ML13108A229

CP-2013-04 Draft Outlines

ML13108A229
Person / Time
Site:	Comanche Peak
Issue date:	04/08/2013
From:	Kelly Clayton Operations Branch IV
To:
laura hurley
References
Download: ML13108A229 (30)

Text

ES-401 PWR Examination Outline FORM ES-401-2 Facility Name:CPNPP Date of Exam:04/01/2013 K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G

Total Total 1

3 3

3 3

3 3

18 6

2 1

1 2

2 2

1 9

4 Tier Totals 4

4 5

5 5

4 27 10 1

2 3

3 3

3 2

3 3

1 3

2 28 5

2 1

1 0

1 1

1 1

1 1

1 1

10 0

2 3

Tier Totals 3

4 3

4 4

3 4

4 2

4 3

38 8

1 2

3 4

1 2

2 2

Note: 1.

Note: 2.

Note: 3.

Note: 4.

Note: 5.

Note: 6.

Note: 7.*

Note: 8.

Note: 9.

ES-401, 21 of 33 3

2 5

3 3

5 2

2 3

7 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.

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.

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.

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

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.

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.

Tier Group SRO-Only Points RO K/A Category Points A 2 G *

1. Emergency Abnormal Plant Evolutions N/A N/A

2.

Plant Systems

3. Generic Knowledge and Abilities Categories 1

3 3

5 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.

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.

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).

4 2

2 10 1

3 2

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)

Q#

E/APE # / Name / Safety Function K

1 K

2 K

3 A

1 A

2 G

K/A Topic(s)

IR 39 000007 Reactor Trip - Stabilization - Recovery / 1 0

3 Reactor trip status panel 3.5 1

40 000008 Pressurizer Vapor Space Accident / 3 0

3 Controllers and positioners 2.5 1

41 000009 Small Break LOCA / 3 3

7 Existence of adequate natural circulation 4.2 1

000011 Large Break LOCA / 3 0

42 000015 RCP Malfunctions / 4 000017 RCP Malfunctions (Loss of RC Flow) / 4 0

4 Basic steady state thermodynamic relationship between RCS loops and S/Gs resulting from unbalanced RCS flow 2.9 1

000022 Loss of Rx Coolant Makeup / 2 0

43 000025 Loss of RHR System / 4 0

3 Immediate actions contained in EOP for Loss of RHRS 3.9 1

000026 Loss of Component Cooling Water / 8 0

44 000027 Pressurizer Pressure Control System Malfunction / 3 0

1 Definition of saturation temperature 3.1 1

000029 ATWS / 1 0

45 000038 Steam Gen. Tube Rupture / 3 04.

18 Knowledge of the specific bases for EOPs.

3.3 1

000040 Steam Line Rupture - Excessive Heat Transfer

/ 4 0

WE12 Uncontrolled Depressurization of all Steam Generators / 4 0

46 000054 Loss of Main Feedwater / 4 0

1 MFW line break depressurizes the S/G (similar to a steam line break) 4.1 1

47 000055 Station Blackout / 6 0

3 Actions necessary to restore power 3.9 1

48 000056 Loss of Off-site Power / 6 0

5 Initiation (manual) of safety injection process 3.8 1

49 000057 Loss of Vital AC Inst. Bus / 6 0

5 Backup instrument indications 3.2 1

50 000058 Loss of DC Power / 6 0

2 Actions contained in EOP for loss of dc power 4.0 1

51 000062 Loss of Nuclear Svc Water / 4 0

3 The valve lineups necessary to restart the SWS while bypassing the portion of the system causing the abnormal condition 2.6 1

52 000065 Loss of Instrument Air / 8 0

4 Cross-over to backup air supplies 3.0 1

54 W/E04 LOCA Outside Containment / 3 0

2 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 3.8 1

56 W/E11 Loss of Emergency Coolant Recirc. / 4 0

3 Desired operating results during abnormal and emergency situations 3.7 1

55 W/E05 Inadequate Heat Transfer - Loss of Secondary Heat Sink / 4 02.

03 Knowledge of the design, procedural, and operational differences between units.

3.8 1

53 000077 Generator Voltage and Electric Grid Disturbances / 6 04.

01 Knowledge of EOP entry conditions and immediate action steps.

4.6 1

K/A Category Totals:

3 3

3 3

3 3

18 ES-401, 22 of 33 Group Point Total:

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)

Q#

E/APE # / Name / Safety Function K

1 K

2 K

3 A

1 A

2 G

K/A Topic(s)

IR 57 000001 Continuous Rod Withdrawal / 1 12 Long-range effects of core quadrant power tilt 2.8 1

000003 Dropped Control Rod / 1 0

000005 Inoperable/Stuck Control Rod / 1 0

000024 Emergency Boration / 1 0

58 000028 Pressurizer Level Malfunction / 2 06 Checking of RCS leaks 3.3 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

59 000060 Accidental Gaseous Radwaste Rel. / 9 04.

45 Ability to prioritize and interpret the significance of each annunciator or alarm.

4.1 1

000061 ARM System Alarms / 7 0

60 000067 Plant Fire On-site / 8 08 Limits of affected area 2.9 1

61 000068 Control Room Evac. / 8 09 Saturation margin 4.1 1

000069 Loss of CTMT Integrity / 5 0

W/E14 High Containment Pressure / 5 0

000074 Inad. Core Cooling / 4 0

W/E06 Degraded Core Cooling / 4 0

W/E07 Saturated Core Cooling / 4 0

62 000076 High Reactor Coolant Activity / 9 01 Process radiation monitors 2.6 1

W/E01 Rediagnosis / 3 0

W/E02 SI Termination / 3 0

W/E13 Steam Generator Over-pressure / 4 0

65 W/E15 Containment Flooding / 5 01 Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and t

ti d

l f t

2.9 1

W/E16 High Containment Radiation / 9 0

W/E03 LOCA Cooldown - Depress. / 4 0

W/E09 Natural Circulation Operations / 4 0

64 W/E10 Natural Circulation with Steam Voide in Vessel with/without RVLIS. / 4 03 Manipulation of controls required to obtain desired operating results during abnormal, and emergency situations 3.4 1

63 W/E08 RCS Overcooling - PTS / 4 01 Facility operating characteristics during transient conditions, including coolant chemistry and the effects of temperature, d

ti it h

d ti li it ti d

3.4 1

K/A Category Totals:

1 1

2 2

2 1

9 ES-401, 23 of 33 Group Point Total:

ES-401 4

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

Q#

System # / Name K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G

K/A Topic(s)

IR 1

003 Reactor Coolant Pump 0

2 CCW pumps 2.5 1

2 004 Chemical and Volume Control 0

6 Control instrumentation 2.6 1

3,4 005 Residual Heat Removal 0

3 0

3 RHR heat exchanger; RHR pump/motor malfunction 2.5; 2.9 2

5,6 006 Emergency Core Cooling 0

6 1

6 Relationship between ECCS flow and RCS pressure; RCS temperature, including superheat, saturation, and subcooled 3.5; 4.1 2

7 007 Pressurizer Relief/Quench Tank 0

1 Containment 3.3 1

8 008 Component Cooling Water 0

1 CCW indications and controls 3.3 1

9 010 Pressurizer Pressure Control 0

1 Pressure detection systems 2.7 1

10 012 Reactor Protection 0

2 125V dc system 3.4 1

11 013 Engineered Safety Features Actuation 0

1 Definitions of safety train and ESF channel 2.8 1

12,13 022 Containment Cooling 0

2 0

1 Containment pressure; CCS fans 3.6; 3.6 2

025 Ice Condenser CPNPP does not have this system 14 026 Containment Spray 0

7 Loss of containment spray pump suction when in recirculation mode, possibly caused by clogged sump screen, pump inlet high temperature exceeded cavitation, 3.6 1

15 039 Main and Reheat Steam 0

5 RCS T-ave 3.2 1

16 059 Main Feedwater 1

6 Automatic trips for MFW pumps 3.1 1

17 061 Auxiliary/Emergency Feedwater 0

3 Pump head effects when control valve is shut 2.6 1

18 062 AC Electrical Distribution 0

2 Circuit breaker automatic trips 2.5 1

19 063 DC Electrical Distribution 0

1 Meters, annunciators, dials, recorders, and indicating lights 2.7 1

20,21 064 Emergency Diesel Generator 1

0 1

8 Automatic load sequencer: blackout; Consequences of premature opening of breaker under load 3.5; 2.6 2

22,23 073 Process Radiation Monitoring 0

1 04.

06 Radioactive effluent releases; Knowledge of EOP mitigation strategies.

3.6; 3.7 2

24,25 076 Service Water 1

5 0

4 FPS; Emergency heat loads 2.5; 3.5 2

26,27 078 Instrument Air 0

1 0

3 Instrument air compressor; Cross-tied units 2.7; 3 2

28 103 Containment 02.

42 Ability to recognize system parameters that are entry-level conditions for Technical Specifications.

3.9 1

0 K/A Category Totals:

2 3

3 3

3 2

3 3

1 3

2 28 ES-401, Page 24 of 33 Group Point Total:

ES-401 5

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

Q#

System # / Name K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G

K/A Topic(s)

IR 29 001 Control Rod Drive 1

2 Location and interpretation of CRDS ac/dc status alarms 2.9 1

002 Reactor Coolant 0

30 011 Pressurizer Level Control 0

3 VCT level 2.8 1

31 014 Rod Position Indication 0

6 Loss of LVDT 2.6 1

015 Nuclear Instrumentation 0

016 Non-nuclear Instrumentation 0

32 017 In-core Temperature Monitor 0

2 Saturation and subcooling of water 3.7 1

33 027 Containment Iodine Removal 0

1 Fans 3.1 1

028 Hydrogen Recombiner and Purge Control 0

34 029 Containment Purge 01.

32 Ability to explain and apply system limits and precautions.

3.8 1

033 Spent Fuel Pool Cooling 0

034 Fuel Handling Equipment 0

35 035 Steam Generator 0

5 Amount of reserve water in S/G 2.9 1

041 Steam Dump/Turbine Bypass Control 0

36 045 Main Turbine Generator 0

5 Electrohydraulic control 2.6 1

37 055 Condenser Air Removal 0

6 PRM system 2.6 1

056 Condensate 0

38 068 Liquid Radwaste 0

4 Automatic isolation 3.8 1

071 Waste Gas Disposal 0

072 Area Radiation Monitoring 0

075 Circulating Water 0

079 Station Air 0

086 Fire Protection 0

K/A Category Totals:

1 1

0 1

1 1

1 1

1 1

1 10 ES-401, Page 25 of 33 Group Point Total:

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)

Q#

E/APE # / Name / Safety Function K

1 K

2 K

3 A

1 A

2 G

K/A Topic(s)

IR 000007 Reactor Trip - Stabilization - Recovery / 1 0

000008 Pressurizer Vapor Space Accident / 3 0

000009 Small Break LOCA / 3 0

76 000011 Large Break LOCA / 3 02.

04 Ability to explain the variations in control board layouts, systems, instrumentation, and procedural actions between units at a facility.

3.6 1

000015 RCP Malfunctions / 4 000017 RCP Malfunctions (Loss of RC Flow) / 4 0

77 000022 Loss of Rx Coolant Makeup / 2 0

2 Charging pump problems 3.7 1

000025 Loss of RHR System / 4 0

79 000026 Loss of Component Cooling Water / 8 0

4 The normal values and upper limits for the temperatures of the components cooled by CCW 2.9 1

000027 Pressurizer Pressure Control System Malfunction / 3 0

78 000029 ATWS / 1 02.

40 Ability to apply Technical Specifications for a system.

4.7 1

000038 Steam Gen. Tube Rupture / 3 0

80 000040 Steam Line Rupture - Excessive Heat Transfer

/ 4 04.

41 Knowledge of the emergency action level thresholds and classifications.

4.6 1

81 WE12 Uncontrolled Depressurization of all Steam Generators / 4 0

2 Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments 3.9 1

000054 Loss of Main Feedwater / 4 0

000055 Station Blackout / 6 0

000056 Loss of Off-site Power / 6 0

000057 Loss of Vital AC Inst. Bus / 6 0

000058 Loss of DC Power / 6 0

000062 Loss of Nuclear Svc Water / 4 0

000065 Loss of Instrument Air / 8 0

W/E04 LOCA Outside Containment / 3 0

W/E11 Loss of Emergency Coolant Recirc. / 4 0

W/E05 Inadequate Heat Transfer - Loss of Secondary Heat Sink / 4 0

000077 Generator Voltage and Electric Grid Disturbances / 6 0

K/A Category Totals:

0 0

0 0

3 3

6 ES-401, 22 of 33 Group Point Total:

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)

Q#

E/APE # / Name / Safety Function K

1 K

2 K

3 A

1 A

2 G

K/A Topic(s)

IR 000001 Continuous Rod Withdrawal / 1 0

000003 Dropped Control Rod / 1 0

000005 Inoperable/Stuck Control Rod / 1 0

000024 Emergency Boration / 1 0

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

82 000037 Steam Generator Tube Leak / 3 02 Agreement/disagreement among redundant radiation monitors 3.9 1

83 000051 Loss of Condenser Vacuum / 4 02.

44 Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and di ti ff t

l t

d t

diti 4.4 1

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

000069 Loss of CTMT Integrity / 5 0

85 W/E14 High Containment Pressure / 5 04.

49 Ability to perform without reference to procedures those actions that require immediate operation of system components and t

l 4.4 1

000074 Inad. Core Cooling / 4 0

W/E06 Degraded Core Cooling / 4 0

W/E07 Saturated Core Cooling / 4 0

000076 High Reactor Coolant Activity / 9 0

W/E01 Rediagnosis / 3 0

W/E02 SI Termination / 3 0

W/E13 Steam Generator Over-pressure / 4 0

W/E15 Containment Flooding / 5 0

W/E16 High Containment Radiation / 9 0

84 W/E03 LOCA Cooldown - Depress. / 4 01 Facility conditions and selection of appropriate procedures during abnormal and emergency operations 4.2 1

W/E09 Natural Circulation Operations / 4 0

W/E10 Natural Circulation with Steam Voide in Vessel with/without RVLIS. / 4 0

W/E08 RCS Overcooling - PTS / 4 0

K/A Category Totals:

0 0

0 0

2 2

4 ES-401, 23 of 33 Group Point Total:

ES-401 4

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

Q#

System # / Name K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G

K/A Topic(s)

IR 003 Reactor Coolant Pump 0

88 004 Chemical and Volume Control 02.

25 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

4.2 1

005 Residual Heat Removal 0

006 Emergency Core Cooling 0

007 Pressurizer Relief/Quench Tank 0

008 Component Cooling Water 0

010 Pressurizer Pressure Control 0

86 012 Reactor Protection 0

1 Faulty bistable operation 3.6 1

013 Engineered Safety Features Actuation 0

022 Containment Cooling 0

025 Ice Condenser 0

026 Containment Spray 0

87 039 Main and Reheat Steam 0

3 Indications and alarms for main steam and area radiation monitors (during SGTR) 3.7 1

89 059 Main Feedwater 0

1 Feedwater actuation of AFW system 3.6 1

061 Auxiliary/Emergency Feedwater 0

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

90 103 Containment 04.

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

4.7 1

0 K/A Category Totals:

0 0

0 0

0 0

0 3

0 0

2 5

ES-401, Page 24 of 33 Group Point Total:

ES-401 5

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

Q#

System # / Name K

1 K

2 K

3 K

4 K

5 K

6 A

1 A

2 A

3 A

4 G

K/A Topic(s)

IR 001 Control Rod Drive 0

002 Reactor Coolant 0

011 Pressurizer Level Control 0

014 Rod Position Indication 0

015 Nuclear Instrumentation 0

016 Non-nuclear Instrumentation 0

017 In-core Temperature Monitor 0

027 Containment Iodine Removal 0

91 028 Hydrogen Recombiner and Purge Control 0

3 The hydrogen air concentration in excess of limit flame propagation or detonation with resulting equipment damage in containment 4.0 1

029 Containment Purge 0

033 Spent Fuel Pool Cooling 0

034 Fuel Handling Equipment 0

035 Steam Generator 0

041 Steam Dump/Turbine Bypass Control 0

045 Main Turbine Generator 0

055 Condenser Air Removal 0

92 056 Condensate 0

4 Loss of condensate pumps 2.8 1

068 Liquid Radwaste 0

93 071 Waste Gas Disposal 04.

35 Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects.

4.0 1

072 Area Radiation Monitoring 0

075 Circulating Water 0

079 Station Air 0

086 Fire Protection 0

K/A Category Totals:

0 0

0 0

0 0

0 2

0 0

1 3

ES-401, Page 25 of 33 Group Point Total:

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

Form ES-401-3 Facility Name:CPNPP Date of Exam:04/01/2013 Q#

IR IR 94 2.1. 45 Ability to identify and interpret diverse indications to validate the response of another indicator.

4.3 1

66 2.1. 2 Ability to interpret and execute procedure steps.

4.6 1

67 2.1. 36 Knowledge of procedures and limitations involved in core alterations.

3.0 1

2.1.

2.1.

2.1.

Subtotal 2

1 95 2.2. 07 Knowledge of the process for conducting special or infrequent tests.

3.6 1

96 2.2. 35 Ability to determine Technical Specification Mode of Operation.

4.5 1

68 2.2. 18 Knowledge of the process for managing maintenance activities during shutdown operations, such as risk assessments, work prioritization, etc.

2.6 1

69 2.2. 38 Knowledge of conditions and limitations in the facility license.

3.6 1

2.2.

2.2.

Subtotal 2

2 97 2.3. 05 Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

2.9 1

98 2.3. 15 Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

3.1 1

70 2.3. 04 Knowledge of radiation exposure limits under normal or emergency conditions.

3.2 1

71 2.3. 12 Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc 3.2 1

72 2.3. 13 Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high radiation areas aligning filters etc 3.4 1

2.3.

Subtotal 3

2 99 2.4. 29 Knowledge of the emergency plan.

4.4 1

100 2.4. 40 Knowledge of SRO responsibilities in emergency plan implementation.

4.5 1

73 2.4. 11 Knowledge of abnormal condition procedures.

4.0 1

74 2.4. 31 Knowledge of annunciator alarms, indications, or response procedures.

4.2 1

75 2.4. 46 Ability to verify that the alarms are consistent with the plant conditions.

4.2 1

2.4.

Subtotal 3

2 10 7

ES-401, Page 26 of 33 RO SRO-Only Tier 3 Point Total

3.

Radiation Control

4.

Emergency Procedures /

Plan

2.

Equipment Control Category Topic

1.

Conduct of Operations K/A #

ES-401 Record of Rejected K/As Form ES-401-4 ES-401, Page 27 of 33 Tier /

Group Randomly Selected K/A Reason for Rejection 2 / 1 073 G 2.4.41 Replacement K/A 073 G 2.4.6.

(Q#23) - This K/A topic (Knowledge of emergency action level thresholds and classifications) is not covered in the RO curriculum. Randomly reselected 073 G 2.4.6.

2 / 1 076 K1.09 Replacement K/A 076 K1.15.

(Q#25) - There is no Reactor Building Closed Cooling Water System at CPNPP. Randomly reselected 076 K1.15.

2 / 1 078 A2.01 Replacement K/A 059 A2.01.

(Q#89) - Randomly reselected 059 A2.01 to avoid three (3) Instrument Air System questions between the RO and SRO exams.

3 / 1 G 2.1.26 Replacement K/A G 2.1.45.

(Q#94) - This K/A does not have a corresponding 10CFR55.43(b) reference for the SRO per ES-401-2 Note #9. Randomly reselected G 2.1.45.

Note to Chief Examiner There are 24 topic areas (procedures) in the Comanche Peak Tier 1 Group 1 setting. There are a total of 24 questions in Tier 1 Group 1 between the RO and SRO exams; therefore, every procedure requires a corresponding question per ES-401-2. Questions 76 through 81 were directly reselected from the CPNPP list of available procedures.

1 / 1 008 G 2.2.4 Replacement K/A 011 G 2.2.4.

(Q#76) - Replaced procedure to meet NUREG 1021, ES-401-2, Note #4 requirements. Reselected 011 G 2.2.4. 008 AK2.03 was used on Tier 1 Group 1 (Q#40).

1 / 1 025 AA2.02 Replacement K/A 022 AA2.02.

(Q#77) - Replaced procedure to meet NUREG 1021, ES-401-2, Note #4 requirements. Reselected 022 AA2.02. 025 AK3.03 was used on Tier 1 Group 1 (Q#43).

1 / 1 056 G 2.2.4 Replacement K/A 029 G 2.2.40.

(Q#78) - Replaced procedure and K/A (there are no differences between units for either LOOP or ATWS) to meet NUREG 1021, ES-401-2, Note #4 requirements. Randomly reselected 029 G 2.2.40. 056 AA1.05 was used on Tier 1 Group 1 (Q#48).

1 / 1 065 AA2.04 Replacement K/A 026 AA2.04.

(Q#79) - Replaced procedure to meet NUREG 1021, ES-401-2, Note #4 requirements. Reselected 026 AA2.04. 065 AK3.04 was used on Tier 1 Group 1 (Q#52).

1 / 1 W/E04 G 2.4.41 Replacement K/A 040 G 2.4.41.

(Q#80) - Replaced procedure to meet NUREG 1021, ES-401-2, Note #4 requirements. Reselected 040 G 2.4.41. W/E04 EK2.02 was used on Tier 1 Group 1 (Q#54).

1 / 1 W/E05 EA2.2 Replacement K/A W/E12 EA2.2.

(Q#81) - Replaced procedure to meet NUREG 1021, ES-401-2, Note #4 requirements. Reselected W/E12 EA2.2. W/E05 G 2.2.23 was used on Tier 1 Group 1 (Q#55).

Note to Chief Examiner These SRO topics were randomly reselected as there are more procedures/systems in Tier 1 Group 2 and Tier 2 Group 2 than available questions.

1 / 2 001 AA2.02 Replacement K/A 037 AA2.02.

(Q#82) - Replaced procedure to meet NUREG 1021, ES-401-2, Note #4 requirements. Randomly reselected 037 AA2.02. 001 AK1.12 was used on Tier 1 Group 2 (Q#57).

1 / 2 076 G 2.4.31 Replacement K/A 051 G 2.2.44.

(Q#83) - Replaced procedure to meet NUREG 1021, ES-401-2, Note #4 requirements. Randomly reselected 051 G 2.2.44. 076 AK2.01 was used on Tier 1 Group 2 (Q#62) and G 2.4.31 was used on Tier 3 (Q#74).

2 / 2 068 A2.02 Replacement K/A 056 A2.04.

(Q#92) - Replaced system to meet NUREG 1021, ES-401-2, Note #4 requirements. Randomly reselected 056 A2.04 (A2.04 is the only K/A that is 2.5 IR). 068 AA2.09 was used on Tier 1 Group 2 (Q#61)

Note to Chief Examiner These topics were reselected to avoid overlap of generic K/As.

2 / 1 076 G 2.2.38 Replacement K/A 004 G 2.2.25.

(Q#88) - Reselected 004 to avoid three (3) Service Water System questions between the RO and SRO exams and G 2.2.25 to avoid generic K/A overlap. G 2.2.38 was used on Tier 3 (Q#69).

1 / 1 W/E05 G 2.2.38 Replacement K/A W/E05 G 2.2.3.

(Q#55) - Reselected W/E05 G 2.2.3 to avoid generic K/A overlap. G 2.2.38 was used on Tier 3 (Q#69).

2 / 1 073 G 2.4.31 Replacement K/A 073 K3.01.

(Q#22) - Reselected 073 K3.01 to avoid generic K/A overlap and balance skyscraper (selected only topic available in K3). G 2.4.31 was used on Tier 3 (Q#74).

1 / 2 060 G 2.4.35 Replacement K/A 060 G 2.4.45.

(Q#59) - Reselected 060 G 2.4.45 to avoid generic K/A overlap. G 2.4.35 was used on Tier 2 Group 2 (Q#93).

Note to Chief Examiner These topics were reselected to avoid overlap on the written exam.

1/1 038 G 2.4.8 Replacement K/A 038 G.2.4.18.

Q#45 - Reselected 038 G 2.4.18 to avoid question overlap. 038 G 2.4.8 overlaps with the K/A used on Q#23 (073 G 2.4.6) since CPNPP has few Process Radiation Monitors that interface with EOPs.

Administrative Topics Outline CPNPP NRC 2013 ES-301-1 RO Admin JPM Outline Rev b.docx Facility:

CPNPP Units 1 and 2 Date of Examination:

04/01/13 Examination Level RO Operating Test Number:

NRC Administrative Topic (see Note)

Type Code*

Describe Activity to be Performed Conduct of Operations (RA1)

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

JPM: Restore Refueling Water Storage Tank Level (RO1307).

Conduct of Operations (RA2)

M, R 2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation (4.3).

JPM: Perform Reactor Coolant System Pressure /

Temperature Verification (RO5115).

Equipment Control (RA3)

M, R 2.2.1 Ability to perform pre-startup procedures including operating those controls associated with plant equipment that could affect reactivity (4.5).

JPM:

Perform a 1/M Plot and Predict Critical Conditions (RO1003).

Radiation Control (RA4)

D, S 2.3.13 Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc. (3.4)

JPM:

Perform Actions for Fuel Handling Accident in the Spent Fuel Pool (RO4504).

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)

Administrative Topics Outline Task Summary CPNPP NRC 2013 ES-301-1 RO Admin JPM Outline Rev b.docx RA1 The applicant will restore Refueling Water Storage Tank (RWST) level when it is determined out of specification per SOP-104A, Reactor Make-Up and Chemical Control System, Section 5.2.7, Makeup to RWST. Critical steps include calculating the required volume of borated water necessary to raise RWST level, Boric Acid Flowrate, total gallons of Boric Acid, and potentiometer settings for the Flow Control Valves. This is a modified bank JPM.

RA2 The applicant will perform a Reactor Coolant System Pressure/Temperature Verification per ABN-905A, Loss of Control Room Habitability, Attachment 7, RCS Pressure/Temperature Verification. Critical steps include calculating saturation temperatures, subcooled margin, and cooldown rate. This is a modified bank JPM.

RA3 The applicant will perform a 1/M plot for a Reactor Startup per IPO-002A, Plant Startup from Hot Standby, Attachment 2, Inverse Count Rate Ratio Calculation.

The critical steps include accurately calculating and plotting 1/M and predicting critical conditions. This is a modified bank JPM.

RA4 The applicant will implement radiological emergency actions per ABN-908, Fuel Handling Accident, Section 3.0, Fuel Handling Accident in the Fuel Building Involving Spent Fuel. The critical steps include initiating local evacuation, activating the Radiological Emergency Alarm, and ensuring proper ventilation alignment. This is a bank JPM.

Administrative Topics Outline CPNPP NRC 2013 ES-301-1 SRO Admin JPM Outline Rev b.docx Facility:

CPNPP Units 1 and 2 Date of Examination:

04/01/13 Examination Level SRO Operating Test Number:

NRC Administrative Topic (see Note)

Type Code*

Describe Activity to be Performed Conduct of Operations (SA1)

M, R 2.1.25 Ability to interpret reference materials such as graphs, curves, tables, etc. (4.2).

JPM: Restore Refueling Water Storage Tank Level and Evaluate Technical Specifications (SO1211).

Conduct of Operations (SA2)

M, R 2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation (4.4).

JPM: Perform Reactor Coolant System Pressure /

Temperature Verification and Evaluate Technical Specifications (SO1005).

Equipment Control (SA3)

M, R 2.2.14 Knowledge of the process for controlling equipment configuration or status. (4.3)

JPM:

Determine Fire Compensatory Measures for an Emergent Condition (New).

Radiation Control (SA4)

M, R 2.3.6 Ability to approve release permits (3.8).

JPM:

Review a Liquid Waste Release Permit (SO1039).

Emergency Plan (SA5)

M, R 2.4.44 Knowledge of emergency plan protective action recommendations. (4.4)

JPM:

Determine Protective Action Requirements (SO1140).

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)

Administrative Topics Outline Task Summary CPNPP NRC 2013 ES-301-1 SRO Admin JPM Outline Rev b.docx SA1 The applicant will restore Refueling Water Storage Tank (RWST) per SOP-104A, Reactor Make-Up and Chemical Control System, Section 5.2.7, Makeup to RWST. Critical steps include calculating the required volume of borated water necessary to raise RWST level, Boric Acid Flowrate, total gallons of Boric Acid, and potentiometer settings for the Flow Control Valves and then evaluating Technical Specifications when it is determined that RWST temperature is out of specification. This is a modified bank JPM.

SA2 The applicant will perform a Reactor Coolant System Pressure/Temperature Verification per ABN-905A, Loss of Control Room Habitability, Attachment 7, RCS Pressure/Temperature Verification. Critical steps include calculating saturation temperatures, subcooled margin, cooldown rate, and evaluating Technical Specifications when the cooldown rate exceeds 100°F/hr. This is a modified bank JPM.

SA3 The applicant will evaluate a Fire Protection Impairment per STA-738, Fire Protection Systems/Equipment Impairments. The critical steps are to determine Fire Watch implementation and other Compensatory Measures. This is a modified bank JPM.

SA4 The applicant will review a Liquid Waste Release Permit per STA-603, Control of Station Radioactive Effluents and STA-603-10, Batch Liquid Radioactive Effluent Release Data Sheet. The critical steps include identifying any errors and actions required prior to approving the release. This is a modified bank JPM.

SA5 The applicant will determine Protective Actions per EPP-304, Protective Action Recommendations. The critical steps include determining the proper Protective Actions, Pasquill Stability Class, and Zones to be evacuated or sheltered. This is a modified bank JPM.

ES-301 Control Room / In-Plant Systems Outline Form ES-301-2 Page 1 of 3 CPNPP NRC 2013 ES-301-2 RO & SRO JPM Outline Rev c.docx Facility:

CPNPP Units 1 and 2 Date of Examination:

04/01/13 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-1 004 - Chemical and Volume Control System (RO1335)

Emergency Boration from the RWST (RO Only)

A, M, S 1

S-2 010 - Pressurizer Pressure Control System (RO1824)

Wide Range Pressure Transmitter Failure in MODE 5 A, L, D, S 3

S-3 003 - Reactor Coolant Pump System (RO1118)

Respond to Reactor Coolant Pump Seal Failure D, S 4P S-4 059 - Main Feedwater System (RO3408)

Respond to a Heater Drain Pump Trip A, N, S 4S S-5 026 - Containment Spray (CS) System (RO2002)

Transfer CS from Injection to Recirculation A, D, EN, S 5

S-6 064 - Emergency Diesel Generator System (RO4302E)

Loss of Both 6900 Volt Safeguards Buses A, EN, N, S 6

S-7 086 - Fire Protection System (RO4405)

Respond to a Fire in the Safeguards Building D, S 8

S-8 060 - Accidental Gaseous Radwaste Release (RO4006)

Perform a Containment Pressure Reduction A, D, S 9

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

P-1 004 - Chemical & Volume Control System (RO5112)

Perform Actions to Restart Positive Displacement Pump D, E, R 2

P-2 015 - Nuclear Instrumentation System (RO1818)

Respond to Loss of Source Range Instrumentation E, N, R 7

P-3 035 - Steam Generator (SG) System (RO5115)

Transfer SG Atmospheric Relief Valve Control D, E, R 4P

ES-301 Control Room / In-Plant Systems Outline Form ES-301-2 Page 2 of 3 CPNPP NRC 2013 ES-301-2 RO & SRO JPM Outline Rev c.docx 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 initiate an emergency boration per ABN-107, Emergency Boration following a Reactor Trip with two stuck Control Rods. The alternate path occurs when Boric Acid flow is not available due to a clogged filter and the Refueling Water Storage Tank must be aligned via Attachment 4, Transfer of Charging Pump Suction to the RWST. This is a modified bank JPM under the Chemical and Volume Control System -

Reactivity Control safety function. This incorporates Operating Experience from Comanche Peak and is a PRA significant action. (K/A 004.A2.14 - IR 3.8 / 3.9)

S-2 The applicant will respond to a Wide Range Pressure Transmitter failure while in MODE 5 per ALM-0053A, 1-ALB-5C, Window 1.4 - PORV 455A/456 NOT CLOSE or ABN-715, Wide Range RCS Pressure Instrument Malfunction. The alternate path occurs when the Power Operated Relief Valve opens and 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.A4.03 - IR 4.0 / 3.8)

S-3 The applicant will respond to a Reactor Coolant Pump Seal failure per ABN-101, Reactor Coolant Pump Trip / Malfunction. This is a bank JPM under the Reactor Coolant Pump System - Primary System Heat Removal from Reactor Core Safety Function.

(K/A 015.AA1.22 - IR 4.0 / 4.2)

S-4 The applicant will respond to a Heater Drain Pump trip per ABN-302, Feedwater, Condensate, Heater Drain System Malfunction, Section 4.0, Heater Drain Pump Trip. The alternate path occurs when the Main Turbine fails to Runback with Control Rods in AUTO and both Main Feedwater Pumps trip on a loss of suction pressure. 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 059.A3.07 - IR 3.4 / 3.5)

ES-301 Control Room / In-Plant Systems Outline Form ES-301-2 Page 3 of 3 CPNPP NRC 2013 ES-301-2 RO & SRO JPM Outline Rev c.docx S-5 The applicant will transfer Containment Spray suction to the Containment Sumps per EOS-1.3A, Transfer to Cold Leg Recirculation, Attachment 1.H, Containment Spray Switchover Criterion. The alternate path occurs when one of the Containment Sump Valves to the Containment Spray Pumps cannot be opened. This is a bank JPM under the Containment Spray System - Containment Integrity Safety Function. This is a PRA significant action. (K/A 026.A4.01 - IR 4.5 / 4.3)

S-6 The applicant will respond to a loss of both 6900 Volt Safeguards Buses per ABN-601, Response to a 138/345 KV System Malfunction, Section 7.0, Loss of Both Safeguards Buses - MODE 1, 2, 3, or 4. The alternate path includes tripping the Reactor, stopping Reactor Coolant Pumps, and closing the Train A Emergency Diesel Generator Output Breaker. This is a new JPM under the Emergency Diesel Generator System - Electrical Safety Function. This is a PRA significant action. (K/A 064.A4.06 - IR 3.9 / 3.9)

S-7 The applicant will respond to a fire in the Safeguards Building per ABN-804A, Respond to a Fire in the Safeguards Building, Section 5.0, Fire Affecting Safeguards Building Fire Area 1SD. This is a bank JPM under the Fire Protection System - Plant Service Systems Safety Function. (K/A 068.AA1.22 - IR 4.0 / 4.3)

S-8 The applicant will perform a Containment Pressure Reduction per SOP-801A, Containment Ventilation System. The alternate path requires closing the Containment release path upon receipt of a high radiation alarm. This is a bank JPM under the Accidental Gaseous Radwaste Release - Radioactive Release safety function.

(K/A 060.AA2.05 - IR 3.7 / 4.2)

P-1 The applicant will perform Unit 2 local actions to restart the Positive Displacement Pump per ABN-301, Instrument Air System Malfunction and SOP-103B, Chemical and Volume Control. This is a bank JPM under the Chemical and Volume Control System - Reactor Coolant System Inventory Control Safety Function. (K/A 004.A4.01 - IR 3.1 / 3.5)

P-2 The applicant will respond to a Loss of Source Range Instrumentation while in MODE 3 and perform actions to isolate potential dilution flowpaths per ABN-701, Source Range Instrument Malfunction, Attachment 1, Actions Required When SR Instrumentation Cannot Be Restored, Step 6. This is a new JPM under the Nuclear Instrumentation System - Instrumentation Safety Function. (K/A G 2.1.30 - IR 4.4 / 4.0)

P-3 The applicant will transfer control of Unit 2 Steam Generator Atmospheric Relief Valves per ABN-905B, Loss of Control Room Habitability, Attachment 9, Control Transfer of Steam Generator Atmospheric Relief Valves. This is a bank JPM under Steam Generator System - Primary System Heat Removal from Reactor Core safety function. This is a PRA significant action. (K/A 068.AA1.01 - IR 4.3 / 4.5)

ES-301 Transient and Event Checklist Form ES-301-5 CPNPP NRC 2013 ES-301-5 Transient and Event Checklist Rev a.docx Facility:

CPNPP 1 and 2 Date of Exam:

04/01/13 Operating Test No.:

NRC A

P P

L I

C A

N T

E V

E N

T T

Y P

E SCENARIOS CPNPP #1 CPNPP #2 CPNPP #3 CPNPP #4 T

O T

A L

MINIMUM(*)

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

SRO-U SRO-I RO RX 4

4 2

1 1

0 NOR 4

4 4

1 4

2 1,2 1

1 1

I/C 1,2,3 1,3,6 2,7 2,3 3,7 2,8 1,2,3, 4

1,3,4 1,2,6, 7

3,4 3,7 4,8 4

4 2

MAJ 5

5 5

5,6 5,6 5,6 5

5 5

6,9 6,9 6,9 2

2 1

TS 1,3,4 3,4 1,2 4,5 0

2 2

RX 1

1 0

NOR 1

1 1

I/C 4

4 2

MAJ 2

2 1

TS 0

2 2

RX 1

1 0

NOR 1

1 1

I/C 4

4 2

MAJ 2

2 1

TS 0

2 2

RX 1

1 0

NOR 1

1 1

I/C 4

4 2

MAJ 2

2 1

TS 0

2 2

RX 1

1 0

NOR 1

1 1

I/C 4

4 2

MAJ 2

2 1

TS 0

2 2

Instructions:

1.

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

2.

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

3.

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

Appendix D Scenario Outline Form ES-D-1 CPNPP NRC 2013 ES-D-1 Simulator Scenario Outlines Rev c.docx Facility:

CPNPP 1 & 2 Scenario No.:

1 Op Test No.:

April 2013 NRC Examiners:

Operators:

Initial Conditions: 100% power MOL - RCS Boron is 908 ppm (by sample).

Turnover: Maintain steady-state power conditions.

Critical Tasks:

Control Pressurizer Pressure to Avoid RPS or ESFAS Actuation per ABN-705, Pressurizer Pressure Malfunction.

Control Steam Generator Level to Avoid RPS or ESFAS Actuation per ABN-708, Feedwater Flow Instrument Malfunction.

Identify and Isolate the Faulted Steam Generator Prior to Exiting EOP-2.0A, Faulted Steam Generator Isolation.

Event No.

Malf. No.

Event Type*

Event Description 1

+10 min RX08A I (RO, SRO)

TS (SRO)

Pressurizer Pressure Channel (PT-455) Fails Low.

2

+20 min RX02A I (BOP, SRO)

Steam Generator (1-01) Steam Flow Instrument (FT-512) Fails High.

3

+25 min CV01B C (RO, SRO)

TS (SRO)

Centrifugal Charging Pump (1-01) Trip.

4

+40 min FW22 R (RO)

N (BOP, SRO)

TS (SRO)

Low Pressure Feedwater Heater Bypass Valve (PV-2286) Fails Open.

5

+45 min FW25A M (RO, BOP, SRO) Feedwater Line Leak to Steam Generator (1-01) Outside Containment After Feedwater Isolation Valve (600 second ramp).

6

+50 min CS02F CS02H C (RO)

Train B Containment Spray Pumps 1-02 & 1-04 Safety Injection Sequencer Start Failure.

7

+50 min FW38A OVRDE I (BOP)

Steam Generator 1-01 Feedwater Isolation Valve (HS-2134)

Actuation Failure.

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications Actual Target Quantitative Attributes 7

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)

Appendix D Scenario Outline Form ES-D-1 CPNPP NRC 2013 ES-D-1 Simulator Scenario Outlines Rev c.docx SCENARIO

SUMMARY

NRC #1 The crew will assume the watch at 100% power with no scheduled activities per IPO-003A, Power Operations. The first event is a Pressurizer Pressure Channel (PT-455) that fails low. The crew will enter ABN-705, Pressurizer Pressure Malfunction, Section 2.0, place the Pressurizer Master Pressure Controller in MANUAL, transfer to an Alternate Channel, and restore Pressurizer Pressure Control to AUTO. The SRO will refer to Technical Specifications.

The next event is a high failure of Steam Generator (1-01) Steam Flow Instrument, FT-512. Operator actions are per ABN-707, Steam Flow Instrument Malfunction, Section 2.0. The crew must manually control Steam Generator level, transfer to an Alternate Channel, and restore Steam Generator (SG)

Feedwater Flow Control to AUTO.

This is followed by a Centrifugal Charging Pump (CCP) trip. The Initial Operator Action is to start the Standby CCP per ABN-105, Chemical and Volume Control System Malfunction, Section 3.0. In the event Letdown is isolated, Letdown flow will be restored per the Control Board Job Aid. The SRO will refer to Technical Specifications.

When Technical Specifications have been referenced, the Low Pressure Heater Bypass Valve fails open. Entry into ABN-302, Feedwater, Condensate, Heater Drain System Malfunction, Section 7.0, is required and Rod Control is returned to AUTO and a Manual Turbine Runback to 900 MWe is performed. During this event, Control Rod position may drop below the Rod Insertion Limit (RIL) and when informed, the SRO will refer to Technical Specifications.

When plant conditions are stable, a Feed Line Break will commence on a 600 second ramp outside Containment downstream of Steam Generator (SG) 1-01 Feed Line Isolation Valve HS-2134. The crew will observe lowering Pressurizer pressure and level and manually initiate a Reactor Trip and Safety Injection. EOP-0.0A, Reactor Trip or Safety Injection, is entered and actions implemented until it is determined that SG 1-01 pressure is lower than the other Steam Generators and a transition into EOP-2.0A, Faulted Steam Generator Isolation, is performed.

The scenario includes a Feedwater Isolation Signal actuation failure on HS-2134, Main Feedwater Line Isolation Valve which fails to close when the P-4 interlock is satisfied. Additionally, the Train B Containment Spray Pumps must be manually started due to a Safety Injection Sequencer failure.

This scenario is terminated when the Faulted Steam Generator is identified and isolated per EOP-2.0A and Letdown flow is established per EOS-1.1A, Safety Injection Termination.

Risk Significance:

Failure of risk important system prior to trip:

Centrifugal Charging Pump Trip Risk significant core damage sequence:

Feed Line Break Outside Containment Risk significant operator actions:

Manually Initiate Turbine Runback Start Train B Containment Spray Pumps Isolate Faulted Steam Generator Terminate Safety Injection Flow

Appendix D Scenario Outline Form ES-D-1 CPNPP NRC 2013 ES-D-1 Simulator Scenario Outlines Rev c.docx Facility:

CPNPP 1 & 2 Scenario No.:

2 Op Test No.:

April 2013 NRC Examiners:

Operators:

Initial Conditions: 100% power MOL - RCS Boron is 908 ppm (by sample).

Turnover: Maintain steady-state power conditions. Positive Displacement Charging Pump and 75 GPM Letdown Orifice in service per Radiation Protection request for maintenance.

Critical Tasks:

Control Steam Generator Level to Avoid RPS or ESFAS Actuation per ABN-709, Feed Header Pressure Instrument Malfunction.

Identify and Isolate the Faulted Steam Generator Prior to Exiting EOP-2.0A, Faulted Steam Generator Isolation.

Trip Reactor Coolant Pumps within 10 minutes upon a Loss of Subcooling per EOP-0.0A, Reactor Trip or Safety Injection, Foldout Page.

Initiate Cooldown of Reactor Coolant System Prior to Exiting ECA-3.1A, SGTR With Loss of Reactor Coolant - Subcooled Recovery Desired.

Event No.

Malf. No.

Event Type*

Event Description 1

+25 min N (RO)

Start Centrifugal Charging Pump 1-01, Raise Letdown Flow to 120 GPM, and Secure Positive Displacement Charging Pump.

2

+30 min RX18 I (BOP, SRO)

Feed Header Pressure Transmitter (PT-508) Fails High.

3

+40 min RP05D I (RO, SRO)

TS (SRO)

Reactor Coolant System Loop (1-04) Narrow Range Cold Leg Temperature Instrument (TI-441A) Fails High.

4

+60 min SG01A R (RO)

N (BOP, SRO)

TS (SRO)

Steam Generator (1-01) Tube Leak at 10 GPM.

Rapid Down Power Required.

5

+65 min SG01A M (RO, BOP, SRO) Steam Generator (1-01) Tube Rupture at 400 GPM (300 second ramp).

6

+65 min MS10A1 MS10A2 M (RO, BOP, SRO) Main Steam Safety Valves (MS-021 & MS-022) on Steam Generator (1-01) Fail Open Upon Reactor Trip.

7

+65 min RX16A C (RO)

Power Operated Relief Valve (PCV-455A) Fails Open Upon Reactor Trip.

8

+65 min RH01C C (BOP)

Residual Heat Removal Pump (1-01) Safety Injection Sequencer Start Failure.

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications Actual Target Quantitative Attributes 6

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

Appendix D Scenario Outline Form ES-D-1 CPNPP NRC 2013 ES-D-1 Simulator Scenario Outlines Rev c.docx SCENARIO

SUMMARY

NRC #2 The crew will assume the watch at 100% power with no scheduled activities per IPO-003A, Power Operations. The Positive Displacement Charging Pump (PDP) is in service and Letdown flow is 75 GPM per Radiation Protection while maintenance is performed. The scenario begins by placing Centrifugal Charging Pump 1-01 in service, raising Letdown flow to 120 GPM, and securing the Positive Displacement Charging Pump (PDP) per SOP-103A, Chemical and Volume Control System.

When the PDP is secured, a Main Feedwater (MFW) Header Pressure Transmitter will fail high. Entry into ABN-709, Feedwater Header Pressure Instrument Malfunction, Section 5.0, is required and the MFW Pump Turbine Master Speed Controller is placed in MANUAL. This controller will remain in MANUAL for the duration of the scenario and require monitoring/adjustment during the subsequent down power.

When plant parameters are restored to normal, 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.

When Technical Specifications have been referenced, a 10 GPM Steam Generator (SG) Tube Leak will ensue. The crew will enter ABN-106, High Secondary Activity, Section 3.0, and determine that a Rapid Downpower is required. The SRO will refer to Technical Specifications. When power has been reduced 3% to 5%, a 400 GPM Steam Generator Tube Rupture will commence on a 300 second ramp.

When control of the power reduction is no longer feasible, the crew will trip the Reactor, initiate a Safety Injection, and enter EOP-0.0A, Reactor Trip or Safety Injection. Two Main Steam Safety Valves and a Power Operated Relief Valve (PORV) will fail open upon Reactor Trip. The PORV should be closed once identified at Step 10 of EOP-0.0A if not already recognized. When SG 1-01 is identified as faulted, the crew will transition from EOP-0.0A to EOP-2.0A, Faulted Steam Generator Isolation. When SG 1-01 is isolated in EOP-2.0A, a transition to EOP-3.0A, Steam Generator Tube Rupture, will be made.

When it is determined that Ruptured Steam Generator 1-01 pressure is less than 420 psig in EOP-3.0A, a transition to ECA-3.1A, SGTR with Loss of Reactor Coolant - Subcooled Recovery Desired, will be made. The scenario includes Train A 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 Atmospheric Relief Valves in ECA-3.1A.

Risk Significance:

Risk significant core damage sequence:

Steam Generator Tube Rupture Failed Open Main Steam Safety Valves Risk significant operator actions:

Restore Steam Generator Level Control Close Pressurizer PORV Start Residual Heat Removal Pump Identify & Isolate Faulted/ Ruptured SG Initiate RCS Cooldown

Appendix D Scenario Outline Form ES-D-1 CPNPP NRC 2013 ES-D-1 Simulator Scenario Outlines Rev c.docx Facility:

CPNPP 1 & 2 Scenario No.:

3 Op Test No.:

April 2013 NRC Examiners:

Operators:

Initial Conditions: 100% power MOL - RCS Boron is 908 ppm (by sample).

Turnover: Maintain steady-state power conditions.

Critical Tasks:

Control Reactor Coolant System Pressure to Avoid RPS or ESFAS Actuation per ABN-603, Loss of Protection or Instrument Bus.

Control Steam Generator Level to Avoid RPS or ESFAS Actuation per ABN-603, Loss of Protection or Instrument Bus.

Restore Component Cooling Water Flow Prior to Tripping the Reactor per ABN-502, Component Cooling Water System Malfunctions.

Initiate Train A and/or Train B Containment Isolation Phase B due to Failure to Automatically Actuate Prior to Exiting FRZ-0.1A, Response to High Containment Pressure.

Event No.

Malf. No.

Event Type*

Event Description 1

+20 min ED07A C (RO, BOP, SRO)

TS (SRO)

Loss of Inverter (IV1PC1).

2

+30 min CC02A CC03A C (BOP, SRO)

TS (SRO)

Train A Component Cooling Water Pump 1-01 Trip.

Train B Component Cooling Water Pump 1-02 Auto Start Failure.

3

+35 min CV16A I (RO, SRO)

Volume Control Tank Level Transmitter (LT-112) Fails Low.

4

+45 min RC17B C (RO, SRO)

Reactor Coolant Leak Inside Containment on Loop 2 Hot Leg of 600 GPM on 600 second ramp.

5

+60 min RC08B2 M (RO, BOP, SRO) Large Break Loss of Coolant Accident Inside Containment on Loop 2 Hot Leg Upon Reset of Containment Spray Signal in EOP-1.0A.

6

+60 min SI04D C (BOP)

Safety Injection Pump (1-02) Auto Start Failure on Safety Injection Signal.

7

+60 min RP10A RP10B I (BOP)

Automatic Train A and B Containment Isolation Phase B Failure.

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications Actual Target Quantitative Attributes 7

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) 1 EOP contingencies requiring substantive actions (0-2) 4 Critical tasks (2-3)

Appendix D Scenario Outline Form ES-D-1 CPNPP NRC 2013 ES-D-1 Simulator Scenario Outlines Rev c.docx 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 loss of Inverter IV1PC1. Actions are per ABN-603, Loss of Protection or Instrument Bus, Section 2.0, and include placing Rod Control in MANUAL, controlling Steam Generator (SG) level and RCS pressure, and adjusting Charging flow due to a loss of Letdown. Affected systems will be restored to normal when the Alternate Power Supply is aligned. The SRO will refer to Technical Specifications.

When Technical Specifications are referenced, the Train A Component Cooling Water (CCW) Pump will trip and the Train B CCW Pump will fail to start. The crew will enter ABN-502, Component Cooling Water System Malfunction, Section 2.0, and transfer CCW flow to Train B. The SRO will refer to Technical Specifications.

The next event is a low failure of the Volume Control Tank Level Transmitter. The crew will reference annunciator ALM-0061A-4.5, VCT LEVEL LO, and ABN-105, Chemical and Volume Control System Malfunction, and establish an Alternate Operating Mode for the Reactor Makeup System.

When plant conditions are stable, a Reactor Coolant Leak inside Containment will commence. Once it is determined that Pressurizer level cannot be maintained, the Reactor must be manually tripped and Safety Injection manually initiated. The crew will enter EOP-0.0A, Reactor Trip or Safety Injection, and then transition to EOP-1.0A, Loss of Reactor or Secondary Coolant.

The scenario is complicated by a Train A and B Containment Isolation Phase B automatic actuation failure. Additionally, the Train B Safety Injection Pump will fail to auto start upon actuation of the Safety Injection Sequencer.

While in EOP-1.0A, a Large Break Loss of Coolant Accident will occur when the Containment Spray Signal is RESET during performance of Step 7. At that point, the Unit Supervisor should recognize that entry into FRZ-0.1A, Response to High Containment Pressure, is required due to a Critical Safety Function Status Tree ORANGE path inside Containment. When the actions of FRZ-0.1A are completed a transition to FRP-0.1A, Response to Imminent Pressurized Thermal Shock Condition, is required.

FRP-0.1A will be exited at Step 1 RNO when it is determined that Reactor Coolant System pressure is less than 425 psig and Residual Heat Removal System flow is greater than 750 GPM.

The crew will return to EOP-1.0A, Loss of Reactor or Secondary Coolant, Step 8, as this is the current procedure and step in effect. This scenario is terminated when the conditions are reached for a Transfer to Cold Leg Recirculation.

Risk Significance:

Failure of risk important system prior to trip:

Loss of Protection System Inverter Loss of Component Cooling Water Risk significant core damage sequence:

Small then Large Break LOCA Risk significant operator actions:

Restore Steam Generator Level Control Restore Pressurizer Pressure Control Start Train B Safety Injection Pump Actuate Phase B Containment Isolation

Appendix D Scenario Outline Form ES-D-1 CPNPP NRC 2013 ES-D-1 Simulator Scenario Outlines Rev c.docx Facility:

CPNPP 1 & 2 Scenario No.:

4 Op Test No.:

April 2013 NRC Examiners:

Operators:

Initial Conditions: ~3% power BOL - RCS Boron is 1659 ppm by Chemistry sample. Steam Dump System in service for Reactor Coolant System Temperature Control.

Turnover: Recirculate the Refueling Water Storage Tank prior to MODE 1 entry then raise Reactor Power from 3% to 8% in preparation for Turbine Startup.

Critical Tasks:

Control Pressurizer Level to Avoid RPS or ESFAS Actuation per ABN-706, Pressurizer Level Instrument Malfunction.

Control Steam Generator Level to Avoid RPS or ESFAS Actuation per ABN-710, Steam Generator Level Instrument Malfunction.

Manually Trip Reactor Due to Reactor Protection System Failure Prior to Exiting EOP-0.0A, Reactor Trip or Safety Injection.

Trip Reactor Coolant Pumps within 10 minutes upon a Loss of Subcooling per EOP-1.0A, Loss of Reactor or Secondary Coolant, Foldout Page.

Event No. Malf. No.

Event Type*

Event Description 1

+10 min N (BOP)

Recirculate the Refueling Water Storage Tank Using Containment Spray Pump 1-01.

2

+30 min R (RO)

N (BOP, SRO)

Raise Power to 6% to 8% in Preparation for Synchronizing Main Generator to Electrical Grid.

3

+40 min RX05A I (RO, SRO)

Pressurizer Level Transmitter (LT-459A) Fails High.

4

+50 min RX04A I (BOP, SRO)

TS (SRO)

Steam Generator (1-01) Level Channel (LT-551) Fails Low.

5

+55 min CS02A TS (SRO)

Containment Spray Pump (1-01) Trip.

6

+60 min RP14B M (RO, BOP, SRO) Spurious Train B Safety Injection Actuation Signal.

7

+60 min RP01 I (RO)

Automatic Reactor Trip Failure.

8

+60 min RP09B C (BOP)

Train B Containment Isolation Phase A Automatic Actuation Failure.

9

+80 min RC08C1 M (RO, BOP, SRO) Small Break Loss of Coolant Accident Inside Containment When Safety Injection Pump 1-02 is Secured in EOS-1.1A.

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

Appendix D Scenario Outline Form ES-D-1 CPNPP NRC 2013 ES-D-1 Simulator Scenario Outlines Rev c.docx Actual Target Quantitative Attributes 7

Total malfunctions (5-8) 3 Malfunctions after EOP entry (1-2) 2 Abnormal events (2-4) 2 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)

Appendix D Scenario Outline Form ES-D-1 CPNPP NRC 2013 ES-D-1 Simulator Scenario Outlines Rev c.docx SCENARIO

SUMMARY

NRC #4 The crew will assume the watch with power at approximately 3% per IPO-003A, Power Operations.

Prior to raising power, the crew will recirculate the Refueling Water Storage Tank (RWST) using Containment Spray Pump 1-01 per SOP-204A, Containment Spray System, Section 5.1.3, Recirculation through the Recirculation Header. When the RWST recirculation is started, the crew will continue with IPO-003A, Section 5.1, Warmup and Synchronization of the Turbine Generator, Step 5.1.16, and perform a power ascension using the Rod Control and Steam Dump Systems.

When power has been raised 3% to 5%, a Pressurizer Level Channel will fail high. Response is per ABN-706, Pressurizer Level Instrumentation Malfunction, Section 2.0, and requires MANUAL control of either the Pressurizer Level Controller or the Charging Flow Controller. If response to this channel failure is not timely, Letdown will isolate. Once an Alternate Channel is selected, Pressurizer Level is restored to AUTO operation.

The next event is a Steam Generator Level Transmitter failure. Actions are per ABN-710, Steam Generator Level Instrumentation Malfunction. The BOP will be required to take manual control of the Feedwater Bypass Control Valve and then select an alternate controlling channel to return the Feedwater System to automatic control. The SRO will refer to Technical Specifications.

When conditions are stable, Containment Spray Pump 1-01 will trip. Actions are per ALM-0022A, 1-ALB-2B, Window 1.3 - ANY CSP OVRLD TRIP. The SRO will refer to Technical Specifications.

When Technical Specifications have been referenced, a spurious Train B Safety Injection Signal will actuate. The crew will determine that the Reactor did not automatically trip and initiate a Reactor Trip and Safety Injection and enter EOP-0.0A, Reactor Trip or Safety Injection. This scenario is complicated by Train B Containment Isolation Phase A Automatic actuation failure.

The crew will exit EOP-0.0A at Step 15 and enter EOS-1.1A, Safety Injection Termination. While in EOS-1.1A, Safeguards Signals are reset, the Charging flow path is realigned, and Safety Injection (SI)

Pumps are stopped. When SI Pump 1-02 is stopped, a Small Break Loss of Coolant Accident will initiate. At this point, the crew will follow guidance on the EOS-1.1A Foldout Page that requires a transition to EOP-1.0A, Loss of Reactor or Secondary Coolant.

The scenario is terminated when it is determined in EOP-1.0A that Pressurizer pressure continues to slowly lower, and a transition to EOS-1.2 A, Post LOCA Cooldown and Depressurization, is required.

Risk Significance:

Failure of risk important system prior to trip:

Containment Spray Pump Trip Risk significant core damage sequence:

Automatic Reactor Trip Failure Small Break Loss of Coolant Accident Risk significant operator actions:

Manually Trip Reactor Manually Initiate Safety Injection Initiate Train B Containment Isolation

ES-301 Transient and Event Checklist Form ES-301-5 CPNPP NRC 2013 ES-301-5 Transient and Event Checklist Rev a.docx Facility:

CPNPP 1 and 2 Date of Exam:

04/01/13 Operating Test No.:

NRC A

P P

L I

C A

N T

E V

E N

T T

Y P

E SCENARIOS CPNPP #1 CPNPP #2 CPNPP #3 CPNPP #4 T

O T

A L

MINIMUM(*)

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

SRO-U SRO-I RO RX 4

4 2

1 1

0 NOR 4

4 4

1 4

2 1,2 1

1 1

I/C 1,2,3 1,3,6 2,7 2,3 3,7 2,8 1,2,3, 4

1,3,4 1,2,6, 7

3,4 3,7 4,8 4

4 2

MAJ 5

5 5

5,6 5,6 5,6 5

5 5

6,9 6,9 6,9 2

2 1

TS 1,3,4 3,4 1,2 4,5 0

2 2

RX 1

1 0

NOR 1

1 1

I/C 4

4 2

MAJ 2

2 1

TS 0

2 2

RX 1

1 0

NOR 1

1 1

I/C 4

4 2

MAJ 2

2 1

TS 0

2 2

RX 1

1 0

NOR 1

1 1

I/C 4

4 2

MAJ 2

2 1

TS 0

2 2

RX 1

1 0

NOR 1

1 1

I/C 4

4 2

MAJ 2

2 1

TS 0

2 2

Instructions:

1.

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

2.

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

3.

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