ML042730577

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Written Exam Impact of SRO Question Deletions on Exam Validity
ML042730577
Person / Time
Site: Oyster Creek
Issue date: 09/24/2004
From: Caruso J
NRC/RGN-I/DRS/OSB
To:
Conte R
Shared Package
ML031620284 List:
References
50-219/04-301
Download: ML042730577 (37)
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Text

EM01 BWR SRBExamination Outline Form ES-401-1 Abilities Categories cf, @q9

Notes for post-exam distribution:

1. WA Change on SRO-7 from 295004/2.1.33 to 2.2.11 (Knowledge of the process for controlling temporary changes)
2. KIA Change on SRO-3from 295024EA2.02 to 295028/EK2.04 (Knowledge of the interrelationship between High Drywell Temperature and Drywell Ventilation)
3. SRO-12 dejeted
4. SRO-23 deleted.
5. SRO-25 deleted
6. It should be noted that all SRO applicants took the RO, 75 question written exam. The RO exam had two questions in Generic Knowledge and Abilities Category 3 "Radiation Control". These were Questions 64 and 65. Questions RO-2, RO-4, RO-23 and RO-31 also had Radiation Control aspects. Specifically, Question RO-23 (Tier 1, Group 1) has two knowledge aspects. One is related to High off Site release rates and the second is related to the biological effects of radioisotope Ingestion. The first one is testing RO SGTS and AOG system knowledge, while the second is really testing the SROs ability to make plant alignment changes based on the effects that the release would have (on the thyroid). This second part has a very close correlation to the SRO Generic Knowledge Category 3.

[In addition, for information, the Operating Test covered the Radiological aspect as follows: two JPMs with Radiation Control aspects; Admin 4s Approve Radioactive Discharge Permit" and PlantlSimulator

  1. !5, "Bypass Isolation Interlock for Torus Vent Valves and Prepare to Vent the Torus". Scenario ##4had a malfunction "RBHVAC Rad Ventiliation monitor fails upscale, RBHVAC trips, SGTS fails to staff in which the operators were required to manually start SGTS.]

Therefore, even though the SRO Written Exam had its Generic Knowledge and Abilities Category 3 "Radiation Control" question deleted (SRO-25) there were adequate opportunities in the remainder of the exam to evaluate the SROs in this generic category.

Note: 1. Ensure that at least two topics from every K/A category are sampled within each tier of the RO outline (Le., the Tier Totals in each K/A category shall not be less than two). Refer to Section D.1.c for additional guidance regarding SRO sampling.

2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by k1 from that specified in the table based on NRC revisions. The final RO exam must total 38875 points and the SRO-only exam must total 25 points.
3. Select topics from many systems and evolutions; avoid selecting more than two*

Driorities. &

tttfee WA topics from a given system or evolution unless they relate o plant-specific

,A$*

4.

5.

. 6.*

7.

8. tings, and point totals 9.

19 of 34 NUREG-1021, Draft Revision 9

I! Emergency anc I

BWRE normal lination Outline utions Tier l/Gmup 1 (RO ISRO) .

Form ES-401-1 1 ~

BAPE # I Name I Safety Function I: C K A

! 3 1 WA TopIc(s) 295001 Partial or Corn lete Loss of Forced AA2. Abiljty to determine andor interpret core FIOW Circulation?j 4 the foltowrng as they a ly to PARTIAL OR COMPLETE 4K:

LOSS FORCED CORE FLOW CIRCULATION :

(CFR: 41.10 / 43.5 / 45.13)

AA2.01 Powedflow map8 295003 Partial or Complete Loss ofAC16 295005 Main Turbine Generator Trip i3 295006 SCRAM I1 AA2. Ability to determine andlor interpret the following as the ap 1ytoSCRAM.

(d! 81.10 / 43.5 /45.13)

AA2.02 Control rod position.

295016 Control Room Abandonment I 7 I 29501 8 Partial or Total Loss of CCW/8 295019 Partial or Total Loss of Inst. Air / 8 I 2.2.27 Knowled e of the reheling process.

(CFR: 43.6 / 45.f3)

I 295021 Loss of Shutdown Cooling 14 ---

l 295023 Refueling Acc Cooling Mode I 8 I EA2.02 D w e l l kmerature.

1 I

~~

295025 High Reactor Pressure / 3 295026 Suppression Pool High Water Temp. / 5 I 295027 High Containment Temperature I 5 3 of 34 NUREG-1021, Draft Revision 9

ES-401 I I I I 11 \

2 2.4.20 Knowle e of operational implications of OP warnings I cautions I and I1 Group Pbint Total:

la l

ES-40I ES-401 camination Outline Form ES-401-1 Emergencyan( i lutions -Tier l/Group 2 (RO I SRO)

UAPE # I Name / Safety Function KIA Topic@)

295002 Loss of Main Condenser 2.4.31 Knowledge of annunciators alarms and Vac I 3 indications/ and use of the response instructions.

(CFR. 41.10 / 45.3) 295007 High Reactor Pressure 13 295008 High Reactor Water Level I2 295009 Low Reactor Water Level AA2.Ability to determine andlor inteIpret the 12 followln as they app&to LOW REACTOR WATER LEVEL (CFR: 41.lo / 43.5 I45.13)

AA2.03 Reactor water cleanup blowdown rate...................

2.9 2.9 295010 High Drywell Pressure I5 295011 High Containment Temp I 5

295012 High D 2.4.44 Ability to recognize abnormal Temperature / P" indications for s stem operating meters which are entry !eve1 conditions KIjmergenc Operatkg Procedures [related to High Drywely tern erature (13% 41.1d/43.5/45.13) 295013 High Suppression Pool Temo. I 5 295014 Inadvertent Reactivity AA2.03 Ability to determineandlor interpret Addition / 1 fhe cause of reactiyity adetion as it applies to madvertent reachvity ad&bon (CFR: 41.10/43.5 /45.13) 295015 Incomplete SCRAM / 1 295017 High Off-site Release Rate I 9 295020 Inadvertent Cont.

IsolationI5 & 7 295022 Loss of CRD PumDs / 1 295029 High Suppression Pool Wtr Lvl 15 5 of 34 NUREG-I 021, Draft Revision 9

ES-401 295032 High Secondary ContainmentArea Temperature /

5 295033 High Secondary Containment Area Radiation Levels / 9 295034 Secondary Containment Ventilation High Radiation I9 295035 Seconda Containment High Differential geswre / 5 295036 Seconda Containment High SumpfArea %ate, Level I 5 500000 High CTMT Hydrogen Conc. I5 WA Category Point Totals: Group Point Total: 7 I

4

ES-401 Ih 1 ES-401 BWR s R ( F E x nination Outline Plant Systf is - Tie p 1 (RO I SRO)

F OES-401-I

~

1 System # I Name WA Topic(s) IR #

I1 203000 RHWLPCI:

Injection Mode 205000 Shutdown Cooling 207000 Isolation 209002 HPCS 211000 SLC 212000 RPS 215003 IRM A2.Ability to (a) predict the impacts of 3.8 1 thefollowm on the IFJT~RMEDIATE RANGE MONITOR (IRM) SYSTEM ;and (b) based on those predictions, use procedures to correc control, or mitigate i le consequences of those abnormal conditions or operations:

(CFR 41.5 / 45.6)

A2.04 Up scale or down scale trips 215004 Source Range 217000 RClC 1 218000 ADS 223002 PCIS/Nuclear Steam SUDDIVShutoff Level Control 1 261000 SGTS -

7 of34 NUREG-I021 Draft Revision 9

ES-401 (Cm.43.5 145.12145.13)

S 2.?.25 Knowledge of bases in TS for 3.7 limiting conditions of operations and safe limits [Related to AC Electrical Distrr ution 2.4.48 Ability to interpret control room 3.8 indications to verify the status and operationof system / and understand how operator action s and directives afect lant and s stem conditions. Related to EDGs]

(EFR: 43.5 145.12f flGroup Point Total:

ES-401 I System # I Name 201001 CRD Hydraulic 201002 RMCS 201003 Control Rod and Drive Mechanism 201004 RSCS 201005 RCIS 201006 RWM 202001 Recirculation 202002 RecirculationFlow Control 204000 RWCU 214000 RPIS 1 215001 Traversing In-core Probe 215002 RBM 216000 Nuclear Boiler Inst.

219000 RHWLPCI:

TorudPool Cooling Mode 223001 Primary CTMT and Aux.

226001 RHWLPCI: CTMT 230000 RHWLPCI:

Torus/PoolSpray Mode 233000 Fuel Pool Cooling/Cleanup 234000 Fuel Handling Equipment 239001 Main and Reheat Steam I 239003 MSlV Leakage Control 241000 Reactor/Turbine 1 245000 Main Turbine Gen. I Aux.

256000 Reactor Condensate 9of34 NUREG-1021,Draft Revision 9

ES-401 259001 Reactor Feedwater 268000 Radwaste 271000 Offgas 272000.Radiation Monitorina 286000 Fire Protection 7 6s A2. Abili to a) redictthe impactso the o owin on the FIRE PROTECTIO~ISYSTEM 3.0

and (b) based on those predictions, useproceduresto correct, control, or mitlgate the consequences of those abnonnal conditions or operations

I (CFR: 41.5 /45.6) 288000 Plant Ventilation A2.03 A.C. distribution failure:

Plant-Specific I

290002 Reactor Vessel 2.2.32 Knowledge ofthe effects 3.3 lnternals of alterations on core configuration.

(CFR:43.6)

I WA Category Point Totals: Group Point Total:

ES-401 11 of34 NUREG-1021, Draft Revision 9

ES-401 b

ES-401 ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-53 Facility: Date of Exam:

I I

~

Category WA # Topic RO 1

II SRC IR # IR 1.

Conduct of Operations

~~

2.1.14 Knowledge of system status criteria which 1 requires notificationof plant personnel (CFR43.5/45.12) 2.1.

2.1.

I 2.2.21 Knowledge ofpre and post maintenance 3.5 operability requirements.

2.

Equipment (CFR 43.2)

Control 2.2.26 Knowledge of refuelingadministrative 3.7 requirements (CFR 45-3/4513)

SS;Rv-I -3 3x

3. (CF 3.4/45.10) 5Lf?TED Radiation Control 2.3. I 2.3. I 13 of 34 NUREG-I021, Draft Revision 9

ES-401 Subtotal 2.4.36 KnowIedge of chemistry /health physics tasks during emergency operations.

(CFR: 43.5) 2.4.41 Knowledge of the emergency action level thresholds and classifications.

(CFR 43.5 / 45.11) 2.4.

2.4.

2.4.

2.4. I Subtotal

11 EM01 Emergency UAPE # I NameI Safety Function ination Outline ubons -Tier VGmup 1 (ROI SRO)

WA Topias)

F m ES-401-1 IR 2.1.33 Ability to recognize indicationsfor 3.4 system operating parameters which are entry-level conditions for technical s ecifications.

(8FR: 43.2 / 43.3 145.3) 295003 Partial or Complete Loss ofAC I 6 AAI. Ability to operate andlor monitor the 4.4 following as '

the ap 1 to PARTIAL OR COMPLETE LO& #AC. POWER :

(CFR: 41.7 145.6)

AA1.03 S stems necessaryto assure safe Y

plant shu down.......

NU.Ability to determine and/or interpret 3.5 the followin as the ap I 9, PARTIAL OR COMPLETE LOgS B4b.C. POWER *

(CFR: 41.10 /43.5/45.13)

AA2.02 Extentof partialor completelossof D.C. power.

295005 Main Turbine Generator AKl . Knowled e of the operational 4.0 Trip i 3 im lietions o&e fo&win concepts as they apply to MAIN TURBI~E GENERATOR TRIP :

(CFR:41.8 to41.10)

AKI .01 Pressure effects on reactor Dower. -

295006 SCRAM I 1 AK2 Knowled e of the interrelations 3.7 beGeen SC& and the following:

(CFR: 41.7 I45.8)

AK2.03 CRD hydraulic system...

295016 Control Room AK2. Knowled e of the interrelations 4.0 Abandonment I7 between CONTROL ROOM ABANDONMENT and the following:

(CFR 41.7 I45.8)

AK2.02 Local control stations: Plant-Specific ...

ES-401 AK1, Knowled e of the operational 3.9 im lications o A e 3.4 Folbwin conce ts as they a ply to PART~AY OR ~OMPLETE L O ~ SOF COMPONENT COOLINGWATER :

(CFR: 41.8 to 41.IO)

AK1.01 Effects on componentkystem operations.6 AK2. Knowled e of the interrelations between PARTIIL OR COMPLETE LOSS OF COMPONENT COOLINGWATERand the foiowin * ~

(CFR 41.77.45.8)

AK2.02 Plant operations.

AK3. Knowledge of the reasons for the 3.5 following responses as they a ply to PARTIAL OR COMPLETE LOSS OF INSTRUMENTAIR :

(CFR: 41.5 145.6)

AK3.02 Standby air compressor operation.

295021 Loss of Shutdown Cooling 2.1.22 Ability to determine Mode of 2.8 0 eration.

(8FR: 43.5 / 45.13)

[I4 -

295023 RefuelingAcc Cooling Mode / 8 EA2. Abililty to determine and/or interpret 3.9 the followin as j y to HIGH DRYWELL L%SSagFE (CFR: 41.10j43.5 /45.13)

EA2.04 Suppression chamber pressure:

Plant-Specific.

295025 High Reactor Pressure/ 3 EK3. Knowledge of the reasons for the 4.2' following responses as the apply to HIGH REACTOR PRESS~RE (CFR: 41.5 / 45.6)

EK3.06 Alternate rod insertion: Plant-Specific..

5 of 34 NUREG-I 021, Draft Revision 9

ES-40 I 295026 Suppression Pool High Water Temp. I5 1 EK3. Knowledae of the reasons for the EK3.02 Suppression pool cooling..

2.4.21 Knowledge of the parameters and POOL logic'used to assess the status of safety functions includin

  • 3.91 3.7
1. Reacfvity.control
2. Core coolin and heat remova!
3. Reactor c d a n t s stem integrity
4. Containment conJtions
5. Radioactivi release control.

(CFR 43.5 14 .12)

I 295027 High Containment Temperature I 5 3.9 EA1.02 Drywell ventilation system..

3.5 1 EK2.08 SRV discharge submergence..

EA1. Ability to operate and/or monitor the 4.4*

following as they apply to REACTOR LOW WATER LEVEL :

(CFR:41.7 I45.6)

EAI.03 Low pressure Core Spray System.

295037 SCRAM Condition Present .

EKI Knowled e of the operational irn lications odhe 4.1*

and Power Above APRM Downscale or Unknown I1 folLwin conce ts as the a ~y to s c d COND~ION REACTOR POWER ABOVE APRM PRE~ENF AND DOWNSCALEOR UNKNOWN :

(CFR: 41.8 to41.10)

EK1.02 Reactor water level effects on reactor power.

EK?. Knowled e of the operational 2.5 im licptions oaile folkwin conce ts as the apply to HIGH OFF-SITE RELEASE (CFR 41.8 to 41.10)

R A ~ :E EKI .Ol Biological effects of radioisotope ingestion.

ES-401 7 of 34 NUREG-1021, Draft Revision 9

ES-401 ES-401 BWR fR8 nination Outline I IK O

F ES-401-I Emeraencv and Abnormal Plan -

dutrons Tier I/Group 2 (RO/ SRO)

UAPE # / Name / Safety Function KIA Topic@)

295002 Loss of Main Condenser Vac I 3 295007 High Reactor Pressure I 3 295008 High Reactor Water Level 12 295009 Low Reactor Water Level 12 295010 High Drywell Pressure / 5 295011 High ContainmentTemp I 5

295012 High D AA2. Abilitv to determine andlor intemret 3.9 Temperature I the followifig as ap I t o HIGH DRYWELL

%dPERA?LYRE -

(CFR: 41.I 0 / 43.5 / 45.1 3)

AA2.02 Drywell pressure.

7 295013 High Suppression Pool AK3. Knowledge of the reasons for the 3.6 Temp. I5 following responses as the a 1 to HIGH SUPPRESSION POOL +Ed82 (CFR: 41.5 145.6)

RATURE :

AK3.02 Limiting heat additions.

295014 Inadvertent Reactivity Addition I 1 -

295015 IncompleteSCRAM I 1 295017 High Off-site Release A B . Knowled e of the interrelations 3.3 Rate I 9 between HIGaOFF-SITE RELEASE RATE and the following:

(CFR: 41.7 145.8)

AK2.03 Off-gas system.

295020 Inadvertent Cont.

Isolation / 5 & 7 295022 Loss of CRD PumDs I 1 295029 High Suppression Pool wtr Lvl / 5

ES-401 295032 High Secondary I EA1.Ability to operate andor monitor the 3.7 Containment Area Temperature I 5

EA1.03 Secondary containment ventilation..

295033 High Seconday EK3. Knowledge of the reasons for 3.3 Containment Area Radiation Emergency Depressurizationas it applies Levels I 9 to (CFR: 41.7 I45.8)

EK3.01 High Secondary Containment Radiation Levels.

295034 Secondary Containment EK1. Knowled e of the operational 3.8 Ventilation High RadiationI9 im tications o?ae folkwing concepts as the s E c o ND A R Y C o N T X i VENTILATION HIGH RADIATION :

N W at N?

(CFR: 41.8 to41.10)

EKI .01 Personnel protection.

295035 Secondary Containment High Differential Pressure / 5 I following EK3. Knowledae of the reasons for the respcjnses 2.8 as they a ly to SECONDARY CONTAINMEJ?

PRESSURE -

HIGH DIFFERENTIAL (CFR: 41.5 145.6)

I EK3.01 Blow-out panel operation: Plant-Specific..

I 295036 Seconda Containment High Sump/Area r;i((aterLevel 1 5 I 500000 High CTMT Hydrogen Conc. I5 1I WA Category Point Totals: Group Point Total:

9 of 34 NUREG-1021, Draft Revision 9

ES-401 ES-401 BWR m x a m i n a in Outline FOITYI ES-401-I ES-401 Plant S terns Tier ZGrc p I {RO 1 SRO) 1-System # / Name WA Topic(s) IR #

203000 RHWLPCI:

Injection Mode 205000 Shutdown K1. Knowledge of the physical 3.11 2 Cooling connections and/or causeeffect 3.4 relationships between SHUTDOWN COOLING SYSTEM(RHR SHUTDOWN COOLING MODE) and the followin :

(CFR: 41270 41.9 / 45.7 to 45.8)

K?.05 Component cooling water systems.

A2.Abili to redict the impacton the SH"TD8W8 cO-OiiNGTSYSTEM RHR SHUTDOWN COOLING t h D E )

0 eration from:

(8FR: 41.5 145.5)

A2.09 Low Reactor water level 206000 HPCl ~~~~ ~ ~

207000 Isolation KI. Knowledae of the DhYSiCal 3.8/ 2 (Emergency) Condenser connections Znd/or caus'eeffect 3.5 relationshi s between ISOLATION EMERG&CWCONOENSER and the followin : '

(CFR:41.2?041.9145.7 to 45.8)

KI.01 Reactor vessel: BWR-2,3 .

K2. Knowledge of electrical power su lies to the following:

(C%: 41.7)

K2.02 Initiationlogic: BWR-2,3 11 of34 NUREG-I021, Drafl Revision 9

209001 LPCS KI. Knowledgeof the physical 3.71 connections andlor causeeffect 3.3 relationshi s between LOW PRESSURE~ORESPRAYSYSTEM and the followin -

(CFR: 41.2 to 48:9 I45.7 to 45.8)

K1.05 Automatic depressurization system ..

use piobdures to correcf, control, or mitigate the consequences of those abnormal conditions or operations:

(CFR: 41.5 I 45.6)

A2.05 Core spray line break .

7 209002 HPCS 7 21 1000 SLC K2. Knowledgeof electrical power 3.11 su plies to the 2.6*

folkwing:

(CFR: 41.7)

K2.02 Explosivevalves K3. Knowledge of the effect that a loss or malfunction of the STANDBY LIQUID CONTROL SYSTEM will have on followin (CFR: 41.7 I4 5 J j K3.03 Core plate differential pressure indication .

212000 RPS K3. Knowledge of the effect that a 3.7 loss or malfunctionof the REACTOR PROTECTION SYSTEM will have on following:

(CFR: 41.7 / 45.4)

K3.05 RPS logic channels 215003 IRM K4. Knowled e of INTERMEDIATE 3.7 RANGE MO~ITOR(IRM) SYSTEM desi n feature@) and/or interlocks whica provide for the following:

(CFR: 41.7)

K4.01 Rod withdrawal blocks.

7 215004 Source Range K5. Knowledgeof the operational 2.8 Monitor imolications of the followin conce ts as the ap I to S O U R C ~R A N ~ MONIT&R E (!%M)

SYSTEM :

(CFR: 41.5 I45.3)

K5.03 Changing detector position L_

ES-40l K6. Knowledge of the effect that a 3.7 loss or malfunctionof the following will have on the AVERAGE POWER RANGE MONlTORlLOCAL POWER RANGE MONITOR SYSTEM :

(CFR: 41.7 1 45.7)

K6.01 RPS .

A2. Abili to (a) predict the impacts 4.2

  • Y

. owin on of the fol t h e A S T O M A T I C DEPRESSURIZATION SYSTEM and (b) based on those rediction;,

use Drocedures to corre2 control, or mitigatethe consequences of those abnormal conditions or o erations:

&FR: 41.5 / 45.6)

A2.06 ADS initiation signals present A4. Ability to manually operate 3.6 and/or monitor in the control room:

(CFR: 41.7 / 45.5 to 45.8)

A4.03 Reset system isolations 1

K5. Knowledge of the operational 3.3/

im lications of the 3.1 folkwin mnce ts as they apply to RELlEF%AF&

VALVES -

(CFR: 41.*5 / 45.3)

K5.04 . Tail pipe temperature monitoring .

2.4.6 Knowledge symptom based EOP mitigation strate ies.

(CFR:41.10/43.5/4%.13)

A3. Abilitv to monitor automatic operatioris of I 3.0 the REACTOR WATER LEVEL CONTROL SYSTEM including:

(CFR:41.5 / 45.6)

A3.06 Reactor. Water level setpoint setdown following SCRAM .

13 of 34 NUREG-I021,Draft Revision 9

ES-401 A I . Abilitv to Dredict and/or monitor 3.0 changes i n '

arametersassociated with o eratin ke STANDBY GAS TREA%IEN?

SYSTEM controls including:

(CFR: 41.5 145.5)

A I .04 Secondary containment differential pressure .

A4. Ability to manually operate 3.2 andlor monitor in the control room:

(CFR: 41.7 I45.5 to 45.8)

A4.03 Local operation of breakers 21-30 Ability to locate and operate 3.9 components I including local controls.

(CFR: 41.7 145.7)

A.4 Ability to manually operate 2.7 andlor monitor in the control room:

(CFR 41.7145.5)

A4.03 Battery discharge rate: Plant-Specific

. - - Knowledoe K? ~

.- - . of the ohvsical 2.91 connections Zndlor cabs6 effect 3.8 relationshi s between EMERGENCY GENEfWf0RS (D IESEUJET) and the followin :

(CFR: 41.290 41.9 145.7 to 45.8)

KI .03 Fire protection system .

K6. Knowledae of the effect that a ibis -ormalfijjction of the followin will have on the EMERGENCf GENERATORS DIESEUJET) :

(CFR: 41.71454)

K6.01 Starting air K4. Knowled e of (INSTRUMENT 3.0 I(

AIR SYSTEd design featuye(s r'

and oKinterloc s which provide or the following:

(CFR: 41.7)

K4.02 Cross-overto other air systems

I I I I 1I 400000 Component K3. Knowledge of the effectthat a 2.91 2 Cooling Water loss or malfunction of 2.5 the CCWS will have on the following:

(CFR: 41.7 I 45.6)

K3.01 Loads cooled by CCWS 2.2.25 Knowled e of bases. in I I technical s ecficayions for limiting wnditions r!i operations and safety WA Category Point Totals:

I limits.

(CFR. 43.2)

Group Point Total: 2 6

I d

15 of 34 NUREG-1021, Draft Revision 9

ES-401 ES-401 B' 1 Outline Form ES-401-1 P nt SI I i ( R 0 I SRO)

I I

I WA Topic(s)

I IR K6. Knowled e of the effect that 3.0 a loss or qattnction of the followin will have on the CONTROL WOD DRIVE HYDRAULIC S stem :

(CFR 41.7 1427)

I K6.02 Condensatestorage tanks I I I 1 design interlocks which provide for the I feature(s1 and/or I followin *

(CFR4Y.7)

K4. Knowled e of ROD WORTH 3.2 MINIMIZER ~ Y S T E M(RWM)

&!%%IC) design feature(s) and/or interlocks which P rovtde for the following:

CFR 41.7)

I K4.06 Correction of out of sequence rod positions: P-Spec K3. Knowled e of the effect that 3.9 a loss or malknction of the RECIRCULATION SYSTEM will have on following:

(CFR: 41.7 / 45.4)

K3.03 Reactor power.

I K3. Knowled e of the effect that 4.0 a loss or rnal%nction of the RECIRCULATION FLOW CONTROLSYSTEMwill have on followin (CFR: 4y.7 I45.4)

K3.02 Reactor power.

I K1. Knowledge of the physical 3.1 connections and/or causeeffect relationshi s between REACTOR WATERCrEANUP SYSTEMand the followin :

(CFR: 41.270 41.9 / 45.7 to 45.8)

I K1.Ol Reactor vessel . I

ES-401 K5. Knowledge of the 2.6 I operational implications of the followin concepts as the a I to RHFSRPCI: CONTAI~MR~I SPRAY SYSTEM MODE :

(CFR:41.5 / 45.3)

K5.06 Vacuum breakeroperation AI. Ability to predict andlor 3.5 1 monitor changes in parameters associated with o eratin the MAIN AND

&HEAS STEAM SYSTEM controls includin :

(CFR: 41.5 I45.8)

A I -09 Main Steam Flow .

R A3. Ability to monitor automatic 3.0 I operations of the REACTOR CON DE NSAT E SYSTEM includin (CFR: 4!:7

- /45.7)

A3.06 Hotwell level 17 of 34 NUREG-1021, Draft Revision 9

ES-401 1 268000 Radwaste A3. Ability to monitor automatic o erations of the OFFGAS S!'STEM including:

(CFR: 41.7 1 45.7)

A3.03 System temperatures tV A2. Ability to (d redict the impacts of the o lowin on the RADIATION MOWITORING SYSTEM ;and (b) based on those predictions,use procedures to correct, control, or mitigate the consequences of those abnormal conditions or o erations:

(CFR: 41.5 I4Z.6)

A2.07- H dro en injection operation: Aant-%pecific .

286000 Fire Protection 288000 Plant Ventilation 290001 Secondary CTMT 290003 Control Room HVAC 290002 ReactorVessel lntemals K/A Category Point Totals: Group Point Total:

ES-401 Date of Exam:

I KIA # Topic 2.1.2 knowledge of operator responsibilities during all modes of operation.

(CFR: 41.IO / 45.13) 2.1.12 Ability to apply TS for a system.

(CFR: 43.1 I45.13) plant procedures during different modes of operation.

(CFR: 41-10/ 43.5 / 45.12)

I Subtotal controls associated with plant equipment that could affect reactivity.

(CFR:45.1) 2.2.11 Knowledge of the process for controlling temporary changes.

(CFR: 41.1 0 / 43.3 / 45.13) 2.2.26 Knowledge of refueling administration requirements (43345.13)

Subtotal 2.3.2 Knowledge of facility ALARA program.

(CFR: 41.12/43.4/45.9/45.10) 2.3.9 Knowledge of the process for performing a containmentpurge.

(CFR: 43.4 I45.10)

I 9 of 34 NUREG-1021, Draft Revision 9

ES-401 QUESTION #2 The Reactor is at 10% power when an unisolable leak occurs in the Reactor Water Cleanup System. Radiation levels at the Cleanup Pumps and the S/D HX area are > 1 WHR.

Answer the following as required by EMG- 3200.11 "Secondary Containment Control" a) Which of the following action(s) idare required for this, specific condition?

b) How does this action mitigate the effects of the unisolable leak?.

a) b)

A. Shutdown the Reactor Shutting down reduces the source of the radiation.

B. Emergency Depressurization Depressurizationreduces the driving head for flow from the leak.

C. Emergency Depressurization Depressurizationwill result in cooler RCS fluid, thus keeping radioactive gases in solution.

D. Shutdown the Reactor Shutting down reduces the driving head for flow from the leak.

ANSWER: 6 U(PLANATI0N:

Emergency Depressurization is required since two areas in Table 12 have exceeded their max safe radiation levels. Shutting down does nothing to reduce the flow from the leak. The main effect from ED is to reduce the driving head for the leak.

TECHNICAL REFERENCE(S): EOP Bases (Attach if not previously provided)

Proposed references t o be provided t o applicants during examhation: EOPs (without entrv)

Learning Objective: (As available)

Examination Outline Cross-reference: Level. RO SRO Tier # -1 -

Group # -2 KiA # 295033/EK3.0?-

Importance Rating 3.3 -

KIA Topic

Description:

Knowledge of the reasons for Emergency Depressurization as it applies to High Secondary Containment Radiation Levels Question Source: Bank #

Modified Bank # X (Note changes or attached parent}

New Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis I O CFR Part 55 Content: 55.41 X 55.43 Comments: Changed KIA based on oversampiing of SGTS. Modified 1996 Quad Cities.

QUESTION #4 In procedure 205.0, Reactor Refueling, the operator is directed to IMMEDIATELY EVACUATE the area if refuel floor radiation levels begin to increase following a drop of a fuel assembly from the refueling equipment.

What is the basis for evacuating?

A. The direct radiation from the dropped fuel bundle could cause unplanned radiation exposure.

B. The dropped fuel bundle may breach the cavity seal and drain the cavity.

C. The dropped bundle may release radioactive gasses that could will cause unplanned radiation exposure.

D. The dropped bundle may create a criticality event in the reactor.

ANSWER: C U(PLANATI0N:

This is the classic fuel handling accident in which the fission product gasses contained within the fuel cladding are released and bubble up through the water. Personnel remaining over the cavity can be exposed to doses approximating 10CFRI 00 limits. The approximately 25 feet of water provide adequate shielding from direct radiation and allow time for action to restore level.

Criticality is avoided by refueling interlocks and refueling patterns, though answer D could be considered a plausible distractor for someone unfamiliar with refueling interlocks and core design.

TECHNICAL REFRENCE(S): RG 1.25: Procedure 205.0 DU 5.0 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # -1 -

Group # -2 K/A # 295034/EKl . O T Importance Rating 3.8 -

WA Topic

Description:

Knowledge of the operational implications of Personnel Protection as it applies to Secondary Containment High Radiation Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

QUESTION#23 Given the following plant conditions:

0 Reactor is at 100% power 0 AOG is in service 0 Main Steam Line Radiation Monitors all at approximately 550 mrhr 0 Stack Effluent HI alarm Reactor Bldg Vent Radiation at 8 mrhr 0 RCS activity at 90% of TS limit 0 6"IC isolated for maintenance 0 Significant/visible packing leak from "A" IC outboard steam isolation valve a NO leaks in the "A" IC tube bundle What action(s) would result in having the greatest reduction in the thyroid damage for the public?

A. Close "A" IC outboard steam isolation valve B. Reduce reactor power until stack effluent HI alarm clears C. Start SGTS and shutdown Reactor Building HVAC D. Close "A" IC vent valve ANSWER: C EXPLANAT10N:

Starting SGTS is the only action that will remove radioactive iodine being released from the steam leak. The AOG will remove all iodine from the off gas regardless of reactor power so reducing power will not result in a reduction in iodine.

TECHNICAL REFERENCE(S): (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # - i -

Group # - 1 KIA # 295038/EKlx Importance Rating 2.5 -

WA Topic

Description:

Knowledge of the operational implications of the biological effects of radioactive ingestion as it applies to Off Gas Release rate.

Question Source: Bank # -

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge -

Comprehensive or Analysis X

QUESTION #31 A reactor startup is in progress with the following conditions:

The previous shift had increased power from 60% to 80%.

0 Your shift just increased power from 80% to 90% power when all operable Main Steam Line Rad Monitors are verified to read greater than 550 mr/hr but less than 800 rnrhr.

Within one minute, area radiation monitors in the vicinity of the Moisture Separators experience at least a doubling of their readings.

The Off Gas monitors are observed to be unaffected.

What actions are required to mitigate the consequences of these abnormal conditions?

A. SCRAM the reactor in accordance with ABN-1, Reactor Scram and CLOSE MSlVs and IC Vents.

B. Reduce Hydrogen Injection flow to between 5 and 6 scfm and monitor the effect on Main Steam Line Activity for the next 10 minutes.

C. Direct Chemistry to sample Off Gas (Technical Specification 4.6.E) and request guidance from Reactor Engineering.

D. Start plant shutdown in accordance with Procedure 203, Plant Shutdown ANSWER: B EXPLANATION:

Since the radiation increase is not observed in the off Gas System it can be concluded that the radiation is from an isotope with a very short half life. Hydrogen Injection produces, almost exclusively, N16 with a half life of approximately 7 seconds. Per "Plant Startup" procedure Hydrogen Injection is increased in 10% power increments and ,given that there has been no increase in Off Gas activity it is reasonable to assume the increase is attributed to hydrogen Injection. The reference procedure has the operator first reduce Hydrogen Injection before taking other action. If the MSL radiation decreases in response to reducing Hydrogen Injection then only monitoring of Off Gas is required. The other actions are specified if the HI MSL condition is attributed to fuel failure.

TECHNICAL REFERENCE(S): ABN-26 Rev 0 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: ABN-26 Learning Objective: (As available}

ExaminationOutline Cross-reference: Level RO SRO Tier # -2 -

Group # -2 WA # 272000/A2.0r Importance Rating 2.6 -

QUESTION # 64 You are about to sign on to an RWP to operate a valve in the Radwaste Building and notice there is no requirement to wear a respirator. You recall from your tour this morning that the area you will enter is posted as a High Airborne Activity Area. You ask the HP tech about this and are told it is consistent with ALARA to NOT wear a respirator.

Why would NOT wearing a respirator in a High Airborne Activity Area be consistent with ALARA?

A. The respirator is only effective on particulates so there will be NO difference in TEDE.

B. The respirator will limit your vision and may be a safety hazard in "tight" spaces.

C. Wearing a respirator may increase your stay time and actually increase TEDE.

D. The Airborne Activity is short lived and the respirator will have NO impact on ALARA.

ANSWER: C EXPIANATION:

TEDE includes internal and external dose. If the HP calculates TEDE to be lower without a respirator then that is consistent with ALARA.

TECHNICAL REFERENCE(S): (Attach if not previously provided)

Proposed references to be provided to applicants during examination:

Learning Objective: (As available)

Examination Outline Cross-reference: Level RO SRO Tier # - 3 -

Group # - 3 -

WA # 2.3.2 Importance Rating 2.5 -

WA Topic

Description:

Knowledge of the facility AMRA program.

Question Source: Bank #

Modified Bank # (Note changes or attached parent)

New X Question Cognitive Level: Memory or Fundamental Knowledge X Comprehensive or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 x Comments: This goes beyond what is taught in GET.

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