ML14155A502

2014-02-Final Outlines

ML14155A502
Person / Time
Site:	Arkansas Nuclear
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: ML14155A502 (28)

Text

Revision 2 ES-401 PWR Examination Outline FORM ES-401-2 Facility Name:Arkansas Nuclear One Unit 2 Date of Exam:2/21/2014 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 4 3 4 1 18 3 3 6 Abnormal 2 2 1 2 N/A 2 1 N/A 1 9 3 1 4 Plant Evolutions 6 4 Tier Totals 6 3 6 5 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 0 1 1 2 1 10 1 1 1 3 Plant Systems Tier Totals 4 3 4 5 3 2 3 5 2 4 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 Revision 2

Revision 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)

K K K A A 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 CE/E02 Reactor Trip Recovery / 1 3.0 3 actions associated with the (Reactor Trip Recovery).

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

000009 Small Break LOCA / 3 0 01.

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

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

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

0 The automatic actions (alignments) within the CCWS resulting 000026 Loss of Component Cooling Water / 8 3.6 1 2 from the actuation of the ESFAS 000027 Pressurizer Pressure Control System 0 Controllers and positioners 2.6 1 Malfunction / 3 3 1

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

4 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 CE/E05 Excessive Steam Demand / 4 3.4 2 limitations in the Facilitys license and amendments.

000054 Loss of Main Feedwater / 4 1

Components, and functions of control and safety systems, 0

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

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

4 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 000057 Loss of Vital AC Inst. Bus / 6 3.7 1 4 indicators 0

000058 Loss of DC Power / 6 Actions contained in EOP for loss of dc power 4.0 1 2

000062 Loss of Nuclear Svc Water / 4 0 0 Knowing effects on plant operation of isolating certain equipment 000065 Loss of Instrument Air / 8 2.9 1 3 from instrument air 000077 Generator Voltage and Electric 0 Under-excitation 3.3 1 Grid Disturbances / 6 3 K/A Category Totals: 4 2 4 3 4 1 Group Point Total: 18 ES-401, 22 of 33 Revision 2

Revision ES-4012 Form ES-401-2 3

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

K K K A A 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 05 Reactor power - turbine power 4.1 1 000005 Inoperable/Stuck Control Rod / 1 0 000024 Emergency Boration / 1 04 Pumps 2.6 1 000028 Pressurizer Level Malfunction / 2 0 000032 Loss of Source Range NI / 7 01 Startup termination on source-range loss 3.2 1 000033 Loss of Intermediate Range NI / 7 0 000036 Fuel Handling Accident / 8 02 SDM 3.4 1 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 000067 Plant Fire On-site / 9 8 procedures.

4.2 1 31 Actions contained in EOP for control room evacuation emergency 000068 Control Room Evac. / 8 18 task 4.2 1 000069 Loss of CTMT Integrity / 5 0 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 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 Revision 2 ES-401, 23 of 33

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

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

1 2 3 4 5 6 1 2 3 4 0

003 Reactor Coolant Pump S/G 3.5 1 2

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

0 0 005 Residual Heat Removal RHR pumps; Heatup/cooldown rates 3; 3.5 2 1 1 0 1 Valve positioning on safety injection signal; Inadvertent SIS 3.9; 006 Emergency Core Cooling actuation 2

9 3 3.9 0

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

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

008 Component Cooling Water indications to verify the status and operation of a system, 3; 4.2 2 5 44 and understand how operator actions and directives affect 0

010 Pressurizer Pressure Control Pressure detection systems 2.7 1 1

0 012 Reactor Protection Channel blocks and bypasses 3.6 1 3

013 Engineered Safety Features 0 Fuel 4.4 1 Actuation 1 0

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

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

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

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; 062 AC Electrical Distribution fault on it 2

1 2 3.2 0

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

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

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

0 076 Service Water Emergency heat loads 3.7 1 2

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

4.4 1 30 0

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

Revision 2 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 System # / Name G K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 001 Control Rod Drive 0 04.

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

011 Pressurizer Level Control 2

PZR heaters 3.1 1 014 Rod Position Indication 0 1

015 Nuclear Instrumentation 9

Heat balance 2.9 1 0

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 0

028 Hydrogen Recombiner and Purge Control 3

propagation or detonation with resulting equipment damage in 3.4 1 containment 0

029 Containment Purge 3

Automatic purge isolation 3.2 1 033 Spent Fuel Pool Cooling 0 0

034 Fuel Handling Equipment 1

Radiation levels 3.3 1 035 Steam Generator 0 0

041 Steam Dump/Turbine Bypass Control 1

RCS T-ave. meter (cooldown rate) 3.2 1 0

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

086 Fire Protection Deluge valves 3.0 1 5

K/A Category Totals: 1 1 1 1 1 0 0 1 1 2 1 Group Point Total: 10 ES-401, 25 of 33 Revision 2

Revision 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 (SRO)

K K K A A 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

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 000040 Steam Line Rupture / 4 4.6 02 associated with EOP entry conditions.

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

CE/E06 Loss of Feedwater / 4 0

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

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

01.

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.

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 Revision 2

Revision 2 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 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.

000005 Inoperable/Stuck Control Rod / 1 Knowledge of limiting conditions for operations and safety limits. 4.7 1 22 Whether boron flow and/or MOVs are malfunctioning, from plant 000024 Emergency Boration / 1 01 conditions 4.1 1 000028 Pressurizer Level Malfunction / 2 0 000032 Loss of Source Range NI / 7 0 000033 Loss of Intermediate Range NI / 7 0 000036 Fuel Handling Accident / 8 0 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 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 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 3 1 Group Point Total: 4 Revision 2 ES-401, 23 of 33

Revision 2 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 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

012 Reactor Protection Incorrect channel bypassing 3.7 1 3

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

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

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

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

062 AC Electrical Distribution 0 063 DC Electrical Distribution 0 0

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 Revision 2

Revision 2 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 System # / Name G K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 1

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.

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 Control 0 029 Containment Purge 0 033 Spent Fuel Pool Cooling 0 0

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 Revision 2

Revision 2 ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility Name:Arkansas Nuclear One Unit 2 Date of Exam:2/21/2014 RO SRO-Only Category K/A # Topic IR # IR #

Ability to evaluate plant performance and make operational judgments based on operating 2.1. 07 characteristics, reactor behavior, and instrument interpretation. 2.9 1 2.1. 21 Ability to verify the controlled procedure copy. 3.9 1 Knowledge of RO duties in the control room during fuel handling such as responding to alarms from the

1. 2.1. 44 fuel handling area, communication with the fuel storage facility, systems operated from the control room 3.9 1 in support of fueling operations, and supporting instrumentation.

Conduct of 2.1.

Operations 2.1. 01 Knowledge of conduct of operations requirements. 4.2 1 2.1. 39 Knowledge of conservative decision making practices. 4.3 1 Subtotal 3 2 2.2. 06 Knowledge of the process for making changes to procedures. 3.0 1 2.2. 07 Knowledge of the process for conducting special or infrequent tests. 2.9 1

2. 2.2. 43 Knowledge of the process used to track inoperable alarms. 3.0 1 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, 2.2. 17 such as risk assessments, work prioritization, and coordination with the transmission 3.8 1 system operator.

2.2.

Subtotal 3 2 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 2.3. 13 radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked 3.4 1 high-radiation areas, aligning filters, etc.

3. 2.3.

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, 2.4. 05 and emergency evolutions. 3.7 1 2.4. 32 Knowledge of operator response to loss of all annunciators. 3.6 1

4. 2.4.

Emergency Procedures 2.4. 14 Knowledge of general guidelines for EOP usage. 4.5 1

/ Plan 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 Revision 2

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

T1/G1 0022 G2.4.1 associated with Loss of Reactor Coolant Makeup (Loss of Charging AOP)

QID # 5 Selected 0022 K1.03 as the replacement K/A RO Overlap with Operating exam (there is a SGTR event with Tech Spec Calls)

T1/G1 0038 G2.2.42 Selected 0038 A1.44 as the replacement K/A QID #10 RO 058 A1.01 Question does not apply. Unit does not have the ability to cross tie Vital DC T1/G1 buses. Selected 058 AK3.02 as the replacement K/A QID #16 RO 033 AK 3.01 Startup Channels serve the purpose of Intermediate range instrumentation T1/G2 during a reactor startup. Selected 032 AK3.01 as the replacement K/A QID #21 RO System over sample concerns between Tier 1 and Tier 2. Selected 026 T2/G1 0013 A3.02 K4.08 as the replacement K/A QID #39 RO System over sample concerns. Selected 005 A1.01 as the replacement K/A T2/G1 061 A3.03 QID #46 RO System over sample concerns. Selected 039 A2.05 as the replacement K/A T2/G1 064 A4.06 QID #51 RO Unit 2 does not control the stations fire pumps or have indications in the T2/G2 086 A1.01 control room for the fire water system. The fire water system is operated by QID #65 Unit 1. Selected 086 A 4.05 as the replacement K/A RO Rejected original G2.1.5 due to being an SRO duty. Selected G2.1.7 as the T3 G2.1.5 replacement K/A QID #66 RO Does not lend itself to a generic question (directs monitoring plant T3 G2.1.19 components or systems). Selected G2.1.21 as the replacement K/A QID #67 RO Does not lend itself to a generic question (specific system parameters that T3 G2.2.42 are entry level conditions for Tech Specs). Selected G2.2.43 as the QID #71 replacement K/A SRO There is not an E-Plan associated with Emergency Boration. Selected 024 T1/G2 024 G2.4.41 AA2.01 as the replacement K/A QID #83 SRO Difficultly of matching K/A. Selected G2.4.14 as the replacement K/A T3 G2.4.23 QID #99 Rev 2 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*

Spent Fuel Pool Makeup Calculation 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 Perform Evolution 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 1

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

Determine Shutdown Operations Protection Plan Condition 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 A8. Radiation Control M/R ANO-2-JPM-NRC-ADMIN-RWP3 2.3.7 SRO (3.6)

EOF Evacuation Determination A9. Emergency N/R ANO-2-JPM-NRC-ADMIN-EOFEVAC Procedures/Plan 2.4.38 SRO (4.4)

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

• 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 2

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 Verify Containment Coolers in Emergency Mode 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 4 061 A1.01 RO-3.9/SRO4.2 D/EN/L/S Heat Removal Shutdown EFW Train A with EFAS Signal Present Secondary S5. ANO-2-JPM-NRC-FWCS1 4 035 A4.01 RO-3.7/SRO-3.6 Heat Removal Place Feedwater Control system in Automatic D/S Primary S6. ANO-2-JPM-NRC-CVCS12 2 004 A4.06 RO-3.6/SRO-3.1 Verification of 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 Instrumentation Manually Trip the Reactor S8. ANO-2-JPM-NRC-PZR01 3 010 A4.01 RO-3.7/SRO-3.5 D/S Pressure Control Equalize RCS and 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 Perform Local Operations of the Proportional Heaters Pressure Control P2. ANO-2-JPM-NRC-EDDCS 6 064 A4.01 RO-4.0/SRO-4.3 D/E/L Electrical Startup Diesel Generator Without DC Control Power (2K-4A)

P3. ANO-2-JPM-NRC-WGDTR 9 071 A2.02 RO-3.3/SRO-3.6 A/N/R Rad Control Perform Waste Gas Decay Tank Release

@ 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 2

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 Verify Containment Coolers in Emergency Mode 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 4 061 A1.01 RO-3.9/SRO4.2 D/EN/L/S Heat Removal Shutdown EFW Train A with EFAS Signal Present Secondary S5. ANO-2-JPM-NRC-FWCS1 4 035 A4.01 RO-3.7/SRO-3.6 Heat Removal Place Feedwater Control system in Automatic D/S Primary S6. ANO-2-JPM-NRC-CVCS12 2 004 A4.06 RO-3.6/SRO-3.1 Verification of 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 Instrumentation Manually Trip the Reactor 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 Perform Local Operations of the Proportional Heaters Pressure Control P2. ANO-2-JPM-NRC-EDDCS 6 064 A4.01 RO-4.0/SRO-4.3 D/E/L Electrical Startup Diesel Generator Without DC Control Power (2K-4A)

P3. ANO-2-JPM-NRC-WGDTR 9 071 A2.02 RO-3.3/SRO-3.6 A/N/R Rad Control Perform Waste Gas Decay Tank Release

@ 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 2

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 Verify Containment Coolers in Emergency Mode S2.

S3.

S4. ANO-2-JPM-NRC-EFW01 4 061 A1.01 RO-3.9/SRO4.2 D/EN/L/S Heat Removal Shutdown EFW Train A with EFAS Signal Present Secondary 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 Perform Local Operations of the Proportional Heaters Pressure Control P2. ANO-2-JPM-NRC-EDDCS 6 064 A4.01 RO-4.0/SRO-4.3 D/E/L Electrical Startup Diesel Generator Without DC Control Power (2K-4A)

P3. ANO-2-JPM-NRC-WGDTR 9 071 A2.02 RO-3.3/SRO-3.6 A/N/R Rad Control Perform Waste Gas Decay Tank Release

@ 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 2

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%, 260 EFPD. RED Train Maintenance Week.

Turnover:

EOOS indicates Minimal Risk.

Evolution scheduled: Shift Control Element Drive Mechanism (CEDM) fans from 2VSF-35D to 2VSF-35C IAW 2104.033 starting with 10.6.

Event Malf. No. Event Type* Event No. Description 1 N (BOP) Shift Control Elements Drive Mechanism (CEDM) fans.

N (SRO) OP-2104.033, Containment Atmosphere Control.

2 XCVLDNHXOU I (ATC) The temperature input to the letdown HX temperature K12D01 I (SRO) controller (2TIC-4815) fails Hi.

OP-2203.012L, Annuciator 2K-12 Corrective Action (ACA) 3 CT2VSF1D C (BOP) 2VSF-1D Containment cooler trips. TS for SRO.

C (SRO) OP-2203.012D/E, 2K-04 and 2K05 ACAs TS (SRO) 4 CEA43DROP R (ATC) CEA 43 fully inserts. TS for SRO.

C (BOP) OP-2203.003, CEA Malfunction AOP C (SRO)

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

C (SRO) OP-2203.025, RCP Emergencies AOP 6 MSSGBLK M (ALL) Excess Steam Demand inside containment on B SG.

OP-2202.001, Standard Post Trip Actions (SPTAs) EOP and OP-2202.009, Functional Recovery EOP.

7 CV4652 C (ATC) B RCP normal spray valve fails open.

C (SRO) OP-2202.010, Standard Attachments EOP or OP-2203.0028, Pressurizer System Malfunction AOP 8 EFW2P7BFLT M (ALL) 2P-7B EFW pump motor fault on start, 2P-7A EFW pump EFW2P7ACOU coupling failure. OP-2202.009, Functional Recovery EOP.

9 CV0760 C (BOP) The selected AFW flow path discharge valve (2CV-0760 or DO_CV_0760_1 C (SRO) 2CV-0761) breaker trip.

DO_CV_0760_2 OP-2202.010, Standard Attachments EOP.

CV0761 DO_CV_0760_1 DO_CV_0760_2 End Feedwater is restored to A SG.

point

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Revision 3 Page 1 of 51

Appendix D Scenario 1 Form ES-D-1 Target Quantitative Attributes (Section D.5.d) Actual Attributes Total Malfunctions (5-8) 8 Malfunctions after EOP entry (1-2) 2 Abnormal Events (2-4) 2 Major Transients (1-2) 2 EOPs entered requiring substantive actions (1-2) 1 EOP contingencies requiring substantive actions (0-1) 1 Critical Tasks (2-3) 3 Critical Task Justification References A RCP must be secured within Exceeding operating limits has 1015.050 Time Critical 10 min of the reactor trip. the potential to degrade the RCS Operation action program, pressure boundary. RCPs should Attachment C be maintained in an available CE EPGB Simulator CTs:

condition for last-resort use if CT-23, Trip any RCP exceeding needed. operating limits (ESDE-03, FRG-04)

Stabilize and control RCS If RCS heatup is allowed after CE EPGB Simulator CTs:

temperature after the ESD SG blowdown, the RCS could CT-07, Establish RCS blowdown terminates. Maintain over pressurize and result in temperature Control (SPTA-07, RCS pressure within the lifting PZR and SG safeties. ESDE-05, HR-05)

Pressure-Temperature limits of These pressure stresses added to 200°F and 30°F Margin to thermal stresses of rapid Saturation throughout cooldown could present PTS implementation of SPTAs and concerns.

Functional Recovery EOP.

Restore Feedwater prior to both Inventory in the unaffected SG is CE EPGB Simulator CTs:

SG levels reaching 70 wide required to remove decay heat CT-08, Establish RCS Heat range. from the reactor core (core melt Removal (ESDE-08, HR-01) potential). EOP 2202.009 Functional Recovery EOP 2202.006 Loss of Feedwater EOP Tech Guide Scenario #1 Objectives

1) Evaluate individual ability to transfer CEDM fans.

2) Evaluate individual response to a failure of a temperature input to the letdown heat exchanger and ability to manually control temperature.

3) Evaluate individual response to a trip of a Containment Cooling fan.

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.

10) Evaluate individual ability to respond to RCP spray valve failing open.

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Appendix D Scenario 1 Form ES-D-1 SCENARIO #1 NARRATIVE Simulator session begins with the plant at 100% power steady state.

When the crew has completed their control room walk down and brief, the BOP will shift Control Elements Drive Mechanism (CEDM) fans from 2VSF-35D to 2VSF-35C.

When the CEDM fans have been shifted or cued by lead examiner, the temperature input (2TE-4815) 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 due to excessive cooling flow. The SRO will direct the ATC to take manual control of the Letdown heat exchanger temperature control valve and manually control temperature for the duration of the scenario.

After the letdown temperature controller has been placed in manual and cued by the lead examiner, 2VSF-1D containment cooler will trip. The BOP will determine that 2VSF-1D containment cooler has tripped and refer to OP-2203.012D/E, 2K04 and 2K05 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. [Site OE: CR-ANO-2-2006-2444, 2VSF-1A motor failure and breaker trip.]

After the BOP has started the idle containment cooling fan and cued by lead examiner, CEA 43 will drop into the core due to faulty timing card. The SRO will enter OP-2203.003, CEA malfunction AOP. 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 (3.1.3.1 Action d) and Aztilt (3.2.3). [Site and industry OE: CR-ANO-2-2007-0127 dropped CEA, and NRC Event # 49601 Palo Verde dropped CEA.]

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 that causes oil level to lower and bearing temperatures to rise. The CRS will enter OP-2203.025, RCP Emergencies AOP. The crew will monitor the A RCP oil level trend and bearing temperatures. After bearing temperatures begin to rise (trip criteria >180F/min.) 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. Securing a RCP not satisfying operating limits is a time critical operator action per OP-1015.050 Time Critical Operator Action Program. [Site OE: RCP oil leaks CR-ANO-2-2013-1602, CR-ANO-2-2013-587, CR-ANO-2-2013-58.]

The crew will implement OP-2202.001, Standard Post Trip Actions (SPTA) EOP. After the reactor trips a Main Steam line break (B SG) inside containment 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 to start and 2P-7A EFW pump coupling will break causing a loss of feedwater event. The ATC will secure all the Reactor Coolant pumps due to the Containment Spray actuation. When the B RCP spray valve (2CV-4652) handswitch is placed in manual, the valve will fail open. The ATC must recognize this and isolate the spray valve using the associated block valve. If the spray valve is not isolated, the ATCs ability to control RCS pressure will be limited. [Industry OE for Excess Steam Demand, SOER 82-7, Reator Vessel Pressurized Thermal Shock.]

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

After completing SPTAs, The SRO will diagnose an Excess Steam Demand and Loss of Feedwater event and enter OP-2202.009, Functional Recovery EOP. 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 valve from AFW will trip its breaker when the valve is opened requiring use of the alternate flow path. [Loss of feedwater events industry OE: SOER 86-01 Reliability of PWR Auxiliary feedwater systems, and PRA operator action # 3 Establish flow to SGs from AFW to the SGs given a los of both EFW and MFW flow to the SGs.]

Revision 3 Page 4 of 51

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 %. MOL. C channel Excore has failed and PPS points 1 through 4 are in bypass. RED Train Maintenance Week. B Component Cooling Water CCW pump in service.

Turnover:

EOOS indicates Minimal Risk. Hold power 39- 41 % 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 (TS 3.3.1.1 action 2 entered).

Evolution scheduled: Perform Red Train Proportional Heater test starting with step 2.1.

Event Malf. No. Event Type* Event No. Description 1 N (ATC) Perform Red Train Proportional Heater test.

N (SRO) OP-2103.005 Pressurizer Operations.

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

C (SRO) OP-2203.026, NI malfunction AOP.

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

I (BOP) OP-2203.028, Pressurizer System Malfunction AOP I (SRO) 4 CCW2P33BPWR C (BOP) 2P-33B CCW pump trips and 2P-33C CCW pump fails to CCW2P33CPWR C (SRO) start. OP-2203.025, RCP Emergencies AOP 5 RCP2P32CSLK R (ATC) C Reactor Coolant Pump (RCP) develops an intersystem N (BOP) LOCA from the RCS to CCW of 15 gpm. TS for SRO.

N (SRO) OP-2203.016, Excess RCS leakage AOP TS (SRO) 6 RCP2P32CSLK M (All) C RCP intersystem LOCA degrades to 250 gpm. CCW to ESFK202AAF RCPs fail to auto close on CIAS.

ESFK202BAF OP-2202.001, Standard Post Trip Actions (SPTA), and OP-2202.003, Loss of Coolant Accident EOP 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) 2203.012B, Annuciator 2K-02 Corrective Action (ACA)

End CCW to RCP has been isolated, a RCS cooldown has been point started and condenser vacuum maintain by operation of 2CV-0233 gland seal regulator.

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Revision 3 Page 1 of 41

Appendix D Scenario 2 Form ES-D-1 Target Quantitative Attributes (Section D.5.d) Actual Attributes Total Malfunctions (5-8) 7 Malfunctions after EOP entry (1-2) 2 Abnormal Events (2-4) 4 Major Transients (1-2) 1 EOPs entered requiring substantive actions (1-2) 1 EOP contingencies requiring substantive actions (0-1) 0 Critical Tasks (2-3) 3 Critical Task Justification Isolate RCS from leaving Isolating CCW to Containment CE EPGB Simulator CTs: CT-Containment by closing CCW to prevents release of radioactivity 09, Establish Containment containment valves within 10 by bypassing Containment. Isolation (LOCA-07) min. of the Reactor Trip. Failure to establish a containment 10CFR20 boundary could result in violating 10CFR100 exposure limits.

Commence an RCS cooldown Cooling down and depressurizing CE EPGB Simulator CTs: CT-within 30 minutes of entry into the RCS removes decay heat and 20, Cool down and depressurize OP-2202.003, LOCA EOP. lowers the DP at the break, RCS (LOCA-09) slowing the leak rate and reducing CR-ANO-2-2010-948, Critical makeup volume required. SDC task times entry conditions are also required for long-term cooling.

Establish RCS pressure control to Once RCS subcooling is lost, CE EPGB Simulator CTs: CT-maintain RCS subcooling. PZR level is no longer a valid 06, Establish RCS Pressure Maintain pressure and indication of RCS inventory. A Control (LOCA-12) temperature within the PT limits reactor head void can form, and if of <2000 F and >300F MTS left uncontrolled, could result in throughout implementation of core uncovery and fuel damage.

OP-2202.003, LOCA EOP.

Scenario #2 Objectives

1) Evaluate individual ability to perform Proportional heater surveillance.

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 40% power steady.

When the crew has completed their control room walk down and brief, they will perform the Red Train Proportional Heater surveillance.

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

When the Red Train Proportional Heater has been placed to auto or when cued by the lead examiner, Channel B Excore upper chamber will fail high. The SRO will enter the OP-2203.026, NI Malfunction AOP and the crew should determine that B channel linear power is failed but log power is still functional 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. The points must be tripped because Channel C is in bypass. [Site OE: CR-ANO-2-2002-693, D Excore failure.]

When the SDBCS permissives have been aligned and cued by the 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 OP-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. [Site OE: CR-ANO-2-2011-1605, Pressurizer pressure failing high.]

After the BOP has tripped points 1, 3, and 4, and cued by lead examiner, 2P-33B CCW pump will trip and 2P-33C CCW pump will fail to start automatically or manually. The SRO will enter OP-2203.025, RCP Emergencies AOP. The BOP should call NLOs to investigate the CCW pump trip.

The SRO should direct the BOP to start 2P-33C CCW pump but it will fail to start. The SRO will then direct opening all CCW cross-tie valves and start 2P-33A CCW pump. [Site OE: CR-ANO 2007-313, Trip of 2P-33B CCW pump with 2P-33C out of service for maintenance.]

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. The crews recognition of the leak may be delayed because the B Surge Tank level would normally rise from the different pump configuration. Also the CCW letdown radiation monitor will alarm indicating RCS to CCW leakage. The SRO will enter OP-2203.016, Excess RCS Leakage AOP, 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 of Excess RCS Leakage, align the CCW surge tanks to the gas collection header and direct the NLO 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 Action a 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. [Industry OE: NRC information notice 92-36 Intersystem LOCA outside containment. Industry OE: SEN-220, SEN-216, & SEN-182, RCS leakage events.]

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 CCW to RCPs valves will fail to auto close on a valid CIAS. The SRO should enter and direct the actions of SPTAs.

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

The crew will implement OP-2202.001, Standard Post Trip Actions (SPTA) EOP. 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 and enter OP-2202.003, Loss of Coolant Accident EOP. After the crew has entered the LOCA EOP and cued by the lead examiner, 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. [Site OE: for 2PCV-0231 gland seal pressure control valve CR-ANO-2-2009-719, CR-ANO-2-2009-311, and CR-ANO-2-2006-1406.]

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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; RED Train Maintenance Week.

Turnover:

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

EOOS indicates Minimal Risk.

Evolution scheduled: Shift running vacuum pumps.

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

N (SRO) OP-2106.010 Condenser Vacuum System.

2 XRRPZRLSP I (ATC) Reactor Reg. output to PZR level control program fails I (SRO) to 41%. OP-2203.028, Pressurizer System Malfunction AOP 3 DO_HS_8259_G C (BOP) 2RITS-8271-2 Containment Atmosphere Monitor CV82591 C (SRO) (CAMS) coupling fails and 2RITS-8231-1 CAMS particulate detector fails. TS for SRO.

XRI2RITS8231A TS (SRO)

OP-2203.012J, Annunciator 2K-10 Corrective Action DO_RITS8231_10 (ACA), and OP-2203.012K, 2K-11 ACA 4 MFWPMPBTRP R (ATC) B Main Feed Water pump trips. TS (Tcold out of C (BOP) range high) for SRO.

OP-2203.027, Loss of Main Feedwater pump AOP C (SRO)

TS (SRO) 5 SGBTUBE M (ALL) B Steam Generator Tube Rupture ramps up to 300 TS (SRO) gpm over 20 min. TS for SRO OP-2203.038, Primary to Secondary leakage AOP, OP-2201.001 Standard Post Trip Actions EOP, and 2202.004 Steam Generator Tube Rupture EOP 6 ESFSIAS2 C (ATC) Green Train SIAS fails to actuate and letdown isolation CV48211 C (BOP) 2CV-4821-1 fails open.

C (SRO) OP-2202.010 Standard Attachments EOP 7 CV0302 C (ATC) Steam dump turbine bypass valve fails closed.

CV0303 C (SRO) OP-2105.008, SDBCS operations CV0306 End B Steam Generator is isolated.

Point

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Revision 3 Page 1 of 44

Appendix D Scenario 3 Form ES-D-1 Target Quantitative Attributes (Section D.5.d) Actual Attributes Total Malfunctions (5-8) 6 Malfunctions after EOP entry (1-2) 2 Abnormal Events (2-4) 3 Major Transients (1-2) 1 EOPs entered requiring substantive actions (1-2) 1 EOP contingencies requiring substantive actions (0-1) 0 Critical Tasks (2-3) 2 Critical Task Justification Perform one or more of the Once RCS subcooling is lost, CE EPGB Simulator CTs:

following to maintain/restore PZR level is no longer a valid CT-06, Establish RCS Pressure Margin to Saturation (MTS) > 30 indication of RCS inventory. A Control (SGTR-10) degrees F. reactor head void can form, and if 1015.050 Time Critical Start the Green train HPSI left uncontrolled, could result in Operation Actions, pump and open HPSI injection core uncovery and fuel damage. Attachment C valve(s)

Isolate letdown RCP operating limits require Adjust RCS Cooldown rate MTS to be >300F.

MTS must be restored >30 degrees F within 10 min.

Conduct an RCS cooldown to Reduce Thot below 5350F is CE EPGB Simulator CTs:

Thot <535 0F and isolate B SG necessary to prevent a MSSV CT-20, Cooldown and (2202.010 Attachment 10 from lifting (5350F), thus depressurize RCS (SGTR-05) completed) within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after the preventing an offsite release and CT-14, Isolate most affected SG Reactor trip. exceeding 10CFR100 exposure (SGTR-09).

limits at the site boundary. SAR Section 15.1.18 Assumption is that the operator 1015.050 Time Critical will diagnose within 30 minutes Operation Actions, and then isolate within next 30 Attachment C minutes after entry into 2202.004, EOP 2202.004, SGTR Tech SGTR EOP Guide 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 sample pump.

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.

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Appendix D Scenario 3 Form ES-D-1 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 Reactor Reg pressurizer level program output will fail to minimum (41%). The SRO will enter the OP-2203.028, PZR System Malfunctions AOP. 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 or cued by lead examiner, the in-service Containment Air Monitor System (CAMS) unit 2RITS-8271-2 coupling will fail. 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 Action a. [Site OE: CR-ANO-2-2013-1880, CAMS particulate detector failure, CR-ANO-2-2011-2691, for CAMS unit low air flow and CR-ANO 2006-1191, for sample motor failure.]

When all actions due to the CAMS failure have been completed, or cued by the lead examiner, B MFWP will trip. The SRO will enter and implement OP-2203.027, Loss of Main Feedwater Pump AOP. This will result in steam flow exceeding feed flow and SG levels lowering. 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. The SRO will be required to enter Tech Spec 3.2.6 for Tc out of range high. Then start a normal boration power reduction to less than 80%. [Industry OE: INPO event # OE31445 Loss of a Main feedwater pump.

Site OE: CR-ANO-2-2009-3744, B Main Feedwater pump trip.]

After the crew has restored feedwater flow greater than steam flow or cued by lead examiner, a Steam Generator Tube Leak will occur on B Steam Generator. The SRO will enter OP 2203.038, Primary to Secondary Leakage AOP. The SRO should enter TS 3.4.6.2 Action a, RCS leakage, and TS 3.7.1.2 for EFW when steam is isolated to 2P-7A EFW pump. When the leak rate exceeds 44 gpm, the crew should determine that the leak rate is greater than 44 gpm. They will trip the reactor, actuate SIAS, and CCAS. [Industry OE: SOER 83-2, Steam Generator Tube Ruptures. Steam Generator Tube Rupture response is a time critical operator action per OP-1015.050 Time Critical Operator action program.]

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

The Crew will implement OP-2202.001, Standard Post Trip Actions (SPTA) EOP. 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 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 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. [Industry OE: SOER 83-9, Valve inoperability cause by motor operator failures for 2CV-4821-1.]

The SRO will diagnose and enter OP-2202.004, 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 Steam dump valve (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.

[Site OE: CR-ANO 2010-558, CR-ANO 2009-3780, CR-ANO 2008-1190, Steam dump valve failure to open.]

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