ML041890331

Draft Outlines

ML041890331
Person / Time
Site:	Oyster Creek
Issue date:	04/19/2004
From:	Hackenberg J AmerGen Energy Co
To:	Conte R NRC/RGN-I/DRS/OSB
Conte R
References
50-219/04-301 50-219/04-301
Download: ML041890331 (45)

Text

J. Caruso 1/20/04 EXAM OUTLINE COMMENTS Comments provided per telecom to licensee Greg Young on 1/20/04.

Admin All but one JPM item is new (A.3 SRO JPM - bank item, last NRC exam). Discussed actual content of each JPM task 0 RO Admin JPM A.3 - JPM involves a hand held. Comment GET level can we make more operationaVlicense oriented. Made several suggestions. Licensee will consider developing a new JPM to evaluate stay times to accomplish a task in-plant based on current exposure and specified survey map radiation levels.

0 SRO Admin JPM A.4 - Although not specifically listed on outline, task includes notification and classification. This will be done as a follow-up JPM to the dynamic simulator portion of the exam.

JPMs Only one generic outline submitted need to have a least one separate outline for the SROUs or at least provide a subset listing of those JPMs that will on the SROU exam.

Reminder 80% limit on bank JPMs.

JPM #a & #b discussed proposed JPMs and proposed scenario events to ensure that JPMs do not overlap with scenario events and determined that overlap does not appear to exist.

JPM #g, cautioned Shutdown SGTS may be testing the same system knowledge/ability that is being examined on scenario #2 when SBGT fails to start (Le., operating the same equipment - may not sufficiently diverse)

JPM outline for both the RO and SROl do not meet NUREG 1021 in that for JPMs performed in simulator Le., the licensee has proposed that two safety functions are repeated - there are two safety function #9s and two safety function Ws. Also in-plant JPM#b should be safety function #I vice #2.

In-plant JPM #a, Line-up fire water to Core Spray not sufficiently modified only changing task to a different train. Licensee will replace Scenario Outlines 0 Licensee verified all 4 scenarios are new.

0 Critical tasks are not identified so couldnt comment but discussed expectations per Appendix D.

0 Scenarios #2,3, and 4 may not meet minimum number of total malfunctions 5-8per ES 301-4.

0 Scenario # ,Ievent 5 only RO credit.

0 Scenario #2,event 6 only RO credit, event #5 no operator actions required besides reducing power counts one time as reactivity.

0 Scenario #3, event 2 is a component failure not instrument.

0 Scenario #4, event 7, both RO and BOP have separate actions can count for both.

Scenario #4 is 10% power - told licensee we are considering less than 5% as our standard for low power. Also this should be one of the scenarios scheduled for use and not the spare.

To: John Caruso/Gil Johnson From: Greg Young

Subject:

Submittal of Operating Exam outlines John/Gil, Enclosed you will find the outlines for the Admin and PlantISimulator JPMs along with 4 exam scenario outlines for your review and comment. Please call me if you have any questions or comments.

Enclosures:

Form ES-201-2 Form ES-301-5 Form ES-301-1 Form ES-301-2 Form ES-D-1 (4)

Greg Young 609-971-4196

From: <gpyoung@amergenenergy.com>

To: Gilbert Johnson <GXJ@nrc.gov>

Date: 1/15/04 1:57PM

Subject:

Re: exam Here are the outlines for scenarios (4) and the admin and sim/plantjpms.

The crew make-up for the exam, at this time, is as follows:

n Crew 1 1 RO, 1 SRO-I, and 1 SRO-U I at- q Crew 2 1 RO, 1 SRO-I, and 1 SRO-U

\ -4 CI Crew 3 IRO, 2SRO-I (See attached file: ES-D-I SR04.doc)(See attached file: ES-301-2 NRC.doc)

(See attached file: ES-D-1 SROl .doc)(See attached file: ES-D-1 SR02.doc)

(See attached file: ES-D-I SR03.doc)(Seeattached file: ES-301-1 NRC.doc)

This E-mail and any of its attachments may contain Exelon Corp.

proprietary information, which is privileged, confidential, or subject to copyright belonging to the Exelon Corp. Family of Companies.

This E-mail is intended solely for the use of the individual or entity to which it is addressed. If you are not the intended recipient of this E-mail, you are hereby notified that any dissemination, distribution, copying, or action taken in relation to the contents of and attachments to this E-mail is strictly prohibited and may be unlawful. If you have receivedthis E-mail in error, please notify the sender immediately and permanently delete the original and any copy of this E-mail and any printout. Thank You.

ES-301 Administrative Topics Outline Form ES-301-1 Administrative Topic Describe activity to be performed Conduct of Operations Calculate DW bulk temperature (JPM, new, SP-26)

[RO & SRO]

RO - Calculate Unidentified Leakage (JPM, new, 200.0F)

SRO - Approve Unidentified Leakrate Calculation (JPM, new)

Equipment Control Surveillance Test prerequisites for Core Spray - Alternate Path (JPM, new, 200.OG)

[RO & SRO]

RO - Use Hand-held Frisker to perform whole body frisk (JPM, Radiation Control new)

SRO - Approve Radioactive Discharge Permit - Alternate Path (JPM, 200.06, last NRC) 1l Emergency Plan Make an Emergency Classification (JPM)

Fire/Natural Hazard based (new)

[for SRO only]

are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.

- - 1 .

ES-301 Transient and Event Checklist Form ES-301-5 OPERATING TEST NO.:

NRC Reviewer: --___

NUREG-1021, Draft Revision 9 26 of 27

Scenario Outline ES-D-1 Simulation Facility Oyster Creek Scenario No. SRO #1 Op Test No. 2004NRC Examiners Operators CRS PRO i

URO Scenario The scenario begins with the reactor at 100% power with the A CRD Pump out of service. The crew Summary will begin by removing the EPR from service. A reference leg leak will develop in a RPV Level instrument. The crew will take manualcontrol of RPV level and transfer to the alternate signal. A loss of power to VMCC 1A2 will result in the crew restoring RPS and resettingthe half scram. The C Feedwater Pump trips requiring the crew to reduce power to maintain reactor level. The only available CRD pump trips, which will require the crew to scram the reactor. A steam leak develops from the A Isolation Condenser and the crew will be unable to isolate the leak. This will require entry into Secondary Containment Control and eventually result in the need to Emergency Depressurize. One EMRV will not open.

Event Malfunction Event Event No. No. Type* Description SRO 1 N BOP Remove EPR from service for maintenance MAL-NSSO11C 1 I RO 1 Level instrument reference leg leak develops, take manual control of RPV level, swap instrument signals, return to auto control 3 MAL-EDS004A C Loss of Power to VMCC 1A2, restore RPS (Tech Spec)

BOP SRO 4 MAL-CFWOOGC C C Feedwater Pump Trips BOP SRO 5 R RO 43m-SRO 6 BKR CRD001 C RO Only available CRD pumktrips - Results in Plant Scram vLv-Icsoo5 Steam leak from the A Isolation Condenser, isolation valves fail to VLV-ICs006 MAL-ICS003A-8 I MAL-NSS023A IC I One of Five EMRVs fails to open

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

Scenario Outline ES-D-1 Simulation Facility Oyster Creek Scenario No. SRO #2 Op Test No. 2004NRC Examiners Operators CRS PRO URO Scenario The scenario begins with the reactor at 99% power with the A CRD pump out of service. The crew Summary will begin by placing an alternate Reactor Building HVAC fan in service and secure the running fan.

The RBHVAC ventilation radiation monitor will fail upscale, causing RBHVAC to trip, but the SGTS will fail to start. The crew will start the SGTS manually. A control rod drifts out and it will be restored to its programmed position. The running service Water pump trips requiring the standby pump to be started. The rod drift will cause a small fuel failure. Power will be reducedto lower radiation levels. A leak in the Torus will require the Reactor to be scrammed and eventually this will lead to Emergency Depressurization. Five rods will fail to insert on the scram.

I( Initial Condition 99% power 1 Turnover: See Attached Shift Turnover Sheet 11 Event I Malfunction I Event Event No. Type* Description SRO N Swap Reactor Building HVAC Supply fans BOP SRO 2 MAL-RMS005M I RBHVAC Rad Ventilation monitor Fails Upscale, RBHVAC trips, SGTS MAL-SCN005 BOP fails to start. (Tech Spec)

SRO 3 MAL- C RO Control Rod Drifts Out CRD005-2239 SRO 4 MAL-SWS001B C BOP Running Service Water pump trips Small Fuel Failure MAL RXSl .001/120s Power Reduction to Lower adiat n Lev Is.

fl&uL./ &&.A A!*

(1 1 I

L 7 8000,900s MAL-CRD022 lM RO BOP SRO Torus Water Leak

-1039,-4211, C RO Five Rods Fail to Insert on the Scram (

-2635, -1 423,

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

Scenario Outline ES-D-1 Simulation Facility Oyster Creek Scenario No. SRO #3 Op Test No. 2004NRC Examiners Operators CRS PRO URO Scenario The scenario begins with the reactor at 95% power with the A Isolation Condenser out of service.

Summary The crew will begin by swapping the TBCCW Pumps. The in service B CRD pump will trip requiring the crew to start the other pump and evaluate Tech Specs. The 1-2 Circulating Water Pump trips requiring the crew to reduce power to maintain condenser vacuum. A Recirculation loop leak will require the crew to manually scram the reactor due to increasing Drywell pressure. An AlWS will occur, but Alternate Rod Injection will be successful. Four Containment Isolationvalves will remain open and will be manually closed by the crew. Containment conditions will result in initiation of Drywell Sprays to control Drywell pressure between 4 - 12 psig.

Event Malfunction Event Event No. No. Type* Description SRO 1 N BOP Swap Turbine Building CCW pumps SRO BKR-CRDO01 Db RO Running CRD pump trips (Tech Specs) 3 MAL-CWS001B C BOP I Circulating Water pump trips 4 R RO Power Reduction to maintain condenser vacuum BOP SRO 5 MAL-NSS004D M RO Recirculation loop leak - Results in Plant Scram BOP SRO 6 MAL-RPS005 I RO ATWS - ARI is successful MAL-RPSOOG SRO 7 MAL-RPS007A C BOP Containment Isolation is incomplete, sump isolationvalves remain open, manually closed

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

L Simulation Facility Oyster Creek Scenario Outline Scenario No.

&;&/

SRO #4 Y

Op Test No. 2004NRC ES-D-1 I( Examiners Operators CRS PRO URO Scenario The scenario begins with the reactor startup in pr 10% power with mode switch in RUN.

Level Instrument RE02A Fails and idle. Core Spray will be m upscale requiring it to be Initial Conditi 2 1 N BOP SRO I I

/

Swap Servi e water pumps 3 MAL-NSS007A I BOP Reactor el Instrument Fails Downscale, Core Spray starts but EDG MAL-DGN003B I

4 MAL-NIS001A I ctor fails upscale SRO 5 MAL-RBC001A C BOP Running RBCCW pump trips 6 MAL-NSSOl7A M RO Steam leak develops in the Drywell BOP SRO 7 MAL-EDSOOSF C RO Loss of DC-F causes multiple recirc pump trip and Isolation BOP Condensers to initiate - results in'plant scram and ICs must be secured /

/

/

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

ES-401 ES-401 BWR SlW-Examination Outline Form ES-401-1 I Facility: Date of Exam: 1.

I.

Tier 1.

Emergency 84 Abnormal Plant Evolutions 2.

Plant Systems

3. Generic Knowledge and Abilities Categories

ES-401 Note: 1. Ensure that at least two topics from every WA category are sampled within each tier of the RO outline (i.e., the Tier Totals in each WA category shall not be less than two). Refer to Section D.I.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 *I from that specified in the table based on NRC revisions. The finalRO 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-er tiwee WA topics from a given system or evolution unless they relate to plant-specific priorities.

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

5. The shaded areas are not applicable to the categoryltier.

6.* The generic (G) WAS in Tiers Iand 2 shall be selected from Section 2 of the WA Catalog, but the topics must be relevant to the applicable evolution or system. The SRO K/As must also be linked to 10 CFR 55.43 or an SRO-level learning objective.

7. On the following pages, enter the WA numbers, a brief description of each topic, the topics importance ratings (IR) for the -applicable license level, and the point totals for each system and c

. . y

. . . Enter the group and tier totals for0 eachrcategory a t e g .

in the v

table above; summarize all the SRO-only knowledge and non-A2 ability categories in the columns labeled K and A. Use duplicate pages for RO and SRO-only exams.

h. For Tier 3, enter the WA numbers, descriptions, importance ratings, and point totals on Form ES-401-3.

i. Refer to ES-401,Attachment 2, for guidance regarding the elimination of inappropriate WA statements.

3 of 34 NUREG-1021, Draft Revision 9

ES-401 ES-401 INF! WExamination Outline Form ES-401-1 Emergency id Ab rmal Plant Evolutions -Tier l/Group 1 (RO I SRO) -

1 WAPE # / Name I Safety Function WA Topic@)

IR 295001 Partial or Com lete Loss of Forced 2.1.33 Ability to recognize indications for 3.4 PI core FIOW Circulation & 4 system operating pa.rameters which are entry-level conditions for technical s ecifications.

(8FR: 43.2 / 43.3 / 45.3) 295003 Partial or Complete Loss of AC I 6 AA1. Ability to operate and/or monitor the 4.4 following as the ap I to PARTIAL OR COMPLETE Logs 8qA.C. POWER :

(CFR: 41.7 / 45.6)

AA1.03 S stems necessaryto assure safe Y

plant shu down.......

295004 Partial or Total Loss of DC AA2. Ability to determine and/or interpret 3.5 hr/6 the followin as the ap I ?o PARTIAL OR COMPLETE LOgS 8qD.C. POWER (CFR: 41.10 143.5 /45.13)

AA2.02 Extentof partial or complete loss of D.C. power.

295005 Main Turbine Generator AK1. Knowled e of the operational 4.0 Trip / 3 im lications o?the folLwin concepts as they apply to MAIN TURBI~E GENERATOR TRIP :

(CFR: 41.8 to41.10)

AKI . O l Pressureeffects on reactor power. -

295006 SCRAM / 1 AK2. Knowled e of the interrelations 3.7 between SCRWM and the following:

(CFR: 41.7 / 45.8)

AK2.03 CRD hydraulic system...

295016 Control Room AK2. Knowled e of the interrelations 4.0 Abandonment I 7 between CONTROL ROOM ABANDONMENT and the following:

(CFR: 41.7 145.8)

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

ES-401 AK1. Knowled e of the operational 3.51 imDlications ohhe 3.4 followin conce ts as they a ply to PARTIATOR ~OMPLETE L O ~ SOF COMPONENT COOLINGWATER :

(CFR: 41.8 to 41.10)

AKI .01 Effects on componentlsystem operations.6 AK2. Knowled e of the interrelations between PARTIIL OR COMPLETE LOSS OF COMPONENT COOLINGWATER and the followin .

(CFR: 41.7745.8)

AK2.02 Plant operations. -

AK3. Knowledae of the reasonsfor the 3.5 AK3.02 Standby air compressor operation.

295021 Loss of Shutdown Cooling 2.1.22 Abilitv to determine Mode of 2.8 14 0 eration.

(d3FR: 43.5 145.13) 295023 Refueling Acc Cooling Mode 18 -

295024 High Drywell Pressure / 5 EA2. Ability to determine and/or interpret 3.9 the followin as the Qy to HIGH DRYWELL PRZSS~RE-(CFR: 41.10'143.5 145.13)

EA2.04 Suppression chamber pressure:

Plant-Specific.

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-1021, Draft Revision 9

ES-401 EK3. Knowledge of the reasons for the 3.91 following res 6nses 3.7 as they a HIGH WA#&

P to SUPPRESSION POOL TEMPERATURE:

(CFR: 41.5 / 45.6)

EK3.02 Suppression pool cooling..

2.4.21 Knowledge of the parameters and logic used to assess the status of safety functions Yk!kK,ity. control

2. Core coolin and heat removal 9

3. Reactorcoo ant s stem integrity

4. Containment condlitions

5. Radioactivi release control.

(CFR: 43.5 14 .12) 295027 High Containment Temperature 1 5 -

295028 High Drywell Temperature EA1. Abilitv to oDerate andlor monitor the 3.9 I5 EA1.02 Drywell ventilation system..

295030 Low Suppression Pool Wtr EK2. Knowledae of the interrelations 3.5 Lvl 1 5 between LOW SUPPRESSION POOL WATER LEVEL and the followin .

(CFR: 41.7 145%)

EK2.08 SRV discharge submergence..

295031 Reactor Low Water Level I EAI. Ability to operate andlor monitor the 4.4*

2 following as they apply to REACTOR LOW WATER LEVEL :

(CFR: 41.7 145.6)

EA1-01 Low ressure coolant injection IRHRI: Plant-ZDecific.

It 295037 SCRAM Condition Present EK1. Knowled e of the operational 4.1*

and Power Above APRM im lications o8he Downscale or Unknown I 1 folKwin conce ts as the a ly to S C R A ~C O N D ~ O N REACTOR POWER ABOVE APRM PRE~ENY' AND DOWNSCALE OR UNKNOWN :

(CFR: 41.8 to41.10)

EK1.02 Reactor water level effects on reactor power.

ES-401 295038 High Off-site Release Rate EKI. Knowled e of the operational 2.5 19 P im lications o the folEwin conce ts as the apply to HIGH OFF-SISE RELEASE (CFR: 41.8 to41.10)

RA~E  :

EKI.01 Biological effects of radioisotope ingestion.

600000 Plant Fire On Site 1 8 I AK3 Knowledae of the reasonsfor the I 2.8 following respijnses as they apply to PLANT FIRE ON SITE:

AK3.04 Actions contained in the abnormal procedurefor plant fire on site .

v KIA Category Totals:

M Group Point Total: 2 I

0 8

7 of 34 NUREG-1021, Draft Revision 9

ES-401 1 ES-401 BWR SRSExamination Outline Emeraencv and Abnormal Plant Evolutions - Tier IIGrouD 2 IRO / SRO)

Form ES-401-1 E/APE # / Name I Safety Function WA Topic@)

It 295002 Loss of Main Condenser Vac I 3 295007 High Reactor PressureI 3 295008 High Reactor Water Level I2 295009 Low Reactor Water Level 12

~ 295010 High Drywell Pressure/ 5 -

~ 295011 High ContainmentTemp /

5 -

295012 High D well AA2. Abilitv to determine and/or intermet 3.9 Temperature/  !? the followifig as ap I to HIGH DRYWELL

hEelXPERA?3KE.

(CFR: 41.10143.5145.13)

AA2.02 Drywell pressure.

AK3. Knowledge of the reasons for the 3.6 following responses as the a I to HIGH SUPPRESSION POOL ~ E ~ ~ ! & A T U: R E (CFR: 41.5 145.6)

AK3.02 Limiting heat additions.

295014 InadvertentReactivity Addition / 1 -

AK2. Knowled e of the interrelations 3

between HIG OFF-SITE RELEASE RATE and the following:

(CFR: 41.7 145.8)

AK2.03 Off-gas system.

295020 InadvertentCont.

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

ES-401 295032 High Secondary EAI. Ability to operate and/or monitor the 3.7 Containment Area Temperature / following as 5 the a ply to HIGH SECONDARY CO~AINMENT TEMPERATURE.

AREA (CFR: 41.7 / 45.6)

EA1.03 Secondary containment ventilation.. I I

295033 High Secondary EK2. Knowled e of the interrelations 3.9 Containment Area Radiation between HIGt?SECONDARY Levels / 9 CONTAINMENTAREA RADIATIONLEVELS and the followin .

(CFR: 41.7 / 45%)

EK2.04 Standby gas treatment system/FRVS..

295034 Secondary Containment EKI. Knowled e of the operational 3.8 Ventilation High Radiation / 9 im lications oAhe folKwrng concepts as they appl to SECONDARY CONTAINMENT VENTILATION HIGH RADIATION :

(CFR: 41.8 to 41.10)

EKI. O l Personnelprotection.

I 295035 Secondary Containment EK3. Knowledae of the reasons for the I 2.8 High Differential Pressure/ 5 EK3.01 Blow-out panel operation: Plant-Specific..

500000 High CTMT Hydrogen WA Category Point Totals:

c I

Group Point Total:

~

9 of 34 NUREG-1021, Draft Revision 9

ES-401

~

ES-401 I ES-401 BWR SF28 nt S ation Outline p 1 (RO I SRO)

Form ES-401-c _

System # / Name WA Topic(s) IR 203000 RHRILPCI:

- Injection Mode -

205000 Shutdown K1. Knowledge of the physical 3.1/

Cooling connectionsand/or causeeffect 3.4 relationships between SHUTDOWN COOLING SYSTEM(RHR SHUTDOWN COOLING MODE) and the followin :

(CFR: 41.290 41.9 I45.7 to 45.8)

K1.05 Component cooling water systems.

A I . Ability to predict and/or monitor chanaes in arafieters associatedwith operating L e SHUTDOWN COOLING SYSTEM RHR SHUTDOWN COOLING hODE) controls includin :

(CFR: 41.5 / 45.g)

A I .05 Reactor water level 206000 HPCl c _

207000 Isolation K1. Knowledge of the physical 3.8/

(Emergency) Condenser connections and/or causeeffect 3.5 relationshi s between ISOLATION (EMERGf&CY)CONDENSER and the followin :

(CFR: 41.290 41.9 / 45.7 to 45.8)

K1.O1 Reactorvessel: BWR-2,3 .

K2. Knowledge of electrical power lies to the following:

?8%: 41.7)

K2.02 Initiationlogic: BWR-2,3 11 of 34 NUREG-1021, Draft Revision 9

ES-401 K1. Knowledge of the physical 3.7/

connectionsand/or causeeffect 3.3 relationshi s between LOW PRESSURE ORE SPRAY SYSTEM and the followin (CFR: 41.2 to 4?.9 / 45.7 to 45.8)

K1.05 Automatic depressurization system. .

A2.Abili to (a) predictthe impacts of

• *Y the fo lowin PRESSUREC~~ESPRAY on the LOW SYSTEM

and (b) based on those redictions, P

use Proceduresto correc , control. or mitigate the consequences of those abnormal conditions or operations:

(CFR: 41.5 / 45.6)

A2.05 Core spray line break.

K2. Knowledge of electrical power 3.1i suudies to the 2.6*

following:

(CFR: 41.7)

K2.02 Explosivevalves K3. Knowledge of the effect that a loss or malfunction of the STANDBY LIQUID CONTROL SYSTEM will h i e on followin (CFR: 41.7 45Jj K3.03 Core plate differential pressure indication .

K3. Knowledae of the effect that a 3.7 loss or malfulnctionof the REACTOR PROTECTION SYSTEM will have on following:

(CFR: 41.7 / 45.4)

K3.05 RPS logic channels K4. Knowled e of INTERMEDIATE 3.7 RANGE M O ~ I T O R(IRM) SYSTEM desi n feature(s) andlor interlocks whica provide for the following:

(CFR: 41.7)

K4.01 Rod withdrawal blocks.

K5. Knowledge of the operational 2.8 im lications of the folKwin conce ts as the ap I to SOUR& R A N ~ E M O N I T ~ ~&&M)

R SYSTEM -

(CFR: 41.5 / 45.3)

K5.03 Changing detector position

ES-401 215005 APRM / LPRM K6. Knowledge of the effect that a 3.7 loss or malfunctionof the following will have on the AVERAGE POWER RANGE MONITOFULOCAL POWER RANGE MONITOR SYSTEM :

(CFR: 41.7 / 45.7)

K6.01 RPS .

217000 RClC 218000 ADS A2. Abili to (a) predict the impacts 4.2 t h e Y

of the fol owing on A U T O M A T I C DEPRESSURIZATION SYSTEM .

and (b) based on those rediction;,

use procedures to correcr control, or mitigatethe consequences of those abnormal conditions or o erations:

(8FR: 41.5 / 45.6)

A2.06 ADS initiation signals present 223002 PCIS/Nuclear A4. Ability to manually operate 3.6 Steam Supply Shutoff and/or monitor in the control room:

(CFR: 41.7 / 45.5 to 45.8)

A4.03 Reset system isolations 239002 SRVs K5. Knowledge of the operational 3.3/

im lications of the 3.1 folkwin conce ts as they apply to RELIE#SAFET!

VALVES *

(CFR: 41.%/ 45.3)

K5.04 Tail pipe temperature monitoring .

2.4.6 Knowledge symptom based EOP mitigation strate ies.

(CFR: 41.10 I43.5 / 4t.13) 259002 Reactor Water A2. Abili to (a) predict the impacts 3.6 Level Control of the fol

• ?owin on the REACT8R WATER LEVEL CONTROL SYSTEM : and (bl based on those predictions, use pt-odedures to correct, control, or mitigate the consequences of those abnormal conditions or o erations:

(CFR: 41.5)6/! !4 A2.03 Loss of reactor water level input.

13 of 34 NUREG-1021, Draft Revision 9

ES-401 AI. Ability to predict and/or monitor chanaes in ara6eters associatedwith o eratin the STANDBY GAS TREATMEN?

SYSTEM controls including:

(CFR: 41.5 / 45.5)

A I .04 Secondary containment differential pressure .

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

(CFR: 41.7 / 45.5 to 45.8)

A4.03 Local operation of breakers 2.1.30 Ability to locate and operate comnonents / includina local 1

controls.

(CFR: 41.7 / 45.7)

A.4 Ability to manually operate and/or monitor in the control room:

(CFR 41.7/45.5)

A4.03 Battery discharge rate: Plant-Specific K1. Knowledge of the physical connectionsand/or cause effect relationshi s between EMERGENCY GENERAfORS (DIESEUJET) and the followin :

(CFR: 41.290 41.9 / 45.7 to 45.8)

K1.03 Fire protection system .

K6. Knowledge of the effect that a loss or malfunction of the followin will have on the EMERGENCf GENERATORS DIESEUJET) :

(CFR: 41.7 / 45. )f K6.01 Starting air K4. Knowled e of (INSTRUMENT L

AIR SYSTEd design feature@

and or interloc s which providef)or the following:

(CFR: 41.7)

K4.02 Cross-overto other air systems

ES-401 400000 Component K3. Knowledge of the effect that a 2.91 Cooling Water loss or malfunctionof 2.5 the CCWS will have on the following:

(CFR: 41.7 145.6)

K3.01 Loads cooled by CCWS 2.2.25 Knowled e of bases. in P  ?

technical s ecifica ions for limiting conditions or operations and safety limits.

(CFR: 43.2)

WA Category Point Group Point Total:

Totals:

15 of 34 NUREG-I 021, Draft Revision 9

ES-401 ES-401 BWR sR8 ition Outline Form ES-401-'

2 IRO / SRO) -

System# / Name WA Topic(s) IR K6. Knowled e of the effect that 3.0 a loss or rnal?unction of the followin will have on the CONTROL #OD DRIVE HYDRAULIC S stem :

(CFR: 41.7 / 427)

K6.02 Condensate storaae tanks K4. Knowled e of REACTOR 3.3 MANUAL CO~TROLSYSTEM design feature(s) andlor interlocks which provide for the following:

(CFR: 41.7)

K4.05 "Notch override" rod withdrawal. -

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

(PLANT SPECIFIC) design feature(s) and/or interlocks which provide for the following:

(CFR: 41.7)

K4.06 Correction of out of sequence rod positions: P-Spec K3. Knowled e of the effect that 3.9

?

a loss or mal unction of the RECIRCULATION SYSTEM will have on following:

(CFR: 41.7 / 45.4)

K3.03 Reactor power.

K3. Knowled e of the effect that 4.0 a loss or rnal?unction of the RECIRCULATION FLOW CONTROLSYSTEMwill haveon following:

(CFR: 41.7 I45.4)

K3.02 Reactor power.

204000 RWCU K1. Knowledge of the physical 3.1 connections and/or causeeffect relationshi s betweenREACTOR WATERCkANUP SYSTEMand the followin :

(CFR: 41.2?041.9/45.7 to45.8)

K1.01 Reactor vessel . -

ES-401 214000 RPlS 215001 Traversing In-core Probe 215002 RBM 216000 Nuclear Boiler Inst.

219000 RHWLPCI:

ToruslPool Cooling Mode 223001 Primary CTMT and Aux.

226001 RHR/LPCI: CTMT K5. Knowledge of the 2.6 1 Spray Mode operational implications of the followin concepts as the a to RHFQ/LPCi:CoNTAirYhnRY SPRAY SYSTEM MODE :

(CFR: 41.5 145.3)

K5.06 Vacuum breaker operation 230000 RHR/LPCI:

Torus/Pool Spray Mode a K6. Knowled e of the effect that 3.5 a loss or mal nction of the following will have on the 1

R H R / L P C I -

TORUS/SUPPRESSION POOL SPRAY MODE :

(CFR: 41.7 I45.7)

K6.09 Reactor building to suppression pool vacuum breakers.

233000 Fuel Pool CoolinglCleanup 234000 Fuel Handling EauiDment 239001 Main and Reheat 3.6 1 Steam IIII Al.05 Main steam line radiation monitors.

239003 MSlV Leakage Control 241000 Reactor/Turbine Pressure Reaulator 245000 Main Turbine Gen. /

Aux.

17 of 34 NUREG-1021, Draft Revision 9

ES-401 R A2.Abilitv to (dl Dredict the 2.6 1 impacts 6f the followin on the RADIATION MOdITORING 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 / 4516)

IIII Group Point Total:

ES-401 I ES-401 Facility:

Generic Knowledge and Abilities Outline (Tier 3)

Date of Exam:

Form ES-401-53 Category WA# Topic F I SRC 3nly IR # IR #

2.1.1 Knowledge of conduct of operations requirements. 3.7 1 (CFR: 41. I O / 45.13)

1. -

Conduct of 2.1.10 Knowledge of conditions and limitations in the 2.7 1 Operations facility license.

(CFR: 43.1 / 45.13)

~

2.1.20 Ability to execute proceduresteps. 4.3 1 (CFR: 41.10 / 43.5 / 45.12)

Subtot: -

2.2.1 Ability to perform pre-startup procedures for the 3.7 1 facility / including operating those controls associated with plant equipment that could

2. affect reactivity.

Equipment (CFR: 45.1)

Control ~

2.2.2 Ability to manipulate the console controls as 4.0 1 required to operate the facility between shutdown and designated power levels.

(CFR: 45.2) 2.2.11 Knowledge of the process for controlling temporary 2.5 1 changes.

(CFR: 41.10/43.3/45.13)

Subtotal 2.3.2 Knowledge of facility ALARA program. 2.5 1 (CFR: 41.12 /43.4/45.9/45.10) 3.

Radiation Control 19 of 34 NUREG-1021, Draft Revision 9

ES-401 2.3.9 Knowledge of the process for performing a 2.5 containment purge.

+

(CFR: 43.4 / 45.10)

I Subtotal 2.4.29 Knowledge of the emergency plan. 2.6 (CFR: 43.5 / 45.11) 2.4.50 Ability to verify system alarm setpoints and operate 3.3 controls identified in the alarm response manual.

(CFR: 45.3)

I Subtotal 3-27

ES-401 ES-401 BWR ffte-Examination Outline Form ES-401-1 I.

- - Points RO WA Category K K K A A A A G M 4 5 6 1 2 3 4

• Total

3. Generic Knowledge and Abilities Categories

ES-401 ES-401 BWRZ iination Outline FOIVIES-401-'

Emergency I utions -Tier 1IGroup 1(RO I SRO)

UAPE # IName ISafety Function WA Topic(s) 295001 Partial or Corn lete Loss of Forced AA2. Abilitv to determine andor interpret 3.8 core FIOW Circulation PI 4 the followink as they a ly to PARTIAL OR COMPLETE LOSS 6% FORCED CORE FLOW CIRCULATION :

(CFR 41.10 /43.5 /45.13)

AA2.01 Power/flow map8 295003 Partial or Complete Loss ofAC16 295004 Partial or Total Loss of DC 2.1.12 Ability to apply technical 4.0 Pwr I 6 s ecifications for a system.

(8FR 43.2 / 43.5 145.3) 1 295005 Main Turbine Generator T r i I~3 1 295006 SCRAM 1 I AA2. Ability to determine andor interpret the following as 4.4*

the ap ly to SCRAM *

(CJR: fl.10 / 43.5 / 45: 13)

AA2.02 Control rod position.

I 295016 Control Room Abandonment 17 I 295018 Partial or Total Loss of CCWl8 295019 Partial or Total Loss of 2.2.27 Knowled e of the refueling process.

Inst. Air 18 (CFR 43.6 / 45.f3) 1 14295021 Loss of Shutdown Cooling I/ 295023 Refueling Acc Cooling Model8 295024 High Drywell Pressure 1 5 EA2. Abilitv to determine andor interpret the followin a i the ap fy to HIGH DRYWELL PRESSURE:

(CJR fl.10 /43.5 /45.13)

EA2.02 Drywell temperature.

1 295025 Hiah Reactor Pressure I 3 295026 Suppression Pool High Water Temp. I 5 3 of 34 NUREG-1021, Draft Revision 9

ES-401 295028 High Drywell Temperature 2.4.20 Knowledge of operational 4.0 15 imdications of EOP warnings noies.

- / cautions / and (CFR 41.10 /45.13) 295030 Low Suppression Pool Wtr Lvl I 5 -

295031 Reactor Low Water LevelI EA2. Ability to determine andor interpret the 4.2*

2 following -

as the apply to REACTOR LOW WATER LEVE~.

(CFR: 4i.10 143.5 145.13)

EA2.02 Reactor power.

295037 SCRAM Condition Present and Power Above APRM Downscale or Unknown I 1

~~

295038 High Off-site Release Rate I9 600000 Plant Fire On Site / 8 2.4.27 Knowledge of fire in the plant 3.5 procedure.

(CFR: 41.10 / 43.5 / 45.13)

KIA Category Totals: Group Point Total:

ES-401 ES-401 BWR S Xxamination Outline Form ES-401-1 Emergency ant ial F int Evolutions - Tier 11Group 2 (RO I SRO) 1-EIAPE # I Name I Safety Function KIA Topic(s) IR 295002 Loss of Main Condenser 2.4.31Knowledge of annunciators alarms and 3.4 Vac I 3 indications / and use of the response instructions.

(CFR 41.10/45.3) 11 295007 High Reactor Pressure I 3 I 295008 High Reactor Water Level 12 295009 Low Reactor Water Level 12 1: AA2. Ability to determine andor interpret the followin as they app& to LOW REACTOR WATER LEVEL (CFR 41.10 / 43.5 / 45.13)

AA2.03 Reactor water cleanup blowdown rate................... 2.9 2.9 295010 High Drywell Pressure 1 5 29501 1 High Containment Temp I 295012 High D 2.4.16 Knowledge of EOP im lementation 4.0 Temperature I r l hierarchv and coordination wit[ other support procedutes.

(CFR 41.10/43.5 /45.13) 1 295013 High Suppression Pool Temp. 1 5 295014 Inadvertent Reactivity Addition I I 295015 Incomplete SCRAM I 1 AA2. Ability to determine andor interpret the followin as the ap fy to INCOMPLETE SCRAM :

(ClR:11.10/43.5/45.13)

AA2.02 Control rod position.

1 295017 High Off-site Release Ratel9 l 295020 Isolation InadvertentCont.

15 & 7 I 295022 Loss of CRD Pumps / 1 I 295029 High Suppression Pool Wtr Lvl 1 5 295032 High Secondary Containment Area Temperature I 5 of 34 NUREG-1021, Draft Revision 9

ES-401 295033 High Secondary Containment Area Radiation Levels / 9 295034 Secondary Containment Ventilation Hiah RadiationI 9 295035 Secondary Containment High Differential Pressure/ 5 295036 Seconda Containment High SumplArea Kater Level I 5 1

KIA Category Point Totals:

ES-401 ES-401 BWR SRB-Examination Outline Form ES-401-1 Plant Svstems - Tier 2IGroun 1 (RO / SROI System # / Name 203000 RHR/LPCI:

Injection Mode 205000 Shutdown Coolincl 5 K/A Topic@)

206000 HPCl 207000 Isolation (Emergency) Condenser 209001 LPCS 209002 HPCS 21I000 SLC 212000 RPS 215003 IRM A2. Ability to (a) predict the impacts of 3.8 I the followln on t h e I N T f R M E D I A T E RANGE MONITOR fIRM) SYSTEM : and fb)

I

\ I based on thosk predictibns, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

(CFR 41.5 / 45.6)

I A2.04 Up scale or down scale trips I I 215004 Source Range Monitor 215005 APRM / LPRM I

217000 RClC 218000 ADS 223002 PCIS/Nuclear Steam SUDD~V Shutoff 239002 SRVs 259002 ReactorWater A2. Ability to (a) predict the impacts of 3.7 1 Level Control the followln on the REACFOR WATER LEVEL CONTROL SYSTEM ;and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

(CFR 41.5 / 45.6)

A2.03 Loss of reactor water level innut I 7 of 34 NUREG-1021, Draft Revision 9

ES-401 11 261000 SGTS 262001 AC Electrical S 2.1.7 Ability to evaluate plant 4.4 Distribution performance and make operational judgments based on operating characteristics / reactor behavior / and instrument interpretation.

(CFR 43.5 /45.12/45.13) 262002 UPS (ACIDC) l 263000 DC Electrical Distribution 264000 EDGs -

i 300000 InstrumentAir 2.4.48 Ability to interpret control room 3.8 indications to verify the status and operation of system / and understand how operator action s and directives affect plant and system conditions.

(CFR: 43.5 / 45.12)

I 400000 Component Coolina Water WA Category Point Group Point Total:

Totals:

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

ES-401

~

286000 Fire Protection A2. Abili to a) redict the 3.0 7 fP impactso the olowin on the FIRE PR0TECTIOf-l SYSTEM

and (b) based on those predictions, useprocedures to correct, control, or mitigate the consequences of those abnormal conditions or operations

(CFR 41.5 / 45.6)

A2.03 A.C. distribution failure:

Plant-Specific 288000 Plant Ventilation I 290001 Secondary CTMT 290003 Control Room HVAC 290002 Reactor Vessel 2.2.32 Knowledge of the effects lnternals of alterations on core configuration.

(CFR 43.6)

WA Category Point Totals: Group Point Total:

ES-401 11 of 34 NUREG-I 021, Draft Revision 9

ES-401 b

ES-401 ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-40143 I Facility: Date of Exam:

Category Topic I - SR(

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

Conduct of (CFR: 41.10/43.5 /45.12)

Operations 2.1.25 Ability to obtain and interpret station 3.1 reference materials such as graphs /

monographs / and tables which contain performance data.

(CFR 41.10 /43.5 /45.12) 2.1.

2.1.

2.1.

2.1.

Subtotal 2.2.16 Knowledge of the process for making of field changes.

(CFR: 41.10 /45.13) 2.

Equipment Control 2.2.21 Knowledge of pre and post maintenance operability requirements.

(CFR: 43.2) 2.2.

2.2.

2.2.

2.2.

Subtotal 13 of 34 NUREG-1021, Draft Revision 9

3.

Radiation Control 4.

Emergency Procedures 2.4.41 Knowledge of the emergency action level 4.1 I Plan thresholds and classifications.

(CFR 43.5 / 45.1 1) 2.4.

2.4.

2.4.

2.4. I I I I I

ES-401 Record of Reiected WAS Form ES-401-#4 Oyster Creek Written Exam; RO Portion

~~

Randomly Reason for Rejection Selected K/A 264000/K1.03 There is no fire protection associated with the emergency diesel aenerators.

295031/EA1.01 There is no LPCl at OC. The only low pressure ECCS is Core Spray. Discussed with Fred Guenther and he agreed it was amrowiate to use Core SDrav in place of LPCI.

230000/K6.09 Torus sprays are not distinguished from Drywell Sprays in OC EOPs. The only modes of containment spray identified in EOPs are Drywell Spray and Torus Cooling. In addition this KIA was, essentially, a duplicate of 226001/K5.06.

21.10 The examiner reviewed the License Conditions and determined there were very few. None of these appeared to yield good testable material at the RO level. The licensee had requested this WA be reiected based on a Dreviouslv amroved exam outline.

2.2.2 This K&A is more appropriately tested during the Operating Test.

259002/A2.03 This K&A is more appropriately tested during the Operating Test.

The Licensee agreed to include a Feedwater failure from level input in one of the scenarios.

205000/A1.05 This K&A is appropriate for a plant having Shutdown Cooling mode of RHR. Specifically where misoperation of RHR can effect RPV level (from opening the min flow valve, for example). No such relationship exists at OC which has a separate Shutdown Cooling svstem with no min flow valve discharclinn to the torus.

239001/A1.05 Main Steam Line Radiation no longer causes auto closure of MSIVs.

There is a question regarding manual response to increasing steam line radiation in Turbine Building.

ES-401 Record of Reiected WAS Form E S - 4 0 1 4 4 Oyster Creek Written Exam; SRO Portion

~~ ~ ~~~

Tier I Randomly Reason for Rejection Group ~~

Selected WA ~~~

1I1 2950061AA2.06 Verifying rod position following a Scram is done during the operating

~

test. This WA would be oversamdina in the written exam. ~~ ~

212 286OOOlA2.01 No plant-specific logic failure identified as testable material.

314 2.4.41 Classification for EP done on Operating Test.

112 29501212.4.16 This WA is better tested on Operating Test and is expected to be part of one of the scenarios.

311 2.1.5 This KIA is better tested on a JPM.

311 2.1.25 This KIA is better tested on a JPM.

211 259002fA2.03 This KIA is better tested during the Operating Test. Licensee agreed to have a level input event in one of the scenarios.

II2 2950151AA2.02 Oversampling of ATWS between RO and SRO exams.

312 2.2.16 This is sampled in RO and is of low significance for SRO (2.6). In

~~

addition most, if not all the SRO applicants will take both exams.

313 2.3.7 This is not an SRO function. Even non-licensed personnel are aware of RWP meoaration.