ML102320373

R.E. Ginna - Final Outlines (Folder 3)

ML102320373
Person / Time
Site:	Ginna
Issue date:	05/25/2010
From:	Constellation Energy Nuclear Group
To:	Fish T H Operations Branch I
Hansell S
Shared Package
ML092470059	List: IR 05000244/2010301 ML100400205 ML101680575 ML101720347 ML101720553 ML101720560 ML101720564 ML102320373 ML102320386 ML102320393 ML102320408 ML102520045 ML102520054 ML102710006
References
TAC U01766
Download: ML102320373 (44)

Text

Written Examination Outline Form ES-401-2 Facility:

R.E. Ginna Date of Exam: 06/21/10 RO KJA Category Points SRO-Only Points Tier Group K K K K K K A A A A G Total A2 G* Total 1 2 3 4 5 6 1 2 3 4

• 1. 3 3 4 2 3 3 18 3 3 6 Emergency

& 2 1 1 2 2 1 2 9 2 2 4 Plant Evaluations Tier 4 4 6 4 4 5 27 5 5 10 Totals 1 2 2 3 2 3 2 2 3 3 3 3 28 3 2 5 2. 2 2 1 2 1 0 0 1 1 1 1 0 10 0 2 1 3Systems 4 3 5 3 3 2 3 4 4 4 3 38 5 3 8 Totals 1 2 3 4 1 2 3 4 3. Generic Knowledge

& Abilities 10 7 3 3 2 2 2 1 2 2 Note Ensure that at least two topics from every applicable KiA category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each KiA category shall not be less than two). 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. 3Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems that are not included on the outline should be added. Refer to section D.1.b of ES-401, for guidance regarding elimination of inappropriate KiA statements.

4Select 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 (lR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively. 6Select SRO topics for Tiers 1 and 2 from the shaded systems and KiA categories. The generic (G) KiAs in Tiers 1 and 2 shall be selected from Section 2 of the KiA Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable KiA's On the following pages, enter the KiA 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 l eft 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 KiAs that are linked to 10CFR55.43 1

/6 ES-401 Form ES-401-2 R.E. 2010 NRC Written Examination Emergency and Abnormal Plant Evolutions

-Tier 1 Group EAPE#/Name Safety Function 056 / Loss of Off-site Power 055 / Station Blackout / 6 026 / Loss of Component Cooling Water / 8 015 / RCP Malfunctions (Loss of RCS Flow) / 3 E12 / Steam Line Rupture-Excessive Heat Transfer / 4 025 / Loss of Residual Heat Removal System / 4 009 / Small Break LOCA / 3 X KIA Topic(s) AA2.23 -Ability to determine and interpret the following as X they apply to the Loss of Offsite Power: Turbine trip-reactor button and indicator EA2.04 -Ability to determine or interpret the following as X they apply to a Station Blackout:

Instruments and controls operable with only dc battery power available AA2.06 -Ability to determine and interpret the following as they apply to the Loss of Component Cooling Water: X The length of time after the loss of CCW flow to a component before that component may be damaged 2.4.4 -Emergency Procedures

/ Plan: Ability to recognize abnormal X indications for system operating parameters that are entry level conditions for emergency and abnormal 2.2.22 -Equipment Control: X Knowledge of limiting conditions for operations and safety limits. 2.4.3 -Emergency X Procedures

/ Plan: Ability to identify post-accident instrumentation.

EK1.02 -Knowledge of the operational implications of the following concepts as they apply to the small break LOCA: Use of steam tables 3.9 76 4.1 77 3.1 78 4.7 79 4.7 80 3.9 81 3.5 39 ES-401 Form ES-401-2 R.E. 2010 NRC Written Examination Emergency and Abnormal Plant Evolutions

-Tier 1 Group EAPE#/Name Safety Function 007 / Reactor Trip x Stabilization

-Recovery / 1 E05/ Inadequate Heat Transfer -Loss of x Secondary Heat Sink / 4 E 11 / Loss of Emergency x Coolant Recirculation / 4 E04 / LOCA Outside x Containment

/ 3 077 / Generator Voltage and Electric Grid x Disturbances KIA Topic(s) EK1.04 -Knowledge of the operational implications of the following concepts as they apply to the reactor trip: Decrease in reactor power following reactor trip (prompt and su uent '"',,, ....., EK1.1 -Knowledge of the operational implications of the following concepts as they apply to the (Loss of Secondary Heat Sink) Components, capacity, and function of emergency EK2.2 -Knowledge of the interrelations between the (Loss of Emergency Coolant Recirculation) and the following:

Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the faci EK2.1 -Knowledge of the interrelations between the (LOCA Outside Containment) and the following:

Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes , and automatic and manual features.

AK2.07 -Knowledge of the interrelations between Generator Voltage and Electric Grid Disturbances and the following:

Turbine / control 3.6 40 3.8 41 3.9 42 3.5 43 3.6 44 Form ES-401-2 R.E. Ginna 2010 NRC Written Examination Outline Emergency and Abnormal Plant Evolutions

-Tier 1 Group 1 EAPE#/Name Safety KIA Topic(s) AK3.03 -Knowledge of the reasons for the following responses as they apply to 022 I Loss of the Loss of Reactor Coolant x Coolant Makeup / Makeup: Performance of lineup to establish excess letdown after determining need AK3.03 -Knowledge of the reasons for the following responses as they apply to 026 I Loss of the Loss of Component x Cooling Water I Cooling Water: Guidance actions contained in EOP for Loss of CCW/nuclear service water AK3.02 -Knowledge of the reasons for the following 040 / Steam Line Rupture / responses as they apply to 4 the Steam Line Rupture: ESF AS initiation AA 1.24 -Ability to operate and I or monitor the following as they apply to the Loss of 056 I Loss of Off-site Power Offsite Power: Plant 16 computer, to call up in-core temperature monitoring AA 1.21 -Ability to operate and / or monitor the following as they apply to the Reactor 5 / 17 1 Reactor Coolant Coolant Pump Malfunctions Pump Malfunctions 14 (Loss of RC Flow): Development of natural circulation flow EK3.02 -Knowledge of the reasons for the following responses as they apply to 055 / Station Blackout / 6 Loss of Off-Site and On-Site Power: Actions contained in EOP for loss of off-site and 3.1 45 4.0 46 4.4 47 2.9 48 4.4 49 4.3 50 Form ES-401-2 R.E. 2010 NRC Written Examination Emergency and Abnormal Plant Evolutions

-Tier 1 Group EAPE#/Name Safety KIA Topic(s) AA2.17 -Ability to determine and interpret the following as 057 1 Loss of Vital AC they apply to the Loss of Electrical Instrument Bus 1 Vital AC Instrument Bus: 3.1 51 6 System and component status, using local or remote controls AA2.12 -Ability to determine 027 1 Pressurizer Pressure and interpret the following as Control System Malfunction they apply to the 3.7 52 13 Pressurizer Pressure Control Malfunctions:

PZR level EA2.01 -Ability to determine 038 1 Steam Generator or interpret the following as 4.1 53 Tube Rupture 1they apply to a SGTR: When to isolate one or more S/Gs 2.1.23 -Conduct of Operations:

Ability to perform specific system and 011 1 Large Break LOCA 1 4.3 54 integrated plant procedures during all modes of plant 054 1 Loss of Main 4.2 55 Feedwater 1 4 0081 Pressurizer Vapor 3.8 56 Space Accident 1 3 KIA Category Totals 1816 Form ES-401-2 R.E. 2010 NRC Written Examination Emergency and Abnormal Plant Evolutions

-Tier 1 Group EAPE#/Name Safety KIA Topic(s) AA2.03 -Ability to determine and interpret the following as 036 I Fuel Handling they apply to the Fuel 4.2 82 Incidents Handling Incidents:

Magnitude of potential radioactive release AA2.12 -Ability to determine and interpret the following as 0331 Loss of Intermediate they apply to the Loss of Range Nuclear X Intermediate Range Nuclear 3.1 83 Instrumentation 17 Instrumentation:

Maximum allowable channel disagreement 2.4.30 -Emergency Procedures I Plan; Knowledge of events related to system operation I status E03 I LOCA Cooldown that must be reported to 4.1 84 Depressurization internal organizations or external agencies, such as the state, the NRC, or the transmission system operator.

2.4.6 -Emergency E141 High Containment Procedures I Plan:4.7 85 Pressure Knowledge of EOP mitigation strategies. AK1 .01 -Knowledge of the operational implications of the following concepts as 032 I Loss of Source Range they apply to Loss of Source 2.5 57 Instrumentation 17 Range Instrumentation

Effects of voltage changes on performance EK2.08 -Knowledge of the interrelations between the 0741 Inadequate Core following Inadequate Core 2.5 58 Cooling 14 Cooling: Sensors and detectors 6

Form ES-401-2 R.E. 2010 NRC Written Examination Emergency and Abnormal Plant Evolutions

-Tier 1 Group EAPE#/Name Safety Function I K1 I K2 I K3 I A1 I A2 I G KIA Topic(s) 061 1 Area Radiation Monitoring (ARM) System Alarms/7 x E02/ SI Termination 13 E 13 1 Steam Generator Overpressure I 4 069 / Loss of Containment Integrity / 5 076 I High Reactor Coolant Activity /9 X

• E 1 0 I Natural Circulation with Steam Void in Vessel with/without RVLlS /4 AK3.02 -Knowledge of the reasons for the following responses as they apply to the Area Radiation Monitoring (ARM) System Alarms: Guidance contained in alarm response for ARM EA 1.3 -Ability to operate and / or monitor the following as they apply to the (SI Termination)

Desired operating results during abnormal and emergency situations.

EA2.2 -Ability to determine and interpret the following as they apply to the (Steam Generator Overpressure)

Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments.

2.2.39 -Equipment Control: Knowledge of less than or equal to one hour technical specification action statements for AK3.05 -Knowledge of the reasons for the following responses as they apply to the High Reactor Coolant Activity : Corrective actions as a result of high fission-product radioactivity level in the RCS 2.2.37 -Equipment Control: Ability to determine operability and I or availability of safety related I Imp. I Q# I 3.4 59 3.8 60 3.0 61 3.9 62 2.9 63 3.6 64 ES-401 Form ES-401-2 R.E. 2010 NRC Written Examination Emergency and Abnormal Plant Evolutions

-Tier 1 Group EAPE#/Name Safety Function KiA Topic(s) 003 1 Dropped Control Rod 11 AA 1.01 -Ability to operate and 1 or monitor the following as they apply to the Dropped Control Rod: Demand position counter and converter KiA Category Totals 2 Group Point Total: 2.9 65 9/4 ES-401 Form ES-401-2 R.E. Ginna 2010 NRC Written Examination Outline Plant Systems -Tier 2 Group 1 System #/Name KIA Topic(s) A2.04 -Ability to (a) predict the impacts of the following malfunctions or operations on the CSS; and (b) based on those 026 Containment Spray X predictions, use procedures to correct, 4.2 86 control, or mitigate the consequences of those malfunctions or operations:

Failure of spray pump A2.04 -Ability to (a) predict the impacts of the following malfunctions or operations on the Main Feedwater System; and (b) based on those 059 Main Feedwater X predictions , use 3.4 87 procedures to correct , control , or mitigate the consequences of those malfunctions or operations

Feeding a dry SG 2.2.38 -Equipment 010 Pressurizer Pressure Control X Control: Knowledge of conditions and limitations 4.5 88 in the facility license. 2.4.9 -Emergency Procedures I Plan: Knowledge of low power 004 Chemical and Volume Control X I shutdown implications in accident (e.g., loss of 4.2 89 coolant accident or loss of residual heat removal) mitigation strategies. 9 ES-401 Form ES-401-2 R.E. Ginna 2010 NRC Written Examination Outline Plant Systems -Tier 2 Group 1 System #/Name KIA Topic(s) A2.11 -Ability to (a) predict the impacts of the following malfunctions or operations on the ECCS; and (b) based on those 006 Emergency Core Cooling X predictions, use procedures to correct , 4.4 90 control, or mitigate the consequences of those malfunctions or operations:

Rupture of ECCS header K1.01 -Knowledge of the physical connections 022 Containment Cooling X and/or cause-effect relationships between the CCS and the 3.5 1 following systems: SWS/cooling system K1.02 -Knowledge of the physical connections 008 Component Cooling Water X and/or cause-effect relationships between the CCWS and the 3.3 2 following systems: Loads cooled by CCWS K2.02 -Knowledge of 064 Emergency Diesel Generator X bus power supplies to the following:

Fuel oil 2.8 3 pumps K2.02 -Knowledge of 078 Instrument Air X bus power supplies to the following Emergency 3.3 4 air compressor K3.01 -Knowledge of 026 Containment Spray X the effect that a loss or malfunction of the CSS will have on the 3.9 5 following:

CCS K3.07 -Knowledge of the effect that a loss or 076 Service Water X malfunction of the SWS 3.7 6 will have on the following:

ESF loads 10

Form R.E. Ginna 2010 NRC Written Examination Outline Plant Systems Tier 2 Group 1 System #/Name KiA Topic(s) K4.06 Knowledge of MRSS design feature(s) 039 Main and Reheat Steam x and/or interlock(s) which provide for the following:

3.3 7 Prevent reverse steam flow on steam line break K4.0B Knowledge of 013 Engineered ESFAS design feature(s)

Safety Features and/or interlock(s) which 3.1 B !J""\I.,LUCUIUn provide for the following Redund K5.44 Knowledge of the operational implications of the 004 Chemical and ume Control following concepts as they apply to the CVCS: 3.2 9 Pressure response in PZR during K5.01 Knowledge of the operational Emergency Core Cooling x implications of the following concepts as they apply to ECCS: 2.B 10 Effects of temperatures on water level indications K6.03 Knowledge of 12 Reactor . Protection the effect of a loss or malfunction of the following will have on the 3.1 11 RPS: Tri ic circuits K6.02 -Knowledge of the effect of a loss or 003 Reactor Coolant jPump malfunction on the following will have on the 2.7 12 RCPS: RCP Seals and seal water su A 1.02 -Ability to predict and/or monitor changes in parameters (to prevent 007 Pressurizer Relief/Quench Tank exceeding design limits) associated with operating the PRTS 2.7 13 controls including:

Maintaining quench tank ES-401 Form ES-401-2 System #lName 063 DC Electrical Distribution 005 Residual Heat Removal 010 Pressurizer Pressure Control 059 Main Feedwater 103 Containment R.E. Ginna 2010 NRC Written Examination Outline Plant Systems -Tier 2 Group 1 KIA Topic(s) A 1.01 -Ability to predict and/or monitor changes in parameters associated with operating the dc electrical system controls including:

Battery capacity as it is affected lilq,!"h:::lrn", rate A2.01 -Ability to (a) predict the impacts of the following malfunctions or operations on the RHRS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Failure modes for pressure, flow, pump motor amps, motor temperature, and tank level instrumentation A2.01 -Ability to (a) predict the impacts of the following malfunctions or operations on the PZR PCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Heater failures A3.06 -Ability to monitor automatic operation of the MFW, including:

Feedwater isolation A3.01 -Ability to monitor automatic operation of the containment system, including:

Containment isolation 2.5 14 2.7 15 3.3 16 3.2 17 3.9 18 ES-401 Form ES-401-2 System #!Name 062 AC Electrical Distribution 073 Process Radiation Monitoring 1 Auxiliary/Emergency Feedwater 076 Service Water 013 Engineered Safety Features Actuation 010 Pressurizer Pressure Control 007 Pressurizer Relief/Quench Tank R.E. Ginna 2010 NRC Written Examination Outline Plant Systems -Tier 2 Group 1 KIA Topic(s) A4.04 -Ability to manually operate and/or monitor in the control 2.6 19 room: Local operation of breakers A4.02 -Ability to manually operate and/or monitor in the control room Radiation 3.7 20 monitoring system control .34 -Emergency Procedures

/ Plan: Knowledge of RO tasks performed outside the main control room during 4.2 21 an emergency and the resultant operational effects. 2.1.27 -Conduct of Operations:

Knowledge of system purpose and / 3.9 22 or function.

K3.01 -Knowledge of the effect that a loss or malfunction of the 4.4 23 ESFAS will have on the followi : Fuel 2.4.20 -Emergency Procedures

/ Plan; Knowledge of operational 3.8 24 implications of EOP warnings, cautions, and notes. A3.01 -Ability to monitor automatic operation of the PRTS, including:

2.7 25 Components which discha to the PRT ES-401 Form ES-401-2 R.E. 2010 NRC Written Examination Plant Systems -Tier 2 Group System #/Name 039 Main and Reheat Steam 063 DC Electrical Distribution 003 Reactor Coolant x Pump KIA Category Totals 2 2 3 2 322 KIA Topic(s) A2.05 -Ability to (a) predict the impacts of the following mal-functions or operations on the MRSS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Increasing steam demand, its relationship to increases in reactor r A4.01 -Ability to manually operate and/or monitor in the control room: Major breakers and control fuses K5.02 -Knowledge of the operational implications of the following concepts as they apply to the RCPS: Effects of RCP coastdown on RCS Group Point Total: 3.3 26 2.8 27 2.8 28 28/5 ES-401 Form ES-401-2 R.E. 2010 Written Examination Plant Systems -Tier 2 Group System #/Name KIA Topic(s) 068 Liquid Radwaste X 2.1.32 Ability to explain and apply all system limits and precautions 2.6 91 2.4.21 -Emergency Procedures

/ Plan: Knowledge of the parameters and logic used to assess the status of safety 011 Pressurizer Level Control X functions, such as reactivity control, core 4.6 92 cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc. A2.19 -Ability to (a) predict the impacts of the following malfunction or operations on the CRDS-and (b) based on those 001 Control Rod Drive X predictions, use procedures to correct, 4.0 93 control, or mitigate the consequences of those malfunctions or operations:

Axial flux distribution K3.01 -Knowledge of 011 Pressurizer Level Control X the effect that a loss or malfunction of the PZR LCS system will have on 3.2 29 the following:

CVCS A4.02 -Ability to 015 Nuclear Instrumentation X manually operate and/or monitor in the control 3.9 30 room: NIS indicators A3.01 -Ability to monitor automatic operation of 016 Non-nuclear Instrumentation X the NNIS, including:

Automatic selection of 2.9 31 NNIS inputs to control systems ES-401 Form ES-401-2 System #/Name 001 Control Rod Drive 002 Reactor Coolant 071 Waste Gas Disposal 072 Area Radiation Monitoring 041 Steam DumplTurbine Bypass Control 055 Condenser Air !Removal R.E. Ginna 2010 Written Examination Outline Plant Systems -Tier 2 Group 2 KiA Topic(s) K2.05 -Knowledge of x bus power supplies to 3.1 32 the followi : MIG sets K1.07 -Knowledge of the physical connections and/or cause-effect relationships between the RCS and the 3.5 33 following systems: Reactor vessel level indication K4.01 -Knowledge of design feature(s) and/or interlock(s) which x provide for the following:

2.6 34 Pressure capability of the waste gas decay tank K3.02 -Knowledge of the effect that a loss or x malfunction of the ARM system will have on the 3.1 35 following:

Fuel handling A2.02 -Ability to (a) predict the impacts of the following malfunctions or operations on the SDS; and (b) based on those 3.6 36 predictions or mitigate the consequences of those malfunctions or operations:

Steam valve stuck 0 K1.06 -Knowledge of the physical connections and/or cause effect x relationships between 2.6 37 the CARS and the following systems: PRM ES-401 Form ES-401-2 R.E. 2010 Written Examination Plant Systems -Tier 2 Group System #/Name 086 Fire Protection KIA Category Totals KIA Topic(s) A 1.04 -Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the Fire Protection System controls including:

Fire dam 2.7 38 10/3 ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: R.E. Ginna Category KA# 2.1.5 2.1.8 2.1.3 1. Conduct of Operations 2.1.25 2.1.42 Subtotal 2.2.37 2.2.35 2.2.13 2. Equipment Control 2.2.44 Subtotal Date: 6/21/2010 Topic Abil i ty to use procedures related to shift staffing, such as minimum crew complement , overtime l imitations, etc. Ability to coordinate personnel activities outside the control room. Knowledge of shift or short-term relief turnover practices.

Ability to interpret reference materials , such as graphs , curves, tables , etc. Knowledge of new and spent fuel movement procedures. Ability to determine operabili t y and / or avai l ability of safety related equipment.

Ability to determine Technical Specification Mode of Operation. Knowledge of tagging and clearance I procedures Ability to interpret control room indications to verify the status and operation of a system , and understand how operator actions and directives affect plant and system conditions. IR 2.9 3.4 3.7 3.6 3.6 4.1 RO SRO-Only Q# IR Q# 66 67 75 4.2 94 3.4 1 00 3 2 68 69 74 4.4 95 3 1 ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 2.3.14 2.3.13 3. Radiation Control 2.3.5 2.3.4 Subtotal 2.4.30 2.4.39 4. Emergency Procedures

/ Plan 2.4.41 2.4.38 Subtotal Tier 3 Point Total: Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

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. 3.4 3.4 70 71 Ability to use radiation monitoring systems , such as fixed radiation monitors and alarms , portable survey instruments , personnel monitoring equipment, etc. Knowledge of radiation exposure limits under normal or emergency conditions. 2.9 3.7 96 98 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.

Knowledge of the RO's responsibilities in emergency plan implementation.

2.7 3.9 2 72 73 2 Knowledge of the emergency act i on level thresholds and classifications. Ability to take actions called for in the facility emergency plan, including supporting or acting as emergency coordinator if required. 4.6 4.4 97 99 2 10 2 7 19 ES-401 Record of Rejected KiA's Form ES-401-4 Tier / Group 3/3 3/1 2/2 2/1 2/1 1 / 1 2/1 3/2 2/2 2/2 3/2 3/4 2/1 2/2 2/2 1 / 1 1 /2 Randomly Selected KA G3 / 2.3.4 replaced by G3 /2.3.14 G2 / 2.2.35 replaced by G2 /2.2.44 028/ K6.01 replaced by 002 / K6.04 007 / A2.01 replaced by 059 / A2.04 012 /G2.4.9 replaced by 004/ G2.4.9 040/ AK3.03 replaced by 040 / AK3.02 010/ G2.4.30 r eplaced by 010/ G2.4.20 G2 / 2.2.15 replaced b y G2 / 2.2.7 028/ A2.01 replaced b y 001/ A2.19 068/ A2.02 replaced by068/2.1.32 G /2.2.7 rep l aced by G /2.2.13 062 G2.4.34 replaced by 015 G2.4.4 003 K6.14 replaced by 003 K6.02 01 1 K5.12 replaced by 011 K3.01 002 K6.04 replaced by 002 K1 .07 055EA 1.01 replaced by 055 EK3.02 024 AK1.01 replaced by 032 AK1 .01 Reason for Rejection This topic is duplicated on the SRO Examinat i on Duplicated on RO examination and is more suitable to RO examination System is no longer used or referenced by any procedures at t he facility Excessive overlap for system. Questions 7 and 47 also test same topic. Reactor Protection has already performed its function in lower mode, making it impossible to develop a test item to evaluate knowledge of this KA topic for this system Direct overlap with question 7. Selected this item for remova l because it has the lower KA importance value of the 2 items For this system , this generic KA will not yield an RO test i tem. This information is confined to respons i bility of SRO. Direct overlap with topic on Operating Examination. System is no longer used or referenced by any procedures at the facility No facility reference for selected KA and no known action for condition that KA presented. KA topic would not yield a topic at RO level. Corporate procedure had criteria that were variable and would require significant interpreta t ion by SROs Unable to develop SRO question to the selected topic because no RO actions outside of the control room would be taken for the event Unable to develop an operationally sound test item to meet the KA using a val i d faci l ity reference. Topic selected would only yield GFES type test item Unable to develop an operationally sound test item to match required KA topic. Unable to develop a test item that would avoid overlap with other test items and be higher than LOD =1 Unable to develop an operationally relevant question with supporting facility references 20 ES-401 Record of Rejected KIA's Form ES-401-4 II Administrative Topics Outline Form ES-301-1 Facility:

Ginna Date of Examination:

6/21/10 Examination Level: RO Operating Test Number: NRC Administrative Topic Type Code* Describe activity to be performed (see Note) 2.1.25 (3.9) Ability to interpret reference material, such as Conduct of Operations graphs, curves, tables, etc. M,R JPM: Given a Set of Conditions, Perform a Critical Rod Position Calculation In Accordance With 1.2.2, Critical Rod Position Calculation.

2.1.18 (3.6) Ability to make accurate, clear, and concise Conduct of Operations M, S logs, records, status boards, and reports. JPM: Perform a Daily Surveillance Log 2.2.13 (4.1) Knowledge of tagging and clearance Equipment Control procedures.

M,R JPM: Perform the RO Review of a Tagout For V-3968, 4B Condensate Heater Discharge Check Valve. Radiation N/A CATEGORY NOT SELECTED . 2.4.21 (4.0) Knowledge of the parameters and logic used to Emergency assess the status of safety functions, such as reactivity control, core COOling and heat removal, reactor coolant system integrity, containment M,R conditions, radioactivity release control, etc. JPM: Monitor Critical Safety Function Status Trees All items (5 total are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.

• Type Codes & (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (::> 3 for ROs; ::> 4 for SROs & RO retakes) (0) (N)ew or (M)odified from bank (> 1) (4) (P)revious 2 exams (::> 1; randomly selected)

(0)

Administrative Topics Outline Form ES-301-1 RO Admin JPM Summary This JPM is modified.

This JPM is considered modified because there are existing Critical Rod Position calculations in the exam bank, but there are not any for the conditions specified in this JPM. Using 0-1.2.2, Critical Rod Position Calculation, the candidate will follow the procedural steps and use the attached graphs and tables to determine the critical rod position.

This JPM will be accomplished in the classroom. This JPM is modified.

This JPM is considered modified because there are existing JPMs requiring the performance of a daily log surveillance in the exam bank, but there are not any for the conditions specified in this JPM. The candidate will be required to use 0-6.13, Daily Surveillance Log, Attachment 1, to log plant parameters displayed on the Main Control Board. Parameters out ()f specification will require recommendations to enter Technical Specifications.

This JPM will be accomplished in the simulator. This JPM is modified.

This JPM is considered modified because there are existing JPMs requiring the review of a tagout in the exam bank, but there are not any for the component specified in this JPM. The candidate will be required to use system drawings and CNG-OP-1.01-1007, Clearance and Safety Tagging, to review the isolation points required to replace the discharge check valve for the 4B Low Pressure Heater and then review the complete forms associated with the tagout. This JPM will be conducted in the classroom. This JPM is modified.

This JPM is considered modified because there are existing ..IPMs requiring the monitoring of the Critical Safety Function Status Trees, but there are not any for the conditions specified in this ..IPM. The candidate will be given plant data to review and be required to identify the status each Critical Safety Function Status Trees and make a recommendation on what procedure to enter. This JPM will be accomplished in the classroom.

Administrative Topics Outline Form ES-301-1 Facility:

Ginna of Examination:

6/21/10 Examination Level: SRO Operating Test Number: NRC Administrative Topic (see Note) Type Code"' Describe activity to be performed Conduct of Operations M,R 2.1.25 (4.2) JPM: Ability to interpret reference material, such as graphs, curves, tables, etc. Independently Verify a Critical Rod Position Calculation Conduct of Operations M,R 2.1.18 (3.8) JPM: Ability to make accurate, clear, and concise logs. records, status board, and reports. Perform the Shift Manager review of the 0-6.13. DAILY SURVEILLANCE LOG Equipment Control D,R 2.2.37 (4.6) Ability to determine operability and/or availability of safety related equipment.

JPM: Perform a Safety Function Determination Radiation Control P,R 2.3.14 (3.8) Knowledge of radiation or contamination hazards that may arise during normal. abnormal, or emergency conditions or activities.

JPM: Respond to a Contaminated Injured Person Emergency Plan D.R 2.4.41 (4.6) Knowledge of the emergency action level thresholds and classifications.

JPM: Determine the EAL For an Event All items (5 total are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.

• Type Codes & (C)ontrol room. (S)imulator.

or Glass(R)oom (D)irect from bank (.,; 3 for ROs; s 4 for SROs & RO retakes) (2) (N)ew or (M)odified from bank (> 1) (2) (P)revious 2 exams (.,; 1; randomly selected)

(1)

Administrative Topics Outline Form ES-301-1 SRO Admin JPM Summary This JPM is modified.

This is considered modified because there are existing Critical Rod Position calculations in the exam bank, but there are not any for the conditions specified in this JPM. The candidate will review a completed 0-1.2.2, Critical Rod Position Calculation, and identify errors that result in the estimated and actual critical rod position to differ by approximately 200 pcm. This JPM will be accomplished in the classroom. This JPM is modified.

This JPM is considered modified because there are existing 6.13, Daily Surveillance Logs, in the exam bank, but errors in the JPM have been changed. The candidate will review the completed 0-6.13 and identify three errors, one of which will require entry into a Technical Specification LCO. This JPM will be accomplished in the classroom. This JPM is directly from our bank. The candidate will perform a safety function determination using Attachment 1 and 2 of A-S2.4, Control of Limiting Conditions for Operating Equipment.

While performing the Safety Function Determination the candidate will determine that the plant will be in a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Limiting Condition for Operation.

This JPM will be accomplished in the classroom. This JPM was randomly selected from the 2007 NRC Exam. The operator will be required to respond to a contaminated, injured person using A-7, Procedure for Handling Injuries/Medical Emergencies at Ginna Station. Also CNG-NL-1.01-1004, Regulatory Reporting, will be used to report the incident to the NRC within the required time period. This JPM will be accomplished in the classroom. This JPM is considered direct from the exam bank since it was developed for use on the 2010 Licensed Operator Requalification Exam. The candidate will evaluate plant conditions and determine a Site Area EmergE!ncy classification is required.

After classification the candidate will fill out the EPIP-1.S, Notifications, Attachment 3a, New York State Radiological Emergency Data, for transmission.

This is a time critical JPM. This will be accomplished in the classroom.

-2 Control Room/In-Plant Systems Outline Form Facility: Date of Examination: Exam Level (circle one): Operating Test No.: Control Room Systems@ (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U, including 1 ESF) System I JPM Title 064 Emergency Diesel Generator Start an Emergency Diesel Generator During a Loss of All AC Power 006 Emergency Core Cooling System In accordance with FR-C.1, RESPONSE TO INADEQUATE CORE COOLING. Isolate an SI Accumulator That Has Discharged 003 Reactor Coolant Pump Start a Reactor Coolant Pump During a Plant Startup and Then Respond to an Alarm Associated With the Pump Just Started APE 068 Control Room Evacuation Respond To a Control Room Evacuation 010 Pressurizer Pressure Control System Place L TOP in service 004 Chemical and Volume Control System Initiate Rapid 80ration in Preparation For Proceeding to Colc! Shutdown 026 Containment Spray Evaluate Containment Spray (CS) Flow Requirements and Reduce Flow Accordance With E-1, LOSS OF REACTOR OR SECONDARY 061 Auxiliary I Emergency Feedwater System Place the Standby Auxiliary Feedwater System in service during a loss of heat sink In-Plant Systems@ (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U) 071 Waste Gas Disposal System Release the D Gas Decay Tank APE 067 Plant Fire On Site Secure Ventilation System During a Fire Type Code* Safety Function S, A, M, E, EN 6 S,A,E,EN,N 2 S,A, L,M 4P S,A,E.M 8 S. M, L 3 S.L,N 1 S,E,EN,N 5 S, M, E 4S A,M,R 9 D,E,R 8 Control Room/In-Plant Systems Outline Form ES-301-2 k. 039 Main and Reheat Steam System A,P,E 4S Locally Operate the ARVs 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-II SRO-U (A)lternate path 4-6 (6) / 4-6/2-3 (C)ontrol room (D)irect from bank ::o9(1)/s:8/s:4 (E)mergency or abnormal in-plant 1 (7) 11 1?:. 1 (EN)gineered safety feature -/ -1 1 (control room system) (L)ow-Power 1 Shutdown 1 (3) 11 I?:. 1 (N)ew or (M)odified from bank including 1 (A) 2 (9) I ?:. 2 I?:. 1 (P)revious 2 exams s: 3 (1) I s: 3 I ::;; 2 (randomly selected) (R)CA ?:. 1 (2) 11 I 1 (S)imulator

-2 ES-301 Control Room/In-Plant Systems Outline JPM Summary JPM A This is a modified JPM. This JPM is considered modified because there are existing ,IPMs where an Emergency Diesel Generator is started during a loss of all AC power in the exam bank, but there are not any with the malfunctions that are in this JPM. The candidate will be placed in a loss of all AC power when the Control Room Supervisor directs them to start the "A" Emergency Diesel Generator.

This is an alternate path JPM because when the Emergency Diesel Generator starts the candidate will find the Emergency Diesel Generator is not being cooled. After energizing the required electrical bus and starting the required pumps, an Auxiliary Operator will call the control room and inform the Candidate the Emergency Diesel is still not being cooled. Based on the report from the Auxiliary Operator the Candidate will need to make the decision to trip the Emergency Diesel Generator.

JPM B This is a new JPM. The candidate will be placed in an inadequate core cooling situation and the crew has just reduced Reactor Coolant System pressure to allow the Safety Injection Accumulators to inject. The Control Room Supervisor will direct the candidate to isolate the "An and "B" Safety Injection Accumulators.

The "An Safety Injection Accumulator Isolation Valve will not close. The alternate path will consist of the candidate having to vent the "A" Safety Injection Accumulator.

JPM C This is a modified JPM. This JPM is considered modified because there are existing JPMs where a Reactor Coolant Pump is required to be started, but there are not any with the malfunctions that are in this JPM. The candidate will be placed in a plant startup. While heating up from cold shutdown to hot shutdown following a refueling outage the Control Room Supervisor will direct the candidate to start the "A" Reactor Coolant Pump. During the Reactor Coolant Pump start, the operator will have to control Reactor Coolant System pressure using the Letdown Line Pressure Control Valve and coordinating with another board operator to control charging line flow. Shortly after starting the Reactor Coolant Pump, the bearing oil system will loose oil and require the pump to be tripped. ,IPM D This is a modified JPM. This JPM is considered modified because there are existing JPMs where the performance of the immediate actions of AP-CR.1 is required, but there are not any with the malfunctions that are in this JPM. The operator will take the watch as the Head Control Operator.

As soon as the JPM begins the examiner will inform the Candidate that noxious fumes are in the Control Room. The candidate will begin performing the immediate actions of the control room evacuation procedure.

Each immediate action step performed by the candidate will require alternate path actions due to equipment failure. JPM E This is a modified JPM. This JPM is considered modified because there are existing JPMs where L TOP is placed in service, but there are not any with the malfunctions that are in this JPM. The candidate will be placed in a plant shutdown.

While cooling down from Hot Shutdown to Cold Conditions, the crew is ready to place L TOP in service. The Shift Manager will direct the candidate to align one Pressure Operated Relief Valve for Low Temperature Over Pressure control. JPM F This is a new JPM. The candidate will be placed in a plant shutdown.

While cooling down from Hot Shutdown to Cold Shutdown the crew will be required to borate to cold shutdown using rapid boration.

The candidate will be directed by the Control Room Supervisor to coordinate with the Auxiliary Operators to perform the rapid boration.

JPM G This is a new JPM. The candidate will be in a Design Basis Loss of Coolant Accident.

After the Containment Recirculation Fan Coolers and Containment Spray System have reduced Containment pressure to less than 4 PSIG. the candidate will be directed to secure two Containment Spray Pumps JPM H This is a modified JPM. This JPM is considered modified because there are existing JPMs where Standby Auxiliary Feedwater is placed in service, but there are not any with the plant conditions that are in this JPM. The candidate will be in a Loss of Secondary Heat Sink with a ruptured Steam Generator.

While attempting to restore feedwater to the Steam Generators the Control Room Supervisor will direct the Candidate to align and start the Standby Auxiliary Feedwater system. Based on plant conditions the Candidate will have to cross connect the "A" train of Standby Auxiliary Feedwate:r to the "B" train of Standby Auxiliary Feedwater.

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 JPM I This is a modified JPM. This JPM is considered modified because there are existing JPMs where a Gas Decay Tank is released, but there are not any with the mailfunctions that are in this JPM. The candidate will be informed that the plant is operating at 100% power and the "0" Gas Decay Tank is full. When given the order to release the "0" Gas Decay Tank, they will go into the Auxiliary Building, a Radiological Controlled Area, and manually release the tank. After the release has been initiated, the Control Room will contact the Candidate and inform them the radiation monitor monitoring the release is in alarm. Based on the Control Room's report, the Candidate will secure the release. JPM J This is a Bank JPM. The candidate will be informed that a fire has occurred on the intermediate floor of the Auxiliary Building, a Radiological Controlled Area. The Shift Manager will direct the candidate to secure ventilation in the intermediate floor of Auxiliary Building.

JPM K This JPM was on the 2008 NRC Exam. The candidate will be informed that the Control Room has been evacuated due to a fire. The Shift Manager will direct the Candidate to locally operate the "A" Atmospheric Relief Valve. After initially opening the "A" Atmospheric Relief Valve the Shift Manager will direct the Candidate throttle closed the valve. While throttling the valve it will become stuck. The Candidate will have to isolate the "A" Atmospheric Relief Valve's root valve.

Control Roomlln-Plant Systems Outline Form ES-301-2 Facility:

Ginna Date of Examination:

6/21/10 Exam Level: SRO(I) Operating Test No.: NRC Control Room Systems@ (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U, including 1 ESF) Type Code* Safety System 1 JPM Title Function a. 064 Emergency Diesel Generator S,A,M,E,EN 6 Start an Emergency Diesel Generator During a Loss of All AC Power 006 Emergency Core Cooling S, A, E, EN, N In accordance with FR-C.1, RESPONSE TO INADEQUATE COOLING, Isolate an SI Accumulator That Has i 003 Reactor Coolant S, A, L, M Start a Reactor Coolant Pump During a Plant Startup and Then Respond an Alarm Associated With the Pump Just d. APE 068 Control Room Evacuation S,A,E,M 8 Respond To a Control Room Evacuation

e. 010 Pressurizer Pressure Control System S, M, L 3 Place L TOP in service ! f. 004 Chemical and Volume Control System S,L,N 1 Initiate Rapid Boration in Preparation For Proceeding to Cold Shutdown 026 Containment S,E,EN,N Evaluate Containment Spray (CS) Flow Requirements and Reduce Flow Accordance With E-1, LOSS OF REACTOR OR SECONDARY In-Plant Systems@ (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U) i. 071 Waste Gas Disposal System A,M,R 9 Release the D Gas Decay Tank j. APE 067 Plant Fire On Site D,E,R 8 Secure Ventilation System During a Fire k. 039 Main and Reheat Steam System A,P,E 4S Locally Operate the ARVs .@ 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.

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

• Type Codes (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power 1 Shutdown (N)ew or (M)odified from bank including 1 (A) (P)revious 2 exams (R)CA (S)imulator Criteria for RO 1 SRO-II SRO-U 4-6/4-6 (6)/2-3 :>9/::;8(1)/::;4 1 11 (6)1 1 1 (2)(control room 1 11 (3)1 2 12 (8)/ ::; 3 1::; 3 (1)1 ::; 2 (O)(randomly 1 / 1 (2)/

ES-301 Control Roomll n-Plant Systems Outline Form ES-301-2 JPMSummary JPM A This is a modified JPM. This JPM is considered modified because there are existing JPMs where an Emergency Diesel Generator is started during a loss of all AC power in the exam bank, but there are not any with the malfunctions that are in this JPM. The candidate will be placed in a loss of all AC power when the Control Room Supervisor directs them to start the "N Emergency Diesel Generator.

This is an alternate path JPM because when the Emergency Diesel Generator sta,1s the candidate will find the Emergency Diesel Generator is not being cooled. After energizing the required electrical bus and starting the required pumps, an Auxiliary Operator will call the control room and inform the Candidate the Emergency Diesel is still not being cooled. Based on the report from the Auxiliary Operator the Candidate will need to make the decision to trip the Emergency Diesel Generator.

JPM B This is a new JPM. The candidate will be placed in an inadequate core cooling situation and the crew has just reduced Reactor Coolant System pressure to allow the Safety Injection Accumulators to inject. The Control Room Supervisor will direct the candidate to isolate the "An and "B" Safety Injection Accumulators.

The "An Safety Injection Accumulator Isolation Valve will not close. The alternate path will consist of the candidate having to vent the "An Safety Injection Accumul;ator.

JPM C This is a modified JPM. This .IPM is considered modified because there are existing JPMs where a Reactor Coolant Pump is required to be started, but there are not any with the malfunctions that are in this JPM. The candidate will be placed in a plant startup. While heating up from cold shutdown to hot shutdown following a refueling outage the Control Room Supervisor will direct the candidate to start the "An Reactor Coolant Pump. During the Reactor Coolant Pump start, the operator will have to control Reactor Coolant System pressure using the Letdown Line Pressure Control Valve and coordinating with another board operator to control charging line flow. Shortly after starting the Reactor Coolant Pump, the bearing oil system will loose oil and require the pump to be tripped. JPM D This is a modified JPM. This JPM is considered modified because there are existing JPMs where the performance of the immediate actions of AP-CR.1 is required, but there are not any with the malfunctions that are in this JPM. The operator will take the watch as the Head Control Operator.

As soon as the JPM begins the examiner will inform the Candidate that noxious fumes are in the Control Room. The candidate will begin performing the immediate actions of the control room evacuation procedure.

Each immediate action step performed by the candidate will require alternate path actions due to eqUipment failure .

• IPM E This is a modified JPM. This JPM is considered modified because there are existing JPMs where L TOP is placed in service, but there are not any with the malfunctions that are in this .IPM. The candidate will be placed in a plant shutdown.

While cooling down from Hot Shutdown to Cold Conditions, the crew is ready to place L TOP in service. The Shift Manager will direct the candidate to align one Pressure Operated Relief Valve for Low Temperature Over Pressure control. .IPM F This is a new JPM. The candidate will be placed in a plant shutdown.

While cooling down from Hot Shutdown to Cold Shutdown the crew will be required to borate to cold shutdown using rapid boration.

The candidate will be directed by the Control Room Supervisor to coordinate with the AUXiliary Operators to perform the rapid boration.

JPM G This is a new .IPM. The candidate will be in a Design Basis Loss of Coolant Accident.

After the Containment Recirculation Fan Coolers and Containment Spray System have reduced Containment pressure to less than 4 PSIG, the candidate will be directed to secure two Containment Spray Pumps JPM I This is a modified JPM. This JPM is considered modified because there are existing JPMs where a Gas Decay Tank is released, but there are not any with the malfunctions that are in this JPM. The candidate will be informed that the plant is operating at 100% power and the "D" Gas Decay Tank is full. When given the order to release the "D" Gas Decay Tank, they will go into the Auxiliary Building, a Radiological Controlled Area, and manually release the tank. After the release has been initiated, the Control Room will contact the Candidate and inform them the radiation monitor monitoring the release is in alarm. Based on the Control Room's report, the Candidate will secure the release. JPM J This is a Bank JPM. The candidate will be informed that a fire has occurred on the intermediate floor of the Auxiliary Building, a Radiological Controlled Area. The Shift Manager will direct the candidate to secure ventilation in the intermediate floor of Auxiliary Building.

Control Room/In-Plant Systems Outline Form ES-301-2 JPM This JPM was on the 2008 NRC Exam. The candidate will be informed that the Control Room has been evacuated due to a fire. The Shift Manager will direct the Candidate to locally operate the "AU Atmospheric Relief Valve. After initially opening the "AU Atmospheric Relief Valve the Shift Manager will direct the Candidate throttle closed the valve. While throttling the valve it will become stuck. The Candidate will have to isolate the "AU Atmospheric Relief Valve's root valve ..

Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Date of Examination:

6/21/10 Exam Level: SRO Operating Test No.: NRC Control Room Systems@ (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U, including 1 ESF) Type Code* Safety System 1 JPM Title Function a. 064 Emergency Diesel Generator S,A,M,E,EN 6 Start an Emergency Diesel Generator During a Loss of All AC Power b. 006 Emergency Core Cooling System S,A,E,EN,N 2 In accordance with FR-C.1, RESPONSE TO INADEQUATE COOLING, Isolate an SI Accumulator That Has i c. 003 Reactor Coolant Pump I i S,D,L 4P Start a Reactor Coolant Pump During a Plant Startup I In-Plant Systems@ (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U) i. 071 Waste Gas Disposal System A,M,R 9 Release the D Gas Decay Tank j. APE 067 Plant Fire On Site D,E,R 8 Secure Ventilation System During a Fire I 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-II SRO-U (A)lternate path 4-6/4-6/2-3 (C)ontrol (D)irect from bank :$ 9 / :::; 8 / :<:; 4 (E)mergency or abnormal in-plant 1 11 11 (EN)gineered safety feature -/ -1 1 (2){control room (L)ow-Power 1 Shutdown (N)ew or (M)odified from bank including 1 (A) 21 2 1'2. 1 (P)revious 2 exams :::; 3 / .,; 3 / :<:; 2 (O)(randomly

{R)CA ;;;: 1 / 1 J;;;: 1 ES-301 Control Roomlln-Plant Systems Outline Form ES-301-2 JPMSummary JPM A This is a modified JPM. This JPM is considered modified because there are existing JPMs where an Emergency Diesel Generator is started during a loss of all AC power in the exam bank, but there are not any with the malfunctions that are in this JPM. The candidate will be placed in a loss of all AC power when the Control Room Supervisor directs them to start the UN Emergency Diesel Generator.

This is an alternate path JPM because when the Emergency Diesel Generator starts the candidate will find the Emergency Diesel Generator is not being cooled. After energizing the required electrical bus and starting the required pumps, an Auxiliary Operator will call the control room and inform the Candidate the Emergency Diesel is still not being cooled. Based on the report from the Auxiliary Operator the Candidate will need to make the decision to trip the Emergency Diesel Generator.

JPM B This is a new JPM. The candidate will be placed in an inadequate core cooling situation and the crew has just reduced Reactor Coolant System pressure to allow the Safety Injection Accumulators to inject. The Control Room Supervisor will direct the candidate to isolate the "AU and "B" Safety Injection Accumulators.

The "A" Safety Injection Accumulator Isolation Valve will not close. The alternate path will consist of the candidate having to vent the "AU Safety Injection Accumulator.

JPM C This is a bank JPM. While heating up from cold shutdown to hot shutdown following a refueling outage the Control Room Supervisor will direct the candidate to start the "AU Reactor Coolant Pump. During the Reactor Coolant Pump start, the operator will have to control Reactor Coolant System pressure using the Letdown Line Pressure Control Valve and coordinating with another board operator to control charging line flow. JPM I This is a modified JPM. This JPM is considered modified because there are existing .IPMs where a Gas Decay Tank is released, but there are not any with the malfunctions that are in this JPM. The candidate will be informed that the plant is operating at 100% power and the "0" Gas Decay Tank is full. When given the order to release the "Du Gas Decay Tank, they will go into the Auxiliary Building, a Radiological Controlled Area, and manually release the tank. After the release has been initiated, the Control Room will contact the Candidate and inform them the radiation monitor monitoring the release is in alarm. Based on the Control Room's report, the Candidate will secure the release. JPM J This is a bank -'PM. The candidate will be informed that a fire has occurred on the intermediate floor of the Auxiliary Building, a Radiological Controlled Area. The Shift Manager will direct the candidate to secure ventilation in the intermediate floor of Auxiliary Building.

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

Ginna Scenario No.: 1 Op Test No.: N10-1 Examiners:

Operators: (SRO) (RO) (BOP) Initial Conditions:

The plant is at 49% power (BOL). The plant power was reduced several days ago due to a malfunction on the A MFW Pump. Corrective Maintenance has been completed, and the pump is ready to be restarted.

RG&E Energy Control Center has requested that the electric plant be aligned to a 0/100 configuration on circuit 7T to allow the RG&E personnel to perform an insulator inspection on the 767 Line. Per Chemistry direction Normal is at 60 gpm. Turnover:

The following equipment is Out-Of-Service:

Replacement.

The B AFW Pump is OOS for Bearing Event No. Malt. No. Event Type* Event Description 1 2 3 4 NA EDS07B PZR02D TUR05E TUR09D N-BOP N-SRO C-RO C-BOP C(TS)-SRO I-RO I-BOP I(TS)-SRO R-RO C-BOP C-SRO Shift Electric Plant Loss of B Instrument Bus Pressurizer Pressure (PT -449) fails High Main Turbine High Vibration/EHC control failure 5 FDW09A M-RO M-BOP M-SRO Feed Line Rupture Inside Containment 6 7 TUR02 RPS07K C-BOP C-BOP Main Turbine Failure to Auto Trip A AFW Pump Fails after start 8 FDW42A FDW15B C-RO Standby AFW fails to function * (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor -1 Appendix D Scenario Outline Form ES-D-1 Ginna 2010 NRC Scenario #1 The plant is at 49% power (BOL). The plant powm was reduced several days ago due to a malfunction on the A MFW Pump. Corrective Maintenance has been completed, and the pump is ready to be restarted.

RG&E Energy Control Center has requested that the electric plant be aligned to a 0/100 configuration on circuit 7T to allow the RG&E personnel to perform an insulator inspection on the 767 Line. Per Chemistry direction Normal Letdown is at 60 gpm. The following equipment is Out-Of-Service:

The B AFW Pump is OOS for Bearing Replacement.

Shortly after taking the watch, the operator will shift the Electric Line-up from SO/50 to 0/100 in accordance with 0-6.9.2, Establishing and/or Transferring Offsite Power to Bus 12A112B. Shortly afterwards, a loss of the B Instrument Bus will occur. The operator will respond in accordance with AR-E-14, LOSS B INSTR. BUS. Power will be restored to the bus per guidance in ER-INST.3, Instrument Bus Power Restoration, which will include the isolation and re-establishment of Normal Letdown in accordance with S-3.2E, Placing In or Removing From Service Normal Letdown/Excess Letdown. The operator will address two additional MCB Annunciators; AR-E-20, CNMT OR PLANT VENT RAD MON PUMP TRIP, and AR-F-6, PRESSURIZER HEATER BREAKER TRIP, while restoring from the transient.

The operator will address Technical Specification 3.8.7, AC Instrument Bus Sources Modes 1-4, and 3.8.9, Distribution Systems -Modes 1,2, 3 and 4. After this, the controlling Pressurizer Pressure Transmitter will fail High, causing the Spray Valves to open. The operator will respond in accordance with AR-F-2, PRESSURIZER HIGH PRESS 2310 PSI and AR-F-10, PRESSURIZER LO PRESS 2205 PSI, and enter AP-PZR.1, Abnormal PZR Pressure.

AP-PZR.1 will refer the operator to ER-INST.1, Reactor Protection Bistable Defeat After Instrumentation Loop Failure, for the defeat of PT-449. The operator will address Technical Specification 3.4.1, RCS Pressure, Temperature, and Flow Departure From Nucleate Boiling (DNB) Limits; 3.3.1, Reactor Trip System (RTS) Instrumentation; and 3.3.2, ESFAS Instrumentation.

Following this, a turbine high vibration condition on Bearing #5 will develop within about 60 seconds, and an EHC failure will occur causing the turbine to shift to manual. The operator will respond in accordance with AR-*1-27, ROTOR ECCENTRICITY OR VIBRATION; and enter AP-TURB.3, Turbine Vibration; and then AP-TURB.5, Rapid Downpower.

The operator will need to lower the Turbine Load using Manual EHC control. During the load reduction, a feed line rupture inside Containment will occur. The Reactor will trip and Safety Injection will actuate causing the operator to enter E-O, Reactor Trip or Safety Injection.

Auto turbine trip will fail to occur, but manual trip will be successful.

On the Reactor Trip the A AFW Pump will fail to Autostart, then trip after it is manually started, and the TDAFW Pump will triip on overspeed.

The operator will transition from E-O to FR-H.1, Response to a Loss of Secondary Heat Sink. -2 Appendix D Scenario Outline Form ES-D-1 The operator will unsuccessfully attempt to place the Standby AFW System in service, and then attempt to restore a Secondary Heat Sink using the MFW System. Once the Secondary Heat Sink is re-established using MFW, the operator will transition back to O, and then transition to E-2, Faulted Steam Generator Isolation.

The scenario will terminate at Step 9 of E-2, after the crew has determined that a transition to E-1, Loss of Reactor or Secondary Coolant, is required.

Critical Tasks: FR-H.1 Establish feedwater flow into at least one Steam Generator before RCS Bleed and Feed is required.

Safety Significance:

Failure to establish feedwater flow into at least one Steam Generator results in the crew having to rely upon the lower-priority action of having to initiate RCS Bleed and Feed to minimize the possibility of core uncovery.

Failure to perform this task, when able to do so, constitutes incorrect performance that leads to degradation of the RCS and/or fuel cladding fission product barriers.

E-2 A Isolate the Faulted Steam Generator before transitioning out of E-2. Safety Significance:

Failure to isolate a Faulted SG that can be isolated causes challenges to the Critical Safety Functions that would not otherwise occur. Failure to isolate flow could result in an unwarranted Orange or Red Path condition on RCS Integrity, Subcriticality (if cooldown is allowed to continue uncontrollably) and/or Containment (if the break is inside Containment).

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

Ginna Scenario 1\10.: 2 Op Test No.: N10-1 Examiners:

Operators: (SRO) (RO) (BOP) Initial Conditions:

The Plant is at 100% power (EOl). Per the daily work schedule, CROI-7, Swapping Service Water Pumps, is to be p43rformed this shift, swapping to A and D Service Water pumps. Turnover: The following equipment is Out-Of-Service:

The B SI Pump is OOS for Bearing Replacement.

Event Malf. No. Event Type* Event No. Description 1 ClG09C N-BOP Swap Service Water Pumps/D Service Water Pump Trip ClG01D C(TS)-SRO 2 CVC10A I-RO VCT level 112 Fails HIGH I-SRO 3 CVC12C C-RO C Charging Pump trips C-SRO 4 TUR16B I-BOP Turbine Impulse Pressure (PT-486) fails High I-SRO 5 A-FDW30 R-RO A MFW Pump Oil Sump HI-lO level/Rapid DownpowerlStuck Rod C-BOP G11 C(TS)-SRO 6 SGN04B M-RO SGTR M-BOP M-SRO 7 RPS07A C-RO A SI Pump fails to start in AUTO RPS07C C SI Pump fails to start in AUTO RPS07D * (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor -1 Appendix D Scenario Outline Form ES-D-1 Ginna 2010 NRC Scenario #2 The Plant is at 100% power (EOL). Per the dailly work schedule, CROI-7, Swapping Service Water Pumps, is to be performed this shift, swapping to A and D Service Water pumps. The following equipment is Out-Of-Service:

The B SI Pump is OOS for Bearing Replacement.

Shortly after taking the watch, the operator will start the A Service Water pump and then stop the B Service Water Pump in accordance with P-17, Operations Control Room Operating Instructions.

Afterwards, the operator will start the D Service Water pump and then stop the C Service Water Pump. When operator stops the C Service Water Pump, its Discharge Check Valve will stick Open, and the pump shaft will rotate backwards.

The operator will restart the C Service Water Pump, and the D Service Water Pump will trip. The operator will respond in accordance with AR-J-9, SAFEGUARDS BREAKER TRIP. The operator will address Technical Specification 3.7.8, Service Water (SW) System. Subsequently, VCT Level transmitter LT-112 will fail High. The operator will respond in accordance with AR-A-2, VCT LEVEL 14 % 86, and enter ER-CVCS.1, Reactor Makeup Control Malfunction.

Following this, the C Charging Pump will trip. The operator will respond in accordance with AR-G-25, MOTOR OFF CTR SECT PMPS EXCEPT MAIN & AUX FEED PMPS and start the B Charging Pump. Then, Main Turbine 1 st Stage Pressure Instrument PT-486 will fail High. The operator will respond in accordance with AR-G-22, ADFCS SYSTEM TROUBLE, and enter INST.1, Reactor Protection Bistable Defeat After Instrumentation Loop Failure. After this, the MFW pump Oil Sump HI-LO level annunciator will alarm (L-21). The operator will respond in accordance with AR-L-21, MAIN FEEDWATER PUMP OIL SUMP HI-LO LEVEL, enter AP-TURB.5, Rapid Downpower, and remove the affected MFP from service. During the downpower Control Rod G-11 will stick in its original position.

The operator will respond in accordance with AR-C-5, PPCS ROD SEQUENCE OR ROD DEVIATION/PPCS LTOP HI-LOW TEMPERATURE, and enter AP-RCC.2, RCC/RPI Malfunction.

The operator will address Technical Specification 3.1.4, RCC/RPI Malfunction, and 3.2.4, Quadrant Power Tilt Ratio. Shortly afterwards, a Steam Generator Tube Hupture will occur on the B Steam Generator.

The operator will respond in accordance with AR-PPCS-R47AR, SGTL INDICATED, and enter AP-SG.1, Steam Generator Tube Leak. Ultimately, the leak will be determined to be beyond the capability of the Charging Pump capacity, and the Reactor will trip and Safety Injection will actuate. The operator will enter E-O, Reactor Trip or Safety Injection.

On the Safety Injection, the A and C SI Pumps will fail to start in Auto, and the operator will be required to manually start the pumps. -2 Appendix D Scenario Outline Form ES-D-1 The operator will transition from E-O to E-3, Steam Generator Tube Rupture, isolate the flow into and out of the B Steam Generator, and then conduct a plant cooldown and depressurization.

The scenario will terminate at Step 36 of E-3, after the crew has controlled ReS pressure and charging flow to minimize primary-to-secondary leakage. Critical Tasks: E-OI Establish flow from at least one high-head ECCS Pump before transition out of O. Safety Significance:

Failure to start at least two SI Pumps, when they are available to start, results in a violation of the Facility License Condition.

The FSAR analysis results are predicted on the assumption of a minimum ECCS flowrate, which includes two SI Pumps. Failure to perform this task will leave the plant in an unanalyzed condition.

E-3A Isolate feedwater flow into and steam flow from the ruptured SG before a transition to ECA-3.1 occurs. Safety Significance:

Failure to isolate the ruptured SG causes a loss of .1.P between the ruptured SG and the intact SGs. Upon a loss of .1.P, the crew must transition to a contingency procedure that constitutes an incorrect performance that "necessitates the crew taking compensating action which complicates the event mitigation strategy." If the crew fails to isolate steam from the SG, or feed flow into the SG the ruptured SG pressure will tend to decrease to the same pressures as the intact SGs, requiring a transition to a contingency procedure, and delaying the stopping of Res leakage into the SG. E-3C Depressurize the RCS to meet SI termination criteria before ruptured SG level reaches 100% Wide Range Level. Safety Significance:

Failure to stop the reactor coolant leakage into a ruptured SG by depressurizing the RCS (when it is possible to do so) complicates the mitigation of the event. It also constitutes a "Significant reduction of Safety Margin beyond that irreparably introduced by the scenario.

If RCS depressurization does NOT occur, the inventory in the secondary side of the ruptured SG will occur leading to water release through the SG PORV or Safety Valve, which could cause and unisolable fault in the ruptured SG.

Appendix Scenario Outline Form ES-D-1 Facility: Scenario No.: 4 Op Test No.: N10-1 Operators: (SRO) (RO)l1 (BOP) Initial The Plant is at 100% power (MOIL). It is expected that immediately after turnover the crew will swap Condensate Pumps per Maintenance Dept Work Order, and conduct routine Rod Control exercises on Control Bank D.

equipment is Out-Of-Service:

The B AFW Pump is OOS for Bearing Event Malt. No. Event Event No. Type* Description 1 CND04B P Swap Condensate Pumps (Start B, Secure C)/B Condensate C-SRO Pump Trip 2 I-RO Rod Control Exercise/MRPI Failure K7 I (TS)-SRO 3 PZR03C I-RO Pressurizer Level (L T -428) Fails HIGH I (TS)-SRO Heater Drain Pump A trips/Rapid Downpower 4 R-RO C-BOP C-SRO 5 C-BOP B SG ARV Controller (AOV-3410) fails in AUTO C-SRO STM05A M-RO MSIVs Close and SG Safeties lift/fail OPEN (1 per SG) STM05B M-BOP STM09A M-SRO STM09B TDAFW Pump Steam Supply Valves Fail to Open in AUTORPS07M C-BOP RPS07N * (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor -1 Appendix D Scenario Outline Form ES-D-1 Ginna 2010 NRC Scenario #4 The Plant is at 100% power (MOL). It is expectE!d that immediately after turnover the crew will swap Condensate Pumps in preparation for taking C Condensate Pump OOS for maintenance, and conduct routine Rod Control surveillance on Control Bank D. The following equipment is Out-Of-Service:

The B AFW Pump is OOS for Bearing Replacement.

Shortly after taking the watch, the operator will swap Condensate Pumps in accordance with T-5G, Swapping Condensate Pumps With At Least One MFP Running. After the B Condensate Pump is started, the motor will trip on overload in 60 seconds. The operator will address AR-G-8, 4KV MOTOR OVERLOAD, and AR-G-25, MOTOR OFF SECT PMPS EXCEPT MAIN & AUX FEED PMPS, and secure the evolution to swap condensate pumps. After this, the operator will conduct a routine Rod Control surveillance on Control Bank D in accordance with STP-0-1, Rod Control System. When the rods are being returned to their normal position, a MRPI detector coil stack f;ailure will occur which will result in an indication that one of the exercised Control Rods appears to have dropped into the core. The operator will respond in accordance with AR-C-14, ROD BOTTOM ROD STOP and AR-C-29, MRPI SYSTEM FAILURE, and enter AP-RCC.2, RCC/RPI Malfunction.

The operator will address Technical Specification 3.1.7, Rod Position Indication, and 3.14, Rod Group Alignment Limits. Next, Pressurizer Level Transmitter LT-428 will fail High. The operator will respond in accordance with AR-F-4, PRESSURIZER LEVEL DEVIATION

-5 NORMAL +5, and F-28, PRESSURIZER HI LEVEL CHANNEL ALERT 87%, and enter ER-INST.1, Reactor Protection Bistable Defeat After Instrumentation Loop Failure. The operator will address Technical Specification 3.3.1, Reactor Trip System Instrumentation, and 3.3.3, Post Accident Monitoring Instrumentation.

After this, the A Heater Drain Pump will trip. The operator will respond in accordance with AR-G-25, MOTOR OFF CTR SECT PMPS EXCEPT MAIN & AUX FEED PMPS, and/or AR-H-17, FEED PUMP NET POSITIVE SUCTION HEAD, and enter AP-FW.1, Abnormal MFW Pump Flow or NPSH. The clperator will reduce load to 70% in accordance with AP-TURB.5, Rapid Load Reducti()n.

During the downpower, the B Steam Generator ARV (AOV-3410)

Controller will fail in Auto such that the valve goes fully Open. The operator will respond in accordance with A-503.1, Emergency and Abnormal Operating Procedures Users Guide, and close the valve manually.

At a Turbine load of about 550 MWe, both MSIVs will inadvertently fail shut. The Reactor will trip, and the operator will enter E-O, Reactor Trip or Safety Injection.

On the plant trip one or more SG Safety Valves will open, and the lowest set valve will stick in the OPEN position on each SG. Additionally, the A AFW Pump will trip upon an automatic start signal, and the TDAFW Pump Steam Supply Valves will fail to open upon an automatic signal. The operator will be required to manually start the TDAFW Pump to restore Secondary Heat Sink. -2 Appendix D Scenario Outline Form ES-D-1 The operator will transition from E-O to E-2, Faulted Steam Generator Isolation, and then to ECA-2.1, Uncontrolled Depressurization of All Steam Generators.

The scenario will terminate at Step 16 of ECA-2.1, after the crew has correctly determined whether plant conditions meet SI Termination criteria.

Critical Tasks: E-OF Establish 230 gpm of AFW Flow to the Steam Generators before transition out of E-O, unless the transition is made to FR-H.1, and then before the RCPs are manually tripped to limit heat input to the RCS. Safety Significance:

Failure to establish a Secondary Heat Sink through the initiation of AFW flow unnecessarily challenges both the HEAT SINK and the CORE COOLING Critical Safety Functions.

Additionally, the FSAR Safety Analysis results are predicated on the assumption that at least one train of safeguards actuates and delivers a minimum amount of AFW flow to the Steam Generators.

Failure to perform this task, when the ability to do so exists, results in a violation of the Facility License Condition and places the plant in an unanalyzed condition.

ECA-2.1 A Control the AFW flowrate to 50 gpm per SG in order to minimize the RCS Cooldown rate before a severe challenge (Orange Path) develops to the integrity CSF. Safety Significance:

Failure to control the AFW flow rate to the SGs leads to an unnecessary and avoidable severe challenge to the integrity CSF. Also, failure to perform the Critical Task increases challenges to the SUBCRITICAUTY and CONTAINMENT Critical Safety Functions which otherwise would not occur. -