ML050480131
| ML050480131 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 12/30/2003 |
| From: | Gody A Operations Branch IV |
| To: | Ray H Southern California Edison Co |
| References | |
| 50-361/03-301, 50-362/03-301 50-361/03-301, 50-362/03-301 | |
| Download: ML050480131 (27) | |
Text
ES-401 PWR RO Examination Outline Form ES-401-4 (R8, S1)
Facility: San Onofre Date of Exam: 12 December 2003 Exam Level: RO Tier Group K/A Category Points Point Total K
1 K
2 K
3 K
4 K
5 K
6 A
1 A
2 A
3 A
4 G
1 3
3 3
3 2
2 16 2
3 3
3 3
3 2
17 3
0 0
1 1
0 1
3
- 1. Emergency
& Abnormal Plant Evolutions Tier Totals 6
6 7
7 5
5 36 1
2 2
2 3
2 1
2 2
3 2
2 23 2
3 2
2 1
2 1
2 2
2 1
2 20 3
0 1
1 1
1 1
1 1
1 0
0 8
- 2. Plant Systems Tier Totals 5
5 5
4 6
3 5
5 6
3 4
51 Cat 1 Cat 2 Cat 3 Cat 4 13
- 3. Generic Knowledge and Abilities 4
3 3
3 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 +/-1 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 1 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 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.
ES-401 PWR RO Examination Outline Form ES-401-4 (R8, S1)
Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G
K/A Topic(s)
Imp.
Points 000005 Inoperable/Stuck Control Rod / 1 1
AK3.05 3.9 1
000015/17 RCP Malfunctions / 4 1
2.1.28 3.2 1
BW/E09; CE/A13; W/E09&E10 Natural Circ. / 4 0
000024 Emergency Boration / 1 1
AK1.02 3.6 1
000026 Loss of Component Cooling Water / 8 1
AK3.03 4.0 1
000027 Pressurizer Pressure Control System Malfunction / 3 1
AK1.02 2.8 1
000040 (BW/E05; CE/E05; W/E12) Steam Line Rupture - Excessive Heat Transfer / 4 0
CE/A11; W/E08 RCS Overcooling - PTS / 4 1
EK2.2 3.6 1
000051 Loss of Condenser Vacuum / 4 1
AA2.02 3.9 1
000055 Station Blackout / 6 1
EA1.07 4.3 1
000057 Loss of Vital AC Elec. Inst. Bus / 6 1
AA2.19 4.0 1
000062 Loss of Nuclear Service Water / 4 1
AA1.02 3.2 1
000067 Plant Fire On-site / 9 1
AA1.08 3.4 1
000068 (BW/A06) Control Room Evac. / 8 1
AK2.02 3.7 1
000069 (W/E14) Loss of CTMT Integrity / 5 1
AK1.01 2.6 1
000074 (W/E06&E07) Inad. Core Cooling / 4 1
EK3.11 4.0 1
BW/E03 Inadequate Subcooling Margin / 4 1
EK2.1 3.6 1
000076 High Reactor Coolant Activity / 9 1
2.1.32 3.4 1
BW/A02&A03 Loss of NNI-X/Y / 7 0
K/A Category Totals:
3 3
3 3
2 2
Group Point Total:
16
ES-401 PWR RO Examination Outline Form ES-401-4 (R8, S1)
Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G
K/A Topic(s)
Imp.
Points 000001 Continuous Rod Withdrawal / 1 1
2.2.25 1
000003 Dropped Control Rod / 1 1
AA1.06 1
000007 (BW/E02&E10; CE/E02) Reactor Trip -
Stabilization - Recovery / 1 1
EK1.2 1
BW/A01 Plant Runback / 1 0
BW/A04 Turbine Trip / 4 0
000008 Pressurizer Vapor Space Accident / 3 1
AK2.02 1
000009 Small Break LOCA / 3 1
EK3.26 1
000011 Large Break LOCA / 3 0
W/E04 LOCA Outside Containment / 3 1
EA2.1 1
BW/E08; W/E03 LOCA Cooldown/Depress. / 4 1
EK3.3 1
W/E11 Loss of Emergency Coolant Recirc. / 4 1
EK2.2 1
W/EO1 & E02 Rediagnosis & SI Termination / 3 1
EK1.1 1
000022 Loss of Reactor Coolant Makeup / 2 1
AK3.06 1
000025 Loss of RHR System / 4 1
AA2.02 1
000029 Anticipated Transient w/o Scram / 1 1
EA1.08 1
000032 Loss of Source Range NI / 7 1
AK1.01 1
000033 Loss of Intermediate Range NI / 7 1
AA2.07 1
000037 Steam Generator Tube Leak / 3 0
000038 Steam Generator Tube Rupture / 3 1
EA1.01 1
000054 (CE/E06) Loss of Main Feedwater / 4 0
BW/E04; W/E05 Inadequate Heat Transfer - Loss of Secondary Heat Sink / 4 1
2.4.14 1
000058 Loss of DC Power / 6 1
AA2.03 1
000059 Accidental Liquid RadWaste Rel. / 9 0
000060 Accidental Gaseous Radwaste Rel. / 9 0
000061 ARM System Alarms / 7 0
W/E16 High Containment Radiation / 9 0
CE/E09 Functional Recovery 0
K/A Category Point Totals:
3 2
3 3
4 2
Group Point Total:
17
ES-401 PWR RO Examination Outline Form ES-401-4 (R8, S1)
Emergency and Abnormal Plant Evolutions - Tier 1/Group 3 E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G
K/A Topic(s)
Imp.
Points 000028 Pressurizer Level Malfunction / 2 0
000036 (BW/A08) Fuel Handling Accident / 8 0
000056 Loss of Off-site Power / 6 1
AK3.02 1
000065 Loss of Instrument Air / 8 0
BW/E13&E14 EOP Rules and Enclosures 0
BW/A05 Emergency Diesel Actuation / 6 0
BW/A07 Flooding / 8 0
CE/A16 Excess RCS Leakage / 2 0
W/E13 Steam Generator Over-pressure / 4 1
2.4.47 1
W/E15 Containment Flooding / 5 1
EA1.3 1
K/A Category Point Totals:
0 0
1 1
0 1
Group Point Total:
3
ES-401 PWR RO Examination Outline Form ES-401-4 (R8, S1)
Plant Systems - Tier 2/Group 1 System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G
K/A Topic(s)
Imp.
Points 001 Control Rod Drive 1
AA2.03 3.5 1
001 Control Rod Drive 1
AA3.06 3.9 1
003 Reactor Coolant Pump 1
K6.14 2.6 1
003 Reactor Coolant Pump 1
K1.06 3.3 1
004 Chemical and Volume Control 1
K2.03 3.3 1
004 Chemical and Volume Control 1
K1.02 3.5 1
013 Engineered Safety Features Actuation 1
A1.03 2.6 1
013 Engineered Safety Features Actuation 1
A3.02 4.1 1
015 Nuclear Instrumentation 1
A3.03 3.9 1
015 Nuclear Instrumentation 1
2.2.2 4.0 1
017 In-core Temperature Monitor 1
A4.01 3.8 1
017 In-core Temperature Monitor 1
K3.01 3.5 1
022 Containment Cooling 1
A4.05 3.8 1
022 Containment Cooling 1
A1.02 3.6 1
025 Ice Condenser 0
056 Condensate 0
059 Main Feedwater 1
2.4.4 4.0 1
059 Main Feedwater 1
A2.12 3.1 1
061 Auxiliary/Emergency Feedwater 1
K4.06 4.0 1
061 Auxiliary/Emergency Feedwater 1
K2.02 3.7 1
068 Liquid Radwaste 1
K5.04 3.2 1
071 Waste Gas Disposal 1
K5.04 2.7 1
071 Waste Gas Disposal 1
K3.04 2.5 1
072 Area Radiation Monitoring 1
K4.03 2.5 1
072 Area Radiation Monitoring 1
K5.02 3.2 1
K/A Category Point Totals:
2 2
2 3
2 1
2 2
3 2
2 Group Point Total:
23
ES-401 PWR RO Examination Outline Form ES-401-4 (R8, S1)
Plant Systems - Tier 2/Group 2 System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G
K/A Topic(s)
Imp.
Points 002 Reactor Coolant 1
K1.08 4.5 1
006 Emergency Core Cooling 1
K6.03 3.6 1
010 Pressurizer Pressure Control 1
K4.03 3.8 1
011 Pressurizer Level Control 1
A3.02 2.6 1
012 Reactor Protection 1
K3.02 3.2 1
014 Rod Position Indication 1
K5.04 4.3 1
016 Non-nuclear Instrumentation 1
A2.03 3.0 1
026 Containment Spray 1
A3.01 4.3 1
029 Containment Purge 1
K3.02 2.9 1
033 Spent Fuel Pool Cooling 1
A1.02 2.8 1
035 Steam Generator 1
K1.01 4.2 1
039 Main and Reheat Steam 1
K1.05 2.5 1
055 Condenser Air Removal 1
K3.01 2.5 1
062 AC Electrical Distribution 1
2.4.16 3.0 1
063 DC Electrical Distribution 1
K2.01 2.9 1
064 Emergency Diesel Generator 1
A1.03 3.2 1
073 Process Radiation Monitoring 1
A4.01 3.9 1
075 Circulating Water 1
K2.03 2.6 1
079 Station Air 1
2.1.27 2.8 1
086 Fire Protection 1
A2.02 3.0 1
K/A Category Point Totals:
3 2
3 1
1 1
2 2
2 1
2 Group Point Total:
20
ES-401 PWR RO Examination Outline Form ES-401-4 (R8, S1)
Plant Systems - Tier 2/Group 3 System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G
K/A Topic(s)
Imp.
Points 005 Residual Heat Removal 1
A1.01 3.5 1
007 Pressurizer Relief/Quench Tank 0
008 Component Cooling Water 1
K4.09 2.7 1
027 Containment Iodine Removal 1
K2.01 3.1 1
028 Hydrogen Recombiner and Purge Control 0
034 Fuel Handling Equipment 1
A2.03 3.3 1
041 Steam Dump/Turbine Bypass Control 1
K6.03 2.7 1
045 Main Turbine Generator 1
K3.01 2.9 1
076 Service Water 0
078 Instrument Air 1
K1.05 3.4 1
103 Containment 1
A3.01 3.9 1
K/A Category Point Totals:
1 1
1 1
0 1
1 1
1 0
0 Group Point Total:
8 Plant-Specific Priorities System / Topic Recommended Replacement for...
Reason Points Plant-Specific Priority Total: (limit 10)
y Category K/A #
Topic Imp.
Points 2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation 3.7 1
2.1.12 Ability to apply Tech. Spec for a system 2.9 1
2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operations 3.9 1
2.1.32 Ability to explain and apply all system limits and precautions 3.4 1
Conduct of Operations Total 2.2.12 Knowledge of surveillance procedures 3.0 1
2.2.25 Knowledge of bases in technical specifications for LOCs and safety limits 2.5 1
2.2.27 Knowledge of refueling process 2.6 1
Equipment Control Total 2.3.1 Knowledge of 10CFR 20 and related facility rad con requirements.
2.6 1
2.3.9 Knowledge of the process for performing a containment purge.
2.5 1
2.3.11 Ability to control radiation releases 2.7 1
Radiation Control Total 2.4.6 Knowledge of the symptoms based EOP mitigation strategies.
3.1 1
2.4.24 Knowledge of Loss of Cooling water procedure.
3.3 1
2.4.27 Knowledge of fire procedures 3.0 1
Emergency Procedures/
Plan Total Tier 3 Point Total RO 13 8 of 45 NUREG-1021, Revision 8
ES-401 PWR SRO Examination Outline Form ES-401-3 (R8, S1)
Facility: San Onofre Nuclear Generating Station Date of Exam: 12/11/03 Exam Level: SRO Tier Group K/A Category Points Point Total K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G
- 1. Emergency
& Abnormal Plant Evolutions 1
4 3
4 5
4 4
24 2
3 2
3 3
3 2
16 3
0 0
1 1
0 1
3 Tier Totals 7
5 8
9 7
7 43
- 2. Plant Systems 1
1 2
1 2
2 1
2 2
3 2
1 19 2
3 2
2 1
1 1
2 2
1 1
1 17 3
0 1
1 1
0 0
1 0
0 0
0 4
Tier Totals 4
5 4
4 3
2 5
4 4
3 2
40
- 3. Generic Knowledge and Abilities Cat 1 Cat 2 Cat 3 Cat 4 4
4 4
5 17 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 +/-1 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 1 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 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.
ES-401 PW R SR O Examination Outline Form ES-401-3 (R8, S1)
Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G
K/A Topic(s)
Imp.
Points 000001 Continuous Rod W ithdrawal / 1 1
2.2.25 3.7 1
000003 Dropped Control Rod / 1 1
AA1.06 4.1 1
000005 Inoperable/Stuck Control Rod / 1 1
AK3.05 4.2 1
000011 Large B reak LO CA / 3 1
EA2.14 4.0 1
W /E04 LO CA Outside Containment / 3 1
EA2.1 4.2 1
W /EO 1 & E02 Rediagnosis & SI T ermination / 3 1
EK1.1 3.8 1
000015/17 R CP Malfunctions / 4 1
2.1.28 3.3 1
BW /E09; CE/A13; W /E09&E10 Natural Circ. / 4 1
EK3.4 3.6 1
000024 Em ergency Boration / 1 1
AK1.02 3.9 1
000026 Loss of Com ponent Cooling W ater / 8 1
AK3.03 4.2 1
000029 Anticipated T ransient w/o Scram / 1 1
EA1.08 4.5 1
000040 (BW /E05; CE/E05; W /E12) Steam Line Rupture
- Excessive Heat Transfer / 4 1
AK1.07 4.2 1
CE /A11; W /E08 RCS Overcooling - PT S / 4 1
EK2.2 3.1 1
000051 Loss of Condenser Vacuum / 4 1
AA2.02 4.1 1
000055 Station Blackout / 6 1
EA1.07 4.5 1
000057 Loss of Vital AC E lec. Inst. Bus / 6 1
AA2.19 4.3 1
000059 Accidental Liquid R adW aste Rel. / 9 1
2.4.50 3.3 1
000062 Loss of Nuclear Service W ater / 4 1
AA1.02 3.3 1
000067 Plant Fire On-site / 9 1
AA1.08 3.7 1
000068 (BW /A06) Control Room Evac. / 8 1
EK2.1 4.0 1
000069 (W /E14) Los s of CT MT Integrity / 5 1
Ak1.01 3.1 1
000074 (W /E06&E07) Inad. Core Cooling / 4 1
EK3.11 4.1 1
BW /E03 Inadequate Subcooling Margin / 4 1
Ek2.1 4.0 1
000076 High Reactor C oolant Activity / 9 1
2.1.32 3.8 1
BW /A02&A03 Loss of NN I-X/Y / 7 K/A Category Totals:
4 3
4 5
4 4
Group Point Total:
24
ES-401 PW R SR O Examination Outline Form ES-401-3 (R8, S1)
Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G
K/A Topic(s)
Imp.
Points 000007 (BW /E02&E10; CE/E02) Reactor Trip -
Stabilization - Recovery / 1 1
K1.2 3.9 BW /A01 Plant Runback / 1 BW /A04 Turbine Trip / 4 000008 Pressurizer Vapor Space Accident / 3 1
AK2.02 2.7 000009 Sm all Break LO CA / 3 1
EK3.26 4.5 BW /E08; W /E03 LO CA Cooldown - Depress. / 4 1
EK3.3 3.9 W /E11 Loss of Em ergency Coolant R ecirc. / 4 1
EK2.2 4.3 000022 Loss of Reactor Coolant Makeup / 2 1
AK3.06 3.3 000025 Loss of RH R S ystem / 4 1
AA2.02 3.8 000027 Press urizer Pressure Control System Malfunction / 3 1
AK1.02 3.1 000032 Loss of Source R ange N I / 7 1
AK1.01 3.1 000033 Loss of Intermediate Range NI / 7 1
AA2.07 4.2 000037 Steam Generator Tube Leak / 3 1
2.1.7 4.4 000038 Steam Generator Tube R upture / 3 1
EA1.01 4.4 000054 (CE/E06) Loss of Main Feedwater / 4 1
AA1.01 4.4 BW /E04; W /E05 Inadequate Heat Transfer - Loss of Secondary Heat Sink / 4 1
2.4.14 3.9 000058 Loss of DC Power / 6 1
AA2.03 3.9 000060 Accidental Gaseous Radwaste Rel. / 9 000061 AR M S ystem Alarms / 7 W /E16 High Containment Radiation / 9 000065 Loss of Instrument Air / 8 1
AA1.03 3.1 CE/E09 Functional Recovery K/A Category Point Totals:
3 2
3 3
3 2
Group Point Total:
16
ES-401 PW R SR O Examination Outline Form ES-401-3 (R8, S1)
Emergency and Abnormal Plant Evolutions - Tier 1/Group 3 E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G
K/A Topic(s)
Imp.
Points 000028 Pressurizer Level Malfunction / 2 000036 (BW /A08) Fuel Handling Accident / 8 000056 Loss of Off-site Power / 6 1
AK3.02 4.7 BW /E13&E14 EOP R ules and Enclosures BW /A05 Emergency Diesel Actuation / 6 BW /A07 Flooding / 8 CE /A16 Excess R CS Leakage / 2 W /E13 Steam G enerator O ver-pressure / 4 1
2.4.47 3.7 W /E15 Containment Flooding / 5 1
EA1.3 3.0`
K/A Category Point Totals:
0 0
1 1
0 1
Group Point Total:
3
ES-401 PW R SR O Examination Outline Form ES-401-3 (R8, S1)
Plant Systems - Tier 2/Group 1 System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G
K/A Topic(s)
Imp.
Points 001 Control Rod Drive 1
A2.03 4.2 003 Reactor Coolant Pump 1
K6.14 2.9 003 Reactor Coolant Pump 1
K1.03 3.6 004 Chemical and Volume Control 1
K2.03 3.5 013 Engineered Safety Features Actuation 1
A1.03 2.6 013 Engineered Safety Features Actuation 1
A3.02 4.2 014 Rod Position Indication 1
K5.02 3.3 015 Nuclear Instrumentation 1
A3.03 3.9 017 In-core Temperature Monitor 1
A4.01 4.1 022 Containment Cooling 1
A4.05 3.8 022 Containment Cooling 1
A1.02 3.8 025 Ice Condenser 026 C ontainm ent Spray 1
A3.01 4.5 056 Condensate 1
A2.04 2.8 059 Main Feedwater 1
2.4.4 4.3 061 Auxiliary/Emergency Feedwater 1
K4.06 4.2 063 DC Electrical Distribution 1
K2.01 3.1 068 Liquid Radwaste 1
K5.04 3.5 071 W aste G as D isposal 1
K3.04 2.9 072 Area Radiation Monitoring 1
K4.03 3.6 K/A Category Point Totals:
1 2
1 2
2 1
2 2
3 2
1 Group Point Total:
19
ES-401 PW R SR O Examination Outline Form ES-401-3 (R8, S1)
Plant Systems - Tier 2/Group 2 System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G
K/A Topic(s)
Imp.
Points 002 Reactor Coolant 1
K1.08 4.6 006 Emergency Core Cooling 1
K6.03 3.9 010 Pressurizer Pressure Control 1
K4.03 4.1 011 Pressurizer Level Control 1
A3.02 2.8 012 Reactor Protection 1
K5.01 3.8 016 Non-nuclear Ins trumentation 1
A2.03 3.3 027 Containment Iodine R emoval 1
K2.01 3.4 028 Hydrogen R ecombiner and Purge Control 029 Containment Purge 1
K3.02 3.5 033 Spent Fuel Pool Cooling 1
A1.01 3.3 034 Fuel Handling Equipment 035 Steam Generator 1
K1.01 4.5 039 Main and R eheat Steam 1
K1.05 2.6 055 Condenser Air R emoval 1
K3.01 2.7 062 AC Electrical Distribution 1
2.4.16 4.0 064 Em ergency Diesel Generator 1
A1.03 3.3 073 Process Radiation Monitoring 1
A4.01 3.9 075 Circulating W ater 1
K2.03 2.7 079 Station Air A2.02 3.3 086 Fire Protection 1
103 Containment K/A Category Point Totals:
3 2
2 1
1 1
2 2
1 1
1 Group Point Total:
17
ES-401 PW R SR O Examination Outline Form ES-401-3 (R8, S1)
Plant Systems - Tier 2/Group 3 System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G
K/A Topic(s)
Imp.
Points 005 Res idual Heat Rem oval 1
A1.01 3.6 007 Pressurizer Relief/Quench Tank 008 Com ponent Cooling W ater 1
K4.09 2.9 041 Steam Dump/Turbine Bypass Control 1
K2.01 3.4 045 Main Turbine Generator 1
K3.01 3.2 076 Service W ater 078 Instrument Air K/A Category Point Totals:
0 1
1 1
0 0
1 0
0 0
0 Group Point Total:
4 Plant-Specific Priorities System / Topic Rec omm ended Replacement for...
Reason Points Plant-Specific Priority Total: (limit 10)
ES-401Generic Knowledge and Abilities Outline (Tier 3)Form ES-401-5 (R8, S1)
Facility: Date of Exam:
Exam Level:
Category K/A #
Topic Imp.
Points Conduct of Operations 2.1.12 Ability to apply tech spec for a system 4.0 1
2.1.16 Ability to operate any phone or pager system and 2-way radio 2.8 1
2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation 4.0 1
2.1.32 Ability to explain and apply all system limits and precautions 3.8 1
Total 4
Equipment Control 2.2.6 Knowledge of the process for making changes in procedures as described in the safety analysis report.
3.3 1
2.2.12 Knowledge of surveillance procedures 3.4 1
2.2.25 Knowledge of the bases in technical specifications for limiting conditions for operations and safety limits 2.6 1
2.2.27 Knowledge of the refueling process 3.8 1
Total 4
Radiation Control 2.3.1 Knowledge of 10 CFR: 20 and related facility radiation control requirements 3.0 1
2.3.6 Knowledge of the requirements for reviewing and approving release permits 3.1 1
2.3.9 Knowledge of the process for performing a planned gaseous radioactive release 3.4 1
2.3.11 Ability to control radiation releases 3.2 1
Total 4
Emergency Procedures/
Plan 2.4.1 Knowledge of EOP entry conditions and immediate action steps 4.6 1
2.4.6 Knowledge of symptom based EOP mitigation strategies 4.0 1
2.4.24 Knowledge of loss of cooling water procedures 3.7 1
2.4.25 Knowledge of fire protection procedures 3.4 1
2.4.46 Ability to verify that the alarms are consistent with the plant conditions 3.6 1
Total 5
Tier 3 Point Total (RO/SRO) 17
ES-301 Administrative Topics Outline Form-ES-301-1 Form ES-301-1 RO ADM Outline NUREG-1021, Revision 8 Facility:
SONGS Date of Examination:
12/15/2003 Examination Level:
RO Operating Test Number:
1 Describe method of evaluation:
- 2. TWO Administrative Questions A.1a Conduct of Operations 2.1.21 Ability to obtain and verify controlled procedure copy. (3.1)
JPM: Prepare a procedure change request using a Procedure Modification Permit A.1b Conduct of Operations 2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. (3.7)
JPM: Determine Shutdown Margin A.2 Equipment Control 2.2.13 Knowledge of tagging and clearance procedures. (3.6)
JPM: Review a Boundary Verification A.3 Radiation Exposure Control 2.3.2 Knowledge of facility ALARA program. (2.5)
JPM: Determine stay time for work to be performed A.4 Emergency Plan 2.4.43 Knowledge of RO responsibilities in E-Plan implementation.
(3.3)
JPM: Initiate Initial Site Accountability Forms
ES-301 Administrative Topics Outline Form-ES-301-1 Form ES-301-1 RO ADM Outline NUREG-1021, Revision 8 Task Summary A1a Applicant will be given a set of conditions requiring a change to an existing procedure. The task will be to initiate a procedure change, choosing between using a Procedure Modification Permit (PMP) or using a Temporary Change Notice (TCN)
A1b Applicant will be given a set of plant conditions that include a CEA anomaly such as a stuck CEA. The task will be to determine Shutdown Margin by manually performing a calculation given the available conditions A2 Applicant will be given a tagging boundary to verify. The tagging request will contain a critical error that the applicant must identify and correct prior to completing the verification A3 Applicant will be given a task that will be performed in the RCA. The task is to determine stay time for the work performed based on conditions provided, including REP and task with applicable dose limitations, and a survey map of the area where the work will be performed A4 Applicant will be given conditions requiring site accountability during an emergency event. The task is to initiate the Site Accountability form in accordance with the Emergency Plan.
ES-301 Administrative Topics Outline Form-ES-301-1 Form ES-301-1 SRO ADM Outline NUREG-1021, Revision 8 Facility:
SONGS Date of Examination:
12/15/2003 Examination Level:
SRO Operating Test Number:
1 Describe method of evaluation:
- 2. TWO Administrative Questions A.1a Conduct of Operations 2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation (4.0)
JPM: Determine Time to Boil Margin for Shutdown Nuclear Safety A.1b Conduct of Operations 2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. (4.4)
JPM: Review a Shutdown Margin calculation and direct appropriate actions A.2 Equipment Control 2.2.13 Knowledge of tagging and clearance procedures (3.8)
JPM: Evaluate (for approval) a work authorization request A.3 Radiation Exposure Control 2.3.2 Knowledge of facility ALARA program. (2.9)
JPM:
Determine stay time for work to be performed and take action for change in radiological conditions A.4 Emergency Plan 2.4.41 Knowledge of the emergency action level thresholds and classifications. (4.1)
JPM: Perform Event Classification
ES-301 Administrative Topics Outline Form-ES-301-1 Form ES-301-1 SRO ADM Outline NUREG-1021, Revision 8 Task Summary A1a Applicant will be given plant conditions where the RCS is in a Drained Down condition. The task will be to complete a calculation of RCS Time to Boil Margin for Shutdown Nuclear Safety, using applicable correction factors for each parameter or factor affecting the Time to Boil Margin A1b Applicant will be required to review a Shutdown Margin calculation that will require action in accordance with Technical Specifications. The task is to determine that Shutdown Margin does not meet requirements and determine the appropriate action in accordance with Technical Specifications A2 Applicant will be required to review a work authorization for approval. The work authorization will contain critical errors that must be identified and corrected prior to approval.
A3 Applicant will be given a task that will be performed in the RCA. The task is to determine stay time for the work performed based on conditions provided, including REP and task with applicable dose limitations, and a survey map of the area where the work will be performed. When the candidate performs the task, a change in radiological conditions will require the candidate to take action to minimize exposure A4 Applicant will be given a set of plant conditions requiring evaluation of the site emergency plan for event classification. The task is to determine the appropriate criteria and EAL and make the correct event classification in accordance with the emergency plan.
ES-301 Control Room Systems and Facility Walk-Through Test Outline Form-301-2 Facility:
SONGS Units 2 and 3 Date of Examination:
12/15/2003 Exam Level:
RO/SROI Operating Test No.:
1 B.1: Control Room Systems System JPM Description Type Code*
Safety Function S1 001 CEDMCS Control Large ASI Oscillations MSA 1
S2 006 ECCS Drain SIT 2T007 DS 2
S3 041 SBCS Cooldown to 530°F on SBCS following SGTR NSE 4S S4 003 RCP Start RCP P-004 LMSA 4P S5 026 CSS Verify RAS Initiation Repeat From Last NRC DASE 5
S6 062 AC Distribution Restore Bus 2A06 From 1E Cross Tie Operations MS 6
S7 012 RPS Set a CEAC INOP Flag in a CPC DS 7
B.2 Facility Walk-Through P1 001 CEDMCS Perform Local ATWS Actions Repeat From Last NRC DER 1
P2 039 MRSS Manually open ADV DA 4S P3 064 EDG Shutdown EDG and Transfer Control to Control Room DE 6
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, (E)OP/AB NOTES S1 When candidate gains control of ASI, 2 CEAs will drop, requiring manual reactor trip S3 New JPM to examine ability to use SBCS instead of ADVs for RCS C/D following SGTR S4 Modified to provide for High Upper Thrust Bearing temperature requiring RCP Trip IAW ARPs S5 Verification of RAS will indicate that equipment has failed to start. Manipulations of actuated equipment must be performed manually S6 Similar to 2000 NRC Exam JPM but simulator mods have made it easier to complete task on 2A06 without excessive cueing P2 ADV will not open manually. Isolation valve must be throttled to minimize DP to allow for ADV opening
Appendix D Scenario Outline Form ES-D-1 Facility:
SONGS Scenario No.:
1 Op Test No.:
1 Examiners:
Candidates:
53% Power, BOC. Dilution and power increase in progress Train A CCW in service HPSI P-017 OOS MDAFW P-141 OOS RM-7818 OOS Turnover:
Continue to raise power at 3% per hour Critical Tasks:
Start HPSI Re-establish RCS Subcooling Throttle HPSI Event No.
Malf.
No.
Event Type*
Event Description 1
(R) CO (N) ACO (N) CRS Raise Power 2
TU10A (C) ALL Main Turbine HP Governor Valve fails closed 3
CC03A (C) ACO Train A CCW Header Leak 4
CV02A CV03A (C) CO RCP P-001 Seal Failure 5
RC05A (M) ALL RCP P-001 Shaft Shear 6
RC03 SBLOCA upon Reactor Trip 7
RP01C EC08D (C) CO HPSI Actuation Failure (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
Scenario Event Description NRC Scenario 1 The crew will assume the shift at 53% power, Beginning of Cycle. A power increase is in progress in accordance with SO23-5-1.7, Power Operations.
After the crew has demonstrated control of the plant power change, a Main Turbine HP Governor Control valve will fail closed, requiring the crew to stabilize the plant and respond to the failure in accordance with Annunciator Response Procedures and SO23-10-3, Operation of the Turbine Control and Protection System.
When the plant is stable and action for the failed governor control valve is complete, a rupture of the operating CCW loop will occur. The crew will diagnose the failure using Annunciator Response Procedures and SO23-13-7, Loss of Component Cooling Water/Salt Water Cooling. The crew will be required to start the idle CCW and SWC pumps and transfer cooling loads to the operable header. The ruptured header will be removed from service. The CRS will be required to evaluate Technical Specifications.
The following event is an RCP seal failure. The crew will respond in accordance with Annunciator Response Procedures and SO23-13-6, RCP Seal Failure. The crew will diagnose a failure of the lower and middle seals on RCP P-001, and the CRS will determine the need for a plant shutdown in accordance with SO23-5-1.7, Power Operations.
As the crew prepares to initiate a plant shutdown, RCP P-001 shaft shears, and the reactor will trip on low RCS loop flow.
A Small Break LOCA develops, and the remaining RCPs must be tripped due to CIAS actuation or loss of RCS subcooling. Additionally, the crew must manually start the available HPSI pump due to an automatic actuation failure.
The scenario is terminated upon initiation of HPSI Throttle/Stop when the criteria are met in SO23-12-3, Loss of Coolant Accident.
Risk Significance:
Risk important components out of service:
HPSI P-017, MDAFW P-141 Failure of risk important system prior to trip:
CCW Rupture Risk significant core damage sequence:
SBLOCA with HPSI failure Risk significant operator actions:
Manual HPSI initiation
Appendix D Scenario Outline Form ES-D-1 Facility:
SONGS Scenario No.:
2 Op Test No.:
1 Examiners:
Candidates:
100% Power, MOC Train A CCW in service HPSI P-017 OOS MDAFW P-141 OOS RM-7818 OOS Turnover:
CREACUS surveillance is in progress. Complete the surveillance Critical Tasks:
Depressurize RCS HPSI Throttle Stop Event No.
Malf. No.
Event Type*
Event Description 1
(N) ACO CREACUS alignment 2
RC15B (I) CO Pressurizer Pressure Transmitter PT-100Y Fails High 3
SG04C (I) ACO SG-E089 level transmitter fails high 4
FW09A (C) ALL (R) CO Main Feed Pump A Trip - Load Reduction required 5
SG01B SG02B (C) ALL SG Tube Leak SG-E089 6
FW09B (M) ALL Main Feed Pump B Trip - Reactor Trip required 7
TU07 (C) ACO Turbine fails to trip. Manual trip required 8
MS05B (M) ALL ESDE outside containment SG-E089 (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
Scenario Event Description NRC Scenario 2 The crew will assume the shift at 100% power, Middle of Cycle. Initial conditions will require the crew to secure CREACUS from a surveillance run when they assume the shift.
The controlling Pressurizer Pressure Channel, PT-0100-Y, will fail high. The crew will take action in accordance with Annunciator Response Procedures and SO23-3-1.10, Pressurizer Pressure and Level Control. Channel X will be selected for control and the CRS will evaluate Technical Specifications.
When Pressurizer Pressure Control is restored, SG E089 level transmitter LT-1105 will fail high. The crew must take action to restore feed flow in accordance with Annunciator Response Procedures and SO23-13-24, Feedwater Control System Malfunction. SG E089 level will ultimately be restored to automatic control by contacting I&C.
When SG level control is restored to automatic, Main Feedwater Pump P-062 will trip on high vibration, requiring a rapid power reduction to approximately 70% power in accordance with SO23-13-24 and SO23-5-1.7, Power Operations.
When the plant is stabilized, a steam generator tube leak will develop on SG E089. The crew will respond in accordance with Annunciator Response Procedures, SO23-13-14, Reactor Coolant Leak, and Technical Specifications. The CRS will determine that a reactor shutdown is required due to excessive primary to secondary leakage.
When preparations for plant shutdown are being made, Main Feedwater Pump P-063 will trip, requiring a reactor trip.
The Main Turbine must be manually tripped. The steam generator tube leak will increase in severity, and after the Main Turbine is manually tripped an ESDE will occur on the SG with the Tube Rupture.
The scenario is terminated when RCS pressure is reduced and HPSI is throttled.
Risk Significance:
Risk important components out of service: HPSI P-017, MDAFW P-141 Risk significant initiating event:
Loss of Main Feedwater Risk significant operator actions:
Manual Turbine Trip, RCS depressurization
Appendix D Scenario Outline Form ES-D-1 Facility:
SONGS Scenario No.:
3 Op Test No.:
1 Examiners:
Candidates:
100% Power, MOC.
Train A CCW in service HPSI Pump P-017 OOS MDAFW Pump P-141 OOS RM-7818 OOS Turnover:
Maintain current plant conditions Critical Tasks:
Restore AFW flow Initiate Emergency Boration Event No.
Malf. No.
Event Type*
Event Description 1
RC16B (I) CO Controlling PZR Level Channel Y Fails Low 2
CW05B (C) ACO Circulating Water Pump P-116 Trips 3
(R) CO (N)
ACO/CRS Load Reduction 4
CV17A (C) CO BAMU Pump P-174 Trip 5
TP02B (C) ACO TPCW Pump P-120 Trip. P-119 Auto Start Failure.
6 PG05 (M) ALL Main Generator trip on Main Transformer Sudden pressure 7
RD0242 RD0248 (C) CO Two Stuck CEAs. Emergency Boration required 8
FW02B MDAFW P-504 Trip 9
FW25 SDAFW Overspeed (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
Scenario Event Description NRC Scenario 3 The crew will assume the shift at 100% power, Middle of Cycle.
The controlling Pressurizer Level Channel, LI-0110, Channel Y, fails low. The crew will take action to stabilize Charging and Letdown flow, and transfer pressurizer level control and heater operation to channel X, in accordance with SO23-3-1.10, Pressurizer Pressure and Level Control. The CRS will evaluate Technical Specifications.
When Pressurizer Level Control is restored to automatic, Circulating Water Pump P-116 trips on overcurrent. The crew will respond in accordance with Annunciator Response Procedures to determine that Main Condenser Vacuum remains in the unrestricted area of operation, and SO23-2-5, Circulating Water System Operation, to remove the Circulating Water Pump and condenser section from service. Based on the expected outage period, the crew will determine that a power reduction is required.
When the CO initiates RCS boration for the power reduction, the in-service BAMU Pump will trip, requiring manual operation to start the standby BAMU Pump.
When the RCS boration is in progress, TPCW Pump P-120 will trip, and P-119 fails to automatically start. The crew must manually start TPCW Pump P-119 in accordance with Annunciator Response Procedures to prevent a plant trip on High Stator Cooling Water temperature.
When TPCW is restored and the load decrease is in progress, a Main Transformer Fault will occur. The Main Generator will trip, causing a Main Turbine trip. When the reactor trips, two CEAs will be stuck, requiring Emergency Boration of the RCS.
Additionally, Main Condenser Vacuum is lost on the reactor trip, making Main Feedwater unavailable.
The crew will transition to SO23-13-2, Reactor Trip Recovery. Subsequent AFW failures will require entry to SO23-12-6, Loss of Feedwater.
AFW will be restored by manual reset and start of the Turbine Driven AFW pump.
The scenario is terminated when feedwater flow is restored.
Risk Significance:
Risk important components out of service: HPSI P-017, MDAFW P-141 Risk significant initiating event:
Turbine Trip, Loss of Off Site power source Risk significant operator actions: