: 41.5 / 45.5)

2.9 37

012 Reactor Protection X K3.03 Knowledge of the effect that a loss or malfunction of the RPS will have on the following: SDS.

(CFR: 41.7 / 45.6)

3.1 38

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. (CFR: 41.7 / 45.5 to 45.8)

2.7 39

022 Containment Cooling X A4.01 Ability to manually operate and/or monitor in the control room: CCS fans. (CFR: 41.7 / 45.5 to 45.8)

3.6 40

025 Ice Condenser

026 Containment Spray X K4.04 Knowledge of CSS design feature(s) and/or interlock(s) which provide for the following: Reduction of temperature and pressure in containment after a LOCA by condensing steam, to reduce radiological hazard, and protect equipment from corrosion damage (spray).

(CFR: 41.7)

3.7 41

Rev 1

039 Main and Reheat Steam X K5.08 Knowledge of the operational implications of the following concepts as the apply to the MRSS: Effect of steam removal on reactivity.

(CFR: 441.5 / 45.7)

3.6 42

059 Main Feedwater X A3.04 Ability to monitor automatic operation of the MFW, including: Turbine driven feed pump.

(CFR: 41.7 / 45.5)

2.5 43

061 Auxiliary/Emergency Feedwater

X 2.2.38 Knowledge of conditions and limitations in the facility license. (CFR: 41.7 / 41.10 / 43.1 / 45.13)

3.6 44


X

K1.01 Knowledge of the physical connections and/or causeeffect relationships between the AFW and the following systems: S/G system.

(CFR: 41.2 to 41.9 / 45.7 to 45.8)

4.1 45

062 AC Electrical Distribution X K4.03 Knowledge of ac distribution system design feature(s) and/or interlock(s) which provide for the following: Interlocks between automatic bus transfer and breakers.

(CFR: 41.7)

2.8 46

062 AC Electrical Distribution X K2.01 Knowledge of bus power supplies to the following: Major system loads. (CFR: 41.7)

3.3 47

063 DC Electrical Distribution X A3.01 Ability to monitor automatic operation of the DC electrical system, including: Meters, annunciators, dials, recorders, and indicating lights. (CFR: 41.7 / 45.5)

2.7 48

064 Emergency Diesel Generator X K6.07 Knowledge of the effect of a loss or malfunction of the following will have on the ED/G system: Air receivers.

(CFR: 41.7 / 45.7)

2.7 49

064 Emergency Diesel Generator X K3.02 Knowledge of the effect that a loss or malfunction of the ED/G system will have on the following: ESFAS controlled or actuated systems. (CFR: 41.7 / 45.6)

4.2 50

Rev 1

073 Process Radiation Monitoring

X K1.01 Knowledge of the physical connections and/or cause-effect relationships between the PRM system and the following systems: Those systems served by PRMs. (CFR: 41.2 to 41.9 / 45.7 to 45.8)

3.6 51

076 Service Water X A3.02 Ability to monitor automatic operation of the SWS, including: Emergency heat loads.

(CFR: 41.7 / 45.5)

3.7 52

076 Service Water X K2.01 Knowledge of bus power supplies to the following: Service water.

(CFR: 41.7)

2.7 53

078 Instrument Air X K4.02 Knowledge of IAS design feature(s) and/or interlock(s) which provide for the following: Cross-over to other air systems.

(CFR: 41.7)

3.2 54

103 Containment X A4.01 Ability to manually operate and/or monitor in the control room: Flow control, pressure control, and temperature control valves, including pneumatic valve controller..

(CFR: 41.7 / 45.5 to 45.8)

3.2 55

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

Rev 1

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 K1

K2

K3

K4

K5

K6

A1

A2

A3

A4

G K/A Topic(s) IR #

001 Control Rod Drive X K4.14 Knowledge of CRDS design feature(s) and/or interlock(s) which provide for the following: Operation parameters, including proper rod speed (CFR: 41.7)

2.6 56

002 Reactor Coolant

011 Pressurizer Level Control

014 Rod Position Indication

015 Nuclear Instrumentation

016 Non-nuclear Instrumentation

017 In-core Temperature Monitor X K1.02 Knowledge of the physical connections and/or causeeffect relationships between the ITM system and the following systems: RCS.

(CFR: 41.2 to 41.9 / 45.7 to 45.8)

3.3 57

027 Containment Iodine Removal

028 Hydrogen Recombiner and Purge Control

X K5.01 Knowledge of the operational implications of the following concepts as they apply to the HRPS: Explosive hydrogen concentration.

(CFR: 41.5 / 45.7)

3.4 58

029 Containment Purge X A1.02 Ability to predict and/or monitor changes in parameters to prevent exceeding design limits) associated with operating the Containment Purge System controls including: Radiation levels.

(CFR: 41.5 / 45.5)

3.4 59

033 Spent Fuel Pool Cooling X K4.01 Knowledge of design feature(s) and/or interlock(s) which provide for the following: Maintenance of spent fuel level.

(CFR: 41.7)

2.9 60

034 Fuel Handling Equipment

Rev 1

035 Steam Generator X K6.03 Knowledge of the effect of a loss or malfunction on the following will have on the S/GS: S/G level detector.

(CFR: 41.7 / 45.7)

2.6 61

041 Steam Dump/Turbine Bypass Control

X K3.01 Knowledge of the effect that a loss or malfunction of the SDS will have on the following: S/G.

(CFR: 41.7 / 45.6)

3.2 62

045 Main Turbine Generator

055 Condenser Air Removal

056 Condensate

068 Liquid Radwaste

071 Waste Gas Disposal

072 Area Radiation Monitoring X A4.03 Ability to manually operate and/or monitor in the control room: Check source for operability demonstration.

(CFR: 41.7 / 45.5 to 45.8)

3.1 63

075 Circulating Water X A2.02 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 Circulating water pumps.

(CFR: 41.5 / 43.5 / 45.3 / 45.13)

2.5 64

079 Station Air

086 Fire Protection X A3.03 Ability to monitor automatic operation of the Fire Protection System including: Actuation of fire detectors. (CFR: 41.7 / 45.5)

2.9 65

K/A Category Point Totals: 1 0 1 2 1 1 1 1 1 1 0 Group Point Total: 10

Rev 1

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


Category K/A # Topic RO SRO-Only

IR # IR #

1. Conduct of Operations

2.1.5 Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc. (CFR: 41.10 / 43.5 / 45.12)

2.9 66

2.1.26 Knowledge of industrial safety procedures (such as rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and hydrogen). (CFR: 41.10 / 45.12)

3.4 67

2.1.34 Knowledge of primary and secondary plant chemistry limits. (CFR: 41.10 / 43.5 / 45.12)

2.7 68

Subtotal 3

2. Equipment Control

2.2.7 Knowledge of the process for conducting special or infrequent tests. (CFR: 41.10 / 43.3 / 45.13)

2.9 69

2.2.13 Knowledge of tagging and clearance procedures. (CFR: 41.10 / 45.13)

4.1 70

2.2.44 Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions. (CFR: 41.5 / 43.5 / 45.12)

4.2 71

Subtotal 3

3. Radiation Control

2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions. (CFR: 41.12 / 43.4 / 45.10)

3.2 72

2.3.11 Ability to control radiation releases. (CFR: 41.11 / 43.4 / 45.10)

3.8 73

Subtotal 2

4. Emergency Procedures / Plan

2.4.34 Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects. (CFR: 41.10 / 43.5 / 45.13)

4.2 74

2.4.43 Knowledge of emergency communications systems and techniques. (CFR: 41.10 / 45.13)

3.2 75

Subtotal 2

Tier 3 Point Total 10 7

Rev 0 Page 1 of 2

ES-401 Record of Rejected K/As (RO) Form ES-401-4

• Random Selection method was performed for re-selecting K/As using pennies with dates ending in 0-9. The pennies were drawn in sequence to designate K-1 thru A4, then drawn to pick the number of the K/A in sequence. Pennies 1-9 were drawn from, then 0-9, then 0-9 again. Generic K/A replacements were drawn from the generic section using the same number selection method. If a K/A could be selected from within the same System and K/A section (ie. A1, K2, etc.) then only the last two numbers were reselected. If a K/A could not be replaced within the same section, then a new section and K/A number were selected within the same system using the method listed above. If a no usable K/A could be selected within the system, then a new system, and K/A was selected from within the Same Tier and Group ensuring no resampling of systems within the Tier and Group unless all Systems were already sampled.

Tier / Group

Randomly Selected K/A (NEW K/A)

Reason for Rejection

(OLD K/A)

1 / 1 009 EK3.15 009 EK3.06 - Unable to write question to meet K/A with plausible distractors. Random selection from within 009 EK3.

1 / 1 011 EK2.02 011 EK1.01 - Audit Overlap. Unable to write another question to K/A and maintain separation from NRC and AUDIT exams. Random selection from within topic 00011.

1 / 1 025 AK1.01 025 AK3.03 – Site specific procedure for Loss of RHR does not have any Immediate Operator Actions. Cannot write a question for this K/A. Random selection from within topic 00025

1 / 2 032 AA2.02 032 AA2.03 – Unable to write discriminating question pertaining to the automatic removal of high voltage to SRNIs. Random selection from within 032 AA2.

1 / 2 003 AK1.02 W/E15 EK1.2 – Rejected due to oversampling. Multiple other K/As all apply to the Containment. Random selection from within unsampled Tier 1/ Group 2 topics, with random K/A sampled within that Topic.

2 / 1 008 A4.01 008 A4.07 – Question redundancy on this sample plan. CCW surge tank control is sampled twice on this sample plan. Random selection from within 008 A4.

2 / 1 022 A4.01 022 A4.02 – Rejected due to equipment not applicable to Callaway. The site does not have CCS pumps. Random selection from within 022 A4.

2 / 1 061 2.2.38 061 2.4.2 – Unable to write question to this generic K/A as it would apply to the AFW system. Random selection from Generic K/A section.

2 / 1 062 K4.03 062 K4.07 – Unable to write question to meet K/A with plausible distractors. Random selection from within 062 K4.

Rev 0 Page 2 of 2

ES-401 Record of Rejected K/As (RO) Form ES-401-4

Tier / Group

Randomly Selected K/A


2 / 1 103 A4.01 103 A4.04 – Audit Overlap. Unable to write another question to K/A and maintain separation from NRC and AUDIT exams. Random selection from within 103 A4.

2 / 2 001 K4.14 001 K4.08 - Rejected due to equipment not applicable to Callaway. The site does not have interlocks preventing excessive rod movement. Interlocks for Rod control that were looked at for question writing could be better tied to other K/As. Random selection from within 001 K4.

2 / 2 072 A4.03 072 A2.02 – Rejected due to equipment not applicable to Callaway. Area Radiation Monitors do not cause any automatic actuations; therefore a failure of an ARM has no impact on plant operation. Random selection from within topic 072.

2 / 2 075 A2.02 075 K2.03 – Rejected due to oversampling. Other K/A in sample plan tests the same knowledge, Power supply to Service Water Pumps. Random selection from within topic 075.

G G 2.2.44 G 2.2.41 – Unable to write question to meet K/A. K/A is better suited to a JPM. Random selection from within Generic K/As.

Rev 1

ES-401 PWR Examination Outline (SRO) Form ES-401-2


Tier

Group

RO K/A Category Points SRO-Only Points

K1

K2

K3

K4

K5

K6

A1

A2

A3

A4

G *

Total

A2 G* Total

1. Emergency &

Abnormal Plant

Evolutions

1

N/A

N/A

18 3 3 6

2 9 2 2 4

Tier Totals 27 5 5 10

2.

Plant Systems

1 28 3 2 5

2 10 2 1 3

Tier Totals 38 5 3 8

3. Generic Knowledge and Abilities Categories

1 2 3 4 10 1 2 3 4 7

2 2 1 2

Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outlines (i.e., 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.

Rev 1

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 (SRO)


K2

K3

A1

A2

G K/A Topic(s) IR #

000007 (BW/E02&E10; CE/E02) Reactor Trip - Stabilization - Recovery / 1

X 2.4.45 Ability to prioritize and interpret the significance of each annunciator or alarm. (CFR: 41.10 / 43.5 / 45.3 / 45.12)

4.3 76

000008 Pressurizer Vapor Space Accident / 3

000009 Small Break LOCA / 3

000011 Large Break LOCA / 3

000015/17 RCP Malfunctions / 4 X 2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. (CFR: 41.5 / 43.5 / 45.12 / 45.13)

4.7 77

000022 Loss of Rx Coolant Makeup / 2

000025 Loss of RHR System / 4

000026 Loss of Component Cooling Water / 8

000027 Pressurizer Pressure Control System Malfunction / 3

000029 ATWS / 1

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 2.2.12 Knowledge of surveillance procedures. (CFR: 41.10 / 45.13)

4.1 78

000057 Loss of Vital AC Inst. 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 . (CFR: 43.5 / 45.13)

4.3 79

000058 Loss of DC Power / 6

000062 Loss of Nuclear Svc Water / 4

000065 Loss of Instrument Air / 8 X AA2.04 Ability to determine and interpret the following as they apply to the Loss of Instrument Air: Typical conditions which could cause a compressor trip (ie. high temperature) (CFR: 43.5 / 45.13)

2.7 80

Rev 1

W/E04 LOCA Outside Containment / 3 X EA2.1 Ability to determine and interpret the following as they apply to the (LOCA Outside Containment): Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

(CFR: 43.5 / 45.13)

4.3 81

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: 6

Rev 1

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 (SRO)


K2

K3

A1

A2

G K/A Topic(s) IR #

000001 Continuous Rod Withdrawal / 1

000003 Dropped Control Rod / 1

000005 Inoperable/Stuck Control Rod / 1

000024 Emergency Boration / 1

000028 Pressurizer Level Malfunction / 2

000032 Loss of Source Range NI / 7

000033 Loss of Intermediate Range NI / 7 X AA2.09 Ability to determine and interpret the following as they apply to the Loss of Intermediate Range Nuclear Instrumentation: Conditions which allow bypass of an intermediate-range level trip switch. (CFR: 43.5 / 45.13)

3.7 82

000036 (BW/A08) Fuel Handling Accident / 8 X 2.4.41 Knowledge of the emergency action level thresholds and classifications. (CFR: 41.10 / 43.5 / 45.11)

4.6 83

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

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 X 2.4.18 Knowledge of the specific bases for EOPs. (CFR: 41.10 / 43.1 / 45.13)

4.0 84

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

Rev 1

BW/E03 Inadequate Subcooling Margin / 4

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 X EA2.2 Ability to determine and interpret the following as they apply to the (Pressurized Thermal Shock): Adherence to appropriate procedures and operation within the limitations in the facility’s license and amendments. (CFR: 43.5 / 45.13)

4.1 85

CE/A16 Excess RCS Leakage / 2

CE/E09 Functional Recovery

K/A Category Point Totals: 2 2 Group Point Total: 4

Rev 1

ES-401 4 Form ES-401-2

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

System # / Name K1

K2

K3

K4

K5

K6

A1

A2

A3

A4

G K/A Topic(s) IR #

003 Reactor Coolant Pump X 2.4.6 Knowledge of EOP Mitigation strategies.

(CFR: 41.10 / 43.5 / 45.13)

4.7 90

004 Chemical and Volume Control

005 Residual Heat Removal X A2.01 Ability to (a) predict the impacts of the following malfunctions or operations on the RHRS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Failure modes for pressure, flow, pump motor amps, motor temperature, and tank level instrumentation

(CFR: 41.5 / 43.5 / 45.3 / 45.13)

2.9 86

006 Emergency Core Cooling X A2.02 Ability to (a) predict the 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: Loss of flow path.

(CFR: 41.5 / 45.5)

4.3 87

007 Pressurizer Relief/Quench Tank

008 Component Cooling Water

010 Pressurizer Pressure Control

012 Reactor Protection

013 Engineered Safety Features Actuation

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

(CFR: 41.5 / 43.5)

4.7 88

022 Containment Cooling

025 Ice Condenser

Rev 1

026 Containment Spray

039 Main and Reheat Steam

059 Main Feedwater


062 AC Electrical Distribution X A2.10 Ability to (a) predict the impacts of the following malfunctions or 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: Effects of switching power supplies on instruments and controls

(CFR: 41.5 / 43.5 / 45.3 / 45.13)

3.3 89

063 DC Electrical Distribution

064 Emergency Diesel Generator

073 Process Radiation Monitoring

076 Service Water

078 Instrument Air

103 Containment


Rev 1

ES-401 5 Form ES-401-2

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

System # / Name K1

K2

K3

K4

K5

K6

A1

A2

A3

A4

G K/A Topic(s) IR #

001 Control Rod Drive

002 Reactor Coolant

011 Pressurizer Level Control

014 Rod Position Indication X 2.2.38 Knowledge of conditions and limitations in the facility license.

(CFR: 41.7 / 41.10 / 43.1)

4.5 91

015 Nuclear Instrumentation

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 X A2.01 Ability to (a) predict the impacts of the following malfunctions or operations on the Spent Fuel Pool Cooling System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Inadequate SDM (CFR: 41.5 / 43.5 / 45.3 / 45.13)

3.5 92

034 Fuel Handling Equipment X A2.01 Ability to (a) predict the impacts of the following malfunctions or operations on the Fuel Handling System, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Dropped Fuel Element.

(CFR: 41.5 / 43.5 / 45.3 / 45.13)

4.4 93

035 Steam Generator

041 Steam Dump/Turbine Bypass Control

045 Main Turbine Generator

055 Condenser Air Removal

056 Condensate

Rev 1

068 Liquid Radwaste

071 Waste Gas Disposal

072 Area Radiation Monitoring

075 Circulating Water

079 Station Air

086 Fire Protection


Rev 1

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


Category K/A # Topic RO SRO-Only

IR # IR #

1. Conduct of Operations

2.1.35 Knowledge of the fuel-handling responsibilities of SROs. (CFR: 41.10 / 43.7)

3.9 94

2.1.13 Knowledge of facility requirements for controlling vital/controlled access. (CFR: 41.10 / 43.5 / 45.9 / 45.10)

3.2 95

Subtotal 2

2. Equipment Control

2.2.20 Knowledge of the process for managing troubleshooting activities. (CFR: 41.10 / 43.5 / 45.13)

3.8 96

2.2.5 Knowledge of the process for making design or operating changes to the facility. (CFR: 41.10 / 43.3 / 45.13)

3.2 97

Subtotal 2

3. Radiation Control

2.3.6 Ability to approve release permits. (CFR: 41.13 / 43.4 / 45.10)

3.8 98

Subtotal 1

4. Emergency Procedures / Plan

2.4.27 Knowledge of “fire in the plant” procedures. (CFR: 41.10 / 43.5 / 45.13)

3.9 99

2.4.37 Knowledge of the lines of authority during implementation of the emergency plan. (CFR: 41.10 / 45.13)

4.1 100

Subtotal 2

Tier 3 Point Total 7

Rev 0 Page 1 of 1

ES-401 Record of Rejected K/As (SRO) Form ES-401-4

• Random Selection method was performed for re-selecting K/As using pennies with dates ending in 0-9. The pennies were drawn in sequence to designate K-1 thru A4, then drawn to pick the number of the K/A in sequence. Pennies 1-9 were drawn from, then 0-9, then 0-9 again. Generic K/A replacements were drawn from the generic section using the same number selection method. If a K/A could be selected from within the same System and K/A section (ie. A1, K2, etc.) then only the last two numbers were reselected. If a K/A could not be replaced within the same section, then a new section and K/A number were selected within the same system using the method listed above. If a no usable K/A could be selected within the system, then a new system, and K/A was selected from within the Same Tier and Group ensuring no resampling of systems within the Tier and Group unless all Systems were already sampled.

Tier / Group

Randomly Selected K/A


1 / 1 065 AA2.04 065 AA2.05 – Rejected due to not having direction to shut down the plant within the Loss of Instrument Air procedure at Callaway. Random selection from within 065 AA2.

1 / 2 033 AA2.09 005 AA2.03 – Overlap with Audit Exam. Unable to write another question to meet K/A and maintain separation from NRC and AUDIT exams. Random selection from within unsampled topics within Tier 1 / Group 2.

2 / 1 006 A2.02 008 A2.02 – Rejected due to oversampling of the CCW system. Random selection from within unsampled topics within Tier 2 / Group 1.

2 / 1 003 2.4.6 103 2.4.6 – Rejected due to oversampling of the Containment system. Using the same Generic K/A, Random selection from the unsampled topics within Tier 2 / Group 1.

2 / 2 034 A2.01 068 2.1.23 – Rejected due to overlap with generic K/A section. Limited required knowledge at Callaway pertaining to release permits. Unable to write question to this topic and K/A. Random selection from the unsampled topics within Tier 2 / Group 2, with random selection of new K/A number within Generic or A2 sections.

ES-301 Administrative Topics Outline Form ES-301-1 Rev 0

NUREG-1021, Revision 9 09/14/2014 9:45 AM

Facility: Callaway Date of Examination: 12/8/2014

Examination Level: RO Operating Test Number: 2014-1

Administrative Topic (see Note)

Type Code* Describe activity to be performed

Conduct of Operations

A.1.a R, M

2.1.25 (3.9) Ability to interpret reference materials, such as graphs, curves, tables, etc.

JPM: Determine Time to Boil for a Loss of Shutdown

Cooling


A.1.b R, D

2.1.18 (3.6) Ability to make accurate, clear, and concise logs, records, status boards, and reports.

JPM: Complete RCS Inventory Balance

Equipment Control

A.2 R, D, P*

2.2.12 (3.7) Knowledge of surveillance procedures. JPM: Review Pump Run Data and Determine if

Acceptance Criteria are Met

Radiation Control

A.3 R, M

2.3.14 (3.4) Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

JPM: Determine if Respirator Needs to be Worn

NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking

only the administrative topics, when 5 are required.

*Type Codes & Criteria: (C)ontrol room 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) (S)imulator *The JPMs from the last 2 exams were randomly selected by placing 8 slips of paper labeled “A1.a 2011” through “A4 2013” in a hardhat. “A2 2011”was drawn from the hardhat.


NUREG-1021, Revision 9 09/14/2014 9:45 AM

ADM A1.a This is a MODIFIED JPM. The applicant will be assigned the task of determining time to

boil given a loss of RHR IAW OTO-EJ-00003, Loss of RHR While Operating at Reduced Inventory or Mid-Loop Conditions. Upon completion of this JPM the operator will have determined the Time to Boil for Condition 1 and Condition 2.

ADM A1.b This is a BANK JPM. The applicant will be assigned the task of determining the RCS

leakage using OSP-BB-00009 ADD 1, RCS Inventory Balance Excessive Leakage or Manual Calculation. Upon completion of this JPM the operator will have identified Total RCS leakage is approximately 0.6821 gpm, Identified RCS Leakage is approximately 0.0764 gpm, and Unidentified RCS Leakage is approximately 0.6057 gpm.

ADM A2 This BANK JPM was used on the 2011 ILT NRC Exam. The applicant will be assigned

the task of completing the calculations and determining if the Acceptance Criteria has been satisfied in accordance with OSP-EN-P001B, TRAIN B CONTAINMENT SPRAY PUMP INSERVICE TEST. Upon completion of this JPM the operator will have calculated pump flow and D/P, completed Attachment 4 of OSP-EN-P001B, TRAIN B CONTAINMENT SPRAY PUMP INSERVICE TEST, and determined that the acceptance criteria is met for the ‘B’ Containment Spray Pump.

ADM A3 This is a MODIFIED JPM. The applicant will be assigned the task of determining if

personnel working on the Spent Fuel Pool rack should wear a respirator. Upon completion of this JPM the operator will have calculated dose with and without a respirator and selected performing the job with a respirator.


Page 1 of 2

Facility: Callaway Date of Examination: 12/8/2014

Examination Level: SRO Operating Test Number: 2014 - 1

Administrative Topic (see Note)

Type Code* Describe activity to be performed


A.1.a R, N

2.1.18 (3.8) Ability to make accurate, clear, and concise logs, records, status boards, and reports.

JPM: Determine Reportability for a Required

Shutdown


A.1.b R, D, P*

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

JPM: Review ECP Calculation

Equipment Control

A.2 R, D

2.2.40 (4.7) Ability to apply Technical Specifications for a system.

JPM: Determine Applicable Technical Specifications

Radiation Control

A.3 R, M

2.3.14 (3.8) Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

JPM: Determine if Respirator Needs to be Worn

Emergency Procedures/Plan

A.4 R or S, D

2.4.43 (3.8) Knowledge of emergency communications systems and techniques

JPM: Make appropriate Offsite Notifications during an

ALERT

NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking

only the administrative topics, when 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)

*The JPMs from the last 2 exams (the 2013 re-exam was also included) were randomly selected by placing 15 slips of paper labeled “A1.a 2011” through “A4 2013R” in a hardhat. “A1.a 2013R”was drawn from the hardhat.


Page 2 of 2

ADM A1.a This is a NEW JPM. The applicant will be assigned the task of determining the initial reportability requirements of a TS required shutdown per APA-ZZ-00520, REPORTING REQUIREMENTS AND RESPONSIBILITIES. Upon completion of this JPM the operator will have determined that a 4 hour report is required to be made to the NRC Operation Center.

ADM A1.b This BANK JPM was used on the 2013 ILT NRC Re-Exam. The applicant will be

assigned the task of reviewing and approving the completed Attachment 1 of OSP-SF-00005, Estimated Critical Position Calculation using the printouts and information provided by the Reactor engineer. Upon completion of this JPM the operator will have identified the errors on Attachment 1 of OSP-SF-00005 “Estimated Critical Position Calculation”.

ADM A2 This is a BANK JPM. The applicant will be assigned the task of determining the impact of

hanging a Hold Off tag on plant safety systems and identifying all applicable Technical Specifications Limiting Condition for Operation (LCO) entries that will apply. Upon completion of this JPM the operator will have determined that the tagging will remove the safety related nitrogen to TDAFP flow control valve ALHV0008 and to Atmospheric Steam Dump ABPV0001. The applicable T/S LCOs are 3.7.4.A and 3.7.5.C.

ADM A3 This is a MODIFIED JPM. The applicant will be assigned the task of determining if

personnel working on the Spent Fuel Pool rack should wear a respirator. Upon completion of this JPM the operator will have calculated dose with and without a respirator and selected performing the job with a respirator.

ADM 4 This is a BANK JPM. The applicant will be assigned the task of determining the

Emergency Event Classification, completing the Sentry Notification form, and performing the notification of off-site agencies within the allotted amount of time. Upon completion of this JPM the operator will have determined the event classification to be an Alert based on EAL HA-1.2, Tornado and damage to table H-1 system, within 15 minutes of start and completed the Sentry notification form and sent it within 15 minutes of the event classification time.

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

Page 1 of 3

Facility: ____Callaway_____________________ Date of Examination: _12/8/14_____ Exam Level: RO SRO-I SRO-U Operating Test No.: __2014-1____

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

System / JPM Title Type Code* Safety Function

S1 004 CVCS (BG) / Perform a Dilution N, S 1

S2 006 ECCS (SB) / Initial RCS Depressurization and SI Block L, M, S 2

S3 010 Reactor Coolant System (BB) / Perform System Surveillance – BBHV8000A Stroke Test

D, P*, S 3

S4 003 Reactor Coolant Pump System (BB) / Start Reactor Coolant Pump during RCS Natural Circulation Cooldown

D, S 4P

S5 026 Containment Spray (EN) / Align Containment Spray for Recirculation A, D, EN, S 5

S6 064 SR Elect Gen and Dist (NE) / Manually Start Diesel Generators A, M, S 6

S7 015 Nuclear Instrumentation System (SE) / NIS Gain Adjustment N, S 7

S8 008 Component Cooling Water System (EG) / Respond to a loss of CCW to the RCPs

A, D, S 8

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

P1 055 Station Blackout / Locally Start (NE01) Emergency Diesel A, D, P*, E 6

P2 054 Loss of Main feedwater / Local TDAFP Start Assuming Loss of AC and DC Power

D, E 4S

P3 068 Control Room Evacuation / Locally Trip Reactor Trip and Bypass Breakers

E, M, R 1

@ 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


Page 2 of 3

(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

*The JPMs from the last 2 exams were randomly selected by placing 22 slips of paper labeled “a” through “k” for the 2011 exam and “S1” through “P3” from the 2013 exam in a hardhat. Two of these items (S2 and P1) were drawn from the hardhat. P1 had been modified when it was placed in the JPM bank, the modified JPM from the bank was used. S1 This is a NEW JPM. The applicant will be assigned the task of diluting the RCS

100 gallons by performing a nominal 120 gpm dilution per OTN-BG-00002, Reactor Makeup Control and Boron Thermal Regeneration System. Upon completion of this JPM, the applicant will have diluted the RCS 100 gallons.

S2 This is a MODIFIED JPM in a SHUTDOWN condition. The applicant will be

assigned the task of depressurizing the RCS and blocking the SI signal, per OTG-ZZ-00006 Addendum 04, Initial RCS Depressurization and SI Block. Upon completion of this JPM, the applicant will have blocked both trains of Pressurizer Pressure and Steam Line Pressure Safety Injection.

S3 This BANK JPM was used on the 2013 ILT NRC Exam. It was randomly selected

using the method described above. The applicant will be assigned the task of performing the Stroke Time Test for BBHV8000A, RCS PZR OUT PWR OPER RLF HV, per OSP-BB-V0001, RCS Valve Inservice Test, Section 6.1. Upon completion of this JPM, the applicant will have completed the Stroke Time Test for BBHV8000A per OSP-BB-V0001 and reported to the CRS that the Acceptance Criteria was not satisfied.

S4 This is a TIME CRITICAL, BANK JPM. The applicant will be assigned the task of

starting ‘D’ Reactor Coolant Pump in accordance with EOP Addendum 3, Starting An RCP. Upon completion of this JPM, the applicant will have started ‘D’ RCP.

S5 This is an ALTERNATE PATH, BANK JPM. The applicant will be assigned the task of completing the actions of ES-1.3, Transfer to Cold Leg Recirculation, Step 6. Upon completion of this JPM, the applicant will have aligned the train ‘B’ Containment Spray Pump to a recirculating lineup and secured the train ‘A’ Containment Spray Pump.


Page 3 of 3

S6 This is an ALTERNATE PATH, MODIFIED JPM. The applicant will be assigned

the task of completing the actions of Step 5 of ECA-0.0, Loss of All AC Power. Upon completion of this JPM, the applicant will have manually started NE01 and manually started ‘A’ ESW Pump.

S7 This is a NEW JPM. The applicant will be assigned the task of performing gain

adjustment for NIS Power Range Channel N41 in accordance with Attachment 1, NIS Power Range Channel N41 Gain Adjustment At SE054A, of OSP-SE-00004, NIS POWER RANGE HEAT BALANCE. Upon completion of this JPM, the operator will have correctly adjusted indicated Reactor Power to a value greater than calculated RTP, but not greater than 1% above calculated RTP.

S8 This is an ALTERNATE PATH, TIME CRITICAL, BANK JPM. The applicant will

be assigned the task of performing Attachment C of OTO-BB-00002, RCP Off-Normal to address a loss of CCW to the RCPs. Upon completion of this JPM, the operator will have restored CCW to Containment and the RCPs.

P1 This ALTERNATE PATH, BANK JPM was used on the 2013 ILT NRC Exam. It

was randomly selected using the method described above. The JPM had been modified after the 2013 ILT NRC Exam when it was placed in the exam bank. The applicant will be assigned the task of locally starting Emergency Diesel Generator NE01 per EOP Addendum 21. Upon completion of this JPM, the applicant will have started Emergency Diesel Generator NE01 and placed the Master Transfer Switch, KJ-HS-9, in AUTO.

P2 This is a BANK JPM. The applicant will be assigned the task of locally starting

TDAFP per EC Supplemental Guideline, Attachment R, Starting TDAFP On Loss of AC and DC Power. Upon completion of this JPM, the applicant will have started the TDAFP and notified the EC that flow to the Steam Generators is available.

P3 This is a MODIFIED JPM. The applicant will be assigned the task of performing

OTO-ZZ-00001, Control Room Inaccessibility Attachment A, Safe Shutdown Operator Actions with Fire. Upon completion of this JPM the operator will have isolated loads on MCC NG04C, placed class 1E electrical equipment A/C unit in ISO/RUN, aligned CCW and locally tripped both reactor trip breakers and bypass breakers.


Page 1 of 3

Facility: ____Callaway_____________________ Date of Examination: _12/8/14_____ Exam Level: RO SRO-I SRO-U Operating Test No.: __2014-1____

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

System / JPM Title Type Code* Safety Function

S1 004 CVCS (BG) / Perform a Dilution N, S 1

S2 006 ECCS (SB) / Initial RCS Depressurization and SI Block L, M, S 2

S3 010 Reactor Coolant System (BB) / Perform System Surveillance – BBHV8000A Stroke Test

D, P*, S 3

S4 003 Reactor Coolant Pump System (BB) / Start Reactor Coolant Pump during RCS Natural Circulation Cooldown

D, S 4P

S5 026 Containment Spray (EN) / Align Containment Spray for Recirculation A, D, EN, S 5

S6 064 SR Elect Gen and Dist (NE) / Manually Start Diesel Generators A, M, S 6

S8 008 Component Cooling Water System (EG) / Respond to a loss of CCW to the RCPs

A, D, S 8

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

P1 055 Station Blackout / Locally Start (NE01) Emergency Diesel A, D, P*, E 6

P2 054 Loss of Main feedwater / Local TDAFP Start Assuming Loss of AC and DC Power

D, E 4S

P3 068 Control Room Evacuation / Locally Trip Reactor Trip and Bypass Breakers

E, M, R 1

@ 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


Page 2 of 3

*The JPMs from the last 2 exams were randomly selected by placing 22 slips of paper labeled “a” through “k” for the 2011 exam and “S1” through “P3” from the 2013 exam in a hardhat. Two of these items (S2 and P1) were drawn from the hardhat. P1 had been modified when it was placed in the JPM bank, the modified JPM from the bank was used. S1 This is a NEW JPM. The applicant will be assigned the task of diluting the RCS

100 gallons by performing a nominal 120 gpm dilution per OTN-BG-00002, Reactor Makeup Control and Boron Thermal Regeneration System. Upon completion of this JPM, the applicant will have diluted the RCS 100 gallons.

S2 This is a MODIFIED JPM in a SHUTDOWN condition. The applicant will be

assigned the task of depressurizing the RCS and blocking the SI signal, per OTG-ZZ-00006 Addendum 04, Initial RCS Depressurization and SI Block. Upon completion of this JPM, the applicant will have blocked both trains of Pressurizer Pressure and Steam Line Pressure Safety Injection.

S3 This BANK JPM was used on the 2013 ILT NRC Exam. It was randomly selected

using the method described above. The applicant will be assigned the task of performing the Stroke Time Test for BBHV8000A, RCS PZR OUT PWR OPER RLF HV, per OSP-BB-V0001, RCS Valve Inservice Test, Section 6.1. Upon completion of this JPM, the applicant will have completed the Stroke Time Test for BBHV8000A per OSP-BB-V0001 and reported to the CRS that the Acceptance Criteria was not satisfied.

S4 This is a TIME CRITICAL, BANK JPM. The applicant will be assigned the task of

starting ‘D’ Reactor Coolant Pump in accordance with EOP Addendum 3, Starting An RCP. Upon completion of this JPM, the applicant will have started ‘D’ RCP.

S5 This is an ALTERNATE PATH, BANK JPM. The applicant will be assigned the

task of completing the actions of ES-1.3, Transfer to Cold Leg Recirculation, Step 6. Upon completion of this JPM, the applicant will have aligned the train ‘B’ Containment Spray Pump to a recirculating lineup and secured the train ‘A’ Containment Spray Pump.

S6 This is an ALTERNATE PATH, MODIFIED JPM. The applicant will be assigned

the task of completing the actions of Step 5 of ECA-0.0, Loss of All AC Power. Upon completion of this JPM, the applicant will have manually started NE01 and manually started ‘A’ ESW Pump.

S8 This is an ALTERNATE PATH, TIME CRITICAL, BANK JPM. The applicant will

be assigned the task of performing Attachment C of OTO-BB-00002, RCP Off-Normal to address a loss of CCW to the RCPs. Upon completion of this JPM, the operator will have restored CCW to Containment and the RCPs.


Page 3 of 3

P1 This ALTERNATE PATH, BANK JPM was used on the 2013 ILT NRC Exam. It

was randomly selected using the method described above. The JPM had been modified after the 2013 ILT NRC Exam when it was placed in the exam bank. The applicant will be assigned the task of locally starting Emergency Diesel Generator NE01 per EOP Addendum 21. Upon completion of this JPM, the applicant will have started Emergency Diesel Generator NE01 and placed the Master Transfer Switch, KJ-HS-9, in AUTO.

P2 This is a BANK JPM. The applicant will be assigned the task of locally starting

TDAFP per EC Supplemental Guideline, Attachment R, Starting TDAFP On Loss of AC and DC Power. Upon completion of this JPM, the applicant will have started the TDAFP and notified the EC that flow to the Steam Generators is available.

P3 This is a MODIFIED JPM. The applicant will be assigned the task of performing

OTO-ZZ-00001, Control Room Inaccessibility Attachment A, Safe Shutdown Operator Actions with Fire. Upon completion of this JPM the operator will have isolated loads on MCC NG04C, placed class 1E electrical equipment A/C unit in ISO/RUN, aligned CCW and locally tripped both reactor trip breakers and bypass breakers.

ES-301-5

Facility: Callaway Date of Exam: 12/08/2014 Operating Test No.: 2014-1

Crew A

A P P L I C A N T

E V E N T

T Y P E

Scenarios 2 3

4 T

O T A L

M I N I M U

M(*)

CREW POSITION CREW POSITION CREW POSITION

I1 I2 R1 I1 R1 I2 I2 I1 R1

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

R I U

SRO-I

(I1)

RX 3 4 2 1

NOR 1 1 2 1

I/C 3,4,5,6 1,2,3,

4 2,4 10 4

MAJ 7 5,6 5 4 2

TS 2,3 1,2 4 2

SRO-I

(I2)

RX 4 1 1

NOR 1 1 1

I/C 5,6 2,3,4 2,3,4 8 4

MAJ 7 5,6 5 4 2

TS 3,4 2 2

RO

(R1)

RX 3 1 1

NOR 1 1 1

I/C 4,5,6 1,4 3,4 7 4

MAJ 7 5,6 5 4 2

TS 0 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 applicant’s competence count toward the minimum requirements specified for the applicant’s license level in the right-hand columns.

ES-301-5


Crew B

A P P L I C A N T

E V E N T

T Y P E

Scenarios 2 3

4 5 T

O T A L

M I N I M U

M(*)

CREW POSITION

CREW POSITION

CREW POSITION

CREW POSITION

I3 I5 I6 I4 I6 I3 I5 I3 I4 I6 I4 I5

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

R I U

SRO-I

(I3)

RX 3 1 1

NOR 0* 1

I/C 1,2,3,4

3,4

2,3 8 4

MAJ 5,6 5 4 4 2

TS 1,2 2 2

SRO-I

(I4)

RX 4 1 1

NOR 1 1 2 1

I/C 5,6 2,3,4

5 4

MAJ 7 5 4 3 2

TS 3,4 2 2

SRO-I

(I5)

RX 3 1 1

NOR 1 1 2 1

I/C 4,5,6

1,4 2,3 7 4

MAJ 7 5,6 4 4 2

TS 2,4 2 2

SRO-I

(I6)

RX 4 1 1

NOR 1 1 2 1

I/C 3,4,5,6

2,3,4

2,4 9 4

MAJ 7 5,6 5 4 2

TS 2,3 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




3. Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to

the applicant’s competence count toward the minimum requirements specified for the applicant’s license level in the right-hand columns.

ES-301-5


A P P L I C A N T

E V E N T

T Y P E

Scenarios 1

T

O T A L

M I N I M U

M(*)

CREW POSITION

CREW POSITION

CREW POSITION

CREW POSITION

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

R I U

SPARE

RX

NOR

I/C 1,2,3,4

1,3,4

2,4

MAJ 5 5 5

TS 1,3 Instructions: 3. 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




3. Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to

the applicant’s competence count toward the minimum requirements specified for the applicant’s license level in the right-hand columns.

SRO ATC BOP Scenario 5 12/9/14 0730-0900 Crew B I6 - I4 - I5 Scenario 2 12/10/14 0730-0900 Crew A I1 - I2 - R1 - 12/10/14 0945-1200 Crew B I3 - I5 - I6 - Scenario 3 12/11/14 0730-0900 Crew A I1 - R1 - I2 - 12/11/14 0945-1200 Crew B I4 - I6 - I3 - Scenario 4 12/12/14 0730-0900 Crew A I2 - I1 - R1 - 12/12/14 0945-1200 Crew B I5 - I3 - I4 -

Appendix D Scenario Outline Form ES-D-1

Facility: Callaway Scenario No.: 1, Rev 0 Op-Test No.: 2014-1 Examiners: ____________________________ Operators: _____________________________

____________________________ _____________________________ ____________________________ _____________________________

Initial Conditions: 100% Power, Steady State Conditions. Turnover: ‘B’ CCP is tagged out due to oil leak

Event

No.

Malf. No. Event Type*

Event Description

1 BB/PT0456 SRO (I) RO (I) PZR Pressure instrument Fail high (Tech Spec)

2 (AB) ABFT0542 SRO (I) BOP (I) SG Steam Flow Inst Fail

3 SF/SFB08_DR SRO (C) RO (C) Dropped rod (Tech Spec)

4 SF/SF2CD2/2

SRO (C) RO (C) BOP (C)

Second Dropped Rod / Rx Trip

5 AB/AB001_D

SRO (M) RO (M) BOP (M)

Steam Line Break inside containment

6 SAS9XX_1, 2, 3, 4 SRO (C) BOP (C) MSIVs fail to close

7 (EN) PEN01A_2 and

EN08RL_420TVSP

SRO (C) BOP (C)

Containment Spray Pump Auto Start Failure and Containment Spray Pump Discharge Valve Failure

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

Target Quantitative Attributes (Per Scenario; See Section D.5.d)

Actual Attributes

1. Total malfunctions (5-8) 7 2. Malfunctions after EOP entry (1-2) 2 3. Abnormal events (2-4) 3 4. Major transients (1-2) 1 5. EOPs entered/requiring substantive actions (1-2) 1 6. EOP contingencies requiring substantive actions (0-2) 0 7. Critical tasks (2-3) 2

Scenario Event Description Callaway 2014 NRC Exam Scenario #1, rev. 0

The plant is at 100% with the ‘B’ CCP tagged out After the crew takes the watch, the lower selected PZR Pressure Channel BB PT-456 fails high. Annunciator PZR PRESS HI and RX PARTIAL TRIP will alarm to alert the operator of the malfunction. The CRS should enter OTO-BB-00006, Pressurizer Pressure Control Malfunction, and transfer pressure control to remove the failed channel. After the Technical Specifications from the previous event have been identified, a Steam Flow Channel on ‘D’ S/G Fails Low. The crew will respond using, OTO-AE-00002, Steam Generator Water Level Control Malfunctions and select SG steam flow channel selector to an operable channel on AB FS-542C and DFWCS. After the crew has bypassed the failed channel, a dropped rod occurs as indicated by DRPI and Control Rod Alarms. The ROs will perform the immediate actions of OTO-SF-00001, Rod Control/Malfunctions, to place rods in manual. The crew will stabilize the plant and identify Technical Specifications. After Tech Specs have been identified, two more control rods drop into the core. The operators will take immediate action to trip the reactor. After the Reactor Trip, a Steam Line Break inside containment will occur. The MSIVs will fail to auto close. The crew will respond IAW E-0, Reactor Trip or Safety Injection, and transition to E-2, Faulted Steam Generator Isolation. The ‘A’ Containment Spray Pump fails to auto start and the ‘B’ Containment Spray Pump discharge valve fails to open. The crew can manually start the ‘A’ Containment Spray Pump and/or direct an OT to open EN-HV-12, CS PUMP B DISCH, in the field while performing the actions of E-0. The crew will continue to perform the actions of E-2, and isolate the ‘D’ SG, then transition to ES-1.1, SI Termination. The scenario can be terminated after the transition to ES-1.1 occurs

Callaway 2014 NRC Exam Scenario #1, rev. 0 Critical Tasks: Event #5 Isolate the faulted ‘D’ SG before transition out of E-2

CT-17 Isolate faulted SG Event #7 Manually actuate containment cooling by:

1) Starting the 'A' Containment Spray Pump before a RED path develops to the containment CSF

OR 2) Directing an equipment operator to locally OPEN EN-HV-12, CS PUMP B

DISCH before a RED path develops to the containment CSF CT-3 Manually actuate containment cooling

References OTO-BG-00001, Pressurizer Level Control Malfunction OTO-AE-00002, Steam Generator Water Level Control Malfunctions OTO-SF-00001, Rod Control/Malfunctions E-0, Reactor Trip or Safety Injection E-2, Faulted Steam Generator Isolation ES-1.1, SI Termination Tech Spec 3.3.1.E & M for Reactor Trip System Instrumentation Tech Spec 3.3.2.D and L for Reactor Trip System Instrumentation Tech spec 3.1.4.B for Rod Group Alignment Limits ODP-ZZ-00025, EOP/OTO User's Guide


Facility: Callaway Scenario No.: 2, rev. 0 Op-Test No.: 2014-1

Examiners: ____________________________ Operators: _____________________________

____________________________ _____________________________ ____________________________ _____________________________

Initial Conditions: 100% Power, Steady State Conditions. Turnover: NE02 is tagged out for an LCO. Currently in hour 20 of 30 hours scheduled duration. On coming crew will be swapping Service Water Pumps for maintenance.

Event No.


Event Description

1 N/A SRO (N) BOP (N)

Start Service Water Pump ‘C’ Secure Service Water Pump ‘A’

2 N/A SRO RO

RCS High Activity / Place 120 gpm Letdown in-service (Tech Spec)

3 BNLT0932 SRO (I) Refueling Water Storage Tank (RWST) Level Channel Fails Low (Tech Spec)

4 PCD01_Trip SRO (C) BOP (C)

Loss of Main Seal Oil Pump

5 NB01_F SRO (C) RO (C) BOP (C)

Loss of NB01

6 KAC01_1 SRO (C) RO (C) BOP (C)

Turbine Trip / Reactor Trip

7 Lossofswitchyard.lsn


Loss of all AC Power

8 PAL02_3 SRO (C) BOP (C)

TDAFW fails to Auto Start



Actual Attributes


Scenario Event Description Callaway 2014 Scenario #2, rev. 0

The plant is operating at 100%, steady state power. NE02 is tagged out for an LCO. Currently in hour 20 of 30 hours scheduled duration. On coming crew will be swapping Service Water Pumps for maintenance. After the crew takes the watch, the crew will start ‘C’ Service Water Pump and secure ‘A’ Service Water Pump IAW OTN-EA-00001, Service Water System. After the Service Water Pumps are swapped, a report from Chemistry indicates high activity in the RCS. The Crew will enter OTO-BB-00005, RCS High Activity and establish 120 gpm letdown flow IAW OTN-BG-00001 Addendum 04, Operation of CVCS Letdown. Tech Spec 3.4.16 applies. After Tech Specs have been addressed for the RCS High Activity, a Refueling Water Storage Tank (RWST) level channel fails low. The crew will respond IAW OTO-BN-00001, RWST Level Channel Malfunction. Tech Spec 3.3.2 applies. After Tech Specs have been addressed for the RWST Level Channel Malfunction, the Main Seal Oil Pump will trip with a failure of the Emergency Seal Oil Pump to automatically start. The crew will respond IAW OTO-MA-00002, Generator Seal Oil System Malfunction and manually start the Emergency Seal Oil Pump. After the crew has addressed the Main Seal Oil Pump trip, the NB01 bus will trip and lockout. The crew should enter OTO-NB-00001, Loss of Power to NB01. OTO-NB-00001 will direct the crew to secure NE01 and verify the electric plant line up. After the Technical Specifications for the loss of NB01 have been identified a Turbine Trip will occur. This will cause a Reactor Trip. The crew should respond by entering E-0, Reactor Trip or Safety Injection. The crew will then transition to ES-0.1, Reactor Trip Response. After the crew has transition to ES-0.1 and prior to Step 6 a Loss of Off Site Power will occur. The crew will transition to ECA-0.0, Loss of All AC Power. The crew will manually start the TDAFW and restore power using EOP Addendum 39, Alternate Emergency Power Supply. The scenario can be terminated after power has been restored to NB02.

Callaway 2014 NRC Scenario #2, rev. 0 Critical Tasks: Event #8 Establish greater than 285,000 lbm/hr AFW flow rate to the SGs prior to SG level

becoming less than 10% wide range CT - 23, Establish AFW flow during SBO Event #7 Energize at least one AC Emergency Bus prior to RCP seal degradation CT - 24, Energize at least one ac emergency bus

References OTN-EA-00001, Service Water System OTO-BB-00005, RCS High Activity OTN-BG-00001 Addendum 04, Operation of CVCS Letdown OTO-BN-00001, RWST Level Channel Malfunction OTO-MA-00002, Generator Seal Oil System Malfunction OTO-NB-00001, Loss of Power to NB01 E-0, Reactor Trip or Safety Injection ES-0.1, Reactor Trip Response ECA-0.0, Loss of All AC Power EOP Addendum 39, Alternate Emergency Power Supply



____________________________ _____________________________ ____________________________ _____________________________

Initial Conditions: 100% Power, Steady State Conditions. Turnover: ‘B’ CCP is tagged out due to oil leak

Event No.


Event Description

1 SEN0044 SRO (I) RO (I) Nuclear Instrument Fails. (Tech Spec)

2 AELT0519 SRO (I) BOP (I) SG Level Instrument Fail (Tech Spec)

3 PAE01A_2 PAE01A_3 PAE01A_4

SRO (C)(R) RO (R) BOP (C)

MFP Vibrations Forcing Downpower

4 PAE01a_1 SRO (C) RO (C) BOP (C)

MFP Trip / Rx Trip

5 SF006 SRO (M) RO (M) BOP (M)

Reactor Trip Failure / ATWS

6 TBB03 SRO (M) RO (M) BOP (M)

Pressurizer Steam Space Leak

7 SIS_A SIS_B

SRO (C) RO (C) Automatic SI Actuation Failure Both Trains



Actual Attributes


Scenario Event Description Callaway 2014-1 NRC Scenario #3, rev. 0

The plant is steady at 100% power. After the crew takes the watch, Power Range NI, N44 will fail high. The crew will respond with OTO-SE-00001, Nuclear Instrument Malfunction, to select an operable channel and bypass the failed instrument. The CRS will evaluate Tech Spec 3.3.1. After the failed NI has been bypassed and the CRS has identified the applicable Tech Specs, Steam Generator ‘A’ controlling level channel slowly fails to 75%. The crew will respond using OTO-AE-00002, Steam Generator Water Level Control Instrument Malfunctions, to control SG level. The CRS will evaluate Tech Spec 3.3.1. and 3.3.2. After Tech Specs have been addressed, ‘A’ MFP will develop vibrations. The crew will respond with OTO-MA-00008, Rapid Load Reduction to lower power to below 65%. The crew will lower turbine load and borate to lower power. After the crew has lowered power 5 to 10 MWe and prior to reducing power less than 70%, ‘A’ MFP will trip. The crew will respond IAW OTO-AE-00001, Feedwater System Malfunction and attempt to trip the reactor. When the crew attempts to trip the reactor rods will not insert. The crew will respond by entering E-0, Reactor Trip or Safety Injection and transitioning to FR-S.1, Response to Nuclear Power Generation/ATWS. The crew will be able to successfully de-energize PG19 and 20, insert control rods, and emergency borate the RCS. After rods have been inserted and the reactor has been shutdown the crew will transition back to E-0, Reactor Trip or Safety Injection. After the transition back to E-0, a leak in the PZR steam space develops. When the RCS depressurizes SI fails to automatically initiate. The crew will manually initiate SI and transition to E-1, Loss of Reactor or Secondary Coolant. The scenario can be terminated after the EDGs are secured in E-1.

Callaway 2014-1 NRC Scenario #3, rev. 0 Critical Tasks: Event #5 Insert negative reactivity into the core by at least one of the following methods before

completing the immediate action steps of FR-S.1 • Deenergize PG19 and PG20 • Insert Control Rods • Establish emergency boration flow to the RCS

CT- 52, Insert negative reactivity into the core Event #7 Manually actuate at least one train of SIS-actuated safeguards equipment before

transition to E-1 or completion of ES-0.1 step 3 CT-2, Manually actuate Sl

References OTO-SE-00001, Nuclear Instrument Malfunction OTO-AE-00002, Steam Generator Water Level Control Instrument Malfunctions OTA-RK-00026 Add 120E, MFP A VIB ALERT OTO-MA-00008, Rapid Load Reduction OTO-AE-00001, Feedwater System Malfunction E-0, Reactor Trip or Safety Injection E-1, Loss of Primary or Secondary Reactor coolant FR-S.1, Response to Nuclear Power Generation/ATWS Tech Spec 3.3.1 for Reactor Trip System Instrumentation Tech spec 3.3.2 for ESFAS Instrumentation ODP-ZZ-00025, EOP/OTO User's Guide



____________________________ _____________________________ ____________________________ _____________________________

Initial Conditions: 80% Power, Steady State Conditions.

Turnover: ‘A’ MDAFWP is tagged out for Breaker PMTs

Event No.

Malf. No. Event Type* Event Description

1 N/A SRO (N) RO (N)

Reduce Letdown to 75 gpm.

2 WEIUBGTC0130TVSPNO

AUTO

SRO (C) RO (C)

LTDN HX Temp Cont Valve Fails

3 (AB)PV004A SRO (C) BOP (C)

‘D’ SG ASD Fails Open (Tech Spec)

4 (BB) EBB01D

SRO (C)(R) RO (C)(R) BOP (C)

SG Tube leak / Downpower (Tech Spec)

5 (BB) EBB01D


SG Tube Rupture

6 (BB)BBPCV0455A

SRO (C) BOP(C)

PZR PORV Failure



Actual Attributes


Scenario Event Description Callaway 2014 NRC Scenario #4, rev. 0

The plant is steady at 80% power steady state. After the crew takes the watch, they will perform OTN-BG-00001 ADD 4, Changing Letdown from 120 gpm to 75 gpm. After letdown flow has been reduced, the Letdown HX temperature control valve will fail closed. The crew will take actions IAW OTA-RK-00018 ADD 39B. The crew will take manual control of the Letdown HX temperature control valve. After the crew has addressed the letdown HX temperature control valve failure the SG ‘D’ ASD fails open. The crew will respond IAW OTO-AB-00001, Steam Dump Malfunction. The crew will close the ‘D’ ASD. The CRS will evaluate Tech Spec 3.7.4.A. After Tech Specs have been evaluated, ‘D’ SG will develop a tube leak. The crew will respond IAW OTO-BB-00001, Steam Generator Tube Leak. The Crew will lower power and the CRS will evaluate Tech Spec 3.4.13, RCS Operational Leakage. After the power reduction has been commenced, the tube leak will increase. The crew will trip the reactor due to the increased leak and respond IAW E-0, Reactor Trip or Safety Injection. The crew will transition to E-3, Steam Generator Tube Rupture and isolate the ruptured ‘D’ SG. Instrument Air cannot be restored to containment due to a valve failure, therefore normal pressurizer spray is NOT available. When the crew opens the PZR PORV to lower pressure the PORV will stick open. The crew will then have to close the PORV block valve. The scenario can be terminated after the cooldown is in progress and the PORV has been isolated.

Callaway 2014-1 NRC Scenario #4, rev. 0 Critical Tasks: Event #5 Isolate feedwater flow into and steam flow from the ruptured SG before a transition to

ECA-3.1 occurs. CT 18, Isolate the Ruptured SG. Event #6 Terminate the RCS depressurization by closing PORV block Valve prior to

transitioning to ECA-3-1. CT 10, Close PRZ PORV Block Valve.

References OTN-BG-00001 ADD 4, Operation of CVCS Letdown OTA-RK-00018, ADD 39B, Letdown Heat Exchanger Discharge Temperature High. OTO-AB-00001, Steam Dump Malfunction OTO-BB-00001, Steam Generator Tube Leak OTO-MA-00008, Rapid Load Reduction OTN-BG-00002, Attachment 8, Borate Mode of RMCS Operation E-0, Reactor Trip or Safety Injection E-3, Steam Generator Tube Rupture Tech Spec 3.7.4.A, Atmospheric Steam Dump Valves (ASDs) Tech Spec 3.4.13, RCS Operational Leakage ODP-ZZ-00025, EOP/OTO User's Guide



____________________________ _____________________________ ____________________________ _____________________________

Initial Conditions: 30% Power, Preparing to Shutdown. Turnover: ‘B’ CCP is tagged out due to oil leak.

Event No.


Event Description

1 N/A SRO (N) BOP (N) Place Mini-Purge in service

2 BBTE0421B SRO (I) RO (I) Tavg Channel Failure – Loop 2 Tc (Tech Spec)

3 BGLT0149 SRO (I) RO (I) VCT Level Transmitter Failure

4 BB002_A SRO (M) RO (M) BOP (M)

LOCA (Tech Spec)

5 PBG05A_1 SRO (C) RO (C) ‘A’ CCP failure to auto start

6

CPIS_A_BLOCK_AUTO CPIS_B_BLOCK_AUTO

SRO (C) BOP (C) Mini-Purge Failure to isolate



Actual Attributes


Scenario Event Description Callaway 2014-1 NRC Scenario #5, rev. 0

The plant is in a shutdown at 30% power. After the crew takes the watch, they will place Mini-Purge in service in IAW OTN-GT-00001, Containment Purge System, in preparation for Containment entry. After the Mini-Purge system has been placed in service Tcold Channel BB TE-421 B fails high. The crew will respond using OTO-BB-00004, RTD Channel Failure, and select an operable channel. The CRS will evaluate Tech Spec 3.3.1. After the CRS has identified the Tech Spec for the RTD channel failure, BG LT-149 will fail high. The crew will respond IAW OTO-BG-00004, VCT Level Channel Failures, and place the letdown divert valve in the VCT position. After the crew has taken the action for the VCT level channel failure, a leak in the RCS will develop. The crew will respond by entering OTO-BB-00003, RCS Excessive Leakage. The crew should determine that a reactor trip is required due to the leak rate. The CRS will evaluate Tech Spec 3.4.13.A. The crew will trip the reactor and enter E-0, Reactor Trip or Safety Injection. The crew will identify that the ‘A’ CCP failed to start and manually start the ‘A’ CCP. The crew will transition to E-1, Loss of Reactor or Secondary Coolant. The crew will secure all RCPs and identify that Mini-Purge failed to isolate. The crew will manually isolate Mini-Purge. The scenario can be terminated after the crew transitions to ES-1.2, Post LOCA Cooldown and Depressurization.

Callaway 2014-1 NRC Scenario #5, rev. 0 Critical Tasks: Event #4 Trip all RCPs within 5 minutes of meeting RCP trip criteria. CT- 16, Manually trip RCPs Event #5 Establish flow from ‘A’ CCP before completion of E-0 Attachment A. CT- 6, Establish flow from at least one Charging/S1 pump Event #6 Ensure that at least one CTMT Minipurge Damper is closed on each penetration –

Manually close GT-HZ-12 before completion of E-0, Attachment A. CT- 14, Close containment minipurge isolation valves

References OTN-GT-00001, Containment Purge System OTO-BB-00004, RTD Channel Failure OTO-BG-00004, VCT Level Channel Failures OTO-BB-00003, RCS Excessive Leakage E-0, Reactor Trip or Safety Injection E-1, Loss of Primary or Secondary Reactor coolant Tech Spec 3.3.1 for Reactor Trip System Instrumentation ODP-ZZ-00025, EOP/OTO User's Guide

es -401 pwr examination outline (ro) form es-401-2 · the generic (g) k/as in tiers 1 and 2 shall...

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