ML023460236
| ML023460236 | |
| Person / Time | |
|---|---|
| Site: | Kewaunee  |
| Issue date: | 10/11/2002 |
| From: | NRC/RGN-III/DRS/OLB |
| To: | Short P Nuclear Management Co |
| References | |
| 50-305/02-301 | |
| Download: ML023460236 (46) | |
Text
OUTLINE SUBMITTAL WITH NRC COMMENTS FOR THE KEWAUNEE INITIAL EXAMINATION - AUG/SEP 2002
Kewaunee Nuclear Power Plant N490 Highway 42 Kewaunee, WI 54216-9511 920.388.2560 Point Beach Nuclear Plant 6610 Nuclear Road Two Rivers, WI 54241 920.755.2321 Committed to Nuclear &cellence Kewaunee / Point Beach Nuclear Operated by Nuclear Management Company, LLC NRC-02-049 May 20, 2002 Mr. D. E. Hills, Chief Operations Branch U. S. Nuclear Regulatory Region III 801 Warrenville Road Lisle, IL 60532-4351 Commission
Dear Mr. Hills:
Docket Number 50-305 Operating License DPR-43 Kewaunee Nuclear Power Plant Initial Operator Licensing Examination Outlines In response to your letter dated February 22, 2002, enclosed are the initial operator licensing examination outlines. As confirmed with your staff, the examinations are currently scheduled for the weeks of August 26, September 3, and September 9, 2002. NUREG 1021 physical security requirements state that the enclosed examination materials shall be withheld from public disclosure until after the examination is complete.
Please contact Mr. Chuck Sizemore at 920/388-8873 or Mr. Phillip Short at 920/388-8229 if you have questions regarding the examination outlines or require additional information.
Sincerely, Thomas Coutu Plant Manager Enclosures bcc w/oe:
C. Sizemore File K. Davidson G. Harrington
- a2124
ES-201 Examination Outline Form ES-201-2 Quality Checklist Facility:
Date of Examination:
Initials Item Task Description a
I4 1.
- a. Verify that the outline(s) fit(s) the appropriate model per ES-40 1.
rd W
- b.
Assess whether the outline was systematically and randomly prepared in accordance with R
Section D. I of ES-401 and whether all K/A categories are appropriately sampled.
T
- c. Assess whether the outline over-emphasizes any systems, evolutions, or generic topics.
T N
- d. Assess whether the justifications for deselected or rejected K/A statements are appropriate.
3 l
- a. Using Form ES-301-5, verify that the proposed scenario sets cover the required number of normal x
l
- 2.
evolutions, instrument failures, and major transients.
- b. Assess whether there are enough scenario sets (and spares) to test the projected number and mix of S
applicants in accordance with the expected crew composition and rotation schedule without compromising X
I exam integrity; ensure each applicant can be tested using at least one new or significantly modified M
scenario, that no scenarios are duplicated from the applicants' audit test(s)*, and scenarios will not be repeated over successive days.
_/X(@
- c. To the extent possible, assess whether the outline(s) conform(s) with the qualitative and quantitative criteria specified on Form ES-301-4 and described in Appendix D.
- 3.
- a.
Verify that:
(I) the outline(s) contain(s) the required number of control room and in-plant tasks, W
(2) no more than 30% of the test material is repeated from the last NRC examination,
/
(3) *no tasks are duplicated from the applicants' audit test(s), and T
(4) no more than 80% of any operating test is taken directly from the licensee's exam banks.
- b.
Verify that:
(I) the tasks are distributed among the safety function groupings as specified in ES-301, (2) one task is conducted in a low-power or shutdown condition, (3) 40% of the tasks require the applicant to implement an alternate path procedure, (4) one in-plant task tests the applicant's response to an emergency or abnormal condition, and (5) the in-plant walk-through requires the applicant to enter the RCA.
v A
- c. Verify that the required administrative topics are covered, with emphasis on performance-based activities.
A Y548
- d. Determine if there are enough different outlines to test the projected number and mix of applicants and ensure that no items are duplicated on successive days.
A t
4.
- a. Assess whether plant-specific priorities (including PRA and IPE insights) are covered in the appropriate X,
G exam section.
E
- b. Assess whether the 10 CFR 55.41/43 and 55.45 sampling is appropriate.
NE
- c. Ensure that K/A importance ratings (except for plant-specific priorities) are at least 2.5.
R
- d. Check for duplication and overlap among exam sections.
A
- e. Check the entire exam for balance of coverage.
L
_t
/
Printed Name/ Signature Date
- a.
Author PA__,_ A SA.rt -
Al p
_/l )_
- b.
Facility /Reviewer(*)
Daze1 e'eke. /
1-02
- c.
NRC Chief Examiner (#)
IfA,,-/,
/ E Age IA.
J '
- d.
NRC Supervisor
_)
A i/
Note:
Not applicable for NRC-developed examinations.
Independent NRC reviewer initial items in Column c;" chief examiner concurrence required.
NUE(12,eeio 8Z SuP NUREG 1021, Revision 8, Supplement I
ES-301 Administrative Topics Outline Form ES-301-1 Facility: KEWAUNEE NUCLEAR PLANT Date of Examination: 8/26/02 Examination Level (circle one):
RO / SRO Operating Test Number: 2002301 Administrative Describe method of evaluation:
Topic/Subject ONE Administrative JPM, OR Description TWO Administrative Questions A. I Conduct of JPM: Perform a Pre-Critical Checklist.
Operations/
Reactor Plant Startup Requirements JPM: Perform an Estimated Critical Position (ECP).
Plant Parameter Verification A.2 Equipment JPM: Review a tagout for accuracy.
Control/
Tagging &
Clearances N/A A.3 Radiation Control/
JPM: Perform a Radiation Monitor Functional Test.
Ability to perform procedures to guard against N/A personnel exposure.
A.4 Emergency Plan/
JPM: Manually start the Control Room Post Accident Recirculation system in Emergency response to a security threat.
Facility N/A NUREG-1021, Revision 8
ES-301 Administrative Topics Outline Form ES-301-1 Facility: KEWAUNEE NUCLEAR PLANT Date of Examination: 8/26/02 Examination Level (circle one):
RO / SRO Operating Test Number: 2002301 Administrative Describe method of evaluation:
Topic/Subject
- l. ONE Administrative JPM, OR Description 2 TWO Administrative Questions A.l Conduct of JPM: Perform a Pre-Critical Checklist.
Operations/
Reactor Plant Startup Requirements JPM: Perform an Estimated Critical Position (ECP).
Plant Parameter Verification A.2 Equipment JPM: Review a tagout for accuracy.
Control/
Tagging &
Clearances N/A A.3 Radiation Control/
JPM: Review/Authorize an Emergency Radiation Work Permit (RWP).
Knowledge of Radiation Exposure Limits l
N/Al A.4 Emergency Plan/
JPM: Make an Emergency Plan Classification, including Protective Action Emergency Recommendations (PARS).
Action Levels and Classifications NUREG-1021, Revision 8
ES-301 Control Room Systems and Facility Walk-Through Test Outline Form ES-301-2 Facility: KEWAUNEE NUCLEAR PLANT Date of Examination: 8/26/02 Exam Level (circle one):RO / SRO (I) / SRO(U)
Operating Test No: 2002301 B. I Control Room Systems System JPM Title Type Safety Code*
Function
- a. Control Rod Drive System / Perform Actions To Stop A Continuous Rod M, A, S I
Withdrawal.
- b. Emergency Core Cooling System / With A Fire In A Dedicated Zone -
D, S, L 2
Restore RCS Inventory using the SI System.
- c. Reactor Coolant Pump System / Start a Reactor Coolant Pump.
D, S, L 4
- d. Containment Spray System / Secure Containment Spray Pumps.
NSL 5
(ESF)
- e. A.C. Electrical Distribution / Shift Bus 5 From TAT To The RAT N, S 6
- f. Nuclear Instrumentation / Place An Excore Nuclear Instrumentation D, S 7
Channel Out of Service.
- g. Component Cooling Water System / Shift Component Cooling Water Pumps M, A, S 8
(Loss of CC).
B.2 Facility Walk-Through
- a. Chemical and Volume Control System / Perform Actions Necessary For Control Room Evacuation - Establish Letdown (performed from dedicated D, A, L 2
shutdown panel!.
- b. Main Steam System / Locally Operate the S/G PORV.
D, L, A, R (normal mode 4
used on last NRC exam)
- c. Emergency Diesel Generators/Operate the Diesel Generator (locally).
D, L (alternate mode 6
used on last NRC exam)
- Type Codes: (D)irect from bank, (M)odified from bank, (N)ew, (A)lternate path, (C)ontrol room, (S)imulator, (L)ow-Power, (R)CA NUREG-1021, Revision 8
ES-301 Control Room Systems and Facility Walk-Through Test Outline Form ES-301-2 Facility: KEWAUNEE NUCLEAR PLANT Date of Examination: 8/26/02 Exam Level (circle one):RO / SRO (I) / SRO(U)
Operating Test No: 2002301 B. I Control Room Systems Type Safety System / JPM Title Cype futi Code*
Function
- a. Control Rod Drive System / Perform Actions To Stop A Continuous Rod M, A, S Withdrawal.
- b. Emergency Core Cooling System / With A Fire In A Dedicated Zone -
D, S, L 2
Restore RCS Inventory using the SI System.
- c. Reactor Coolant Pump System / Start a Reactor Coolant Pump.
D, S, L 4
- d. Containment Spray System / Secure Containment Spray Pumps.
N, SL 5
(ESF)
- e. A.C. Electrical Distribution / Shift Bus 5 From TAT To The RAT N, S 6
- f. Nuclear Instrumentation / Place An Excore Nuclear Instrumentation D, S 7
Channel Out of Service.
- g. Component Cooling Water System / Shift Component Cooling Water Pumps 8
CmoetCoig
- atrM, A,
8 (Loss of CC).
B.2 Facility Walk-Through
- a. Chemical and Volume Control System / Perform Actions Necessary For Control Room Evacuation - Establish Letdown (performed from dedicated D, A, L 2
shutdown panel).
- b. Main Steam System / Locally Operate the S/G PORV.
D, L, A, R (normal mode 4
used on last NRC exam)
- c. Emergency Diesel Generators /Operate the Diesel Generator (locally).
D, L (alternate mode 6
used on last NRC exam)
- Type Codes: (D)irect from bank, (M)odified from bank, (N)ew, (A)lternate path, (C)ontrol room, (S)imulator, (L)ow-Power, (R)CA NUREG-1021, Revision 8
ES-301 Control Room Systems and Facility Walk-Through Test Outline Form ES-301-2
[
Facility: KEWAUNEE NUCLEAR PLANT Date of Examination: 8/26/02 Exam Level (circle one):RO / SRO (I) / SRO(U)
Operating Test No: 2002301 l
B. I Control Room Systems System / JPM Title Type Safety Code*
Function
- b. Emergency Core Cooling System / With A Fire In A Dedicated Zone -
D, S, L 2
Restore RCS Inventory using the SI System.
- d. Containment Spray System / Secure Containment Spray Pumps.
N, S, L 5
(ESF)
- g. Component Cooling Water System / Shift Component Cooling Water Pumps M, A, S 8
(Loss of CC).
B.2 Facility Walk-Through D, A, L, R
- b. Main Steam System / Locally Operate the S/G PORV.
(normal mode 4
used on last NRC exam)
D, L
- c. Emergency Diesel Generators/Operate the Diesel Generator (locally).
(alternate mode 6
used on last NRC exam)
- Type Codes: (D)irect from bank, (M)odified from bank, (N)ew, (A)lternate path, (C)ontrol room, (S)imulator, (L)ow-Power, (R)CA NUREG-1021, Revision 8
Appendix D Scenario Outline Form ES-D-1 It Facility: KNPP Examiners:
Scenario No.: I Operators:
OP-Test No.: 2002301 Initial Conditions: The plant is at 50% power, MOC, equilibrium xenon conditions. Power was reduced two days ago for repairs to the 'A' Main Feedwater Pump. The pump has been repaired and the plant is ready to raise power to 100%. Today is Sunday, present clock time is real time. A normal shift complement is available.
Turnover: The following equipment is inoperable and has been properly removed from service:
'A' Residual Heat Removal Pump - tagged out 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> ago due to high bearing temperatures and vibration during a scheduled surveillance run. Maintenance is on-site and a crew is working to repair the pump.
'B' Motor Drive Auxiliary Feedwater Pump - tagged out last shift due to indications of a significant amount of water in the lube oil. A lube oil cooler leak is suspected. Maintenance has been notified.
LT-472 'B' S/G Water Level Channel - tagged out due to suspected transmitter failure. A plan for transmitter replacement is being developed. A-MI-87 has been completed to remove this channel from service.
The goal for the shift is begin the power ramp to 1 00 per N-0-03.
Event Maf.
Event Event No.
No.
Type*
Description I
R - RO Perform a power increase per N-0-03.
I - BOP LT-461 'A' S/G Water level channel fails low (controlling channel).
_SRO 3
C - BOP Trip of running CW pump.
SRO C - ALL RCS leak develops on 'A' RCS loop requiring reactor trip.
5M
- ALL RCS leak increases to a large break LOCA following reactor trip.
C - RO RHR Pump 'B' fails to auto-start.
SRO 7
C - RO Sump B suction valve SI-35 I B will not open firom control room, must be opened SRO locally.
(Note: Time compression required to accelerate lowering level in RWST) tN)ormal, (K)eactuvity, (I)nstrument, (C)omponent, (M)ajor NUREG-1021, Revision 8, Supplement I
Appendix D Scenario Outline Form ES-D-]
I Facility: KNPP Examiners:
Scenario No.: 2 Operators:
OP-Test No.: 2002301 Initial Conditions: The plant is at 100% power, MOC, equilibrium xenon conditions. Testing of the turbine stop and lgovernor valves is scheduled for later in the shift. Today is Sunday, present clock time is real time. A normal shift complement is available.
Turnover: The following equipment is inoperable and has been properly removed from service:
'A' Residual Heat Removal Pump - tagged out 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> ago due to high bearing temperatures and vibration during a scheduled surveillance run. A maintenance crew is on-site and working to repair the pump.
'B' Motor Drive Auxiliary Feedwater Pump - tagged out last shift due to indications of a significant amount of water in the lube oil. A lube oil cooler leak is suspected. Maintenance has been notified.
LT-472 'B' S/G Water Level Channel - tagged out due to suspected transmitter failure. A plan for transmitter replacement is being developed. A-MI-87 has been completed to remove this channel from service.
The goal for the shift is reduce power to < 390 Mwe per N-0-03 for testing of the turbine stop and governor valves CsP-54-086)
Event Malf.
Event Event No.
No.
Type*
Description R - RO Perform a power reduction per N-0-03 N - BOP SRO 2
1 - RO Controlling pressurizer pressure channel PT-43 I fails high.
SRO 3
I - BOP Generator Hydrogen temperature controller drifts shut.
l SRO C - RO S/G 'A' tube leak develops leading to reactor trip.
SRO 5
C - BOP S/G 'A' blowdown fails to isolate.
SRO 6
M - ALL Tube leak increases to rupture following reactor trip.
7 C - BOP Main turbine fails to auto-trip.
SRO l________
(N)ormai, (K)eactivity, (I)nstrument, (C)omponent, (M)ajor NUREG-1021, Revision 8, Supplement I
F -I I Appendix D Scenario Outline Form ES-D-l Facility: KNPP Examiners:
Scenario No.: 3 Operators:
OP-Test No.: 2002301 Initial Conditions: The plant is at 20% power, BOC, with a startup in progress. The plant tripped from 100% power 5 days ago due to failure of the 'A' Main Feedwater Regulating Valve. Today is Sunday, present clock time is real time.
A normal shift complement is available.
Turnover: The following equipment is inoperable and has been properly removed from service:
'A' Residual Heat Removal Pump - tagged out 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> ago due to high bearing temperatures and vibration during a scheduled surveillance run. A maintenance crew is on-site and working to repair the pump.
'B' Motor Drive Auxiliary Feedwater Pump - tagged out last shift due to indications of a significant amount of water in the lube oil. A lube oil cooler leak is suspected. Maintenance has been notified.
LT-472 'B' S/G Water Level Channel - tagged out due to suspected transmitter failure. A plan for transmitter replacement is being developed. A-MI-87 has been completed to remove this channel from service.
The goal for the shift is to continue with plant startup per N-0-02 at step 4.34 (SP-54-064 is NOT required).
Event l Maf.
Event Event No.
No.
Type* ]
Description R - RO Perform a power increase per N-0-02.
I - BOP Steam Generator 'A' Pressure transmitter PT-468 fails low.
l SRO 3
1 - RO VCT level transmitter LT-141 fails high (divert).
SRO I - RO S/G 'B' level transmitter LT-473 fails low (auto trip should occur, but does not).
SRO 5
C - ALL Both Reactor Trip Breakers fail to open.
M - ALL S/G 'B' fault occurs inside containment.
7 C - BOP
'A' Motor Driven Auxiliary Feedwater Pump fails to auto-start.
SRO I
(IN)ormaj, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor NUREG-1021, Revision 8, Supplement I
ES-401 PWR RO Examination Outline Form ES-401-4 Facility: Kewaunee Nuclear Plant Date of Exam: 8/26/02 Exam Level: RO K/A Category Points Tier Group K
K K
K K
K A
A G
Toinalt 1
2 3
4 5
6 1
2 I.
1 3
3 5
I 3
1 o 1
1 6 Emergency &
'I*i~______
Abnormal 2
4 4
3
- goid 4
l
.I..,..
1 17 Pat3 0
1 1
0
~
F"~~
0 3
Evolutions Tier
~li' Totl 7
8 8
8 3
i:
0 1i2 36 Totals.,, i:.X:_
X,,..=
1 2
2 3
2 2
2 2
2 2
2 2
23
- 2.
2 2
2 2
3 2
2 1
2 2
1 1
20 Plant Systems 3
1 1
1 0
0 1
1 0
2 1
0 8
Toitearls 5
5 6
5 4
5 4
4 6
4 3
51
- 3. Generic Knowledge and Abilities Cat I { Cat2 Cat3 Cat4 l
13 3
1 4
2 4
Note:
- 1.
Ensure that at least two topics from every K/A category are sampled within each tier (i.e., the "Tier Totals" in each K/A category shall not be less than two).
- 2.
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 exam must total 100 points.
- 3.
Select topics from many systems; avoid selecting more than two or three K/A topics from a given system unless they relate to plant-specific priorities.
- 4.
Systems/evolutions within each group are identified on the associated outline.
- 5.
The shaded areas are not applicable to the category/tier.
6.*
The generic K/As in Tiers I and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system.
- 7.
On the following pages, enter the K/A numbers, a brief description of each topic, the topics' importance ratings for the RO license level, and the point totals for each system and category. K/As below 2.5 should be justified on the basis of plant-specific priorities. Enter the tier totals for each category in the table above.
NUREG-1021, Revision 8, Supplement I
ES-401 PWR RO Examination Outline Form ES-401-4 Emergency and Abnormal Plant Evolutions - Tier 1/Group I E/APE 4/Name / Safety Function K l K2 K3 Al A2 C
K/A Topic(s)
FImp.
Points 005.AKI.06 Knowledge of thc operational implications of the 000005 Inoperable/Stuck Control Rod / I x
bases for a power limit. for rod misalignment, as they apply to 2.9 1
the inoperable/stuck control rod.
0000 15/17 RCP Malfunctions / 04 Topic not randomly selected.
N/A N/A x
W/E09.EK2.2 Knowledge of the interrelations between Natural 3.6 I
Circulation Operations and the f'acility's heat removal systems.
including primary coolant, emergency coolant, the decay heat removal systems. and relations between the proper operation of these systems to the operation of the facility.
BW/E09: CE/A13: W/E09&EIO Natural Circ. /4 X
W/E09.EK3.4 Knowledge of the reasons for RO or SRO 3.4 I
function within the control room team. as appropriate to the assigned position. in such a way that procedures are adhered to and the limitations in the facilities license and amendments are not violated as applied to Natural CirCUla1tion Operations.
000024 E.mergency Boration / I X
024.AAI.04 Ability to operate and/or monitor the manual 3.6 1
boration valve as applied to an emergency boration.3.
000026 Loss of Component Cooling Water / 8 Topic not randomly selected.
N/A N/A X
027.AK2.03 Knowledge oflthe interrelations between the P/.R 2.6 Pressure Control System malfunctions and controllers and 000027 Pressurizer Pressure Control System positioners.
Malfunction / 3 X
I 027.AK3.03 Knowledge of the reasons for actions contained in 3.7 EOI' for lIZR Pressure Control System malfunctions.
X 040.AAI.13 Ability to operate and/or monitor Steam [.me 4.2 I
isolation valve indications as they apply to Steam Line Rupture.
000040 (BW/E05: CE/E05; W/E12) Steam Line Rupture - EIxcessive I leat Transfer / 4 x
W/E12.EKI.2 Knowledge of the operational implications of 3.5 1
normal, abnormal, and emergency procedures associated with an uncontrolled depressurization oflall steam generators.
W/E08.EKI.2 Knowledge of the operational implications of' CE/A I l W/E08 RCS Overcooling - PTS / 4 X
normal, abnormal. and emergency procedures associated with 3.4
]
_I'P S.
ES-401 IWR RO Examination Outline Form ES-401-4 Emergency and Abnormal Plant Evolutions- '[Tier 1/Group I (CON'INElI..I))
E/APE K
/ Name / Safety Function KI K2 K3 Al A2 G
K/A Topic(s)
[ Imp.
IPoints X
051.AK3.01 Knowledge of the reasons for loss of steam dump 2.8 capability upon loss of condenser vacuum.
000051 Loss of Condenser Vacuum / 4 051.AAI.04 Ability to operate and/or monitor rod position as 2.5 1
X applied to a loss of condenser vacuum.
055.EK3.01 Knowledge of the reasons for the length ofttime for 000055 Station Blackout / 6 X
.2.7 which battery capacity is designed as applied to a S13.
057.AA2.04 Ability to determuine/interpret the ESI system panel 000057 Loss of Vital AC Elec. Inst. Bus / 6 X
alarm annunciators and channel status indicators as they apply to 3.7 a l.oss of AC vital electrical instrument bus.
000062 Loss of Nuclear Service Water / 4 Topic not randomly selected.
N/A N/A 000067 Plant Fire On-site / 9 Topic not randomly selected.
N/A N/A 000068 (BW/A06) Control Room Evacuation /8 Topic not randomly selected.
N/A N/A 00009 (WE14 Los ofTMT nterity/
~069
-2.4.2 Knowledge of systemn setpoints. interlocks, and3.
000069 (W/E14) Loss of CTMTr Integrity / 5 X
automatic actions associated with EOP entry conditions.
3 9 X
074.EK2.08 Knowledge of the interrelations between Inadequate 2.5 Core Cooling and sensors and detectors.
000074 (W/E06&E07) Inadequate Core Cooling / 4 X
W/E06.EK3.4 Knowledge of the reasons for RO or SRO 3.5 I
function within the control room team as appropriate to the assigned position, in such a way that procedures are adhered to and the limitations in the 'acilities license and amendments are not violated as applied to Degraded Core Cooling.
BW/E03 Inadequate Subcooling Margin / 4 Suppressed - Not applicable to facility N/A N/A 000076 1-ligh Rcactor Coolant Activity / 9
'Iopic not randomly selected.
N/A N/A 13W/A02&A03 Loss of NNI-X/Y / 7 Supprcssed - Not applicable to facility N/A N/A K/A Category Totals:
3 3
5 3
1 I
Group Point Total:
16
lS-4()
PWR RO Examination Outline Form lS-401-4 Emerg-ency and Abnormal Plant Evolutions - Tier l/Group 2 FI/APE # / Name / Safety Function K I K2 K3 A I A2 G
K/A 'I'opic(s)
[ imp.
Points 000001 Continuous Rod Withdrawal / I X
OO.AAl.01 Ability to operate and/or monitor the Bank Select 3.5 1
Switch as applied to a Continuous Rod Withdrawal.
003.AK1.04 Knowledge of the operational implications of the 000003 Dropped Control Rod / I X
effects of power level and control position on flux as applied to a 3.1 I
Dropped Control Rod.
000007 (BW/E02&E 10: CE/E02) Reactor Trip -
X 007.EAI.05 Ability to operate and monitor Nuclear 4.0 I
Stabilization - Recoverv / I Instrumentation as it applies to a reactor trip.
BW/AOI Plant Runback / I Suppressed - Not applicable to facility N/A N/A BW/A04 Turbine Trip / 4 Suppressed - Not applicable to facility N/A N/A 000008 Pressurizer Vapor Space Accident /3 Topic not randomly selected.
N/A N/A 000009 Small Break LOCA / 3 Topic not randomly selected.
N/A N/A 000091 Large Break LOCA / 3 Topic not randomly selected.
N/A N/A W/E04.EK1.2 Knowledge of the operational implications of W/E04 LOCA Outside Containment / 3 X
normal, abnormal. and emergency operating procedures 3.5 1
associated with a L.OCA outside containment.
X W/E03.EK2.2 Knowledge of the interrelations between the 3.7 i
LOCA Cooldown and Depressuriiation and the facility's heat removal systems. including primary coolant, emergency coolant.
the decay heat removal systems. and relations between the proper BW/EO8: W/E03 LOCA Cooldown/Depress. /4 operation of these systems to the operation of the facility.
X W/E03.EK3.2 Knowledge of the reasons tor normal, abnormal.
3.4 1
and emergency operating procedures associated with a LOCA Cooldown and Depressurization.
X W/EI I.EK2.1 Knowledge of the interrelations between the Loss 3.6 1
of Emergency Coolant Recirculation and components. and functions of control and safety systems, including instrumentation. signals. interlocks. failure modes, and automatic W/f. I I Loss of Emergency Coolant Recirc. / 4 and manual features.
X W/E1 1.EA1.3 Ability to operate and/or monitor the desired 3.7 operating results dulring abnormal and emergency situations as applied to a Loss of Emergency Coolant Recirculation.
ES-401 PWR RO Examination Outline Form ES-401-4 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 I
-(CONTINUED) l/APE # / Name / Safety Function KI K2 K3 Al A2 G
K/A Topic(s)
Imp. I Points W/EO.EK2.2 Knowledoe of the interrelations between the (Reactor Trip or Safety In jection / Re-diagnosis) and the lW/E01 & E02 Rediagnosis & SI Termination / 3 flacility's heat removal systems. incluhding primary coolant.
emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.
000022 Loss of Reactor Coolant Makeup / 2 Topic not randomly selected.
N/A N/A 000025 Loss of RI-IR System / 4 Topic not randomly selected.
N/A N/A X
029.EK2.06 Knowledge of the interrelations between an ATWS 2.9 and breakers, relays, and disconnects.
000029 Anticipated 'c'ransient w/o Scram / I X
029.EA2.07 Ability to determine or interpret the reactor trip 4.2 1
breaker indicating lights as applied to an ATWS.
000032 Loss of Source Range NI / 7 Topic not randomly selected.
N/A N/A 2.4.49 Ability to perform without reference to procedures those 000033 Loss of Intermediate Range NI /7 X
actions that require immediate operation of system components 4.0 1
and controls.
000037 Steam Generator Tube Leak / 3 X
037.AK3.04 Knowledge of the reasons for use of the "feed and 2.5 bleed"' process as they apply to a S/G Tube Leak.
,000038 Steam Generator 'ube Rupture / 3 038.EA1.44 Ability to operate and monitor the level operating limits for S/Gs as applied to a Si'R.
.4 054.AKI.01 Knowledge of the operational implications of a 000054 (CE/E06) Loss of Main Feedwater / 4 X
MFW line break depressurizing the S/G as related to a loss of 4.1 I
main feedwater.
BW/E04~ W/E05 Inadequate Heat 'Iransfer - Loss Topic not randomly selected.
N/A N/A of Secondary Heat Sink /4 Topi not randomly selected.
N/A N/A 4
10)0058 Loss of D)C Power / 6 Topic not randomly selected.
N/A N/A 000059 Accidental Liquid RadWaste Rel. / 9
_opic not randomly selected.
N/A N/A 000060 Accidental Gaseous Radwaste Rel. / 9
'ITopie not randomly selected.
N/A N/A 000061 ARM System Alarms / 7 T
Topie not randomly selected.
N/A N/A
ES-401 PWR R() Examination Outline Form 1S-4()1-4 Emergency and Abnormal Plant Evolutions - Tier I/Group 2 I
-(CONTIINUED)
E/APE # / Name / Safety Function KI K2 K3 Al A2 G
K/A Topic(s)
Imp.
Points X
W/16.EK1.3 Knowledge of'the operational implications of' 3.0 annunlciators and conditions indicating signals. and remedial actions associated with High Containment Radiation.
W/E 16 High Containment Radiation / 9 X
W/16.EK3.3 Knowledge of the reasons for manipulation of 3.0 controls required to obtain the desired operating results during abnormal and emergency situations.
CE/E09 Functional Recovery
=
Suppressed - Not applicable to facility N/A N/A K/A Category Totals:
l 4 l 4 3 l 4 l I GI roup Point Total:
N/A 17
ES-401 I'WR R() Ixamination Outline For ES-401-4 Emergency and Abnormal Plant Evolutions - Tier I/Group 3 F`/API 4 / Name / Safety Function KI K2 J K3 l Al A2 ] G K/A Topic(s)
Im Points X
028.AK2.02 Knowledge oftthe interrelations between Pressurizer 2.6 1
Level Control Malfunctions and sensors and detectors.
000028 Pressurizer Level Malfunction / 2 X
028.AA2.11 Ability to determine and interpret a leak in the PZR 3.2 as applied to a pressurizer level control malfunction 000036 (13W/A08) Fuel I landling Accident / 8 Topic not randomly selected.
N/A N/A (1()()()56 Loss of Of'f-site Pover / 6 X056.AA 1.04 Ability to operate and/or monitor the addjustment of' 3.
speed of EDG to maintain fiequency and voltage levels.
000(065 Loss of Instrument Air / 8 Topic not randomly selected.
N/A N/A 13W/E.13&14 EOP Rules and Enclosures Suppressed - Not applieable to taeitity N/A N/A BW/A05 Emergency Diesel Actuation / 6 Suppressed - Not applicable to facility N/A N/A BW/A07 Flooding / 8 Suppressed - Not applicable to facility N/A N/A Cl/A16 Excess RCS Leakage / 2 Suppressed -Not applicable to facility N/A N/A W/E 1 3 Steam Generator Over-pressure / 4 Topic not randomly selected.
N/A N/A W/E 15 Containment Flooding / 5
'I'opic not randomly selected.
N/A N/A K/A Category Iotals:
l I
0 1
l 0
Group Point Total:
3 3
ES-401 PWR RO E~xamination Outline lor ES-401-4 Plant Systems - Tier 2/Group I _
I/API 4 / Name / Safety Function l K1 l K2 l K3 l K4 l K5 K6 l Al A2 A3 A4 l (
K/A lopic(s)
Imp.
Points X
OOI.K5.04 Knowledge of rod insertion limits 4.3 as they apply to the Control Rod Drive System.
001 Control Rod Drivc X
OOI.A4.15 Ability to manually operate 3.1 and/or monitor in the control room. stopping boration/dilotion or other means of reactivity change. while adjusting either rod position or
'rave.
X 003.K1.01 Knowledge of the physical 2.6 I
connections and/or cause-cffect relationships between the RCPS and RCP lube oil.
003 Reactor Coolant Pump x
003.K6.04 Knowledge of the effect of a loss 2.8 1
or malfunction of the containment isolation valves affecting RCP operation will have on the RCPS.
X 004.K4.01 Knowledge of CVCS design 2.8 feature(s) and/or interlock(s) which provide for oxygen control in the RCS.
0(04 Chemical and Volume Control X
004.A4.19 Ability to manually operate 3.1 and/or monitor in the control room the CVCS letdown orifice isolation valve and valve control switches.
X 013.K3.02 Knowledge oflthe effect that a 4.3 loss or malfunction of the ESFAS will have 013 Engineering Safety Features Actuation on the RCS.
X 013.K2.01 Knowledge of bus power supplies 3.6 I
to ESFAS/safeguards equipment control.
X 015.K2.01 Knowledge of bus power supplies 3.3 to NIS channels, components. and intercotmections.
015 Nuclear Instrumentation X
015.KI.03 Knowledge oflthe physical 3.1 connections and/or cause-effect relationships hetween the NIS and CRDS.
ES-401 PWR RO Examination Outline Form ES-401-4 Plant Systems - Tier 2/Group I (CONTINIJED))
IF/API
- / Name / Safety Function KI K2 K3 K4 K5 K6 Al A2 A3 A4 G
K/A Topic(s)
Imp.
Points X
017.A2.02 Ability to predict the impacts of 3.6 1
core damage on the ITM system. and based on those predictions. use procedures to correct. control, or mitigate the 017 In-core Temperature Monitor consequences.
X 017.K5.02 Knowledge ofthe operational 3.7 l
implications of saturation and subcooling of water as applied to the ITM.
X 022.K3.01 Knowledge ofthe effect that a 2.9 l
loss or malfunction of the Containment Cooling System will have on containment equipment subiect to damage by high or low 022 Containment Cooling temperature. humidity. and pressure.
X 2.1.23 Ability to perform specific system and 3.9 integrated plant procedures during all modes of plant operation.
(025 Ice Condenser S-ppressed Not applicable to facility N/A N/A 056.A2.04 Ability to predict the impacts of a loss of condensate pumps on the condensate 056 Condensate X
system. and based on those predictions. usc 2.6 procedures to correct, control, or mitigate the consequences.
059.A 1.03 Ability to predict and/or monitor changes in parameters (to prevent exceeding 059 Main Feedwater design limits) associated with operating the 2.7 MaW controls including power level restrictions for operation of MFW pumps and valves.
X 061.K6.02 Knowledge ofthe effect of a loss 2.6 or malfunction of' pumps will have on the AFW components.
061 Auxiliary/Emergency Feedwater X
061.A3.01 Ability to monitor automatic 4.2 I
operation of the AFW. including AFW startup and flows.
lE-401 PWR K() Examination Outline Form ES-401-4 Plant Systems - Tier 2/Group I (CONTINUEI))
E/APIE # / Name / Safety F'unction KI K2 K3 K4 K5 K6 Al A2 A3 A4 G
K/A Topic(s)
Imp.
Points X
068.A3.02 Ability to monitor automatic 3.6 operation of'the liquid Radwaste System including automatic isolation.
068 Liquid Radwaste X
2.1.14 Knowledge of system status criteria 2.5 I
which require the notification of plant personnel.
071.K4.06 Knowledge of design feature(s) 071 Waste Gas Disposal and/or interlock(s) which provide for the 2 7 l
sampling and monitoring of waste gas release tanks.
X 072.A1.01 Ability to predict and/or monitor 3.4 changes in parameters (to prevent exceeding design limits) associated with operating the ARM system Controls including radiation 072 Area Radiation Monitoring levels.
X 072.K3.02 Knowledge of the effect that a 3.1 I loss or malfunction of the ARMV system will have on fuel handling operations.
K/A Category Totals:
2 213 2
2 212 2
J 2 lGrouP Point Total:
23
ES-401 PWR RO Examination Outline For ES-40 1-4 Plant Systems - Tier 2/Group 2 1
/A'PI#/ Name / Safety lunction KI K2 l K3 l K4 l K5 l K6 l Al l A2 l A3 l A4 l G l K/A Topic(s) l Imp. l Points 002.K6.12 Knowledge of the effect of a loss 002 Reactor Coolant X
or malfunction of thc RCS Code Safety 3.0 1
valves.
006 Emergency Core Cooling Iopic not randomly selected.
N/A N/A 010 Pressurizer Pressure Control Topic not randomly selected.
N/A N/A 01 I.K4.04 Kriowledge of PZR LCS design 011 Pressurizer Level Control X
feature(s) and/or interlock(s) which provide 3.()
for PZR level inputs.
012.K1.04 Knowledge of the physical 012 Reactor Protection X
connections and/or cause effect relationships 3.2 between the RPS and RPIS.
014 Rod l'osition Indication Topic not randomly selected.
N/A N/A X
016.K5.01 Knowledge of the operational 2.7 1
implication of the separation of control and protection circuits.
016 Non-Nuclear Instrumentation X
016.K3.04 Knowledge of the effect that a 2.6 loss or malfunction of the NNIS will have on the MFW system.
X 026.K1.01 Knowledge ofthe physical 4.2 connections and/or cause effect relationships between the CSS and ECCS.
026 Containment Spray X
026.A2.04 Ability to predict the impacts of a 3.9 I
failure of a spray pump on the CSS, and based on those predictions. use procedures to correct. control, or mitigate the consequences.
029 Containment Purge Topic not randomly selected.
N/A N/A 033 Spent Fuel Pool Cooling
_opic not randomly selected.
N/A N/A
IS-401 PWR RO Examination Outline Form ES-401-4 Plant Systems - I[ier 2/Group 2 (CONTINUED)
E/AIE# /Name / Safety Function lKI lK2 K3 K4 K5 lK6 Al A2 A3 A4 G
lKATopic(s)
[
Imp.
Points X
035.K6.01 Knowledge of the effect of a loss 3.2 I
or malfunction of the MSIVs will have on the S/G system.
035 Steam Generator X
035.A.01 Abilitv to predict and/or monitor 3.6 changes in parameters (to prevent exceeding design limits) associated with operating the S/G system controls including S/G wide and narrow range Ievel during startup. shutdown.
and normal operations.
039.A4.04 Ability to manually operate 039 Main and Reheat Steam X
and/or monitor in the control room 3.8 emergency feedwater pump turhines.
055.A3.03 Ability to monitor automatic 055 Condenser Air Removal X
operation of the CARS. including automatic 2.5 1
diversion of the CARS exhaust.
X 062.K3.03 Knowledge of the eflect of a loss 3.7 or malfunction of the AC Distribution system will have on the DC sytem.tlbtil sse 062 AC Electrical Distribution X
062.K2.01 Knowledge of bus power supplies 3.3 to major system loads.
X 063.K4.02 Knowled-e of l)C electrical 2.9 system (lesign feature(s) and/or interlock(s) which provide for breaker interlocks.
permissives, bypasses. and cross-ties.
063 DC Electrical Distribution X
063.A3.01 Ability to monitor automatic 2.7 operation of the DC electrical system including meters, annunciators, dials, recorders. and indicating lights.
064 Emergency l)iesel Generator X
064.o2.01 Knowledge ofbusL power Supplies 2.7
__ _to the E'D G air com pressor.
I
_ __I_
l ES-4()1 IWR O lxamination Outline Form ES-401-4 Plant Systcms - Tier 2/Group 2 (CONTINUIED)
I/API # / Name / Safety Function Kl K2 K3 K4 K5 K6 Al A2 A3 A4 G Kf A'I'opic(s)
Imp.
Points 073.K4.0I Knowledge of PRM system 073 Process Radiation Monitoring X
design feature(s) and/or interlock(s) which 4.0 1
provide for release termination when radiation exceeds setpoint.
075.A2.02 Ability to predict the impacts of a loss of circulating water pumps on the CW 075 Circulating Water X
system, and based on those predictions. use 2.5 I
procedures to correct, control, or mitigate the consequences.
079 Station Air
'opic not randomly selected.
N/A N/A X
2.4.48 Ability to interpret control room 3.5 indications to verify the status and operation of system, and understand how operator actions and directives affect plant and system 086 Fire Protection conditions.
X 086.K5.03 Knowledge of the operational 3.1 implication of the effect of water spray on electrical components as applied to the Fire Protection System.
[K/A Category Totals:
{ 2 212 f3 J2 i2 lr2i 2
l1lIroGroupnt nt'Iotal:
-20
ES-401 I'WR RO Examination Outline For ES-401-4 Plant Systems -- Tier 2/Group 3 Imp.
Points E/APE # / Name / Safety Function K I K2 K3 K4 KS K6 Al A2 A3 A4 G
KJA Topic(s)
Imp.
Poits 005.K6.03 Knowledge of the effect of a loss 005 Residual H1eat Removal X
or malfunction of the RI IR heat exchanger 2.5 I
will have on the RIIR system.
007 Pressurizer RelielYQuench Tank Topic not randomly selected.
N/A N/A 008.A4.10 Ability to manually operate 008 Component Cooling Water X
and/or monitor in the control room 3.1 I
conditions that require the operation of two CCW coolers.
027 Containment Iodine Removal Suppressed - Not applicable to facility N/A N/A 028 Hydrogen Recombiner and lPurge Control Topic not randomly selected.
N/A N/A 034.A3.0 1 Ability to monitor automatic 034 Fuel Handling Equipment X
operation of the fuel handling system.
2.5 1
including travel limits.
041 Steam Dump/Turbine Bypass Control Topic not randomly selected.
N/A N/A 045.K3.01 Knowledge of the eflect that a 045 Main Turbine Generator X
loss or malfunction of the main turbine 2.9 9 generator will have on the remainder of the plant.
076 Service Water T'opic not randomly selected.
N/A N/A X
078.K2.01 Knowledge ol'bus power supplies 2.7 to the instrument air compressors.
078 Instrument Air X
078.A3.01 Ability to monitor automatic 3.1 operation of the instrument air system,
_including air pressure.
ES-401 PWR RO Examination Outline For ES-401-4 Plant Systems - [icr 2/Group 3 (CONTINUED)
X 103.KI.01 Knowledge of the physical 3.6 connections and/or cause effect relationships between the containment system and the containment cooling system.
103 Containment X
103.AI.0l Ability to predict and/or monitor 3.7 changes in parameters (to prevent exceedin.n design limits) associated with opcr-atin-g the containment systemn controls including containment pressure. temperature. and umiity 1KACaeory Totals:
JI l I O1 OII 1I0l2I I
0 GrO LIu p oint Totl-
Plant-Specific Priorities System / Topic Recommended Replacement for...
Reason Points NONE I'lant-Spceific Priority lotal: (limit 10)
ES-401 Generic Knowledge and Abilities Outline (Tier 3)
ES-401-5 Facility: KNP Date of Exam: 8/26/02 Exam Level: RO l
Category K/A #
Topic Imp.
Points 2.1.9 Ability to direct personnel activities inside the control room.
2.5 I
Conduct of 2.1.10 Knowledge of conditions and limitations in the facility license.
2.7 I
Operations 2.1.22 Ability to determine Mode of Operation 2.8 1
Total 3
Ability to perform pre-startup procedures for the facility, including 2.2.1 operating those controls associated with plant equipment that could 3.7 I
affect reactivity.
Equipment 2.2.12 Knowledge of surveillance procedures.
3.0 l
Control 2.2.13 Knowledge of tagging and clearance procedures.
3.6 2.2.27 Knowledge of the refueling process.
2.6 l
Total 4
2.3.1 Knowledge of 10 CFR 20 and related facility radiation control 2.6 Radiation requirements.
Control 2.3.9 Knowledge of the process for performing a containment purge.
2.5 I
Total 2
2.4.6 Knowledge of symptom based EOP mitigation strategies.
3 2.4.14 Knowledge of generic guidelines for EOP flowchart use.
3.0I Emergencv Procedures/
2.4.25 Knowledge of fire protection procedures.
2.9 Plan 2.4.49 Ability to perform without reference those actions that require 4.0 I immediate operation of system components and controls.
Total 4
Tier 3 Point Total (RO) 13
ES-401 PWR SRO Examination Outline ES-401-3 Facility: Kewaunee Nuclear Plant Date of Exam: 8/26/02 Exam Level: SRO K'A Category Points l
Point Tier Group K
K K
K K
K A
A A
G Total 1
2 1
2 3
4 5
l I.
1 4
4 4
iJI 5
43 24 Emergency&
IL l
3 1
l 3
16 Abnormal 2
2 3
4 16~i Plant
'll Evolutions 3
0 1
0 0
1 1
3 Tier il' 9i:
6 8
8 ir 0
8 6
l 7
43 Totals
~i 1
2 1
2 2
1 2
2 2
2 1
2 19
- 2.
2 2
1 1
2 2
2 2
1 2
1 1
17 Plant 30 1
0 0
0 1
0 0
1 0
1 4
Tier Totals 4
3 4
3 5
4 3
5 2
4 40
- 3. Generic Knowledge and Abilities Cat I Cat2 Cat 3 Cat4 17 4
4 4
5 t Note:
I.
Ensure that at least two topics from every K/A category are sampled within each tier (i.e., the "Tier Totals" in each K/A category shall not be less than two).
- 2.
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 +/- 1 firom that specified in the table based on NRC revisions. The final exam must total 100 points.
- 3.
Select topics from many systems; avoid selecting more than two or three K/A topics firom a given system unless they relate to plant-specific priorities.
- 4.
Systems/evolutions within each group are identified on the associated outline.
- 5.
The shaded areas are not applicable to the category/tier.
6.*
The generic K/As in Tiers I and 2 shall be selected firom Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system.
- 7.
On the following pages, enter the K/A numbers, a brief description of each topic, the topics' importance ratings for the SRO license level, and the point totals for each system and category. K/As below 2.5 should be justified on the basis of plant-specific priorities. Enter the tier totals for each category in the table above.
NUREG-1021, Revision 8, Supplement I
ES-401 I'WR SR() Ixamination Outline Form ES-401-3 Emergency and Abnormal Plant Evolutions-lTier l/Group I lI/APE # / Name / Safcty Function KI K2 K3 Al I A2 G
K/A 'opic(s)
Poinp.
Poits 000001 Continuous Rod Withdrawal /1 X
001.AAI.01 Ability to operate and/or monitor the Bank Select 3.2 Switch as applied to a Continuous Rod Withdrawal.
003.AKI.04 Knowledge of the operational implications of the 000003 Dropped Control Rod / I X
effects of power level and control position on flux as applied to a 3.7 Dropped Control Rod.
005.AA2.03 Ability to determine and interpret the required 000005 Inoperable/Stuck Control Rod / I X
actions if more than one rod is stuck or inoperable as applied to 4.4 the Inoperable/Stuck Control Rod.
))001 I Large Break LOCA / 3 Topic not randomly selected.
N/A N/A W/E04.EKI.2 Knowledge of the operational implications of W/E04 I.OCA Outside Containment / 3 X
normal, abnormal, and emergency operating procedures 4.2
_associated with a LOCA outside containment.
X E01 - 2.2.22 Knowledge of Ilimiting conditions for operations 4.1 and safety limits.
X W/EO0.EK2.2 Knowledge of the interrelations between the 3.8 W/EO1 & E02 Rc-diagnosis & SI Termination /3 (Reactor Irip or Sal'ety Injection / Rc-diagnosis) and the facility's heat removal systems, including primary coolant.
emergency coolant, the decay heat removal systems. and relations between the proper operation of these systems to the opcration of the facility.
X 015/017.AA2.08 Ability to determine and interpret when to 3.5 secure RCPs on high bearing temperature as applied to a RCI' 000015/17 RCP Malfunctions / 4 malfunction.
X 2.1.12 Ability to apply Technical Specifications for a system.
4.0 X
W/E09.EK2.2 Knowledgc of the interrelations between Natural 3.9 Circulation Operations and the facility's heat removal systems.
including primary coolant. emergency coolant, the decay heat removal systems. and relations between the proper operation of these systems to the operation of the facility.
BW/E09: CE/A13: W/E09 & E10 Natural Circ. /4 X
W/E09.EK3.4 Knowledge of the reasons for RO or SRO 3.6 function within the control room team as appropriate to the assigned position. in such a way that procedures are adhered to and the limitations in the facilities license and amendments are not violated as applied to Natural Circulation Operations.
ES-401 PWR SRO Examination Outline Form ES-401-3 Emergency and Abnormal plant Evolutions - Tier 1/Group I (CONTINUJEI))
E/AIPE 4 / Name / Safety lunction K]
K2 K3 Al A2 G
K/A Topic(s)
Imp. [
'oits 000024 Emergency Boration / I X
024.AAI.04 Ability to operate and/or monitor the manual 3.7 1
boration valve as applied to an emergency boration.
000026 Loss of Component Cooling Water / 8 T'opie not randomly selected.
N/A N/A 000029 Anticipated Transient w/o Scram / I X
029.EK2.06 Knowledge of the interrelations between an ATWS 3.1 and breakers, relays, and disconnects.
X 040.AA1.13 Ability to operate and/or monitor Steam Line 4.2 1
isolation valve indications as they apply to Steam Line Rupture.
000040 (BW/E(5) CE/E)05 W/1.12) Steam Line Rupture - Excessive Heat Transfer / 4 X
W/E12.EKI.2 Knowledge of the operational implications of 3.8 1
normal, abnormal, and emergency procedures associated with an uncontrolled depressurization of all steam generators.
W/E08.EK1.2 Knowledge of the operational implications of CE/All. W/E08 RCS Overcooline - PTS /4 X
normal, abnormal, and emergency procedures associated with 4.0 I-'S.
X 051.AK3.01 Knowledge of the reasons for loss of steam dump 3.1 capability upon loss oflcondenser vacuum.
000051 Loss of Condenser Vacuum / 4 X
051.AA1.04 Ability to operate and/or monitor rod position as 2.5 applied to a loss of condenser vacuum.
000055 Station Blackout / 6 X
055.EK3.01 Knowledge of the reasons for the length of time for 3.4 1
which battery capacity is designed as applied to a SBO.
000057 Loss of Vital AC Flee. Inst. Bus / 6 Topic not randomly selected.
N/A N/A 00()059 Accidental Liquid RadWaste Release / 9 Topic not randomly selected.
N/A N/A 000062 Loss of Nuclear Service Water / 4
=opic not randomly selected.
N/A N/A 000067 Plant Fire On-site / 9 X
067.AA2.15 Ability to determine and interpret the requirements 3.9 for establishing a lire watch as applied to the plant fire on site.
000068 (BW/A06) Control Room Evacuation / 8 X
068.AA 1.31 Ability to operate and/or monitor the ED/G as they 4.0 apply to the Control Room Evacuation.
000069 (W/E 14) Loss of CTMT Integrity / 5 X
069.AA2.01 Ability to determine and interpret a loss of' 43 containment intCgrity.
4_3
IS-4()
PWR SRO Examination Outline Form ES-401-3 Emergcncy and Abnormal Plant Elvolutions - Tier 1/Group I (CONTINUED)
F/APE# / Name / Safety Function l K I K2 [ K3 Al l A2 l C K/A Topic(s)
] Imp. ]
Poits X
074.EK2.08 Knowledge of the interrelations between Inadequate 2.5 I
Core Cooling and sensors and detectors.
000074 (W/E06&E07) Inadequate Core Cooling /4 X
W/E06.EK3.4 Knowledge of the reasons for RO or SRO 3.7 function within the control room team as appropriate to the assigned position, in such a way that procedures are adhered to and the limitations in the facilities license and amendments are not violated as applied to Degraded Core Cooling 13W/1.(03 Inadequate Subcooling Margin / 4 Suppressed -- Not applicable to iacility N/A N/A 000076 Fligh Reactor Coolant Activity / 9 X
2.2.22 Knowledge of limiting conditions for operations and 4.1 safety limits.
13W/A(2&A(3 Loss of NNI-XY / 7 Suppressed - Not applicable to facility N/A N/A K/A Category Totals:
4 4
4 l 5 4
3 GroupPointI'otal:
24
ES-401 PWR SRO Examination Outline Form ES-401-3 Emergency and Abnormi-al Plant Elvolutions- 'Tier I/Group 2 I/APE 4 / Name / Safety Function l KI l K2 K3[ Al A2 G
K/A Topic(s)
Impf.
Points 000007 (BW/E02&E 10; CE/E02) Reactor Irip -
X 007.EAI.05 Ability to operate and monitor the Nuclear 4.1 I
Stabilization - Recovery / I Instrumentation as they apply to a reactor trip.
BW/AO1 Plant Runback / I Suppressed - Not applicable to facility N/A N/A 13W/A04 'T'urbine Trip / 4 Suppressed - Not applicable to facility N/A N/A 000008 Pressurizer Vapor Space Accident / 3 Topic not randomly selected.
N/A N/A 000009 Small Break l.OCA / 3 X
009.E(3.21 Knowlede of the reasons for actions contained in 4.5 N/A the EOP for small break LOCA/leak-.
W/E03.EK2.2 Knowledge of the interrelations between the L.OCA Cooldown and l)epressurization and the facility's heat BW/E08: W/E03 LOCA Cooldown - Depress. /4 X
removal systems. including primary coolant. emergency coolant.
4.0 1
the decay heat removal systems. and relations betweeni the proper operation oflthese systems to the operation of the facility.
X W/E11.EK2.1 Knowledge of the interrelations between the Loss 3.9 of ELmergency Coolant Recirculation and components. and functions oflcontrol and safety systems. including instrumentation. signals. interlocks. failure modes. and automatic W/E II Loss of Emergency Coolant Recirc. / 4 and manual features.
X W/EI I.EA 1.3 Ability to operate and/or monitor the desired 4.2 I
operating results during abnormal and emergency situations as applied to a Loss of Emergency Coolant Recirculation.
000022 Loss of Reactor Coolant Makeup / 2 Topic not randomly selected.
N/A N/A 0010(25 Loss of RI IR System / 4 Topic not randomly selected.
N/A N/A X
027.AK2.03 Knowledge of the interrelations between the PZR 2.8
]
PressuLe ('ontrol System malfunctions and controllers and 000027 Pressurizer Prcssurc Control System positioners.
Malfunction / 3 X
027.AK3.03 Knowledge of the reasons for actions contained in 4.1 I
_1.()1 for PlR P'ressure Control System malfunctions.
000032 Loss of Source Range NI / 7 Topic not randomly selected.
N/A N/A
ES-401 PWR SRO Examination Outline Form ES-401-3 Emergency and Abnormal Plant Evolutions - Tier I/Group 2 (CONTINUED)
I./AI'I 4 / Name / Safety Function KI K2 K3 Al A2 G
K/A 'opic(s)
Imp.
P'oints X
033.AA2.08 Ability to determine and interpret intermediate 3.4 range channel operability as applied to a loss of intermediate range nuclear instrumentation.
000033 Loss of Intermediate Range NI / 7 X
2.4.49 Ability to perform without reference to procedures those 4.0 I
actions that require immediate operation of'system components and controls.
X 2.1.12 Ability to apply lechnical Specifications for a system.
4.0 00(037 Steam Generator Tube Leak / 3 X
037.AK3.04 Knowledge of the reasons for use of the "feed and 2.9 i
bleed" process as they apply to a S/(i lube Leak.
000038 Steam Generator Tube Rupture / 3 X038.EAI.44 Ability to operate and monitor the level operating 3.4 limits for S/Gs as applied to a SGTR.
054.AKI.01 Knowledge of the operational implications of a 000054 (CE/E06) Loss of Main Feedwater /4 X
MFW line break depressurizing the S/G as related to a loss of 4.3 1
main feedwater.
BW/E04: W/E05 Inadequate Heat Transfer - Loss T
Topic not randomly selected.
N/A N/A of Secondamy I lcat Sink / 4 000058 Loss of D)C l'ower / 6 Topic not randomly selected.
N/A N/A 000060 Accidental Gaseous Radwaste Release. / 9 Topic not randomly selected.
N/A N/A 000061 ARM System Alarms / 7 X
2.1.32 Ability to explain and apply all system limits aid 3.8
_____________ARM
________________Alarms/_______7__precautions.38
ES-401 PWR SRO Examination Outline Form ES-401-3 Emergency and Abnormal plant Evolutions - Tier I/Group 2 (CONTINUED) lI/AI'E # / Name / Safety Function KI K2 K3 Al A2 C K/A Topic(s)
Imp.
Points X
W/16.EKI.3 Knowledge of the operational implications of 3.3 annunciators and conditions indicating signals. and remedial actions associated with I ligh Containment Radiation.
W/E16 High Containment Radiation /9 X
W/16.EK3.3 Knowledge of the reasons for manipulation of 3.0 1
controls required to obtain the desired operating results during abnormal and emergency situations involving high Containment Radiation.
000065 Loss of Instrument Air / 8 Topic not randomly selected.
N/A N/A Cl./1.09 Functional Recovery Suppressed - Not applicable to facility N/A N/A K/A Category Totals:
2 l3 l4 3 3 l I l_ 3 1
l ou irp Point Total:
16
ES-401 I'WR SR() Examination Outline For ES-401-3 Emergency and Abnormal Plant Envolutions -T'ier 1I/Group 3 EI/AI'E # / Name / Safety Function KI K2 K3 Al A2
(,
K/A Topic(s)
Points X
028.AK2.02 Knowledue of the interrelations between Pressurizer 2.7 Level Control Malfunctions and sensors and detectors.
000028 Prcssurizer Level Malfunction / 2 X
028.AA2.11 Ability to determine and interpret a leak in the P/ZR 3.6 as applied to a pressurizer level control malfunction 000036 (BW/A08) Fuel Handling Accident /8 ropic not randomly selected.
N/A N/A 000056 Loss of Off-Site Power / 6
'I'opic not randomly selected.
N/A N/A 13W/F1 3& 14 EOP Rules and Enclosures Suppressed - Not applicable to facility N/A N/A BW/A05 Emergency Diesel Actuation / 6 Suppressed - Not applicable to facility N/A N/A L3W/A07 Flooding / 8 Suppressed - Not applicable to facility N/A N/A CE/A 16 Excess RCS Leakage / 2 Suppressed - Not applicable to facility N/A N/A W/1.3 Steam Generator Over-pressuIe / 4 X
2.4.18 Knowledge ofthc specific bases for EOPs.
3.6 W/E 1 5 Containment Flooding / 5
'Topic not randomly selected.
N/A N/A K/A Category otals:
l 0 1 I l 0 l I
l 1 l Group lPoint'otal:
I
ES-401 PWR SRO Examiniation Outline For ES-401-3 Plant Systems - Tier 2/Group I l
E/APE 4 / Name / Safety Function KlI K2 K3 K4 K5 K6 Al A2 A3 A4 G
K/A
'opic(s)
Imp. IPoints X
001.K5.04 Knowledge of rod insertion limits 4.7 as they apply to the Control Rod Drive System.
001 Control Rod Drive X
OO1.A4.15 Ability to manually operate 3.1 I
and/or monitor in the control room stopping boration/dilution or other means of reactivity change while adjusting either rod position or Tave.
X 003.KI.01 Knowledge ofthe physical 2.8 connections and/or cause-effect relationships between the RCPS and RCP' lube oil.
003 Reactor Coolant Pump X
003.K6.04 Knowledge of the effect of a loss 3.1 or malfunction of the containment isolation valves affecting RCP operation will have on the RCPIS.
004.K4.01 Knowledge ofCVCS design 004 Chemical and Volume Control X
feature(s) and/or interlock(s) which provide 3.3 1
for oxygen control in the RCS.
013.K3.02 Knowledge of'the ef1ect that a 013 Engineering Safety Features Actuation X
loss or malfunction of the F.SFAS will have 4.5 on the RCS.
014 Rod Position Indication Topic not randomly selected.
N/A N/A X
2.4.11 Knowledge of abnormal condition 3.6 1
procedures.
015 Nuclear Instrumentation x
015.K2.01 Knowledge of bus power supplies 3.7 to NIS channels, components. and interconnections
ES-401 IPWR SRO Examination Outline Form ES-401-3 Plant Systems - Tier 2/Group I (CONTINUJEI))
E/APE #/ Name / Safety Function K] [ K2 K3 ] K4 K5 K6 Al A2 J A3 [ A4 l G K/A'l'opic(s)
[ Imp.
Points 017.A2.02 Ability to predict the impacts of 017 In-core Temperature Monitor X
core damage on the ITM system. and based 4.1 1
on those predictions, use procedures to correct, control, or mitigate the consequences 022.K3.01 Knowledge of the effect that a loss or mallunction of the Containment 022 Containment Cooling X
Cooling System will have on containment 3.2 I
equipment subiect to damage by high or low temperature. humidity, and pressure.
025 Iee Condenser Suppressed -Not applicable to facility N/A N/A X
026.KI.01 Knowledge oflthe physical 4.2 1
connections and/or cause effect relationships between the CSS and ECCS.
026 Containment Spray X
026.A2.04 Ability to predict the impacts of a 4.2 1
failure of a spray pump on the CSS. and based on those predictions. use procedures to correct, control, or mitigate the consequences.
056 Condensate Topic not randomly selected.
N/A N/A X
059.A1.03 Ability to predict and/or monitor changes in parameters (to prevent exceeding 059 Main Feedwater design limits) associated with operating the 2.9 1
MFW controls including power level restrictions for operation of MFW pumps and valves.
X 061.K6.02 Knowledge of the effect of a loss 2.7 or malfunction of pumps will have on the AlW components.
061 Auxiliary/Emergency Feedwater X
061.A3.01 Ability to monitor automatic 4.2 1
operation of the AFW. including AFW I
_startup and flows.
ES-401 PWR SRO Examination Outline Form ES-401-3 Plant Systems - Tier 2/Group I (CONTINU El))
E/APE # /Name / Safety Function lK1K2lK3lK4 K5 I K6lAl A2 A3A4lG l K/ATopic(s)
] Imp. [ Points X
063.K4.02 Knowledge of DC electrical 3.2 system design feature(s) and/or interlock(s) which provide for breaker interlocks.
permissives, bypasses. and cross-ties.
063 DC Electrical Distribution X
063.A3.01 Ability to monitor automatic 3.1 operation of the DC electrical system including meters, annunciators, dials, recorders, and indicating lights.
068 Liquid Radwaste Topic not randomly selected.
N/A N/A 071 Waste (ias Disposal X
2.4.11 Knowledge of abnormal condition 3.6 procedures.
072.AI.01 Ability to predict and/or monitor changes in parameters (to prevent exceeding 072 Area Radiation Monitoring x
design limits) associated with operating the 3.6 ARM system controls including radiation levels.
K/A Category'l'otals:
2 2 1
2 2
I 2
2 212 J
2 Group Point Total:
19
ES-401 PWR SRO Examination Outline For ES-401-3 Plant Systems - Tier 2/Group 2 I/API'I 4 / Name / Safety Function KI l K2 l K3 l K4 l K5 [ K6 Al ] A2 l A3 A4 G,
K/A Topic(s)
Imp. l Points 002.K6.12 Knowledge of the eflect ol'a loss 002 Reactor Coolant X
or malfunction of the RCS Code Safety 3.5 I_
valves.
006 Emergency Core Cooling
'Topic not randomly selected.
N/A N/A 010 Pressurizer Pressure Control Topic not randomly selected.
N/A N/A 01 I.K4.04 Knowledge of PZR LCS design 011 Pressurizer Levcl Control X
feature(s) and/or interlock(s) which provide 3.3 for PZR level inputs.
012.KI.04 Knowledge of the physical 012 Reactor Protection X
connections and/or cause effect relationships 3.3 I
016.K5.01 Knowledge of the operational 016 Non-Nuclear Instrumentation x
implication of the separation of control and 2.8 protection circuits.
027 Containment Iodine Removal Suppressed - Not applicable to fucility N/A N/A 028.K5.04 Knowledge of the operational 028 H-ydrogen Recombiner and Purge Control X
implications of the selective removal of 3.2 hydrogcen as applied to the I IRI'S.
029 Containment Purge
_'Iopic not randomly selected.
N/A N/A 033 Spent Fuel l'ool Cooling Topic not randomly selected.
N/A N/A 034.A3.01 Ability to monitor automatic 034 Fuel Hlandling Equipment X
operation of the fuel handling system.
3.1 I
including travel limits.
X 035.K6.01 Knowledge of the effect of a loss 3.6 or malfunction of the MSIVs will have on the S/G system.
035 Steam Generator X
035.A1.01 Ability to predict and/or monitor 3.8 changcs in parameters (to prevent exceeding desig hlimits) associated with operating the S/G system controls includintg S/G wide and narrow range level during startup. shutdowt, and normal operations.
039.A4.04 Ability to manually operate 039 Main and Rcheat Steam X
and/or monitor in the control room 3.9
________________________emergenlcy feedwalter p~ump) turbines
ES-401 PWR SRO Examination Outline Form ES-40(1-3 Plant Systems - Tier 2/Group 2 (CON'INUED)
E/APE
/ Name / Safety Function IKI lK2 lK3 K4 K5 K6 Al A2 A3 IA4 G
I KATopic(s) l Imp.
Points 055.A3.03 Ability to monitor automatic 055 Condenser Air Removal X
operation of the CARS, including automatic 2.7 I
diversion of'the CARS exhaust.
062.K3.03 Knowledge oftthe effect of a loss (162 AC Electrical Distribution X
or malfunction of the AC Distribution system 3.9 1
will have on the DC system.
(64 Emergency Diesel Generator 064.K2.01 Knowledge of bus power supplies 3.1 to the EDG air compressor.
073.K4.01 Knowledge of l'RM system 073 Process Radiation Monitoring X
design feature(s) and/or interlock(s) which 4.3 1
provide for release termination when radiation exceeds setpoint.
075.A2.02 Ability to predict the impacts ofta loss of circulating water pumps on the CW 075 Circulating Water X
system, and based on those predictions. use 2.7 i
procedures to correct, control, or mitigate the
______consequences.
079 Station Air Topic not randomly selected.
N/A N/A 2.4.30 Knowledge of which events related to 086 Fire Protection X
system operations/status should be reported 3.6 I
to outside agencies.
X 103.KI.01 Knowledge oftthe physical 3.9 1
connections and/or cause effect relationships between the containment system and the containment cooling system.
103 Containment X
103.Al.01 Ability to predict and/or monitor 4.1 1
changes in parameters (to prevent exceeding design limits) associated with operating the containment system controls including containment pressure. temperature. and
=___ humidity.
K/A Category Totals:
2 lI lI 2
l2 l2 l2 1
2 I
I Group Point otal:
l 17
ES-401 PWR SRO Examination Outline For ES-401-3 Plant Systems - Tier 2/Group 3 I/APE # / Name / Safety Function K I K2 K3 K4 K5 K6 A I A2 A3 A4 IG K/A Topic(s)
I Imp.
Points 005.K6.03 Knowledge of the effect of a loss 005 Residual Heat Removal X
or malfunction of the RHR heat exchanger 2.6 will have on the RI IR system.
007 I'ressurizer Relief/Quench Tank Topic not randomly selected.
N/A N/A 008 Component Cooling Water Topic not randomly selected.
N/A N/A 041 Steam Dump/Turbine Bypass Control Topic not randomly selected.
N/A N/A 045 Main 'lurbine Generator Topic not randomly selected.
N/A N/A 076 Service Water 2.4.24 Knowledge of loss of coolino water 37 procedures.
X 078.K2.01 Knowledge of bus power supplies 2.9 to the instrument air compressors.
078 Instrument Air X
078.A3.01 Ability to monitor automatic 3.2 I
operation of the instrument air system.
ineluding air pressure.
K/A Category Iotals:
0 l
0 0
0 I
0
_O I
0 r
Group Point Total 4 4
ES-401 Generic Knowledge and Abilities Outline (Tier 3)
ES-401-5 Facility: KNP Date of Exam: 8/26/02 Exam Level: SRO Category K/A #
Topic Imp.
Points Ability to evaluate plant performance and make operational 2.1.7 judgements based on operating characteristics, reactor behavior, and 4.4 I
instrument interpretation.
Conduct of 2.1.9 Ability to direct personnel activities inside the control room.
4.0 l Operations 2.1.12 Ability to apply Technical Specifications for a system.
4.0 I
2.1.22 Ability to determine Mode of Operation.
3.3 I
Total 4
Ability to perform pre-startup procedures for the facility, including 2.2.1 operating those controls associated with plant equipment that could 3.6 I
affect reactivity.
Equipment 2.2.13 Knowledge of tagging and clearance procedures.
3.8 I
Control 2.2.22 Knowledge of limiting conditions for operations and safety limits.
4.1 2.2.34 Knowledge of the process for determining the internal and external 3I effects on core reactivity.
Total 4
2.3.1 Knowledge of 10 CFR 20 and related facility radiation control 0
requirements.
2.3.2 Knowledge of facility ALARA program.
2.9 I
Radiation Knowledge of the requirements for reviewing and approving release Control 2.3.6 permits.
2.3.9 Knowledge of the process for performing a containment purge.
3.4 I
Total 4
2.4.6 Knowledge of symptom based EOP mitigation strategies.
4.0 2.4.14 Knowledge of general guidelines for EOP flowchart use.
3.9 I
Emergency 2.4.25 Knowledge of fire protection procedures.
3.4 I
Procedures/
2436 Knowledge of Chemistry / Health Physics tasks during emergency
.8 Plan operations.
2.4.49 Ability to perform without reference those actions that require 4.0 I
immediate operation of system components and controls.
Total 5
Tier 3 Point Total (SRO) 17
Plant-Spccific Prioritics System / Topic Recommended Replacement for...
Reason Points NONE Plant-Specific Priority Total: (limit 10)
Kewaunee Outline Review NRC Comments/ LIC Response 5/22/02 WRITTEN:
- 1. NRC: What computer program do you use to randomly select KAs?
LIC: WD Associates. Description in package.
- 2. NRC: Were any KAs suppressed/rejected?
LIC: Yes. Suppressed KA list submitted with package.
(Not applicable questions: Were justification statements prepared? Were KAs suppressed/rejected/justified on a case-by-case basis? Which ones? Why? How many? We need to review the suppressed/rejected/justified KA information.)
ADMIN JPMs:
- 1. Make sure the admin JPMs have significant, verifiable consequences such that if they are performed incorrectly, the task cannot be successfully completed.
- 3. Make sure the JPM is significant enough to make a pass/fail license decision.
OPERATING JPMs:
General: Want alternate path JPMS to follow guidance in Appendix C, ie, procedurally driven (ARPs or ABNs are good), completes the task or mitigates the problem without reliance on actions by other control room operators...
- 1. I need to review a list of audit exam JPMs (to verify none of those JPMs are repeated on the NRC exam).
- 2. Make sure none of JPMs are performed in the scenarios. What about B.1.c (perform emergency boration) and B.1.d (place excess letdown in service)?
SCENARIOS:
- 1. Verify each scenario has Tech Specs
- 2. Verify each scenario has 2-3 critical tasks.
- 3. Verify 1-2 malfunctions after EOP entry.
Page 1 of 2
Kewaunee Outline Review NRC Comments! LIC Response 5/22/02
- 4. Scenario 1:
Event 7, Sump B suction valve S1-351 B will not open from control room, must be opened locally.
If opened locally, no measurable feedback for the applicant and cannot count this as a "C" failure for RO/SRO.
- 5. Scenario 2:
- 6. Scenario 3:
Event 1, Perform a power increase per N-0-02.
Does the power increase include rods and boron?
Page 2 of 2