calvert cliffs - draft - outlines (folder 2). · 6. select sro topics for tiers i and 2 from the...

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 of the 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 '~~iE/APE #/Name/Safety Function KA Topic Imp Pts1 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 J13 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 ISystem 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


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

2.9they apply to Dropped Control Rod:

AK1.l6 - MTC

AK2 - Knowledge of the interrelations between the Accidental Liquid Radwaste

2.7Release 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

2.6Activity 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


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 of the 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


Plant Systems - Tier 2 / Group 1 - REACTOR OPERATOR


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 3.9will have on the following:

K3.01 - CRDS

2.4 - Emergency Procedures / Plan

2.4.31 - Knowledge of annunciator 4.2 alarms, indications, or response procedures.

][(5 - Knowledge of the operational implications of the following concepts as they apply

2.9to the ESFAS:

K5.02 Safety system logic and reliability

A3 - Ability to monitor automatic operation of the ESF AS

3.7including:

A3.01 - Input channels and logic

Al - Ability to predict and/or monitor changes in parameters (to prevent exceeding design

3.6limits) 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 2.8will have on the following:

K3.06 - SDS


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.


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


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


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


2.2.42 - 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


Emergency & Abnormal Plant Evolutions - Tier 1 I Group 2 - Senior Reactor Operator


005 Inoperable/Stuck Control Rod /1

024 Emergency Soration /1

028 Pressurizer Level Malfunction / 2

037 Steam Generator Tube Leak! 3


Kl KA Topic


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 of the specific bases for EOPs.

Point Total:

Imp Pts

4.2

3.9

3.5

4.0

4

3.7


Plant Systems Tier 2 I Group 1 Senior Reactor Operator


004 CVCS

006 ECCS

059 Main Feedwater

063 DC Electrical Distribution

103 Containment


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 of ECCS header


2.4.4 - Ability to recognize 4.6abnormal indications for system

operating parameters that are entry-level conditions for emergency and abnormal operating procedures.


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


Plant Systems Tier 2 / Group 2 -- Senior Reactor Operator


029 Containment Purge

034 Fuel Handling Equipment

041 Steam Dumpffurbine Bypass Control

KJA Category Totals:

KA Topic


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


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 of plant 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 of the process for managing maintenance activities during 2.2.18

shutdown opcrations, such as risk assessments, work prioritization, etc.

Subtotal

Knowledge of radiation exposure limits under normal or emergency2.3.4

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 of the organization of the operating procedures network for 2.4.5

normal, abnormal, and emergency evolutions.

2.4.29 Knowledge of the emergency plan,

2.4.1 Knowledge of EOP entry conditions and immediate action steps.

2.4.11 Knowledge of abnormal 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-40 I contains guidance, in the form of a 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 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 2.4.4, 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

TypeAdministrative 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 of the 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 #: 2012I ~U

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

SafetySystem 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 8And 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


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,RPlan 2.4.41 Knowledge of the emergency action level thresholds and

classifications (2.9, 4.6).



(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)


.... ~ ~.

'J t Date of Examination: 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


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 ;:::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


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

TypeAdministrative 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).

~



(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)


. ~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)

SafetySystem / 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.


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 8And 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

• Communicates with 2C43 (Safe Shutdown Panel) '-"

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

• Disables electrical inputs to the AFW Pp Trip Solenoid valves

• Resets 11 AFW Pp Throttle/Stop Valve

SROI SROUI 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)

• Aligns FIC-306 for restoration of Shutdown Cooling flow

• Starts a LPSI Pp

• 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 of the 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

• Verify CSAS

• 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

• Opens Boric Acid Direct Makeup valve l-CVC-514

• Starts all available Boric Acid Pumps

• 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-ATC2 rcs026 01 Failure ofLT-IIOXT-SRO

R-ATC3 None Commences power escalation N - BOP/SRO

C - BOP/SRO4 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-ALLrps005, 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


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 hours 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-ATC4 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 017 C-ALL AF AS "A" & "B" failureafw004-02

* (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



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/SRO1 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/SRODownpower Reduce TAvE to less than 53TFT-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



Initial Conditions: 75% MOC 10,885 MWDIMTU. Power was reduced from 100% 3 hours 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/SRO1 rcs024 01 RCS Pressure transmitter PT -1 02A fails high T-SRO

2 480vOOl 04 C - BOP/SRO 12B 480V bus failure

cw002 01 thru3 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 016 C-ALL SIAS "A" & "B" failureesfaOOl 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

calvert cliffs - draft - outlines (folder 2). · 6. select sro topics for tiers i and 2 from the...

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