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, 1 OCFR55.40(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 1 OCFR 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 I permits 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: 763-295-51 51 Fax: 763-295-1454

USNRC Paae 2 Nuclear Management Company, LLC 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 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 copy for the SRO examination)

If you have any questions regarding the enclosed information please contact Gerald M.

Allex, Operations Training Instructor, (763-271 -2654 or 763-295-1 563), or John Fields (763-295-1 663).

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 - 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 and I copy for the SRO examination)

Item ES-201 Examination Outline FOIITI ES-201-2 Quality Checklist Initials Task Description I

I j

i I,

-96 H

F I

1 -

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

2.

S

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.

I M

b. 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 cornpromisingexam integrjty; 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.

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

3.

a. Venfythat:

W

/

T (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.

I

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

4.

G E

N E

R A

L -

. 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 <e 4@d4Ll&LJ,J@4d45hFd@

MONTICELLO INITIAL EXAM OUTLINE REVIEW COMMENTS Review of Q/A Checklist ES-201-2 Item 1.a:

Item I

.b:

Item 1.c:

Item 2:

Item 3.a(2):

Item 3.a(3):

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.

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

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.

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.

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.

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 A. I A. 2 A. 3 A.4 Administrative TopidSubject Descriotion Plant Parameter Verification Conduct Valve Lineup Surveillance Testing Radiation Control Calculating Exposure Emergency Communications Describe method of evaluation

1. ONE Administrative JPM, OR

2. TWO Administrative Questions JPM Perform Core Thermal Limits Monitoring Procedure C.2-05 Generic 2.1.19 (3.0)

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

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

JPM Expected Dose Determination to Inspect Equipment Procedure 4AWI-08.04.01 Generic 2.3.1 (2.6)

JPM Public Address Announcement of a General Emergency 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 4.1 4.2 9.3 4.4 Administrative TopidSu bject Description Plant Parameter Verification Crew Composition Surveillance Testing Radiation Control Calculating Exposure Emergency Plan Protective Action Recommendation Describe method of evaluation

1. ONE Administrative JPM, OR

2. TWO Administrative Questions JPM Perform Core Thermal Limits Monitoring Procedure C.2-05 Generic 2.1.I 9 (3.0), 2.1.33 (4.0)

JPM Crew Staffing Determination Procedure OWI-01.06; Procedure 4 AWI-08.01.01 Generic 2.1.4 (3.4)

JPM Perform the Daily Off-Gas Hydrogen Analyzer Checks Procedure 0000-H; Test 0209; ODCM-03.01 Generic 2.2.12 (3.4)

JPM Expected Dose Determination to Inspect Equipment Procedure 4AWI-08.04.01 Generic 2.3.1 (3.0)

JPM Classify an Event and Determine Protective Action Recommendations Procedure A.2-101; Procedure A.2-204 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 D, s Date of Facility: MNGP Examination: 9115103 Exam Level (circle one) @) I Operating Test No.: 03-01 6

I B.I Control Systems

a.

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

Emergency Startup the Reactor Protection Motor Generators JPM-B. 09.1 2-00 1 212000 K1.04 (3.413.6); 2.1.30 (3.9/3.4),

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

b.

D, R

c.

System 1 JPM Title 1

7 9

a.

Inadvertent Control Rod Insertion J PM-B.05.05-006 201003 A3.01 (3.713.6)

b.

Reject Water From Rx Vessel Using RWCU to Radwaste 204000 A I.07 (2.912.9)

J PM-B. 02.02-005 JPM-B. 02.04-004 239001 A4.01 (4.214.0)

d.

Core Spray Isolation After Pump Trip 209001 A2.01 (3.413.4)

JPM-B.03.01-006

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 262001 A2.07 (3.013.2)

JPM-E.2-05-001

g.

IRM Functional JPM-B.05.01.01-002 21 5003 A4.07 (3.613.6)

Function D,A1 I

1 I

2 I

D,A,S I 3

4

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 QUALITATWE AllRIBUTES 11 Facility:

MONTICELLO Date of Exam:

9115103 Scenario Numbers:

4 I I

Operating Test No. 03-01 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 /l/

k I,,!

A 11 TARGET QUANTITATIVE ATTRIBUTES (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 TY Pe RO Evolution Minimum Scenario Number 1

2 3

4 Type Number 1

2 2

5 3

1 1

1 5, 61 5, 6, 5961 41 6, Reactivity 1

Normal 1

4 2, 4, 31 4, 3, 4, 21 3, Instrument /

Component 9

9 8

8 7, 8 7, 8 7, 9 7

Maior 1

As RO Reactivity Normal Instrument /

Component Major As SRO 11 12 2

5

"/A N/A N/A I - 1 1 1 I

Reactivity 1

1 2

2 5

3 1

1 1

294, 314, 3,4, 2,3, 5161 59 61 5, 61 4,6i 9

9 8

8 7, 8 7, 8 7, 9 7

0 1

2 1

NIH N,

Normal 0

Instrument /

2 2, 5 3,

Reactivity Normal SRO-U Instrument I Component Major I

SRO-I I

N/A NIA N/A N/A NIA N/A N/A NIA NIA NIA N/A N/A N/A N/A N/A N/A 0

1 2

1 Instructions: (1)

(2)

(3)

Author:

NRC Reviewer:

Enter the operating test number and Form ES-D-1 event numbers for each evolution type 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.

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.

/

I n

NUREG-I 021, Revision 8, Supplement 1

Amendix D Scenario Outline Form ES-D-1 I

Facility: MNGP Scenario No. NRC 03-01 I

Op-Test No: 03-01 Operators:

I Examiners:

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)

1. I1 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 A SRV 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

ED1 2 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) 9 RC03 BOP (C)

RClC Turbine Trip (simulated via small Recirc line break)

EquiDment Out of Service

1. I 3 Service Water Pump for coupling repair.

1. I2 IRM is bwassed for erratic indication.

I

1. I 2 SBLC pimp 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 03-03 Op-Test No:

03-01 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. I2 IRM is bypassed for erratic indication.

1. I2 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 A RHR 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. I 2 IRM 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 0511 312003 K2K3 X

X X

X X

Facility:

Monticello Nuclear Generating X

X X

ES - 401 Emergenc I

I I

295007 295009 E/APE # I I

E/APE Name / Safety Function High Reactor Pressure 13 Low Reactor Water Level 12 K1 -

2950 13 295013 2950 15 295007 I High Reactor Pressure 13 I

High Suppression Pool Temperature / 5 High Suppression Pool Temperature / 5 Incomplete SCRAM I 1 295016 1295010 I High Drywell Pressure I 5 I

Control Room Abandonment I 7 295016 Control Room Abandonment 17 2950 17 295017 I High Off-Site Release Rate / 9 I

I High Off-Site Release Rate I 9 1295024 I High Drywell Pressure I 5 Ix and Abnorm A1 -

X X -

11 Plant Evolutions - Tier 1 / Group 1 EK1.O 1 - Drywell integrity: Plant-Specific Form ES-401-1

-I- +

3.a I

1 4.0 1

4.2*

1 4.1*

1 3.7*

1 4.6*

1 3.7 1

4.2* 1 1

1

BWR SRO Examination Outline Facility: Monticello Nuclear Generating A1 X

X X

ES - 401 Emereencv and Abnormal Plant Evolutions - Tier 1 / GrouD 1 A2 X

X X

E/APE ## I E/APE Name / Safety Function I K1 IK21K3 I

I I

I 295025 High Reactor Pressure / 3 295025 High Reactor Pressure / 3 295025 I High Reactor Pressure / 3 X M 295025 I High Reactor Pressure / 3

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~

295026 295026 Suppression Pool High Water Temperature 1 5 Suppression Pool High Water Temperature / 5 X

3.5 3.8 4.2*

3.7 4.2*

4.1 295030 I Low Sumression Pool Water Level / 5 I

l l

1 1

1 1

1 1

295030 I Low Suppression Pool Water Level / 5 I

1x1 29503 1 295037 295037 295038 500000 29503 1 I Reactor Low Water Level / 2 I I Ix Reactor Low Water Level / 2 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown / 1 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown / 1 High Off-Site Release Rate 19 High Containment Hydrogen Concentration / 5 X

WACategory Totals:

3 6

5 5

5 Printed 05/13/2003 G I KA Topic I

characteristics, reactor behavior, and instrument I EK1.05 - ?Exceeding safety limits EA2.02 - Suppression pool level EK3.04 - TSBLC injection I EA1.05 - HPCI I EK2.08 - SRV discharge submergence I EK3.01 - Automatic demessurization system actuation EA 1.10 - Control rod drive EA2.02 - Reactor water level EA2.06 - Reactor pressure EA1.07 - Control room ventilation: Plant-Specific I EKl.O 1 - Containment integrity Form ES-401-1 4.4 I

~

4.1*

I 3.9 I 1 2

Group Point Total:

26 2

BWR SRO Examination Outline Printed 05/13/2003 Facility:

Monticello Nuclear Generating Eme ind Abnormal Plant Evolutions - Tier 1 / Group 2 Form 5-401-:

Points ES - 401 E/APE #

295001

~

E/APE Name / Safety Function Partial or Complete Loss of Forced Core Flow Circulation / 1 Loss of Main Condenser Vacuum I 3 KA Tonic

[mp.

3.3 AK2.04 - Reactor/turbine pressure regulating system:

Plant-Specific 1

295002 AA2.02 - Reactor power: Plant-Specific 3.3 1

295004 AA1.02 - Systems necessary to assure safe plant shutdown 4.1 1

Partial or Complete Loss of D.C. Power / 6 295005 AA1.02 - RPS 3.6 3.3 4.3 3.5 Main Turbine Generator Trip / 3 295008 High Reactor Water Level / 2 High Drywell Temperature / 5 AK2.11 - Main steam 2.4.21 - Knowledge of the parameters and logic used to 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.

AK1.O 1 - Pressure/temperature relationship 295012 High Drywell Temperature / 5 295012 Partial or Complete Loss of Component Cooling Water / 8 295018 AK3.05 - Placing standby heat exchanger in service 3.3 1

2950 19 AA1.03 - Instrument air compressor power supplies 3.O 4.0 Partial or Complete Loss of Instrument Air / 8 2.4.16 - Knowledge of EOP implementation hierarchy and coordination with other sumort mocedures.

High Drywell Temperature / 5 295028 295028 EAl.05 - ADS 3.7 1

High Drywell Temperature / 5 1

BWR SRO Examination Outline Facility:

Monticello Nuclear Generating A1 A2 X

Printed:

0511312003 G KATopic Imp.

Points EA2.03 - DwelVcontainment water level 3.5 1

ES - 401 Emereencv and Abnormal Plant Evolutions - Tier 1 / Groun 2 Form ES-401-E/APE #

295029 295029 295033 295034 295035 295036 E/APE Name / Safety Function K1 K2 K3 High Suppression Pool Water Level I 5 High Suppression Pool Water Level 15 X

High Secondary Containment Area Radiation Levels /

X 9

Secondary Containment Ventilation High Radiation I X

9 Secondary Containment High Differential Pressure I 5 Secondary Containment High SumpIArea Water X

Level I 5 EK3.03 - Reactor SCRAM EK1.02 - Personnel protection EK3.02 - Starting SBGTIFRVS: Plant-Specific 3.5 1

4.2*

1 4.1 1

X EK3.O 1 - Emergency depressurization I 2.8 I 1 I I I EA1.02 - SBGTIFRVS 3.8 1

I I

I I

I 5

2 2

Group Point Total:

17 2

BWR SRO Examination Outline K6 Printed 05/13/2003 A1 X

Facility:

42 X

X X

Monticello Nuclear Generating A3 X

X ES - 401 I K4 X

S s/Ev#

203000 t 203000 K5

/209001 G

(209001 KATopic Imp.

Points K2.03 - Initiation logic 2.9*

1 K4.05 - Pump minimum flow A2.05 - Core spray line break A3.04 - System status lights and alarms System I Evolution Name K1 RHRILPCI: Injection Mode (Plant Specific) 12 RHR/LPCI: Injection Mode (Plant Saecific) 12 2.6 1

3.6 1

3.8 1

K2 -

212000 215004 2 15005 216000 2 16000 223001 223001 223002 X -

K1.04 - LPRM channels Low Pressure Core Spray System 12 I I

3.6 1

Low Pressure Core Spray System 12 Reactor Protection System 17 Source Range Monitor (SRM) System 17 A1.O1 - Recorders and meters K1.03 - Containment/drywell atmosphere control A2.03 - Safetylrelief valve leaking or stuck open Average Power Range MonitorLocal Power Range Monitor System / 7 Nuclear Boiler Instrumentation I 7 3.3 1

3.3 1

4.2*

1 Nuclear Boiler Instrumentation / 7 2.8*

Primary Containment System and X

Auxiliaries 15 Primary Containment System and Auxiliaries I 5 Primary Containment Isolation SystemINuclear Steam Supply Shut-Off I 5 1

Plant ystems - Tier 2 I Group 1 Form ES-401-A2.17 - Keep fill system failure I 3.5 1 1

IK3*01 - RPs 1 3. 4 1 1 K3.14 - High pressure coolant injection:

Plant-Specific I 4.2* I 1

A3.03 - SPDSIERISICRIDSIGDS:

Plant-Specific I

I I

1

BWR SRO Examination Outline System / Evolution Name Primary Containment Isolation System/Nuclear Steam Supply Shut-Off / 5 RHR/LPCI: Containment Spray System Mode / 5 Printed 05/13/2003 K1 K2 K3 Facility:

Monticello Nuclear Generating 24 1000 ES - 401 ReactorITurbine Pressure Regulating System 13 Plant Svstems - Tier 2 / Grouo 1 Form ES-401-1 241 000 259002 26 1000 262001 264000 s y s m #

223002 Reactor/Turbine Pressure Regulating System / 3 Reactor Water Level Control System /

2 Standby Gas Treatment System / 9 A.C. Electrical Distribution 16 X

Emergency Generators (DieseVJet) / 6 22600 1 290001

~

Secondary Containment / 5 239002 X

I RelieflSafety Valves / 3 I

l l

A3.02 - Normal building differential pressure:

3.5 1

Plant-Specific 290001 I Secondary Containment / 5 I

l l

I l l I

I I

I X I I

I I I I I IK4.03 -Fluid leakage collection I 2.9 I 1

2

3 B

Y R rn

Printed:

05/13/2003 K1 X

BWR SRO Examination Outline K2 X

Facility:

Monticello Nuclear Generating ES - 401 K5 X

Plant Systems - Tier 2 / Group 2 Form ES-401-1 K6 X

41 X

X A2 X

43 X

A4 X

K3 X

K4 Imp.

2.9 KA Topic Al.03 - Rod movement sequence lights Points 1 -

1 K5.02 - Valve operation 2.9 205000 Shutdown Cooling System (RHR Shutdown Cooling Mode) 14 Shutdown Cooling System (RHR Shutdown Cooling Mode) 14 Al.05 - Reactor water level 3.4 1

205000 2 15002

~~~

~

Rod Block Monitor System / 7 1

A2.01-Withdrawal of control rod in high 215002 Rod Block Monitor System 17 RHR/LPCI: Torus/Suppression Pool Cooling Mode I 5 RHR/LPCI: TorusISuppression Pool Smav Mode I 5 K2.03 - APRM channels: BWR-3,4,5 2.9 A3.01 - Valve operation 3.3 A4.09 - Indicating lights and alarms 3.3 219000 230000 3.3 230000 RHlULPCI: TorusISuppression Pool Spray Mode I5 K1.05 - A.C. electrical 1

234000 Fuel Handling Equipment I 8 K6.04 - ?Refueling platform air system:

Plant-Specific 2.1.33 - Ability to recognize indications for system operating parameters which are entry-level conditions for technical specifications.

3.7 4.0 234000 Fuel Handling Equipment I 8 K3.02 - ?Off-site radioactive release rate I 3.9 27 1000 Offgas System / 9 1

1

Facility:

Monticello Nuclear Generating ES - 401 Sys/Ev #

27 1000 B W R SRO Examination Outline Plant Systems - Tier 2 / Group 2 System / Evolution Name K1 K2 Offgas System / 9 Printed: 05/13/2003 Form ES-401-1 K3 K4 K5 K6 A1 A2 A3 A4 G KATopic Imp.

Points X

K4.05 - Redundancy 2.6 1

400000

~~

~~~

K/A Category Totals:

1 1

Component Cooling Water System (CCWS) / 8 X 2.2.2 - Ability to manipulate the console 3.5 1

controls as required to operate the facility between shutdown and designated power levels.

1 1

1 1

2 1

1 1

2

~

~~

Group Point Total:

13 2

Facility:

Monticello Nuclear Generating ES - 401 KA Topic A1.03 - Valve status: Mark-I&II(Not-BWR1)

BWR SRO Examination Outline Plant Systems - Tier 2 / Group 3 Imp.

Points 2.8 1

Printed: 05/13/2003 Form ES-401-1 A4.01 - Sump integrators K4.04 - Moisture removal fiom generated steam WACategory Totals:

0 0

1 1

0 0

1 0

0 1

0 3.6 1

2.8 1

K3.01 - Reactor water level I 3.3 1 1

~~

Group Point Total:

4 1

Generic Knowledge and Abilities Outline (Tier 3)

BWR SRO Examination Outline for auxiliary systems thatare outside the control room handling systems).

radiation releases.

ALARA program.

Facilitv:

Monticello Nuclear Generating 2.9 2.9 Printed 05/13/2003 Form ES-401-5 Generic Category Conduct of Operations Equipment Control KA KATopic Imp.

Points 2.1.7 2.1.11 2.1.32 2.1.18 2.1.22 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

Knowledge of less than one hour technical specification action statements for systems.

Ability to explain and apply system limits and precautions.

Ability to make accurate, clear and concise logs, records, status boards, and reports.

Ability to determine Mode of Operation.

4.4 3.8 3.8 3.0 3.3 Category Total:

5 2.2.13 2.2.26 2.2.24 2.2.32 1

1 1

1

1. -

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

Knowledge of tagging and clearance procedures.

Knowledge of refueling administrative requirements.

Ability to analyze the affect of maintenance activities on LCO status.

Knowledge of the effects of alterations on core configuration.

3.3 3.8 3.7 3.8 3.3 1

1 1

1 1

Category Total:

5 Radiation Control Category Total:

3 1

Generic Knowledge and Abilities Outline (Tier 3)

BWR SRO Examination Outline 2.4.24 2.4.5 2.4.6 2.4.30 Printed: 05/13/2003 Form ES-401-5 Knowledge of loss of cooling water procedures.

3.7 1

Knowledge of the organization of the operating procedures network for normal, abnormal, Knowledge symptom based EOP mitigation strategies.

4.0 1

Knowledge of which events related to system operations/status should be reported to 3.6 1

and emergency evolutions.

3.6 1

outside agencies.

Facility:

Monticello Nuclear Generating Generic Category Emergency Plan KA KATopic Imp.

Points 2

ES-401 BWR SRO Examination Outline Facility:

Monticello Nuclear Generating Plant Exam Date: 09/15/2003 Printed: 05/13/2003 Form ES-401-1 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 Facility:

Monticello Nuclear Generating Printed 0511312003 Emergency and Abnormal Plant Evolutions - Tier 1 / Grouu 1 Form ES-401-2 ES - 401 Imp.

3.2

?oints 1 -

G I KA Topic I

E/APE #

295006 AK3.06 - Recirculation pump speed reduction:

Plant-Suecific SCRAM/ 1 295006 1

1 -

AA2.06 - Cause of reactor SCRAM AK2.03 - RHR/LPCI: Plant-SDecific SCRAM1 1 3.5 3.1 295007 High Reactor Pressure 13 I

Ix I

295007 1

High Reactor Pressure / 3 Low Reactor Water Level 12 X

AA2.02 - Reactor power AK2.03 - Recirculation system AK2.02 - DrywelVsuppression chamber differential pressure: Mark-I&II AA1.05 - Drywelllsuppression vent and purge 4.1*

3.1 3.3 3.1 4.1 295009 High Drywell Pressure 15 I

I x 2950 10 295010 1

1 -

High Drywell Pressure / 5 295024 EKl.01 - Drywell integrity: Plant-Specific High Drywell Pressure 15 I X I 295025 High Reactor Pressure 13 X 2.1.7 - Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

1 3.7 3.9 3.6 4.0 3.3 -

EK3.O 1 - Automatic depressurization system actuation EA 1.10 - Control rod drive Reactor Low Water Level 12 Reactor Low Water Level 12 29503 1 29503 1 1

1 -

295037 1

SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown / 1 EA2.06 - Reactor pressure 1 -

EK1.O1 - Containment integrity 500000 WACategoryTotals:

2 3

2 2

3 1

Group Point Total:

13 1

B W R RO Examination Outline Printed 0511312003 E/APE Name I Safety Function Partial or Complete Loss of Forced Core Flow Circulation I 1 Loss of Main Condenser Vacuum 13 High Reactor Water Level 12 Facility:

Monticello Nuclear Generating K1 K2 X

X ES - 401

~

High Drywell Temperature I 5 High Suppression Pool Temperature / 5 High Suppression Pool Temperature 1 5 Control Room Abandonment / 7 Control Room Abandonment I 7 High Off-Site Release Rate I 9

~

X X

X Form ES-401-2 I Plant Evolutions - Tier 1 / Group 2 I

I I

I KA Topic I Imp. I Points I

I EIAPE #

29500 1 3.3 I '

AK2.04 - Reactorlturbine pressure regulating system:

Plant-Specific X I AA2.02 - Reactor power: Plant-Specific I 3.2 I 1 295002 295008 1

3.1 3.3 3.6 3.8 4.0*

3.5 AK2.11 - Main steam AK1.01 - Pressureltemperature relationship AK3.01 - Suppression pool cooling operation AA2.0 1 - Suppression pool temperature 295012 1

X X

X X

X X

295013 295013 1

295016 AK2.02 - Local control stations: Plant-Specific AK3.03 - Disabling control room controls 1

2950 16 295017 1 -

AK2.12 - Standby gas treatmentRRVS 3.4 3.2 AK3.05 - Placing standby heat exchanger in service 2.1.20 - Ability to execute procedure steps.

4.3 2950 18 1

Partial or Complete Loss of Component Cooling Water I 8 Partial or Complete Loss of Component Cooling Water I 8 Partial or Complete Loss of Instrument Air / 8 2950 18 1

1 1 -

AA1.03 - Instrument air compressor power supplies EK3.04 - tSBLC injection 295019 295026 Suppression Pool High Water Temperature / 5 High Drywell Temperature / 5 I (EAl.05-ADS 1

295028 1

BWR RO Examination Outline Facility:

Monticello Nuclear Generating A1 X

X ES - 401 EIAPE #

295030 295030 295033 A2 G 295034 K1 X

295038 K2 K3 X

X Emer EIAPE Name I Safety Function Low Suppression Pool Water Level I 5 Low Suppression Pool Water Level / 5 High Secondary Containment Area Radiation Levels I 9

Secondary Containment Ventilation High Radiation I 9

High Off-Site Release Rate 19 ormal Plant Svolutions - Tier 1 / Grom 2 Printed 05/13/2003 KA Topic EA1.05 - HPCI EK2.08 - SRV discharge submergence EK1.02 - Personnel protection EK3.02 - Starting SBGTRRVS: Plant-Specific EA1.07 - Control room ventilation: Plant-SDecific Form ES-401-2 3.5 3.5 I )

3.6 11 WACategory Totals:

2 5

5 4

2 1

Group Point Total:

19 2

Facility:

Monticello Nuclear Generating ES - 401 295032 295035 BWR RO Examination Outline Emereencv and Abnormal Plant Evolutions - Tier 1 / Groua 3 High Secondary Containment Area Temperature I 5 X

Secondary Containment High Differential Pressure 15 r

I I

I I

X E/APE # 1 E/APE Name / Safety Function IK11K21K3 I

I I

I EK2.05 - Temperature sensitive instrumentation 3.2 EAl.02 - SBGTIFRVS 3.8 295036 Secondary Containment High SumpIArea Water Level 15 295036 I Secondary Containment High Sump/Area Water I

I I X I Level 1 5 I

I I

Printed:

05/13/2003 Form ES-401-2 A1 IA2 IGIKATopic I Imp.

I 1

1 I

1 X I I EA2.03 - Cause of the high water level 1 3.4 I

I 1

I I

I I

I I I I EK3.01 - Emergency depressurization I 2.6 Points 1

1 WACategoryTotals:

0 1

1 1

1 0

Group Point Total:

4 1

BWR RO Examination Outline 43 X

X Printed:

05/13/2003 A4 X

X X

Facility:

Monticello Nuclear Generating US Form ES-401-2 Plant Svstems - Tier 2 / Grow 1 K6 ES - 401

<3 X

K4 X

X 41 !A2

~

KA Topic A3.03 - System pressure i p. I Points 2.7 I 1

1201002 I Reactor Manual Control System / 1 AI.03 - Rod movement sequence lights X

X I

1202002 1 Recirculation Flow Control System / 1 K4.06 - Recirculation pump adequate NPSH:

Plant-Specific K2.03 - Initiation logic A2.17 - Keep fill system failure A4.10 - System pumps: BWR-2,3,4 K4.05 - Pump minimum flow 203000 RHR/LPCI: Injection Mode (Plant Specific) 12 I

3*3 I 203000 RHIULPCI: Injection Mode (Plant Specific) 12 206000 High Pressure Coolant Injection System / 2 Low Pressure Core Spray System 12 Reactor Protection System / 7 Intermediate Range Monitor (IRM)

System I 7 Source Range Monitor (SRM) System I 7 209001 2 12000 2 15003 215004 Y-2.6 I 1

3.9* I 1

A3.04 - System status lights and alarms 3.4 I 1

A4.05 - Trip bypasses t

K3.01 - RPS 12 15004 I Source Range Monitor (SRM) System 17 A4.04 - SRM drive control switches K1.04 - LPRM channels Power Range Monitor System 17 1

BWR RO Examination Outline 43 X

Printed: 05/13/2003 A4 X

Facility:

ES - 401 System / Evolution Name Monticello Nuclear Generating K1 sys/Ev #

2 15005 K2 216000 K3 K4 X

X X

216000 2 17000 3.4 3.O 3.2 4.0 223001 1

1 1

1 223001 2.5*

2.9 2.7*

3.9 223002 1

1 1

1 223002 239002 239002 241000 Average Power Range Monitor/Local I

Nuclear Boiler Instrumentation / 7 Nuclear Boiler Instrumentation / 7 Reactor Core Isolation Cooling System (RCIC) / 2 Primary Containment System and Primary Containment System and Auxiliaries / 5 Primary Containment Isolation SystemJNuclear Steam Supply Shut-Off / 5 Primary Containment Isolation System/Nuclear Steam Supply Shut-Off / 5 Relieusafety Valves 13 I

Relieusafety Valves / 3 ReactodTurbine Pressure Regulating System I 3 Plant Svstems - Tier 2 / Group 1 m A1.05 - Lights and alarms K3.14 - High pressure coolant injection:

Plant-Specific A 1.O 1 - Recorders and meters K4.04 - Prevents turbine damage: Plant-Specific 1

I K1.03 - Containment/drywell atmosphere control A2.03 - Safetyhelief valve leaking or stuck lopen A3.03 - SPDS/ERIS/CRIDS/GDS:

Plant-Specific K4.05 - Single failures will not impair the function ability of the system K6.03 - A.C. power: Plant-Specific A4.06 - Reactor water level I

Al.07 - Bypass valve position Form ES-401-2 Imp. j Points 3.3 I 1

3.8 1 1

1 2

BWR RO Examination Outline K4 Printed:

05l1312003 K5 K6 A1 X

X Facility:

Monticello Nuclear Generating ES - 401 42 X

I 7

I l

l A3 A4 Sys/Ev # I System / Evolution Name IK1 IK2 IK3 I

I I

I 2

Imp.

Points KA Topic WACategory Totals:

2 2

2 259001 259002 26 1000 Plant Reactor Feedwater System I 2 Reactor Water Level Control System I 2

Standby Gas Treatment System I 9 4

1 2

4 ystems - Tier 2 / Group 1 Form ES-401-2 3

3 4

I 3-3 I K5.04 - Turbine inlet pressure vs. reactor I Dressure K2.0 1 - Reactor feedwater pump(s):

A2.05 - Loss of applicable plant air systems 1

Group Poi It Total:

28 3

I 43 X

BWR RO Examination Outline A4 X

X Printed:

0511312003 I

I Facility:

ES - 401 1

1 Monticello Nuclear Generating I I Plant Systems - Tier 2 / Group 2 Form ES-401-2 1

1 Al.02 - System flow 3.5 sysmv #

20 1006 G I KA Topic I

Imp.

System /Evolution Name K1 Points 1

1 1

1 1

1 1

1 1

1 1

I Rod Worth Minimizer System (RWM)

(Plant SDecific) I 7 2.9 -

Al.03 - Latched group indication:

K5.02 - Valve operation 205000 2.8 Shutdown Cooling System (RHR Shutdown Cooling Mode) 14 Shutdown Cooling System (RHR Shutdown Cooling Mode) 14 Rod Block Monitor System I 7 205000 Al.05 - Reactor water level I

3.4 2 15002 3.3 A2.01-Withdrawal of control rod in high 2 15002 2.8 219000

3.3 RHRLPCI

TorusISuppression Pool cool in^ Mode I 5 A3.0 1 - Valve operation 2 19000 RHRILPCI: TorusISuppression Pool Cooling Mode I 5 22600 1 3.3 RHR/LPCI: Containment Spray System Mode I 5 A4.13 - Containment/drywell temperature 230000 RHR/LPCI: TorusISuppression Pool X

Spray Mode I 5 K1.05 - A.C. electrical 3.2 2.6 3.3 245000 Main Turbine Generator and Auxiliary A4.10 - Hydrogen gas pressure K2.01 - Off-site sources of power 26200 1 1

BWR RO Examination Outline I

I I

I Printed: 05/13/2003 I

I I

i i

Facility:

Monticello Nuclear Generating ES - 401 A3 X

~~

I I

I I

I A4 G X

lK1 I= IK3 lK4 KA Topic 2.4.10 - Knowledge of annunciator response procedures.

A2.0 1 - Under voltage Sys/Ev #

262001 I I I I D.C. Electrical Distribution / 6 263000 I System / Evolution Name A.C. Electrical Distribution / 6 262002 I

I 1

I I

I Unintermptable Power Supply (A.C./D.C.) / 6 Cooling Water System X

I I I I I

K5.O 1 - Hydrogen generation during battery charging K3.02 - ?Off-site radioactive release rate K4.05 - Redundancy K4.03 - Fluid leakage collection A3.02 - Noma1 building differential pressure:

Plant-Specific 2.2.2 - Ability to manipulate the console controls as required to operate the facility between shutdown and designated power levels.

KlACategory Totals:

1 2

1 2

271000 271000 290001 29000 1 Offgas System / 9 X

Offgas System / 9 X

Secondary Containment I 5 X

Secondary Containment I 5 Imp.

3.O -

2.6 2.6 3.3 2.6 2.8 3.5 4.0 -

Points 1

1 1

1 1

1 2

0 3

2 2

2 2

Group Point Total:

19 2

I m w

d I

a Y z 8

c1 C

.I n

a 8

0 1

0 0

w w

0 1

0 0

0 I

v1 Y

m z e e

d 0

Y 3

Generic Knowledge and Abilities Outline (Tier 3)

BWR RO Examination Outline Knowledge of less than one hour technical specification action statements for systems.

Ability to explain and apply system limits and precautions.

Ability to make accurate, clear and concise logs, records, status boards, and reports.

Ability to determine Mode of Operation.

Printed 05/13/2003 Form ES-401-5 3.O 3.4 2.9

2.8 Facilitv

Monticello Nuclear Generating Ability to control radiation releases.

Knowledge of facility ALARA program.

Knowledge of the process for performing a containment purge.

Generic Category Conduct of Operations 2.7 1

2.5 1

2.5 1

KA 2.1.1 1 2.1.32 2.1.18 2.1.22 Equipment Control 2.2.30 2.2.13 Radiation Control I Radiation Control I 2.3.1 1 2.3.2 2.3.9 2.3.1 1 2.3.2 2.3.9 Emergency Plan 2.4.24 2.4.5 2.4.6 2.4.17 KA Topic Imp.

Points -

1 1

1 1

Category Total:

4 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.

Knowledge of tagging and clearance procedures.

~~

~~

Category Total:

3 Knowledge of loss of cooling water procedures.

Knowledge of the organization of the operating procedures network for normal, abnormal, and emergency evolutions.

Knowledge symptom based EOP mitigation strategies.

Knowledge of EOP terms and definitions.

3.3 2.9 3.1 3.1 1

1 1

1 Category Total:

4 Generic Total: 13 1

ES-401 BWR RO Examination Outline Facilitv:

Monticello Nuclear Generating Plant Printed: 05/13/2003 Form ES-401-2 Exam Date: 094 512003 Exam Level: RO I

I WA Category Points I

1 c:1

~

Ca:

1 Ca13

3. Generic Knowledge And Abilities 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