ML022240096

Outline Submittal for the Dresden Initial Exam - June 2002

ML022240096
Person / Time
Site:	Dresden
Issue date:	06/03/2002
From:	Otten M Exelon Generation Co, Exelon Nuclear
To:	Office of Nuclear Reactor Regulation
References
50-237/02-301, 50-249/02-301, D-1, ES-201, ES-201-1, ES-201-2, ES-301-1, ES-301-2, ES-301-5, ES-301-6, ES-401-1, ES-401-10, ES-401-2
Download: ML022240096 (36)

Text

Outline Submittal FOR THE DRESDEN INITIAL EXAM - JUNE 2002 Contains the following:

Outline Submittal Letter from Licensee ES-201 -1 Examination Preparation Checklist Letter Exelon cover letter transmitting the Outline ES-201-2 Examination Outline Quality Checklist ES-301-1 Administrative Topics Outline (RO)

ES-301-1 Administrative Topics Outline (SRO)

ES-301-2 Control Room and Facility Walk-Through Test Outline (RO/SRO(l))

ES-301-5 Transient and Event Checklist ES-301-6 Competencies Checklist D-1 Dynamic Simulator Scenario Outline for 4 scenarios ES-401-1 BWR SRO Examination Outline ES-401-2 BWR RO Examination Outline ES-401 -10 Record of Rejected K/As Admin There were no NRC Comments on the submitted test outlines

ES-201 Examination Preparation Checklis: Form ES-201-1 Facility: Dresden Nuclear Station U2/U3 Date :f Examir  :  : June 3, 2002 Examinations Developed by: Facility / NRC (circle one)

Target Chief Date* Task Description / Reference Examiner's Initials

-180 1. Examination administration date confirmed (C.l.a; C 2.a & b) drm

-120 2. NRC examiners and facility contact assigned (C.1 .d.: :.2.e) drm

-120 3. Facility contact briefed on security & other requireme-ts (C.2.c drm

-120 4. Corporate notification letter sent (C.2.d) drm

[-90] [5. Reference material due (C.l.e; C.3.c)] n/a

-75 6. Integrated examination outline(s) due (C.l.e & f; C.3 d) drm

-70 7. Examination outline(s) reviewed by NRC and feedba:k provide: drm to facility licensee (C.2.h; C.3.e)

-45 8. Proposed examinations, supporting documentation, and drm reference materials due (C.l.e, f, g & h; C.3.d)

-30 9. Preliminary license applications due (C.1.1; C.2.g; ES-202) drm

-14 10. Final license applications due and assignment shee: prepared drm (C.1.1; C.2.g; ES-202)

-14 11. Examination approved by NRC supervisor for facilit., licensee drm review (C.2.h; C.3.f)

-14 12. Examinations reviewed with facility licensee (C.1.j; C.2.f & h; C g) drm

-7 13. Written examinations and operating tests approved oy drm NRC supervisor (C.2.i; C.3.h)

-7 14. Final applications reviewed; assignment sheet upda:ed; waive drm letters sent (C.2.g, ES-204)

15. Proctoring/written exam administration guidelines reviewed wi

-7 facility licensee and authorization granted to give wrtten examsz drm (if applicable) (C.3.k)

-7 16. Approved scenarios, job performance measures, ard questior drm distributed to NRC examiners (C.3.i)

• Target dates are keyed to the examination date identified in -.he corpora*e notification letter.

They are for planning purposes and may be adjusted on a case-by-case :)asis in coordination with the facility licensee.

Applies only to examinations prepared by the NRC.

Exeln IxlnGnrto o pay IC Dresden Nuclear Power Station V%%1Cocn(rP.Corn1 N uclear 6500 North Dresden Road Morris. IL 60450-9765 10 CFR 55.40 January 07, 2002 PSLTR: #02-0001 U. S. Nuclear Regulatory Commission Region III ATTN: Operator Licensing Branch 801 Warrenville Road Lisle, IL 60532-4351 Dresden Nuclear Power Station Units 2 and 3 Facility Operating License Nos. DPR-19 and DPR-25 Docket Nos. 50-237 and 50-249

Subject:

Initial License Examination Integrated Examination Outline Enclosed is the integrated examination outline, which Dresden Nuclear Power Station (DNPS) is submitting for review, comment, and approval for the Initial License Examination, scheduled for the week of June 6, 2002, at DNPS.

This submittal includes outlines for the senior reactor operator and reactor operator written examinations, the job performance measure walk-through, the administrative job performance measure walk-through, and the integrated operational scenarios.

This outline has been developed in accordance with NUREG-1021, "Operator Licensing Examiner Standards," Revision 8, Supplement 1.

In accordance with NUREG 1021, Section ES-201, please ensure that these materials are withheld from public disclosure until after the examinations are complete.

January 7, 2002 U. S. Nuclear Regulatory Commission Page 2 Should you have any questions concerning this letter, please contact Mr. D. F. Ambler, Regulatory Assurance Manager, at (815) 416-2800.

Respectfully, Preston Swafford Site Vice President Dresden Nuclear Power Station

Enclosures:

ES-201-2 Examination Outline Quality Checklist ES-201-3 Examination Security Agreements ES-301-1 Administrative Topics Outline ES-301-2 Control Room Systems and Facility Walk-Through Test Outline ES-301-5 Transient and Event Checklist ES-301-6 Competencies Checklist ES-401-1 BWR SRO Examination Outline ES-401-2 BWR RO Examination Outline ES-401-10 Record of Rejected K/As ES-D-1 Scenario Outlines cc: NRC Document Control Desk - w/o enclosures Region II NRC Regional Administrator - w/o enclosures NRC Senior Resident Inspector - Dresden Nuclear Station - w/o enclosures

ES-201 Examination Outline Form ES-201-2 Quality Checklist Facilit Dale of Examination Item Task Descnpion Inilials a b" cot 1 a Verify that the outline(s) fit(s) the appropriate model per ES-401 R b Assess whether the outline was systematically and randomly prepared in accordance with I Section D I of ES-401 and whether all K/A categones are appropnately sampled A0 T c Assess whether the outline over-emphasizes any systems, evolutions, or generic topics.

E N d Assess whether the justifications for deselecled or rejected K/A statements are appropriate . 4..

2. a. Using Form ES-301-5, verify that the proposed scenario sets cover the required number of normal evolutions. instrument and component failures. and major transients.

S I b. Assess whether there are enough scenario sets (and spares) to lest the projected number and M mix of applicants in accordance with the expected crew composition and rotation schedule without compromising exam integrity; ensure each applicant can be tested using at least one new or A0 significantly modified scenario, that no scenarios are duplicated from the applicants' audit test(s)',

and scenarios will not be repeated over successive days.

c. To the extent possible, assess whether the outline(s) conform(s) with the qualitative and quantitative criteria specified on Form ES-301-4 and described in Appendix D.

Alt.

3. a. Verify that:

(1) the outline(s) contain(s) the required number of control room and in-plant tasks.

W (2) no more than 30% of the test material is repeated from the last NRC examination, Ac ,

I (3)* no tasks are duplicated from the applicants* audit test(s), and T (4) no more than 80% of any operating test is taken directly from the licensee's exam banks.

b. Verify that:

(1) the tasks are distributed among the safety function groupings as specified in ES-301.

(2) one task is conducted in a low-power or shutdown condition, (3) 40% of the tasks require the applicant to implement an alternate path procedure, (4) one in-plant task tests the applicant's response to an emergency or abnormal condition, and (5) the in-plant walk-through requires the applicant to enter the RCA.

c. Verify that the required administrative topics are covered, with emphasis on performance based activities.

d. Determine if there are enough different outlines to test the projected number and mix of applicants and ensure that no items are duplicated on successive days.

4. a. Assess whether plant-specific priorities (including PRA and IPE insights) are covered in the appropriate exam section. M G 4-r E b. Assess whether the 10 CFR 55.41/43 and 55.45 sampling is appropriate. A N N

E c. Ensure that K/A importance ratings (except for plant-specific priorities) are at least 2.5. 4 4 R

A d. Check for duplication and overlap among exam sections..

L e. Check the entire exam for balance of coverage.

f. Assess whether the exam fits the appropriate job level (RO or SRO). 11419 UA Pridted Name I Si gature D te

a. Author47 Z 7 -e , tu P

b. Facility Reviewer

(')

c. NRC Chief Examiner (#) _,/'i04': / X ,' ..

d. NRC Supervisor .e * ' -___. /___ __,,

Note:

• Not applicable for NRC-developed examinations.

1. Independent NRC reviewer initial items in Column c: chief examiner concurrence required.

23 of 24 NUREG-1021, Revision 8, Supplement 1 NRC Copy

ES-301 Adminstrative To*,cs Out -e Form ES-301-1 Facility: Dresden 1:e of Examination: May 27, 2002 Examination Level (circle one)0 SRO Dperating Test Number: ILT 01-1 Administrative Describe method of evalustion:

Topic/Subject 1. ONE Administrative DM, OF Description 2. TWO Administrative -uestio- S A. 1 Safety-Related JPM: Determine if Jet Pimp F': ,,, Meets Requirements Surveillance REF: DOS 0202-02, Je- Pump z:erability and Degradation K/A: 2.1.25 [Ability to obtin and -:erpret station reference material such as graphs / monograohs / a -: tables which contain performance data] RO IMPORTANCE: 2.8 Shutdown JPM: Verify Off-Site Power Sou:es Available Power Sources REF: DOS 0040-10, Urt 2 Shu::own Power Sources and Distribution K/A: 2.1.31 [Ability to locate contr: room switches / controls and indications and to determihe that t-ey are correctly reflecting the desired plant lineup] RO II.PORTANCE: 4.2 A.2 Tracking of JPM: Log and Track Short Dura:.on Timeclock Limiting Conditions for REF: OP-AA-108-104, Technics Specification Compliance Operations K/A: 2.2.23 [Ability to track limiting conditions for operations]

RO IMPORTANCE: 2.6 A.3 Radiation JPM: Locate Valve 2-1201-122 -'id Determine Requirements for Control Entering RCA Requirements REF: RP-AA-460, Contrmls for F gh and Very High Radiation Areas K/A: 2.3.1 [Knowledge of 10 CFR 20 and related facility radiation control requirements] RO APORT'-.NCE: 2!.6 A.4 Post-Accident JPM: Estimating the Post Accide't Noble; Gas Activity Instrumentation REF: DOP 1700-10, Estmating :-e Post Accident Noble Gas Activity Release W.th/With-:-t the E berline SPING-4 Monitor Available K/A: 2.4.3 [Ability to identie! post-a:cident instrumentation]

RO IMPORTANCE: 3.5 NUREG-1021, Revision 8 NRC Copy

ES-301 Adminstrative Topics Outline Form ES-301-1f Facility: Dresden Date of Examination May 27, 2002 Examinaton Leve - rcle one): RO Operating Test Number ILT 01-1 Administrative Describe method of evaluation:

Topic"Subjec: 1. ONE Administrative JPM, OR Description 2. TWO Administrative Questions A.1 Safety-Rela:ed JPM: Review Faulted Jet Pump Operability Surveillance Surveillance Review REF: DOS 0202-02, Jet Pump Operability and Degradation K/A: 2.1.12 [Ability to apply technical specifications for a system]

SRO IMPORTANCE: 4.0 Overtime JPM: Evaluate Overtime of Operators and Complete Required Limitations Documentation for Exceeding Allowable Limits REF: LS-AA-119, Overtime Controls K/A: 2.1.5 [Ability to locate and use procedures and directives related to shift staffing and activities] SRO IMPORTANCE: 3.4 A.2 Controlling JPM: Review and Approve Temporary Modification Temporary Changes REF: CC-AA-112, Temporary Configuration Changes K/A: 2.2.11 [Knowledge of the process for controlling temporary changes] SRO IMPORTANCE: 3.4 A.3 Reviewing JPM: Review Liquid Radwaste Discharge Permit Liquid Release Permits REF: DOP 2000-110, Waste Surge Tank Radwaste Discharge to River with the Off Stream Liquid Effluent Monitor Operable K/A: 2.3.6 [Knowledge. of the requirements for reviewing aand approving release permits] SRO IMPORTANCE: 3.1 A.4 Emergency JPM: Prepare a NARS Form for TransmittalIncluding Determination Plan Off-Site of PARS Notifications REF: EP-AA-1 14, Notifications; EP-AA-1 13, Protective Actions I

K/A: 2.4.38 [Ability to take actions called for in the facility emergency plan / including (if required) supporting or acting as emergency coordinator] SRO IMPORTANCE: 4.0 NUREG-1021, Revision 8 NRC Copy

ES-301 Control Room Systems and Facility Walk-Through Test Outline Facility: Dresden Exam Level (circle one) Form ES-301-2 QO~ SRO(U) Date of Examination:

Operating Test May 27, 2002 No.: ILT 01-1 System / JPM Title Code*

a. Recirculation Type Recirculation Flow Control Safety Function Flow Control System /Transfer from Individual Man. to Master D, S

b. High Pressure (1)

Coolant Injection Reactivity Control Start HPCI Alarm, DOS for Surveillance System 2300-03; K/A: with Exhaust / Manually c.

206000A4.12, Pot Drain N, A, S, (2)

Main Turbine 4.0 / 3 .9 Synchronize Generator and Reactor Water the Main Generator Auxiliary Systems Inventory Control K/A: 245000A4.02, to the Grid, /

3.1 / 2.9 DGP 01-01; N, 5, L

d. Safety Relief (4)

Heat Removal High Pressure,Valves / Relief Valve frm DOS 0250-04; Testing at Reactor Core 4.4 /4.4 K/A: 239002A4.01, Low and at N, A, S, L

e. A. C. Electrical (3)

Distribution Reactor Pressure 1 using the crosstie breakers, I Crosstie Bus 23-1 Control DOP 6500-30; and 33-K/A: M, 5, L f.Rod Worth (6)

Position Data,Minimizer Electrical System I Enter DOP 0400-02 Substitute Rod 26A:20 01 6A4.01 D, S

g. Standby ,3 /3.27 . (7)

MaintenanceGas Treatment Instrumentation Testing with System / SBGT Post Signal, DOS 7500-02; Receipt of Auto K/A: 295020AK2. Initiation D, A, S 11, 3.2 / 3.4 (9)

Radioactivity

a. Release Isolation Condenser Pump Start with Faulted / Isolation Condenser D SS P 10 0-C Lube Oil Pressure, Makeup R ; K/A : 2 95 D, A 0 16AA 1.

b. Reactor Protection 4 0 9,

. '0/4

.0 (4)

"R Heat Removal Power Supply, System / Transfer frm K/A:

DOP 0500-03 RPS to Reserve eactor Core 212000K4.03,  ;

3.0 /3.1 D

c. Instrument '(7)

Connect Air System / Unit Instrumentation Operation, 2/3 K/A: 295019AA1.02, DOP 4700-~03 Instrument Air Cross-3.3  ; D, R

• Type Codes: /3.1 (D)irect from (8) bank, (M)odified Plant Service .

from bank, (N)ew, Systems (A)lternate path, (S)imulator, (C)ontrol room, (L)ow-Power, (R)CA NUREG-1 021, Revision 8 NRC Copy

Appendix D Scenario Summary Attachment 1 Facility: Dresden Scenario No: ILT-N-1 Op-Test No: ILT 01-1 Summary:

The crew assumes the shift with a unit startup in progress and the reactor at about 15% power.

In accordance with the unit startup procedure, the SRO directs the Assistant NSO (ANSO) transfer auxiliary electrical power from transformer to 22 to transformer 21. The NSO, as directed by the SRO, then increases reactor power by control rod withdrawal. Following the power increase, the main turbine bypass valve #1 opens spuriously. The valve is closed when ANSO takes manual action at the EHC control the panel. Circulating water pump 2B then trips overload and the ANSO manually starts circulating on vacuum. APRM channel 5 fails downscale water pump 2C to maintain condenser followed by a companion IRM 15 spike upscale partial half-scram occurs. The NSO inserts and a a complete half-scram. The SRO addresses technical specification requirements for the the ARPM channel. The APRM channel is bypassed the half-scram cannot be reset by the NSO. but A spurious RPS "A" channel half scram occurs resulting in a full reactor scram. Several control rods fail to insert and an ATWS occurs. During actions to recover from the ATWS an ECCS suction line break occurs resulting in a lowering water level. An emergency depressurization torus is then performed as directed by the DEOP LL primary containment control. The scenario is for terminated when the reactor is depressurized, reactor pressure is being controlled, and actions are taken to address the lowering torus water level.

',PC Copy

Appendix D Scenario Outline Form E_.2-1 Facility: Dresden Scenario No: ILT-N-2 Op-Test Noý ILT Examiners:

Operators:

Initial Conditions: Unit in Mode 2 at approximately 1% reactor power; IRM channel 12 out of service; Unit 3 is in Mode 5.

Turnover: Unit startup in progress; return RBCCW pump 2B to service following maintenanc then continue power ascension Event Malf. Event No. Event No. Type* Description N/A N ANSO SRO sa swap RBCCW BC pumps up 2N/A R NSO SRO raise reactor power by withdrawing S

RODxxxDN NSO control rods SRO control rod double notches during withdrawal 4 NII12POT I NSO SRO IRM channel fails upscate 5 PCP85401 ANSO SRO drywell pressure controller failure 6 K49 T12 ANSO main feed breaker to MCC 24-1 trips with failure of C SRO emergency diesel generator to start automatically 131 ALL steam line break in X-area (outside drywell) from main steam line 2A M

8 112 /116 ALL partial failure of group 1 isolation; steam discharge into X-area continues at reduced rate (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor NUREG-1021, Revision 8, Supplemer*

NRC Copy

Appendix D Scenario S -mary Attachment 1 Facility. Dresden Sce'ario No. -T-N-2 Op-Test No: ILT 01-1 Summary:

The crew assumes the shift with a unit sta-:up in proc'ess and the reactor in Mode 2 at about 1% power. Maintenance has been completed on RBCSW pump 2B and the SRO directs the ANSO to switch running pumps and place :he RBCC",',' pump 2B in service. The NSO, as directed by the SRO, then continues the power ascers on for unit startup by control rod withdrawal. During the control rod withdrawal, a contr: rod double notches beyond the withdraw limit and must be repositioned. IRM channel 12 then :'is upscale and a half-scram occurs on the RPS "A" channel. The NSO bypasses the failed IRM crannel and the SRO addresses the technical specification requirements for the failure. Dr,well pressure then begins to decrease and pressure control is regained when the ANSO takes m; rual control of the drywell pressure controller. The main feed breaker to MCC 24-1 then tr- s and the U2 EDG fails to automatically start. The U2 EDG does run when manually started b'. :he ANSO and power is restored to MCC 24-1. Temperature alarms in the X-area are received ,,,hen a break in main steam line 2A occurs outside of the drywell. As temperatures increase, a grcup 1 isolation and reactor scram are automatically initiated, The MSIVs in main steam line 2A do not close fully and steam discharge into the X-area continues at a reduced rate, An emergency depressurization is then conducted as directed by the DEOP for secondary containment control. The scenario terminates when reactor pressure has been reduced and is under ccntrol.

NFRC COPY

Appendix - Scenario Outline Form ES-D-1 Facility: E[>eden Scenario No: ILT-N-3 Op-Test No: ILT 01-1 Examiners Operators:

Initial Conc :'.s: Approximately 40% reactor power; IRM channel 12 out of service; Unit 3 is in Mode 5.

Turnover:. -it shutdown in progress for forced outage; shutdown reactor feed pump 2B, then continue pc3/4,,, =r reduction for unit shutdown Event Yalf. Event Event No. 14o. Type* Description ANSO 1 N/A N SO shutdown reactor feed pump for unit shutdown SRO NSO 2 '/A R SO lower reactor power by reducing recirculation flow SRO 3 R__MLS I NSO feedwater level control system setpoint drifts high SRO SE?1371 NSO 4 SE:,DOP1 C SO reactor feed pump 2A failure FVYDOP1 SRO 5 AS~xSDANSO 5AC*3xSD ANSRO spurious ADS valve opening 6 -11 ANSO stator cooling water pump trips on overload and MG: GCBTR SRO standby pump fails to start automatically 7 "21 ANSO Bus 22 normal feed breaker trips on overload SRO 8 - M ALL feed line break inside of drywell 9 HP-,'BKR ANSO HPCI injection valve failure SRO (N)ormal, (F eactivity, (I)nstrument, (C)omponent, (M)ajor NUREG-1021, Revision 8, Supplement I NPC COpY

Appendix D Scenario Summary Attachment 1 Facility: Dresden Scenario No. ILT-N-3 Op-Test No: ILT 01-1 Summary:

The crew assumes the shift with a unit shutdown in progress and the reactor at about 40% power.

The SRO, in accordance with the unit shutdown procedure, directs the ANSO to shutdown a reactor feed pump. The NSO, as directed by the SRO, then lowers reactor power by reducing recirculation flow. Manual control of the feedwater level control system is taken by the NSO after it is observed that the setpoint begins to drift high. A low oil pressure alarm is received for reactor feed pump 2A and the auxiliary oil pump cannot be started. The NSO then starts the standby reactor feed pump and pump 2A is shutdown. An ADS valve then spuriously opens and is manually closed by the ANSO. The SRO addresses the technical specification requirements for the ADS valve failure. The running stator water cooling pump trips and the standby pump fails to start automatically. The standby pump is manually started by the ANSO. The normal feeder breaker to Bus 22 then trips on overload resulting in a loss of all reactor feed pumps. The reactor scrams (or is scrammed) on low water level and shortly thereafter, a feed line break occurs inside the drywell. When HPCI is initiated the injection valve fails to open. An emergency depressurization is performed as directed by the DEOPs due to the inability to maintain reactor water level. The scenario terminates when reactor water level is restored.

NRC COPY

Appendix D Scenario Outline --:rm ES-D-I Facility: Dresden Scenario No: ILT-N-5 Op-Tes: ',D: ILT01-1 Examiners: Operators.

Initial Conditions: Approximately 70% reactor power; IRM channel 12 out of service Jnit 3 is in Mode 5.

Turnover: Power reduction in progress for drywell entry.

Event Malf. Event Eve,,

No. No. Type* Descripon ANSO 1 N/A N SO establish drywell de-inerting Ineup SRO NSO 2 N/A R SO lower reactor power by reducrg recirculatio- :ow SRO 3 RRMAFDBK NSO recirculation pump controller speed signal fa -,,e SRO MGGH2CON ANSO main generator hydrogen temperature contro er output SRO fails low 5 ICTUBLK C ANSO isolation condenser tube leak SRO N/A NSO 6 C SRO CRD pump failure 7 CIGP1 ALL spurious group 1 actuation and reactor scram 8 RDFHYLK M ALL SDV full hydraulic lock (ATWS) 9 SCRLFVAD SCRLFVBD NSO SRO SBLC pump relief valves fail open SCRLFVBD SRO (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor NUREG-1021, Revision 8, Su-:_ement 1 NRC COPY

Appendix D -':ario Summary Attachment 1 Facility: Dresden Sc-- *-io No: ILT-N-5 Op-Test No: ILT 01-1 Summary:

The crew assumes the shift with reactor p: .,,er at about 70% and a power reduction in progress to conduct a drywell entry for leakage insp-:-:ons. The ANSO, as directed by the SRO, lines up systems for drywell de-inerting. The NSO :--en lowers reactor power by reducing recirculation flow following direction by the SRO. Durinn -e power reduction, the speed control signal fails low for recirculation pump 2A and the pump fic,,, -eduction is stopped when the NSO locks out the scoop tube. Alarms are then received due :-- high main generator hydrogen temperature resulting from a failed controller- Hydroge- -÷mperature is restored after the controller is placed in manual by the ANSO. Alarms are then re:.eived due to an isolation condenser tube leak. The isolation condenser is manually isolated bý e ANSO. The SRO addresses the technical specification requirements for the inoperab e solation condenser. A field report is received that the 2A CRD pump is failing due to rapid oil $sfrom a leak. The NSO shutdowns the 2A CRD pump and starts the 2B CRD pump. During ',ID work on main steam line flow transmitters, a spurious group I isolation and a reactor scr;m occurs. A full hydraulic lock of the scram discharge volume results in little inward rod -otion and an ATWS. When boron injection is initiated, the SBLC pumps do not inject borc- into the reactor due to the pump relief valves failing open. The crew then initiates actions for ah;e- ate SBLC injection. The scenario terminates after manual driving in of control rods is in progrez-:S and a scram/reset has been successfully initiated.

NRC COPY

1 ",-401 13NNIZ. SRO Examination Outline Pirinted: 01/04'2002 F cilitv: Dresden Forni FS-40 1-1 Exam D)ate: 05'27.."2002 Exam Level: SRO K/A Category Points Group Point

'rot a I KI K2 K3 K4 K5 K6 AI A2 A3 A4 G 1 4 5 3 5 5 4 26 E-7.rgency

& 2 3 2 3 3 3 3 17

-normal I

?Pant Tier E%-Iutions Totals 7 7 6 8 8,7 43 1 2 2 2 2 2 2 2 2 2 2 3 23 2.

2 1 1 2 1 1 2 1 1 1 0 2 13 P lant ___ ______

S stems 3 0 0 0 0 1 0 0 1 0 0 2 4 Tier Totals 3 3 4 3 4 4 3 4 3 2 7 40 Cat I Cat 2 Cat 3 Cat 4 2 Generic Knowledge And Abilities 4 5 4 4 17 No:ý:

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

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

3. Select topics from many systems; avoid selecting more than two or thlree KIA topics from a given system 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 category tter.

I NRC COPy

BWR SP" "xamination Outline Printed: 01 Facility: .Sden ES - 401 Emergency and Abnormal Plant Evolutions - Tier 1 / Group 1 Form ES-401-1 E/APE # E/APE Name / Safety Function K1 K2 K3 Al A2 G KA Topic Imp. Points 295003 Partial or Complete Loss of A.C. Power /6 X AA1.03 - Systems necessary to assure safe plant 4.4* 1 shutdown 295003 Partial or Complete Loss of A.C. Power / 6 X AK2.03 - A.C. electrical distribution system 3.9 1 295009 Low Reactor Water Level / 2 X AA2.02 - Steam flow/feedflow mismatch 3.7 1 295010 High Drywell Pressure / 5 X AA2.06 - Drywell temperature 3.6 1 295013 High Suppression Pool Temperature 5 X AA2.01 - Suppression pool temperature 4.0 1 295014 Inadvertent Reactivity Addition / -1 . X AK2.01 - RPS 4.1 1 295014 Inadvertent Reactivity Addition! I.. X AK3.02 - Control rod blocks 3.7 1 295015 Incomplete SCRAM / I X 2.4.30 - Knowledge of which events related to system 3.6 1 operations/status should be reported to outside agencies.

295015 Incomplete SCRAM / I X AK1.04 - Reactor pressure: Plant-Specific 3.8 1 295016 Control Room Abandonment! 7 . X 2.1.32 - Ability to explain and apply system limits and 3.8 I precautions.

295016 Control Room Abandonment /7 X AA1.04 - A.C. electrical distribution 3.2 1 295017 High Off-Site Release Rate 9 X AK2,04 - Plant ventilation systems 3.3

BWR SP' xamination Outline Printed: 01 Facility: -oden ES - 401 Emergency and Abnormal Plant Evolutions - Tier 1 / Group 1 Form ES-401-1 E/APE # E/APE Name / Safety Function Ki K2 K3 Al A2 G KA Topic Imp. Points 295023 Refueling Accidents / 8 X 2.1.14 - Knowledge of system status criteria which 3.3 1 require the notification of plant personnel.

295023 Refueling Accidents / 8 X AAI.03 - Fuel handling equipment 3.6 1 295024 High Drywell Pressure / 5 X EK1.01 - Drywell integrity: Plant-Specific 4.2* 1 295025 High Reactor Pressure / 3 X 2.4.4 - Ability to recognize abnormal indications for 4.33 system operating parameters which are entry-level conditions for emergency and abnormal operating procedures.

295025 High Reactor Pressure /3 X EK3.04 - Isolation condenser initiation: Plant-Specific 4.7* 1 295030 Low Suppression Pool Water Level / 5 X EA2.04 - Drywell! suppression chamber differential 3.7 pressure: Mark-I&II 295031 Reactor Low Witer Level 12' ' X EA2,01 - Reactor wutcr level 4. 6 295031 Reactor Low Water Level 2 X EK2.16 - Reactor water level control 4.1 1

\tvc AI l Domviscalc tc Unknown II 295037 SCRAM Condition Present and Reactor Power X EKI.02 - Reactor water level effects on reactor power 4.3*

Above APRM Downscale or Unknown / I 295038 High Off-Site Release Rate /9 X EA1.03 - Process liquid radiation monitoring system 3.9 295038 High Off-Site Release Rate / 9 X EKI.03 - tMeteorological effects on off-site release 3.8

z Facility: Aden BWR SPr 'xamination Outline C3 Printed: 01 0 ES - 401 0

A Category tl'otals: 4 5 3 5 5 4 Group Point Total: 26

-3

7

'den RW~R SP~r 'Xam iination Outline i,1iV{r M

()

n ES -401 0

'0 and Abnormal Plant Evolutions - Tier 1

/ Group 2 K(2 K31]AI ýA2 GýKA Topic I

I

z 1) 0 Facility:. -0den BWR SP ` xamination Outline Printed: 01, 0 ES- 401 Emergency and Abnormal Plant Evolutions - Tier 1 / Group I 2 Form ES-401-1 E/APE # E/APE Name / Safety Function K1 K2K3 Al A2 G KATopic i Imp. Points 295032 High Secondary Containment Area Temperatur e/5 X 2.4.49 - Ability to perform without reference to 4.0 1 procedures those actions that require immediate operation of system components and controls.

295032 High Secondary Containment Area Temperature / 5 X EA 1.03 - Secondary containment ventilation 3.7 1 295033 High Secondary Containment Area Radiation Leevels X EA 1.01 - Area radiation monitoring system

/9 4.0 1 295034 Secondary Containment Ventilation High Radiat ion /

x rTIrl ()I T) 9 ,- rcinuiue, protection 4.1 1 F

295035 Secondary Containment High Differential Pressure /

5 x EK3.02 - Secondary containment ventilation response 3.5 1 K/A Category Totals: 3 2 3 3 3 3 Group Point Total: 17

BWR SRO I- imination Outline Printed: C '002 Facility: ;esden ES - 401 Plant Systems - Tier 2 / Group 1 Form ES-401-1 Sys/Ev # System / Evolution Name KI K2 K3 K4 K5 K6 Al A2 A3 A4 G KA Topic Imp. Points 202002 Recirculation Flow Control System / I X K4.05 - Limiting recirculation pump speed 34 1 mismatch: Plant-Specific 202002 Recirculation Flow Control System / 1 X K5.01 - Fluid coupling: BWR-3, 4 2.8 206000 High Pressure Coolant Injection X K6.09 - Condensate storage and transfer 3.5 System / 2 system: BWR-2, 3, 4 206000 High Pressure Coolant Injection X A 1.06 - System flow: BWR-2, 3, 4 3.7 System / 2 209001 Low Pressure Core Spray System / 2 X K1 .10 - Emergencv vcnerntor 18 209001 Low Pressure Core Spray System / 2 X K3.03 - Emergency generatorts 3.>

I 1000 I iqlid ('

1RImlb\ riit' s.u-,01 I X "..I.o IK1miiucdgc ',)Ilyiufl p i d 1(k91' I .UP mitigation strategies.

215004 Source Range Monitor (SRM) System X A4.04 - SRM drive control switches 3.2

/7 215005 Average Power Range Monitor/Local X A3.07 - RPS status 3.8 1 Power Range Monitor System / 7 216000 Nuclear Boiler Instrumentation / 7 X K6.01 - A.C. electrical distribution 3.3 218000 Automatic Depressurization System / X K3.01 - Restoration of reactor water level after 4.4*

3 a break that does not depressurize the reactor when required

BWR SRO I -mination Outline Printed: C '002 Facility: :esden ES - 401 Plant Systems - Tier 2 / Group 1 Form ES-401-1 Sys/Ev# System / Evolution Name KI K2 K3 K4 K5 K6 Al A2 A3 A4 G KA Topic Imp. Points 218000 Automatic Depressurization System X K4 01 - Prevent infivcr-cni n il iif d , 1.-thI 223001 Primary Containment System and X A3.02 - Vacuum breaker/relief valve operation 3.4 1 Auxiliaries / 5 223002 Primary Containment Isolation X 2.4.4 - Ability to recognize abnormal 4.3 SystemlNuclear Steam Supply indications for system operating parameters Shut-Off/ 5 which are entry-level conditions for emergency and abnormal operating procedures.

223002 Primary Containment Isolation X K 1.19 - Component cooling water systems 2.9 System/Nuclear Steam Supply Shut-Off / 5 226001 RHR/LPCI: Containment Spray X K2.02 - Pumps 2.9*

System Mode / 5 ... .. . ...

226001 RHR/LPCI: Containment Spray . . . X K5,02 - Water hammer 2.7 System Mode / 5 259002 Reactor Water Level Control System./ X A2.01 - Loss of any number of main steam flo 3.4 2 inputs 259002 Reactor Water Level Control System X A4.06 - DP/Single/three element control 3.2 1 2 selector switch: Plant-Specific 261000 Standby Gas Treatment System / 9 X 2.4.30 - Knowledge of which events related to 3.6 1 system operations/status should be reported to outside agencies.

z BWR SRO T" imination Outline 0 Facility: Printed: C '002

esden 0

0

ý) ES -401 1< F-K/A Category Totals: 2 2 2 2 2 2 2 2 2 2 3 Group Point Total: 23

z BWR SRO' " mination Outline 0 Printed: C '002 Facility: iuresden 0

ES - 401 Plant Systems - Tier 2 / Group 2 I 5 Al A2A 4GKTopic K2.05 - Alternate rod insertion valve solenoids:

Plant-Specific I

I I

262002 Uninterruptable Power Supply X K3.17 - Process monitoring: Plant-Specific (A.C./D.C.) / 6 3.1 I

zn BWR SRO V amination Outline Printed: C '002 n Facility: .)resden 0

1< ES -401 ------

K/A Category Totals: 1 1 2 1 1 2 1 1 1 0 2 Group Point Total: 13 2

BWR SRO r imination Outline Printed: "002 z

n Facility: resden 0 ES - 401 Plant Systems - Tier 2 / Group 3 Form ES-40 I-1 C,,

Sys/Ev # System / Evolution Name K11 K2 K3 K4 K5 K6 Al A2 A3 A4 G KA Topic Imp. Points 239001 Main and Reheat Steam System / 3 X K5.06 - Air operated MSIV's 2.9 1 288000 Plant Ventilation Systems / 9 X A2.01 - High drywell pressure: Plant-Specific 3.4 I 288000 Plant Ventilation Systems / 9 X 2.1.33 - Ability to recognize indications for 4.0 1 system operating parameters which are entry-level conditions for technical Ispecifications.

290002 Reactor Vessel Internals / 5 X 2.1.32 - Ability to explain and apply system 3.8 1 1limits and precautions.

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

,',..UowICuge and ADilities Outline (Tier 3)

Printed: 01/04/2, BWR SRO L..diination Outline Form ES-401-5 Facility: Dresden Generic Category KA KA Topic Imp. Points Conduct of Operations 2.1.13 Knowledge of facility requirements for controlling vital / controlled access. 2.9 1 2.1.11 Knowledge of less than one hour technical specification action statements for systems. 3.8 1 2.1.22 Ability to determine Mode of Operation. 3,3 1 2.1.8 Ability to coordinate personnel activities outside the control room. 3.6 t Category Total: 4 Equipment Control 2.2.3 (multi-unit) Knowledge of the design, procedural, and operational differences between 3.3 1 units.

2.2.8 Knowledge of the process for determining if the proposed change, test, or experiment 3.3 1 involves an unreviewed safety question.

2.2.26 Knowledge of refueling administrative requirements. 3.7 1 2.2.2 Ability to manipulate the console controls as required to operate the facility between 3.5 1

. shutdown and designated power levels.

2.2.34 Knowledge of the process for determining the internal and external effects on core 3.2" I reactivity.

Category Total: 5 Radiation Control 2.3.9 Knowledge of the process for performing a containment purge. 3A4 1 3.1 1 2.3.6 Knowledge of the requirements for reviewing and approving release permits.

2.9 1 2.3.2 Knowledge of facility ALARA program.

3,1 1 2.3.4 Knowledge of radiation exposure limits and contamination control, including permissible levels in excess of those authorized.

(ltiegoly IuiItI: -1

z .... , I-iumIes vutlne (Tier 3)

"BWR SRO L- -dination Outline Printed: 01/04/2, 0

0D Faeilit:- Dresden 1< Form ES-401-5 Generic Category KA KA Topic "E m rg n cy- Pln243 24 .- --

2 --- Knowledge

' Im p .

of operator response to loss of all annunciators. P oin ts 3

2.4.7 Knowledge of event based EOP Mitigation strategies.

3.8 2.4.35 Knowledge of local auxiliary operator tasks during emergency system geography and system operations including 2.4.45 implications.

Ability to prioritize and interpret the significance of each annunciator or alarm.3.

Category Total:

4 Generic Total: 17

FS-40 I 13WRI R() Exa n1iin alion Out line PacilitN : I't'iltcd: 0 1 04 '2002

[DI e-cncc 1"0"11 F S-4 1-2 fDatc: 052~7/200-2

-EXIM FA,"" Levecl: RO NRC Copy I

z 0 Facility: ,den BWR R-" amination Outline Printed: 01/ 12 0

o ES- 401 0 and Abnormal Plant Evolutions - Tier 1 / Group 1 K2 K3 A1 IA 2 IG KAToni, K/A Category Totals: 4 3 3 2 1 0 Group Point Total: 13

z R R (" ' am ination O utline SBW 01/

Printed: -2 0 FSly0d Emergency and Abnormal Plant Evolutions

- Tier 1 / Group 2 E/APE # E/APE Name / Safety Function K1 K2 K3 Al A2 G KA Topic Form ES-401-2 Fmp PS4 295002 Loss of Main Condenser Vacuum / 3 x AK2.04 - Reactor/turbine pressure regulating system 3. 1 295003 Partial or Complete Loss of A.C. Power / 6 X AK2.03 - A.C. electrical distribution system 3.7 295003 Partial or Complete Loss of A.C. Power / 6 X AA 1.03 - Systems necessary to assure safe plant 44 shutdown 295004 Partial or Complete Loss of D.C. Power / 6 XAK3.02 - Ground isolation/fault determination 295008 High Reactor Water Level / 2 xAK3.04 - Reactor feed pump trip: Plant-Specific3.

295013 High Suppression Pool Temperature /5 X AA2.01 - Suppression pool temperature3.

295016 Control Room A~bandonment ...... AAI.04 - A.C. electrical distribution 295017 High Off-Site Release Rate / 9 X AK2.04 - Plant ventilation systems 295017 High Off-Site Release Rate / _9_.AK3.03

- tImplementation of site emergency plan 3.3 295018 Partial or Complete Loss Water /8 of Component Cooling p....X2.1.14 require -the Knowledge of system status criteria which2.

notification of plant personnel.

295019 Partial or Complete Loss of Instrument Air/ 8 S.. . i o AK2.17 - High pressure coolant injecction,:

Plant-Specific 295020 Inadvertent Continment Isolation/L5 X AA2.0l - Drywell/containment pressure 6

? S00S ftilh \\\C lZ l pr ' rc 9

, I 1r,\. .;ll. .li B '

295029 High Suppression Pool Water Level 5 X EA2.02 - Reactor pressure

.5 I I

z SFacility: ,den BWR RU" amination Outline Printed: O0/( y) o ES- 401 0o Emergency and Abnormal Plant Evolutions K l3 r

- Tier 1 / Group 2 Form ES-401-2 KI1 K2 K3 Al A2 G KATopic evels X EA 1.01 - Area radiation monitoring system K/A Category Totals: 4 4 3 4 3 1

roiup Point T(otal

J

z Fi n Facility:

,den BWR PR' aamination Outline Printed: 01, 0 ES -401 Emergency and Abnormal Plant Evolutions - Tier 1 /

0 _

K1 IK2 K3 Al IA2 IG IKAToDiC I

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

z SBWR RO *' mination Outline 0 Facility: resden 0

SES - 401

BWR RO F "mination Outline Printed: 0 002 Facility: esden ES - 401 Plant Systems - Tier 2 / Group 1 Form ES-401-2 Sys/Ev# System / Evolution Name Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G KATopic Imp. Points 215004 Source Range Monitor (SRM) System X A4.04 - SRM drive control switches 3.2

/7 215005 Average Power Range Monitor/Local X A3.07 - RPS status 3.8 1 Power Range Monitor System / 7 215005 Average Power Range Monitor/Local X K5.06 - Assignment of LPRM's to specific 2.5* 1 Power Range Monitor System / 7 APRM channels 216000 Nuclear Boiler Instrumentation / 7 X K6.01 - A.C. electrical distribution 3.1 1 216000 Nuclear Boiler Instrumentation / 7 X K4.01 - Reading of nuclear boiler parameters 3.6 1 outside the control room 218000 Automatic Depressurization System / X K3.01 - Restoration of reactor water level after 4.4 1 3 a break that does not depressurizc the reactor

\vhl i i 218000 Automatic Depressurization System! X K4.01 - Prevent inadvertent initiatior of ADS 3.7 1 3 logic 223001 Primary Containment System and X A3.02 - Vacuum breaker/relief valve operation 3.4 1 Auxiliaries / 5 223002 Primary Containment Isolation X K1.1 - (Crmporcni (o ,lit!. NI,.li ..i' .

Sv'tcrn/1Nuclcar Stc;imn Siply Shut-Olf/ 5