ML043170090

Draft - Outlines (Folder 2)

ML043170090
Person / Time
Site:	Indian Point
Issue date:	10/22/2004
From:	Christman R Entergy Nuclear Indian Point 2, Entergy Nuclear Northeast
To:	Conte R NRC/RGN-I/DRS/OSB
Conte R
References
Download: ML043170090 (29)

Text

ES-401 PWR Examination Outline Form ES-401-2

-acility:

Indian Point 2 Date of Exam:

10/22/2004 Exam Level:

RO iote:

1.

2.

3.

4.

5.

6.*

7.

8.

9.

Ensure that at least two topics from every WA category are sampled within each tier of the RO outline (i.e., the Tier Totals in each WA category shall not be less than two). Refer to Section D. 1.c for additional guidance regarding SRO sampling.

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 f 1 from that specified in the table based on NRC revisions. The final exam must total 75 points and the SRO-only exam must total 25 points.

Select topics from many systems; avoid selecting more than two WA topics from a given system unless they relate to plant-specific priorities.

Systems/evolutions within each group are identified on the associated outline.

The shaded areas are not applicable to the category/tier.

The generic (G) WAS in tiers 1 and 2 shall be selected from Section 2 of the WA Catalog, but the topics must be relevant to the applicable evolution or system.

The SRO WAS must also be linked to 10 CFR 55.43 or an SRO-level learning objective.

On the following pages, enter the WA numbers, a brief description of each topic, the topics importance ratings (IR) for the applicable license level, and the point totals for each system and category. Enter the group and tier totals for each category in the table above; summarize all the SRO-only knowledge and non-A2 ability categories in the columns labeled K and A. Use duplicate pages for RO and SRO-only exams.

For Tier 3, enter the WA numbers, descriptions, importance ratings, and point totals on form ES-401-3.

Refer to ES-401, Attachment 2, for guidance regarding the elimination of inappropriate WA statements.

NUREG-1021, Draft Revision 9 Page 1 of 12

Emergency and Abnormal Plant Evolutions - Tier I/Group 1 (RO / SRO) 11 E/APE#/Name/SafetyFunctlon I K1 I K2 I K3 I A I I A2 I G I Number I KIA Topic(@

I IR I

Q#

II 000007 (BWlEO2 & EIO; CEIEO2) I Reactor Trip -

Stabilization - Recovery I 1

000008 I Pressurizer Vapor Space Accident I 3

R I

I I

I 000009 / Small Break LOCA I3 I

I I

I 00001 I I Large Break LOCA 1 3 R

I I

I 00001 5/17 RCP Malfunctions 14 I

R 000022 I Loss of Reactor Coolant Makeup I 2

000025 I Loss of RHR System I 4 R

R 000026 I Loss of Component Cooling Water I 8 000029 I Anticipated Transient w/o Scram I 1

I I

000038 I Steam Generator Tube Rupture 13 R

000040 (BW/EO5; CEIEO5; WIE12) I Steam Line Rupture - Excessive Heat Transfer 1 4 000054 (CEIEO6) I Loss of Main Feedwater I 4 000055 I Station Blackout I 6

000056 I Loss of Off-site Power / 6 000057 I Loss of Vial AC Elec. Inst. Bus I 6 IIH 000058 I Loss of DC Power I 6 000062 I Loss of Nuclear Service Water I 4 1

I I

I Not Selected AK3.03 EA2.34 EK2.02 AK3.02 Knowledge of the reasons for actions contained in EOP for PZR vapor space accidentlLOCA Ability to determine or interpret conditions for throttling or stopping HPI as they apply to a small break LOCA Knowledge of the interrelatlons between pumps and a Large Break LOCA Knowledge of the reasons for responses of CCW lineup and flow paths to RCP oil coolers during RCP malfunctions 4.1 3.6 2.6 3.0 AKI.02 Knowledge of the operational implications of the relationship of charging flow to pressure differential between charging and RCS as they apply to Loss of Reactor Coolant Pump during all modes of operations 2.7 AKI -01 Knowledge of the operational implications of a loss of RHRS 3.9 MI.07 Abirity to operate andlor monitor flow rates to the components and systems that are serviced by the CCWS; interactions among the components pressure instrument fails high EK2.06 Knowledge of the interrelations between the breakers, relays, 2.9 and disconnects following an ATWS EA1.I 1

Ability to operate and monitor SG level indicators as they apply 3.8 to a SGTR Not selected 2.9 R

A42.15 Ability to determine and interpret the actions to be taken if PZR 3.7 R

G2.4.2 Knowledge of system set points, interlocks and automatic 3.9 actions associated with EOP entw conditions EK3.02 Knowledge of the reasons for the actions contained in EOP for 4.3 loss of offsite and onsite power Knowledge of the reasons for the following responses as they apply to the Loss of Vital AC Instrument Bus Actions contained in EOP for loss of vital AC electrical instrument bus R

G2.1.20 Ability to execute procedure steps 4.3 AK3.01 4.1 Ability to determine and interpret DC loads lost; impact on ability to operate and monitor plant systems as they apply to the loss 3.5 l

l I

I I of DC Power NUREG-1021, Draft Revision 9 Page 2 of 12

Emergency and Abnormal Plant Evolutions - Tier I/Group I (RO / SRO)

WIE04 I LOCA Outside Containment I3 Containment and the components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes and automatic and manual features 1

I K/A Category Point Totals:

1 3 1 3 1 4 1 2 1 3 1 3 1 I Group Point Total:

I 18 NUREG-1021, Draft Revision 9 Page 3 of 12

Emergency and Abnormal Plant Evolutions - Tier I/Group 2 (RO / SRO)

I l

l I Not Selected I

I 300001 I Continuous Rod Withdrawal I I 360003 I Dropped Control Rod I 1

I l

l i Not Selected Not Selected Not Selected Control Malfunctions and sensors and detectors AK2.02 Knowledge of the interrelations between the Pressurizer Level 2.6 30005 InoperableIStuck Control Rod I I DO0024 Emergency Boration I 1

DO0028 I Pressurizer Level Malfunction I 2 DO0032 I Loss of Source Range NI I 7 Knowledge of the Interrelations between the Loss of Source Range Nuclear Instrumentation and the power supplies, including proper switch positions Not Selected 2.7 000033 I Loss of Intermediate Ranae NI I 7 000036 (BWIAO8) I Fuel Handling Accident I 8

I l

l I Not Selected I

7 000037 I Steam Generator Tube Leak I 3 Knowledge of the operational implications of the leak rate vs.

3.5 pressure drop concept as it applies to a Steam Generator Tube teak Not Selected 000051 I Loss of Condenser Vacuum I 4

~ _ _ _ _ _

000059 I Accidental Liquid Radwaste Rel. I 9

AK3.01 Knowledge of the reasons for the termination of a release of 3.5 radioactive liquid as it applies to the Accidental Liquid Radwaste Release Not Selected 000060 I Accidental Gaseous Radwaste Rei. I 9

000061 I ARM System Alarms I 7

I l

l k

t Selected I

000067 I Plant Fire On-site / 9 Not Selected Ability to operate and I or monitor the SIG levels as they apply to the Control Room Evacuation Not Selected

4. I 000068 (BWIA06) I Control Room Evac. I 8

000069 (WIEI4) I Loss of CTMT lntearitv I 5 AA1.03 0000?4 WE06 & E07) I Inad. Core Cooling I 4 I

I R I G2.4.18 I Knowledge of the specific bases for EOPs R

000076 I High Reactor Coolant Activity I 9

WE/OI & 02 I Rediagnosis & SI Termination I 3

WIE13 I Steam Generator Over-pressure I 4

W/E25 I Containment Flooding I 5

Not Selected Not Selected Not Selected Knowledge of the operational implications of the normal, abnormal and emergency operating procedures associated with Containment flooding Ability to determine and interpret adherence to appropriate procedures and operation within the limitations in the facility's license and amendments as they apply to High containment Radiation 2.7 3.0 WE1 5 EKI.2 WIE16 EA2.2 WIEI6 I High Containment Radiation I 9 Not Selected Not Selected I

I BWIAOI I Plant Runback I 1

BWIAO2 & A03 I LOSS Of NNI-WY I 7

NUREG-1021, Draft Revision 9 Page 4 of 12

I ES-401 BW/AO4 /Turbine Trip 14 PWR Examination Outline Emergency and Abnormal Plant Evolutions - Tier l/Group 2 (RO / SRO)

I I

I I

I Not Selected 1

UAPE # I Name I Safety Functlon I

BW/AO5 / EmerQency Diesel Actuation I 6 1 K1 I K2 1 K3 1 A I 1 A2 1 G 1 Number 1 I

I I Not Selected 1

KIA topic(8)

BW/A07 / Flooding I 8 BWE03 I Inadequate Subcooling Margin I4 BW/E08; WE03 / LOCA Cooldown I Depress. I 4 Not Selected Not Selected operating procedures associated with (LOCA Cooldown and Depressurization).

R W/E03.

Knowledge of the reason for normal, abnormal and emergency 3.4 EK3.2 BW/EO9; CEIA13; WE09 & 10 Natural Circ.14 BW/E13 8 E14 / EOP Rules and Enclosures CE/AI 1; W/E08 / RCS Overcooling - PTS / 4 CEIA16 I Excess RCS Leakane I 2 Not Selected Not Selected Not Selected Not Selected CUE09 I Functional Recovery K/A Cateaorv Point Totals:

NUREG-1021, Draft Revision 9 Not Selected 2

2 2

1 1

1 GrOUb Point Total:

9 Page 5 of 12

Plant Systems - Tier 2/

003 Reactor Coolant Pump 003 Reactor Coolant Pump 004 Chemical and Volume Control 004 Chemical and Volume Control 005 Residual Heat Removal R

A 006 Emergency Core Cooling 006 Emergency Core Cooling 008 Component Cooling Water I 010 Pressurizer Pressure Control 012 Reactor Protection I

I 022 Containment Cooling NUREG-1021, Draft Revision 9 Page 6 of 12 K5.05 A1.10

-r-K3.02 K6.18 A I.01 A3.02 K6.01 K1.02 A4.04 Knowledge of the operational implications of the dependency of RCS flow rates upon the number of oaeratina RCPs Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RCPs controls including RCP standpipe levels 2.5 Knowledge of the effect that a loss or malfunction of the CVCS will have on RCP

---++

Ability Seal Injection to manually operate and/or monitor in the control room Boron concentration Knowledge of the bus power supplies to the RCS pressure boundary motorsperated valves Knowledge of the effect that a loss or malfunction of the ECCS will have on the fuel Knowledge of the effect that a loss or malfunction of the ECCS will have on Subcooling Margin Indicators --

Ability parameters to predict (to prevent and/or exceeding monitor changes design in limits) associated with maintaining quench tank water level within limits Knowledge of the physical connections and /

or cause-effect relationships between the CCWS PRMS Ability to monitor automatic operation of PZR PCS. includina: PZR aressure.

I 31 Knowledge of the effect that a loss or malfunction of the pressure detection systems will have on the PZR PCS Knowledge of the physical connections and /

or cause-effect relationships between the RPS and the 125VDC Svstem Knowledge of bus power supplies to the ESFAS/safeguards equipment Ability to manually operate and/or monitor in the Control Room: Valves in the CCS.

Plant Systems - Tier 2/Group 1 (RO / SRO) 025 Ice Condenser 026 Containment Spray 039 Main and Reheat Steam 056 Condensate 059 Main Feedwater R

061 Auxiliary / Emergency Feedwater I

062 AC Electrical Distribution R

063 DC Electrical Distribution R

064 Emergency Diesel Generator 064 Emergency Diesel Generator

/

I

/

073 Process Radiation Monitoring R

076 Service Water NUREG-1021, Draft Revision 9 f

K4.06 NIA Knowledge of the CSS design feature(@

and/or interlock(s) which provide for lodlne scavenging via the CSS

~

~~

Ability to (a) predicthe impacts of Increasing steam demand, its relationship to increases in reactor power operation on the MRSS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or oDerations R

A2.04 K3.04 K5.01 K2.01 K3.02 G2.1.28 Ability to (a) predict the impacts of loss of condensate pumps, Condensate System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations Knowledge of the effect that a loss or malfunction of the MFW will have of the RCS Knowledge of the operational implications of the relationship between AFW flow and RCS heat transfer Knowledge of bus power supplies to the major system loads Knowledge of the effect that a loss or malfunction of the DC Electrical System will have on the following: Components using dc control power.

Knowledge of the purpose and function of the major system components and controls t

A4.01 K1.01 Ability to manually operate andlor monitor in the control room local and remote operation of the ED/G Knowledge of the physical connections and/or cause-effect relationships between the PRM system and those systems served bv PRMs R

A4.01 Ability to manually operate and/or monitor in the control room SWS Pumps Page 7 of 12

Plant Systems - Tier 2/Group 1 (RO / SRO) 076 Service Water 078 Instrument Air 103 Containment +

I K/A Category Point Totals:

1 4 1 3 1 5 1 1 1 2 NUREG-1021, Draft Revision 9 K1.05 Knowledge of the physical connections 3.8 andlor cause-effect relationships between the SWS and the D/G containment systems including containment I

I I

I I

I I

I 2

2 2

2 4

1 Group PointTotal:

28 Page 8 of 12

Not selected Knowledge of design features@) andlor interlock(s) which provide for anti-siphon devices 2.6 035 Steam Generator R

K5.01 068 Liquid Radwaste 071 Waste Gas Disposal Not selected Not selected Plant Systems - Tier 2/Group 2 (RO / SRO) 001 Control Rod Drive I

I I I

I R I I K4.07 Knowledge of the CRDS design feature(@

andlor interlock@) which provide for the rod stops Not selected Knowledge of the operational implications of the PZR level indication when RCS is saturated K5.15 AI.03 Ability to predict andlor monitor changes in parameters associated with operating the RPlS controls, including PDIL, PPDIL Not selected Not selected A2.02 Ability to (a) predict the impacts of core damage on the ITM system; and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of core damage Not selected

-~

~

1 028 Hydrogen Recombiner and Purge Control I I

I I

I I

Not selected K4.03 R

~~

1 034 Fuel Handling Equipment I

I I

I I

I I

1 1

1 1

1 I Not selected I

1 I 3.4 I I

Knowledge of operational implications of the effect of secondary parameters, pressure, and tm G2.1.10 Knowledge of conditions and limitations in the facility license.

2.7 041 Steam Dumpnurbine Bypass Control A I.06 Ability to predict andlor monitor changes in parameters (to prevent exceeding design limits) associated with operating the MT/G system controls including expected response of secondary plant parameters following T/G trip 3.3 I

1 I

I I

I I

l l

I I

I I Not selected I

I 11 055 Condenser Air Removal

Plant Systems - Tier 2/Group 2 (RO / SRO)

~-

072 Area Radiation Monitoring 075 Circulating Water 079 Station Air 086 Fire Protection

~-

Not selected the control room Not selected Fire Protection System including actuation of the FPS i

I R

2.1.8 Ability to coordinate personnel activities outside 3.8 I

R A3.02 Ability to monitor automatic operation of the 2.9 I

KIA Category Point Totals:

2 2

2 1

1 2

Group Point Total:

10 i

NUREG-1021, Draft Revision 9 Page 10 of 12

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

Form ES-401-3 Facility:

Indian Point Unit 2 Date of Exam:

10/22/2004 RO Category WA #

Topic IR Q#

2.1.29 Knowledge of how to conduct and vet-@ valve 3.4 Conduct of 2.1 -7 Ability to evaluate plant performance and make 3.7 lineups.

operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

Operations 2.1.20 Ability to execute procedure steps.

4.3 Total 3

Equipment 2.2.33 Knowledge of control rod programming.

2.5 Control 2.2.22 Knowledge of limiting conditions for operations 3.0 and safety limits.

Total 3

Radiation Control 2.3.1 I Ability to control radiation releases.

2.7 2.3.1 Knowledge of 10 CFR: 20 and related facility 2.6 radiation control reauirements.

I Total I

3 2.4.16 Knowledge of EOP implementation hierarchy and 3.0 coordination with other s u ~ ~ o r t Drocedures Emergency Procedures /

Plan 2.4.6 Knowledge symptom based EOP mitigation strategies 3.8 Knowledge of EOP entry conditions and immediate action stem.

I 4.3 I I Total 1

3 Tier 3 Point Total RO NUREG-1021, Draft Revision 9 Page 11 of 12

~

ES-401 Record of Rejected WAS Form ES-401-4 Tier I Group Randomly Selected KIA Reason for Rejection 211 010 K3 Too many K3s, Not enough K6's 211 103 K3 Too many K3s, Not enough A3's NUREG-I 021, Draft Revision 9 Page 12 of 12

ES-401 PWR Examination Outline Form ES-401-2

=acility:

Indian Point 2 Date of Exam:

10/22/2004 Exam Level:

SRO Abnormal Plant dote:

1.

2.

3.

4.

5.

6.*

7.

8.

9.

Ensure that at least two topics from every WA category are sampled within each tier of the RO outline (Le., the "Tier Totals" in each K/A category shall not be less than two). Refer to Section D. 1.c for additional guidance regarding SRO sampling.

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 f 1 from that specified in the table based on NRC revisions. The final exam must total 75 points and the SRO-only exam must total 25 points.

Select topics from many systems; avoid selecting more than two WA topics from a given system unless they relate to ptant-specific priorities.

Systems/evolutions within each group are identified on the associated outline.

The shaded areas are not applicable to the categoryher.

The generic (G)

WAS in tiers 1 and 2 shall be selected from Section 2 of the WA Catalog, but the topics must be relevant to the applicable evolution or system.

The SRO WAS must also be linked to 10 CFR 55.43 or an SRO-level learning objective.

On the following pages, enter the WA numbers, a brief description of each topic, the topics' importance ratings (IR) for the applicable license level, and the point totals for each system and category. Enter the group and tier totals for each category in the table above; summarize all the SRO-only knowledge and non-A2 ability categories in the columns labeled "K" and "A." Use duplicate pages for RO and SRO-only exams.

For Tier 3, enter the KIA numbers, descriptions, importance ratings, and point totals on form ES-401-3.

Refer to ES-401, Attachment 2, for guidance regarding the elimination of inappropriate WA statements.

I NUREG-1021, Draft Revision 9 Page 1 of 9

Emergency and Abnormal Plant Evolutions - Tier I/Group 1 (SRO) 000007 (BWIE02 & E10; CWEO2) I Reactor Trip -

Stabilization - Recovery I I 000008 I Pressurizer Vapor Space Accident I 3

000009 I Small Break LOCA I 3

S EK2.03 00001 1 I Large Break LOCA I 3

00001 5/17 RCP Malfunctions I 4 S

AA2.02 000022 I Loss of Reactor Coolant Makeup I 2

000025 I Loss of RHR System I 4 000026 I Loss of Component Cooling Water I 8 000027 I Pressurizer Pressure Control System Malfunction I 3 000029 I Anticipated Transient w/o Scram I 1

000038 I Steam Generator Tube Rupture I 3 S

EK2.02 000040 (BWIE05; CEIE05; WE12) I Steam Line S

G2.4.6 Rupture -Excessive Heat Transfer I 4 000054 (CEIEO6) I Loss of Main Feedwater I 4 AKI.01 S

000055 I Station Blackout I 6

000056 I Loss of Off-site Power I 6 S

AK3.02 000057 I Loss of Vial AC Elec. Inst. Bus I 6 000058 I Loss of DC Power I 6 000062 I Loss of Nuclear Service Water I 4 I

l l

l I

I I

000065 I Loss of Instrument Air I 8

WIEO4 I LOCA Outside Containment I 3 WIEI 1 I Loss of Emergency Coolant Recirc. I 4

S E l 1 I I I I I I lEA2.2 Knowledne of the interrelations between the small break LOCA 3.3 and the gteam Generators I

I I

Ability to determine and interpret abnormalities in RCP air vent flow paths andlor oil cooling system as they apply to the reactor Coolant Pump Malfunction 3.0 I

Knowledge of the interrelations between sensors and detectors and a SGTR Knowledge of symptom based EOP mitigation strategies 2.5 4.0 Knowledge of the operational implications of a MFW line break 4.3 depressurizes the SIG concepts as they apply to Loss of Main Feedwater I

NUREG-1021, Draft Revision 9 Page 2 of 9

Emergency and Abnormal Plant Evolutions - Tier l/Group 2 (SRO)

WE/OI & 02 I Rediagnosis & SI Termination I3 NUREG-1021, Draft Revision 9 Page 3 of 9

Emergency and Abnormal Plant Evolutions - Tier l/Group 2 (SRO)

BW/EO9; CE/AI 3; W/E09 & I O Natural Circ./ 4 S

EA2.2 Ability to detenine and interpret adherence to appropriate 3.8 procedures and operation within the limitations in the facility's iicense and amendments as they apply to Natural Circulation Operations BWIE13 & E14 I EOP Rules and Enclosures CE/AI 1; W/E08 / RCS Overcooling - PTS I 4 CUA16 I Excess RCS Leakage I 2 NUREG-1021, Draft Revision 9 Page 4 of 9

Plant Systems - Tier 2/Group 1 (SRO) 013 Engineered Safety Features Actuation 022 Containment Cooling 025 Ice Condenser 026 Containment Spray 004 Chemical and Volume Control I

NIA 005 Residual Heat Removal I

006 Emergency Core Cooling 007 Pressurizer Relief/Quench Tank 008 Component Cooling Water pump/motor malfunctions or operations on the RHRS and based on those predictions use procedures to correct, control or mitigate the consequences of those malfunctions or 1 010 Pressurizer Pressure Control I

I 1 056 Condensate 1

1 1

1 1

1 1

1 1

1 1

1 I

I I

059 Main Feedwater 061 Auxiliary I Emergency Feedwater 062 AC Electrical Distribution S

K4.01 063 DC Electrical Distribution 064 Emergency Diesel Generator 073 Process Radiation Monitoring features andlor interlocks which provide for Bus lockouts NUREG-1021, Draft Revision 9 Page 5 of 9

Plant Systems - Tier 2/Group I (SRO) 076 Service Water 078 Instrument Air 103 Containment WA Category Point Totals:

1 1

2 Group Point Total:

4 V

NUREG-1021, Draft Revision 9 Page 6 of 9

Plant Systems - Tier 2/Group 2 (SRO) 001 Control Rod Drive 002 Reactor Coolant S

K5.07 Knowledge of the operatlonal implications of 3.9 reactivity effects of RCS boron, pressure and temperature as they apply to the RCS 01 1 Pressurizer Level Control Ij 014 Rod Position Indication I

I I

I I

I I I - l I I I m

a r

Instrumentation 01 6 Non-nuclear Instrumentation 01 7 In-Core Temperature Monitor 027 Containment Iodine Removal 028 Hydrogen Recombiner and Purge Control 029 Containment Purge 033 Spent Fuel Pool Cooling 034 Fuel Handling Equipment S

142.01 Ability to predict the impacts of a dropped fuel 4.4 element on the Fuel Handling System and based on those predictions, use procedures to correct, control or mitigate the consequences of those malfunctions or operations 11 035 Steam Generator I

I I

I I

I I

I I

I I

I I

041 Steam Dumpnurbine Bypass Control 045 Main Turbine Generator 055 Condenser Air Removal b&

quid Radwaste I

I I

I I

I I

I I

1 1

1 071 Waste Gas Disposal 072 Area Radiation Monitoring 075 Circulating Water 079 Station Air 086 Fire Protection WA Category Point Totals:

1 1

Group Point Total:

2 b

NUREG-1021, Draft Revision 9 Page 7 of 9

ES-401 Generic Knowledge and Abilities Outline (Tier 3) (SRO) Form ES-401-3 Facility: Indian Point Unit 2 Conduct of I

II Date of Exam:

10/22/2004 SRO Topic IR Q#

Ability to recognize indications for system operating parameters which are entry-level 4.0 conditions for technical specifications Ability to supervise and assume a management role during r>lant transients and umet conditions 4.3 I

2 Equipment Control Knowledge of bases in technical specifications Knowledge of refueling administrative 3.7 3.7 for limiting conditions for operations and safety limits requirements I Total I

2 Radiation 2.3.4 Knowledge of radiation exposure limits and 3.1 Control contamination control, including permissible levels in excess of those authorized Total I

Emergency 2.4.27 Knowledge of fire in the plant procedures 3.5 Procedures /

2.4.6 Knowledge of symptom based EOP mitigation 4.0 Plan strategies I

I Total 2

Tier 3 Point Total SRO 7

NUREG-1021, Draft Revision 9 Page 8 of 9

ES-401 Record of Rejected WAS (SRO)

Form ES-401-4 NUREG-1021, Draft Revision 9 Page 9 of 9

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Indian Point 2 Date of Examination: October 11, 2004 Exam Level (circle on O(I) / SRO(U)

Operating Test No.: 1 Administrative Topic (see Note)

Conduct of Operations Conduct of Operations Equipment Control Radiation Control Emergency Plan Describe activity to be performed:

Interpretation and application of overtime guidelines Application of Technical Specifications, determine that tripping bistables will cause a reactor trip.

(PERFORM AFTER JPM Sim-C)

Conduct an emergency tagout removal Calculate and Record a Liquid Radioactive Release for #14 Liquid Waste Distillate Storage Tank Not applicable NOTE: All items (5 total) are required for SROs. RO applicants required only 4 items unless they are retaking only the administrative topics, when 5 are required.

NUREG-1021, Draft Revision 9

ES-301 Administrative ToDics Outline Form ES-301-1 Date of Examination: October 11, 2004 Exam Level (circle one): RO SRO(U)

Operating Test No.: 1 Facility: Indian Point Unit Administrative Topic (see Note)

Conduct of Operations Conduct of Operations Equipment Control Radiation Control Emergency Plan Describe activity to be performed:

Interpretation and Application of overtime guidelines/Replace watchstander due to illness Apply Technical Specifications, Monitor RPS, Place a RPS Channel in the tripped condition (one channel already failed, placing second channel will cause reactor trip.) OTDT logic - Pressure channel failure and concurrent Temperature failure.

(PERFORM AFTER JPM S.C)

Conduct an emergency tagout removal Review and Approve a Liquid Radioactive Release Emergency Plan Classification (following scenario) (Time critical, 15 mins)

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

NUREG-1021, Draft Revision 9

ES-301 Control Room/ln-Plant Systems Outline Form ES-301-2 Facility: Indian Point U Exam Level (circle on&RO(I)

/ SRO(U)

Control Room Systems (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U)

Date of Examination: October 11,2004 Operating Test No.: 1

- - P System / JPM Title Type Safety Code*

Function

a. Perform the required actions for a malfunction of rod position N, s 1

b. Align SI pump and header during LOCA with RCS N, A, S 2

indicator temperature <350F

c. Perform the required action for PZR PRESSURE CHANNEL D, s 3

FAILURE (Control pressure manually)

d. Verify RCP operation as per EOP E-0, Step 9 N, A, s 4 pri

e. Start 21 and 23 ABFP from the control room and supply A W flow to the SGs during plant shutdown N, A, S, L 4 sec

f.

Manually initiate containment spray when actuation is N, A, s 5

required

g.

Energize 6.9 kv from 13.8 kv backup power N, S 6

h. Remove an Intermediate Range Channel from service N, S 7

ln-Plant Systems (3 for RO; 3 for SRO-I; 3 or 3 for SRO-U)

i.

Manually start 21 Emergency Diesel Generator l

D 1

6

~~~~

j.

Lineup alternate cooling to the SIS and RHR Pumps

k. Align 24 Large Gas Decay Tank for start of discharge 1

M,R I

9

• Type Codes: (D)irect from bank, (M)odified from bank, (N)ew, (A)lternate path, (C)ontrol room, (S)imulator, (L)ow-Power, (R)CA NUREG-1021. Draft Revision 9

ES-301 Control Room/ln-Plant Systems Outline Form ES-301-2 Control Room Systems (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U)

I N,S I

a.

Perform the required actions for a malfunction of rod position I

indicator 1

b. Align SI pump and header during LOCA with RCS temperature c350F 2

c.

Perform the required action for PZR PRESSURE CHANNEL D, s 3

FAILURE (Control pressure manually)

d. Verify RCP operation as per EOP E-0, Step 9 N, A, S 4 pri

e. Start 21 and 23 ABFP from the control room and supply AFW flow to the SGs during plant shutdown N, A, S, L 4 sec

f.

Manually initiate containment spray when actuation is N, A, S 5

required

g.

Not Required

h. Remove an Intermediate Range Channel from service 7

I In-Plant Systems (3 for RO; 3 for SRO-I; 3 or 3 for SRO-U)

i.

Manually start 21 Emergency Diesel Generator D

6

j.

Lineup alternate cooling to the SIS and RHR Pumps D, R 8

k. Align 24 Large Gas Decay Tank for start of discharge M, R 9

• Type Codes: (D)irect from bank, (M)odified from bank, (N)ew, (A)lternate path, (C)ontrol room, (SNmulator. (L)ow-Power. (RICA NUREG-1 021, Draft Revision 9

Appendix D Scenario Outline Form ES-D-1 Facility: Indian Point 2 Scenario No.: NRC#l Op-Test No.: 1 Examiners:

Operators:

Initial Conditions: 3% Rated Thermal Power, MOL Turnover: Unit 2 is at 3% power, recovering from a 7 day forced outage to repair body to bonnet leak on PRZR Spray Loop 23 Bypass Valve 524. Shift orders are to continue the startup in accordance with Pop 1.3 Plant Startup, Mode 2 to Mode 1. The previous shift completed POP 1.3 though step 4.23. The Operations Manager, Reactor Engineering and Power Marketing have authorized a rate of power increase of 200 MWe per hour to 100% RTP.

Event No.

1 -

Malf. No.

Event Type*

N CRS/BOP R

RO Event Description Raise reactor power 2

XMT-RCS020A I

ALL Pressurizer Level Channel 2 (LT-460) Fails Low (TS CRS) 3 MAL-RCSOl4B C

ALL 22 SG Tube Leak (5 gpm) (TS CRS) 4 MAL-RCP007C C

CRS/BOP 23 RCP High Vibration 5

MAL-ATS007B C

CRS/RO

~~

22 Main Boiler Feed Pump Trip (Manual reactor trip required) 6 MAL-RCSO14B M

ALL SGTR with subsequent Loss of Offsite Power MAL-EPS001 7

MOC-SIS001 C

CRSBOP 21 SI Pump Fails to Auto Start

'ity, (I)nstrument, (C)omponent, (M)ajor NUREG-1021, Draft Revision 9 1 of 1

Appendix D Scenario Outline Form ES-D?

2 Facility: Indian Point 2 Scenario No.: NRC#2 Op-Test No.: 1 MAL-EPS007D Examiners:

Operators:

4 Initial Conditions: 100% Rated Thermal Power, MOL. 21 EDG is out of Service. 22 Sharging Pump is out of service.

MAL-EPS001 MAL-DSG003B Turnover: Unit 2 is at 100% Power steady state conditions 340 EFPD. 21 EDG is out of service and has been inoperable for 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br />. Maintenance is currently performing

?epairs.

n addition, 22 Charging Pump was removed from service for corrective maintenance 18 lours ago. Expected return to service in 35 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br />.

swso10 sws-011 (N)ormal (R)eac Event Type*

N BOP/CRS R

RO C

ALL C

ALL M

ALL Event Description 23 EDG inoperable due to 86 Lockout Relay tripped Begin TS required plant shutdown Loss of Bus 6A (Lose 23 CHG Pump, Starts 21 CHG Pump) 21 Charging Pump Trips Manual Reactor Trip Loss of all AC Station Aux Xfmr fails 22 EDG fails to start

~~

25 SW Pump Fails to auto start following start of 22 EDG and energizing associated 480V buses rity, (I)nstrument, (C)omponent, (M)ajor NUREG-1 021, Draft Revision 9 1 of 1

Appendix D Scenario Outline Form ES-D-1 Facility: Indian Point 2 Scenario No.: NRC#3 Op-Test No.: 1 Examiners:

Operators:

Initial Conditions: 30% Rated Thermal Power, MOL Turnover: Unit 2 is at 30% Power with power ascension to 100% is in progress following a forced outage. No equipment 00s.

Malf. No.

Event Type*

N CRS/BO P R

RO Event No.

1 -

Event Description Raise reactor power 2

MAL-CRFOOl AY C

ALL Stuck Control Rod (P-1 0) 3 XMT-SG N026A I

RO/CRS LT 447 24 SG Controlling Level Channel fails low.

4 MAL-CCWOO1 D C

BOP/CRS RCP Upper Bearing Oil Cooler CCW leak

~

LBLOCA MAL-RCS001 C M

ALL 5

6 RLY-PPL487 PPL488 RLY-C RO Safety Injection auto actuation failure, RO manually actuates 7

MOC-RHR001 RHR002 MOC-C ALL Both Recirculation Pumps both fail to start

• (N)ormal (R)eacti\\ I, (I)nstrument, (C)omponent, (M)ajor NUREG-1 021, Draft Revision 9 1 of 1

Appendix D Scenario Outline Form ES-D-1 Event Malf. No.

Event No.

Type*

I MOT-C CFW003A BOP/CRS

~~

~

Facility: Indian Point 2 Event Description Condensate Pump trips Reduce steam flowefeed Flow Scenario No.: NRC#4 Op-Test No.: 1 2

3 4

Examiners:

Operators:

XMT-I VCT Level Transmitter fails low CVCOl9A ALL MAL-M Faulted Steam Generator SGN002B ALL Reactor auto and manual trips fail to actuate Bat FailRxTrips. bat AOV-C PORV Fails Open. Block valve used to isolate it.

RCS002A RO/CRS Initial Conditions: 100% Rated Thermal Power, MOL. 21 EDG is out of Service. 21 Charging Pump is out of service.

Turnover: Unit 2 is at 100% Power steady state conditions 340 EFPD. 21 EDG is out of service and has been inoperable for 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br />. Maintenance is currently performing repairs.

In addition, 22 Charging Pump.was removed from service for corrective maintenance 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> ago. Expected return to service in 35 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br />.

• (N)ormal (R)eactivity, (I)nstrument, (C)omponent, (M)ajor NUREG-1021, Draft Revision 9 1 of 1