ML12072A351
| ML12072A351 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 12/09/2011 |
| From: | Laing D Entergy Nuclear Northeast, Entergy Nuclear Operations |
| To: | Peter Presby Operations Branch I |
| Jackson D | |
| Shared Package | |
| ML113070677 | List: |
| References | |
| TAC U01843, JTRG-11-007 | |
| Download: ML12072A351 (32) | |
Text
Entergy Nuclear Northeast Entergy Nuclear Operations, Inc.
James A. Fitzpatrick NPP P.O. Box 110 Lycoming, NY 13093 Tel 315 342 3840
-===-Entergy December 9,2011 JTRG-11-007 Mr. Peter Presby, NRC Chief Examiner United States Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406-1415
SUBJECT:
LICENSED OPERATOR INITIAL EXAMINATION OUTLINES
Dear Mr. Presby,
In accordance with NUREG-1021 Revision 9, please find the attached Initial License Examination Outlines and Exam Outline Quality Checklist Random sampling for written test items was completed per ES-401 Attachment 1. The attached materials shall be withheld from public disclosure until the examinations are complete.
If you have any questions, please do not hesitate to contact me at (315) 349 6023.
DANIEL LAING TRAINING MANAGER Attachment CC:
Ops Initial Training Superintendent JTRG File
-75 Day NRC Outline Submittals
ES-401 BWR Examination Outline Form ES-401-1 Date of Exam: 3/8/12 RO KIA Category Points SRO-Only Points Tier Group K K K K K K A A A A G A2 G*
Total 1 2 3 4 5 6 1 2 3 4 Total
- 1.
1 2
4 5
2 4
3 20 5
2 7
Emergency &
Abnormal Plant 2
0 2
2 N/A 1
0 N/A 2
7 2
1 3
Evolutions Tier Totals 2
6 7
3 4
5 27 7
3 10 1
3 2
3 1
2 5
1 2
2 2
3 26 3
2 5
II
- 2.
')
II 2
3 1
1 2
0 1
2 0
2 0
0 12 0
1 Systems Plant Tier Totals 6
3 4
3 2
6 3
2 4
2 3
38 4
- 3. Generic Knowledge and Abilities 1
2 3
4 10 1
2 3
4 7
Categories 3
2 2
3 2
1 2
Note:
- 1.
Ensure that at least two topics from every applicable KJA category are sampled within each tier of the RO and SRO-only outlines (I.e., except for one category in Tier 3 of the SRO-only outline, the "Tier Totals" in each KJA 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 RO exam must total 75 points and the SRO-only exam must total 25 points.
- 3.
Systems/evolutions within each group are identified on the associated outline: systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate KIA statements.
- 4.
Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.
- 5.
Absent a plant-specific priority, only those KJAs having an importance rating (IR) of 2.5 or higher shall be selected.
Use the RO and SRO ratings for the RO and SRO-only portions, respectively.
- 6.
Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.
7.*
The generiC (G) KJAs in Tiers 1 and 2 shall be selected from Section 2 of the KJA Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable KJAs.
- 8.
On the following pages, enter the KJA numbers, a brief description of each topic, the topics' importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams.
- 9.
For Tier 3, select topics from Section 2 of the KJA catalog, and enter the KJA numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to KJAs that are linked to 10 CFR 55.43.
~S-401 2
Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emeraencv and Abnormal Plant Evolutions - Tier 1/Group 1 (RO I SRO)
EIAPE # I Name I Safety Function 295001 Partial or Complete Loss of Forced Core Flow Circulation 11 &4 295003 Partial or Complete Loss of AC 16 295004 Partial or Total Loss of DC Pwr / 6
- 295005 Main Turbine Generator Trip / 3 295006 SCRAM 1 1 295016 Control Room Abandonment 17 295018 Partial or Total Loss of CCW /8 K
KIA Topic(s)
IR 1
AK 2,01 Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF FORCED 3,61 3.7 1
CORE FLOW CIRCULATION and the following:
(CFR: 41.7 I 45.8)
Recirculation system X 2,4.50 Emergency Procedures I Plan 4.21 4.0 2
Ability to verify system alarm setpoints and operate controls identified in the alarm response manual. (CFR: 41.10 I 43.5 / 45.3)
X AA 2.03 Ability to determine and/or interpret the 2.8/
3 following as they apply to PARTIAL OR 2.9 COMPLETE LOSS OF D.C. POWER:
(CFR: 41.10/43.5/45.13)
Battery voltage X
AK 3.07 Knowledge of the reasons for the following responses as they apply to MAIN 3.81 3.8 4
TURBINE GENERATOR TRIP: (CFR: 41.5/45.6)
Bypass valve operation X
AK 3.03 Knowledge of the reasons for the following responses as they apply to SCRAM:
3.81 3.9 5
(CFR: 415/45.6)
Reactor pressure response X
AA 2.03 Ability to determine and/or interpret the following as they apply to CONTROL ROOM 4.31 4.4 6
ABANDONMENT: (CFR: 41.10/43.5/45.13)
Reactor pressure X
AA 2,03 Ability to determine and/or interpret the following as they apply to PARTIAL OR 3.21 3.5 7
COMPLETE LOSS OF COMPONENT COOLING WATER: (CFR: 41.10/43.5/45.13)
Cause for partial or complete loss
If 295019 Partial or Total Loss of Inst. Air 18 3.51 8
X AA 2.01 Ability to determine andlor interpret the following as they apply to PARTIAL OR 3.6 COMPLETE LOSS OF INSTRUMENT AIR:
(CFR: 41.10/43.5/45.13)
Instrument air system pressure 295021 Loss of Shutdown Cooling 14 X
AK 3.04 Knowledge of the reasons for the 3.31 9
following responses as they apply to LOSS OF 3.4 SHUTDOWN COOLING: (CFR: 41.5 /45.6)
Maximizing reactor water cleanup flow 295023 Refueling Acc I 8 X
AK 1.02 Knowledge of the operational 3.21 10 implications of the following concepts as they 3.6 apply to REFUELING ACCIDENTS:
(CFR: 41.8 to 41.10)
Shutdown margin 295024 High Drywell Pressure 15 X
3.91 11 EA 1.17 Ability to operate andlor monitor the following as they apply to HIGH DRYWELL
3.9 PRESSURE
(CFR: 41.7/45.6)
Containment spray: Plant*Specific 295025 High Reactor Pressure / 3 X
3.5/
12 EK 3.08 Knowledge of the reasons for the following responses as they apply to HIGH 3.5 REACTOR PRESSURE: (CFR: 41.5/45.6)
Reactor/turbine pressure regulating system operation 295026 Suppression Pool High Water X
3.91 13 EK 3.02 Knowledge of the reasons for the Temp. 15 4.0 following responses as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: (CFR: 41.5 1 45.6)
Suppression pool cooling 295027 High Containment Temperature 1 5 295028 High Drywell Temperature I 5 X 2.4.35 Emergency Procedures I Plan 3.81 14 4.0 Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects. (CFR: 41.10/43.5/45.13) 295030 Low Suppression Pool Wtr Lvii 5 3.71 15 X
EK 2.04 Knowledge of the interrelations between LOW SUPPRESSION POOL WATER LEVEL and 3.8 the following: (CFR: 41.7 I 45.8)
RHRlLPCI
295031 Reactor Low Water Levell 2 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown 11 295038 High Off-sile Release Rate 19 600000 Plant Fire On Site 18 700000 Generator Voltage and Electric Grid Disturbances 1 6 KIA Category Totals:
X 2.4.18 Emergency Procedures I Plan 3.31 16 4.0 Knowledge of the specific bases for EOPs.
(CFR: 41.10/43.1 145.13)
X EK 1.06 Knowledge of the operational implications of the following concepts as they 4.01 42 17 apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN:
(CFR: 41.8 to 41.10)
Cool down effects on reactor power X
EA 1.03 Ability to operate and/or monitor the following as they apply to HIGH OFF-SITE 3.71 3.9 18 RELEASE RATE: (CFR: 41,7 1 45.6)
Process liquid radiation monitoring system X
AK 2.01 Knowledge of the interrelations between PLANT FIRE ON SITE and the following:
2.61 2,7 19 Sensors 1detectors and valves X
AK 2.03 Knowledge of the interrelations between GENERATOR VOLTAGE AND ELECTRIC GRID 3,01 3,1 20 DISTURBANCES and the following:
(CFR:41A,41~,41~,41.10/45.8)
Sensors, detectors, indicators 2
4 5
2 4
3 Group Point Total:
20
ES-401 3
Form ES401-1 I ES-401 BWR Examination Outline Form ES-401-1
~
! EIAPE # I Name I Safety Function 295002 Loss of Main Condenser Vac I 3 295007 High Reactor Pressure I 3 295008 High Reactor Water Level "
295009 Low Reactor Water Levell 2 295010 High Drywell Pressure I 5 295011 High Containment Temp 15 295012 High Drywell Temperature 15 295013 High Suppression Pool Temp, 15 295014 Inadvertent Reactivity Addition 11 295015 Incomplete SCRAM 11 295017 High Oft-site Release Rate I 9 295020 Inadvertent Cont. Isolation 15 & 7 295022 Loss of CRD Pumps I 1 295029 High Suppression Pool Wtr Lvl/5 295032 High Secondary Containment Area Temperature 1 5 K
K K
1 2
3 A
1 I~
G X
X X
X X
)
KIA Topic(s)
AK 2.06 Knowledge of the interrelations between HIGH REACTOR PRESSURE and the following:
(CFR: 41.7 145.8)
PCIS AA 1.04 Ability to operate andlor monitor the following as they apply to HIGH DRYWELL PRESSURE: (CFR: 41,7/456)
Drywell sampling system AK 3,01 Knowledge of the reasons for the following responses as they apply to HIGH DRYWELL TEMPERATURE: (CFR: 41,5/45,6)
Increased drywell COOling AK 2.03 Knowledge of the interrelations between INCOMPLETE SCRAM and the following: (CFR: 41,7 145,8)
Rod control and information system 2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation.
(CFR: 41.10 I 43.51 45,2/45,6)
IR 3.51 3.7 3,11 3,0 3,51 3,6 3.21 3,6 4.31 4.4 23 I
24 27 22 26
295033 High Secondary Containment Area Radiation Levels I 9 295034 Secondary Containment X 2.1.28 Knowledge of the purpose and function of 4.11 21 Ventilation High Radiation I 9 major system components and controls. (CFR: 41.7) 4.1 295035 Secondary Containment High I I Differential Pressure / 5 295036 Secondary Containment High Sump/Area Water Levell 5 X
EK 3.02 Knowledge of the reasons for the following responses as they apply to SECONDARY 2.81 28 25 CONTAINMENT HIGH SUMP/AREA WATER LEVEL: (CFR: 41.5/45.6)
Reactor SCRAM 500000 Hi2h CTMT Hydrogen Conc. I 5 fBI 0
2 2
1 0 2
Group Point Total:
7
ES-401 4
Form ES-401-1 ES-401 BWR Examination Outline Plant Systems - Tier 2/Group 1 (RO I SRO)
Form ES-401-1 System # I Name K
1 K
2 K
3 K
4 K
5 K
6 A
1 A
2 A
3 A
4 G
KIA Topic(s)
IR 203000 RHR/LPCI: Injection Mode X
A 2.10 Ability to (a) predict the impacts of the following on the RHR/LPCI:
INJECTION MODE (PLANT SPECIFIC) ;
and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 I 45.6) 3.31 3.5 28 Nuclear boiler instrument failures 205000 Shutdown Cooling X
K 3.03 Knowledge of the effect that a loss or malfunction of the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) will have on following:
(CFR:41.7/45.4) 3.81 3.9 29 K3.03 Reactor temperatures (moderator, vessel, flange) 206000 HPCI X 2.4.6 Knowledge of EOP mitigation strategies. (CFR: 41.10 I 43.5 I 45.13) 3.71 4.7 30 207000 Isolation (Emergency)
Condenser 209001 LPCS X
A4.01 Ability to manually operate and/or monitor in the control room:
(CFR: 41.7 I 45.5 to 45.8) 3.81 3.6 32 Core spray pump 209002 HPCS 211000 SLC X
K 1.05 Knowledge of the physical connections and/or cause effect relationships between STANDBY LIQUID CONTROL SYSTEM and the following:
(CFR: 41.2 to 41.9 I 45.7 to 45.8) 3.41 3.6 31 RWCU 212000 RPS X
K 6.02 Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR PROTECTION SYSTEM:
(CFR:41.7/45.7) 3.71 3.9 33 Nuclear instrumentation II
2150031RM X
K 3.04 Knowledge of the effect that a loss 3.6/
34 or malfunction of the INTERMEDIATE 3.6 RANGE MONITOR (IRM) SYSTEM will have on following: (CFR: 41.7/45.4)
Reactor power indication i
215004 Source Range Monitor X
2.6/
35 K 2.01 Knowledge of electrical power supplies to the following: (CFR: 41.7) 2.8 SRM channels/detectors 215005 APRM / LPRM X
K 1.02 Knowledge of the physical 3.7/
36 connections and/or cause effect 3.7 relationships between AVERAGE POWER RANGE MONITORILOCAL POWER RANGE MONITOR SYSTEM and the following: (CFR: 41.2 to 41.9/45.7 to 45.8)
K 5.01 Knowledge of the operational 2.6/
37 implications of the following concepts as 2.6 they apply to REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) :
(CFR: 41.5/45.3)
Indications of pump cavitation 218000 ADS X
X 3.9/
38/39 K 6.01 Knowledge of the effect that a loss or malfunction of the following will have on 4.1 the AUTOMATIC DEPRESSURIZATION SYSTEM: (CFR: 41.7 145.7)
RHRlLPCI system pressure: Plant-Specific 3.9/
2.1.27 Knowledge of system purpose 4.0 and/or function. (CFR: 41.7) 223002 PCIS/Nuclear Steam X
3.7/
40 K 3.07 Knowledge of the effect that a loss Supply Shutoff 3.8 or malfunction of the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF will have on following:
(CFR: 41.7/45.4)
Reactor pressure
'I 239002 SRVs 259002 Reactor Water Level X
Control 261000 SGTS 262001 AC Electrical Distribution 262002 UPS (AC/DC) 263000 DC Electrical Distribution
,I X
X K 3.01 Knowledge of the effect that a loss or malfunction of the RELIEF/SAFETY 3.9/
4.0 41/42 VALVES will have on following: (CFR: 41.7
/45.4)
Reactor pressure control A 2.03 Ability to (a) predict the impacts of the following on the RELIEF/SAFETY VALVES; and (b) based on those 4.1/
4.2 predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
(CFR: 41.5/45.6)
Stuck open SRV K 1.14 Knowledge of the physical connections and/or cause effect 2.9/
3.0 43 relationships between REACTOR WATER LEVEL CONTROL SYSTEM and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8)
X K 6.03 Knowledge of the effect that a loss or malfunction of the following will have on 3.0/
3.1 44/45 the STANDBY GAS TREATMENT SYSTEM: (CFR: 41.7/45.7)
Emergency diesel generator system A 4.06 Ability to manually operate and/or monitor in the control room:
3.3/
3.6 Reactor building differential pressure X
K 6.01 Knowledge of the effect that a loss or malfunction of the following will have on 3.1/
3.4 46 the A.C. ELECTRICAL DISTRIBUTION:
(CFR: 41.7/45.7)
D.C. power X
A 3.01 Ability to monitor automatic operations of the UNINTERRUPTABLE 2.8/
3.1 47 POWER SUPPLY (A.C'/D.C.) including:
(CFR: 41.7/45.7)
Transfer from preferred to alternate source X
4.6/
48 2.1.20 Ability to interpret and execute 4.6 procedure steps. (CFR: 41.10/43.5/
45.12)
264000 EDGs X
A 3.05 Ability to monitor automatic operations of the EMERGENCY GENERATORS (DIESEUJET) including:
(CFR: 41.7/45.7) load shedding and sequencing 3.41 3.5 49 300000 Instrument Air X
X K 2.01 Knowledge of electrical power supplies to the following: (CFR: 41.7)
Instrument air compressor 2.8/
2.8 SO/51 K 5.13 Knowledge of the operational implications of the following concepts as they apply to the INSTRUMENT AIR SYSTEM: (CFR: 41.5/ 45.3)
Filters 2.9/
2.9 400000 Component Cooling Water X X K 6.05 Knowledge of the effect that a loss or malfunction of the following will have on the CCWS: (CFR: 41.7 / 45.7)
Motors 2.8/
2.9 52/53 A 1.04 Ability to predict and lor monitor changes in parameters associated with operating the CCWS controls including:
(CFR: 41.5/45.5)
Surge Tank level 2.8/
2.8 KJA Category Point Totals:
3 21 31 1 2 5
1 Total:
ES-401 5
Form ES-401-1 I ES-401 BWR Examination Outline Form ES-401-1 System # I Name K
1 K
2 K
3 KIA Topic(s)
IR 201001 CRD Hydraulic X
K 1.03 Knowledge of the physical connections andlor cause effect relationships between CONTROL ROD DRIVE HYDRAULIC SYSTEM and the following: (CFR: 41.2 to 41.9
/45.7 to 45.8) 3.11 3.1 54 Recirculation pumps (seal purge): Plant-Specific 201002 RMCS X
K 4.05 Knowledge of REACTOR MANUAL CONTROL SYSTEM design feature(s) and/or interlocks which provide for the following: (CFR: 41.7)
"Notch override" rod withdrawal 3.31 33 55 201003 Control Rod and Drive Mechanism 201004 RSCS 201005 RCIS 201006 RWM 202001 Recirculation X
A 1.01 Ability to predict and/or monitor changes in parameters associated With operating the RECIRCULATION SYSTEM controls including: (CFR: 41.5/45.5)
Recirculation pump flow: Plant-Specific 3.61 3.5 56 202002 Recirculation Flow Control 204000 RWCU 214000 RPIS X
K 4.02 Knowledge of ROD POSITION INFORMATION SYSTEM design feature(s) and/or interlocks which provide for the following: (CFR: 41.7) 2.5/
2.5 57 Thermocouple 215001 Traversing In-core Probe X
K 1.05 Knowledge of the physical connections and/or cause effect relationships between TRAVERSING IN-CORE PROBE and the following: (CFR: 41.2 to 41.9 I 45.7 to 45.8)
Primary containment isolation system 3.31 3.4 58 215002 RBM 216000 Nuclear BOiler Insl.
X K 1.05 Knowledge of the physical connections and/or cause effect relationships between NUCLEAR BOILER INSTRUMENTATION and the following: (CFR: 412 to 41.91 45.7 to 45.8) 3.7/
3.9 59 Residual heat removal: Plant-Specific
219000 RHR/LPCI: Torus/Pool Cooling Mode 223001 Primary CTMT and AUK 226001 RHR/LPCI: CTMT Spray Mode 230000 RHRlLPCI: Torus/Pool Spray Mode 233000 Fuel Pool Cooling/Cleanup X
2.8/
60 K 2.02 Knowledge of electrical power supplies 2.9 to the following: (CFR: 41.7)
RHR pumps 234000 Fuel Handling Equipment X
3.1/
61 A 3.02 Ability to monitor automatic operations of the FUEL HANDLING EQUIPMENT
3.7 including
(CFR: 41.7/45.7)
Interlock operation 239001 Main and Reheat Steam X
2.8/
62 A 3.03 Ability to monitor automatic operations of the MAIN AND REHEAT STEAM SYSTEM
2.8 including
(CFR: 41.7 f 45.7)
Moisture separator reheat steam supply:
Plant-S pecific 239003 MSIV Leak§lf!e Control 241000 ReactoriTurbine Pressure Regulator 245000 Main Turbine Gen. I Aux.
I I
256000 Reactor Condensate X
2.81 63 K 604 Knowledge of the effect that a loss or 28 malfunction of the following will have on the REACTOR CONDENSATE SYSTEM: (CFR:
41.7/45.7)
A.C. power 259001 Reactor Feedwater 268000 Radwaste 271000 Offgas X
3.31 64 =
A 1.01 Ability to predict and/or monitor 3.2 changes in parameters associated with operating the OFFGAS SYSTEM controls including: (CFR: 41.5/45.5)
Condenser vacuum 272000 Radiation Monitoring 286000 Fire Protection X
3.61 65 K 3.03 Knowledge of the effect that a loss or malfunction of the FIRE PROTECTION 3.8 SYSTEM will have on following: (CFR: 41.7 {
45.4)
Plant protection 288000 Plant Ventilation 290001 Secondary CTMT 290003 Control Room HVAC 290002 Reactor Vessel Internals Rffi
~
I II 2
0 1
21 Point Total:
12
- Gtt:
ES-401 Generic Knowledge and Abilities Outline (Tier 3)
Form ES-401-3 Facil'
~"
Date of Exam:
3/8/12 Category KIA #
Topic RO/SRO SRO-Only IR IR 2.1.2 Knowledge of operator responsibilities during all modes of 4.1/4.4 66
- 1.
plant operation. (CFR: 41.10/45.13)
Conduct of Operations 2.1.1 Knowledge of conduct of operations requirements.
3.8/4.2 67 (CFR: 41.10/45.13) 2.1.45 4.3/4.3 68 Ability to identify and interpret diverse indications to validate the response of another indication.
(CFR: 41.7/43.5/45.4) I 2.1.
2.1.
2.1.
Subtotal 3
2.2.6 Knowledge of the process for making changes to 3.0/3.6 69 procedures.
- 2.
(CFR: 41.10 143.3/45.13)
Equipment Control 2.2.35 Ability to determine Technical Specification Mode of 3.6/4.5 70 Operation.
(CFR:41.7/41.10/43.2/45.13) 2.2.
2.2.
2.2.
2.2.
Subtotal 2
2.3.11 Ability to control radiation releases.
3.8/4.3 71 (CFR:41.11 143.4/45.10)
- 3.
I Radiation Control 2.3.13 Knowledge of radiological safety procedures pertaining to 3.4/3.8 72 I
licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.
(CFR: 41.12/43.4/45.9/45.10)
2.3.
2.3.
2.3.
2.3.
Subtotall 2.4.27
- 4.
Emergency Procedures 1 Plan 2.4.5 2.4.49 2.4.
2.4.
2.4.
Subtotal Tier 3 Point Total Knowledge of "fire in the plant" procedures.
(CFR: 41.10/43.5/45.13) I Knowledge of the organization of the operating procedures network for normal, abnormal, and emergency evolutions.
(CFR: 41.10/43.5/45.13)
Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.
(CFR: 41.10 143.2 145.6) 3.4/3.9 3.7/4.3 4.6/4.4 2
73 74 75 3
10
ES-401 2
Form ES-401-1 CC An BWR Examination Outline Emergency and Abnormal Plant Evolutions* Tier lIGroup 1 (RO I SRO)
EIAPE # I Name / Safety Function 295001 Partial or Complete Loss of Forced Core Flow Circulation 11 & 4 295003 Partial or Complete Loss of AC 16 295004 Partial or Total Loss of DC Pwr 16 295005 Main Turbine Generator Trip 1 3 295006 SCRAM / 1 295016 Control Room Abandonment /7 295018 Partial or Total Loss of CCW /8 29501 9 Partial or Total Loss of Insl. Air 18 295021 Loss of Shutdown Cooling 14 295023 Refueling Acc /8 295024 High Drywell Pressure 15 295025 High Reactor Pressure I 3 295026 Suppression Pool High Water Temp. /5 295027 High Containment Temperature / 5 295028 High Drywell Temperature I 5 295030 Low Suppression Pool Wtr LviI 5 K K K A A G
KIA Topic(s}
IR 1
2 3
1 2
X AA2.04 Ability to determine and/or interpret the 3.7 81 following as they apply to PARTIAL OR COMPLETE LOSS OF AC. POWER: (CFR: 41.10/43.5/45.13)
System lineups X
AA2.07 Ability to determine and/or interpret the 3.1 77 following as they apply to LOSS OF SHUTDOWN COOLING: (CFR: 41.10 J43.5 /45.13)
Reactor recirculation now m
X EA2.02 Ability to determine and/or interpret the ~
following as they apply to HIGH DRYWELL PRESSURE: (CFR: 4110! 43.5/45.13)
Drywell temperature X 2.1.7 Conduct of Operations 4.7 78 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. (CFR: 41.5/43.5/45.12/45.13)
X EA2.03 Ability to determine and/or interpret the 3.9 79 following as they apply to LOW SUPPRESSION POOL WATER LEVEL: (CFR 41.10/435 I 45.13)
Reactor pressure
'I Level/2 X
EA204 Ability to determine and/or interpret the following as they apply to REACTOR LOW WATER LEVEL: (CFR: 41.10143.5145.13) 4.8 76 Adequate core cooling 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown I 1 295038 High Off-site Release Rate I 9 X 2.4.4 Emergency Procedures / Plan 4.7 80 Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures (CFR: 41.10 I 43.2 I 45.6) 600000 Plant Fire On Site I 8 700000 Generator Voltage and Electric Grid Disturbances I 6 II KIA Category Totals:
5 2
Group Point Total:
7
ES-401 3
Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (RO / SRO)
E/APE # / Name / Safety Function K
K K A A G KIA Topic(s)
IR 1
2 3
1 2
295014 Inadvertent Reactivity Addition /1 295002 Loss of Main Condenser Vac / 3 295007 High Reactor Pressure / 3 295008 High Reactor Water Level / 2 295009 Low Reactor Water Level / 2 295010 High Drywell Pressure / 5 295011 High Containment Temp / 5 295012 High Drywell Temperature / 5 295013 High Suppression Pool Temp. / 5 295015 Incomplete SCRAM / 1 X
AA2.02 Ability to determine and/or interpret the 4.2 83 following as they apply to INCOMPLETE SCRAM:
(CFR 41.10/43.5/45.13)
Control rod position 295017 High Off-site Release Rate / 9 295020 Inadvertent Cont. Isolation / 5 &
295032 High Secondary Containment 7
295022 Loss of CRD Pumps / 1 295029 High Suppression Pool Wtr Lvi / 5 Area Temperature / 5 295033 High Secondary Containment X 2.1.32 Ability to explain and apply system limits and 4.0 85 Area Radiation Levels / 9 precautions. (CFR 41.10 I 43.2 / 45.12) 295034 Secondary Containment Ventilation High Radiation / 9 295035 Secondary Containment High Differential Pressure / 5 295036 Secondary Containment High Sump/Area Water Level / 5 500000 High CTMT Hydrogen Conc. / 5 X
3.8 84 EA2.03 Ability to determine and I or interpret the following as they apply to HIGH PRIMARY CONTAINMENT HYDROGEN CONCENTRATIONS (CFR 41.10 I 43.5 145.13)
Combustible limits for drywell KIA Category Point Totals:
2 1
Group Point Total:
3
ES-401 4
Form ES-401-1 ES-401 System # I Name 203000 RHR/LPCI: Injection Mode 205000 Shutdown Cooling 206000 HPCI 207000 Isolation (Emergency)
Condenser 209001 LPCS 209002 HPCS 211000SLC 212000 RPS 2150031RM 215004 Source Range Monitor 215005 APRM I LPRM 217000 RCIC 218000 ADS 223002 PCISINuclear Steam Supply Shutoff 239002 SRVs 259002 Reactor Water Level Control 261000 SGTS BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 1 (RO I SRO)
K K K K K K A A A A G KIA Topic(s)
IR 1
2 3
4 5
6 1
2 3
4 X
3.5 90 A2.17 Ability to (a) predict the impacts of the following on the RHRILPCI: INJECTION MODE (PLANT SPECIFIC) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 I 45.6)
Keep fill system failure X 2.4.31 Knowledge of annunciator alarms, 4.1 86 indications, or response procedures.
(CFR:41.10/45.3)
X A2.17 Ability to (a) predict the impacts of the 3.4 87 following on the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) ;
and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 I 45.6)
High suppression pool level II
II 262001 AC Electrical Distribution 262002 UPS (ACIDe) 263000 DC Electrical Distribution 264000 EDGs 300000 Instrument Air 400000 Component Cooling Water KJ X
3.2 88 A2.01 Ability to (a) predict the impacts of the followmg on the D.C. ELECTRICAL DISTRIBUTION; and (b) based on those predictions, use procedures to correct control, or mitigate the consequences of those abnormal conditions or operations:
(CFR 41.5/456)
Grounds I
X 2.2.44 Ability to interpret control room 4.4 89 indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.
(CFR: 41.5/43.5/45.12) 1:+/-:t+/-+/-:Q 5
II
ES-401 System # / Name 201001 CRD Hvdraulic 201002 RMCS 201003 Control Rod and Drive Mechanism 201004 RSCS 201005 RCIS 201006 RWM 202001 Recirculation 202002 Recirculation Flow Control 204000 RWCU 214000 RPIS 215001 Traversing In-core Probe 215002 RBM 216000 Nuclear Boiler Inst.
219000 RHR/LPCI: Torus/Pool Cooling Mode 223001 Prima~ CTMT and Aux.
226001 RHR/LPCI: CTMT Spray Mode 230000 RHR/LPCI Torus/Pool Spray Mode 233000 Fuel Pool Cooling/Cleanup 234000 Fuel Handlina Eauipment 239001 Main and Reheat Steam 239003 MSIV Leakalle Control 241000 ReactorlTurbine Pressure Reoulator 245000 Main Turbine Gen. I Aux.
256000 Reactor Condensate
~
sle Offaas 272000 Radiation Monitorina 286000 Fire Protection 288000 Plant Ventilation 290001 Secondary CTMT 5
Form ES-401-1 BWR Examination Outline Form ES-401-1
~"'m'-n" 21G,,", 2 IRO I SROI K
A A
A A
G KIA Topic(s)
IR 1
6 1
2 3
4 EE X
42 91 2.4.11 Knowledge of abnormal condition procedures.
(CFR: 41.10 i 43.5/4513) if X
4.8 92 2.4.1 Knowledge of EO? entry conditions and immediate action steps.
(CFR 41.1 0 i 43.51 45.13)
II 290003 Control Room HVAC X
A202 Ability to (e) predict the impacts of the following on the CONTROL ROOM HVAC :
and (b) based on those predictions. use procedures to correct, control. or mitigate the consequences of Ihose abnormal conditions or operations (CFR: 41.S/45.6) 3.4 93 Extreme environmental conditions 290002 Reactor Vessel Internals KIA Category Point Totals:
1 2
Group Point Total:
3
ES-401 Generic Knowledge and Abilities Outline (Tier 3)
Form ES-401-3 Facility:
James A. Fitzpatrick Date of Exam:
3/8/12 Category KIA #
2.1.34
- 1.
Conduct of Operations 2.1.35 2.1.
2.1.
2.1.
2.1.
Subtotal 2.2.40
- 2.
Equipment Control 2.2.
2.2.
2.2.
2.2.
2.2.
Subtotal 2.3.14
- 3.
Radiation Control 2.3.4 2.3.
2.3.
2.3.
2.3.
Subtotall Topic Knowledge of primary and secondary plant chemistry limits.
(CFR: 41.10/43.5/45.12)
IR RO SRO-Only IR 3.5 94 Knowledge of the fuel-handling responsibilities of SROs.
(CFR: 41.10/43.7) 3.'1 95 Ability to apply Technical Specifications for a system.
4.7 96 (CFR: 41.10/43.2/43.5/45.3)
Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.
(CFR: 41.12/43.4 145.10) 3.8 I
97 Knowledge of radiation exposure limits under normal or emergency conditions.
(CFR: 41.12/43.4 /45.10) 3.7 98 2
2.4.28 I 4.1 99 Knowledge of procedures relating to a security event (non-safeguards information).
- 4.
(CFR: 41.10 143.5/45.13)
Emergency Procedures I Plan 2.4.44 4.4 100 Knowledge of emergency plan protective action recommendations.
(CFR: 41.10/41.12/43.5/45.11) I 2.4.
2.4.
2.4.
2.4.
Subtotal 2
Tier 3 Point Total E
ES*401 Tier I Randomly Group Selected KIA Tier 1 /
295001 Group 1 AK 2.08 Tier 1 /
295001 Group 1 AK 2.05 Tier 1 /
295026 Group 1 EK 3.03 Tier 1 /
295027 Group 1 Tier 1 /
295011 Group 2 Tier 1 /
295007 Group 2 AK 2.01 Record of Rejected KlAs Form ES*401*4 Reason for Rejection JAF is a BWR-4 (K&A specific to BWR-1)
LPCI loop select logic not applicable at JAF K&A not applicable to JAF operating procedures for high torus temperature All K&A's applicable to Mark III containment only All K&A's applicable to Mark III containment only Rejected K&A AK 2.01 on basis of previously sampled K&A similarities. Similarities are:
- 012: EPE 295025 High Reactor Pressure, EK 3.08:
Reactor/turbine I2ressure regulating system ol2eration.
- 023: APE 295007: High Reactor Pressure, AK 2.01:
Reactor/turbine I2ressure regulating system ol2eration.
- Replaced AK 2.01 with randomly selected KA:
APE 295007, AK 2.06, PCIS/NSSSS The guidance for this rejection is stated in ES-401 Section D 1.d: ensure that no EPE/APE, system, or KIA category is over-sampled...
Page 1 of2
Tier 2/
207000 K&A's not applicable to JAF (Iso/Emergency Condensers)
Group 1 Tier 2/
209002 K&A's not applicable to JAF (HPCS)
Group 1 Tier 3 2.3.5 Could not write a discriminating, operationally oriented question.
Randomly selected 2.3.11.
Tier 2/
203000 A2.01 Rejected K&A A2.01 on basis of previously sampled K&A Group 1 similarities. Similarities are:
- Replaced with 203000 A2.01 with randomly selected 203000 A2.10 (NBI failures).
Page 2 of2
ES-301 Administrative Topics Outline Form ES-301-1 I Facility: James A. Fitzpatrick Examination Level: RO SRO Administrative Topic (see Note)
Conduct of Operations Conduct of Operations Equipment Control Radiation Control Emergency Procedures/Plan X
Type Code*
M,S M,R N,R D,R P,D Date of Examination: 2/27/12 Operating Test Number:
Describe activity to be performed Control Panel Walkdown Determine Shift Staffing ST-26K Recirc Loop Startup Differential Temperature Check Evaluate an RWP and Survey Map EAP-17 Pager and CAl\\! Activation During Security Event NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.
- Type Codes & Criteria:
(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (S 3 for ROs; S 4 for SROs & RO retakes)
(N)ew or (l\\t1)odified from bank (~ 1)
(P)revious 2 exams (S 1; randomly selected)
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: James A. Fitzpatrick Date of Examination: 2/27/12 Exam Level: RO SRO-I X SRO-U Operating Test No.:
Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)
System !,JPM Title I
Type Code* I Safety Function I
I I
I
- e. Transfer 10300 (10400) bus from T-4 to Reserve Transformer A/MIS 6
- f. Restore RB Ventilation following isolation DIS 5
- g.
In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)
I j. Vent Torus to lower Primary Containment pressure AlE/LIM 5
i k. Supply cooling water to EDG 'A' ! 'C' from 46P-2B D
8 All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
- Type Codes Criteria for RO I SRO-II SRO-U (A)lternate path 4-6 /4-6 I 2-3 (C)ontrol room
{D)irect from bank
~9/s8/S4 (E)mergency or abnormal in-plant
~1/~1/~1 (EN)gineered safety feature
- I -
I
~1 (control room system)
{L)ow-Power I Shutdown
~1/~1/~1 (N)ew or (M)odified from bank including 1 (A)
~2/~2/~1 (P)revious 2 exams
~ 3 I s 3 I s 2 (randomly selected)
.* (R)CA
~1f~1/~1 II (S)imulator
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: James A. Fitzpatrick Date of Examination: 2/27/12 Exam Level: RO SRO-I SRO-U X Operating Test No.:
Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)
System I JPM Title i a. Secure 'B' SBGT train following auto initiation
- b. Place HPCI in pressure control mode
- d.
- e.
- f.
i g.
- h.
Type Code*
Safety Function 01 LI S 9
AI EN I LIM 1S 4
D/E/LIS 2
In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)
I I i. Vent the Scram Air Header D/E/R 1
I j. Vent Torus to lower Primary Containment pressure AlE/LIM 5
All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
- Type Codes (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power 1Shutdown (N)ew or (M)odified from bank including 1 (A)
(P)revious 2 exams (R)CA (S)imulator Criteria for RO I SRO-I I SRO-U 4-6 I 4-6 I 2-3 S9/S8/S4
~1/~1/~1 1 ;:::1 (control room system)
- 1/;
- :1/;::1
- 2/;
- :2/~1 s 3 1s 3 1S 2 (randomly selected)
~1/~1/~1
Appendix D Scenario Outline Form ES-D-1 Facility:
Fitzpatrick Scenario No.: _2_
Op-Test No.:
Examiners:
Operators:
Initial Conditions: Reactor at 75% power Turnover: Reactor is at 75% power. No equipment is OOS. A rod sequence exchange is in progress.
There are 3 rods that are to moved from position 00 to position 48. After the exchange is completed, reactor power is to be returned to 100% using Recirc.
Event Malf.
Event Event No.
No.
Type*
Description 1
NA N - ATC, 2 Rods are moved to position 48 and coupling check performed.
SRO 2
RD07 C - ATC, 3rd Rod moved to position 48 but drifts in and then driven full in. (AOP-27)
SRO Reactor power reduced with recirc to < 25% pre-transient or 42.4 mlbm/hr core R
- BOP, flow.
RMOI:29 ODCM-Reactor Building Rad Monitor Fails downscale SRO 4
DG06A C
All Loss of 10500 Bus with slow start of EDG "An and failure of EDG "C" to start.
I DGOIC TS" SRO RPS is lost and restored with MG set or Alternate power supply. The Y2 scram reset. (AOP 18 and AOP 59) is :
I ZDILLHO NC05 TRIP 5
MS02 C
- BOP, Small leak in drywell. Torus is vented. Reactor scram is attempted prior to SRO reaching 2.7 psig. (AOP-39) 6 RPO IAlB M-All High power A TWS, injection is terminated and prevented. Level lowered to RP09 110" to prevent power oscillations. (EOP-3)
RDIO:AlI 7
SLOIA/B C
I SL05A/B 8
RDB C
- ATC, Removal ofRPS fuses partially works. The remainder of the eontrol rods will be SRO inserted by inserting another manual scram or driving the rods in.
E(N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor
Appendix 0 Scenario Outline Form ES-D-1 Facility: __ Fitzpatrick Scenario No.:
3 Op-Test No.:
Examiners:
Operators:
Initial Conditions: 68% power.
Turnover: The plant is shutting down for a refueling outage. "9" CRD is tagged out for maintenance.
Continue to cold shutdown lAW OP-65.
I Event Malf.
Event Event No.
No.
Type*
Description 1
NA R-ATC, Lower power with recire to 65% with control rods.
SRO i
2 NA N
AN932:01 TRM-Core Spray Pipe Break Detector Alarm SRO 4
AD05 C-BOP, SR Y fails open, then closes when fuses pulled (AOP-36).
ED43A TR.\\1-SRO Loss of 115kv Line # 3. (AOP-72) 6 EGO!
M-All Generator Trip i
i I 7 RPOIA/8 C - ATC, RPS fails to scram reactor ARI works (EOP-2).
SRO 11 8
ED44 C - ATC, Loss of Offsite Power - Manually close MSIYs due to loss of Circulating Water SRO 9
RRI5 C-80P, Coolant leak in drywell requiring initiation of drywell sprays SRO I
10 HPOI C-BOP, HPCI fails to automatically initiate. After manual initiation, HPCI trips.
SRO HP02 II RRI5 M - All Leak greater than capacity of RCIC and CRD. Level lowers, Alternate Level Control entered and Emergency Depressurize at -19". Level to be restored i
.g;reater than T AF with low pressure systems.
I i
(N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor