ML081220228

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Outline Submittal and NRC Comments for the Davis-Besse Initial Examination February 2008
ML081220228
Person / Time
Site: Davis Besse Cleveland Electric icon.png
Issue date: 02/25/2008
From: Valos N
Operations Branch III
To:
FirstEnergy Nuclear Operating Co
Shared Package
ML080920714 List:
References
50-346/08-301
Download: ML081220228 (41)
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Text

OUTLINE SUBMITTAL AND NRC COMMENTS FOR THE DAVIS-BESSE INITIAL EXAMINATION FEBRAURY 2008

FENOC 1

5501 Norill Sale Houlc 2 Oak Hnrboi Ohio 43449 419321 7676 F8x 410 327 i5R2

- WHEN SEPARATED FROM ENCLOSURE 1, HANDLE THIS Mark B Benlla V m Presrdenf N ~ k d i DOCUMENT AS UNRESTRICTED -

Docket Number 50-346 License Number NPF-3 Serial Number 1 - 1509 November 2 6, 2007 Mr. Nicholas Valos Senior Operations Engineer - Region 111 United States Nuclear Regulatory Commission 2443 Warrenville Road Suite 210 Lisle, IL 60532-4352 I O CFR 55

Subject:

Operator License Examination Outline

Dear Mr. Valos:

Enclosed is the operator license examination outline required to support the operator license examinations being administered at the Davis-Besse Nuclear Power Station (DBNPS) during the weeks of February 18 and February 25,2008. The operator license examination outline shall be withheld from public disclosure until after the scheduled examinations are complete.

Mr. Paul F. Timmerman, Senior Nuclear Operations Instructor, can respond to questions with regard to the submitted operator license examination outline, at (419) 321-75 10.

If you require additional information, please contact Mr. Raymond A. Hruby, Jr., Manager - Site Regulatory Compliance, at (419) 321-8000.

Sincerely yours, w6y LJS/s Attachment Enclosure cc:

Regional Administrator, NRC Region 111 w/o Chief - Operations Branch, NRC Region 111 w/o DB-1 NRC/NRR Senior Project Manager w/o DB-I Senior Resident Inspector w/o USNRC Document Control Desk wio Utility Radiological Safety Board w/o NOV 21 2307

- WHEN SEPARATED FROM ENCLOSURE 1, HANDLE THIS DOCUMENT AS UNRESTRICTED -

Docket Number 50-346 License Number NPF-3 Serial Number 1-1509 Attachment Page 1 of 1 COMMITMENT LIST The following list identifies those actions committed to by the Davis-Besse Nuclear Power Station in this document. Any other actions discussed in the submittal represent intended or planned actions by Davis-Besse. They are described only as information and are not regulatory commitments. Please notify the Manager - Site Regulatory Compliance (419) 321-8000 at Davis-Besse of any questions regarding this document or associated regulatory commitments.

COMMITMENTS DUE DATE None NIA

- OPERATOR LICENSE EXAMINATION OUTLINE -

WITHHOLD FROM PUBLIC DISCLOSURE UNTIL AFTER THE SCHEDULED EXAMINATIONS ARE COMPLETE Docket Number 50-346 License Number NPF-3 Serial Number 1-1509 Operator License Examination Outline (42 pages follow)

11 i a Verity that the outlme(s) fit(s) the appropriate model, In accordance mth ES-401 I,Y hole u ndependenl hRC rev ewer n,La lems n C o. n n c' Cn et exam ner concdrrence reqt. reo n, 7' b Assess rrnetner Ine OLI ne was s{slemal cally an0 ranoom i prepared n acwroancr.%.In Secl on D 1 01 ES-401 and nnelner a r( A calegor es are appropr alely samp eo T

T E

N

2.

S and major transients.

i M

U L

A T

R

c. Assess whether the outline over-emphasizes any systems, evoiutions. or generic topics.

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d. Assess whether the justifications for deselected or rejected WA statements are appropriate.
a. Using Form ES-301-5. verify that the proposed scenario sets mver the required number of normal evolutions. instrument and component failures, technical specifications.
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 compromising exam integrity, and ensure that each applicant can be tested using at least one new or significantly modified scenario, that no scenarios are duplicated from the applicants' audit test@), and that scenarios will not be repeated on subsequent 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.
a. Verify that the systems walk-through outline meets the criteria specified on Form ES-301-2:

(1) the outline@) contain(s) the required number of control room and in-plant tasks (2) task repeh'tion from the last two NRC examinations is within the limits specified on the form (3) no tasks are duplicated from the applicants'audit test@)

(4) the number of new or modified tasks meets or exceeds the minimums specified on the form (5) the number of alternate path, low-power, emergency, and RCA tasks meet the criteria on the 2.b distributed among the safety functions as specified on the form

3.

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form.

b. Verify that the administrative outline meets the criteria specified on Form ES-301-1:
11) the tasks are distributed amono the tooics as soecified on the form

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(2) at least one task is new or s i g r k n t l y modified (3) no more than one task is repeated from the last two NRC licensing examinations

c. Determine if there are enouah different outlines to lest the oroiected number and mix I,*,"

of applicants and ensure that no items are duplicated on subiequent days.

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a. Assess whether plant-specific priorities (including PRA and IPE insights) are covered
4.

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L in the appropriate exam sections.

b. Assess whether the 10 CFR 55.41143 and 55.45 sampling is appropriate.
c. Ensure that WA importance rah'ngs (except for plant-specific priorities) are at least 2.5.

I d. Check for duplication and overlap amonq exam sections.

ES-201, Rev. 9 Examination Outline Quality Checklist Form ES-201-2

a. Author
b. Facility Reviewer (*)
c. NRC Chief Examiner I#)
d. NRC Supervisor
e. Check the entire exam for balance of coverage.
f. Assess whether the exam fits the appropriate job level (RO or SRO).

pfi py:

Printed NamelSignature

Examination Outline Quality Checklist Form ES-201-2 ES-201, Rev. 9

acility: Davis-Besse Date of Examination: 211812008 Item
1.

W R

I T

T E

N

2.

S I

M U

L A

T 0

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3.

W I

T

4.

G E

N E

R A

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L Task Description a

a. Verify that the outline(s) fit(s) the appropriate model, in accordance with ES-401
b. Assess whether the outline was systematically and randomly prepared in accordance wlth Section D.l of ES-401 and whether all KIA categorles are appropriately sampled.

c Assess whether the Outline over-emphasizes any systems, evolutions. or generlc topics

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d. Assess whether the justifications for deselected or rejected KIA Statements are appropriate.

e out ine s con orm(s) with the qualitative distributed among the safety functions as specified on the form (2)task repetition from the last two NRC examinations is within the limits specified on the form (3)no tasks are duplicated from the applicants' audit test(s)

(4)the number of new or modified tasks meets or exceeds the minimums specified on [he form

b. Verify that the administrative outline meets the criteria specified on Forni ES-301-1.

m sec ions.

e. Check the entire exam for balance of coverage.

g4 I. Assess whether the exam fits the aPDroDriate iob level IRO or SROI.

wh Facility Reviewer (+)

1 !zlo.a NRC Chief Examine orI2rio0 318'11 Independent NRC reviewer initlal Items In Column "c'*, cnief examiner concdrrence req, red

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

Davis-Besse Date of Examination:

211 812008 Examination Level (circle one):

RO Operating Test Number:

NRC Administrative Topic Type (see Note)

Code' Conduct of Operations D

Conduct of Operations NIA Equipment Control M

Radiation Control I

N

I-Emergency Plan Describe activity to be performed JPM 227 - Calculate RCS with F755 unavailable 2.2.18 Knowledge of the process for managing maintenance activities NOT SELECTED JPM 148 - Review a tagout and determine it is incorrect.

2.2.13 Knowledge of tagging and clearance procedure ANI - Calculate Radiation Release using SGTL Abnormal Procedure, DB-OP-02531, Attachment 1 2.3.8 Knowledge of the process for performing a planned gaseous radioactive release JPM 241 - Make EP Offsite Notification 2.4.43 Knowledge of emergency communications and techniques NOTE:

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

"Type Codes & Criteria:

(C)ontrol room (D)irect from bank ( I 3 for ROs; 5 4 for SROs & RO retakes)

(N)ew or (M)odified from bank (> 1)

(P)revious 2 exams ( I 1; randomly selected)

(S)imulator NUREG-1021, Revision 9

ES-301 Administrative Topics

... Outline Form.

ES-301-1 DAVIS-BESSE RO NRC ADMINISTRATIVE TOPICS OUTLINE

SUMMARY

1. JPM 227 - Calculate RCS with F755 unavailable The candidate will be directed to perform DB-OP-03006, Misc. Instrument Shift Checks,, Calculation of RCS Total Flow Computer Point F577 Unavailable SETTING:

Classroom or Plant

2. JPM 148 (Modified) - Review a tagout and determine it is wrong.

The candidate will be provided a copy of a tagout and determine what is incorrect.

SETTING:

Classroom or Simulator

3. AN-I - Calculate Radiation Release using SGTL Abnormal Procedure, DE-OP-02531,, SGTL Rate Calculation Determine primary to secondary tube leak using the Steam Jet Air Ejector Radiation Monitors, RE 1003A and RE 1003B, and chemistry sheet.

SETTING:

Simulator

4. JPM 241 - Make EP Offsite Notification An ALERT has been declared. The candidate will initiate offsite notifications using the dedicated (4-way ringdown) phone. The State of Ohio Highway Patrol will not answer.

The candidate will provide the information to Ottawa and Lucas Counties. The candidate will then find the Highway Patrols phone number and call using a non-dedicated phone line.

SETTING:

Classroom or Simulator NUREG-1021, Revision 9

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

Davis-Besse Examination Level (circle one):

SRO Administrative Topic (see Note)

Conduct of Operations Conduct of Operations Equipment Control Radiation Control Emergency Plan Type Code*

D N

M N

D Date of Examination:

211 812008 Operating Test Number:

NRC Describe activity to be performed JPM 227 - Calculate RCS with F755 unavailable 2.2.18 Knowledge of the process for managing maintenance activities AN2 - Review an Auxiliary Feedwater Surveillance Test and determine Operability JPM 148 - Review a tagout and determine it is incorrect.

2.2.13 Knowledge of tagging and clearance procedure ANI - Calculate Radiation Release using SGTL Abnormal Procedure, DB-OP-02531, Attachment 1 2.3.8 Knowledge of the process for performing a planned gaseous radioactive release JPM 178 - Security Event Classification and Notification 2.4.43 Knowledge of emergency communications and techniques NOTE:

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

Type Codes & Criteria:

(C)ontrol room (D)irect from bank (5 3 for ROs; 5 4 for SROs & RO retakes)

(N)ew or (M)odified from bank (> 1)

(P)revious 2 exams (2 1; randomly selected)

(S)imulator NUREG-1021. Revision 9

ES-301 Administrative Topics Outline Form ES-301-1 DAVIS-BESSE RO NRC ADMINISTRATIVE TOPICS OUTLINE

SUMMARY

1. JPM 227 - Calculate RCS with F755 unavailable The candidate will be directed to perform DE-OP-03006, Misc. Instrument Shift Checks,, Calculation of RCS Total Flow Computer Point F577 Unavailable SETTING:

Classroom or Plant

2. AN The candidate will be provided with the Auxiliary Feedwater Surveillance Test Acceptance Criteria and determine that the AFW Pump is not operable Setting:

Classroom or Plant

3. JPM 148 (Modified) - Review a tagout and determine it is wrong.

The candidate will be provided a copy of a tagout and determine what is incorrect SETTING:

Classroom or Simulator

4. AN Calculate Radiation Release using SGTL Abnormal Procedure, DE-OP-02531. SGTL Rate Calculation Determine primary to secondary tube leak using the Steam Jet Air Ejector Radiation Monitors, RE 1003A and RE 10038, and chemistry sheet.

SETTING:

Simulator

5. JPM 178 - Security Event Classification and Notification -Time Critical The Security Supervisor reports that a Credible Threat has been reported. The candidate will classify the event, complete forms, and make NRC notification.

SETTING:

Classroom or Simulator NUREG-1021, Revision 9

ES-301 Control Room/ln-Plant Systems Outline Form ES-301-2

e.
f.
g.

Facility:

DAVIS - BESSE Date of Examination:

211 8/08 Exam Level (circle one)

RO I SRO(I)

Operating Test No.:

NRC JPM 85 - Purge Containment D, L 8

JPM - 221 - CRD Sequence Fault Reset D

1 JPM 215 - Respond to a high Station Vent radiation alarm A, D, C 9

1 Control Room Systems@ (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U)

/I 1 i.

j.

1 Type Code' I Safety Function System I JPM Title JPM 127 -Actions for steam binding of the Motor Driven D

4s Feedpump JPM 11 5 - Emergency Shutdown of EDG A, D. E, M 6

I a. 1 JPM 14 - Loss of Service Water Loop 1 to Primary loads 1

A, D, M n

11 b.

1 JPM 33 -Transfer LPI Suction to the CTMT Emergency Sump I

A, D I

3 I/ c.

1 JPM 48 Exchange RCS Flow amphynol I

D I

7 I1 I

A.D I

JPM 97 -manually Actuate SFAS after some components blocked 1 h. I NEW - Synch the Main Generator to the Grid (RO ONLY) b 1

6 U

1 Implant Systems" (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U)

Y 11 k. I New - Primary Side RO CTRM Evacuation I E, N, R I z

11 All control room (and in-plant) systems must be different and serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

Type Codes i

Criteria for RO I SRO-I / SRO-U (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (L)ow-Power (N)ew or (M)odified from bank including l(A)

(P)revious 2 exams

( R K A 4-6 14-6 / 2-3 9 1 81 4 1 / I / 1 1 / 1 / 1 2/ 2/ 1 3 I 3 I 2 (randomly selected) 1 / 1 1 1 NUREG-1021, Revision 9

ES-301

.. Control Roomlln-Plant Systems Outline Form ES-30?-'2::

a.
b.

C.

d.
e.
f.
9.
h.

I.

j.
k.

From a Mode 1 condition, the candidate will be directed to respond to a loss of Service Water to Primary loads.

Bank JPM 14 WA:

055A4.01 From a Large Break LOCA condition, the candidate will be directed to transfer the LPI Suction the Containment Emergency Sump Bank JPM 33 WA:

01 1 EA 1. I 1 From Mode 3, 2. or 1. the candidate will be directed to RCS flow instruments Bank JPM 98 WA:

039A3.02 From a LOCA condition, the candidate will be directed to actuate SFAS do to changing plant conditions and some SFAS equipment out of their SFAS position Bank JPM 97 WA:

013A4.02 From a Mode 1 condition, the candidate will be direct to start the a purge on Containment.

Bank JPM 85 KA:

029A2.03 From a plant startup condition, the candidate will be directed to recover from a Control Rod Drive Sequence Fault.

Bank JPM. 221 KA:

001 A2.14 From a any Mode condition, the candidate will be directed to respond alarm procedure, DB-OP-02009 for high station vent radiation.

JPM 215 WA:

2.3.11.2 From a Mode 1 condition, the candidate will be directed to synchronize the Main Generator to the Grid.

NEW JPM WA:

045A4.11 From a Mode 1 condition, the candidate will be directed to relieve the steam binding of the Motor Driven Feedhater Pump.

Bank JPM 127 WA:

061 A2.04 From a shutdown condition, the candidate will be direct to emergency shutdown the Emergency Diesel Generator.

Bank JPM 1 15 WA:

064.A4.06 From any plant condition, the candidate will be directed to restore the makeup system from outside the Control Room.

NEW JPM KIA:

068 A l. l NUREG-1021. Revision 9

Form ES-301-2 Control Room/ln-Plant Systems Outline

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ES-301 Facility:

DAVIS - BESSE Date of Examination:

211 8108 Exam Level (circle one)

RO I SRO(I)

Operating Test No.:

NRC Control Room Systems' (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U)

I Type Code' 1 Safety Function Svstem I JPM Title

a.

JPM 14 - Loss of Service Water Loop 1 to Primary loads A. D, M 4s

b.

JPM 33 -Transfer LPI Suction to the CTMT Emergency Sump A, D 3

c.

1 JPM 48 Exchange RCS Flow amphynol 7

d.

JPM 97 - manually Actuate SFAS after some components 2

blocked

e. I JPM 85 - Purge Containment
f.

JPM - 221 - CRD Sequence Fault Reset D

1

g.

JPM 215 -Respond to a high Station Vent radiation alarm A, D, C 9

h.

ROONLY In-Plant Systems" (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U)

I.

JPM 127 -Actions for steam binding of the Motor Driven Feedpump All control room (and in-plant) systems must be different and serve different safety functions; In-plant systems and functions may overlap those tested in the control room.

  • Type Codes Criteria for RO I SRO-I I SRO-U
A)lternate path 4-6 14-6 12-3
C)ontrol room

'D)irect from bank 9 1. 8 1 4

E)mergency or abnormal in-plant

' 1 1 ' 1 1 ' 1

L)ow-Power 1 1. 1 1 1 N)ew or (M)odified from bank including 1(A)

- 2 1 21 1 P)revious 2 exams

,R)CA

~ 1 1 ~ 1 1 1

'S)imulator

... 3 I ' 3 I, ' 2 (randomly selected)

WREG-1021. Revision 9

ES-301

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... Control Roorn/ln-Piant

- Systems-Outline.

Form ES-301-2

a.
b.

C.

d.
e.
f.
9.
h.

I.

1.
k.

From a Mode 1 condition, the candidate will be directed to respond to a loss of Service Water to Primary loads.

Bank JPM 14 WA:

055A4.01 From a Large Break LOCA condition, the candidate will be directed to transfer the LPI Suction the Containment Emergency Sump Bank JPM 33 KIA:

011 E A 1. l l From Mode 3, 2, or 1, the candidate will be directed to RCS flow instruments Bank JPM 98 WA:

039A3.02 From a LOCA condition, the candidate will be directed to actuate SFAS do to changing plant conditions and some SFAS equipment out of their SFAS position Bank JPM 97 WA:

013A4.02 From a Mode 1 condition, the candidate will be direct to start the a purge on Containment.

Bank JPM 85 KA:

029A2.03 From a plant startup condition, the candidate will be directed to recover from a Control Rod Drive Sequence Fault.

Bank JPM. 221 KA:

001 A2.14 From a any Mode condition, the candidate will be directed to respond alarm procedure, DB-OP-02009 for high station vent radiation.

JPM 215 WA:

2.3.11.2 RO ONLY JPM WA:

From a Mode 1 condition, the candidate will be directed to relieve the steam binding of the Motor Driven Feedwater Pump.

Bank JPM 127 WA:

061 A2.04 From a shutdown condition, the candidate will be direct to emergency shutdown the Emergency Diesel Generator.

Bank JPM 115 WA:

064.A4.06 From any plant condition, the candidate will be directed to restore the makeup system from outside the Control Room.

NEW JPM KIA:

068 A l. l NUREG-1021. Revision 9

ES-301, Rev. 9 Transient and Event Checklist Form ES-301-5

acility: Davis-Besse Date of Exam: 2/18/2008 Operating Test No.:

I

=

A P

P L

I C

A N

T -

LO

RO-I (1 8RO-U -
0 RO-I (21 RO-U 0

RO-I (3)

RO-U 0

RO-I (4)

RO-U E

v E

N T

T Y

P E -

X OR C

IAJ S

X OR T'

M U

Scenarios 4

2.4 2.4 4

Page 1 of 4

ES-301, Rev. 9 Transient and Event Checklist Form ES-301-5 1

I 2

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11 Faci ity Dav s-Besse Date of Exam 2/18/2008 Operating Test No 3

4 P

P L

I C

A N

T RO SRO-I (5)

SRO-U RO SRO-l (6; SRO-U RO SRO-I (71 SRO-U RO SRO-I (81 SRO-U I CREW CREW CREW POSITION POSITION POSITION E

V E

CREW POSITION N

T 4

2 0

T Y

P 1

1 1

4 2

2 1

2 2

1 1

0 E -

l X 4OR IC AAJ 4

4 2

8 2

2 Page 2 Of 4

ES-301, Rev. 9 Transient and Event Checklist Form ES-301-5 Facility Dai Besse Date of Exam 2/18/2008 Operating Test No

~

A P

P L

I C

A N

T RO SRO-I (9)

SRO-U RO SRO-I (10)

SRO-U RO (1)

SRO-l SRO-U RO (2)

SRO-I SRO-U Page 3 of 4

ES-301, Rev. 9 Transient and Event Checklist Form ES-301-5

acility: Davis-Besse Date of Exam: 2/18/2008 Operating Test No.:

0 istructions M

I N

I M

U 7 2

1 2

2 1

1 0

1 1

4 4

2 2

1 2

2

)

Check the applicant level and enter the operating test number and Form ES-D-1 event numbers for each event type: TS are not applicable for RO applicants. ROs must serve in both the "at-the-controls (ATC)"

and "balance-of-plant (BOP)" positions; Instant SROs must do one scenario, including at least two instrument or component (IC) malfunctions and one major transient, in the ATC position.

Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.5.d) but must be significant per Section C.2.a of Appendix D. (*) Reactivity and normal evolutions may be replaced with additional instrument or component malfunctions on a 1-for-1 basis.

Whenever practical, both instrument and component malfunctions should be included: only those that require verifiable actions that provide insight to the applicant's competence count toward the minimum requirements specified for the applicant's license level in the right-hand columns.

)

Page 4 of 4

ES-401 PWR Examination Outline Form ES-40?;2 -.

Facility:

Davis Besse 2008 NRC Exam Date of Exam:

2/18/2008 Ensure that at least two topics from every applicable WA category are sampled within each tier of the RO and SRO-only outlines (Le., except for one category in Tier 3 of the SRO-only outline, the "Tier Totals" in each WA category shall not be less than two).

The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +I from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points.

Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems that are not included on the outline should be added. Refer to ES-401, Attachment 2, for guidance regarding elimination of inappropriate WA statements.

Select topics from as many systems and evolutions as possible; sample every system or evolution in the arouD before selectina a second tooic for anv svstem or evolution.

Absent a plant specific priority, only those KAs having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only oortions. resoectivelv.

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

The generic (G) WAS in Tiers 1 and 2 shall be selected from Section 2 of the WA Catalog, but the topics must be relevant to the applicable evolution or system.

On the following pages, enter the WA numbers, a brief description of each topic, the topics' importance ratings (IR) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above. Use duplicate pages for RO and SRO-only exams.

For Tier 3, select topics from Section 2 of the WA Catalog, and enter the WA numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to WAS that are linked to 10CFR55.43 VUREG-1021 1

Davis Besse 2008 NRC Written Exam Written Examination Outline Emergency and Abnormal Plant Evolutions -Tier 1 Group 1 Form ES-401-2 ES-401 X

AK3,02 Ability lo determine and interpret the following as they apply to the (Post-Trip Stabiiization) Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

Knowledge of the reasons for the foliowing responses as they apply to the Pressurizer Vapor Space Accident: Why PORV or code safety exit temperature is below RCS or PZR temperature Ability to determine or interpret the following as they apply to a small break LOCA: Reactor trio setooints 2.1.23 5 4.0

~

4.7

~

3.3 -

3.4 -

3.8 Conduct of Operations Ability to perform specific system and integrated plant procedures during ail modes of plant operation Ability to determine and interpret the following as they apply to the Pressurizer Vapor Space Accident Inadequate wre coo11no 007 I Reactor Trip I 1 008 I Pressurizer Vapor Space Acddent I 3 022 I Loss of Reactor Coolant Makeup I 2 2.4.50 Emergency Procedures I Pian Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.

Ability to detenine and interpret the following as they apply to the Loss of Residual Heat Removal System: Existence of proper RHR overpressure orotection 025 I Loss of Residual Heat Removal Svstem i 4 065 I Loss of Instrument Air / 8 Conduct of Operations: Ability lo explain and apply ail system IimitS and precautions.

I 1

2.1.32 E l 0 Post-Trio St2biliration 4.0

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3.6

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3.9

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3.3 008 / Pressurizer Vapor Space Accident / 3 X

X X

009 / Small Break LOCA 1 3 X

01 1 I Large Break LOCA 13 EA2.05 Ability to determine or interpret the following as they apply to a Large Break LOCA: Significance of charging pump operation 015 I 1 7 I Reactor Coolant Pump Malfunctions 14 Knowledge of the interrelations between the Reactor Cwlant Pump Malfunctions (Loss of RC Flow) and the following: RCP indicators and wntrols Conduct of Operations: Ability lo perform specific System and integrated plant procedures during all t

modes of olant ooeralion.

2.8

~

3.9

~

3.9

~

AU.10 2.1.23 022 I Loss of Reactor Coolant Makeup 12 025 I Loss of Residual Heat Removal System 14 AK1.O1 rtnoweage of the opera1 ona mp cal ons tl inr foi owing wncepts as tne) app, IL Loss 01 ResdLa meal Removal Sbslcm Lossof R I I R ~

I during all modes of operaion NUREG-1021 2

Form ES-401-2 ES-401 Davis Besse 2008 NRC Written Exam Written Examination Outline Emergency and Abnormal Plant Evolutions -Tier 1 Group 1 026 I Loss of Component Cwling Water I 8

~

~

027 I Pressurizer Pressure Control Sy5km Malfunction 13 029 /Anticipated Transient Without Scram (ATWS) I 1 I 038 I Steam Generator Tube Rupture / 3 I

054 I Lms of Main Feedwater 14 055 I Station Blackout1 6 056 I Loss of Ofkite Power 16 062 I Loss of Nuclear Service. Water I 4 X

065 I Loss of Instrument Air I8 E04 Inadequate Heat Transfer 14 E05 Excessive Heat Transfer 14 I

Ability to determine and interpret the following as they apply to the Lo55 of Component Cooling Water: The length of time after the loss of CCW flow to a onnponent before that uxnpnent may be damaged Conduct of Operations: Knowiedge of operator responsibilities during all modes of plant operation.

Knowledge of the interrelations between the ATWS and the following: Breakers, relays, and disconnects Knowledge of the operational implications of the fouowing cancepts as they apply to the SGTR. Use of steam tables Ability to operate and I or monitor the following as they apply to the Loss of Main Feedwater (MFW):

HPI, under total feedwater loss conditions Ability to operate and monitor me following as they apply to a Station Blackout: Reduction of loads on the battery Knowledge of the reasons for the following responses as they apply to the Loss of Offsite Power: Order and time to initiation of power for the load sequencer Emergency ProceduresPlan: Ability to recognize abnormal indications for system operating parameters which are entry level conditions for emergency and abnormal operating procedures.

Knowledge of the reasons for the following responses as they apply to the LOSS of Instrument Air: Cross-over to backup air supplies Knowledge of the operational implications of the following concepls as they apply to the (Inadequate Heat Transfer): Annunciators and conditir-I 4.4 49 3.5 I 50 4 I

NUREG-I 021 3

Form ES-401-2 ES-401 Davis Besse 2008 NRC Written Exam Written Examination Outline Emergency and Abnormal Plant Evolutions -Tier 1 Group 1 E10 Post-Trip Stabilization KIA Category Point Totals:

X 3/3 3

3 3

3 3/3 EK2 2 Grow Point Total:

I hnowleage 01 lne inlerre at ons beween lhe (Post-1~

p S t a ~ zauon) ana the follofflng Faulty's neat removal systems. tncluamg pnmary m l a n l emergency m a n !. !he aecay heal removal s)51ems. ana relations beween lhe proper operauon 01 Ihese systems 10 Ihe operauon 01 Ihe 13Clllry 3.5 56 1 8/6 NUREG-1021 4

ES-401 Davis Besse 2008 NRC Written Exam Form ES-401-2 Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 2 032 I Loss of Source Range Nuclear Instrumentatton 17 X

X m. 2 X

EA2.2 I

2.1.23 4.0 Conduct of Operations: Ability to perform specific system and integrated plant procedures during ail modes of plant operation Ability to determine and interpret the following as they apply to the (Shutdown Outside Control room)

Adherence to appropriate procedures and operation within the limitations in the fauliys license and amendments.

4.2 Ability to determine and interpret the following as they apply to the (Natural Csrculatkon Cooldown)

Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments 4.0 EO9 I Natural Circulation Operations 14 2.4.31 --L 3.4

~

4.5 Emergency Procedures i Plan Knowledge of annunciators alarms and indications, and use of the response instructions.

Ability to determine and interpret the following as they apply to the Continuous Rod Withdrawal:

Proper actions to be taken if automatic safety functions have not taken place Knowledge of the reasons for the following responses as they apply to the Inoperable I Stuck Control Rod: Tech-Sow limits for rod mismatch E13 I Steam Generator Overpressure I 4 001 I Continuous Rod Withdrawal I 1 I

AA2.03 005 I InoperablelStuck Control Rod I 1

3.6 AK3.03 AK2.02 EA1.09 Knowledge of the interrelations between the Accidental Gaseous Radwaste Release and the following: Auxiliary building ventilation system 2.1 060 I Accidental Gaseous RadWaste Release 19 074 I Inadequate Core Cooling I 4 Ability to operate and monitor Me following as they amlv to a lnadeauate Core Coolina: CVCS 3.7 AK1.2 Anow edge of h e operaiional mpllwtons 01 me Iohowing mcepls as lney apply to me (Plant R..nhachJ hormal aonormal ana emergency opera1 ng proceades associalea Yntn (Pant K..nhachl 3.5 A07 Plant Runback Knowledge of the reasons for the following responses as they apply to the (Turbine Trip)

Normal, abnormal and emergency operating orocedures asscciated with Turbine Triol.

3.4 62 A04 Turbine Trip AK3.2 1 068 Control Room Evacuation X

~

1 1

3 2

I (I K',>I OlPralions AD I l y 10 explain and appl)

.: q. w m,mts ana precautions 3.4 NUREG-I021 5

ES-401 Davis Besse 2008 NRC Written Exam Form ES-401-2 Written Examination Outline Emergency and Abnormal Plant Evolutions -Tier 1 Group 2 E14 Natural Circulation Cwldown X

EKl 3 E<2 2 NUREG-1021 Knowleage of tne operat onal implicat ons 01 Ihe followng mnCeptS as they apply to the (haldra C ICL alion Cwloom) Annmciatots and mno tions ndicaling scgnals. and remeaal acuons asSOC ateo wlh the (Fralural Cirwlatton Cootdown)

<noMedge of the interrelations bebeen the (EOP enclosures) and me followng Faulity's heal rernova systems. including pnmary Molani.

emergency coolant. me dewy heal removal 3 8 65 systems. and re ations between the proper opera1 on of Inese systems 10 !he operaLon of me lac ddy 3 5 64 G~oLP Point T ~ U I I 9/4 EOP enclosures KIA Category Point Total:

X 1/2 2

2 2

1

Form ES-401-2 ES-401 Davis Besse 2008 NRC Written Exam Written Examination Outline Plant Systems -Tier 2 Group 1 Conduct of Operations: Ability to recognize indications for system operating parameten which are entrylevel conditions for technical specifications.

4.0 004 Chemical and Volume Control 2.1.33 2.1.32

~

M.03 3.8 Conduct of Operations: Ability to explain and apply all system limits and precautions.

Ability to (a) predict the impacts of the following malfunctions or operations on the P S ; and (b) based on those predictions. use procedures to correct. control, or mitigate the consequences of those malfunctions or operations:

Overpressurization of the PZR Ability to (a) predict the impacts of the following malfunctions or operations on the RPS: and (b) based on those predictions, use procedures to correct, control. or mitigate the consequences of those malfunctions or operations: Failure of RPS siqnal to trip the reactor 3.9 4.7 005 Residual Heat Removal 007 Pressurizer RelieflQuench Tank X

X X

A2.06 012 Reactor Protection 2.1.12 -

A4.08 Conduct of Operations: Ability to apply technical specifications for a system Ability to manually operate andlor monitor in the control r w m : RCP uxling water supplies Knowledge of the effect of a loss of malfunction on the following CVCS mponents:

Demineralizers and im exchangers Knowledge of CVCS design feature@) andlor interlock(s) which provide for the following:

Temperaturelpressure wntrol in letdown line:

prevent boiling, lifting reliefs, hydraulic shock, piping damage, and burst Knowledge of the effect of a loss or malfunction on the following will have on the RHRS: RHR heat exchanaer 4,0 3.2 2.5 3.1 2.5 026 Containment Spray 003 Reactor Coolant Pump 004 Chemical and Volume Control K6.02 -

K4.11 004 Chemical and Volume Control 005 Residual Heat Removal K6.03 Ab ly 10 pred CI and or monitor manges In aarxnelefs (to prevent exceeang des.gn Itmts) x 5 o c ated rntn operating the ECCS conlro s.

ncldo "9 Boron wnCentra1,on m accumdlaIo(.

w o n storage tan*s 6nonteoge of tne pnys cai wnnectons and or CdLsetflecI relationships Detween me PRTS and IIW lo own9 s>stems RCS A1.02

~

K1.03

~

006 Emergency Core Cwling 007 Pressurizer RelieflQuench Tank 7

NUREG-I 021

ES-401 Davis Besse 2008 NRC Written Exam Form ES-401-2 Written Examination Outline Plant Systems - Tier 2 Group 1 3.0 Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS, and (b) based on those predictions. use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Highllow CCW temperature Emergency Procedures I Plan Knowledge of annunciators alarms and indications, and use of the respmse instructions.

Knowledge of bus power supplies to the following:

Controller for PZR spray valve 008 Component Cwling Water A2.03 2.4.31 008 Component Cooling Water 3.3 -

2.5 010 Pressurizer Pressure Control K2.02 010 Pressurizer Pressure Control A4.02 Ability to manually operate andlor monitor in the control room: PZR heaters 3.6 012 Reactor Protection 3.3 3.6 -

Knowledge of bus power supplies to the following:

RPS channels, components. and interconnections Ability to monitor automatic operation of the RPS.

including: Bistables K2.01 A3.02 012 Reactor Protection 013 Engineered Safety Features Actuation A3.01 Ability to monitor automatic operation of the ESFAS including: Input channels and logic 3.7 Aoility to (a) preo CI !lie mpacts of !tie fodomng ma funct ons or operalions on me CCS. and (0) based on those pred.clions &e procedures 10 correct control or m ligate me consequences 01 those malfunctlons oi operatons Ma.or ledk in ccs Knowleoge of DUS power SJPP es to Ihe lo ow ng Containment spray p m p s 022 Containment Cooling A2.05 3.1 -

3.4 K2.01 026 Containment Spray 039 Main and Reheat Steam 3.6 Knowledge of the operational implications of the following concepts s the apply to the MRSS: Effect of steam removal on reactivity Knowledge of MFW design feature(s) andlor inter(ock(s) which provide for the following:

Feedwater regulatoly valve operation (on basis of steam flow, feed flow mismatch)

K5.08 K4.08 A1.O1 2.5 059 Main Feedwater 3.9 061 AuxillarylEmergency Feedwater parameten (to prevent exceeding design limits) asscciated with operating the AFW controls induding: SIG level NUREG-I021

ES-401 Davis Besse 2008 NRC Written Exam Form ES-401-2 Written Examination Outline Plant Systems - Tier 2 Group 1 Knowledge of the physical connections andlor cause-effect relationships between the ac distribution system and the following systems: DC distribution Knowledge of dc electrical system design followino: Manuallautomatic transfers of control 3.5 K1'03 X

K4.01 feature@) andlor interlwk(s) which provide for the 2.7 I I x 062 AC Electrical Distnbution 19 20 063 DC Electrical Distribution I

I X

X X

3 3

064 Emergency Diesel Generator I X I 2.4'6 l

l 064 Emergency Diesel Generator 3.1 21 Emergency Procedures I Plan Knowledge symptom based EOP mitigation strategies.

073 Process Radiation Monitoring 076 Service Water X

076 Service Water Knowledge of the effect that a loss or malfunction EDlG (manual loads)

Knowledge of the operational implications as they apply to concepts as they apply to the PRM system: Radiation intensity changes with source distance K3.03 of the EDIG system will have on the following:

3.6 22 2.5 23 K5.02 I

I x A3'02 X

K1.16 1 x 1 078 Instrument fur Ability to monitor automatic operation of the SWS, 3.7 24 including: Emergency heat loads Knowledge of the physical connections andlor cause-effect relationships between the SWS and 3.6 25 078 Instrument Air K3.01 I

I 3.4 26 Conduct of Operations: Ability to explain and apply all system limits and precautions.

Knowledge of the effect that a 105s or malfunction of the IAS will have on the following: Containment 3.1 27 103 Containment 3

2 2

I

/

Knowledge of the effect that a loss or malfunction of the containment system will have on the following: Loss of containment integrity under shutdown conditions 3.3 28 K3.01 2

2

/

2 3

2 Group Point Total:

2815 IUA Cateowv Point Totals:

1 313 I 3 I

I I

I I

I I

I the followina svstems: ESF I

I I

I I

I I

I I

I I air svstem I

I I

NUREG-1021 9

ES-401 Davis Besse 2008 NRC Written Exam Form ES-401-2 Written Examination Outline Plant Systems -Tier 2 Group 2 2.8 2.7 3.1 2.5 3.5 2.7 2.6 2.5 016 Non-nuclear Instrumentation X

30 31 32 33 34 35 36 37 I x I I 033 Spent Fuel P w l Cooling 001 Control Rod Dave 015 Nuclear Insbumenlation 01 6 Non-nuclear Instrumentation 01 1 Pressurizer Level Control System 034 Fuel Handling Equipment 035 Steam Generator 041 Steam Dumprrurbine Bypass Control 056 Condensate System 075 Circulating Water X

X X

2.4.31 -

2.4.30 A2.03 K3.02 K5.10 A4.02 K2.02 -

A3.01 A1.02 -

K6.03 K1.03 K4.01 Emergency Procedures I Plan Knowledge of annmciatoffi alarms and indications. and use of

~~

~-

the re4ponse instructions.

Emergency Procedures I Plan Knowledge of which events related to system operationslslatus should be reported to outside agencies.

Ability to (a) predict the impacts of Me following mal-functions or operations on the GS: and (b) based on those predictions, use procedures to wrrect. Wntrol, or mitigate the wnsequences of those malfunctions or operations: Pressurellevel transmitter failure Knowledge of the effect that a loss or malfunction of the CRDS will have on the following: RCS Knowledge of the operational implications of the following concepts as they apply to the NIS:

Exwre Detector operation Ability to manually operate andlor monitor in the control room: Recorders Knowledqe of bus Dower supplies to the folloMng:

Pressurizer heaters

~~~~

~~

~

~

Ability to monitor automatic operation of the Fuel Handling System. including: Travel limits

~

~~~

Ability to predict andlor monitor changes in parameterr (to prevent exceeding design limits) associated with operating the SlGS controls including: S/G pressure Knowledge of the effect of a loss or malfunction on the following will have on the SDS: Controller and positioners. including ICs, SIG, CROS Knowledge of the physical connections andlor cause-effect relationships between the Condensate System and the following systems:

LI3\\1, 1.11 I..

Knowledge of circulating water system design feature(s) and interlock(s) which provide for the following: Heat sink NUREG-1021 10

ES-401 Davis Besse 2008 NRC Written Exam Form ES-401-2 Written Examination Outline Plant Systems -Tier 2 Group 2 079 Station Air NUREG-I021 11

Facility:

Davis Besse 2008 NRC Exam I Date of Exam:

Category

1.

Conduct of Operations 2.

Iquipment Control

3.

iadiation Control I.

mergency

'rocedures / Plan

'ier 3 Point Total WA #

Topic 2.1.20 2.1.33 Ability to execute procedure steps.

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

Ability to use plant computer to obtain and evaluate parametric information on system or component STATUS 2.1.19 Subtotal 2.2.33 2.2.7 2.2.27 Knowledge of control rod programming.

Knowledge of the process for conducting tests or experiments not described in the safety analysis report.

Knowledge of the refueling process Ability to perform pre-startup procedures for the facility, including operating those controls associated with plant equipment that could affect reactivity.

2.2.1 Subtotal A

I Knowledge of 10 CFR: 20 and related facilitv personnel exposure.

2.3.1 1 Ability to control radiation releases.

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

Knowledge of the process for performing a 2.3.9 containment purge.

Subtotal 2'4'36 2'4'28 2.4.5 Knowledge of chemistry / health physics tasks during emergency operations.

Knowledge of procedures relating to emergency response to sabotage.

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

Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emeraencv and abnormal 2.4.4 I

I operating proceddres I Aoilily to dlagnose and recognize trends in an 2.4.47 accurate and timely manner utilizing the appropriate control room reference material.

I I

Subtotal 2/18/2008 I

IR -

3.4 -

3.0

~

~

2.6 -

3.7 2.9 -

4.0 3.4 NUREG-I021 12

roup 211 rier I Randomly Reason for Rejection Selected WA 003 A4.07 I Q # I Reselected 003 A4.08 as RCPs do not have a seal bypass at this facility 21 1 211 212 2 1 2 1 I 1 1 11 112 211 211 212 008 G2.4.30 I Q #8 Reselected WA to 2.4.31 due to WA importance rating of <2.5 for RO Q #25 Reselected K1.16 because system relationship does not exist for selected topic Q #30 Reselected 015 WA 5.10 used since WA 5.01 is deleted from KA 076 K1 '05 n, E vc n, catalogue Q #32 Reselected 01 1 K2.02 because equipment is not used at facility Q #52 TS Basis for APE not applicable for RO knowledge. Reselected G2.4.4 ul.Jr\\.J.ul 028 K2.01 062 G2.2.25 I

Q #77 Difficulty in developing an SRO test item, and excessive overlap with Oo8 AA2'20 I other topics on exam. Reselected 008 AA2.30 032 G2.1.27 I Q #82 Difficulty in developing SRO topic to match KA. Reselected G2.1.23 005 G2.1.27 026 G2'1'14 017 G2.4.31 Q #87 Difficulty in developing SRO topic to match KA. Reselected G2.1.32 Q #90 No condition specific to system that could yield an SRO level test item.

Reselected G2.1.12 Q #91 No action required for condition related to selected system. Reselected Stem 016 NUREG-I021 13

/ F F e i b i x D Scenario Outline Form ES-D-1 I Malf.

Event No.

Type*

N - BOP HU21A C - RO TS - SRO R - R O R - SRO Facility:

DAVIS-BESSE Scenario No.:

1 Op Test No.:

NRC 2008 Examiners:

Operators:

Event Description Swap Turbine Plant Cooling Water Pumps RCP 1 Loss of oil to upper bearing Power reduction prior to stopping RCP 1-1 Initial Conditions:

100% power, MOL High Pressure Injection 1 pump is out of service HH43 Turnover:

Event M - All I Reactor Coolant System leak - 2500 gpm 7

LGLE C - R O I SFAS Modules L231 fail to trip L1621 Stuck Control Rod at -75% powei IksR:Ro I

LlTL28 I-BOP SFFD I C - BOP I MFPT 2 Vibration II SG 1 level transmitter fails mid-scale 8

9 (N)ormal, (R)eactivity, (1)nstrument.

(C)omponent, (M)ajor Appendix D NUREG 1021 Revision 9

1 Appendix D Scenario Outline Form ES-D-I 1 DAVIS-BESSE 2008 NRC EXAM SIMULATOR SCENARIO 1 GENERAL DESCRIPTION The crew will assume control with power holding at 100% power.

The Lead Evaluator will cue the swapping of the Turbine Plant Cooling Water Pumps in accordance with DB-OP-06263, TURBINE PLANT COOLING WATER.

The Lead Evaluator will cue the leak for RCP 1-1 upper bearing. The oil leak will cause high bearing temperature. The crew should respond to alarm 6-1-A, in accordance with DB-OP-02006, REACTOR COOLANT PUMP ALARM PANEL 6 ANNUNCIATORS, and then enter DB-OP-02515 will require the crew to reduce power to 572% in accordance with DB-OP-02504.

RAPID SHUTDOWN, and stop the affected RCP. The SRO should enter the proper TS after the RCP is stopped.

The Lead Evaluator will cue the stuck rod. The crew should respond to alarm 5-2-E, CRD ASYM-METRIC ROD, in accordance with DB-OP-02005, PRIMARY INSTRUMENTATION ALARM PANEL 5 ANNUNCIATORS and then enter DB-OP-02516, CRD MALFUNCTIONS.

The SRO should enter TS 3.1.3.1. TS Limit with a stuck rod and 3 RCPs is 320 MWE (45%

DB-OP-02515, REACTOR COOLANT PUMP AND MOTOR ABNORMAL OPERATION.

power). The crew should perform reduce power in accordance with DB-OP-02504, RAPID SHUTDOWN.

The Lead Evaluator will cue the MFPT 2 vibration failure during the power reduction. Crew should recognize problem with MFPT 2 and investigate. Power level should be low enough to trip MFPT 2.

The Lead Evaluator will cue the RCS leak. This will lower Pressurizer level and the crew should trip the reactor when 1 0 0 is reached in accordance with DB-OP-02522, SMALL RCS LEAKS.

The crew will transition to the Emergency Operating Procedure, DB-OP-02000, when the reactor trips.

RCS pressure will lower and a LOSS OF SUBCOOL MARGIN will occur. The crew will trip all running RCPs and route to DB-OP-02000, SECTION 5.0, LACK OF ADEQUATE SUBCOOLING MARGIN. The RCS is large enough to cause an SFAS actuation.

Failure of SFAS L231 will be put in at the start of the scenario. The failure of the SFAS module will prevent Component Cooling Water Pump from operating and SG level will not control at the higher SG level (124). The crew should start the CCW Pump and manually control level at the higher level.

The failure of the SG 1 level transmitter to mid-scale will cause level to rise in the SG. The crew should take manual control of the SG level control valve to maintain proper level in the SG.

The Lead Evaluator can terminate the scenario when all high level activities have been completed and the evaluators agree the crew can be properly evaluated.

Appendix D NUREG 1021 Revision 9

[Appendix D Scenario Outline Form ES-D-1 I Initial Conditions:

50 - 60% power, 3 RCP. MOL No equipment out of service

=acility:

DAVIS-ESSE Scenario No.:

Malf.

No.

HI70 L153B Operators:

rxaminers:

Event Event Type' Description N - All TS - SRO C-RO R-RO Control Rod drop Transfer Gland Steam from Main Steam to Auxiliary Steam The crew will be notified that Auxiliary Feedwater Pump 1 has no Governor oil Make-up valve (MU 32) fails to operate in auto NP03 NP05 PLZZ G529B SFERE W

l Turnover:

TS - SRO I - BOP Loss of NNI X DC N - SRO M -All C-RO C - BOP Loss of Offsite AC Emergency Diesel Generator 1 fails to auto start AFW Pump 2 governor valve closes P

Event No.

1 2

3 4

5 6

7 8

(N)ormal.

(R)eactivity.

(1)nstrument.

(C)ornponent.

(M)ajor Appendix D NUREG 1021 Revision 9

1 Appendix D Scenario Outline FormES-D-1 I DAVIS-BESSE 2008 NRC EXAM SIMULATOR SCENARIO 2 GENERAL DESCRIPTION The crew will assume control with power at 50 - 60% power and 3 Reactor Coolant Pumps in operations.

The Lead Evaluator will cue the transfer of Gland Steam from Main Steam to Auxiliary Steam The Lead Evaluator will cue the Auxiliary Feedwater Pump 1 oil problem. The crew should review T.S. 3.7.1.2 and declare AFW 1 inoperable. This will put them in a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Action Statement.

The MU32 failure to operate in Auto will be inserted during the setup of the scenario. The failure will not be detected until the control rod drops. The RCS temperature will lower and the Pressurizer level will drop due to the dropped control rod. If Pressurizer level drops to 200 inches alarm 4-2-E. PZR LVL LO, on REACTOR COOLANT ALARM PANEL 4 ANNUNCIATORS will alarm. The crew should identify that MU32 is not responding. The crew should refer to DB-OP-02512, LOSS OF RCS MAKEUP and take the MU32 Station to hand and control Pressurizer level by adjusting MU32 position manually.

The Lead Evaluator will cue the dropped control rod. The crew should respond to alarm 5-1-E, CRD LCO. and.5-2-E, CRD ASYM-METRIC ROD, in accordance with DB-OP-02005, PRIMARY INSTRUMENTATION ALARM PANAL 5 ANNUNCIATORS and then enter DB-OP-02516, CRD MALFUNCTIONS. The SRO should enter TS 3.1.3.1. TS Limit with a dropped rod and 3 RCPs is 320 MWE (45% power). The crew should reduce power in accordance with DB-OP-02504, RAPID SHUTDOWN.

The Lead Evaluator will cue the LOSS OF NNI X DC during control rod recovery. Annunciator 14-1-D, NNI-X 24 VDC BUS TRIP, on MSR/ICS ALARM PANEL 14 ANNUNCIATORS will alarm. The crew should enter DB-OP-2532. LOSS OF NNVICS POWER. The crew should recognize a minor transient is in progress due to the midscale failure of Turbine Throttle Pressure by transferring the Turbine to MANUAL and transferring the SG/Rx Demand Station to HAND, and lowering the Turbine load.

The Lead Evaluator will cue the Loss of Offsite AC power. The crew will enter DB-OP-02000, RPS, SFAS, SFRCS Trip, or SG Tube Rupture, when the reactor trips. Emergency Diesel Generator (EDG) 1 will fail to auto start. Essential electrical bus D1 will be powered from their respective Emergency Diesel Generators. EDG 1 should be started manually.

The Lead Evaluator will cue the AFW Pump 2 governor valve closes. The crew should respond by entering DB-OP-02000, SECTION 6, LACK OF HEAT TRANSFER. The crew should energize 132. non-essential electrical bus, from the Station Blackout Diesel Generator and align the Motor Driven Feedwater Pump to supply the Steam Generator.

The Lead Evaluator can terminate the scenario when all high level activities have been completed and the evaluators agree the crew can be properly evaluated.

Appendix D NUREG 1021 Revision 9

TA. en.a~x-D PP Scenario Outline Form ES-D-I

-acility:

DAVIS-BESSE Scenario No.:

3 Op Test No.:

NRC 2008 Lxaminers:

Operators:

nitial Conditions:

100% power, MOL No equipment out of sewice rurnover:

Event No.

1 2

3 4

5 6

7 9

Malf.

No.

B2MlN BV24B HH50 UFO9A UFOBA UFllA UFO6A SFDPC F30AB F30AC Event Type*

N - BOP C-RO TS - SRO C-RO R-RO TS - SRO C-RO M -All M -All C -BOP Event Description Perform TG OversDeed Test. DB-SS-04154 Make-up Pump 1 trips MU 1903 fails closed SG 1 tube leak and plant shutdown Turbine Vibration MFPT 2 weed lowers and plant triD SG Tube Rupture SFRCS Ch 1 fails to actuate (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Appendix D NUREG 1021 Revision 9

1 Appendix D Scenario Outline Form ES-D-I 1 DAVIS-BESSE 2008 NRC EXAM SIMULATOR SCENARIO 3 GENERAL DESCRIPTION The crew will assume control with power holding at 100% power The Lead Evaluator will cue the TG Overspeed Test, DB-SS 04154 The Lead Evaluator will cue the loss of the running Makeup Pump. The crew should respond to annunciator 6-6-C, SEAL INJ TOTAL FLOW, in accordance with DB-OP-02006, REACTOR COOLANT PUMP ALARM PANEL 6 ANNUNCIATORS, and then enter DB-OP-02515, REACTOR COOLANT PUMP AND MOROR ABNORMAL OPERTIONS, and DB-OP-02512, LOSS OF RCS MAKEUP. The crew should close MU19, Seal Injection Flow Control Valve, and start the standby Makeup Pump. The crew should restore Pressurizer level and RCP Seal Injection flow.

The Lead Evaluator will cue the MU1903, Letdown Dernin Inlet. failure. The crew will respond to annunciator 2-2-A, LETDOWN PRESS HI, in accordance with DB-OP-02002, LETDOWN/MAKEUP ALARM PANEL 2 ANNUNCIATORS. The crew will isolate Letdown. An Equipment Operator will call up and identify that MU1903 had been inadvertently closed.

Letdown will be re-established in accordance with DB-OP-06006. MAKEUP AND PURIFICATION SYSTEM The Lead Evaluator will cue the Steam Generator 1 Tube Leak. The crew should respond to annunciator 12-1-8, MN STM LINE 2 RAD HI, in accordance with DB-OP-06012. STM GEN/SFRCS ALARM PANEL 12 ANNUNCIATOR and then enter DB-OP-02531, STEAM GENERATOR TUBE LEAK. The crew should start a rapid shutdown in accordance with DB-OP-02504, RAPID SHUTDOWN. The crew will evaluate the SG leakage and determine is in excess of T.S. 3.4.6.2.

The Lead Evaluator will cue the rise in SG tube leakage. This rise will be larger than Makeup capacity and after the reactor trip the crew will enter DB-OP-02000, Section 8, SG TUBE RUPTURE.

After the reactor trip, the MFPT 2 speed will lower. The crew should respond to the lowering MFPT meed indication and initiate SFRCS. SFRCS Ch. 1 will not work and the crew will reposition the SFRCS valves in accordance with the DB-OP-02000 Table 1, SFRCS ACTUATED EQUIPMENT.

The Lead Evaluator can terminate the scenario when all high level activities have been completed and the evaluators agree the crew can be properly evaluated.

Appendix D NUREG 1021 Revision 9

Appendix D..

- scenari6-6uiTine

-... Form'ES-D-l].

Event Event Type*

Description

~~

N - BOP I-RO TS-SRO Manual Voltage Regulator operations per DB-OP-06301, step 3.4 SFAS Containment Pressure transmitter fails low Facility:

DAVIS-BESSE Scenario No.:4 Op Test No.:

NRC 2008 Examiners:

Operators:

~~

Initial Conditions:

100% power, MOL No equipment out of service Turnover:

1 2

3 4

5 6

7 L6P1 D H l C l C HDP309 HN29A SFEF FKMID I -RO I Pressurizer Temperature fails mid-scale II c - BOP R -RO TS - SRO M -All C - BOP Turbine Plant Cooling Water Pump 3 trips RCP 2-1 Seal Cooler leak Steam Leak in Fan Allev AF 3872. AFW level control valve. fails oDen (N)ormal, (R)eactivity, (1)nstrument.

(C)omPonent, (M)ajor Appendix D NUREG 1021 Revision 9

I Appendix D Scenario Outline Form ES-D-1 1 DAVIS-BESSE 2008 NRC EXAM SIMULATOR SCENARIO 4 GENERAL DESCRIPTION The crew will take the watch with power holding at 100% power.

The Lead Evaluator will cue the operation of the Voltage Regulator.

The Lead Evaluator will cue the Containment Pressure transmitter failure. The crew should respond to alarm 5-4-8, SFAS CTMT PRESS LO CH TRIP, in accordance with DB-OP-02005, PRIMARY INSTRUMENTATION ALARM PANEL 5 ANNUNCIATORS. The SRO should enter the correct TS and direct the RO to trip the SFAS channel.

The Lead Evaluator will cue the Pressurizer Temperature mid-scale failure The crew should respond to annunciator 4-2-E, PZR LVL LO, in accordance with DB-OP-02004, REACTOR COOLANT ALARM PANEL 4 ANNUNCIATORS, and then enter DB-OP-02513, PRESSURIZER SYSTEM ABNORMAL OPERATION. The crew should select the alternate temperature instrument and return to normal operations.

The Lead Evaluator will cue tripping of TPCW Pump 3. The crew should respond to annunciator 11-I-F, TPCW HI LVL TK LVL, in accordance with DB-OP-02011. HEAT SINK ALARM PANEL 11 ANNUNCIATORS, and then enter DB-OP-02514, LOSS OF TURBINE PLANT COOLING WATER PUMP. The crew should start TPCW Pump 2.

The Lead Evaluator will cue RCP 2-1 seal cooler failure. The crew should respond to annunciator 11-4-A, CCW SURGE TK LVL HI, in accordance with DB-OP-02011, HEAT SINK ALARM PANEL 11 ANNUNCIATORS, and then enter DB-OP-DB-OP-02523. COMPONENT COOLING WATER SYSTEM MALFUNCTIONS. Seal Return temperature will rise. The crew should enter DB-OP-02515, REACTOR COOLANT PUMP AND MOTOR ABNORMAL OPERATION, and lower power to-72% and turn RCP 2-1 off.

The Leak Evaluator will cue the steam leak. The crew should respond to annunciator 12-2-8.

SG 1 TO AFPT 2 MN STM PRESS LO, in accordance with DB-OP-02012, STM GENBFRCS ALARM PANEL 12 ANNUNCIATORS, and then enter DB-OP-02525, STEAM LEAKS. The crew will attempt to isolate the steam leak. The crew should trip the reactor and enter DB-OP-02000. RPS, SFAS, SFRCS, or SG Tube Rupture and blowdown the Steam Generator.

AF 3872, AFW TO SG 2, will fail open. This will feed SG 2, the SG with the steam leak. The crew should respond by isolating AFW to SG 2 in accordance with DB-OP-02000.

The Lead Evaluator can terminate the scenario when all high level activities have been completed and the evaluators agree the crew can be properly evaluated.

Appendix D NUREG 1021 Revision 9

DAVIS-BESSE 2008 INITIAL LICENSE EXAM OPERATING OUTLINE COMMENTS Source

1.

Admin JPM A (RO and SRO) - JPM 227 -

Calculate RCS with F755 Unavailable

2.
3.

Admin JPM B (SRO) -

AN2 - Review an Auxiliary Feedwater Surveillance Test and Determine Operability Admin JPM D (RO) -

AN1 -Calculate Radiation Release Using SGTL Abnormal Procedure, DE-OP-02531, Control Room JPM G

- JPM 215 - Respond to a High Station Vent Radiation Alarm

4.

Comment Control Room JPM A Service Water Loop 1 to Primary Loads

- JPM 14 -LOSS Of

1. The WA for this JPM is listed as 2.2.18.

However, the WA for a Conduct of Operations JPM needs to come from Section 2.1 instead of 2.2.

2. Expand the title of the JPM to "Calculate RCS Flow with F755 Unavailable."

Need WA listed for this JPM from Section 2.1.

5.
1. The WA for this JPM is listed as 2.3.8.

However, this WA is not appropriate, since the WA is for a planned gaseous release, whereas the JPM is associated with a release due to a SGTL. Suggest possible use of WA 2.3.10 (2.9/3.3) for this JPM.

2. For an RO, the WA value for 2.3.8 is only 2.3 (i.e., < 2.5). Need to provide some justification for using this JPM with a WA < 2.5.

Control Room JPM C

- JPM 48 -Exchange RCS Flow Amphynol The WA for this JPM is listed as 055 A4.01.

However, this WA is for the Condenser Air Removal System (with WA values of 1.E, 1.9) that have nothing to do with the JPM.

The WA for this JPM is listed as 039 A3.02.

However, this WA is for the Main and Reheat Steam System that has nothing to do with the JPM. The WA is associated with Safety Function 4 (heat removal), whereas the JPM is associated with Safety Function 7 (instrumentation).

The WA for this JPM is listed as 2.3.1 1.2, which is not a valid WA.

Resolution

1. Comment incorporated.

The WA for the JPM was changed to 2.1 25.

2. Comment incorporated.

Comment incorporated.

The WA for the JPM was changed to 2.1.33.

1. Comment incorporated.

The WA for the JPM was changed to 2.3.10.

2. Comment incorporated.

The WA for the JPM was changed to 2.3.10, with a values of 2.713.3.

Comment incorporated.

The WA for the JPM was changed to 062 AA1.02.

The JPM was replaced per facility discretion. The new JPM has a WA of 012 A4.03, which is associated with Safety Function 7.

Comment incorporated.

The WA for the JPM was changed to 071 A3.03.

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7.
0.
9.
10. -
11.
12.
13.

DAVIS-BESSE 2008 INITIAL LICENSE EXAM OPERATING OUTLINE COMMENTS Control Room JPM H

-Sync the Main Generator to the Grid (RO only)

In-Plant JPM K -

Primary Side RO CTRM Evacuation Scenario 1 Outline Scenario 2 Outline Scenario 3 Outline General Comment 1 for all Scenarios General Comment 2 for all Scenarios The WA for this JPM is listed as 045 A4.11 associated with Safety Function 4 (Heat Removal) (2.4, 2.3). For this JPM to be associated with Safety Function 6 (Electrical),

the WA should be changed to 062 K4.05 or some other WA associated with Safety Function 6.

The WA for this JPM is listed as 068 Al.1, which is associated with Safety Function 9 (Radioactivity Release), which is not the correct WA. A WA associated Safety Function 2 (Inventory Control) should be listed.

Editorial: Delete Event No. 9 from sheet.

1. For Event 1, it appears that only the BOP should be given credit for this Normal evolution.
2. For Event 5, delete "N-SRO" under Event TvDe.
1. For Event 5, Turbine Vibration, there is no mention of this event in the Scenario 3 General Description.
2. For Event 5, change the Event Type such that the BOP gets credit for the Event, instead of the RO.

Only Scenario 1 has any major equipment initially 00s. The other scenarios should have some equipment initially 00s.

None of the scenarios have a Turnover shown on the outline sheets. The crews should have a Turnover for each scenario.

Comment incorporated.

The WA for the JPM was changed to 062 A4.07.

Comment incorporated.

The WA for the JPM was changedto 068AA1.13, which is associated with Safety Function 2 (Inventory Control) (i.e., Makeup Pump control during a Control Room Evacuation).

Comment incorporated.

1. Comment incorporated.
2. Comment incorporated
1. Comment incorporated
2. Comment incorporated Comment incorporated.

Scenarioschanged such that the Initial Conditions specity equipment initially 00s.

Comment incorporated.

Scenarioschangedsuch that a Turnover is shown on the outline sheet.

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