ML15007A176

2014-12 Final Outlines

ML15007A176
Person / Time
Site:	River Bend
Issue date:	12/12/2014
From:	Vincent Gaddy Operations Branch IV
To:
laura hurley
References
Download: ML15007A176 (32)

Text

ES-401 BWR Examination Outline Form ES-401-1 Facility: River Bend Station Date of Exam: December 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 4 4 3 3 3 3 20 3 4 7 Emergency & N/A N/A Abnormal Plant 2 1 2 1 1 1 1 7 2 1 3 Evolutions Tier Totals 5 6 4 4 4 4 27 5 5 10 1 2 3 2 2 3 2 2 2 3 3 2 26 3 2 5 2.

Plant 2 1 1 1 2 1 1 1 1 1 1 1 12 0 2 1 3 Systems Tier Totals 3 4 3 4 4 3 3 3 4 4 3 38 5 3 8

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

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.

ES-401 2 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 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 AK2.02 Knowledge of the interrelations between 295001 Partial or Complete Loss of Forced R PARTIAL OR COMPLETE LOSS OF FORCED 3.2 1 Core Flow Circulation / 1 & 4 CORE FLOW CIRCULATION and the following:

Nuclear boiler instrumentation 295003 Partial or Complete Loss of AC / 6 AK1.04 Knowledge of the operational implications 295004 Partial or Total Loss of DC Pwr / 6 R of the following concepts as they apply to PARTIAL 2.8 2 OR COMPLETE LOSS OF DC POWER : Effect of battery discharge rate on capacity G2.2.44 Ability to interpret control room indications S to verify the status and operation of a system, and 4.4 76 understand how operator actions and directives affect plant and system conditions.

AK2.07 Knowledge of the interrelations between 295005 Main Turbine Generator Trip / 3 R MAIN TURBINE GENERATOR TRIP and the 3.6 3 following: Reactor pressure control AK3.03 Knowledge of the operational implications 295006 SCRAM / 1 R of the following concepts as they apply to SCRAM: 3.8 4 Reactor pressure response S G2.4.34 Knowledge of RO tasks performed outside 4.2 77 the main control room during an emergency and the resultant operational effects. (SCRAM)

A1.06 Knowledge of the operational implications of 295016 Control Room Abandonment / 7 R the following concepts as they apply to CONTROL 4.0 5 ROOM ABANDONMENT : Reactor water level A2.04 Ability to determine and/or interpret the 295018 Partial or Total Loss of CCW / 8 R following as they apply to PARTIAL OR 2.9 6 COMPLETE LOSS OF COMPONENT COOLING WATER : System flow G2.4.47 Ability to diagnose and recognize trends in 295019 Partial or Total Loss of Inst. Air / 8 R an accurate and timely manner utilizing the 4.2 7 appropriate control room reference material.

AK1.03 Knowledge of the operational implications 295021 Loss of Shutdown Cooling / 4 R of the following concepts as they apply to LOSS 3.9 8 OF SHUTDOWN COOLING : Adequate core cooling S 3.6 78 AA2.04 Ability to determine and/or interpret the following as they apply to LOSS OF SHUTDOWN COOLING : Reactor water temperature AK2.05 Knowledge of the interrelations between 295023 Refueling Acc / 8 R REFUELING ACCIDENTS and the following: 3.5 9 Secondary containment ventilation EK3.06 Knowledge of the reasons for the following 295024 High Drywell Pressure / 5 R responses as they apply to HIGH DRYWELL 4.0 10 PRESSURE: Reactor Scram EA1.05 Ability to operate and/or monitor the 295025 High Reactor Pressure / 3 R following as they apply to HIGH REACTOR 3.7 11 PRESSURE: RCIC: Plant-Specific EA2.03 Ability to determine and/or interpret the 295026 Suppression Pool High Water R following as they apply to SUPPRESSION POOL 3.9 12 Temp. / 5 HIGH WATER TEMPERATURE: Reactor pressure G2.4.20 Knowledge of the operational implications 295027 High Containment Temperature / 5 R of EOP warnings, cautions, and notes. 3.8 13

EK1.01 Knowledge of the operational implications 295028 High Drywell Temperature / 5 R of the following concepts as they apply to HIGH 3.5 14 DRYWELL TEMPERATURE: Reactor water level measurement G2.4.21 Knowledge of the parameters and logic S used to assess the status of safety functions, such 4.6 79 as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

EK2.08 Knowledge of the interrelations between 295030 Low Suppression Pool Wtr Lvl / 5 R LOW SUPPRESSION POOL WATER LEVEL and 3.5 15 the following: SRV discharge submergence EK2.03 Knowledge of the interrelations between 295031 Reactor Low Water Level / 2 R REACTOR LOW WATER LEVEL and the 4.2 16 following: Low pressure core spray S EA2.04 Ability to determine and/or interpret the 4.8 80 following as they apply to REACTOR LOW WATER LEVEL: Adequate core cooling EA1.06 Ability to operate and/or monitor the 295037 SCRAM Condition Present R following as they apply to SCRAM CONDITION 4.1 17 and Reactor Power Above APRM PRESENT AND REACTOR POWER ABOVE Downscale or Unknown / 1 APRM DOWNSCALE OR UNKNOWN: Neutron monitoring system EA2.03 Ability to determine and/or interpret the 295038 High Off-site Release Rate / 9 R following as they apply to HIGH OFF-SITE 3.5 18 RELEASE RATE: Radiation levels AA2.16 Ability to determine and interpret the 600000 Plant Fire On Site / 8 S following as they apply to PLANT FIRE ON SITE: 3.5 81 Vital equipment and control systems to be maintained and operated during a fire R G2.4.11 Knowledge of abnormal condition 4.0 19 procedures AK1.03 Knowledge of the operational implications 700000 Generator Voltage and Electric Grid R of the following concepts as they apply to 3.3 20 Disturbances / 6 GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES: Under-excitation G2.2.37 Ability to determine operability and/or S availability of safety related equipment. 4.6 82 K/A Category Totals (RO): 4 4 3 3 3 3 Group Point Total: 20 K/A Category Totals (SRO): 3 4 Group Point Total: 7

ES-401 3 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 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 295002 Loss of Main Condenser Vac / 3 AK2.06 Knowledge of the interrelations between 295007 High Reactor Pressure / 3 R HIGH REACTOR PRESSURE and the following: 3.7 21 PCIS/NSSSS: Plant Specific AA2.05 Ability to determine and/or interpret the 295008 High Reactor Water Level / 2 R following as they apply to HIGH REACTOR WATER 2.9 22 LEVEL: Swell 295009 Low Reactor Water Level / 2 295010 High Drywell Pressure / 5 G2.4.04 Ability to recognize abnormal indications for 295011 High Containment Temp / 5 R system operating parameters that are entry level 4.5 23 conditions for emergency and abnormal operating procedures: High Containment Temperature.

AK1.02 Knowledge of the operational implications of 295012 High Drywell Temperature / 5 R the following concepts as they apply to HIGH 3.1 24 DRYWELL TEMPERATURE : Reactor power level control AK2.01 Knowledge of the interrelations between 295013 High Suppression Pool Temp. / 5 R HIGH SUPPRESSION POOL TEMPERATURE and 3.6 25 the following: Suppression pool cooling 295014 Inadvertent Reactivity Addition / 1 AA2.02 Ability to determine and/or interpret the 295015 Incomplete SCRAM / 1 S following as they apply to INCOMPLETE SCRAM : 4.2 83 Control rod position 295017 High Off-site Release Rate / 9 295020 Inadvertent Cont. Isolation / 5 & 7 295022 Loss of CRD Pumps / 1 EK3.03 Knowledge of the reasons for the following 295029 High Suppression Pool Wtr Lvl / 5 R responses as they apply to HIGH SUPPRESSION 3.4 26 POOL WATER LEVEL : Reactor SCRAM S G2.2.25 Knowledge of the bases in Technical 4.2 84 Specifications for limiting conditions for operations and safety limits.

EA1.03 Ability to operate and/or monitor the 295032 High Secondary Containment R following as they apply to HIGH SECONDARY 3.7 27 Area Temperature / 5 CONTAINMENT AREA TEMPERATURE :

Secondary Containment Ventilation 295033 High Secondary Containment Area Radiation Levels / 9 295034 Secondary Containment Ventilation High Radiation / 9 295035 Secondary Containment High Differential Pressure / 5 295036 Secondary Containment High Sump/Area Water Level / 5 EA2.03 Ability to determine and / or interpret the 500000 High CTMT Hydrogen Conc. / 5 S following as they apply to HIGH PRIMARY 3.8 85 CONTAINMENT HYDROGEN CONCENTRATIONS: Combustible limits for drywell

K/A Category Point Totals (RO): 1 2 1 1 1 1 Group Point Total: 7 K/A Category Point Totals (SRO): 2 1 3

ES-401 4 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 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 K4.03 Knowledge of RHR/LPCI: INJECTION 203000 RHR/LPCI: Injection R MODE (PLANT SPECIFIC) design feature(s) 3.2 28 Mode and/or interlocks which provide for the following: pump minimum flow protection K5.02 Knowledge of the operational 205000 Shutdown Cooling R implications of the following concepts as 2.8 29 they apply to SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) : Valve operation S A2.03 Ability to (a) predict the impacts of the 3.2 86 following on the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: A.C.

failure 206000 HPCI N/A for RBS 207000 Isolation (Emergency) N/A for RBS Condenser K6.03 Knowledge of the effect that a loss or 209001 LPCS R malfunction of the following will have on the 3.3 30 LOW PRESSURE CORE SPRAY SYSTEM :

Torus/suppression pool water level A3.03 Ability to monitor automatic R operations of the LOW PRESSURE CORE 3.5 31 SPRAY SYSTEM including: System pressure A1.08 Ability to predict and/or monitor 209002 HPCS R changes in parameters associated with 3.1 32 operating the HIGH PRESSURE CORE SPRAY SYSTEM (HPCS) controls including:

System lineup: BWR-5,6 K5.06 Knowledge of the operational 211000 SLC R implications of the following concepts as 3.0 33 they apply to STANDBY LIQUID CONTROL SYSTEM: Tank level measurement A2.03 Ability to (a) predict the impacts of the R following on the STANDBY LIQUID 3.2 34 CONTROL SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

A.C. power failures 212000 RPS R A3.06 Ability to monitor automatic 4.2 35 operations of the REACTOR PROTECTION SYSTEM including: Main turbine trip: Plant-Specific G2.2.12 Knowledge of surveillance S procedures. 4.1 87 215003 IRM R A4.05 Ability to manually operate and/or 3.4 36 monitor in the control room: Trip bypasses

215004 Source Range Monitor R K2.01 Knowledge of electrical power 2.6 37 supplies to the following: SRM channels/detectors G2.4.31 Knowledge of annunciator alarms, indications, or response procedures. SRMs R 4.2 38 215005 APRM / LPRM R K1.09 Knowledge of the physical 3.6 39 connections and/or cause-effect relationships between AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM and the following: Reactor recirculation system:

BWR-5,6 A2.02 Ability to (a) predict the impacts of the S following on the AVERAGE POWER 3.7 88 RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Upscale or downscale trips 217000 RCIC R K2.01 Knowledge of electrical power 2.8 40 supplies to the following: Motor operated valves 218000 ADS R K3.02 Knowledge of the effect that a loss or 4.5 41 malfunction of the AUTOMATIC DEPRESSURIZATION SYSTEM will have on following: Ability to rapidly depressurize the reactor A4.02 Ability to manually operate and/or R monitor in the control room: ADS logic 4.2 42 initiation 223002 PCIS/Nuclear Steam R K4.01 Knowledge of PRIMARY 3.0 43 Supply Shutoff CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF design feature(s) and/or interlocks which provide for the following:

Redundancy 239002 SRVs R K5.04 Knowledge of the operational 3.3 44 implications of the following concepts as they apply to RELIEF/SAFETY VALVES :

Tail pipe temperature monitoring A2.06 Ability to (a) predict the impacts of the S following on the RELIEF/SAFETY VALVES ; 4.3 89 and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Reactor high pressure 259002 Reactor Water Level R K6.03 Knowledge of the effect that a loss or 3.1 45 Control malfunction of the following will have on the REACTOR WATER LEVEL CONTROL SYSTEM : Main steam flow input 261000 SGTS R A1.01 Ability to predict and/or monitor 2.9 46 changes in parameters associated with operating the STANDBY GAS TREATMENT SYSTEM controls including: System flow

262001 AC Electrical R A2.11 Ability to (a) predict the impacts of the 3.2 47 Distribution following on the A.C. ELECTRICAL DISTRIBUTION ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Degraded system voltages G2.2.25 Knowledge of the bases in S Technical Specifications for limiting 4.2 90 conditions for operations and safety limits.

262002 UPS (AC/DC) R A3.01 Ability to monitor automatic 2.8 48 operations of the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) including:

Transfer from preferred to alternate source 263000 DC Electrical R A4.01 Ability to manually operate and/or 3.3 49 Distribution monitor in the control room: Major breakers and control power fuses: Plant- Specific 264000 EDGs R K1.05 Knowledge of the physical 3.2 50 connections and/or cause-effect relationships between EMERGENCY GENERATORS (DIESEL/JET) and the following: Emergency generator fuel oil supply system R G2.1.31 Ability to locate control room 4.6 51 switches, controls, and indications, and to l determine that they correctly reflect the desired plant lineup.

300000 Instrument Air R K2.01 Knowledge of electrical power 2.8 52 supplies to the following: Instrument air compressor 400000 Component Cooling R K3.01 Knowledge of the effect that a loss or 2.9 53 Water malfunction of the CCWS will have on the following: Loads cooled by CCWS K/A Category Point Totals (RO): 2 3 2 2 3 2 2 2 3 3 2 Group Point Total: 26 K/A Category Point Totals 3 2 Group Point Total: 5 (SRO):

ES-401 5 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 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 201001 CRD Hydraulic 201002 RMCS N/A for RBS 201003 Control Rod and Drive R A4.02 Ability to manually operate 3.5 54 Mechanism and/or monitor in the control room:

CRD mechanism position: Plant-Specific 201004 RSCS N/A for RBS 201005 RCIS R A1.01 Ability to predict and/or monitor 3.2 55 changes in parameters associated with operating the ROD CONTROL AND INFORMATION SYSTEM (RCIS) controls including: First stage shell pressure/turbine load: BWR-6 G2.1.23 Ability to perform specific system and integrated plant procedures S 4.4 91 during all modes of plant operation.

201006 RWM 202001 Recirculation R K4.02 Knowledge of RECIRCULATION 3.1 56 System design feature(s) and/or interlocks which provide for the following: Adequate recirculation pump NPSH 202002 Recirculation Flow Control R K1.09 Knowledge of the physical 3.1 57 connections and/or cause-effect relationships between RECIRCULATION FLOW CONTROL SYSTEM and the following: Reactor water level 204000 RWCU S A2.14 Ability to (a) predict the impacts 3.2 92 of the following on the REACTOR WATER CLEANUP SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

System high temperature 214000 RPIS N/A for RBS 215001 Traversing In-core Probe 215002 RBM N/A for RBS 216000 Nuclear Boiler Inst. R K6.01 Knowledge of the effect that a 3.1 58 loss or malfunction of the following will have on the NUCLEAR BOILER INSTRUMENTATION: A.C. electrical distribution 219000 RHR/LPCI: Torus/Pool Cooling R K2.02 Knowledge of electrical power 3.1 59 Mode supplies to the following: Pumps 223001 Primary CTMT and Aux. R G2.1.28 Knowledge of the purpose and 4.1 60 function of major system components and controls.

226001 RHR/LPCI: CTMT Spray Mode N/A for RBS

230000 RHR/LPCI: Torus/Pool Spray N/A for RBS Mode 233000 Fuel Pool Cooling/Cleanup 234000 Fuel Handling Equipment 239001 Main and Reheat Steam R K3.15 Knowledge of the effect that a 3.5 61 loss or malfunction of the MAIN AND REHEAT STEAM SYSTEM will have on following: Reactor water level control 239003 MSIV Leakage Control 241000 Reactor/Turbine Pressure R A2.01 Ability to (a) predict the impacts 3.5 62 Regulator of the following on the REACTOR/TURBINE PRESSURE REGULATING SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Loss of turbine inlet pressure signal 245000 Main Turbine Gen. / Aux.

256000 Reactor Condensate 259001 Reactor Feedwater 268000 Radwaste 271000 Offgas 272000 Radiation Monitoring R K4.02 Knowledge of RADIATION 3.7 63 MONITORING System design feature(s) and/or interlocks which provide for the following: Automatic actions to contain the radioactive release in the event that the predetermined release rates are exceeded 286000 Fire Protection 288000 Plant Ventilation R A3.01 Ability to monitor automatic 3.8 64 operations of the PLANT VENTILATION SYSTEMS including:

Isolation/initiation signals 290001 Secondary CTMT S A2.04 Ability to (a) predict the impacts 3.7 93 of the following on the SECONDARY CONTAINMENT ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: High airborne radiation 290003 Control Room HVAC R K5.01 Knowledge of the operational 3.2 65 implications of the following concepts as they apply to CONTROL ROOM HVAC: Airborne contamination (e.g.,

radiological, toxic gas, smoke) control 290002 Reactor Vessel Internals K/A Category Point Totals (RO): Group Point Total: 12 K/A Category Point Totals (SRO): Group Point Total: 3

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: River Bend Station Date of Exam: December 2014 Category K/A # Topic RO SRO-Only IR # IR #

2.1.7 Ability to evaluate plant performance and make 4.4 66 operational judgments based on operating characteristics,

1. reactor behavior, and instrument interpretation.

Conduct 2.1.31 Ability to locate control room switches, controls, and 4.6 67 of Operations indications, and to determine that they correctly reflect the desired plant lineup.

2.1.40 Knowledge of refueling administrative requirements. 2.8 68 2.1.20 Ability to interpret and execute procedure steps. 4.6 94 2.1.36 Knowledge of procedures and limitations involved in core 4.1 95 alterations.

2.2.6 Knowledge of the process for making changes to 3.0 69 procedures.

2.2.14 Knowledge of the process for controlling equipment 3.9 70 2.

configuration or status.

Equipment Control 2.2.35 Ability to determine Technical Specification Mode of 3.6 71 Operation.

2.2.11 Knowledge of the process for controlling temporary 3.3 96 design changes.

2.2.38 Knowledge of conditions and limitations in the facility 4.5 97 license.

2.3.11 Ability to control radiation releases. 3.8 72 2.3.13 Knowledge of radiological safety procedures pertaining to 3.4 73 licensed operator duties, such as response to radiation 3.

Radiation monitor alarms, containment entry requirements, fuel Control handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

2.3.14 Knowledge of radiation or contamination hazards that 3.8 98 may arise during normal, abnormal, or emergency conditions or activities.

2.4.12 Knowledge of general operating crew responsibilities 4.0 74 during emergency operations.

4.

2.4.26 Knowledge of facility protection requirements, including 3.1 75 Emergency fire brigade and portable fire fighting equipment usage.

Procedures /

Plan 2.4.16 Knowledge of EOP implementation hierarchy and 4.4 99 coordination with other support procedures or guidelines such as, operating procedures, abnormal operating procedures, and severe accident management guidelines.

2.4.44 Knowledge of emergency plan protective action 4.4 100 recommendations.

Subtotal Tier 3 Point Total

ES-401 Record of Rejected K/As Form ES-401-4 Tier / Randomly Reason for Rejection Group Selected K/A Q#10 KA asked for reason to ED on High DW Pressure; 1/1 295024 EK3.04 RBS has no guidance to ED for this condition.

Randomly selected another KA in the EK3 category - EK3.06 Q#21 KA could not be written to non-GFES level.

1/2 295007 AK2.01 Randomly selected another KA in the AK2 category - AK2.06 Q#23 KA asked for ARP actions and setpoints for high cont.

temperature; there is no alarm/annunciator associated with 1/2 295011 G2.4.50 containment high temperature.

Randomly selected G2.4.04 from the remaining G2.4 category Q#27 KA asked for monitor/operate area temperature monitoring 1/2 295032 EA1.01 for secondary containment. We were unable to come up with three credible distractors; Randomly selected EA1.03 Q#28 KA asked for design features/interlocks that provide for Surveillance for all operable components; We could not write 2/1 203000 K4.09 a question to this KA without it being an SRO question Randomly selected K4.03 from the remaining 14 in the K4 category Q#38 KA rejected due to RBS has no AOP/EOP entry for SRMs; 2/1 215004 G2.4.4 Randomly selected G2.4.31 Q#52 KA asked for electrical power supply to the emergency air compressor; at RBS, the emergency air compressor is diesel 2/1 300000 K2.02 operated.

Selected the only other KA in the K2 category - K2.01 Q#77 KA asked for setpointsassociated with EOP entry conditions for a SCRAM. We could not write an SRO level 1/1 295006 G2.4.2 question to this KA.

Randomly selected G2.4.34 from the remaining G2.4 category and wrote a new question.

Q#92 KA asked about a system RBS does not have (Low Pressure 2/2 204000 A2.02 RWCU)

Randomly selected another KA in the A2 category - A2.14

NRC ES-301 Administrative Topics Outline Form ES-301-1 Facility: RIVER BEND STATION Date of Examination: 12/8/2014 Examination Level: RO SRO Operating Test Number: __________

Administrative Topic Type Describe activity to be performed (see Note) Code*

(A1) Determine the amount of Decay Heat in the Core Conduct of Operations R, D per OSP-0041 KA 2.1.20 (4.6) Ability to interpret and execute procedure steps (A2) Determine corrected Fuel Zone Level Indication Conduct of Operations R, D KA 2.1.23 (4.3) Ability to perform specific system and integrated plant procedures during all modes of plant operation.

(A3) Use plant drawings to determine effect(s) of fuse R, N removal Equipment Control KA 2.2.15 (3.9) Ability to determine the expected plant configuration using design and configuration control documentation, such as drawings, line-ups, tag-outs, etc.

Radiation Control (A4) Determine Containment Water Level During Containment Flooding Emergency Procedures/Plan R, M KA 2.4.21 (4.0) Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

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

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

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

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

NRC RO ADMIN JPMS A1 - This task will have the applicant calculate core decay heat using the charts from Attachments 1 and 6 of OSP-0041.

A2 - This task will have the applicant plot RPV level using a set of correction curves and then determine is adequate core cooling is assured.

A3 - The applicant will use plant drawings to determine the effects of removing a fuse to support a maintenance activity. Three effects are asked for: (1) the fail position of an exhaust filter damper; (2) any other components that are affected and the associated effect; and (3) identification of any alarms or indications that would be seen in the Control Room.

A4 - This task will have the applicant use given data and a procedure to determine the water level in the containment. This data will be further used to determine correlated water level in the RPV. The task is being performed because of a plant emergency that required entry into the Severe Accident Procedures (SAPs).

JPM A4 was re-selected due to validation comments about the original JPM not being an RO task in our plant.

NRC ES-301 Administrative Topics Outline Form ES-301-1 Facility: RIVER BEND STATION Date of Examination: 12/8/2014 Examination Level: RO SRO Operating Test Number: __________

Administrative Topic Type Describe activity to be performed (see Note) Code*

(A5) Determine stay time for a hot environment Conduct of Operations R, M KA 2.1.26 (3.4) Knowledge of industrial safety procedures (such as rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen, and hydrogen).

R, M (A6) Determine Actions for Reactor Water Chemistry Conduct of Operations KA 2.1.34 (3.5) Knowledge of primary and secondary plant chemistry limits.

R, N (A7) Determine Secondary Containment Operability Equipment Control KA 2.2.12 (4.1) Knowledge of surveillance procedures.

(A8) Determine required actions upon Radioactive Radiation Control R, D Effluent Monitor Failure KA 2.3.11 (4.3) Ability to control radiation releases.

(A9) Determine PAR Emergency Procedures/Plan R, M KA 2.4.44 (4.4) Knowledge of emergency plan protective action recommendations.

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

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

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

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

NRC SRO ADMIN JPMs A5 - This task will have the applicant determine the type of work demand and the maximum stay time for an evolution involving two nuclear equipment operators.

Use of EN-IS-108, Working in Hot Environments will be required A6 - This task will require an SRO to review data and AOP-0058 to determine actions after a report from the Chemistry Department. In part, the SRO is determining if the Mode Switch can be taken to RUN.

A7 - This task will require the SRO to use Daily Logs and given data to determine corrected annulus pressure and secondary containment operability.

A8 - The applicant reviews given plant and environmental conditions, as well as the TRM to determine potential actions for allowing a radiological effluent discharge.

A9 - The applicant reviews given plant and environmental conditions to determine a Protective Action Recommendation. This is a TIME CRITICAL JPM.

NRC ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: RIVER BEND STATION Date of Examination: 12/8/2014 Exam Level: RO SRO-I SRO-U Operating Test No.: ______________

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

a. (S1) Pull Rods to Achieve Criticality - w/ trip of CRD pumps S,A,E,L,N 1

b. (S2) Shutdown RCIC using SOP- w/ Drain Trap Level High S,A,N 2

c. (S3) Perform Main Turbine BPV Cycle Test (STP-509-0101) S,D 3

d. (S4) Reduce SDC to only RHR-A - w/ overload pre-trip alarm S,A,L,N 4

e. (S5) Stop Cont High Volume Purge - w/Decay Heat Fan trip S,A,D,E,EN 9

f. (S6) Place Suppression Pool Cooling System in Service (SOP-140) S,D 5

g. (C1) Shed DC Loads for Station Blackout (AOP-50, Attach 3) C,D,E,L 6

h. (C2) Bypass Control Rod in RACS (STP-500-0705) C,D 7 In-Plant Systems (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

i. (P1) Vent Scram Air Header (Encl 11) R,D,E,EN 1

j. (P2) Supply Fire Protection Water for CB via SSW (SOP-37, sect 5.3) E,N 8

k. (P3) Perform UO Actions of AOP-31 to Operate the Div 1 EDG A,E,L,N 6

@ 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 9/8/46 (E)mergency or abnormal in-plant 1/1/16 (EN)gineered safety feature - / - / 1 (control room system) 2 (L)ow-Power / Shutdown 1/1/14 (N)ew or (M)odified from bank including 1(A) 2/2/15 (P)revious 2 exams 3 / 3 / 2 (randomly selected) 0 (R)CA 1/1/11 (S)imulator Revision 3

NRC S1 - The applicant is directed to withdraw control rods to achieve criticality.

Before criticality is reached however, at the evaluators discretion, a CRD pump trip occurs. As an alternate path, the standby CRD pump is started but also trips. The operator will be required to place the mode switch in SHUTDOWN after receiving an accumulator fault associated with a withdrawn control rod and RPV pressure less than 600 psig.

S2 - This is a replacement JPM for Starting the Condenser Air Removal Pumps.

A post-maintenance run of RCIC is ongoing. Direction will be given to secure RCIC using the SOP. An alarm alerts the applicant of an abnormal condition in the RCIC exhaust drain system. Because of plant conditions, and direction in the Alarm Response Procedure, the applicant will restart the RCIC Gland Seal Compressor.

S3 - This is a replacement JPM for Performing a Pressure Set Adjustment.

There were too few steps in the previous JPM.

The plant is operating at power; this JPM is a surveillance to verify the full stroke operation of the Main Turbine Bypass Valves.

S4 - This is a replacement JPM for placing the Turbine on the turning gear.

The reactor is shutdown with two loops of shutdown cooling in operation.

This task will have the applicant secure the B loop of SDC. As an alternate path, a low voltage condition will cause elevated motor amps on the A RHR Pump; the applicant will have to use the ARP to reduce load on the pump by reducing flow.

S5 - This JPM was adjusted to allow the evolution at higher power so there would be less low power JPMs. (Containment purge instead of drywell purge.)

The plant had experienced high containment radiation levels and high volume purge, using Standby Gas Treatment, was placed in service and rad levels reduced. This task will have the applicant secure the containment purge using SOP-0059. As an alternate Path, a failure of the decay heat removal fan occurs, and the applicant will be required to re-start the standby gas treatment system with a suction from outside air using SOP-0043.

S6 - This JPM was adjusted to allow the evolution at higher power so there would be less low power JPMs.

The plant is performing a startup per GOP-0001. This task has the applicant start the Suppression Pool Cooling and Cleanup System using SOP-0140, Suppression Pool Cleanup and Alternate Decay Heat Removal section 5.2.

NRC C1 - This task will remove DC loads from the station batteries in order to reduce heat load in the control room. Without operator action, during a station blackout, Control Room temperatures can reach greater than 120F in less than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

C2 - This task will have the applicant bypass a control rod to allow movement of the rod without position indication.

P1 - This task will have the applicant simulate venting the scram air header during an ATWS.

P2 - This task will have the applicant align the Standby Service Water System to the Fire Protection Water System Hose Racks. This would be performed provide water for firefighting via manual hose streams in the Control Building.

P3 - This task will have the applicant simulate performing a post-maintenance startup of the RPS Motor Generator Set B using SOP-0079, Reactor Protection System. An Alternate Path is taken when the MG fails to self-excite and the applicant must use guidance in the procedure to reset the over excitation trip to allow the motor generator to achieve proper voltage.

NRC ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: RIVER BEND STATION Date of Examination: 12/8/2014 Exam Level: RO SRO-I SRO-U Operating Test No.: ______________

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

a. (S1)

b. (S2) Shutdown RCIC using SOP- w/ Drain Trap Level High S,A,N 2

c. (S3) Perform Main Turbine BPV Cycle Test (STP-509-0101) S,D 3

d. (S4) Reduce SDC to only RHR-A - w/ overload pre-trip alarm S,A,L,N 4

e. (S5) Stop Cont High Volume Purge - w/Decay Heat Fan trip S,A,D,E,EN 9

f. (S6) Place Suppression Pool Cooling System in Service (SOP-140) S,D 5

g. (C1) Shed DC Loads for Station Blackout (AOP-50, Attach 3) C,D,E,L 6

h. (C2) Bypass Control Rod in RACS (STP-500-0705) C,D 7 In-Plant Systems (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

i. (P1) Vent Scram Air Header (Encl 11) R,D,E,EN 1

j. (P2) Supply Fire Protection Water for CB via SSW (SOP-37, sect 5.3) E,N 8

k. (P3) Perform UO Actions of AOP-31 to Operate the Div 1 EDG A,E,L,N 6

@ 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 4 (C)ontrol room (D)irect from bank 9/8/46 (E)mergency or abnormal in-plant 1/1/15 (EN)gineered safety feature - / - / 1 (control room system) 2 (L)ow-Power / Shutdown 1/1/13 (N)ew or (M)odified from bank including 1(A) 2/2/14 (P)revious 2 exams 3 / 3 / 2 (randomly selected) 0 (R)CA 1/1/11 (S)imulator Revision 2

NRC S2 - This is a replacement JPM for Starting the Condenser Air Removal Pumps.

A post-maintenance run of RCIC is ongoing. Direction will be given to secure RCIC using the SOP. An alarm alerts the applicant of an abnormal condition in the RCIC exhaust drain system. Because of plant conditions, and direction in the Alarm Response Procedure, the applicant will restart the RCIC Gland Seal Compressor.

S3 - This is a replacement JPM for Performing a Pressure Set Adjustment.

There were too few steps in the previous JPM.

The plant is operating at power; this JPM is a surveillance to verify the full stroke operation of the Main Turbine Bypass Valves.

S4 - This is a replacement JPM for placing the Turbine on the turning gear.

The reactor is shutdown with two loops of shutdown cooling in operation.

This task will have the applicant secure the B loop of SDC. As an alternate path, a low voltage condition will cause elevated motor amps on the A RHR Pump; the applicant will have to use the ARP to reduce load on the pump by reducing flow.

S5 - This JPM was adjusted to allow the evolution at higher power so there would be less low power JPMs. (Containment purge instead of drywell purge.)

The plant had experienced high containment radiation levels and high volume purge, using Standby Gas Treatment, was placed in service and rad levels reduced. This task will have the applicant secure the containment purge using SOP-0059. As an alternate Path, a failure of the decay heat removal fan occurs, and the applicant will be required to re-start the standby gas treatment system with a suction from outside air using SOP-0043.

S6 - This JPM was adjusted to allow the evolution at higher power so there would be less low power JPMs.

The plant is performing a startup per GOP-0001. This task has the applicant start the Suppression Pool Cooling and Cleanup System using SOP-0140, Suppression Pool Cleanup and Alternate Decay Heat Removal section 5.2.

NRC C1 - This task will remove DC loads from the station batteries in order to reduce heat load in the control room. Without operator action, during a station blackout, Control Room temperatures can reach greater than 120F in less than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

C2 - This task will have the applicant bypass a control rod to allow movement of the rod without position indication.

P1 - This task will have the applicant simulate venting the scram air header during an ATWS.

P2 - This task will have the applicant align the Standby Service Water System to the Fire Protection Water System Hose Racks. This would be performed provide water for firefighting via manual hose streams in the Control Building.

P3 - This task will have the applicant simulate performing a post-maintenance startup of the RPS Motor Generator Set B using SOP-0079, Reactor Protection System. An Alternate Path is taken when the MG fails to self-excite and the applicant must use guidance in the procedure to reset the over excitation trip to allow the motor generator to achieve proper voltage.

NRC ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: RIVER BEND STATION Date of Examination: 12/8/2014 Exam Level: RO SRO-I SRO-U Operating Test No.: ______________

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

a. (S1)

b. (S2) Shutdown RCIC using SOP- w/ Drain Trap Level High S,A,N 2

c. (S3)

d. (S4)

e. (S5) Stop Cont High Volume Purge - w/Decay Heat Fan trip S,A,D,E,EN 9

f. (S6)

g. (C1)

h. (C2) Bypass Control Rod in RACS (STP-500-0705) C,D 7 In-Plant Systems (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

i. (P1) Vent Scram Air Header (Encl 11) R,D,E,EN 1

j. (P2)

k. (P3) Perform UO Actions of AOP-31 to Operate the Div 1 EDG A,E,L,N 6

@ 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 9/8/43 (E)mergency or abnormal in-plant 1/1/13 (EN)gineered safety feature - / - / 1 (control room system) 2 (L)ow-Power / Shutdown 1/1/11 (N)ew or (M)odified from bank including 1(A) 2/2/12 (P)revious 2 exams 3 / 3 / 2 (randomly selected) 0 (R)CA 1/1/11 (S)imulator Revision 3

December 2014 NRC Exam Appendix D Scenario Outline Form ES-D-1 Facility: River Bend Station Scenario No.: NRC-2 IC No.: 252 Examiners: ________________ Operators: _____________________________

Initial Conditions: 100% reactor power.

Turnover Shift priorities: 1) STP-309-0203 in progress; ready to unload and shutdown the HPCS DG per step 7.3 Event Malf. Event Type* Event No. No. Description 1 NA N (SRO, BOP) Unload the HPCS DG per STP-309-0203 Steps 7.3.1 through 7.3.4 2 ED004R TS (SRO) E22-ACB03 trips after DG output breaker opens (Tech Spec) zdi5(497)!=0 3 T CCS001B C (SRO, BOP) Component Cooling Water pump CCS-P1B trips and pump CCS-P1C fails to CCS003C auto start (AOP-0012) 4 T CNM004B C (SRO,ATC) Condensate pump CNM-P1B trips (RX) (AOP-0006) 5 T N/A TS(SRO) Control Building Fire Door (CB098-10) Failure to close and latch (Tech Spec) 6 T ED010 M (ALL) Loss RSS#1 result in loss of normal feedwater Reactor Scram 7 RCIC002 C (SRO, BOP) Reduce RPV injection Sources E51MOVF013P

• RCIC fails to auto start and must be manually started LPCS002

• RCIC Injection Valve limited to approximately 100-200 gpm flow RHR001A

• Loss of Low Pressure ECCS automatic injection ED004Q 8 MSS001 M (ALL) Steam Leak in Drywell Must ED on lowering water level prior to -186 RPV level Must restore RPV level with low pressure ECCS to > -186

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Total Malfunctions (5-8) (7) E22-ACB03, CCS-P1B, CNM-P1B,CB098-10, Loss RSS#1, DW Leak, Reduce RPV injection sources Malfunctions after EOP entry (1-2) (2) Reduce RPV injection sources, DW Leak Abnormal Events (2-4) (2), AOP-0012, AOP-0006 Major Transients (1-2) (2) Reactor scram due to loss of normal feedwater, DW Leak

December 2014 NRC Exam EOPs entered (1-2) (2) EOP-0001, EOP-0002.

EOP contingencies (0-2) (1) Alternate level control Critical Tasks (2-3) (2) ED prior to RPV level going below -186; Restore RPV water level to -186 within 30 minutes after RPV pressure reaching 200 psig General Scenario Outline Event 0 - The team assumes the shift with reactor power at 100% and the division 3 diesel generator month operability surveillance test in progress.

Event 1 - The team unloads and opens the output breaker for the division 3 diesel generator (Normal evolution)

Event 2 - When the output breaker is opened the supply breaker (E22-ACB03)for the division 3 480 volt switchgear trips on overcurrent resulting in Tech Spec 3.8.9 entry and HPCS becoming unavailable Event 3 - Next one of the operating turbine plant component cooling water pumps (CCS-P1B) trips with the standby pump (CCS-P1C) failing to auto start, the operators will manually start the standby pump to restore normal cooling water flow.

Event 4 - Following the temporary reduction in cooling water flow one the three operating condensate pumps (CNM-P1B) trips resulting in AOP-0006 entry and required power reduction to 90%

Event 5 - A report from a roving security officer results in TRM 3.7.9.6 entry for a failed fire barrier (door CB098-10)

Event 6 - A partial loss of offsite power (RSS#1) occurring resulting in a loss of the remaining condensate pumps and a loss of normal feedwater injection requiring the plant to be shutdown (Scrammed)

Event 7 - The only high pressure injection system available is RCIC and it fails to auto start. When manually started the RCIC injection valve fails to fully open result in reduced injection (around 100-200gpm).

At -143 inches water level in the RPV the Division 2 480 VAC bus loses power (EJS*SWG2B trips open) preventing injection for division 2 low pressure ECCS systems. The RHR A injection valve (E12-MOVF042A) also loses power preventing injection. The LPCS injection valve (E21-MOVF005) fails to automatically open, requiring the BOP operator to open the valve.

Event 8 - Following the scram a steam leak begins in the drywell resulting in loss of coolant accident requiring emergency depressurization due to lowering reactor water level and restoration of RPV water level utilizing low pressure injection systems.

(Critical Task to emergency depressurize prior to RPV water level dropping below -186 inches)

(Critical Task to restore RPV water level above -186 inches within 30 minutes of RPV pressure reaching 200 psig)

December 2014 NRC Exam Appendix D Scenario Outline Form ES-D-1 Facility: River Bend Station Scenario No.: NRC-3 IC No.: 253 Examiners: ________________ Operators: _____________________________

Initial Conditions: 93% reactor power. RCIC tagged out for Line Fill Pump Maintenance.

Turnover Shift priorities: (1) Swap Stator Water Cooling Pumps (for tagout of GMC-P1A);

(2) Raise Reactor Power IAW RMP (using flow)

Event Malf. Event Event No. No. Type* Description 1 NA N (SRO,BOP) Swap Stator Water Cooling Pumps (from A to B) 2 NA R (ATC) Raise Reactor Power with Flow 3 T P863_75A:F_6 I (SRO,BOP) SSW SWGR Room Temp Switch, HVY-ESX25B fails high 4 T B21005 TS (SRO) RPV pressure transmitter B21-PTN078A, fails high (TS) 5 T GMC002B C (SRO,BOP) Trip of Stator Water Cooling Pump B, Standby pump fails to start GMC001A 6 T RPS003A C (SRO,BOP) Loss of RPS-A (AOP-0010) 7 SWPMOV4AP TS (SRO) Failure to isolate (SWP-MOV 5B) (TS) 8 T GMC003A M (ALL) Shear of Stator Water Cooling Pump A Shaft - Requires Scram RPS-001B RPS fails to Automatically and Manually Scram - Rods inserted by ARI RPS001C 9 MGEN003 C (ATC,BOP) Main Generator Reverse Power Relay Failure 10 CNM006 Condensate Filter High D/P - Loss of Feed (recoverable)

C (SRO,BOP)

HPCS002 HPCS injection valve fails to open

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Total Malfunctions (5-8) (8)HVY-ESX25B fails high GMC-B Trip, Press Transmitter, RPS-A, SWPMOV5B failure to isolate, GMC-A shear, Reverse Power Failure, Loss of Feed Malfunctions after EOP entry (1-2) (2) Reverse Power, HPCS inject failure Abnormal Events (2-4) (3), AOP-0010, AOP-0001, AOP-0002 Major Transients (1-2) (1) Reactor scram with Loss of Stator Water Cooling EOPs entered (1-2) (2) EOP-0001 EOP contingencies (0-2) (1) EOP-0001, Alternate Level Control Critical Tasks (2-3) (2) Initiate ARI; RPV Level Restored

December 2014 NRC Exam General Scenario Outline Event 0 - The team assumes the shift with reactor power 93% following a control rod sequence exchange. RCIC is out of service for line fill pump maintenance.

Event 1 - The team alternates the generator stator cooling water pump in preparation for oil change on the A pump.

Event 2 - Following the stator pump swap the team will raise reactor power with recirculation flow as directed by reactor engineering to 97%

Event 3 - HVY-ESX25B temperature switch fails high bringing in switchgear high temperature annunciator and failing to start HVY-FN25A and HVY25B the operator will start the fans per the alarm response procedure Event 4 - Next the A RPS RPV pressure transmitter fails high causing a 1/2 scram condition resulting in Tech Spec 3.3.1.1 entry Event 5 - The B stator cooling water pump trips next with failure of the A standby pump to auto start. The resulting turbine-generator run back is stopped when the operators manually start the A pump.

Event 6 - Following recovery of stator cooling a loss of RPS A occurs requiring the entry into the abnormal operating procedure and recovery of isolated systems Event 7 - Primary containment isolation SWP-MOV5B fails to automatically isolate, requiring entry into T.S. 3.6.1.3 Event 8 - At this time the shaft of the restarted stator cooling water pump A shears resulting in a turbine generator run back. With no stator cooling water available the team is required to shutdown the reactor (Scram). An RPS failure then requires alternate rod insertion to be initiated to insert all control rods.

(Critical Task to imitate ARI prior to RPV level two)

Event 9 - Following reactor shutdown the main generator fails to trip on reverse power. The team will transfer house power prior to opening the main generator output breakers Event 10 - Following reactor shutdown the condensate full flow filters become clogged resulting in a loss of normal feedwater injection. The HPCS injection valve fails to open and the condensate full flow filter bypass valve must be manually opened to recover feed water injection (Critical Task to restore and maintain RPV water level to > -186 inches prior to ED required at -186 inches)

December 2014 NRC Exam Appendix D Scenario Outline Form ES-D-1 Facility: River Bend Station Scenario No.: NRC-4 IC No.: 254 Examiners: ________________ Operators: _____________________________

Initial Conditions: 80% reactor power. Performing Sequence Exchange; RHR-B is in Supp Pool Cooling and Containment Purge is in service for post-maintenance RCIC testing; Turnover Shift priorities: Start RCIC for testing Adjust MVARs Raise Reactor power in accordance with the RMP for sequence exchange; Event Malf. Event Event No. No. Type* Description 1 NA N (BOP) Start RCIC IAW SOP-0035, Section 4.2 2 NA N Adjust MVAR on Main Generator (SRO,ATC) 3 NA R (ATC) Raise Power IAW RMP for sequence exchange 4 T DI-HVR-UC1A C HVR-UC1A trip (TS)

(SRO,BOP)

LO_HVR-UC1A-A P863_71a:f_3 5 T NMS011F I (SRO,ATC) APRM F fails upscale (TS) 6 T MSS010 I (SRO,BOP) Loss of Steam Seal Header Pressure 7 T RCIC004 M (ALL) Steam Leak in RCIC Room - Spreads to RHR-C RMS215/219 RCIC Steam Supply Isol Valves (F063 & 64) fail to shut RCIC007/008 Rx Scram 8 MSS008D,O,G C SRV Failures B21RV41B, D, F (SRO,BOP) 9 RMS110A C Auxiliary Building Isolation Damper (HVR-AOD164) fails to auto shut RMS110B (SRO,BOP)

LO-HVR-AOD-164-G LO-HVR-AOD-164-R

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Total Malfunctions (5-8) (6) HVR-UC1A Trip, APRM upscale, SSE Press, Steam Leak, SRV Failures, HVR-AOD164 fail Malfunctions after EOP entry (1-2) (2) SRVs, HVR-AOD164 isolation failure Abnormal Events (2-4) (3) AOP-0001, AOP-0002, ARP for SSE Press Major Transients (1-2) (1) Steam Leak in RCIC Room - Spreading EOPs entered (1-2) (2) EOP-0001, EOP-0003 EOP contingencies (0-2) (1) EOP-0003 Critical Tasks (2-3) (2) ED within 20 minutes following exceeding max safe Table H in two areas, Isolate the auxiliary building within 25 minutes of RMS-RE110 reaching 6.02E-3 uCi /ml

December 2014 NRC Exam General Scenario Outline Event 0 - The team assumes the shift with reactor power at 80% for control rod sequence exchange Event 1 - After taking the shift the team will start RCIC CST to CST for post maintenance testing Event 2 - After the start of RCIC, the System Operations Center (SOC) requests that main generator MVAR be raised prior to increasing reactor power Event 3 - The team will raise reactor power with control rods as part of the control rod sequence exchange Event 4 - Following control rod with draw Containment Unit Cooler A trips requiring start of the non-safety related unit cooler C. Entry is made into Tech Spec 3.6.1.7 Event 5 - Next APRM F fails upscale resulting in a 1/2 scram. Entry into Tech Spec 3.3.1.1 will be made and the APRM will be bypassed and the 1/2 scram reset Event 6 - The inlet valve to the turbine steam seal pressure regulator fails next requiring the team to utilize the bypass valve to restore steam seal header pressure prior to a loss of condenser vacuum Event 7 - A steam leak then occurs in the RCIC room. RCIC room temperature and radiation levels begin to rise. The unit is taken offline (Scrammed) when the leak is unable to be isolated. As the steam leak continues radiation levels continue to rise, because the door between the RCIC room and RHR C was left open as the maintenance personnel exited the room, RHR C radiation levels begin to rise. The team will enter EOP-0003 for secondary containment control and track radiation levels.

When radiation level reach the MAX SAFE values in both rooms the RPV will be emergency depressurized (Critical Task to emergency depressurize within 20 minutes of exceeding max safe in two or more areas in Table H- additional SRVs must be opened to ensure RPV is depressurized)

Event 8 - During the emergency depressurization three ADS safety relief valves will fail to open requiring the team to open additional safety relief valves to ensure the reactor is depressurized Event 9 - Auxiliary building exhaust radiation levels require manual isolation of the auxiliary building. A failure to automatically close on HVR-AOD164 requires it to be manually closed via the hand switch (Critical Task to isolate the auxiliary building within 25 minutes of RMS-RE110 reaching 6.02E-3 uCi/ml)

December 2014 NRC Exam Appendix D Scenario Outline Form ES-D-1 Facility: River Bend Station Scenario No.: NRC-1 (Spare) IC No.: 251 Examiners: ________________ Operators: _____________________________

Initial Conditions: 84% reactor power B21-SRV51C is leaking - resulting in slowly rising suppression pool level ADS-SRV Leaking Alarm is in CWS-P1D tagged out for motor rewind Turnover Shift priorities: 1) Plant Shutdown in-progress due to leaking SRV; Reactor Engineering directs perform Step 2 of the RMP and then contact RE further guidance. 2) Perform SOP-0031, Section 4.8, Suppression Pool Reject to Radwaste Event Malf. Event Event No. No. Type* Description 0 MSS007K NA SRV 51C is leaking (Initial Condition) 1 NA R (ATC) Insert Control Rods per RMP step 2 (Reactivity Manipulation) 1a CRDM3633 TS (SRO) Control Rod Accumulator Fault (Tech Spec) 2 NA N (SRO, BOP) Perform SOP-0031 Section 4.8, Suppression Pool Reject to Radwaste 3 T RHR010A C (SRO, BOP) RHR A Pump Shaft Shear (Tech Spec) 4 T FWS017A I (SRO, ATC) Steam Line Flow Transmitter A fails downscale (AOP-0006) 5 NA N (SRO,ATC) Perform SOP-0009 Section 6.3, Three Element to Single Element Control Transfer (Booth call if necessary) 6 T ED003D C (SRO, ATC) Loss of NNS-SWG2A (AOP-0005) 7 CRDM2017 M (ALL) Reactor scram with turbine trip due to loss of vacuum CRDM3641 ATWS - 3 Control Rods CRDM4425 8 MSS005K C (SRO, BOP) Safety Relief Valve 51C fails open (AOP-0035) 9 RHS- C (BOP) RHS-AOV64 Fails to auto isolate. (AOP-0003)

AOV64P

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Total Malfunctions (5-8) (7) Accum Fault, RHR A Shear, MSL A Flow XMT, Loss NNS-SWG2A, SRV open, ATWS, RHS-AOV64 failure Malfunctions after EOP entry (1-2) (2) SRV open, failure to isolate Abnormal Events (2-4) (4), AOP-0006, AOP-0005, AOP-0035, AOP-0003 Major Transients (1-2) (1) Reactor scram with turbine trip due to loss of vacuum / 3 Control Rod ATWS

December 2014 NRC Exam EOPs entered (1-2) (2) EOP-0001/ EOP-0001A EOP contingencies (0-2) (1) Enter EOP-0001A Critical Tasks (2-3) (2) SRV closed; All control rods inserted General Scenario Outline Event 0 - The crew takes the shift with a plant shutdown (84%) in progress due to a leaking SRV. CWS-P1D is tagged out for motor rewind.

Event 1 - The team reduces reactor power with control rods per Step 2 of the shutdown reactivity maneuvering plan, which lowers reactor power to 80%. Following movement of the last control rod an accumulator fault occurs due to high moisture resulting in TRM 3.1.5.1 entry Event 2 - The team now starts RHR A in suppression pool cooling mode and begins reject of suppression pool to the radioactive waste system per SOP-0031 section 4.8.

Event 3 - During establishment of suppression pool reject the shaft shears on RHR pump A requiring entry into Tech Spec 3.5.1 and 3.6.2.3. The system lineup is secured Event4 - The team next combats a failure of the A steam flow transmitter (down-scale) for the feedwater level control system. Manual control is established to stabilize and restore RPV water level per Abnormal Operating Procedure AOP-0006.

Event5 - Following failure of the steam flow transmitter Single Element Control is established by the team to allow for automatic RPV level control per SOP-0009 Section 6.3.

Event6 - Next a loss of circulating water switchgear NNS-SWG2A bus occurs, resulting in a loss of most of the condenser cooling pumps requiring entry into abnormal operating procedure AOP-0005 for a loss of condenser vacuum and the plant will have to be shutdown (Scrammed) due to loss of vacuum.

Event 7 - After the mode switch is taken to shutdown, three control rods fail to insert resulting in a low power (0%) ATWS condition. EOP-0001A is entered and action taken to insert the control rods (Critical task to insert all control rods prior exiting EOP-1A, RPV CONTROL, ATWS)

Event 8 - Also following plant shutdown the leaking safety relief valve (SRV 51C) opens fully resulting in an uncontrolled RPV pressure drop, actions are taken per abnormal operating procedure AOP-0035 and the SRV is successfully closed.

(Critical task to close the SRV prior to exceeding the Heat Capacity Temperature Limit)

Event 9 - The final event is a failure of secondary containment isolation valve for the suppression pool cleanup system (RHS-AOV64) to isolate on the low RPV water level 3 received (as expected) during plant shutdown, requiring entry into AOP-0003.