ML16188A234

Draft - Outlines (Folder 2)

ML16188A234
Person / Time
Site:	Calvert Cliffs
Issue date:	03/11/2016
From:	Jaeger J Exelon Generation Co
To:	David Silk Operations Branch I
Shared Package
ML15308A607	List: ML16012A008 ML16155A226 ML16188A232 ML16188A233 ML16188A234 ML16188A235 ML16188A237 ML16188A251 ML16188A279 ML16188A304 ML16229A510 ML16229A511
References
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Download: ML16188A234 (31)

Text

ES-401 PWR Examination Outline Form ES-401-2 I Facility: CALVERT CLIFFS NUCLEAR POWER PLANT Date of Exam: 06/13/2016 I RO Category Kl A Points SRO Only Points I Group Tier K

K K

K K

K A

A A

A G

Total A2 G

Total I

2 3

4 5

6 I

2 3

4

1. Emergency &

I 3

3 3

3 3

3 18 3

3 6

Abnormal Plant 2

I 2

I NIA 2

2 NIA I

9 2

2 4

Evolutions Tier Totals 4

5 4

5 5

4 27 5

5 10 I

3 2

3 3

2 2

3 3

2 2

3 28 3

2 5

2. Plant Systems 2

0 I

I I

I I

I I

I I

I 10 0

2 I

3 Tier Totals 3

3 4

4 3

3 4

4 3

3 4

38 5

3 8

I 2

3 4

I 2

3 4

3. Generic Knowledge & Abilities Categories 10 7

2 3

2 3

2 I

2 2

Note:

I. 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 Kl A category shall not be less than two).

2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by+/- 1 from that specified in the table based on NRC revisions.

The final RO exam must total 75 points and the SRO-only exam must total 25 points.

3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate Kl A statements.

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

5. Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/ A categories.

7. *The generic (G) Kl As in Tiers 1 and 2 shall be selected from Section 2 of the K/ A Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D. I.b of ES-401 for the applicable K/As.

8. On the following pages, enter the K/ A numbers, a brief description of each topic. the topics' importance ratings

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

9.

For Tier 3, select topics from Section 2 of the K/ A catalog, and enter the K/ A numbers, descriptions, IRs, and point totals(#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.

ES-401 PWR Examination Outline Form ES-401-2 Emergency & Abnormal Plant Evolutions - Tier 1 I Group 1 - REACTOR OPERA TOR E/APE #/Name/Safety Function K

K K

A A

G KA Topic Imp Pts 1

2 3

1 2

007 Rx Trip - Stabilization - Recovery EA2 - Ability to determine or interpret the x

following as they apply to a reactor trip:

4.3 1

(I)

EA2.06 - Occurrence of a reactor trip 2.1-Conduct of Operations 008 Pressurizer Vapor Space Accident x

2.1.28 - Knowledge of the purpose and 4.1 1

(3) function of major system components and controls.

EKl - Knowledge of the operational 009 Small Break LOCA implications of the following concepts as (3) x they apply to the small break LOCA:

4.2 1

EK 1.01 - Natural circulation and cooling, including reflux boiling AK2 - Knowledge of the interrelations 0000151000017 RCP Malfunctions between the Reactor Coolant Pump x

Malfunctions (Loss of RC Flow) and the 2.9 I

(4) following:

AK2.07 - RCP seals AK3 - Knowledge of the reasons for the 022 Loss of Rx Coolant Makeup x

following responses as they apply to the 3.2 I

(2)

Loss of Reactor Coolant Makeup:

AK3.04 - Isolating letdown AK2 - Knowledge of the interrelations 025 Loss of RHR System between the Loss of Residual Heat (4) x Removal System and the following:

2.6 1

AK2.05 - Reactor building sump AAI - Ability to operate and I or monitor the following as they apply to the Loss of 026 Loss of Component Cooling Water x

Component Cooling Water:

2.9 1

(8)

AA I.07 - Flow rates to the components and systems that are serviced by the CCWS; interactions among the components AK2 - Knowledge of the interrelations 027 Pzr Press Control Sys Malfunction between the Pressurizer Pressure Control (3) x Malfunctions and the following:

2.6 I

AK2.03 - Controllers and positioners 029 ATWS EA2-Ability to determine or interpret the (I) x following as they apply to a ATWS:

4.4 I

EA2.01 - Reactor nuclear instrumentation EK I - Knowledge of the operational 038 Steam Gen. Tube Rupture x

implications of the following concepts as 3.1 I

(3) they apply to the SGTR:

EK 1.0 I - Use of steam tables

ES-401 PWR Examination Outline Form ES-401-2 Emergency & Abnormal Plant Evolutions - Tier 1 I Group 1 - REACTOR OPERA TOR E/APE #/Name/Safety Function K

K K

A A

G KA Topic Imp Pts 1

2 3

1 2

AK3 - Knowledge of the reasons for the 000054 Loss of Main Feedwater following responses as they apply to the x

Loss of Main Feedwater:

3.4*

I (4)

AK3.02 - Matching offeedwater and steam flows 055 Station Blackout EA2 - Ability to determine or interpret the (6) x following as they apply to a SBO:

3.9 I

EA2.03 - Actions necessary to restore power 2.4 - Emergency Procedures I Plan 057 Loss of Yitai AC Instrument Bus 2.4.49 - Ability to perform without reference (6) x to procedures those actions that require 4.6 I

immediate operation of system components and controls.

2.4 - Emergency Procedures I Plan 058 Loss of DC Power 2.4.49 - Ability to perform without reference (6) x to procedures those actions that require 4.6 I

immediate operation of system components and controls.

AAl - Ability to operate and/or monitor 062 Loss of Nuclear Service Water the following as they apply to the Loss of (4) x Nuclear Service Water):

3.2 I

AA 1.02 - Loads on the S WS in the control room AA 1-Ability to operate and I or monitor 065 Loss of Instrument Air x

the following as they apply to the Loss of 3.5*

I (8)

Instrument Air:

AA 1.04-Emergency Air Compressor AKI - Knowledge of the operational 077 Generator Voltage and Electric Grid implications of the following concepts as x

they apply to Generator Voltage and 3.3 I

Disturbances (6)

Electric Grid Disturbances:

AK 1.02 - Over-excitation EK3 - Knowledge of the reasons for the following responses as they apply to the CE/E05 Excess Steam Demand (4) x (Excess Steam Demand) 3.8 I

EK3.3 - Manipulation of controls required to obtain desired operating results during abnormal, and emergency situations I

KIA Category Totals:

3 3

3 3

3 3

Group Point Total:

18 I

ES-401 PWR Examination Outline Form ES-401-2 Emergency & Abnormal Plant Evolutions - Tier 1 I Group 2 - REACTOR OPERA TOR E/APE #/Name/Safety Function K

K K

A A

G KA Topic Imp Pts 1

2 3

1 2

AK2 - Knowledge of the interrelations 003 Dropped Control Rod between the Dropped Control Rod and the x

following:

2.5 1

( 1)

AK2.05 - Control rod drive power supplies and logic circuits AKI - Knowledge of the operational 028 Pressurizer Level Malfunction implications of the following concepts as x

they apply to Pressurizer Level Control 2.8 1

(2)

Malfunctions:

AK 1.01 - PZR reference leak abnormalities AA2 - Ability to determine and interpret the following as they apply to an 060 Accidental Gaseous Radwaste Rel.

Accidental Gaseous Radwaste release:

(9) x 3.1 1

AA2.02 - The possible location of a radioactive-gas leak, with the assistance of PEO, health physics and chemistry personnel AAl - Ability to operate and I or monitor 061 ARM System Alarms x

the following as they apply to Area 3.6 1

(7)

Radiation Monitoring System Alarms:

AA 1.01 - Automatic actuation 068 Control Room Evacuation 2.1-Conduct of Operations (8) x 2.1.20 - Ability to interpret and execute 4.6 1

procedure steps.

EK2 - Knowledge of the interrelations 074 Inadequate Core Cooling x

between Inadequate Core Cooling and the 3.9 I

(4) following:

EK2.04 - HPI pumps AK3 - Knowledge of the reasons for the following responses as they apply to RCS CE/Al I RCS Overcooling x

Overcooling:

2.9 1

(4)

AK3.2 - Normal, abnormal and emergency operating procedures associated with (RCS Overcool ing)

AAl - Ability to operate and I or monitor the following as they apply to Natural CE/ A 13 Natural Circulation Circulation Operations:

(4) x AA 1.1 - Components, and functions of 3.3 I

control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features AA2 - Ability to determine and interpret the following as they apply to Excess RCS CE/ A 16 Excess RCS Leakage x

Leakage:

2.9 1

(2)

EA2.2 - Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments KIA Category Totals:

I 2

I 2

2 1

Group Point Total:

9

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

K K

K K

K A

A A

A G

KA Topic Imp Pts 1

2 3

4 5

6 1

2 3

4 K2 - Knowledge of bus power 003 Reactor Coolant Pump x

supplies to the following:

3.1 1

K2.01 -RCPS I*

2.2 - Equipment Control 003 Reactor Coolant Pump x 2.2.12 - Knowledge of surveillance 3.7 1

. **. procedures.

KS - Knowledge of the operational implications of the 004 Chemical and Volume x

following concepts as they apply 3.2 1

Control to the eves:

K5.16 - Source of TA VE. and TREF signals to control and RPS K3 - Knowledge of the effect that 005 Residual Heat Removal a loss or malfunction of the x

RHRS will have on the following:

3.1

• 1 K3.06 - CSS K6 - Knowledge of the effect of a 006 Emergency Core Cooling x

loss or malfunction on the following will have on the ECCS:

4.2 I

K6.13 - Pumps 2.4 - Emergency Procedures/Plan 006 Emergency Core Cooling x 2.4.31 - Knowledge of annunciator 3.4 1

alarms, indications, or response procedures.

Kl - Knowledge of the physical connections and/or cause-effect 007 Pressurizer Quench Tank x

relationships between the PRTS 3.0 1

and the following systems:

Kl.03 - RCS Kl - Knowledge of the physical connections and/or cause-effect 008 Component Cooling Water x

relationships between the CCWS 3.0 I

and the following systems:

K 1.05 - Sources of makeup water A4 - Ability to manually operate and/or monitor in the control 008 Component Cooling Water room:

x 2.5*

I A4.06 - Remote operation of hand-operated throttle valves to regulate CCW flow rate 010 Pressurizer Pressure K2 - Knowledge of bus power Control x

supplies to the following:

3.0 I

K2.0 I - PZR heaters

ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2 I Group I - REACTOR OPERA TOR System/Evolution #/Name K

K K

K K

K A

A A

A G

KA Topic Imp Pts 1

2 3

4 5

6 1

2 3

4

', 2.4 - Emergency 010 Pressurizer Pressure Procedures/Plan x

3.3 I

Control 2.4.18 - Knowledge of the specific bases for EOPs.

K4 - Knowledge of RPS design feature(s) and/or interlock(s) which provide for the following:

012 Reactor Protection x

K4.02 - Automatic reactor trip 3.9 I

when RPS setpoints are exceeded for each RPS function; basis for each K4 - Knowledge of ESF AS design feature(s) and/or 013 ESFAS x

interlock( s) which provide for 4.3*

I the following:

K4.04 - Auxiliary feed actuation signal A2 - Ability to (a) predict the impacts of the following malfunctions or operations on the CCS; and (b) based on those 022 Containment Cooling x

,. predictions, use procedures to 2.6 1

correct, control, or mitigate the consequences of those malfunctions or operations:

A2.03 - Fan motor thermal overload/high-speed operation Al - Ability to predict and/or monitor changes in parameters (to prevent exceeding design 026 Containment Spray x

limits) associated with operating 2.7 I

the CSS controls including:

A 1.06 - Containment spray pump cooling K4 - Knowledge of MRSS design feature(s) and/or interlock(s) 039 Main and Reheat Steam x

which provide for the following:

3.7 I

K4.05 - Automatic isolation of steam line A4 - Ability to manually operate 059 Main Feedwater x

and monitor in the control room:

3.1

• I A4.0 I - SGFPT trip indication

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

K K

K K

K A

A A

A G

KA Topic Imp Pts 1

2 3

4 5

6 1

2 3

4

.,:t A2 - Ability to (a) predict the

1J

impacts of the following r.. *

f malfunctions or operations on the MFW; and (b) based on I

,

i.;L those predictions, use 059 Main Feedwater

.X procedures to correct, 2.9*

I

'* control, or mitigate the

1 consequences of those malfunctions or operations

• i: A2.04 - Feeding a dry SIG u

• ~~~ K6 - Knowledge of the effect of a

'J;tt loss or malfunction of the 061 Auxiliary/Emergency following will have on the AFW Feed water x

tq2 components:

2.5 I

I:

K6.0 I - Controllers and I

., positioners

• f:p Kl - Knowledge of the physical

+*:. connections and/or cause-effect 062 AC Electrical Distribution x

relationships between the ac 3.7 I

'/ distribution system and the following systems:

i.t KI.04 - Off-site power sources

.::/

'r>

including:

'Y5; A I.OJ - Containment pressure,

<,{

temperature, and humidity KIA Category Totals:

3 2

3 3

2 2

3 '3 2

2

.. 3 Group Point Total:

28

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

K K

K K

K A

A A

A G

KA Topic Imp Pts 1

2 3

4 5

6 1

2 3

4 Lt A4 - Ability to manually operate and/or monitor in the control 002 Reactor Coolant x

room:

3.4*

I A4.08 - Safety parameter display systems 011 Pressurizer Level Control K2 - Knowledge of bus power x

supplies to the following:

3.1 I

System K2.0I - Charging pumps

,:k KS - Knowledge of the operational implications of the 016 Non-nuclear x

following concepts as they apply 2.7*

Instrumentation

\\:'

to the NNIS system:

K5.0 I - Separation of control and i'.

protection circuits A3 - Ability to monitor

' automatic operation of the ITM, 017 In-core Temperature x

including:

3.6*

1 Monitor 0':~,;

A3.0 I - Indications of normal, r!r;:,.;

,~ '

natural, and interrupted circulation of RCS Ht Al - Ability to predict and/or monitor changes in parameters I*'***\\*

to prevent exceeding design 029 Containment Purge x

. limits) associated with operating 3.4 I

the Containment Purge System controls including:

I*\\,

A 1.02 - Radiation levels K6 - Knowledge of the effect of a 035 Steam Generator loss or malfunction on the x

following will have on the S/GS:

2.6 I

(

K6.03 - SIG level detector K4 - Knowledge of SDS design 041 Steam Dump/Turbine x

feature(s) and/or interlock(s) 2.6*

Bypass Control which provide for the following:

K4. I 6 - Low main steam pressure A2 - Ability to (a) predict the impacts of the following malfunctions or operations on the Condensate System; and (b) based on those predictions, use 056 Condensate x

Procedures to correct, control, 2.6 I

or mitigate the consequences of those malfunctions or operations:

A2.04 - Loss of condensate pumps

ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2 I Group 2 - REACTOR OPERA TOR System/Evolution #/Name K

K K

K K

K A

A A

A G

KA Topic Imp Pts 1

2 3

4 5

6 1

2 3

4 2.2 - Equipment Control 2.2.44 - Ability to interpret control room indications to verify 068 Liquid Radwaste x

the status and operation of a 4.2 I

system, and understand how operator actions and directives affect plant and system conditions.

K3 - Knowledge of the effect

\\:.:,

that a loss or malfunction of the Fire Protection System will have 086 Fire Protection x

on the following:

2.7 I

':'i:";::

K3.01 - Shutdown capability with redundant equipment I

KIA Category Totals:

0 1

1 1

1 1

1 1

1 1

l Group Point Total:

10 I

/.:>

ES-401 PWR Examination Outline Form ES-401-2 Emergency & Abnormal Plant Evolutions - Tier I I Group I - Senior Reactor Operator E/APE #/Name/Safety K

K K

A A

G KA Topic Imp Pts Function I

2 3

I 2

2.4 - Emergency Procedures I Plan 01 I Large Break LOCA x

2.4.35 - Knowledge of local auxiliary 4.0 I

(3) operator tasks during an emergency and the resultant operational effects.

I**

AA2 - Ability to determine and interpret 040 Steam Line Rupture the following as they apply to the Steam x

Line Rupture:

4.7 I

(4)

Ii?

AA2.02 - Conditions requiring a reactor If trip I

AA2 - Ability to determine and interpret the following as they apply to the Loss of 056 Loss of Off-site Offsite Power:

4.3 I

Power (6)

AA2.32 - Transient trend of coolant temperature toward no-load TAVE 2.1 - Conduct of Operations 062 Loss of Nuclear

',\\;,:

2. I.23 - Ability to perform specific system

• x.

4.4 I

Service Water (4) and integrated plant procedures during all modes of plant operation.

• i/. /.*

EA2 - Ability to determine and interpret the following as they apply to the CE/E02 Reactor Trip -

• (Reactor Trip Recovery)

Stabilization - Recovery iX EA2.2 - Adherence to appropriate 4.0 I

( 1) procedures and operation within the limitations in the facility's license and amendments 2.4 - Emergency Procedures/Plan 2.4.4 - Ability to recognize abnormal CE/E06 Loss of Main x>

indications for system operating 4.7 I

Feedwater (4) parameters that are entry-level conditions for emergency and abnormal operating procedures.

I KIA Category Totals:

0 0

0 0

3 3

Group Point Total:

6

ES-401 PWR Examination Outline Form ES-401-2 Emergency & Abnormal Plant Evolutions - Tier I I Group 2 - Senior Reactor Operator E/APE #/Name/Safety Kl K2 K3 Al A2 G

KA Topic Imp Pts Function AA2 - Ability to determine and interpret 00 I Continuous Rod the following as they apply to the

,x Continuous Rod Withdrawal:

4.6 I

Withdrawal (I )

I<',

AA2.05 - Uncontrolled rod withdrawal, from available indications 2.4 - Emergency Procedures I Plan

'; 2.4.21 - Knowledge of the parameters and 032 Loss of Source logic used to assess the status of safety Range NI (7) x functions, such as reactivity control, core 4.6 I

cooling and heat removal, reactor coolant system integrity, containment conditions,

~:,,

radioactivity release control, etc.

I~

2.2 - Equipment Control 036 Fuel Handling

' ;,,:x ; 2.2.39 - Knowledge ofless than or equal to 4.5 I

Accident (8) l:i one hour Technical Specification action statements for systems.

Ability to determine and interpret the 051 Loss of Condenser x

following as they apply to the Loss of 2.7*

I Vacuum (4)

Condenser Vacuum:

' ;; AA2.0 I - Cause for low vacuum condition Kl A Category Totals:

0 0

0 0

2

';)

Group Point Total:

4

ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2 I Group 1 - Senior Reactor Operator System/Evolution K

K K

K K

K A

A A

A G

KA Topic Imp Pts

1. /Name I

2 3

4 5

6 I

2 3

4 A2 - Ability to (a) predict the impacts of the following malfunctions or operations on the RHRS, and (b) based on 005 Residual Heat those predictions, use Xi procedures to correct, control, 3.1 I

Removal or mitigate the consequences of those malfunctions or operations:

(

A2.03 - RHR pump/motor malfunction 2.4 - Emergency Procedures I

• ~.;

Plan 012 Reactor Protection x

2.4.49 - Ability to perform without reference to procedures 4.4 I

those actions that require immediate operation of system components and controls.

A2 - Ability to (a) predict the impacts of the following malfunctions or operations on the CCS; and (b) based on 022 Containment

• x those predictions, use 2.6 I

Cooling procedures to correct, control, or mitigate the consequences of those malfunctions or i

operations:

A2.02 - Fan motor vibration A2 - Ability to (a) predict the impacts of the following malfunctions or operations on the MRSS; and (b) based on 039 Main and Reheat predictions, use procedures to x

correct, control, or mitigate 3.6 I

Steam the consequences of those malfunctions or operations:

A2.05 - Increasing steam demand, its relationship to increases in reactor power 061 2.2 - Equipment Control Auxiliary/Emergency x

2.2.38 - Knowledge of 4.5 I

Feedwater conditions and limitations in the facility license.

I KIA Category Totals:

0 0

0 0

0 0

0 3

0 0

2 Group Point Total:

5

ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2 I Group 2 - Senior Reactor Operator System/Evolution K

K K

K K

K A

A A

A G

KA Topic Imp Pts

1. /Name 1

2 3

4 5

6 1

2 3

4 2.2 - Equipment Control 015 Nuclear 2.2.42 - Ability to recognize Instrumentation x system parameters that are 4.6 1

entry-level conditions for Technical Specifications.

A2 - Ability to (a) predict the impacts of the following malfunctions or operations on the Waste Gas Disposal System; and (b) based on 071 Waste Gas x

those predictions, use 2.5 1

Disposal procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.06 - Supply failure to the isolation valve A2 - Ability to (a) predict the impacts of the following malfunctions or operations on the circulating water system; and (b) based on those predictions, use procedures to 075 Circulating Water x

correct, control, or mitigate 2.7*

1 the consequences of those malfunctions or operations:

A2.03 - Safety features and relationship between condenser vacuum, turbine trip, and steam dump I

KIA Category Totals:

0 0

0 0

0 0

0 2

0 0

1 Group Point Total: j 3

ES-401 PWR Examination Outline Form ES-401-3 Tier 3 Generic Knowledge & Abilities Outline - RO & SRO I Facility: CALVERT CLIFFS NUCLEAR POWER PLANT Date of Exam: 06/13/2016 RO SRO Category KIA#

Topic IR Pts IR Pts Knowledge of procedures, guidelines, or limitations 2.1.37 associated with reactivity management.

4.3 I

. )~ii$

,\\'

2.1.45 Ability to identify and interpret diverse indications to 4.3 I

]<": ; J*~i~~,'.,

Conduct validate the response of another indication.

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

4.4 I

r.:...

2.1.40 Knowledge of refueling administrative requirements.

3.9 I

Subtotals:

2 2

Ability to perform pre-startup procedures for the facility, 2.2.1 including operating those controls associated with plant 4.5 I

equipment that could affect reactivity.

j<;'J.

Ability to manipulate the console controls as required to

<.'.i>

i Equipment 2.2.2 operate the facility between shutdown and designated power 4.6 I

levels.

Control

.i' 2.2.13 Knowledge of tagging and clearance procedures.

4.1 I

2.2.7 Knowledge of the process for conducting special or 3.6 I

infrequent tests.

Subtotals:

3 I

2.3.11 Ability to control radiation releases.

3.8 I I Knowledge of radiation or contamination hazards that may 2.3.14 arise during normal, abnormal, or emergency conditions or 3.4 I

activities.

Radiation Ability to use radiation monitoring systems, such as fixed Control 2.3.5 radiation monitors and alarms, portable survey instruments, 2.9 I

personal monitoring equipment, etc.

)

2.3.6 Ability to approve release permits.

3.8 I

Subtotals:

2 2

2.4.13 Knowledge of crew roles and responsibilities during EOP 4.0 I

usage.

2.4.25 Knowledge of fire protection procedures.

3.3 I

Emergency 2.4.43 Knowledge of emergency communications systems and 3.2 I

,.. /

techniques.

Procedures/Plan 2.4.6 Knowledge of EOP mitigation strategies.

4.7 I

2.4.40 Knowledge of SRO responsibilities in emergency plan 4.5 I

implementation.

Subtotals:

3 2

I Tier 3 Totals I 10 I 7 I

ES-401 Record of Rejected Kl As Form ES-401-4 Tier I Group Randomly Selected Kl A Reason for Rejection I

This KIA is not applicable to plant since CCNPP utilizes Component RO 022 Loss of Reactor Cooling Water to cool the thermal barrier. KIA would be applicable at 111 Coolant Makeup AK3.06 a site that utilizes seal injection to help cool the barrier. Kept system and replaced KIA with one that had not been sampled. KIA AK3.04 was randomly selected using numbered poker chips.

RO 003 Reactor Coolant This Kl A is not applicable to the RO. Knowledge of bases in Technical Pump Specifications is a SRO responsibility. Kept system and replaced KIA 211 2.2.25 with one that had not been sampled. KIA 2.2.12 was randomly selected using numbered poker chips.

RO 0 I 0 Pressurizer Pressure This Kl A is not applicable to plant and previously approved for Control suppression. Kept system and replaced Kl A with one that had not been 211 AK2.04 sampled. KIA AK2.0I was randomly selected using numbered poker chips.

This KIA is not applicable to plant since Containment Iodine Removal RO 027 Containment Iodine units are not monitored for temperature and there are no actions for the Removal operator to perform. Replaced system since there was only one KIA 212 A2.0I choice under A2 for the Containment Iodine Removal system.

Replaced with System 056 Condensate and KIA A2.04. System and Kl A were randomly selected using numbered poker chips.

RO 068 Liquid Radwaste This KIA is not applicable to the Liquid Radwaste System since Liquid Radwaste is common between the two units. Kept system 068 and 212 2.2.3 replaced KIA with one that had not been sampled. KIA 2.2.44 was randomly selected, using numbered poker chips.

RO This Kl A is not applicable to the RO. Management of shutdown 2.2.18 maintenance activities is a SRO responsibility. Replaced KIA with one 3/NA that had not been sampled. Kl A 2.2. I was randomly selected, using numbered poker chips.

SRO 032 Loss of Source Spent several unsuccessful hours attempting to develop a question for Range NI (7) this combination ofsystemlKIA. Kept system 032 and replaced KIA 112 2.4.6 with one that had not been sampled. KIA 2.4.21 was randomly selected, using numbered poker chips.

Exam Outline Statement The methodology used to generate the Written Exam Outline was Westinghouse NRC K/A Exam Generator Software.

CCNPP Lead Exam Author

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Calvert Cliffs Nuclear Power Plant Date of Examination: 611312016 Exam Level: RO Operating Test#: 2016 Administrative Topic Type Describe activity to be performed (see Note)

Code*

Respond to Abnormal RCS Chemistry Conduct of Operations R,N Kl A 2.1.34 Knowledge of primary and secondary plant chemistry limits Importance 2.7 Verify the adequacy of mechanical boundaries for a safety tagout Equipment Control R,N Kl A 2.2.41 Ability to obtain and interpret station electrical and mechanical drawings Importance 3.5 Recall Emergency Response Organization Emergency Plan S, N Kl A 2.4.39 Knowledge of RO responsibilities in emergency plan implementation Importance 3.9 Respond to Condenser Vacuum reduction (Evaluate U-2 Load versus Vacuum)

Conduct of Operations R,N Kl A 2.1.25 Ability to interpret reference materials, such as graphs, curves, tables, etc.

Importance 3.9 NOTE: All items (five total) are required for SROs. RO applicants require only four 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 (2'. I)

(P)revious 2 exams (S I; randomly selected)

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Calvert Cliffs Nuclear Power Plant Date of Examination: 6/13/2016 Exam Level: RO Operating Test#: 2016 Control Room Systems:

• 8 for RO; 7 for SRO-I; 2 or 3 for SRO-U System I JPM Title Type Code*

Safety Function

a.

SIM-I Respond to CEA(s) Misaligned by >7.5" but <15" 001 Control Rod Drive System A,E,N,S I

K4.02 Control rod mode select control (movement control)

b.

SIM-2 Verify RAS A,E,EN, 013 Engineered Safety Features Actuation System L,N, S 2

A3.02 Operation of actuated equipment

c.

SIM-3 Respond to an uncontrolled loss of R WT level 006 Emergency Core Cooling System E,L,N, S 3

A2.03 System leakage

d.

SIM-4 Restart RCPs 002 Reactor Coolant System D,E,L,S 4P A2.03 Loss of forced circulation

e.

SIM-5 Verify the Containment Environment Safety Function is satisfied A,E,EN, 027 Containment Iodine Removal System L, N, S 5

A4.03 CIRS fans

f.

SIM-6 Restore Auxiliaries (Restore 21 4K v bus using Offsite power) 062 AC Electrical Distribution System E, L, N, S 6

A2.05 Methods for energizing a dead bus

g.

SIM-7 Deenergize an RPS Cabinet for maintenance 012 Reactor Protection System A,N,S 7

A2.04 Erratic power supply operation

h.

SIM-8 Respond to a Fuel Handling incident in the Containment 034 Fuel Handling Equipment System D,E,L,S 8

A2.0l Dropped fuel element In-Plant Systems

• 3 for RO; 3 for SRO-I; 3 or 2 for SRO-U

a.

PL T-1 Start 11 & 12 Containment Air Coolers 022 Containment Cooling System D,E,L,R 5

068A1.28 Transfer of controls from control room to shutdown panel or local control

b.

PLT-2 Deenergize CV-517/518/519/PZR/RVH Vents and Trip MCC-114 Load Center Breaker D,E,L 6

063 DC Electrical Distribution System 068 A 1.28 PZR level control and pressure control

c.

PL T-3 Control RCS and SIG Inventory from 2C43 035 Steam Generator System D,E,L 4P Auxiliary shutdown panel controls and indicators

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

• 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)ltemate path 4-6 I 4-6 I 2-3 (C)ontrol room (D)irect from bank

S9/<::8/<::4 (E)mergency or abnormal in-plant 2': II 2': I I 2': I (EN)gineered safety feature

- I - I 2':1 (control room system)

(L)ow-Power I Shutdown 2': II 2': I I 2': I (N)ew or (M)odified from bank including 1 (A) 2:2/?_2/?_]

(P)revious 2 exams

S 3 I<:: 3 I <:: 2 (randomly selected)

(R)CA 2':Il2':1/?_]

(S)imulator

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Calvert Cliffs Nuclear Power Plant Date of Examination: 611312016 Exam Level: SRO-I Operating Test#: 2016 Administrative Topic Type Describe activity to be performed (see Note)

Code*

Complete the Shift Turnover Information Sheet (Minimum Essential Conduct of Operations S,N Equipment List Attachment)

Kl A 2.1.3 Knowledge of shift or short-term relief turnover practices Importance 3.9 Respond to Abnormal RCS Chemistry Conduct of Operations R,N Kl A 2.1.34 Knowledge of primary and secondary plant chemistry limits Importance 3.5 Verify the adequacy of mechanical boundaries for a safety tagout Equipment Control R,N Kl A 2.2.41 Ability to obtain and interpret station electrical and mechanical drawings Importance 3.9 Determine radiological-based emergency response actions per the ERPIP while maintaining an overview of plant conditions Radiation Control R,N KIA 2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions Importance 3.7 Recommend Protective Action Guidelines to Public Officials Emergency Plan R,N Kl A 2.4.44 Knowledge of emergency plan protective action recommendations Importance 4.4 NOTE:

All items (five total) are required for SROs. RO applicants require only four 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 (:'.'. l)

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

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Calvert Cliffs Nuclear Power Plant Date of Examination: 6/1312016 Exam Level: SRO-I Operating Test#: 2016 Control Room Systems:

• 8 for RO; 7 for SRO-I; 2 or 3 for SRO-U System I JPM Title Type Code*

Safety Function

a.

SIM-I Respond to CEA(s) Misaligned by >7.5" but <15" 001 Control Rod Drive System A,E, N, S 1

K4.02 Control rod mode select control (movement control)

b.

SIM-2 Verify RAS A,E,EN, 013 Engineered Safety Features Actuation System L, N, S 2

A3.02 Operation of actuated equipment

c.

SIM-3 Respond to an uncontrolled loss of RWT level 006 Emergency Core Cooling System E,L, N, S 3

A2.03 System leakage

d.

SIM-4 Restart RCPs 002 Reactor Coolant System D, E, L, S 4P A2.03 Loss of forced circulation

e.

SIM-5 Verify the Containment Environment Safety Function is satisfied A,E,EN, 027 Containment Iodine Removal System L, N, S 5

A4.03 CIRS fans

f.

SIM-7 Deenergize an RPS Cabinet for maintenance 012 Reactor Protection System A,N,S 7

A2.04 Erratic power suooly operation

g.

SIM-8 Respond to a Fuel Handling incident in the Containment 034 Fuel Handling Equipment System D, E, L, S 8

A2.0 I Dropped fuel element In-Plant Systems

• 3 for RO; 3 for SRO-I; 3 or 2 for SRO-U

a.

PL T-1 Start 11 & 12 Containment Air Coolers 022 Containment Cooling System D,E,L,R 5

068 A 1.28 Transfer of controls from control room to shutdown panel or local control

b.

PLT-2 Deenergize CV-517/518/519/PZR/RVH Vents and Trip MCC-114 Load Center Breaker D,E,L 6

063 DC Electrical Distribution System 068 A 1.28 PZR level control and pressure control

c.

PL T-3 Control RCS and S/G Inventory from 2C43 035 Steam Generator System D,E,L 4P Auxiliary shutdown panel controls and indicators

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

• 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 I 4-6 I 2-3 (C)ontrol room (D)irect from bank

S9/:S8/:S4 (E)mergency or abnormal in-plant 2:1/2:1/2:1 (EN)gineered safety feature

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

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

S 3 I :S 3 I :S 2 (randomly selected)

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

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Calvert Cliffs Nuclear Power Plant Date of Examination: 6/13/2016 Exam Level: SRO-U Operating Test#: 2016 Administrative Topic Type Describe activity to be performed (see Note)

Code*

Complete the Shift Turnover Information Sheet (Minimum Essential Conduct of Operations S, N Equipment List Attachment)

Kl A 2.1.3 Knowledge of shift or short-term relief turnover practices Importance 3.9 Respond to Abnormal RCS Chemistry Conduct of Operations R,N Kl A 2.1.34 Knowledge of primary and secondary plant chemistry limits Importance 3.5 Verify the adequacy of mechanical boundaries for a safety tagout Equipment Control R,N Kl A 2.2.41 Ability to obtain and interpret station electrical and mechanical drawings Importance 3.9 Determine radiological-based emergency response actions per the ERPIP while maintaining an overview of plant conditions Radiation Control R,N KIA 2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions Importance 3.7 Recommend Protective Action Guidelines to Public Officials Emergency Plan R,N Kl A 2.4.44 Knowledge of emergency plan protective action recommendations Importance 4.4 NOTE:

All items (five total) are required for SROs. RO applicants require only four 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 (2 I)

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

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Calvert Cliffs Nuclear Power Plant Date of Examination: 6/13/2016 Exam Level: SRO-U Operating Test#: 2016 Control Room Systems:

• 8 for RO; 7 for SRO-I; 2 or 3 for SRO-U System I JPM Title Type Code*

Safety Function

a.

SIM-1 Respond to CEA(s) Misaligned by >7.5" but <15" 00 I Control Rod Drive System A, E, N, S 1

K4.02 Control rod mode select control (movement control)

b.

SIM-2 Verify RAS A, E, EN, 013 Engineered Safety Features Actuation System L, N, S 2

A3.02 Operation of actuated equipment

c.

SIM-8 Respond to a Fuel Handling incident in the Containment 034 Fuel Handling Equipment System D,E,L,S 8

A2.0 I Dropped fuel element In-Plant Systems* 3 for RO; 3 for SRO-I; 3 or 2 for SRO-U

a.

PL T-1 Start 11 & 12 Containment Air Coolers 022 Containment Cooling System D,E,L,R 5

068 A 1.28 Transfer of controls from control room to shutdown panel or local control

b.

PL T-2 Deenergize CV-517/518/5 I 9/PZR/RVH Vents and Trip MCC-114 Load Center Breaker D,E,L 6

063 DC Electrical Distribution System 068 A 1.28 PZR level control and pressure control

c.

NI A since using 3 Simulator JPMs

• 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 I 4-6 I 2-3 (C)ontrol room (D)irect from bank

S9/:S8/:S4 (E)mergency or abnormal in-plant

11::::11::::1 (EN)gineered safety feature

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

(L)ow-Power I Shutdown 2:1/2:1/2:1 (N)ew or (M)odified from bank including 1 (A) 2:2/2:2/21 (P)revious 2 exams

S 3 I :S 3 I :S 2 (randomly selected)

(R)CA

11::::1121 (S)imulator

Appendix D Scenario Outline Form ES-D-1 Calvert Cliffs Nuclear Power Plant Scenario #: 1 OP-Test#: 2016 Examiners:

Operators:

Initial Conditions: Unit-I is at 100% power, MOC. Unit-2 is at 100% power.

Turnover: 11 AFW Pump is OOS, 11 SLC Pump is OOS, 13 CCW Pp breaker to 14 480 bus is OOS, 12 SGFP MOP-Bis OOS.

Instructions to the crew: Maintain current power level.

Event#

Malfunction #

Event Type*

Event Description 1

PIC07 lTIC C.:ATC TIC-223 Setpoint Fails High 223 PTSTPT 2

ccw002 02 C-BOP/SRO 12 CCW Pump Breaker Failure/AOP-7C T-SRO 3

msOl 9 03 I - BOP/SRO 12 S/G LT-l 123C Fails Low T-SRO 4

cvcs023 01 C-ATC/SRO 11 Charging Pump Failure T-SRO rcsOl 1 03 C-BOP/SRO 5

R-ATC l 2A RCP Lower and Middle Seal Failures rcs012 03 6

1HS2328 C-ATC/SRO Turbine Generator Runback/ AOP-7F /EOP-0 rcs008 03 7

rcs013 03 M-ALL 12A RCP Locked Rotor/350 GPM RCS Leak rcs014 03 8

esfaOOl 01 I-ATC SIAS A Auto Failure 9

ms003 02 C-BOP 12 SG Steam Leak in Turbine Building msOl 1 02 EOP-8 (N)ormal (R)eactivity (l)nstrument (C)omponent (M)ajor (T)ech Spec Critical Tasks:

1. Trips I IA & 12B or 1 IB & 12A RCPs when RCS pressure decreases to <1725 PSIA prior to RCS pressure reaching 1300 PSIA.

2. Identifies 12 S/G as the faulted S/G and isolates 12 SIG.

Appendix D Scenario Outline Calvert Cliffs Nuclear Power Plant Scenario #: 1 Event 1 SCENARIO OVERVIEW MULTIPLE RCP SEAL FAILURE I LOCA/ESDE Form ES-D-1 OP-Test#: 2016 Letdown HX Temperature Controller, TIC-223, will malfunction. Letdown temperatures downstream of the Non-Regenerative Heat Exchanger will rise. The crew will use 1C07 Alarm Manual to ensure CVCS IX's are bypassed and will shift the controller to manual to restore Non-Regenerative Heat Exchanger outlet temperatures to normal.

Event 2 12 Component Cooling (CCW) Pump trips. The crew will implement AOP-7C and determine that a common mode failure does not exist. Either 11 or 13 CCW Pump will be started and the RCPs will be monitored to ensure bearing temperatures and flows are returning to normal. Tech Spec 3.7.5 will be evaluated for applicability.

Event3 12 S/G Level Transmitter, LT-1123C, will fail low. The crew will respond per the 1C03 and 1 C05 Alarm Manuals and determine that LT failure impacts both RPS and ESF AS. They may use the N0-1-200, Common Tap Analysis (Att. 12) to determine that the failure is isolated to L T-1123C only. The crew will bypass the affected RPS Channel "C" Trip unit using OI-6 and direct the bypassing of the affected ESFAS Sensor "ZF" modules per OI-34. Tech Spec 3.3.1 will be evaluated for applicability.

Event 4 The running Charging Pump will trip due to a breaker failure. The crew will respond per the 1 C07 Alarm Manuals. The crew will start a backup Charging Pump per OI-2A. Technical Requirements Manual 15.1.2 will be evaluated for applicability.

Event 5 Two seals will fail on l 2A RCP. The crew will use the 1 C06 Alarm Manual and determine that both the 12A RCP lower and middle seals have failed and determine an expeditious downpower is required. The crew will use OP-3 and commence a downpower at< 30%/hr.

Event 6 As power is being lowered, a Generator Run back will occur due to the loss of 12 Stator Liquid Cooling Pump. The crew should recognize that Generator load is rapidly lowering and that RCS pressure and temperature are both rising beyond their control. The crew is expected to manually trip the Unit before receipt of any automatic RPS trip.

Event 7 The crew should implement EOP-0 after the reactor is tripped. Immediately after the trip occurs, 12A RCP will suffer a catastrophic seal assembly failure resulting in the failure of all remaining seals and a seized rotor event. The result will be a LOCA into Containment. The crew should identify PZR level is lowering and should isolate letdown. The crew should identify degrading Containment pressures and temperatures and take actions to maximize Containment Cooling.

Event 8 When SIAS is received, the crew should identify the failure of SIAS-A to actuate and manually initiate SIAS-A. The crew is also expected to complete RCP trip strategy once SIAS is received, Page 3 of22

Appendix D Scenario Outline Form ES-D-1 Calvert Cliffs Nuclear Power Plant Scenario #: 2 OP-Test#: 2016 Examiners:

Operators:

Initial Conditions: U-1 is at 100% power, MOC. U-2 sat 100%.

Turnover: 23 AFW Pump is OOS, 11 CAC is OOS, MCClOlAT and lOlBT are tied, IA DG running for STP 0-8A-l, Max Emergency Generation is in effect.

Instructions to the crew: Maintain current power level. Complete STP 0-8A-l.

Event#

Malfunction #

Event Type*

Event Description 1

cvcs006 C-ATC/SRO CVC-516 Fails Shut T-SRO 2

PlC18 M06 C-BOP lA DG Low Lube Oil Pressure T-SRO 3

rcs023 02 I-ATC PT-lOOY Fails Low 4

cw004 01 C-BOP/SRO 15 CWP Exciter Failure/ AOP-7G/ AOP-7L R-ATC 5

P 1 C 1 8 KE 11 02 C-ALL Loss ofU-4000-11 Transformer/EOP-0 6

afw004 01 C-BOP AFAS Failure afw004 02 7

l-AFW-161 M-ALL AFW Suction Blockage EOP-3 8

P1C06 1HS151 C-ATC l lA RCP Trip Failure (N)ormal (R)eactivity (I)nstrument (C)omponent (M)ajor (T)ech Spec Critical Tasks:

1. Establishes AFW flow to at least one SIG and establishes a rising level trend in at least one SIG prior to SIG levels going below (-) 350 inches.

2. Commences OTCC when both SIG levels are below(-) 350 inches or TCOLD rises uncontrollably 5 °F or greater. (Must be initiated prior to CET temperature reaching 560°F.)

Appendix D Scenario Outline Form ES-D-1 Calvert Cliffs Nuclear Power Plant Scenario#: 2 OP-Test#: 2016 SCENARIO OVERVIEW LOSSOFVACUUM/LOSSOFPOWER/LOSSOFALLFEEDWATER Event 1 Letdown CV -516 fails shut, isolating letdown. The crew will respond using the 1 C07 Alarm Manual. The crew will secure Charging and align Charging Pumps per OI-2A to control PZR level. Tech Spec 3.4.9 and Technical Requirements Manual 15.1.2 will be evaluated for applicability.

Event 2 The Control Room will get a radio notification that a lube oil leak has developed on the IA DG.

The leak will be large enough that the DG should be secured per OI-21A. Tech Spec 3.3.1 will be evaluated for applicability.

Event 3 Pressurizer Pressure Controller PT-1 OOY fails low. The crew will respond using the 1 C06 Alarm Manual. The crew will shift pressure control to Channel X.

Event4 15 Circulating Water Pump (CWP) will trip. The crew will respond using AOP-7L. The crew will secure the 13A Waterbox (associated with 15 CWP). Vacuum will begin lowering and the crew will respond using AOP-70. A rapid downpower per OP-3 will begin due to the lowering vacuum.

Event 5 The plant will trip due to the loss ofU-4000-11. The crew will respond using EOP-0. The crew will restore PZR level by restarting Charging Pumps, shut the MSIV s due to the loss of power to the Moisture Separator Reheater Steam Source MOVs, restart a Component Cooling Water Pump to ensure cooling to the Reactor Coolant Pumps, and establish an RCS heat sink using the Atmospheric Dump Valves.

Event 6 AF AS will not automatically actuate due to a failure. The crew will need to initiate AFW flow using either 13 AFW Pump or 11 AFW Pump to restore S/G levels. Initiating AFW is a Critical Task since Main Feedwater has been lost due to the loss of power.

Event 7 A complete blockage of AFW will occur. The crew should identify the loss of all feed water and implement EOP-3.

Event 8 The crew will attempt to secure all the Reactor Coolant Pumps. 11 A RCP will fail to trip from 1 C06 requiring the crew to trip the RCP using an alternate breaker location. The crew will continue with actions in EOP-3 including isolating S/G blowdown, commencing boration, and commencing a rapid cooldown. Once S/G levels lower below -350", the crew will initiate Once-Through-Core-Cooling (OTCC) to ensure RCS heat removal. The scenario can be terminated once the critical task associated with establishing OTCC is completed.

Page 3 of21

Appendix D Scenario Outline Form ES-D-1 Calvert Cliffs Nuclear Power Plant Scenario #: 3 OP-Test#: 2016 Examiners:

Operators:

Initial Conditions: U-1 is at 10-4% power, MOC. Unit-2 is in Mode 5.

Turnover: 12 SGFP is OOS (tags being cleared), 11 ADV is wisping a small amount of steam.

Instructions to the crew: Prepare to raise power to 1 % per OP-2.

Event#

Malfunction #

Event Type*

Event Description 1

1HS5010 C-BOP/SRO 11 SGFP Trip/AOP-3G 2

N-SRO Raise Reactor Power to 1 %/OP-2 R-ATC 3

ni002 03 I-BOP/SRO NI-RPS Ch C Wide Range NI High Volt Power T-SRO Supply fails 4

cd002 C-BOP/SRO Hotwell Level Transmitter L T-4405 Fails Low/AOP-3G 5

ceds012 06 C-ATC/SRO Dropped CEA/AOP-lB T-SRO 6

ceds012 58 C-ATC/SRO Second Dropped CEA/EOP-0 7

ms002 01 M-ALL 11 S/G Tube Rupture 8

rmsOOl I-BOP Failure of RI-4014 to Isolate Blowdown/EOP-6 9

P1C09 1HS5 C-ATC SIAS block failure (N)ormal (R)eactivity (I)nstrument (C)omponent (M)ajor (T)ech Spec Critical Tasks: (shaded)

1. Trips 11 A & 12B or 11 B & 12A RCPs when RCS pressure decreases to <1725 PSIA prior to RCS pressure reaching 1300 PSIA.

2. Identifies 11 S/G as the faulted S/G. Isolates 11 S/G.

Appendix D Scenario Outline Calvert Cliffs Nuclear Power Plant Scenario #: 3 Event 1 SCENARIO OVERVIEW DROPPED CEA I SG TUBE RUPTURE Form ES-D-1 OP-Test#: 2016 11 SGFP will trip. The crew will respond using AOP-3G. The crew will reset the SGFP and restore feedflow to the S/Gs.

Event 2 The crew will be requested to raise reactor power to 1 % power using OP-2. Power will be raised to the Point of Adding Heat.

Event 3 A High Volt power supply for Channel C WRNI will fail low. The crew will bypass the affected RPS Channel "C" Trip units using OI-6. Tech Spec 3.3.1 will be evaluated for applicability.

Event4 A Level Transmitter input to the Hotwell Level Controller will fail low. The crew will respond using AOP-3G. The crew will shift Hotwell Level Control to manual to restore Hotwell levels.

Event 5 A CEA will drop. The crew will respond using AOP-lB. The crew will recognize that the reactor has become subcritical and will insert all Regulating Group CEAs. Tech Specs 3.1.4 and 3.1.5 will be evaluated for applicability.

Event 6 A second CEA will drop while the Regulating Group CEAs are being inserted. The crew will respond using AOP-1 B. The crew will manually trip the reactor due to two dropped CEAs. The crew will respond using EOP-0.

Event 7 Immediately after the reactor is manually tripped, a S/G tube rupture in 11 S/G will begin. The crew will isolate letdown and start all Charging Pumps, and use Turbine Bypass Valves to establish an RCS heat sink.

Event 8 The S/G Blowdown RMS will fail to isolate the S/G Blowdown isolation valves. The crew will manually isolate S/G Blowdown. The crew will implement EOP-6. The crew will manually initiate High Pressure Safety Injection, commence RCS boration, and commence a rapid cooldown to 5 l 5°F.

Event 9 During the cooldown, Pressurizer Pressure Block will not function, and a SIAS will actuate. The crew will trip 2 RCPs once RCS pressure lowers below 1725 PSIA to meet the Trip 2/Leave 2 Critical Task,. The crew will continue the cooldown to a Thot of 515°F and then perform Critical Task to isolate 11 S/G. The scenario can be terminated once 11 S/G is isolated.

Page 3 of21