ML17172A692

Section 1 - Draft Administrative Documents

ML17172A692
Person / Time
Site:	North Anna
Issue date:	06/22/2017
From:	Virginia Electric & Power Co (VEPCO)
To:	NRC/RGN-II
References
Download: ML17172A692 (44)

Text

ES-301 Administrative Topics Outline Form ES-301-1 Facility: North Anna Power Station Date of Examination: 6/20/2016 Examination Level: Combined (See Below) Operating Test Number: 1 Administrative Topic Type Describe activity to be performed (see Note) Code*

Determine minimum RHR flow based on time after M R Conduct of Operations shutdown and determine minimum allowable level 2

• 125 based on the calculation (1-AP-11 Loss of RHR Flow)

(ALL) (RO 3.9 I SRO 4.2)

Determine Quadrant Power Tilt Ratio by hand

. D,R calculation(1-PT-23) and determine maximum allowable Conduct of Operations 2.1.7 power level based on the calculation (ALL) (RO 4.4 I SRO 4.7)

Calculate maximum CC temperature for refueling in N R Equipment Control accordance with 1-OP-4.1.

2.2.44 (ALL) (RO 4.2 I SRO 4.4)

Select appropriate Radiation Work Permit and calculate M R Radiation Control stay time 2.3.7 (ALL) (RO 3.5 I SRO 3.6)

Determine Protective Action Request and update M R Emergency Procedures/Plan requirement fEPIP-1.06) 2.4.44 (SRO Only) (SRO 4.4)

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

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

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

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

ES-301 Control Roomlln-Plant Systems Outline Form ES-301-2 Facility: North Anna Power Station Date of Examination: 6/20/20 16 Exam Level: RO SRO-l SRO-U Operating Test No.: 1 Control Room Systems:* 8 for RO; 7 for SRO-l; 2 or 3 for SRO-U System I JPM Title (KA) Type Code*

Function

a. Verify safety injection flow (1-E-0). . (ALL) AD,E,EN,L 2 S

01 1 EA1 .13 Alt Path: Boron injection tank isolation valves will not open

b. Establish redundant cold leg injection flow paths (1-E-1) (RO and SRO-l) D,EEN,L,S 3 006A4.05

c. Respond to a loss of reactor coolant pump seal cooling (1-AP-33.2) (ALL) A,D,E,S 4(pti)

O15AA1.07 Alt Path: Seal return valve fails to close

d. Reduce containment pressure to subatmospheric (1-FR-Z.4) (RO and SRO-I) D,E,L,S 5 WE14EA1 .1

e. Respond to voltage regulator failure. (1-AP-26) (RO and SRO-I) A,E,S 6 077AA1 .03 Alt Path: Voltage does not respond requiring unit trip

f. Adjust Power Range NIs. (1-PT-24.1) (ALL) A,M,E,S 7 015A1.01 Alt Path: Rods step in when placed back in auto

g. Respond to CW flooding in the turbine building (0-AP-39.1) (RO and SRO-l) A,E,S 8 075A2.02 Alt Path: CW pumps fail to trip

h. Place Waste Gas Decay Tank on bleed. (RO) N,S 9 071 A4. 27

[ In-Plant Systems* (3 for RO); (3 for SRO-l); (3 or 2 for SRO-U) 7

i. Place a failed turbine stop valve position in TEST.(MOP-55.80)(RO and SRO-I) D 01 6A2 .01

j. Establish Normal Charging flow locally. (1-FCA-2) (ALL) D,E,R 1 004A1.11

k. Cross connect the 480 volt electrical busses. (1-OP-26.2) (ALL) D 6 062A4.07 All RD and SRO-l control room (and in-plant) systems must be different and serve different safety functions; all five SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

Type Codes Criteria for RD / SRO-l I S RD-U A)lternate path 4-6 I 4-6 I 2-3 (C)ontrol room (D)irectfrombank 9/814 (E)mergency or abnormal in-plant 1 / 1 I 1 (EN)gineered safety feature 1 / 1 / 1 (control room system)

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

(R)CA (S)imulator

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

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

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

Event Maif. Event Event Type* Description No. No.

I C (R);(S) Pressurizer spray valve fails open. Can be closed with SOV.(CT)

RC2002 l(B)(S) Running SW pump trips 2 SWO1O4 3 RDO7 C (R) (S) Continuous automatic control rod insertion which can be stopped in manual (CT) 4 MSO1O3 C (B) (S) Selected steam flow channel fails high on B SG TS_(S) 5 CH16OJ C (R) (S) Charging pump trip with failure of discharge check valve (CT)

CH21O1 TS(S) 6 MS1 002 C (ALL) Steam leak requiring power reduction 6a R (B) (5) Power reduction N (B) 7 MS1 002 M(ALL) Steam break requiring unit trip (2 CTs) 8 TUO3 C (B) (S) No automatic turbine trip will occur/A MSTV will not close automatically (CT) 9 C (B)(S) Turbine-driven AFW pump 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 BIT has been isolated in 1-ES-il.

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

?

SIMULATOR EXAMINATION GUIDE EVENT DESCRIPTION

1. Pressurizer spray valve fails open. Can be closed with SOV switch (CT)

2. Running SW pump trips

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

4. Selected steam flow channel fails high on B SG

5. Charging pump trip with failure of discharge check valve (CT)

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

7. Steam break requiring reactor trip (CTs)

8. No automatic turbine trip (CT),

9. Turbine driven AFW pump doesnt automatically start Scenario Recapitulation:

/

Malfunctions after EOP entry 2 No automatic turbine trip/A MSTV doesnt close automatically, turbine-driven AFW pump doesnt automatically start Total Malfunctions 9 Failed pressurizer spray valve, SW pump trip, continuous rod insertion, failed steam flow channel, charging pump trip/discharge check valve failure, steam leak, steamline break, no automatic turbine trip/A MSTV doesnt close automatically, turbine-driven AFW pump doesnt automatically start Abnormal Events 6 Failed pressurizer spray valve, SW pump trip, continuous rod insertion, failed steam flow channel, charging pump trip/discharge check valve failure, steam leak Major Transients 1 Faulted SG EOPs Entered 2 E-2, ES-1.1 EOP Contingencies 0 Critical Tasks 6 SCENARIO DURATION

1. Minutes 2016 NRC 1 Page 2 Revision 0

SIMULATOR EXAMINATION SCENARIO

SUMMARY

SCENARIO 2016 NRC 1 The scenario begins with the unit at 100% power, MOL. I-SW-P-lA, Unit I A SW pump, is tagged out for major repairs. 1-BC-P-YB 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 will fail open. The crew will respond in accordance with l-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. Once the crew has stabilized the unit, or at the direction of the lead evaluator, the next event can occur.

The 2-SW-P-IA 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-YB. The unit supervisor will consult TS and enter the action of 3.7.$B. 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. Once the crew has stabilized the unit, or at the direction of the lead evaluator, the next event can occur.

At this time channel ifi steam flow on B SG will fail high, the crew will enter 1-AP-3, Loss of Vital Instrumentation, and take manual control of B main feed regulation valve (MFRV). The crew will swap instrumentation to an operable channel. The US will consult TSs for the failure.

Once the channels have swapped and TS consulted, or at the direction of the lead evaluator, the next event can occur.

The running charging pump will trip and the standby pump will auto start. The discharge che*

valve on the previously running pump will stick open. The crew will enter 1-AP-49, Loss of Normal Charging, and close the discharge MOVs on the previously running pump. The crew will restore letdown flow and the US will consult TS 3.5.2 and make arrangements to swap to the C charging pump. 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. 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 will fail to start automatically and will have to be manually started.

The crew will proceed through 1-E-0 and transition to l-E-2, Faulted Steam Generator Isolation, and isolate the faulted SG. The crew will transition to 1-ES-I.1, SI Termination, and isolate the BIT. The scenario can be terminated at this time with direction form the lead evaluator.

2016 NRC 1 Page 3 Revision 0

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

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

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

Event Mall. Event Event Type* Description No. No.

I C (B) (5) Loss of IA (CT) 2 CH27 C (R) (S) 1-CH-TE-1I44 failure 3 CNO9OJ C (ALL) Main Condenser vacuum leak 3a R (R) (5) Unit power reduction due to vacuum leak N_(B) 3b RD14 C (R) (S) Rods stop stepping in automatic 4 RC0803 I (R) fS) Selected pressurizer level channel fails low (CT)

TS_(5) 4a N (B) (5) Restore letdown 5 TUJ1O1 C (B)(S) EHC pump trips 6 RCO4 C (ALL) RCS leak TS_(S) 7 RCOJOJ M (ALL) LBLOCA 8 C (ALL) Loss of emergency recirc (CT) 9 QSO3 C (ALL) Containment Depressurization Actuation does not work automatically (CT) 70 C (B) LHSI pumps dont automatically start (CT) 1-SI-P-lA shaft shears when started Events 8, 9 and 10 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

SIMULATOR EXAMINATION GUIDE EVENT DESCRIPTION

1. Loss of instrument air

2. l-CH-TE-1144 failure

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

4. Selected pressurizer level channel fails low 4a. Letdown is restored 5 EHC pump trips/standby pump does not auto start

6. RCS leak that eventually requires a unit trip

7. LBLOCA

8. NoautoCDA

9. LHSI pumps dont auto start/sheared shaft on 1-SI-P-lA

10. Loss of Emergency recirc Scenario Recapitulation:

Malfunctions after EOP entry 3 failure of automatic CDA, LHSI pumps do not automatically start/shaft shears on 1-SI-P-lA, loss of emergency recirc Total Malfunctions 11 Loss of Instrument Air, 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-IA, loss of emergency recirc Abnormal Events 6 Loss of Instrument Air, 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-l.l EOP Contingencies 1 ECA-1.l Critical Tasks 5 SCENARIO DURATION

1. Minutes 2016NRC2 Page 2 Revision 0

SIMULATOR EXAMINATION SCENARIO

SUMMARY

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

The first event will be a loss of instrument air. The running instrument air compressor will trip and the standby compressor will fail to start. The crew will enter 1-AP-2$, Loss of Instrument Air, and start the standby compressor. Once instrument air pressure has been restored and the next event can occur.

Next, 1-CH-TE-1144, low pressure letdown 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 tmatic..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 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-1 122 in manual and redue chargingfiuwtp zero. The RO will then swap to an operable pressurizer level channel. The crew will restore letdown at this time (Normal event). 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 will trip and the standby pump 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. 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.

i-SI-P-lA will start, but will have a sheared shaft. 1-SI-P-YB will start and flow. Also, when a CDA is required it will not happen automatically and the crew will have to manually actuate CDA.

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-MO V-1860B has no power. An operator sent to locally open the valve will 2016 NRC 2 Page 3 Revision 0 N ii

report that it is bound and will not open manually. At this time the crew will transition to 1-ECA-1.1, Loss of Emergency Coolant Recirculation. Once the crew has pressed the SI RECJRC MODE reset buttons and established a single train of SI flow, the scenario can be stopped with direction from the lead evaluator.

2016 NRC 2 Page 4 Revision 0

Appendix D Scenario Outline Form ES-D-1 Facility: North Anna Power Station Scenario No.: (2016) NRC -3 Op-Test No.: 1 Examiners: Operators:

Initial Conditions: 74% power MOL, Power was reduced several days ago to repair tubes in the 3A feedwater heater. Work has been completed and the feed train returned to service. Xenon is at equilibrium. 1-SW-P-lA is tagged out for major repairs. 1-BC-P-lB is tagged out for shaft replacement.

2H is the protected train.

Turnover: Ramp unit to 100% power. Assist maintenance with work on i-SW-P-iA, and 1-BC-P-i B, as required.

Event Maif. Event Event No. No. Type* Description I N (R) (S) Swap to UFM calorimetric 2 TS (5) 1-FW-MOV-154A loses power R(R)(S) Ramp the unit up 3

4 CNO3O1 C (B) A condensate pump degrade due to strainer clogging, standby pump_will_not_auto-start_if_required.

5 CHO8 I (R) (5) 1-CH-FT-1122 fails high 6 ASO1 C (B) (5) Aux steam pressure transmitter fails low 7 MSO2O1 I (R)(S) Selected first stage pressure transmitter failure (CT)

IS(S) 7a N (R) (5) Place steam dumps in steam pressure mode 8 ELOJ M (ALL) Loss of A RCP 9 RD32 C (ALL) Reactor does not trip automatically (CT) 10 C (R) Stuck control rods requite emergency boration (CT) 11 C (R) Emergency borate valve does not open from control room Events 9-i 1 will happen during event 8 and are numbered only for use_on_subsequent_forms.

The scenario can be terminated once emergency boration has been_started_for_the_stuck_control_rods.

(N)ormal, (R)eactivity, (l)nstrument, fC)omponent, (M)ajor

SIMULATOR EXAMINATION GUIDE EVENT DESCRIPTION

1. Swap to UFM calorimetric

2. l-FW-MOV-154A loses power

3. Ramp the unit up

4. A condensate pump degrades due to strainer clogging, standby pump will not auto-start if required.

5. l-CH-FT-1122 fails high

6. Aux steam pressure transmitter fails low

7. Selected first stage pressure transmitter failure 7a Steam dumps are transferred to Steam Pressure Mode

8. Loss of A RCP

9. Reactor does not trip automatically

10. Stuck control rods requiring emergency boration

11. Emergency borate valve fails to open from control room Scenario Recapitulation:

Malfunctions after EOP entry 3 Reactor does not trip automatically, three rods stuck out, emergency borate valve does not open from control room)

Total Malfunctions 9 fW-MOV loses power, A condensate pump degrades due to strainer clogging/standby pump does not start automatically, CH-FT-1122 fails high, AS pressure transmitter fails low, first stage pressure channel fails, loss of A RCP, reactor does not automatically trip, three rods stuck out, emergency borate valve does not open from control room Abnormal Events 4 A/condensate pump degrades due to strainer clogging/standby pump does not start automatically, CH-FT-1122, leak on CC pump, first stage pressure failure Major Transients 1 Loss of switchyard EOPs Entered 1 ES-O.l EOP Contingencies 0 Critical Tasks 3 SCENARIO DURATION

1. Minutes 2016NRC3 Page 2 Revision 0

I SIMULATOR EXAMINATION SCENARIO

SUMMARY

SCENARIO 2016 NRC 3 The scenario begins with the unit at 74% power, MOL. 1-SW-P-lA, Unit 1 A SW pump, is tagged out for major repairs. 1-BC-P-lB 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 normal event of swapping the calorimetric of record to the UFM. Once this is completed the next event can occur. This event can be pre-briefed.

Next, power will be lost to 1-FW-MOV-154A, A Main Feed Reg isolation MOV. The crew will receive a computer alarm and lose indication on the valve. They will enter the action for TS 3.7.3 and dispatch an operator to the power supply. The operator will report back that the breaker is open. Electricians are working at an adjacent breaker and may have inadvertently opened the breaker. The breaker can be re-closed, allowing the crew to exit the action. The next event can occur once the US has reviewed the TS, and the crew has either re-closed the breaker, or requested input from Ops management on the incident.

The next 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 suction strainer on a running Condensate Pump will clog causing Condensate Pump discharge pressure to decrease. The crew should start the standby Condensate pump (which will not auto-start) per the AR or 1-AP-31, Loss of Main Feedwater, to restore Main Feed Pump suction pressure. Once the standby Condensate Pump is running and MFW suction pressure is restored, the next event can occur.

The charging flow transmitter will fail high causing charging flow to decrease. Per the AR, the RO will place 1-CH-LCV-1 122 in manual and control charging flow. Once charging flow has been restored, the next event can occur.

At this time 1-AS-PT-105, Aux Steam pressure transmitter, will fail low. This will cause the AS relief valve to lift. The crew will enter 1-AP-38, Excessive Load Increase. The crew will ramp the unit, as necessary to stabilize reactor power. 1-AS-PCV 105 will be closed. Once the PCV is closed with power stable, and with direction from the lead evaluator, the next event can occur.

Next, the selected first stage pressure channel will fail low. The crew will be expected to respond JAW 1-AP-3, Loss of Vital Instrumentation. The RO will place rod control in manual. The BOP will take manual control of SG level, piallow SG level to control at 33% in automatic as directed by the US. The crew will place steam dumps in steam pressure mode and swap channels to the operable channel. The US/STA should refer to technical specifications and determine that permissives must be checked within one hour and 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 the crew has determined the appropriate MOP for placing the channel in trip and checked the one-hour permissives, the next event will occur.

At this time, the A reactor coolant pump will trip, but the reactor will not trip. Several control rods will stick out on the manual trip. The crew will enter 1 -E-0, Reactor Trip or Safety Injection, and then l-ES-0.1, Reactor Trip Response. 0-AP-lO, Loss of Electrical 2016 NRC 3 Page 3 Revision 0 j

Power, will also be entered. When borating for the stuck rods, the emergency borate valve will not open from the control room and will have to be locally opened. The scenario may be terminated after the crew has started an adequate emergency boration for the stuck control rods.

1 2016 NRC 3 Page 4 Revision 0

Appendix 0 Scenario Outline Form ES-D-1 Facility: North Anna Power Station Scenario No.: (2016) NRC-4 Op-Test No.: j.

Examiners: Operators:

Initial Conditions: 69% power, MDL. 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-iA is tagged out for major repairs.

1-BC-P-lB 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-IA and 1-BC-P-i B, as required.

Event Maif. Event Event Type* Description No. No.

I R (R) (5) Ramp the unit up N (B) 2 C (R) (5) Median/Tave unit fails high(CT) 2a N (R) (5) Steam dumps are placed in steam pressure mode) 3 TS(S) IRPlfails low 4 C (B) (5) SG PORV opens unexpectedly. Can be closed from the control room 5 I (R) (5) Letdown leak, isolable from control room 5a N (S)(B) Place excess letdown in service 6 I (B) (5) Selected feed flow transmitter fails 6a S (IS) RWST level transmitter fails downscale 7 M (ALL) Loss of Main Feedwater/ATWS 8 C (R) Control rods will not insert in auto or manual 9 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 i-E-0.

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

SIMULATOR EXAMINATION GUIDE EVENT DESCRIPTION

1. Ramp unit up

2. Median/Tave unit fails high 2a. Steam dumps to steam pressure mode

3. IRPI fails low

4. 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 feed flow transmitter fails 6a. RW$T level channel fails downscale

7. Loss of Main feedwater/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 10 IRPI fails low, Tave unit fails low, SG PORV opens, letdown leak, selected feed flow channel fails RWST level channel failure, loss of main feed, ATWS, rods do not insert in auto or manual, turbine will not trip and stop valves will not close Abnormal Events 5 Tave unit fails low, SG PORV opens, letdown leak, selected feed flow channeL fails RWST level channel fails Major Transients 1 Loss of Main feed/ATWS EOPs Entered 1 PR-S.l EOP Contingencies 1 FR-S.l Critical Tasks 3 SCENARIO DURATION

1. Minutes 2016 NRC 4 Page 2 Revision 0

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-lA, Unit 1 A SW pump, is tagged out for major repairs. 1-BC-P-lB 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 lAW 1-AP-l.1, Continuous Uncontrolled Rod Motion, and place rod control in MANUAL. Also, crew should address annunciators B-A7, MEDIAN/HI TAVG <>TREF DEVIATION, and B-A8, LOOP lA-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 SO PORV failing open due to the failure of the E/P. The crew may reduce power per I -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 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. At this time, 1-QS-LT-100A, will also fail downscale. This failure wiLij2e covered by 1-AP-3. The crew will swap steam generator water level control channels to channel ifi. Once the crew has identified MOPs, consulted Tech Specs, and with direction of the lead evaluator, the next event can occur.

A fault will occur on breaker 15A2, A station service bus normal supply breaker. 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 0

turbine stop valves will not close. The crew will close the MSTVs, and inject the BIT to have sufficient negative reactivity addition. At this time the reactor will be tripped locally. The crew will transition back to I -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 0

Appendix D Scenario Outline Form ES-D-1 Facility: North Anna Power Station Scenario No.: (2016) NRC-5 (SPARE) Op-Test No.: 1 Examiners: Operators:

Initial Conditions: Unit is at 100% power. i-SW-P-lA is tagged out for major repairs. i-BC-P-lB is tagged out for shaft replacement. 2H is the protected train.

Turnover: Ramp the unit down to 70% in preparation for removing a feed train from service. Support maintenance on repair of 1-SW-P-lA and 1-BC-P-i B, as required.

Event MaIf. Event Event No. No. Type* Description I R (R) (S Ramp the unit down in preparation for removing a feed train from N (B) service Ia FW3301 C (B) (S) A MFRV is erratic in automatic, must be placed in manual control 2 I (R) (S) 1-RC-LC-1459G fails high causing charging flow to increase 3 RC1 102 15 (S) 1-RC-Fl-14i5, B Loop flow Channel II 4 CCO7O1 C (B) (S) Leak on running CC pump 5 RD16i8 C(R)S) Dropped rod (CT)

IS(S) 6 FW18Oi C (ALL) Feedback arm falls off A MFRV requiring reactor trip 7 RD2128 M (ALL) Ejected control rod on reactor trip (2 CT5)

8. S108 C (ALL) Failure of automatic safety injection (CT) 9 C (R) Failure of BOP SI switch. (RD must actuate SI) 10 S11303 C (ALL) No auto Phase A isolation (CT) 511 304 (Events 8, 9, and 10 happen during event 7 and are numbered only for use on subsequent forms.)

The scenario can be terminated once the crew has performed actions in i-E-1.

(N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor 4 1

SIMULATOR EXAMINATION GUIDE EVENT DES CRWTION

1. Ramp the unit down in preparation for removing a feed train from service la A: MFRV is erratic in automatic requiring manual operation

2. 1-RC-LC-1459G fails high causing charging flow to increase

3. 1-RC-FI-1415, B Loop flow Channel II

4. Leak on running CC pump

5. Dropped rod 6.17. A MFRV feedback arm falls off requiring reactor trip! Ejected control rod on reactor trip

8. Failure of automatic safety injection (CT)

9. Failure of BOP SI switch. (RO must actuate SI) io. No auto Phase A isolation Scenario Recapitulation:

Malfunctions after EOP entry 4 Ejected rod, failure of automatic SI, failure of BOP SI switch, failure of auto Phase A Total Malfunctions 10 A MFRV is erratic in automatic, RC-LC-1459G fails low, 1-RC-FI-1415 fails low, leak on running CC pump, dropped rod, A MfRV feedback arm falls off, ejected rod, failure of automatic SI, failure of BOP SI switch, failure of auto Phase A Abnormal Events 5 A MFRV is erratic in automatic, RC-LC-1459G fails low, 1-RC-FI-1415 fails low, leak on running CC pump, dropped rod Major Transients 1 Ejected rod EOPs Entered 1 E-1 EOP Contingencies 0 Critical Tasks 5 SCENARIO DURATION

1. Minutes 2016 NRC 5 Page 2 Revision 0

I..

SIMULATOR EXAMINATION SCENARIO

SUMMARY

SCENARIO 2016 NRC 5 The scenario begins with the unit at 100% power, MOL. 1-SW-P-IA, Unit 1 A SW pump, is tagged Out for major repairs. 1-BC-P-lB is tagged for shaft replacement, not expected back for several days. 2H is the protected train. Shift orders are to ramp the unit down to less than 70%

power in preparation for taking A feed train out of service for work on 1-SD-LCV-l2lA, the 3A feedwater heater high level divert to the condenser.

The crew will begin a 3%/mm ramp to 69% power in accordance with l-OP-2.2, Unit Power Operation from Mode 1 to Mode 2. This evolution can be pre-briefed. When enough of a power decrease has occurred and with the direction of the lead evaluator, the next part of the event can occur. As the ramp continues, the A MFRV will begin to act erratically. The BOP will place the valve in manual and restore level to normal. The valve will have to remain in manual, if it is placed back in automatic it will still act erratically. The next event can occur once SG level is returned to normal in manual, and with direction of the lead evaluator.

At this time, 1-RC-LC-1459G will fail high causing charging flow to increase. The crew will respond taking manual control of l-CH-FCV-1122 to restore charging flow to normal. Charging control will remain in manual for the remainder of the scenario. Once charging flow has been restored, and with direction form the lead evaluator, the next event can occur.

Next, 1-RC-FI-1415, Channel II of B RCS loop flow, will fail low. The crew will enter 1-AP-3, Loss of Vital Instrumentation and the US will 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 />. This event is for a TS call, so once TS have been evaluated, the next event can occur.

A CC leak will develop on the seal of the running CC pump. CC surge tank level will decrease and AB sump level will increase. Personnel dispatched to the area will identify the leak location.

The crew will direct a makeup to the CC surge tank, swap CC pumps, and direct isolation of the affected pump. Once the leak is isolated and head tank level is returned to service, the next event can occur.

At this time a control rod will drop into the core. The crew will enter 1-AP-1.2, Dropped Rod, and the RO will place rods in Manual to prevent rods from moving to correct for Tavg/Tref differences. Once the US has entered TS 3.2.4 for QPTR, and with the direction of the lead evaluator, the next event can occur.

The A MFRV feedback arm will fall off and the valve will fail full open. The crew will trip the reactor (or it will automatically trip). When the reactor trips a control rod will be ejecicd. The crew will be required to manually initiate safety injection, the BOP switch will not work, requiring that the RO turn his switch (the crew may not be aware of this failure). Phase A isolation will also fail to automatically actuate and will be actuated by the crew using the 1-E-0 attachment. Once the crew has transitioned to l-E-l, Loss of Primary or Secondary Coolant, and performed actions, the scenario may be terminated.

2016 NRC 5 Page 3 Revision 0

ES-401 PWR Examination Outline Form ES-401-2 Facility: NORTH ANNA POWER STATION Date of Exam: JUNE 2016 RO K/A Category Points SRO-Only Points Tier Group K K K K K K A A A A G* A2 G* Total 1 2 3 4 5 6 1 2 3 4 Total

1. 1 3 3 3 3 3 3 18 3 3 6 Emergency &

Abnormal 2 1 1 2 N/A 2 2 N/A 1 9 2 2 4 Plant Evolutions Tier Totals 4 4 5 5 5 4 27 5 5 10 1 3 2 2 2 3 3 2 3 2 3 3 28 3 2 5 2.

Plant 2 1 1 1 1 1 1 1 0 1 1 1 10 0 1 2 3 Systems Tier Totals 4 3 3 3 4 4 3 3 3 4 4 38 4 4 8

3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 2 2 3 3 1 2 2 2 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each K/A category shall not be less than two). (One Tier 3 Radiation Control K/A is allowed if the K/A is replaced by a K/A from another Tier 3 Category).

2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points.

3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted with justification; operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.

4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.

5. Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.

7. The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As.

8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in a category other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams.

9. For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.

G* Generic K/As

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

E/APE # / Name / Safety Function K K K A A G* K/A Topic(s) IR #

1 2 3 1 2 000007 (BW/E02&E10; CE/E02) Reactor X 007EA2.06: Ability to determine or interpret the following 4.3 Trip - Stabilization - Recovery / 1 as they apply to a reactor trip: Occurrence of a reactor trip 000008 Pressurizer Vapor Space X 008AK1.01: Knowledge of the operational implications of 3.2 Accident / 3 the following concepts as they apply to a Pressurizer Vapor Space Accident: Thermodynamics and flow characteristics of open or leaking valves.

008AG2.4.20: 008AK1.01: Knowledge of the operational X implications of the following concepts as they apply to a 4.3 Pressurizer Vapor Space Accident: Knowledge of the operational implications of EOP warnings, cautions, and notes 000009 Small Break LOCA / 3 X 009EG2.4.21: Knowledge of the parameters and logic 4.0 used to assess the status of safety lfunctions, such as reactivity control, core cooling and heat removal, reactor lcoolant system integrity, containment conditions, radioactivity release control, letc.

000011 Large Break LOCA / 3 X 011EA1.09: Ability to operate and monitor Core flood tank 4.3 initiation as they apply to a Large Break LOCA 000015/17 RCP Malfunctions / 4 000022 Loss of Rx Coolant Makeup / 2 X 022A2.02: Knowledge of the interrelations between the 3.2 Loss of Reactor Coolant Makeup and Charging pump problems.

000025 Loss of RHR System / 4 X 025AK1.01: Knowledge of the operational implications of 3.9 the Loss of RHR System during all modes of operation as they apply to Loss of Residual Heat Removal System:

000026 Loss of Component Cooling Water / 8 000027 Pressurizer Pressure Control X 027AK2.03: Knowledge of the interrelations between the 2.6 System Malfunction / 3 Pressurizer Pressure Control Malfunctions and Controllers and positioners 000029 ATWS / 1 X 029EK3.03: Knowledge of the reasons for Opening BIT 3.7 inlet and outlet valves as the apply to the ATWS.

000038 Steam Gen. Tube Rupture / 3 X 038EK3.03: Knowledge of the reasons for Automatic 3.6 actions associated with high radioactivity in S/G sample lines as the apply to the SGTR.

000040 (BW/E05; CE/E05; W/E12) X 040AK2.01: Knowledge of the interrelations between the 2.6 Steam Line Rupture - Excessive Heat Steam Line Rupture and the valves.

Transfer / 4 000054 (CE/E06) Loss of Main X 054AG2.4.20: Knowledge of the operational implications of 3.8 Feedwater / 4 EOP warnings, cautions, and notes.

054AG2.1.19: Ability to use plant computers to evaluate X system or component status. 3.8 000055 Station Blackout / 6 X 055EA1.02: Ability to operate and monitor Manual ED/G 4.3 start as applied to a Station Blackout 055EA2.03: Ability to determine or interpret actions X necessary to restore power as they apply to a Station 4.7 Blackout.

Tier 1 / Group 1 (RO / SRO) Page 1 of 2

000056 Loss of Off-site Power / 6 X 056AK3.01: Knowledge of the reasons order and time to 3.5 initiation of power for the load sequencer as they apply to the Loss of Offsite Power:

000057 Loss of Vital AC Inst. Bus / 6 X 057AA2.02: Ability to determine and interpret Core flood 3.7 tank pressure and level indicators as they apply to the Loss of Vital AC Instrument Bus.

000058 Loss of DC Power / 6 X 058AG2.4.11: Knowledge of abnormal condition 4.0 procedures.

058AG2.1.20: Ability to interpret and execute procedure X steps. 4.6 000062 Loss of Nuclear Svc Water / 4 000065 Loss of Instrument Air / 8 W/E04 LOCA Outside Containment / 3 X WE04EK1.1: Knowledge of the operational implications of 3.5 the Components, capacity, and function of emergency systems as they apply to the LOCA Outside Containment.

W/E11 Loss of Emergency Coolant X WE11EK2.2: Knowledge of the interrelations between the 3.9 Recirc. / 4 Loss of Emergency Coolant Recirculation and facility*s heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.

BW/E04; W/E05 Inadequate Heat X WE05EA1.1: Ability to operate and / or monitor 4.1 Transfer - Loss of Secondary Heat Sink / 4 components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features. as they apply to the Loss of Secondary Heat Sink.

X WE05EA2.2: Ability to determine and interpret adherence to appropriate procedures and operation within the 4.3 limitations in the facility*s license and amendments as they apply to the Loss of Secondary Heat Sink.

000077 Generator Voltage and Electric X 077AA2.05: Ability to determine and interpret operational 3.2 Grid Disturbances / 6 status of offsite circuit as they apply to Generator Voltage and Electric Grid Disturbances.

K/A Category Totals: 3 3 3 3 3/3 3/3 Group Point Total: 18/6 Tier 1 / Group 1 (RO / SRO) Page 2 of 2

ES-401 3 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (RO / SRO)

E/APE # / Name / Safety Function K K K A A G* K/A Topic(s) IR #

1 2 3 1 2 000001 Continuous Rod Withdrawal / 1 X 001AA2.04: Ability to determine and interpret 4.3 reactor power and its trend as they apply to the Continuous Rod Withdrawal 000003 Dropped Control Rod / 1 000005 Inoperable/Stuck Control Rod / 1 000024 Emergency Boration / 1 000028 Pressurizer Level Malfunction / 2 X 028AK1.01: Knowledge of the operational 2.8 implications of PZR reference leak abnormalities as they apply to PZR Level Control Malfunctions:

000032 Loss of Source Range NI / 7 X 032AA2.04: Ability to determine and interpret 3.1 satisfactory source-range/intermediate-range overlap as they apply to the Loss of Source Range Nuclear Instrumentation 000033 Loss of Intermediate Range NI / 7 X 033AK3.01: Knowledge of the reasons for the 3.2 termination of startup following loss of intermediate range instrumentation .

000036 (BW/A08) Fuel Handling Accident / 8 000037 Steam Generator Tube Leak / 3 000051 Loss of Condenser Vacuum / 4 000059 Accidental Liquid Radwaste Rel. / 9 000060 Accidental Gaseous Radwaste Rel. / 9 X 060AK2.01: Knowledge of the interrelations 2.6 between the Accidental Gaseous Radwaste Release and the ARM system, including the normal radiation-level indications and the operability status.

060AG2.4.30: Knowledge of events related to X system operation/status that must be reported to 4.1 internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.

000061 ARM System Alarms / 7 X 061AA2.03: Ability to determine and interpret 3.3 setpoints for alert and high alarms as they apply to the Area Radiation Monitoring (ARM) System Alarms.

000067 Plant Fire On-site / 8 X 067AA1.05: Ability to operate and / or monitor 3.0 plant and control room ventilation systems as they apply to the Plant Fire on Site.

000068 (BW/A06) Control Room Evac. / 8 X 068AK3.17: Knowledge of the reasons for 3.7 injection of boric acid into the RCS as they apply to the Control Room Evacuation.

000069 (W/E14) Loss of CTMT Integrity / 5 X WE14EG2.4.21: Knowledge of the parameters 4.6 and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

000074 (W/E06&E07) Inad. Core Cooling / 4 000076 High Reactor Coolant Activity / 9 X 076AA2.03: Ability to determine and interpret RCS 2.5 radioactivity level meter as they apply to the High Reactor Coolant Activity.

W/EO1 & E02 Rediagnosis & SI Termination / 3 W/E13 Steam Generator Over-pressure / 4 W/E15 Containment Flooding / 5 Tier 1 / Group 2 (RO / SRO) Page 1 of 2

W/E16 High Containment Radiation / 9 BW/A01 Plant Runback / 1 BW/A02&A03 Loss of NNI-X/Y / 7 BW/A04 Turbine Trip / 4 BW/A05 Emergency Diesel Actuation / 6 BW/A07 Flooding / 8 BW/E03 Inadequate Subcooling Margin / 4 BW/E08; W/E03 LOCA Cooldown - Depress. / 4 BW/E09; CE/A13; W/E09&E10 Natural Circ. / 4 X WE09EG2.4.31: Knowledge of annunciator alarms, indications, or response procedures.

BW/E13&E14 EOP Rules and Enclosures CE/A11; W/E08 RCS Overcooling - PTS / 4 X WE08EA1.1: Ability to operate and / or monitor 3.8 components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features as they apply to Pressurized Thermal Shock.

CE/A16 Excess RCS Leakage / 2 CE/E09 Functional Recovery K/A Category Point Totals: 1 1 2 2 2/2 1/2 Group Point Total: 9/4 Tier 1 / Group 2 (RO / SRO) Page 2 of 2

ES-401 4 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 1 (RO / SRO)

System # / Name K K K K K K A A A A G* K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 003 Reactor Coolant Pump X 003K5.02: Knowledge of the operational 2.8 implications of effects of RCP coastdown on RCS parameters as they apply to the RCPS.

X 003K6.14: Knowledge of the effect of a loss 2.6 or malfunction on the starting requirements will have on the RCPS.

004 Chemical and Volume X 004A2.22: Ability to (a) predict the impacts of 3.2 Control a mismatch of letdown and changing flows on the CVCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of the mismatch.

005 Residual Heat Removal X 005A1.07: Ability to predict and/or monitor 2.5 changes in parameters (to prevent exceeding design limits) associated with operating the RHRS controls including Determination of test acceptability by comparison of recorded valve response times with Tech-Spec requirements.

005A2.01: Ability to (a) predict the impacts of X failure modes for pressure, flow, pump motor 2.7 amps, motor temperature, and tank level instrumentation the RHRS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations.

006 Emergency Core Cooling X 006K5.07: Knowledge of the operational 2.7 implications expected temperature levels in various locations of the RCS due to various plant conditions as they apply to ECCS.

006K6.05: Knowledge of the effect of a loss X or malfunction of HPI/LPI cooling water will 3.0 have on the ECCS.

007 Pressurizer Relief/Quench X 007G2.1.20: Ability to interpret and execute 4.6 Tank procedure steps.

008 Component Cooling Water X 008K2.02: Knowledge of bus power supplies 3.0 to CCW pump, including emergency backup.

010 Pressurizer Pressure Control X 010K6.02: Knowledge of the effect of a loss 3.2 or malfunction of the PZR will have on the PZR PCS.

012 Reactor Protection X 012K4.04: Knowledge of RPS design 3.1 feature(s) and/or interlock(s) which provide for redundancy.

X 012A2.04: Ability to (a) predict the impacts of 3.2 faulty or erratic operation of detectors and function generators on the RPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Tier 2 / Group 1 (RO / SRO) Page 1 of 3

013 Engineered Safety Features X 013A4.01: Ability to manually operate and/or 4.5 Actuation monitor in the control room: ESFAS-initiated equipment which fails to actuate.

X 013A2.04: Ability to (a) predict the impacts of 4.2 the loss of instrument bus on the ESFAS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations.

022 Containment Cooling X 022K4.05: Knowledge of CCS design 2.6 feature(s) and/or interlock(s) which provide for the following: Containment cooling after LOCA destroys ventilation ducts 025 Ice Condenser 026 Containment Spray X 026A1.06: Ability to predict and/or monitor 2.7 changes in parameters (to prevent exceeding design limits) associated with operating the CSS controls including: Containment spray pump cooling 039 Main and Reheat Steam X 039K5.08: Knowledge of the operational 3.6 implications of the effect of steam removal on reactivity as applied to the MRSS.

X 039A2.03: Ability to (a) predict the impacts of 3.7 Indications and alarms for main steam and area radiation monitors (during SGTR) on the MRSS; and (b) based on predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

059 Main Feedwater X 059A3.02: Ability to monitor automatic 2.9 operation of the MFW, including programmed levels of the S/G.

061 Auxiliary/Emergency X 061K2.03: Knowledge of bus power supplies 4.0 Feedwater to the following: AFW diesel driven pump 062 AC Electrical Distribution X 062A2.11: Ability to (a) predict the impacts of 3.7 the aligning standby equipment with correct emergency power source (D/G) on the ac distribution system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations.

X 062G2.4.11: Knowledge of abnormal 4.0 condition procedures.

063 DC Electrical Distribution X 063K3.01: Knowledge of the effect that a loss 3.7 or malfunction of the DC electrical system will have on the following: ED/G.

064 Emergency Diesel Generator X 064A4.01: Ability to manually operate and/or 4.0 monitor in the control room: Local and remote operation of the ED/G.

X 064K3.03: Knowledge of the effect that a 3.6 loss or malfunction of the ED/G system will have on the following: ED/G (manual loads).

073 Process Radiation Monitoring X 073K1.01: Knowledge of the physical 3.6 connections and/or cause-effect relationships between the PRM system and those systems served by PRMs.

Tier 2 / Group 1 (RO / SRO) Page 2 of 3

076 Service Water X 076A3.02: Ability to monitor automatic 3.7 operation of the SWS, including emergency heat loads.

X 076A4.01: Ability to manually operate and/or 2.9 monitor in the control room: SWS pumps 076G2.1.25: Ability to interpret reference X materials, such as graphs, curves, tables, etc. 4.2 078 Instrument Air X 078G2.1.31: Ability to locate control room 4.6 switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup.

078K1.05: Knowledge of the physical X connections and/or cause-effect relationships 3.4 between the IAS and MSIV air.

103 Containment X 103K1.03: Knowledge of the physical 3.1 connections and/or cause-effect relationships between the containment system and shield building vent system.

103G2.2.38: Knowledge of conditions and X limitations in the facility license. 4.5 K/A Category Point Totals: 3 2 2 2 3 3 2 3/3 2 3 3/2 Group Point Total: 28/5 Tier 2 / Group 1 (RO / SRO) Page 3 of 3

ES-401 5 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 2 (RO / SRO)

System # / Name K K K K K K A A A A G* K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 001 Control Rod Drive 002 Reactor Coolant X 002K5.10: Knowledge of the operational 3.6 implications of the relationship between reactor power and RCS differential temperature as they apply to the RCS 011 Pressurizer Level Control X 011K4.05: Knowledge of PZR LCS design 3.7 feature(s) and/or interlock(s) which provide for PZR level inputs to RPS.

014 Rod Position Indication X 014A1.02: Ability to predict and/or monitor 3.2 changes in parameters (to prevent exceeding design limits) associated with operating the RPIS controls, including Control rod position indication on control room panels.

015 Nuclear Instrumentation X 015G2.2.25: Knowledge of the bases in 4.2 Technical Specifications for limiting conditions for operations and safety limits.

016 Non-Nuclear Instrumentation 017 In-Core Temperature Monitor X 017K6.01: Knowledge of the effect of a loss or 2.7 malfunction of the following ITM system components: Sensors and detectors .

027 Containment Iodine Removal X 027K2.01: Knowledge of bus power supplies to 3.1 the following: Fans.

028 Hydrogen Recombiner and X 028A2.01: Malfunctions or operations on the 3.6 Purge Control HRPS; and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of those malfunctions or operations: Hydrogen recombiner power setting, determined by using plant data book 029 Containment Purge X 029K1.05: Knowledge of the physical connections 2.9 and/or cause-effect relationships between the Containment Purge System and Containment air cleanup and recirculation system.

033 Spent Fuel Pool Cooling X 033G2.2.3: Knowledge of the design, procedural, 2.7 and operational differences between units.

034 Fuel Handling Equipment 035 Steam Generator X 035A4.01: Ability to manually operate and/or 2.7 monitor in the control room: Fill of dry S/G 041 Steam Dump/Turbine Bypass Control 045 Main Turbine Generator 055 Condenser Air Removal X 055K3.01: Knowledge of the effect that a loss or 2.5 malfunction of the CARS will have on the Main condenser 056 Condensate Tier 2 / Group 2 (RO / SRO) Page 1 of 2

068 Liquid Radwaste X 068A3.01: Ability to monitor automatic operation 2.5 of the Liquid Radwaste System including Evaporator pressure control.

071 Waste Gas Disposal X 071G2.4.21: Knowledge of the parameters and 4.6 logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

072 Area Radiation Monitoring 075 Circulating Water 079 Station Air 086 Fire Protection K/A Category Point Totals: 1 1 1 1 1 1 1 0/1 1 1 1/2 Group Point Total: 10/3 Tier 2 / Group 2 (RO / SRO) Page 2 of 2

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: NORTH ANNA POWER STATION Date of Exam: JUNE 2016 Category K/A # Topic RO SRO-Only IR # IR #

2.1.14 Knowledge of criteria or conditions that require plant-wide 3.1 announcements, such as pump starts, reactor trips, mode changes, etc.

1. 2.1.36 Knowledge of procedures and limitations involved in core 3.0 Conduct of alterations.

Operations 2.1.40 Knowledge of refueling administrative requirements. 3.9 2.1.

Subtotal 2 1 2.2.20 Knowledge of the process for managing troubleshooting activities 2.6 2.2.40 Ability to apply Technical Specifications for a system. 3.4 2.

2.2.14 Knowledge of the process for controlling equipment configuration 4.3 Equipment or status Control 2.2.21 Knowledge of pre- and post-maintenance operability 4.1 requirements.

Subtotal 2 2 2.3.11 Ability to control radiation releases 3.8 2.3.12 Knowledge of radiological safety principles pertaining to licensed 3.2 operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

2.3.15 Knowledge of radiation monitoring systems, such as fixed 2.9

3. radiation monitors and alarms, portable survey instruments, Radiation personnel monitoring equipment, etc.

Control 2.3.14 Knowledge of radiation or contamination hazards that may arise 3.8 during normal, abnormal, or emergency conditions or activities.

2.3.15 Knowledge of radiation monitoring systems, such as fixed 3.1 radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

2.3.

Subtotal 3 2 2.4.14 Knowledge of general guidelines for EOP usage. 3.8 2.4.26 Knowledge of facility protection requirements, including fire 3.1 brigade and portable firefighting equipment usage.

2.4.39 Knowledge of RO responsibilities in emergency plan 3.9

4. implementation. l Emergency Procedures / 2.4.27 Knowledge of fire in the plant procedures. 3.9 Plan 2.4.30 Knowledge of events related to system operation/status that 4.1 must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.

2.4.

Subtotal 3 2 Tier 3 Point Total 10 7

ES-401, REV 9 TIGI PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME/SAFETY FUNCTION: IR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO 007EA2.06 -

Reactor Trip Stabilization - Recovery 4.3 4.5 Ii Occurrence of a reactor trip 008AK1 .01 Pressurizer Vapor Space Accident / 3 3.2 3.7 Thermodynamics and flow characteristics of open or leak- ing valves 009EG2.4.21 Small Break LOCA/3 4.0 4.6 Knowledge of the parameters and logic used to assess the status of safety functions 011 EA1 .09 Large Break LOCA / 3 4.3 4.3 Core flood tank initiation 022AA2.02 Loss of Rx Coolant Makeup/ 2 3.2 3.7 Charging pump problems 025AK1.Ol LossofRHRSystem/4 3.9 4.3 E D D E Loss of RHRS during all modes of operation 027AK2.03 Pressurizer Pressure Control System 2.6 2.8 Malfunction / 3 Controllers and positioners 029EK3.03 ATWS! 1 3.7 3.6 fl Opening BIT inlet and outlet valves 038EK3.03 Steam Gen. Tube Rupture! 3 3.64 Automatic actions associated with high radioactivity in S!G sample lines 040AK2.01 -

Steam Line Rupture Excessive Heat 2.6 2.5 Valves Transfer / 4 054AG2.4.20 Loss of Main Feedwater / 4 3.8 4.3 Knowledge of operational implications of EOP warnings, cautions and notes.

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ES-401, REV 9 TIGI PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RD SRO 055EA1.02 Station Blackout? 6 4.3 4.4 El Manual ED/G start 056AK3.0l Loss of Off-site Power? 6 3.5 3.9 Order and time to initiation of power for the load sequencer 057AA2.02 Loss of Vital AC Inst. Bus? 6 3.7 3.8 Core flood tank pressure and level indicators 058AG2.4.11 Loss of DC Power? 6 4.0 4.2 Knowledge of abnormal condition procedures.

WEO4EK1.l LOCA Outside Containment? 3 3.5 3.9 Components, capacity and function of emergency systems.

WEO5EA1.1 Inadequate Heat Transfer- Loss of 4.1 4.0 fl Secondary Heat Sink / 4 Components and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes and automatic and manual features.

WE11EK2.2 Loss of Emergency Coolant Recirc.?4 3.9 4.3 D D Facilitys heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems and relations between the proper operation of these systems to the operation of the facility.

Page 2 of 2 06/24/2015 1:29 PM

ES-401, REV9 TIG2 PWR EXAMINATION OUTLINE KA FORM ES-401-2 NAME I SAFETY FUNCTION: IR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO 028AK1.01 Pressurizer Level Malfunction/2 2.8 3.1 PZR reference leak abnormalities 032AA2.04 Loss of Source Range NI I 7 3.1 3.5 Satisfactory source-range/intermediate-range overlap 033AK3.01 Loss of Intermediate Range NI / 7 3.2 3.6 Termination of startup following loss of intermediate-range instrumentation 060AK2.0l Accidental Gaseous Radwaste Rel. /9 2.6 2.9 ARM system, including the normal radiation-level indications and the operability status 067AA1 .05 Plant Fire On-site / 8 3 3.1 Plant and control room ventilation systems 068AK3.17 Control Room Evac. / 8 3.7 4 Injection of boric acid into the RCS 076AA2.03 High Reactor Coolant Activity / 9 2.5 3 RCS radioactivity level meter WEO8EA1.1 RCS Overcooling - PTS / 4 3.8 3.8 Components and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes and automatic and manual feature weO9EG2.4.31 Natural Circ. / 4 s.

4.2 4.1 Knowledge of annunciators alarms, indications or response procedures Page 1 of 1 06/24/2015 1:29 PM

ES-401, REV 9 T2GI PWR EXAMINATION OUTLINE KA FORM ES-401-2 NAME I SAFETY FUNCTION: IR Kl K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO 003K5.02 Reactor Coolant Pump 2.8 3.2 Effects of RCP coastdown on RCS parameters 003K6.14 Reactor Coolant Pump -

2.6 2.9 fl fl Starting requirements 004A2.22 Chemicai and Volume Control 3.2 3.1 Mismatch of letdown and changing flows 005A1.07 Residual Heat Removal 2.5 3.1 Determination of test acceptability by comparison of recorded valve response times with Tech-Spec requirements 005A2.01 Residual Heat Removal 2.7 2.9 -

E Failure modes for pressure, flow, pump motor amps, motor temperature and tank level instrumentation 006K5.07 Emergency Core Cooling 2.7 3.0 Expected temperature levels in various locations of the RCS due to various plant conditions 006K6.05 Emergency Core Cooling 3.0 3.5 HPI/LPI cooling water 007G2.1 .20 Pressurizer Relief/Quench Tank 4.6 4.6 Ability to execute procedure steps.

008K2.02 Component Cooling Water 3.0 3.2 CCW pump, including emergency backup 010K6.02 Pressurizer Pressure Control 3.2 3.5 PZR 012K4.04 Reactor Protection 3.1 3.3 Redundancy Page 1 of 3 06/24/2015 1:29 PM

ES-401, REV 9 T2GI PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR KI K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO 013A4.0l Engineered Safety Features Actuation 4.5 4.8 El ESFAS-initiated equipment which fails to actuate 022K4.05 Containment Cooling 2.6 2.7 Containment cooling after LOCA destroys ventilation ducts 026A1 .06 Containment Spray -

2.7 3.0 Containment spray pump cooling 039K5.08 Main and Reheat Steam 3.6 3.6 Effect of steam removal on reactivity 059A3.02 Main Feedwater 2.9 3.1 Programmed levels of the S/G 061 K2.03 Auxiliary/Emergency Feedwater 4.0 3.8 El AFW diesel driven pump 062A2.l 1 AC Electrical Distribution 3.7 4.1 fl Aligning standby equipment with correct emergency power source (DIG) 062G2.4.l 1 AC Electrical Distribution 4.0 4.2 Knowledge of abnormal condition procedures.

063K3.0l DC Electrical Distribution 3.7 4.1 ED/G 064A4 01 Emergency Diesel Generator 4.0 4.3 Local and remote operation of the EDIG 064K3.03 Emergency Diesel Generator 3.6 3.9 EDIG (manual loads)

Page 2 of 3 06/24/2015 1:29 PM

ES-401, REV 9 T2GI PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR Kl K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RD SRO 073K1.01 Process Radiation Monitoring 3.6 3.9 i Those systems served by PRMs D D E 076A3.02 Service Water 3.7 3.7 Emergency heat loads 076A4.01 Service Water 2.9 2.9 J SWS pumps 078G2.l.3l Instrument Air 4.6 4.3 Ability to locate control room switches, controls and indications and to determine that they are correctly reflecting the desired plant lineup.

078K1.05 InstrumentAir 3.4 3.5 MSlVair 103K1.03 Containment 3.1 3.5 fl E Shield building vent system Page 3 of 3 06/24/2015 1:29 PM

ES-401, REV 9 T2G2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO 002K5.l0 Reactor Coolant 3.6 4.1 Relationship between reactor power and RCS differential temperature 011K4.05 Pressurizer Level Control 3.7 4.1 PZR level inputs to RPS 014A1.02 Rod Position Indication 3.2 3.6 Control rod position indication on control room panels 017K6.01 In-core Temperature Monitor 2.7 3.0 Sensors and detectors 027K2.0l Containment Iodine Removal 3.1 3.4 Fans 029K1 .05 Containment Purge 2.9 3.1 Containment air cleanup and recirculation system 033G2.2.3 Spent Fuel Pool Cooling 3.8 3.9 (multi-unit license) Knowledge of the design, procedural and operational differences between units.

035A4.02 Steam Generator 2.7 2.8 Fill of dry S/G 055K3.D1 Condenser Air Removal 2.5 2.7 Main condenser 068A3.O1 Liquid Radwaste 2.5 2.4 Evaporator pressure control Page 1 of 1 06/24/2015 1:29PM

ES-401, REV 9 T3 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO G2.1.14 Conduct of operations 3.1 3.1 Knowledge of criteria or conditions that require plant-wide announcements, such as pump starts, reactor trip, mode changes, etc.

G2. 1.36 Conduct of operations 3.0 4.1 Knowledge of procedures and limitations involved in core alterations G2.2.20 Equipment Control 2.6 3.8 Knowledge of the process for managing troubleshooting activities.

G2.2.40 Equipment Control 3.4 4.7 Ability to apply technical specifications for a system.

G2.3.1i Radiation Control 3.8 4.3 Ability to control radiation releases.

G2.3.12 Radiation Control 3.2 3.7 Knowledge of radiological safety principles pertaining to licensed operator duties G2.3.15 Radiation Control 2.9 3.1 Knowledge of radiation monitoring systems G2.4. 14 Emergency Procedures/Plans 3.8 4.5 Knowledge of general guidelines for EOP usage.

G2.4.26 Emergency Procedures/Plans 3.1 3.6 Knowledge of facility protection requirements including fire brigade and portable fire fighting equipment usage.

G2.4.39 Emergency Procedures/Plans 3.9 3.8 Knowledge of the ROs responsibilities in emergency plan implementation.

Page 1 of 06/24/2015 1:29PM

ES-401, REV 9 SRO TIGI PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO 008AG2.4.20 Pressurizer Vapor Space Accident / 3 3.8 4.3 Knowledge of operational implications of EOP warnings, cautions and notes.

054AG2.1 .19 Loss of Main Feedwater 14 3.9 3.8 Ability to use plant computer to evaluate system or component status.

055EA2.03 Station Blackout! 6 -

3.9 4.7 Actions necessary to restore power 058AG2.1 .20 Loss of DC Power! 6 4.6 4.6 Ability to execute procedure steps.

077AA2.05 Generator Voltage and Electric Grid 3.2 3.8 Disturbances ! 6 Operational status of offsite circuit WEO5EA2.2 -

Inadequate Heat Transfer Loss of 3.7 4.3 Secondary Heat Sink ! 4 Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments.

Page 1 of 1 06/24/2015 1:29PM

ES-401, REV 9 SRO TIG2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: )R Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO OO1AA2.04 Continuous Rod Withdrawal / 1 4.2 4.3 Reactor power and its trend 060AG2.4.30 Accidental Gaseous Radwaste Rel. / 9 2.7 4.1 Knowledge of events related to system operations/status that must be reported to internal orginizations or outside agencies.

061AA2.03 ARM System Alarms! 7 3 3.3 Setpoints for alert and high alarms wel4EG2.4.21 Loss of CTMT Integrity / 5 4.0 4.6 Knowledge of the parameters and logic used to assess the status of safety functions Page 1 of 1 06/24/2015 1:29 PM

ES-401, REV 9 SRO T2GI PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO 012A2.05 Reactor Protection 3.1 3.2 Faulty or erratic operation of detectors and function generators 013A2.04 Engineered Safety Features Actuation 3.6 4.2 Loss of instrument bus 039A2.03 Main and Reheat Steam 3.4 3.7 Indications and alarms for main steam and area radiation monitors (during SGTR) 076G2.1 .25 Service Water 3.9 4.2 Ability to interpret reference materials such as graphs monographs and tables which contain performance data.

103G2.2.38 Containment 3.6 4.5 Knowledge of conditions and limitations in the facility license.

Page 1 of 1 06/24/2015 1:29PM

ES-401, REV 9 SRO T2G2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO 015G2.2.25 Nuclear Instrumentation 3.2 4.2 fl Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

028A2.01 Hydrogen Recombiner and Purge 3.4 3.6 Control Hydrogen recombiner power setting, determined by using plant data book 071 G2.4.21 Waste Gas Disposal 4.0 4.6 Knowledge of the parameters and logic used to assess the status of safety functions Page 1 of 1 06/24/2015 1:29PM

ES-401, REV 9 SRO T3 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: PR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RD SRO G2.l.40 Conduct of operations 2.8 3.9 fl Knowledge of refueling administrative requirements G2.2.14 Equipment Control 3.9 4.3 Knowledge of the process for controlling equipment configuration or status G2.2.21 Equipment Control 2.9 4.1 Knowledge of pre- and post-maintenance operability requirements.

G2.3.14 Radiation Control 3.4 3.8 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities G2.3.15 Radiation Control 2.9 3.1 Knowledge of radiation monitoring systems G2.4.27 Emergency Procedures/Plans 3.4 3.9 Knowledge of fire in the plant procedures.

G2.4.30 Emergency Procedures/Plans 2.7 4.1 Knowledge of events related to system operations/sta tus that must be reported to internal orginizations or outside agencies.

Page 1 of 1 06/24/2015 1:29 PM