ML14059A255

Draft - Outlines (Folder 2)

ML14059A255
Person / Time
Site:	Nine Mile Point
Issue date:	01/16/2014
From:	Exelon Generation Co
To:	Fish T H, Fuller B J Operations Branch I
Jackson D E
Shared Package
ML13295A030	List: ML14059A010 ML14059A254 ML14059A255 ML14059A256 ML14059A258 ML14076A120 ML14076A121 ML14076A122 ML14076A123
References
TAC U01875
Download: ML14059A255 (42)

Text

ES-401 Written Outline Form ES-401-1 Facility:

Nine Mile Point Unit 2 Date of Exam: March 2014 RO KIA 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 4 4 3 3 3 3 20 3 4 7 1. Emergency 2 1 1 2 ,* 1 1 :. 1 7 2 1 3 & .. *,*** Plant Tier *.,; Evolutions Total 5 5 5 ... 4 4 4 27 5 5 10 s 1 2 2 3 3 3 2 2 2 2 3 2 26 2 3 5 2. 2 1 1 Plant 1 1 1 1 1 1 1 1 2 12 0 1 2 3 Systems Tier Total 3 3 4 4 4 3 3 3 3 4 4 38 3 5 8 s 3. Generic Knowledge

& Abilities 1 2 3 4 1 2 3 4 10 7 Categories 2 3 3 2 2 1 2 2 Note 1. Ensure that at least two topics from every applicable KIA 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 KIA 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 that are not included on the outline should be added. Refer to section D.1.b of ES-401, for guidance regarding elimination of inappropriate KIA 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 KAs having an importance rating (IR) of 2.5 or higher shall be selected.

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

6. Select SRO topics for Tiers 1 and 2 from the shaded systems and KIA categories.

7.* The generic (G) KIAs in Tiers 1 and 2 shall be selected from Section 2 of the KIA 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 KIA's 8. On the following pages, enter the KIA numbem, a brief description of each topic, the topics' importance ratings (IR) 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 f</A Catalog, and enter the KIA numbers, descriptions, IRs, and point totals(#)

on Form ES-401-3.

Limit SRO selections to KIAs that are linked to 1 OCFR55.43 ES-401 Nine Mile Point Unit 2 Written Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions -Tier 1 Group 1 EAPE #I Name Safety Function Gi I KIA Topic(s) Jlmp.J Q# AA2.03 -Ability to determine 295004 Partial or and/or interpret the following Complete Loss of DC X as they apply to PARTIAL OR 2.9 76 Power I 6 COMPLETE LOSS OF D.C. POWER: Battery voltage AA2.03 -Ability to determine 295023 Refueling and/or interpret the following Accidents I 8 X as they apply to REFUELING 3.8 77 ACCIDENTS:

Airborne contamination levels AA2.01 -Ability to determine and/or interpret the following 295021 Loss of X as they apply to LOSS OF 3.6 78 ShutdoW:n Cooling I 4 SHUTDOWN COOLING: Reactor water heatup/cooldown rate 2.4.46 -Emergency 295018 Partial or Procedures I Plan: Ability to Complete Loss of CCW X verify that the alarms are 4.2 79 /8 consistent with the plant conditions.

2.2.25 -Equipment Control: 295025 High Reactor Knowledge of the bases in X Technical Specifications for 4.2 80 Pressure I 3 limiting conditions for operations and safety limits. 295003 Partial or 2.4.41 -Emergency Complete Loss of AC X Procedures I Plan: Knowledge 4.6 81 Power I 6 of the emergency action level thresholds and classifications.

2.4.8-Emergency Procedures 700000 Generator I Plan: Knowledge of how Voltage and Electric X abnormal operating 4.5 82 Grid Disturbances procedures are used in conjunction with EOPs. AK1.02-Knowledge of the operational implications of the 295006 SCRAM I 1 X following concepts as they 3.4 39 apply to SCRAM: Shutdown margin AK1.04-Knowledge of the operational implications of the 295001 Partial or following concepts as they Complete Loss of X apply to PARTIAL OR 2.5 40 Forced Core Flow COMPLETE LOSS OF Circulation I 1 & 4 FORCED CORE FLOW CIRCULATION:

Limiting cycle oscillation:

Plant-Specific ES-401 Nine Mile Point Unit 2 Written Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions -Tier 1 Group 1 EAPE #I Name Safety Function KIA Topic(s) AK1.01 -Knowledge of the operational implications of the 295018 Partial or following concepts as they apply to PARTIAL OR 3.5 41 Complete Loss of CCW X COMPLETE LOSS OF /8 COMPONENT COOLING WATER: Effects on component/system operations EK2.03-Knowledge of the 295038 High Off-site interrelations between HIGH X OFF-SITE RELEASE RATE 3.6 42 Release Rate I 9 and the following:

Plant ventilation systems EK2.04 -Knowledge of the interrelations between HIGH 295028 High Drywell X DRYWELL TEMPERATURE 3.6 43 Temperature I 5 and the following:

Drywell ventilation AK2.06-Knowledge of the interrelations between 295003 Partial or PARTIAL OR COMPLETE 3.4 44 Complete Loss of AC X LOSS OF A.C. POWER and Power I 6 the following:

D.C. electrical loads AK3.02-Knowledge of the reasons for the following responses as they apply to 700000 Generator GENERATOR VOLTAGE AND Voltage and Electric X ELECTRIC GRID 3.6 45 Grid Disturbances DISTURBANCES:

Actions contained in abnormal operating procedure for voltaae and arid disturbances AK3.02-Knowledge of the reasons for the following 295021 Loss of X responses as they apply to 3.3 46 Shutdown Cooling I 4 LOSS OF SHUTDOWN COOLING: Feeding and bleedinq reactor vessel EK3.04-Knowledge of the reasons for the following 295024 High Drywell X responses as they apply to 3.7 47 Pressure /5 HIGH DRYWELL PRESSURE:

Emergency depressurization AA 1.01 -Ability to operate 295016 Control Room and/or monitor the following as X they apply to CONTROL 3.8 48 Abandonment I 7 ROOM ABANDONMENT:

RPS ES-401 Nine Mile Point Unit 2 Written Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions -Tier 1 Group 1 EAPE #I Name Safety Function G I KIA Topic(s) limp. I Q# II AA 1. 08 -Ability to operate and I or monitor the following as 600000 Plant Fire On-X they apply to PLANT FIRE ON 2.6 49 site I 8 SITE: Fire fighting equipment used on each class of fire AA 1.07-Ability to operate and/or monitor the following as 295005 Main Turbine X they apply to MAIN TURBINE 3.3 50 Generator Trip I 3 GENERATOR TRIP: A.C. electrical distribution EA2.02 -Ability to determine and/or interpret the following 295026 Suppression X as they apply to 3.8 51 Pool High Water SUPPRESSION POOL HIGH Temperature I 5 WATER TEMPERATURE:

Su__QQ_ression pool level EA2.04-Ability to determine and/or interpret the following 295025 High Reactor X as they apply to HIGH 3.9 52 Pressure I 3 REACTOR PRESSURE:

Suppression pool level EA2.02 -Ability to determine and/or interpret the following 295031 Reactor Low X as they apply to REACTOR 4.0 53 Water Level I 2 LOW WATER LEVEL: Reactor power 295019 Partial or 2.1.20 -Conduct of Complete Loss of X Operations:

Ability to interpret 4.6 54 Instrument Air I 8 and execute procedure steps. 2.1.27 -Conduct of 295030 Low Operations:

Knowledge of 3.9 55 Suppression Pool Water X system purpose and I or Level I 5 function.

2.1.32 Conduct of 295023 Refueling X Operations:

Ability to explain 3.8 56 Accidents I 8 and apply system limits and precautions.

AK1.05-Knowledge of the operational implications of the following concepts as they 295004 Partial or apply to PARTIAL OR COMPLETE LOSS OF D.C. 3.3 57 Complete Loss of DC X POWER: Loss of breaker Power/6 protection ES-401 Nine Mile Point Unit 2 Written Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions -Tier 1 Group 1 EAPE #I fl!ame Safety Function G I KIA Topic(s) limp. I Q# EK2.11 -Knowledge of the 295037 SCRAM interrelations between SCRAM Condition Present and CONDITION PRESENT AND Reactor Power Above X REACTOR POWER ABOVE 3.8 58 APRM Downscale or APRM DOWNSCALE OR Unknown I 1 UNKNOWN and the following:

RMCS: Plant-Specific KIA Category Totals: 4 4 3 3 3/3 3/4 Group Point Total: 1 20/7 ES-401 Nine Mile Point Unit 2 Written Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions -Tier 1 Group 2 EAPE #I Name Safety Function KIA Topic(s) EA2.04-Ability to determine and I or interpret the following 500000 High as they apply to HIGH Containment Hydrogen X PRIMARY CONTAINMENT 3.3 83 Concentrations I 5 HYDROGEN CONCENTRATIONS:

Combustible limits for wetwell 2.4.47-Emergency Procedures 295029 High I Plan: Ability to diagnose and Suppression Pool Water X recognize trends in an accurate 4.2 84 Level I 5 and timely manner utilizing the appropriate control room reference material.

AA2.01 -Ability to determine 295022 Loss of CRD X and/or interpret the following as 3.6 85 Pumps I 1 they apply to LOSS OF CRD PUMPS: Accumulator pressure AK1.05-Knowledge of the operational implications of the 2950201nadvertent following concepts as they apply Containment Isolation I 5 X to INADVERTENT 3.3 59 &7 CONTAINMENT ISOLATION:

Loss of drywell/containment cooling EK2.08 -Knowledge of the 295029 High interrelations between HIGH SUPPRESSION POOL WATER Pool Water X LEVEL and the following:

2.6 60 Level I 5 Drywell/suppression chamber ventilation AK3.01 -Knowledge of the 295022 Loss of CRD reasons for the following Pumps I 1 X responses as they apply to 3.7 61 LOSS OF CRD PUMPS: Reactor SCRAM AA 1.03 -Ability to operate 295009 Low Reactor and/or monitor the following as Water Level I 2 X they apply to LOW REACTOR 3.0 62 WATER LEVEL: Recirculation system: Plant-Specific EA2.01 -Ability to determine 295032 High Secondary and/or interpret the following as Containment Area X they apply to HIGH 3.8 63 Temperature I 5 SECONDARY CONTAINMENT AREA TEMPERATURE:

Area temperature ES-401 Nine Mile Point Unit 2 Written Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions -Tier 1 Group 2 EAPE #I Name Safety Function G I KIA Topic(s) Imp. I Q# 295010 High Drywell 2.4.11 -Emergency Procedures X I Plan: Knowledge of abnormal 4.0 64 Pressure I 5 condition procedures.

EK3.03-Knowledge of the reasons for the following 500000 High responses as they apply to HIGH PRIMARY Containment Hydrogen X CONTAINMENT HYDROGEN 3.0 65 Concentrations I 5 CONCENTRATIONS:

Operation of hydrogen and oxygen recombiners KIA Category Totals: 1 1 2 1 11 11 Group Point Total: 1 713 2 1 ES-401 System # I Name K K K 1 2 3 209001 LPCS 212000 RPS 262002 UPS (AC/DC) 218000 ADS 261000 SGTS 215004 Source Range X Monitor Form ES-401-1 Nine Mile Point Unit 2 Written Examinat1ion Outline Plant Systems -Tier 2 Group 1 K K K A A2 A G I 4 5 6 1 3 A2.03-Ability to (a) predict the impacts of the following on the LOW PRESSURE CORE SPRAY SYSTEM; and (b) X based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

A.C. failures A2.18-Ability to (a) predict the impacts of the following on the REACTOR PROTECTION SYSTEM; and (b) based on those X predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

SCRAM air header low pressure 2.1.23-Conduct of Operations:

Ability to X perform specific system and integrated plant procedures during all modes of plant operation.

2.4.6 -Emergency X Procedures I Plan: Knowledge of EOP mitigation strategies.

2.2.12 -Equipment X Control: Knowledge of surveillance procedures.

K1.01 -Knowledge of the physical connections and/or cause-effect relationships between SOURCE RANGE MONITOR (SRM) SYSTEM and the following:

Reactor protection system I'm' I a* I 3.6 86 3.9 87 4.4 88 4.7 89 4.1 90 3.6 1 ES-401 400000 Component X Cooling Water 212000 RPS X 262001 AC Electrical X Distribution 264000 EDGs X 215005 APRM I LPRM X 203000 RHR/LPCI:

Injection Mode Form ES-401-1 Nine Mile Point Unit 2 Written Examination Outline Plant Systems -Tier 2 Group 1 K1.02-Knowledge of the physical connections and I or cause-effect relationships between CCWS and the following:

Loads cooled by CCWS K2.01 -Knowledge of electrical power supplies to the following:

RPS sets K2.01 -Knowledge of electrical power supplies to the following:

Off-site sources of power K3.02-Knowledge of the effect that a loss or malfunction of the EMERGENCY GENERATORS (DIESEL/JET) will have on following:

A.C. electrical distribution K3.07-Knowledge of the effect that a loss or malfunction of the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM will have on following:

Rod block monitor: Plant-Specific K4.07-Knowledge of RHR/LPCI:

INJECTION MODE (PLANT SPECIFIC) design X feature(s) and/or interlocks which provide for the following:

Emergency generator load sequencing 3.2 2 3.2 3 3.3 4 3.9 5 3.2 6 3.7 7 ES-401 I Sy*tem #I N*mo 215003 IRM 300000 Instrument Air 259002 Reactor Water Level Control 223002 PCIS/Nuclear Steam Supply Shutoff 262002 UPS (AC/DC) Form ES-401-1 Nine Mile Point Unit 2 Written Examination Outline Plant Systems -ner 2 Group 1 K4.05-Knowledge of INTERMEDIATE RANGE MONITOR (IRM) SYSTEM design X feature(s) and/or interlocks which provide for the following:

Changing detector position K5.13-Knowledge of the operational implications of X the following concepts as they apply to the INSTRUMENT AIR SYSTEM: Filters K5.01 -Knowledge of the operational implications of the following concepts as they apply to REACTOR X WATER LEVEL CONTROL SYSTEM: GEMAC/Foxboro/Bailey controller operation:

Plant-Specific K6.02-Knowledge of the effect that a loss or malfunction of the following will have on the PRIMARY X CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF: D.C. electrical distribution K6.02-Knowledge of the effect that a loss or malfunction of the X following will have on the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.):

D.C. electrical power 2.9 8 2.9 9 3.1 10 3.0 11 2.8 12 ES-401 System # I Name K K K 1 2 3 218000 ADS 211000 SLC 205000 Shutdown Cooling 209002 HPCS 209001 LPCS Form ES-401-1 Nine Mile Point Unit 2 Written Examination Outline Plant Systems -Tier 2 Group 1 K K K A A2 A G I 4 5 6 1 3 A1.04-Ability to predict and/or monitor changes in parameters associated with operating the X AUTOMATIC DEPRESSURIZATION SYSTEM controls including:

Reactor pressure A 1.07-Ability to predict and/or monitor changes in parameters associated X with operating the STANDBY LIQUID CONTROL SYSTEM controls including:

Reactor power A2.09-Ability to (a) predict the impacts of the following on the SHUTDOWN COOLING SYSTEM(RHR SHUTDOWN COOLING X MODE); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Reactor low water level A2.02 -Ability to (a) predict the impacts of the following on the HIGH PRESSURE CORE SPRAY SYSTEM (HPCS); and (b) based on X those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Pump trips: BWR-5,6 A3.01 -Ability to monitor automatic operations of X the LOW PRESSURE CORE SPRAY SYSTEM including:

Valve operation I'm' I Q# I 4.1 13 4.3 14 3.6 15 3.6 16 3.6 17 ES-401 263000 DC Electrical Distribution 261000 SGTS 217000 RCIC 264000 EDGs 223002 PCIS/Nuclear Steam Supply Shutoff 211000 SLC 212000 RPS 239002 Safety Relief Valves Nine Mile Point Unit 2 Written Examination Outline Plant Systems -Til3r 2 Group 1 Form ES-401-1 G Imp Q# X X X X X X X X A3.01 -Ability to monitor automatic operations of the D.C. ELECTRICAL DISTRIBUTION including:

3.2 18 Meters, dials, recorders, alarms, and indicating lights A4.06-Ability to manually operate and/or monitor in the control room: Reactor 3.3 19 building differential pressure A4.09-Ability to manually operate and/or monitor in the control room: System pressure 2.4.9-Emergency Procedures I Plan: Knowledge of low power I shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.

2.4.6 -Emergency Procedures I Plan: Knowledge of EOP mitiqation strateoies.

K4.02-Knowledge of STANDBY LIQUID CONTROL SYSTEM design feature(s) and/or interlocks which provide for the following:

Component and system testing K3.07-Knowledge of the effect that a loss or malfunction of the REACTOR PROTECTION SYSTEM will have on following:

Reactor power (thermal heat flux) A4.01 -Ability to manually operate and/or monitor in the control room: SRVs 3.7 20 3.8 21 3.7 22 3.0 23 3.8 24 4.4 25 ES-401 System # I Name K K K 1 2 3 262001 AC Electrical Distribution KJA Category Totals: 2 2 3 Form ES-401-1 Nine Mile Point Unit 2 Written Examination Outline Plant Systems -Tier 2 Group 1 K K K A A2 A A G 4 5 6 1 3 4 K5.01 -Knowledge of the operational implications of the following concepts as X they apply to A. C. ELECTRICAL DISTRIBUTION:

Principle involved with paralleling two A. C. sources 3 3 2 2 21 2 3 2/ Group Point Total: 2 3 Imp Q# 3.1 26 I 26/5 ES-401 I System# t'Nam* 215002 RBM 272000 Radiation Monitoring 241000 Reactor/Turbine Pressure Regulating System 245000 Main Turbine Generator and Auxiliary Systems 286000 Fire Protection 202002 Recirculation Flow Control Form ES-401-1 Nine Mile Point Unit 2 Written Examination Outline Plant Systems -Tier 2 Group 2 G I A2.04 -Ability to (a) predict the impacts of the following on the ROD BLOCK MONITOR SYSTEM; and (b) based X on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Power supply losses: BWR-3,4,5 2.2.22 -Equipment Control: Knowledge of X limiting conditions for operations and safety limits. 2.1.31 -Conduct of Operations:

Ability to locate control room X switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup. K1.06-Knowledge of the physical connections and/or cause-effect relationships between X MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS and the following:

Component cooling water systems K2.02-Knowledge of X electrical power supplies to the following:

Pumps K3.04-Knowledge of the effect that a loss or malfunction of the X RECIRCULATION FLOW CONTROL SYSTEM will have on following:

Reactor/turbine pressure regulation system 2.8 91 4.7 92 4.3 93 2.6 27 2.9 28 2.9 29 ES-401 K K K System # I Name 1 2 3 204000 RWCU 271000 Offgas 230000 RHR/LPCI:

Torus/Pool Spray Mode 226001 RHR/LPCI:

Containment Spray Mode 202001 Recirculation 259001 Reactor Feedwater Form ES-401-1 Nine Mile Point Unit 2 Written Examination Outline Plant Systems -Tier 2 Group 2 K K K A A2 A G I 4 5 6 1 3 K4.04-Knowledge of REACTOR WATER CLEANUP SYSTEM X design feature(s) and/or interlocks which provide for the following:

System isolation upon receipt of isolation signals K5.04-Knowledge of the operational implications of the following concepts X as they apply to OFFGAS SYSTEM: Hydrogen concentration measurement K6.01 -Knowledge of the effect that a loss or malfunction of the following will have on the X RHR/LPCI:

TORUS/SUPPRESSION POOL SPRAY MODE: I A.C. electrical i A1.01 -Ability to predict and/or monitor changes in parameters associated with operating the ' RHR/LPCI:

X ! CONTAINMENT SPRAY SYSTEM MODE controls including:

Containmentldrywell pressure A2.23 -Ability to (a) predict the impacts of the following on the RECIRCULATION SYSTEM; and (b) based X on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Valve closures A3.03-Ability to monitor automatic operations of X the REACTOR FEEDWATER SYSTEM including:

System flow limp I 3.5 30 2.9 31 3.3 32 3.6 33 3.2 34 3.3 35 ES-401 System #I Name K K K 1 2 3 223001 Primary Containment and Auxiliaries 268000 Radwaste 239001 Main and Reheat Steam KIA Category Totals: 1 1 1 Form ES-401-1 Nine Mile Point Unit 2 Written Examination Outline Plant Systems -Tier 2 Group 2 K K K A A2 A A G 4 5 6 1 3 4 A4.12 -Ability to manually operate and/or X monitor in the control room: Drywell coolers/chillers 2.1.7-Conduct of Operations:

Ability to evaluate plant performance and make X operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

2.2.44 -Equipment Control: Ability to interpret control room I indications to verify the ' X status and operation of a i system, and understand I how operator actions and directives affect plant and system conditions. 1 1 1 1 1/ 1 1 21 Group Point Total: 1 'I 2 Imp. Q # 3.5 36 4.4 37 4.2 38 1 12/3 Facility:

Nine Mile Point Unit 2 Date: March 2014 RO SRO-Only Category KIA# Topic IR Q# IR Q# 2.1.20 Ability to interpret and execute procedure 4.6 94 steps. 2.1.42 Knowledge of new and spent fuel movement 3.4 98 procedures.

Knowledge of RO duties in the control room during fuel handling such as responding to 1. alarms from the fuel handling area, Conduct 2.1.44 communication with fuell storage facility, 3.9 66 of Operations systems operated from the control room in support of fueling operations, and supporting instrumentation.

Ability to identify and interpret diverse 2.1.45 indications to validate the response of 4.3 67 another indicator.

Subtotal 2 2 2.2.6 Knowledge of the process for making 3.6 95 changes to procedures.

2. 2.2.43 Knowledge of the process used to track 3.0 68 Equipment inoperable alarms. Control Knowledge of the process for conducting 2.9 69 2.2.7 special or infrequent tests. 2.2.37 Ability to determine operability and I or 3.6 74 availability of safety related equipment.

Subtotal 3 1 Knowledge of radiological safety principles pertaining to licensed operator duties, such 2.3.12 as containment entry requirements, fuel 3.7 96 handling responsibilities, access to locked high-radiation

areas, filters, etc. Knowledge of radiation monitoring systems, 3. 2.3.15 such as fixed radiation monitors and alarms, 3.1 100 Radiation portable survey instruments, personnel Control . monitoring equipment, etc. 2.3.11 Ability to control radiation releases.

3.8 70 2.3.4 Knowledge of radiation exposure limits 3.2 71 under normal or emergency conditions.

Knowledge of radiation monitoring systems, 2.3.15 such as fixed radiation monitors and alarms, 2.9 75 portable survey instruments, personnel monitoring equipment, etc. Subtotal 3 2 Knowledge of facility protection 2.4.26 requirements, including fire brigade and 3.6 97 portable fire fighting equipment usage. Knowledge of the parameters and logic used to assess the status of safety functions, 2.4.21 such as reactivity control, core cooling and 4.6 99 heat removal, reactor coolant system 4. integrity, containment conditions, radioactivity release control, etc. Emergency Procedures I Plan 2.4.31 Knowledge of annunciator alarms, 4.2 72 indications, or response procedures.

2.4.41 Knowledge of the action level 2.9 73 thresholds and classifications.

Subtotal 2 2 Tier 3 Point Total 10 7 ES-401 . 1 /1 . 1 /2 2/1 2/1 2/2* 2/2 G G Record of Rejected K/As Form ES-401-4 Randomly Selected KIA 295027 High Containment Temperature 295011 High Containment Temp 206000 HPCI 207000 Isolation (Emergency)

Condenser 201005 RCIS 239003 MSIV Leakage Control 2.2.3 Knowledge of the design, procedural, and operational differences between units. 2.2.4 Ability to explain the variations in control board/control room layouts, systems, instrumentation, and procedural actions between units at a facility.

Reason for Rejection This topic applies to plants with Mark Ill containments only. The facility has a Mark II containment.

This topic applies to plants with Mark Ill containments only. The facility has a Mark II containment.

This system is not installed at the facility.

This system is not installed at the facility.

This system is not installed at the facility.

This system is not installed at the facility.

This KIA applies to multi-unit facilities only. This KIA applies to multi-unit facilities only.

Question 56 A discriminating question at the appropriate license level could not be developed for the 295023 Refueling Accidents randomly sampled generic KIA. Additionally, 2.4.41 -Emergency Procedures the randomly sampled generic KIA overlapped I Plan: Knowledge of the with that of question #73. 1 I 1 emergency action level Randomly re-selected KIA 295023 Refueling thresholds and classifications.

Accidents 2.1.32-Conduct of Operations:

Ability to explain and apply system limits and precautions.

Question 10 The facility does not have a FWCI mode of the 259002 Reactor Water Level Feedwater system. Control Randomly re-selected KIA 259002 Reactor Water Level Control K5.01 -Knowledge of the K5.08-Knowledge of the operational implications of the following 2 I 1 operational implications of the concepts as they apply to REACTOR WATER following concepts as they LEVEL CONTROL SYSTEM: apply to REACTOR WATER GEMAC/Foxboro/Bailey controller operation:

LEVEL CONTROL SYSTEM: Plant-Specific Heat removal mechanisms:

FWCI Question 7 An adequate question meeting the KIA could not be constructed, as the facility ultimately 203000 RHR/LPCI:

Injection ensures pump runout protection by adhering to Mode procedural pump limitations, not a design K4.15 -Knowledge of feature or interlock.

RHR/LPCI:

INJECTION MODE Randomly re-selected KIA 203000 RHR/LPCI:

2 I 1 (PLANT SPECIFIC) design lnjeGtion Mode K4.07-Knowledge of feature(s) and/or interlocks RHFULPCI:

INJECTION MODE (PLANT which provide for the following:

SPECIFIC) design feature(s) and/or interlocks Pump runout protection:

Plant-which provide for the following:

Emergency Specific generator load sequencing Question 84 A discriminating question at the appropriate license level could not be developed for the 295029 High Suppression Pool randomly sampled generic KIA. Water Level Randomly re-selected KIA 295029 High 1 I 2 2.1.28-Conduct of Operations:

Suppression Pool Water Level 2.4.47-Ability Knowledge of the purpose and to diagnose and recognize trends in an function of major system accurate and timely manner utilizing the components and controls.

appropriate control room reference material.

Question 92 A discriminating question at the appropriate license level could not be developed for the 272000 Radiation Monitoring randomly sampled generic KIA. 2.1.28 -Conduct of Operations:

Randomly re-selected KIA 272000 Radiation 2/2 Knowledge of the purpose and Monitoring 2.2.22 -Knowledge of limiting function of major system conditions for operations and safety limits. components and controls.

Question 88 A discriminating question at the appropriate license level could not be developed for the 262002 UPS (AC/DC) randomly sampled generic KIA. 2.1.19 -Conduct of Operations:

Randomly re-selected KIA 262002 UPS 2/1 Ability to use plant computers to (AC/DC) 2.1.23 -Ability to perform specific . evaluate system or component system and integrated plant procedures dunng status. all modes of plant operation.

Question 89 A discriminating question at the appropriate license level could not be developed for the 218000 ADS randomly sampled generic KIA. 2.4.1 -Emergency Procedures I Randomly re-selected KIA 218000 ADS 2.4.6-2/1 Plan: Knowledge of EOP entry Knowledge of EOP mitigation strategies.

conditions and immediate action steps. Question 38 A discriminating question at the appropriate license level could not be developed for the 239001 Main and Reheat randomly sampled generic KIA. Steam Randomly re-selected KIA 239001 Main and 2.2.40 -Equipment Control: Reheat Steam 2.2.44-Ability to interpret 2/2 Ability to apply technical control room indications to verify the status and specifications for a system. operation of a system, and understand how operator actions and directives affect plant and system conditions.

Question 96 The randomly sampled generic KIA overlaps 2.3.4 -Knowledge of radiation with Question 71. exposure limits under normal or Randomly re-selected KIA 2.3.12 -Knowledge emergency conditions.

of radiological safety principles pertaining to 3 licensed operator duties, such as containment entrv requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

Question 80 The randomly sampled generic KIA is identical to that for Question 90 and not well suited for a 295025 High Reactor Pressure discriminating and valid question.

2.2.12 -Equipment Control: Randomly re-selected KIA 295025 High 1 11 Knowledge of surveillance Reactor Pressure 2.2.25-Equipment Control: procedures.

Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits. Question 81 The randomly sampled generic KIA is identical to that for Question 90 and not well suited for a 295003 Partial or Complete discriminating and valid question.

Loss of AC Power Randomly re-selected KIA 295003 Partial or 1 11 2.2.12-Equipment Control: Complete Loss of AC Power 2.4.41 -Knowledge of surveillance Emergency Procedures I Plan: Knowledge of procedures.

the emergency action level thresholds and classifications.

Question 47 The facility does not have an Auxiliary Building.

295024 High Drywell Pressure Randomly re-selected KIA 295024 High Drywell Pressure EK3.04-Knowledge of the reasons EK3.09-Knowledge of the for the following responses as they apply to reasons for the following HIGH DRYWELL PRESSURE:

Emergency 1 11 responses as they apply to depressurization.

HIGH DRYWELL PRESSURE:

Auxiliary building isolation:

Plant-Specific Question 37 There are no specific bases in the EOPs that relate to Radwaste to support construction of a 268000 Radwaste valid question with the randomly sampled KIA. 2.4.18 -Emergency Procedures Randomly re-selected KIA 268000 Radwaste I Plan: Knowledge of the 2.1. 7 -Conduct of Operations:

Ability to 212 specific bases for EOPs. evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

Question 57 An acceptable question could not be developed without either significant overlap with Question 295004 Partial or Complete 44 or testing Generic Fundamentals.

Loss of DC Power Randomly re-selected KIA 295004 Partial or AK 1 . 04 -Knowledge of the Complete Loss of DC Power AK1.05-1 I 1 operational implications of the Knowledge of the operational implications of the following concepts as they following concepts as they apply to PARTIAL apply to PARTIAL OR OR COMPLETE LOSS OF D.C. POWER: Loss COMPLETE LOSS OF D. C. of breaker protection POWER: Effect of battery discharge rate on capacity Question 65 An acceptable question could not be developed 500000 High Containment without significant overlap with either Question 60 or Question 83. Hydrogen Concentrations Randomly re-selected KIA 500000 High EK3.06 -Knowledge of the Containment Hydrogen Concentrations EK3.03 reasons for the following

-Knowledge of the reasons for the following 1 I 2 responses as they apply to responses as they apply to HIGH PRIMARY HIGH PRIMARY CONTAINMENT HYDROGEN CONTAINMENT HYDROGEN CONCENTRATIONS:

Operation of hydrogen CONCENTRATIONS:

and oxygen recombiners Operation of wet well vent Question 45 An acceptable question could not be developed 700000 Generator Voltage and because the procedures associated with Generator voltage and electrical grid Electric Grid Disturbances disturbance (N2-SOP-70, Major Grid AK3.01 -Knowledge of the Disturbances, and associated alarm response reasons for the following procedures) do not contain explicit Reactor and responses as they apply to Turbine trip criteria.

GENERATOR VOLTAGE AND Randomly re-selected KIA 700000 Generator 1 I 1 ELECTRIC GRID Voltage and Electric Grid Disturbances AK3.02 DISTURBANCES:

Reactor and -Knowledge of the reasons for the following turbine trip criteria responses as they apply to GENERA TOR VOLTAGE AND ELECTRIC GRID DISTURBANCES:

Actions contained in abnormal operating procedure for voltage and grid disturbances ES-301 Administrative Topics Outline Form ES-301-1 Facility:

__ __,_N=M.!.!'-'P2=--..:...:Nc.:..:R=C

__ _ Date of Examination:

March 2014 Examination Level: SRO Administrative Topic (see Note) Conduct of Operations Conduct of Operations Equipment Control Operating Test Number: NRC Type Code* Describe activity to be performed R,D R,M R,N Determine the Severity of a Reactivity Event and Actions Required Given a mispositioned control rod, the applicant will assess a the Reactivity Severity Level and take appropriate corrective actions KIA 2.1.37 (4.6) Knowledge of procedures, guidelines, or limitations associated with reactivity management.

CNG-OP-3.01-1 000 and N2-0P-96 Determine Plant Impact for Inoperable Unit Cooler Given a failed closed service water inlet valve to 2HVC*UC1 03A, the applicant will determine the effect on the unit cooler and Division 1 Chiller operability per N2-0P-53E and Technical Specifications 2.1.32 (4.0) Ability to explain and apply system limits and precautions.

N2-0P-53E and Technical Specifications Determine Components Which Need Protection The applicant will review plant conditions and determine which components need to be protected 2.2.14 (4.3) Knowledge of the process for controlling equipment configuration or status. S-ODP-OPS-0122 Radiation Control R,D, P Emergency Plan R,D Inspection of High Radiation Areas Given radiological conditions related to an area where work is to be performed as shown on a survey map, and other applicable conditions such as the RWP, ensure the appropriate radiological aspects of the job are met prior to sending the operator into the area. 2.3.12 (3. 7) Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc GAP-RPP-0'1, 02, 07 and 08; S-RAP-RPP-0703 Event Classification and Notifications Given a plant event, the applicant will determine classification and notification requirements. (Time Critical) 2.4.41 (4.6) Knowledge of the emergency action level thresholds and classifications.

EPI P-EPP-0:2 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 for ROs; 4 for SROs & RO retakes) (N)ew or (M)odified from bank (P)revious 2 exams randomly selected)

ES-301 Administrative Topics Outline Form ES-301-1 Facility:

__ ___,N_,_,M..:..:..:....P==-2

_-:...:.N.:..:.RC::::__

__ Date of Examination:

March 2014 Examination Level: RO Administrative Topic (see Note) Conduct of Operations Conduct of Operations Equipment Control Operating Test Number: NRC Type Code* Describe activity to be performed M,R N, R N,R Determine Containment Water Level The applicant will calculate Containment Water Level and take actions based on the results 2.1.25 (3.9) Ability to interpret reference materials, such as graphs, curv1es, tables, etc. N2-EOP-6.2:3 Develop and get approval for a Temporary Note The applicant will develop and get approval for a Temporary Note for a malfunctioning control switch. 2.1.15 (2. 7) Knowledge of administrative requirements for temporary management directives, such as standing orders, night orders, operations memos, etc. CNG-OP-1.01-1 005 Defeat the Reactor Building Ventilation LOCA Isolation Signals The applicant will use prints and drawings to explain how to defeat the Reactor Building Ventilation LOCA Isolation Signals. 2.2.41 (3.5) Ability to obtain and interpret station electrical and mechanical drawings N2-EOP-6.216 Fire Fighting Response for a Fire in the Protected Area The applicant will make the appropriate announcements for a fire in the protected area. Emergency Plan D,S 2.4.39 (3.9) Knowledge of RO responsibilities in emergency plan implementation EPIP-EPP-28 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 for ROs; 4 for SROs & RO retakes) (N)ew or (M)odified from bank (P)revious 2 exams (<1; randomly selected)

ES-301 Control Room/In-Plant Systems Outline Facility:

Nine Mile Point Unit 2 NRC Date of Examination:

Exam Level: RO/SRO Operating Test No.: Control Room Systems@ (8 for RO; 2 or 3 for SRO-U, including 1 ESF) S-1 S-2 S-3 S-4 System I JPM Title Rotate Drywall Unit Coolers from UC3A to 38 The applicant will start UC3B and secure When UC3A is shutdown, UC3B will develop high vibration which will require executing ARP 873213. ARP 873213 will direct shutting down UC3B by placing the control switch in Pull-To Lock. KIA 223001, A4.12. 3.5/3.6 N2-0P-60, Section 2.0 and ARP 873213 Perform Weekly RPS Surveillance The applicant will perform a RPS Weekly Surveillance on RPS Channels A. KIA 212000, A4.02 3.6/3.7 N2-0SP-RPS-W002 Maximize RDS Flow After Scram The applicant will maximize RDS flow by starting RDS-P1 B and opening both the Flow Control Valve and Drive Water Control Valve. Once flow is maximized, suction strainer clogging will cause annunciator 603318 to alarm. The applicant will call for an operator to open the bypass lines around the filters per N2-0P-30 KIA 201001, A4.01 3.1/3.1 N2-0P-30, Section H.3.0 and ARP 603318 Reset L V1 OB Lockout and Place FWLC Back in Automatic The applicant will reset a lockout on LV10B and place FWLC back in automatic.

Once the lockout is reset, the FWLC Master controller will fail causing RPV Level to ChangE!. The applicant will be required to take manual control of FWLC and restore level to normal band. KIA 259001, A4.05, 4.0/3.9 N2-SOP-06, Attachment 1, Section 1.3 Type Code* A,N,S D,S A, M, L, S A,N,S Form ES-301-2 March 2014 NRC Safety Function 5 CONTAINMENT INTEGRITY 7 INSTRUMENTATION 1 REACTIVITY CONTROL 2 REACTOR WATER INVENTORY CONTROL ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 S-5 Energize 2ENS*SWG1 03 from Division 2 EDG and D, P, L, S 6 Energize 2NNS-SWG015 (Stub Bus) from 2ENS*SWG103 ELECTRICAL During a Station Blackout The applicant will energize 2ENS*SWG1 03 from Division 2 EDG and energize 2NNS-SWG014 (Stub Bus) from 2ENS*SWG1

03. KJA 262001, A4.01 3.4/3.7 N2-SOP-03, Attachment 5, Section 5.3 S-6 Bypass RCIC Room High Temperature Isolation A, D, L, S 4 The applicant will bypass the RCIC Room High Temperature HEAT REMOVAL FROM THE CORE Isolation during a station blackout.

RCIC will trip requiring the trip to be reset. KJA 217000, A4.02 3.9/3.9 N2-SOP-02, Attachment 4 S-7 Override the Control Room Envelope ACU Cross-D,S 9 Divisional Operating Interlock RADIOACTIVITY RELEASE The applicant will override the Division 1 Control Room Envelope ACU Cross-Divisional Operating Interlocks.

KJA 290003 A3.01 3.3/3.5 N2-0P-53A, H.15.0 S-8 Respond to an Inadvertent Closure of 2SWP*MOV50B N,S 8 (RO PLANT SERVICE The applicant will respond to an inadvertent closure of SYSTEMS Only) 2SWP*MOV508.

KJA 400000, A4.01 3.1/3.0 N2-SOP-11, Flowchart and Attachment

1. In-Plant Systems@ (3 for RO; 3 or 2 for SRO-U) P-1 Local Start of Division 1 Diesel Generator A,D 6 The applicant will locally start the Division 1 Emergency ELECTRICAL Diesel Generator.

After locally starting, a low lube oil pressure alarm will require the EDG to be shutdown.

Initial efforts to shutdown the EDG will not be successful requiring the applicant to perform an Emergency Shutdown.

KJA 264000, A4.04 3.7/3.7 N2-0P-100A, Section F.5.0 and H.1.0 ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 P-2 Isolate a Hydraulic Control Unit with Cooling Water D,R 1 The applicant will isolate an HCU with cooling1 water per N2-REACTIVITY CONTROL OP-30. KIA 201003, A2.01 3.413.6 N2-0P-30, Section F.8.0 P-3 Align Firewater to RHS B D,E,R 2 The applicant will align Firewater to RHS B per N2-EOP-6.6 REACTOR WATER INVENTORY CONTROL KIA 203000, A2.02 3.513.5 N2-EOP-6.6, Section 6.2 @ All RO and SR0-1 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 I SR0-1 I SRO-U (A)Iternate path 4-6 I 4-6 I 2-3 (C)ontrol room (D)irect from bank $;91$;81$;4 (E)mergency or abnormal in-plant <:11<:11<:1 (EN)gineered safety feature -I -I <:1 (control room system) (L)ow-Power I Shutdown <:11<:11<:1 (N)ew or (M)odified from bank including 1 (A) <:21<:21<:1 (P)revious 2 exams $; 3 I$; 3 I $; 2 (randomly selected) (R)CA <:11<:11<:1 (S)imulator Appendix D Scenario Outline Form ES-D-1 Facility:

Nine Mile Point 2 Scenario No.: NRC-1 Op-Test No: March 2014 Examiners:

Operators:

Initial Conditions:

Simulator IC-150 1. Reactor Power is at 90% 2. 2WCS-P1 B is out of service for maintenance.

Turnover:

1. The crew will be required to raise power to 95% using recirculation flow Event Malf. Event Type* Event No. No. Description 1 N/A R (SRO) The crew will raise reactor power to 95% using recirculation flow. R (RO) N2-0P-101D

.* '*:;*.*. :* 2 CW01A C (BOP) Service Water Pump 1A trips on motor electric fault requiring the CW10E C (SRO) crew to manually start a standby service water pump. While TS (SRO) starting the standby pump, the associated discharge valve will fail to automatically open requiring the operator to manually open the valve. The CRS will declare the pump inoperable and evaluate TS3.7.1. ARP's, N2-0P-11, TS 3.7.1 . :C!/*. : .**; !.. ; .**::*. *< . ' ... :"' . .* ..

• 3 CS01B C (BOP) HPCS inadvertently initiates and injects into the core. FWLC will C (SRO) respond correctly and maintain RPV level below the Level 8 trip TS (SRO) setpoint.

The HPCS malfunction will prevent the system from being returned to standby which will required the crew to place the HPCS pump in pull to lock (PTL). With HPCS in PTL, the CRS will declare HPCS inoperable and evaluate TS 3.5.1. N2-0P-33, TS 3.5.1 . ... * . **. 4 RD12A C (RO) Suction strainer clogging will cause the running RDS pump to trip. C (SRO) The crew will respond per the SOP to swap suction strainers and restart the RDS pump. N2-SOP-30 . .{ " 5 ED02A C (BOP) A complete loss of offsite power will occur. The Division 1 EDG ED02B C (SRO) will auto start, however it will not close in on the bus because of DG04B C (RO) an electrical fault on the Division 1 Switchgear.

The Division 2 ED05A EDG will fail to automatically start. The crew must take action to manually start the Division 2 EDG (CRITICAL TASK) and power the Division 2 switchgear.

Due to the loss of Division 1 switchgear, the crew will manually scram the reactor, trip the turbine, and trip both recirculation pumps per N2-SOP-11.

N2-SOP-03, N2-SOP-11

... :' . ,. .'"/'"' : . NRC Scenario 1 March 2014 6 RR20 M (All) A small LOCA will occur causing drywell pressure to rise. The crew will respond to control RPV level and Pressure and begin actions to control Primary Containment (PC) pressure.

N2-EOP-RPV, N2-EOP-PC

.. ' i,;: ***********

.**.

• "co* * .. :* i * . . > *.** .* i . .*: . 7 CS05 C (RO/BOP) While attempting to maintain RPV level above the TAF, the crew SL03A C(SRO) will attempt to restart HPCS which will trip a few seconds after the RC01 crew takes the pump out of PTL. RCIC will fail to automatically start once the manual initiation switch is depressed.

The crew will RC06 manually start RCIC. Once RCIC is started, RCIC will trip due to an instrument failure. N2-EOP-RPV, N2-EOP-HC 8 N/A C (RO/BOP) Due to the LOCA and failure of adequate high pressure injection C(SRO) sources, the crew will blowdown the reactor (CRITICAL TASK) once RPV level reaches the TAF. The crew will then align appropriate low pressure ECCS injection sources to raise RPV level above the TAF. Once RPV level has recovered sufficiently, the scenario may be terminated.

N2-EOP-RPV, N2-EOP-C2

• (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor NRC Scenario 1 March 2014 Facility:

Nine Mile Point 2 Scenario No: NRC-1 Op-Test No: March 2014 TARGET QUANTITATIVE ATTRIBUTES ACTUAL (PER SCENARIO; SEE SECTION D.5.d) ATTRIBUTES

1. Total malfunctions (5-8) 6 Events 2, 3, 4, 5, 6, 7 2. Malfunctions after EOP entry (1-2) 1 Event 7 3. Abnormal events (2-4) 2 Events 4, 5 4. Major transients (1-2) 1 EventS 5. EOPs entered/requiring substantive actions (1-2) 2 Event6,8 6. EOP contingencies requiring substantive actions (0-2) 2 EventS 7. Critical tasks (2-3) 2 CRITICAL TASK DESCRIPTIONS:

CRITICAL TASK JUSTIFICATION:

CT -1.0: Given a failure of the Division 2 EDG to start, the crew will take This task is identified as critical because without action to manually start the Division 2 EDG lAW N2-SOP-03 operator action to manually start the Division 2 EDG, the station would be in Station Blackout conditions.

CT-2.0: Given RPV water level at or below the TAF but above the This task is identified as critical because without MSCWL, the crew will open 7 ADS valves lAW N2-EOP-C2 operator action to open the 7 ADS valves, RPV water level would continue to lower until the fuel is no longer adequately cooled. NRC Scenario 1 March 2014 SCENARIO

SUMMARY

The plant is operating at 90% power with 2WCS-P113 out of service for maintenance.

The crew will take the shift and raise reactor power to 95% using recirculation flow. After the reactivity maneuver, Service Water Pump 1A will trip on motor electric fault. The crew will take action to start a standby service water pump per N2-0P-11.

When starting the standby pump, the discharge valve will fail to automatically open requiring the crew to manually open the valve. Once the standby service water pump is started, HPCS inadvertently initiates and injects into the RPV. FWLC will respond and automatically maintain RPV level below the Level 8 setpoint.

The crew will attempt to reset HPCS and place it back in standby per N2-0P-33, however the HPCS malfunction will prevent the system from being returned to standby. The crew will be required to place HPCS in pull to lock (PTL). After the HPCS pump is in PTL, suction strainer clogging will cause the running Control Rod Drive Pump to trip. The crew will take action per N2-SOP-30 and swap suction strainers.

Once the suction strainers have been swapped, the crew will restart the Rod Drive pump. Following the restoration of the Control Rod Drive Pump, a loss of both Line 5 and 6 will occur. An electrical fault will cause a complete loss of the Division 1 electrical switchgear.

A fault on the Division 2 EDG will prevent it from automatically starting.

The crew will take actions per N2-SOP-03 and manually start the Division 2 EDG and power the Division 2 electrical switchgear (CRITICAL TASK). The loss of Division 1 switchgear will require the crew to manually scram the reactor, trip the turbine, and trip both recirculation pumps. The crew will enter N2-EOP-RPV and begin actions to stabilize RPV pressure and level. After the reactor is scrammed, a small LOCA will occur. The LOCA will cause RPV level to lower. The crew will attempt to maintain level using HPCS, however once HPCS is taken out of PTL, it will trip on motor electric fault. The crew will attempt to start RCIC for level control, however it will fail to automatically start using the initiation pushbutton.

Once the crew manually starts up RCIC, it will trip on a failed pressure transmitter.

The crew will manually start the Standby Liquid Control system (SLS). When RPV level reaches the TAF, the CRS will enter N2-EOP-C2 and direct all 7 ADS valves be opened. The crew will open the ADS valves and blowdown the reactor (CRITICAL TASK). As RPV pressure lowers, the low pressure ECCS systems will begin to inject and recover RPV level. The crew will control the low pressure ECCS systems to raise RPV level back to the normal band. The scenario will be terminated when RPV level has been recovered sufficiently.

Termination Criteria:

RPV water level has been recovered to above the top of active fuel. Major Procedures N2-SOP-30, N2-SOP-03, N2-SOP-11, N2-EOP-RPV, Exercised:

PC, N2-EOP-C2, Mitigation Strategy:

RL 2-Small break LOCA or loss of high pressure injection, RPV level cannot be maintained above the top of active fuel, RPV Slowdown, recover level above TAF with low pressure systems and I or alternate coolant injection systems. NRC Scenario 1 March 2014 Appendix D Facility:

Nine Mile Point 2 Examiners:

Initial Conditions:

Simulator IC-151 Scenario Outline Scenario No.: NRC-2 Operators:

Form ES-D-1 Op-Test No: March 2014 1. Reactor power is -65% in the process of shutting down 2. The shutdown is on hold awaiting the results of surveillance results 3. Instrument Air Compressor C is out of service due to maintenance.

Turnover:

1. lAS Compressor C is out of service due to maintenance.

2. The Reactor shutdown will recommence following surveillance testing. Event No. Malf. No. Event ED01 3 4 RP06B NRC Scenario 2 Type* R (RO) R (SRO) C (BOP) C (SRO) TS (SRO) Event Desc A malfunction in the normal station service transformer cooling system causes transformer temperatures to rise. The crew will coordinate with a Plant Operator at the transformer and lower power using N2-SOP-1 01 D to stabilize transformer temperatures.

The RPS B Motor Generator will trip causing a half scram on RPS B side. The crew will enter N2-SOP-97 and align the RPS B solenoids to their alternate power supply. While resetting the EPA's per N2-SOP-97, the Plant Operator will report that the undervoltage trip relay had to be bypassed in order to reset one of the EPA's. The CRS will declare the associated EPA inoperable and evaluate TS 3.3.8.3. March 2014 5 6 7 8 9 CW09 C (BOP) CW26 C (SRO) FW03A M (All) FW03B RD12 C (RO) C (SRO) RR20 C (RO/BOP) RH14A C (SRO) RHS10B C (RO/BOP) C (SRO) A clogging of the Service Water Traveling Screens will cause service water intake bay level to lower. The crew will take action per N2-SOP-11 and attempt to clean the traveling screens. Intake bay will continue to lower to 234 feet. The intake bay bypass valves 2SWP*MOV77 AlB will fail to automatically open requiring the crew to take manual action to open the valves (CRITICAL TASK). Once MOV77A and Bare open, intake bay level will recover. N2-SOP-11 A loss of all feed pumps will occur. The crew will place the mode switch in shutdown and begin taking action to stabilize RPV level and pressure.

N2-SOP-06 RPV level control will be complicated by malfunctions in the RDS and Feedwater systems which will force the crew to use HPCS, RCIC, and/or FW Booster Pumps to maintain RPV level. N2-EOP-RPV, N2-SOP-101C A LOCA will occur. The Division 1 ECCS system fails to automatically initiate.

The crew will manually initiate Division 1 ECCS system, however RHS Pump A will not start due to a broken control switch. Primary containment parameters will continue to degrade. Suppression chamber sprays will be initiated.

Once suppression chamber pressure reaches 10 psig, the crew will attempt to spray the drywell using RHS B. While the crew is attempting to align drywell sprays, 2RHS*MOV25B will stick shut. Plant Operators will be dispatched in an attempt to manually open MOV25B. While the POs are attempting to manually open MOV25B, primary containment parameters continue to degrade. The CRS will determine that Suppression Chamber Pressure cannot be restored and maintained within the Pressure Suppression Limit and will enter N2-EOP-C2 and direct 7 ADS valve be opened. The crew will open 7 ADS valves and blowdown the reactor (CRITICAL TASK). The scenario may be terminated once 7 ADS valves are opene!d. * (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor NRC Scenario 2 March 2014 Facility:

Nine Mile Point 2 Scenario No: NRC-2 Op-Test No: March 2014 TARGET QUANTITATIVE ATTRIBUTES ACTUAL (PER SCENARIO; SEE SECTION 0.5.d) ATTRIBUTES

1. Total malfunctions (5-8) 8 Events 1, 2, 3, 4, 5, 6, 7, 8 2. Malfunctions after EOP entry (1-2) 2 Event7,8 3. Abnormal events (2-4) 2 Events 4, 5 4. Major transients (1-2) 1 EventS 5. EOPs entered/requiring substantive actions (1-2) 2 Events 7, 9 6. EOP contingencies requiring substantive actions (0-2) 1 Event9 7. Critical tasks (2-3) 2 CRITICAL TASK DESCRIPTIONS:

CRITICAL TASK JUSTIFICATION:

CT-1.0: Given service water intake bay level less than 234ft and a This task is identified as critical because without failure of 2SWP*MOV77A

& 776 to automatically open, the crew will operator action, the plant will lose its ultimate take action to manually open 2SWP*MOV77 A & 77B per N2-SOP-11.

heat sink. CT-2.0: Given Suppression Chamber Pressure unable to be restored This task is identified as critical because without and maintained within the Pressure Suppression Limit, the crew will operator action, the primary containment open 7 SRV's JAW N2-EOP-C2.

pressure suppression function would continue to degrade and would not be able to accept a full blowdown of the reactor. NRC Scenario 2 March 2014 SCENARIO

SUMMARY

The crew will take the shift at -65% power. The shutdown is on hold pending surveillance results. Instrument Air Compressor Cis out of service due to maintenance.

After the crew takes the shift, a malfunction in the normal station service transformer cooling system causes transformer temperatures to rise. The crew will dispatch a PO to investigate.

The PO will report that several of the transformer cooling fans are not running and cannot be started. The crew will coordinate with the PO and lower power per N2-SOP-1 01 D to stabilize transformer temperatures.

Once temperatures are under control, APRM 2 will fail upscale. The crew will take action per the ARP's and N2-0P-92 to bypass APRM 2. The CRS will evaluate TS 3.3.1.1 for the inoperable APRM. Once APRM 2 is bypassed, the running Gland Seal Exhaust Fan will trip on motor electric fault. The crew will take action per the ARP's and N2-0P-25 to isolate the tripped Gland Seal Exhaust Fan and start a standby fan. After the Gland Seal Exhaust Fan is started, the RPS B Motor Generator will trip causing a half scram on RPS B side. The crew will enter N2-SOP-97 and align the RPS B solenoids to their alternate power supply. While resetting the EPA's per N2-SOP-97, the Plant Operator will report that the undervoltage trip relay had to be bypassed in order to reset one of the EPA's. The CRS will declare the associated EPA inoperable and evaluate TS 3.3.8.3. After the RPS B solenoids are powered from their alternate power supply, Service Water intake clogging will occur causing Service Water intake bay level to lower. The crew will take action per N2-SOP-11 and attempt to clean the traveling screens. Intake bay will continue to lower to 234 feet. The intake bay bypass valves 2SWP*MOV77 AlB will fail to automatically open requiring the crew to take manual action to open the valves (CRITICAL TASK). Once MOV77A and Bare open, intake bay level will recover. Once Service Water intake bay level is restored, a loss of all feed pumps will occur. The loss will require the crew to place the Mode Switch in shutdown.

Once RPV level has been stabilized using alternate level control systems, a LOCA will occur. The LOCA will cause Primary Containment (PC) parameters to degrade and the crew will enter N2-EOP-PC to stabilize PC parameters.

Malfunctions in the Division 1 RHS systems will prevent RHS A from being used for primary containment control and the crew will be required to use RHS B to spray the suppression chamber. As PC conditions continue to degrade, the crew will attempt to spray the drywell using RHS B. While the crew is attempting to align drywell sprays, 2RHS*MOV25B (Drywell Spray Valve) will stick shut. Plant Operators will be dispatched in an attempt to manually open MOV25B. While the PO's are attempting to manually open MOV25B, primary containment parameters will continue to degrade. The CRS will determine that Suppression Chamber Pressure cannot be restored and maintained within the Pressure Suppression Limit and will enter N2-EOP-C2 and direct 7 ADS valve be opened. The crew will open 7 SRV's and blowdown the reactor (CRITICAL TASK). The scenario may be terminated once 7 SRV's are opened. Termination Criteria:

7 SRV's are open Major Procedures N2-SOP-97, N2-SOP-11, N2-EOP-RPV, N2-EOP-PC, Exercised:

C2 Mitigation Strategy:

PC 4, High containment pressure approaching PCPL, exceeds PSP, RPV Slowdown required NRC Scenario 2 March 2014 Appendix D Scenario Outline Form ES-D-1 Facility:

Nine Mile Point 2 Scenario No.: NRC-4 Operators:

Examiners:

________ _ Op-Test No: March 2014 Initial Conditions:

Simulator IC-153 1. Reactor power is -3% 2. Reactor startup is in progress Turnover:

1. Continue reactor startup and raise power to 1 0% Event No. Malf. No. Event 2 RD07 3 PC10A 4 PC01 5 IA02A IA04A IA04B 6 RH13A NRC Scenario 4 Type* R (RO) R (SRO) C (RO) C (SRO) TS (SRO) C (BOP) C (SRO) C (BOP) C (SRO) C (BOP) C (SRO) TS (SRO) Event Description The crew will assume the watch and continue the startup by withdrawing rods per N2-0P-101A.

While withdrawing rods, one of the control rods will stick. The crew will respond per N2-0P-30 and raise drive water pressure.

Once drive water pressure is raised, the rod will become unstuck. N2-0P-30 One pair of Suppression Chamber to Drywell Vacuum Breakers will fail open. There are no operator actions; however the CRS will evaluate TS 3.6.1.7. TS 3.6.1.7 A fault will cause a loss of all drywell cooling. The crew will respond per N2-SOP-60 and restart the drywell cooling fans in "Fan Only" mode!. N2-SOP-60 A fault will occur on Instrument Air Compressor A. Compressors Band C will fail to auto start. The crew will take action per N2-SOP-19 and manually start either Compressor B or C (CRITICAL TASK) to restore air header pressure.

N2-SOP-19 An electrical failure causes a spurious initiation of Division 1 ECCS systems. The crew will take action to shutdown the Division 1 ECCS systems. The CRS will evaluate TS 3.5.1 and 3.8.1. N2-0P-31, TS 3.5.1, TS 3.8.1 March 2014 7 9 MT01 FW08 RC12 C (RO) C (SRO) Overrides C (BOP) C (SRO) A Seismic Event will occur. The event will cause a FWLC failure. The crew will take manual control of FWLC. N2-SOP-90 A RCIC Steam Leak will require the crew to manually scram the reactor (CRITICAL TASK). The crew will take action to stabilize RPV level and pressure.

N2-EOP-RPV, N2-EOP-SC Secondary Containment conditions will degrade requiring the crew to either anticipate RPV blowdown per N2-EOP-RPV, or perform a blowdown per N2-EOP-C2 (CRITICAL TASK). The scenario may be terminat,ed when the RPV is being depressurized.

N2-EOP-SC, N2-EOP-C2

• (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor NRC Scenario 4 March 2014 Facility:

Nine Mile Point 2 Scenario No: NRC-4 Op-Test No: March 2014 TARGET QUANTITATIVE ATTRIBUTES ACTUAL (PER SCENARIO; SEE SECTION D.5.d) ATTRIBUTES

1. Total malfunctions (5-8) i' Events 2, 4, 5, 6, 7, 8, S 2. Malfunctions after EOP entry (1-2) 1 EventS 3. Abnormal events (2-4) Events 4, 5, 7 4. Major transients (1-2) 1 EventS 5. EOPs entered/requiring substantive actions (1-2) ,, .. Events 8, S 6. EOP contingencies requiring substantive actions (0-2) 1 EventS 7. Critical tasks (2-3) <I CRITICAL TASK DESCRIPTIONS:

CRITICAL TASK JUSTIFICATION:

CT-1.0: Given a trip ofthe running instrument air compressor and a This task is identified as critical because without failure of the lag and backup air compressors to automatically start, the operator action to start the lag or backup air crew will take action to manually start the lag or backup air compressor, instrument air header pressure will compressor. degrade until the reactor scrams due to low RPV level and/or loss of scram air header pressure.

CT-2.0, Given secondary containment temperatures approaching a This task is identified as critical because without maximum safe value in one area, the crew will initiate a manual r<eactor operator action to scram, the reactor will scram lAW N2-EOP-RPV continue to provide energy to the RCIC steam line break and cause increased secondary containment temperatures and radiation levels. CT-3.0A Given secondary containment temperatures approaching or This task is identified as critical because without above maximum safe values in one area, the crew will open 5 main operator action to depressurize the reactor, turbine bypass valves lAW N2-EOP-RPV secondary containment integrity, the integrity of equipment located in the secondary containment, and continued safe operation of the plant cannot be assured. Note: The crew may choose to wait until two or more areas are above maximum safe values before depressurizing the reactor. If the crew chooses to depressurize the reactor via the SRVs, then CT-2.0A does not have to be evaluated.

CT-3.08 Given secondary containment temperatures above maximum This task is identified as critical because without safe values in two areas, the crew will open 7 ADS valves lAW N2-EOP-operator action to depressurize the reactor, C2 secondary containment integrity, the integrity of equipment located in the secondary containment, and continued safe operation of the plant cannot be assured. Note: The crew may choose to "anticipate blowdown" and depressurize the reactor to the main condenser.

If the crew chooses to depressurize the reactor to the main condenser and are successful in preventing two areas from exceeding the maximum safe temperatures, then CT-2.08 does not have to be evaluated.

NRC Scenario 4 March 2014 SCENARIO

SUMMARY

The crew will take the shift at -3% power. The RO will raise power using rods. While withdrawing rods, a control rod will stick. The crew will take action to raise drive water pressure per N2-0P-30.

Raising drive water pressure will free the stuck rod and allow the startup to continue.

After power has been sufficiently raised, an instrument failure will cause one pair of Suppression Chamber to Drywell Vacuum Breakers to fail open. There are no operator actions; however the CRS will evaluate TS 3.6.1.7. Once TS 3.6.1.7 has been evaluated, an electrical fault will cause a loss of all drywell cooling. The crew will respond per N2-SOP-60 and restart the drywell cooling fans in "Fan Only" mode. After Drywell Cooling has been restored, an electrical fault will occur on Instrument Air Compressor A. Compressors B and C will fail to auto start. The crew will take action per N2-SOP-19 and manually start either Compressor B or C (CRITICAL TASK) to restore air header pressure.

After the loss of instrument air, an electrical failure will cause a spurious initiation of Division 1 ECCS systems. The crew will take action to shutdown the Division 1 ECCS systems. The CRS will evaluate TS 3.5.1 and 3.8.1. Following the inadvertent initiation of Division 1 ECCS systems, a seismic event occurs. The event will cause an unisolable RCIC steam leak and a FWLC failure. The crew will take action per N2-EOP-SC and enter N2-EOP-RPV to manually scram the reactor (CRITICAL TASK). RPV level control will be complicated by the FWLC failure. Due to the RCIC steam leak, Secondary Containment conditions will continue to degrade requiring the crew to either anticipate RPV blowdown per N2-EOP-RPV, or perform a blowdown per N2-EOP-C2 (CRITICAL TASK). The scenario may be terminated when the RPV is being depressurized.

Termination Criteria:

RPV pressure lowering due to anticipating blowdown or actually blowing down the reactor Major Procedures N2-SOP-60, N2-SOP-19, N2-SOP-90, N2-EOP-RPV, Exercised:

SC, N2-SOP-C2.

Mitigation Strategy:

SC1-Secondary containment leak. Slowdown Required.

NRC Scenario 4 March 2014