ML21056A306

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Final Written Examination and Operating Test Outlines (Folder 3)
ML21056A306
Person / Time
Site: FitzPatrick Constellation icon.png
Issue date: 02/10/2020
From:
Exelon Generation Co
To: Todd Fish
Operations Branch I
Shared Package
ML19105A122 List:
References
CAC 000500
Download: ML21056A306 (28)
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Text

Appendix D Scenario Outline Form ES-D-1 Facility: James A. Fitzpatrick Scenario No.: NRC-1 Op-Test No.: 18-1 Examiners: Operators:

Initial Conditions: The plant is operating at approximately 93% power. RBCLC pump A is out of service for maintenance. SRV A is inoperable. Line 3 is out of service and ready to be restored.

Turnover: Restore Line 3 to service per OP-44 G.9. Then, raise Reactor power to 98% using Recirculation flow.

Event Malf. Event Event No. No. Type* Description N- Restore Line 3 to Service 1 N/A BOP, SRO OP-44 R- Raise Reactor Power with Recirculation Flow 2 N/A ATC, SRO OP-27 C- Loss of Offsite Power 3 ED44 SRO AOP-72, Technical Specifications C- SRV Inadvertently Opens 4 AD06:C BOP, SRO AOP-36, Technical Specifications FW05:A Feedwater Pump Vibration and Delayed Pump Trip 5 C-AII FW01:A ARP 09-6-4-11, AOP-41 Steam Leak in Drywell 6 MS02:A C-AII AOP-39, AOP-1 Loss of Coolant Accident 7 RR15:A M-AII EOP-2, EOP-4 C- First Torus Spray Valve Used Fails to Open 8 Remote ATC, SRO EOP-4 HPCI Trips 9 HP02 C-AII EOP-2

  • (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor

Appendix D Scenario Outline Form ES-D-1 Facility: James A. Fitzpatrick Scenario No.: NRC-1 Op-Test No.: 18-1

1. Malfunctions after EOP entry (1-2) 2 Events 8 & 9
2. Abnormal events (2-4) 4 Events 3, 4, 5, 6
3. Major transients ( 1-2) 1 Event 7
4. EOPs entered/requiring substantive actions (1-2) 2 EOP-2, EOP-4
5. Entry into a contingency EOP with substantive 2 actions (.::1 per scenario set)

EOP-2 Alternate Level Control Leg EOP-2 Emergency Depressurization Leg

6. Pre-identified critical tasks (~2) 2 CRITICAL TASK DESCRIPTIONS:

CT-1: Given a coolant leak inside the Containment, the crew will spray the Drywell, in accordance with EOP-4. Drywell spray must be initiated within 15 minutes of Torus pressure exceeding 15 psig.

CT-2: Given a coolant leak, a loss of high pressure injection systems, and the inability to restore and maintain Reactor water level above the Top of Active Fuel {TAF), the crew will initiate actions for an Emergency RPV Depressurization, in accordance with EOP-2. Reactor water level must be restored and maintained above TAF within thirty minutes of lowering below TAF.

Appendix D Scenario Outline Form ES-D-1 Facility: James A. Fitzpatrick Scenario No.: NRC-2 Op-Test No.: 18-1 Examiners: Operators:

Initial Conditions: The plant is operating at approximately 70% power. RBCLC pump A is out of service for maintenance.

Turnover: Swap Feedwater level control from level column A to B per OP-2A section G.27. Then, perform a Control Rod pattern ad*ustment in accordance with Reactor Engineering instructions.

Event Malf. Event Event No. No. Type* Description N- Swap Feedwater Level Control Instruments 1 N/A BOP, SRO OP-2A R- Perform Control Rod Pattern Adjustment 2 N/A ATC, SRO OP-26 C- CRD Flow Control Valve Fails Closed 3 RD03:A ATC, SRO OP-25 Recirc Flow Unit Failure 4 RR23:B I-All OP-16, Technical Specifications C- Loss of RBCLC Flow to Recirculation Pump B 5 SW01:B BOP, SRO OP-27, AOP-8, Technical Specifications Thermal Hydraulic Instability 6 RX03 C-AII AOP-8, AOP-1 Hydraulic Failure to Scram 7 RD22 M-AII EOP-3 C- First SLC Pump Delayed Trip 8 SL01 ATC, SRO EOP-3

  • (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor

Appendix D Scenario Outline Form ES-D-1 Facility: James A. Fitzpatrick Scenario No.: NRC-2 Op-Test No.: 18-1

1. Malfunctions after EOP entry (1-2) 1 Event 8
2. Abnormal events (2-4) 2 Events 5 & 6
3. Major transients (1-2) 1 Event 7
4. EOPs entered/requiring substantive actions (1-2) 1 EOP-2
5. Entry into a contingency EOP with substantive 1 actions (~1 per scenario set)

EOP-3

6. Pre-identified critical tasks (~2) 2 CRITICAL TASK DESCRIPTIONS:

CT-1: Given a failure to scram with Reactor power above 2.5%, the crew will lower Reactor power by one or*inore of the following methods, in accordance

.with EOP-3:

Terminating and preventing all RPV injection except SLC, RCIC, and CRD Tripping Recirculation pump A Injecting boron The Reactor power reduction must be initiated within five 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 EOP-3. All insertable control rods must be inserted within one hour of the start of the failure to scram.

Appendix D Scenario Outline Form ES-D-1 Facility: James A. Fitzpatrick Scenario No.: NRC-3 Op-Test No.: 18-1 Examiners: Operators:

Initial Conditions: The plant is operating at approximately 100% power. RBCLC pump A is out of service for maintenance.

Turnover: Swap TBCLC pumps per OP-41 section G.1.

Event Malf. Event Event No. No. Type* Description N- Swap TBCLC Pumps 1 N/A BOP, SRO OP-41 R- Condensate Booster Pump Trips 2 FW25:C ATC, SRO ARP 09-6-3-22, AOP-41, RAP-7.3.16 I-ATC, APRM Fails As-ls 3 NM14:A SRO ARP 09-5-2-2(44), OP-16 HP0S I-BOP, HPCI Inadvertently Initiates; Turbine Trip Pushbutton Fails 4

Override SRO AOP-77, AOP-32, Technical Specifications EPIC Fails 5 YC01 I-SRO ARP 09-8-1-4, ST-40C Remaining Condensate Booster Pumps Trip 6 FW25 C-AII AOP-41, AOP-1, EOP-2 RR33 Steam Leak in Drywell 7 M-AII MS02 EOP-2, EOP-4 I-BOP, RCIC Fails to Automatically Start 8 RC02 SRO EOP-2 RR33 RR16 Multiple Level Instrument Failures 9 RR17 I-All EOP-2, EOP-7 RR19 RR31

  • (N}ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor

Appendix D Scenario Outline Form ES-D-1 Facility: James A. Fitzpatrick Scenario No.: NRC-3 Op-Test No.: 18-1

1. Malfunctions after EOP entry (1-2) 2 Events & 9
2. Abnormal events (2-4) 3 Events 2, 4, 6
3. Major transients (1-2) 1 Event 7
4. EOPs entered/requiring substantive actions (1-2) 2 EOP-2, EOP-4
5. Entry into a contingency EOP with substantive 1 actions (~1 per scenario set)

EOP-7

6. Pre-identified critical tasks (~2) 2 CRITICAL TASK DESCRIPTIONS:

CT-1: Given a coolant leak inside the Containment, the crew will spray the Drywell, in accordance with EOP-4. Drywell spray must be initiated within 15 minutes of Torus pressure exceeding 15 psig.

CT-2: Given an unknown Reactor water level, the crew will flood the RPV, in accordance with EOP-7.

The RPV must be flooded to the Main Steam Lines within 30 minutes of when Reactor water level becomes unknown.

Appendix D Scenario Outline Form ES-D-1 Facility: James A. Fitzpatrick Scenario No.: NRC-4 Op-Test No.: 18-1 Examiners: Operators:

Initial Conditions: The plant is operating at approximately 90% power. RBCLC pump A is out of service for maintenance. Torus Cooling is in service following completion of RCIC surveillance testing.

Turnover: Secure Torus Cooling per OP-13B. Then, raise Reactor power using control rods and Recirculation flow per the provided reactivity instructions.

Event Malf. Event Event No. No. Type* Description N- Secure Torus Cooling 1 N/A BOP, SRO OP-13B C- RHRSW Pumps Trip 2 SW04:A SRO ARP 09-3-1-25, Technical Specifications R- Raise Reactor Power with Control Rods and Recirculation Flow 3 N/A ATC, SRO OP-65, OP-26 C- Stuck Control Rod 4 RD10 ATC, SRO OP-25 C- RCIC Steam Leak, Fails to Automatically Isolate 5 RC09 BOP, SRO ARPs, EOP-5, Technical Specifications Loss of Main Condenser Vacuum 6 MC01 C-AII AOP-31, AOP-1, EOP-2 Coolant Leak in Drywell 7 CU09 M-AII EOP-2, EOP-4 C- Ten Control Rods Fail to Scram; CRD Drive Water Pressure 8 RD13 ATC, Control Valve Fails As-ls SRO EOP-3 FW01 C- Feedwater Pumps Trip; HPCI Fails to Automatically Start 9 BOP, HP01 SRO EOP-2/3

  • (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor

Appendix D Scenario Outline Form ES-D-1 Facility: James A. Fitzpatrick Scenario No.: NRC-4 Op-Test No.: 18-1

1. Malfunctions after EOP entry (1-2) 2 Events 8 & 9
2. Abnormal events (2-4) 2 Events 5 & 6
3. Major transients (1-2) 1 Event7
4. EOPs entered/requiring substantive actions (1-2) 2 EOP-2, EOP-4
5. Entry into a contingency EOP with substantive 1 actions (.?:1 per scenario set)

EOP-3

6. Pre-identified critical tasks (.?:2) 3 CRITICAL TASK DESCRIPTIONS:

CT-1: Given a coolant leak inside the Containment, the crew will spray the Drywell, in accordance with EOP-4. Drywell spray must be initiated within 15 minutes of Torus pressure exceeding 15 psig.

CT-2: Given a coolant leak inside the Containment and degraded high pressure injection sources, the crew will establish injection to the Reactor, in accordance with EOP-2 and/or EOP-3. Injection must be established such that Reactor water level does not lower below -19".

CT-3: Given a failure to scram, the crew will initiate Control Rod insertion, in accordance with EOP-3. All insertable control rods must be inserted within one hour of the start of the failure to scram.

ES-301 Administrative Topics Outline Form ES-301-1 Facility: James A. Fitzpatrick Date of Examination: Feb 2020 Examination Level: RO Operating Test Number: 18-1 Administrative Topic (see Type Describe activity to be performed Note) Code*

Core Thermal Heat Balance Verification Using Turbine Conduct of Operations P,D,R Steam Pressure 17-1 NRC KIA 2.1.19 (3.9), OP-65, RAP-7.3.03 Perform RCIC Lineup Verification Per ST-24A Conduct of Operations N,S KIA 2.1.31 (4.6), ST-24A Explain RPS Operation Using Electrical Drawings Equipment Control M,R KIA 2.2.41 (3.5), 1.60-17, 1.67-97, 1.67-99, 1.67-101 Determine Worker Exposure for Emergent Work Radiation Control D,R KIA 2.3.4 (3.2), RP-AA-203

,:w ,, ,,, "

4 ' \, '<"'(  ;;;*

~  ;'" 11 "'
x E;er'g~~~y Plan ~,; 'IJ NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).
  • Type Codes and Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (S 3 for ROs; :s; 4 for SROs and RO retakes)

(N)ew or (M)odified from bank (~ 1)

(P)revious 2 exams (S 1, randomly selected)

ES-301 Administrative Topics Outline Form ES-301-1 Facility: James A. Fitzpatrick Date of Examination: Feb 2020 Examination Level: SRO Operating Test Number: 18-1 Administrative Topic (see Type Describe activity to be performed Note) Code*

Core Thermal Heat Balance Verification Using Turbine Conduct of Operations P,D,R Steam Pressure 17-1 NRC KIA 2.1.19 (3.8), OP-65, RAP-7.3.03 Determine Reportability Requirements - Technical Conduct of Operations N,R Specification Shutdown and Group I Isolation KIA2.1.18 (3.8), NUREG 1022, LS-AA-1400 Explain RPS Operation Using Electrical Drawings and Determine Technical Specification Impact of lnoperability Equipment Control M,R KIA2.2.41 (3.9), 1.60-17, 1.67-97, 1.67-99, 1.67-101, Technical Specification 3.3.1.1 Determine Worker Exposure for Emergent Work and Radiation Control D,R Required Actions KIA 2.3.4 (3.7), EN-RP-201 Determine Emergency Classification and Initiate Event Emergency Plan M,R Notification KIA 2.4.40 (4.5), EP-AA-1014, EP-CE-114-100 NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).

  • Type Codes and Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (S 3 for ROs; s 4 for SROs and RO retakes)

(N)ew or (M)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: Feb 2020 Exam Level: RO / SRO-I Operating Test Number: 18-1 Control Room Systems: 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Code* Safety Function

a. Control Room Heating, Ventilation and Air Conditioning/ Isolate Control Room and Relay Room Ventilation P,D,EN,S 9

17-1 NRC KIA 290003 A4.01 (3.2/3.2), OP-55B

b. Reactor Feedwater System / Transfer Feedwater Level Control to Master-Auto, Level Drifts Low M,A, L, S 2 KIA 259001 A4.01 (3.6/3.5), OP-2A
c. Low Pressure Core Spray I Shutdown Core Spray Following Terminate/Prevent and Subsequent Injection N, L, EN, S 4 KIA 209001 A4.01 (3.8/3.6), EP-5, OP-14
d. AC. Electrical Distribution/ Restore L34 to Normal Power Source, Loss of 10400 Bus D,A,S 6 KIA 262001 A4.01 (3.4/3. 7), OP-46A, AOP-17
e. Primary Containment Isolation System /Nuclear Steam Supply Shut-Off/ Reset Group 2 Isolation, Restore RWCU, RWCU Steam Leak M,A,EN,S 5 KIA 223002 A4.03 (3.6/3.5), AOP-15, OP-28
f. Control Rod Drive Hydraulic System / Restore CRD to Normal Alignment Following A TWS, CRD Controller Fails in Automatic D,A,S 1 KIA 201001 A2.07 (3.2/3.1 ), EP-3, OP-25
g. Main and Reheat Steam System / Perform Main Steam Shutdown Lineup N,L,S 3 KIA 239001 A4.01 (4.2/4.0), OP-1
h. Average Power Range Monitor/Local Power Range Monitor/

Bypass LPRM (RO Only)

KIA 215005 A4.06 (3.6/3.8), OP-16 D,S 7

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 In-Plant Systems: 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U

i. Primary Containment Isolation System / Nuclear Steam Supply Shut-Off/ Restore H2/O2 Monitors Following Isolation P, D,A, E 5

17-1 NRC KIA 223002 A2.09 (3.6/3.7), EP-2, OP-37

j. D.C. Electrical Distribution/ Perform In-Plant Actions for Station Blackout D,R,E 6 KIA 295003 AA1 .04 (3.6/3.7), AOP-49
k. Control Rod Drive Hydraulic System/ Swap CRD Flow Control Valves D,R 1 KIA 201001 2.1.20 (4.6/4.6), OP-25
  • All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions, all five SRO-U systems must serve different safety functions, and in-plant systems and functions may overlap those tested in the control room.
  • Type Codes Criteria for R /SRO-1/SRO-U (A)lternate path 4--6/4-6 /2-3 (C)ontrol room (D)irect from bank S 9/S 8/S 4 (E)mergency or abnormal in-plant ~ 1/~ 1/~ 1 (EN)gineered safety feature ~ 1/~ 1/~ 1 (control room system)

(L)ow-Power/Shutdown ~ 1/~ 1/~ 1 (N)ew or (M)odified from bank including 1(A) ~ 2/~ 2/~ 1 (P)revious 2 exams s 3/S 3/s 2 (randomly selected)

(R)CA ~ 1/~ 1/~ 1 (S)imulator

ES-401 1 Form ES-401-1

& -* Patrick Date of Exam: February 2020 Tier Group

.

  • RO KIA Category Points SRO-Only Points K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G* Total A2 G* Total
1. 1 4 4 3 3 3 3 20 4 3 7 Emergency and N/A N/A 2 1 1 1 2 1 1 7 2 1 3 Abnormal Plant Evolutions Tier Totals 5 5 4 5 4 4 27 6 .4 10 1 2 2 3 2 2 3 2 2 3 2 3 26 3. 2 5 2.:

Plant 2 2 0 1 1 1 1 1 1 1 2 1 12 0 1 2 3 Systems Tier Totals . 4 2 4 3 3 4 3 3 4 4 4 38 4 4 8

3. Generic Knowledge and Abilities ~ 2 3 .. 4 10 1 2 3 4 7 Categories 3 3 1 3 2 2 1 2 Note: 1. Ensure that at least two topics from every applicable KIA category are sampled within each tier of the RO and
SRO-only outline sections (i.e., except for one category in Tier 3 of the SRO-only section, th.e "Tier Totals" in each KIA category shall not be less than two). (One Tier 3 radiation control KIA 1s allowed if it is replaced by a:

KIA from another Tier 3 category.)

2. -The point total for each group arid 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 NRG revisions.

The final RO exam must total 75 points, and the SRO-orily exam must total 25 points:

3. Systems/evolutions within each group are identified on the* outline. Systems or evolutions that do not apply at the facility should be deleted with justification. 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 qf 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 KIAs having an importance rating (IR) of 2.5 or higher. . Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

6 . Select SRO topics for Tiers 1 and 2from the shaded systems and KIA categories.

7. . The generic (G) KIAs in Tiers 1 and 2 shall be selected from Section 2 of the KIA 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 KlAs.
8. On the following pages, enter the KIA numbers, a bdef description of each topic, the topics' I Rs for ttie

.applicabl~ 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 a category 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 KIA catalog, .and enter the KIAnµmbers, descriptions, IRs, and

. point totals(#) on Form ES-401-3. Limit SRO selections to KIAs that are linked to 10 CFR 55.43, G* Generic.KIAs

  • These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the KIA catalog is used to develop. the sample plan. They are riot required to be included when using earlier

( of revisions the KIA catalog. *

    • These systems/evolutions may be eliminated from the sample (as applicable to the facility) when Revision 3 of the KIA catalog is used to develop the sample plan.

ES-401 2 Form ES-401-1 IES-401 BWR Examination Outline Form"'" *~

  • Emergency and Abnormal Plant Evolutions-Tier 1/Group 1 (RO/SRO)

E/APE # / Name I Safety Function K1 K2 K3 A1 A2 G* KIA Topic(s) IR Q#

295021 (APE 21) Loss of Shutdown Cooling/ X 2.1 :23 Ability to perform specific system and 4A 76 4 integrated plant procedures during all .

modes of plant operation.

295023 (APE 23) Refueling Accidents / 8 X AA2.03 Ability to determine anclior interpret 3.8 77

  • the following as they apply to REFUELING ACCIDENTS: Airborne contamination levels.

295018 (APE 18) Partial or Complete Loss of X AA2.02 Ability to determine and/or interpret 3.2 78 CCW/8 the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLINGWATER: Cooling wat~r

.. temperature:

295030 (EPE 7) Low Suppression Pool Water X 2.2.38 Knowledge of conditions and

  • 4.5 79 Level/ 5 limitations in the facilitv license.
  • 295037 (EPE 14) Scram Condition Present X EA2,01 Ability to determine and/or interp'ret .4,3 80 and Reactor Power Above APRM Downscale the following as they apply to SCRAM or Unknown / *1 CONDITION PRESENT AND REACTOR POWERABOVE APRM DOWNSCALE OR UNKNOWN: Reactor Power 295004 Partial or Complete Loss of D.C. X AA2:03 Ability to determine and/or interpret 2.9 81 Power the following as they apply fo PARTIAL OR COMPLETE LOSS OF D.C. POWER:

Batterv voltaae 295038 (EPE 15) High Offsite Radioactivity X 2.4:20 Knowledge of.the operational 4:3 . 82, Release Rate I 9 implications of EOP warnings,- cautions, and notes.

295001 (APE 1) Partial or Complete Loss of X AK2.07 Knowledge of.the interrelations 3.4 1 Forced Core Flow Circulation/ 1 & 4 between PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION and the followinci: Core flow indication 295003 (APE 3).Partial or Complete Loss of X AA2.02 Ability to determine and/or interpret 4.2 2-AC Power/6 the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER:

Reactor power I pressure I and level 295004 (APE 4) Partial or Total Loss of DC X AA2.03 Ability to determine and/or interpret 2.8 3 Power/6

  • the. following as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER:

Batterv voltaae 295005 (APE 5) Main Turbine Generator Trip / X 2.4.31 Knowledge of annunciator alarms, 4.2 4

  • 3 indications, or response procedures.

295006 (APE 6) Scrarn / 1 X AK3.06 Kriowl~dge of the reasons forthe 3:2 5*

following responses as they apply to SCRAM: Recirculation pump speed reduction: Plant-Specific 295016 (APE 16) Control Room Abandonment X AK3.01 Knowledge of the reasons for the 4.1

  • 6 17 following responses as they apply t.o CONTROL ROOM ABANDONMENT:

Reactor SCRAM 295018 (APE 18) Partial or Complete Loss of . X AK1 :01 Knowledge of the operational 3.5 7 CCW/8 ,. implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER: Effects on component/system operations

ES-401 3 Form ES-401-1 295019 (APE 19) Partial or Complete Loss of . X AA 1.01 Ability to operate and/or monitor the 3.5 8 Instrument Air/ 8 following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR: .

Backuo air supply 295021 (APE 21) Loss of Shutdown Cooling/ X AA 1.02 Ability to operate and/or monitor the 3.5 9 4 following as they apply to LOSS OF SHUTDOWN COOLING: RHR/shutdown cooling 295023 (APE 23) Refueling Accidents / 8 X AK2.05 Knowledge of the interrelations J:5 10 between REFUELING ACCIDENTS and the.

following: Secondary containment ventilation 295024 High DrywellPressure / 5 X EK1 .01 Knowledge of the operational 4.1 11 implications of the following concepts as they apply to HIGH DRYWELL PRESSURE Drywell integrity: Plant-Specific 295025 (EPE 2) High Reactor Pressure / 3 X EA 1;04 Ability to operate and/or monitor the 3.8 12 following as they apply to HIGH REACTOR PRESSURE: HPCI: Plant-Specific 295026 (EPE 3) Suppression Pool High Water X EK3.01 Knowledge of the reasons for the 3.8 13 Temperature I 5 following responses as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Emergency/normal depressurization 295028 (EPE 5) High Drywell Temperature X 2.2.42 Ability to recognize system 3.9 14 (Mark I and Mark II only) / 5 parameters that are entry-level conditions for Technical Specifications.

295030 (EPE 7) Low Suppression Pool Water X EK1 .02 Knowledge of the operational .

  • 3.5 15
  • Level/ 5 implications of the following concepts as they apply to LOW SUPPRESSION POOL WATER LEVEL: Pump NPSH 29503HEPE 8) Reactor Low Water Level/ 2 *x 2.4.34 Knowledge of RO tasks performed 4.2 16 outside the main control room during.an emergency and the resultant operational effects.

295037 (EPE 14) Scram Condition Present X EK2.03 Knowledge of the interrelations 4.1 17 and Reactor Power Above APRM Downscale between SCRAM CONDITION PRESENT or Unknown / 1 AND REACTOR POWERABOVE APRM DOWNSCALE OR UNKNOWN and the followini:i: ARI/RPT/ATWS: Plant-Soecific 295038 (EPE.15) High Offsite Radioactivity X EK1 .02 Knowledge of the operational 4:2 18 Release Rate I 9 implications of the following concepts as they apply to HIGH OFF-SITE RELEASE RATE: Protection of the general public 600000 (APE 24) Plant Fire On Site/ 8 X AA2.14 Ability to determine and interpret the 3.0 19 following as they apply to PLANT FIRE ON SITE: Equipment that will be affected by fire suppression activities in each zone AK2:02 Knowledge of the interrelations 700000 (APE 25) Generator Voltage and. )( between GENERATOR VOLTAGE AND 3.1 20 Electric Grid Disturbances / 6 ELECTRIC GRID DISTURBANCES and the following: Breakers, relays KIA Categorv Totals: 4 4 3 3 3/4 3/3 Group Point Total: 20/7

ES-401 4 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emeraencv and Abnormal Plant Evolutions-Tiet 1/Grouo 2 (RO/SRO)

E/APE # / Name I Safetv Function K1 K2 K3 A1 A2 G* KIA Topic(s) IR Q#

295033 (EPE 10) High Secondary X EA2.03 Ability.to determine and/or interpret 4:2 83 Containment Area Radiation Levels/ 9 the following as they apply to HIGH SECONDARY CONTAINMENT AREA RADIATION LEVELS: tCause of high area radiation 295014 (APE 14) Inadvertent Reactivity X 2.1. 7 Ability to evaluate plant performance 4.7 84 Addition/ 1 and make operational judgments *based on operating characteristics, reactor behavior, and instrument interoretation.

295010 (APE 10) High Drywell Pressure I 5 X AA2.01 Ability to determine andfor interpret 3:8 85 the following as they apply fo HIGH DRYWELL PRESSURE: tLeak rates AK2.01 Knowledge of the interrelations 295013 (APE 13) High Suppression Pool X

  • between HIGH SUPPRESSION POOL 3.6 21.

WaterTemperature .f 5 WATER TEMPERATURE and the.

  • following: Suooressicin pool cooling
  • 295020 (APE 20) Inadvertent Containment X AK1 .02 Knowledge of the operational 3.5 22 Isolation / 5 & 7 implications of the following concepts as they apply to INADVERTENT CONTAINMENT ISOLATIO.N: Power/

reactivitv control 295029 (EPE 6) High Suppression Pool Water X EA1 .03 Ability to operate and/or monitor 2.9 23 Level/ 5 the following as they apply to HIGH SUPPRESSION POOL WATER LEVEL:

RHR/LPCI 295032 (EPE 9) High Secondary Containment X EK3.01 Knowledge of the reasons for the 3.5 24 Area Temperature/ 5 following responses as they apply to HIGH SECONDARY CONTAINMENTAREA TEMPERATURE: Emergency/riormal deoressurization 295009 Low . Reactor Water

. Level X AA2.01 Ability to determine and/or interpret 4.2 25

. the following as they apply to LOW

  • REACTOR WATER LEVEL: Reactor water level 295035 (EPE 12) Secondary Containment X . 2.4.1 Knowledge of EOP entry conditions 4.6 26 ..

High Differential Pressure / 5 and immediate action steps.

295036 (EPE 13) Secondary Containment X EA 1.01 Ability to operate and/or monitor 3.2 27 High Sump/Area Water Level/ 5 the following as they *apply to SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL: Secondary

  • containment equipment and floor drain systems KIA Category Point Totals: 1 1 1 2 1/2 1/1 Group Point Total: 7/~

ES-401 5 Form ES-401-1 I\ES-401 BWR Examination Outline Form ES-401-1 Plant Svstems-Tier 2/Grou J 1 RO/SRO)

System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G* KIA Topic(s) IR Q#

262001 (SF6 AC) AC Electrical X A2.06 Ability to.(a) predict the impacts of the 2.9 86 Distribution following on the A.C. ELECTRICAL DISTRIBUTION; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or-operations: De-energizing a plant bus 203000 (SF2, SF4 RHR/LPCI) X 2.1.31 Ability to locate control room switches, 4.3 87 RHR/LPCI: Injection Mode controls, and indications, and to determine that they correctly reflect the desired plant lineup.

211000 (SF1 SLCS) Standby Liquid X A2.01 Ability to (a) predict"the impacts of the 3.8 88 Control following on the STANDBY LIQUID CONTROL SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Pump trip 215005 (SF? PRMS) Average Power X A2.07 Ability to (a) predict the impacts of the 3.4 89 Range Monitor/Local Power Range following on the AVERAGE POWER RANGE Monitor MONITOR/LOCAL POWER RANGE .

MONITOR SYSTEM; and (b) based on those predictions, use procedures to*correct, control, or mitigate the consequences of those abnormal conditions or operations:

Recirculation flow channels flow mismatch 209001 (SF2, SF4 LPCS) X 2.4.9 Knowledge of low power/shutdown 4.2 90 Low-Pressure Core Spray implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.

203000 (SF2, SF4 RHR/LPCI) X K1 .17 Knowledge of the physical connections 4.0 28 RHR/LPCI: Injection Mode and/or cause/effect relationships between RHR/LPCI: INJECTION MODE (PLANT SPECIFIC) and the following: Reactor pressure 205000 (SF4 SCS) Shutdown Cooling X A1 .01 Ability to predict and/or monitor 3.3 29 changes in parameters associated with operating the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) controls including: Heat exchanger cooling flow 205000 (SF4 SCS) Shutdown Cooling X K3.02 Knowledge of the effect that a loss or 3.2 30 malfunction of the SHUTDOWN COOLiNG SYSTEM (RHR SHUTDOWN COOLING MODE) will have on following: Reactor water level: Plant-Specific 206000 (SF2, SF4 HPCIS) X A 1.06 Ability to predict and/or monitor 3.8 31 High-Pressure Coolant Injection changes in parameters associated with operating the HIGH PRESSURE COOLANT INJECTION SYSTEM controls including:

System flow

ES-401 Form .ES-401-1 209001 (SF2, SF4 LPCS) X l*v"  ::,;z+ K5.04 Knowledge of the operational 2,8. 32 Low-Pressure Core Spray *. implications of the *folJowing concepts as they .

i<t:1>:' apply to LOW PRESSURE CORE SPRAY

  • ./ SYSTEM: Heat re_mova_l (transfer)
  • 33

.. rj;:} SYSTEM including: Flow indication: Plant-Specific * * * * *

r. *.

212000 (SF? RPS) Reactor Protection X K4.02 Knowledge of REACTOR *. . . 3.5. 34 it;,;; PROTECTION SYSTEM design feature(s) .

l)1i  :

and/or interlocks which provide.for the.

following: The prevention of a reactor SCRAM following a single component fajJure 215003 (SF? IRM) . X

., .. , K2:01 Knowledge of electricai power supplies 2.5 35 Intermediate-Range Monitor to the following: IRM channels/detectors 215003 (SF? IRM) X ::. A4.01 Ability to manually operate and/or 3.3 36 Intermediate-Range Monitor .. monitor in the control room: IRM*recorder indication 215004 (SF? SRMS) Source-Range 1:i; X A3.04 Ability to monitor automaUc operations 3.6 37 Monitor of the SOURCE RANGE MONITOR (SRM) .

. i~f-; SYSTEM including: Control.rod block status.

215005. (SF? PR.MS) Average Power 2.1.28 Knowledge of the purpose an_d ..

  • 4.1 38

. Range Monitor/Local Power Range  ;'t; function of major system components arid.

Monitor \'~h .. controls. . . .

217000 (SF2, SF4 RCIC) Reactor ~i A2.08 Ability to (a) predict the impacts of the 3.b 39.

';iil following on the REACTOR CORE

~

,,,'\,

Core Isolation Cooling .*

  • ISOLATION COOLING SYSTEM (RCIC); and j

(b) based on those predictions, use

  • procedures to correct, control, or mitigate the*

x:-r

  • t consequences of those abnormal *conditions or operations: Loss of lube oil cooling

. 218000 (SF3 ADS) Automatic X . :hr K3.02 Knowledge of the effect that a loss or 4.5

  • 40 Depressurization  ::*" malfunction of the AUTOMATIC DEPRESSU.RIZATION.SYSTEM will have on following: Ability to rapidlydepressurize the.

> reactor

,::,;J: < malfunction K3.19 Knowledge of the effect that a loss or 2.8 of the PRIMARY CONTAINMENT 41

. Supply Shutoff . ISOLATION SYSTEM/NUCLEAR STEAM

,., ..,,;--. i>i;J! SUPPLY SHUT-OFF will have orifollow'ing: .

"*r,;*. Containment atmosphere sampfjng

  • \**.  ;,:*

239002 (SF3 SRV) Safety Relief X *.r K6.05 Knowledge of the effect that a loss or 3.o*- 42 Valves malfunction of the following will have ori the_ *

i!'
  • REL_IEF/SAFETY VALVES
    Discharge line
  • , :JJ ,,;:!'¥* vacuum breaker 259002 (SF2 RWLCS) Reactor Water X i'.

,,.,. K1 .16 Knowledge ()f the physicai connections 3.4 43.

Level Control . *./:, and/or cause-effect.relationships between 1/;l

>:::;. .REACTOR. WATER LEVEL CONTROL. . . . * .

SYSTEM and fhe following: HPCI, Plant- . >

!1?l Specific . . . ** * *. *

!::JJ' X 261000 (SF9 SGTS) Standby Gas Treatment

,
::' A3.01 Ability to monitor automatic *operations of the STANDBY GASTREATMENT ..

3.2 44

\':': SYSTEM including: System flow

  • 261000 (SF9 SGTS) Standby Gas

' .j 3/4 components, 2.1.30 Ability to locate and operate 4.4 45 Treatment , .. '* including local controls .

ES-401 7 Form ES-401-1 262001 (SF6 AC) AC Electrical X A4.01 Ability to manually operate and/or 3.4 46 Distribution monitor in the control room: All breakers and disconnects (including available switch yard):

Plant-Specific .

262002 (SF6.UPS) Uninterruptable X A2.02 Ability to (a) predict the impacts of the 2.5 47 Power Supply (AC/DC) following .on the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Over voltage 263000 (SF6 DC) DC Electrical X K2.01 Knowledge of electrical power supplies 3.1 48 Distribution to the following: Major D.C. loads 264000 (SF6 EGE) Emergency X K6.08 Knowledge of the effect that a loss or 3.6 49 Generators (Diesel/Jet) EOG malfunction of the following will have on the EMERGENCY GENERATORS (DIESEUJET): A.C. power 264000 (SF6 EGE) Emergency X K5.05 Knowledge of the. operational 3.4 50 Generators (Diesel/Jet) EDG implications of the following concepts as they apply to EMERGENCY GENERATORS (DIESEUJET): Paralleling A.C. power sources 300000 (SF8 IA) Instrument Air X K4.03 Knowledge of INSTRUMENT AIR 2.8 51 SYSTEM design feature(s) and or interlocks which provide for the following: Securing of IAS upon loss of cooling water 300000 (SF8 IA) Instrument Air X 2.4.49 Ability to perform without reference to 4.6 52 procedures those actions that require immediate operation of system components and controls.

400000 (SF8 CCS) Component X K6.05 Knowledge of the effect that a loss or 2.8 53 Cooling Water malfunction of the following will have on the CCWS: Motors KIA Category Point Totals: 2 2 3 2 2 3 2 2/ 3 3

2 3 Group Point Total:

/2 91

ES-401 8 Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Svstems Tier 2/Grouo 2 (RO/SRO\

System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G* KIA Topic(s) IR Q#

290002 (SF4 RVI) Reactor Vessel Internals X 2.2.25 Knowledge of the bases in 4.2 91 Technical Specifications for limiting conditions for operations and safetv limits.

234000 Fuel Handling Equipment X A2.01 Ability to (a) predict the impacts of 3.7 92 the following on the FUEL HANDLING EQUIPMENT;.and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Interlock failure 202002 (SF1 RSCTL) Recirculation Flow Control X 2.2.40 Ability to apply Technical 4.7 93 Specifications for a system.

201001 (SF1 CRDH) CRD Hydraulic X 2.2.12 Knowledge of surveillance 3.7 54 I orocedures.

201003 (SF1 CROM) Control Rod and Drive X A 1.03 Ability to predict and/or monitor 2.9 55 Mechanism changes in parameters associated with operating the CONTROL ROD AND DRIVE MECHANISM controls including:

CRD drive water flow 201002 (SF1 RMCS) Reactor Manual Control X K1 .05 Knowledge of the physical 3.4 56 connections and/or cause-effect relationships between REACTOR MANUAL CONTROL SYSTEM and the following: Rod worth minimizer: Plant-Soecific 202001 (SF1, SF4 RS) Recirculation X K5.1 o Knowledge of the operational 2.8 57 implications of the following concepts as they apply to RECIRCULATION SYSTEM: Motor generator set operation:

Plant-Specific .

204000 (SF2 RWCU) Reactor Water Cleanup X K4.06 Knowledge of REACTOR WATER 2.6 58 CLEANUP SYSTEM design feature(s) and/or interlocks which provide for the following: Maximize plant efficiency (use of reaenerative heat exchanaer\

214000 (SF7 RPIS) Rod Position Information X K6.02 Knowledge of the effect that a loss 2.7 59 or malfunction of the following will have on the ROD POSITION INFORMATION SYSTEM: Position indication orobe 215001 (SF7 TIP) Traversing _In-Core Probe X A2.07 Ability to (a) predict the impacts of 3.4 60 the following on the TRAVERSING IN-CORE PROBE; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Failure to retract during accident conditions: Mark-I&11 215002 (SF7 RBMS) Rod Block Monitor X A4.03 Ability to manually operate and/or 2.8 61 monitor in the control room: Trio bvoasses 219000 (SF5 RHR SPC) RHR/LPCI:

X A4.04 Ability to manually operate and/or 3.0 62 Torus/Suppression Pool Cooling Mode monitor in the control room: Minimum flow valves 223001 (SF5 PCS) Primary Containment and X A3.04 Ability to monitor automatic 4.2 63 Auxiliaries operations of the PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES including: Containment I drvwell resoonse durina LOCA 256000 Reactor Condens_ate System X K3.02 Knowledge of the effect that a loss 3.2 64 or malfunction of the REACTOR CONDENSATE SYSTEM will have on following: CRD hydraulics system

-ES-401 9 Form ES-401-1 241000 (SF3 RTPRS) Reactor/Turbine Pressure .

X K1.. 31 Knowledge cif the ptiysical 3:1 65 Regulating .

connections and/or cause-effect*

relationships between .

REACTOR/TURBINE PRESSURE.*

REGULATING SYSTE.Mand the followin!l: Turbine protection .

KIA Category Point Totals: 2 0 1 1 1 1 1 1 1 2 1 Group Point Tota.I: 12/3 .

/1 /2

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: James A. FitzPatrick Date of Exam: February 2020 Category KIA# Topic RO SRO-only IR # IR #

2.1.26 Knowledge of industrial safety procedures (such as rotating 3.4 66 equipment, electrical, high temperature, high pressure, caustic, chlorine, oxvaen and hvdrogen).

2.1.44 Knowledge of RO duties in the control room during fuel handling 3.9 67 such as responding to alarms from the fuel handling area, communication with the fuel storage facility, systems operated from the control room in support of fueling operations, and supporting

1. Conduct of instrumentation.

Operations 2.1.37 Knowledge of procedures, guidelines, or limitations associated with 4.3 68 reactivity manaaement.

2.1.13 Knowledge of facility requirements for controlling vital/controlled 3.2 94 access.

2.1.41 Knowledge of the refueling process. 3.7 95 Subtotal 3 2 2.2.14 Knowledge of the process for controlling equipment configuration or 3.9 69 status.

2.2.39 Knowledge of less than or equal to one hour Technical Specification 3.9 70 action statements for systems.

2. Equipment 2.2.22 Knowledge of limiting conditions for operations and safety limits. 4.0 71 Control 2.2.7 Knowledge of the process for conducting special or infrequent tests. 3.6 96 2.2.18 Knowledge of the process for managing maintenance activities 3;9 97 during shutdown operations, such as risk assessments, work prioritization, etc.

Subtotal 3 2 2.3.5 Ability to use radiation monitoring systems, such as fixed radiation 2.9 72 monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

3. Radiation 2.3.12 Knowledge of radiological safety principles pertaining to licensed 3.7 98 Control operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

Subtotal 1 1 2.4.41 Knowledge of the emergency action level thresholds and 2.9 73 classifications.

2.4.3 Ability to identify post-accident instrumentation. 3.7 74

4. 2.4.26 Knowledge of facility protection requirements, including fire brigade 3.1 75 Emergency and portable fire fighting equipment usage.

Procedures I 2.4.18 Knowledge of the specific bases for EOPs. 4.0 99 Plan 2.4.30 Knowledge of events related to system operation/status that must be 4.1 100 reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.

Subtotal 3 2 Tier 3 Point Total 10 7

.ES-401. Record. of Rejected K/As Form ES-401".'4 Tier/ Randomly Selected KIA* Reason for Rejection Group The following topics/ KIAs were excluded from the systematic and random sampling process:

295027 High Containment This topic applies to plants with Mark Ill Temperature

  • containments only~ *The facility has a* Mark I .

1 /1 containment.*

  • 295011
  • High Containment This topic applies to plants with Mark Ill Temperature containments only. The facility has a Mark I 1/2 containment. .

207000 Isolation (Emergency) This system is not installed at the facility; 2/1 Condenser 209002 HPCS This system is not installed at the facility.

2/1 201004 RSCS

  • This system is no longer installed at the facility:

2/2 201005 RCIS This system is not installed at the facility, 2/2 2.2.3 Knowl.edge of the design, . This KIA applies to multi-unit facilities only, procedural, and operational G differences between units.

2.2.4 Ability to explain the This KIA applies to multi-unit facilities only.

vadations .in control board/control room layouts, G systems, instrumentation, and procedural actions between units at a facility.

ES-401 Record of Rejected K/As Form ES-401-4 The following KIAs were rejected followinQ the systematic and random samoling process:

Question 5 An acceptable question could not be dev~loped due to lack ofa direct turbine generator trip on

.295006 Scram a Reactor scram at the facility.

AK3.05 - Knowledge of the Randomly resampled KIA 295006 Scram reasons for the following 1/ 1 AK3.06 - Knowledge of the reasons for the responses as they apply to following responses as they apply to SCRAM:

SCRAM: Direct turbine Recirculation pump speed reduction: Plant-generator trip: Plant-Specific Specific.

Question 6 An acceptable question could not be developed without too much overlap from the last two NRC 295016 Control Room exams.

Abandonment Randomly resampled KIA 295016 Control AK3.03 - Knowl~dge of the Room Abandonment AK3.01 - Knowledge of 1/ 1 reasons for the following the reasons for the following responses as they responses as they apply to apply to CONTROL ROOM ABANDONMENT:

CONTROL ROOM Reactor SCRAM.

ABANDONMENT: Disabling control room controls Question 13 An acceptable question could not be developed due to simplistic KIA, lack of specific bases . *.

295026 Suppression Pool High related to KIA, and similarity to Question 21 ..

Water Temperature Randomly resampled KIA 295026 Suppression EK3.02 - Knowledge of the Pool High Water Temperature EK3.01 -

1/ 1 reasons for the following Knowledge of the reasons for the following

.responses as they apply to responses as they apply to SUPPRESSION SUPPRESSION POOL HIGH POOL HIGH WATER TEMPERATURE:

WATER TEMPERATURE:

Emergency/normal depressurization.

  • Suppression pool cooling Question 14 An acceptable question could not be developed due to lack of system set points, interlocks .and 295028 High Drywell automatic actions associated with the EOP Temperature entry condition for high Drywell temperature.

2.4.2.:. Knowledge of system set 1/ 1 Randomly resampled KIA 295028 High Drywell points, interlocks and automatic Temperature 2.2.42 - Ability to recognize actions associated with EOP system parameters that are entry-level entry conditions.

conditions for Technical Specifications.

ES-401 Record of Rejected K/As Form ES-401-4 Question 18 An acceptable question could not be developed due to lack of testable RO level material. .

295038 High Offsite Radioactivity Release Rate Ran?omly _resampled KIA 295038 High Offsite Rad1oact1v1ty* Release Rate EK1 .02 -

  • EK1 .01 - Knowledge of the Know~edge of the operational implications of the 1/ 1 operational implications of the following concepts as they apply to HIGH OFF-following concepts as they SITE RELEASE RATE: Protection of the apply to HIGH OFF-SITE general public.

RELEASE RATE: Biological effects of radioisotope ingestion Question 25

  • A~ acceptable question could not be developed without oversampling Secondary Containment 295034 Secondary ..

Control concepts (see also Questions #24 26 .

Containment Ventilation High

  • 27, and 83). . * .' '

Radiation

  • Randomly resampled KIA 295009 Low Reactor EA2.02 -Ability to determine Water Level AA2.01 ~ Ability to determine. -*

.1 / 2 and/or interpret the following as and/or interpret the following as they apply to they apply to SECONDARY LOW REACTOR WATER LEVEL: Reactor -

CONTAINMENT VENTILATION water level.

HIGH RADIATION: Cause of high radiation levels Question 41 An accepta~le_question could not be developed

?ue t? s1mphstIc nature of KIA (any failure of 223002 Primary Containment 1solatIon will rather obviously lead to only high Isolation/Nuclear Steam_ Supply

  • radiation levels). * . * . * * . . .

Shutoff . * .

Rand~mly resampled KIA 223002 Primary

  • K3.06 - Knowledge of the effect Containment lsolaUon/Nuclear Steam Supply that a loss or malfunction of the Shutoff K3.19 - Knowledge of the effect that a 2/1 PRIMARY CONTAiNMENT
  • loss or malfunction of the PRIMARY .

ISOLATION CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SYSTEM/NUCLEAR STEAM SUPPLY SHUT-SUPPLY SHUT-OFF will have OFF will have on following: Containment*

on following: Turbine building atmosphere sampling. *

  • radiation . . * >. .

. ES-401. .Record of Rejected K/As Form ES-401'.'4 Question 46 A~ acce~table question could not be developed without either low LOO or testing GFES type 262001 AC Electrical knowledge .. * *

  • Distribution Ra.ndomly resampled KIA 262001 AC Electrical A4.02 -AbiMy to manually 2/1 Distribution _A4.~1 -Ability to manually operate operate and/or monitor.in the*

and/or monitor in the control room: All breakers control room: Synchroscope..

and disconnects (including available switch *.

  • including understanding of . '

yard): Plarit-Specifici

  • running and incoming voltages Question 50 A~ acceptable question could not b~ developed without too much overlap from the last two NRC 264000 Emergency Generators exams.

(Diesel/Jet) EOG * . ..

Randomly res~mpled KIA 264000 Emergency .

  • KS.06 - Knowledge of the Generators (Diesel/Jet) EDG KS.OS*_ * .. *
  • 2 / 1 . operational implications .of the Know~edge of the operational implicationSof the foUowing concepts as they . following concepts as they apply to ** * . * **

apply to EMERGENCY EMER~ENCY GENERATOR.S (DIESEL/JET):

GENERATORS (DIESEL/JET): . Paralleling A.C. power sources. . .* .*

  • Load sequencing . * .

Question 59 A~ acceptable question could not be developed

  • without too much overlap from the last two NRC 214000 Rod Position exams.

Information ..

Randomly resampled KIA 214000 Rod.Position K6.01 - Knowledge of the effect 212 *. Information K6.02 - Knowledge ofthe effect that that a loss or malfunction of the

~ loss or malfunction of the following will have following will have on the ROD on the ROD POSITION INFORMATION *

  • POSITION INFORMATION SYSTEM: .Position indication* probe. *
  • SYSTEM: A.C. electrical power Question 64 A~ acceptable question could riot be developed without oversampling RHR concepts across the 230000 RHR/LPCI:

exam.

Torus/Suppression Pool Spray Mode *

  • Randomly resampled KIA 256000 Reactor ..

Condensate System K3.02 - Knowledge of the K3.02 - Knowledge of the effect effect that a loss. or malfunction of the

  • that a loss or malfunction of the
  • 2/2 RHR/LPCI:

REACTOR CONDENSATE SYSTEM wHI have onfollowing: CRD hydraulics system. * * * * *.

TORUS/SUPPRESSION POOL SPRAY MODE will have on .

.following: Suppression pool temperature * *

  • Record of Rejected K/As Form ES-401-4 Question 68 ..
  • An acceptable question could not be developed at a discriminating level due to limited related *.

2.1.15 ;_. Knowledge of administrative requirements for ROs (they just

  • administrative requirements for review these as part of turnovert temporary m,magement 3

directives, such as standing Randomly resampled KIA 2.1.37 - knowledge orders,* night orders, operations of procedures, guidelines, or limitations *

  • memos; etc.* associated with reactivity management.

Question 81

  • An acceptable qUesUon could not be developed
  • at the SRO level without being low LOO and
  • 295006 Scram*

overlapping with operating exam.

M2.03 - Ability to d.etermine . Randomly resampled KIA 295004 Partial or .

  • 1/ 1 and/or interpret the following as Complete Loss of D.C. Power M2.03 -Ability.

they apply to SCRAM: Reactor to determine and/or interpret the following as water level they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER Battery voltage. .

Question 86 An acceptable quesUon could not be developed at the SRO level without being lciw LOD due to 262001 AC Electrical . the simplistic* nature of the KIA.

Distribution Randomly resampled KIA 262001 AC Electrical A2.07 - Ability to (a) predict the. Distribution A2.06 -Ability to (a) predict the

  • impacts of the following on the impacts of the following*on the AC. *.

AC. ELECTRICAL ELECTRICAL DISTRIBUTION; and (b) based 2/1 DISTRIBUTION; and (b) based on those predictions, use procedures to correct, on those predictions, use control, or mitigate the consequences of those*

procedures to correct, control, abnormal conditions or operations: De-

  • or mitigate the consequences of energizing a plant .bus.

those abnormal conditions or operations: Energizing a dead bus

ES-401

  • Record of Rejected K/As Form .ES-401-4 Question 92 An acceptable question could not be developed at the SRO level without being low LOO and *

.259001. Feedwater overlapping the operating exam;. *

  • A2.03 -Ability to (a) predict the Randomly resampled KIA 234000 Fuel impacts of the *following on the Handling Equipment A2.01 - Ability to (a) .

.REACTOR FEED WATER .*

predict the impacts of the following on the. FUEL.

SYSTEM; and (b) based on 2/2 *HANDLING EQUIPMENT: arid (b) based on

  • those predictions, use
  • those predictions; use procedures to correGt, * .

procedures to correct, control, control, or mitigate the consequences of those or mitigate the consequences.of abnormal conditions or operations: Interlock those abnormal conditions or failure.

operations: Loss ofcondensate .

pump(s)

. '

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