ML103500252

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Nine Mile Point, Unit 1 - Final Outlines (Folder 3)
ML103500252
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Site: Nine Mile Point Constellation icon.png
Issue date: 10/27/2010
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Operations Branch I
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ES-401 Written Examination Outline Form ES-401-1 Facility:

Nine Mile Point Unit 1 Date of Exam: November 2010 RO KIA Category Po i nts SRO-Only Points Tier Group K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4 G

  • Total A2 G* Total 1. Emergency

& Plant Evaluations 1 2 Tier Totals 3 1 4 3 1 4 3 1 4 4 1 5 4 2 6 3 1 4 20 7 27 3 2 5 4 1 5 7 3 10 1 2 2 2 2 2 2 3 3 3 3 2 26 2 3 5 2. Plant Systems 2 Tier Totals 1 3 2 4 1 3 1 3 1 3 1 3 1 4 1 4 1 4 1 4 1 3 12 38 0 3 1 2 5 3 8 3. Generic Knowledge

&Abilities 1 2 3 4 10 1 2 3 4 7 2 2 3 3 2 2 1 2 Note 1. Ensure that at least two topics from every applicable KIA category are sampled within each tier of the RO and SRO-on l y 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 ident i fied on the assoc i ated outl i ne; systems or evolutions that do not apply at the facility should be deleted and j u stified; 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 rega r ding elimination of i nappropriate KIA statements.

4. Select topics from as many systems and evolutions as possible; samp l e every system or evolution in the group be f ore 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 ratings for the RO and SRO-only portions , respectively. 6. Select SRO to p ics for Tiers 1 and 2 from the shaded systems and KIA categories. 7.* The generic (G) KlAs 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 B. On the following pages , enter the KIA numbers , a brief description of each topic , the topics' importance ratings (IR) for the license level , and the point tota l s (#) 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 Tie r 3 , select topics from Section 2 of the KIA Cata l og , and enter the KIA numbers , descriptions , IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to KlAs that are linked to 10CFR55.43 Form ES-401-1 Nine Mile Point Unit Written Examination Emergency and Abnormal Plant Evolutions

-Tier 1 Group EAPE#/Name Safety KJA Topic (s) EA2.02 -Ability to determine and/or interpret the following 295024 High Drywell X as they apply to HIGH 4.0 76 Pressure 15 DRYWELL PRESSURE: Drywell temperature AA2.04 -Ability to determine and/or interpret the following 295004 Partial or as they apply to PARTIAL 3.3 77 Loss of DC Pwr I OR COMPLETE LOSS OF D.C. POWER: System lineups AA2.02 -Ability to determine and/or interpret the following 295001 Partial or Complete as they apply to PARTIAL Loss of Forced Core Flow X OR COMPLETE LOSS OF 3.2 78 Circulation I 1 & 4 FORCED CORE FLOW CIRCULATION:

Neutron monitoring 295038 High Off-site 2.4.18, Knowledge of the4.0 79 Release Rate I specific bases for EOPs. 2.2.38 -Equipment Control: 295026 Suppression Pool Knowledge of conditions and 4.5 80 High Water Temp. limitations in the facility license. 2.2.39 -Equipment Control: 295037 SCRAM Conditions Knowledge of less than or Present and Reactor Power X equal to one hour technical 4.5 81 Above APRM Downscale or specification action Unknown 11 statements for systems. 2.2.37 -Equipment Control: Ability to determine 295005 Main Turbine X operability and I or 4.6 82 Generator Trip I 3 availability of safety related equipment.

EK1.03 -Knowledge of the operational implications of the following concepts as 295030 Low Suppression they apply to LOW 3.8 39 Pool Water Levell 5 SUPPRESSION POOL WATER LEVEL: Heat capacity ES-401 Form ES-401-1 Nine Mile Point Unit Written Examination Emergency and Abnormal Plant Evolutions

-Tier 1 Group e Safety Function 295024 High Drywell Pressure /5 x 295005 Main Turbine Generator Trip / 3 x 295028 High Drywell Temperature

/5 295006 SCRAM / 1 295025 High Reactor Pressure /3 x x x 700000 Generator Voltage "pC":",," c';'''c

<", and Electric Grid x Disturbances 295004 Partial or Total x "/"""c ;'c'.c.'Loss of DC Pwr / 6

.. l KIA Topic(s) EK1.01 -Knowledge of the operational implications of the following concepts as they apply to HIGH DRYWELL PRESSURE:

Drywell integrity: Plant-S ecific AK 1 .03 -Knowledge of the operational implications of the following concepts as they apply to MAl N TURBINE GENERATOR TRIP: Pressure effects on reactor level EK2.04 -Knowledge of the interrelations between HIGH DRYWELL TEMPERATURE and the following:

Drywell ventilation AK2.06 -Knowledge of the interrelations between SCRAM and the following:

Reactor ower EK2.01 -Knowledge of the interrelations between HIGH REACTOR PRESSURE and the followin : RPS AK3.02 -Knowledge of the reasons for the following responses as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES:

Actions contained in abnormal operating procedure for voltage and grid disturbances.

AK3.02, Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER: Ground isolation/fault determination.

4.1 40 3.5 41 3.6 42 4.2 43 3.6 45 2.9 46 ES-401 Form ES-401-1 Nine Mile Point Unit Written Examination Emergency and Abnormal Plant Evolutions

-Tier 1 Group EAPE#lName Safety Function I K1 I K2 I K3 I A1 295016 Control Room x iAbandonment / 7 295031 Reactor Low Water Level/2 295037 SCRAM Conditions Present and Reactor Power Above APRM Downscale or Unknown /1 295021 Loss of Shutdown .Cooling /4 295003 Partial or Complete Loss of AC 16 295019 Partial or Total *Loss of Inst. Air /8 295026 Suppression Pool High Water Temp. /5 I A2 I G KIA Topic(s) AK3.03 -Knowledge of the reasons for the following responses as they apply to

  • 3.5 47 CONTROL ROOM ABANDONMENT:

Disabling control room controls EA 1.10 -Ability to operate and/or monitor the following as they apply to REACTOR *3.6 148 LOW WATER LEVEL: Control rod drive EA 1.10 -Ability to operate and/or monitor the following as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER 3.7 149 ABOVE APRM DOWNSCALE OR i UNKNOWN: Alternate boron* injection methods: Specific AA 1.02 -Ability to operate and/or monitor the following as they apply to LOSS OF 3.5

  • 50 SHUTDOWN COOLING: RHR/shutdown coolin AA2.04 -Ability to determine*

and/or interpret the following i as they apply to PARTIAL 3.5 51 OR COMPLETE LOSS OF A.C. POWER: System lineu s AA2.01 -Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF 3.5 52 i INSTRUMENT AIR: Instrument air system ressure EA2.03 -Ability to determine and/or interpret the following as they apply to SUPPRESSION POOL

  • 3.9 53 HIGH WATER TEMPERATURE:

Reactor ES-401 Form ES-401-1 Nine Mile Point Unit Written Examination Emergency and Abnormal Plant Evolutions

-Tier 1 Group EAPE#lName Safety Function I K1 I K2 I K3 I A1 I A2 I G KIA Topic(s) 2.4.21 -Emergency Procedures I Plan: Knowledge of the parameters and logic used to assess the status of safety 295023 Refueling functions, such as reactivity Accidents I 8 control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc. 2.4.46 -Emergency 295018 Partial or Total Loss of CCW I 8 Procedures I Plan: Ability to verify that the alarms are consistent with the plant conditions.

2.1.27 -Conduct of 295038 High Off-site Operations:

Knowledge of

  • Release Rate I 9 system purpose and I or function.

AA 1 .09 -Ability to operate and / or monitor the following 1600000 Plant Fire On-site I *8 as they apply to PLANT FIRE ON SITE: Plant fire zone panel (including detector location AA2.06 -Ability to determine and/or interpret the following 295001 Partial or Complete as they apply to PARTIAL Loss of Forced Core Flow OR COMPLETE LOSS OF Circulation f 1 & 4 FORCED CORE FLOW CIRCULATION:

Nuclear boiler instrumentation KIA CategoryTotals 3 Group Point Total: I Imp. I Q# i 4.0 54 4.2 55. 3.9

  • 56 2.5 571 3.2 58. 2017 Form ES-401-1 Nine Mile Point Unit Written Examination Emergency and Abnormal Plant Evolutions

-Tier 1 Group EAPE#/Name Safety KIA Topic (s) AA2.03 -Ability to determine and/or interpret the following 295020 Inadvertent Cont. as they apply to 3.7 83 Isolation I 5 & INADVERTENT CONTAINMENT ISOLATION:

Reactor power 295007 High 2.4.6, Knowledge of EOP4.7 84 Pressure mitiQation strateQies.

AA2.03 -Ability to determine and/or interpret the following 295014 as they apply to 4.3 85 Reactivity Addition I INADVERTENT REACTIVITY ADDITION:

Cause of reactivity addition AK1.02 -Knowledge of the operational implications of the following concepts as 295017 High Off-site they apply to HIGH OFF-3.8 59 Release Rate I 9 SITE RELEASE RATE: Protection of the general public AK2.01 -Knowledge of the interrelations between LOW 295009 Low Reactor Water REACTOR WATER LEVEL 3.9 60 Level 12 and the following:

Reactor water level indication EK3.01 -Knowledge of the reasons for the following responses as they apply to 295032 High Secondary HIGH SECONDARY Containment Area 3.5 61 CONTAINMENT AREA Temperature I 5 TEMPERATURE

Emergency/normal depressurization EA 1.04 -Ability to operate and/or monitor the following as they apply to 295036 Secondary SECONDARY Containment High 3.1 62 CONTAINMENT HIGH Sump/Area Water Levell 5 SUMP/AREA WATER LEVEL: Radiation monitoring:

Plant-Specific ES-401 Form ES-401-1 Nine Mile Point Unit Written Examination Emergency and Abnormal Plant Evolutions

-Tier 1 Group EAPE#/Name Safety Function I K1 I K2 I K3 I A1 I A2 I G I KIA Topic(s) AA2.01 -Ability to determine and/or interpret the following 295012 High Drywell as they apply to HIGH Temperature 15 DRYWELL TEMPERATURE:

Drywell tern erature 2.4.1 -Emergency 295029 High Suppression Pool Water Levell 5 Procedures I Plan: Knowleqge of EOP entry conditions and immediate action ste s. AA2.01 -Ability to determine and/or interpret the following i 295002 Loss of Main as they apply to LOSS OF Condenser Vac I 3 MAIN CONDENSER VACUUM: Condenser vacuum/absolute ressure KIA CategoryTotals Group Point Total: I Imp. I Q# I 3.8 63 4.6 64 2.9 65 7/3 ES-401 Form ES-401-1 System #/Name 259002 Reactor Water Level Control 211000 SLC 205000 Shutdown Cooling 264000 EDGs 207000 Isolation (Emergency)

Condenser Nine Mile Point Unit 1 Written Examination Outline Plant Systems -Tier 2 Group 1 KIA Topic(s) A2.07 -Ability to (a) predict the impacts of the following on the REACTOR WATER LEVEL CONTROL SYSTEM; and (b) based X on those predictions, use 2.5 86 procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Loss of comparator bias signal A2.07 -Ability to (a) predict the impacts of the following on the STANDBY LIQUID CONTROL SYSTEM; and (b) based on those X predictions, use 3.2 87 procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Valve closures 2.2.12 -Equipment X Control: Knowledge of 4.1 88 surveillance procedures.

2.1.32 -Conduct of Operations:

Ability to X explain and apply all 4.0 89 system limits and precautions.

2.2.25 -Equipment Control: Knowledge of X bases in technical specifications for limiting 4.2 90 conditions for operations and safety limits.

Form ES-401-1 System #/Name 262001 AC Electrical Distribution x 212000 RPS x 300000 Instrument Air 206000 HPCI x x 239002 SRVs x 262002 UPS (AC/DC) x Nine Mile Point Unit Written Examination Plant Systems -Tier 2 Group KiA Topic(s) K 1.02 -Knowledge of the physical connections and/or cause-effect relationships between AC. ELECTRICAL DISTRIBUTION and the following:

D.C. electrical distribution K1.02 -Knowledge of the physical connections and/or cause-effect relationships between REACTOR PROTECTION SYSTEM and the following: . Nuclear boiler instrumentation , .. K2.02 -Knowledge of electrical power supplies to the following:

Emergency air com ressor K2.01 -Knowledge of electrical power supplies to the following:

System valves: BWR-2,3,4 K3.03 -Knowledge of the effect that a loss or malfunction of the RELI EF/SAFETY VALVES will have on following:

Ability to rapidly depressurize the reactor K3.08 -Knowledge of the effect that a loss or malfunction of the UNINTERRUPTABLE POWER SUPPLY (AC.lD.C.)

will have on following:

Computer operation:

Plant-S ecific 3.3 1 3.7 2 3.0 3 3.2 4 4.3 5 2.7 6 Form ES-401-1 Nine Mile Point Unit Written Examination Plant Systems -Tier 2 Group System KiA Topic(s) I> K4.09 -Knowledge of { REACTOR WATER I: LEVEL CONTROL SYSTEM design 259002 Reactor :..*.* feature(s) and/or x 7*Water Level '. interlocks which provide for the following:

Single element control (reactor water level provides the .A only

....: K4.01 -Knowledge of CCWS design feature(s) 400000 Component and or interlocks which x 8 Cooling Water provide for the following:

I::. . Automatic start of [;:!:;:;" standby pump K5.01 -Knowledge of : the operational

.* implications of the following concepts as 215005 APRM I ;. they apply to AVERAGE 2.8 9POWER RANGE MONITOR/LOCAL POWER RANGE . ....... MONITOR SYSTEM: Ii; '::. LPRM detector operation

K5.03 -Knowledge of

'. the operational implications of the 207000 Isolation following concepts as (Emergency) x they apply to 2.7 10 Condenser ISOLATION (EMERGENCY)

.:: transfer:

BWR-2,3 ';j:; K6.02 -Knowledge of .* the effect that a loss or malfunction of the following will have on the 205000 Shutdown SHUTDOWN COOLING 2.7 11 Cooling . { SYSTEM (RHR I.*** ;'. SHUTDOWN COOLING I'e. MODE): D.C. electrical

....* ! I:' power Form ES-401-1 System #/Name 215004 Source Range Monitor .263000 DC *Electrical Distribution 261000 SGTS 209001 LPCS Nine Mile Point Unit Written Examination Plant Systems -Tier 2 Group KIA Topic(s) K6.01 -Knowledge of the effect that a loss or malfunction of the following will have on the SOURCE RANGE MONITOR (SRM) SYSTEIVI:RPS A1.01 -Ability to predict and/or monitor changes in parameters associated with operating the D.C. ELECTRICAL DISTRIBUTION controls including:

Battery charging/discharging rate A1.06 -Ability to predict and/or monitor changes in parameters associated with operating the STANDBY GAS TREATMENT SYSTEM controls including:

Drywell and suppression chamber differential ressure: Mark-I A2.06 -Ability to (a) predict the impacts of the following on the LOW PRESSURE CORE SPRAY SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Inadequate s stem flow 3.2 12 2.5 13 2.7 14 3.2 15 Form ES-401-1 Nine Mile Point Unit Written Examination Plant Systems -Tier 2 Group System #/Name 2150031RM 218000 ADS 223002 PCIS/Nuclear

  • Steam Supply I Shutoff 211000 SLC 264000 EDGs .' I ." X ii/J';;!l I,> .... " '. !/ X , ,i.., I .I'i" "; . < i.* ..*.*.. X ' .. I I " , . I' " I KIA Topic(s) A2.05 -Ability to (a) , " .' predict the impacts of the I following on the INTERMEDIATE li sIii ,.. RANGE MONITOR (IRM) SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those

,> abnormal conditions or :' YY, operations:

Faulty or '.' ,

c i*i" " I'" , .....* . " . I' '; \,'j' .' X 1'\;\'. Cd erratic operation of detectors/system A3.04 -Ability to monitor automatic operations of the AUTOMATIC DEPRESSURIZATION SYSTEM including:

Primary containment pressure A3.02 -Ability to monitor automatic operations of the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY OFF including:

Valve closures A4.05 -Ability to manually operate and/or monitor in the control room: Flow indication:

Plant-Specific

'V)..; A4.01 -Ability to manually operate and/or X I> ' monitor in the control room: Adjustment of le'D" '. exciter voltage 3.3 16 3.7 17 3.5 18 4.1 19 3.3 20

  • ******* Form ES-401-1 Nine Mile Point Unit Written Examination Plant Systems -Tier 2 Group System #/Name 218000 ADS 209001 LPCS 223002 PCIS/Nuclear Steam Supply Shutoff 300000 Instrument Air 262002 UPS (AC/DC) " , ',', " ' . ,

,',(,:.., ,. *'j k ' 'j ,. I :;L' i. 1**')

I', ," iii. '. .. ii' ': .: I*.***. .'<' KJA Topic(s) X 2.4.8 -Emergency Procedures

/ Plan: Knowledge of how .i" abnormal operating procedures are used in i conjunction with EOP's. 2.4.4 -Emergency Procedures

/ Plan: Ability to recognize abnormal .' indications for system J<:' operating parameters which are entry-level conditions for emergency and abnormal operating procedures . A2.06 -Ability to (a) predict the impacts of the following on the .. PRIMARY CONTAINMENT ISOLATION . SYSTEM/NUCLEAR STEAM SUPPLY , ," OFF; and (b) based on those predictions, use ,'. procedures to correct, control, or mitigate the consequences of those !',::; '. '.' abnormal conditions or operations:

Containment

.*..!E*; instrumentation failures ;./. A3.02 -Ability to monitor automatic operations of X 'j the INSTRUMENT AIR SYSTEM including:

Air temperature A4.01 -Ability to manually operate and/or monitor in the control Xf .; room: Transfer from ,'. alternative source to preferred source 3.8 21 4.5 22 3.0 23 2.9 24 2.8 25 ES-401 Form ES-401-1 Nine Mile Point Unit 1 Written Examination Outline Plant Systems -Tier 2 Group 1 System #/Name KIA Topic(s) 2150031RM A 1.05 -Ability to predict and/or monitor changes in parameters associated with operating the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM controls including:

SCRAM and rod block tri set oints 3.9 26 KIA Category Totals ES-401 Form ES-401-1 Nine Mile Poin t Unit 1 Written Examination Outline Plant Systems -Tier 2 Group 2 System #/Name KJA Topic(s) A2.08 -Ability to (a) predict the impacts of the following on the FIRE PROTECTION SYSTEM; and (b) based 286000 Fire Protection X on those predictions , use procedures to correct, 3.3 91 control , or mitigate the consequences of those abnormal conditions or operations

Failure to actuate when required 2.4.31 -Emergency 201003 Control Rod and Drive Mechanism X Procedures

/ Plan: Knowledge of annunciator alarms , indications , or response 4.1 92 procedures.

2.1.32 -Conduct of 234000 Fuel Handling Equipment X Operations

Ability to explain and apply all system limits and 4.0 93 precautions. K 1 .05 -Knowledge of the physical connections and/or cause-effect 259001 Reactor Feedwater X relationships between REACTOR 3.2 27 FEEDWATER SYSTEM and the following:

Condensate system 219000 RHR/LPCI : K2.01 -Knowledge of Torus/Pool Cooling X electrical power supplies 2.5 28 Mode to the following:

Valves K3.01 -Knowledge of the effect that a loss or 271000 Off-gas X malfunction of the OFFGAS SYSTEM will 3.5 29 have on following:

Condenser vacuum Form ES-401-1 System #/Name 272000 Radiation Monitoring 288000 Plant Ventilation 201002 RMCS 204000 RWCU 201003 Control Rod and Drive Mechanism Nine Mile Point Unit Written Examination Plant Systems -Tier 2 Group KIA Topic(s)

  • i K4.03 -Knowledge of RADIATION .L MONITORING System I design feature(s) and/or interlocks which provide 3.6 30 ....... for the following:

Fail safe tripping of process ". *.. radiation monitoring logic . during conditions of .........

>:' , instrument failure K5.01 -Knowledge of p y " . the operational implications of the I'i',. following concepts as X 3.1 31;; they apply to PLANT , VENTILATION SYSTEMS: Airborne contamination control ?c:<

K6.01 -Knowledge of the effect that a loss or

." malfunction of the I',.;X /:>< following will have on the. 2.5 32 REACTOR MANUAL ..... CONTROL SYSTEM: ." .. / .**;N./* Select matrix power . ,,;i*,." A 1.03 -Ability to predict .. ' and/or monitor changes in parameters associated . with operating the X REACTOR WATER 2.8 33

_ CLEANUP SYSTEM controls including:

Reactor water ',. temperature A2.04 -Ability to predict .... and/or monitor changes .'.' in parameters associated with operating the 3.5 34 CONTROL ROD AND '.,.: DRIVE MECHANISM . , controls including:

Single : control rod SCRAM *

  • ES-401 Form ES-401-1 System #/Name 223001 CTMT and 1201006 RWM 214000 RPIS 256000 Reactor Condensate KIA Category Totals Nine Mile Point Unit Written Examination Plant Systems -Tier 2 Group KIA Topic(s) A3.05 -Ability to monitor automatic operations of the PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES including:

o ell ressure A4.04 -Ability to manually operate and/or monitor in the control room: Rod withdrawal error indication: S ec Not-BWR6) 2.2.44 -Equipment Control: Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

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

System Group Point Total: 4.3 35 i 3.3

  • 36 4.2 37 2.7 38 12/3 ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility:

Nine Mile Point Unit 1 Date: November 2010 Category KA# Topic RO SRO-Only IR Q# IR Q# 2.1 .34 Knowledge of primary and secondary plant chemistry limits. 2.7 66 2.1.8 Ability to coordinate personnel activities outside the control room. 3.4 67 1. Conduct of Operations 2.1.40 Knowledge of refueling administrative requirements 3.9 94 2.1 .13 Knowledge of facility requirements for controlling vital/controlled access. 3.2 98 Subtotal 2 2 2.2.13 Knowledge of tagging and clearance procedures.

4.1 68 2.2.20 Knowledge of the process for managing troubleshooting activities.

2.6 69 2. Equipment Ability to recognize system parameters Control 2.2.42 that are entry-level conditions for 4.6 95 Technical Specifications. 2.2.40 Ability to apply technical specifications for a system. 4.7 100 Subtotal 2 2 Ability to use radiation monitoring systems, such as fixed radiation 2.3.5 monitors and alarms, portable survey 2.9 70 instruments, personnel monitoring equipment , etc. Knowledge of radiation or 2.3.14 contam i nation hazards that may arise during normal , abnormal , or emergency 3.4 71 3. Radiation conditions or activities.

Contro l 2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions. 3.2 75 2.3.11 Ability to control radiation releases.

4.3 96 Subtotal 3 1 ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401 -3 2.4.45 2.4.34 4. Emergency Procedures I Plan 2.4.26 2.4.17 Subtotal Tier 3 Point Total: Knowledge of the emergency action level thresholds and classifications. Ability to prioritize and i nterpret the signif i cance of each annunciator o r alarm. Knowledge of RO tasks performed outside the main control room during an emergency and the resultant opera t ional effects. 2.9 4.1 4.2 72 73 74 Knowledge of facility protection requirements, including fire brigade and portable fire fighting equipment usage. Knowledge of EOP terms and definitions. 3.6 4.3 97 99 3 10 2 7 ES-401 Record of Rejected KIA's Form ES-401-4 Randomly Selected Tier 1 Group Reason for Rejection KA Question 76, Ability to determine and/or interpret the following as they apply to HIGH DRYWELL PRESSURE: Containment radiation levels: Mark-III.

Nine Mile Point Unit 1 has a Mark-I containment, not a Mark-III containment.

1 11 2950241 EA2.07 Randomly selected EA2.02 -Ability to determine and/or interpret the following as they apply to HIGH DRYWELL PRESSURE:

Drywell temperature.

Question 44 , Knowledge of the interrelations between HIGH REACTOR PRESSURE and the following: RCIC: Plant-Specific. Nine Mile Point Unit 1 does not have RCIC. 1 11 2950251 EK2.07 Randomly selected EK2.01 -Knowledge of the interrelations between HIGH REACTOR PRESSURE and the following: RPS. Question 2 , Knowledge of the physical connections and/or cause-effect relationships between REACTOR PROTECTION SYSTEM and the following

Relief/safety valves (low-low-set logic): Plant-Specif i c. Nine M i le Point Unit 1 does not have low-low set logic associated with 2/1 2120001 K1 .07 relief/safety valves. Randomly selected K1 .02 -Knowledge of the physical connections and/or cause-effect relationships between REACTOR PROTECTION SYSTEM and the following:

Nuc l ear boiler instrumentation. Question 15 , Ability to (a) predict the impacts of the following on the LOW PRESSURE CORE SPRAY SYSTEM ; and (b) based on those predictions , use procedures to correct , control , or mitigate the consequences of those abnormal conditions or operations

Loss of fire protection:

BWR-1. Nine Mile Point Unit 1 is a BWR-2 , not a BWR-1 . 2/1 209001 1 A2.11 Randomly selected A2.06 -Ability to (a) predict the impacts of the following on the LOW PRESSURE CORE SPRAY SYSTEM ; and (b) based on those predictions , use procedures to correct , control , or mitigate the consequences of those abnormal conditions or operations

Inadequate system flow.

ES-401 Record of Rejec t ed KIA's Form ES-401-4 1 11 295038 1 2.1.;30 2/1 2610001 A1 .05 2/2 234000/2.1.19 2/1 300000 1 2.2.38 Question 79, Conduct of Operations:

Ability to locate and operate components, including local controls (High Off-site Release Rate). This KIA involves asking an SRO about the location and operation of local controls.

Writing a question on this topic and meeting SRO question requirements would be difficult.

Randomly selected 2.1.6 -Conduct of Operations:

Ability to manage the control room crew during plant transients.

Question 14 , Ability to predict and/or monitor changes in parameters associated with operating the STANDBY GAS TREATMENT SYSTEM controls including: Primary conta i nment oxygen level: Mark-I&II.

This KIA involves the re l ationship between SGTS Controls and 02 leve l s. There is no procedural refe r ence avai l able to write a question on this relationship. Randomly selected A 1.06 -Abi l ity to predict and/or monitor changes in parameters associated w i th operating the STANDBY GAS TREATMENT SYSTEM controls including: Drywell and suppression chamber differentia l pressure: Mark-I. Question 93 , Conduct of Operations

Ability to use plant computers to evaluate system or component status (Fuel Handling Equipment).

This KIA involves the relationship between Fuel Handling Equipment and the plant process computer. There is no direct relationship at Nine Mile Point Unit 1. Randomly selected 2.1 .32 -Conduct of Operations:

Ability to explain and apply all system limits and precautions.

Question 90, Equipment Control: Knowledge of conditions and limitations in the facility license (Instrument Air). There is no direct relationship between Instrument Air and the facility license. Additionally, this is one of four Instrument Air KlAs. Randomly selected 207000 Isolation (Emergency)

Condenser, 2.2.25 -Equipment Control: Knowledge of bases in technical specifications for limiting conditions for operations and safety limits.

ES-401 Record of Rejec t ed KIA's Form ES-401-4 2/2 215002 / K1.02 3/3 G3 / 2.3.11 1 / 1 295038/2.1.6 Question 27 , Knowledge of the physical connections and/or cause-effect relationships between ROD BLOCK MONITOR SYSTEM and the following:

LPRM: BWR-3 , 4 , 5. Nine Mile Point Unit 1 does not have a Rod Block Monitor. Randomly selected another Tier 2 System and KIA. 259001 Reacto r Feedwater, K1 .05 -Knowledge of the physical connections and/or cause effect relationships between REACTOR FEEDWATER SYSTEM and the following: Condensate System. Question 70 , Ability to control rad i ation releases.

This KIA is identical w i th the KIA for question 96. This topic is also covered in other KlAs in the exam . To prevent a potential double jeopardy question for an SRO candidate anothe r Gene r ic KIA will be randomly added. Randomly selected 2.3.5 , Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms , portable survey instruments , personnel monitoring equipment , etc. Question 79, Conduct of Operations

Ability to manage the control room crew during plant transients. This is not an acceptable KIA for a Tier 1 or Tier 2 topic. Randomly selected 2.4.18, Knowledge of the specific bases for EOPs.

1 / 1 295004 / AK3.03 2/2 201006/ A4.02 1 /2 295012/2.4.47 3/4 G3 / 2.4.25 Record of Rejected KIA's Form Question 46 , Knowledge of the reasons for the responses as they apply to PARTIAL OR LOSS OF D.C. POWER: Reactor SCRAM: Plant-SpecificThere are no procedura l references regarding a loss of and a reactor scramRandomly selected AK3.02 , Knowledge of the reasons the following responses as they apply to PARTIAL COMPLETE LOSS OF D.C. POWER: isolation/fault Question 36 , Ability to monitor automatic operations of ROD WORTH MINIMIZER SYSTEM (RWM) SPECIFIC) incl u ding: Pushbutton indicating switchesBased on lim i ted function of RWM pushbutton switches at Nine Mile Po i nt Unit 1 , th i s KIA has operational validityRandomly selected A4.04 , Ability to monitor operations of the ROD WORTH MINIMIZER (RWM) (PLANT SPECIFIC) including: Rod withdrawal indication

P-Spec Question 84 , Emergency Procedures

/ Plan: Ability diagnose and recognize trends in an accurate and time lmanner utilizing the appropriate control room material (High Drywell Temperature).

This is the 4th dealing with High Drywell Temperature (Questions 42 , and 76). Since Drywell Cooling, HCTL and CSIL have been tested, there is not a suitable SRO question to the Randomly selected from the untested Tier 1 Group 2 KlAs295007 , High Reactor Pressure , 2.4.6 , Knowledge of mitigation strategiesQuestion 99 , Knowledge of fire protection proceduresThis is the th i rd fire protection KIA on the SRO exam (a l#91 and #97). Re-sampling for better balance of Randomly selected 2.4.17 -Knowledge of EOP terms ES-301 Administrative Topics Outline Form ES-301-1 Facility:

Nine Mile Point Unit 1 Examination Level: RO. SRO o Administrative Topic Type Code* (see Note) Conduct of Operations M,S Conduct of Operations M,R Equipment Control N,R Date of Examination:

11/10 Operating Test Number: 1 Describe activity to be performed PERFORM RPV LEVEL INSTRUMENT CHECKS PER ST-DO, DAILY CHECKS Take control room reactor water level instrument readings for various daily checks required by Technical Specifications, enter the instrument readings into the applicable sections of the Daily Checks and take appropriate actions based on those checks. 2.1.7 (4.4) Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

N1-ST-DO PERFORM OWED AND DWFD LEAK RATE CALCULATIONS USING INTEGRATOR READINGS Given the OWED and DWFD integrator readings determine the identified and unidentified leak rates lAW Att 6 of 8. 2.1.18 (3.6) Ability to make accurate, clear, and concise logs, records, status boards, and reports. N1-0P-8 PREPARE A TAGOUT FOR RBCLC PUMP 13 Identify the isolations required to tagout RBCLC pump 13 for the shaft seal replacement.

Record the required isolations using CNG-OP-1.01-1007 attachment

8. 2.2.13 (4.1) Knowledge oftagging and clearance procedures.

CNG-OP-1.01-1007, N1-0P-11, P&IDC-18022-C, EWD C-19436-C ACTIONS FOR EXTERNAL SECURITY THREATS Given plant conditions, respond to a security threat per EPP-10, Attachment 2, Security Contingency Event (CSa Checklist)

Emergency M,S 2.4.28 (3.2) Knowledge of procedures relating to a security event (non-safeguards information).

EPIP-EPP-10 Attachment 2 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) (Nlew or (M)odified from bank (P)revious 2 exams (:::.1; randomly selected)

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

Nine Mile Point Unit 1 Examination Level: RO SRO

  • Administrative Topic Type Code* (see Note) Conduct of Operations D,R Conduct of Operations M,R Equipment Control D,R Date of Examination:

11/10 Operating Test Number: 1 Describe activity to be performed DETERMINE THERMAL LIMITS WITH INOPERABLE PRESSURE REGULATOR Given plant parameters including an inoperable reactor pressure regulator, determine the adjusted thermal limit values. Core Operating Limit Report graphs and a 3D Monicore printout are used to evaluate conditions against the adjusted thermal limits. 2.1.19 (3.8) Ability to use plant computers to evaluate system or component status. N1-RESP-1, Core Operating Limits Report, Technical Specifications ASSESS REPORTABILITY REQUIREMENTS Given a series of plant events, determine the reporting requirements per 10 CFR 50.72. 2.1.18 (3.8) Ability to make accurate, clear, and concise logs, records, status boards, and reports. 10 CFR 50.72, NUREG 1022, CNG-NL-1.01-1004 EVALUATE A COMPLETED SURVEILLANCE TEST AND TAKE THE REQUIRED ACTIONS Given a completed Surveillance Test, N1-ST-M1A, Liquid Poison Pump #11 Operability Test, complete the "Acceptance Criteria" and "SM Review" sections.

2.2.12 (4.1) Knowledge of surveillance procedures.

N1-ST-M1A, Technical Specifications GENERATE AND APPROVE AN EMERGENCY EXPOSURE AUTHORIZATION Radiation Control D,R Given a work activity, area dose rates and personnel dose history, determine the need for an emergency exposure authorization and select the appropriate person to perform the task. 2.3.4 (3.7) Knowledge of radiation exposure limits under normal and emergency conditions.

EPIP-EPP-15 CLASSIFY EMERGENCY EVENT AND PERFORM INITIAL NOTIFICATIONS Emergency Plan M,R Given plant conditions, determine event classification and complete initial notifications.

2.4.41 (4.6) Knowledge of the emergency action level thresholds and classifications.

EAL Matrix, EPIP-EPP-18, EPIP-EPP-20 All items (S total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all S are required.

  • Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (S3 for ROs; S 4 for SROs & RO retakes) (N)ew or (M)odified from bank (P)revious 2 exams (Si; randomly selected)

II ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility:

Nine Mile Point Unit 1 Date of Examination:

November 2010 Exam Level: RO/SRO Operating Test No.: 1 Control Room Systems@ (8 for RO; 2 or 3 for SRO-U, including 1 ESF) System 1 JPM Title Type Code* Safety Function S-1 Respond to a Loss of Service Water D,A,S 8 PLANT SERVICE The candidate will start the standby Service water pump. SYSTEMS The pump then trips, requiring override actions lAW 18.1. KIA 295018 AA.01 (3.3/3.4)

S-2 Bypass LPRM Input To APRM D,S 7 INSTRUMENTATION The candidate will bypassLPRM 20-25A input to its associated APRM lAW N1-0P-38C.

KIA 215005 A4.04 Synchronize Main Generator to Grid, Main Generator M,A,S S-3 Locks Out HEAT REMOVAL FROM CORE The candidate will complete synchronizing the Main Generator to the grid lAW N1-0P-32 and a generator lockout will occur, requiring N1-S0P-31.1 actions. KIA 245000 A4.02 (3.1/2.9)

D, L, S S-4 Rapid RWCU System Restoration for Level Control 2 REACTOR WATER INVENTORY The candidate will perform rapid RWCU system restoration CONTROL for RPV level control and establish reject flow to the condenser to lower level lAW N1-0P-3. KIA 204000 A4.06 (3.0/2.9)

S-5 Start the RB Emergency Ventilation System Loop D,EN,S 9 11 RADIOATIVITY RELEASE The candidate will manually start Reactor Building Emergency Ventilation System Loop 11 lAW N1-0P-10.

KIA 288000 A4.01 (3.1/2.9)

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 S-6 MSIV Stroke Test and Limit Switch Test The candidate will perform the MSIV Stroke Test and Limit Switch Test lAW N1-ST-Q26 for MSIV 112. P,S 3 REACTOR PRESSURE CONTROL S-7 KIA 239001 A4.01 (4.2/4.1)

NRC 2009 Perform Rod Block Withdrawal Test The candidate will select and withdraw a control rod and perform an over-travel check lAW N1-ST-R4.

The rod will be uncoupled.

The candidate will re-couple the control rod lAW N1-0P-5 and complete the test. , I N,A, L, S 1 REACTIVITY CONTROL KIA 201003 A2.02 (3.7/3.8)

S-8 Vent the Drywell Prior to Personnel Entry N,S 5 The candidate will lineup and vent the Drywell to lower pressure prior to personnel entry lAW N1-0P-9. PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES KIA 223001 A4.03 (3.4/3.4)

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

  • P-1 Lineup Lake Water to Supply the Emergency Condenser Makeup Tanks using the Electric Fire Pump M,A,E,R 4 HEAT REMOVAL FROM REACTOR CORE The candidate will attempt to lineup the Diesel Fire Pump to supply lake water to the Emergency Condenser Makeup Tanks lAW N1-S0P-21.2.

The Diesel Fire Pump will fail, requiring use of the Electric Fire Pump. KIA 207000 2.1.30 (4.4/4.0)

P-2 Transfer RPS Bus 11 from UPS 162A to UPS 162B The candidate will place UPS 1628 in service and place UPS 162A in standby lAW N1-0P-40.

D, R 6 ELECTRICAL P-3 KIA 262002 2.1.20 (4.6/4.6)

Inject Boron Into the Reactor Using the Hydro Pump The candidate will lineup and inject boron using the Hydro Pump lAW N1-EOP-3.2.

D,E,R 1 REACTIVITY CONTROL KIA 295037 EA1.10 (3.7/3.9)

II Control Room/In-Plant Systems Outline Form ES-301-2 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 Criteria for RO I SRO-II SRO-U (A)ltemate path 4-6 1 4-6 1 2-3 (C)ontrol room (D)irect from bank S9/s8/S4 (E)mergency or abnormal in-plant ::::1/::::1/::::1 (EN)gineered safety feature I ::::1 (control room system) (L)ow-Power 1 Shutdown ::::1/::::1/;::1 (N)ew or (M)odified from bank including 1 (A) ;::2/;::2/;::

1 (P)revious 2 exams s 3 1:;; 31 :;; 2 (randomly selected) (R)CA ::::1/::::1/;::1 (S)imulator Appendix Scenario Outline Form ES-O-1 Facility:

Nine Mile Point Unit 1 Scenario No.: NRC-01 Op-Test No.: 11/10 Examiners:

Operators:

__________ Initial Conditions:

Simulator IC-151 1. Reactor power is approximately 85% 2. APRM 14 is bypassed 3. CRD Pump 11 is out of service Turnover:

1. Return APRM 14 to service 2. Raise power to 100% with recirculation flow Malf. No. Event No. Type*

Place APRM 14 in service N (BOP) 1 N/A N (SRO) OP-38C Raise power with recirculation flow R (RO)2 N/A R (SRO) OP-43B EPR 3 4 ED07 7 Containment Spray Raw Water pump 121 trip 8 ERV 111 fails to open 9 AD07A C (ALL) EOP-8 * (N)ormal, (R)eactivity, (M)ajor NRC Scenario 1 -1 November 2010 Facility:

Nine Mile Point Unit 1 Scenario No.: NRC-01 Op-Test No.: 11/10 1. Total malfunctions (5-8) 7 Events 3-9 I 2. Malfunctions after EOP entry (1-2) 2 Events 8 and 9 3. Abnormal events (2-4) 4 Events 3-6 4. Major transients (1-2) Event 5. EOPs entered/requiring substantive actions (1-2) 2 I Events 6-9 (EOP-2, EOP-4) 6. EOP contingencies requiring substantive actions (0-2) Events 8 and 9 7. Critical tasks (2-3) CRITICAL TASK CT -1.0 Given a LOCA in the Drywell, the crew initiate Containment Sprays prior to exceeding Pressure Suppression Pressure limit, in with CT -2.0 Given a lowering torus water level, the crew will execute N1-EOP-8, RPV Blowdown, when it ! determined Torus water level cannot be above eight (8) feet, in accordance with NRC Scenario 1 November 2010 Appendix Scenario Outline Form ES-D-1 Facility:

Nine Mile Point Unit 1 Scenario No.: NRC-02 Op-Test No.: Examiners:

Operators:

_________Initial Conditions:

Simulator 1. Reactor power is approximately 100% 2. EDG 102 is ready for start Turnover:

1. Complete surveillance test N 1-ST -M4A 2. Lower power to 95% with recirculation flow Malf. No. Event No. Complete N1-ST-M4A, Emergency Diesel Generator 102 and PB 102 Operability Test 2 Override Lower power to 95% with recirculation flow R (RO) 3 NIA R (SRO) OP-43B RR pump 12 MIA station failure and delayed pump trip RR68B I (BOP)4 RR01B I (SRO) SOP-1.3 5 Override 6 RP01B 7 8 RD33 M (ALL) 9 Overrides C (ALL) * (N)ormal, (C)omponent, (M)ajor NRC Scenario 2 -1 November 2010 i I Facility:

Nine Mile Point Unit 1 Scenario No.: NRC-02 Op-Test No.: 11/10 1. Total malfunctions (5-8) Events 2, 2. Malfunctions after EOP entry (1-2) 1 I Event 9 3. Abnormal events (2-4) Events 4. Major transients (1-2) Event 5. EOPs entered/requiring substantive actions (1-2) 1 Event 7 and 8 (EOP-2) I 6. EOP contingencies requiring substantive actions (0-2) 1 . Events 8 and 9 (EOP-3) I 7. Critical tasks (2-3) CRITICAL TASK CT-1.0 Given lowering CRD system air pressure, crew will insert a manual reactor scram control rods begin drifting, in accordance with ARP..f3 and/or CT -2.0 Given a failure of the reactor to scram power above 6% and RPV water level above inches, the crew will terminate and prevent injection except boron and CRO, in accordance CT -3.0 Given a failure of the reactor to scram with power above 6%, the crew will lower reactor power by inserting control rods or injecting boron, in accordance with N1-EOP-3.

NRC Scenario 2 November 2010 Appendix D Scenario Outline Form ES-D-1 Facility:

Nine Mile Point Unit 1 Scenario No.: NRC-03 Op-Test No.: Examiners:

Operators:

_________Initial Conditions:

Simulator 1. Reactor power is approximately 100% Turnover:

1. Transfer Powerboard 101 supply from R1014 to R1011 in accordance with N1-0P-30 section H.8.0. Previous shift has completed step H.8.1. 2. Feedwater 11 is out of service for maintenance.

Malf. No. Event NRC Scenario 3 November 2010 Facility:

Nine Mile Point Unit 1 Scenario No.: NRC-03 Op-Test No.: 11120110 1. Total malfunctions (5-8) Events 2. Malfunctions after EOP entry (1-2) 1 Event 7 3. Abnormal events (2-4) 4 Events 2-5 4. Major transients (1-2) 5. EOPs entered/requiring substantive actions (1-2) 2 Events 6 and 7 (EOP*2, EOP-4) 6. EOP contingencies requiring substantive actions (0-2) Event 7 7. Critical tasks 3 CRITICAL TASK DESCRIPTIONS:

CT*1.0 Given an inadvertently open ERV at power, crew will close the ERV or insert a manual scram prior torus temperature exceeding 11 OUF, in accordance CT-2.0 Given a LOCA in the Drywell, the crew will Containment Sprays prior to exceeding the Suppression Pressure limit, in accordance with 4. CT-3.0 Given a LOCA with degraded high injection capability, the crew will depressurize the and inject with Preferred and Altemate Injection to restore and maintain RPV water level above inches, in accordance with I NRC Scenario 3 November 2010 Appendix D Scenario Outline Form ES-D-1 Facility:

Nine Mile Point Unit 1 Scenario No.: NRC-04 Op-Test No.: 11/10 Examiners:

Operators:

__________ Initial Conditions:

Simulator IC-154 1. Reactor power is approximately 85% 2. Containment Spray Pump 122 is OOS for repair (TS 3.3.7.b, day 1 of 15 day LCO). Turnover:

1. Shutdown Condensate Pump 13 for maintenance due to a motor oil leak 2. Perform a Rod uence Excha Malt. No. Event Type* NRC Scenario 4 November 2010 I Facility' Nine Mile Point Unit 1 Scenario No ' .. NRC-04 Op-Test No'.. 11/10 1. Total malfunctions (5-8) Events 3-8 2. Malfunctions after EOP entry (1-2) Events 7 and 3. Abnormal events (2-4) 3 ! Events 3-5 4. Major transients (1-2) Event 5. EOPs entered/requiring substantive actions (1-2) Events 6 and 7 (EOP-2, 6. EOP contingencies requiring substantive actions (0-2) Event 7 7. Critical tasks (2-3) CRITICAL TASK CT-1.0 Given an un-isolable RWCU leak outside containment and one general area temperature the maximum safe limit, the crew will insert a reactor scram, in accordance with CT-2.0 Given an un-isolable RWCU leak outside primary containment and two general area temperatures above the maximum safe limit, the crew will execute N1-EOP-8, RPV Slowdown, in accordance with N1-EOP-5.

NRC Scenario 4 November 2010

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