ML041760159

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Examination Outline Submittal with NRC Comments for D.C. Cook Initial Examination - March 2004
ML041760159
Person / Time
Site: Cook  American Electric Power icon.png
Issue date: 03/15/2004
From: Bailey R
American Electric Power Co
To:
NRC/RGN-III
References
50-315/OL04-301, 50-316/OL04-301
Download: ML041760159 (51)
v • d • e


Text

EXAMINATION OUTLINE SUBMITTAL WITH NRC COMMENTS FOR THE D. C. COOK INITIAL EXAMINATION - MARCH 2004

ES-201 Examination Outline Form ES-201-2 Quality Checklist 3

Facility: d & m k

,!-4d L!+Vf? / tdul7 2 Date of Examination: 03/15/-.2dhGp 1

I

1.

W Item a

b' c#

a. Verify that the outiine(s) fit@) the appropriate model per ES-401.

I

,Initial:

Task Description I

T

b. Assess whether the outline was systematically and randomly prepared in accordance with Section D.1 of ES-401 and whether all WA categories are appropriately sampled.

f E

c. Assess whether the outline over-emphasizes any systems, evolutions, or generic topics.
d. Assess whether the justifications for deselected or rejected WA statements are appropriate.

. /"d 4

@'6/.&

2.

S

a. Using Form ES-301-5, verify that the proposed scenario sets cover the required number of normal evolutions, instrument and component failures, and major transients.
c. To the extent possible, assess whether the outline(s) conform(s) with the qualitative and quantitative criteria specified on Form ES-301-4 and described in Appendix D.

I M

3.
b. Assess whether there are enough scenario sets (and spares) to test the projected number and mix of applicants in accordance with the expected crew composition and rotation schedule without compromising exam integrity; ensure each applicant can be tested using at least one new or significantly modified scenario, that no scenarios are duplicated from the applicants' audit test(s)',

and scenarios will not be repeated bver successive days.

2% u w

1 T

a. Verifythat (1) the outJine(s) contain&) the required number of control room and in-plant tasks, (2) no more than 30% of the test material is repeated from the last NRC examination, (3)' no tasks are duplicated from the applicants' audit test(s), and (4) no more than 80% of any operating test is taken directly from the licensee's exam banks.

among the safely function groupings as specified in ES-301, t Author I. Facility Reviewer r)

/

. NRC Chief Examiner (#)

/2/3//$3

i. NRC Supervisor

\\/lrsloy J

L dote:

  • Not applicable for NRC-developed examinations.

t: Independent NRC reviewer initial items in Column "c:" chief examiner concurrence required.

I I

J I

I 23 of 24 NUREG-1021, Revision 8, Supplement 1

NRC Outline Comments ADMIN JPMS A.2(RO), Determine Isolation Boundaries for the Refueling Water Purification Pump. KA 2.1.24 (2.8)

NRC Comment:

Licensee Action:

Change KA from 2.1.24 to Agree. KA changed.

2.2.13, knowledge of tagging and clearance procedures.

SYSTEM JPMs

~

There are 5 alternate path JPMs on the SRO(I) outline, ES-301 requires only 4.

B.le(SRO), Dilute the RCS, WA 004 A4.12, DSL, 1 B.2a(*SRO), Locally Restore CR Ventilation, WA APE 068 AA1.24, NA, 7; B.2c(*SRO), Local Operation and 243), WA APE 068 AA1.O1 4.3/4.5; DR; 4s Of U-I SG PORVS (MRV-213 SCENARIOS Will this evolution be performed in scenarios (ie, power change)?

B.2a(*SRO) and B.2crSRO) have same APE.

Will rewrite or change one alternate path to "non-alternate path."

JPM only performed by SROs.

When SROs are in the RO position, they will not perform this task.

Will rewrite or replace one of JPMs as appropriate.

I Who gets credit for N and R events (not designated in scenarios)?

Identify Tech Specs. Need at least two per scenario.

No 00s equipment identified. Is there some equipment 00s at beginning of scenarios (ie, affect mitigation strategy after EOP entry)?

Generally, BOP will perform the N, and RO will perform the R (ie, does both rods/CVCS and MT load).

Agree.

Generally the same equipment is 00s for scenarios in a set, sometimes equipment will affect the mitigating strategy (not always same equipment).

Page 1 of 2

Scenario 1, Event 3 Scenario 2, Event 3 Scenario 4 (spare):

Scenario 6, Events 4a and 4b WRITTEN EXAM NRC Outline Comments May ask for changes during the validation week to evaluate competencies.

Some events that merely require action to start or shutdown another component may provide adequate evaluation.

Will applicants place excess UD in service?

Only requires RO to identify dropped rod (not able to recover). Replace, or add more analysis/diagnosis/required action.

No N event.

Both are part of the same event, can't credit two events.

No, applicant will have to restore UD.

Agree. Will review during validation.

I ~ocomments I

Page2of 2

ES-401 PWR RO Examination Outline Form ES-401-4 11 Facility: Cook Plant Unit 1 8, Unit 2 Date of Exam: 03/15 - 26/2004 Exam Level: RO I

I I

I Note:

1.
2.

Ensure that at least two topics from every WA category are sampled within each tier (i.e., the Tier Totals in each WA category shall not be less than two).

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 exam must total 100 points.

Select topics from many systems; avoid selecting more than two or three WAtopics from a given system unless they relate to plant-specific priorities.

Systems/evolutions within each group are identified on the associated outline.

The shaded areas are not applicable to the categoryhier.

The generic WAS in Tiers 1 and 2 shall be selected from Section 2 of the WA Catalog, but the topics must be relevant to the applicable evolution or system.

On the following pages, enter the WA numbers, a brief description of each topic, the topics importance ratings for the SRO license level, and the point totals for each system and category. WAS below 2.5 should be justified on the basis of plant-specific priorities. Enter the tier totals for each category in the table above.

3.
4.
5.

6.*

7.

I NUREG-1021, Revision 8, Supplement 1

PWR RO Examination Outline Facility:

DC Cook Nuclear Station E/APE Name / Safety Function K1 ES - 401 K2 K3 A1 A2 G E/APE #

015 024 027 027 040 040 062 074 E06 E06 KA Topic AK1.02 - Consequences of an RCPS failure AKl.04 - Low temperature limits for boron concentration 2.4.1 1 - Knowledge of abnormal condition procedures.

AA2.12 - PZR level AA2.02 - Conditions requiring a reactor trip AK2.01 - Valves AA1.02 - Loads on the SWS in the control room Printed:

12/15/2003 Imp.

Points 3.7 1

2.8 1

3.4 1

3.7 1

4.6 1

2.6*

1 3.2 1

Emergencv and Abnormal Plant Evolutions - Tier 1 / GrouD 1 Form ES-401-4 EK2.05 - LPI pumps 3.9 1

2.1.7 - Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

emergency systems EK1.l - Components, capacity, and function of 3.7 1

3.6 1

1

PWR RO Examination Outline Printed:

12/1 512003 Facility:

DC Cook Nuclear Station ES - 401 E/APE #

E07 E07 E08 E09 EO9 E12 Emei E/APE Name / Safety Function Saturated Core Cooling / 4 Saturated Core Cooling 14 Pressurized Thermal Shock / 4 Natural Circulation Operations / 4 Natural Circulation Operations / 4 Uncontrolled Depressurization of all Steam Generators 14 K/A Category Totals:

3 I

3 3

G KATopic EA1.2 - Operating behavior characteristics of the facility EA2.1 - Facility conditions and selection of appropriate procedures during abnormal and emergency operations EK3.3 - Manipulation of controls required to obtain desired operating results during abnormal, and emergency situations EAI.3 - Desired operating results during abnormal and emergency situations EK3.4 - RO or SRO function within the control room team as appropriate to the assigned position, in such a way that procedures are adhered to and the limitations in the facilities license and amendments are not violated EK2.2 - Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility 2

mP.

3.2 3.2 3.7 3.5 3.4 3.6 Points 1

Group Point Total:

16 2

Facility:

DC Cook Nuclear Station ES - 401 Emereenc EIAPE #

00 1 00 1 007 008 01 1 022 029 037 03 8 03 8

~

EIAPE Name I Safety Function K1 Continuous Rod Withdrawal / 1 Continuous Rod Withdrawal I 1 Reactor Trip / 1 Pressurizer (PZR) Vapor Space Accident (Relief Valve Stuck Open) / 3 Large Break LOCA / 3 Loss of Reactor Coolant Makeup 12 X

Anticipated Transient Without Scram (ATWS) I 1 X

Steam Generator (S/G) Tube Leak / 3 X

PWR RO Examination Outline Printed:

12/15/2003 md Abnormal Plant Evolutions - Tier 1 I Group 2 Form ES-401-4 KA Topic AA2.01 - Reactor tripped breaker indicator AK3.01 - Manually driving rods into position that existed before start of casualty EK2.03 - Reactor trip status panel AA1.01 - PZR spray block valve and PORV block valve EA1.05 - Manual and/or automatic transfer of suction of charging pumps to borated source AK1.04 - Reason for changing from manual to automatic control of charging flow valve controller EK1.01 - Reactor nucleonics and thermo-hydraulics behavior AK1.02 - Leak rate vs. pressure drop EK 1.03 - Natural circulation EK3.06 - Actions contained in EOP for RCS water inventory balance, S/G tube rupture, and plant shutdown procedures mP-4.2 3.2 3.5 4.2 4.3 2.9 2.8 3.5 3.9 4.2

'oints 1

1 1

PWR RO Examination Outline Facility:

DC Cook Nuclear Station ES - 401 Emereencv and Abnormal Plant Evolutions - Tier 1 / Grour, 2 A1 X

X Printed:

12/15/2003 A2 Form ES-401-4 2.7 3.8 3.1 3.3 3.9 3.9 U

I 1

1 1

1 1

1 EIAPE #

EIAPE Name I Safety Function K1 K2 054 Loss of Main Feedwater (MFW) / 4 059 Accidental Liquid Radwaste Release / 9 X

059 Accidental Liquid Radwaste Release I 9 X

I I

I 061 1 Area Radiation Monitoring (ARM) System Alarms / I I

2.4.6 - Knowledge symptom based EOP mitigation strategies.

17 El 1 Loss of Emergency Coolant Recirculation I 4 X

K/A Category Totals:

4 3

G - KA Topic AA 1.04 - HPI, under total feedwater loss conditions AK2.02 - Radioactive-gas monitors AK3.04 - Actions contained in EOP for accidental liquid radioactive-waste release EA1.1 - Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features EK2.2 - Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility

'mp; I,,:,,,

1 Group Point Total:

17 2

PWR RO Examination Outline Facility:

DC Cook Nuclear Station K2 Printed 12/15/2003 K3 A1 A2 G KATopic Imp.

Points AK1.02 - SDM 3.4 1

X AA2.01 - Cause and effect of low-pressure instrument 2.9 1

air alarm X

EK3.4 - RO or SRO function within the control room 3.1 1

team as appropriate to the assigned position, in such a ES - 401 Emergenr E/APE #

036 E/APE Name / Safety Function K1 X

Fuel Handling Incidents / 8 065 E13 Loss of Instrument Air / 8 Steam Generator Overpressure I 4 WACategory Totals:

1 0

1 0

1 0

Group Point Total:

3 1

PWR RO Examination Outline Printed:

1211512003 Form Imp.

4.2 2.G*

2.8 3.3 2.8 4.4 2.7*

2.6 3.3 3.7

3.1 Facility

DC Cook Nuclear Station ES-401-4 Points 1

1 1

1 1

1 1

1 1

1 1

ES - 401 I

I K5 X

(RCPS) 14 Reactor Coolant Pump System I

K6 X

X I (Rcps)14 Chemical and Volume Control System (CVCS) / 1 Chemical and Volume Control System (CVCS) I 1 0 13 Engineered Safety Features Actuation System (ESFAS) I 2 Engineered Safety Features Actuation System (ESFAS) / 2 0 13 0 15 Nuclear Instrumentation System I 7 0 15 Nuclear Instrumentation System I 7 In-Core Temperature Monitor (ITM)

System 1 7 In-Core Temperature Monitor (ITM)

roup 1 KA Topic A 1.09 - Location and interpretation of RCS temperature and pressure indications A4.07 - RCP seal bypass A4.03 - RCP lube oil and lift pump motor controls K2.03 - Charging pumps K4.01-Oxygen control in RCS K3.01 - Fuel KG.01 - Sensors and detectors K6.02 - Discriminator/compensation circuits K2.0 1 - NIS channels, components, and interconnections K5.02 - Saturation and subcooling of water A2.01 - Thermocouple open and short circuits L

1

PWR RO Examination Outline Printed:

12/15/2003

roup 1 Form KA Topic Imp.

A2.04 - Loss of service water 2.9*

A3.01 - Refrigerant system 3.0*

2.1.33 - Ability to recognize indications for 3.4 system operating parameters which are entry-level conditions for technical specifications.

A1.03 - Power level restrictions for operation o MFW pumps and valves K3.01-RCS 4.4 2.7*

2.4.3 1 - Knowledge of annunciators alarms and

3.3 Facility

DC Cook Nuclear Station ES-401-,

Points 1

1 1

1 1

1 jySmv #

022 System I Evolution Name K1 Containment Cooling System (CCS) I 025 025 059 061 061 068 068 071 071 indications, and use of the response instructions.

Ice Condenser System I 5 Ice Condenser System I 5 Main Feedwater (MFW) System 14 Auxiliary / Emergency Feedwater (AFW) System 14 Auxiliary I Emergency Feedwater (AFW) System 14 Liquid Radwaste System (LRS) 19 Liquid Radwaste System (LRS) / 9 X

Waste Gas Disposal System (WGDS) 1 9 Waste Gas Disposal System (WGDS) 19 A3.02 - Automatic isolation K1.07 - Sources of liquid wastes for LRS K4.04 - Isolation of waste gas release tanks K5.04 - Relationship of hydrogerdoxygen concentrations to flammability 2

3.6 1

2.7 1

2.9 1

2.5 1

PWR RO Examination Outline Printed:

12/15/2003 sySmv #

072 072 DC Cook Nuclear Station System / Evolution Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G KATopic Imp.

Points System I 7 Area Radiation Monitoring (ARM)

X A1.O1 - Radiation levels 3.4 1

Area Radiation Monitoring (ARM)

X K1.04 - Control room ventilation 3.3*

1 System 17 WACategory Totals:

2 2

2 2

2 2

3 2

2 2

2 Group Point Total: 23 3

PWR RO Examination Outline Printed:

1211 512003 K5 K6 X

Facility:

Plant A1 X

X CS - 401 iysmv #

002 K3 X

006 K4 006 K1 X

010 K2 X

012 014 026 026 029 035 DC Cook Nuclear Station System I Evolution Name Reactor Coolant System (RCS) I 2 Emergency Core Cooling System (ECCS) 12 Emergency Core Cooling System (ECCS) I 2 Pressurizer Pressure Control System (PZR PCS) I 3 Reactor Protection System I 7 Rod Position Indication System (RPIS) / 1 Containment Spray System (CSS) I 5 Containment Spray System (CSS) / 5 Containment Purge System (CPS) / 8 Steam Generator System (SIGS) I 4 ns - Tiel K3.03 - Containment 4.2 1

I K2.02 - Valve operators for accumulators 2.5*

1 K5.O 1 - Determination of condition of fluid in PZR, using steam tables 3.5 1

A 1.O 1 - Trip setpoint adjustment 2.9*

1 K1.02 - NIS 3.0 1

A2.03 - Failure of ESF 4.1 1

Al.05 - Chemical additive tank level and 3.1 1

concentration X 2.1.23 - Ability to perform specific system and 3.9 1

integrated plant procedures during all modes of plant operation.

A3.01 - S/G water level control 4.0 1

1

PWR RO Examination Outline Printed:

1211 512003 System / Evolution Name K1 K2 K3 K4 K5 Main and Reheat Steam System X

(MRSS) / 4 A.C. Electrical Distribution System 1 6 X A.C. Electrical Distribution System 1 6 X

D.C. Electrical Distribution System / 6 X Emergency Diesel Generator (EDIG)

System 1 6 Emergency Diesel Generator (EDIG)

System I 6 Process Radiation Monitoring (PRM)

X System I 7 Process Radiation Monitoring (PRM)

System I 7 Station Air System (SAS) I 8 X

Fire Protection System (FPS) I 8 X

Facility:

Plant Systems - Tier 2 / Group 2 K6 A1 A2 A3 A4 G KATopic K5.05 - Bases for RCS cooldown limits K1.02 - EDIG K4.03 - Interlocks between automatic bus transfer and breakers K1.03 - Battery charger and battery X

A2.19 - Consequences of high VARS on EDIG integrity X

K6.07 - Air receivers K1.O1 - Those systems served by PRMs X

A4.02 - Radiation monitoring system control panel K4.01 - Cross-connect with IAS K3.01 - Shutdown capability with redundant equipment

S - 401 lys/Ev I#

039 062 062 063 064 064 073 073 079 086 Form Imp.

2.7 4.1 2.8*

2.9 2.5 2.7 3.6 3.7 2.9 2.7 S-401-4 Points 1

1 1

1 1

1 1

1 1 -

1 WACategory Totals:

4 1

2 2

2 1

2 3

1 1

1 Group Point Total: 20 2

PWR RO Examination Outline Printed 12/15/2003

$1 A2 A3 A4 Facility:

ES - 401 G KATopic sys/Ev #

007 008 028 034 04 1 078 078 103 K5 X

DC Cook Nuclear Station K6 I

I

<3 X

System / Evolution Name Pressurizer Relief Tank/Quench Tank System (PRTS) / 5 K4 X

I I

Component Cooling Water System I

X PRT A2.07 - Consequences of high or low CCW Turbine Bypass Control / 4 (CCWS) / 8 Hydrogen Recombiner and Purge Control System (HRF'S) / 5 Fuel Handling Equipment System (FHES) 1 8 Instrument Air System (IAS) / 8 X

Instrument Air System (IAS) / 8 Containment System 1 5 X

WA Category Totals:

0 1

A4.02 - Neutron levels X

Al.01 - Containment pressure, temperature, an humidity I

I I

I 1

1x1 I IA3.01 - Components which discharge to the flow rate and temperature; the flow rate at which the CCW standby pump will start I I I I I K2.01 - Hydrogen recombiners K5.02 - Use of steam tables for saturation temperature and pressure K3.01 - Containment air system K4.02 - Cross-over to other air systems Form Imp.

2.7* -

2.5*

2 9 3.5 2.5 3.1*

3.2 3.7 5-40 1-4 Points 1

1 1

1 1

1 1

1 1

1 1

1 0

Group Point Total:

8 1

Generic Knowledge and Abilities Outline (Tier 3)

PWR RO Examination Outline Conduct of Operations Facility:

DC Cook Nuclear Station Generic Category KA KATopic

2.1.8 Printed

1211 512003 Form ES-401-5 Ability to coordinate personnel activities outside the control room.

Knowledge of less than one hour technical specification action statements for systems.

Ability to apply technical specifications for a system.

Ability to make accurate, clear and concise verbal reports.

Imp.

Points 3.8 1

3.0 1

2.9 1

3.5 1

2.2.12 Equipment Control Knowledge of surveillance procedures.

3.0 1

2.2.24 Ability to analyze the affect of maintenance activities on LCO status.

2.2.27 I 2.6 1 1

Knowledge of the refueling process.

2.6 1

Knowledge of 10 CFR 20 and related facility radiation control requirements.

2.3.9 2.3.10 2.6 1

Knowledge of the process for performing a containment purge.

~ Ability to perform procedures to reduce excessive levels of radiation and guard against personnel exposure.

Emergency ProceduredPlan I

2.5 I

2.9 1

Ability to identify post-accident instrumentation.

Knowledge of the bases for prioritizing emergency procedure implementation during Knowledge of loss of cooling water procedures.

emergency operations.

3.5 1

2.8 1

3.3 1

Generic Total:

13 1

PWR SRO Examination Outline Form ES-401-3 ES-401 1

Facility: Cook Plant Unit 1 & Unit 2 Date of Exam: 03/15 --26/2004 Exam Level: SRO I

I

~~~~

Note:

1.
2.

Ensure that at least two topics from every WA category are sampled within each tier (Le., the Tier Totals in each WA category shall not be less than two).

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 exam must total 100 points.

Select topics from many systems; avoid selecting more than two or three WAtopics from a given system unless they relate to plant-specific priorities.

Systems/evolutions within each group are identified on the associated outline.

The shaded areas are not applicable to the categoryltier.

The generic WAS in Tiers 1 and 2 shall be selected from Section 2 of the WA Catalog, but the topics must be relevant to the applicable evolution or system.

On the following pages, enter the WA numbers, a brief description of each topic, the topics importance ratings for the SRO license level, and the point totals for each system and category. WAS below 2.5 should be justified on the basis of plant-specific priorities. Enter the tier totals for each category in the table above.

3.
4.
5.

6.*

7.

I1 NUREG-I 021, Revision 8, Supplement 1

Facility:

DC Cook Nuclear Station ES - 401 KA Topic AK3.O 1 - Manually driving rods into position that existed before start of casualty AA2.03 - Dropped rod, using in-corelex-core instrumentation, in-core or loop temperature measurements 2.2.25 - Knowledge of bases in technical specifications for limiting conditions for operations and safety limits.

EA2.08 - Conditions necessary for recovery when accident reaches stable phase EA 1.05 - Manual and/or automatic transfer of suction Emer Imp.

3.6 3.8 3.7 3.9*

3.9 SIAPE #

00 1 AK1.02 - Consequences of an RCPS failure AK1.04 - Low temperature limits for boron concentration 003 003 4.1 3.6 01 1 01 1 behavior AA2.05 - When ESFAS systems may be secured AK2.01 - Valves 015 4.5 2.5 024 029 040 040 EIAPE Name I Safety Function Continuous Rod Withdrawal I 1 Dropped Control Rod I 1 Dropped Control Rod I 1 Large Break LOCA I 3 Large Break LOCA 13 Reactor Coolant Pump (RCP) Malfunctions 14 Emergency Boration I 1 Anticipated Transient Without Scram (ATWS) I 1 Steam Line Rupture 14 Steam Line Rupture 14 PWR SRO Examination Outline encv and Abnormal Plant Evolutions - Tier 1 / Grow 1 Printed:

1211 512003 Form ES-401-3 I

I EKl.01 - Reactor nucleonics and thermo-hydraulics I 3.1

'oints 1

1 1

1 1

1

Facility:

DC Cook Nuclear Station ES - 401

(/APE #

059 E/APE Name / Safety Function Accidental Liquid Radwaste Release / 9 Accidental Liquid Radwaste Release / 9 059 062 Loss of Nuclear Service Water / 4 062 Loss of Nuclear Service Water I 4 PWR SRO Examination Outline Emergencv and Abnormal Plant Evolutions - Tier 1 / Grow 1 I

K1 K2 K3 A1 X

X X

I KA Topic AK2.02 - Radioactive-gas monitors AK3.04 - Actions contained in EOP for accidental liquid radioactive-waste release AA2.05 - The normal values for SWS-header flow rate and the flow rates to the components cooled by the sws AA1.02 - Loads on the SWS in the control room EK2.05 - LPI pumps EK3.1 - Facility operating characteristics during transient conditions, including coolant chemistry and the effects of temperature, pressure, and reactivity changes and operating limitations and reasons for these operating characteristics Imp.

2.7 4.3 2.5*

3.3 4.1 3.6 074 I

Inadequate Core Cooling 14 X

Printed:

12/15/2003 Form ES-401-3 E02 SI Termination I 3 X

E04 E06 E06 LOCA Outside Containment / 3 Degraded Core Cooling 14 X

Degraded Core Cooling 14 2.1.25 - Ability to obtain and interpret station reference materials such as graphs, monographs, and tables which contain performance data.

Points 3.1 1

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

4.0 2

2.4.20 - Knowledge of operational implications of EOP warnings, cautions, and notes.

4.0 Printed

1211 512003 PWR SRO Examination Outline K3 X

Facility:

DC Cook Nuclear Station A1 X

X ES - 401 KA Topic EA1.2 - Operating behavior characteristics of the SIAPE ##

E07 E08 E09 E09 E12 Imp.

Points 3.7 1

Emer EIAPE Name I Safety Function Saturated Core Cooling 14 2.4.21 - Knowledge of the parameters and logic used to assess the status of safety functions including: 1.

Reactivity control; 2. Core cooling and heat removal;

3. Reactor coolant system integrity; 4. Containment conditions; 5. Radioactivity release control.

EK3.4 - RO or SRO function within the control room team as appropriate to the assigned position, in such a way that procedures are adhered to and the limitations in the facilities license and amendments are not violated emergency situations EA1.3 - Desired operating results during abnormal and EK2.2 - Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relanons between the proper Pressurized Thermal Shock I 4 4.3 1

3.6 1

3.8 1

3.9 1

Natural Circulation Operations 14 Natural Circulation Operations 1 4 Uncontrolled Depressurization of all Steam Generators 14 enc!

K1 -

WA Category Totals:

4 4

4 I

I facility I

1 4

Group Point Total:

24 3

PWR SRO Examination Outline Facility:

DC Cook Nuclear Station ES - 401 E/APE #

007 007 008 009 009 022 037 037 03 8 03 8 Printed:

1211 512003 Emergencv and Abnormal Plant Evolutions - Tier 1 / Groua 2 Form ES-401-3 I

E/APE Name / Safety Function K1 K2 K3 A1 A2 Reactor Trip / 1 X

Reactor Trip / 1 X

I I

1 I

1 Pressurizer (PZR) Vapor Space Accident (Relief K A Topic Imp.

Points EA2.03 - Reactor trip breaker position 4.4 1

EK2.03 - Reactor trip status panel 3.6 1

AA 1.O 1 - PZR spray block valve and PORV block 4.0 1

valve l

l EK3.20 - Tech-Spec leakage limits 4.3 1

2.1.20 - Ability to execute procedure steps.

4.2 1

AK1.04 - Reason for changing from manual to 3.0 1

automatic control of charging flow valve controller AA2.02 - Agreemenvdisagreement among redundant 3.9 1

radiation monitors AK1.02 - Leak rate vs. pressure drop 3.9 1

EK3.06 - Actions contained in EOP for RCS water inventory balance, SIG tube rupture, and plant shutdown procedures EK1.03 - Natural circulation 1

Facility:

DC Cook Nuclear Station ES - 401 E/APE Name / Safety Function Loss of Main Feedwater (MFW) 14 Loss of Main Feedwater (MFW) 14 PWR SRO Examination Outline Emerpencv and Abnormal Plant Evolutions - Tier 1 / GrouD 2 K1 E/AF'E #

054 054 065 El 1 El 1 E16 G

X KATopic 2.4.46 - Ability to verify that the alarms are consistent with the plant conditions.

AA1.04 - HPI, under total feedwater loss conditions AA2.01 - Cause and effect of low-pressure instnunent air alarm EK2.2 - Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility EA1.1 - Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features EK2.1 - Components, and functions of control and safety systems, including instrumentation, signals, Interlocks, iailure modes, and automatic and manual Loss of Instrument Air I 8 Loss of Emergency Coolant Recirculation 14 1 Loss of Emergency Coolant Recirculation 14 High Containment Radiation / 9 K2 -

K3 -

A1 -

A2 Printed:

12/15/2003 Form ES-401-3 Imp.

3.6 4.5 3.2 4.3 4.0 3.3 Points 1

1 1

WACategory Totals:

3 3

2 3

3 2

Group Point Total:

16 2

Facility:

DC Cook Nuclear Station ES - 401 KA Topic AK1.02 - SDM AA2.13 - Operational status of ventilation supply fans for the service water building, control room and battery Emergenc:

Imp.

Points 3.8 1

2.6 1

E/APE #

036 2.4.4 - Ability to recognize abnormal indications for system operating parameters which are entry-level 056 4.3 1

E15 Loss of Offsite Power I 6 Containment Flooding 1 5 WA Category Totals:

1 PWR SRO Examination Outline Printed:

1211 512003 and Abnormal Plant volutions - Tier 1 / Group 3 Form ES-401-:

I I

room I

I 1

Group Point Total:

3 1

PWR SRO Examination Outline Printed:

12/15/2003 System / Evolution Name Reactor Coolant Pump System (RCPS) 14 Reactor Coolant Pump System (RCPS) / 4 Chemical and Volume Control System (CVCS) I 1 (CVCS) I 1 Chemical and Volume Control System Facility:

K1 ZS - 401

$ys/Ev #

003 003 Engineered Safety Features Actuation System (ESFAS) / 2 Rod Position Indication System (RPIS) / 1 004 X

004 013 013 014 015 017 017 025 DC Cook Nuclear Station Engineered Safety Features Actuation System (ESFAS) / 2 Nuclear Instrumentation System / 7 I In-Core Temperature Monitor (ITM)

System / 7 In-Core Temperature Monitor (ITM)

System I 7 Ice Condenser System / 5

roup 1 KA Topic A4.07 - RCP seal bypass 2.2.17 - Knowledge of the process for managing maintenance activities during power operations.

K2.03 - Charging pumps K4.01 - Oxygen control in RCS K3.01 - Fuel K6.01 - Sensors and detectors K1.02 - NIS K6.02 - Discriminatorlcompensation circuits K5.02 - Saturation and subcooling of water A2.01 - Thermocouple open and short circuits A3.01 - Refrigerant system Form

[mp.

2.6 -

3.5 3.5 -

3.3 4.7 -

3.1*

3.3 2.9 4.0 -

3.5 3.0*

S-401-:

Points 1

1 1

1 1

1 1

PWR SRO Examination Outline Printed:

1211 512003 ystems i 2 X

Facility:

DC Cook Nuclear Station A3 X

ES - 401 sysmv I#

026 059 06 1 06 1 063 068 07 1 072

\\4 Plan1 G KATopic Imp.

4.4 A2.03 - Failure of ESF WACategory Totals:

2 1

2 2

1 2

2 I Pits.

K1.03 - Battery charger and battery A3.02 - Automatic isolation ;

K4.04 - Isolation of waste gas release tanks I

I I

A 1.O 1 - Radiation levels 3.6 S-401-.

Points 1

1 2

Group Point Total: 19 2

PWR SRO Examination Outline Printed:

12/15/2003 K3 X

Facility:

DC Cook Nuclear Station ES - 401 K4 K5 X

X K6 Plant A1 X

I 016 I Non-Nuclear Instrumentation System K1 K2 X

Steam Generator System (SIGS) / 4

/

sySrnv 002 006 010 0 10 System / Evolution Name Reactor Coolant System (RCS) I 2 Emergency Core Cooling System (ECCS) I 2 Pressurizer Pressure Control System (PZR PCS) / 3 Pressurizer Pressure Control System (PZR PCS) 1 3 012 Reactor Protection System I 7 028 034 K3.03 - Containment 4.4

("IS)

/ 7 Hydrogen Recombiner and Purge Control System (HRPS) / 5 Fuel Handling Equipment System (FHES) / 8 X 2.2.18 - Knowledge of the process for 3.6 managing maintenance activities during shutdown operations.

K5.01 - Determination of condition of fluid in PZR, using steam tables 4.0 A 1.O 1 - Trip setpoint adjustment 3.4*

039 X 2.4.45 - Ability to prioritize and interpret the 3.6 significance of each annunciator or alarm.

Main and Reheat Steam System (MRSS) I 4 K2.01 - Hydrogen recombiners 2.8*

A4.02 - Neutron levels 3.9 A3.01 - S/G water level control K5.05 - Bases for RCS cooldown limits S-401-:

Points 1

1 1

1

Printed:

12/15/2003 PWR SRO Examination Outline 062 Facility:

DC Cook Nuclear Station A.C. Electrical Distribution System / 6 ES - 401 I

I 064 I Emergency Diesel Generator (ED/G) I I

I System/6 1 073 I Process Radiation Monitoring (PRM) I X I

I 073 I Process Radiation Monitoring (PRM) I I

I System/7 $

Containment System I 5 WA Category Totals:

2 1

2 1

2 0

2 2

1 2

2 Group Point Total: 17 2

PWR SRO Examination Outline Printed:

12/15/2003 Imp.

4.0 2.8*

2.8 Facility

ES - 401 Points 1

1 1

SysrEv #

008 008 04 1 078 K5 X

K/A Category Totals:

0 K6 DC Cook Nuclear Station K2 (CCWS) 1 8 K3 K4 X

Turbine Bypass Control 1 4 Instrument Air System (US) I 8 il A2 X

rier 2 / Group 3 I

I KA Topic 2.1.33 - Ability to recognize indications for system operating parameters which are entry-level conditions for technical specifications.

A2.07 - Consequences of high or low CCW flow rate and temperature; the flow rate at which the CCW standby pump will start K5.02 - Use of steam tables for saturation temperature and pressure I

I K3.01 - Containment air system 0

1 Group Point Total:

4 1

Generic Knowledge and Abilities Outline (Tier 3)

PWR SRO Examination Outline 2.2.22 Printed: 12/15/2003 Form ES-401-5 Knowledge of limiting conditions for operations and safety limits.

Facility:

DC Cook Nuclear Station Ability to analyze the affect of maintenance activities on LCO status.

~ Knowledge of bases in technical specifications for limiting conditions for operations and safety limits.

Knowledge of the refueling process.

Generic Category Conduct of Operations Equipment Control Radiation Control KA KATopic Imp.

Points 2.1.4 2.1.10 2.1.13 2.1.17 Knowledge of shift staffing requirements.

Knowledge of conditions and limitations in the facility license.

Knowledge of facility requirements for controlling vital / controlled access.

Ability to make accurate, clear and concise verbal reports.

3.4 3.9 2.9 3.6 Category Total:

4 2.2.24 2.2.25 2.2.27 1

1 1

1 -

I 2.3.2 2.3.3 2.3.4 2.3.10 4.1 3.8 3.7 3.5 1

1 1

1 Category Total:

4 Knowledge of facility ALARA program.

Knowledge of SRO responsibilities for auxiliary systems that are outside the control room (e.g., waste disposal and handling systems).

Knowledge of radiation exposure limits and contamination control, including permissible levels in excess of those authorized.

Ability to perform procedures to reduce excessive levels of radiation and guard against personnel exposure.

2.9 2.9 3.1 3.3 1

1 1

1 Category Total:

4 1

Facility:

DC Cook Nuclear Station 2.4.3 2.4.8 2.4.18 2.4.23 2.4.24 Generic Category Emergency ProceduredPlan Ability to identify post-accident instrumentation.

Knowledge of how the event-based emergencylabnomal operating procedures are used in conjunction with the symptom-based EOPs.

Knowledge of the specific bases for EOPs.

Knowledge of the bases for prioritizing emergency procedure implementation during emergency operations.

Knowledge of loss of cooling water procedures.

Generic Knowledge and Abilities Outline (Tier 3)

PWR SRO Examination Outline Printed: 1211 512003 Form ES-401-5 I

mp.

Points 3.8 3.7 3.6 3.8 3.7 1

1 1

1 1 -

Category Total:

5 Generic Total: 17 2

ES-301 Administrative Topics Outline Form ES-301-1 Facility:

DC COOK Date of Examination:

03/15/2004 Examination Level (circle one): @ SRO Operating Test Number: -

Administrative TopiclSubject Description A. 1 A.2 A. 3 A.4 Parameter Verification KIA SYS 001 A4.11 (3.5)

Parameter Verification KIA 2.1.33 (3.4)

Use of P&IDs KA 2.1.24 (2.8)

Control of Radiation Releases KA 2.3.1 1 (2.7)

Emergency Communication KA 2.4.39 (3.3)

Describe method of evaluation:

1. ONE Administrative JPM, OR
2. TWO Administrative Questions NRC2004-Ala Determination of Reactor Shutdown Margin per 02-bHP-4021-001-012.

Classroom / Simulator NRC2004-A1 b Perform Unit 1 LTOP Verification per 01 -0HP-4030-114-030.

Classroom / Simulator NRC2004-A2 Determine Isolation Boundaries for the Refueling Water Purification Pump.

Classroom I Simulator NRC2004-A3 Perform Containment Pressure Relief per 02-OHP-4021-028-004 Simulator NRC2004-A4 Complete EMD-32A Nuclear Plant Event Notification Form per PMP-2080-EPP-107.

Classroom / Simulator NUREG-1 02 1 Revision 8

ES-301 Control Room Svstems and Facilitv Walk-Throuah Test Outline Form ES-301-2 System I JPM Title

a. [NRC2004-SIMOI] Perform Hot Shutdown Panel Pressurizer Backup Heater Operability Test per 02-OHP-4030-214-049. WA SYS 01 OA4.02 3.613.4
b. [NRC2004-SIM02] Perform Subcooling Margin Determination per 02-OHP-4023-SUP-OOI. WA EPE TY Pe Safety Code*

Function N,S 3

D,S,L 7

01 1 EA1.14 3.9/4.1

c. [NRC2004-SIM03] Isolate Ruptured Steam Generator per 02-OHP-4023-E-3. WA EPE 038 EA1.32 4.6/4.7 D S
d. [NRC2004-SIM04] Place Excess Letdown in Service per02-OHP-4021-003-001. WA SYS 004 A4.06 3.6/3.1 4P M,A,S,L 2

N,S

e. [NRC2004-SIM05] Perform Hydrogen Recombiner Functional Test per 02-OHP-4030-STP-013A. WA SYS 028 A4.01 4.0/4.0
f. [NRC2004-SIM06] Synchronize and Load 2AB EDG per 01 -OHP-4030-STP-O27AB, WA SYS 064 A4.06 D,A,S 5

6 3.913.9 (From Cook 2002 NRC.exam)

Alarm per 12-OHP-4022-057-001. WA SYS 075 A2.01 N,A,S,L

g. [NRC2004-SIM09] Respond to Traveling Screen DP 8

D,A

c. [NRC2004-INP06] Perform Local Diesel Generator Trip and Isolation per 02-OHP-4025-LTI-3 (DG Fails to Trip with Pushbutton) APE AA1.31 3.9/4.0 (From Cook 2002 NRC exam)

D D, R

a. [NRC2004-INP03] Shift TDAFP Suction to ESW per
b. [NRC2004-INP05] RCP Seal Injection via CVCS 02-4022-055-003. WA APE 054 AA1.01 4.5/4.4 Cross-Tie to Maintain Pressurizer Level per 01 -0HP-4025-LS-6. APE 022 AAI.01 3.4/3.3 6

4s 2

  • Type Codes: (D)irect from bank, (M)odified from bank, (N)ew, (A)lternate path, (C)ontrol room, (S)imulator, (L)ow-Power, (R)CA NUREG-102 1, Revision 8

ES-301 Administrative Tonics Outline Form ES-301-1 03/15/2004 Examination Level (circle one): RO Operating Test Number Facility:

DC Cook Administrative TopiclSubject Description A. 1 A.2 A.3 A.4 Plant Parameter Verification KIA 2.1.25 (3.1)

EPE 009 EA2.38 14.31 TS Action KA 2.1.I2 (4.0)

Maintenance KA 2.2.17 (3.5)

Control of Radiation Releases KA 2.3.6 (3.1)

EALs &

Class if ications KA 2.4.44 (4.0)

~

Describe method of evaluation:

1. ONE Administrative JPM, OR
2. TWO Administrative Questions NRC2004-A5a Calculate Reactor Vessel Void Vent Time per 02-OHP-4023-FR-I.3.

Classroom / Simulator NRC2004-A5b Verify Appropriate LCO Action for Inoperable Radiation Monitors.

Simulator NRC2004-A6 Perform Screen House Vulnerability Determination per 12-OHP-5030-057-001.

Classroom / Simulator NRC2004-A7 Review Waste Gas Decay Tank Release for SRO Authorization per 12-OHP-4021-023-002.

Classroom / Simulator NRC2004-A8 Perform an Emergency Plan Classification with PAR per PMP-2080-EPP-101.

Classroom / Simulator NUREG-102 1 Revision 8

ES-301 Control Room Systems and Facility Walk-Through Test Outline Form ES-301-2

a. [NRC2004-SIM02] Perform Subcooling Margin Determination per 02-OHP-4023-SUP-OOI. KIA EPE 01 1 EA1.14 3.9/4.1
b. [NRC2004-SIM03] Isolate Ruptured Steam Generator per 02-OHP-4023-E-3. WA EPE 038 EA1.32 4.614.7
c. [NRC2004-SIM05] Perform Hydrogen Recombiner Functional Test per 02-OHP-4030-STP-013A. WA SYS 028 A4.01 4.0/4.0
d. [NRC2004-SIM06] Synchronize and Load 2AB EDG per OI-OHP-4030-STP-O27AB, WA SYS 064 A4.06 3.9/3.9 (From Cook 2002 NRC Exam)
e. [NRC2004-SIM07] Dilute the RCS per 12-OHP-4021-
f. [NRC2004-SIM08] Adjust Pressure in an SI 005-001. WA SYS 004 A4.12 3.3/3.8 Accumulator per 02-OHP-4021-008-006. WA SYS 006 A I.I 3 3.513.7 Alarm per 12-OHP-4022-057-001. WA SYS 075 A2.01 3.013.2
g. [NRC2004-SIM09] Respond to Traveling Screen DP Facility: DC COOK Date of Examination: 03/15/2004 Exam Level (circle one): RO /

B.1 Control Room Systems Operating Test No.:

D,S,L 7

4P N S 5

M,A,S,L D,A, S 6

DS,L I

DS,L 3

N,A,S,L 8

System / JPM Title N,A D,A D,R

a. [NRC2004-INPOI] Locally Restore CR Ventilation per
b. [NRC2004-INP02] Locally Restore CRlD Inverters per
c. [NRC2004-INP04] Local Operation of U-I SG PORVs (MRV-213 and 243) per 01-OHP-4025-LS-4. WA APE 068 AA1.01 4.3/4.5 01-OHP-4025-R-14. APE 068 AA1.24 3.0/3.6 01-OHP-4021-082-008. APE 057 AA1.01 3.7/3.7 1

Type 1

Safety Code*

Function 7

6 4s

  • Type Codes: (D)irect from bank, (M)odified from bank, (N)ew, (A)lternate path, (C)ontrol

// room, (S)imulator, (L)ow-Power, (R)CA NUREG-1021, Revision 8

~

orm 03/15/2004 Examination Level (circle one): RO Operating Test Number Facility:

DC Cook Administrative TopiclSubject Description A. 1 A.2 A.3 A.4 Plant Parameter Verification WA 2.1.25 (3.1)

EPE 009 EA2.38 (4.3)

TS Action KA 2.1.I2 (4.0)

Maintenance KA 2.2.1 7 (3.5)

Control of Radiation Releases KA 2.3.6 (3.1)

EALs &

Classifications KA 2.4.44 (4.0)

Describe method of evaluation:

1. ONE Administrative JPM, OR
2. TWO Administrative Questions NRC2004-A5a Calculate Reactor Vessel Void Vent Time per 02-OHP-4023-FR-1.3.

Classroom / Simulator NRC2004-A5b Verify Appropriate LCO Action for Inoperable Radiation Monitors-Simulator NRC2004-A6 Perform Screen House Vulnerability Determination per 12-OHP-5030-057-001.

Classroom / Simulator NRC2004-A7 Review Waste Gas Decay Tank Release for SRO Authorization per 12-OHP-4021-023-002.

Classroom / Simulator NRC2004-A8 Perform an Emergency Plan Classification with PAR per PMP-2080-EPP-101.

Classroom / Simulator NUREG-102 1 Revision 8

~ ~ _ _ _ _ _ _

~ _ _ _ _

ES-303 Control Room Systems and Facility Walk-Through Test Outline Form ES-301-2 1

Facility: DC COOK Date of Examination:

03/15/2004 Exam Level (circle one): RO I SRO(I) / @

Operating Test No.:

B.2 Facility Walk-Through I/ B.l Control Room Systems N,A

a. [NRC2004-1NP01] Locally Restore CR Ventilation per 01 -0HP-4025-R-14. APE 068 AA1.24 3.0/3.6 System I JPM Title
a. [NRC2004-SIM07] Dilute the RCS per 12-OHP-4021-005-001. WA SYS 004 A4.12 3.3/3.8
b. [NRC2004-SIM08] Adjust Pressure in an SI Accumulator per 02-OHP-4021-008-006. WA SYS 006 A I.I 3 3.513.7
c. [NRC2004-SIM09] Respond to Traveling Screen DP Alarm per 12-OHP-4022-057-001. WA SYS 075 A2.01 3.013.2 7

D,R

b. [NRC2004-1NP04] Local Operation of U-1 SG PORVs (MRV-213 and 243) per 01-OHP-4025-LS-4. WA APE 068 AAI.01 4.3/4.5 Type Code*

4s Safety Function 1

3 8

I I

  • Type Codes: (D)irect from bank, (M)odified from bank, (N)ew, (A)lternate path, (C)ontrol room, (S)imulator, (L)ow-Power, (R)CA NUREG-1021, Revision 8

Cook Plant Scenario Outline Form ES-D-1 Facility: Cook Plant Unit 1 & Unit 2 Scenario No.: COOK04-01 Op-Test No.: Set 1 Event Malf No.

No.

1 2

3 RXO5A@O 11 Examiners:

Event Event Type*

Description N

R I-RO Start South Hotwell pump and Stop Middle Hotwell Pump Raise Reactor Power and Turbine Load Controlling Pressurizer Level Channel (NLP-151) fails LOW Operators:

5 6

CCOIA C-RO East CCW pump trips on overcurrent CC02B West CCW pump fails to AUTO start SW07A C-BOP Main Turbine Oil Cooler Controller (WRV-970) fails LOW (I

4 I RX17F @ 1200 I I-BOP I #22 SG Pressure Instrument (MPP-222) fails HIGH 9

RP16A RP16B C-RO Both trains of CTS/Phase B isolation fails to AUTO actuate 11 7 I RCOlB@50 I

M I RCS Loop 2 Large Break LOCA RP07A II I RPO7B 1 C-BOP I Both trains of Main Steamline Isolation fails to AUTO actuate Cook Plant Unit 1 & Unit 2 NUREG-I 021, Revision 8, Supplement 1

Summary The crew is directed to start the South Hotwell Pump and shutdown the Middle Hotwell Pump. After the Hotwell pumps have been swapped, the crew is required to raise power to 99% for continued operations.

The first event will involve the controlling Pressurizer Level channel instrument (NLP-151) failing LOW. This results in charging flow rising, pressurizer level rising, pressurizer heaters OFF, and letdown isolation. RO will be required to restore normal letdown and charging flow conditions. Crew will be required to implement AOP actions to stabilize the plant and trip Bistables.

The second event will involve the #22 SG Pressure instrument (MPP-222) failing HIGH. This will result in the opening of #22 SG PORV (MRV-220) releasing steam to the atmosphere. BOP will be required to manually close the #22 SG PORV. Crew will be required to implement AOP actions to stabilize the plant, trip bistables, and declare

  1. 22 SG PORV Radiation Monitor Inoperable.

The third event will involve an overcurrent trip of the East CCW pump with failure of the West CCW pump to AUTO start. This will result in a loss of CCW flow until the West CCW pump is started. RO will be required to manually start the West CCW pump.

The fourth event will involve the Main Turbine Oil Cooler Controller failing LOW in AUTO. This will result in the loss of adequate cooling to the Main Turbine Oil Cooler and temperature alarms. BOP will be required to take manual control of WRV-970 and restore temperature control to the Main Turbine Oil Cooler.

The main event will involve a RCS Loop 2 LBLOCA The unit will trip and a Safety Injection will actuate. The CTSRhase B and the Main Steamline Isolation circuits will fail to automatically actuate requiring manual actuations. The crew will transition to from E-0 to E-1 and then to ES-1.3. The scenario will terminate when the crew has aligned RHR and CTS to the containment sump.

Critical Tasks Initiate Containment Sprayphase B Isolation Transfer to Cold Leg Recirculation Procedures E-0, Reactor Trip or Safety Injection E-1, Loss of Reactor or Secondary Coolant ES-1.3, Transfer to Cold Leg Recirculation FR-Z. 1, Response to High Containment Pressure

Cook Plant Scenario Outline Form ES-D-1 Event Malf. No.

Event No.

Type*

1 N

Facility: Cook Plant Unit I 81 Unit 2 Scenario No.: COOK04-02 Op-Test No.:

Set 1 Event Description Start Middle CB pump and place North CB pump in standby Examiners:

2 3a 3b 4

5 6

Operators:

RX20G@5E6 I-BOP

  1. 24 SG Steam Flow Transmitter (MFC-140) fails HIGH NIIOB@200 I-RO Power Range N142 fails HIGH RD0445 C-RO CBD Rod D-4 partial insertion/mechanically bound RDPROD(45) 195 R

C-BOP Lower reactor power and turbine load Main Turbine Steam seal controller (SRV-22) fails in AUTO MS16A@75 30 sec ramp RC02C@30 M

RCS Loop 3 SBLOCA (300 gpm) 2 min ramp Initial Conditions: IC-37, MOL; 99% power, 967 ppm Boron, 8 GWD, Equilibrium Xenon RPI OA RPIOB C-RO AUTO Safety Injection actuation failure RP03A 7 T RP03B

~

[C-RO

~ I AUTO/MANUAL Reactor trip actuation failure (ATWS)

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Cook Plant Unit I

& Unit 2 NUREG-1021, Revision 8, Supplement 1

Summarv The crew is directed to start the Middle Condensate Booster Pump and shutdown the North Condensate Booster Pump. After the Condensate Booster pumps have been swapped, the crew is required to lower power to 95% for Main Turbine Control Valve testing.

The first event will involve the #24 SG Steam Flow instrument (MFC-140) failing HIGH. This will result in the opening of #24 SG FWRV (FRV-240) to raise feedwater flow. BOP will be required to take manual control and regulate FRV-240. Crew will be required to implement AOP actions to stabilize the plant and trip bistables.

The second event will involve the Power Range N42 detector failing HIGH. This results in automatic control rod insertion. CBD Rod D-4 will partially drop to 195 steps and mechanically bind. RO will be required to place the rod control switch in MANUAL and identify misaligned rod. Crew will be required to implement AOP actions to stabilize the plant, trip Bistables, and implement a rapid power reduction to < 75% per Tech Spec.

The third event will involve the Main Turbine Steam Seal controller (SRV-22) failing in AUTO. This will result in an opening of the Steam Dump valve (SRV-25) causing a loss of sealing steam and lowering condenser vacuum. BOP will be required to take manual control, restore sealing steam and condenser vacuum.

The main event will involve an Failure of the RPS function (ATWS) with SBLOCA event. The reactor must be locally tripped and a manual Safety Injection will be required. The crew will implement FR-S.1 actions until reactor is subcritical. The crew will perform the actions of E-0, then transition to E-1. The scenario will terminate when the crew has transitioned to ES-1.2.

Critical Tasks Insert Negative Reactivity Manually actuate Safety Injection Procedures E-0, Reactor Trip or Safety Injection FR-S. 1, Response to Anticipated Trip without Scram E-1, Loss of Reactor or Secondary Coolant NUREG-1021, Revision 8, Supplement 1

Cook Plant Scenario Outline Form ES-D-1 Event No.

1 Facility: Cook Plant Unit 1 & Unit 2 Scenario No.: COOK04-03 Op-Test No.: Set 2 ll Malf. No.

Event Event Type*

Description N

Perform ESW pump strainer functional test Examiners:

5 6

Operators:

CV16A I-RO VCT Level Transmitter (QLC-451) fails HIGH FW05B C-BOP West Main Feed Pump Trip 9

11 2

I RX23B@O I I-BOP I #21 SG Level transmitter fails LOW RPI OB RP11 B RP11 D RCR20 @ 3 I C-RO I PRZ PORV (NRV-153) Leak by (3 gpm)

II I 101 NMOI 53 NMO-153 fails OPEN C-RO II 4

I I R I Reduce reactor power and turbine load for CSD Train B Safety Injection signal failure to actuate in AUTO or Manual mode PRZ PORV (NRV-153) fails OPEN l M I

11 7 I RC17C@95 11 8

1 RP19C I C-RO I Train A K600 Relay failure: CCP to BIT Inlet valve (IMO-255) fails OPEN and 2CD EDG fails to start

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Cook Plant Unit 1 & Unit 2 NUREG-I 021, Revision 8, Supplement 1

Summary The crew is directed to perform the ESW Pump Strainer hctional test. After the ESW Pump Strainer test has been completed, the crew is required to respond to various events.

The first event will involve the #21 Steam Generator Level instnunent (BLP-111) failing LOW. This results,in a feedwater flow and Steam Generator level rise. BOP will be required to take manual control and restore normal level. Crew will be required to implement AOP actions to stabilize the plant and trip Bistables.

The second event will involve the Pressurizer PORV NRV-153 drifting partially open with the Block valve o\\JMO-153) failing to close. This will result in the loss of capability to stop NRV-153 leakage. Crew will be required to implement a power reduction to cold shutdown conditions.

The third event will involve the VCT Level instrument (QLC-45 1) failing HIGH. This will result in the VCT High Level alarm. RO will be required to perform a manual blend to RCS.

The fourth event will involve a trip of the West Main Feed Pump. This will result in a rapid power reduction to less than 60%. RO will be required to control reactivity while BOP lowers turbine load. Crew will be required to monitor plant conditions during power reduction.

The main event will involve the opening of NRV-153 to fill open. A unit trip will be required. As the crew performs the actions of E-0, the Train B Safety Injection signal will fail to AUTO/MANUAL actuate. In addition, a K600 relay failure will prevent High Head Injection flow requiring manual actions to align the BIT. The crew will transition to E-1. The scenario will terminate when the crew has transitioned to ES-1.2 for Post LOCA cooldown.

Critical Tasks Establish injection flow from at least one (1) high head ECCS pump Stop all running RCPs when RCS pressure is below 1300 psig Procedures E-0, Reactor Trip or Safety Injection E-1, Loss of Reactor or Secondary Coolant

Cook Plant Scenario Outline Form ES-D-1 Facility: Cook Plant Unit 1 & Unit 2 Scenario No.: COOK04-04 Op-Test No.:

Spare Examiners:

Operators:

Initial Conditions: IC-36, MOL; 79% power, 1013 ppm Boron, 8 GWD, Xenon building in Turnover: Power Escalation in progress; PRZ PORV NRV-153 leakage with Block valve NMO-153 closed Event 1

2 3

4a 4b 5a 5b 7

l 8 Malf. No.

RX18 @ 1200 FW34A FW58B RXOSA @ 10 ED07B RC23C @ 30 2 min ramp MS04D @ 20 2 min delay RPO7A RP07B RP20B (I

Event Type*

R I-BOP C-BOP I-RO C-RO M

M C-BOP C-BOP 1)nstrume Event Description

~

Raise reactor power and turbine load Main Steam Bypass Header Pressure transmitter (UPC-101) fails HIGH North CB pump trip; Middle CB pump fails to start in AUTO PRZ Pressure Master Controller fails LOW Loss of Bus 21 PHC SGTR on #23 SG (300 gpm)

Steam Line Break outside Containment (#24 SG) - downstream of MSlV Steam Line Isolation fails to AUTO actuate SI Train B - K609 Relay fails to actuate

, (C)omponent, (M)ajor Cook Plant Unit 1 8. Unit 2 NUREG-I 021, Revision 8, Supplement 1

Summary The crew is directed to raise power to 100% for continued operations.

The first event will involve the Main Steam Bypass Header pressure instrument (UPC-101) failing HIGH. This results in raising both Feed pumps speed and flow output with corresponding rise in steam generator level. BOP.

will be required to take manual control of feed pump speed.

The second event will involve a trip of the North Condensate Booster pump. This will result in reduced feedwater capability. BOP will be required to manually start the Middle Condensate Booster pump. Crew will be required to implement compensatory actions to stabilize the plant.

The third event will involve the Pressurizer Pressure Master Controller failing LOW and a loss of Bus 21PHC. This will result in < 50% Pressurizer Heaters capability. RO will be required to take manual control of Pressurizer Pressure Master Controller and restore normal pressure control. The Crew will be required to implement compensatory actions to restore power to Bus 2 lPHC and recover full heater capability.

The main event will involve a Main Steamline Break outside containment on the #24 SG (isolatable). Failure of the Main Steamline Isolation actuation circuit will require a manual actuation. The unit will trip and a Safety Injection will actuate. As the crew performs the actions of E-0, they should identify the Steam Generator Tube Rupture on

  1. 23 SG and the failure of the 2AB EDG to auto start. The crew will transition to E-3 to isolate #23 SG. The scenario will terminate when the crew has terminated SI injection.

Critical Tasks Isolate #23 Steam Generator Cooldown and Depressurize RCS to stop #23 SGTR Terminate SI injection Procedures E-0, Reactor Trip or Safety Injection E-3, Steam Generator Tube Rupture

Cook Plant Scenario Outline Form ES-D-1

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Malf. No.

Facility: Cook Plant Unit 1 & Unit 2 Scenario No.: COOK04-05 Op-Test No.:

Set I Examiners:

Operators:

Initial Conditions: IC-16, BOLI: 79% power, 150 MWD, 779 ppm Boron, Xenon Equilibrium Turnover: Thunderstorm Warnings with 2CD EDG Out of Service for fuel rack repairs; Swap NESW

~~

~

~~

~

~

Event Event Type*

Description N

Start South NESW pump and Stop North NESW pump pumps.

Event No.

1 M

M 2

Electrical Grid Load Rejection Loss of All AC Power (2 min delay) 3 4

5 6

7a 7b 8

9 RX19A @ 0 1

I-RO I Turbine Impulse Transmitter (MPC-253) fails LOW cvo9 I

C-RO I Letdown Temperature Controller fails LOW RX25A @ 100 I I-BOP I #21 Main Feed Reg Valve (FRV-210) controller oscillation FW12A @ 10 I C-BOP I East MFP lube oil leak 1

R 1 Perform Rapid Power Reduction ED19 @ 25 ED01 ED25 EGO1 EG08A 1 C-BOP I 2AB EDG Speed Governor failure FW48C 1 C-BOP I TD AFW pump fails to AUTO start I

I Cook Plant Unit I & Unit 2 NUREG-I 021, Revision 8, Supplement 1

Summary The crew is directed to start the South NESW Pump and shutdown the North NESW Pump. After the NESW pumps have been swapped, the crew is required to maintain power at 79%.

The first event will involve the Turbine Impulse instrument (MPC-253) failing LOW. This results in a control rod insertion. RO will be required to place rod control in manual. Crew will be required to implement AOP actions to stabilize the plant and trip Bistables.

The second event will involve the Letdown Temperature Controller failing LOW in AUTO. This will result in the loss of letdown cooling. RO will be required to take manual control of CRV-470 and restore temperature to normal band.

The third event will involve the #21 SG Main Feedwater Regulating Valve (FRV-210) controller oscillating. This will result in the opening and closing of FRV-210. BOP will be required to take manual control of FRV-210 to restore normal feed flow capability and restore SG level to program. The Crew will be required to implement actions to stabilize the unit.

The fourth event will involve an oil leak on the East MFP. This will result in a requirement to shutdown the pump.

The Crew will be required to implement a power reduction to lower power to < 60%.

The main event will involve a GRID load rejection leading to a total loss of all AC power. The unit will trip. Failure of the only available EDG will require entry into ECA-0.0 actions. Failure of the TD AFW pump to auto start will require a manual start to restore feedwater flow. The crew will be required to take actions to restore emergency (EP) power. The crew should transition to ECA-0.1 once power has been restored to one Safeguards Bus. The scenario will terminate when the crew has restored emergency power and transitioned to ECA-0.1.

Critical Tasks Establish AFW (Start TDAFWP)

Restore Emergency Power to Safeguards Bus Procedures E-0, Reactor Trip or Safety Injection ECA-0.0, Loss of All AC Power

~~~

Cook Plant 9

Scenario Outline Form ES-D-1 FW51 C-BOP TDAFW pump T&W failure due to blown control power fuse Facility: Cook Plant Unit 1 & Unit 2 Scenario No.: COOK04-06 Op-Test No.:

Set 2 Examiners:

Operators:

Initial Conditions: IC-17, BOLI: 99% power, 150 MWD, 729 ppm Boron, Xenon building in Turnover: Swap CCP pumps 1

R I Lower Reactor Power and Turbine Load ZD I 1 0 I TA2 4a I CLOSE C-BOP West MD AFW pump inadvertent start l

l 4b I RX21A @ 0 I I-BOP I SG Feed Flow Instrument (FFC-210) fails LOW 5

I CV13A I C-RO 1 East CCPpumptrips EDOSH EDOSD Loss of Bus T21 D due to fault on Bus 2D 2D RCP Bus Fault (Loss of RCS Flow in Loop I ) - momentary i

M l

RPOIA RPOI B I C-RO I Reactor Trip (RTA) Fails to AUTO actuate Reactor Trip (RTB) Fails to AUTO actuate 8

1 TC03 I C-BOP I Turbine trip fails to AUTO actuate Cook Plant Unit 1 & Unit 2 NUREG-1021, Revision 8, Supplement 1

Summary The crew is directed to start the East Charging Pump and shutdown the West Charging Pump. After the Charging pumps have been swapped, the crew is directed to reduce power to 80%.

The first event will involve the RCS Loop 4 Cold Leg Temperature instrument (NTP-241) failing HIGH.' This'will result in the AUTO insertion of control rods and a lower trip setpoint value for OPAT and OTAT. The RO will be required to take manual control of rods to stop insertion. Crew will be required to implement AOP actions to stabilize the plant and trip bistables.

After a small power change, the next event will involve the inadvertent start of the West MD AFW pump and a corresponding failure of #21 Steam Generator Feed Flow instrument (FFC-210) LOW. This will result in a rise in feedwater flow to #21 & #24 SGs with corresponding SG level rise. The BOP will be required to take manual control of FRV-210. Crew will be required to implement AOP actions to stabilize the plant, stop the West MD AFW pump locally, and trip bistables.

The fourth event will involve the East Charging Pump trip. This will result in the loss of charging and letdown. RO will be required to manually start the West Charging Pump to restore charging flow. Crew will be required to implement AOP actions to stabilize the plant.

The main event will involve the loss of RCP Bus 2D and Safety Bus T21D requiring a reactor trip. Failure of the Reactor Protection circuit will require a manual reactor trip and failure of the main turbine trip circuit will require a manual turbine trip. As the crew performs the actions of E-0, they should identify the failure to provide adequate feed flow. The crew should transition to ES-0.1 with CSFST for Heat Sink being RED. The crew will immediately transition to FR-H. 1 and re-establish feedwater flow using the TD AFW pump. The scenario will terminate when the crew has established adequate feed flow.

Critical Tasks Perform Manual Reactor Trip Establish AFW flow for Secondary Heat Sink Procedures E-0, Reactor Trip or Safety Injection FR-H. 1, Loss of Secondary Heat Sink

TierIGroup Randomly Selected Reason For Rejection WA 1 / 2 1 / 2 000001 AA2.01 ## Unable to write question of appropriate difficulty. Replace With WA 000001 AA2.03 000037 AK 1.O 1

  • This WA is too close to 04 1000 K5.O 1. Would have resulted in testing same knowledge. Replaced with WA 000037 AK1.02 I / 2 2 / 2 2 / 2 2 / 2 2 / 3 1 / 1 1 / 1 2 / 1 2 / I 2 / 1 000038 EKI.01
  • 01000 K5.02
  • 055000 A3.03
  • 027000 K2.01
  • 034000 K6.02
  • 00WE06 2.1.2 000027 2.4.14 #

022000 A2.06 025000 2.1.18 #

061000 2.4.17 #

This WA is too close to 04 1000 K5.O 1. Would have resulted in testing same knowledge. Replaced with WA 000038 AK1.03 This WA appeared on the Audit examination. The knowledge required is narrowly defined and so the question would be a repeat. Replaced with WA 0 10000 K5.O 1.

DC Cook does NOT auto divert CARS Exhaust. Replaced with WA 035000 A3.01 Containment Iodine Removal System NOT Used at DC Cook (CTS Only). Replaced with WA 028000 K2.01 Fuel Handling System is not tied to Radiation monitors at Cook. Impact is to Stop. Unable to develop a question with plausible distracters of sufficient difficulty (>2). Replaced with WA 034000 A4.02 Unable to develop a question with plausible distracters of sufficient difficulty (>2). Replaced with WA OOWE06 2.1.7 WA addresses EOP flowchart but event is AOP. Replaced with WA 000027 2.4.11 DC Cook Containment Cooling Systems does NOT have a pump. Replaced with KIA 022000 A2.04 Ice Condenser And Logs. Unable to develop a question with plausible distracters of sufficient difficulty (>2).

Replaced with WA 025000 2.1.33 AFW and EOP terms. Unable to develop a question with plausible distracters of sufficient difficulty (>2).

Redaced with WA 061000 2.4.3 I NLTREG - 1021 Revision 8, Supplement 1 2 / 2 026000 K2.02 CTS MOV power supply knowledge. Unable to develop a question with plausible distracters of sufficient difficultv (>2). Replaced with WA 026000 Al.05 2 / 2 029000 2.1.24 #

Containment Purge and prints. Unable to develop a question with plausible distracters of sufficient difficulty 1(>2). Replaced with WA 02900 2.1.23 2 / 2 3

079000 A4.01 IAS is Tied to SAS with Check Valves that can NOT be monitored from Control Room. Replaced with WA 079000 K4.01.

Question #98 was at SRO level. Use Question on SRO exam and replace with SRO question #13 1 with WA 2.4.18 that was written at RO Level.

  • = Common to Both RO and SRO exams
  1. = Justification not required per NRC website WA Guidance.
    1. = Changed based on written Examination Review.

2.2.12 ##

Form ES-401-10 ES-401 SRO Exam Record of Rejected KlAs Tier/Group 1 / 2 1 / 2 2 / 2 2 / 2 2 / 2 2 / 3 1 / 1 1 / 1 1 / 1 1 / 2 1 / 3 Randomly Selected Reason For Rejection WA 000037 AKl.01

  • 000038 EK1.O1
  • 01000 K5.02
  • 034000 K6.02
  • 055000 A3.03
  • 027000 K2.01
  • 00WE07 2.2.17 #

This WA is too close to 041000 K5.01. Would have resulted in testing same knowledge. Replaced with WA 000037 AK1.02.

This WA is too close to 041000 K5.01. Would have resulted in testing same knowledge. Replaced with WA 000038 AK1.03.

This WA appeared on the Audit examination. The knowledge required is narrowly defrned and so the question would be a repeat. Replaced with WA 0 10000 K5.O 1.

Fuel Handling System is not tied to Radiation monitors at Cook. Impact is to Stop. Unable to develop a question with plausible distracters of sufficient difficulty (>2). Replaced with WA 034000 A4.02 DC Cook does NOT auto divert CARS Exhaust. Replaced with WA 035000 A3.01 Containment Iodine Removal System NOT Used at DC Cook (CTS Only). Replaced with WA 028000 K2.01.

Saturated Core Cooling with Managing Maintenance. Replaced with WA 00WE06 2.4.20. (Saturated Core Cooling procedure does NOT contain Notes or Cautions)

PTS with configuration changes. Replaced with WA OOWEO8 2.4.21.

Large Break LOCA with Temporary Changes. Replaced with WA OOWE04 2.1.25 Small Break LOCA with Temporary Changes. Replaced with WA 000009 2.1.20 Unable to develou SRO level Question. Reulaced with WA 000056 A2.13 OOWEOS 2.2.14 #

00WE04 2.2.1 1 ##

000009 2.2.11 #

000056 A2.05 11 2 / 1 IO03000 2.1.13 #

IRCPS and vital area access. Reulaced with WA 003000 2.2.17 2 / 1 3

3 3

061000 2.2.27 #

2.1.20 2.4.34 2.4.18 ##

AFW and Refueling process. Replaced with WA 061000 2.2.3 Unable to develop SRO level question. (Also 2.1.20 already used). Replaced with WA 2.1.13 Unable to develop SRO level question. (Also 2.4.34 appears on Audit Examination) Replaced with WA 2.4.18.

Question #13 1 was at RO level. Use Question on RO exam and replace with RO question #98 with WA 2.2.12 that was written at SRO Level

  • = Common to Both RO and SRO exams # = Justification not required per NRC website WA Guidance 11 I

I## = Changed based on written Examination Review.

2 / 1 3

3 3

NUREG - 1021 Revision 8, Supplement 1 061000 2.2.27 #

2.1.20 2.4.34 2.4.18 ##

AFW and Refueling process. Replaced with WA 061000 2.2.3 Unable to develop SRO level question. (Also 2.1.20 already used). Replaced with WA 2.1.13 Unable to develop SRO level question. (Also 2.4.34 appears on Audit Examination) Replaced with WA 2.4.18.

Question #13 1 was at RO level. Use Question on RO exam and replace with RO question #98 with WA 2.2.12 that was written at SRO Level

  • = Common to Both RO and SRO exams # = Justification not required per NRC website WA Guidance
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