ML112940418

2011 Duane Arnold Energy Center Initial Examination Outline Submittal

ML112940418
Person / Time
Site:	Duane Arnold
Issue date:	06/27/2011
From:	Costanzo C R Nextera Energy, NextEra Energy Duane Arnold
To:	McNeil D R Operations Branch III
Shared Package
ML11167A125	List: NG-11-0123, 2011 Duane Arnold Energy Center Initial Examination Outline Submittal
References
NG-11-0123, NUREG-1021
Download: ML112940418 (39)

Text

2011 DUANE ARNOLD ENERGY CENTER INITIAL EXAMINATION OUTLINE SUBMITTAL M NEXTeraDUANE ARNOLD April 7, 2011 NG-11-0123 NUREG 1021 U.S. Nuclear Regulatory Commission, Region III Attention:

Dell McNeil 2443 Warrenville Road Suite 210 Lisle, IL 60532-4352 Duane Arnold Energy Center Docket 50-331 License No. DPR-49 Examination Material for Duane Arnold Energy Center Initial License Examination Week of June 27,2011 In accordance with the guidelines of NUREG 1021, "Operating License Examination Standard for Power Reactors," Revision 9, we are sending you the integrated examination outlines for the initial license examinations to be administered at our facility the week of June 27,2011. NUREG 1021 physical security requirements state that the enclosed examination materials shall be withheld from public disclosure until after the examination is complete.

You may direct any questions or comments regarding this material to Curtis Hansen or Wayne Render at 319-851-7268.

This letter contains no new commitments.

Christopher R. Costanzo Vice President, Duane Arnold Energy Center NextEra Energy Duane Arnold, LLC Enclosure Chief Operator License Branch, Region III NRC Resident Inspector NRC Document Control Desk NextEra Energy Duane Arnold, LLC, 3277 DAEC Road, Palo, IA 52324 BLIND CARBON COpy LIST FOR April 7, D. Curtland w/o G. Pryw/o C. Hansen w/o W. Render w/o IRMS w/o S. Catron w/o W. Simmons w/o Licensing CTS Project w/o Examination Material for Duane Arnold Energy Center Initial License Examination Week of June 27,2011 FILE A117a Examination Outline Quality Checklist Facility:

DAEC Date of Examination:

6/2011 a. b. c . . d. 1 Item 1. W R I T T E N 2. S I M U L A T o R 3. W / T 4. G E N E R A L Author Task Description

a. Verify that the outline(s) fit(s) the appropriate model per ES-401. b. Assess whether the outline was systematically and randomly prepared in accordance with Section D.1 of ES-401 and whether all KiA categories are appropriately sampled. c. Assess whether the outline over-emphasizes any systems, evolutions, or generic topics. d. Assess whether the justifications for deselected or rejected KiA statements are appropriate.

Initials a b* c# a. Using Form ES-301-5, verify that the proposed scenario sets cover the required number of normal evolutions, instrument and component failures, technical specifications, and I U i major transients.

,;v-Assess whether there are enough scenario sets (and spares) to test the projected number and mix of applicants in accordance with the expected crew composition and rotation schedule without compromising exam integrity; and ensure that each applicant can be tested using at least one new or significantly modified scenario, that no scenarios are duplicated from the applicants' audit test(s), and scenarios will not be repeated on subsequent days. c. To the extent possible, assess whether the outline(s) conform(s) with the qualitative and ;1. 1lII..t..

I J A/ ..quantitative criteria specified on Form ES-301-4 and described in Appendix D. .II'"' a. Verify that systems walk-through outline meets the criteria specified on Form ES-301-2:

(1) the outline(s) contain(s) the required number of control room and in-plant tasks, distributed among the safety functions as specified on the form (2) task repetition from the last two NRC examinations is within the limits specified on the form, (3) no tasks are duplicated from the applicants' audit test(s) (4) the number of new or modified tasks meets or exceeds the minimums specified on the form (5) the number of alternate path, low-power, emergency, and RCA tasks meet the criteria on the form. b. Verify that the administrative outline meets the criteria specified on Form ES-301-1:

(1) the tasks are distributed among the topics as specified on the form (2) at least one task is new or significantly modified (3) no more than one task is repeated from the last two NRC licensing examinations

c. Determine if there are enough different outlines to test the projected number and mix of applicants and ensure that no items are duplicated on subsequent days. i a. Assess whether plant-specific priorities (including PRA and IPE insights) are covered in the appropriate exam section. b. Assess whether the 10CFR 55.41/43 and 55.45 sampling is appropriate.

c. Ensure that KiA importance ratings (except for plant-specific priorities) are at least 2.5. d. Check for duplication and overlap among exam sections.

e. Check the entire exam for balance of coverage.

f. Assess whether the exam fits the appropriate job level (RO or SRO). /) A Printed Name I Signature Facility Reviewer (*) NRC Supervisor NOTE: # Independent NRC reviewer initial items in Column "c', chief examiner concurrence required.

• Not applicable for NRC-prepared examination outlines Facility:

DAEC NRC Date of Examination:

6/2011 Examination Level (circle one): Operating Test Number: 1 Conduct of Operations N,R Administrative Topic (see Note) Type Code* Describe activity to be performed Conduct of Operations D, S Performance of Attachment

10. Volume Weighted Drywell Average Air Temperature, of STP 3.0.0-01.

Instrument Checks. The candidate will determine the weighted volume average air temperature is above 135°F. KIA: 2.1.18 (3.6) Ability to make accurate, clear and concise logs, records, status boards, and reports. Verification of License Requirements Given information related to maintenance of active license status for three operators, the candidate will determine which operator(s), if any, is(are) qualified to relieve the watch. (the candidate will be given the status of 3 operators in regard to last medical exam. hours worked in last quarter, SCBA fit test latest date etc. AND given the procedure that describes the requirements, determine if anyone meets eligibility requirements)

KIA: 2.1.4 (3.3) Knowledge of individual licensed operator responsibilities related to shift staffing.

such as medical requirements, "no-solo" operation, maintenance of active license status, 10CFR55. etc. The candidate will determine blocking points, tag types, and component position for a clearance on a RBCCW pump. KIA: 2.2.13 (4.3) Knowledge oftagging and clearance procedures Equipment Control M,R Radiation Control D,R Inspection Of High Radiation Areas The candidate will determine the expected exposure for the task and select the operator who can perform the task without exceeding DAEC limits. KIA 2.3.12 (3.7) Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities.

access to locked radiation areas, aligning filters. etc. All items: 5 total are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.

• Type Codes & (C)ontrol room, (S)imulator, or Class(R}oom (D)irect from bank (::::; 3 for ROs; s; 4 for SROs & RO retakes) (N)ew or (M)odified from bank (Ii!: 1) (P)revious 2 exams (::::; 1; randomly selected)

ES-301 Administrative Topics Outline Form ES-301-1 Facility:

DAEC NRC Date of Examination:

6/2011 Examination Level (circle one): Operating Test Number: 1 ROe Administrative Topic Type Code* Describe activity to be performed (see Note) Performance of Attachment 11, Core Thermal Limits Check, of Conduct of Operations M,R STP 3.0.0-01, Instrument Checks. The candidate will perform the Core Thermal Limits check, determine that 2 values are OOS and reference the appropriate TS and recommend actions as required.

KIA: 2.1.18 (3.6) Ability to make accurate, clear and concise logs, records, status boards, and reports. Verification of License Reguirements Conduct of Operations N,R Given information related to maintenance of active license status for three operators, the candidate will determine which operator(s), if any, is(are) qualified to relieve the watch. (the candidate will be given the status of 3 operators in regard to last medical exam, hours worked in last quarter, SCBA fit test latest date etc. AND given the procedure that describes the requirements, determine if anyone meets eligibility requirements)

KIA: 2.1.4 (3.8) Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, "no-solo" operation, maintenance of active license status, 10CFR55, etc. Review the isolation

(:2oints for RWS Ball "8" Diver

&"8" Equipment Control M,S Core S(:2rall Pum(:2 Motor Ins(:2egion by Electrical maintenance.

Evaluate TS im(:2lications.

The candidate will review the clearance requests for the work above and find an error on each one. Determines TS applicability.

KIA: 2.2.13 (4.3) Knowledge of tagging and clearance procedures Dose and EXQosure Authorization Review Radiation Control P, R The candidate will determine the dose for a given task the need for an Emergency Exposure Authorization.

Complete required paperwork KIA 2.3.4 (3.7) Knowledge of radiation exposure limits under normal or emergency conditions Emergencll ClassificationlReclassification Emergency Plan D,R Given a set of plant conditions determine the appropriate EAL classification.

Then, based on new information, reclassify the event. TIME CRITICAl.

KIA 2.4.41 (4.6) Knowledge of the emergency action level thresholds and classifications.

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

• Type Codes &Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (s 3 for ROs; s 4 for SROs &RO retakes) (N)ew or (M)odified from bank 1) (P)revious 2 exams (s 1; randomly selected)

Control Room/In-Plant Systems Outline Facility:

DAEC NRC Date of Examination:

6/2011 Exam LevelG SRO-I SRO-U Operating Test No.: 2011 NRC Control Room Systems@ (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U, including 1 ESF) System I ,IPM Title Type Code* Safety Function S-1 Reset a SCOOll Tube LoCkull on the A and 8 Recirc MG D,S Reactivity Sets Control The operators will reset the Recirc MG Set Scoop tubes lAW 01 264. (1 ) KIA 202002 A4.01 3.3/3. S-2 Swa(2 from the "8" FRV to the Startu(2 FRV. Failure of the M,A,L,S Reactor Water S/U FRV Inventory The operator will shift feedwater level control from the B FRV to the Control S/U FRV lAW 01 644. The S/U FRV will fail requiring the B FRV to be placed back in service. (2) KIA 259002 A4.03 3.8/3.6 S-3 Establish a Leakage Path to the Main Condenser lAW AOP D,S Radioactivity 672.2 Release The operator will establish a leakage path with the MSIVs closed (9) following a fuel failure. KIA 239003 A4.08 3.1/2.9 S-4 Manual Startu(2 Using The Test Pot To Control HPClln N,A,S Heat Removal pressure control mode From the The operator will place HPCI in service using the test pot and in Reactor Core the pressure control mode lAW 01 152, Section 5.1. During the evolution the min flow valve will fail to close. (4) KIA 206000 A4.01 3.8/3.7 S-5 Defeat Containment Atmosphere Monitoring Sample Line N,S Containment Isolation and Place HgO£ in Service Integrity The operator will open Drywell and Torus sample lines after a (5) PCIS 3 isolation and re-establish H2-0Z Analyzers.

KIA 223001 A4.04 3.5/3.6 S-6 Main Generator Synch to Grid M,A,S Electrical The operator will perform sections of 01 698 section 3.3 (6) complete thru step 2, to synch the turbine to the grid. While loading the generator a primary lockout will occur and the generator will fail to trip. The generator trip PB will also fail requiring the operator to manually open the outputs and exciter field breaker (1 C08C A-1) KIA 262001 A4.04 3.613.7 S-7 Perform Downscale/U(2scale Trill Operational Check of D,S Instrumentation ARMs (7) The operator will perform the Downscale/Upscale trip operational check of ARM RI-9167, Reactor Building Railroad Access Area lAW 01-879.2.

KIA 272001 A4.02 3.0/3.0)

Control Room/In-Plant Systems Outline Form ES-301-2 Facility:

DAEC NRC Date of Examination:

6/2011 SRO*I SRO-U Operating Test No.: 2011 NRC Control Room Systems@ (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U, including 1 ESF) System 1 JPM Title Type Code'" Safety Function S-8 Install EOP Defeat 4 wilh a GrouQ 7 Isolation D,A,EN,S Plant Service Install EOP Defeat 4 to restore Drywell cooling. (Alternate Path Systems requires Well Water to be secured prior to installing Defeat 4, (8)and then Well Water must be restarted).

KIA 400000 A4.01 3.1/3.0 P,A,R Plant Systems The operator will initiate Cable Spreading Room C02 Flood P-1 Manuall!llnitiate Cable SQreading Room CO,z Flood (8) System but the alternate means will be required.

KIA 286000 2.1.30 4.4/4.0 P-2 StartuQ the "An RPS generator set lAW 01358 Instrumentation The operator will perform the in-plant actions to start up M (7) the "An RPS MG set. EPA reset will be required KIA 2120002.1.20 4.6/4.6 P-3 Maximize CRD Injection l AW AlP 407 N,E,R Reactivity The operator will maximize the CRD System as a means Control of injecting water into the RPV when normal injection (1 ) systems are inadequate or unavailable.

KIA 201001 2.1.23 4.3/4.4 All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room. .. Type Codes Criteria for RO I SRO-II SRO-U {A)lternate path 4-6 14-6 12-3 (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant 2!1/2!1/2!1 (EN)gineering Safeguards Feature 2! 1/2! 1/2! 1 (control room system) (L)ow-Power 1 Shutdown 2!1/2!1/2!1 (N)ew or (M)odified from bank including 1(A) 2!2/2!2/2!1 (P)revious 2 exams 3 I 3 I 2 (randomly selected) (R)CA 2!1/2!1/2!1 lS)imulator NUREG-1021, Revision 9 ES-301 Control Roomlln-Plant Systems Outline Form ES-301-2 Facility: NRC Date of Examination:

6/2011 Exam Level: RO lSRO-Y SRO-U Operating Test No.: 2011 NRC Control Room for RO; 7 for SRO-I; 2 or 3 for SRO-U, including 1 ESF) System 1 JPM Title Type Code* Safety Function .

'!:i:,'::{

.. , ........ ....... ;.;;i .1.', *.. < . ,.. ..

.... ., '. S-2 Swa(2 from the "B" FRV to the Startu(2 FRV. Failure of the M,A, L, S StU FRV The operator will shift feedwater level control from the B FRV to the SlU FRV lAW 01 644. The SIIJ FRV will fail requiring the B FRV to be placed back in service. Reactor Water Inventory Control (2) KIA 259002 A4.03 3.813.6 S-3 Esti!blish a Leaki!ge Path to Main Condenser lAW AOP 672.2 The operator will establish a leakage path with the MSIVs closed D,S Radioactivity Release (9) following a fuel failure. KIA 239003 A4.08 3.112.9 S-4 Manui!1 StartuQ Using The Test Pot To Control HPCIIC gressure control mage The operator will place HPCI in service using the test pot and in the pressure control mode lAW 01152, Section 5.1. During the evolution the min flow valve will fail to close. KIA 206000 A4.01 3.813.7 S-5 Defea! Containment Atmosl2here Monitoring S!lIWl2le Line Isolation and Place HzOz in N,A,S Heat Removal From the Reactor Core (4) N,S Containment Integrity The operator will open Drywell and Torus sample lines after a PCIS 3 isolation and re-establish H2-02 Analyzers.

(5) S-6 KIA 223001 A4.04 3.5/3.6 Main Generator to Grid The operator will perform sections of 01 698 section 3.3 complete thru step 2, to synch the turbine to the grid. While loading the generator a primary lockout will occur and the generator will fail to trip. The generator trip PB will also fail requiring the operator to manually open the outputs and exciter field breaker (1C08C A-1) M,A,S Electrical (6) S-7 KIA 262001 M.D4 3.613.7 Perform DowDsCale/Ul2scale Trig Ogerational Check of ARMs The operator will perform the Downscale/Upscale trip operational check of ARM RI-9167. Reactor Building Railroad Access Area lAW 01-879.2.

D,S Instrumentation (7) KIA 272001 A4.02 3.0/3.0) NUREG-1021, Revision 9 Control Roomlln-Plant Systems Outline Form ES-301-2 Facility: NRC Date of Examination:

6/2011 Exam Level: RO SRO-U Operating Test No.: 2011 NRC Control Room RO; 7 for SRO-I; 2 or 3 for SRO-U, including 1 ESF) System / JPM Title Type Code* Safety Function S-8 Install EOP Defeat 4 with a Groul2 7 Isolation D,A,EN,S Plant Service Install EOP Defeat 4 to restore Drywell cooling. (Alternate Path Systems requires Well Water to be secured prior to installing Defeat 4, (8)and then Well Water must be restarted).

KIA 400000 A4.01 3.1/3.0 P,A,R Plant Systems The operator will initiate Cable Spreading Room C02 Flood P-1 Manuallx Initiate Cable Sl2reading Room CO 2 Flood Sxstem (8) System but the alternate means will be required.

KIA 286000 2.1.30 4.414.0 Instrumentation The operator will perform the in-plant actions to start up P-2 Startul2 the "A" RPS generator set lAW 01 358 M (7) the "A" RPS MG set. EPA reset will be required KIA 212000 2.1.20 4.6/4.6 Reactivity Control P-3 Maximize CRD lAW AlP 407 N,E,R The operator will maximize the CRD System as a means of injecting water into the RPV when normal injection (1 ) systems are inadequate or unavailable.

KIA 201001 2.1.234.3/4.4 All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

• Type Codes Criteria for RO 1 SRO-II SRO-U (A)lternate path 4-6 / 4-6/ 2-3 (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant ;:::1/;:::1/;:::1 (EN)gineering Safeguards Feature ;::: 11;::: 11;::: 1 (control room system) (L)ow-Power

/ Shutdown ;:::1/;:::1/;:::1 (N)ew or (M)odified from bank including 1(A) ;:::2/;:::2/;:::1 (P)revious 2 exams 3/ 3/:s; 2 (randomly selected)

{R)CA ;:::1/;::::1/;:::1 (S}imulator NUREG-1021 , Revision 9 Control Room/In-Plant Systems Outline NRC Date of Examination:

6/2011Exam Level: RO SRO-U Operating Test No.: 2011 NRC Control Room for RO; 7 for SRO-I; 2 or 3 for SRO-U, including 1 ESF) System / JPM Title Type Code* Safety Function -S-8 Install EOP Defeat 4 with a GrouQ 7 Isolation D,A,EN,S Plant Service Install EOP Defeat 4 to restore Drywell cooling. (Alternate Path Systems requires Well Water to be secured prior to installing Defeat 4, and then Well Water must be restarted).

(8) KIA 400000 M.01 3.1/3.0 D,A,R Plant Systems The operator will initiate Cable Spreading Room C02 Flood P-1 Initiate Cable Sr;!reading Room COg Flood (8) System but the alternate means will be required.

KIA 2860002.1.304.4/4.0 P-2 Startur;!

the "A" RPS generator set lAW 01358 M,A Instrumentation The operator will perform the in-plant actions to start up (7) the "A" RPS MG set. EPA reset will be required KIA 212000 2.1.20 4.6/4.6 P-3 Maximize CRD Injection lAW AlP 407 N,E, R Reactivity The operator will maximize the CRD System as a means Control of injecting water into the RPV when normal injection (1) systems are inadequate or unavailable.

KlA2010012.1.234.3/4.4 All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

• Type Codes Criteria for RO / SRO-II SRO-U (A)lternate path 4-6/4-6/2-3 (C)ontrol room (D)irect from bank 9/ 81 4 (E)mergency or abnormal in-plant 1 I 1 / 1 (EN)gineering Safeguards Feature 1/;>11 1 (control room system) (L)ow-Power

/ Shutdown 1 / 1 / 1 (N)ew or (M)odified from bank including 1 (A) 2/ 2/ 1 (P)revious 2 exams 31 31 2 (randomly selected) (R)CA 1 I 1 I 1 (S)imulator Facility:

DAEC NRC Date of Examination:

6/2011 Exam Level: RO SRO-I Operating Test No.: 2011 NRC Control Room Systems@ (8 for SRO-I; 2 or 3 for SRO-U, including 1 ESF) System I JPM Title $-1 S-2 SwaQ from the "8" FRV to the StartuQ FRV. Failure of the S/U FRV The operator will shift feedwater level control from the B FRV to the S/U FRV lAW 01 644. The SIU FRV will fail requiring the B FRV to be placed back in service. KIA 259002 A4.03 3.8/3.6 S-3 Establish a Leakage Path to the Main Condenser lAW AOP 672.2 The operator will establish a leakage path with the MSIVs closed following a fuel failure. KIA 239003 A4.08 3.1/2.9 S-4 ; S-5 '7 .. S-8 Install EOP Defeat 4 with a GrouQ 7 Isolation Install EOP Defeat 4 to restore Drywell cooling. (Alternate Path requires Well Water to be secured prior to installing Defeat 4, and then Well Water must be restarted).

KIA 400000 A4.01 3.1/3.0 P-1 P-2 , StartuQ the "A" RPS generator set lAW 01 358 The operator will perform the in-plant actions to start up the "An RPS MG set. EPA reset will be required KIA 212000 2.1.20 4.6/4.6 P-3 Maximize CRD Injection lAW AlP 407 The operator will maximize the CRD System as a means of injecting water into the RPV when normal injection systems are inadequate or unavailable.

KIA 201001 2.1.234.3/4.4 Type Code* Safety Function M,A, L, S Reactor Water Inventory Control (2) D,S Radioactivity Release (9) ; D,A,EN,S Plant Service Systems (8) M,A Instrumentation (7) N,E,R Reactivity Control (1) @ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

• Type Codes (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineering Safeguards Feature (L)ow-Power

/ Shutdown (N)ew or (M)odified from bank including 1 (A) (P)revious 2 exams (R)CA (S)imulator s; 9 / 81 4 ?: 1 I 11 1 1 1 1 / 1 (control room system) 2' 11 1 / 1 :' 2/ 21 1 31 3/ 2 (randomly selected) 11 1 I 1 I Criteria for RO I SRO-II SRO-U I 4-6 1 4-6 1 2-3 Date of Exam: June 2011 Operating Test No.: 1 A E Scenarios P V 1 2 3 4 T MP E L N 0 I I T T N CREW CREW POSITION CREW POSITION CREW I C POSITION POSITION A M A T L U N Y S A B S A B S A B S A B T P R T 0 R T 0 R T 0 R T 0 M(*) E 0 C P 0 C P 0 C P 0 C P R RX -1 1 NOR 1 1 -2 1 1 1 RO 1 IIC 3,5,7 2,4 4,5 7 4 4 2 MAJ 6 6 6 I I 13 2 2 1 TS ---0 0 RX 1 1 NOR 1

• 1 2 1 1 1 R02 I 4,5 I/C 3,5,7 5,6,9 8 4 4 2 MAJ 6 6 8 3 2 2 1 TS ---0 0 2 2 . 2 1 1 1 0 NOR 1

• 1 1 1 1 R03 I/C 3,5,7 4,5 5 4 4 2 MAJ 6 6 2 2 2 1 TS -* 0 0 2 2 RX 2 . 1 1 1 0 NOR 1 1 2 1 1 1 R04 IIC 4,5,7 2,4 3,5,7,8 9 4 4 2 MAJ 6 6 6 3 2 2 1 TS --0 0 2 2 RX -3 1 1 1 0 NOR 1 1 I -2 1 1 1 RO*5 I/C 14,6,7 2,4 3,5,7,8 4 4 2 MAJ 6 6 8 3 2 2 1 TS ---0 0 2 2 RX 3 2 2 1 1 0 NOR 1 1 2 1 1 1 SROU 1 IIC 2,4, 5,7 9 4 4 2 MAJ 6 2 2 2 1 TS 2,4 I 4,5 14 0 2 2 ES-301 Transient and Event Checklist Form ES*301*5 Date of Exam: June 2011 Operating Test No.: 1 A E Scenarios p v 1 2 3 4 T M p E L N CREW 0 I CREW CREW CREW N I T POSITION POSITION POSITION POSITION T I C A M A T S A B S A B S A B S A B L N Y R T 0 R T 0 R T 0 R T 0 U T P 0 C P 0 C P 0 C P 0 C p M(*) E R I U RX 2 2 2 1 1 0 NOR 1 1 2 1 1 1 SROU-2 IIC 3,4,5, 3,4,5, 9 4 4 2 7 7,8 MAJ 6 6 2 2 2 1 TS 3,5 4,5 4 0 2 2 RX 2 3 -2 1 SROI-1 NOR 1 1 2 1 IIC 3,4,5, 5,7 3,5,7, 10 4 4 2 7 8 MAJ 6 6 6 3 2 2 1 TS 3,5 -2 0 2 RX 2 3 3 3 1 1 NOR 1 -1 2 1 1 1 SROt-2 IIC 3,4,5, 4,5,6, 7 5,7 7,9 11 4 4 2 MAJ 6 6 8 3 2 2 1 TS 3,5 -2,4,6 5 0 2 2 RX 3 2 2 1 1 0 NOR -1 1 1 1 1 SROI-3 IIC 3,4,5, 5,7 7,8 7 4 4 2 MAJ .6 6 2 2 2 1 TS -4,5 2 0 2 2 RX 2 3 2 1 1 0 NOR -1 1 1 1 1 SROI-4 IIC 2,4,5, 4,5,7 7 7 4 4 2 MAJ 6 6 2 2 2 1 TS -2,4 2 RX 2 3 0 NOR -1 11 1 1 1 SROI-5 IIC 4,5,7 2,4,5, 7 MAJ 6 6 TS 2,4 Facility: Date of Exam: June 2011 Operating Test No.: 1 Instructions: Check the applicant level and enter the operating test number and Form ES-D-1 event numbers for each event type; TS are not applicable for RO applicants.

ROs must serve in both the "at-the-controls (ATC)" and "balance-of-plant (BOP)" positions; Instant SROs must serve in both the SRO and the ATC positions, including at least two instrument or component (l/C) malfunctions and one major transient, in the ATC position.

If an Instant SRO additionally serves in the BOP position, one IIC malfunction can be credited toward the two I/C malfunctions required for the ATC position. Reactivity manipulations may be conducted under normal or control/ed abnormal conditions (refer to Section D.5.d) but must be significant per Section C.2.a of Appendix D. (*) Reactivity and normal evolutions may be replaced with additional instrument or component malfunctions on a 1-for-1 basis. Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicant's competence count toward the minimum requirements specified for the applicant's license level in the right-hand columns.

Facility:

DAEC 2011 NRC Date of Exam: 5/2011 Tier Group RO KIA Category Points SRO-Only Points K 1 K 2 K 3 K 4 K 5 K 6 A 1 A 2 A 3 A 4 G

• Total A2 G* Total 1. Emergency

& Abnormal Plant Evolutions 1 5 3 3 (}':: f," ,i.. .. ,: . 3 3 , I ..... 3 20 3 4 7 2 1 1 1 {,' I:>: ', .ii. fl*".,. 1:'5. 2 1 , 1 7 2 1 3 Tier Totals 6 4 4 .,t

t,* ....,i!*

I"* .{ 5 4 4 27 5 5 10 2. Plant Systems 1 3 3 2 2 2 2 2 2 3 3 2 26 2 3 5 2 1 1 1 1 2 1 1 1 1 1 1 12 0 2 1 3 Tier Totals 4 4 3 3 4 3 3 3 4 4 3 38 4 4 8 3. Generic Knowledge

&Abilities Categories 1 2 3 4 10 1 2 3 4 7 3 2 2 3 2 2 1 2 Note 1. Ensure that at least two topics from every applicable KiA category are sampled within each tier of the RO and SRO-only outlines (Le., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each KiA category shall not be less than two). 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 1 from that specified in the table based on NRC revisions.

The final RO exam must total 75 pOints and the SRO-only exam must total 25 points. 3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems that are not included on the outline should be added. Refer to section D.1.b of ES-401, for guidance regarding elimination of inappropriate KiA statements.

4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.

5. Absent a plant specific priority, only those KAs having an importance rating (IR) of 2.5 or higher shall be selected.

Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

6. Select SRO topics for Tiers 1 and 2 from the shaded systems and KiA categories.

7.* The generic (G) KiAs in Tiers 1 and 2 shall be selected from Section 2 of the KiA Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable KiA's B. On the following pages, enter the KiA 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 category.

Enter the group and tier totals for each category in the table above. If fuel handling equipment is sampled in other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams. 9. For Tier 3, select topics from Section 2 of the KiA Catalog, and enter the KiA numbers, descriptions, IRs, and point totals (#) on Form ES-401-3.

Limit SRO selections to KiAs that are linked to 10CFR55.43 BWR Examination Emergency and Abnormal Plant Evolutions

-Tier 1 Group 1 EAPE#/Name Safety Function 295030 Low Suppression Pool Water Levell 5 295001 Partial or Complete Loss of Forced Core Flow Circulation

/1 &4 295018 Partial or Total Loss of CCW / 8 295025 High Reactor Pressure 13 295019 Partial or Total Loss of Inst. Air 18 295023 Refueling Accidents / 8 295021 Loss of Shutdown Cooling 14 KIA Topic(s) EA2.01 -Ability to determine!

and/or interpret the following

• as they apply to LOW SUPPRESSION POOL 4.2 WATER LEVEL: Su ression 001 level M2.06 -Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION:

Nuclear boiler instrumentation M2.01 -Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF 3.4 COMPONENT COOLING WATER: Component tern eratures 2.4.45 -Emergency Procedures

/ Plan: Ability to prioritize and interpret the 4.3 significance of each annunciator or alarm. 2.4.35 -Emergency Procedures I Plan: Knowledge of local auxiliary operator tasks during 4.0 emergency and the resultant o erational effects. 2.2.22 -Equipment Control: Knowledge of limiting conditions for operations and 4.7 safet limits. 2.4.9 -Emergency Procedures

/ Plan: Knowledge of low power I shutdown implications in accident (e.g., loss of 4.2 coolant accident or loss of residual heat removal) miti ation strate ies. 76 78 79 80 81 82 BWR Examination Emergency and Abnormal Plant Evolutions

-Tier 1 Group 1 e Safety Function 295025 High Reactor Pressure 13 295024 High Drywell Pressure 15 295005 Main Turbine Generator Trip I 3 295003 Partial or Complete Loss of AC 16 295001 Partial or Complete Loss of Forced Core Flow Circulation 11 & 4 600000 Plant Fire On-site 1 8 x x x x x x KIA Topic(s) AK1.01 -Knowledge of the operational implications of the following concepts as they apply to REFUELING ACCIDENTS:

Radiation ex osure hazards EK1.06 -Knowledge of the operational implications of the following concepts as they apply to HIGH REACTOR PRESSURE:

Pressure effects on reactor water level EK1.01 -Knowledge of the operational implications of the following concepts as they apply to HIGH DRYWELL PRESSURE:

Drywell integrity: S ecific AK2.05 -Knowledge of the interrelations between MAIN TURBINE GENERATOR TRIP and the following:

Extraction steam s stem AK2.04 -Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF AC. POWER and the following:

A.C. electrical loads AK2.03 -Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION and the following:

Reactor water level AK3.04 -Knowledge of the reasons for the following responses as they apply to PLANT FIRE ON SITE: Actions contained in the abnormal procedure for plant fire on site 3.6 39 3.5 40 4.1 41 2.6 42 3.4 43 3.6 44 2.8 45 BWR Examination Emergency and Abnormal Plant Evolutions

-Tier 1 Group 1 Ir EAPE#/Name Safety Function I K1 I K2 I K3J A1 I A2 I G I KIA Topic(s) limp. I295037 SCRAM Conditions Present and Reactor Power. Above APRM Downscale or* Unknown 11 295019 Partial or Total Loss of Inst. Air I 8 295018 Partial or Total Loss of CCW 1 8 700000 Generator Voltage and Electric Grid Disturbances 295006 SCRAM /1 295030 Low Suppression Pool Water Level/ 5 295021 Loss of Shutdown Cooling 14 x x EK3.05 -Knowledge of the reasons for the following responses as they apply to SCRAM CONDITION PRESENT AND REACTOR 3.2 46 POWER ABOVE APRM DOWNSCALE OR UNKNOWN: Cold shutdown boron wei ht: Plant-S ecific AK3.02 -Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE 3.5 47 LOSS OF INSTRUMENT AIR: Standby air compressor o eration AA 1.02 -Ability to operate and/or monitor the following as they apply to PARTIAL 3.3 48 OR COMPLETE LOSS OF COMPONENT COOLING WATER: S stem loads AA 1 .04 -Ability to operate and/or monitor the following as they apply to GENERATOR VOLTAGE 4.1 49 AND ELECTRIC GRID DISTURBANCES:

Reactor controls.

AA 1.02 -Ability to operate and/or monitor the following as they apply to SCRAM: 3.9 50 Reactor water level control s stem EA2.02 -Ability to determine and/or interpret the following as they apply to LOW SUPPRESSION POOL 3.9 51 WATER LEVEL: Suppression pool tem erature AA2.03 -Ability to determine and/or interpret the following as they apply to LOSS OF 3.5 52 SHUTDOWN COOLING: Reactor water level BWR Examination Outline Emergency and Abnormal Plant Evolutions

-Tier 1 Group 1 (RO/SRO) EAPE#lName Safety Function 295038 High Off-site Release Rate I 9 KIA Topic(s) EA2.04 -Ability to determine andlor interpret the following as they apply to HIGH SITE RELEASE RATE: Source of off-site release EK1.03 -Knowledge of the operational implications of the following concepts as 295031 Reactor Low Water Level 12 X they apply to REACTOR 3.7 54 295026 Suppression Pool High Water Temp. 15 295004 Partial or Total Loss of DC Pwr 16 LOW WATER LEVEL: Water level effects on reactor ower 2.1.28 -Conduct of Operations:

Knowledge of the purpose and function of major system components and controls.

2.4.31 -Emergency Procedures I Plan: Knowledge of annunciator alarms, indications, or onse rocedures.

2.1.2 -Knowledge of 295016 Control Room bandonment I 7 295028 High Drywell Temperature 15 KIA Category Totals X 5 3 3 operator responsibilities 1',,<1 during all modes of plant o eration. EK1.02 -Knowledge of the operational implications of the following concepts as they apply to HIGH DRYWELL TEMPERATURE:

Equipment environmental qualification Group Point Total: 4.1 55 4.2 56 4.1 57 2.9 58 2017 BWR Examination Outline Emergency and Abnormal Plant Evolutions

-Tier 1 Group 2 (RO/SRO) EAPE#/Name Safety Function I K1 I K2 I K3 I A1 I A2 I G I KIA Topic(s) limp. I Q# AA2.04 -Ability to determine and/or interpret the following 295020 Inadvertent Cont. Isolation I 5 &7 as they apply to INADVERTENT CONTAINMENT ISOLATION:

Reactor 3.9 83 ressure 295017 High Off-site Release Rate / 9 2.1.19 -Conduct of Operations:

Ability to use plant computers to evaluate 3.8 84 s stem or com onent status. AA2.04 -Ability to determine and/or interpret the following 295002 Loss of Main Condenser Vac / 3 as they apply to LOSS OF MAIN CONDENSER VACUUM: Offgas system flow 2.9 85 EK1.02 -Knowledge of the operational implications of 295035 Secondary Containment High Differential Pressure / 5 x the following concepts as they apply to SECONDARY CONTAINMENT HIGH DIFFERENTIAL PRESSURE:

Radiation release 3.7 59 EK2.03 -Knowledge of the 500000 High CTMT Hydrogen Conc. I 5 x interrelations between HIGH. CONTAINMENT HYDROGEN CONCENTRATIONS the following:

Containment Atmos here Control S stem 3.3 .60 .295007 High Reactor

• Pressure / 3 295013 High Suppression Pool Temperature / 5 x AK3.05 -Knowledge of the reasons for the following responses as they apply to HIGH REACTOR PRESSURE:

Low pressure s stem isolation AA 1.01 -Ability to operate and/or monitor the following as they apply to HIGH SUPPRESSION POOL 3.0 61 3.9 62 TEMPERATURE:

BWR Examination Emergency and Abnormal Plant Evolutions

-Tier 1 Group 2 PE#/Name Safety Function .295036 Secondary Containment High Sump/Area Water Level/ 5 295022 Loss of Control Rod Drive Pumps /1 295008 High Reactor Water Level/2 KIA Category Totals KIA Topic(s) EA2.03 -Ability to determine and/or interpret the following as they apply to SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL: Cause of the high water level 2.1.20 -Conduct of Operations:

Ability to interpret and execute rocedure ste s. AA 1.03 -Ability to operate and/or monitor the following as they apply to HIGH REACTOR WATER LEVEL: 4.6 3.3 64 65 Group Point Total:

System #INa me KiA Topic(s) A2,03 -Ability to (a) predict the impacts of the following on the AUTOMATIC DEPRESSURIZATION 218000 ADS SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or 3,6 86 operations:

Loss of air supply to ADS valves: Plant-S ecific . A2,06 -Ability to (a) predict the impacts of the following on the REACTOR WATER LEVEL CONTROL 259002 Reactor Water Level Control SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or 3.4 87 operations:

Loss of controller signal out ut 206000 HPCI 2.4.18 -Emergency Procedures I Plan: Knowledge of the s ecific bases for EOPs, 4,0 88 262001 AC Electrical Distribution 2.4.50 -Emergency Procedures I Plan: Ability to verify system alarm setpoints and operate controls identified in the 4,0 89 . alarm res onse manual. 2,1 ,7 -Conduct of 300000 Instrument Air Operations:

Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor 4,7 ! 90 behavior, and instrument inter retation, BWR Examination Outline 1i==_____---r---r--rP_la.,..nt_S..,.:y;...s.,..te_ms

-Tier 2 Gr_0=Fup,====1

______...,...._,.-..., System #JName KIA Topic(s) 206000 HPCI x j 211000 SLC x 400000 Component Cooling Water 300000 Instrument Air x x 215005 APRM / LPRM .x K1.10 -Knowledge of the physical connections and/or cause-effect relationships between HIGH PRESSURE 3.4 1 COOLANT SYSTEM and following:

storage and s stem:

K1.01 -Knowledge the physical and/or cause-relationships STANDBY 3.0 2 . CONTROL SYSTEM and the following:

Core spray line break detection:

Plant-S ecific K2.02 -Knowledge of electrical power supplies 2.9 3 to the following: K2.02 -Knowledge electrical power to the 3.0 4 Emergency com K3.08 -Knowledge the effect that a loss malfunction of AVERAGE 3.0 5 POWER MONITOR SYSTEM will have on following:

core thermal calculations BWR Examination Plant Systems -Tier 2 Group 1 IF-============r=r=

System #lName KJA Topic(s) K3.10 -Knowledge of the effect that a loss or malfunction of the REACTOR PROTECTION SYSTEM 212000 RPS x will have on following:

The ability of the core cooling systems to provide adequate core cooling during loss of coolant accidents K4.04 -Knowledge of REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) design 217000 RCIC x feature(s) and/or interlocks which provide for the following:

Prevents turbine dama e: Plant-S ecific K4.03 -Knowledge of SHUTDOWN COOLING SYSTEM (RHR 205000 Shutdown Cooling IX SHUTDOWN COOLING MODE) design feature(s) and/or interlocks which provide for the following:

Low reactor water level: Plant-S ecific K5.01 -Knowledge of the operational implications of the following concepts as 218000 ADS X they apply to AUTOMATIC DEPRESSURIZATION SYSTEM: ADS logic o eration K5.02 -Knowledge of the operational implications of the 203000 RHRlLPCI:

following concepts as X I Injection Mode they apply to RHRlLPCI:

INJECTION MODE (PLANT SPECIFIC):

Core coolin methods 3.5 6 3.0 .7 3.8 8 3.8 9 3.5 10 BWR Examination Outline Grou System #/Name X 209001 LPCS 261000 SGTS 2150031RM 223002 PCIS/Nuclear Steam Supply Shutoff KIA Topic(s) K6.11 -Knowledge of the effect that a loss or malfunction of the following will have on the LOW PRESSURE CORE SPRAY . SYSTEM: ADS K6.03 -Knowledge of the effect that a loss or malfunction of the following will have on the STANDBY GAS TREATMENT SYSTEM: Emergency diesel enerator s stem A1.02 -Ability to predict and/or monitor changes in parameters associated with operating the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM controls including:

Reactor power indication response to rod osition chan es A 1.03 -Ability to predict and/or monitor changes in parameters associated with operating the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPL Y OFF controls including:

SPDS/ERIS/CRI DS/GDS : Plant-S ecific 11 '3.6 12 3.0 3.7 13 2.5 14 BWR Examination 11=-______r==r==P,...la_n,...t_S,.:;.y_st,...e_m,...s_-

Tier 2 System #lName 239002 SRVs 259002 Reactor Water Level Control 264000 EDGs 262002 UPS (AC/DC) 262001 AC Electrical Distribution KIA Topic(s) A2.01 -Ability to (a) predict the impacts of the following on the RELIEF/SAFETY VALVES; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Stuck open vacuum breakers A2.05 -Ability to (a) predict the impacts of the following on the REACTOR 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:

Loss of applicable plant air s stems A3.06 -Ability to monitor automatic operations of the EMERGENCY GENERATORS (DIESEL/JET) including:

Cooling water system eration . A3.01 -Ability to monitor automatic operations of the UNINTERRUPTABLE POWER SUPPLY (A.C.lD.C.)

including:

Transfer from preferred to alternate source A4.04 -Ability to manually operate and/or . monitor in the control room: Synchronizing and paralleling of different A.C. su lies 3.0 15 3.2 16 3.1 17 2.8 18 3.6 19 BWR Examination Outline Plant Systems -Tier 2 Group 1 (RO/SRO)

System #lName KIA Topic(s)

I e' A4.06 -Ability to 215004 Source X I' manually operate and/or 3.2 20 Range Monitor i' monitor in the control i . ,'. ;c ***. room: Alarms and lights I I',* ;i'" I ". 2.2.22 -Equipment 263000 DC t;:,;* ;. Control: Knowledge of Electrical

.;, .......,... X limiting conditions for 4.0 21 Distribution l'rt}t' ". operations and safety I to' limits. I'"' "'. .; 2.1.30 -Conduct of ,'. Operations:

Ability to 215005 APRM I t X LPRM iii ., locate and operate 4.4 22 l*.. .'** components, including iJ:' ;;.' local controls.

_. ; .,*... K 1.0 1 -Knowledge of. " Ii ". the physical connections ft.;. i and/or cause-effect Ie' I* relationships between 218000 ADS X I' ***.. AUTOMATIC 4.0 23 \ .....'.. , DEPRESSURIZATION

H*;*..* SYSTEM and the , .;, following

RHRlLPCI:

> '.' '. Plant-Specific i *fj>* <<"'. A3.05 -Ability to monitor f: ,..*..* ..... '*r automatic operations of I' . the REACTOR 212000 RPS 'J,'; X .... ,. 3.9 24

'. ;..... PROTECTION SYSTEM including:

SCRAM

.... , "'. instrument volume level ,;" c' ..... A4.02 -Ability to 1209001 LPCS X ........ manually operate and/or 3.5 25; , monitor in the control '.' .', . room: Suction valves :'.' ... :'; .' K2.03 -Knowledge of 203000 RHRlLPCI: .i*; / ; electrical power supplies Injection Mode X ;X"Jr , to the following:

Initiation 2.7 26 ;", t**) ",'" logic KIA Category Totals 3 3 2 2 2 2 2 3 3'!¥"'J '." Group Point Total: I 26/5 BWR Examination Outline

____-r==;==Plant Systems -Tier 2 Group 2 (R_O_'_S_R_O..;.}

_____-r-_,......-ll KIA Topic{s) A2.06 -Ability to (a) predict the impacts of the following on the RECIRCULATION FLOW CONTROL 202002 Recirculation Flow Control SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Low reactor water level: Plant-S ecific 2.1.30 -Conduct of Operations:

Ability to 204000 RWCU locate and operate components, including local controls.

A2.05 -Ability to (a) predict the impacts of the following on the ROD WORTH MINIMIZER SYSTEM (RWH) (PLANT SPECIFIC);

and (b) based on those 201006 RWM predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Out of sequence rod , movement; BWR6 K1.04 -Knowledge of the physical connections and/or cause-effect relationships between 201002 RMCS x REACTOR MANUAL CONTROL SYSTEM and the following:

Rod block monitor: Plant-S ecific 3.3 4.0 3.5 91 92 93 3.5 27 256000 Reactor Condensate 290003 Control Room HVAC 233000 Fuel Pool Cooling/Cleanup 201006 RWM i 245000 Main Turbine Gen. / Aux. 1202001 Recirculation BWR Examination Outline Plant Systems -Tier 2 Group 2 (RO/SRO) x x X. x x limp. I Q# K2.01 -Knowledge of electrical power supplies 2.7 28 to the following:

System um s K3.01 -Knowledge of the effect that a loss or malfunction of the CONTROL ROOM HVAC will have on following:

Control room habitabilit K4.06 -Knowledge of FUEL POOL COOLING AND CLEAN-UP design feature(s) and/or interlocks which provide for the following:

Maintenance of ade uate 001 level K5.01 -Knowledge of ROD WORTH MINIMIZER SYSTEM (RWM) (PLANT SPECIFIC) design feature(s) and/or interlocks which provide for the following:

Minimize clad damage if a control rod drop accident (CRDA) occurs: P-S ec Not-BWR6 K6.10 -Knowledge of the effect that a loss or malfunction of the following will have on the MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS: 3.5 29 2.9 30 3.3 31 2.8

• 32 3.4 33 BWR Examination Outline Plant Systems -Tier 2 Group 2 (RO/SRO) KJA Topic(s) [.'. A2.15 -Ability to predict '. the impacts of the f following on the I.: . .'. RADIATION 1\; MONITORING SYSTEM; , and (b) based on those 272000 Radiation 2.5 34 predictions, use Monitoring procedures to correct, control, or mitigate the ... consequences of those .'. abnormal conditions or operations:

Maintenance operations A3.04 -Ability to monitor automatic operations of 286000 Fire 3.2 35 the FIRE PROTECTION Protection I;. SYSTEM including:

System initiation A4.01 -Ability to manually operate and/or X .'monitor in the control 215002 RBM 2.8 36 room: IRM/RBM recorder/switch: 3,4,5 r. 2.4.21 -Emergency

[1Procedures

/ Plan: Knowledge of the parameters and logic ..' used to assess the < status of safety " .functions, such as 268000 Radwaste 4.0 37reactivity control, core "'\ ,. cooling and heat <:j, .. ." removal, reactor coolant ;:,,' system integrity, ...*.....

". 'is . containment conditions, I;. '. radioactivity release l' . . control, etc.

• • • • • • BWR Examination Plant Systems -Tier 2 Group 2 System limp. I Q# 216000 Nuclear Boiler Inst. KIA Category Totals 1 1 1 1 X 2 1 I*::':" I;:. Ii .

r* ..... fro 1 1

.*. . ....... I* I< : .." '." ... 1 111 KS.10 -Knowledge of the operational implications of the following concepts as they apply to NUCLEAR BOILER 3.1 38 INSTRUMENTATION:

Indicated level versus actual vessel level during vessel heatups or cooldowns Group Point Total: I 12/3 Facility:

Duane Arnold 2010 Audit Date: 6/2011 Category KA# Topic 2.1.1 Knowledge of conduct of operations requirements Knowledge of industrial safety procedures (such as rotating 2.1.26 equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and hydrogen).

1. Conduct of 2.1.40 Knowledge of refueling administrative Operations requirements 2.1.35 Knowledge of the fuel-handling responsibilities of SRO's. 2.1.36 Knowledge of procedures and limitations involved in core alterations.

Subtotal Knowledge of less than or equal to one 2.2.39 hour4.513889e-4 days <br />0.0108 hours <br />6.448413e-5 weeks <br />1.48395e-5 months <br /> technical specification action statements for systems. Knowledge of the process for 2.2.14 controlling equipment configuration or status. 2. Equipment Control Knowledge of the process for making 2.2.5 design or operating changes to the facility.

2.2.19 dge of maintenance work order rements. Subtotal RO IR Q# 3.8 66 3.4 67 2.8 75 3 3.9 68 3.9 69 2 SRO-Only IR Q# 3.9 94 : 4.1 99 2 3.2 95 3.4 100 2 Ability to comply with radiation work 2.3.7 permit requirements during normal or 3.5 70 abnormal conditions.

2.3.4 Knowledge

of radiation exposure limits 3.2 ! 71 under normal or emergency conditions.

3. Radiation Control 2.3.11 Ability to control radiation releases.

3.8 96 Subtotal 2 1 Knowledge of procedures relating to a 2.4.28 security event (non-safeguards 3.2 72 information).

Knowledge of EOP implementation hierarchy and coordination with other 2.4.16 I support procedures or guidelines such 3.5 73 as, operating procedures, AOP's and 4. Emergency SAMG's. Knowledge of how abnormal operating I Procedures I 2.4.8 3.8 74 Plan procedures are used in ! conjunction with EOPs 2.4.29 Knowledge of the emergency plan. 4.4 97 2.4.46 Ability to verify that the alarms are 4.2 98 consistent with the plant conditions.

Subtotal 3 2 3 Point Total: 10 7 Based on the validation week and the number of candidates, we did not use Scenarios 1 and 2 for the exam. In addition, the licensee requested that we avoid putting these in ADAMS so that they could still be used for exam material.

Scenario Event DAEC 2011 NRC Scenario Facility: Scenario No.: 3 Op Test No.: 2011 NRC Examiners: BOP-Initial Conditions: 100% Reactor Power RCIC TS LCO 3.5.3A -Day 1 of 14 day LCO

• RCIC was operating in CST-CST for baseline vibration testing. An Engineer accidently bumped the mechanical overspeed lever and RCIC tripped. A visual inspection determined no damage to RCIC and the trip is ready to be reset and placed in standby lAW 01-150, Section 3.3. The 2 nd Assistant and the System Engineer are standing by in the RCIC room to observe the reset. Once RCIC is back in Standby, continue the to raise power with Control Rods. RE direction is to pull two steps of rods, then wait for conditioning.

Critical 1. With a* Primary System discharging into secondary containment, insert a manual scram before any parameter reaches the Max Safe Operating Limit. 2. With a Primary System discharging into secondary containment and the same parameter exceeding the Max Safe Operating Limit in more than one area perform an Emergency Depressurization.

Event No. 1 2 Malf. No. N/A N/A 3 SW21D 4 RR17A 5 RP02B 6 HP05.08 ED06C 7 STRC01 8 Override DI-AD-19 Event Type* C-RO C-SRO TS-SRO C-ALL TS-SRO M-ALL C-BOP C-SRO C-BOP C-SRO Unisolate and reset RCIC Raise reactor power with control rods. D Well Water Pump trip AOP 408 liB" Reactor Recirc Pump speed controller fails downscale AOP 255.2 TS 3.4.1 RPS "A" EPA Breaker Trip AOP 358 TS 3.3.8.2., TS 3.6.1.3 Unisolable HPCI leak in secondary containment, Startup XFMR Lockout EOP-1, EOP-2.

EOP-ED RCIC Fails to auto start "One ADS SRV fails to open * (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M}ajor Scenario Event DAEC 2011 NRC Scenario DAEC 2011 NRC Scenario #3 The scenario begins with reactor power at 90%. RCIC was operating in CST for baseline vibration testing. An Engineer accidently bumped the mechanical overspeed lever and RCIC tripped. A visual inspection determined no damage to RCIC and the trip is ready to be reset and placed in standby lAW 150, Section 3.3. Once RCIC is placed in standby, the crew will continue to raise reactor power with control rods. Then the "8" Recirc Pump flow controller will fail downscale and the operators will lock up the scoop tube lAW the AOP. The SRO will address Technical Specifications (TS) for the speed mismatch.

An EPA breaker will trip on "A" RPS resulting in Y2 scram. The crew will enter the AOP, transfer RPS to the alternate supply and the Y2 scram will be reset. The SRO will address TS for the EPA trip. The Main Turbine will experience high vibration and the crew will be required to lower reactor power. This results in lowering vibration on the Main Turbine. A leak will develop on the HPCI steam line in Secondary Containment and will not be isolable.

Area temperatures will rise requiring EOP entry and a reactor scram (CRITICAL TASK 1.0). Additionally, as RPV level lowers RCIC will fail to initiate at its auto setpoint and must be placed in service manually for RPV level control. As area temperatures continue to rise, an Emergency Depressurization(ED) must be performed when the max safe operating limit is exceeded in more than one area (CRITICAL TASK 2.0). During the ED, one ADS SRV will fail to open and another SRV must be opened. The scenario ends with the ED completed and RPV level at >170 inches.

Scenario Event DAEC 2011 NRC Scenario Facility: Scenario No.: 4 Op Test No.: 2011 NRC Examiners: BOP-Initial Conditions:

• Plant shutdown for refueling outage in progress

• Currently in IPOI-3, Section 5.0 Step (8)

• Step 31 of the Pull Sheet, 2 Rods at position 30

• Reactor Power 59% Turnover:

• Perform STP 3.3.1.1-17

-MSIV Function Test

• Continue lowering power as directed with control rods

• Secure the second Feedwater and Condensate pumps at 50% power Critical 1. Following' a LOOP, manually start the "B" SBDG to re-power the Essential Bus. 2. Recover RPV level prior to an Emergency Depressurization required.

Event No. Malf. No. 1 N/A 2 Override DI-MS-058 3 N/A 4 NM08C 5 Overrides An1c03b(2}

6 ED08A 7 8 9 FW02B RP05A OVERRIDE S ED01A,B,C RR15B DG02A STDG02 RC05 HP03 Event Type* N-BOP N-SRO TS-SRO R-RO R-SRO C-.RO C-SRO TS C-BOP C-SRO C-ALL TS C-RO C-SRO M -ALL C-BOP C-SRO MSIV Trip/Close Functional test MSIV fails closed TS 3.6.1.3 Lower power with Control Rods "C" APRM Upscale, Y:z Scram Briefly in TS 3,3.1.1 "A" RHR pump start, fails to trip on overcurrent Loss of Bus 1A1 AOP304.1, AOP 264 TS -3.4.1 Loss of Condensate/Feedwater-Reactor Scram -RPS PB Failure IPOI-5, EOP-1 Loss of Offsite Power, "A" SBDG Output breaker will not close, "B" SBDG fails to Auto-Start, Small Recirc Break EOP-1, EOP-2 RCIC Fails, HPCI Controller fails in auto * (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Scenario Event DAEC 2011 NRC Scenario OAEC 2011 NRC Scenario #4 The scenario begins with the reactor at 60% power and a shutdown in progress. The crew will perform the MSIV Trip/Close Functional Test. During the test, one MSIV will fail closed. The SRO will address Technical Specifications (TS) for the valve failure. The crew will continue the shutdown by inserting control rods. While moving control rods, an APRM will fail upscale and the crew will take procedural actions to bypass the APRM and reset the Y2 scram. Then, a RHR pump will spuriously start and its minimum flow valve will fail to open. The SRO will address TS and direct removing the pump from service. Once TS are addressed, a loss of Non Essential 4160v Bus 1A 1 will occur. The crew will take actions lAW the AOP and address the resultant Recirc Pump trip and also swap bus power supplies.

The SRO will address TS for single loop operation.

Then, the condensate pumps will trip causing a loss of Feedwater.

The crew will insert a manual scram and the SRO will enter IPO-5 due to the scram and EOP-1 due to low RPV level. After the initial scram actions are performed, a LOOP will occur with a small drywell leak. The output breaker on the "A" SBDG will fail to close and cannot be closed. The "B" SBDG will fail to start and must be started manually to power an Essential Bus (CRITICAL TASK 1.0). As the operators attempt to recover RPV level, RCIC will fail and the HPCI flow controller must be operated in manual to permit level recovery before an Emergency Depressurization is required (CRITICAL TASK 2.0). The scenario may be terminated when the "B" essential bus is recovered and RPV level is >170inches.