ML23122A098
| ML23122A098 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 05/02/2023 |
| From: | Brian Fuller NRC/RGN-I/DORS/OB |
| To: | Dougherty M Constellation Energy Generation |
| References | |
| Download: ML23122A098 (1) | |
Text
Form 3.2-1 Administrative Topics Outline Facility: Nine Mile Point Unit 1 Date of Examination: Jan 2023 Examination Level:
RO Operating Test Number: 2023 Administrative Topic (Step 1)
Activity and Associated K/A (Step 2)
Type Code (Step 3)
Conduct of Operations Perform Daily Checks (Partial - Rad Monitors)
K/A 2.1.19 (3.9), N1-ST-D0 M, S Conduct of Operations Perform Heat-up Rate Determination K/A 2.1.7 (4.4), N1-OP-43A D, R Equipment Control Perform Daily Thermal Limit Surveillance K/A 2.2.12 (3.7), N1-RESP-1A D, R Radiation Control Application of Radiation Exposure Limits -
SDC Room K/A 2.3.12 (3.2), RP-AA-203 P, D, R 2018 NRC Emergency Plan
Location:
(C)ontrol room, (S)imulator, or Class(R)oom Source and Source Criteria:
(P)revious two NRC exams (no more than one JPM that is randomly selected from last two NRC exams)
(D)irect from bank (no more than three for ROs, no more than four for SROs and RO retakes)
(N)ew or Significantly (M)odified from bank (no fewer than one)
Form 3.2-1 Administrative Topics Outline Facility: Nine Mile Point Unit 1 Date of Examination: Jan 2023 Examination Level:
SRO Operating Test Number: 2023 Administrative Topic (Step 1)
Activity and Associated K/A (Step 2)
Type Code (Step 3)
Conduct of Operations Review Daily Checks (Partial - Rad Monitors)
K/A 2.1.19 (3.8), N1-ST-D0 N, R Conduct of Operations Perform Time to Boil Calculation for Reactor Coolant System K/A 2.1.40 (3.9), OP-NM-108-117-1002 P, D, R 2018 NRC Equipment Control Perform Daily Thermal Limit Surveillance K/A 2.2.12 (4.1), N1-RESP-1A D, R Radiation Control Application of Radiation Exposure Limits -
SDC Room K/A 2.3.12 (3.7), RP-AA-203 M, R Emergency Plan Determine Protective Action Recommendations and Complete Part 1 Notification Fact Sheet K/A 2.4.44 (4.4), EP-CE-111, EP-CE-114 M, R
Location:
(C)ontrol room, (S)imulator, or Class(R)oom Source and Source Criteria:
(P)revious two NRC exams (no more than one JPM that is randomly selected from last two NRC exams)
(D)irect from bank (no more than three for ROs, no more than four for SROs and RO retakes)
(N)ew or Significantly (M)odified from bank (no fewer than one)
Form 3.2-2 Control Room/In-Plant Systems Outline Facility: Nine Mile Point Unit 1 Date of Examination: Jan 2023 Operating Test Number: 2023 Exam Level: RO / SRO-I System/JPM Title Type Code Safety Function Control Room Systems
- a. Control Rod Exercising Operability Test, Loss of RPIS K/A 214000 A2.01 (3.3/3.3), N1-ST-W1, N1-OP-5 P, D, A, S 2018 NRC 7
- c. Start a Reactor Recirculation Pump, Low Frequency K/A 202001 A4.01 (4.0), N1-OP-1 M, A, S, L 4
- d. Rapid RWCU System Restoration for Level Control, RWCU Steam Leak in Reactor Building K/A 204000 A4.02 (3.3), N1-EOP-HC M, A, S, L 2
- e. Verify Auto-Start of RBEVS and Secure One Train K/A 261000 A4.03 (3.7), N1-OP-10 N, S, EN 9
- f. Synchronize Main Generator to the Grid, Main Generator Locks Out K/A 262001 A4.04 (3.9), N1-OP-32, N1-SOP-31.1 D, A, S 6
- g. Vent the Drywell K/A 223001 A4.03 (3.1), N1-OP-9 D, S, L 5
K/A 239001 A4.01 (4.2/4.0), N1-EOP-HC D, S, L 3
In-Plant Systems
- i. Swap CRD Flow Control Valves K/A 201001 A2.07 (4.0/3.6)
D, R 1
- j. Diesel Fire Pump Start with No Control Power K/A 286000 A2.02 (2.6/2.8)
D, E 8
- k. Remove ERV Fuses in the Plant K/A 239002 A2.03 (4.6/4.4), N1-SOP-1.4 D, E, R 3
Code License Level Criteria RO SRO-I SRO-U (A)lternate path 4-6 4-6 2-3 (C)ontrol room (D)irect from bank 9
8 4
(E)mergency or abnormal in-plant 1
1 1
(EN)gineered safety feature (for control room system) 1 1
1 (L)ow power/shutdown 1
1 1
(N)ew or (M)odified from bank (must apply to at least one alternate path JPM) 2 2
1 (P)revious two exams (randomly selected) 3 3
2 (R)adiologically controlled area 1
1 1
(S)imulator
Form 3.3-1 Scenario Outline
Facility:
Nine Mile Point Unit 1 Scenario #:
NRC-1 Scenario Source:
New Op. Test #:
2023 Examiners:
Applicants/
Operators:
Initial Conditions:
The plant is operating at approximately 85% power. RBCLC pump 12 is out of service for maintenance.
Turnover:
Swap Service Water pumps per N1-OP-18 section F.2. The procedure is in progress up to step F.2.2.b. Then, perform a control rod pattern adjustment per the provided instructions.
Critical Tasks:
CT-1: Given a failure to scram with Reactor power above 6%,
the crew will lower Reactor power by one or more of the following methods, in accordance with N1-EOP-C5:
Terminating and preventing all injection into the RPV except boron and CRD Tripping all Recirc pumps Injecting boron The Reactor power reduction must be initiated within ten minutes of the start of the failure to scram.
CT-2: Given a failure to scram, the crew will initiate control rod insertion, in accordance with N1-EOP-C5. All control rods must be inserted within one hour of the start of the failure to scram.
Event No.
Malf.
No.
Event Type*
Event Description 1
N/A N - BOP, SRO Swap Service Water Pumps N1-OP-18 2
N/A R - ATC, SRO Perform a Control Rod Pattern Adjustment N1-OP-5 3
SRO Control Rod Drive Pump Trip N1-SOP-5.1, Technical Specifications 4
ATC TS -
SRO EPR Oscillation N1-SOP-31.2, Technical Specifications 5
TU02E TU02F C - All Main Turbine Vibrations N1-ARP-A2-3-5, N1-SOP-1, N1-EOP-RPV 6
RP05A RP09 M - All MC -
- ATC, BOP RPS and ARI Fail to Insert Control Rods N1-EOP-C5 7
LP01 C - All First Liquid Poison Pump Trips After Start N1-EOP-C5 8
TC01 C - All Main Turbine Inadvertently Trips N1-EOP-C5
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Tech Spec, (MC)Manual Control
Facility: Nine Mile Point Unit 1 Scenario No.: NRC-1 Op-Test No.: 2023
- 1. Malfunctions after EOP entry (1-2)
Events 7 & 8 2
- 2. Abnormal events (2-4)
Events 3, 4, 5 3
- 3. Major transients (1-2)
Event 6 1
- 4. EOPs entered/requiring substantive actions (1-2)
EOP-RPV 1
- 5. Entry into a contingency EOP with substantive actions (1 per scenario set)
EOP-C5 1
- 6. Pre-identified critical tasks (2) 2
Form 3.3-1 Scenario Outline
Facility:
Nine Mile Point Unit 1 Scenario #:
NRC-3 Scenario Source:
New Op. Test #:
2023 Examiners:
Applicants/
Operators:
Initial Conditions:
The plant is operating at approximately 100% power. RBCLC pump 12 is out of service for maintenance.
Turnover:
Shift Powerboard 101 from R1014 to R1011 in accordance with N1-OP-30 section H.8. Then, lower Reactor power to 85%
using Recirculation flow in accordance with the provided instructions.
Critical Tasks:
CT-1: Given high Drywell pressure, a failure of the Reactor Protection System to process an automatic scram, and failure of the manual scram pushbuttons, the crew will manually scram the Reactor, in accordance with N1-SOP-1 and/or N1-EOP-RPV. The scram must be initiated within 5 minutes of Drywell pressure exceeding the automatic scram setpoint.
CT-2: Given an un-isolable Torus water leak and the inability to maintain Torus water level above 8.0, the crew will perform an RPV Blowdown, in accordance with N1-EOP-PC. The depressurization must be initiated before Torus water level lowers below 8.0.
Event No.
Malf.
No.
Event Type*
Event Description 1
N/A N - BOP, SRO Swap Powerboard 101 Supply N1-OP-30 2
N/A R - ATC, SRO Lower Reactor Power with Recirculation Flow N1-OP-43B 3
SRO Core Spray Pump 111 Suction Leak N1-EOP-SC, Technical Specifications 4
SRO Reactor Recirculation Pump 13 Trip N1-SOP-1.3, Technical Specifications 5
PC05 MS04 C - All MC -
Override RP04A/
ATC RPS Pushbuttons Fail, RPS Fails to Process Automatic Scram N1-SOP-1, N1-EOP-RPV 7
PC04 M - All Torus Leak N1-EOP-PC, N1-EOP-C2 8
Override C - All Bypass Opening Jack Fails N1-EOP-RPV
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Tech Spec, (MC)Manual Control
Facility: Nine Mile Point Unit 1 Scenario No.: NRC-3 Op-Test No.: 2023
- 1. Malfunctions after EOP entry (1-2)
Events 6 & 8 2
- 2. Abnormal events (2-4)
Events 3, 4, 5 3
- 3. Major transients (1-2)
Event 7 1
- 4. EOPs entered/requiring substantive actions (1-2)
- 5. Entry into a contingency EOP with substantive actions (1 per scenario set)
EOP-C2 1
- 6. Pre-identified critical tasks (2) 2
Form 3.3-1 Scenario Outline
Facility:
Nine Mile Point Unit 1 Scenario #:
NRC-4 Scenario Source:
New Op. Test #:
2023 Examiners:
Applicants/
Operators:
Initial Conditions:
The plant is operating at approximately 3% power during a startup. IRM 11 is bypassed. RBCLC pump 12 is out of service for maintenance.
Turnover:
Return IRM 11 to service in accordance with N1-OP-38B section H.2. Then, continue control rod withdrawals.
Critical Tasks:
CT-1: Given lowering Circulating Water intake level, the crew will start at least one Emergency Service Water pump, in accordance with N1-SOP-18.1. The Emergency Service Water pump must be started within 15 minutes of Circulating Water intake level lowering below 238.8 (Annunciator H2-1-3).
CT-2: Given a coolant leak inside the Containment, lowering Reactor water level, and failure of injection systems to automatically inject to the Reactor, the crew will inject to the Reactor to restore and maintain Reactor water level above -84, in accordance with N1-EOP-RPV. Reactor water level must not be below -84 for more than 10 minutes.
Event No.
Malf.
No.
Event Type*
Event Description 1
N/A N - BOP, SRO Return IRM to Service N1-OP-38B 2
N/A R - ATC, SRO Raise Reactor Power with Control Rods N1-OP-43A, N1-OP-5 3
RD02 I - ATC, SRO Control Rod Drift Out N1-SOP-5.2 4
ED02A C - SRO TS -
SRO Loss of Line 1 N1-ARP-A8-1-1, N1-OP-33A, Technical Specifications 5
SRO Emergency Condenser Tube Leak N1-ARP-K1-1-2, Technical Specifications 6
CW17 C - All Intake Structure Clogging N1-SOP-18.1, N1-SOP-1, N1-EOP-RPV 7
RR29 M - All MC -
BOP HPCI Fails to Automatically Initiate, Core Spray Pumps Fail to Automatically Start N1-EOP-RPV
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Tech Spec, (MC)Manual Control
Facility: Nine Mile Point Unit 1 Scenario No.: NRC-4 Op-Test No.: 2023
- 1. Malfunctions after EOP entry (1-2)
Event 8 1
- 2. Abnormal events (2-4)
Events 3, 5, 6 3
- 3. Major transients (1-2)
Event 7 1
- 4. EOPs entered/requiring substantive actions (1-2)
EOP-RPV 1
- 5. Entry into a contingency EOP with substantive actions (1 per scenario set) 0
- 6. Pre-identified critical tasks (2) 2
Form 4.1-BWR Boiling-Water Reactor Examination Outline Facility:
Nine Mile Point 1 K/A Catalog Rev. 3 Rev.
2 Date of Exam:
01/30/2023 Tier Group RO K/A Category Points SRO-Only Points K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
Total A2 G*
Total
- 1.
Emergency and Abnormal Plant Evolutions 1
3 3
4 3
4 3
20 3
4 7
2 1
1 1
1 1
1 6
2 1
3 Tier Totals 4
4 5
4 5
4 26 5
5 10
- 2.
Plant Systems 1
3 2
1 2
3 2
2 3
2 4
2 26 2
3 5
2 1
1 2
1 1
1 1
1 0
1 1
11 1
1 1
3 Tier Totals 4
3 3
3 4
3 3
4 2
5 3
37 4
4 8
- 3.
Generic Knowledge and Abilities Categories CO EC RC EM 6
CO EC RC EM 7
2 2
1 1
2 2
1 2
- 4. Theory Reactor Theory Thermodynamics 6
3 3
Notes: CO =
EM =
Conduct of Operations; EC = Equipment Control; RC = Radiation Control; Emergency Procedures/Plan These systems/evolutions may be eliminated from the sample when Revision 2 of the K/A catalog is used to develop the sample plan.
These systems/evolutions are only included as part of the sample (as applicable to the facility) when Revision 2 of the K/A catalog is used to develop the sample plan.
ES-4.1-BWR BWR Examination Outline (Nine Mile Point 1)
Emergency and Abnormal Plant EvolutionsTier 1/Group 1 (RO/SRO)
Item E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G*
K/A Topic(s)
IR Q#
1 (295001) (APE 1)
PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION X
(295001) (APE 1) PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION (G2.2.23) Ability to track technical specification limiting conditions for operation (CFR: 41.10 / 43.2 /
45.13) 3.1 1
2 (295003) (APE 3)
PARTIAL OR COMPLETE LOSS OF AC POWER X
(295003) (APE 3) PARTIAL OR COMPLETE LOSS OF AC POWER (G2.4.32) Knowledge of operator response to loss of annunciators (CFR: 41.10 / 43.5 / 45.13) 3.6 2
3 (295004) (APE 4)
PARTIAL OR COMPLETE LOSS OF DC POWER X
(295004AK1.06) Knowledge of the operational implications and/or cause and effect relationships of the following concepts as they apply to the (APE 4) PARTIAL OR COMPLETE LOSS OF DC POWER: (CFR: 41.5 / 41.7 / 45.7 / 45.8) Prevention of inadvertent system(s) actuation upon restoration of DC power 3.7 3
4 (295005) (APE 5) MAIN TURBINE GENERATOR TRIP X
(295005AK1.02) Knowledge of the operational implications and/or cause and effect relationships of the following concepts as they apply to the (APE 5) MAIN TURBINE GENERATOR TRIP: (CFR:
41.5 / 41.7 / 45.7 / 45.8) Core thermal limits 3.8 4
5 (295006) (APE 6)
SCRAM X
(295006) (APE 6) SCRAM (G2.4.50) Ability to verify system alarm setpoints and operate controls identified in the alarm response procedure (CFR: 41.10 / 43.5 / 45.3) 4.2 5
6 (295016) (APE 16)
CONTROL ROOM ABANDONMENT X
(295016AA2.04) Ability to determine or interpret the following as they apply to (APE 16) CONTROL ROOM ABANDONMENT: (CFR:
41.10 / 43.5 / 45.13) Suppression pool temperature 4.0 6
7 (295018) (APE 18)
PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER (CCW)
X (295018AA2.06) Ability to determine or interpret the following as they apply to (APE 18) PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER (CCW): (CFR: 41.10 / 43.5 /
45.13) Surge tank level 3.6 7
8 (295019) (APE 19)
PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR X
(295019AK2.09) Knowledge of the relationship between the (APE
- 19) PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR and the following systems or components: (CFR: 41.8 / 41.10 / 45.3)
Primary containment and auxiliaries 3.4 8
9 (295021) (APE 21)
LOSS OF SHUTDOWN COOLING X
(295021AK3.05) Knowledge of the reasons for the following responses or actions as they apply to (APE 21) LOSS OF SHUTDOWN COOLING: (CFR: 41.5 / 41.10 / 45.6 / 45.13)
Establishing alternate heat removal flow paths 4.0 9
10 (295023) (APE 23)
REFUELING ACCIDENTS X
(295023AK2.10) Knowledge of the relationship between the (APE
- 23) REFUELING ACCIDENTS and the following systems or components: (CFR: 41.8 / 41.10 / 45.3) Nuclear instrumentation 3.5 10 11 (295024) (EPE 1) HIGH DRYWELL PRESSURE X
(295024EK3.08) Knowledge of the reasons for the following responses or actions as they apply to (EPE 1) HIGH DRYWELL PRESSURE: (CFR: 41.5 / 41.10 / 45.6 / 45.13) Containment spray 4.3 11 12 (295025) (EPE 2) HIGH REACTOR PRESSURE X
(295025EK2.12) Knowledge of the relationship between the (EPE
- 2) HIGH REACTOR PRESSURE and the following systems or components: (CFR: 41.8 / 41.10 / 45.3) Main and reheat steam system 3.1 12 13 (295026) (EPE 3)
SUPPRESSION POOL HIGH WATER TEMPERATURE X
(295026EA1.08) Ability to operate or monitor the following as they apply to (EPE 3) SUPPRESSION POOL HIGH WATER TEMPERATURE: (CFR: 41.5 / 41.7 / 45.5 to 45.8) LPCS 3.8 13 14 (295028) (EPE 5) HIGH DRYWELL TEMPERATURE (MARK I AND MARK II ONLY)
X (295028EA1.02) Ability to operate or monitor the following as they apply to (EPE 5) HIGH DRYWELL TEMPERATURE (MARK I AND MARK II ONLY): (CFR: 41.5 / 41.7 / 45.5 to 45.8) Drywell ventilation system 3.6 14 15 (295030) (EPE 7) LOW SUPPRESSION POOL WATER LEVEL X
(295030EA2.01) Ability to determine or interpret the following as they apply to (EPE 7) LOW SUPPRESSION POOL WATER LEVEL: (CFR: 41.10 / 43.5 / 45.13) Suppression pool level 4.1 15
16 (295031) (EPE 8)
REACTOR LOW WATER LEVEL X
(295031EA1.11) Ability to operate or monitor the following as they apply to (EPE 8) REACTOR LOW WATER LEVEL: (CFR: 41.5 /
41.7 / 45.5 to 45.8) Condensate system 3.6 16 17 (295037) (EPE 14)
SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN X
(295037EK3.06) Knowledge of the reasons for the following responses or actions as they apply to (EPE 14) SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: (CFR: 41.5 / 41.10 / 45.6 / 45.13)
Maintaining heat sinks external to the containment 4.1 17 18 (295038) (EPE 15)
HIGH OFFSITE RADIOACTIVITY RELEASE RATE X
(295038EK3.05) Knowledge of the reasons for the following responses or actions as they apply to (EPE 15) HIGH OFFSITE RADIOACTIVITY RELEASE RATE: (CFR: 41.5 / 41.10 / 45.6 /
45.13) Reactor shutdown/SCRAM 4.1 18 19 (600000) (APE 24)
PLANT FIRE ON SITE X
(600000AA2.06) Ability to determine or interpret the following as they apply to (APE 24) PLANT FIRE ON SITE: (CFR: 41.10 / 43.5 /
45.13) Need for pressurizing control room (recirculating mode) 3.4 19 20 (700000) (APE 25)
GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES X
(700000AK1.05) Knowledge of the operational implications and/or cause and effect relationships of the following concepts as they apply to the (APE 25) GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES: (CFR: 41.5 / 41.7 / 45.7 / 45.8) Voltage disturbance 3.5 20 21 (295001) (APE 1)
PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION X
(295001AA2.12) Ability to determine or interpret the following as they apply to (APE 1) PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: (CFR: 41.10 / 43.5 /
45.13) Thermal-hydraulic instabilities 4.2 76 22 (295005) (APE 5) MAIN TURBINE GENERATOR TRIP X
(295005) (APE 5) MAIN TURBINE GENERATOR TRIP (G2.4.45)
Ability to prioritize and interpret the significance of each annunciator or alarm 4.3 77 23 (295006) (APE 6)
SCRAM X
(295006) (APE 6) SCRAM (G2.2.44) Ability to interpret control room indications to verify the status and operation of a system and understand how operator actions and directives affect plant and system conditions 4.4 78 24 (295019) (APE 19)
PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR X
(295019) (APE 19) PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR (G2.1.20) Ability to interpret and execute procedure steps 4.6 79 25 (295024) (EPE 1) HIGH DRYWELL PRESSURE X
(295024EA2.12) Ability to determine or interpret the following as they apply to (EPE 1) HIGH DRYWELL PRESSURE: (CFR:
41.10 / 43.5 / 45.13) Safety/relief valves 3.7 80 26 (295028) (EPE 5) HIGH DRYWELL TEMPERATURE (MARK I AND MARK II ONLY)
X (295028EA2.02) Ability to determine or interpret the following as they apply to (EPE 5) HIGH DRYWELL TEMPERATURE (MARK I AND MARK II ONLY): (CFR: 41.10 / 43.5 / 45.13)
Reactor pressure 3.7 81 27 (295030) (EPE 7) LOW SUPPRESSION POOL WATER LEVEL X
(295030) (EPE 7) LOW SUPPRESSION POOL WATER LEVEL (G2.4.16) Knowledge of emergency and abnormal operating procedures implementation hierarchy and coordination with other support procedures or guidelines such as, operating procedures, abnormal operating procedures, or severe accident management guidelines.
4.4 82 K/A Category Totals:
3 3
4 3
7 7
Group Point Total:
27
ES-4.1-BWR BWR Examination Outline (Nine Mile Point 1)
Emergency and Abnormal Plant EvolutionsTier 1/Group 2 (RO/SRO)
Item E/APE # / Name / Safety Function K1 K2 K3 A1 A2 G*
K/A Topic(s)
IR Q#
28 (295002) (APE 2) LOSS OF MAIN CONDENSER VACUUM X
(295002AA1.05) Ability to operate or monitor the following as they apply to (APE 2) LOSS OF MAIN CONDENSER VACUUM: (CFR:
41.5 / 41.7 / 45.5 to 45.8) Main turbine generator and auxiliaries system 3.4 21 29 (295008) (APE 8) HIGH REACTOR WATER LEVEL X
(295008) (APE 8) HIGH REACTOR WATER LEVEL (G2.4.18)
Knowledge of the specific bases for emergency and abnormal operating procedures 3.3 22 30 (295013) (APE 13)
HIGH SUPPRESSION POOL TEMPERATURE.
X (295013AA2.02) Ability to determine or interpret the following as they apply to (APE 13) HIGH SUPPRESSION POOL TEMPERATURE.: (CFR: 41.10 / 43.5 / 45.13) Localized heating/stratification 3.5 23 31 (295020) (APE 20)
INADVERTENT CONTAINMENT ISOLATION & 7 X
(295020AK1.06) Knowledge of the operational implications and/or cause and effect relationships of the following concepts as they apply to the (APE 20) INADVERTENT CONTAINMENT ISOLATION & 7: (CFR: 41.5 / 41.7 / 45.7 / 45.8) Loss of reactor building HVAC 3.3 24 32 (295022) (APE 22)
LOSS OF CONTROL ROD DRIVE PUMPS X
(295022AK2.09) Knowledge of the relationship between the (APE
- 22) LOSS OF CONTROL ROD DRIVE PUMPS and the following systems or components: (CFR: 41.8 / 41.10 / 45.3) NBI 3.1 25 33 (500000) (EPE 16)
HIGH CONTAINMENT HYDROGEN CONCENTRATION X
(500000EK3.08) Knowledge of the reasons for the following responses or actions as they apply to (EPE 16) HIGH CONTAINMENT HYDROGEN CONCENTRATION: (CFR: 41.5 /
41.10 / 45.6 / 45.13) Operation of drywell nitrogen purge system 3.6 26 34 (295009) (APE 9) LOW REACTOR WATER LEVEL X
(295009) (APE 9) LOW REACTOR WATER LEVEL (G2.4.21)
Knowledge of the parameters and logic used to assess the status of emergency operating procedures critical safety functions or shutdown critical safety functions 4.6 83 35 (295017) (APE 17)
ABNORMAL OFFSITE RELEASE RATE X
(295017AA2.01) Ability to determine or interpret the following as they apply to (APE 17) ABNORMAL OFFSITE RELEASE RATE: (CFR: 41.10 / 43.5 / 45.13) Offsite release rate 4.2 84 36 (295035) (EPE 12)
SECONDARY CONTAINMENT HIGH DIFFERENTIAL PRESSURE X
(295035EA2.02) Ability to determine or interpret the following as they apply to (EPE 12) SECONDARY CONTAINMENT HIGH DIFFERENTIAL PRESSURE: (CFR: 41.10 / 43.5 / 45.13)
Radiation release rate 3.9 85 (295007) (APE 7) HIGH REACTOR PRESSURE
/ 3 (295010) (APE 10)
HIGH DRYWELL PRESSURE / 5 (295012) (APE 12)
HIGH DRYWELL TEMPERATURE / 5 (295014) (APE 14)
INADVERTENT REACTIVITY ADDITION
/ 1 (295029) (EPE 6) HIGH SUPPRESSION POOL WATER LEVEL / 5 (295032) (EPE 9) HIGH SECONDARY CONTAINMENT AREA TEMPERATURE / 5 (295033) (EPE 10)
HIGH SECONDARY CONTAINMENT AREA RADIATION LEVELS / 9
(295034) (EPE 11)
SECONDARY CONTAINMENT VENTILATION HIGH RADIATION / 9 (295036) (EPE 13)
SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL / 5 K/A Category Totals:
1 1
1 1
3 2
Group Point Total:
9
ES-4.1-BWR BWR Examination Outline (Nine Mile Point 1)
Emergency and Abnormal Plant EvolutionsTier 2/Group 1 (RO/SRO)
Item System / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
K/A Topic(s)
IR Q#
37 (205000) (SF4 SCS)
SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE)
X (205000K5.02) Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the (SF4 SCS) SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE):
(CFR: 41.5 / 45.3) Valve operation 3.5 27 38 (205000) (SF4 SCS)
SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE)
X (205000K5.03) Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the (SF4 SCS) SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE):
(CFR: 41.5 / 45.3) Decay heat removal 3.9 28 39 (206000) (SF2, SF4 HPCI) HIGH PRESSURE COOLANT INJECTION SYSTEM X
(206000A4.02) Ability to manually operate and/or monitor the (SF2, SF4 HPCI) HIGH PRESSURE COOLANT INJECTION SYSTEM in the control room: (CFR: 41.7 / 45.5 to 45.8) Flow controller 4.3 29 40 (207000) (SF4 IC)
ISOLATION (EMERGENCY)
CONDENSER X
(207000K6.09) Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the (SF4 IC) ISOLATION (EMERGENCY) CONDENSER:
(CFR: 41.7 / 45.7) Reactor protection system 3.3 30 41 (209001) (SF2, SF4 LPCS) LOW PRESSURE CORE SPRAY SYSTEM X
(209001A3.01) Ability to monitor automatic operation of the (SF2, SF4 LPCS) LOW PRESSURE CORE SPRAY SYSTEM including:
(CFR: 41.7 / 45.7) Valve operation 4.2 31 42 (211000) (SF1 SLCS)
STANDBY LIQUID CONTROL SYSTEM X
(211000K4.02) Knowledge of (SF1 SLCS) STANDBY LIQUID CONTROL SYSTEM design features and/or interlocks that provide for the following: (CFR:
41.7) Component and system testing 3.0 32 43 (212000) (SF7 RPS)
(212000A2.09) Ability to (a) predict the impacts of the following on the (SF7 RPS) REACTOR PROTECTION SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: (CFR: 41.5 /
45.6) High containment/drywell pressure 4.3 33 44 (215003) (SF7 IRM)
INTERMEDIATE RANGE MONITOR SYSTEM X
(215003K2.01) (SF7 IRM)
INTERMEDIATE RANGE MONITOR SYSTEM Knowledge of electrical power supplies to the following: (CFR: 41.7) IRM channels/detectors 3.4 34
45 (215004) (SF7 SRMS)
SOURCE RANGE MONITOR SYSTEM X
(215004A2.05) Ability to (a) predict the impacts of the following on the (SF7 SRMS) SOURCE RANGE MONITOR SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: (CFR: 41.5 /
45.6) Faulty or erratic operation of detectors/system 3.7 35 46 (215004) (SF7 SRMS)
SOURCE RANGE MONITOR SYSTEM X
(215004K6.01) Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the (SF7 SRMS) SOURCE RANGE MONITOR SYSTEM: (CFR: 41.7 /
45.7) RPS 3.3 36 47 (215005) (SF7 PRMS)
AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR X
(215005A4.04) Ability to manually operate and/or monitor the (SF7 PRMS) AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR in the control room:
(CFR: 41.7 / 45.5 to 45.8) LPRM back panel switches, and indicating lights 3.4 37 48 (218000) (SF3 ADS)
AUTOMATIC DEPRESSURIZATION SYSTEM X
(218000K1.02) Knowledge of the physical connections and/or cause and effect relationships between the (SF3 ADS) AUTOMATIC DEPRESSURIZATION SYSTEM and the following systems: (CFR:
41.2 to 41.9 / 45.7 to 45.8) LPCS system 4.2 38 49 (510000) Service Water System X
(510000A1.03) Ability to predict and/or monitor changes in parameters associated with operation of the Service Water System, including: Service water pressures 3.6 39 50 (223002) (SF5 PCIS)
PRIMARY CONTAINMENT ISOLATION SYSTEM /
NUCLEAR STEAM SUPPLY SHUTOFF X
(223002A4.08) Ability to manually operate and/or monitor the (SF5 PCIS) PRIMARY CONTAINMENT ISOLATION SYSTEM / NUCLEAR STEAM SUPPLY SHUTOFF in the control room: (CFR: 41.7 / 45.5 to 45.8) Group isolations 4.2 40 51 (239002) (SF3 SRV)
(239002K2.01) (SF3 SRV) SAFETY RELIEF VALVES Knowledge of electrical power supplies to the following: (CFR: 41.7) SRV solenoids 3.7 41 52 (259002) (SF2 RWLCS)
REACTOR WATER LEVEL CONTROL SYSTEM X
(259002K1.17) Knowledge of the physical connections and/or cause and effect relationships between the (SF2 RWLCS) REACTOR WATER LEVEL CONTROL SYSTEM and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8)RWCU 3.0 42 53 (261000) (SF9 SGTS)
STANDBY GAS TREATMENT SYSTEM X
(261000A3.02) Ability to monitor automatic operation of the (SF9 SGTS) STANDBY GAS TREATMENT SYSTEM including:
(CFR: 41.7 / 45.7) Fan start 3.8 43 54 (261000) (SF9 SGTS)
STANDBY GAS TREATMENT SYSTEM X
(261000K1.01) Knowledge of the physical connections and/or cause and effect relationships between the (SF9 SGTS) STANDBY GAS TREATMENT SYSTEM and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8) Plant ventilation systems 3.5 44
55 (262001) (SF6 AC) AC ELECTRICAL DISTRIBUTION X
(262001K5.02) Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the (SF6 AC) AC ELECTRICAL DISTRIBUTION: (CFR: 41.5 / 45.3)
Breaker control power 3.5 45 56 (262002) (SF6 UPS)
UNINTERRUPTABLE POWER SUPPLY (AC/DC)
X (262002) (SF6 UPS)
UNINTERRUPTABLE POWER SUPPLY (AC/DC) (291008K1.06)
BREAKERS, RELAYS, AND DISCONNECTS (CFR: 41.7)
Interpreting one-line diagram of control circuitry 3.6 46 57 (263000) (SF6 DC) DC ELECTRICAL DISTRIBUTION X
(263000) (SF6 DC) DC ELECTRICAL DISTRIBUTION (G2.1.7) 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) 4.4 47 58 (263000) (SF6 DC) DC ELECTRICAL DISTRIBUTION X
(263000K4.03) Knowledge of (SF6 DC) DC ELECTRICAL DISTRIBUTION design features and/or interlocks that provide for the following: (CFR: 41.7) Ground detection 2.9 48 59 (264000) (SF6 EGE)
EMERGENCY GENERATORS (DIESEL/JET)
X (264000A4.04) Ability to manually operate and/or monitor the (SF6 EGE) EMERGENCY GENERATORS (DIESEL/JET) in the control room: (CFR: 41.7 / 45.5 to 45.8) 4.1 49 60 (300000) (SF8 IA)
INSTRUMENT AIR SYSTEM X
(300000A2.03) Ability to (a) predict the impacts of the following on the (SF8 IA) INSTRUMENT AIR SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: (CFR: 41.5 / 45.6) Low instrument air pressure 3.9 50 61 (300000) (SF8 IA)
INSTRUMENT AIR SYSTEM X
(300000K3.06) Knowledge of the effect that a loss or malfunction of the (SF8 IA) INSTRUMENT AIR SYSTEM will have on the following systems or system parameters:
(CFR: 41.7 / 45.4) Component cooling water system 3.0 51 62 (400000) (SF8 CCS)
COMPONENT COOLING WATER SYSTEM X
(400000A1.01) Ability to predict and/or monitor changes in parameters associated with operation of the (SF8 CCS)
COMPONENT COOLING WATER SYSTEM including: (CFR: 41.5 /
45.5) CCW flow rate 3.0 52 63 (211000) (SF1 SLCS)
STANDBY LIQUID CONTROL SYSTEM X
(211000A2.09) Ability to (a) predict the impacts of the following on the (SF1 SLCS)
STANDBY LIQUID CONTROL SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: (CFR: 41.5 / 45.6)
Automatic or manual initiation failure 4.1 86
64 (212000) (SF7 RPS)
(212000) (SF7 RPS) REACTOR PROTECTION SYSTEM (G2.4.41)
Knowledge of the emergency action level thresholds and classifications (SRO Only) 4.6 87 65 (215005) (SF7 PRMS)
AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR X
(215005A2.03) Ability to (a) predict the impacts of the following on the (SF7 PRMS)
AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: (CFR:
41.5 / 45.6) Inoperable trip 3.9 88 66 (239002) (SF3 SRV)
(239002) (SF3 SRV) SAFETY RELIEF VALVES (G2.1.8) Ability to coordinate personnel activities outside the control room.
4.1 89 67 (259002) (SF2 RWLCS)
REACTOR WATER LEVEL CONTROL SYSTEM X
(259002) (SF2 RWLCS) REACTOR WATER LEVEL CONTROL SYSTEM (G2.4.3) Ability to identify post-accident instrumentation (CFR: 41.6 / 45.4) 3.9 90 (203000) (SF2, SF4 RHR/LPCI) RHR/LPCI:
INJECTION MODE (209002) (SF2, SF4 HPCS) HIGH PRESSURE CORE SPRAY SYSTEM (217000) (SF2, SF4 RCIC) REACTOR CORE ISOLATION COOLING SYSTEM (510000) (SF4 SWS*)
SERVICE WATER SYSTEM K/A Category Totals:
3 2
1 2
3 2
2 5
2 4
5 Group Point Total:
31
ES-4.1-BWR BWR Examination Outline (Nine Mile Point 1)
Emergency and Abnormal Plant EvolutionsTier 2/Group 2 (RO/SRO)
Item System / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
K/A Topic(s)
IR Q#
68 (202001) (SF1, SF4 RS)
RECIRCULATION SYSTEM X
(202001) (SF1, SF4 RS)
RECIRCULATION SYSTEM (G2.1.30) Ability to locate and operate components, including local controls (CFR: 41.7 / 45.7) 4.4 53 69 (214000) (SF7 RPIS)
ROD POSITION INFORMATION SYSTEM X
(214000K2.01) (SF7 RPIS) ROD POSITION INFORMATION SYSTEM Knowledge of electrical power supplies to the following:
(CFR: 41.7) RPIS 3.0 54 70 (215001) (SF7 TIP)
TRAVERSING IN CORE PROBE X
(215001K5.01) Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the (SF7 TIP) TRAVERSING IN CORE PROBE: (CFR: 41.5 / 45.3)
Flux detection 3.2 55 71 (216000) (SF7 NBI)
NUCLEAR BOILER INSTRUMENTATION X
(216000K4.16) Knowledge of (SF7 NBI) NUCLEAR BOILER INSTRUMENTATION design features and/or interlocks that provide for the following: (CFR:
41.7) RPV level instrumentation design calibration conditions 3.3 56 72 (219000) (SF5 RHR SPC) RHR/LPCI:
TORUS/SUPPRESSION POOL COOLING MODE X
(219000A4.02) Ability to manually operate and/or monitor in the control room: Valve lineup 4.2 57 73 (230000) (SF5 RHR SPS) RHR/LPCI:
TORUS/SUPPRESSION POOL SPRAY MODE X
(230000K3.05) Knowledge of the effect that a loss or malfunction of the (SF5 RHR SPS) RHR/LPCI:
TORUS/SUPPRESSION POOL SPRAY MODE will have on the following systems or system parameters: (CFR: 41.7 / 45.4)
Primary containment 3.7 58 74 (241000) (SF3 RTPRS)
REACTOR/TURBINE PRESSURE REGULATING SYSTEM X
(241000K3.17) Knowledge of the effect that a loss or malfunction of the (SF3 RTPRS)
REACTOR/TURBINE PRESSURE REGULATING SYSTEM will have on the following systems or system parameters: (CFR: 41.7 / 45.4)
Turbine acceleration 2.9 59 75 (245000) (SF4 MTGEN)
MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS X
(245000K6.01) Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the (SF4 MTGEN) MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS: (CFR: 41.7 / 45.7)
Gland seal 3.1 60 76 (259001) (SF2 FWS)
FEEDWATER SYSTEM X
(259001A1.08) Ability to predict and/or monitor changes in parameters associated with operation of the (SF2 FWS)
FEEDWATER SYSTEM including:
(CFR: 41.5 / 45.5) Feedwater control valve position 3,6 61
77 (271000) (SF9 OG)
OFFGAS SYSTEM X
(271000K1.15) Knowledge of the physical connections and/or cause and effect relationships between the (SF9 OG) OFFGAS SYSTEM and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8) Hydrogen water chemistry system 2.7 62 78 (286000) (SF8 FPS)
FIRE PROTECTION SYSTEM X
(286000A2.02) Ability to (a) predict the impacts of the following on the (SF8 FPS) FIRE PROTECTION SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: (CFR: 41.5 / 45.6) DC electrical distribution failure 2.6 63 79 (201002) (SF1 RMCS)
REACTOR MANUAL CONTROL SYSTEM X
(201002) (SF1 RMCS) REACTOR MANUAL CONTROL SYSTEM (G2.2.14) Knowledge of the process for controlling equipment configuration or status 4.3 91 80 (223001) (SF5 PCS)
PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES X
(223001A2.11) Ability to (a) predict the impacts of the following on the (SF5 PCS)
PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: (CFR: 41.5 / 45.6)
Abnormal suppression pool level 3.7 92 81 (234000) (SF8 FH)
FUEL HANDLING X
(234000K4.06) Knowledge of (SF8 FH) FUEL HANDLING design features and/or interlocks that provide for the following: (CFR:
41.7) Protection from dropping a fuel assembly 3.5 93 (201001) (SF1 CRDH)
CRD HYDRAULIC SYSTEM (201003) (SF1 CRDM)
CONTROL ROD AND DRIVE MECHANISM (201004) (SF7 RSCS)
ROD SEQUENCE CONTROL SYSTEM (201005) (SF1, SF7 RCIS) ROD CONTROL AND INFORMATION SYSTEM (201006) (SF7 RWMS)
ROD WORTH MINIMIZER SYSTEM (202002) (SF1 RSCTL)
RECIRCULATION FLOW CONTROL SYSTEM (204000) (SF2 RWCU)
REACTOR WATER CLEANUP SYSTEM (215002) (SF7 RBMS)
ROD BLOCK MONITOR SYSTEM (226001) (SF5 RHR CSS) RHR/LPCI:
CONTAINMENT SPRAY MODE SYSTEM MODE
(233000) (SF9 FPCCU)
FUEL POOL COOLING/CLEANUP (239001) (SF3, SF4 MRSS) MAIN AND REHEAT STEAM SYSTEM (239003) (SF9 MSVLCS) MAIN STEAM ISOLATION VALVE LEAKAGE CONTROL SYSTEM (256000) (SF2 CDS)
CONDENSATE SYSTEM (268000) (SF9 RW)
RADWASTE SYSTEM (272000) (SF7, SF9 RMS) RADIATION MONITORING SYSTEM (288000) (SF9 PVS)
PLANT VENTILATION SYSTEMS (290001) (SF5 SC)
SECONDARY CONTAINMENT (290003) (SF9 CRV)
CONTROL ROOM VENTILATION (290002) (SF4 RVI)
REACTOR VESSEL INTERNALS (510001) (SF8 CWS*)
CIRCULATING WATER SYSTEM K/A Category Totals:
1 1
2 2
1 1
1 2
0 1
2 Group Point Total:
14
Form 4.1-COMMON Common Examination Outline ES-4.1-COMMON COMMON Examination Outline (Nine Mile Point 1)
Facility:
Nine Mile Point 1 Date of Exam:
01/30/2023 Generic Knowledge and Abilities Outline (Tier 3) (RO/SRO)
Category K/A #
Topic RO SRO-Only Item #
IR Q#
IR Q#
- 1.
Conduct of Operations G2.1.4 Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, no-solo operation, maintenance of active license status, 10 CFR Part 55 82 3.3 64 G2.1.43 (G2.1.43) Ability to use an online power distribution monitoring system and/or procedures to determine the effects on reactivity of plant changes, such as RCS temperature, secondary plant, or fuel depletion (CFR: 41.10 / 43.6 / 45.6) 83 4.1 65 G2.1.1 (G2.1.1) Knowledge of conduct of operations requirements (CFR: 41.10 /
43.10 / 45.13) 84 4.2 94 G2.1.27 (G2.1.27) Knowledge of system purpose and/or function (CFR: 41.7) 85 4.0 95 Subtotal N/A 2
N/A 2
- 2.
Equipment Control G2.2.1 (G2.2.1) Ability to perform pre-startup procedures for the facility, including operating those controls associated with plant equipment that could affect reactivity (CFR: 41.5 / 41.10 / 43.5 / 43.6 / 45.1) 86 4.5 66 G2.2.20 (G2.2.20) Knowledge of the process for managing troubleshooting activities (CFR: 41.10 / 43.5 / 45.13) 87 2.6 67 G2.2.13 (G2.2.13) Knowledge of tagging and clearance procedures (CFR: 41.10 /
43.1 / 45.13) 88 4.3 96 G2.2.25 (G2.2.25) Knowledge of the bases in technical specifications for limiting conditions for operation and safety limits (SRO Only) (CFR: 43.2) 89 4.2 97 Subtotal N/A 2
N/A 2
- 3.
Radiation Control G2.3.5 (G2.3.5) Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms or personnel monitoring equipment (CFR: 41.11 / 41.12 /
43.4 / 45.9) 90 2.9 68 G2.3.5 (G2.3.5) Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms or personnel monitoring equipment (CFR:
41.11 / 41.12 / 43.4 / 45.9) 91 2.9 98 Subtotal N/A 1
N/A 1
- 4.
Emergency Procedures / Plan G2.4.14 (G2.4.14) Knowledge of general guidelines for emergency and abnormal operating procedures usage 92 3.8 69 G2.4.28 (G2.4.28) Knowledge of procedures relating to a security event (ensure that the test item includes no safeguards information) (CFR: 41.10 / 43.5 /
45.13) 93 4.1 99 G2.4.44 (G2.4.44) Knowledge of emergency plan implementing procedures protective action recommendations (SRO Only) (CFR: 41.10 / 41.12 / 43.5 /
45.11) 94 4.4 100 Subtotal N/A 1
N/A 2
Tier 3 Point Total N/A 6
N/A 7
Form 4.1-COMMON Common Examination Outline ES-4.1-COMMON COMMON Examination Outline (Nine Mile Point 1)
Facility:
Nine Mile Point 1 Date of Exam:
01/30/2023 Theory (Tier 4) (RO)
Category K/A #
Topic RO Item #
IR Q#
Reactor Theory 292002 (292002K1.09) NEUTRON LIFE CYCLE (CFR: 41.1) Define K-excess (excess reactivity) 95 2.6 70 292003 (292003K1.05) REACTOR KINETICS AND NEUTRON SOURCES (CFR: 41.1)
Define reactor period 96 3.7 71 292007 (292007K1.01) FUEL DEPLETION AND BURNABLE POISONS (CFR: 41.1)
Define burnable poison and state its use in the reactor 97 3.1 72 Subtotal 3
Thermodynamics 293003 (293003K1.07) STEAM (CFR: 41.14) Define the following term: Saturated liquid 98 2.8 73 293008 (293008K1.32) THERMAL HYDRAULICS (CFR: 41.14) (CORE ORIFICING)
Describe core bypass flow 99 2.6 74 293010 (293010K1.04) BRITTLE FRACTURE AND VESSEL THERMAL STRESS (CFR: 41.14) State how the possibility of brittle fracture is minimized by operating limitations 100 3.2 75 Subtotal 3
Tier 3 Point Total N/A 6
Form 4.1-1 Record of Rejected Knowledge and Abilities Refer to Examination Standard (ES)-4.2, Developing Written Examinations, Section B.3, for deviations from the approved written examination outline.
Tier/Group Randomly Selected K/A Reason for Rejection 1 / 1 Question #8 295019 PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR AK2.03 - Knowledge of the relationship between the PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR and the following systems or components:
Feedwater system An acceptable question could not be developed without overlapping the concepts tested in Question #79.
Randomly reselected K/A 295019 PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR AK2.09 - Knowledge of the relationship between the PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR and the following systems or components: Primary containment and auxiliaries.
1 / 1 Question #19 600000 PLANT FIRE ON SITE AA2.07 - Ability to determine or interpret the following as they apply to PLANT FIRE ON SITE: Whether malfunction(s) are due to common-mode electrical failures An acceptable question could not be developed without becoming an overly simplistic power supply question.
Randomly reselected K/A 600000 PLANT FIRE ON SITE AA2.06 - Ability to determine or interpret the following as they apply to PLANT FIRE ON SITE: Need for pressurizing control room (recirculating mode).
1 / 2 Question #22 295008 HIGH REACTOR WATER LEVEL 2.2.42 - Ability to recognize system parameters that are entry-level conditions for technical specifications An acceptable question could not be developed because the generic K/A did not fit well with the given evolution.
Randomly reselected K/A 295008 HIGH REACTOR WATER LEVEL 2.4.18 - Knowledge of the specific bases for emergency and abnormal operating procedures.
2 / 1 Question #39 223002 PRIMARY CONTAINMENT ISOLATION SYSTEM / NUCLEAR STEAM SUPPLY SHUTOFF A1.03 - Ability to predict and/or monitor changes in parameters associated with operation of the PRIMARY CONTAINMENT ISOLATION SYSTEM / NUCLEAR STEAM SUPPLY SHUTOFF including: Plant process computer/parameter display systems The system 510000 was inadvertently skipped during the sampling process. This question was an instance where another system was double sampled, therefore it was selected to be replaced by system 510000 Substituted K/A 510000 Service Water System A1.03 - Ability to predict and/or monitor changes in parameters associated with operation of the Service Water System, including: Service water pressures.
2 / 1 Question #42 259002 REACTOR WATER LEVEL CONTROL SYSTEM K1.10 - Knowledge of the physical connections and/or cause and effect relationships between the REACTOR WATER LEVEL CONTROL SYSTEM and the following systems: EMERGENCY GENERATORS (DIESEL)
An acceptable question could not be developed because there is little to no relationship between the Reactor Water Level Control system and the Emergency Diesel Generators at this facility.
Randomly reselected K/A 259002 REACTOR WATER LEVEL CONTROL SYSTEM K1.17 -
Knowledge of the physical connections and/or cause and effect relationships between the REACTOR WATER LEVEL CONTROL SYSTEM and the following systems: RWCU.
2 / 1 Question #46 262002 UNINTERRUPTABLE POWER SUPPLY (AC/DC) 291008 Breakers, Relays, and Disconnects K1.05 - Function of thermal overload protection device An acceptable question could not be developed due to lack of thermal overloads related to the UPSs at this facility.
Randomly reselected K/A 262002 UNINTERRUPTABLE POWER SUPPLY (AC/DC) 291008 Breakers, Relays, and Disconnects K1.06
- Interpreting one-line diagram of control circuitry.
2 / 2 Question #57 219000 RHR/LPCI:
TORUS/SUPPRESSION POOL COOLING MODE A3.01 - Ability to monitor automatic operation of the RHR/LPCI:
TORUS/SUPPRESSION POOL COOLING MODE including: Valve operation An acceptable question could not be developed because there is no automatic valve operation associated with this system.
Randomly reselected K/A 219000 RHR/LPCI:
TORUS/SUPPRESSION POOL COOLING MODE A4.02 - Ability to manually operate and/or monitor in the control room: Valve lineup.
2 / 2 Question #58 230000 RHR/LPCI:
TORUS/SUPPRESSION POOL SPRAY MODE K3.04 - Knowledge of the effect that a loss or malfunction of the RHR/LPCI:
TORUS/SUPPRESSION POOL SPRAY MODE will have on the following systems or system parameters: Suppression chamber air temperature An acceptable question could not be developed without testing minutia or operationally irrelevant information due to lack of indication and procedural guidance related to Torus air temperature.
Randomly reselected K/A 230000 RHR/LPCI:
TORUS/SUPPRESSION POOL SPRAY MODE K3.05 - Knowledge of the effect that a loss or malfunction of the RHR/LPCI:
TORUS/SUPPRESSION POOL SPRAY MODE will have on the following systems or system parameters: Primary containment.
2 / 2 Question #63 286000 FIRE PROTECTION SYSTEM A2.03 - Ability to (a) predict the impacts of the following on the FIRE PROTECTION SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: AC electrical distribution failure An acceptable question could not be developed that tested both the first and second parts of the K/A due to the simplicity of AC electrical supplies to Fire Protection and lack of procedural references.
Randomly reselected K/A 286000 FIRE PROTECTION SYSTEM A2.02 - Ability to (a) predict the impacts of the following on the FIRE PROTECTION SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: DC electrical distribution failure.
3 Question #69 2.4.23 - Knowledge of the bases for prioritizing emergency operating procedures implementation An acceptable question could not be developed at the RO license level because prioritizing EOP implementation is an SRO responsibility.
Randomly reselected K/A 2.4.14 - Knowledge of general guidelines for emergency and abnormal operating procedures usage.
1 / 1 Question #77 295005 MAIN TURBINE GENERATOR TRIP 2.4.22 - Knowledge of the bases for prioritizing safety functions during abnormal/emergency operations An acceptable question could not be developed because the generic K/A did not fit well with the given evolution.
Randomly reselected K/A 295005 MAIN TURBINE GENERATOR TRIP 2.4.45 - Ability to prioritize and interpret the significance of each annunciator or alarm.
1 / 1 Question #78 295006 SCRAM 2.2.19 - Knowledge of maintenance work order requirements An acceptable question could not be developed because the generic K/A did not fit well with the given evolution.
Randomly reselected K/A 295006 SCRAM 2.2.42
- Ability to recognize system parameters that are entry-level conditions for technical specifications.
1 / 1 Question #82 295030 LOW SUPPRESSION POOL WATER LEVEL 2.3.11 - Ability to control radiation releases An acceptable question could not be developed because the generic K/A did not fit well with the given evolution.
Randomly reselected K/A 295030 LOW SUPPRESSION POOL WATER LEVEL 2.4.16 -
Knowledge of emergency and abnormal operating procedures implementation hierarchy and coordination with other support procedures or guidelines such as, operating procedures, abnormal operating procedures, or severe accident management guidelines.
1 / 2 Question #83 295009 LOW REACTOR WATER LEVEL 2.4.43 - Knowledge of emergency communications systems and techniques An acceptable question could not be developed because the generic K/A did not fit well with the given evolution.
Randomly reselected K/A 295009 LOW REACTOR WATER LEVEL 2.4.21 - Knowledge of the parameters and logic used to assess the status of emergency operating procedures critical safety functions or shutdown critical safety functions.
2 / 1 Question #87 212000 REACTOR PROTECTION SYSTEM 2.4.20 - Knowledge of the operational implications of emergency and abnormal operating procedures warnings, cautions, and notes An acceptable question could not be developed because the generic K/A did not fit well with the given evolution.
Randomly reselected K/A 212000 REACTOR PROTECTION SYSTEM 2.4.41 - Knowledge of the emergency action level thresholds and classifications (SRO Only).
2 / 1 Question #89 239002 SAFETY RELIEF VALVES 2.1.29 - Knowledge of how to conduct system lineups, such as valves, breakers, or switches An acceptable question could not be developed because the generic K/A did not fit well with the given evolution.
Randomly reselected K/A 239002 SAFETY RELIEF VALVES 2.1.8 - Ability to coordinate personnel activities outside the control room.
2 / 2 Question #91 201002 REACTOR MANUAL CONTROL SYSTEM 2.2.5 - Knowledge of the process for making design or operating changes to the facility, such as 10 CFR 50.59, Changes, Tests and Experiments, screening and evaluation processes, administrative processes for temporary modifications, disabling annunciators, or installation of temporary equipment An acceptable question could not be developed because the generic K/A did not fit well with the given evolution.
Randomly reselected K/A 201002 REACTOR MANUAL CONTROL SYSTEM 2.2.14 -
Knowledge of the process for controlling equipment configuration or status.
3 Question #64 2.1.32 - Ability to explain and apply system precautions, limitations, notes, or cautions An acceptable question could not be developed for the generic K/A that fit with Tier 3.
Randomly reselected K/A 2.1.4 - Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, no-solo operation, maintenance of active license status, 10 CFR Part 55.
1 / 1 Question #78 295006 SCRAM 2.2.42 - Ability to recognize system parameters that are entry-level conditions for technical specifications An acceptable question could not be developed for the generic K/A that fit with Tier 1.
Randomly reselected K/A 295006 SCRAM 2.2.44
- Ability to interpret control room indications to verify the status and operation of a system and understand how operator actions and directives affect plant and system conditions.
1 / 1 Question #79 295019 PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR 2.2.36 - Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operation An acceptable question could not be developed for the generic K/A that fit with Tier 1.
Randomly reselected K/A 295019 PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR 2.1.20
- Ability to interpret and execute procedure steps.