ML20132F542

Forwards NRC Approved Operator Licensing Exam (Facility Outline & Initial Exam Submittal & as-given Operating Exam) for Tests Administered on 961007-11

ML20132F542
Person / Time
Site:	Columbia
Issue date:	12/19/1996
From:	Hurley L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20132F547	List: GO2-96-176, Informs That Pilot Initial License Exam Outline to Be Administered by NRC on 961007,completed ML20132F542
References
NUDOCS 9612260007
Download: ML20132F542 (1)

Text

1. ""%g UNITED STATES gy ,j\ NUCLEAR REGULATORY COMMISSION 4 :,

g R E GION IV

?* h [ 611 RYAN PLAZA DRIVE, SUITE 400 k9 , AR LINGTON, TEXAS 76011-8064 December 19, 1996 NOTE T0: NRC Document Control Desk Mail Stop 0-5-D-24 FROM: Laura Hurley, Licensing Assistant Operations Branch, Region IV

SUBJECT:

OPERATOR LICENSING EXAMINATIONS ASMINISTERED ON OCTOBER 7-11, 1996, AT WASHINGTON PUBLIC POWER SUPPLY SYSTEM DOCKET #50-397 On October 7-11, 1996, Operator Licensing Examinations were administered at the referenced facility. Attached you will find the following information for I processing through NUDOCS and distribution to the NRC staff, including the NRC PDR:

Item #1 - a) Facility submitted outline and intial exam submittal, designated for distribution under RIOS Code A070.

b) As given operating examination, designated for distribution under RIDS Code A070.

Item #2 - Examination Report with the as given written examination attached, designated for distribution under RIDS Code IE42.

If you have any questions, please contact Laura Hurley, Licensing Assistant, Operations Branch, Region IV at (817) 860-8253.

9612260007 961219 PDR V ADOCK 05000397 PDR e -

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T '

Cs lto. Box 968

• 3000 GeorSe Wsahington WWASlilNGTON PU July 26,1996 ay

• Richland. \nishiriston 993S2 0968 = (509) 3 72 5000 GO2-96-148 Docket No. 50-397 Mr. T.P. Gwynn, Director Division of Reactor Safety U.S. NRC, Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064

SUBJECT:

WNP-2 OPERATING LICENSE NPF-21 PROPOSED PILOT INITIAL INATION LICENSE OUTLINE EXAM The proposed pilot initial license examination Nuclear Regulatory Commission on October 7 199 outline fo outline is being mailed to Mr. Howard Bundy at R evaluation and approval. 6, has been completed. The examination egion IV on Friday, July 26,1996, for Per the requirement in Examiner Standards 201 to Mr. Bundy in a double envelope re ma k d OPENED BY ADDRESSEE ONLY" "FOR OFFICIAL USE ONLY", the propo public disclosure until afternthe examination completed.

has bee. WNP-2 re and "TO BE Operations Training at (509) 377-8266If you have any com

, please contact W.D. Shaeffer, Superintendent Respectfully, ,

J.I. Albers

\ Nuclear Training Manager (MD 1027) i cc:

{ TO McKernon - NRC/RIV I

JL Pellet - NRC/RIV Document Control Desk - NRC i TG Colburn - NRR ,

NS Reynolds - Winston and Strawn NRC Sr. Resident Inspector - 927N DL Williams - BPA/399 (1P/

j U/

s y  %

Cs WASHINGTON PUBLIC POWER SUPPLY SYSTEM P.o. Box 968

• 3000 Georse 1Vashington \Vay

• Richland 1Vashington 993S2-0968 * (509) 372 5000 July 26,1996  ;

GO2-96-148 Docket No. 50-397 i

Mr. T.P. Gwynn, Director Division of Reactor Safety U.S. NRC, Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064

SUBJECT:

WNP-2 OPERATING LICENSE NPF-21 PROPOSED PILOT INITIAL LICENSE EXAMINATION OUTLINE The proposed pilot initial license examination outline for the exam to be administered by the Nuclear Regulatory Commission on October 7,1996, has been completed. The examination outline is being mailed to Mr. Howard Bundy at Region IV on Friday, July 26,1996, for evaluation and approval.

Per the requirement in Examiner Standards-201, the proposed examination outline is being sent to Mr. Bundy in a double envelope marked "FOR OFFICIAL USE ONLY" and "TO BE OPENED BY ADDRESSEE ONLY", WNP-2 requests that these materials be withheld from public disclosure until after the examination has been completed.

If you have any comments or concerns, please contact W.D. Shaeffer, Superintendent, Operations Training at (509) 377-8266. I Respectfully,

]

J.P. Albers Nuclear Training Manager (MD 1027) cc: TO McKernon - NRC/RIV }

JL Pellet - NRC/RIV Document Control Desk - NRC 0 )

TG Colburn - NRR NS Reynolds - Winston and Strawn NRC Sr. Resident Inspector - 927N ,

DL Williams - BPA/399

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e Facility: WNP2 Knowledge and Ability Record Fonn Exam date: October 7,1996 ref: NUREG/BR-0122, Rev. 5 COUNT MATRIX l Summarizing Counts by K/A Group for BWR - Senior Reactor Operator Total Plant Wide Generics 14 KI K2 K3 K4 K5 K6 Al A2 A3 A4 SG Plant Systems I 2 1 2 4 3 2 2 2 2 2 1 23 Plant Systems II 1 1 4 4 0 0 0 2 0 0 1 13 Plant Systems 111 0 0 0 1 0 1 0 1 0 0 1 4 Emergency /Abn I 3 6 4 - --- - 4 4 - --

5 26 Emergency /Abn 11 2 6 2 -- - -

3 1 - ---

3 17 l

Totals 8 14 12 9 3 3 9 10 2 2 11 -

Model Total 97 1

I Phility: WNP2 Kn:wledge and Ability Record Fcrm Exam date: October 7,1996 i ret: NUREG/BR-0122, Rev. 5 .)

I PLANT-WIDE GENERIC RESPONSIBILITIES l

BWR - Senior Reactor Operator Target: 1 's.

• Actual: 17 %

l 1

K/A Topic SRO Ratmg f

1. 29400lKl.01 Knowledge of how to conduct and verify valve lineups. 3.7 l l

2. 29400lKl.02 Knowledge of tagging and clearance procedures. 4.5 j

3. 29400lKl.03 Knowledge of 10 CFR 20 and related facility radiation control requirements. 3.8 j 4, l 29400lKl.04 Knowledge of Facility ALARA program. 3.6 l S. 29400lKl.05 Knowledge of Facility requirements for controlling access to vital / control areas. 3.7 i

6. 294001Kl.13 Knowledge of safety procedures related to oxygen-deficient environment. 3.6 j

7. 29400lKl.16 Knowledge of facility protection requirements including fire brigade and 3.8  ;

portable fire fighting equipment usage.

8. 294001 A1.01 Ability to obtain and verify control procedure copy. 3.4

9. 294001 Al.02 Ability to execute procedural steps. 4.2 10 294001 A1.03 Ability to locate and use procedures and directives related to shift staffmg and 3.7  ;

activities.

1I 294001 AI.05 Ability to make accurate, clear, and concise verbal reports. 3.8 j 12 294001 Al.09 Ability to coordinate personnel activities inside the Control Room. 4.2 f

13 294001 Al.14 Ability to maintain primary and secondary Plant chemistry within allowable 3.4  ;

limits.  !

14 294001 A1.16 Ability to take actions called for in the Facility Emergency Plan, including (if 4.7  ;

required) supporting or acting as the Emergency Coordinator.

' NOTE: It is intended that 17 of 100 questions be written utilizing the above 14 selected KAs.

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Facility: WNP2 Knzyl:dge cnd Ability Reccrd Form Exam date: October 7,1996 l ref: NUREG/BR-0122 Rev. 5 E '

PLANT SYSTEMS - GROUP II 4

BWR - Reactor Operator Target: 19 % Actual: 19 %

4 K/A Topic (Systems - Group 11) RO Rating  !

1. 201003A2.01 Ability to predict the impact of the following on the Control Rod and Drim 3.4 Mechanism: Stuck Rod.

2. 201003K4.04 Control Rod and Driw Mechanism design feature (s) and/or interlocks which 3.6 .

provide for the use of either accumulator or Reactor water to Scram the Control Rod, j 3. 20200lGK.06 Knowledge of bases in Technical Specifications for limiting conditions for 3.8 operations and safety limits.

Ability to predict the impact of the following on the Recirculation System: High 3.9

) 4. 202001 A2.14 i Reactor Pressure (A1WS) Initiation.

S. 204000K4.04 Reactor Water Cleanup System design feature (s) and/or interlocks which provide 3.6 l

for System isolation signals.

4 2

6. 214000K3.03 Knowledge of the effect that a loss or malfunction of Rod Position Infonnation 3.2 System will have on RMCS.

7. 219000Kl.06 Physical connections and/or cause-effect relationships between RHR/LPCit 3.2  :

i Torus / Suppression Pool Cooling Mode and the Keep Fill System.

> 8. 219000K4.09 RHR/LPCit Torus / Suppression Pool Cooling Mode design feature (s) and/or 3.3

interlocks which provide for # cat Erchanger cooling.

9. 22600lGK.07 Purpose and function of major components and controls in the RHR/LPCit 3.5 Containment Spray System Mode,

10. 230000A4.02 Manually operate and/or monitor RRR/LPCit Torus / Suppression Pool Spray 3.8 i'

Valves.

I1. 245000K4.09 Main Turbine Generator and Auxiliary Systems design feature (s) and/or 3.2 interlocks which provide for Turbine Control.

Effect that a loss or malfunction of Circulating Water System will have on the 3.1

12. 256000K6.02 Renaer Condensate System.

13. 256000GK.07 Purpose and function of major components and controls in the Reactor 3.4 l

Condensate System.

262002K4.01 Uninterruptable Power Supply design feature (s) and/or interlocks which provide 3.4 l 14 i for transferfrom preferred power to alternate power supplies.

i Effect that a loss or malfunction of the D.C. Electrical Distribution System will 3.8 i 15. 263000K3.03 j have on systems with D.C. components.

3.5 4 16. 272000Kl.02 Physical connections and/or cause-effec _t relationships between Radiation Monitoring System and the Ofgas System.

17, Ability to predict the impact of the following on the fire Protection System: 3.1

) 286000A2.06 1 Low Fire Main Pressure.

Operational implications of Vacuum Brealer operation as it applies to 3.3

18. 29000lK5.01 Secondary Containment.

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Facility: WNP2 Kumledge end Ability Rec rd Form Exam date: October 7,1996 ref: NUREG/BR4122, Rev. 5 K/A Topic (Systems - Group II) RO Rating

19. 290003A3.01 Monitor automatic operations of the Control Room #VAC including: 3.3 inislation' Reconfiguration.

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! Facility: WNP2 Knowledge end Ability Rec::rd Form Exam date: October 7,1996  :

ref: NUREG/BR-0122, Rev. 5 PLANT SYSTEMS - GROUP Ill i BWR - Reactor Operator Target: 4 % Actual: 4%

l K/A Topic (Systems - Group !!!) RO Rating l 1. 21500lK4.01 Traversing In-Core Probe design feature (s) and/or interlocks which provide for 3.4 1 Primary Containment isolation.

2. 233000A2.07 Ability to predict the impact of the following on the fuel Pool Cooling and 3.0

Cleanup System

High Fuel Pool Temperature.

f 3. 288000A2.03 Ability to predict the impact of the following on the Plant Ventilation System: 3.5 N

Loss of Coolant Accident (s).

4. 290002K5.01 Operational implications of 7hennal Limits as it applies to Reactor vessel 3.5

internals.

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. Facility: WNP2 Knowledge end Ability Reccrd Form Exam date: October 7,1996 a

ref: NUREG/BR-0122 Rev. 5 4 EMERGENCY & ABNORMAL PLANT EVOLUTIONS - GROUP I f BWR - Reactor Operator Target: 13 % Actual: 13 %

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, K/A Topic (Emerg. & Abn. - Group I) RO Rating

1. 295006G.005 Knowledge of the annunciator alarms and indications, and use of the response 4.0 instructions.

l 2. 295006A1.06 Operate and/or monitor CRD Hydraulic System as applied to SCRAM. 3.6 1

l 3. 295009K2.03 Interrelations between Low Reaaor Water Lewl and the following: Reactor 3.2 Recirculation System.

4

4. 295009A1.03 Operate and/or monitor let Pump Net Positiw Suaion Head as applied to Low 3.1 1 Reactor Water Lewl.

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j 5. 295014A2.01 Determine and/or interpret the following as they apply to inadertent Reactivity 4.2 j Addition: Reactor Power.

6. 295015A1.02 Operate and/or monitor Cooldown efects on Reactor Power as applied to 4.2 l

l Incomplete SCRAM.

7. 295015K2.08 Interrelations between /ncomplete Scram and the following: Neutron monitoring 3.7 system.

8. 295024K3.07 Reason (s) for Drywell Venting as applied to High Dowell Pressure. 4.0 j 9. 295024G.011 Ability to recognize abnormal indications for system operating parameters which 4.5 l are entry-level conditions for emergency and abnormal operating procedures.

i'

10. 295025A2.03 Determine and/or interpret the following as they apply to High Reactor 4.1 i Pressure: Suppression Pool Temperature.

4 i1. 295031Kl.01 Operational iraplications of adequate core cooling as applied to Reactor Low 4.7 l Water Lewl.

t2. 295037K2.13 Interrelations between Scram comfitions present and Reactor Power abow 4.1 l

APRM downscale or unknown and alternate Boron injection methods.

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13. 295037A2.01 Determine and/or interpret the following as they apply to Scram conditions 4.3 l

f present and Reactor Power abow APRM downscale or unknown: Reactor i Power.

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i Facility: WNP2 Knowledge end Ability Rec:rd Form Exam date: October 7,1996 ref: NUREG/BR-0122, Rev. 5 EMERGENCY & ABNORMAL PLANT EVOLUTIONS - GROUP II BWR - Reactor Operator Target: 19 % Actual: 19 %

K/A Topic (Emerg. & Abn. - Group 11) RO Rating

1. 295001G.010 Ability to perform without reference to procedures those actions that require 3.7 immediate operation of system components or controls.

2. 295002K3.01 Reasons for the following responses as they apply to Loss of Main Condenser 3.8 Vacuum: Reactor Scram.

3. 295003Al.03 Operate and/or monitor Systems necessary to assure safe Plant Shutdown as 4.4 applied to Partial or Complete Loss of A.C. Power.

4. 295004K2.03 Interrelations between Partial or complete loss ofD.C. Power and D.C. Bus 3.3 Loads.

5, 295008A1.08 Operate and/or monitor feedwater System as applied to High Reactor Water 3.5 Lewl.

6. 295013K2.01 Interrelations between High Suppression Pool Temperature and the following: 3.6 Suppression Pool Cooling.

7, 295016K3.01 Reason (s) for disabling Control Room controls as applied to Control Room 4.1 Abandonment.

8. 295017K3.01 Reason (s) for System Isolations as applied to High Of-Site Release Rate. 3.6

9. 295018K2.02 Interrelations between Partial or complete Loss of Component Cooling Water 3.6 and Plant operations.

10. 295019A2.01 Determine and/or interpret the following as they apply to Partial or complete 3.6 Loss ofInstrument Air: Instrument Air system pressure.

I1. 295019G.005 Knowledge of the annunciator alarms and indications, and use of the response 3.3 instructions.

12, 295020K2.04 Interrelations between inadwnent Containment Isolation and the following: 3.1 RWCU system.

13. 295022Al.01 Operate and/or monitor Reactor Pressure vs. Rod Insertion as applied to Loss of 3.4 CRD Pumps.

14. 295026K3.01 Reason (s) for Emergency /Nonnal depressurization as applied to Suppression 3.8 Pool High Water Temperature.

15. 295028Kl.01 Operational implications of Reactor Water Lewt measurement as applied to 3.7 High Drywell Temperature.

16. 295029K2.06 Interrelations between High Suppression Pool Water Lewl and the following: 3.5 SRV's and discharge piping.

17. 295030G.007 Ability to explain and apply all system limits and precautions. 3.6 18, 295033K2.03 Interrelations between High Secondary Containment Area Radiation Lewis and 3.9 the following: Secondary Containment Ventilation.

19. 295034K2.02 Interrelations between Secundary Containment Ventilation High Radiation and 3.9  !

the following: Area Radiation Monitoring System.

I8

Facility: WNP2 Knswledge and Ability Rec:rd Form Exam date: October 7,1996 ref: NUREG/BR-0122 Rev. 5 -

EMERGENCY & ABNORMAL PLANT EVOLUTIONS - GROUP III BWR - Reactor Operator Target: 4 % Actual: 4% ,

i K/A Topic (Emerg. & Abn. - Group III) RO Rating i

1. 295021G.008 Ability to recognize indications for system operating parameters which are 3.2 entry-level conditions for Technical Specifications. ,

2. 295023Kl.01 Operational implications of Radiation caposure hazards as applied to Refueling 3.6 Accidents.

i

3. 295032K3.03 Reason (s) for Isolating gected systems as applied to High Secondary 3.8 i Containment Area Temperature.

4. 295035A1.02 Operate and/or monitor Secondary Containment Ventilation System as applied to 3.6 1 7

Secondary Containment high diferentialpressure.

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Facility: WNP2 Kn wledge end Ability Record Form Exam date: October 7,1996 ref: NUREG/BR-0122, Rev. 5 PLANT SYSTEMS - GROUP I BWR - Senior Reactor Operator Target: 23 % Actual: 23 %

K/A Topic (Systems - Group I) SRO Rating

1. 202002A2.07 Ability to predict the impact of the following on the Recirculation Flow Control 3.3 System: Loss ofA.C. power.

2. 202002Kl.02 Physical connections and/or cause-effect relationships between Recirc. Flow 4.2 Control System and Reactor Power.

3. 203000Kl.13 Physical connections and/or cause-effect relationships between RiiR/LPCI: Inj. 4.0 mode and Drywell Pressure.

4. 209002A3.01 Monitor automatic operations of the liigh Pressure Core Spray System (HPCS) 3.3 including: Valve Operation.

3. 209002K4.02 liigh Pressure Core Spray System (liPCS) design feature (s) and/or interlocks 3.5 which provide for preventing overfilling the Reactor Vessel.

6. 211000K6.03 Effect that a loss or malfunction of A.C. Power will have on the Standby Liquid 3.3 Control System.

7. 212000Al.11 Predict and/or monitor changes in parameters associated with operating the 3.3 Reactor Protection System status lights and alanns.

8. 212000K2.01 Electric Power supplies to the Reactor Protection System. 3.3

9. 212000K3.06 Knowledge of the effect that a loss or malfunction of RPS will have on the 4.1 scram air header solenoid operated valves.

10. 215004A4.01 Manually operate and/or monitor SRM count rate and period. 3.8

11. 216000K5.07 Operational implications of elevated Containment temperature efects on lewt 3.8 indication as it applies to NBI.

12, 217000A1.03 Predict and/or monitor changes in Reactor Water Level parameters associated 4.0 with ope.nting RCIC.

13, 218000K5.01 Operational implications of ADS Logic Operation as it applies to the Automatic 3.8 Depressurization System.

14. 223001 A4.13 Manually operate and/or monitor Hydrogen Recombiners. 3.4

15. 22300lK4.06 Primary Containment System and Auxiliaries design feature (s) and/or interlocks 3.3 which provide for maintaining proper containment / secondary containment to Drywell diferentialpressure.

16. 22600lGK.07 Purpose and function of major components and controls in the RHR/LPCh 3.5 Containment Spray System Mode.

17. 239002K4.05 Relief / Safety Valves design feature (s) and/or interlocks which provide for SRV 3.7 operationfrom more than one location.

18. 241000K3.02 Knowledge of the effect that a loss or malfunction of Reactor / Turbine Pressure 4.3 Regulating System will have on Reactor Pressure.

3

i Facility: WNP2 Kn:wledge end Ability Record Form Exam date: October 7,1996 ref: NUREG/BR-0122, Rev. 5 K/A Topic (Systems - Group I) SRO Rating

19. 241000A2.03 Ability to predict the impact of the following on the Reactor / Turbine Pressure 4.2 Regulating System: Failed open/ closed bypass valve (s).

20. 261000K4.01 Standby Gas Treatment System design feature (s) and/or interlocks which 3.8 provide for automatic system initiation.

21. 264000K6.08 Effect that a loss or malfunction of A.C. Power will have on the Emergency 3.7 i Diesel Generators.

22. 264000A3.05 Monitor automatic operations of the Emergency Diesel Generators including: 3.5 Load shedding and sequencing.

23. 29000lK5.01 Operational implications of Vacuum Breater operation as it applies to 3.4 Secondary Containment, j

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Facility: WNP2 Kn::wledge end Ability Reccrd Form Exam date: October 7,1996 ref: NUREG/BR-0122, Rev. 5 PLANT SYSTEMS - GROUP 11 BWR - Senior Reactor Operator Target: 13 % Actual: 13 %

i i K/A Topic (Systems - Group II) SRO Rating i 1, 201001A2.04 Ability to predict the impact of the following on the CRD Hydraulic System: 3.9 Scram Conditions.

1

2. 20100lK2.05 Electric Power supplies to the Alternate Rod Insenion Valves. 4.5 l 3. 201002K4.02 Reanor Manual Control System design feature (s) and/or interlocks which 3.5 1 provide for Control Rod Blocks.

4. 201002K3.01 Knowledge of the effect that a loss or malfunction of Reactor Manual Control 3.4 4 System will have on the ability to move ControlRods.

, 5. 202.001 A2.14 Ability to predict the impact of the following on the Recirculation System: lligh 4.2 Reactor Pressure (AIWS) initiation.

l 6. 20200lGK.06 Knowledge of bases in Technical Specifications for limiting conditions for 4.1 operations and safety limits.

7. 204000K4.04 Reactor Water Cleanup System design feature (s) and/or interlocks which provide 3.6 2 for System Isolation signals.

8. 214000K3.03 Knowledge of the effect that a loss or malfunction of Rod Position Information 3.2 j System will have on RMCS. i 1

9. 245000K4.09 Main Turbine Generator and Auxiliary Systems design feature (s) and/or 3.2 interlocks which provide for Turbine Control.

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l 10. 25900lK3.01 Knowledge of the effect that a loss or malfunction of Reanor Feedwater System 3.9 will have on Reactor Water level, d

i j 11. 262002K4.01 Uninterruptable Power Supply design feature (s) and/or interlocks which provide 3.4 >

l for transferfrom preferred power to alternate power supplies.

12. 263000K3.03 Effect that a loss or mi. function of the D.C. Electrical Distribution System will 3.8 l have on syst:ms w!!h D.C. components, t

13. 272000Kl.02 Physical connections and/or cause-effect relationships between Radiation 3.5 i Monitoring System and the Ofgas System, s

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Facility: WNP2 Kn wledge end Ability Reccrd Form Exam date: October 7,1996 ref: NUREG/BR-0122, Rev. 5 PLANT SYSTEMS - GROUP III BWR - Senior Reactor Operator Target: 4 % Actual: 4%

K/A Topic (Systems - Group III) SRO Rating

1. 201003K4.04 Control Rod and Drim Mechanism design feature (s) and/or interlocks which 3.7 provide for the following: 1he use of either accumulator or Reactor Water to Scram the control rod.

2. 201003A2.01 Ability to predict the impact of the following on the Control Rod and Drive 3.6 Mechanism: Stuck Rod.

3. 256000K6.02 Effect that a loss or malfunction of Circulating Water System will have on the 3.1 Reactor Condensate System.

4. 256000GK.07 Purpose and function of major components and controls in the Reactor 3.4 ,

Condensate System. l l

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Facility

WNP2 - Knsuledge end Ability Rec:rd Form Exam date: October 7,1996

' ref: NUREG/BR-0122. Rev. 5 1 EMERGENCY & ABNORMAL PLANT EVOLUTIONS - GROUP I l i

BWR - Senior Reactor Operator Target: 26% Actual: 26 %  ;

I l K/A Topic (Emerg. & Abn. - Group I) SRO Rating

1. 295003K2.02 Interrelations between Panial or Complete Loss of A.C. Power and the 4.2 Emergency Generators.

2. 295003A1.03 Operate and/or monitor Systems neussary to assure safe Plant Shutdown as 4.4 l

applied to Partial or Complete Loss ofA.C. Power.

3. 295006G.005 Knowledge of the annunciator alarms and indications, and use of the response 4.0 instructions.

7

4. 295006A1.06 Operate and/or monitor CRD Hydraulic System as applied to SCRAM. 3.6 j

5. 295009K2.03 Interrelations between Low Reactor Water Level and the following: Reactor 3.2 Recirculation System.

Operate and/or monitor Jet Pump Net Positiw Suction Head ax applied to Low 3.1  ;

6. 295009A1.03 Reador Water Level.

7, Interrelations between High Suppression Pool Temperature and tise following: 3.7 295013K2.01 Suppression Pool Cooling.

8. 295014A2.01 Determine and/or interpret the following as they apply to Inadwrient Reactivity 4.2 Addition: Reactor Power.

9, 295015A1.02 Operate and/or monitor Cooldown egects on Reactor Power as applied to 4.2 l Incomplete SCRAM.

10, 295015K2.08 Interrelations between incomplete Scram and the following: Neutron monitoring 3.7 system.

Ability to locate and operate components, including local controls. 4.1 I1. 2950160.006 Reason (s) for disabling Control Room controls as applied to Control Room 4.2 i

12. 295016K3.01 Abandonment.

3.9 j

13. 295017K3.01 Reason (s) for System Isolations as applied to High O.f-Site Release Rate.

Operational implications of Radiation exposure hazards as applied to Refueling 4.1

14. 295023Kl.01 Accidents.

I Reason (s) for Drywell Venting as applied to High Drywell Pressure. 4.0

15. 295024K3.07 Ability to recognize abnormal indications for system operating parameters which 4.5

16. 295024G.011 are entry-level conditions for emergency and abnormal operating procedures.

Interrelations between High Reactor Pressure and the following: RPS. 4.1

17. 295025K2.01 Determine and/or interpret the following as they apply to High Reanor 4.1

18. 295025A2.03 Pressure: Suppression Pool Temperature.

Ability to utilize symptom based procedures. 4.5

19. 295026G.012 Reason (s) for Emergency /Nonnal depressurization as applied to Suppression 4.1

20. 295026K3.01 Pool High Water Temperature.

7

Facility: WNP2 Knswiedge end Ability Recsrd Form Exam date: October 7,1996 ref: NUREG/BR-0122, Rev. 5 j K/A Topic (Emerg. & Abn. - Group 1) SRO Rating i 1

21. 295030Kl.03 Operational implications of Heat Capacity as applied to Low Suppression Pool 4.1 Water Lewl.

22. 295030G.007 Ability to explain and apply all system limits and precautions. 3.9

23. 295031 A2.01 Determine and/or interpret the following as they apply to Reactor Low Water 4.6 l Level: Reaaor Water Lewl.

24. 295031Kl.01 Operational implications of adequate core cooling as applied to Reactor Low 4.7 Water Level.

25. 295037K2.13 ' Interrelations between Scram conditions present and Reactor Power above 4.1 APRM downscale or unknown and alternate Baron injection methods.

4

26. 295037A2.01 Determine and/or interpret the following as they apply to Scram conditions 4.3

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present and Reactor Power above APRM downscale or unknown: Reactor i Power.

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_ .- . . - , . -~ . .- - - . . _ - - - - . - . - - . - .. - - .

Facility: WNP2 Knowledge::nd Ability Rec:rd Forrn Exam date: October 7,1996 ref: NUREG/BR-0122, Rev. 5 i

EMERGENCY & ABNORMAL PLANT EVOLUTIONS - GROUP 11 BWR - Senior Reactor Operator Target: 17 % Actual: 17 %

K/A Topic (Emerg. & Abn. - Group II) SRO Rating

1. 29500lG.010 Ability to perform without reference to procedures those actions that require 3.7 immediate operation of system components or controls.

2. 295002K3.01 Reasons for the following responses as they apply to Loss of Main Condenter 3.8 Vacuum: Reactor Scram.

3. 295004K2.03 Interrelations between Partial or complete Loss ofD.C. Power and D.C. Bus 3.3 Loads.

4, 295005Kl.02 Operational implications of Core 7hermal Limit mnsiderations as applied to 3.6 Main Turbine Generator Trip.

5. 295008A1.08 Operate and/or monitor Peedwater System as applied to High Reanor Water 3.5 Lesel.

6. 29501BK2.02 Interrelations between Partial or complete Loss of Component Cooling Water 3.6 and Plant operations.

7. 295019A2.01 Determine and/or interpret the following as they apply to Partial or complete 3.6 Loss ofinstrument Air: Instrument Air system pressure.

8, 295019G 005 Knowledge of the annunciator alarms and indications, and use of the response 3.3 instmetions.

9. 295020K2.04 Interrelations between Inadiertent Containment isolation and the following: 3.1 RWCU system.

10. 295021G.008 Ability to recognize indications for system operating parameters which are 3.9 entry-level conditions for Technical Specifications.

I1. 295022A1.01 Operate and/or monitor Reactor Pressure vs. Rod Insertion as applied to Loss of 3.4 CRD Pumps.

12, 295028Kl.01 Operational implications of Reactor Water Level measurement as applied to 3.7 High Drywell Temperature.

13. 295029K2.06 Interrelations between High Suppression Pool Water Lesel and the following: 3.5 SRV's and discharge piping.

14. 295032K3.03 Reasons for the following responses as they apply to High Secondary 3.9 Containment Area Temperature: Isolating afected Systems.

15. 295033K2.03 Interrelations between High Semndary Containment Area Radiation Leirls and 3.9 the following: Secondary Containment Ventilation.

16. 295034K2.02 Interrelations between Secondary Containment Ventilation High Radiation and 3.9 the following: Area Radiation Monitoring System.

I

17. 295035Al.02 Operate and/or monitor SBUT as applied to Secondary Containment High 3.8 Diferential Pressure.

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Facility: WNP2 Kn:wledge end Ability Rec rd Form Exam date: October 7,1996 ref: NUREG/BR-0122, Rev. 5 COUNT MATRIX Summarizing Counts by KIA Group for BWR - Reactor Operator Total Plant Wide Generics 13 K1 K2 K3 K4 K5 K6 Al A2 A3 A4 SG Plant Systems ! 2 2 4 5 2 3 2 4 2 2 0 28 Plant Systems li 2 0 2 5 1 1 0 3 1 1 3 19 Plant Systems III 0 0 0 1 1 0 0 2 0 0 0 4 Emergency /Abn 1 1 3 1 - -- -- 3 3 - --

2 13 Emergency /Abn 11 1 7 4 --- - -

3 1 -- - 3 19 Emergency /Abn Ill 1 0 1 -- - - 1 0 -- -

1 4 Totals 7 12 12 11 4 4 9 13 3 3 9 -

Model Total 100 l

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Facility: WNP2 Kn::wledge end Ability Rec:rd Ferm Exam date: October 7,1996

< ref: NUREG/BR4122, Rev. 5 ,

I a PLANT-WIDE GENERIC RESPONSIBILITIES  !

1 BWR - Reactor Operator Target: 13 % Actual: 13 %

, K/A Topic RO Rating

1. 29400lKl.02 Knowledge of tagging and clearance procedures. 3.9

2. 29400lKl.03 Knowledge of 10CFR20 and related Facility radiation control requirements. 3.3

! 3. 29400lKl.04 Knowledge of Facility ALARA program. 3.3

5. 294001XI.05 Knowledge of Facility requirements for controlling access to vital / control areas. 3.2

6. 29400lKl.09 Knowledge of safety procedures related to high pressure. 3.4 7, 29400lKl.14 Knowledge of safety procedures related to confined spaces. 3.2 1

8. 29400lKl.16 Knowledge of facility protection requirements including fire brigade and 3.5 portable fire fighting equipment usage.

9. 294001 Al.02 Ability to execute procedural steps. 4.2 10 294001A1.05 Ability to make accurate, clear, and concise verbal reports. 3.4

]

294001 Al.09 Ability to coordinate personnel activities inside the Control Room. 3.3 l 11 12 294001 A1.13 Ability to locate Control Room switches, controls and indications, and to 4.5 I i detennine that they are correctly reflecting the desired Plant kneup.

i 13 294001 Al.15 Ability to use Plant Computer to obtain and evaluate parametric information on 3.2

. system and component status.

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Facility: WNP2 Kn:wledge end Ability Rectrd Form Exam date: October 7,1996 ref: NUREG/BR-0122, Rev. 5 l

i PLANT SYSTEMS - GROUP 1 l BWR - Reactor Operator Target: 28% Actual: 28 %  ;

I K/A Topic (Systems - Group I) RO Rating

1. 201001A2.04 Ability to predict the impact of the following on the CRD liydraulic System: 3.8 Scram Conditions.  ;

2. 201001K2.05 Electric Power supplies to the Alternate Rod Insenion Valms. 4.5 f

3. 201002K3.01 Knowledge of the effect that a loss or malfunction of Reactor Manual Control 3.4 ,

System will have on the ability to mow Control Rods. l 3.5

4. 201002K4.02 Reactor Manual Control System design feature (s) and/or interlocks which }

provide for Control Rod Blocks. l

5. 202002A2.07 Ability to predict the impact of the following on the Recirculation Flow Control 3.3 ]:

System: Loss of A.C. power.

I

6. 202002Kl.02 Physical connections and/or cause-effect relationships between Recire. Flow 4.2  :

Control System and Reactor Power. l t

7. 203000Kl.13 Physical connections and/or cause-effect relationships between RHR/LPCI: Inj. 3.9 [

I mode and Drywell Pressure.

8. 209002A3.01 Monitor automatic operations of the High Pressure Core Spray System (HPCS) 3.3 including: Valw Operation.

9. 209002K4.02 High Pressure Core Spray System (HPCS) design feature (s) and/or interlocks 3.4 i

which provide for prewnting owrfilling the Reanor Vessel.

l

10. 211000K6.03 Effect that a loss or malfunction of A.C. Power will have on the Standby Liquid 3.2 l Control System.

i 3.4

11. 212000Al.11 Predict and/or monitor changes in parameters associated with operating the [

Reactor Protection System status lights and alanns.

212000K2.01 Electric Power supplies to the Reactor Protection System. 3.2

12. [t Knowledge of the effect that a loss or malfunction of RPS will have on the 4.0  ;

13. 212000K3.06 scram air header solenoid operated valves. l

14. 215004A4.01 Manually operate and/or monitor SRM muni rate and period. 3.9 l Operational implications of elevated Containment temperature efeas on lewi 3.6

15. 216000K5.07 indication as it applies to NBl.

16, 217000A1.03 Predict and/or monitor changes in Reactor Water Lewl parameters associated 4.0 with operating RCIC.

17, Ope.ational implications of ADS Logic Operation as it applies to the Automatic 3.8 218000K5.01 Depressurization System.

Manually operate and/or monitor Hydrogen Recombiners. 3.4

18. 223001 A4.13 P imary Containment System and Auxiliaries design feature (s) and/or interlocks 3.1

19. 22300lK4.06 which provide for maintaining proper containment /sewndary containment to i Drywr!! differentialpressure.

12

2 .

Facility: WNP2 Kurledge cnd Ability Rec:rd Form Exam date: October 7,1996 ref: NUREG/BR-0122, Rev. 5 K/A Topic (Systems - Group I) RO Rating

20. 223002K6.07 Effect that a loss or malfunction of Essential A.C. Power will have on the 3.2 Primary Containment Isolation / Nuclear Steam Supply Shut-off System.

21. 223002A2.10 Ability to predict the impact of the following on the Primary Containment 3.9 Isolation /Nuc! car Steam Supply Shut-ofSystem: Loss of Coolant Accident (s).

22. 239002K4.05 Relief / Safety Valves design feature (s) and/or interlocks which provide for SRV 3.6 operationfrom enore than one location.

23. 241000K3.02 Knowledge of the effect that a loss or malfunction of Reactor / Turbine Pressure 4.2 Regulating System will have on Reactor Pressure.

24. 241000A2.03 Ability to predict the impact of the following on the Reactor / Turbine Pressure 4.1 Regulating System: Failed open/ closed bypass valve (s).

25, 259001K3.01 Knowledge of the effect that a loss or malfunction of Reactor Fredwater System 3.9 will have on Reactor Water Level.

26. 261000K4.01 Standby Gas Treatment System design feature (s) and/or interlocks which 3.7 l provide for automatic system initiation. ,

3.6  !

27. 264000K6.08 Effect that a loss or malfunction of A.C. Power will have on the Emergency Diesel Generators.

28. 264000A3.05 Monitor automatic operations of the Emergency Diesel Generators including: 3.4 Load shedding and sequencing.

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.- . - - _ - . -. _ _ . - -_ _ .. .__ ._.= _ ~ -_ _ ~--. - _ -.

e e i Facility: WNP2 INDIVIDUAL WALK-THROUGH TEST -OUTLINE ES-301-2 I'

Examination Level: SRO (Upgrade)

Week of Examination: October 7,1996 I I

Examiner's Name: j

\

l J JPM# System Safety Area Description of JPM i Function 1 I I

] 3 Control Room Ventilation 9 Sim. Startup Control Room Ventilation.

j 290003G A.09 (3.6/3.5) (A/P) a i

2 RPS 7 Plant Restart of RPS-MG-1.

l 212000GA9 (4.2/4.2)

I 7 Emerg. Diesel Generator 6 Plant Perform MANUAL Start of HPCS DG from the 264000A4.04 (3.7/3.7) local Panel.

{ l Suppression Pool 5 Plant Reduce SUPPRESSION POOL Level from the 219000A4.13 (3.9/3.8) (S/D) Remote Shutdown Panel.

5 EOP E/A CR Override ECCS Valve Logic to throttle RPV 295015G A.06 (4.1/3.9) Injection.

1I ' Conduct of Operations' Admin. Sim. Generator Exitation Curve (Spider Curve)

, 294001 A1.08 (3.1/3.6) interpretation with given situation.

Examiner: Chief Examiner:

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1 2 JPM ChmAllet per l'5 388

1. 10 $ROG)*O applaams JPMs wt 7 esmnd man sai 3 In ples. 2 S $ ROM JPMs wt 2 or 3 Caravai . sun ami 2 er 3 in.ples.

3 At naasi 7 difteam usiety functuwe few EROd6*O's, 4. As iness 5 ddimm salsty festume far $ ROM spplicemas.

1

5. I Ceasel same JPM mmee be en l'1F. 6. For each eveism selected. esima l samtug OR dewlap I arw JPM

]

1. At best t JPM missed to elemhme er lne pneer casinian. 8. I er 2 JPMe sequim 'sisernmar poise *,

9 At base t 'in p6mm' JPM mousse EDP er Abusenal eduwe 10. At luost t 'in pines

• JPM ergaises esson into red. sammlied ama j

14. *!Mwmify* er pmeceipeed speetiam enuuse s Ka, As. mal (e 12. lass than 305 overisp fiam lust NRC laem.

4 83. As basi 2 NrW ar signirummly alwsed JPMs for $ rod)%0's 14. An enesi i NEW er sigsfuusly ahesed JPM for SROM.

i 15. Adnumstatew inpas shneid to evehened in IPMe w4enever pramihin,

, miner een pseecnpied eastennn.

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. m Facility: WNP2 INDIVIDUAL WALK-TIIROUGli TEST -OUTLINE ES-301-2 l

l 1 Examination Level: SRO (Instant) i Week of Examination: October 7,1996 Examiner's Name:

l J

JPM# System Safety Area Description of JPM Function

! 6 A.C. Electrical Dist. 6 Sim. Synchronize the MAIN GENERATOR with the j I 262001 A4.04 (3.6/3.7) GRID. l 1

4 3 Control Room Ventilation 9 Sim. Startup Control Room Ventilation. )

4 290003GA 09(3.6/3.5) (A/P) 4 Standby Liquid Coatrol Sys. 1 Sim. Operate SLC BORON Sytem for RPV Injection.

, 211000A4.04(4.5/4.6) (A/P) 8 RCIC 2 Sim. initiate RCIC for RPV injection " ARM" and

! 217000A2.10 (3.1/3.1)

• DEPRESS".

l 217000A2.ll (3.1/3.2) 9 RSCS 7 Sim. Place RSCS into service.

201004A402 (3.5/3.2) i 10 LPCI 4 Sim. Align LPCI-C to Standby status.

203000A4.02 (4.1/4.1) j 2 RPS 7 Plant Restart of RPS-MG-1.

212000G A9 (4.2/4.2)

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! 7 Emerg. Diesel Generator 6 Plant Perform MANUAL Stan of HPCS DG from the 264000A4.04 (3.7/3.7) Local Panel.

1 Suppression Pool 5 Plant Reduce SUPPRESSION POOL level from the 219000A4.13 (3.9/3.8) (S/D) Remote Shutdown Panel.

I 5 EOP E/A CR Override ECCS Valve Logic to throttle RPV 295015G A.06 (4.!/3.9) Injection.

11

• Conduct of Operations" Admin. Sim. Generator Exitation Curve (Spider Curve) 294001 A1.08 (3.1/3.6) interpretation with given situation.

l Alternate RFW 2 Sim. Reactor Feed Pump Quick Start following a (12) 259001 A4.02 (3.9/3.7) manual trip.

4

Examiner

Chief Examiner:

JPM Chattlal per 53-301 I

l 1. 10 EROGn%O sppliinvue JPMs of 1 Cansmi sesn end 3 twelma. 2 5 EROfU) JPMs wt 2 or 3 Cereral siman and 2 or 3 trepiam. l

3. As best 7 d4Nes safety fweakas fw SHO(1)*0's 4 As besi 5 dibseis enfeiy funnuw, im SROru) eppliaaans-

6. I Canind raurn JPM smet be en flr. 6. l'ar neeb sysum seisceed, selmet I samting OR aleveinp i nrw JPM.

)

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7. At best 1 JPM selsed to elmadown er k=> pner ammiitm S. I er 2 JPMs apa 'elsemmis pods'. ~

< 9. At best I *Li plem* JPM segusse EOP or Amerwmal amenes. 10 At best I *si plam* JPM equeos esann kno end. ernandled esos.

j al. *Demesnify* cs ; W qwetwas enue:g es he. As. eral Gs 12 Imes them 305 enerlap fnwn bei NRC Enem.

13 As best 2 NrW or signibently elessed JPMe fur EROGvkO's. 14. M b=si I N!'W or swenkenly alwed JPM for ERO(U).  ;

ll. Admaususine tap es seenuld be evolmend in JPMe wtonswer presibis,  !

suiker than pnewtpeed quetkos 21

, a e Facility: WNP2 INDIVIDUAL WALK-TIIROUGil TEST -OUTLINE ES-301-2 1

Examination Level: RO 1 Week of Examination: October 7e 1996

)

{ Examiner's Name:

JPM# System Safety Area Description of JPM Function

] ,

4

) 6 A.C. Electrical Dist. 6 Sim. Synchronize the MAIN GENERATOR with the 262001A4.04 (3.6/3.7) GRID.

3 Control Room Ventilation 9 Sim. Startup Control Room Ventilation.

290003GA.09(3.6/3.5) (A/P) s 4 Standby Liquid Control Sys. 1 Sim. Operate SLC BORON Sytem for RPV Injection.

j 211000A4.04(4.5/4.6) (A/P) l 8 RCIC 2 Sim. Initiate RCIC for RPV injection " ARM" and 217000A2.10 (3.1/3.1)

• DEPRESS *.

{

217000A2.ll (3.1/3.2) 9 RSCS 7 Sim. Place RSCS into service.

1 201004A402 (3.5/3.2) i

~

10 LPCI 4 Sim. Align LPCI-C to Standby status.

. 203000A4.02 (4.1/4.1) i

! 2 RPS 7 Plant Restart of RPS-MG-1.

212000GA9 (4.2/4.2) 1 i 7 Emerg. Diesel Generator 6 Plant Perform MANUAL Start of HPCS DG from the i 264000A4.04 (3.7/3.7) Local Panel.

l 1 Suppression Pool 5 Plant Reduce SUPPRESSION POOL Level from the

219000A4.13 (3.9/3.8) (S/D) Remote Shutdown Panel, i 5 EOP E/A CR Override ECCS Valve logic to throttle RPV

} 295015G A.06 (4.1/3.9) Injection.

11

• Conduct of Operations

• Admin. Sim. Generator Exitation Curve (Spider Cun'e)

294001 A1.08 (3.1/3.6) interpretation with given situation.

, Alternate RFW 2 Sim. Reactor Feed Pump Quick Start following a (12) 259001 A4.02 (3.9/3.7) manual trip.

5 Examiner: Chief Examiner:

l

, JPM Chetthet per FS-3et

l. 10 EROGvit0 spphcoms JPMs w/ 7 Cauvel gaan and 3 in.ples. 2. 3 $8 TOM JPMs w/ 2 or 3 Conval santo sul 2 or 3 in-P i ani.

j i 3 As laest 7 dhient esseey femt=== for ERom'Ito's. 4. At best 5 d&ves sety factumn far $ ROM emplaams.

7. 1 Caraval eoran JPM nuan te en 13F. 6. For end syneem seir=1sd. who i saistus OR elewice I arw IPM.

7. At inest i JPM slated to shnadamn er km preer ennditian, 8 l or 2 JPMs sospaiss *almnets paths".

5 9. Al inest 8 'in pies

• JPM sequises E.OP er Aheennst actions. 10. At lomst 1 *in plant

• JPM napans esamt into ved, controlled anos.

II, "Divefeify' to poempeed treet' ee w onweg the lis. As. and Gs. 12. Ims than Mll overisp frara bat NRC r.aarn.

13, At best 2 Niv er significamly alweed IPMs for 1RomitO's. 14. At inest 1 NIV or eisrurummly abseed IPM far $ ROM.

I l $. Adnumstentew capus newnald be evolmend b IPMs whearver pnesihic.

sother than ymsmpeed speerismo.

22

- - . - . - . . - - - __ __~ - - - - _ - . ~ -. . . . . . . - .

, n. e Facility: WNP2 INDIVIDUAL WALK-THROUGII TEST -OUTLINE ES-301-2 JPM CROSS REFERENCE Pilot Exam JPM # Reaual JPM # Validation Times I LR000145/82-RJE-0004 5 Minutes 1 2- LR000251/82-RJE-0043 6 Minutes )

3 LR000209/82-RJE-0106 6 Minutes 4 LR000217/82-RJE-0133 5 Minutes 5 LR000233/82-RJE-0174 9 Minutes 6 LR000172/82-RJE@41 9 Minutes 7 LR000199/82-RJE-0087 7 Minutes 8 LR000302/82-RJE-0054 3 Minutes 9 LR000195/82-RJE-0082 9 Minutes 10 LR000159/82-RJE-0024 8 Minutes 11 (Admin.) To Be Developed 12 (Alternate) LR000131/NONE 16 Minutes TUESDAY SRO(U): JPM # 1 Plant SRO(l): JPM # 1 Plant RO: JPM # 1 Plant l JPM # 2 Plant JPM # 2 Plant JPM # 2 Plant 1 JPM # 7 Plant JPM # 7 Plant JPM # 7 Plant JPM # 5 CR JP!' # 5 CR JPM # 5 CR JPM # 3 Sim. JPM # 3 Sim. JPM # 3 Sim.

JPM # 11 Sim. JPM # 9 Sim. JPM # 9 Sim.

JPM # 11 Sim. JPM # 11 Sim.

SimulatorJPM's 3 & 9 were chosen such that they can be perfonned with two Candidates at a time verformine alternatine JPM's.

WEDNESDAY SRO(l): JPM # 4 Sim. RO: JPM # 4 Sim.

JPM # 10 Sim. JPM # 10 Sim.

1 JPM # 6 Sim. JPM # 6 Sim. j JPM # 8 Sim. JPM # 8 Sim. j JPM's for this day are crouped such that they can be perfonned with two Candidates at a time performine alternatine JPM's.

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• Facility: WNP2 ADMINISTRATIVE TOPICS - OUTLINE ES-301-1 Examination level: RO/SRO - (Circle one)

Administrative Describe method of evaluation:

Topic / Subject 1. ONE Administrative JPM, OR Description 2. TWO Administrative Questions A.1

• Conduct of Operations" 29400lKl.07 (3.3/3.6) - Planned follow-up questions for in-plant JPM.

294001Kl.07 (3.3/3.6) - Planned follow-up questions for in-plant JPM.

' Conduct of Operations" 294001 A1.08 (3.1/3.6) - Simulator JPM.

A.2 ' Equipment Control" 294001 A1.07 (3.0/3.7) - Planned follow-up questions on Simulator scenarios.

294001 A1.07 (3.0/3.7) - Planned follow-up questions on Simulator scenarios.  ;

1 A.3

• Radiation Control' 29400lKl.04 (3.3/3.6) - Planned follow-up questions for In-plant JPM.

29400!Kl.04 (3.3/3.6) - Planned follow-up questions for in-plant JPM.

A4 " Emergency Plan" 294001 A1.16 (2.9/4.7) - Planned follow-up questions on Simulator scenarios.

294001 A1.16 (2.9/4.7) - Planned follow-up questions on Simulator scenarios.

I Examiner: Chief Examiner; i l

l 1

JPM CherLlist per 11301

1. i 80 ERo(I)/Ro apphcants JPMs w/ 7 Control roons and 3 int ant. 2. 5 SRo(U) JPMs w/ 2 or 3 Control rman and 2 or 3 in-plant.

3. At least 7 different safety functions for SRo(I)/Ro's. 4. At least 5 different safety functions for SRo(U) applicants.

8, I Control rown JPM must be an ESF. 6. For each system sekcted, seket ! existing OR develop 1 new JPM.

7. At kast 1 JPM related to shutdown or km power condition. 8. I or 2 JPMs require ' alternate paths".

9. At kast I 'm plant" JPM requires 13or or Abnormal actums. 10. At Icast 1 *in plant

• JPM requires cacost mto red. controlled area.

II. *!hversify" the prescripted questions among the Ks. As, and Os. 12. Less than 30% overlap from last NRC Exam.

13. At least 2 NEW or signifwantly altered JPMs for sRo(T)/Ro's. 14. At kast I NEw or significantly altered JPM for SRo(U).

15. Administrative topics should be evaluated in JPMs whenever possible, rather than prescripted questxms.

24

Facility: WNP2 SCENARIO EVENTS - OUTLINE ES-301-3 SCENARIO NO. 01 - Small Steam Leak in the Drywell.

(ref: minor modification of LR000120 rev. 0)

Initial conditions: Plant operating at 100% power. No major evolutions planned /no major equipment is out of service. Plant is in full compliance with all T.S. and regulatory requirements.

Event no. Type

• Event Description

1. I Blown fuse in RC-1.

2. M Small Steam leak in the Drywell.

3. I/C APRM "E" failure.

4. N/R Scram Reactor per PPM 3.3.1 prior to high D/W pressure. Potential entry into PPM 5.1.1 and PPM 5.2.1.

5. N Initiate WW/DW Sprays as required.

6. C/M MSIV isolation on high temp in the Steam Tunnels establish RPV/P control with SRV's and RPV/L control with CBP's and/or RCIC.

Simulator Checkhd per ES-301

1. Each SRO-I and RO appbcant rotates to *kad" RO positmn.

2. Ini6al conditions should include: STARTUP, LOW POWER and FULL POWER.

3. EACH scenare must excersiae EACH applicant on: Normal evolution, reachvity manipulation, inst. failure, component failure and major plant transient

4. Each SRO-! MUST have a significant reactivity change.

5. Target Quantitative Attributea (per Scenario)

a. Total Malfuncuans = 5 to 8

b. Malfunctions after EOP entry = i to 2

c. Abnormal Events = 2 to 4

d. Major Transients = 1 to 2

c. EOP s entered /requinns substantive actmn = 1 to 2

f. EOP cantagencies requinns substantive action = 0 to 2 Examiner: Chief Examiner:

• (N)ormal (R)eactivity (I)nstrument (C)omponent (M)ajor 25

a

,. e j Facility: WNP2 SCENARIO EVENTS - OUTLINE ES-301-3 4

SCENARIO NO. 02 - ATWS, Emergency Depress.

j (ref: significant modification of LR000134 rev. 0) 1 l Initial conditions: 60% Power, RCIC OOS for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for small steam leak repair.

Event no. Type

• Event Description l 1. R Continue Plant Shutdown per Imad Dispatch.

, 2. N Remove second Reactor Feedwater Pump per Procedure.

! 3. C LPCS-P-2 shaft shear.

l 4. C RHR IAop 'A' suction line break, unisolable.

5. M Manual scram, Hydraulic ATWS.

j 6. I LPCS-V-5 fails to open forcing RHR B to be used per PPM 5.5.26 and

manual operation of RHR-V-3B and RHR-V-48 B.

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Enulator Checklist per ES-301

1. Each SRO-l and RO applicant rotates to "Acad" RO positics

2. Inkial conditiot.3 should include: STARTUP, LOW IOWER and FULL IV%T.R.

3. EACH scenario must excersise EACH applicant on: Nonnal evolution, reactivity manipula6an, inst failure, component failure and major plant transient.

4. Each SRO-l MUST have a significant reactivity change.

5. Targes Quanutative Attributes (per Scenario) a- Total Malfunctions = 5 to 8

b. Malfunctions after EOP entry = a to 2

c. Abnonnal Events = 2 to 4

d. Major Transients = 1 to 2

c. EOP's entered / requiring substantsvc action = 1 to 2

f. EOP contingencies requiring substantive action = 0 to 2 Examiner: Chief Examiner:

• (N)ormal- (R)eactivity (I)nstrument (C)omponent (M)ajor 26

i Facility: WNP2 SCENARIO EVENTS - OUTLINE ES-301-3 SCENARIO NO. 03 - Single Loop Operation, Loss of CAS.

(ref: significant modification of LR000096 rev. 0)

Initial conditions: Reactor Plant Startup is in progress, PPM 3.1.2 completed through section 4.8, Reactor power = 24%, SLC-P-1B is OOS.

Event no. Tvoe* Event Description

1. R/N Continue Start-up per PPM 3.1.2.

2. I RFW-LI-606A, RPV/L Monitor narrow range fails low.

3. C RRC "A" trips (PPM 2.2.1).

4. C Loss of CAS.

5. M Control Rod Drift /MSIV closure, entry into PPM 5.1.1, Electric ATWS, entry into PPM 5.1.2.

6. C RFW-V-10A & B fail closed.

I l

Simulator Checklist per FS-301

1. Each SRO-! and RO applicant rotates to

• lead" RO posi6an.

2. Initial conditions simuld include; STARTUP, LOW POWER and FULL PO%TE.  !

3. EACH scenario must eacessise EACH applicant on: Normal evolunon, reac6vity manipula6on, inst. failure, component failure and major plant transient. )

4. Each SRO-1 MUST have a sigruficant reac6vity change.  !

5. Target Quandtative Auributes (per Scenaru$

a. Total Malfunchans = 5 to 8

b. Malfunc6ans after EOP entry = 1 to 2

c. Abnormal Eventa = 2 to 4

d. Major Transients = I to 2

c. EOP's entered /requems substanuve acuan = 1 to 2 l

f. EOP contangencica requiring substantive ac6an = 0 to 2 I

i Examiner: Chief Examiner:

• (N)ormal (R)eactivity (I)nstrument (C)omponent (M)ajor 27

l

, h  % l Facility: WNP2 SCENARIO EVENTS - OUTLINE ES-301-3 l

SCENARIO NO. 64 - Small LOCA, unisolable RCIC steam leak.

(ref: significant modification of LR000140 rev. 0) l Initial conditions: Plant Turnover during Start-up (Power = 4%). APRM "C" OOS, I&C working.

Event no. Type

• Event Description ,

l 1

1. N RCIC full flow surveillance. l

2. R Rod withdrawl for plant start-up.

3. I/C Loss of SM-8 due to a ground.

4. C/M RCIC steam leak downstream of RCIC-V-45.

5. C RCIC-V-8 Failure to close.

6. M Small break LOCA.

7. I HPCS-V-4 fails to auto open.

Sissolater ChecLra, pee ESJ01

1. Each SRO-! and RO applicant rotates to 'lcad" RO posioon.

2. Initial conditions should include; STARTUP, LOW POWER and FULL PO%T.R.

3. EACH scenario must excersise EACH sppikant on: Normal evolution, reacnvity manipulatam, inst. failure, component failure and major plant transient.

4. Each SRO-I MUST have a significant reactivity change.

5. Target Quanutative Attritustes (per Scenario)

n. Total Malfunctions = 5 to 8

b. Malfunctions aher EOP entry = 1 to 2

c. Abnormal Events = 2 to 4

d. Major Transients = 1 to 2

c. l'OP's entered / requiring substannve acnon = 1 to 2 ]

f. EOP contingencies requirmg substantive actson = 0 to 2 l

Examiner: Chief Examiner:

• (N)ormal (R)eactivity (I)nstrument (C)omponent (M)ajor 1 28

.J e Facility: WNP2 SCENARIO EVENTS - OUTLINE ES-301-3 i SCENARIO NO. 05 - Loss of Control Room Annunciation, LOCA, ED on TAF.

, (ref: significant modification of LR000125)

! Initial conditions: Power = 96%, EOC coast down. Suppression Pool temp = 81*F. Maintenance just completed on CRD-P-1 A.

Event no. Type

• Event Descrintion

] 1. N Surveillance 7.4.3.1.1.50, RPS & Isolation Reactor Vessel Level IAw, I2 vel

, 3; RCIC Isolation, Level 8 - CFT is in progress currently at Step 7.1.13.

2. N Start CRD-P-1 A and secure CRD-P-1B.

3. I/C less of annunciators on P601,602 ar.d 603 due to ground on SI-2.

. 4. I/R Inadvertent initiation of HPCS due to instrument failure.

! 5. M Recire. suction line break.

4

6. C RHR-P-2A Shaft Shear.

i i

a 1

i J

4 i

i Simulator Cbedlist per f1301

l. Esch SRO-1 and RO applicant rotates to

• lead" RO posithm.

2. Initial condinons should include; STARTUP, LOW POWER and Full POWER.

3. EACH scenario must excersise EACH applicant on: Nonna! evolution, reac6vity manipula6cn, inst. failure, component fadure and major plant transient.

4. Each SRO-! MUST have a signicerant reac6vity change

5. Target Quanutadve Attributes (per Scenario)

a. Total Malfuncthms = 5 to 8

b. Malfunctions aher EOP entry = 1 to 2

c. Abnormal Events = 2 to 4

d. Major Transients = 1 to 2

c. EOP's entered /requinns substantive action

• I to 2

f. EOP contingencies requirmg substanuve acoun = 0 to 2 Examiner: Chief Examiner:

• (N)ormal (R)eactivity (I)nstrument (C)omponent (M)ajor 29