ML102160314

Draft - Outlines (Facility Letter Dated 4/9/2010) (Folder 2)

ML102160314
Person / Time
Site:	Indian Point
Issue date:	04/09/2010
From:	Caruso J G Operations Branch I
To:	Ferrick J Entergy Nuclear Northeast
Hansell S
Shared Package
ML092470061	List: IR 05000247/2010301 ML101950011 ML102160274 ML102160314 ML102160334 ML102170047 ML102510238 ML102640428 ML102640438 ML102640564 ML102640579 ML102640589 ML102640597
References
TAC U01787
Download: ML102160314 (43)

Text

April 19, 2010 T-75 day Outline Submittal Comments Written Exam Outline o 76, 77, 78 & 081 -KIA not SRO level topic o 9 -Pretty basic system knowledge Q 80 -borderline RO level knowledge Q 85 -Rad exposure limits GET level except under emergency conditions Q88 -verify alarms sounds like RO level topic Q39 &41 testing CCS seems like a lot of questions on this system 047 & 51 -could be SRO level topic depends on Q asked. o 53 -too simple Q 59 -maintenance of spent fuel pool may be too simplistic 092 -looks like RO topic General comment alarm response extensively tested on operating test Q 68 -industrial safety important but not license level topic Q 94 & 95 not SRO level topics 069 -knowledge of process for conducting special tests -appears more like an open reference requal topic. o 96 -ok if evaluating TS or risk Q71 -ability to use rad monitoring equip -GET level Q98 -ok for EOP condition Q99 -already tested fire protection procedure on earlier O. No comments on Operating Test Outlines Entergy Nuclear Northeast Entergy Nuclear Operations, Inc. IPEG Training P.O. Box 308 ---::-Entergy Buchanan, NY 10511 914-788-2604 April 9, 2010 Indian Point Unit No.2 Docket No. 50-247 NL-10-039 IP-TNG-10-01 Mr. Samuel J. Collins Regional Administrator Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, Pennsylvania 19406-1415

Subject:

Indian Point Unit 2 Initial Licensed Operator Examination Outline

Dear Mr. Collins,

In preparation for the Indian Point Unit 2 Reactor Operator and Senior Reactor Operator initial licensing examinations, scheduled to start on July 12, 2010, Entergy Nuclear Operations, Inc. (Entergy) is providing the enclosed Reactor Operator and Senior Reactor Operator Written Examination Outlines to Mr. John Caruso of your staff. The examination outlines are being provided in accordance with the instructions of 1021, "Operator Licensing Examination Standards for Nuclear Power Reactors," Rev 9. In accordance with 10 CFR 55.49 and the Examination Security and Integrity Considerations in Examiner Standard ES-201, Attachment 1, the attached materials should be withheld from public disclosure until after the examinations are complete.

Entergy is making no commitments in this letter. Should you have any questions regarding this matter, please contact Mr. Stephen Davis, Superintendent, fLO Operations Training at (914) 788-2904, Mr. Arthur Singer, Superintendent, LRQ Operations Training, at (914) 788-2942, Mr. Tim Jenkins, Senior Instructor Examination author at (914) 788-2630, or Mr. Charlie Kocsis, Senior Instructor Examination author at (914) 788-2065.

SincerelYC Manager, Training and Development dian Point Energy Center Signed per NUREG 1021, ES-201, C.1.g Docket No. 50-247 NL-10-039 Page 2 of 2 Certification of systematic and random NUREG-1 021, Rev. 9, Form ES-201-2, "Examination Outline Quality NUREG-1021, Rev. 9, Form ES-301-1, "RO Administrative Topics NUREG-1021 , Rev. 9, Form ES-301-1, "SRO Administrative Topics NUREG-1021, Rev. 9, Form ES-301-2, "RO Control Room/ln-Plant Systems NUREG-1 021, Rev. 9, Form ES-301-2, "SRO-I Control Room/ln-Plant Systems NUREG-1 021, Rev. 9, Form ES-301-5, Transient and Event NUREG-1021 , Rev. 9, Form ES-D-1, "Scenario 1

NUREG-1 021, Rev. 9, Form ES-D-1, "Scenario 2

NUREG-1021, Rev. 9, Form ES-D-1, "Scenario 3

NUREG-1021, Rev. 9, Form ES-D-1, "Scenario 4

NUREG-1021 , Rev. 9, Form ES-D-1, "Scenario 5

NUREG-1 021, Rev. 9, Form ES-401-2, "PWR Examination Outline" -NUREG-1021, Rev. 9, Form ES-401-3, "Generic Knowledge and Abilities Outline" -NUREG-1021 , Rev. 9, Form ES-401-4, "Record of Rejected KlAs" -cc: Document Control Desk U.S. Nuclear Regulatory Commission Mail Stop O-P1-17 Washington, DC 20555-0001 w/o Enclosures Mr. Samuel Hansell Jr. Chief, Operational Safety Branch Division of Reactor Safety Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, Pennsylvania 19406-1415 w/o Enclosures Mr. John Caruso Chief Examiner Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, Pennsylvania 19406-1415 with Enclosures Resident Inspector's Office U.S. Nuclear Regulatory Commission Indian Point Unit 2 P.O. Box 59 Buchanan, N.Y. 10511-0059 w/o Enclosures Description of program used to generate IPEC Unit 2 July 2010 Written Exam KlAs Generated the RO and SRO sample plan using the "NKEG" Database Program, version 1.1, developed by Westinghouse Electric Company. This program will automatically produce a Random Sample Plan based on NUREG 1122, Rev. 2, Supplement 1 KlAs. K1As were suppressed prior to the outline generation process as provided for in the examiner standard, the list of suppressed KlAs is provided as required by the examiners standard.

Inappropriate and inapplicable KlAs were discarded during the outline development process and are included in the record of rejected KJAs. The replacement KJAs were replaced using the random sample function of the NKEG database program. NUREG 1021 ES-401 PWR Examination Outline Form ES-401-2 Facllitll:

Indian Point Unit 2 Printed: 03/16/2010 Date Of Exam: 07/12/2010 Generic Knowledge And RO KIA Category Points SRO-Only Points Tier 1. Group 1 7tW: K4 K5 K6 A1 3 A2 3 A3 A4 G* 3 Total 18 A2 3 G* 3 Total 3 Emergency

& 2 1 2 1 N/A 1 2 N/A =1 9 2 2 4 Abnormal Tier Plant Evolutions Totals 4 5 4 4 5 5 27 5 5 10 2. 1 3 2 3 3 3 2 3 3 2 2 2 28 3 2 5 Plant 2 1 1 1 1 1 1 1 1 1 1 0 10 0 I 2 1 3 Systems 4 3 4 4 4 3 4 4 3 3 2 38 5 3 8 1 2 3 4 1 2 3 4 3. 7 Abilities Categories 2 3 3 2 2 2 1 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 (Le., 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. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-speCific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate KiA statements. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution. Absent a plant-specific priority, only those KiAs having an importance rating (IR) of 2.5 or higher shall be selected.

Use the RO and SRO ratings for the RO and SRO-only portions, respectively. Select SRO topics for Tiers 1 and 2 from the shaded systems and KiA categories. The generic (G) 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 KfAs. On the following pages, enter the KiA numbers, a brief description of each topic, the topics' importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category.

Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-onlyexams. 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 10 CFR 55.43. NUREG PWR RO/SRO Examh. .ton Outline Facility:

Indian Pont Unit 2 ES-401 E/APE # I Name I Safety Function 1 Form ES-401-2 i '(i# 000007 Reactor Trip -Stabilization I X Knowledge of the operational 3.4 I 1 Recovery 11 implications of the following concepts as they apply to the reactor trip: -Shutdown margin 000008 Pressurizer Vapor Space Accident Ability to determine and interpret the I 3.9 I 76 13 following as they apply to the Pressurizer Vapor Space Accident:

-High-pressure safety injection pump flow indicator, and controller 000009 Small Break LOCA I 3 Conduct of Operations

-Ability to I 4.2 I 81 interpret reference materials such as tables etc. 000011 Large Break LOCA I 3 Emergency ProceduresIPlan

-Knowledge 4.5 I 77 of how abnormal operating procedures are llsed in conjunction with EOPs. 000011 Large Break LOCA I 3 Ability to determine and interpret the 3.7 I 78 following as they apply to a Large Break LOCA: -Difference between overcooling and LOCA indications 0000151000017 RCP Malfunctions 14 Equipment Control -Ability to analyze I 3.1 I 3 the effect of maintenance activities, such as degraded power sources, on the status oflimiting conditions for p 000025 Loss of RHR System I 4 Knowledge of the reasons for the I 3.1 I 2 following responses as they apply to the Loss of Residual Heat Removal System: Shift to alternate 000026 Loss of Component Cooling X Knowledge of the reasons for the I 3.6 I 4 Water 18 following responses as they apply to the Loss of Component Cooling Water: The automatic actions (alignments) within the CCWS resulting from the actuation of the ESFAS NUREG 1021 2 PWR RO/SRO

_.on Outline Facility:

Indian Pont Unit 2 NRC Writteu Examiuation Outline ES-401 Emer enc lantEvolutions

-Tier 1/ Group 1 . .. Form ES-401-2 J E/APE#/Namel Safetv Function KI K2 K3 [Ai] Au:=9 I Number _ _ _ ] Imp. I Q# 000027 Pressurizer Pressure Ability to determine and interpret the I 3.7 5 System Malfunction I following as they apply to the Pressurizer Pressure Control Malfunctions:

-Actions to be taken ifPZR pressure instrument fails 000029 ATWS 11 x Knowledge of the interrelations between I 2.9 I 6 the ATWS and the following:

-Breakers, and disconnects 000038 Steam Gen. Tube Rupture Ability to determine and interpret the I 4.1 I 7 following as they apply to a SGTR: When to isolate one or more SIGs 000040 Steam Line Rupture -Excessive Ability to determine and interpret the I 4.7 I 79 Heat Transfer / 4 following as they apply to the Steam Line Rupture: -Conditions requiring a reactor 000054 Loss of Main Feedwater Ability to operate andlor monitor the 4.4 I 8 following as they apply to the Loss of Main Feedwater (MFW): -Manual startup of electric and steam-driven AFW 000055 Station Blackout I Ability to determine and interpret the I 3.4 I 9 following as they apply to a Station Blackout:

-Existing valve positioning on a loss of instrument air "",,,tE>,.,., 000056 Loss of Off-site Power I 6 ! X Knowledge of the operational I 3.1 I 10 implications of the following concepts as they apply to Loss of Offsite Power: Definition of saturation conditions, for the """tp,.,..., 000057 Loss of Vital AC Inst. Bus 16 Ability to operate andlor monitor the 3.5 11 following as they apply to the Loss of Vital AC Instrument Bus: -RWST and VCTvalves NUREG Facility:

Indian Pont Unit 2 ES-401 # I Name I Safety Function 000058 Loss 2.2.36 I Equipment Control-Ability to analyze I 3.1 I 12 the effect of maintenance activities, such as degraded power sources, on the status I of limiting conditions for 000058 Loss of DC Power /6 _2.4.46 Emergency ProcedureslPlan

-AbIlity to I 4.2 I 80 verify that the alarms are consistent with I the :e lant conditions.

000062 Loss of Nuclear Svc Water / 4 Equipment Control -Abihty to recognIze I 3.9 I 13 system parameters that are entry level conditions for Technical 000077 Generator Voltage and Knowledge of the operational I 3.3 I 14 Ixr Grid Disturbances / implications of the following concepts as they apply to Generator Voltage and Electric Grid Disturbances:

-excitation lu_W IE04 LOCA Outside Containment / 3 I X I _EK3.2 I Knowledge of the reasons for the I 3.4 I 15 following responses as they apply to the LOCA Outside Containment:

-Normal, abnormal and emergency operating procedures associated with LOCA Outside Containment W IE05 Inadequate Heat Transfer -Loss I Knowledge of the interrelations between I 3.7 I 16 of Secondary Heat Sink the Loss of Secondary Heat Sink and the following:

-Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features W/Ell Loss of Emergency Ability to operate andlor monitor the I 3.9 I 17 Recirc. following as they apply to the Loss of Emergency Coolant Recirculation: Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features NUREG PWR RO/SRO Examb..__ .ton Outline Facility:

Indian Pont Unit ES-401 Emergency and Abnormal Plant Evolutions

-Tier 11 Group 1 _____E/APE # I Name I Safety Function ---i I I ----I -

n =--i I I I n_ n_ W/E12 -Steam Line Rupture -Excessive x Knowledge of the interrelations between I 3.6 18 Heat Transfer I 4 the Uncontrolled Depressurization of all Steam Generators 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 ,mctPTYI" to the operation of the NUREG 1021 PWR RO/SRO Outline Facility:

Indian Pont Unit 2 ES-401 E/APE # I Name I Safety Function 000001 Continuous Rod Withdrawal

/1 E AA2.01 000003 Dropped Control Rod / 1 000003 Dropped Control Rod / 1 000028 Pressurizer Level Malfunction

/ 2 I _2.4.31 _2.2.22 2.2.25 000032 Loss of Source Range NI / 7 _AA2.03 000033 Loss of Intermediate Range NI / 7 I I X _AAl.03 000024 Emergency Boration / 1 _AA2.02 000036 Fuel Handling Accident / 8 I X I _AKI.02 I Ability to determine and interpret the I 4.2 I 19 following as they apply to the Continuous Rod Withdrawal:

-Reactor tripped breaker indicator I Emergency ProceduresIPlan

-Knowledge 4.2 23 of annunciators alarms, indications, or instructions.

I Control -Knowledge of 4.7 I 82 limiting conditions for operations and limits. I Equipment Control-Knowledge of bases I 3.2 I 24 in technical specifications for limiting conditions for operations and safety limits. I Ability to determine and interpret the I 2.8 I 20 following as they apply to the Loss of Source Range Nuclear Instrumentation: Expected values of source range indication when high voltage is removed I Ability to operate and/or monitor the I 3.0 I 21 following as they apply to the Loss of Intermediate Range Nuclear Instrumentation:

-Manual restoration of )wer I Ability to determine and interpret the I 4.4 I 83 following as they apply to the Emergency Boration:

-When use of manual boration valve is needed I Knowledge of the operational I 3.4 I 22 implications of the following concepts as they apply to Fuel Handling Incidents: SDM -NUREG 1021 PWR RO/SRO Examh. _,on Outline Facility:

Indian Pont Unit 2 ES-401 Emer!!encv and Abnormal Plant Evolutions

-Tier 11 GrouD 2 Form ES-401-2 I E/APE #1 Name I Safety Function I KI I K2 I K3 I Al I A2 I G Number' KIA Topic I Imp. IQ# 000037 Steam Generator Tube Leak 13 AA2.02 I Ability to determine and interpret the 3.9 I 84 following as they apply to the Steam Generator Tube Leak: Agreement/disagreement among redundant radiation monitors W 1E03 LOCA Cooldown -Depress. 14 x I-E-K-2-.-}-+IKnowledge ofthe interrelations between I 3.6 I 25 the LOCA Cooldown and Depressurization and the following:

-Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features W 1E06 Inad. Core Cooling I 4 x EK2.2 Knowledge of the interrelations between I 3.8 I 26 the Degraded Core Cooling 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 W IE08 RCS Overcooling

-PTS / 4 x EK3.1 Knowledge of the reasons for the I 3.4 I 27 following responses as they apply to the Pressurized Thermal Shock: -Facility operating characteristics during transient conditions, including coolant chemistry and the effects of temperature, pressure, and reactivity changes and operating limitations and reasons for these operating characteristics W/E13 Steam Generator Over-pressure

/ 2.3.4 Radiological Controls -Knowledge of I 3.7 I 85 4 radiation exposure limits under normal and emenzencv conditions.

NUREG 1021 PWR RO/SRO ExaL _.ation Outline Facility:

Indian Pont Unit 2 ES-401 Syste11:1 c:.::.-.:..:::::==-

___L.::::.:J 003 Reactor Coolant x Knowledge of the physical connections I 3.0 I 28 Pump and/or cause-effect relationships between the RCPS and the following systems: CCWS 004 Chemical and x Knowledge of the operational I 2.8 I 29 Volume Control implications of the following concepts as they apply to the CVCS: -Relationship ___+I SUR and 004 Chemical and Ability to (a) predict the impacts of the 3.9 I 86 Volume Control following malfunctions or operations on the CVCS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

-Low RWST 005 Residual Heat x Ability to predict and/or monitor changes I 3.3 I 30 Removal in parameters (to prevent exceeding design limits) associated with operating the RHRS controls including:

-Detection of and response to presence of water in RHR 005 Residual Heat Ability to manually operate and/or I 3.1 I 31 Removal monitor in the control room: -Controls and indication for closed cooling water 006 Emergency Core Ability to monitor automatic operation of I 4.1 I 32 Cooling the ECCS, including:

-Pumps 006 Emergency Core x Knowledge of the operational I 2.9 I 33 Cooling implications of the following concepts as they apply to the ECCS: -Brittle fracture; causes and oreventative actions NUREG 1021 PWR RO/SRO Exa.aation Outline Facility:

Indian Pont Unit 2 ES-401

.. . I -----I -_. u_ I u_ I I I 007 Pressurizer I Knowledge of the physical connections I 2.9 I 34 Relief/Quench and/or cause-effect relationships between the PRTS and the following systems: Containment

"",,,,It,,,..,, 008 Component Equipment Control --A--=-b--:-:i1-c-ity-to---CCde-t-erm-i:-n-e-+I-'3-.6-,-t1--:-3 sc:--ll Cooling Water operability and/or availability of safety related 008 Component 1-----+-'1 Ability to (a) predict the impacts of the I 3.3 I 36 Cooling Water following malfunctions or operations on the CCWS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

-PRMS alarm 010 Pressurizer I X Knowledge ofthe effect of a loss or I 3.2 I 37 Pressure Control malfunction of the following will have on r:--:-:c--:---c:--I1tl1e PZR PCS: -PZRsprays and heaters 010 Pressurizer Conduct of Operations

-Ability to I 4.4 I 87 Pressure Control perform specific system and integrated plant procedures during all modes of plant 012 Reactor Protection I I X Knowledge of the operational I 3.1 I 38 implications of the following concepts as to the RPS: Power 012 Reactor Emergency ProceduresIPlan

-Ability to I 4.0 I 88 verify system alarm setpoints and operate controls identified in the alarm response manual. 013 Engineered Safety I X Knowledge of the effect of a loss or 2.7 40 Features Actuation malfunction of the following will have on L-___.L 1

___,,--_-,---_--, NUREG PWR RO/SRO Exa.__.J.ation Outline Facility:

Indian Pont 022 Containment

• K3.01 IX I I I I

• 022 Containment .K4.04 IX I I I I

• 026 Containment X K2.01 Spray 039 Main and K4.05Steam 059 Main Feedwater Ix I K3.02 059 Main Feedwater

.-A2.03 061 Al.OIEmergency Feedwater I I I I I Knowledge of the effect that a loss or malfunction of the CCS will have on the following:

-Containment equipment subject to damage by high or low and --'Knowledge ofCCS design feature(s) and/or interlock(s) which provide for the following:

-Cooling of control rod drive motors . Knowledge of bus power supplies to the following:

-Containment spray pumps Knowledge ofMRSS design feature(s) and/or interlock(s) which provide for the following:

-Automatic isolation of steam line I Knowledge of the effect that a loss or malfunction of the MFW System will have on the following:

-AFW Y I Ability to Ca) predict the impacts of the following malfunctions or operations on the MFW System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: eventAbility to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the AFW System controls including: S/G level I 2.9 I I 2.8 I I 3.0 I I 3.7 I I 3.6 I I 3.1 I I 3.9 I NUREG PWR RO/SRO Ex:... .nation Outline Facility:

lndian Pont Unit 2 ES-401 062 AC Electrical Distribution 062 AC Electrical Distribution 063 DC Electrical Distribution 064 Emergency Diesel Generator 073 Process Radiation Monitoring 076 Service Water 076 Service Water NUREG I Ability to (a) predict the impacts of the* A2.06 following malfunctions or operations on the A.C. Distribution System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

-Keeping the safeguards I Ability to predict and/or monitor IX *

• Al.OI in parameters (to prevent exceeding design limits) associated with operating the A.C. Distribution System controls including:

-Significance of DIG load limits I Ability to manually operate

• A4.03 IXI monitor in the control room: -Battery rateI I Knowledge of the physical connections

• Kl.05and/or cause-effect relationships between the ED/G System and the following

-aIr . IX I

• K3.01 Knowledge of the effect that a loss or malfunction of the PRM System will have I I I I I on the following:

-Radioactive effluent

.-_ 2.4.25 I Emergency Procedure sIP Ian -

of fire .. ..A2.01 I Ability to (a) predict the impacts of the following malfunctions or operations on the SWS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

-Loss 11 I 3.4 I 47 I 3.4 I 46 I 3.4 I I 3.6 I 3.3 3.5 I 51 PWR RO/SRO Ex...__.nation Outline Facility:

Indian Pont Unit 2 . -078 Instrument Air IX _ K2.01 I Knowledge of bus power supplies to the 2.7 53 -Instrument air -078 Instrument Air _X _ A3.01 I Ability to monitor automatic operation of 3.1 54 the lAS, including:

-Air Eressure _ K4.06 -103 Containment IX I Knowledge of Containment System 3.1 55 design feature(s) and/or interlock(s) which provide for the following: Containment isolation . _ A2.02 103 Containment I Ability to (a) predict the impacts of the 3.2 I 90 following malfunctions or operations on the Containment System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

-Necessary plant conditions for work in containment NUREG 1021 PWR RO/SRO ...lation Outline Facility:

Indian Pont Unit 2

• 002 Reactor Coolant -__ A2.01 I Ability to (a) predict the impacts of the I 4.3 I 56 following malfunctions or operations on the RCS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

-Loss of coolant inventory 011 Pressurizer Level I Ability to predict and/or monitor changes 3.3 I 58 IX *in parameters (to prevent exceeding design limits) associated with operating the PZR LCS controls including:

• and letdown flows 015 Nuclear K2.01 I Knowledge of bus power supplies to the I 3.3 I 57 Instrumentation I IX I I I I I following:

-NIS channels, components, and interconnections 033 Spent Fuel I Knowledge of Spent Fuel Pool Cooling I 2.5 I 59* K4.02 IX ISystem design feature(s) and/or I I I I I .X interlock( s) which provide for the following:

-Maintenance of spent fuel cleanliness 034 Fuel I Ability to monitor automatic operation of I 2.9 I 60* A3.03 the Fuel Handling System, including: I I I I I I I I flux at shutdown 035 Steam Generator IX I _ K6.01 I Knowledge of the effect of a loss or I 3.2 I 61 malfunction of the following will have on -the S/GS: -MSNs 041 Steam I K3 Knowledge of the effect that a loss or I 3.8 I 62I <02 DumplTurbine Bypass malfunction of the SDS will have on the Control following:

-RCS NUREG

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PWR RO/SRO Ex.._.fnation Outline Facility:

Indian Pont Unit 2 045 Main Turbine IX I Generator

• 045 Main Turbine Generator I

• I 055 Condenser Air Removal

• 071 Waste Gas

• IX Disposal I I I I I I I 072 Area Radiation IX I Monitoring I I I I I

• I 086 Fire Protection

-NUREG 1021 I Knowledge of the operational I 2.7 I 65

• K5.23 implications of the following concepts as they apply to the MTIG System: Relationship between rod control and RCS boron concentration during TIG load increases I Ability to (a) predict the impacts ofthe 2.9 91

• A2.17 following malfunctions or operations on the MT IG System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Malfunction of control I Emergency ProcedureslPlan

-Ability to I 4.0 I 92.2.4.50 verifY system alarm setpoints and operate controls identified in the alarm response manual.I Ability to manually operate and/or I 3.1 I 63

• A4.26 monitor in the control room: -Authorized waste gas release, conducted in compliance with radioactive gas discharge nit I Knowledge of the physical connections 13.1 1 64

• Kl.OI and/or cause-effect relationships between the ARM system and the following

-Plant ventilation

_ A2.02 I Ability to (a) predict the impacts of the I 3.3 I 93 following malfunctions or operations on the Fire Protection System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or ons: -Low FPS header Facility I Indian Point Unit 2 I Date of Exam 7/12/2010 Category KIA # 2.1.26 2.1.42 1. Conduct of 2.1.25 Operations 2.1.45 i Subtotal 2.2.1 2.2.7 2. Equipment 2.2.41 Control 2.2.19 2.2.21 Subtotal NUREG 1021 Topic Knowledge of industrial safety procedures (such as rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and hydrogen).

Knowledge of new and spent fuel movement procedures.

Ability to interpret reference materials, such as graphs, curves, tables, etc. Ability to identify and interpret diverse indications to validate the response of another indication.

Ability to perform pre-startup procedures for the facility, including operating those controls associated with plant equipment that could affect reactivity.

Knowledge of the process for conducting special or infrequent tests. Ability to obtain and interpret station electrical and mechanical drawings.

Knowledge of maintenance work order requirements.

Knowledge of pre-and post-maintenance operability requirements.

RO SRO-Only IR Q# IR Q# 3.4 68 2.5 67 I 4.2 94 i 4.3 95 i 2 2 4.5 66 2.9 69 3.5 70 I 3.4 96 J 4.1 97 I 3 Facility I Indian Point Unit 2 I Date of Exam 7112/2010 Category KJA# Topic RO IR Q# IR Ability to use radiation monitoring 2.9 systems, such as fixed monitors and alarms, survey instruments, personnel monitoring equipment, etc. Ability to comply with radiation 3.5 work permit requirements normal and abnormal Knowledge of radiological safety 3.2 3. principles pertaining to licensed Radiological operator duties, such as containment Controls entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning etc. ! Knowledge of radiation or 3.8 contamination hazards that arise during normal, abnormal, emergency conditions or activities.

3 Knowledge ofEOP layout, symbols, 3.4 and Knowledge ofthe emergency plan. 3.1 Knowledge of fire protection 3.7

4. Emergency Procedures/plan 2.4.30 Knowledge of which events related 4.1 . to system operations/status I must be reported to organizations or external agencies, such as State, the NRC, or the transmission system

! 2 ! Tier 3 Point 10 NUREG

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ES-401 Record of Rejected Form ES-401-4 Tier I Randomly Selected KIA Reason for R-1/1 000009 Knowledge of the reasons for the following System does not exist at responses as they apply to the small break Monitoring of in-core T-cold R-1/1 000017 G2.1.29 Reactor Coolant Pump (RCP)

This Generic KA is not applicable to normal Knowledge of how to conduct system lineups, such as valves, breakers, switches, etc. R-1/1 000026 Knowledge of the reasons for the following There is no automatic function at responses as they apply to the Loss of Cooling The conditions that will initiate the automatic opening and closing of the SWS isolation valves to the CCW/nuclear service water coolers R-1/2 000067 AK3.04 067 Plant Fire on Site EOPs do not contain actions for Actions contained in EOP for plant fire on R-1/2 W/E01 G2.4.25 E01 Rediagnosis EOPs do not contain actions for Knowledge of fire protection R-2/1 012000 Knowledge of the effect of a loss or malfunction of Equipment not applicable at the following will have on the Core protection

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R-2/1 013000 K4.22 R-2/1 063000 K4.01 R-2/1 013000 G2.1.21 R-2/1 078000 K2.02 R-2/2 028000 AK1.02 R-2/2 033000 G2.2.40 R-2/2 079000 K4.01 R-3 2.1.9 Knowledge of ESFAS design feature(s) and/or interlock(s) which provide for the following:

Reason for shut safety injection pump discharge valve of train to be tested Knowledge of D.C. Electrical System design feature(s) and/or interlock(s) which provide for the following:

Manual/automatic transfers of control Engineered Safety Features Actuation Ability to verify the controlled procedure Knowledge of bus power supplies to the Emergency air Ability to predict and/or monitor changes parameters (to prevent exceeding design limits) associated with operating the HRPS controls including:

Containment Spent Fuel Pool Cooling System Abilitytoapply technical specifications for a Knowledge of SAS design feature(s) interlock(s) which provide for the Cross-connect with Ability to direct personnel activities inside the REJECT due to inability to write discriminatory RO level question Rejected due to similarities with RO question #48 Rejected due to inability to write discriminatory RO level question Unit 2 does not have an emergency air compressor REJECT due to inability to write discriminatory RO level question Unit 2 does not have Tech Spec for SFP Cooling Unit 2 does not have auto function/interlock for lAS & SAS ROs do not direct activities in control room

---

R-3 2.1.21 Ability to verify the controlled procedure Rejected due to inability to discriminatory RO level R-3 Knowledge of radiation monitoring systems, such as Rejected due to similarities with RO question fixed radiation monitors and alarms, portable survey #71 instruments, personnel monitoring equipment, etc. S-1/1 000022 G2.1.4 Loss of Rx Coolant Generic KA does not apply to system specific E/APE Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, "no solo" operation, maintenance of active license status, 10CFR55, etc. S-1/1 000065 G2.4.16 Loss of Instrument Rejected due to similarities with question Knowledge of EOP implementation hierarchy and coordination with other support procedures or guidelines such as, operating procedures, abnormal operating procedures, severe accident management guidelines.

5-1/2 000068 G2.4.8 Control Room Rejected due to similarities with question Knowledge of how abnormal operating procedures are used in conjunction with EOPs. S-1/2 000076 Ability to determine and interpret the following as Rejected due to over sampling of they apply to the High Reactor Coolant Activity:

monitors (See SRO questions 9 & Process effluent radiation chart recorder S-211 056000 G2.3.15 Condensate Rejected Unit 2 does not have system radiation Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

I!S3 2.4.5 Knowledge of the organization of the operating Rejected due to similarities with SRO -procedures network for normal, abnormal, and question evolutions.

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

Indian Point Unit 2 Date of Examination:

12, 2010 Exam Level: RO X SRO-I D SRO-U D 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 :a.Realign a Misaligned Rod Terminate Safety Injection after Main Steam Une Break Depressurize the RCS during SGTR using Aux Spray Transfer from AFW to Low Flow Bypass Feed. Align Recirculation Spray Restore Power to Bus 2A using 22 EDG Perform Required Actions for 23 SG Pressure (439B) Failing ! Adjust the Alarm setpoints for R-44 in preparation for a gaseous release In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U) i. Perform Required Actions to Dump Steam Locally Using the Atmospheric Steam Dump Valve for 21 SG Start the Appendix R SBO EDG Perform the Required Actions to Establish Backup Cooling to the Charging Pumps Type Code* Safety Function M,A,S 1 N,S,EN 2 D,A,S,EN 3 M, L, P, S 4-S N,S,EN 5 A,N,S 6 D,A,S 7 N,S 9 D,A, E 4-P N 6 D,R,E 8 All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; ailS SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

• Type Criteria for RO I SRO-II SRO-U Control Roomlln-Plant Systems Outline Form ES-301-2 Realign a Misaligned Control Bank Rod in accordance with 2-AOP-ROD-1, Rod Control Malfunctions.

One Control Bank Rod has become misaligned during power ascension with power level approximately 50%. When the candidate releases the HOLD-OUT switch, the control rod will continue to withdraw.

The candidate should enter 2-AOP-ROD-1, trip the reactor and perform the Immediate Operator Actions for a Reactor Trip. This is an Alternate Path JPM. This is a Modified Bank JPM Failure to properly perform this task will result in violation of TS and possible exceeding hot channel factors. Terminate Safety Injection after Main Steam Line Break. The plant experienced a steam break outside containment upstream of the MSIVs. Following the isolation of the faulted SG, the crew would transition to E-1; then, the crew would transition to ES-1.1, SI Termination using the foldout page criteria.

This procedure flowpath does not have SI, or Phase A reset prior to entry. The candidate will be required to perform all actions to Reset SI and Phase A Signals, then Terminate SI by securing the pumps. This is a new JPM. Failure to properly perform this JPM will result in SI flow continuing and possible PTS condition. Depressurize the RCS during a SGTR using Aux Spray. A SGTR of adequate size to cause an SI has occurred.

6.9 KV Bus 3 tripped on fault resulting in a loss of 23 RCP. All actions up to depressurize to refill the pressurizer and minimize break flow will have been completed.

The PORV Block Valves will be danger tagged shut and PCV-455A (Loop 24 Spray Valve) will not open. The candidate will continue in 2-E-3, Steam Generator Tube Rupture and perform depressurization using Aux Spray. This is an Alternate Path JPM. This JPM directly from the JPM bank; however, it has not been used on the previous 2 NRC Exams. Failure to properly perform this task will result in excessive loss of RCS inventory and possible SG overfill. Transfer from AFW to Low Flow Bypass Feed. The plant is at approximately 2-3% power. One MBFP has been started and is ready to provide flow to the SGs. In accordance with 2-S0P-21.1, Main Feedwater System, the candidate will transfer steam generator feedwater from the Auxiliary Feedwater System to the Main Feedwater Low Flow Bypass valves. A similar JPM was used on the last Unit 3 exam; however the method used was different from the method used in this JPM. Failure to properly perform this task will result in possible reactor trip on SG level. Align Recirculation Spray Flow. The plant has experienced a Large Break LOCA. Transfer to recirculation has been accomplished.

When the RWST has decreased to 2 feet the operating Containment Spray pump must be secured and transfer to recirculation spray flow must be accomplished in accordance with 2 -ES-1. 3 Transfer to Cold Leg Recirculation.

This JPM requires the candidate to ensure proper core flow while Recirculation Spray flow is established since the Recirculation Pumps will be providing both core cooling flow and containment spray flow. This is a new JPM. Failure to properly perform this task will result in failure to meet FSAR assumptions for Iodine removal.

Administrative Topics Outline Form ES-301-1 Facility:

Indian Point Unit 2 Examination Level: RO X SRO Date of Examination:

Jull! 12, 2010 Operating Test Number: Administrative Topic (see Note) Type Code* Describe activity to be performed Calculate Shutdown Margin Conduct of Operations N,R 2.1.25 Ability to interpret reference materials such as graphs, curves, tables etc. 2.1.19 Ability to use plant computers to evaluate system or component status Perform IR NIS COL (Control Room) Conduct of Operations N,S 2.1.29 Knowledge of how to conduct system lineups, such as valves, breakers, switches, etc. Tagout BA Transfer Pump Discharge Valve Equipment Control N,R 2.2.13 Knowledge of tagging and clearance procedures.

Determine Protective Clothing and Stay Time Radiation Control N,R 2.3.7 Ability to comply with radiation work permit requirements during normal and abnormal conditions.

Emergency Procedures/Plan Not Applicable for RO All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.

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

(A)lternate path 4-6 1 4-6 1 2-3 (C)ontrol room (D)irect from bank s9/s8/s4 (E)mergency or abnormal in-plant (EN)gineered safety feature

• 1 1 (control room system) (L)ow-Power 1 Shutdown (N)ew or (M)odified from bank including 1 (A) (P)revious 2 exams s 3 1 s 3 1 s 2 (randomly selected) (R)CA ;::1/2:1/2:1 . (S)imulator ES-301 Administrative Topics Outline Form ES-301-1 INDIAN POINT UNIT 2 NRC RO EXAMINATION CONDUCT OF OPERATIONS:

Calculate Shutdown Margin -The candidate will be given a set of conditions and asked to calculate Shutdown Margin. This is accomplished using any computer with access to the IPEC intranet to obtain current plant data from the On-Line NuPOP. The data is entered in the SDM calculation section of WRC-1.

• This is a New JPM

• RO Only CONDUCT OF OPERATIONS:

Perform IR NIS COL (Control Room) -The candidate is directed to perform the control room section of the IR NIS Check Off List. One switch will be out of position.

The candidate must identify this switch and inform the CRS.

• This is a New JPM

• RO Only EQUIPMENT CONTROL: Tagout BA Transfer Pump Discharge Valve -The candidate will be given plant prints and associated procedures and directed to prepare a manual tagout for the 21 Boric Acid Transfer Pump Discharge Valve. NOTE: Manual tagout JPMs exist in the JPM Bank; however, this component (BA pump discharge valve) is new and has not been used before.

• This is a New JPM

• ROOnly RADIATION CONTROL: Determine Protective Clothing and Stay Time -The candidate will be given an RWP for a task that must be performed in a Locked High Radiation Area. The candidate will also be given the exposure history for an individual.

The candidate must determine the appropriate Protective Clothing required for the task and the allowed stay time for the individual without exceeding administrative limits.

• This is a New JPM

• RO Only Control Room/In-Plant Systems Outline Form ES-301-2 Facility:

Indian Point Unit 2 Date of Examination:

12, 2010 Exam Level: RO D SRO-I X SRO-U Operating Test No.: Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF) System I J PM Title Realign a Misaligned Rod Terminate Safety Injection after Main Steam Line Break Depressurize the RCS during SGTR using Aux Spray Transfer from AFW to Low Flow Bypass Feed. Align Recirculation Spray NA for SROs Perform Required Actions for 23 SG Pressure Channel (439B) Failing Low Adjust the Alarm setpoints for R-44 in preparation for a gaseous release In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U) Perform Required Actions to Dump Steam Locally Using the Atmospheric Steam Dump Valve for 21 SG Start the Appendix R SBO EDG Type Code* Safety Function M,A,S 1 N,S,EN 2 D,A,S,EN 3 M, L, P, S 4-S N,S,EN 5 D,A,S 7 N,S 9 D,A, E 4-P N 6 k. Perform the Required Actions to Establish Backup Cooling to the Charging Pumps D,R,E 8 @ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

• Type Codes I Criteria for RO I SRO-II SRO-U (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power

/ Shutdown (N)ew or (M)odified from bank including 1 (A) (P)revious 2 exams (R)CA (S)imulator 4-6 / 4-6 / 2-3 S9/58/54 ;::1/;::1/;::1 -/ -/ ;::1 (control room system) ;::1/;::1/;::1

2/;

:2/;::1 S 3 / S 3 I S 2 (randomly selected)

1/;

:1/;::1 Restore power to bus 2A using 22 EDG. Bus 2A normal supply breaker will be tripped on overcurrent.

The candidate will use 2-S0P-480V-1, Loss Of Normal Power To Any 480v Bus. All 3 EDGs will have automatically started and be running unloaded.

All of the loads on the bus will be removed and a visual inspection of the bus performed (Local action). The bus will be re-energized from the control room using the EDG supply breaker. This is a New JPM. This is an Alternate path JPM. Failure to properly perform this task will result in reduction in redundant power supplies for safeguards equipment. Perform Required Actions for 23 SG Pressure Channel (439B) Failing Low (alternate Path). The affected Steam Pressure Transmitter provides density compensation for the steam flow channel used in the Steam Generator Water Level Control System. Steam Pressure failing Low will result in Steam Flow failing Low. The Immediate Operator actions will attempt to place the unaffected steam flow transmitter in service. The switch will not function (stuck contacts) requiring the candidate to take manual control of the feedwater regulating valve and controlling level. Additional actions include tripping bistables to remove the channel from service. This is a Bank JPM. This JPM has never been used on an ILO NRC exam. This is an Alternate path JPM. Failure to properly perform this task will result in loss of control of SG level and possible Reactor Trip. Adjust the Alarm setpoints for R-44 in preparation for a gaseous release. In preparation for a gaseous waste release, the Warn and Alarm setpoint for Radiation Monitor 44, Plant Vent Radio Gas, must be changed. A Gaseous Waste Release Permit calculation indicates that the Alarm and Warn setpoint must be reset prior to the actual release. The candidate must change the Alarm and Warn setpoint to the values calculated on the Release Permit. This is a New JPM. Failure to properly perform this task may result in excessive release of radioactive gas to the environment. Perform Required Actions to Dump Steam Locally Using the Atmospheric Steam Dump Valve for 21 SG. The JPM is part of the Appendix R actions. Instrument Air will not be available for the Atmospheric Steam Dump Valve. The candidate will be required to simulate connecting the alternate Nitrogen supply tank to the valve and control steam flow locally. In Plant JPM This is a Bank JPM. This JPM has not been used at Unit 2 for initial NRC exams. The Nitrogen Bottles were recently added and this JPM was written for annual requal operating exam. Failure to properly perform this task will result in inability to control RCS temperature during a control room evacuation. Start the Appendix R Emergency Diesel Generator.

USing 2-S0P-27.6, Unit 2 Appendix R Diesel Generator Operation, Start the Appendix R EDG Normal Engine Start (Parallel Mode).This is a relatively new piece of equipment.

This EDG was not installed during the last NRC exam. In Plant JPM This is a New JPM. Failure to properly perform this task will result in not supplying electrical power during control room evacuation event.

Perform the Required Actions to Establish Backup Cooling to the Charging Pumps -This JPM is part of the Appendix R actions. The control room is evacuated and CCW is not available to the charging pumps. The candidate is directed to align backup city water cooling to the charging pumps. In Plant JPM This JPM is directly from existing bank. This JPM has not been used on the previous last 2 Unit 2 Initial NRC examinations.

A similar JPM was used on the last Unit 3 Initial exam; however, the methodology is significantly different between units. Failure to properly perform this task will result in inability to maintain RCS inventory and possible core damage.

Control Room/In-Plant Systems Outline Form ES-301-2 Realign a Misaligned Control Bank Rod in accordance with 2-AOP-ROD-1, Rod Control Malfunctions.

One Control Bank Rod has become misaligned during power ascension with power level approximately 50%. When the candidate releases the HOLD-OUT switch, the control rod will continue to withdraw.

The candidate should enter 2-AOP-ROD-1, trip the reactor and perform the Immediate Operator Actions for a Reactor Trip. This is an Alternate Path JPM. This is a Modified Bank JPM Failure to properly perform this task will result in violation of TS and possible exceeding hot channel factors. Terminate Safety Injection after Main Steam Line Break. The plant experienced a steam break outside containment upstream of the MSIVs. Following the isolation of the faulted SG, the crew would transition to E-1, then, the crew would transition to ES-1.1, SI Termination using the foldout page criteria.

This procedure flowpath does not have SI, or Phase A reset prior to entry. The candidate will be required to perform all actions to Reset SI and Phase A signals, then Terminate SI by securing the pumps. This is a new JPM. Failure to properly perform this .IPM will result in SI flow continuing and possible PTS condition. Depressurize the RCS during a SGTR using Aux Spray. A SGTR of adequate size to cause an SI has occurred.

6.9 KV Bus 3 tripped on fault resulting in a loss of 23 RCP. All actions up to depressurize to refill the pressurizer and minimize break flow will have been completed.

The PORV Block Valves will be danger tagged shut and PCV-455A (Loop 24 Spray Valve) will not open. The candidate will continue in 2-E-3, Steam Generator Tube Rupture and perform depressurization using Aux Spray. This is an Alternate Path JPM. This JPM directly from the JPM bank; however, it has not been used on the previous 2 NRC Exams. Failure to properly perform this task will result in excessive loss of RCS inventory and possible SG overfill. Transfer from AFW to Low Flow Bypass Feed. The plant is at approximately 2-3% power. One MBFP has been started and is ready to provide flow to the SGs. In accordance with 2-S0P-21.1, Main Feedwater System, the candidate will transfer steam generator feedwater from the Auxiliary Feedwater System to the Main Feedwater Low Flow Bypass valves. A similar JPM was used on the last Unit 3 exam; however the method used was different from the method used in this JPM. Failure to properly perform this task will result in possible reactor trip on SG level. Align Recirculation Spray Flow. The plant has experienced a Large Break LOCA Transfer to recirculation has been accomplished.

When the RWST has to 2 feet the operating Containment Spray pump must be secured and transfer to recirculation spray flow must be accomplished in accordance with 2-ES-1.3 Transfertopold Leg Recirculation.

This JPM requires the candidate to ensure proper core flow while Recirculation Spray flow is established since the Recirculation Pumps, will be providing both core cooling flow and containment spray flow. ' This is a new JPM. Failure to properly perform this task will result in failure to meet FSAR assumptions for Iodine removal.

Not Used for SROI candidates Perform Required Actions for 23 SG Pressure Channel (439B) Failing Low (alternate Path). The affected Steam Pressure Transmitter provides density compensation for the steam flow channel used in the Steam Generator Water Level Control System. Steam Pressure failing Low will result in Steam Flow failing Low. The Immediate Operator actions will attempt to place the unaffected steam flow transmitter in service. The switch will not function (stuck contacts) requiring the candidate to take manual control of the feedwater regulating valve and controlling level. Additional actions include tripping bistables to remove the channel from service. This is a Bank JPM. This .IPM has never been used on an ILO NRC exam. This is an Alternate path JPM. Failure to properly perform this task will result in loss of control of SG level and possible Reactor Trip. Adjust the Alarm setpoints for R-44 in preparation for a gaseous release. In preparation for a gaseous waste release, the Warn and Alarm setpoint for Radiation Monitor 44, Plant Vent Radio Gas, must be changed. A Gaseous Waste Release Permit calculation indicates that the Alarm and Warn setpoint must be reset prior to the actual release. The candidate must change the Alarm and Warn setpoint to the values calculated on the Release Permit. This is a New JPM. Failure to properly perform this task may result in excessive release of radioactive gas to the environment. Perform Required Actions to Dump Steam Locally USing the Atmospheric Steam Dump Valve for 21 SG. The JPM is part of the Appendix R actions. Instrument Air will not be available for the Atmospheric Steam Dump Valve. The candidate will be required to simulate connecting the alternate Nitrogen supply tank to the valve and control steam flow locally. In Plant JPM This is a Bank JPM. This JPM has not been used at Unit 2 for initial NRC exams. The Nitrogen Bottles were recently added and this JPM was written for annual requal operating exam. Failure to properly perform this task will result in inability to control RCS temperature during a control room evacuation. Start the Appendix R Emergency Diesel Generator.

USing 2-S0P-27.6, Unit 2 Appendix R Diesel Generator Operation, Start the Appendix R EDG Normal Engine Start (Parallel Mode).This is a relatively new piece of equipment This EDG was not installed during the last NRC exam. In Plant JPM This is a New JPM. Failure to properly perform this task will result in not supplying electrical power during control room evacuation event. Perform the Required Actions to Establish Backup Cooling to the Charging Pumps -This JPM is part of the Appendix R actions. The control room is evacuated and CCW is not available to the charging pumps. The candidate is directed to align backup city water cooling to the charging pumps. In Plant JPM This JPM is directly from existing bank.

This .IPM has not been used on the previous last 2 Unit 2 Initial NRC examinations.

A similar JPM was used on the last Unit 3 Initial exam; however, the methodology is significantly different between units. Failure to properly perform this task will result in inability to maintain RCS inventory and possible core damage.

Administrative Topics Outline Form ES-301-1 Facility:

Indian Point Unit 2 Date of Examination:

12, 2010 Examination Level: ROO SRO X Operating Test Number: Administrative Topic Type Describe activity to be performed (see Note) Code* Review WCR-1 N,R 2.1.25 Ability to interpret reference materials, such as Conduct of Operations graphs, curves, tables, etc. Determine Location for Spent Fuel Assembly N,R 2.1.42 Knowledge of new and spent fuel movement Conduct of Operations procedures.

Initiate a Temporary Procedure Change N,R 2.2.6 Knowledge of the process for making changes to Equipment Control procedures.

Review/Approve a Liquid Radiation Release Permit M,R,P 2.3.11 Ability to control radiation releases.

Radiation Control Classify Security Event 7 ,D Knowledge of procedures relating to a security Emergency Procedures/Plan event (non-safeguards information).

D All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.

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

ES-301 Form ES-301-1 INDIAN POINT UNIT 2 NRC EXAMINATION CONDUCT OF OPERATIONS:

Review WCR-1 The candidate will be given a copy of WCR-1 , Reactivity Summary Sheet prepared by an RO for review. The candidate will review the calculation and find an error. The candidate should NOT sign Reviewed By and return the form.

• This is a New JPM

• SRO Only CONDUCT OF OPERATIONS:

Determine Location for Spent Fuel Assembly This JPM gives the candidate a spent fuel assembly with initial enrichment and burnup. The spent fuel assembly must be moved in the pit. The candidate must determine the spent fuel pit location zone.

• This is a New JPM

• The SRO Only EQUIPMENT CONTROL: Initiate a Temporary Procedure Change. This JPM will initiate a Temporary Procedure Change for 2-AOP-ROD-1.

The procedure is missing a step to place the P-A converter in Manual before depressing the UP-Down buttons. (Note this step does exist in other locations in the procedure)

• This is a New JPM

• SRO Only. RADIATION CONTROL: Review a Liquid Radiation Release Permit. This JPM has modified values from the existing bank version. The permit will have inaccurate information.

• This is a Modified Bank JPM

• SRO Only. EMERGENCY PROCEDURES/PLAN:

Respond to a Security Event. The candidate will be notified that a security event is in progress.

The candidate must respond in accordance with AOP-SEC-1 and classify the event. Once the event is classified, the candidate will have 15 minutes to complete the NY State Part 1 form.

• This is a Modified Bank JPM.

• SRO Only.

Facility:

Indian Point 2 Scenario Op-Test No.:

Operators:

Initial Reset simulator to IC-118 Load Simulator The Plant is in a 100% normal full power PORV PCV-456 and associated block valve are tagged out due to PCV-456 blowing Turnover:

Maintain 100% Power Critical Tasks: Establish greater than 400 gpm AFW flow to the SGs before transition out of E-O or tripping the RCPs in the FR-H.1. (E-O--F) Isolate Feedwater flow to and steam flow from the ruptured SG before transition to occurs. Establish/Maintain RCS temperature to ensure transition out of E-3 does not occur due either of the RCS temperature too high to maintain required subcooling RCS temperature too low resulting in severe challenge to the subcriticality or integrity CSF . (E-3-B) When SI termination criteria are met, stop SI pumps before completion of ECA-3.3 step Event Malf. Event No. No. Type* CNH-C(ATC) 23 MFRV fails closed in auto with manual available ramped over 10 PCS008D minutes. C(CRS) XMT-I(CRS) VCT level instrument fails low causing automatic makeup and charging CVC019 pump suction to swap to the RWST. I(BOP)A MAL-R(ATC) SGTL on 23 SG 900 gpd. This will require a downpower and eventual RCS014 shutdown.

N(CRS)C N(BOP) I 4 .,ALL) Loss of Station Auxiliary Transformer will result in loss of 6.9 KV Buses EPSOOI 5 and 6 causing 480 V Buses 5A and 6A to be powered from their EDGs 1 MAL-M(ALL) SGTR on 23 SG grows to 280 gpm. This will lead'to team performing a RCSOl4 manual reactor trip and Sf. C MAL-C(CRS) Fault on 480 V Bus 6A during the SI loading sequence.

This will result in EPS007D having no available PORVs. MOC-C(CRS) 21 AFW pump will not auto start (inserted at setup). This along with loss AFWOOI of 6A will result in inadequate heat sink until addressed by team. C(BOP} (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor U2 NRC 2010 Scenario I: FRV failure, VCT level transmitter failure, SGTL, LOOP, SGTR w/o pressure control. Page 1 of2 Session Outline: The evaluation begins with the plant at 100% power steady state operation.

Shortly after the team takes the watch, 23 MFRV will slowly fail closed. The team should recognize the failure and the ATC should transfer control of the valve to manual per administrative guidance ofEN-OP-II5, Conduct of Operations.

The team will enter 1, Loss of Main Feedwater.

While the team is progressing through 2-AOP-FW-I (or after exit), VCT level instrument 112 will fail low. This will cause an automatic makeup and charging pump suction to swap to the RWST. The team will respond per 2-AOP-CVCS-I, CVCS Malfunctions.

After the team has stabilized charging pump suction, a 900 gpd steam generator tube leak will develop on 23 SG. The team will implement 2-AOP-SG-I, Steam Generator Tube Leak, and begin a shutdown.

After the team starts the load reduction, a fault will occur on the Station Auxiliary Transformer (SAT). 6.9 KV Busses 5 and 6 will de-energize and 21 and 23 EDGs will auto start to re-power 480V Busses 5A and 6A. The team will implement 2-AOP-138KV-l, Loss of Power to 6.9 KV Bus 5 and/or 6. While progressing with the shutdown and addressing the loss of SAT, the tube leakage in 23 SG will increase to 280 gpm. The team will diagnose the increase in leak rate and trip the reactor and actuate SI. Following the reactor trip and SI, the team will have to establish AFW because 23 does not have power and 21 will not auto start (malfunction).

22 AFW pump must be placed in service. to feed 23 and 24 SGs. 21 AFW may be manually started to feed 21 and 22 SGs or 22 AFW pump may be used to feed all four SGs. The team will progress through E-O, Reactor Trip or Safety Injection and transition to E-3, Steam Generator Tube Rupture. 23 SG will be isolated and the team will cool down the RCS in preparation to depressurize.

The team will be unable to depressurize the RCS using E-3. Normal spray cannot be used because no RCPs are in service. Auxiliary spray will not be available because instrument air to containment will not be available (PCV-1228 will not open). Neither PORV will be available; one is tagged out, and the other's closed block valve does not have power. The team will transition to 2-ECA-3.3, SGTR without Pressurizer Pressure Control. The scenario will be terminated when SI pumps have been stopped after RCS depressurization in ECA-3.3. Procedural flow path: 2-AOP-FW-I, 2-AOP-CVCS-I, 2-AOP-SG-I, (2-POP-2.l, I, or 2-AOP-RLR-I), 2-AOP-138KV-l, E-O, E-3, ECA-3.3 U2 NRC 2010 Scenario 1: FRY failure, YCT level transmitter failure, SGTL, LOOP, SGTR w/o pressure control. Page 2 of2 Facility:

Indian Point 2 Scenario No.: Op-Test No.:_1 Examiners:

___________

_ Operators:

Initial Conditions: Reset simulator to IC-287 Load Simulator Schedule-Scenari04 is in Mode 1 just above 5% power preparing to come ..

.0Raise power to approximately 8-10% to place MTG in service. \'

() Critical Insert negative reactivity into the core by at least one of the following methods before completing FR-S.1 step 4: (FR-S.1-C)

• De-energize the control rod drive MG sets

• Place rod control in manual and insert RCCAs

• Establish emergency boration flow to the RCS Establish at least 800 gpm AFW flow to the SGs before completion of FR-S.1 step 3. (E-O --F) Manually actuate at least one train of SIS actuated safeguards before completion of E-O step 4. (E-O --D) Event Malf. Event Event No. No. Type* Description N/A R(ATC) Power escalation.

N(CRS) N(BOP) 2 C(CRS) 22 Service Water Pump trip. SWS007 C(BOP) Tech Spec for CRS I(ALL) Controlling PZR Pressure transmitter fails Tech Spec for CRS A C(CRS) FCV-625 spurious C(BOP) Z, Steam leak in the Turbine Building leading to plant trip. C(ATC) Entered at setup, Reactor Trip Breakers will not open causing PPL003/4 to enter C(CRS) Entered at setup, SI does not automatically actuate. PPL487/8 actuation will be C(BOP) ! * {N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor U2 NRC 2010 Scenario 4: Power escalation from 5%, SWP Trip, PT-455 Failure, FCV-625 Closure, Steam Line Rupture, A TWS, Faulted SG. Page 1 of 13 Session Outline: The scenario begins with the plant at 5% power with no equipment is out of service. The team has been instructed to raise power to 8-10% and place MTG in service.

taking the watch, the crew will commence raising ri:e power escalation has 22 SWP will trip. The team will start another pump per 2-ARP-SJF.

C ollowing the restoration ofSW, a failure high ofPT-455 will occur. The team will respond using 2-AOP-INST-l "Instrument or Controller Failures." The channel will be removed from service. After the channel is removed from service, FCV -625 will go closed with no apparent reason. The team should respond per 2-ARP-SGF and re-open the valve. If the team elects to not re-open the valve, the scenario can continue.

Prior to completion of the Subsequent Actions of2-AOP-CCW-l, a steam break will occur in the Turbine Building.

The team will attempt to manually trip the plant but the reactor trip breakers will not open. The reactor will not trip from the Control Room and the team will respond per 2-FR-S.l, "Response to Nuclear Power Generation

/ ATWS," and will shutdown the reactor by manually inserting control rods and initiating Emergency Boration.

The reactor trip breakers will not be locally opened after an NPO is dispatched, until after emergency Boration has been aligned. One MSIV will fail to close from the control switches.

The team will proceed through 2-FR-S.l until transition to 2-E-O, "Reactor Trip or Safety Injection." After the transition to 2-E-0 is made, the team will determine that three SGs are intact and 23 SG is faulted. The Team will also determine that SI did not automatically actuate and must manually actuate SI. The team will transition to 2-E-2, Faulted Steam Generator Isolation and isolate 23 SG. The scenario is terminated after the actions of2-E-2 are complete and a transition to 2-E-l, Loss of Reactor or Secondary Coolant is announced or at the discretion of the lead exammer. Procedure flow path: 2-POP-1.3, 2-ARP-SJF, 2-AOP-INST-l, 2-ARP-SGF, 2-AOP-UC-l, 0, 2-FR-S.l, 2-E-0, 2-E-2 U2 NRC 2010 Scenario 4: Power escalation from 5%, SWP Trip, PT-455 Failure, FCV-625 Closure, Steam Line Rupture, ATWS, Faulted sa. Page 2 of 13 Facility:

Indian Point 2 Scenario No.: * (spare) .... Op-Test No.:_1 Examiners:

____________

_ Operators:

Initial Reset simulator to IC-118 Load Simulator The Plant is in a 100% normal full power 22 AFW Pump has been OOS for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> due to bearing oil line Maintain 100% Critical Establish RCS bleed and feed when the average of the three lowest S/G levels reach 41 % WR. (FR-H.1-F) Open all reactor vessel head vent valves before resetting SI in FR-H.1 (non-WOG) Close all reactor vessel head vent valves before exitin 2-FR-H.1.

FR-H.1-G Event Malf. Event No. No. T e'" Descri XMT-I(ALL) 21 SG B Channel of Steam Flow Fails high. Team will place Channel .J. in SGNOO2 service and enter 2-AOP-INST

-1. r A Tech ec Evaluation for CRS MAL-C(ALL) Loss of 480V Bus 3A. Team will enter 2-AOP-480V

-\ and diagnose EPS007B a T.S. shutdown is required due to having 2 inoperable AFW N(CRS) .Tech Spec Evaluation for CRS MAL-2 AFW pump will not start. T is will lead the team to a loss of heat sink. C(ALL) PORV PCV-455C will not open. This will require the team to open the RCS002 reactor head vent valves to perform bleed and feed. A Instrument, U2 NRC 2010 Loss of Secondary Heat Sink, Bleed and Feed required, following SG Scenario 5 Flow Failure, Loss of 480V Bus, Failure ofRCP #1 Seal, Turbine trip Page 1 of 15 Session Outline: The evaluation begins with the plant at 100% power steady state operation.

The following equipment is out of service: 22 AFW pump has been out-of-service for bearing oil line repair for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. It is expected back within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> (ITS 3.7.5 -72 hr AOT). 21 and 23 AFW pumps are protected equipment.

After taking the watch, 21 SO Steam Flow Channel B transmitter will fail high. The ATC will switch to the A channel and the team will enter 2-AOP-INST-1.

The CRS will refer to Tech Spec Table 3.3.2-1 and bistables will be tripped. After bistables are tripped, a fault will occur on 480V Bus 3A. The team will take actions in accordance with AOP-480V-l, "Loss of Normal Power to any Safeguards 480V Bus." Due to the fault on Bus 3A, 22 EDO cannot re-energize the bus. TS require plant shutdown due to 2 trains of AFW inoperable (TS 3.7.5 condition C). After team has begun shutdown, 21 RCP will experience

1. 1 seal degradation.

The team will perform actions of AOP-RCP-l, Reactor Coolant Pump Malfunctions." The #1 seal degradation severity will then increase requiring reactor trip. When the reactor is tripped, the turbine upper left stop and control valve pair fail to close. £-.15J7 MSIV's must be manually closed to trip the tiirbme. {/Y'\-""D...... l 23 ABFP will not auto start and will not be able to be manually started from the Control Room due to 480V circuit breaker failure. (21 ABFP is de-energized due to fault on bus 3A, and 22 ABFP is out of service.)

The team will subsequently transition to FR-H.l, "Loss of Secondary Heat Sink" due to a loss of AFW flow. SO WR levels will lower until bleed and feed is required.

One PRZR PORV will not open when required.

The crew will open the Reactor Head Vent valves. 21 AFW pump will then be successfully started from its ASSS supply, or 23 AFW pump from its normal supply after swapping 480V breakers with the spare breaker. The scenario can be terminated after the head vent valves have been closed, or at the discretion of the lead evaluator.

Procedure flow path: 2-AOP-fNST-l, 2-AOP-480V-l, 2-POP-2.1 or 2-AOP-RSD-l, RCP-I, 2-E-O, 2-FR-H.l U2 NRC 2010 Loss of Secondary Heat Sink, Bleed and Feed required, following SG Scenario 5 Flow Failure, Loss of 480V Bus, Failure of RCP #1 Seal, Turbine trip Page 2 of 15