ML12277A382

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Draft - Outlines (Folder 2)
ML12277A382
Person / Time
Site: Calvert Cliffs  Constellation icon.png
Issue date: 09/12/2012
From: Moore R
Constellation Energy Nuclear Group
To: D'Antonio J
Operations Branch I
Jackson D
Shared Package
ML12067A049 List:
References
TAC U01849
Download: ML12277A382 (31)
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ES-40l PWR Examination Outline Form ES-401-2 II Facility: Calvert Cliffs Nuclear Power Plant Date of Exam: 08/13120121 RO Category KIA Points SRO Only Points Tier Group K

1 K

2 K

3 K

4 K

5 K

6 A

I A

2 A

3 A

4 G

Total A2 G

Total

1. Emergency &

Abnormal Plant 1

2 3

2 3

2 3

1 N/A 3

2 3

1 N/A q 18 9

3 2

3 2

6 4

Evolutions Tier Totals 5

5 4

5 4

4 27 5

5 10 I

2 2

3 3

2 2

3 2

3 3

28 3

2 5

2. Plant Systems 2

1 1

1 1

1 1

0 1

1 1

10 1

1 1

3 Tier Totals 3

3 4

4 3

3 3

3 4

4 38 5

3 8

3. Generic Knowledge & Abilities Categories I

3 2

2 3

3 4

2 10 1

2 2

I 3

2 4

2 7

Note:

1.

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

2.

The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-I 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.l.b of ES-401 for guidance regarding the elimination of inappropriate KIA statements.

4.

Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.

5.

Absent a plant-specific priority, only those K/As having an importance rating (IR) of2.5 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 I and 2 from the shaded systems and KIA categories.

7.*

The generic (G) KlAs in Tiers I and 2 shall be selected from Section 2 ofthe KIA Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.I.b ofES-40 I for the applicable KIAs.

8.

On the following pages, enter the KIA numbers, a brief description of each topic, the topics' importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams.

9.

For Tier 3, select topics from Section 2 of the KIA catalog, and enter the KIA numbers, descriptions, IRs, and point totals

(#) on Form ES-401-3. Limit SRO selections to KlAs that are linked to 10 CFR 55.43.

ES-40I PWR Examination Outline Form ES-40I-2 Emergency & Abnormal Plant Evolutions - Tier 11 Group I - REACTOR OPERATOR K

K K A

A '~~i E/APE #/Name/Safety Function KA Topic Imp Pts 1

2 3

1 2

I I

I EKI - Knowll!dge of the operational implications of the following concepts as 007 Reactor Trip - Stabilization -

[;iv.* ~fu1I~ they apply to the reactor trip:

X 3.6 I

Recovery 1 I EKl.04 - Decrease in reactor power following reactor trip (prompt drop and subsequent decay)

AK2 - Knowlledge of the interrelations 008 Pressurizer Vapor Space Accident

~..

between the F'ressurizer Vapor Space X

2.7 J

13 Accident and the following:

AK2.02 - Sensors and detectors

[p:f31i;;'; EK3 - Knowledge of the reasons for the I!~;J;J; ;;;~~ following responses as the apply to the 009 Small Break LOCA 1 3 X

small break LOCA:

4.1 I

~~,,~ EK3.24 - ECCS throttling or termination criteria

,i,.;',

EA2 - Ability to determine or interpret the

.~\\ following as tllIey apply to a Large Break oII Large Break LOCA 1 3

,"': LOCA:

3.7 I

EA2.13 - Difference between overcooling

,,' and LOCA indlications AK3 - Knowlledge of the reasons for the following responses as they apply to the 022 Loss ofRx Coolant Makeup 1 2 X

Loss of Reactor Coolant Makeup:

3.1 1

AK3.03 - Perfi:lrmance oflineup to establish excess letdown after determining need AK2 - Knowlledge of the interrelations

.;.~I between the Loss of Residual Heat 025 Loss of RHR System 1 4 X

Removal System and the following:

2.7 I

}~;;

AK2.03 - Service water or closed cooling iKe water pumps AK3 - Knowlledge of the reasons for the

':':::', following responses as they apply to the 026 Loss of Component Cooling X

Loss of Component Cooling Water:

3.5 I

Water 18

/.ffi';1 AK3.04 - Effect on the CCW flow header of a loss ofCCW

>.*.. AK2 - Knowlledge of the interrelations 027 Pressurizer Pressure Control

.".;:: between the F'ressurizer Pressure Control X

2.6 I

System Malfunction 13 Malfunctions and the following:

AK2.03 - COllitrollers and positioners 2.4 - Emergel1lcy Procedures 1 Plan 029 ATWS 11 X 2.4.45 - Ability to prioritize and interpret the 4.1 1

significance of each annunciator 1alarm.

ES-401 PWR Examination Outline Form ES-401-2 Emergency & Abnormal Plant Evolutions - Tier 1 / Group 1 - REACTOR OPERATOR E/APE #/Name/Safety Function 038 Steam Gen. Tube Rupture I 3 040 Steam Line Rupture - Excessive Heat Transfer I 4 055 Station Blackout 16 056 Loss of Off-site Power I 6 058 Loss of DC Power 16 062 Loss ofNucIear Svc Water I 4 065 Loss of Instrument Air! 8 077 Generator V oltage and Electric Grid Disturbances 16 CE/E06 Loss of Main Feedwater ! 4 KIA Category Totals:

KA Topic EA2 - Ability to determine or interpret the following as they apply to a SGTR:

EA2.10 - Flowpath for charging and letdown flows AKI - Knowlj~dge of the operational implications of the following concepts as they apply to Steam Line Rupture:

AKI.03 - RCS shrink and consequent depressurization EAt - Ability to operate and monitor the following as they apply to a SBO:

EA1.0 1 - In-ecre thermocouple temperatures 2.2 - Equipment Control 2.2.3 - Knowledge of the design, procedural, and operational differences between units.

AAI - Ability to operate and I or monitor the following as they apply to the Loss of DC Power:

AA 1.01 - Cross-tie of the affected de bus with the alternate supply AA2 - Ability to determine and interpret the following ns they apply to the Loss of Nuclear Service Water:

AA2.03 - The valve lineups necessary to restart the SWS while bypassing the portion of the system causing the abnormal condition AA1-Ability to operate and lor monitor the following as they apply to the Loss of Instrument Air:

AAI.05 - RPS 2.1-Conduct of Operations 2.1.23 - Ability to perfonn specific system and integrated plant procedures during all modes of plant operation.

EKI. Knowledge of the operational implications of the following concepts as they apply to the (Loss of Feedwater)

EKI.2 - Normal, abnormal and emergency operating proct~dures associated with (Loss of Feedwater)

Group Point Total:

Imp Pts 3.1 3.8 3.7 3.8 3.4 2.6 3.3 4.3 3.2 18

4.2 ES-401 PWR Examination Outline Form ES-401 Emergency & Abnormal Plant Evolutions Tier 1 / Group 2 - REACTOR OPERATOR E/APE #iName/Safety Function 00 I Continuous Rod Withdrawal/I 003 Dropped Control Rod 11 059 Accidental Liquid RadWaste ReI.

19 061 ARM System Alarms I 7 067 Plant Fire On-site 19 069 Loss ofCTMT Integrity / 5 076 High Reactor Coolant Activity 19 CE/A 16 Excess RCS Leakage / 2 CE/E09 Functional Recovery KIA Category Totals:

KA Topic Imp Pts AA2 - Ability to determine and interpret the following as they apply to the Continuous Rod Withdrawal:

AA2.04 - Reactor power and its trend AKI - Knowlj~dge of the operational implications of the following concepts as 2.9 they apply to Dropped Control Rod:

AK1.l6 - MTC AK2 - Knowledge of the interrelations between the Accidental Liquid Radwaste 2.7 Release and the following:

AK2.01 - Radioactive-liquid monitors AAI - Ability to operate and I or monitor the following as they apply to the Area Radiation Monitoring (ARM)System

3.6 Alarms

AAl.OI - Automatic actuation AK3. Knowledge of the reasons for the following responses as they apply to the Plant Fire on Site:

3.3 AK3.04 - Actions contained in EOP for plant fire on site AAI. Ability.0 operate and lor monitor the following as they apply to the Loss of Containment Integrity:

2.8 AA 1.03 - Fluid systems penetrating containment AK2' Knowledge of the interrelations between the High Reactor Coolant 2.6 Activity and the following:

AK2.01 - Process radiation monitors 2.1 Conduct of Operations 2.1.20 - Ability to interpret and execute 4.6 procedure steps.

EKI. Knowledge of the operational implications of the following concepts as they apply to the (Functional Recovery) 3.2 EK1.2 - Normal, abnormal and emergency operating procedures associated with (Functional Recovery)

Group Point Total:

9

ES-401 PWR Examination Outline Form ES-401-2 Plant Systems Tier 21 Group 1 - REACTOR OPERATOR System/Evolution #lName 003 Reactor Coolant Pump 003 Reactor Coolant Pump 004 Chemical and Volume Control 005 Residual Heat Removal 005 Residual Heat Removal 006 Emergency Core Cooling 007 Pressurizer Relief/Quench Tank 007 Pressurizer Relief/Quench Tank K K 1

2 x

KA Topic Imp Pts K6 - Knowledge of the effect of a loss or malfunction on the following will have on the RCPS:

2.6 K6.14 - Starting requirements A4 - Ability to manually operate nnd/or monitor in the Control Room:

3.1 A4.04 - RCP seal differential pressure instrumentation K6 - Knowledge of the effect of a loss or malfunction on the following CVCS components:

3.1 K6.13 - Purpose and function ofthe boration/dilution batch controller K5 - Knowledge of the operational implications of the following concepts as they apply the RHRS:

2.9 K5.03 - Reactivity effects ofRHR till water A4 - Ability to manually operate and/or monitor in the control room:

3.4 A4.02 - Heat exchanger bypass flow control K5 - Knowledge of the operational implications of the following concepts as they apply to ECCS:

K5.07 - Expected temperature levels in various locations ofthe RCS due to various plant

(:onditions 2.7 KI - Knowledge of the physical

<:onnections and/or cause/effect relationships between the PRTS and the following systems:

3.0 KI.03 - RCS A4 - Ability to manually operate

~lDd/or monitor in the control room:

2.7 A4.01 PRT spray supply valve

ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2 / Group 1 - REACTOR OPERATOR System/Evolution #lName 008 Component Cooling Water OlD Pressurizer Pressure Control 012 Reactor Protection 012 Reactor Protection 013 ESFAS 013 ESFAS 022 Containment Cooling 026 Containment Spray 026 Containment Spray 039 Main and Reheat Steam KKK K KA Topic Imp Pts 1 234 K4 - Knowledge of CCWS design feature(s) and/or interlock(s) x which provide for the following:

2.7 K4.09 - The "standby" feature for the CCW pumps K2 - Knowledge of bus power supplies to the following:

x 2.7 K2.04 - Indicator for code safety position K3 - Knowledge of the effect that a loss or malfunction of the RPS x

3.9 will have on the following:

K3.01 - CRDS 2.4 - Emergency Procedures /

Plan 2.4.31 - Knowledge ofannunciator 4.2 alarms, indications, or response procedures.

][(5 - Knowledge of the operational implications of the following concepts as they apply 2.9 to the ESFAS:

K5.02 Safety system logic and reliability A3 - Ability to monitor automatic operation of the ESF AS

3.7 including

A3.01 - Input channels and logic Al - Ability to predict and/or monitor changes in parameters (to prevent exceeding design 3.6 limits) associated with operating jhe CCS controls including:

A1.0 I - Containment temperature A3 - Ability to monitor automatic operation of the CSS, including:

A3.02 - Verification that cooling 3.9 water is supplied to the cntmt spray heat exchanger 2.1 - Conduct of Operations 2.1.28 - Knowledge of the purpose 4.1 and function of major system components and controls.

)(3 Knowledge of the effect that a loss or malfunction of the MRSS x

2.8 will have on the following:

K3.06 - SDS

ES-401 PWR Examination Outline Form ES-401-2 Plant Systems Tier 2 I Group 1 - REACTOR OPERATOR SystemlEvolution #lName 059 Main Feedwater 061 Auxiliary/Emergency Feedwater 062 AC Electrical Distribution 063 DC Electrical Distribution 064 Emergency Diesel Generator 073 Process Radiation Monitoring 073 Process Radiation Monitoring K

I K

2 K

3 K

4 K

5 KA Topic Imp Pts A2 - Ability to (a) predict the impacts of the following malfunctions or operations on the MFW; and (b) based on those predictions, use procedures to 2.7 correct, control, or mitigate the consequences of those malfunctions or operations:

A2.03 - Overfeeding event A2 - Ability to (a) predict the impacts of the following malfunctions or operations on the AFW; and (b) based on those predictions, use procedures to 3.4 eorrect, control, or mitigate the eonsequences of those malfunctions or operations:

A2.07 - Air or MOY failure K3 - Knowledge of the effect that II loss or malfunction of the ac X

distribution system will have on 3.5 the following:

K3.01 - Major system loads Al - Ability to predict and/or monitor changes in parameters associated with operating the DC electrical system controls

2.5 including

A 1.0 1 - Battery capacity as it is affected by discharge rate Kl - Knowledge of the physical

(~onnections and/or cause/effect X

.*elationships between the ED/G 3.6 system and the following systems:

K1.04 - DC distribution system 1(4 - Knowledge of PRM system design feature(s) and/or X

interlock(s) which provide for the following:

K4.01 - Release termination when radiation exceeds setpoint 2.4 - Emergency Procedures /

)Ian 2.4.49 - Ability to perform without reference to procedures those 4.6 actions that require immediate operation of system components and controls.

ES-401 PWR Examination Outline Form ES-401-2 Plant Systems Tier 2 / Group 1.. REACTOR OPERATOR System/Evolution #/Name K

I K

3 K

4 K

5 KA Topic Imp Pts 076 Service Water X

K2. Knowledge of bus power supplies to the following:

2.7 K2.0I. Service water 078 Instrument Air X

]((4 - Knowledge of lAS design lfeature(s) and/or interlock(s) which provide for the following:

3.1 K4.03. Securing of SAS upon loss of cooling water 103 Containment A2 - Ability to (a) predict the impacts of the following malfunctions or operations on the containment system and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations 3.5 A2.03 Phase A and B isolation KIA Category Totals:

2 2

3 3

P Point Total:

28

ES-401 PWR Examination Outline Form ES-40l-2 Plant Systems - Tier 2 / Group 2 - REACTOR OPERATOR System/Evolution #/Name 001 Control Rod Drive 002 Reactor Coolant 015 Nuclear instrumentation 017 In-core Temperature Monitor 027 Containment Iodine Removal 028 Hydrogen Recombiner and Purge Control 045 Main Turbine Generator 072 Area Radiation Monitoring 075 Circulating Water K

1 K

2 KA Topic Imp Pts K4 - Knowledge of CRDS design IFeature(s) and/or interlock(s) which provide for the following:

3.4 K4.23 Rod motion inhibit

]1(1 - Knowledge of the physical connections and/or cause-effect x

relationships between the RCS 3.7 and the following systems:

K1.06 - CVCS A3 - Ability to monitor automatic operation of the N IS, including:

A3.05 - Recognition of audio 2.6 output expected for a given plant condition

]1(6 - Knowledge of the effect of a loss or malfunction of the following ITM system 2.7

<<:omponents:

K6.0 1 - Sensors and detectors A4 - Ability to manually operate lind/or monitor in the control I'oom:

3.3 A4.0J - CIRS controls K2 - Knowledge of bus power x

supplies to the following:

2.5 K2.01 Hydrogen recombiners KS - Knowledge of the operational implications of the following concepts as the apply to the MT/B System:

2.5 KS.17 - Relationship between moderator temperature coefficient and boron concentration in RCS as T/G load increases Al - Ability to predict andlor monitor changes in parameters (to prevent exceeding design limits) associated with operating 3.4 the ARM system controls including:

Al.OI Radiation levels

2. t - Conduct of Operations 2.1.32 - Ability to explain and apply system limits and 3.8 precautions.

ES-401 PWR Examination Outline Fonn ES-401-2 Plant Systems - Tier 2 I Group 2-REACTOR OPERATOR System/Evolution #fName KKK 123 KA Topic Imp Pts 086 Fire Protection x

)(3 Knowledge of the effect that a loss or malfunction of the Fire Protection System will have on the following:

K3.01 - Shutdown capability with redundant equipment 2.7 KIA Category Totals:

1 Group Point Total:

10

ES-401 PWR Examination Outline Form ES-401 Emergency & Abnormal Plant Evolutions Tier 1 I Group 1 Senior Reactor Operator E/APE #/Name/Safety Function 015/017 RCP Malfunctions / 4 038 Steam Gen. Tube Rupture 1 3 054 Loss of Main Feedwater / 4 057 Loss of Vital AC lnst. Bus 16 CE/E02 Reactor Trip Stabilization - Recovery I 1

CE/E05 Steam Line Rupture - Excessive Heat Transfer / 4 KIA Category Totals:

KKK 123 KA Topic Imp Pts 2.4 - Emergency Procedures I Plan 2.4.21 - Knowledge ofthe parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

4.6 2.1 - Conduct of Operations 2.1.19 - Ability to use plant computers to evaluate system or component status.

3.8 AA2 - Ability to determine and interpret the following as they apply to the Loss of Main Feedwater (MFW):

4.2 AA2.03 - Conditions and reasons for AFW pump startup AA2 - Ability to determine and interpret the following as they apply to the Loss of Vital AC Instrument Bus:

3.8 AA2.0 1 - Safety injection tank pressure and level indicators 2.2 - Equipment Control 2.2.42 - Ability to recognize system parameters that are' entry-level conditions for Technical Specifications.

4.6 EA2 - Ability to determine and interpret the following as they apply to the (Excess Steam Demand)

EA2.1 - Facility conditions and selection of appropriate procedures during abnormal and emergency operations 4.0 oup Point Total:

6

ES-401 PWR Examination Outline Form ES-401-2 Emergency & Abnormal Plant Evolutions - Tier 1 I Group 2 - Senior Reactor Operator E/APE #/Name/Safety Function 005 Inoperable/Stuck Control Rod /1 024 Emergency Soration /1 028 Pressurizer Level Malfunction / 2 037 Steam Generator Tube Leak! 3 KIA Category Totals:

Kl KA Topic 2.4 - Emergency Procedures I Plan 2.4.9 - Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.

AA2 - Ability to determine and interpret the following as they apply to the Emergency Boration:

AA2.05 - Amount of boron to add to achieve required SDM AA2 - Ability to determine and interpret the following as they apply to the Pressurizer Level Control AA2.08 - PZR kvel as a function of 2.4 - Emergency Procedures / Plan 2.4.18 - Knowledge ofthe specific bases for EOPs.

Point Total:

Imp Pts 4.2 3.9 3.5 4.0 4

3.7 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems Tier 2 I Group 1 Senior Reactor Operator System/Evolution #lName 004 CVCS 006 ECCS 059 Main Feedwater 063 DC Electrical Distribution 103 Containment KIA Category Totals:

KA Topic Imp Pts A2 - Ability to (a) predict the ilmpacts of the following malfunctions or operations on the evcs; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.17 Low PZR pressure A2 - Ability to (a) predict the impacts of the following malfunctions or operations on the ECCS; and (b) based on those Ilredictions, use procedures to 4.4 I:orrect, control, or mitigate the I:onsequences of those malfunctions or operations:

A2.ll - Rupture ofECCS header 2.4 - Emergency Procedures I Plan 2.4.4 - Ability to recognize 4.6 abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.

2.2 - Equipment Control 2.2.42 - Ability to recognize system parameters that are entry-level 4.6 conditions for Technical Specifications.

A2 - Ability to (a) predict the impacts of the following malfunctions or operations on the containment system; and (b) based on those predictions, use 2.6 procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.0 I - Integrated leak rate test Group Point Total:

5

4.6 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems Tier 2 / Group 2 -- Senior Reactor Operator System/Evolution #lName 029 Containment Purge 034 Fuel Handling Equipment 041 Steam Dumpffurbine Bypass Control KJA Category Totals:

KA Topic 2.2 - Equipment Control 2.2.37 - Ability to determine operability and/or availability of safety related equipment.

Kl - Knowledge of the physical I:onnections and/or cause-effect relationships between the Fuel Handling System and the following systems:

Kl.04 - NIS A2 Ability to (a) predict the impacts of the following malfunctions or operations on the SDS; and (b) based on those predictions or mitigate the consequences of those malfunctions or operations:

A2.03 - Loss of lAS Group Point Total:

Imp Pts 3.5 3.1 3

ES-401 PWR Examination Outline Form ES-401-3 Tier 3 Generic Knowledge & Abilities Outline - RO & SRO Facility: Calvert Cliffs Nuclear Power Plant Category Conduct of Operations Equipment Control Radiation Control Emergency Procedures/Plan Tier 3 TotaI(s)

KJA#

Topic 2.t.3 Knowledge of shift or short-tenn relief turnover practices.

Ability to perform specific system and integrated plant procedures uring all modes of plant operation.

Ability to use procedures to detern1ine the effects on reactivity ofplant 2.1.43 changes, such as reactor coolant system temperature, secondary plant, fuel depletion, etc.

2.1.20 Ability to interpret and execute procedure steps.

2.1.35 Knowledge of the fuel-handling responsibilities of SROs.

Subtotal (multi-unit license) Ability to explain the variations in control 2.2.4 board/control room layouts, systems, instrumentation and procedural actions between units at a facility.

Ability to recognize system parameters that are entry-level conditions 2.2.42 for Technical Specifications.

Knowledge ofthe process for managing maintenance activities during 2.2.18 shutdown opcrations, such as risk assessments, work prioritization, etc.

Subtotal Knowledge ofradiation exposure limits under normal or emergency 2.3.4 conditions.

Ability to use radiation monitoring systems, such as fixed radiation 2.3.5 monitors and alarms, portable survey instruments, personal monitoring equipment, etc.

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

2.3.6 Ability to approve release permits.

2.3.11 Ability to control radiation releases.

Subtotal Knowledge ofthe organization ofthe operating procedures network for 2.4.5 normal, abnormal, and emergency evolutions.

2.4.29 Knowledge ofthe emergency plan, 2.4.1 Knowledge of EOP entry conditions and immediate action steps.

2.4.11 Knowledge ofabnormal condition procedures.

Subtotal

ES-401 Tier / Group RO - III RO - III RO - 112 RO - 2/1 RO - 211 RO - 2/2 SRO - III SRO 1/1 SRO 211 SRO 2/2 Randomly Selected KIA 056 Loss of Off-site Pwr, KIA 2.2.2 077 Generator Voltage &

Grid Disturbances, KiA 2.1.26 CE/AI6 Excess RCS Lkg, KIA 2.1.26 012 Reactor Protection, KIA 2.4.42 026 Containment Spray, KiA 2.1.21 075 Circulating Water, KIA 2.1.1 015/017 RCP Malfs, KIA 2.4.5 038 SGTR KiA 2.3.6 059 Main Feedwater, KIA 2.4.44 029 Containment Purge, KIA 2.2.18 Record of Rejected KJAs Fonn ES-40 1-4 Reason for Rejection ES-401 contains guidance, in the form ofa list, on generic KlAs for use with Tiers I & 2. The randomly selected KIA is not on the ES-401 list.

Replaced with KIA 2.2.3, which was randomly drawn, using numbered poker chips.

ES-40I contains guidance, in the form ofa list, on generic KIAs for use with Tiers 1 & 2. The randomly selected KiA is not on the ES-401 list.

Replaced with KIA 2.1.23, which was randomly drawn, using numbered poker chips.

ES-401 contains guidance, in the form ofa list, on generic KiAs for use with Tiers 1 & 2. The randomly selected KiA is not on the ES-40l list.

Replaced with KIA 2.1.20, which was randomly drawn, using numbered poker chips.

ES-40 I contains guidance, in the form of a list, on generic KIAs for use with Tiers I & 2. The randomly selected KiA is not on the ES-401 list.

Replaced with KIA 2.4.31, which was randomly drawn, using

  • numbered poker chips.

ES-401 contains guidance, in the form ofa Jist, on generic KiAs for use with Tiers I & 2. The randomly selected KiA is not on the ES-40l list.

Replaced with KIA 2.1.28, which was randomly drawn, using numbered poker chips.

ES-401 contains guidance, in the form ofa list, on generic KiAs for use with Tiers 1 & 2. The randomly selected KIA is not on the ES-40 1 list.

Replaced with KIA 2.1.32, which was randomly drawn, using numbered poker chips.

ES-401 contains guidance, in tht: form ofa list, on generic KiAs for use with Tiers 1 & 2. The randomly selected KiA is not on the ES-40 I list.

Replaced with KIA 2.4.21, whiich was randomly drawn, using numbered poker chips.

ES-40] contains guidance, in the form of a list, on generic KIAs for use with Tiers I & 2. The randomly selected KiA is not on the ES-401 list.

Replaced with KIA 2.1.19, which was randomly drawn, using numbered poker chips.

ES-401 contains guidance, in the form ofa list, on generic KIAs for use with Tiers I & 2. The randomly selected KiA is not on the ES-401 list.

Replaced with KIA 2.4.4, which was randomly drawn, using numbered poker chips.

ES-40 1 contains gu idance, in the: form of a I ist, on generic KIAs for use with Tiers 1 & 2. The randomly selected KIA is not on the ES-40 1 list.

Replaced with KIA 2.2.37, which was randomly drawn, using numbered poker chips.

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Calvert Cliffs Nuclear Power Plant Date of Examination: 8/3/12 thru 8/2311 2 Exam Level: RO I SRO-I / SRO-U Operating Test #: 2012 Type Administrative Topic (see Note)

Describe activity to be performed Code*

Estimate Time to Boiling and Core Uncovery (RO-Admin-l )

Conduct of Operations R,M 2.1.20 - Ability to interpret and execute procedure steps (4.6,4.6).

Calculate BAST volume required to raise RWT to refueling boron concentration Conduct ofOperations R,N 2.1.25 - Ability to interpret reference materials, such as graphs, curves, tables, etc. (3.9,4.2)

Equipment Control Determine protective clothing and limits associated with Radiation Control M,R performance of a task in the RCA 2.3.7 (3.5, 3.6)

Perform an Independent Assessment of an Event Using the EOP-O Diagnostic Flowchart and Recommend the Correct Recovery Procedure.

(RO-ADMIN-4)

Emergency Procedures / Plan D, S 2.4.21 - Knowledge ofthe parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc. (4.0,4.6)

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; :S 4 for SROs & RO retakes)

(N)ew or (M)odified from bank (2: I)

(P)revious 2 exams I; randomly selected)

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

Date of Examination: 8/3/12 thru 8/23/12 Operating Test #: 2012 I

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

Safety System 1JPM Title Type Code*

Function

a.

SIM-I, Respond to a Loss of SDC with the RCS Open A,

P,S 4 (primary)

b.

SIM-2, Respond to a Pressurizer Spray Valve Failure A,P,S 3

c.

SIM-3, Respond to a FRV or FRV controller malfunction D,S 4 (secondary)

d.

SIM-4, Verify Recirculation Actuation Signal A,EN, P,S 2

e.

SIM-5, Verify the Containment Environment Safety Function is satisfied.

A, 5

f.

SIM-6, Null NI Pots to DeltaT Pots D,S 7

g.

SIM-7, Respond to Inadvertent Dilution During Reactor Startup D, L, S 1

h.

SIM-8, Unload and shutdown the OC DG D,S 6

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

a.

PL T -1, Obtain Safe Shutdown Equipment then Strip MCC-114R D,E,L,R 6

b.

PLT-2, Isolate Di Water And Condensate Makeup To The Service Water E,L,N 8

And Component Cooling Head Tanks

~T ~ Align AFW Pump Speed Control to I C43 A,E,L,M 4 (secondary)

All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

  • Type Codes Criteria for RO / SRO-I / SRO-U (A)ltemate path 4-6/4-6 /2-3 (C)ontrol room (D)irect from bank

~9/~8/~4 (E)mergency or abnormal in-plant 2':112'1/2:1 (EN)gineered safety feature

/ - I 2:1 (control room system)

(L)ow-Power / Shutdown 2':112'1/2'1 (N)ew or (M)odified from bank including I(A) 2':2/2:2/2'1 (P)revious 2 exams

~ 3/ :::: 3/:::: 2 (randomly selected)

(R)CA 2':112::1/2'1 (S)imulator

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Calvert Cliffs Nuclear Power Plant Date of Examination: 8/3112 thru 8/23/12 Exam Level: RO / SRO-I / SRO-U Operating Test #: 2012 Administrative Topic Type Describe activity to be performed (see Note)

Code*

Determine Status of Safety Functions for the FRP Conduct ofOperations N,S 2.1.20 - Ability to interpret and execute procedure steps (4.6, 4.6)

Evaluate the need for plant cooldown.

Conduct of Operations R,M 2.1.25 - Ability to interpret reference materials, such as graphs, curves, tables, etc. (3.9,4.2).

Equipment Control D,R Ability to apply T.S.s for a system Determine protective clothing requirements and dose limits associated with performance ofa task in the RCA Radiation Control M, R 2.3.7 Ability to comply with radiation work permit requirements during normal or abnormal conditions (3.5,3.6).

Determine the appropriate emergency response actions per the ERPIP while Emergency Procedures I maintaining an overview of plant conditions. (SRO-Admin-5)

D,R Plan 2.4.41 Knowledge of the emergency action level thresholds and classifications (2.9, 4.6).

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

  • Type Codes & Criteria:

(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (s 3 for ROs; s 4 for SROs & RO retakes)

(N)ew or (M)odified from bank

1)

(P)revious 2 exams I; randomly selected)

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

~

~.

'J t

Date ofExamination: 8/3/12 thru 8/23/12 Exam Level: RO / SRO-I / SRO-U Operating Test #: 2012 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including] ESF)

Safety System 1JPM Title Type Code*

Function

a.

SIM-3, Respond to a FRV or FRV controller malfunction D,S 4 (secondary)

b.

SIM-4, VerifY Recirculation Actuation Signal A, EN, L, P, S 2

c.

SIM-5, VerifY the Containment Environment Safety Function is satisfied.

A,M,S 5

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

a.

PLT-I, Obtain Safe Shutdown Equipment then Strip MCC-114R D,E,L,R 6

b.

PLT-2, Isolate Di Water And Condensate Makeup To The Service Water E,L,N 8

And Component Cooling Head Tanks All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; ailS SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

  • Type Codes Criteria for RO / SRO-I 1SRO-U (A)ltemate path 4-6 14-6 I 2-3 (C)ontrol room (D)irect from bank
9/
:;8/::;4 (E)mergency or abnormal in-plant
1/:::11;:::1 (EN)gineered safety feature 1 ;:::1 (control room system)

(L)ow-Power 1Shutdown

11:::11:::1 (N)ew or (M)odified from bank including I (A)
21 21
::1 (P)revious 2 exams
3 1
:; 3 / ::; 2 (randomly selected)

(R)CA 11:::11;:::1 (S)imulator i

i ES-301 Administrative Topics Outline Form ES-301-1 Facility: Calvert Cliffs Nuclear Power Plant Date of Examination: 8/3/12 thru 8123II2 Exam Level: RO 1SRO*I I SRO-U Operating Test #: 2012 Type Administrative Topic (see Note)

Describe activity to be performed Code*

Determine Status of Safety Functions for the FRP Conduct of Operations N,S 2.1.20 - Ability to interpret and execute procedure steps (4.6, 4.6)

Evaluate the need for plant cooldown.

Conduct of Operations R,M 2.1.25 - Ability to interpret reference materials, such as graphs, curves, tables, etc. (3.9,42).

D,R

'l_ i I Equipment Control Ability to apply T.S.s for a system Determine protective clothing requirements and dose limits associated with performance ofa task in the RCA Radiation Control M,R 2.3.7 Ability to comply with radiation work permit requirements during nonnal or abnormal conditions (3.5, 3.6).

Determine the appropriate emergency response actions per the ERPIP while maintaining an overview of plant conditions.

Emergency Procedures I Plan D,R (SRO-Admin-5) 2.4.41 - Knowledge of the emergency action level thresholds and classifications (2.9, 4.6).

~

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

  • Type Codes & Criteria:

(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (:S 3 for ROs; :s 4 for SROs & RO retakes)

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

(P)revious 2 exams (:s I ; randomly selected)

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

. ~r\\/l ') th.... 12, Facility: Calvert Cliffs Nuclear Power Plant Date ofT"'

Exam Level: RO / SRO-I I SRO-U Operating Test #: 2012 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

Safety System / JPM Title Type Code*

Function

a.

SIM-l, Respond to a Loss of SDC with the RCS Open A,L,P,S 4 (primary)

b. SIM-2, Respond to a Pressurizer Spray Valve Failure A,P,S 3
c. SIM-3, Respond to a FRV or FRV controller malfunction D,S 4 (secondary)
d. SIM-4, VerifY Recirculation Actuation Signal A, EN, L,P, S 2
e. SIM-5, VerifY the Containment Environment Safety Function is satisfied.

A,M,S 5

f.

SIM-6, Null NI Pots to DeltaT Pots D,S 7

SIM-8, Unload and shutdown the OC DO D,S 6

  • g.

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

a.

PLT-I, Obtain Safe Shutdown Equipment then Strip MCC-114 R D,E,L,R 6

b. PLT-2, Isolate Di Water And Condensate Makeup To The Service Water E,L,N 8

And Component Cooling Head Tanks

c. PLT-3, Align AFW Pump Speed Control to lC43 L,M 4(se~1 All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
  • Type Codes Criteria for RO I SRO-I I SRO-U (A)lternate path 4-6/4..6/2-3 (C)ontrol room (D)irect from bank s9/s8/s4 (E)mergency or abnormal in-plant 2:112:112:1 (EN)gineered safety feature

- / - / 2:1 (control room system)

(L)ow-Power / Shutdown 2:112:112:1 (N)ew or (M)odified from bank including leA) 2:2/2:2/2:1 (P)revious 2 exams S 3/:S 3/ S 2 (randomly selected)

(R)CA 2:1/2:]/2:1 (S)imulator

NRC EXAM PLANT JPM DESCRIPTIONS JPM#

Brief description of JPM In-Plant ~JPMs Obtain Safe Shutdown Equipment Then Strip MCC 114R (RCA)

The operator will perform the actions required to prepare key Plant-l safety related MCC loads for being powered from the OC DG.

Critical tasks are:

  • Place ALL 480V breakers on MCC 114R to OFF.
  • Place selected 480V breakers on MCC 114R to ON.

Isolate DI Water and Condensate MfU to the SRW and CCW Head Tanks The operator will perform the actions required to align the Fire Main to supply makeup water to the Service Water and Component Cooling head tanks.

Critical tasks are:

Plant-2

  • Isolates Condensate and DI Water supplies to the Head Tanks.
  • Installs hose between Fire Main and Condensate Header
  • Opens isolation valves

Align AFW Pump Speed Control to 1 C43 (Alt Path)

The operator will perform support actions required to align AFW Pp control to I C43.

Critical tasks are:

Plant-3

  • Locally aligns hand valves to shift control to 1 C43
  • Resets 11 AFW Pp Throttle/Stop Valve SROI SROU I RO X

X X

X X

X X

X N/A

I NRC EXAM SIMULATOR JPMDESCRIPTIONS JPM#

Sim-l Sim-2 Sim-3 SimA Brief description of.JPM Simulator JPMs Respond to a Loss Of Shutdown Cooling with the Reactor Coolant System Open (Alt Path)

Critical tasks are:

  • Respond to indication ofoperating Low Pressure Safety Injection (LPSI) Pump LPSI cavitation by securmg pump
  • Restores suction flowpath (opens SI -651)
  • Restores Shutdown Cooling flowrate to desired value Respond to a Pressurizer Spray Valve Failure (Alt Path)

Critical tasks are:

  • Attempt to shut the failed open I'zr Spray Vlv
  • Trip the Reactor
  • Perform post-trip immediate actions for reactivity control - Secure the RCP associated with the failed open Pzr Spray Vlv Respond to a FRV or FRV controller malfunction Critical tasks are:
  • Performs actions required to shift positioners for the MFRV Verify the Recirculation Actuation Signal (Alt Path)

...~-

Critical tasks are:

  • Secures 12 Low Pressure Safety Injection Pump
  • Throttles High Pressure Safety Injection flow to achieve 250 GPM through each ofthe four headers RO ISROI SROU

-V

-V N/A

-V

-V N/A I

-V

-V

-V i

I

-V

-V

-V

NRC EXAM SIMULATOR JPM DESCRIPTIONS JPM#

Brief description of.JPM RO SROI SROU Simulator JPMs (continued)

Sim-5 Verify the Containment Environment Safety Function is satisfied. (Alt Path)

Critical tasks are:

\\ '

  • Verify CIS

,j

,j

,j

  • Align IRUs I

Null NI Pots to Delta T Pots Critical tasks are:

Sim-6

  • Bypasses RPS Trip Units 1,2, 7, 8 & 10 per 01-30
  • Adjusts NI Power on RPS Channel "A" per 01-30
  • Unbypasses RPS Trip Units 1,2, 7, 8 & 10 per 01-30

,j

,j N/A Respond to an Inadvertent Dilution During Reactor Startup and Restore Required Shutdown Margin Critical tasks are:

Sim-7

  • Opens l-CVC-520 to bypass the CVCS Ion Exchangers
  • Starts all available Charging Pumps

,j N/A N/A Perform a normal shutdown, from the Control Room, of the OCDG.

Critical tasks are:

Sim-8 I

  • Open the OC DG 4KV Bus feeder breaker
  • Open the 07 4KV BUS TIE breaker
  • Adjust volts/frequency

\\

  • Open OC DG Output breaker
  • Secure OC DG

,j

,j N/A Totals 8

7 3

NRC EXAM RO ADMIN JPM DESCRIPTIONS JPM#

RO-Admin-l RO-Admin-2 RO-Admin-3 RO-Admin-4 Brief description of J:PM RO RO ADMIN JPMs Respond to a complete loss of SDC (Estimate Time to Boiling & Core Uncovery)

Critical tasks are:

  • Correctly determines Time-to-Boil using multipliers for days

.J shutdown and post-refueling multiplier

  • Correctly determines Time-*to-Core Uneovery using multipliers for days shutdown and post-refueling multiplier Calculate BAST volume required to raise R WT to refueling boron concentration Critical tasks are:
  • Correctly determines Refueling Boron Concentration
  • Calculates ilB
  • Determines required volume of Boric Acid solution Determine protective clothing requirements and dose limits associated with performance of a task in the RCA Critical tasks are:
  • Determines required protective clothing
  • Determines area hotspot
  • Calculates expected dose based on assignment
  • Identifies dose rate alarm setpoint
  • Identifies low dose waiting area Use the Diagnostic Flowchart in EOP-O and Select the Appropriate Recovery Procedure Critical tasks are:
  • Determines RCS Pressure and Inventory Safety Function not met
  • Determines Core and RCS Heat Removal Safety Function not met
  • Recommends implementation of EOP-4, Excess Steam Demand Event I

NRC EXAM SRO ADMIN JPM DESC~RIPTIONS J..

Brief description of JPM

~KUl SRO ADMIN JPMs Determine Status of Safety Functions for the FRP Critical tasks are:

SRO-Admin-l X

X

  • Correctly determines SUCCf:SS Path, status (met/not met) and priority for each of the six safety functions Evaluate the need for plant cooldown.

Critical tasks are:

SRO-Admin-2

  • Determines there is NOT adequate inventory to X

X perform a cooldoWll.

  • Determines how long hot standby can be maintained with the current inventory Apply Technical Specifications to a plant condition Critical tasks are:

SRO-Admin-3 X

X

  • Determines applicable LCO(s)
  • Determines required actions
  • Determines Completion time Determine protective clothing requirements and dose limits associated with performance of a task in the RCA Critical tasks are:

SRO-Admin-4

  • Determines required protective clothing X

X

  • Determines area hotspot
  • Calculates expected dose based on assignment
  • Identifies dose rate alarm setpoint
  • Identifies low dose waiting area Determine Appropriate Emergency Response Actions Critical tasks are:

'\\.

SRO-Admin-5

  • Evaluates conditions against ERPIP Attachment 1, X

X Emergency Action Level (EAL) criteria

  • Makes correct Protective Action Recommendation
  • Correctly completes Attachment 3, within 15 minutes

Appendix D Scenario Outline Form ES-D-1 Facility: Calvert Cliffs Nuclear Power Plant Scenario #:1 OP-Test #: CCNPP 2012 Examiners:

Operators:

Initial Conditions: Unit-l is at 65% Power with a core burnup of 10,885 MWDIMTU. Unit-2 is in Mode 5. 12 CS pump is OOS for repairs. 23 Auxiliary Feedwater Pump is out of service for overhaul of the coupling (expected back in three shifts).

Turnover: Place 12 SGFP in parallel operation with 11 SGFP and return power to 100%.

Event Malfunction #

Event Type*

Event Description I

None N - BOP/SRO Aligns 12 SGFP for parallel operation will SGFP I-ATC 2

rcs026 01 Failure ofLT-IIOX T-SRO R-ATC 3

None Commences power escalation N - BOP/SRO C - BOP/SRO 4

srw003 01 11 Service Water Pump Bkr Failure T-SRO 5

cd005 01 C - BOP/SRO Condensate Booster Pump trip Condensate Booster Pump discharge header rupture

cd008, (25%). Reactor will not trip automatically or with 6

M-ALL rps005, rps006 Rx Trip pushbuttons. CEDM MG Sets must be de-energized to trip the reactor.

7 Various C-ALL Loss of All Feedwater / Once-Thru-Cooling (N)ormal (R)eactivity (I)nstrument

( C)omponent (M)ajor (T)ech Spec Iminut~>of anc}):isting Re~ctor Trip Condition.

(report not critical). N/A if RPS setpoint not reached.

Trip aU RCPs priOl":Jo commenCing RCS Cooldown.

3. Initiates OTCC when both S/G levels are <-350" or TCOLD rises > 5°F uncontrollably (must be commenced prior to CETtemperatures reaching 560°F)

AppendixD Scenario Outline Form ES-D-1 Facility: Calvert Cliffs Nuclear Power Plant Scenario #: 2 OP-Test #: CCNPP 2012 Examiners:

Operators:

Initial Conditions: Unit-1 is at 100% power. Con~ Burnup is 17,885 MWDIMTU. Unit-2 is in Mode 5.

Turnover: 12 AFW Pump is OOS due to a governor problem (has been OOS for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and is due back by end of shift). 14 CAR is OOS due to a failed motor bearing. The 2A DG was removed from service yesterday for scheduled maintenance. 11 ADV is wisping a small amount of steam. Instructions to the crew are to maintain full power.

,.,.-. H"_

Eve Event Type*

Event Description C - ATC/SRO I

1 cvcs003 01 11 Charging Pump coupling failure T-SRO 2

rcs023 01 I -ATC PT-I00X fails high C - BOP/SRO 3

swOO} 02 Saltwater Leak T-SRO R-ATC 4

tg004_01 MTCV-l Fails shut C - BOP/SRO 5

swyd002 M-ALL Loss Of Offsite Power resulting in a reactor trip 6

dgOO2_02 C -ALL No 4KV SR Busses resulting in Station Blackout afw004 01 7

C-ALL AF AS "A" & "B" failure afw004-02 (N)onnal (R)eactivity (I)nstrument (C)omponent (M)ajor (T)ech Spec by securingl~$;~> *.~por by isolating tripping th¢SRWp' *~ipigh rOom level.

3.

Ke~ltorc:!s po~ to 11 or 14 4KV bus prior to 125V DC bus voltage going below 106V.

Appendix D Scenario Outline Form ES-D-l Facility: Calvert Cliffs Nuclear Power Plant Scenario #:3 OP-Test #: CCNPP 2012 Examiners:

Operators:

Initial Conditions: Unit-1 is at 100% with a core burnup of 10,885 MWDIMTU. Unit-2 is in Mode 5.

Turnover: PORV-402 is isolated due to leakby and 11 4 KV Bus alternate feed is tagged out for breaker PMs. The crew is instructed to maintain 100% power.

Event #

Malfunction #

Event Type*

Event Description C - BOP/SRO 1

cntmOOI 01 11 Containment Air Cooler (CAC) trips T-SRO 2

ms015 I -ALL ADV Controller, I-HIC-4056, fails in automatic

'\\{

R - ATC \\

msOOI 02 11 S/G Tube leak (0 - 65 GPM over 2 minutes) 3 C -NOP/SRO Downpower Reduce TAvE to less than 53TF T-SRO I

4 12 02 M-ALL 11 S/G Tube rupture - one tube Gf C - BOP/SRO 11 AFW Pp trip I *

(N)ormal (R)eactivity (I)nstrument (C)omponent (M)<tior (T)ech Spec

Appendix D Scenario Outline Form ES-D-l Facility: Calvert Cliffs Nuclear Power Plant Scenario #:4 OP-Test #: CCNPP 2012 Examiners:

Operators:

Initial Conditions: 75% MOC 10,885 MWDIMTU. Power was reduced from 100% 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> ago by ESOD request and for Main Turbine valve testing Turnover: 12A Waterbox is being cleaned. 11 Condensate Pump tagged out for oil change out. 23 Aux Feed Pump is out of service for overhaul of the coupling (expected back in three shifts). Instructions for the shift are to maintain power @ 75%.

Event #

Malfun

'T'ype*

Event Description 1-BOP/SRO 1

rcs024 01 RCS Pressure transmitter PT -1 02A fails high T-SRO 2

480vOOl 04 C - BOP/SRO 12B 480V bus failure cw002 01 thru 3

C - BOP/SRO Intake Structure Malfunction cw002 06 R-ATC Rapid Power Reduction due to securing 11 CW Pp 4

Downpower N - BOP/SRO with 13 Circ Water Pump already secured Manual Reactor Trip due to securing 12 CW Pp.

msOI6 09 5

M-ALL Stuck open S/G Safety Valve on Reactor trip, 350 rcs002 GPM RCS leak ramps in over 4 minutes esfaOOl 01 6

C-ALL SIAS "A" & "B" failure esfaOOl 02 (N)ormal (R)eactivity (I)nstrurnent (C)ornponent (M)ajor (T)ech Spec r~

prio~

~xit.

2.

RCS'pressllre falls B'tHow 172S PSIA prior to

3. Establishes RCStemperature control
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