ML20207J468

Forwards NRC Approved Operator Licensing Exam (Facility Outline & Initial Exam Submittal & as-given Operating Exam) for Tests Administered on 980313 at Farley NPP

ML20207J468
Person / Time
Site:	Farley
Issue date:	03/01/1999
From:	Michael B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9903160291
Download: ML20207J468 (10)

Text

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

s i

l 1

March 1,1999 NOTE TO: NRC Document Control Desk Mail Stop 0-5s.)-24 FROM:

MW Beverly Michael, Licensi@ Assistant, Operator Licensing and Human Performance Branch, Division of Reactor Safety, Region 11

SUBJECT:

OPERATOR LICENSING RETAKE EXAMINATIONS ADMINISTERED ON MARCH 13,1998 AT THE FARLEY NUCLEAR PLANT DOCKET NOS. 50-348 AND 50-364 On March 13,1998, Operator Licensing Examinations were administered at the referenced facility. Attached, you will find the following information for processing through NUDOCS ar.d distribution to the NRC staff, including the NRC PDR:

Item #1 - a) Facility submitted outline and initial exam submittal, designated for distribution undar RIDS 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.

Attachments: As stated 9903160291 990301 PDR ADOCK 05000348 V pm

/9mr *W ES-301 Individual Walk-through Test Outline 1 Form ES 301-2 Examination Level (Circle One): RO/ ROU / SRO(U)

Facility: b /b'g/ I E Week of Examination:

3 //J M /

Examiner's Name (print):

System / JPM Safety Planned Follow-up Questions:

Function K/A/G // Importance // Description S/C 1. CROS: Recover a masabgned Control Rod (CRO 338) i 1 Reactivity A. KA 001K1.0514.5/4.41 Determine the cause of a reactor trip given specific plant Control conditions.

ABN,F B. KA: 001 K6.11 12.0/3.21 Determine the expected indications for a given set of conditions. .

SIC 2. ECCS: Raise the "A" 2. RCS inventory Accumulator Pressure (CRO- Control A. KA: 011E A1.0514.3/3.9) Determine the consequences for inadegaste procedure 076) ESF compliance B. K A: 2.4.414.0/4.31 Determine expected system auto configuration for a set of off-normal conditions.

S 3. PZR FCS: Respond to e stuck 3. heactor Press open spray valve. (New - ARP- Control A. KA: 008AK3.02 (3.6/4.11 Determine expected system parameters for a given set 1.8 C1 CRO-998) ABN,F of conditions.

B. KA: 015/017AA1.0513.8/3.81 Ability to operaate or monitor RCS flow indication S/C 4. RCP; Start an RCP (CRO- 4. Rx Core Heat 043A) Removal A. KA: 002K1.1314.1/4.21 Determine the correct operator response to a given set of F off-normai conditions.

LP B. KA: 003K1.0313.3/3.61 Determine expected system response to a given change in plant conditions.

S/C 5. CS: Ahgn Contasnment Spray for post accident 5. Containment A. KA: 022A1.1013.6/3.71 Determine the offects on system operations for a given recircdation (CRO.346) Integrity rnalfunction.

ABN B. K A: 026K4.02 (3.1/3.61 Determine affect of insufficient recirculation time S 6. AC; Manually synchronize the T/G to the grid. (New - SOP. 6. Electrical A. KA: 045K3.0112.8/3.1] Determine equipment / component malfunction from a given 28.1, sec 4.8 CRO-999) set of g.lant conditions. I l

B. KA: 2.1.2 (3.0/4.01 Determine limiting conditions and constraints of operating plant equipment ,

S/C 7. WGDS: Control room )

operations for Liquid Waste 9. Radioactivity A. KA: 073K1.01 13.1/3.61 Dete mine the source of a radioactive release from a given i Release (CRO.277) Release set of off-normalindications. '

P 8. IA: Restore Compressed Air 8. Plant Service systems af ter and auto-isolation. Systems A. KA: 078K3.0213.4/3.61 Determine expected system response to a given set of off.

150-444) normal conditions.

{

B. KA: 078A3.01 (3.1/3.21 Determine expected system response to f ailed system component j P 9. AC: Rack in an Alles Chalmers  !

4160V Breaker 6. Electrical A. KA: 058AA2.0313.5/3 91 Determine expected system response to a loss of power. j i

B. KA: 2.1.33 [3.4/4.01 Determine Technical Specification requirements in response to i a givens set of off-normal cond.tions. l P 10. CVCS: Place 1 A BAT O/S 1. Reactivity and 18 BAT on Recirculation Control A. KA: 2.1.2313 9/4.01 Determine the required operator actions for a given set of off-(S0-090) RCA normal conditions.

B. KA: 004 A1.1013.7/3.91 Deterrnme expected system response to a givens set of plant condrtions. i 1

S = Simulator, P = Plant, L = Laboratory, C = Control Room, F = Faulted / Alt. Path. T = Time critical, RCA= Radiological Control Area, = Low Power, ESF = Eng'neered Safety Features, ABN Examiner: O/ Mhief Examiner: - . . l y g  ;

DISTRIBUTION CODE A070

, ES-401 PWR SRO Examination Outline Form ES-401-3 0 .

{

Facility: Farley Nuclear Plant Date ofExam: March 2,1998 Exam Level: SRO K/A Category Points Tier Group Point Total K K K K :l K-n K. A A A' X G 1 2 3 74 1 152 ?6? 1 2 635 24?

1. 1 3 1 7 2 6 5 24 Emergency &

Abnonnal Plant Evolutions 2 2 1 2 3 4 4 16 3 0 1 0 0 1 1 3 Tier 5 3 9 5 11 10 43 Totals 1 3 2 1 2 2 1 1 2 2 2 1 19 2.

Plant Systems 2 4 'l 1 2 2 2 1 1 0 1 2 17 3 1 0 1 0 0 0 0 0 1 1 0 4 Tier 8 3 3 4 4 3 2 3 3 4 3 40 Totals

3. Generic Knowledge and Abilities Cat 1 Cat 2 Cat 3 Cat 4 17 5 5 4 3 l Note:

. . Attempt to distribute topics among all K/A categories; select at least one topic from every K/A category within each tier.

• Actual point totals must match those specified in the table.

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

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

. The shaded areas are not applicable to the category / tier.

NUREG-1021 - 23 of 39 Interim Rev. 8, January 1997

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

Emergency and Abnormal Plant Evolutions - Tier 1/ Group 1 E/ APE # / NameI Safety Function K1 K2 K3 A1 A2 G K/A Topic (s) Imp. Points '

000001 Continuous Rod Withdrawal /I 01 Bank select switch 3.2 1 000003 Dropped Control Rod 11 411 Knowledge of abnormal procedure conditions 3.6 1 000006 Inoperable / Stuck Control Rod II 06 Bases for power limit, for rod misalignment 3.6 1 000011 Large Break LOCAIlit 10 Verification of adequate core cooling 4.7 1 WE02 Si Termination f Ill 2 Facliity's heat removal system 3.9 1 000015/17 RCP Malfunctions / N 02 Consequences of an RCPS failure 4.1 1 E10 Natural Cire.I N 3 Manipulation of controls 3.6 1 000024 Emergency Boration Ii 01 Relationship between boron addition and change in Teve 3.6 1 000026 Loss of Component Cooling WateriVill 03 Guidance actions contained in EOP for Loss of CCW/ nuclear service 4.2 1 ,

water 000029 Anticipated Transient wfo Scram f l 11 Initiating emergency boration 4.3 1 000040 (WE12)lVSteam Une Rupture - Excessive 03 Difference between steamline rupture and LOCA 4.7 1 Heat Transferi 000040 (WE12)lVSteam Line Rupture - Excessive 10 AFW System 4.1 1 Heat Transfer i WE08 RCS Overcooling - PTS / N 1 Selection of appropriate procedures 4.2 1 000051 Loss of Condenser Vacuurn / N 02 Conditions requiring reactor and/or turbine trip 4.1 1 000055 Station Blackout t VI 02 Actions contained in EOP for loss of offsite and onsite power 4.4 1 000055 Station Blackout iVI 416 Knowledge of EOP implementation hierachy and coordination with 4.0 1 other support procedures 000057 Loss of Vital AC Elec. Inst. Bus / VI 19 The plant auto actions that will occur on the loss of a vital ac 4.3 1 electricalinstrument bus 000059 Accidental Liquid RadWaste Rel.IIX 311 Ability to control radiation releases 3.2 1 000062 Loss of Nuclear Service Water / IV 06 The length of time after the loss of CCW flow to a component before 3.1 1 that component may be damaged 000067 Plant Fire On. site f lX 04 Actions contained in EOP for plant fire on site 4.1 1 000068 Control Room Evac.IVill 03 Transfer of AFW flow control valves and pumps to local control 4.3 1  !

000069 (W/E14) Loss of CTMT IntegrityiV 222 Knowledge of limiting conditions for operations and safety limits. 4.1 1 000074 (W!E06&E07) Inad. Core Cooling / IV 44 Recognize abnormalindications for system operating parameters 4.3 1 which are entry-level conditions for emergency & abnormal procedures.

000076 High Reactor Coolant Activity f lX 05 Corrective actions as a result of high fission-product radioactivity 3.6 1 levelin the RCS.

KIA Category Totals: 3 1 7 2 6 6 Group Point Total: 24 i

s ES-401 PWR SRO Esarnination Outline Form ES-4013 Emergency and Abnormal Plant Evolutions - Tier 1/ Group 2 E/ APE #1 Name i Safety Function K1 K2 K3 A1 A2 G K/A Topic (s) knp. Points 000007 Reactor Trip - Stabilization - Recovery 11 449 Ability to perform without reference to procedures those actions that 4.0 1 require immediate operation of system components and controls.

000007 Reactor Trip - Stabilization - Recovery 11 02 MFW System 3.7 1 000008 Pressurtzer Vapor Space Accident Itil 02 Sensors and detectors 2.7 1 000009 Small Break LOCAIIII 23 RCP tripping requirements 4.3 1 W!Ett Loss of Emergency Coolant Rectre.IN 1 Facility .onditions and selection of appropriate procedures during 4.2 1 abnormal and emergency operations 000022 Loss of Reactor Coolant Makeup Ill 08 VCT level 3.3 1 000026 Loss of RHR SystemIIV 01 Loss of RHRS during all modes of operation 4.3 1 000027 Pressurizer Pressure Control System 02 Normal values fx RCS pressure 3.3 1 Malfunction I111 000032 Loss of Source Range NilVII 133 Ability to ree nize indications for system operatino parameters 4.0 1 which are ent -level condet60ns for technical specifications 000033 Loss of Intermediate Range NI / VII 44 Ability to recognize abnormal indications for system operating 4.3 1 parameters wiiech are entry-level conditions for emergency and abnormal operating proceaures.

000037 Steam Generator Tube LeakIlli 07 Actions contained in EOP for S/G tube leak 4.4 1 000038 Steam Generator Tube Rupture Ilil 38 Cooldown of RCS to specified temperature 4.5 1 000054 Loss of Main Feedwater f lV 02 Effects of feedwater introduction on dry S/G 4.2 1 000054 Loss of Main Feedwater / IV - 449 Ability to perform without reference to procedures those actions that 4.0 1 require immediate operation of system components and controls.

W/E05 Inadequate Heat Transfer - Loss of 1 Facility conditions and selection of appropriate procedures during 4.4 1 Secondary Heat SinkIIV abnormaland emergency operations.

000065 Loss of instrument Air / Vill 08 Failure modes of air-operated equipment 3.3 1 K/A Category Point Totals: 2 1 2 3 4 4 Group Point Total: 16

,fi-.  ! t! ,.lh, i!!, ?; i e I Et[f!!iiI:! :!t L[itIlji>{ . Ih ILL -

a 3- s

. 1 t

n 4 io 3 P

S 1 1 1 E

m .

o r p F 0 0 1 Im 2 4 4 sr t

e e

m ra a

p c

e p

s

)

s m

( e i

c p sr o ro T of A

I f

si s

3 K nn p s iok tit p u r e ad i r

o r n cn t r

o io s i l

dc n t o

1 t s c r e a e o e eT i

d p

IW r.c y

r a

l t

a n.

es a n -

r f o T o

tn uo sr x'tne i

e c

t n

ore n i Oiu t le e o nl l o ta r P oo r yi tl r

u p iv t

t n ic l

c u aE nt o ii bh c o r

n C Aw O G ima al xP 3 Ela G 3 Om Rr 1 1 Som 2 Rt 2 A 4 1 WA Pd n 1 a A y 0 c

n e

g 3

r K 0 e

m E 2 3 K 0 1 1

K 0 n

o i

I I

I t

c n n u o l l

it i F e V I V

y m i i

t t e t f

l n r e

f a a e S M id w  :

i l c o s e e c P l a

v A e t m e g t o a L n is- T N r i f t 1 e l d f n

1. iz n O io r a E u f o P P s H s y A s l e s r

/ e u o o E r g P F L t

e 1 0 4 6 a 0 2 3 6 4- 0 0 0 C S 0 0

0 0

0 0 A I

E 9 0 9 K

4

, t a

+

ES401 PWR SRO Examination Outline Form ES.4413 .

Plant Systems . Tier 2/ Group 1 System #1 Name K1 K2 K3 K4 K5 KS A1 A2 A3 A4 G' K/A Topic (s) Imp. Points f a31 Control Rod Drive 01 T. ave and NoJoad T ave - 4.2 1 001 Control Rod Drive 128 Knowledge of the purpose and function of 3.3 1 i major system components and controis j 8B03 Reactor Coolant Pump 87 Minimizing RCS leakage beschanical seals) 3.4 1 I 004 Chemical and Volume Control 98 Sensors and detectors 2.8 1 l 004 Chemical and Volume Control 04 RCPs, including sealinjection flow 3.8 1 I 013 Engineered Safety Features Actuation 02 Reset of ESFAS channels 4.4 1 014 Rod Position Indication - 02 Loss of Power to the RPIS 3.6 1 f

015 Nuclear instrumentation 03 Verification of proper 7-- ." _ , 3.9 1 waiaty 017 In. core Temperature Monitor 02 Saturation and -- '- "._, of water 4.0 1 f

022 Containment Cooling et initiation of safeguards mode of operation 4.3 1 [

026 Containment Spray 01 Containment spray pumps 3.8 1 f

068 Condensate 04 Loss of condensas e pumps 2.8 1 069 Main Feedwater 04 SM3s water level control system 3.4 1 l 061 Auxiliary / Emergency Feedwater 01 RCS 4.8 1 f 061 Auxiliary / Emergency Feedwater 02 AFW automatic start loss of MFW 4.6 1  :

gg/GS 8'v't. N or seMy 063 D.C. Electrical Distribution 01 Major de loads 3.1 1 068 Liquid Radwaste 04 Reactor drain tank - 2.5 1 071 Waste Gas Disposal 94 Relat of 3.1 1 073 Process Radiation Monitoring System - 02 Radiation Monitor System Control Panel 3.7 1

[

I KIA Category Point Totals: 3 2 1 2 2 1 1 2 2 2 1 Group Point Total: 19  !

e i

t i

n

i

. t t

+

ES-401 PWR SRO Examination Outline Form ES-401-3 Plant Systems -Tier 2fGroup 2 i t .

l System # / Name K1 K2 K3 K4 K5 KS A1 A2 A3 A4 G  ! K/A Topic (s) knp. Points 002 Reactor Coolant 12 RW9 of temp. ave. and loop diff. 3.9 1 Temp. to loop hot-leg anil cold-leg temp  ;

indecations l

006 Emergency Core Cooding 16 RWST Levet and T_.. .,.;_ 3.9 - 1 010 Pressurizer Pressure Control 08 PZR LCS 3.5 1 j 011 Pressurizer Level Control 10 Failure of PZR level instrument - high 3.8 1 i 012 Reactor Protection 06 Automatic or manual enable /disade of RPS 3.5 1 trips 012 Reactor Protection 11 Trip setpoint calculators 2.9 1 016 Non-nuclear Instrumentation 12 S/G 3.5 1 i i 028 Hydrogen Recombiner and Purge Control 02 Flammable hydrogen concentration 3.9 1 f

, 033 Spent Fuel Pool Cooling 222 Knowledge of Ilmiting conditions for 4.1 - 1 .

operations and safety limits (

034 Fuel Handling Equipment 130 Abil 3.4 1 locate andyste components,.

036 Steam Generator 01 MFW/AFW systems 4.5 1 039 Main and Reheat Steam 06 RCS 3.7 l

1 i 062 AC ElectricalDistribution 01 Major system loads 3.4 1 064 Emergency DieselGenerator 97 Air receivers 2.9 1 f

064 Emergency Diesel Generator et Local and Remote operation of the EDfG 4.3 1 073 Process Radiation Monitoring 01 Those systems served by PRMs 3.9 1 075 Circulating Water System 01 Heat Sink 2.8 1 i

i K/A Category Point Totals: 4 1 1 2 2 2 1 1 0 1 2 Group Point Total: 17 4

i t

i r

&  ; jt iiL ,Ii r u .f!lI:!t iir i tI !i l !,  : 4 !l l t t , i E

3-t s _

. 1 0 i n

4 o S P 1 1 1 1 4 E

m r .

o p F m 3 9 6 1 I

3 2 3 3 I

R P

e h

t t

o e

)

g r

s a c h i

p s c S o i d W T C s A h e

/ ic C v l:

K h y l a b a t o

w v T s d p t e t n l o m n e o u i

o n c d P o

p s p O d m u S m a a o C o o t e r R C L S G G

0 4 t e3 np A e 1 i

ltu uo 3 1 OrG A 0 1 nt o2 ir t 2 ae A iTni 0 m-as 1 xm A Ee t 0

Os Ry 6 SSt K 0 Rna Wl PP 6 K 0 4

K 0 3 1 K 0 1 2

K 0 1

9 K 0 1 l

o r

- t n

k o n C a s e T s h r ap m c e a t y n a N

1 la v e u W Be  :

ts

1. o Q g in t a

m m e

f f in br o e R ie lo u T t

s l

e o T t n

y t a R C /

p S e r t io H e n m P iz e u y l

a r n D r u u o p o s m g id s a e s e m e t 1

0 e r o t a 4- R P C S C S 6 0

7 0

5 0

1 4

A

/

E 0 0 0 0 K

1 ES-401 Generic Knowledee and Abilities Outline (Tier 3) Form ES-401-5 a

~

Fae:iity: Farley Nuclear Plant Date of Exam: March 2,1998 Exam Level: SRO Category K/A # Topic Imp. Points Conduct of 2.1.13 Knowledge of facility requirements for controlling 2.9 1 Operations vital / controlled access.

2.1.5 Ability to locate and use procedures and directives 3.4 1 related to shift staffing and activities.

2.1.4 Knowledge of shift staffmg requirements. 3.4 1 2.1.21 Ability to obtain and verify controlled procedure 3.2 1 copy.

2.1.9 Ability to direct personnel activities inside the control 4.0 1 room.

Total 5 Equipment 2.2.13 Knowledge of tagging and clearance procedures. 3.8 1 Control 2.2.20 Knowledge of the process for managing 3.3 1 troubleshooting activities.

2.2.19 Knowledge of maintenance work order requirements. 3.1 1 2.2.17 Krawledge of the process for managing maintenance 3.5 1 activities during power operations.

i 2.2.26 Kno vledge of refueling administrative requirements. 3.7 1 Total 5 1 R*.diation 2.3.1 Knowledge of 10 CFR: 20 and related facility 3.0 1 Control radiation control requirements.

2.3.2 Knowledge of facility ALARA program. 2.9 1 2.3.7 Knowledge of the process for preparing a radiation 3.3 1 work permit.

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

Total 4 Emergency 2.4.38 Ability to take actions called for in the facility 4.0 1 Procedures emergency plan, including (if required) supporting or and Plan acting as emersency coordinator.

2.4.44 Knowledge of emergency plan protective action 4.0 1 recommendations.

2.4.27 Knowledge of fire in the plant procedure. 3.5 1 Torri 3 F Tier 1 Target Point Total (RO/SRO) 13/17

, .