ML14155A496

From kanterella
Jump to navigation Jump to search
02-Draft Outlines
ML14155A496
Person / Time
Site: Arkansas Nuclear  Entergy icon.png
Issue date: 02/27/2014
From: Vincent Gaddy
Operations Branch IV
To:
Entergy Operations
laura hurley
References
50-313/14-02, 50-368/14-02 50-313/OL-14, 50-368/OL-14
Download: ML14155A496 (28)
v • d • e


Text

ES-401 PWR Examination Outline FORM ES-401-2 Facility Name:Arkansas Nuclear One Unit 2 Date of Exam:2/7/2014 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. Emergency 1 4 2 3 4 4 1 18 3 3 6 Abnormal 2 2 1 2 N/A 2 1 N/A 1 9 2 2 4 Plant Evolutions 5 5 Tier Totals 6 3 5 6 5 2 27 10 1 3 2 3 4 2 2 3 4 1 2 2 28 3 2 5 2.

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

3. Generic Knowledge and Abilities 10 7 Categories 3 3 2 2 2 2 1 2 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each K/A category shall not be less than two).

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 Rev 0

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 000007 Reactor Trip / 1 1

0 Annunciators and conditions indicating signals, and remedial 1 CE/E02 Reactor Trip Recovery / 1 actions associated with the (Reactor Trip Recovery).

3.0 3

3 2 000008 Pressurizer Vapor Space Accident / 3 Inadequate core cooling 4.3 1 0

000009 Small Break LOCA / 3 0 01.

3 000011 Large Break LOCA / 3 Ability to interpret and execute procedure steps. 4.6 1 20 000015 RCP Malfunctions / 4 0 4 Natural circulation in a nuclear reactor power plant 4.4 1 000017 RCP Malfunctions (Loss of RC Flow) / 4 1 0

5 000022 Loss of Rx Coolant Makeup / 2 Relationship between charging flow and PZR level 3.0 1 3

0 6 000025 Loss of RHR System / 4 LPI pumps 3.4 1 3

0 The automatic actions (alignments) within the CCWS resulting 7 000026 Loss of Component Cooling Water / 8 from the actuation of the ESFAS 3.6 1 2

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

9 000029 ATWS / 1 M/G set power supply and reactor trip breakers 4.1 1 2

4 10 000038 Steam Gen. Tube Rupture / 3 Level operating limits for S/Gs 3.4 1 4

000040 Steam Line Rupture / 4 1

0 Adherence to appropriate procedures and operation within the 11 CE/E05 Excessive Steam Demand / 4 limitations in the Facilitys license and amendments.

3.4 2

000054 Loss of Main Feedwater / 4 1

Components, and functions of control and safety systems, including 0

12 CE/E06 Loss of Feedwater / 4 instrumentation, signals, interlocks, failure modes, and automatic and manual 3.3 1 features.

0 13 000055 Station Blackout / 6 Actions contained in EOP for loss of offsite and onsite power 4.3 1 2

4 14 000056 Loss of Off-site Power / 6 Proper operation of the ED/G load sequencer 3.8 1 7

0 ESF system panel alarm annunciators and channel status 15 000057 Loss of Vital AC Inst. Bus / 6 indicators 3.7 1 4

0 16 000058 Loss of DC Power / 6 Cross-tie of the affected dc bus with the alternate supply 3.4 1 1

000062 Loss of Nuclear Svc Water / 4 0 0 Knowing effects on plant operation of isolating certain equipment 17 000065 Loss of Instrument Air / 8 from instrument air 2.9 1 3

000077 Generator Voltage and Electric 0 18 Under-excitation 3.3 1 Grid Disturbances / 6 3 K/A Category Totals: 4 2 3 4 4 1 Group Point Total: 18 ES-401, 22 of 33 Rev 0

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 000001 Continuous Rod Withdrawal / 1 0 19 000003 Dropped Control Rod / 1 05 Reactor power - turbine power 4.1 1 000005 Inoperable/Stuck Control Rod / 1 0 20 000024 Emergency Boration / 1 04 Pumps 2.6 1 000028 Pressurizer Level Malfunction / 2 0 000032 Loss of Source Range NI / 7 0 Termination of startup following loss of intermediate-range 21 000033 Loss of Intermediate Range NI / 7 01 instrumentation 3.2 1 22 000036 Fuel Handling Accident / 8 02 SDM 3.4 1 23 000037 Steam Generator Tube Leak / 3 11 When to isolate one or more S/Gs 3.8 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

04. Knowledge of annunciator alarms, indications, or response 24 000067 Plant Fire On-site / 9 8 procedures.

4.2 1 31 Actions contained in EOP for control room evacuation emergency 25 000068 Control Room Evac. / 8 18 task 4.2 1 000069 Loss of CTMT Integrity / 5 0 26 000074 Inad. Core Cooling / 4 27 ECCS valve control switches and indicators 4.2 1 000076 High Reactor Coolant Activity / 9 0 CE/A13 Natural Circ. / 4 0 CE/A11 RCS Overcooling / 4 0 CE/A16 Excess RCS Leakage / 2 0 Normal, abnormal and emergency operating procedures 27 CE/E09 Functional Recovery 02 associated with (Functional Recovery).

3.2 1 0

0 0

0 0

0 0

0 0

K/A Category Totals: 2 1 2 2 1 1 Group Point Total: 9 ES-401, 23 of 33 Rev 0

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 0

28 003 Reactor Coolant Pump S/G 3.5 1 2

3 Relationship between temperature and pressure in CVCS 29 004 Chemical and Volume Control components during solid plant operation 3.8 1 0

0 0 30/46 005 Residual Heat Removal RHR pumps; Heatup/cooldown rates 3; 3.5 2 1 1 0 1 3.9; 31/32 006 Emergency Core Cooling 2 9 3 3.9 0

33 007 Pressurizer Relief/Quench Tank Maintaining quench tank pressure 2.7 1 2

Sources of makeup water; Ability to interpret control room indications to 0 02.

34/35 008 Component Cooling Water verify the status and operation of a system, and understand how 3; 4.2 2 5 44 operator actions and directives affect plant and system conditions.

0 36 010 Pressurizer Pressure Control Pressure detection systems 2.7 1 1

0 37 012 Reactor Protection Channel blocks and bypasses 3.6 1 3

013 Engineered Safety Features 0 38 Fuel 4.4 1 Actuation 1 0

40 022 Containment Cooling Cooling of control rod drive motors 2.8 1 4

025 Ice Condenser 0 0 0 Cooling water; Automatic swapover to containment sump suction for 4.1; 41/39 026 Containment Spray recirculation phase after LOCA (RWST low-low level alarm) 2 2 8 4.1 0 0 Increasing steam demand, its relationship to increases in 3.3; 51/42 039 Main and Reheat Steam reactor power; Emergency feedwater pump turbines 2

5 4 3.8 0 0 43/44 059 Main Feedwater S/Gs; Tripping of MFW pump turbine 3.5; 3 2 3 7 0

45 061 Auxiliary/Emergency Feedwater Decay heat sources and magnitude 3.2 1 2

0 1 Major system loads; Restoration of power to a system with a 3.3; 47/48 062 AC Electrical Distribution fault on it 2

1 2 3.2 0

49 063 DC Electrical Distribution Breaker interlocks, permissives, bypasses and cross-ties 2.9 1 2

0 50 064 Emergency Diesel Generator Fuel oil storage tanks 3.2 1 8

0 52 073 Process Radiation Monitoring Those systems served by PRMs 3.6 1 1

0 53 076 Service Water Emergency heat loads 3.7 1 2

01. Ability to locate and operate components, including local 54 078 Instrument Air controls.

4.4 1 30 0

55 103 Containment Containment pressure, temperature, and humidity 3.7 1 1

K/A Category Totals: 3 2 3 4 2 2 3 4 1 2 2 Group Point Total: 28 ES-401, 24 of 33 Rev 0

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 4 001 Control Rod Drive 0 04.

56 002 Reactor Coolant Knowledge of abnormal condition procedures. 4.0 1 11 0

57 011 Pressurizer Level Control 2 PZR heaters 3.1 1 014 Rod Position Indication 0 1

58 015 Nuclear Instrumentation 9 Heat balance 2.9 1 0

59 016 Non-nuclear Instrumentation 2 PZR LCS 3.4 1 017 In-core Temperature Monitor 0 027 Containment Iodine Removal 0 The hydrogen air concentration in excess of limit flame 028 Hydrogen Recombiner and Purge 0 60 3 propagation or detonation with resulting equipment damage in 3.4 1 Control containment 0

61 029 Containment Purge 3 Automatic purge isolation 3.2 1 033 Spent Fuel Pool Cooling 0 0

62 034 Fuel Handling Equipment 1 Radiation levels 3.3 1 035 Steam Generator 0 0

63 041 Steam Dump/Turbine Bypass Control 1 RCS T-ave. meter (cooldown rate) 3.2 1 0

64 045 Main Turbine Generator 6 RCS, during steam valve test 2.6 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 Fire header pressure 2.9 1 K/A Category Totals: 1 1 1 1 1 0 1 1 1 1 1 Group Point Total: 10 ES-401, 25 of 33 Rev 0

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 / 1 0

CE/E02 Reactor Trip Recovery / 1 000008 Pressurizer Vapor Space Accident / 3 0 0

76 000009 Small Break LOCA / 3 Whether PZR water inventory loss is imminent 4.3 1 6

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 000038 Steam Gen. Tube Rupture / 3 0

04. Knowledge of system set points, interlocks and automatic actions 77 000040 Steam Line Rupture / 4 associated with EOP entry conditions.

4.6 02 1

CE/E05 Excessive Steam Demand / 4 000054 Loss of Main Feedwater / 4 0

CE/E06 Loss of Feedwater / 4 0

78 000055 Station Blackout / 6 Existing valve positioning on a loss of instrument air system 3.7 1 1

000056 Loss of Off-site Power / 6 0 000057 Loss of Vital AC Inst. Bus / 6 0 0

79 000058 Loss of DC Power / 6 DC loads lost; impact on to operate and monitor plant systems 3.9 1 3

01.

80 000062 Loss of Nuclear Svc Water / 4 Ability to explain and apply system limits and precautions. 4.0 1 32 000065 Loss of Instrument Air / 8 0 Ability to analyze the effect of maintenance activities, such as 000077 Generator Voltage and Electric 02.

81 degraded power sources, on the status of limiting conditions for 4.2 1 Grid Disturbances / 6 36 operations.

K/A Category Totals: 0 0 0 0 3 3 Group Point Total: 6 ES-401, 22 of 33 Rev 0

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 02.

82 000005 Inoperable/Stuck Control Rod / 1 Knowledge of limiting conditions for operations and safety limits. 4.7 1 22

04. Knowledge of the emergency action level thresholds and 83 000024 Emergency Boration / 1 classifications.

4.6 1 41 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 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 000067 Plant Fire On-site / 9 8 0 000068 Control Room Evac. / 8 0 84 000069 Loss of CTMT Integrity / 5 02 Verification of automatic and manual means of restoring integrity 4.4 1 000074 Inad. Core Cooling / 4 0 000076 High Reactor Coolant Activity / 9 0 CE/A13 Natural Circ. / 4 0 Facility conditions and selection of appropriate procedures during 85 CE/A11 RCS Overcooling / 4 01 abnormal and emergency operations.

3.3 1 CE/A16 Excess RCS Leakage / 2 0 CE/E09 Functional Recovery 0 K/A Category Totals: 0 0 0 0 2 2 Group Point Total: 4 ES-401, 23 of 33 Rev 0

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 4 003 Reactor Coolant Pump 0 004 Chemical and Volume Control 0 005 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 0

86 012 Reactor Protection Incorrect channel bypassing 3.7 1 3

013 Engineered Safety Features 02. Knowledge of limiting conditions for operations and safety 87 limits.

4.7 1 Actuation 22 022 Containment Cooling 0 025 Ice Condenser 0 01.

88 026 Containment Spray Ability to interpret and execute procedure steps. 4.6 1 20 039 Main and Reheat Steam 0 059 Main Feedwater 0 0

89 061 Auxiliary/Emergency Feedwater pump failure or improper operation 3.8 1 4

062 AC Electrical Distribution 0 063 DC Electrical Distribution 0 0

90 064 Emergency Diesel Generator Failure modes of water, oil, and air valves 3.3 1 1

073 Process Radiation Monitoring 0 076 Service Water 0 078 Instrument Air 0 103 Containment 0 K/A Category Totals: 0 0 0 0 0 0 0 3 0 0 2 Group Point Total: 5 ES-401, 24 of 33 Rev 0

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 1

91 001 Control Rod Drive 8 Incorrect rod stepping sequence 3.8 1 002 Reactor Coolant 0 011 Pressurizer Level Control 0 014 Rod Position Indication 0 015 Nuclear Instrumentation 0 Ability to evaluate plant performance and make operational 01.

92 016 Non-nuclear Instrumentation judgments based on operating characteristics, reactor 4.7 1 07 behavior, and instrument interpretation.

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 0

93 034 Fuel Handling Equipment 1 Fuel protection from binding and dropping 3.4 1 035 Steam Generator 0 041 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 0 086 Fire Protection 0 K/A Category Totals: 0 0 0 1 0 0 0 1 0 0 1 Group Point Total: 3 ES-401, 25 of 33 Rev 0

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility Name:Arkansas Nuclear One Unit 2 Date of Exam:2/7/2014 RO SRO-Only Category K/A # Topic Q# IR # IR #

Ability to use procedures related to shift staffing, such as minimum crew complement, 66 2.1. 05 overtime limitations, etc. 2.9 1 67 2.1. 19 Ability to use plant computers to evaluate system or component status. 3.9 1 Knowledge of RO duties in the control room during fuel handling such as responding to alarms from the fuel handling area, 68

1. 2.1. 44 communication with the fuel storage facility, systems operated from the control room in support of fueling operations, and supporting instrumentation.

3.9 1 Conduct of 2.1.

Operations 94 2.1. 39 Knowledge of conservative decision making practices. 4.3 1 95 2.1. 01 Knowledge of conduct of operations requirements. 4.2 1 Subtotal 3 2 69 2.2. 06 Knowledge of the process for making changes to procedures. 3.0 1 70 2.2. 07 Knowledge of the process for conducting special or infrequent tests. 2.9 1 Ability to recognize system parameters that are entry-level conditions for Technical 71

2. 2.2. 42 Specifications. 3.9 1 96 Equipment 2.2. 14 Knowledge of the process for controlling equipment configuration or status. 4.3 1 Control Knowledge of the process for managing maintenance activities during power operations, such as risk assessments, work 97 2.2. 17 prioritization, and coordination with the transmission system operator. 3.8 1 2.2.

Subtotal 3 2 72 2.3. 11 Ability to control radiation releases. 3.8 1 Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor 73 2.3. 13 alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

3.4 1

3. 2.3.

98 Radiation 2.3. 14 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, 3.8 1 or emergency conditions or activities.

Control 2.3.

2.3.

Subtotal 2 1 Knowledge of the organization of the operating procedures network for normal, abnormal, 74 2.4. 05 and emergency evolutions. 3.7 1 75 2.4. 32 Knowledge of operator response to loss of all annunciators. 3.6 1

4. 2.4.

Emergency Knowledge of the bases for prioritizing emergency procedure implementation during 99 Procedures 2.4. 23 emergency operations. 4.4 1

/ Plan 100 2.4. 28 Knowledge of procedures relating to a security event. 4.1 1 2.4.

Subtotal 2 2 Tier 3 Point Total 10 7 ES-401, Page 26 of 33 Rev 0

ES-401 Record of Rejected K/As Form ES-401-4 Tier / Randomly Reason for Rejection Group Selected K/A 1/1 (RO) 0022 G 2.4.1 There are no Immediate actions (as defined by plant procedures)

QID #5 associated with Loss of Rx coolant Makeup (Loss of Charging AOP)

Selected 0022 K 1.03 as the replacement K/A 1/1 (RO) 0038 G 2.2.42 The operating Exam has a SGTR event with Tech Spec calls. Selected QID #10 0038 A 1.44 as the replacement K/A 2/1 0013 A 3.02 System over sample concerns between Tier 1 and Tier 2. Selected 026 K QID #39 4.08 as the replacement K/A 2/1 061 A 3.03 System over sample concerns. Selected 005 A1.01 as the replacement K/A QID #46 2/1 064 A 4.06 System over sample concerns. Selected 039 A 2.05 as the replacement QID #51 K/A Rev 0 ES-401, Page 27 of 33

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Arkansas Nuclear One Unit 2 Date of Examination: 02/10/2014 Examination Level: RO X SRO Operating Test Number: 2014-1 Administrative Topic Type Describe activity to be performed (see Note) Code*

Determine Boric Acid and DI water volume for make up to the Spent Fuel Pool A1. Conduct of Operations D/R ANO-2-JPM-NRC-ADMIN-SFPMU2 2.1.20 RO(4.6)

Calculate Time to Boil using computer program A2. Conduct of Operations D/P/R ANO-2-JPM-NRC-ADMIN-TTBCRO 2.1.23 RO (4.3)

Evaluate Containment atmospheric conditions.

N/R ANO-2-JPM-NRC-ADMIN-CNTMT A3. Equipment Control 2.2.12 RO (3.7)

Review emergency RWP and determine stay time based on rad levels A4. Radiation Control D/R ANO-2-JPM-NRC-ADMIN-RWP2 2.3.7 RO (3.5)

Emergency Procedures/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; 4 for SROs & RO retakes)

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

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

Revision 0

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Arkansas Nuclear One Unit 2 Date of Examination: 02/10/2014 Examination Level: RO SRO X Operating Test Number: 2014-1 Administrative Topic Type Describe activity to be performed (see Note) Code*

Review and approve calculation of volume needed to raise SFP level.

A5. Conduct of Operations D/R ANO-2-JPM-NRC-ADMIN-SFPMU 2.1.20 SRO (4.6)

Determine the Shutdown Operations protection condition and if requirements are met.

A6. Conduct of Operations N/R ANO-2-JPM-NRC-ADMIN-SOPP1 2.1.40 SRO (3.9)

Supervisory review of maintenance activities for configuration control A7. Equipment Control D/P/R ANO-2-JPM-NRC-ADMIN-MAINT 2.2.14 SRO (4.3)

Review emergency RWP and determine stay time for operators A8. Radiation Control M/R ANO-2-JPM-NRC-ADMIN-RWP3 2.3.7 SRO (3.6)

Determine EAL classification A9. Emergency M/R ANO-2-JPM-NRC-ADMIN-EAL13 Procedures/Plan 2.4.41 SRO (4.6)

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

  • Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank ( 3 for ROs; 4 for SROs & RO retakes)

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

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

Revision 0

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Arkansas Nuclear One Unit 2 Date of Examination: 02/10/2014 Exam Level: RO X SRO-I SRO-U Operating Test No.: 2014-1 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System / JPM Title Type Code* Safety Function S1. ANO-2-JPM-NRC-CNTCL 5 022 A4.03 RO-3.2/SRO-3.2 A/D/EN/L/S Containment Actuate Containment Cooling S2. ANO-2-JPM-NRC-ELEC06 6 062 A4.01 RO-3.3/SRO-3.1 A/M/S Electrical Transfer Auxiliaries from SU#2 to SU#3 for 2A-1 S3. ANO-2-JPM-NRC-CVCS2 1 004 A4.07 RO-3.9/SRO3.7 A/D/L/S Reactivity control Perform Emergency Boration S4. ANO-2-JPM-NRC-EFW01 4 Heat Removal 061 A1.01 RO-3.9/SRO4.2 D/EN/L/S Secondary Shutdown an EFW train with EFAS present S5. ANO-2-JPM-NRC-FWCS1 4 Heat Removal 035 A4.01 RO-3.7/SRO-3.6 D/S Primary Place Feedwater Control system in Automatic S6. ANO-2-JPM-NRC-CVCS12 2 004 A4.06 RO-3.6/SRO-3.1 Verify minimum letdown flow N/S Inventory Control S7. ANO-2-JPM-NRC-EOP6 7

012 A2.06 RO-4.4/SRO-4.7 A/D/S Manually trip the reactor Instrumentation S8. ANO-2-JPM-NRC-PZR01 3 010 A4.01 RO-3.7/SRO-3.5 D/S Pressure Control Equalize Pressurizer Boron In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

P1. ANO-2-JPM-NRC-PRHTR 3 068 AA1.07 RO-4.1/SRO-4.2 D/E/L Pressure Control Local operation of proportional heaters P2. ANO-2-JPM-NRC-AUADV 4 041 A4.08 RO-3.0/SRO-3.1 D/E/L/P Heat Removal Operate SDBCS valves locally P3. ANO-2-JPM-NRC-WGDTR 9

071 A2.02 RO-3.3/SRO-3.6 A/N/R Perform a Waste gas tank release Rad Control

@ 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 Revision 0

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Arkansas Nuclear One Unit 2 Date of Examination: 02/10/2014 Exam Level: RO SRO-I X SRO-U Operating Test No.: 2014-1 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System / JPM Title Type Code* Safety Function S1. ANO-2-JPM-NRC-CNTCL 5 022 A4.03 RO-3.2/SRO-3.2 A/D/EN/L/S Containment Actuate Containment Cooling S2. ANO-2-JPM-NRC-ELEC06 6 062 A4.01 RO-3.3/SRO-3.1 A/M/S Electrical Transfer Auxiliaries from SU#2 to SU#3 for 2A-1 S3. ANO-2-JPM-NRC-CVCS2 1 004 A4.07 RO-3.9/SRO3.7 A/D/L/S Reactivity control Perform Emergency Boration S4. ANO-2-JPM-NRC-EFW01 4 Heat Removal 061 A1.01 RO-3.9/SRO4.2 D/EN/L/S Secondary Shutdown an EFW train with EFAS present S5. ANO-2-JPM-NRC-FWCS1 4 Heat Removal 035 A4.01 RO-3.7/SRO-3.6 D/S Primary Place Feedwater Control system in Automatic S6. ANO-2-JPM-NRC-CVCS12 2 004 A4.06 RO-3.6/SRO-3.1 Verify minimum letdown flow N/S Inventory Control S7. ANO-2-JPM-NRC-EOP6 7

012 A2.06 RO-4.4/SRO-4.7 A/D/S Manually trip the reactor Instrumentation S8.

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

P1. ANO-2-JPM-NRC-PRHTR 3 068 AA1.07 RO-4.1/SRO-4.2 D/E/L Pressure Control Local operation of proportional heaters P2. ANO-2-JPM-NRC-AUADV 4 041 A4.08 RO-3.0/SRO-3.1 D/E/L/P Heat Removal Operate SDBCS valves locally P3. ANO-2-JPM-NRC-WGDTR 9

071 A2.02 RO-3.3/SRO-3.6 A/N/R Perform a Waste gas tank release Rad Control

@ 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 Revision 0

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Arkansas Nuclear One Unit 2 Date of Examination: 02/10/2014 Exam Level: RO SRO-I SRO-U X Operating Test No.: 2014-1 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System / JPM Title Type Code* Safety Function S1. ANO-2-JPM-NRC-CNTCL 5 022 A4.03 RO-3.2/SRO-3.2 A/D/EN/L/S Containment Actuate Containment Cooling S2. ANO-2-JPM-NRC-ELEC06 6 062 A4.01 RO-3.3/SRO-3.1 A/M/S Electrical Transfer Auxiliaries from SU#2 to SU#3 for 2A-1 S3.

S4.

S5.

S6.

S7.

S8.

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

P1. ANO-2-JPM-NRC-PRHTR 3 068 AA1.07 RO-4.1/SRO-4.2 D/E/L Pressure Control Local operation of proportional heaters P2. ANO-2-JPM-NRC-AUADV 4 041 A4.08 RO-3.0/SRO-3.1 D/E/L/P Heat Removal Operate SDBCS valves locally P3. ANO-2-JPM-NRC-WGDTR 9

071 A2.02 RO-3.3/SRO-3.6 A/N/R Perform a Waste gas tank release Rad Control

@ 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 Revision 0

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Arkansas Nuclear One Date of Exam: 2-10-14 Operating Test No.: 2014-1 A E Scenarios P V 1 2 3 T M 4 (currently P E selected as spare) O I L N T N CREW CREW CREW CREW I I T A POSITION POSITION POSITION POSITION M C L U A T S A B S A B S A B S A B M(*)

R T O R T O R T O R T O N Y R I U O C P O C P O C P O C P T P E

RO RX 4 3 1 1 1 0 NOR 1 1 2 1 1 1 SRO-I X I/C 2,3,4, 2,4,5,8 3,6,7 4,8 12 4 4 2 5,8 SRO-U MAJ 6,7 6 5 6 4 2 2 1 TS 2,4 2 0 2 2 RX 4 1 1 1 0 RO NOR 1 1 1 2 1 1 1 SRO-I 3,5 2,3,4, 2,4,6 2,3,5, I/C 11 4 4 2 X 5,7,8 7 SRO-U 6,7 6 5 6 MAJ 4 2 2 1 TS 2,5 2 0 2 2 RO RX 5 1 1 1 0 NOR 1 1 1 2 1 1 1 SRO-I X I/C 2,4,8 3,7 2,3,4, 2,3,4, 10 4 4 2 6,7 5,7,8 SRO-U MAJ 6,7 6 5 6 4 2 2 1 TS 2,4 2,3 2 0 2 2 RO RX 1 1 0 NOR 1 1 1 SRO-I I/C 4 4 2 SRO-U 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.

Revision 0 Page 3 of 3

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Arkansas Nuclear One Date of Exam: 2-10-14 Operating Test No.: 2014-1 A E Scenarios P V 1 2 3 T M 4 (currently P E selected as spare) O I L N T N CREW CREW CREW CREW I I T A POSITION POSITION POSITION POSITION M C L U A T S A B S A B S A B S A B M(*)

R T O R T O R T O R T O N Y R I U O C P O C P O C P O C P T P E

RO (1,5) RX 4 3 1 1 1 0 X 1 1 SRO-I NOR 1 1 1 1 I/C 3,5 2,4,5,8 2,4,6 4,8 6 4 4 2 SRO-U 6,7 6 5 6 MAJ 3 2 2 1 TS 0 0 2 2 RX 5 4 1 1 1 0 RO (2,6)

X NOR 1 1 1 1 1 1 SRO-I I/C 2,4,8 3,7 3,6,7 2,3,5, 5 4 4 2 7

SRO-U 6,7 6 5 6 MAJ 3 2 2 1 TS 0 0 2 2 RO (3) RX 4 3 1 1 1 0 X 1 1 SRO-I NOR 1 1 1 1 I/C 3,5 2,4,5,8 2,4,6 4,8 5 4 4 2 SRO-U MAJ 6,7 6 5 6 3 2 2 1 TS 0 0 2 2 RO (4) RX 5 4 1 1 1 0 X 1 1 SRO-I NOR 1 1 1 1 I/C 2,4,8 3,7 3,6,7 2,3,5, 6 4 4 2 7

SRO-U MAJ 6,7 6 5 6 3 2 2 1 TS 0 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.

Revision 0 Page 1 of 3

ES-301 Transient and Event Checklist Form ES-301-5 Facility: Arkansas Nuclear One Date of Exam: 2-10-14 Operating Test No.: 2014-1 A E Scenarios P V 1 2 3 T M 4 (currently P E selected as spare) O I L N T N CREW CREW CREW CREW I I T A POSITION POSITION POSITION POSITION M C L U A T S A B S A B S A B S A B M(*)

R T O R T O R T O R T O N Y R I U O C P O C P O C P O C P T P E

RO (7) RX 4 5 3 1 1 1 0 X 1 SRO-I NOR 1 1 1 1 I/C 3,5 3,7 2,4,6 4,8 5 4 4 2 SRO-U 6,7 6 5 6 MAJ 2 2 2 1 TS 0 0 2 2 RX 4 1 1 1 0 RO (8)

X NOR 1 1 1 1 1 1 1 SRO-I I/C 2,4,8 2,4,5,8 3,6,7 2,3,5, 7 4 4 2 7

SRO-U 6,7 6 5 6 MAJ 2 2 2 1 TS 0 0 2 2 RO RX 0 1 1 0 NOR 1 1 1 1 3 1 1 1 SRO-I I/C 2,3,4, 2,3,4, 2,3,4, 2,3,4, 16 4 4 2 5,8 5,7,8 6,7 5,7,8 SRO-U MAJ 6,7 6 5 6 4 2 2 1 (1,2)

X TS 2,4 2,5 2,4 2,3 6 0 2 2 RO RX 0 1 1 0 NOR 1 1 1 1 2 1 1 1 SRO-I I/C 2,3,4, 2,3,4, 2,3,4, 2,3,4, 11 4 4 2 5,8 5,7,8 6,7 5,7,8 SRO-U MAJ 6,7 6 5 6 3 2 2 1 (3)

X TS 2,4 2,5 2,4 2,3 4 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.

Revision 0 Page 2 of 3

Appendix D Scenario 1 Form ES-D-1 Facility: ANO-2 Scenario No.: 1 (New) Op-Test No.: 2014-1 Examiners: Operators:

Initial Conditions:

100% MOL; All Engineered Safety Features systems are in standby. RED Train Maintenance Week.

Loop 1 Service Water returns aligned to the Emergency Cooling Pond (ECP).

Turnover:

100%. 260 EFPD. EOOS indicates Minimal Risk. . RED Train Maintenance Week.

Evolution scheduled: Shift Loop1 Service Water returns from ECP to Lake Event Malf. No. Event Event No. Type* Description 1 N (BOP) Shift Loop 1 service water returns from ECP to Lake.

N (SRO) 2 CT2VSF1D C (BOP) 2VSF-1D Containment cooler trips. TS for SRO.

C (SRO) 3 XCVLDNHXOU I (ATC) The temperature input to the letdown HX temperature K12D01 I (SRO) controller fails Hi causing excessive cooling flow.

4 CEA43DROP R (ATC) CEA 43 fully inserts due to faulty timing card. TS for C (BOP) SRO.

C (SRO) 5 RCP2P32ALOS C (ATC) A RCP oil leak.

C (SRO) 6 MSSGBLK M (ALL) Excess steam demand inside containment on B Steam generator.

7 EFW2P7BFLT M (ALL) 2P-7B EFW pump motor fault on start, 2P-7A EFW EFW2P7ACOU pump coupling failure.

8 CV0760 C (BOP) The selected AFW flow path discharge valve will trip DO_CV_0760_1 C (SRO) its breaker requiring the other path to be used. If B EFW header is chosen 2CV-0760 will fail and if A DO_CV_0760_2 EFW header is chosen 2CV-0761 will fail.

CV0761 DO_CV_0760_1 DO_CV_0760_2

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Total malfunctions. = 7, Malfunctions after EOP entry = 1, Abnormal events = 2, Major transient = 2, EOPs with substantive actions =1, EOP Contingencies = 1, Critical tasks = 3.

Revision 0 Page 1 of 44

Appendix D Scenario 1 Form ES-D-1 Scenario #1 Objectives

1) Evaluate individual ability to control Service Water return valves.
2) Evaluate individual response to a trip of a Containment Cooling fan.
3) Evaluate individual response to a failure of a temperature input to the letdown heat exchanger and ability to manually control temperature.
4) Evaluate individual response to a CEA Malfunction.
5) Evaluate individual response to a Reactor Coolant pump oil leak (RCP emergencies).
6) Evaluate crew ability to mitigate an Excess Steam Demand.
7) Evaluate crew ability to mitigate a Loss of Feedwater.
8) Evaluate individual ability to combat events using the Functional Recovery procedure.
9) Evaluate individual ability to respond to a failure of an AFW pump discharge valve.

SCENARIO #1 NARRATIVE Simulator session begins with the plant at 100% power steady.

When the crew has completed their control room walk down and brief, the BOP will shift the Loop 1 service water return from ECP to Lake.

When the Loop 1 service water return is aligned to the lake and cued by lead examiner, 2VSF-1D containment cooling will trip. The BOP will determine that 2VSF-1D has tripped and refer to Annunciator corrective actions. The BOP will start the idle containment cooler to maintain containment temperature and pressure in the acceptable region of operation. The SRO will enter Tech Spec 3.6.2.3 action a.

After the BOP has started the idle containment cooling fan and cued by lead examiner, the temperature input to the letdown heat exchanger temperature controller will fail high. The ATC will report that the letdown heat exchanger temperature is reading high on the hand indicating controller but the computer point and control board indication are reading lower than normal. The SRO will direct the ATC to take manual control of the Letdown heat exchanger temperature control valve and manually control temperature.

After the letdown temperature controller has been placed in manual and cued by the lead examiner, CEA 43 will drop into the core. The SRO will enter the CEA malfunction AOP, OP 2203.003. The SRO should check that less than 2 CEAs are inserted and then commence a down power within 15 minutes. The BOP should complete attachment C DNBR/LPD log. The SRO will enter Tech Specs for CEA position (3.1.3.1d) and Aztilt (3.2.3).

After the crew has completed the required reactivity manipulation, entered the appropriate tech specs, and cued by the lead examiner, A RCP oil leak will start which cause oil level to lower and bearing temperatures to raise. The ATC should trip the reactor and secure the A RCP. The crew may elect to secure a RCP in the B S/G loop to balance flows.

Revision 0 Page 2 of 44

Appendix D Scenario 1 Form ES-D-1 SCENARIO #1 NARRATIVE (continued)

The Crew will implement Standard Post Trip Actions (SPTA), OP 2202.001. After the reactor trips a Main Steam line break inside contentment will cause an Excess Steam Demand. Main Steam Isolation (MSIS) and Containment Spray (CSAS) will actuate tripping Main Feedwater pumps, Condensate pumps, AFW pump, closing the MSIVs and feedwater block valves. The 2P-7B EFW pump motor will fail and 2P-7A coupling will break causing a loss of feedwater event. The ATC will secure all the Reactor Coolant pumps due to the Containment Spray actuation.

The SRO will diagnose an Excess Steam Demand and Loss of Feedwater event and enter OP 2202.009, Functional Recovery. The crew will maintain post blowdown temperature and pressure of the RCS to prevent pressurized thermal shock. The BOP will steam A S/G using the upstream Atmospheric Dump valve when B S/G blows dry. The ATC should use Auxiliary Spray to maintain RCS pressure. The Crew will restore Feedwater from the AFW pump 2P-75 after removing the MSIS and CSAS trip. The selected feed path from AFW will trip its breaker when the valve is opened requiring use of the alternate flow path.

Revision 0 Page 3 of 44

Appendix D Scenario 2 Form ES-D-1 Facility: ANO-2 Scenario No.: 2 (New) Op-Test No.: 2014-1 Examiners: Operators:

Initial Conditions:

~40 % due to elevated S/G Chloride. MOL. All Engineered Safety Features systems are in standby. Hold power ~ 40 % until S/G Chloride less than 10 ppb. C channel Excore has failed and PPS points 1 through 4 are in bypass. RED Train Maintenance Week.

Turnover:

RED Train Maintenance Week. 260 EFPD. EOOS indicates Minimal Risk. Hold power ~ 40 % until S/G Chloride less than 10 ppb. SG blowdown ~120 gpm per SG for cleanup. Reactor Engineering is developing reactivity plan for power escalation. C channel Excore has failed and PPS points 1 through 4 are in bypass and all required actions are complete.

Evolution scheduled: Swap component cooling water pumps from 2P-33C to 2P-33B for maintenance on 2P-33C.

Event Malf. No. Event Event No. Type* Description 1 N (BOP) Swap running CCW pumps from 2P-33C to 2P-33B.

N (SRO) 2 NIBUPPER C (BOP) B channel Excore upper chamber fails high. TS for C (SRO) SRO.

3 XRCCHBPCNT I (ATC) B Pressurizer control channel pressure fails high.

I (SRO) 4 CCW2P33BPWR C (BOP) 2P-33B CCW pump trips and 2P-33C CCW pump fails CCW2P33CPWR C (SRO) to re-start.

5 RCP2P32CSLK R (ATC) C RCP develops an intersystem LOCA from the RCS C (BOP) to CCW of 15 gpm. TS for SRO.

C (SRO) 6 RCP2P32CSLK M (All) C RCP intersystem LOCA degrades to 250 gpm. If ESFK202AAF crew does not isolate CCW to RCPs then when CIAS actuates CCW to RCP valves fail to close.

ESFK202BAF 7 RCSHTRON C (ATC) Pressurizer Backup Heaters fail to de-energize on low C (SRO) pressurizer level.

8 CV0231 C (BOP) Gland seal regulator 2PCV-0231 fails closed.

C (SRO)

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Total malfunctions. = 7, Malfunctions after EOP entry = 2, Abnormal events = 4, Major transient = 1, EOPs with substantive actions =1, EOP Contingencies = 0, Critical tasks = 3.

Revision 0 Page 1 of 34

Appendix D Scenario 2 Form ES-D-1 Scenario #2 Objectives

1) Evaluate individual ability to swap running Component Cooling water pumps.
2) Evaluate individual response to a failure of a Nuclear Instrument.
3) Evaluate individual response to a Pressurizer System Malfunction (Pressure channel failure).
4) Evaluate individual response to a failure of a Component Cooling water pump.
5) Evaluate individual response to an intersystem Loss of Coolant Accident. (LOCA)
6) Evaluate crew ability to mitigate an intersystem LOCA.
7) Evaluate individual response to failure of a gland seal regulator.
8) Evaluate individual response to a failure of the pressurizer backup heaters to de-energize on low level.

SCENARIO #2 NARRATIVE Simulator session begins with the plant at 100% power steady.

When the crew has completed their control room walk down and brief, the BOP will swap CCW pumps placing 2P-33B in service.

When 2P-33B CCW pump has been placed in service and cued by lead examiner; Channel B Excore upper chamber will fail high. The SRO will enter the NI malfunction AOP and the crew should determine that B channel linear power is failed but log power is still function by monitoring output for the three chambers. The SRO will also enter Tech Spec 3.3.1.1 action 3 for Reactor Protection System. The BOP will trip points 1, 3, and 4 on channel B by using the linear calibrate switch.

After the BOP has tripped points 1, 3, and 4, and cued by lead examiner, the B pressurizer pressure control channel will fail high causing the spray valves to open and RCS pressure to lower.

The CRS should enter the 2203.028 Pressurizer System malfunction AOP. The crew will place the other pressurizer pressure controller in service, verify that both spray valves close, and the pressurizer heaters restore RCS pressure. The BOP will place a maximum of one Steam Dump and Bypass Control System (SDBCS) valve permissive in manual and all other permissives to off.

When the SDBCS permissives have been aligned and cued by the lead examiner, 2P-33B will trip and 2P-33C will fail to start automatically or manually. The SRO will enter the RCP emergencies AOP, OP 2203.025. The BOP should call NLOs to investigate the CCW pump trip. The SRO should direct the BOP to start 2P-33C but it will fail to start. The SRO will then direct opening all CCW cross-tie valves and start 2P-33A.

Revision 0 Page 2 of 34

Appendix D Scenario 2 Form ES-D-1 SCENARIO #2 NARRATIVE (continued)

After the crew has restored CCW flow to the RCPs, and cued by the lead examiner, A 15 gpm RCS to CCW leak will start. The crew should notice that CCW Surge Tank level is rising. Also the CCW letdown radiation monitor will alarm indicating RCS to CCW leakage. The SRO will enter the Excess RCS leakage, 2203.016 and direct the board operator actions. The crew should perform leak rates, isolate letdown to verify the leak is not in letdown and determine the need for a plant shutdown using normal boration. The SRO should enter Attachment A, align the CCW surge tanks to the gas collection header and direct the AO to control surge tank level. The crew will perform a power reduction such that the plant will be taken off line. The SRO should enter Tech Spec 3.4.6.2 for RCS leakage. The ATC will borate the RCS and reduce turbine load to maintain Tave-Tref within 2°F. The BOP will make preparations to remove secondary plant equipment from service as power is reduced.

After the required reactivity manipulations are complete and cued by the lead examiner, the RCS to CCW will degrade to 250 gpm. The SRO will direct the reactor to be tripped, actuate SIAS &

CCAS, secure RCPs, and isolate CCW to the RCPs. The SRO should enter and direct the actions of SPTAs.

The crew will implement Standard Post Trip Actions (SPTA), OP 2202.001. The ATC should recognize that the pressurizer backup heaters failed to de-energize on low pressurizer level. Also, the crew should place the SDBCS master controller in Auto Local and lower the set point to maintain margin to saturation.

The SRO will diagnose a LOCA and enter OP 2202.003, Loss of Coolant Accident. After the crew has entered the LOCA EOP, 2PCV-0231 gland seal pressure control valve will fail closed. The BOP will manually control 2CV-0233 gland seal bypass valve to maintain gland seal header pressure and condenser vacuum. The crew will commence a cooldown to allow depressurization and refilling the pressurizer. The BOP will restore Service Water to Component Cooling Water and Auxiliary Cooling water.

Revision 0 Page 3 of 34

Appendix D Scenario 3 Form ES-D-1 Facility: ANO-2 Scenario No.: 3 (New) Op-Test No.: 2014-1 Examiners: Operators:

Initial Conditions:

98% MOL; All Engineered Safety Features systems are in standby. RED Train Maintenance Week.

Mabelvale transmission line out of service and Unit 2 output is limited to 1035 MW gross, 995 MW net.

Turnover:

260 EFPD. 98%. Mabelvale transmission line out of service and Unit 2 output is limited to 1035 MW gross, 995 MW net. Reactor engineering is developing reactivity plans. EOOS indicates Minimal Risk. RED Train Maintenance Week.

Evolution scheduled: Shift running vacuum pumps.

Event Malf. No. Event Event No. Type* Description 1 N (BOP) Shift running vacuum pumps.

N (SRO) 2 DO_HS_8259_G C (BOP) 2RITS-8271-2 coupling fails and 2RITS-8231-1 CV82591 C (SRO) particulate detector fails high when started. Tech Spec for SRO.

XRI2RITS8231A DO_RITS8231_10 3 XRRPZRLSP I (ATC) Reactor Reg. output to PZR level control program fails I (SRO) to 41%.

4 MFWPMPBTRP R (ATC) B main feed water pump trips due to thrust bearing C (BOP) wear. TS (Tcold out of range high) for SRO.

C (SRO) 5 SGBTUBE M (ALL) B Steam Generator Tube Rupture ramps up to 300 gpm over 20 min. Manual reactor trip criteria when exceeds 44 gpm. TS for SRO 6 ESFSIAS2 C (ATC) Green Train SIAS fails to actuate and letdown isolation CV48211 C (BOP) 2CV-4821-1 fails open.

C (SRO) 7 CV0302 C (ATC) Steam dump turbine bypass valve being used for the CV0303 C (SRO) cooldown will fail closed during the RCS cooldown.

CV0306 8

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Total malfunctions. = 6, Malfunctions after EOP entry = 2, Abnormal events = 3, Major transient = 1, EOPs with substantive actions =1, EOP Contingencies = 0, Critical tasks = 3.

Revision 0 Page 1 of 34

Appendix D Scenario 3 Form ES-D-1 Scenario #3 Objectives

1) Evaluate individual ability to perform a vacuum pump swap.
2) Evaluate individual response to a failure of a Containment Air monitor.
3) Evaluate individual response to a failure of a Containment Air monitor radiation monitor.
4) Evaluate individual response to a Pressurizer system malfunction involving pressurizer level failing high.
5) Evaluate individual response to a failure of loss of main feedwater pump.
6) Evaluate crews ability to mitigate a Steam Generator Tube Rupture.
7) Evaluate individual response to Green Train SIAS failure to actuate.
8) Evaluate individual response to a failure of letdown to automatically isolate.
9) Evaluate individual response to a steam dump turbine bypass valve failing closed.

SCENARIO #3 NARRATIVE Simulator session begins with the plant at ~98% power.

When the crew has completed their control room walk down and brief, they will shift running vacuum pumps.

When the vacuum pumps have been shifted or when cued by the lead examiner, the in-service CAMS unit coupling will fail and when the back up CAMS unit is started the particulate detector will fail high. The ATC should report the 2K-11 H10 CNTMT Air Monitor trouble alarm and refer to the ACA. The BOP should investigate and determine that 2RITS-8271-2 has low flow. When contacted, the NLO will report the coupling has failed. The BOP should use OP-2104.033 ventilation system operations to place the standby CAMS unit in service. When the standby CAMS unit is placed in service, the particulate detector will fail requiring entry into Tech Spec 3.4.6.1 When all actions due to the CAMS failure have been completed, and cued by the lead examiner, the reactor reg pressurizer level program output will fail to minimum (41%). The SRO will enter the PZR system malfunctions AOP, OP 2203.028. The ATC will take manual control of letdown to control pressurizer level. The ATC must take control of PZR heaters to control RCS pressure (All heaters will be energized) The ATC should place the PZR level controller to Auto and Local then adjust the setpoint to programmed setpoint. Then Letdown should be placed back in automatic.

This failure will also prevent manual start of back up charging pumps if needed to control PZR level.

When letdown has been restored to automatic and cued by the lead examiner, or cued by lead examiner, B MFWP will trip. The SRO will enter and implement Loss of Main feedwater pump AOP, 2203.027. This will result in Steam Flow exceeding Feed flow and SG levels dropping. The crew will manually and rapidly reduce turbine load, insert group 6 and group P CEAs, borate using emergency boration to the RCS until Feed flow is greater than Steam Flow. Then continue with a normal boration power reduction to less than 80%.

After the crew has restored feedwater flow greater than steam flow or cued by lead examiner, a Steam Generator Tube Rupture (SGTR) will occur on B Steam Generator. The SRO will enter primary to secondary leakage AOP, OP 2203.038. The crew should determine that the leak rate is greater than 44 gpm. They will trip the reactor, actuate SIAS, and CCAS. The SRO should enter TS 3.4.6.2.

Revision 0 Page 2 of 34

Appendix D Scenario 3 Form ES-D-1 SCENARIO #3 NARRATIVE (continued)

The Crew will implement Standard Post Trip Actions (SPTA) EOP, 2202.001. When SIAS is actuated the green train components will fail to reposition. The crew should recognize the failure of green train SIAS to actuate. The BOP should have a NLO to check the breaker and pump motors for the green train High Pressure Safety Injection (HPSI) and Low Pressure Safety Injection pumps (LPSI) pumps. After the NLO report, the BOP should manually start 2P-89B HPSI pump and open all injection valves. Also, 2CV-4821-1 red train letdown isolation valve will fail to close leaving letdown aligned. The ATC should recognize that letdown is aligned and close a green train isolation (that failed to actuate) to help maintain RCS inventory. The crew will align service water to CCW to maintain forced circulation. The crew may lower Steam Dump master controller setpoint during SPTAs to aid in maintaining margin to saturation.

The SRO will diagnose and enter the Steam Generator Tube Rupture (SGTR) EOP. The ATC should commence cool down of the RCS to allow isolation of B steam generator. The BOP will override Service Water to Auxiliary Cooling Water to maintain condenser vacuum. During the cooldown, 2CV-0303, 2CV-0302, or 2CV-0306 (depending on which is being used) will fail closed impacting the cooldown rate the ATC will notice that the cooldown has stopped and adjust the cooldown rate to ensure the steam generator is isolated within the 30 minute required time. Once Thot is less than 535 degrees F, the BOP should isolate B steam generator.

Revision 0 Page 3 of 34

Retrieved from "https://kanterella.com/w/index.php?title=ML14155A496&oldid=2452458"