ML083230507

Final Outlines (Folder 3)

ML083230507
Person / Time
Site:	Nine Mile Point
Issue date:	10/01/2008
From:	Nine Mile Point
To:	D'Antonio J M Operations Branch I
Hansell S
Shared Package
ML081060454	List: IR 05000220/2008302 ML081620386 ML082770233 ML082980037 ML083040099 ML083040103 ML083040108 ML083040111 ML083040117 ML083120374 ML083220034 ML083230106 ML083230120 ML083230370 ML083230376 ML083230387 ML083230427 ML083230440 ML083230507 ML090330469
References
TAC U01690
Download: ML083230507 (36)

Text

ES-401 NMPI Written Examination Outline Form ES-401-1 Note 1. 2. 3. 4. 5. 6. 7.

• 8. 9. Ensure that at least two topics from every applicable WA 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 WA category shall not be less than two). 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 +I 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 that are not included on the outline should be added. Refer to section D.l .b of ES-401, for guidance regarding elimination of inappropriate WA 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 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.

Select SRO topics for Tiers 1 and 2 from the shaded systems and KIA categories. The generic (G) WAS in Tiers 1 and 2 shall be selected from Section 2 of the WA Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.l .b of ES-401 for the applicable WAS On the following pages, enter the WA numbers, 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. I does not apply). Use duplicate pages for RO and SRO-only exams. For Tier 3, select topics from Section 2 of the WA Catalog, and enter the WA numbers, descriptions, IRs, and point totals

(#) on Form ES-401-3.

Limit SRO selections to WAS that are linked to

.. 10CFR55.43 ES-401 EAPE#/NameSafetyFunction K1 K2 K3 AI A2 G K/A Topic(s) Imp. Form ES-401-1 Q# NMPI Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 4.2 3.5 4.2 4.1 4.3 4.3 4.1 3.4 3.3 3.3 3.3 3.2 76 77 78 79 80 81 82 39 40 41 42 43 295031 Reactor Low Water Level / 2 2950 16 Control Room Abandonment

/ 7 295028 High Drywell Temperature

/ 5 295006 SCRAM 1 I 295001 Partial or Complete Loss of Forced Core Flow Circulation

/ 1 & 4 295003 Partial or Complete Loss of AC / 6 Rate / 9 295037 SCRAM Conditions Present and Reactor Power Above APRM Downscale or Unknown /I X X 295005 Main Turbine Generator Trip I3 I lXl 295024 High Drywell Pressure

/ 5 295003 Partial or Complete Loss of AC / 6 e followmg as they to REACTOR LOW WATER mdcations, or response AbMy to pr/ont/ze and interpret the siqnificance of each annunciator or EKI .07 - Knowledge of the operational implications of the following concepts as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: Shutdown margin interrelations between MAIN TURBINE GENERATOR TRIP and AK2.04 - Knowledge of the the following:

Main generator concepts as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER Loss of breaker ES-401 EAPE#INameSafetyFunction Form ES-401-1 K1 K2 K3 A1 A2 G KIA Topic(s) Imp. Q# N MP 1 Written Exam inat ion Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 295026 Suppression Pool High Water Temp.

I5 295006 SCRAM I 1 I 295030 Low Suppression Pool Water Level I5 I 295028 High Drywell Temperature 15 295025 High Reactor Pressure I 3 295016 Control Room Abandonment 17 295038 High Off-site Release Rate I9 600000 Plant Fire On-site 18 700000 Generator Voltage and Electric Grid Disturbances I 295001 Partial or Complete Loss of Forced Core Flow Circulation

/ 1 & 4 I 295021 Loss of Shutdown Cooling I4 Level I2 EK3.06 - Knowledge of the reasons for I the following responses as they monitor the following as they apply to HIGH REACTOR PRESSURE:

1 4.5 I 48 I Condenser:

Plant- logic used to assess the status of control, core cooling and heat removal, reactor coolant system ES-401 EAPE#INarneSafetyFunction K1 K2 K3 AI A2 G WA Topic(s) Imp. Form ES-401-1 Q# N M P 1 Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 WA Category Totals:

2 295019 Partial or Complete Loss of Inst. Air I8 295023 Refueling Acc I 8 295018 Partial or Complete Loss of ccw I8 ES-401 EAPE#INameSafetyFunction K1 K2 K3 AI A2 G KIA Topic(s) Form ES-401-1 Imp. Q# NMPI Written Examination Outline Emergency and Abnormal Plant Evolutions -Tier 1 Group 2 3.4 4.2 4.0 3.8 3'1 83 84 85 59 60 295020 Inadvertent Coni.

Isolation

/ 5 8. 7 I 3.4 295007 High Reactor Pressure / 3 61 295010 High Drywell Pressure / 5 3.7 I 62 lx 295015 Incomplete SCRAM I 1 3.8 I 63 295008 High Reactor Water Level I2 3.3 295002 Loss of Main Condenser Vacuum

/ 3 64 295007 High Reactor Pressure 13 3.1 295033 High Secondary Containment Area Radiation Levels I9 295032 High Secondary Containment Area Temperature I5 500000 High CTMT Hydrogen Conc.

I5 KIA Category Totals:

1 65 Ability to verify system alarm setpoints and operate controls identified in the alarm response concepts as they apply to INCOMPLETE SCRAM

Reactor monitor the following as they apply to HIGH REACTOR PRESSURE : Reactorlturbine pressure regulating SECONDARY CONTAINMENT AREA RADIATION LEVELS and wing: Area Rad Monitoring ES-401 KKKKKKAN 1234561 System #I Name NMPI Written Examination Outline Plant Systems - Tier 2 Group 1 AA G Imp Q# 34 Form ES-401-1 300000 Instrument Air 218000ADS 205000 Shutdown Cooling 207000 Isolation (Emergency)

Condenser 263000 DC Electrical Distribution 209001 LPCS 239002 SRVs abnormal conditions or operations: ADS failure to Plan: Knowledge of low power/shutdown implications in accident . LOCA or loss of RH K1.05 - Knowledge of the physical connections and/or cause- effect relationships between LOW PRESSURE CORE SPRAY SYSTEM and the 2.8 4.2 - 4.6 - 4.7 - 2.9 - 3.7 3.6 - 86 87 89 1 2 -

ES-401 AA G KKKKKKAA2 1234561 34 System # / Name Form ES-401-1 Imp Q# NMPI Written Examination Outline Plant Systems - Tier 2 Group 1 259002 Reactor Water Level Control 218000 ADS 261 000 SGTS 205000 Shutdown Cooling 300000 Instrument Air physical connections and/or relationships between STANDBY GAS TREATMENT SYSTEM I I control 1 K4.11 - Knowledqe of ELECTRICAL DISTRIBUTION design feature(s) and/or interlocks which provide for the - 3.8 3.1 - 2.9 - 3.6 3.3 - 2.9 - 2.8 2.5 - 3 4 6 7 9 10 ES-40 1 KKKKKKAN AA G 1234561 34 System #I Name Form ES-401-1 Imp Q# NMPI Written Examination Outline Plant Systems - Tier 2 Group I 262002 UPS (ACIDC) 206000 HPCl 263000 DC Electrical Distribution 21 1000 SLC 400000 Component Cooling Water 207000 Isolation (Emergency)

Condenser 223002 PClSlNuclear Steam Supply Shutoff that a loss or malfunction of the following will have on the HIGH PRESSURE COOLANT INJECTION on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal oDeration:

I I Highllow surge tank level I A2.06 - Abilitv to (a) predict the 2.7 - 2.9 - 2.5 3.6 - 2.8 3.3 3.4 11 13 14 - 15 16 17 KKKKKAA2 AA 234561 34 G Imp Q# -I " Form ES-401-1 ES-40 1 NMPI Written Examination Outline Plant Systems - Tier 2 Group 1 System #I Name I 19 3.2 - 3.4 - 3.2 262001 AC Electrical Distribution 215004 Source Range Monitor 215005 APRM I LPRM 212000 RPS 4.2 - 4.4 21 - 22 264000 EDGs 215003 IRM 2.5 - 4.4 23 - 24 207000 Isolation (Emergency)

Condenser 400000 Component Cooling Water 2.7 - 3.2 I 25 223002 PCIS/Nuclear Steam Supply Shutoff WA Category Totals: I 2615 ES-40 1 KKKKKKAM 1234561 System #I Name Form ES-401-1 Q # AA G Imp. 34 NMPI Written Examination Outline Plant Systems - Tier 2 Group 2 286000 Fire Protection 215001 Traversing In- core Probe 202001 Recirculation c System I 245000 Main Turbine Gen. I Aux. 239001 Main and Reheat Steam I I 286000 Fire Protection I 204000 RWCU I 201001 CRD Hydraulic 202001 Recirculation impacts of the following on the FlRE PROTECTlON SYSTEM ; and (b) based on those predictions, use physical connections andlor cause- effect relationships between MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS and the following: reactor water to various operational implications of the following concepts as they apply to CONTROL ROD DRIVE HYDRAULIC SYSTEM : Solenoid 2.9 91 4.6 92 4.7 93 2.6 27 3.2 28 3.6 29 2.9 30 2.5 31

, .,". , .", . KAA2 34 G Imp. 61 ES-40 1 Form ES-401-1 NMPI Written Examination Outline Plant Systems - Tier 2 Group 2 System # / Name - 2.7 - 2.6 34 271 000 Off-gas the PLANT VENTILATTON SYSTEMS ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Extreme outside weather conditions:

288000 Plant Ventilation 219000 RHWLPCI: ToruslPool Cooling Mode 2.9 - 2.9 35 - 36 201001 CRD Hydraulic Ability to perfom- specific system and integrated plant procedures during all modes of plant operation.

K3.01 - Knowledge of the effect that a loss or malfunction of the SECONDARY CONTAINMENT will have on following: Off-site radioactive release rates Group Point Total: 201002 RMCS 4.3 37 290001 Secondary CTMT 4.0 --I- 38 KIA Category Totals:

I 1213 Form ES-401-3 NMPI Generic Knowledge and Abilities Outline (Tier 3) ES-401 I NMPI Date: October 2008 RO SRO-On1 y IR Q# IR Q# 3.2 94 4.6 99 KIA # Topic Knowledge of facility requirements for controlling vital

/ controlled access.

Ability to interpret and execute procedure steps. 2' l3 2. 2.2.12 Knowledge of surveillance procedures.

3.7 68 2.2.39 3.9 69 Knowledge of less than one hour technical specification action statements for systems.

Ability to manipulate the console controls as required to operate the facility between shutdown and designated power levels. 4.6 74 2'2'2 duties such as response to radiation monitor Ability to comply with radiation work permit conditions.

2.3.7 requirements

during normal or abnormal 3.5 70 ES-40 1 4. Emergency Procedures

/ Plan Tier 3 Point Tok NMPI Generic Knowledge and Abilities Outline (Tier 3) 2.3.13 Knowledge of Radiological Safety Procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high radiation areas, aligning filters, etc. Form ES-401-3 T emergency plan implementation.

2-4*40 I Knowledge of operator response to loss of all 3.6 I 72 Subtotal 4.5 __. 97 98 - 2 7 -

ES-401 Tier / Group 111 111 111 111 111 111 111 1 I2 112 112 NMPI Record of Rejected WAS Form ES-401-4 Randomly Selected WA 295005 I AKI .01 Knowledge of the operational implications of the following concepts as they apply to MAIN TURBINE GENERATOR TRIP

Pressure effects on reactor power. 295004 I AKI .01 Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER:

Automatic load sheeding 295024 I EK2.17 Knowledge of the interrelations between HIGH DRYWELL PRESSURE and the following:

Aux Bldg isolation logic 295006 I AK3.03 Knowledge of the reasons for the following responses as they apply to SCRAM : Reactor pressure response 295030 I EK3.02 Knowledge of the reasons for the following responses as they apply to LOW SUPPRESSION POOL WATER LEVEL: HPCl operation 295025 I EA1.04 Ability to operate and/or monitor the following as they amlv to HIGH REACTOR PRE'SSURE:

HPCI 29501 9 I 2.4.47 Partial or Complete Loss of Inst. Air / Ability to diagnose and recognize trends in an accurate and timely manner I utilizing the appropriate control room reference material.

29501 7 I AK2.12 Knowledge of the interrelations between HIGH and the following: Standby gas treatmenVFRVS 295009 I AK3.01 OFF-SITE RELEASE RATE Knowledge of the reasons for the following responses as they apply to LOW REACTOR WATER LEVEL

Recirculation pump run back: Plant-Specific 295033 I EA2.02 Ability to determine and/or Reason for Rejection

(#40) Topic oversampled (see # 62) Randomly selected AK 2.04 (#41) Topic does not apply to NMPl. Randomly selected AKI .05 (M2) Topic does not apply to NMPI. Randomly selected EK2.18

(#45) Generic Fundamental Topic. Randomly selected AA1.02 (#46) Topic does not apply at NMPI. Randomly selected EK3.01 (#48) Topic does not apply to NMPI. Randomly selected EA1.06 (#56) Topic not related to EPE. Randomly selected 2.4.49 (#60) Oversampled (see #38). Randomly selected 295008 AK2.09 (#61) Topic does not apply at NMPl. Randomly selected 295002 AK3.02 (#63) Topic does not apply at NMPI for R ES-401 1 /2 1/2 1/2 1/2 1 /2 2/1 2/1 2/1 2/1 NMPI Record of Rejected WAS Form ES-401-4 interpret the following as they apply to HIGH SECONDARY CONTAINMENT AREA RADIATION LEVELS

Equipment operability 295032 / 2.4.30 High Secondary Containment Area Temperature

/ Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.

295029 / 2.2.25 High suppression pool water

/eve/ 2950 12 / 2.4.50 High Drywell temperature

/ Ability to verify system alarm setpoints and operate controls identified in tthe alarm response manual.

295036 /AA2.03 Ability to determine and/or interpret the following as they apply to SECONDARY CONTAINMENT HlGH SUMP/AREA WATER LEVEL: Cause of high water level 500000 / EK3.04 Knowledge of the reasons for the following responses as they apply to HIGH PRIMARY CONTAINMENT HYDROGEN CONCENTRATIONS: Emergency depressurization 259002 / K2.02 Knowledge of electrical power supplies to the following: Feedwater coolant injection (FWCI) initiation logic:

FWCI/HPCI . 21 5003 / K3.05 Knowledge of the effect that a loss or malfunction of the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM will have on following:

APRM: Plant-Specific 300000 / K5.13 Knowledge of the operational implications of the following concepts as they aDDh to the INSTRUMENT AIR kYSTEM: Filters 261000 / K3.05 Knowledge of the effect that a loss or malfunction of the STANDBY GAS TREATMENT SYSTEM will acted EK2.01 (#64) Topic not related to APE for RO. Randomly selected 2.4.18 (#84) Topic not addressed in TS bases. Randomly selected 295007

(#85) Topic tested in operating portion of exam. Random1.y selected 2950 IO (#83) Similar EOP-5 concepts are tested throughout the exam. Randomly selected 295020 AA2.06

(#65) Topic does not apply to NMP 1, due to EOP change. Randomly selected EA2.01 (#3) Topic was oversampled (power supplies)

Randomly selected K1.03 (#6) Topic does not apply at NMPI. Randomly selected K3.02 (#I 0) Oversampled (see

1. 86). Randomly selected K5.01 (#5) Oversampled (see

1. 38). Randomly selected K1.03 ES-401 211 211 2/1 2/1 212 212 212 3 3 3 NMPI Record of Rejected KlAs Form ES-401-4 have on following:

Secondary containment contaminationlradiation levels 21 7000 I K4.05 Knowledge of REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) design feature(s) and/or interlocks which provide for the following: Prevents radioactivity release to auxiliary/reactor building 206000 I K6.08 Knowledge of the effect that a loss or malfunction of the following will have on the HIGH PRESSURE COOLANT INJECTION SYSTEM : Reactor pressure:

BWR-2,3,4 203000 I A3.09 Ability to monitor automatic operations of the RHWLPCI: INJECTION MODE (PLANT SPECIFIC) including:

Emergency generator load sequencing 205000 / 2.4.4 1 Shutdown Cooling 1 Knowledge of the emergency action level thresholds and classifications.

21 9000 I A3.01 Ability to monitor automatic operations of the RHWLPCI: TORUS/SUPPRESSION POOL COOLING MODE including: Valve operation 201 004 I A4.02 Ability to manually operate andlor monitor in the control room: RSCS console switches and indicators:

BWR- 4.5 201002 12.1.27 Reactor manual control system / system purpose 2.4.40 - Knowledge of SRO responsibilities in emergency plan implementation.

2.2.4 - (multi-unit license)

Ability to explain the variations in control board/control room layouts, systems, instrumentation, and procedural actions between units at a facility.

2.3.15 - Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring J equipment, etc. (#7) System does not exist at NMPI. Randomly selected 212000 K4.11 (#I 2) Topic oversampled. Randomly selected K6.03

(#I 8) System does not exist at NMPI. Randomly selected 262001 A3.02 (#88) Topic covered in operating exam. Randomly seiected 2.4.9 (#35) Topic does not apply at NMPI. Randomly selected K5.04

(#36) System does not exist at NMPI. Randomly selected 201 001 A4.03 (#37) Topic does not lend itself to a discriminating question (system function) Randomly selected 2.1 23 (#72) Not an RO level topic. Randomly selected 2.4.32 (#74) Not a multi unit license. Randomly selected 2.2.2 (#98) Topic covered in Admin JPM. Randomly selected 2.4.40 ES-401 NMPI Record of Rejected WA's Form ES-401-4 I 245000 Main Turbine and I Gen AUX., 2.1.30 - Conduct Of Operations:

Ability f0 locate and operate components, (#93) Not an SRO /eve/ topic. Reselected per NRC direction, 20200 7 - 2.2.22 including local controls.

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

SRO Nine Mile Point Unit 1 I Administrative Topic (see Note) Conduct of Operations Conduct of Operations Equipment Control Radiation Control TY Pe Code* N N D P Date of Examination:

October 2008 Operating Test Number: 1 ~~~~ ~ Describe activity to be performed PERFORM A TIME TO BOIL CALCULATION FOR THE SPENT FUEL POOL Given shutdown conditions perform a time to boil calculation N1-ODP-OPS-0108 2.1.37 (4.6) Knowledge of procedures, guidelines, or limitations associated with reactivity management DETERMINE PERSONNEL OVERTIME AVAILABILITY Given a list of personnel and their previous work hours, determine who is available for overtime and why others are not available based on Tech Spec and administrative requirements.

GAP-FFD-02 2.1.5 (3.9) Ability to use procedures related to shift staffing, such as minimum crew requirements, overtime limitations, etc.

REVIEW SURVEILLANCE DATA INCLUDING ACTIONS FOR UNSATISFACTORY CONDITIONS Review and evaluate surveillance acceptance criteria including TS implication for unsatisfactory conditions.

N1 -ST-Q19; Technical Specifications 2.2.1 2 (3.4) Knowledge of surveillance procedures.

2.2.24 (3.8) Ability to analyze the effect of maintenance activities on LCO status. DETERMINE ACTIONS REQUIRED FOR AN INOPERABLE EFFLUENT RADIATION MONITOR Given plant conditions, determine operability of an effluent radiation monitor and apply action statements contained in the station ODCM. (CR NM-2004-976)

ARP HI-4-5, ODCM 2.3.1 1 (4.3) Ability to control radiation releases.

N Emergency Plan SROs. RO applicants require only 4 items unless they topics, when all 5 are required.

CLASSIFY EMERGENCY EVENTS AND COMPLETE NOTIFICATION FACT SHEET Classify emergency events based on plant conditions and complete the appropriate notification form(s). Given further degraded plant conditions, reclassify the emergency event. EPIP-EPP-01 , EPIP-EPP-01 -EAL, EPIP-EPP-20 2.4.40 (4.5)

Knowledge of SRO responsibilities in emergency plan implementation 2.4.41 (4.6) Knowledge of the emergency action level thresholds and classifications

• Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (53 for ROs; 5 4 for SROs & RO retakes) (N)ew or (M)odified from bank (>1) (P)revious 2 exams (51; randomly selected)

ES-301 Administrative Tapics Outline Form ES-301-1 TY Pe Code* D N N D Date of Examination:

October 2008 Operating Test Number:J Describe activity to be performed PERFORM CONTROL SWITCH LINEUP VERIFICATION While performing N1 -PM-DO02 lineup verification, identify system components that are not in the correct lineup N1 -PM-DO02 2.1.29 (4.1) Knowledge of how to conduct system lineups, such as valves, breakers, switches, etc.

DETERMINE PERSONNEL OVERTIME AVAILABILITY Given a list of personnel and their previous work hours, determine who is available for overtime and why others are not available based on Tech Spec and administrative requirements GAP-FFD-02 2.1.5 (3.9) Ability to use procedures related to shift staffing, such as minimum crew requirements, overtime limitations, etc. PERFORM DAILY THERMAL LIMIT SURVEILLANCE Perform the Daily Thermal Limit Surveillance and identify discrepancies N1 -RESP-I, 30 Monicore 2.2.1 2 (3.7) Knowledge of surveillance procedures PERFORM ACTIONS FOR A MEDICAL EMERGENCY WITH AN INJURED, CONTAMINATED PERSON Given a report of a medical emergency with an injured, contaminated person, perform the actions of the Chief Shift Operator Medical Emergency Checklist.

EPI P-EPP-04 2.4.39 (3.9) Knowledge of RO responsibilities in emergency plan' implementation.

~~ ~ 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 (53 for ROs; 5 4 for SROs & RO retakes) (N)ew or (M)odified from bank (51) (P)revious 2 exams (51; randomly selected)

ES-30 1 Control Room/ln-Plant Systems Outline Form ES-30 1 -2 Facility:

Nine Mile Point Unit 1 Date of Examination: October 2008 Exam Level: RO/SRO-VSRO-U Opera ti ng Test No. : 1 Control Room SystemsQ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including I ESF) SRO-U in BOLD - #s S-I ,3,7/P-1,2 Type Code*

Safety Function System / JPM Title

~ D,A,EN,S 1 S-I Initiate Liquid Poison Injection, RWCU Fails to Isolate WA 21 1000 AI .08 3.7/3.8 5 S-2 Transfer Torus Water to the Waste Collector Tank Using Containment Spray Loop 11 1 WA 295029 EA1.03 2.9/3.0 2 S-3 Transfer Load from #I 1 and #I 2 Feedwater Pumps to #I 3 Feedwater Pump, #I 3 Feedwater FCV fails closed WA 259001 A2.07 3.7/3.8 9 S-4 Startup Control Room Ventilation System WA 290003 A4.01 3.2/3.2 6 S-5 EDG 103 S/D - PB 103 Return to Normal Power WA 264000 A4.05 3.6/3.7 Perform RWM Diagnostic

& Rod Block Tests WA 201006 A4.01 thru A4.06, 2.9/2.9 to 3.3/3.4 S-6 7 4 S-7 Remove the Generator from the Grid and Perform Emergency Governor Trip Test KIA 245000 A4.02 (3.1/2.9), A4.06 (2.712.6)

I 3 S-8 RO ONLY Alternate RPV Blowdown Through Emergency Condenser Vents to Torus K/A 207000 AI .05 (4.0/4.2), A4.05 (3.5/3.7), A4.07 (4.2/4.3)

I implant Systems@ (3 for RO; 3 or 2 for SRO-U) P-I Air Start the Diesel Fire Pump WA 286000 A3.01 3.4/3.4 8 P-2 Initiation of Emergency Condensers from Remote Shutdown Panel 11 WA 29501 6 AA1.09 4.0/4.0 3 P-3 Place UPS 162A in Standby from Shutdown Condition and Transfer to Supply RPS 11 WA 212000 AI .04 (2.8/3.0), AI .05 (2.6/2.7) 6 ES-301 Control RoomAn-Plant Systems Outline Form ES-30 1 -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. @ 1 Type Codes Criteria for RO I SRO-I / SRO-U (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power I Shutdown (N)ew or (M)odified from bank including 1(A) (P)revious 2 exams (WCA (S)imulator 4-6 I 4-6 I 2-3 191sa114 27 I21 I2 1 21 I21 I21 22122121 I 3 I I 3 I I 2 (randomly selected) 21121 121 - I - I 21 (control room system)

ES-301 Control Room/ln-Plant Systems Outline Form ES-301-2 2007 NRC Examination Summary Description of JPMs s-I s-2 s-3 s-4 s-5 S-6 5-7 5-8 P-I P-2 P-3 This is an alternate path bank JPM in the Reactivity Control Safety Function area. The applicant will inject Liquid Poison N1-OP-12 and Reactor Water Cleanup will fail to isolate requiring manual actions. This is a bank JPM in the Containment Integrity Safety Function area.

The applicant will transfer torus water to the Waste Collector Tank using Containment Spray Loop 11 1 IAW N1-EOP-1, Att.15. This is a new alternate path JPM in the Rx Water Inventory Control Safety Function area. The applicant will transfer load from

1. I 1 and #I2 Feedwater Pumps to #I3 Feedwater Pump IAW NI- OP-I 6 and the #I 3 pump flow control valve will malfunction requiring manual actions to control vessel level. This is a bank JPM in the Radioactivity Release Safety Function area. The applicant will startup Control Room Ventilation IAW N1-OP-49. This is a new alternate path JPM in the Electrical Safety Function area. The applicant will shutdown Emergency Diesel Generator 103 and return Powerboard 103 to Normal Power IAW N1-OP-45, section G.2.0. The Emergency Diesel Generator will fail to stop after a cooldown period, requiring a manual trip to be performed.

This is a new JPM in the Instrumentation Safety Function area. The applicant will perform Rod Worth Minimizer Post Maintenance Tests IAW Nl-ST-V3, Section 8.2 thru 8.4. This is a new JPM in the Heat Removal Safety Function area.

The applicant will perform the Emergency Governor Trip Test and Remove the Generator from the Grid IAW N1-OP-31, Section G.2.0 and Nl-PM-V7, Section 8.1. This is a bank JPM in the Reactor Pressure Control Safety Function area. The applicant will perform an Alternate RPV Blowdown Through the Emergency Condenser Vents to Torus IAW N1-EOP-1, Att.14. This is a bank JPM in the Plant Service Systems Safety Function area.

The applicant will perform an Air Start of the Diesel Fire Pump IAW NI-OP-21A, Section H.4.4. This is an alternate path bank JPM in the Reactor Pressure Control Safety Function area.

The applicant will perform an Initiation of ECs from Remote Shutdown Panel 11 IAW N1-SOP-21.2. Additional actions will be required to control the reactor pressure.

This is a modified bank JPM in the Electrical Safety Function area. The applicant will place UPS 162A in Standby from a Shutdown Condition and Transfer the supply to RPS 11 IAW N1-OP-40, Section E.l .O.

Appendix D Scenario Outline Form ES-D-1 Facility:

Nine Mile Point 1 Examiners: Operators: Initial Conditions:

Simulator IC 171 1. Reactor Power approximately 4% Turnover:

1. The crew is directed to shutdown the reactor by inserting control rods

2. Crew is directed to perform N1-OP-09, N2 lnerting and H2-02 Monitoring Systems step G.l to de-inert the Primary Containment with Rx Coolant Temp >212"F Scenario No.: NRC-01 Op-Test No.: October 2008 Event Malf. No. Event Event Description No. I I (7%) I I Radiation, requires a reactor scram (SOP-25.2)

I * (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor NRC Scenario 1 October 2008 Facility:

Nine Mile Point 1 Scenario No.: NRC-01 TARGET QUANTITATIVE ATTRIBUTES I ACTUAL 1. Total malfunctions (5-8) Events 2, 4, 5, 6, 7, 9 I (PER SCENARIO; SEE SECTION D.5.d) 6 2. Event 9 Malfunctions after EOP entry (1 -2) 1 3. Abnormal events (2-4) Events 4, 5, 6,7 4. Major transients (1-2) Event 8 5. EOPs enteredhequiring substantive actions (1 -2) Event 8 (EOP-6) 4 1 1 6. EOP contingencies requiring substantive Events 9 (EOP-8) actions (0-2)

7. Critical tasks (2-3) CRITICAL TASK DESCRIPTIONS:

CT-1.0 Given a fuel failure, the crew will insert a manual reactor scram as Main Steam Line radiation levels rise. CT-2.0 Given unisolable primary system leak, indications of fuel failure and rising off-site release rates approaching the General Emergency level, the crew will perform an RPV Blowdown.

Op-Test No.: October 2008 NRC Scenario 1 October 2008 SCENARIO

SUMMARY

Length: 90 minutes Initial Power Level: 4% with plant shutdown in progress Mitigating Strategy Code:

RR4, fuel leak with a failure of EC tubes and EC fails to isolate, requires RPV Blowdown to stop release The crew assumes the shift with the plant being shutdown.

The crew is directed to de-inert the containment in accordance with N 1 -0P-9, N2 lnerting and H2-02 Monitoring Systems. When drywell pressure is lowered to 0 psig, the operator will secure the lineup, but one of the containment isolation valves will fail to fully close. This will require entry into Technical Specifications and ensuring a second valve in the line is isolated. Then the crew will continue the shutdown by inserting control rods. Next Reactor Building Radiation Monitor 12 will fail upscale causing a trip of RBVS and a start of RBEVS. Additionally there will be a failure of the Reactor Building to isolate. The crew must isolate the Reactor Building to restore Secondary Containment and the SRO must address Technical Specifications.

When these actions are complete, both seals on the 11 Recirculation Pump will fail requiring the crew to shutdown and isolate the pump. Following the loss of the Recirculation Pump, a fuel failure will cause offgas and main steam line radiation levels to rise, requiring a reactor scram and vessel isolation. Multiple control rods will fail to fully insert during the scram requiring the crew to enter N1-SOP-1 and take alternate actions to insert the control rods. The rods are inserted using RMCS. Following the scram, the crew will diagnose an Emergency Condenser tube leak. They will try to isolate the affected EC but the isolation valves will fail to fully close.

Rising off site radiation levels will require an RPV blowdown before General Emergency levels are reached.

Major Procedures: N1-SOP-1.2, N1-SOP-25.2, N1-SOP-1 .I, N1-SOP-1, N1-EOP-2, N1-EOP-6, and N1-EOP-8 EAL Classification:

Site Area Emergency, EALs 3.4.1, 5.1.3 and 5.2.4 Termination Criteria: RPV Blowdown in progress, RPV water level controlled in assigned band NRC Scenario 1 October 2008 Appendix D Scenario Outline Form ES-D-1 Facility: Nine Mile Point 1 Examiners:

Operators:

Initial Conditions:

Simulator IC 172 Scenario No.: NRC-02 Op-Test No.: October 2008 1. Reactor Power approximately 90% 2. Four Recirculation Loops in service Turnover: 1. Recirc Pump 15 MG set has been repaired and should be returned to service. 2. After starting Recirc Pump 15 MG set operate it for one hour while maintenance takes readings before returning to 100% power.

Event Malf. No. Event Event DescriDtion No. requiring shifting to the alternate FCV (SOP-5.1) (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor NRC Scenario 2 October 2008 Facility:

Nine Mile Point 1 Scenario No.: NRC-02 TARGET QUANTITATIVE ATTRIBUTES (PER SCENARIO; SEE SECTION D.5.d) 1. Total malfunctions (5-8)

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

CT-1.0 Given an un-isolable RWCU leak outside primary containment and one general area temperature above the maximum safe limit, the crew will insert a manual reactor scram. CT-2.0 Given a failure of RPS to de- energize when a scram is required, the crew will insert control rods by initiating manual Alternate Rod Insertion (ARI). CT-3.0 Given an un-isolable RWCU leak outside primary containment and two general area temperatures above the maximum safe limit, the crew will perform an RPV Blowdown.

ACTUAL ATTRIBUTES 5 1 4 1 3 OD-Test No.:

October 2008 NRC Scenario 2 October 2008 SCENARIO

SUMMARY

Length: 90 minutes Initial Power Level: 90%, 4 Loop Operation Mitigating Strategy Code:

SCI, un-isolable primary system leak in the Secondary Containment, RPV Blowdown required The crew assumes the shift with the plant operating at 90% power and four recirculation loops in service. Immediately after assuming the shift the crew will be directed to restore Recirculation Pump 15 to service and return to full power. The crew will assess plant conditions and lower power with Recirculation Flow until flow is less than 50 Mlbm/hr. They will then return Recirculation Pump 15 to service.

After the crew has placed the pump in service, the Main Generator Auto Voltage Regulator will fail. The crew will diagnose the failure and take manual control of generator voltage and restore the correct generator output. When a normal generator output is established, the Control Rod Drive Flow Control Valve fails closed, requiring shifting to the alternate FCV. After CRD flow is returned to normal, a loss of power to Power Board 11 occurs. The SRO will address Technical Specifications.

A Reactor Water Cleanup system line break will occur in the Secondary Containment downstream of the Supply Isolation Valves. Reactor Water Cleanup will fail to isolate on high area temperature. The crew will attempt to isolate the system, but the valves will fail to fully close. This break will require a scram and RPV blowdown due to exceeding the Maximum Safe Value for general area temperatures.

When the Mode Switch is placed in SHUTDOWN and/or the Reactor Trip pushbuttons on the E Panel are pushed the reactor will NOT scram. ARI must be manually initiated to scram the control rods. Major Procedures:

N1-SOP-1, N1-SOP-1.1, N1-SOP-1.3, N1-SOP-5.1, N1-SOP-30.1, N1-EOP-2, N1-EOP-3, N1-EOP-5, and N1-EOP-8 EAL Classification:

Site Area Emergency, EALs 3.4.1, 4.1 .I Termination Criteria: All control rods are in, RPV Blowdown in progress, RPV water level controlled in assigned band NRC Scenario 2 October 2008 Appendix D Scenario Outline Form ES-D-1 I Facility:

Nine Mile Point 1 Scenario No.: NRC-03 Op-Test No.: October 2008 -1 Examiners: Operators: Initial Conditions: Simulator IC 173 1. Reactor Power approximately 100% (CRD Pump 12 must be in service) Turnover:

1. Turbine Surveillance Testing, NI-PM-Q7, to be performed

2. Feed Pump 12 is out of service because of a burned out motor Event Descridion

! I I Override I C(SR0) 1 pump to auto start TS fSRO1 .- I I cso3c I I Spray 121 fails to start I I .i * (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor NRC Scenario 3 October 2008 TARGET QUANTITATIVE ATTRIBUTES (PER SCENARIO; SEE SECTION D.5.d) 1. Total malfunctions (5-8)

Events 2, 3, 4, 5, 6, 8, 9 ACTUAL ATTRIBUTES 7 2. Malfunctions after EOP entry (1 -2) Events 8,9 3. Abnormal events (2-4) Events 2,3,4, 5, 6,8 2 6 4. Major transients (1 -2) Event 7 I 6. EOP contingencies requiring substantive Events 7,9 (EOP-2 Alternate Level Leg, actions (0-2) EOP-8) 5. EOPs enteredhequiring substantive actions (1 -2) Events 7,9 (EOP-2, EOP-4) 2 2 7. Critical tasks (2-3) CRITICAL TASK DESCRIPTIONS:

CT-1.0 Given a LOCA with a loss of high pressure injection, the crew will execute N1-EOP-8, RPV Blowdown when RPV water level drops below -84 inches. CT-2.0 Given a LOCA with a loss of high pressure injection and Core Spray, the crew will inject to the RPV with Condensate and Feedwater Booster pumps. CT-3.0 Given a LOCA in the Drywell, the crew will initiate Containment Sprays to prevent exceeding PSP. NRC Scenario 3 3 October 2008 SCENARIO

SUMMARY

Length: 90 minutes Initial Power Level: Approximately 1 OO%, above 100%

rodline Mitigating Strategy Code: RL2, Small LOCA, RPV Blowdown required to permit injection with low pressure systems to recover RPV water level above TAF SUM MARY The crew assumes the shift with the plant at 100% power with Feedwater Pump 12 under clearance for maintenance. The crew will perform N1 -PM-Q7, Turbine Thrust Bearing Test from the Control Room.

Next, APRM 13 fails. The crew will bypass the APRM and reset the half scram. Next, Powerboard 103 trips on fault. The crew will take action to secure EDG 103 and attempt to restore Powerboard 176. Powerboard 103 and Powerboard 17B are both faulted and are not restored. The trip of CRD Pump 12 (PB 178) will require starting CRD Pump 11 and the SRO must address Technical Specifications. When the necessary steps for the loss of Powerboard 103 are completed, Feedwater Booster Pump 11 will trip with a failure of the standby pump to start. The standby pump can be manually started. The SRO must again address Technical Specifications. When the standby Feedwater Booster Pump is manually started, the Master Feedwater Controller will fail as-is. RPV water level will slowly deviate from the set level. The crew must diagnose the failure and the BOP operator will be required to take manual control of RPV level.

With RPV water level in manual control, Feedwater Pump 11 will trip because of delayed effects from the earlier Feedwater Booster Pump trip. This will require an entry into N1-SOP-1 .I, Emergency Power Reduction to lower power to within the capacity of Feedwater Pump

13. While troubleshooting the electrical faults and troubles with the Feedwater system, the crew recognizes a coolant leak in the containment.

Drywell pressure and temperature rise, requiring the crew to insert a manual SCRAM on rising drywell pressure.

When the turbine trips, Powerboards 1 1 and 12 fail to automatically transfer. This results in a loss of feedwater, condensate, circulating water and other loads. Operators are able to restore these power boards. RPV water level continues to drop with only one liquid poison pump and CRD pump 1 I available for injection. The crew will determine they cannot maintain level above -109" and enter N1-EOP-8, RPV Blowdown.

While blowing down the crew must diagnose that the inboard IV for Core Spray 11 1 fails to open and Core Spray pump 121 fails to start. With Core Spray unavailable for injection, the crew will inject with the feedwater booster pumps using N1-EOP-1, Att 25 or 26. Major Procedures:

N1-SOP-1, N1-SOP-1.1, N1-SOP-5.1, N1-SOP-16.1, N1-SOP-30.1, NI- SOP-30.2, N1-EOP-1, N1-EOP-2, N1-EOP-4, N1-EOP-8 Termination Criteria: RPV Blowdown in progress, RPV water level above TAF and controlled in assigned band, containment pressure controlled in accordance with N1- EOP-1 Att 17 EAL Classification: Alert, EAL 3.1.1 NRC Scenario 3 October 2008 Appendix D Scenario Outline Form ES-D-1 Facility:

Nine Mile Point 1 Examiners: Operators:

Initial Conditions:

Simulator IC 174 1. Reactor Power approximately 90% for a rod pattern adjustment Turnover:

1. Maintenance completed work on TBCLC pump 12 2. APRM 13 bypassed due to failed power supply 3. Recirc Pump 14 OOC due to high vibrations Scenario No.: NRC-04 Op-Test No.: October 2008 Event Malf. No.

Event Event Type* Description No. 4 I RMlA I TS (SRO) 1 Main Steam Line Radiation Monitor 11 1 fails 1 RR92 I TS(SR0) I (same instrument line) fail low, requires manual FWLC (SOP-I 6.1 ) emergency power reduction required (SOP-I 8.1 ) 9 FW24 C (BOP) The crew will be unable to re-inject with feedwater/

FW28 c (SRO) condensate because the valves they used to terminate and prevent will fail closed (EOP-8) (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Overrides NRC Scenario 4 October 2008 Facilitv:

Nine Mile Point I Scenario No.: NRC-04 OD-Test No.: October 2008 TARGET QUANTITATIVE ATTRIBUTES (PER SCENARIO; SEE SECTION D.5.d)

1. Total malfunctions (5-8)

Events 3, 4, 5, 6, 8, 9 - 2. Malfunctions after EOP entry (1-2)

Events 8, 9 3. Abnormal events (2-4) Event 3, 4, 5, 6 4. Major transients (1 -2) Event 7 5. EOPs enteredhequiring substantive actions (1 -2) Event 8 (EOP-4) 6. EOP contingencies requiring substantive Events 7,9 (EOP-3, EOP-8) actions (0-2) 7. Critical tasks (2-3) CRITICAL TASK DESCRIPTIONS:

CT-1 .O Given a failure of the reactor to scram with power above 6% or unknown and RPV water level above -41 inches, the crew will terminate and prevent all injection except boron and CRD. CT-2.0 Given a failure of the reactor to scram with RPV water level unable to be restored and maintained above

-109 inches with CondensatelFeedwater and CRD, the crew will perform an RPV Blowdown and re-establish injection with Core Spray. CT-3.0 Given a LOCA in the Drywell, the crew will initiate Containment Sprays to Prevent exceedinn PSP. ACTUAL ATTRIBUTES 6 2 4 1 1 NRC Scenario 4 October 2008 SCENARIO

SUMMARY

Length: 90 minutes Initial Power Level:

Approximately 90%, 4 loop operation Mitigating Strategy Code:

AT3, high power ATWS with small LOCA, Blowdown required, re- inject with Core Spray The scenario begins with the crew performing a control rod pattern adjustment.

Next, the crew will be directed to return TBCLC Pump 12 to service and secure TBCLC Pump 11. Next the crew must respond to high D/P across one of the Service Water Pump Discharge Strainers.

This will require placing another Service Water Pump in service. Once the standby Service Water Pump has been started, Main Steam Line Radiation Monitor 11 1 will become inoperable.

The SRO will determine the Technical Specification implications.

When this is complete, an RPS pressure transmitter will fail low, followed closely by the in- service feedwater system pressure transmitter also failing low. The crew will be required to shift to manual feedwater level control.

The crew may then shift reactor pressure/level columns and return to automatic feedwater level control. Technical Specifications must be addressed due to the RPS pressure transmitter failure. Next the intake structure traveling screens clog causing high D/Ps. This will eventually result in a low level in the intake structure with the subsequent tripping of the Circulating Water pumps. This will require entering N1-SOP-18.1, Service Water Failure/Low Intake Level. As intake level continues to lower, the crew will insert a manual scram. When the scram occurs the control rods will not insert. This ATWS is complicated by the total loss of the normal heat sinks. Additionally, following the ATWS, a Recirculation Line break will cause RPV water level to lower, requiring the crew to re-establish injection. When the crew attempts to re-establish Feedwater flow, the Feedwater isolation valves will not re-open. When it is determined that RPV water level cannot be restored and maintained above -109 inches, the crew will perform an RPV Blowdown, and re-inject with Core Spray. Major Procedures: N1-SOP-1.1, N1-SOP-16.1, N1-SOP-18.1, N1-EOP-1, N1-EOP-2, N 1 -EOP-3, N 1 -EOP-3.1, N 1 -EOP-4, N 1 -EOP-8 EAL Classification:

Site Area Emergency, EAL 2.2.2 Termination Criteria: RPV Blowdown in progress, RPV water level above -109 inches and controlled in assigned band, containment pressure controlled in accordance with N1-EOP-1 Att I NRC Scenario 4 October 2008