ML21033A538

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Final Written Examination and Operating Test Outlines (Folder 3)
ML21033A538
Person / Time
Site: Nine Mile Point Constellation icon.png
Issue date: 02/02/2021
From: Brian Fuller
Operations Branch I
To: Isham P
Exelon Nuclear Generation Corp
Fuller B
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ML19309G108 List:
References
CAC 00500
Download: ML21033A538 (32)
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ES-301 Administrative Topics Outline Form ES-301-1 Facility: Nine Mile Point Unit 1 Date of Examination: December 2020 Examination Level: RO Operating Test Number: LC1 19-1 NRC Administrative Topic (see Note)

Type Code*

Describe activity to be performed Conduct of Operations P, D, R 2017 NRC Perform Control Rod Position Verification and Determine Reactivity Event Severity N1-OP-42, OP-AA-300, N1-OP-5, K/A 2.1.37 (4.3)

Conduct of Operations N, R Develop and get Approval for an Operator Aid OP-AA-115-101, KA 2.1.15 (2.7)

Equipment Control D, R Explain RPS Manual Scram Circuit Using Prints C-19859-C, K/A 2.2.41 (3.5)

Radiation Control D, R Determine Radiological and Heat Stress Requirements Related to Operator Work in High Radiation Areas -

Steam Leak in ECIV Room RP-AA-10/11/12/403/460, SA-AA-111, K/A 2.3.7 (3.5)

Emergency Procedures/Plan 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 2020 Examination Level: SRO Operating Test Number: LC1 19-1 NRC Administrative Topic (see Note)

Type Code*

Describe activity to be performed Conduct of Operations P, D, R 2017 NRC Perform Control Rod Position Verification and Determine Reactivity Event Severity and Notification Requirements OP-AA-300, N1-OP-5, N1-OP-42, K/A 2.1.37 (4.6)

Conduct of Operations D, S Determine Reportability Requirements for Loss of Offsite Power with EDG Failure LS-AA-1400, NUREG 1022, K/A 2.1.18 (3.8)

Equipment Control N, R Review and Approval of Completed Surveillance Test, N1-ST-Q6A, Containment Spray System Loop 111 Quarterly Operability Test N1-ST-Q6A, KA 2.2.12 (4.1)

Radiation Control D, R Determine Radiological and Heat Stress Requirements Related to Operator Work in High Radiation Areas -

Steam Leak in ECIV Room RP-AA-10/11/12/403/460, SA-AA-111, K/A 2.3.7 (3.6)

Emergency Procedures/Plan M, S Emergency Event Classification & Notification EP-AA-1013, K/A 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 ( 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 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Nine Mile Point Unit 1 Date of Examination: December 2020 Exam Level: RO/SRO-I/SRO-U Operating Test No.: LC1 19-1 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. Shift Reactor Building Operating Exhaust and Supply Fans K/A 288000 A4.01 (3.1/2.9), N1-OP-10 D, S, A 9
b. Shift Feedwater Pressure and Level Channels K/A 259002 A4.06 (3.1/3.2), N1-OP-16 N, S 2
c. Respond to a Recirculation Pump Seal Failure K/A 202001 A2.10 (3.5/3.9) N1-SOP-1.2 M, S, A 1
d. Place Containment Spray in Torus Cooling K/A 219000 A4.02 (3.7/3.5) N1-EOP-1 D, S, A 5
e. Alternate RPV Blowdown Through the Reactor Head Vent Valves K/A 239001 A2.09 (3.4/3.7), N1-EOP-1, N1-EOP-8 M, S, A 3
f. Place 11 Shutdown Cooling Loop in Service K/A 205000 A4.01 (3.7/3.7), N1-OP-4 D, S, A, L 4
g. Channel 11 Non-Coincident Scram Test (RO Only)

K/A 215004 A4.05 (3.1/3.2), N1-ST-R4, N1-OP-5 P, D, S, L (2017 NRC) 7

h. EDG 103 Control Room Start Following Station Blackout K/A 295003 AA1.02 (4.2/4.3), N1-OP-45 P, D, S, EN (2017 NRC) 6 In-Plant Systems* (3 for RO); (3 for SRO-I) ; (3 or 2 for SRO-U)
i. Respond to CLC Makeup Tank Level Alarm K/A 295018 AA2.04 (2.9/2.9), N1-ARP-H1 D, A, R 8
j. Initiate Emergency Condenser Locally K/A 207000 A2.08 (3.8/3.8), N1-OP-13 D, L, E, R 4
k. Lineup to Flood the Reactor Vessel Using the Diesel Fire Pump K/A 295031 EA1.08 (3.8/3.9)

M, E, R 2

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

  • Type Codes Criteria for RO / SRO-I / SRO-U (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 4-6 / 4-6 / 2-3 9 / 8 / 4 1 / 1 / 1 1 / 1 / 1 (control room system) 1 / 1 / 1 2 / 2 / 1 3 / 3 / 2 (randomly selected) 1 / 1 / 1 Pairings:

a then b f then g

Appendix D Scenario Outline Form ES-D-1 Facility: Nine Mile Point Unit 1 Scenario No.: NRC-1 Op-Test No.: LC1 19-1 NRC Examiners: ____________________________ Operators:

Initial Conditions: The plant is operating at approximately 50% power. Reactor Building Exhaust Fan 12 is out of service for maintenance. IRM 11 is bypassed due to spiking.

Turnover: Start Circulating Water pump 11. Raise Reactor power with Recirculation flow.

Event No.

Malf.

No.

Event Type*

Event Description 1

N/A N -

BOP, SRO Start Circulating Water Pump 11 N1-OP-19 2

N/A R -

ATC, SRO Raise Reactor Power with Recirculation Flow N1-OP-1, N1-OP-43B 3

HV01A HV02 C-BOP TS-SRO Reactor Building Exhaust Fan 11 trips. Requires RBEVS initiation.

L1-3-4, L1-1-5, N1-EOP-5, N1-OP-10, Technical Specifications 4

RR65E RR09E I -All TS-SRO RR pump 15 Blind Controller failure and delayed pump trip.

Discharge valve fails to shut.

N1-SOP-1.3, Tech Spec 3.1.7 5

RX01 C - All Fuel Failure N1-SOP-25.2, N1-SOP-1.1, N1-SOP-1 6

ED27 FW03A C - BOP Powerboard 12 Fails to Auto Transfer and Feedwater Pump 11 Trips N1-SOP-30.2, N1-SOP-1 7

MS01 M - All Main Steam Line Break in Turbine Building N1-EOP-2, N1-EOP-6 8

MS13A MS13C C - All Two MSIVs Fail to Close N1-EOP-6, N1-EOP-8 9

Overrides C - ATC Turbine Building Ventilation Exhaust Fans Trip N1-EOP-6, N1-EOP-8 (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

Facility: Nine Mile Point Unit 1 Scenario No.: NRC-1 Op-Test No.: LC1 19-1 NRC

1. Malfunctions after EOP entry (1-2)

Event 6, 7, 8, 9 4

2. Abnormal events (2-4)

Events 3, 4, 5, 6 4

3. Major transients (1-2)

Event 7 1

4. EOPs entered/requiring substantive actions (1-2)

N1-EOP-2, N1-EOP-6 2

5. Entry into contingency EOP with substantive actions (at least 1 per scenario set)

N1-EOP-8 1

6. Preidentified Critical tasks (at least 2) 2 CRITICAL TASK DESCRIPTIONS:

CRITICAL TASK JUSTIFICATION:

CT-1.0: Given fuel failure causing elevated Main Steam Line radiation levels, scram the Reactor within 15 minutes of exceeding 3.75 times normal full power background, in accordance with N1-SOP-25.2.

High Main Steam Line radiation levels indicate fuel failure and release of fission products to the Reactor coolant. A Reactor scram is required by N1-SOP-25.2 and reduces the rate of energy production and thus the heat input, radioactivity release, and flow down the Main Steam Lines. Scramming the Reactor also allows further mitigating actions, such as Reactor isolation and depressurization.

CT-2.0: Given an un-isolable primary system discharging outside of primary and secondary containments, commence N1-EOP-8, RPV Blowdown, before off-site release rate exceeds the General Emergency level, in accordance with N1-EOP-6.

An un-isolable primary system discharging outside of Primary and Secondary Containments resulting in off-site release rates approaching the General Emergency limit indicates a significant problem posing a direct and immediate threat to the health and safety of the public. 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. This will lower the release of radioactivity to the environment and lower the dose received by the public.

SCENARIO

SUMMARY

The scenario begins at approximately 50% power. IRM 11 is bypassed due to spiking and Reactor Building Exhaust Fan 12 is out of service for maintenance. Circulating Water pump 11 is out of service following maintenance. The crew will start Circulating Water pump 11, then raise Reactor power with recirculation flow.

After the crew has raised reactor power, #11 RB exhaust fan will trip. The crew will diagnose the fan trip and a positive RB pressure. With #12 RB exhaust fan OOS, the crew will start the Reactor Building Emergency Ventilation System (RBEVS) to restore a negative RB pressure. One RBEVS train will trip.

Entry into N1-EOP-5, Secondary Containment Control is required. SRO determines TS 3.4.4 must be entered for the inoperable RBEVS system.

Next, RRP 15 flow rises due to a blind controller failure. The crew will take the M/A station to manual, and the rise will stop. RRMG 15 will develop a high slot temperature, requiring the crew to remove it from service.

Next, fuel failure will occur due to the previous transients. The crew will respond per N1-SOP-25.2, Fuel Failure or High Activity in Rx Coolant or Off-Gas. This includes performing an emergency power reduction per N1-SOP-1.1, and eventually scramming the Reactor per N1-SOP-1 (Critical Task). When the Generator trips after the scram, Powerboard 12 will fail to transfer to reserve power. The crew will execute N1-SOP-30.2, Loss of Powerboard 12, to re-energize the powerboard. Feedwater Pump 11 will trip shortly after the reactor scram.

Following the scram, a Main Steam line break will occur. The MSIVs will fail to close both automatically and manually, leading to an un-isolable leak into the Turbine Building. The running Turbine Building ventilation exhaust fan will trip. The crew will start the standby Turbine Building ventilation exhaust fan, however it will trip after a short time delay. This will allow an un-monitored, ground level release from the Turbine Building. The crew will enter N1-EOP-6, Radioactivity Release Control. Field reports will indicate off-site release rates approaching the General Emergency level. The crew will perform an RPV Blowdown per N1-EOP-8 (Critical Task).

Appendix D Scenario Outline Form ES-D-1 Facility: Nine Mile Point Unit 1 Scenario No.: NRC-2 Op-Test No.: LC1 19-1 NRC Examiners: ____________________________ Operators:

Initial Conditions: The plant is operating at approximately 85% power. Reactor Building Exhaust Fan 12 is out of service for maintenance. IRM 11 is bypassed due to spiking.

Turnover: Shutdown Condensate Pump 11 for maintenance due to a motor oil leak and place in Pull-To-Lock. Then Perform a Rod Sequence Exchange.

Event No.

Malf.

No.

Event Type*

Event Description 1

N/A N -BOP TS-SRO Condensate Pump 11 shutdown N1-OP-15A, Technical Specifications 2

N/A R -ATC, SRO Rod Sequence Exchange N1-OP-5, ReMA 3

RD04 C-ATC Stuck control rod (2017 NRC Scenario 1)

N1-OP-5 4

EC03B C-BOP, TS-SRO Emergency Condenser 12 Inadvertent Initiation ARP K1-1-5, N1-OP-13, Technical Specifications 5

RD34 IA01 C -All Instrument air leak, Reactor scram required N1-SOP-20.1, N1-SOP-1 6

RD33 M -All ATWS N1-EOP-2, N1-EOP-3 7

Overrides C -All Feedwater Isolation Valves 11 and 12 fail to isolate N1-EOP-3 (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

Facility: Nine Mile Point Unit 1 Scenario No.: NRC-2 Op-Test No.: LC1 19-1 NRC

1. Malfunctions after EOP entry (1-2)

Event 6, 7 2

2. Abnormal events (2-4)

Events 3, 4, 5 3

3. Major transients (1-2)

Event 6 1

4. EOPs entered/requiring substantive actions (1-2)

N1-EOP-2 1

5. Entry into contingency EOP with substantive actions (at least 1 per scenario set)

N1-EOP-3 1

6. Preidentified Critical tasks (at least 2) 3 CRITICAL TASK DESCRIPTIONS:

CRITICAL TASK JUSTIFICATION:

CT-1.0: Given a failure of the reactor to scram with power above 6% and RPV water level above -41 inches, the crew will terminate and prevent all injection except boron and CRD, within 15 minutes of failure to scram indications, in accordance with N1-EOP-3.

High Reactor power after a scram represents a challenge to nuclear fuel and to plant heat sinks. In the event of a loss of the normal heat sink, this may result in adding heat to the Torus and challenging the Primary Containment.

Lowering Reactor power reduces these challenges.

CT-2.0 Given a failure of the reactor to scram with power above 6%, the crew will lower reactor power by inserting control rods or injecting boron, within 15 minutes of failure to scram indications, in accordance with N1-EOP-3.

Inserting control rods lowers Reactor power, which reduces challenges to the plant during a failure to scram.

Additionally, inserting control rods ultimately provides a long-term, stable core shutdown. Boron injection will lower power, however, alone may not provide a stable shutdown condition.

SCENARIO

SUMMARY

The scenario begins at approximately 85% power. IRM 11 is bypassed due to spiking and Reactor Building Exhaust Fan 12 is out of service for maintenance. The crew is directed to remove Condensate Pump 11 from service immediately for maintenance due to a motor oil leak. This requires entry into TS 3.1.8 for a redundant HPCI component.

After the pump has been removed from service, the crew will conduct a rod pattern exchange. During the rod pattern exchange, a control rod becomes stuck. The crew will enter N1-OP-5, Section H.13 and raise drive water pressure to move the control rod. While the control rod is stuck, entry into Tech Spec 3.1.1.a(2) is required.

Then, an inadvertent EC initiation occurs. The crew will respond to isolate the EC and the SRO will determine Tech Spec 3.1.3.b requires a 7 day LCO.

Next, an Instrument Air leak will occur in the piping to the CRD system. The crew will insert a manual Reactor scram as CRD air pressure lowers below 60 psig (Critical Task).

When the scram occurs the control rods will not fully insert. The crew must terminate and prevent injection (Critical Task). When the operator attempts to close Feedwater Isolation Valves 11 and 12, the valves will fail to isolate Feedwater flow. The crew must diagnose the failure and place the Feedwater pumps in Pull-To-Lock to terminate feeding the RPV. The crew will lower Reactor power by inserting control rods per EOP-3.1 and/or using Liquid Poison (Critical Task).

Appendix D Scenario Outline Form ES-D-1 Facility: Nine Mile Point Unit 1 Scenario No.: NRC-3 Op-Test No.: LC1 19-1 NRC Examiners: ____________________________ Operators:

Initial Conditions: The plant is operating at approximately 87% power. Reactor Building Exhaust Fan 12 is out of service for maintenance. IRM 11 is bypassed due to spiking.

Turnover: Recirc Pump 11 MG set has been repaired and is ready to be returned to service. Restore 11 recirc pump to service.

After starting Recirc Pump 11 MG set and placing in service, operate it for one hour while maintenance takes readings before returning to 100% power.

Event No.

Malf.

No.

Event Type*

Event Description 1

N/A N -BOP, SRO Restore Recirc Pump 15 to service 2

NM19C I-ATC, SRO APRM 13 fails upscale ARP (2017 NRC Scenario 4) 3 AD05A C -BOP R-ATC TS-SRO ERV Inadvertently opens (2017 NRC Scenario 3)

N1-SOP-1.4, N1-SOP-1.1, Technical Specifications 4

ED05 C-All TS-SRO Powerboard 12 Electrical Fault N1-SOP-30.2, N1-SOP-1.3, N1-SOP-1.1, Technical Specifications 5

EC01 M -All Steam leak inside Drywell N1-EOP-2, N1-EOP-4 6

PC10A PC10C C-All Failed open Torus to Drywell vacuum breaker N1-EOP-4 (2017 NRC Scenario 3) 7 FW28A FW28B CS07 C -BOP, SRO HPCI fails to initiate, Core Spray fails to auto-inject N1-EOP-2 (2017 NRC Scenario 3) 8 CT01A CT01B C-ATC, SRO Containment Spray pumps 111 and 112 trip N1-EOP-8, N1-EOP-4 (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

Facility: Nine Mile Point Unit 1 Scenario No.: NRC-3 Op-Test No.: LC1 19-1 NRC

1. Malfunctions after EOP entry (1-2)

Event 6, 7, 8 3

2. Abnormal events (2-4)

Events 2, 3, 4 3

3. Major transients (1-2)

Event 5 1

4. EOPs entered/requiring substantive actions (1-2)

N1-EOP-2, N1-EOP-4 2

5. Entry into contingency EOP with substantive actions (at least 1 per scenario set)

N1-EOP-8 1

6. Preidentified Critical tasks (at least 2) 3 CRITICAL TASK DESCRIPTIONS:

CRITICAL TASK JUSTIFICATION:

CT-1.0: Given an inadvertently open ERV at power, close the ERV or insert a manual scram prior to Torus temperature exceeding 110oF, in accordance with N1-SOP-1.4 A manual Reactor scram is required before Torus temperature exceeds 110oF. This reduces the rate of energy production and thus heat input to the Torus. Additionally, this allows evaluating the success of the Reactor scram before boron injection would be required due to Torus temperature in the event of a failure to scram. Closing the ERV prior to the need for the scram avoids the need for these more substantial actions, prevents challenging the plant with a scram, and stops heat input to the Torus.

CT-2.0: Given a LOCA in the Drywell and a failure of HPCI to initiate, the crew will inject with preferred and alternate injection systems to restore and maintain RPV water level above -84 inches, in accordance with N1-EOP-2.

Injection with preferred and alternate injection systems may be initiated before RPV level lowers to -84 inches, but must be initiated within 15 minutes of RPV water level lowering below -84 inches.

Maintaining Reactor water level above -

84 inches ensures adequate core cooling through the preferred method of core submergence. This protects the integrity of the fuel cladding.

CT-3.0: Given a LOCA in the Drywell and degraded Containment Spray capability, the crew will execute N1-EOP-8, RPV Blowdown, when it is determined Torus pressure cannot be maintained inside the Pressure Suppression Pressure (PSP) limit, in accordance with N1-EOP-4. N1-EOP-8 may be entered prior to exceeding PSP, but must be executed within 15 minutes of exceeding PSP.

A Blowdown is required to limit further release of energy into the Primary Containment and to ensure that the RPV is depressurized while pressure suppression capability is still available.

This protects the integrity of the Primary Containment.

SCENARIO

SUMMARY

The scenario begins at approximately 87% power. IRM 11 is bypassed due to spiking and Reactor Building Exhaust Fan 12 is out of service for maintenance.

Immediately after assuming the shift the crew will be directed to restore Recirculation Pump 15 to service and return to full power. The crew will assess plant conditions and verify Recirculation Flow is less than 50 Mlbm/hr. They will then return Recirculation Pump 15 to service.

After the crew has placed the recirc pump in service, APRM 13 will fail upscale causing a half scram. The crew will bypass the APRM and reset the half scram.

When the half scram is reset, ERV 111 will inadvertently open. The crew will enter N1-SOP-1.4, Stuck Open ERV. The crew will perform an emergency power reduction to approximately 85% power, then take actions to close ERV 111 (Critical Task). These actions will close the ERV, but leave it inoperable. The SRO will determine the Tech Spec impact.

Next, Powerboard 12 will de-energize due to an electrical fault. This will cause loss of multiple major loads, including a second Recirculation pump, a Service Water pump, and a Circulating Water pump.

The crew will respond per N1-SOP-30.2. This will include lowering Reactor power to restore the plant within single Circulating Water pump operating limitations. The SRO will determine the Tech Spec impact of this power loss.

A steam leak will then develop in the Primary Containment. The crew will insert a scram. Following the scram, HPCI will fail to initiate, requiring manual action to establish injection with preferred and/or alternate injection systems to maintain RPV water level (Critical Task).

When the crew attempts to spray the Containment, Containment Spray pumps 111 and 112 will trip. The two remaining Containment Spray pumps will be insufficient to avoid violating PSP, and the crew will perform an RPV Blowdown (Critical Task).

Appendix D Scenario Outline Form ES-D-1 Facility: Nine Mile Point Unit 1 Scenario No.: NRC-4 Op-Test No.: LC1 19-1 NRC Examiners: ____________________________ Operators:

Initial Conditions: The plant is operating at approximately 100% power. Reactor Building Exhaust Fan 12 is out of service for maintenance. IRM 11 is bypassed due to spiking.

Turnover: Lower Torus water level to 10.8 feet per N1-OP-14 using Containment Spray 111.

Event No.

Malf.

No.

Event Type*

Event Description 1

N/A N - BOP Transfer Torus Water to the Waste Collection Tank N1-OP-14 2

CT01A C-BOP TS-SRO Containment Spray Pump 111 Trip Technical Specifications 3

TC03A C-ATC, SRO EPR Fails High N1-SOP-31.2 4

RP20B TS-SRO Drywell Pressure transmitter Failed Low Technical Specifications 5

PC05 CW06B C -

BOP, SRO R-ATC Seismic Event with Circulating Water Pump trip.

N1-SOP-28, N1-SOP-1.1 6

EG11 C -All Degraded 345KV Grid conditions N1-SOP-33B.1, N1-SOP-1 7

RR29 M -All Coolant leak in Drywell N1-EOP-2, N1-EOP-4 8

FW03A FW03B C -All Trip of Feedwater Pumps N1-EOP-2, N1-EOP-8 (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

Facility: Nine Mile Point Unit 1 Scenario No.: NRC-4 Op-Test No.: LC1 19-1 NRC

1. Malfunctions after EOP entry (1-2)

Events 7, 8 2

2. Abnormal events (2-4)

Events 2, 3, 4, 5, 6 5

3. Major transients (1-2)

Event 7 1

4. EOPs entered/requiring substantive actions (1-2)

N1-EOP-2, N1-EOP-4 2

5. Entry into contingency EOP with substantive actions (at least 1 per scenario set)

N1-EOP-8 1

6. Preidentified Critical tasks (at least 2) 2 CRITICAL TASK DESCRIPTIONS:

CRITICAL TASK JUSTIFICATION:

CT-1.0: Given a LOCA in the Drywell with Drywell temperature approaching 300F or Torus pressure exceeding 13 psig, initiate Containment Sprays prior to exceeding the Pressure Suppression Pressure limit, in accordance with N1-EOP-4.

Initiating Containment Sprays reduces Primary Containment pressure. This 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 a LOCA with degraded high pressure injection capability, the crew will depressurize the RPV and inject with Preferred and Alternate Injection Systems to restore and maintain RPV water level above -84 inches, in accordance with N1-EOP-2. Injection with preferred and/or alternate injection systems must be performed within 15 minutes of performing an RPV blowdown.

Maintaining Reactor water level above -

84 inches ensures adequate core cooling through the preferred method of core submergence. This protects the integrity of the fuel cladding.

SCENARIO

SUMMARY

The scenario begins at approximately 100% power. IRM 11 is bypassed due to spiking and Reactor Building Exhaust Fan 12 is out of service for maintenance. Torus water level is at the high end of the normal band. The crew will lower Torus Water Level in accordance with N1-OP-14 using 111 Containment Spray System. Containment Spray pump 111 will trip requiring the SRO to make a Tech Spec call.

Then, the EPR fails high. The MPR automatically takes control of pressure at a value about 5 psig above the initial pressure. The crew will enter N1-SOP-31.2, remove the EPR from service and return reactor pressure to the initial value.

Next, one of the four drywell pressure transmitters fails downscale, preventing that channel from actuating protective functions. The transmitter inputs to RPS, Core Spray, Containment Spray and Automatic Depressurization Systems (ADS). Tech Spec 3.6.2 entry is required.

Next, a seismic event occurs causing one of the circulating water pumps to trip. The crew will respond by lowering power per N1-SOP-1.1 in order to maintain condenser vacuum. Then, a grid disturbance develops, resulting in lowering frequency and voltage on the 345KV power lines. The crew will enter N1-SOP-33B.1 and monitor grid frequency to determine action times for tripping the turbine. As the grid continues to degrade, the crew will scram the Reactor.

A coolant leak in the Drywell will develop following the scram. The crew will enter N1-EOP-4 and re-enter N1-EOP-2. The crew will initiate Containment Sprays to prevent exceeding Pressure Suppression Pressure, in accordance with N1-EOP-4 (Critical Task). The remaining high pressure Feedwater pump will trip, causing RPV water level to lower to the top of active fuel (TAF). With the degraded high pressure injection capability, the crew will enter RPV Blowdown before RPV water level drops below -109 inches, in accordance with N1-EOP-2 (Critical Task).

ES-401 Page 1 of 16 Form ES-401-1 Facility:

Nine Mile Point Unit 1 (Rev. 1)

Date of Exam: December 2020 Tier Group RO K/A Category Points SRO-Only Points K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*

Total A2 G*

Total

1.

Emergency and Abnormal Plant Evolutions 1

3 3

3 N/A 3

4 N/A 4

20 4

3 7

2 1

1 2

1 1

1 7

2 1

3 Tier Totals 4

4 5

4 5

5 27 6

4 10

2.

Plant Systems 1

3 1

2 2

2 2

2 3

3 3

3 26 3

2 5

2 1

1 1

1 1

1 1

1 2

1 1

12 0

2 1

3 Tier Totals 4

2 3

3 3

3 3

4 5

4 4

38 5

3 8

3. Generic Knowledge and Abilities Categories 1

2 3

4 10 1

2 3

4 7

3 2

2 3

2 2

1 2

Note: 1. Ensure that at least two topics from every applicable K/A 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 K/A category shall not be less than two). (One Tier 3 radiation control K/A is allowed if it is replaced by a K/A 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 NRC revisions. The final RO exam must total 75 points, and the SRO-only exam must total 25 points.
3. Systems/evolutions within each group are identified on the 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 K/A 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 K/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 K/A categories.
7. The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As.
8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics IRs for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above. If fuel-handling equipment is sampled in 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 K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.

G* Generic K/As These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan. They are not required to be included when using earlier revisions of the K/A catalog.

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

ES-401 Page 2 of 16 Form ES-401-1 Nine Mile Point Unit 1 (Rev. 1)

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

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

K/A Topic(s)

IR Q#

21 295001 (APE 1) Partial or Complete Loss of Forced Core Flow Circulation / 1 & 4 X

(295001G2.1.19) PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION / 1 & 4: Ability to use plant computers to evaluate system or component status.

3.8 76 22 295003 (APE 3) Partial or Complete Loss of AC Power / 6 X

(295003AA2.01) Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER: Cause of partial or complete loss of A.C. power 3.7 77 23 295005 (APE 5) Main Turbine Generator Trip /

3 X

(295005AA2.08) Ability to determine and/or interpret the following as they apply to MAIN TURBINE GENERATOR TRIP: Electrical distribution status 3.3 78 24 295016 (APE 16)

Control Room Abandonment / 7 X

(295016AA2.03) Ability to determine and/or interpret the following as they apply to CONTROL ROOM ABANDONMENT: Reactor pressure 4.4 79 25 295019 (APE 19) Partial or Complete Loss of Instrument Air / 8 X

(295019G2.4.9) PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR / 8: 2.4.9 Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.

4.2 80 26 295037 (EPE 14) Scram Condition Present and Reactor Power Above APRM Downscale or Unknown / 1 X

(295037G2.4.41) SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN / 1: Knowledge of the emergency action level thresholds and classifications.

4.6 81 27 700000 (APE 25)

Generator Voltage and Electric Grid Disturbances / 6 X

(700000AA2.01) Ability to determine and/or interpret the following as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES: Operating point on the generator capability curve 3.6 82 1

295001 (APE 1) Partial or Complete Loss of Forced Core Flow Circulation / 1 & 4 X

(295001AK1.04) Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: Limiting cycle oscillation: Plant-Specific 2.5/3.3 39 2

295003 (APE 3) Partial or Complete Loss of AC Power / 6 X

(295003AK2.04) Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF A.C. POWER and the following: A.C. electrical loads 3.4/3.5 40 3

295004 (APE 4) Partial or Total Loss of DC Power / 6 X

(295004AK3.01) Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER: Load shedding:

Plant-Specific 2.6/3.1 41 4

295005 (APE 5) Main Turbine Generator Trip /

3 X

(295005AA1.04) Ability to operate and/or monitor the following as they apply to MAIN TURBINE GENERATOR TRIP: Main generator controls 2.7/2.8 42 5

295006 (APE 6) Scram /

1 X

(295006AA2.06) Ability to determine and/or interpret the following as they apply to SCRAM: Cause of reactor SCRAM 3.5/3.8 43 6

295016 (APE 16)

Control Room Abandonment / 7 X

(295016G2.1.32) CONTROL ROOM ABANDONMENT /

7: Ability to explain and apply system limits and precautions.

3.8/4.0 44 7

295018 (APE 18) Partial or Complete Loss of CCW / 8 X

(295018AA2.05) Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER: System pressure 2.9/2.9 45 8

295019 (APE 19) Partial or Complete Loss of Instrument Air / 8 X

(295019G2.4.47) PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR / 8: Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.

4.2/4.2 46 9

295021 (APE 21) Loss of Shutdown Cooling / 4 X

(295021AK1.04) Knowledge of the operational implications of the following concepts as they apply to LOSS OF SHUTDOWN COOLING: Natural circulation 3.6/3.7 47 10 295023 (APE 23)

Refueling Accidents / 8 X

(295023AK2.01) Knowledge of the interrelations between REFUELING ACCIDENTS and the following: Fuel handling equipment 3.3/3.7 48 11 295024 (EPE 1) High Drywell Pressure / 5 X

(295024EK3.06) Knowledge of the reasons for the following responses as they apply to HIGH DRYWELL PRESSURE: Reactor SCRAM 4.0/4.1 49 12 295025 (EPE 2) High Reactor Pressure / 3 X

(295025EA1.06) Ability to operate and/or monitor the following as they apply to HIGH REACTOR PRESSURE:

Isolation condenser: Plant-Specific 4.5/4.5 50

ES-401 Page 3 of 16 Form ES-401-1 Nine Mile Point Unit 1 (Rev. 1)

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

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

K/A Topic(s)

IR Q#

13 295026 (EPE 3)

Suppression Pool High Water Temperature / 5 X

(295026EA2.03) Ability to determine and/or interpret the following as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Reactor pressure 3.9/4.0 51 14 295028 (EPE 5) High Drywell Temperature (Mark I and Mark II only)

/ 5 X

(295028G2.4.3) HIGH DRYWELL TEMPERATURE (MARK I AND MARK II ONLY) / 5: Ability to identify post-accident instrumentation.

3.7/3.9 52 15 295030 (EPE 7) Low Suppression Pool Water Level / 5 X

(295030EK1.02) Knowledge of the operational implications of the following concepts as they apply to LOW SUPPRESSION POOL WATER LEVEL: Pump NPSH 3.5/3.8 53 16 295031 (EPE 8) Reactor Low Water Level / 2 X

(295031EK2.02) Knowledge of the interrelations between REACTOR LOW WATER LEVEL and the following:

Reactor pressure 3.8/3.9 54 17 295037 (EPE 14) Scram Condition Present and Reactor Power Above APRM Downscale or Unknown / 1 X

(295037EK3.07) Knowledge of the reasons for the following responses as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: Various alternate methods of control rod insertion: Plant-Specific 4.2/4.3 55 18 295038 (EPE 15) High Offsite Radioactivity Release Rate / 9 X

(295038EA1.06) Ability to operate and/or monitor the following as they apply to HIGH OFF-SITE RELEASE RATE: Plant ventilation 3.5/3.6 56 19 600000 (APE 24) Plant Fire On Site / 8 X

(600000AA2.03) Ability to determine and interpret the following as they apply to PLANT FIRE ON SITE: Fire alarm 2.8/3.2 57 20 700000 (APE 25)

Generator Voltage and Electric Grid Disturbances / 6 X

(700000G2.2.12) GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES / 6: Knowledge of surveillance procedures.

3.7/4.1 58 K/A Category Totals:

3 3

3 3

8 7

Group Point Total:

20/7

ES-401 Page 4 of 16 Form ES-401-1 Nine Mile Point Unit 1 (Rev. 1)

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

Item E/APE # / Name /

Safety Function K1 K2 K3 A1 A2 G*

K/A Topic(s)

IR Q#

35 295013 (APE 13) High Suppression Pool Temperature. / 5 X

(295013AA2.01) Ability to determine and/or interpret the following as they apply to HIGH SUPPRESSION POOL TEMPERATURE: Suppression pool temperature 4.0 83 36 295014 (APE 14)

Inadvertent Reactivity Addition / 1 X

(295014G2.1.23) INADVERTENT REACTIVITY ADDITION / 1: Ability to perform specific system and integrated plant procedures during all modes of plant operation.

4.4 84 37 295029 (EPE 6) High Suppression Pool Water Level / 5 X

(295029EA2.03) Ability to determine and/or interpret the following as they apply to HIGH SUPPRESSION POOL WATER LEVEL: Drywell/containment water level 3.5 85 28 295012 (APE 12) High Drywell Temperature / 5 X

(295012AK3.01) Knowledge of the reasons for the following responses as they apply to HIGH DRYWELL TEMPERATURE: Increased drywell cooling 3.5/3.6 59 29 295015 (APE 15)

Incomplete Scram / 1 X

(295015AA1.02) Ability to operate and/or monitor the following as they apply to INCOMPLETE SCRAM: RPS 4.0/4.2 60 30 295022 Loss of Control Rod Drive Pumps X

(295022AA2.01) Ability to determine and/or interpret the following as they apply to LOSS OF CRD PUMPS:

Accumulator pressure 3.5/3.6 61 31 295032 (EPE 9) High Secondary Containment Area Temperature / 5 X

(295032G2.1.20) HIGH SECONDARY CONTAINMENT AREA TEMPERATURE / 5: Ability to interpret and execute procedure steps.

4.6/4.6 62 32 295035 (EPE 12)

Secondary Containment High Differential Pressure / 5 X

(295035EK1.01) Knowledge of the operational implications of the following concepts as they apply to SECONDARY CONTAINMENT HIGH DIFFERENTIAL PRESSURE: Secondary containment integrity 3.9/4.2 63 33 295036 (EPE 13)

Secondary Containment High Sump/Area Water Level / 5 X

(295036EK2.01) Knowledge of the interrelations between SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL and the following: Secondary containment equipment and floor drain system 3.1/3.2 64 34 500000 (EPE 16) High Containment Hydrogen Concentration / 5 X

(500000EK3.06) Knowledge of the reasons for the following responses as they apply to HIGH PRIMARY CONTAINMENT HYDROGEN CONCENTRATIONS:

Operation of wet well vent 3.1/3.7 65 K/A Category Totals:

1 1

2 1

3 2

Group Point Total:

7/3

ES-401 Page 5 of 16 Form ES-401-1 Nine Mile Point Unit 1 (Rev. 1)

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

Item E/APE # / Name / Safety Function K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*

K/A Topic(s)

IR Q#

64 206000 (SF2, SF4 HPCIS) High-Pressure Coolant Injection X

(206000A2.17) Ability to (a) predict the impacts of the following on the HIGH PRESSURE COOLANT INJECTION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

HPCI inadvertent initiation:BWR-2,3,4 4.3 86 65 209001 (SF2, SF4 LPCS) Low-Pressure Core Spray X

(209001G2.1.30)

LOW-PRESSURE CORE SPRAY: Ability to locate and operate components, including local controls.

4.0 87 66 211000 (SF1 SLCS)

Standby Liquid Control X

(211000A2.04) Ability to (a) predict the impacts of the 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: Inadequate system flow 3.4 88 67 261000 (SF9 SGTS)

Standby Gas Treatment X

(261000G2.2.25) STANDBY GAS TREATMENT:

Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

4.2 89 68 218000 Automatic Depressurization System X

A2.06 - Ability to (a) predict the impacts of the 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: ADS initiation signals present.

4.3 90 38 205000 (SF4 SCS)

Shutdown Cooling X

(205000A2.08) Ability to (a) predict the impacts of 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: Loss of heat exchanger cooling 3.3/3.5 1

39 206000 (SF2, SF4 HPCIS) High-Pressure Coolant Injection X

(206000A3.05) Ability to monitor automatic operations of the HIGH PRESSURE COOLANT INJECTION SYSTEM including: Reactor water level:BWR-2,3,4 4.3/4.3 2

ES-401 Page 6 of 16 Form ES-401-1 Nine Mile Point Unit 1 (Rev. 1)

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

Item E/APE # / Name / Safety Function K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*

K/A Topic(s)

IR Q#

40 207000 (SF4 IC)

Isolation (Emergency)

Condenser X

(207000A4.04) ISOLATION (EMERGENCY)

CONDENSER: Ability to manually operate and/or monitor in the control room:

Vent line radiation levels:BWR-2,3 3.8/4.0 3

41 209001 (SF2, SF4 LPCS) Low-Pressure Core Spray X

(209001G2.4.50)

LOW-PRESSURE CORE SPRAY: Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.

4.2/4.0 4

42 211000 (SF1 SLCS)

Standby Liquid Control X

(211000K1.06) Knowledge of the physical connections and/or cause-effect relationships between STANDBY LIQUID CONTROL SYSTEM and the following: Reactor vessel 3.7/3.7 5

43 212000 (SF7 RPS)

Reactor Protection X

(212000K2.01) REACTOR PROTECTION SYSTEM:

Knowledge of electrical power supplies to the following: RPS motor-generator sets 3.2/3.3 6

44 215003 (SF7 IRM)

Intermediate-Range Monitor X

(215003K3.02) Knowledge of the effect that a loss or malfunction of the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM will have on following:

Reactor manual control 3.6/3.6 7

45 215004 (SF7 SRMS)

Source-Range Monitor X

(215004K4.01) Knowledge of SOURCE RANGE MONITOR (SRM) SYSTEM design feature(s) and/or interlocks which provide for the following: Rod withdrawal blocks 3.7/3.7 8

46 215005 (SF7 PRMS)

Average Power Range Monitor/Local Power Range Monitor X

(215005K5.01) Knowledge of the operational implications of the following concepts as they apply to AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM: LPRM detector operation 2.8/2.9 9

47 218000 (SF3 ADS)

Automatic Depressurization X

(218000K6.06) Knowledge of the effect that a loss or malfunction of the following will have on the AUTOMATIC DEPRESSURIZATION SYSTEM: D.C. power:

Plant-Specific 3.4/3.6 10 48 223002 (SF5 PCIS)

Primary Containment Isolation/Nuclear Steam Supply Shutoff X

(223002A1.04) Ability to predict and/or monitor changes in parameters associated with operating the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF controls including:

Individual system relay status 2.6/2.8 11

ES-401 Page 7 of 16 Form ES-401-1 Nine Mile Point Unit 1 (Rev. 1)

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

Item E/APE # / Name / Safety Function K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*

K/A Topic(s)

IR Q#

49 223002 (SF5 PCIS)

Primary Containment Isolation/Nuclear Steam Supply Shutoff X

(223002A2.10) Ability to (a) predict the impacts of 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: Loss of coolant accidents 3.9/4.2 12 50 239002 (SF3 SRV)

Safety Relief Valves X

(239002A2.05) Ability to (a) predict the impacts of the 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:

Low reactor pressure 3.2/3.4 13 51 239002 (SF3 SRV)

Safety Relief Valves X

(239002A3.01) Ability to monitor automatic operations of the RELIEF/SAFETY VALVES including: SRV operation after ADS actuation 3.8/3.9 14 52 259002 (SF2 RWLCS)

Reactor Water Level Control X

(259002A3.07) Ability to monitor automatic operations of the REACTOR WATER LEVEL CONTROL SYSTEM including: FWRV lockup 3.5/3.6 15 53 259002 (SF2 RWLCS)

Reactor Water Level Control X

(259002A4.06) REACTOR WATER LEVEL CONTROL SYSTEM: Ability to manually operate and/or monitor in the control room:

DP/Single/three element control selector switch:

Plant-Specific SYSTEM:259002 Reactor Water Level Control System Plant-Specific 3.1/3.2 16 54 261000 (SF9 SGTS)

Standby Gas Treatment X

(261000A4.09) STANDBY GAS TREATMENT SYSTEM: Ability to manually operate and/or monitor in the control room:

Ventilation valves/dampers 2.7/2.7 17 55 212000 Reactor Protection System X

Knowledge of EOP mitigation strategies.

3.7/4.7 18 56 262001 (SF6 AC) AC Electrical Distribution X

(262001G2.1.7) AC ELECTRICAL DISTRIBUTION: 2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

4.4/4.7 19

ES-401 Page 8 of 16 Form ES-401-1 Nine Mile Point Unit 1 (Rev. 1)

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

Item E/APE # / Name / Safety Function K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*

K/A Topic(s)

IR Q#

57 262001 (SF6 AC) AC Electrical Distribution X

(262001K1.02) Knowledge of the physical connections and/or cause-effect relationships between A.C.

ELECTRICAL DISTRIBUTION and the following: D.C. electrical distribution 3.3/3.6 20 58 262002 (SF6 UPS)

Uninterruptable Power Supply (AC/DC)

X (262002K1.17) Knowledge of the physical connections and/or cause-effect relationships between UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) and the following: Scram solenoid valves: Plant-Specific 3.1/3.3 21 59 263000 (SF6 DC) DC Electrical Distribution X

(263000K3.02) Knowledge of the effect that a loss or malfunction of the D.C.

ELECTRICAL DISTRIBUTION will have on following: Components using D.C. control power (i.e. breakers) 3.5/3.8 22 60 264000 (SF6 EGE)

Emergency Generators (Diesel/Jet) EDG X

(264000K4.01) Knowledge of EMERGENCY GENERATORS (DIESEL/JET) design feature(s) and/or interlocks which provide for the following: Emergency generator trips (normal) 3.5/3.7 23 61 300000 (SF8 IA)

Instrument Air X

(300000K5.01) Knowledge of the operational implications of the following concepts as they apply to the INSTRUMENT AIR SYSTEM: Air compressors 2.5/2.5 24 62 400000 (SF8 CCS)

Component Cooling Water X

(400000K6.06) Knowledge of the effect that a loss or malfunction of the following will have on the CCWS:

Heat exchangers and condensers 2.9/2.9 25 63 207000 (SF4 IC)

Isolation (Emergency)

Condenser X

(207000A1.01) Ability to predict and/or monitor changes in parameters associated with operating the ISOLATION (EMERGENCY)

CONDENSER controls including: Isolation condenser level:BWR-2,3 3.7/3.8 26 K/A Category Totals:

3 1

2 2

2 2

2 6

3 3

5 Group Point Total:

26/5

ES-401 Page 9 of 16 Form ES-401-1 Nine Mile Point Unit 1 (Rev. 1)

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

Item E/APE # / Name /

Safety Function K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*

K/A Topic(s)

IR Q#

81 214000 (SF7 RPIS) Rod Position Information X

(214000A2.01) Ability to (a) predict the impacts of the following on the ROD POSITION INFORMATION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Failed reed switches 3.3 91 82 234000 (SF8 FH)

Fuel-Handling Equipment X

(234000G2.4.31)

FUEL-HANDLING EQUIPMENT: Knowledge of annunciator alarms, indications, or response procedures.

4.1 92 83 245000 (SF4 MTGEN)

Main Turbine Generator/Auxiliary X

(245000A2.03) Ability to (a) predict the impacts of the following on the MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of condenser vacuum 3.6 93 69 201001 (SF1 CRDH)

CRD Hydraulic X

(201001A3.08) Ability to monitor automatic operations of the CONTROL ROD DRIVE HYDRAULIC SYSTEM including: Drive water flow 3.0/2.9 27 70 201003 (SF1 CRDM)

Control Rod and Drive Mechanism X

(201003A4.02) CONTROL ROD AND DRIVE MECHANISM: Ability to manually operate and/or monitor in the control room:

CRD mechanism position:

Plant-Specific 3.5/3.5 28 71 204000 (SF2 RWCU)

Reactor Water Cleanup X

(204000G2.2.42)

REACTOR WATER CLEANUP: Ability to recognize system parameters that are entry-level conditions for Technical Specifications.

3.9/4.6 29 72 215001 (SF7 TIP)

Traversing In-Core Probe X

(215001K1.05) Knowledge of the physical connections and/or cause-effect relationships between TRAVERSING IN-CORE PROBE and the following:

Primary containment isolation system:(Not-BWR1) 3.3/3.4 30 73 216000 (SF7 NBI)

Nuclear Boiler Instrumentation X

(216000K2.01) NUCLEAR BOILER Instrumentation:

Knowledge of electrical power supplies to the following: Analog trip system: Plant-Specific 2.8/2.8 31

ES-401 Page 10 of 16 Form ES-401-1 Nine Mile Point Unit 1 (Rev. 1)

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

Item E/APE # / Name /

Safety Function K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*

K/A Topic(s)

IR Q#

74 223001 (SF5 PCS)

Primary Containment and Auxiliaries X

(223001K3.03) Knowledge of the effect that a loss or malfunction of the PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES will have on following:

Containment/drywell pressure: Plant-Specific.

3.4/3.5 32 75 233000 (SF9 FPCCU)

Fuel Pool Cooling/Cleanup X

(233000K4.06) Knowledge of FUEL POOL COOLING AND CLEAN-UP design feature(s) and/or interlocks which provide for the following: Maintenance of adequate pool level 2.9/3.2 33 76 239001 (SF3, SF4 MRSS) Main and Reheat Steam X

(239001K5.09) Knowledge of the operational implications of the following concepts as they apply to MAIN AND REHEAT STEAM SYSTEM: Decay heat removal 3.4/3.5 34 77 241000 (SF3 RTPRS)

Reactor/Turbine Pressure Regulating X

(241000K6.08) Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR/TURBINE PRESSURE REGULATING SYSTEM: Reactor power 3.6/3.7 35 78 256000 Reactor Condensate System X

A1.03 - Ability to predict and/or monitor changes in parameters associated with operating the REACTOR CONDENSATE SYSTEM controls including: System pressure 2.8/2.8 36 79 288000 (SF9 PVS)

Plant Ventilation X

(288000A2.05) Ability to (a) predict the impacts of the following on the PLANT VENTILATION SYSTEMS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Extreme outside weather conditions: Plant-Specific 2.6/2.7 37 80 290003 (SF9 CRV)

Control Room Ventilation X

(290003A3.01) Ability to monitor automatic operations of the CONTROL ROOM HVAC including:

Initiation/reconfiguration 3.3/3.5 38 K/A Category Totals:

1 1

1 1

1 1

1 3

2 1

2 Group Point Total:

12/3

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

Form ES-401-3 Nine Mile Point Unit 1 (Rev. 1)

BWR Examination Outline Form ES-401-3 Plant Systems - Tier 3 (RO/SRO)

Category K/A #

Topic RO SRO-only Item IR Q

IR Q#

1. Conduct of Operations G2.1.8 Ability to coordinate personnel activities outside the control room.

84 3.4 66 G2.1.28 Knowledge of the purpose and function of major system components and controls.

85 4.1 67 G2.1.2 Knowledge of operator responsibilities during all modes of plant operation.

86 4.1 68 G2.1.40 Knowledge of refueling administrative requirements.

87 3.9 94 G2.1.5 Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.

88 3.9 95 Subtotal 3

2

2. Equipment Control G2.2.35 Ability to determine Technical Specification Mode of Operation.

89 3.6 69 G2.2.2 Ability to manipulate the console controls as required to operate the facility between shutdown and designated power levels.

90 4.6 70 G2.2.17 Knowledge of the process for managing maintenance activities during power operations, such as risk assessments, work prioritization, and coordination with the transmission system operator.

91 3.8 96 G2.2.25 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

92 4.2 97 Subtotal 2

2

3. Radiation Control G2.3.13 Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

93 3.4 71 G2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions.

94 3.2 72 G2.3.11 Ability to control radiation releases.

95 4.3 98 Subtotal 2

1

4. Emergency Procedures/Plan G2.4.8 Knowledge of how abnormal operating procedures are used in conjunction with EOPs.

96 3.8 73 G2.4.35 Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects.

97 3.8 74 G2.4.34 Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects.

98 4.2 75 G2.4.19 Knowledge of EOP layout, symbols, and icons.

99 4.1 99 G2.4.22 Knowledge of the bases for prioritizing safety functions during abnormal/emergency operations.

100 4.4 100 Subtotal 3

2 Tier 3 Point Total 10 7

ES-401 Record of Rejected K/As Form ES-401-4 Tier / Group Randomly Selected K/A Reason for Rejection The systematic and random sampling process utilized the pre-approved Nine Mile Point Unit 1 K/A suppression list.

The following K/As were rejected following the systematic and random sampling process:

2 / 1 Question 12 223002 Primary Containment Isolation/Nuclear Steam Supply Shutoff A2.02 - Ability to (a) predict the impacts of 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: D.C. electrical distribution failures Randomly resampled the K/A to limit overlap with other questions and previous NRC exam.

Randomly reselected K/A 223002 Primary Containment Isolation/Nuclear Steam Supply Shutoff A2.10 - Ability to (a) predict the impacts of 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: Loss of coolant accidents.

2 / 1 Question 18 261000 Standby Gas Treatment 2.4.3 - Ability to identify post-accident instrumentation.

Resampled to limit overlap (SGTS sampled three times, generic K/A sampled twice).

Randomly resampled K/A 212000 Reactor Protection System 2.4.6 - Knowledge of EOP mitigation strategies.

2 / 2 Question 27 201001 CRD Hydraulic A3.06 - Ability to monitor automatic operations of the CONTROL ROD DRIVE HYDRAULIC SYSTEM including: Reactor power An acceptable question could not be developed for the K/A due to lack of automatic system response based on Reactor power.

Randomly resampled K/A 201001 CRD Hydraulic A3.08 - Ability to monitor automatic operations of the CONTROL ROD DRIVE HYDRAULIC SYSTEM including: Drive water flow.

ES-401 Record of Rejected K/As Form ES-401-4 2 / 2 Question 32 223001 Primary Containment and Auxiliaries K3.01 - Knowledge of the effect that a loss or malfunction of the PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES will have on following: Secondary containment A question with an appropriate level of difficulty could not be developed for the K/A.

Randomly resampled K/A 223001 Primary Containment and Auxiliaries K3.03 - Knowledge of the effect that a loss or malfunction of the PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES will have on following:

Containment/drywell pressure: Plant-Specific.

2 / 2 Question 36 268000 Radwaste A1.02 - Ability to predict and/or monitor changes in parameters associated with operating the RADWASTE controls including: Off-site release An operationally relevant question could not be developed for the K/A because the facility does not typically conduct off-site releases from this system.

Randomly resampled K/A 256000 Reactor Condensate System A1.03 - Ability to predict and/or monitor changes in parameters associated with operating the REACTOR CONDENSATE SYSTEM controls including:

System pressure.

1 / 1 Question 46 295019 Partial or Complete Loss of Instrument Air 2.4.2 - Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions.

An acceptable question could not be developed for the K/A due to lack of EOP entry conditions associated with the evolution.

Randomly resampled K/A 295019 Partial or Complete Loss of Instrument Air 2.4.47 - Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.

1 / 1 Question 49 295024 High Drywell Pressure EK3.05 - Knowledge of the reasons for the following responses as they apply to HIGH DRYWELL PRESSURE: RPV flooding An acceptable question could not be developed for the K/A because RPV flooding is not a response to high Drywell pressure.

Randomly resampled K/A 295024 High Drywell Pressure EK3.06 - Knowledge of the reasons for the following responses as they apply to HIGH DRYWELL PRESSURE: Reactor SCRAM.

ES-401 Record of Rejected K/As Form ES-401-4 1 / 1 Question 57 600000 Plant Fire On Site AA2.06 - Ability to determine and interpret the following as they apply to PLANT FIRE ON SITE:

Need for pressurizing control room (recirculating mode)

An acceptable question could not be developed for the K/A due to low operational importance /

minutia.

Randomly resampled K/A 600000 Plant Fire On Site AA2.03 - Ability to determine and interpret the following as they apply to PLANT FIRE ON SITE: Fire alarm.

1 / 2 Question 65 500000 High Containment Hydrogen Concentration EK3.04 - Knowledge of the reasons for the following responses as they apply to HIGH PRIMARY CONTAINMENT HYDROGEN CONCENTRATIONS:

Emergency depressurization An acceptable question could not be developed for the K/A due to lack of emergency depressurization based on hydrogen concentration at the facility.

Randomly resampled K/A 500000 High Containment Hydrogen Concentration EK3.06 -

Knowledge of the reasons for the following responses as they apply to HIGH PRIMARY CONTAINMENT HYDROGEN CONCENTRATIONS: Operation of wet well vent.

3 Question 67 2.1.19 - Ability to use plant computers to evaluate system or component status.

Resampled for better balance of coverage because the generic K/A was already used on the SRO exam.

Randomly resampled K/A 2.1.28 - Knowledge of the purpose and function of major system components and controls.

3 Question 70 2.2.42 - Ability to recognize system parameters that are entry-level conditions for Technical Specifications.

Resampled for better balance of coverage because the generic K/A was already used on another question.

Randomly resampled K/A 2.2.2 - Ability to manipulate the console controls as required to operate the facility between shutdown and designated power levels.

3 Question 75 2.4.32 - Knowledge of operator response to loss of all annunciators.

An acceptable question could not be developed for the K/A without overlapping the previous NRC exam.

Randomly resampled K/A 2.4.34 - Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects.

ES-401 Record of Rejected K/As Form ES-401-4 1 / 2 Question 83 295013 High Suppression Pool Temperature AA2.02 - Ability to determine and/or interpret the following as they apply to HIGH SUPPRESSION POOL TEMPERATURE:

Localized heating/stratification An acceptable question could not be developed for the K/A at the SRO level.

Randomly resampled K/A 295013 High Suppression Pool Temperature AA2.01 - Ability to determine and/or interpret the following as they apply to HIGH SUPPRESSION POOL TEMPERATURE: Suppression pool temperature.

2 / 1 Question 90 262001 AC Electrical Distribution A2.07 - Ability to (a) predict the impacts of the 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: Energizing a dead bus Resampled for better balance of coverage due to many other AC power questions.

Randomly resampled K/A 218000 Automatic Depressurization System A2.06 - Ability to (a) predict the impacts of the 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: ADS initiation signals present.

2 / 2 Question 91 214000 Rod Position Information A2.02 - Ability to (a) predict the impacts of the following on the ROD POSITION INFORMATION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Reactor SCRAM An acceptable question could not be developed for the K/A at the SRO level.

Randomly resampled K/A 214000 Rod Position Information A2.01 - Ability to (a) predict the impacts of the following on the ROD POSITION INFORMATION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Failed reed switches.

3 Question 95 2.1.9 - Ability to direct personnel activities inside the control room.

Resampled because the K/A is already tested extensively on the operating exam.

Randomly resampled K/A 2.1.5 - Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.

ES-401 Record of Rejected K/As Form ES-401-4 3

Question 100 2.4.20 - Knowledge of the operational implications of EOP warnings, cautions, and notes.

An acceptable question could not be developed for the K/A generically and at the SRO level.

Randomly resampled K/A 2.4.22 - Knowledge of the bases for prioritizing safety functions during abnormal/emergency operations.

1 / 2 Question 61 295017 Abnormal Offsite Release Rate AA2.01 - Ability to determine and/or interpret the following as they apply to HIGH OFF-SITE RELEASE RATE: Off-site release rate: Plant-Specific Resampled to limit overlap with the operating exam.

Randomly resampled K/A 295022 Loss of Control Rod Drive Pumps AA2.01 - Ability to determine and/or interpret the following as they apply to LOSS OF CRD PUMPS: Accumulator pressure.

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