ML081220228
| ML081220228 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| 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) | |
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
- WHEN SEPARATED FROM ENCLOSURE 1, HANDLE THIS 419321 7676 Mark B Benlla F8x 410 327 i 5 R 2 V m Presrdenf N ~ k d i DOCUMENT AS UNRESTRICTED -
Docket Number 50-346 I O CFR 55 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
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 111w/o Chief - Operations Branch, NRC Region 111w/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)
ES-201, Rev. 9 Examination Outline Quality Checklist Form ES-201-2 1 i a Verity that the outlme(s) fit(s) the appropriate model, In accordance mth ES-401 I,Y n,
b Assess rrnetner Ine OLIne 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 7'
T T
- c. Assess whether the outline over-emphasizes any systems, evoiutions. or generic topics. m E d. Assess whether the justifications for deselected or rejected WA statements are appropriate.
N ,;fl
- 2. 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.
S and major transients.
i b. Assess whether there are enough scenario sets (and spares) to test the projected number M and mix of applicants in accordance with the expected crew composition and rotation schedule U without compromising exam integrity, and ensure that each applicant can be tested using A
L 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. , 2.b T
- c. To the extent possible, assess whether the outline(s) conform(s) with the qualitative R 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 distributed among the safety functions as specified on the form (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@)
- 3. (4) the number of new or modified tasks meets or exceeds the minimums specified on the form W
(5) the number of alternate path, low-power, emergency, and RCA tasks meet the criteria on the form. &;
I
- 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
~~~~~ ~~ ~ ,
(2) at least one task is new or s i g r k n t l y modified on the form (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. WJ
- 4. in the appropriate exam sections. J1 E
- b. Assess whether the 10 CFR 55.41143 and 55.45 sampling is appropriate.
kw N c. Ensure that WA importance rah'ngs (except for plant-specific priorities) are at least 2.5. d2:
E I d. Check for duplication and overlap amonq exam sections. IC0 A
L
- e. Check the entire exam for balance of coverage. pfi
__ f. Assess whether the exam fits the appropriate job level (RO or SRO).
py:
Printed NamelSignature
- a. Author
- b. Facility Reviewer (*)
- c. NRC Chief Examiner I#)
- d. NRC Supervisor hole u ndependenl hRC rev ewer n,La lems n C o . n n c' Cn et exam ner concdrrence reqt. reo
ES-201, Rev. 9 Examination Outline Quality Checklist Form ES-201-2
- acility: Davis-Besse Date of Examination: 211812008 L
Item
- Task Description a W
- 1. 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 R
I Section D . l of ES-401 and whether all KIA categorles are appropriately sampled.
T T
c Assess whether the Outline over-emphasizes any systems, evolutions. or generlc topics
].E E d. Assess whether the justifications for deselected or rejected KIA Statements are appropriate.
- N 2.
S I
M U
L A
T 0 e out ine s con orm(s) with the qualitative
- R 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) 3.
(4)the number of new or modified tasks meets or exceeds the minimums specified on [he form W
I T b. Verify that the administrative outline meets the criteria specified on Forni ES-301-1.
4.
G E
N E
R m sec ions.
A L
- 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: 211812008 Examination Level (circle one): RO Operating.Test Number: NRC Administrative Topic Type Describe activity to be performed (see Note) Code' JPM 227 - Calculate RCS with F755 unavailable Conduct of Operations D 2.2.18 Knowledge of the process for managing maintenance activities NOT SELECTED Conduct of Operations NIA JPM 148 - Review a tagout and determine it is Equipment Control M incorrect.
2.2.13 Knowledge of tagging and clearance procedure ANI - Calculate Radiation Release using SGTL Radiation Control I N Abnormal Procedure, DB-OP-02531, Attachment 1
I-2.3.8 Knowledge of the process for performing a planned gaseous radioactive release JPM 241 - Make EP Offsite Notification Emergency Plan 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, Attachment 7, 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, Attachment 1, 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 Date of Examination: 211812008 Examination Level (circle one): SRO Operating Test Number: NRC Administrative Topic Type Describe activity to be performed (see Note) Code*
JPM 227 - Calculate RCS with F755 unavailable Conduct of Operations D 2.2.18 Knowledge of the process for managing maintenance activities AN2 - Review an Auxiliary Feedwater Surveillance Conduct of Operations N Test and determine Operability JPM 148 - Review a tagout and determine it is Equipment Control M incorrect.
2.2.13 Knowledge of tagging and clearance procedure ANI - Calculate Radiation Release using SGTL Radiation Control N 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 Emergency Plan D 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, Attachment 7, 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 Attachment 1. 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 Facility: DAVIS - BESSE Date of Examination: 211 8/08 Exam Level (circle one) RO I SRO(I) Operating Test No.: NRC 1 Control Room Systems@(8 for RO; 7 for SRO-I; 2 or 3 for SRO-U) /I System I JPM Title 1 Type Code' I Safety Function I 1a. JPM 14 - Loss of Service Water Loop 1 to Primary loads 1 A, D, M 1 b. 1 JPM 33 -Transfer LPI Suction to the CTMT Emergency Sump I A, D I 3 n I/ c. 1 JPM 48 Exchange RCS Flow amphynol I D I 7 I1 JPM 97 -manually Actuate SFAS after some components blocked I A.D I
- e. JPM 85 - Purge Containment D, L 8
- g. JPM 215 - Respond to a high Station Vent radiation alarm A, D, C 9 1 Ih. NEW - Synch the Main Generator to the Grid (RO ONLY) b 1 6 U 1 Implant Systems" (3for RO; 3 for SRO-I; 3 or 2 for SRO-U) Y 1 i. JPM 127 -Actions for steam binding of the Motor Driven D 4s Feedpump
- j. JPM 115 - Emergency Shutdown of EDG A, D. E, M 6 1 Ik. New - Primary Side RO CTRM Evacuation I E , N , R I z 1 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.
(A)lternate path Type Codes i Criteria for RO I SRO-I / SRO-U 4-614-6/ 2-3 (C)ontrol room (D)irect from bank 9 1 81 4 (E)mergency or abnormal in-plant 1/ I/ 1 (L)ow-Power 1 / 1/ 1 (N)ew or (M)odified from bank including l ( A ) 2/ 2/ 1 (P)revious 2 exams 3 I 3 I 2 (randomly selected)
(RKA 1/ 1 1 1 NUREG-1021, Revision 9
- ES-301
.. .Control
. Roomlln-Plant
. .. . Systems Outline Form ES-30?-'2::
- a. 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
- b. 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 C. From Mode 3, 2. or 1. the candidate will be directed to RCS flow instruments Bank JPM 98 WA: 039A3 .02
- d. 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
- e. From a Mode 1 condition, the candidate will be direct to start the a purge on Containment.
Bank JPM 85 KA: 029A2.03
- f. 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
- 9. 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
- h. From a Mode 1 condition, the candidate will be directed to synchronize the Main Generator to the Grid.
NEW JPM WA: 045A4.11 I. 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
- j. From a shutdown condition, the candidate will be direct to emergency shutdown the Emergency Diesel Generator.
Bank JPM 1 15 WA: 064.A4.06
- k. 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 Control Room/ln-Plant Systems Outline Form ES-301-2
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Facility: DAVIS - BESSE Date of Examination: 2118108 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)
Svstem I JPM Title I Type Code' 1 Safety Function
- a. JPM 14 - Loss of Service Water Loop 1 to Primary loads A. D, M 4s
- d. JPM 97 - manually Actuate SFAS after some components 2 blocked
- e. I JPM 85 - Purge Containment
- 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 91.81 4
- E)mergency or abnormal in-plant '11 '11 ' 1
- L)ow-Power 11.11 1 N)ew or (M)odified from bank including 1(A) - 2 1 21 1 P)revious 2 exams ...3 I ' 3 I , ' 2 (randomly selected)
,R)CA ~ 1 1 ~ 11 1
'S)imulator WREG-1021. Revision 9
.~ . .~ .. . Control Roorn/ln-Piant
.. - Systems- Outline . Form ES-301-2
- a. 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
- b. 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 C. From Mode 3, 2, or 1, the candidate will be directed to RCS flow instruments Bank JPM 98 WA: 039A3 .02
- d. 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
- e. From a Mode 1 condition, the candidate will be direct to start the a purge on Containment.
Bank JPM 85 KA: 029A2.03
- f. From a plant startup condition, the candidate will be directed to recover from a Control Rod Drive Sequence Fault.
Bank JPM. 221 KA: 001 A 2 . 1 4
- 9. 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
I. 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
- 1. From a shutdown condition, the candidate will be direct to emergency shutdown the Emergency Diesel Generator.
Bank JPM 115 WA: 064.A4.06
- k. 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.:
= -
A E Scenarios I P v C
A N
P L
I E
N T
T Y
T' M
U T P
- RO-I (1 -
C 8RO-U -
IAJ
- S
- 0 X OR RO-I (21 RO-U 0
RO-I (3)
RO-U 0
RO-I (4)
RO-U 4
- 2.4 2.4 4 Page 1 of 4
ES-301, Rev. 9 Transient and Event Checklist Form ES-301-5 1 Faci ity Dav s-Besse
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Date of Exam 2/18/2008 Operating Test No E
P V P E 1 I 2 3 4 L N CREW CREW CREW CREW POSITION POSITION POSITION POSITION I T C
A T N Y T P
- E RO lX SRO-I (5)
SRO-U RO 4OR SRO-l (6; -
IC 4 4 2 SRO-U -
AAJ RO 1 1 1 SRO-I (71 4 4 2 SRO-U 2 2 1 0 2 2 8
RO 1 1 0 SRO-I (81 SRO-U 2 2 I
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
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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 7
- acility: Davis-Besse Date of Exam: 2/18/2008 Operating Test No.:
M I
N I
M U
2 1 2 2 1 1 0 1 1 4 4 2 2 1 0 2 2 istructions
) 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 ( I C ) 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
ES-401 Davis Besse 2008 NRC Written Exam Form ES-401-2 Written Examination Outline Emergency and Abnormal Plant Evolutions -Tier 1 Group 1 007 I Reactor Trip I 1 008 I Pressurizer Vapor Space Acddent I 3 022 I Loss of Reactor Coolant Makeup I 2 5 2.1.23 2.4.50 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 w r e coo11no Emergency Procedures I Pian Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.
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4.0 4.7 3.3 Ability to detenine and interpret the following as 025 I Loss of Residual Heat Removal Svstem i 4 they apply to the Loss of Residual Heat Removal System: Existence of proper RHR overpressure 3.4 orotection 065 I Loss of Instrument Air / 8 1 2.1.32 I Conduct of Operations: Ability lo explain and apply ail system IimitS and precautions. 3.8 Ability lo determine and interpret the following as E l 0 Post-Trio St2biliration X they apply to the (Post-Trip Stabiiization) Facility conditions and selection of appropriate procedures 4.0 during abnormal and emergency operations.
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Knowledge of the reasons for the foliowing 008 / Pressurizer Vapor Space Accident / 3 X AK3,02 responses as they apply to the Pressurizer Vapor Space Accident: Why PORV or code safety exit 3.6 temperature is below RCS or PZR temperature
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009 / Small Break LOCA 1 3 Ability to determine or interpret the following as they X 3.9 apply to a small break LOCA: Reactor trio setooints
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Ability to determine or interpret the following as they 01 1 I Large Break LOCA 13 EA2.05 apply to a Large Break LOCA: Significance of 3.3 charging pump operation t
Knowledge of the interrelations between the 015 I 1 7 I Reactor Coolant Pump Malfunctions 14 X Reactor Cwlant Pump Malfunctions (Loss of RC AU.10 2.8 Flow) and the following: RCP indicators and wntrols
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Conduct of Operations: Ability lo perform specific 022 I Loss of Reactor Coolant Makeup 12 2.1.23 System and integrated plant procedures during all 3.9 modes of olant ooeralion.
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rtnoweage of the opera1 ona mp cal ons t l inr 025 I Loss of Residual Heat Removal System 14 X AK1.O1 foi owing wncepts as tne) app IL Loss 01 3.9 ResdLa meal Removal Sbslcm Lossof R I I R ~
I during all modes of operaion
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ES-401 Davis Besse 2008 NRC Written Exam Form ES-401-2 Written Examination Outline Emergency and Abnormal Plant Evolutions -Tier 1 Group 1 I
Ability to determine and interpret the following as they apply to the Lo55 of Component Cooling 026 I Loss of Component Cwling Water I 8 Water: The length of time after the loss of CCW flow to a onnponent before that uxnpnent may be damaged
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027 I Pressurizer Pressure Control Sy5km Malfunction Conduct of Operations: Knowiedge of operator 13 responsibilities during all modes of plant operation.
029 /Anticipated Transient Without Scram (ATWS) I 1 I Knowledge of the interrelations between the ATWS and the following: Breakers, relays, and disconnects 038 I Steam Generator Tube Rupture / 3 I 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 054 I Lms of Main Feedwater14 they apply to the Loss of Main Feedwater (MFW): 4.4 49 HPI, under total feedwater loss conditions 055 IStation Blackout1 6 Ability to operate and monitor me following as they apply to a Station Blackout: Reduction of loads on the battery 3.5 I 50 4
Knowledgeof the reasons for the following responses as they apply to the Loss of Offsite 056 I Loss of Ofkite Power 16 Power: Order and time to initiation of power for the load sequencer Emergency ProceduresPlan: Ability to recognize abnormal indications for system operating 062 ILoss of Nuclear Service. Water I 4 X parameters which are entry level conditions for emergency and abnormal operating procedures.
Knowledge of the reasons for the following 065 I Loss of Instrument Air I8 responses as they apply to the LOSSof Instrument Air: Cross-over to backup air supplies Knowledge of the operational implications of the following concepls as they apply to the (Inadequate E04 Inadequate Heat Transfer 1 4 Heat Transfer): Annunciators and conditir-E05 Excessive Heat Transfer 14 I I NUREG-I 021 3
ES-401 Davis Besse 2008 NRC Written Exam Form ES-401-2 Written Examination Outline Emergency and Abnormal Plant Evolutions -Tier 1 Group 1 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 E10 Post-Trip Stabilization X EK2 2 emergency m a n ! . !he aecay heal removal 3.5 56 s)51ems. ana relations beween lhe proper operauon 01 Ihese systems 10 Ihe operauon 01 Ihe 13Clllry KIA Category Point Totals: 3/3 3 3 3 3 3/3 Grow Point Total: I 18/6 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 1 7 X I 2.1.23 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 4.0 they apply to the (Shutdown Outside Control room)
X m . 2 Adherence to appropriate procedures and operation 4.2 within the limitations in the fauliys license and amendments.
Ability to determine and interpret the following as they apply to the (Natural Csrculatkon Cooldown)
EO9 I Natural Circulation Operations 14 X EA2.2 Adherence to appropriate procedures and operation 4.0 within the limitations in the facility's license and amendments
--L Emergency Procedures i Plan Knowledge of E13 I Steam Generator Overpressure I 4 2.4.31 annunciators alarms and indications, and use of the 3.4 response instructions.
~
Ability to determine and interpret the following as they apply to the Continuous Rod Withdrawal:
001 I Continuous Rod Withdrawal I 1 I AA2.03 Proper actions to be taken if automatic safety functions have not taken place 4.5 Knowledge of the reasons for the following 005 I InoperablelStuckControl Rod I1 AK3.03 responses as they apply to the Inoperable I Stuck 3.6 Control Rod: Tech-Sow limits for rod mismatch Knowledge of the interrelations between the 060 I Accidental Gaseous RadWaste Release 1 9 AK2.02 Accidental Gaseous Radwaste Release and the 2.1 following: Auxiliary building ventilation system Ability to operate and monitor Me following as they 074 IInadequate Core Cooling I 4 EA1.09 3.7 amlv to a lnadeauate Core Coolina: CVCS Anow edge of h e operaiional mpllwtons 01 me Iohowing m c e p l s as lney apply to me (Plant A07 Plant Runback AK1.2 R..nhachJ hormal aonormal ana emergency 3.5 opera1 ng proceades associalea Yntn (Pant K..nhachl Knowledge of the reasons for the following responses as they apply to the (Turbine Trip)
A04 Turbine Trip AK3.2 3.4 62 Normal, abnormal and emergency operating orocedures asscciated with Turbine Triol.
1 068 Control Room Evacuation
~
X 1 1 3 2
- OlPralions AD I l y 10 explain and appl)
I (I K' ,>I
.: q . w m , m t s 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 Knowleage of tne operat onal implicat ons 01 Ihe followng mnCeptS as they apply to the (haldra E14 Natural Circulation Cwldown X EKl 3 C ICL alion C w l o o m ) Annmciatots and mno tions 35 64 ndicaling scgnals. and remeaal acuons asSOC ateo wlh the (Fralural Cirwlatton Cootdown)
<noMedge of the interrelationsbebeen the (EOP enclosures) and me followng Faulity's heal rernova systems. including pnmary Molani.
EOP enclosures X E<2 2 emergency coolant. me dewy heal removal 38 65 systems. and re ations between the proper opera1on of Inese systems 10 !he operaLon of me lac ddy KIA Category Point Total: 1/2 2 2 2 1 G ~ o L PPoint T ~ U I I 9/4 NUREG-1021
ES-401 Davis Besse 2008 NRC Written Exam Form ES-401-2 Written Examination Outline Plant Systems -Tier 2 Group 1 Conduct of Operations: Ability to recognize 004 Chemical and Volume indications for system operating parameten which 2.1.33 4.0 Control are entrylevel conditions for technical specifications.
Conduct of Operations: Ability to explain and 005 Residual Heat Removal 2.1.32 3.8 apply all system limits and precautions.
~
Ability to (a) predict the impacts of the following malfunctions or operations on the P S ; and (b) 007 Pressurizer RelieflQuench based on those predictions. use procedures to X M.03 3.9 Tank 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 012 Reactor Protection X A2.06 4.7 correct, control. or mitigate the consequences of those malfunctions or operations: Failure of RPS siqnal to trip the reactor Conduct of Operations: Ability to apply technical 026 Containment Spray 2.1.12 4,0 specifications for a system Ability to manually operate andlor monitor in the 003 Reactor Coolant Pump X A4.08 3.2 control r w m : RCP uxling water supplies Knowledge of the effect of a loss of malfunction 004 Chemical and Volume K6.02 on the following CVCS m p o n e n t s : 2.5 Control Demineralizers and im exchangers Knowledge of CVCS design feature@)andlor interlock(s) which provide for the following:
004 Chemical and Volume K4.11 Temperaturelpressure wntrol in letdown line: 3.1 Control prevent boiling, lifting reliefs, hydraulic shock, piping damage, and burst Knowledge of the effect of a loss or malfunction 005 Residual Heat Removal K6.03 on the following will have on the RHRS: RHR heat 2.5 exchanaer Ab l y 10 pred CI and or monitor manges In aarxnelefs (to prevent exceeang des.gn Itmts) 006 Emergency Core Cwling A1.02 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 007 Pressurizer RelieflQuench K1.03 CdLsetflecI relationships Detween me PRTS and Tank IIW lo own9 s>stems RCS
~
NUREG-I 021 7
ES-401 Davis Besse 2008 NRC Written Exam Form ES-401-2 Written Examination Outline Plant Systems - Tier 2 Group 1 Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS, and (b) based on those predictions. use procedures to 008 Component Cwling Water A2.03 3.0 correct, control, or mitigate the consequences of those malfunctions or operations: Highllow CCW temperature Emergency Procedures I Plan Knowledge of 008 Component Cooling Water 2.4.31 annunciators alarms and indications, and use of 3.3 the respmse instructions.
Knowledge of bus power supplies to the following:
010 Pressurizer Pressure Control K2.02 2.5 Controller for PZR spray valve Ability to manually operate andlor monitor in the 010 Pressurizer Pressure Control A4.02 3.6 control room: PZR heaters Knowledge of bus power supplies to the following:
012 Reactor Protection K2.01 3.3 RPS channels, components. and interconnections Ability to monitor automatic operation of the RPS.
012 Reactor Protection A3.02 3.6 including: Bistables 013 Engineered Safety Features Ability to monitor automatic operation of the A3.01 ESFAS including: Input channels and logic 3.7 Actuation 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 10 022 Containment Cooling A2.05 3.1 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 026 Containment Spray K2.01 3.4 Containment spray p m p s Knowledge of the operational implications of the 039 Main and Reheat Steam K5.08 following concepts s the apply to the MRSS: Effect 3.6 of steam removal on reactivity Knowledge of MFW design feature(s) andlor inter(ock(s) which provide for the following:
059 Main Feedwater K4.08 2.5 Feedwater regulatoly valve operation (on basis of steam flow, feed flow mismatch) 061 AuxillarylEmergency parameten (to prevent exceeding design limits)
A1.O1 3.9 Feedwater 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 I Ix Knowledge of the physical connections andlor cause-effect relationships between the ac 3.5 19 062 AC Electrical Distnbution K1'03 distribution system and the following systems: DC distribution 063 DC Electrical Distribution I I X K4.01 Knowledgeof dc electrical system design feature@)andlor interlwk(s) which provide for the followino: Manuallautomatictransfers of control 2.7 20 064 Emergency Diesel Generator I IX 2.4'6 Emergency Procedures IPlan Knowledge symptom based EOP mitigation strategies.
3.1 21 l l Knowledgeof the effect that a loss or malfunction 064 Emergency Diesel Generator X K3.03 of the EDIG system will have on the following: 3.6 22 EDlG (manual loads)
Knowledge of the operational implications as they apply to concepts as they apply to the PRM 2.5 23 073 Process Radiation Monitoring X K5.02 system: Radiation intensity changes with source distance Ability to monitor automatic operation of the SWS, 3.7 24 076 Service Water X A3'02 including: Emergency heat loads I
Knowledge of the physical connections andlor 076 Service Water K1.16 cause- effect relationships between the SWS and 3.6 25 I x I I I I I I I I the followina svstems: ESF I I I Conduct of Operations: Ability to explain and 3.4 26 078 Instrument fur 1 x 1 apply all system limits and precautions.
I I Knowledge of the effect that a 105s or malfunction 078 Instrument Air X K3.01 of the IAS will have on the following: Containment 3.1 27 I I I I I I I I air svstem I I I I /
Knowledgeof the effect that a loss or malfunction of the containment system will have on the 103 Containment X K3.01 3.3 28 following: Loss of containment integrity under shutdown conditions IUA Cateowv Point Totals: 1 313 I 3 3 3 3 2 2 2 2 / 2 3 2 Group Point Total: 2815 NUREG-1021 9
ES-401 Davis Besse 2008 NRC Written Exam Form ES-401-2 Written Examination Outline Plant Systems -Tier 2 Group 2 Emergency Procedures I Plan Knowledge of 016 Non-nuclear Instrumentation X 2.4.31 annmciatoffi
~- alarms and indications. and use of
~~
the re4ponse instructions.
033 Spent Fuel P w l Cooling IxI I 2.4.30 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 X A2.03 wrrect. Wntrol, or mitigate the wnsequences of those malfunctions or operations: Pressurellevel transmitter failure Knowledge of the effect that a loss or malfunction 001 Control Rod Dave K3.02 of the CRDS will have on the following: RCS Knowledge of the operational implications of the 015 Nuclear Insbumenlation K5.10 following concepts as they apply to the NIS: 2.8 30 Exwre Detector operation Ability to manually operate andlor monitor in the 016 Non-nuclear Instrumentation A4.02 2.7 31 control room: Recorders 011 Pressurizer Level Control Knowledqe of bus Dower supplies to the folloMng:
K2.02 Pressurizer heaters 3.1 32 System
~~~~ ~~ ~ ~
Ability to monitor automatic operation of the Fuel 034 Fuel Handling Equipment X A3.01 2.5 33 Handling System. including: Travel limits
~ ~~~
Ability to predict andlor monitor changes in parameterr (to prevent exceeding design limits) 035 Steam Generator X A1.02 3.5 34 associated with operating the SlGS controls
- including: S/G pressure Knowledge of the effect of a loss or malfunction 041 Steam Dumprrurbine Bypass K6.03 on the following will have on the SDS: Controller 2.7 35 Control and positioners. including ICs, SIG, CROS Knowledge of the physical connections andlor cause-effect relationships between the 056 Condensate System K1.03 2.6 36 Condensate System and the following systems:
LI3\1, 1.11 I..
Knowledge of circulating water system design 075 Circulating Water K4.01 feature(s) and interlock(s) which provide for the 2.5 37 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: 2/18/2008 I Category WA # Topic IR 2.1.20 Ability to execute procedure steps.
Ability to recognize indications for system 2.1.33 operating parameters which are entry-level 3.4
- 1. conditions for technical specifications. -
Conduct of Ability to use plant computer to obtain and Operations 2.1.19 evaluate parametric information on system or 3.0 component STATUS Subtotal 2.2.33 Knowledge of control rod programming.
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Knowledge of the process for conducting tests or 2.2.7 experiments not described in the safety analysis report. ~
2.2.27 Knowledge of the refueling process 2.6
- 2. -
Iquipment Control Ability to perform pre-startup procedures for the facility, including operating those controls 2.2.1 3.7 associated with plant equipment that could affect reactivity.
Subtotal A I Knowledge of 10 CFR: 20 and related facilitv personnel exposure.
2.3.1 1 Ability to control radiation releases.
- 3. Knowledge of radiation exposure limits and iadiation Control 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 Knowledge of chemistry / health physics tasks 2'4'36 during emergency operations.
Knowledge of procedures relating to emergency 2'4'28 response to sabotage. -
Knowledge of the organization of the operating 2.4.5 procedures network for normal, abnormal, and 2.9 emergency evolutions. -
I. Ability to recognize abnormal indications for mergency system operating parameters which are entry-2.4.4 4.0
'rocedures / Plan level conditions for emeraencv and abnormal I
" I operating proceddres I Aoilily to dlagnose and recognize trends in an 2.4.47 accurate and timely manner utilizing the 3.4 appropriate control room reference material.
II Subtotal
'ier 3 Point Total NUREG-I021 12
rier I Randomly Reason for Rejection
- roup Selected WA 211 003 A4.07 I Q # I Reselected 003 A4.08 as RCPs do not have a seal bypass at this facility 21 1 008 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 211 076 K1 '05 selected topic n,E v c n, Q #30 Reselected 015 WA 5.10 used since WA 5.01 is deleted from KA 212 ul.Jr\.J.ul catalogue 212 028 K2.01 Q #32 Reselected 01 1 K2.02 because equipment is not used at facility 1I 1 062 G2.2.25 Q #52 TS Basis for APE not applicable for RO knowledge. Reselected G2.4.4 I
1 11 Oo8 AA2'20 I Q #77 Difficulty in developing an SRO test item, and excessive overlap with other topics on exam. Reselected 008 AA2.30 112 032 G2.1.27 I Q #82 Difficulty in developing SRO topic to match KA. Reselected G2.1.23 211 005 G2.1.27 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.
211 026 G2'1'14 Reselected G2.1.12 Q #91 No action required for condition related to selected system. Reselected 212 017 G2.4.31 Stem 016 NUREG-I021 13
/FFeibix D Scenario Outline Form ES-D-1 I Facility: DAVIS-BESSE Scenario No.: 1 Op Test No.: NRC 2008 Examiners: Operators:
Initial Conditions: 100% power, MOL High Pressure Injection 1 pump is out of service Turnover:
Event Malf. Event Event No. Type* Description N - BOP Swap Turbine Plant Cooling Water Pumps HU21A C - RO RCP 1 Loss of oil to upper bearing TS - SRO R-RO Power reduction prior to stopping RCP 1-1 R - SRO L1621 SFFD IksR:Ro I C - BOP I Stuck Control Rod at -75% powei I MFPT 2 Vibration I HH43 M - All I Reactor Coolant System leak - 2500 gpm 7 LGLE C-RO I SFAS Modules L231 fail to trip 8 LlTL28 I-BOP SG 1 level transmitter fails mid-scale 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, REACTOR COOLANT PUMP AND MOTOR ABNORMAL OPERATION.
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%
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
=acility: DAVIS-ESSE Scenario No.:
-rxaminers: Operators:
Initial Conditions: 50 - 60% power, 3 RCP. MOL No equipment out of service W
Turnover: l P
Event Malf. Event Event No. No. Type' Description 1 N - All Transfer Gland Steam from Main Steam to Auxiliary Steam 2 TS - SRO The crew will be notified that Auxiliary Feedwater Pump 1 has no Governor oil 3 HI70 C-RO Make-up valve (MU 32) fails to operate in auto 4 L153B R-RO Control Rod drop TS - SRO 5 NP03 I - BOP Loss of NNI X DC NP05 N - SRO 6 PLZZ M -All Loss of Offsite AC 7 G529B C-RO Emergency Diesel Generator 1 fails to auto start 8 SFERE C - BOP AFW Pump 2 governor valve closes
- (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 Malf. Event Event No. No. Type* Description 1 N - BOP Perform TG OversDeed Test. DB-SS-04154 2 B2MlN C-RO Make-up Pump 1 trips TS - SRO 3 BV24B C-RO MU 1903 fails closed -.
4 HH50 R-RO SG 1 tube leak and plant shutdown TS - SRO 5 UFO9A C-RO Turbine Vibration UFOBA UFllA UFO6A 6 SFDPC M -All MFPT 2 weed lowers and plant triD 7 M -All SG Tube Rupture F30AB C -BOP SFRCS Ch 1 fails to actuate F30AC 9
- (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] . .
Facility: DAVIS-BESSE Scenario No.:4 Op Test No.: NRC 2008 Examiners: Operators:
~~
Initial Conditions: 100% power, MOL No equipment out of service Turnover:
Event Event Type* Description
~~
1 N - BOP Manual Voltage Regulator operations per DB-OP-06301, step 3.4 2 L6P1D I-RO SFAS Containment Pressure transmitter fails low TS-SRO 3 HlClC I -RO I Pressurizer Temperature fails mid-scale I 4 HDP309 c - BOP Turbine Plant Cooling Water Pump 3 trips 5 HN29A R -RO RCP 2-1 Seal Cooler leak TS - SRO 6 SFEF M -All Steam Leak in Fan Allev 7 FKMID C - BOP 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 Comment Resolution
and SRO) - JPM 227 - However, the WA for a Conduct of Operations The WA for the JPM was Calculate RCS with JPM needs to come from Section 2.1 instead of changed to 2.1 25.
F755 Unavailable 2.2.
- 2. Expand the title of the JPM to "Calculate 2. Comment incorporated.
RCS Flow with F755 Unavailable."
AN2 - Review an The WA for the JPM was Auxiliary Feedwater changed to 2.1.33.
Surveillance Test and Determine Operability
AN1 -Calculate However, this WA is not appropriate, since the The WA for the JPM was Radiation Release WA is for a planned gaseous release, whereas changed to 2.3.10.
Using SGTL the JPM is associated with a release due to a Abnormal Procedure, SGTL. Suggest possible use of WA 2.3.10 DE-OP-02531, (2.9/3.3) for this JPM.
Attachment 1
- 2. For an RO, the WA value for 2.3.8 is only 2.3 2. Comment incorporated.
(i.e., < 2.5). Need to provide some justification The WA for the JPM was for using this JPM with a WA < 2.5. changed to 2.3.10, with a values of 2.713.3.
- JPM 14 -LOSS Of However, this WA is for the Condenser Air The WA for the JPM was Service Water Loop 1 Removal System (with WA values of 1.E,1.9) changed to 062 AA1.02.
to Primary Loads that have nothing to do with the JPM.
- JPM 48 -Exchange However, this WA is for the Main and Reheat facility discretion. The new RCS Flow Amphynol Steam System that has nothing to do with the JPM has a WA of 012 JPM. The WA is associated with Safety A4.03, which is associated Function 4 (heat removal), whereas the JPM is with Safety Function 7.
associated with Safety Function 7 (instrumentation).
Control Room JPM G The WA for this JPM is listed as 2.3.11.2, which Comment incorporated.
- JPM 215 - Respond is not a valid WA. The WA for the JPM was to a High Station Vent changed to 071 A3.03.
Radiation Alarm 1 of2
DAVIS-BESSE 2008 INITIAL LICENSE EXAM OPERATING OUTLINE COMMENTS
-Sync the Main associated with Safety Function 4 (Heat The WA for the JPM was Generator to the Grid Removal) (2.4, 2.3). For this JPM to be changed to 062 A4.07.
(RO only) associated with Safety Function 6 (Electrical),
the WA should be changed to 062 K4.05 or some other WA associated with Safety Function 6.
Primary Side RO is associated with Safety Function 9 The WA for the JPM was CTRM Evacuation (Radioactivity Release), which is not the correct changedto 068AA1.13, WA. A WA associated Safety Function 2 which is associated with (Inventory Control) should be listed. Safety Function 2 (Inventory Control) (i.e., Makeup Pump control during a Control Room Evacuation).
Scenario 1 Outline Editorial: Delete Event No. 9 from sheet. Comment incorporated.
9.
- 10. Scenario 2 Outline 1. For Event 1 , it appears that only the BOP 1. Comment incorporated.
should be given credit for this Normal evolution.
- 2. For Event 5, delete "N-SRO" under Event 2. Comment incorporated
- TvDe.
- 11. Scenario 3 Outline 1. For Event 5, Turbine Vibration, there is no 1. Comment incorporated mention of this event in the Scenario 3 General Description.
- 2. For Event 5, change the Event Type such that 2. Comment incorporated the BOP gets credit for the Event, instead of the RO.
- 12. General Comment 1 Only Scenario 1 has any major equipment Comment incorporated.
for all Scenarios initially 00s. The other scenarios should have Scenarioschanged such some equipment initially 00s. that the Initial Conditions specity equipment initially 00s.
- 13. General Comment 2 None of the scenarios have a Turnover shown Comment incorporated.
for all Scenarios on the outline sheets. The crews should have a Scenarioschangedsuch Turnover for each scenario. that a Turnover is shown on the outline sheet.
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