ML21032A323

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Draft Written Examination and Operating Test Outlines (Folder 2)
ML21032A323
Person / Time
Site: Nine Mile Point 
Issue date: 02/01/2021
From: Brian Fuller
Operations Branch I
To:
Exelon Nuclear Generation Corp
Fuller B
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ML19309G108 List:
References
CAC 00500
Download: ML21032A323 (35)
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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 NRG Administrative Topic Type Describe activity to be performed (see Note)

Code*

Perform Control Rod Position Verification and Conduct of Operations P,D,R Determine Reactivity Event Severity 2017 NRG N1-OP-42, OP-AA-300, N1-OP-5, KIA 2.1.37 (4.3)

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

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

Determine Radiological and Heat Stress Requirements Related to Operator Work in High Radiation Areas -

Radiation Control D,R Steam Leak in ECIV Room RP-AA-10/11/12/403/460, SA-AA-111, KIA 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 (:5 3 for ROs; s 4 for SROs & RO retakes)

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

(P)revious 2 exams (:5 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 Type Describe activity to be performed (see Note)

Code*

Perform Control Rod Position Verification and P,D,R Determine Reactivity Event Severity and Notification Conduct of Operations 2017 NRC Requirements OP-AA-300, N1-OP-5, N1 -OP-42, KIA 2.1.37 (4.6)

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

Review and Approval of Completed Surveillance Test, N 1-ST-Q6A, Containment Spray System Loop 111 Equipment Control N,R Quarterly Operability Test N1-ST-Q6A, KA 2.2.12 (4.1)

Determine Radiological and Heat Stress Requirements Related to Operator Work in High Radiation Areas -

Radiation Control D,R Steam Leak in ECIV Room RP-AA-10/11/12/403/460, SA-AA-111, KIA 2.3.7 (3.6)

Post-Scenario Emergency Event Classification Emergency Procedures/Plan M,S EP-AA-1013, KIA 2.4.29 (4.4)

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

  • Type Codes & Criteria:

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

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

(P)revious 2 exams (s 1; randomly selected)

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: 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 I JPM Title Type Code*

Safety Function

a. Shift Reactor Building Operating Exhaust and Supply Fans D,S,A 9

KIA 288000 A4.01 (3.1/2.9), N1-OP-10

b. Shift Feedwater Pressure and Level Channels N,S 2

KIA 259002 A4.06 (3.1/3.2), N1-OP-16

c. Respond to a Recirculation Pump Seal Failure M,S,A 1

KIA 202001 A2.10 (3.5/3.9) N1-SOP-1.2

d. Place Containment Spray in Torus Cooling D,S,A 5

KIA 219000 A4.02 (3.7/3.5) N1-EOP-1

e. Alternate RPV Slowdown Through the Reactor Head Vent Valves M,S,A 3

KIA 239001 A2.09 (3.4/3.7), N1-EOP-1, N1-EOP-8

f. Place 11 Shutdown Cooling Loop in Service D,S,A,L 4

KIA 205000 A4.01 (3.7/3.7), N1-OP-4

g. Channel 11 Non-Coincident Scram Test (RO Only)

P, D,S,L 7

KIA 215004 A4.05 (3.1/3.2), N1-ST-R4, N1-OP-5 (2017 NRC)

h. EOG 103 Control Room Start Following Station Blackout P,D,S,EN 6

KIA 295003 AA1.02 (4.2/4.3), N1-OP-45 (2017 NRC)

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

i. Respond to CLC Makeup Tank Level Alarm D,A,R 8

KIA 295018 AA2.04 (2.9/2.9), N1-ARP-H1

j. Initiate Emergency Condenser Locally D,L,E,R 4

KIA 207000 A2.08 (3.8/3.8), N1-OP-13

k. Lineup to Flood the Reactor Vessel Using the Diesel Fire Pump M,E,R 2

KIA 295031 EA1.08 (3.8/3.9)

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 f then g Criteria for RO/ SRO-I / SRO-U 4-6 I 4-6 I 2-3

S9/S8/S4 2:1/2:1/2:1 2: 1 / 2:1 / 2:1 (control room system) 2:1/2:1/2:1 2:2/2:2/2:1 s 3 / s 3 / s 2 (randomly selected) 2:1/2:1/2:1

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

Event Event No.

No.

Type*

Description N-Start Circulating Water Pump 11 1

N/A

BOP, SRO N1-OP-19 R-Raise Reactor Power with Recirculation Flow 2

N/A

ATC, SRO N1-OP-1, N1 -OP-43B 3

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

HV02 TS-SRO L 1-3-4, L 1-1-5, N1-EOP-5, N1-OP-10, Technical Specifications RR65E I-All RR pump 15 Blind Controller failure and delayed pump trip 4

TS-SRO RR09E N1-SOP-1.3, Tech Spec 3.1.7 Fuel Failure 5

RX01 C-AII N1-SOP-25.2, N1-SOP-1.1, N1-SOP-1 ED27 Powerboard 12 Fails to Auto Transfer and Feedwater Pump 11 6

C-BOP Trips FW03A N1-SOP-30.2, N1-SOP-1 7

MS01 M-AII Main Steam Line Break in Turbine Building N1-EOP-2, N1-EOP-6 MS13A Two MSIVs Fail to Close 8

C-AII MS13C 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, (l)nstrument, (C)omponent, (M)ajor

Facility: Nine Mile Point Unit 1 Scenario No.: NRC-3 Op-Test No.: LC119-1 NRC

1. Malfunctions after EOP entry (1-2) 4 Event6,7,8,9
2. Abnormal events (2-4) 4 Events 3, 4, 5, 6
3. Major transients (1-2) 1 Event7
4. EOPs entered/requiring substantive actions (1-2) 2 N1-EOP-2, N1-EOP-6
5. Entry into contingency EOP with substantive actions (at least 1 per scenario set) 1 N1-EOP-8
6. Preidentified Critical tasks (at least 2) 2 CRITICAL TASK DESCRIPTIONS:

CRITICAL TASK JUSTIFICATION:

CT-1.0: Given fuel failure causing Main Steam Line radiation levels to High Main Steam Line radiation levels approach and/or exceed 3.75 times normal full power background, scram the indicate fuel failure and release of fission Reactor, in accordance with N1-SOP-25.2.

products to the Reactor coolant. A Reactor scram is required by N 1-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 An un-isolable primary system primary and secondary containments and off-site release rate approaching discharging outside of Primary and the General Emergency level, perform an RPV Slowdown, in accordance Secondary Containments resulting in off-with N1-EOP-6.

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 N 1-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 N 1-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 N 1-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 Malf.

Event Event No.

No.

Type*

Description 1

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

N/A R-ATC, Rod Sequence Exchange SRO N1-OP-5, ReMA Stuck control rod (2017 NRC Scenario 1) 3 RD04 C-ATC N1-OP-5 4

EC03B C-BOP, Emergency Condenser 12 Inadvertent Initiation TS-SRO ARP K1-1-5, N1-OP-13, Technical Specifications RD34 Instrument air leak, Reactor scram required 5

C-AII IA01 N1-SOP-20.1, N1-SOP-1 ATWS 6

RD33 M-AII N1-EOP-2, N1-EOP-3 Feedwater Isolation Valves 11 and 12 fail to isolate 7

Overrides C-AII N1-EOP-3 (N)ormal, (R)eactivity, (l)nstrument, (C)omponent (M)ajor

Facility: Nine Mile Point Unit 1 Scenario No.: NRC-2 Op-Test No.: LC119-1 NRC

1. Malfunctions after EOP entry (1-2) 2 Event 6, 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) 1 N1-EOP-2
5. Entry into contingency EOP with substantive actions (at least 1 per scenario set) 1 N1-EOP-3
6. Preidentified Critical tasks (at least 2) 3 CRITICAL TASK DESCRIPTIONS:

CRITICAL TASK JUSTIFICATION:

CT-1.0: Given lowering Instrument Air System pressure, the crew will insert Lowering instrument air pressure and a manual reactor scram when air pressure lowers to <70 psig, in accordance CRD air pressure can result in various with N1-ARP-F3 and/or N1-SOP-20.1.

rods drifting to unpredictable locations in the core. This can lead to an unanalyzed rod pattern and localized power peaking resulting in fuel damage.

CT-2.0: Given a failure of the reactor to scram with power above 6% and High Reactor power after a scram RPV water level above -41 inches, the crew will terminate and prevent all represents a challenge to nuclear fuel injection except boron and CRD, in accordance with N1-EOP-3.

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-3.0 Given a failure of the reactor to scram with power above 6%, the crew Inserting control rods lowers Reactor will lower reactor power by inserting control rods or injecting boron, in power, which reduces challenges to the accordance with N1-EOP-3.

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 N1-0P-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 Malf.

Event Event No.

No.

Type*

Description 1

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

NM19C I-ATC, APRM 13 fails upscale SRO ARP (2017 NRC Scenario 4)

C-BOP ERV Inadvertently opens (2017 NRC Scenario 3) 3 AD05A R-ATC N 1-SOP-1.4, N 1-SOP-1.1, Technical Specifications TS-SRO C-AII Powerboard 12 Electrical Fault 4

ED05 TS-SRO N1-SOP-30.2, N1-SOP-1.3, N1-SOP-1.1, Technical Specifications EC01 Steam leak inside Drywell 5

M-AII N1-EOP-2, N1-EOP-4 6

PC10A C-AII Failed open Torus to Drywell vacuum breaker PC10C N1-EOP-4 (2017 NRC Scenario 3)

FW28A C-BOP, HPCI fails to initiate, Core Spray fails to auto-inject 7

FW28B SRO N1-EOP-2 (2017 NRC Scenario 3)

CS07 8

CT01A C-ATC, Containment Spray pumps 111 and 112 trip CT01B SRO N1 -EOP-8, N1-EOP-4 (N)ormal, (R)eactivitv, (l)nstrument, (C)omponent, (M)aior

Facility: Nine Mile Point Unit 1 Scenario No.: NRC-3 Op-Test No.: LC119-1 NRC

1. Malfunctions after EOP entry (1-2) 3 Event6,7,8
2. Abnormal events (2-4) 3 Events 2, 3, 4
3. Major transients (1-2) 1 Event 5
4. EOPs entered/requiring substantive actions (1-2) 2 N1-EOP-2, N1-EOP-4
5. Entry into contingency EOP with substantive actions (at least 1 per scenario set) 1 N1-EOP-8
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 A manual Reactor scram is required manual scram prior to Torus temperature exceeding 110°F, in accordance before Torus temperature exceeds with N1-SOP-1.4 110°F. 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 Maintaining Reactor water level above -

crew will inject with preferred and alternate injection systems to restore and 84 inches ensures adequate core cooling maintain RPV water level above -84 inches, in accordance with N1-EOP-2.

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 A Slowdown is required to limit further capability, the crew will execute N1-EOP-8, RPV Slowdown, when it is release of energy into the Primary determined Torus pressure cannot be maintained inside the Pressure Containment and to ensure that the RPV Suppression Pressure limit, in accordance with N1-EOP-4.

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 N 1-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 Slowdown (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 Malf.

Event Event No.

No.

Type*

Description Transfer Torus Water to the Waste Collection Tank 1

N/A N-BOP N1-OP-14 2

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

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

RP20B TS-SRO Drywell Pressure transmitter Failed Low Technical Specifications C-PC05

BOP, Seismic Event with Circulating Water Pump trip.

5 SRO CW06B N1-SOP-28, N1-SOP-1.1 R-ATC 6

EG11 C-AII Degraded 345KV Grid conditions N 1-SOP-33B.1, N 1-SOP-1 7

RR29 M-AII Coolant leak in Drywell N1-EOP-2, N1-EOP-4 FW03A Trip of Feedwater Pumps 8

C-AII FW03B N1-EOP-2, N1-EOP-8 (N)ormal, (R)eactivity, (l)nstrument, (C)omoonent, (M)aior

Facility: Nine Mile Point Unit 1 Scenario No.: NRC-4 Op-Test No.: LC119-1 NRC

1. Malfunctions after EOP entry (1-2) 2 Events 7, 8
2. Abnormal events (2-4) 5 Events 2, 3, 4, 5, 6
3. Major transients (1-2) 1 Event 7
4. EOPs entered/requiring substantive actions (1-2) 2 N1-EOP-2, N1-EOP-4
5. Entry into contingency EOP with substantive actions (at least 1 per scenario set) 1 N1-EOP-8
6. Preidentified Critical tasks (at least 2) 2 CRITICAL TASK DESCRIPTIONS:

CRITICAL TASK JUSTIFICATION:

CT-1.0: Given a LOCA in the Drywell, the crew will initiate Containment Initiating Containment Sprays reduces Sprays prior to exceeding the Pressure Suppression Pressure limit, in Primary Containment pressure. This accordance with N1-EOP-4.

reduces stresses on the Drywall 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 Maintaining Reactor water level above -

crew will depressurize the RPV and inject with Preferred and Alternate 84 inches ensures adequate core cooling Injection Systems to restore and maintain RPV water level above -84 inches, through the preferred method of core in accordance with N1-EOP-2.

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 Slowdown before RPV water level drops below -109 inches, in accordance with N 1-EOP-2 (Critical Task).

Appendix D Scenario Outline Form ES-D-1 Facility: Nine Mile Point Unit 1 Scenario No.: NRC-5 Op-Test No.: LC1 19-1 NRC Examiners:

Operators:

Initial Conditions: The plant is operating at approximately 2-3% power. Reactor Building Exhaust Fan 12 is out of service for maintenance. IRM 11 is bypassed due to spiking.

Turnover: Power ascension is on hold to place the EPR in Service.

Event Malf.

Event Event No.

No.

Type*

Description N/A N-BOP, Place the EPR in control 1

SRO N1-OP-43A, N1-OP-31, 2

N/A R-ATC, Raise reactor power with control rods SRO 3

NM01C I-ATC, SRM Fails Upscale SRO N1-OP-38A (2017 NRC Scenario 1) 4 RD05 C-ATC Control Rod Uncouples During Coupling Check N1-OP-5, Technical Specifications Main Steam Line Radiation Monitor Failure 5

RM51A TS-SRO ARP F1 7, Technical Specifications CW02A Sequential Loss of Service Water 6

C-AII CW02B N1-SOP-18.1 RP05A I-BOP, RPS fails to scram the reactor, ARI is successful 7

RP05B SRO N1-EOP-2, N1-SOP-1 (2017 NRC Scenario 1) 8 EC02 M-AII EC 11 Steam Leak in the Reactor Building (2017 NRC Scenario 1)

ARP K1-1-1, K1-3-4, N1-SOP-1, N1-EOP-2, N1-EOP-5 EC07A EC 11 will not isolate, requiring a RPV Slowdown 9

EC0BA C-AII N1-EOP-8 (2017 NRC Scenario 1)

EC0BB (N)ormal, (R)eactivitv, (l)nstrument, (C)omponent, (M)aior

Facility: Nine Mile Point Unit 1 Scenario No.: NRC-5 Op-Test No.: LC1 19-1

1. Malfunctions after EOP entry (1-2) 3 Events 7, 8, 9
2. Abnormal events (2-4) 5 Events 3,4,5,6, 7
3. Major transients (1-2) 1 Events
4. EOPs entered/requiring substantive actions (1-2) 2 N1-EOP-2, N1-EOP-5
5. Entry into contingency EOP with substantive actions (at least 1 per scenario set) 1 N1-EOP-8
6. Preidentified Critical tasks (at least 2) 3 CRITICAL TASK DESCRIPTIONS:

CRITICAL TASK JUSTIFICATION:

CT-1.0: Given the plant at power with a trip of both Service Water Pumps, Service Water is the ultimate heat sink for the crew will start ESW pumps to restore cooling to vital plant loads in the RBCLC system. Failure to restore accordance with N1-SOP-18.1.

cooling to this system will result in a loss of cooling to equipment important to safe plant operation, such as recirculation pump seals and RWCU, and would result in unnecessary challenge to plant safety.

CT-2.0: Given a LOCA in the Drywell and a failure of HPCI to initiate, the Inserting control rods during a transient crew will inject with preferred and alternate injection systems to restore and lowers reactor power, which reduces maintain RPV water level above -84 inches, in accordance with N1-EOP-2.

challenges to the plant during the transient. With RPS failing to trip, the crew must rely on backup shutdown methods to ensure control rods are inserted to provide long term, stable, core shutdown.

CT-3.0: Given a LOCA in the Drywell and degraded Containment Spray An un-isolable primary system capability, the crew will execute N1-EOP-8, RPV Slowdown, when it is discharging outside of Primary determined Torus pressure cannot be maintained below the Pressure Containment resulting in two general area Suppression Pressure limit, in accordance with N1-EOP-4.

temperatures above the maximum safe limit indicates a wide-spread problem posing a direct and immediate threat to 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.

SCENARIO

SUMMARY

The scenario begins with the plant in a startup at approximately 2-3% power and raising power by control rod withdrawal. Control rod withdrawal has been stopped while the crew places the Electric Pressure Regulator (EPR) in service per N1-0P-43A and N1 -OP-31. When the EPR is in service the crew will continue the startup by withdrawing control rods. While withdrawing control rods, an SRM upscale trip will occur. The crew will respond per ARP F3-4-1 and N1-0P-3BA to bypass the SRM.

While continuing to withdraw control rods and performing a coupling check on a fully withdrawn rod, the rod becomes uncoupled. The crew will recouple the control rod per N1 -OP-5. While the control rod is uncoupled, Tech Spec 3.1.1.a(2) applies.

After the control rod is recoupled, a Main Steam Line Rad monitor will fail inoperable. The SRO will address Tech Specs.

Next, a sequential loss of Service Water occurs. SWP 11 trips first. The crews tries to restore Service Water with SWP 12, but that also eventually fails, resulting in a loss of all Service Water. The crew will exercise the override in N1-SOP-18.1, including starting ESW Pumps (Critical Task) and scramming the reactor. The Mode Switch will fail to scram the Reactor, however manual ARI actuation will result in successful control rod insertion (Critical Task).

Next, a steam leak will develop from Emergency Condenser 11. The crew will attempt to isolate the leak, however the Emergency Condenser will fail to isolate both automatically and manually. Two General Areas of the Reactor Building will exceed the maximum safe temperatures. The crew will blowdown the Reactor per N1-EOP-B (Critical Task).

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

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

Total A2 G*

Total

1.

1 3

3 3

3 4

4 20 4

3 7

Emergency and 2

1 1

2 N/A 1

1 N/A 1

7 2

1 3

Abnormal Plant Evolutions Tier Totals 4

4 5

4 5

5 27 6

4 10

2.

1 3

1 2

2 2

2 2

3 3

3 3

26 3

2 5

Plant 2

1 1

1 1

1 1

1 1

2 1

1 12 0

2 1

3 Systems Tier Totals 4

2 3

3 3

3 3

4 5

4 4

38 5

3 8

1 2

3 4

1 2

3 4

3. Generic Knowledge and Abilities Categories 10 7

3 2

2 3

2 2

1 2

Note: 1.

Ensure that at least two topics from every applicable KIA category are sampled within each tier of the RO and SRO-only outline sections (i.e., except for one category in Tier 3 of the SRO-only section, 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 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 O.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) Kl As 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 O.1.b of ES-401 for the applicable Kl As.
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, I Rs, and point totals {#} on Form ES-401-3. Limit SRO selections to Kl As that are linked to 10 CFR 55.43.

G* Generic KIAs

. These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the KIA catalog is used to develop the sample plan. They are 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 Page 2 of 16 Form ES-401-1 Nine Mile Point Unit 1 (Rev. 1)

BWR Examination Outline Fo1111 Erner :1encv and Abnormal Plant Evolutions - Tier 1/Group 1 (RO/SRO)

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

KIA Topic(s)

IA Q#

Function 295001 (APE 1) Partial (295001 G2.1.19) PARTIAL OR COMPLETE LOSS OF 21 or Complete Loss of X

FORCED CORE FLOW CIRCULATION / 1 & 4: Ability to 3.8 76 Forced Core Flow use plant computers to evaluate system or component Circulation / 1 & 4 status.

295003 (APE 3) Partial (295003AA2.01) Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE 22 or Complete Loss of AC X

LOSS OF A.C. POWER: Cause of partial or complete 3.7 77 Power / 6 loss of A.C. power 295005 (APE 5) Main (295005AA2.08) Ability to determine and/or interpret the 23 Turbine Generator Trip /

X following as they apply to MAIN TURBINE GENERATOR 3.3 78 3

TRIP: Electrical distribution status 295016 (APE 16)

(295016AA2.03) Ability to determine and/or interpret the 24 Control Room X

following as they apply to CONTROL ROOM 4.4 79 Abandonment I 7 ABANDONMENT: Reactor pressure (295019G2.4.9) PARTIAL OR COMPLETE LOSS OF 295019 (APE 19) Partial INSTRUMENT AIR / 8: 2.4.9 Knowledge of low 25 or Complete Loss of X

power/shutdown implications in accident (e.g., loss of 4.2 80 Instrument Air / 8 coolant accident or loss of residual heat removal) mitioation strateoies.

295037 (EPE 14) Scram (295037G2.4.41) SCRAM CONDITION PRESENT AND Condition Present and REACTOR POWER ABOVE APRM DOWNSCALE OR 26 Reactor Power Above X

UNKNOWN / 1: Knowledge of the emergency action level 4.6 81 APRM Downscale or Unknown / 1 thresholds and classifications.

700000 (APE 25)

(700000AA2.01) Ability to determine and/or interpret the 27 Generator Voltage and X

following as they apply to GENERATOR VOLTAGE AND 3.6 82 Electric Grid ELECTRIC GRID DISTURBANCES: Operating point on Disturbances / 6 the aenerator caoabilitv curve 295001 (APE 1) Partial (295001 AK1.04) Knowledge of the operational implications of the following concepts as they apply to 1

or Complete Loss of X

PARTIAL OR COMPLETE LOSS OF FORCED CORE 2.5/3.3 39 Forced Core Flow FLOW CIRCULATION: Limiting cycle oscillation: Plant-Circulation/ 1 & 4 Specific 295003 (APE 3) Partial (295003AK2.04) Knowledge of the interrelations between 2

or Complete Loss of AC X

PARTIAL OR COMPLETE LOSS OF A.C. POWER and 3.4/3.5 40 Power/ 6 the followina: A.C. electrical loads 295004 (APE 4) Partial (295004AK3.01) Knowledge of the reasons for the 3

or Total Loss of DC X

following responses as they apply to PARTIAL OR 2.6/3.1 41 Power/ 6 COMPLETE LOSS OF D.C. POWER: Load shedding:

Plant-Specific 295005 (APE 5) Main (295005AA 1.04) Ability to operate and/or monitor the 4

Turbine Generator Trip/

X following as they apply to MAIN TURBINE GENERATOR 2.7/2.8 42 3

TRIP: Main generator controls 295006 (APE 6) Scram /

(295006AA2.06) Ability to determine and/or interpret the 5

X following as they apply to SCRAM: Cause of reactor 3.5/3.8 43 1

SCRAM 295016 (APE 16)

(295016G2.1.32) CONTROL ROOM ABANDONMENT/

6 Control Room X

7: Ability to explain and apply system limits and 3.8/4.0 44 Abandonment/ 7 precautions.

295018 (APE 18) Partial (295018AA2.05) Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE 7

or Complete Loss of X

LOSS OF COMPONENT COOLING WATER: System 2.9/2.9 45 CCW/8 pressure 295019 (APE 19) Partial (295019G2.4.47) PARTIAL OR COMPLETE LOSS OF 8

or Complete Loss of X

INSTRUMENT AIR/ 8: Ability to diagnose and recognize 4.2/4.2 46 trends in an accurate and timely manner utilizing the Instrument Air/ 8 aoorooriate control room reference material.

295021 (APE 21) Loss (295021 AK1.04) Knowledge of the operational 9

X implications of the following concepts as they apply to 3.6/3.7 47 of Shutdown Cooling / 4 LOSS OF SHUTDOWN COOLING: Natural circulation 295023 (APE 23)

(295023AK2.01) Knowledge of the interrelations between 10 X

REFUELING ACCIDENTS and the following: Fuel 3.3/3.7 48 Refueling Accidents / 8 handlinQ equipment 295024 (EPE 1) High (295024EK3.06) Knowledge of the reasons for the 11 X

following responses as they apply to HIGH DRYWELL 4.0/4.1 49 Drywall Pressure/ 5 PRESSURE: Reactor SCRAM I

295025 (EPE 2) High (295025EA 1.06) Ability to operate and/or monitor the 12 X

following as they apply to HIGH REACTOR PRESSURE:

4.5/4.5 50 Reactor Pressure I 3 Isolation condenser: Plant-Specific

ES-401 Page 3 of 16 Form ES-401-1

_ ru,m u11u 1 (Rev. 1)

BWR Examination Outline Form ES-401-1 Emer ~ency and Abnormal Plant Evolutions - Tier 1/Group 1 (RO/SRO) 11 E/APE #/Name/Safety K1 K2 K3 A1 A2 G*

KIA Topic(s)

IR Q#

Item Function 295026 (EPE 3)

(295026EA2.03) Ability to determine and/or interpret the 13 Suppression Pool High X

following as they apply to SUPPRESSION POOL HIGH 3.9/4.0 51 Water Temperature/ 5 WATER TEMPERATURE: Reactor pressure 295028 (EPE 5) High (295028G2.4.3) HIGH DRYWELL TEMPERATURE 14 Drywell Temperature X

(MARK I AND MARK II ONLY)/ 5: Ability to identify post-3.7/3.9 52 (Mark I and Mark II only) accident instrumentation.

/5 295030 (EPE 7) Low (295030EK1.02) Knowledge of the operational implications of the following concepts as they apply to 15 Suppression Pool Water X

LOW SUPPRESSION POOL WATER LEVEL: Pump 3.5/3.8 53 Level/ 5 NPSH 295031 (EPE 8) Reactor (295031 EK2.02) Knowledge of the interrelations between 16 X

REACTOR LOW WATER LEVEL and the following:

3.8/3.9 54 Low Water Level/ 2 Reactor pressure 295037 (EPE 14) Scram (295037EK3.07) Knowledge of the reasons for the Condition Present and following responses as they apply to SCRAM 17 Reactor Power Above X

CONDITION PRESENT AND REACTOR POWER 4.2/4.3 55 APRM Downscale or ABOVE APRM DOWNSCALE OR UNKNOWN: Various Unknown/ 1 alternate methods of control rod insertion: Plant-Specific 295038 (EPE 15) High (295038EA 1.06) Ability to operate and/or monitor the 18 Offsite Radioactivity X

following as they apply to HIGH OFF-SITE RELEASE 3.5/3.6 56 Release Rate / 9 RATE: Plant ventilation 600000 (APE 24) Plant (600000AA2.03) Ability to determine and interpret the 19 X

following as they apply to PLANT FIRE ON SITE: Fire 2.8/3.2 57 Fire On Site/ 8 alarm 700000 (APE 25)

(700000G2.2.12) GENERATOR VOLTAGE AND 20 Generator Voltage and X

ELECTRIC GRID DISTURBANCES/ 6: Knowledge of 3.7/4.1 58 Electric Grid Disturbances / 6 surveillance procedures.

KIA Cateaorv 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 /

K1 K2 K3 A1 A2 G*

KIA Topic(s)

IR Q#

Safety Function 295013 (APE 13) High (295013AA2.01) Ability to determine and/or interpret the 35 Suppression Pool X

following as they apply to HIGH SUPPRESSION POOL 4.0 83 Temperature. 15 TEMPERATURE: Suppression pool temperature 295014 (APE 14)

(295014G2.1.23) INADVERTENT REACTIVITY ADDITION/ 1: Ability to perform specific system and 36 Inadvertent Reactivity X

integrated plant procedures during all modes of plant 4.4 84 Addition/ 1 operation.

295029 (EPE 6) High (295029EA2.03) Ability to determine and/or interpret the 37 Suppression Pool Water X

following as they apply to HIGH SUPPRESSION POOL 3.5 85 Level/ 5 WATER LEVEL: Drywall/containment water level 295012 (APE 12) High (295012AK3.01) Knowledge of the reasons for the 28 Drywall Temperature/ 5 X

following responses as they apply to HIGH DRYWELL 3.5/3.6 59 TEMPERATURE: Increased drywall cooling 29 295015 (APE 15)

X (295015AA 1.02) Ability to operate and/or monitor the 4.0/4.2 60 Incomplete Scram/ 1 following as they apply to INCOMPLETE SCRAM: RPS 295022 Loss of Control (295022AA2.01) Ability to determine and/or interpret the 30 Rod Drive Pumps X

following as they apply to LOSS OF CAD PUMPS:

3.5/3.6 61 Accumulator pressure 295032 (EPE 9) High (295032G2.1.20) HIGH SECONDARY CONTAINMENT 31 Secondary Containment X

AREA TEMPERATURE/ 5: Ability to interpret and 4.6/4.6 62 Area Temperature I 5 execute procedure steps.

295035 (EPE 12)

(295035EK1.01) Knowledge of the operational 32 Secondary Containment X

implications of the following concepts as they apply to 3.9/4.2 63 High Differential SECONDARY CONTAINMENT HIGH DIFFERENTIAL Pressure/ 5 PRESSURE: Secondary containment intearity 295036 (EPE 13)

(295036EK2.01) Knowledge of the interrelations between 33 Secondary Containment X

SECONDARY CONTAINMENT HIGH SUMP/AREA 3.1/3.2 64 High Sump/Area Water WATER LEVEL and the following: Secondary Level/ 5 containment equipment and floor drain system 500000 (EPE 16) High (500000EK3.06) Knowledge of the reasons for the 34 Containment Hydrogen X

following responses as they apply to HIGH PRIMARY 3.1/3.7 65 CONTAINMENT HYDROGEN CONCENTRATIONS:

Concentration / 5 Operation of wet well vent II KIA 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 K1 K2 K3 K4 KS K6 A1 A2 A3 A4 G*

KIA Topic(s)

IA Q#

Function (206000A2.17) Ability to (a) predict the impacts of the following on the HIGH PRESSURE COOLANT INJECTION SYSTEM; and 206000 (SF2, SF4 (b) based on those 64 HPCIS) High-Pressure X

predictions, use procedures 4.3 86 Coolant Injection to correct, control, or mitigate the consequences of those abnormal conditions or operations:

HPCI inadvertent initiation:BWR-2,3,4 (209001 G2.1.30) 209001 (SF2, SF4 LOW-PRESSURE CORE 65 LPCS) Low-Pressure X

SPRAY: Ability to locate and 4.0 87 Core Spray operate components, includina local controls.

(211 000A2.04) Ability to (a) predict the impacts of the following on the STANDBY LIQUID CONTROL SYSTEM; and (b) based on 66 211000 (SF1 SLCS)

X those predictions, use 3.4 88 Standby Liquid Control procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Inadequate system flow (261000G2.2.25) STANDBY GAS TREATMENT:

67 261000 (SF9 SGTS)

X Knowledge of the bases in 4.2 89 Standby Gas Treatment Technical Specifications for limiting conditions for ooerations and safetv limits.

A2.06 - Ability to (a) predict the impacts of the following on the AUTOMATIC DE PRESSURIZATION 218000 Automatic SYSTEM; and (b) based on 68 Depressurization X

those predictions, use 4.3 90 System procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: ADS initiation sianals present.

(205000A2.08) Ability to (a) predict the impacts of the following on the SHUTDOWN COOLING SYSTEM(RHR SHUTDOWN COOLING 38 205000 (SF4 SCS)

X MODE); and (b) based on 3.3/3.5 1

Shutdown Cooling those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of heat exchanaer coolino (206000A3.05) Ability to monitor automatic 206000 (SF2, SF4 operations of the HIGH 39 HPCIS) High-Pressure X

PRESSURE COOLANT 4.3/4.3 2

II Coolant Injection INJECTION SYSTEM including: Reactor water level:BWR-2,3,4

ES-401 Page 6 of 16 Form ES-401-1

- *****- r umt Unit 1 (Rev. 1)

BWR Examination Outline Form ES-401-1 Plant Svstems - Tier 2/Group 1 (RO/SRO) 11 item E/APE #/Name/Safety K1 K2 K3 K4 KS K6 A1 A2 A3 A4 G*

KIA Topic(s)

IA Q#

Function (207000A4.04) ISOLATION (EMERGENCY) 207000 (SF4 IC)

CONDENSER: Ability to 40 Isolation (Emergency)

X manually operate and/or 3.8/4.0 3

Condenser monitor in the control room:

Vent line radiation levels:BWR-2,3 (209001 G2.4.50) 209001 (SF2, SF4 LOW-PRESSURE CORE 41 LPCS) Low-Pressure X

SPRAY: Ability to verify 4.2/4.0 4

Core Spray system alarm setpoints and operate controls identified in the alarm response manual.

(211 000K1.06) Knowledge of the physical connections and/or cause-effect 42 211000 (SF1 SLCS)

X relationships between 3.7/3.7 5

Standby Liquid Control STANDBY LIQUID CONTROL SYSTEM and the following: Reactor vessel (212000K2.01) REACTOR PROTECTION SYSTEM:

43 212000 (SF? RPS)

X Knowledge of electrical 3.2/3.3 6

Reactor Protection power supplies to the following: RPS motor-oenerator sets (215003K3.02) Knowledge of the effect that a loss or 215003 (SF? 1AM) malfunction of the 44 Intermediate-Range X

INTERMEDIATE RANGE 3.6/3.6 7

Monitor MONITOR (1AM) SYSTEM will have on following:

Reactor manual control (215004K4.01) Knowledge of SOURCE RANGE 215004 (SF? SAMS)

MONITOR (SAM) SYSTEM 45 Source-Range Monitor X

design feature(s) and/or 3.7/3.7 8

interlocks which provide for the following: Rod withdrawal blocks (21 S00SKS.01) Knowledge of the operational 215005 (SF? PRMS) implications of the following Average Power Range concepts as they apply to 46 X

AVERAGE POWER 2.8/2.9 9

Monitor/Local Power RANGE MONITOR/LOCAL Range Monitor POWER RANGE MONITOR SYSTEM: LPRM detector ooeration (218000K6.06) Knowledge of the effect that a loss or 218000 (SF3 ADS) malfunction of the following will have on the 47 Automatic X

AUTOMATIC 3.4/3.6 10 Depressurization DEPRESSURIZATION SYSTEM: D.C. power:

Plant-Specific (223002A 1.04) Ability to predict and/or monitor changes in parameters 223002 (SFS PCIS) associated with operating the PRIMARY 48 Primary Containment X

CONTAINMENT 2.6/2.8 11 Isolation/Nuclear Steam ISOLATION Supply Shutoff SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF controls including:

Individual system relay status

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 K1 K2 K3 K4 KS K6 A1 A2 A3 A4 G*

KIA Topic(s)

IA Q#

Function (223002A2.10) Ability to (a) predict the impacts of the following on the PRIMARY CONTAINMENT ISOLATION 223002 (SFS PCIS)

SYSTEM/NUCLEAR STEAM SUPPLY SHUT-49 Primary Containment X

OFF; and (b) based on 3.9/4.2 12 Isolation/Nuclear Steam those predictions, use Supply Shutoff procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of coolant accidents (239002A2.05) Ability to (a) predict the impacts of the following on the RELIEF/SAFETY VALVES; 239002 (SF3 SRV) and (b) based on those 50 X

predictions, use procedures 3.2/3.4 13 Safety Relief Valves to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Low reactor pressure (239002A3.01) Ability to monitor automatic 51 239002 (SF3 SRV)

X operations of the 3.8/3.9 14 Safety Relief Valves RELIEF/SAFETY VALVES including: SRV operation after ADS actuation (259002A3.07) Ability to 259002 (SF2 RWLCS) monitor automatic operations of the REACTOR 52 Reactor Water Level X

WATER LEVEL CONTROL 3.5/3.6 15 Control SYSTEM including: FWRV lockup (259002A4.06) REACTOR WATER LEVEL CONTROL SYSTEM: Ability to manually operate and/or 259002 (SF2 RWLCS) monitor in the control room:

53 Reactor Water Level X

DP/Single/three element 3.1/3.2 16 Control control selector switch:

Plant-Specific SYSTEM:259002 Reactor Water Level Control System Plant-Specific (261 000A4.09) STANDBY GAS TREATMENT 54 261000 (SF9 SGTS)

X SYSTEM: Ability to 2.7/2.7 17 Standby Gas Treatment manually operate and/or monitor in the control room:

Ventilation valves/dampers 55 212000 Reactor X

Knowledge of EOP 3.7/4.7 18 Protection System mitigation strategies.

(262001 G2.1.7) AC ELECTRICAL DISTRIBUTION: 2.1.7 Ability to evaluate plant 56 262001 (SF6 AC) AC X

performance and make 4.4/4.7 19 Electrical Distribution operational judgments based on operating characteristics, reactor behavior, and instrument II interpretation.

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

KIA Topic(s)

IA Q#

Function (262001 K1.02) Knowledge of the physical connections and/or cause-effect 57 262001 (SF6 AC) AC X

relationships between A.G.

3.3/3.6 20 Electrical Distribution ELECTRICAL DISTRIBUTION and the following: D.C. electrical distribution (262002K1.17) Knowledge of the physical connections and/or cause-effect 262002 (SF6 UPS) relationships between 58 Uninterruptable Power X

UNINTERRUPTABLE 3.1/3.3 21 Supply (AC/DC)

POWER SUPPLY (A.C./D.C.) and the following: Scram solenoid valves: Plant-Specific (263000K3.02) Knowledge of the effect that a loss or malfunction of the D.C.

59 263000 (SF6 DC) DC X

ELECTRICAL 3.5/3.8 22 Electrical Distribution DISTRIBUTION will have on following: Components using D.C. control power (i.e. breakers)

(264000K4.01) Knowledge of EMERGENCY 264000 (SF6 EGE)

GENERATORS 60 Emergency Generators X

(DIESEUJET) design 3.5/3.7 23 (Diesel/Jet) EOG feature(s) and/or interlocks which provide for the following: Emergency generator trips (normal)

(300000K5.01) Knowledge of the operational 61 300000 (SF8 IA)

X implications of the following 2.5/2.5 24 Instrument Air concepts as they apply to the INSTRUMENT AIR SYSTEM: Air compressors (400000K6.06) Knowledge 400000 (SF8 CCS) of the effect that a loss or malfunction of the following 62 Component Cooling X

will have on the CCWS:

2.9/2.9 25 Water Heat exchangers and condensers (207000A 1.01) Ability to predict and/or monitor changes in parameters 207000 (SF4 IC) associated with operating 63 Isolation (Emergency)

X the ISOLATION 3.7/3.8 26 Condenser (EMERGENCY)

CONDENSER controls including: Isolation condenser level:BWR-2,3 KIA 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

_ nJlnt Unit 1 (Rev. 1)

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

(

II Item E/APE #/Name /

K1 K2 K3 K4 KS Safetv Function K6 A1 A2 A3 A4 G*

KIA Topic(s)

IA Q#

(214000A2.01) Ability to (a) predict the impacts of the following on the ROD POSITION INFORMATION SYSTEM; and (b) based on 81 214000 (SF? RPIS) Rod X

those predictions, use 3.3 91 Position Information procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Failed reed switches (234000G2.4.31) 234000 (SF8 FH)

FUEL-HANDLING EQUIPMENT: Knowledge of 82 Fuel-Handling X

annunciator alarms, 4.1 92 Equipment indications, or response orocedures.

(245000A2.03) Ability to (a) predict the impacts of the following on the MAIN TURBINE GENERATOR AND AUXILIARY 245000 (SF4 MTGEN)

SYSTEMS; and (b) based 83 Main Turbine X

on those predictions, use 36 93 Generator/ Auxiliary procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of condenser vacuum (201001 A3.08) Ability to monitor automatic 69 201001 (SF1 GROH)

X operations of the CONTROL 3.0/2.9 27 CAD Hydraulic ROD DRIVE HYDRAULIC SYSTEM including: Drive water flow (201003A4.02) CONTROL ROD AND DRIVE 201003 (SF1 CROM)

MECHANISM: Ability to 70 Control Rod and Drive X

manually operate and/or 3.5/3.5 28 Mechanism monitor in the control room:

CAD mechanism position:

Plant-Specific (204000G2.2.42)

REACTOR WATER 204000 (SF2 RWCU)

CLEANUP: Ability to 71 Reactor Water Cleanup X

recognize system 3.9/4.6 29 parameters that are entry-level conditions for Technical Specifications.

(215001 K1.05) Knowledge of the physical connections and/or cause-effect 215001 (SF? TIP) relationships between 72 Traversing In-Core X

TRAVERSING IN-CORE 3.3/3.4 30 Probe PROBE and the following:

Primary containment isolation system:(Not-BWR1)

(216000K2.01) NUCLEAR 216000 (SF? NBI)

BOILER Instrumentation:

73 Nuclear Boiler X

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

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/

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

KIA Topic(s)

IR Q#

Safety Function (223001 K3.03) Knowledge of the effect that a loss or malfunction of the 223001 (SF5 PCS)

PRIMARY CONTAINMENT 74 Primary Containment X

SYSTEM AND 3.4/3.5 32 and Auxiliaries AUXILIARIES will have on following:

ContainmenVdrywell pressure: Plant-Specific.

(233000K4.06) Knowledge of FUEL POOL COOLING 233000 (SF9 FPCCU)

AND CLEAN-UP design 75 Fuel Pool X

feature(s) and/or interlocks 2.9/3.2 33 Cooling/Cleanup which provide for the following: Maintenance of adequate pool level (239001 K5.09) Knowledge of the operational 239001 (SF3, SF4 implications of the following 76 MASS) Main and X

concepts as they apply to 3.4/3.5 34 Reheat Steam MAIN AND REHEAT STEAM SYSTEM: Decay heat removal (241 000K6.08) Knowledge of the effect that a loss or 241000 (SF3 RTPRS) malfunction of the following 77 Reactor/Turbine X

will have on the 3.6/3.7 35 Pressure Regulating REACTOR/TURBINE PRESSURE REGULATING SYSTEM: Reactor power A 1.03 - Ability to predict and/or monitor changes in 256000 Reactor parameters associated with 78 Condensate System X

operating the REACTOR 2.8/2.8 36 CONDENSATE SYSTEM controls including: System pressure (288000A2.05) Ability to (a) predict the impacts of the following on the PLANT VENTILATION SYSTEMS; and (b) based on those 79 288000 (SF9 PVS)

X predictions, use procedures 2.6/2.7 37 Plant Ventilation to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Extreme outside weather conditions: Plant-Specific (290003A3.01) Ability to 290003 (SF9 CRV) monitor automatic 80 Control Room X

operations of the CONTROL 3.3/3.5 38 Ventilation ROOM HVAC including:

Initiation/reconfiguration KIA 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 BWR Examination Outline Form ES-401-3 (Rev. 1)

Plant Systems - Tier 3 (RO/SRO)

RO SRO-only Category KIA#

Topic Item a

IR IR Q#

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 85 4.1 67 controls.

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

86 4.1 68

1. Conduct of Operations 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 88 3.9 95 complement, overtime limitations, etc.

Subtotal 3

2 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 90 4.6 70 between shutdown and designated power levels.

G2.2.17 Knowledge of the process for managing maintenance activities during power 91 3.8 96 operations, such as risk assessments, work prioritization, and coordination

2. Equipment with the transmission system operator.

Control G2.2.25 Knowledge of the bases in Technical Specifications for limiting conditions for 92 4.2 97 operations and safety limits.

Subtotal 2

2 G2.3.13 Knowledge of radiological safety procedures pertaining to licensed operator 93 3.4 71 duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

G2.3.4 Knowledge of radiation exposure limits under normal or emergency 94 3.2 72

3. Radiation conditions.

Control G2.3.11 Ability to control radiation releases.

95 4.3 98 Subtotal 2

1 G2.4.8 Knowledge of how abnormal operating procedures are used in conjunction 96 3.8 73 with EOPs.

G2.4.35 Knowledge of local auxiliary operator tasks during an emergency and the 97 3.8 74 resultant operational effects.

4. Emergency G2.4.34 Knowledge of RO tasks performed outside the main control room during an 98 4.2 75 emeraencv and the resultant operational effects.

Procedures/Plan 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 100 4.4 100 abnormal/emergency operations.

Subtotal 3

2 Tier 3 Point Total 10 7

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

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

Question 12 Randomly resampled the KIA to limit overlap with other questions and previous NRG exam.

223002 Primary Containment Randomly reselected KIA 223002 Primary Isolation/Nuclear Steam Containment Isolation/Nuclear Steam Supply Supply Shutoff Shutoff A2.1 O - Ability to (a) predict the impacts of the following on the PRIMARY A2.02 - Ability to (a) predict CONTAINMENT ISOLATION the impacts of the following SYSTEM/NUCLEAR STEAM SUPPLY SHUT-on the PRIMARY OFF; and (b) based on those predictions, use CONTAINMENT procedures to correct, control, or mitigate the 2 I 1 ISOLATION consequences of those abnormal conditions or SYSTEM/NUCLEAR operations: Loss of coolant accidents.

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 Question 18 Resampled to limit overlap (SGTS sampled three times, generic KIA sampled twice).

261000 Standby Gas Treatment Randomly resampled KIA 212000 Reactor 2 / 1 Protection System 2.4.6 - Knowledge of EOP 2.4.3 - Ability to identify mitigation strategies.

post-accident instrumentation.

Question 27 An acceptable question could not be developed for the Kl A due to lack of automatic system 201001 CRD Hydraulic response based on Reactor power.

A3.06 - Ability to monitor Randomly resampled KIA 201001 CRD Hydraulic automatic operations of the A3.08 - Ability to monitor automatic operations of CONTROL ROD DRIVE the CONTROL ROD DRIVE HYDRAULIC 2/2 HYDRAULIC SYSTEM SYSTEM including: Drive water flow.

including: Reactor power

ES-401 Record of Rejected K/ As Form ES-401-4 Question 32 A question with an appropriate level of difficulty could not be developed for the Kl A.

223001 Primary Containment and Randomly resampled KIA 223001 Primary Auxiliaries Containment and Auxiliaries K3.03 - Knowledge of the effect that a loss or malfunction of the K3.01 - Knowledge of the PRIMARY CONTAINMENT SYSTEM AND 2/2 effect that a loss or AUXILIARIES will have on following:

malfunction of the Containment/drywell pressure: Plant-Specific.

PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES will have on following: Secondary containment Question 36 An operationally relevant question could not be developed for the Kl A because the facility does 268000 Radwaste not typically conduct off-site releases from this system.

A 1.02 - Ability to predict 2/2 and/or monitor changes in Randomly resampled KIA 256000 Reactor parameters associated with Condensate System A 1.03 - Ability to predict operating the RADW ASTE and/or monitor changes in parameters controls including: Off-site associated with operating the REACTOR release CONDENSATE SYSTEM controls including:

System pressure.

Question 46 An acceptable question could not be developed for the KIA due to lack of EOP entry conditions 295019 Partial or Complete associated with the evolution.

Loss of Instrument Air Randomly resampled KIA 295019 Partial or 1 / 1 2.4.2 - Knowledge of Complete Loss of Instrument Air 2.4.47 - Ability system set points, to diagnose and recognize trends in an accurate interlocks and automatic and timely manner utilizing the appropriate actions associated with control room reference material.

EOP entry conditions.

Question 49 An acceptable question could not be developed for the KIA because RPV flooding is not a 295024 High Drywell response to high Drywell pressure.

Pressure Randomly resampled KIA 295024 High Drywell EK3.05 - Knowledge of the Pressure EK3.06 - Knowledge of the reasons for reasons for the following the following responses as they apply to HIGH 1 / 1 responses as they apply to DRYWELL PRESSURE: Reactor SCRAM.

HIGH DRYWELL PRESSURE: RPV flooding

ES-401 Record of Rejected K/ As Form ES-401-4 Question 57 An acceptable question could not be developed for the KIA due to low operational importance/

600000 Plant Fire On Site minutia.

AA2.06 - Ability to Randomly resampled KIA 600000 Plant Fire On 1 / 1 determine and interpret the Site AA2.03 - Ability to determine and interpret following as they apply to the following as they apply to PLANT Fl RE ON PLANT FIRE ON SITE:

SITE: Fire alarm.

Need for pressurizing control room (recirculating mode)

Question 65 An acceptable question could not be developed for the Kl A due to lack of emergency 500000 High Containment depressurization based on hydrogen Hydrogen Concentration concentration at the facility.

EK3.04 - Knowledge of the Randomly resampled KIA 500000 High reasons for the following Containment Hydrogen Concentration EK3.06 -

1 / 2 responses as they apply to Knowledge of the reasons for the following HIGH PRIMARY responses as they apply to HIGH PRIMARY CONTAINMENT CONTAINMENT HYDROGEN HYDROGEN CONCENTRATIONS: Operation of wet well vent.

CONCENTRATIONS:

Emergency depressurization Question 67 Resampled for better balance of coverage because the generic KIA was already used on 2.1.19 - Ability to use plant the SRO exam.

3 computers to evaluate system or component Randomly resampled KIA 2.1.28 - Knowledge of status.

the purpose and function of major system components and controls.

Question 70 Resampled for better balance of coverage because the generic KIA was already used on 2.2.42 - Ability to recognize another question.

system parameters that are 3

entry-level conditions for Randomly resampled KIA 2.2.2 - Ability to Technical Specifications.

manipulate the console controls as required to operate the facility between shutdown and designated power levels.

Question 75 An acceptable question could not be developed for the KIA without overlapping the previous NRC 2.4.32 - Knowledge of exam.

operator response to loss 3

of all annunciators.

Randomly resampled KIA 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 Question 83 An acceptable question could not be developed for the KIA at the SRO level.

(

295013 High Suppression Pool Temperature Randomly resampled KIA 295013 High Suppression Pool Temperature AA2.01 - Ability AA2.02 - Ability to to determine and/or interpret the following as 1 / 2 determine and/or interpret they apply to HIGH SUPPRESSION POOL the following as they apply TEMPERATURE: Suppression pool temperature.

to HIGH SUPPRESSION POOL TEMPERATURE:

Localized heating/stratification Question 90 Resampled for better balance of coverage due to many other AC power questions.

262001 AC Electrical Distribution Randomly resampled KIA 218000 Automatic Depressurization System A2.06 - Ability to (a)

A2.07 - Ability to (a) predict predict the impacts of the following on the the impacts of the following AUTOMATIC DEPRESSURIZATION SYSTEM; on the A.G. ELECTRICAL and (b) based on those predictions, use 2 / 1 DISTRIBUTION; and (b) procedures to correct, control, or mitigate the based on those predictions, consequences of those abnormal conditions or use procedures to correct, operations: ADS initiation signals present.

control, or mitigate the consequences of those abnormal conditions or operations: Energizing a dead bus Question 91 An acceptable question could not be developed for the KIA at the SRO level.

214000 Rod Position Information Randomly resampled KIA 214000 Rod Position Information A2.01 - Ability to (a) predict the A2.02 - Ability to (a) predict impacts of the following on the ROD POSITION the impacts of the following INFORMATION SYSTEM; and (b) based on on the ROD POSITION those predictions, use procedures to correct, 2/2 INFORMATION SYSTEM; control, or mitigate the consequences of those and (b) based on those abnormal conditions or operations: Failed reed predictions, use procedures switches.

to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Reactor SCRAM Question 95 Resampled because the KIA is already tested extensively on the operating exam.

2.1.9 - Ability to direct 3

personnel activities inside Randomly resampled KIA 2.1.5 - Ability to use the control room.

procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.

ES-401 Record of Rejected K/ As Form ES-401-4 Question 100 An acceptable question could not be developed for the KIA generically and at the SRO level.

2.4.20 - Knowledge of the 3

operational implications of Randomly resampled KIA 2.4.22 - Knowledge of EOP warnings, cautions, the bases for prioritizing safety functions during and notes.

abnormal/emergency operations.

Question 61 Resampled to limit overlap with the operating exam.

295017 Abnormal Offsite Release Rate Randomly resampled KIA 295022 Loss of Control Rod Drive Pumps AA2.01 - Ability to 1 / 2 AA2.01 - Ability to determine and/or interpret the following as they determine and/or interpret apply to LOSS OF CRD PUMPS: Accumulator the following as they apply pressure.

to HIGH OFF-SITE RELEASE RATE: Off-site release rate: Plant-Specific

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