ML040160078

Examination Outline and NRC Comments for the Monticello Initial Examination - September 2003

ML040160078
Person / Time
Site:	Monticello
Issue date:	09/23/2003
From:	Allex G Nuclear Management Co
To:	NRC/RGN-III
References
50-263/03-301 50-263/03-301
Download: ML040160078 (41)

Text

EXAMINATION OUTLINE AND NRC COMMENTS FOR THE MONTICELLO INITIAL EXAMINATION - SEP 2003

Monticello Nuclear Generating Plant Operated by Nuclear Management Company, LLC May 14,2003 L-MT-03-036 10 CFR Part 55.40 Regional Administrator, Region Ill US Nuclear Regulatory Commission 801 Warrenville Road Lisle, Illinois 60532-4351 Attention: Hinori (Pete) Peterson MONTICELLO NUCLEAR GENERATING PLANT DOCKET 50-263 LICENSE NO. DPR-22 EXAMINATION OUTLINES FOR THE INITIAL LICENSING EXAMINATION TO BE CONDUCTED THE WEEK OF SEPTEMBER 15,2003.

Reference 1: NUREG 1021, Operator Licensing Examination Standards for Power Reactors, Revision 8, Supplement 1 In accordance with the requirements of 10 CFR 55.40(b)(4), a power reactor facility licensee must receive NRC approval of their proposed written examination and operating tests. Further, 10CFR55.40(a) requires that examinations meet the requirements of Reference 1. Therefore, enclosed for your review are the proposed examination outlines for the initial license examinations for our operator license applicants.

In accordance with 10CFR 55.49, Integrity of Examinations and Tests and Reference 1, Section ES-201, Attachment 1, Examination Security and Integrity Guidelines, the Nuclear Management Company, LLC requests that the enclosed materials be withheld from public disclosure until after the examinations are complete, and further that the enclosed materials only be viewed by the NRC examiner, Mr. Hinori (Pete) Peterson.

The proposed examination outlines were prepared per the guidelines of Reference 1, sections ES-301 and ES-401. Proposed outlines have been prepared to support development, by the Monticello facility, of examinations for seven Reactor Operator license candidate and four instant Senior Reactor Operator license candidates.

The proposed examination outlines prepared for the initial site-specific written examinations were randomly generated using the software application titled BWR WA Database Program, version 1.07.

Reference Ipermits licensees to screen the entire Knowledge and Abilities (WA) catalog to eliminate inapplicable WA statements. A listing of the WA statements that were suppressed from selection for outline generation is provided in the enclosed report entitled, Monticello Suppressed WA Report.

2807 West County Road 75 Monticello, Minnesota 55362-9637 Telephone: 763-295-5151 Fax: 763-295-1454

USNRC Nuclear ManagementCompany, LLC Paae 2 Enclosed are the following specific items for your review.

ES-201-2, Examination Outline Quality Checklist ES-301-1, Administrative Topics Outline (1 copy for the RO examination and 1 copy for the SRO examination)

ES-301-2, Control Room Systems and Facility WaIk-Through Test Outline ES-301-4, Simulator Scenario Quality Checklist ES-301-5, Transient and Event Checklist ES-D-1, Scenario Outline (1 for each scenario for 4 total)

ES-401-1, BWR SRO Examination Outline ES-401-2, BWR RO Examination Outline Monticello Suppressed WA Report If you have any questions regarding the enclosed information please contact Gerald M.

Allex, Operations Training Instructor, (763-271-2654 or 763-295-1563), or John Fields (763-295-1663).

David L. Wilson Site Vice President Monticello Nuclear Generating Plant cc: USNRC Document Control Desk (w/o attachments)

NRR Project Manager, NRC (w/o attachments)

Sr. Resident Inspector, NRC (w/o attachments)

Gerard Lashinski, General Supervisor Operations Training (w/o attachments) - ES-201-2, Examination Outline Quality Checklist - ES-301-1, Administrative Topics Outline ( I copy for the RO examination and Icopy for the SRO examination) - ES-301-2, Control Room Systems and Facility Walk-Through Test Outline - ES-301-4, Simulator Scenario Quality Checklist - ES-301-5, Transient and Event Checklist - ES-D-I, Scenario Outline (I for each scenario for 4 total) - ES-401-1, BWR SRO Examination Outline - ES-401-2, BWR RO Examination Outline - Monticello Suppressed WA Report

ES-201 Examination Outline FOITI ES-201-2 Quality Checklist j

i Item Task Description I Initials I

I ,

-96 H

F I d. Assess whether the justifications for deselected or rejected WA statements are appropriate.

I 1- I

2. a. Using Form ES-301-5, veriiy 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 test the projected number and M mix of applicants in accordance with the expected crew composition and rotation schedule without cornpromisingexamintegrjty; ensure 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 over successive days.

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

3. a. Venfythat:

(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,

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

b. Verifythat:

(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 condtior, (3) 40% of the tasks require the applicant to implement an alternate Dath orocedure.

(4) one implant task tests the applikant's response to an emergency'or abnormal condition, and (5) the in-plant walk-through requires the applicant to enter the RCA.

c. Venfy 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.

I 4.

G E

N E

R A

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. Author

. Facility Reviewer r)

. NRC Chief Examiner (#)

. NRC Supervisor ote: " Not applicable for NRCdevelooed examinations. ~ - - 1 I

1. Independent NRC reviewer initial items in Column 'c;" chief examiner concurrence required. Il 23 of 24 NUREG-1021, Revision 8, Supplement 1 dc <e4@d4Ll&LJ,J@4d45hFd@

MONTICELLO INITIAL EXAM OUTLINE REVIEW COMMENTS Review of Q/A Checklist ES-201-2 Item 1.a: The licensee was informed that the written exam WA outline did not use the appropriate Rev 8, Supplement I data form cover sheet. The cover sheet submitted was from an older version. The note on the cover sheet allowed a minimum of one topic from every WA category within each tier; however, the most recent revision requires a minimum of two topics. The overall WA distribution and point totals are correct. The licensee was informed to ensure updating of their documents.

Item I.b: Verified that the licensees WA program selects the WA to the specific WA number, i.e., K3.10, etc..

Item 1.c: Identified several duplicate use of WAS; however, it meets the minimum requirement of NUREG 1021, I... avoid selecting more than two or three WAS topics from a given system.... Relayed to the licensee to be extra careful in developing questions where there are double use of WAS (i.e., two WA topics from a given system) so as not to cover similar knowledge items.

Item 2: In general, the scenarios meet the minimum requirements. However, the overall selection of malfunctions, although without review of the actual material, appears to be minimal in difficulty. Several malfunctions appear to be less discriminating. The licensee was informed to ensure good operations validation and focus on the guidelines in Appendix D of NUREG 1021. Actual review of the exam quality will be performed during the on-site validation.

In addition, the scenarios did not indicate if they were bank, modified, or new.

The licensee verbally informed the NRC that all four scenarios were new.

Item 3.a(2): Based only on the titles of the Admin JPMs, it appeared that there were greater than 30% duplication from the last NRC exam. After verbal confirmation by the licensee, it appears that only one Admin JPM (A.l.a)is duplicated exactly from the last NRC exam. Others which indicated duplication only based on the title description, both RO & SRO JPMs A3 and A4, will be verified during the on-site validation.

Item 3.a(3): Verified verbally with licensee that no duplication from the audit exam. The licensee who is developing the NRC exam was the same individual who developed the outline for the audit exam. The individual assured that there are no duplication; however, the licensee will again verify once the exam materials, both audit and NRC, are developed that there are no duplications. The licensee Page 1 of 2

was also requested by the Chief Examiner to submit the audit outline with the NRC exam material for double assurance.

Item 3.a(4): The licensee did not indicate on the Admin JPMs which were bank, modified, or new. The licensee verbally indicated that only one JPM was from the bank and one was duplicate from the last NRC exam.

Item 3.b(2): From the tile of the systems JPM, B.1.b and B.l.g, did not appear to be a low power or shutdown JPM. Licensee was questioned, and verbally the licensee assured that these JPMs must be conducted at low power. Actual material verification will be conducted during the on-site validation.

Item 3.b(3): The licensee originally submitted systems JPMs with 5 alternate path JPMs.

The licensee was instructed to change one of the five to an normal, non-alternate path JPM. The licensee'indicated that JPM B.1.d will be changed to an non-alternate path, Item 3.c: The licensee was informed that the RO JPM A4 (make a PA announcement of an emergency) was not discriminating. The licensee indicated they will review and attempt to identify a better discriminating JPM. Also, the licensee was informed that the NRC will also attempt to identify and suggest an appropriate replacement.

Page 2 of 2

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

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Facility: MNGP Date of Examination: 9/15/03 Exam Level (circle one) @/ SRO Operating Test Number: 03-01 Administrative Describe method of evaluation TopidSubject 1. ONE Administrative JPM, OR Descriotion 2. TWO Administrative Questions A. I Plant JPM Perform Core Thermal Limits Monitoring Parameter Procedure C.2-05 Verification Generic 2.1.19 (3.0)

Conduct JPM Independent Verification Valve Procedure 4AWI-04.04.02; Procedure 0255-06-IA-I Lineup Generic 2.1 2 9 (3.4)

A. 2 Surveillance JPM Perform the Daily Off-Gas Hydrogen Analyzer Checks Testing Procedure 0000-H; Test 0209 Generic 2.2.12 (3.0)

A. 3 Radiation JPM Expected Dose Determination to Inspect Equipment Control Procedure 4AWI-08.04.01 Calculating Generic 2.3.1 (2.6)

Exposure A.4 Emergency JPM Public Address Announcement of a General Emergency Communications Declaration Procedure A.2-105 Generic 2.4.43 (2.8)

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

=acility: MNGP Date of Examination: 9/15/03 ixam Level (circle one) RO I @ Operating Test Number: 03-01 Administrative Describe method of evaluation TopidSubject 1. ONE Administrative JPM, OR

- Description 2. TWO Administrative Questions 4.1 Plant JPM Perform Core Thermal Limits Monitoring Parameter Procedure C.2-05 Verification Generic 2.1 .I9 (3.0), 2.1.33 (4.0)

Crew JPM Crew Staffing Determination Composition Procedure OWI-01.06; Procedure 4 AWI-08.01.01 Generic 2.1.4 (3.4) 4.2 Surveillance JPM Perform the Daily Off-Gas Hydrogen Analyzer Checks Testing Procedure 0000-H; Test 0209; ODCM-03.01 Generic 2.2.12 (3.4) 9.3 Radiation JPM Expected Dose Determination to Inspect Equipment Control Procedure 4AWI-08.04.01 Calculating Generic 2.3.1 (3.0)

Exposure 4.4 Emergency Plan JPM Classify an Event and Determine Protective Action Protective Recommendations Action Procedure A.2-101; Procedure A.2-204 Recommendation Generic 2.4.38 (4.0), 2.4.44 (4.0)

ES-301 Control Room Systems and Facility Walk-Through Test Outline Form ES-301-2 Date of Facility: MNGP Examination: 9115103 Exam Level (circle one) @) I Operating Test No.: 03-01 I B.I Control Systems System 1 JPM Title Function

a. Inadvertent Control Rod Insertion J PM-B.05.05-006 201003 A3.01 (3.713.6) D,A1 I 1

b. Reject Water From Rx Vessel Using RWCU to Radwaste JPM-B.02.02-005 I 2 204000 A I .07 (2.912.9)

I d.

JPM-B.02.04-004 239001 A4.01 (4.214.0)

Core Spray Isolation After Pump Trip D,A,S I 3 JPM-B.03.01-006 4 209001 A2.01 (3.413.4)

e. Reset a Group II Isolation JPM-C.4-B.04.01 .B - Part A 223002 A4.03 (3.6135)

f. Restore Bus 15 from Bus 13 JPM-E.2-05-001 D, s 6 262001 A2.07 (3.013.2)

g. IRM Functional JPM-B.05.01.01-002 215003 A4.07 (3.613.6)

a. Depressurize the Scram Air Header JPM-(2.5-3101-007 D, A, R 1 295037 EA1.03 (4.114.1) Emergency

b. Startup the Reactor Protection Motor Generators JPM-B.09.12-00 1 D, R 7 212000 K1.04 (3.413.6); 2.1.30 (3.9/3.4),

c. Manual Initiation of EFT in the High Radiation Mode JPM-B.08.13-05-002 N, A, R 9 288000 A2.04 (3.713.8)

ES-301 Simulator Scenario Quality Checklist Form ES-301-4

) is incorporated into the scenario without tion team to obtain NUREG-1021, Revision 8, Supplement 1 24 of 26

ES-301 Simulator Scenario Quality Checklist Form ES-301-4 1 Facility: MONTICELLO Date of Exam: 9115103 Scenario Numbers: 4 I I OperatingTest No. 03-01 QUALITATWE AllRIBUTES Initial a b' dc

• %/*>

3. Eache

7. If time compression techniques are used, the scenario summary clearly so indicates, Operators have sufficient time to carry out expected activities without undue time constraints. Cues are I/lk/ I!,, A 1 TARGET QUANTITATIVEATTRIBUTES (PER SCENARIO; SEE SECTION D.4.D) I ActualAttributes I - I - I -

NUREG-1021, Revision 8, Supplement I 24 of 26

ES-301 Transient and Event Checklist Form ES-301-5 OPERATING TEST NO.: 03-01 Applicant Evolution Minimum Scenario Number TYPe Type Number 1 2 3 4 Reactivity 1 1 2 2 5 Normal 1 3 1 1 1 RO Instrument / 4 2, 4, 314, 3, 4, 213, Component 5, 61 5,6, 5961 416, 9 9 8 8 Maior 1 7,8 7,8 7,9 7 Reactivity 1 11 12 2 5 I I Normal 0

- 1 1 1 NIH "/A N, N/A N/A As RO Instrument / 2 2, 5 3, I SRO-I I 1 2 2 5 Reactivity 0 Normal 1 3 1 1 1 As SRO Instrument / 2 294, 314, 3,4, 2,3, Component 5161 59 61 5, 61 4,6i 9 9 8 8 Major 1 7,8 7,8 7,9 7 Reactivity 0 N/A NIA N/A N/A Normal 1 NIA N/A N/A NIA SRO-U InstrumentI 2 NIA NIA N/A N/A Component Major 1 N/A N/A N/A N/A Instructions: (1) Enter the operating test number and Form ES-D-1 event numbers for each evolutiontype (2) Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.4.d) but must be significant per Section C.2.a of Appendix D.

(3) 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 requirement.

Author: I n -

/

NRC Reviewer:

NUREG-I021, Revision 8, Supplement 1

Amendix D Scenario Outline Form ES-D-1 I

II Facility: MNGP Scenario No. NRC Op-Test No: 03-01 03-01 Examiners: Operators:

Initial Conditions: A reactor startup is in progress from a refueling outage. The crew will assume the shift with the plant operating at 13% power making preparations for rolling the Main Turbine. Operations Manual C. 1, Reactor Startup, has been completed up through step V1.A. 18.

Turnover: Withdraw control rods to achieve 1-1/2 Turbine Bypass Valves open in accordance with step C.l Step VI.B.l and then roll the Main Turbine per C.l Step VI1.B and continue with plant startup.

Event Malf. Event Event No. No. DescriDtion 1 N/A RO (R)- Withdraw control rods to establish 1-1/2 Bypass Valves 2 RWOI RO (I) RWM Equipment Failure (Tech Spec Call) 3 N/A BOP (N) Roll the Main Turbine 4 TCOSB BOP (C) Pressure Regulator Oscillations, MPR 5 SWOIA RO (C) # I 1 RBCCW PumpTrip 6 IA04 BOP (C) Instrument Air Header Failure 7 Note 1 M Failure to Scram (West SDV hydraulic lock) 8 RM03K M Rupture of West SDV I IRM03H I I RXOI 9 SO2501 BOP(C) Failure of ASRV to Open EcluiDment Out of Service

1. I 3 Service Water Pump for coupling repair.

1. I2 IRM is bypassed for erratic indication.

1. I 2 SBLC PumD is out for oil reDlacement.

Note 1: This event requires 61 malfunctions to be inserted to have 61 control rods associated with the West SDV stuck.

I I I I

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

Appendix D Scenario Outline Form ES-D-1 Facility: MNGP Scenario No. NRC Op-Test No: 03-01 03-02 Examiners: Operators:

Initial Conditions: Reactor power was reduced to 50% last shift to perform some on-line maintenance on the No. 11 RFP. The maintenance is complete and the RFP is ready to start.

Turnover: Start the No. 11 RFP and return to 100% power.

Event I Malf. 1 Event 1 Event to Auto Start 6 A16507 RO (I) Failure of RPV Level Control Setpoint 7 ED12 M Loss of Off-Site Power DG02B Failure of No. 12 EDG to Start 8 RROIB M HPCl Steam Line Break on Start (inside drywell)

(simulated via small Recirc line break) 9 RC03 BOP (C) RClC Turbine Trip EquiDment Out of Service

1. I 3 Service Water Pump for coupling repair.

1. I 2 IRM is bwassed for erratic indication.

I #I 2 SBLC p i m p is out for oil replacement.

I

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

Amendix D Scenario Outline Form ES-D-1 Facility: MNGP Scenario No. NRC Op-Test No: 03-01 03-03 Examiners: Operators:

Initial Conditions: The plant is operating at approximately 14% power. The Main Turbine has been placed on-line and the next step is to place the Main Turbine Generator in service. Ops Man C.l is complete through step VI.C.4.

Turnover: Place the Main Turbine Generator in service and continue power increase.

Event I Malf. I Event I Event I I Equbment Out of Service

1. I 3 Service Water Pump for coupling repair.

1. I 2 IRM is bypassed for erratic indication.

1. I 2 SBLC Pump is out for oil replacement.

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

Appendix D Scenario Outline Form ES-D-I Facility: MNGP Scenario No. NRC Op-Test No: 03-01 03-04 Examiners: Operators:

Initial Conditions: The crew will take the duty with the plant at 100% reactor power. HPCl has been inoperable for the past 5 days for repairs. The off-going crew is in the progress of performing procedure 0255-06-IA-I to complete the PMT requirements to restore HPCl to operable status. The next step in the procedure is to secure Torus cooling. The off-going crew has secured ARHR from Torus Cooling.

Turnover: Secure B RHR from the Torus Cooling Mode per step 81 of 0255-06-IA-1.

Event I Malf. I Event Event Equipment Out of Service

1. I 3 Service Water Pump for coupling repair.

1. 2IIRM is bwassed for erratic indication.

I HPCl is inopyor performance of 0255-06-IA-1.

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

BWR SRO Examination Outline Printed 0511312003 Facility: Monticello Nuclear Generating ES 401 Emergenc and Abnorm 11Plant Evolutions - Tier 1 / Group 1 Form ES-401-1 I I I E/APE # I E/APE Name / Safety Function K1

- K2K3 - A1 I

295007 I High Reactor Pressure 1 3 I X

X

-+

I 295007 High Reactor Pressure 1 3 295009 Low Reactor Water Level 1 2 X 1295010 I High Drywell Pressure I 5 I X 2950 13 High Suppression Pool Temperature / 5 X 3.a I 1 295013 High Suppression Pool Temperature/ 5 4.0 1 2950 15 Incomplete SCRAM I 1 X 4.2* 1 295016 Control Room Abandonment I 7 X 4.1* 1 295016 Control Room Abandonment 1 7 X 3.7* 1 295017 I High Off-Site Release Rate / 9 I

I X 4.6* 1 2950 17 High Off-Site Release Rate I 9 X 3.7 1 1295024 I High Drywell Pressure I 5 Ix EK1 .O 1 - Drywell integrity: Plant-Specific 4.2* 1 1 1

BWR SRO Examination Outline Printed 05/13/2003 Facility: Monticello Nuclear Generating ES 401 -

Emereencv and Abnormal Plant Evolutions Tier 1 / GrouD 1 Form ES-401-1 E/APE ## I E/APE Name / Safety Function I K1 IK21K3 I A1 A2 G KA Topic I I I I I 295025 295025

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295026 295026

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High Reactor Pressure / 3 High Reactor Pressure / 3 Suppression Pool High Water Temperature 1 5 Suppression Pool High Water Temperature / 5 M X X

X characteristics,reactor behavior, and instrument IEK1.05 - ?Exceeding safety limits EA2.02 - Suppression pool level EK3.04 - TSBLC injection

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4.4 4.1*

I I

295030 I Low Sumression Pool Water Level / 5 I l l X I EA1.05 - HPCI 3.5 1 295030 I Low Suppression Pool Water Level / 5 I 1x1 I EK2.08 - SRV discharge submergence 3.8 1 295031 I Reactor Low Water Level / 2 I I Ix I -

EK3.01 Automatic demessurization system actuation 4.2* 1 295031 Reactor Low Water Level / 2 X EA 1.10 - Control rod drive 3.7 1 295037 SCRAM Condition Present and Reactor Power Above X EA2.02 - Reactor water level 4.2* 1 APRM Downscale or Unknown / 1 295037 SCRAM Condition Present and Reactor Power Above X -

EA2.06 Reactor pressure 4.1 1 APRM Downscale or Unknown / 1 295038 High Off-Site Release Rate 1 9 X EA1.07 - Control room ventilation: Plant-Specific 500000 High Containment Hydrogen Concentration / 5 X IEKl .O 1 - Containment integrity 3.9 I 1 WACategory Totals: 3 6 5 5 5 2 Group Point Total: 26 2

BWR SRO Examination Outline Printed 05/13/2003 Facility: Monticello Nuclear Generating ES - 401 ~

Eme -

ind Abnormal Plant Evolutions Tier 1 / Group 2 Form 5-401-:

E/APE # E/APE Name / Safety Function KA Tonic [mp. Points 295001 Partial or Complete Loss of Forced Core Flow AK2.04 - Reactor/turbinepressure regulating system: 3.3 1 Circulation / 1 Plant-Specific 295002 Loss of Main Condenser Vacuum I 3 AA2.02 - Reactor power: Plant-Specific 3.3 1 295004 Partial or Complete Loss of D.C. Power / 6 AA1.02 - Systems necessary to assure safe plant 4.1 1 shutdown 295005 Main Turbine Generator Trip / 3 AA1.02 - RPS 3.6 295008 High Reactor Water Level / 2 AK2.11 - Main steam 3.3 295012 High Drywell Temperature / 5 -

2.4.21 Knowledge of the parameters and logic used to 4.3 assess the status of safety functions including:

1.Reactivity control 2.Core cooling and heat removal 3.Reactor coolant system integrity 4.Containment conditions 5.Radioactivity release control.

295012 High Drywell Temperature / 5 -

AK1.O 1 Pressure/temperature relationship 3.5 295018 Partial or Complete Loss of Component Cooling AK3.05 - Placing standby heat exchanger in service 3.3 1 Water / 8 2950 19 Partial or Complete Loss of Instrument Air / 8 -

AA1.03 Instrument air compressor power supplies 3.O 295028 High Drywell Temperature / 5 2.4.16 - Knowledge of EOP implementation hierarchy 4.0 and coordination with other sumort mocedures.

295028 High Drywell Temperature / 5 -

EAl.05 ADS 3.7 1 1

BWR SRO Examination Outline Printed: 0511312003 Facility: Monticello Nuclear Generating ES 401 -

Emereencv and Abnormal Plant Evolutions Tier 1 / Groun 2 Form ES-401-E/APE # E/APE Name / Safety Function K1 K2 K3 A1 A2 G KATopic Imp. Points 295029 High Suppression Pool Water Level I 5 X EA2.03 - DwelVcontainment water level 3.5 1 295029 High Suppression Pool Water Level 1 5 X EK3.03 - Reactor SCRAM 3.5 1 295033 High Secondary Containment Area Radiation Levels / X EK1.02 - Personnel protection 4.2* 1 9

295034 Secondary Containment Ventilation High Radiation I X EK3.02 - Starting SBGTIFRVS: Plant-Specific 4.1 1 9

295035 Secondary Containment High Differential Pressure I 5 X EA1.02 - SBGTIFRVS 3.8 1 295036 Secondary Containment High SumpIArea Water Level I 5 X

I II I I I EK3.O 1 - Emergency depressurization I I I

2.8 I

1 5 2 2 Group Point Total: 17 2

BWR SRO Examination Outline Printed 05/13/2003 t

Facility: Monticello Nuclear Generating ES 401 -

Plant ystems Tier 2 I Group 1 Form ES-401-I S s/Ev# System I Evolution Name K2 K1 - K4 K5 K6 A1 42 A3 G KATopic Imp. Points 203000 RHRILPCI: Injection Mode (Plant X -

K2.03 Initiation logic 2.9* 1 Specific) 1 2 203000 RHR/LPCI: Injection Mode (Plant Saecific) 1 2 X A2.17 - Keep fill system failure I 13.5 1

/209001 Low Pressure Core Spray System 1 2 I I X K4.05 - Pump minimum flow 2.6 1 (209001 Low Pressure Core Spray System 1 2 X A2.05 - Core spray line break 3.6 1 212000 Reactor Protection System 1 7 X A3.04 - System status lights and alarms 3.8 1 215004 Source Range Monitor (SRM) System 17 -

IK3*01 RPs 13.41 1 2 15005 Average Power Range MonitorLocal K1.04 - LPRM channels 3.6 1 Power Range Monitor System / 7 216000 Nuclear Boiler InstrumentationI 7 -

K3.14 High pressure coolant injection:

Plant-Specific I I4.2* 1 2 16000 Nuclear Boiler Instrumentation / 7 X A1.O1 - Recorders and meters 3.3 1 223001 Primary Containment System and X K1.03 - Containment/drywellatmosphere 3.3 1 Auxiliaries 1 5 control 223001 Primary Containment System and X A2.03 - Safetylrelief valve leaking or stuck 4.2* 1 Auxiliaries I 5 open 223002 Primary Containment Isolation X A3.03 - SPDSIERISICRIDSIGDS: 2.8* 1 SystemINuclear Steam Supply Plant-Specific Shut-Off I 5

- I I I 1

BWR SRO Examination Outline Printed 05/13/2003 Facility: Monticello Nuclear Generating ES 401 Plant Svstems - Tier 2 / Grouo 1 Form ES-401-1 sysm# System / Evolution Name K1 K2 K3 223002 Primary Containment Isolation System/Nuclear Steam Supply Shut-Off / 5 22600 1 RHR/LPCI: Containment Spray System Mode / 5 239002 I RelieflSafety Valves / 3 I l l 24 1000 ReactorITurbine Pressure Regulating System 1 3 241000 Reactor/Turbine Pressure Regulating System / 3 259002 Reactor Water Level Control System /

2 26 1000 Standby Gas Treatment System / 9 262001 A.C. Electrical Distribution 1 6 X 264000 Emergency Generators (DieseVJet) / 6 290001 I SecondaryContainment / 5

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I l l X I I I I I I I IK4.03 -Fluid leakage collection I 2.9 I 1 290001 I

Secondary Containment / 5 Ill I I I X A3.02 - Normal building differential pressure:

Plant-Specific 3.5 1 2

3 "

B Y

R rn

BWR SRO Examination Outline Printed: 05/13/2003 Facility: Monticello Nuclear Generating ES - 401 Plant Systems - Tier 2 / Group 2 Form ES-401-1 K1 K2 K3 K4 K5 K6 41 A2 43 A4 KA Topic Imp. Points Al.03 - Rod movement sequence lights X 2.9

- 1 205000 Shutdown Cooling System (RHR X K5.02 - Valve operation 2.9 1 Shutdown Cooling Mode) 1 4 205000 Shutdown Cooling System (RHR X Al.05 - Reactor water level 3.4 1 Shutdown Cooling Mode) 1 4

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2 15002 Rod Block Monitor System / 7 X A2.01- Withdrawal of control rod in high 1 215002 Rod Block Monitor System 1 7 X K2.03 - APRM channels: BWR-3,4,5 2.9 219000 RHR/LPCI: Torus/Suppression Pool X A3.01 - Valve operation 3.3 Cooling Mode I 5 230000 RHR/LPCI: TorusISuppression Pool X A4.09 - Indicating lights and alarms 3.3 Smav Mode I 5 230000 RHlULPCI: TorusISuppression Pool X K1.05 - A.C. electrical 3.3 1 Spray Mode I 5 234000 Fuel Handling Equipment I 8 X K6.04 - ?Refueling platform air system: 3.7 Plant-Specific 234000 Fuel Handling Equipment I 8 2.1.33 - Ability to recognize indications for 4.0 system operating parameters which are entry-level conditions for technical specifications.

27 1000 Offgas System / 9 X -

K3.02 ?Off-site radioactive release rate I 3.9 1 1

B W R SRO Examination Outline Printed: 05/13/2003 Facility: Monticello Nuclear Generating ES - 401 -

Plant Systems Tier 2 / Group 2 Form ES-401-1 Sys/Ev # System / Evolution Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G KATopic Imp. Points 27 1000 Offgas System / 9 X K4.05 - Redundancy 2.6 1 400000 Component Cooling Water System X 2.2.2 - Ability to manipulate the console 3.5 1 (CCWS) / 8 controls as required to operate the facility

~~ ~~~

between shutdown and designated power levels. ~ ~~ -

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

BWR SRO Examination Outline Printed: 05/13/2003 Facility: Monticello Nuclear Generating ES - 401 -

Plant Systems Tier 2 / Group 3 Form ES-401-1 KA Topic Imp. Points A1.03 - Valve status: Mark-I&II(Not-BWR1) 2.8 1 A4.01 - Sump integrators 3.6 1 K4.04- Moisture removal fiom generated steam 2.8 1 K3.01 - Reactor water level I 13.3 1

~~

WACategory Totals: 0 0 1 1 0 0 1 0 0 1 0 Group Point Total: 4 1

Generic Knowledge and Abilities Outline (Tier 3)

Printed 05/13/2003 BWR SRO Examination Outline Form ES-401-5 Facilitv: Monticello Nuclear Generating Generic Category KA KATopic Imp. Points 2.1.7 Ability to evaluate plant performance and make operationaljudgments based on operating 4.4

- 1 Conduct of Operations characteristics,reactor behavior, and instrument interpretation.

2.1.11 Knowledge of less than one hour technical specification action statements for systems. 3.8 1 2.1.32 Ability to explain and apply system limits and precautions. 3.8 1 2.1.18 Ability to make accurate, clear and concise logs, records, status boards, and reports. 3.0 1 2.1.22 Ability to determine Mode of Operation. 3.3 1.

Category Total: 5 Equipment Control 2.2.8 IKnowledge of the process for determining if the proposed change, test, or experiment 3.3

- 1 involves an unreviewed safety question.

2.2.13 Knowledge of tagging and clearance procedures. 3.8 1 2.2.26 Knowledge of refueling administrativerequirements. 3.7 1 2.2.24 Ability to analyze the affect of maintenance activities on LCO status. 3.8 1 2.2.32 Knowledge of the effects of alterations on core configuration. 3.3 1 Category Total: 5 Radiation Control for auxiliary systems thatare outside the control room 2.9 handling systems).

radiation releases.

ALARA program. 2.9 Category Total: 3 1

Generic Knowledge and Abilities Outline (Tier 3)

Printed: 05/13/2003 BWR SRO Examination Outline Form ES-401-5 Facility: Monticello Nuclear Generating Generic Category KA KATopic Imp. Points Emergency Plan 2.4.24 Knowledge of loss of cooling water procedures. 3.7 1 2.4.5 Knowledge of the organization of the operating procedures network for normal, abnormal, 3.6 1 and emergency evolutions.

2.4.6 Knowledge symptom based EOP mitigation strategies. 4.0 1 2.4.30 Knowledge of which events related to system operations/statusshould be reported to 3.6 1 outside agencies.

2

ES-401 BWR SRO Examination Outline Printed: 05/13/2003 Facility: Monticello Nuclear Generating Plant Form ES-401-1 Exam Date: 09/15/2003 Exam Level: SRO l l WA Category Points Point Total 26 17 43 23 13 Note:

1. Attempt to distribute topics among all WA Categories; select at least one topic from every WA 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 three WA topics from agiven 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 tier.

1

BWR RO Examination Outline Printed 0511312003 Facility: Monticello Nuclear Generating ES 401 -

Emergency and Abnormal Plant Evolutions Tier 1 / Grouu 1 Form ES-401-2 E/APE # I G KA Topic Imp. -

?oints I

295006 SCRAM/ 1 -

AK3.06 Recirculation pump speed reduction: 3.2 1 Plant-Suecific 295006 SCRAM1 1 AA2.06 - Cause of reactor SCRAM 3.5 1 295007 High Reactor Pressure 1 3 I Ix -

AK2.03 RHR/LPCI: Plant-SDecific 3.1 1 I

295007 High Reactor Pressure / 3 AA2.02 - Reactor power 4.1* 1 295009 Low Reactor Water Level 1 2 X AK2.03 - Recirculation system 3.1 2950 10 High Drywell Pressure 1 5 I Ix -

AK2.02 DrywelVsuppression chamber differential pressure: Mark-I&II 3.3 295010 High Drywell Pressure / 5 AA1.05 - Drywelllsuppression vent and purge 3.1 1 295024 High Drywell Pressure 1 5 IXI -

EKl.01 Drywell integrity: Plant-Specific 4.1 1 295025 High Reactor Pressure 1 3 -

X 2.1.7 Ability to evaluate plant performance and make 3.7 1 operationaljudgments based on operating characteristics, reactor behavior, and instrument interpretation.

29503 1 Reactor Low Water Level 1 2 EK3.O1 - Automatic depressurizationsystem actuation 3.9 1 29503 1 Reactor Low Water Level 12 EA 1.10 - Control rod drive 3.6 1 295037 SCRAM Condition Present and Reactor Power Above EA2.06 - Reactor pressure 4.0 1 APRM Downscale or Unknown / 1 500000 -

EK1.O1 Containment integrity -3.3 - 1 WACategoryTotals: 2 3 2 2 3 1 Group Point Total: 13 1

B W R RO Examination Outline Printed 0511312003 Facility: Monticello Nuclear Generating ES 401 - - -

I Plant Evolutions Tier 1 / Group 2 Form ES-401-2 I I I I EIAPE # E/APE Name I Safety Function K1 K2 KA Topic IImp. I Points I I 29500 1 Partial or Complete Loss of Forced Core Flow Circulation I 1 X AK2.04 Reactorlturbine pressure regulating system:

Plant-Specific 3.3 I' 295002 Loss of Main Condenser Vacuum 13 X I AA2.02 - Reactor power: Plant-Specific I 3.2 I 1 295008 High Reactor Water Level 1 2

____ ~

X ~

AK2.11 Main steam 3.1 1 295012 High Drywell Temperature I 5 X -

AK1.01 Pressureltemperaturerelationship 3.3 1 295013 High Suppression Pool Temperature / 5 X AK3.01 - Suppression pool cooling operation 3.6 295013 High Suppression Pool Temperature 1 5 AA2.0 1 - Suppression pool temperature 3.8 295016 Control Room Abandonment / 7 X AK2.02 - Local control stations: Plant-Specific 4.0* 1 2950 16 Control Room Abandonment I 7 X AK3.03 - Disabling control room controls 3.5 1 295017 High Off-Site Release Rate I 9 X AK2.12 - Standby gas treatmentRRVS 3.4 1 2950 18 Partial or Complete Loss of Component Cooling X AK3.05 - Placing standby heat exchanger in service 3.2 1 Water I 8 2950 18 Partial or Complete Loss of Component Cooling -

2.1.20 Ability to execute procedure steps. 4.3 1 Water I 8 295019 Partial or Complete Loss of Instrument Air / 8 X -

AA1.03 Instrument air compressor power supplies 1 295026 Suppression Pool High Water Temperature/ 5 X -

EK3.04 tSBLC injection 1 295028 High Drywell Temperature/ 5 X I (EAl.05-ADS 1 1

BWR RO Examination Outline Printed 05/13/2003 Facility: Monticello Nuclear Generating ES 401 Emer -

ormal Plant Svolutions Tier 1 / Grom 2 Form ES-401-2 EIAPE # EIAPE Name I Safety Function K1 K2 K3 A1 A2 G KA Topic 295030 Low Suppression Pool Water Level I 5 X EA1.05 - HPCI 3.5 295030 Low Suppression Pool Water Level / 5 X -

EK2.08 SRV discharge submergence 3.5 I) 295033 High Secondary Containment Area Radiation Levels I X -

EK1.02 Personnel protection 9

295034 Secondary Containment Ventilation High Radiation I X EK3.02 - Starting SBGTRRVS: Plant-Specific 9

295038 High Off-Site Release Rate 19 X EA1.07 - Control room ventilation: Plant-SDecific 3.6 11 WACategory Totals: 2 5 5 4 2 1 Group Point Total: 19 2

BWR RO Examination Outline Printed: 05/13/2003 Facility: Monticello Nuclear Generating r

ES 401 Emereencv -

- " and Abnormal Plant Evolutions Tier 1 / Groua 3 Form ES-401-2 I I I I E/APE # 1 E/APE Name / Safety Function IK11K21K3 A1 IA2 IGIKATopic IImp. Points I I I I I 1 1 I 295032 High Secondary Containment Area Temperature I 5 X EK2.05 - Temperature sensitive instrumentation 3.2 1 295035 Secondary Containment High Differential Pressure 1 5 X EAl.02 - SBGTIFRVS 3.8 1 295036 Secondary Containment High SumpIArea Water Level 1 5 1 II I

X I 1 EA2.03 - Cause of the high water level 1 I

3.4 I II I I I I I 295036 I Secondary Containment High Sump/Area Water I I IX -

EK3.01 Emergency depressurization 2.6 I Level 1 5 I I I WACategoryTotals: 0 1 1 1 1 0 Group Point Total: 4 1

BWR RO Examination Outline Printed: 05/13/2003 Facility: Monticello Nuclear Generating ES - 401 -

I Plant Svstems Tier 2 / Grow 1 Form ES-401-2

• ~

<3 K4 US K6 41 !A2 43 A4 KA Topic ip. Points X -

A3.03 System pressure 2.7 I 1 1201002 I Reactor Manual Control System / 1 X AI .03 - Rod movement sequence lights 1202002 1 Recirculation Flow Control System / 1 X K4.06 - Recirculation pump adequate NPSH:

Plant-Specific 203000 RHR/LPCI: Injection Mode (Plant K2.03 - Initiation logic Specific) 12 I

203000 RHIULPCI: Injection Mode (Plant Specific) 12 X A2.17 - Keep fill system failure 3*3 I Y-206000 High Pressure Coolant Injection X A4.10 - System pumps: BWR-2,3,4 System / 2 209001 Low Pressure Core Spray System 1 2 X K4.05 - Pump minimum flow 2.6 I 1 2 12000 Reactor Protection System / 7 X A3.04 - System status lights and alarms 3.9* I 1 2 15003 Intermediate Range Monitor (IRM)

System I 7 X A4.05 - Trip bypasses 3.4 I 1 215004 1215004 I Source Range Monitor (SRM) System I7 Source Range Monitor (SRM) System 17 Power Range Monitor System 17 X

I X

K3.01 - RPS A4.04 - SRM drive control switches K1.04 - LPRM channels t 1

BWR RO Examination Outline Printed: 05/13/2003 Facility: Monticello Nuclear Generating ES 401 sys/Ev #

2 15005 216000 System / Evolution Name Average Power Range Monitor/Local Nuclear Boiler Instrumentation / 7 I

K1 K2 K3 K4 X

43 A4 m1 Plant Svstems Tier 2 / Group 1 A1.05 Lights and alarms K3.14 - High pressure coolant injection:

Plant-Specific Form ES-401-2 Imp.

3.3 3.8 j

I Points 1

1 216000 Nuclear Boiler Instrumentation / 7 -

A 1.O 1 Recorders and meters 3.4 1 2 17000 Reactor Core Isolation Cooling X K4.04 - Prevents turbine damage: Plant-Specific 3.O 1 System (RCIC) / 2 1

223001 Primary Containment System and IK1.03 - Containment/drywell atmosphere control 3.2 1 223001 Primary Containment System and A2.03 - Safetyhelief valve leaking or stuck 4.0 1 Auxiliaries / 5 lopen 223002 Primary Containment Isolation X -

A3.03 SPDS/ERIS/CRIDS/GDS: 2.5* 1 SystemJNuclearSteam Supply Plant-Specific Shut-Off / 5 223002 Primary Containment Isolation X K4.05 - Single failures will not impair the 2.9 1 System/Nuclear Steam Supply function ability of the system Shut-Off / 5 239002 Relieusafety Valves 1 3 -

K6.03 A.C. power: Plant-Specific 2.7* 1 239002 Relieusafety Valves / 3 I X IA4.06 - Reactor water level 3.9 1

1 241000 ReactodTurbine Pressure Regulating Al.07 - Bypass valve position System I 3 2

BWR RO Examination Outline Printed: 05l1312003 Facility: Monticello Nuclear Generating ES - 401 -

Plant ystems Tier 2 / Group 1 Form ES-401-2 I 7 I l l I

Sys/Ev # System / Evolution Name IK1 IK2 IK3 K4 K5 K6 A1 42 A3 A4 2 KA Topic Imp. Points I I I I X

IK5.04- Turbine inlet pressure vs. reactor Dressure I I3-3 K2.0 1 Reactor feedwater pump(s):

259001 Reactor Feedwater System I 2 X A2.05 - Loss of applicable plant air systems 259002 Reactor Water Level Control System I 2

26 1000 Standby Gas Treatment System I 9 X WACategory Totals: 2 2 2 4 1 2 4 3 3 4 1 Group Poi It Total: 28 3

I BWR RO Examination Outline Printed: 0511312003 Facility: Monticello Nuclear Generating ES - 401 -

Plant Systems Tier 2 / Group 2 Form ES-401-2 sysmv # System /Evolution Name K1 I 43 A4 G KA Topic Imp. Points I

20 1006 Rod Worth Minimizer System (RWM)

(Plant SDecific) I 7 I Al.03 - Latched group indication:

2.9 1 205000 Shutdown Cooling System (RHR -

K5.02 Valve operation 2.8 1 Shutdown Cooling Mode) 1 4 I 1 I 1 205000 Shutdown Cooling System (RHR Shutdown Cooling Mode) 1 4 I I

I Al.05 - Reactor water level 1

1 3.4 1 2 15002 Rod Block Monitor System I 7 A2.01- Withdrawal of control rod in high 3.3 1 2 15002 2.8 1 219000 RHRLPCI: TorusISuppression Pool X -

A3.0 1 Valve operation 3.3 1 cool in^ Mode I 5 2 19000 RHRILPCI: TorusISuppression Pool Al.02 - System flow 3.5 1 Cooling Mode I 5 22600 1 RHR/LPCI: Containment Spray X A4.13 - Containment/drywelltemperature 3.3 1 System Mode I 5 230000 RHR/LPCI: TorusISuppressionPool X K1.05 - A.C. electrical 3.2 1 Spray Mode I 5 245000 Main Turbine Generator and Auxiliary X A4.10 - Hydrogen gas pressure 2.6 1 262001 K2.01 - Off-site sources of power 3.3 1 1

BWR RO Examination Outline Printed: 05/13/2003 Facility: Monticello Nuclear Generating ES 401

~~

I I I I I I I I I I I I i i Sys/Ev # System / Evolution Name 262001 A.C. Electrical Distribution / 6 lK1 I= IK3 lK4 A3 A4 G KA Topic X 2.4.10 Knowledge of annunciator response procedures.

Imp.

3.O Points 1

262002 Unintermptable Power Supply A2.0 1 - Under voltage 2.6 1 (A.C./D.C.) / 6 K5.O1 - Hydrogen generation during battery 263000 I D.C. Electrical Distribution / 6 I I I I charging 2.6 1 271000 Offgas System / 9 X K3.02 - ?Off-site radioactive release rate 3.3 1 271000 Offgas System / 9 X K4.05 - Redundancy 2.6 290001 Secondary Containment I 5 X K4.03 - Fluid leakage collection 2.8 29000 1 Secondary Containment I 5 X A3.02 - Noma1 building differential pressure: 3.5 1 I

I 1 I I I Plant-Specific Cooling Water System X 2.2.2 - Ability to manipulate the console 4.0 1 controls as required to operate the facility I I I I I between shutdown and designated power levels. -

KlACategory Totals: 1 2 1 2 2 0 3 2 2 2 2 Group Point Total: 19 2

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Generic Knowledge and Abilities Outline (Tier 3)

Printed 05/13/2003 BWR RO Examination Outline Form ES-401-5 Facilitv: Monticello Nuclear Generating Generic Category KA KA Topic Imp. Points Conduct of Operations 2.1.1 1 Knowledge of less than one hour technical specification action statements for systems. 3 .O

-1 2.1.32 Ability to explain and apply system limits and precautions. 3.4 1 2.1.18 Ability to make accurate, clear and concise logs, records, status boards, and reports. 2.9 1 2.1.22 Ability to determine Mode of Operation. 2.8 1 Category Total: 4 Equipment Control 2.2.30 Knowledge of RO duties in the control room during fuel handling such as alarms from fuel handling area / communication with fuel storage facility / systems operated from the control room in support of fueling operations / and supporting instrumentation.

2.2.13 Knowledge of tagging and clearance procedures.

I Radiation Control 2.3.1 1 Ability to control radiation releases. 2.7 1 2.3.2 Knowledge of facility ALARA program. 2.5 1 2.3.9 Knowledge of the process for performing a containment purge. 2.5 1

~~ ~~

Category Total: 3 Emergency Plan 2.4.24 Knowledge of loss of cooling water procedures. 3.3 1 2.4.5 Knowledge of the organization of the operating procedures network for normal, abnormal, 2.9 1 and emergency evolutions.

2.4.6 Knowledge symptom based EOP mitigation strategies. 3.1 1 2.4.17 Knowledge of EOP terms and definitions. 3.1 1 Category Total: 4 Generic Total: 13 1

ES-401 BWR RO Examination Outline Printed: 05/13/2003 Facilitv: Monticello Nuclear Generating Plant Form ES-401-2 Exam Date: 094 512003 Exam Level: RO I I I WA Category Points

3. Generic Knowledge And Abilities 1 c1: ~ Ca: 1 Ca13 4 I 13 Note:

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

1