ML083230507

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Final Outlines (Folder 3)
ML083230507
Person / Time
Site: Nine Mile Point Constellation icon.png
Issue date: 10/01/2008
From:
Nine Mile Point
To: D'Antonio J
Operations Branch I
Hansell S
Shared Package
ML081060454 List:
References
TAC U01690
Download: ML083230507 (36)
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ES-401 NMPI Written Examination Outline Form ES-401-1 Note 1. 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).

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 +from I 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/evolutionswithin 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 regardingelimination of inappropriateWA statements.
4. Select topics from as many systems and evolutionsas possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.
5. Absent a plant specific priority, only those KAs having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.
6. Select SRO topics for Tiers 1 and 2 from the shaded systems and KIA categories.
7. The generic (G) WASin 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 W A S
8. On the following pages, enter the WA numbers, a brief descriptionof 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 # I does not apply). Use duplicate pages for RO and SRO-only exams.
9. For Tier 3, select topics from Section 2 of the WA Catalog, and enter the W A numbers, descriptions, IRs, and point totals (#)

. . on Form ES-401-3. Limit SRO selections to WASthat are linked to 10CFR55.43 ES-401 Form ES-401-1 NMPI Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 EAPE#/NameSafetyFunction K1 K2 K3 AI A2 G K/A Topic(s) Imp. Q#

295031 Reactor Low Water e followmg as they 4.2 76 Level / 2 to REACTOR LOW WATER 2950 16 Control Room 3.5 77 Abandonment / 7 295028 High Drywell 4.2 78 Temperature / 5 295006 SCRAM 1 I 4.1 79 mdcations, or response 295001 Partial or Complete AbMy to pr/ont/zeand interpret the Loss of Forced Core Flow 4.3 80 siqnificance of each annunciator or Circulation / 1 & 4 295003 Partial or Complete 4.3 81 Loss of AC / 6 4.1 82 Rate / 9 EKI .07 - Knowledge of the operational implicationsof the following 295037 SCRAM Conditions concepts as they apply to SCRAM Present and Reactor Power X CONDITION PRESENT AND 3.4 39 Above APRM Downscaleor REACTOR POWER ABOVE Unknown / I APRM DOWNSCALEOR UNKNOWN: Shutdown margin AK2.04 - Knowledge of the interrelationsbetween MAIN 295005 Main Turbine Generator X TURBINE GENERATOR TRIP and 3.3 40 Trip I 3 the following: Main generator concepts as they apply to PARTIAL OR COMPLETE LOSS 3.3 41 OF D.C. POWER Loss of breaker 295024 High Drywell Pressure /

5 I lXl 3.3 42 295003 Partial or Complete 3.2 43 Loss of AC / 6 ES-401 Form ES-401-1 NMP1 Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 EAPE#INameSafetyFunction K1 K2 K3 A1 A2 G KIA Topic(s) Imp. Q#

295026 Suppression Pool High Water Temp. I 5 295006 SCRAM I1 I295030 Low Suppression Pool Water Level I 5 I EK3.06 - Knowledge of the reasons for 295028 High Drywell Temperature 15 I the following responses as they 295025 High Reactor PressureI 3

I monitor the following as they apply to HIGH REACTOR PRESSURE:

Condenser: Plant-1 I 4.5 48 295016 Control Room Abandonment 17 295038 High Off-site Release Rate I 9 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 I295021 Loss of Shutdown Cooling I 4 logic used to assess the status of control, core cooling and heat removal, reactor coolant system Level I 2 ES-401 Form ES-401-1 NMP1 Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 EAPE#INarneSafetyFunction K1 K2 K3 AI A2 G WA Topic(s) Imp. Q#

295019 Partial or Complete Loss of Inst. Air I 8 295023 Refueling Acc I 8 295018 Partial or Complete Loss of ccw I 8 WA Category Totals: 2 ES-401 Form ES-401-1 NMPI Written Examination Outline Emergency and Abnormal Plant Evolutions -Tier 1 Group 2 EAPE#INameSafetyFunction K1 K2 K3 AI A2 G KIA Topic(s) Imp. Q#

I 295020 Inadvertent Coni.

Isolation / 5 8. 7 295007 High Reactor 3.4 83 4.2 84 Pressure / 3 Ability to verify system alarm 295010 High Drywell setpoints and operate controls 4.0 85 Pressure / 5 identified in the alarm response I

I 295015 Incomplete SCRAM I 1 l x concepts as they apply to INCOMPLETESCRAM : Reactor 3.8 59 295008 High Reactor 31 Water Level I 2 60 295002 Loss of Main 3.4 61 Condenser Vacuum / 3 monitor the following as they apply 295007 High Reactor to HIGH REACTOR PRESSURE : 3.7 62 Pressure 13 Reactorlturbinepressure regulating 295033 High Secondary SECONDARY CONTAINMENT Containment Area 3.8 63 AREA RADIATION LEVELS and Radiation Levels I 9 wing: Area Rad Monitoring 295032 High Secondary Containment Area 3.3 64 Temperature I 5 500000 High CTMT 3.1 65 Hydrogen Conc. I 5 KIA Category Totals: 1 ES-401 Form ES-401-1 NMPI Written Examination Outline Plant Systems - Tier 2 Group 1 K K K K K K A N A A G Imp Q#

System # I Name 3 4 1 2 3 4 5 6 1 300000 Instrument Air 2.8 86 218000ADS 4.2 87 abnormal conditions or operations: ADS failure to Plan: Knowledge of low 205000 Shutdown power/shutdown 4.6 88 Cooling implications in accident

. LOCA or loss of RH 207000 Isolation (Emergency) 4.7 89 Condenser --

263000 DC Electrical 2.9 90 Distribution K1.05 - Knowledge of the physical connections and/or cause- effect relationships between LOW 209001 LPCS 3.7 1 PRESSURE CORE SPRAY SYSTEM and the 239002 SRVs 3.6 2 ES-401 Form ES-401-1 NMPI Written Examination Outline Plant Systems - Tier 2 Group 1 K K K K K K A A 2 System # / Name A A G Imp Q#

1 2 3 4 5 6 1 3 4 259002 Reactor Water Level Control 3.8 3 218000 ADS 3.1 4 physical connections and/or relationships 261000 SGTS between STANDBY GAS 2.9 5 TREATMENT SYSTEM 3.6 6 I I control 1 K4.11 - Knowledqe of 3.3 7 ELECTRICAL DISTRIBUTION design feature(s) and/or interlocks 2.9 8 which provide for the 205000 Shutdown Cooling 2.8 9 300000 InstrumentAir 2.5 10 ES-401 Form ES-401-1 NMPI Written Examination Outline Plant Systems - Tier 2 Group I System # I Name K K K K K K A N A A G Imp Q#

1 2 3 4 5 6 1 3 4 262002 UPS (ACIDC) 2.7 11 that a loss or malfunction of the following will have on 206000 HPCl 2.9 12 the HIGH PRESSURE COOLANT INJECTION 263000 DC Electrical 2.5 13 Distribution 21 1000 SLC 3.6 14 on those predictions, use 400000 Component procedures to correct, 2.8 15 Cooling Water control, or mitigate the consequences of those abnormal oDeration:

I I Highllow surge tank level I A2.06 - Abilitv to (a) predict the 207000 Isolation (Emergency) 3.3 16 Condenser 223002 PClSlNuclear 3.4 17 Steam Supply Shutoff

-I "

ES-401 Form ES-401-1 NMPI Written Examination Outline Plant Systems - Tier 2 Group 1 I System # I Name K K K K K A A 2 2 3 4 5 6 1 A

3 A

4 G Imp Q#

262001 AC Electrical 3.2 18 Distribution 215004 Source Range 3.4 19 Monitor 215005 APRM ILPRM 3.2 20 212000 RPS 4.2 21 264000 EDGs 4.4 22 215003 IRM 2.5 23 207000 Isolation (Emergency) 4.4 24 Condenser 400000 Component 2.7 25 Cooling Water 223002 PCIS/Nuclear 3.2 26 Steam Supply Shutoff I-WA Category Totals: I 2615 ES-401 Form ES-401-1 NMPI Written Examination Outline Plant Systems - Tier 2 Group 2 K K K K K K A M A A G Imp.

Q System # I Name 1 2 3 4 5 6 1 3 4 #

impacts of the following on the FlRE PROTECTlON SYSTEM ;and (b) based on those predictions, use 286000 Fire Protection 2.9 91 c

I 215001 Traversing In-core Probe 202001 Recirculation System 245000 Main Turbine Gen. I Aux.

physical connections andlor cause- effect relationships between MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS 4.6 4.7 2.6 92 93 27 and the following:

I 239001 Main and Reheat Steam 3.2 28 I 286000 Fire Protection 3.6 29 I

204000 RWCU 2.9 30 reactor water to various I

operational implications of the following concepts as 201001 CRD Hydraulic they apply to CONTROL 2.5 31 ROD DRIVE HYDRAULIC SYSTEM : Solenoid 202001 Recirculation ES-401 Form ES-401-1 NMPI Written Examination Outline Plant Systems - Tier 2 Group 2 System # / Name K A A 2 3 4 G Imp.

6 1 271000 Off-gas 2.7 33 the PLANT VENTILATTON SYSTEMS ; and (b) based on those predictions, use 288000 Plant procedures to correct, 2.6 34 Ventilation control, or mitigate the consequences of those abnormal conditions or operations: Extreme outside weather conditions:

219000 RHWLPCI:

ToruslPool Cooling 2.9 35 Mode 201001 CRD Hydraulic 2.9 36 Ability to perfom- specific system and integrated 201002 RMCS 4.3 37 plant procedures during all modes of plant operation.

K3.01 - Knowledge of the effect that a loss or malfunction 290001 Secondary of the SECONDARY 4.0 38 CTMT CONTAINMENT will have on following: Off-site radioactive release rates

--I-KIA Category Totals: Group Point Total: I 1213 NMPI Generic Knowledge and Abilities Outline ES-401 Form ES-401-3 (Tier 3)

NMPI Date: October 2008 I RO SRO-On1y KIA # Topic IR Q# IR Q#

2' l3 Knowledge of facility requirements for controlling vital / controlled access.

3.2 94 Ability to interpret and execute procedure

2. 4.6 99 steps.

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

Ability to manipulate the console controls as required to operate the facility between 2'2'2 4.6 74 shutdown and designated power levels.

duties such as response to radiation monitor Ability to comply with radiation work permit 2.3.7 requirements during normal or abnormal 3.5 70 conditions.

NMPI Generic Knowledge and Abilities Outline ES-401 Form ES-401-3 (Tier 3)

T Knowledge of Radiological Safety Procedures pertaining to licensed operator duties, such as response to radiation monitor 2.3.13 alarms, containment entry requirements, fuel handling responsibilities, access to locked high radiation areas, aligning filters, etc.

97 2-4*40 I emergency plan implementation. 4.5 98 4.

Emergency Procedures /

Plan Knowledge of operator response to loss of all 3.6 I 72 Subtotal - 2 Tier 3 Point Tok 7 ES-401 NMPI Record of Rejected WAS Form ES-401-4 Tier / Group Randomly Selected WA Reason for Rejection 295005 I AKI .01 Knowledge of the operational implications of the following concepts as they (#40)Topic oversampled (see # 62) Randomly 111 apply to MAIN TURBINE selected AK 2.04 GENERATOR TRIP :

Pressure effects on reactor power.

295004 I AKI .01 Knowledge of the operational implications of the following (#41) Topic does not apply to NMPl. Randomly 111 concepts as they apply to PARTIAL OR COMPLETE selected AKI .05 LOSS OF D.C. POWER:

Automatic load sheeding 295024 I EK2.17 Knowledge of the interrelations between HIGH (M2) Topic does not apply to NMPI. Randomly 111 DRYWELL PRESSURE and selected EK2.18 the following: Aux Bldg isolation logic 295006 I AK3.03 Knowledge of the reasons for (#45)Generic Fundamental Topic. Randomly selected 111 the following responses as they apply to SCRAM : AA1.02 Reactor pressure response 295030 I EK3.02 Knowledge of the reasons for the following responses as (#46) Topic does not apply at NMPI. Randomly 111 they apply to LOW SUPPRESSION POOL selected EK3.01 WATER LEVEL: HPCl operation 295025 I EA1.04 Ability to operate and/or (#48) Topic does not apply to NMPI. Randomly 111 monitor the following as they amlv to HIGH REACTOR selected EA1.06 PRE'SSURE: HPCI 295019 I 2.4.47 Partial or Complete Loss of Inst. Air / Ability to diagnose and recognize trends in an (#56) Topic not related to EPE. Randomly selected 111 accurate and timely manner I 2.4.49 utilizing the appropriate control room reference material.

295017 I AK2.12 Knowledge of the interrelations between HIGH (#60) Oversampled (see #38). Randomly selected 1I 2 OFF-SITE RELEASE RATE 295008 AK2.09 and the following: Standby gas treatmenVFRVS 295009 I AK3.01 Knowledge of the reasons for the following responses (#61) Topic does not apply at NMPl. Randomly 112 as they apply to LOW REACTOR WATER LEVEL : selected 295002 AK3.02 Recirculation pump run back:

Plant-Specific 112 295033 I EA2.02 (#63) Topic does not apply at NMPI for RO. Randomly Ability to determine and/or ES-401 NMPI Record of Rejected WAS Form ES-401-4 interpret the following as a c t e d EK2.01 they apply to HIGH SECONDARY CONTAINMENTAREA RADIATION LEVELS :

Equipment operability 295032 / 2.4.30 High Secondary Containment Area Temperature /

Knowledge of events related to system operation/status (#64) Topic not related to APE for RO. Randomly 1/ 2 that must be reported to selected 2.4.18 internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.

295029 / 2.2.25 (#84) Topic not addressed in TS bases. Randomly 1/2 High suppression pool water

/eve/ selected 295007 2950 12 / 2.4.50 High Drywell temperature /

Ability to verify system alarm (#85) Topic tested in operating portion of exam.

1/2 setpoints and operate controls Random1.y selected 2950 IO identified in tthe alarm response manual.

295036 /AA2.03 Ability to determine and/or interpret the following as (#83) Similar EOP-5 concepts are tested throughout 1/2 they apply to SECONDARY CONTAINMENT HlGH the exam. Randomly selected 295020 AA2.06 SUMP/AREA WATER LEVEL:

Cause of high water level 500000 / EK3.04 Knowledge of the reasons for the following responses as (#65) Topic does not apply to NMP 1, due to EOP 1/ 2 they apply to HIGH PRIMARY CONTAINMENT HYDROGEN change. Randomly selected EA2.01 CONCENTRATIONS:

Emergency depressurization 259002 / K2.02 Knowledge of electrical power supplies to the (#3) Topic was oversampled (power supplies) 2/1 following: Feedwater coolant injection (FWCI)

Randomly selected K1.03 initiation logic:

FWCI/HPCI .

215003 / K3.05 Knowledge of the effect that a loss or malfunctionof (#6) Topic does not apply at NMPI. Randomly 2/1 the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM will selected K3.02 have on following: APRM:

Plant-Specific 300000 / K5.13 Knowledge of the operational implications of the (#I0) Oversampled (see #86). Randomly selected 2/1 following concepts as they K5.01 aDDh to the INSTRUMENT AIR kYSTEM: Filters 261000 / K3.05 Knowledge of the effect that a (#5)Oversampled (see #38). Randomly selected 2/1 loss or malfunction of the STANDBY GAS K1.03 TREATMENT SYSTEM will ES-401 NMPI Record of Rejected KlAs Form ES-401-4 have on following: Secondary containment contaminationlradiationlevels 217000 I K4.05 Knowledge of REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) design (#7) System does not exist at NMPI. Randomly 211 feature(s) and/or interlocks which provide for the selected 212000 K4.11 following: Prevents radioactivity release to auxiliary/reactorbuilding 206000 I K6.08 Knowledge of the effect that a loss or malfunctionof 211 the following will have on the (#I 2) Topic oversampled. Randomly selected K6.03 HIGH PRESSURE COOLANT INJECTION SYSTEM :

Reactor pressure: BWR-2,3,4 203000 I A3.09 Ability to monitor automatic operations of the (#I 8) System does not exist at NMPI. Randomly 2/1 RHWLPCI: INJECTION MODE (PLANT SPECIFIC) selected 262001 A3.02 including: Emergency generator load sequencing 205000 / 2.4.4 1 Shutdown Cooling 1 (#88)Topic covered in operating exam. Randomly 2/1 Knowledge of the emergency action level thresholds and seiected 2.4.9 classifications.

219000 I A3.01 Ability to monitor automatic operations of the (#35) Topic does not apply at NMPI. Randomly 212 RHWLPCI:

selected K5.04 TORUS/SUPPRESSION POOL COOLING MODE including: Valve operation 201004 I A4.02 Ability to manually operate andlor monitor in the control (#36) System does not exist at NMPI. Randomly 212 room: RSCS console selected 201001 A4.03 switches and indicators: BWR-4.5 201002 12.1.27 (#37) Topic does not lend itself to a discriminating 212 Reactor manual control system / system purpose question (system function) Randomly selected 2.12 3 2.4.40 Knowledge of SRO 3 responsibilitiesin emergency (#72) Not an RO level topic. Randomly selected 2.4.32 plan implementation.

2.2.4 (multi-unit license)

Ability to explain the variations in control board/controlroom 3 layouts, systems, (#74) Not a multi unit license. Randomly selected 2.2.2 instrumentation,and procedural actions between units at a facility.

2.3.15 Knowledge of radiation monitoring systems, such as fixed radiation (#98) Topic covered in Admin JPM. Randomly 3 monitors and alarms, portable survey instruments, personnel selected 2.4.40 monitoring J equipment, etc.

ES-401 NMPI Record of Rejected WAs Form ES-401-4 I 245000 Main Turbine and I Gen AUX., 2.1.30 - Conduct (#93) Not an SRO /eve/ topic. Reselected per Operations: Ability f0 locate Of and operate components, NRC direction, 20200 7 - 2.2.22 including local controls.

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Nine Mile Point Unit 1 Date of Examination: October 2008 Examination Level: SRO Operating Test Number: 1 I ~~~~ ~

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

PERFORM A TIME TO BOIL CALCULATION FOR THE SPENT FUEL POOL Given shutdown conditions perform a time to boil calculation Conduct of Operations N 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.

Conduct of Operations N 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 UNSATISFACTORYCONDITIONS Review and evaluate surveillance acceptance criteria including TS implication for unsatisfactory conditions.

D Equipment Control 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 P radiation monitor and apply action statements contained in the Radiation Control station ODCM. (CR NM-2004-976)

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

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.

Emergency Plan N

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

  • Type Codes & Criteria: (C)ontrol room, (S)imulator, o r Class(R)oom (D)irect from bank (53for 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 Date of Examination: October 2008 Operating Test Number:J TY Pe Describe activity t o b e performed Code*

PERFORM CONTROL SWITCH LINEUP VERIFICATION While performing N1-PM-DO02 lineup verification, identify system components that are not in the correct lineup D

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 N

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 N

N1-RESP-I, 3 0 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.

D EPIP-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 administrativetopics, when all 5 are required.

  • Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (53for ROs; 5 4 for SROs & RO retakes)

(N)ew or (M)odified from bank (51)

(P)revious 2 exams (51; randomly selected)

ES-301 Control Room/ln-Plant Systems Outline Form ES-301-2 Facility: Nine Mile Point Unit 1 Date of Examination: October 2008 Exam Level: RO/SRO-VSRO-U Operati ng Test No.: 1 Control Room SystemsQ ( 8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including IESF)

SRO-U in BOLD - # s S-I ,3,7/P-1,2 Type Code* Safety System / JPM Title Function

~

S-I Initiate Liquid Poison Injection, RWCU Fails to Isolate D,A,EN,S 1 WA 21 1000 A I .08 3.7/3.8 S-2 Transfer Torus Water to the Waste Collector Tank Using 5 Containment Spray Loop 111 WA 295029 EA1.03 2.9/3.0 S-3 Transfer Load from # I 1 and # 2IFeedwater Pumps to # I 3 2 Feedwater Pump, # I 3 Feedwater FCV fails closed WA 259001 A2.07 3.7/3.8 S-4 Startup Control Room Ventilation System 9 WA 290003 A4.01 3.2/3.2 S-5 EDG 103 S/D - PB 103 Return to Normal Power 6 WA 264000 A4.05 3.6/3.7 S-6 Perform RWM Diagnostic & Rod Block Tests 7

WA 201006 A4.01 thru A4.06, 2.9/2.9 to 3.3/3.4 I 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) 4 S-8 Alternate RPV Blowdown Through Emergency Condenser Vents 3 RO to Torus ONLY K/A 207000 A I .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 8 WA 286000 A3.01 3.4/3.4 P-2 Initiation of Emergency Condensers from Remote Shutdown 3 Panel 11 WA 295016 AA1.09 4.0/4.0 P-3 Place UPS 162A in Standby from Shutdown Condition and 6 Transfer to Supply RPS 11 WA 212000 A I .04 (2.8/3.0), A I .05 (2.6/2.7)

ES-301 Control RoomAn-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.

1 Type Codes Criteria for RO I SRO-I / SRO-U (A)lternate path 4-6 I 4-6 I 2-3 (C)ontrol room (D)irect from bank 191sa114 (E)mergency or abnormal in-plant 27 I 2 1 I 2 1 (EN)gineered safety feature - I - I 21 (control room system)

(L)ow-PowerI Shutdown 21I21 I21 (N)ew or (M)odified from bank including 1(A) 22122121 (P)revious 2 exams I3 I I3 I I2 (randomly selected)

(WCA 21121 1 2 1 (S)imulator

ES-301 Control Room/ln-Plant Systems Outline Form ES-301-2 2007 NRC Examination Summary Description of JPMs s-I 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.

s-2 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 111 IAW N1-EOP-1, Att.15.

s-3 This is a new alternate path JPM in the Rx Water Inventory Control Safety Function area. The applicant will transfer load from # I 1 and # I 2 Feedwater Pumps to # I 3 Feedwater Pump IAW N I -

OP-I 6 and the # I 3 pump flow control valve will malfunction requiring manual actions to control vessel level.

s-4 This is a bank JPM in the Radioactivity Release Safety Function area. The applicant will startup Control Room Ventilation IAW N1-OP-49.

s-5 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.

S-6 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.

5-7 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.

5-8 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.

P-I 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.

P-2 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.

P-3 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 Scenario No.: NRC-01 Op-Test No.: October 2008 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 H2-02Monitoring Systems step G.l to de-inert the Primary Containment with Rx Coolant Temp >212"F Event Malf. No. Event Event No.

Description 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 Op-Test No.: October 2008 TARGET QUANTITATIVE ATTRIBUTES I ACTUAL (PER SCENARIO; SEE SECTION D.5.d)

1. Total malfunctions (5-8)

I 6 Events 2, 4, 5, 6, 7, 9

2. Malfunctions after EOP entry (1-2) 1 Event 9
3. Abnormal events (2-4) 4 Events 4, 5, 6,7
4. Major transients (1-2) 1 Event 8
5. EOPs enteredhequiringsubstantive 1 actions (1-2)

Event 8 (EOP-6)

6. EOP contingencies requiring substantive actions (0-2)

Events 9 (EOP-8)

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.

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 N1-0P-9, N2 lnerting and H2-02Monitoring 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 Scenario No.: NRC-02 Op-Test No.: October 2008 Examiners: Operators:

Initial Conditions: Simulator IC 172

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 No.

DescriDtion 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 OD-Test No.: October 2008 TARGET QUANTITATIVE ATTRIBUTES ACTUAL (PER SCENARIO; SEE SECTION D.5.d) ATTRIBUTES

1. Total malfunctions (5-8) 5 Events 3,4, 5, 6, 8
2. Malfunctions after EOP entry (1-2) 1 Event 8
3. Abnormal events (2-4) 4 Events 3,4, 5, 6
4. Major transients (1-2) 1 Event 7
5. EOPs enteredhequiringsubstantive actions (1-2)

Events 6 and 7 (EOP-5)

6. EOP contingencies requiring substantive actions (0-2)

Events 7,8 (EOP-3, EOP-8)

7. Critical tasks (2-3) 3 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.

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)

TS fSRO1 1 pump to auto start 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 ACTUAL (PER SCENARIO; SEE SECTION D.5.d) ATTRIBUTES

1. Total malfunctions (5-8) 7 Events 2, 3, 4, 5, 6, 8, 9
2. Malfunctions after EOP entry (1-2) 2 Events 8,9
3. Abnormal events (2-4) 6 Events 2,3,4, 5, 6,8
4. Major transients (1-2) I Event 7
5. EOPs enteredhequiringsubstantive 2 actions (1-2)

Events 7,9 (EOP-2, EOP-4)

6. EOP contingencies requiring substantive 2 actions (0-2)

Events 7,9 (EOP-2 Alternate Level Leg, EOP-8)

7. Critical tasks (2-3) 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 October 2008

SCENARIO

SUMMARY

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

SUMMARY

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 11 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 1I 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 111 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, N I -

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 Scenario No.: NRC-04 Op-Test No.: October 2008 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 Event Malf. No. Event Event No.

Type* Description 4 I RMlA I TS (SRO) 1 Main Steam Line Radiation Monitor 111 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)

Overrides (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor NRC Scenario 4 October 2008

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

1. Total malfunctions (5-8) 6 Events 3, 4, 5, 6, 8, 9
2. Malfunctions after EOP entry (1-2) 2 Events 8, 9
3. Abnormal events (2-4) 4 Event 3, 4, 5, 6
4. Major transients (1-2) 1 Event 7
5. EOPs enteredhequiring substantive 1 actions (1-2)

Event 8 (EOP-4)

6. EOP contingencies requiring substantive actions (0-2)

Events 7 , 9 (EOP-3, EOP-8)

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.

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 111 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, N1-EOP-3, N1-EOP-3.1, N1-EOP-4, N1-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

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