ML14174B409

2014-06 Final Outlines

ML14174B409
Person / Time
Site:	Comanche Peak
Issue date:	06/16/2014
From:	Vincent Gaddy Operations Branch IV
To:	Luminant Generation Co
laura hurley
References
Download: ML14174B409 (27)

Text

ES-401 PWR Examination Outline FORM ES-401-2 Facility Name: Date of Exam:

RO K/A Category Points SRO-Only Points Tier Group K K K K K K A A A A G Total A2 G* Total 1 2 3 4 5 6 1 2 3 4

• 1 3 2 4 3 3 3 18 4 2 6

1. Emergency Abnormal 2 1 1 2 N/A 2 2 N/A 1 9 4 0 4 Plant Evolutions Tier Totals 4 3 6 5 5 4 27 8 2 10 1 3 2 2 3 2 3 2 4 3 2 2 28 3 2 5 2.

2 1 1 1 2 1 1 1 0 0 1 1 10 0 2 1 3 Plant Systems Tier Totals 4 3 3 5 3 4 3 4 3 3 3 38 5 3 8 1 2 3 4 1 2 3 4

3. Generic Knowledge and Abilities 10 7 Categories 3 2 3 2 1 2 2 2 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outlines (i.e., 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).

Note: 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.

Note: 3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.

Note: 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.

Note: 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.

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

Note: 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.

Note: 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.

Note: 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, 21 of 33

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)

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

1 2 3 1 2 0

1 000007 Reactor Trip - Stabilization - Recovery / 1 Shutdown margin 3.4 1 2

0 2 000008 Pressurizer Vapor Space Accident / 3 PRT level pressure and temperature 3.8 1 8

1 3 000009 Small Break LOCA / 3 Actions to be taken if PTS limits are violated 3.8 1 4

0 4 000011 Large Break LOCA / 3 Consequences of managing LOCA with loss of CCW 3.7 1 3

000015 RCP Malfunctions / 4 02.

5 Knowledge of limiting conditions for operations and safety limits. 4.0 1 000017 RCP Malfunctions (Loss of RC Flow) / 4 22 000022 Loss of Rx Coolant Makeup / 2 0 0

6 000025 Loss of RHR System / 4 RHR heat exchangers 2.9 1 1

0 7 000026 Loss of Component Cooling Water / 8 Control of flow rates to components cooled by the CCWS 2.9 1 6

000027 Pressurizer Pressure Control System 0 8 Controllers and positioners 2.6 1 Malfunction / 3 3 0

9 000029 ATWS / 1 Reactor nucleonics and thermo-hydraulics behavior 2.8 1 1

000038 Steam Gen. Tube Rupture / 3 0 000040 Steam Line Rupture - Excessive Heat Transfer 0 10 Containment temperature and pressure considerations 3.4

/4 6 1

WE12 Uncontrolled Depressurization of all Steam Generators / 4 000054 (CE/E06) Loss of Main Feedwater / 4 0 0

11 000055 Station Blackout / 6 Actions contained in EOP for loss of offsite and onsite power 4.3 1 2

1 12 000056 Loss of Off-site Power / 6 Operational status of PZR backup heaters 3.4 1 7

000057 Loss of Vital AC Inst. Bus / 6 0 0

13 000058 Loss of DC Power / 6 Vital and battery bus components 3.1 1 3

0 The automatic actions (alignments) within the nuclear service 14 000062 Loss of Nuclear Svc Water / 4 water resulting from the actuation of the ESFAS 3.6 1 2

04.

15 000065 Loss of Instrument Air / 8 Knowledge of abnormal condition procedures. 4.0 1 11 W/E04 LOCA Outside Containment / 3 0

04. Knowledge of the operational implications of EOP warnings, 16 W/E11 Loss of Emergency Coolant Recirc. / 4 cautions, and notes.

3.8 1 20 BW/E04; W/E05 Inadequate Heat Transfer - Loss of 0 Annunciators and conditions indicating signals, and remedial 17 actions associated with the Loss of Secondary Heat Sink 3.9 1 Secondary Heat Sink / 4 3 000077 Generator Voltage and Electric 0 18 Reactor and turbine trip criteria 3.9 1 Grid Disturbances / 6 1 K/A Category Totals: 3 2 4 3 3 3 Group Point Total: 18 ES-401, 22 of 33

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)

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

1 2 3 1 2 Proper actions to be taken if automatic safety functions have not 19 000001 Continuous Rod Withdrawal / 1 03 taken place 4.5 1 20 000003 Dropped Control Rod / 1 03 Relationship of reactivity and reactor power to rod movement 3.5 1 000005 Inoperable/Stuck Control Rod / 1 0 21 000024 Emergency Boration / 1 02 Actions contained in EOP for emergency boration 4.2 1 000028 Pressurizer Level Malfunction / 2 0 000032 Loss of Source Range NI / 7 0 000033 Loss of Intermediate Range NI / 7 0 000036 Fuel Handling Accident / 8 0 22 000037 Steam Generator Tube Leak / 3 01 Maximum controlled depressurization rate for affected S/G 3.7 1 000051 Loss of Condenser Vacuum / 4 0 000059 Accidental Liquid RadWaste Rel. / 9 0 000060 Accidental Gaseous Radwaste Rel. / 9 0 000061 ARM System Alarms / 7 0 000067 Plant Fire On-site / 8 0 000068 Control Room Evac. / 8 0 23 000069 Loss of CTMT Integrity / 5 03 Personnel access hatch and emergency access hatch 2.8 1

W/E14 High Containment Pressure / 5 24 000074 Inad. Core Cooling / 4 06 RCPS 3.6 W/E06 Degraded Core Cooling / 4 1 W/E07 Saturated Core Cooling / 4 25 000076 High Reactor Coolant Activity / 9 06 Actions contained in EOP for high reactor coolant activity 3.2 1 W/E01 Rediagnosis / 3 1

Adherence to appropriate procedures and operation within the 26 W/E02 SI Termination / 3 02 limitations in the facility's license and amendments 3.5

04. Knowledge of local auxiliary operator tasks during an emergency 27 W/E13 Steam Generator Over-pressure / 4 and the resultant operational effects.

3.8 1 35 W/E15 Containment Flooding / 5 0 W/E16 High Containment Radiation / 9 0 W/E03 LOCA Cooldown - Depress. / 4 0 W/E09 Natural Circulation Operations / 4 0

W/E10 Natural Circulation with Steam Voide in Vessel with/without RVLIS. / 4 W/E08 RCS Overcooling - PTS / 4 0 K/A Category Totals: 1 1 2 2 2 1 Group Point Total: 9 ES-401, 23 of 33

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

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

1 2 3 4 5 6 1 2 3 4 Effects of RCP shutdown on secondary parameters, such 0 3.2; 28,49 003 Reactor Coolant Pump 04 as steam pressure, steam flow, and feed flow; RCP pump 2 2 and motor bearing temperatures 2.9 1

50 004 Chemical and Volume Control PZR pressure and temperature 3.5 1 5

1 30 005 Residual Heat Removal RWST 3.5 1 1

1 1 Interlocks between RHR valves and RCS; Inadvertent SIS 3.2; 55,31 006 Emergency Core Cooling actuation 2

6 3 3.9 0 Relationships between PZR level and changing levels of 32 007 Pressurizer Relief/Quench Tank the PRT and bleed holdup tank 2.5 1 9

0 0 RCS, in order to determine source(s) of RCS leakage into 52,33 008 Component Cooling Water the CCWS; CCW pump, including emergency backup 3.3; 3 2 4 2 0

34 010 Pressurizer Pressure Control Pressure detection systems 2.7 1 1

0 01. Loss of instrument power; Ability to explain and apply 3.6; 51,35 012 Reactor Protection system limits and precautions.

2 2 32 3.8 013 Engineered Safety Features 0 36 Fuel 4.4 1 Actuation 1 0

37 022 Containment Cooling Automatic containment isolation 3.6 1 3

025 Ice Condenser 0 Prevention of path for escape of radioactivity from 0 0 3.7; 29,38 026 Containment Spray containment to the outside (interlock on RWST isolation 2 9 6 after swapover); Containment spray pump cooling 2.7 0

39 039 Main and Reheat Steam Bases for RCS cooldown limits 2.7 1 5

0 40 059 Main Feedwater Feeding a dry S/G 2.9 1 4

0 41 061 Auxiliary/Emergency Feedwater Pumps 2.6 1 2

0 42 062 AC Electrical Distribution All breakers (including available switchyard) 3.3 1 1

0 43 063 DC Electrical Distribution Major DC loads 2.9 1 1

Fuel oil storage tanks; Knowledge of RO tasks performed 0 04. 3.2; 44,53 064 Emergency Diesel Generator outside the main control room during an emergency and 2 8 34 the resultant operational effects. 4.2 0

45 073 Process Radiation Monitoring Detector failure 2.7 1 2

0 46 076 Service Water ESF loads 3.7 1 7

0 0 2.7; 47,54 078 Instrument Air Service air; Air pressure 2 2 1 3.1 0

48 103 Containment Containment isolation 3.9 1 1

0 K/A Category Totals: 3 2 2 3 2 3 2 4 3 2 2 Group Point Total: 28 ES-401, Page 24 of 33

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

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

1 2 3 4 5 6 1 2 3 4 0

56 001 Control Rod Drive 1 One-line diagram of power supply to M/G sets 3.5 1 0

57 002 Reactor Coolant 5 Detection of RCS leakage 3.8 1 1

58 011 Pressurizer Level Control 2 Criteria and purpose of PZR level program 2.7 1 014 Rod Position Indication 0 0

59 015 Nuclear Instrumentation 2 Discriminator/compensation circuits 2.6 1 0

60 016 Non-nuclear Instrumentation 4 MFW System 2.6 1 017 In-core Temperature Monitor 0 027 Containment Iodine Removal 0 028 Hydrogen Recombiner and Purge 0

Control 029 Containment Purge 0

04. Knowledge of annunciator alarms, indications, or response 61 033 Spent Fuel Pool Cooling 4.2 1 31 procedures.

034 Fuel Handling Equipment 0 1

62 035 Steam Generator 2 RPS 3.7 1 0

63 041 Steam Dump/Turbine Bypass Control 2 Steam pressure 3.1 1 0

64 045 Main Turbine Generator 6 Turbine stop valves 2.8 1 055 Condenser Air Removal 0 056 Condensate 0 068 Liquid Radwaste 0 071 Waste Gas Disposal 0 072 Area Radiation Monitoring 0 075 Circulating Water 0 079 Station Air 0 0

65 086 Fire Protection 1 Adequate supply of water for FPS 3.1 1 K/A Category Totals: 1 1 1 2 1 1 1 0 0 1 1 Group Point Total: 10 ES-401, Page 25 of 33

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)

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

1 2 3 1 2 000007 Reactor Trip - Stabilization - Recovery / 1 0 000008 Pressurizer Vapor Space Accident / 3 0 0

76 000009 Small Break LOCA / 3 Possible leak paths 3.8 1 2

000011 Large Break LOCA / 3 0 000015 RCP Malfunctions / 4 0

000017 RCP Malfunctions (Loss of RC Flow) / 4 000022 Loss of Rx Coolant Makeup / 2 0 000025 Loss of RHR System / 4 0 000026 Loss of Component Cooling Water / 8 0 000027 Pressurizer Pressure Control System 0

Malfunction / 3 000029 ATWS / 1 0 0

78 000038 Steam Gen. Tube Rupture / 3 Existence of natural circulation, using plant parameters 4.2 1 9

Ability to evaluate plant performance and make operational 000040 Steam Line Rupture - Excessive Heat Transfer 01.

79 judgments based on operating characteristics, reactor behavior, 4.7

/4 07 and instrument interpretation. 1 WE12 Uncontrolled Depressurization of all Steam Generators / 4 0

80 000054 (CE/E06) Loss of Main Feedwater / 4 Conditions and reasons for AFW pump startup 4.2 1 3

000055 Station Blackout / 6 0 000056 Loss of Off-site Power / 6 0 02.

81 000057 Loss of Vital AC Inst. Bus / 6 Knowledge of limiting conditions for operations and safety limits. 4.7 1 22 000058 Loss of DC Power / 6 0 000062 Loss of Nuclear Svc Water / 4 0 000065 Loss of Instrument Air / 8 0 0 Facility conditions and selection of appropriate procedures 77 W/E04 LOCA Outside Containment / 3 during abnormal and emergency operations 4.3 1 1

W/E11 Loss of Emergency Coolant Recirc. / 4 0 BW/E04; W/E05 Inadequate Heat Transfer - Loss of 0

Secondary Heat Sink / 4 000077 Generator Voltage and Electric 0

Grid Disturbances / 6 K/A Category Totals: 0 0 0 0 4 2 Group Point Total: 6 ES-401, 22 of 33

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)

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

1 2 3 1 2 000001 Continuous Rod Withdrawal / 1 0 000003 Dropped Control Rod / 1 0 000005 Inoperable/Stuck Control Rod / 1 0 000024 Emergency Boration / 1 0 83 000028 Pressurizer Level Malfunction / 2 08 PZR level as a function of power level 3.5 1 000032 Loss of Source Range NI / 7 0 000033 Loss of Intermediate Range NI / 7 0 000036 Fuel Handling Accident / 8 0 000037 Steam Generator Tube Leak / 3 0 000051 Loss of Condenser Vacuum / 4 0 000059 Accidental Liquid RadWaste Rel. / 9 0 000060 Accidental Gaseous Radwaste Rel. / 9 0 000061 ARM System Alarms / 7 0 85 000067 Plant Fire On-site / 8 15 Requirements for establishing a fire watch 3.9 1 000068 Control Room Evac. / 8 0 000069 Loss of CTMT Integrity / 5 0

W/E14 High Containment Pressure / 5 000074 Inad. Core Cooling / 4 Adherence to appropriate procedures and operation within the 84 W/E06 Degraded Core Cooling / 4 02 4.1 1 limitations in the facility's license and amendments W/E07 Saturated Core Cooling / 4 000076 High Reactor Coolant Activity / 9 0 W/E01 Rediagnosis / 3 1

Facility conditions and selection of appropriate procedures 82 W/E02 SI Termination / 3 01 during abnormal and emergency operations 4.2 W/E13 Steam Generator Over-pressure / 4 0 W/E15 Containment Flooding / 5 0 W/E16 High Containment Radiation / 9 0 W/E03 LOCA Cooldown - Depress. / 4 0 W/E09 Natural Circulation Operations / 4 0

W/E10 Natural Circulation with Steam Voide in Vessel with/without RVLIS. / 4 W/E08 RCS Overcooling - PTS / 4 0 K/A Category Totals: 0 0 0 0 4 0 Group Point Total: 4 ES-401, 23 of 33

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

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

1 2 3 4 5 6 1 2 3 4 0

86 003 Reactor Coolant Pump Effects of VCT pressure on RCP seal leakoff flows 2.8 1 5

004 Chemical and Volume Control 0

02. Knowledge of the bases in Technical Specifications for 87 005 Residual Heat Removal limiting conditions for operations and safety limits.

4.2 1 25 006 Emergency Core Cooling 0 007 Pressurizer Relief/Quench Tank 0 Results of excessive exit temperature from the letdown 0

88 008 Component Cooling Water cooler, including the temperature effects on ion-exchange 2.8 1 9 resins 010 Pressurizer Pressure Control 0 012 Reactor Protection 0 013 Engineered Safety Features 0

Actuation 022 Containment Cooling 0 025 Ice Condenser 0 0

89 026 Containment Spray Failure of chemical addition tanks to inject 4.1 1 5

039 Main and Reheat Steam 0 059 Main Feedwater 0 04.

90 061 Auxiliary/Emergency Feedwater Knowledge of the specific bases for EOPs. 4.0 1 18 062 AC Electrical Distribution 0 063 DC Electrical Distribution 0 064 Emergency Diesel Generator 0 073 Process Radiation Monitoring 0 076 Service Water 0 078 Instrument Air 0 103 Containment 0 0

K/A Category Totals: 0 0 0 0 0 0 0 3 0 0 2 Group Point Total: 5 ES-401, Page 24 of 33

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

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

1 2 3 4 5 6 1 2 3 4 001 Control Rod Drive 0 002 Reactor Coolant 0 011 Pressurizer Level Control 0 014 Rod Position Indication 0 0

91 015 Nuclear Instrumentation 1 Power supply loss or erratic operation 3.9 1 016 Non-nuclear Instrumentation 0 017 In-core Temperature Monitor 0 027 Containment Iodine Removal 0 028 Hydrogen Recombiner and Purge 0

Control 029 Containment Purge 0 033 Spent Fuel Pool Cooling 0 034 Fuel Handling Equipment 0 0

92 035 Steam Generator 1 Faulted or ruptured S/Gs 4.6 1 041 Steam Dump/Turbine Bypass Control 0 045 Main Turbine Generator 0 055 Condenser Air Removal 0 056 Condensate 0 068 Liquid Radwaste 0 071 Waste Gas Disposal 0 072 Area Radiation Monitoring 0 075 Circulating Water 0 079 Station Air 0 04.

93 086 Fire Protection Knowledge of abnormal condition procedures. 4.2 1 11 K/A Category Totals: 0 0 0 0 0 0 0 2 0 0 1 Group Point Total: 3 ES-401, Page 25 of 33

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility Name: Date of Exam:

RO SRO-Only Category K/A # Topic Q# IR # IR #

Knowledge of industrial safety procedures (such as rotating equipment, electrical, high temperature, 66 2.1. 26 high pressure, caustic, chlorine, oxygen and hydrogen). 3.4 1 67 2.1. 03 Knowledge of shift or short-term relief turnover practices. 3.7 1 Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical 68 1. 2.1. 04 requirements, no-solo operation, maintenance of active license status, 10CFR55, etc. 3.3 1 Conduct of 94 Operations 2.1. 34 Knowledge of primary and secondary plant chemistry limits. 3.5 1 2.1.

2.1.

Subtotal 3 1 69 2.2. 39 Knowledge of less than or equal to one hour Technical Specification action statements for systems. 3.9 1 70 2.2. 37 Ability to determine operability and/or availability of safety related equipment. 3.6 1 95 2.2. 42 Ability to recognize system parameters that are entry-level conditions for Technical Specifications. 4.6 1 2.

Equipment 96 2.2. 20 Knowledge of the process for managing troubleshooting activities. 3.8 1 Control 2.2.

2.2.

Subtotal 2 2 Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable 71 2.3. 05 survey instruments, personnel monitoring equipment, etc. 2.9 1 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or 72 2.3. 14 emergency conditions or activities. 3.4 1 Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment 73 2.3. 12 entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, 3.2 1

3. etc.

Radiation 97 Control 2.3. 11 Ability to control radiation releases. 4.3 1 Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable 98 2.3. 15 survey instruments, personnel monitoring equipment, etc. 3.1 1 2.3.

Subtotal 3 2 74 2.4. 06 Knowledge of EOP mitigation strategies. 3.7 1 75 2.4. 32 Knowledge of operator response to loss of all annunciators. 3.6 1 Knowledge of facility protection requirements, including fire brigade and portable fire fighting equipment 99 4. 2.4. 26 usage. 3.6 1 Emergency Knowledge of EOP implementation hierarchy and coordination with other support procedures or 100 Procedures / 2.4. 16 guidelines such as, operating procedures, abnormal operating procedures, and severe accident 4.4 1 management guidelines.

Plan 2.4.

2.4.

Subtotal 2 2 Tier 3 Point Total 10 7 ES-401, Page 26 of 33

ES-401 Record of Rejected K/As Form ES-401-4 Tier / Randomly Reason for Rejection Group Selected K/A 1/1 011 EA2.03 Question 4 Could not write a question with regard to existence of natural circulation during a Large Break LOCA as there is no natural circulation in the RCS with a large break LOCA. Replaced K/A 011 EA2.09 with K/A 011 EA2.03.

1/2 003 AK1.03 Question 20 Could not write an operationally valid question for an Inoperable/Stuck Rod Xenon transient as CPNPP does not allow extended operation with an inoperable or stuck control rod. Replaced K/A 005 AK1.03 with 003 AK1.03.

1/2 W/E 13 G.2.4.35 Question 27 Could not write a question with regard to immediate operator actions without reference to procedures during a steam generator over-pressure event as there are no immediate operator actions which would be performed without procedure reference when addressing steam generator over-pressure.

Steam generator over-pressure is a YELLOW path FRG at CPNPP. Replaced K/A W/E 13 G.2.4.49 with W/E 13 G.2.4.35.

2/1 003 K5.04 Question 28 Could not write an operationally valid question on the operational implications of the effects of RCP shutdown on TAVE as CPNPP trips the Reactor at all times in Modes 1 and 2 when an RCP trips, thus the effects on TAVE are only evaluated with respect to a reactor trip where temperature control is what is monitored versus loop variations. Replaced K/A 003 K5.03 with 003 K5.04.

2/1 004 K4.09 Question 29 Could not develop operationally valid distracters for the effect of a loss or malfunction on the spray/heater combination in the pressurizer to assure uniform boron concentration. Replaced K/A 004 K6.01 with 026 K4.09.

ES-401 Record of Rejected K/As Form ES-401-4 Tier / Randomly Reason for Rejection Group Selected K/A 2/1 007 A4.09 Question 32 Identified that the Recognition of a Leaking PORV or Safety Valve via PRT indications was an overlap to Question 2 which was PRZR Vapor Space Accident monitoring of the PRT pressure and temperature. Replaced K/A 007 A4.10 with 007 A4.09.

2/1 061 K6.02 Question 41 Questions 41 and 52 both sampled system 061 in the K1 category. Replaced K/A 061 K1.02 with K/A 061 K6.02 to preventing over sampling of the K1 category for auxiliary/emergency feedwater. K6 was selected to provide balance in outline.

2/2 033 G.2.4.31 Question 61 Could not write a question with regard to immediate operator actions performed without reference to procedures for the Spent Fuel Pool Cooling System as there are no immediate operator actions which would be performed without procedure reference when addressing Spent Fuel Pool Cooling system malfunctions. This is addressed in an AOP at CPNPP.

Replaced K/A 033 G.2.4.49 with 033 G.2.4.31.

2/2 086 K4.01 Question 65 Could not write a question as the loss or malfunction of the Fire Protection system would not affect, fire, smoke and heat detectors. Replaced K/A 086 K6.04 with 086 K4.01.

1/1 W/E 04 EA2.01 Question 77 Could not write an operationally valid question on calculation of expected values of flow in the RCS loop with RCP secured.

CPNPP trips the Reactor at all times in Modes 1 and 2 when an RCP trips, thus the resulting loop flow is not considered procedurally as forced flow exist in the remaining three loops.

Replaced K/A 015 AA2.07 with W/E 04 EA2.01.

ES-401 Record of Rejected K/As Form ES-401-4 Tier / Randomly Reason for Rejection Group Selected K/A 1/1 054 AA2.03 Question 80 Could not write an SRO only question with regard to immediate operator actions during a loss of main feedwater event. The immediate operator actions in the AOP for a Main Feedwater pump trip are actions the ROs perform from memory at CPNPP so this is not an SRO only topic for CPNPP. Additionally, other areas which could make the question an SRO level are lacking in loss of main feedwater as nearly the entire system is non safety-related and procedure selection is at the RO knowledge level. Replaced K/A 054 G.2.4.49 with K/A 054 AA2.03.

1/1 057 G.2.2.22 Question 81 Could not write a question as system limits and precautions do not apply to Station Blackout. Replaced K/A 055 G.2.1.32 with 057 G.2.2.22.

1/2 005 AA2.03 Question 82 Generic K/A 2.1.43 is not one of the Tier 1 and 2 generic K/As as listed in NUREG 1021 ES-401. Replaced K/A 003 G.2.1.43 with 005 AA2.03.

1/2 074 EA2.02 Question 84 Could not write an SRO level question on loss of condenser vacuum as this is a non-safety system without Technical Specifications. Replaced K/A 051 G.2.4.46 with 074 EA2.02.

2/1 061 G.2.4.18 Question 90 Generic K/A 2.4.23 is not one of the Tier 1 and 2 generic K/As as listed in NUREG 1021 ES-401. Replaced K/A 061 G.2.4.23 with 061 G.2.4.18.

2/2 035 A2.01 Question 92 Could not write a Technical Specification question on Spent Fuel Pool Cooling System as there are no Technical Specifications for this system. A similar topic was sampled on Question 61. Replaced K/A 033 G.2.2.25 with K/A 035 A2.01.

ES-401 Record of Rejected K/As Form ES-401-4 Tier / Randomly Reason for Rejection Group Selected K/A 2/2 086 G.2.4.11 Question 93 Generic K/A 2.4.25 is not one of the Tier 1 and 2 generic K/As as listed in NUREG 1021 ES-401. Replaced K/A 086 G.2.4.25 with 086 G.2.4.11.

3 G.2.1.34 Question 94 Generic K/A 2.1.15 has no 10CFR43 tie in NUREG 1122.

Replaced K/A G.2.1.15 with K/A G.2.1.34 1/1 077 AK3.01 Question 18 Could not develop an operationally valid question including distracters for the effect of generator voltage and electric grid disturbances on the turbine/generator control. Replaced K/A 077 AK2.07 with 077 AK3.01.

2/2 015 K6.02 Question 59 Could not develop an operationally valid question on disagreement between channels without discussing annunciators or alarms. Replaced K/A 015 A3.04 with 015 K6.02.

3 G.2.3.15 Question 98 Could not develop an SRO level question on knowledge of radiation exposure limits. Replaced K/A G.2.3.04 with G.2.3.15.

2/1 010 K6.01 Question 34 Could not develop an operationally valid question on the ability to monitor the Pressurizer Pressure Control System that did not conflict with the operational exam. Replaced K/A 010 A3.02 with 010 K6.01.

2/1 008 K1.04 Question 52 Auxiliary Feedwater System was over sampled. Replaced K/A 061 K1.04 with 008 K1.04.

2/2 016 K3.04 Question 60 Auxiliary Feedwater System was over sampled. Replaced K/A 016 K3.06 with 016 K3.04.

Administrative Topics Outline Facility: CPNPP Units 1 and 2 Date of Examination: 06/09/14 Examination Level RO Operating Test Number: NRC Administrative Topic Type Code* Describe Activity to be Performed (see Note) 2.1.25 Ability to interpret reference materials such as graphs, curves, tables, etc. (3.9).

Conduct of Operations N, R (RA1)

JPM: Determine Minimum and Maximum RHR Flow Rate (RO1404).

2.1.23 Ability to perform specific system and integrated plant procedures during all modes Conduct of Operations of plant operation. (4.3)

M, R (RA2)

JPM: Perform a Shutdown Margin Calculation (RO1010D).

2.2.12 Knowledge of surveillance procedures (3.7).

Equipment Control M, R (RA3) JPM: Perform a Manual Quadrant Power Tilt Ratio Calculation (RO1803D).

2.3.12 Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling Radiation Control responsibilities, access to locked high M, R radiation areas, aligning filters, etc. (3.2).

(RA4)

JPM: Determine Radiation Doses during System Alignment (RWT029D).

Emergency Plan 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; for 4 for SROs & RO retakes)

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

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

CPNPP NRC 2014 ES-301-1 RO Admin JPM Outline.docx

Administrative Topics Outline Task Summary RA1 The applicant will determine minimum and maximum Residual Heat Removal (RHR) System flow per ABN-104, Residual Heat Removal System Malfunction, Attachment 3, Minimum RHR Flow Guideline for Decay Heat Removal, Attachment 4, RHR Maximum Flow Limit, and Attachment 16, Actual Versus Indicated Reactor Vessel Level. Critical steps include calculating the minimum RHR flow rate given time after shutdown and maximum RHR flow rate given recalculated RCS level. This is a new JPM.

RA2 The applicant will perform a Shutdown Margin Calculation per OPT-301, Shutdown Margin Calculation, Attachment 10.1.1, Manual Generation of OPT-301-9, Shutdown Margin. Critical tasks include identifying individual parameters associated with the calculation and determining the Shutdown Margin. This is a modified bank JPM.

RA3 The applicant will perform a manual Quadrant Power Tilt Ratio calculation per OPT-302, Calculating Power Tilt Ratio, and determine whether Acceptance Criteria are met. The critical steps include recording data, accurately performing calculations and applying Acceptance Criteria. This is a modified bank JPM.

RA4 The applicant will determine radiation doses during a system alignment per STA-657, ALARA Job Planning/Debriefing. The critical steps include calculating the dose received in a radiation field with and without protective shielding. This is a modified bank JPM.

CPNPP NRC 2014 ES-301-1 RO Admin JPM Outline.docx

Administrative Topics Outline Facility: CPNPP Units 1 and 2 Date of Examination: 06/09/14 Examination Level SRO Operating Test Number: NRC Administrative Topic Type Code* Describe Activity to be Performed (see Note) 2.1.1 Knowledge of conduct of operations requirements. (4.2)

Conduct of Operations N, R (SA1)

JPM: Determine Technical Specification and Event Reportability (SO1005D).

2.1.23 Ability to perform specific system and integrated plant procedures during all modes Conduct of Operations of plant operation. (4.4)

M, R (SA2)

JPM: Perform a Shutdown Margin Calculation (SO1002A).

2.2.12 Knowledge of surveillance procedures (4.1).

Equipment Control M, R JPM: Perform a Manual Quadrant Power Tilt Ratio (SA3)

Calculation and Evaluate Technical Specifications (SO1202C).

2.3.6 Ability to approve release permits. (3.8)

Radiation Control D, R (SA4) JPM: Review a Gaseous Waste Release Permit (SO1039C).

2.4.41 Knowledge of the emergency action level Emergency Plan thresholds and classifications. (4.6)

N, R (SA5) JPM: Classify an Emergency Plan Event (SO1136G).

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; for 4 for SROs & RO retakes)

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

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

CPNPP NRC 2014 ES-301-1 SRO Admin JPM Outline.docx

Administrative Topics Outline Task Summary SA1 The applicant will identify impacted Technical Specification Limiting Conditions for Operations and determine Event Reportability per STA-501, Non-Routine Reporting for the Station Electrical System. The critical steps include identifying the Technical Specification Limiting Condition for Operation, Required Action, Completion Time and determining the Oral and Written Reporting Requirements.

This is a new JPM.

SA2 The applicant will perform a Shutdown Margin Calculation per OPT-301, Shutdown Margin Calculation, Attachment 10.1.1, Manual Generation of OPT-301-9, Shutdown Margin. Critical tasks include reviewing individual parameters, verifying adequate Shutdown Margin, and identifying any required action when the Acceptance Criteria is not met. This is a modified bank JPM.

SA3 The applicant will perform a manual Quadrant Power Tilt Ratio calculation per OPT-302, Calculating Power Tilt Ratio, and determine whether Acceptance Criteria are met. The critical steps include recording data, accurately performing calculations, applying Acceptance Criteria, and identifying any impacted Technical Specification CONDITION, REQUIRED ACTION, and COMPLETION TIME. This is a modified bank JPM.

SA4 The applicant will review a Gaseous Waste Release Permit per STA-603, Control of Station Radioactive Effluents. The critical steps include identifying any errors and actions required prior to approving the release. This is a bank JPM.

SA5 The applicant will classify an Emergency Plan event per EPP-201, Assessment of Emergency Action Levels, Emergency Classification, and Plan Activation.

Critical steps include determining the Event Category and Event Classification using the Hot, Common, and Cold Emergency Action Level Classification Charts.

This is a new JPM.

CPNPP NRC 2014 ES-301-1 SRO Admin JPM Outline.docx

ES-301 Control Room / In-Plant Systems Outline Form ES-301-2 Facility: CPNPP Units 1 and 2 Date of Examination: 06/09/14 Exam Level: RO SRO(I) SRO (U) Operating Test No.: NRC 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 S-1 001 - Control Rod Drive System (RO1026A) (RO Only) A, L, M, S 1 Respond to Reactor Startup Continuous Rod Insertion S-2 004 - Chemical and Volume Control System (RO1309A) N, S 2 Lower Letdown Flow and Adjust Charging Flow S-3 010 - Pressurizer Pressure Control System (RO1222) A, D, S 3 Respond to a Pressurizer Spray Valve Failure S-4 005 - Residual Heat Removal System (RO1507A) A, D, EN, L, S 4P Transfer RHR & SI Pumps to Hot Leg Recirculation S-5 061 - Auxiliary Feedwater System (RO3516A) A, EN, N, S 4S Respond to Inadvertent Start of TDAFW Pump S-6 064 - Emergency Diesel Generator System (RO4215B) A, D, S 6 Restore Safeguards Bus 1EA1 to Offsite Power S-7 016 - Non-Nuclear Instrumentation System (RO1833) D, S 7 Respond to Feedwater Flow Instrument Failure S-8 086 - Fire Protection System (RO4406C) M, S 8 Respond to a Fire in the Control Room In-Plant Systems (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U)

P-1 045 - Main Turbine Generator System (RO4217D) D, E 4S Align a Main Generator Vent and Purge P-2 103 - Containment System (RO2119A) D, E, R 5 Perform Containment Phase A Local Isolation P-3 071 - Waste Gas Disposal System (RO4006) N, R 9 Terminate Release of Radioactive Gas Page 1 of 3 CPNPP NRC 2014 ES-301-2 RO & SRO JPM Outline.docx

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

@ 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 NRC JPM Examination Summary Description S-1 The applicant will perform actions for a continuous rod insertion during a Reactor Startup.

The Control Rods will initially be withdrawn in accordance with IPO-002A, Plant Startup from Hot Standby, Step 5.4.1. The alternate path requires operator action per ABN-712, Rod Control System Malfunction, Section 2.0, Abnormal Control Rod Response in MODE 1 or 2 and/or EOP-0.0A, Reactor Trip or Safety Injection, Step 1. This is a modified bank JPM under Control Rod Drive System - Reactivity Control Safety Function. This is a PRA significant action. (K/A 001.A2.11 - IR 4.4 / 4.7)

S-2 The applicant will lower Letdown flow and adjust Charging flow in accordance with SOP-103A, Chemical and Volume Control System, Sections 5.2.4, Lowering/Securing Letdown Flow and 5.2.2, Raising/Lowering Charging with a CCP in Operation. This is a new JPM under the Chemical and Volume Control System - Reactor Coolant System Inventory Control Safety Function. (K/A 004.A4.19 - IR 3.1 / 2.8)

S-3 The applicant will perform the actions for a Pressurizer Spray Valve failing open in accordance with ABN-705, Pressurizer Pressure Malfunction. The alternate path includes actions to trip the Reactor and Reactor Coolant Pump when the Pressurizer Spray Valve fails to close. This is a bank JPM under the Pressurizer Pressure Control System -

Reactor Pressure Control Safety Function. This is a PRA significant action.

(K/A 010.A2.02 - IR 3.9 / 3.9)

S-4 The applicant will transfer Residual Heat Removal Pumps and Safety Injection Pumps to Hot Leg Recirculation per EOS-1.4A, Transfer to Hot Leg Recirculation. The alternate path occurs when the Safety Injection to Hot Legs 2 & 3 Injection Isolation Valve fails to open at Step 2d. This is a bank JPM under the Residual Heat Removal System - Primary System Heat Removal from Reactor Core Safety Function. This is a PRA significant action. (K/A 011.EA1.11 - IR 4.2 / 4.2)

Page 2 of 3 CPNPP NRC 2014 ES-301-2 RO & SRO JPM Outline.docx

ES-301 Control Room / In-Plant Systems Outline Form ES-301-2 S-5 The applicant will respond to an inadvertent opening of the steam admission valve to the Turbine Driven Auxiliary Feedwater Pump in accordance with ALM-0082A, 1-ALB-8B, Window 4.5 - TD AFWP STM SPLY VLV LEAKING HV-2452-1/2 and then by ABN-305, Auxiliary Feedwater Malfunction, Section 6.0, Inadvertent Turbine Driven AFW Pump Start (Steam Supply Valve Fails Open). The alternate path occurs when the steam supply valve fails to close. This is a new JPM under the Main Feedwater System - Secondary System Heat Removal from Reactor Core Safety Function. This is a PRA significant action. (K/A 061.A2.07 - IR 3.4 / 3.5)

S-6 The applicant will restore Offsite Power to 6.9 kV Safeguards Bus 1EA1 from Transformer XST1 in accordance with SOP-609A, Diesel Generator System, Section 5.7, Transferring From DG Supplying Alone to Normal or Alternate Supply. The alternate path occurs when a lowering frequency requires separating the Emergency Diesel Generator from the grid. This is a bank JPM under the Emergency Diesel Generator System -

Electrical Safety Function. (K/A 064.A4.07 - IR 3.4 / 3.4)

S-7 The applicant will respond to a Feed Flow Instrument failure in accordance with ALM-0081A, 1-ALB-8A, Window 2.8 - SG 2 STM FLO & FW FLO MISMATCH and ABN-708, Feed Flow Instrument Malfunction. This is a bank JPM under the Non-Nuclear Instrumentation System - Instrumentation Safety Function. (K/A 059.A2.11 - IR 3.0 / 3.3)

S-8 The applicant will perform the actions for a Control Room fire in accordance with ABN-803A, Respond to a Fire in the Control Room or Cable Spreading Room, Attachment 1, Reactor Operator Actions to Achieve Hot Shutdown. This is a modified bank JPM under the Fire Protection System - Plant Service Systems Safety Function.

(K/A 068.AK3.12 - IR 4.1 / 4.5)

P-1 The applicant will align a Main Generator Vent and Purge per ABN-601, Response to a 138/345 KV System Malfunction, Attachment 11, Main Generator Hydrogen Vent and Argon Purge. This is a bank JPM under the Main Turbine Generator System - Secondary System Heat Removal from Reactor Core Safety Function. This is a PRA significant action. (K/A G 2.4.35 - IR 3.8 / 4.0)

P-2 The applicant will perform local actions to isolate Containment Phase A Isolation Valves in accordance with EOP-0.0A, Reactor Trip or Safety Injection, Attachment 4, Phase A Isolation. This is a bank JPM under the Containment System - Containment Integrity Safety Function. (K/A 069.AA1.01 - IR 3.5 / 3.7)

P-3 The applicant will perform actions to terminate a radioactive gas release in accordance with RWS-201, Gaseous Waste Processing System, Section 5.4.5, Gas Decay Tank X-01 Discharge to the Ventilation System, Step 5.4.5.N. This is a new JPM under the Waste Gas Disposal System - Radioactivity Release Safety Function.

(K/A 060.AA2.06 - IR 3.6 / 3.8)

Page 3 of 3 CPNPP NRC 2014 ES-301-2 RO & SRO JPM Outline.docx

Appendix D Scenario Outline Form ES-D-1 Facility: CPNPP 1 & 2 Scenario No.: 1 Op Test No.: June 2014 NRC Examiners: Operators:

Initial Conditions: 100% power MOL - RCS Boron is 924 ppm (by sample).

Turnover: Maintain steady-state power conditions. Tornado Warning from the National Weather Service.

Pressurizer Steam Space Sample is in progress by Chemistry.

Critical Tasks: Emergency Stop Train B Diesel Generator within 15 minutes of Breaker Failure in accordance with ECA-0.0A, Loss of All AC Power. (Event 6)

Isolate Reactor Coolant System Leakage Paths in accordance with ECA-0.0A, Loss of All AC Power prior to initiating the Steam Generator depressurization. (Event 7)

Initiate an Operator Induced Cooldown in accordance with ECA-0.0A, Loss of All AC Power.

(Event 8)

Event No. Malf. No. Event Type* Event Description 1 RP05D I (RO, SRO) Reactor Coolant System Loop (1-04) Narrow Range Cold Leg

+10 min TS (SRO) Temperature Instrument (TI-441A) Fails High.

2 RP03J I (BOP, SRO) Main Steam Line (1-04) Pressure Transmitter (PT-544) Channel I

+20 min TS (SRO) Fails Low.

3 RX05A I (RO, SRO) Pressurizer Level Channel (LT-459A) Fails Low.

+30 min TS (SRO) 4 FW14B R (RO) Heater Drain Pump (1-02) Trip.

+45 min TC09I N (BOP, SRO) Automatic Turbine Runback Failure.

TS (SRO) 5 ED01 M (RO, BOP, SRO) Loss of All AC Power Due to Loss of Offsite Power.

+50 min 6 EG06A C (BOP) Emergency Diesel Generator (1-01) Air Start Failure.

+50 min EG16B Emergency Diesel Generator (1-02) Output Breaker Failure.

7 WDR04 C (RO) Pressurizer Steam Space Sample Valves (1/1-4165A & 1/1-4176A)

+55 min Fail to Auto Close. Manual Closure of 1-HV-4165A Required.

8 MS13A I (BOP) Atmospheric Relief Valve (1-01) Fails Closed due to Steam

+75 min Pressure Instrument (PT-2325) Failure.

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications Actual Target Quantitative Attributes 9 Total malfunctions (5-8) 4 Malfunctions after EOP entry (1-2) 4 Abnormal events (2-4) 1 Major transients (1-2) 1 EOPs entered/requiring substantive actions (1-2) 1 EOP contingencies requiring substantive actions (0-2) 3 Critical tasks (2-3)

CPNPP NRC 2014 ES-D-1 Simulator Scenario Outlines

Appendix D Scenario Outline Form ES-D-1 SCENARIO

SUMMARY

NRC 1 The crew will assume the watch at 100% power with no scheduled activities per IPO-003A, Power Operations. A Tornado Warning from the National Weather Service is in progress. ABN-907, Acts of Nature, Section 5.0, Severe Weather, has been performed and includes placing the High Flux at Shutdown switches in BLOCK and realigning Control Room Ventilation to the Emergency Recirculation Mode. A Pressurizer Steam Space sample is in progress.

In the first event a Reactor Coolant System (RCS) Loop 4 TCOLD Instrument will fail high. The crew enters ABN-704, TC/N-16 Instrumentation Malfunction, Section 2.0, places Rod Control in MANUAL and defeats the affected channel. The SRO will refer to Technical Specifications.

The next event is a Main Steam Line 1-04 Channel I Pressure Transmitter failing low. Operator actions are per ABN-709, Steam Line Pressure Instrument Malfunction, Section 2.0, and require taking MANUAL control of the Feedwater Flow Control Valve, transferring to an Alternate Channel, and restoring Feedwater Flow Control to AUTO. The SRO will refer to Technical Specifications.

When plant conditions are stable a low failure of Pressurizer Level Channel, LT-459A, will occur.

Operator actions are per ABN-706, Pressurizer Level Instrument Malfunction, Section 2.0. The crew must manually control either the Charging Flow Controller or the Pressurizer Level Controller, transfer to an Alternate Channel, and restore Reactor Coolant System (RCS) Letdown flow. The SRO will refer to Technical Specifications.

The next event is a trip of a Heater Drain Pump with an automatic Turbine Runback failure. The crew responds per ABN-302, Feedwater, Condensate, Heater Drain System Malfunction, Section 4.0. When it is determined that automatic plant response has not activated, Control Rods are placed in AUTO and a manual Turbine Runback will be initiated. The crew will stabilize load at 700 MWe. During this event, Control Rod position may drop below the Rod Insertion Limit (RIL) and when informed, the SRO will refer to Technical Specifications. Additionally, ABN-401, Main Turbine Malfunction, Section 8.0, must be entered to reset the Turbine Runback.

The major event is a Loss of Offsite Power with an air start failure of Train A Diesel Generator and a breaker failure of Train B Diesel Generator resulting in a Total Loss of All AC Power. The crew enters either EOP-0.0A, Reactor Trip or Safety Injection and then exits to ECA-0.0A, Loss of All AC Power, or enters ECA-0.0A directly. While in ECA-0.0A, Reactor Coolant System leakage paths are isolated and a cooldown is initiated to 270°F. When the cooldown is commenced the crew will restore power to the Train B Safeguards Bus.

The event is complicated by failure of a Pressurizer Steam Space Sample Valve to automatically close and an Atmospheric Relief Valve that must be manually opened in order to facilitate Reactor Coolant System cooldown. This scenario is terminated when power is restored to the Train B Safeguards Bus and transition to ECA-0.0A, Step 26, Stabilize Steam Generator Pressures is performed.

Risk Significance:

Failure of risk important system prior to trip: Turbine Runback Failure Risk significant core damage sequence: Loss of All AC Power Risk significant operator actions: Stop Train B Diesel Generator Isolate RCS Leakage Paths Initiate RCS Cooldown Restore Safeguards Bus Power CPNPP NRC 2014 ES-D-1 Simulator Scenario Outlines

Appendix D Scenario Outline Form ES-D-1 Facility: CPNPP 1 & 2 Scenario No.: 2 Op Test No.: June 2014 NRC Examiners: Operators:

Initial Conditions: 100% power MOL - RCS Boron is 924 ppm (by sample).

Turnover: Maintain steady-state power conditions.

Critical Tasks: Trip Reactor Coolant Pumps within 10 minutes upon a Loss of Subcooling per EOP-0.0A, Reactor Trip or Safety Injection or EOP-1.0A, Loss of Reactor or Secondary Coolant, Foldout Pages. (Event 5)

Initiate Cooldown of Reactor Coolant System Prior to Exiting EOS-1.2A, Post LOCA Cooldown and Depressurization. (Event 6)

Manually Initiate Train A and Train B Safety Injection Signal Prior to Exiting EOP-0.0A, Reactor Trip or Safety Injection. (Event 7)

Event No. Malf. No. Event Type* Event Description 1 RX12 C (BOP, SRO) Main Steam Header Pressure Transmitter (PT-507) Fails Low.

+5 min 2 RX08A I (RO, SRO) Pressurizer Pressure Channel (PT-455) Fails High.

+15 min TS (SRO) 3 RX04C I (BOP, SRO) Steam Generator (1-03) Level Transmitter (LT-553) Fails High.

+25 min TS (SRO) 4 AFP N (RO, SRO) Fire in Auxiliary Building Fire Area AC.

+40 min 13_89 TS (SRO) Centrifugal Charging Pump (1-02) Manual Start Required.

5 RX16A C (RO) Power Operated Relief Valves (PCV-455A/456) Fail Open.

+45 min RX16B Reactor Trip Required.

6 RCR23 M (RO, BOP, SRO) Power Operated Relief Valve Block Valve (1/1-8000A) Fails Open

+50 min upon Breaker Trip.

7 RP07A C (RO) Train A Safety Injection Fails to Automatically Actuate.

+55 min RP07B Train B Safety Injection Fails to Automatically Actuate.

8 RH01D C (BOP) Residual Heat Removal Pump (1-02) Safety Injection Sequencer

+55 min Start Failure.

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications Actual Target Quantitative Attributes 8 Total malfunctions (5-8) 2 Malfunctions after EOP entry (1-2) 4 Abnormal events (2-4) 1 Major transients (1-2) 2 EOPs entered/requiring substantive actions (1-2) 0 EOP contingencies requiring substantive actions (0-2) 3 Critical tasks (2-3)

CPNPP NRC 2014 ES-D-1 Simulator Scenario Outlines

Appendix D Scenario Outline Form ES-D-1 SCENARIO

SUMMARY

NRC 2 The crew will assume the watch at 100% power with no scheduled activities per IPO-003A, Power Operations.

The first event is a low failure of Main Steam Header Pressure Transmitter (PT-507). The crew enters ABN-709, Steam Header Pressure Instrument Malfunction, Section 3.0, and places the Main Feedwater Pump Master Speed Control in MANUAL and restore required Steam Generator levels. The controller will remain in MANUAL and may require adjustment when the Steam Generator Level Transmitter fails later in the scenario.

The next event is a high failure of a Pressurizer Pressure Channel. Operator actions are per ABN-705, Pressurizer Pressure Malfunction, Section 2.0, and require closing the Power Operated Relief Valve (PORV) and its associated Block Valve, placing the Pressurizer Master Pressure Controller in MANUAL, selecting an alternate controlling Channel, and restoring Pressurizer pressure to normal. The SRO will refer to Technical Specifications.

When plant conditions are stable a high failure of a Steam Generator (1-03) Level Transmitter will occur. Crew actions are per ABN-710, Steam Generator Level Instrumentation Malfunction, Section 2.0, and include placing Steam Generator (SG) Level Control in MANUAL, stabilizing the plant, aligning an Alternate Channel, and transferring SG Level Control back to AUTO. The SRO will refer to Technical Specifications.

When plant parameters are restored to normal, a fire alarm in Auxiliary Building Fire Area AC will be initiated. The crew enters ABN-805A, Response to Fire in the Auxiliary Building or the Fuel Building, Section 3.0. Actions include placing the Train B Centrifugal Charging Pump in service. The fire will continue to spread in the Auxiliary Building resulting in the inadvertent opening of both Power Operated Relief Valves (PORVs) as addressed in ABN-805A, Section 7.0, and will require a manual Reactor Trip.

The crew will enter and perform actions of EOP-0.0A, Reactor Trip or Safety Injection. The Power Operated Relief Valves will fail to close and one PORV Block Valve breaker will trip before the valve closes resulting in a Small Break Loss of Coolant Accident. The crew will then transition from EOP-0.0A to EOP-1.0A, Loss of Reactor or Secondary Coolant, in preparation for an eventual cooldown and depressurization.

The scenario includes a failure of both trains of Safety Injection to automatically actuate along with a Train B Residual Heat Removal Pump that fails to start upon initiation of the Safety Injection Sequencer.

This scenario is terminated when a cooldown is commenced via the Steam Dump Valves in EOS-1.2A, Post LOCA Cooldown and Depressurization.

Risk Significance:

Failure of risk important system prior to trip: Train A Centrifugal Charging Pump Risk significant core damage sequence: Small Break LOCA Risk significant operator actions: Manually Initiate Safety Injection Trip Reactor Coolant Pumps Manually Start Train B RHR Pump Initiate RCS Cooldown CPNPP NRC 2014 ES-D-1 Simulator Scenario Outlines

Appendix D Scenario Outline Form ES-D-1 Facility: CPNPP 1 & 2 Scenario No.: 3 Op Test No.: June 2014 NRC Examiners: Operators:

Initial Conditions: 100% power MOL - RCS Boron is 924 ppm (by sample).

Turnover: Maintain steady-state power conditions.

Critical Tasks: Manually Trip Reactor Due to Reactor Protection System Failure Prior to Exiting EOP-0.0A, Reactor Trip or Safety Injection. (Event 5)

Initiate Train A and/or Train B Containment Isolation Phase A due to Failure to Automatically Actuate Prior to Exiting EOP-0.0A, Reactor Trip or Safety Injection. (Event 8)

Identify and Isolate Ruptured Steam Generator Prior to Commencing an Operator Induced Cooldown per EOP-3.0A, Steam Generator Tube Rupture. (Event 6)

Initiate Cooldown of Reactor Coolant System Prior to Exiting EOP-3.0A, Steam Generator Tube Rupture. (Event 6)

Event No. Malf. No. Event Type* Event Description 1 SG01A C (SRO) Steam Generator (1-01) Tube Leak at 50 gpd (0.0347 gpm).

+10 min 2 RX09A I (RO, BOP, SRO) Main Turbine 1st Stage Pressure Transmitter (PT-505) Fails High.

+20 min TS (SRO) 3 SW01B C (BOP, SRO) Station Service Water Pump 1-02 Trip.

+30 min TS (SRO) 4 FW16 R (RO) Lowering Condenser Vacuum Requiring Power Reduction Followed

+50 min C (BOP, SRO) by Total Loss of Condenser Vacuum.

5 RP01 I (BOP) Automatic Reactor Trip Failure.

+55 min RP13A Manual Reactor Trip Failure from CB-07.

6 SG01A M (RO, BOP, SRO) Steam Generator (1-01) Tube Rupture at 500 gpm (600 second

+65 min ramp) upon Turbine Trip.

7 OVRDE C (RO/BOP) Steam Generator (1-01) Blowdown Isolation Valve (1-HS-2397)

+70 min Fails to Close on Safety Injection.

8 RP09A C (BOP) Containment Isolation Phase A Train A and Train B Auto Actuation

+70 min RP09B Failure.

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications Actual Target Quantitative Attributes 8 Total malfunctions (5-8) 3 Malfunctions after EOP entry (1-2) 4 Abnormal events (2-4) 1 Major transients (1-2) 1 EOPs entered/requiring substantive actions (1-2) 0 EOP contingencies requiring substantive actions (0-2) 4 Critical tasks (2-3)

CPNPP NRC 2014 ES-D-1 Simulator Scenario Outlines

Appendix D Scenario Outline Form ES-D-1 SCENARIO

SUMMARY

NRC 3 The crew will assume the watch at 100% power with no scheduled activities per IPO-003A, Power Operations.

The first event is a Steam Generator #1 tube leak of 50 gpd. Crew actions are per ABN-106, High Secondary Activity, Section 2.0, and include verification of leakage rate to ensure proper procedural action. The SRO will refer to Technical Specifications.

The next event is a Main Turbine 1st Stage Pressure Transmitter failure. The crew responds per ABN-709, Steam Line Pressure, Steam Header Pressure, Turbine 1st-Stage Pressure and Feed Header Pressure Instrument Malfunction, Section 4.0. Several actions are required on the part of the RO and BOP to stabilize plant conditions. The SRO will refer to Technical Specifications.

When conditions are stable, Station Service Water Pump 1-02 will trip. The crew will enter ABN-501, Station Service Water System Malfunction, Section 2.0. Initial operator actions include placing the Train B Emergency Diesel Generator in PULLOUT. The SRO will refer to Technical Specifications.

When Pressurizer conditions are stable, a loss of Condenser vacuum will occur. The crew will respond per ABN-304, Main Condenser and Circulating Water System Malfunction, Section 3.0. Actions include lowering of Main Turbine load in 50 MWe increments in an attempt to maintain Condenser vacuum.

Once a load reduction is performed, Condenser vacuum will continue to deteriorate until a Reactor Trip is required.

Once it is determined that Condenser vacuum cannot be maintained, the Reactor must be manually tripped and EOP-0.0A, Reactor Trip or Safety Injection, entered. The automatic Reactor Trip function and one manual Reactor Trip switch are disabled to ensure appropriate crew actions and communication. The crew will transition from EOP-0.0A to EOS-0.1A, Reactor Trip Response and then return to EOP-0.0 when Foldout Page criteria for Pressurizer level or Reactor Coolant System (RCS) subcooling are not met. A Steam Generator Tube Rupture is diagnosed in EOP-0.0A and a transition to EOP-3.0A, Steam Generator Tube Rupture will be performed.

The scenario is complicated by a Train A and B Containment Isolation Phase A failure and a Steam Generator Blowdown Valve that fails to automatically isolate.

While in EOP-3.0A, the ruptured Steam Generator will be isolated and a RCS cooldown commenced using the Atmospheric Relief Valves. The scenario is terminated when steps to cooldown the Reactor Coolant System are reached in EOP-3.0A.

Risk Significance:

Failure of risk important system prior to trip: Steam Generator Tube Leak Train B Emergency Diesel Generator Risk significant core damage sequence: Automatic Reactor Trip Failure Steam Generator Tube Rupture Risk significant operator actions: Manually Initiate Reactor Trip Isolate SG Blowdown Flow Initiate Containment Phase A Isolation Identify and Isolate Ruptured SG CPNPP NRC 2014 ES-D-1 Simulator Scenario Outlines