ML17172A720

Section 10 - Final Simulator Scenarios - North Anna Power Station

ML17172A720
Person / Time
Site:	North Anna
Issue date:	06/21/2017
From:	NRC/RGN-II
To:	Virginia Electric & Power Co (VEPCO)
References
Download: ML17172A720 (147)

Text

Appendix D Scenario Outline Form ES-D-1 Facility: North Anna Power Station Scenario No.: (2016) NRC -1 Op-Test No.: 1 Examiners: Bruno Cabarello______________ Operators: _____________________________

Newton Lacy________________ _____________________________

Gary Callaway_______________ _____________________________

Initial Conditions: 100% MOL, 1-SW-P-1A is tagged out for major repairs. 1-BC-P-1B is tagged out for shaft replacement. 2H is the protected train.

Turnover: Maintain current plant conditions. Assist maintenance with work on 1-SW-P-1A, and 1-BC-P-1B.

Event Malf. Event Event No. No. Type* Description 1 RC2002 I/C (RO) "B" Pressurizer spray valve (1-RC-PCV-1455B) fails open. Can be closed with SOV. (CT)

I/C (B) 2 SW0104 SW pump, 2-SW-P-1A, trips TS (S) 3 RD07 I/C (RO) Continuous automatic control rod insertion which can be stopped in manual (CT) 4 MS0103 I/C (B) Selected Ch III steam flow channel (1-MS-FI-1484) fails low on "B" SG (CT)

TS (S) 5 CH1601 I/C (RO) Charging pump "A" trips with failure of standby pump ("B") to auto CH2101 TS (S) start and discharge check valve on "A" pump sticks open (CT) 6 MS1002 Steam leak outside containment requiring power reduction 6a R (RO) Power reduction (with control rods in manual from event 3)

N (B) 7 MS1002 M(ALL) Steam break outside containment requiring unit trip (CT) 8 TU03 I/C (B) No automatic turbine trip will occur/"A" MSTV will not close automatically 9 I/C (B) Turbine-driven AFW pump (1-FW-P-2) fails to start in auto Events 8 and 9 are part of event 7 and are numbered only for use on subsequent forms.

The scenario can be terminated once the crew transitions out of 1-E-2.

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

DOMINION NORTH ANNA POWER STATION LICENSED OPERATOR REQUALIFICATION EXAMINATION SIMULATOR EXAMINATION GUIDE SCENARIO 2016 NRC 1

SIMULATOR EXAMINATION GUIDE EVENT DESCRIPTION

1. "B" Pressurizer spray valve (1-RC-PCV-1455B) fails open. Can be closed with SOV. (CT)

2. SW pump, 2-SW-P-1A, trips

3. Continuous automatic rod insertion which will stop when rods in manual (CT)

4. Selected Ch III steam flow channel (1-MS-FI-1484) fails low on "B" SG (CT)

5. Charging pump "A" trips with failure of standby pump ("B") to auto start and discharge check valve on "A" pump sticks open (CT)

6. Steam leak outside containment requiring a power reduction 6a. Power reduction

7. Steam break outside containment requiring reactor trip (CT)

8. No automatic turbine trip/ "A" MSTV doesn't close automatically

9. Turbine-driven AFW pump (1-FW-P-2) fails to start in auto Scenario Recapitulation:

Malfunctions after EOP entry 2 No automatic turbine trip/"A" MSTV doesn't close automatically, turbine-driven AFW pump doesn't automatically start Total Malfunctions 9 Failed pressurizer spray valve, SW pump trip, continuous rod insertion, failed steam flow channel, charging pump trip/failure of standby pump/discharge check valve failure, steam leak, steamline break, no automatic turbine trip/"A" MSTV doesn't close automatically, turbine-driven AFW pump doesn't automatically start Abnormal Events 6 Failed pressurizer spray valve, SW pump trip, continuous rod insertion, failed steam flow channel, charging pump trip/failure of standby/discharge check valve failure, steam leak Major Transients 1 Faulted SG EOPs Entered 1 E-2 EOP Contingencies 0 Critical Tasks 5 SCENARIO DURATION 90 Minutes 2016 NRC 1 Page 2 Revision 1

SIMULATOR EXAMINATION SCENARIO

SUMMARY

SCENARIO 2016 NRC 1 The scenario begins with the unit at 100% power, MOL. 1-SW-P-1A, Unit 1 "A" SW pump, is tagged out for major repairs. 1-BC-P-1B is tagged for shaft replacement, not expected back for several days. 2H is the protected train.

Once the crew has taken the unit one of the pressurizer spray valves (1-RC-PCV-1455B) will fail open. The crew will respond in accordance with 1-AP-44, "Loss of RCS Pressure," and the RO will be required to use the remote close SOV in order to close the spray valve (CT). Once the crew has stabilized the unit, or at the direction of the lead evaluator, the next event can occur.

2-SW-P-1A will trip, leaving no pump running on the "B" SW header. The crew will enter 0-AP-12, "Loss of Service Water," and start 1-SW-P-1B. The unit supervisor will consult TS and enter the action of 3.7.8B. Once SW flow has been restored and TS reviewed, or at the direction of the lead evaluator, the next event can occur.

Next, the control rods will begin to insert for no reason. The crew will enter 1-AP-1.1, "Uncontrolled Continuous Rod Motion," and place control rods in manual (CT). Once the crew has stabilized the unit, or at the direction of the lead evaluator, the next event can occur.

At this time channel III steam flow (1-MS-FI-1484) on "B" SG will fail low, the crew will enter 1-AP-3, "Loss of Vital Instrumentation," and take manual control of "B" main feed regulation valve (MFRV) (CT). The crew will swap instrumentation to an operable channel. The US will review Tech Specs for the failure. Once the channels have swapped and TS reviewed, or at the direction of the lead evaluator, the next event can occur.

1-CH-P-1A, "A" charging pump, will trip and the standby pump ("B") will not auto start. The discharge check valve on the previously running pump ("A") will stick open. The crew will start a standby pump (either "B" or "C") and enter 1-AP-49, "Loss of Normal Charging," and close the discharge MOVs on the "A" pump (CT) The crew will restore letdown flow and the US will consult TS 3.5.2 and may make arrangements to swap to the "C' charging pump (if not already running).

Once letdown is restored and TS have been reviewed, or at the direction of the lead evaluator, time the next event can occur.

A steam leak will develop on the "B" steam line outside containment. The crew will enter 1-AP-38, "Excessive load Increase," and begin reducing turbine power. The RO will be required to insert rods in manual. Once a sufficient load decrease has occurred, the next event can occur.

A main steamline break will occur outside containment. The crew will enter 1-E-0, "Reactor Trip or Safety Injection." The turbine will not automatically trip, but will trip when the pushbuttons are pressed . Also, "A" MSTV will not close automatically when required, but can be closed manually.

The turbine driven AFW pump, 1-FW-P-2, will fail to start automatically and will have to be manually started. The crew will proceed through 1-E-0 and transition to 1-E-2, "Faulted Steam Generator Isolation," and isolate the faulted SG (CT). The crew will announce transition to 1-ES-1.1, "SI Termination," the scenario can be terminated at this time with direction from the lead 2016 NRC 1 Page 3 Revision 1

evaluator.

2016 NRC 1 Page 4 Revision 1

SCENARIO TURNOVER SHEET Read the following to the crew:

Purpose:

This examination is intended to evaluate the crews performance of various tasks associated with the Initial License Operator Training Program. All activities should be completed in accordance with approved operations standards.

1. You are on a day shift during the week.

2. A rough log should be maintained to aid in making reports and to help during briefs.

3. Respond to what you see. In the unlikely event that the simulator fails such that illogical indications result, the session will be terminated and the crew informed.

Unit Status:

Unit 1 is at 100% power. RCS boron is 1096 ppm and core age is 9,000 MWD/MTU. Aux steam is on unit 2.

Unit 2 is at 100% power.

Equipment Status:

1-SW-P-1A tagged out for major repairs. 1-BC-P-1B is tagged for shaft replacement, not expected back for several days. Maintenance rule window is green. 2H is the protected train.

1-BC-P-1A and both spent fuel pit cooling pumps are protected.

Shift Orders:

Maintain current plant conditions. Assist maintenance with work on 1-SW-P-1A and 1-BC-P-1B, as required.

2016 NRC 1 Page 5 Revision 1

EVENT 1: Given that the unit is at power and "B" PRZR spray valve has failed open, the crew will be expected to respond in accordance with 1-AP-44, "Loss of Reactor Coolant Pressure."

TIME EXPECTED ACTION INSTRUCTOR REMARKS SPD Verified: __________ (Initials)

PCS alarms for "B" spray valve PRZR spray valve 1-RC-PCV-1455B has full open indication.

Master pressure controller output decreases.

PRZR pressure decreases.

Annunciators B-F7and B-H6 illuminate RO identifies annunciator B-F7, PRZ HI-LO PRESS.

RO identifies RCS pressure decreasing.

US directs crew to enter 1-AP-44.

RO monitors RCS pressure greater than 1870 psig.

RO checks PRZR PORVs closed. (YES)

RO checks master pressure controller not failed. (YES)

RO checks spray valves closed. (NO)

NOTE: Valve cannot be manually closed.

Crew must use SOV.

CT1 Crew stops RCS pressure decrease: Critical task RO closes REMOTE CLOSE SOV *Prior to reaching an automatic reactor for spray valve. trip on low pressure RO verifies all PRZR heaters are energized.

RO checks that aux spray valve is closed.

RO checks PRZR safety valves closed and PRZR PORVs closed or isolated.

RO verifies RCS pressure stable or increasing.

RO verifies RCS pressure returned to normal.

RO adjusts sprays and heaters, as required, to maintain normal pressure.

US references Technical Specification *Not counted as TS 3.4.1, Action A (2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />), if pressure went below 2205 psig. (TS 3.4.11 and 3.4.13 are not applicable.

2016 NRC 1 Page 6 Revision 1

EVENT 1: Given that the unit is at power and "B" PRZR spray valve has failed open, the crew will be expected to respond in accordance with 1-AP-44, "Loss of Reactor Coolant Pressure."

TIME EXPECTED ACTION INSTRUCTOR REMARKS US requests Work Control Center supervisor to inform the OMOC of the failure and initiate CR.

NOTE: The next event can occur after the Validation time: 7 minutes crew has returned RCS pressure to normal, and as directed by the lead evaluator.

2016 NRC 1 Page 7 Revision 1

EVENT 2: Given the plant in Mode 1 and a trip of 2-SW-P-1A has occurred, the crew will respond in accordance with 0-AP-12, "Loss of Service Water."

TIME EXPECTED ACTION INSTRUCTOR REMARKS SPD Verified: __________ (Initials)

Annunciators J-D3, J-B3, then B-B7, B-E8, and B-C8 illuminate Unit 2 "A" SW pump has amber and green lights lit "B" SW header flow decreases Crew identifies annunciator J-D3, SW PP 1-P1A, 2-P1A AUTO TRIP..

BOP identifies that unit 2 "A" SW pump has tripped.

US directs entry into 0-AP-12.

BOP checks SW reservoir level > 310 feet.

(SG panels ~314.5)

Crew checks SW system for integrity and flooding:

AB sump level normal (H&J SG panel)

Chiller room sump level normal (J)

Turbine building valve pit sump level normal (no alarms)

No reports of flooding.

Crew verifies SW supply headers are intact. (Safeguards panels)

Crew verifies at least one SW pump running on each supply header. (NO)

BOP starts 1-SW-P-1B. (J Safeguards panel)

Crew verifies return header flows indicated.

Crew verifies SW system is stable:

SW pump amps SW pump discharge pressure Both supply headers in service Both return header flows normal.

Crew verifies operability of equipment.

US refers to Tech Spec 3.7.8B and enters Throttling was previously verified for action to verify SW throttled within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> tagout of 1-SW-P-1A.

and restore one pump to service within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> BOP checks SW reservoir level between 313' and 315'.

2016 NRC 1 Page 8 Revision 1

EVENT 2: Given the plant in Mode 1 and a trip of 2-SW-P-1A has occurred, the crew will respond in accordance with 0-AP-12, "Loss of Service Water."

TIME EXPECTED ACTION INSTRUCTOR REMARKS NOTE: The next event may occur once Validation time: 10 minutes SW pump has been started, and at the direction of the lead evaluator.

2016 NRC 1 Page 9 Revision 1

EVENT 3: Given that the unit is operating at power and control rods are inserting for no apparent reason, the crew will be expected to respond in accordance with 1-AP-1.1, "Continuous Uncontrolled Rod Motion."

TIME EXPECTED ACTION INSTRUCTOR REMARKS SPD Verified: __________ (Initials)

Rods step in at maximum speed RO identifies control rods stepping in at maximum speed.

RO/BOP identify no known cause of rod insertion. (NIs, 1st stage pressure, Tave/Tref mismatch)

US directs crew to enter AP-1.1.

CT2 Crew takes action to stop rod motion and Critical Task stabilize the unit. *Prior to automatic reactor trip RO places rod control in MANUAL.

RO verifies rod motion stopped.

RO maintains the following using rods/boration:

Rods above Lo/Lo-Lo limit AFD in spec (A-H7 not LIT).

RO checks RCS Tave > 541F, above min and below max of attachment 2. Adjusts as directed by the US.

RO checks PRZR pressure stable or trending to 2235 psig. Adjusts heaters and spray as required.

RO checks PRZR level stable. Adjusts charging, if required.

Crew checks controls rods above the lo insertion limit.

Crew maintains stable plant conditions.

US notifies I&C to investigate.

The US reports the failure to the Work Control Center and requests that the reactivity management admin procedure be referenced, appropriate notifications made, and Condition Report be initiated.

NOTE: The next event can occur after the Validation time 8 minutes crew has stabilized the plant, and as directed by the lead evaluator.

2016 NRC 1 Page 10 Revision 1

EVENT 4: Given that the unit is at power, and a selected steam flow channel (1-MS-FI-1484) has failed low, the crew will be expected to respond in accordance with 1-AP-3, "Loss of Vital Instrumentation."

TIME EXPECTED ACTION INSTRUCTOR REMARKS SPD Verified: __________ (Initials)

Annunciators F-E2 and F-F2 illuminate 1-MS-FI-1484 fails low in range "B" MFRV demand decreases "B" SG feed flow and level decrease BOP identifies annunciator F-E2, STM GEN .

1A FW > STM FLOW CH III-IV, and informs US.

Crew identifies "B" SG steam flow channel III failed low.

US directs crew to perform immediate actions of 1-AP-3:

Crew verifies SG level controlling channels normal (NO)

CT3 BOP places "B" MFRV in Critical Task MANUAL and adjusts to control *Prior to a reactor trip due to SG level "B" SG level Crew verifies turbine first-stage pressure channels normal RO verifies PRZR level indications are normal.

Crew verifies redundant instrument channels normal RO verifies systems affected by PRZR level channels normal:

Operable channel selected Emergency bus B/U heaters restored L/D in service Level control in automatic Control group heaters not tripped.

Crew verifies both turbine first stage pressure channels normal.

Crew verifies operable channels selected for SGWLC. (NO) 2016 NRC 1 Page 11 Revision 1

EVENT 4: Given that the unit is at power, and a selected steam flow channel (1-MS-FI-1484) has failed low, the crew will be expected to respond in accordance with 1-AP-3, "Loss of Vital Instrumentation."

TIME EXPECTED ACTION INSTRUCTOR REMARKS Crew swaps SGWLC channels:

RO verifies rod control to MANUAL RO turns steam dumps off or swaps to steam pressure mode as directed by the US BOP verifies all FRV B/P in MANUAL and places MFRVs in MANUAL Crew swaps all Steam flow/Feed flow Seven switches are placed to yellow and First-stage pressure channels to dots channel IV (Vertical board)

BOP verifies SG level median controlling channels are functional BOP returns MFRVs to AUTOMATIC RO verifies Steam dumps are available RO returns steam dumps to TAVE mode Checks both 1st stage pressure operable Verifies/places interlock switches in Off/Reset Checks P-E4 not lit Verifies/places mode selector in Tave Verifies steam dump demand is zero Place interlock switches in ON RO checks Tave/Tref within 1.5F and leaves rods in manual due to earlier failure.

NOTE: During this time the WCC will call and inform the control room that the instrument techs have found a problem in auto rod control, rods will work in manual.

Crew verifies no other instrument channels are failed.

Crew refers to 1-MOP-55.77 for placing the failed channel in trip.

US refers to Technical Specifications:

3.3.2 Functions 1f, 1g, 4d, and 4e -

Condition D Determines the failed channel must be placed in trip within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

2016 NRC 1 Page 12 Revision 1

EVENT 4: Given that the unit is at power, and a selected steam flow channel (1-MS-FI-1484) has failed low, the crew will be expected to respond in accordance with 1-AP-3, "Loss of Vital Instrumentation."

TIME EXPECTED ACTION INSTRUCTOR REMARKS NOTE: The next event can occur after the Validation time 15 minutes crew swaps channels and the US has reviewed Tech Specs, as directed by the lead evaluator.

2016 NRC 1 Page 13 Revision 1

EVENT 5: Given that the unit is at power and a loss of the running charging pump ("A") and a failure of the auto-start of the standby pump ("B") concurrent with a failed open discharge check valve ("A") has occurred, the crew will be expected to respond in accordance with 1-AP-49, "Loss of Normal Charging."

TIME EXPECTED ACTION INSTRUCTOR REMARKS SPD Verified: __________ (Initials)

Annunciators C-A5, C-B5, C-C5, and C-G6 are illuminated 1-CH-P-1A has amber light lit on breaker indication "B" charging pump does not auto-start 1-CH-FI-1122A is off-scale low 1-CH-FCV-1122 ramps open Charging pump discharge pressure decreases Letdown isolates RO identifies annunciator C-A5, CH PP 1A 15H6 LOCKOUT.

Crew identifies loss of running charging with no auto start of standby pump and no charging flow.

NOTE: Crew may start either the "B" or "C" charging pump based on the fact that it should have auto-started (per the DNOS), or start the pump in 1-AP-49 step 20. These steps were not included.

RO starts either 1-CH-P-1B or 1-CH-P-1C.

RO identifies there is still no charging flow indicated.

US directs crew to enter 1-AP-49.

Crew checks the charging pump that auto started for gas binding. (NO)

Crew identifies that a charging pump manipulation has taken place.

CT4 Crew closes discharge MOVs for Critical Task the previously running pump *Prior to Safety Injection being required

("A"). by degraded plant conditions.

(E-0, Step 4) 1-CH-MOV-1286A (Vertical Board) 1-CH-MOV-1287A (Vertical Board)

RO verifies charging conditions returning to normal:

Discharge pressure Charging flow Motor amps.

2016 NRC 1 Page 14 Revision 1

EVENT 5: Given that the unit is at power and a loss of the running charging pump ("A") and a failure of the auto-start of the standby pump ("B") concurrent with a failed open discharge check valve ("A") has occurred, the crew will be expected to respond in accordance with 1-AP-49, "Loss of Normal Charging."

TIME EXPECTED ACTION INSTRUCTOR REMARKS NOTE: If sent, operator will report that the breaker for the "A" Charging pump (15H6) has a timed overcurrent drop.

Crew returns letdown to service:

Crew controls charging Crew puts 1-CH-PCV-1145 in manual and opens to 100%

Crew verifies/opens letdown isolation valves:

1-CH-TV-1204A/B (Safeguards panels) 1CH-LCV-1460A/B 1-CH-HCV-1200B Crew opens letdown orifice valve Crew adjusts 1-CH-PCV-1145 to establish 300 psig letdown pressure and places valve in AUTO When level is returned to normal, crew places charging back in automatic.

RO checks standby charging pumps in auto-after-stop. ("C" has no auto)

US verifies that CRs are being submitted.

US reviews TS 3.5.2A for having only one At minimum, once discharge valves are HHSI pump (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />). closed, SRO should note entry into TS 3.5.2A NOTE: Crew may also discuss an entry into TS 3.0.3, which was applicable until the discharge MOVs were closed.

NOTE: Crew may discuss starting the "C" pump on the 1H bus.

NOTE: Once the US has referred to Tech Validation time 10 minutes Specs, pressurizer level has stabilized, and as directed by the Lead Evaluator, the next event can occur.

2016 NRC 1 Page 15 Revision 1

EVENT 6: Given a faulted steam generator outside containment the crew will respond in accordance with "1-AP-38," Excessive Load Increase."

TIME EXPECTED ACTION INSTRUCTOR REMARKS SPD Verified: __________ (Initials)

Annunciator D-C8 illuminates Reactor power increases RCS temperature decreases Megawatts decrease RO identifies 1D-C8, SMOKE DET SYS SMOKE INDICATION TROUBLE, annunciator and notifies US.

Crew identifies smoke detectors alarming in MSVH.

Crew identifies reactor power increasing and temperature decreasing.

US directs crew to enter 1-AP-38.

RO checks steam dumps closed.

BOP checks SG PORVs closed.

Crew begins ramping the turbine down to stabilize power 100%.

NOTE: Step directs reducing reactor Attached power using either OPERATOR AUTO or TURBINE MANUAL. BOP may use either Attach. 3 of 1-AP-38 or the RO turnover GOP to prepare the turbine for ramping in automatic.

BOP places turbine in OPER AUTO and IMP-IN.

BOP removes turbine from limiter.

BOP begins ramping turbine down.

RO verifies proper auto rod control. (NO) Rods are in manual.

RO uses control rods in manual to control Annunciator B-A7 alarms at 5 degrees Tave within 1.5F of Tref.

RO energizes additional PRZR heaters, as required.

BOP checks reactor power reduced to the power level before the event started.

2016 NRC 1 Page 16 Revision 1

EVENT 6: Given a faulted steam generator outside containment the crew will respond in accordance with "1-AP-38," Excessive Load Increase."

TIME EXPECTED ACTION INSTRUCTOR REMARKS RO maintains rods above limits and AFD within limits.

BOP checks main generator output stable.

NOTE: When sent, an operator will report that steam is coming from the upper louvers on the MSVH.

RO checks Tave on program with Tref (RO may start a boration.)

Crew checks steam flow channel indications are normal.

BOP checks turbine in OPER AUTO.

Crew checks SG PORVS are closed.

Crew checks SG safety valves are closed.

BOP checks MSR inlet FCV operation is normal.

BOP checks 1-AS-PCV-105 is operating normally. (MS to AS PCV)

Crew checks plant steam systems are intact.

NOTE: Once enough of a load reduction Validation time 8 minutes has been seen, and with direction of the lead evaluator, the next event can occur.

If the crew decides to trip the reactor, the next event can be inserted at that time.

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EVENT 7: Given that the unit is at power and a steam break has occurred, the unit will respond in accordance with 1-E-0, "Reactor Trip or Safety Injection," 1-E-2, "Faulted Steam Generator," and 1-ES-1.1, "SI Termination.

TIME EXPECTED ACTION INSTRUCTOR REMARKS SPD Verified: __________ (Initials)

"B" SG pressure rapidly decreases RCS pressure and temperature decrease Turbine does not trip automatically Turbine-driven AFW pump does not start automatically "A" MSTV does not close automatically RO reports that RCS conditions are rapidly degrading.

BOP reports that "B" SG pressure is decreasing.

US determines that the unit should be tripped due to the severity of the steam leak.

US directs crew to enter 1-E-0.

Crew manually trips the reactor.

Rx trip and bypass breakers open Rod bottom lights lit Flux lowering..

BOP trips turbine:

Presses manual trip pushbuttons Turbine stop valves closed NOTE: "A" MSTV will not Reheaters reset automatically close.

Reheater FCVs closed After 30 sec: G-12 open RO verifies emergency busses are energized.

Crew checks if safety injection has actuated, or should have actuated.

SI first out or LHSI pumps running (YES)

RO/BOP manually safety inject.

RO verifies that no CAP items 1-5 are applicable.

US hold brief and hands out attachments Attached 4(5), 7.

Crew verifies SI flow indicated. ("C" charging pump may be started at this time)

Crew checks RCS pressure < 225 psig.

(NO) 2016 NRC 1 Page 18 Revision 1

EVENT 7: Given that the unit is at power and a steam break has occurred, the unit will respond in accordance with 1-E-0, "Reactor Trip or Safety Injection," 1-E-2, "Faulted Steam Generator," and 1-ES-1.1, "SI Termination.

TIME EXPECTED ACTION INSTRUCTOR REMARKS NOTE: Attachment 7 of 1-E-0 will isolate "B" SG Crew checks AFW flow indicated to all SGs. (NO)

BOP starts 1-FW-P-2. Could also be done by attachment 4 BOP verifies that total AFW flow is > 340 gpm.

• RO checks RCS average temperature stable at or trending to 547F. (NO)

RO verifies that no steam is being dumped.

BOP/crew adjust AFW flows to maintain >

340 gpm to "A" and "C" SGs until NR level in at least one SG is > 11%.

• RO checks pressurizer PORVS are closed.

• RO checks that pressurizer spray valves are responding to control pressure at 2235 psig, or are closed with zero demand.

• RO checks at least one PORV block valves is open.

RO checks that RCS subcooling based on CETCs is < 25F. (NO)

Crew checks if all SG pressures are > 80 Decision making step psig and under control of operator. (NO)

US directs transition to 1-E-2.

NOTE: Some of the following actions will have been done by attachment 7 of 1-E-0.

BOP verifies MSTVs and MSTV bypass valves closed. (attach 7) (NO)

BOP closes "A" MSTV using either PB on safeguards panels or using APP R switch on BB2.

BOP checks pressures in all SGs:

Any > 80 psig and stable.

Crew identifies that "B" SG is faulted.

(attach 7) 2016 NRC 1 Page 19 Revision 1

EVENT 7: Given that the unit is at power and a steam break has occurred, the unit will respond in accordance with 1-E-0, "Reactor Trip or Safety Injection," 1-E-2, "Faulted Steam Generator," and 1-ES-1.1, "SI Termination.

TIME EXPECTED ACTION INSTRUCTOR REMARKS CT5 Crew isolates the faulted SG. Critical Task

• Items above line are done by attach 7 of 1-E- Prior to transitioning to 1-E-1 0 (Complicating mitigating strategy as BOP verifies all 1-FW-MOV-154A/B/C transition should be to 1-ES-1.1).

closed BOP verifies all MFRVs and bypass MFRVs BOP closes 1-FW-MOV-100B.

BOP verifies 1-FW-HCV-100B closed.

BOP verifies "B" SG PORV closed.

Verifies attachment 7 of 1-E-0 is complete.

BOP verifies all SG blowdown trip valves closed.

Crew dispatches an operator to locally close 1-MS-56. (attach 1 of 1-E-2)

Crew dispatches an operator to verify closed 1-MS-58. (attach 1 of 1-E-2)

BOP checks ECST level.

BOP verifies outside instrument air is supplying containment.

Crew checks for secondary radiation.

RO resets SI RO resets Phase A RO resets AMSAC Crew verifies secondary radiation indications are normal on AE, SG blowdown, SG main steam lines and TT exhaust.

Crew checks if SI can be terminated.

RCS subcooling based on CETCs > 25F Total AFW flow > 340 gpm or one intact SG > 11%

RCS pressure stable or increasing PRZR level > 21%

US directs crew to transition to 1-ES-1.1.

NOTE: Scenario can be ended when Validation time: 14 minutes crew transitions out of 1-E-2, or as directed by the lead evaluator.

2016 NRC 1 Page 20 Revision 1

REFERENCES PROCEDURE REV.

Abnormal Procedure 1-AP-44, "Loss of Reactor Coolant System Pressure." 19 Abnormal Procedure 0-AP-12, "Loss of Service Water." 39 Abnormal Procedure 1-AP-1.1, "Continuous Uncontrolled Rod Motion." 9 Abnormal Procedure 1-AP-3, "Loss of Vital Instrumentation." 28 Abnormal Procedure 1-AP-49, "Loss of Normal Charging." 14 Abnormal Procedure 1-AP-38, "Excessive Load Increase." 19 Emergency Procedure 1-E-0, "Reactor Trip or Safety Injection." 49 Emergency Procedure 1-E-2, "Faulted Steam Generator Isolation." 14 Emergency Procedure 1-ES-1.1, "SI Termination." 23 Station Annunciator Response Procedures. N/A Administrative Procedure PI-AA-5000, "Human Performance." 8 INPO, Guideline for Teamwork and Diagnostic Skill Development: INPO 88-003, Jan. 1988 INPO, ACAD 07-002 Simulator Training Guidelines Jan. 2007 2016 NRC 1 Page 21 Revision 1

ATTACHMENT 1 SIMULATOR OPERATOR'S COMPUTER PROGRAM 2016 NRC 1 Page 22 Revision 1

SIMULATOR OPERATOR'S COMPUTER PROGRAM 2016 NRC 1 Initial conditions

1. Recall IC 321

2. Ensure Tave, Tref, PDTT level, and VCT level are selected on trend recorders.

3. 2H is the protected train.

4. Place red stickers on 1-BC-P-1B and 1-SW-P-1A.

5. Have copy of 1-E-2, Attachment 1 for booth.

PRELOADS PRIOR TO SCENARIO START CONDITION MALFUNCTION/OVERRIDE/ETC.

Tagout 1-SW-P-1A Place pump in PTL.Rack out breaker and close discharge valve.

Remote functions:

SWP1A_RACKIN SW_6 = 0 Tagout 1-BC-P-1B Place pump in PTL and Rack out breaker.

Remote function:

BCP1B_RACKIN = RACKOUT Failure of 1-CH-P-1B to start Switch override:

in AUTO CHP1B_ASTP = False Failure of "A" MSTV to Malfunction:

close automatically MS0501 Set up trigger 30 to allow valve to be close manually TVMS101A_1_CLOSE .OR. TVMS101A_2_CLOSE .OR. MSTV_APP_R_CLOSE DMF MS0501 Failure of turbine to trip Malfunction:

automatically TU03 Terry Turbine fails to start Remote Function:

automatically FWP2_AUTO_DEFEAT = T Monitor:

MTC_FLAG = F 2016 NRC 1 Page 23 Revision 1

SCENARIO EVENTS EVENT 1 "B" spray valve fails open MALFUNCTIONS/OVERRIDES Malfunction:

RC2002, Delay time = 5, Value = True, Trigger = 1 The next event will occur after the crew has returned RCS pressure to normal, or at the direction of the lead evaluator.

COMMUNICATIONS 2016 NRC 1 Page 24 Revision 1

EVENT 2 Loss of 2-SW-P-1A MALFUNCTIONS/OVERRIDES Malfunction:

SW0104, Delay time = 5, Trigger = 2 The next event may occur once SW pump has been started, and at the direction of the lead evaluator.

COMMUNICATIONS Outsides operator can report after 10 minutes that there is nothing wrong with 2-SW-P-1A locally and that 1-SW-P-1B is running normally.

If sent, a Safeguards operator can report after 3 minutes an overcurrent drop on breaker (25H5) for 2-SW-P-1A. Unit 2 SRO and WCC have been informed.

2016 NRC 1 Page 25 Revision 1

EVENT 3 Rod Insertion MALFUNCTIONS/OVERRIDES Malfunction:

RD07, Delay time = 5, Trigger = 3 The next event will occur after the crew has stabilized the plant, and as directed by the lead evaluator.

COMMUNICATIONS If operator is sent to rod drive, wait 3 minutes and then report back that you do not see anything obvious on the cabinets.

If permission is requested for moving rods:

If through WCC be sure you have their recommendation. Then call back in 3-5 minutes and say the OMOC concurs.

If OMOC directly contacted: Ask for their recommendation, then concur.

2016 NRC 1 Page 26 Revision 1

EVENT 4 "B" SG Steam flow fails low MALFUNCTIONS/OVERRIDES Malfunction:

MS0103, Delay time = 5, Severity = -0.5, Trigger = 4 The next event can occur after the crew swaps channels and the US has reviewed Tech Specs, as directed by the lead evaluator.

COMMUNICATIONS After crew has swapped control channels call as WCC and report that the Instrument shop has found a problem that affects auto rod control only. Control rods will operate normally in manual.

2016 NRC 1 Page 27 Revision 1

EVENT 5 Failure of 1-CH-P-1A with a stuck open check valve "B" charging pump fails to auto-start MALFUNCTIONS/OVERRIDES Malfunctions:

CH1601, Delay time = 5, Trigger = 5 CH2101, Delay time = 5, Trigger = 5 Once the US has referred to Tech Specs, pressurizer level has stabilized, and as directed by the Lead Evaluator, the next event can occur.

COMMUNICATIONS If sent to investigate the charging pumps and breaker, after 5 minutes report the "B" Charging pump is running normally, "A" Charging pump is not running and looks normal.

After 3 minutes report that the breaker for the "A" Charging pump (15H6) has a timed overcurrent drop.

2016 NRC 1 Page 28 Revision 1

EVENT 6 Steam leak outside containment on "B" SG MALFUNCTIONS/OVERRIDES Malfunction:

MS1002, Delay time = 5, Ramp = 600, Severity = 4, Trigger = 6 Once enough of a load reduction has been seen, and with direction of the lead evaluator, the next event can occur. If the crew decides to trip the reactor, the next event can be inserted at that time.

COMMUNICATIONS When sent to look for steam leak: wait 5 minutes and then report that steam is coming from the upper louvers (only) on the MSVH. Will remain in area to keep personnel from entering.

If asked, verify with lead evaluator and report that you cannot safely enter the building.

(If lead evaluator wants to delay this communication, first report that you will get a "buddy" and see if you can get in.)

2016 NRC 1 Page 29 Revision 1

EVENT 7 Large steam break outside containment (turbine fails to trip automatically/"A" MSTV does not close automatically, TT fails to auto start)

MALFUNCTIONS/OVERRIDES Update Steam break by inserting trigger 7 On trigger screen:

Setup trigger 7 to increase severity of steam leak.

IMF MS1002 (7 5) 40 30 Remote function:

MS_57 = 0, Delay = 30, Ramp = 30 Trigger = 10 CH_217 = 0, Delay = 30, Ramp = 30, Trigger = 20 NOTE: Scenario can be ended when the crew announces transition to 1-ES-1.1, and as directed by the lead evaluator.

COMMUNICATIONS When called to close 1-CH-217, use trigger 20 and call back 3-5 minutes later to report it is done.

When sent to do attachment 1 of 1-E-2, use trigger 10 and then call back in 5-6 minutes to report the attachment is done.

2016 NRC 1 Page 30 Revision 1

ATTACHMENT 3 SCENARIO PERFORMANCE OBJECTIVES 2016 NRC 1 Page 31 Revision 1

SIMULATOR REQUALIFICATION EXAMINATION TERMINAL PERFORMANCE OBJECTIVE Given equipment failures and operational situations, operate the plant in accordance with Technical Specifications to bring the unit to a safe condition, using applicable procedures, and applying effective teamwork, communication, and diagnostic skills.

GENERIC PERFORMANCE OBJECTIVES A. During shift operations the shift manager will take a conservative course of action, especially when uncertain conditions exist, when dealing with core cooling or heat sink availability, primary system and containment integrity, and reactivity control associated with plant evolutions.

B. During shift operations the shift manager will provide overall crew guidance by prioritizing and integrating the actions of the shift crew in accordance with administrative procedures.

C. During shift operations each crew member will participate in a team effort that resolves conflicts, provides input into the team decision and communicates all the necessary information to enhance teamwork in accordance with administrative procedures.

D. During shift operations the Shift Technical Advisor will independently assess events and based on those assessments make recommendations to the crew regarding mitigation strategy.

E. During shift operations each crew member will utilize operator fundamentals to ensure Teamwork Effectiveness, High Standards for Controlling Evolutions, Indications Monitored Closely, a Natural Bias for Conservatism, and Knowledge of Plant Design and Theory.

2016 NRC 1 Page 32 Revision 1

EVENT 1 PERFORMANCE OBJECTIVES EVENT GOAL: Given that the unit is at power and "B" PRZR spray valve has failed open, the crew will be expected to respond in accordance with 1-AP-44, "Loss of Reactor Coolant Pressure."

NORTH ANNA SPECIFIC TASKS:

R634 Respond to a loss of Reactor Coolant System pressure CRITICAL TASK:

See next page 2016 NRC 1 Page 33 Revision 1

CT Statement:

Crew stops RCS pressure decrease.

Safety Significance:

Failure to close the RCS spray valve under the postulated plant conditions constitutes "mis-operation or incorrect crew performance which leads to degradation of any barrier to fission product release." In this case, DNBR is reduced. Therefore, failure to close the spray valve represents a "demonstrated inability by the crew to take an action or combination of actions that would prevent a challenge to plant safety."

Cues:

Valid indication of pressure decreasing by the presence of various annunciators, indication of RCS spray valve open, and RCS pressure indication decreasing and procedurally directed by 1-AP-44.

Performance Indicator:

RO places REMOTE CLOSE SOV in CLOSE for associated spray Feedback:

RCS pressure decrease stopped.

WOG

Reference:

Based on Appendix B CT-10 Conditions:

Prior to reaching an automatic reactor trip on low pressure.

2016 NRC 1 Page 34 Revision 1

EVENT 2 PERFORMANCE OBJECTIVES EVENT GOAL: Given the plant in Mode 1 and a trip of 2-SW-P-1A has occurred, the crew will respond in accordance with 0-AP-12, "Loss of Service Water."

NORTH ANNA SPECIFIC TASKS:

R653 Respond to a loss of a service water pump.

S70 Evaluate compliance with technical specifications.

CRITICAL TASK:

N/A 2016 NRC 1 Page 35 Revision 1

EVENT 3 PERFORMANCE OBJECTIVES EVENT GOAL: Given that the unit is operating at power and control rods are inserting for no apparent reason, the crew will be expected to respond in accordance with 1-AP-1.1, "Continuous Uncontrolled Rod Motion."

NORTH ANNA SPECIFIC TASKS:

R475 Perform the immediate operator actions in response to a continuous uncontrolled rod motion.

CRITICAL TASK:

See next page 2016 NRC 1 Page 36 Revision 1

CT Statement:

Crew takes action in accordance with AP-1.1, to stop rod motion and stabilize the unit.

Safety Significance:

Core reactivity is not under control of the operator due to the failed control channel. "It is expected that the operator will attempt to take manual actions to correct for anomalous conditions during power operation."

Cues:

Continuous inward control rod motion with TAVE and TREF matched.

Performance Indicator:

RO places rod control to manual.

Feedback:

Rod motion stops WOG

Reference:

None Conditions:

Prior to automatic reactor trip 2016 NRC 1 Page 37 Revision 1

EVENT 4 PERFORMANCE OBJECTIVES EVENT GOAL: Given that the unit is at power, and a selected steam flow channel (1-MS-FI-1484) has failed low, the crew will be expected to respond in accordance with 1-AP-3, "Loss of Vital Instrumentation."

NORTH ANNA SPECIFIC TASKS:

R626 Respond to a steam generator water level control channel failure S70 Evaluate compliance with technical specifications CRITICAL TASK:

See next page 2016 NRC 1 Page 38 Revision 1

CT Statement:

Crew takes manual control of main feed reg valve to control steam generator level Safety Significance:

Failure to control steam generator level can result in a reactor trip (on either high or low SG level) which will complicate plant recovery actions and could result in damage to SG components or MS lines.

Cues:

Indication of loss of steam generator level control include:

Annunciators. (i.e. SG level error, FW < (or >) steam flow)

Indications that a controlling steam flow, steam pressure, or feed flow instrument has failed hi or lo Associated MFRV opening or closing to control level due to failed instrument Steam generator level increasing or decreasing to reactor trip setpoint Performance Indicator:

Crew takes manual control of associated MFRV and returns level to program Feedback:

Steam generator level returns to program WOG

Reference:

None Conditions:

Prior to receiving an automatic reactor trip on high or low steam generator level 2016 NRC 1 Page 39 Revision 1

EVENT 5 PERFORMANCE OBJECTIVES EVENT GOAL: Given that the unit is at power and a loss of the running charging pump

("A") and a failure of the auto-start of the standby pump ("B") concurrent with a failed open discharge check valve ("A") has occurred, the crew will be expected to respond in accordance with 1-AP-49, "Loss of Normal Charging."

NORTH ANNA SPECIFIC TASKS:

R704 Respond to a loss of normal charging.

S70 Evaluate compliance with technical specifications.

CRITICAL TASK:

See next page 2016 NRC 1 Page 40 Revision 1

CT Statement:

Crew takes action to prevent charging pump run-out due to a stuck open discharge check valve on a non-running charging pump.

Safety Significance:

Failure to prevent charging pump run-out constitutes a "mis-operation or incorrect crew performance which leads to degraded ECCS capacity."

Cues:

Indication/annunciation that one charging pump has tripped or been shutdown with a stuck open discharge check valve.

High amps on the running charging pump.

Low/no charging flow or seal injection indicated.

Performance Indicator:

Crew closes charging pump discharge MOVs on the previously running charging pump.

Feedback:

Discharge MOVs for the previously running pump indicate closed and charging and seal injection flow returns to normal.

WOG

Reference:

None.

Conditions:

Prior to Safety Injection being required by degraded plant conditions.

2016 NRC 1 Page 41 Revision 1

EVENT 6 PERFORMANCE OBJECTIVES EVENT GOAL: Given a faulted steam generator outside containment the crew will respond in accordance with "1-AP-38," Excessive Load Increase."

NORTH ANNA SPECIFIC TASKS:

R539 Perform the immediate operator actions in response to an excessive load increase CRITICAL TASK:

N/A 2016 NRC 1 Page 42 Revision 1

EVENT 7 PERFORMANCE OBJECTIVES EVENT GOAL: Given that the unit is at power and a steam break has occurred, the unit will respond in accordance with 1-E-0, "Reactor Trip or Safety Injection,"

1-E-2, "Faulted Steam Generator," and 1-ES-1.1, "SI Termination.

NORTH ANNA SPECIFIC TASKS:

R185 Perform the immediate operator actions in response to a reactor trip or safety injection R183 Identify and isolate a faulted steam generator R189 Terminate safety injection CRITICAL TASK:

See next page 2016 NRC 1 Page 43 Revision 1

CT Statement:

Crew isolates faulted Steam Generator.

Safety Significance:

Failure to isolate a faulted SG that can be isolated causes challenges to the integrity CSF beyond those irreparably introduced by the postulated conditions. For the reference plant, neither of these transients (blowdown of a single SG with or without RCPs running) constitutes an orange-path challenge to the integrity CSF. However, if the faulted SG is not isolated, the cooldown transient for reactor vessel inlet temperature could result in an orange-path challenge to the integrity CSF, especially if RCPs are not running.

Cues:

"B" SG is depressurizing in an uncontrolled manner or is completely depressurized and Valve position and flow rate indication that AFW continues to be delivered to the faulted SG Performance Indicator:

BOP closes 1-FW-MOV-100B to secure AFW flow to "B" steam generator.

Feedback:

AFW flow indication to "B" steam generator decreases to zero.

WOG

Reference:

Appendix B CT-17 Conditions:

Prior to transitioning to 1-E-1. (Complication of mitigating strategy) 2016 NRC 1 Page 44 Revision 1

ATTACHMENT 2 SIMULATOR PERFORMANCE DATASHEET

Scenario Performance Datasheet EVENT 1: Given that the unit is at power and "B" PRZR spray valve has failed open, the crew will be expected to respond in accordance with 1-AP-44, "Loss of Reactor Coolant Pressure."

SPD Verified: __________ (Initials)

PCS alarms for "B" spray valve PRZR spray valve 1-RC-PCV-1455B has full open indication.

Master pressure controller output decreases.

PRZR pressure decreases.

Annunciators B-F7and B-H6 illuminate EVENT 2: Given the plant in Mode 1 and a trip of 2-SW-P-1A has occurred, the crew will respond in accordance with 0-AP-12, "Loss of Service Water."

SPD Verified: __________ (Initials)

Annunciators J-D3, J-B3, then B-B7, B-E8, and B-C8 illuminate Unit 2 "A" SW pump has amber and green lights lit "B" SW header flow decreases EVENT 3: Given that the unit is operating at power and control rods are inserting for no apparent reason, the crew will be expected to respond in accordance with 1-AP-1.1, "Continuous Uncontrolled Rod Motion."

SPD Verified: __________ (Initials)

Rods step in at maximum speed EVENT 4: Given that the unit is at power, and a selected steam flow channel (1-MS-FI-1484) has failed low, the crew will be expected to respond in accordance with 1-AP-3, "Loss of Vital Instrumentation."

SPD Verified: __________ (Initials)

Annunciators F-E2 and F-F2 illuminate 1-MS-FI-1484 fails low in range "B" MFRV demand decreases "B" SG feed flow and level decrease 2016 NRC 1 Date _________ Revision 1

EVENT 5: Given that the unit is at power and a loss of the running charging pump ("A") and a failure of the auto-start of the standby pump ("B") concurrent with a failed open discharge check valve ("A") has occurred, the crew will be expected to respond in accordance with 1-AP-49, "Loss of Normal Charging."

SPD Verified: __________ (Initials)

Annunciators C-A5, C-B5, C-C5, and C-G6 are illuminated 1-CH-P-1A has amber light lit on breaker indication "B" charging pump does not auto-start 1-CH-FI-1122A is off-scale low 1-CH-FCV-1122 ramps open Charging pump discharge pressure decreases Letdown isolates EVENT 6: Given a faulted steam generator outside containment the crew will respond in accordance with "1-AP-38," Excessive Load Increase."

SPD Verified: __________ (Initials)

Annunciator D-C8 illuminates Reactor power increases RCS temperature decreases Megawatts decrease EVENT 7: Given that the unit is at power and a steam break has occurred, the unit will respond in accordance with 1-E-0, "Reactor Trip or Safety Injection," 1-E-2, "Faulted Steam Generator,"

and 1-ES-1.1, "SI Termination.

SPD Verified: __________ (Initials)

"B" SG pressure rapidly decreases RCS pressure and temperature decrease Turbine does not trip automatically Turbine-driven AFW pump does not start automatically "A" MSTV does not close automatically 2016 NRC 1 Date _________ Revision 1

Appendix D Scenario Outline Form ES-D-1 Facility: North Anna Power Station Scenario No.: (2016) NRC-2 Op-Test No.: 1 Examiners: Bruno Cabarello______________ Operators: _____________________________

Newton Lacy________________ _____________________________

Gary Callaway_______________ _____________________________

Initial Conditions: 100% MOL, 1-SW-P-1A is tagged out for major repairs. 1-BC-P-1B is tagged out for shaft replacement. 2H is the protected train.

Turnover: Maintain current plant conditions. Assist maintenance with work on 1-SW-P-1A, and 1-BC-P-1B.

Event Malf. Event Event No. No. Type* Description 1 I/C (B) Running Service Air compressor (2-SA-C-1) trips/ standby compressor (1-SA-C-1) fails to start automatically (CT) 2 CH27 I/C (RO) Failure of 1-CH-TE-1144, Non-regen HX temperature element 3 CN0901 Main Condenser vacuum leak 3a R (RO) Unit power reduction due to vacuum leak N (B) 3b RD14 I/C (RO) Rods stop stepping in automatic 4 RC0803 I/C (RO) Selected pressurizer level channel (1-RC-LT-1461) fails low (CT)

TS (S) 4a N (RO) Restore letdown 5 TU1101 I/C (B) EHC pump (1-TM-P-3) trips/standby pump (1-TM-P-4) fails to auto-start (CT) 6 RC04 TS (S) RCS leak 7 RC0101 M (ALL) LBLOCA 8 QS03 I/C (ALL) Containment Depressurization Actuation does not work automatically (CT) 9 I/C (B) LHSI pumps don't automatically start (CT) 1-SI-P-1A shaft shears when started 10 I/C (ALL) Loss of emergency recirc (loss of 1 LHSI pump and 1 containment sump suction valve) (CT) 11 I/C (B) 1-CH-MOV-1381 (Seal return isolation) does not close automatically Events 8, 9, 10 and 11 happen during event 7 and are numbered for use on subsequent forms.

The scenario can be terminated once a charging pump has been stopped in 1-ECA-1.1.

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

DOMINION NORTH ANNA POWER STATION LICENSED OPERATOR REQUALIFICATION EXAMINATION SIMULATOR EXAMINATION GUIDE SCENARIO 2016 NRC 2

SIMULATOR EXAMINATION GUIDE EVENT DESCRIPTION

1. Service Air compressor (2-SA-C-1) trips/standby fails to auto-start (CT)

2. 1-CH-TE-1144 failure (Non-regen HX temperature element)

3. Condenser vacuum leak 3a. Power reduction to stabilize vacuum 3b. Rods stop working in automatic during ramp

4. Selected pressurizer level channel (1-RC-LT-1461) fails low (CT) 4a. Letdown is restored

5. EHC pump trips (1-TM-P-3)/standby pump (1-TM-P-4) does not auto start (CT)

6. RCS leak that eventually requires a unit trip

7. LBLOCA

8. No auto CDA (CT)

9. LHSI pumps don't auto start/sheared shaft on 1-SI-P-1A (CT) 11 1-CH-MOV-1381 does not close on SI/Phase A

10. Loss of Emergency recirc (CT)

Scenario Recapitulation:

Malfunctions after EOP entry 4 Failure of automatic CDA, LHSI pumps do not automatically start/shaft shears on 1-SI-P-1A, loss of emergency recirc, 1-CH-MOV-1381 does not close on SI/phase A Total Malfunctions 12 Service Air compressor trips/standby compressor fails to auto-start, 1-CH-TE-1144 failure, Condenser vacuum leak, rods stop working in auto, pressurizer level channel failure, EHC pump trips/standby pump fails to start, RCS leak, LBLOCA, failure of automatic CDA, LHSI pumps do not automatically start/shaft shears on 1-SI-P-1A, loss of emergency recirc, 1-CH-MOV-1381 does not close on SI/phase A Abnormal Events 6 Service Air compressor trips/standby compressor fails to auto-start, 1-CH-TE-1144 fails, condenser vacuum leak, pressurizer level channel failure, EHC pump trips/standby pump fails to start, RCS leak, Major Transients 1 LBLOCA EOPs Entered 2 E-1, ECA-1.1 EOP Contingencies 1 ECA-1.1 Critical Tasks 6 SCENARIO DURATION 95 Minutes 2016 NRC 2 Page 2 Revision 1

SIMULATOR EXAMINATION SCENARIO

SUMMARY

SCENARIO 2016 NRC 2 The scenario begins with the unit at 100% power, MOL. 1-SW-P-1A, Unit 1 "A" SW pump, is tagged out for major repairs. 1-BC-P-1B is tagged for shaft replacement, not expected back for several days. 2H is the protected train.

The first event will be the trip of the running service air compressor (2-SA-C-2) and the failure of the standby compressor (1-SA-C-1) to start.) The crew will enter 1-AP-28, "Loss of Instrument Air," and start the standby compressor. (CT) Once instrument air pressure has been restored and the next event can occur.

Next, 1-CH-TE-1144, Non-regen HX temperature element, will fail low causing letdown temperature to increase. The crew will use the AR for C-C6 to take manual control of 1-CC-TCV-106 and restore letdown temperature. When temperature has decreased sufficiently, the crew will restore flow through the demin train. Once letdown temperature has been restored, the next event can occur.

At this time a condenser vacuum leak will ramp in due to the failure of the loop seals. The crew will identify the loss of condenser vacuum and enter 1-AP-14, "Loss of Condenser Vacuum." The crew will begin a load reduction to try to stabilize vacuum. The control rods will fail to move in automatic when required and the RO will have to insert rods in manual. The operator sent to the turbine building will report the loop seal problem, and state that the isolation valve will not move and ask for permission to use a valve leverage device to assist in closing the valve. Once the valve is closed condenser vacuum will recover and the crew will hold the ramp. Once vacuum has improved and the ramp stopped, the next event can occur.

A selected pressurizer level channel (1-RC-LT-1461) will fail low causing letdown to isolate. The crew will enter 1-AP-3, "Loss of Vital Instrumentation," and take actions to place 1-CH-LCV-1122 in manual and reduce charging flow to zero. The RO will then swap to an operable pressurizer level channel. The crew will restore letdown at this time (Normal event) (CT). The SRO will review TS and note that the channel must be placed in trip within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. After letdown has been restored and TS actions reviewed, the next event can occur.

Now, the operating EHC pump (1-TM-P-3) will trip and the standby pump (1-TM-P-4) will not start. The BOP will recognize the loss of EHC and start the standby pump based either on the DNOS, or the AR for low EHC pressure (CT). During this time, a RCS leak (approximately 60 gpm) will occur inside containment. The crew should respond in accordance with 1-AP-16, "Increasing Primary Plant Leakage," and 1-AP-5, "Unit 1 Radiation Monitoring System." The US should refer to Technical Specifications and either direct the crew to commence a unit shutdown or make preparations for a containment entry due to excessive RCS leakage.

The crew will receive indications of a LBLOCA and will enter 1-E-0, "Reactor Trip or Safety Injection." No LHSI pumps will start and the crew will attempt to manually start the pumps.

1-SI-P-1A will start, but will have a sheared shaft. 1-SI-P-1B will start and flow (CT). 1-CH-MOV-1381 will not close on the SI/Phase A signal and will have to closed manually. Also, when a CDA is required it will not happen automatically and the crew will have to manually actuate CDA (CT).

2016 NRC 2 Page 3 Revision 1

The crew will transition to 1-E-1 and when they check power available to the "B" train of LHSI, they will find that 1-SI-MOV-1860B has no power. An operator sent to the valve will report that the actuator is broken on the MOV. At this time the crew will transition to 1-ECA-1.1, "Loss of Emergency Coolant Recirculation." Once the crew has pressed the SI RECIRC MODE reset buttons and established a single train of SI flow (CT), the scenario can be stopped with direction from the lead evaluator.

2016 NRC 2 Page 4 Revision 1

SCENARIO TURNOVER SHEET Read the following to the crew:

Purpose:

This examination is intended to evaluate the crews performance of various tasks associated with the Initial License Operator Training Program. All activities should be completed in accordance with approved operations standards.

1. You are on a day shift during the week.

2. A rough log should be maintained to aid in making reports and to help during briefs.

3. Respond to what you see. In the unlikely event that the simulator fails such that illogical indications result, the session will be terminated and the crew informed.

Unit Status:

Unit 1 is at 100% power. RCS boron is 1096 ppm and core age is 9,000 MWD/MTU. Aux steam is on unit 2.

Unit 2 is at 100% power.

Equipment Status:

1-SW-P-1A tagged out for major repairs. 1-BC-P-1B is tagged for shaft replacement, not expected back for several days. Maintenance rule window is green. 2H is the protected train.

1-BC-P-1A and both Spent Fuel Pit Cooling pumps are protected Shift Orders:

Maintain current plant conditions. Assist maintenance with work on 1-SW-P-1A and 1-BC-P-1B, as required.

2016 NRC 2 Page 5 Revision 1

EVENT 1: Given that the unit is at power, and a loss of an air compressor (2-SA-C-1) has occurred, the crew will be expected to respond in accordance with 1-AP-28, "Loss of Instrument Air."

TIME EXPECTED ACTION INSTRUCTOR REMARKS SPD Verified: __________ (Initials) 2-SA-C-1 trips Annunciator J-D1, and possibly F-F8, J-E8 illuminate Instrument air pressure decreases 1-SA-C-1 does not start automatically BOP identifies annunciator 1J-D1, SERVICE AIR COMPRESSOR TROUBLE.

Crew identifies lowering IA pressure.

US directs crew to enter 1-AP-28.

CT1 BOP starts all available air Critical Task compressors in HAND *Prior to IA pressure decreasing to 1-IA-C-1 <70 psig, requiring a reactor trip 1-SA-C-1 1-IA-C-2A 1-IA-C-2B

• Crew checks IA pressure outside containment BOP checks IA pressure < 70 PSIG.

(NO)

Crew determines cause of loss of IA to be a failed compressor with an auto-start failure of 1-SA-C-1.

Crew verifies cause of the loss of IA is corrected.

BOP verifies IA pressure > 94 psig or trending to 94 psig.

Crew checks if either 1-IA-TV-102A or 102B is closed. (NO)

Crew returns air compressors to normal.

Crew dispatches operator to ensure 2-IA-TV-211 is closed.

RO checks letdown in service.

Crew verifies attachment 6 was not initiated.

2016 NRC 2 Page 6 Revision 1

EVENT 1: Given that the unit is at power, and a loss of an air compressor (2-SA-C-1) has occurred, the crew will be expected to respond in accordance with 1-AP-28, "Loss of Instrument Air."

TIME EXPECTED ACTION INSTRUCTOR REMARKS US directs WCC to submit a CR.

NOTE: The next event can occur once IA Validation time: 9 minutes is verified to be increasing.

2016 NRC 2 Page 7 Revision 1

EVENT 2: Given that the unit is at power and 1-CH-TE-1144 has failed low causing letdown temperature to increase, the crew will be expected to respond in accordance with the AR for C-C6, DEMIN INLET DIVERT HI TEMP.

TIME EXPECTED ACTION INSTRUCTOR REMARKS SPD Verified: __________ (Initials) 1-CH-TI-1143 indication increases 1-CH-TI-1144 indication decreases 1-CC-TCV-106 output decreases Annunciator C-C6 illuminates 1-CH-TCV-1143 bypasses the demin train RO identifies annunciator C-C6, DEMIN INLET DIVERT HI TEMP.

Crew identifies 1-CH-TCV-1143 is bypassing the demins due to a high temperature indicated on 1-CH-TI-1143.

NOTE: VCT temperature could increase C-A3 VCT high temperature may alarm to high alarm setpoint.

RO takes manual control of 1-CC-TCV-106 to stabilize temperature, per the AR.

Crew verifies that no more SW flow to the CC HXs is required.

Crew monitors reactor power.

RO ensures adequate charging flow to cool letdown flow.

RO verifies that temperature has decreased to < 115F on 1-CH-TI-1143, and then holds 1-CH-TCV-1143 in the IX position until red light is lit and amber light is not lit. RO releases switch to auto position RO verifies that flow continues through the demins when the switch is released.

NOTE: The next event can occur once Validation time: 9 minutes letdown temperature has been restored to normal.

2016 NRC 2 Page 8 Revision 1

EVENT 3/3a/3b: Given that the unit is at power and loss of condenser vacuum is occurring, the crew will respond in accordance with 1-AP-14, "Low Condenser Vacuum."

TIME EXPECTED ACTION INSTRUCTOR REMARKS SPD Verified: __________ (Initials)

Condenser vacuum degrades Annunciator A-G1 alarms if vacuum reaches setpoint Rods do not insert automatically to control temperature NOTE: The following condenser vacuum limits/setpoints apply:

G-F3 (Vacuum pre-trip) < 25"HgV (~5" HgA)

A-G1 (Loss of C-9) > 4"HgA E-D2 (Turbine trip) 20"-23"HgV (6.9-9.8"HgA) AND ASO pressure < 45 psig Monitor > 5.5"HgA - Trip reactor Crew identifies degrading condenser vacuum.

US directs crew to enter 1-AP-14.

Crew checks that generator output breaker is closed.

RO initiates RCS boration using either Attach 4 and GOP attached attachment 4 or the standard ramp plan and 1-GOP-8.3.4.

2016 NRC 2 Page 9 Revision 1

EVENT 3/3a/3b: Given that the unit is at power and loss of condenser vacuum is occurring, the crew will respond in accordance with 1-AP-14, "Low Condenser Vacuum."

TIME EXPECTED ACTION INSTRUCTOR REMARKS

• Reduce plant load at 5%/min until vacuum is stable BOP places turbine in OPERATOR AUTO:

Verifies GV tracking meter is ~ zero Ensures turbine reference and setter are matched Presses OPER AUTO pushbutton.

BOP removes turbine from valve position limiter, as required, using attachment 3 or turnover GOP:

Uses Down PB to lower Reference control Sets ramp rate thumbwheel to desired rate Presses GO When VPL light goes out - presses HOLD BOP places turbine control in IMP-IN:

Ensures turbine reference and setter are matched Presses IMP-IN pushbutton.

BOP commences manual turbine load reduction per attachment 3:

Sets Reference control using down PB Sets designated ramp rate using thumbwheel Presses GO.

RO verifies rods are in automatic.

RO/ energizes pressurizer heaters as required to maintain RCS pressure> 2205 psig.

NOTE: During the ramp the RO will Verify crew manually moves rods to determine that control rods are not control Tavg within 5 of Tref.

inserting in automatic.

• RO verifies proper auto control rod insertion. (NO)

RO places control rods in manual and adjusts as required to maintain Tavg within 5F of Tref.

• Crew monitors condenser vacuum 3.5" hg abs or less. If not and reactor power > 30%

then verifies vacuum < 5.5" hg abs.

2016 NRC 2 Page 10 Revision 1

EVENT 3/3a/3b: Given that the unit is at power and loss of condenser vacuum is occurring, the crew will respond in accordance with 1-AP-14, "Low Condenser Vacuum."

TIME EXPECTED ACTION INSTRUCTOR REMARKS Crew dispatches operator with attachment for local turbine building operations.

BOP verifies condenser vacuum breaker, 1-AS-MOV-100, is closed.

BOP inlet and outlet water box MOVs are full open.

Crew checks CW pump discharge MOVs are open (or throttled open) on all running pumps.

BOP verifies air ejector lineup.

AS pressure between 180 and 210 psig 1-AS-FCV-100A/B open 1-SV-TV-102-2, open 1-SV-TV-103, 102-1 closed BOP verifies gland steam operation:

Gland steam header pressure 120-130 psig HP TB GS pressure 1.5 to 15 psig NOTE: When dispatched, the field Verify crew manually moves rods to operator will try to isolate the loop seal control Tave before allowing vacuum per the AP-14 attachment and inform loss to be "fixed" the control room that the isolation valve will not turn. Will get permission to use a valve leverage device.

Once enough of a power decrease has been observed, the operator will report that the valve is now closed. At this time vacuum will begin to improve.

Crew checks vacuum stable or improving.

If so, US directs crew to hold the ramp.

US directs entry into either 1-OP-2.2 or 1-AP-2.2 while continuing with 1-AP-14.

Crew checks condensate system:

Condensate hotwell level normal Condensate pumps running Sends operator to check system operation.

Crew checks circ water systems:

Water boxes primed Running SW pump amps normal..

Crew verifies vacuum stable or improving.

Crew verifies main turbine on line 2016 NRC 2 Page 11 Revision 1

EVENT 3/3a/3b: Given that the unit is at power and loss of condenser vacuum is occurring, the crew will respond in accordance with 1-AP-14, "Low Condenser Vacuum."

TIME EXPECTED ACTION INSTRUCTOR REMARKS Crew checks cause of vacuum loss identified and repaired.

Crew returns main condenser to normal, as required.

Crew goes to 1-OP-2.1 to restore power to 100%.

US makes report to Work Control Center and requests that CR be submitted and management notifications be made.

NOTE: Once the vacuum leak is stopped Validation time: 15 minutes and the crew has stabilized the unit, and with direction of the lead evaluator, the next event can occur.

2016 NRC 2 Page 12 Revision 1

EVENT 4/4a: Given that the unit is at power and a selected pressurizer level channel, 1-RC-LI-1461, has failed low, the crew will respond in accordance with 1-AP-3, "Loss of Vital Instrumentation."

TIME EXPECTED ACTION INSTRUCTOR REMARKS SPD Verified: __________ (Initials)

Annunciators B-F8, B-G7, and B-E2 are illuminated 1-RC-LI-1461 fails low Letdown isolates RO identifies annunciators B-F8, PRZ LO LEVEL, and B-G7, PRZ LO LEV HTRS OFF - LETDWN ISOL.

US directs crew to enter 1-AP-3.

RO identifies that the selected pressurizer level channel 1461 has failed low.

BOP verifies SG level controlling channels are normal.

BOP verifies turbine first stage pressure indication normal.

RO verifies pressurizer level indications are normal. (NO)

CT2 RO places 1122 in manual and minimizes Part of CT2 (next page) charging. (May go to zero due to no letdown.)

Crew verifies redundant instrument channel indication normal.

RO verifies systems affected by pressurizer level channels are normal. (NO)

CT2 RO selects operable channel of pressurizer Part of CT2 (next page) level for control (459/460).

RO verifies proper annunciator configuration.

RO verifies emergency backup heater configuration.

RO verifies letdown in service. (NO) 2016 NRC 2 Page 13 Revision 1

EVENT 4/4a: Given that the unit is at power and a selected pressurizer level channel, 1-RC-LI-1461, has failed low, the crew will respond in accordance with 1-AP-3, "Loss of Vital Instrumentation."

TIME EXPECTED ACTION INSTRUCTOR REMARKS CT2 RO restores letdown using attach 2. Critical task Ensures charging flow at least 25 gpm Normal evolution Ensures 1-CH-LCV-1460A/B are *Prior to a PRZR hi level reactor trip open Ensures 1-CH-TV-1204A/B are open Places 1-CH-PCV-1145 in manual &

opens Opens 1-CH-HCV-1200B (or 1200A or C)

Adjusts 1-CH-PCV-1145 to obtain 300 psig letdown pressure and places in Auto Adjusts charging and letdown to maintain program PRZR level.

RO verifies PRZR level control in auto (NO):

RO restores pressurizer level to This will likely take a few minutes.

program Want to return to automatic for RO adjusts 1-RC-LCV-1459G, if identification of next event.

required RO places 1-CH-FCV-1122 in automatic RO resets PRZR control group heaters.

BOP verifies both turbine first stage pressure channels normal.

BOP verifies operable channels selected for SGWLC.

Crew verifies operation of other vital instrumentation.

Crew refers to 1-MOP-55.72 for placing the failed channel in trip.

US refers to TS-3.3.1 Function 9, Condition L and determine that the channel must be placed in trip within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

NOTE: The next event will occur after the Validation time: 19 minutes crew identifies the applicable MOP and pressurizer level control is back in automatic, with the direction of the lead evaluator.

2016 NRC 2 Page 14 Revision 1

EVENT 5/6: Given that the unit is at power and there are no EHC pumps running, the crew will respond in accordance with the applicable AR. Also, when indications exist of RCS leakage greater than TS limits, the crew will be expected to respond in accordance with 1-AP-16, "Excessive Primary Plant Leakage," and 1-AP-5, "Unit 1 Radiation Monitoring System."

TIME EXPECTED ACTION INSTRUCTOR REMARKS SPD Verified: __________ (Initials)

Annunciators K-F5 and T-B3 will illuminate "A" EHC pump trips "B" EHC pump does not auto-start RCS leak Charging flow increasing, VCT level decreasing Containment sump level increasing Containment gaseous radiation (RM-160) increases (also RM-163 and 164)

Annunciators K-D2, K-D4, and possible J-F2 will illuminate.

NOTE, Both events will be put in on the same trigger.

BOP identifies K-F5, TURB SUPERV PANEL TROUBLE.

Crew identifies T-B4, EH FLUID RESERVOIR LOW- PRESSURE.

BOP identifies no EHC pump running.

US directs BOP to start the standby EHC Per DNOS, since it should auto start, or AR.

pump, 1-TM-P-4.

CT3 BOP manually starts 1-TM-P-4. Critical task

• Prior to reactor trip caused by turbine trip caused by low EHC pressure Annunciator alarms at 1600 psig and auto-start should occur at 1400 psig NOTE: If the crew dispatches an operator to look at EHC pumps, the operator will report that 1-TM-P-3 is unusually hot, and 1-TM-P-4 appears normal.

US requests Work Control Center supervisor to make notifications of the failure and initiate WR and CR.

Crew identifies charging flow increasing, RCS leak and/or increasing containment rad levels, AP-5 steps included starting on next and/or increasing containment sump level. page US directs entry into 1-AP-16. Final leak value is 40 5 gpm 2016 NRC 2 Page 15 Revision 1

EVENT 5/6: Given that the unit is at power and there are no EHC pumps running, the crew will respond in accordance with the applicable AR. Also, when indications exist of RCS leakage greater than TS limits, the crew will be expected to respond in accordance with 1-AP-16, "Excessive Primary Plant Leakage," and 1-AP-5, "Unit 1 Radiation Monitoring System."

TIME EXPECTED ACTION INSTRUCTOR REMARKS Crew verifies unit in mode 1.

• RO verifies pressurizer level stable or increasing. If not, then RO will take manual control of charging. May also isolate letdown per RNO.

• Verify VCT level stable.

Crew starts a VCT makeup, as required. 1-GOP-8.3.3 attached RO checks 1-CH-LCV-1115A not diverted:

VCT position indicated In-service gas stripper level is stable (with PDTT pump secure)

In-service gas stripper flow is not indicated.

RO verifies letdown in service with normal indications:

Letdown flow Non-regen temperature Regen outlet temperature VCT temperature VCT pressure AB general exhaust radiation AB central exhaust radiation Vent stack A radiation Excess letdown temperature and pressure.

RO checks excess letdown parameters.

RO checks charging system parameters normal:

CHP discharge pressure Charging flow Regen outlet temp Seal injection flows AB sump level.

2016 NRC 2 Page 16 Revision 1

EVENT 5/6: Given that the unit is at power and there are no EHC pumps running, the crew will respond in accordance with the applicable AR. Also, when indications exist of RCS leakage greater than TS limits, the crew will be expected to respond in accordance with 1-AP-16, "Excessive Primary Plant Leakage," and 1-AP-5, "Unit 1 Radiation Monitoring System."

TIME EXPECTED ACTION INSTRUCTOR REMARKS Crew checks containment conditions:

Sump pumping rate Temperature Pressure Radiation Crew identifies increasing containment radiation and sump pumping rate.

Crew goes to attachment 5 for AP-5 1-RM-RMS-160 alarms.

Crew informs HP of alarm(s) and asks them to determine if Containment gaseous samples are required. If so, request them to obtain and analyze samples.

Crew verifies not in Modes 5 or 6. Also, that there is no obvious malfunction of rad monitor, nor any fuel handling accident.

Crew has SEM refer to EALs.

Crew determines that the alarms are due to a RCS leak in containment.

Crew performs an RCS leakrate PT, when possible.

Crew requests samples of containment sump.

Crew stops containment stop pumps, if AP-16 continued desired.

US initiates attachment for containment walkdown.

US enters Tech. Spec. 3.4.13A. Reduce leakage to within limits with completion time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or Mode 3 in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Mode 5 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

NOTE: Crew may decide to make a containment entry and look for leak as they have 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to stop the leakage per the T.S.

US directs crew to either commence a unit shutdown or make preparations for a containment entry.

2016 NRC 2 Page 17 Revision 1

EVENT 5/6: Given that the unit is at power and there are no EHC pumps running, the crew will respond in accordance with the applicable AR. Also, when indications exist of RCS leakage greater than TS limits, the crew will be expected to respond in accordance with 1-AP-16, "Excessive Primary Plant Leakage," and 1-AP-5, "Unit 1 Radiation Monitoring System."

TIME EXPECTED ACTION INSTRUCTOR REMARKS NOTE: The next event can occur after the Validation time: 16 minutes crew either makes preparations to ramp, makes preparations for a containment entry, and as directed by the lead evaluator.

2016 NRC 2 Page 18 Revision 1

EVENT 7: Given that the unit is at power, and a large break LOCA has occurred inside containment, the crew will be expected to respond in accordance with 1-E-0, "Reactor Trip or Safety Injection," and 1-E-1, "Loss of Secondary or Reactor Coolant."

TIME EXPECTED ACTION INSTRUCTOR REMARKS SPD Verified: __________ (Initials)

Pressurizer pressure and level rapidly decreasing 1-SI-P-1A and 1-SI-P-1B fail to auto-start 1-SI-P-1A shears its shaft when started 1-CH-MOV-1381 does not close automatically RO identifies PRZR pressure and level rapidly decreasing.

US directs crew to enter 1-E-0.

Crew manually trips the reactor.

Rx trip and bypass breakers open Rod bottom lights lit Flux lowering..

BOP verifies turbine trip.

Turbine trip PBs pressed Turbine stop valves closed Reheaters reset Reheater FCVs closed After 30 sec. G-12 open.

RO verifies AC emergency busses energized.

Crew checks if safety injection has actuated, or should have actuated. (YES)

SI first out (YES) or LHSI pumps running (NO)

Crew manually actuates SI.

RO identifies that CAPs 1,2,3, and 5 apply.

Crew discusses that adverse numbers are in CAP 1 effect.

CT4 Crew manually starts SI pumps. CAP 2 RO manually starts "B" LHSI pump. Critical task Notes that "A" has a sheared shaft. *Prior to entering FR-C.1 (3 chances in E-0, CAP, procedure step, attachment 4)

RO stops all RCPs and closes charging CAP 3 pump recircs.

2016 NRC 2 Page 19 Revision 1

EVENT 7: Given that the unit is at power, and a large break LOCA has occurred inside containment, the crew will be expected to respond in accordance with 1-E-0, "Reactor Trip or Safety Injection," and 1-E-1, "Loss of Secondary or Reactor Coolant."

TIME EXPECTED ACTION INSTRUCTOR REMARKS CT5 RO/BOP manually actuates CDA CAP 5 RO/BOP actuate CDA by Critical Task simultaneously turning 2 CDA *Prior to transitioning to FR-Z.1 switches (Only one set required) (two chances in 1-E-0, CAP and BO ensures CC pumps are stopped attachment 4)

RO verifies RCPs are stopped BOP ensures QS pumps are running with discharge MOVs open US initiates attachments 4 (5, 2, and 3). Attached NOTE: 1-CH-MOV-1381 does not close automatically and will be closed manually per page 6 of attachment 5.

Crew verifies SI flow (HHSI and LHSI).

(This step will also cover starting LHSI pumps.)

BOP verifies AFW flow.

Flow to all SG indicated Total flow > 340 gpm.

RO checks RCS Tave stable at or trending to desired temperature. (NO)

BOP adjusts AFW flow.

RO checks PRZR PORVs and spray valves closed and block valves at least one open.

RO checks RCS subcooling <25F [85F]. [ ] adverse number Crew checks at least one charging pump running and flowing to the RCS.

RO stops all RCPs (previously done by CAP).

RO checks RCS pressure < 1275 psig

[1475 psig] and verifies RCPs stopped.

RO verifies charging pump recirc valves are closed.

BOP checks SGs not faulted. (YES) Decision making step BOP checks SGs not ruptured. (YES) Decision making step Crew checks if RCS is intact inside Decision making step containment. (NO) 2016 NRC 2 Page 20 Revision 1

EVENT 7: Given that the unit is at power, and a large break LOCA has occurred inside containment, the crew will be expected to respond in accordance with 1-E-0, "Reactor Trip or Safety Injection," and 1-E-1, "Loss of Secondary or Reactor Coolant."

TIME EXPECTED ACTION INSTRUCTOR REMARKS US directs crew to transition to 1-E-1.

Crew begins monitoring Critical Safety Function Status Trees.

US directs crew to transition to 1-FR-P.1 FR-P.1 (when applicable).

RO check RCS pressure greater than 225

[450] psig. (NO)

BOP verifies either low-head flow greater than 1000 gpm. (YES)

US directs crew to transition to 1-E-1.

RO checks PRZR PORVs and spray valves closed and block valves at least one open.

RO checks RCS subcooling <25F [85F]/

Crew checks at least one charging pump running and flowing to the RCS.

RO stops all RCPs (previously done by CAP).

RO checks RCS pressure < 1275 psig

[1475 psig] and verifies RCPs stopped.

RO verifies charging pump recirc valves are closed.

Crew checks SGs not faulted.

All pressures > 80 psig and all under the con troll of theoperator. (YES)

• Crew checks intact SG levels >11%

[22%].

Crew checks secondary radiation.

RO resets SI, Phase A Crew verifies IA TV open (NO)

Crew resets Phase B Crew opens IA TVs Crew opens IA TVs RO resets AMSAC Crew checks radiation monitors Crew lines up for SG samples Crew calls chemistry.

• RO checks PRZR PORVs and block valves.

2016 NRC 2 Page 21 Revision 1

EVENT 7: Given that the unit is at power, and a large break LOCA has occurred inside containment, the crew will be expected to respond in accordance with 1-E-0, "Reactor Trip or Safety Injection," and 1-E-1, "Loss of Secondary or Reactor Coolant."

TIME EXPECTED ACTION INSTRUCTOR REMARKS

• Crew checks if SI can be terminated:

RCS subcooling > 25F [75F](NO)

• Crew checks if manual CDA is required:

QS pumps not running (NO)

• Crew checks QS pump status:

QS pumps - any running Both of the Following RWST level < 3% (NO)

QS pump amps fluctuating (NO)

• Crew checks if Casing Cooling pumps should be stopped:

Casing cooling tank level < 10% (NO)

• Crew checks if redundant RS pumps can be stopped:

RS pumps any running Containment pressure <13 psia If so:

SM approval Verify/reset CDA Crew secures one train of RS pumps Crew closes SW supply to RSHX isolation valves for pumps that were stopped

• Crew maintains containment pressure:

Checks RS sump level > 4' 10" If so:

Operates at least one train of RS to maintain containment pressure < 13 psia.

• Crew checks if LHSI pumps should be stopped due to RCS pressure >225 psig

[450 psig].(NO)

Crew checks if EDGs should be stopped:

AC emergency busses energized by off-site power (YES)

Initiates attachment 2 for stopping Attached EDGs.

2016 NRC 2 Page 22 Revision 1

EVENT 7: Given that the unit is at power, and a large break LOCA has occurred inside containment, the crew will be expected to respond in accordance with 1-E-0, "Reactor Trip or Safety Injection," and 1-E-1, "Loss of Secondary or Reactor Coolant."

TIME EXPECTED ACTION INSTRUCTOR REMARKS Crew checks recirculation capability: Reports on the MOV will come in once Verify at least one train of CL recirc is operators are dispatched. This will take available: a few minutes due to having to access 1-SI-P-1A (NO) the 2nd level of Safeguards.

1-SI-P-1B 1-SI-MOV-1860B (NO) has no lights Breaker 1J1-2N J3 will be checked and Try to restore at least one train of CL reported as tripped with acrid smell in recirc area If power not available: check if equipment can be manually operated.

(NO)

US announces transition to 1-ECA-1.1. Validation time: 27 minutes 2016 NRC 2 Page 23 Revision 1

EVENT 8: Given that a loss of emergency coolant recirc has occurred with a LBLOCA in progress, the crew will respond in accordance with 1-ECA-1.1, "Loss of Emergency Coolant Recirculation."

TIME EXPECTED ACTION INSTRUCTOR REMARKS SPD Verified: __________ (Initials)

None RO resets both trains of SI, if required.

(Already reset.)

BOP presses both SI RECIRC MODE RESET buttons.

Crew verifies emergency coolant recirculation equipment available:

LHSI pumps LHSI pump suctions from containment.

(NO)

Crew tries to restore one train of emergency coolant recirculation equipment.

CT6 Crew establishes one train of SI Critical task flow: *Prior to 8% RWST level Verifies only one charging pump running (NO)

Stops one charging pump Places all non-running charging pumps in PTL Verifies RCS pressure < 225 psig [450 psig]

Verifies only one LHSI pump running NOTE: Scenario may be stopped once a Validation time: 5 minutes charging pump has been stopped, with concurrence from the lead evaluator.

2016 NRC 2 Page 24 Revision 1

REFERENCES PROCEDURE REV.

Abnormal Procedure 1-AP-28, "Loss of Instrument Air." 36 Abnormal Procedure 1-AP-14, "Low Condenser Vacuum." 31 Abnormal Procedure 1-AP-3, "Loss of Vital Instrumentation." 28 Abnormal Procedure 1-AP-16, "Increasing Primary Plant Leakage." 30 Abnormal Procedure 1-AP-5, "Unit One Radiation Monitoring System." 39 Emergency Procedure 1-E-0, "Reactor Trip or Safety Injection." 49 Emergency Procedure 1-E-1, "Loss of Reactor or Secondary Coolant." 27 Functional Restoration Procedure 1-FR-C.2, "Response to Degraded Core 14 Cooling."

Functional Restoration Procedure 1-FR-P.1, "Response to Imminent Pressurized 20 Thermal Shock."

Emergency Contingency Action 1-ECA-1.1, "Loss of Emergency Coolant 20 Recirculation."

Station Annunciator Response Procedures. N/A Administrative Procedure PI-AA-5000, "Human Performance." 8 INPO, Guideline for Teamwork and Diagnostic Skill Development: INPO 88-003, Jan. 1988 INPO, ACAD 07-002 Simulator Training Guidelines Jan. 2007 2016 NRC 2 Page 25 Revision 1

ATTACHMENT 1 SIMULATOR OPERATOR'S COMPUTER PROGRAM 2016 NRC 2 Page 26 Revision 1

SIMULATOR OPERATOR'S COMPUTER PROGRAM 2016 NRC 2 Initial conditions

1. Recall IC 322

2. Ensure Tave, Tref, PDTT level, and VCT level are selected on trend recorders.

3. 2H is the protected train.

4. Place red stickers on 1-SW-P-1A and 1-BC-P-1B.

5. Have copy of AP-14 attachment 2, page 1 for booth.

PRELOADS PRIOR TO SCENARIO START CONDITION MALFUNCTION/OVERRIDE/ETC.

Tagout 1-SW-P-1A Place pump in PTL. Rack out breaker and close discharge valve.

Remote functions:

SWP1A_RACKIN = 0 SW_6 = 0 Tagout 1-BC-P-1B Place pump in PTL and Rack out breaker.

Remote function:

BCP1B_RACKIN = RACKOUT 1-SA-C-1 fails to start Switch override:

automatically SAC1_AUTO = OFF Auto start failure of EHC Switch override:

pump TMP4_ASTP = Off Failure of auto CDA Malfunction:

QS03 "A" LHSI pump sheared Malfunction:

shaft SI0901 Low head SI pumps fail to Switch override:

auto start SIP1A_AUTO = OFF SIP1B_AUTO = OFF 1-CH-MOV-1381 fails to Monitor:

close automatically but will CHMOV381_RATE = 0 work when close button pressed On trigger screen for trigger 22:

MOV381_C SET CHMOV381_RATE = 180 2016 NRC 2 Page 27 Revision 1

SCENARIO EVENTS EVENT 1 Instrument air compressor trips MALFUNCTIONS/OVERRIDES Remote function:

U2_SAC1_FAULT = TRUE, Delay time = 5, Trigger = 1 Malfunction:

CA0402, Delay time = 5, Severity = 15, Trigger = 1 Set up Trigger 30 to delete leak when IA compressor started SAC1_HAND DMF CA0402 The next event can occur once the crew has restored instrument air pressure, and with direction from the lead evaluator.

COMMUNICATIONS If sent to 2-SA-C-1: wait 5 minutes and then report a Motor Overload alarm.

(If electricians are sent can report a fault in compressor motor.) This has been reported to Unit 2 and WCC.

If called to check 2-IA-TV-211 closed, report that red light is NOT lit and dryer appears to be operating normally.

2016 NRC 2 Page 28 Revision 1

EVENT 2 Failure of 1-CH-TE-1144 MALFUNCTIONS/OVERRIDES Malfunction:

CH27, Delay time = 5, Severity = -1, Trigger = 2 COMMUNICATIONS If I&C sent to cabinet and/or AB operator sent to transmitter: wait 5 minutes and report back that you see nothing obviously wrong.

2016 NRC 2 Page 29 Revision 1

EVENT 3/3a/3b Loss of Condenser Vacuum/Unit ramp/Auto rod control fauls MALFUNCTIONS/OVERRIDES Malfunction:

CN0901, Delay time = 5, Ramp = 120, Severity = 100, Trigger = 3 RD14, Delay time = 5, Trigger = 3 When the vacuum leak is stopped and the crew has stabilized the unit, and with direction of the lead evaluator, the next event can occur.

COMMUNICATIONS If would take at least 5 minutes to get attachment and find this.

When sent to perform attachment, verify that enough of a ramp has occurred, then delete malfunction CN0901. (If no one dispatched with attachment in hand, than initially report the loop seal is hot and get direction to isolate.)

If enough of a ramp has NOT occurred, than first report back that the isolation valve, 1-VP-22, is difficult to turn and request permission to use a valve wrench (torque amplifying device)

You will slowly reopen in 15 minutes and report back any problems.

2016 NRC 2 Page 30 Revision 1

EVENT 4 Pressurizer level channel 1461 fails MALFUNCTIONS/OVERRIDES Malfunction:

RC0803, Delay time = 5, Ramp = 1, Severity = -1, Trigger = 4 The next event can occur after the crew identifies the applicable MOP and pressurizer level control is back in automatic, and as directed by the lead evaluator.

COMMUNICATIONS 2016 NRC 2 Page 31 Revision 1

EVENT 5/6 EHC pump trips/RCS leak MALFUNCTIONS/OVERRIDES Malfunction:

TU1101, Delay time =5, Trigger = 5 RC04, Delay time = 5, Ramp = 300, Severity = 10, Trigger = 5 The next event will occur after the crew either makes preparations to ramp, makes preparations for a containment entry, and as directed by the lead evaluator.

COMMUNICATIONS If the crew dispatches an operator to look at EHC pumps, wait 3 minutes and then report that 1-TM-P-3 is unusually hot, and 1-TM-P-4 appears normal. No EHC leaks and pressure is normal.

If consulted about decision to either ramp or make containment entry: ask what they recommend and agree with it.

2016 NRC 2 Page 32 Revision 1

EVENT 7 LBLOCA MALFUNCTIONS/OVERRIDES Malfunction:

RC0101, Delay time = 5, Ramp = 60, Severity = 30, Trigger = 7 MOV control:

SIMOV860B_RACKIN = RACKOUT, Delay time = 120, Trigger = 7 WILL GET A U-2 ALARM WHEN CDA ACTUATED. SILENCE IT ON EXTREMEVIEW.

Remote function:

CH_217 = 0, Delay time = 30, Ramp = 30, Trigger = 20 The next event occurs when the crew reaches the point in the procedure to check containment recirc capability COMMUNICATIONS If called: SM gives permission to stop redundant RS pumps.

If sent: 1-SI-P-1A has a sheared shaft.

If sent to breaker for 1-SI-MOV-1860B wait 3 minutes and then report that 1J1-2N J3 has tripped open. There is an acrid smell in the area.

If sent to check 1-SI-MOV-1860B, you will need HP coverage. Report that the actuator is broken on the MOV. Pieces of it are on the floor.

2016 NRC 2 Page 33 Revision 1

EVENT 8 Loss of containment Recirc MALFUNCTIONS/OVERRIDES Preloaded.

Scenario may be stopped once a charging pump has been stopped, with concurrence from the lead evaluator.

COMMUNICATIONS 2016 NRC 2 Page 34 Revision 1

ATTACHMENT 3 SCENARIO PERFORMANCE OBJECTIVES 2016 NRC 2 Page 35 Revision 1

SIMULATOR REQUALIFICATION EXAMINATION TERMINAL PERFORMANCE OBJECTIVE Given equipment failures and operational situations, operate the plant in accordance with Technical Specifications to bring the unit to a safe condition, using applicable procedures, and applying effective teamwork, communication, and diagnostic skills.

GENERIC PERFORMANCE OBJECTIVES A. During shift operations the shift manager will take a conservative course of action, especially when uncertain conditions exist, when dealing with core cooling or heat sink availability, primary system and containment integrity, and reactivity control associated with plant evolutions.

B. During shift operations the shift manager will provide overall crew guidance by prioritizing and integrating the actions of the shift crew in accordance with administrative procedures.

C. During shift operations each crew member will participate in a team effort that resolves conflicts, provides input into the team decision and communicates all the necessary information to enhance teamwork in accordance with administrative procedures.

D. During shift operations the Shift Technical Advisor will independently assess events and based on those assessments make recommendations to the crew regarding mitigation strategy.

E. During shift operations each crew member will utilize operator fundamentals to ensure Teamwork Effectiveness, High Standards for Controlling Evolutions, Indications Monitored Closely, a Natural Bias for Conservatism, and Knowledge of Plant Design and Theory.

2016 NRC 2 Page 36 Revision 1

EVENT 1 PERFORMANCE OBJECTIVES EVENT GOAL: Given that the unit is at power, and a loss of an air compressor (2-SA-C-1) has occurred, the crew will be expected to respond in accordance with 1-AP-28, "Loss of Instrument Air."

NORTH ANNA SPECIFIC TASKS:

R530 Respond to a loss of instrument air outside of containment.

CRITICAL TASK:

See next page 2016 NRC 2 Page 37 Revision 1

CT Statement:

Crew starts all available air compressors.

Safety Significance:

Failure to start all available air compressors under the postulated plant conditions constitutes mis-operation or incorrect crew performance which leads to degradation of plant conditions which could result in a unit trip and/or safety injection. In this case, the instrument air pressure can be maintained above the trip set point by starting the air compressors.

Therefore, failure to start the air compressors also represents a "demonstrated inability by the crew to take an action or combination of actions that would prevent a challenge to plant safety."

Cues:

Instrument air low pressure alarm.

Meter indication of low instrument air pressure.

Performance Indicator:

BOP starts all available air compressors.

Feedback:

Instrument air pressure stabilizes above the trip set point.

WOG

Reference:

None.

Conditions:

Prior to reaching the trip set point of 70 PSIG.

2016 NRC 2 Page 38 Revision 1

EVENT 2 PERFORMANCE OBJECTIVES EVENT GOAL: Given that the unit is at power and 1-CH-TE-1144 has failed low causing letdown temperature to increase, the crew will be expected to respond in accordance with the AR for C-C6, DEMIN INLET DIVERT HI TEMP.

NORTH ANNA SPECIFIC TASKS:

None CRITICAL TASK:

N/A 2016 NRC 2 Page 39 Revision 1

EVENT 3 PERFORMANCE OBJECTIVES EVENT GOAL: Given that the unit is at power and loss of condenser vacuum is occurring, the crew will respond in accordance with 1-AP-14, "Low Condenser Vacuum."

NORTH ANNA SPECIFIC TASKS:

R518 Respond to a partial loss of condenser vacuum.

CRITICAL TASK:

N/A 2016 NRC 2 Page 40 Revision 1

EVENT4 PERFORMANCE OBJECTIVES EVENT GOAL: Given that the unit is at power and a selected pressurizer level channel, 1-RC-LI-1461, has failed low, the crew will respond in accordance with 1-AP-3, "Loss of Vital Instrumentation."

NORTH ANNA SPECIFIC TASKS:

R633 Respond to a failure of the controlling pressurizer level channel.

S70 Evaluate compliance with technical specifications.

CRITICAL TASK:

See next page 2016 NRC 2 Page 41 Revision 1

CT Statement:

Crew takes manual control of PRZR level, minimizes charging, and restores letdown.

Safety Significance:

Failure to take manual control of PRZR level constitutes a "mis-operation or incorrect crew performance" which will result in an unnecessary reactor trip on high PRZR level.

Cues:

The controlling PRZR level channel is failed low, letdown is isolated.

Performance Indicator:

RO verifies/places controller for 1-CH-FCV-1122 in manual and controls PRZR level.

RO performs RNO step and selects operable channel.

Crew restores charging and letdown as directed by the US.

Feedback:

PRZR level does not exceed trip setpoint.

WOG

Reference:

None.

Conditions:

Prior to a PRZR high level reactor trip.

2016 NRC 2 Page 42 Revision 1

EVENT 5/6 PERFORMANCE OBJECTIVES EVENT GOAL: Given that the unit is at power and there are no EHC pumps running, the crew will respond in accordance with the applicable AR. Also, when indications exist of RCS leakage greater than TS limits, the crew will be expected to respond in accordance with 1-AP-16, "Excessive Primary Plant Leakage," and 1-AP-5, "Unit 1 Radiation Monitoring System.

NORTH ANNA SPECIFIC TASKS:

R520 Respond to increasing primary plant leakage.

CRITICAL TASK:

See next page 2016 NRC 2 Page 43 Revision 1

CT Statement:

Crew starts the standby EHC pump Safety Significance:

Failure to maintain/recover EHC pressure will cause an unnecessary turbine trip/reactor trip that could have been prevented by starting the standby pump.

Cues:

Annunciators (Turb superv panel trouble, EH fluid reservoir lo pressure)

Indications that there is EHC pump running Performance Indicator:

Crew manually starts standby EHC pump Feedback:

EH fluid reservoir low pressure alarm clears WOG

Reference:

None Conditions:

Prior to receiving an automatic reactor trip due to a turbine trip caused by low EHC pressure 2016 NRC 2 Page 44 Revision 1

EVENT 7 PERFORMANCE OBJECTIVES EVENT GOAL: Given that the unit is at power, and a large break LOCA has occurred inside containment, the crew will be expected to respond in accordance with 1-E-0, "Reactor Trip or Safety Injection," and 1-E-1, "Loss of Secondary or Reactor Coolant."

NORTH ANNA SPECIFIC TASKS:

R185 Perform the immediate operator actions in response to a reactor trip or safety injection.

S69 Identify a reportable occurrence and make appropriate notifications.

S85 Notify the appropriate personnel of emergency events.

CRITICAL TASK:

See Following Pages 2016 NRC 2 Page 45 Revision 1

CT Statement:

Crew starts at least one LHSI pump.

Safety Significance:

Failure to manually start at least one low-head ECCS pump under the postulated conditions constitutes "mis-operation or incorrect crew performance which leads to degraded ECCS...capacity." In this case, at least one low-head ECCS pump can be manually started from the control room. Therefore, failure to manually start a low-head ECCS pump also represents a "demonstrated inability by the crew to:

• Recognize a failure/incorrect auto actuation of an ESF system or component

• Effectively direct/manipulate ESF controls" Cues:

Indication and/or annunciation that low-head ECCS pumped injection is required

• SI actuation

• RCS pressure below the shutoff head of the low-head ECCS pumps and Indication and/or annunciation that no low-head ECCS pump is injecting into the core

• Control switch indication that the circuit breakers or contactors for both low-head ECCS pumps are open

• All low-head ECCS pump discharge pressure indicators read zero

• All flow rate indicators for low-head pumped injection read zero Performance Indicator:

BOP manually starts an available Low Head SI pump.

Feedback:

Indication and/or annunciation that at least one low-head ECCS pump is injecting Flow rate indication of injection from at least one low-head ECCS pump WOG

Reference:

Appendix B CT 5 Conditions:

Prior to entering 1-FR-C.1.

2016 NRC 2 Page 46 Revision 1

CT Statement:

Crew manually actuates CDA.

Safety Significance:

Failure to manually actuate CDA under the postulated conditions constitutes a "demonstrated inability by the crew to recognize a failure of an ESF system or component."

Cues:

Indication/annunciation that containment pressure has exceed the CDA setpoint with indication that CDA did not automatically initiate.

Performance Indicator:

RO/BOP manually actuates CDA.

Feedback:

Indication/annunciation that CDA has actuated.

WOG

Reference:

Appendix B CT 3 Conditions:

Prior to transitioning to 1-FR-Z.1.

2016 NRC 2 Page 47 Revision 1

EVENT 8 PERFORMANCE OBJECTIVES EVENT GOAL: Given that a loss of emergency coolant recirc has occurred with a LBLOCA in progress, the crew will respond in accordance with 1-ECA-1.1, "Loss of Emergency Coolant Recirculation."

NORTH ANNA SPECIFIC TASKS:

None CRITICAL TASK:

See next page 2016 NRC 2 Page 48 Revision 1

CT Statement:

Crew establishes one train of SI flow.

Safety Significance:

Failure to stop the ECCS pumps taking a suction on the RWST before it empties results in cavitation, air binding, and loss of suction. All of these results can lead to pump damage sufficient to [reduce the availability of the pumps when a suction source subsequently becomes available].

Cues:

Procedurally directed by ECA-1.1.

Performance Indicator:

RO stops 1 HHSI pump.

Feedback:

Decreased RCS makeup flow.

Pump amps at zero.

WOG

Reference:

Appendix B CT 28 Conditions:

Prior to 8% RWST level 2016 NRC 2 Page 49 Revision 1

ATTACHMENT 2 SIMULATOR PERFORMANCE DATASHEET

Scenario Performance Datasheet EVENT 1: Given that the unit is at power, and a loss of an air compressor (2-SA-C-1) has occurred, the crew will be expected to respond in accordance with 1-AP-28, "Loss of Instrument Air."

SPD Verified: __________ (Initials) 2-SA-C-1 trips Annunciator J-D1, and possibly F-F8, J-E8 illuminate Instrument air pressure decreases 1-SA-C-1 does not start automatically EVENT 2: Given that the unit is at power and 1-CH-TE-1144 has failed low causing letdown temperature to increase, the crew will be expected to respond in accordance with the AR for C-C6, DEMIN INLET DIVERT HI TEMP.

SPD Verified: __________ (Initials) 1-CH-TI-1143 indication increases 1-CH-TI-1144 indication decreases 1-CC-TCV-106 output decreases Annunciator C-C6 illuminates 1-CH-TCV-1143 bypasses the demin train EVENT 3/3a/3b: Given that the unit is at power and loss of condenser vacuum is occurring, the crew will respond in accordance with 1-AP-14, "Low Condenser Vacuum."

SPD Verified: __________ (Initials)

Condenser vacuum degrades Annunciator A-G1 alarms if vacuum reaches setpoint Rods do not insert automatically to control temperature EVENT 4/4a: Given that the unit is at power and a selected pressurizer level channel, 1-RC-LI-1461, has failed low, the crew will respond in accordance with 1-AP-3, "Loss of Vital Instrumentation."

SPD Verified: __________ (Initials)

Annunciators B-F8, B-G7, and B-E2 are illuminated 1-RC-LI-1461 fails low Letdown isolates 2016 NRC 2 Date _________ Revision 1

EVENT 5/6: Given that the unit is at power and there are no EHC pumps running, the crew will respond in accordance with the applicable AR. Also, when indications exist of RCS leakage greater than TS limits, the crew will be expected to respond in accordance with 1-AP-16, "Excessive Primary Plant Leakage," and 1-AP-5, "Unit 1 Radiation Monitoring System.

SPD Verified: __________ (Initials)

Annunciators K-F5 and T-B3 will illuminate "A" EHC pump trips "B" EHC pump does not auto-start RCS leak Charging flow increasing, VCT level decreasing Containment sump level increasing Containment gaseous radiation (RM-160) increases (also RM-163 and 164)

Annunciators K-D2, K-D4, and possible J-F2 will illuminate.

EVENT 7: Given that the unit is at power, and a large break LOCA has occurred inside containment, the crew will be expected to respond in accordance with 1-E-0, "Reactor Trip or Safety Injection," and 1-E-1, "Loss of Secondary or Reactor Coolant."

SPD Verified: __________ (Initials)

Pressurizer pressure and level rapidly decreasing 1-SI-P-1A and 1-SI-P-1B fail to auto-start 1-SI-P-1A shears its shaft when started 1-CH-MOV-1381 does not close automatically EVENT 8: Given that a loss of emergency coolant recirc has occurred with a LBLOCA in progress, the crew will respond in accordance with 1-ECA-1.1, "Loss of Emergency Coolant Recirculation."

SPD Verified: __________ (Initials)

None 2016 NRC 2 Date _________ Revision 1

Appendix D Scenario Outline Form ES-D-1 Facility: North Anna Power Station Scenario No.: (2016) NRC-4 Op-Test No.: 1 Examiners: Bruno Cabarello______________ Operators: _____________________________

Newton Lacy________________ _____________________________

Gary Callaway_______________ _____________________________

Initial Conditions: 79% power, MOL. Power was held here due to a severe thunderstorm warning for the area. The warning has now been lifted. The unit is being returned to service after maintenance on the voltage regulator following a unit trip. Xenon is at equilibrium. 1-SW-P-1A is tagged out for major repairs.

1-BC-P-1B is tagged out for shaft replacement. 2H is the protected train.

Turnover: Ramp the unit to 100% power. Support maintenance on repair of 1-SW-P-1A and 1-BC-P-1B, as required.

Event Malf. Event Event No. No. Type* Description 1 R (RO) Ramp the unit up N (B) 2 I/C (RO) Median/Tave unit fails high (CT) 3 TS (S) IRPI (K2) fails low 4 I/C (B) "B" SG PORV opens unexpectedly. Can be closed from the control room 5 I/C (RO) Letdown leak, isolable from control room TS (S) 5a N (B) Place excess letdown in service 6 I/C (B) (Selected) "A" Ch. IV feed flow transmitter, 1-FW-FT-1476, fails low 7 M (All) Trip of normal feeder breaker for "A" station service bus (15A2)

(swaps to alternate power source 15A1)

Trip of "A" Main Feedwater pump with failure of standby pump ("B")

to start ATWS (CT) 8 I/C (RO) Control rods will not insert in auto or manual 9 I/C (B) Turbine stop valves will not close, must close MSTVs Events 8 and 9 are part of event 7 and are numbered for use on subsequent forms.

The scenario can be terminated once crew transitions back to 1-E-0.

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

DOMINION NORTH ANNA POWER STATION LICENSED OPERATOR REQUALIFICATION EXAMINATION SIMULATOR EXAMINATION GUIDE SCENARIO 2016 NRC 4

SIMULATOR EXAMINATION GUIDE EVENT DESCRIPTION

1. Ramp unit up

2. Median/Tave unit fails high

3. IRPI (K2) fails low

4. "B" SG PORV opens unexpectedly. Can be closed from the control room

5. Letdown leak, isolable from control room 5a. Place excess letdown in service

6. (Selected) "A" Channel IV feed flow transmitter, 1-FW-FT-1476, fails low

7. "A" SS swap to alternate power/Trip of MFP/failure of standby pump to start/ATWS

8. Control rods will not work in automatic or manual.

9. Turbine stop valves will not close Scenario Recapitulation:

Malfunctions after EOP entry 3 ATWS, rods do not insert in auto or manual, turbine will not trip and stop valves will not close Total Malfunctions 8 IRPI fails low, Tave unit fails low, SG PORV opens, letdown leak, selected feed flow channel fails, swap of "A" SS to alternate/trip of main feed pump with failure of standby pump to start/ATWS, rods do not insert in auto or manual, turbine will not trip and stop valves will not close Abnormal Events 4 Tave unit fails low, SG PORV opens, letdown leak, selected feed flow channel fails Major Transients 1 Main feed pump trip with failure of standby pump to start/ATWS EOPs Entered 1 FR-S.1 EOP Contingencies 1 FR-S.1 Critical Tasks 2 SCENARIO DURATION 110 Minutes 2016 NRC 4 Page 2 Revision 1

SIMULATOR EXAMINATION SCENARIO

SUMMARY

SCENARIO 2016 NRC 4 The scenario begins with the unit at 79% power, MOL. Power was held here due to a severe thunderstorm warning for the area. The warning has now been lifted. The unit is being returned to service after maintenance on the voltage regulator following a unit trip. Xenon is at equilibrium. 1-SW-P-1A, Unit 1 "A" SW pump, is tagged out for major repairs. 1-BC-P-1B is tagged for shaft replacement, not expected back for several days. 2H is the protected train. Shift orders are to continue the unit ramp to 100% power.

The first event will be a ramp up in power. This event can be pre-briefed. Once enough of a ramp has been seen, the next event can occur.

The next event will be the failure of the median/select Tave unit high. The crew will be expected to respond IAW 1-AP-1.1, "Continuous Uncontrolled Rod Motion," and place rod control in MANUAL (CT). Also, crew should address annunciators B-A7, MEDIAN/HI TAVG

< > TREF DEVIATION, and B-A8, LOOP 1A-B-C TAVG DEVIATION, take manual control of charging flow, and place steam dumps in steam pressure mode. After these actions have been completed and plant conditions are stable, or as directed by the lead evaluator, the next event will occur.

At this time, the IRPI for rod K-2 in control bank "A" will drop to zero. The US will review technical specification 3.1.7 and notify the instrument shop. Once this failure has been addressed, the next event can occur.

The next failure to occur will be the "B" SG PORV failing open due to the failure of the E/P. The crew may reduce power per 1-AP-38, "Excessive Load Increase." They will close the PORV using the controller and stabilize the unit. The next event will occur after the crew has stabilized the unit, and at the direction of the lead evaluator.

Next, there will be a leak on the letdown line in the Auxiliary building. The crew will enter 1-AP-16, "Increasing Primary Plant Leakage," and isolate the leak. They will place excess letdown in service using 1-OP-8.5, "Operation of Excess Letdown." The US will review Tech Specs for primary plant leakage.

The selected feed flow channel (1476) on "A" steam generator will fail low. The crew should respond in accordance with 1-AP-3, "Loss of Vital Instrumentation," and place the "A" steam generator level control in manual to restore normal operating level. The crew will swap steam generator water level control channels to channel III. Once the crew has identified the MOP, and with direction of the lead evaluator, the next event can occur.

Breaker 15A2, "A" station service bus normal supply breaker, will open unexpectedly. Breaker 15A1, RSST supply to station service will close in to supply power to the "A" station service bus.

The crew should notify the Electrical Department to investigate the fault. At this time the "A" Main Feed pump will trip and the standby pump will not auto-start. The crew will attempt to trip the reactor in accordance with 1-E-0, but the reactor will not trip. The crew will enter 1-FR-S.1, "Response to Nuclear Power Generation/ATWS." Rods will not insert in auto or manual and the 2016 NRC 4 Page 3 Revision 1

turbine stop valves will not close. The crew will close the MSTVs (CT) and inject the BIT to have sufficient negative reactivity addition (CT). At this time the reactor will be tripped locally. The crew will transition back to 1-E-0, "Reactor Trip or Safety Injection," and perform the immediate actions. At this time the scenario may be terminated with the direction of the lead evaluator.

2016 NRC 4 Page 4 Revision 1

SCENARIO TURNOVER SHEET Read the following to the crew:

Purpose:

This examination is intended to evaluate the crews performance of various tasks associated with the Initial License Operator Training Program. All activities should be completed in accordance with approved operations standards.

1. You are on a day shift during the week.

2. A rough log should be maintained to aid in making reports and to help during briefs.

3. Respond to what you see. In the unlikely event that the simulator fails such that illogical indications result, the session will be terminated and the crew informed.

Unit Status:

Unit 1 is at approximately 79% power by calorimetric. Power was held here due to a severe thunderstorm warning for the area. The warning has now been lifted. The unit is being returned to service after maintenance on the voltage regulator following a unit trip. Xenon is at equilibrium.

RCS boron is 1171 ppm and core age is 9,000 MWD/MTU. Aux steam is on unit 2.

Unit 2 is at 100% power.

Equipment Status:

1-SW-P-1A is tagged out for major repairs. 1-BC-P-1B is tagged for shaft replacement, not expected back for several days. 2H is the protected train. Maintenance rule window is green.

1-BC-P-1A and both Spent Fuel Pit Cooling pumps are protected.

Shift Orders:

Shift orders are to continue the unit ramp to 100% power using auto rod control. Assist maintenance with work on 1-SW-P-1A and 1-BC-P-1B, as required.

2016 NRC 4 Page 5 Revision 1

EVENT 1: Given that the unit is at approximately 79% power and the crew has been instructed to increase power, the crew will ramp the unit up in accordance with 1-OP-2.1, "Unit Startup from Mode 2 to Mode 1."

TIME EXPECTED ACTION INSTRUCTOR REMARKS SPD Verified: __________ (Initials)

Reactor power increases Turbine power increases Tavg/Tref increase Generator megawatts increase NOTE: The crew may raise primary Copy will be attached temperature prior to ramping the turbine. Turbine operation is done using attachment 9 (Guidance for main turbine operations) of 1-OP-2.1.

BOP verifies turbine in Auto.

BOP verifies/sets desired ramp rate (0.3%

per minute).

BOP adjusts limiter position, as required.

BOP increases turbine setter to desired position. (.3%/min)

BOP presses GO on turbine.

BOP monitors turbine ramp.

RO starts a dilution when required using 1-GOP-8.3.1 or 1-GOP-8.3.2.

RO monitors control rods to maintain Tave Rods were directed to be operated in within 1.5F of Tref with rods above automatic per pre-brief.

insertion limits.

NOTE: The next event can occur once Validation time: 27 minutes the crew has ramped a sufficient amount, and as directed by the lead evaluator.

2016 NRC 4 Page 6 Revision 1

EVENT 2: Given that the unit is operating at power, and the Tave MSS unit has failed high, the crew will be expected to respond in accordance with 1-AP-1.1, "Continuous Uncontrolled Rod Motion."

TIME EXPECTED ACTION INSTRUCTOR REMARKS SPD Verified: __________ (Initials)

Rods stepping in at maximum speed Annunciators B-A7 and B-A8 illuminate Annunciators B-F8 and C-C5 may also illuminate.

RO identifies rods stepping in at maximum speed.

RO identifies median/select Tave unit failed.

NOTE: Crew may reference 1-AP-3 US directs crew to enter AP-1.1.

US directs crew to hold the ramp.

CT1 Crew takes action to stop rod Critical task motion and stabilize the unit: *Prior to reactor trip on PRZR low RO places rod control in MANUAL pressure RO verifies rod motion stopped.

RO verifies 1-RC-TI-1408A Median/Hi MEDIAN/HI TAVG <> TREF Tavg is normal. (NO) DEVIATION Crew initiates actions of annunciator B-A7 See next page for actions while continuing with the AP.

• RO maintains the following using control *Continuous action rods and boration:

Rod bank LO/LOLO limit annunciators NOT lit AFD monitor annunciator NOT lit RO checks RCS Tave > 541F, above min and below max of attachment 2. Adjusts as directed by the US.

RO will need to find a method of monitoring Tavg other than the benchboard indication. (Can use indications on VB or PCS.)

Crew adjusts temperature, as required, using control rods/boration, or turbine load.

RO checks PRZR pressure stable or trending to 2235 psig. Adjusts heaters and spray as required.

2016 NRC 4 Page 7 Revision 1

EVENT 2: Given that the unit is operating at power, and the Tave MSS unit has failed high, the crew will be expected to respond in accordance with 1-AP-1.1, "Continuous Uncontrolled Rod Motion."

TIME EXPECTED ACTION INSTRUCTOR REMARKS RO checks PRZR level stable, and adjusts charging flow as required.

Crew checks controls rods above the LO insertion limit.

Crew maintains stable plant conditions.

US requests Work Control Center supervisor to inform the OMOC of the failure and to notify Instrument Department to investigate the cause of the failure. Requests initiation of CR.

NOTE: The following actions are IAW MEDIAN/HI TAVG <> TREF AR for annunciators B-A7. DEVIATION.

Attached Crew verifies 1-RC-TI-1408A is normal.

(NO)

Crew verifies control rods are in manual.

Crew places 1-CH-FCV-1122 in MANUAL and controls PRZR level at program.

RO transfers condenser steam dumps to steam pressure mode:

Places both interlock switches to OFF/RESET Places steam dump controller to MANUAL Places mode selector switch to steam pressure Ensures steam dump demand is zero Places steam dump controller to AUTO Verifies steam dump demand is zero Places both interlock switches to ON.

Crew verifies unit conditions are stable or under control:

Reactor power Secondary load Steam Generator levels.

NOTE: The next event can occur after the Validation time: 14 minutes crew stabilizes the unit, or as directed by the lead evaluator.

2016 NRC 4 Page 8 Revision 1

EVENT 3: Given that the unit is at power and indications of failed IRPI (K2) exist, the crew will respond in accordance with annunciator response A-G2 and technical specifications.

TIME EXPECTED ACTION INSTRUCTOR REMARKS SPD Verified: __________ (Initials)

IRPI for rod K-2 in Control Bank A will drop to zero Rod Bottom Light for K-2 will illuminate Annuciators A-G2 and A-F1 will illuminate RO identifies annunciators A-G2, RPI ROD BOT ROD DROP, and A-F1, CMPTR ALARM ROD DEV/SEQ.

NOTE: If unsure of conditions, the crew may enter 1-AP-1.2 for a dropped rod.

Steps in this procedure were not included.

RO identifies IRPI K-2 in control bank "A" is reading zero.

RO checks for other indications of a dropped rod.

RO identifies that no rod has dropped, IRPI problem.

Crew notifies I&C of IRPI problem.

NOTE: If I&C is asked to investigate the K-2 rod, they will report that it is an IRPI problem.

US reviews Technical Specification 3.1.7A and determines that a flux map must be done within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (or power reduced to <50%).

NOTE: The next event can occur once Validation time: 8 minutes the IRPI failure has been addressed, or at the direction of the lead examiner.

2016 NRC 4 Page 9 Revision 1

EVENT 4: Given that the unit is at power and "B" SG PORV has failed open, the crew will respond in accordance with 1-AP-38, "Excessive Load Increase."

TIME EXPECTED ACTION INSTRUCTOR REMARKS SPD Verified: __________ (Initials)

Reactor power slowly increases.

Megawatts slowly decrease PCS alarm PCV-MS101B Low Flow is received.

RO/BOP identifies increase in reactor power/decrease in MW or PCS alarm on MS-PCV-101B.

US directs crew to enter 1-AP-38.

RO verifies all steam dumps closed.

NOTE: Crew may identify "B" PORV open from PCS alarm and/or PNID screen.

BOP verifies all SG PORVs indicate closed. (NO)

BOP places SG PORV in manual and closes it.

BOP verifies turbine load normal.

RO verifies reactor power is less than or equal to 100% power and stable.

NOTE: If crew decides that power is not stable they may ramp the unit down slightly. These steps are not included.

RO verifies proper rod control (NO, rods are in manual)

RO controls Tave within 1.5F of Tref.

RO energizes additional PRZR heaters to maintain pressure. (Already energized)

• Crew checks turbine load control:

Reactor power reduced to the power level before the event occurred Valve Position Limit light - Off Turbine in operator auto Turbine in IMP-IN.

• RO maintains rod bank LO/LOLO limit annunciators NOT lit and AFD annunciator NOT lit.

2016 NRC 4 Page 10 Revision 1

EVENT 4: Given that the unit is at power and "B" SG PORV has failed open, the crew will respond in accordance with 1-AP-38, "Excessive Load Increase."

TIME EXPECTED ACTION INSTRUCTOR REMARKS Crew checks plant status is stable:

Main generator output stable Tave on program with Tref.

BOP checks steam flow channel indications are normal.

BOP checks turbine is in operator auto.

Crew checks plant steam systems:

SG PORVs SG safeties MSR inlet FCVs AS PCV.

Crew checks for RCS dilution due to improper demin operation, 1-CC-TCV-106 operation, or PG water leakby.

Crew verifies cause of load increase has been corrected.

Crew checks steam dumps in OFF. (NO)

Crew enters a CR to document the reactivity management event.

US consults Technical Specifications:

3.7.4 Condition A and determines that the PORV is operable.

NOTE: The next event can occur after Validation time: 8 minutes the crew has stabilized the unit, and at the direction of the lead evaluator.

2016 NRC 4 Page 11 Revision 1

EVENT 5/5a: Given that the unit is at power, and indications exist of a letdown leak outside containment, the crew will be expected to respond in accordance with 1-AP-16, "Increasing Primary Plant Leakage."

TIME EXPECTED ACTION INSTRUCTOR REMARKS SPD Verified: __________ (Initials)

Letdown flow decreases on 1-CH-FI-1150 VCT level steadily decreases Auxiliary building sump level steadily increases PCS alarm Non-regen Hx Letdown out F is received RO identifies decrease in letdown flow.

RO identifies decrease in VCT level.

BOP identifies increase in Aux building sump level.

Crew directs an operator to walkdown the Auxiliary Building and look for primary leaks.

US directs entry into 1-AP-16.

Crew verifies unit in mode 1.

• RO verifies PRZR level stable or Already in manual increasing. If not, the RO adjusts charging flow to control PRZR level.

• RO verifies VCT level stable. If not, the crew starts a makeup from the blender as required.

RO checks LCV-1115A not diverted.

NOTE: If sent, operator will report that a leak exists on the letdown piping in the Auxiliary Building penetration area. If the leak is already isolated, he will report that the floor is wet back at the letdown penetration.

RO verifies letdown in service with normal indications. (NO)

RO isolates letdown by closing 1-CH-HCV-1200B and 1-CH-LCV-1460A and B Minimizes/isolates charging.

Crew verifies that the leak stopped.

2016 NRC 4 Page 12 Revision 1

EVENT 5/5a: Given that the unit is at power, and indications exist of a letdown leak outside containment, the crew will be expected to respond in accordance with 1-AP-16, "Increasing Primary Plant Leakage."

TIME EXPECTED ACTION INSTRUCTOR REMARKS NOTE: Crew will initiate 1-OP-8.5 while continuing with 1-AP-16.

5a Crew places excess letdown in service Normal for BOP (or possibly RO) using 1-OP-8.5:

Verifies initial conditions Reviews Ps&Ls Checks calorimetric power, VCT level, Seal injections flows, PRZR heater status, CC flow to excess L/D HX Closes 1-CH-HCV-1137 Checks 1-CH-MOV 1380 and 1381 are open Has operator energize loop drains Places 1-CH-HCV-1389 in the PDTT position to flush the heat exchanger Deletes F0134A point from processing, as necessary Opens a loop drain valve and verifies the other 2 loop drains are closed Opens 1-CH-HCV-1201 Slowly opens 1-CH-HCV-1137 Adjusts 1137 as required to maintain temperatures and pressure When PDTT level has increased at least 14% - throttles 1-CH-HCV-1137 closed and places 1-CH-HCV-1389 in VCT position Maintains pressurizer level Logs appropriate parameters Places 1-CH-PCV-1145 in manual Verifies letdown is isolated.

Closes 1-CH-FCV-1122 in manual (already done)

Fully opens 1-CH-PCV-1145 Monitors accumulator levels.

Makes notifications..

US directs WCC to make notifications and initiate WR and CR.

2016 NRC 4 Page 13 Revision 1

EVENT 5/5a: Given that the unit is at power, and indications exist of a letdown leak outside containment, the crew will be expected to respond in accordance with 1-AP-16, "Increasing Primary Plant Leakage."

TIME EXPECTED ACTION INSTRUCTOR REMARKS US reviews RCS leakage Tech Spec 3.4.13A Post exam comments: deleted TS (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />); applicable until letdown was 3.6.1A (ML ML16214A301) isolated and the leak stopped.

Also TS 3.6.1A for containment being inoperable. Restore within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

May Also reference 3.6.3A for containment isolation valves. This one is 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Crew initiates leak rate PT.

NOTE: Once excess letdown is in Validation time: 26 minutes service, the next event can occur.

2016 NRC 4 Page 14 Revision 1

EVENT 6: Given that the unit is at power and "A" SG selected feed flow transmitter, 1-FW-FT-1476, has failed, the crew will be expected to respond in accordance with 1-AP-3, "Loss of Vital Instrumentation."

TIME EXPECTED ACTION INSTRUCTOR REMARKS SPD Verified: __________ (Initials)

Annunciators F-E1, F-D1 and possibly F-F1 illuminate "A" MFRV ramps open "A" SG Channel IV feed flow is off-scale low BOP identifies annunciator F-E1, SG 1A FW > STM FLOW CH III - IV.

BOP identifies "A" SG feedwater flow channel IV has failed low.

US directs crew to perform immediate actions of 1-AP-3:

Crew verifies SG level controlling channels normal (NO)

BOP places "A" MFRV in MANUAL and adjusts to control "A" SG level Crew verifies turbine first-stage pressure channels normal RO verifies PRZR level indications are normal.

Crew verifies redundant instrument channels normal Crew verifies first stage pressure indications AP-3 continues normal.

RO verifies pressurizer level indications are normal.

RO verifies systems affected by pressurizer level channels are normal:

RO verifies operable pressurizer level channels are selected RO verifies emergency bus backup heaters are restored Ro verifies letdown is in service RO verifies pressurizer level control is in automatic RO verifies pressurizer control group heaters are not tripped.

Crew verifies both first stage pressure channels normal.

Crew verifies all SGWLC channels selected to an operable channel. (NO) 2016 NRC 4 Page 15 Revision 1

EVENT 6: Given that the unit is at power and "A" SG selected feed flow transmitter, 1-FW-FT-1476, has failed, the crew will be expected to respond in accordance with 1-AP-3, "Loss of Vital Instrumentation."

TIME EXPECTED ACTION INSTRUCTOR REMARKS Crew swaps SGWLC channels:

RO verifies rod control to MANUAL RO reports steam dumps in steam pressure mode BOP places all MFRVs in MANUAL and verifies FRV B/Ps in MANUAL Crew swaps all Steam flow/Feed flow Seven switches to blue dot and First-stage pressure channels to channel III BOP verifies median levels are functional BOP returns MFRVs to AUTOMATIC RO leaves steam dumps in steam pressure mode RO leaves rod control in manual.

Crew checks for other failed channels.

Crew enters 1-MOP-55.78.

NOTE: The next event may occur after Validation time: 17 minutes the crew has identified the MOP and as directed by the lead evaluator.

2016 NRC 4 Page 16 Revision 1

EVENT 7: Given that the unit is at power and a loss of the normal feeder for "A" station service bus and then a loss of two feedwater pumps has occurred and the reactor has will not trip, the crew will respond in accordance with 1-AP-31,"Loss of Main Feedwater," 1-FR-S.1, "Response to Nuclear Power Generation/ATWS," and 1-E-0, "Reactor Trip or Safety Injection."

TIME EXPECTED ACTION INSTRUCTOR REMARKS SPD Verified: __________ (Initials)

"A" SS bus normal feeder breaker opens "A" MFP trips and "B" MFP does not auto-start Reactor will not trip automatically or manually Rods will not step in auto or manual Turbine does not trip or runback Crew identifies annunciator 1H-C8, SS BUSSES NOR SUP BKR AUTO TRIP.

Crew identifies breaker 15A2 has tripped 15A2 is the normal feeder breaker to open and breaker 15A1 has automatically "A" SS bus closed to supply the "A" station service bus 15A1 is the alternate feed from "A" via the RSST. RSST NOTE: Feed pump will trip 85 seconds later. This is unexpected.

Crew identifies that "A" FW pump has tripped with no auto-start of "B" FW pump.

US directs crew to enter 1-AP-31.

Crew determines that power is >70% with only 1 MFP running.

US directs crew to enter 1-E-0.

RO/BOP attempt to trip the reactor.

2016 NRC 4 Page 17 Revision 1

EVENT 7: Given that the unit is at power and a loss of the normal feeder for "A" station service bus and then a loss of two feedwater pumps has occurred and the reactor has will not trip, the crew will respond in accordance with 1-AP-31,"Loss of Main Feedwater," 1-FR-S.1, "Response to Nuclear Power Generation/ATWS," and 1-E-0, "Reactor Trip or Safety Injection."

TIME EXPECTED ACTION INSTRUCTOR REMARKS Crew identifies reactor did not trip, transitions to 1-FR-S.1, Nuclear Power Generation/ATWS, and takes actions to bring the reactor subcritical.

RO/BOP attempt to manually trip the reactor.

RO places rods in auto and determines they are not inserting RO attempts to manually insert control rods. (NO)

BOP attempts to manually trip the turbine (NO)

BOP attempts to runback the turbine to close stop valves (NO)

BOP closes the MSTVs BOP verifies/starts all AFW pumps running.

RO verifies at least one charging pump running.

RO places in-service boric acid transfer pump in fast speed.

RO opens emergency borate valve 1-CH-MOV-1350 RO verifies adequate negative reactivity insertion. (NO)

NOTE: Injecting the BIT or locally tripping the reactor is part of the same CT as above. Trip breakers will not be opened until BIT is being injected unless directed by lead evaluator.

2016 NRC 4 Page 18 Revision 1

EVENT 7: Given that the unit is at power and a loss of the normal feeder for "A" station service bus and then a loss of two feedwater pumps has occurred and the reactor has will not trip, the crew will respond in accordance with 1-AP-31,"Loss of Main Feedwater," 1-FR-S.1, "Response to Nuclear Power Generation/ATWS," and 1-E-0, "Reactor Trip or Safety Injection."

TIME EXPECTED ACTION INSTRUCTOR REMARKS Crew manually injects BIT: Critical Task RO opens charging pump suction *Before returning to step 5 from step 15 valves from the RWST (i.e. not subcritical)

RO closes charging pump suction valves from the VCT BOP closes BIT recirc valves BOP opens at least one BIT outlet CT2 valve BOP opens at least one BIT inlet valve RO closes letdown valves RO closes normal charging line isolations RO checks pressurizer pressure less than 2335 PSIG.

CT2 Crew checks if reactor trip has occurred. Only critical if BIT not injected (NO) *Before returning to step 5 from step 15 Crew dispatches operator to trip reactor locally.

Reactor trip breakers are opened locally NOTE: Reactor will be tripped after crew starts injecting the BIT.

NOTE: Crew may choose to perform Crew may have chosen to remove attachment for isolating the BIT and excess letdown from service. If so, they restoring charging and letdown since will be required to place it back in they have emergency boration flow. If service per this attachment.

so let them at least isolate the BIT before continuing.

RO verifies reactor subcritical.

US directs transition to 1-E-0.

RO verifies reactor tripped.

BOP verifies turbine tripped.

RO verifies emergency busses energized.

Crew verifies SI is not actuated or required.

2016 NRC 4 Page 19 Revision 1

EVENT 7: Given that the unit is at power and a loss of the normal feeder for "A" station service bus and then a loss of two feedwater pumps has occurred and the reactor has will not trip, the crew will respond in accordance with 1-AP-31,"Loss of Main Feedwater," 1-FR-S.1, "Response to Nuclear Power Generation/ATWS," and 1-E-0, "Reactor Trip or Safety Injection."

TIME EXPECTED ACTION INSTRUCTOR REMARKS US directs transition to 1-ES-0.1.

NOTE: Scenario can be terminated at Validation time: 10 minutes this time with direction of the lead evaluator.

2016 NRC 4 Page 20 Revision 1

REFERENCES PROCEDURE REV.

Operating Procedure 1-OP-2.1, "Unit Startup from Mode 2 to Mode 1."

Abnormal Procedure 1-AP-1.1, "Continuous Uncontrolled Rod Motion." 9 Abnormal Procedure 1-AP-38, "Excessive Load Increase." 19 Abnormal Procedure 1-AP-16, Increasing Primary Plant Leakage." 30 Operating Procedure 1-OP-8.5, "Operation of Excess Letdown." 23 Abnormal Procedure 1-AP-3, "Loss of Vital Instrumentation." 28 Emergency Procedure 1-E-0, "Reactor Trip or Safety Injection." 49 Functional Restoration Procedure1-FR-S.1, "Response to Nuclear Power 17 Generation/ATWS."

Emergency Procedure 1-ES-0.1 "Reactor Trip Response." 32 Station Annunciator Response Procedures. N/A Administrative Procedure PI-AA-5000, "Human Performance." 8 INPO, Guideline for Teamwork and Diagnostic Skill Development: INPO 88-003, Jan. 1988 INPO, ACAD 07-002 Simulator Training Guidelines Jan. 2007 2016 NRC 4 Page 21 Revision 1

ATTACHMENT 1 SIMULATOR OPERATOR'S COMPUTER PROGRAM 2016 NRC 4 Page 22 Revision 1

SIMULATOR OPERATOR'S COMPUTER PROGRAM 2016 NRC 4 Initial conditions

1. Recall IC 324

2. Ensure Tave (570-580), Tref, PDTT level, and VCT level are selected on trend recorders.

3. 2H is the protected train.

4. Place instrument channels in channel IV position.

5. Rods at 189 steps on D.

6. Calorimetric = 2

7. Verify 1-OP-8.5 copies in drawer PRELOADS PRIOR TO SCENARIO START CONDITION MALFUNCTION/OVERRIDE/ETC.

Tagout 1-SW-P-1A Place pump in PTL. Rack out breaker and close discharge valve.

Remote functions:

SWP1A_RACKIN = RACKOUT SW_6 = 0 Tagout 1-BC-P-1B Place pump in PTL and Rack out breaker.

Remote function:

BCP1B_RACKIN = RACKOUT "B" FW pump does not start Switch overrides:

in auto FWP1B1_ASTOP, Override = OFF FWP1B2_ASTOP, Override = OFF Auto trip failure/ Manual trip Malfunctions:

failure RD32 RD38 Remote function (RPS)

AMSAC_DEFEAT = T 2016 NRC 4 Page 23 Revision 1

CONDITION MALFUNCTION/OVERRIDE/ETC.

Failure of auto/manual Malfunctions:

turbine trip/failure of stop TU03 valves to runback TU02 Switch Overrides:

GV_FAST = OFF, Trigger = 7 GV_DOWN = OFF, Trigger = 7 VV_POS_LIM_DOWN = OFF, Trigger = 7 2016 NRC 4 Page 24 Revision 1

SCENARIO EVENTS EVENT 1 Unit ramp MALFUNCTIONS/OVERRIDES COMMUNICATIONS Answer calls to SOC, MOC, chemistry, etc, as required.

2016 NRC 4 Page 25 Revision 1

EVENT 2 Tave unit failure MALFUNCTIONS/OVERRIDES Malfunction:

RC1501, Delay time = 5, Ramp = 10, Severity = 1, Trigger = 2 The next event will occur after the crew stabilizes the unit, and as directed by the lead evaluator.

COMMUNICATIONS If OMOC asked about withdrawing rods to match temperature: wait 3 minutes and then call back and give permission.

If asked any other way: Ask for recommendation. Call back after 3 minutes (if WCC) and concur with their recommendation.

2016 NRC 4 Page 26 Revision 1

EVENT 3 IRPI failure MALFUNCTIONS/OVERRIDES Malfunction:

RD0121, Delay time = 5, Severity = -1, Trigger = 3 The next event can occur once the IRPI failure has been addressed and at the direction of the lead examiner.

COMMUNICATIONS If I&C is contacted to check rod K-2, wait 10 minutes and then report back that it is an IRPI problem.

If I&C is contacted to tell them of the IRPI failure. Just acknowledge the communication.

2016 NRC 4 Page 27 Revision 1

EVENT 4 Failure of "B" SG PORV MALFUNCTIONS/OVERRIDES Controller override:

PCVMS101B, Analog value = 0, Trigger = 4 COMMUNICATIONS When sent to locally check PORVs and safeties: wait 5 minutes and report that there is no steam from any of them.

If sent to look at PORVS (before procedure step) then wait 3 minutes and report that there is no steam coming from the roof of the MSVH.

2016 NRC 4 Page 28 Revision 1

EVENT 5 Letdown leak MALFUNCTIONS/OVERRIDES Malfunction:

CH02, Delay time = 5, Ramp = 120, Severity = 100, Trigger = 5 Remote function:

HCV1557_ENERGIZE = T, Trigger = 20 The next event can occur once excess letdown is in service, and with the direction of the lead evaluator.

COMMUNICATIONS When sent to penetration area, wait 5 minutes and then give a cue for current plant conditions, if leak not isolated then water is leaking out, if isolated then floor is wet and pipes are dripping water.

Can later report that leak was Upstream of 1-CH-TV-1204B (or between containment wall and trip valve).

NOTE: Must be reported as upstream to get correct TS call.

When sent to close breaker for loop drains: base time on whether you were sent to stand by. Put in trigger 20 and then report back that 1-EP-CB-26B BKR 22 is closed.

When sent to get numbers for 1-OP-8.5: wait 5 minutes and then give the following values:

1-CH-TI-1136 (Seal Return HX Outlet) is 87 degrees 1-CC-FI-102 (Seal water HX CC outlet flow) 183 GPM (Note the above numbers are from a unit 1 procedure done 1/28/16 2016 NRC 4 Page 29 Revision 1

EVENT 6 "A" SG channel IV feed flow (1476) fails low MALFUNCTIONS/OVERRIDES Malfunctions:

FW1202, Delay time = 5, Ramp = 20, Severity = -1, Trigger = 6 After the crew has identified MOPs, consulted TS, and with direction of the lead evaluator, the next event may occur.

COMMUNICATIONS If dispatched to check MFRV operation due to failure and valve in manual, wait 5 minutes and report that all MFRVs look normal.

If dispatched to check RWST level transmitter, wait at least 5 minutes and then report that there is nothing obviously wrong at the transmitter.

2016 NRC 4 Page 30 Revision 1

EVENT 7 Loss of "A" SS bus normal feed/Loss of 2 MFW pumps/ATWS/Rod control failure MALFUNCTIONS/OVERRIDES ATWS is preloaded, failure of turbine to trip is preloaded, failure of turbine TVs to close is preloaded Remote function:

EL15A2_BKR = OPEN, Delay time = 5, Trigger = 7 Malfunctions:

FW2201, Delay time = 90, Trigger = 7 RD14, Delay time = 30, Trigger = 7 RD15, Delay time = 30, Trigger = 7 Use the following to trip the reactor. MAKE SURE CREW HAS STARTED INJECTING BIT (at least has BIT inlets and outlets open. Unless directed otherwise by the lead evaluator.

Remote functions:

SP_RTA_BKR = F, Delay time = 0, Trigger = 10 SP_RTB_BKR = F, Delay time = 2, Trigger = 10 Scenario can be terminated once the crew has announced transition to 1-ES-0.1 and as directed by the lead evaluator.

COMMUNICATIONS When sent to trip reactor locally: once the crew is injecting the BIT put in trigger 10, and then report back that you have opened all the breakers.

NOTE: If crew does not begin injecting the BIT in a timely manner, lead evaluator can direct the opening of the reactor trip breakers if he desires.

2016 NRC 4 Page 31 Revision 1

ATTACHMENT 3 SCENARIO PERFORMANCE OBJECTIVES 2016 NRC 4 Page 32 Revision 1

EVENT 1 PERFORMANCE OBJECTIVES EVENT GOAL: Given that the unit is at approximately 79% power and the crew has been instructed to increase power, the crew will ramp the unit up in accordance with 1-OP-2.1, "Unit Startup from Mode 2 to Mode 1."

NORTH ANNA SPECIFIC TASKS:

R705 Dilute the Reactor Coolant System using the blender.

CRITICAL TASK:

N/A 2016 NRC 4 Page 33 Revision 1

EVENT 2 PERFORMANCE OBJECTIVES EVENT GOAL: Given that the unit is operating at power, and the Tave MSS unit has failed high, the crew will be expected to respond in accordance with 1-AP-1.1, "Continuous Uncontrolled Rod Motion."

NORTH ANNA SPECIFIC TASKS:

R475 Perform the immediate operator actions in response to a continuous uncontrolled rod motion.

CRITICAL TASK:

See next page 2016 NRC 4 Page 34 Revision 1

CT Statement:

Crew takes action in accordance with AP-1.1, to stop rod motion and stabilize the unit.

Safety Significance:

Core reactivity is not under control of the operator due to the failed control channel. "It is expected that the operator will attempt to take manual actions to correct for anomalous conditions during power operation."

Cues:

Indication of a failed MMS Unit.

Continuous inward control rod motion with Tave and Tref matched.

Performance Indicator:

RO places rod control in manual.

Feedback:

Rod motion stops WOG

Reference:

None Conditions:

Prior to reactor trip on low pressurizer pressure.

2016 NRC 4 Page 35 Revision 1

EVENT 3 PERFORMANCE OBJECTIVES EVENT GOAL: Given that the unit is at power and indications of failed IRPI (K2) exist, the crew will respond in accordance with annunciator response A-G2 and technical specifications.

NORTH ANNA SPECIFIC TASKS:

S70 Evaluate compliance with technical specifications.

CRITICAL TASK:

N/A 2016 NRC 4 Page 36 Revision 1

EVENT 4 PERFORMANCE OBJECTIVES EVENT GOAL: Given that the unit is at power and "B" SG PORV has failed open, the crew will respond in accordance with 1-AP-38, "Excessive Load Increase."

NORTH ANNA SPECIFIC TASKS:

R539 Perform the immediate operator actions in response to an excessive load increase.

CRITICAL TASK:

N/A 2016 NRC 4 Page 37 Revision 1

EVENT 5 PERFORMANCE OBJECTIVES EVENT GOAL: Given that the unit is at power, and indications exist of a letdown leak outside containment, the crew will be expected to respond in accordance with 1-AP-16, "Increasing Primary Plant Leakage."

NORTH ANNA SPECIFIC TASKS:

R520 Respond to increasing primary plant leakage.

CRITICAL TASK:

N/A 2016 NRC 4 Page 38 Revision 1

EVENT 6 PERFORMANCE OBJECTIVES EVENT GOAL: Given that the unit is at power and "A" SG selected feed flow transmitter, 1-FW-FT-1476, has failed, the crew will be expected to respond in accordance with 1-AP-3, "Loss of Vital Instrumentation."

NORTH ANNA SPECIFIC TASKS:

R626 Respond to a steam generator water level control channel failure.

S70 Evaluate compliance with technical specifications.

CRITICAL TASK:

N/A 2016 NRC 4 Page 39 Revision 1

EVENT 7 PERFORMANCE OBJECTIVES EVENT GOAL: Given that the unit is at power and a loss of the normal feeder for "A" station service bus and then a loss of two feedwater pumps has occurred and the reactor has will not trip, the crew will respond in accordance with 1-AP-31,"Loss of Main Feedwater," 1-FR-S.1, "Response to Nuclear Power Generation/ATWS," and 1-E-0, "Reactor Trip or Safety Injection."

NORTH ANNA SPECIFIC TASKS:

R224 Perform immediate operator actions in response to a nuclear power generation/ATWS.

S69 Identify a reportable occurrence and make appropriate notifications.

S85 Notify the appropriate personnel of emergency events.

CRITICAL TASK:

See next Page 2016 NRC 4 Page 40 Revision 1

CT Statement:

Crew identifies reactor did not trip, transition to 1-FR-S.1, Nuclear Power Generation/ATWS, and take actions to bring the reactor subcritical.

Safety Significance:

Failure to insert negative reactivity under the postulated plant conditions results in an unnecessary situation in which the reactor remains critical. Failure to insert negative reactivity constitutes "mis-operation or incorrect crew performance which leads to incorrect reactivity control."

Cues:

Valid indication of a required reactor trip by the presence of a first out annunciator, with a failure of the reactor to trip automatically or manually from the control room.

Performance Indicator:

Crew opens valves to inject the BIT.

OR Reactor trip breakers are opened locally per crew direction Feedback:

BIT flow is indicated WOG

Reference:

Appendix B CT 52 Conditions:

Prior to returning to step 5 from step 15 (i.e. not subcritical).

2016 NRC 4 Page 41 Revision 1

ATTACHMENT 2 SIMULATOR PERFORMANCE DATASHEET

Scenario Performance Datasheet EVENT 1: Given that the unit is at approximately 79% power and the crew has been instructed to increase power, the crew will ramp the unit up in accordance with 1-OP-2.1, "Unit Startup from Mode 2 to Mode 1."

SPD Verified: __________ (Initials)

Reactor power increases Turbine power increases Tavg/Tref increase Generator megawatts increase EVENT 2: Given that the unit is operating at power, and the Tave MSS unit has failed high, the crew will be expected to respond in accordance with 1-AP-1.1, "Continuous Uncontrolled Rod Motion."

SPD Verified: __________ (Initials)

Rods stepping in at maximum speed Annunciators B-A7 and B-A8 illuminate Annunciators B-F8 and C-C5 may also illuminate.

EVENT 3: Given that the unit is at power and indications of failed IRPI (K2) exist, the crew will respond in accordance with annunciator response A-G2 and technical specifications.

SPD Verified: __________ (Initials)

IRPI for rod K-2 in Control Bank A will drop to zero Rod Bottom Light for K-2 will illuminate Annuciators A-G2 and A-F1 will illuminate EVENT 4: Given that the unit is at power and "B" SG PORV has failed open, the crew will respond in accordance with 1-AP-38, "Excessive Load Increase."

SPD Verified: __________ (Initials)

Reactor power slowly increases.

Megawatts slowly decrease PCS alarm PCV-MS101B Low Flow is received.

EVENT 5: Given that the unit is at power, and indications exist of a letdown leak outside containment, the crew will be expected to respond in accordance with 1-AP-16, "Increasing Primary Plant Leakage."

SPD Verified: __________ (Initials)

Letdown flow decreases on 1-CH-FI-1150 VCT level steadily decreases Auxiliary building sump level steadily increases PCS alarm Non-regen Hx Letdown out F is received 2016 NRC 4 Date _________ Revision 1

EVENT 6: Given that the unit is at power and "A" SG selected feed flow transmitter, 1-FW-FT-1476, has failed, the crew will be expected to respond in accordance with 1-AP-3, "Loss of Vital Instrumentation."

Annunciators F-E1, F-D1 and possibly F-F1 illuminate "A" MFRV ramps open "A" SG Channel IV feed flow is off-scale low EVENT 7: Given that the unit is at power and a loss of the normal feeder for "A" station service bus and then a loss of two feedwater pumps has occurred and the reactor has will not trip, the crew will respond in accordance with 1-AP-31,"Loss of Main Feedwater," 1-FR-S.1, "Response to Nuclear Power Generation/ATWS," and 1-E-0, "Reactor Trip or Safety Injection."

SPD Verified: __________ (Initials)

"A" SS bus normal feeder breaker opens "A" MFP trips and "B" MFP does not auto-start Reactor will not trip automatically or manually Rods will not step in auto or manual Turbine does not trip or runback 2016 NRC 4 Date _________ Revision 1