ML041830296

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Final Outlines for Davis-Besse Initial Examination - May 2004
ML041830296
Person / Time
Site: Davis Besse Cleveland Electric icon.png
Issue date: 05/03/2004
From: House J
FirstEnergy Nuclear Operating Co
To:
Office of Nuclear Reactor Regulation
Shared Package
ML041600289 List:
References
50-346/04-301
Download: ML041830296 (16)
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Text

FINAL OUTLINES FOR THE DAVIS-BESSE INITIAL EXAMINATION - MAY 2004

- ES-301 Administrative Topics Outline Form ES-30 I -I Facility: Davis-Besse NPS Date of Examination: 5/3/04 - 5/7/04 Examination Level (circle one): RO / SRO Operating Test Number: 1,2 and 3 Administrative Topic (see Note)

Conduct of Operations New Not used on previous exam WA - GEN 2.1.17 3.5/3.6 Conduct of Operations New Not used on previous exam WA - GEN 2.1.12 2.9/4.0 Equipment Control Modified lised on 2002 exam WA - GEN 2.2.12 3.0/3.4 3adiation Control Modified Jsed on 2000 exam

.</A - GEN 2.3.1 1 2.7/3.2 Zmergency Plan

\\lew

\\lot used on previous exam UA-GEN 2.4.432.8/3.5 Describe activity to be performed:

Notification to the Federal Aviation Administration for Cooling Tower Aviation Light Failure Review and Correct a Shutdown Margin Calculation Control of Locked Valves during Post Maintenance Testing Review of Containment Pressure Reduction Release Manually Activate the Emergency Response Organization 3roup Page 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.

NUREG-1021, Draft Revision 9 22 of 27

ES-301 Administrative Totics Outline Form ES-301-1 3",,>.<

Facility: Davis-Besse NPS Date of Examination: 5/3/04 - 5/7/04 Examination Level (circle one): RO / SRO Operating Test Number: 1 and 2 Administrative Topic (see Note)

Conduct of Operations New Not used on previous exam WA - 192002 K1.13 3.5/3.7 Conduct of Operations Equipment Control Modified Used on 2002 exam WA - GEN 2.2.12 3.0/3.4 Radiation Control New Not used on previous exam WA - GEN 2.3.1 1 2.7i3.2 Emergency Plan Modified Not used on previous exam WA - GEN 2.4.39 3.3/3.1 Describe activity to be performed:

Shutdown Margin Calculation with > I Known Stuck Rod Control of Locked Valves during Post Maintenance Testing Containment Pressure Reduction Release Offsite Dose Assessment Using a Nomogram 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.

NUREG-1021, Draft Revision 9 22 of 27

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

Nl A1 Sl L Exam Level (circle one): RO / SRO(I) / SRO(U)

Operating Test No: 2 and 3 Safety Function I

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

System/ JPM Title

a. Obtain 0% Lights and In-Limit Lights for All Control Rods in a Group
b. Manual SFAS Level 1 and 2 Actuation
2. Spray the Pressurizer for Boron Equalization
3. 1 Hour Shutdown of Decay Heat Pumps to Support Core Alterations
3. Operate Containment Hydrogen Purge/Dilution Systems J. Restore a tripped RPS Channel to Service
7. Shift from 4 to 2 Circulating Water Pump Operation n-Plant Systems (3 for RO; 3 for SRO-I: 3 or 2 for SRO-U) 4P 5

. High Pressure Injection Alternate Minimum Recirc Flowpath

. Emergency Operation of the Startup Feedwater Pump D

4 s

c. Station Blackout Diesel Generator Emergency Shutdown 6

. Type Codes: (D)irect from bank, (M)odified from bank, (N)ew, (A)Iternate path, (C)ontrol oom, (S)imulator, (L)ow-Power, (R)CA 23 of 27 NUREG-I 021 Draft Revision 9

Control Room/ln-Plant Systems Outline Form ES-301-2 N, A, s, L U

t/

8 Facility: Davis Besse Nuclear Power Station Date of Examination: 5/3/04 - 5/10/04 Exam Level (circle one): RO / SRO(I) / SRO(U)

Operating Test No:

1 Nl R

i. High Pressure Injection Alternate Minimum Recirc Flowpath Control Room Systems (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U) 3 System/ JPM Title
j. Emergency Operation of the Startup Feedwater Pump
k.
a. Obtain 0% Lights and In-Limit Lights for All Control Rods in a Group D

4s

b. Manual SFAS Level 1 and 2 Actuation
j. Emergency Operation of the Startup Feedwater Pump
k.

~

C.

D 4s

d.
e.
f.
h. Shift from 4 to 2 Circulating Water Pump Operation Safety Function N, AI s, L
  • Type Codes: (D)irect from bank, (M)odified from bank, (N)ew, (A)Iternate path, (C)ontrol room, (S)imulator, (L)ow-Power, (R)CA 23 of 27 NUREG-1021, Draft Revision 9

Control Roomlln-Plant Systems Outline Form ES-301-2 Type Code*

Facility: Davis Besse Nuclear Power Station Date of Examination: 5/3/04 - 511 0104 Exam Level (circle one): RO / SRO(I) / SRO(U)

Operating Test No: 1 and 2 Safety Function Control Room Systems (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U)

System/ JPM Title

a. Obtain 0% Lights and In-Limit Lights for All Control Rods in a Group 1

l 2

b. Manual SFAS Level I and 2 Actuation
s. Spray the Pressurizer for Boron Equalization 3
j. 1 Hour Shutdown of Decay Heat Pumps to Support Core Alterations 4P I
3. Operate Containment Hydrogen Purge/Dilution Systems

'. Energize D I Bus from Emergency Diesel Generator 1

1. Restore a tripped RPS Channel to Service 7

I. Shift from 4 to 2 Circulating Water Pump Operation I

N,A, s, L 8

I I

n-Plant Systems (3 for RO; 3 for SRO-I: 3 or 2 for SRO-U)

. High Pressure Injection Alternate Minimum Recirc Flowpath

. Emergency Operation of the Startup Feedwater Pump D

I 4s

. Station Blackout Diesel Generator Emergency Shutdown Type Codes: (D)irect from bank, (M)odified from bank, (N)ew, (A)lternate path, (C)ontrol oom, (S)imulator, (L)ow-Power, (R)CA 23 of 27 NUREG-1021, Draft Revision 9

Appendix D Scenario Outline Form ES-D-1 Event No.

I 2

I, Malf.

Event Event No.

Type*

Description R (RO)

Lower Reactor power using the Reactor Demand Station KEP2E C (BOP)

Inadvertent Service Water Pump 1 trip Facility: Davis-Besse NPS Scenario No.:

1 Op-Test No.: 1, 2, and 3 5

6 7

8 Examiners:

Operators:

SFEG C (BOP) lsolable Main Steam Line leak (Tech. Spec.)

P8RFC M (All)

Loss of Offsite Power GF32B B2M21 C (RO)

Makeup Pump 2 fails to start C (RO)

Emergency Diesel Generator 2 fails to automatically start Initial Conditions: 28% Reactor Power, Emergency Diesel Generator 1 Out of Service 11 Turnover:

Plant Shutdown in progress due to Tech Spec 3.8.1.1 3

I R3N5 I I (RO) I Power Range Nuclear Instrument 5 fails high (Tech Spec) 11 4 I UKRF I C (BOP) I Main Turbine Quill Shaft failure (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor NUREG-1021, Draft Revision 9 40 of 41

Appendix D Scenario Outline Form ES-D-I

' (Ro)

C (BOP)

C (SRO)

I (BOP) d Safety Features Actuation System RCS pressure transmitter fails high (Tech. Spec.)

High Pressure Feedwater Heater tube leak EDG 1 Trouble Alarm (Tech. Spec.)

Main Steam Header Pressure transmitter fails mid-scale Facility: Davis-Besse NPS Scenario No.:

2 Op-Test No.: 1, 2, and 3 R (RO)

I (BOP)

C (RO)

Examiners:

Operators:

Power reduction Loss of all Condensate Pumps ATWS Initial Conditions: 100% Reactor Power, Makeup Pump 1 Out of Service Turnover:

Event Malf. No.

7 I

-qTz FAKMD G530A L1 T2V or L1 T2N F4 I S I 595 L4, L8 9

I FKMID 10 I "46

  • (N)o rm al,

Event Type*

I Event Description C (RO) I Small RCS leak C (BOP) I Auxiliary Feedwater Target Rock Valve fails open M (All) I Small Break LOCA (R)eactivity, (I)nstrument, (C)omponent, (M)ajor NUREG-1021, Draft Revision 9 40 of 41

Appendix D Scenario Outline Form ES-D-1 Malf.

No.

Facility: Davis-Besse NPS Scenario No.:

3 Op-Test No.:

3 Examiners:

Operators:

Event Event Type*

Description Initial Conditions: 50% Reactor Power, both Main Feedwater Pumps in service I

R (RO)

Turnover:

increase Main Feedwater Pump 2 has been placed in service, readv to beain power Raise reactor power HlClC Fail Pressurizer Temperature Transmitter mid-scale (Tech.

l(R0) ec DCMI "51 LITL20 BMFI 1 C (RO) I Makeup Filter differential pressure high C (BOP)

M (All)

I (BOP)

Loss of Condenser Vacuum Steam Generator Tube Rupture Steam Generator Level Transmitter fails mid-scale r

1 C (SRO) 1 Loss of Shield Building Integrity (Tech Spec)

Event i;;7 1

2 3

4 5

6 7

NUREG-1021, Draft Revision 9 40 of 41

ES-401 PWR Examination Outline Form ES 401-2

3. Generic Knowledge and Abilities Categories Note:
1.
2.
3.
4.
5.
6. *
7.
8.
9.

Ensure that at least two topics from every WA category are sampled within each tier of the RO outline (i.e., the "Tier Totals" in each WA category shall not be less than two). Refer to Section D.l.c for additional guidance regarding SRO sampling.

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 fl from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO-only must total 25 points.

Select topics from many systems and evolutions; avoid selecting more than two WA topics from a given system or evolution unless they relate to plant-specific priorities.

Systems/evolutions within each group are identified on the associated outline.

The shaded areas are not applicable to the categorykier.

The generic (G) WAS in Tier 1 and 2 shall be selected from Section 2 of the WA Catalog, but the topics must be relevant to the applicable evolution of system. The SRO WAS must be linked to 10 CFR 55.43 or an SRO-level learning objective.

On the following pages, enter the WA numbers, a brief description of each topic, the topics' importance rating (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; summarize all the SRO-only knowledge and non-A2 ability categories in the columns labeled "K" and "A." Use duplicate pages for RO and SRO-only exams.

For Tier 3, enter the WA numbers, descriptions, importance ratings, and point totals on Form Refer to ES-401, Attachment 2, for guidance regarding the elimination of inappropriate WA statements.

ES-401-3.

NUREG-1021, Draft Revision 9 22 of 34

ES-401 PWR Examination Outline Form ES-401-:

Emergency i K

1 E/APE # / Name / Safety Functions 000007 (BW/E02&ElO; CE/E02) Reactor Trip -

Stabilization - Recovery / 1 X

000009 Small Break LOCA I 3 000017/17 RCP Malfunctions / 4 000026 Loss of Component Cooling Water I 8 000027 Pressurizer Pressure Control System Malfunction I 3 000029 Anticipated Transient w/o Scram 11 000038 Steam Generator Tube Rupture I 3 X

000040 (BW/E05; CE/E05; W/E12) Steam Line Rupture - Excessive Heat Transfer / 4 000054 (CE/EO6) Loss of Main Feedwater I 4 000055 Station Blackout I 6 300056 Loss of Off-site Power I 6 300057 Loss of Vital AC Elec. Inst. Bus / 6 300058 Loss of DC Power 1 6 300062 Loss of Nuclear Service Water / 4 300065 Loss of Instrument Air / 8 ME04 LOCA Outside Containment / 3 WE1 1 Loss of Emergency Coolant Recirc. 14 3WlE04; WlE05 Inadequate Heat Transfer -

-oss of Secondary Heat Sink / 4 id Abnormal Plant Evolul Ins -Tier l/Group 1 (RO / SRO)

WA Topic(s)

IR EK1.02 - Shutdown Margin I 3.4 I 1

AA1.05 - Operate or monitor LPI System I 3.4 I 1

SRO AA1.07 - Reseating of code safety and PORV I 4.2 I I

I 1

EK3.13 - Stopping the affected RCP 3.4 1

EK2.02 - Interrelations between pumps and LBLOCAs 2.6 1

SRO 2.4.31 - Knowledae of annunciators and alarms 3.4 1

AK1.03 - Relationship between charging flow and Pzr level 3.0 SRO AA2.01 Whether MU line leak exists 3.8 2

AA2.05 - Limitations on LPI flow I 3.1 I 1

AK3.03 - Actions in EOP for loss of CCW

.^

1 4.u 1

SRO AA2.04 - Normal values and upper limits of components cooled by CCW AK3.04 - Why level recovers slower on loss of Pzr 2.9 level 2.8 1

EA1.08 - 0 erate reactor tri ushbuttons SRO EK1.02 - Leak rate vs. ressure dro 2.2.22 - Knowledge of LCOs and safety limits 3.4 1

EA1.04 - Reduction of loads on the battery 3.5 1

AA2.43 - Determine occurrence of a turbine trip 3.9 1

AK3.01 - EOP actions for loss of a vital bus 4.1 1

2.4.10 - Annunciator response procedure 3.0 1

SRO AA2.01 - Verify loss of DC has occurred 4.1 1

2.1.28 - purpose of major components 3.2 1

SRO 2.1.32 - Ability to explain and apply limits and 3.8 1

precautions AA2.05 - Determine when to shutdown 3.4 1

EK2.2 - Interrelation between heat removal systems and inadequate heat transfer 4.2 1

18/7 Group Point Total:

NUREG-1021, Draft Revision 9 23 of 34

PWR Examination Outline d Abnormal Plant Evolutions -Tier 11Group 2 (RO / SRO)

Form ES-4014 ES-401 EIAPE # I Name / Safety Functions 000028 Pressurizer Level Malfunction 12 000032 Loss of Source Range NI 17 000033 Loss of Intermediate Range NI / 7 X

000036 (BW/AO8) Fuel Handling Accident I 8 000037 Steam Generator Tube Leak 1 3 000051 Loss of Condenser Vacuum 14 000059 Accidental Liquid RadWaste Rel. I 9 000060 Accidental Gaseous Radwaste Rel. I 9 000061 ARM System Alarms I 7 000067 Plant Fire On-Site I 9 000068 (BWIAO6) Control Room Evac. / 8 000069 (W/E14) Loss of CTMT Integrity 000074 (W/E06&E07) Inad. Core Cooling 1 4 000076 High Reactor Coolant Activity I 9 W/EOl & E02 Rediagnosis & SI Termination WlE13 Steam Generator Over-pressure I 4 W/E15 Containment Flooding / 5 W/E 16 High Containment Radiation I 9 BW/AOl Plant Runback I 1 BW/A02&A03 LOSS of NNI-W / 7 BW/A04 Turbine Trip 14 BW/A05 Emergency Diesel Actuation 16 BW/AO7 Flooding 18 BWlEO3 Inadequate Subcooling Margin 14 BWlE08; WlE03 LOCA Cooldown - Depress. 14 WA Category Point Totals:

( 0 WA Topic@)

K K

A A

G 2

3 1

2 SRO AK2.02 - Breakers, relays, disconnects and AK1.01 - Relationship between boron addition and X

Ctrm switches Tave SRO AA1.10 - CVCS centrifugal charging pumps X

X AK3.01 - Startup termination on source range loss AK1.O1 - Effect of voltage change on IR Nl's SRO AA2.10 -Tech Spec limits if both IR channels X

fail SRO 2.1.7 - Ability to evaluate plant performance and X

make judgements X

AK3.01 - EOP guidance for loss of Ctmt integrity AA1.2 - Monitor operating behavior characteristics of X

facility X

2.4.48 - Interpret Control Room indications AK2.1 - Interrelation between EDG and safety X

components AK2.2 - Interrelation between flooding and heat removal systems SRO AA2.2 - Adherence to procedures and operation X

X within limits X

EA2.2 - Adherence to procedures 1 I 0 I 1 I 2 I 1 I Group Point Total:

NUREG-1021, Draft Revision 9 24 of 34

ES-401 PWR Examination Outline Plant Systems -Tier 2/Group 1 (RO / SRO)

Form ES-401-2 WA Topic@)

GI A1.O1 - Monitor RCP vibrations I K2.02 - Knowledae of Dower sumlies to MU Pumos 1 SRO A4.13 - Moiitor MUT levei and pressure coitrol I A4.01 - Operate/monitor RHR Pumps I I K1.04 - Causeleffect between ECCS and Aux Spray I K3.01 - Effect malfunction will have on Ctmt K4.01 - Quench Tank cooling K1.02 - Cause/effect between loads and CCW K2.02 - Power supply for spray valve A3.01 Monitor PRT during PORV test SRO K6.02 - Effect of Pzr malf on pressure control K1.01 - Cause/effect between RPS and 120 vac K4.04 - Knowledge of RPS redundancy K6.01 - Effect of loss of detector on SFAS K3.02 - Effect of loss of CAC on Ctmt instruments X SRO 2.1.20 - Ability to execute procedure steps K1.01 - Cause/effect between CSS and ECCS K5.05 - Basis for RCS cooldown limits A4.04 - Operate AFP turbines SRO A2.01 - Predict impact on steam flow paths during LOCA A2.04 - Impact of loss of Cond Pumps K3.03 - Effect of M!W malfunction on SGs K5.05 - FW line voiding & water hammer Al.02 - Monitor changes in SG pressure A3.04 - Monitor auto operation of inverter K2.01 - Power supplies for major DC loads I K1.05 - Cause/effect between EDG and starting air K3.01 - Effect of malfunction on a release I Al.02 - Monitor TPCW temperatures 3.6 q-+

3.0 2.7 3.0 4.2 2.6 3.2 2.8 I

1 I Group Point Total:

NUREG-1021, Draft Revision 9 25 of 34

?

ES-401 PWR Examination Outline Plant Systems - Tier 2IGroup 2 (RO / SRO)

Form ES-401-2 System # I Name K

K K

K K

K A

A A

1 2

3 4

5 6

1 2

3 001 Control Rod Drive X

002 Reactor Coolant I

I I

I I

I I

I I

041 Steam Dumpnurbine Bypass Control X

045 Main Turbine Generator X

055 Condenser Air Removal 068 Liquid Radwaste KIA Topic(s)

IR K1.07 - Cause/effect between RCS and vessel level 3.5 1

K6.04 - effect of level ctlr malf on level control 3.1 1

A4.01 - Operate rod selection control 3.3 1

K2.01 - NI power supplies 3.3 1

K5.02 - Concept of saturation and subcooling 3.7 1

K3.03 - Effect of SFPCS malfunction on spent fuel temperature 3.0 1

A2.03 - Predict pressurellevel transmitter failure 3.4 1

A3.05 - Monitor automatic operation to control MS pressure 2.9 1

2.4.1 1 - Knowledge of abnormal condition procedures 3.4 1/

SRO A2.15 - Impact of turbine overspeed 2.6 1

K4.02 - ARM design for Fuel Building isolation 3.2 1

SRO 2.4.4 - Recognize abnormal conditions 4.3 1

NUREG-1 021, Draft Revision 9 26 of 34

ES-401 Generic Knowledge and Abilities Outline (Tier 3)

Form ES-401-3 P

Facility: Dav Category

1.

Conduct of Operations

2.

Equipment Control

3.

Radiation Control

4.

Emergency Procedure

/ Plan

-Besse NPS Date of Exam
5/10/04 2.2.1 Ability to perform pre-start procedures 3.7 1

2.2.1 1 Knowledge of process for temp changes 2.5 1

2.2.28 Knowledge of new and spent fuel movement 2.6 1

2.2.1 8 Knowledge of process for managing maintenance 3.6 1

1 1

1 2.4.1 Knowledge of EOP entry conditions 4.3 1

2.4.43 Knowledge of emergency communications 2.8 1

2.4.45 Ability to prioritize annunciator alarms 3.3 1

2.4.22 Knowledge of basis for prioritizing safety functions 4.0 1

2.4.28 Knowledge of procedures related to sabotage 3.3 1

27 of 34 NURG-1021. Draft Revision 9

1 ES-401 Record of Rejected WAS Form ES-401-4 Reason for Rejection WA, since malfunction of the Pzr level - heater interlock has no effect on Pzr level control svstem.

i NUREG-1021, Draft Revision 9 28 of 34

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