ML21137A169

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Final Written Examination and Operating Test Outlines (Folder 3)
ML21137A169
Person / Time
Site: Nine Mile Point Constellation icon.png
Issue date: 05/17/2021
From: Alexander M
Exelon Nuclear Generation Corp
To: Brian Fuller
Operations Branch I
Shared Package
ML20233A469 List:
References
EPID L-2021-OLL-0027
Download: ML21137A169 (30)
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Text

ES-301 Administrative Topics Outline Form ES-301-1 Facility:

Nine Mile Point Unit 2 Date of Examination:

April 2021 Examination Level: RO Operating Test Number:

LC2 20-1 NRC Administrative Topic (see Note)

Type Code*

Describe activity to be performed Conduct of Operations D, R Perform Jet Pump Surveillance N2-OSP-LOG-D001, K/A 2.1.18 (3.6)

Conduct of Operations M, R Perform N2-RESP-001, Core Operating Limits Verification N2-RESP-001, K/A 2.1.7 (4.4)

Equipment Control N, R Evaluate Injection System Vortex Limits N2-EOP-6.29, K/A 2.2.44 (4.2)

Radiation Control Emergency Plan D, S Fire Fighting Response for a Fire in the Protected Area OP-NM-201-005, K/A 2.4.27 (3.4)

NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).

  • Type Codes and Criteria:

(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank ( 3 for ROs; 4 for SROs and 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 2 Date of Examination:

April 2021 Examination Level: SRO Operating Test Number:

LC2 20-1 NRC Administrative Topic (see Note)

Type Code*

Describe activity to be performed Conduct of Operations D, R Determine the Plant Impact for Inoperable Unit Cooler N2-OP-53E, Tech. Specs., K/A 2.1.32 (4.0)

Conduct of Operations D, R Determine Core Thermal Power IAW N2-REP-11 N2-REP-11, K/A 2.1.7 (4.7)

Equipment Control N, R Respond to Notification of a Safety Limit Violation Technical Specifications, K/A 2.2.22 (4.7)

Radiation Control D, P, R (NRC 16-

1)

Offsite Dose Calculation Manual (ODCM)

Assessment for Inoperable Equipment N2-OP-42, ODCM, K/A 2.3.15 (3.1)

Emergency Plan N, R Emergency Plan Classification (Alert, EAL CA5)

NMP Unit 2 EAL Wallboard - EP-AA-1013 Addendum 4 Appendix 1, EP-AA-1013 Addendum 4, EP-AA-112-100-F-01, K/A 2.4.41 (4.6)

NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).

  • Type Codes and Criteria:

(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank ( 3 for ROs; 4 for SROs and 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 2 Date of Examination: April 2021 Exam Level: RO/SRO-I Operating Test No.: LC2 20-1 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. Transfer RCIC Lineup Post-Scram (Alternate Path)

K/A 217000 A4.07 (3.9/3.8), N2-EOP-RPV, N2-OP-35 A,D,S,EN 2

b. Override the Control Room Envelope ACU Cross-Divisional Operating Interlock K/A 290003 A3.01 (3.3/3.5) N2-OP-53A D, S 9
c. Depressurizing the RPV to the Main Condenser K/A 239001 A4.09 (3.9/3.9) N2-EOP-6.18 M,L,S 3
d. HPCS Pump Run Following Maintenance (Alternate Path)

K/A 209002 A4.01 (3.7/3.7) N2-OP-33 A,D,P,S (LC2-NRC16-1) 4

e. Rotate Drywell Unit Coolers (Alternate Path)

K/A 223001 A4.12 (3.5/3.6) N2-OP-60 A,D,S 5

f. Energize Reserve Station Transformer 1B, NPS-SWG003 and NNS-SWG015 K/A 262001 A4.01 (3.4/3.7) N2-SOP-03 D,S 6
g. Perform Weekly RPS Surveillance (Alternate Path)

K/A 212000 A4.01 (4.6/4.6) N2-OSP-RPS-W002, N2-SOP-08 A,D,S 7

h. Secure 2SWP*P1A K/A 400000 A4.01 (3.1 / 3.0) N2-OP-11, N2-OP-58 N,S 8

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

i. Lineup for Boron injection with the Hydro Pump K/A 295037 EA1.10 (3.7/3.9) N2-EOP-6.15 D,E,R 1
j. Reduce Lighting Loads During Station Blackout (2LAC-PNLU02 only)

K/A 295003 AA1.01, (3.7/3.8) N2-SOP-02 N,E 6

k. Align Fire Water System to Inject through RHR B (Alternate Path)

K/A 295031 EA1.08 (3.8/3.9) N2-EOP-6.6 A,N,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 alone c alone d and e f alone g then b h alone

Appendix D Scenario Outline Form ES-D-1 Facility: Nine Mile Point Unit 2 Scenario No.: NRC-1 Op-Test No.: LC2 20-1 Examiners: ____________________________ Operators:

Initial Conditions: The plant is operating at approximately 4.5% power with power ascension in progress. RHR B is in suppression pool cooling due to a completed RCIC run that raised suppression pool temperature Turnover: 1.

Reactor power is approximately 4.5%.

2.

The crew will transfer reboiler steam supply to main steam and recommence the startup to raise power to 8%.

Critical Tasks: See page 2 Event No.

Malf.

No.

Event Type*

Event Description 1

N/A N-BOP, SRO Transfer reboiler steam supply to main steam.

N2-OP-25, Sect. F.5.0 2

N/A R-ATC SRO Recommence the reactor startup and raise power using control rods per N2-OP-101A.

N2-OP-101A 3

NM07 I-ATC, SRO IRM failure downscale while raising power.

ARP's, N2-OP-92 4

RD08 C-ATC, SRO TS-SRO Control Rod Overtravel.

ARP's, N2-OP-30, T.S. 3.1.3 5

RH13A, RH15 (ILT 16-1 Scenario #1 Event 6)

C-BOP, SRO TS-SRO Inadvertent Initiation of ECCS w/ Min Flow Failure.

ARP's, N2-OP-100A, T.S. 3.5.1, 3.6.1.6, 3.6.2.3, 3.6.2.4 6

Remote RP04 C-BOP SRO Loss of RPM-MG1A.

N2-SOP-97 7

RR:PA:MT:I RR:PB:MT:I MS12 M-All A spurious trip of both RCS pumps and an isolation of the MSIVs occurs. The reactor will fail to automatically scram on MSIV position or RPV high pressure and will fail to scram when the RPS pushbuttons are armed and depressed. Scram using RRCS.

N2-EOP-RPV 8

OVR-Dis C-All One SRV 'C' solenoid fuse blows.

N2-EOP-RPV 9

CW08 C-All Isolable SWP break occurs in the 'B' RHR heat exchanger room that results in one area above the maximum safe value.

N2-EOP-SC (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

Facility: Nine Mile Point Unit 2 Scenario No.: NRC-1 Op-Test No.: LC2 20-1

1. Malfunctions after EOP entry (1-2)

Event 8, 9 2

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)

N2-EOP-RPV, N2-EOP-SC 2

5. Entry into a contingency EOP with substantive actions (>1 per scenario set) 0
6. Pre-identified Critical Tasks (> 2) 2 CRITICAL TASK DESCRIPTIONS:

CRITICAL TASK JUSTIFICATION:

CT-1.0, Given the plant with a failure of RPS to trip, the crew will manually scram the reactor using the RRCS manual initiation push buttons within 10 minutes of the indications that RPS failed to trip in accordance with N2-SOP-101C.

Critical Task 1.0 is identified as critical because with a failure of an automatic RPS scram to occur due to RPV high pressure, the Reactor must be scrammed. This malfunction represents a failure of an automatic actuation of an ESF system. Manually performing the action reduces the rate of energy production and thus the heat input into the Secondary Containment.

CT-2.0, Given an isolable service water leak in the RHR B Pump room that threatens ECCS capacity, the crew will isolate the leak within 15 minutes of the indications of the pipe break in accordance with N2-EOP-SC.

Critical Task 2.0 is identified as critical because failure to isolate the SWP leak into the RHR B pump room could cause damage to and loss of RHR B resulting in degraded emergency core cooling system (ECCS) capacity.

SCENARIO

SUMMARY

The scenario begins at 4.5% reactor power with a reactor startup in progress and no equipment out of service. RHR B is in suppression pool cooling due to a completed RCIC run that raised suppression pool temperature.

Event 1 is the normal evolution performed by the BOP operator to transfer Reboiler steam supply to main steam.

Event 2 is a reactivity evolution. The ATC operator will recommence the reactor startup by raising power using control rods per N2-OP-101A.

Event 3 occurs during the control rod withdrawal, when an IRM fails downscale. The crew will respond per the ARPs and bypass the IRM.

Event 4 occurs when control rod 26-43 is fully withdrawn per the control rod sequence to the full out position and overtravels. The SRO will declare the control rod inoperable and evaluate T.S.

3.1.3. The crew will perform the actions of the ARP and N2-OP-30 to recouple and recover the control rod to be back in accordance with the approved control rod sequence.

Event 5 occurs after the control rod has been recoupled, when an inadvertent Division I ECCS signal is received. This causes the CSL and RHR 'A' pumps to automatically start and run on minimum flow. During the transient 2RHS*MOV4A (2RHS*P1A minimum flow valve) fails closed. The crew will evaluate using redundant and independent indications that the ECCS signal is not valid and determine that 2RHS*P1A is running at shutoff head. The crew will then place 2RHS*P1A in P-T-L. The crew will evaluate Technical Specifications for the inoperability of two ECCS injection systems.

Event 6 occurs when 2RPS-MG1 spuriously trips off. A silent half scram occurs on RPS 'A' requiring the crew to enter and execute the actions of N2-SOP-97. The crew will dispatch an Equipment Operator to verify the condition of 2RPM-MG1. The crew will then reposition the power source selector switch to the alternate 'A' position and direct the field operator to reset the EPAs. With the EPAs reset, the crew will recognize that the silent half scram has been reset and exit N2-SOP-97.

Event 7 & 8 start when a spurious trip of both reactor recirculation pumps and an isolation of the MSIVs occurs. The reactor will fail to automatically scram on RPV high pressure and will fail to scram when RPS is manual trip is attempted using the arm and depress pushbuttons on panel 603 and the reactor mode switch. The crew will recognize the failure of the reactor to scram and manually initiate RRCS using the arm and depress pushbuttons on panel 603 (Critical Task 1.0). The SRO will classify the failure to scram as an Unusual Event. During the vessel isolation transient, SRVs will open causing heat addition to the suppression pool. The crew may start RCIC to augment RPV pressure control. The pressure transient caused by manual SRV operation will add heat to the suppression pool. The crew will use SRVs to control pressure and when the first operated SRV solenoid keylock switch is placed in open, the 'C' solenoid fuses will blow requiring the use of other SRVs for pressure control.

Event 9 occurs when an isolable SWP break occurs in the 'B' RHR heat exchanger room that results in one area above the maximum safe value. The crew will enter N2-EOP-SC and take required actions. The crew will diagnose the source of the leak and determine that it is from the service water system in the RHR 'B' heat exchanger room. The crew will also determine that N2-EOP-SC requires the system isolated (close SWP*MOV90B) (Critical Task 2.0). The

scenario concludes when RPV pressure and level are stabilized and are being controlled in the ordered band.

Appendix D Scenario Outline Form ES-D-1 Facility: Nine Mile Point Unit 2 Scenario No.: NRC-2 Op-Test No.: LC2 20-1 Examiners: ____________________________ Operators:

Initial Conditions: The scenario begins at 85% reactor power in preparation to raise load line. Reactor power has been adjusted to complete ReMA step 1. Load Line Adjustment is on hold until further direction from Reactor Engineering.

Turnover: 1.

The BOP operator to swap Reactor Recirculation Pump HPU A subloops to 1 in lead and 2 in standby in accordance with N2-OP-29.

2.

The ATC operator will recommence the load line adjustment by raising power using control rods per the approved ReMA.

Critical Tasks: See page 2 Event No.

Malf.

No.

Event Type*

Event Description 1

N/A N-BOP, SRO Swap Recirc Pump HPU subloops.

N2-OP-29 2

RD07 (ILT 16-1 Scenario #1 Event 2)

R-ATC SRO Stuck Control Rod.

N2-OP-30 3

RR08B (ILT 16-1 Scenario #3 Event 3)

I-ATC, SRO Loss of recirculation flow input to APRM #2.

ARP's, N2-OP-92 4

RC14 C-BOP, SRO TS-SRO ICS*P2 Trip w/ Indications Discharge Piping Not Full.

ARP's, N2-OP-35, T.S. 3.5.3 5

RR15A, RH16A C-All TS-SRO Failure of 2RCS*P1A Inner and Outer Seal.

N2-SOP-29.1, T.S. 3.4.1 6

RR20 M-All LOCA with one pair of Drywell to Suppression Chamber vacuum breakers failing open. Inability to stay below PSP RPV Blowdown.

N2-EOP-RPV, N2-EOP-PC, N2-EOP-C2 7

Remote RH27 Malfunction RH10B C-All RHS*MOV15A 600V Bkr Trips and RHS*MOV25B Jammed.

N2-EOP-PC 8

RH22B C-All RHS*MOV4B Fails to Auto-Close.

N2-EOP-PC (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

Facility: Nine Mile Point Unit 2 Scenario No.: NRC-2 Op-Test No.: LC2 20-1

1. Malfunctions after EOP entry (1-2)

Event 7, 8 2

2. Abnormal events (2-4)

Events 2, 3, 4, 5 4

3. Major transients (1-2)

Event 6 1

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

N2-EOP-RPV, N2-EOP-PC 2

5. Entry into a contingency EOP with substantive actions (>1 per scenario set)

N2-EOP-C2 1

6. Pre-identified Critical Tasks (> 2) 2 CRITICAL TASK DESCRIPTIONS:

CRITICAL TASK JUSTIFICATION:

CT-1.0, Given the plant with an isolable primary system leak into the drywell, the crew will manually isolate the A recirculation loop within 30 minutes of the indications of rising drywell pressure in accordance with N2-SOP-29.1.

Critical Task 1.0 is identified as critical because without operator action to trip and isolate the Recirc pump, Drywell pressure would continue to rise until the reactor automatically scrams. This also represents a degradation of a fission product barrier.

CT-2.0, Given the plant with rising primary containment pressure due to a LOCA, the crew will perform a RPV blowdown within 15 minutes of determining that PSP is exceeded with both drywell spray flow paths unavailable from the control room in accordance with N2-EOP-C2.

Critical Task 2.0 is identified as critical because without operator action there would be a continued release of energy from the RPV into the containment. The action serves to terminate, or reduce as much as possible, any continued primary containment pressure increase. The decision to perform a RPV blowdown will be made based on an evaluation of available containment spray systems and may be made prior to exceeding PSP if it is determined that the given rate of change of suppression chamber pressure will cause PSP to be exceeded before additional (outside control room) spray systems can be lined up.

SCENARIO

SUMMARY

The scenario begins at 85% reactor power with 2RDS-P1B out of service in preparation to raise load line. Reactor power has been adjusted to complete ReMA step 1. Load Line Adjustment is on hold until further direction from Reactor Engineering.

Event 1 is the normal evolution performed by the BOP operator to swap Reactor Recirculation Pump HPU A subloops to 1 in lead and 2 in standby in accordance with N2-OP-29.

Event 2 is a reactivity evolution. The ATC operator will recommence the load line adjustment by raising power using control rods per the approved ReMA. While the RO is raising power using rods a control rod will stick. The crew will take action to raise drive water pressure per N2-OP-

30. Raising drive water pressure will free the stuck rod and allow the load line adjustment to continue.

Event 3 occurs when the recirculation flow input to APRM #2 fails downscale. The crew will verify all other APRMs are reading normal and determine that a scram should not have occurred. The crew will follow up with ARP actions and local panel indications and determine that APRM #2 is required to be bypassed per N2-OP-92. The crew will also evaluate T.S.

3.3.1.1.

Event 4 occurs when RCIC keepfill pump 2ICS*P2 trips on motor overload. The crew will also receive annunciator alarms for 2ICS*P2 low discharge pressure and RCIC discharge piping not full. The crew is expected to carry out the ARP actions, close 2ICS*MOV150, and declare RCIC inoperable. The crew will be required to evaluated T.S. 3.5.3.

Event 5 occurs after 2ICS*MOV150 has been closed, when the inner and outer seals on 2RCS*P1A slowly degrade. The crew will enter N2-SOP-29.1. RCS seal pressure will exceed the danger limit established in N2-SOP-29.1 and Drywell pressure will begin to rise. The crew will then trip 2RCS*P1A and close 2RCS*P1A suction and discharge isolation valves (Critical Task 1.0) and enter N2-SOP-29. N2-SOP-29 will require the first four cram rods inserted.

Additionally, N2-SOP-29.1 will require WCS flow lowered to <450 gpm and the cleanup suction valve from recirc. 'A' closed.

Event 6, 7 & 8 start when a small LOCA causes primary containment parameters to degrade.

The LOCA will cause drywell pressure to rise relatively rapidly, forcing the crew to determine whether or not an emergency power reduction can be performed before the scram occurs. The crew should verify the validity of the event, using redundant diverse indications. Following the scram the crew will maintain reactor water level with condensate and feedwater. The drywell pressure rise will require entry into N2-EOP-PC. One pair of Drywell to Suppression Chamber Vacuum breakers will fail open causing drywell and suppression chamber pressure to rise at nearly the same rate and be at approximately the same pressure. The crew will then evaluate that RHR A, 'B' and service water spray through RHR'B' are available for containment spray systems. When containment spray is attempted with RHR 'B', the containment spray valve (RHS*MOV25B) will stick shut and will not be able to be opened. This failure of the RHR 'B' containment spray valve will render the RHR B drywell spray system unavailable. When containment spray is attempted with RHR 'A', the containment spray valve (RHS*MOV15A) power supply breaker will trip open and will not be able to be opened from the control room.

Pressure Suppression Pressure will be exceeded and an RPV blowdown is required (Critical Task 2.0). After the blowdown the crew will be notified that RHS*MOV25B can be opened in the field and Drywell sprays will be available. When RHS*MOV25B has been opened in the field and drywell spray follow is established, RHS*MOV4B will fail to auto close requiring the crew to

manually close RHS*MOV4B in order to get full drywell spray flow. The crew will continue with containment spray to reduce containment pressure.

Appendix D Scenario Outline Form ES-D-1 Facility: Nine Mile Point Unit 2 Scenario No.: NRC-4 Op-Test No.: LC2 20-1 Examiners: ____________________________ Operators:

Initial Conditions: The scenario begins at rated reactor power with 2WCS*P1B out of service for seal leakage.

Turnover: 1.

Event 1 is the normal evolution performed by the BOP operator to start 2RHS*P1C in Full Flow Test Mode Operation in accordance with N2-OP-31 Section H.12.0 and run for 5 minutes.

Critical Tasks: See page 2 Event No.

Malf.

No.

Event Type*

Event Description 1

N/A N-BOP, SRO Start RHR in full flow test mode.

N2-OP-31 2

DI-4567 (ILT 16-1 Scenario #2 Event 3)

C-ATC, SRO TS-SRO A RCS FCV fails open.

N2-SOP-08, N2-SOP-101D, T.S. 3.4.1 3

ED16 R-ATC, SRO Main Transformer Loss of Cooling.

ARP's, N2-SOP-101D 4

RC16 RC11 I-BOP, SRO TS-SRO Isolable RCIC Steam Leak with failure of Automatic Isolation.

ARP's, N2-EOP-SC, T.S. 3.5.3, 3.3.6.1 5

CW12B CW13A C-BOP, SRO IAS Mini Loop Cooling Wtr Pmp Trip, Standby Fails To Auto Start.

ARP's, N2-OP-13 6

TU02 C-ATC, SRO Rising Main Turbine Vibration.

ARP's, N2-SOP-101C 7

RD17Z MS04 RR27 M-All Low power ATWS, loss of valid level indication, RPV flooding (ATWS leg).

N2-EOP-PC, N2-EOP-C4 8

AD08 C-All One ADS Valve Nitrogen Supply Severed.

N2-EOP-C4 9

DG04A C-All EDG1 Fail to UV / LOCA Auto Start.

N2-EOP-C5 (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

Facility: Nine Mile Point Unit 2 Scenario No.: NRC-4 Op-Test No.: LC2 20-1

1. Malfunctions after EOP entry (1-2)

Event 8, 9 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)

N2-EOP-PC 1

5. Entry into a contingency EOP with substantive actions (>1 per scenario set)

N2-EOP-C5, N2-EOP-C4 1

6. Pre-identified Critical Tasks (> 2) 4 CRITICAL TASK DESCRIPTIONS:

CRITICAL TASK JUSTIFICATION:

CT-1.0, Given the plant with an isolable RCIC steam leak in the reactor building, the crew will manually isolate RCIC prior to any area temperature reaching 212°F in accordance with N2-EOP-SC.

Critical Task 1.0 is identified as critical because with no operator action the primary system will continue to discharge into the Secondary Containment. An area temperature above its isolation setpoint is an indication that steam from a primary system may be discharging into the Secondary Containment. As temperatures continue to rise, the continued operability of equipment needed to carry out EOP actions may be compromised.

CT-2.0, Given a condition where RPV Flooding is warranted with all control rods not full in, the crew will terminate and prevent all RPV injection except boron, CRD and RCIC prior to opening ADS valves and until RPV pressure lowers below the MSCP in accordance with N2-EOP-C4.

Critical Task 2.0 is identified as critical because without operator action, the manual RPV blowdown combined with an ATWS in progress would cause the uncontrolled injection of relatively cold water which would result in fuel damage.

CT-3.0, Given a condition where RPV Flooding is warranted with all control rods not full in, the crew will open 7 SRVs within 15 minutes of the indications of unknown RPV water level in accordance with N2-EOP-C4.

Critical Task 3.0 is identified as critical because without operator action, reactor pressure would remain too high to facilitate the only remaining preferred injection source to inject into the vessel.

This would prevent RPV water level from being restored and therefore prevent adequate core cooling from being assured. The intent is to get at least 7 SRVs (ADS or non-ADS) open.

CT-4.0, Given a condition where RPV Flooding is warranted with all control rods not full in, 7 SRVs open, and RPV pressure < 178 psig; the crew will slowly raise injection to flood the RPV to the main steam lines in accordance with N2-EOP-C4.

Critical Task 4.0 is identified as critical because with Reactor water level unknown, the status of core cooling is unknown. RPV flooding is required to establish conditions to cool the core.

This protects the fuel cladding integrity.

SCENARIO

SUMMARY

The scenario begins at rated reactor power with 2WCS*P1B out of service for seal leakage.

Event 1 is the normal evolution performed by the BOP operator to start 2RHS*P1C in Full Flow Test Mode Operation in accordance with N2-OP-31 Section H.12.0.

Event 2 occurs when reactor recirculation flow control valve 'A' begins to drift open. The crew will examine reactor power and MWe output and determine an unplanned power change is occurring. The crew will enter and take the actions on N2-SOP-8. N2-SOP-8 will require the crew to depress the HPU shutdown pushbutton to lock up the flow control valve and close the associated hydraulic fluid outside isolation valve. The crew will reduce reactor power to restore and maintain reactor power < 3988 MWth using either cram rods or recirculation flow. The crew will investigate the cause of the transient and evaluate required Tech. Specs.

Event 3 occurs when main transformer XM1A begins to overheat. The operators will be alerted to this condition when annunciator 852618 and corresponding computer point SPMTC01 go into alarm. The crew will dispatch a field operator to investigate local indications including a general visual inspection, cooling pump operation, cooling fan operation and local temperature readings.

The field operator will report back that some cooling fans are not running and that local temperatures are rising. The crew will determine that local temperature will exceed 110°C, requiring the crew to reduce MVAR loading in addition to reducing reactor power. When MVAR load is reduced and reactor power reduction has been performed the field operator will report that main transformer temperatures are lowering and are below 110°C.

Event 4 occurs when a steam leak in the RCIC Pump Room occurs. The crew will enter and execute the actions of N2-EOP-SC. The crew may order a Reactor Building Evacuation to protect station personnel. RCIC room temperatures will initiate an automatic isolation of RCIC; however RCIC will fail to isolate automatically, requiring the crew to recognize the failure and take actions to manually isolate the RCIC system (Critical Task 1.0). RCIC will isolate manually using the keylock Containment Isolation Valve control switches on Panel 601. The crew will then monitor Secondary Containment (RCIC room) and RCIC system parameters to verify the steam leak was successfully isolated and evaluate Tech. Specs.

Event 5 occurs when the instrument air mini loop cooling water pump trips. The standby mini loop cooling water pump will fail to auto start. The crew will perform the appropriate ARP/SOP actions and manually start the standby mini loop cooling water pump to restore cooling water to the operating IAS compressors. The crew may enter N2-SOP-19 and start the standby mini loop cooling water pump. The crew will be forced to closely monitor IAS system parameters to ensure a loss of instrument air does not occur.

Event 6 starts when a rise in Main Turbine Vibration occurs caused by foreign material in one of the bearing oil lines. The crew will be forced to monitor main turbine vibration and determine from 851140, Turbine Generator Vibration High when the threshold for tripping the main turbine has been reached. The crew will then scram the reactor and trip the turbine in accordance with N2-SOP-21.

Events 7, 8 & 9 start after the scram when the control rods only insert to position 02 with reactor power remaining at approximately 1%. The crew will enter N2-EOP-RPV and transition to N2-EOP-C5. The crew will inhibit ADS and place HPCS in PTL. A small steam leak in the drywell will cause drywell pressure to rise. The crew will enter N2-EOP-PC. The Division I diesel generator will fail to start on the LOCA (high drywell pressure) signal. RPV pressure will be

controlled automatically by EHC. Control rods will be inserted in accordance with N2-EOP-6.14 initially, until SRMs and IRMs are inserted and it is determined that the reactor is shutdown with no boron injected, then the crew may exit the power leg of N2-EOP-C5 and insert control rods per N2-SOP-101C and N2-OP-30. When average drywell temperature is above 250°F, all RPV water level instruments will experience reference leg flashing. The crew will exit N2-EOP-C5 and enter N2-EOP-C4. All RPV injection will be terminated and prevented except for boron, CRD, and RCIC (Critical Task 2.0) and then 7 ADS valves will be opened (Critical Task 3.0) to flood the RPV to the main steam lines. One ADS SRV will fail to open, requiring the crew to open one additional non-ADS SRV to achieve a total of 7 SRVs open. As RPV pressure lowers all RPV level instruments will fail upscale. When RPV pressure is < 178 psig, RPV injection will be commenced to flood the RPV to the main steam lines (Critical Task 4.0).

ES-401 1

Form ES-401-1 Facility: Nine Mile Point Unit 2 Date of Exam: April 2021 Tier Group RO K/A Category Points SRO-Only Points K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*

Total A2 G*

Total

1.

Emergency and Abnormal Plant Evolutions 1

3 3

4 N/A 4

3 N/A 3

20 3

4 7

2 1

1 1

1 2

1 7

2 1

3 Tier Totals 4

4 5

5 5

4 27 5

5 10

2.

Plant Systems 1

2 2

2 3

1 3

2 3

3 3

2 26 2

3 5

2 1

2 1

1 1

1 1

1 1

1 1

12 0

1 2

3 Tier Totals 3

4 3

4 2

4 3

4 4

4 3

38 3

5 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 2

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

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

K/A Topic(s)

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

2.2.40 Ability to apply Technical Specifications for a system.

4.7 76 295004 (APE 4) Partial or Complete Loss of DC Power / 6 X

AA2.02 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER: Extent of partial or complete loss of D.C. power 3.9 77 295006 (APE 6) Scram / 1 X

2.1.25 Ability to interpret reference materials, such as graphs, curves, tables, etc.

4.2 78 295019 (APE 19) Partial or Complete Loss of Instrument Air / 8 X

AA2.01 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR:

Instrument air system pressure 3.6 79 295024 High Drywell Pressure / 5 X

2.4.41 Knowledge of the emergency action level thresholds and classifications.

4.6 80 295028 (EPE 5) High Drywell Temperature (Mark I and Mark II only) / 5 X

EA2.05 Ability to determine and/or interpret the following as they apply to HIGH DRYWELL TEMPERATURE: Torus/suppression chamber pressure: Plant-Specific 3.8 81 295038 (EPE 15) High Offsite Radioactivity Release Rate / 9 X

2.4.30 Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.

4.1 82 295001 (APE 1) Partial or Complete Loss of Forced Core Flow Circulation / 1 & 4 X

AK3.06 Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: Core flow indication 2.9 1

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

AA1.03 Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER: Systems necessary to assure safe plant shutdown 4.4 2

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

AA2.01 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER: Cause of partial or complete loss of D.C. power 3.2 3

295005 (APE 5) Main Turbine Generator Trip /

3 X

2.4.47 Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.

4.2 4

295006 (APE 6) Scram / 1 X

AK1.03 Knowledge of the operational implications of the following concepts as they apply to SCRAM:

Reactivity control 3.7 5

295016 (APE 16) Control Room Abandonment

/ 7 X

AK2.02 Knowledge of the interrelations between CONTROL ROOM ABANDONMENT and the following: Local control stations: Plant-Specific 4.0 6

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

AK3.02 Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER: Reactor power reduction 3.3 7

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

AA1.03 Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR:

Instrument air compressor power supplies 3.0 8

295021 (APE 21) Loss of Shutdown Cooling /

4 X

AA2.05 Ability to determine and/or interpret the following as they apply to LOSS OF SHUTDOWN COOLING: Reactor vessel metal temperature 3.4 9

295023 (APE 23) Refueling Accidents / 8 X

2.4.18 Knowledge of the specific bases for EOPs. 3.3 10

ES-401 3

Form ES-401-1 295024 High Drywell Pressure / 5 X

EK1.01 Knowledge of the operational implications of the following concepts as they apply to HIGH DRYWELL PRESSURE: Drywell integrity: Plant-Specific 4.1 11 295025 (EPE 2) High Reactor Pressure / 3 X

EK2.01 Knowledge of the interrelations between HIGH REACTOR PRESSURE and the following:

RPS 4.1 12 295026 (EPE 3) Suppression Pool High Water Temperature / 5 X

EK3.05 Knowledge of the reasons for the following responses as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Reactor SCRAM 3.9 13 295028 (EPE 5) High Drywell Temperature (Mark I and Mark II only) / 5 X

EA1.03 Ability to operate and/or monitor the following as they apply to HIGH DRYWELL TEMPERATURE: Drywell cooling system 3.9 14 295030 (EPE 7) Low Suppression Pool Water Level / 5 X

EA2.02 Ability to determine and/or interpret the following as they apply to LOW SUPPRESSION POOL WATER LEVEL: Suppression pool temperature 3.9 15 295031 (EPE 8) Reactor Low Water Level / 2 X

2.1.20 Ability to interpret and execute procedure steps.

4.6 16 295037 (EPE 14) Scram Condition Present and Reactor Power Above APRM Downscale or Unknown / 1 X

EK1.06 Knowledge of the operational implications of the following concepts as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN:

Cooldown effects on reactor power 4.0 17 295038 (EPE 15) High Offsite Radioactivity Release Rate / 9 X

EK2.05 Knowledge of the interrelations between HIGH OFF-SITE RELEASE RATE and the following: Site emergency plan 3.7 18 600000 (APE 24) Plant Fire On Site / 8 X

AK3.04 Knowledge of the reasons for the following responses as they apply to PLANT FIRE ON SITE:

Actions contained in the abnormal procedure for plant fire on site 2.8 19 700000 (APE 25) Generator Voltage and Electric Grid Disturbances / 6 X

AA1.05 Ability to operate and/or monitor the following as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES: Engineered safety features 3.9 20 K/A Category Totals:

3 3

4 4

3/3 3/4 Group Point Total:

20/7

ES-401 4

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

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

K/A Topic(s)

IR 295017 (APE 17) Abnormal Offsite Release Rate / 9 X

AA2.03 Ability to determine and/or interpret the following as they apply to HIGH OFF-SITE RELEASE RATE: Radiation levels: Plant-Specific 3.9 83 295033 (EPE 10) High Secondary Containment Area Radiation Levels / 9 X

2.4.6 Knowledge of EOP mitigation strategies.

4.7 84 295036 (EPE 13) Secondary Containment High Sump/Area Water Level / 5 X

EA2.02 Ability to determine and/or interpret the following as they apply to SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL: Water level in the affected area 3.1 85 295002 (APE 2) Loss of Main Condenser Vacuum / 3 X

AA2.04 Ability to determine and/or interpret the following as they apply to LOSS OF MAIN CONDENSER VACUUM: Off-gas system flow 2.8 21 295008 (APE 8) High Reactor Water Level / 2 X

AK1.02 Knowledge of the operational implications of the following concepts as they apply to HIGH REACTOR WATER LEVEL:

Component erosion/damage 2.8 22 295010 (APE 10) High Drywell Pressure / 5 X

AK2.01 Knowledge of the interrelations between HIGH DRYWELL PRESSURE and the following:

Suppression pool level 3.2 23 295032 High Secondary Containment Area Temperature X

EK3.01 Knowledge of the reasons for the following responses as they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURE: Emergency/normal depressurization 3.5 24 295014 (APE 14) Inadvertent Reactivity Addition / 1 X

AA1.06 Ability to operate and/or monitor the following as they apply to INADVERTENT REACTIVITY ADDITION: Reactor/turbine pressure regulating system 3.3 25 295020 (APE 20) Inadvertent Containment Isolation / 5 & 7 X

AA2.03 Ability to determine and/or interpret the following as they apply to INADVERTENT CONTAINMENT ISOLATION: Reactor power 3.7 26 295022 (APE 22) Loss of Control Rod Drive Pumps / 1 X

2.2.44 Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

4.2 27 K/A Category Point Totals:

1 1

1 1

2/2 1/1 Group Point Total:

7/3

ES-401 5

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

System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*

K/A Topic(s)

IR 209002 (SF2, SF4 HPCS)

High-Pressure Core Spray X

2.2.44 Ability to interpret control room indications to verify the status and operation of a system and understand how operator actions and directives affect plant and system conditions.

4.4 86 212000 (SF7 RPS) Reactor Protection X

A2.08 Ability to (a) predict the impacts of the following on the REACTOR PROTECTION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Low reactor level 4.2 87 259002 (SF2 RWLCS) Reactor Water Level Control X

2.2.37 Ability to determine operability and/or availability of safety related equipment.

4.6 88 262002 (SF6 UPS) Uninterruptable Power Supply (AC/DC)

X A2.01 Ability to (a) predict the impacts of the following on the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Under voltage 2.8 89 400000 (SF8 CCS) Component Cooling Water X

2.1.27 Knowledge of system purpose and/or function.

4.0 90 203000 (SF2, SF4 RHR/LPCI)

RHR/LPCI: Injection Mode X

K1.11 Knowledge of the physical connections and/or cause effect relationships between RHR/LPCI: INJECTION MODE (PLANT SPECIFIC) and the following: Nuclear boiler instrumentation 3.7 28 205000 (SF4 SCS) Shutdown Cooling X

K2.01 Knowledge of electrical power supplies to the following: Pump motors 3.1 29 205000 (SF4 SCS) Shutdown Cooling X

K3.01 Knowledge of the effect that a loss or malfunction of the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) will have on following: Reactor pressure 3.3 30 209001 (SF2, SF4 LPCS)

Low-Pressure Core Spray X

K4.02 Knowledge of LOW-PRESSURE CORE SPRAY SYSTEM design feature(s) and/or interlocks which provide for the following: Prevents water hammer 3.0 31 209002 (SF2, SF4 HPCS)

High-Pressure Core Spray X

K5.01 Knowledge of the operational implications of the following concepts as they apply to HIGH PRESSURE CORE SPRAY SYSTEM (HPCS):

Adequate core cooling: BWR-5,6 3.8 32 211000 (SF1 SLCS) Standby Liquid Control X

K6.03 Knowledge of the effect that a loss or malfunction of the following will have on the STANDBY LIQUID CONTROL SYSTEM: A.C.

power 3.2 33 211000 (SF1 SLCS) Standby Liquid Control X

A1.04 Ability to predict and/or monitor changes in parameters associated with operating the STANDBY LIQUID CONTROL SYSTEM controls including: Valve operations 3.6 34 212000 (SF7 RPS) Reactor Protection X

A2.15 Ability to (a) predict the impacts of the following on the REACTOR PROTECTION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Load rejection 3.7 35 215003 (SF7 IRM)

Intermediate-Range Monitor X

A3.01 Ability to monitor automatic operations of the INTERMEDIATE RANGE MONITOR (IRM)

SYSTEM including: Meters and recorders 3.3 36 215004 (SF7 SRMS) Source-Range Monitor X

A4.04 Ability to manually operate and/or monitor in the control room: SRM drive control switches 3.2 37

ES-401 6

Form ES-401-1 215004 (SF7 SRMS) Source-Range Monitor X

2.1.32 Ability to explain and apply system limits and precautions.

3.8 38 215005 (SF7 PRMS) Average Power Range Monitor/Local Power Range Monitor X

K1.12 Knowledge of the physical connections and/or cause effect relationships between AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM and the following: Full core display 3.2 39 217000 (SF2, SF4 RCIC) Reactor Core Isolation Cooling X

K2.01 Knowledge of electrical power supplies to the following: Motor operated valves 2.8 40 217000 (SF2, SF4 RCIC) Reactor Core Isolation Cooling X

K3.03 Knowledge of the effect that a loss or malfunction of the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) will have on following:

Decay heat removal 3.5 41 218000 (SF3 ADS) Automatic Depressurization X

K4.01 Knowledge of AUTOMATIC DEPRESSURIZATION SYSTEM design feature(s) and/or interlocks which provide for the following:

Prevent inadvertent initiation of ADS logic 3.7 42 223002 (SF5 PCIS) Primary Containment Isolation/Nuclear Steam Supply Shutoff X

K4.06 Knowledge of PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF design feature(s) and/or interlocks which provide for the following: Once initiated, system reset requires deliberate operator action 3.4 43 239002 (SF3 SRV) Safety Relief Valves X

K6.02 Knowledge of the effect that a loss or malfunction of the following will have on the RELIEF/SAFETY VALVES: Air (Nitrogen) supply:

Plant-Specific 3.4 44 259002 (SF2 RWLCS) Reactor Water Level Control X

A1.04 Ability to predict and/or monitor changes in parameters associated with operating the REACTOR WATER LEVEL CONTROL SYSTEM controls including: Reactor water level control controller indications 3.6 45 261000 (SF9 SGTS) Standby Gas Treatment X

A2.01 Ability to (a) predict the impacts of the following on the STANDBY GAS TREATMENT SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Low system flow 2.9 46 262001 (SF6 AC) AC Electrical Distribution X

A3.01 Ability to monitor automatic operations of the A.C. ELECTRICAL DISTRIBUTION including:

Breaker tripping 3.1 47 262001 (SF6 AC) AC Electrical Distribution X

A4.04 Ability to manually operate and/or monitor in the control room: Synchronizing and paralleling of different A.C. supplies 3.6 48 262002 (SF6 UPS) Uninterruptable Power Supply (AC/DC)

X K6.02 Knowledge of the effect that a loss or malfunction of the following will have on the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.): D.C. electrical power 2.8 49 263000 (SF6 DC) DC Electrical Distribution X

A2.01 Ability to (a) predict the impacts of the following on the D.C. ELECTRICAL DISTRIBUTION; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Grounds 2.8 50 264000 (SF6 EGE) Emergency Generators (Diesel/Jet) EDG X

A3.02 Ability to monitor automatic operations of the EMERGENCY GENERATORS (DIESEL/JET) including: Minimum time for load pick up 3.1 51 300000 (SF8 IA) Instrument Air X

A4.01 Ability to manually operate and / or monitor in the control room: Pressure gauges 2.6 52 400000 (SF8 CCS) Component Cooling Water X

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.

3.8 53 K/A Category Point Totals:

2 2

2 3

1 3

2 3/2 3 3

2/3 Group Point Total:

26/5

ES-401 7

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

System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*

K/A Topic(s)

IR 223001 Primary Containment System and Auxiliaries X

2.2.38 Knowledge of conditions and limitations in the facility license.

4.5 91 234000 (SF8 FH) Fuel Handling Equipment X

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

Interlock failure 3.7 92 288000 (SF9 PVS) Plant Ventilation X

2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation.

4.4 93 201001 (SF1 CRDH) CRD Hydraulic X

K2.03 Knowledge of electrical power supplies to the following: Backup SCRAM valve solenoids 3.5 54 201003 (SF1 CRDM) Control Rod and Drive Mechanism X

K3.03 Knowledge of the effect that a loss or malfunction of the CONTROL ROD AND DRIVE MECHANISM will have on following: Shutdown margin 3.2 55 202002 (SF1 RSCTL) Recirculation Flow Control X

K4.07 Knowledge of RECIRCULATION FLOW CONTROL SYSTEM design feature(s) and/or interlocks which provide for the following: Minimum and maximum pump speed setpoints 2.9 56 204000 (SF2 RWCU) Reactor Water Cleanup X

K5.04 Knowledge of the operational implications of the following concepts as they apply to REACTOR WATER CLEANUP SYSTEM: Heat exchanger operation 2.7 57 214000 (SF7 RPIS) Rod Position Information X

K6.02 Knowledge of the effect that a loss or malfunction of the following will have on the ROD POSITION INFORMATION SYSTEM: Position indication probe 2.7 58 215002 (SF7 RBMS) Rod Block Monitor X

A1.01 Ability to predict and/or monitor changes in parameters associated with operating the ROD BLOCK MONITOR SYSTEM controls including: Trip reference: BWR-3,4,5 2.7 59 219000 (SF5 RHR SPC) RHR/LPCI:

Torus/Suppression Pool Cooling Mode X

A2.04 Ability to (a) predict the impacts of the following on the RHR/LPCI:

TORUS/SUPPRESSION POOL COOLING MODE; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Valve openings 3.1 60 226001 (SF5 RHR CSS) RHR/LPCI: Containment Spray Mode X

A3.01 Ability to monitor automatic operations of the RHR/LPCI:

CONTAINMENT SPRAY SYSTEM MODE including: Valve operation 3.0 61 230000 (SF5 RHR SPS) RHR/LPCI:

Torus/Suppression Pool Spray Mode X

A4.07 Ability to manually operate and/or monitor in the control room: System flow 3.6 62 245000 (SF4 MTGEN) Main Turbine Generator/Auxiliary X

2.1.27 Knowledge of system purpose and/or function.

3.9 63 259001 (SF2 FWS) Feedwater X

K1.05 Knowledge of the physical connections and/or cause effect relationships between REACTOR FEEDWATER SYSTEM and the following:

Condensate system 3.2 64 272000 (SF7, SF9 RMS) Radiation Monitoring X

K2.03 Knowledge of electrical power supplies to the following: Stack gas radiation monitoring system 2.5 65

ES-401 8

Form ES-401-1 XK/A Category Point Totals:

1 2

1 1

1 1

1 1/1 1 1

1/2 Group Point Total:

12/3

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

Form ES-401-3 Facility: Nine Mile Point Unit 2 Date of Exam: April 2021 Category K/A #

Topic RO SRO-only IR IR

1. Conduct of Operations 2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

4.7 94 2.1.42 Knowledge of new and spent fuel movement procedures.

3.4 95 2.1.45 Ability to identify and interpret diverse indications to validate the response of another indication.

4.3 66 2.1.4 Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, no-solo operation, maintenance of active license status, 10CFR55, etc.

3.3 67 2.1.1 Knowledge of conduct of operations requirements.

3.8 68 Subtotal 3

2

2. Equipment Control 2.2.43 Knowledge of the process used to track inoperable alarms.

3.3 96 2.2.36 Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

4.2 97 2.2.38 Knowledge of conditions and limitations in the facility license.

3.6 69 2.2.42 Ability to recognize system parameters that are entry-level conditions for Technical Specifications.

3.9 70 Subtotal 2

2

3. Radiation Control 2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions.

3.7 98 2.3.14 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

3.4 71 2.3.15 Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

2.9 72 Subtotal 2

1

4. Emergency Procedures/

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

4.6 99 2.4.37 Knowledge of the lines of authority during implementation of the emergency plan.

4.1 100 2.4.11 Knowledge of abnormal condition procedures.

4.0 73 2.4.43 Knowledge of emergency communications systems and techniques.

3.2 74 2.4.22 Knowledge of the bases for prioritizing safety functions during abnormal/emergency operations.

3.6 75 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 2 K/A suppression list.

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

1 / 1 Question 7 295018 Partial or Complete Loss of CCW AK3.07 - Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER: Cross-connecting with backup systems Resampled to limit overlap with last two NRC exams.

Randomly reselected K/A 295018 Partial or Complete Loss of CCW AK3.02 - Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER: Reactor power reduction 1 / 2 Question 24 295012 High Drywell Temperature AK3.01 - Knowledge of the reasons for the following responses as they apply to HIGH DRYWELL TEMPERATURE:

Increased drywell cooling An acceptable question could not be developed for the randomly sampled K/A at a high enough level of difficulty.

Randomly reselected K/A 295032 High Secondary Containment Area Temperature EK3.01 - Knowledge of the reasons for the following responses as they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURE: Emergency/normal depressurization 1 / 2 Question 27 295022 Loss of Control Rod Drive Pumps 2.2.36 - Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

The combination of system and generic K/A is a poor match for an RO question and the generic K/A is also used on Question 97.

Randomly reselected K/A 295022 Loss of Control Rod Drive Pumps 2.2.44 - Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

ES-401 Record of Rejected K/As Form ES-401-4 2 / 1 Question 30 205000 Shutdown Cooling K3.03 - Knowledge of the effect that a loss or malfunction of the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) will have on following: Reactor temperatures (moderator, vessel, flange)

Resampled to limit overlap with Question 9 Randomly reselected K/A 205000 Shutdown Cooling K3.01 - Knowledge of the effect that a loss or malfunction of the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) will have on following:

Reactor pressure 2 / 1 Question 47 262001 AC Electrical Distribution A3.03 - Ability to monitor automatic operations of the A.C. ELECTRICAL DISTRIBUTION including:

Load shedding Resampled to limit overlap with Question 20.

Randomly reselected K/A 262001 AC Electrical Distribution A3.03 - Ability to monitor automatic operations of the A.C. ELECTRICAL DISTRIBUTION including: Breaker tripping 2 / 2 Question 60 219000 RHR/LPCI:

Torus/Suppression Pool Cooling Mode A2.02 - Ability to (a) predict the impacts of the following on the RHR/LPCI:

TORUS/SUPPRESSION POOL COOLING MODE; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Pump trips Due to plant design, an acceptable question could not be developed for the randomly sampled K/A at a high enough level of difficulty.

Randomly reselected K/A 219000 RHR/LPCI:

Torus/Suppression Pool Cooling Mode A2.04 -

Ability to (a) predict the impacts of the following on the RHR/LPCI: TORUS/SUPPRESSION POOL COOLING MODE; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Valve openings

ES-401 Record of Rejected K/As Form ES-401-4 2 / 2 Question 62 230000 RHR/LPCI:

Torus/Suppression Pool Spray Mode A4.03 - Ability to manually operate and/or monitor in the control room: Keep fill system Resampled to limit overlap with Question 31.

Randomly reselected K/A 230000 RHR/LPCI:

Torus/Suppression Pool Spray Mode A4.07 -

Ability to manually operate and/or monitor in the control room: System flow 2 / 2 Question 64 259001 Feedwater K1.14 - Knowledge of the physical connections and/or cause effect relationships between REACTOR FEEDWATER SYSTEM and the following:

RCIC: Plant-Specific An acceptable question could not be developed for the randomly sampled K/A due to limited connection between Feedwater and RCIC at this facility.

Randomly reselected K/A 259001 Feedwater K1.14 - Knowledge of the physical connections and/or cause effect relationships between REACTOR FEEDWATER SYSTEM and the following: Condensate system 1 / 1 Question 82 295038 High Offsite Radioactivity Release Rate 2.4.3 - Ability to identify post-accident instrumentation.

Technical Specifications do not identify any release rate instrumentation as post-accident monitoring instrumentation.

Randomly reselected K/A 295038 High Offsite Radioactivity Release Rate 2.4.30 - Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.

1 / 2 Question 83 295017 High Off-site Release Rate AA2.02 - Ability to determine and/or interpret the following as they apply to HIGH OFF-SITE RELEASE RATE: Total number of curies released:

Plant-Specific The randomly sampled K/A is of low operational relevance.

Randomly reselected K/A 295017 High Off-site Release Rate AA2.03 - Ability to determine and/or interpret the following as they apply to HIGH OFF-SITE RELEASE RATE: Radiation levels: Plant-Specific 1 / 2 Question 84 There are no immediate operator actions in the associated emergency operating procedure.

ES-401 Record of Rejected K/As Form ES-401-4 295033 High Secondary Containment Area Radiation Levels 2.4.49 - Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.

Randomly reselected K/A 295033 High Secondary Containment Area Radiation Levels 2.4.6 - Knowledge of EOP mitigation strategies.

2 / 1 Question 88 259002 Reactor Water Level Control 2.4.2 - Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions.

An acceptable SRO level question could not be developed for the randomly sampled combination of system and generic K/A.

Randomly reselected K/A 259002 Reactor Water Level Control 2.2.37 - Ability to determine operability and/or availability of safety related equipment.

2 / 2 Question 91 216000 Nuclear Boiler Instrumentation 2.4.31 - Knowledge of annunciator alarms, indications, or response procedures.

An acceptable SRO level question could not be developed for the randomly sampled combination of system and generic K/A.

Randomly reselected K/A 216000 Nuclear Boiler Instrumentation 2.2.38 - Knowledge of conditions and limitations in the facility license.

2 / 2 Question 92 234000 Fuel Handling Equipment A2.03 - Ability to (a) predict the impacts of the following on the FUEL HANDLING EQUIPMENT; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of electrical power An acceptable question could not be developed for the randomly sampled K/A at the SRO level and a high enough level of difficulty.

Randomly reselected K/A 234000 Fuel Handling Equipment A2.01 - Ability to (a) predict the impacts of the following on the FUEL HANDLING EQUIPMENT; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Interlock failure

ES-401 Record of Rejected K/As Form ES-401-4 2 / 2 Question 93 288000 Plant Ventilation 2.4.34 - Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects.

An acceptable question could not be developed for the randomly sampled K/A due to lack of RO tasks performed outside the main control room for the given system.

Randomly reselected K/A 288000 Plant Ventilation 2.1.23 - Ability to perform specific system and integrated plant procedures during all modes of plant operation.

3 Question 100 2.4.2 - Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions.

An acceptable SRO level question could not be developed for the randomly sampled K/A.

Randomly reselected K/A 2.4.37 - Knowledge of the lines of authority during implementation of the emergency plan.

2 / 2 Question 91 216000 Nuclear Boiler Instrumentation 2.2.38 - Knowledge of conditions and limitations in the facility license.

An acceptable SRO level question could not be developed for the randomly sampled combination of system and generic K/A.

Randomly reselected K/A 223001 Primary Containment System and Auxiliaries 2.2.38 -

Knowledge of conditions and limitations in the facility license.

3 Question 98 2.3.6 - Ability to approve release permits.

An acceptable SRO level question could not be developed for the randomly sampled generic K/A because the facility does not typically perform releases.

Randomly reselected K/A 2.3.4 - Knowledge of radiation exposure limits under normal or emergency conditions.

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