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

Task Description I

I

,Initial:

a b' c#

I

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

W T

f E

I

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.
c. Assess whether the outline over-emphasizes any systems, evolutions, or generic topics.

. /"d 4 I

d. Assess whether the justifications for deselected or rejected WA statements are appropriate.

@'6/ .& , J

2. 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.

S I b. Assess whether there are enough scenario sets (and spares) to test the projected number and M 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

c. To the extent possible, assess whether the outline(s) conform(s) with the qualitative and quantitative criteria specified on Form ES-301-4and described in Appendix D.
3. a. Verifythat (1) the outJine(s) contain&) the required number of control room and in-plant tasks, w (2) no more than 30% of the test material is repeated from the last NRC examination, 1 (3)' no tasks are duplicated from the applicants' audit test(s), and T (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
i. NRC Supervisor J L \/lrsloy dote: Not applicable for NRC-developed examinations.

t: Independent NRC reviewer initial items in Column "c:" chief examiner concurrence required. I 23 of 24 NUREG-1021, Revision 8, Supplement 1

NRC Outline Comments ADMIN JPMS NRC Comment: Licensee Action:

A.2(RO), Determine Isolation Change KA from 2.1.24 to Agree. KA changed.

Boundaries for the Refueling 2.2.13, knowledge of Water Purification Pump. KA tagging and clearance 2.1.24 (2.8) procedures.

SYSTEM JPMs

~

There are 5 alternate path Will rewrite or change one JPMs on the SRO(I) alternate path to "non-outline, ES-301 requires alternate path."

only 4.

B.le(SRO), Dilute the RCS, Will this evolution be JPM only performed by SROs.

WA 004 A4.12, DSL, 1 performed in scenarios (ie, When SROs are in the RO power change)? position, they will not perform this task.

B.2a(*SRO), Locally Restore B.2a(*SRO) and Will rewrite or replace one of CR Ventilation, WA APE 068 B.2crSRO) have same JPMs as appropriate.

AA1.24, NA, 7; APE.

B.2c(*SRO), Local Operation Of U-I SG PORVS (MRV-213 and 243), WA APE 068 AA1.O1 4.3/4.5; DR; 4s SCENARIOS I

Who gets credit for N and Generally, BOP will perform R events (not designated in the N, and RO will perform the scenarios)? R (ie, does both rods/CVCS and MT load).

Identify Tech Specs. Need Agree.

at least two per scenario.

No 00s equipment Generally the same identified. Is there some equipment is 00s for equipment 00s at scenarios in a set, sometimes beginning of scenarios (ie, equipment will affect the affect mitigation strategy mitigating strategy (not always after EOP entry)? same equipment).

Page 1 of 2

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.

Scenario 1, Event 3 Will applicants place No, applicant will have to excess UD in service? restore UD.

Scenario 2, Event 3 Only requires RO to Agree. Will review during identify dropped rod (not validation.

able to recover). Replace, or add more analysis/diagnosis/required action.

Scenario 4 (spare): No N event.

Scenario 6, Events 4a and Both are part of the same 4b event, can't credit two events.

WRITTEN EXAM I ~ocomments I Page2of 2

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

Note: 1. 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).

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 exam must total 100 points.
3. Select topics from many systems; avoid selecting more than two or three WAtopics from a given system unless they relate to plant-specific priorities.
4. Systems/evolutionswithin each group are identified on the associated outline.
5. The shaded areas are not applicable to the categoryhier.

6.* 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.

7. 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.

I NUREG-1021, Revision 8, Supplement 1

PWR RO Examination Outline Printed: 12/15/2003 Facility: DC Cook Nuclear Station ES - 401 Emergencv and Abnormal Plant Evolutions - Tier 1 / GrouD 1 Form ES-401-4 E/APE # E/APE Name / Safety Function K1 K2 K3 A1 A2 G KA Topic Imp. Points 015 AK1.02 - Consequences of an RCPS failure 3.7 1 024 AKl.04 - Low temperature limits for boron 2.8 1 concentration 027 2.4.1 1 - Knowledge of abnormal condition procedures. 3.4 1 027 AA2.12 - PZR level 3.7 1 040 AA2.02 - Conditions requiring a reactor trip 4.6 1 040 AK2.01 - Valves 2.6* 1 062 AA1.02 - Loads on the SWS in the control room 3.2 1 074 EK2.05 - LPI pumps 3.9 1 E06 2.1.7 - Ability to evaluate plant performance and make 3.7 1 operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

E06 EK1.l - Components, capacity, and function of 3.6 1 emergency systems 1

PWR RO Examination Outline Printed: 12/1512003 Facility: DC Cook Nuclear Station ES 401 Emei I

E/APE # E/APE Name / Safety Function G KATopic mP. Points E07 Saturated Core Cooling / 4 EA1.2 - Operating behavior characteristics of the 3.2 1 facility E07 Saturated Core Cooling 1 4 EA2.1 - Facility conditions and selection of appropriate 3.2 procedures during abnormal and emergency operations E08 Pressurized Thermal Shock / 4 EK3.3 - Manipulation of controls required to obtain 3.7 desired operating results during abnormal, and emergency situations E09 Natural Circulation Operations / 4 EAI.3 - Desired operating results during abnormal and 3.5 emergency situations EO9 Natural Circulation Operations / 4 EK3.4 - RO or SRO function within the control room 3.4 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 E12 Uncontrolled Depressurization of all Steam EK2.2 - Facility's heat removal systems, including 3.6 Generators 1 4 primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility K/A Category Totals: 3 3 3 2 Group Point Total: 16 2

PWR RO Examination Outline Printed: 12/15/2003 Facility: DC Cook Nuclear Station ES - 401 Emereenc md Abnormal Plant Evolutions - Tier 1 I Group 2 Form ES-401-4

~

EIAPE # EIAPE Name I Safety Function K1 KA Topic mP- 'oints 00 1 Continuous Rod Withdrawal / 1 AA2.01 - Reactor tripped breaker indicator 4.2 1 00 1 Continuous Rod Withdrawal I 1 AK3.01 - Manually driving rods into position that 3.2 1 existed before start of casualty 007 Reactor Trip / 1 EK2.03 - Reactor trip status panel

- 3.5 008 Pressurizer (PZR) Vapor Space Accident (Relief AA1.01 - PZR spray block valve and PORV block 4.2 Valve Stuck Open) / 3 valve 01 1 Large Break LOCA / 3 EA1.05 - Manual and/or automatic transfer of suction 4.3 of charging pumps to borated source 022 Loss of Reactor Coolant Makeup 1 2 X AK1.04 - Reason for changing from manual to 2.9 automatic control of charging flow valve controller 029 Anticipated Transient Without Scram (ATWS) I 1 X EK1.01 - Reactor nucleonics and thermo-hydraulics 2.8 behavior 037 Steam Generator (S/G) Tube Leak / 3 X AK1.02 - Leak rate vs. pressure drop 3.5 038 EK 1.03 - Natural circulation

- 3.9 038 EK3.06 - Actions contained in EOP for RCS water 4.2 inventory balance, S/G tube rupture, and plant shutdown procedures 1

PWR RO Examination Outline Printed: 12/15/2003 Facility: DC Cook Nuclear Station ES - 401 Emereencv and Abnormal Plant Evolutions - Tier 1 / Grour, 2 U "

Form ES-401-4 EIAPE #

054 I

EIAPE Name I Safety Function Loss of Main Feedwater (MFW) / 4 K1 K2 A1 A2 G KA Topic X

AA 1.04 - HPI, under total feedwater loss conditions

'mp; I,,:,,,

059 Accidental Liquid Radwaste Release / 9 X AK2.02 - Radioactive-gas monitors 2.7 1 059 Accidental Liquid Radwaste Release I 9 AK3.04 - Actions contained in EOP for accidental 3.8 1 liquid radioactive-waste release I I I 061 1 Area Radiation Monitoring (ARM) System Alarms / I I X 2.4.6 - Knowledge symptom based EOP mitigation 3.1 1 strategies.

17 3.3 1 X EA1.1 - Components, and functions of control and 3.9 1 safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features El 1 Loss of Emergency Coolant Recirculation I 4 X EK2.2 - Facility's heat removal systems, including 3.9 1 primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility K/A Category Totals: 4 3 1 Group Point Total: 17 2

PWR RO Examination Outline Printed 12/15/2003 Facility: DC Cook Nuclear Station ES - 401 Emergenr E/APE # E/APE Name / Safety Function K1 K2 K3 A1 A2 G KATopic Imp. Points 036 Fuel Handling Incidents / 8 X AK1.02 - SDM 3.4 1 065 Loss of Instrument Air / 8 X AA2.01 - Cause and effect of low-pressure instrument 2.9 1 air alarm E13 Steam Generator Overpressure I 4 X EK3.4 - RO or SRO function within the control room 3.1 1 team as appropriate to the assigned position, in such a WACategory Totals: 1 0 1 0 1 0 Group Point Total: 3 1

PWR RO Examination Outline Printed: 1211512003 Facility: DC Cook Nuclear Station ES 401 I ;roup 1 Form ES-401-4 I I K5 K6 KA Topic Imp. Points A 1.09 - Location and interpretation of RCS 4.2 1 temperature and pressure indications A4.07 - RCP seal bypass 2.G* 1 (RCPS) 1 4 Reactor Coolant Pump System A4.03 - RCP lube oil and lift pump motor 2.8 1 I (Rcps)14 Chemical and Volume Control System controls K2.03 - Charging pumps 3.3 1 (CVCS) / 1 Chemical and Volume Control System K4.01- Oxygen control in RCS 2.8 1 (CVCS) I 1 0 13 Engineered Safety Features Actuation K3.01 - Fuel 4.4 1 System (ESFAS) I 2 0 13 Engineered Safety Features Actuation X KG.01 - Sensors and detectors 2.7* 1 System (ESFAS) / 2 0 15 Nuclear Instrumentation System I 7 X K6.02 - Discriminator/compensationcircuits 2.6 1 L

0 15 Nuclear Instrumentation System I 7 K2.0 1 - NIS channels, components, and 3.3 1 interconnections In-Core Temperature Monitor (ITM) X K5.02 - Saturation and subcooling of water 3.7 1 System 1 7 In-Core Temperature Monitor (ITM) A2.01 - Thermocouple open and short circuits 3.1 1 1

PWR RO Examination Outline Printed: 12/15/2003 Facility: DC Cook Nuclear Station

roup 1 Form ES-401-,

jySmv# System I Evolution Name K1 KA Topic Imp. Points 022 Containment Cooling System (CCS) I A2.04 - Loss of service water 2.9* 1 025 Ice Condenser System I 5 A3.01 - Refrigerant system 3.0* 1 025 Ice Condenser System I 5 2.1.33 - Ability to recognize indications for 3.4 1 system operating parameters which are entry-level conditions for technical specifications.

059 Main Feedwater (MFW) System 1 4 A1.03 - Power level restrictions for operation o 2.7* 1 MFW pumps and valves 061 Auxiliary / Emergency Feedwater K3.01- RCS 4.4 1 (AFW) System 1 4 061 Auxiliary I Emergency Feedwater 2.4.3 1 - Knowledge of annunciators alarms and 3.3 1 (AFW) System 1 4 indications, and use of the response instructions.

068 Liquid Radwaste System (LRS) 19 A3.02 - Automatic isolation 3.6 1 068 Liquid Radwaste System (LRS) / 9 X K1.07 - Sources of liquid wastes for LRS 2.7 1 071 Waste Gas Disposal System (WGDS) K4.04 - Isolation of waste gas release tanks 2.9 1 19 071 Waste Gas Disposal System (WGDS) K5.04 - Relationship of hydrogerdoxygen 2.5 1 19 concentrations to flammability 2

PWR RO Examination Outline Printed: 12/15/2003 DC Cook Nuclear Station sySmv# System / Evolution Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G KATopic Imp. Points 072 Area Radiation Monitoring (ARM) X A1.O1 - Radiation levels 3.4 1 System I 7 072 Area Radiation Monitoring (ARM) X K1.04 - Control room ventilation 3.3* 1 System 1 7 WACategory Totals: 2 2 2 2 2 2 3 2 2 2 2 Group Point Total: 23 3

PWR RO Examination Outline Printed: 1211512003 Facility: DC Cook Nuclear Station CS - 401 Plant ns - Tiel iysmv # System I Evolution Name K1 K2 K3 K4 K5 K6 A1 002 Reactor Coolant System (RCS) I 2 006 Emergency Core Cooling System X K3.03 - Containment 4.2 1 (ECCS) 12 I

006 Emergency Core Cooling System X K2.02 - Valve operators for accumulators 2.5* 1 (ECCS) I 2 010 Pressurizer Pressure Control System X K5 .O 1 - Determination of condition of fluid in 3.5 1 (PZR PCS) I 3 PZR, using steam tables 012 Reactor Protection System I 7 X A 1.O 1 - Trip setpoint adjustment 2.9* 1 014 Rod Position Indication System X K1.02 - NIS 3.0 1 (RPIS) / 1 026 Containment Spray System (CSS) I 5 A2.03 - Failure of ESF 4.1 1 026 Containment Spray System (CSS) / 5 X Al.05 - Chemical additive tank level and 3.1 1 concentration 029 Containment Purge System (CPS) / 8 X 2.1.23 - Ability to perform specific system and 3.9 1 integrated plant procedures during all modes of plant operation.

035 Steam Generator System (SIGS) I 4 A3.01 - S/G water level control 4.0 1 1

PWR RO Examination Outline Printed: 1211512003 Facility:

S - 401 Plant Systems - Tier 2 / Group 2 Form S-401-4 System / Evolution Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G KATopic Imp. Points lys/Ev I#

039 Main and Reheat Steam System X K5.05 - Bases for RCS cooldown limits 2.7 1 (MRSS) / 4 062 A.C. Electrical Distribution System 1 6 X K1.02 - EDIG 4.1 1 062 A.C. Electrical Distribution System 1 6 X K4.03 - Interlocks between automatic bus 2.8* 1 transfer and breakers 063 D.C. Electrical Distribution System / 6 X K1.03 - Battery charger and battery 2.9 1 064 Emergency Diesel Generator (EDIG) X A2.19 - Consequences of high VARS on EDIG 2.5 1 System 1 6 integrity 064 Emergency Diesel Generator (EDIG) X K6.07 - Air receivers 2.7 1 System I 6 073 Process Radiation Monitoring (PRM) X K1.O1 - Those systems served by PRMs 3.6 1 System I 7 073 Process Radiation Monitoring (PRM) X A4.02 - Radiation monitoring system control 3.7 1 System I 7 panel 079 Station Air System (SAS) I 8 X K4.01 - Cross-connect with IAS 2.9 1 086 Fire Protection System (FPS) I 8 X K3.01 - Shutdown capability with redundant 2.7

- 1 equipment 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 Facility: DC Cook Nuclear Station ES - 401 Form 5-40 1-4 I I System / Evolution Name <3 K4 K5 K6 $1 A2 A3 A4 G KATopic Imp. Points sys/Ev # I I I 1x1 I I 1 IA3.01 - Components which discharge to the

-2.7* 1 007 Pressurizer Relief Tank/Quench Tank System (PRTS) / 5 PRT I I 008 Component Cooling Water System I X A2.07 - Consequences of high or low CCW 2.5* 1 028 (CCWS) / 8 Hydrogen Recombiner and Purge X IIIII flow rate and temperature; the flow rate at which the CCW standby pump will start K2.01 - Hydrogen recombiners

-2 9 1 Control System (HRF'S) / 5 Fuel Handling Equipment System X A4.02 - Neutron levels

- 3.5 1 034 (FHES) 1 8 04 1 X K5.02 - Use of steam tables for saturation 2.5 1 Turbine Bypass Control / 4 temperature and pressure 078 Instrument Air System (IAS) / 8 X K3.01 - Containment air system 3.1* 1 078 Instrument Air System (IAS) / 8 X K4.02 - Cross-over to other air systems 3.2 1 103 Containment System 1 5 X Al.01 - Containment pressure, temperature, an 3.7 1 humidity WA Category Totals: 0 1 1 1 1 1 0 Group Point Total: 8 1

Generic Knowledge and Abilities Outline (Tier 3)

Printed: 1211512003 PWR RO Examination Outline Form ES-401-5 Facility: DC Cook Nuclear Station Generic Category KA KATopic Imp. Points Conduct of Operations 2.1.8 Ability to coordinate personnel activities outside the control room. 3.8 1 Knowledge of less than one hour technical specification action statements for systems. 3.0 1 Ability to apply technical specifications for a system. 2.9 1 Ability to make accurate, clear and concise verbal reports. 3.5 1 Equipment Control 2.2.12 Knowledge of surveillance procedures. 3.0 1 2.2.24 Ability to analyze the affect of maintenance activities on LCO status. I 12.6 1 2.2.27 Knowledge of the refueling process. 2.6 1 Knowledge of 10 CFR 20 and related facility radiation control requirements. 2.6 1 2.3.9 Knowledgeof the process for performing a containment purge. 2.5 I 2.3.10 ~ Ability to perform procedures to reduce excessive levels of radiation and guard against 2.9 1 personnel exposure.

I Emergency ProceduredPlan Ability to identify post-accident instrumentation.

Knowledge of the bases for prioritizing emergency procedure implementation during 3.5 2.8 1

1 emergency operations.

Knowledge of loss of cooling water procedures. 3.3 1 Generic Total: 13 1

ES-401 PWR SRO Examination Outline Form ES-401-3 1 Facility: Cook Plant Unit 1 & Unit 2 I

Date of Exam: 03/15 --26/2004 Exam Level: SRO I

~~~~

Note: 1. 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).

2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +I from that specified in the table based on NRC revisions. The final exam must total 100 points.
3. Select topics from many systems; avoid selecting more than two or three WAtopics from a given system unless they relate to plant-specific priorities.
4. Systems/evolutions within each group are identified on the associated outline.
5. The shaded areas are not applicable to the categoryltier.

6.* 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.

7. 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.

1 I

NUREG-I 021, Revision 8 , Supplement 1

PWR SRO Examination Outline Printed: 1211512003 Facility: DC Cook Nuclear Station ES - 401 Emer encv and Abnormal Plant Evolutions - Tier 1 / G r o w 1 Form ES-401-3 I

SIAPE # EIAPE Name I Safety Function KA Topic Imp. 'oints 00 1 Continuous Rod Withdrawal I 1 AK3 .O 1 - Manually driving rods into position that 3.6 1 existed before start of casualty 003 Dropped Control Rod I 1 AA2.03 - Dropped rod, using in-corelex-core 3.8 1 instrumentation, in-core or loop temperature measurements 003 Dropped Control Rod I 1 2.2.25 - Knowledge of bases in technical specifications 3.7 1 for limiting conditions for operations and safety limits.

01 1 Large Break LOCA I 3 EA2.08 - Conditions necessary for recovery when 3.9* 1 accident reaches stable phase 01 1 Large Break LOCA 1 3 EA 1.05 - Manual and/or automatic transfer of suction 3.9 1 015 Reactor Coolant Pump (RCP) Malfunctions 1 4 AK1.02 - Consequences of an RCPS failure 4.1 024 Emergency Boration I 1 AK1.04 - Low temperature limits for boron 3.6 concentration I

029 Anticipated Transient Without Scram (ATWS) I 1 EKl.01 - Reactor nucleonics and thermo-hydraulics I 3.1 behavior 040 Steam Line Rupture 1 4 AA2.05 - When ESFAS systems may be secured 4.5 040 Steam Line Rupture 1 4 AK2.01 - Valves 2.5 1

PWR SRO Examination Outline Printed: 12/15/2003 Facility: DC Cook Nuclear Station ES - 401 Emergencv and Abnormal Plant Evolutions - Tier 1 / Grow 1

- I Form ES-401-3

(/APE # E/APE Name / Safety Function K1 K2 K3 A1 KA Topic Imp. Points 059 Accidental Liquid Radwaste Release / 9 X AK2.02 - Radioactive-gas monitors 2.7 1 059 Accidental Liquid Radwaste Release / 9 X AK3.04 - Actions contained in EOP for accidental 4.3 1 liquid radioactive-waste release 062 Loss of Nuclear Service Water / 4 AA2.05 - The normal values for SWS-header flow rate 2.5*

and the flow rates to the components cooled by the sws 062 Loss of Nuclear Service Water I 4 X AA1.02 - Loads on the SWS in the control room 3.3 I . ,

I 074 Inadequate Core Cooling 1 4 X EK2.05 - LPI pumps 4.1 E02 SI Termination I 3 X EK3.1 - Facility operating characteristics during 3.6 transient conditions, including coolant chemistry and the effects of temperature, pressure, and reactivity changes and operating limitations and reasons for these operating characteristics E04 LOCA Outside Containment / 3 2.1.25 - Ability to obtain and interpret station reference 3.1 materials such as graphs, monographs, and tables which contain performance data.

E06 Degraded Core Cooling 1 4 X EK1.1- Components, capacity, and function of 4.0 emergency systems E06 Degraded Core Cooling 1 4 2.4.20 - Knowledge of operational implications of EOP 4.0 warnings, cautions, and notes.

2

PWR SRO Examination Outline Printed: 1211512003 Facility: DC Cook Nuclear Station ES - 401 Emer - enc!

I I SIAPE ## EIAPE Name I Safety Function K1 K3 A1 KA Topic Imp. Points E07 Saturated Core Cooling 1 4 X EA1.2 - Operating behavior characteristics of the 3.7 1 E08 Pressurized Thermal Shock I 4 facility 2.4.21 - Knowledge of the parameters and logic used to I 1 4.3 1 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.

E09 Natural Circulation Operations 1 4 X EK3.4 - RO or SRO function within the control room 3.6 1 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 E09 Natural Circulation Operations 1 4 X EA1.3 - Desired operating results during abnormal and 3.8 1 emergency situations E12 Uncontrolled Depressurization of all Steam EK2.2 - Facility's heat removal systems, including 3.9 1 Generators 1 4 primary coolant, emergency coolant, the decay heat removal systems, and relanons between the proper WA Category Totals: 4 4 4 4 Group Point Total: 24 3

PWR SRO Examination Outline Printed: 1211512003 Facility: DC Cook Nuclear Station ES - 401 Emergencv and Abnormal Plant Evolutions - Tier 1 / Groua 2

- I Form ES-401-3 E/APE # E/APE Name / Safety Function K1 K2 K3 A1 A2 K A Topic Imp. Points 007 Reactor Trip / 1 X EA2.03 - Reactor trip breaker position 4.4 1 007 Reactor Trip / 1 X EK2.03 - Reactor trip status panel 3.6 1 I I 1 I 1 008 Pressurizer (PZR) Vapor Space Accident (Relief AA 1.O 1 - PZR spray block valve and PORV block 4.0 1 009 valve EK3.20 - Tech-Spec leakage limits l l4.3 1 009 2.1.20 - Ability to execute procedure steps. 4.2 1 022 AK1.04 - Reason for changing from manual to 3.0 1 automatic control of charging flow valve controller 037 AA2.02 - Agreemenvdisagreement among redundant 3.9 1 radiation monitors 037 AK1.02 - Leak rate vs. pressure drop 3.9 1 038 EK3.06 - Actions contained in EOP for RCS water inventory balance, SIG tube rupture, and plant shutdown procedures 038 EK1.03 - Natural circulation 1

PWR SRO Examination Outline Printed: 12/15/2003 Facility: DC Cook Nuclear Station ES - 401 Emerpencv-and -

Abnormal Plant Evolutions - Tier 1 / GrouD 2 Form ES-401-3 E/AF'E # E/APE Name / Safety Function K1 K2 K3 A1 A2 G KATopic Imp. Points 054 Loss of Main Feedwater (MFW) 1 4 X 2.4.46 - Ability to verify that the alarms are consistent 3.6 1 with the plant conditions.

054 Loss of Main Feedwater (MFW) 1 4 AA1.04 - HPI, under total feedwater loss conditions 4.5 1 065 Loss of Instrument Air I 8 AA2.01 - Cause and effect of low-pressure instnunent 3.2 air alarm El 1 El 1 1

Loss of Emergency Coolant Recirculation 1 4 Loss of Emergency Coolant Recirculation 14 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 4.3 4.0 E16 High Containment Radiation / 9 EK2.1 - Components, and functions of control and 3.3 1 safety systems, including instrumentation, signals, Interlocks, iailure modes, and automatic and manual WACategory Totals: 3 3 2 3 3 2 Group Point Total: 16 2

PWR SRO Examination Outline Printed: 1211512003 Facility: DC Cook Nuclear Station ES - 401 Emergenc: and Abnormal Plant volutions - Tier 1 / Group 3 Form ES-401-:

I I E/APE # KA Topic Imp. Points 036 AK1.02 - SDM 3.8 1 056 Loss of Offsite Power I 6 AA2.13 - Operational status of ventilation supply fans 2.6 1 for the service water building, control room and battery E15 Containment Flooding 1 5 room 2.4.4 - Ability to recognize abnormal indications for I 4.3 I 1 system operating parameters which are entry-level WA Category Totals: 1 1 Group Point Total: 3 1

PWR SRO Examination Outline Printed: 12/15/2003 Facility: DC Cook Nuclear Station ZS - 401 ;roup 1 Form S-401-:

System / Evolution Name K1 KA Topic [mp. Points

$ys/Ev #

003 Reactor Coolant Pump System A4.07 - RCP seal bypass

-2.6 1 (RCPS) 1 4 003 Reactor Coolant Pump System 2.2.17 - Knowledge of the process for 3.5 1 (RCPS) / 4 managing maintenance activities during power operations.

004 Chemical and Volume Control System K2.03 - Charging pumps 3.5 1 (CVCS) I 1 004 Chemical and Volume Control System K4.01 - Oxygen control in RCS 3.3 1 (CVCS) I 1 013 Engineered Safety Features Actuation K3.01 - Fuel 4.7 1 System (ESFAS) / 2 013 Engineered Safety Features Actuation K6.01 - Sensors and detectors 3.1* 1 System (ESFAS) / 2 014 Rod Position Indication System X K1.02 - NIS 3.3 (RPIS) / 1 I

015 Nuclear Instrumentation System / 7 K6.02 - Discriminatorlcompensationcircuits 2.9 017 In-Core Temperature Monitor (ITM) K5.02 - Saturation and subcooling of water 4.0 System / 7 017 In-Core Temperature Monitor (ITM) A2.01 - Thermocouple open and short circuits 3.5 System I 7 025 Ice Condenser System / 5 A3.01 - Refrigerant system 3.0*

1

PWR SRO Examination Outline Printed: 1211512003 Facility: DC Cook Nuclear Station ES - 401 Plan1 ystems S-401-.

sysmv I# i 2 A3 \4 G KATopic Imp. Points 026 X A2.03 - Failure of ESF 4.4 1 059 06 1 06 1 I Pits. -

K1.03 Battery charger and battery 063 068 07 1 072 X

I I A3.02 - Automatic isolation K4.04 - Isolation of waste gas release tanks A 1.O 1 - Radiation levels I

3.6 WACategory Totals: 2 1 2 2 1 2 2 1 2 Group Point Total: 19 2

PWR SRO Examination Outline Printed: 12/15/2003 Facility: DC Cook Nuclear Station ES - 401 Plant S-401-:

sySrnv # System / Evolution Name K1 K2 K3 K4 K5 K6 A1 Points 002 Reactor Coolant System (RCS) I 2 1 006 Emergency Core Cooling System X K3.03 - Containment 4.4 1 (ECCS) I 2 010 Pressurizer Pressure Control System X 2.2.18 - Knowledge of the process for 3.6 1 (PZR PCS) / 3 managing maintenance activities during shutdown operations.

0 10 Pressurizer Pressure Control System X K5.01 - Determination of condition of fluid in 4.0 (PZR PCS) 1 3 PZR, using steam tables 012 Reactor Protection System I 7 X A 1.O 1 - Trip setpoint adjustment 3.4*

I 016 I Non-Nuclear Instrumentation System X 2.4.45 - Ability to prioritize and interpret the 3.6

("IS) /7 significance of each annunciator or alarm.

028 Hydrogen Recombiner and Purge X K2.01 - Hydrogen recombiners 2.8*

Control System (HRPS) / 5 034 Fuel Handling Equipment System A4.02 - Neutron levels 3.9 (FHES) / 8

/

Steam Generator System (SIGS) / 4 A3.01 - S/G water level control 039 Main and Reheat Steam System X K5.05 - Bases for RCS cooldown limits (MRSS) I 4 1

PWR SRO Examination Outline Printed: 12/15/2003 Facility: DC Cook Nuclear Station ES - 401 062 A.C. Electrical Distribution System / 6 I I 064 I Emergency Diesel Generator (ED/G) I I1 073 I System/6 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 Facility: DC Cook Nuclear Station ES - 401 rier 2 / Group 3 I I SysrEv # K2 K3 K4 K5 K6 i l A2 KA Topic Imp. Points 008 2.1.33 - Ability to recognize indications for 4.0 1 (CCWS) 1 8 system operating parameters which are entry-level conditions for technical specifications.

008 X A2.07 - Consequences of high or low CCW 2.8* 1 flow rate and temperature; the flow rate at which the CCW standby pump will start 041 X K5.02 - Use of steam tables for saturation 2.8 1 Turbine Bypass Control 1 4 temperature and pressure I I 078 Instrument Air System (US) I 8 X K3.01 - Containment air system K/A Category Totals: 0 0 1 Group Point Total: 4 1

Generic Knowledge and Abilities Outline (Tier 3)

Printed: 12/15/2003 PWR SRO Examination Outline Form ES-401-5 Facility: DC Cook Nuclear Station Generic Category KA KATopic Imp. Points Conduct of Operations 2.1.4 Knowledge of shift staffing requirements. 3.4 1 2.1.10 Knowledge of conditions and limitations in the facility license. 3.9 1 2.1.13 Knowledge of facility requirements for controlling vital / controlled access. 2.9 1 2.1.17 Ability to make accurate, clear and concise verbal reports. 3.6 1 Category Total: 4 Equipment Control 2.2.22 Knowledge of limiting conditions for operations and safety limits. 4.1

- 1 2.2.24 Ability to analyze the affect of maintenance activities on LCO status. 3.8 1 2.2.25 Knowledge of bases in technical specifications for limiting conditions for operations and

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3.7 1 safety limits.

2.2.27 Knowledge of the refueling process. 3.5 1 I

Category Total: 4 Radiation Control 2.3.2 Knowledge of facility ALARA program. 2.9 1 2.3.3 Knowledge of SRO responsibilities for auxiliary systems that are outside the control 2.9 1 room (e.g., waste disposal and handling systems).

2.3.4 Knowledge of radiation exposure limits and contamination control, including permissible 3.1 1 levels in excess of those authorized.

2.3.10 Ability to perform procedures to reduce excessive levels of radiation and guard against 3.3 1 personnel exposure.

Category Total: 4 1

Generic Knowledge and Abilities Outline (Tier 3)

Printed: 1211512003 PWR SRO Examination Outline Form ES-401-5 Facility: DC Cook Nuclear Station Generic Category :mp. Points Emergency ProceduredPlan 2.4.3 Ability to identify post-accident instrumentation. 3.8 1 2.4.8 Knowledge of how the event-based emergencylabnomal operating procedures are used 3.7 1 in conjunction with the symptom-based EOPs.

2.4.18 Knowledge of the specific bases for EOPs. 3.6 1 2.4.23 Knowledge of the bases for prioritizing emergency procedure implementation during 3.8 1 emergency operations.

2.4.24 Knowledge of loss of cooling water procedures. 3.7 1 I -

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 Describe method of evaluation:

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

Verification KIA SYS 001 A4.11 (3.5) Classroom / Simulator NRC2004-A1b Perform Unit 1 LTOP Verification per Parameter Verification 01-0HP-4030-114-030.

KIA 2.1.33 (3.4)

Classroom / Simulator A.2 NRC2004-A2 Determine Isolation Boundaries for the Refueling Use of P&IDs Water Purification Pump.

KA 2.1.24 (2.8)

Classroom I Simulator A. 3 NRC2004-A3 Perform Containment Pressure Relief per Control of Radiation 02-OHP-4021-028-004 Releases KA 2.3.1 1 (2.7) Simulator A.4 NRC2004-A4 Complete EMD-32A Nuclear Plant Event Notification Emergency Communication Form per PMP-2080-EPP-107.

KA 2.4.39 (3.3)

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 TY Pe Safety Code* Function

a. [NRC2004-SIMOI] Perform Hot Shutdown Panel N,S 3 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 D,S,L 7 Determination per 02-OHP-4023-SUP-OOI . WA EPE 01 1 EA1.14 3.9/4.1
c. [NRC2004-SIM03] Isolate Ruptured Steam Generator 4P M,A,S,L per 02-OHP-4023-E-3. WA EPE 038 EA1.32 4.6/4.7
d. [NRC2004-SIM04] Place Excess Letdown in Service 2 DS per02-OHP-4021-003-001. WA SYS 004 A4.06 3.6/3.1
e. [NRC2004-SIM05] Perform Hydrogen Recombiner 5 N,S 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 6 D,A,S per 01-OHP-4030-STP-O27AB, WA SYS 064 A4.06 3.913.9 (From Cook 2002 NRC.exam)
g. [NRC2004-SIM09] Respond to Traveling Screen DP 8 N,A,S,L Alarm per 12-OHP-4022-057-001. WA SYS 075 A2.01
a. [NRC2004-INP03] Shift TDAFP Suction to ESW per D 4s 02-4022-055-003. WA APE 054 AA1.01 4.5/4.4
b. [NRC2004-INP05] RCP Seal Injection via CVCS 2 D,R Cross-Tie to Maintain Pressurizer Level per 01-0HP-4025-LS-6. APE 022 AAI .01 3.4/3.3
c. [NRC2004-INP06] Perform Local Diesel Generator 6 D,A 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)
  • 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

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

~

Administrative Describe method of evaluation:

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

KIA 2.1.25 (3.1)

EPE 009 EA2.38 Classroom / Simulator 14.31 NRC2004-A5b Verify Appropriate LCO Action for Inoperable TS Action Radiation Monitors.

KA 2.1 .I2 (4.0)

Simulator A.2 NRC2004-A6 Perform Screen House Vulnerability Determination Maintenance per 12-OHP-5030-057-001.

KA 2.2.17 (3.5)

Classroom / Simulator A.3 NRC2004-A7 Review Waste Gas Decay Tank Release for SRO Control o f Radiation Authorization per 12-OHP-4021-023-002.

Releases KA 2.3.6 (3.1) Classroom / Simulator A.4 NRC2004-A8 Perform an Emergency Plan Classification with PAR EALs &

Classifications per PMP-2080-EPP-101.

KA 2.4.44 (4.0)

Classroom / Simulator NUREG-102 1 Revision 8

ES-301 Control Room Systems and Facility Walk-Through Test Outline Form ES-301-2 Facility: DC COOK Date of Examination: 03/15/2004 Exam Level (circle one): RO / Operating Test No.:

B.1 Control Room Systems System / JPM Title 1 Type Code*

1 Safety Function

a. [NRC2004-SIM02] Perform Subcooling Margin D,S,L 7 Determination per 02-OHP-4023-SUP-OOI . KIA EPE 01 1 EA1.14 3.9/4.1
b. [NRC2004-SIM03] Isolate Ruptured Steam Generator 4P M,A,S,L per 02-OHP-4023-E-3. WA EPE 038 EA1.32 4.614.7
c. [NRC2004-SIM05] Perform Hydrogen Recombiner NS 5 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 D,A, S 6 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- DS,L I 005-001. WA SYS 004 A4.12 3.3/3.8
f. [NRC2004-SIM08] Adjust Pressure in an S I DS,L 3 Accumulator per 02-OHP-4021-008-006. WA SYS 006 A I .I 3 3.513.7
g. [NRC2004-SIM09] Respond to Traveling Screen DP N,A,S,L 8 Alarm per 12-OHP-4022-057-001. WA SYS 075 A2.01 3.013.2
a. [NRC2004-INPOI] Locally Restore CR Ventilation per 7 N,A 01-OHP-4025-R-14. APE 068 AA1.24 3.0/3.6
b. [NRC2004-INP02] Locally Restore CRlD Inverters per 6 D,A 01-OHP-4021-082-008. APE 057 AA1.01 3.7/3.7
c. [NRC2004-INP04] Local Operation of U - I SG PORVs 4s D,R (MRV-213 and 243) per 01-OHP-4025-LS-4. WA APE 068 AA1.01 4.3/4.5

/

  • 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 - -

Facility: DC Cook 03/15/2004 Examination Level (circle one): RO Operating Test Number Administrative Describe method of evaluation:

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

WA 2.1.25 (3.1)

EPE 009 EA2.38 Classroom / Simulator (4.3)

NRC2004-A5b Verify Appropriate LCO Action for Inoperable TS Action Radiation Monitors-KA 2.1 .I2 (4.0)

Simulator A.2 NRC2004-A6 Perform Screen House Vulnerability Determination Maintenance per 12-OHP-5030-057-001.

KA 2.2.1 7 (3.5)

Classroom / Simulator A.3 NRC2004-A7 Review Waste Gas Decay Tank Release for SRO Control of Radiation Authorization per 12-OHP-4021-023-002.

Releases KA 2.3.6 (3.1) Classroom / Simulator A.4 NRC2004-A8 Perform an Emergency Plan Classification with PAR EALs &

Classifications per PMP-2080-EPP-101.

KA 2.4.44 (4.0)

Classroom / Simulator NUREG- 1021 Revision 8

~ ~ _ _ _ _ _ _ ~ _ _ _ _

ES-303 Control Room Systems and Facility Walk-Through Test Outline Form ES-301-2 Facility: DC COOK Date of Examination: 03/15/2004 Exam Level (circle one): RO I SRO(I) / @ Operating Test No.:

I/ B.l Control Room Systems System I JPM Title Type Safety Code* Function

a. [NRC2004-SIM07] Dilute the RCS per 12-OHP-4021- 1 005-001. WA SYS 004 A4.12 3.3/3.8
b. [NRC2004-SIM08] Adjust Pressure in an SI 3 Accumulator per 02-OHP-4021-008-006. WA SYS 006 A I .I3 3.513.7
c. [NRC2004-SIM09] Respond to Traveling Screen DP 8 Alarm per 12-OHP-4022-057-001. WA SYS 075 A2.01 3.013.2 1 B.2 Facility Walk-Through
a. [NRC2004-1NP01] Locally Restore CR Ventilation per N,A 7 01-0HP-4025-R-14. APE 068 AA1.24 3.0/3.6
b. [NRC2004-1NP04] Local Operation of U-1 SG PORVs 4s D,R (MRV-213 and 243) per 01-OHP-4025-LS-4. WA APE 068 AAI .01 4.3/4.5 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 1 Examiners: Operators:

Event Malf Event Event No. No. Type* Description 1 N Start South Hotwell pump and Stop Middle Hotwell Pump 2 R Raise Reactor Power and Turbine Load 3 RXO5A@O I-RO Controlling Pressurizer Level Channel (NLP-151) fails LOW (I 4 I RX17F @ 1200 I I-BOP I #22 SG Pressure Instrument (MPP-222) fails HIGH 5 CCOIA C-RO East CCW pump trips on overcurrent CC02B West CCW pump fails to AUTO start 6 SW07A C-BOP Main Turbine Oil Cooler Controller (WRV-970) fails LOW 1 7 I RCOlB@50 I M I RCS Loop 2 Large Break LOCA I I RP07A RPO7B 1 C-BOP I Both trains of Main Steamline Isolation fails to AUTO actuate 9 RP16A C-RO Both trains of CTS/Phase B isolation fails to AUTO actuate RP16B 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 Facility: Cook Plant Unit I81Unit 2 Scenario No.: COOK04-02 Op-Test No.: Set 1 Examiners: Operators:

Initial Conditions: IC-37, MOL; 99% power, 967 ppm Boron, 8 GWD, Equilibrium Xenon Event Malf. No. Event Event No. Type* Description 1 N Start Middle CB pump and place North CB pump in standby 2 RX20G@5E6 I-BOP #24 SG Steam Flow Transmitter (MFC-140) fails HIGH 3a NIIOB@200 I-RO Power Range N142 fails HIGH 3b RD0445 C-RO CBD Rod D-4 partial insertion/mechanically bound RDPROD(45) 195 4 R Lower reactor power and turbine load 5 MS16A@75 C-BOP Main Turbine Steam seal controller (SRV-22) fails in AUTO 30 sec ramp 6 RC02C@30 M RCS Loop 3 SBLOCA (300 gpm) 2 min ramp 7T RP03A RP03B ~

[C-RO ~ I AUTO/MANUAL Reactor trip actuation failure (ATWS)

RPI OA RPIOB C-RO AUTO Safety Injection actuation failure (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 l Facility: Cook Plant Unit 1 & Unit 2 Examiners:

Scenario No.: COOK04-03 Operators:

Op-Test No.: Set 2 Event Malf. No. Event Event No. Type* Description 1 N Perform ESW pump strainer functional test 1 2 I RX23B@O I I-BOP I #21 SG Level transmitter fails LOW I I RCR20 @ 3 101NMOI 53 I C-RO I PRZ PORV (NRV-153) Leak by (3 gpm)

NMO-153 fails OPEN I 4 I IR I Reduce reactor power and turbine load for CSD 5 CV16A I-RO VCT Level Transmitter (QLC-451) fails HIGH 6 FW05B C-BOP West Main Feed Pump Trip 1 7 I RC17C@95 lI M II PRZ PORV (NRV-153) fails OPEN 1 1 8 RP19C C-RO Train A K600 Relay failure: CCP to BIT Inlet valve (IMO-255) fails OPEN and 2CD EDG fails to start 9 RPI OB C-RO Train B Safety Injection signal failure to actuate in AUTO or RP11B Manual mode RP11D

  • (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 h c t i o n a l 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-451) 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 Malf. No. Event Event Type* Description

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1 R Raise reactor power and turbine load 2 RX18 @ 1200 I-BOP Main Steam Bypass Header Pressure transmitter (UPC-101) fails HIGH 3 FW34A C-BOP North CB pump trip; Middle CB pump fails to start in AUTO FW58B 4a RXOSA @ 10 I-RO PRZ Pressure Master Controller fails LOW 4b ED07B C-RO Loss of Bus 21PHC 5a RC23C @ 30 M SGTR on #23 SG (300 gpm) 2 min ramp 5b MS04D @ 20 M Steam Line Break outside Containment (#24 SG) - downstream of 2 min delay MSlV 7 RPO7A C-BOP Steam Line Isolation fails to AUTO actuate RP07B SI Train B - K609 Relay fails to actuate l 8 RP20B C-BOP

( I 1)nstrume , (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 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 pumps.

~ ~~ ~ ~~ ~ ~

Event Malf. No. Event Event No. Type* Description 1 N Start South NESW pump and Stop North NESW pump 2 RX19A @ 0 1 I-RO I Turbine Impulse Transmitter (MPC-253) fails LOW 3 cvo9 I C-RO I Letdown Temperature Controller fails LOW 4 RX25A @ 100 I I-BOP I #21 Main Feed Reg Valve (FRV-210) controller oscillation 5 FW12A @ 10 I C-BOP I East MFP lube oil leak 6 1 R 1 Perform Rapid Power Reduction 7a ED19 @ 25 M Electrical Grid Load Rejection 7b ED01 M Loss of All AC Power (2 min delay)

ED25 EGO1 8 EG08A 1 C-BOP I 2AB EDG Speed Governor failure 9 FW48C 1 I

C-BOP I I

TD AFW pump fails to AUTO start 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

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Cook Plant Scenario Outline Form ES-D-1 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 4a I ZD I10 ITA2 CLOSE l C-BOP l West MD AFW pump inadvertent start 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 RPOIA iI M lI C-RO Loss of Bus T21 D due to fault on Bus 2D 2D RCP Bus Fault (Loss of RCS Flow in Loop I ) - momentary Reactor Trip (RTA) Fails to AUTO actuate RPOI B Reactor Trip (RTB) Fails to AUTO actuate 8 1 TC03 I C-BOP I Turbine trip fails to AUTO actuate 9 FW51 C-BOP TDAFW pump T & W failure due to blown control power fuse 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 000001 AA2.01 ## Unable to write question of appropriate difficulty. Replace With WA 000001 AA2.03 1/ 2 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 000038 EKI .01
  • This WA is too close to 04 1000 K5 .O 1. Would have resulted in testing same knowledge. Replaced with WA 000038 AK1.03 2/ 2 01000 K5.02
  • 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.

2/ 2 055000 A3.03

  • DC Cook does NOT auto divert CARS Exhaust. Replaced with WA 035000 A3.01 2/2 027000 K2.01
  • Containment Iodine Removal System NOT Used at DC Cook (CTS Only). Replaced with WA 028000 K2.01 2/3 034000 K6.02
  • 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 1/ 1 00WE06 2.1.2 Unable to develop a question with plausible distracters of sufficient difficulty (>2). Replaced with WA OOWE06 2.1.7 1/ 1 000027 2.4.14 # WA addresses EOP flowchart but event is AOP. Replaced with WA 000027 2.4.11 2/ 1 022000 A2.06 DC Cook Containment Cooling Systems does NOT have a pump. Replaced with KIA 022000 A2.04 2/ I 025000 2.1.18 # Ice Condenser And Logs. Unable to develop a question with plausible distracters of sufficient difficulty (>2).

Replaced with WA 025000 2.1.33 2/ 1 061000 2.4.17 # AFW and EOP terms. Unable to develop a question with plausible distracters of sufficient difficulty (>2).

Redaced with WA 061000 2.4.3I 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 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.

3 2.2.12 ## 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 # = Justification not required per NRC website WA Guidance.
    1. = Changed based on written Examination Review.

NLTREG - 1021 Revision 8, Supplement 1

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

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

1/ 2 000038 EK1.O1

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

2/2 01000 K5.02

  • 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.

2/2 034000 K6.02

  • 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 2/2 055000 A3.03
  • DC Cook does NOT auto divert CARS Exhaust. Replaced with WA 035000 A3.01 2/3 027000 K2.01
  • Containment Iodine Removal System NOT Used at DC Cook (CTS Only). Replaced with WA 028000 K2.01.

1/ 1 00WE07 2.2.17 # Saturated Core Cooling with Managing Maintenance. Replaced with WA 00WE06 2.4.20. (Saturated Core Cooling procedure does NOT contain Notes or Cautions) 1/ 1 OOWEOS 2.2.14 # PTS with configuration changes. Replaced with WA OOWEO8 2.4.21.

1/ 1 00WE04 2.2.1 1 ## Large Break LOCA with Temporary Changes. Replaced with WA OOWE04 2.1.25 1/ 2 000009 2.2.11 # Small Break LOCA with Temporary Changes. Replaced with WA 000009 2.1.20 1/ 3 000056 A2.05 Unable to develou SRO level Question.Reulaced with WA 000056 A2.13 1 2/ 1 IO03000 2.1.13 # IRCPS and vital area access. Reulaced with WA 003000 2.2.17 2/ 1 061000 2.2.27 # AFW and Refueling process. Replaced with WA 061000 2.2.3 3 2.1.20 Unable to develop SRO level question. (Also 2.1.20 already used). Replaced with WA 2.1.13 3 2.4.34 Unable to develop SRO level question. (Also 2.4.34 appears on Audit Examination) Replaced with WA 2.4.18.

3 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 1 I I## = Changed based on written Examination Review.

NUREG - 1021 Revision 8, Supplement 1

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