ML12214A258

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Initial Exam 2012-301 Draft Administrative Documents
ML12214A258
Person / Time
Site: Farley  Southern Nuclear icon.png
Issue date: 07/23/2012
From:
NRC/RGN-II
To:
Southern Nuclear Operating Co
Shared Package
ML12214A278 List:
References
Download: ML12214A258 (35)
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ES-401 PWR Examination Outline Form ES-401-2 Facility: FARLEY Date of Exam: JUNE 2012 yPohts 1

ROKJACapo SRO-Only_Points Tier Group T G*

KIKKK KKAAAAG A2 Total 112 3 4 5 6 1 2 3 4

  • Total
1. 1 3 3 3 3 3 3 18 3 3 6 Emergency &

Abnormal 2 j_ j... . N/A .1 i... N/A 2 9 2 2 4 Evolutions Tier Totals 4 4 5 5 4 5 27 5 5 10 1 33332322322 28 3 2 5 2.

Plant 2 11111101111 10 1 1 1 3 Systems Tier Totals 4 4 4 4 3 4 2 3 4 3 3 38 5 3 8

3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 2 2 3 3 1 2 2 2 Note: 1. Ensure that at least two topics from every applicable K/A 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 K/A 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 K/A 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 K/As 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 K/A categories.

7* The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A 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 K/As.

8. On the following pages, enter the K/A 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 K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401 -3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.

ES-401 2 Form ES-401-2 ES-401 PWR Examination OutHne Form ES-401-2 Emergency and Abnorrna! Plant Evolutions Tier 1/Group 1

- 0 / SRO)

E/APE#/Name/SafetyFunction K K K A A G K/ATopic(s) IR #

12312 000007 (BW/E02&E1 0; CE/E02) Reactor Trip Stabilization Recovery / 1 7 1. 0 4, 000008 Pressurizer Vapor Space R 0 0 g A 1<3. 04 -

Accident! 3 000009 Small Break LOCA /3 R 009 EE 1< 03 000011 Large Break LOCA/3 R O( EA 000015/17 RCP Malfunctions /4 000022 Loss of Rx Coolant Makeup / 2 R 0 2 G 4. 34 4..

000025 Loss of RHR System / 4 R

0 S A K 3, 0_ a3 000026 Loss of Component Cooling Water/8 000027 Pressurizer Pressure Control System Malfunction / 3 o2 AKi o ag 000029ATWS/1 R O EKJ..o3 000038 Steam Gen. Tube Rupture / 3 R U3 4, 4?

000040 (BW/E05; CE/E05 W/E12 Steam Line Rupture Excessive eat R 040 V)/& 12 A A , II 2 Transfer/4 000054 (CE/E06) Loss of Main 0 Feedwater/4 .

0 4- A ,

000055 Station Blackout / 6 R 0 A , 0 4,4 000056 Loss of Off-site Power / 6 000057 Loss of Vital AC Inst. Bus / 6 057 AA I .0 3.7 000058 Loss of DC Power! 6 A I DI 000062 Loss of Nuclear Svc Water.! 4 0( 2 AA 1 0 7 000065 LossoflnstrumentAir/8 R Qc5 A<3, o3 2 W/E04 LOCA Outside Containment/3 W1D4 1.. g 1

3 W/E1 1 Loss of Emergency Coolant Recirc. / 4 ransfercs of Secondary Heat Sink/4 T

0 II 37 000077 Generator Voltage and Electric Grid Disturbances / 6 K/A Category Totals:

  • I 3 3 Group Point Total; 14)6

ES-401 3 Form ES-401-2 ES-401 PWR Examination Outline ,... Form ES-401-2 Emergency and Abnormal Plant Evolutions Tier 1/Group 2! SRO)

E/APE # / Name / Safety Function K K K A A G K/A Topic(s) IR #

12312 000001 Continuous Rod Withdrawal / 1 000003 Dropped Control Rod / 1 R 003 AA I. O7 3,g Q 000005 Inoperable/Stuck Control Rod / 1 R 0 05 A 2..,

. 000024 Emergency Boration / I R O4 AA I 000028 Pressurizer Level Malfunction / 2 000032 Loss of Source Range NI/7 R V3 AAD 000033 Loss of Intermediate Range NI / 7 g 03 A K3 0 3.2-000036 (BW/A08) Fuel Handling Accident! 8 000037 Steam Generator Tube Leak / 3 000051 Loss of Condenser Vacuum / 4 000059 Accidental Liquid RadWaste Rel. / 9 000060 Accidental Gaseous Radwaste Rel. / 9 000061 ARM System Alarms / 7 g &( A K 1,01 000067 Plant Fire On-site/8 R D7 A ,4.

000068 (BW/A06) Control Room Evac. / 8 000069 (W/E14) Loss of CTMT Integrity / 5 000074 (W/E06&E07) mad. Core Cooling /4 000076 High Reactor Coolant Activity / 9 W/EO1 & E02 Rediagnosis & SI Termination / 3 W/E13 Steam Generator Over-pressure/4 vJ/-L3 W/E1 5 Containment Flooding / 5 W/E16 High Containment Radiation / 9 BW/A01 Plant Runback / 1 BW/A02&A03 Loss of NNI-X/Y / 7 BW/A04 Turbine Trip /4 BW/A05 Emergency Diesel Actuation / 6 BW/A07 Flooding / 8 BW/E03 Inadequate Subcooling Margin / 4 BW/E08; W/E03 LOCA Cooldown - Depress. / 4 BW/E09; CE/A13;E ElO Natural Circ. /4 P W/O9 BW/E13&E14 EOP Rules and Enclosures CE/All; W/E08 RCS Overcooling - PTS / 4 CE/Al 6 Excess RCS Leakage /2 CE/E09 Functional Recovery K/A Category Point Totals: t} I I 1 QI Group Point Total: K4

ES-401 4 Form ES-401-2 ES-401 PWR Examination Outli Form ES-401-2

=

Rant Svsterns..11er2/Grouo6)SRO)

System # / Name K K K K K K A A A A G K/A Topic(s) IR #

1234561234 4

003 Reactor Coolant Pump g 3 2. 23 004 Chemical and Volume p Control o4 AhiO 3.7 f<5.o II 005 Residual Heat Removal 0O A 006 Emergency Core Cooling 00 K* 1/

007 Pressurizer Relief/Quench Tank KSIOZ 3!j I 008 Component Cooling Water OL) A 3, o , K3,o 010 Pressurizer Pressure Control

0) A 4 03 012 Reactor Protection oi2 k) 6 Ob 013 Engineered Safety Features .-y / /

Actuation 1 34 O22ContainmentCooling 1<.0Q- 3,0 025 Ice Condenser AJ/,t 026 Containment Spray 2

  • 0 f 4.2 039 Main and Reheat Steam K )4oZ 3I 059 Main Feedwater 0 9 A 3.

061 Auxiliary/Emergency Feedwater Ic 0 2 I 062 AC Electrical Distribution 0b2 Ak,1 03 3 02.

063 DC Electrical Distribution 1 t73 2, 2 44 42

/A 171 4 3%

064 Emergency Diesel Generator O Ni. 07 K4 0+

073 Process Radiation Monitoring R V73 076 Service Water V7 jD 3.1 II O78lnstrumentAir g o7 A4.; , K0 103 Containment R 103 AZ. 0 K/A Category Point Totals: .. I I Group Point Total:

ES-401 5 Form ES-401-2 ES-401 PWR Examination Outr Form ES-401-2 PlantSystems-Tier2/Group CRC SRO)

System #1 Name K K K K K K A A A A G K/A Topic(s) IR #

1234561234 001 Control Rod Drive 002 Reactor Coolant R OO- K 0 2S I 011 Pressurizer Level Control 01 / K I 3,1 014 Rod Position Indication /4 A, P4-015 Nuclear Instrumentation 016 Non-nuclear Instrumentation 017 In-core Temperature Monitor 917 KS 02 37 027 Containment Iodine Removal 028 Hydrogen Recombiner and Purge Control 029 ContainmentPurge O2 2239 3,9 033 Spent Fuel Pool Cooling 33 A 01 034 Fuel Handling Equipment R. 034 X h 01 25 035 Steam Generator 03 5 )<3 0/ 44-041 Steam Dump/Turbine Bypass Control 045 Main Turbine Generator 055 Condenser Air Removal 056 Qondensate 068 Liquid Radwaste 4, °4 3$

071 Waste Gas Disposal 072 Area Radiation Monitoring 075 Circulating Water 079 StationAir 086 Fire Protection K/A Category Point Totals: L I. .L I I 1.. 2 1 ui j Group Point Total:

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: rritt/ Date of Exam: JuA ti 2.

Category K/A# Topic SRO-Only 2.1.1 JCt Oj tQ t

1 5 3.g 2.1.4 w-So&.. 3 1.

Conduct 2.1.

of Operations 2.1.

2.1.

2.1.

Subtotal 2.2.4O ppLj %cs 4k.

t 4

sr 3.+

2.2.41

2. 2.2.

Equipment Control 2.2.

2.2.

2.2.

Subtotal 2 2.3. 1Z 5-JL iri 2 2.3.5 4/ih -b rt

3. 2.3. ti RJP Radiation Control 2.3.

2.3.

2.3.

Subtotal 3 2.4. 14- 6c-uvtiz ,

g

4. 2.4.43 Ccnio 3,2 Emergency 2.4. 2 pOic1 /opak- 4 2.

Procedures I )

Plan 2.4.

2.4.

2.4.

Subtotal 3 Tier3PointTotal 7

ES-401 2 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions-Tier 1/Group 1 (RO SRO E/APE #1 Name / Safety Function K K K A A G K/A Topic(s) IR #

12312 000007 (BW/E02&E10; CE/E02) Reactor Trip Stabilization Recovery / 1 000008 Pressurizer Vapor Space Accident / 3 000009 Small Break LOCA / 3 000011 Large Break LOCA/3 0 1 EAa_

000015/17 RCP Malfunctions / 4 000022 Loss of Rx Coolant Makeup / 2 000025 Loss of RHR System / 4 0O25 AA 2 000026 Loss of Component Cooling 000027 Pressurizer Pressure Control System Malfunction I 3 000029 ATWS/1 000038 Steam Gen. Tube Rupture / 3 000040 (BW/E05; CE/E05; W/E12)

Steam Line Rupture Excessive Heat Transfer / 4 000054 (CE/E06) Loss of Main Feedwater / 4 000055 Station Blackout / 6 000056 Loss of Off-site Power / 6 O& A I 7 000057 Loss of Vital AC Inst. Bus / 6 000058 Loss of DC Power / 6 000062 Loss of Nuclear Svc Water / 4 000065 Loss of Instrument Air / 8 W/E04 LOCA Outside Containment / 3 Loss of Emergency Coolant L 4,2 BW/E04; W/E05 Inadequate Heat Transfer- Loss of Secondary Heat Sink/4 000077 Generator Voltage and Electric Grid Disturbances / 6 K/A Category Totals: = = Group Point Total: 1861

ES-401 3 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions Tier 1/Group 2 (R

- /SRO E/APE # / Name / Safety Function K K K A A G K/A Topic(s) IR #

12312 000001 Continuous Rod Withdrawal / 1 000003 Dropped Control Rod/I 003 A 22 z2 1 4;r 000005 Inoperable/Stuck Control Rod / 1 000024 Emergency Boration/ 1 024 4 2, 4-. 4 42-000028 Pressurizer Level Malfunction / 2 000032 Loss of Source Range NI / 7 000033 Loss of Intermediate Range NI / 7 000036 (BW/A08) Fuel Handling Accident / 8 000037 Steam Generator Tube Leak / 3 000051 Loss of Condenser Vacuum / 4 000059 Accidental Liquid RadWaste Rel. / 9 000060 Accidental Gaseous Radwaste Rel. / 9 000061 ARM System Alarms / 7 000067 Plant Fire On-site / 8 o7 A/\ OX 000068 (BW/A06) Control Room Evac. / 8 000069 (W/E14) Loss of CTMT Integrity / 5 000074 (W/E06&E07) mad. Core Cooling / 4 000076 High Reactor Coolant Activity / 9 W/EO1 & E02 Rediagnosis & SI Termination / 3 WIE13 Steam Generator Over-pressure /4 W/E1 5 Containment Flooding / 5 W/E16 High Containment Radiation / 9 BW/A0I_Plant_Runback_/_1 BW/A02&A03 Loss of NNI-X1Y /7 BW/A04 Turbine Trip / 4 BW/A05 Emergency Diesel Actuation / 6 BW/A07 Flooding / 8 BW/E03 Inadequate Subcooling Margin / 4 BW/E08; W/E03 LOCA Cooldown - Depress. / 4 BW/E09; CE/Al 3; W/E0Q atural Circ. / 4 t) 1 I0 -.2.

BW/El3&E14 EOP Rules and Enclosures CE/All; W/E08 RCS Overcooling - PTS / 4 CE/Al 6 Excess RCS Leakage / 2 CE/E09 Functional Recovery K/A Category Point Totals: = = = = dZ i Group Point Total:

ES-401 4 Form ES-401-2 ES-401 PWR Examination Outline Form ES-4012 Plant Systems Tier 2/Group 1(RO

- i(

System #1 Name K K K K K K A A A A G K/A Topic(s) IR #

123456 1234 003 Reactor Coolant Pump 004 Chemical and Volume Control 005 Residual Heat Removal OOS r4 47 Oo6EmergencyCoreCooling OO& 413 007 Pressurizer Relief/Quench Tank 008 Component Cooling Water A . oI 3.6 010 Pressurizer Pressure Control O1 43 012 Reactor Protection 013 Engineered Safety Features Actuation 022 Containment Cooling 025 Ice Condenser 026 Containment Spray 2L AI,t/

039 Main and Reheat Steam 059 Main Feedwater 061 Auxiliary/Emergency Feedwater 062 AC Electrical Distribution 063 DC Electrical Distribution 064 Emergency Diesel Generator 073 Process Radiation Monitoring 076 Service Water 078 Instrument Air 103 Containment K/A Category Point Totals:

f 3

Group Point Total: 281

ES-401 5 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems Tier 2/Group 2 (RO / SO System # / Name K K K K K K A A A A G K/A Topic(s) IR #

1234561234 001 Control Rod Drive 002 Reactor Coolant 01 1 Pressurizer Level Control 014 Rod Position Indication 015 Nuclear Instrumentation 016 Non-nuclear Instrumentation 017 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 034- k4 01 035 Steam Generator 041 Steam Dump/Turbine Bypass Control 045 Main Turbine Generator 055 Condenser Air Removal 056 Condensate 068 Liquid Radwaste 071 Waste Gas Disposal 072 Area Radiation Monitoring O12 1.2 075 Circulating Water 079 Station Air 086 Fire Protection 3,3 K/A Category Point Totals:

I ] Group Point Total: 1 q1)

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: ALY Date of Exam: I.UVE 201Z Category KIA# Topic RO SRO-Only IR lR #

2.1. 7d4L udsi& CR.

2.1.

1.

Conduct 2.1.

of Operations 2.1.

2.1.

2.1.

Subtotal 2.2. 1 Pre prrck; raviti.i afr& 44-2.2. ie aa d!t d4 4

2. 2.2.

Equipment Control 2.2.

2.2.

2.2.

Subtotal 2.3./ 1 12%44 C4>?f It&44 4 2.3. 4 &I%Ø ti I

3. 2.3.

Radiation Control 2.3.

2.3.

2.3.

Subtotal 2.4. / 0

&d 4- /_jc

4. 2.4.3 U p Emergency 2.4.

Procedures I Plan 2.4.

2.4.

2.4.

Subtotal Tier 3 Point Total .7

ES-401, REV 9 T1G1 EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IA Kl K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO 007EG2.1 .20 -

Reactor Trip Stabilization - Recovery 4.6 4.6 Ability to execute procedure steps.

El

/1 008AK3.04 Pressurizer Vapor Space Accident / 3 4.2 4.6 RCP tripping requirements 009EK2.03 Smat Break LOCA/3 3 3.3 [] El [] [] [ S/Gs Oil EA2.08 Large Break LOCA /3 3.4 3.9 [] [] [ [ Conditions necessary for recovery when accident reaches stable phase 022AG2.4.34 Loss of Rx Coolant Makeup / 2 4.2 4.1 El El El El El El El El El El Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects 025AK3.02 Loss of RHR System / 4 3.3 3.7 Isolation of RHR low-pressure piping prior to pressure increase above specified level 027AK1.02 Pressurizer Pressure Control System 2.8 3.1 El J Expansion of liquids as temperature increases MaIf unction / 3 029EK1 .03 ATWS / 1 3.6 3.8 Effects of boron on reactivity 038EG2.4.47 Steam Gen. Tube Rupture / 3 4.2 4.2 El Ability to chagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.

O4OAA1.1l -

Steam line Rupture Excessive Heat 3.2 3.1 MFW system El El El El El El El El El El Transfer / 4 054AA2.03 Loss of Main Feedwater / 4 4.1 4.2 El El El El El El El El El El Conditions and reasons for AFW pump startup Page 1 of 2 9/29/2011 8:10AM

ES-401, REV 9 T1G1 P R EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR Kl K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO 055EA2.02 Station Blackout / 6 4.4 4.6 RCS core cooling through natural circulation cooling to S/G cooling 057AA1 .01 Loss of Vital AC Inst. Bus / 6 3.7 3.7 fl .

LI Manual inverter swapping O5BAK1.0l Loss of DC Power! 6 2.8 3.1 [] fl [] Battery charger equipment and instrumentation 062AA1 .07 Loss of Nuclear Svc Water! 4 2.9 3 Flow rates to the components and systems that are serviced by the SWS; interactions among the components 065AK3.03 Loss of Instrument Air /8 2.9 3.4 El fl j Knowing effects on plant operation of isolating certain equipment from instrument air WEO4EK2.2 LOCA Outside Containment! 3 3.8 4.0 fl fl fl Facilitys 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 facility.

WEO5EK2.l Inadequate Heat Transfer Loss of 3.7 3.9 Components and functions of control and safety systems, Secondary Heat Sink /4 including instrumentation, signals, interlocks, failure modes and automatic and manual features.

Page 2 of 2 9/29/2011 8:10AM

ES-401, REV 9 T1G2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME/SAFETY FUNCTION: IR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO OO3AA1 .07 Dropped Control Rod / 1 3.8 3.8 El El In-core and ex-core instrumentation 005AK2.02 Inoperable/Stuck Control Rod / 1 2.5 2.6 El J Breakers, relays, disconnects and control room switches 024AA1 .06 Emergency Boration I 1 3.2 3.1 El BWST temperature 032AA2.05 Loss of Source Range NI /7 2.9 3.2 [] [ El El El El El Nature of abnormality, from rapid survey of control room data 033AK3.01 Loss of Intermediate Range NI /7 3.2 3.6 El El El El El El El El El El Termination of startup following loss of intermediate-range instrumentation 061 AKi .01 ARM System Alarms/ 7 2.5 2.9 El El El El El El El El El El Detector limitations 067AG2.4.8 Plant Fire On-site / 8 3.8 4.5 El El El El El El El El El El Knowledge of how abnormal operating procedures are used in conjunction with EOPs.

weO9EG2.4.20 Natural Circ. / 4 3.8 4.3 El El El El El El El El El El Knowledge of operational implications of EOP warnings, cautions and notes.

WE13EK3.4 Steam Generator Over-pressure/4 3.1 3.3 El El El El El El El El El El RO or SRO function within the control room team as appropriate to the assigned position, in such a way that procedures are adhered to and the limitations in the facilities license and amendments are not violated.

Page 1 of 1 9/29/2011 8:10AM

ES-401, REV 9 T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME/SAFETY FUNCTION: IR Kl K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO 003G2.l .23 Reactor Coolant Pump 4.3 4.4 Ability to perform specific system and integrated plant procedures during all modes of plant operation.

003K6.02 Reactor Coolant Pump 2.7 3.1 RCP seals and seal water supply 004A1.10 Chemical and Volume Control 3.7 3.9 Reactor power 005A2.02 Residual Heat Removal 3.5 3.7 Pressure transient protection during cold shutdown 005K5.02 Residual Heat Removal 3.4 3.5 Need for adequate subcooling 006K4.l 1 Emergency Core Cooling 3.9 4.2 Reset of SS 007K5.02 Pressurizer Relief/Quench Tank 3.1 3.4 Method of forming a steam bubble in the PZR 008A3.06 Component Cooling Water 2.5 2.5 LI LI LI LI LI LI LI LI LI Typical CCW pump operating conditions, including vibra tion and sound levels and motor current 008K3.03 ComponentCooling Water 4.1 4.2 LI LI LI LI LI LI LI LI LI LI RCP 010A4.03 Pressurizer Pressure Control 4.0 3.8 LI LI LI LI LI LI LI LI LI LI PORV and block valves 012K1.06 ReactorProtection 3.1 3.1 LI LI LI LI LI LI LI LI LI LI T/G Page 1 of 3 9/29/2011 8:10AM

ES-401, REV 9 T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR Kl K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SAC 012K6.06 Reactor Protection 2.7 2.8 Sensors and detectors ESFAS/safeguards equipment control -

013K2.01 Engineered Safety Features Actuation 3.6 3.8 022K3.02 Containment Cooling 3.0 3.3 El Containment instrumentation readings 026K1 .01 Containment Spray 4.2 4.2 ECCS 039K4.02 Main and Reheat Steam 3.1 3.2 Utilization of T-ave. program control when steam dumping through atmospheric relief/dump valves, including T-ave. limits 059A3.02 Main Feedwater 2.9 3.1 Programmed levels of the SIG 061 K6.02 Auxiliary/Emergency Feedwater 2.6 2.7 Pumps 062A1 .03 AC Electrical Distribution 2.5 2.8 Effect on instrumentation and controls of switching power supplies 062K3.02 AC Electrical Distribution 4.1 4.4 J fl ED/G 063G2.2.44 DC Electrical Distribution 4.2 4.4 D Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions 064K1 .04 Emergency Diesel Generator 3.6 3.9 LI LI El LI LI LI LI LI LI DC distribution system Page 2 of 3 9/29/2011 8:10AM

ES-401, REV 9 T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IA Kl K2 K3 K4 K5 KG Al A2 A3 A4 G TOPIC:

RO SRO 064K4.04 Emergency Diesel Generator 3.1 3.7 Overload ratings 073A2.01 Process Radiation Monitoring 2.5 2.9 Erratic or failed power supply 076K2.08 Service Water 3.1 3.3 El [1 El E] El ESF-actuated MOVs 078A4.0l Instrument Air 3.1 3.1 El El El El El El El El El El Pressure gauges 078K2.01 Instrument Air 2.7 2.9 El El El El El El El El El El Instrument air compressor 103A3.01 Containment 3.9 4.2 El El El El El El El El El El Containment isolation Page3of3 9/29/2011 8:10AM

(

ES-401, REV 9 TG2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR Ki K2K3K4K5K6A1 A2 A3 A4 G TOPIC:

RO SRQ 002K6.06 Reactor Coolant 2.5 2.8 LI El LI LI LI LI LI LI LI LI Sensors and Detectors 011K2.01 Pressurizer Level Control 3.1 3.2 Li El LI LI LI LI LI LI LI LI Charging pumps 014A2.04 Rod Position Indication 3.4 3.9 LI LI LI LI LI LI LI ] LI LI LI Misaligned rod 017K5.02 In-core Temperature Monitor 3.7 4.0 LI LI LI LI LI LI LI LI LI LI Saturation and subcooling of water 029G2.2.39 Containment Purge 3.9 4.5 LI LI LI fl LI LI LI LI LI LI Knowledge of less than one hour technical specification action statements for systems.

033A3.01 Spent Fuel Pool Cooling 2.5 2.7 LI LI LI LI LI LI LI LI J LI LI Temperature control valves 034K1.O1 Fuel Handling Equipment 2.5 3.2 1 LI LI LI LI LI LI LI LI LI LI RCS 035K3.01 Steam Generator 4.4 4.6 LI LI LI LI LI LI LI LI LI LI RCS 068A4.04 Liquid Radwaste 3.8 3.7 LI LI LI LI LI LI LI LI LI LI Automatic isolation 079K4.01 Station Air 2.9 3.2 LI LI LI LI LI LI LI LI LI LI Cross-connect with AS Page 1 of 1 9/29/2011 8:10AM

ES-401, REV 9 T3 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO G2.l.l Conductof operations 3.8 4.2 LI LI Knowledge of conduct of operations requirements.

G2l.4 Conductof operations 3.3 3.8 LI LI LI LI LI LI LI Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, no-soIo operation, maintenance of active license statur, 10CFR55 etc.

G2.2.40 Equipment Control 3.4 4.7 LI LI LI LI LI LI LI 1 Ability to apply technical specifications for a system.

G2.2.41 Equipment Control 3.5 3.9 LI LI LI l Ability to obtain and interpret station electrical and mechanical drawings G2.3.12 Radiation Control 3.2 3.7 LI LI LI LI LI LI LI LI LI LI 1 Knowledge of radiological safety principles pertaining to licensed operator duties G2.3.5 Radiation Control 2.9 2.9 LI LI LI LI LI LI LI LI LI LI Ability to use radiation monitoring systems G2.3.7 Radiation Control 3.5 3.6 LI LI LI LI LI LI LI LI LI LI Ability to comply with radiation work permit requirements during normal or abnormal conditions G2.4.14 Emergency Procedures/Plans 3.8 4.5 Knowledge of general guidelines for EOP usage.

LI LI LI LI LI LI LI LI LI LI G2.4.43 Emergency Procedures/Plans 3.2 3.8 LI LI LI LI LI LI LI LI LI LI Knowledge of emergency communications systems and techniques.

G2.4:50 Emergency Procedures/Plans 4.2 4.0 LILILILILILILILILILI Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.

Page 1 of 1 9/29/2011 8:10AM

ES-401, REV 9 (y1G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR Kl K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO 011 EA2.02 Large Break LOCA /3 3.3 3.7 Consequences to RHR of not resetting safety injection 025AA2.05 Loss of RHR System / 4 3.1 3.5 ELE1L1r Limitations on LPI flow and temperature rates of change 026AG2.2.25 Loss of Component Cooling Water /8 3.2 4.2 DflLL1flDD Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

056AG2.1 .7 Loss of Off-site Power / 6 4.4 4.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior and instrument interpretation.

weO4EG2.4.9 LOCA Outside Containment! 3 3.8 4.2 ri LI LI Fii Knowledge of low power I shutdown implications in accident (e.g. LOCA or loss of RHR) mitigation strategies.

WE1 1 EA2.l Loss of Emergency Coolant Recirc. / 4 3.4 4.2 LI LI LI LI LI LI LI LI LI LI Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

Page 1 of 1 9/29/2011 8:10AM

ES-401, REV 9 ()1G2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: H9 K1K2K3K4K5K6A1A2A3A4G TOPIC:

RO SRO 003AG2.2.22 Dropped Control Rod / 1 4.0 4.7 Knowledge of limiting conditions for operations and safety limits.

024AG2.4.46 Emergency Boration / 1 4.2 4.2 Ability to verify that the alarms are consistent with the plant conditions.

067AA2.08 Plant Fire On-site! 8 2.9 3.6 Limits of affected area WE1 OEA2.2 Natural Circ. With Seam Void! 4 3.4 3.9 D Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments.

Page 1 of 1 9/29/2011 8:10AM

ES-401, REV 9 T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: HR Kl K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO 005G2.4.30 Residual Heat Removal 2.7 4.1 Knowledge of events related to system operations/status that must be reported to internal orginizations or outside agencies.

006A2.02 Emergency Core Cooling 3.9 4.3 Loss of flow path 008A2.0l Component Cooling Water 3.3 3.6 E H:  : Loss of CCW pump 010G2.4.45 Pressurizer Pressure Control 4.1 4.3 Ability to prioritize and interpret the significance of each annunciator or alarm.

026A2.01 Containment Spray 2.7 3.0 H E Reflux boiHng pressure spike when first going on recirculation Page 1 of 1 9/29/2011 8:10AM

ES-401, REV 9 ()r2G2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: Kl K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO 034K4.01 Fuel Handling Equipment 2.6 3.4 Fuel protection from binding and dropping 072G2.2.36 Area Radiation Monitoring 3.1 4.2 LI LI LI fl LI LI LI fl LI LI Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions of operations 086A2.02 Fire Protection 3.0 3.3 LI LI LI LI LI LI LI LI LI LI Low FPS header pressure Pagelof 1 9/29/2011 8:10AM

ES-401, REV 9 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IA Kl K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO G2.1 .8 Conduct of operations 3.44.1 Ability to coordinate personnel activities outside the control room.

G2.2.1 Equipment Control 4.5 4.4 Ability to perform pre-startup procedures for the facility, including operating those controls associated with plant equipment that could affect reactivity.

G2.2.15 Equipment Control Ability to determine the expected plant configuration using design and configuration control documentaion G2.3.1 1 Radiation Control 3.8 4.3 Ability to control radiation releases G2.3.4 Radiation Control 3.2 3.7 Knowledge of radiation exposure limits under normal and emergency conditions G2.4.18 Emergency Procedures/Plans 3.3 4.0 Knowledge of the specific bases for EOP5.

G2.4.32 Emergency Procedures/Plans 3.6 4.0 Knowledge of operator response to loss of all annunciators.

Page 1 of 1 9/29/2011 8:10AM

ES-301-i Administrative Topics Outline Draft Submittal Facility: Farley Nuclear Plant Date of Examination: June 18, 2012 Examination Level: RO X SRO X Operating Test Number: FA2012-301 Administrative Topic Type Describe activity to be performed (see Note) Code*

a. A.1 .a

Title:

Determine maximum RHR flow and time to Conduct of Operations N, R saturation for a loss of RHR event while at midloop.

(RO and SRO)

G2.1.25 39I4.2

b. A.1 .b

Title:

Determine if shift manning requirements are met per Conduct of Operations D, P SOP-0.0, General Instruction to Operations Personnel.

(RO and SRO)

G2.1 .5 2.9*1 39

c. A.2

Title:

Perform STP-9.0, RCS Leakage Test, and evaluate Equipment Control M, R Acceptance Criteria.

(RO only)

G2.2.12 3.7/4.1

c. A.2

Title:

Perform STP-9.0, RCS Leakage Test, evaluate Equipment Control M, R Acceptance Criteria and evaluate TS.

(SRO only)

G2.2.12 3.7/4.1

d. A.3

Title:

Calculate dose from local equipment operation and Radiation Control M, R determine if it can be performed by a Relocation per EIP (RO only) 14, Personnel Movement, Relocation, Re-entry and Site Evacuation.

G2.3.14 3.4/3.8

d. A.3

Title:

Calculate dose from local equipment operation and Radiation Control M, R determine if a volunteer is required to perform the (SRO only) emergency actions.

G23.14 3.4/38

e. A.4

Title:

Classify an Emergency Event.

Emergency Procedures/Plan N, R (SRO only) G2.4.41 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 ( 3 for ROs; 4 for SROs & RO retakes) 1/1 (N)ew or (M)odified from bank ( 1) 3/4 (P)revious 2 exams ( 1; randomly selected) 0/0

ES-301-1 Administrative Topics Outline Explanation sheet Draft Submittal Facility: Farley Nuclear Plant Date of Examination: June 18, 2012 Examination Level: RO X SRO X Operating Test Number: FA2012-301 Administrative Topic Type Describe activity to be performed (see Note) Code*

a. A.1 .a Determine maximum RHR flow and time to saturation Conduct of Operations for a loss of RHR event while at midloop.

N R AOP-12, Residual Heat Removal System Malfunction entry:

1) determine maximum RHR flow for current RCS level of 1228.5 to prevent cavitation and
2) If both RHR pumps are tripped, determine time to saturation with RCS level at 1228.5, after refueling with 1/3 new core, 350 hours0.00405 days <br />0.0972 hours <br />5.787037e-4 weeks <br />1.33175e-4 months <br /> after shutdown. Candidate will have to make determinations from tables and graphs in AOP-12.

G2.1 .25 Ability to interpret reference materials, such as graphs, curves, tables, etc.

3.9/4.2

b. A.1 .b Determine if shift manning requirements are met per Conduct of Operations Scip-o.o, General Instruction to Operations Personnel.

(RO and SRO)

D, R Due to a family emergency, an OPS shift person has to go home. The candidate will determine if adequate personnel are available to staff the shift for both Unit 1 and Unit 2, from the personnel that remain.

G2.1 .5 Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.

2.9*! 39 C. A.2 Perform STP..9.0, RCS Leakage Test, and evaluate Equipment Control Acceptance Criteria.

(RO only)

M,R Candidate will be required to perform STP-9.0 manually (with the plant computer unavailable). He will be given a table of parameters for the start of the leak rate, and then for parameters two hours later. Candidate must calculate the leakrate from the parameters given. The calculation will show 1.1 gpm unidentified leakage.

Candidate has to evaluate the test and determine if Acceptance Criteria is met (Acceptance Criteria is not met).

G2.2.12 Knowledge of surveillance procedures.

3.7! 4.1

C. A.2 Perform STP-9.O, RCS Leakage Test, evaluate Equipment Control Acceptance Criteria and evaluate TS.

(SRO only)

M,R Candidate will be required to perform STP-9.O manually (with the plant computer unavailable). He will be given a table of parameters for the start of the leak rate, and then for parameters two hours later. Candidate must calculate the leakrate from the parameters given. The calculation will show 1.1 gpm unidentified leakage.

Candidate has to evaluate the test and determine if Acceptance Criteria is met (Acceptance Criteria is not met). The second part of the JPM is to evaluate Tech Specs for the RCS Leakage.

G2.2.12 Knowledge of surveillance procedures.

3.7/ 4.1

d. A.3 Calculate dose from local equipment operation and Radiation Control determine if it can be performed by a Relocation per (RO only) M, R EIP-14, Personnel Movement, Relocation, Re-entry and Site Evacuation.

Candidate will be required to calculate dose received from emergency field actions associated with isolating a failed open atmospheric relief valve with a SGTR in progress. Determine if the action can be performed without exceeding dose limits per EIP-14.O.

G2.3.14 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

3.4/ 3.8

d. A.3 Calculate dose from local equipment operation and Radiation Control determine if a volunteer is required to perform the (SRO only) M,R emergency actions.

Calculate dose from emergency field actions associated with isolating a failed open atmospheric relief valve with a SGTR in progress and failed fuel.

SRO part- Determine if the individual performing the task is required to be a willing volunteer, per EIP-14.O.

G2.3.14 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

3.4/3.8

e. A.4 N, R Classify an Emergency Event.

Emergency Procedures/Plan (SRO only) Classify an event from given information and complete Checklist 1, Classification Determination (page 1 of 1) within 15 minutes.

G2.4.41 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 ( 3 for ROs; 4 for SROs & RO retakes) i/l (N)ew or (M)odified from bank ( 1) 3/4 (P)revious 2 exams ( 1; randomly selected) 0/0 Notes: There will be a total of 4 SRO events on the JPM exams.

One on the simulator, JPM F. to evaluate Tech Specs for a failed NI.

3 from this page; A.2 will evaluate Tech specs for the leak in progress.

A.3 will have the SRO determine if the task during the emergency can be accomplished if the person does not volunteer, which will have the SRO calculate the dose for the event in progress and then determine from EIP-14 if the person is required to sign as a volunteer or not.

A.4 will be the classification of an event. At Southern company plants, with the new NMPs that our SROs use to classify the event, the ED is required to classify the event. Then the SRO will get an extra person to fill out the Emergency Notification form (NMP-EP-1 11 Figure 1) on the computer and then have the SRO review it and then the extra person will send it to the states.

The ED will return to oversight of the facility while the form is being filled out and then review and approve it prior to sending it out. On Checklist 1 Classification Determination, all the information that is required to be filled out on the Emergency Notification form is there so the extra will know what to put on the form.

ES-301-2 Control Room/In-Plant Systems Outline Draft Submittal Facility: Farley Nuclear Plant Date of Examination: June 18, 2012 Exam Level: RO SRO-l Operating Test No.: FA2012-301 Control Room Systems@ (8 for RO); (7 for SRO-l)

System / JPM Title Type Code* Safety Function

a. CRO-033E Perform Corrective Actions In Response To a Malfunction of the Rod Control System for a Continuous Rod A D S Withdrawal.

001A2.11 3.4/3.7 APEOO1AA2.05 4.4/4.6 EPEOO7EA2.02 4.3/4.6 2

b. CR0-NEW 1- Re-establish HHSI flow due to a SGTR per N, L, 5 Attachment 3 of ESP-1 .1, SI Termination.

006A4.01 4.1/3.9 006A4.02 4.0/3.8 006A4.07 4.4/4.4

c. CR0-NEW 2- Perform lOAs of AOP-4.0 while at 25% power A, N, S 4P for a RCP trip.

003A1.07 3.4/3.4 003A2.02 3.7/3.9 APEO15/O17AA1 .08 3.0/2.9 APEO15/O17AA1 .09 3.1/3.2

d. CR0-NEW 3- Start Containment spray system during the A, N, L, 5 5 performance of EEP-0 during a LB LOCA.

026A2.03 4.1/4.4 026A2.04 3.9/4.2

e. CRO-258 modified Start the 1 B DG and energize the 1 G

- A, M, S 6 4160V Bus.

064A3.06 3.3/3.4 064A4.06 3.9/3.9 EPEO55EA1.06 4.1/4.5 EPEO55EA2.03 3.9/4.7

f. CR0-i 27A - Perform actions of AOP-i00 for a Nl-42 failure. D, 5 7 015A2.01 3.5/3.9 015A3.02 3.7/3.9 0i5A4.03 3.8/3.9 015G2.2.40 SRO 4.7
g. CRO-091A Place the Standby CCW Hx on service on the A D, 5 8 Train.

008A4.01 3.3/3.1 008A4.09 3.0/2.9

h. CRO-350A Operate Post LOCA Hydrogen Recombiner

- D, E, C 5 028A2.01 3.4/3.6 (RO only) 028A4.01 4.0/4.0 In-Plant Systems (3 for RO); (3 for SRO-l)

i. SO-430B and C (combined) Mechanically align the 1 C SW

- D 4S pump to the B Train and select it for B Train autostart.

076A2.01 3.5/3.7 076A2.02 2.7/3.1

j. SO-351A Start the 1-2A DG from the DLCP in Mode 4.

- D, E 6 064A2.02 2.7/2.9 APEO56AA2.21 3.6/3.8 EPEO55EA1 .02 4.3/4.4

k. SO-372B modified - Perform a #2 WMT release. A, M, E, R 9 068A2.04 3.3/3.3 068A3.02 3.6/3.6

@ All RO and SRO-l 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 AC I SAC-I ACTUAL (A)lternate path 4-6 I 4-6 (5/5)

(C)ontrol room (1/0)

(D)irect from bank 9/ 8 (6/5)

(E)mergency or abnormal in-plant 1 / 1 (2/2)

(EN)gineered safety feature - I -

(L)ow-Power I Shutdown 1 / 1 (2/2)

(N)ew or (M)odified from bank including 1 (A) 2/ 2 (5/5)

(P)revious 2 exams 3 / 3 (randomly selected) 0 (R)CA 1l1 (1/1)

(S)imulator (7/7)

ES-301-2 Control Room/In-Plant Systems Outline Explanation sheet Draft Submittal Facility: Farley Nuclear Plant Date of Examination: June 18, 2012 Exam Level: RO SRO-l Operating Test No.: FA2012-301 Control Room Systems@ (8 for RO); (7 for SRO-l)

System / JPM Title Type Code* Safety Function

a. CRO-033E Perform Corrective Actions In Response To a Malfunction of the Rod Control System for a Continuous Rod A D S 1

Withdrawal.

Reactor power is at apprx 72% with a significant oil leak on the 1 B SGFP, Candidate is to ramp down at 4 MW/mm to remove the SGFP from service. A malfunction will occur in the Rod Control System such that rods will start stepping out continuously (regardless of Auto/Manual position). AOP-19 actions will require a Reactor trip due the malfunction of rod control with a continuous rod withdrawal. The Rx trip handswitches will not work and the Rx will be tripped by opening CRDM MG set breakers. A manual turbine trip will be required due to no P-4 signal. An SI may occur due to the Main Turbine remaining online when the Reactor is tripped. Terminate JPM when candidate verifies IOAs.

001A2.11 3.4/3.7 APEOO1AA2.05 4.4/4.6 EPEOO7EA2.02 4.3/4.6

b. CR0-NEW 1 Re-establish HHSI flow due to a SGTR per

- N, L, S 2 Attachment 3 of ESP-1 .1, SI Termination.

Even though A SG Tube Rupture occurs while in ESP 1 1, SI Termination this meets KA procedure Candidate must perform actions of the foldout page we02eal 1 in (Attachment 3) to Re establish HHSI flow SF3, I believe it 006A4 01 4 1/3 9 safety function 2 better fits SF2 006A4.02 4.0/3.8 safety function 2 from inventory 006A4.07 4.4/4.4 safety function 2 control W/EO2EA1 1 4 0/3 9 safety function 3

c. CR0-NEW 2 Perform lOAs of AOP-4.0 while at 25% power A, N, S 4P for a RCP trip.

Reactor power is apprx 25%. 1 B RCP will trip when an adjustment is made to the Main FRV during the transfer, requiring AOP-4.0 entry (Rx trip is not required due to one RCP trip when <

30% power). AOP-4.0 actions are required (to close the Pressurizer spray valve and isolate FW flbw) when the RCP trips.

The 1 B Main FRV will stick open in its current position.. The candidate will have to perform RNO actions to close M0V3232B to isolate feed flow to 1 B SG to prevent overfill. In addition, AOP 4.0 will require Reactor trip due to RCS temp <541°F. If candidates response is inadequate, the SG level will go high, and the Turbine will trip. The SGFPs should trip, but that trip will be blocked so that if the candidate does not control FW flow, a SG overfill will occur. The automatic Reactor trip will be blocked, but manual will work by operating the Rx Trip HS. SGFP(s) will not automatically trip. (Closing M0V3232B or tripping SGFPs is critical to prevent overfilling the SG.) Terminate when lOAs of EEP-0 are complete.

Discussion: The simulator exam has a RCP high vibrations at >30% Rx power and requires a Rx trip. The Audit exam has a mode 3 JPM that has the student secure the RCP and then implement AOP-4.O which has different actions from this JPM.

003A1 .07 3.4/3.4 003A2.02 3.7/3.9 APEO1 5/01 7AA1 .08 3.0/2.9 APEO15/O17AA1 .09 3.1/3.2

d. CR0-NEW 3 Start Containment spray system during the

- A, N, L, S 5 performance of EEP-0 during a LB LOCA.

A LOCA has occurred and the candidate is at step 6 of EEP-0.

Containment pressure is 30 psi, requiring RNO actions. Phase B actuates, CS does not, perform Attachment 5 is required. Both trains of CS will have to be started manually and discharge MOVs opened, M0V8820B, the discharge on 1 B CS pump will not open. Candidate will have to reset the CS actuation signal and S/D the 1 B CS pump.

Discussion: This is alternate path cluetotheft4hat during the performance of the RNO coIumnfaiIures occur that require alternate decisions within the RNO column.

026A2.03 4.1/4.4 026A2.04 3.9/4.2

e. CRO-258 modified Start the lB DG and energize the 1 G

- A, M, 5 6 4160V Bus.

There has been a Loss of all A/C power. ECP-0.0, Loss of All A/C Power has been entered. 2C DG is Tagged Out. 1 -2A DG has tripped due to an electrical issue. 1C DG is supplying U-2. The candidate is to perform ECP-0.0 starting at step 5.2. He will start the 1 B DG (the only one available to U-i). When the DG is started, the output breaker fails to close automatically and actions are required per the RNO column of ECP-0.0 to close the output breaker. Once the output breaker is closed from the EPB, the sequencer will run and start loads. ECP-0.0 will be exited at step 5.8.

064A3.06 3.3/3.4 064A4.06 3.9/3.9 EPE055EA1 .06 4.1/4.5 EPE055EA2.03 3.9/4.7

f. CRO-127A Perform actions of AOP-100 for a Nl-42 failure.

D, S 7 UOP-1 .2, 4% power and ramping up, N-42 has just failed.

Perform AOP-i00 section 1.12. Terminate JPM at step i2of section 1.12 of AOP-100 when the candidate states that he will submit a CR at step 12. The simulator does not provide the capability of tripping bistables for OT deltaT at step 13. Bank JPM is CR0-i 27A.

Allow the candidate to complete actions in the Simulator, and then for SRO only have the candidate determine if a Mode change can be performed. If so, what actions are required to ramp up to 12%

power?

015A2.0i 3.5/3.9 015A3.02 3.7/3.9 015A4.03 3.8/3.9 015G2.2.40 SRO 47

g. CRO-091A Place the Standby CCW Hx on service on the A D, 5 8 Train.

100% power, swap from the 1 B CCW Hx o/s to the 1 C Hx 0/s.

Requires unisolating and throttling open flow to one Hx while throttling closed and isolating the other Hx. The 3009 FCVs are reverse acting valves (100% demand is full closed), requiring increased attention to detail. This action is performed per SOP 23.0 section 4.4.

008A4.01 3.3/3.1 008A4.09 3.0/2.9

h. (RO only) CRO-350A Operate Post LOCA Hydrogen

- D, F, C 5 Recombiner.

A LOCA has occurred with an increase in Containment H2 concentration. Step 6.5 RNO of EEP-1 directs performance of Attachment 3 to place the Post LOCA H2 Recombiners in service.

The candidate is directed to place 1A H2 Recombiner in service.

The task requires manipulation of control switches on the Recombiner panel and use of Figure 1 of Attachment 3 to determine the correct setting. Bank JPM is CRO-350A.

This task will be simulated in the actual Control Room, this eQuipment is not available in the Simulator 028A2.01 3.4/3.6 028A4.01 4.0/4.0 In-Plant Systems@ (3 for RO); (3 for S no-I)

SO-430 B and C (combined) Mechanically align the 1C SW

- D 4S pump to the B Train and select it for B Train autostart.

Unit 1 has experienced an LOSP. The 1C SW pump is aligned to the A Train and the 1 D SW pump has tripped. Two SW pumps are required to support operation of the 1 B DG. Actions have been completed to align the 1C SW pump to the B Train electrically. The candidate is to align the 1C SW pump to the B Train mechanically and select it to autostart for the 1 D SW pump.

This is per SOP-24.OD (a checklist in the back of SOP-24.0).

Candidate will start at step 2.0 and complete the rest of the checklist. Then Appendix 6 will have to be performed to align the SW pump for autostart.

Discussion: this JPM will be performed at the SWIS which is in an outside structure away from the main part of the plant. This JPM has never been done for any HLT class at FNP due to the location of the JPM. Using station keeping methods this should not present too much of a challenge to accomplish.

076A2.01 3.5/3.7 076A2.02 2.7/3.1

j. SO-351A -Start the 1-2A DG from the DLCP in Mode 4. D, E 6 Candidate will perform FNP-0-SOP-38.1 to start the 1-2A DG from the Local panel in Mode 4 (section 4.10 performed, not 4.11). Voltage and Frequency will have to be adjusted to bring it in spec. Once DC is running with voltage and frequency in spec, control will be shifted to the Control Room EPB operator Discussion: This JPM will be performed in the Diesel Building and is relevant to a Loss of ALL AC event similar to Fukashima.

064A2.02 2.7/2.9 APEO56AA2.21 3.6/3.8 EPEO55EA1 .02 4.3/4.4

k. SO-372B modified - Perform a #2 WMT release. A, M, E, R 9 Start at step 4.4.9 of SOP-50.1, Appendix 2, to open RCV-18. The candidate will initiate the release and then a Water Discharge Line Hi-Rad alarm will come in on the Liquid Waste Panel.

Candidate must perform actions of ARP-13.2, annunciator 14, to stop the release per the ARP. Terminate the JPM when the release is stopped. Modified from SO-372B to make it an alternate path.

The release can be stopped by one of the following methods:

1) Closing RCV-18
2) Closing Vi 14
3) Closing V1O8A
4) Stopping the WMT pump Discussion: This JPM has never been performed in this manner at FNP.

068A2.04 3.3/3.3 068A3.02 3.6/3.6

@ All RO and SRO-l 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-l (A)lternate path 4-6 / 4-6 (5/5)

(C)ontrol room (1/0)

(D)irect from bank 9/ 8 (6/5)

(E)mergency or abnormal in-plant 1 / 1 (2/2)

(EN)gineered safety feature - / -

(L)ow-Power/Shutdown 1 / 1 (2/2)

(N)ew or (M)odified from bank including 1 (A) 2/ 2 (5/5)

(P)revious 2 exams 3 / 3 (randomly selected) 0 (R)CA i/i (1/1)

(S)imulator (7/7)

DG starts on the simulator exam -

2C DG started in one scenario, small blackout DG that is started differently than Large DGs.

Difference is the flowpath and method for starting and aligning will be different.

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