ML19119A177

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Final Written Examination and Operating Test Outlines (Folder 3)
ML19119A177
Person / Time
Site: Nine Mile Point Constellation icon.png
Issue date: 11/30/2018
From: Peter Presby
Operations Branch I
To: Isham P
Exelon Generation Co
Shared Package
ML18338A499 List:
References
CAC 00500, EPID L-2018-OLL-0007
Download: ML19119A177 (26)
v • d • e


Text

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Nine Mile Point Unit 1 Date of Examination: December 2018 Examination Level: RO Operating Test Number: LC1 17-1 NRG Administrative Topic Type Describe activity to be performed (see Note)

Code*

Verification Of Active License Status Conduct of Operations M,R OP-AA-105-101, OP-AA-105-102, KIA 2.1.4 (3.3)

DWFDT / DWEDT Leak Rate Determination and Evaluation Conduct of Operations D,R N1-0P-8, KIA 2.1.18 (3.6)

Develop a clearance boundary for the Liquid Poison Equipment Control D,R Test Tank OP-CE-109-.01 KA 2.2.13 (4.1)

P,D,R Application of Radiation Exposure Limits IAW RP-AA-Radiation Control 203 - SOC Room (2017 NRG)

RP-AA-203, KIA 2.3.4 (3.2)

Emergency Procedures/Plan I

NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.

  • Type Codes & Criteria:

(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (:,; 3 for ROs; ::; 4 for SROs & RO retakes)

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

(P)revious 2 exams (:,; 1; randomly selected)

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Nine Mile Point Unit 1 Date of Examination: December 2018 Examination Level: SRO Operating Test Number: LC1 17-1 NRC Administrative Topic Type Describe activity to be performed (see Note)

Code*

Reactivate SRO Licenses Conduct of Operations D,R OP-AA-105-102, KA 2.1.4 (3.8)

Perform Time to Boil Calculation for Reactor Coolant Conduct of Operations D,R System OP-NM-108-117-1002, KIA 2.1.40 (3.9)

Review and Approval of Completed Surveillance Test, N1-ST-Q13, Emergency Service Water Pump and Equipment Control N,R Check Valve Operability Test N1-ST-Q13, KIA 2.2.12 (4.1)

P,D,R Determine Actions for Inoperable Service Water Radiation Control Radiation Monitor (2017 NRG)

N1-ARP-H1, ODCM, KIA 2.3.15 (3.1)

I Emergency Event Reclassification and Notification Emergency Procedures/Plan D,R EP-CE-111, EPIP-EPP-01 EAL Flowchart, KIA 2.4.41 (4.6)

NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.

  • Type Codes & Criteria:

(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (:5 3 for ROs; :5 4 for SROs & RO retakes)

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

(P)revious 2 exams (:5 1; randomly selected)

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Nine Mile Point Unit 1 Date of Examination: December 2018 Exam Level: RO/SR0-1/SRO-U Operating Test No.: LC1 17-1 NRC Control Room Systems* (8 for RO); (7 for SRO-I} ; (2 or 3 for SRO-U)

System/ JPM Title Type Code*

Safety Function

a. Swap CRD Pumps M,A,S 1

KIA 201001 A4.01 (3.1/3.1), N1-0P-5

b. Perform N1-ST-M8, Reactor Building Emergency Ventilation System Operability Test N,S,EN 9

KIA 288000 A4.01 (3.1/2.9), N1-ST-M8

c. Vent the Drywell Prior to Personnel Entry >212 M,S,L,A 5

KIA 223001 A4.03 (3.4/3.4), N1-0P-9

d. Rapid RWCU System Restoration for Level Control (RO Only)

D,S,L 2

KIA 204000 A4.06 (3.0/2.9), N1-EOP-HC

e. Restore Emergency Condenser To Service D,A,EN,S 4

KIA 207000 A4.05 (3.5/3.7), N1-0P-13

f. Swap PB 101 from 1014 to R1011 D,S 6

KIA 262001 A4.01 (3.4/3.7), N1-0P-30

g. Control Rod Exercising Operability Test P,D,A,S 7

KIA 214000 A4.02 (3.8/3.8), N1-ST-W1, N1-0P-5 (2015 NRC)

h. MSIV Stroke test and Limit Switch Test P,S, D 3

KIA 239001 A4.01 (4.2/4.0), N1-ST-Q26 (2015 NRC)

In-Plant Systems* (3 for RO); (3 for SRO-I) ; (3 or 2 for SRO-U)

i. Swap CRD Stabilizing Valves D,R 1

KIA 201001 A2.08 (2.8/2.8), N1-0P-5

j. Lineup Lake Water to Supply the Emergency Condenser D,E,A,R Makeup Tanks using the Electric Fire Pump 4

KIA 207000 A1.01 (3.7/3.8), N1-SOP-21.2

k. Supply Emergency Cooling Water to EDG from the Diesel Fire Pump D,E,R 8

KIA 400000 K1.02 (3.2/3.4), N1-0P-45

All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

  • Type Codes (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power / Shutdown (N)ew or (M)odified from bank including 1 (A)

(P)revious 2 exams (R)CA (S)imulator Pairings:

A then B Ethen F Criteria for RO / SRO-I / SRO-U 4-6 I 4-6 I 2-3 S9/S8/S4

~1/~1/~1

~ 1 / ~1 / ~1 (control room system)

~1/~1/~1

~2/~2/~1 s 3 Is 3 Is 2 (randomly selected)

~1/~1/~1

Appendix D Scenario Outline Form ES-D-1 Facility: Nine Mile Point Unit 1 Scenario No.: NRC-2 Op-Test No.: LC1 17-1 NRC Examiners:

Operators:

Initial Conditions: The plant is operating at approximately 90% power. Containment Spray Pump 112 is out of service for maintenance. Steam Packing Exhauster 12 is out of service due to high vibrations.

PB 11 is aligned to reserve power in preparation for cross-tying PB 16.

Turnover: Cross-tie PB 16A to PB 16B with PB 168 supplying. Power board 11 will remain aligned to reserve power. Then, raise reactor power to 95% using recirc flow.

Event Malf.

Event Event No.

No.

Type*

Description N-BOP, Cross-tie PB 16A to PB 16B 1

N/A SRO N1-0P-30 2

N/A R-ATC, Raise reactor power with recirc.

SRO N1-0P-1 3

RD02 C-ATC, Control Rod 26-35 Drift~ Out SRO N1-SOP-5.2 C-Feedwater Booster Pump 11 Trips with Failure of standby FW02A

BOP, 4

SRO Feedwater Booster Pump to Auto-start Override N 1-SOP-16.1, Technical Specifications TS-SRO C-AII Respond to trip of Reactor Protection System (RPS) UPS 172 5

RP25 TS-SRO Technical Specification N 1-SOP-40.1 RWCU break in the Secondary Containment requiring scram; 6

CU11 M-AII RWCU Isolation Valves to isolate N1-EOP-2, N1-EOP-5, N1-EOP-8 C-Mode Switch Fails to Scram 7

Overrides

ATC, SRO N1-SOP-1 (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor

Facility: Nine Mile Point Unit 1 Scenario No.: NRC-2 Op-Test No.: LC117-1 NRC

1. Malfunctions after EOP entry (1-2) 1 Events 7
2. Abnormal events (2-4) 3 Events 3, 4, 5
3. Major transients (1-2) 1 Event 6
4. EOPs entered/requiring substantive actions (1-2) 2 N1-EOP-2, N1-EOP-5
5. EOP contingencies requiring substantive actions (0-2) 1 N1-EOP-8
6. Preidentified Critical tasks (2-3) 2 CRITICAL TASK DESCRIPTIONS:

CRITICAL TASK JUSTIFICATION:

CT-1.0: Given an un-isolable RWCU leak outside primary containment and With an un-isolable primary system one general area temperature above the maximum safe limit, the crew will discharging outside of Primary insert a manual reactor scram, in accordance with N1-EOP-5.

Containment resulting in general area temperature above the maximum safe limit, the Reactor must be scrammed.

This reduces the rate of energy production and thus the heat input, radioactivity release, and break flow into the Secondary Containment. This also ensures the Reactor is shutdown prior to need for a blowdown.

CT-2.0: Given an un-isolable RWCU leak outside primary containment and An un-isolable primary system two general area temperatures above the maximum safe limit, the crew will discharging outside of Primary execute N1-EOP-8, RPV Slowdown, in accordance with N1-EOP-5.

Containment resulting in two general area temperatures above the maximum safe limit indicates a wide-spread problem posing a direct and immediate threat to I

Secondary Containment. A blowdown minimizes flow through the break, rejects heat to the suppression pool in preference to outside the containment, and places the primary system in the lowest possible energy state.

Appendix D Scenario Outline Form ES-D-1 Facility: Nine Mile Point Unit 1 Scenario No.: NRC-3 Op-Test No.: LC1 17-1 NRC Examiners:

Operators:

Initial Conditions: The plant is operating at approximately 100% power. Containment Spray Pump 112 is out of service for maintenance. Steam Packing Exhauster 12 is out of service due to high vibrations.

Turnover: Reduce reactor power to 98% with recirc flow. Then, start TBCLC Pump 12 and secure TBCLC Pump 11.

Event Malf.

Event Event No.

No.

Type*

Description 1

N/A R-ATC, Lower reactor power to 98% with recirc flow SRO N1-0P-1 N-Swap Running TBCLC Pumps 2

N/A

BOP, SRO (2017 Scenario 4), N 1-0P-24 C-BOP, Powerboard 101 fault 3

ED06 SR::>

N1-SOP-1.3 I

I-SF 0 Reactor Pressure Instrument 36-0?C Fails Low I

4 RP17B TS-SRO Technical Specifications I-ATC, EPR Oscillation 5

TC06 SRO TS-SRO N1-SOP-31.1, Technical Specifications CW04A All RBCLC Pumps Trip, Motor Driven Feedwater Pumps Fail to 6

CW04B C-AII Operate and 13 FW Pump clutch disengages CW04C (2015 Scenario 5), N1-SOP-11.1, N1-SOP-1, N1-EOP-2 CU01 Coolant Leak Inside Primary Containment 7

M-AII EC01 (2015 Scenario 5), N 1-EOP-2, N 1-EOP-4 8

VICP201 M-AII Fuel Zone Level Instrument Sporadic Indication 68/69 (2015 Scenario 5), N1-EOP-2, N1-EOP-7 (N)ormal, (R)eactivitv, (l)nstrument, (C)omoonent, (M)aior

Facility: Nine Mile Point Unit 1 Scenario No.: NRC-3 Op-Test No.: LC117-1 NRC

2. Malfunctions after EOP entry (1-2) 2 Events 7, 8
3. Abnormal events (2-4) 3 Events 3, 5, 6
4. Major transients (1-2) 2 Events 7 & 8
5. EOPs entered/requiring substantive actions (1-2) 2 N1-EOP-2, N1-EOP-4
6. EOP contingencies requiring substantive actions (0-2) 1 N1-EOP-7
7. Preidentified Critical tasks (2-3) 2 CRITICAL TASK DESCRIPTIONS:

CRITICAL TASK JUSTIFICATION:

CT-1.0: Given a LOCA in the Drywell with the inability to maintain Initiating Containment Sprays reduces containment parameters within the Pressure Suppression Pressure limit, Primary Containment pressure. This initiate Containment Sprays, in accordance with N1-EOP-4.

reduces stresses on the Drywell and Torus, assists in avoiding "chugging" that may cause fatigue failure of the LOCA downcomers, and avoids the need for a blowdown. These benefits reduce challenges to the fuel cladding, the RPV, and the Primary Containment.

CT-2.0: Given the plant with RPV water level unknown, execute N1-EOP-7, With Reactor water level unknown, the RPV Flooding, in accordance with N1-EOP-2.

status of core cooling is unknown. RPV flooding is required to establish conditions to cool the core. This protects the fuel cladding integrity.

Appendix D Scenario Outline Form ES-D-1 Facility: Nine Mile Point Unit 1 Scenario No.: NRC-4 Op-Test No.: LC1 17-1 NRC Examiners:

Operators:

Initial Conditions: A plant startup is in progress with reactor power approximately 2-3%. Containment Spray Pump 112 is out of service for maintenance. Steam Packing Exhauster 12 is out of service due to high vibrations.

Turnover: Continue power ascension by withdrawing control rods.

nt Malf.

Event Event I

No.

No.

Type*

Description N/A R-ATC, Raise power with control rods 1

SRO N 1-0P-43A, N 1-0P-5 C-ATC, Control Rod Double Notches 2

RD42 SRO N1-0P-5 RR06A I-ATC, IRM Downscale Failure 3

SRO N 1-SOP-1.2, RR07A C-B01, Powerboard 16A Electrical Fault 4

ED12A SRO (2015 Scenario 5), ARP L4-3-6, N 1-EOP-4 I

RR06A C-BOP, Recirc Pump 11 seal failure requiring isolation of the pump 5

SRO N1-SOP-1.2, Technical Specification 3.2.5, 3.1.7.e RR07A TS-SRO PCOS C-BOP, Seismic Event; lsolable Leak on Containment Spray Suction Line 6

SRO CT04A TS-SRO N1-SOP-28, N1-EOP-5, Technical Specifications Second Seismic Event; Torus Break; Multiple Control Rods Fail to PCOS Insert 7

PC04 M-AII N1-EOP-5, N1-EOP-4, N1-SOP-1, N1-EOP-2, N1-EOP-8, N1-EOP-3 CT02B Containment Spray Raw Water Pumps 112 and 121 Trips 8

C-AII CT02C N1-EOP-4 (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor

Facility: Nine Mile Point Unit 1 Scenario No.: NRC-4 Op-Test No.: LC117-1 NRC

1. Malfunctions after EOP entry (1-2) 1 Events 8
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) 3 N1-EOP-2, N1-EOP-4, N1-EOP-5
5. EOP contingencies requiring substantive actions (0-2) 2 N1-EOP-3, N1-EOP-8
6. Preidentified Critical tasks (2-3) 2 CRITICAL TASK DESCRIPTIONS:

CRITICAL TASK JUSTIFICATION:

CT-1.0: Given an un-isolable Torus leak exceeding makeup capacity, scram Lowering Torus water level challenges the Reactor, in accordance with N1-EOP-4.

the pressure suppression function of the Primary Containment. Continued Reactor operation is not allowed with an inoperable Primary Containment. A Reactor scram also allows subsequent mitigating actions, such as Reactor cooldown and/or blowdown.

CT-2.0: Given an un-isolable Torus leak exceeding makeup capacity, If torus water level lowers below the perform an RPV Slowdown, in accordance with N1-EOP-4.

elevation of the ERV discharge holes, opening ERVs would discharge steam directly into the torus airspace. The resulting pressure increase could exceed the maximum pressure capability of the Primary Containment. Since the RPV i

may not be kept at pressure under these conditions, a blowdown is required.

ES-401 1

Form ES-401-1 Facility:

Nine Mile Point Unit 1 Date of Exam:

December 2018 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 3

4 2

4 3

4 20 3

4 7

Emergency and 2

1 1

1 NIA 2

1 N/A 1

7 2

1 3

Abnormal Plant Evolutions Tier Totals 4

5 3

6 4

5 27 5

5 10

2.

1 3

2 3

3 2

2 2

3 2

2 2

26 3

2 5

Plant 2

1 1

1 1

1 2

0 1

2 1

1 12 0

1 2

3 Systems Tier Totals 4

3 4

4 3

4 2

4 4

3 3

38 4

4 8

3. Generic Knowledge and Abilities 1

2 3

4 10 1

2 3

4 7

Categories 3

2 2

3 2

1 2

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, the "Tier Totals" in each KIA category shall not be less than two). (One Tier 3 radiation control KIA is allowed if it is replaced by a KIA from another Tier 3 category.)

2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRG revisions.

The final RO exam must total 75 points, and the SRO-only exam must total 25 points.

3. Systems/evolutions within each group are identified on the 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 of inappropriate KIA statements.
4. Select topics from as many systems and evolutions as possible. Sample every system or evolution in the group before selecting a second topic for any system or evolution.
5.

Absent a plant-specific priority, only those Kl As 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 2 from 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 KIAs.
8. On the following pages, enter the KIA numbers, a brief description of each topic, the topics' I Rs for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above. If fuel-handling equipment is sampled in 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 KIA numbers, 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 Kl As

. 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 not required to be included when using earlier revisions of 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 ES-401 BWR Examination Outline F

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

E/APE #/Name / Safetv Function K1 K2 K3 A1 A2 G*

KIA ToolCISl IA Q#

295001 (APE 1) Partial or Complete Loss of X

AA2.02, Ability to determine and/or interpret 3.1 27 Forced Core Flow Circulation/ 1 & 4 the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: Neutron monitorina 295003 (APE 3) Partial or Complete Loss of X

AA 1.01, Ability to operate and/or monitor the 3.7 29 AC Power /6 following as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER: A.C.

electrical distribution system 295004 (APE 4) Partial or Complete Loss of X

AK1.04, Knowledge of the operational 2.8 30 DC Power /6 implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER: Effect of battery discharge rate on capacity 295005 (APE 5) Main Turbine Generator Trip/

X G.2.4.31, Knowledge of annunciator alarms, 4.2 31 3

indications, or response procedures.

295006 (APE 6) Scram/ 1 X

AK2.06, Knowledge of the interrelations 4.2 32 between SCRAM and the following: Reactor power X G2.1.19, Ability to use plant computers to 3.8 76 evaluate system or component status.

295016 (APE 16) Control Room Abandonment X G2.4.35, Knowledge of local auxiliary 3.8 33

/7 operator tasks during an emergency and the resultant operational effects.

295018 (APE 18) Partial or Complete Loss of X

AK 1.01, Knowledge of the operational 3.5 34 CCW/8 implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER: Effects on componenVsystem operations 295019 (APE 19) Partial or Complete Loss of X

AK3.03, Knowledge of the reasons for the 3.2 35 Instrument Air/ 8 following responses as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR: Service air isolations:

Plant-Specific 295021 (APE 21) Loss of Shutdown Cooling/

X G2.2.37, Ability to determine operability 3.6 36 4

and/or availability of safety related equipment.

X AA2.06, Ability to determine and/or 3.3 n

Interpret the following as they apply to LOSS OF SHUTDOWN COOLING: Reactor pressure 295023 (APE 23) Refueling Accidents / 8 X

AK1.01, Knowledge of the operational 3.6 37 implications of the following concepts as they apply to REFUELING ACCIDENTS:

Radiation exposure hazards X

G2.2.37, Ability to determine operability 4.6 78 and/or availability of safety related equipment.

295024 High Drywall Pressure/ 5 X

EK2.15, Knowledge of the interrelations 3.8 38 between HIGH DR'l'WELL PRESSURE and the following: Containment spray logic:

Plant-Specific

ES-401 3

Form ES-401-1 295025 (EPE 2) High Reactor Pressure/ 3 X G2.4.8, Knowledge of how abnormal 3.8 39 operating procedures are used in conjunction with EOPs.

295026 (EPE 3) Suppression Pool High Water X

EA 1.03, Ability to operate and/or monitor the 3.9 28 Temperature/ 5 following as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE:

Temperature monitoring X

EA2.01, Ability to determine and/or 4.2 79 interpret the following as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Suppression pool water temperature 295028 (EPE 5) High Drywell Temperature X

EK3.06, Knowledge of the reasons for the 3.4 40 (Mark I and Mark II only)/ 5 following responses as they apply to HIGH DRYWELL TEMPERATURE: ADS EA2.01, Ability to determine and/or X

interpret the following as they apply to 4.1 80 HIGH DRYWELL TEMPERATURE:

Drywell temperature 295030 (EPE 7) Low Suppression Pool Water X

EA2.04, Ability to determine and/or Interpret 3.5 41 Level/ 5 the following as they apply to LOW SUPPRESSION POOL WATER LEVEL:

Drywall/ suppression chamber differential pressure: Mark I & II 295031 (EPE 8) Reactor Low Water Level/ 2 X

EA 1. 13, Ability to operate and/or monitor the 4.3 42 following as they apply to REACTOR LOW WATER LEVEL: Reactor water level control 295037 (EPE 14) Scram Condition Present X

EA 1.01, Ability to operate and/or monitor the 4.6 43 and Reactor Power Above APRM Downscale following as they apply to SCRAM or Unknown / 1 CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: Reactor Protection System 295038 (EPE 15) High Offsite Radioactivity X

EK2.08, Knowledge of the interrelations 2.6 44 Release Rate I 9 between HIGH OFF-SITE RELEASE RATE and the following: SPDS/ERIS/CRIDS/GDS:

Plant-Specific.

X G2.4.41, Knowledge of the emergency 4.6 81 action level thresholds and classifications.

600000 (APE 24) Plant Fire On Site / 8 X

M2.05, Ability to determine and interpret 2.9 45 the following as they apply to PLANT FIRE ON SITE: Ventilation alignment necessary to secure affected area X

2.1.7, Ability to evaluate plant 4.7 82 performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

700000 (APE 25) Generator Voltage and X

AK2.07, Knowledge of the interrelations 3.6 46 Electric Grid Disturbances / 6 between GENERA TOR VOLTAGE AND ELECTRIC GRID DISTURBANCES and the followinq: Turbine/generator control Totals:

3 4

2 4

3/3 4/4 RO/SRO Group Point Total:

20n

ES-401 4

Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Emeraencv and Abnormal Plant Evolutions-Tier 1/Group 2 (RO/SRO)

E/APE #/Name I Safetv Function K1 K2 K3 A1 A2 G*

KIA Topic(s)

IR Q#

295002 (APE 2) Loss of Main Condenser X

AK1.03, Knowledge of the operational 3.6 47 Vacuum/3 implications of the following concepts as they apply to LOSS OF MAIN CONDENSER VACUUM: Loss of heat sink 295007 (APE 7) High Reactor Pressure/ 2 X

G2.2.42, Ability to recognize system 4.6 83 parameters that are entry-level conditions for Technical Specifications.

295009 (APE 9) Low Reactor Water Level/ 2 X

AA 1.04, Ability to operate and/or monitor 2.7 48 the following as they apply to LOW REACTOR WATER LEVEL: Reactor water cleanup 295012 (APE 12) High Drywell Temperature I X

AK3.01, Knowledge of the reasons for the 3.5 49 5

following responses as they apply to HIGH DRYWELL TEMPERATURE: Increased drywall cooling 295013 (APE 13) High Suppression Pool X

G2.4.20, Knowledge of the operational 3.8 50

'Temperature. I 5 implications of EOP warnings, cautions, and notes.

295015 (APE 15) Incomplete Scram/ 1 X

AA 1.05, Ability to operate and/or monitor 2.5 51 the following as they apply to INCOMPLETE SCRAM: Rod worth minimizer: Plant-Specific 295022 (APE 22) Loss of Control Rod Drive X

AK2.07, Knowledge of the interrelations 3.4 52 Pumps/ 1 between LOSS OF CAD PUMPS and the following: Reactor pressure (SCRAM assist): Plant-Specific 295029 (EPE 6) High Suppression Pool Water X

EA2.03, Ability to determine and/or 3.4 53 Level/ 5 interpret the following as they apply to HIGH SUPPRESSION POOL WATER LEVEL: DrvwelVcontainment water level 295033 (EPE 10) High Secondary X

EA2.03, Ability to determine and/or 4.2 84 Containment Area Radiation Levels / 9 interpret the following as they apply to HIGH SECONDARY CONTAINMENT AREA RADIATION LEVELS: Cause of high area radiation 295036 (EPE 13) Secondary Containment X

EA2.03, Ability to determine and/or 3.8 85 High Sump/Area Water Level/ 5 interpret the following as they apply to SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL: Cause of the high water level KJA Category Point Totals:

1 1

1 2

1/2 1/1 RO/SRO Group Point Total:

7/3

ES-401 5

Form ES-401-1 ES-401 BWR Examination Outline Form ES-401-1 Plant Systems-Tier 2/Group 1 (RO/SRO)

System # / Name K K K K4 K

K A A A A G KIA Topic(s)

IR Q#

1 2

3 5

6 1

2 3

4.

205000 (SF4 SCS) Shutdown Cooling X

K3.02, Knowledge of the effect that a loss or 3.2 6

malfunction of the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) will have on following: Reactor water level: Plant-Specific X

A2.02, Ability to (a) predict the impacts of 2.7 86 the following on the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Low shutdown cooling suction pressure:

Plant-Specific 206000 (SF2, SF4 HPCIS)

X K2.01, Knowledge of electrical power 3.2 3

High-Pressure Coolant Injection supplies to the following: System valves:

BWR-2,3,4 X

K4.07, Knowledge of HIGH PRESSURE 4.3 24 COOLANT INJECTION SYSTEM design feature(s) and/or interlocks which provide for the following: Automatic system initiation:

BWR-2,3,4 207000 (SF4 IC) Isolation X

K6.04, Knowledge of the effect that a loss or 3.2 11 (Emergency) Condenser malfunction of the following will have on the ISOLATION (EMERGENCY) CONDENSER:

Plant air systems: BWR-2,3 209001 (SF2, SF4 LPCS)

X K1.09, Knowledge of the physical 3.2 1

Low-Pressure Core Spray connections and/or cause-effect relationships between LOW PRESSURE CORE SPRAY SYSTEM and the following: Nuclear boiler instrumentation X

A2.01, Ability to (a) predict the impacts of 3.4 87 the following on the LOW PRESSURE CORE SPRAY SYSTEM; and {b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Pump trips 21 iOOO (SFi SLCS) Standby Liquid X

A 1.03, Ability to predict and/or monitor 3.6 13 Control changes in parameters associated with operating the STANDBY LIQUID CONTROL SYSTEM controls including: Pump discharge pressure

ES-401 6

Form ES-401-1 212000 (SF7 RPS) Reactor X

K5.02, Knowledge of the operational 3.3 9

Protection implications of the following concepts as they apply to REACTOR PROTECTION SYSTEM:

Specific logic arrangements X 2.4.46, Ability to verify that the alarms are 4.2 88 consistent with the plant conditions.

215003 (SF7 IRM)

X K5.01, Knowledge of the operational 2.6 10 I nterrnediate-Range Monitor implications of the following concepts as they apply to INTERMEDIATE RANGE MONITOR (1AM) SYSTEM: Detector operation 215004 (SF? SAMS) Source-Range X

K4.02, Knowledge of SOURCE RANGE 3.4 7

Monitor MONITOR (SAM) SYSTEM design feature(s) and/or interlocks which provide for the following: Reactor SCRAM signals 215005 (SF7 PRMS) Average Power X

A 1.07, Ability to predict and/or monitor 3.0 14 Range Monitor/Local Power Range changes in parameters associated with Monitor operating the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM controls including:

APRM (gain adjustment factor)

X G2.1.32, Conduct of Operations: Ability to 3.8 22 explain and apply all system limits and precautions.

218000 (SF3 ADS) Automatic X

A2.01, Ability to (a) predict the impacts of the 4.1 16 Depressurtzation following on the AUTOMATIC DEPRESSURIZATION SYSTEM; and (b) based on those predictions. use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Small steam line break LOCA 223002 (SF5 PCIS) Primary X

A3.01, Ability to monitor automatic operations 3.4 17 Containment Isolation/Nuclear Steam of the PRIMARY CONTAINMENT Supply Shutoff ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF including: System indicating lights and alarms X

A2.04, Ability to (a) predict the impacts of 3.2 89 the following on the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Process radiation monitoring system failures

ES-401 7

Form ES-401-1 239002 (SF3 SRV) Safety Relief X

A2.01, Ability to (a) predict the impacts of the 3.0 15 Valves following on the RELIEF/SAFETY VALVES; and {b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Stuck open vacuum breakers X

K1.07, Knowledge of the physical 3.6 23 connections and/or cause-effect relationships between RELIEF/SAFETY VALVES and the following: Suppression Pool 259002 (SF2 RWLCS) Reactor Water X

A4.01, Ability to manually operate and/or 3.8 20 Level Control monitor in the control room: All individual component controllers in the manual mode X G2.4.9, Knowledge of low power/

4.2 90 shutdown implications in accident (e.g.,

loss of coolant accident or loss of residual heat removal) mitigation strategies.

261000 (SF9 SGTS) Standby Gas X

K6.03, Knowledge of the effect that a loss or 3.0 12 Treatment malfunction of the following will have on the STANDBY GAS TREATMENT SYSTEM:

Emergency diesel generator system X

A2.07, Ability to (a) predict the impacts of the 2.7 25 following on the STANDBY GAS TREATMENT SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

A.G. electrical failure 262001 (SF6 AC) AC Electrical X

K4.03, Knowledge of A.G. ELECTRICAL 3.1 8

Distribution DISTRIBUTION design feature(s) and/or interlocks which provide for the following:

Interlocks between automatic bus transfer and breakers 262002 (SF6 UPS) Uninterruptable X

K3.1 O - Knowledge of the effect that a loss or 2.7 5

Power Supply (AC/DC) malfunction of the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) will have on following: Containment isolation: Plant-Specific 263000 (SF6 DC) DC Electrical X

A3.01, Ability to monitor automatic operations 3.2 18 Distribution of the D.C. ELECTRICAL DISTRIBUTION including: Meters, dials, recorders, alarms, and indicating lights X

K3.03, Knowledge of the effect that a loss or 3.4 26 malfunction of the D.C. ELECTRICAL DISTRIBUTION will have on following:

Systems with D.C. components {I.e. valves, motors, solenoids, etc.)

ES-401 8

Form ES-401 *1 264000 (SF6 EGE) Emergency X

A4.04, Ability to manually operate and/or 3.7 19 Generators (DieseVJet) EOG monitor in the control room: Manual start, loading, and stopping of emergency generator: Plant-Specific 300000 (SFB IA) Instrument Air X

K2.02, Knowledge of electrical power 3.0 4

supplies to the following: Emergency air compressor X 2.1.30, Conduct of Operations: Ability to 4.4 21 locate and operate components, including local controls.

400000 (SFB CCS) Component X

K1.02, Knowledge of the physical 3.2 2

Cooling Water connections and I or cause-effect relationships between CCWS and the following: Loads cooled by CCWS KIA Category Point Totals:

3 2

3 3

2 2

2 3/ 2 2 21 RO/SRO Group Point Total:

26/5 3

2

ES-401 9

Form ES-401-1 ES-401 BWR Examination Outline Fo-- r-.-. in....

Plant Svstems-Tier 2/Grouo 2 (RO/SRO)

System # / Name K

K K

K K

K A A A A G*

KIA Topic(s)

IA Q#

1 2

3 4

5 6

1 2

3 4

201002 (SF1 RMCS) Reactor Manual Control X

A2.01, Ability to (a) predict the 2.8 91 Impacts of the following on the REACTOR MANUAL CONTROL SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Rod movement sequence timer malfunctions 201003 (SF1 CROM) Control Rod and Drive X G2.2.38, Knowledge of conditions 4.5 92 Mechanism and limitations In the facility license.

201006 (SF7 RWMS) Rod Worth Minimizer X

A2.01, Ability to (a) predict the 2.5 54 impacts of the following on the ROD WORTH MINIMIZER SYSTEM (RWM) (PLANT SPECIFIC); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Power supply loss: P-Spec(Not-BWR6) 202001 (SF1, SF4 RS) Recirculation X G2.2.36, Ability to analyze the 4.2 93 effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

202002 (SF1 RSCTL) Recirculation Flow X

A3.03, Ability to monitor automatic 3.1 55 Control operations of the RECIRCULATION FLOW CONTROL SYSTEM including: Scoop tube operation:

BWR-2,3,4 215001 (SF7 TIP) Traversing In-Core Probe X

K4.01, Knowledge of TRAVERSING 3.4 56 IN-CORE PROBE design feature(s) and/or interlocks which provide for the following: Primary containment isolation: Mark I & II (Not-BWR1) 223001 (SF5 PCS) Primary Containment and X G2.4.3, Ability to identify post-3.7 57 Auxiliaries accident instrumentation.

226001 (SFS AHR CSS) RHR/LPCI:

X K6.11, Knowledge of the effect that 2.8 58 Containment Spray Mode a loss or malfunction of the following will have on the RHR/LPCI:

CONTAINMENT SPRAY SYSTEM MODE: Component cooling water svstems 245000 (SF4 MTGEN) Main Turbine X

A4.09, Ability to manually operate 2.6 59 Generator/ Auxiliary and/or monitor in the control room:

Hydrogen seal oil pressure

ES-401 10 Form ES-401-1 259001 (SF2 FWS) Feedwater X

K6.06, Knowledge of the effect that 2.7 60 a loss or malfunction of the following will have on the REACTOR FEEDWATER SYSTEM: Plant service water 268000 (SF9 AW) Radwaste X

K1.05, Knowledge of the physical 2.9 61 connections and/or cause-effect relationships between RADWASTE and the following: Drywall equipment drains 272000 (SF7, SF9 RMS) Radiation Monitoring X

K2.05, Knowledge of electrical*

2.6 62 power supplies to the following:

Reactor building ventilation monitors: Plant-Specific 288000 (SF9 PVS) Plant Ventilation X

KS.02, Knowledge of the operational 3.2 63 implications of the following concepts as they apply to PLANT VENTILATION SYSTEMS:

Differential pressure control 290001 (SF5 SC) Secondary Containment X

A3.01, Ability to monitor automatic 3.9 64 operations of the SECONDARY CONTAINMENT including:

Secondary containment isolation 290002 (SF4 RVI) Reactor Vassel Internals X

K3.01, Knowledge of the effect that 3.2 65 a loss or malfunction of the REACTOR VESSEL INTERNALS will have on following: Reactor water level KIA Category Point Totals:

1 1

1 1

1 2

0 1/ 2 1

1/ RO/SRO Group Point Total:

12/3 1

2

ES-401 Generic Knowledge and Abilities Outline (Tier 3)

Form ES-401-3 Facility: Nine Mile Point Unit 1 Date of Exam: January 2019 Category KIA#

Topic RO SRO-only IR Q#

IA Q#

2.1.4 Knowledge of individual licensed operator 3.3 66 responsibilities related to shift staffing, such as medical requirements, "no-solo" operation, maintenance of active license status, 10CFR55, etc.

2.1.37 Knowledge of procedures, guidelines, or limitations 4.3 67 associated with reactivit~ management.

2.1.31 Ability to locate control room switches, controls, 4.3 94

1. Conduct of and indications, and to determine that they Operations correctly reflect the desired plant lineup.

2.1.40 Knowledge of refueling administrative 3.9 95 requirements.

2.1.43 Ability to use procedures to determine the effects on 4.1 68 reactivity of plant changes, such as reactor coolant system temperature, secondary plant, fuel depletion, etc.

Subtotal 3

2 2.2.2 Ability to manipulate the console controls as required 4.6 69 to operate the facility between shutdown and desionated power levels.

2. Equipment 2.2.21 Knowledge of pre-and post-maintenance 4.1 96 Control operability requirements.

2.2.22 Knowledge of limiting conditions for operations and 4.0 70 safety limits.

Subtotal 2

1 2.3.5 Ability to use radiation monitoring systems, such as 2.9 71 fixed radiation monitors and alarms, portable survey instruments, personnel monitorinq equipment, etc.

2.3.7 Ability to comply with radiation work permit 3.6 97 requirements during normal or abnormal conditions.

3. Radiation 2.3.13 Knowledge of radiological safety procedures pertaining 3.4 72 Control to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked hiqh-radiation areas, alionina filters, etc.

2.3.14 Knowledge of radiation or contamination hazards 3.8 98 that may arise during normal, abnormal, or emergency conditions or activities.

Subtotal 2

2

4. Emergency 2.4.14 Knowledae of aeneral auidelines for EOP usage.

3.8 73 Procedures/Plan 2.4.17 Knowledge of EOP terms and definitions.

4.3 99

ES-401 Generic Knowledge and Abilities Outline (Tier 3)

Form ES-401-3 2.4.21 Knowledge of the parameters and logic used to 4.6 100 assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

2.4.32 Knowledge of operator response to loss of all 3.6 74 annunciators.

2.4.22 Knowledge of the bases for prioritizing safety functions 3.6 75 during abnormal/emergency operations.

Subtotal 3

2 Tier 3 PointT otal 10 7

ES-401 Record of Rejected K/A's Form ES-401-4 The following Kl As were rejected following the systematic and random sampling process:

---~-

Question 9 A discriminating question could not be developed without testing generic fundamentals 212000 RPS knowledge.

2/1 K5.01 - Knowledge of the Randomly reselected KIA 212000 RPS K5.02 -

operational implications of Knowledge of the operational implications of the the following concepts as following concepts as they apply to REACTOR they apply to REACTOR PROTECTION SYSTEM: Specific logic PROTECTION SYSTEM:

arrangements.

Fuel thermal time constant

-~ *-* -*------*-*-**------**-' **-*------ --------*-

Question 14 This facility does not have recirculation flow control valves.

215005 APRM / LPRM Randomly reselected Kl A 215005 A 1.07 - Ability A 1.06 - Ability to predict to predict and/or monitor changes in parameters and/or monitor changes in associated with operating the AVERAGE parameters associated with POWER RANGE MONITOR/LOCAL POWER 2 / 1 operating the AVERAGE RANGE MONITOR SYSTEM controls including:

POWER RANGE APRM (gain adjustment factor).

MONITOR/LOCAL POWER RANGE MONITOR SYSTEM controls including:

Recirculation flow control valve position: Plant-Specific Question 23 An acceptable question could not be written for the randomly selected KIA due to limited 239002 SRVs interrelations between SRVs and plant air systems.

K1.05 - Knowledge of the i

physical connections Randomly reselected KIA 239002 K1.07 -

and/or cause-effect Knowledge of the physical connections and/or relationships between cause-effect relationships between 2 / 1 RELIEF/SAFETY VALVES RELIEF/SAFETY VALVES and the following:

and the following: Plant air Suppression Pool.

systems: Plant-Specific

ES-401 Record of Rejected K/A's Form ES-401-4 Question 25 There are no interlocks *or initiations for RBEVS (SGTS) related to high system pressure at this 261000 SGTS facility. An acceptable question could not be developed without testing minutia.

A2.14 - Ability to (a) predict the impacts of the following Randomly reselected KIA 261000 A2.07 - Ability on the STANDBY GAS to (a) predict the impacts of the following on the 2 / 1 TREATMENT SYSTEM; STANDBY GAS TREATMENT SYSTEM; and (b) and (b) based on those based on those predictions, use procedures to predictions, use procedures correct, control, or mitigate the consequences of to correct, control, or those abnormal conditions or operations: A.C.

mitigate the consequences electrical failure.

of those abnormal conditions or operations:

High system pressure:

Plant-Specific Question 28 295001 was inadvertently sampled twice on the RO exam prior to sampling 295026.

295001 Partial or Complete Loss of Forced Core Flow Reselected 295026 Suppression Pool High Circulation Water Temperature and randomly reselected EA 1.03 - Ability to operate and/or monitor the 1 / 1 AA 1.03 - Ability to operate following as they apply to SUPPRESSION POOL and/or monitor the following HIGH WATER TEMPERATURE: Temperature as they apply to PARTIAL monitoring.

OR COMPLETE LOSS OF FORCED CORE FLOW Cl RCULATION: RMCS:

Plant-Specific Question 36 An acceptable question could not be developed for the randomly sampled KIA due to lack of 295021 Loss of Shutdown Technical Specification bases related to loss of Cooling Shutdown Cooling that are RO level.

1 / 1 2.2.25 - Knowledge of the Randomly reselected KIA 295021 Loss of bases in Technical Shutdown Cooling 2.2.37 - Ability to determine Specifications for limiting operability and/or availability of safety related conditions for operations equipment.

and safety limits.

Question 40 An acceptable question could not be developed for the randomly sampled KIA without 295028 High Drywell overlapping Question 49.

Temperature Randomly reselected KIA 295028 High Drywall 1 / 1 EK3.04 - Knowledge of the Temperature EK3.06 - Knowledge of the reasons reasons for the following for the following responses as they apply to responses as they apply to HIGH DRYWELL TEMPERATURE: ADS.

HIGH DRYWELL TEMPERATURE:

Increased drywall cooling

ES-401 Record of Rejected K/A's Form ES-401-4 Question 63 An acceptable question could not be developed for the randomly sampled KIA without testing 288000 Plant Ventilation minutia due to a lack of operationally relevant references related to Plant Ventilation K5.03 - Knowledge of the temperature control.

2/2 operational implications of the following concepts as Randomly reselected KIA 288000 Plant they apply to PLANT Ventilation KS.02 - Knowledge of the operational VENTILATION SYSTEMS:

implications of the following concepts as they Temperature control apply to PLANT VENTILATION SYSTEMS:

Differential pressure control.

Question 64 An acceptable question could not be developed for the randomly sampled KIA due to lack of 290001 Secondary Secondary Containment controls and associated Containment system lineups.

2/2 A 1.01 - Ability to predict Randomly reselected KIA 290001 Secondary and/or monitor changes in Containment A3.01 - Ability to monitor automatic parameters associated with operations of the SECONDARY CONTAINMENT operating the including: Secondary containment isolation.

SECONDARY CONTAINMENT controls including: System lineups Question 67 An acceptable question could not be developed for the randomly sampled KIA without 2.1.19 ~ Ability to use plant oversampling plant computer topics (see computers to evaluate Questions 44 & 76). Use of plant computers is 3

system or component also tested extensively on the operating exam.

status.

Randomly reselected KIA 2.1.37 - Knowledge of procedures, guidelines, or limitations associated with reactivity management.

Question 75 The randomly sampled generic KIA is also tested on Question 33. Reselecting for better balance 2.4.35 - Knowledge of local of coverage.

3 auxiliary operator tasks during an emergency and Randomly reselected KIA 2.4.22 - Knowledge of the resultant operational the bases for prioritizing safety functions during effects.

abnormal/emergency operations.

Question 88 An acceptable question could not be developed for the randomly sampled KIA due to lack of RO 212000 Reactor Protection tasks performed outside the control room related to the Reactor Protection System.

2 / 1 2.4.34 - Knowledge of RO tasks performed outside Randomly reselected KIA 212000 Reactor the main control room Protection 2.4.46 - Ability to verify that the alarms during an emergency and are consistent with the plant conditions.

the resultant operational effects.

ES-401 Record of Rejected K/A's Form ES-401-4 Question 91 An acceptable question could not be developed for the randomly sampled KIA without 201002 Reactor Manual overlapping the operating exam. Additionally, the Control KIA did not readily support testing at the SRO level.

A2.02 - Ability to (a) predict the impacts of the following Randomly reselected KIA 201002 Reactor on the REACTOR Manual Control A2.01 - Ability to (a) predict the 2/2 MANUAL CONTROL impacts of the following on the REACTOR SYSTEM; and (b) based on MANUAL CONTROL SYSTEM; and (b} based on those predictions, use those predictions, use procedures to correct, procedures to correct, control, or mitigate the consequences of those control, or mitigate the abnormal conditions or operations: Rod consequences of those movement sequence timer malfunctions.

abnormal conditions or operations: Rod drift alarm Question 93 An acceptable question could not be developed for the randomly sampled Kl A due to lack of 202001 Recirculation surveillance procedures for the Recirculation system.

2/2 2.2.12 - Knowledge of surveillance procedures.

Randomly reselected KIA 202001 Recirculation 2.2.36 - Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

Question 83 Editorial error - APE 295008 does not coincide with High Reactor Pressure.

1 / 2 295008 High Reactor Pressure Conferred with Chief Examiner to change 295008 to 295007 to coincide with High Reactor Pressure

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