ULNRC-03510, Forwards Exam Outline as Required by NUREG-1021, Operator Licensing Examiner Stds

Forwards Exam Outline as Required by NUREG-1021, Operator Licensing Examiner Stds

ML20137Y326
Person / Time
Site:	Callaway
Issue date:	12/17/1996
From:	Biermann F UNION ELECTRIC CO.
To:	Bundy H NRC
Shared Package
ML20137Y179	List: ML20137Y426 ULNRC-03510, Forwards Exam Outline as Required by NUREG-1021, Operator Licensing Examiner Stds
References
RTR-NUREG-1021 ULNRC-03510, ULNRC-3510, NUDOCS 9704230023
Download: ML20137Y326 (29)

Text

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Canaway Plant Post 01 face Box E20 fulton. Missoun 65251 9

Union Euccrmc

$3 December 17,1996 4-Howard Bundy, Lead Examiner U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 400 j

Arlington, TX 76011-8064 ULNRC-03510

~

SUBJECT:

PILOT INITIAL EXAMINATION OUTLINE On December 12,1996, we discussed exam sample plan materials with you at the Callaway Plant. Per our discussion, enclosed you will find the examination outline as required by NUREG-1021, " Operator Licensing Examiner Standards". If you have any questions regarding this material, please contact me at (573) 676-8404 or Bob Nelson at (573) 676-8402.

Respectfiilly, 4

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wcw (om,e j

rancis X. Biermanni Operating Supervisor, Training FXB/cjs

Enclosures:

Admin Section Plant Walkthrough (JPMs)

Simulator Scenarios Written Outlines Computer Disc with above information Ih QTd [

483 V

SCENARIO EVENTS Callaway Plant Initial License Exam - February 1997 SCENARIO # 1 EXAMINERS:

APPLICANTS:

1 1

INITIAL CONDITIONS 100% Steady State / BOIL Equilibrium Xenon TURNOVER Ultimate licat Sink (UllS) Level 111 Alarm IN - Need to Pump Down EVENTIMALF Event Event Description No.

No.

Type

• I N

UllS Pump Down Essential Service Water to Service Water

)

2 R

'C' RCP liigh Vibration / Power Reduction to < 48% Power 3

I Steam Pressure Channel Failure 4

I PZR Pressure Channel 455 Failure 5

C

'C' RCP Shaft Break - Pas tial Loss of Flow Reactor Trip - Failure of Main Turbine AUTO Trip.

6 M

Large Break LOCA 7

C LOCA Sequencer Failure 8

C Failure of AUTO Swap of RIIR Suetion Valves from RWST to CTMT Sump

• (N) Normal (R) Reactivity (1) In'strument (C) Component (M) Major gpsawn.ewwon

.)

=

SCENARIO EVENTS Callaway Plant Initial I,1 cense Exam - February 1997 SCENARIO # 2 EXAMINERS:

APPLICANTS:

l l

INITI AL CONDITIONS 30% Power Steady State TURNOVER Need To Remove Normal Charging Pump (NCP) from Service For WPA MD AFP Out Of Service EVENT MALF Event Event Description No.

No.

Type

• 1 N

Shift from the NCP to a Centrifugal Charging Pump i

2 I

PZR Level Channel Failure 3

I S/G Level Channel Failure 4

C S/G Tube Leak / Power Reduction for S/G Tube Leak 5

C/R Failure of Control Rod Bank D to Move / Auto and Manual /

Perform Plant Shutdown using Boration Only 6

M S/G Tube Rupture (-400 gpm)- Plant Trip 7

C Turbine Driven AFP Fails to Start Automatically 8

C Failure ofIligh IIcad injection (CCP Lockout / BI IIcader Valve)

• (N) Normal (R) Reactivity (I) Instrument (C) Component (M) Major 3013S2 IK1C/l 18 P4t/12/26%

feb

i SCENARIO EVENTS 1

Callaway Plant Initial License Exam - February 1997 j

SCENARIO # ll/U EXAMINERS:

APPLICANTS:

i INITIAL CONDITIONS 100% Steady State / HOU Equilibrium Xenon TURNOVER MD AFW Out of Service EVENT MALF Event Event Description No.

No.

Type

• i 1

I VCT Level Channel Failure (112) 2 N

Place Excess Letdown Into Service J

3 I

Feed Flow Channel Failure 4

R MFP Oil Leak / Downpower to Reciove MFP From Service 5

C SG PORY Failure Open During Downpower in Automatic /

ManualControl Works 6

C/M MFP Trip with AUTO Reactor Trip Failure / RO Ilandswitch Works 7

C/M Overspeed Trip of Turbine Driven AFP / Loss ofIIcatsink

• (N) Normal (R) Reactivity (I) Instrument (C) Component (M) Major 101-1580 DOCA 52 PM/12/26No fsb

l i

WALKTHROUGH TEST - OUTLINE i

Callaway Plant Initial License Exam - February 1997 EXAM LEVEL-SRO(l)

System / JPM Area Safety Questions - K/A // Importance // Description Function i

I-Emergency lloration / FR S.1 Plant I

000024K302 (4.4) Actions in EOP FR-S 1.

M l

URO-CTK07PCO24J (RCA) 000024G05 (3.6) Knowledge of Annunciators /Use of Procedures 2.

Shift inst. Ilus from II/U to Norm Plant VII 000057K301 (4.1) Reasons for actions on loss of NN bus

]

EOS SNNO3PA010J 000057A102 0.8) PZR Level control on loss of NN bus 3-Emer. Feedwater from Firewater Plant V

000054KiO2 (4.2) Feed ilot Dry Steam Generator M

EOE-SAE03PA0383 (RCA) 059000K419 (3.4) Temp Correct FWlV Accum Pressure 4-Place 112 Recombiner in Serv.

Plant VI 028MA401 (44) Manually operate llRPS Controls URO-SGS0lPFC28J (C/R) 0280M Gl1 0.5) Tech Spec as sciated with 112 Control 5.

Cooldown from Aux S/D Panet Sim V]ll

@@68All2 (4.4) ASP Controls indications

]

(NEW) 000068K312 (4.5) Sequence of Actions for Evacuation Swap CCW Service Loop Sim X

008000K401 (3.1) Auto Start of CCW pumps M

j 6.

URO-SEG02PBC21J 008030A304 (3.6) Auto Actions of System on St M

Horation of RilR System Sim IV 005000K407 (3.2) Systemlnterlocks M

7.

URO SEJ02PCC23J 005000K408 0,1) Line up RilR for Recire Makeup to RWST Sim

))

006000K204 (3.6) Power Supply to ESFAS valves 8-(NEW) 000022A108 (3.4) Monitor VCTievel M

Respond to Process Rad Monitor.

Sim X]

073000A402 (3.7) Monitor System with RMll inoperable 9.

URO-SSP 04PBC86)(F)

(A/P) 073000K401 (4.3) Release termination due to fligh Rad j

Source Range Inst. Failure Sim IX 015000K501 (2.9) SR Detector Operation -

M 10.

Use 4 inst. readings determine faulty instrument 015000K506 0 4)

URO-SSE03 PAC 851 Use 1/M plot to predict criticality.

wom p m2N mesemen wemance II Parallel Emer D/G to Bus Sim V]l

• 0] 'f}4 URO-SNE0lPCC093 Alternate g

JPM Checuist i

10 SRO(l)/RO JOMs w/ 7 CR and 3 in-plant 9

5 SRO(U) JPMs w/2-3 CR and 2-3 in-plant 2

7 Different Safety Functions for SRO(1) and RO 10 $ Different Safety Functions for SRO(U) 3 CR JPM must be and ESF 11 Each System Selected - 1 JPM developed or selected 4

1 JPM low power or Shutdown 12 1-2 JPMs require alternate path 5

I in plant JPM EOP or Abnormal Actions 13 lin-plant JPM requires escort into RCA 6 " Diversify" questions among Ks, As. and Gs 14 Less than 30% overlap from last NRC exam 7

i New or Sir.ificantly Altered JPMs 15 l new or sigmficantly altered JPM for SRO(U) 8 Admin Topics Evaluated in JPMs when possible sRol1012 DOCD st AM/12!I2N6 hb

WALKTHROUGH TEST - OUTLINE a

Callaway Plant Initial License Exam - February 1997 EXAM LEVEL-RO

)

System / JPM Area Safety Questions - K/A // Importance // Description Function I-Emergeracy Boration / FR S.!

Plant

[

000024K302 (4 2) Actions in EOP FR S.I.

M URO-CTK07PCO24j (RCA) 000024G05 (3.6) Knowledge of Annunciators /Use of Procedures 1

2.

Shift Inst. Bus from II/U to Norm Plant Vij 000057K301 (4.1) Reasons for actions on loss of NN bus EOS-SNNO3PA010J 000057A102 (3.8) PZR Level control on loss of NN bus I-Emer. Feedwater from Firewater Plant V

000054KiO2 (3.6) Feed ilot Dry Steam Generator M

EOE-SAE03PA0383 (RCA) 059000K419 (3.4) Temp Correct FWlV Accum Pressure 4.

Place 112 Recombiner in Serv.

Plant Vj 028000A401 (4.0) Manually operate !!RPS Controls URO-SGS0lPFC28)

(C/R) 028000Gil (3.5) Tech Spec associated with il2 Control l

5.

Cooldown from Aux S/D Panel Sim VIII MM68Al12 (4.4) ASP ControlsIndications j

1 000068K312 (4.5) Sequence of Actions for Evacustion (NEW)

Swap CCW Service Loop Sim X

008000K401 (3.1) Auto Start of CCW pumps M

6.

URO-SEG02PBC2]J 008030A304 (3.6) Auto Actions of System on SI M

Sim IV 005000K407 (3.2) System Interiocks M

7.

Boration of RilR System 00$000K408 (3.1) Line up RilR for Recire URO-SEJ02PCC23J Makeup to RWST Sim Il 006000K204 (3.6) Power Supply to ESFAS valves 8.

i (NEW) 000022A108 (3.4) Monitor VCT level M

Respond to Process Rad Monitor.

Sim X[

073000A402 (3.7) Monitor System with RMil Inoperable 9.

URO-SSP 04PBC86J(F)

(A/P) 073000K401 (4.3) Release termination due to liigh Rad Source Range Inst. Failure Sim IX Ol5000K501 (2.9) SR Detector Operation.

M 10.

Use d inst. readings determine faulty instrument 015000K506 (3.4)

URO-SSE03 PAC 85j Use 1/M plot to predict criticality.

II Parallel Emer D/O to Bus Sim V((

064050G01 (3.6) SNE02C06C DieselGen Surveillance 0

35 URO-SNE0lPCC09J Alternate JPM Checklist 10 SRO(1)/RO JOMs w/ 7 CR and 3 in-plant 9

5 SRO(U) JPMs w/2-3 CR and 2-3 in-plant 2

7 Different Safety Functions for SRO(l) and RO 10 $ DifTerent Safety Functions for SRO(U) 3 CR JPM must be and ESF 11 Each System Selected 1 JPM developed or selected 4

1 JPM low power or Shotdown 12 1-2 JPMs require alternate path 5

1 in plant JPM EOP or Abnormal Actions 13 lin-plant JPM requires escort into RCA 6 " Diversify" questions among Ks, As, and Gs 14 less than 30% overlap from last NR" exam 7

i New or Sigoilicantly Altered JPMs 15 1 new or significantly altered JPM for SkCG;)

8 Admin Topics Evaluated in JPMs when possible Ro-3012 DOC /1 $1 AM/12/12/%

re

i WALKTHROUGH TEST - OUTLINE Callaway Plant initial License Exam - February 1997 EXAM LEVEL-SRO(U) l System / JPM Area Safety Questions - K/A // Importance // Description Function I.

Emergency Boration / FR-S.1 Plant 1

000024K302 (4.4) Actionsin EOPIR S.I.

M URO-CTK07PCO24J (RCA) 000024G05 (3 6) Knowledge of Annunciators /Use of Procedures 3.

Emer. Feedwater from Firewater Plant V

000054Klo2 (4.2) Feed liot Dry Steam Generator M

EOE-SAE03PA0383 (RCA) 059000K419 (3.4) Temp Correct FWlV Accum Pressure 4.

Place 112 Recombiner in Serv.

Plant VI 028000A401 (4.0) Manually operate ilRPS Controls URO-SGS0lPFC28J (C/R) 028000Gil (3.5) Tech Spec associated with 112 Control

]

l 5.

Cooldown from Aux S/D Panel Sim VIII 000068Al12 (4.4) ASP Controts Indications f

(NEW) 000068K312 (4.5) Sequence of Actions for Evacuation 9.

Respond to Process Rad Monitor.

Sim XI 073000.1.402 (3.7) Monitor System with RMil Inoperable i

URO-SSP 04PBC86J(F)

(A/P) 073000K401 (4,3) Release termination due to fligh Rad 4

1 i

i 0640ml (4.0) mon 6C Diccien suncinance II Parallel Emer D/O to ilus Sim VIJ Alternate 064000K410 (4.0)

URO-SNE0lPCC09]

Monthly Surv. on IX)- Effect ofloss of Normal Feeder to ND Bus

]

JPM Checklist i

10 SRO(!)/RO JOMs w/ 7 CR and 3 in-plant 9

5 SRO(U) JPMs w/2-3 CR and 2-3 in-plant 2

7 Different Safety Functions for SRO(l) and RO 10 5 Different Safety Functions for SRO(U) 3 CR JPM must be and ESF 11 Each System Selected - 1 JPM developed or, elected j

4 1 JPM low power or Shutdown 12 12 JPMs requi c alternate path 5

I in plant JPM EOP or Abnormal Actions 13 lin-plant JPM requires escort into RCA 6 Diversify" questions among Ks, As, and Gs 14 Less than 30% overlap from last NRC exam 7

i New or Significantly Altered JPMs 15 1 new or significantly altered JPM for SROtU) 8 Admin Topics Evaluated in JPMs when possible sRoV3012 tioCf7 51 awl 2/12?%

fab l

ADMIN TOPICS - OUTLINE l

Callaway Plan Initial License Exam - February 1997 EXAM LEVEL-SRO(I)

Administrative Method of Evaluation:

Topic / Discussion

1. ONE Admin JPM, or Description

2. TWO Admin Questions Al Shift 194001 A.I.03 Q-Shift manning requirements Manning (3.4)

Activity 194001 A.l.11 Q-Shift activity briefing requirements.

a Coordination (4 l)

Plant Parameter 194001 A1.02 JPM - Calculate QPTR 2

Verification (3.9)

A2 Surveillance 194001^1.08 Q. Heat Balance Surveillance acceptance criteria Procedures (3.1) review.

Locked 194001 A1.06 Q-Requirements for altering position oflocked Components (3.4) components.

1 194001Kl.03 Q-High Rad Key control - Access to incore A3 Knowledge of (3.4) detectors.

Significant Radiation Hazards 194001Kl.03 Q-Dose limits and authorization to exceed.

(3.4)

A4 Emergency Action 194001 A1.16 JPM -Classify EVENT -(IN CONTROL ROOM)

Levels (44) 1 l

l i

JPM Checklist J

l 10 SRo(I)Ro JPMs w/ 7 CR and 3 in-plant 9

5 SRo(U) JPMs w/2-3 CR and 2-3 in-plant 2

7 Different Safety Functions for SRo(l) and Ro 10 5 DifTerent Salcty Functions for SRo(U) i 3

CR JPM must be and ESF ll Each System Selected - 1 JPM developed or selected 4

1 JPM low power or Shutdown 12 1-2 JPMs require attemate path 5

1 in plant JPM EoP or Abnormal Actions 13 lin-plant JPM requires escort into RCA 6 " Diversify" questions among Ks, As. and Gs 14 Less than 30% overlap from last NRC exam 7

i New or Significantly Altered JPMs 15 1 new or significantly altered JPM for SRo(U)

R Admin Topics Evaluated in JPMs when possible sROlloll DOCM 49 AW11/11%

rub

e ADMIN TOPICS - OUTLINE Callaway Plan Initial License Exam - February 1997 EXAM LEVEL-SRO(U)

Administrative Method of Evaluation:

Topic / Discussion

1. ONE Admin JPM, or Description 2.TWO Admin Questions Al Shift 194001 A.I.03 Q-Shift manning requirements Manning (3.4)

Activity 194001 A.I.11 Q-Shift activity briefing requirements.

Coordination (41) 1 Plant Parameter 194001 A1.02 JPM - Calculate QPTR d

Verification (3.9) j i

A2 Surveillance 194001^1.08 Q.11 eat Balance Surveillance acceptance criteria Procedures (3.1) review.

i Locked 194001 A1.06 Q-Requirements for altering position oflocked Components (3.4) components.

194001Kl.03 Q-liigh Rad Key control - Access to incore A3 Knowledge of (3.4) detectors.

Significant Radiation Ilazards 194001Kl.03 Q-Dose limits and authorization to exceed.

(3.4)

A4 Emergency Action 194001 A1.16 JPM -Classify EVENT - (IN CONTROL ROOM)

Levels (4 4)

JPM Checklist i

10 SRO(lyRo JPMs w/ 7 CR and 3 in-plant 9

5 sRO(U) JPMs w/2-3 CR and 2-3 in-plant 2

7 Different safety Functions for SRO(1) and RO 10 $ Different safety Functions for SRo(U) 3 CR JPM must be and ESF 11 Each system Selected - 1 JPM developed or selected 4

1 JPM low power or Shutdown 12 1-2 JPMs require alternate path 5

I in plant JPM 00P or Abnormal Actions 13 lin-plant JPM requires escort into RCA 6 " Diversify questions among Ks, As, and Gs 14 Less than 30% overlap from last NRC exam 7

i New or Significantly Altered JPMs 15 1 new or significantly altered JPM for SRo(U) 8 Admin Topics Evaluated in JPMs when possible sROL90ll DOC /9 49 AM/t2/1796 rsk

c >

ADMIN TOPICS - OUTLINE Callaway Plan Initial License Exam - February 1997 EXAM LEVEL-RO Administrative Method of Evaluation:

Topic / Discussion

1. ONE Admin JPM, or Description

2. TWO Admin Questions Al Shift 194001 A1.03 Q-Requirements for watch relief / turnover.

Turnover (2.5)

Procedure 194001A.01 Q-Use of Continuous Use Procedures.

Use (3.3)

Plant Paranster 194001 A1.02 JPM - Calculate QPTR Verification (41)

A2 Equipment 194001 A1.13 Q-Determine Proper ECCS valve alignment.

Control (4.3)

Tagging &

194001Kl.02 Q-Use of Valving / Switching Orders Clearances (3.7) 194001Kl.04 Q-Operation of Electronic Dosimeter.

A3 Knowledge of (3.3)

Significant Radiation Hazards 19400lKl.03 Q-Dose limits and authorization to exceed.

(2.8)

A4 Emergency Action 194001 A1.16 Q-E-Plan Notification requirements Levels (3.1) 194001 Al.16 Q-Lines of authority during RERP activation.

(3.1)

JPM Checklist i

10 SRo(l)/Ro JPMs w/ 7 CR and 3 in-plant 9

5 SRo(U) JPMs w/2-3 CR and 2 3 in-plant 2

7 Difrerent Safety Functions for SRo(l) and Ro 10 5 Different Safety Functions for sRo(U) 3 CR JPM must be and ESF 11 Each System Selected - 1 JPM developed or selected 4

1 JPM low power or Shutdown 12 1-2 JPMs require alteniate path 5

I in plant JPM EoP or Abnormal Actions 13 lin-plant JPM requires escort into RCA 6

• Diversify" questions among Ks, As. and Gs 14 Less than 30% overlap from last NRC exam 7

i New cy Significantly Altered JPMs 15 I new or significantly altered JPM for SRo(U) 8 Admin Topics Evaluated in JPM5 when possible RO-Ml-1IXK/911 AWil/2nM6 rub

,s CALLAWAY PLANT - WRITTEN EXAM OUTLINE REACTOR OPERATOR Kl K2 K3 K4 K5 K6 Al A2 A3 A4 SG Systems I 1

2 2

3 3

1 3

1 3

1 3

23 Systems 11 4

1 3

1 1

2 0

2 1

3 2

20 Systems 111 1

0 0

1 1

0 0

2 1

2 0

8 Emer/Abn I 2

1 4

3 4

2 16 Emer/Abn 11 1

1 3

4 3

5 17 Emer/Abn III O

0 1

0 1

1 3

TOTALS 9

5 13 5

5 3

10 13 5

6 13 87 Plant Generics =

13 TEST TOTAL 100 1

l 1

i 1

12/17/96 ROOUT. DOC /12/17/96/9.55 AM FXB

,u I

Knowledge and Abilities Record Form PLANT SYSTEMS PWR - Reactor Operator - 51%

Check if 194001 I

included K/A #

Statement Rating

1. Kn wiedge of how to conduct and verify valve lineups.

3.6 X

Kl.01

2. Knowledge of tagging and clearance procedures.

X Kl.02 3.7

3. Knowledge of 10 CFR 20 and related facility radiation X

Kl.03 control requirements.

2.8 4.

Knowledge of safety procedures related electrical j

X Kl.07 equipment.

3.6*

]

1

5. Knowledge of facility protection requirements, X

Kl.16 including fire brigade and portable fire-fighting 3.5 l

equipment usage.

6. Ability to execute procedural steps.

X A1.02

4. l

• 7.

Ability to locate and use procedures and station X

Al.03 directives related to shift staffing and activities.

2.5 I

8.

Ability to make accurate, clear, and concise verbal X

Al.05 reports.

3.6

9. Ability to maintain accurate, clear and concise logs, j

X Al.06 records, status boards and reports.

3.4

10. Ability to obtain and interpret station reference X

Al.08 material such as graphs, monographs, and tables which 2.6 contain system performance data.

j i1. At>ility to locate control room switches, controls, and X

Al.13 indications, and to determine that they are correctly 4.3

• reflecting the desired plant lineup.

1 1.A ty to use plant computer to oMain and evaluate X

Al.15 3.1 parametnc mformation on system and component status.

i

13. Ability to take actions called for in the Facility X

A 1.16 Emergency Plan, including (if required) supporting or 3.1 4

acting as the Emergency Coordinator.

~

2

,u Knowledge and Abilities Record Form PLANT SYSTEMS l

PWR - Reactor Operator - 51%

l Plant Specific Priorities System #

K/A #

K/A Topic Rating 013 MSFIS - Modification Purpose and Operations 059 LER 96-03 FWIS due to Trip of MFP Group I Plant Systems - 23%

j 001 Control Rod Drive System 022 Containment Cooling System 003 Reactor Coolant Pump System 056 Condensate System 004 Chemical and Volume Control System 059 Main Feedwater System 013 ESFAS 061 Auxiliary / Emergency Feedwater System 015 Nuclear Instrumentation System 068 Liquid Radwaste System 017 In-Core Temperature Monitor System 071 Waste Gas Disposal System 072 Area Radiation Monitoring System System #

K/A #

K/A Topic Ratina UE Task 001 010K5.21 1.

Prediction in change in boron concentration due to 3.4 CTK-03 power operation, dilution, or boration.

001 000K2.02 2.

Power supply flowpath to Reactor Trip Breakers 3.6 SSB-02A 001 000K4.02 3.

CRDS interlocks for control rod mode select.

3.8 SSF-Ol A 003 000A1.07 4.

Predict changes in RCS temperature and pressure 3.4 SBB-02A associated with operating a RCP.

003 000G.10 5.

Recognize abnormal RCP indications as entry level 4.1 SBB-04B conditons for OTO procedure.

004 000K6.01 6.

Spray / Heater performance requirements for Boron 3.1 SBB-02F Mixing.

004 000K2.02 7.

Knowledge of bus power supply operation to a CCP 3.3 SBG-00 013 000K4.03 8.

ESFAS design feature which provides MSLIS.

3.9 SAB-05A 4

001 000GK.04 9.

Purpose and function of major components and 3.7 SSS-01 controls in the AMSAC system 013 000K5.02

10. ESFAS system logic and reliability.

2.9 SSA-02 013 000A2.01

11. Predict the impact of a LOCA on the ESFAS 4.6

. 3 A-03 A 015 000A3.03

12. NIS permissives status during plant startup.

3.9 CTK-02 015 000K4.02

13. NIS interlocks to ROD control 3.7 SSE-02 017 020A4.02

14. Temperature values used for RCP operations in 3.8 CRK-08 Inadequate Core Cooling 022 000A1.01

15. Monitor CTMT temp while operating Cont. Cooling 3.6 SGN-02 022 000A3.01

16. Auto operation of the Cont. Cooling due to SIS 4.1 SGN-04 056 020G.10

17. Recognize abnormal indications for Condensate Pump 3.2 SAD-03 Operation 059 000Kl.04

18. Operation of the Feedwater Level Control Sys.

3.4 SAE-02C 059 000K3.02

19. Effect of Loss of MFP on the AFW system.

3.6 SAE-04 061 000K5.01

20. AFW flowrates required for RCS heat transfer 3.6 SAE-02K 061 000A1.04 21. Predict char.ge in CST level for AFW operation 3.9 SAL-02 061 000A3.03

22. Operation of AFW system on Auto Start 3.9 SAL-03 072 000K3.02 23. Loss of ARM effect on fuel handling 3.1 SKE-01 3

12/17/96

,Y Knowledge and Abilities Record Form PLANT SYSTEMS PWR - Reactor Operator - 51%

Group II Plant Systems - 20%

002 Reactor Coolant System 006 Emergency Core Cooling System 039 Main and Reheat Steam System 010 Pressurizer Pressure Control System 055 Condenser Air Removal System 011 Pressurizer Level Control System 062 AC Electrical Distribution System 012 Reactor Protection System 063 DC Electrical Distribution System 014 Rod Position Indication System 064 Emergency Diesel Generator System 026 Containment Spray System 073 Process Radiation Monitoring System 029 Containment Purge System 075 Circulating Water System 033 Spent Fuel Pool Cooling System 079 Station Air System 035 Steam Generator System 086 Fire Protection System l

System #

K/A #

K/A Topic Ratina UE Task 035 010K3.01 1.

Effect of a loss of Steam Generators on the RCS 4.4 SAB-04 039 000A4.04 2. Monitor operations of the TD AFP Turbine 3.8 SAL-02B 002 000K5.09 3.

Relationship between Pres and Temp for Satutation 3.7 SBB-00 and Subcooled Conditions i

011 000A2.03 4.

Loss of PZR Level effect on PZR Level Control 3.8 SBB-04G 033 000G.11 5.

Immediate actions for loss of SFP level 3.6 SEC-02 075 000A4.01 6.

Manual operation of the ESW pumps 3.2 SEF-02C 006 000K6.03 7.

Perfomance capability of SI pumps.

3.6 SEM-03 026 000K4.01 8.

Interlocks associated with Cont. Spray Ops in LOCA 4.2 SEP-02 029 000K1.03 9.

Cause/effect of CPIS on Containment Purge 3.6 SGT-02A 079 000Kl.01

10. Cause/Effect relationship between SA and IAS 3.0 SKA-00 086 000A4.02

11. Operate / Monitor KC008 Panel in the Control Room 3.5 SKC-01 062 000K2.01

12. Knowledge of NB/NG Power Supply to Major Loads 3.3 SNB-01 064 050G.07

13. Explain limits and precautions of DG operation.

3.6 SNE-OlB 064 000A3.07

14. Monitor Load Sequencer Operation on DG 3.6 SNE-02C 010 000Kl.02

15. Effect of a load shed signal on PZR Press. Control 3.9 SNF-04 063 000K3.02

16. Effect of Loss of DC Control Power to Components 3.5 SNK-01 012 000K6.03

17. Design of RPS Trip Logic Circuitry 3.1 SSB-00 l

012 000A4.03

18. Monitor Bypass Breaker Operation 3.6 SSB-04 014 000A2.03

19. Respond to a Dropped Rod / OTO-SF-0000x 3.6 SSF-02C 073 000Kl.01

20. Cause/effect relationship between Process RAD 3.6 SSP-04B Monitoring to Steam Generator Sampling 4

i l

~.

Knowledge and Abilities Record Form PLANT SYSTEMS PWR - Reactor Operator - 51%

Group III Plant Systems - 8%

005 Residual lieat Removal System 034 Fuel fiandling Equipment System 007 Pressurizer Relief Tank / Quench Tank 041 Steam Dump System 008 Component Cooling Water System 045 Main Turbine Generator 027 Containment Iodine Removal System 076 Service Water System 028 Ilydrogen Recombiner and Purge Control 078 Instrument Air System 103 Containment System System #

K/A #

K/A Topic Ratine UE Task 041 020K4.18 1.

Steam Dump operation with Turbine Trip 3.4 SAB-04 007 000A2.01 2.

Predict Impact of Stuck Open PORV on PRT 3.9 SBB-02 045 000A4.01 3.

Turbine SV,CV,CIV operation and indication 3.1 SCH-03 076 000Kl.19 4.

Physical Connections between Serv. water and 3.6 SEA-02A emergency heat loads.

008 010A3.01 5.

Monitor Operation of the CCW pump / interlocks 3.2 SEG-020 005 000A2.02 6.

Impact of Press Transient on RIIR in MODE 5 3.5 SEJ-03 028 000K5.01 7.

Explosive 112 Limits applied to H2 Recombiner 3.4 SGS-01D 078 000K3.02 8.

Effect on plant systems on loss ofInst. Air 3.4 SKA-02A l

i i

l i

5 a

l l

,O' j

Knowledge and Abilities Record Form EMERGENCY PLANT EVOLUTIONS 1

PWR - Reactor Operator - 36%

L Plant Specific Priorities System #

K/A #

K/A Tonic Rating

]

j Group I Emergency and Abnormal Plant Evolutions - 16%

l' 000005 Inoperable / Stuck Control Rod

'000055 Loss of Offsite and Onsite Power 2

000015 RCP Motor Malfunction 000057 Loss of Vital AC ElectricalInstrument Bus j

i 000024 Emergency Boration 000067 Plant Fire On Site 000026 Loss of Component Cooling Water 000068 Control Room Evacuation 2

000027 Pressurizer Pressure Control System 000069 Loss of Containment Integrity

. Malfunction 000074 Inadequate Core Cooling 4

000040' Steam Line Rupture 000076 High Reactor Coolant Activity 00005i Loss of Condenser Vacuum K/A #

K/A Topic Rating UE Task 000005EK3.01 1.

Basis for Emergency Boration Stuck Rod on Rx Trip 4.0 CRK-01 A 000040EAl.03 2.

Isolation of one Steam Line from Header on Steam 4.3 CRK-03E j

Break j

000055EKl.01 3.

Effect of Batt Disch Rates on Capacity Loss of AC 3.3 CRK-05B 000055EK3.02 4.

Basis for actions on loss of all AC in ECA-0.0 4.3 CRK-05C j-000074EA1.01 5.

Monitor RCS water inventory for Inadequate Core 4.2 CRK-08B j

Cooling j

i 000040EKl.01 6.

Steam line rupture consequences of PTS on the RCS 4.1 CRK-10B i

000067EK3.04 7.

Reasons for Actions for Fire in CR 3.3 CTS-02A i

000068EK2.01

8. Knowledge of the ASP layout 3.9 CTS-02B 000051EA2.02 9.

Determine when to Trip Turbine on Loss of Vacuum 3.9 SAC-04 000027EA1.01

10. Operation of PZR Heaters, Sprays, and PORVs 4.0 SBB-02r j

000015EA2.10

11. Determine when to Trip RCP on loss of seal inj or 3.7 SBB-04B cooling 000076G.08

12. Determine Tech Spec Actions for High RCS Activity 3.1 SBB-04E j

000026EK3.02

13. Auto actions of CCW system from an ESFAS 3.6 SEG-04 actuation 000069EA2.02

14. Evaluate Containment Integrity meets requirements 3.7 SGT-03E 000057EA2.19

15. Auto Actions that occur on loss of Vital AC Inst Bus 4.0 SNB-04B 000024G.10

16. Recognize entry conditions for Loss of Shutdown 4.1 SSF-02G Margin 6

l

, c'.

Knowledge and Abilities Record Form EMERGENCY PLANT EVOLUTIONS l

PWR - Reactor Operator - 36%

Group II Emereeney and Abnormal Plant Evolutions - 17%

00000)

Continuous Rod Withdrawal 000033 Loss ofIntermediate-Range 000003 Dropped Control Rod 000037 Steam Generator Tube Leak 000007 Reactor Trip 000038 Steam Generator Tube Rupture 4

000008 Pressurizer Vapor Space Accident 000054 Loss of Main Feedwater 000009 Small Break LOCA

'000058 Loss of DC Power 000011 Large Break LOCA 000059 Liquid Radioactive-Waste Release 000022 Loss of Reactor Coolant Makeup 000060 Gaseous-Waste Release 000025 Loss of Residual lleat Removal System 000061 Area Radiation Monitoring System Alarms j

000029 Anticipated Transient Without Scram 000032. Loss of Source-Range Nuclear j

K/A #

K/A Tonic Ratine UE Task I

000008EA1.06 1.

Pressurizer level response to a Vapor Space Leak 3.6 CRK-02 000009EK3.24 2.

Basis for S1 Terminations on Small Break LOCA 4.2 CRK-02A j

00001IG.11 3.

Immediate actions for Large Break LOCA 4.5 CRK-02E j

000038EA1.32 4.

Isolation of ruptured S/G per E-3 4.6 CRK-04 000038EK3.06

5. Actions contained in E-3 for inventory balance, 4.2 CRK-04B SGTR, and plant shutdown.

000029EK3.12 6.

Basis for actions in EOPs for ATWS 4.4 CRK-07A 000001EKl.03 7.

Relationship between reactivity and power to rod 4.0 CTK-04 movement 000054G.09 8.

Verify System alarm setpoints per OTA for Feedwater 3.6 SAE-04 loss 000037EA2.12 9.

Determine flow rate of S/G tube leak 3.3 SBB-04A 000022EA1.01

10. CVCS letdown and charging Operation for loss of 3.4 SBG-06A Charging 000025G.10

11. Recognize abnormal indications for Loss of RHR 4.2 SEJ-03A 000058EA1.01

12. Cross tie NK bus to alternate supply 3.4 SNK-01 000007EK2.03

13. Knowledge of the Rx Trip Status Panel SB069 3.5 SSB-04 0000610.09

14. Respond to an Area Radiaition Monitor Actuation 3.5 SSD-02A 000033EA2.02

15. Determine unreliable IR channel based on indicaitons 3.3 SSE-02C 000032G.11

16. Actions for Source Range Channel Failure 4.1 SSE-03A 000059EA2.01

17. Determine failure mode for a liquid process monitor 3.2 SSP-04C 7

J e

Knowledge and Abilities Record Form EMERGENCY PLANT EVOLUTIONS PWR - Reactor Operator - 36%

Group III Emereency and Abnormal Plant Evolutions - 3%

000028 Pressurizer Level Malfunction 000056 Loss of Offsite Power 000036 Fuel liandling Incident 000065 Loss ofInstrument Air K/A #

K/A Tepic Ratina UE Task 000028EA2.01 1.

Determine PZR level as a function of Power 3.4 SBB-02 000056K3.02 2.

Basis for actions contained in EOPs for loss of Offsite 4.4 SPA-02 Power 000065G.1C 3.

Respond to a loss ofInst Air per OTO-KA-00001 3.8 SKA-02 1

8

..a RD SUPPLEMENTAL ITEM SIIEET

.I System #

K/A #

K/A Topic Ratina UE Task 1.

2.

3.

4.

t 5.

6.

7.

8.

I 9.

1 10.

I 1.

l 12.

13.

14.

15.

16.

17.

18.

19.

20.

21.

22.

1 23.

J 9

,?'.

CALLAWAY PLANT - WRITTEN EXAM OUTLINE SENIOR REACTOR OPERATOR Kl K2 K3 K4 K5 K6 Al A2 A3 A4 SG Systems 1 0

2 2

2 3

1 2

0 4

1 2

19 Systems 11 3

0 0

1 2

3 0

1 1

4 2

17 Systems Ill 0

0 1

1 0

0 0

1 1

0 0

4 Emer/Abn I 2

1 6

MM iR=4 9e 4

6 NR re-v 5

24 Emer/Abn 11 1

0 2

iien iw

e;1 5

4 144 e 4

16 Emer/Abn 111 1

0 1

cisei i:;--E ucj 0

1 M:

Twir 0

3 TOTALS 7

3 12 4

5 4

11 13 6

5 13 83 Plant Generics =

17 TEST TOTAL 100 l

l 1

12/17/96 SROOUT. DOC /12/17/999:57 Ah!

FXB

~*

Knowledge and Abilities Record Form GENERIC RESPONSIBILITES PWR - Senior Reactor Operator - 17%

Check if 194001 included K/A #

Statement Rating UE TASK X

K1.01 1.

Knowledge of how to conduct and verify valve lineups.

3.7 X

Kl.02 2. Knowledge of tagging and clearance procedures.

4.1 X

Kl.03 3. Knowledge of 10 CFR 20 and related facility radiation control 3.4 requirements.

X Kl.07 4.

Knowledge of safety procedures related electdcal equipment.

3.7

• X Kl.16 5.

Knowledge of facility protection requirements, including Gre 4.2 brigade and portable fire-fighting equipment usage.

X A1.02 6. Ability to execute procedural steps.

3.9 X

A1.05 7. Ability to make accurate, clear, and concise verbal reports.

3.8 X

A1.06 8.

Ability to maintain accurate, clear and concise logs, records, status 3.4 boards and reports.

X Al.08 9. Ability to obtain and interpret station reference material such as 3.1 graphs, monographs, and tables which contain system performance data.

X A1.13 10. Ability to locate control room switches, controls, and indications, 4.1 and to determine that they are correctly reflecting the desired plant lineup.

X A1.15 11. Ability to use plant computer to obtain and evaluate parametric 3.4 information on system and component status.

X A1.03 12. Ability to locate and use procedures and station directives related 3.4 to shift staffing and activities.

X.

Al.1613. Ability to take actions called for in the Facility Emergency Plan, 4.4 including acting as the Emergency Coordinator.

X Kl.05 14. Knowledge of facility requirements for controlling access to

~

vital / control areas.

X Kl.13 15. Knowledge of safety procedures related to oxygen-deficient 3.6 environment.

X A1.01 16. Ability to obtain and verify control procedure copy.

3.4 Al.12 17. Ability to direct personnel activities outside the control room.

4. l

• X 2

12/17/96 SROOUT. DOC /12/17/96/9:57 AM FXB

4 a

Knowledge and Abilities Record Fonn PLANT SYSTEMS PWR - Senior Reactor Operator - 40%

Plant Specific Priorities System #

K/A #

K/A Tonic Ratina 013 MSFIS - Modification Purpose and Operations 059 LER 96-03 FWIS due to Trip of MFP Group I Plant Systems - 19%

001 Control Rod Drive System 026 Uxtainment Spray System 003 Reactor Coolant Pump System 056 Condensate System 004 Chemical and Volume Control System 059 Main Feedwater System 013 Engineered Safety Features Actuation 061 Auxiliary / Emergency Feedwater System System 063 DC Electrical Distribution System 014 Rod Position Indication System 068 Liquid Radwaste System 015 Nuclear Instrumentation System 071 Waste Gas Disposal System 017 In-Core Temperature Monitor System 072 Area Radiation Monitoring System 022 Containment Cooling System System K/A #

K/A Tonic Ratine UE TASK 001 000K2.02

1. Power supply flowpath to Reactor Trip Breakers 3.6 001 010KS.21 2.

Prediction in change in boron concentration due to 3.9 power operation, dilution, or boration.

003 000G.16

3. Recognize abnormal RCP indications as entry level 4.4 cedhns for OTO procedure.

004 000K6.01 4.

Spray /IIeater perfomrance requirements for Boron 3.3 Mixing.

004 000K2.02 5. Knowledge of bus power supply operation to a CCP 3.5 013 000A2.01

6. Predict the impact of a LOCA on the ESFAS 4.8 013 000K5.02 7.

ESFAS system logic and reliability.

3.3 013 000K4.03 8.

ESFAS design feature which provides MSLIS.

4.5 015 000A3.03

9. NIS permissives status during plant startup.

3.9 017 020A4.02

10. Temperature values used for RCP operations in 4.1 Inadequate Core Cooling 022 000A3.01

11. Auto operation of the Cont. Cooling due to SIS 4.3 022 000Al.01

12. Monitor CTMT temp while operating Cont. Cooling 3.7 001 000GK.04

13. Purpose and function of major components and 4.0 SSS-01 controls in the AMSAC system 061 000A1.04

14. Predict change in CST level for AFW operation 3.9 061 000A3.03

15. Operation of AFW system on Auto Start 3.9 061 000K5.01

16. AFW flowrates required for RCS heat transfer 3.9 072 000KS.02

17. Loss of ARM effect on fuel handling 3.5 026 000A3.01

18. Monitor Auto Operation of CS Pump & Valves 4.5 063 000K4.02

19. Interlocks /Permissives for NK Bus Operation 3.2 3

12/17/96 sROOUT. DOC /12/17/96/12:53 PM FXB

s's Knowledge and Abilities Record Form PLANT SYSTEMS PWR - Senior Reactor Operator - 40%

Group II Plant Systems - 17%

002 Reactor Coolant System 034 Fuel Handling Equipment System 006 Emergency Core Cooling System 035 Steam Generator System 010 Pressurizer Pressure Control System 039 Main and Reheat Steam System 011 Pressurizer Level Control System 055 Condenser Air Removal System 012 Reactor Protection System 062 AC Electrical Distribution System 016 Non-Nuclear Instrumentation System 064 Emergency Diesel Generator System 028 Ilydrogen Recombiner and Purge Control 073 Process Radiation Monitoring System 029 Containment Purge System 075 Circulating Water System 033 Spent Fuel Pool Coolint Systern 079 Station Air System 086 Fire Protection System System K/A #

K/A Tonic Ratine UE TASK 002 000K5.09 1.

Relationship between Pres and Temp for Satutation 4.2 ana Subcooled Conditions 006 000K6.03 2.

Perfomance capability of SI pumps.

3.9 011 000A2.03 3.

Loss of PZR Level effect on PZR Level Control 3.9 012 000A4.03 4.

Monitor Bypass Breaker Operation 3.6 012 000K6.03,5. Design of RPS Trip Logic Circuitry 3.5 028 000K5.01 6.

Explosive H2 Limits applied to H2 Recombiner 3.9 j

029 000Kl.03

7. Cause/effect of CPIS on Containment Purge 3.8

]

033 000G.ll 8.

Immediate actions for loss of SFP level 3.8 039 000A4.04 9.

Monitor operations of the TD AFP Turbine 3.9 064 000A3.07

10. Monitor Load Sequencer Operation on DG 3.7 073 000K 1.01

11. Cause/effect relationship between Process RAD 3.9 Monitoring to Steam Generator Sampling 075 000A4.01

12. Manual operation of the ESW pumps 3.2 079 000K1.01

13. Cause/Effect relationship between SA and IAS 3.1 086 000A4.02

14. Operate / Monitor KC008 Panel in the Control Room 3.5 010 000K4.03

15. PZR Design for Overpressure Control 4.1 035 010K6.01

16. Performance & Design Attributes of MSIVs 3.6 062 000G.08

17. Recognize entry actions for AC Dist per Tech Specs 4.3 4

12/17/96 SROOUT. DOC /12/17/96/9.57 AM FXB

N v

Knowledge and Abilities Record Fomi PLANT SYSTEMS PWR - Senior Reactor Operator - 40%

_ Group III Plant Systems - 4%

005 Residual lleat Removal System 041 Steam Dump System /

007 Pressurizer Relief Tank System 045 Main Turbine Generator System 008 Component Cooling Water System 076 Service Water System 078 Instrument Air System Systeml K/A #

K/A Topic Ratine UE TASK 007 000A2.01 1.

Predict Impact of Stuck Open PORV on PRT 4.2 041 020K4.18 2.

Steam Dump operation with Turbine Trip 3.6 078 000K3.02 3.

Effect on plant systems on loss ofInst. Air 3.6 008 010A3.03 4.

Requirements for CCW system for different plant 3.2 conditions 5

12/17/96 SROOUT. DOC /12/17/96/9:57 AM FXB

e-Knowledge and Abilities Record Form EMERGENCY PLANT EVOLUTIONS PWR - Senior Reactor Operator - 43%

l Group I Emergency and Abnormal Plant Evolutions -24%

j 000001 Continuous Rod Withdrawal 000051 Loss of Condenser Vacuum 000003 Dropped Control Rod 000055 Loss of Offsite and Onsite Power 000005 Inoperable / Stuck Control Rod 000057 Loss of Vital AC Electrical Instrument Bus 000011 Large Break LOCA 000059 Accidental Liquid Radioactive - Release 000015 RCP Motor Malfunction 000067 Plant Fire On Site 000024 Emergency Boration 000068 Control Room Evacuation 000026 Loss of Component Cooling Water 000069 Loss of Containment Integrity 000029 Anticipated Transient Without Scram 000074 Inadequate Core Cooling 000040 Steam Line Rupture 000076 High Reactor Coolant Activity K/A K/A Tonic Ratina UE TASK 000005EK3.01 1.

Basis for Emergency Boration Stuck Rod on Rx Trip 4.3 000024G.10 2.

Recognize entry conditions for Loss of Shutdown 4.2 Margin j

000026EK3.02 3.

Auto actions of CCW system from an ESFAS actuation 3.9 000029EK3.12 4.

Basis for actions in EOPs fer ATWS 4.7 000040EAl.03 5.

Isolation of one Steam Line from Header on Steam 4.3 Break l

000040EKl.01 6.

Steam line rupture consequences of PTS on the RCS 4.4 l

000051EA2.02 7.

Determine when to Trip Turbine on Loss of Vacuum 4.1 000055EKl.01 8.

Effect of Batt Disch Rates on Capacity -Loss of AC 3.7 l

000055EK3.02 9.

Basis for actions on loss of all AC in ECA-0.0 4.7 000057EA2.19 10. Auto Actions that occur on loss of Vital AC Inst Bus 4.3 000059EA2.01

11. Determine failure mode for a liquid process monitor 3.5 000067EK3.04

12. Reasons for Actions in Procedures for Fire on Site 4.1 000068EK2.01

13. Knowledge of the ASP layout 4.0 000069EA2.02 14. Evaluate Containment Integrity meets requirements 4.4 000074EA1.01

15. Monitor RCS water inventory for inadequate Core 4.4 Cooling l

000076G.08

16. Determine Tech Spec Actions for High RCS Activity 4.0 000015EA2.10 17. Determine when to Trip RCP on loss of seal inj or 3.7 cooling 000001EK3.02

18. Basis for Tech Spec Limit on control rod operability 4.3 I

000003G.07

19. Explain limits and precautions for Dropped Rod 4.2 Procedure 0000llEA1.14 20. Monitor Subcooling Meters during Large Break LOCA 4.1 00001 IG.07

21. Apply precautions and limitatic:ns during LOCA 4.2 000024EA2.04 22. Determine the availability of Borated water sources 4.2 4

000029G.11 -

23. Ability to perform immediate actions for ATWS 4.7 000057EAl.01 24. Monitor Inverter Swapping on Loss of Vital AC 3.7 6

12/17/96 SROOUT. DOC /12/17/96/9.57 AM FXB

......_ _ _ s.. _ _.. _.... _... _

i Knowledge and Abilities Record Form PLANT SYSTEMS PWR - Senior Reactor Operator - 40%

1 1

25.

l I

j i

i l

4 t

l r

I t

)

l I

[.

7 12/17/96 SROOUT. DOC /12/17/96/9;57 AM FXB f

,~

Emergency Plant Evolutions (Continued)

Group H Emercency and Abnormal Plant Evolutions - 16%

000007 Reactor Trip 000037 Steam Generator Tube Leak 000008 Pressurizer Vapor Space Accident 000038 Steam Generator Tube Rupture 000009 Small Break LOCA 000054 Loss of Main Feedwater 000022 Loss of Reactor Coolant Makeup 000058 Loss of DC Power 000025 Loss of Residual lleat Removal System 000060 Accidental Gaseous-Waste Release 000027 PZR Pressure Cont System Malfunction 000061 ARM System Alarms 000032 Loss of Source-Range Instrumentation 000065 Loss ofInstrument Air 000033 Loss ofI-R Instrumentation K/A_

K/A Topic Ratina UE TASK 000008EAl.06 1.

Pressurizer level response to a Vapor Space Leak 3.6 000009EK3.24 2.

Basis for Si Terminations on Small Break LOCA 4.6 000022EAl.01 3.

CVCS letdown and charging Operation for loss of 3.3 Charging 000025G.10 4.

Recognize abnormal indications for Loss of RHR 4.4 000027EAl.01 5. Operation of PZR Heaters, Sprays, and PORVs 3.9 000033EA2.02 6.

Determine unreliable IR cha.mel based on indicaitons 3.6 000032G.11 7.

Actions for Source Range Channel Failure 4.2 000038EA1.32 8.

Isolation of ruptured S/G per E-3 4.7 000038EK3.06 9.

Actions contained in E-3 for inventory balance, SGTR, 4.5 and plant shutdown.

000061G.09

10. Respond to an Area Radiaition Monitor Actuation 3.6 000007EAl.03

11. Monitor RCS pressure and temp on Reactor Trip 4.1 0000370.05

12. Tech Spec Basis for S/G Tube Leaks and LCO 4.0 000054EKl.02 13. Knowledge of Effect of Feeding a Dry S/G 4.2 000058EA2.01

14. Determine a loss of DC power has occurred 4.1 000065EA2.06 15. Determine when to Trip Reactor on Decreasing IA 4.2 Pressure 000022EA2.01

16. Determine failure of flow control valve in CVCS 3.6 8

12/17/96 SROOUT. DOC /12/17/96/9.57 AM FXB

L-s Em:rgency Plant Evolutions (Continued)

Group III Emercency and Abnormal Plant Evolutions - 3%

000028 Pressurizer Level Malfunction 000056 Loss of Offsite Power 000036 Fuel liandling Incident K/A K/A Tonic Ratina UE TASK 000028EA2.01 1.

Determine PZR level as a function of Power 3.6

)

000056K3.02

2. Basis for actions contained in EOPs for loss of Offsite 4.7 l

Power 000036EK1.01 3.

Knowledge of Criticality during Fuel liandling.

4.3 i

l 9

12/17/96 SROOUT. DOC /12/17/96/9:57 AM FXB

I l

Knowledge and Abilities Record Form PWR-Supplemental item Sheet K/A K/A Toni[

Ratina UE TASK 1.

2.

3.

4.

5.

6.

7,

)

8.

9.

i l

10.

l l

l 1

l i

10 12/17/96 SROOUT. DOC /12/17/96/9:57 AM FXB r

_