ML12116A252

Final Outlines (Folder 3)

ML12116A252
Person / Time
Site:	FitzPatrick
Issue date:	02/22/2012
From:	Presby P A Operations Branch I
To:	Entergy Nuclear Generation Co
Jackson D E
Shared Package
ML113070677	List: IR 05000333/2012301 ML113330864 ML120610564 ML12072A344 ML12072A351 ML12072A353 ML12072A356 ML12116A225 ML12116A242 ML12116A252 ML12116A259 ML12116A275 ML12116A290
References
TAC U01843
Download: ML12116A252 (21)

Text

BWR Examination Outline Form ES*401*1 James A.

Date of Exam: 3/8/12 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. 1 2 4 5 2 4 3 20 5 ! 2 7 Emergency

& Abnormal 2 0 2 2 N/A 1 0 N/A 2 7 2 1 Tier Totals 6 7 3 4 5 27 3 10 1 3 I 2 I 3 1 2 5 1 2 2 2 3 26 3 2 5 2. 2 3 1 1 2 0 1 2 0 2 0 0 12 0 1 2 3 Systems Plant Tier Totals 3 4 3 2 6 3 2 4 2 3 38 4 4 8 3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 3 2 2 3 2 1 2 2 Note: Ensure that at least two topics from every applicable KIA category are sampled within each tier of the and SRO-only outlines (Le.* except for one category in Tier 3 of the SRO-only outline. the "Tier in each KIA category shall not be less than 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. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important.

site-specific systems/evolutions that are not included on the outline should be added, Refer to Section O.1.b of ES-401 for guidance regarding the elimination of inappropriate KIA statements. 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. Absent a plant-specific priority. only those KlAs 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. Select SRO topics 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 0.1 ,b of ES-401 for the applicable KlAs. On the following pages. enter the KIA numbers. a brief description of each topiC. the topics' importance ratings (IRs) for the applicable license level. and the point totals (#) for each system and category, Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in other than Category A2 or G* on the SRO-only exam. enter it on the left side of Column A2 for Tier 2. Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams. For Tier 3. select topics from Section 2 of the KIA catalog. and enter the KIA numbers. descriptions.

IRs. and point totals (#) on Form ES-401-3.

Limit SRO selections to KlAs that are linked to 10 CFR 55.43.

2 Form ES-401-1 I ES-401 EIAPE # I Na'ne I Safety Function 295001 Parti<!ll or Complete Loss of Forced Core Flow Citculation 11 & 4 295003 Partial or Complete Loss of AC I 6 295004 Parti lor Total Loss of DC Pwr 16 295005 Main TurbinE! Generator Trip I 3 295006 SCR j\M f 1 295016 Room Abandonment 1 7 295018 Parti<ill or Total Loss of CCW 18 BWR Examination Form ES ",.

• KIA IR #ffiillili G 3.61 1 PARTIAL OR COMPLETE LOSS OF FORCED X AK 2.01 Knowledge of the interrelations between 3.7 CORE FLOW CIRCULATION and the following: (CFR: 41.7/45.8)

Recirculation system X 2.4.50 Emergency Procedures

/ Plan 4.2/ 2 4.0 Ability to verify system alarm setpoints and operate controls identified in the alarm response manual. (CFR: 41.10 I 43.5/ 45.3) AA 2.03 Ability to determine and/or interpret the 2.8/ 3 following as they apply to PARTIAL OR 2.9 COMPLETE LOSS OF D.C. POWER: (CFR: 41.10/43.5/45.13) Battery voltage AK 3.07 Knowledge of the reasons for the 3.81 4 following responses as they apply to MAIN 3.8 TURBINE GENERATOR TRIP: (CFR: 41.5/45.6)

Bypass valve operation AK 3.03 Knowledge of the reasons for the 3.81 5 following responses as they apply to SCRAM: 3.9 (CFR: 41.5 f 45.6) Reactor pressure response AA 2.03 Ability to determine and/or interpret the 4.31 6 following as they apply to CONTROL ROOM 4.4 ABANDONMENT: (CFR: 41.10/43.5/45.13) Reactor pressure AA 2.03 Ability to determine and/or interpret the 3.21 7 following as they apply to PARTIAL OR 3.5 COMPLETE LOSS OF COMPONENT COOLING WATER: (CFR: 41.10/43.5/45.13) Cause for partial or complete loss 295019 Partie lor Total Loss of Ins!. Air /8 295021 Loss of Shutdown Cooling / 4 295023 Refueling Ace / 8 295024 High Drywell Pressure / 5 295025 High Reactor Pressure / 3 295026 Suppression Pool High Water Temp, / 5 295027 High Containment Temperature / 5 295028 High Drywell Temperature

/5 295030 Low Suppression Pool Wtr Lvi / 5 X AA 2.01 Ability to determine and/or interpret the following as they apply to PARTIAL OR 3.5/ 3.6 8 COMPLETE LOSS OF INSTRUMENT AIR: (CFR: 41.10/43.5/45.13)

Instrument air system pressure X AK 3.04 Knowledge of the reasons for the following responses as they apply to LOSS OF 3.3/ 3.4 9 SHUTDOWN COOLING: (CFR: 41.5/45.6)

Maximizing reactor water cleanup flow X AK 1.02 Knowledge of the operational implications of the following concepts as they 3.2/ 3.6 10 apply to REFUELING ACCIDENTS: (CFR: 41.8 to 41.10) Shutdown margin X EA 1.17 Ability to operate and/or monitor the following as they apply to HIGH DRYWELL 3.9/ 3.9 11 PRESSURE: (CFR: 41.7/45.6)

Containment spray: Plant-Specific X EK 3.08 Knowledge of the reasons for the following responses as they apply to HIGH 3.5/ 3,5 12 REACTOR PRESSURE: (CFR: 41.5/45,6)

Reactor/turbine pressure regulating system operation X EK 3.02 Knowledge of the reasons for the following responses as they apply to 3,9/ 4,0 13 SUPPRESSION POOL HIGH WATER TEMPERATURE: (CFR: 41.5/45.6)

Suppression pool cooling X 2.4.35 Emergency Procedures

/ Plan 3.8/ 14 4,0 Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects, (CFR: 41.10/43.5/45.13)

X EK 2,04 Knowledge of the interrelations between LOW SUPPRESSION POOL WATER LEVEL and 3,7/ 3,8 15 the following: (CFR: 41,7/45,8)

RHR/LPCI 295031 Reactor Low Water Level 12 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown 11 295038 High Off-site Release Rate I 9 600000 Plant Fire On Site I 8 700000 Generator Voltage and Electric Grid Disturbances I 6 X 2.4.18 Emergency Procedures 1 Plan Knowledge of the specific bases for EOPs. (CFR: 41.10 143.1/45.13) EK 1.06 Knowledge of the operational implications of the following concepts as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: (CFR: 41.8 to 41.10) Cooldown effects on reactor power EA 1.03 Ability to operate and/or monitor the following as they apply to HIGH OFF-SITE RELEASE RATE: (CFR: 41.71 45.6) Process liquid radiation monitoring system AK 2.01 Knowledge of the interrelations between PLANT FIRE ON SITE and the following:

Sensors I detectors and valves AK 2.03 Knowledge of the interrelations between GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES and the following: (CFR: 41.4,41.5,41.7,41.10/45.8)

Sensors, detectors, indicators I 2 4 5 2 4 Group Point Total: 3.31 16 4.0 4.01 17 4.2 3.71 18 3.9 2.6/ 19 2.7 3.0/ 20 3.1 3 Form ES*401*1 ES-401 BWR Examination Outline Form Emergency and Abnormal Plant Evolutions

-Tier 1/Group 2 (RO I EIAPE # I Name I Safety Function K K K A A G KIA Topic(s) IR # 1 2 3 1 2 295002 Loss of Main Condenser Vac I 3 295007 High Reactor Pressure I 3 X AK 2.06 Knowledge of the interrelations between HIGH REACTOR PRESSURE and the following:

3.51 3.7 23 (CFR: 41.7 I 45.8) PCIS 295008 High Reactor Water Levell 2 295009 Low Reactor Water Levell 2 295010 High Drywell Pressure 15 X AA 1.04 Ability to operate andlor monitor the 3.11 24 following as they apply to HIGH DRYWELL 3.0 PRESSURE: (CFR: 41.7/45.6)

Drywell sampling system 295011 High Containment Temp 15 295012 High Drywell Temperature 15 AK 3.01 Knowledge of the reasons for the following 3.51 27 responses as they apply to HIGH DRYWELL 3.6 TEMPERATURE: (CFR: 41.5/45.6)

Increased drywell cooling 295013 High Suppression Pool Temp. I 5 295014 Inadvertent Heactivity Addition 11 295015 Incomplete SCRAM 11 3.21 22AK 2.03 Knowledge of the interrelations between INCOMPLETE SCRAM and the following: (CFR: 41.7 3.6 145.8) Rod control and information system 295017 High Off-site Release Rate I 9 295020 Inadvertent Cont. Isolation I 5 & 7 295022 Loss of CRD Pumps 11 4.31 262.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant 4.4 operation. (CFR: 41.10 I 43.5 I 45.2 I 45.6) 295029 High Suppression Pool Wtr Lvii 5 295032 High Secondary Containment Area T em perature I ei 295033 High Secondary Containment Area Radiation Levels / 9 295034 Secondary Containment Ventilation High Radiation / 9 X 2.1.28 Knowledge of the purpose and function of major system components and controls. (CFR: 41.7) 4.1/ 4.1 21 295035 Secondary Containment High Differential Pressure I 5 295036 Secondary Containment High Sump/Area Water Levell 5 X EK 3.02 Knowledge of the reasons for the following responses as they apply to SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL: (CFR: 41.5/45.6) 2.8/ 2.8 25 Reactor SCRAM 500000 High CTMT Hydrogen Conc. I 5 M 01 21 2 I 1 o ? nt Total: 7 ES*401 4 Form ES*401*1 ES-401 BWR Examination Outline Plant Systems -Tier 2/GrouD 1 (RO 1 SRO) Form ES-401-1 , , K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4 G KIA Topic(s) IR # 203000 RHR/LPCI:

Injection Mode X A 2.10 Ability to (a) predict the impacts of the following on the RHRlLPCI:

INJECTION MODE (PLANT SPECIFIC)

and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations

(CFR: 41.5/45.6) 3.31 3.5 28 Nuclear boiler instrument failures 205000 Shutdown Cooling X K 3.03 Knowledge of the effect that a loss or malfunction of the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) will have on following: (CFR: 41.7/45.4) 3.81 3.9 29 K3.03 Reactor temperatures (moderator, vessel, flange) 206000 HPCI X 2A.6 Knowledge of EOP mitigation strategies. (CFR: 41.10/43.5/45.13) 3.71 4,7 30 207000 Isolation (Emergency)

Condenser 209001 LPCS X A4.01 Ability to manually operate and/or monitor in the control room: (CFR: 41.7/45.5 to 45.8) 3.81 3.6 32 Core spray pump 209002 HPCS 211000 SLC X K 1.05 Knowledge of the physical connections andlor cause effect relationships between STANDBY LIQUID CONTROL SYSTEM and the following: (CFR: 41.2 to 41.9/45.7 to 45.8) 3.4/ 3.6 31 RWCU 212000 RPS X K 6,02 Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR PROTECTION SYSTEM: (CFR: 41.7/45.7) 3.71 3.9 33 Nuclear instrumentation 2150031RM 215004 Source Rangle Monitor 215005 APRM / LPRM 217000 RCIC 218000 ADS 223002 PCIS/Nuclear Steam Supply Shutoff X K 3.04 Knowledge of the effect that a loss 3.6/ 34 or malfunction of the INTERMEDIATE 3.6 RANGE MONITOR (IRM) SYSTEM will have on following: (CFR: 41.7/45.4)

Reactor power indication X K 2.01 Knowledge of electrical power supplies to the following: (CFR: 41.7) 2.6/ 2.8 35 SRM channels/detectors X K 1.02 Knowledge of the physical 3.7/ 36 connections and/or cause effect 3.7 relationships between AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM and the following: (CFR: 41.2 to 41.9/45.7 to 45.8) IRM X K 5.01 Knowledge of the operational implications of the following concepts as 2.6/ 2.6 37 they apply to REACTOR CORE ISOLATION COOLING SYSTEM (RCIC): (CFR: 41.5/45.3)

Indications of pump cavitation X X K 6.01 Knowledge of the effect that a loss or malfunction of the following will have on 3.9/ 4.1 38/39 the AUTOMATIC DEPRESSURIZATION SYSTEM: (CFR: 41.7/45.7)

RHRlLPCI system pressure:

Plant-Specific 2.1.27 Knowledge of system purpose and/or function. (CFR: 41.7) 3.9/ 4.0 X K 3.07 Knowledge of the effect that a loss or malfunction of the PRIMARY 3.7/ 38 40 CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF will have on following: (CFR: 41.7/45.4)

Reactor pressure 239002 SRVs 259002 Reactor Water Level X Control 261000 SGTS 262001 AC Electrical Distribution 262002 UPS (AC/DC) 263000 DC Electrical Distribution X X K 3.01 Knowledge of the effect that a loss or malfunction of the RELIEF/SAFETY 3.91 4.0 41/42 VALVES will have on following: (CFR: 41.7 145.4) Reactor pressure control A 2.03 Ability to (a) predict the impacts of 4.11 the following on the RELIEF/SAFETY 4.2 VALVES; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5/45.6)

Stuck open SRV K 1 .14 Knowledge of the physical connections and/or cause effect 2.9/ 3,0 43 relationships between REACTOR WATER LEVEL CONTROL SYSTEM and the following: (CFR: 41,2 to 41,9/45.7 to 45.8) Main turbine X X K 6.03 Knowledge of the effect that a loss or malfunction of the following will have on 3.01 3,1 44/45 the STANDBY GAS TREATMENT SYSTEM: (CFR: 41.7/45.7)

Emergency diesel generator system A 4.06 Ability to manually operate and/or monitor in the control room: 3.31 3,6 Reactor building differential pressure X K 6.01 Knowledge of the effect that a loss or malfunction of the following will have on 3,11 34 46 the A.C, ELECTRICAL DISTRIBUTION: (CFR: 41,7/45.7)

D.C. power X A 3.01 Ability to monitor automatic operations of the UNINTERRUPTABLE 2.8/ 3.1 47 POWER SUPPLY (A.C.lD.C.)

including: (CFR: 41.7/45,7)

Transfer from preferred to alternate source X 4.61 48 2.1.20 Ability to Interpret and execute 4.6 procedure steps. (CFR: 41.10/43.51 45.12) 264000 EDGs X A 3.05 Ability to monitor automatic operations of the EMERGENCY GENERATORS

{DIESEUJET}

including: (CFR: 41.7/45.7)

Load shedding and sequencing 300000 Instrument Air X X K 2.01 Knowledge of electrical power supplies to the following: (CFR: 41.7) Instrument air compressor K 5.13 Knowledge of the operational implications of the following concepts as they apply to the INSTRUMENT AIR SYSTEM: (CFR: 41.5/45.3)

Filters 400000 Component Gaoling Water X X K 6.05 Knowledge of the effect that a loss or malfunction of the following will have on the CCWS: (CFR: 41.7/45.7)

Motors A 1.04 Ability to predict and / or monitor changes in parameters associated with operating the CCWS controls including: (CFR: 41.51 45.5) Surge Tank Level I KIA Category Point Totals: 5 1 3.41 3.5 49 2.81 2.8 2.9/ 2.9 2.8/ 2.9 2.8/ 2.8 50151 52153 26 ES-401 5 Form ES-401*1 ES-401 System # / Name K 1 BWR Examination Outline K G 234 KIA Topic(s) Form ES-401-1 IR # 201001 CRD Hydraulic X K 1.03 Knowledge of the physical connections andlor cause effect relationships between CONTROL ROD DRIVE HYDRAULIC SYSTEM and the following: (CFR: 41.2 to 41.9 145.7 to 45.8) 3.11 3.1 54 Recirculation pumps (seal purge): Plant-Specific 201002 RMCS X K 4.05 Knowledge of REACTOR MANUAL CONTROL SYSTEM design feature(s) and/or interlocks which provide for the following: (CFR: 41.7) 3.31 3.3 55 "Notch override" rod withdrawal 201003 Control Rod and Drive Mechanism 201004 RSCS P++H 201005 RCIS 201006 RWM 202001 Recirculation X A 1.01 Ability to predict andlor monitor changes in parameters associated with operating the RECIRCULATION SYSTEM controls including: (CFR: 41.5/45.5) 3.61 3.5 56 Recirculation pump flow: Plant-Specific 202002 Recirculation Flow Control 204000 RWCU 214000 RPIS X H K 4.02 Knowledge of ROD POSITION INFORMATION SYSTEM design feature(s}

and/or interlocks which provide for the following: (CFR: 41.7) 2.51 2.5 57 Thermocouple 215001 Traversing In-cc,re Probe X K 1.05 Knowledge of the physical connections and/or cause effect relationships between TRAVERSING IN-CORE PROBE and the following: (CFR: 41.2 to 41.9/45.7 to 45.8) 3.31 3.4 58 Primary containment isolation system 215002 RBM 216000 Nuclear Boiler Inst. X K 1.05 Knowledge of the physical connections and/or cause effect relationships between NUCLEAR BOILER INSTRUMENTATION and the following: (CFR: 41.2 to 41.9/45.7 to 45.8) 3.71 39 59 Residual heat removal: Plant-Specific 219000 RHRlLPCI:

TorLlslPool 223001 Primllf}'

CTMT clnd Aux, I tjl 226001 RHRlLPCI:

CTMT Spray Mode = 230000 RHRlLPC1:

ToruslPool Spray Mode 233000 Fuel Pool CoolinglCleanup X 2,81 60 K 2,02 Knowledge of electrical power supplies 2,9 to the following: (CFR: 41,7) RHR pumps 234000 Fuel Handling X 3,11 61 A 3,02 Ability to monitor automatic operations 3,7 of the FUEL HANDLING EQUIPMENT including: (CFR: 41,7145,7)

Interlock operation 239001 Main and Reheat Steam X 2,81 62 A 3,03 Ability to monitor automatic operations 2,8 of the MAIN AND REHEAT STEAM SYSTEM including: (CFR: 41,7145,7)

Moisture separator reheat steam supply: Plant-S pecific 239003 MSIV Leakage Control II 241000 ReactorlTurbine 245000 Main Turbine Gen, I 256000 Reactor Condensate X 2,81 K 6,04 Knowledge of the effect that a loss or 28 malfunction of the following will have on the REACTOR CONDENSATE SYSTEM: (CFR: 41,7/45,7)

A.C, power I 259001 Reactor Feedwater 268000 Radwaste ttt 271000 Offgas X 3,31 64 A 1,01 Ability to predict and/or monitor 3.2 changes in parameters associated with operating the OFFGAS SYSTEM controls including: (CFR: 41.5/45,5)

Condenser vacuum 272000 Radiation 286000 Fire Protection X 3,61 K 3,03 Knowledge of the effect that a loss or 3.8 malfunction of the FIRE PROTECTION SYSTEM will have on following: (CFR: 41.71 45.4) Plant protection 288000 Plant Ventilation 290001 290003 Control Room 290002 Reactor Vessel 3 0 0 Total:

ES*401 Generic Knowledge and Abilities Outline (Tier 3) Form ES*401*3 .. es A. Fitzpatrick Date of Exam: 3/8/12 Cate90ry KIA # 2.1.34 1. Conduct of Operations 2.1.35 2.1. 2.1. 2.1. 2.1. Subtotal 2.2.40 2. Equipment Control 2.2. 2.2. 2.2. 2.2. 2.2. Subtotal 2.3.14 3. Radiation Control 2.3.4 2.3. 2.3. 2.3. 2.3. Subtotal!

Topic Knowledge of primary and secondary plant chemistry limits. (CFR: 41.10/43.5/45.12)

IR RO # SRO-Only IR # 3.5 94 Knowledge of the fuel-handling responsibilities of SROs. 3.9 95 (CFR: 41.10 143.7) -Ability to apply Technical Specifications for a system. 4.7 96 (CFR: 41.10 143.2/43.5/45.3)

I Knowledge of radiation or contamination hazards that 3.8 97 may arise during normal, abnormal, or emergency conditions or activities. (CFR: 41.12/43.4145.10) 3.7 9i) Knowledge of radiation exposure limits under normal or emergency conditions. (CFR: 41.12/43.4/45.10) 2 2.4.28 Knowledge of procedures relating to a security event 4, I 99 4. Emergency (non-safeguards information). (CFR: 41.10/43.5/45.13)

Procedures I Plan 2.4.44 Knowledge of emergency plan protective action 4,4 100 recommendations. (CFR: 41.10/41.12/435/45.11)

I 2.4. 2.4. 2.4. 2.4. Subtotal 2 Tier 3 Point Total 7 ES-40:...:1

_________.:..:R:..:.e.::.co::.:r:..:d:...:o:..:f..:,R.:.:e:!;ie:.:c:.:te::.:d:....:.,:KI:.:.A.::s=------

_____-=F0iiiii:lr=mi:l:::iEE=S-=4:=0:::i::1-=4 Tier I Randomly Group Selected KIA Tier 1 I 295001 Group 1 AK 2.08 Tier 1 I 295001 Group 1 AK 2.05 Tier 1 I 295026 Group 1 EK 3.03 Tier 1 I 295027 Group 1 Tier 1 I 295011 Group 2 Tier 1 I 295007 Group 2 AK 2.01 Reason for Rejection JAF is a BWR-4 (K&A specific to BWR-1) LPCI loop select logic not applicable at JAF K&A not applicable to JAF operating procedures for high torus temperature All K&A's applicable to Mark III containment only All K&A's applicable to Mark III containment only Rejected K&A AK 2.01 on basis of previously sampled K&A similarities.

Similarities are: 012: EPE 295025 High Reactor Pressure, EK 3.08: Reactor/turbine Qressure regulating s)!stem oQeration. 023: APE 295007: High Reactor Pressure, AK 2.01: Reactor/turbine Qressure regulating s)!stem oQeration. Replaced AK 2.01 with randomly selected APE 295007, AK2.06, The guidance for this rejection is stated in ES-401 Section D 1.d: ensure that no EPE/APE, system, or KIA category is over-sampled

... Page 1 of2 Tier 2 / 207000 K&A's not applicable to JAF (Iso/Emergency Condensers)

Group 1 Tier 2 / 209002 K&A's not applicable to JAF (HPCS) Group 1 Tier 3 Could not write a discriminating, operationally oriented question.

Randomly selected 2.3.11. Tier 2 / 203000 A2.01 Rejected K&A A2.01 on basis of previously sampled K&A Group 1 similarities.

Similarities are: Q 15 was already written testing the concept of inadequate NPSH in LPCI mode. 203000 A2.01 addressed inadequate NPSH in LPCI mode also. Replaced with 203000 A2.01 with randomly selected 203000 A2.1 0 (NBI failures).

Page 2 of2 ES-301 Administrative Topics Outline -Rev 2 Form ES-301-1 Facility:

James A. Fitzpatrick Examination Level: RO S RO X Date of Examination:

2/27/12 Operating Test Number: Administrative Topic Type (see Note) Code* Describe activity to be performed D,R Verify F uel Movement Sheets Conduct of Operations 2.1.35 SRO 3.9 D,R NPO Task -Qualification Check Conduct of Operations 2.1.8 SRO 4.1 N,R ST-26K Recirc Loop Startup Differential Temperature Equipment Control SRO Review 2.2.12 SRO 4.1 N,R Determine Requirements for Drywell Entry and Tech Radiation Control Spec LCO 2.3.12 SRO 3.7 Determine Protective Action Recommendations and P,D,R Emergency Procedures/Plan Complete Event Notification Form 2.4.38 SRO 4.4 EE 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. (C)ontrol room, (S)imulator, or Class(R)oom I'Type Codes & Criteria: (D}irect from bank (:$ 3 for ROs; s 4 for SROs & RO (N}ew or (M)odified from bank (;:: (P)revious 2 exams (:$ 1; randomly

____

__

Facility:

James A. Fitzpatrick Date of Examination: Exam Level: RO SRO-I X SRO-U Operating Test ! Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF) r System I JPM Title Type Code* Safety Function a. Switch Relay Room Ventilation N/S 9 b. Reset RPS scram with scram valve fail to A/D/LIS 7 c. Place HPCI in pressure control A I EN I LIM I S 4 : d. Re-open MSIVs with RPV DIS 3 e. Transfer 10300 (10400) bus from T-4 to Reserve Transformer A/M/S 6 f. Une-up 'A' RHR keep full for injection to D/E/LIS 2 g. Restore RB Ventilation following DIS 5 In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for i. Vent the Scram Air D/E/R 1 j. Vent Torus to lower Primary Containment AIEILIM 5 k. Supply cooling water to EDG 'B' I '0' from Fire 0 8 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 I 4-6 / 2-3 (C)ontrol room (D)irect from bank S9/S8/S4 (E)mergency or abnormal in-plant 2:1/2:1/2:1 (EN)gineered safety feature -/ -/ 2:1 (control room system) (L)ow-Power I Shutdown 2:1/2:1/2:1 (N)ew or (M)odified from bank including 1 (A) 2:2/2:2/2:1 (P)rEwious 2 exams s 31 s 3 1s 2 (randomly selected) (R)CA 2:1/2:1/2:1 (S)imulator i

____

James A. Fitzpatrick Date of Examination:

2/27/12 Exam Level: RO SRO-I SRO-U Operating Test No.: Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF) System I ,I PM Title Type Code* Safety Function a. Switch Relay Room Ventilation N/S 9 b. c. Place HPCI in pressure control AI EN I LI MIS 4 d. e. f. Line-up 'A' RHR keep full for injection to D/E/LIS 2 g. In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for i. Veint the Scram Air D/E/R 1 j. Vent Torus to lower Primary Containment A/E/LIM 5 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 1 SRO-I 1 SRO-U (A)lternate path 4-6 1 4-6 1 (C)ontrol (D)irect from bank (E)mI3rgency or abnormal in-plant safety feature 1 (control room (L)ow-Power 1 Shutdown (N)ew or (M)odified from bank including 1 (A) (P)revious 2 exams => 3 1=> 3 1=> 2 (randomly (R)CA Appendix D Scenario Outline Form ES-D-1 Facility:

Fitzpatrick Scenario No.: 1 Op-Test No.: Rev 2 Examiners:

Operators:

Initial Conditions:

Reactor at 75% power Turnover:

Reactor is at 75% power. No equipment is OOS. A rod sequence exchange is in progress.

There are 3 rods that are to be moved from position 00 to position 48. After the exchange is completed, reactor power is to be returned to 100% using Recirc. EV4:mt Malf. Nlo. No. I NA 2 RD07 3 HP05 4 DG06A DGOIC ZDILLHO NC05 TRIP 5 MS02 6 RPOIAIB RP09 RDIO:AII 7 SLOIAIB SL05AIB 8 RDI3 Event Type* N -ATC, SRO C -ATC, SRO R-BOP, SRO TS-SRO C-ATC, TS-SRO C-AII TS-SRO C-BOP, SRO M-AII C-ATC, SRO C-ATC, SRO 2 Rods are moved to position 48 and coupling check performed.

3 rd Rod moved to position 48 but drifts in and then driven full in. (AOP-27) Reactor power reduced with Recirc to < 25% pre-transient or 42.4 Mlbmlhr core flow. HPCI inadvertent start Loss of 10500 Bus with slow start of EDG "A" and failure of EDG "c" to start. RPS is lost and restored with MG set or Alternate power supply. The 12 scram is reset. (AOP 18 and AOP 59) Small leak in drywell. Reactor scram is attempted prior to reaching 2.7 psig. (AOP-39) High power A TWS, injection is terminated and prevented.

Level lowered to 110" to prevent power oscillations. (EOP-3) Failure of first SLC pump, 2 nd SLC pump works for seconds and then trips. Removal ofRPS fuses partially works. The remainder of the control rods will be inserted by inserting another manual scram or driving the rods in. * (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Appendix Scenario Outline Form ES-D-1 Facility:

Scenario No.:2 Op-Test No.: Rev 2 Operators:

Initial Conditions:

68% power.

The plant is shutting down due to Tech Spec LCO Action Statement.

"8" CRO is tagged out of service, AOS '0' is inoperable and RHR Loop 'A' is out of service due to Heat Exchanger issue. Continue to cold shutdown lAW OP-65. Event No. NA NA AD05 AN932:01 5 ED43A 6 FW19B\C 7 OS RPS3A\B TCII ED44 RRI5 HPOI HP02 RRl5 EDlSA CS02B Event Type* R-ATC, SRO N-BOP, SRO C BOP, SRO TS-SRO TRM SRO TRM-SRO M-All C -ATC, SRO C-ATC M-All M-AII C-BOP, SRO M -All C-BOP, SRO Lower power to 65% with control rods. Remove Condensate and Condensate booster pump from service SRV fails open and then closes when fuses pulled (AOP-36).

Core Spray Pipe Break Detector Alarm Loss of 115kv Line # 3.

Condensate Pump 'B' trip, 'C' fails for RPS fails to scram reactor -Mode Switch works Main Turbine fails to auto Loss of Off site Power -Manually close MSIVs due to loss of Circulating Coolant leak in drywell requiring initiation of drywell HPCI fails to automatically initiate.

After manual initiation, HPCI Leak greater than capacity ofRCIC and CRD. Level lowers, Alternate Level Control entered and Emergency Depressurize at -19". Level to be restored greater than T AF with low pressure systems. Loss 10500 bus Core Spray iniection valve fails to auto open * (R)eactivity, (I )nstrument, (C)omponent, (M)ajor