ML13274A443

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Initial Exam 2012-302 Draft Administrative Documents
ML13274A443
Person / Time
Site: Sequoyah  Tennessee Valley Authority icon.png
Issue date: 09/27/2013
From:
NRC/RGN-II
To:
Tennessee Valley Authority
References
Download: ML13274A443 (25)
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Text

ES-401, Rev. 9 PWR Examination Outline Form ES-401-2 Facility: Sequoyah Date of Exam: December 2012 RO K/A Category Points SRO-Only Points Tier Group K K K K K K A A A A G A2 G* Total 1234561234

  • Total
1. 1 33 318 3 3 6 Emergency &

Abnormal ant 2 1 2 T2 2 1 1 9 2 2 4 N/A N/A Tier Totals 4 5 5 5 4 4 27 5 5 10 1 22323323323 28 2 3 5 2.

Plant 2 11111111101 10 1 2 3 Systems TierTotals 33434434424 38 3 5 8

3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 2 3 2 3 2 2 1 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 that are not included on the outline should be added.

Refer to ES-401, Attachment 2, 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.
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. 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, Rev. 9 2 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions Tier 1/Group I (RO / SRO)

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

12312 000007 (BW/E02&E1 0; CE/E02) Reactor Trip X 2.1.30 Ability to locate and operate components, 4.0 76

- Stabilization Recovery / 1

- including local controls.

(CFR: 41.7 / 45.7) 000008 Pressurizer Vapor Space Accident / 3 X AKI. Knowledge of the operational implications 3.1 of the following concepts as they apply to a Pressurizer Vapor Space Accident:

(CFR 41.8 /41.10! 45.3)

AK1 .02 Change in leak rate with change in pressure 000009 Small Break LOCA / 3 X 2.1.23 Ability to perform specific system and 4.4 77 integrated plant procedures during all modes of plant operation.

(CFR: 41.10/43.5/45.2! 45.6) 000011 Large Break LOCA / 3 X EKI Knowledge of the operational implications 4.1 2 of the following concepts as they apply to the Large Break LOCA:

(CFR41.8/41.10/45.3)

EK1 .01 Natural circulation and cooling, including reflux boiling 00001 5/17 RCP Malfunctions! 4 X AA2. Ability to determine and interpret the 3.4 3 following as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow):

(CFR: 43.5/45.13)

AA2.1 1 When to jog RCPs during ICC 000022 Loss of Rx Coolant Makeup / 2 X AAI. Ability to operate and I or monitor the 3.4 4 following as they apply to the Loss of Reactor Coolant Makeup:

(CFR 41.7! 45.5! 45.6)

AA1 .01 CVCS letdown and charging 000025 Loss of RHR System /4 X AAI. Ability to operate and / or monitor the 3.2 5 following as they apply to the Loss of Residual Heat Removal System:

(CFR 41.7 / 45.5 / 45.6)

AA1.09 LPI pump switches, ammeter, discharge pressure gauge, flow meter, and indicators 000026 Loss of Component Cooling Water I 8 X AK3. Knowledge of the reasons for the following 4.0 6 responses as they apply to the Loss of Component Cooling Water:

(CFR 41.5,41.10 / 45.6 145.13)

AK3.03 Guidance actions contained in EOP for Loss of CCW 000027 Pressurizer Pressure Control System X AK2. Knowledge of the interrelations between 2.6 7 Malfunction / 3 the Pressurizer Pressure Control Malfunctions and the following:

(CFR 41.71 45.7)

AK2.03 Controllers and positioners

ES-401, Rev. 9 3 Form ES-401-2 000029 ATWS / 1 X 2.1.25 Ability to interpret reference materials, 3.9 8 such as graphs, curves, tables, etc.

(CFR: 41.10/43.5/45.12) 000038 Steam Gen. Tube Rupture! 3 X 2.2.12 Knowledge of surveillance procedures. 3.7 9 (CFR: 41.10/45.13) 000040 (BW/E05; CE/E05; W/E12) Steam X AK3. Knowledge of the reasons for the following 4.5 10 Line Rupture Excessive Heat Transfer / 4

- responses as they apply to the Steam Line Rupture:

(CFR 41.5,41.10145.6! 45.13)

AK3.04 Actions contained in EOPs for steam line rupture 000054 (CE/E06) Loss of Main Feedwater / 4 X AA2. Ability to determine and interpret the 3.4 11 following as they apply to the Loss of Main Feedwater (MFW):

(CFR: 43.5145.13)

AA2.07 Reactor trip first-out panel indicator 000055 Station Blackout/6 X EA2 Ability to determine or interpret the 3.9 12 following as they apply to a Station Blackout:

(CFR 43.5 / 45.13)

EA2.03 Actions necessary to restore power 000056 Loss of Off-site Power! 6 X AKI. Knowledge of the operational implications 3.1 13 of the following concepts as they apply to Loss of Offsite Power:

CFR 41.8 141.10! 45.3)

AK1 .04 Definition of saturation conditions, implication for the systems X AA2. Ability to determine and interpret the 4.1 78 following as they apply to the Loss of Offsite Power:

(CFR: 43.5145.13)

AA2.42 Occurrence of a reactor trip 000057 Loss of Vital AC Inst. Bus! 6 X AA2. Ability to determine and interpret the 4.3 79 following as they apply to the Loss of Vital AC Instrument Bus:

(CFR: 43.5145.13)

AA2.19 The plant automatic actions that will occur on the loss of a vital ac electrical instrument bus 000058 Loss of DC Power! 6 X 2.2.38 Knowledge of conditions and limitations 3.6 14 in the facility license.

(CFR:41.7/41.10!43.1 /45.13) 000062 Loss of Nuclear Svc Water! 4 X 2.4.8 Knowledge of how abnormal operating 4.5 80 procedures are used in conjunction with EOPs.

(CFR: 41.10/43.5/45.13) 000065 Loss of Instrument Air! 8 X AA2. Ability to determine and interpret the 2.6 81 following as they apply to the Loss of Instrument Air:

(CFR: 43.5145.13)

AA2.02 Relationship of flow readings to system operation

ES-401. Rev. 9 4 Form ES-401-2 W/E04 LOCA Outside Containment / 3 X EK2. Knowledge of the interrelations between 3.5 15 the (LOCA Outside Containment) and the following:

(CFR: 41.7/ 45.7)

EK2.1 Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

WE1 1; Loss of Emergency Coolant X EK3. Knowledge of the reasons for the following 3.5 16 Recirculation /4 responses as they apply to the (Loss of Emergency Coolant Recirculation)

(CFR: 41.5/41.10,45.6, 45.13)

EK3.2 Normal, abnormal and emergency operating procedures associated with (Loss of Emergency Coolant Recirculation).

BW/E04; W/E05 Inadequate Heat Transfer - X EAI. Ability to operate and I or monitor the 4.1 17 Loss of Secondary Heat Sink / 4 following as they apply to the (Loss of Secondary Heat Sink)

(CFR: 41.7! 45.5 / 45.6)

EA1 .1 Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features 000077 Generator Voltage and Electric Grid X AK2. Knowledge of the interrelations between 3.0 18 Disturbances /6 Generator Voltage and Electric Grid Disturbances and the following:

(CFR: 41.4, 41.5, 41.7, 41.101 45.8)

AK2.03 Sensors, detectors, indicators K/A Category Totals: 3T 3 3 3 3 3 Group Point Total: 18 j

[ SRO K/A Category Totals:

==== 3 3 Group Point Total: 6

ES-401, Rev. 9 5 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions Tier 1/Group 2 (RO / SRO)

E/APE#/Name/Safety Function A A G K/A Topic(s) IR 1 2312 000001 Continuous Rod Withdrawal / 1 000003 Dropped Control Rod / I 000005 Inoperable/Stuck Control Rod / I 000024 Emergency Boration / I 000028 Pressurizer Level Malfunction / 2 000032 Loss of Source Range NI / 7 X 2.1.23 Ability to perform specific system 4.4 82 and integrated plant procedures during all modes of plant operation.

(CFR: 41.10 / 43.5 / 45.2 / 45.6) 000033 Loss of Intermediate Range NI / 7 000036 (BW/A08) Fuel Handling Accident / 8 X AK3. Knowledge of the reasons for the 2.9 19 following responses as they apply to the Fuel Handling Incidents:

(CFR 41.5,41.10145.6! 45.13)

AK3.02 Interlocks associated with fuel handling equipment 000037 Steam Generator Tube Leak / 3 X AAI. Ability to operate and I or monitor 3.7 20 the following as they apply to the Steam Generator Tube Leak:

(CFR 41.7 /45.5! 45.6)

AA1.01 Maximum controlled depressurization rate for affected S/G 000051 Loss of Condenser Vacuum/4 X AA2. Ability to determine and interpret 4.1 83 the following as they apply to the Loss of Condenser Vacuum:

(CFR: 43.5145.13)

AA2.02 Conditions requiring reactor and/or turbine trip 000059 Accidental Liquid RadWaste Rel. / 9 X AK2. Knowledge of the interrelations 2.7 21 between the Accidental Liquid Radwaste Release and the following:

(CFR 41.7 / 45.7)

AK2.01 Radioactive-liquid monitors 000060 Accidental Gaseous Radwaste Rel. /9 X AK3. Knowledge of the reasons for 3.3 22 the following responses as they apply to the Accidental Gaseous Radwaste:

(CFR 41.5,41.10! 45.6 / 45.13)

AK3.02 Isolation of the auxiliary building ventilation 000061 ARM System Alarms /7 000067 Plant Fire On-site / 8 X 2.2.37 Ability to determine operability 3.6 23 and/or availability of safety related equipment.

(CFR: 41 .7/43.5/45.12) 000068 (BW/A06) Control Room Evac. / 8 000069 (W/E14) Loss of CTMT Integrity / 5

ES-401, Rev. 9 6 Form ES-401-2 000074 CiN/E06&E07) Inadequate Core Cooling / 4 X 2.2.44 Ability to interpret control room 4.4 84 indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

(CFR: 41 .5/43.5/45.12) 000076 High Reactor Coolant Activity /9 X AK2. Knowledge of the interrelations 2.6 24 between the High Reactor Coolant Activity and the following:

(CFR 41.7 I 45.7)

AK2.01 Process radiation monitors W/EOl & E02 Rediagnosis & SI Termination / 3 X EA2. Ability to determine and interpret 4.0 85 the following as they apply to the (Reactor Trip or Safety Injection Rediagnosis)

(CFR: 43.5/45.13)

EA2.1 Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

W/E13 Steam Generator Over-pressure /4 X EA2. Ability to determine and interpret 2.9 the following as they apply to the (Steam Generator Overpressure)

(CFR: 43.5/45.13)

E.A2.1 Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

W/E15 Containment Flooding / 5 W/E16 High Containment Radiation / 9 X EK1. Knowledge of the operational 2.7 26 implications of the following concepts as they apply to the (High Containment Radiation)

(CFR: 41 .8/41.10, 45.3)

EK1.1 Components, capacity, and function of emergency systems.

BW/A01 Plant Runback / I BW/A02&A03 Loss of NNI-XIY / 7 BW/A04 Turbine Trip / 4 BW/A05 Emergency Diesel Actuation / 6 BWIAO7 Flooding / 8 BW/E03 Inadequate Subcooling Margin / 4 BW/E08; W/E03 LOCA Cooldown Depress. / 4 X EAt Ability to operate and! or monitor 3.7 27 the following as they apply to the (LOCA Cooldown and Depressurization)

(CFR: 41.7 / 45.5/ 45.6)

EA1 .3 Desired operating results during abnormal and emergency situations.

BWIEO9; CE/Al 3; W/E09&E1 0 Natural Circ. / 4 BW/E13&E14 EOP Rules and Enclosures CE/All; W/E08 RCS Overcooling PTS / 4 CEIA16 Excess RCS Leakage /2 CE/E09 Functional Recovery K/A Category Point Totals: 1 2 2 2 1 Group Point Total: 9 K/A Category Point Totals: (SRO)

= == 2 2 Group Point Total:

1

ES-401, Rev. 9 7 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems Tier 2/Group 1(RO / SRO)

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

1234561234 003 Reactor Coolant Pump X K5 Knowledge of the operational 3.2 28 implications of the following concepts as they apply to the RCPS:

(CFR: 41.5 / 45.7)

K5.04 Effects of RCP shutdown on secondary parameters, such as steam pressure, steam flow, and feed flow X 2.2.41 Ability to obtain and interpret 3.9 86 station electrical and mechanical drawings.

(CFR: 41.10 / 45.12 / 45.13) 004 Chemical and Volume Control X K5 Knowledge of the operational 2.6 29 implications of the following concepts as they apply to the CVCS:

(CFR: 41.5/45.7)

K5.08 Estimation of subcritical multiplication factor (K-eff) by means other than the 6-factor formula: relationship of count rate changes to reactivity changes 005 Residual Heat Removal X KI Knowledge of the physical 3.1 30 connections and/or cause-effect relationships between the RHRS and the following systems:

(CFR: 41.2 to 41.9/ 45.7 to 45.8)

K1.12 Safeguard pumps X K6 Knowledge of the effect of a loss or 2.5 31 malfunction on the following will have on the RHRS:

(CFR: 41.7/45.7)

K6.03 RHR heat exchanger X 2.4.50 Ability to verify system alarm 4.0 87 setpoints and operate controls identified in the alarm response manual.

(CFR: 41.10/43.5/45.3) 006 Emergency Core Cooling X 2.4.20 Knowledge of the operational 3.8 32 implications of EOP warnings, cautions, and notes.

(CFR: 41.10 / 43.5/ 45.13)

ES-401, Rev. 9 8 Form ES-401-2 007 Pressurizer Relief/Quench Tank X K5 Knowledge of the operational 3.1 33 implications of the following concepts as the apply to PRTS:

(CFR: 41.5/45.7)

K5.02 Method of forming a steam bubble in the PZR X A2 Ability to (a) predict the impacts of 2.6 88 the following malfunctions or operations on the P S; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

(CFR: 41.5 / 43.5/45.3 / 45.13)

A2.07 Recirculating quench tank 008 Component Cooling Water X A3 Ability to monitor automatic 3.0 34 operation of the CCWS, including:

(CFR: 41.7/45.5)

A3.05 Control of the electrically operated, automatic isolation valves in the CCWS X A4 Ability to manually operate and/or 3.0 35 monitor in the control room:

(CFR: 41.7/45.5)

A4.09 CCW temperature control valve X A2 Ability to (a) predict the impacts of 3.2 89 the following malfunctions or operations on the CCWS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

(CFR: 41.5 /43.5 /45.3 /45.13)

A2.03 High/low CCW temperature 010 Pressurizer Pressure Control X KI Knowledge of the physical 3.6 36 connections and/or cause-effect relationships between the PZR PCS and the following systems:

(CFR: 41.2 to 41.9 / 45.7 to 45.8)

Ki .03 RCS 012 Reactor Protection X A3 Ability to monitor automatic 3.7 37 operation of the RPS, including:

(CFR: 41.7/45.5)

A3.06 Trip logic

ES-401. Rev. 9 9 Form ES-401-2 013 Engineered Safety Features X A2 Ability to (a) predict the impacts of 3.7 38 Actuation the following malfunctions or operations on the ESFAS; and (b) based Ability on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations; (CFR: 41.5 / 43.5 / 45.3 / 45.13)

A2.05 Loss of do control power 022 Containment Cooling X A4 Ability to manually operate and/or 3.2 39 monitor in the control room:

(CFR: 41.7 1 45.5 to 45.8)

A4.03 Dampers in the CCS 025 Ice Condenser X K6 Knowledge of the effect of a loss or 3.4 40 malfunction of the following will have on the ice condenser system:

(CFR: 41.7 /45.7)

K6.O1 Upper and lower doors of the ice condenser 026 Containment Spray X A2 Ability to (a) predict the impacts of 4.1 41 the following malfunctions or operations on the CSS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

(CFR: 41.5 / 43.5 / 45.3 / 45.13)

A2.03 Failure of ESF 039 Main and Reheat Steam X Al Ability to predict and/or monitor 3.0 42 changes in parameters (to prevent exceeding design limits) associated with operating the MRSS controls including:

(CFR: 41.5 / 45.5)

A1.06 Main steam pressure 059 Main Feedwater X K3 Knowledge of the effect that a loss 3.6 43 or malfunction of the MFW will have on the following:

(CFR: 41.7/45.6)

K3.04 RCS X A2 Ability to (a) predict the impacts of 2.7 44 the following malfunctions or operations on the MFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

(CFR: 41.5 / 43.5 / 45.3 / 45.13)

A2.03 Overfeeding event

ES-401, Rev. 9 10 Form ES-401-2 061 Auxiliary/Emergency Feedwater X K2 Knowledge of bus power supplies to 4.0 45 the following:

(CFR: 41.7)

K2.03 AFW diesel driven pump x K4 Knowledge of AFW design feature(s) 2.7 46 and/or interlock(s) which provide for the following:

(CFR: 41.7)

K4.13 Initiation of cooling water and lube oil 062 AC Electrical Distribution X Al Ability to predict andlor monitor 2.5 changes in parameters (to prevent exceeding design limits) associated with operating the ac distribution system controls including:

(CFR: 41.5/45.5)

Al .03 Effect on instrumentation and controls of switching power supplies X 2.4.49 Ability to perform without 4.6 48 reference to procedures those actions that require immediate operation of system components and controls.

(CFR: 41.10/ 43.2 / 45.6) 063 DC Electrical Distribution X AS Ability to monitor automatic 2.7 49 operation of the DC electrical system, including:

(CFR: 41.7/45.5)

A3.0l Meters, annunciators, dials, recorders, and indicating lights 064 Emergency Diesel Generator x K6 Knowledge of the effect of a loss or 2.7 50 malfunction of the following will have on the EDIG system:

(CFR: 41.7 /45.7)

K6.07 Air receivers x 2.2.4 (multi-unit license) Ability to explain 3.6 51 the variations in control board/control room layouts, systems, instrumentation, and procedural actions between units at a facility.

(CFR: 41.6/41.7/41.10/45.1 /45.13) 073 Process Radiation Monitorin K3 Knowledge of the effect that a loss 3.6 52 or malfunction of the PRM system will have on the following:

(CFR: 41.7/45.6)

K3.01 Radioactive effluent releases

ES-401, Rev. 9 11 Form ES-401-2 076 Service Water X K4 Knowledge of SWS design feature(s) 2.9 53 and/or interlock(s) which provide for the following:

(CFR: 41/7)

K4.02 Automatic start features associated with SWS pump controls X 2.4.34 Knowledge of RO tasks 4.1 90 performed outside the main control room during an emergency and the resultant operational effects.

(CFR: 41.10 /43.5 / 45.13) 078 Instrument Air X K2 Knowledge of bus power supplies to 3.3 54 the following:

(CFR: 41.7)

K2.02 Emergency air compressor 103 Containment X K3 Knowledge of the effect that a loss 3.3 55 or malfunction of the containment system will have on the following:

(CFR: 41.7/45.6)

K3.01 Loss of containment integrity under shutdown conditions K/A Category Point Totals: 3 2 3 3 3 Group Point Total: 28 K/A Category Point Totals: (SRO) ]_ = =

2 3 Group Point Total: 1

ES-401, Rev. 9 12 Form ES-401-2

[L___________________

ES-401 PWR Examination Outline PlantSjstems-ller2/Group2(RO/SRO)

Form ES-401-2 System # / Name K K K K K K A A A A G K/A Topic(s) IR 1234561234 001 Control Rod Drive X K3 Knowledge of the effect that a loss or 3.4 -g malfunction of the CRDS will have on the following:

(CFR: 41.7/45.6)

K3.02 RCS 002 Reactor Coolant X A3 Ability to monitor automatic 3.7 57 operation of the RCS, including:

(CFR: 41.7/45.5)

A3.01 Reactor coolant leak detection system 011 Pressurizer Level Control X Al Ability to predict and!or monitor 3.1 58 changes in parameters (to prevent exceeding design limits) associated with operating the PZR LCS controls including:

(CFR: 41.5 / 45.5)

Al .04 T-ave 014 Rod Position Indication X KI Knowledge of the physical 3.0 59 connections andlor cause-effect relationships between the RPIS and the following systems:

(CFR: 41.3 to 41.9 / 45.7 to 45.8)

Kl.02 NIS 015 Nuclear Instrumentation X A2 Ability to (a) predict the impacts of 3.9 91 the following malfunctions or operations on the NIS; and (b based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

(CFR: 41.5 1 43.5 / 45.3 / 45.5)

A2.01 Power supply loss or erratic operation 016 Non-nuclear Instrumentation 017 In-core Temperature Monitor 028 Hydrogen Recombiner and X K2 Knowledge of bus power supplies to 2.5 60 Purge Control the following:

(CFR: 41.7)

K2.0l Hydrogen recombiners 029 Containment Purge 033 Spent Fuel Pool Cooling X 2.4.46 Ability to verify that the alarms 4.2 61 are consistent with the plant conditions.

(CFR: 41.10/43.5/45.3/45.12) 034 Fuel Handling Equipment

ES-401, Rev. 9 13 Form ES-401-2 035 Steam Generator X K6 Knowledge of the effect of a loss or 2.6 62 malfunction on the following will have on the SIGS:

(CFR: 41.7/45.7)

K6.03 S/G level detector 041 Steam DumplTurbine Bypass Control 045 Main Turbine Generator 055 Condenser Air Removal 056 Condensate X A2 Ability to (a) predict the impacts of 2.6 63 the following malfunctions or operations on the Condensate System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

(CFR: 41.5/43.5/ 45.3 / 45.13)

A2.04 Loss of condensate pumps 068 Liquid Radwaste 071 Waste Gas Disposal X K5 Knowledge of the operational 2.5 64 implication of the following concepts as they apply to the Waste Gas Disposal System:

(CFR: 41.5/45.7)

K5.04 Relationship of hydrogen/oxygen concentrations to flammability 072 Area Radiation Monitoring X 2.2.37 Ability to determine operability 4.6 92 andlor availability of safety related equipment.

(CFR: 41.7/43.5/45.12) 075 Circulating Water X K4 Knowledge of circulating water 2.5 65 system design feature(s) and interlock(s) which provide for the following:

(CFR: 41.7)

K4.01 Heat sink 079 Station Air 086 Fire Protection X 2.4.11 Knowledge of abnormal condition 4.2 93

. procedures.

(CFR: 41.10/43.5/45.13)

KIACategoryPontTotals 1 1 iji 1 1 iiNi1OII GroupPointTotal K/A Category Point Totals: (SRO)

=

i I.iJ 11 Group Point Total: 3

ES-401, Rev. 9 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Sequoya Date of Exam: December2012 RO SRO-Only Category K/A # Topic IR Q# IR Q#

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

2 1

  • 25 (CFR: 41.10/43.5/45.12) 3.9 66 Ability to identify and interpret diverse indications to validate the response of another indication.

2.1.45 4.3 61 (CFR: 41 .7/43.5/45.4)

Knowledge of individual licensed operator responsibilities related to shift staffing, such as Conduct of medical requirements, no-solo operation, 2.1.4 maintenance of active license status, IOCFR55, 3.8 94 Operations etc.

(CFR: 41.10 / 43.2)

Knowledge of criteria or conditions that require plant-wide announcements, such as pump starts, 2.1.14 reactor trips, mode changes, etc. 3.1 95 (CFR: 41.10 /43.5/45.12)

Subtotal 2 2 Ability to perform pre-startup procedures for the facility, including operating those controls associated with plant equipment that could affect 2.2.1 reactivity. 4.5 68 (CFR: 41.5/41.10/43.5/43.6/45.1)

Knowledge of limiting conditions for operations and safety limits.

2.2.22 4.0 69 (CFR: 41.5 / 43.2 / 45.2)

Ability to interpret control room indications to verify the status and operation of a system, and

2. understand how operator actions and directives Equipment Control 2.2.44 4.2 70 affect plant and system conditions.

(CFR: 41 .5/43.5/45.12)

Knowledge of the process for making design or operating changes to the facility.

2.2.5 3.2 96 (CFR: 41.10 /43.3/45.13)

Ability to track Technical Specification limiting conditions for operations.

2.2.23 4.6 97 (CFR: 41.10/43.2/45.13)

Subtotal 3 2 Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring 2 3 equipment, etc. 2.9 71 (CFR: 41.11 /41.12/43.4/45.9) 3.

Knowledge of radiological safety principles Radiation Control pertaining to licensed operator duties, such as containment entry requirements, fuel handling 2.3.12 responsibilities, access to locked high-radiation 3.2 72 areas, aligning filters, etc.

(CFR: 41.12/45.9/45.10)

ES-401, Rev. 9 15 Form ES-401-2 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or 2.314 emergency conditions or activities. 3.8 98 (CFR: 41.12/43.4/45.10)

Subtotal 2 1 Knowledge of the emergency plan.

2.4.29 (CFR: 41.10/43.5/45.11) 3.1 73 Knowledge of emergency communications systems and techniques.

2 4 43 (CFR: 41.10/45.13) 3.2 74 Ability to perform without reference to procedures those actions that require immediate operation of

4. 2.4.49 system components and controls. 4.6 75 Emergency (CFR: 41.10 / 43.2 / 45.6)

Procedures! Plan Knowledge of crew roles and responsibilities during EOP usage.

2 4 13 (CFR:41.10145.12) 4.6 99 Ability to take actions called for in the facility emergency plan, including supporting or acting as 2.4.38 emergency coordinator if required. 44 100 (CFR: 4110/43.5/45.11)

Subtotal 3 2 Tier3PointTotal 10 7

ES-301 Administrative Topics Outline Form ES-301-1 Facility: SeQuoyah Nuclear Station 1 & 2 Date of Examination:12/3-7 2012 Exam Level: RO X SRO LI Operating Test No: 201 2-302 Administrative Topic (see Type Describe activity to be performed Code*

[ Note)

Conduct of Operations R, D K/A 2.1.26 (3.4) Perform a Containment Formaldehyde Stay Time Calculation.

Conduct of Operations R, M K/A 2.1.7 (4.4) Monitor Critical Safety Status Trees for Degraded Core Cooling.

Equipment Control R, M K/A 2.2.12 (3.7) Perform a Reactivity Balance Calculation for a Power Ascension.

Radiation Control N/A Not Examined.

Emergency Procedures/Plan R, N K/A 2.4.39 (3.9) Complete and Review a State Notification Form.

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)

(N)ew or (M)odified from bank ( 1)

(P)revious 2 exams ( 1; randomly selected)

A.1.a While acting as an extra RO, the examinee will perform a Containment Formaldehyde Stay Time Calculation using O-Tl-OPS-000-001 .0, Containment Formaldehyde Stay Time Calculation in preparation for a Containment entry.

A.1.b While acting as a member of the operating crew and given the conditions of a plant emergency, the examinee will interpret and apply plant conditions using 1-FR-0 UNIT 1 STATUS TREES to determine that the Safety Functions for Core Cooling, Pressurized Thermal Shock and Containment are all severely challenged.

A.2 While acting as an extra RO, the examinee will perform a Reactivity Balance for a power ascension using 0-SO-62-7, Boron Concentration Control Appendix E, Reactivity Balance.

A.3 Not examined.

A.4 While acting as the Site Communicator and given data for a plant emergency, will use EPIP-5 GENERAL EMERGENCY and complete a state notification.

ES-301 Administrative Topics Outline Form ES-301-1 Facility: SeQuoyah Nuclear Station 1 & 2 Date of Examination:12/3-7 2012 Exam Level: RO SRO X Operating Test No: 201 2-302 Administrative Topic (see Type Describe activity to be performed Note) Code*

Conduct of Operations R, D K/A 2.1 .25 (4.2) Determine the Operability of a Boric Acid Tank before use.

Conduct of Operations R, D K/A 2.1.7 (4.4) Perform a RCS Void Determination and Apply the Result to Determine RCS Pump Sweep Requirements.

Equipment Control R, N K/A 2.2.14 (4.3) Perform a Risk Assessment Using EQOS.

Radiation Control R, N K/A 2.3.14 (3.8) Determine Reporting Requirements for a Contaminated and Injured person.

Emergency Procedures/Plan R, M K/A 2.4.41 (4.6) Classify the Event using the EPIP-1 and Complete a TVA INITIAL NOTIFICATION.

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)

(N)ew or (M)odified from bank ( 1)

(P)revious 2 exams ( 1; randomly selected)

C

A.1.a While acting as the on-shift Unit Supervisor, the examinee will interpret Boric Acid Tank level and boron concentration and apply the data using the Technical requirements Manual Section 3.1.2.6 figure to determine if the Boric Acid Storage Tank may be placed in service.

A.1.b While acting as an on-shift Unit Supervisor during a refueling outage, the examinee will calculate the volume of water required to fill the voids in the RCS and to raise pressure and determine that individual RCP Pump sweeps are required to restore the RCS from the outage condition.

A.2 While acting as the Operations Work Control SRO and given a hypothetical maintenance situation, the examinee will perform a Risk Assessment using the Equipment Out of Service computer program. As a result of performing the test case, the examinee will conclude the Core Damage Frequency (CDF) and Large Early R&ease Frequency (LERF) Multipliers change and the LERF indicator changes from green to yellow.

A.3 While acting as an on-shift Unit Supervisor during a refueling outage, the examinee will receive the report of a contaminated injured man. The examinee will be directed to list all required reports to outside agencies and to list by title, the internal TVA notification requirements.

A.4 While acting as the Site Emergency Director and given data for a plant emergency, the examinee will use EPIP-1 EMERGENCY PLAN CLASSIFICATION MATRIX and interpret the data to determine the correct Emergency Classification of General Emergency. The examinee will then use EPIP-5 GENERAL EMERGENCY to determine the correct Protective Action Recommendation.

ES-3O1 Control Room!In-Plant Systems Outline Form ES-301-2 Facility: Seguoyah Nuclear Station 1 & 2 Date of Examination: 12/3-7 2012 Exam Level: RO X SRO-l El SRO-U El Operating Test No: 2012-302 Control Room Systems (8 for RO); (7 for SRO-l); (2 or 3 for SRO-U, including 1 ESF)

System / JPM Title Type Code* Safety Function

a. KIA EPE 029 EA 1.02 (3.6/3.3) Establish Emergency Boration Using The S, A, M 1 Normal Boration Flowpath.
b. K/A 004 A4.06 (3.6/3.1) Remove Excess Letdown from Service. 5, D 2
c. K/A EPE 011 EA1 .11 (4.2/4.2) Transfer To Hot Leg Recirculation With A S, A, M, EN, L 3 Safety Injection Pump Trip.
d. K/A APE 025 AK1 .01 (3.9/4.3) Respond To A RHR Pump Trip And Establish S, N, L 4P RHR Flow Using The Standby RHR Pump.
e. KJA EPE E05 EA2.1 (3.4/4.4) Establish Secondary Heat Sink Using Main 5, M, L 4S Feedwater.
f. K/A EPE E14 EA 1.1 (3.7/3.7) Respond To High Containment Pressure S, A, D, L 5 Condition By Placing RHR Spray In Service.
g. K/A APE 056 AA1 .04 (3.2/3.1) Restore Offsite Power From The Emergency 5, M, L 6 Diesel Generator To The Shutdown Board.
h. K/A 016 A2.01 (3.0/3.1) Remove A Steam Pressure Channel From Bypass On C The Digital Control System.

In-Plant Systems (3 for RO); (3 for SRO-l); (3 or 2 for SRO-U) 5, D 7

i. K/A EPE 029 EA 1.11 (3.9/4.1) Locally Open The Reactor Trip Breakers A, R, D, E 1 During An ATWS.
j. K/A APE 068 Ml .01 (4.3/4.5) Operate Steam Generator Atmospheric Dump A, N, L, E 8 PCV-1-12 Locally. (New Plant Modification)
k. K/A EPE 055 EA2.03 (3.9/4.7) Reset The lA-A EDG Overspeed Trip N, L, E 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-l / SRO-U (A)Iternate path 4-6 / 4-6 / 2-3 (C)ontrol room (D)irectfrombank 9/<_81<_4 (E)mergency or abnormal in-plant 1 / 1 I 1 (EN)gineered safety feature - / - / 1 (control room system)

(L)ow-Power / Shutdown 1 / 1 / I (N)ew or (M)odified from bank including 1(A) 2/ 2 / 1 C (P)revious 2 exams (R)CA (S)imulator 3 I 3 / 2 (randomly selected)

a. The examinee will assume the shift as the ATC during an ATWS condition. The examinee will be directed to initiate Emergency Boration using EA-68-4 EMERGENCY BORATION section 4.2. During the alignment the emergency borate valve will fail to operate. The examinee uses the alternate path to establish emergency boration flow through the normal boration flowpath.
b. The examinee will assume the shift with Excess Letdown in service. The examinee will remove Excess Letdown from service using 1 -SO-62-6, Excess Letdown Section 7.0.
c. The examinee will assume the shift five hours after a LOCA. The ECCS was placed in Cold Leg Recirculation using ES-i .3, Transfer to RHR Containment Sump. The examinee will be directed to place the ECCS in Hot Leg Recirculation using ES-i .4, Transfer to Hot Leg Recirculation. During the alignment, the 1A Safety Injection pump trips. The examinee uses the alternate path to establish Hot Leg recirculation flow through the 1 B Safety Injection pump.
d. The examinee will assume the shift in MODE 5 and the A Train RHR pump in service in the Shutdown Cooling mode with a slight cooldown in progress. While monitoring plant operation the A Train RHR pump trips and the examinee will respond to the condition using AOP-R.03, RHR SYSTEM MALFUNCTION.

The examinee will place the B Train RHR pump in service and re-establish control of RHR temperature with a slight cooldown.

e. The examinee will assume the shift in MODE 3 with no AFW flow available. The examinee will use EA 2, Establishing Secondary Heat sink using Main Feedwater or Condensate section 4.4 and start a Main Feed pump and control feed flow using a Main Feed Regulating Bypass Valve.
f. The examinee will assume the shift after a LOCA and a High Containment Pressure condition. The examinee will use FR-Z. 1, High Containment Pressure, starting at step 13 to establish RI-IR Spray flow to reduce Containment pressure. During the alignment, the B Train RHR Spray flow isolation valve FCV 41 fails to operate. The examinee uses the alternate path to re-establish B Train RHR to the injection mode and establishes A Train RHR to the Containment Spray mode.
g. The examinee will assume the shift in MODE 3 following a Loss of Power event to the 1A 6.9kv Shutdown Board with the 1A EDG supplying power to the board. The exam inee will restore power using EA-202-1 Restoring Offsite Power to 6900V Shutdown Boards. The examinee will parallel the 1A EDG to offsite power using the normal supply breaker and ultimately unload stop the 1A EDG.
h. The examinee will assume the shift in MODE 1 and a Steam Pressure detector in BYPASS on the Digital Control System. The examinee will use 1-SO-98-1 DISTRIBUTED CONTROL SYSTEM to remove the Steam Pressure detector from BYPASS and restore the detector to service.

The examinee will assume the shift in MODE 1 with an ATWS condition in progress. The examinee will be directed to locally trip the Reactor. The Reactor Trip breakers will fail to open locally. The examinee uses the alternate path to locally open both CRDM MG set output breakers.

j. The examinee will assume the shift in MODE 3 and a Control Room Evacuation in progress. The examinee will be directed to take local control of PCV-1 -5, S/G #1 Atmospheric Dump valve using AOP C.04, SHUTDOWN FROM AUXILIARY CONTROL ROOM, Appendix K.1. The examinee attempts to locally operate PCV-1 -5, S/G #1 Atmospheric Dump valve but discovers the valve to be immoveable. The examinee uses the alternate path to take local control of PCV-1 -1 2, S/G #2 Atmospheric Dump valve.

The local controls for PCV-1-12, SIG #2 Atmospheric Dump valve are a recently added plant modification.

k. The examinee will assume the shift in MODE 3 following a major grid disturbance and Unit 1 in a Loss of All AC power. The examinee will be directed to reset the overspeed trip mechanisms on the 1A EDG. The examinee will reset both overspeed trip mechanisms and subsequently reset the 1A EDG lockout relay to start the IA EDG.

ES-30 1 ControLRoornitn-Plant Systems Outline Form ES4OI-2 Facility: Seguoyah Nuclear Station 1 & 2 Date of Examination: 12/3-7 2012 Exam Level: RO SRO-I X SRO-U Operating Test No: 2012-302 Control Room Systems (8 for RO); (7 for SRO-l); (2 or 3 for SRO-U, including 1 ESF)

System / JPM Title Type Code* Safety Function

a. K/A EPE 029 EA 1.02 (3.6/3.3) Establish Emergency Boration Using The 5, A, M 1 Normal Boration Flowpath.
b. K/A 004 A4.06 (3.6/3.1) Remove Excess Letdown from Service. 5, D 2
c. K/A EPE 011 EAI .11 (4.2/4.2) Transfer To Hot Leg Recirculation With A 5, A, M, EN, L 3 Safety Injection Pump Trip.
d. N/A
e. K/A EPE E05 EA2.1 (3.4/4.4) Establish Secondary Heat Sink Using Main S, M, L 4S Feedwater.
f. K/A EPE E14 EA 1.1 (3.7/3.7) Respond To High Containment Pressure 5, A, D, L 5 Condition By Placing RHR Spray In Service.
g. K/A APE 056 AA1 .04 (3.2/3.1) Restore Offsite Power From The Emergency 5, M, L 6 Diesel Generator To The Shutdown Board.
h. K/A 016 A2.01 (3.0/3.1) Remove A Steam Pressure Channel From Bypass On C The Digital Control System.

In-Plant Systems (3 for RO); (3 for SRO-l); (3 or 2 for SRO-U) 5, D 7 I. K/A EPE 029 EA 1.11 (3.9/4.1) Locally Open The Reactor Trip Breakers A, R, D, E 1 During An ATWS.

j. K/A APE 068 AA1 .01 (4.3/4.5) Operate Steam Generator Atmospheric Dump A, N, L, E 8 PCV-1-12 Locally. (New Plant Modification)
k. K/A EPE 055 EA2.03 (3.9/4.7) Reset The lA-A EDG Overspeed Trip N, L, E 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 / SRO-U (A)lternate path 4-6 I 4-6 / 2-3 (C)ontrol room (D)irect from bank 91814 (E)mergency or abnormal in-plant 1 I 1 I 1 (EN)gineered safety feature - / - / 1 (control room system)

(L)ow-Power / Shutdown 1 I 11>_i (N)ew or (M)odified from bank including 1(A) 21>_2/ 1 C (P)revious 2 exams (R)CA (S)lmuator 3 / 3 I 2 (randomly selected) 1 I 1 / 1

a. The examinee will assume the shift as the ATC during an ATWS condition. The examinee will be directed to initiate Emergency Boration using EA-68-4 EMERGENCY BORATION section 4.2. During the alignment the emergency borate valve will fail to operate. The examinee uses the alternate path to establish emergency boration flow through the normal boration flowpath.
b. The examinee will assume the shift with Excess Letdown in service. The examinee will remove Excess Letdown from service using 1-SO-62-6, Excess Letdown Section 7.0.
c. The examinee will assume the shift five hours after a LOCA. The ECCS was placed in Cold Leg Recirculation using ES-i .3, Transfer to RHR Containment Sump. The examinee will be directed to place the ECCS in Hot Leg Recirculation using ES-i .4, Transfer to Hot Leg Recirculation. During the alignment, the 1A Safety Injection pump trips. The examinee uses the alternate path to establish Hot Leg recirculation flow through the 1 B Safety Injection pump.
d. Not examined.
e. The examinee will assume the shift in MODE 3 with no AFW flow available. The examinee will use EA 2, Establishing Secondary Heat sink using Main Feedwater or Condensate section 4.4 and start a Main Feed pump and control feed flow using a Main Feed Regulating Bypass Valve.
f. The examinee will assume the shift after a LOCA and a High Containment Pressure condition. The examinee will use FR-Z. 1, High Containment Pressure, starting at step 13 to establish RHR Spray flow to reduce Containment pressure. During the alignment, the B Train RHR Spray flow isolation valve FCV 41 fails to operate. The examinee uses the alternate path to re-establish B Train RHR to the injection mode and establishes A Train RHR to the Containment Spray mode.
g. The examinee will assume the shift in MODE 3 following a Loss of Power event to the 1A 6.9kv Shutdown Board with the 1A EDG supplying power to the board. The exam inee will restore power using EA-202-i Restoring Offsite Power to 6900V Shutdown Boards. The examinee will parallel the iA EDG to offsite power using the normal supply breaker and ultimately unload stop the IA EDG.
h. The examinee will assume the shift in MODE 1 and a Steam Pressure detector in BYPASS on the Digital Control System. The examinee will use 1-SO-98-1 DISTRIBUTED CONTROL SYSTEM to remove the Steam Pressure detector from BYPASS and restore the detector to service.
i. The examinee will assume the shift in MODE 1 with an ATWS condition in progress. The examinee will be directed to locally trip the Reactor. The Reactor Trip breakers fail to open locally. The examinee uses the alternate path to locally open both CRDM MG set output breakers.
j. The examinee will assume the shift in MODE 3 and a Control Room Evacuation in progress. The examinee will be directed to take local control of PCV-1-5, S/G #1 Atmospheric Dump valve using AOP C.04, SHUTDOWN FROM AUXILIARY CONTROL ROOM, Appendix K.1. The examinee attempts to locally operate PCV-1-5, S/G #1 Atmospheric Dump valve but discovers the valve to be immoveable. The examinee uses the alternate path to take local control of PCV-1 -12, S/G #2 Atmospheric Dump valve.

The local controls for PCV-i-i2, S/G #2 Atmospheric Dump valve are a recently added plant modification.

k. The examinee will assume the shift in MODE 3 following a major grid disturbance and Unit 1 in a Loss of All AC power. The examinee will be directed to reset the overspeed trip mechanisms on the 1A EDG. The examinee will reset both overspeed trip mechanisms and subsequently reset the 1A EDG lockout relay to start the IA EDG.

ES-301 Control Roomlln-Plant Systems Outline form_ES-301 -2 Facility: Seguoyah Nuclear Station 1 & 2 Date of Examination: 12/3-7 2012 Exam Level: RO El SRO-l El SRO-U X Operating Test No: 2012-302 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. N/A
b. K/A 004 A4.06 (3.6/3.1) Remove Excess Letdown from Service. S, D 2
c. K/A EPE 011 EAI .11 (4.2/4.2) Transfer To Hot Leg Recirculation With A S, A, M, EN, L 3 Safety Injection Pump Trip.
d. N/A
e. N/A
f. N/A
g. N/A
h. K/A 016 A2.01 (3.0/3.1) Remove A Steam Pressure Channel From Bypass On 5, D 7 The Digital Control System.

In-Plant Systems (3 for RO); (3 for SRO-l); (3 or 2 for SRO-U)

i. K/A EPE 029 EA 1.11 (3.9/4.1) Locally Open The Reactor Trip Breakers A, R, D, E 1 During An A1WS.
j. K/A APE 068 AA1 .01 (4.3/4.5) Operate Steam Generator Atmospheric Dump A, N, L, E 8 PCV-1-12 Locally. (New Plant Modification)
k. N/A 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-l / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 (C)ontrol room (D)irect from bank 9/ 8I 4 (E)mergency or abnormal in-plant 1 / 1 / I (EN)gineered safety feature - / - I 1 (control room system)

(L)ow-Power I Shutdown 1 / 1 / 1 (N)ew or (M)odified from bank including 1 (A) 2I 2/ 1 (P)revious 2 exams 3I 3 I 2 (randomly selected)

(R)CA (S)imulator

a. Not examined.
b. The examinee will assume the shift with Excess Letdown in service. The examinee will remove Excess Letdown from service using 1-SO-62-6, Excess Letdown Section 7.0.
c. The examinee will assume the shift five hours after a LOCA. The ECCS was placed in Cold Leg Recirculation using ES-i .3, Transfer to RHR Containment Sump. The examinee will be directed to place the ECCS in Hot Leg Recirculation using ES-i .4, Transfer to Hot Leg Recirculation. During the alignment, the 1A Safety Injection pump trips. The examinee uses the alternate path to establish Hot Leg recirculation flow through the 1 B Safety Injection pump.
d. Not examined.
e. Not examined.
f. Not examined.
g. Not examined.
h. The examinee will assume the shift in MODE 1 and a Steam Pressure detector in BYPASS on the Digital Control System. The examinee will use 1-S0-98-i DISTRIBUTED CONTROL SYSTEM to remove the Steam Pressure detector from BYPASS and restore the detector to service.
i. The examinee will assume the shift in MODE I with an ATWS condition in progress. The examinee will be directed to locally trip the Reactor. The Reactor Trip breakers fail to open locally. The examinee uses the alternate path to locally open both CRDM MG set output breakers.

The examinee will assume the shift in MODE 3 and a Control Room Evacuation in progress. The examinee will be directed to take local control of PCV-i -5, S/G #1 Atmospheric Dump valve using AOP C.04, SHUTDOWN FROM AUXILIARY CONTROL ROOM, Appendix K.1. The examinee attempts to locally operate PCV-1-5, S/G #1 Atmospheric Dump valve but discovers the valve to be immoveable. The examinee uses the alternate path to take local control of PCV-1-12, S/G #2 Atmospheric Dump valve.

The local controls for PCV-1-i2, S/G #2 Atmospheric Dump valve are a recently added plant modification.

k. Not examined.
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