ML21131A189

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Draft Written Examination and Operating Test Outlines (Folder 2)
ML21131A189
Person / Time
Site: Nine Mile Point Constellation icon.png
Issue date: 05/11/2021
From: Alexander M
Exelon Nuclear Generation Corp
To: Brian Fuller
Operations Branch I
Fuller B
Shared Package
ML20233A469 List:
References
EPID L-2021-OLL-0027
Download: ML21131A189 (34)
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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 Describe activity to be performed Code*

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

Perform APRM Gain Adjustment D, S Conduct of Operations (NRC 16-1)

N2-OSP-NMS-@004, K/A 2.1.31 (4.6)

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

Radiation Control Fire Fighting Response for a Fire in the Protected Area Emergency Plan D, S 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 Describe activity to be performed Code*

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

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

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

Offsite Dose Calculation Manual (ODCM)

P, R Assessment for Inoperable Equipment Radiation Control (NRC 16-

1) N2-OP-42, ODCM, K/A 2.3.15 (3.1)

Emergency Plan Classification (Alert, EAL CA5)

Emergency Plan N, R 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 for Pressure Control (Alternate Path)

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

b. Override the Control Room Envelope ACU Cross-Divisional Operating Interlock D, S 9 K/A 290003 A3.01 (3.3/3.5) N2-OP-53A
c. Venting the RPV to the Condenser M,L,S 3

K/A 239001 A4.09 (3.9/3.9) N2-EOP-6.18 (LC2-NRC16-1)

d. HPCS Pump Run Following Maintenance (Alternate Path) A,D,P,S 4

K/A 209002 A4.01 (3.7/3.7) N2-OP-33 (LC2-NRC16-1)

e. Rotate Drywell Unit Coolers (Alternate Path)

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

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

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

h. Secure 2SWP*P1A N,S 8 K/A 400000 A4.01 (3.1 / 3.0) N2-OP-11, N2-OP-58 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 D,E,R 1 K/A 295037 EA1.10 (3.7/3.9) N2-EOP-6.15
j. Reduce Lighting Loads During Station Blackout (2LAC-PNLU02 only)

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

k. Align Fire Water System to Inject through RHR B (Alternate Path) A,N,E,R 2 K/A 295031 EA1.08 (3.8/3.9) N2-EOP-6.6
  • 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 4-6 / 4-6 / 2-3 (C)ontrol room (D)irect from bank 9/8/4 (E)mergency or abnormal in-plant 1/1/1 (EN)gineered safety feature 1 / 1 / 1 (control room system)

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

(R)CA 1/1/1 (S)imulator Pairings:

a alone b alone c alone d and e f alone g alone 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 Malf. Event Event No. No. Type* Description N/A N-BOP, Transfer reboiler steam supply to main steam.

1 SRO N2-OP-25, Sect. F.5.0 N/A R-ATC Recommence the reactor startup and raise power using control SRO rods per N2-OP-101A.

2 N2-OP-101A NM07 I-ATC, IRM failure downscale while raising power.

3 SRO ARP's, N2-OP-92 RD08 C-ATC, Control Rod Overtravel.

4 SRO TS-SRO ARP's, N2-OP-30, T.S. 3.1.3 RH13A, C-BOP, Inadvertent Initiation of ECCS w/ Min Flow Failure.

RH15 SRO 5 (ILT 16-1 TS-SRO Scenario #1 Event 6) ARP's, N2-OP-100A, T.S. 3.5.1, 3.6.1.6, 3.6.2.3, 3.6.2.4 Remote C-BOP Loss of RPM-MG1A.

6 RP04 SRO N2-SOP-97 RR:PA:MT:I M-All A spurious trip of both RCS pumps and an isolation of the MSIVs RR:PB:MT:I occurs. The reactor will fail to automatically scram on MSIV MS12 position or RPV high pressure and will fail to scram when the RPS 7 pushbuttons are armed and depressed. Scram using RRCS.

N2-EOP-RPV OVR- C-All MSS*PSV128 'C' solenoid fuses blow.

13F21DI2546 8 13F22DI2547 N2-EOP-RPV CW08 C-All Isolable SWP break occurs in the 'B' RHR heat exchanger room that results in one area above the maximum safe value.

9 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) 2 Event 8, 9
2. Abnormal events (2-4) 4 Events 3, 4, 5, 6
3. Major transients (1-2) 1 Event 7
4. EOPs entered/requiring substantive actions (1-2) 2 N2-EOP-RPV, N2-EOP-SC
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 Critical Task 1.0 is identified as critical scram the reactor using the RRCS manual initiation push buttons within 10 because with a failure of an automatic minutes of the indications that RPS failed to trip in accordance with N2-SOP- RPS scram to occur due to a full MSIV 101C. isolation, 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 Critical Task 2.0 is identified as critical threatens ECCS capacity, the crew will isolate the leak within 15 minutes of because failure to isolate the SWP leak the indications of the pipe break in accordance with N2-EOP-SC. 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 18-35 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 MSIV position or 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 2MSS*PSV128 'C' solenoid keylock switch is placed in open, '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 Malf. Event Event No. No. Type* Description N/A N-BOP, Swap Recirc Pump HPU subloops.

1 SRO N2-OP-29 RD07 R-ATC Stuck Control Rod.

2 (ILT 16-1 Scenario #1 SRO Event 2) N2-OP-30 RR08B I-ATC, Loss of recirculation flow input to APRM #2.

3 (ILT 16-1 SRO Scenario #3 Event 3) ARP's, N2-OP-92 RC14 C-BOP, ICS*P2 Trip w/ Indications Discharge Piping Not Full.

4 SRO TS-SRO ARP's, N2-OP-35, T.S. 3.5.3 RR15A, C-All Failure of RCS-P1A Inner and Outer Seal.

5 RH16A TS-SRO N2-SOP-29.1, T.S. 3.4.1 RR20 M-All LOCA with one pair of Drywell to Suppression Chamber vacuum breakers failing open. Inability to stay below PSP RPV Blowdown.

6 N2-EOP-RPV, N2-EOP-PC, N2-EOP-C2 Remote C-All RHS*MOV15A 600V Bkr Trips and RHS*MOV25B Jammed.

RH27 7 Malfunction RH10B N2-EOP-PC RH22B C-All RHS*MOV4B Fails to Auto-Close.

8 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) 2 Event 7, 8
2. Abnormal events (2-4) 4 Events 2, 3, 4, 5
3. Major transients (1-2) 1 Event 6
4. EOPs entered/requiring substantive actions (1-2) 2 N2-EOP-RPV, N2-EOP-PC
5. Entry into a contingency EOP with substantive actions (>1 per scenario set) 1 N2-EOP-C2
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, Critical Task 1.0 is identified as critical the crew will manually isolate the A recirculation loop within 30 minutes of because without operator action to trip the indications of rising drywell pressure in accordance with N2-SOP-29.1. 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 Critical Task 2.0 is identified as critical LOCA, the crew will perform a RPV blowdown within 15 minutes of because without operator action there determining that PSP is exceeded with both drywell spray flow paths would be a continued release of energy unavailable from the control room in accordance with N2-EOP-C2. 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-3 Op-Test No.: LC2 20-1 Examiners: ____________________________ Operators: _____________________________

Initial Conditions: The scenario begins at rated reactor power with 2RHS*MOV24A and 2TMB-P1B out of service for corrective maintenance.

Turnover: 1. The BOP operator to add water to the Suppression Pool using the High Pressure Core Spray System (CSH) per normal operating procedure N2-OP-33, High Pressure Core Spray.

Critical Tasks: See page 2 Event Malf. Event Event No. No. Type* Description N/A N-BOP, Add water to the suppression pool with CSH.

1 SRO N2-OP-33 RD:PB:MT:I C-ATC, CRD Pump Trip P1B on Motor Electrical Fault with accumulator low RD06-30-51 (ILT 16-1 SRO pressure.

2 Scenario #4 TS-SRO Event 4)

N2-SOP-30, T.S. 3.1.5 OV0720 I-ATC, FWLC Feed Flow Instrument Fails Midscale.

3 SRO ARP's MS20A C-BOP, Gland seal exhaust fan TME-FN1A trip.

4 (ILT 16-1 SRO Scenario #2 Event 4) N2-OP-25 MC01 R-ATC, Loss of Main Condenser Vacuum that can be stabilized.

5 SRO ARP's, N2-SOP-09 ED02B C-BOP, Loss of Off-Site 115KV Line 6.

6 SRO TS-SRO ARP's, N2-SOP-03, T.S. 3.7.1, 3.8.1 MC01 M-All Further Main Condenser vacuum degradation occurs. Manual SL01B Reactor Scram, control rods stick at position 18. Reduced boron 7 injection due a tripped SLS pump.

RD17Z N2-EOP-C5, N2-EOP-PC RP11A I-All RRCS Failure to trip the RCS pumps.

8 RP11B N2-EOP-C5 MS12 I-All Reactor Vessel Isolation.

9 N2-EOP-C5, N2-EOP-PC

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

Facility: Nine Mile Point Unit 2 Scenario No.: NRC-3 Op-Test No.: LC2 20-1

1. Malfunctions after EOP entry (1-2) 2 Event 8, 9
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) 1 N2-EOP-PC
5. Entry into a contingency EOP with substantive actions (>1 per scenario set) 1 N2-EOP-C5
6. Pre-identified Critical Tasks (> 2) 5 CRITICAL TASK DESCRIPTIONS: CRITICAL TASK JUSTIFICATION:

CT-1.0, Given a loss of Offsite 115KV Line 6, the crew will restore flow Critical Task 1.0 is identified as critical through the service water non-essential headers within 15 minutes of event because without operator action the initiation in accordance with N2-SOP-03. service water cooled components such as Turbine Building Closed Loop Cooling and Reactor Building Closed Loop Cooling will heat up and cause a main turbine trip and reactor scram.

CT-2.0, Given a failure of the reactor to scram with power above 4% or Critical Task 2.0 is identified as critical unknown and reactor water level above 100 inches; the crew will terminate because the action of lowering level and prevent injection except for boron, CRD and RCIC prior to any places the feedwater spargers in the indications of core thermal-hydraulic instabilities in accordance with N2- steam space providing effective heating EOP-C5. of the relatively cold feedwater and eliminating the potential for high core inlet subcooling. This prevents or mitigates the consequences of any large irregular neutron flux oscillations induced by neutronic / thermal-hydraulic instabilities which would result in fuel damage.

CT-3.0, Given a failure of the reactor to scram with reactor water level Critical Task 3.0 is identified as critical intentionally lowered; the crew will inject with preferred injection sources to because after RPV water level is restore and maintain RPV level above the Minimum Steam Cooling Water intentionally lowered to at least 100" but Level (-39 inches actual) prior to RPV Blowdown requirements being met in above -39", maintenance of water level accordance with N2-EOP-C5. above -39" (the minimum steam cooling water level) assures that the core will remain adequately cooled.

SCENARIO

SUMMARY

The scenario begins at rated reactor power with 2RHS*MOV24A and 2TMB-P1B out of service for corrective maintenance.

Event 1 is the normal evolution performed by the BOP operator to add water to the Suppression Pool using the High Pressure Core Spray System (CSH) per normal operating procedure N2-OP-33, High Pressure Core Spray.

Event 2 occurs when CRD-P1B trips on motor electrical fault. The crew will enter and perform the actions of N2-SOP-30. The crew will shift 2RDS-FC107 to manual, close 2RDS-FC107 to minimum position and then start the standby CRD pump (CRD-P1A). Once CRD-P1A is running the crew will adjust 2RDS-FC107 to 63 gpm and place it back in auto. During the event, control rod accumulator 30-51 will experience a low pressure condition requiring the crew to evaluate technical specification 3.1.5 and recharge the HCU.

Event 3 occurs when the analog Feed Flow Line 'B' flow transmitter fails to about 50% of scale, forcing DFWLCS into SINGLE ELEMENT. The crew will enter N2-ARP-603100 and execute steps per alarm 603143, DFWLCS ALT CTION to place DFWLC in SINGLE ELEMENT LOCKED.

Event 4 occurs when the in-service steam packing exhauster (TME-FN1A) trips on motor electrical fault. The crew will be forced to monitor turbine steam seal indications to ensure a loss of the main turbine does not occur. The crew will follow the appropriate ARP actions and start a standby steam packing exhauster in accordance with N2-OP-25. The crew will also dispatch field operators to perform visual inspections in order to determine the cause of the event.

Event 5 occurs when main condenser air in-leakage occurs. Main condenser vacuum will lower slowly and will force the crew to enter N2-SOP-9 and N2-SOP-101D to lower power to stabilize vacuum. Vacuum differential between water boxes will remain less than 2 inches. The crew will be able to stabilize vacuum by lowering reactor power using recirculation flow and/or CRAM rods. The crew will send operators to investigate the vacuum loss and will be required to monitor main condenser vacuum as a critical parameter to ensure further degradation does not occur. Reactor power is now stabilized at 95%.

Event 6 occurs when a loss of offsite 115 KV Line 6 occurs requiring the crew to enter and execute SOP-3, Loss of AC Power. The crew will perform N2-SOP-3 actions to stabilize and restore Service Water flow through the Reactor Building and Turbine Building non-essential headers (Critical Task 1.0). Per N2-SOP-03, Attachment 1, the crew will now restore Drywell cooling, pneumatics to the drywell, and the Containment Monitoring System. The crew may also enter N2-SOP-38 for the loss of Spent Fuel Pool Cooling. The crew will contact Power Control to determine cause and the expected duration of outage. The crew will evaluate Technical Specifications actions for the line loss.

Event 7, 8 & 9 start when the Main Condenser vacuum begins to further degrade. The crew will recognize the loss of vacuum and attempt to Scram the reactor. When the RPS manual scram pushbuttons are depressed and the Mode Switch is placed in SHUTDOWN, control rods will fail to fully insert and be stuck at position 18. The crew will enter N2-EOP-RPV and transition to N2-EOP-C5. The following events will now occur:

1) RRCS will initiate SLS after the 98 second time delay but fail to trip the RCS pumps to OFF
2) When SLS is initiated, 2SLS*P1B will trip
3) Seven (7) minutes after RRCS is manually initiated, all MSIVs will inadvertently close resulting in an isolated, high power ATWS condition. N2-EOP-PC will be entered when suppression pool temperature exceeds 90°F.

The crew will inhibit ADS and prevent HPCS injection. RPV water level will be lowered by terminating and preventing all injection except for boron, CRD, and RCIC to reduce reactor power (Critical Task 2.0). The final RPV water level band should be established above -39 inches (MSCWL) using Condensate / Feedwater (Critical Task 3.0).

After the MSIVs close, RPV pressure will now be controlled using SRV keylock switches at 2CEC*PNL601. RPV pressure may be adjusted to a lower pressure band to prevent violation of HCTL. Control rods are inserted in accordance with N2-EOP-6.14. Both loops of RHR will be placed in Suppression Pool cooling to reduce suppression pool temperature.

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 Malf. Event Event No. No. Type* Description N/A N-BOP, Start RHR in full flow test mode.

1 SRO N2-OP-31 DI-4567 C-ATC, A RCS FCV fails open.

2 (ILT 16-1 SRO Scenario #2 Event 3) TS-SRO N2-SOP-08, N2-SOP-101D, T.S. 3.4.1 ED16 R-ATC, Main Transformer Loss of Cooling.

3 SRO ARP's, N2-SOP-101D RC16 I-BOP, Isolable RCIC Steam Leak with failure of Automatic Isolation.

4 RC11 SRO TS-SRO ARP's, N2-EOP-SC, T.S. 3.5.3, 3.3.6.1 CW12B C-BOP, IAS Mini Loop Cooling Wtr Pmp Trip, Standby Fails To Auto Start.

5 CW13A SRO ARP's, N2-OP-13 TU02 C-ATC, Rising Main Turbine Vibration.

6 SRO ARP's, N2-SOP-101C RD17Z M-All Low power ATWS, loss of valid level indication, RPV flooding MS04 (ATWS leg).

7 RR27 N2-EOP-PC, N2-EOP-C4 AD08 C-All One ADS Valve Nitrogen Supply Severed.

8 N2-EOP-C4 DG04A C-All EDG1 Fail to UV / LOCA Auto Start.

9 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) 2 Event 8, 9
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) 1 N2-EOP-PC
5. Entry into a contingency EOP with substantive actions (>1 per scenario set) 1 N2-EOP-C5, N2-EOP-C4
6. Pre-identified Critical Tasks (> 2) 5 CRITICAL TASK DESCRIPTIONS: CRITICAL TASK JUSTIFICATION:

CT-1.0, Given the plant with an isolable RCIC steam leak in the reactor Critical Task 1.0 is identified as critical building, the crew will manually isolate RCIC prior to any area temperature because with no operator action the reaching 212°F in accordance with N2-EOP-SC. 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 Critical Task 2.0 is identified as critical rods not full in, the crew will terminate and prevent all RPV injection except because without operator action, the boron, CRD and RCIC prior to opening ADS valves and until RPV pressure manual RPV blowdown combined with an lowers below the MSCP in accordance with N2-EOP-C4. 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 Critical Task 3.0 is identified as critical rods not full in, the crew will open 7 SRVs within 15 minutes of the because without operator action, reactor indications of unknown RPV water level in accordance with N2-EOP-C4. 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 Critical Task 4.0 is identified as critical rods not full in, 7 SRVs open, and RPV pressure < 178 psig; the crew will because with Reactor water level slowly raise injection to flood the RPV to the main steam lines in accordance unknown, the status of core cooling is with N2-EOP-C4. 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.

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. 1 3 3 4 4 3 3 20 3 4 7 Emergency and N/A N/A 2 1 1 1 1 2 1 7 2 1 3 Abnormal Plant Evolutions Tier Totals 4 4 5 5 5 4 27 5 5 10 1 2 2 2 3 1 3 2 3 3 3 2 26 2 3 5 2.

Plant 2 1 2 1 1 1 1 1 1 1 1 1 12 0 1 2 3 Systems Tier Totals 3 4 3 4 2 4 3 4 4 4 3 38 3 5 8

3. Generic Knowledge and Abilities 1 2 3 4 1 2 3 4 Categories 3 2 2 3 10 2 2 1 2 7 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 X 2.2.40 Ability to apply Technical Specifications for 4.7 76 Forced Core Flow Circulation / 1 & 4 a system.

295004 (APE 4) Partial or Complete Loss of X AA2.02 Ability to determine and/or interpret the 3.9 77 DC Power / 6 following as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER: Extent of partial or complete loss of D.C. power 295006 (APE 6) Scram / 1 X 2.1.25 Ability to interpret reference materials, such 4.2 78 as graphs, curves, tables, etc.

295019 (APE 19) Partial or Complete Loss of X AA2.01 Ability to determine and/or interpret the 3.6 79 Instrument Air / 8 following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR:

Instrument air system pressure 295024 High Drywell Pressure / 5 X 2.4.41 Knowledge of the emergency action level 4.6 80 thresholds and classifications.

295028 (EPE 5) High Drywell Temperature X EA2.05 Ability to determine and/or interpret the 3.8 81 (Mark I and Mark II only) / 5 following as they apply to HIGH DRYWELL TEMPERATURE: Torus/suppression chamber pressure: Plant-Specific 295038 (EPE 15) High Offsite Radioactivity X 2.4.30 Knowledge of events related to system 4.1 82 Release Rate / 9 operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.

295001 (APE 1) Partial or Complete Loss of X AK3.06 Knowledge of the reasons for the following 2.9 1 Forced Core Flow Circulation / 1 & 4 responses as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: Core flow indication 295003 (APE 3) Partial or Complete Loss of X AA1.03 Ability to operate and/or monitor the 4.4 2 AC Power / 6 following as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER: Systems necessary to assure safe plant shutdown 295004 (APE 4) Partial or Complete Loss of X AA2.01 Ability to determine and/or interpret the 3.2 3 DC Power / 6 following as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER: Cause of partial or complete loss of D.C. power 295005 (APE 5) Main Turbine Generator Trip / X 2.4.47 Ability to diagnose and recognize trends in 4.2 4 3 an accurate and timely manner utilizing the appropriate control room reference material.

295006 (APE 6) Scram / 1 X AK1.03 Knowledge of the operational implications 3.7 5 of the following concepts as they apply to SCRAM:

Reactivity control 295016 (APE 16) Control Room Abandonment X AK2.02 Knowledge of the interrelations between 4.0 6

/7 CONTROL ROOM ABANDONMENT and the following: Local control stations: Plant-Specific 295018 (APE 18) Partial or Complete Loss of X AK3.02 Knowledge of the reasons for the following 3.3 7 CCW / 8 responses as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER: Reactor power reduction 295019 (APE 19) Partial or Complete Loss of X AA1.03 Ability to operate and/or monitor the 3.0 8 Instrument Air / 8 following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR:

Instrument air compressor power supplies 295021 (APE 21) Loss of Shutdown Cooling / X AA2.05 Ability to determine and/or interpret the 3.4 9 4 following as they apply to LOSS OF SHUTDOWN COOLING: Reactor vessel metal temperature 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 4.1 11 of the following concepts as they apply to HIGH DRYWELL PRESSURE: Drywell integrity: Plant-Specific 295025 (EPE 2) High Reactor Pressure / 3 X EK2.01 Knowledge of the interrelations between 4.1 12 HIGH REACTOR PRESSURE and the following:

RPS 295026 (EPE 3) Suppression Pool High Water X EK3.05 Knowledge of the reasons for the following 3.9 13 Temperature / 5 responses as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Reactor SCRAM 295028 (EPE 5) High Drywell Temperature X EA1.03 Ability to operate and/or monitor the 3.9 14 (Mark I and Mark II only) / 5 following as they apply to HIGH DRYWELL TEMPERATURE: Drywell cooling system 295030 (EPE 7) Low Suppression Pool Water X EA2.02 Ability to determine and/or interpret the 3.9 15 Level / 5 following as they apply to LOW SUPPRESSION POOL WATER LEVEL: Suppression pool temperature 295031 (EPE 8) Reactor Low Water Level / 2 X 2.1.20 Ability to interpret and execute procedure 4.6 16 steps.

295037 (EPE 14) Scram Condition Present X EK1.06 Knowledge of the operational implications 4.0 17 and Reactor Power Above APRM Downscale of the following concepts as they apply to SCRAM or Unknown / 1 CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN:

Cooldown effects on reactor power 295038 (EPE 15) High Offsite Radioactivity X EK2.05 Knowledge of the interrelations between 3.7 18 Release Rate / 9 HIGH OFF-SITE RELEASE RATE and the following: Site emergency plan 600000 (APE 24) Plant Fire On Site / 8 X AK3.04 Knowledge of the reasons for the following 2.8 19 responses as they apply to PLANT FIRE ON SITE:

Actions contained in the abnormal procedure for plant fire on site 700000 (APE 25) Generator Voltage and X AA1.05 Ability to operate and/or monitor the 3.9 20 Electric Grid Disturbances / 6 following as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES: Engineered safety features 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 X AA2.03 Ability to determine and/or interpret the 3.9 83 Rate / 9 following as they apply to HIGH OFF-SITE RELEASE RATE: Radiation levels: Plant-Specific 295033 (EPE 10) High Secondary X 2.4.6 Knowledge of EOP mitigation strategies. 4.7 84 Containment Area Radiation Levels / 9 295036 (EPE 13) Secondary Containment X EA2.02 Ability to determine and/or interpret the 3.1 85 High Sump/Area Water Level / 5 following as they apply to SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL: Water level in the affected area 295002 (APE 2) Loss of Main Condenser X AA2.04 Ability to determine and/or interpret the 2.8 21 Vacuum / 3 following as they apply to LOSS OF MAIN CONDENSER VACUUM: Off-gas system flow 295008 (APE 8) High Reactor Water Level / 2 X AK1.02 Knowledge of the operational 2.8 22 implications of the following concepts as they apply to HIGH REACTOR WATER LEVEL:

Component erosion/damage 295010 (APE 10) High Drywell Pressure / 5 X AK2.01 Knowledge of the interrelations between 3.2 23 HIGH DRYWELL PRESSURE and the following:

Suppression pool level 295032 High Secondary Containment Area X EK3.01 Knowledge of the reasons for the 3.5 24 Temperature following responses as they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURE: Emergency/normal depressurization 295014 (APE 14) Inadvertent Reactivity X AA1.06 Ability to operate and/or monitor the 3.3 25 Addition / 1 following as they apply to INADVERTENT REACTIVITY ADDITION: Reactor/turbine pressure regulating system 295020 (APE 20) Inadvertent Containment X AA2.03 Ability to determine and/or interpret the 3.7 26 Isolation / 5 & 7 following as they apply to INADVERTENT CONTAINMENT ISOLATION: Reactor power 295022 (APE 22) Loss of Control Rod Drive X 2.2.44 Ability to interpret control room indications 4.2 27 Pumps / 1 to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

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) X 2.2.44 Ability to interpret control room indications 4.4 86 High-Pressure Core Spray to verify the status and operation of a system and understand how operator actions and directives affect plant and system conditions.

212000 (SF7 RPS) Reactor X A2.08 Ability to (a) predict the impacts of the 4.2 87 Protection 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 259002 (SF2 RWLCS) Reactor X 2.2.37 Ability to determine operability and/or 4.6 88 Water Level Control availability of safety related equipment.

262002 (SF6 UPS) Uninterruptable X A2.01 Ability to (a) predict the impacts of the 2.8 89 Power Supply (AC/DC) following on the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Under voltage 400000 (SF8 CCS) Component X 2.1.27 Knowledge of system purpose and/or 4.0 90 Cooling Water function.

203000 (SF2, SF4 RHR/LPCI) X K1.11 Knowledge of the physical connections 3.7 28 RHR/LPCI: Injection Mode and/or cause effect relationships between RHR/LPCI: INJECTION MODE (PLANT SPECIFIC) and the following: Nuclear boiler instrumentation 205000 (SF4 SCS) Shutdown X K2.01 Knowledge of electrical power supplies to 3.1 29 Cooling the following: Pump motors 205000 (SF4 SCS) Shutdown X K3.01 Knowledge of the effect that a loss or 3.3 30 Cooling malfunction of the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) will have on following: Reactor pressure 209001 (SF2, SF4 LPCS) X K4.02 Knowledge of LOW-PRESSURE CORE 3.0 31 Low-Pressure Core Spray SPRAY SYSTEM design feature(s) and/or interlocks which provide for the following: Prevents water hammer 209002 (SF2, SF4 HPCS) X K5.01 Knowledge of the operational implications 3.8 32 High-Pressure Core Spray of the following concepts as they apply to HIGH PRESSURE CORE SPRAY SYSTEM (HPCS):

Adequate core cooling: BWR-5,6 211000 (SF1 SLCS) Standby Liquid X K6.03 Knowledge of the effect that a loss or 3.2 33 Control malfunction of the following will have on the STANDBY LIQUID CONTROL SYSTEM: A.C.

power 211000 (SF1 SLCS) Standby Liquid X A1.04 Ability to predict and/or monitor changes in 3.6 34 Control parameters associated with operating the STANDBY LIQUID CONTROL SYSTEM controls including: Valve operations 212000 (SF7 RPS) Reactor X A2.15 Ability to (a) predict the impacts of the 3.7 35 Protection 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 215003 (SF7 IRM) X A3.01 Ability to monitor automatic operations of 3.3 36 Intermediate-Range Monitor the INTERMEDIATE RANGE MONITOR (IRM)

SYSTEM including: Meters and recorders 215004 (SF7 SRMS) Source-Range X A4.04 Ability to manually operate and/or monitor in 3.2 37 Monitor the control room: SRM drive control switches

ES-401 6 Form ES-401-1 215004 (SF7 SRMS) Source-Range X 2.1.32 Ability to explain and apply system limits 3.8 38 Monitor and precautions.

215005 (SF7 PRMS) Average Power X K1.12 Knowledge of the physical connections 3.2 39 Range Monitor/Local Power Range and/or cause effect relationships between AVERAGE POWER RANGE MONITOR/LOCAL Monitor POWER RANGE MONITOR SYSTEM and the following: Full core display 217000 (SF2, SF4 RCIC) Reactor X K2.01 Knowledge of electrical power supplies to 2.8 40 Core Isolation Cooling the following: Motor operated valves 217000 (SF2, SF4 RCIC) Reactor X K3.03 Knowledge of the effect that a loss or 3.5 41 Core Isolation Cooling malfunction of the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) will have on following:

Decay heat removal 218000 (SF3 ADS) Automatic X K4.01 Knowledge of AUTOMATIC 3.7 42 Depressurization DEPRESSURIZATION SYSTEM design feature(s) and/or interlocks which provide for the following:

Prevent inadvertent initiation of ADS logic 223002 (SF5 PCIS) Primary X K4.06 Knowledge of PRIMARY CONTAINMENT 3.4 43 Containment Isolation/Nuclear Steam ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF design feature(s) and/or Supply Shutoff interlocks which provide for the following: Once initiated, system reset requires deliberate operator action 239002 (SF3 SRV) Safety Relief X K6.02 Knowledge of the effect that a loss or 3.4 44 Valves malfunction of the following will have on the RELIEF/SAFETY VALVES: Air (Nitrogen) supply:

Plant-Specific 259002 (SF2 RWLCS) Reactor X A1.04 Ability to predict and/or monitor changes in 3.6 45 Water Level Control parameters associated with operating the REACTOR WATER LEVEL CONTROL SYSTEM controls including: Reactor water level control controller indications 261000 (SF9 SGTS) Standby Gas X A2.01 Ability to (a) predict the impacts of the 2.9 46 Treatment 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 262001 (SF6 AC) AC Electrical X A3.01 Ability to monitor automatic operations of 3.1 47 Distribution the A.C. ELECTRICAL DISTRIBUTION including:

Breaker tripping 262001 (SF6 AC) AC Electrical X A4.04 Ability to manually operate and/or monitor in 3.6 48 Distribution the control room: Synchronizing and paralleling of different A.C. supplies 262002 (SF6 UPS) Uninterruptable X K6.02 Knowledge of the effect that a loss or 2.8 49 Power Supply (AC/DC) malfunction of the following will have on the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.): D.C. electrical power 263000 (SF6 DC) DC Electrical X A2.01 Ability to (a) predict the impacts of the 2.8 50 Distribution 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 264000 (SF6 EGE) Emergency X A3.02 Ability to monitor automatic operations of 3.1 51 Generators (Diesel/Jet) EDG the EMERGENCY GENERATORS (DIESEL/JET) including: Minimum time for load pick up 300000 (SF8 IA) Instrument Air X A4.01 Ability to manually operate and / or monitor 2.6 52 in the control room: Pressure gauges 400000 (SF8 CCS) Component X 2.4.9 Knowledge of low power/shutdown 3.8 53 Cooling Water implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.

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 #

216000 (SF7 NBI) Nuclear Boiler Instrumentation X 2.2.38 Knowledge of conditions and 4.5 91 limitations in the facility license.

234000 (SF8 FH) Fuel Handling Equipment X A2.01 Ability to (a) predict the impacts of 3.7 92 the following on the FUEL HANDLING EQUIPMENT; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Interlock failure 288000 (SF9 PVS) Plant Ventilation X 2.1.23 Ability to perform specific system 4.4 93 and integrated plant procedures during all modes of plant operation.

201001 (SF1 CRDH) CRD Hydraulic X K2.03 Knowledge of electrical power 3.5 54 supplies to the following: Backup SCRAM valve solenoids 201003 (SF1 CRDM) Control Rod and Drive X K3.03 Knowledge of the effect that a loss 3.2 55 Mechanism or malfunction of the CONTROL ROD AND DRIVE MECHANISM will have on following: Shutdown margin 202002 (SF1 RSCTL) Recirculation Flow Control X K4.07 Knowledge of RECIRCULATION 2.9 56 FLOW CONTROL SYSTEM design feature(s) and/or interlocks which provide for the following: Minimum and maximum pump speed setpoints 204000 (SF2 RWCU) Reactor Water Cleanup X K5.04 Knowledge of the operational 2.7 57 implications of the following concepts as they apply to REACTOR WATER CLEANUP SYSTEM: Heat exchanger operation 214000 (SF7 RPIS) Rod Position Information X K6.02 Knowledge of the effect that a loss 2.7 58 or malfunction of the following will have on the ROD POSITION INFORMATION SYSTEM: Position indication probe 215002 (SF7 RBMS) Rod Block Monitor X A1.01 Ability to predict and/or monitor 2.7 59 changes in parameters associated with operating the ROD BLOCK MONITOR SYSTEM controls including: Trip reference: BWR-3,4,5 219000 (SF5 RHR SPC) RHR/LPCI:

X A2.04 Ability to (a) predict the impacts of 3.1 60 Torus/Suppression Pool Cooling Mode 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 226001 (SF5 RHR CSS) RHR/LPCI: Containment X A3.01 Ability to monitor automatic 3.0 61 Spray Mode operations of the RHR/LPCI:

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

X A4.07 Ability to manually operate and/or 3.6 62 Torus/Suppression Pool Spray Mode monitor in the control room: System flow 245000 (SF4 MTGEN) Main Turbine X 2.1.27 Knowledge of system purpose 3.9 63 Generator/Auxiliary and/or function.

259001 (SF2 FWS) Feedwater X K1.05 Knowledge of the physical 3.2 64 connections and/or cause effect relationships between REACTOR FEEDWATER SYSTEM and the following:

Condensate system 272000 (SF7, SF9 RMS) Radiation Monitoring X K2.03 Knowledge of electrical power 2.5 65 supplies to the following: Stack gas radiation monitoring system

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 #

2.1.7 Ability to evaluate plant performance and make operational 4.7 94 judgments based on operating characteristics, reactor behavior, and instrument interpretation.

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 4.3 66

1. Conduct of response of another indication.

Operations 2.1.4 Knowledge of individual licensed operator responsibilities 3.3 67 related to shift staffing, such as medical requirements, no-solo operation, maintenance of active license status, 10CFR55, etc.

2.1.1 Knowledge of conduct of operations requirements. 3.8 68 Subtotal 3 2 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 4.2 97 degraded power sources, on the status of limiting conditions for operations.

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 3.9 70 conditions for Technical Specifications.

2. Equipment Subtotal 2 2 Control 2.3.6 Ability to approve release permits. 3.8 98 2.3.14 Knowledge of radiation or contamination hazards that may 3.4 71 arise during normal, abnormal, or emergency conditions or activities.

2.3.15 Knowledge of radiation monitoring systems, such as fixed 2.9 72 radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

Subtotal 2 1 2.4.21 Knowledge of the parameters and logic used to assess the 4.6 99 status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

2.4.37 Knowledge of the lines of authority during implementation of 4.1 100 the emergency plan.

4. Emergency Procedures/Plan 2.4.11 Knowledge of abnormal condition procedures. 4.0 73 2.4.43 Knowledge of emergency communications systems and 3.2 74 techniques.

2.4.22 Knowledge of the bases for prioritizing safety functions during 3.6 75 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 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:

Question 7 Resampled to limit overlap with last two NRC exams.

295018 Partial or Complete Loss of CCW Randomly reselected K/A 295018 Partial or Complete Loss of CCW AK3.02 - Knowledge of AK3.07 - Knowledge of the the reasons for the following responses as they reasons for the following apply to PARTIAL OR COMPLETE LOSS OF 1/1 responses as they apply to COMPONENT COOLING WATER: Reactor PARTIAL OR COMPLETE power reduction LOSS OF COMPONENT COOLING WATER: Cross-connecting with backup systems Question 24 An acceptable question could not be developed for the randomly sampled K/A at a high enough 295012 High Drywell level of difficulty.

Temperature Randomly reselected K/A 295032 High AK3.01 - Knowledge of the Secondary Containment Area Temperature 1/2 reasons for the following EK3.01 - Knowledge of the reasons for the responses as they apply to following responses as they apply to HIGH HIGH DRYWELL SECONDARY CONTAINMENT AREA TEMPERATURE: TEMPERATURE: Emergency/normal Increased drywell cooling depressurization Question 27 The combination of system and generic K/A is a poor match for an RO question and the generic 295022 Loss of Control K/A is also used on Question 97.

Rod Drive Pumps Randomly reselected K/A 295022 Loss of Control 2.2.36 - Ability to analyze Rod Drive Pumps 2.2.44 - Ability to interpret the effect of maintenance control room indications to verify the status and activities, such as operation of a system, and understand how degraded power sources, operator actions and directives affect plant and 1/2 on the status of limiting system conditions.

conditions for operations.

ES-401 Record of Rejected K/As Form ES-401-4 Question 30 Resampled to limit overlap with Question 9 205000 Shutdown Cooling Randomly reselected K/A 205000 Shutdown Cooling K3.01 - Knowledge of the effect that a K3.03 - Knowledge of the loss or malfunction of the SHUTDOWN effect that a loss or COOLING SYSTEM (RHR SHUTDOWN malfunction of the COOLING MODE) will have on following:

2/1 SHUTDOWN COOLING Reactor pressure SYSTEM (RHR SHUTDOWN COOLING MODE) will have on following: Reactor temperatures (moderator, vessel, flange)

Question 47 Resampled to limit overlap with Question 20.

262001 AC Electrical Randomly reselected K/A 262001 AC Electrical Distribution Distribution A3.03 - Ability to monitor automatic operations of the A.C. ELECTRICAL 2/1 A3.03 - Ability to monitor DISTRIBUTION including: Breaker tripping automatic operations of the A.C. ELECTRICAL DISTRIBUTION including:

Load shedding Question 60 Due to plant design, an acceptable question could not be developed for the randomly sampled 219000 RHR/LPCI: K/A at a high enough level of difficulty.

Torus/Suppression Pool Cooling Mode Randomly reselected K/A 219000 RHR/LPCI:

Torus/Suppression Pool Cooling Mode A2.04 -

A2.02 - Ability to (a) predict Ability to (a) predict the impacts of the following the impacts of the following on the RHR/LPCI: TORUS/SUPPRESSION on the RHR/LPCI: POOL COOLING MODE; and (b) based on those TORUS/SUPPRESSION predictions, use procedures to correct, control, or POOL COOLING MODE; mitigate the consequences of those abnormal and (b) based on those conditions or operations: Valve openings 2/2 predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Pump trips

ES-401 Record of Rejected K/As Form ES-401-4 Question 62 Resampled to limit overlap with Question 31.

230000 RHR/LPCI: Randomly reselected K/A 230000 RHR/LPCI:

Torus/Suppression Pool Torus/Suppression Pool Spray Mode A4.07 -

Spray Mode Ability to manually operate and/or monitor in the 2/2 control room: System flow A4.03 - Ability to manually operate and/or monitor in the control room: Keep fill system Question 64 An acceptable question could not be developed for the randomly sampled K/A due to limited 259001 Feedwater connection between Feedwater and RCIC at this facility.

K1.14 - Knowledge of the physical connections Randomly reselected K/A 259001 Feedwater 2/2 and/or cause effect K1.14 - Knowledge of the physical connections relationships between and/or cause effect relationships between REACTOR FEEDWATER REACTOR FEEDWATER SYSTEM and the SYSTEM and the following: following: Condensate system RCIC: Plant-Specific Question 82 Technical Specifications do not identify any release rate instrumentation as post-accident 295038 High Offsite monitoring instrumentation.

Radioactivity Release Rate Randomly reselected K/A 295038 High Offsite 1/1 2.4.3 - Ability to identify Radioactivity Release Rate 2.4.30 - Knowledge post-accident of events related to system operation/status that instrumentation. must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.

Question 83 The randomly sampled K/A is of low operational relevance.

295017 High Off-site Release Rate Randomly reselected K/A 295017 High Off-site Release Rate AA2.03 - Ability to determine AA2.02 - Ability to and/or interpret the following as they apply to 1/2 determine and/or interpret HIGH OFF-SITE RELEASE RATE: Radiation the following as they apply levels: Plant-Specific to HIGH OFF-SITE RELEASE RATE: Total number of curies released:

Plant-Specific Question 84 There are no immediate operator actions in the associated emergency operating procedure.

1/2 295033 High Secondary Containment Area Randomly reselected K/A 295033 High

ES-401 Record of Rejected K/As Form ES-401-4 Radiation Levels Secondary Containment Area Radiation Levels 2.4.6 - Knowledge of EOP mitigation strategies.

2.4.49 - Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.

Question 88 An acceptable SRO level question could not be developed for the randomly sampled combination 259002 Reactor Water of system and generic K/A.

Level Control Randomly reselected K/A 259002 Reactor Water 2/1 2.4.2 - Knowledge of Level Control 2.2.37 - Ability to determine system set points, operability and/or availability of safety related interlocks and automatic equipment.

actions associated with EOP entry conditions.

Question 91 An acceptable SRO level question could not be developed for the randomly sampled combination 216000 Nuclear Boiler of system and generic K/A.

Instrumentation Randomly reselected K/A 216000 Nuclear Boiler 2/2 2.4.31 - Knowledge of Instrumentation 2.2.38 - Knowledge of conditions annunciator alarms, and limitations in the facility license.

indications, or response procedures.

Question 92 An acceptable question could not be developed for the randomly sampled K/A at the SRO level 234000 Fuel Handling and a high enough level of difficulty.

Equipment Randomly reselected K/A 234000 Fuel Handling Equipment A2.01 - Ability to (a) predict the A2.03 - Ability to (a) predict impacts of the following on the FUEL HANDLING the impacts of the following EQUIPMENT; and (b) based on those on the FUEL HANDLING predictions, use procedures to correct, control, or 2/2 EQUIPMENT; and (b) mitigate the consequences of those abnormal based on those predictions, conditions or operations: Interlock failure use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of electrical power Question 93 An acceptable question could not be developed for the randomly sampled K/A due to lack of RO 2/2 288000 Plant Ventilation tasks performed outside the main control room for the given system.

ES-401 Record of Rejected K/As Form ES-401-4 2.4.34 - Knowledge of RO tasks performed outside Randomly reselected K/A 288000 Plant the main control room Ventilation 2.1.23 - Ability to perform specific during an emergency and system and integrated plant procedures during all the resultant operational modes of plant operation.

effects.

Question 100 An acceptable SRO level question could not be developed for the randomly sampled K/A.

2.4.2 - Knowledge of system set points, Randomly reselected K/A 2.4.37 - Knowledge of 3

interlocks and automatic the lines of authority during implementation of the actions associated with emergency plan.

EOP entry conditions.

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