outline submittal for the monticello initial examination ...es-201-2, examination outline quality...

OUTLINE SUBMITTAL

FOR THE MONTICELLO INITIAL EXAMINATION - OCT 2002

Outline Submittal

Contains the following:

Outline Submittal Letter from Licensee

ES-201-1 Letter ES-201-2 ES-301 -1 ES-301 -1 ES-301-2 ES-301-4 ES-301-5 D-1 ES-401 -1 ES-401-2 ES-401 -10 Admin

Examination Preparation Checklist Exelon cover letter transmitting the Outline Examination Outline Quality Checklist Administrative Topics Outline (RO) Administrative Topics Outline (SRO) Control Room and Facility Walk-Through Test Outline (RO/SRO(l)) Simulator Scenario Quality Checklist Transient and Event Checklist Dynamic Simulator Scenario Outline for 2 scenarios BWR SRO Examination Outline BWR RO Examination Outline Record of Rejected K/As There were no NRC Comments on the submitted test outlines

ES-201 Examination Preparation Checklist Form ES-201-1

Facility: Monticello Nuclear Generating Station Date of Examination: October 14, 2002

Examinations Developed by: [ Facilityl / NRC (circle one)

Target Chief Date* Task Description / Reference Examiner's

Initials

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

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

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

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

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

-75 6. Integrated examination outline(s) due (C.1.e & f; C.3.d) drm

-70 7. Examination outline(s) reviewed by NRC and feedback provided drm to facility licensee (C.2.h; C.3.e)

-45 8. Proposed examinations, supporting documentation, and drm reference materials due (C.1 .e, f, g & h; C.3.d)

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

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

-14 11. Examination approved by NRC supervisor for facility licensee drm

review (C.2.h; C.3.f)

-14 12. Examinations reviewed with facility licensee (C.I.j; C.2.f & h; C.3.g) drm

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

-7 14. Final applications reviewed; assignment sheet updated; waiver drm letters sent (C.2.g, ES-204)

15. Proctoring/written exam administration guidelines reviewed with drm -7 facility licensee and authorization granted to give written exams

(if applicable) (C.3.k) -7 16. Approved scenarios, job performance measures, and questions drm

distributed to NRC examiners (C.3.i)

* Target dates are keyed to the examination date identified in the corporate notification letter. They are for planning purposes and may be adjusted on a case-by-case basis in coordination with the facility licensee. Applies only to examinations prepared by the NRC.

NMC' Monticello Training Center Committed to Nuclear Excellence Operated by Nuclear Management Company, LLC

April 30, 2002

Attention: Dell McNeil US Nuclear Regulatory Commission, Region III 801 Warrenville Road Lisle, Illinois 60532-4351

Subject: Monticello Nuclear Generating Plant (Docket No. 50-263) Initial Licensing Examination to be Conducted the Week of September 9, 2002.

In accordance with NUREG 1021, Operator Licensing Examination Standards for Power Reactors, Revision 8, Supplement 1, enclosed for your review are the proposed examination outlines for the initial license examinations for our operator license applicants. In accordance with NUREG 1021, ES-201, Attachment 1, "Examination Security and Integrity Guidelines," the enclosed materials are to be withheld from public disclosure until after the examinations are complete.

The proposed examination outlines were prepared per the guidelines of NUREG 1021, sections ES-301 and ES-401. Proposed outlines have been prepared to support development, by the Monticello facility, of examinations for one Reactor Operator license candidate and two "instant" Senior Reactor Operator license candidates.

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 Walk-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 3 total) ES-401-1, BWR SRO Examination Outline ES-401-2, BWR RO Examination Outline Monticello Suppressed K/A Report

The proposed outlines prepared for the initial site-specific written examinations were randomly generated using the software application developed by the BWR Owners' Group Generic Cross Reference Committee, titled "BWR K/A Database Program," version 1.07.

Prior to generating the examination outlines, the K/A catalog was screened to eliminate K/A statements that were not applicable to the Monticello facility. A listing of the K/A statements that were suppressed from selection for outline generation. is provided in the enclosed report, "Monticello Suppressed K/A Report."

2100 West River Streete Monticello, Minnesota 55362-9637 Telephone: 763.295.4141 * Fax: 763.295.1592

NMC Committed to Nuclear Excellence

Monticello Training Center Operated by Nuclear Management Company, LLC

If you have any questions regarding the enclosed information please contact Mary Engen, Operations Training Instructor, at 763-271-2622 or 763-295-1563, or via electronic mail at [email protected].

Kurt Markling Operations Shift Supervisor Operations Department Training Liaison

2100 West River Street* Monticello, Minnesota 55362-9637 Telephone: 763.295.4141 e Fax: 763.295.1592

ES-201 Examination Outline Form ES-201-2 Quality Checklist

Facility: Date of Examination:

Initials Item Task Description

a b* c#

1. a. Verify that the outline(s) fit(s) the appropriate model per ES-401. J(, W R b. Assess whether the outline was systematically and randomly prepared in accordance with

Section D.1 of ES-401 and whether all K/A categories are appropriately sampled.,, __, _ T T c. Assess whether the outline over-emphasizes any systems, evolutions, or generic topics. Ow l E N d. Assess whether the justifications for deselected or rejected K/A statements are appropriate.

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

S I b. Assess whether there are enough scenario sets (and spares) to test the projected number and M mix of applicants in accordance with the expected crew composition and rotation schedule without

compromising exam integrity; ensure each applicant can be tested using at least one new or Jy M significantly modified scenario, that no scenarios are duplicated from the applicants' audit test(s)*, and scenarios will not be repeated over successive days.

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

3. a. Verify that: (1) the outline(s) contain(s) the required number of control room and in-plant tasks,

W (2) no more than 30% of the test material is repeated from the last NRC examination, Y Tem / (3)* no tasks are duplicated from the applicants' audit test(s), and T (4) no more than 80% of any operating test is taken directly from the licensee's exam banks.

b. Verify that: (1) the tasks are distributed among the safety function groupings as specified in ES-301, (2) one task is conducted in a low-power or shutdown condition, (3) 40% of the tasks require the applicant to implement an alternate path procedure, I V (4) one in-plant task tests the applicant's response to an emergency or abnormal condition, and (5) the in-plant walk-through requires the applicant to enter the RCA.

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

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

4. a. Assess whether plant-specific priorities (including PRA and IPE insights) are covered in the R Vappropriate exam section.

G E b. Assess whether the 10 CFR 55.41/43 and 55.45 sampling is appropriate. N E c. Ensure that K/A importance ratings (except for plant-specific priorities) are at least 2.5. . _

R A d. Check for duplication and overlap among exam sections. L e. Check the entire exam for balance of coverage. f "

f. Assess whether the exam fits the appropriate job level (RO or SRO).

Printed Name.I Signatj reDate a. Author 0-4_v,,,/ V I'jKrK. , -2

b. Facility Reviewer (*) M.A t R,, I A; 6- o c. NRC Chief Examiner(#) "-tt .N . 4 ,<,.4t_ Z., -Vv L d. NRC Supervisor 6r -A

Note: Not applicable for NRC-developed examinations. # Independent NRC reviewer initial items in Column "c;" chief examiner concurrence required.

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

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

Facility: Monticello Date of Examination: 0910912002 Exam Level (circle one) RO / SRO Operating Test Number: 02-01

Administrative Describe method of evaluation Topic/Subject 1. ONE Administrative JPM, OR

Description 2. TWO Administrative Questions

A.1 Plant Parameter JPM - Perform Core Thermal Limits monitoring Verification Procedure: C.2-05, Rev. 13

Generic 2.1.19 (3.0)

Shift JPM - Perform a Control Room Shift Turnover Checklist and Turnover panel walkdown and identify any discrepancies.

Procedure: Checklist 3139, Rev. 24

Generic 2.1.31 (4.2)

A.2 Surveillance JPM - Perform a Daily Jet Pump Operability test Testing Procedure: 0000-B, Rev. 83

Generic 2.2.12 (3.0)

A.3 Radiation JPM - Expected dose determination to inspect reactor leak Control Procedure: 4 AWI-08.04.01, Rev. 12

Calculating Exposure

Generic 2.3.4 (2.5)

A.4 Emergency JPM - Make a Public Address (PA) announcement of a General Communications Emergency declaration.

Procedure: A.2-105, Rev. 15

Generic 2.4.43 (2.8)

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

Facility: Monticello Date of Examination: 09/09/2002 Exam Level (circle one) RO I O Operating Test Number: 02-01

Administrative Describe method of evaluation Topic/Subject 1. ONE Administrative JPM, OR

Description 2. TWO Administrative Questions

A.1 Plant Parameter JPM - Perform Core Thermal Limits Monitoring and identify actions Verification that need to be taken.

Procedure: C.2-05, Rev. 13

Generic 2.1.19 (3.0), 2.1.33 (4.0)

Temporary JPM - Temporary Change Request Procedure Procedure: 4 AWI-02.02.05, Rev. 9; 0255-04-lA-1, Rev. 52

Modification

Generic 2.1.20 (4.2)

A.2 Temporary JPM - Initiate a Jumper Bypass Form Modification Procedure: 4AWI-04.04.03, Rev. 16; B.03.02-05, Rev. 14; of Systems Jumper Bypass Form 3034, Rev. 20

(Bank) Generic 2.2.11 (3.3)

A.3 Radiation JPM - Expected dose determination to inspect reactor leak Control Procedure: 4 AWI-08.04.01, Rev. 12

Calculating Exposure

Generic 2.3.4 (3.1)

A.4 Emergency Plan JPM - Classify an event and determine Protective Action Protective Action Recommendations Recommendation Procedure: A.2-101, Rev. 27; A.2-204, Rev. 14

(Modified Bank) 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

Facility: Monticello Nuclear Date of Examination: 09/09/2002 Exam Level (circle one) R- SRO / SRO(U) Operating Test No.: 02-01

B.1 Control Systems

System / JPM Title Type Safety Code* Function

a. Control Rod Drive (CRD) Operability D, A, S 1 JPM-B.01.03-006 201003, A2.01 (3.4/3.6), A4.02 (3.5/3.5)

b. Start a Reactor Feedpump D, S 2 JPM-B.06.05-001 259001, A4.02 (3.9/3.7)

c. MSIV Exercise Test D, A, S 3 JPM-B.02.04-003 239001, A4.01 (4.2/4.0)

d. HPCI Reactor Vessel Pressure Control N, A, S 4 JPM B.03.02-xxx 206000, A1.02 (4.2/4.2)

e. Reset a Group 2 Isolation N, S 5 JPM C.4-B.04.011.B3-xxx

223002, A4.03 (3.6/3.5)

f. Manually Start #11 EDG (Control Room Actions) D, S 6 JPM-B.09.08-001 264000, A4.04 (3.7/3.7)

g. Rod Worth Minimizer Operability D, S 7 JPM-B.05.02-001 201006, A3.02 (3.5/3.4)

B.2 Facility Walk-Through

a. Manual Control of RHRSW CV-1 729 D, A, L, R 8 JPM B.08.01.03-001 400000, A1.01 (2.8/2.8) Abnormal

b. FWCVControl D, R 2 JPM-B.05.07-002 259001, A2.07 (3.7/3.8)

c. Start a Mechanical Vacuum Pump D, R 9 JPM-B.06.03-002 271000, A2.01 (3.1/3.1) {295002, AA1.02 (2.9/2.9)} Abnormal

*Type Codes: (D)irect from bank, (M)odified from bank, (N)ew, (A)lternate path, (C)ontrol room, S)imulator, (L)ow-Power, (R)CA

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

Facility: MONTICELLO Date of Exam: 9/9/2002 Scenario Numbers: 1 I 2 I 3 Operating Test No. 02-01

QUALITATIVE ATTRIBUTES Initial

a b* c#

1. The initial conditions are realistic, in that some equipment and/or instrumentation may be out of . •

service, but does not cue the operators into expected events.

2. The scenarios consist mostly of related events. _ _ ,

3. Each event description consists of * the point in the scenario when it is to be initiated * the malfunction(s) that are entered to initiate the event MI'A-/f. * the symptoms/cues that will be visible to the crew * the expected operator actions (by shift position) • the event termination point (if applicable)

4. No more than one mechanistic failure (e.g., pipe break) is incorporated into the scenario without a credible preceding incident such as a seismic event.

5. The events are valid with regard to physics and thermodynamics. P et, A 6. Sequencing and timing of events is reasonable, and allows the examination team to obtain

complete evaluation results commensurate with the scenario objectives.

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 V4p, W/4,4 tA given.

8. The simulator modeling is not altered.

9. The scenarios have been validated. Any open simulator performance deficiencies have been 'V41 evaluated to ensure that functional fidelity is maintained while running the planned scenarios.

10. Every operator will be evaluated using at least one new or significantly modified scenario. All c J other scenarios have been altered in accordance with Section D.4 of ES-301.

11. All individual operator competencies can be evaluated, as verified using Form ES-301-6 (submit ,/* the form alongl with the simulator scenarios).

12. Each applicant will be significantly involved in the minimum number of transients and events cht specified on Form ES-301-5 (submit the form with the simulator scenarios.

13. The level of difficulty is appropriate to support licensing decisions for each crew position. • % TARGET QUANTITATIVE ATTRIBUTES (PER SCENARIO; SEE SECTION D.4.D) Actual Attributes -

1. Total malfunctions (5-8) 7 / 5 / 7__ _

2. Malfunctions after EOP entry (1-2) 1 / 1 / 1 A 3. Abnormal events (2-4) 2 / 2 / 2 $ g=

4. Major Transients (1-2) 2 / 1 / 2 *t, 5. EOPs entered/requiring substantive actions (1-2) 2 / 2 / 1 ,--,

6. EOP contingencies requiring substantive actions (0-2) 2 / 1 / 1

7. Critical tasks (2-3) 2 / 2 / 2

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

ES-301 Transient and Event Checklist Form ES-301-5

OPERATING TEST NO.: 02-01

Applicant Evolution Minimum Scenario Number Type Type Number 1 2 3 4

Reactivity 1 2 6 2

Normal 1 1 1 1

RO Instrument/ 4 3, 4, 2, 3, 3, 4,

Component 5,6, 4,5, 5,6, 9 8,9 9

Major 7,8 7 7,8

Reactivity 1 2 6 2

Normal 0 NA NA NA

As RO Instrument! 2 3.5 3,5, 4,6,

Component 8, 9 9

Major 1 7,8, 7 7,8

SRO-]

Reactivity 0 2 6 2

Normal 1 1 1 1

As SRO Instrument! 2 3, 4, 2, 3, 3,4,

Component 5,6, 4,5, 5,6, 9 9 9

Major 1 7,8, 7 7,8

Reactivity 0 NA NA NA

Normal 1 NA NA NA

SRO-U Instrument 2 NA NA NA

Component

Major 1 NA NA NA

Instructions:

Author:

NRC Reviewe

C1) Enter the operating test number and each evolution type

Form ES-D-1 event numbers for

(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.

r:

NUREG-1021, Revision 8, Supplement 1

Appendix D Scenario Outline Form ES-D-1

Facility: Monticello Scenario No. #2

Examiners:

Op-Test No:

Operators:

Initial Conditions: The plant is at 100% power. RCIC surveillance is being completed for PMT. The "B" CRD FCV is out of service to repair an air line leak. No. 11 RHR and RHRSW pumps are in service for Torus cooling, and LPCI LCO has been entered. All other systems are normal.

Turnover: Complete RCIC surveillance.

Event Malf. Event Event No. No. Type* Description 01 BOP-N Secure from RCIC Surveillance 02 BOP-C Trip of No. 11 RHR Pump causing RHR to drain down to the

Torus and making RHR unavailable until filled and vented. 03 RO-C CRD Pump Failure 04 BOP-I Hydrogen Cooler Water Controller fails partially closed 05 RO-C Leak into RBCCW from RWCU causing the manual isolation

of RWCU. 06 RO-R Loss of No. 11 Condensate Pump with loss of Bus 13

causing loss of No. 11 Circ Water Pump. Essential bus 15 will automatically transfer to the 1AR. No. 11 Feed Pump also trips.

07 M Loss of No. 12 Condensate Pump causes loss of No. 12 Feed Pump with loss of feed and reactor scram with ATWS.

08 C "A" CRD FCV fails closed. 09 RO-C Selected SBLC Pump fails to start, and other SBLC must be

started.

T t f

t I +

t I ±

i i t

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

02-01

Appendix D Scenario Outline Form ES-D-1

Facility: Monticello Scenario No. #3 Op-Test No: 02-01

Operators:

Initial Conditions: The plant is approximately 18% 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.1 is complete through step VI.C.4. No. 12 Reactor Feedpump is out of service, but should be returned to service in time to support power increase. No. 13 Service Water Pump is out of service to repair coupling.

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

Event Malf. Event Event No. No. Type* Description 01 BOP-N Synchronize Main Generator to the Grid 02 RO-R Increase power by pulling rods 03 BOP-C No. 11 Service Water Pump Trip 04 RO-C Hotwell Low Level 05 BOP-C Steam Packing Exhauster High Temperature 06 RO-I Channel "4" APRM Fails High 07 M Design Basis Earthquake 08 M Nuclear Boiler reference line breaks, causing loss of all level

indication. 09 RO-C Loss of No. 13 Bus

Note 1 Event 5: If required to get into Tech Spec more could have a failure of B RPS to trip. Problem is that simulator requires a full scram failure (PP04)and 5-B-5 does not come in. If PP04 is removed, to allow a full scram, then 5-B-5 would come in which could be confusing.

t 1 1-

r 1 1-

t I ±

F I +

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

Examiners:

Facility:

ES - 401

BWR SRO Examination OutlineMonticello Nuclear Generating

Emergency and Abnormal Plant Evolutions - Tier 1 / Groun I

E/APE # E/APE Name / Safety Function KI K2 K3 Al A2 G KA Topic Imp. Points

295003 Partial or Complete Loss of A.C. Power / 6 X AK2.02 - Emergency generators 4.2* 1

295003 Partial or Complete Loss of A.C. Power / 6 X AA1.01 - A.C. electrical distribution system 3.8 1

295006 SCRAM / 1 X AK1.01 - Decay heat generation and removal. 3.9 1

295006 SCRAM / I X AK3.03 - Reactor pressure response 3.9* 1

295010 High Drywell Pressure / 5 X 2.4.8 - Knowledge of how the event-based 3.7 1 emergency/abnormal operating procedures are used in conjunction with the symptom-based EOPs.

295013 High Suppression Pool Temperature / 5 X AA2.01 - Suppression pool temperature 4.0 1

295013 High Suppression Pool Temperature / 5 X AK2.01 - Suppression pool cooling 3.7 1

295014 Inadvertent Reactivity Addition / 1 X AK1.01 - Prompt critical 3.8 1

295015 Incomplete SCRAM / I X AK3.01 - Bypassing rod insertion blocks 3.7 1

295017 High Off-Site Release Rate / 9 X AK3.01 - System isolations 3.9 1

295023 Refueling Accidents / 8 X 2.2.21 - Knowledge of pre and post maintenance 3.5 1 operability requirements.

295023 Refueling Accidents / 8 X AAI.02 - Fuel pool cooling and cleanup system 3.1 1

295024 High Drywell Pressure / 5 X EK1.01 - Drywell integrity: Plant-Specific 4.2* 1

295024 High Drywell Pressure / 5 X EA2.02 - Drywell temperature 4.0 1

Page 1 of 13

Printed: 04/25/2002

Monticello Nuclear GeneratingBWR SRO Examination Outline

Emergency and Abnormal Plant Evolutions - Tier 1 / Groun I

E/APE # E/APE Name / Safety Function KI K2 K3 Al A2 G KA Topic Imp. Points

295025 tligh Reactor Pressure / 3 X EK1.04 - Decay heat generation 3.9 1

295025 High Reactor Pressure / 3 X EA2.04 - Suppression pool level 3.9 1

295026 Suppression Pool High Water Temperature / 5 X EA1.03 - Temperature monitoring 3.9 1

295026 Suppression Pool High Water Temperature / 5 X 2.1.10 - Knowledge of conditions and limitations in the 3.9 1 facility license.

295030 Low Suppression Pool Water Level / 5 X EA2.03 - Reactor pressure 3.9 1

295031 Reactor Low Water Level / 2 X 2.1.32 - Ability to explain and apply system limits and 3.8 1 precautions.

295031 Reactor Low Water Level / 2 X EA 1.06 - Automatic depressurization system 4.4* 1

295037 SCRAM Condition Present and Reactor Power Above X EK2.07 - Neutron monitoring system 4.0 1 APRM Downscale or Unknown / 1

295037 SCRAM Condition Present and Reactor Power Above X EK3.04 - Hot shutdown boron weight: Plant-Specific 3.7 1 APRM Downscale or Unknown / 1

295038 High Off-Site Release Rate / 9 X EK2.03 - Plant ventilation systems 3.8 1

295038 High Off-Site Release Rate / 9 X EK3.03 - Control room ventilation isolation: 3.9 1 Plant-Specific

500000 High Containment Hydrogen Concentration / 5 X EA2.03 - Combustible limits for drywell 3.8 1

K/A Category Totals: 4 4 5 4 5 4 Grdup Point Total: 26

Page 2 of 13

Facility:

ES - 401

Printed: 04/25/2002

I•n rrn WqANII

Monticello Nuclear GeneratingBWR SRO Examination Outline

Emergency and Abnormal Plant Evolutions - Tier I / GrouD 2

E/APE # E/APE Name / Safety Function K1 K2 K3 Al A2 G KA Topic Imp. Points

295001 Partial or Complete Loss of Forced Core Flow X AK3.06 - Core flow indication 3.0 1 Circulation / 1

295002 Loss of Main Condenser Vacuum / 3 X AA1.01 - Condensate system 2.6 1

295008 High Reactor Water Level / 2 X AK 1.03 - Feed flow/steam flow mismatch 3.2 1

295008 High Reactor Water Level / 2 X 2.1.6 - Ability to supervise and assume a management 4.3 1 role during plant transients and upset conditions.

295012 High Drywell Temperature / 5 X AK2.02 - Drywell cooling 3.7 1

295018 Partial or Complete Loss of Component Cooling X AA2.02 - Cooling water temperature 3.2 1 Water / 8

295021 Loss of Shutdown Cooling / 4 X AA2.05 - Reactor vessel metal temperature 3.3 1

295022 Loss of CRD Pumps / 1 X AK2.02 - CRD mechanism 3.1 1

295028 High Drywell Temperature / 5 X EA1.02 - Drywell ventilation system 3.8 1

295029 High Suppression Pool Water Level / 5 X EA2.02 - Reactor pressure 3.6 1

295029 High Suppression Pool Water Level / 5 X EK1.01 - Containment integrity 3.7 1

295032 High Secondary Containment Area Temperature / 5 X EK3.01 - Emergency/normal depressurization 3.8 1

295033 High Secondary Containment Area Radiation Levels / X EK2.01 - Area radiation monitoring system 4.0 1 __________ 9

Page 3 of 13

Facility:

ES - 401

Printed: 04/25/2002

]?nrrn I•_A[ilI

Facility: Monticello Nuclear Generating

ES - 401

BWR SRO Examination Outline

Emergency and Abnormal Plant Evolutions - Tier 1 / Group 2

K/A Category Totals: 3 3 3 2 3 3

E/APE # E/APE Name / Safety Function KI K2 K3 Al A2 G KA Topic imp. Points

295034 Secondary Containment Ventilation High Radiation / X 2.4.10 - Knowledge of annunciator response procedures. 3.1 1 9

295034 Secondary Containment Ventilation High Radiation / X EK3.05 - Manual SCRAM and depressurization: 3.9 1 9 Plant-Specific

295035 Secondary Containment High Differential Pressure / 5 X EKI .01 - Secondary containment integrity 4.2* 1

600000 Plant Fire On Site / 8 X 2.3. 10 - Ability to perform procedures to reduce 3.3 1 excessive levels of radiation and guard against personnel exposure.

17

Page 4 of 13

Group Point Total:

Printed: 04/25/2002

Fnrm F.N-Afi I-1

Monticello Nuclear Generating


Plant Systems - Tier 2 / Group I

Printed: 04/25/2002

Sys/Ev # System / Evolution Name KI K2 K3 K4 K5 K6 Al A2 A3 A4 G KATopic Imip. Points

202002 Recirculation Flow Control System / 1 X K1.03 - Reactor core flow 3.7 1

203000 RHR/LPCI: Injection Mode (Plant X K6.05 - Condensate storage and transfer system: 2.5 1 Specific) / 2 Plant-Specific

203000 RHR/LPCI: Injection Mode (Plant X 2.1.7 - Ability to evaluate plant performance 4.4 Specific) / 2 and make operational judgments based on

operating characteristics, reactor behavior, and _instrument interpretation.

211000 Standby Liquid Control System / 1 X A2.01 - Pump trip 3.8*

215004 Source Range Monitor (SRM) System X K2.01 - SRM channels/detectors 2.8 1 /7

215004 Source Range Monitor (SRM) System X A4.01 - SRM count rate and period 3.8 1 /7

215005 Average Power Range Monitor/Local X K2.02 - APRM channels 2.8 Power Range Monitor System / 7

215005 Average Power Range Monitor/Local X K4.01 - Rod withdrawal blocks 3.7 Power Range Monitor System / 7

216000 Nuclear Boiler Instrumentation / 7 X K1.09 - Redundant reactivity control/ alternate 4.0 rod insertion; Plant-Specific

218000 Automatic Depressurization System / X A1.03 - ADS valve air supply pressure: 3.4* 3 Plant-Specific

223001 Primary Containment System and X K6.12 - D.C. electrical distribution 3.0 Auxiliaries / 5

Page 5 of 13

Facility:

ES - 401 IPnrm I•.•.A[I1 - 1



Plant Systems - Tier 2 / Group 1

Sys/Ev # System / Evolution Name K1 K2 K3 K4 K5 K6 Al A2 A3 A4 G KATopic Imp. Points

223002 Primary Containment Isolation X K4.05 - Single failures will not impair the 3.1 1 System/Nuclear Steam Supply function ability of the system Shut-Off/ 5

223002 Primary Containment Isolation X A3.01 - System indicating lights and alajms 3.4 1 System/Nuclear Steam Supply Shut-Off/ 5

226001 RHR/LPCI: Containment Spray X A3.05 - Containment pressure 4.0* 1 System Mode / 5

239002 Relief/Safety Valves / 3 X K5.06 - Vacuum breaker operation 3.0 1

241000 Reactor/Turbine Pressure Regulating X K3.02 - Reactor pressure 4.3" System / 3

241000 Reactor/Turbine Pressure Regulating X K5.04 - Turbine inlet pressure vs. reactor 3.3 1 System / 3 pressure

259002 Reactor Water Level Control System / X A1.05 - FWRV/startup level control position: 2.9 1 2 Plant-Specific

259002 Reactor Water Level Control System / X A2.05 - Loss of applicable plant air systems 3.4 1 2

262001 A.C. Electrical Distribution / 6 X 2.2.21 - Knowledge of pre and post 3.5 1 maintenance operability requirements.

262001 A.C. Electrical Distribution / 6 X A4.01 - All breakers and disconnects (including 3.7 available switch yard): Plant-Specific

290001 Secondary Containment / 5 X I ! f KK3.01 - tOff-site radioactive release rates 4.4"

Page 6 of 13

Facility:

ES - 401

Printed: 04/25/2002

Inrm l•£-l[lll


ES - 401



Printed: 04/25/2002

K/A Category Totals: 2 2 2 2 2 2 2 2 2 2 3 Group Point Total: 23

Page 7 of 13

Sys/Ev# System / Evolution Name KI 1K2 K3 K4 K5 K6 Al A2 A3 A4 G KATopic Imp. Points

290001 Secondary Containment / 5 X 2.4.48 - Ability to interpret control room 3.8 1 indications to verify the status and operation of system, and understand how operator action s and directives affect plant and system conditions.

•nrm IT•~AilII


ES - 401



Printed: 04/25/2002

Form ES-401-1

Sys/Ev # System / Evolution Name KI K2 K3 K4 K5 K6 Al A2 A3 A4 G KA Topic Imp. Points

201001 Control Rod Drive Hydraulic System / X K4. 11 - Protection against filling the SDV 3.6 1 1 during non-SCRAM conditions

202001 Recirculation System / 1 X 2.1.33 - Ability to recognize indications for 4.0 1 system operating parameters which are entry-level conditions for technical specifications.

215002 Rod Block Monitor System / 7 X K5.01 - Trip reference selection: Plant-Specific 2.8 1

215002 Rod Block Monitor System / 7 X A1.01 - Trip reference: BWR-3, 4, 5 2.8 1

230000 RHR/LPCI: Tonis/Suppression Pool X K3.02 - Suppression pool temperature 3.5 1 Spray Mode / 5

230000 RHR/LPCI: Torus/Suppression Pool X K5.01 - System venting 2.7 1 Spray Mode / 5

245000 Main Turbine Generator and Auxiliary X A2.09 - Turbine vibration 2.8 1 Systems / 4

272000 Radiation Monitoring System / 7 X K2.05 - Reactor building ventilation monitors: 2.9 1 Plant-Specific

272000 Radiation Monitoring System / 7 X A3.09 - Containment isolation indications 3.5

286000 Fire Protection System / 8 X K 1.09 - Emergency generator rooms: 3.3 1 Plant-Specific

290003 Control Room HVAC / 9 X 2.3.10 - Ability to perform procedures to reduce 3.3 1 excessive levels of radiation and guard against personnel exposure.

Page 8 of 13



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


Page 9 of 13


400000 Component Cooling Water System X K6.04 - Pumps 3.1 1 (CCWS) / 8

400000 Component Cooling Water System X A4.01 - CCW indications and control 3.0 1 (CCWS) / 8

Printed: 04/25/2002



Plant Systems -Tier 2 / Group 3

Printed: 04/25/2002

Form ES-401-1


Page 10 of 13

Facility:

ES - 401

Sys/Ev # System / Evolution Name K1 K2 K3 K4 K5 K6 Al A2 A3 A4 G KA Topic Imp. Points

201003 Control Rod and Drive Mechanism / I X 2.2.25 - Knowledge of bases in technical 3.7 1 specifications for limiting conditions for operations and safety limits.

201003 Control Rod and Drive Mechanism / 1 X K1.01 - Control rod drive hydraulic system 3.3 1

233000 Fuel Pool Cooling and Clean-up / 9 X K4.07 - Supplemental heat removal capability 2.9

233000 Fuel Pool Cooling and Clean-up / 9 X A1.07 - System temperature 2.8 1

Generic Knowledge and Abilities Outline (Tier 3)

BWR SRO Examination OutlinePrinted: 04/25/2002

Form ES-401-5

Imp. PointsKA Topic

Conduct of Operations 2.1.4 Knowledge of shift staffing requirements. 3.4 1

2,1.22 Ability to determine Mode of Operation. 3.3 1

2.1.33 Ability to recognize indications for system operating parameters which are entry-level 4.0 1 conditions for technical specifications.

2.1.31 Ability to locate control room switches, controls and indications and to determine that 3.9 1 they are correctly reflecting the desired plant lineup. I

Category Total: 4

Equipment Control 2.2.27 Knowledge of the refueling process. 3.5 1

2.2.32 Knowledge of the effects of alterations on core configuration. 3.3 1

2.2.30 Knowledge of RO duties in the control room during fuel handling such as alarms from 3.3 1 fuel handling area / communication with fuel storage facility / systems operated from the control room in support of fueling operations / and supporting instrumentation.

2.2.11 Knowledge of the process for controlling temporary changes. 3.4* 1

2.2.2 Ability to manipulate the console controls as required to operate the facility between 3.5 1

shutdown and designated power levels.

Category Total: 5

Radiation Control 2.3.9 Knowledge of the process for performing a containment purge. 3.4 1

2.3.2 Knowledge of facility ALARA program. 2.9 1

2.3.1 Knowledge of 10 CFR 20 and related facility radiation control requirements. 3.0 1

2.3.7 Knowledge of the process for preparing a radiation work permit. 3.3 1 11Category Total: 4

Page 11 of 13


Generic Category KA




Generic Category

Printed: 04/25/2002

Form ES-401-5

- Imp. Points

Category Total:

Generic Total:

4

17

Page 12 of 13

KA KA Topic

Emergency Plan 2.4.38 Ability to take actions called for in the facility emergency plan, including (if 4.0 1 required)supporting or acting as emergency coordinator.

2.4.17 Knowledge of EOP terms and definitions. 3.8 1

2.4.8 Knowledge of how the event-based emergency/abnormal operating procedures are used in 3.7 1 conjunction with the symptom-based EOPs.

2.4.50 Ability to verify system alarm setpoints and operate controls identified in the alarm 3.3 1 response manual. I

ES-401 BWR SRO Examination Outline

Facility: Monticello Nuclear Generating Plant

Exam Date: 09/09/2002

Printed: 04/25/2002

Form ES-401-1

Exam Level: SRO

K/A Category Points

Tier Group

KI

�1 I r

K2 K3 K4 K5 I K6 Al

4 4 5 4 5

Emergency & 2 3 3 3 2 3

Abnormal Plant Tier

Evolutions Totals 7 7 8 6 8

2.

Plant Systems

2 2 2 2 2 2 2 2

A2 A3 A4

2 2

2 1 1 1 1 2 1 1 1 1 1 2 13

3 1 0 0 1 0 0 1 0 0 0 1 4

Tier Totals 4 3 3 4 4 3 4 3 3

Cat I Cat 2 Cat 3 Cat 4 3. Generic Knowledge And Abilities

4 5 4 4 17

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

within each tier.

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

3. Select topics from many systems; avoid selecting more than two or 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.

Page 13 of 13

Point Total

26

17

43

23

G

4

3

7

3

3 6 40

I

Facility:

ES - 401

BWR RO Examination OutlineMonticello Nuclear Generating

Emergency and Abnormal Plant Evolutions - Tier 1 / Gr•rnin 1

E/APE # E/APE Name / Safety Function K1 K2 K3 Al A2 G KATopic Imp. Points

295006 SCRAM / 1 X AK1.01 - Decay heat generation and removal. 3.7 1

295006 SCRAM / 1 X AK3.03 - Reactor pressure response 3.8 1

295014 Inadvertent Reactivity Addition / 1 X AK 1.01 - Prompt critical 3.7 1

295015 Incomplete SCRAM / 1 X AK3.01 - Bypassing rod insertion blocks 3.4 1

295024 High Drywell Pressure / 5 X EK.OI - Drywell integrity: Plant-Specific 4.1 1

295024 High Drywell Pressure / 5 X EA2.02 - Drywell temperature 3.9 1

295025 High Reactor Pressure / 3 X EK 1.04 - Decay heat generation 3.6 1

295031 Reactor Low Water Level / 2 X EA1.06 - Automatic depressurization system 4.4* 1

295031 Reactor Low Water Level / 2 X EK2.10 - Redundant reactivity control: Plant-Specific 4.0 1

295037 SCRAM Condition Present and Reactor Power Above X EK2.07 - Neutron monitoring system 4.0* 1 APRM Downscale or Unknown / 1

295037 SCRAM Condition Present and Reactor Power Above X EK2.04 - SBLC system 4.4* 1 APRM Downscale or Unknown / I

500000 High Containment Hydrogen Concentration / 5 X EA2.03 - Combustible limits for drywell 3.3 1

500000 High Containment Hydrogen Concentration / 5 X EA 1.05 - Wetwell sprays 3.3 1

K/A Category Totals: 4 3 2 2 2 0 Group Point Total: 13

Page 1 of 12

Printed: 04/25/2002

Monticello Nuclear GeneratingBWR RO Examination Outline

Emergency and Abnormal Plant Evolutions - Tier I / C~ronnn-___ -. - -~-1- -~ ul11 Emrgnc an bora lntEouios-TirI/Gru

E/APE Name / Safety Function

Facility:

ES - 401

E.. Im. Points 295001 Partial or Complete Loss of Forced Core Flow X AK3.06 - Core flow indication 2.9 1

Circulation / I

295002 Loss of Main Condenser Vacuum / 3 X AA1.01 - Condensate system 2.6 1

295003 Partial or Complete Loss of A.C. Power / 6 X AK2.02 - Emergency generators 4.1* 1

295003 Partial or Complete Loss of A.C. Power / 6 X AA1.01 - A.C. electrical distribution system 3.7 1

295008 High Reactor Water Level / 2 X AK1.03 - Feed flow/steam flow mismatch 3.2 1 295012 High Drywell Temperature / 5 X AK2.02 - Drywell cooling 3.6 1

295013 High Suppression Pool Temperature / 5 X AK2.01 - Suppression pool cooling 3.6 1

295017 High Off-Site Release Rate /9 X AK3.01 - System isolations 3.6 1

295017 High Off-Site Release Rate / 9 X AKI.02 - tProtection of the general public 3.8* 1

295022 Loss of CRD Pumps / 1 X AK2.02 - CRD mechanism 3.1 1

295026 Suppression Pool High Water Temperature / 5 X I BEA1.03 - Temperature monitoring 3.9* 1

295026 Suppression Pool High Water Temperature/5 X 2.1.10 - Knowledge of conditions and limitations in the 2.7 1 _ facility license.

295028 High Drywell Temperature / 5 X EA1.02 - Drywell ventilation system 3.9 1

295029 High Suppression Pool Water Level / 5 X EK1.01 - Containment integrity 3.4 1

Page 2 of 12

ES-401-2

E/APE #

Printed: 04/25/2002

Y


ES - 401

BWR RO Examination Outline


K/A Category Totals: 3 7 2 6 0 1 Group Point Total: 19

Page 3 of 12

Form ES-401-2

E/APE # E/APE Name / Safety Function KI1 K2 K3 Al A2 G KA Topic Imp. Points

295033 High Secondary Containment Area Radiation Levels / X EK2.01 - Area radiation monitoring system 3.8 1 9

295033 High Secondary Containment Area Radiation Levels / X EAl.04 - SBGT/FRVS 4.2* 9

295034 Secondary Containment Ventilation High Radiation I X EK2.04 - Secondary containment ventilation 3.9 1 9

295038 High Off-Site Release Rate / 9 X EK2.03 - Plant ventilation systems 3.6 1

295038 High Off-Site Release Rate / 9 X EA1.03 - Process liquid radiation monitoring system 3.7 1

Printed: 04/25/2002


ES - 401



K/A Category Totals: 1 0 1 1 0 1

E/APE # E/APE Name / Safety Function K1 K2 K3 Al A2 G KA Topic Imp. Points

295021 Loss of Shutdown Cooling / 4 X AK3.01 - Raising reactor water level 3.3 1

295023 Refueling Accidents / 8 X AAI.02 - Fuel pool cooling and cleanup system 2.9 1

"295032 High Secondary Containment Area Temperature / 5 X 2.4.48 - Ability to interpret control room indications to 3.5 1 verify the status and operation of system, and understand how operator action s and directives affect plant and system conditions.

295035 Secondary Containment High Differential Pressure / 5 X EK1.01 - Secondary containment integrity 3.9 1

4

Page 4 of 12

Group Point Total:

Printed: 04/25/2002

Form E•-4fll-•


ES - 401


Plant Systems - Tier 2 / Group I

Printed: 04/25/2002

Form ES-401-2

Sys/Ev # System / Evolution Name KI K2 K3 K4 KS K6 Al A2 A3 A4 G KA Topic Imp. Points

201001 Control Rod Drive Hydraulic System / X K4.1 I - Protection against filling the SDV 3.6 1 1 during non-SCRAM conditions

201002 Reactor Manual Control System / 1 X A3.01 - Control rod block actuation 3.2 1

202002 Recirculation Flow Control System / I X K 1.03 - Reactor core flow 3.7 1

202002 Recirculation Flow Control System / I X K3.02 - Reactor power 4.0 1

203000 RHR/LPCI: Injection Mode (Plant X K6.05 - Condensate storage and transfer system: 2.5 1 Specific) / 2 Plant-Specific

203000 RHR/LPCI: Injection Mode (Plant X 2.4.27 - Knowledge of fire in the plant 3.0 1 Specific) / 2 procedure.

209001 Low Pressure Core Spray System / 2 X A4.01 - Core spray pump 3.8 1

211000 Standby Liquid Control System / I X A2.01 - Pump trip 3.5 1

211000 Standby Liquid Control System / 1 X A1.02 - Explosive valve indication 3.8 1

215004 Source Range Monitor (SRM) System X K2.01 - SRM channels/detectors 2.6 1 /7

215004 Source Range Monitor (SRM) System X A4.01 - SRM count rate and period 3.9 1 /7

215005 Average Power Range Monitor/Local X K2.02 - APRM channels 2.6 1 Power Range Monitor System /7 1 1 1 17 1

Page 5 of 12


ES - 401



- Printed: 04/25/2002

Form ES-401-2

Sys/Ev# System / Evolution Name Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G KA Topic Imp. Points 215005 Average Power Range Monitor/Local X - K4.01 - Rod withdrawal blocks 3.7 1

Power Range Monitor System / 7

216000 Nuclear Boiler Instrumentation / 7 X K1.09 - Redundant reactivity control/ alternate 3.7 1 rod insertion; Plant-Specific

216000 Nuclear Boiler Instrumentation / 7 X K3.30 - Recirculation system 3.2 1

218000 Automatic Depressurization System / X A 1.03 - ADS valve air supply pressure: 3.2 1 3 Plant-Specific

223001 Primary Containment System and X K6.12 - D.C. electrical distribution 2.7 1 Auxiliaries / 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

223002 Primary Containment Isolation X A3.01 - System indicating lights and alarms 3.4 1 System/Nuclear Steam Supply Shut-Off / 5

239002 Relief/Safety Valves / 3 X K5.06 - Vacuum breaker operation 2.7 1

241000 Reactor/Turbine Pressure Regulating X K3.02 - Reactor pressure 4.2* 1 System / 3

241000 Reactor/Turbine Pressure Regulating X K5.04 - Turbine inlet pressure vs. reactor 3.3 1 System / 3 pressure

259002 Reactor Water Level Control System j X A1.05 - FWRV/startup level control position: 2.9 1 - 2 1 1 1 1 1 1 1 Plant-SpecificI I

Page 6 of 12



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


Page 7 of 12

Sys/Ev System / Evolution Name KI K2 K3 K4 K5 K6 Al A2 A3 A4 G KA Topic imp. Points

259002 Reactor Water Level Control System / X A2.05 - Loss of applicable plant air systems 3.2 1 2

261000 Standby Gas Treatment System / 9 X K1.02 - Drywell 3.2 1

261000 Standby Gas Treatment System /9 X A2.02 - High system flow 2.9 1

264000 Emergency Generators (Diesel/Jet) / 6 X K5.05 - Paralleling A.C. power sources 3.4 1

264000 Emergency Generators (Diesel/Jet) / 6 X A4.05 - Transfer of emergency generator (with 3.6 1 _ _load) to grid

Printed: 04/25/2002





201003 Control Rod and Drive Mechanism / I X KI1.01 - Control rod drive hydraulic system 3.2 1

204000 Reactor Water Cleanup System / 2 X A4.09 - Reactor water temperature 2.9 1

205000 Shutdown Cooling System (RHR X A 1.08 - Heat exchanger temperatures 3.1 Shutdown Cooling Mode) / 4

215002 Rod Block Monitor System / 7 X K5.01 - Trip reference selection: Plant-Specific 2.6 6

215002 Rod Block Monitor System / 7 X A1.01 - Trip reference: BWR-3, 4, 5 2.7 1

219000 RHR/LPCI: Torus/Suppression Pool X K4.10 - Prevention of leakage to the 3.3 1 Cooling Mode / 5 environment through system heat exchanger:

__Plant-Specific

226001 RHR/LPCI: Containment Spray X A3.05 - Containment pressure 4.0* 1 System Mode / 5

230000 RIfR/LPCI: Torus/Suppression Pool X K3.02 - Suppression pool temperature 3.3 1 Spray Mode / 5

230000 RHRILPCI: Torus/Suppression Pool X K5.01 - System venting 2.6 1 Spray Mode / 5

245000 Main Turbine Generator and Auxiliary X A2.09 - Turbine vibration 2.5 Systems / 4

262001 A.C. Electrical Distribution / 6 X A4.01 - All breakers and disconnects (including 3.4 available switch yard): Plant-Specific

271000 (Offgas System / 9 X K4.01 - Dilution of hydrogen gas concentration 2.9

Page 8 of 12

Facility:

ES - 401

Printed: 04/25/2002

Form E•-Afll.9




K/A Category Totals: 2 1 2 2 2 2 2 1 2 3 0

Form FS-401l-2

Group Point Total: 19

Page 9 of 12

Facility:

ES - 401


272000 Radiation Monitoring System / 7 X K2.05 - Reactor building ventilation monitors: 2.6 1 _ Plant-Specific

272000 Radiation Monitoring System / 7 X A3.09 - Containment isolation indications 3.6 1

286000 Fire Protection System / 8 X K1.09 - Emergency generator rooms: 3.2 1 Plant-Specific

290001 Secondary Containment / 5 X K3.01 - tOff-site radioactive release rates 4.0 1

290003 Control Room HVAC / 9 X K6.02 - Component cooling water systems 2.7 1

400000 Component Cooling Water System X K6.04 - Pumps 3.0 1 (CCWS) / 8

400000 Component Cooling Water System X A4.01 - CCW indications and control 3.1 (CCWS) / 8

Printed: 04/25/2002



ES - 401 Plant Svstems - Tier 2 G roun 3


Page 10 of 12

Sys/Ev# System / Evolution Name KI K2 K3 K4 K5 K6 Al A2 A3 A4 G KA Topic Imp. Points

233000 Fuel Pool Cooling and Clean-up / 9 X K4.07 - Supplemental heat removal capability 2-7 1

233000 Fuel Pool Cooling and Clean-up / 9 X AI.07 - System temperature 2.7 1

288000 Plant Ventilation Systems / 9 X KI.06 - Plant air systems 2.7 1

290002 Reactor Vessel Internals / 5 X 2.2.23 - Ability to track limiting conditions for 2.6 1 operations.

Printed: 04/25/2002

I•nrm I• qdfl I _')


Printed: 04/25/2002BWR RO Examination Outline

FacilitV: Monticello Nuclear Generating

KA KA Topic

Form ES-401-5

Imp. Points

Conduct of Operations 2.1.33 Ability to recognize indications for system operating parameters which are entry-level - 3.4 1 conditions for technical specifications.

2.1.31 Ability to locate control room switches, controls and indications and to determine that 4.2 1 they are correctly reflecting the desired plant lineup.

2.1.29 Knowledge of how to conduct and verify valve lineups. 3.4 1

Category Total: 3

Equipment Control 2.2.30 Knowledge of RO duties in the control room during fuel handling such as alarms from 3.5 1 fuel handling area / communication with fuel storage facility / systems operated from the control room in support of fueling operations / and supporting instrumentation.

2.2.11 Knowledge of the process for controlling temporary changes. 2.5 1

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

2.2.22 Knowledge of limiting conditions for operations and safety limits. 3.4 1

Category Total: 4

Radiation Control 2.3.2 Knowledge of facility ALARA program. 2.5 1

2.3.1 Knowledge of 10 CFR 20 and related facility radiation control requirements. 2.6 1

2.3.10 Ability to perform procedures to reduce excessive levels of radiation and guard against 2.9 1 personnel exposure.

Category Total: 3

Emergency Plan 2.4.17 Knowledge of EOP terms and definitions. 3.1 1

2.4.8 Knowledge of how the event-based emergency/abnormal operating procedures are used in 3.0 1 conjunction with the symptom-based EOPs.

2.4.50 Ability to verify system alarm setpoints and operate controls identified in the alarm 3.3 1 response manual.

Category Total:

Generic Total:

3

13

Page 11 of 12

Generic Category

ES-401 BWR RO Examination Outline

Facility: Monticello Nuclear Generating Plant

Exam Date: 09/09/2002

Printed: 04/25/2002

Form ES-401-2

Exam Level: RO

K/A Category Points

Tier Group KI K3

1 4 3 2 1. _____ ____

Emergency 2 3 7 2 &

Abnormal Plant 3 1 0 1

Evolutions

Totals 8 10 5 Tier

2.

Plant Systems

3 2 3 3 2

K4 K5 K6 Al A2 A3 A4 G Point Total

2 2 0 13

6 0 1 19

1 0 1 4

9 2 2 36

3 2 3 I 28

2 2 1 2 2 2 2 2 1 2 3 0 19

3 1 0 0 1 0 0 1 0 0 0 1 4

Tier Totals 6 3 5 6 5 4 6 4 4 6 2 51

Cat I Cat 2 Cat 3 Cat 4 3. Generic Knowledge And Abilities

3 4 3 3 13

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

within each tier.

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

3. Select topics from many systems; avoid selecting more than two or 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.

Page 12 of 12

If I • T

K2

I

(

ES-401 Record of Rejected K/As (RO Outline) Form ES-401 -10

Tier/Group Randomly Selected K/A Reason for Rejection 2/1 261000A2.02 261000A2.02 was replaced by 261000A2.01 because the High

System Flow condition does not exist at MNGP, thus the KA is not applicable to the selected system (SBGT).

i _ _ _ _ _ _ _ _ _ _ _ _

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1 1

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outline submittal for the monticello initial examination ...es-201-2, examination outline quality...

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