ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 · I ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Calvert Cliffs Nuclear Power Plant Date of Examination:

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

cility: Calvert Cliffs Nuclear Power Plant Date of Examination: 8/6/2018 am Level: SRO-I Operating Test#: 2018

Control Room Systems: * 8 for RO; 7 for SRO-I; 2 or 3 for SRO-U

System/ JPM Title Type Code* Safety

Function

a. SIM-I Dilution of the RCS/Respond to loss of I YI 0 004 Chemical and Volume Control System A,N,S I A2.06 Inadvertent boration/dilution (SR0-4.3)

b. SIM-2 Reset SIAS following a SG Tube Rupture 013 Engineered Safety Features Actuation System A4.02 Reset of ESF AS channels (SR0-4.4)

EN,N,S 2

C. SIM-3 Fill Safety Injection Tank/Respond to lowering pressure 006 Emergency Core Cooling System A, L, N, S 3 Al.13 Accumulator pressure (level, boron concentration) (SR0-3.7)

d. SIM-4 Raise 500 KV Voltage/Respond to Generator trip 045 Main Turbine Generator System A, N, S 4S A3. l l Generator trip (SR0-2.9)

e. SIM-5 Verify the Containment Environment Safety Function is satisfied 026 Containment Spray System EN,N,S 5 A 1.01 Containment pressure (SR0-4.2)

f. SIM-6 Transfer ESF Bus from SMECO to BGE Grid 062 AC Electrical Distribution System N, S 6 A4.0I All breakers (including available switchyard) (SR0-3.1)

g.

h. SIM-8 Raise LPSI Flow/Loss of RFP Level 029 Containment Purge System A,N,S 8 Al.02 Radiation levels (SR0-3.4)

In-Plant Systems * 3 for RO; 3 for SRO-I; 3 or 2 for SRO-U

i. PL T-1 Secure the MGs/V erify Reactor shutdown/Isolate SG BO/Conserve RCS inventory during fire in CSR

E, N I 001 Control Rod Drive System A 1.06 Reactor power (SR0-4.4)

j. PLT-2 Start I I SW AC/Isolate 11 Cavity Cooling Fan Damper/Start SWGR HY AC during Control Room Evacuation 068 Control Room Evacuation D,E,R 8 AA 1.21 Transfer of controls from control room to shutdown panel or local control (SR0-4.1)

k. PL T-3 Place 21 Battery Charger in service 063 DC Electrical Distribution System

N 6 A3.0 I Meters, annunciators, dials, recorders, and indicating lights (SR0-3.1)


* 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/ SRO-I / SRO-U

(A)ltemate path 4-6 I 4-6 I 2-3 [ 4]

(C)ontrol room

(D)irect from bank :S9/:S8/:S4 [I]

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

(EN)gineered safety feature - I - I 2: I ( contro I room system) [2]

(L)ow-Power I Shutdown 2:1/2:1/2:1 [1]

(N)ew or (M)odified from bank including l(A) 2: 2 I 2: 2 / 2: I [9 including 4(A)]

(P)revious 2 exams :S 3 / :S 3 / :S 2 (randomly selected) [OJ

(R)CA 2:1/2:1/2:1 [I]

(S)imulator

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

Facility: Calvert Cliffs Nuclear Power Plant Date of Examination: 8/6/2018 Exam Level: RO Operating Test#: 2018


System / JPM Title Type Code* Safety

Function

a. SIM-] Dilution of the RCS/Respond to loss of I YI 0 004 Chemical and Volume Control System A,N,S 1 A2.06 Inadvertent boration/dilution (R0-4.2)

b. SIM-2 Reset SIAS following a SG Tube Rupture 013 Engineered Safety Features Actuation System EN,N,S 2 A4.02 Reset ofESFAS channels (R0-4.3)

C. SIM-3 Fill Safety Injection Tank/Respond to lowering pressure 006 Emergency Core Cooling System A,L,N,S 3 Al.13 Accumulator pressure (level, boron concentration) (R0-3.5)

d. SIM-4 Raise 500 KV Voltage/Respond to Generator trip 045 Main Turbine Generator System A,N,S 4S A3.11 Generator trip (R0-2.6)

e. SIM-5 Verify the Containment Environment Safety Function is satisfied 026 Containment Spray System EN,N,S 5 Al.OJ Containment pressure (R0-3.9)

f. SIM-6 Transfer ESF Bus from SMECO to BGE Grid 062 AC Electrical Distribution System N, S 6 A4.0l All breakers (including available switchyard) (R0-3.3)

g. SIM-7 Nulling NI Pots to Delta-T Pots 015 Nuclear Instrumentation System D, S 7 A3.03 Verification of proper functioning/operability (R0-3.9)

h. SIM-8 Raise LPSI Flow/Loss ofRFP Level 029 Containment Purge System A,N,S 8 A 1.02 Radiation levels (R0-3.4)

In-Plant Systems* 3 for RO; 3 for SRO-I; 3 or 2 for SRO-U

I. PLT-1 Secure the MGs/Verify Reactor shutdown/Isolate SGBD/Conserve RCS inventory during fire in CSR

E,N I 00 I Control Rod Drive System A 1.06 Reactor power (R0-4.1)

j. PL T-2 Start 11 SW AC/Isolate 11 Cavity Cooling Fan Damper/Start SWGR HVAC during Control Room Evacuation 068 Control Room Evacuation D,E,R 8 AA 1.21 Transfer of controls from control room to shutdown panel or local control (R0-3.9)


N 6 A3.01 Meters, annunciators, dials, recorders, and indicating lights (R0-2.8)



* Type Codes Criteria for RO/ SRO-I/ SRO-U

(A)lternate path 4-6 I 4-6 I 2-3 [4]

(C)ontrol room

(D)irect from bank :S9/:S8/:S4 [2]

(E)mergency or abnormal in-plant 2:112:112:1 [2]

(EN)gineered safety feature - I - I 2: I ( control room system) [2]

(L)ow-Power I Shutdown 2:112:1/2:1 [1]

(N)ew or (M)odified from bank including l(A) 2:2/2:2/2: I [9 including 4(A)]

(P)revious 2 exams :S 3 I :S 3 / :S 2 (randomly selected) [OJ

(R)CA 2:112:112:1 [ 1]

(S)imulator

.. ES-301 Administrative Topics Outline Form ES-301-1

Facility: Calvert Cliffs Nuclear Power Plant Date of Examination: 8/6/2018 Exam Level: SRO-U Operating Test#: 2018

Administrative Topic Type Describe activity to be performed

(see Note) Code*

Complete an OP-7 Core Alt Checklist Conduct of Operations R,N

G2. I .36 Knowledge of procedures and limitations involved in core alterations (SR0-4. I)

Verify an Estimated Critical Condition Calculation Conduct of Operations R,D

G2. I .25 Ability to interpret reference materials, such as graphs, curves, tables, etc. (SR0-4.2)

Evaluate and then independently verify AFW Pump performance data Equipment Control R,N during a Surveillance Test, then determine Tech Spec actions

G2.2. 12 Knowledge of surveillance procedures (SR0-4. I)

Approve Steam Generator Slowdown Release Permit and determine Radiation Control R,N ODCM actions ifRI-4014/4095 are taken OOS during the release

G2.3.6 Ability to approve release permits (SR0-3.8)

Interpret Seismic Event results and evaluate Emergency Plan and Emergency Plan R,N Reportability requirements

G2.4.29 Knowledge of the emergency plan (SR0-4.4)

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) [I]

(N)ew or (M)odified from bank (2: I) [4]

(P)revious 2 exams (:'.S I; randomly selected) [OJ


Facility: Calvert Cliffs Nuclear Power Plant Date of Examination: 8/6/2018 Exam Level: SRO-U Operating Test#: 2018


System/ JPM Title Type Code* Safety

Function

a.

b. SIM-2 Reset SIAS following a SG Tube Rupture 0 I 3 Engineered Safety Features Actuation System A4.02 Reset ofESFAS channels (SR0-4.4)

EN,N,S 2

C. SIM-3 Fill Safety Injection Tank/Respond to lowering pressure 006 Emergency Core Cooling System A, L, N, S 3 A 1.13 Accumulator pressure (level, boron concentration) (SR0-3 .7)

d. SIM-4 Raise 500 KV Voltage/Respond to Generator trip 045 Main Turbine Generator System A,N,S 4S A3. I l Generator trip (SR0-2.9)

e.

f.

0 e,•

h.

In-Plant Systems * 3 for RO; 3 for SRO-I; 3 or 2 for SRO-U

I.

j. PL T-2 Start 11 SW AC/Isolate 11 Cavity Cooling Fan Damper/Start SWGR HVAC during Control Room Evacuation 068 Control Room Evacuation D,E,R 8 AA 1.21 Transfer of controls from control room to shutdown panel or local control (SR0-4.1)


N 6 A3.0I Meters, annunciators, dials, recorders, and indicating lights (SR0-3.1)


* Type Codes Criteria for RO/ SRO-I / SRO-U

(A)lternate path 4-6 I 4-6 I 2-3 [2]

(C)ontrol room

(D)irect from bank :S9/:S8/:S4 [I]

(E)mergency or abnormal in-plant 21/21/c::1 [I]

(EN)gineered safety feature - I - I c::I ( control room system) [I]

(L)ow-Power / Shutdown 211211::::1 [l]

(N)ew or (M)odified from bank including I (A) 2 2 I 2 2 I 2 1 [4 including 2 (A)]

(P)revious 2 exams :S 3 / :S 3 / :S 2 (randomly selected) [O] (R)CA 211211::::1 [I]

(S)imulator

ES-401 PWR Examination Outline Form ES-401-2

Facility: CAL VERT CLIFFS NUCLEAR POWER PLANT Date of Exam: 08/06/2018 I RO Category KIA Points SRO Only Points

Group Tier

K K K K K K A A A A G Total A2 G Total

1 2 3 4 5 6 1 2 3 4

1. Emergency & 1 3 3 3 3 3 3 18 3 3 6

Abnormal Plant 2 2 2 2 NIA 1 1 NIA 1 9 2 2 4

Evolutions Tier Totals 5 5 5 4 4 4 27 5 5 10

1 2 3 2 3 2 3 2 3 3 2 3 28 3 2 5 2. Plant Systems 2 1 1 1 1 I 1 1 1 1 0 I IO 0 2 1 3

Tier Totals 3 4 3 4 3 4 3 4 4 2 4 38 5 3 8

1 2 3 4 I 2 3 4 3. Generic Knowledge & Abilities Categories IO 7

3 2 3 2 2 1 2 2

Note: I. Ensure that at least two topics from every applicable Kl A category are sampled within each tier of the RO and SRO-

only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the "Tier Totals" in each 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. 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 Kl 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 Kl A categories. 7. The generic (G) K/As in Tiers 1 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.l.b ofES-401 for the applicable K/As. 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 K/As that are linked to 10 CFR 55.43.

G* Generic K/As * These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the

Kl A catalog is used to develop the sample plan. They are not required to be included when using earlier revision of the KIA catalog.

** These system/evolutions may be eliminated from the sample (as applicable to the facility) when Revision 3 of the KIA catalog is used to develop the sample plan.


Emergency & Abnormal Plant Evolutions-Tier 1 / Group I -REACTOR OPERATOR

E/APE #/Name/Safety Function K K K A A G KA Topic Imp Pts 1 2 3 1 2

EK2 Knowledge of the interrelations 007 Rx Trip - Stahilization - Recovery/I X between a reactor trip and the following: 2.6 I

EK2.02 Breakers, relays and disconnects

AK3 Knowledge of the reasons for the following responses as they apply to the

008 Pressurizer Vapor Space Accident/3 X Pressurizer Vapor Space Accident:

3.6 I AK3.02 Why PORV or code safety exit temperature is below RCS or PZR temperature

EKl Knowledge of the operational implications of the following concepts as

009 Small Break LOCA/3 X they apply to the small break LOCA: 4.2 I EKI.01 Natural circulation and cooling, including reflux boiling

EK2 Knowledge of the interrelations

011 Large Break LOCA/3 X between the and the following Large Break LOCA: 2.6 I

EK2.02 Pumps

AKI Knowledge of the operational implications of the following concepts as

015 RCP Malfunctions/4 X they apply to Reactor Coolant Pump

4.4 l Malfunctions (Loss of RC Flow): AKI.01 Natural circulation in a nuclear reactor power plant

2.4 Emergency Procedures / Plan

AG2.4.21 Knowledge of the parameters and

022 Loss of Rx Coolant Makeup/2 X logic used to assess the status of safety functions, such as reactivity control, core 4.0 1 cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

AAl Ability to operate and/ or monitor

025 Loss of RHR Systern/4 X the following as they apply to the Loss of

3.6 I Residual Heat Removal System:

AA 1.12 RCS temperature indicators

AAl Ability to operate and/ or monitor the following as they apply to the Loss of

026 Loss of Component Cooling Water/8 X Component Cooling Water:

AA 1.05 The CCWS surge tank, including 3.1 1

level control and level alarms, and radiation alarm

AK3 Knowledge of the reasons for the following responses as they apply to the

027 Pzr Press Control Sys Malfunction/3 X Pressurizer Pressure Control Malfunctions: 3.7 l

AK3.03 Actions contained in EOP for PZR PCS malfunction


Emergency & Abnormal Plant Evolutions - Tier 1 / 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

EA2 Ability to determine or interpret the 029 ATWS/1 X following as they apply to a ATWS: 4.4 I

EA2.0I Reactor nuclear instrumentation

EKl - Knowledge of the operational

038 Steam Gen. Tube Rupture/3 X implications of the following concepts as

3.9 I they apply to the SGTR:

EKI.03 Natural circulation

AA2 Ability to determine and interpret

040 Steam Line Rupture -Excessive Heat the following as they apply to the Steam

X Line Rupture: 4.1 I Transfer/ ( 4)

AA2.05 When ESFAS systems may be secured

EAl Ability to operate and monitor the following as they apply to a Station

055 Station B1ackouU6 X Blackout: 3.3 I EAI.05 Battery, when approaching fully discharged

2.4 Emergency Procedures / Plan 056 Loss of Offsite Power/6 X AG2.4.3 I Knowledge of annunciator alarms, 4.2 I

indications, or response procedures

AA2 Ability to determine and interpret the following as they apply to the Loss of

058 Loss of DC Power/6 X DC Power: 3.5 I AA2.03 DC loads lost; impact on ability to operate and monitor plant systems

2.4 Emergency Procedures / Plan 062 Loss of Nuclear Service Water/4 X AG2.4. I 8 Knowledge of the specific bases 3.3 l

forEOPs

AK3 Knowledge of the reasons for the following responses as they apply to

077 Generator Voltage and Electric Grid Generator Voltage and Electric Grid

Disturbances/6 X Disturbances: 3.6 I

AK3.02 Actions contained in abnormal operating procedure for voltage and grid disturbances

EK2 Knowledge of the interrelations between the (Loss of Feedwater) and the following:

CE/E06 Loss of Main Feedwater/4 X EK2. I Components, and functions of control 3.3 I and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features

I KIA Category Totals: 3 3 3 3 3 3 Group Point Total: 1s I


Emergency & Abnormal Plant Evolutions - Tier 1 / 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

AKl Knowledge of the operational implications of the following concepts as

001 Continuous Rod Withdrawal/I X they apply to Continuous Rod

4.0 1 Withdrawal:

AKl.06 Relationship of reactivity and reactor power to rod movement

AKl Knowledge of the operational implications of the following concepts as

003 Dropped Control Rod/1 X they apply to Dropped Control Rod: 2.8 1 AK 1.15 Definition and application of power defect

AK2 Knowledge of the interrelations between the Loss of Source Range Nuclear

032 Loss of Source Range Nl/7 X Instrumentation and the following: 2.7 1 AK2.0l Power supplies, including proper switch positions

2.1 Conduct of Operations 051 Loss of Condenser Vacuum/4 X AG2. l .27 Knowledge of system purpose 3.9 1

and/or function.

AAl Ability to operate and / or monitor

061 ARM System Alarms/7 X the following as they apply to Area

3.6 I Radiation Monitoring System Alarms:

AAl.01 Automatic actuation

EA2 Ability to determine or interpret the following as they apply to a Inadequate

074 Inadequate Core Cooling/4 X Core Cooling:

EA2.07 The difference between a LOCA and 4.1 1

inadequate core cooling, from trends and indicators

AK2 Knowledge of the interrelations between the (RCS Overcooling) and the following:

CE/Al 1 RCS Overcooling/4 X AK2.1 Components, and functions of control 3.2 1 and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features

AK3 Knowledge of the reasons for the following responses as they apply to the (Natural Circulation Operations)

AK3.4 RO or SRO function within the CE/Al3 Natural Circulation/4 X control room team as appropriate to the 3.1 1

assigned position, in such a way that procedures are adhered to and the limitations in the facilities license and amendments are not violated


Emergency & Abnormal Plant Evolutions - Tier 1 / Group 2 - REACTOR OPERATOR

E/APE #/Name/Safety Function K K K A A G KA Topic Imp Pts

1 2 3 1 2

EK3 Knowledge of the reasons for the following responses as they apply to the (Functional Recovery)

EK3.4 RO or SRO function within the CE/E09 Functional Recovery/None X control room team as appropriate to the 3.3 1

assigned position, in such a way that procedures are adhered to and the limitations in the facilities license and amendments are not violated

KIA Category Totals: 2 2 2 1 1 1 Group Point Total: 9


Plant Systems-Tier 2 / 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 s 6 1 2 3 4

A4 Ability to manually operate

003 Reactor Coolant Pump X and/or monitor in the control room: 2.9 I

A4.02 RCP motor parameters

K2 Knowledge of bus power 003 Reactor Coolant Pump X supplies to the following: 2.5 I

K2.02 CCW pumps

K3 Knowledge of the effect that a 004 Chemical and Volume

X loss or malfunction of the eves

Control will have on the following: 3.7 I

K3.04 RCPS

A4 Ability to manually operate and/or monitor in the control

005 Residual Heat Removal X room: 3.6 I A4.0l Controls and indication for RHRpumps

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

006 Emergency Core Cooling X limits) associated with operating

3.1 I the ECCS controls including:

A 1.0 I Avoidance of thermal and pressure stresses due to pump startup

2.1 Conduct of Operations 007 Pressurizer Quench Tank X 02.1.20 Ability to interpret and 4.6 I

execute procedure steps

K4 Knowledge of PRTS design

007 Pressurizer Quench Tank X feature(s) and/or interlock(s)

2.6 I which provide for the following:

K4.0l Quench tank cooling

K2 Knowledge of bus power

008 Component Cooling Water X supplies to the following:

K2.02 CCW pump, including 3.0 I

emergency backup

K6 Knowledge of the effect of a

0 IO Pressurizer Pressure loss or malfunction of the

X following will have on the PZR 2.7 I Control PCS:

K6.0l Pressure detection systems


Plant Systems -Tier 2 / 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

A2 Ability to (a) predict the impacts of the following malfunctions or operations on the RPS; and (b) based on those predictions, use procedures to

012 Reactor Protection X correct, control, or mitigate the 3.1 I consequences of those malfunctions or operations:

A2.05 Faulty or erratic operation of detectors and function generators

KS Knowledge of the operational implications of the

013 ESFAS X following concepts as they apply

2.9 1 to the ESFAS:

K5.02 Safety system logic and reliability

A3 Ability to monitor automatic

022 Containment Cooling X operation of the CCS, including:

A3.0l Initiation of safeguards 4.1 1

mode of operation

K2 Knowledge of power supplies 022 Containment Cooling X to the following: 3.0 1

K2.01 Containment cooling fans

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

026 Containment Spray X malfunctions or operations:

3.6 1 A2.07 Loss of containment spray pump suction when in recirculation mode, possibly caused by clogged sump screen, pump inlet high temperature exceeded cavitation, voiding), or sump level below cutoff interlock) limit

K3 Knowledge of the effect that

026 Containment Spray X a loss or malfunction of the CSS will have on the following: 4.2 1

K3.02 Recirculation spray system


Plant Systems - Tier 2 / 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

A2 Ability to (a) predict the impacts of the following malfunctions or operations on the MRSS; and (b) based on

039 Main and Reheat Steam X predictions, use procedures to

3.4 1 correct, control, or mitigate the consequences of those malfunctions or operations:

A2.04 Malfunctioning steam dump

2.1 Conduct of Operations 059 Main Feedwater X G2.1.27 Knowledge of system 3.9 1

purpose and/or function

Kl Knowledge of the physical connections and/or cause effect

059 Main Feedwater X relationships between the MFW

3.4 1 and the following systems:

Kl.04 S/GS water level control system

KS Knowledge of the operational implications of the

061 Auxiliary/Emergency X

following concepts as the apply 3.6 1 Feed water to theAFW:

K5.0l Relationship between AFW flow and RCS heat transfer

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

062 AC Electrical limits) associated with operating

Distribution X the ac distribution system 2.5 1 controls including:

Al.03 Effect on instrumentation and controls of switching power supplies

K4 Knowledge of ac distribution system design feature(s) and/or

062 AC Electrical interlock(s) which provide for

Distribution X the following: 2.8 1 K4.03 Interlocks between automatic bus transfer and breakers


063 DC Electrical X

relationships between the DC 2.9 I Distribution electrical system and the

following systems:

Kl.03 Battery charger and battery


Plant Systems - Tier 2 / Group 1 - REACTOR OPERATOR


G KA Topic Imp Pts 1 2 3 4 5 6 1 2 3 4

K6 Knowledge of the effect of a loss or malfunction of the

064 Emergency Diesel X following will have on the ED/G 2.7 1

Generators system:

K6.07 Air receivers

K6 Knowledge of the effect of a

064 Emergency Diesel loss or malfunction of the

X following will have on the ED/G 3.2 1 Generators system:

K6.08 Fuel oil storage tanks

K4 Knowledge of PRM system design feature(s) and/or

073 Process Radiation X

interlock(s) which provide for 3.3 1 Monitoring the following:

K4.02 Letdown isolation on high-RCS activity

2.1 Conduct of Operations

02.1.31 Ability to locate control

076 Service Water X room switches, controls, and 4.6 1 indications, and to determine that they correctly reflect the desired plant lineup

A3 Ability to monitor automatic 078 Instrument Air X operation of the IAS, including: 3.1 1

A3.01 Air pressure


103 Containment X operation of the containment

3.9 I system, including:

A3.01 Containment isolation

I KIA Category Totals: 2 3 2 3 2 3 2 3 3 2 3 Group Point Total: 28 I


Plant Systems -Tier 2 / Group 2 - REACTOR OPERATOR


G KA Topic Imp Pts 1 2 3 4 5 6 1 2 3 4

K6 Knowledge of the effect or a

002 Reactor Coolant X loss or malfunction on the following RCS components: 3.0 I

K6. l 2 Code Safety valves

011 Pressurizer Level Control K2 Knowledge of bus power

System X supplies to the following: 3.1 I K2.02 PZR heaters

KS - Knowledge of the operational implications of the

014 Rod Position Indication X

following concepts as they apply 2.7 I System to the RPIS:

K5.0l Reasons for differences between RPIS and step counter

Al Ability to predict and/or monitor changes in parameters

015 Nuclear Instrumentation X

to prevent exceeding design 2.5 1 System limits) associated with operating

the NIS controls including:

A 1.06 Fuel bum up


016 Non-nuclear operation of the NNIS,

Instrumentation X including: 2.9 I A3.01 Automatic selection of NNIS inputs to control systems


029 Containment Purge X relationships between the

3.6 1 Containment Purge System and the following systems:

Kl.03 Engineered safeguards

K4 Knowledge of design

033 Spent Fuel Pool Cooling feature(s) and/or interlock(s)

System X which provide for the following: 2.9 I K4.0l Maintenance of spent fuel level

A2 Ability to (a) predict the impacts of the following malfunctions or operations on the Fuel Handling System ; and

034 Fuel Handling Equipment X

(b) based on those predictions, 3.4 1 System use procedures to correct,

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

A2.02 Dropped cask


Plant Systems - Tier 2 / Group 2 - REACTOR OPERA TOR


G KA Topic Imp Pts 1 2 3 4 5 6 1 2 3 4

K3 Knowledge of the effect that a loss or malfunction of the

045 Main Turbine Generator X MT/G system will have on the 2.9 1 System following:

K3.01 Remainder of the plant

2.4 Emergency Procedures / Plan

G2.4.4 Ability to recognize 055 Condenser Air Removal X abnormal indications for system 4.5 1 System operating parameters that are

entry-level conditions for emergency and abnormal operating procedures

I KIA Category Totals: 1 1 1 1 1 1 1 1 1 0 1 Group Point Total: 10


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

E/APE #/Name/Safety K K K A A G KA Topic Imp Pts Function 1 2 3 1 2

AA2 Ability to determine and interpret

008 Pressurizer Vapor the following as they apply to the

X Pressurizer Vapor Space Accident: 3.6 1 Space Accident/3

AA2. l 9 PZR spray valve failure, using plant parameters

AA2 . Ability to determine and

015 Reactor Coolant interpret the following as they apply to

X the Reactor Coolant Pump 3.5 1 Pump Malfunctions/4 Malfunctions (Loss of RC Flow):

AA2.0l Cause of RCP failure

2.2 Equipment Control 040 Steam Line

X AG2.2.25 Knowledge of the bases in 4.2 I Rupture (4) Technical Specifications for limiting conditions for operations and safety limits

057 Loss of Vital AC 2.4 Emergency Procedures / Plan Electrical Instrument X AG2.4.3 Ability to identify post-accident 3.9 1 Bus/6 instrumentation

2.4 Emergency Procedures / Plan 077 Generator Voltage AG2.4.47 Ability to diagnose and and Electric Grid X recognize trends in an accurate and 4.2 1 Disturbances/6 timely manner utilizing the appropriate

control room reference material

EA2. Ability to determine and interpret the following as they apply to

CE/E06 Loss of Main X

the (Loss ofFeedwater) 3.9 1

Feedwater/4 EA2. l Facility conditions and selection of appropriate procedures during abnormal and emergency operations

KIA Category Totals: 0 0 0 0 3 3 Group Point Total: 6


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

E/APE #/Name/Safety Kl K2 K3 Al A2 G KA Topic Imp Pts Function

2.1 Conduct of Operations 037 Steam Generator

X AG2. l .28 Knowledge of the purpose and 4.1 l Tube Leak/3 function of major system components and

controls

2.2 Equipment Control

060 Accidental Gaseous AG2.2.25 Knowledge of the bases in

Radwaste Retease/9 X Technical Specifications for limiting 4.2 I

conditions for operations and safety limits

AA2 Ability to determine and interpret the following as they apply to the (RCS

CAI I RCS X Overcooling)

3.3 l Overcooling/4 AA2. l Facility conditions and selection

of appropriate procedures during abnormal and emergency operations

EA2 Ability to determine and interpret the following as they apply to the

CE09 Functional (Functional Recovery)

Recovery/None X EA2.2 Adherence to appropriate 4.0 l procedures and operation within the limitations in the facility's license and amendments

I KIA Category Totals: 0 0 0 0 2 2 Group Point Total: 4 I


Plant Systems - Tier 2 / Group 1 - Senior Reactor Operator

System/Evolution K K K K K K A A A A G KA Topic Imp Pts #/Name 1 2 3 4 5 6 1 2 3 4

A2 Ability to (a) predict the impacts of the following malfunctions or operations on the CVCS; and (b) based on

004 Chemical and X

those predictions, use 3.7 1 Volume Control System procedures to correct, control,

or mitigate the consequences of those malfunctions or operations:

A2.17 Low PZR pressure

2.4 - Emergency Procedures/ Plan

008 Component X Cooling Water System

02.4.50 Ability to verify system alarm setpoints and operate

4.0 1

controls identified in the alarm response manual

2.2 Equipment Control 010 Pressurizer 02.2.25 Knowledge of the bases Pressure Control X in Technical Specifications for 4.2 1 System limiting conditions for

operations and safety limits

A2 - Ability to (a) predict the impacts of the following malfunctions or operations on

013 Engineered Safety the ESFAS; and (b) based on Features Actuation X predictions, use procedures to 4.8 1 System correct, control, or mitigate

the consequences of those malfunctions or operations:

A2.01 LOCA

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

022 Containment X

predictions, use procedures to 3.0 I Cooling System correct, control, or mitigate

the consequences of those malfunctions or operations:

A2.03 Fan motor thermal overload/high-speed operation

KIA Category Totals: 0 0 0 0 0 0 0 3 0 0 2 Group Point Total: 5


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

System/Evolution K K K K K K A A A A G KA Topic Imp Pts #/Name 1 2 3 4 5 6 1 2 3 4

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

016 Non-Nuclear X

those predictions, use 3.1 1

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

A2.01 Detector failure

2.4 - Emergency Procedures / Plan

G2.4.21 Knowledge of the parameters and logic used to

027 Containment Iodine assess the status of safety

Removal System X functions, such as reactivity 4.6 I

control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

A2 · Ability to (a) predict the impacts of the following malfunctions or operations on the ARM system; and (b)

072 Area Radiation X

based on those predictions, 2.9

Monitoring System use procedures to correct, I

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

A2.02 Detector failure

KIA Category Totals: 0 0 0 0 0 0 0 2 0 0 1 Group Point Total: 3


Tier 3 Generic Knowledge & Abilities Outline - RO & SRO

Facility: CAL VERT CLIFFS NUCLEAR POWER PLANT Date of Exam: 08/06/2018

RO SRO Category KIA# Topic

IR Pts IR Pts

2.1.1 Knowledge of conduct of operations requirements 3.8 l

Knowledge of criteria or conditions that require plant-wide 2.1.14 announcements, such as pump starts, reactor trips, mode 3.1 l

changes, etc

Conduct 2.1.8 Ability to coordinate personnel activities outside the control

3.4 1 of room

Operations 2.1.23

Ability to perform specific system and integrated plant 4.4 l

procedures during all modes of plant operation.

2.1.38 Knowledge of the station's requirements for verbal

3.8 l communications when implementing procedures

Subtotals: 3 2

2.2.14 Knowledge of the process for controlling equipment

3.9 l configuration or status

Equipment 2.2.43 Knowledge of the process used to track inoperable alarms 3.0 1

Control Knowledge of limiting conditions for operations and safety 2.2.22

limits 4.7 l

Subtotals: 2 1

2.3.11 Ability to control radiation releases. 3.8 1

Knowledge of radiation or contamination hazards that may 2.3.14 arise during normal, abnormal, or emergency conditions or 3.4 1

activities.

Radiation Ability to use radiation monitoring systems, such as fixed

2.3.5 radiation monitors and alarms, portable survey instruments, 2.9 I Control personal monitoring equipment, etc.

2.3.11 Ability to control radiation releases 4.3 1

2.3.7 Ability to comply with radiation work permit requirements

3.6 1 during normal or abnormal conditions

Subtotals: 3 2

2.4.17 Knowledge of EOP terms and definitions 3.9 l

2.4.43 Knowledge of emergency communications systems and

3.2 1 techniques

Emergency Knowledge of EOP implementation hierarchy and

Procedures/Plan 2.4.16 coordination with other support procedures or guidelines

4.4 1 such as, operating procedures, abnormal operating procedures, and severe accident management guidelines

2.4.19 Knowledge of EOP layout, symbols, and icons 4.1 1

Subtotals: 2 2

I Tier 3 Totals I 10 I 7 I

ES-401 Record of Rejected Kl As Form ES-401-4

I Tier/ Group Randomly Selected KIA Reason for Rejection

RO 022 Loss of Reactor This KIA is not applicable to the RO. Knowledge of Emergency

Coolant Makeup Action Levels and Classification is a SRO responsibility. Kept system

1/1 and replaced KIA with one that had not been sampled. 02.4.21 was 02.4.41 randomly selected using numbered poker chips.

RO 025 Loss of Residual This KIA is not applicable to plant and previously approved for

1/1 Heat Removal System suppression. Kept system and replaced KIA with one that had not been

AAl.19 sampled. AAI.12 was randomly selected using numbered poker chips.

RO This Kl A is similar to test item generated for Kl A 055EA 1.05 and

058 Loss of DC Power would give aid in answering this test item, lowering the level of

1/1 AA2.02 difficulty. Kept system and replaced KIA with one that had not been sampled. AA2.03 was randomly selected using numbered poker chips.

RO 00 I Continuous Rod Spent several unsuccessful hours attempting to develop a question for

Withdrawal this combination of system/KIA. Kept system and replaced KIA with

1/2 AKI.12 one that had not been sampled. AKI.06 was randomly selected using numbered poker chips.

RO Spent several unsuccessful hours attempting to develop a question for

012 Reactor Protection this combination of system/KIA. Kept system and replaced KIA with

2/1 K5.02 one that had not been sampled. A2.05 was randomly selected using numbered poker chips.

RO 013 Engineered Safety KIA category A2 was oversampled due to random selection of KIA

Features Actuation 012A2.04. Kept system and replaced KIA with K5 category to balance

2/1 A2.0l sample plan. K5.02 was randomly selected using numbered poker chips.

This KIA is not applicable to the RO. Knowledge of events that must RO 055 Condenser Air be reported to internal organizations or outside agencies is a SRO

Removal responsibility. Kept system and replaced KIA with one that had not 2/2 02.4.30 been sampled. 02.4.4 was randomly selected using numbered poker

chips.

This KIA is similar to test item generated for Ol5AA2.0I and would SRO 003 Reactor Coolant give aid in answering this test item, lowering the level of difficulty.

Pump Replaced KIA with system and A2 category that had not been sampled. 2/1 A2.0l 004 Chemical and Volume Control System and A2. l 7 was randomly

selected using numbered poker chips.

SRO 027 Containment Iodine Spent several unsuccessful hours attempting to develop a question for

Removal System this combination of system/KIA. Kept system and replaced KIA with

2/2 02.4.4 one that had not been sampled. 02.4.21 was randomly selected by NRC.

Documents

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 · I ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Calvert Cliffs Nuclear Power Plant Date of Examination: