ML14140A144

04-FINAL Outlines

ML14140A144
Person / Time
Site:	Waterford
Issue date:	04/02/2014
From:	Chris Steely Operations Branch IV
To:	Entergy Operations
laura hurley
References
50-382/14-004
Download: ML14140A144 (112)

Text

ES-401 PWR Examination Outline Form ES-401-2 Facility: Waterford 3 (RO Exam Outline) Date of Exam: April 2, 2014 RO K/A Category Points SRO-Only Points Tier Group K K K K K K A A A A G A2 G* Total 1 2 3 4 5 6 1 2 3 4

• Total

1. 1 3 3 3 3 3 3 18 6 Emergency &

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

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

3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 3 2 2 3 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each K/A category shall not be less than two).

2. The point total for each group and tier in the proposed outline must match that specified in the table.

The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions.

The final RO exam must total 75 points and the SRO-only exam must total 25 points.

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

4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.

5. Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected.

Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

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

7.* The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As.

8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams.

9. For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.

2014 NRC Revision 1

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

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

1 2 3 1 2 EK2.1 - Knowledge of the interrelations between the 000007 (BW/E02&E10; CE/E02) Reactor X (Reactor Trip Recovery) and the following: Components, 3.3 1 Trip - Stabilization - Recovery / 1 and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

AA2.01 - Ability to determine and interpret the following 000008 Pressurizer Vapor Space X as they apply to the Pressurizer Vapor Space Accident: 3.9 2 Accident / 3 RCS pressure and temperature indicators and alarms EK2.03 - Knowledge of the interrelations between the 000009 Small Break LOCA / 3 X small break LOCA and the following: S/Gs 3.0 3 2.4.20 -Knowledge of the operational implications of 000011 Large Break LOCA / 3 X EOP warnings, cautions, and notes. 3.8 4 AK3.04 - Knowledge of the reasons for the following 000015/17 RCP Malfunctions / 4 X responses as they apply to the Reactor Coolant Pump 3.1 5 Malfunctions (Loss of RC Flow): Reduction of power to below the steady state power- to-flow limit AA1.01 -Ability to operate and / or monitor the following 000022 Loss of Rx Coolant Makeup / 2 X as they apply to l the Loss of Reactor Coolant Makeup: 3.4 6 CVCS letdown and charging AK1.01 - Knowledge of the operational implications of 000025 Loss of RHR System / 4 X the following concepts as they apply to Loss of Residual 3.9 7 Heat Removal System: Loss of RHRS during all modes of operation 000026 Loss of Component Cooling X AK3.03 Knowledge of the reasons for the following 4.0 16 Water / 8 responses as they apply to the Loss of Component Cooling Water: Guidance actions contained in EOP for Loss of CCW AA2.15 - Ability to determine and interpret the following 000027 Pressurizer Pressure Control X as they apply to the Pressurizer Pressure Control 3.7 8 System Malfunction / 3 Malfunctions: Actions to be taken if PZR pressure instrument fails high EA1.12 - Ability to operate and monitor the following as 000029 ATWS / 1 X they apply to a ATWS: M/G set power supply and reactor 4.1 9 trip breakers EK1.02 - Knowledge of the operational implications of 000038 Steam Gen. Tube Rupture / 3 X the following concepts as they apply to the SGTR: Leak 3.2 10 rate vs. pressure drop 2.1.27 - Knowledge of system purpose and/or function.

000040 (BW/E05; CE/E05; W/E12) X 3.9 11 Steam Line Rupture - Excessive Heat Transfer / 4 EK2.1 - Knowledge of the interrelations between the 000054 (CE/E06) Loss of Main X (Loss of Feedwater) and the following: Components, and 3.3 12 Feedwater / 4 functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features EK1.01 - Knowledge of the operational implications of 000055 Station Blackout / 6 X the following concepts as they apply to the Station 3.3 13 Blackout : Effect of battery discharge rates on capacity AA2.40 - Ability to determine and interpret the following 000056 Loss of Off-site Power / 6 X as they apply to the Loss of Offsite Power: Service water 3.3 14 pump ammeter and flowmeter 2014 NRC Revision 1

AK3.01 - Knowledge of the reasons for the following 000057 Loss of Vital AC Inst. Bus / 6 X responses as they apply to the Loss of Vital AC 4.1 15 Instrument Bus: Actions contained in EOP for loss of vital ac electrical instrument bus 000058 Loss of DC Power / 6 2.2.36 - Ability to analyze the effect of maintenance 000062 Loss of Nuclear Svc Water / 4 X activities, such as degraded power sources, on the status 3.1 17 of limiting conditions for operations.

AA1.01 - Ability to operate and / or monitor the 000065 Loss of Instrument Air / 8 X following as they apply to the Loss of Instrument Air: 2.7 18 Remote manual loaders W/E04 LOCA Outside Containment / 3 W/E11 Loss of Emergency Coolant Recirc. / 4 BW/E04; W/E05 Inadequate Heat Transfer - Loss of Secondary Heat Sink / 4 000077 Generator Voltage and Electric Grid Disturbances / 6 K/A Category Totals: 3 3 3 3 3 3 Group Point Total: 18/6 2014 NRC Revision 1

ES-401 3 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (RO / SRO)

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

1 2 3 1 2 000001 Continuous Rod Withdrawal / 1 AA1.06 - Ability to operate and / or monitor the 000003 Dropped Control Rod / 1 X 4.0 19 following as they apply to the Dropped Control Rod: RCS pressure and temperature AA2.03 - Ability to determine and interpret the 000005 Inoperable/Stuck Control Rod / 1 X 3.5 20 following as they apply to the Inoperable / Stuck Control Rod: Required actions if more than one rod is stuck or inoperable 000024 Emergency Boration / 1 000028 Pressurizer Level Malfunction / 2 2.1.25 - Ability to interpret reference materials, 000032 Loss of Source Range NI / 7 X 3.9 21 such as graphs, curves, tables, etc.

000033 Loss of Intermediate Range NI / 7 000036 (BW/A08) Fuel Handling Accident / 8 AA2.11 - Ability to determine and interpret the 000037 Steam Generator Tube Leak / 3 X 3.8 22 following as they apply to the Steam Generator Tube Leak: When to isolate one or more S/Gs 000051 Loss of Condenser Vacuum / 4 000059 Accidental Liquid RadWaste Rel. / 9 000060 Accidental Gaseous Radwaste Rel. / 9 AK1.01 - Knowledge of the operational 000061 ARM System Alarms / 7 X 2.5 23 implications of the following concepts as they apply to Area Radiation Monitoring (ARM)

System Alarms: Detector limitations 000067 Plant Fire On-site / 8 000068 (BW/A06) Control Room Evac. / 8 000069 (W/E14) Loss of CTMT Integrity / 5 EK2.09 - Knowledge of the interrelations 000074 (W/E06&E07) Inad. Core Cooling / 4 X 2.6 24 between the and the following Inadequate Core Cooling: Controllers and positioners AK3.05 - Knowledge of the reasons for the 000076 High Reactor Coolant Activity / 9 X 2.9 25 following responses as they apply to the High Reactor Coolant Activity : Corrective actions as a result of high fission-product radioactivity level in the RCS W/EO1 & E02 Rediagnosis & SI Termination / 3 W/E13 Steam Generator Over-pressure / 4 W/E15 Containment Flooding / 5 W/E16 High Containment Radiation / 9 BW/A01 Plant Runback / 1 BW/A02&A03 Loss of NNI-X/Y / 7 BW/A04 Turbine Trip / 4 BW/A05 Emergency Diesel Actuation / 6 BW/A07 Flooding / 8 BW/E03 Inadequate Subcooling Margin / 4 2014 NRC Revision 1

BW/E08; W/E03 LOCA Cooldown - Depress. / 4 BW/E09; CE/A13; W/E09&E10 Natural Circ. / 4 BW/E13&E14 EOP Rules and Enclosures CE/A11; W/E08 RCS Overcooling - PTS / 4 AK2.2 - Knowledge of the interrelations between CE/A16 Excess RCS Leakage / 2 X 3.0 26 the (Excess RCS Leakage) and the following:

Facilitys heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.

EA1.3 -Ability to operate and / or monitor the CE/E09 Functional Recovery X 3.6 27 following as they apply to the (Functional Recovery): Desired operating results during abnormal and emergency situations.

K/A Category Point Totals: 1 2 1 2 2 1 Group Point Total: 9/4 2014 NRC Revision 1

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

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

1 2 3 4 5 6 1 2 3 4 K6.02 - Knowledge of the effect of a loss or 003 Reactor Coolant Pump X malfunction on the following will have on 2.7 28 the RCPS: RCP seals and seal water supply A2.18 - Ability to (a) predict the impacts of 004 Chemical and Volume X the following malfunctions or operations on 3.1 29 Control the CVCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: High VCT level K6.03 - Knowledge of the effect of a loss or 005 Residual Heat Removal X malfunction on the following will have on 2.5 30 the RHRS: RHR heat exchanger K3.02 - Knowledge of the effect that a loss 006 Emergency Core Cooling X X or malfunction of the ECCS will have on the 4.3 31 following: Fuel K4.16 - Knowledge of ECCS design 3.2 32 feature(s) and/or interlock(s) which provide for the following: Interlocks between RHR valves and RCS K1.03 - Knowledge of the physical 007 Pressurizer Relief/Quench X X connections and/or cause-effect relationships 3.0 33 Tank between the PRTS and the following systems: RCS K4.01 - Knowledge of PRTS design feature(s) and/or interlock(s) which provide 2.6 34 for the following: Quench tank cooling K3.01 - Knowledge of the effect that a loss 008 Component Cooling Water X X or malfunction of the CCWS will have on 3.4 35 the following: Loads cooled by CCWS A1.02 - Ability to predict and/or monitor 2.9 36 changes in parameters (to prevent exceeding design limits) associated with operating the CCWS controls including: CCW temperature A2.01 - Ability to (a) predict the impacts of 010 Pressurizer Pressure Control X the following malfunctions or operations on 3.3 37 the PZR PCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Heater failures K5.01 - Knowledge of the operational 012 Reactor Protection X implications of the following concepts as the 3.3 38 apply to the RPS: DNB 013 Engineered Safety Features X A4.03 - Ability to manually operate and/or 4.5 39 Actuation monitor in the control room: ESFAS initiation 022 Containment Cooling X X K2.01 - Knowledge of power supplies to the 3.0 40 following: Containment cooling fans A3.01 - Ability to monitor automatic operation of the CCS, including: Initiation of 4.1 41 safeguards mode of operation 2014 NRC Revision 1

025 Ice Condenser This system is not part of plant design, but included in random selection process.

A1.03 - Ability to predict and/or monitor 026 Containment Spray X changes in parameters (to prevent exceeding 3.5 42 design limits) associated with operating the CSS controls including: Containment sump level K4.04 - Knowledge of MRSS design 039 Main and Reheat Steam X feature(s) and/or interlock(s) which provide 2.9 43 for the following: Utilization of steam pressure program control when steam dumping through atmospheric relief/dump valves, including T-ave. limits A3.03 - Ability to monitor automatic 059 Main Feedwater X X operation of the MFW, including: Feedwater 2.5 44 pump suction flow pressure 2.1.28 - Knowledge of the purpose and 4.1 45 function of major system components and controls.

A1.05 -Ability to predict and/or monitor 061 Auxiliary/Emergency X changes in parameters (to prevent exceeding 3.6 46 Feedwater design limits) associated with operating the AFW controls including: AFW flow/motor amps A4.01 - Ability to manually operate and/or 062 AC Electrical Distribution X monitor in the control room: All breakers 3.3 47 (including available switchyard) 2.4.8 - Knowledge of how abnormal 063 DC Electrical Distribution X operating procedures are used in conjunction 3.8 48 with EOPs.

K2.03 - Knowledge of bus power supplies to 064 Emergency Diesel Generator X the following: Control power 3.2 49 K5.03 - Knowledge of the operational 073 Process Radiation X implications as they apply to concepts as 2.9 50 Monitoring they apply to the PRM system: Relationship between radiation intensity and exposure limits K1.20 - Knowledge of the physical 076 Service Water X X connections and/or cause- effect 3.4 51 relationships between the SWS and the following systems: AFW K3.07 - Knowledge of the effect that a loss or malfunction of the SWS will have on the 3.7 52 following: ESF loads K1.02 - Knowledge of the physical 078 Instrument Air X X connections and/or cause-effect relationships 2.7 53 between the IAS and the following systems:

Service air A4.01 - Ability to manually operate and/or monitor in the control room: Pressure gauges 3.1 54 A3.01 - Ability to monitor automatic 103 Containment X operation of the containment system, 3.9 55 including: Containment isolation K/A Category Point Totals: 3 2 3 3 2 2 3 2 3 3 2 Group Point Total: 28/5 2014 NRC Revision 1

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

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

1 2 3 4 5 6 1 2 3 4 K4.05 - Knowledge of CRDS design feature(s) 001 Control Rod Drive X and/or interlock(s) which provide for the 3.9 56 following: Boration and dilution.

002 Reactor Coolant A3.02 - Ability to monitor automatic operation 011 Pressurizer Level Control X of the PZR LCS, including: Reactor power 2.6 57 014 Rod Position Indication K6.03 - Knowledge of the effect of a loss or 015 Nuclear Instrumentation X malfunction on the following will have on the 2.6 58 NIS: Component interconnections K5.01 - Knowledge of the operational 016 Non-nuclear Instrumentation X implication of the following concepts as they 2.7 59 apply to the NNIS: Separation of control and protection circuits K1.02 - Knowledge of the physical connections 017 In-core Temperature Monitor X and/or cause-effect relationships between the 3.3 60 ITM system and the following systems: RCS 027 Containment Iodine Removal A1.02 - Ability to predict and/or monitor 028 Hydrogen Recombiner X changes in parameter (to prevent exceeding 3.4 61 and Purge Control design limits) associated with operating the HRPS controls including: Containment pressure K3.01 - Knowledge of the effect that a loss or 029 Containment Purge X malfunction of the Containment Purge System 2.9 62 will have on the following: Containment parameters 033 Spent Fuel Pool Cooling 034 Fuel Handling Equipment Randomly selected this system for a RO question. Accounted for in SRO exam per ES-401-2 instructions.

035 Steam Generator 041 Steam Dump/Turbine Bypass Control 045 Main Turbine Generator 055 Condenser Air Removal 056 Condensate 068 Liquid Radwaste A2.02 - Ability to (a) predict the impacts of the 071 Waste Gas Disposal X following malfunctions or operations on the 3.3 63 Waste Gas Disposal System ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Use of waste gas release monitors, radiation, gas flow rate, and totalizer 072 Area Radiation Monitoring 2014 NRC Revision 1

K2.03 - Knowledge of bus power supplies to 075 Circulating Water X the following: Emergency/essential SWS 2.6 64 pumps 079 Station Air A4.02 - Ability to manually operate and/or 086 Fire Protection X monitor in the control room: Fire Detection 3.5 65 Panels K/A Category Point Totals: 1 1 1 1 1 1 1 1 1 1 0 Group Point Total: 10/3 2014 NRC Revision 1

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Waterford 3 (RO Exam Outline) Date of Exam: April 2, 2014 Category K/A # Topic RO SRO-Only IR # IR #

Knowledge of individual licensed operator responsibilities related to 2.1.4 shift staffing, such as medical requirements, no-solo operation, 3.3 66 maintenance of active license status, 10CFR55, etc.

1. Ability to coordinate personnel activities outside the control room.

Conduct 2.1.8 3.4 67 of Operations Knowledge of RO duties in the control room during fuel handling, such 2.1.44 as responding to alarms from the fuel handling area, communication 3.9 68 with the fuel storage facility, systems operated from the control room in support of fueling operations, and supporting instrumentation.

2.1.

Subtotal 3 Ability to perform pre-startup procedures for the facility, including 2.2.1 operating those controls associated with plant equipment that could 4.5 69 affect reactivity.

Knowledge of pre- and post-maintenance operability requirements.

2. 2.2.21 2.9 70 Equipment 2.2.

Control Subtotal 2 Ability to use radiation monitoring systems, such as fixed radiation 2.3.5 monitors and alarms, portable survey instruments, personnel monitoring 2.9 71 equipment, etc.

Knowledge of radiation monitoring systems, such as fixed radiation

3. 2.3.15 monitors and alarms, portable survey instruments, personnel monitoring 2.9 72 Radiation equipment, etc.

Control 2.3.

Subtotal 2 Knowledge of the organization of the operating procedures network for 2.4.5 normal, abnormal, and emergency evolutions. 3.7 73 Knowledge of low power/shutdown implications in accident (e.g., loss

4. 2.4.9 of coolant accident or loss of residual heat removal) mitigation 3.8 74 Emergency strategies.

Procedures / Knowledge of EOP implementation hierarchy and coordination with Plan 2.4.16 other support procedures or guidelines such as, operating procedures, 3.5 75 abnormal operating procedures, and severe accident management guidelines.

2.4.

Subtotal 3 Tier 3 Point Total 10 7 2014 NRC Revision 1

ES-401 Record of Rejected K/As Form ES-401-4 Tier / Randomly Reason for Rejection Group Selected K/A 2/1 025 Ice Condenser This system was selected for a question in the RO exam. Since this system is not part of the Waterford 3 design, another system was randomly selected for the question.

1/1 022 Loss of Reactor Originally selected AK2 area for a question. None of the K/A Coolant Makeup statements had an Importance Rating above 2.5. Reselected 026 Loss of Component Cooling Water in the AK2 area. Reselected an AA1 area question for 022 Loss of Reactor Coolant Makeup.

1/1 026 Loss of Reselected an AK2 area for a question (see previous item). There Component Cooling are no specific KA statements provided. Reselected AK2 question Water (CCW) under 009 Small Break LOCA.

1/1 065 Loss of Originally selected an AK2 area for the question. There are no KA Instrument Air statements that are above a 2.5 Importance Rating. Reselected an EK2 KA statement under E06 Loss of Main Feedwater. Reselected an AA1 KA statement for 065 Loss of Instrument Air.

2/1 073 Process Originally selected an A3 area for the question. There are no KA Radiation statements available. Reselected an A3 KA statement under 103 Monitoring Containment. Reselected a K5 statement under 073 Process Radiation Monitoring.

2/2 055 Condenser Air Originally selected an A2 area for the question. There are no KA Removal statements that are above a 2.5 Importance Rating. Reselected the A2 area under 071 Waste Gas Disposal.

2/2 086 Fire Protection Originally selected a K2 area for the question. There are no KA statements available. Reselected the K2 KA statement under 075 Circulating Water. Reselected the A4 KA statement under 086 Fire Protection.

2/1 076 Service Water Originally selected K1.15 for Service Water. There is no System relationship at Waterford 3 between TCCW/ACC and the FPS.

Randomly selected another system for the K1 Statement for 076 Service Water. The system selected was K1.20, AFW.

2/2 086 Fire Protection Originally selected A4.06 for the Halon System. There is no Halon System. Systems at Waterford 3. Randomly selected another K/A for the A4 Statement for 086, Fire Protection System. The K/A selected was A4.02, Fire Detection Panels.

1/1 029 Anticipated Originally selected EA1.06 for Operating switches for normal Transient Without charging header isolation valves. Waterford 3 does not operate Scram (ATWS) charging isolation valves for an ATWS event. Randomly selected another K/A for the EA1 Statement for 029 Anticipated Transient Without Scram (ATWS). The K/A selected was EA1.12, M/G set power supply and Reactor Trip Breakers 2014 NRC Revision 1

1/1 026 Loss of Originally selected 000058 AK3.02, Loss of DC power. It was Component Cooling determined that there were too many questions already in the RO exam for Water a loss of DC. This system was oversampled. Randomly selected another K/A for Tier 1 Group 1. The K/A selected is AK3.03 for 026, Loss of Component Cooling Water.

1/2 076 High Reactor Originally selected 076 AK3.06, Actions contained in the EOP for Coolant Activity high RCS activity. W-3 has no actions in the EOP for high RCS Activity, W-3 has only AOPs for this event. Randomly selected another AK3 K/A from 076. The K/A selected is 076 AK 3.05 2/1 006 Emergency Originally selected 006 K4.25. Knowledge of interlocks associated Core Cooling with concentrated boric acid supply to the RWST. At W-3, the valves used to fill the RWSP with boric acid have no interlocks. .

Randomly selected another K4 K/A from 006. The K/A selected is 006 K4.16.

2/1 008 Component Originally selected 008 K3.02. Knowledge of the effect that a loss Cooling Water of CCWS will have on CRDS. There is no procedural requirements at W-3 for a loss of CCW to CRDS. Randomly selected another K3 K/A from 008. The K/A selected is 008 K3.01.

1/1 011 Large Break Originally selected 011 G2.4.11. There is nothing in Waterford 3 LOCA AOPs about a Large Break LOCA. Randomly re-selected in the G2.4 field for a subject in EOP space for Waterford 3. Selected G2.4.20. Knowledge of the operational implications of EOP warnings, cautions and notes.

1/2 037 AA2.11 Originally selected 000037 AA2.11. The original question to this K/A was unsat on NRC review and it was determined that an RO question was not able to be written to this K/A. Could not develop a question for the magnitude of the release. Randomly selected a K/A for 037 in the AA2 field. Randomly selected AA2.10 for 037.

2014 NRC Revision 1

ES-401 PWR Examination Outline Form ES-401-2 Facility: Waterford 3 (SRO Exam Outline) Date of Exam: April 2, 2014 RO K/A Category Points SRO-Only Points Tier Group K K K K K K A A A A G A2 G* Total 1 2 3 4 5 6 1 2 3 4

• Total

1. 1 18 3 3 6 Emergency &

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

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

3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 2 2 2 1 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each K/A category shall not be less than two).

2. The point total for each group and tier in the proposed outline must match that specified in the table.

The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions.

The final RO exam must total 75 points and the SRO-only exam must total 25 points.

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

4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.

5. Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected.

Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

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

7.* The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As.

8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams.

9. For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.

2014 NRC Revision 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 / SRO)

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

1 2 3 1 2 2.4.4 - Ability to recognize abnormal indications for 000007 (BW/E02&E10; CE/E02) Reactor X system operating parameters that are entry-level 4.7 SRO Trip - Stabilization - Recovery / 1 conditions for emergency and abnormal operating 1 procedures.

000008 Pressurizer Vapor Space Accident / 3 2.2.40 - Ability to apply Technical Specifications for a 000009 Small Break LOCA / 3 X system. 4.7 SRO 2

EA2.04 - Ability to determine or interpret the following 000011 Large Break LOCA / 3 X as they apply to a Large Break LOCA: Significance of 3.9 SRO PZR readings 3 000015/17 RCP Malfunctions / 4 000022 Loss of Rx Coolant Makeup / 2 000025 Loss of RHR System / 4 AA2.03 - Ability to determine and interpret the following 000026 Loss of Component Cooling X as they apply to the Loss of Component Cooling Water: 2.9 SRO Water / 8 The valve lineups necessary to restart the CCWS while 4 bypassing the portion of the system causing the abnormal condition 000027 Pressurizer Pressure Control System Malfunction / 3 000029 ATWS / 1 2.2.38 - Knowledge of conditions and limitations in the 000038 Steam Gen. Tube Rupture / 3 X facility license. 4.5 SRO 5

000040 (BW/E05; CE/E05; W/E12)

Steam Line Rupture - Excessive Heat Transfer / 4 000054 (CE/E06) Loss of Main Feedwater / 4 000055 Station Blackout / 6 000056 Loss of Off-site Power / 6 000057 Loss of Vital AC Inst. Bus / 6 000058 Loss of DC Power / 6 000062 Loss of Nuclear Svc Water / 4 000065 Loss of Instrument Air / 8 W/E04 LOCA Outside Containment / 3 W/E11 Loss of Emergency Coolant Recirc. / 4 BW/E04; W/E05 Inadequate Heat Transfer - Loss of Secondary Heat Sink / 4 AA2.05 - Generator Voltage and Electric Grid 000077 Generator Voltage and Electric X Disturbances: Operational Status of Offsite Circuit 3.8 SRO Grid Disturbances / 6 6 2014 NRC Revision 0

K/A Category Totals: 3 3 Group Point Total: 18/6 2014 NRC Revision 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 / SRO)

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

1 2 3 1 2 000001 Continuous Rod Withdrawal / 1 000003 Dropped Control Rod / 1 000005 Inoperable/Stuck Control Rod / 1 000024 Emergency Boration / 1 000028 Pressurizer Level Malfunction / 2 X AA2.02 - Ability to determine and interpret the 3.8 SRO following as they apply to the Pressurizer Level 7 Control Malfunction: PZR level as a function of power level or T-ave. including interpretation or malfunction 000032 Loss of Source Range NI / 7 000033 Loss of Intermediate Range NI / 7 000036 (BW/A08) Fuel Handling Accident / 8 000037 Steam Generator Tube Leak / 3 000051 Loss of Condenser Vacuum / 4 000059 Accidental Liquid RadWaste Rel. / 9 2.4.50 -Ability to verify system alarm setpoints 000060 Accidental Gaseous Radwaste Rel. / 9 X 4.0 SRO and operate controls identified in the alarm 8

response manual.

000061 ARM System Alarms / 7 000067 Plant Fire On-site / 8 000068 (BW/A06) Control Room Evac. / 8 X 2.4.28 Knowledge of procedures relating to a 4.1 SRO security event. 9 000069 (W/E14) Loss of CTMT Integrity / 5 000074 (W/E06&E07) Inad. Core Cooling / 4 000076 High Reactor Coolant Activity / 9 W/EO1 & E02 Rediagnosis & SI Termination / 3 W/E13 Steam Generator Over-pressure / 4 W/E15 Containment Flooding / 5 W/E16 High Containment Radiation / 9 BW/A01 Plant Runback / 1 BW/A02&A03 Loss of NNI-X/Y / 7 BW/A04 Turbine Trip / 4 BW/A05 Emergency Diesel Actuation / 6 BW/A07 Flooding / 8 BW/E03 Inadequate Subcooling Margin / 4 BW/E08; W/E03 LOCA Cooldown - Depress. / 4 BW/E09; CE/A13; W/E09&E10 Natural Circ. / 4 BW/E13&E14 EOP Rules and Enclosures 2014 NRC Revision 0

AA2.2 - Ability to determine and interpret the CE/A11; W/E08 RCS Overcooling - PTS / 4 X 3.4 SRO following as they apply to the (RCS 10 Overcooling) Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments.

CE/A16 Excess RCS Leakage / 2 CE/E09 Functional Recovery K/A Category Point Totals: 2 2 Group Point Total: 9/4 2014 NRC Revision 0

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

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

1 2 3 4 5 6 1 2 3 4 003 Reactor Coolant Pump 004 Chemical and Volume Control 2.4.45 - Ability to prioritize and interpret the 005 Residual Heat Removal X significance of each annunciator or alarm. 4.3 SRO 12 006 Emergency Core Cooling 007 Pressurizer Relief/Quench Tank 008 Component Cooling Water 2.1.25 -Conduct of Operations: Ability to 010 Pressurizer Pressure Control X interpret reference materials, such as graphs, 4.2 SRO curves, tables, etc. 11 A2.01 - Ability to (a) predict the impacts of 012 Reactor Protection X the following malfunctions or operations on 3.6 SRO the RPS; and (b) based on those predictions, 13 use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Faulty bistable operation A2.05 - Ability to (a) predict the impacts of 013 Engineered Safety Features X the following malfunctions or operations on 4.2 SRO Actuation the ESFAS; and (b) based Ability on those 14 predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of dc control power 022 Containment Cooling 025 Ice Condenser 026 Containment Spray 039 Main and Reheat Steam 059 Main Feedwater 061 Auxiliary/Emergency Feedwater A2.12 - Ability to (a) predict the impacts of 062 AC Electrical Distribution X the following malfunctions or operations on 3.6 SRO the ac distribution system; and (b) based on 15 those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Restoration of power to a system with a fault on it 063 DC Electrical Distribution 2014 NRC Revision 0

064 Emergency Diesel Generator 073 Process Radiation Monitoring 076 Service Water 078 Instrument Air 103 Containment K/A Category Point Totals: 3 2 Group Point Total: 28/5 2014 NRC Revision 0

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

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

1 2 3 4 5 6 1 2 3 4 001 Control Rod Drive X 2.2.12 - Knowledge of surveillance procedures. 4.1 SRO 17 002 Reactor Coolant 011 Pressurizer Level Control 014 Rod Position Indication 015 Nuclear Instrumentation 016 Non-nuclear Instrumentation 017 In-core Temperature Monitor 027 Containment Iodine Removal 028 Hydrogen Recombiner and Purge Control 029 Containment Purge 033 Spent Fuel Pool Cooling A2.02 - Ability to (a) predict the impacts of the 034 Fuel Handling Equipment X following malfunctions or operations on the 3.9 SRO Fuel Handling System ; and (b) based on those 16 predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Dropped Cask 035 Steam Generator 041 Steam Dump/Turbine Bypass Control A2.15 - Ability to (a) predict the impacts of the 045 Main Turbine Generator X following malfunctions or operation on the 2.6 SRO MT/G system; and (b) based on those 18 predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Turbine overspeed 055 Condenser Air Removal 056 Condensate 068 Liquid Radwaste 071 Waste Gas Disposal 072 Area Radiation Monitoring 075 Circulating Water 079 Station Air 086 Fire Protection 2014 NRC Revision 0

K/A Category Point Totals: 2 1 Group Point Total: 10/3 2014 NRC Revision 0

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Waterford 3 (SRO Exam Outline) Date of Exam: April 2, 2014 Category K/A # Topic RO SRO-Only IR # IR #

Ability to use procedures related to shift staffing, such as 2.1.5 3.9 SRO minimum crew complement, overtime limitations, etc.

19

1. Ability to identify and interpret diverse indications to validate 2.1.45 4.3 SRO Conduct the response of another indication.

20 of Operations 2.1.

2.1.

2.1.

2.1.

Subtotal 2 2.2.13 Knowledge of tagging and clearance procedures. 4.3 SRO 21 2.2.14 Knowledge of the process for controlling equipment 4.3 SRO 2.

configuration or status. 22 Equipment Control 2.2.

2.2.

2.2.

2.2.

Subtotal 2 2.3.4 Knowledge of radiation exposure limits under normal or 3.7 SRO emergency conditions. 23 2.3.6 Ability to approve release permits. 3.8 SRO 3.

24 Radiation Control 2.3.

2.3.

2.3.

2.3.

Subtotal 2 2.4.42 Knowledge of emergency response facilities. 3.8 SRO 25 4.

Emergency 2.4.

Procedures / 2.4.

Plan 2.4.

2.4.

2.4.

Subtotal 1 Tier 3 Point Total 10 7 2014 NRC Revision 0

ES-401 Record of Rejected K/As Form ES-401-4 Tier / Randomly Reason for Rejection Group Selected K/A 1/2 00001 AA2.01 Originally selected AA2.01 for Continuous Rod Withdrawal. Could not develop an SRO question for this K/A. At W-3, the steps in the offnormal (OP-901-102) for this event are limited. Randomly selected another K/A for Tier 1 group 2. AA2.02 for 028, Pressurizer Level Malfunction was selected.

2/1 004 G2.4.1 Originally selected G2.4.1 for Chemical and Volume Control.

Could not develop an SRO question for this K/A. This K/A was knowledge of immediate action steps, which is not SRO level.

Randomly selected another K/A for Tier 2 group 1. G2.1.25 for 010, Pressurizer Pressure control was selected.

2/1 012 A2.06 Originally selected A2.06 for Reactor Protection. Could not develop an SRO question for this K/A The action for this K/A is an immediate action to trip the reactor. Randomly selected another K/A for Tier 2 group 1. A2.01 for 012, Reactor Protection was selected 3 G2.2.35 Originally selected G2.2.35. Could not develop an SRO question for this K/A that did not conflict with the previous two previous NRC exams, the operating tests, or the audit exam. Randomly selected another K/A Tier 3 for Equipment Control. G2.2.14 was selected.

1/1 077 AA2.08 Originally selected AA2.08 for Generator Voltage and Electric Grid Disturbances. Waterford 3 does not have criteria for tripping the reactor or turbine on an electric grid disturbance. Randomly selected another K/A for Tier 1 group 1. AA 2.05 Operational status of offsite circuit was selected.

1/2 060 G2.4.6 Originally selected G2.4.6 for Accidental Gaseous Radwaste Rel.

Tier 1 Group 2 are abnormal operating procedures at Waterford 3.

Could not tie EOP mitigation strategies to any of the categories.

Randomly selected G2.4.50 for 060, Accidental Gaseous Radwaste Rel.

2/2 041 G2.2.12 Originally selected G2.2.12 for Steam Dump/Turbine Bypass Control. There is no surveillance procedures performed on Steam Bypass at W-3. A Question had already been developed in the RO Exam for the ADV. Randomly selected another Tier 2/ Group 2 system for G2.2.12. 0001, Control Rod Drive was selected.

2/2 034 A2.03 Originally selected A2.03 for Fuel Handling Equipment System.

There is no procedural guidance for W-3 for a mis-positioned fuel element. Randomly selected A2.02 for Fuel Handling Equipment System.

2014 NRC Revision 0

3 G2.2.6 Originally selected G2.2.6. A question was written for this K/A, but a procedure change made to the plant procedure to simplify the revision process invalidated the question. An SRO question could no longer be developed for this K/A. Randomly selected another K/A Tier 3 for Equipment Control. G2.2.13 was selected 1/2 CE/A13 2.4.35 Originally selected CE/A13 2.4.35. After NRC review of the question submitted the question was deemed unsat. It was determined that an SRO level question on local AO actions could not be written. Randomly selected another system from Tier 1 group 2. Randomly selected 068, Control Room Evacuation.

Randomly selected another generic topic from G4 that a question could be written on for Control Room Evacuation. G2.4.28 was randomly selected.

2014 NRC Revision 0

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Waterford 3 Date of Examination: 4/07/2014 Examination Level: RO SRO Operating Test Number: 1 Administrative Topic Type Describe activity to be performed (see Note) Code*

A1 2.1.25, Ability to interpret reference materials, such as graphs, curves, tables, etc.

Conduct of Operations Calculate required and available Feedwater to place the M,R plant on Shutdown Cooling and determine the time K/A Importance: requirement to be on Shutdown Cooling.

3.9 A2 2.1.23, Ability to perform specific system and integrated plant procedures during all modes of plant operation.

Conduct of Operations Complete a calculation for determining the amount of pure water that may be added to the Refuel Cavity without D,P,R K/A Importance: dilution to below shutdown margin requirements in accordance with OP-010-006, Outage Operations, section 4.3 9.24, Refueling Cavity Boron Concentration.

(Used on 2012 NRC Exam)

A3 2.2.12, Knowledge of surveillance procedures Equipment Control Complete surveillance OP-903-008, Reactor Coolant N,R System Isolation Leakage Test, Attachment 10.9 RC Loop 2 Hot Leg Injection Inside Containment Check SI-510B Leak K/A Importance:

Rate Data.

3.7 A4 2.3.11, Ability to control radiation releases.

Radiation Control Evaluate Meteorological conditions for gaseous release M,R from the Gaseous Waste Management System in accordance with OP-007-003, Gaseous Waste K/A Importance:

Management.

3.8 Emergency Plan Not Selected NOTE: All items (5 total are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.

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

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

(P)revious 2 exams ( 1; randomly selected) 2014 NRC Revision 1

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Waterford 3 Date of Examination: 4/07/2014 Examination Level: RO SRO Operating Test Number: 1 Administrative Topic Type Describe activity to be performed (see Note) Code*

A5 2.1.25, Ability to interpret reference materials, such as Conduct of Operations graphs, curves, tables, etc.

N,R Review and approve a completed calculation for required and available Feedwater to place the plant on Shutdown K/A Importance: Cooling and review recommendation for the maximum time 4.2 requirement to be on Shutdown Cooling.

A6 2.1.23, Ability to perform specific system and integrated plant procedures during all modes of plant operation.

Conduct of Operations Review and approve a completed calculation for determining the amount of pure water that may be added to K/A Importance: D,P,R the Refuel Cavity without dilution to below shutdown margin 4.4 requirements in accordance with OP-010-006, Outage Operations, section 9.24, Refueling Cavity Boron Concentration.

(Used on 2012 NRC Exam)

A7 2.2.40, Ability to apply Technical Specifications for a Equipment Control system.

N,R Review and approve surveillance OP-903-008, Reactor K/A Importance: Coolant System Isolation Leakage Test, Attachment 10.9 4.7 RC Loop 2 Hot Leg Injection Inside Containment Check SI-510B Leak Rate Data.

A8 2.3.4, Knowledge of radiation exposure limits under normal Radiation Control or emergency conditions.

Calculate dose and assign non-licensed operators to vent D,R Safety Injection piping in Safeguards Room A. Given dose K/A Importance: rate with and without shielding installed, time to install 3.7 shielding, and job completion time using 1 operator or using 2 operators, determine proper job assignment.

A9 2.4.41, Knowledge of the emergency action level thresholds Emergency Plan and classifications.

M,R Determine appropriate Emergency Plan EAL.

K/A Importance:

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 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) 2014 NRC Revision 0

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Waterford 3 Date of Examination: 4/07/14 Exam Level RO SRO-I SRO-U Operating Test No.: NRC Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System / JPM Title Type Code* Safety Function S1 004 Chemical and Volume Control System Secure Emergency Boration in accordance with OP-901-103, Emergency Boration. (Used on 2012 NRC Exam) D,L,P,S 1 A4.07 Boration/Dilution RO - 3.9, SRO - 3.7 S2 006 Emergency Core Cooling System Mitigate the release of radioactivity through RWSP following RAS Actuation in accordance with OP-902-002, Loss of Coolant accident Recovery.

Fault: LPSI pump continues to run after RAS actuates, requiring the A,D,EN,L,S 2 applicant to manually stop the running LPSI pump after additional valve manipulations.

011 EA1.12 Long term containment of RO - 4.1, SRO - 4.4 radioactivity S3 003 Reactor Coolant Pump System Perform a Reactor Coolant Pump Shutdown in accordance with OP-001-002, Reactor Coolant Pump Operation.

Fault: Reactor Coolant pump reverse rotates requiring stopping of A,D,L,S 4P remaining Reactor Coolant Pumps.

A2.02 Conditions which exist for an abnormal shutdown of an RCP in comparison to a RO - 3.7, SRO - 3.9 normal shutdown of an RCP S4 076 Service Water System Perform a manual start of the Auxiliary Component Cooling water System in accordance with OP-002-001, Auxiliary Component Cooling D,L,S 4S Water.

A4.01 SWS pumps RO - 2.9, SRO - 2.9 S5 026 Containment Spray System Realign Containment Spray for auto initiation following a Containment Spray actuation signal in accordance with OP-902-009, Standard D,EN,L,S 5 Appendices (Appendix 5-E).

A4.01 CSS Controls RO - 4.5, SRO - 4.3 S6 064 Emergency Diesel Generators Restore Power to Safety Bus 3B in accordance with OP-902-000, Standard Post Trip Actions. A,EN,N,S 6 Fault: Tie breaker fails to open automatically. (WF3 OE)

A3.03 Indicating lights, meters, and recorders RO - 3.4, SRO - 3.3 S7 008 Component Cooling Water System Start Component Cooling Water Pump in accordance with OP-901-311, Loss of Train B Safety Bus. D,S 8 A4.01 CCW indications and Controls RO - 3.3, SRO - 3.1 1 2014 NRC Revision 1

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 S8 068 Liquid Radwaste System Discharge Waste Condensate Tank A to the Circulating water System in accordance with OP-007-004, Liquid Waste Management System.

Fault: Upon initiation of flow, LWM flow controller output fails high, A,D,L,S 9 raising flow beyond what is permitted by the release permit.

A4.03 Stoppage of releases if limits RO - 3.9, SRO - 3.8 exceeded In-Plant Systems@ (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U)

P1 061 Auxiliary/Emergency Feedwater (AFW) System Establish Local Manual Control of EFW Flow in accordance with OP-009-003, Emergency Feedwater. (PRA Top Operator Action List) D,E,L 4S K1.01 S/G System RO - 4.1, SRO - 4.1 P2 062 A.C. Electrical Distribution Transfer SUPS 014AB from Alternate to Normal AC Power in accordance with OP-006-005, Inverters and Distribution.

A,D,P 6 Fault: After alignment, voltage will not be indicated on SUPS 014 AB inverter. (Used on 2012 NRC Exam)

A3.04 Operation of inverter RO - 2.7, SRO - 2.9 P3 033 Spent Fuel pool Cooling System Restore Spent Fuel Cooling Pump to Operation via the Emergency Diesel Generators in accordance with OP-901-513, Spent Fuel pool E,N,R 8 Cooling Malfunction.

K3.03 Spent fuel temperature RO - 3.0, SRO - 3.3

@ 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 5 (C)ontrol room (D)irect from bank 9/ 8/ 4 9 (E)mergency or abnormal in-plant 1/ 1/ 1 2 (EN)gineered safety feature - / - / 1 (control room system) 3 (L)ow-Power / Shutdown 1/ 1/ 1 7 (N)ew or (M)odified from bank including 1(A) 2/ 2/ 1 2 (P)revious 2 exams 3/ 3/ 2 (randomly selected) 2 (R)CA 1/ 1/ 1 1 (S)imulator 2 2014 NRC Revision 1

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Waterford 3 Date of Examination: 4/07/14 Exam Level RO SRO-I SRO-U Operating Test No.: NRC Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System / JPM Title Type Code* Safety Function S1 004 Chemical and Volume Control System Secure Emergency Boration in accordance with OP-901-103, Emergency Boration. (Used on 2012 NRC Exam) D,L,P,S 1 A4.07 Boration/Dilution RO - 3.9, SRO - 3.7 S2 006 Emergency Core Cooling System Mitigate the release of radioactivity through RWSP following RAS Actuation in accordance with OP-902-002, Loss of Coolant accident Recovery.

Fault: LPSI pump continues to run after RAS actuates, requiring the A,D,EN,L,S 2 applicant to manually stop the running LPSI pump after additional valve manipulations.

011 EA1.12 Long term containment of RO - 4.1, SRO - 4.4 radioactivity S3 003 Reactor Coolant Pump System Perform a Reactor Coolant Pump Shutdown in accordance with OP-001-002, Reactor Coolant Pump Operation.

Fault: Reactor Coolant pump reverse rotates requiring stopping of A,D,L,S 4P remaining Reactor Coolant Pumps.

A2.02 Conditions which exist for an abnormal shutdown of an RCP in comparison to a RO - 3.7, SRO - 3.9 normal shutdown of an RCP S4 076 Service Water System Perform a manual start of the Auxiliary Component Cooling water System in accordance with OP-002-001, Auxiliary Component Cooling D,L,S 4S Water.

A4.01 SWS pumps RO - 2.9, SRO - 2.9 S5 S6 064 Emergency Diesel Generators Restore Power to Safety Bus 3B in accordance with OP-902-000, Standard Post Trip Actions. A,EN,N,S 6 Fault: Tie breaker fails to open automatically. (WF3 OE)

A3.03 Indicating lights, meters, and recorders RO - 3.4, SRO - 3.3 S7 008 Component Cooling Water System Start Component Cooling Water Pump in accordance with OP-901-311, Loss of Train B Safety Bus. D,S 8 A4.01 CCW indications and Controls RO - 3.3, SRO - 3.1 3 2014 NRC Revision 1

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 S8 068 Liquid Radwaste System Discharge Waste Condensate Tank A to the Circulating water System in accordance with OP-007-004, Liquid Waste Management System.

Fault: Upon initiation of flow, LWM flow controller output fails high, A,D,L,S 9 raising flow beyond what is permitted by the release permit.

A4.03 Stoppage of releases if limits RO - 3.9, SRO - 3.8 exceeded In-Plant Systems@ (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U)

P1 061 Auxiliary/Emergency Feedwater (AFW) System Establish Local Manual Control of EFW Flow in accordance with OP-009-003, Emergency Feedwater. (PRA Top Operator Action List) D,E,L 4S K1.01 S/G System RO - 4.1, SRO - 4.1 P2 062 A.C. Electrical Distribution Transfer SUPS 014AB from Alternate to Normal AC Power in accordance with OP-006-005, Inverters and Distribution.

A,D,P 6 Fault: After alignment, voltage will not be indicated on SUPS 014 AB inverter. (Used on 2012 NRC Exam)

A3.04 Operation of inverter RO - 2.7, SRO - 2.9 P3 033 Spent Fuel pool Cooling System Restore Spent Fuel Cooling Pump to Operation via the Emergency Diesel Generators in accordance with OP-901-513, Spent Fuel pool E,N,R 8 Cooling Malfunction.

K3.03 Spent fuel temperature RO - 3.0, SRO - 3.3

@ 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 5 (C)ontrol room (D)irect from bank 9/ 8/ 4 8 (E)mergency or abnormal in-plant 1/ 1/ 1 2 (EN)gineered safety feature - / - / 1 (control room system) 2 (L)ow-Power / Shutdown 1/ 1/ 1 6 (N)ew or (M)odified from bank including 1(A) 2/ 2/ 1 2 (P)revious 2 exams 3/ 3/ 2 (randomly selected) 2 (R)CA 1/ 1/ 1 1 (S)imulator 4 2014 NRC Revision 1

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Waterford 3 Date of Examination: 4/07/14 Exam Level RO SRO-I SRO-U Operating Test No.: NRC Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System / JPM Title Type Code* Safety Function S1 S2 006 Emergency Core Cooling System Mitigate the release of radioactivity through RWSP following RAS Actuation in accordance with OP-902-002, Loss of Coolant accident Recovery.

Fault: LPSI pump continues to run after RAS actuates, requiring the A,D,EN,L,S 2 applicant to manually stop the running LPSI pump after additional valve manipulations.

011 EA1.12 Long term containment of RO - 4.1, SRO - 4.4 radioactivity S3 003 Reactor Coolant Pump System Perform a Reactor Coolant Pump Shutdown in accordance with OP-001-002, Reactor Coolant Pump Operation.

Fault: Reactor Coolant pump reverse rotates requiring stopping of A,D,L,S 4P remaining Reactor Coolant Pumps.

A2.02 Conditions which exist for an abnormal shutdown of an RCP in comparison to a RO - 3.7, SRO - 3.9 normal shutdown of an RCP S4 S5 S6 064 Emergency Diesel Generators Restore Power to Safety Bus 3B in accordance with OP-902-000, Standard Post Trip Actions. A,EN,N,S 6 Fault: Tie breaker fails to open automatically. (WF3 OE)

A3.03 Indicating lights, meters, and recorders RO - 3.4, SRO - 3.3 S7 S8 5 2014 NRC Revision 1

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 In-Plant Systems@ (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U)

P1 061 Auxiliary/Emergency Feedwater (AFW) System Establish Local Manual Control of EFW Flow in accordance with OP-009-003, Emergency Feedwater. (PRA Top Operator Action List) D,E,L 4S K1.01 S/G System RO - 4.1, SRO - 4.1 P2 P3 033 Spent Fuel pool Cooling System Restore Spent Fuel Cooling Pump to Operation via the Emergency Diesel Generators in accordance with OP-901-513, Spent Fuel pool E,N,R 8 Cooling Malfunction.

K3.03 Spent fuel temperature RO - 3.0, SRO - 3.3

@ 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 3 (C)ontrol room (D)irect from bank 9/ 8/ 4 3 (E)mergency or abnormal in-plant 1/ 1/ 1 2 (EN)gineered safety feature - / - / 1 (control room system) 2 (L)ow-Power / Shutdown 1/ 1/ 1 3 (N)ew or (M)odified from bank including 1(A) 2/ 2/ 1 2 (P)revious 2 exams 3/ 3/ 2 (randomly selected) 0 (R)CA 1/ 1/ 1 1 (S)imulator 6 2014 NRC Revision 1

Appendix D Scenario Outline Form ES-D-1 Facility: Waterford 3 Scenario No.: 1 Op Test No.:

Examiners: Operators:

Initial Conditions: Reactor power is 100%, MOC Turnover:

Protected Train is B, AB Busses are aligned to Train B, Nothing is OOS, maintain 100%

power Event Malf. Event Event No. No. Type* Description (A) VCT level instrument CVC-ILT-0227 Fails high diverting I - ATC letdown to the Boron Management system. OP-901-113, 1 CVC12A1 I - SRO Volume Control Tank Makeup Control Malfunction C - BOP (A) Atmospheric Dump Valve on Steam Generator #1 C - SRO spuriously opens requiring closing in accordance with OP-2 MS23A TS - SRO 901-221, Secondary System Transient. (TS 3.7.1.7)

R - ATC N - BOP (A) Control Element Assembly CEA #9 drops into the core N - SRO requiring a rapid plant down power to 70%.

3 RD02 A09 TS - SRO OP-901-102, CEA or CEDMCs Malfunction. (TS 3.1.3.1)

Main Feedwater Pump B loss of oil pressure trip results in a Reactor Power Cutback and unanalyzed rod configuration 4 FW32 B M - ALL and reactor trip. OP-902-000, Standard Post Trip Actions.

3 Control Element Assemblies fail to insert into the core RD11 A07 following the reactor trip, Emergency Boration (Critical Task RD11 A37 1, commence emergency boration prior to exiting step 1 5 RD11 A39 C - ATC of OP-902-000)

Main Turbine fails to trip automatically requiring manual action. OP-902-000, Standard Post Trip Actions. (Critical Task 2, trip the main turbine prior to TCOLD lowering to 6 RP03 C - BOP 443 F)

Main Feedwater Pump A overspeed trip resulting in a loss of C - ATC Main Feedwater requiring stopping of 2 RCPs. OP-902-006, 7 FW03 A C - SRO Loss of Main Feedwater Recovery Procedure RP05 A6/A7 Emergency Feedwater fails to actuate requiring manual RP05 B6/B7 action to initiate Emergency Feedwater. (Critical Task 3, RP05 C6/C7 manually actuate EFW prior to SG level lowering to <

8 RP05 D6/D7 I - BOP 20% WR) (#2 Dominant Accident Sequence)

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor 2014 NRC Exam Scenario 1 D-1 Rev 2

Scenario Event Description NRC Scenario 1 The crew assumes the shift at 100% power with instructions to maintain 100% power.

After taking the shift, Volume Control Tank level instrument CVC-ILT-0227 fails high resulting in valve CVC-169 diverting letdown to the Boron Management System. The ATC should recognize this through indications on the Plant Monitoring Computer. The SRO should enter into procedure OP-901-113, Volume Control Tank Makeup Control Malfunction, and direct the ATC to place valve CVC-169 to the VCT position.

After the procedure has been reviewed and a brief has occurred, the setpoint for #1 Steam Generator Atmospheric Dump Valve (ADV) fails low and the controller MS-IPIC-0303-A1 demands the ADV fully open. The SRO should enter into procedure OP-901-221, Secondary System Transient and direct the BOP to manually close the valve by placing controller MS-IPIC-0303-A1 in manual control and lowering the output to minimum. The SRO should review Technical Specification 3.7.1.7 and determine that the ADV will need to be restored to operable status in automatic within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be below 70% Rated Thermal Power within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

After the SRO addresses Technical Specifications of the above event, Control Element Assembly #9 (Regulating group 4) drops into the core. The SRO should enter into procedure OP-901-102, CEA or CEDMCS Malfunction and proceed to section E1, CEA Misalignment Greater than 7 inches. The SRO will direct the BOP to adjust turbine load to match T AVG to TREF initially and then perform a rapid plant downpower (commence direct boration to the RCS) within 15 minutes of the dropped CEA to comply with Technical Specification 3.1.3.1 and the COLR. The SRO should evaluate and enter Technical Specifications 3.1.3.1 action c.

After the crew has performed a significant portion of the downpower and at lead Examiner discretion, Main Feedwater Pump B will develop an oil leak that will lead to a loss of oil pressure and a trip of the pump. The BOP will note the tripped pump and a Reactor Power Cutback will occur as expected. The Reactor Power cutback combined with a previous dropped CEA will result in an unanalyzed rod configuration and the ATC will conduct a manual Reactor Trip. During the trip 3 CEAs remain fully withdrawn and the ATC will commence emergency boration to the RCS in accordance with OP-902-000, Standard Post Trip Actions (Critical Task 1). While the ATC is emergency borating the RCS, the turbine will fail to automatically trip and MSIS will fail to actuate requiring the BOP to manually trip the turbine or close the Main Steam Isolation Valves in accordance with OP-902-000, Standard Post Trip Actions (Critical Task 2).

While the crew is progressing through the action steps of OP-902-000, Standard Post Trip Actions, Main Feedwater Pump A will trip on an overspeed condition resulting in a complete loss of Main Feedwater.

The crew will complete Standard Post Trip Actions and enter diagnostics. The crew will enter Appendix 1 of OP-902-009, Standard Appendices and diagnose into OP-902-006, Loss of Main Feedwater Recovery Procedure. The SRO will direct the BOP to ensure that Emergency Feedwater is available to the Steam Generators. At this point, or earlier depending on when the narrow range level in the Steam Generators reaches 27.4% and lowering, Emergency Feedwater actuation failed to occur. The BOP must take action to ensure that a heat sink exists by manually actuating Emergency Feedwater Actuation Signal, establishing Auxiliary Feedwater flow, or depressurizing the Steam Generators and injecting with Condensate pumps (Critical Task 3). In Waterford 3 Probalistic Risk Analysis, the #2 Dominant Accident Sequence that would lead to core damage is a Loss of Main Feedwater with an EFW failure. This accident would (statistically speaking) constitute 20.86% of all core damage events at WF3. This is mitigated by performing Critical Task 3. The crew will continue on in OP-902-006 and stop 1 RCP in each loop.

The scenario may be terminated once all critical tasks have been completed or performance standards exceeded and when the crew stops 2 RCPs in accordance with OP-902-006, Loss of Main Feedwater Recovery or at the discretion of the lead examiner.

2014 NRC Exam Scenario 1 D-1 Rev 2

NRC Scenario 1 CRITICAL TASKS

1. ESTABLISH REACTIVITY CONTROL This task is satisfied by commencing Emergency Boration flow by either Boric Acid makeup pumps or Gravity Feed valves in accordance with OP-902-000, Standard Post Trip Actions step 1 prior to exiting the step to verify Reactivity Control. This task becomes applicable following the initiation of a Reactor Trip.

2. ESTABLISH REACTIVITY CONTROL This task is satisfied by stopping the steam flow to the Main Turbine by either tripping the Main Turbine manually or by closing the Main Steam Isolation valves before TCOLD lowers to below 443 F. This task becomes applicable following the initiation of the Reactor Trip signal.

3. ESTABLISH RCS HEAT REMOVAL This task is satisfied by manually actuating Emergency Feedwater Actuation System, manually starting at least 1 Emergency Feedwater pump, aligning and starting the Auxiliary Feedwater pump, or depressurizing at least 1 Steam Generator and injecting the Steam Generator with Condensate flow prior to Steam Generator levels reducing below 20% Wide Range in both Steam Generators. This task becomes applicable once Steam Generator water level reduces below 27.4% Narrow Range in one or both Steam Generators.

Scenario Quantitative Attributes

1. Total malfunctions (5-8) 8

2. Malfunctions after EOP entry (1-2) 4

3. Abnormal events (2-4) 3

4. Major transients (1-2) 1

5. EOPs entered/requiring substantive actions (1-2) 1

6. EOP contingencies requiring substantive actions (0-2) 0

7. Critical tasks (2-3) 3 2014 NRC Exam Scenario 1 D-1 Rev 2

NRC Scenario 1 SCENARIO SETUP A. Reset Simulator to IC-187.

B. Verify Scenario Malfunctions, Remotes, and overrides are loaded, as listed in the Scenario Timeline.

C. Ensure Protected Train B sign is placed in SM office window.

D. Verify EOOS is 10.0 Green with nothing out of service E. Complete the simulator setup checklist.

F. Start Insight, open file Crew Performance.tis.

2014 NRC Exam Scenario 1 D-1 Rev 2

NRC Scenario 1 SIMULATOR BOOTH INSTRUCTIONS Event 1 VCT level instrument, CVC-ILT-0227, Fails High

1. On Lead Examiner's cue, initiate Event Trigger 1.

2. If Work Week Manager or PMI are called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

Event 2 ADV #1 Controller MS-IPIC-0303A SETPOINT FAILURE

1. On Lead Examiner's cue, initiate Event Trigger 2.

2. If Work Week Manager or PMI are called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

Event 3 CEA #9 Falls into the core

1. On Lead Examiner's cue, initiate Event Trigger 3.

2. If Work Week Manager or PMI are called, inform the caller that a work package will be assembled and a team will be sent to CEDMCS Alley.

3. If called as RAB and directed to CEDMCs Alley, respond in 3 minutes that you have arrived. If asked, report that there is no apparent cause for the dropped CEA.

4. If Chemistry is called to sample the RCS for Dose Equivalent Iodine due to the down power, acknowledge and report that samples will be taken 2-6 hours from notification time and if asked tell the caller your name is Joe Chemist.

5. If notified as Load Dispatcher (Woodlands) acknowledge the communications and inform the caller that the grid will remain stable with available backup generation.

6. If requested to monitor Polisher Vessel D/P and remove as necessary, acknowledge the report.

Event 4 Main Feedwater pump B Lube oil leak/trip

1. On Lead Examiner's cue, initiate Event Trigger 4.

2. After event is initiated, wait 2 minutes and report as NAO that a significant oil leak is coming from B Main Feedwater pump.

3. If directed as NAO to open the breakers from the oil pumps, wait 2 minutes and initiate Event Trigger 14.

Event 5-6 3 CEAs stuck out on Reactor Trip and Main Turbine Fails to Trip

1. No actions for these events.

Event 7 Main Feedwater Pump A overspeed trip

1. On lead examiner's cue, initiate Event Trigger 7.

2. If called as NAO to check status of Main Feedwater Pump A, wait 2 minutes and report that you are unable to determine why it has tripped.

Event 8 Emergency Feedwater Actuation System fails to actuate

1. No actions for this event.

2. At the end of the scenario, before resetting, end data collection and save the file as 2014 Scenario 1-(start-end time).tid. Export to .csv file. Save the file into the folder for the appropriate crew..

2014 NRC Exam Scenario 1 D-1 Rev 2

NRC Scenario 1 SCENARIO TIMELINE DELAY RAMP EVENT KEY DESCRIPTION TRIGGER FINAL HH:MM:SS HH:MM:SS EVENT DESCRIPTION 8 RP05B6 FAILS TO TRIP CH B LO S/G LEVEL #1(EFAS-1) N/A 00:00:00 00:00:00 ACTIVE EFAS FAILS TO ACTUATE ON LOW S/G LEVEL 8 RP05B7 FAILS TO TRIP CH B LO S/G LEVEL #2(EFAS-2) N/A 00:00:00 00:00:00 ACTIVE EFAS FAILS TO ACTUATE ON LOW S/G LEVEL 8 RP05C6 FAILS TO TRIP CH C LO S/G LEVEL #1(EFAS-1) N/A 00:00:00 00:00:00 ACTIVE EFAS FAILS TO ACTUATE ON LOW S/G LEVEL 8 RP05C7 FAILS TO TRIP CH C LO S/G LEVEL #2(EFAS-2) N/A 00:00:00 00:00:00 ACTIVE EFAS FAILS TO ACTUATE ON LOW S/G LEVEL 8 RP05D6 FAILS TO TRIP CH D LO S/G LEVEL #1(EFAS-1) N/A 00:00:00 00:00:00 ACTIVE EFAS FAILS TO ACTUATE ON LOW S/G LEVEL 8 RP05D7 FAILS TO TRIP CH D LO S/G LEVEL #2(EFAS-2) N/A 00:00:00 00:00:00 ACTIVE EFAS FAILS TO ACTUATE ON LOW S/G LEVEL 6 RP08G RELAY K305 FAILED, MSISTRAIN A (MS/FW) N/A 00:00:00 00:00:00 ACTIVE MAIN STEAM ISOLATION SIGNAL FAILS TO ACTUATE 2014 NRC Exam Scenario 1 D-1 Rev 2

NRC Scenario 1 DELAY RAMP EVENT KEY DESCRIPTION TRIGGER FINAL HH:MM:SS HH:MM:SS EVENT DESCRIPTION 6 RP09F RELAY K305 FAILED, MSISTRAIN B (MS/FW) N/A 00:00:00 00:00:00 ACTIVE MAIN STEAM ISOLATION SIGNAL FAILS TO ACTUATE 6 RP08H RELAY K313 FAILED, MSISTRAIN A (MS/FW) N/A 00:00:00 00:00:00 ACTIVE MAIN STEAM ISOLATION SIGNAL FAILS TO ACTUATE 6 RP09H RELAY K313 FAILED, MSISTRAIN B (MS/FW) N/A 00:00:00 00:00:00 ACTIVE MAIN STEAM ISOLATION SIGNAL FAILS TO ACTUATE 5 RD11A07 CEA 07 MECHANICALLY STUCK N/A 00:00:00 00:00:00 ACTIVE CEA 07 FAILS TO INSERT 5 RD11A37 CEA 37 MECHANICALLY STUCK N/A 00:00:00 00:00:00 ACTIVE CEA 37 FAILS TO INSERT 5 RD11A39 CEA 39 MECHANICALLY STUCK N/A 00:00:00 00:00:00 ACTIVE CEA 39 FAILS TO INSERT 6 RP03 REACTOR TRIP WITH NO AUTO TURBINE TRIP N/A 00:00:00 00:00:00 ACTIVE TURBINE FAILS TO TRIP AUTOMATICALLY 1 CV12A1 VCT LEVEL XMTR CVC-ILIC-0227 FAILS HI 1 00:00:00 00:00:00 ACTIVE VCT LEVEL TRANSMITTER FAILS HIGH 2 MS23A ADV 1 CNTRLR MS-IPIC-0303A SETPT FAILURE (0-1200 PSIG) 2 00:00:00 00:00:00 0 ATMOSPHERIC DUMP VALVE CONTROLLER FAILURE 2014 NRC Exam Scenario 1 D-1 Rev 2

NRC Scenario 1 DELAY RAMP EVENT KEY DESCRIPTION TRIGGER FINAL HH:MM:SS HH:MM:SS EVENT DESCRIPTION 3 RD02A09 DROPPED CEA 09 3 00:00:00 00:00:00 ACTIVE DROPPED CEA 09 4 FW32B MFW PUMP B LUBE OIL PIPE BREAK 4 00:00:00 00:00:00 65 MFW PUMP B LUBE OIL PIPE BREAK 7 FW03A MFW PUMP A OVERSPEED TRIP 7 00:00:00 00:00:00 ACTIVE MFW PUMP A OVERSPEED TRIP 8 RP05A6 FAILS TO TRIP CH A LO S/G LEVEL #1(EFAS-1) N/A 00:00:00 00:00:00 ACTIVE EFAS FAILS TO ACTUATE ON LOW S/G LEVEL 8 RP05A7 FAILS TO TRIP CH A LO S/G LEVEL #2(EFAS-2) N/A 00:00:00 00:00:00 ACTIVE EFAS FAILS TO ACTUATE ON LOW S/G LEVEL 4 FWR67 SGFP B MOP-1 BREAKER 14 00:00:00 00:00:00 OPEN OIL PUMP BREAKER REMOTE FUNCTION 4 FWR68 SGFP B MOP-2 BREAKER 14 00:00:00 00:00:00 OPEN OIL PUMP BREAKER REMOTE FUNCTION 4 FWR69 SGFP B EOP BREAKER 14 00:00:00 00:00:00 OPEN OIL PUMP BREAKER REMOTE FUNCTION 2014 NRC Exam Scenario 1 D-1 Rev 2

NRC Scenario 1 REFERENCES Event Procedures 1 OP-901-113, Volume Control Tank Makeup Control Malfunction, Rev. 301 2 OP-901-221, Secondary System Transient, Rev. 0 Technical Specification 3.7.1.7 3 OP-901-102, CEA or CEDMCS Malfunction, Rev. 301 OP-901-212, Rapid Plant Power Reduction, Rev. 6 OP-901-501, PMC or Core Operating Limit Supervisory System Malfunction, Rev. 13 Technical Specification 3.1.3.1 4 OP-901-101, Reactor Power Cutback, Rev. 8 5 OP-902-000, Standard Post Trip Actions, Rev. 15 OP-901-103, Emergency boration, Rev. 2 6 OP-902-000, Standard Post Trip Actions, Rev. 15 7 OP-902-009, Standard Appendices, Rev. 309, Appendix 1, Diagnostic Flow Chart OP-902-006, Loss of Main Feedwater Recovery Procedure, Rev. 15 8 OP-902-006, Loss of Main Feedwater Recovery Procedure, Rev. 15 2014 NRC Exam Scenario 1 D-1 Rev 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 1 Page 1 of 30 Event

Description:

VCT level instrument CVC-ILT-0227 Fails high.

Time Position Applicants Actions or Behavior Examiner Note Cue the Simulator Operator when ready for Event 1 ATC Recognizes and reports indications of failed channel.

Alarms:

PMC PID D39302 NT VCT PMC PID D39303 FLASH TK Indications:

Valve CVC-169, VCT Inlet/Bypass To Holdup Tanks indicates BMS VCT level indication CVC-ILT-0226 slowly lowering VCT level indication PMC-IUR-0001 Green Pen (Flashing)

Note The SRO may direct the ATC to take manual control of CVC-169, VCT Inlet/Bypass To Holdup Tanks and direct valve to the VCT position prior to entering procedure.

SRO Enter and direct the implementation of OP-901-113, Volume Control Tank Makeup Control Malfunction OP-901-113, Volume Control Tank Makeup Control Malfunction NOTE Failure low of VCT level instrument CVC-ILT-0227(PID A39401) will cause RWSP TO CHARGING PUMPS (CVC 507) to open and VCT DISCH VALVE (CVC 183) to close. Failure of VCT level instrument CVC-ILT-0226 (PID A39400) affects CP-4 level indication and Auto makeup to the VCT.

SRO 1. IF a VCT level instrument fails, THEN perform the following:

N/A a. IF level instrument CVC-ILT-0227 fails low causing Charging Pump suction source to swap to RWSP, THEN perform the following:

2014 NRC Exam Scenario 1 D-2 Rev 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 1 Page 2 of 30 Event

Description:

VCT level instrument CVC-ILT-0227 Fails high.

Time Position Applicants Actions or Behavior ATC b. IF level instrument CVC-ILT-0227 fails high causing the VCT INLET/BYPASS VALVE (CVC 169) to divert to BMS, THEN perform the following:

1) Align VCT INLET/BYPASS VALVE (CVC 169) to VCT.

2) Makeup to VCT as required to restore level in accordance with OP-002-005, CHEMICAL AND VOLUME CONTROL.

3) Initiate corrective action to repair level instrument.

5) WHEN level instrument CVC-ILT-0227 is repaired, THEN restore VCT INLET/BYPASS VALVE (CVC 169) to AUTO.

N/A c. IF level instrument CVC-ILT-0226 fails, THEN secure auto makeup to the VCT in accordance with OP-002-005, CHEMICAL AND VOLUME CONTROL, and monitor VCT level using PMC PID A39401, CVCS VOL CONT TK LVL 1.

CAUTION DIRECT LOCAL DILUTION/BORATION OPERATIONS MUST BE DIRECTED FROM THE CONTROL ROOM. THIS REQUIRES THAT CONTINUOUS COMMUNICATIONS BE ESTABLISHED AND MAINTAINED BETWEEN THE CONTROL ROOM AND THE OPERATOR STATIONED IN BAMT ROOM A. IF COMMUNICATION IS LOST AT ANY TIME, THEN THE LOCAL OPERATOR SHALL IMMEDIATELY SECURE DILUTION/BORATION BY CLOSING MANUAL DIRECT BORATION ISOLATION(BAM 138) AND PMU TO CHG PMP SUCT HDR ISOL (PMU 140).

N/A 2. IF boric acid OR primary water flow can NOT be established to VCT, THEN makeup directly to Charging Pump suction as follows:

Examiner Note This event is complete after the ATC places CVC-169 to the VCT position or At Lead Examiners Discretion Examiner Note Cue the Simulator Operator when ready for Event 2 2014 NRC Exam Scenario 1 D-2 Rev 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 2 Page 3 of 30 Event

Description:

Atmospheric Dump Valve #1 spuriously opens Time Position Applicants Actions or Behavior BOP Recognize and report indications of ADV open.

Alarms:

SG 1 ATMOSPHERIC DUMP VLV OPEN (Cabinet M, G-1)

COLSS MASTER (Cabinet L, A-6)

LETDOWN FLOW HI/LO (Cabinet G, C-1)

Indications RCS temperature dropping ADV 1 setpoint indicates 0 ADV 1 output is 100%

Reactor Power Rising above 100%

Note The SRO may direct the BOP to take manual control and close ADV 1 upon the initial report, prior to entering the off normal.

SRO Enter and direct the implementation of OP-901-221, Secondary System Transient.

OP-901-221, Secondary System Transient, E0 - General NOTE (1) Some steps of this procedure may not be applicable due to plant conditions. In these cases SM/CRS may NA the step.

(2) Steps within this procedure may be performed concurrently or out of sequence with SM/CRS concurrence.

N/A 1. If Reactor trip occurs, then go to OP-902-000, Standard Post Trip Actions.

N/A 2. If Reactor Power Cutback occurs, then perform OP-901-101, Reactor Power Cutback, concurrently with this procedure.

BOP 3. If an Atmospheric Dump Valve fails or begins to fail Open, then place the respective controller to MANUAL with minimum output.

N/A 4. If a Steam Bypass Valve fails or begins to fail Open, then perform any of the following (in preferred order) to close the valve.

N/A 5. If an uncontrollable RCS cooldown exists, then perform the following:

2014 NRC Exam Scenario 1 D-2 Rev 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 2 Page 4 of 30 Event

Description:

Atmospheric Dump Valve #1 spuriously opens Time Position Applicants Actions or Behavior NOTE (1) The following are initial turbine load reductions to be considered depending on the current power level, time in core life, and equipment malfunction.

Transient Initial Load Reduction Rate Two or more Heater Drain Pumps Tripping FW Heater #1 or #2 Extraction Steam 100 MW 40 MW/min Valve Closure Atmospheric Dump Valve Fails Open Steam Bypass Valve Fails Open (2) With COLSS in service, utilize the following to observe instantaneous power changes for power levels 40%. Reference Attachment 3, COLSS Maneuvering Power Indications, for all other power levels.

Reactor Power UFM in service UFM not in service MSBSRAW 95% (PMC PID C24631)

USBSRAW FWBSRAW

< 95% and 40% (PMC PID C24629) (PMC PID C24630)

BOP 6. If Main Turbine is available, then adjust Turbine load as necessary to maintain the following:

Reactor Power 100%

Match Tavg with Tref FWPT Suction Pressure > 300 PSIG (monitored on CP-1 via CD IPI1280, IP Htrs Outlet Hdr)

RCS Tcold 536F - 549F 2014 NRC Exam Scenario 1 D-2 Rev 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 2 Page 5 of 30 Event

Description:

Atmospheric Dump Valve #1 spuriously opens Time Position Applicants Actions or Behavior OP-005-007, Main Turbine and Generator CAUTION THE FOLLOWING SECTION HAS THE POTENTIAL TO AFFECT CORE REACTIVITY. REACTOR POWER, RCS TEMPERATURE, AND MAIN GENERATOR MW LOAD SHOULD BE CLOSELY MONITORED DURING PERFORMANCE OF THIS SECTION.

BOP 6.2.1 To change Load/Rate perform the following:

6.2.1.1 Depress LOAD/RATE MW/MIN pushbutton.

6.2.1.2 Depress appropriate numerical pushbuttons for desired load rate.

6.2.1.3 Depress ENTER pushbutton.

NOTE Prior to changing Reference Demand, Main Turbine load must not be changing.

BOP 6.2.2 To change Main Turbine load, perform the following:

6.2.2.1 Depress REF pushbutton.

6.2.2.2 Depress appropriate numerical pushbuttons for desired MW load.

6.2.2.3 Depress ENTER pushbutton.

6.2.2.4 Depress GO pushbutton.

6.2.2.5 Verify Turbine load change stops at the desired MW load.

N/A 7. If needed, then concurrently perform OP-901-212, Rapid Plant Down Power, until a power level is reached in which the plant can be stabilized.

N/A 8. If a loss of Feedwater preheating occurs, then go to E1, Loss of Feedwater Preheating.

SRO 9. If a loss of secondary inventory occurs (Condensate, Feedwater or Steam leak), then go to E2, Loss of Secondary Inventory.

2014 NRC Exam Scenario 1 D-2 Rev 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 2 Page 6 of 30 Event

Description:

Atmospheric Dump Valve #1 spuriously opens Time Position Applicants Actions or Behavior OP-901-221, Secondary System Transient, E2 - Loss of Secondary Inventory ATC/BOP

1. Verify the following:

1.1 Pressurizer Pressure Control System maintaining or restoring Pressurizer pressure to 2250 PSIA.

1.2 Pressurizer Level Control System maintaining or restoring Pressurizer level to program level.

1.3 Steam Generator levels being maintained or restored to 50% to 70% Narrow Range level.

1.4 Steam Bypass Control System responding to maintain Steam Generator pressure.

N/A 2. Dispatch an operator and determine leak severity and location.

SRO 3. Attempt to safely isolate the leak.

3.1 If a Atmospheric Dump Valve is isolated or associated controller placed in manual, then refer to the Technical Specification 3.7.1.7, Atmospheric Dump Valves.

SRO Reviews the following Technical Specifications and determines applicable actions:

3.7.1.7 action a Examiner Note This event is complete after the BOP places controller MS-IPIC-0303-A1 in manual and closes ADV #1 and the SRO has evaluated Technical Specifications or At Lead Examiners Discretion Examiner Note Cue the Simulator Operator when ready for Event 3 2014 NRC Exam Scenario 1 D-2 Rev 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 3 Page 7 of 30 Event

Description:

Control Element Assembly #9 drops into the core Time Position Applicants Actions or Behavior ATC / BOP Recognize and report indications of dropped CEA Alarms RPS CHANNEL TRIP LOCAL PWR DENSITY HI (Cabinet K, A-11)

RPS CHANNEL TRIP DNBR LO (Cabinet K, A-12)

LOCAL PWR DENSITY HI PRETRIP B/D (Cabinet K, C-11)

DNBR LO PRETRIP B/D (Cabinet K, C-12)

CEA CHANNEL B DEVIATION (Cabinet K, H-12)

CEA CHANNEL C DEVIATION (Cabinet K, K-13)

CEA CALCULATOR CHNL B TROUBLE (Cabinet K, K-15)

CEA CALCULATOR CHNL C TROUBLE (Cabinet K, K-16)

RPS CHANNEL D TROUBLE (Cabinet K, H-18)

Indications Rod bottom light for CEA 9 TCOLD dropping LPD and DNBR trips on Channel D (targeted channel)

Examiner Note The CRS may direct a Main Turbine load reduction before entering the off normal procedure to raise T COLD. It is not required to do this before entering OP-901-102. This is not a deficiency if the CRS does not perform this step at this time. If the CRS performs this step, the guidance is located on page 8.

CRS Enter and direct the implementation of OP-901-102, CEA or CEDMCS Malfunction.

OP-901-102, CEA or CEDMCS Malfunction, Immediate Operator Actions Examiner Note This is the immediate action for this procedure. This is not applicable to this scenario.

N/A 1. If in Mode 1 and two or more Control Element Assemblies drop or are misaligned by >19 inches, then manually trip the Reactor and go to OP-902-000, Standard Post Trip Actions.

OP-901-102, CEA or CEDMS Malfunction, E0, General N/A 2. If any of the following occur, then manually trip the Reactor and go to OP-902-000, Standard Post Trip Actions:

SRO 3. If Control Element Assembly is misaligned >7 inches, then go to section E1, CEA Misalignment Greater Than 7 Inches.

2014 NRC Exam Scenario 1 D-2 Rev 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 3 Page 8 of 30 Event

Description:

Control Element Assembly #9 drops into the core Time Position Applicants Actions or Behavior OP-901-102, CEA or CEDMS Malfunction, E1, CEA Misalignment Greater than 7 inches SRO 1. Match Tavg and Tref by performing the following:

Adjust Turbine load in accordance with OP-010-004, Power Operations.

Adjust RCS boron concentration in accordance with OP-002-005, Chemical and Volume Control.

OP-005-007, Main Turbine and Generator CAUTION THE FOLLOWING SECTION HAS THE POTENTIAL TO AFFECT CORE REACTIVITY. REACTOR POWER, RCS TEMPERATURE, AND MAIN GENERATOR MW LOAD SHOULD BE CLOSELY MONITORED DURING PERFORMANCE OF THIS SECTION.

BOP 6.2.1 To change Load/Rate perform the following:

6.2.1.1 Depress LOAD/RATE MW/MIN pushbutton.

6.2.1.2 Depress appropriate numerical pushbuttons for desired load rate.

6.2.1.3 Depress ENTER pushbutton.

NOTE Prior to changing Reference Demand, Main Turbine load must not be changing.

BOP 6.2.2 To change Main Turbine load, perform the following:

6.2.2.1 Depress REF pushbutton.

6.2.2.2 Depress appropriate numerical pushbuttons for desired MW load.

6.2.2.3 Depress ENTER pushbutton.

6.2.2.4 Depress GO pushbutton.

6.2.2.5 Verify Turbine load change stops at the desired MW load.

OP-901-102, CEA or CEDMS Malfunction, E1, CEA Misalignment Greater than 7 inches SRO 2. Notify Duty Plant Manager and Duty Engineering.

SRO 3. Record time of CEA misalignment >7 inches in Station Log.

SRO 4. If CEA misalignment >19 inches, then go to step 8.

2014 NRC Exam Scenario 1 D-2 Rev 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 3 Page 9 of 30 Event

Description:

Control Element Assembly #9 drops into the core Time Position Applicants Actions or Behavior CAUTION A POWER REDUCTION MUST BE STARTED WITHIN 15 MINUTES OF CEA MISALIGNMENT

>7 INCHES TO COMPLY WITH TECH SPEC 3.1.3.1.

Examiner Note The SRO should determine that a power reduction is required within 60 minutes.

NOTE (1) Complete the required down-power prior to withdrawal of the affected CEA.

(2) If a Shutdown Bank CEA has dropped or is misaligned and is <145 inches withdrawn, then CEA Group Out of Sequence annunciator (A-7 on Cabinet L) is an expected annunciator when performing ASI Control using Regulating Group CEAs.

SRO 8. If misalignment >19 inches or affected CEA is not aligned to within 7 inches of all other CEAs in the same group within 15 minutes, then perform the following:

Reduce power in accordance with OP-901-212, Rapid Plant Power Reduction to comply with Technical Specification 3.1.3.1.

Maintain Tavg at Tref by adjusting turbine load If PMC is Operable, then verify CEA Pulse Counter indication is correct or enter the correct CEA position in the PMC database.

Declare COLSS Inoperable and enter OP-901-501, PMC or COLSS Inoperable and perform concurrently with this procedure due to COLSS being Inoperable.

Use SEC CAL PWR (C24230), CBTFSP (C24102), BDELT (C24104), CBDELT (C24103), or TURB PWR (C24101) for indication during power reduction.

Examiner Note When a full crew complement is present, OP-901-501, PMC or COLSS inoperable would be performed by the STA.

2014 NRC Exam Scenario 1 D-2 Rev 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 3 Page 10 of 30 Event

Description:

Control Element Assembly #9 drops into the core Time Position Applicants Actions or Behavior OP-901-212, Rapid Plant Power Reduction, E0, General Actions Examiner Note A Rapid downpower does not have to be started at 30MW/min, but should attempt to eventually reach that value. The crew will likely not start the load reduction at that high of a rate.

NOTE (1) A rapid power reduction is defined as approximately 30 MW/minute load reduction on the main turbine.

(2) Power Reduction may be stopped at any point.

(3) Some Steps of this procedure may not be applicable due to plant conditions. In these cases SM/CRS may NA the step.

(4) Steps within this procedure may be performed concurrently or out of sequence with SM/CRS concurrence.

(5) During power reduction PMC PID C24650, COLSS DESCENDING PWR TRACK (DUMOUT19),

will automatically select and display the correct power indication.

OP-010-003, Plant Startup, provides greater detail on which power indications are displayed by PID C24650 based on power level and whether or not the UFM is in service.

(6) Volume Control Tank (VCT) level may lower during the down power. Charging pump suction swaps to the RWSP at 8% VCT level. Makeup to the VCT in accordance with OP-002-005, Chemical and Volume Control, may be necessary if boration from the RWSP is not desired.

ATC 1. Begin RCS Boration by one of the following methods:

1.1 Direct Boration or 1.2 Borate from the RWSP using one or two Charging Pump as follows:

1.2.1 Open RWSP to Charging Pumps Suction Isolation, CVC-507.

1.2.2 Close Volume Control Tank Outlet Isolation, CVC-183.

1.2.3 If necessary, then start another Charging pump 2014 NRC Exam Scenario 1 D-2 Rev 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 3 Page 11 of 30 Event

Description:

Control Element Assembly #9 drops into the core Time Position Applicants Actions or Behavior OP-005-005, Chemical and Volume Control CAUTION (1) THIS SECTION AFFECTS REACTIVITY. THIS EVOLUTION SHOULD BE CROSS-CHECKED AND COMPLETED PRIOR TO LEAVING CP-4.

(2) AT LEAST ONE REACTOR COOLANT PUMP IN EACH LOOP SHOULD BE OPERATING PRIOR TO PERFORMING DIRECT BORATION OPERATIONS TO ENSURE PROPER CHEMICAL MIXING.

ATC 6.7.1 Inform SM/CRS that this Section is being performed.

NOTE When performing a Plant down power where final RCS Boron Concentration needs to be determined, the following Plant Data Book figure(s) will assist the Operator in determining the required RCS Boron PPM change.

1.2.1.1 Power Defect Vs Power Level 1.4.3.1 Inverse Boron Worth Vs. Tmod at BOC (<30 EFPD) 1.4.4.1 Inverse Boron Worth Vs. Tmod at Peak Boron (30 EFPD up to 170 EFPD) 1.4.5.1 Inverse Boron Worth Vs. Tmod at MOC (170 EFPD up to 340 EFPD) 1.4.6.1 Inverse Boron Worth Vs. Tmod at EOC ( 340 EFPD)

ATC 6.7.2 At SM/CRS discretion, calculate volume of Boric Acid to be added on Attachment 11.6, Calculation of Boric Acid Volume for Direct Boration or VCT Borate Makeup Mode.

ATC 6.7.3 Set Boric Acid Makeup Batch Counter to volume of Boric Acid desired.

ATC 6.7.4 Verify Boric Acid Makeup Pumps selector switch aligned to desired Boric Acid Makeup Pump A(B).

ATC 6.7.5 Place Direct Boration Valve, BAM-143, control switch to AUTO.

ATC 6.7.6 Place Makeup Mode selector switch to BORATE.

ATC 6.7.7 Verify selected Boric Acid Makeup Pump A(B) Starts.

ATC 6.7.8 Verify Direct Boration Valve, BAM-143, Opens.

2014 NRC Exam Scenario 1 D-2 Rev 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 3 Page 12 of 30 Event

Description:

Control Element Assembly #9 drops into the core Time Position Applicants Actions or Behavior NOTE The Boric Acid Flow Totalizer will not register below 3 GPM. The Boric Acid Flow Totalizer is most accurate in the range of 10 - 25 GPM.

ATC 6.7.9 If manual control of Boric Acid flow is desired, then perform the following:

6.7.9.1 Verify Boric Acid Flow controller, BAM-IFIC-0210Y, in Manual.

6.7.9.2 Adjust Boric Acid Flow controller, BAM-IFIC-0210Y, output to >3 GPM flow rate.

N/A 6.7.10 If automatic control of Boric Acid flow is desired, then perform the following:

6.7.10.1 Place Boric Acid Flow controller, BAM-IFIC-0210Y, in Auto.

6.7.10.2 Adjust Boric Acid Flow controller, BAM-IFIC-0210Y, setpoint potentiometer to >3 GPM flow rate.

ATC 6.7.11 Verify Boric Acid Makeup Control Valve, BAM-141, Intermediate or Open.

ATC 6.7.12 Observe Boric Acid flow rate for proper indication.

ATC 6.7.13 When Boric Acid Makeup Batch Counter has counted down to desired value, then verify Boric Acid Makeup Control Valve, BAM-141, Closed.

2014 NRC Exam Scenario 1 D-2 Rev 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 3 Page 13 of 30 Event

Description:

Control Element Assembly #9 drops into the core Time Position Applicants Actions or Behavior Examiner Note The step below is only applicable if the ATC adds acid in small batches.

NOTE Step 6.7.14 may be repeated as necessary to achieve desired total boron addition for plant conditions.

ATC 6.7.14 If additional boric acid addition is required and with SM/CRS permission, then perform the following:

6.7.14.1 Reset Boric Acid Makeup Batch Counter.

6.7.14.2 Verify Boric Acid Makeup Control Valve, BAM-141, Intermediate or Open.

6.7.14.3 Observe Boric Acid flow rate for proper indication.

6.7.14.4 When Boric Acid Makeup Batch Counter has counted down to desired value, then verify Boric Acid Makeup Control Valve, BAM-141, Closed.

ATC 6.7.15 Verify Boric Acid Flow controller, BAM-IFIC-0210Y, in Manual.

ATC 6.7.16 Verify both Boric Acid Flow controller, BAM-IFIC-0210Y, output and setpoint potentiometer set to zero.

ATC 6.7.17 Place Makeup Mode selector switch to MANUAL.

ATC 6.7.18 Verify Selected Boric Acid Makeup Pump A (B) Stops.

ATC 6.7.19 Verify Direct Boration Valve, BAM-143, Closed.

ATC 6.7.20 Place Direct Boration Valve, BAM-143, control switch to CLOSE.

2014 NRC Exam Scenario 1 D-2 Rev 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 3 Page 14 of 30 Event

Description:

Control Element Assembly #9 drops into the core Time Position Applicants Actions or Behavior OP-901-212, Rapid Plant Power Reduction, E0, General Actions NOTE To prevent Pressurizer heater cutout, avoid operating with Pressurizer pressure near the heater cutout pressure of 2270 PSIA while on Boron Equalization.

ATC 2. Perform Boron Equalization as follows:

2.1 Place available Pressurizer Pressure Backup Heater Control Switches to ON.

2.2 Reduce Pressurizer Spray Valve Controller (RC-IHIC-0100) setpoint potentiometer to establish spray flow and maintain RCS pressure 2250 PSIA (2175 - 2265).

CAUTION REFER TO TECHNICAL SPECIFICATION 3.1.3.6 FOR TRANSIENT INSERTION LIMITS.

ATC 3. Operate CEAs in accordance with OP-004-004, Control Element Drive, to maintain ASI using CEA Reg. Group 5, 6 or Group P Control Element Assemblies in accordance with OP-010-005, Plant Shutdown, Attachment 9.10, Axial Shape Control Guidelines.

SRO 4. Notify the Woodlands System Load Dispatcher that a rapid power reduction is in progress.

BOP 5. Announce to Station Personnel over the Plant Paging System that a rapid plant power reduction is in progress.

ATC 6. Maintain RCS Cold Leg Temperature 536 F to 549 F.

BOP 7. Commence Turbine load reduction by performing the following:

7.1 Depress LOAD RATE MW/MIN pushbutton.

7.2 Set selected rate in Display Demand Window.

7.3 Depress ENTER pushbutton.

7.4 Depress REFERENCE pushbutton.

7.5 Set desired load in Reference Demand Window.

7.6 Depress ENTER pushbutton.

7.7 Depress GO pushbutton.

2014 NRC Exam Scenario 1 D-2 Rev 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 3 Page 15 of 30 Event

Description:

Control Element Assembly #9 drops into the core Time Position Applicants Actions or Behavior NOTE If USBSCAL is not in service, the COLSS Steam Calorimetric will be automatically disabled when MSBSCAL, PMC PID C24246, drops below 95% Power, and will revert back to FWBSCAL, PMC PID C24235. This may result in a step change in COLSS indicated Plant Power (BSCAL) of up to 1.0% when this occurs.

ATC 8. When Reactor Power consistently indicates less than 98% power, as indicated on PMC PID C24631 [MAIN STEAM RAW POWER (MSBSRAW)], or an alternate point provided by Reactor Engineering, then verify the value of C24648 [BSCAL SMOOTHING VAL. APPLD (DUMOUT17)] automatically changes to 1.

N/A 9. If C24648 does not automatically change to 1, then inform Reactor Engineering and set the value of 1 for COLSS power smoothing constant K24250, [ADDRSSBL SMOOTHING FOR BSCAL (ALPHA)] in accordance with OP-004-005, Core Operating Limits Supervisory System.

BOP 10. Following a Reactor Power change of >15% within a one hour period, direct Chemistry Department to sample Reactor Coolant System (RCS) for an isotopic iodine analysis two to six hours later.

BOP 11. When Condensate flow is <18,000 gpm, verify Gland Steam Condenser Bypass, CD-154, Closed (PMC PID D02404).

BOP 12. Monitor Condensate Polisher differential pressure and remove Polishers from service to maintain system pressure in accordance with OP-003-031, Condensate Polisher/Backwash Treatment.

N/A 13. When Reactor Power is approximately 70% or Heater Drain Pump flow is unstable, then remove Heater Drain Pumps from service by taking pump control switches to Stop.

OP-901-102, CEA or CEDMS Malfunction, E1, CEA Misalignment Greater than 7 inches SRO 9. If either of the following occur, then within one hour declare CEA Inoperable and verify acceptable Shutdown Margin in accordance with OP-903-090, Shutdown Margin:

Misaligned CEA is in Shutdown Bank and cannot be withdrawn to 145 inches [T.S. 3.1.3.5]

or Any CEA is misaligned from its group by >7 inches and cannot be aligned [T.S. 3.1.3.1]

2014 NRC Exam Scenario 1 D-2 Rev 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 3 Page 16 of 30 Event

Description:

Control Element Assembly #9 drops into the core Time Position Applicants Actions or Behavior SRO Reviews the following Technical Specifications and determines applicable actions:

3.1.3.1 action c ATC 10. Monitor the following for compliance with Technical Specifications:

Linear Heat Rate (LPD) [T.S. 3.2.1]

Departure from Nucleate Boiling Ratio [T.S. 3.2.4]

Azimuthal Power Tilt [T.S. 3.2.3]

ASI [T.S. 3.2.7]

ATC 11. Perform the following to verify compliance with Technical Specification 3.2.3:

Verify COLSS is detecting Azimuthal Power Tilt as influenced by the misaligned CEA.

If measured Azimuthal Power Tilt exceeds 0.03, then verify Azimuthal Power Tilt within limits of Technical Specification 3.2.3 If measured Azimuthal Power Tilt exceeds CPC Power Tilt Allowance, then adjust CPC Power Tilt allowance to the measured value in accordance with OP-004-006, Core Protection Calculator System.

NOTE If the CEA misalignment is due to a dropped rod then the ACTM card for the dropped rod may need to be reset if the CEA is not responding to a withdrawal demand.

ATC 12. Maintain ASI within 0.05 of target ESI to limit potential impact of transient Xenon on core peaking using Manual Group or Manual Sequential.

N/A 13. Correct cause of CEA misalignment.

BOP 14. If initial Reactor power is 75 %, then notify Chemistry Department to sample Reactor Coolant System for an isotopic iodine analysis within two to six hours due to a power reduction of 15 % in one hour.

Examiner Note This event is complete after the Reactivity Manipulation is satisfied and the SRO has evaluated Technical Specifications OR At Lead Examiners Discretion 2014 NRC Exam Scenario 1 D-2 Rev 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 3 Page 17 of 30 Event

Description:

Control Element Assembly #9 drops into the core Time Position Applicants Actions or Behavior Examiner Note Cue the Simulator Operator when ready for Event 4 Examiner Note Event 5 and 6 will occur automatically after the crew takes action responding to Event 4.

2014 NRC Exam Scenario 1 D-2 Rev 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 4 Page 18 of 30 Event

Description:

Main Feedwater Pump Trip / Reactor Cutback leads to Reactor Trip Time Position Applicants Actions or Behavior ATC / BOP Recognize and report indications of MFP Trip and Reactor Cutback Alarms FWPT B OIL PUMP TROUBLE (Cabinet F, R-18)

REACTOR PWR CUTBACK ACTUATION (Cabinet H, K-5)

FWPT B FLOW LO (Cabinet F, M-20)

FWPT B FLOW LO-LO (Cabinet F, N-20)

FWPT B DISCH PRESS/FLOW MONITOR FAIL (Cabinet F, P-20)

SG 2 LEVEL HI/LO (Cabinet F, U-18)

SG1 LEVEL HI/LO (Cabinet F, U-14)

HTR DRAIN PUMP A(B)(C) TRIP/TROUBLE [Cabinet F, P-11(12)(13)]

HTR DRAIN PUMP A(B)(C) SUCTION PRESS LO [Cabinet F, Q-11(12)(13)]

HTR DRAIN PUMP A FLOW LO (Cabinet F, R-11(12)(13)

Indications Rod bottom lights for CEAs 44, 21, 45, 22, 46, 23, 47, 20 TCOLD dropping Reactor Power Dropping CRS Enter and direct the implementation of OP-901-101, Reactor Power Cutback.

Examiner Note The SRO should direct a manual Reactor Trip due to incorrect rod configuration following the Reactor Power Cutback ATC Trip the Reactor and enter OP-902-000, Standard Post Trip Actions Examiner Note The subsequent actions contained in Event 4 will be contained after events 5 and 6 which occur in order of how they will be performed during the scenario.

2014 NRC Exam Scenario 1 D-2 Rev 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 5 Page 19 of 30 Event

Description:

3 CEAs fail to insert on the Reactor Trip Time Position Applicants Actions or Behavior ATC Recognize and report indications of 3CEAs fail to insert Alarms N/A Indications Rod bottom lights for CEAs 7, 37, 39 extinguished CEA position for CEA 7, 37, 39 showing withdrawn OP-902-000, Standard Post Trip Actions ATC 1. Determine Reactivity Control acceptance criteria are met:

a. Check reactor power is dropping.

b. Check startup rate is negative.

c. Check less than TWO CEAs are NOT fully inserted.

ATC c.1 Commence emergency boration.

CRITICAL TASK ESTABLISH REACTIVITY CONTROL This task is satisfied by commencing Emergency Boration flow by either Boric Acid makeup pumps or Gravity Feed valves in accordance with OP-902-000, Standard Post Trip Actions step 1 prior to exiting the step to verify Reactivity Control.

This task becomes applicable following the initiation of a Reactor Trip.

Examiner Note The steps below are contained on an Operator Aid on the RTGB ATC 1. Place Makeup Mode selector switch to MANUAL. (CRITICAL)

ATC 2. Align borated water source by performing one of the following (a. or b.):

(CRITICAL)

a. Initiate Emergency Boration using Boric Acid Pump as follows:

Open Emergency Boration Valve, BAM-133.

Start one Boric Acid Pump.

Close recirc valve for Boric Acid Pump started:

BAM-126A Boric Acid Makeup Pump Recirc Valve A or BAM-126B Boric Acid Makeup Pump Recirc Valve B 2014 NRC Exam Scenario 1 D-2 Rev 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 5 Page 20 of 30 Event

Description:

3 CEAs fail to insert on the Reactor Trip Time Position Applicants Actions or Behavior ATC b. Initiate Emergency Boration using Gravity Feed as follows:

Open the following Boric Acid Makeup Gravity Feed valves:

BAM-113A Boric Acid Makeup Gravity Feed Valve A BAM-113B Boric Acid Makeup Gravity Feed Valve B ATC 3. Close VCT Disch Valve, CVC-183. (CRITICAL)

ATC 4. Verify at least one Charging Pump operating and Charging Header flow

>40 GPM.

Examiner Note This event is complete after the ATC has established emergency boration OR At Lead Examiners Discretion 2014 NRC Exam Scenario 1 D-2 Rev 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 6 Page 21 of 30 Event

Description:

Turbine Fails to Automatically Trip Time Position Applicants Actions or Behavior BOP Recognize indications of Main Turbine failing to trip Alarms N/A Indications Main Turbine Governor and Throttle valves indicate open TCOLD lowering RCS pressure lowering OP-902-000, Standard Post Trip Actions BOP 2. Determine Maintenance of Vital Auxiliaries acceptance criteria are met:

a. Check the Main Turbine is tripped:

Governor valves closed Throttle valves closed CRITICAL TASK ESTABLISH REACTIVITY CONTROL This task is satisfied by stopping the steam flow to the Main Turbine by either tripping the Main Turbine manually or by closing the Main Steam Isolation valves before TCOLD lowers to below 443 F.

This task becomes applicable following the initiation of the Reactor Trip signal.

BOP a.1 Perform ANY of the following: (CRITICAL)

1) Manually trip the Main Turbine using TURBINE TRIP and THINK pushbuttons.

2) Close BOTH MSIVs.

Examiner Note This event is complete after the BOP has stopped steam flow to the Main Turbine OR At Lead Examiners Discretion 2014 NRC Exam Scenario 1 D-2 Rev 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 4 Page 22 of 30 Event

Description:

Main Feedwater Pump Trip / Reactor Cutback leads to Reactor Trip Time Position Applicants Actions or Behavior BOP b. Check the Main Generator is tripped:

GENERATOR BREAKER A tripped GENERATOR BREAKER B tripped EXCITER FIELD BREAKER tripped BOP c. Check station loads are energized from offsite electrical power as follows:

Train A A1, 6.9 KV non safety bus A2, 4.16 KV non safety bus A3, 4.16 KV safety bus A-DC electrical bus A or C vital AC Instrument Channel Train B B1, 6.9 KV non safety bus B2, 4.16 KV non safety bus B3, 4.16 KV safety bus B-DC electrical bus B or D vital AC Instrument Channel ATC 3. Determine RCS Inventory Control acceptance criteria are met:

a. Check that BOTH the following conditions exist:

Pressurizer level is 7% to 60%

Pressurizer level is trending to 33% to 60%

b. Check RCS subcooling is greater than or equal to 28ºF.

Examiner Note If the Main Turbine manual trip is delayed in Event 6, this may have an effect on RCS Pressure and Temperature and PZR level which will make some of the following steps applicable. It is possible that a Safety Injection and Containment isolation could occur, but this will not impact the flow of the scenario unless a subsequent misdiagnosis is made by the crew.

ATC 4. Determine RCS Pressure Control acceptance criteria are met by checking that BOTH of the following conditions exist:

Pressurizer pressure is 1750 psia to 2300 psia Pressurizer pressure is trending to 2125 psia to 2275 psia 2014 NRC Exam Scenario 1 D-2 Rev 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 4 Page 23 of 30 Event

Description:

Main Feedwater Pump Trip / Reactor Cutback leads to Reactor Trip Time Position Applicants Actions or Behavior ATC 5. Determine Core Heat Removal acceptance criteria are met:

a. Check at least one RCP is operating.

b. Check operating loop T is less than 13ºF.

c. Check RCS subcooling is greater than or equal to 28ºF.

Examiner Note At any point prior to event 8, the crew may notice that Steam Generator levels have reduced below 27% NR and an Emergency Feedwater Actuation Signal failed to occur. If the operator manually initiates EFAS at this time, it will satisfy the critical task in Event 8. This is not required to be performed at this time but only prior to lowering level to below 20% WR.

BOP 6. Determine RCS Heat Removal acceptance criteria are met:

a. Check that at least one steam generator has BOTH of the following:

Steam generator level is 10% to 76% NR Main Feedwater is available to restore level within 55%-70%

NR.[60% to 80% NR]

ATC b. Check RCS TC is 530 ºF to 550 ºF BOP c. Check steam generator pressure is 885 psia to 1040 psia.

BOP d. Check Feedwater Control in Reactor Trip Override:

MAIN FW REG valves are closed STARTUP FW REG valves are 13% to 21% open Operating main Feedwater pumps are 3800 rpm to 4000 rpm BOP e. Reset moisture separator reheaters, and check the temperature control valves closed.

ATC 7. Determine Containment Isolation acceptance criteria are met:

a. Check containment pressure is less than 16.4 psia.

b. Check NO containment area radiation monitor alarms OR unexplained rise in activity.

c. Check NO steam plant activity monitor alarms OR unexplained rise in activity.

BOP 8 Determine Containment Temperature and Pressure Control acceptance criteria are met:

a. Check containment temperature is less than or equal to 120ºF.

b. Check containment pressure is less than 16.4 psia.

2014 NRC Exam Scenario 1 D-2 Rev 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 4 Page 24 of 30 Event

Description:

Main Feedwater Pump Trip / Reactor Cutback leads to Reactor Trip Time Position Applicants Actions or Behavior N/A 9. IF ALL safety function acceptance criteria are met, AND NO contingency actions were performed, THEN GO TO OP-902-001, "Reactor Trip Recovery" procedure.

SRO 10. IF ANY safety function acceptance criteria are NOT met, OR ANY contingency action was taken, THEN GO TO Appendix 1, "Diagnostic Flowchart.

Examiner Note Cue the Simulator Operator when ready for Event 7 Examiner Note This event is complete after the SRO has commenced diagnosis in OP-902-009, Standard Appendices OR At Lead Examiners Discretion 2014 NRC Exam Scenario 1 D-2 Rev 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 7 Page 25 of 30 Event

Description:

Feedwater Pump A overspeed trip / Loss of Main Feedwater Time Position Applicants Actions or Behavior BOP Recognize and report indications of Feedwater Pump A trip Alarms FWPT A TRIP OVERSPEED (Cabinet F, K-15)

FWPT A FLOW LO (Cabinet F, M-16)

FWPT A FLOW LO-LO (Cabinet F, N-16)

Indications Steam Generator levels lowering Feedwater flow lowering Feedwater Pump speed lowering OP-902-009, Standard Appendices, Appendix 1 Examiner Note Appendix 1 is a flow chart used to diagnose to the correct recovery procedure for the event in progress. The steps below will be followed by a YES or NO to indicate proper flow path.

ATC Rx Pwr dropping, SUR negative, and < two CEAs NOT fully inserted or Emergency Boration in progress (YES)

ATC Pressurizer pressure dropping rapidly and Pressurizer level changing (NO)

SRO Page 2 BOP At least one 7 KV non safety bus and one 4 KV safety bus powered from offsite (same train) (YES)

BOP MFW restoring or maintaining at least one SG level 55% to 70% NR (NO)

SRO GO TO OP-902-006 Loss of Main Feedwater Recovery OP-902-006, Loss of Main Feedwater Recovery SRO Confirm Diagnosis

• 1. Confirm diagnosis of a Loss of Main Feedwater by checking Safety Function Status Check Acceptance Criteria are satisfied.

BOP Announce the Event

2. Announce the Loss of Main Feedwater in progress using the plant page.

2014 NRC Exam Scenario 1 D-2 Rev 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 7 Page 26 of 30 Event

Description:

Feedwater Pump A overspeed trip / Loss of Main Feedwater Time Position Applicants Actions or Behavior N/A Classify the Event

• 3. Advise the Shift Manager to REFER TO EP-001-001, "Recognition &

Classification of Emergency Condition", and implement the Emergency Plan.

N/A Restore Operation of DCT Sump Pumps

• 4. IF power has been interrupted to either 3A or 3B safety buses, THEN perform Appendix 20, "Operation of DCT Sump Pumps".

SRO Implement Placekeeping

5. REFER TO Section 6.0, "Placekeeper", and record the time of the reactor trip.

ATC Verify ONE RCP In Each Loop Secured

6. Verify no more than two RCPs are operating.

N/A Stop ALL RCPs

• 7. IF ANY of the following conditions exist, THEN perform the following:

a. IF ALL feedwater is lost, THEN stop ALL RCPs.

b. IF MFW is lost AND the only feedwater available is ONE motor driven EFW Pump, THEN within the following 30 minutes stop ALL RCPs.

BOP Verify Proper CCW Operation

• 8. Check a CCW pump is operating for each energized 4.16 KV safety bus.

BOP Conserve Steam Generator Inventory

• 9. Conserve steam generator inventory:

a. Verify the following steam generator blowdown isolation valves are closed:

• BD 102A, SG BLOWDOWN ISOL STM GEN 1 (IN)

• BD 102B, SG BLOWDOWN ISOL STM GEN 2 (IN)

• BD 103A, SG BLOWDOWN ISOL STM GEN 1 (OUT)

• BD 103B, SG BLOWDOWN ISOL STM GEN 2 (OUT) 2014 NRC Exam Scenario 1 D-2 Rev 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 7 Page 27 of 30 Event

Description:

Feedwater Pump A overspeed trip / Loss of Main Feedwater Time Position Applicants Actions or Behavior BOP b. Verify the following steam generator sampling valves are closed:

• SSL 8006A, SAMPLING ISOLATION SG 1

• SSL 8006B, SAMPLING ISOLATION SG 2

• SSL 301A, SAMPLING ISOLATION MAIN STM LINE 1

• SSL 8004A, SAMPLING ISOLATION SG 1

• SSL 8004B, SAMPLING ISOLATION SG 2

• SSL 301B, SAMPLING ISOLATION MAIN STM LINE 2 N/A Reset MSIS

• 10. IF MSIS has actuated AND opening the MSIVs or MFIVs will aid in the restoration of feedwater, THEN REFER TO Appendix 5-B, "MSIS Main Steam Pressure Reset Procedure", and reset the MSIS.

Examiner Note The Loss of Feedwater event is complete after the crew has stopped 2 Reactor Coolant Pumps and has completed Event 8 OR At Lead Examiners Discretion 2014 NRC Exam Scenario 1 D-2 Rev 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 8 Page 28 of 30 Event

Description:

Emergency Feedwater Fails to Actuate Time Position Applicants Actions or Behavior BOP Recognize and report indications of EFAS fail to actuate Alarms Absence of Alarms:

EFAS-1 TRAIN A LOGIC INITIATED (Cabinet K, M-19)

EFAS-2 TRAIN A LOGIC INITIATED (Cabinet K, N-19)

EFAS-1 TRAIN B LOGIC INITIATED (Cabinet K, M-20)

EFAS-2 TRAIN B LOGIC INITIATED (Cabinet K, N-20)

Indications Steam Generator levels <27.4% NR EFW Pump A, B, AB not operating CRITICAL TASK ESTABLISH RCS HEAT REMOVAL This task is satisfied by manually actuating Emergency Feedwater Actuation System, manually starting and aligning at least 1 Emergency Feedwater pump, aligning and starting the Auxiliary Feedwater pump, or depressurizing at least 1 Steam Generator and injecting the Steam Generator with Condensate flow prior to Steam Generator levels reducing below 20% Wide Range in both Steam Generators.

This task becomes applicable once Steam Generator water level reduces below 27.4% Narrow Range in one or both Steam Generators.

BOP Restore Steam Generator Inventory

• 11. Replenish inventory in at least one steam generator by performing ANY of the following:

a. Check emergency Feedwater available to at least one steam generator.

b. REFER TO Appendix 32, "Establishing Main Feedwater" and restore main feedwater flow to at least one steam generator.

BOP Actuate EFAS - 1 and EFAS - 2. (CRITICAL)

Examiner Note The following steps are added as a reference if the crew continues on in OP-902-006 and has not yet actuated Emergency Feedwater. These steps describe alternative methods to establishing RCS heat removal. (Appendix 32 A & B) 2014 NRC Exam Scenario 1 D-2 Rev 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 8 Page 29 of 30 Event

Description:

Emergency Feedwater Fails to Actuate Time Position Applicants Actions or Behavior OP-902-009, Standard Appendices, Appendix 32A Starting Auxiliary Feedwater Pump NOTE When AFW is being used to feed the steam generators, consideration should be given to aligning the condenser to the CST.

BOP 1. Verify MAIN FW ISOL VALVES open:

FW-184A FW-184B BOP 2. Place the following controllers to "MAN" and set to minimum:

FW-IFIC-8202, AUX FEEDWATER FLOW FW-IHIC-1105, SG 1 STARTUP FW REG FW-IHIC-1106, SG 2 STARTUP FW REG BOP 3. Verify the following valves closed:

FW-170A, SG 1 MAIN FW REG ISOLATION FW-170B, SG 2 MAIN FW REG ISOLATION BOP 4. Verify the following valves open:

FW-163A, SG 1 STARTUP FW REG ISOLATION FW-163B, SG 2 STARTUP FW REG ISOLATION BOP 5. Start Auxiliary Feedwater Pump.

BOP 6. Adjust the following controllers to 5% output:

FW-IHIC-1105, SG 1 STARTUP FW REG FW-IHIC-1106, SG 2 STARTUP FW REG NOTE AFW pump discharge pressure should be monitored closely while adjusting flow.

BOP 7. Adjust the following controllers and valve to achieve desired flow rates:

• FW-IHIC-1105, SG 1 STARTUP FW REG

• FW-IHIC-1106, SG 2 STARTUP FW REG

• FW-IFIC-8202, AUX FEEDWATER FLOW AFW-125, AUXILIARY FEEDWATER PRESSURE RELIEF OP-902-009, Standard Appendices, Appendix 32B Aligning Condensate to Feed the Steam Generators BOP 1. Verify MFW pumps are tripped.

BOP 2. Verify AFW pump secured.

2014 NRC Exam Scenario 1 D-2 Rev 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 8 Page 30 of 30 Event

Description:

Emergency Feedwater Fails to Actuate Time Position Applicants Actions or Behavior BOP 3. Verify at least one Condensate Pump operating.

BOP 4. Open FW 125, FW PUMPS BYPASS valve.

N/A 5. IF MSIS has initiated, THEN REFER TO Appendix 5-B, "MSIS Main Steam Pressure Reset Procedure" and reset MSIS.

BOP 6. Verify MFIV open for selected steam generator.

BOP 7. Maintain SG level 55%-70% NR [60-80% NR] by adjusting the following controllers:

FW-IHIC-1105, SG 1 STARTUP FW REG FW-IHIC-1106, SG 2 STARTUP FW REG Examiner Note This event is complete after the crew has actuated EFAS OR At Lead Examiners Discretion 2014 NRC Exam Scenario 1 D-2 Rev 2

Appendix D Scenario Outline Form ES-D-1 Facility: Waterford 3 Scenario No.: 3 Op Test No.:

Examiners: Operators:

Initial Conditions: Reactor power is ~1%, MOC Turnover:

Protected Train is B, AB Bus is aligned to Train B, Raise power to 5%.

Event Malf. Event Event No. No. Type* Description R - ATC Secure the Auxiliary Feedwater Pump and raise power N - BOP to 5% in accordance with OP-003-035, Auxiliary 1 N/A N - SRO Feedwater and OP-010-003, Plant Startup.

Steam generator 1 Pressure Instrument SG-IPT-I - BOP 1013C, fails low requiring Technical Specification I - SRO entry and bypass of multiple PPS bistables. (TS 3.3.1, 2 SG04G TS - SRO 3.3.2)

C - BOP (A) Reactor Coolant Pump 1A lower seal fails. OP-3 RC08A C - SRO 901-130, Reactor Coolant Pump Malfunction I - ATC (A) PZR level instrument, RC-ILI-0110X, fails low I - SRO affecting letdown and heaters. OP-901-110, 4 RC15A2 TS - SRO Pressurizer Level Control Malfunction. (TS 3.3.3.5)

RP02A RP02B RCP 1A sustains a locked rotor and an automatic RP02C reactor trip does not occur. Manual action is needed RP02D C - ATC to trip the reactor (Critical Task 1, manually trip 5 RC03A M - ALL reactor prior to exiting step 1 of OP-902-000)

Loss of Coolant Accident occurs inside containment RC23A and valve CS-125A (Containment Spray Hdr A 6 CS04A M - ALL Isolation) fails closed Containment Spray Pump B trips and cannot be restarted requiring entry into OP-902-008, Functional Recovery and action taken to align Low Pressure Safety Injection pump to provide Containment Spray DI-08A04S22-1 C - BOP (Critical Task 2, align LPSI pump to replace CS 7 CS01B C - SRO pump prior to exiting Appendix 28 of OP-902-009)

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor 2014 NRC Exam Scenario 3 D-1 Rev 3

Scenario Event Description NRC Scenario 3 The crew assumes the shift at ~ 1% power with instructions to stop the running Auxiliary Feedwater Pump and raise power to 5%. All requirements have been met to change modes from MODE 2 to MODE 1.

The Shift Manager has given permission to change modes. The SRO should direct raising power using Control Element Assemblies in accordance with OP-010-003, Plant Startup.

After conducting the required reactivity manipulation, Steam Generator 1 Pressure Instrument, SG-IPT-1013C, fails low. The SRO should review and enter Technical Specifications 3.3.1 action 2 and 3.3.2 action 13 and direct the BOP to bypass the Steam Generator 1 Pressure Lo, Steam Generator 1 P, and Steam Generator 2 P trip bistables in Plant Protection System Channel C within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, in accordance with OP-009-007, Plant Protection System. The SRO should review Technical Specifications 3.3.3.5 and 3.3.3.6 and OP-903-013, Monthly Channel Checks, and determine that Technical Specification entry for 3.3.3.5 and 3.3.3.6 is not required.

After Technical Specifications are addressed, Reactor Coolant Pump 1A lower seal fails requiring entry into OP-901-130, Reactor Coolant Pump Malfunction and implement Section E1, Seal Failure. The SRO should direct the BOP to lower Controlled Bleed off temperature by lowering Component Cooling Water temperature.

After the crew is in Section E1 of OP-901-130 AND the BOP has adjusted Component Cooling Water Temperature, PZR level instrument, RC-ILI-0110X, fails low. The crew will implement OP-901-110, Pressurizer Level Control malfunction and select the non-faulted channel and reenergize Pressurizer heaters. The SRO should review Technical Specifications 3.3.3.5 and 3.3.3.6 and OP-903-013, Monthly Channel Checks, and determine that Technical Specification 3.3.3.5 action a is applicable and the channel needs to be restored within 7 days. The SRO should determine that Technical Specification 3.3.3.6 is not applicable as long as QSPDS channel for Pressurizer level is credited in accordance with OP-903-013, Monthly Channel Checks.

Once the SRO has addressed Technical Specifications, Reactor Coolant Pump 1A sustains a locked rotor and low Reactor Coolant flow bistables actuate. The Reactor Protection System fails to open the required Reactor Trip Breakers and an ATWS condition exists. The ATC should recognize that an automatic protection system has failed to occur and manually trip the reactor by depressing both Reactor Trip pushbuttons (A and D) on CP-2 (CRITICAL TASK 1). The Reactor will be successfully tripped from the RTGB and the crew will enter OP-902-000, Standard Post Trip Actions.

While the crew is performing Standard Post Trip Actions (RCS Heat Removal checks), a Loss of Coolant Accident will occur inside Containment. Safety Injection, Containment Isolation, and Containment Spray will all actuate. The ATC should secure all running Reactor Coolant Pumps when Containment Spray Actuates. When Containment Spray actuates, CS-125A (Containment Spray Hdr A Isolation) fails to open and cannot be opened from the RTGB or locally as it is mechanically stuck. This will result in Containment Spray Pump B as the only source of Containment Spray. The crew should enter OP-902-009, Standard Appendices Appendix 1, Diagnostic Flowchart and diagnose to OP-902-002, Loss of Coolant Accident Recovery.

Once the crew has entered OP-902-002, Containment Spray Pump B will trip and will not be able to be restarted. The crew should determine that Containment isolation and Containment Pressure and Temperature Control Safety Functions are not being met and diagnose into OP-902-008, Functional Recovery. The SRO should prioritize Containment Isolation first due to CS-125B being open and Containment Pressure and Temperature Control second. The crew will perform steps in OP-902-008, Functional Recovery and align the LPSI pump B to replace CS Pump B to establish Containment Temperature and Pressure Control (CRITICAL TASK 2).

The scenario can be terminated after the established Containment Spray flow from the Low Pressure Safety Injection Pump or at the lead examiners discretion.

2014 NRC Exam Scenario 3 D-1 Rev 3

NRC Scenario 3 CRITICAL TASKS

1. ESTABLISH REACTIVITY CONTROL This task is satisfied by shutting down the Reactor by either depressing both Reactor Trip pushbuttons on CP-2 or CP-8, initiating Diverse Reactor Trip Pushbuttons on CP-2, or by deenergizing the rod drive MG sets by deenergizing busses 32A and 32B on CP-1 prior to exiting the step to Verify Reactivity Control (Step 1) of OP-902-000, Standard Post Trip Actions. This task becomes applicable following the tripped Reactor Coolant Pump 1A.

2. ESTABLISH CONTAINMENT TEMPERATURE AND PRESSURE CONTROL This task is satisfied by aligning Low Pressure Safety Injection Pump B to replace Containment Spray Pump B in accordance with Appendix 28 of OP-902-009, Standard Appendices prior to exiting OP-902-009 Appendix 28. This task becomes applicable following the initiation of a Containment Spray Actuation Signal.

Scenario Quantitative Attributes

1. Total malfunctions (5-8) 6

2. Malfunctions after EOP entry (1-2) 1

3. Abnormal events (2-4) 2

4. Major transients (1-2) 2

5. EOPs entered/requiring substantive actions (1-2) 1

6. EOP contingencies requiring substantive actions (0-2) 1

7. Critical tasks (2-3) 2 2014 NRC Exam Scenario 3 D-1 Rev 3

NRC Scenario 3 SCENARIO SETUP A. Reset Simulator to IC-189.

B. Verify Scenario Malfunctions, Remotes, and overrides are loaded, as listed in the Scenario Timeline.

C. Ensure Protected Train B sign is placed in SM office window.

D. Verify PMC is set to MODE 2.

E. Verify EOOS is 10.0 Green F. Complete the simulator setup checklist.

G. Start Insight, open file Crew Performance.tis.

H. If scenario is allowed to run for 10 minutes, the last alarm should be HTR 4B Alt drain valve open (K-

6) on CP-13. Approximately 10 minutes later, another alarm will occur for HTR 6C (H-10) on CP-13 followed by SG level HI/LO. Allow simulator to run as long as desired and then place the simulator in freeze prior to starting the scenario.

2014 NRC Exam Scenario 3 D-1 Rev 3

NRC Scenario 3 SIMULATOR BOOTH INSTRUCTIONS Event 1 Secure AFW Pump and Raise power to 5%

1. No actions for this event.

Event 2 Steam Generator 1 Pressure Instrument SG-IPT-1013C, fails low

1. On Lead Examiner's cue, initiate Event Trigger 2.

2. If Work Week Manager or PMI are called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

Event 3 RCP 1A Lower Seal Fails

1. On Lead Examiner's cue, initiate Event Trigger 3.

2. If the Duty Engineering or RCP Engineer is called inform the caller that you will monitor RCP 1A for further degradation.

3. If the Work Week Manager or PMM are called, inform the caller that a work package will be assembled for the next forced outage.

Event 4 PZR level instrument, RC-ILI-0110X, fails low

1. On Lead Examiner's cue, initiate Event Trigger 4.

2. If called as NAO to check LCP-43, wait 3 minutes and report RC-ILI-0110-X1 is failed downscale.

3. If directed to check a started Charging pump for proper operation following start, wait 3 minutes and report the following:

a. Suction and discharge valves are open

b. Proper oil levels exist

c. No abnormal vibrations or noises present

4. If Work Week Manager, Computer Technician, or PMI are called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

Event 5 Reactor Coolant Pump 1A locked rotor

1. On Lead Examiner's cue, initiate Event Trigger 5.

Event 6 Loss of Coolant Accident and valve CS-125A fails closed

1. On Lead Examiner's cue, initiate Event Trigger 6

2. If directed to override CS-125A, wait 3 minutes and initiate Event Trigger 16 and report that the keyswitch for CS-125A is in override.

3. If directed to place the keyswitch for CS-125A in NORMAL, wait 1 minute and initiate Event Trigger 26 and report that the keyswitch for CS-125B is in NORMAL

4. If called as NAO to verify proper operation of unloaded Emergency Diesel Generators, then wait 2 minutes and manually initiate Event Trigger 10. Wait an additional minute and manually initiate Event Trigger 11 to acknowledge local EDG panels. Report that both A and B EDGs are running properly unloaded.

2014 NRC Exam Scenario 3 D-1 Rev 3

NRC Scenario 3 Event 7 Containment Spray Pump B Trip

1. After the crew enters OP-902-002 and on Lead Examiner's cue, initiate Event Trigger 7.

2. If directed to override CS-125B, wait 3 minutes (1 minute if already in the area) and initiate Event Trigger 17 and report that the keyswitch for CS-125B is in OVERRIDE

3. If directed to place the keyswitch for CS-125B in NORMAL, wait 1 minute and initiate Event Trigger 27 and report that the keyswitch for CS-125B is in NORMAL

4. At the end of the scenario, before resetting, end data collection and save the file as 2012 Scenario 1-(start-end time).tid. Export to .csv file. Save the file into the folder for the appropriate crew 2014 NRC Exam Scenario 3 D-1 Rev 3

NRC Scenario 3 SCENARIO TIMELINE DELAY RAMP EVENT KEY DESCRIPTION TRIGGER FINAL HH:MM:SS HH:MM:SS EVENT DESCRIPTION 5 RP02A RPS CH A AUTO TRIP FAILURE N/A 00:00:00 00:00:00 ACTIVE NO AUTOMATIC REACTOR TRIP 5 RP02B RPS CH B AUTO TRIP FAILURE N/A 00:00:00 00:00:00 ACTIVE NO AUTOMATIC REACTOR TRIP 5 RP02C RPS CH C AUTO TRIP FAILURE N/A 00:00:00 00:00:00 ACTIVE NO AUTOMATIC REACTOR TRIP 5 RP02D RPS CH D AUTO TRIP FAILURE N/A 00:00:00 00:00:00 ACTIVE NO AUTOMATIC REACTOR TRIP 6 CS04A CS TRAIN A CS-125A FAILS TO AUTO OPEN N/A 00:00:00 00:00:00 ACTIVE CS-125A FAILS SHUT 2 SG04G MS LINE IPT-1013C FAIL (0-100%) 2 00:00:00 00:00:00 0 SG-IPT-1013C FAILS LOW 3 RC08A RCP 1A LOWER SEAL FAILURE (0-100%) 3 00:00:00 00:00:00 100 PRESSURIZER CODE SAFETY, RC-317A, FAILS OPEN, 4 RC15A2 PZR CONTROL LT 0110X FAILS LO 4 00:00:00 00:00:00 ACTIVE PZR LT 0110X FAILS LOW 5 RC03A RCP RC-MPMP-0001A SHAFT SEIZURE 5 00:00:00 00:00:00 ACTIVE RCP 1A SHAFT SEIZURE 2014 NRC Exam Scenario 3 D-1 Rev 3

NRC Scenario 3 DELAY RAMP EVENT KEY DESCRIPTION TRIGGER FINAL HH:MM:SS HH:MM:SS EVENT DESCRIPTION 6 RC23A RCS COLD LEG 1A RUPTURE 6 00:00:00 00:00:00 3 COLD LEG LOSS OF COOLANT ACCIDENT 7 CS01B LOSS OF CNTMT SPRAY PUMP B 7 00:00:00 00:00:00 ACTIVE CS PUMP B TRIP 6 CSR13A CS-125A REMOTE KEY SW TO CLOSE VALVE 16 00:00:00 00:00:00 OVERRIDE CS-125A KEY SWITCH 6 CSR13A CS-125A REMOTE KEY SW TO CLOSE VALVE 26 00:00:00 00:00:00 NORMAL CS-125A KEY SWITCH 7 CSR13B CS-125B REMOTE KEY SW TO CLOSE VALVE 17 00:00:00 00:00:00 OVERRIDE CS-125B KEY SWITCH 7 CSR13B CS-125B REMOTE KEY SW TO CLOSE VALVE 27 00:00:00 00:00:00 NORMAL CS-125B KEY SWITCH 6 EGR26 EDG A LOCAL ANNUN ACK 10 00:00:00 00:00:00 ACKN LOCAL EDG ANNUNCIATOR ACKNOWLEDGE 6 EGR27 EDG B LOCAL ANNUN ACK 11 00:00:00 00:00:00 ACKN LOCAL EDG ANNUNCIATOR ACKNOWLEDGE 7 DI-08A04S22-1 CC-125A CNTMT SPRAY HEADER A ISOL N/A 00:00:00 00:00:00 CLOSE OVERRIDE CS-125A CONTROL SWITCH TO CLOSE 2014 NRC Exam Scenario 3 D-1 Rev 3

NRC Scenario 3 REFERENCES Event Procedures 1 OP-003-035, Auxiliary Feedwater, Rev. 304 OP-010-003, Plant Startup, Rev. 332 (Copy marked up through Step 9.4.54) 2 OP-009-007, Plant Protection System, Rev. 15 OP-903-013, Monthly Channel Checks, Rev. 16 Technical Specification 3.3.1 Technical Specification 3.3.2 3 OP-901-130, Reactor Coolant Pump Malfunction, Rev. 7 4 OP-901-110, Pressurizer Level Control Malfunction, Rev. 5 Technical Specification 3.3.3.5 OP-903-013, Monthly Channel Checks, Rev. 16 5 OP-902-000, Standard Post Trip Actions, Rev. 13 6 OP-902-009, Standard Appendices, Rev. 309, Appendix 1, Diagnostic Flow Chart OP-902-002, Loss of Coolant Accident Recovery Procedure, Rev. 18 OP-902-009, Standard Appendices, Rev. 309, Appendix 2, Figures 7 OP-902-008, Functional Recovery, Rev. 22 OP-902-009, Standard Appendices, Rev. 309, Appendix 28, Aligning LPSI Pump to Replace CS Pump 2014 NRC Exam Scenario 3 D-1 Rev 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 1 Page 1 of 32 Event

Description:

Secure AFW Pump and raise power to 5%

Time Position Applicants Actions or Behavior OP-010-003, Plant Startup, Section, 9.4 SRO 9.4.52.2 Secure Auxiliary Feedwater Pump in accordance with OP-003-035, Auxiliary Feedwater.

OP-003-035, Auxiliary Feedwater, Section, 7.0 N/A 7.1 Shutdown of Auxiliary Feedwater Pump 7.1.1 If recircing Steam Generators, then re-align Blowdown to Condenser A by performing the following:

BOP 7.1.2 Secure Auxiliary Feedwater Pump.

BOP 7.1.3 Place Auxiliary Feedwater Controller, FW-IFIC-8202, in Manual (MAN).

BOP 7.1.3.1 Set controller to minimum setting.

BOP 7.1.4 Open Auxiliary Feedwater Pump Discharge Pressure Cntrl, AFW-125.

OP-010-003, Plant Startup, Section, 9.4 BOP 9.4.52.3 Maintain Steam Generator levels 50 to 70% NR.

N/A 9.4.53 Adjust Steam Generator Blowdown flow as recommended by Chemistry Department.

ATC 9.4.54 Begin raising Reactor power by CEA withdrawal or boron dilution to 5% full power.

ATC Commences raising power per Reactivity plan and per OP-004-004, Control Element Drive OP-004-004, Control Element drive, Section, 6.7 N/A 6.7 Operation of CEAs in Manual Group (MG) Mode (c) 2014 NRC Exam Scenario 3 D-2 Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 1 Page 2 of 32 Event

Description:

Secure AFW Pump and raise power to 5%

Time Position Applicants Actions or Behavior CAUTION (1) CRITICALITY SHALL BE ANTICIPATED ANY TIME CEAS ARE WITHDRAWN AND THE REACTOR IS NOT CRITICAL.

(2) OBSERVE APPLICABLE GROUP INSERTION LIMITS IN ACCORDANCE WITH TECHNICAL SPECIFICATION 3.1.3.6 (REG GROUP), AND TECHNICAL SPECIFICATION 3.1.3.5 (SHUTDOWN BANKS).

(3) IMPROPER OPERATION OF CEAS IN MANUAL GROUP MODE MAY CAUSE A REACTOR TRIP BASED ON AN OUT-OF-SEQUENCE CONDITION.

(4) CEA INITIALIZATION PROGRAM MUST BE RUNNING IN THE PLANT MONITORING COMPUTER TO HAVE GROUP STOPS AND SEQUENTIAL PERMISSIVES AVAILABLE.

ATC 6.7.1 Verify Plant Monitoring Computer operable in accordance with OP-004-012, Plant Monitoring Computer.

ATC 6.7.2 Position Group Select switch to desired group (Group P).

ATC 6.7.3 Place Mode Select switch to MG and verify the following:

White lights Illuminated on Group Selection Matrix for selected group MG light Illuminates ATC 6.7.4 Operate CEA Manual Shim switch to WITHDRAW or INSERT group to desired height while monitoring the following:

CEA Position Indicator selected CEA group is moving in desired direction If Reactor is critical, then monitor the following:

o Reactor Power o Reactor Coolant System (RCS) temperature o Axial Shape Index (ASI)

NOTE The Operator should remain in the area in front of the CEA Drive Mechanism Control Panel when the Mode Select switch is not in OFF.

ATC 6.7.5 When desired set of moves have been completed, then place Mode Select switch to OFF.

OP-010-003, Plant Startup, Section, 9.4 ATC 9.4.55 Prior to exceeding 5% power, verify Linear Power Channels are on scale.

2014 NRC Exam Scenario 3 D-2 Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 1 Page 3 of 32 Event

Description:

Secure AFW Pump and raise power to 5%

Time Position Applicants Actions or Behavior Examiner Note This event is complete when the AFW pump is stopped and Reactivity Manipulation is satisfied or At Lead Examiners Discretion Examiner Note Cue the Simulator Operator when ready for Event 2 2014 NRC Exam Scenario 3 D-2 Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 2 Page 4 of 32 Event

Description:

Steam Generator Pressure Instrument SG-IPT-1013C fails low Time Position Applicants Actions or Behavior ATC Recognize and report indications of failed channel.

Alarms:

RPS CHANNEL TRIP SG 1 PRESSURE LO (Cabinet K, E-15)

SG 1 PRESSURE LO PRETRIP A/C (Cabinet K, F-15)

ESFAS CHANNEL TRIP PRESS SG 2 > SG 1 (Cabinet K, L-16)

PRESS SG 2 > SG 1 ESFAS PRETRIP A/C (Cabinet K, M-16)

RPS CHANNEL C TROUBLE (Cabinet K, G-18)

Indications:

SG #1 Pressure Safety Channel C fails downscale Pre-Trip indication Channel C LO SG-1 PRESS bistable Trip indication Channel C LO SG-1 PRESS bistable Pre-Trip indication Channel C HI SG-2 P bistable Trip indication Channel C HI SG-2 P bistable Examiner Note All BOP manipulations for OP-009-007 are located at CP-10 except as noted.

OP-009-007, Plant Protection System ,Section 6.2, Trip Channel Bypass Operation SRO 6.2.1 Refer to Attachment 11.11, PPS Bistable Bypass Chart to assist in determination of Trip Channels requiring placement in bypass.

Determines the following bistables are affected and need to be bypassed:

11 - LO SG-1 PRESS 19 - HI SG-1 P 20 - HI SG-2 P SRO Directs BOP to bypass the LO SG-1 PRESS, HI SG-1 P, and HI SG-2 P bistables in PPS Channel C within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in accordance with OP-009-007, Plant Protection System.

BOP 6.2.2 To place a bistable in or remove a bistable from bypass, go to Attachment 11.10, Trip Channel Bypass Operation.

2014 NRC Exam Scenario 3 D-2 Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 2 Page 5 of 32 Event

Description:

Steam Generator Pressure Instrument SG-IPT-1013C fails low Time Position Applicants Actions or Behavior OP-009-007, Plant Protection System ,Attachment 11.10, Trip Channel Bypass Operation CAUTION (1) ATTEMPTING TO PLACE MORE THAN ONE TRIP CHANNEL IN BYPASS REMOVES BOTH TRIP CHANNELS FROM BYPASS.

(2) PRIOR TO PLACING ANY TRIP CHANNEL IN BYPASS, VERIFY BYPASS PUSH BUTTONS ON DE-ENERGIZED PPS BAY NOT DEPRESSED.

BOP 11.10.1 To Bypass a Trip Channel, perform the following:

11.10.1.1 Circle the bistable numbers selected for bypass under Step 11.10.1.4. (Circles bistable numbers 11, 19 and 20 in Step 11.10.1.4 table)

BOP 11.10.1.2 Check desired Trip Channel is not Bypassed on another PPS Channel.

BOP 11.10.1.3 Open key-locked portion of BCP in desired PPS Channel.

BOP 11.10.1.4 Depress Bypass push buttons for the desired Trip Channels BOP 11.10.1.5 Check all selected bistable Bypass push buttons remain in a Depressed state.

BOP 11.10.1.6 Check all selected bistable Bypass lights Illuminate on BCP for the desired Trip Channels.

CREW 11.10.1.7 Check all selected bistable Bypass lights Illuminate on ROM for the desired Trip Channels. (Verifies correct bistables lit on CP-7 PPS Channel C Remote Operator Module)

SRO Reviews the following Technical Specifications and determines applicable actions:

3.3.1 - action 2 3.3.2 - action 13 3.3.3.5 - no actions required 3.3.3.6 - no actions required 2014 NRC Exam Scenario 3 D-2 Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 2 Page 6 of 32 Event

Description:

Steam Generator Pressure Instrument SG-IPT-1013C fails low Time Position Applicants Actions or Behavior Examiner Note This event is complete when bistables are bypassed and Technical Specifications have been addressed or At Lead Examiners Discretion Examiner Note Cue the Simulator Operator when ready for Event 3 2014 NRC Exam Scenario 3 D-2 Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 3 Page 7 of 32 Event

Description:

RCP 1A Lower Seal Failure Time Position Applicants Actions or Behavior ATC / BOP Recognize and report indications of RCP Seal failure Alarms PMC alarm RCP 1A CONTROLLED BLEEDOFF TEMP HI (Cabinet H, F-4) (if unattended)

Indications Middle Seal Pressure rising Upper Seal Pressure rising CCW return temperature rising Controlled bleed off temperature rising CRS Enter and direct the implementation of OP-901-130, Reactor Coolant Pump Malfunction.

OP-901-130, Reactor Coolant Pump Malfunction, E0, Subsequent Operator Actions N/A 1. IF Reactor Coolant Pump trips, THEN verify Reactor tripped AND GO TO OP-902-000, STANDARD POST TRIP ACTIONS.

N/A 2. IF loss of Component Cooling Water to Reactor Coolant Pumps occurs, THEN GO TO OP-901-510, COMPONENT COOLING WATER SYSTEM MALFUNCTION.

SRO 3. IF Reactor Coolant Pump Seal has failed, THEN GO TO section E1, Seal Failure.

OP-901-130, Reactor Coolant Pump Malfunction, E1, Seal Failure NOTE

1. RCP Seal pressure and Control Bleedoff temperature and flow are normally as follows (assuming normal operating RCS temperature and pressure):

Vapor Seal pressure: 25 to 45 PSIG Upper Seal pressure: 585 to 915 PSIG Middle Seal pressure: 1237 to 1815 PSIG CBO temperature: 135° to 190°F CBO flow: 1.2 to 1.8 GPM

2. (If only one Reactor Coolant Pump Seal has failed on a Reactor Coolant Pump, THEN pump operation may continue provided the seal package is monitored for further degradation.

BOP 1. Inform System Engineer of Reactor Coolant Pump Seal failure.

2014 NRC Exam Scenario 3 D-2 Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 3 Page 8 of 32 Event

Description:

RCP 1A Lower Seal Failure Time Position Applicants Actions or Behavior CAUTION (1) CCW TEMPERATURES OF <75 F COULD LEAD TO ESSENTIAL CHILLER TRIPS ON EVAPORATOR LOW REFRIGERANT PRESSURE.

(2) CCW TEMPERATURE SHOULD BE CHANGED AT A RATE OF <10 F IN ONE HOUR TO PREVENT DEGREDATION OF THE REACTOR COOLANT PUMP SEALS.

BOP 2. IF Controlled Bleedoff temperature is rising, THEN lower Component Cooling Water temperature by ANY of the following:

Start Dry Cooling Tower Fans Start Auxiliary Component Cooling Water Pump(s) AND associated Wet Cooling Tower Fans.

Start Auxiliary Component Cooling Water Pump(s) AND lower ACC-126A(B) setpoint.

N/A 3. IF TWO OR MORE seals fail in rapid succession, (within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />) THEN perform the following:

3.1 Trip the Reactor.

3.2 Secure affected Reactor Coolant Pump.

3.3 GO TO OP-902-000, STANDARD POST TRIP ACTIONS N/A 4. IF a SECOND seal fails on the same Reactor Coolant Pump, THEN commence a controlled Plant shutdown in accordance with OP-010-005, Plant Shutdown.

N/A 5. IF during the controlled Plant shutdown Controlled Bleedoff temperature rises to >200 F OR a THIRD seal fails on the same Reactor Coolant Pump, THEN perform the following:

NOTE Manual override is accomplished by positioning control switch to CLOSED, then to OPEN. If after 100 seconds CCW Return temperature is <145 F, Seal Cooler isolations will remain open.

CAUTION IF COMPONENT COOLING WATER IS LOST TO REACTOR COOLANT PUMP SEALS FOR >10 MINUTES, THEN RESTORING COMPONENT COOLING WATER TO REACTOR COOLANT PUMPS MAY RESULT IN SEAL FAILURE.

N/A 6. IF Component Cooling Water to a Reactor Coolant Pump Seal Cooler isolates, THEN within 3 minutes restore Component Cooling Water flow to Reactor Coolant Pump by opening the Reactor Coolant Pump Component Cooling Water Isolation valves:

N/A 7. IF CCW flow is lost to a Reactor Coolant Pump Seal AND can NOT be restored within 3 minutes, THEN perform the following:

2014 NRC Exam Scenario 3 D-2 Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 3 Page 9 of 32 Event

Description:

RCP 1A Lower Seal Failure Time Position Applicants Actions or Behavior Examiner Note This event is complete after the BOP takes action to lower Component Cooling Water temperature OR At Lead Examiners Discretion Examiner Note Cue the Simulator Operator when ready for Event 4 2014 NRC Exam Scenario 3 D-2 Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 4 Page 10 of 32 Event

Description:

Pressurizer Level Channel RC-ILI-0110X fails low Time Position Applicants Actions or Behavior ATC / BOP Recognize and report indications of Channel Failure Alarms PRESSURIZER LEVEL HI/LO (Cabinet H, B-1)

PRESSURIZER LEVEL LO-LO (Cabinet H, C-1)

LETDOWN FLOW HI/LO (Cabinet G, C-1)

Indications PZR Level RC-ILR-0110-X fails downscale All PZR Heaters OFF All Charging Pumps running Letdown Flow lowering Actual PZR level RC-ILR-0110-X rising SRO Enter and direct the implementation of OP-901-110, Pressurizer Level Control Malfunction.

OP-901-110, Pressurizer Level Control Malfunction, E0, General Actions N/A 1. Stop Turbine load changes.

N/A 2. IF malfunction is due to failure of Letdown Flow Control valve, THEN GO TO OP-901-112, CHARGING/LETDOWN MALFUNCTION.

SRO 3. IF malfunction is due to failure of Pressurizer Level Control Channel (incorrect readings on EITHER RC-ILI-0110X OR RC-ILI-0110Y), THEN GO TO Subsection E1, Pressurizer Level Control Channel Malfunction.

OP-901-110, Pressurizer Level Control Malfunction, E1, Pressurizer Level Control Channel Malfunction NOTE Selecting the non-faulted channel may cause automatic actions to occur if actual level is not at program level.

ATC 1. Place Pressurizer Level Controller (RC-ILIC-0110) in MAN AND adjust OUTPUT to slowly adjust letdown flow to restore Pressurizer level.

2014 NRC Exam Scenario 3 D-2 Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 4 Page 11 of 32 Event

Description:

Pressurizer Level Channel RC-ILI-0110X fails low Time Position Applicants Actions or Behavior ATC 2. Transfer Pressurizer Level Control CHANNEL SELECT switch to non-faulted channel.

ATC 3. Transfer Pressurizer CHANNEL SELECT LO LEVEL HEATER CUTOFF switch to non-faulted channel.

ATC 4. Verify desired backup Charging pumps in AUTO.

ATC 5. Verify ALL PROPORTIONAL AND BACKUP HEATER BANKS reset.

ATC 6. Place Pressurizer Level Controller (RC-ILIC-0110) in AUTO and verify Pressurizer Level is being restored to setpoint.

ATC 7. Verify Pressurizer level controlling at program setpoint in accordance with Attachment 1, Pressurizer Level Versus Tave Curve.

SRO 8. Refer to Technical Specifications 3.3.3.5 and 3.3.3.6 for Remote Shutdown and Accident Monitoring operability determination.

2014 NRC Exam Scenario 3 D-2 Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 4 Page 12 of 32 Event

Description:

Pressurizer Level Channel RC-ILI-0110X fails low Time Position Applicants Actions or Behavior SRO Reviews the following Technical Specifications and determines applicable actions:

3.3.3.5 - action a 3.3.3.6 - no actions required Examiner Note This event is complete after the ATC has restored automatic Pressurizer level control and the SRO has evaluated Technical Specifications OR At Lead Examiners Discretion Examiner Note Cue the Simulator Operator when ready for Event 5 2014 NRC Exam Scenario 3 D-2 Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 5 Page 13 of 32 Event

Description:

RCP 1A locked rotor no automatic Reactor Trip Time Position Applicants Actions or Behavior ATC Recognize and report indications of RCP trip and no Auto RX Trip Alarms:

RCP 1A TRIP/TROUBLE (Cabinet H, A-3)

RCP 1A LUBE OIL PRESSURE LO (Cabinet H, E-3)

RCP 1A VIBRATION HI (Cabinet H, A-4)

RCP 1A CCW FLOW LO (Cabinet H, H-4)

RCP 1A CCW FLOW LOST (Cabinet SA(SB), A-1(6)

RPS CHANNEL TRIP LOCAL PWR DENSITY HI (Cabinet K, A-11)

RPS CHANNEL TRIP DNBR LO (Cabinet K, A-12)

LOCAL PWR DENSITY HI PRETRIP A/C(B/D) (Cabinet K, B(C)-11)

DNBR LO PRETRIP A/C(B/D) (Cabinet K, B(C)-12)

RPS CHANNEL TRIP COOLANT FLOW LOST (Cabinet K, D-12)

RPS CHANNEL A(B)(C)(D) TROUBLE (Cabinet K, E(F)(G)(H)-18)

Indications:

RCP 1A STOP light and Yellow Trip indication RCP 1A Oil Lift Pumps running RCP 1A motor amps indicate 0amps RCP 1A P lowering Trip indication Channel A(B)(C)(D) HI LOCAL POWER bistables Pre-Trip indication Channel A(B)(C)(D) HI LOCAL POWER bistables Trip indication Channel A(B)(C)(D) LOW DNBR bistables Pre-Trip indication Channel A(B)(C)(D) LOW DNBR bistables Trip indication Channel A(B)(C)(D) SG LO FLOW bistables All Reactor Trip breakers remain closed All CEAs remain withdrawn CRITICAL TASK ESTABLISH REACTIVITY CONTROL This task is satisfied by shutting down the Reactor by either depressing both Reactor Trip pushbuttons on CP-2 or CP-8, initiating Diverse Reactor Trip Pushbuttons on CP-2, or by deenergizing the rod drive MG sets by deenergizing busses 32A and 32B on CP-1 prior to exiting the step to Verify Reactivity Control (Step 1) of OP-902-000, Standard Post Trip Actions.

This task becomes applicable following the tripped Reactor Coolant Pump 1A.

ATC Manually Trip the Reactor (CRITICAL) 2014 NRC Exam Scenario 3 D-2 Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 5 Page 14 of 32 Event

Description:

RCP 1A locked rotor no automatic Reactor Trip Time Position Applicants Actions or Behavior OP-902-000, Standard Post Trip Actions ATC 1. Determine Reactivity Control acceptance criteria are met:

a. Check reactor power is dropping.

b. Check startup rate is negative.

c. Check less than TWO CEAs are NOT fully inserted.

BOP 2. Determine Maintenance of Vital Auxiliaries acceptance criteria are met:

a. Check the Main Turbine is tripped:

Governor valves closed Throttle valves closed BOP b. Check the Main Generator is tripped:

GENERATOR BREAKER A tripped GENERATOR BREAKER B tripped EXCITER FIELD BREAKER tripped BOP c. Check station loads are energized from offsite electrical power as follows:

Train A A1, 6.9 KV non safety bus A2, 4.16 KV non safety bus A3, 4.16 KV safety bus A-DC electrical bus A or C vital AC Instrument Channel Train B B1, 6.9 KV non safety bus B2, 4.16 KV non safety bus B3, 4.16 KV safety bus B-DC electrical bus B or D vital AC Instrument Channel ATC 3. Determine RCS Inventory Control acceptance criteria are met:

a. Check that BOTH the following conditions exist:

Pressurizer level is 7% to 60%

Pressurizer level is trending to 33% to 60%

b. Check RCS subcooling is greater than or equal to 28ºF.

2014 NRC Exam Scenario 3 D-2 Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 5 Page 15 of 32 Event

Description:

RCP 1A locked rotor no automatic Reactor Trip Time Position Applicants Actions or Behavior ATC 4. Determine RCS Pressure Control acceptance criteria are met by checking that BOTH of the following conditions exist:

Pressurizer pressure is 1750 psia to 2300 psia Pressurizer pressure is trending to 2125 psia to 2275 psia ATC 5. Determine Core Heat Removal acceptance criteria are met:

a. Check at least one RCP is operating.

b. Check operating loop T is less than 13ºF.

c. Check RCS subcooling is greater than or equal to 28ºF.

Examiner Note This event is complete after the crew has reached the steps to check RCS Heat Removal OR At Lead Examiners Discretion Examiner Note Cue the Simulator Operator when ready for Event 6 2014 NRC Exam Scenario 3 D-2 Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 6 Page 16 of 32 Event

Description:

Loss of Coolant Accident CS-125A fails closed Time Position Applicants Actions or Behavior BOP Recognize and report indications of LOCA Alarms CONTAINMENT WATER LEAKAGE HI (Cabinet N, L-20)

CONTAINMENT WATER LEAKAGE HI-HI (Cabinet N, K-20)

PRESSURIZER PRESSURE HI/LO (Cabinet H, E-1)

CIAS TRAIN A(B) LOGIC INITIATED [Cabinet K, E-19(20)]

SIAS TRAIN A(B) LOGIC INITIATED [Cabinet K, G-19(20)]

CSAS TRAIN A(B) LOGIC INITIATED [Cabinet K, H-19(20)]

MSIS TRAIN A(B) LOGIC INITIATED [Cabinet K, L-19(20)]

SUBCOOLED MARGIN LO [Cabinet M(N), M-7(17)]

RPS CHANNEL TRIP PZR PRESSURE LO (Cabinet K, A-16)

RPS CHANNEL TRIP CNTMT PRESSURE HI (Cabinet K, A-17)

ESFAS CHANNEL TRIP CNTMT PRESSURE HI (Cabinet K, L-17)

CNTMT SPRAY HDR A FLOW LO (Cabinet M, F-4)

Indications RCS pressure lowering PZR level lowering Containment Pressure and Temperature rising Containment activity rising Containment water level rising ESFAS components actuating CS-125A indicates in CLOSE (GREEN)

Examiner Note The crew should report indications but continue in OP-902-000, Standard Post Trip Actions BOP 6. Determine RCS Heat Removal acceptance criteria are met:

a. Check that at least one steam generator has BOTH of the following:

Steam generator level is 10% to 76% NR Main Feedwater is available to restore level within 55%-70%

NR.[60% to 80% NR]

ATC b. Check RCS TC is 530 ºF to 550 ºF BOP c. Check steam generator pressure is 885 psia to 1040 psia.

2014 NRC Exam Scenario 3 D-2 Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 6 Page 17 of 32 Event

Description:

Loss of Coolant Accident CS-125A fails closed Time Position Applicants Actions or Behavior BOP d. Check Feedwater Control in Reactor Trip Override:

MAIN FW REG valves are closed STARTUP FW REG valves are 13% to 21% open Operating main Feedwater pumps are 3800 rpm to 4000 rpm BOP e. Reset moisture separator reheaters, and check the temperature control valves closed.

ATC 7. Determine Containment Isolation acceptance criteria are met:

a. Check containment pressure is less than 16.4 psia.

b. Check NO containment area radiation monitor alarms OR unexplained rise in activity.

c. Check NO steam plant activity monitor alarms OR unexplained rise in activity.

ATC a.1 IF containment pressure is greater than or equal to 17.1 psia, THEN verify the following:

• CIAS is initiated

• SIAS is initiated

• MSIS is initiated BOP 8 Determine Containment Temperature and Pressure Control acceptance criteria are met:

a. Check containment temperature is less than or equal to 120ºF.

b. Check containment pressure is less than 16.4 psia.

BOP 8.1 Verify at least three containment fan coolers are operating.

BOP 8.2 IF containment pressure is greater than or equal to 17.1 psia, THEN verify ALL available containment fan coolers are operating in emergency mode.

BOP/ATC 8.3 IF containment pressure is greater than or equal to 17.7 psia, THEN verify ALL of the following:

• CSAS is initiated

• ALL available containment spray pumps are delivering flow greater than 1750 gpm

• ALL RCPs are secured N/A 9. IF ALL safety function acceptance criteria are met, AND NO contingency actions were performed, THEN GO TO OP-902-001, "Reactor Trip Recovery" procedure.

2014 NRC Exam Scenario 3 D-2 Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 6 Page 18 of 32 Event

Description:

Loss of Coolant Accident CS-125A fails closed Time Position Applicants Actions or Behavior SRO 10. IF ANY safety function acceptance criteria are NOT met, OR ANY contingency action was taken, THEN GO TO Appendix 1, "Diagnostic Flowchart.

OP-902-009, Standard Appendices, Appendix 1 Examiner Note Appendix 1 is a flow chart used to diagnose to the correct recovery procedure for the event in progress. The steps below will be followed by a YES or NO to indicate proper flow path.

ATC Rx Pwr dropping, SUR negative, and < two CEAs NOT fully inserted or Emergency Boration in progress (YES)

ATC Pressurizer pressure dropping rapidly and Pressurizer level changing (YES)

BOP Steam Generator Pressure Abnormally Low (NO)

N/A Primary Break ATC Containment Pressure and Temperature Abnormally High (YES)

ATC Activity in the Steam Plant (NO)

N/A LOCA inside Containment BOP At least one 4KV safety bus energized (YES)

SRO Go To OP-902-002, LOCA Recovery OP-902-002, Loss of Coolant accident Recovery NOTE The Shift Chemist should be notified if a SIAS or CIAS has occurred. The secondary sampling containment isolation valves should not be opened following an SIAS or CIAS until directed by the Shift Chemist.

BOP Confirm Diagnosis

• 1. Confirm diagnosis of a LOCA:

a. Check Safety Function Status Check acceptance criteria are satisfied.

b. IF steam generator sample path is available, THEN direct Chemistry to sample BOTH steam generators for activity.

2014 NRC Exam Scenario 3 D-2 Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 6 Page 19 of 32 Event

Description:

Loss of Coolant Accident CS-125A fails closed Time Position Applicants Actions or Behavior BOP Announce the Event

2. Announce a Loss of Coolant Accident is in progress using the plant page.

N/A Classify the Event

• 3. Advise the Shift Manager to REFER TO EP-001-001, "Recognition &

Classification of Emergency Condition", and implement the Emergency Plan.

N/A Restore Operation of DCT Sump Pumps

• 4. IF power has been interrupted to either 3A or 3B safety buses, THEN perform Appendix 20, "Operation of DCT Sump Pumps".

SRO Implement Placekeeping

5. REFER TO Section 6.0, "Placekeeper", and record the time of the reactor trip.

ATC Verify SIAS Initiation

6. IF pressurizer pressure is less than 1684 psia, THEN check SIAS has initiated.

BOP *7. IF SIAS has initiated, THEN:

a. Verify safety injection pumps have started.

b. Check safety injection flow is within the following:

Appendix 2-E, "HPSI Flow Curve" Appendix 2-F, "LPSI Flow Curve" Examiner Note Appendix 2-E and Appendix 2-F are contained on the next 2 pages. Injection flow will be meeting all requirements.

2014 NRC Exam Scenario 3 D-2 Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 6 Page 20 of 32 Event

Description:

Loss of Coolant Accident CS-125A fails closed Time Position Applicants Actions or Behavior 2014 NRC Exam Scenario 3 D-2 Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 6 Page 21 of 32 Event

Description:

Loss of Coolant Accident CS-125A fails closed Time Position Applicants Actions or Behavior ATC c. Verify ALL available charging pumps are operating.

N/A RCP Trip Strategy

8. IF pressurizer pressure is less than 1621 psia, AND SIAS is actuated, THEN:

a. Verify no more than two RCPs are operating.

b. IF pressurizer pressure is less than the minimum RCP NPSH of Appendix 2A-D, "RCS Pressure and Temperature Limits", THEN stop ALL RCPs.

2014 NRC Exam Scenario 3 D-2 Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 6 Page 22 of 32 Event

Description:

Loss of Coolant Accident CS-125A fails closed Time Position Applicants Actions or Behavior ATC Verify RCP Operating Limits

• 9. IF RCPs are operating, THEN:

a. Verify CCW available to RCPs.

b. IF a CSAS is initiated, THEN stop ALL RCPs.

c. IF RCS TC is less than 382ºF, THEN verify no more than two RCPs are operating.

BOP Verify Proper CCW Operation

• 10. Check a CCW pump is operating for each energized 4.16 KV safety bus.

ATC Isolate the LOCA

11. Isolate the LOCA:

a. Verify the following letdown containment isolation valves are closed:

CVC 101, LETDOWN STOP VALVE CVC 103, LETDOWN ISOL VALVE CVC 109, LETDOWN ISOL VALVE BOP b. Verify the following RCS sampling containment isolation valves are closed:

Train A PSL 107, HOT LEG PSL 204, PZR SURGE PSL 304, PZR STEAM Train B PSL 105, HOT LEG PSL 203, PZR SURGE PSL 303, PZR ISOL VLV Examiner Note This event is complete after the crew has entered OP-902-002, Loss of Coolant Accident Recovery and all RCPs are secured OR At Lead Examiners Discretion Examiner Note Cue the Simulator Operator when ready for Event 7 Examiner Note It takes approximately 30 seconds from when the simulator operator initiates Event 7 for the first indication to be present to the applicants.

2014 NRC Exam Scenario 3 D-2 Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 7 Page 23 of 32 Event

Description:

Containment Spray Pump B Trips Time Position Applicants Actions or Behavior BOP Recognize and report indications of B Spray Pump Trip Alarms CNTMT SPRAY PUMP B TRIP/TROUBLE (Cabinet N, B-14)

Indications Containment Spray Pump B indicates OFF (GREEN)

Containment Spray flow indicates 0 GPM OP-902-002, Loss of Coolant accident Recovery SRO Confirm Diagnosis

• 1. Confirm diagnosis of a LOCA:

a. Check Safety Function Status Check acceptance criteria are satisfied.

b. IF steam generator sample path is available, THEN direct Chemistry to sample BOTH steam generators for activity.

SRO Evaluate Safety Functions SRO Determine Containment Pressure and Temperature Control Safety Function not met SRO SAFETY FUNCTION:

8. Containment Temperature and Pressure Control Condition 2

a. Containment fan coolers > 1 operating in emerg mode

b. Containment spray pump > 1 operating with flow > 1750 gpm

c. Containment pressure < 50 psia SRO a.1 GO TO ONE of the following:

Appendix 1, "Diagnostic Flowchart" OP-902-008, "Functional Recovery Procedure" Examiner Note Appendix 1 Diagnostic Flowchart will not provide guidance for this event. OP-902-008, Functional Recovery Procedure is the procedure to recover from this event.

OP-902-008, Functional Recovery BOP Announce the Event

1. Announce that the Functional Recovery Procedure is in progress using the plant page.

2014 NRC Exam Scenario 3 D-2 Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 7 Page 24 of 32 Event

Description:

Containment Spray Pump B Trips Time Position Applicants Actions or Behavior N/A Classify the Event

• 2. Advise the Shift Manager to REFER TO EP-001-001, "Recognition &

Classification of Emergency Condition", and implement the Emergency Plan.

SRO Implement Placekeeping

3. REFER TO the "Placekeeper", and record the time of the reactor trip.

ATC RCP Trip Strategy

4. IF pressurizer pressure is less than 1621 psia, AND SIAS is actuated, THEN:

a. Verify no more than two RCPs are operating.

b. IF pressurizer pressure is less than the minimum RCP NPSH of Appendix 2A-D, "RCS Pressure and Temperature Limits", THEN stop ALL RCPs.

N/A Verify RCP Operating Limits

• 5. IF RCPs are operating, THEN:

a. IF a CSAS is initiated, THEN stop ALL RCPs.

b. Verify CCW available to RCPs. c. IF RCS TC is less than 382ºF

[384ºF], THEN verify no more than two RCPs are operating.

NOTE The Shift Chemist should be notified if a SIAS or CIAS has occurred. The secondary sampling containment isolation valves should not be opened following an SIAS or CIAS until directed by the Shift Chemist.

BOP Sample BOTH Steam Generators

6. Direct Chemistry to sample BOTH steam generators for activity and boron.

N/A Verify Equipment Ventilation

• 7. IF power is lost to both 3A and 3B safety buses and NOT expected to be restored within 30 minutes, THEN perform the following:

NOTE Portable emergency lighting is available in Appendix R lockers located at LCP-43, Remote Shutdown Panel and +35 RAB Relay Room.

N/A Reduce Battery Loads

• 8. IF power is lost to both 3A and 3B safety buses and NOT expected to be restored within 30 minutes, THEN perform the following to reduce unnecessary station loads:

2014 NRC Exam Scenario 3 D-2 Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 7 Page 25 of 32 Event

Description:

Containment Spray Pump B Trips Time Position Applicants Actions or Behavior N/A Restore Operation of DCT Sump Pumps

• 9. IF power has been interrupted to either 3A or 3B safety buses, THEN perform Appendix 20, "Operation of DCT Sump Pumps".

BOP Place Hydrogen Analyzers in Service

10. Place Hydrogen Analyzers in service:

Train A

a. Place Train A H2 ANALYZER CNTMT ISOL VALVE keyswitch to "OPEN". (Key 216)

b. Place H2 ANALYZER A POWER to "ON".

c. Check H2 ANALYZER A PUMP indicates ON.

BOP Train B

a. Place Train B H2 ANALYZER CNTMT ISOL VALVE keyswitch to "OPEN". (Key 217)

b. Place H2 ANALYZER B POWER to "ON".

c. Check H2 ANALYZER B PUMP indicates ON.

SRO Identify Success Paths

• 11. Identify success paths to be used to satisfy each safety function using BOTH of the following:

Resource Assessment Trees Safety Function Tracking Sheet Note 1. Reactivity Control RC-1

2. Maintenance of Vital Auxiliaries (DC) MVA-DC-1

3. Maintenance of Vital Auxiliaries (AC) MVA-AC-1

4. RCS Inventory Control IC-2

5. RCS Pressure Control PC-2

6. RCS and Core Heat Removal HR-2

7. Containment Isolation CI-1

8. Containment Pressure and Temperature Control CPTC-2 SRO Perform Safety Function Status Checks

• 12. REFER TO Section 6.0, "Safety Function Status Check", and perform Safety Function Status Checks.

2014 NRC Exam Scenario 3 D-2 Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 7 Page 26 of 32 Event

Description:

Containment Spray Pump B Trips Time Position Applicants Actions or Behavior Examiner Note

1. RC Met

2. MVA-DC Met

3. MVA-AC Met

4. IC Met

5. PC Met

6. HR Met

7. CI NOT Met

8. CPTC NOT Met SRO Maintain Success Paths

• 13. Prioritize Safety Functions based on ALL of the following:

a. Instructions for those Safety Functions which do NOT meet any success path.

b. Instructions for those Safety Functions for which success path one criteria is NOT met.

c. Instructions for Safety Functions for which success path one criteria is met.

Note 1. Containment Isolation CI-1

2. Containment Pressure and Temperature Control CPTC-2

3. RCS Inventory Control IC-2

4. RCS Pressure Control PC-2

5. RCS and Core Heat Removal HR-2

6. Reactivity Control RC-1

7. Maintenance of Vital Auxiliaries (DC) MVA-DC-1

8. Maintenance of Vital Auxiliaries (AC) MVA-AC-1 SRO Implement Success Paths

14. Implement success paths based on prioritization from previous step.

Examiner Note The SRO may decide that by establishing Containment Pressure and Temperature control, that this would satisfy the Containment Isolation Safety Function. The SRO may decide to implement CPTC-2 to restore both Safety Functions. However this guide contains CI-1 steps below followed by CPTC-2 steps.

2014 NRC Exam Scenario 3 D-2 Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 7 Page 27 of 32 Event

Description:

Containment Spray Pump B Trips Time Position Applicants Actions or Behavior OP-902-008, Functional Recovery, CI-1 SRO Verify Containment Isolation

• 1. IF ANY of the following conditions exist:

Containment pressure is greater than 17.1 psia Pressurizer pressure is less than 1684 psia Containment area radiation monitors greater than the Hi Alarm setpoint THEN perform BOTH of the following:

a. Verify CIAS is initiated.

b. Verify that an isolation valve is closed for each containment penetration required to be closed.

N/A b.1 IF ANY containment fan cooler is NOT operating AND containment pressure is greater than 17.1 psia, THEN REFER TO Appendix 21-B, "CFC CCW Override" and close the associated Containment Fan Cooler CCW Isolation Valves.

SRO b.2 IF ANY CS-125, Containment Spray Header Isolation is open AND the associated CS pump is NOT operating, THEN REFER TO Appendix 21-A, "CS-125 Override," and close the valve.

OP-902-009, Standard Appendices, Appendix 21-A BOP 1.1 Place Containment Spray Pump A(B) control switch to "OFF."

Examiner Note CNTMT SPRAY HDR B ISOL VLV IN CSAS OVERRIDE (Cabinet N, K-14) will actuate upon the execution of the step below to place the keyswitch for CS-125B in OVERRIDE.

BOOTH 1.2 Place the keyswitch for CS-125A(B), Containment Spray Isolation Valve, located on the side of Auxiliary Panel 1(Auxiliary Panel 2) to "OVERRIDE."

BOP 1.3 Place CS-125A(B), CNTMT SPRAY HEADER A(B) ISOL valve to "OPEN" and then to "CLOSE."

SRO Transition back to OP-902-008, Functional Recovery 2014 NRC Exam Scenario 3 D-2 Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 7 Page 28 of 32 Event

Description:

Containment Spray Pump B Trips Time Position Applicants Actions or Behavior OP-902-008, Functional Recovery, CI-1 (cont)

N/A Determine Most Affected SG

2. IF a SGTR is indicated by ANY of the following:

Steam generator activities Main Steam Line radiation levels Steam generator Blowdown radiation monitor readings Steam generator level rise when NOT feeding One steam generator level rising faster that the other with feed and steaming rates being essentially the same for both Feed flow mismatch between steam generators Steam flow verses feed flow mismatch in a steam generator prior to the reactor trip THEN REFER TO Heat Removal success path in use and isolate the most affected steam generator:

BOP CCW Leakage

3. Check the following CCW Radiation Monitor AB indications:

Hi Alarm clear No abnormal rise in radiation monitor reading SRO Check CI-1 Acceptance Criteria

• 4. Check CI-1, Automatic/Manual Isolation is satisfied by ANY of the following:

Condition 1

a. IF steam plant activity monitor alarms or an unexplained rise in steam plant activity exists, THEN ALL release paths from the most affected steam generator to the environment are closed.

b. Containment pressure is less than 17.1 psia.

c. NO Containment area radiation monitor alarms or unexplained rise.

d. IF SGTR diagnosed, THEN steam generator pressure less than 1000 psia

[960 psia] AND NOT steaming via ADV.

SRO Condition 2

a. IF steam plant activity monitor alarms or an unexplained rise in steam plant activity exists, THEN ALL release paths from the most affected steam generator to the environment are closed.

b. Each containment penetration required to be closed for current plant conditions has an isolation valve closed.

c. IF SGTR diagnosed, THEN steam generator pressure less than 1000 psia

[960 psia] AND NOT steaming via ADV.

2014 NRC Exam Scenario 3 D-2 Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 7 Page 29 of 32 Event

Description:

Containment Spray Pump B Trips Time Position Applicants Actions or Behavior N/A Continue Efforts to Restore Safety Functions

1. IF the Containment Isolation safety function is still NOT met, THEN continue efforts to establish Containment Isolation by evaluating further actions based on ALL of the following:

The urgency of other safety functions that are NOT met The risk to plant personnel and the public of leaving certain containment penetrations unisolated The feasibility of isolating containment penetration(s) by alternate means OP-902-008, Functional Recovery, CPTC-2 BOP Verify Containment Spray Actuation

• 1. IF containment pressure is greater than 17.7 psia, THEN perform BOTH of the following:

a. Verify CSAS is initiated.

b. Verify ALL containment spray pumps are delivering flow greater than 1750 gpm.

BOP a.1 IF ANY CS-125, Containment Spray Header Isolation is open AND the associated CS pump is NOT operating, THEN REFER TO Appendix 21-A, "CS-125 Override," and close the valve.

OP-902-009, Standard Appendices, Appendix 21-A BOP 1.1 Place Containment Spray Pump A(B) control switch to "OFF."

Examiner Note CNTMT SPRAY HDR B ISOL VLV IN CSAS OVERRIDE (Cabinet N, K-14) will actuate upon the execution of the step below to place the keyswitch for CS-125B in OVERRIDE.

BOOTH 1.2 Place the keyswitch for CS-125A(B), Containment Spray Isolation Valve, located on the side of Auxiliary Panel 1(Auxiliary Panel 2) to "OVERRIDE."

BOP 1.3 Place CS-125A(B), CNTMT SPRAY HEADER A(B) ISOL valve to "OPEN" and then to "CLOSE."

SRO Transition back to OP-902-008, Functional Recovery 2014 NRC Exam Scenario 3 D-2 Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 7 Page 30 of 32 Event

Description:

Containment Spray Pump B Trips Time Position Applicants Actions or Behavior OP-902-008, Functional Recovery, CPTC-2 (cont)

N/A Containment Spray Termination

• 2. IF CS pumps are operating AND ALL of the following conditions are satisfied:

Containment pressure is less than 16.4 psia and stable or lowering RAS has NOT actuated Containment Spray is NOT required for iodine removal THEN REFER TO Appendix 5-E, "CSAS Reset Procedure",and reset CSAS actuation.

SRO Check CTPC-2 Acceptance Criteria

3. Check CTPC-2, Containment Spray is satisfied by ANY of the following:

Condition 1

a. BOTH containment spray pumps delivering flow greater than or equal to 1750 gpm.

b. Containment pressure is less than 50 psia.

Condition 2

a. At least one containment fan cooler is operating in the emergency mode.

b. At least one containment spray pump delivering flow greater than or equal to 1750 gpm.

c. Containment pressure is less than 50 psia.

SRO 3.1 IF the Containment Temperature and Pressure safety function is still NOT met, THEN REFER TO CTPC Continuing Actions.

OP-902-008, Functional Recovery, CPTC Continuing Actions SRO Continue Efforts to Restore Safety Functions

1. IF the Containment Temperature and Pressure safety function is still NOT met, THEN continue efforts to establish Containment Temperature and Pressure by evaluating further actions using ALL of the following:

a. The urgency of other safety functions that are NOT met.

b. The rate of change of containment temperature and pressure and potential for damage to the containment.

c. The feasibility of restoring a success path by restoring ALL of the following:

Vital auxiliaries necessary to operate systems or components in the success path Manual operation of valves

d. Use of alternate components to implement a success path:

Appendix 28, "Aligning LPSI to Replace CS" 2014 NRC Exam Scenario 3 D-2 Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 7 Page 31 of 32 Event

Description:

Containment Spray Pump B Trips Time Position Applicants Actions or Behavior OP-902-009, Standard Appendices, Appendix 28-B CRITICAL TASK ESTABLISH CONTAINMENT TEMPERATURE AND PRESSURE CONTROL This task is satisfied by aligning Low Pressure Safety Injection Pump B to replace Containment Spray Pump B in accordance with Appendix 28 of OP-902-009, Standard Appendices prior to exiting OP-902-009 Appendix 28.

This task becomes applicable following the initiation of a Containment Spray Actuation Signal.

SRO 1.1 Obtain TSC concurrence prior to performing this evolution.

BOP 1.2 Verify LPSI Pump B control switch in "OFF." (CRITICAL)

BOP 1.3 Verify Containment Spray Pump B control switch in "OFF."

BOP 1.4 Place SI-129B, LPSI FLOW CONTROL VALVE to "AUTO." (Key 146)

(CRITICAL)

BOP 1.5 Place SI-IFIC-0306, LPSI FLOW CONTROLLERS HEADER 1A/1B in "MAN." (CRITICAL)

BOP 1.6 Adjust SI-IFIC-0306, LPSI FLOW CONTROLLERS HEADER 1A/1B to 0% output. (CRITICAL)

BOP 1.7 Verify the following valves Closed: (CRITICAL)

SI-415B, LPSI SHUTDOWN TEMP CONTROL valve (Key 147)

SI-138B, LPSI FLOW CONTROL COLD LEG 1B SI-139B, LPSI FLOWCONTROL COLD LEG 1A BOP 1.8 Open SI-125B/SI-412B, SHDN HX B ISOL valves. (Key 145)

(CRITICAL)

BOP 1.9 Verify CS-125B, CNTMT SPRAY HEADER B ISOL valve open.

(CRITICAL)

BOP 1.10 Start LPSI Pump B. (CRITICAL)

BOP 1.11 Verify Containment Spray Header B flow.

2014 NRC Exam Scenario 3 D-2 Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 7 Page 32 of 32 Event

Description:

Containment Spray Pump B Trips Time Position Applicants Actions or Behavior Examiner Note This event is complete after the BOP has established Containment Spray flow in accordance with OP-902-009, Standard appendices, Appendix 28 OR At Lead Examiners Discretion 2014 NRC Exam Scenario 3 D-2 Rev 1