ML13066A287

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Draft - Outlines (Folder 2)
ML13066A287
Person / Time
Site: Indian Point 
Issue date: 02/27/2013
From:
Entergy Nuclear Indian Point 3
To: D'Antonio J
Operations Branch I
Jackson D
Shared Package
ML12214A189 List:
References
TAC U01866
Download: ML13066A287 (33)
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Text

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

Indian Point Unit 3 Printed: 9/14/2012 Date Of Exam:

02/11/2013

3. Generic Knowledge And RO KIA Category Points SRO-Only Points Tier Group K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G* Total A2 G*

Total

1.

1 3

3 3

3 3

3 18 3

3 6

Emergency 2

2 2

2 N/A 1

1 N/A 1

9 2

2 4

Abnormal Tier Plant Evolutions Totals 5

5 5

4 4

4 27 5

5

\\0

2.

1 3

2 3

3 3

2 3

3 2

3 1

28 3

2 5

Plant 2

1 1

0 1

1 1

1 1

1 1

1 10 0 I 2

1 3

Systems Tier Totals 4

3 3

4 4

3 4

4 3

4 2

38 5

3 8

1 2

3 4

1 2

3 4

10 7

Abilities Categories 3

3 2

2 2

2 1

2 Note:

1.

Ensure that at least two topics from every applicable KiA category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the "Tier Totals" in each KiA category shall not be less than two).

2.

The point total for each group and tier in the proposed outline must match that specified in the table.

The final point total for each group and tier may deviate by +/-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.

3.

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.

4.

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.

5.

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.

6.

Select SRO topics for Tiers 1 and 2 from the shaded systems and KiA categories.

7.* 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 KiAs.

8. 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-only exams.
9.

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 1021

PWR RO/SRO Examination Outline Facility: Indian Pont Unit 3 NRC Written Examination Outline

~401 Emerg.ejcy ajd Abnormal Plant Evolutions - Tier 1/ Group 1 Form ES-401-2 E/APE #1 Name I Safety FUDction

[KI K2tQJ.~

Number LKlATopic

[Imp./ Q#

Knowledge of the interrelations between 000008 Pressurizer Vapor Space the Pressurizer Vapor Space Accident x

2.7 1

Accident 13 and the following: - Sensors and

~______+I~D~e~te~Qr~s______________________-+__~

Knowledge of the reasons for the following responses as they apply to the 000009 Small Break LOCA J 3 x

4.1 2

small break LOCA: - ECCS throttling or termination criteria Ability to determine and interpret the following as they apply to a Large Break 000011 Large Break LOCA J3 LOCA: - Conditions necessary for 3.4 3

recovery when accident reaches stable NUREG 1021 2

PWR RO/SRO Examination Outline Facility: Indian Pont Unit 3 ES-401 EJAPE #J 1

000026 Loss of Component Cooling Water /8 000027 Pressurizer Pressure Control System Malfunction / 3 000029 ATWS / 1 x

x Ability to determine and interpret the following as they apply to the Loss of Component Cooling Water: - Location of a I-__-----ilf-I--'-ea=k-'-i~nthe CCWS I

Knowledge of the operational implications of the following concepts as they apply to Pressurizer Pressure Control Malfunctions: - Expansion of liquids as increases Knowledge of the interrelations between the A TWS and the following: - Breakers, and discon 2.9 2.8 2.9 7

8 9

000038 Steam Gen. Tube Rupture I 3 000040 Steam LineRu:pture - J;xcesSfyf;l Heat (ransfer14' 000055 Station Blackout I 6 000056 Loss of Off-Site Power I 6 000057 Loss of Vital AC Inst. Bus 16 x

x Conduct of '""..............,'<> """"'Y'V """'....'. I 46 and execute I2rocedure steQs.

Equipment Control - Knowledge of the process for managing maintenance activities during shutdown operations, I 2.6 such as risk assessments, work etc.

Ability to operate and/or monitor the following as they apply to the Loss of I 3.6 Offsite Power: - Auxiliary/emergency feedwater Ability to operate and/or monitor the following as they apply to the Loss of Vital I 3 6 AC Instrument Bus: - Feedwater pump control oressure and level in S/G I

I I

I 11 12 4

13 NUREG 1021 3

PWR RO/SRO Examination Outline Facility:

Pont 3

NRC Written Examination Outline ES-401

_______..:::-::.::m::rl:.=-e=-.!rg"-'e:..:;:~=-=c..Ly--=a:..:;:~=-=d=-=A:.::.::;~normaI Plant Evolutions -

II G 1

Form ES-401-2 E/APE # I Name I 00005~tQSS Of;J.;lQPower/6

, 1':;: 1 "

/"~;"""<,'".,

000062 Loss of Nuclear Svc Water / 4 000077 Generator Voltage and Electric x

Grid Disturbances / 6 W/E04 LOCA Outside Containment I 3 W/E05lnadequate Heat Transfer - Loss x

of Secondary Heat Sink I 4 W/E11 Loss of Emergency Coolant X

EA1.3 Recirc./4 rv_" '- v~'~o IV;:>', IJIlt-'Clv\\'VI,l \\1,,1

,80

,!ilt.and ' ' "tor lant s stems Conduct of Operations* Ability to interpret and execute I2rocedure ste~s.

4.6 14 Knowledge of the reasons for the following responses as they apply to Generator Voltage and Electrical Grid Disturbances:. Actions contained in I 3.6 I 15 abnormal operating procedure for voltage and grid distu Ability to determine and interpret the following as they apply to the LOCA Outside Containment: - Adherence to appropriate procedures and operation I 3.6 I 16 within the limitations in the facility's license and amendments Knowledge of the interrelations between the Loss of Secondary Heat Sink and the following: - Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat I 3.9 I 17 removal systems, and relations between the proper operation of these systems to the ol2eration of the Ability to operate and/or monitor the following as they apply to the Loss of I Emergency Coolant Recirculation:

I 3.7 I 18 Desired operating results during abnormal and emeraencv situations NUREG 1021 4

PWR RO/SRO Examination Outline Facility: Indian Pont Unit 3 NRC Written Examination Outline 2

Knowledge of the reasons for the following responses as they apply to the 000033 Loss of Intermediate Range NI/7 x

Loss of Intermediate Range Nuclear Instrumentation: - Guidance contained in 3.6 19 EOP for loss of intermediate-range 1-__---lIJnstrum_e_ntGition_c-:--___--:-__--I__-t__-l Ability to operate and/or monitor the 000036 Fuel Handling Accident /8 x

following as they apply to the Fuel Handling Incidents: - Fuel handling 3.1 20 durina an incident Emergency Procedures/Plan - Knowledge 000068 Control Room Evac. / 8 of RO tasks performed outside the main control room during an emergency and 4.2 21 the resultant operational effects.

Knowledge of the interrelations between 000076 High Reactor Coolant Activity I 9 x

the High Reactor Coolant Activity and the 2.6 22 1-------11 fQllowing: - Process radiation monitors Ability to determine and interpret the following as they apply to the 81 W/E02 81 Termination / 3 x

Termination: - Facility conditions and selection of appropriate procedures 3.8 23 during abnormal and emergency NUREG 1021 5

PWR RO/SRO Examination Outline Facility: Indian Pont Unit 3 2

Function Knowledge of the interrelations between the LOCA Cooldown and Depressurization and the following: -

Facility's heat removal systems, including W/E03 LOCA Cooldown - Depress. 14 x

primary coolant, emergency coolant, the I 3.7 124 decay heat removal systems, and relations between the proper operation of these systems to the operation of the Knowledge of the operational implications of the following concepts as they apply to W/E06 Inad. Core Cooling I 4 x

the Degraded Core Cooling: - Normal, abnormal and emergency operating I 3.5 125 procedures associated with Degraded Core determine and. interpret the

.a$they~ply'to the Natural Wir:.10Nat~~1 Cir~, 14 wit,h Steam Void inVessel wiU!llwithPut R.\\IlIS: - Facil~QOff.llol!ls

~nd $$1~on()f.approP\\~epr~UJfes.

during ~al and e~v.'

Knowledge of the operational implications of the following concepts as they apply to W/E14 Loss of CTMT Integrity 15 x

the High Containment Pressure: -

Annunciators and conditions indicating I 3.3 126 signals, and remedial actions associated with the High Containment Pressure Knowledge of the reasons for the following responses as they apply to the W/E15 Containment Flooding 15 x

Containment Flooding: - Normal, abnormal and emergency operating I 2.8 127 procedures associated with Containment Floodi NUREG 1021 6

PWR RO/SRO Examination Outline Facility: Indian Pont Unit 3 NRC Written Examination Outline ES-401 S},stem #/Name 003 Reactor Coolant Pump 004 Chemical and Volume Control 005 Residual Heat Removal 005 Residual Heat Removal 006 Emergency Core Cooling 006 Emergency Core Cooling 007 Pressurizer Relief/Quench Tank 001 Pressurizer Relief/Quench Tank DOS Component Cooling Water

..... u~ill1iiiS - Tier 2 / Group 1 Form ES-401-2 lKluL~2 LR~J K41 KS] K§ u,t\\J.n,t\\3~Nu~er nl KIA Topic I pi

-. AbilitY to (a) predict the impacts of the following malfunctions or operations on the RCPS and (b) based on those Ipredictions, use procedures to correct, A2.05 12.5 128 control, or mitigate the consequences of those malfunctions or operations: - Effects of VCT pressure on RCP seal leakoff flows Knowledge of the effect of a loss or Imalfunction on the following CVCS I 2.7 I 29 Ix I K6.29 components: - Reason for excess letdown and jts r~IC3Jipnshie to CCWS Knowledge of the physical connections Iand/or cause-effect relationships between I 3 2 I 30 Ix I K1.01 the RHRS and the following systems: -

CCWS I Knowledge of bus power supplies to the following: - RCS pressure boundary 12.7 131 K2.03 Ix I I I motor-o~erated valves Knowledge of the effect of a loss or Ix K6.05 Imalfunction of the following will have on 13.0 132 I

I I

I the ECCS: - HPIILPI cooling water Knowledge of the operational implications Iof the following concepts as they apply to 12.S 133 I I Ix I I K5.01 the ECCS: - Effects of temperatures on water level indications Knowledge of the physical connections Iand/or cause-effect relationships between I 3 0 I 34 Ix I K1.03 the PRTS and the following systems: -

RCS I

I r,-CftlJtl

. *A~t'IiilI1i!',

Knowledge of the physical connections

~d10r cause-effect relationships between I 3 1 01 135 Ix I I I the CCWS and the following systems: -

~~~'"

NUREG 1021 7

PWR RO/SRO Examination Outline Facility: Indian Pont Unit 3 NRC Written Examination Outline U

~::t!~~ #lName

.... I KITI<.2TK3T ](11I<.5JKlta:i_~;y;:/.o_~_~_!--,b-,-er_LI~K_**I_A_T_O.......P_iC_________--L.

010 Pressurizer Pressure ContrOl x

Knowledge of PZR PCS design feature(s) and/or interlock(s) which provide for the followina: - Over pressure control I 3.8 I 36 Ability to predict and/or monitor changes 010 Pressurizer Pressure Control x

in parameters (to prevent exceeding design limits) associated with operating the PZR PCS controls including: - RCS I 3.7 I 37 012 Reactor Protection automatic operation of the RPS. includina: - Trio breakers 4.0 38 Knowledge of RPS design feature(s) and/or interlock(s) which provide for the 012 Reactor Protection x

following: - Automatic reactor trip when 3.9 39 RPS setpoints are exceeded for each 1-___+I..CcR~P....:S::...::=fu::.::.n9.tion; basis for each Knowledge of ESFAS design feature(s) 013 Engineered Safety Features Actuation x

and/or interlock(s) which provide for the following: - Safeguards equipment control 3.3 40 reset prO(:edI,HieSitoCQEfect.

mltlgat~ the.consequences Inclinn~ or nt'IAr:;ltinn~'

Ability to predict and/or monitor changes 022 Containment Cooling x

in parameters (to prevent exceeding design limits) associated with operating the CCS controls including: - Cooling 3.2 142 water flow 026 Containment Spray x

~~~~~--~--------I----+---~

Knowledge of the effect that a loss or malfunction of the CSS will have on the following: - CCS 3.9 143 Knowledge of the operational implications 039 Main and Reheat Steam x

of the following concepts as they apply to the MRSS: - Effect of steam removal on 3.6 44 NUREG 1021 8

PWR RO/SRO Examination Outline Facility: Indian Pont Unit 3 059 Main Feedwater 059 Main Feedwater 061 Auxiliary/Emergency Feedwater 061 Auxiliary/Emergency Feedwater 061 Auxiliary/Emer9ef1ey~

Fe~watet..

002.AC*Electrical.

Distribution 063 DC Electrical Distribution X

A1.03 x

Ability to predict and/or monitor changes in parameters (to prevent exceeding I design limits) associated with operating 12.7 145 the MFW System controls including:

Power level restrictions for operation of I\\i1FW ~um~s and valves Ability to manually operate and/or monitor in the control room: - Feedwater control 12.9 146 increase and decrease Knowledge of the operational implications of the following concepts as they apply to 12.6 147 the AFW System: - Pump head effects when control valve is shut Ability to (a) predict the impacts of the following malfunctions or operations on the AFW System and (b) based on those predictions, use procedures to correct, 13.2 141 control, or mitigate the consequences of those malfunctions or operations: - Loss of air to steam su~~ly valve Cond~,~ Qper:ations.,. Ability tot~l~i.,.

86 48 Ability to manually operate and/or monitor in the control room: - Battery discharge 1 3.0 1 49 rate NUREG 1021 9

PWR RO/SRO Examination Outline Facility: Indian Pont Unit 3 064 Emergency Diesel Generator 076 Service Water

.. A2.02 076 Service Water Ix K2.01

_X 078 Instrument Air A3.01 103 Containment Ix I

.. K3.01 3.7 150 3.7 151 Ability to (a) predict the impacts of the following malfunctions or operations on the SWS and (b) based on those I predictions, use procedures to correct, 12.7 152 control, or mitigate the consequences of those malfunctions or operations:

Service water header ~ressure I Knowledge of bus power supplies to the following: - Service water 12.7 153 I Ability to monitor automatic operation of I 3.1 154 the lAS, including: - Air ~ressure Knowledge of the effect that a loss or malfunction of the Containment System I will have on the following: - Loss of 13.3 155 containment integrity under shutdown conditions NUREG 1021 10

Facility: Indian Pont Unit 3 001 Control Rod Drive 002 Reactor Coolant 002 RuctorCoolantiJ*.. :4t,,: 1['.'

011 Pressurizer Level Control I

I I 015 Nuclear Instrumentation 017 In-core Temperature Monitor 027 Containment Ix I Iodine Removal PWR RO/SRO Examination Outline Ability to (a) predict the impacts of the following malfunctions or operations on the RCS and (b) based on those

.. A2.04 I predictions, use procedures to correct, 14.3 157 control, or mitigate the consequences of those malfunctions or operations: - Loss of heat sinks Conduct of

  • ... I..li.J*lT;f*..*t)r~.'!'.j;2.1.42i/!:newaoo....!;

.i*. '.

procedures.

ic.

Knowledge of the operational implications Iof the following concepts as they apply to I 3 6 I 58 I Ix I I

K5.15 the PZR LCS: - PZR level indication when I

RCS is saturated IAbility to monitor automatic operation of X

the NIS, including: - Annunciator and 13.7 159 A3.02 alarm Ability to manually operate and/or monitor Iin the control room: - Temperature values used to determine RCS/RCP operation I 3.8 I 60 Ix

  • A4.02 during inadequate core cooling (Le., if ble, average of five highest va Knowledge of the physical connections Iand/or cause-effect relationships between I 34 I 61 K1.01 the CIRS and the following systems: -

CSS NUREG 1021 11

PWR RO/SRO Examination Outline Facility: Indian Pont Unit 3 NRC Written Examination Outline ES-401 uu mm un~lant_ms - Tier 21 Group 2m u.

Form E~401-2n_

System #lName I KI I K2 I K3 I K4 [K5J K6 I Al A3 I A4

  • Number I.KlA Topic Um}U Q#

Knowledge of Containment Purge System 029 Containment design feature(s) and/or interlock(s) which x

3.2 62 Purge provide for the following: - Automatic isolation Knowledge of the effect of a loss or 034 Fuel Handling malfunction of the following will have on x

2.6 63 Equipment the Fuel Handling System: - Radiation Conduct of Operations - Knowledge of 3.9 164 and or function.

Ability to predict and/or monitor changes in parameters (to prevent exceeding 086 Fire Protection x

design limits) associated with operating 2.9 165 the Fire Protection System controls includina: - FPS lineu NUREG 1021 12

Generic Knowledge and Abilities Outline (Tier 3)

PWR RO/SRO Examination Outline Facility: Indian Pont Unit 3 Form ES-401-3 Facili Indian Point Unit 3 Date of Exam 2/11/2013 RO SRO-On Category KIA #

Topic IR Q#

IR Q#

Ability to use procedures related to shift staffing, such as minimum crew 2.1.5 2.9 66 complement, overtime limitations, etc.

Ability to perform specific system and integrated plant procedures during all 4.3 67

  • modes of plant operation.
1. Conduct of Knowledge of procedures, guidelines, Operations or limitations associated with reactivity 2.1.37 4.3 68 management.

Know:ledgflofrefue,u"gadministrat~~..,.

Equipment Control-Ability to perform pre-startup procedures for the facility, 2.2.1 including operating those controls 4.5 69 associated with plant equipment that could affect reactivit.

Ability to manipulate the console controls as required to operate the 2.2.2 4.0 70 facility between shutdown and desi nated ower levels.

2. Equipment Ability to analyze the effect of Control maintenance activities, such as 2.2.36 3.1 71 degraded power sources, on the status of Iimitin conditions for 0 erations.

Knowledg~ of Iimitingcond!tiQns.fo" oerations and safe limits'..

  • ~O\\Nff?4g. ofQas6$ in t '.

.$~cificati<?nsf~f timiting' Co 0

.ratiOns and' sa limits.

3 3

2 NUREG 1021 13 2

Generic Knowledge and Abilities Outline (Tier 3)

PWR RO/SRO Examination Outline Facility: Indian Pont Unit 3 Facilit Indian Point Unit 3 Date of Exam Category KJA #

Topic Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment 2.3.12 entry requirements, fuel handling responsibilities, access to locked high radiation areas, ali nin filters, etc.

Knowledge of radiological safety procedures pertaining to licensed

3.

operator duties, such as response to Radiological 2.3.13 radiation monitor alarms, containment Controls entry requirements, fuel handling responsibilities, access to locked high I radiation areas, ali nin filters, etc w

~AbUity to 1J~'radiati9fl

!";.'.\\,

2.3.5 s~rps. $~9~

moniror$"al1'd*****

5, partablesf.!rvey instruments, personal monitoring e uiment, etc Subtotal Knowledge of the organization of the operating procedures network for 2.4.5 normal, abnormal, and emergency evolutions.

Knowledge of abnormal condition 2.4.11 rocedures.

4. Emergency Knowl~ge of the bases for prior!~!~nl'.

Procedures/plan safetyfunction5 du~il1g e

Tier 3 Point Totals Form ES-401-3 2/11/2013 RO SRO-Only IR Q#

IR Q#

3.2 72 3.4 73 2

3.7 74 100 2

10 10 7

7 NUREG 1021 14

Randomly Selected KIA

,~~

~ ~ -----~ ~

Conduct of Operations - Knowledge of facility requirements for controlling vitali controlled access.

Ability to operate andlor monitor the following as they apply to the Reactor Coolant Pump Malfunctions: - RCP on/off and run indicators

-~~

~-----~~

Ability to determine and interpret the following as they apply to the Loss of DC Power: - That a loss of dc power has occurred; verification that substitute po",!~r sources have come on line Conduct of Operations - Knowledge of non-nuclear safety procedures (e.g. rotating equipment.

electrical, high temperature, high pressure, caustic,

~~Iorine, oxygen and hydrogen.

Ability to operate and/or monitor the following as they apply to the Reactor Coolant Pump Malfunctions: - RCP bearing temperature indicators I---~~

Knowledge of the reasons for the following responses as they apply to the Emergency Boration:

- Actions contained in EOP for emergency boration Conduct of Operations - Knowledge of refueling administrative requirements.

Knowledge of the operational implications of the following concepts as they apply to the Steam Generator Overpressure: - Annunciators and conditions indicating signals, and remedial actions associated with the Steam Generator Overpressure

~---.-

~----.-

Reason for Rejection REJECTED Generic KA is not applicable to Large Break LOCA.

REJECTED This KA is evaluated during Simulator Operational Examination.

REJECTED Not applicable to Indian Point 3.

Unit 3 does not have any automatic swap to alternate DC source functions.

REJECTED Generic KA is not applicable to LOCA Outside Containment RREJECTED This KA was rejectd due to overlap with SRO Only Question 76.

REJECTED This KA overlaps with a currently selected JPM.

REJECTED Over sample, This Generic KA selected for SRO Only Question 94 REJECTED This is a Yellow Path Functional Restoration Procedure. The condition is generally corrected without entering the procedure. Not able to write a discriminatory RO Question that is not too trivial.

0030002314 Radiological Controls - Knowledge of radiation or REJECTED Generic KA not applicable to the contamination hazards that may arise during normal, R 2/1 Reactor Coolant Reactor Coolant Pump System.

abnormal, or emergency conditions or activities.

Pump Knowledge of the physical connections and/or 004000K134 cause-effect relationships between the CVCS and REJECTED Plant configuration NA at IPEC.

R-2/1 Chemical and the following systems: - Interface between CVCS KA is Not Applicable Volume Control and reactor coolant drain tank; and PZR PCS

~-

Radiological Controls - Ability to use radiation monitoring 0220002305 systems, such as fixed radiation monitors and alarms, REJECTED Over sample Radiation Monitors R-2/1 Containment portable survey instruments, personnel monitoring (Questions 22, 51, 63, 90, 98) equipment, etc.

r CqQ!iD9 Conduct of Operations - Ability to locate control 0390002131 room switches, controls and indications and to REJECTED This Generic KA is evaluated R 2/1 Main and Reheat determine that they are correctly reflecting the during Simulator Operational Examination.

Steam System desired plant lineup:

064000K301 Knowledge of the effect that a loss or malfunction of REJECTED Indian Point 3 does not have an the ED/G System will have on the following:

R 2/1 Emergency Diesel automatic loader. KA is Not Applicable.

Systems controlled by automatic loader Generator 073000A403 Ability to manually operate and/or monitor in the Process Radiation control room: - Check source for operability R 211 REJECTED Overlap with SRO question 90 Monitoring demonstration System Emergency Procedures/Plan - Knowledge of EOP 0150002416 implementation hierarchy and coordination with REJECTED Generic KA only applicable to other support procedures or guidelines such as, Nuclear R2/2 A TWS condition at Indian Point. This would operating procedures, abnormal operating Instrumentation be overlap with SRO question 78 procedures, severe accident management System guidelines.

033000A301 Ability to monitor automatic operation of the Spent REJECTED Indian Point 3 does not have Fuel Pool Cooling System, including: - Temperature R 2/2 Spent Fuel Pool automatic temperature control valves.

control valves CqQliD9

R3 1940012144 Conduct or Operations R3 1940012305 000029A206 S 1/1 S 1/1 Anticipated Transient Without Scram (A TWS) 0000652105 Loss of Instrument Air S 1/1 00WE042220 LOCA Outside Containment S 1/1 0000112104 Large Break LOCA 00WE112450 S 1/1 Loss of Emergency Coolant Recirculation Conduct of Operations - Knowledge of RO duties in the CR during fuel handling, such as responding to alarms from the fuel handling area, communication with the fuel storage facility, systems operated from the CR in support of fueling operations, and supporting instrumentation.

Radiological Controls - Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

Ability to determine and interpret the following as they apply to a A TWS: - Main turbine trip switch position indication Conduct of Operations - Ability to use procedures related to shift staffing such as minimum crew compliment, overtime limitations, etc.

Equipment Control - Knowledge of the process for managing troubleshooting activities.

Conduct of Operations - Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, "no solo" operation, maintenance of active license status, 10CFR55!~tc.

Emergency Procedures/Plan - Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.

REJECTED At Indian Point 3 the ROs in the control room respond to plant alarms as necessary. All other activities are performed by refueling team members not necessarily any RO REJECTED Overlap with SRO Question 98.

REJECTED Equipment (Main Turbine Trip Switch) does not exist at IPEC.

REJECTED Generic KA is not applicable to Loss of Instrument Air REJECTED Generic KA is not applicable to LOCA Outside Containment REJECTED Generic KA is not applicable to Large Break LOCA REJECTED. This Generic KA is best evaluated during Simulator Scenarios. Not able to write a discriminatory SRO question.

00WE062142 S 1/2 Degraded Core Gooling 003000221 S 2/1 Reactor Coolant Pump 0060002141 S 2/1 Emergency Core Cooling System 0110002314 S 2/2 Pressurizer Level Control System Conduct of C spent fuel me perations - Knowledge of new and vement procedures.

REJECTED Generic KA is not applicable to Degraded Core Cooling Equipment C procedures f(

controls assc affect reactivi ontrol - Ability to perform pre-startup r the facility, including operating those ciated with plant equipment that could ty.

REJECTED Generic KA refers to subsequent step information in an SOP that requires contacting Reactor Engineering for guidance.

Unable to write a discriminatory SRO Conduct of 0 perations - Knowledge of the refueling REJECTED Generic KA is not applicable to process.

Radiological contaminatio abnormal, or I-----

Emergency Core Cooling System Controls - Knowledge of radiation or n hazards that may arise during normal, IREJEC~ED Generic KA is not applicable to emergency conditions or activities.

Pressurizer Level Control System

ES-301 Administrative Topics Outline Form ES~301 ~1 Facility:

IPEe Unit 3 Examination Level: Ro8 Administrative Topic (see Note)

Conduct of Operations Conduct of Operations Equipment Control Radiation Control Emergency Procedures/Plan Date of Examination:

2l:l 1l2Q13 SRO D Operating Test Number:

l Type Describe activity to be performed Code*

Perform a QPTR Calculation 2.1.23 - Conduct of Operations - Ability to perform M

specific system and integrated plant procedures during all modes of plant operation.

RO-4.3 Determine Deborating Demineralizer In Service Time per 3-S0P-CVCS-4 N

2.1.25 - Conduct of Operations - Ability to interpret reference materials such as graphs, curves, tables etc.

RO-3.9 Prepare a Manual Taggout of 31 Recirc Pump N

2.2.13 - Equipment Control-Knowledge of tagging and clearance procedures.

RO -4.1 Prepare a Manual Gaseous Waste Release Permit Calculation M

2.3.11 - Radiological Controls - Ability to control radiation releases.

RO-3.B NA for ROs NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.

  • Type Codes & Criteria:

(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (: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)

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

I~EC Unit 3 Examination Level: RO D Administrative Topic (see Note)

Conduct of Operations Conduct of Operations Equipment Control Radiation Control Emergency Procedures/Plan SRO 8 Date of Examination:

2l:1 :1l2Q:13 Operating Test Number:

1 Type Code*

Describe activity to be performed Review a QPTR Calculation M

2.1.23 - Conduct of Operations - Ability to perform specific system and integrated plant procedures during all modes of plant operation.

SRO-4.4 N

Review Deborating Demineralizer In Service Time Calculation per 3-S0P-CVCS-4 2.1.25 - Conduct of Operations - Ability to interpret reference materials such as graphs, curves, tables etc.

SRO-4.2 Review a Manual Taggout of 31 Recirc Pump N

2.2.13 - Equipment Control - Knowledge of tagging and clearance procedures.

SR04.3 Review a Manual Gaseous Waste Release Permit Calculation M

2.3.11 - Radiological Controls - Ability to control radiation releases.

SRO -4.3 Classify E-Plan Event and Complete Part 1 Form M

2.4.41 - Emergency Procedures/Plan - Knowledge of the emergency action level thresholds and classifications.

SRO-4.6 NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.

  • Type Codes & Criteria:

(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (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)

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

IPEC UNIT 3 Date of Examination:

2-7-2013 Exam Level: RO ~ SRO-I D SRO-U D Operating Test No.:

1 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System I JPM Title Type Code*

Safety Function

a. Transfer to Hot Leg Recirculation A,EN 2

000011 A.1.11 RO-4.2 SRO-4.2

b. Respond to Failed Open Spray Valve A,N 3

000027A1.01 R04.0 SRO-3.9

c. Respond to #1 Rep Seal Failure D,P 4P 003000A2.01 RO-3.S SRO-3.9
d. Transfer from Low Flow Bypass Valves to Main Feed Reg Valves N, L 4S OS9000A4.03 RO-2.9 SRO-2.9
e. Start a Fan Cooler Unit N

S 022000A4.01 RO - 3.6 SRO - 3.6

f. Bring in 13.8 KV Power D

6 062000A4.01 RO - 3.3 SRO - 3.1

g. 33 SG Pressure Channel Fails Low (Alt Path)

A 7

03S000K401 RO - 3.6 SRO - 3.8

h. Reset Rad Monitor 27 Alarm Setpoint Bantam 11 M

9 073000A402 RO - 3.7 SRO - 3.7 In-Plant Systems(g/ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

i. Locally Emergency Borate A,E,R 1

000024A104 RO - 3.6 SRO - 3.7

j. Align City Water to RHR D,E, R 8

OOSOOOK101 RO - 3.2 SRO - 3.1

k. Secure release lineup for 31 SGDT and lineup 32 SGDT for N,R 9

release 071000A40S RO - 2.6 SRO - 2.6 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 Criteria for RO I SRO-II SRO-U I

(A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power 1Shutdown (N)ew or (M)odified from bank including 1(A)

(P)revious 2 exams (R)CA (S)imulator 4-6 14-6 12-3 s9/s8/s4

2:1/:2:1/:2:1 1 :2:1 (control room system)
2:1/:2:1/:2:1
2:2/:2:2/:2:1 s 3 1s 3 1s 2 (randomly selected)
2:1/:2:1/:2:1

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

IPEC UNIT 3 Date of Examination:

2-7-2013 Exam Level: RO D SRO-I ~ SRO-U D Operating Test No.:

1 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System / JPM Title Type Code*

Safety Function

a. Transfer to Hot Leg Recirculation A,EN 2

000011 A.1.11 RO-4.2 SRO-4.2

b. Respond to Failed Open Spray Valve A,N 3

000027A1.01 RO-4.0 SRO-3.9

c. Respond to #1 RCP Seal Failure D,P 4P 003000A2.01 RO-3.S SRO-3.9
d. Transfer from Low Flow Bypass Valves to Main Feed Reg Valves N, L 4S OS9000A4.03 RO-2.9 SRO-2.9
e. Start a Fan Cooler Unit N

S 022000A4.01 RO - 3.6 SRO - 3.6

f. Bring in 13.8 KV Power D

6 062000A4.01 RO - 3.3 SRO - 3.1

g. 33 SG Pressure Channel Fails Low (Alt Path)

A 7

03S000K401 RO - 3.6 SRO - 3.8 M

9

h. NASROI In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)
i. Locally Emergency Borate A, E, R 1

000024A104 RO - 3.6 SRO - 3.7

j. Align City Water to RHR D,E,R 8

00SOOOK101 RO - 3.2 SRO - 3.1

k. Secure release lineup for 31 SGDT and lineup 32 SGDT for N,R 9

release 071000A40S RO - 2.6 SRO - 2.6 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 Criteria for RO / SRO-I / SRO-U I

(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/s8/S4

~1/~1/~1

- / -

/

~1 (control room system)

~1/~1/~1

~2/~2/~1 s 3 / s 3 / s 2 (randomly selected)

~1/~1/~1

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

IPEG Unit 3 Scenario No. 1 Op-Test No.: 1 Examiners:

Operators:

Initial Conditions:

100% power steady state. 32 AFW Pump is out of service for bearin9 repair.

Need to have 35 SWP in service.

Turnover:

32 AFW Pumps is expected back in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Team has been asked to reduce power to 90% for a Turbine StOI2 and Control Valve Test.

Event Malt.

Event Event No.

No.

Type*

Description 1

N/A R(ATC)

Perform power reduction.

N (BOP)

N (CRS) 2 MAL C (CRS) 33 MFRV fails open in auto manual available.

CFW013 E

C (ATC) 33 MFRV will be placed in manual to control 33 SG level.

3 MAL-e (ALL)

NRHX Tube Leak.

CFW011 Letdown will be isolated and excess letdown will be placed in service.

4 MAL C (CRS)

Loss of 480V Bus 3A.

EPS05B C (BOP) 34 or 36 SWP will be started.

TS CRS will determine T.S. shutdown is required due AFW pumps. -

(CRS) 5 MAL C (CRS)

Inadvertent main generator trip. Reactor will not automatically trip.

GEN002 C (ATC)

Manual trip will be required.

6 MAL M (ALL) 33 AFW pump will not start causing a loss of heat sink. Heat sink CFW001 will be restored using condensate.

C 7

MAL TS PORV 455C leaks by requiring isolation. CRS can be asked about PRS003 (CRS)

TS requirements at end of scenario.

0 C (ATC)

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

The evaluation begins with the plant at 100% power steady state operation.

  • 32 ABFP 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 /> (TS 3.7.5 - 72 hr AOT). 31 and 33 ABFPs are protected equipment.
  • The team has been requested to reduce power to 90% in preparation for a turbine stop and control valve test.

After the team has begun reducing reactor power, 33 Main Feed Regulating valve (FCV-417) fails open. The team will take actions per AOP-FW-1, "Loss of Feedwater" and place the MFRV in manual to control SG level.

During subsequent actions of AOP-FW-1, a leak develops in the !'Jon-Regen heat exchanger.

The team will respond per AOP-CVCS-1 and isolate Letdown.

While putting in Excess Letdown, an electrical fault occurs on 480V Bus 3A. The team will have to start an essential service water pump per AOP-480V-1. The CRS will determine that two AFW Pumps are inoperable necessitating a shutdown per T.S. 3.7.5 or safety function determination per 3.0.6.

The MTG will then trip inadvertently. The reactor will not trip automatically and will require manual trip at the flight panel. After the trip, 33 AFW Pump will not start, and one of the PORVs begins to leak by creating an isolable SBLOCA. The team will respond to the PORV failure by closing its block valve. The team will initiate performance of E-O, "Reactor Trip or Safety Injection".

Since no motor-driven AFW pump is available, the team will transition to FR-H.1, "Loss of Secondary Heat Sink" due to inadequate AFW flow.

The scenario will terminate when Condensate flow is established to the SGs Procedure Flowpath: AOP-FW-1, AOP-CVCS-1, AOP-480V-1, E-O, FR-H.1 Critical Tasks are as follows:

E-O - A Manually trip reactor from the CCR before completing E-O, Step 1 E-O - M Close block MOV upstream of stuck-open Pzr PORV by completion of the first step in the ERG network that directs the crew to close the block MOV FR-H.1 - A Establish Condensate or Main Feed flow to at least one SG before 3 out 4 SG WR levels < 20%

Appendix 0 Scenario Outline Facility:

Indian Point 3 Scenario No.: _2_

Op-Test No.:

Examiners:

Operators:

Initial Conditions:

Reset simulator to 5% Load Simulator Schedule-Scenari02 The Plant is in Mode 1 just above 9% power preparing to come on line.

Turnover:

Raise power to approximately 12% to place MTG in service. No equipment is out of service.

Event Malf.

Event Event No.

No.

Type*

Description 1

N/A R(ATC}

Power ascension, maintain SG levels in manual.

N(CRS}

N(BOP}

2 MOC C(CRS) 31 CCW Pump trip with failure of 32 to auto start.

SWSOO7 C(BOP)

TS(CRS}

3 XMT I(ALL)

Controlling PZR Pressure transmitter fails high.

RCS028 A

TS(CRS) 4 MOV C(CRS)

FCV-625 spurious closure.

CCWOO 8

C(BOP) 5 MAL ~

Steam leak in the Turbine Building leading to plant triD. 33 1\\11 c::: 1\\1 SGNOO fails to close.

6 BKR C(ALL)

Entered at setup, Reactor Trip Breakers will not open causing PPLOO31 team to enter FR-S.1.

4 7

RLY C(CRS)

Entered at setup, SI does not automatically actuate. Manual PPL4871 8

C(BOP) actuation will be required.

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

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.

After taking the watch, the crew will commence raising power. After the power escalation has progressed, 31 CCW Pump will trip. The team will start another pump per ARP-010.

Following the restoration of CCW, a failure high of PT-455 will occur. The team will respond using AOP-INST-1 "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 ARP-01 0 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 of AOP-CCW-1, 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 FR-S.1, "Response to Nuclear Power Generation I ATWS," and will shutdown the reactor by manually inserting control rods and initiating Emergency Boration. The reactor trip breakers will be locally opened after an NPO is dispatched. 33 MSIV will fail to close from the control switches. The team will proceed through FR-S.1 until transition to E-O, "Reactor Trip or Safety Injection."

After the transition to E-O is made, the team will determine that three SGs are intact and 33 SG is faulted. The Team will also determine that SI did not automatically actuate and must manually actuate SI. The team will transition to E-2, Faulted Steam Generator Isolation and isolate 33 SG. At this point the scenario is terminated.

Procedure flow path: POP-1.3, ARP-010, AOP-INST -1, ARP-010, AOP-UC-1, E-O, FR S.1, E-O, E-2 Critical Tasks:

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

De-energize the control rod drive MG sets Place rod control in manual and insert RCCAs Establish emergency boration flow to the RCS.

CT-2 Establish at least 686 gpm AFW flow to the SGs before completion of FR-S.1 step

3.

CT-3 Manually actuate at least one train of SIS actuated safeguards before completion of E-O step 4.

Appendix D Scenario Outline Form ES-D-1 1

Facility: IPEC Unit 3 Scenario No.:

3 Op-Test No.:

Examiners:

Operators:

Initial Conditions:

34% Power IC-153 (T.J J't../

l C {l-L C-'l-.f:::.Att Turnover:

Raise Power to 100%.

Even Malf. No.

Event Event t No.

Type*

Description 1

N/A R (ATC)

N (CRS)

Power Increase N (BOP) 2 MAL PRS006B I (ALL)

TS (CRS)

Controllill9 Pressurizer Level Channel fails low 3

MAL TUR010B I (ALL)

TS (CRS)

PT-412B failure causing AIVISAC to be inoperable and requiring tripping of bistables 4

MAL M (ALL)

Loop Flow Transmitter penetration failure (SBLOCA)

RCS017A 5

MAL CFW001A C (BOP)

C (CRS) 31 AFW Pump Trip 6

MAL SGN005B M (ALL)

SG Tube Leak to Rupture 32 SG (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

~.:Jfl-J

[{; r'ilt.1

.>4 c.,('~" (

r,A ( (.,

r" i

Session Outline:

The evaluation begins with power is at 34% steady state. The crew will be directed to raise power to 100%.

After adequate power change has occurred, Pressurizer Level Channel L T-460 will fail low resulting in a loss of letdown. The crew will respond in accordance with 3-AOP-INST-1, Instrument and Controller Failures. The procedure will address both the instrumentationlTech Spec actions and restoration of charging and letdown. This is a Tech Spec for Instrumentation.

At lead examiner discretion, PT-4128 will fail low requiring bypassing AMSAC and tripping bistables. This is a Tech Spec for Instrumentation.

At lead examiner discretion, 31 Loop Flow Transmitter Penetration will fail resulting in a small break LOCA. The crew may respond using 3-AOP-LEAK-1. The crew will likely trip the reactor and manually actuate Safety Injection.

Following the Safety Injection 31 AFW pump will trip on overcurrent. This will require the crew to start 32 AFW pump to supply flow to 31 and 32 SG.

The crew will proceed through E-O to E-1. When the crew enters E-1 a SGTR will occur in 32 SG. The crew will transition to E-3 when diagnostic check for SGTR or on Foldout page based on uncontrolled level increase.

The crew will perform actions in E-3 and transition to ECA-3.1 due to LOCA with SGTR.

The scenario will be terminated when the team demonstrates they control plant Cooldown or at the lead examiners discretion.

Procedure Flowpath: POP-1.3, AOP-INST-1, AOP-INST-1, AOP-LEAK-1, E-O, E-1, E-3, ECA 3.1 Critical Tasks:

CT-1 Establish at least 365 gpm AFW flow using 32 AFW Pumps prior to exiting E-O Step 5.

CT-2 Isolate feed to 32 SG prior to completion of E-3 Step 3.

CT-3 Initiate RCS Cooldown of <100 °F/hr to prevent SG overfill in ECA-3.1

Appendix 0 Scenario Outline Form ES-O-1 Facility: IPEC Unit 3 Scenario No. 4 Op-Test No.:

1 Examiners:

Operators:

Initial Conditions:

100% power steady state. 32 HHSI pump is OOS, expected to return to service in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Turnover: Maintain 100% Power Event Malt.

Event Event No.

No.

Type*

Description 1

MAL-I (CRS)

PT-404 failure requiring manual control of Main Boiler Feed Pumps.

MSS003 I (ATC) 2 MAL-C (ALL)

Fault on 480V Bus 6A. This will require starting an ESW pump and EPS005 charging pump.

TS D

(CRS) 3 N/A R (ATC)

T.S. load reduction.

N (CRS)

N (BOP) 4 MAL-I (ALL) 31 SG Press Channel Fails High PT-412B failure causing AM SAC

(/'-

SGN002 to be inoperable and requiring tripping of bistables J

C TS 51' iL"? I (CRS)

Vr'lJ(JI--'

~ ".

5 MAL-M (ALL)

SBLOCA RCS001 (t.? '7 A

6 MAL-C (ALL)

ATWS RPS002 AlB 7

MAL-C (BOP No Auto SI SIS001A C (CRS)

/B (N)ormal, (R)eaciivity, (I)nstrument.

(C)omponent.

(M)aior

Summary The evaluation begins with the plant at 100% power steady state operation. 32 SI pump is out of service for bearing inspection. 32 SI Pump was removed from service 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> ago, expected return in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

After the crew assumes the watch, PT ~404 Steam Header Pressure fails high. ATC diagnoses failure and places Main Boiler Feed Pump speed control to manual. CRS directs crew to perform AOP~INST-1, Instrument/Controller Failures.

Next, a fault occurs on Bus 6A. Team enters AOP-480V~1, Loss of 480 V Bus to stabilize the plant. The crew will start a charging pump, ESW Pump and determine that a shutdown must commence.

During the shutdown, 31 SG C Channel Pressure instrument will fail high causing the Atmospheric Dump Valve to go open. The crew will respond in accordance with 3~AOP-INST-1 JInstrument and Controller Failures.

After sufficient shutdown, a SBLOCA occurs. Auto reactor trip is demanded but the Reactor trip breakers do not open. The reactor will not trip from the Control Room and the team will respond per FR-S.1, "Response to Nuclear Power Generation I A TWS" and will SID the reactor by manually inserting control rods and initiating Emergency Boration via an alternate method since MOV-333 cannot be opened from the Control Room. The turbine will be manually tripped. The reactor trip breakers will be locally opened and the team will transition to E~OJ "Reactor Trip or Safety Injection". The SM will declare a Site Area Emergency.

An Auto SI is required, but will not actuate. The RO will manually actuate SI. 33 SI pump cannot be powered. 32 SI pump is OOS. 31 SI pump initially starts but trips. 31 and 32 Charging Pumps are eventually started after manual action to align suction to the RWST. The team will diagnose RCS not intact and transition to EOP E-1, Loss of Reactor or Secondary Coolant.

As the team progresses through E-1, Reactor Vessel level will continue to lower. When level lowers to 44% RCP running range (after about 30 minutes following the ATWS), an orange path on core cooling will require a transition to FR-C.2, Response to Degraded Core Cooling.

The team will initiate depressurization of SGs per FR-C.2. The scenario is terminated after SG depressurization has been initiated.

Procedure Flowpath: AOP-INST-1, AOP-480V-1, AOP-INST-1, FR-S.1, EOP E-O, EOP E-1, FR-C.2 Critical Tasks are as follows:

FR-S.1 - C Insert negative reactivity into the core by at least one of the following methods before step 4 of FR-S.1 is complete:

o De-energize the Rod Drive MG Sets o

Manually insert the rods o

Establish Emergency Boration FR-S.1 -- B Start AFW pumps and establish 686 gpm before transition out of FR-S.1 E~O -- D Manually actuate at least one train of SIS actuated safeguards before completion of FR S.1 step 6 FR-C.2 - A Depressurize SGs to atmospheric pressure (at < 100°F/hr) to inject ECCS accumulators and establish low-head injection flow before a Core Cooling red path develops.

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