ML102870152

Initial Exam 2010-301 Draft RO Written Exam

ML102870152
Person / Time
Site:	Vogtle
Issue date:	04/12/2010
From:	NRC/RGN-II
To:	Southern Nuclear Operating Co
References
50-424/10-301, 50-425/10-301
Download: ML102870152 (298)

Text

Vogtle Initial Exam February 2010 Draft RO ito 40

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HL-15R RO NRC Exam

1. 001K2.05 001/2/2/RODS-MG SETS/MEM- 2.9 / 3.2/NEW/HL-15RNRC/RO/TNT/DS An ATWT is in progress.

- The Unit Operator has left control room to locally trip the reactor.

If unable to successfully open the reactor trip breakers, per 19211-C, FR-S.l Response To Nuclear Power Generation/ATWT, the UO should open (1) MG sets__(2)_ breakers.

A. (1)both (2) supply B (1) both (2) output C. (1) either (2) supply D. (1)either (2) output KIA 001 Control Rod Drive System K2.05 Knowledge of bus power supplies to the following:

M/G sets.

K/A MATCH ANALYSIS The question presents a plausible scenario where an ATWT is in progress, the student must know to open both MG sets output breakers.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Both MG sets output breakers must be opened to trip the reactor as they are in parallel and one MG set can supply the rod control power cabinets. The output hrkrs to onnd vrsis thR siinnlv hrkrs flnninn thc otitniit hrkrs I

HL-I5RRONRCExam immediately cuts power to the rod control power cabinets and all rods immediately drop in. Opening the supply breakers would result in rods dropping in erratically as the MG sets coast down, this could result in possible flux peaking resulting in fuel damage.

B. Correct. Both MG sets output breakers must be opened to trip the reactor as they are in parallel and one MG set can supply the rod control power cabinets. The output breakers are to be opened versus the supply breakers. Opening the output breakers immediately cuts power to the rod control power cabinets and all rods immediately drop in. Opening the supply breakers would result in rods dropping in erratically as the MG sets coast down, this could result in possible flux peaking resulting in fuel damage.

C. Incorrect. Both MG sets output breakers must be opened to trip the reactor as they are in parallel and one MG set can supply the rod control power cabinets. The output breakers are to be opened versus the supply breakers. Opening the output breakers immediately cuts power to the rod control power cabinets and all rods immediately drop in. Opening the supply breakers would result in rods dropping in erratically as the MG sets coast down, this could result in possible flux peaking resulting in fuel damage.

D. Inorrect. Both MG sets output breakers must be opened to trip the reactor as they are in parallel and one MG set can supply the rod control power cabinets. The output breakers are to be opened versus the supply breakers. Opening the output breakers immediately cuts power to the rod control power cabinets and all rods immediately drop in. Opening the supply breakers would result in rods dropping in erratically as the MG sets coast down, this could result in possible flux peaking resulting in fuel damage.

REFERENCES 19211-C, FR-S.1 Response to Nuclear Power Generation/ATWT V-LO-PP-28101, Solid State Protection System, slide #67.

VEGP learning objectives:

LO-PP-27101-02, State the power supplies for the Rod Control System.

2

Approved By Procedure Number Rev J. D.WiIIiams Vogtle Electric Generating Plant 19211-C 20.1 Date Approved FR-S.1 RESPONSE TO NUCLEAR POWER Page Number 1-23-2007 GENERATION/ATWT 6of20 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

• 7 Check for SI:

a. SI signal EXISTS OR

_a. IF an SI signal is actuated ACTUATED. during this procedure, THEN initiate ATTACHMENT A.

Go to Step 8

b. Initiate ATTACHMENT A.

8. Check the following trips have occurred:

a. Reactor trip. _a. Locally trip the Reactor trip and Bypass breakers.

IF the trip breakers will NOT open, THEN trip the Control Rod Drive MG Set output breakers at the Reactor Trip Switchgear.

b. Turbine trip. _b. Dispatch operator to trip turbine at the HP Turbine front standard.

• 9 Check Reactor power:

a. LESS THAN 5%. a. GotoSteplO.

_b. IR SUR LESS THAN 0 DPM.

b. GotoSteplO.

c. Go to Step 24.

Printed November 9, 2009 at 08:16

Reactor Trip Reactor Trip Bypass Breaker Bypass Breaker A B MIG Set Output Breaker Rod Control Reactor Trip Reactor Trip Cabinets Breaker A Breaker B M/G Set Output Breaker 1

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HL-15R RO NRC Exam

2. 002A2.03 OO1/2!2/RCS-LOSS CIRCULATION/C/A -4.1 / 4.3/NEW/HL-15R NRC/RO!DS/TNT Initial conditions:

The unit was operating at 100% power The reactor was manually tripped 2NAA and 2NAB de-energize on the reactor trip The crew implements 19001-C, ES-O.l Reactor Trip Recovery Current conditions:

2NAB has been re-energized RCS WR Tcold - 558 F and rising AFW flow - 150 GPM per SG SG NR levels - 18% and slowly rising The UO and OATC should raise...

A. AFW flow to maintain RCS WR cold leg temperatures at 557 F while attempting to start RCP 4.

B. AFW flow to maintain RCS WR cold leg temperatures at 557 F while attempting to start RCP 1.

C the steaming rate with SG ARVs to maintain RCS WR cold leg temperatures at 557 F while attempting to start RCP 4.

D. the steaming rate with Steam Dumps to maintain RCS WR cold leg temperatures at 557 F while attempting to start RCP 1.

K/A 002 Reactor Coolant System (RCS)

A2.03 Ability to (a) predict the impacts of the following malfunctions or operations on the RCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Loss of forced circulation.

KIA MATCH ANALYSIS The question presents a natural circulation scenario where the student must use the correct actions of porcedure 19001-C and use systems knowledge to determine the impacts on the steam dump system to properly apply the actions in 19001-C.

3

HL-15R RO NRC Exam A. Incorrect. Raising AFW flow is plausible since SG NR levels are only at 18% and RCS temperature is increasing. Raising the AFW flow will introduce more cold feed water into the SGs and temporarily lower RCS temperatures. Step 15 of 19001-C requires starting an RCP (4 or 1 is preferred) to go back to forced flow conditions if possible.

B. Incorrect. Raising AFW flow is plausible since SQ NR levels are only at 18% and RCS temperature is increasing. Raising the AFW flow will introduce more cold feed water into the SG5 and temporarily lower RCS temperatures. Step 15 of 19001-C requires starting an RCP (4 or 1 is preferred) to go back to forced flow conditions if possible.

C. Correct. 2NAA and 2NAB power the RCPs and the circulating water pumps. C-9 signal will not be present, preventing the use of the steam dumps. With WR Tcold>

557 and increasing the correct action per 19001-C step 4 continuous action is to increase the rate of dumping steam. This will have to be done using the SG ARV5 since the steam dumps will not arm for these conditions. Step 15 of 19001-C requires starting an RCP to go back to forced flow conditions if possible.

D. Incorrect. Raising the steaming rate is the correct action per 19001-C step 4, this will be possible with the steam dumps while 2NAB re-energized. Starting RCP 1 is an incorrect action due to still being de-energized.

REFERENCES 19001-C, ES-O.l Reactor Trip Response steps 4 and 15 V-LO-PP-01101 Electrical Distribution presentation, slide 22 L-LO-PP-21201-12 Steam Dumps presentation, slide 91 VEGP learning objectives:

V-LO-PP-21 201-14:

Explain the operation of the steam dump system arming circuit.

V-LO-LP-3701 1-02:

State how the following control systems are employed to automatically stabilize the plant after a reactor trip:

a. steam dumps

b. feedwater

c. pressurizer level and pressure

d. auxiliary feedwater V-LO-LP-3701 1-04:

4

HL-15R RO NRC Exam State and describe the major action categories of 19001, Reactor Trip Recovery.

5

Approved By Procedure Number Rev J.B. Stanley Vogtle Electric Generating Plant 19001-C 31 Date Approved Page Number ES 0.1 REACTOR TRIP RESPONSE ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

• 4 *4 Check RCS temperature stable at IF temperature is less than 557°F or trending to 557°F. and lowering, THEN perform the following as

_With RCP(s) running RCS - necessary:

AVERAGE TEMPERATURE.

a. Stop dumping steam.

-OR-Without RCP(s) running RCS WR

- b. Perform the following as COLD LEG TEMPERATURES. appropriate:

IF at least one SG NR level greater than 10%,

THEN lower total feed flow.

-OR-IF all SG NR levels less than 10%,

THEN lower total feed flow to NOT less than 570 gpm.

_c. IF cooldown continues, THEN close MSlVs and BSlVs.

d. IF temperature less than 557°F and NOT trending to 557°F, THEN borate as necessary to maintain shutdown margin by initiating 13009, CVCS REACTOR MAKUP CONTROL SYSTEM.

_e. IF temperature greater than 557°F and rising, THEN dump steam.

Printed November 11,2009 at 15:25

Approved By . Procedure Number Rev J.B. Stanley Vogtle Electric Generating Plant 19001-C 31 DateApproved PageNumber ES 0.1 REACTOR TRIP RESPONSE ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

e. Control Tavg:

Manual control

-OR-Auto control

15. Check RCP status:

a. RCPs ALL STOPPED.

a. Go to Step 16.

_b. Start an RCP using _b. IF an RCP can NOT be started, ATTACHMENT A. (RCP 4 or THEN verify natural circulation RCP 1 preferred). using ATTACHMENT B.

IF natural circulation NOT established, THEN raise rate of dumping steam using Steam Dumps.

H Steam Dumps NOT available, THEN dump steam using SG ARVs.

Printed November 11,2009 at 15:25

Electrical Distribution OFFSI1E POWER EThMOddi.dRinRB R SOUROENO.2 I SOURCENOI TRANSFORMER UNIT AUX

Fkv 2NUA To4lGkV °4 ESF 2NAB 1-E BRURS .L I I I 1-B BRKR V-LO-PP-O11O1 Ver-02 22 This drawing shows the interconnection of the UATs and RATs with the 13.8kV distribution system. Note to the students how the Non-I E buses have the closest breaker to the RCPs classified as I -E. Also, the 13.8kV buses receive power from the RATs until power (turbine power 12004-C) gets to 30%,

then the buses are transferred to the UATs.

The next few slides will deal with the RX trips associated with the I 3.8kV buses. A drop in frequency would decrease RCP rotor-power output and the developed torque necessary to supply rated flow. A drop in line voltage would result in excessive reactor coolant pump motor currents which could damage the motor. Time delays on the UV and UF relays will prevent spurious signals from tripping the reactor.

V-LO-PP-OIIOI Ver-02 22

Steam Dumps OBJECTIVE LO-PP-21201-12 If no circulating water pumps are running or insufficient vacuum exists in the main condenser, dumping steam into the condenser can cause an overpressure condition which can damage the condenser. To protect against this, a permissive circuit prevents arming of the steam dumps. As can be seen in the figures below, the permissive circuit is composed of contacts which involve condenser vacuum and the circulating water pump breakers and their associated switchgear voltages (at 0 volts for 5.75 seconds). The condenser vacuum contact will be closed as long as two separate condenser pressure transmitters provide a signal indicating that at least 24.92 inches of mercury vacuum exists in the condenser. Each of the circulating water pump breaker contacts will be closed as long as the breaker for the associated pump is closed and the associated switchgear voltages are present. All of the above contacts are arranged so that the condenser must have at least 24.92 inches of mercury vacuum and at least one circulating water pump breaker closed with voltage applied. This will energize the permissive relay which closes the permissive contact in the arming circuit to permit arming of the steam dump valves. The permissive relay then energizes a C-9 Condenser Available permissive status light.

V-LO-PP-21 201 Rev-02 91

HL-15R RO NRC Exam

3. 003AK3.10 001/1/2/DROP ROD-RIL/PDIL/MEM 3.2 / 4.2/NEW/HL-15R NRC/RO/TNT/DS Given the following conditions:

- A dropped rod recovery is in progress while at power.

- 14915, Special Condition Surveillance Logs, Data Sheet 5 for Rod Insertion Limit Monitor Inoperable is being performed by the OATC.

Which ONE of the following choices correctly lists the bank and group of the dropped rod that will render the RIL monitor inoperable during the rod recovery?

A Control Bank C, Group 1 B. Control Bank D, Group 2 C. Shutdown Bank A, Group 1 D. Shutdown Bank B, Group 2 KIA 003 Dropped Control Rod AK3.10 Knowledge of the reasons for the following responses as they apply to the Dropped Control Rod.

RIL and PDIL.

K/A MATCH ANALYSIS The question presents a plausible scenario where a dropped rod recovery is in progress. Special Condition Surveillance Date Sheet 5 for RIL Monitor inoperable is being performed. The student must recall which rods (Control Banks group 1) input into the RIL monitor via the P to A Converter.

ANSWER I DISTRACTOR ANALYSIS A. Correct. Group 1 Control Banks input to the P to A converter which would render the RIL monitor inoperable during recovery. This requires a special condition surveillance to be performed.

B. Incorrect. Group 2 Control Banks do not input to the P to A converter. Plausible the candidate may think that Group 2 Control Banks input to the P to A converter versus the group 1 or that both groups input to the P to A converter.

C. Incorrect. Shutdown Banks Group 1 or Group 2 do not input to the P to A converter.

Plausible that the student could think either group inputs to the P to A converter to 6

HL-15R RO NRC Exam RIL.

D. Incorrect. Shutdown Banks Group 1 or Group 2 do not input to the P to A converter.

Plausible that the student could think either group inputs to the P to A converter to affect RIL since all the Shutdown Banks are required to be fully withdrawn to meet RIL.

REFERENCES V-LO-PP-27201, Digital Rod Position Indication System, slide # 8 V-LO-PP-27101, Rod Control System, slide #93, 100, 101, 102, and 103.

18003, Rod Control Malfunction, section A for Dropped Rods in Mode 1, page # 8, step #A18 and #A19.

14915-1, Special Conditions Surveillance Logs pages # 5 and # 15.

V-LO-TX-27101, Rod Control System, page #44.

VEGP learning objectives:

LO-PP-27101-15, State the inputs to the P to A Converter to inlude:

a. Inputs LO-LP-60303-04, Describe the effects of failing to reset the P to A Converter (Bank Demand Position Display) following a dropped rod retrieval.

LO-LP-60303-19, Given the entire AOP, describe:

a. Purpose of each step.

b. How and why each step is performed.

7

Digital Rod Position Indication System Control Shutdown ALARMS GW GW GW GW GW BANK A BANK B BANK C BANK D BANK A BANK B BANK C BANK i: BANK E CENTRAL 228 ee ce 228 228 ooooQø 228 228 **G Ge. Ge,. *OGG CONTROL 216 216 216 216 216 FAILURE 192 192 192 192 192 123 168 168 168 168 168 URGENT ALARM 144 144 144 144 144 123 120 120 120 DATA A FAILURE 96 96 96 120 96 120 96 7

123 72 72 72 72 72 DATA B 48 48 48 48 48 FAILURE 24 24 24 24 24 123 RB____ RB RB RB RB CSC c Dip2i 5Gftei SPB&V V - La - -

pp 5

I SCDE LOGIC ALARMS I URGENT AND NON URGENT ALARM POWER CABINET NON URGENTj LOGIC CABIMT ALARM POWER CABINET URGENT CIRCUITRY I

INHIBIT STEP GROUP STEPS COUNTERS SUPERVISORY IIo.

DATA LOGGING RELAY SHUTDOWN BANKS SUPERVISORY CIRCTRY BANK STEPS DRWERS i RODS MOVE OUT 7

BANK OVERLAP I STEP COMPLETE SLAVE CONTROL BANKS I UNIT H F CYCLERS I I P-A CONVERTER RELAYS BANK STEP = GROUP I POSiTION IPC ROD SUPERVISION QMCB GROUP STEP COUNTERS QMCB ALARMS V-LO-PP-27101 Rev-2.O 93 CA/y @-Pc) I (RL)

Group 1, rod M12 has been withdrawn from 180 to 190 steps to realign it to the bank. Which of the following is true?

A. Rod Deviation Monitoring and RIL Monitoring are both operable.

77.!

B. Rod Deviation Monitorin, Monitoring are botIj C. Rod Deviation Monitoring is inoperable and RIL Monitoring is operable.

D. Rod Deviation Monitoring is operable and RIL Monitoring is inoperable.

1 oft N V-LO-PP-27101 Rev-2 0 r 00 I A/It lIL

Group 2, rod M4 has been withdrawn from 180 to 190 steps to realign it to the bank. Which of the following is true?

A. When the P-A Converter is checked it should indicate 180 steps.

C. When the P-A Converter is checked it should indicate 200 (190 + 10) steps.

c P A V-LO-PP-27101 Rev-2.0 101

Group 1, rod M12 has been withdrawn from 180 to 190 steps to realign it to the bank. Which of the following is true?

A. When the P-A Converter and the demand position in the IPC are checked they should indicate 180 steps.

B. When the P-A Converter and the demand position in the IPC are checked they should indicate 190 steps.

C. When the P-A Converter and the demand position in the IPC are checked they should indicate 200 (190 + 10) steps.

V-LO-PP-27101 Rev-2.O 102

/

Group 2, rod M4 has been withdrawn from 180 to 190 steps to realign it to the bank. Which of the following is true?

A. When the demand position in the IPC is checked it should indicate 180 steps.

B. Whent indpositio T

chei C, When the demand position in the IPC is checked it should indicate 200 (190 + 10) steps. jD4 .

10 2 /V S V-LOPP271O1 Rev-2.0 103

PROCEDURE NO. REVISION NO. IPAGE NO.

VEGP 18003-C 23 8 of 27 A DROPPED RODS IN MODE 1 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED DA18. Check Rod being realigned - LJA18. Go to Step A20.

GROUP 1 CONTROL OR SHUTDOWN BANK ROD A19. Initiate 14915, SPECIAL CONDITIONS SURVEILLANCE LOGS:

4 9-(- 4-f- $ (-

LI. Rod Insertion Limit Monitor (if control bank) RL M J F LI. Rod Position Deviation Monitor 4 OU,tt2ci1 aVj EIA2O. Check Unit operation at or LIA2O. Limit dropped Rod withdrawal above 75% for at least 72 to 3 steps per hour.

cumulative hours in a 7 day period.

LIA21. Record the affected banks group step counter positions in the Unit Control Log.

Approved By I Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant 14915-1 45 Date Approved Page Number SPECIAL CONDITIONS SURVEILLANCE LOGS 6/29/09 I 5 of 40 Sheet 1 of 2 Table I Special Conditions Surveillance Requirements APPLICABLE DATA SPECIAL CONDITION SURVEILLANCE MODE TECH SPEC FREQUENCY SHEET Reactor Critical, Tavg- Minimum 1 and 2 SR 3.4.2.1 Once per 30 1 Tref Dev Alarm not Temperature for mm reset and any RCS Criticality loop Tavg <561°F Rod Position Deviation Rod Group 1 and 2 SR 3.1.4.1 Once per 4 3 Monitor inoperable Alignment Limits hours Rod Position Rod Position 1 and 2 LCO 3.1.7 Once per 8 4a Indication System Indication System hours (Actions Al, inoperable Operating B.1)

Rod Demand Rod Demand 1 and 2 LCO 3.1.7 Once per 8 4b Indication System Indication System - (Actions C.1.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> inoperable Operating C.1.2 Rod Insertion Limit Control Rod 1 and 2 SR 3.1.6.2 Once per 4 5 Monitor inoperable Insertion Limits hours Axial Flux Difference Axial Flux 1 >50% SR 3.2.3.1 Once per hour 6 Monitor Alarm Difference inoperable Quadrant Power Tilt Quadrant Power 1 >50% SR 3.2.4.1 Once per 12 7 Monitor Alarm Tilt Ratio hours inoperable Quadrant Power Tilt Quadrant Power 1 >50% LCO 3.2.4 Once per 12 7 Ratio >1.02 Tilt Ratio (Action A.2.1) hours One Power Range NI Quadrant Power 1 >75% SR 3.2.4.2 Once per 12 7 Channel inoperable Tilt Ratio hours Primary or Secondary S/G & RCS Press At all times TRS 13.7.1.1 Once per hour 8 Temperature <70°F Limitations Two Source Range NI RCS Boron 6 LCO 3.9.3 Once per 12 9 Channels inoperable hours (Action B.2)

Printed October 15, 2009 at 13:21

Approved By Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant 14915-1 45 Date Approved Page Number 6/29/09 SPECIAL CONDITIONS SURVEILLANCE LOGS 15 of 40 DATA SHEET 5 Sheet 1 of 1 CONTROL ROD INSERTION LIMITS WITH ROD INSERTION LIMIT MONITOR INOPERABLE

1. Record each Control Rod Bank Step Counter Demand position and each Rod Bank Insertion limit once every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

2. Verify each Rod Bank Step Counter Demand position is greater than or equal to the Bank Insertion limit.

3. If any Rod Bank Step Counter Demand position is below the Bank Insertion limit, within 15 minutes initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or equivalent until the required shutdown margin is restored. (Technical Requirement 13.1.1)

Date Odly o&(

Time CBA DEMAND CBA INSERTION LIMIT CBB DEMAND CBB INSERTION LIMIT CBC DEMAND CBC INSERTION LIMIT CBD DEMAND CBD INSERTION LIMIT VERIFIED Shift Supervisor Review: /____

Initial Date Time Printed October 15, 2009 at 13:21

The Supervisory Logic receives the signal and generates the following when the In-HoldOut Switch demands movement:

• signal to the Pulser oscillator to generate pulses for 48 steps/mm.

• signal to the Pulser oscillator to allow oscillator to generate timing pulses.

• signal to the Bank Overlap Unit to select which control bank or banks to move. In this case, both control Banks A and B will move.

• signal to the Master Cycler allowing fast pulses.

• signal to the Master Cycler specifying the direction to move.

• signal to the Slave Cycler specifying the direction to move The Bank Overlap Unit will receive the rod motion and direction demand and will generate the following:

• signal to the Master Cycler specifying which shutdown bank to move.

• signal to the Power Cabinet multiplexing relays specifying which group in the cabinet to move.

The Master Cycler fast steps. At step 0 the Master Cycler sends a GO pulse to slave cycler 1AC and 1BD and signal to supervisory logic that rod motion is started. The Supervisory logic then sends a signal to the Master Cycler inhibiting fast pulse and the Master Cycler continues advancing its counter at the pulser rate (48 steps/mm) . At counter step 3, slave cyclers 2AC and 2BD receives a go pulse. This cycle repeats until demanded motion stops.

On receipt of a Go pulse, the Slave cyclers generate current orders for one step of rod motion and send them to their respective Power Cabinet.

They also generate the following:

• signal to the data logging circuits of the supervisory logic when a step is complete. The Supervisory logic sends this signal to the IPC computer for rod deviation monitoring, signals the Demand counters to step, and, if a group 1 rod, sends a signal to the P/A Converter for the RIL monitor.

• Alarm signals if a fault is detected in its monitored circuits.

The Power Cabinets will generate the currents requested by the slave cycler and the rods will move. The group 1 Control Bank A and B rods will move out a step, then the Group two rods will move out.

D. MOVEMENT OF CONTROL BANKS IN AUTO The Bank selector switch must be selected to AUTO. For this discussion assume CBD is to be moved in from 228 steps.

The Supervisory Logic receives the signal and generates the following when the T-avg Control System demands movement:

• signal to the Pulser oscillator to generate pulses for 8 to 72 steps/mm depending on the T-avg control system.

• signal to the Pulser oscillator to allow oscillator to generate timing pulses.

44 Revision 3.2

HL-15R RO NRC Exam

4. 003G2.4.35 OO1/2/l/RCP-LOCAL SO ACTIONS/C/A -3.8 / 4.O/NEW/HL-15RNRC/RO/TNT/DS Given the following plant conditions:

- 19010-C, E-l .0 Response to Loss of Reactor or Secondary Coolant in effect.

- RCS pressure is 1670 psig and stable.

The following annunciators illuminate:

RCP SHAFT RCP SHAFT VIBRATION HI VIBRATION ALERT Which ONE of the following correctly identifies where RCP vibrations are read and the action to take based on a shaft vibration of 20.2 mils?

Location Action to take A. IPC computer points Immediately trip the affected RCP(s)

B. IPC computer points Continue RCP operation and monitor vibrations C Locally in Control Building Immediately trip the affected RCP(s)

D. Locally in Control Building Continue RCP operation and monitor vibrations 8

HL-15R RO NRC Exam KIA 003 Reactor Coolant Pump System (RCPS)

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

KIA MATCH ANALYSIS The question presents a plausible scenario where RCP vibration alarms are received during EOP performance. The student must know where the vibration is monitored and action to take in response to given vibration values.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Monitored locally in the Control Building, plausible due to other large components such as MFPTs and Main Turbine as well as several RCP parameters may be checked on the IPC. High shaft vibration indicates the immediate trip criteria has been exceeded. Tripping the RCP is the correct action.

B. Incorrect. Monitored locally in the Control Building, plausible due to other large components such as MFPTs and Main Turbine as well as several RCP parameters may be checked on the IPC. Both shaft alert and high level alarms indicate the immediate trip criteria is exceeded, allowing the pump to run after the one that is identified to have a problem is an incorrect action.

C. Correct. Control building is correct location to monitor. The high shaft vibration alarm indicates the immediate trip criteria has been exceeded. Once the problem RCP has been identified, it should be tripped.

D. Incorrect. Control building is correct location. Immediate RCP trip is required for a valid hi RCP shaft vibration alarm. The alarm is validated by locally reading the RCP vibrations. To allow the pump to continue to run is an incorrect action.

REFERENCES 17008, ARP for windows E04 and F04. Pages 38 and 39 and pages 43 and 44.

13003, Reactor Coolant Pump Operation Limitation 2.2.10 V-LO-PP-1 6401, Reactor Coolant Pumps slide #26 VEGP learning objectives:

Not applicable.

9

Approved By Procedure Number Rev C. H. Williams, Jr. i Vogtle Electric Generating Plant 17008-1 13.3 DateApproved I ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 08 ON Page Number 1/1/2004 I PANEL 1A2ON MCB 3of46 (1) (2) (3) (4) (5) (6)

A RCP I RCP 1 RCP 1 RCP 1 RCP 1 MTR OVERLOAD STANDPIPE STANDPIPE NO. 2 SEAL LKOF CONTROLLED LKG LO LEVEL HI LEVEL HI FLOW HI/LO FLOW B RCP2 RCP2 RCP2 RCP2 RCP2 MTR OVERLOAD STANDPIPE STANDPIPE NO. 2 SEAL LKOF CONTROLLED LKG LO LEVEL HI LEVEL HI FLOW HI/LO FLOW C RCP3 RCP3 RCP3STANDPIPE RCP3 RCP3 MTR OVERLOAD STANDPIPE HI LEVEL NO. 2 SEAL LKOF CONTROLLED LKG LO LEVEL HF FLOW HI/LO FLOW D RCP4 RCP4 RCP4 RCP4 RCP4 RCP MTR OVERLOAD STANDPIPE STANDPIPE NO. 2 SEAL LKOF CONTROLLED LKG NO. 1 SEAL LO LEVEL HI LEVEL HI FLOW HI/LO FLOW LO i\P E RCP FRAME RCP SHAFT RCP VIBRATION VIBRATION SEAL WATER INJ ALERT ALERT FILTER HI P F RCP FRAME RCP SHAFT RCP HI VIBRATION HI VIBRATION SEAL WATER INJ LO FLOW Printed December 17, 2009 at 13:35

Approved By Procedure Number Rev C. H. Williams, Jr. Vogtle Electric Generating Plant m. 17008-1 13.3 DateApproved ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 08 ON Page Number 1/1/2004 PANEL 1A2ON MOB 38 of 46 WINDOW E04 ORIGIN SETPOINT RCP SHAFT 1-XE-0471C,D 15 MILS VIBRATION 1-XE-0472C,D ALERT 1 -XE-0473C, D 1 -XE-04740, D 1.0 PROBABLE CAUSE

1. RCS operating temperature below 500° F.

2. Pump Bearing failure.

3. Pump Impeller shaft assembly out-of-balance.

4. Misalignment between Pump Shaft and Motor Shaft.

5. Loose connections or disconnected vibration probes.

6. Vibration Monitoring Panel power failure.

7. Local COMMON RESET not cleared.

2.0 AUTOMATIC ACTIONS NONE 3.0 INITiAL OPERATOR ACTIONS NONE Printed October 16, 2009 at 16:38

Approved By Procedure Number Rev C. H. Williams, Jr. Vogtle Electric Generating Plant 17008-1 13.3 DateApproved ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 08 ON Page Number 1/1/2004 PANEL1A2ONMCB 39 of 46 WINDOW E04 (Continued) 4.0 SUBSEQUENT OPERATOR ACTIONSS NOTE The Vibration Monitoring Panel displays auctioneered high vibration levels.

Loc- ( 1. Dispatch an operator to the Vibration Monitoring Panel 1-1201-P5-VMP to:

a. Identify the Reactor Coolant Pump (RCP) causing the alarm.

b. Check both vibration channels and alarm setpoints for shaft and frame of each RCP (32 points in all) to verify no obvious vibration monitoring equipment problems exist.

c. Attempt to reset alarm using COMMON RESET toggle switch.

2. Continue operation of affected RCP and frequently monitor vibration.

tLti Refer to 13003-1, Reactor Coolant Pump Operation and shut down the

/ 3.

affected RCP if rate of increase in vibration exceeds 1 MIL/hour.

crJ+

5.0 i COMPENSATORY OPERATOR ACTIONS QLCUi 5pf END OF SUB-PROCEDURE

REFERENCES:

1 X4DB1 13, 1 X6ABO6-1 19, 1 X3D-BD-M01 A, 1 X3D-CD-M1 OA, 1 X6ABO9-88, CX5DT1O1-176A, CX5DT1O1-176B Printed October 16, 2009 at 15:38

Approved By . Procedure Number Rev C. H. Williams, Jr. Vogtle Electric Generating Plant z 17008-1 13.3 Date Approved ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 08 ON Page Number 1/1/2004 PANEL1A2ONMCB 43 of 46 WINDOW F04 ORIGIN SETPOINT RCP SHAFT 1-XE-0471C,D 20MILS HI VIBRATION 1 -XE-0472C,D 1 -XE-0473C, D 1 -XE-0474C, D 1.0 PROBABLE CAUSE

1. RCS operating temperature below 500°F.

2. Pump Bearing failure.

3. Pump Impeller shaft assembly out-of-balance.

4. Misalignment between Pump Shaft and Motor Shaft.

5. Loose connections or disconnected vibration probes.

6. Vibration Monitoring Panel power failure.

7. Local COMMON RESET not cleared.

2.0 AUTOMATIC ACTIONS NONE Printed December 17, 2009 at 13:38

Approved By Procedure Number Rev C. H. Williams, Jr. Vogtle Electric Generating Plant 17008-1 13.3 Date Approved ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 08 ON Page Number 1/1/2004 PANEL 1A2ON MCB 44of46 WINDOW F04 (Continued)

NOTES

• Prompt action is required to confirm alarm validity and shut down affected RCP if required.

• The Vibration Monitoring Panel displays auctioneered high vibration levels.

3.0 INITiAL OPERATOR ACTIONS

1. Attempt to confirm validity of annunciator through related plant parameters.

2. Dispatch an operator to the Vibration Monitoring Panel 1-1201-P5-.VMP to:

a. Identify the Reactor Coolant Pump (RCP) causing the alarm.

b. Check both vibration channels and alarm setpoints for shaft and frame of each RCP (32 points in all) to verify no obvious vibration monitoring equipment problems exist.

c. Attempt to reset alarm using COMMON RESET toggle switch.

3. Refer to 13003-1, Reactor Coolant Pump Operation and shut down the affected RCP.

4.0 SUBSEQUENT OPERATOR ACTIONSS

• NONE 5.0 COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE

REFERENCES:

1X4DB113, 1X6ABO9-119, 1X3D-BD-MO1A, IX3D-CD-M1OA, 1X6ABO9-88 Printed December 17, 2009 at 13:38

Approved By Procedure Number Rev C. S. Waidrup Vogtle Electric Generating Plant 13003-1 40 Date Approved Page Number 11/5/08 REACTOR COOLANT PUMP OPERATION 5 of 35 2.2.9 During RCS filling and venting, RCS pressure must be greater than 325 psig prior to starting an RCP to verify adequate seal DIP is maintained throughout RCS fill and vent. If necessary, the RCP should be stopped prior to seal D/P dropping less than 200 psid. If the seal D/P goes below 200 psid during pump operation or coast down, the RCP should be evaluated before restarting the RCP.

2.2.10 An RCP shall be stopped if any of the following conditions exist.

• Motor bearing temperature exceeds 195°F.

• Motor stator winding temperature exceeds 311 °F.

• Seal water inlet temperature exceeds 230°F

• Total loss of ACCW for a duration of 10 minutes.

• RCP shaft vibration of 20 mils or greater.

• RCP frame vibration of 5 mils or greater.

• Differential pressure across the number 1 seal of less than 200 psid.

2.2.11 If a loss of RCP seal cooling (Seal Injection and/or ACCW to Thermal barrier) occurs, resulting in RCP shutdown due to exceeding operating limits, then the unit should be cooled down to Mode 5 to facilitate recovery. Upon reaching Mode 5, ACCW to the Thermal barrier should be restored. Seal injection should then be returned to service. This sequence should prevent seal damage, RCP shaft bowing, ACCW System damage, etc. due to excessive thermal stresses.

Printed December 17, 2009 at 13:42

1) RCP Shaft Vibration

• Measured by a vertical and horizontal proximity probes mounted parallel and perpendicular respectively to the pump discharge at a location near the pump coupling.

• Continuous monitoring

• Alarm in control room alert and high (Alert 15 mils / High 20 mils)

• greater than 15 mils could require pump shutdown (indicative of bearing failure)

• Must trip at 20 mils

2) RCP Frame Vibration

• 2 probes 900 apart mounted at the top of the motor frame

• Continuous monitoring

• Alarm in the control room alert and high (Alert 3 mils I High 5 mils)

• greater than 3 mils could require pump shutdown (indicative of misalignment or a out of balance condition)

• Musttripat5mils Vibrations checked on a daily bases on control building rounds for trending purposes.

Vibrations should be monitored locally during RCP startup.

V-LO-PP-1 6401 Rev-03 26

HL-15R RO NRC Exam

5. 004G2. 1.23 001/2/1 /CVCS-PROCEDURES/C/A 4.3 /4.4/NEW/IlL-i 5R NRC/RO/DS/TNT Given the following conditions:

The reactor is at 100% power PRZR level is slowly lowering due to a 50 GPM RCS leak 120 GPM CVCS letdown is in service RCP seal injection flow is 8 GPM per pump Charging flow controller FIC-0121 is in automatic Which one of the following lists the correct system response and procedurally directed operator actions for this condition?

A. Charging flow will automatically increase due to PRZR level dropping below the program level.

The increase in charging flow will automatically maintain PRZR level at the program level.

B Charging flow will automatically increase due to PRZR level dropping below the program level.

Letdown flow will have to be isolated and charging flow will have to be manually adjusted to approximately 62 GPM to maintain a constant PRZR level.

C. Charging flow will automatically increase due to PRZR program level changing.

The increase in charging flow will automatically maintain PRZR level at the program level.

D. Charging flow will automatically increase due to PRZR program level changing.

Letdown flow will have to be isolated and charging flow will have to be manually adjusted to approximately 62 GPM to maintain a constant PRZR level.

KIA 004 Chemical and Volume Control System G2.1 .23 Ability to perform specific system and integrated plant procedures during all modes of plant operation.

KIA MATCH ANALYSIS Thc nustion renuires the student to correctly determine the CVCS system resnonse to 10

HL-15R RO NRC Exam an RCS leak and the appropriate AOP actions for the CVCS system to stabilze PRZR level matching the K/A topic.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Charging flow will increase due to PRZR level dropping below the program level. Since 120 GPM CVCS letdown is in service, the available increase in charging flow will not keep up with the leak. Maximum charging flow will be approximately 135 GPM.

B. Correct. Charging flow will increase due to PRZR level dropping below the program level. Since 120 GPM CVCS letdown is in service, the available increase in charging flow will not keep up with the leak. The AOP will direct the operator to isolate letdown.

Charging flow will then need to be adjusted to 62 GPM to offset the 50 GPM leak and the 12 GPM RCP seal leakoff flow.

C. Incorrect. Charging flow will increase due to the drop in measured PRZR level.

Program level will remain the same because Tave will not change as a result of the RCS leak. Since 120 GPM CVCS letdown is in service, the available increase in charging flow will not keep up with the leak. Maximum charging flow will be approximately 135 GPM.

D. Incorrect. Charging flow will increase due to the drop in measured PRZR level.

Program level will remain the same because Tave will not change as a result of the RCS leak. Since 120 GPM CVCS letdown is in service, the available increase in charging flow will not keep up with the leak. The AOP will direct the operator to isolate letdown. Charging flow will then need to be adjusted to 62 GPM to offset the 50 GPM leak and the 12 GPM RCP seal leakoff flow.

REFERENCES AOP 18004-C, RCS Leakage step A3 V-LO-TX-1 6001, Primary Systems pages 73, 74, AND 75 for PRZR level control VEGP learning objectives:

LO-PP-1 6302-03:

Describe how Pressurizer level control maintains level on program.

LO-LP-60304-09:

Given the entire AOP, describe:

a. Purpose of selected steps.

b. How and why the step is being performed.

c. Expected response of the plant/parameter(s) for the step.

11

Approved By Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant 18004-C 24 Date Approved Page Number REACTOR COOLANT SYSTEM LEAKAGE

( 3/9/09 5 of 80 A. RCS LEAKAGE (MODE 1, 2, AND 3 WITH RCS PRESSURE >1000 PSIG)

ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED Al. Check plant conditions: Al. Go to the appropriate section of this procedure:

In Mode 1 or 2. _SECTION B. RCS LEAKAGE (MODE 3 <1000 PSIG AND 4)

-OR- -OR-In Mode 3 with RCS pressure greater SECTION C. RCS LEAKAGE than 1000 psig. (MODE 5).

A2. Initiate the Continuous Actions Page.

• A3. Maintain PRZR level:

_a. Adjust charging flow as necessary to maintain program level.

_b. Check PRZR level STABLE OR b. Perform the following:

RISING.

1) Isolate letdown by closing:

a) Letdown Orifice Valves.

_b) Letdown Isolation Valves.

_c) Excess Letdown Valves.

2) Start an additional Charging Pump as necessary.

Step 3 continued on next page Printed October 28, 2009 at 14:43

LO _

It is important to understand the principles of operation and the limitations of these levels transmitters.

AP type level transmitters that are calibrated for normal operating conditions may be inaccurate under abnormal conditions such as a LOCA or steam line break in the containment. Specifically, the reference leg piping and condensing pots are exposed to the containment atmosphere. At elevated containment temperatures, the reference leg pipe and the water it contains will heat up, decreasing the density of the reference leg. This causes the indicated level to be greater than the actual level. As the reference leg is heated up, the volume of the water in the reference leg increases and forces some of the liquid from the reference leg. The pressure that the reference exerts on the level transmitter is less than the pressure that the water exerted prior to being heated up. The result is an indicated change in level (increase) although the actual level may not have changed. The severity of the error will depend on the actual containment conditions. Redundant level channels of the pressurizer may uniformly present inaccurate indications and, under such conditions, must be considered unreliable. Other conditions that may affect pressurizer level indication are reference leg leaks or partial draining due to instrument calibration or other maintenance activities, and a phenomena caused by hydrogen gas coming out of solution in the reference leg. Since the reference leg temperature is cooler than the pressurizer, it has a higher affinity for absorbing hydrogen gas. The hydrogen gas could come out of solution during transients. The results from all of the above mention would be reduction in the AP. This reduction in AP would cause the pressurizer level indication being higher than actual.

16-58 Pressurizer Level Control System PRESSURIZER LEVEL II I 1 U 465A 1 U 456A 1 U 461 The Pressurizer Level Control System utilizes three hot calibrated level channels (LT-459, LT-460, and LT-461) for control. Two of these channels are selected at any given time by a three-position selector switch (LS-459D). The possible combinations are: Channels LT-459 and LT-460, Channels LT 461 and LT-460, or Channels LT-459 and LT-461. Only channels LT-459 and LT-461 can be selected for primary level control and only channels LT-460 and LT-461 can be selected for secondary control. The pressurizer level control uses the primary channel input to compare its value to the calculated level set point to control pressurizer level. The secondary channel is used for protection only. A three-position recorder selector switch (LS-459E) is provided to select the actual level to be recorded along with the program level on LR-459.

The reference level signal is generated by auctioneered high Tavg (No-load Tavg 557°F, to 100% Full Power Tavg of 586.4°F) which generates a program level of 25%

to 57.8%, which corresponds to the difference between No-load Tavg and full load Tavg. The program level is compared to one of the selected level channels, LT459 or LT461, to produce a level error signal.

The level error produced is used as input by the master level controller. The master controller is sensitive to both the magnitude of the difference and the time duration that the difference is present. A large level error will result in a large controller output. The integral portion of the controller will also FR LVL CUT I,Ei :c produce a high output for small errors that are present for long time durations. To change the level in the pressurizer, either the temperature or the mass balance of the RCS must change.

The master controller responds to level errors by changing CVCS charging flow.

During steady state operation with no pressurizer level change, CVCS letdown flow is equal to CVCS charging. If charging flow changes and letdown flow remains constant, then the mass 73 Revision 7.2

v- Lo-) /6oO/

balance of the RCS will change. Pressurizer level control operates on this principal. Charging flow is varied by controlling the position of the flow control valve (FCV-1 21).

The demand signal from the level master controller (LIC-459) is sent to the charging flow controller (FIC 121). The charging flow controller compares demand flow from the pressurizer level master controller with actual flow. If there is a difference between the two, the controller will position FCV-121 accordingly to correct the error. Both controller MANUAL/AUTO stations are located on the C panel in the control room.

The Pressurizer Level Control System will automatically isolate CVCS letdown when pressurizer level decreases to 17%. Both letdown isolation valves, LV-459 and LV-460, close as well as the letdown orifice isolation valves. This prevents draining the pressurizer if a leak occurs in CVCS system. Damage would occur if the heaters were energized and not fully immersed in water. Therefore, the level control system also de-energizes the pressurizer heaters when the water level decreases to 17%. This prevents damage to the wall of the pressurizer vessel due to overheating and to the heaters themselves. The heaters would be exposed if the pressurizer level decreased below 14%. (See Pressurizer Level Control Logic Drawing)

The pressurizer Level control system is designed to accommodate the following without a reactor trip:

a. Ramp unloading rate of 5% per minute with auto rod control.

b. Instantaneous load reduction of 10% with auto rod control.

c. Step load reduction of 50% with both auto rod control and steam dump control.

Level control selector switch LS-459D To further explain its operation the following example is given:

Level transmitter 459/460 is selected on LS-459D LT-459 is selected as the primary channel for the master level control. If the level that is sensed by LT 459 drops to = 17% it will cause the following to occur:

a. CVCS Charging Flow increases by opening FCV-1 21

b. All pressurizer heaters will automatically trip.

c. CVCS Letdown Isolation valve LV-459 will automatically close.

d. All three CVCS Letdown Orifice Isolation valves will automatically close.

LT-460 is selected for the secondary channel. If level sensed by LT-460 drops to 17% it will cause the following to occur:

a. All Pressurizer heaters will automatically trip.

b. CVCS Letdown Isolation valve LV-460 will automatically close.

c. All three CVCS Letdown Orifice Isolation valves will automatically close.

If the primary level control channel sense pressurizer level = 5% above program pressurizer level, a signal is generated that energizes the pressurizer backup heaters. Alarm ALBI 1-COl Przr Hi Level Dev and heaters on annunciates. The purpose for this design is to heat the in surge of water to saturation in anticipation of a possible sudden out surge to maintain pressurizer pressure. Typical pressurizer temperatures are as follows:

• Pressurizer Surge Line Temperature 645°F 74 Revision 7.2

v Lo))6ooi

• Pressurizer Liquid Temperature 650°F

• Pressurizer Steam Space Temperature 650-655°F This example applies to all possible selections on LS-459D.

16-59 Pressurizer Level Protection System The Pressurizer Level Protection System also utilizes the same level transmitters as the Control System.

The level indications provide the information to the Reactor Protection System (RPS). The Reactor Protection System will automatically trip the reactor if the pressurizer level reaches a high level set point of 92% when the reactor is above 10% power. This function however looks at all three level channels and is not based on the switch position of LS-459D. Reactor trip will occur if two out of the three level transmitters are indicating =92%. This Reactor trip function protects the RCS from the over pressurization Pressurizer E9 AuctHTavg Level Controlling Channel L E

setpont 5% V E

L Lo Level Deviation Tavg RBRJ.

SETPOINT

[iR Level Rec 4 an

[1

- CLOSE LV. 459

- CLOSE Orifice Isolation VIvs 17% Setpont LtC-459

- Heaters OFF MASTER LEVEL 9 4PRZR Lo CONTROL Level, Htr CntI OFF, Ltdn Secure that might occur if the pressurizer were to go water solid (Loss of bubble). When the plant is shut down, the pressurizer is cooled down and is allowed to go water solid. The high level trip is automatically disabled by the RPS trip permissive P-7, which happens when the reactor power decreases to <10%.

75 Revision 7.2

HL-15R RO NRC Exam

6. 004K6.31 OO1/2/1/CVCS-SEAL INJ LIMITS/C/A -3.1 / 3.5/M- LOIT BANXJHL-15RNRC/RO/TNT/DS Which one of the following choices lists the correct action to take for the given RCP seal injection flow?

RCP Seal Injection Flow Action to take A. 7.8 gpm Press UP arrow to throttle open HV-0182.

B. 13.4 gpm Press UP arrow to throttle closed HV-O1 82.

C 13.4gpm Press DOWN arrow to throttle open HV-0182.

D. 7.8 gpm Press DOWN arrow to throttle closed HV-O1 82.

12

HL-15R RO NRC Exam KIA 004 Chemical and Volume Control System K6.31 Knowledge of the effect of a loss or malfunction on the following CVCS components.

Seal injection system and limits on flow range.

K/A MATCH ANALYSIS The question presents a plausible scenario where seal injection flows are out of limits.

Proper OATC actions and HV-182 response to these actions are required.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. UP arrow raises flow by closing valve. Correct button to depress, however HV-182 closes, not opens. Wrong response to the action.

B. Incorrect. UP arrow raises flow by closing valve. Wrong button to depress, HV-182 response to the action is correct.

C. Correct. DOWN arrow lowers flow by opening valve. Depress the down arrow to open HV-182 and divert less flow to the seals.

D. Incorrect. DOWN arrow lowers flow by opening valve. Wrong button to depress, HV-182 opens, not closes. Wrong response to the action.

REFERENCES 004A4.1 1, LOIT Exam Bank previously used on HL-1 3 RO Retake Exam (not in last 2).

Question used as base for modification.

Vogtle Text Chapter # 9 for CVCS pages # 21 and # 26 SOP-13003-112, Reactor Coolant Pump Operation limitation 2.2.3.

VEGP learning objectives:

LO-PP-16401-03 Describe the control room indications for a failure of a RCP seal.

13

1. 004A4.11 001 Which ONE of the following actions would be CORRECT to take using HC-O1 82 given the provided seal injection flow indication?

A. seal injection flows 13.8 gpm per RCP, depress the red UP arrow to throttle open HV-0182 to lower seal injection flow.

B seal injection flows 7.4 gpm per RCP, depress the red UP arrow to throttle closed HV-0182 to raise seal injection flow.

C. seal injection flows 7.4 gpm per RCP, depress the green DOWN arrow to throttle open HV-0182 to raise seal injection flow.

D. seal injection flows 13.8 gpm per RCP, depress the green DOWN arrow to throttle closed HV-0182 to lower seal injection flow.

e b) a Page: 1 10/15/2009

through the seal leak off isolation valves (HV-Sl4lA, B, C, and D), through a motor-operated isolation valve (HV-8ll2), and then exits the containment building. The

( seal return flow immediately passes through a second motor-operated isolation valve (HV-8l00) upon exiting the containment. Both of these motor-operated isolation valves serve to isolate the containment upon receiving a Containment Isolation Actuation(CIA) signal. Seal water return flow next passes through the seal water return filter which removes any insoluble material picked up as the seal water passed through the reactor coolant pump seals. It is then reduced in temperature from approximately 175°F to 130°F as it passes through the tube side of the seal water heat exchanger before returning through an isolation valve to the suction header of the charging pumps.

FIGURE 9-3 CL CVCS CHARGING LPI and SEAL INJECTION CL LP4 PZR OTHER LVI 12A RHUT RCPS IRC ORG NCP SEAL 21 Revision 4.1 L. 1.

Charging Flow Control Valve FV-l21 The discharge of the NCP and CCP5 combine into a single charging flow path.

Charging flow is controlled by the position of FV-121. In automatic, FV-12l controls the total flow directed toward the normal charging header. The position of FV-l2l is determined by the output error signal from pressurizer level controller FIC-12l. This error signal is determined by the difference between pressurizer program level (determined by auctioneered high Tavg) and actual pressurizer level.

If an output error signal indicates that pressurizer level is below program level, FV-12l will open to provide more charging flow to eliminate that error signal.

Conversely, if pressurizer level is above program level, then FV-l21 valve position will throttle more closed to lower charging flow. Charging flow controller FIC-l2l can be operated locally in the Auxiliary Building. Charging flow indication is provided in the control room on panel A and C.

Seal Flow Control Valve, HV-l82 This hand controller air-operated valve in the charging header maintains sufficient backpressure in the charging header to ensure adequate flow of seal water to the reactor coolant pumps. The flow indicators, (FI-142, -143, -144, and -145) for each RCP seal injection are used to adjust the setting of this valve so that approximately 8-13 gpm seal injection flow is maintained to each RCP. The valve is manually controlled from the main control board. The valve fails open on loss of power or air. If more seal injection flow is required, the operator depresses the UP pushbutton on HC-182. This causes HV-182 to be in a more shut position, thus forcing more flow toward the RCP seal injection line. This has the immediate effect of lowering charging flow directed toward the normal charging header. With letdown flow the same and now less charging flow going through the normal charging header, letdown temperature out of the regenerative heat exchanger will increase.

Consequently, anytime seal injection flow is adjusted, the effect on letdown parameters must be evaluated. Vice versa, anytime charging flow is changed, the effect on seal injection and letdown parameters must be evaluated also and appropriate actions to restore system parameters to their normal operating band.

Normal Charging header Isolation Valves, HV-8106, HV-8105 Charging flow that does not flow toward the seal package (based on the position of HV-l82) flows past HV-182 toward the RCS penetration past series charging isolation valves HV-8l06 (Train A) and HV-8l05 (Train B). These valves are operated from the main control room and from the shutdown panels. Each valve is powered from a 1E MCC. On a Safety Injection actuation signal, these valves receive a CLOSE signal from their respective train related SI signal. This isolates the normal charging header and allows the safety-related CCPs to direct their flow through the BIT into all RCS cold legs.

Normal/Alternate Charging to RCS Isolation Valve (HV-8l46 and HV-8l47)

Control switches for these motor-operated isolation valves are located on the QMCB and the Remote Shutdown panels. These switches are two position (Close/Open)

An additional switch Control Room/Local transfer switch is located on the Remote Shutdown panel. This switch must be in the Control Room position to enable the QMCB switches. To equalize thermal stresses, SOP 13006-1/2 states that normal charging valve HV-8146 should be in service during even numbered fuel cycles and alternate charging valve HV-8147 should be used during odd-numbered fuel cycles. This should be performed at cold shutdown conditions to avoid thermal transients.

CVCS Pressurizer Auxiliary Spray Valve (HV-8l45)

This air-operated isolation valve is operated from the QMCB and the Remote Shutdown panel. These switches are two position (Close/Open) . There is also a 26 Revision 4.1

Approved By Procedure Number Rev Vogtle Electric Generating Plant C. S. Waidrup 13003-1 40 Date Approved Page Number 11/5/08 REACTOR COOLANT PUMP OPERATION 3 of 35 2.1.5 When starting the first RCP with a bubble in the Pressurizer, the additional RCP heat input may cause an insurge of cooler RCS water into the pressurizer.

Surge line temperature may be controlled by monitoring surge line temperature and adjusting RHR cooling and charging flow to verify a net outsurge from the pressurizer.

2.1.6 With Westinghouse and Operations management approval, RCP5 may be started without ACCW flow to perform 30 second and 1 minute air sweeps per 13001, Reactor Coolant System Filling and Venting or to verify proper rotation following electrical maintenance (less than 1 minute). General Manager approval will be required for starting RCPs without ACCW for any other operation. Operation without ACCW in service for more than 10 minutes is prohibited.

2.1.7 Seal Injection flow should be maintained to coupled RCP5 when RCS level is greater than the 190 foot elevation, however, if necessary, seal injection may be secured to RCPs above the 190 foot elevation provided RCS level is maintained constant.

2.1.8 RCPs should NOT be uncoupled and placed on their back seat until the RCS is depressurized and vented.

2.2 LIMITATIONS 2.2.1 If seal injection is NOT in service AND the reactor coolant temperature is greater than 150°F, Auxiliary Component Cooling Water shall be supplied to the thermal barrier.

2.2.2 When the reactor coolant pressure is less than 100 psig, the No. 1 Seal Leakoff Valves should be closed.

2.2.3 The RCP seal injection flow should be maintained greater than 8 gpm and less than 13 gpm any time seal injection is required.

2.2.4 With the reactor coolant temperature greater than 400°F, the seal injection temperature should be maintained less than 135°F.

Printed October 15, 2009 at 16:38

CVCS Charging HV-182 RCP Seal Injection 32 gpm FV-121 Charging 55 gpoi Pwnp Flow Normal chaiging Header to RS V-LO-PP-09200-02. t 151 Charging FV-l21 and HV-182 are air operated valves that FAIL OPEN on loss of air.

Seal injection flow is controlled by manually positioning HV- 182 to provide backpressure against the charging header to force flow towards the RCP seal injection header.

Consequently, when both valves fail open, then charging flow goes to its maximum value.

If this malfunction was due to a loss of instrument air, procedure 18028-C provides guidance to establish Safety Grade Charging per SOP 13006-1/2.

Charging flow is controlled using the Safety Grade charging flow controller(HC- I 90A or HC- I 90B).

Seal injection is controlled locally by locally opening 1208-U6-l51 or 152.

V-LO-PP-09200-02. 1 151

HL-15R RO NRC Exam

7. 005K2.03002/2/1/RHR-MOV POWER/MEM 2.7/ 2.8/M- LOIT BANKIHL-I5RNRC/RO/TNT/DS Which ONE of the following CORRECTLY describes the power supplies to the RHR loop suction isolation valves?

A. All 4 loops suctions are powered from 1 E 480V MCC5.

B. All 4 loops suctions are powered from 1 E 25KVA Inverters.

C. Two loop suctions on one train are powered from 1 E 480V MCCs Two loop suctions on one train are powered from 1 E 25KVA Inverters.

D One loop suction on each train is powered from 1 E 480V MCCs.

One loop suction on each train is powered from 1 E 25KVA lnverters.

14

HL-15R RO NRC Exam KIA 005 Residual Heat Removal System (RHRS)

K2.03 Knowledge of bus power supplies to the following:

RCS pressure boundary motor-operated valves.

KIA MATCH ANALYSIS The question presents a plausible scenario where the candidate must pick the correct power supplies to both trains RHR loop suction isolation valves.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Plausible to think all 4 valves are powered from 480V 1 E MCCs since most safety related valves are powered from 480V 1 E MCCs, however 2 of the loops suctions are powered from 480V 1 E Inverters CD1 15 and DD1 16.

B. Incorrect. Plausible to think all 4 valves are powered from 480V 1 E lnverters. There are 4 green annunciators associated with these valves (2 for starters and 2 inverters)

C. Incorrect. Plausible to think one train powered from inverters and the other powered from MCCs since there are 2 valves powered from each.

D. Correct.

REFERENCES V-LO-PP-12101, Residual Heat Removal System, slide #42.

13011-1, Residual Heat Removal, pages # 15 and # 22. Step s 4.1.4 and 4.2.4.

LOIT BANK QUESTION 005K2.03-001 VEGP learning objectives:

LO-PP-12101-12 Briefly describe the RHR system alignment during normal power operations and during RCS cooldown.

15

1. 005K2.03 001 Regarding the two following valves:

HV-8701B, RHR PMP-A UPSTREAM SUCTION FROM HOT LEG LOOP 1 HV-8702B, RHR PMP-B UPSTREAM SUCTION FROM HOT LEG LOOP 4 Which ONE of the following CORRECTLY describesthe power supplies to the RHR upstream loop suction isolation valves?

A 125V DC lnverterCDll5 and 480V 1E MCC BBE.

B. 480V 1 E SWGR ABO5 and 1 25V DC Inverter DD1 16.

C. 480V MCC NBE and 125V DC Inverter DD1I4.

D. 125V DC Inverter CD1I3 and 480V 1E SWGR BBO7.

V V Page: 1 1/5/2010

RHRS LOOP Suction Valve Hand switches (A Train) 1

Approved By Procedure Number Rev A.S. Parton Vogtle Electric Generating Plant 13011-1 67.1 Date Approved Page Number 4/24/09 RESIDUAL HEAT REMOVAL SYSTEM 15 of 115 INITIALS Critical

p. Open the RWST TO RHR PMP-A SUCTION 1-HV-8812A.

CV

q. Place RHR Pump A 1-HS-0620 in AUTO.

r. If required, open RHR PMP-A TO COLD LEG 1&2 ISO VLV 1-HV-8809A.

4.1.3.2 Restore power to RHR Train A to CCP Suction as follows:

a. Remove tags and close K2 link for breakers 1ABB-05.

b. Close 1ABB-05 to Valve 1-HV-8804A.

4.1.3.3 Align RHR TRN-A for standby per Checklist 3.

4.1.4 IF RHR is being placed in standby for MODE 3 entry, perform the following: (IV REQUIRED)

a. Shut down Inverter 1 CD1 15 per 13405-1, 1 25V DC 1 E Electrical Distribution System.

b. Open and lock the power supplies to the RHR Loop 1 Inlet (1) Turn off 1ABE-15 for 1-HV-8701A. )

(2) Open the K2 links for breaker 1ABE-15.

ALB34 E06 STARTER 1CD1I5N TROUBLE (3) At 1 CD1 15N open and lock the disconnect for 1-HV-8701 B.

(4) Remove and store the annunciator card associated with ALB34 E06 per 10018-C, Annunciator Control.

Printed October 7, 2009 at 16:05

Approved By Procedure Number Rev A. S. Parton Vogtle Electric Generating Plant 13011-1 67.1 Date Approved Page Number 4/24/09 RESIDUAL HEAT REMOVAL SYSTEM 22 of 115 INITIALS

q. Place RHR Pump B Handswitch 1-HS-0621 in AUTO.

r. If required, open RHR PMP-B TO COLD LEG 3&4 ISO VLV 1-HV-8809B.

4.2.3.2 Restore power to RHR Train B to SIP Suction as follows:

a. Remove tags and close K2 link for breakers 1 BBB-05.

b. Close 1BBB-05 to Valve 1-HV-8804B.

4.2.3.3 Align RHR TRN-B for standby per Checklist 4.

4.2.4 IF RHR is being placed in standby for MODE 3 entry, perform the following:

a. Shut down Inverter 1 DD1 16 per 13405-1, 1 25V DC 1 E Electrical Distribution System. (IV REQUIRED)

b. Open and lock the power supplies to the RHR Loop 4 Inlet Isolations:

(1) Open 1BBE-13 for 1-HV-8702B. (IV REQUIRED)

(2) Open K2 links for breaker 1 BBE-1 3.

(IV REQUIRED)

ALB34 E07 STARTER 1 DD1 l6N TROUBLE (3) At 1 DD1 I6N, open and lock the disconnect for 1-HV-8702A. (IV REQUIRED)

(4) Remove and store the annunciator card associated with ALB34 E07 per 10018-C, Annunciator Control.

Printed October 7, 2009 at 16:05

HL-15R RO NRC Exam

8. 005K6.03 OO1/2/IiRHR-HXIC/A 2.5 / 2.6/M LOll BANKIHL-15RNRC/RO/TNI/DS Given the following:

- The plant is in Mode 6.

- RHR Train B is in service.

- RHR Hx Bypass Valve FV-619 is in auto set to maintain minimum Tech Spec flow.

- RHR Hx Outlet Valve HC-607 demand is set at 30%.

- The instrument air supply to HC-607 severs and is completely detached.

- No other air operated valves are impacted by the failure.

Which ONE of the following describes the following system parameter changes from the initial steady state condition to the final steady state condition?

RCS Cooldown Rate RHR Hx Bypass flow A. Lower Lower B. Lower Higher C Higher Lower D. Higher Higher K/A 005 Residual Heat Removal System (RHRS)

K6.03 Knowledge of the effect of a loss or malfunction on the following will have on the RHRS:

RHR heat exchanger.

K/A MATCH ANALYSIS The question presents a plausible scenario where an air line breaks to the RHR Hx outlet valve (HV-607). This would result in maximum cooling water flow through the RHR Hx (higher RCS Cooldown rate) and lower RHR Hx Bypass flow.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. RCS cooldown rate will actually increase due to HV-607 failing open on loss of air. It will cause flow through the Hx to increase. HV-607 failing open will cause 16

HL-15R RO NRC Exam student determines that the RHR valves fail shut on loss of air.

B. Incorrect. RCS cooldown rate will actually increase due to HV-607 failing open on loss of air. It will cause flow through the Hx to increase. HV-607 failing open will cause FV-619 to throttle shut to maintain a constant total RHR flow. This choice is plausible if the student incorrectly determines that HV-607 fails shut on a loss of air.

C. Correct. RCS cooldown rate increases due to HV-607 failing open on loss of air. It will cause flow through the Hx to increase. HV-607 failing open will cause FV-61 9 to throttle shut to attempt maintain a constant total RHR flow.

D. Incorrect. RCS cooldown rate increases due to HV-607 failing open on loss of air. It will cause flow through the Hx to increase. HV-607 failing open will cause FV-61 9 to throttle shut to maintain a constant total RHR flow. This choice is plausible if the student determines that bypass flow goes up with HX flow due to loss of air.

REFERENCES LO-PP-12101-08-002 from LOIT Exam Bank.

1X4-DB-122, Residual Heat Removal System (excerpt included).

18028-C, Loss of Instrument Air, Attachment B, Loss of Instrument Air in Modes 4, 5, 6 VEGP learning objectives:

LO-LP-12102-03, Describe the RHR system response and the appropriate corrective actions for the following.

a. Loss of air or electrical power to FCV 606 or 618.

LO-LP-60321-02, State the fail position of the following valves on a loss ofinstrument air.

i. RHR heat exchanger outlet valve.

17

1. LO-PP-12101-08 002 Which ONE of the following is CORRECT concerning the effect of a loss of instrument air with RHR Train A in service while in Mode 5?

A. The RHR Heat Exchanger Outlet valve, HV-606, would fail SHUT, and the RHR Flow Control valve, FV-61 8, would fail SHUT.

B. The RHR Heat Exchanger Outlet valve, HV-606, would fail OPEN, and the RHR Flow Control valve, FV-618, would fail OPEN.

C. The RHR Heat Exchanger Outlet valve, HV-606, would fail SHUT, and the RHR Flow Control valve, FV-618, would fail OPEN.

D The RHR Heat Exchanger Outlet valve, HV-606, would fail OPEN, and the RHR Flow Control valve, FV-61 8, would fail SHUT.

i (C  ;

Page: 1 10/13/2009

Date: 10/13/2009 Time 120558 PM ti, i

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FORMATION SEE DWGS 1X4DB143, 1441 DB183. 10. PROVIDE 0.375 ID FLOW RESTRICTiON AS SHOWN ON BPC 21. AiR ISOLATED DURING POWER OPERATION TO 29 A SECTTOI DWG CXDGOO1. ENSURE VALVE FAILS TO IFS SAFETY POSITION REMOVED ALVES FURNISHED BY WESTINGHOUSE EXCEPT: (FULL OPEN OR FULL CLOSED).

LL MANUAL VALVES 2 AND SMALLER AND 11. THE ORIFICE SHOULD BE LOCATED ON A HORIZONTAL STRETCH OF PIPE. 22. ABANDON INPLACE INSTRUMENTS XISO14, 30. THESE AR D ALLOW RE

12. REFER TO ELEMENTARY DIAGRAM FOR THE DETAILED XY5O14A,XY5O 1 4B, X150 15, XY5O 1 5A, COOLING WATER CONNECTIONS. SEE DWGS

23. 4 VIEW PORT TO 31. PN SHELL SIDE CONNECTIONS SEE DWG 1X4DB1 37.

LE SPOOL.

13. FOR TYPICAL DETAIL SEE STD. DWG AX4DD000.

14. DELETED.

VALVE STROKE TEST WITHOUT REMOVING THE VESSEL HEAD.

ç A CLASS CHANGE AT FIRST WELD OF 15. HIGH POINT VENT VALVE (BECHTELFURNISHED) FOR 24. VENT HOLE IS DRILLED ON CONTAINMENT N I HE CONTAINMENT. SEAL PIPING SEE 1X6AFO225. SIDE OF DISC.

NGE ON INLET OF EACH 14ø PIPE FOR 16. TE IS A SURFACE TYPE CLAMP ON RTD rW IS 25. FLOW RESTRICTOR PROVIDED AS SHOWN ON

)VIDE ONE MATING BLIND FLANGE TO BE DRILLED INSTALLED BUT NOT USED. 1 J4 1201 24901.

CONNECTION. MATING FLANGE MUST BE REMOVED 26. FLOW RESTRICTOR PROVIDED AS SHOWN ON 17 VENT HO I N SPECTIVE RHR 1J412O1251O1, IN. PUMP SIDE OF DISC 27. 0.25 iNCH FLOW RESTRICTOR AS SHOWN ON ED TO SEISMIC CATEGORY I AND ASME III CLASS MC. 18. FOR ISI TESTING USE ONLY. INSTRUMENT ISOLATION 6 STEM ARE SUSPECTED TO CAUSE STRESS VALVE TO BE NORMALLY CLOSED.

THE APPUCABLE DESIGN DRAWING AND 19. ALARM ON OPEN HOT LEG VALVE AND HIGH RCS PRESSURE. 28. 0.25 INCH FLOW RESTRICTOR AS SHOWN ON OIflCAT1OND NTS. 20. DELE1ED.

THIS DOCUMENT CONTJNS PROPRIETAR SOUTHERN COMPAW OR OF THIRD PAl OF, THE SIJBSIDLARIES OF THE SOUTI-IE NATION, OR DISCLOSURE OF ANY PORI 7 6 5 4

Title:

C:\DATA\HL-15 Recovery References\P&IDS - Unit 1\1X4DB122.cal

Approved By Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant 18028-C 25 DateApproved Page Number LOSS OF INSTRUMENT AIR ATTACHMENT B Sheet 1 of 7 LOSS OF INSTRUMENT AIR IN MODES 4, 5, OR 6 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

_B1. Check Instrument Air supply header Bi. Go to Step BI 1.

pressure on P1-9361 LESS THAN 100 PSIG.

CAUTION Loss of instrument air will cause CHARGING LINE CONTROL FV-0121 and SEAL FLOW CONTROL HV-0182 to fail open.

_B2. Check RCS inventory SOLID. B2. Perform the following:

_a. IF needed to maintain RCS level, THEN establish safety grade charging by initiating 13006, CHEMICAL AND VOLUME CONTROL SYSTEM.

b. GotoStepB5.

_B3. Trip all charging pumps.

• B4 Monitor No. I seal leakoff temperature and flow until charging pump is restarted.

CAUTION Loss of instrument air pressure will cause the RHR HX outlet valves to fail full open and the HX bypass valves to fail fully closed.

_B5. Check plant Mode MODE 4 OR

- B5. Suspend all fuel movement.

MODE 5.

Printed October 13, 2009 at 13:06

HL-15R RO NRC Exam

9. 006K6.02 OO1/2/1/ECCS-ACCUM/C/A 3.4 /3.9/M WATTS BAR 2008/HL-15R NRC/RO/TNT/DS Given the following:

- The ECCS Accumulator isolation valves are CLOSED and in AUTO with power on the valves.

- PRZR pressure has been raised from 930 psig to 2020 psig.

- Safety Injection (SI) is actuated.

Which ONE of the following identifies the effect on the ECCS Accumulator isolation valves:

1. during PRZR pressure increase to 2020 psig, and

2. the SI signal PRZR Pressure Increase SI signal A. valves remain closed. Open signal is generated to the valves.

B. valves remain closed. Open signal is NOT generated to the valves.

C valves automatically open. Open signal is generated to the valves.

D. valves automatically open. Open signal is NOT generated to the valves.

KIA 006 Emergency Core Cooling System (ECCS)

K6.02 Knowledge of the effect of a loss or malfunction on the following will have on the ECCS:

Core flood tanks (accumulators).

K/A MATCH ANALYSIS The question presents a plausible scenario where ECCS Accumulator isolation valves are inadvertently left closed as RCS I PRZR pressure is raised to 2020 psig. Then, a manual SI signal is generated. The student must determine that the isolation valves would have opened on P-li (2000 psig) prior to receipt of the SI signal. Also, the student must determine a manual SI signal would have generated an open signal to the valves.

ANSWER I DISTRACTOR ANALYSIS 18

HL-15R RO NRC Exam choice is incorrect. The second part of the choice is correct as an open signal is generated on receipt of SI signal. Plausible the students may not realize P-il sends an open signal and know that an SI signal does.

B. Incorrect. Valves will automatically open on P-li (2000 psig) so this part of the choice is incorrect. The second part of the choice is incorrect as an open signal is generated on receipt of SI signal. Plausible the students may not realize P-i i or SI sends an open signal. These valves are usually de-energized and open prior to 1000 psig in the RCS.

C. Correct. Pressure> P-i i (2000 psig) and an SI signal will cause the valves to open by system design.

D. Incorrect. Valves will automatically open on P-il (2000 psig) so this part of the choice is correct. The second part of the choice is incorrect as an open signal is generated on receipt of SI signal. Plausible the students may not realize P-il or SI sends an open signal.These valves are usually de-energized and open prior to 1000 psig in the RCS.

REFERENCES Watts Bar 2008 NRC RO Exam question # 31 used as base for modification (included).

V-LO-PP-l 3101, Emergency Core Cooling System (ECCS) slides # ii 2 and # ii 3.

(included).

V-LO-PP-281 03, Reactor Trip and ESFAS Signals, slides # 47 and # 5i (included).

VEGP learning objectives:

LO-PP-i3lOl-i2, Describe the control logic for the accumulator isolation valves in response to:

a. SI signal

b. Permissive P-li.

19

RO Walls Bar 2008 NRC Initial License Exam WRITTEN QUESTION DATA SHEET Question Number: 31 KIA: 006 K6.02 Emergency Core Cooling Knowledge of the effect of a loss or malfunction on the following will have on the EGOS: Core flood tanks (accumulators).

Tier: 2 RO Imp: 3,4 RD Exam: Yes Cognitive Level: HIGH Group: 1 SRO Imp: 3.9 SRO Exam: Yes Source: NEW Applicable 10CFR55 Section: 41.7145.7 Learning Objective: 3-OT-SYSO63A Objective 24: Given a set of plant conditions, determine the correct response of the Emergency Core Cooling System.

References:

l-47W61 1-63-7, Rev 2.

Question:

Given the following plant conditions:

• Plant startup is in progress.

• During performance of GO-i, Unit Startup from Cold Shutdown to Hot Standby, the CLA isolation valves were left CLOSED with povier on the valves as pressurizer pressure was raised from 900 psig to 1900 psi 9

• A manual safety injection (SI) is initiated.

Which ONE of the following identifies the position of the CLA isolation valves before the Slis initiated and how the MANUAL Safety injection will affect the valves?

Before SI Effect of the SI signal A. Valves will have An open signal will be generated to the valves.

automatically opened.

B. Valves will have An open signal will NOT be generated to the valves.

automatically opened.

C. Valves will have An open signal will be generated to the valves, remained closed.

D. Valves will have An open signal will NOT be generated to the valves, remained closed.

DISTRACTOR ANALYSIS

a. Incorrect. The valves would not have automatically opened prior to the St because the P-Il permissive has not been made. However, an open signal would be generated by the SI. Plausible because the valves do automatically open if pressure is greater than P-il, and an SI would generate an open signal.

b. Incorrect. The valves would not have automatically opened prior to the SI because the P-i 1 permissive has not been made and an open signal would be generated by the SI. Plausible because the valves do automatically open if pressure is greater than P-li and the candidate could conclude that since the valves are normally opened manually and power removed that the St does not generate an open signal due to the CLAs being a passive sub-system in the EGOS.

c. CORRECT. The valves would be closed until the pressure rose above P-i 1(1970 psig). When P-il permissive was met the valves would then automatically open. With pressure at 1900 psi 9 the valves would stilt be closed, but would open when the SI was actuated.

d. Incorrect, The valves would have remained closed because the P-Il permissive has not been made and an open signal would be generated by the SI. Plausible because the valves would remain closed with the pressure less than P-il and the candidate could conclude that since the valves are normally opened manually and power removed that the SI does not generate an open signal due to the CLAs being a passive sub-system in the ECCS.

n I 33oft

ACCUMULATOR OUTLET MOV INTERLOCKS If valves have been powered up and are SHUT, they will automatically OPEN when:

P-Il, or

>SI However, they are normally de-energized open.

V-LO-PP-1 3101 -033 112

V-LO-PP-1 3101-03.3 Reactor Trip & ESFAS Signals What will produce a P-il signal?

2/3 pressurizer pressure inst. <2000 psig.

What is the function of P-li?

-Allows manual block of pressurizer low pressure SI.

-Allows manual block of low steam line pressure SI and SLI.

V-LO-PP-281 03-6.2 47 V-LO-PP-281 03-06.1 47

Reactor Trip & ESFAS Signals What automatically occurs when Pressurizer pressure increases above 2000 psig?

• SI Accumulator Outlet Valves receive an Open signal.

• Enables alarms ACCUM TANK 1-4 ISO VLV 8808A-D NOT FULLY OPEN

• Enables RWST TO SI PUMP ISOLATION VALVE 8806 NOT FULL OPEN alarm V-LO-PP-28103-6.2 51 Point out to the students that the Accumulator outlet valves are normally de energized in the open position when at power.

Caution:

When pressurizer pressure goes above the P-Il set point the SI signal from low pressurizer pressure and the SI and SLI from low steam line pressure automatically unblock.

V-L0-PP-281 03-06.1 51

HL-15R RO NRC Exam

10. 007EG2.4.3 1 002/1/1/RX TRIP-ALARMS/ARPS/C/A 4.2 / 4. 1/NEW/IlL-i 5R NRC/RO/TNT/ DS Given the following plant conditions:

- The plant tripped from 20% power during a loss of grid event.

- All RCS FLOW TRIP 90% bistable lights are LIT.

- The following First Out annunciator is 11 gallop flashing.

LOW FLOW I RCP I P7 PERMISSIVE REACTOR TRIP

- All RCP 1 E and non-i E handswitch red lights are LIT.

- 19001-C, ES-O.l Reactor Trip Response is in effect.

Which ONE of the following is CORRECT regarding the:

1) event which caused the reactor trip and

2) indications the operators will use to monitor RCS temperature?

Cause of Reactor Trip Temperature monitoring A. RCP underirequency RCS AVERAGE TEMPERATURE B. RCP underfrequency RCS WR COLD LEG TEMPERATURES C. RCS two loop low flow RCS AVERAGE TEMPERATURE D RCS two loop low flow RCS WR COLD LEG TEMPERATURES KIA 007 Reactor Trip Stabilization

- - Recovery EG2.4.31 Knowledge of annunciator alarms, indications, or response procedures K/A MATCH ANALYSIS The question presents a plausible scenario where a grid disturbance has resulted in a reactor trip from power. The student must determine from given indications the cause of the reactor trip (2 loop low flow) and the proper method to monitor RCS temperature per 19001-C (RCS WR Tcs). Annunciator alarms, control board indications, and Rx.

Trip Response procedure 19001-C are used in the question stem.

At 20% nowr s nivn in th stcm th 138 kV lr.trir.I hiiss nowrd from th 20

HL-15R RO NRC Exam RATs (offsite power) so a loss of offsite power would lead to an RCS Lo Flow Trip when > P-7. At 20% power, the reactor is> P-7 (10% set point).

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. RCP underfrequency would cause the low flow bistables to be LIT.

However, RCP underfrequency would cause all the RCP breakers to trip open so RCP UF is incorrect. RCS Tave is plausible as it is the normal parameter used to monitor RCS temperature, however, with no RCP5, 19001-C specifies RCS WR Tcs.

B. Incorrect. RCP underfrequency would cause the low flow bistables to be LIT.

However, RCP underfrequency would cause all the RCP breakers to trip open so RCP UF is incorrect. RCS WR Tcs is the correct parameter to use per 19001-C.

C. Incorrect. Two loop low flow is correct. RCS Tave is plausible as it is the normal parameter used to monitor RCS temperature, however, with no RCPs, 19001-C specifies RCS WR Tcs.

D. Correct. Two loop low flow would cause the indications and alarms. RCS WR Tcs are the parameter to monitor per 19001-C.

REFERENCES V-LO-PP-281 03, Reactor Trip and ESFAS signals slides # 116, 117, and 120 (included).

V-LO-TX-28101, Reactor Protection System section B for Reactor Trip and ESFAS signals, page # 27 (included).

V-LO-PP-1 6101, RCS Temperature Instrumentation slide # 28 (included).

V-LO-PP-1 6401, Reactor Coolant Pumps slide # 32 and blowup of slide # 32 (included).

V-LO-TX-1 6001, Primary Systems page # 33 for RCP UV included.

19001-C, ES-0.1 Reactor Trip Response step #4 (included).

17009-1, Annunciator Response Procedure for ALB 09 on Panel 1 Cl on MCB, window E03 (included, note this is the first out panel) 12004-C, Power Operation (Mode 1) steps 4.1.39 from power ascent section (included) and 4.2.10 from power descent section (included).

VEGP learning objectives:

LO-PP-28103-03, List all reactor trip signals, set points, coincidences, permissives, and blocks.

21

HL-15R RO NRC Exam LO-PP-16101-03, State the conditions when wide range temperature indications must be used instead of narrow range instrumentation.

LO-PP-16401-09, Describe the following for the RCP supply breakers.

c. Protection features.

LO-PP-61203-O1, Describe the basic steps involved with transfer of the 13.8 kV and 4160 kV buses from the Reserve Auxiliary Transformers (RATs) to the Unit Auxiliary Transformers (UATs) 22

e p P RcLpj euit CHI ) LO FLOVv i4FA JNDIHW LI UNDFHIRFO ri

! c 1 RCLP?

RCP 1W61 LHS L ft ?

JNDF RVOL F UNOLRF REQ i ;

fl9_ _ __

RCP L _P L LP3 BUH LH BUS2 CR1 L H W i.) Ft OW O FLOW JROftVOL1 UNOLRI REQ 3A ft

[F 4 I 4 LP 4 RCP I3U tCH$ HiS? U44 )

UNEtERVOL I t RFRLQ FLL4 SJ-eiw +oN5 b- 1 I r-r L - t.. i -_*. ,, .,

pp- 2 io 51 II n

TWO LOOP LOW FLOW TRIP 2/3 CHANNELS <90% ON 2/4 LOOPS.

AUTO BLOCKED BELOW P-7.

BASES: DNB PROTECTION.

V-LO-PP-281 O3-62 117

Reactor Trip & ESFAS Signals RCP UF TRIP 1/2 13.8 BUSSES <57.3 HZ AUTO BLOCKED BELOW P-7 WILL ALSO TRIP ALL RCPs BASES: DNB PROTECTION.

V-LO-PP-281 03-6.2 120 The bases behind tripping the RCPs during an under frequency condition is that the inertia of the flywheel on each RCP provides enough forced flow for decay heat removal. This holds true if the RCP5 are at their normal speed before the trip occurs. The assumptions are that if the RCP frequency is lowering, then eventually the RCPs are going to trip anyway. So the reactor and RCPs are tripped early to ensure proper decay heat removal is provided.

0 p

V-LO-PP-28103-06.1 120

Bases: Prevents DNB conditions resulting from loss of one or more reactor coolant pumps.

13) Two loop loss of flow 2 out of 3 channels 2 out of4 loops

= 90% flow t2O enabled above P-7; > 10% reactor power Bases: Prevents DNB conditions resulting from loss of two or more reactor coolant pumps.

14) RCP Under Voltage Reactor Trip I out of 2 buses = 9660 V enabled above P-7; > 10% reactor power Bases: Provides protection against DNB as a result of loss of forced coolant flow. Ensures a reactor trip signal occurs before the low flow trip set point is reached.

15) RCP Under Freg. Reactor Trip I out of 2 buses = 57.3 Hz enabled above P-7; > 10% reactor power trips all RCPs feeder breakers. The bases behind tripping the RCPs during an under fequency condition is: The inertia of the flywheel on each RCP provides enough for force flow

. f r decay heat removal. This hold true if the RCPs are at their normal speed before the trip The assumptions are that if the RCP frequency is lowering that eventually the RCPs are (ufr So the reactor and RCPs are tripped early to ensure proper decay heat as jrv Bases: Provides protection against DNB as a result of loss of forced coolant flow. Ensures a reactor trip signal occurs before the low flow trip set point is reached.

16) Safety Injection Reactor Trip Any Safety Injection Signal Bases: Ensures subcriticality during accident conditions. The ECCS only rated for decay heat removal only.

17) Turbine TriplReactor Trip 4 out of 4 Main 96.7 % open Turbine Stop Valves enabled above P-9; > 40% reactor power

-or-2 out of 3 ETS header = 580 psig pressure channels low enabled above P-9; > 40% reactor power Bases: Assures reactor trip upon reduction of turbine power in excess of that which can be handled by steam dumps and rod control system, including single failures of steam dump or pressurizer spray valve controls.

18) SIG Low-Low level reactor trip 2 out of 4 channels =38%

on I out of 4 S/Gs Bases: Protects the reactor from loss of heat sink by tripping with sufficient water level to allow for starting delays of Auxiliary Feed Water System.

V-LO-TX-281 01-08.1 27

RCS Wide Range Temperature Includes hot leg and cold leg temperature indications I each per loop, 2 total per loop (no spare RTD) 0-700°F Range Measured with RTDs located in dry thermowells Provide protection and minimizes system leakage Response fairly slow to change in RCS temperature under natural circulation conditions Used when off scale on narrow range instruments (e.g., plant cool down) or during loss of forced RCS -

loop flow, or where NR inst. otherwise unavailable kf,v f--c& f(c,,

W.R. Temperature indication will continueto be representative of RCS temperature (partial) in the loop at the RTD thermowells LO-PP-1 6101 R-04 28

This drawing shows the frequency reading coming off the 13.8kV Bus for simplicity. In reality, the taps are similar to the UV taps in the previous slide. Where a combination of UF on [(1 OR 2) AND (3 OR 4)] >P7 will give the RX trip and the opening of the 1 E breakers. The drop in frequency would decrease RCP rotor power output and the developed torque necessary to supply rated flow.

A very important note to the students: Watch how only the 1 -E breakers open after the Under frequency trip. You could also talk about the flywheel on the RCPs to keep them pumping momentarily. In the drawing I left then running a little while after the breakers open.

IP eott(e V-LO-PP-1 6401 Rev-03 32

IJi PHBFrBC,D POTENTIAL oo SWGR AD c

,r.

I I

13.8 Bus Transfer LOOP 3 After Generator Output Bkr Trips r

II I, L V-LO-PP-1 6401

or phase differential current. The reactor coolant pump class 1 E motor breakers (HS-0495A, 0496A, 0497A, 0498A) receive their control power from 125 VDC ESF busses train A, B, C, and D, respectively.

Their breakers will automatically trip on under frequency or instantaneous or time delay over current.

16.23 VIBRATION MONITORING Each RCP is equipped with both Frame and Shaft vibration monitoring. The RCP frame vibration monitors consist of two probes that are mounted 908 apart on the top of each RCP motor frame.

The RCP shaft vibration monitors are measured by a vertical and horizontal proximity probe mounted parallel and perpendicular respectively to the pump discharge at a location near the pump coupling. Both frame and shaft vibrations are continuously monitored. Alarms are generated in the control room if the frame vibration exceeds 3 mils and 5 mils. At 5 mils the operators are required to trip the associated RCP. High frame vibrations are indicative of a misalignment or an out of balance condition. Shaft vibration alarms are also generated if they exceed 15 mils and 20 mils.

At 20 mils the operators are required to trip the associated RCP. High shaft vibrations are indicative of a possible bearing failure. Local monitoring at the RCP vibration panel is required to determine which RCP has a vibration problem. The Control room is only provided with common alarms. During normal plant 33 Revision 7.2

Approved By I Procedure Number Rev J.B. Stanley Vogtle Electric Generating Plant 19001-C 31 DateApproved Page Number ES -0.1 REACTOR TRIP RESPONSE 7/22/2008 4 of 25 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

• 4 Check RCS temperature stable at *4 IF temperature is less than 557°F or trending to 557°F. and lowering, THEN perform the following as

_With RCP(s) running RCS - necessary:

AVERAGE TEMPERATURE.

a. Stop dumping steam.

-OR-Without RCP(s) running RCS WR

- b. Perform the following as COLD LEG TEMPERATURES. appropriate:

IF at least one SG NR level greater than 10%,

THEN lower total feed flow.

-OR-IF all SG NR levels less than 10%,

5c e THEN lower total feed flow to NOT less than 570 gpm.

3 IF cooldown continues, THEN close MSIVs and BSIVs.

_d. IF temperature less than 557°F and NOT trending to 557°F, THEN borate as necessary to maintain shutdown margin by initiating 13009, CVCS REACTOR MAKUP CONTROL SYSTEM.

e. IF temperature greater than 557° F and rising, THEN dump steam.

Printed October 2, 2009 at 12:53

Approved By I Procedure Number Rev S. A. Philips I Vogtle Electric Generating Plant 17009-1 11 Date Approved ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 09 ON PANEL 1 Cl Page Number 3/27/08 ON MCB 32 of 38 WINDOW E03 ORIGIN SETPOINT LOW FLOW/RCP/P7 2 of 4 RCP Loops 90% of normal flow PERMISSIVE Low Flow when REACTOR TRIP above P7 1.0 PROBABLECAUSE

(( Ofe 1 W e w L5

1. Lossof13.8kVbus1NAAor1NAB

2. Two or more Reactor Coolant Pump Breakers tripped.

2.0 AUTOMATIC ACTIONS NOTE This trip function is blocked below the P-7 permissive.

1. Reactor Trip.

2. Turbine Trip.

3. Feedwater Isolation if Tavg is less than 564°F.

4. Steam Dump Armed.

3.0 INITIAL OPERATOR ACTIONS Go to 19000-C, E-O Reactor Trip or Safety Injection.

4.0 SUBSEQUENT OPERATOR ACTIONS NONE 5.0 COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE

REFERENCES:

FSAR Section 7.2, 1X6AAO2-229, PLS Printed October 2, 2009 at 12:56

Approved By 4 Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant , 12004-C 83 Date Approved Page Number 6/30/09 POWER OPERATION (Mode 1) 29 of 89 INITIALS I. IF HDT High Level Dump Valves are in MANUAL, restore HDT Level Control to NORMAL.

m. If necessary, adjust SGBD Condensate Return Temperatures for the present power level per Step 4.1.lOb.

n. IF in service, shut down Feed Water preheating using the 5th Stage Feed Water heaters per 13615, Condensate and Feed Water System.

o. Start up and test the second Main Feed Pump per 13615, Condensate And Feed Water System up through and including performance of 14993, Steam Generator Feed Pump Turbine Lube Oil System Test. and 14992, MFPT Trip Mechanism Test.

I ifr 20we4-1 - hi 4.1.39 BETWEEN 30 and 50% Reactor Power perform the following: I WkJ 5( lu

a. At SS direction, transfer the 13.8kV busses per 13420,1 307 J7 LA /ç 13,8kV AC Electrical Distribution System. / / -. j 0 GTe 1

b. At SS direction, transfer the 41 60V AC busses per /13 t<J k2 S Cct 41 60V AC Non 1 E Electrical Distributiori )o jC.-/et._ -i-i;)

RcL ..

f/\1)

c. Verify l&C has reset the PR high level trip bistables ---_

(NC306) for at least 3 of the 4 power range channels, per Step 4.1 .38.i, prior to exceeding 42% reactor power.

d. Check PREFERRED LINE lamp is lit, on the EHC STATIC TRANSFER SWITCH 161 5-D3-001 in CB Rooms A-78(U1) and A-80 (U2). H PREFERRED LINE lamp is NOT lit, notify System Engineer.

4.1.40 IF NOT required, isolate the Auxiliary Steam header per 13761-C, Auxiliary Steam System.

Printed October 2, 2009 at 13:32

Approved By Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant ,. 12004-C 83 Date Approved Page Number 6/30/09 POWER OPERATION (Mode 1) 38 of 89 INITIALS 4.2.10 BETWEEN 50% and 30% Turbine Power, perform the following:

a. Remove one Main Feed Pump from service per 13615, Condensate And Feed Water System.

CufW

b. Transfer the 1 3.8kV busses from the Unit Auxihary Transformers to the Reserve Auxiliary Transformers per ¶ -y2 54 13420, 13.8kV AC Electrical Distribution System.

c. Transfer the 41 60V AC Non 1 E busses to the Reserve Auxiliary Transformers per 13425 A/B/C, 4160V AC Non 1 E Electrical Distribution System.

d. Monitor Circulating Water Tower Basin level during the power decrease.

e. Adjust Tower blowdown and/or Tower makeup as required to control basin level per 13724, Circulating Water System.

4.2.11 WHEN between 50% power and 20% power AND WHEN a calorimetric is performed per 14030, Nuclear Instrument Calorimetric Calibration, IF any PR NIs are adjusted in the downward direction, notify l&C to adjust the PR high level trip bistables (NC306) for channels N41, N42, N43 and N44 to =90%.

/ /___

Person Contacted Date Time 4.2.12 Maintain operation of the Condensate Demineralizer System per 13616, Condensate Filter Demineralizer System.

4.2.13 At approximately 46% Reactor Power, verify 1 LP LO FL TRIP BLKD P-8 illuminates.

4.2.14 At approximately 38% Reactor Power, verify the TURB TRIP/RX-TRIP BLOCKED P-9 illuminates.

4.2.15 At approximately 37% Turbine power, verify C20 clears (automatically blocks AMSAC ATWAS Mitigation System Actuation Circuitry) by observing permissive light AMSAC BYPASSED LO TURBINE LOAD C20 illuminates.

Printed October 2, 2009 at 13:32

HL-15R RO NRC Exam

11. 007K1.03 O01/2/1/PRT-RCS/C/A- 3.0 / 3.2/B-SEQUOYAH 2007/HL-15R NRC/RO/TNT/DS Given the following plant conditions:

- A reactor trip has occurred.

- RCS pressure is 1830 psig and lowering.

- Containment pressure is 2.3 psig and rising.

Which ONE of the following describes the flow path of the RCPs # 1 seal leakoffs?

Seal leakoff flows are currently directed to the...

A. VCT B PRT C. RCDT D. CTMTsump 23

HL-15R RO NRC Exam K/A 007 Pressurizer Relief Tank/Quench Tank System (PRTS)

KI .03 Knowledge of the physical connections andlor cause effect relationships between the PRTS and the following systems:

RCS.

KIA MATCH ANALYSIS The question presents a plausible scenario where a reactor trip and SI have occurred on low PRZR pressure. The candidate must choose the correct RCP # 1 seal leakoff flow path.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. RCP # 1 seal leakoff normally flows to the VCT but on SI due to low PRZR pressure CIA will isolate this flow path and the leakoff will flow to the PRT via a relief valve.

B. Correct. RCP # 1 seal leakoff normally flows to the VCT but on SI due to low PRZR pressure CIA will isolate this flow path and the leakoff will flow to the PRT via a relief valve.

C. Incorrect. RCP # 1 seal leakoff is plausible to align to the RCDT but this is a manual operation on the QMCB. No indication a manual alignment has been performed is provided in the stem.

D. Incorrect. RCP # 3 seal leakoff normally flows to the Containment sumps and sometimes on seal failures # 2 and # 1 seals could flow to the sumps so this path is plausible. However, no indications of seal failure are provided in the stem.

REFERENCES Sequoyah 2007 NRC Exam question # 11 V-LO-PP-1 6401, Reactor Coolant Pumps V-LO-PP-09200, CVCS Charging System VEGP learning obiectives:

V-LO-PP-1 6401-04, State the effects of closing the # 1 seal leakoff valve.

24

Sequoyah Nuclear Plant SRO NRC Examination 05/09/2007

11. Given the following plant conditions:

- A reactor trip has occurred.

- RCS pressure is 1810 psig and lowering.

- Containment Pressure is 1 .5 psig and rising.

Which ONE (1) of the following describes the status of RCP #1 seal leakoff?

Directed to...

A. VCT B. PRT C. RCDT D. Reactor Building Floor and Equipment Sump oc7 id Oi ((

11

CVCS Charging Note: Valves close automatically on train-related Containment Isolation Signal (CIA).

A CIA signal is generated every time by a Safety Injection Signal. Therefore, a SI signal will cause a CIA signal, which will result in RCP #1 seal leakoff flow path being isolated. Overpressurization is prevented by the relief valve PSV-8 121.

eH V-LO-PP-09200-02. 1 119

CVCS Charging PSV-8121 located inside containment prevents over-pressurizing the seal return line if CIA valves located on the line isolate and seal injection flow remains in service. Setpoint is 150 psig. Relieves to PRT V-LO-PP-09200-02. 1 115

HL-15R RO NRC Exam

12. 007K4.O1 OO1/2/1/PRT-COOLING/C/A 2.6/2.9/NEW/HL.-15R NRC/RO/DS / TNT Given the following:

PRT temperature high due to leakage from a PRZR PORV The unit is at 100% power All systems are in their normal alignment The PRT (1) directly cooled by (2) .

A. (1) is automatically; (2) ACCW via the RCDT heat exchanger B. (1) is automatically; (2) NSCW via the RCDT heat exchanger C (1) must be manually aligned to be; (2) ACCW with the RCDT heat exchanger D. (1) must be manually aligned to be; (2) NSCW with the RCDT heat exchanger 25

HL-15R RO NRC Exam KIA 007 Pressurizer Relief Tank/Quench Tank System (PRTS)

K4.01 Knowledge of the PRTS design feature(s) and/or interlock(s) which provide for the following:

Quench tank cooling.

KIA MATCH ANALYSIS The question requires the student to identify if PRT cooling is automatically or manually aligned and what the cooling meduim is, which meets the K/A topic.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Normal at power PRT alignment is with the recirculation and fill valves shut. With these two valves closed cooling to the PRT must be manually aligned.

ACCW is used to cool the PRT contents via the RCDT heat exchanger.

B. Incorrect. Normal at power PRT alignment is with the recirculation and fill valves shut. With these two valves closed cooling to the PRT must be manually aligned.

ACCW is used to cool the PRT contents via the RCDT heat exchanger. ACCW rejects it heat load to the NSCW system making this cooling choice plausible.

C. Correct. Normal at power PRT alignment is with the recirculation and fill valves shut.

With these two valves closed cooling to the PRT must be manually aligned. ACCW is used to cool the PRT contents via the RCDT heat exchanger.

D. Incorrect. Normal at power PRT alignment is with the recirculation and fill valves shut. With these two valves closed cooling to the PRT must be manually aligned.

ACCW is used to cool the PRT contents via the RCDT heat exchanger. ACCW rejects it heat load to the NSCW system making this cooling choice plausible.

REFERENCES 13004-1, Pressurizer Relief Tank Operation pages 3, 27, and 28.

P&ID drawing 1X4DB-112 for the PRT showing the normal at power valve alignments to the PRT.

VEGP learning objectives:

LO-PP-1 6301-09:

Describe the methods for cooling the PRT.

26

Approved By Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant 13004-1 18 Date Approved Page Number 3/22/09 PRESSURIZER RELIEF TANK OPERATION 3 of 40 3.0 PREREQUISITES AND INITIAL CONDITIONS 3.1 The Gaseous Waste Processing System is available to provide processing of gases from the PRT.

3.2 The Auxiliary Gas System-Nitrogen or nitrogen from the Waste Gas Decay Shutdown Tank is available to provide a nitrogen blanket for the PRT.

3.3 Reactor Make-Up Water is available to provide a cooling spray for the PRT.

3.4 ACCW is available if cooling the PRT with the RCDT Heat Exchanger.

Printed October 28, 2009 at 16:07

Approved By Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant a 13004-1 18 Date Approved Page Number 3/22/09 PRESSURIZER RELIEF TANK OPERATION 27 of 40 INITIALS 4.4.3 PRT Cooldown Using Spray And Drain (One Hour Cooldown)

NOTE Two methods for cooling the PRT exist. Cooling the PRT by spray and drain is designed to cool the PRT in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. This method uses makeup water and drains to the Waste Processing System. Cooling the PRT by recirculation through the RCDT Hx is designed to cool the PRT in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. This method minimizes makeup water use and waste processing of liquid. The time required to cool the PRT and water usage should be considered before deciding which method to use.

4.4.3.1 Establish communications between the Liquid Waste Processing System Panel (WPSL) and the Control Room.

4.4.3.2 Verify the PRT pressure less than or equal to 50 psig as indicated by PRESSURIZER RELIEF TANK 1-Pl-0469 to prevent RCDT System over pressurization.

4.4.3.3 Verify open WPSL RCDT PUMPS DISCH TO RECYC EVAP 1-1901 -U6-327.

4.4.3.4 Realign RCDT Pump Suction to the PRT and initiate spray as follows:

a. Stop the running REACTOR COOLANT DRAIN TANK PUMP

1. 1 1HS-1003A (WPSL)

1. 2 1HS-1003B (WPSL)

CAUTION The RCDT level should be monitored to prevent tank flooding.

b. Place REACTOR COOLANT DRAIN TANK LEVEL 1-LC-1 003 in MANUAL and open the valve (WPSL).

c. Close REACTOR COOLANT DRAIN TANK RECIRCULATION VALVE 1-HV-7144 (WPSL).

Printed October28, 2009 at 16:12

Approved By Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant . 13004-1 18 Date Approved Page Number 3/22/09 PRESSURIZER RELIEF TANK OPERATION 28 of 40 INITIALS

d. Close REACTOR COOLANT DRAIN TANK PUMP SUCTION VALVE 1-HV-7127 (WPSL).

e. Open PRT FILL ISO VLV 1-HV-8030.

f. Open PRT RECIRC ISO VLV 1 -HV-8031.

4.4.3.5 Initiate PRT drain and maintain level and pressure as follows:

a. Verify ACCW is available to the RCDT Heat Exchanger.

b. Start REACTOR COOLANT DRAIN TANK PUMP

1. 1 1HS-1003A (WPSL)

1. 2 1HS-1003B (WPSL)

c. During cooldown, maintain PRT level greater than or equal to 58% as indicated by PRESSURIZER RELIEF TANK 1 -LI-0470.

CAUTION During the PRT cooldown, cooling of the liquid will cause a corresponding pressure decrease.

d. During cooldown, maintain PRT N2 pressure at 3 to 5 psig as indicated by PRESSURIZER RELIEF TANK 1-Pl-0469.

4.4.3.6 At a PRT temperature of 110°F as indicated by PRESSURIZER RELIEF TANK 1-TI-0468, secure spray and realign RCDT System to normal as follows:

a. Close the PRT FILL ISO VLV 1-HV-8030.

b. Stop the running REACTOR COOLANT DRAIN TANK PUMP

1. 1 1HS-1003A (WPSL).

1. 2 1HS-1003B (WPSL).

c. Close PRT RECIRC ISO VLV 1-HV-8031 Printed October 28, 2009 at 16:17

Date: 10/28/2009 Time :04:15:18 PM I IB /_,__\

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2. FOR TYPICAL MAKEUP PIPING ARRANGEMENT TO RCP STAND 10. DELETED. ç2 112O1T-PIPE SEE DWG 1X4DB1 14 (D6). ç3 112O1T.

11. PROVIDE 12 REMOVAL SPOOL. (4) 11201T-QMCB TI 3. ALL EQUIPMENT AND VALVES TO BE FURNISHED BY WESTING HOUSE EXCEPT THE FOLLOWING: 12. FOR PTPICAL DETAL SEE STD. DWG. AX4DD000.

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HL-15R RO NRC Exam

13. OO8AA1.04 OO1/1/1/PRZR VAPOR-FW PUMPS/C/A -2.8 / 2.5/NEW/HL-15RNRC/RO/TNT/DS Given the following sequence of events:

- The reactor is tripped due to a PRZR Safety valve failing open.

- Several minutes later Tave lowers to 557°F.

- PRZR pressure drops to 1830 psig and is now slowly rising.

The UO notes the following:

- Both MFPTs are tripped and all FWI valves are closed.

Which ONE of the following choices CORRECTLY lists the first initiating signals for the FWI valves closure and the MFPTs trip?

FWI Valves Closure MFPTs Trip A. Low PRZR pressure SI Low PRZR pressure SI B Reactor trip coincident with low Tave Low PRZR pressure SI C. Low PRZR pressure SI Reactor trip coincident with low Tave D. Reactor trip coincident with low Tave Reactor trip coincident with low Tave KIA 008 Pressurizer Vapor Space Accident AAI .04 Ability to operate and I or monitor the following as they apply to the Pressurizer Vapor Space Accident:

Feedwater pumps K/A MATCH ANALYSIS The question presents a plausible scenario where a PRZR Safety Valve has lifted. The student has to diagnose that an SI would have occurred from given conditions and also determine the reasons for the trip of the MFPTs and the closure of the FWI valves.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. The MFPT5 would have tripped on low PRZR pressure SI, the FWI valves would have closed on P-4 with Lo Tave.

B. Correct. P4 with Lo Tave will cause the FWI valves to close, SI will cause the 27

HL-15R RO NRC Exam C. Incorrect. While PRZR lo pressure SI would cause a FWI closure and is plausible, the first signal would have been from P-4 with Lo Tave. P-4 with Lo Tave does not cause a MFPT trip but causes a FWI. It is plausible the students may think P-4 with Lo Tave could have caused the MFPTs to trip.

D. Incorrect. The FWI valves closure on P-4 with Lo Tave is correct. P-4 with Lo Tave does not cause a MFPT trip but causes a FWI. It is plausible the students may think P-4 with Lo Tave could have caused the MFPTs to trip.

REFERENCES V-LO-PP-18101, Condensate and Feedwater System slide # 146 (included)

V-LO-PP-28193, Reactor Trip and ESFAS Signals slides # 129, 132, 135, and 136 (included)

V-LO-TX-28101, Reactor Protection System, section B for Reactor Trip and ESFAS Signals, pages # 28, 29, and 30. (included)

VEGP learning objectives:

LO-PP-18101-12, Describe the operation of the Main Feedwater Pump Turbine to include:

h. How the Main Feedwater Pump Turbine will respond to a Safety Injection signal.

LO-PP-28103-05, List all ESF actuation signals with applicable set points, coincidences, permissives, blocks, and discuss the systems response to each ESF actuation signal.

28

Condendate and Feedwater System I -i

.0 V-LO-PP-18101 Rev dIIIJL1J13 146 Conditions that will automatically Trip a Main Feedwater Pump

• Safety Injection (either train)

• Steam Generator High-High Level (2 out of 4 instruments >82%

on I out of 4 Steam Generators)

• Turbine overspeed (110%)

• Low suction Pressure (2 out of 3 instruments on I out of 2 Pumps @ 255 psig with a 20 second time delay)

• Turbine exhaust valve not full open Main Feedwater Pump Turbine tripped is permissive to enable Turbine Exhaust Valve operation using local hand switch.

• Low pump bearing oil pressure (4 psig)

• Low turbine bearing oil pressure (4 psig)

• Turbine thrust bearing wear

• Manual trip from the Main Control Room

• Manual trip locally

• Low Main Condenser Vacuum (13.5 inches Hg Vacuum)

Can be overridden to allow the Main Feedwater Pump reset to establish Steam Generator Blowdown Flow.

V-LO-PP-1 8101 Rev-08 146

What actuation signals input to SI?

-High-i containment pressure 2/3 channels > 3.8 psig.

-Low pressurizer pressure 2/4 channels < 1870 psig.

(may be manually blocked below P-u)

-Low steam line pressure 2/3 channels < 585 psig.

(may be manually blocked below P-li)

V-LO-PP-281 03-6.2

1 IA 11 f 1O kAC (CONTAINMENT (MAIN STEAM LINE 1 REACTOR TRIP XI SECOND MANUAL F1 SIGNAL V-LO--21u-o.2 132 (TkL, ( C1Iue t(

ti1ves)

What actuation signals input to Feed Water Isolation?

P14/SI S/G Hi-Hi LVL FWI

-SI

-P-14

-P-4 with low Tavg (2/4 channels <564°F)

LO TAVG AND REACTOR TRIP FW VLVS CLOSE V-LO-PP-281 03-6.2 135

What is the function of the FWI signal generated by SI or P-4 with low T ? avg (ec T&4p 1cd,j Prevents excessive RCS cooldown by isolating feedwater to the Steam Generators FWI auto close: L1t Ic) 1 tiodve, Main and Bypass Feed Water Isolation Valves and the Main and Bypass Feed Water Regulating Valves.

Main Feed Water Regulating Valves require both trains to isolate.

V-LO-PP-28103-6.2 136

19) General Warning Reactor Trip 2 out of 2 SSPS General Warning Alarms Bases: Prevents reactor operation while both trains of SSPS indicate abnormal conditions.

28.14 ESFAS Actuation Signals Safety Injection (SI)

Purpose:

To protect the core from a loss of coolant water to prevent overheating and damage to the fuel and/or fuel cladding and to provide boric acid to the core for emergency boration.

SI Actuation Signals Coincidence Set point Low Pressurizer Pressure SI 2 out of 4 channels = 1870 psig Can be manually blocked below P-i 1 Containment Pressure High I SI 2 out of 3 channels = 3.8 psig Low Steam Pressure SIISLI 2 out of 3 channels = 585 p5jg*

I out of 4 steam lines Can be manually blocked below P-i 1 Manual SI I out of 2 hand switches Set point is rate sensitive Systems affected by Safety Injection Signal

1) Reactor Trip

2) Turbine Trip

3) Main Feed Pump Trip

4) Feed Water Isolation

5) Motor Driven Auxiliary Feed Water Pump Start

6) MDAFW pumps discharge valve open signal

7) Steam Generator Blowdown valves close signal

8) Steam Generator Sample Valves close signal

9) Containment Isolation Phase A (CIA)

10) Containment Ventilation Isolation (CVI)

11) Control Room !solation (CR1)

12) Essential Chillers start

13) Diesel Generator Emergency start

14) CVCS normal charging and safety grade charging isolates

15) Emergency Core Cooling System (ECCS) start:

- CCPs

- SIPs

- RHR pumps

- ECCS valve alignment

16) NSCW pump starts V-LO-TX-281 01-08.1 28

17) NSCW cooling tower blow down isolates.

18) Containment Coolers supply and return valves receive open signals

19) Containment Coolers start in slow stopped

20) Reactor cavity and aux containment coolers supply and return valves receive close signals.

21) Non-1E buses 1NBO1 and 1NB1O load shed (Stub Buses)

Safety Iniection Reset and Block Safety Injection signals can be reset even if the actuation signal is still present if the following is satisfied:

1) 60 seconds has past since the actuation (Timer TD-1 on both trains of SSPS)

2) Both trains P-4 must be present to seal in the reset signal.

(prevents subsequent re-actuation if initiating signal is still present or from another automatic actuation signal)

3) Reset SI by using both A and B train SI reset hand switches (1 for each train of SSPS)

Located on the Main Control Board C panel When the Safety Injection is reset no equipment changes status. The reset only allows the operator to secure equipment auto started by the actuation.

Feed Water Isolation (FWI)

Purpose:

Isolates feed water to the Steam Generators to prevent rapid cool down of the reactor coolant system also prevents overfilling the steam generators and introduction of water in the steam lines and main turbine.

FWI Actuation Signals

1) P-4 (Reactor Trip) Lo Tavg 564°F on 2 out of 4 Loops

2) Safety Injection (also trips Main Feed Pumps and the Main Turbine)

3) P-14 Hi-Hi Steam Generator Level = 82 % on 2 out of 4 channels in I out of 4 S/Gs (also trips Main Feed Pumps and the Main Turbine)

Equipment affected by the Feed Water Isolation signal V-LO-TX-281 01-08.1 29

1) Main Feed Water Isolation Valves receive close signals

2) Bypass Feed Water Isolation Valves receive close signals

3) Main Feed Water Regulating Valves receive close signals *

4) Bypass Feed Water Regulating Valves receive close signals

• Main Feed Water Regulating Valves are unique in the fact that they require a FWI actuation signal from both SSPS trains to auto close. This is due to the two trains of instrument air solenoid valves being arranged in parallel. The parallel arrangement minimizes the chance of inadvertent feed water isolation on a single solenoid failure.

Resetting Feed Water Isolation Depending on what caused the feed water isolation will determine the method in which it can be reset.

Feed Water Isolation due to Lo Tavg I reactor trip can be reset simply by the use of the Feed Water Isolation Reset switches. (one hand switch for each train of SSPS).

Feed Water Isolation due to SI or P-14 (a.k.a. Full FWI) is different.

1) The actuation signal must be reset or clear

2) The P-4 seal in must be taken away. This is performed by what is know as cycling the reactor trip breakers (Closing the reactor trip breakers and allowing them to re-open) Caution: If the FWI is due to a Safety Injection the actuation signal must be cleared before cycling the trip breakers.

If not Safety Injection actuation will re-occur due to the auto SI block being removed.

3) Reset FWI using the FWI Reset Hand Switches Main Steam Line Isolation (MSLI)

Purposes: 1) Prevents excessive cool down of the RCS for main steam line breaks down stream of the isolation valves.

2) Limits the blow down to the affected faulted steam generator this limits the cool down of the RCS and limits the amount of containment pressurization.

MSLI Actuation Signals Coincidence Set point

1) Low Steam Pressure Sl/SLI 2 out of 3 channels = 585 psig I out of 4 steam lines Set point is rate sensitive Can be manually blocked below P-il

2) Hi Steam Pressure negative Rate 2 out of 3 channels = 100 psig V-LO-TX-281 01-08.1 30

HL-15R RO NRC Exam

14. 008K2.02 OO1/2/1/CCW-PMP POWERIMEM 3.O/3.2/NEW/HL-15RNRC/RO/DS / TNT Which one of the following choices lists all the procedurally allowable power supplies for Unit 1 CCW pumps 1,3, and 5 per 13427A-1, 416OVAC Bus 1AAO2 1E Electrical Distribution System?

A RAT-i A, RAT-1B, SAT, EDG-1A B. RAT-iA, UAT back feed, SAT, EDG-2A C. RAT-iA, RAT-2A, SAT, EDG-1A D. RAT-lA, UAT back feed, SAT, EDG-2A K/A 008 Component Cooling Water System (CCWS)

K2.02 Knowledge of the power supplies to the following:

CCW pump, including emergency backup.

K/A MATCH ANALYSIS The question requires the student to identify all possible power sourcecs for COW pumps 1, 3, and 5 on Unit 1, including emergency backup power supplies matching the K/A topic.

ANSWER / DISTRACTOR ANALYSIS A. Correct. All 3 of these pumps are powered from bus 1AAO2. The allowable power feeds to this bus are from its normal source RAT iA, emergency source EDG-iA, and alternate sources RAT-lB and the SAT per procedure i3427A-1.

B. Incorrect. There is no procedurally directed method to connect Unit 2 DGs to any unit 1 1E busses. UAT back feed is used only for Non-1E 416OVAC busses, making this choice plausible but incorrect.

C. Incorrect. There is no procedurally directed method to connect Unit 2 RATs to any unit 1 1 E busses, however there are several phyically possble connections making this choice plausible.

D. Incorrect. There is no procedurally directed method to connect Unit 2 DGs to any unit 11 E busses. The UAT back feed is used only for Non-i E 41 60V AC busses, making this choice plausible but incorrect.

REFERENCES 29

HL-15R RO NRC Exam 13427A-1, 4160V AC Bus 1AAO2 IE Electrical Distribution System pages 2, 5, 7, and 40, 1 3425A-1, 41 60V AC Non I E Bus 1 NAO1 Electrical Distribution System page 2 V-LO-TX-10101 CCW System page 10 VEGP learning objectives:

LO-PP-01 101-01:

List all offsite electrical power sources.

LO-PP-01 101-02:

Describe switchyard configuration (including SAT alignment) when:

a. main generator is on-line

b. main generator is shutdown LO-PP-01 101-07:

Describe the requirements for energizing a de-energized 13.8kV or 4.16 kV bus.

LO-PP-01 101-08:

Describe the requirements for transferring from the alternate incoming to the normal incoming supply for 13.8 kVor4i6kV busses 30

Approved By Procedure Number Rev S. A. Phillips Vogtle Electric Generating. Plant z 1 3427A-1 5 Date Approved Page Number 4/3/08 416OVAC BUS 1AAO2 1E ELECTRICAL DISTRIBUTION SYSTEM 2 of 58 TABLE OF CONTENTS Section Page 1.0 PURPOSE 3 2.0 PRECAUTIONS AND LIMITATIONS 3 3.0 PREREQUISITES OR INITIAL CONDITIONS 7 4.0 INSTRUCTIONS 7 4.1 Startup 7 4.1.1 Energizing 416OVAC Bus 1AAO2 From Normal Incoming Source [RAT or SAT] 7 4.2 System Operation 12 4.2.1 Paralleling Diesel Generator 1A To 4160V AC Bus 1AAO2 12 4.2.2 Paralleling Normal Incoming Source (RAT or SAT) To 4160V AC Bus 1AAO2 Being Supplied From Diesel Generator 1A 18 4.2.3 Paralleling Alternate Incoming Source To 416OVAC Bus 1AAO2 Being Supplied From Diesel Generator 1A 21 4.2.4 Discontinue Parallel Operation by Removing the Normal Incoming Source (RAT or SAT) from Bus 1AAO1 26 4.2.5 Discontinue Parallel Operation by Removing the Diesel Generator 1A from Bus 1AAO2 28 4.2.6 Discontinue Parallel Operation by Removing the Alternate Incoming Source from Bus 1AAO2 30 4.3 Shutdown 32 4.3.1 De-energizing 416OVAC Bus 1AAO2 Bus When Supplied from the Normal Incoming Source (RAT or SAT) or Alternate Incoming Source 32 4.4 Non Periodic Operations 36 4.4.1 Energizing 4160V AC Bus 1AAO2 from the Alternate Incoming Source 36 4.4.2 Transferring 4160V Bus 1AAO2 from the Normal Incoming Source (RAT or SAT) to the Alternate Incoming Source 40 4.4.3 Transferring 4160V Bus 1AAO2 from the Alternate Incoming Source to the Normal Incoming Source (RAT or SAT) 44 4.4.4 Energizing 4160V 1E Bus 1AAO2 from Diesel Generator 1A When the Sequencer is Not Available 46

5.0 REFERENCES

55 Table 1 416OVAC 1E Bus 1AAO2 Load List 57 Figure 1 Emergency Diesel Generator Operating Limits 58 Pdnted October 29, 2009 at 10:09

Approved By Procedure Number Rev S. A. Phillips Vogtle Electric Generating Plant i3427A-1 5 Date Approved Page Number 4/3/08 4160V AC BUS 1AAO2 1 E ELECTRICAL DISTRIBUTION SYSTEM 5 of 58 2.2.3 During MODEs 1, 2, 3, and 4, in order to ensure that one RAT (NOT the SAT) has adequate capacity and capability to start and run both trains of ECCS loads during the connection, the following conditions shall be met:

a. Grid voltage shall be maintained at or above the minimum expected 100%

grid voltage (as determined by Power Control Center [PCC]) while the busses are interconnected to one RAT.

Should grid voltage degrade below minimum, the transfer must be completed as expeditiously as possible or the alignment returned to separate sources.

b. No additional non 1 E 41 60V AC loads, other than those normally fed from the class 1 E 41 60V AC busses, shall be manually connected to the one RAT feeding both class 1 E 41 60V AC busses.

c. During MODE 1, the automatic bus transfer schemes for the non 1 E 41 60V AC busses shall be disabled during the connection of both 1 E 41 60V AC trains to one RAT. (The 1 3.8kV fast and residual voltage bus transfer schemes for the remaining RAT in service need not be disabled.)

2.2.4 During MODEs 5 and 6, one qualified circuit between offsite transmission network and onsite Class 1 E Distribution System shall be operable per Technical Specification LCO 3.8.2 and a minimum of one Class 1 F 41 60V AC bus shall be energized per Technical Specification LCD 3.8.10.

2.2.5 The Standby Auxiliary Transformer (SAT) shall only be used as one offsite power source for only one unit.

2.2.6 During MODEs 5 and 6 both Class 1E 416OVAC buses may be manually connected to the RAT OR SAT. During that configuration the following precautions apply:

a. When connected to the RAT, the loads on the secondary side shall not exceed:

(1) 416OVAC 1E loads less than 1350 amps total.

(2) 41 60V AC Non-i E loads less than 7500 kVA total.

(3) 416OVAC Non-iF loads less than 1000 amps total.

Printed October 29, 2009 at 10:10

Approved By . . Procedure Number Rev S. A. Philhps Vogtle Electric Generating Plant 1 3427A-1 5 Date Approved Page Number 4/3/08 4I6OVAC BUS 1AAO2 1E ELECTRICAL DISTRIBUTION SYSTEM 7 of 58 INITIALS 3.0 PREREQUISITES OR INITIAL CONDITIONS 3.1 Verify 4160V AC Bus 1AAO2 is aligned per 11427-1, 416OVAC 1 E Electrical Distribution System Alignment.

3.2 Verify 125V DC electrical power is available to supply breaker control power.

4.0 INSTRUCTIONS 4.1 STARTUP NOTE Unless otherwise noted, all switch manipulations are performed at the QEAB in the Control Room.

4.1.1 Energizing 4160V AC Bus IAAO2 From Normal Incoming Source [RAT or SAT]

4.1.1.1 Verify a Normal Incoming Source is available:

a. IF 1AAO2 will be energized from RAT 1NXRA, verify applicable sections of 13415-1, Reserve Auxiliary Transformers, have been performed PRIOR to performing this section.

OR

b. IF 1AAO2 will be energized from SAT, verify applicable sections of 13418-C, Standby Auxiliary Transformer, have been performed PRIOR to performing this section.

Printed October 29, 2009 at 10:31

Approved By . Procedure Number Rev S. A. Philips Vogtle Electric Generating Plant 13427A-1 5 Date Approved Page Number 4/3/08 416OVAC BUS 1AA02 1E ELECTRICAL DISTRIBUTION SYSTEM 40 of 58 INITIALS 4.42 Transferring 4160V Bus IAAO2 From The Normal Incoming Source (RAT or SAT) To The Alternate Incoming Source CAUTIONS

• Performance of Section 4.4.2 is allowed during MODEs 1, 2, 3, and 4 by Technical Specification LCO 3.8.9 for transfer of offsite power sources only.

Otherwise, it is prohibited in MODEs 1, 2, 3, and 4 by Technical Specification LCO 3.8.1 and LCO 3.8.9.

• The interconnection of both Class 1 E 41 60V AC busses is intended to be temporary and the time in this configuration will be minimized to the extent necessary in order to achieve a safe transfer of offsite power sources. The 41 60V AC 1 F breakers are expected to remain closed for approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> during this temporary alignment.

4.4.2.1 Verify an Alternate Incoming Source is available.

a. IF 1AAO2 will be energized from RAT 1NXRB, verify applicable sections of 13415-1, Reserve Auxiliary Transformers, have been performed prior to performing this section.

OR

b. IF 1AAO2 will be energized from SAT, verify applicable sections of 13418-C, Standby Auxiliary Transformer, have been performed prior to performing this section.

4.4.2.2 Verify Limitations in Section 2.2 are met before performing this section.

4.4.2.3 Verify 14230-1, AC Source Verification, has been successfully performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> prior to interconnection.

Critical 4.4.2.4 IF Non-Class 1 E 41 60V AC Buses associated with RAT 1 NXRB are energized from UAT5 by backfeed, place handswitch 1 HS-1 NAO4O1 Alternate Incoming Breaker in PULL-TO-LOCK and Caution Tag.

CV Printed October 29, 2009 at 10:24

Approved By . . Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant z. 13425A-1 1.0 DateApproved 416OVAC NON 1E BUS 1NAO1 ELECTRICAL DISTRIBUTION Page Number 3/4/09 SYSTEM 2 of 10 1.0 PURPOSE This procedure provides the instructions for the operation of the 41 60V AC Non 1 F Electrical Distribution System. Instructions are included in the following sections:

4.1.1 Energizing 4160V Bus 1NAOI from Alternate Incoming Source (RAT or SAT) 4.2.1 Transferring 4160V Bus 1NAO1 from Alternate Incoming Source (RAT Or SAT) to Normal Incoming Source (UAT) 4.2.2 Transferring 4160V Bus 1NAO1 from Normal Incoming Source (UAT) to Alternate Incoming Source (RAT/SAT) 4.3.1 De-Energizing 4160V Bus 1NAO1 4.4.1 Energizing 4160V Bus 1NAOI from Normal Incoming Source (UAT) 2.0 PRECAUTIONS AND LIMITATIONS

/MT J4,/ /

Jo 2.1 PRECAUTIONS If a bus is not aligned to its normal supply for current plant conditions, Caution Tags should be generated and installed on the applicable handswitches to alert the operator to the off normal alignment.

2.2 LIMITATIONS 2.2.1 During MODEs 5 and 6 with both Class 1 E 41 60V AC buses manually connected to the Standby Auxiliary Transformer (SAT), all the 41 60V AC Non-Class 1 E buses fed by the SAT are shed and the automatic bus transfer schemes shall be disabled to prevent any auto swaps from the UAT to the SAT, and the automatic Safety Injection Signal from SSPS for one train of ECCS is blocked.

2.2.2 During MODEs 5 and 6 with both Class 1 E 4160V AC buses manually connected to the same Reserve Auxiliary Transformer [RAT] and a UAT backfeed established to the Non-class 1 E buses, the automatic bus transfer schemes shall be disabled for all the Non-Class 1 E busses [13.8 and 4.16 KV] fed by that RAT to prevent any auto swaps from the UATs to the RATs.

Printed October 29, 2009 at 10:34

\/- LoT t\o 10.3.1 Component Cooling Water Pump & Motor There are six component cooling water pumps arranged three in parallel per train (see Figure 10-1). Each train has three pumps for backup protection and redundancy. The pumps are 50% capacity each. Therefore, during normal operation, two pumps are running and one pump is in standby. This ensures sufficient flow for adequate heat removal to the system components during all modes of operation. The standby pump will start automatically on low pressure, when the pump discharge header reaches 65 psig (reference 17003-1 ALE A06), and another pump is running or when a running pump is tripped by protective functions.

The pumps are Ingersoll-Rand single-stage, double-suction centrifugal pumps driven by 300 HP Westinghouse motors at 1761 rpm. They are rated for 5000 gpm each at a discharge head of 160 ft. Mechanical seals are used to prevent leakage from the pump casing. Cooling for the motor is provided by circulating NSCW through the motor cooler and transferring the heat to the Ultimate Heat Sink. A l2OV, 2A space heater is used to keep the motor windings dry when shut down.

Power to the component cooling water pumps is supplied by the 4.16 kV ESF busses. Component cooling pumps 1, 3, and 5 receive power from 4.16 kV ESF bus AAO2 with component cooling water pumps 2, 4, and 6 10 Revision 3.0

HL-15R RO NRC Exam

15. 009EK2.03 001/1/1/SMALL LOCA-S/G/C/A 3.0 / 3.3/NEW/HL-15R NRC/RO/TNT/DS Given the following:

- A 200 gpm RCS leak is in progress.

- RCS pressure is 1465 psig and stable.

- Containment pressure is 2.1 psig and rising very slowly.

- The crew transitions to 19012-C, E-l .2 Post LOCA Cooldown & Depressurization.

Which ONE of the following is CORRECT regarding minimum SIG NR water level required for these plant conditions and why?

A. 10%, ensures SIC tubes are covered to promote reflux boiling.

B. 32%, ensures S/C tubes are covered to promote reflux boiling.

Cs 10%, ensures S/G inventory to ensure a secondary heat sink.

D. 32%, ensures S/C inventory to ensure a secondary heat sink.

31

HL-15R RO NRC Exam KIA 009 Small Break LOCA EK2.03 Knowledge of the interrelations between the small break LOCA and the following:

SIGs K/A MATCH ANALYSIS The question presents a plausible scenario where a small break LOCA is in progress with some given plant parameters. The candidate must determine the minimum SIG NR level required for plant conditions (non-adverse Containment) and the basis for these levels. With a small break LOCA in progress with the given leak rate, the candidate must be aware that SIG levels are necessary for secondary heat sink and reflux cooling would not be heat removal mechanism for a small break LOCA of this size.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. 10% SIG NR level is required due to non-adverse Containment.

Reflux cooling would NOT be a heat removal mechanism for a small break LOCA of this size.

B. Incorrect. 32% SIG NR level is NOT required due to non-adverse Containment.

Ensures adequate feedwater flow or secondary heat sink is basis.

C. Correct. 10% S/G NR level required part is correct. Reflux cooling would NOT be a heat removal mechanism for a small break LOCA of this size.

D. Incorrect. 32% SIG NR level is NOT required and to ensure adequate feedwater flow or secondary heat sink is the basis.

REFERENCES V-LO-HO-371 11-001, Loss of Reactor or Secondary Coolant pages # 4 and # 8 (included).

VEGP learning objectives:

LO-LP-37111-02, State the effect of various size breaks on the primary system with respect to temperatures and pressures.

32

DESCRI PTION Small RCS LOCA with one train of ECCS __

VVC}P (5 One train of safety injection is assumed, and loss of offsite power is assumed to occur at the reactor trip time. The only means of venting steam on the secondary side is through the steam generator safety valves. Minimum auxiliary feedwater is assumed available one minute after the reactor trip time.

For these break sizes the normal makeup system cannot maintain level and pressure. The RCS will depressurize and an automatic reactor trip and safety injection signal will be generated. Provided that a secondary side heat sink exists, the RCS will reach an equilibrium pressure which corresponds to the pressure at which the liquid phase break flow equals the high pressure pumped safety injection flow,

/

which is above the SG pressures.

RCs Core heat is removed through the steam generators by continuous single or two phase natural circulation, ad Once equilibrium pressure is established there is no further net loss of liquid volume in the RCS. The coe &-

/

natural circulation heat removal mode continues until the time that the break can remove all the decay heat. Prior to this time, auxiliary feedwater is required to maintain the heat sink.

Abnormal indications should be present in the containment for this category of LOCA although the response will be slower and milder than for larger break size LOCAs. Containment pressure will probably not reach the containment High I pressure. (\,,

,q 5e /1,k2l 2c 1 e Li, Figure 1 shows the RCS pressure transient for this case. The RCS pressure stabilizes slightly above the steam generator safety valve set pressure. Figures 2 and 3 show the safety injection and break flows which are both stabilized at a flowrate of approximately 340 gpm. Figure 4 shows that the pressurizer empties at approximately 10 minutes and does not refill. The system remains in a stable condition with the core covered and decay heat being adequately removed.

The equilibrium pressure condition is stable for the long term provided that SI and auxiliary feedwater are available. Since the RCS pressure at the equilibrium condition is determined by a balance between break and ECCS flowrate, in order to depressurize to a cold shutdown condition it is necessary to cool the primary fluid further while stepping down the ECCS flowrate. Long term cooldown/depressurization of the plant is performed using 19012-C, ES 1.2 POST LOCACOOLDOWN AND DEPRESSURIZATION.

/- Ld (-(O-37(1/

4

Medium LOCA Breaks (un 2 to I ft )

2 During the early stages of the depressurization, the break flow is not capable of removing all the decay heat. At this pressure, pumped safety injection flow is less than the break flow, and there is a net loss of mass in the RCS. Voiding throughout the primary side occurs and eventually the RCS begins to drain, starting from the top of the steam generator tubes. The rate of RCS drain is determined by the net loss of liquid inventory, a function of both ECCS flow and break size.

Prior to the start of draining, heat is removed from the steam generator through continuous two phase natural circulation, with two phase mixture flowing over the top of the steam generator tubes. As the draining continues, the natural circulation mode of heat removal ceases, and core heat is removed

/

through condensation of s am in the steam enerator. This method of heat removal is called reflux.

J 5 elvu Mqfr The condensation mode of heat removal is almost as efficient as continuous two phase natural circulation in removing heat. However, condensation heat transfer coefficients may be lower than continuous two phase natural circulation heat transfer coefficients. The steam generator secondary side pressurizes to the safety valve set pressure early in the transient, and remains there throughout the natural circulation and steam condensation heat removal modes. Eventually the primary fluid may drop completely below the steam generator tubes and begin to drain other regions in the RCS. Depending on the location of the break, the draining may partially uncover the core.

As soon as the break flow becomes all steam flow for breaks in this range of size, steam generated in the core can exit out the break, and further system depressurization occurs. Safety injection flow increases to greater than the break flow, and there is no longer a net loss of mass from the RCS. No further core uncovery will occur under these conditions. The steam generator may still be relied upon for heat removal by the condensation mode. However, only a small amount of heat removal by the steam generator is necessary and, with minimum auxiliary feedwater available, the steam generator secondary side will now begin to slowly depressurize below the steam generator safety valve set pressure. The primary system will also slowly depressurize along with the secondary side.

The RCS pressure plot, Figure 6, shows a rapid depressurization to approximately 1200 psig at 5 minutes. Immediately after the draining of the crossover leg, the break uncovers and the break flow becomes all steam. This can be seen by a rapid decrease in break flow at that time on the plot in Figure 7, indicating a change from two phase to all steam flow. Because the location of the break in this sample transient is the cold leg, the core level also decreases in conjunction with the crossover leg draining (Figure 8). As steam is relieved out the break, the core pressure decreases relative to the downcomer pressure, and the hydrostatic head in the downcomer recovers the core. Note that this core level HO 8

HL-15R RO NRC Exam

16. O1OK1.08 OO1/2/1/PRZRPRESS-PZRLCS/C/A -3.2 / 3.5INEW/HL-15RNRCIRO/DS /TNT Complete the following statement for the PRZR Pressure Control System:

When PRZR level the PRZR heaters will de-energize.

A. drops below 17% backup only B. drops 5% below program level backup and proportional C drops below 17% backup and proportional D. drops 5% below program level backup only 33

HL-I5RRO NRC Exam KIA 010 Pressurizer Pressure Control System (PZR PCS)

KI .08 Knowledge of the physical connections and/or cause-effect relationships between the PZR PCS and the following systems:

PZR LCS.

K/A MATCH ANALYSIS The question requires to student to determine the effects of high or low PRZR level on the PRZR pressure control system, matching the K/A topic.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Backup heaters would de-energize, however, the proportional heaters would also de-energize.

B. Incorrect. At 5% above program the PRZR backup heaters will automatically energize. At 5% below program PRZR heaters do not change state.

C. Correct. At 17% PRZR level all PRZR heaters are interlocked off to protect the heaters and I or the heater well from damage due the water level dropping below the heaters.

D. Incorrect. At 5% above program the PRZR backup heaters will automatically energize. At 5% below program PRZR heaters do not change state.

REFERENCES 17011-1, Annunciator Response Procedures for ALB 11 on Panel 1 Cl on MCB Windows:

BOl, PRZR LO LEVEL HTR CNTL OFF LTDN SECURED COl, PRZR CONTROL HI LEVEL DEV AND HEATERS ON DOl, PRZR LO LEVEL DEVIATION VEGP learning objectives:

LO-PP-1 6302-04:

Describe the Hi and Low Pressurizer level protection features including the set points, coincidence, and reason for each.

34

Approved By Procedure Number Rev S. E. Prewitt Vogtle Electric Generating Plant 17011-1 14.2 Date Approved ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 11 ON PANEL 101 Page Number 12-12-2005 ON MOB 14of52 WINDOW BOl ORIGIN SETPOINT PRZR LO LEVEL 1-LT-0459 17% HTR CNTL OFF 1-LT-0460 LTDN SECURED 1-LT-0461 1.0 PROBABLE CAUSE

1. Pressurizer level Control System Malfunction.

2. Charging - Letdown System Malfunction.

3. RCS cooldown.

4. Reactor Coolant System leak.

2.0 AUTOMATIC ACTIONS

1. All Pressurizer Heaters turn off.

2. Letdown isolation.

3.0 INITIAL OPERATOR ACTIONS

1. Check pressurizer level instrumentation.

2. IF instrument malfunction is indicated Go To 18001-C, Primary Systems Instrumentation Malfunctions.

3. IF a Reactor Coolant System leak is indicated, Go To 18004-C, Reactor Coolant System Leakage.

Printed October 29, 2009 at 12:24

WINDOW 001 ORIGIN SETPOINT PRZR CONTROL 1-LT-0459 5% above level HI LEVEL DEV 1-LT-0461 program AND HEATERS ON 1.0 PROBABLE CAUSE

1. Pressurizer Level Control System malfunction.

2. Charging-Letdown System malfunction.

3. Rapid reduction in secondary steam demand.

2.0 AUTOMATIC ACTIONS Pressurizer Backup Heaters energize.

3.0 INITIAL OPERATOR ACTIONS Check pressurizer level using 1 -LR-0459 recorder and if a Pressurizer Level Control System malfunction is indicated, initiate 18001-C, Primary Systems Instrumentation Malfunction.

4.0 SUBSEQUENT OPERATOR ACTIONS IF Pressurizer Level Control System is not correcting level, take manual control and adjust as required.

5.0 COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE

REFERENCES:

1X4DB112, 1X6AUO1-183, 168, 1X6AXO1-106, PLS Printed October 29, 2009 at 12:23

Approved By Procedure Number Rev S. E. Prewitt Vogtle Electric Generating Plant . 17011-1 14.2 Date Approved ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 11 ON PANEL 1C1 Page Number 12-12-2005 ON MCB 30 of 52 WINDOW DOl ORIGIN SETPOINT PRZR 1 -LT-0459 5% below level LO LEVEL 1-LT-0461 program DEVIATION 1.0 PROBABLE CAUSE

1. Pressurizer Level Control System malfunction.

2. Charging-Letdown System malfunction.

2.0 AUTOMATIC ACTIONS NONE 3.0 INITIAL OPERATOR ACTIONS Check pressurizer level using 1-LR-0459 recorder and if a Pressurizer Level Control System malfunction is indicated, initiate 18001-C, Primary Systems Instrumentation Malfunction.

4.0 SUBSEQUENT OPERATOR ACTIONS IF Pressurizer Level Control System is not correcting level, take manual control and adjust as required.

5.0 COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE

REFERENCES:

1X4DB112, 1X6AUOI-183, 168, PLS Printed October 29, 2009 at 12:23

HL-15R RO NRC Exam

17. O11EG2.2.38 001/1/1/LARGE LOCA-LICENSE/MEM 3.6 / 4.5INEW/HL-15RNRC/RO/TNT/DS To maintain Containment parameters within the accident analysis assumptions for a DBA LOCA, the Containment Pressure and Containment Air Temperature LCO limits for Mode 1 are...

A +1.8 psig and 120°F B. +1.8 psig and 130°F C. -3.0 psig and 120°F D. -3.0 psig and 130°F KIA 011 Large Break LOCA G2.2.38 Knowledge of conditions and limitations in the facility license.

KIA MATCH ANALYSIS The question presents a question where the student must know the LCO limits of Tech Specs 3.6.4 and 3.6.5 for Containment Pressure and Containment Air Temperature to prevent exceeding DBA LOCA analyzed design limits.

ANSWER I DISTRACTOR ANALYSIS A. Correct. +1.8 psig is the upper Containment Pressure limit and 120°F is the Containment Air Temperature limit.

B. Incorrect. +1 .8 psig is the correct upper Containement Pressure limit but 130°F is NOT the correct containment Air Temperature limit. 130°F was used since it is the temperature where SFP high temperature alarms and would be a plausible figure the students may recall. 130°F is also a common set point for various plant systems such as bypass valves operations for demin diverts, etc C. Incorrect. -3.0 psig is NOT the correct negative pressure limit. -3.0 is plausible as it is the figure used in the Tech Spec bases for the accident analysis but is NOT the LCO limit. Plausible the candidate could confuse this number with the LCO limit or be confused by the -0.3 psig negative LCO limit. 120°F is the correct Containment Air Temperature limit.

D. Incorrect. -3.0 psig is NOT the correct negative pressure limit. -3.0 is plausible as it is the figure used in the Tech Spec bases for the accident analysis but is NOT the LCO limit. Plausible the candidate could confuse this number with the LCO limit or be confused by the -0.3 psig negative LCO limit. 130°F was used since it is the temperature where SFP high temperature alarms and would be a plausible figure 35

HL-15R RO NRC Exam the students may recall. 1300F is also a common set point for various plant systems such as bypass valves operations for demin diverts, etc.

REFERENCES Tech Spec 3.6.4 Containment Pressure and the bases.

Tech Spec 3.6.5 Containment Air Temperature.

17005-1, window A06 for SEP Hi Temp.

VEGP learning objectives:

LO-LP-3921 0-01, For any item in section 3.6 of Tech Specs, be able to:

a. State the LCO LO-LP-39210-02, Given a set of Tech Specs and the bases, determine for a specific set of plant conditions, equipment availability, and operational mode.

a. Whether any LCO limits of section 3.6 of Tech Specs has been exceeded.

36

Containment Pressure 3.6.4 3.6 CONTAINMENT SYSTEMS 3.6.4 Containment Pressure LCO 3.6.4 Containment pressure shall be -0.3 psig and +1 .8 psig.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Containment pressure A.1 Restore containment 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> not within limits, pressure to within limits.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.1 Verify containment pressure is within limits. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Vogtle Units I and 2 3.6.4-1 Amendment No. 96 (Unit 1)

Amendment No. 74 (Unit 2)

Containment Air Temperature 3.6.5 3.6 CONTAINMENT SYSTEMS 3.6.5 Containment Air Temperature LCO 3.6.5 Containment average air temperature shall be 120°F.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Containment average air A.1 Restore containment 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> temperature not within average air temperature limit, to within limit.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.5.1 Verify containment average air temperature is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> within limit.

Vogtle Units 1 and 2 3.6.5-1 Amendment No. 96 (Unit 1)

Amendment No. 74 (Unit 2)

Containment Pressure B 3.6.4 B 3.6 CONTAINMENT SYSTEMS B 3.6.4A Containment Pressure BASES BACKGROUND The containment pressure is limited during normal operation to preserve the initial conditions assumed in the accident analyses for a loss of coolant accident (LOCA) or steam line break (SLB). These limits also prevent the containment pressure from exceeding the containment design negative pressure differential with respect to the outside atmosphere in the event of inadvertent actuation of the Containment Spray System.

Containment pressure is a process variable that is monitored and controlled. The containment pressure limits are derived from the input conditions used in the containment functional analyses and the containment structure external pressure analysis. Should operation occur outside these limits coincident with a Design Basis Accident (DBA), post accident containment pressures could exceed calculated values.

APPLICABLE Containment internal pressure is an initial condition used SAFETY ANALYSES in the DBA analyses to establish the maximum peak containment internal pressure. The limiting DBAs considered, relative to containment pressure, are the LOCA and SLB, which are analyzed using computer pressure transients. The worst case LOCA generates larger mass and energy release than the worst case SLB. Thus, the LOCA event bounds the SLB event from the containment peak pressure standpoint (Ref. 1).

The initial pressure condition used in the containment analysis was 17.7 psia (3.0 psig). This resulted in a maximum peak pressure from a LOCA of 36.5 psig. The containment analysis (Ref. 1) shows that the maximum peak calculated containment pressure, P , results from 8

the limiting LOCA. The maximum containment pressure resulting from the worst case LOCA, 36.5 psig, does not exceed the containment design pressure, 52 psig.

(continued)

Vogtle Units 1 and 2 B 3.6.4-1 Revision No. 0

Containment Pressure B 3.6.4 BASES APPLICABLE The containment was also designed for an external pressure SAFETY ANALYSES load equivalent to -3 psig. The inadvertent actuation of (continued) the Containment Spray System was analyzed to determine the resulting reduction in containment pressure. The initial pressure condition used in this analysis was 14.093 psia. This resulted in a minimum pressure inside containment of 11 .77 psia, which is less than the design load.

For certain aspects of transient accident analyses, maximizing the calculated containment pressure is not conservative. In particular, the cooling effectiveness of the Emergency Core Cooling System during the core reflood phase of a LOCA analysis increases with increasing containment backpressure. Therefore, for the reflood phase, the containment backpressure is calculated in a manner designed to conservatively minimize, rather than maximize, the containment pressure response in accordance with 10 CFR 50, Appendix K (Ref. 2).

Containment pressure satisfies Criterion 2 of 10 CFR 50.36 (c)(2)(ii).

LCO Maintaining containment pressure at less than or equal to the LCO upper pressure limit ensures that, in the event of a DBA, the resultant peak containment accident pressure will remain below the containment design pressure. Maintaining containment pressure at greater than or equal to the LCO lower pressure limit ensures that the containment will not exceed the design negative differential pressure following the inadvertent actuation of the Containment Spray System.

APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material to containment. Since maintaining containment pressure within limits is essential to ensure initial conditions assumed in the accident analyses are maintained, the LCO is applicable in MODES 1, 2, 3 and 4.

In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature (continued)

Vogtle Units 1 and 2 B 3.6.4-2 Rev. 1-10/01

Approved By Procedure Number Rev AS. Parton Vogtle Electric Generating Plant 17005-1 30 Date Approved ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 05 ON Page Number 5/26/09 PANEL 1A2 ON MOB 14of66 WINDOW A06 ORIGIN SETPOINT SPENT FUEL 1-TISH-626 130°F PIT HI TEMP w1J STç

• -L 1.0 PROBABLE CAUSE

1. Spent Fuel Pool Pump trip.

2. Loss of Component Cooling Water (COW) flow to Spent Fuel Pool Heat Exchanger.

2.0 AUTOMATIC ACTIONS NONE 3.0 INITIAL OPERATOR ACTIONS Go To 18030-C, Loss Of Spent Fuel Pool Level Or Cooling.

4.0 SUBSEQUENT OPERATOR ACTIONS NONE 5.0 COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE

REFERENCES:

1X4DB13O, PLS Printed December 17, 2009 at 14:47

HL-15R RO NRC Exam

18. 012K1.05 OO1/2/1/RPS-ESFAS/C/A -.3.8 / 3.9/NEW/HL-15R NRC/RO/DS / TNT Five minutes following a reactor trip and safety injection, the OATC places the Train A SI reset handswitch in the reset position.

P-4 train A will be generated when the Train A reactor trip breaker is open (I)

SI reset will block (2)

A. (1) and the Train B bypass breaker is open (2) only the Train A PRZR Low Pressure SI and Low Steamline Pressure SI signals B (1) and the Train A bypass breaker is open (2) all Train A automatic SI signals C. (1) or the Train A bypass breaker is open (2) only the Train A PRZR Low Pressure SI and Low Steamline Pressure SI signals D. (1) or the Train B bypass breaker is open (2) all Train A automatic SI signals K/A 012 Reactor Protection System KI .05 Knowledge of the physical connections and/or cause effect relationships between the RPS and the following:

ESFAS.

KIA MATCH ANALYSIS The question requires the student to correctly identify the effects on the ESFAS system and the position of the reactor trip and bypass breakers when resetting the SI signal, matching the K/A topic.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Both the Train A reactor trip and bypass breakers must to be open in 37

HL-15R RO NRC Exam would be blocked. The Low PRZR Pressure and Low Steamline Pressure SI signals may be manually blocked when PRZR pressure is < P-il (2000 PSIG). This operation is performed in the UOPs and EOPs.

B. Correct. P-4 Train A signal is generated when both the train A reactor trip and byapss breakers are open. Once SI is reset with the P-4 signal present, all automatic SI signals are blocked.

C. Incorrect. Both the train A reactor trip and bypass breakers must to be open in order to generate a Train A P-4 signal. If P-4 were present all automatic SI signals would be blocked. The Low PRZR Pressure and Low Steamline Pressure SI signals may be manually blocked when PRZR pressure is < P-il (2000 PSIG). This operation is performed in the UOPs and EOPs.

D. Incorrect. Once SI is reset with the P-4 signal present, all automatic SI signals are blocked. Both the Train A reactor trip and bypass breakers must to be open in order to generate a Train A P-4 signal.

REFERENCES V-LO-TX-28101 Reactor Protection System page 18 VEGP learning objectives:

LO-PP-28i 03-07:

Discuss SI reset to include:

a. Time delay

b. SI reset with P-4

c. SI reset without P-4

d. Auto and Manual actuation capabilities following reset 38

SECTION B REACTOR TRIP AND ESFAS SIGNALS 28.11 PERMISSIVE INTERLOCKS Permissive interlocks provide input to the protection systems to allow or prevent protective functions from occurring under certain plant conditions.

P-4 Indicates reactor tripped Set point or conditions that give P-4 RTA and its bypass (BYA) both open give P-4 Train A RTB and its bypass (BYB) both open give P-4 Train B Function:

1) Trips the Main Turbine to limit the RCS cool down

- P-4 Train A generates a Mechanical Turbine Trip

- P-4 Train B generates an Electrical Turbine Trip

2) Steam Dumps

- P-4 Train A generates a Steam Dump Arming signal

- P-4 Train B transfers Steam Dump controllers from Load reject mode to the Plant trip mode

3) Feed Water Isolation (FWI)

- P-4 in conjunction with Lo Tavg of 564°F

4) Seals in FWI if caused by a Safety Injection or Hi-Hi Steam Generator water level (P-14).

5) SI reset logic

- After Safety Injection has been reset, P-4 blocks any future automatic safety injection signals.

P-6 Source Range Block Permissive Set point:

2.0 x 10-5% POWER on any 11 2 IR NIS detector.

Function:

1) Allows the operator to manually block SR high flux trip.

(both TRN A and TRN B switches, QMCB-C)

2) Loss of P-6 (either train no SSPS) will automatically unblock the Source Range Trip Permissive status light on BPLP (QMCB-C) Illuminates when P-6 is present.

P-6 Resets:

When 2/2 IR NIS detector drop <7.0 X 10-6% POWER V-LO-TX-281 01-08.1 18

HL-15R RO NRC Exam

19. 012K5.O1 OO1/2/I/RPS-DNBIMEM 3.3 / 3.8/M CATAWBA 08/HL-I5RNRC/RO/DS/TNT Which one of the following correctly matches the reactor trip signals to their limiting accident I protection?

Reactor Trip Signal Limiting Accident I Protection A. Overpower DT DNBR Overtemperature DT Excessive fuel heat generation rate (kWlft)

PRZR High Pressure RCS integrity PRZR Low Pressure DNBR B. Overpower DT Excessive fuel heat generation rate (kWlft)

Overtemperature DT DNBR PRZR High Pressure RCCA drive housing rupture PRZR Low Pressure Excessive RCS cooldown C Overpower DT Excessive fuel heat generation rate (kW/ft)

Overtemperature DT DNBR PRZR High Pressure RCS integrity PRZR Low Pressure DNBR D. Overpower DT DNBR Overtemperature DT Excessive fuel heat generation rate (kW/ft)

PRZR High Pressure RCCA drive housing rupture PRZR Low Pressure Excessive RCS cooldown K/A 012 Reactor Protection System K5.01 Knowledge of the operational implications of the following concepts as they apply to the RPS:

DNB.

KIA MATCH ANALYSIS The question requires the student to match 4 different reactor trip functions with their correct bases. Two of the trip bases are for DNBR protection matching the K/A topic.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. The bases for the OTDT and OPDT have been reversed, both of these choices are plausibe since the DT trips do provide this protection.

B. Incorrect. The PRZR high pressure trip bases is incorrectly tied to the PRNI high

+rate flux trip. The PRZR low pressure trip is tied to the bases for the FWI ESFAS 39

HL-15R RO NRC Exam be an expected precusor to an RCCA drive housing rupture and low PRZR would be expected as a result of an excessive RCS coolodown.

C. Correct. OPDT trip bases is excessive fuel heat generation rate (kW/ft). OTDT &

PRZR low pressure reactor trips bases are DNBR. PRZR high pressure provide RCS integrity protection.

D. Incorrect The bases for the OTDT and OPDT have been reversed, both of these choices are plausibe since the DT trips do provide this protection. The PRZR high pressure trip bases is incorrectly tied to the PRNI high +rate flux trip. The PRZR low pressure trip is tied to the bases for the FWI ESFAS function on low Tave. Both of these choices are plausible since high pressure would be an expected precusor to an RCCA drive housing rupture and low PRZR would be expected as a result of an excessive RCS coolodown.

REFERENCES VEGP LCO 3.3.1 bases pages 13, 16, 20, and 23 for:

- PRNI - High Positive Rate Trip

- OTDT Trip

- OPDT Trip

- PRZR High Pressure Trip

- PRZR Low Pressure Trip VEGP LCO 3.3.2 bases page 24 for FWI - Low RCS Tavg conincident with reactor trip Catawba 2008 NRC exam question 38 VEGP learning objectives:

LO-PP-281 03-04:

Discuss the bases for each reactor trip signal.

40

RTS Instrumentation B 3.3.1 BASES APPLICABLE b. Power Range Neutron Flux Low (continued)

SAFETY ANALYSES, LCO, and setpoint). This Function is automatically unblocked when APPLICABILITY three out of four power range channels are below the P-1O setpoint. Above the P-b setpoint, positive reactivity additions are mitigated by the Power Range Neutron Flux High trip Function.

In MODE 3, 4, 5, or 6, the Power Range Neutron Flux Low trip Function does not have to be OPERABLE because the reactor is shut down and the NIS power range detectors cannot detect neutron levels in this range. Other RTS trip Functions and administrative controls provide protection against positive reactivity additions or power excursions in MODE 3, 4, 5, or 6.

3. Power Range Neutron Flux High Positive Rate The Power Range Neutron Flux High Positive Rate trip uses the same channels as discussed for Function 2 above.

The Power Range Neutron Flux High Positive Rate trip Function ensures that protection is provided against rapid increases in neutron flux that are characteristic of an RCCA drive rod housing rupture and the accompanying ejection of the RCCA. This Function compliments the Power Range Neutron Flux High and Low Setpoint trip Functions to ensure that the criteria are met for a rod ejection from the power range.

The LCO requires all four of the Power Range Neutron Flux High Positive Rate channels to be OPERABLE.

In MODE 1 or 2, when there is a potential to add a large amount of positive reactivity from a rod ejection accident (REA), the Power Range Neutron Flux High Positive Rate trip must be OPERABLE. In MODE 3, 4, 5, or 6, the Power Range Neutron Flux High Positive Rate trip Function does not have to be (continued)

Vogtle Units 1 and 2 B 3.3.1-13 Revision No. 0

RTS Instrumentation B 3.3.1 BASES APPLICABLE 5. Source Range Neutron Flux (continued)

SAFETY ANALYSES, LCO, and subcritical, boron dilution (see LCO 3.3.8) and control APPLICABILITY rod ejection events. The Function also provides visual neutron flux indication in the control room.

In MODE 2 when below the P-6 setpoint during a reactor startup, the Source Range Neutron Flux trip must be OPERABLE. Above the P-6 setpoint, the Intermediate Range Neutron Flux trip and the Power Range Neutron Flux Low Setpoint trip will provide core protection for reactivity accidents.

Above the P-6 setpoint, the Source Range Neutron Flux trip is blocked.

In MODE 3, 4, or 5 with the reactor shut down, the Source Range Neutron Flux trip Function must also be OPERABLE. If the Rod Control System is capable of rod withdrawal, the Source Range Neutron Flux trip must be OPERABLE to provide core protection against a rod withdrawal accident. If the Rod Control System is not capable of rod withdrawal, the source range detectors are not required to trip the reactor. Source range detectors also function to monitor for high flux at shutdown. This function is addressed in Specification 3.3.8. Requirements for the source range detectors in MODE 6 are addressed in LCO 3.9.3.

6. Overtemperature AT The Overtemperature AT trip Function (TDI-041 1C, TDI-0421C, TDI-0431 C, TDI-0441 C, TDI-041 1A, TDI-0421A, TDI-0431A, TDI-0441A) is provided to ensure that the design limit DNBR is met.

This trip Function also limits the range over which the Overpower AT trip Function must provide protection. The inputs to the Overtemperature AT trip include pressure, coolant temperature, axial power distribution, and reactor power as indicated by loop AT assuming full reactor coolant flow. Protection from violating the DNBR limit is assured for those transients that are slow with respect to delays from the core to the measurement system. The Function monitors both variation in power and flow since a decrease in flow (continued)

Vogtle Units I and 2 B 3.3.1-16 Rev. 1-3/99

RTS Instrumentation B 3.3.1 BASES APPLICABLE 7. Overpower AT SAFETY ANALYSES, LCO, and The Overpower AT trip Function (TDI-041 1 B, TDI-0421 B, APPLICABILITY TDI-0431 B, TDI-0441 B, TDI-041 1A, TDI-0421A, TDI-0431A, (continued) TDI-0441A) ensures that protection is provided to ensure the integrity of the fuel (i.e., no fuel pellet melting and less than 1%

cladding strain) under all possible overpower conditions. This trip Function also limits the required range of the Overtemperature AT trip Function and provides a backup to the Power Range Neutron Flux High Setpoint trip. The Overpower AT trip Function ensures that the allowable heat generation rate (kW/ft) of the fuel is not exceeded. It uses the AT of each loop as a measure of reactor power with a setpoint that is automatically varied with the following parameters:

• reactor coolant average temperature the Trip Setpoint is varied to correct for changes in coolant density and specific heat capacity with changes in coolant temperature; and

• rate of change of reactor coolant average temperature including dynamic compensation for RTD response time delays.

The Overpower AT trip Function is calculated for each loop as per Note 2 of Table 3.3.1-1. Trip occurs if Overpower AT is indicated in two loops. Since the temperature signals are used for other control functions, the actuation logic must be able to withstand an input failure to the control system, which may then require the protection function actuation and a single failure in the remaining channels providing the protection function actuation. This results in a two-out-of-four trip logic. Section 7.2.2.3 of Reference I discusses control and protection system interactions for this function. Note that this Function also provides a signal to generate a turbine runback prior to reaching the Allowable Value. A turbine runback will reduce turbine power and reactor power. A reduction in power will normally alleviate the Overpower AT condition and may prevent a reactor trip.

(continued)

Vogtle Units I and 2 B 3.3.1-20 Revision No. 0

RTS Instrumentation B 3.3.1 BASES APPLICABLE 8. Pressurizer Pressure (continued)

SAFETY ANALYSES, LCO, and the control system, which may then require the protection APPLICABILITY function actuation, and a single failure in the other channels providing the protection function actuation. Section 7.2.2.3 of Reference I discusses control and protection system interactions for this function.

a. Pressurizer Pressure Low The Pressurizer Pressure Low trip Function ensures that protection is provided against violating the DNBR limit due to low pressure.

The [CO requires four channels of Pressurizer Pressure Low to be OPERABLE.

In MODE 1, when DNB is a major concern, the Pressurizer Pressure Low trip must be OPERABLE. This trip Function is automatically enabled on increasing power by the P-7 interlock (NIS power range P-b or turbine impulse pressure greater than approximately 10% of full power equivalent (P-13)). On decreasing power, this trip Function is automatically blocked below P-7. Below the P-7 setpoint, no conceivable power distributions can occur that would cause DNB concerns.

b. Pressurizer PressureHigh The Pressurizer Pressure High trip Function ensures that protection is provided against overpressurizing the RCS. This trip Function operates in conjunction with the pressurizer relief and safety valves to prevent RCS overpressure conditions.

The LCO requires four channels of the Pressurizer Pressure High to be OPERABLE.

The Pressurizer Pressure High LSSS is selected to be below the pressurizer safety valve actuation pressure and above the power operated relief valve (PORV) setting. This setting (continued)

Vogtle Units 1 and 2 B 3.3.1-23 Revision No. 0

ESFAS Instrumentation B 3.3.2 BASES APPLICABLE a. Turbine Trip and Feedwater Automatic Actuation Logic SAFETY ANALYSES, and Actuation Relays LCO, and APPLICABILITY Automatic Actuation Logic and Actuation Relays consist (continued) of the same features and operate in the same manner as described for ESFAS Function 1 .b. Under specific conditions, a single inoperable actuation relay does not require that the affected automatic actuation logic function be declared inoperable. Specific guidance is provided in this section under the heading Actuation Relays.

b. Feedwater Isolation Low RCS T Coincident with Reactor Trip Since Tavg is used as an indication of bulk RCS temperature, this Function meets redundancy requirements with one OPERABLE channel in each loop.

Thus, this function is specified as a total of four channels and not on a per loop basis. The channels are used in a two-out-of-four logic. The Low RCS Tavg signal is interlocked with P-4 to avert or reduce the continued cooldown of the RCS following a reactor trip. An excessive cooldown of the RCS following a reactor trip could cause an insertion of positive reactivity with a subsequent increase in generated power. The P-4 interlock is discussed in Function 8.a.

c. Turbine Trip and Feedwater Isolation Steam Generator Water Level High High (P-14, LOOP I LOOP2 LOOP3 LOOP4 LI-0517 [1-0527 LI-0537 Ll-0547

[1-0518 [1-0528 [1-0538 LI-0548 LI-0519 [1-0529 [1-0539 Ll-0549

[1-0551 [1-0552 Ll-0553 Ll-0554 NOTE: Steam Generator Water Level channels are required OPERABLE by the Post Accident Monitoring Technical Specification.

The setpoints for this Function on Table 3.3.2-1 are in % of narrow range instrument span.

(continued)

Vogtle Units I and 2 B 3.3.2-24 Revision No. 0

FOR REVIEW ONLY DO NOT DISTRIBUTE C

2008 SRO NRC Examination QUESTION 38 LQuestonBank #jKA_system KA_number r

KAdesc SYSOI2 Knowledge of the operational implications of the following concepts as the apply to the RPS: (CFR; 41.5 / 45.7)LIDNB Which one of the following selections correctly matches the reactor trip signals to their limiting accidentlprotection?

Reactor Trip Signal Limiting AccidentlProtection A. OPDT DNB OTDT Excessive fuel centerline temperature Pzr High Level NC system integrity Pzr Low Pressure DNB B. OPDT Excessive fuel centerline temperature OTDT DNB Pzr High Level DNB Pzr Low Pressure NC system integrity C. OPDT Excessive fuel centerline temperature OTDT DNB Pzr High Level NC system integrity Pzr Low Pressure DNB D. OPDT NC System integrity OTDT Excessive fuel centerline temperature Pzr High Level DNB Pzr Low Pressure DNB ATAW R Tuesday, November 18, 2008 Page 75 of 200

HL-1 5R RO NRC Exam

20. 013K5.02 OO1/2!1/ESFAS-LOGIC/C/A -2.9 / 3.3/NEW/HL-15R NRC/RO/DS / TNT A loss of 1 20V AC vital bus 1 BY1 B has occurred with the unit at 100% power. Which one of the following correctly decribes the impact on SSPS?

A. Only SSPS Train A channel II Input relays are de-energized.

The Train B Logic cabinet is de-energized.

The Train B Slave relays are inoperable.

B. Only SSPS Train A channel II input relays are de-energized.

The Train B Logic cabinet is de-energized.

The Train B Slave relays are operable.

C. SSPS Train A and Train B channel II Input relays are de-energized.

SSPS Train A and Train B Logic cabinets are energized.

The Train B Slave relays are operable.

D SSPS Train A and Train B channel II Input relays are de-energized.

SSPS Train A and Train B Logic cabinets are energized.

The Train B Slave relays are inoperable.

K/A 013 Engineered Safety Features Actuation System (ESFAS)

K5.02 Knowledge of the operational implications of the following concepts as they apply to the ESFAS:

Safety system logic and reliability.

KIA MATCH ANALYSIS The question presents a scenario where a loss of a vital AC bus occurs. The student must correctly determine the impact on both trains of ESFAS actuation circuitry including the logic circuits.

41

HL-15RRO NRC Exam ANSWER I DISTRACTOR ANALYSIS A. Incorrect. A loss of 1BY1B will de-energize all channel II input relays on both trains of SSPS. Both trains of logic cabinets will remain energized via redundant power supplies. The Train B only slave relays will all become inoperable. This choice is plausible due to the train B impacts listed.

B. Incorrect. A loss of 1BY1B will de-energize all channel II input relays on both trains of SSPS. Both trains of logic cabinets will remain energized via redundant power supplies. The Train B only slave relays will all become inoperable. This choice is plausible due to the train B impacts listed and that the train B master relays still have power available.

C. Incorrect. A loss of 1BY1B will de-energize all channel II input relayson both trains of SSPS. Both trains of logic cabinets will remain energized via redundant power supplies. The Train B only slave relays will all become inoperable. This choice is plausible due train B master relays still have power available.

D. Correct. A loss of 1BY1B will de-energize all channel II input relays on both trains of SSPS. Both trains of logic cabinets will remain energized via redundant power supplies. The Train B only slave relays will all become inoperable.

REFERENCES AOP 18032-1, Loss of 1 20V AC Instrument Power, page 44 LO-LP-60324, Loss of 1 20V AC Instrument Power, page 11 V-LO-TX-28101, SSPS text, pages 42, 44, and 45 V-LO-PP-28101, SSPS presentation, pages 23, 29, and 30 VEGP learning objectives:

LO-PP-28101-01:

State the sources of power to the SSPS cabinets.

LO-PP-281 01-02:

Determine how the loss of a power supply will affect SSPS.

42

Approved By Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant 18032-1 27 Date Approved Page Number LOSS OF 120V AC INSTRUMENT POWER 3/22/09 44 of 100 ATTACHMENT C Sheet 4 of 5 TABLE 2 PANEL 1BY1B LOAD LIST BREAKER LOAD 03 NIS CHANNEL II CONTROL POWER 04 SOLID STATE PROTECTION SYSTEM OUTPUT CABINET 2, TRAIN B 05 PROCESS RACK PROTECTION SET II 06 SPARE 07 SOLID STATE PROTECTION SYSTEM CABINET II, TRAIN A 08 NIS CHANNEL II INSTRUMENT POWER 09 DIESEL GENERATOR 1 B WHM PULSE AMPLIFIER 10 SOLID STATE PROTECTION SYSTEM CHANNEL II, TRAIN B 11 BOP PROTECTION PANEL CHANNEL II 12CQPP2 12 NFMS AMPLIFIER / NFMS OPTICAL ISOLATOR 13 HVAC PANEL 1BCQHVC2 14 SPARE 15 SAFEGUARD TEST CABINET, TRAIN B 16 SPARE 17 HVAC INSTRUMENT PANEL 1-1500-V7-002-CBB 18 SWITCHGEAR IBD1 TRANSDUCER POWER 19 PLASMA DISPLAY B 20 SPARE 21 SPARE 22 SPARE 23 MISC EQUIPMENT PANEL QPCP XMFR OF ZLB-16 24 DIST PANELS 1BD11, 1BD12 CONTROL POWER Printed November 10, 2009 at 14:54

LO-LP-60324-08-C III. LESSON OUTLINE: NOTES

3) CR1 C. Loss of Vital Instrument Panel I BYI B Symptoms - Same as on loss of IAYIA lust Train B

a. All Channel II trip status lights energized.

1) Probably the most significant and easiest method of determining the loss of 1BY1B.

b. Loss of N32, N36, N42 simultaneously.

c. 1 BY1 B Trouble Alarm

d. Inverter 1BD1I1 Trouble Alarm

2. Concerns and major effects of loss of 1 BY1 B Objective I Same as IAYIA except Train B equipment

a. SG 2 and 3 ARV will not operate from the control room or remote shutdown panel(NOTE).

b. May lose letdown and have max charging if LT-460 is selected for control.

c. Steam Generator levels and steam lows fail Normally all causing significant transient on all the selected to channel I steam generators. instruments

d. If unable to stabilize plant conditions may Objective 3 require plant shutdown or trip.

e. Plant will trip if below P-1O due to N36 trip signalwhen de-energized.

f. General Warning on Train B SSPS. Objective 4 Without power from 1 BY1 B there is no power to Train B SSPS slave relays.

Any actuation signal requiring operation of slave relays will be blocked.

If SI signal present, only Train A equipment will Auto align.

Train B equipment must be manually re-aligned in this case.

g. Plant will trip if another channel instrument bistable is tripped and/or channel is in test.

11

COMMITMENTS:

FE 86.008 FE 89.026 FE 89.021 OTHER:

10CFR5O, APPENDIX A, CRITERIA 20 THROUGH 25 DCPs 94.059 / 060 DCP 2001-032.doc Action Item # 2003202685 DCP105387131 and 2053871401 - Revise P-9 to 40%

SECTION C SOLID STATE PROTECTION SYSTEM 28.15 Layout and description of SSPS Two separate trains of SSPS exist on each unit for reactor protection. SSPS is divided in the following manner: input relay bay, logic bay, output relay bay, and the test panels. This section will discuss the purpose, automatic and testing operation portions of SSPS.

28.16 Input Relay Bays SSPS input relay bays acts as an isolation device between the various plant inputs and SSPS. It is divided into 4 compartments (one for each protection channel) to provide separation between each input channels.

There are three different types of inputs to the input relay bays: (1) NSSS and BOP protection, (2) Nuclear !nstrumentation System (NIS), and (3) Field Contacts. The input relays associated with NIS are supplied with 120 VAC from their respective NI channels. The field contacts are instruments that input directly into SSPS, such as Main Turbine Stop Valve position and RCP under frequency relays. The field contacts are powered directly from the input relay bay itself. Each SSPS input relay bay is supplied from its respective channel 120 VAC power source.

Input Relay Bay Source Channel I 1AY1A Channel II 1BY1B Channel III 1CY1A Channel IV 1DY1B V-LO-TX-281 01-08.1 42

The switches shown are typical for all block and reset switches found on the main control board.

Each hand switch will reset and block its associated train of SSPS only. This noted by the train designator in the bottom right hand corner.

The multiplexer portion of the logic bay monitors the input and output of the logic cards and transfers the status of each parameter to a de-multiplexer for the main control board or computer.

Identical information from both train A and B SSPS is transmitted over a common OR cable.

The multiplexer illuminates the appropriate Irip Status Light Box (TSLB) when its parameter bi stable trips. Also it indicates the status of all permissives and control interlocks on the Bypass Permissive Light Eanel (BPLP).

28.18 OUTPUT RELAYS BAYS There are two types of output relays; (1) Master relays, and (2) Slave relays. The master relays can be energized from either the Safeguard Driver Card from the Logic Bay or directly from the actuation switch located on the main control board. The master relays must be energized to actuate which require 48 VDC that comes from the logic bay. Each master relay can control up to four slave relays that are dedicated for a given actuation.

V-LO-TX-281 01-08.1 44

I5VDC I 48VDC I48VDC II5VDC LOGIC TESTER i rrit PERMISSIVES, MODE MEMORIES, & SELECTOR BLOCKS CHECK FOR PROPER COINCIDENCE MULTIPLEXER SWITCH SAFEGUARDS UNDERVOL TAGE DRIVER CARD DRIVER CARD S

5 T3 MASTER i .

48 VELAYS MODE i. Z P t 8 SELECTOR SWITCH 15 VDC I; .

i.

OPERATE TEST AVIA (BYIB)

MASTER RELAY SLAVE RELAY P-1I&P.12 ENABLE SIGNAL CS TEST PANEL SLAVE TRELA YS The slave relays are used to send actual start/stop signals to specific plant equipment when actuated. Slave relays require 120 VAC to actuate. Many are arranged with dual operating coils.

The latching coil is used to actuate, which occurs when its master relay energizes. This type relay will remain in the latch position even when de-energized. The second coil called the unlatch coil energizes when the operator reset the actuation from the control board. If both coils are actuated at the same time the relay will remain in the Latch (actuate) position.

28.19 TEST CIRCUITS SSPS is just like all other safety related systems, it must be tested to prove its operability. The SSPS test circuits allows testing from an individual slave relay to testing a reactor trip circuit while the unit remains online. This text will cover the different switches and circuits that allow such testing to be performed.

V-LO-TX-281 01-08.1 45

v- LO_-- 101 Outside view of the input bays for A train SSPS.

Input relay bays are divided into 4 isolated compartments (one for each channel) to provide separation between the input channels.

Plant inputs are applied to the input relay coils and the SSPS input comes from the input relay contacts to provide electrical isolation between the two (no electrical connections between coil and contacts).

Three types of inputs:

1) NSSS and BOP protection

2) NIS

3) Field Contacts Field contacts examples are:

1) Main Turbine Stop Valve position

2) RWST level

3) RCP under frequency Relay All field contacts are powered by its associated input bay.

Input Relay Bay power supplies:

Channel I 1/2AY1A Channel II 1/2BY1B Channel Ill 1/2CYIA Channel IV 1/2DYIB 23

\L- LO-&°l Close up picture of slave Relay K622, K623, and K624.

Demonstrate how to determine if a relay is reset or not.

Train A Slave Relays are powered by 1I2AY1A Train B Slave Relays are powered by 1I2BY1B With a loss of slave relay power all actuations from that train that are generated from SSPS will not occur.

29

-Lo LOGIC BAY POWER SUPPLIES A TRAIN AY1A/CY1A B TRAIN BY1B/DY1B 30 Two 15/48 VDC power supplies.

Only one needed to perform its logic operation.

Train A SSPS 15 VDC/48 VDC #1 Power supply is powered from 1/2AY1A 120 VAC 15 VDC/48 VDC #2 Power supply is powered from 1/2CY1A 120 VAC Train B SSPS 15 VDC/48 VDC #1 Power supply is powered from 1/2BY1B 120 VAC 15 VDC/48 VDC #2 Power supply is powered from 1/2DY1B 120 VAC 30

HL-15R RO NRC Exam

21. 015AK2.07 OO1/1/1/RCP MALF-RCP SEALS/C/A -2.9 / 2.9/M-LOIT/HL-15RNRC/RO/TNT/DS Given the following:

- The plant is at 17% power.

The following conditions exist on Reactor Coolant Pump # 2.

- # 1 seal DIP is 190 psig.

- # 1 seak leakoff flow is 5.2 gpm.

Which ONE of the following describes the required sequence I response to these conditions?

A. Shutdown the RCP, enter AOP-1 8005-C, Partial Loss of Flow.

B. Commence a unit shutdown per UOP-12004-C, shutdown the RCP within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

C Trip the Reactor, enter 19000-C, Reactor Trip or Safety Injection, shutdown the RCP.

D. Maintain current power, shutdown the RCP with engineering I management concurrence.

K/A 015 Reactor Coolant Pump (RCP) Malfunctions AK2.07 Knowledge of the interrelations between the Reactor Coolant Pump Malfunctions (Loss of RC Flow) and the following:

RCP Seals K/A MATCH ANALYSIS The question presents a plausible scenario where an RCP seal immediate trip criteria has been reached (seal DIP). Seal leakoff of 5.2 gpm is below the immediate trip criteria set point and the student must recognize the DIP is at an immediate trip set point. The student must determine the correct actions and procedure to use for the listed conditions.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Immediate RCP trip criteria present due to low seal d/p, >15% power requires trip reactor, enter E-0, then trip RCP. These actions are plausible to enter the Partial Loss of Flow AOP if the candidate does not recall power> 15% requires 43

HL15R RO NRC Exam B. Incorrect. Immediate RCP trip criteria present due to low seal dIp, >15% power requires trip reactor, enter E-0, the trip RCP. However, this choice is plausible if the candidate does not recognize seal dP as too low and thinks the RCP does not require an immediate trip. Commence a unit shutdown and stop RCP within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is a choice on the seal abnormality flow chart.

C. Correct. Immediate RCP trip criteria present due to low seal d/p, >15% power requires trip reactor, enter E-0, then trip the RCP.

D. Incorrect. Immediate RCP trip criteria present due to low seal d/p, >15% power requires trip reactor, enter E-0, then trip the RCP. Management and engineering concurrence very plausible if student does not recognize 200 psid as trip criteria and realizes 5.2 gpm below immediate trip threshold.

REFERENCES 015/01 7AK1 .02 used as base for modification from LOIT exam bank (included).

13003-C, Reactor Coolant Pump Operation, Precautions and Limitations, section 4.2 for Pump Operation With A Seal Abnormality, and Figure 1 RCP Seal Abnormalities Decision Tree.

VEGP learning objectives:

LO-PP-16401-03, Describe the Control Room indications for failure of an RCP seal.

44

1. 015!017AK1.02 001 Given the following:

- The plant is at 17% power. P eM+

- RCP #3 seal dip is 190 psig fr ev

- RCP #3 seal leakoff flow is 5.2 gpm i.-Y -ce

+

Which ONE of the following describes the required sequence/response to these conditions?

A Trip the Reactor, trip RCP #

B. Trip RCP # 3, commence a unit shutdown per UOP-1 2004-C, Power Operation C. Commence a unit shutdown per UOP-1 2004-C, stop RCP #3 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

D. Trip RCP # 3, enter AOP-1 8005-C, Partial Loss of Flow.

fl

/

oic 5we e i3o3 ea fly Corea, I(

,o1Ue SWil ge cece(

Laice, j Page: 1 10/5/2009

Approved By Procedure Number Rev C.S. Waidrup Vogtle Electric Generating Plant 13003-1 40 Date Approved Page Number 11/5/08 REACTOR COOLANT PUMP OPERATION 4 of 35 2.2.5 The following primary to secondary temperature limitations apply for RCP start:

• In order to prevent a low temperature RCS overpressure event, Technical Specification LCO 3.4.6, Note 2 requires that during MODE 4 operation below the COPS arming temperature as specified in the PTLR, the secondary side water temperature of each Steam Generator Temperature be less than 50°F above each of the RCS cold leg temperatures prior to the start of an RCP. Additionally, while in MODE 4 with no other RCPs running, this differential temperature limit is reduced to 25°F at an RCS temperature of 350°F and varies linearly to 50°F at an RCS temperature of 200°F as shown in figure 3. This verifies RHR system design pressures are not exceeded when the RHR suction reliefs are used for cold overpressure protection.

• To verify the above limits are not exceeded, an administrative limit, FSAR 5.2.2.10.2.c, is established such that an RCP shall NOT be started if its associated Steam Generator secondary water temperature is greater than 10°F above its RCS cold leg loop temperature.

• SGBD temperatures are preferred to SG skin temperatures when establishing conditions for starting a Reactor Coolant Pump. However, in Mode 5 SGBD is not required to be in service and SG skin temperatures can be used instead.

2.2.6 An RCP should NOT be started with the reactor critical. (Ref 18005-C) 2.2.7 The following conditions for the No. 1 Seal must be established prior to RCP t t.

• 200-psid minimum differential pressure across No. 1 Seal/ 6-EAJ,4

.1,

• A minimum VCT pressure of 18 psig. /9 f) 1L() j

• Minimum No. 1 Seal Leakoff as obtained from Figure 2.

JJ p

2.2.8 The following starting duty cycle for the RCP should be obs d:j,* N

• Only one RCP shall be started at any one time.

• Two successive starts are permitted, provided the motor is permitted to coast to a stop between starts.

• A third start may be made when the winding and core have cooled by running for a period of 20 minutes, or by standing idle for a period of 45 minutes.

Printed October 5, 2009 at 9:19

Approved By Procedure Number Rev C. S. Waidrup Vogtle Electric Generating Plant 13003-1 40 Date Approved Page Number 11/5/08 REACTOR COOLANT PUMP OPERATION 5 of 35 2.2.9 During RCS filling and venting, RCS pressure must be greater than 325 psig prior to starting an RCP to verify adequate seal DIP is maintained throughout RCS fill and vent. If necessary, the RCP should be stopped prior to seal D/P dropping less than 200 psid. If the seal D/P goes below 200 psid during pump operation or coast down, the RCP should be evaluated before restartingthe RCP. /

lt?PI tk EtE nIyppIi k FiI. t V?N+ N 2.2.10 An RCP shall be stopped if any of the following conditions exist.

o Motor bearing temperature exceeds 195°F.

• Motor stator winding temperature exceeds 311°F.

• Seal water inlet temperature exceeds 230° F

• Total loss of ACCW for a duration of 10 minutes.

• RCP shaft vibration of 20 mils or greater.

• RCP frame vibration of 5 mils or greater.

• Differential pressure across the number 1 seal of less, than 200 psid.

cjiI 4 tip C-pj, el L y,wvfr 2.2.11 If a loss of RCP seal cooling (Seal Injection and/or ACCW to Thermal barrier) occurs, resulting in RCP shutdown due to exceeding operating limits, then the unit should be cooled down to Mode 5 to facilitate recovery. Upon reaching Mode 5, ACCW to the Thermal barrier should be restored. Seal injection should then be returned to service. This sequence should prevent seal damage, RCP shaft bowing, ACCW System damage, etc. due to excessive thermal stresses.

Printed October 5, 2009 at 9:19

Approved By I Procedure Number Rev C. S. Waidrup I Vogtle Electric Generating Plant 13003-1 40 Date Approved I Page Number 11/5/08 I REACTOR COOLANT PUMP OPERATION 12 of 35 4.2 SYSTEM OPERATION 4.2.1 Pump Operation With A Seal Abnormality 4.2.1.1 IF the Plant Computer is available, trend the computer data points listed in Table 2.

4.2.1.2 IF the Plant Computer is NOT available, perform the following:

a. Monitor the QMCB indication listed in Table 2 at least hourly for the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

b. IF NO further seal degradation exists after 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, consult with the Shift Supervisor (SS) for less frequent monitoring.

4.2.1.3 Monitor the No. 1 seal for further degradation using Figure 1 and RCP Trip Criteria as follows:

a. Evaluate the monitored indications using Figure 1, RCP Seal Abnormalities Tree.

b. IF evaluation of the monitored indications using Figure 1 requires immediate pump shutdown, Go to Step 4.2.1.4.

r c. IF any of the following RCP Trip Criteria is exceeded, Go To Step 4.2.1.4 forimmediate RCP shutdown.

Motor bearing temperature Motor stator-winding temperature Seal water inlet temperature RCP shaft vibration RCP TRIP CRITERIA

>195°F

>311°F

>230° F

=20 mils RCP Frame vibration =5 mils

\- #1 seal Differential Pressure <200 psid

1. 1 seal leakoff flow (sum of #1 seal < minimum on Figure 2 with pump leakoff as indicated on the MCB and #2 bearing / seal inlet temperature seal leakoff read locally in containment) increasing Total loss of ACCW for a duration of 10 minutes Printed October 5, 2009 at 9:20

Approved By Procedure Number Rev C. S. Waidrup Vogtle Electric Generating Plant 13003-1 40 Date Approved Page Number 11/5/08 REACTOR COOLANT PUMP OPERATION 13 of 35

d. WHEN directed by Figure 1, stop the affected RCP within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> as follows:

(1) Establish 9 gpm or greater seal injection flow to the affected pump.

(2) Stop the affected RCP by continuing with Step 4.2.1.4.

4.2.1.4 WHEN directed by the SS, perform an RCP shutdown as follows:

a. Start the RCP Oil Lift Pump for affected RCP, if available.

b. IF Reactor Power is greater than 15% Rated Thermal Power:

(1) Trip the Reactor and initiate 19000-C, E-O Reactor Trip Or Safety Injection.

(2) WHEN the immediate operator actions of 19000-C p , p

c. IF Reactor Power is less than 15% Rated Thermal P initiate 18005-C, Partial Loss Of Flow.

d. Stop the RCP by placing the RCP Non-i E Control S in STOP and then placing the RCP 1 E Control Switch in STOP:

RCP Non-i E Control Switch 1 E Control Switch

• Loop 1 1-HS-0495B i-HS-0495A

• Loop 2 1-HS-0496B i-HS-0496A

• Loop 3 1-HS-0497B 1-HS-0497A

• Loop 4 i-HS-0498B 1-HS-0498A Printed October 5, 2009 at 9:20

Approved By Procedure Number Rev C. S. Waidrup Vogtle Electric Generating Plant , 13003-1 40 Date Approved Page Number 11/5/08 REACTOR COOLANT PUMP OPERATION 14 of 35 CAUTION If RCP #1 or #4 is stopped, the associated Spray Valve is placed in manual and closed to prevent spray short cycling.

e. IF RCP #1 OR #4 is stopped, verify its associated spray valve is placed in MANUAL AND closed.

• RCP 1: 1-PIC-0455C

• RCP 4: 1-PIC-0455B

f. WHEN the RCP comes to a complete stop (as indicated by reverse flow), close the RCP Seal Leakoff Isolation valve for the affected pump.

• RCP1: 1-HV-8141A

• RCP2: 1-HV-8141B

• RCP3: 1-HV-8141C

• RCP4: 1-HV-8141D

g. Secure the associated RCP Oil Lift Pump.

h. IF RCP shutdown was due to loss of RCP seal cooling, review Limitation 2.2.11 for recovery action.

Printed October 5, 2009 at 9:20

Procedure Number Rev Vogtle Electric Generating Plant 13003-1 40 Page Number REACTOR COOLANT PUMP OPERATION 32of35 FIGURE 1 - RCP SEAL ABNORMALITIES DECISION TREE

- No 0

HAN NORMAN 7

N. FLOWRTE Yes Y Yes Note 1: Abnormal Operating Range of Figure 2 Note 2: Non-operating Range of Figure 2 Note 3: ALBO8 A-04, B-04, C-04 or D-04 Printed

Approved By . Procedure Number Rev C. S. Waidrup Vogtle Electric Generating Plant rn 13003-1 40 Date Approved Page Number 11/5/08 REACTOR COOLANT PUMP OPERATION 33 of 35 FIGURE 2 NO. I SEAL NORMAL OPERATING RANGE 515 I *p1 L-ke;1c

,5 Jci

-J J Cu Cl) 0 z

NOTE I NOTE 2

- 0.8 0.2 0

0 200 500 1,000 1,500 2,000 I 2,500 2,250 No. I Seal Differential Pressure (PSI) NOTE 3 >

1. If the No. 1 seal leak rates are outside the normal (1 .0-5.0 gpm) but within the operating limits ((0.8-5.5 gpm), continue pump operation. VERIFY that seal injection flow exceeds No. 1 seal leak rate for the affected RCP. Closely monitor pump and seal parameters and contact Engineering for further instructions.

2. Minimum startup requirements are 0.2 gpm at 200 PSID differential across the No. 1 seal. For startups at differential pressures greater than 200 PSID, the minimum No. 1 seal leak rate requirements are defined in the NO. 1 SEAL NORMAL OPERATING RANGE (e.g., at 1000 psi differential pressure, do not start the RCP with less than 0.5 gpm).

3. No.1 Seal Differential Press = RCS WR Press VCT Press.

4. Per Westinghouse Technical Bulletin ESBU-TB-93-01-R1, total #1 seal leakoff is the sum of #1 seal leakoff and #2 seal leakoff. #1 seal leakoff is read directly at the MCB and #2 seal leakoff can be obtained from instrumentation in Containment.

Printed October 5, 2009 at 9:20

HL-15R RO NRC Exam

22. 01 6K1 .08 00 1/2/2/JNSTR-PZR PCS/C/A -3.4/3 .4/NEW/HL-1 5R NRC/RO/TNT/DS Given the following conditions:

- The unit is at 100% power.

- PRZR pressure control is selected to the 457 I 456 position.

The OATC determines that the controlling channel for Pressurizer Pressure control has failed.

If NO action is taken by the crew, which ONE of the following describes a CORRECT plant response for the failure given?

A. PT-457 fails high, PRZR pressure cycles between 2185 psig and 2200 psig.

B. PT-456 fails low, PRZR pressure cycles between 2345 psig and 2325 psig.

C. PT-456 fails low, reactor trips on the high PRZR Pressure setpoint.

D PT-457 fails high, reactor trips on the low PRZR Pressure setpoint.

K/A 016 Non-Nuclear Instrumentation System (NNIS)

KI .08 Knowledge of the physical connections and/or cause effect relationships between the NNIS and the following systems:

PZR PCS K/A MATCH ANALYSIS The question presents a plausible scenario where a Pressurizer Pressure channel has failed, the student must determine the correct plant response.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. This is the correct plant response for PT-456 failing high causing PORV 456 to open. As RCS pressure lowers, the block valves would cycle near 2185 psig.

Plausible the student could invert the channels and plant response.

B. Incorrect. This is the correct plant response for PT-455 failing low causing the PORV to cycle around the high pressure setpoint. However, PORV 456 would cycle from 2335 to 2315, this is the setpoint for the wrong PORV. Plausible the student could invert the channels and plant response.

C. Incorrect. PT-456 failing low would have no effect on the plant response. However, 45

HL-15R RO NRC Exam if the student inverts the controlling channels, it is plausible to think controlling channel could rise and cause a reactor trip.

D. Correct. PT-457 failed high would result in spray valves and a PORV failing open.

RCS pressure would lower to the 2185 setpoint at which point the block valves would attempt to stabilize the plant, however, sprays failed open would lower RCS pressure to the low pressure trip setpoint.

REFERENCES V-LO-PP-16303, Pressurizer Pressure Control System, slides 40, 42, 51, 71, 72, 153, 154, 156, and 157.

VEGP learning objectives:

LO-PP-16303-02, Describe how the Pressurizer Pressure Control System responds to the following failures.

a. Controlling (primary and secondary) channel fails low.

b. Controlling (primary and secondary) channel fails high.

46

PIC-455A Przr Press Master

/( CQ( Controller h1-f_ cl +S5 P&I J 2310 psig

>2345 psig (opens)

Heater Przr Hi

<2325 psig (closes) a. A Pressure N

a PORV 455 .

• a. *

• U U U Ua

• U (at+ aa Heaters a 2210 psig U

÷) U a

• .....aj

.DBackup .......

Przr Control a a aU a .. UU

• U

• a

• UU a a a

• a

• a

• a

• a a a U U U

U Lo Pressure U

a U & Htrs on a

U U

PlC 455B -

PlC 455C I a

a a LOOP 4 LOOP I a a

a a

SPRAY SPRAY I a I

U ONTROLLE ONTROLLEF a U a U a a a a

a U

a I a a

a I U U U U U a U

U 40 Pressurizer Controllinj U Pressure Control U DI_AQ ,ni ri

VARI HEAL CONTROL

d. I, (-

c 7, BACKUP HEATERS PRESS DEV PK-455A (MCB)

PZR P0kV (PCV4.5 SA)

(Ii) P.456 a

{IV) P458 PZR SPRAY CONTROL (PCV455B) pr  % CHANNEL.

SiECT ONly 3ce SWITCH

{MC3) pof- +S(e

)-[]).PZRPORV S.P)-I __i (PcV-456)

PK-455 P1<4SSE (M.CB) (PSDA)

Bi;1 iESS PZR SPRAY (PCY*4$SC)

ONE PR-455 (MCi) LOCAL CHANNEL MIA

_____,___/

REMOTE RECOR[)ER PtC455D PK-455C SELECT SWITCH (MC8) 5L (MCR (PWA)

Example:

455/45E PT-455 is the primarj: ch an.n.ei.. 57/43 PT456 is the C0i1toL5 secondary channel.

LO-PP- 16303-03 51

2310 psig 2335 psig PU 0- L55 245 Przr Hi PORV 456 Pressurizer Non Controlling Pressure Pressure Control LO-PP- 16303-03 71

c1 N

457 45:5/456 PT-455fails high, FOR V-455 opens (When Pressurizer pressure lowers to 2185 psig the FOR V interlock fCt(c 1LA function will close the PORV), p1i h i-*

Both Pressurizer proportional spray valves willfully open, Proportional heaters will turn off, Pressurizer High Pressure Alarm, Pressurizer High Pressure Alert, LO-PP-16303-03 153

57 455/456 PT456fails high:

FOR V-456 opens (When Pressurizer pressure lowers to 2185 psig the FOR V interlock 2I Wcc ciisc function will close the PORV), T Pressurizer High Pressure annunicator, Pressurizer High Pressure Alert annunicator, LO-PP-16303-03 154

L7 455/456 (vifr fr( W/j \\

PT-455fails low, 13 715 poj er4/ C(  ;

Proportional heaters will go to maximum output, All Pressurizer backup heaters will energize, Pressurizer Control Low Pressure and Heaters on annunicator, Pressure Low Pressure Channel Alert annunicator, Pressurizer Low Pressure SlAlert LO-PP-16303-03 156 annunicator,

455/456 i C

• PT456 fails low:

U Pressure Low Pressure Channel Alert s e annunicator, Pressurizer Low Pressure SI Alert ann unicator, LO-PP-16303-03 157

HL-15R RO NRC Exam

23. 022A1.O1 OO1!2!1!CNMT COOL-CNMT TEMP/MEM -3.6! 3.7!NEW!HL-15RNRC!RO!TNT/DS CNMT HI TEMP alarm has just annunciated.

The UO notes that CNMT air temperature is rising with CNMT coolers 1, 2, 5, and 6 in service on high speed.

The unit is shutdown with RCS temperature 375 F.

What is the correct action for the UO to take to stop the CNMT air temperature rise?

A Start CNMT coolers 3 and 4 simultaneously on high speed.

B. Start CNMT coolers 3 and 7 simultaneously on high speed.

C. Start CNMT coolers 3 and 4 sequentially on high speed.

D. Start CNMT coolers 3 and 7 sequentially on high speed.

K/A 022 Containment Cooling System (CCS)

A1.01 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CCS controls including:

Containment temperature.

K/A MATCH ANALYSIS The question presents a scenario where the student must determine the correct operation of the containment cooling system to lower containment temperature below its technical specification limit of 120 F as indicated by the high temperature alarm.

ANSWER I DISTRACTOR ANALYSIS A. Correct. Per ARP 17001-1 window E06, an additional pair of containment coolers should be started. SOP 13120-1 furhter states that each pair of coolers should be started simultaneously. This is done to prevent reverse flow through a cooler since each pair shares common discharge plenums and the back draft dampers are locked open.

B. Incorrect. This would be the correct action to take per the ARP 17001-1 however this pair of coolers is not allowed by the SOP 13120-1.

C. Incorrect. This is the correct pair of coolers to start. However they should be started 47

HL-15R RO NRC Exam associated pairs of containment coolers.

D. Incorrect. This would be the correct action to take per the ARP 17001-1 however this pair of coolers is not allowed by the SOP 13120-1. Additionally, the coolers should be started simultaneously rather than sequentially. Sequential starting is required between associated pairs of containment coolers.

REFERENCES ARP 17001-1, Window E06 SOP 13120-1, Containment Building Cooling System pages 8 and 9 LO-PP-29101, Containment HVAC Systems presentation, pages 17 and 18 VEGP learning objectives:

LO-PP-291 01-08:

Describe routine actions taken to adjust Containment pressure and temperature.

48

Approved By Procedure Number Rev S. A. Phillips Vogtle Electric Generating Plant 17001-1 30 Date Approved ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 01 ON PANEL 1A1 Page Number 3/27/08 ONMCB 45 of 46 WINDOW E06 ORIGIN SETPOINT CNMT 1-TSH-2563 120°F HI TEMP 1 -TSH-261 2 1-TSH-261 3 1.0 PROBABLE CAUSE Insufficient number of Containment Building Cooling Units operating.

2.0 AUTOMATIC ACTIONS NONE 3.0 INITIAL OPERATOR ACTIONS NONE 4.0 SUBSEQUENT OPERATOR ACTIONS

1. Start an additional pair of Containment Cooling Units or a Containment Auxiliary Cooling Unit per 13120-1, Containment Building Cooling Systems.

2. Verify Nuclear Service Cooling Water flow to coolers, and if necessary, dispatch an operator to inspect the Containment Heat Removal System.

3. Refer to Technical Specification LCO 3.6.5 and 3.6.6.

4. IF equipment failure is indicated, initiate maintenance as required.

5.0 COMPENSATORYOPERATOR ACTIONS NONE END OF SUB-PROCEDURE

REFERENCES:

1 X4DB21 2, CX5DT1 01-66, CX5DT1 01-71 Printed November 10, 2009 at 16:14

Approved By Procedure Number Rev S. E. Prewitt Vogtle Electric Generating Plant 13120-1 22.2 Date Approved Page Number 4/10/2006 CONTAINMENT BUILDING COOLING SYSTEM 8 of 50 INITIALS 4.2 CONTAINMENT HEAT REMOVAL SYSTEM STARTUP NOTES

• Normal operation is four fans running in high speed. Running all eight fans in high speed or operation of fans in low speed may be performed at SS direction.

• Containment Coolers should be operated in one of the combinations specified.

• After start of the first two Coolers in a specified combination, the operator should wait 20 seconds before starting the second pair in that combination.

This will limit voltage drop on the respective switchgear.

4.2.1 To start fans in High speed perform Step 4.2.3.

4.2.2 To start fans in Low speed perform Step 4.2.4.

4.2.3 Select one of the following four combinations and start the Containment Coolers one pair at a time, in high speed, by simultaneously placing both handswitches to the start position:

a. Combination 1 High speed operation Train A (1) Fan 1, 1-HS-12582D(B24)QHVC Fan 2, 1-HS-2582D (B25) QHVC (2) Fan 5, 1-HS-12584D (D24) QHVC Fan 6, 1-HS-2584D (D25) QHVC Printed November 10, 2009 at 16:24

Approved By Procedure Number Rev S. E. Prewitt Vogtle Electric Generating Plant ,

13120-1 22.2 Date Approved Page Number 4/10/2006 CONTAINMENT BUILDING COOLING SYSTEM 9 of 50 INITIALS

b. Combination 2 High speed operation Train B (1) Fan 3, 1-HS-12583D (B26) QHVC Fan 4, 1-HS-2583D (B27) QHVC (2) Fan 7, 1-HS-12585D (D26) QHVC Fan 8, 1-HS-2585D (D27) QHVC

c. Combination 3 High speed operation Train A (1) Fan 1, 1-HS-12582D(B24)QHVC Fan 2, 1-HS-2582D (B25) QHVC Train B (2) Fan 7, 1-HS-12585D (D26) QHVC Fan 8, 1-HS-2585D (D27) QHVC

d. Combination 4 High speed operation Train A (1) Fan 5, 1-HS-12584D (D24) QHVC Fan 6, 1-HS-2584D (D25) QHVC Train B (2) Fan 3, 1-HS-12583D (B26) QHVC Fan 4, 1-HS-2583D (B27) QHVC Printed November 10, 2009 at 16:24

Containment Cooling Units

• The fans are normally operated in pairs from the QHVC during normal operation with backup capability from their associated shutdown panels.

Pairs are selected on the basis that their discharges are tied together to a common duct and to ensure balance flows to all cooled areas.

17

Containment Cooling Units

• Motorized discharge dampers are de energized and locked open to preclude*

loss of fan capability during an emergency.

v 18

HL-15R RO NRC Exam

24. 022A2.03 OO1/2/1/CNMT COOL-HI SPEED/C/A 2.6 / 3.O/NEW/HL-15R NRC/RO/DS/TNT Initial conditions:

A steamline break inside containment has occurred EOP 19010-C, Loss of Reactor or Secondary Coolant is being implemented The following sequence of events occurs:

The SI signal is reset CNMT pressure is 8.6 psig and lowering A loss of both RATs occurs Both EDGs start and re-energize their respective busses The correct action to take is to...

A. verify the sequencers start all CNMT coolers on low speed.

Then shift the CNMT coolers to high speed after the sequencing is completed.

B. verify the sequencers start all CNMT coolers on low speed.

Then restart the SI and RHR pumps as needed to maintain RCS inventory.

C verify the sequencers start all CNMT coolers on high speed.

Then shift the CNMT coolers to low speed to prevent a CNMT cooler fan overcurrent condition.

D. verify the sequencers start all CNMT coolers on high speed.

Then shift the CNMT coolers to low speed to prevent an EDG overcurrent condition.

K/A 022 Containment Cooling System (CCS)

A2.03 Ability to (a) predict the impacts of the following malfunctions or operations on the CCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Fan motor thermal overloadlhigh-speed operation.

49

HL-15R RO NRC Exam.

The question presents a scenario where a UV condition following a SI actuation and reset occurs. The student must determine the correct action to take to prevent an over current condition on CNMT cooler motors for the conditions given.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. The SI signal was reset prior to the loss of the RATs. This will result in the sequencers running the UV sequence instead of the SI sequence. Since CNMT pressure is > 3.8 psig (adverse) the CNMT coolers should be shifted to low speed to prevent exceeding the motor current limits on the coolers due to the high density fluid conditions in CNMT. This choice is plausible since the CNMT coolers would start in low speed if the SI signal were not reset.

B. Incorrect. The SI signal was reset prior to the loss of the RATs. This will result in the sequencers running the UV sequence instead of the SI sequence. Since CNMT pressure is > 3.8 psig (adverse) the CNMT coolers should be shifted to low speed to prevent exceeding the motor current limits on the coolers due to the high density fluid conditions in CNMT. This choice is plausible since the CNMT coolers would start in low speed if the SI signal were not reset. Additionally the SI and RHR pumps would have to be manually started to help maintain RCS inventory for these conditions.

C. Correct. The SI signal was reset prior to the loss of the RATs. This will result in the sequencers running the UV sequence instead of the SI sequence. Since CNMT pressure is > 3.8 psig (adverse) the CNMT coolers should be shifted to low speed to prevent exceeding the motor current limits on the coolers due to the high density fluid conditions in CNMT.

D. Incorrect. The SI signal was reset prior to the loss of the RATs. This will result in the sequencers running the UV sequence instead of the SI sequence. Since CNMT pressure is > 3.8 psig (adverse) the CNMT coolers should be shifted to low speed to prevent exceeding the motor current limits on the coolers due to the high density fluid conditions in CNMT. This choice is plausible since loading of the EDGs is a concern and the reason for running the SI or UV sequences.

REFERENCES EOP 19010-C, E-l Loss of Reactor or Secondary Coolant page 11 LO-PP-29101, CNMT HVAC Systems presentation, pages 20, 22, and 23 VEGP learning obiectives:

LO-PP-291 01-04:

State how the Containment atmosphere density changes following a LOCA and why.

50

HL-15R RO NRC Exam State why two speeds are provided for the Containment Coolers and when each speed is used.

LO-PP-291 01-14:

State all auto start signals for the Containment Cooling including set points and coincidence where applicable.

51

Approved By Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant 19010-C 32 Date Approved E-1 LOSS OF REACTOR OR SECONDARY Page Number 3/24/09 COOLANT 11 of 25 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED CAUTIONS If offsite power is lost after SI reset, action is required to restart the following ESF equipment if plant conditions require their operation:

• RHR Pumps

• SI Pumps

• Post-LOCA Cavity Purge Units

• Containment Coolers in low speed (Started in high speed on a UV signal).

• ESF Chilled Water Pumps (If CR1 is reset).

13. Check if RHR Pumps should be stopped:

a. RHR Pumps ANY RUNNING

- a. Go to Step 15.

WITH SUCTION ALIGNED TO RWST.

b. RCS pressure:

1) Greater than 300 psig. 1) Go to Step 16.

2) Stable or rising. 2) Go to Step 15.

c. Reset SI.

d. Stop RHR Pumps.

• 14 IF RCS pressure lowers in an uncontrolled manner to less than 300 psig, THEN restart RHR Pumps.

Printed November 10, 2009 at 15:55

Containment Cooling Units Normal Operations

• The Containment Cooling Units help maintain containment temperature with four of the eight units running in fast speed (97,000 cfm).

• The eight fans are located at the 240 feet elevation, four on each side of the containment behind the SG enclosures.

Vcop/o/

Containment Cooling Units Emerqençy Operations Safety Injection All eight fans start in slow speed

• (43,500 cfm)

• Start signal from SI Sequencer

• Slow speed is used rather than fast speed due to the adverse containment conditions. The higher density pressurized air could cause over current conditions (and possible damage) if the fast speed wind ings are energized.

• Four fans (single Train) in slow speed provide adequate heat removal r /

L-

Containment Cooling Units Emgençy Operations Loss of Offsite Power

• Fans will be shed from the affected buss(es) and restarted in fast speed (97,000 CFM)

• All fans get at start signal at 30.5 secs

• Fans 1&2 (Train A) and 3&4 (Train B)

Time Delayed 20 secs Prevents overloading DG & Bus voltage swings oP- i/

HL-15R RO NRC Exam

25. 022AK1.01 001/1/1/LOSS CHARGE-T SHOCK/C/A -2.8 / 2.5/NEW/HL-15RNRC/RO/TNT/DS Given the following conditions:

- A loss of all AC power occurs in Mode 1.

- The plant is currently in Mode 3.

- HV-8103A, B, C, D Seal Injection Isolation Valves are CLOSED.

Which ONE of the following is CORRECT regarding:

1) The Mode in which seal injection will be re-established and

2) the reason for closing the seal injection isolation valves?

A. Maintain Mode 3 To prevent steam binding the charging pumps via back leakage in the seal lines.

B. Maintain Mode 3 To prevent seal damage and RCP shaft bowing due to excessive thermal stresses.

C. Cooldown to Mode 5 To prevent steam binding the charging pumps via back leakage in the seal lines.

D Cooldown to Mode 5 To prevent seal damage and RCP shaft bowing due to excessive thermal stresses.

K/A 022 Loss of Reactor Coolant Makeup AKI.01 Knowledge of the operational implications of the following concepts as they apply to the Loss of Reactor Coolant Makeup:

Consequences of thermal shock to RCP seals.

K/A MATCH ANALYSIS 52

HL-15R RO NRC Exam which is our most likely cause of Loss of Charging and Seal Injection. The question asks the student to recall the strategy for re-establishing seal cooling and the possible consequences if this is not done properly.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Per WOG background documents for Vogtle, a plant cooldown will be performed to reduce RCP seal temperatures. SOP-i 3003-1/2 Limitation 2.2.11 also states the plant should be cooled down to Mode 5 if a loss of RCP seal cooling has occurred. This is to prevent seal damage and RCP shaft bowing due to excessive thermal stresses. The charging pump binding via seal lines is plausible since it sounds like the steam binding of the ACCW system which is mentioned in our procedures and background documents, plausible the student could confuse this.

B. Incorrect. Per WOG background documents for Vogtle, a plant cooldown will be performed to reduce RCP seal temperatures. SOP-13003-1/2 Limitation 2.2.11 also states the plant should be cooled down to Mode 5 if a loss of RCP seal cooling has occurred. This is to prevent seal damage and RCP shaft bowing due to excessive thermal stresses.

C. Incorrect. Per WOG background documents for Vogtle, a plant cooldown will be performed to reduce RCP seal temperatures. SOP-13003-1/2 Limitation 2.2.11 also states the plant should be cooled down to Mode 5 if a loss of RCP seal cooling has occurred. This is to prevent seal damage and RCP shaft bowing due to excessive thermal stresses. The charging pump binding via seal lines is plausible since it sounds like the steam binding of the ACCW system which is mentioned in our procedures and background documents, plausible the student could confuse this.

D. Correct. Plant should be cooled down to Mode 5 to prevent possible seal damage and bowing of RCP shafts.

REFERENCES SOP-13003-i, Reactor Coolant Pump Operation, Limitation 2.2.11.

LO-HO-37031-001, Loss of All AC Power page 1-3.

VEGP learning objectives:

LO-LP-37031 -02, State why the RCP is a primary concern during a Loss of All AC Power condition LO-PP-16401-005, Given a loss of RCP seal injection, describe the indictions that would be monitored and impact to continued operation of the RCP.

53

Approved By Procedure Number Rev C. S. Waidrup Vogtle Electric Generating Plant 13003-1 40 Date Approved Page Number 11/5/08 REACTOR COOLANT PUMP OPERATION 5 of 35 2.2.9 During RCS filling and venting, RCS pressure must be greater than 325 psig prior to starting an RCP to verify adequate seal DIP is maintained throughout RCS fill and vent. If necessary, the RCP should be stopped prior to seal DIP dropping less than 200 psid. If the seal DIP goes below 200 psid during pump operation or coast down, the RCP should be evaluated before restarting the RCP.

2.2.10 An RCP shall be stopped if any of the following conditions exist.

• Motor bearing temperature exceeds 195°F.

• Motor stator winding temperature exceeds 311 °F.

• Seal water inlet temperature exceeds 230°F

• Total loss of ACCW for a duration of 10 minutes.

• RCP shaft vibration of 20 mils or greater.

o RCP frame vibration of 5 mils or greater.

• Differential pressure across the number 1 seal of less than 200 psid.

2.2.11 If a loss of RCP seal cooling (Seal Injection and/or ACCW to Thermal barrier) occurs, resulting in RCP shutdown due to exceeding operating limits, then the unit should be cooled down to Mode 5 to facilitate recovery. Upon reaching Mode 5, ACCW to the Thermal barrier should be restored. Seal injection should then be returned to service. This sequence should prevent seal damage, RCP shaft bowing, ACCW System damage, etc. due to excessive thermal stresses.

Printed October 5, 2009 at 13:11

LO-HO-37031-001 C: Loss Of All AC Power To evaluate the most severe consequences of a loss of all AC. power to the RCP seal system, a conservative maximum RCP leakage rate has been estimated to be 300 gpm. This rate was estimated by assuming that total RCS pressure of 2235 psig exists across the RCP thermal barrier labyrinth seals with the controlled leakage seals totally ineffective in controlling leakage flow.

The high RCS temperatures and pressures characteristic of a plant no-load condition can lead to eventual RCP seal degradation and increased RCS inventory loss. This seal degradation can be mitigated by reducing the RCS pressure and temperature consistent with other plant constraints. Reducing RCS pressure reduces leakage flow through the RCP seals, thereby reducing RCS inventory loss for a given seal condition.

Reducing RCS temperature reduces the thermal degradation of materials and thermal expansion effects that tend to degrade the seal system sealing capability and sealing life. Consequently, any actions to reduce RCS pressure and temperature during a loss of all A.C. power event will reduce RCS inventory loss and will increase the time to core uncovering.

RCP Seal System Cooling NON-Restoration The effect of restoring ACCW to the thermal barrier heat exchanger following an extended loss of seal cooling event cannot be fully understood without performing detailed analyses. Therefore, the only conclusions that can be made is that restoring seal cooling following an extended loss of all ac power event could jeopardize the integrity of the ACCW system. The plant that the generic Westinghouse Emergency Response Guidelines are based on provides thermal barrier cooling by the CCW system. This plant does not have a separate ACCW system. Since CCW is a safety-related system, and while restoring seal cooling is good practice, it is not necessary to ensure the health and safety of the public. Therefore, the Westinghouse Owners Group Operations Subcommittee has taken the position that the integrity of the CCW system should not be jeopardized to restore seal cooling. In that light, we will not restore ACCW cooling the the thermal barrier heat exchangers here at Vogtle.

Without thermal barrier cooling, two options are available to resolve the seal cooling issue; reestablish seal injection regardless of the potential damage that will occur to the RCP seal package and shaft, or find an alternate method of cooling the seal package. The Operations Subcommittee has decided that an alternate method of cooling the seal package should be employed to minimize damage to the RCP. This alternate method will consist of reducing primary system temperature, which will reduce the temperature of the water flowing through the pump seals. Reducing the seal temperature via a controlled RCS cooldown has several advantages First the seal package should cool evenly minimizing thermal gradients placed on the seal package and minimizing the potential for RCP shaft warping. Second, as seal temperatures are reduced, the seal leakage is likely to decrease due to the associated decrease in system pressure and differential pressure across the seals.

Finally, using this alternate seal cooling approach will not jeopardize any plant safety systems.

Note that relying on an RCS cooldown to reduce seal temperatures will result in continued seal leakage until seal injection can be reestablished. This leakage will not result in core uncovery and should be within the capacity of the normal charging system. However, the leakage will be diverted to the PRT and will eventually cause the rupture disc to fail. This will result in the spilling of reactor coolant fluid on to the containment floor. However, this concern would also exist when restoring seal cooling using ACCW to the thermal barrier heat exchanger since the cooldown rate of the seals is limited to 1°F per minute regardless of the method used to cool the seals (i.e.,

thermal barrier cooling would not reduce leakage any faster than an RCS cooldown). Also, the thermal barrier heat exchanger may not even have the capacity to cool the seals to the necessary temperature unless done concurrently with an RCS cooldown. Finally, the consequences of rupturing the PRT are much less severe than the consequences of failing the entire ACCW system. Therefore, cooling the RCP seals via an RCS cooldown instead of using the thermal barrier heat exchanger should not have a significant impact on plant safety, and may actually improve plant safety by maintaining the integrity of the ACCW system.

Based on the above arguments, during the recovery from an extended loss of all ac power event, no attempt will be made to restore seal cooling via the thermal barrier heat exchanger. Instead, seal cooling will be restored via a controlled RCS cooldown. The limits on restoring seal injection contained in the RCP vendors manual will still be observed.

1-3

HL-15R RO NRC Exam

26. 025AA1.19 001/1/1/LOSS RHR-BLOCK VALVE/C/A -2.6! 2.4/NEW/HL-15RNRC/RO/TNT/DS Given the following sequence:

- The plant is in Mode 6 at midloop.

- RHR pump A trips due to a loss of RCS inventory.

- The RCS has been refilled and RHR pump B is ready to be started.

Complete the following two sentences:

1) To start the pump, the RHR Hx Bypass Valve controller (FIC-0619) should be...

2) To ensure compliance with Tech Spec flow requirements, per procedure the potentiometer setting for the RHR Hx Bypass Valve controller (FIC-061 9) should be set at...

Given: Formula for Potentiometer setting in gpm is (desired flow I 5000)2 X 10.

RHR Hx Bypass valve Potentiometer setting A. in automatic. 3.6 B. in automatic. 4.1 C. in manual and closed. 3.6 D in manual and closed. 4.1 K/A 025 Lass of Residual Heat Removal System (RHRS)

AAI.19 Ability to operate and / or monitor the following as they apply to the Loss of Residual Heat Removal System:

Block orifice bypass valve controller and indicators.

K/A MATCH ANALYSIS The question presents a plausible scenario where an RHR pump is to be started while at midloop. The student must know the required position of the RHR Hx Bypass Valve (FIC-0619) during pump start, and the potentiometer setting to ensure 3000 gpm flow to comply with Technical Specifications.

NOTE: A reference with the RHR Hx Bypass valve calculation formula is to be provided to the students. (Page #33 of SOP-13011, Residual Heat Removal) is 54

HL-15R RO NRC Exam where this formula is given.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Pump should be started with the RHR Hx Bypass Valve (FIC-0619) closed. Plausible the student could choose automatic in order to control flow at the Tech Spec requirement. However, AOP-1 8019, Loss of RHR and SOP-i 3011, RHR both specify pump start with the Hx outlet and Bypass valves closed.

The potentiometer setting is calculated by (desired flow I 5000)2 X 10. The # 3.6 is plausible due to (3000 / 5000)2 X 10 = (0.6)2 X 10 = (0.36) X 10 = 3.6.

However, to ensure 3000 gpm flow under all conditions, Tech Spec Rounds and procedures call for 3200 gpm flow, the setting for this is 4.1.

The potentiometer setting is calculated by (desired flow /5000)2 X 10. The #4.1 is correct due to (3200 / 5000)2 X 10 = (0.64)2 X 10 = (0.41) X 10 = 4.1.

The student is required to recall whether flow should be set at 3200 vs 3000 gpm.

B. Incorrect. Pump should be started with the RHR Hx Bypass Valve (FIC-06i9) closed. Plausible the student could choose automatic in order to control flow at the Tech Spec requirement. However, AOP-18019, Loss of RHR and SOP-i 3011, RHR both specify pump start with the Hx outlet and Bypass valves closed.

The potentiometer setting is calculated by (desired flow / 5000)2 X 10. The # 3.6 is plausible due to (3000 / 5000)2 X 10 = (0.6)2 X 10 = (0.36) X 10 = 3.6.

However, to ensure 3000 gpm flow under all conditions, Tech Spec Rounds and procedures call for 3200 gpm flow, the setting for this is 4.1.

The potentiometer setting is calculated by (desired flow / 5000)2 X 10. The #4.1 is correct due to (3200 / 5000)2 X 10 = (0.64)2 X 10 = (0.41) X 10 = 4.1.

The student is required to recall whether flow should be set at 3200 vs 3000 gpm.

C. Incorrect. Pump should be started with the RHR Hx Bypass Valve (FIC-06i9) closed. Plausible the student could choose automatic in order to control flow at the Tech Spec requirement. However, AOP-1 8019, Loss of RHR and SOP-i 3011, RHR both specify pump start with the Hx outlet and Bypass valves closed.

The potentiometer setting is calculated by (desired flow / 5000)2 X 10. The # 3.6 is plausible due to (3000 I 5000)2 X 10 = (0.6)2 X 10 = (0.36) X 10 = 3.6.

However, to ensure 3000 gpm flow under all conditions, Tech Spec Rounds and procedures call for 3200 gpm flow, the setting for this is 4.1.

55

HL-15R RO NRC Exam The potentiometer setting is calculated by (desired flow I 5000)2 X 10. The #4.1 is correct due to (3200/5000)2 X 10 = (0.64)2 X 10 = (0.41) X 10 = 4.1.

The student is required to recall whether flow should be set at 3200 vs 3000 gpm.

D. Correct. Bypass valve should be in manual and closed. Pot setting for Tech Specs is set at 4.1.

REFERENCES SOP-i 3011, Residual Heat Removal System section 4.4 for Placing Tm-B RHR in Service For RCS Cooldown From Standby Readiness.

AOP-1 8019, Loss of RHR, section B for a loss of RHR capability or imminent loss of RHR due to RCS leakage while in Mode 5 or 6 with RCS level below the PRZR indication range or with SG nozzle dams installed. Steps B16 through B19.

OSP-1 4000, Operations Shift and Daily Surveillance Logs, Data Sheet 3 - Modes 5 & 6 sheet 2 of 6 (page 23).

VEGP learning objectives:

LO-PP-12101-09, Describe how the RHR heat exchanger bypass valve automatically controls flow to the RCS.

LO-PP-12101-16, Briefly describe the operator actions required for placing RHR in service for shutdown cooling.

56

Approved By Procedure Number Rev C. R. Dedrickson Vogtle Electric Generating Plant 18019-C 26.2 Date Approved Page Number 4-1 8-200 7 LOSS OF RESIDUAL HEAT REMOVAL 26 of 69 B. LOSS OF RHR MODE 5CR 6 BELOW PRZR IR OR SG NOZZLE DAMS INSTALLED ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

b. Use either of the following:

ATTACHMENT C Section A

- VALVE LINEUP FOR RHR PUMP COLD LEG I NJ ECT ION.

-OR-ATTACHMENT C Section B

- VALVE LINEUP FOR RHR PUMP HOT LEG INJECTION.

c. Check RV Head - REMOVED. c. Go to Step B16.

d. Use the Refueling Water Purification Pump per ATTACHMENT B.

B16. Identify and isolate any RCS leakage.

NOTES

• The time to boiling in the RCS should be taken into consideration when determining how much time should be spent venting the RHR system prior to taking additional actions for alternate cooling sources.

• If adequate time to completely vent the RHR system is not available, air can be swept out of the RHR lines by filling the RCS to 188 feet 3 inches and running an RHR Pump at a flowrate greater than 3000 gpm (3200 gpm indicated.)

B17. Vent any RHR Pump that experienced cavitation:

_a. Maintain RCS level while venting RHR system.

Step 17 continued on next page Printed October 6, 2009 at 15:09

Approved By Procedure Number Rev C. R. Dedrickson Vogtle Electric Generating Plant 18019-C 26.2 Date Approved Page Number 4-18-2007 LOSS OF RESIDUAL HEAT REMOVAL 27 of 69 B. LOSS OF RHR - MODE 5 CR6 BELOW PRZR IR OR SG NOZZLE DAMS INSTALLED ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

b. Vent RHR Pump(s) at high point vent until water is discharged:

UNIT 1:

i-HV-1 0465 RHR SUCT VENT LINE TRNA (AB-B08) i-HV-10466 RHR SUCT VENTLINE TRNB (FHB-Bi 3)

UNIT 2:

2-HV-i 0465 RHR SUCT VENT LINE TRNA (AB-Bi 31) 2-HV-i 0466 RHR SUCT VENTLINE TRNB (FHB-B03)

c. Vent RHR Pump(s) using casing vents until water is discharged:

UNIT 1:

Train A -

i-i 205-U4-235 (AB-D48)

Train B -

i-i 205-U4-236 (AB-D49)

UNIT 2:

Train A -

2-1 205-U4-235 (AB-D22)

Train B -

2-1 205-U4-236 (AB-D2i)

Printed October 6. 2009 at 15:09

Approved By Procedure Number Rev C. R. Dedrickson Vogtle Electric Generating Plant 18019-C 26.2 DateApproved PageNumber 4-18-2007 LOSS OF RESIDUAL HEAT REMOVAL 28 of 69 B. LOSS OF RHR - MODE 5 OR 6 BELOW PRZR IR OR SG NOZZLE DAMS INSTALLED ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED B18. Establish conditions to start RHR Pump:

_a. RCS level GREATER THAN

- _a. Consult TSC if applicable 188 FEET. and return to Step B6.

b. RHR Pump AVAILABLE.

- b. Consult TSC if applicable and return to Step B6.

c. Check RHR system valves - IN _c. Align valves as required.

PROPER ALIGNMENT:

TRAIN A

• HV-8701A OPEN-

• HV-8701B OPEN-

• HV-0606 - CLOSED

. FV-0618 - CLOSED

• HV-0128 - CLOSED

. HV-1 0465 CLOSED

• FV-0610 -OPEN

. HV-8809A OPEN HV-8716A-CLOSED TRAIN B

• HV-8702A OPEN

. HV-8702B OPEN

• HV-0607 - CLOSED

• FV-0619 - CLOSED

• HV-0128 - CLOSED

. HV-1 0466 CLOSED

• FV-0611 -OPEN e HV-8809B OPEN

• HV-8716B CLOSED

_d. Check CCW cooling to RHR d. Restore CCW cooling by system IN SERVICE,

- initiating 18020-C, LOSS OF COMPONENT COOLING WATER.

Printed October 6, 2009 at 15:09

Approved By . Procedure Number Rev C. R. Dedrickson Vogtle Electric Generating Plant 18019-C 26.2 Date Approved Page Number 4-18-2007 LOSS OF RESIDUAL HEAT REMOVAL 29 of 69 B. LOSS OF RHR MODES OR 6 BELOW PRZR IR OR SG NOZZLE DAMS INSTALLED ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED CAUTION Starting an RHR Pump may result in an RCS level reduction due to shrink or void collapse.

Bi 9. Restore RHR flow:

_a. Start one RHR pump.

_b. Control charging flow to maintain RCS level above 188 feet.

_c. Slowly raise RHR bypass flow to 3000 gpm.

_d. Check RHR Pump NOT - d. Perform the following:

CAVITATING.

1) Reduce flow to stop cavitation.

_2) IF flow must be reduced to less than 1500 gpm to stop cavitation, THEN stop RHR Pump and return to Step B6.

e. Check RHR flow RESTORED.

- e. Consult TSC if applicable and return to Step B6.

f. Establish desired RCS cooldown rate.

Printed October 6, 2009 at 15:09

Approved By Procedure Number Rev A. S. Parton Vogtle Electric Generating Plant 13011-1 67.1 Date Approved Page Number 4/24/09 RESIDUAL HEAT REMOVAL SYSTEM 29 of 115 INITIALS 4.4 PLACING TRN-B RHR IN SERVICE FOR RCS COOLDOWN FROM STANDBY READINESS 4.4.1 Notify HP that this RHR system change could affect area radiation levels so that surveys can be taken and personnel made aware of the changed condition.

4.4.2 Restore power to RHR PMP-B SUCTION FROM HOT LEG LOOP 4 Inlet Isolations and air to RHR System Flow Control Valves as follows: (IV REQUIRED)

a. IF shutdown, place Inverter 1 DD1 16 in service per 13405-1, 125V DC 1E Electrical Distribution System.

b. Install the annunciator card associated with ALB34 E07 and check ALB34 E07 illuminates.

c. At 1 DD1 16N unlock and close the disconnect for 1 -HV-8702A.

d. Check ALB34-E07 extinguishes.

e. Close the K2 link for breaker 1 BBE-1 3.

f. Unlock and close RHR PMP B SUCTION FROM HOT LEG LOOP 4, 1-HV-8702B Supply Breakers 1BBE-13.

g. Close INST AIR LINE 136 DRAIN 1-2420-U4-152 (RC-89).

h. Restore air to RHR System Flow Control Valves by opening INSTR AIR ISOLATION TO LINE 136 1-2420-U4-151 (RC-85 overhead).

Printed October 6, 2009 at 15:02

Approved By Procedure Number Rev A. S. Parton Vogtle Electric Generating Plant 13011-1 67.1 Date Approved Page Number 4/24/09 RESIDUAL HEAT REMOVAL SYSTEM 30 of 115 INITIALS NOTES

• When in Mode 1, 2, or 3, 1-HV-8809A/B should not be shut simultaneously.

• One train of RHR at a time should be aligned for shutdown cooling.

• An operator should be stationed at RHR B high point vent 1 -HV-1 0466 to monitor for water flow to the floor drain when Step 4.4.3 h. is performed.

4.4.3 Align the RHR for shutdown cooling as follows:

a. Close the RHR TRN-B TO HOT LEG CROSSOVER ISO 1 -HV-871 6B.

b. Close RHR TRN-B HEAT EXCH OUTLET 1-HV-0607 and check closure at Group 1 MLB 02 2.2 or by computer point UD8703.

c. Close RHR TRN-B HEAT EXCH BYPASS 1-FV-0619 and check closure by computer point UD8698.

d. Verify open RHR PMP-B TO COLD LEG 3&4 ISO VLV 1-HV-8809B. (IV REQUIRED)

e. Place RHR PMP-B 1-HS-0621 in PULL-TO-LOCK.

f. Close the RWST TO RHR PMP-B SUCTION 1-HV-8812B.

g. Open the RHR PMP-B SUCTION FROM HOT LEG LOOP 4 1 -HV-8702A.

h. Open RHR SUCTION VENT LINE TRN-B 1-HV-10466.

WHEN operator at 1 -HV-1 0466 reports water flowing to the floor drain, close 1-HV-10466.

j. Open the RHR PMP-B SUCTION FROM HOT LEG LOOP 4 Valve 1-HV-8702B.

k. Place RHR PMP-B 1-HS-0621 in AUTO position.

Printed October 6, 2009 at 15:02

Approved By Procedure Number Rev A. S. Parton Vogtle Electric Generating Plant 13011-1 67.1 Date Approved Page Number 4/24/09 RESIDUAL HEAT REMOVAL SYSTEM 31 of 115 INITIALS 4.4.4 Remove power from the RHR to SI Pump Isolation Valve as follows:

a. Check 1-HV-8804B CLOSED.

b. Open breaker 1BBB-05 to valve 1-HV-8804B.

c. Open the K2 links for breaker 1 BBB-05 and tag per NMP-AD-003, Equipment Clearance And Tagging.

4.4.5 Verify train related CCW System is in service per 13715-1, Component Cooling Water System.

4.4.6 Start up one train of RHR as follows:

a. Verify open the RHR PMP-B MINIFLOW ISO 1-FV-0611.

CAUTION In order to prevent excessive RHR heat up and possible pump damage, RHR HEAT EXCH OUTLET for Train B 1-HV-0607 and RHR HEAT EXCH BYPASS for Train B 1-FV-0619 must be closed. Actual valve position should be monitored at Group 1 MLB 02 2.2 or by computer point prior to pump start.

b. Start RHR PMP-B.

c. Establish RHR Letdown per Section 4.5.

4.4.7 Warm up the TRN-B RHR as follows:

NOTE Due to leak-by of the RHR Hx Outlet and Bypass Valves, RHR warming will begin as soon as the pump is started. However, due to miniflow cooling back to the suction of the pump, the temperature rise at the Hx inlet is only expected to reach approximately 200°F with the RCS at approximately 350°F. A rapid temperature rise should be expected when the miniflow valve goes closed.

a. Monitor RHR TRN-B Heat Exchanger Inlet Temperature using Plant Computer T0631, until the temperature stabilizes.

Printed October 6, 2009 at 15:02

Approved By Procedure Number Rev A. S. Parton Vogtle Electric Generating Plant z. 13011-1 67.1 Date Approved Page Number 4/24/09 RESIDUAL HEAT REMOVAL SYSTEM 32 of 115 INITIALS CAUTION If the RCS is under vacuum, a minimum flow rate of about 1200 gpm for 3 minutes is needed to refill the voided section of RHR discharge piping. 1500 gpm should NOT be exceeded during the refill period. Flow rates are to be adjusted very SLOWLY any time flow is being increased due to possible water hammer concerns.

b. Throttle open the RHR TRN-B HEAT EXCH BYPASS 1-FV-0619 until RHR PMP-B MINIFLOW ISO VLV 1-FV-0611 closes.

c. Complete RHR warm-up by monitoring RHR Hx Train B Inlet Temperature using Plant Computer T0631, until the temperature stabilizes.

4.4.8 WHEN RHR warm-up is completed, initiate full flow to the RCS as follows:

NOTES

• >3200 gpm indicated flow ensures >3000 gpm actual flow for all temperatures.

• 3000 gpm RHR flow is required for Mode 6.

CAUTION If the RCS is under vacuum, a minimum flow rate of about 1200 gpm for 3 minutes is needed to refill the voided section of RHR discharge piping. 1500 gpm should NOT be exceeded during the refill period. Flow rates are to be adjusted very slowly any time flow is being increased due to possible water hammer concerns.

a. Throttle open the RHR HEAT EXCH BYPASS for Train B 1-FV-0619 to the desired flow rate (nominally 3000 gpm).

b. Verify the RHR PMP-B MINIFLOW ISO VLV 1-FV-061 1 closes.

Printed October 6, 2009 at 15:00

Approved By Procedure Number Rev A. S. Parton Vogtle Electric Generating Plant 13011-1 67.1 Date Approved Page Number 4/24/09 RESIDUAL HEAT REMOVAL SYSTEM 33 of 115 INITIALS CAUTION The RHR Heat Exchanger Train B Bypass Flow Controller Potentiometer should be set for a minimum flow of 3000 gpm (Pot setting: 3.6 for 3000 gpm, 4.1 for 3200 gpm) prior to placing controller in AUTO. The potentiometer settjng for the desired flow rate (gpm) is approximately equal to (Desired Flow/5000) x 10.

c. Place the RHR TRN-B HEAT EXCH BYPASS Flow Controller 1-FIC-0619A in AUTO, if desired.

NOTE During Solid Plant conditions only 1-PIC-0131 should be used for letdown flow control and 1-HV-0128 should remain in the FULL OPEN position.

d. Adjust the LOW PRESSURE LETDOWN Controller 1-PIC-0131 and/or LETDOWN FROM RHR Control Valve 1-HV-0128 as required to maintain desired letdown flow.

e. Slowly throttle open RHR TRN-B HEAT EXCH OUTLET 1-HV-0607 to establish desired RCS cooling.

4.4.9 IF RCS cooling using both RHR trains is desired, place the second train in service:

U RHR A is in STANDBY READINESS, use Section 4.3.

HZ RHR A is NOT in STANDBY READINESS, use Section 5.3.

4.4.10 Establish RCS Cool down per 12006-C, Unit Cool down To Cold Shutdown.

Printed October 6, 2009 at 14:55

Approved By Procedure Number Rev S.E. Prewitt Vogtle Electric Generating Plant 14000-1 83 Date Approved Page Number 11/25/2008 OPERATIONS SHIFT AND DAILY SURVEILLANCE LOGS 23 of 29 DATA SHEET 3 - MODE 5 & 6 MODE Sheet 2 of 6 DATE LCD TECHSPEC INDICATION flj METHOD OF VERIFICATION SURV REQ PARAMETER INSTRUMENT DAY NIGHT TOLERANCE LCOIPROC RHR TRAINS NTER TRAINS (A B) OPERABLE ATLEASTI RHR TRAIN RHR FLOW IFIC-0618A *>3200 SHALL BE IN OPERATION 3.4.7.1 (GPM( 3.4.7 AND THE REQUIRED RHR 3.4.7.2 RHR FLOW IFIC-0619A *

>3200 TRAINS OR THE REQUIRED (MODE 5, GPM) -

SGS OPERABLE LOOPS FILLED) I ILI-0501 VERIFY RHR CIRCULATION STEAM -

AND OR SG LEVELS 3.4.8.1 GENERATOR 2 1LI-0502 3.4.8 (MODE 5 LEVEL -

LOOPS NOT FILLED) )%( 3 1LI-0503 4 1 LI-0S04 SR 3.9.5.1 MODE 5, LOOPS FILLED - AT LEAST I RHR TRAIN IN OPERATION AND ONE ADDITIONAL RHR TRAIN SHALL BE 3.9.5 (MODE 6 >23 OPERABLE OR THE SECONDARY SIDE water level OF AT LEAST TWO STEAM ABOVE FLANGE) GENERATORS 363% WIDE RANGE. Steam Generators may not be used as an option to an RF-IR train unless the RCS is filled greater than 15% Pressurizer level and RCS pressure has SR 3.9.6.1 been maintained >100 psig since the most recent fill & vent. 3.9.6 (MODE 6 <23 ABOVE FLANGE) MODE 5, LOOPS NOT FILLED - AT LEAST 2 RHR TRAINS OPERABLE WITH I TRAIN IN OPERATION.

MODE 6, >23 FT ABOVE FLANGE - AT LEAST I RHR TRAIN OPERATING WITH >3000 GPM# FLOW.

MODE 6, <23 FT ABOVE FLANGE AT LEAST 2 RHR TRAINS OPERABLE WITH I TRAIN IN OPERATION WITH >3000 GPM# FLOW.

>3200 GPM ENSURES >3000 GPM ACTUAL FLOW AT ALL TEMPERATURES RHR FLOW OF 3000 GPM IS ONLY REQUIRED IN MODE 6 AND IS NIA IN MODE 5 COMPLETED BY: DAY: TIME: NIGHT: TIME:

SS REVIEW: DAY: TIME: NIGHT: TIME:

Printed October 6, 2009 at 15:20

HL-15R RO NRC Exam

27. 026A3.02 002!2!1/C.SPRAY-HS COOLING/MEM -3.9 / 4.2/NEW/HL-15RNRC/RO/DS/TNT Which one of the choices correctly lists ALL the locations where the control room crew can monitor and control the CNMT Coolers following an RCS LOCA?

MLBs Monitor Light Boxes on the vertical section of the main control board QMCB (NSCW) Sloping portion of the NSCW section of the main control board QHVC Main Control Room HVAC panel Indications Controls A. Fan speed MLBs, QHVC QHVC Cooling water valves QMCB (NSCW) QMCB (NSCW)

B. Fan speed MLBs, QMCB (NSCW) QMCB (NSCW)

Cooling water valves MLBs QMCB (NSCW)

C. Fan speed QHVC QHVC Cooling water valves QMCB (NSCW) QMCB (NSCW)

D Fan speed MLBs, QHVC QHVC Cooling water valves MLBs, QMCB (NSCW) QMCB (NSCW)

KIA 026 Containment Spray System (CSS)

A3.02 Ability to monitor automatic operation of the CSS, including; Verification that cooling water is supplied to the containment spray heat exchanger.

K/A MATCH ANALYSIS The question requires the student to properly identify what control room cooling water indications are available to monitor the performance of the CNMT coolers.

57

HL-15R RO NRC Exam ANSWER I DISTRACTOR ANALYSIS A. Incorrect. The fan speed indications and controls listed are correct. The NSCW control and indication locations listed are correct making this choice very plausible.

However, the list for NSCW is incomplete, making this choice incorrect.

B. Incorrect. This choice is plausible but incorrect due to the fan speed control being listed as from the NSCW section. The fans controlled from this section of the main control boards are the NSCW cooling tower fans.

C. Incorrect. This choice is plausbie soince the locations listed are all correct.

However, the list is incomplete since the MLBs has not been included in the list.

D. Correct. Fan speeds can be monitored from the MLBs on the main control board.

Fan sppeds can be monitored and controlled on the HVAC panel. The HVAC panel has the high and low speed handswitches as well as the breaker position inidcating lights. The NSCW CNMT coolers cooling water isolation valves are indicated on the MLBs. The NSCW CNMT coolers cooling water isolation valves are controlled from the sloping section of the NSCW system on the main control board with motor operator hand switches and position indicating lights.

REFERENCES VEGP P&lD 1X4DB135-1, NSCW System VEGP P&lD 1X4DB212, CNMT Heat Removal V-LO-TX-06101, NSCW System Text page 13 V-LO-TX-291 01, CNMT HVAC systems text, page 15 VEGP learning objectives:

LO-PP-291 01-10:

State the cooling water systems that support Containment HVAC systems.

58

Date: 12/17/2009 Time:12:31:47 PM C

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B HEADER INFORSNATTON SEE OWES 1040B146I & 1040R1472.

lEADER INFORIAATTON SEE OWE 16400143, 10400144I &

A. FOR CONTAINMENT COOLERS AND COB flux AIR COOLER AIR SIDE DRAINS SEE 0MG. 0400212. rQ

7. FOR REACTOR CASEY COOLER AIR OWE DRAINS SEE OHIO 1XRDB2O41.

9. CHECK VALVE TO HE LOCATED CLOSE TO TEE.

SOUTHERN COMPANY APONENT COOUNO WATER SYSTEM AND SHELL SIDE CORNS 4501381. 9. CHEER VALVE TO HE CLOSE TO PENETRATION. GEORGIA POWER COMPANY MEAT COOLING WATER SYSTEMS MID SHElL SIDE CONSS SEC IT. LINE SPACER TO BE IUSTVLLED. ALVIN W. VOGTLE NUCLEAR PLANT 36.

Ii. FOR NORMIAL OPERATION, VALVE TO RE LOCKED THEN AT SETTNG OF 42.0 104081351 P & I DIAGRAM CAD NUCLEAR SERVICE A TEN. BLDG SONY INFORMATION SEE DWG. 104D81462. DEGREES FROM FELL SLTSED POSITION.

[EAO 002 I NJ TI I COOLING WATER SYSTEM SYSTEM NO. 1202 I I I I THIS DOCUMENT CONTROlS PRSPKETRAO. CONTDDI000. lAD/OR IRROC SECRET INFGRAATON OF THE SADSIDIADIES OF TOE I H

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V L which carry 5 gpm of NSCW from the discharge of the respective pump and return the water directly to the NSCW basin.

Pump Discharge Valves 6.3.2 NSCW Cooling Towers The nuclear service water mechanical draft cooling towers are the ultimate heat sinks for the plant. Each NSCW tower (train) is designed to accommodate the heat load from the unit under both normal and emergency conditions. Normally, both towers will be operating.

Although designed for one in operation and one in standby, both in operation ensures the reliability and readiness of all components served by NSCW, if required.

The mechanical draft cooling tower uses four fans to induce the movement of air up through the cooling tower. The nuclear service cooling water enters near the top of the cooling tower. A distribution pipe disperses the return water over the entire area of the tower. As the water is released from the distribution pipe, it falls through the fill (material used to break the water into fine droplets) . Curved asbestos cement board hung vertically is used for the fill material. Air is drawn through the windows surrounding the 13 Revision 7.0

VLo-TY-I 9 O/

The following systems use NSCW supplied water in cooling coils to remove heat in containment:

a. Reactor Cavity Cooling Fans

b. Containment Fan Coolers

c. Containment Auxiliary Coolers

4. Normal Chilled Water During refueling outages normal chilled water is supplied to the train B supplied Containment Auxiliary Cooler and Reactor Cavity Cooling Fan

5. 480 VAC 1E and Non lE power See the individual component descriptions for power supplies.

29.3 INSTRUMENTATION AND CONTROL Instrumentation and control, indications, alarms and interlocks will be discussed in this section system by system. All hand switch controls stop-auto-start, spring return to auto for fans or close auto-open, spring return to auto for dampers unless otherwise noted.

Containment Coolers System 1501 Controls and Instrimentation The Containment Coolers have hand switch controls on the QHVC panel and their respective Remote shutdown panels. There are separate hand switch controls for the low and high speed fans in both locations. On the QMCB, there are supply and return flow indicators for each pair of coolers. There are also hand switch controls for the NSCW supply and return MOV5 on the QMCB. There are status lights on the monitor light boxes (MLBS) for the Containment Cooler low speed operation, and for the NSCW MOVs.

There are cooler low flow alarms for each pair of coolers on ALBO2 and ALBO3.

Control Functions and Interlocks The fans may be manually started from the control room in either high or low speed. The high and low speed controls are interlocked so only one speed may be energized. The fans must be run in pairs in specific combinations as outlined in the procedure to allow for even cooling and prevent backf low through idle fans.

On a loss of off site power, the fans in auto will be started in high speed. The Sequencer gives all fans a start signal at 30.5 secs, but delay timer delays the start of two fans by 20 seconds to prevent voltage swings on the 416OVAC lE busses.

On an SI signal, SSPS and the sequencer will trip off any fans running in High speed and the SI sequence will restart the fans in low speed.

All fans start at 30.5 secs.

15 Revision 2.0

HL-15R RO NRC Exam

28. 027AA1.05 OO1!1/1/PZRPRESS-HTRPOWERJC/A -3.3 / 3.2/M-ANO 2005/HL-15RNRC/RO/TNT/DS Given the following:

- The plant is at full power when a loss of offsite power causes a plant trip.

- Both EDGs start and tie onto their respective ESF buses. All equipment sequences on as expected.

Which ONE of the following is CORRECT for PRZR heater banks available for RCS pressure control?

A. All Backup Heater Banks only.

B. Proportional Heater Bank only.

C Backup Heater Banks A and B only.

D. All Proportional and Backup Heater Banks.

KIA 027 Pressurizer Pressure Control System (PZR PCS) Malfunctions AAI .05 Ability to operate and I or monitor the following as they apply to the Pressurizer Pressure Control Malfunctions:

Transfer of heaters to backup power supply.

KIA MATCH ANALYSIS The question presents a plausible scenario where a Loss of Offsite Power occurs resulting in a plant trip with the DGs re-energizing the 1 E emergency buses. The student must choose which PRZR heaters are still available for use (powered from the backup power via DGs). The other heaters powered from the non-i E buses will lose power on the reactor trip when fast bus transfer occurs due to the RATs being de-energized.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. All backup heaters are not available, only backup heaters A and B are powered from the stub buses which would re-energize on an LOSP.

B. Incorrect. The proportional heaters are not available, only backup heaters A and B are power from the stub buses which would re-energize on an LOSP.

C. Correct. Backup heaters A and B are powered from the stub buses which would re-energize on an LOSP.

59

HL-15R RO NRC Exam D. Incorrect. All backup heaters are not available, only backup heaters A and B are powered from the stub buses which would re-energize on an LOSP. The proportional heaters are not available either as they are powered from non-i E buses.

REFERENCES ANO 2005 RO NRC exam question # 57 (included).

HL-1 5 RD Audit exam (April 2009) question # 16 (included).

V-LO-PP-i 6303, Pressurizer Pressure Control, slide #25 (included).

One line 1X3D-AA-EO1A (NBO1), Myriad drawings in refereneces.

One line 1X3D-AA-E1OA (NB1O), Myriad drawings in refereneces.

One line 1X3D-AA-EO8A (NBO8), Myriad drawings in refereneces.

One line 1X3D-AA-EO9A (NBO9), Myriad drawings in refereneces.

One line 1X3D-AA-F13A (PRZR heater panels), Myriad drawings.

VEGP learning objectives:

LO-LP-39208-01, For any item in section 3.4 of Tech Specs, be able to:

a. State the LCD 60

Questions For 2005 ANO UNIT 2 ROISRO Exam BANK 0498 Rev 0 Rev Date: 10/29/2004 RO Select: Yes SRO Select: Yes Points: 1.00 Lic Level: RS Difficulty: 2 Taxonomy: K Source: NEW Originator COBLE 10CFR55_41: 41.7 10CFR55_43: NA Section: 3.2 Type RCS INVENTORY System PRESSURIZER LEVEL CONTROL System 011 K/A: K2.02 RO Tier: 2 RO Group: 2 RO Imp: 3.1 SRO Tier: 2 SRO Group: 2 SRO Imp: 3.2 Description Knowledge of bus power supplies to the following: PZR Heaters.

Question # 57 Given the following:

The plant is at full power when a loss of offsite power causes a plant trip Electrical Bus 2A1 has a LOCKOUT alarm in.

Both EDGs start and tie onto their respective ESF buses Which of the following pressurizer heater banks would be available for RCS pressure control?

A. Both Proportional heater banks.

B. All Backup heater banks.

C. Both Backup heater banks #3 and #4.

D. All Proportional and Backup heater banks.

Answer:

A. Both Proportional heater banks Notes:

Distracter B is incorrect because BU heaters banks are 480 non vital powered and there is no power to their bus.

Distracter C is incorrect because all BU heater banks are non vital powered.

References STM 2-03, RCS, Section 2.2.2 A2LP-RO-RCS OBJ I0.c, Describe the following, concerning the RCS pressurizer: Heaters Historical This question has not been used on any previous NRC exams. BNC 10/29/2004.

ft i / I U

1. 011K2.02 001 Given the following:

- The plant is at full power when a loss of offsite power causes a plant trip.

- Both EDGs start and tie onto their respective ESF buses. All equipment sequences on as expected.

Which ONE of the following is CORRECT for PRZR heater banks available for RCS pressure control?

A. All Backup Heater Banks only.

B. Proportional Heater Banks only.

C Backup Heater Banks A and B only.

D. All Proportional and Backup Heater Banks.

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HL15R RO NRC Exam

29. 028K5.O1 OO1/2/2/H2 PURGE-EXPLSVE H2/MEM 3.4 / 3.9/NEW/HL-15R NRC/RO/DS/TNT A large RCS LOCA has occurred The CNMT Hydrogen monitors indicate 5%

Service Air to CNMT is NOT available.

Which one of the following choices correctly desribes the operational implication of the hydrogen monitor readings and methods to reduce hydrogen concentration inside containment?

Operational Implication Method used to reduce hydrogen A. Combustible atmosphere Post-LOCA hydrogen recombiners B Combustible atmosphere Post-LOCA hydrogen purge C. Embrittlement of cladding Post-LOCA hydrogen recombiners D. Embrittlement of cladding Post-LOCA hydrogen purge KIA 028 Hydrogen Recombiner and Purge Control System (HRPS)

K5.01 Knowledge of the operational implications of the following concepts as they apply to the HRPS:

Explosive hydrogen concentration K/A MATCH ANALYSIS The question requires the student to correctly identify the implication of a high containment hydrogen concentration and what method is used to remove the hydrogen from containment.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Hydrogen becomes combustible above 4% in air. The Post-LOCA hydrogen recombiners have been retired in place.

B. Correct. Hydrogen becomes combustible above 4% in air. The Post-LOCA purge system is a method used per 13130 to reduce CNMT hydrogen concentration.

C. Incorrect. High hydrogen concentrations will cause embrittlement of steel and Zirconium. However, this is NOT the issue for high hydrogen in containment atmosphere. The Post-LOCA hydrogen recombiners have been retired in place.

61

HL15R RO NRC Exam CNMT hydrogen concentration.

REFERENCES V-LO-PP-29101 CNMT HVAC Systems presentation, slides 3, 84, and 86 V-LO-LP-35103 Corrosion lesson plan, pages 14 and 15 SOP 13130-1, Post-Accident Hydrogen Control page 1 and 14.

VEGP learning objectives:

V-LO-LP-361 07-02:

State the hazardous concentration ranges of explosive and flammable mixtures of hydrogen in air.

V-LO-LP-361 07-03:

State the means available to measure and control the containment hydrogen concentration.

V-LO-PP-291 01-18:

State the upper and lower limits for an explosive mixture of hydrogen.

V-LO-PP-291 01-20:

List the methods for monitoring and controlling hydrogen inside Containment.

62

Authors Rev o Date Reason for Revision pys

. Initials Initials 0.0 1/7/04 Initial Development PMT DS 1.0 5/04/04 Incorporated Instructor Feedback Action Item 6 2004201028 TLH DS 2.0 10/07/05 Per Al 2005202977 Delete references to CNMT H2 Recoinbiners LPV DS 3.0 10/19/05 Removed H2 Recotnbiners from Picture of Containment internals LPV DS 4.0 11/02/05 Incorporated MDC 2000.048 per Action Item 6 2004201074 TLH DS 5.0 06/12/06 Revised for 06/13/06 class to enitance and cover the objectives. TNT DS 3

HYDROGEN IS FORMED BY:

• Zirconium - 0 Reaction 2

H

• Aluminum/zinc - 0 2

H

• Radiolysis of Water

• H2 From the RCS and PRZR 84 H2 flammability (combustible) limit is 4% to 75% in air.

18% 59% is the explosive range.

I i 84

N co o C 0 0

E 0

C.) 0 ci) O I D U CO ci) ci) Q Q 0 o F

> c C

Cl) ci Z C) 0 o 0 50 _j

0) CD C/) ci)

I- 0-.D 0)

- 0 C,)

- ci) 0 -

o C 0 -o J ci) 0 Iø

= 0 I U U 0

LO-LP-351 03-07-C Ill. LESSON OUTLINE: NOTES

a. In steam generators around the tubes where boiling can cause concentration of hydroxyl ions

b. Nucleate boiling in the core can also lead to a caustic environment

5. The mechanism of caustic stress corrosion is similar to that of chloride stress corrosion Q. Carbide Precipitation (Sensitization)

1. Type 304 stainless steel is susceptible

2. Sensitization is an increased susceptibility to intergranular attack

3. Sensitization is caused by exposure to heat treatments that precipitate chromium-rich carbides along the grain boundaries

a. Carbon atom ties up many chromium atoms

b. Chromium cannot exert corrosion-resistant effect

4. These carbides are precipitated during thermal and strain transients such as multiple pass welding

5. High carbon content of alloys contributes to susceptibility to sensitization

6. Susceptibility is decreased by stabilizing with carbine forming elements (boron, titanium, niobium)

R. Hydrogen Embrittlement LO-TP-35103-021 Objective 7

1. Process by which steel loses ductility and strength due to tiny cracks

- Result from internal pressure of H2 or CH4 which forms at grain boundaries

2. Process

a. Monatomic hydrogen produced in corrosion reaction is absorbed into metal

b. Hydrogen diffuses along grain boundaries

c. Hydrogen can combine with carbon, which is alloyed with iron, to form methane gas 14

LO-LP-351 03-07-C Ill. LESSON OUTLINE: NOTES

d. Gas collects in small voids and builds up enormous pressures which initiate cracks

e. If the hydrogen atoms combine on the surface of the metal, they are released to the environment as Il2 gas

3. Zirconium alloys are susceptible

a. Zr + 0 2

2H ---> 2 Zr0 + 2 ÷ heat 2H

b. H2 diffuses through oxide layer to metal

c. Zircaloy absorbs as much as 50% of 2

corrosion-produced hydrogen

d. Zircaloy absorbs significantly less 4

because of:

1) Lower percentage of nickel

2) Addition of niobium S. Boric Acid Corrosion Commitment start GP-1 2962.000

1. Carbon steel highly susceptible, stainless steel mildly susceptible

2. Carbon steel and stainless steel undergo wastage or general dissolution corrosion

a. Other forms of corrosion; i.e., pitting, stress corrosion cracking, intergranular attack not types of boric acid corrosion

3. Conditions needed for corrosion to occur

a. High boric acid concentration (>1 %)

b. Elevated temperatures (approx 200° or more)

c. Susceptible material

d. Aerated atmosphere

4. Corrosion rates Concentration of Temperature Corrosion rate boric acid (%) Condition (°F) mils/month 25 Aerated 200 400 25 Deaerated 200 250 15 Aerated 200 350-400 15-25 Dripping 210 400 15

Approved By Procedure Number Rev S. A. Phillips Vogtle Electric Generating Plant 131 30-1 18.2 Date Approved Page Number 5/14/07 POST-ACCIDENT HYDROGEN CONTROL 2 of 22 1.0 PURPOSE This procedure provides instructions for operation of the Containment Hydrogen Monitoring System, the Post-LOCA Cavity Purge System, and the Post-LOCA Containment Hydrogen Purge System during normal and post-LOCA conditions.

Instructions are provided in the following sections.

4.1.1 Placing The Containment Hydrogen Monitoring System In Standby 4.1.2 Deleted 4.1.3 Placing The Post-LOCA Cavity Purge And Post-LOCA Containment Hydrogen Purge Systems In Standby 4.2.1 Containment Hydrogen Monitor 1-151 3-P5-HMA Operation (Hydrogen Measurement) 4.2.2 Containment Hydrogen Monitor 1-1513-P5-HMB Operation (Hydrogen Measurement) 4.4.1 Deleted

,4)MT e)pd-c4Lq?

4.4.2 Diluting Containment Hydrogen Concentration Using The Service Air System 4.4.3 Post-LOCA Containment Hydrogen Purge System Operation 4.4.4 Changing 02 Reagent Gas Bottles At The H 2 Monitors 2.0 PRECAUTIONS AND LIMITATIONS 2.1 PRECAUTIONS 2.1.1 Adhere to all applicable radiological controls.

2.1.2 Train A Hydrogen Monitor Supply Valves 1-HV-2792A, 1-HV-2792B, 1-HV-2791B, and Return Valve 1-HV-2793B may be opened in Modes 1,2,3, and 4 under administrative control as described in the basis for Technical Specification LCD 3.6.3.

2.1.3 Train B Hydrogen Monitor Supply Valves 1-HV-2790A, 1-HV-2790B, 1-HV-2791A, and Return Valve 1-HV-2793A may be opened in Modes 1,2,3, and 4 under administrative control as described in the basis for Technical Specification LCD 3.6.3.

Printed November 11,2009 at 16:05

Approved By I Procedure Number Rev S.A. Phillips I Vogtle Electric Generating Plant 13130-1 18.2 Date Approved I POSTACCIDENT HYDROGEN CONTROL Page Number 5/14/07 I 14 of 22 INITIALS 4.43 Post-LOCA Containment Hydrogen Purge System Operation NOTE If plant conditions warrant, the Emergency Director may waive the Gaseous Release Permit requirement.

C) CAUTIONS

• The Post-LOCA Containment Hydrogen Purge System is to be operated ONLY if the containment hydrogen concentration cannot be maintained

-i below 4% by other means.

Vs (5

• The Post-LOCA Containment Hydrogen Purge System is designed to tle operate with a maximum pressure of 3 psi downstream of CNMT POST LOCA PURGE EXH DUCT CONTROL VLV 1-FV-2693.

4.4.3.1 Initiate a Gaseous Release Permit.

4.4.3.2 Verify containment atmosphere is sampled and analyzed.

4.4.3.3 Verify the Service Air System is operating.

4.4.3.4 Verify compliance with the ODCM Section 3.1.1 Table 3-1 for the gaseous effluent monitoring requirements.

4.4.3.5 Verify the Auxiliary Building Heating Ventilation And Air Conditioning System is operating.

4.4.3.6 Place disconnect switch at local Heater Control Panel 1-1508-N7-0O1-HO1 to on.

4.4.3.7 Push RESET button at local Heater Control Panel 1-1508-N7-001-H01 and verify that reset red light is ON.

Critical 4.4.3.8 Due to high radiation area potential, verify Containment Inside Isolation Valves 1-HV-2624A and 1-HV-2624B are closed and remain closed during the performance of the next step and until personnel have exited the area.

CV Printed December 17, 2009 at 15:36

HL15R RO NRC Exam

30. 032AA2.07 001/1/2/LOSS SR NI-CH DISAGR/C/A 2.8 / 3.4/NEW/HL-15R NRC/RO!TNT/DS The plant is in Mode 3.

- SR I IR Signal Processor Channel Operational Tests are in progress.

- Background counts for both SR channels are 1000 cps

- The UO records the counts when the HFASA alarm lights for each SR channel.

N31 - 2080 cps N32 - 2340 cps Which ONE of the following is CORRECT regarding Technical Specification 3.3.8, High Flux At Shutdown Alarm (HFASA)?

A. The LCO is met for both SR NlS HFASA alarms.

B. LCO entry required due to N31 setpoint too low.

C LCO entry required due to N32 setpoint too high.

D. LCO entry required for both SR NIS HFASA alarms.

K/A 032 Loss of Source Range Nuclear Instrumentation AA2.07 Ability to determine and interpret the following as they apply to the Loss of Source Range Nuclear Instrumentation.

Maximum allowable channel disagreement.

KIA MATCH ANALYSIS The question presents a plausible scenario where a SR Signal Processor Channel Operational Test is in progress. The student is given data for the cps where the HFASA alarm illuminates. The student must determine if the LCO for HFASA is met and why.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. The LCO HFASA alarm setpoint is > 2.3 X background. The HFASA LCO is required to be entered.

B. Incorrect. SR N31 is within the allowable range for the HFASA alarm, there is no low alarm setpoint requirement per Tech Specs. Per our procedures, >2 X background is the normal conservative alarm setpoint. Therefore, N31 alarms close to the expected setpoint per our ARPs.

63

HL-15R RO NRC Exam criteria of the procedure. N31 setpoint is set as expected (>2 X background).

D. Incorrect. While N32 setpoint is too high and outside the Tech Spec limits and acceptance criteria, N31 setpoint is set as expected (>2 X background)

REFERENCES OSP-1 4423-i, N31/N35 Signal Processor Channel Operational Test, data sheet 1.

DSP-i 4424-1, N32/N36 Signal Processor Channel Operational Test, data sheet 1.

17010-1, window COl for Source Range Hi Flux Level At Shutdown.

V-LO-PP-1 7201, Source and Intermediate Range NIS.

Technical Specification 3.3.8, High Flux At Shutdown Alarm (HFASA).

VEGP learning objectives:

LO-PP-17201-04, Discuss the operation of the High Flux At Shutdown Alarm LO-PP-17201-05, Discuss all applicable Technical Specifications associated with the Source and Intermediate Range Nuclear Instrumentation (from memory).

a. All LCOs

b. Applicability

c. All 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> actions.

64

Approved By . . Procedure Number Rev

s. A. Phillips Vogtle Electric Generating Plant ,

14424-1 34.2 Date Approved Page Number 12/1/06 N32/N36 SIGNAL PROCESSOR CHANNEL OPERATIONAL TEST 40 of 44 DATASHEET1 Sheet2of3 SECTION 5.1 N321N36 SIGNAL PROCESSOR CHANNEL OPERATIONAL TEST BELOW P-6 FUNCTION ACTION LOWER REQUIRED UPPER AS I IT VALUE LIMIT FOUND 2.3 x SDM SR HF@SD ALARM* indication IR P6 TRIP* 1.5 E-5 2.0 E-5 2.6 E-5 PERMISSIVE RESET 7.6 E-6 1.0 E-5 1.3 E-5 SR HI FLUX TRIP* 7.9 E+4 1.0 E+5 1.3 E+5 RESET 3.9 E+4 5.0 E+4 6.4 E+4 IR HI LEVEL TRIP 1.5 E+1 2.0 E+1 2.6 E+1 ROD STOP RESET 5.4 E+0 1.0 E+1 1.3 E+1 IR HI FLUX TRIP* 1.9 E+1 2.5 E+1 3.3 E+1 RESET 1.4 E+1 1.88 E+1 2.5 E+1 NOTE 1: Calculate HF@SD TRIP setpoint upper limit by multiplying SR count rate recorded in Step 5.1 .3.2 by 2.3 and record in the Upper Limit box for SR HF@SD.

NOTE 2: Attach additional Data Sheet 1 if High Flux At Shutdown Alarm Setpoint Retest is required following calibration.

Technical Specification limit (ACCEPTANCE CRITERIA).

Must be less than or equal to 2.3 times background level recorded in Step 5.1 .3.2 WHEN performing quarterly surveillance (ACCEPTANCE CRITERIA).

Containment Evacuation Siren Test Completed SAT INIT P-6 Permissive Verified RESET NIT P-i 0 Permissive Verified RESET INIT Printed October 14, 2009 at 16:27

Approved By Procedure Number Rev S. A. Phillips Vogtle Electric Generating Plant . 14423-1 35.1 Date Approved Page Number 12/1/06 N31/N35 SIGNAL PROCESSOR CHANNEL OPERATIONAL TEST 39 of 43 DATA SHEET 1 Sheet 2 of 3 SECTION 5.1 N311N35 SIGNAL PROCESSOR CHANNEL OPERATIONAL TEST BELOW P-6 FUNCTION ACTION LOWER REQUIRED UPPER AS LIMIT VALUE LIMIT FOUND 2.3xSDM SR HF@SD ALARM* indication IR P6 TRIP* 1.5 E-5 2.0 E-5 2.6 E-5 PERMISSIVE RESET 7.6 E-6 1.0 E-5 1.3 E-5 SR HI FLUX TRIP* 7.9 E+4 1.0 E+5 1.3 E+5 RESET 3.9 E+4 5.0 E+4 6.4 E+4 IR HI LEVEL TRIP 1.5 E+1 2.0 E+1 2.6 E+1 ROD STOP RESET 5.4 E+0 1.0 E+1 1.3 E+1 IR HI FLUX TRIP* 1.9 E+1 2.5 E+1 3.3 E+1 RESET 1.4 E+1 1.88 E+1 2.5 E+1 NOTE 1: Calculate HF@SD TRIP setpoint upper limit by multiplying SR count rate recorded in Step 5.1.3.2 by 2.3 and record in the upper limit box for SR HF@SD.

NOTE 2: Attach additional Data Sheet I if High Flux At Shutdown Alarm Setpoint Retest is required following calibration.

Technical Specification urn it (ACCEPTANCE CRITERIA).

Must be less than or equal to 2.3 times background level recorded in Step 5.1.3.2 when performing quarterly surveillance. (ACCEPTANCE CRITERIA)

Containment Evacuation Siren Test Completed SAT INIT P-6 Permissive Verified RESET I N IT P-b Permissive Verified RESET INIT Printed October 14, 2009 at 16:23

Approved By Procedure Number Rev S. A. Phillips Vogtle Electric Generating Plant 17010-1 48 DateApproved ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 10 ON PANEL 1.C1 Page Number 5/28/08 ONMCB 24 of 64 WINDOW C01 ORIGIN SETPOINT SOURCE RANGE 1N31AX* (NIS) 2.0 times Shutdown HI FLUX LEVEL 1N32AX* (NIS) Monitor Indication AT SHUTDOWN UNOO31 (IPC) 2.0 times IPC UN0032 (IPC) UN5031/UN5032 NOTES The High Flux At Shutdown Alarm should be in service:

• In Modes 3, 4 and 5.

• In Mode 6 with the Reactor Makeup Water Valves 1-1 208-U4-1 75, 1-1208-U4-176, 1-1208-U4-177, and 1-1 208-U4-1 83 not closed and secured in position (by mechanical stops).

The High Flux At Shutdown Alarm may be blocked:

• In Modes I and 2.

• During fuel movement when the Reactor Makeup Water Valves 1-1 208-U4-1 75, 1-1 208-U4-1 76, 1-1 208-U4-1 77, and 1-1 208-U4-1 83 are closed and secured in position (by mechanical stops) and the SR channels are operable.

• Mode 3 during reactor startup.

1.0 PROBABLE CAUSE Reactivity addition caused by any of the following:

1. RCS dilution.

2. RCS cool down.

3. Xenon decay.

4. Rod withdrawal.

5. Refueling activities.

2.0 AUTOMATIC ACTIONS Containment evacuation horn actuates.

Printed October 14, 2009 at 16:49

Source & Intermediate Range NIS ANN LIX C ATOP LIGHT BOX 1619 Q5ALBO 10 Technical Specification (High Flux At Shutdown)

The primary purpose of the High Flux AT Shutdown Alarm (HFASA) is to warn the operator of an unplanned boron dilution event in sufficient time (15 minutes prior to loss of shutdown margin) to allow manual action to terminate the event. In order to comply with this you must have two separate channels of alarms OPERABLE receiving inputs from their respective Source Range Channels. The Technical Specification required setpoint for this Main Control Room alarm is 2.3 times background, however ours is set a 2.0 times background which is more conservative (T.S./Bases for 3.3.8).

This function is required to be OPERABLE in MODES 3, 4, and 5 (The HFASA may be blocked in MODE 3 during Reactor Startup).

In addition if the plant is in MODE 6 for refueling or MODE 5, Loops not filled, all the unborated water sources to the RCS must be secured in the closed position. The Technical Specification does allow for RCS chemical additions using water from the Reactor Makeup Water Storage System (RMWST) through the chemical mixing tank under these conditions. One of the requirements would be to have the HFASA OPERABLE while adding chemicals. Refer to T.S./Bases for 3.4.8 & 3.9.2 for details.

V-LO-PP-1 7201 Rev-03 46

High Flux at Shutdown Alarm 3.3.8 3.3 INSTRUMENTATION 3.3.8 High Flux at Shutdown Alarm (HFASA)

LCO 3.3.8 Two channels of HFASA shall be OPERABLE. QF? e3 ee acce,4 APPLICABILITY: MODES 3 4 and 5 / \

1Th5 cc NOTE The HFASA may be bcked in MODE 3during reactor startup. Mk ACTIONS COMPLETION CONDITION REQUIRED ACTION TIME A. One channel of HFASA A.1 NOTE inoperable. LCO 3.O.4c is applicable provided Required Actions B.1 and B.2 are met.

Restore channel to OPERABLE 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> status.

B. Required Action and B.1 Perform SR 3.1.1.1 (verify 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> associated Completion SDM).

Time of Condition A not AND met.

Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter OR Two channels of AND HFASA inoperable.

B.2 Perform SR 3.9.2.1 (verify 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> unborated water source isolated).

Once per 14 days thereafter Vogtle Units 1 and 2 3.3.8-1 Amendment No. 137 (Unit 1)

Amendment No. 116 (Unit2)

HFASA Instrumentation B 3.3.8 B 3.3 INSTRUMENTATION B 3.3.8 High Flux at Shutdown Alarm (HFASA)

BASES BACKGROUND The primary purpose of the HFASA is to warn the operator of an unplanned boron dilution event in sufficient time (15 minutes prior to loss of shutdown margin) to allow manual action to terminate the event. The HFASA is used for this purpose in MODES 3 and 4, and MODE 5 with the loops filled.

The HFASA consists of two channels of alarms, with each channel receiving input from one source range channel. An alarm setpoint of 2.3 times background provides at least 15 minutes from the time the HFASA occurs to the total loss of shutdown margin due to an unplanned dilution event. This meets the Standard Review Plan criteria for mitigating the consequences of an unplanned dilution event by relying on operator action.

APPLICABLE The analysis presented in Reference 1 identifies credible SAFETY ANALYSES boron dilution initiators. Time intervals from the HFASA until loss of shutdown margin were calculated. The results demonstrate that sufficient time for operator response is available to terminate an inadvertent dilution event taking credit for one HFASA with a setpoint of 2.3 times background.

The HFASA satisfied Criterion 3 of 10 CFR 50.36 (c)(2)(ii).

LCO The LCO requires two channels of HFASA to be OPERABLE with input from two source range channels to provide protection against single failure.

APPLICABILITY The HFASA must be OPERABLE in MODES 3, 4, and 5.

The Applicability is modified by a Note which allows the HFASA to be blocked in MODE 3 during reactor startup so that spurious alarms are not generated.

(continued)

Vogtle Units 1 and 2 B 3.3.8-1 Rev. 2-10/01

HFASA Instrumentation B 3.3.8 BASES APPLICABILITY In MODES 1 and 2, operators are alerted to an unplanned (continued) dilution event by a reactor trip on overtemperature delta-T or power range neutron flux high, low setpoint, respectively. As a protective measure in addition to HFASA, in MODE 5 with the loops not filled, unplanned dilution events are precluded by requiring the unborated water source (reactor makeup water storage tank (RMWST)) to be isolated.

ACTIONS With one channel of HFASA inoperable, Required Action A.1 requires the inoperable channel to be restored within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

In this condition, one channel of HFASA remains available to provide protection. The 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> Completion Time is consistent with that required for an inoperable source range channel.

Required Action A.1 is modified by a Note stating that LCO 3.O.4c is applicable provided that Required Actions 8.1 and B.2 are met.

When Condition A (and Required Action Al) are applicable, the Note permits MODE changes provided that Required Action B.1 and B.2 are met. Required Action B.1 is a periodic verification of shutdown margin, and Required Action B.2 ensures that the unborated water source isolation valves are shut, precluding a boron dilution event. With one channel of HFASA inoperable, it is prudent to take the compensatory actions of Required Actions 8.1 and 8.2 if MODE changes are desired or required.

B.1 and 8.2 With the Required Action A.1 and associated Completion Time not met, or with both channels of HFASA inoperable, the appropriate ACTIONS are to verify that the required SDM is present and isolate the unborated water source by performing (continued)

Vogtle Units I and 2 B 3.3.8-2 Rev. 3 6/05

HFASA Instrumentation B 3.3.8 BASES ACTIONS B.1 and B.2 (continued)

SR 3.9.2.1. This places the unit in a condition that precludes an unplanned dilution event. The Completion Times of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter for verifying SDM provide timely assurance that no unintended dilution occurred while the HFASA was inoperable and that SDM is maintained. The Completion Times of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and once per 14 days thereafter for verifying that the unborated source is isolated provide timely assurance that an unplanned dilution event cannot occur while the HFASA is inoperable and that this protection is maintained until the HFASA is restored.

SURVEILLANCE The HFASA channels are subject to a COT and a CHANNEL REQUIREMENTS CALIBRATION.

SR 3.3.8.1 SR 3.3.8.1 requires the performance of a COT every 184 days to ensure that each channel of the HFASA and its setpoint are OPERABLE. This test shall include verification that the HFASA setpoint is less than or equal to 2.3 times background. The frequency of 184 days is consistent with the requirements for the source range channels. This Surveillance Requirement is modified by a Note that provides a 4-hour delay in the requirement to perform this surveillance for the HFASA instrumentation upon entering MODE 3 from MODE 2. This Note allows a normal shutdown to proceed without delay for the performance of the surveillance to meet the applicability requirements in MODE 3.

SR 3.3.8.2 SR 3.3.8.2 requires the performance of a CHANNEL CALIBRATION every 18 months. This test verifies that each channel responds to a measured parameter within the necessary range and accuracy. It encompasses the HFASA portion of the instrument loop. The frequency is based on operating experience and consistency with the typical industry refueling cycle.

REFERENCES 1. FSAR, Subsection 15.4.6.

Vogtle Units 1 and 2 B 3.3.8-3 Rev. 2-9/06

HL15R RO NRC Exam

31. 033A1.02 OO1/2/2/SFPCS-RAD MONITORS/MEM -2.8 / 3.3/NEW/HL-15RNRC/RO/TNT/DS Which ONE of the following is CORRECT regarding:

1) The minimum Spent Fuel Pool level (elevation) required by Tech Specs for adequate shielding and design basis fuel handling events.

2) how the FHB crew would be alerted to High radiation on RE-0008, Fuel Handling Building Area radiation monitor.

Tech Spec level RE-0008 high rad alarm A. 214 ft. 6 inch audible horn and blinking strobe light B. 214 ft. 6 inch warble type siren on plant gai-tronics C 217 ft. 0 inch audible horn and blinking strobe light D. 217 ft. 0 inch warble type siren on plant gai-tronics K/A 033 Spent Fuel Pool Cooling System (SFPCS)

Al .02 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with Spent Fuel Pool System operating the controls including:

Radiation monitoring systems.

K/A MATCH ANALYSIS The question asks the candidate the Tech Spec level required for adequate shielding in the Spent Fuel Pool and how the crew would be alerted of high radiation.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. 217 ft is the minimum Tech Spec limit (23 ft above fuel), 214 ft is plausible as this is prominently mentioned in the AOP. However, this is the level to stop the SEP cooling pumps and is plausible they may recall this level. 214 elevation is also the level a SEP cooling system line break can drain the pool down to. RE..0008 sounds an audible horn and strobe light to alert workers of high radiation, this part of the choice is correct.

B. Incorrect. 217 ft is the minimum Tech Spec limit (23 ft above fuel), 214 ft is plausible as this is prominently mentioned in the AOP. However, this is the level to stop the SEP cooling pumps and is plausible they may recall this level. RE-0008 sounds an audible horn and strobe light to alert workers of high radiation. A warble type siren is 65

HL-15R RO NRC Exam C. Correct. 217 ft is the Tech Spec minimum required level (23 ft above fuel) and RE-0008 sounds an audible horn and strobe light.

D. Incorrect. 217 ft is the Tech Spec minimum required level (23 ft above fuel) and RE-0008 sounds an aubible alarm and strobe light. A warble type siren is used on gai-tronics for plant events, but not locally on high radiation.

REFERENCES 171 00-1, ARP for RE-0008 high radiation (included).

18030, Loss of Spent Fuel Pool Cooling pages 3 and 4 (inlcuded)

Tech Spec 3.7.15, Fuel Storage Pool Level and Bases (Background)

V-LO-PP-25102, Spent Fuel Pool Cooling System, page #23 VEGP learning objectives:

LO-LP-251 02-09, Describe the impacts of the following conditions:

a. Low level in the Spent Fuel Pool.

LO-LP-251 02-12, Describe the minimum allowable water level over the spent fuel pool and the basis for this level.

66

ORIGIN SETPOINT 1 -RE-0008 Area Monitor As determined by (High)

Chemistry Department NOTE For other than HIGH conditions see Pages 4 and 5.

1.0 PROBABLE CAUSE Increase in radiation level near Unit 1 Spent Fuel Pool in the Fuel Handling Building.

2.0 AUTOMATIC ACTIONS On the south wall of the Fuel Handling Building Spent Fuel Pool Room near the door

a. Alarm horn on 1-RA-0008 sounds.

b. Strobe light on 1-RA-0008 blinks.

3.0 INITIAL OPERATOR ACTIONS Evacuate the Fuel Handling Building.

Approved By Procedure Number Rev J. B. Stanley Vogtle Electric Generating Pant 18030-C 19.1 DateApproved PageNumber LOSS OF SPENT FUEL POOL LEVEL OR 3/22/09 COOLING 3of18 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

1. Initiate the Continuous Actions Page.

NOTE Local SFP level indicator and FIGURE 1 should be compared to determine actual SFP level.

_*2. Check SFP level LESS THAN OR

• 2 Go to Step 20.

EQUAL TO 217 FT ELEVATION.

Place fuel assembly in transit in a safe place.

_4. Suspend all fuel assembly movement.

_5. operation over the

elo &1)

_6. Makeup to restore level by initiating 13719, SPENT FUEL POOL COOLING AND PURIFICATION SYSTEM.

7. Check conditions requiring 7. Go to Step 9.

emergency makeup:

. Imminent security threat.

. Any other extreme conditions warranting emergency makeup.

8. Perform the following as necessary:

• Contact the TSC for temporary repair options.

• Initiate ATTACHMENT C for SEP makeup sources.

Printed October 16, 2009 at 09:42

Approved By Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant 18030-C 19.1 Date Approved Page Number LOSS OF SPENT FUEL POOL LEVEL OR 3/22/09 COOLING 4 of 18 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

• 9 Check SFP level LESS THAN OR *9

- Go to Step 20.

EQUAL TO SFP PUMP SUCTION.

(214 ft. 6 in.)

_10. Stop SFP cooling pump(s).

UNIT 1 (At door to AB-A53 &

FHB-A06)

Pts( A+4/

UNIT 2 (At door to AB-A91 &

FHB-A05)

/ rV

_1 1. Stop SEP skimmer pump(s).

UNIT I (AB-A53)

UNIT 2 (AB-A91)

_12. Close SEP cooling pump suction valve(s).

UNIT 1

• 1213-U6-001 (FHB-A06)

• 1213-U6-003 (FHB-A07)

UNIT 2

• 1213-U6-001 (EHB-A05)

• 1213-U6-003 (EHB-A04)

13. Close SEP skimmer pump(s) suction isolation.

UNIT 1

• 1213-U6-014 (AB-A53)

UNIT 2

• 1213-U6-014 (AB-A91)

Printed October 16, 2009 at 09:42

Fuel Storage Pool Water Level 3.7.15 3.7 PLANT SYSTEMS 3.7.15 Fuel Storage Pool Water Level LCO 3.7.15 The fuel storage pool water level shall be 23 ft over the top of irradiated fuel assemblies seated in the storage racks.

APPLICABILITY: During movement of irradiated fuel assemblies in the fuel storage pool.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Fuel storage pool water A.1 NOTE level not within limit. LCO 3.0.3 is not applicable.

Suspend movement of Immediately irradiated fuel assemblies in the fuel storage pool.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.15.1 Verify the fuel storage pool water level is 23 ft 7 days above the top of the irradiated fuel assemblies seated in the storage racks.

Vogtle Units 1 and 2 3.7.15-1 Amendment No. 96 (Unit 1)

Amendment No. 74 (Unit 2)

Fuel Storage Pool Water Level B 3.7.15 B 3.7 PLANT SYSTEMS B 3.7.15 Fuel Storage Pool Water Level BASES BACKGROUND The minimum water level in the fuel storage pool meets the assumptions of iodine decontamination factors following a fuel handling accident. The specified water level shields and minimizes the general area dose when the storage racks are filled to their maximum capacity. The water also provides shielding during the movement of spent fuel.

A general description of the fuel storage pool design is given in J the FSAR, Subsection 9.1.2 (Ref. 1). A description of the Spent Fuel Pool Cooling and Cleanup System is given in the FSAR, Subsection 9.1.3 (Ref. 2). The assumptions of the fuel handling accident are given in the FSAR, Subsection 15.7.4 (Ref. 3).

APPLICABLE The minimum water level in the fuel storage pool meets SAFETY ANALYSES the assumptions of the fuel handling accident described in Regulatory Guide 1 .25 (Ref. 4). The resultant 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid dose per person at the exclusion area boundary is a small fraction of the 10 CFR 100 (Ref. 5) limits.

According to Reference 4, there is 23 ft of water between the top of the damaged fuel bundle and the fuel pool surface during a fuel handling accident. With 23 ft of water, the assumptions of Reference 4 can be used directly. In practice, this LCO preserves this assumption for the bulk of the fuel in the storage racks. In the case of a single bundle dropped and lying horizontally on top of the spent fuel racks, however, there may be < 23 ft of water above the top of the fuel bundle and the surface, indicated by the width of the bundle. To offset this small nonconservatism, the analysis assumes that all fuel rods fail, although analysis shows that only the first few rows fail from a hypothetical maximum drop.

The fuel storage pool water level satisfies Criterion 2 of 10 CFR 50.36 (c)(2)(ii).

(continued)

Vogtle Units 1 and 2 B 3.7.15-1 Rev. 1-10/01

Spent Fuel Pool Cooling System

• Piping is arranged so that the failure of any pipeline cannot drain pool below level required for radiation shielding

• Normal level is 24.5 ft above fuel Tech Spec minimum is 23 ft. above fuel Pump Suction Strainers located 4 ft below normal level

- VLO-PP-25TO2 Rev-J.1 V-LO-PP-25102-05, Describe how the Spent Fuel Pool is designed to minimize the occurrence and effects of inventory loss.

24.5 feet above fuel is 2186 23feetabovefuelis2l7 I N al Q 4

PL (

The biggest threat would be a Transfer Canal Gate seatLt &C)1) failure with the transfer tube open and the canal and reactor cavity empty.

We have admin controls in the UOP to handle this concern.

V-LO-PP-25 102 Rev-0 1.0 23

HL-15R RO NRC Exam

32. 03 5G2.2.40 001 /2/2!SG-TECH SPECSJMEM -3.4! 4.7!M- LOIT BANKIHL- 1 5R NRC/RO/TNT!DS SG # 2 ARV PV-301 0 has been declared inoperable due to large hydraulic fluid leaks on the valve operator.

Which ONE of the following is CORRECT regarding

1) real LCO entry and

2) all the applicable modes the ARVs are required to be OPERABLE?

LCO entry Applicable Modes A. required Modes 1, 2, and 3 B. required Modes 1 and 2 C NOT required Modes 1, 2, and 3 D. NOT required Modes 1 and 2 67

HL-I5RRO NRC Exam KIA 035 Steam Generator System (S/GS)

G2.2.40 Ability to apply Technical Specifications for a system.

K/A MATCH ANALYSIS The question presents a plausible scenario where a SG ARV is declared inoperable.

The candidate must determine if LCO entry is required and the applicable modes.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. LCO entry is NOT required but plausible, the student may think all ARVs are required to be operable. Only 3 ARVs are required to meet the LCD. ARVs are required in Modes 1 through 3.

B. Incorrect. LCD entry is NOT required but plausible the student may think all ARVs are required to be operable only 3 ARVs are required to meet the LCD. Part of choice about ARVs required in Modes 1 and 2 is incorrect.

C. Correct. 3 ARVs required in Modes 1, 2, and 3. LCD entry NOT required as the minimum requirements are met.

D. Incorrect. LCO entry is NOT required as only 3 ARVs are required to meet the LCO.

ARVs are required in Modes 1 through 3. Part of choice about ARVs required in Modes 1 and 2 is incorrect.

REFERENCES LO-LP-3921 1-02-006 from LOIT Exam Bank used as base for modification (included).

Tech Spec LCO 3.7.4 for Atmospheric Relief Valves VEGP learning objectives:

LD-PP-21101-10, Discuss the following concerning the Atmospheric Relief Valve(ARV).

d. Technical Specification requirements for operability.

LO-LP-3921 1-01, For any LCD in section 3.7 of Tech Specs, be able to:

a. State the LCD.

68

1. LO-LP-3921 1-02 006 The following conditions exist on Unit Two.

Unit Two Reactor Power is at 100%.

ARVs PV-3000 and PV-3030 have been declared inoperable because of large fluid leaks on the valve operators.

The hand pump for PV-3010 has been tagged out for two weeks awiating parts to repair the pump Based on this information, what Technical Specification actions are required?

A Restore one required ARV line to operable within 30 days B. Restore Both ARV lines to operable within 30 days C. Restore one required ARV line to operable within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> D. Restore at least two ARV lines to operable within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 9e $

Page: 1 10/16/2009

ARVs 3.7.4 3.7 PLANT SYSTEMS 3.7.4 Atmospheric Relief Valves (ARV5)

LCO 3.7.4 Three ARV lines shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required ARV line A.1 Restore required ARV line 30 days inoperable, to OPERABLE status.

B. Two or more required B.1 Restore at least two ARV 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ARV lines inoperable, lines to OPERABLE status.

C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND C.2 Be in MODE 4 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> Vogtle Units I and 2 3.7.4-1 Amendment No. 137 (Unit 1)

AmendmentNo. 116 (Unit2)

HL-15R RO NRC Exam

33. 038EK1.04 OO1!1/1!SGTR-REFLUX BOIL/MEM 3.1 I 3.3INEW/HL-15RNRC/ROITNT!DS Heat removal from the core by boiling, the steam flows to the S/G via the top of the hot legs, transfers heat to the secondary side water, condenses in the SIG tubes and returns to the vessel via the bottom of the hot legs.

This process is known as (1) and is (2) to occur during a Steam Generator Tube Rupture (SGTR) of a single tube.

(1) (2)

A. Reflux Cooling expected B Reflux Cooling NOT expected C. Natural Circulation expected D. Natural Circulation NOT expected K/A 038 Steam Generator Tube Rupture (SGTR)

EKI.04 Knowledge of the operational implications of the following concepts as they apply to the SGTR:

Reflux boiling.

K/A MATCH ANALYSIS The question gives the definition for reflux cooling which the student must recognize as reflux cooling versus natural circulation and whether reflux cooling is expected during a SGTR of a single tube (not).

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Reflux cooling is the correct term, however, it is not expected to occur during a SGTR, it is possible during some small break LOCAs and loss RHR while at midloop.

B. Correct. Reflux cooling is the correct term, it is not expected to occur during a SGTR, it is possible during some small break LOCAs and loss RHR while at midloop.

C. Incorrect. Natural circulation is plausible since this sounds like a circulation flow path, however, natural circulation as described in WOG backgrounds is with loops full. This is also not expected to occur during a SGTR, it is possible during some small break LOCAs and loss RHR while at midloop.

69

HL-15R RO NRC Exam flow path, however, natural circulation as described in WOG backgrounds is with loops full. This is also not expected to occur during a SGTR, it is possible during some small break LOCAs and loss RHR while at midloop.

REFERENCES V-LO-PP-36101, MCD Core Cooling Mechanisms, slides # 14, 15, 16, 17, 20, 26 & 27.

V-LO-HO-3731 1-001, Steam Generator Tube Rupture, no mention of Reflux Cooling since on a SGTR of a single tube, RCP trip criteria is not normally met nd SG tubes remain filled.

VEGP learning objectives:

LO-LP-36101-04, State three conditions required to establish natural circulation flow.

LO-LP-36101-14, State, in order of effectiveness, alternate methods of cooling available tot he operator during accident conditions.

LO-LP-36101-15, State the effectiveness of steam cooling versus water cooling during core heat removal.

70

Natural Circulation

• 7. List and describe three factors that can be used by the operator to enhance natural circulation flow.

Keep PZR pressure> 1920 PSIG keep-voids from forming in head.

PZR level> 25% water inventory for pressure control.

Steam generator NR level in NR of at least one> 10% maintain heat sink.

2/7/01 LO-PP-36101-00 14

Natural Circulation

+8. List and describe three separate indications of natural circulation flow.

RCS AT < full load AT.

RCS or CETCs constant or decreasing.

SG pressure constant or decreasing at RCS temperature rate while maintaining level.

RCS cold leg temperatures at saturation temperature for SG pressure 2/7/01 LO-PP-36101-00 15

Natural Circulation

49. State how the formation of noncondensible gases and/or steam can result in degradation of natural circulation flow.

If gases block flow path, i.e., loops, SG tubes, below vessel nozzles, they can retard/stop NC flow.

• Major concern is during a Large Break LOCA when the Accumulators inject Nitrogen, potential exist for the Nitrogen to accumulate in the SG U-tubes.

2/7/01 LOPP 3610100 16

Noncondensible Gases

• 10. List several (at least 6) sources of noncondensible gases in the RCS.

Dissolution of hydrogen Radiolysis PZR vapor space 2

Zr-H 0 reaction Accumulator nitrogen Fission gases/fuel pin helium Gases from injected water 2/7/01 LO-PP-36101-00 17

Pool Boiling

• 11. Define the term, Pool Boiling.

Stagnant water in the reactor vessel boils and then moves upward carrying away energy.

2/7/01 LO PP 36101 00 20

RefluxCoohng Cooling process that can occur during mid-loop following Loss of RHR.

• Heat is removed from core by boiling RCS water.

• Steam flows to SG Via top of hot legs.

• Steam condenses in SG tubes, transfers heat to SG secondary side water.

2/7/01

• Water from ohdGi1sed steam returns tO LO PP 36101 00 26 Rx via bottom of the same hot IcIe 2-

Rx Lk 5

HL-15R RO NRC Exam

34. 039A2.O1 OO1/2/1/MMN STEAM-LOCA FLOW/C/A -3.1 / 3.2/NEW/HL-15RNRC/RO/TNT/DS Given the following conditions:

- An RCS LOCA has occurred.

- Containment pressure is 15.5 psig and stable.

- A loss of offsite power to 13.8 kV switchgear NAA has occurred.

- RCS pressure is 1080 psig and stable.

- The crew is performing 19012-C, E-l .2 Post LOCA Cooldown and Depressurization.

Which ONE of the following describes the method that will be used to perform the cooldown of the RCS and the rate of the cooldown?

A SIG ARVs.

<100°F per hour.

B. S/GARVs.

Maximum rate attainable.

C. Steam Dumps.

<100°F per hour.

D. Steam Dumps.

Maximum rate attainable.

K/A 039 Main and Reheat Steam System (MRSS)

A2.01 Ability to (a) predict the impacts of the following malfunctions or operations on the MRSS; and (b) based on predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations; 71

HL-15R RO NRC Exam K/A MATCH ANALYSIS The question presents a plausible scenario where an RCS LOCA has occurred and the crew is cooling down the plant per 19012-C, Post LOCA Cooldown and Depressurization, the student must choose the correct method from plant conditions and the cooldown rate.

ANSWER I DISTRACTOR ANALYSIS A. Correct. SL pressure> 14.5 psig would cause an SLI and steam dumps would not be available. SIG ARVs would need to be used and per 19012-C, the RCS cooldown rate is limited to 100°F per hour.

B. Incorrect. SL pressure> 14.5 psig would cause an SLI and steam dumps would not be available. SIG ARVs would need to be used part is correct. However, the RCS cooldown rate is limited to 100°F per hour but this is plausible as other EOPs such as SGTR, Loss of All AC, LOHS, Inadequate Core Cooling use a maximum rate cooldown.

C. Incorrect. Steam dumps would not be available with Hi-2 Ctmt pressure present as this causes an SLI. The RCS cooldown rate limit of 100°F per hour is correct.

D. Incorrect. Steam dumps would not be available with Hi-2 Ctmt pressure present as this causes an SLI. the RCS cooldown rate is limited to 100°F per hour but this is plausible as other EOPs such as SGTR, Loss of All AC, LOHS, Inadequate Core Cooling use a maximum rate cooldown.

REFERENCES 19012-C, Post LOCA Cooldown and Depressurization, pages 9 and 11.

V-LO-PP-28103, Reactor Trip and ESFAS Signals, slide # 140 V-LO-PP-21201, Steam Dumps, slides # 91 (+ blow up of same slide).

VEGP learning objectives:

LO-LP-371 12-01, Using EOP 19012 as a guide, briefly describe how each step is accomplished.

72

Approved By Procedure Number Rev S.A. Phillips Vogtle Electric Generating Plant 19012-C 31.1 Date Approved Page Number ES - 1.2 POST-LOCA COOLDOWN AND 8/2/08 DEPRESSURIZATION 9of41 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE When the low steamline pressure Sl/SLI signal is blocked, main steamline isolation will occur if the high steam pressure rate setpoint is exceeded.

• 10 Check if low steamline pressure SI/SLI should be blocked:

_a. Steam DumpsAVAILABLE. - - _a. GotoStepl2.

pk:Ie h

b. PRZR pressure LESS THAN

- b. WHEN PRZR pressure is I.r4.y1 TJ 2000 PSIG. less than 2000 psig, and the high steam pressure rate alarms are clear, iveLes C-+--Q THEN block low steamline pressure SI/SLI by performing Step 10.d.

Go to Step 11.

c. High steam pressure rate alarms

- CLEAR.

d. Block low steam line pressure SI/SLI using the following:

. HS-40068

. HS-40069 Printed November 10, 2009 at 08:12

Approved By Procedure Number Rev S. A. Phillips Vogtle Electric Generating Plant 19012-C 31.1 DateApproved Page Number ES - 1.2 POST-LOCA COOLDOWN AND 8/2/08 DEPRESSURIZATION 11 of 41 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

• 12 Initiate RCS cooldown to cold shutdown:

a. Monitor shutdown margin by initiating 14005, SHUTDOWN MARGIN AND KEFF CALCULATIONS.

b. Maintain cooldown rate in RCS cold legs LESS THAN

, 5c(

100°F/HR. 4_.. MR

_c. Use RHR system if in service. y Hi (id3

d. Dump steam to Condenser from _d. Dump steam from intact intact SG(s) using Steam Dumps: SG(s) using SG ARV(s).

_1) Place PIC-507 in Manual.

_2) Match demand on SG Header Pressure Controller PIC-507 and SD demand meter Ul-500.

3) Transfer Steam Dumps to STM PRESS mode.

4) Open available Steam Dumps by slowly raising demand on PIC-507.

_13. Check RCS subcooling GREATER 13. Go to Step 36.

THAN 24°F [38°F ADVERSE].

Printed November 10, 2009 at 08:12

What actuation signals input to SLI?

-Low steam line pressure, I 2/3 channels < 585 psig, on 1/4 loops.

(may be manually blocked below P-il, also rate compensated)

I -High-2 containment pressure $/ /I 5LS 2/3 channels> 14.5 psig.

1Vc71 Lvif/e

-High steam line negative rate, 2/3 channels on 1/4 loops, 100 psig with a 50 second time constant.

(Requires that P-li be present and SI/SLI on low steam line pressure manually blocked).

V-LO-PP-281 03-6.2 140

Steam Dumps OBJECTIVE LO-PP-21201-12 If no circulating water pumps are running or insufficient vacuum exists in the main condenser, dumping steam into the condenser can cause an overpressure condition which can damage the condenser. To protect against this, a permissive circuit prevents arming of the steam dumps. As can be seen in the figures below, the permissive circuit is composed of contacts which involve condenser vacuum and the circulating water pump breakers and their associated switchgear voltages (at 0 volts for 5.75 seconds). The condenser vacuum contact will be closed as long as two separate condenser pressure transmitters provide a signal indicating that at least 24.92 inches of mercury vacuum exists in the condenser. Each of the circulating water pump breaker contacts will be closed as long as the breaker for the associated pump is closed and the associated switchgear voltages are present. All of the above contacts are arranged so that the condenser must have at least 24.92 inches of mercury vacuum and at least one circulating water pump breaker closed with voltage applied. This will energize the permissive relay which closes the permissive contact in the arming circuit to permit arming of the steam dump valves. The permissive relay then energizes a C-9 Condenser Available permissive status light.

,iL/ /Vth eee ed Vce >$

p 4 ,z.., c tJ V-LO-PP-21 201 Rev-02 (e ii

227 device 227 device Undervoltage Undervoltage Circulating Water Pump Running Main Condenser Main Condenser 252 contact Vacuum Vacuum

>24.92 in Hg >24.92 in Hg PERMISBD/E IRODODDLOOI if n

L!!L YJU ADWED ORDD

,Dt, 1

Tt F

frDD.tYpRR TRAIUDD L_i!a2?_ D DPDUDDUDUD DDDDRDD UOQZO PAD D LJ!. 1&! MDDRR I DR P P*GDWPrD 0L014V .REARDDDM 4 1PWDPADR TflDDR STM DUMP PDDOPA TRDPMWR D4DDU IDITLIC PIP P%?MDED IjWii3Dk PR IMIUDfl%T PDRP4PRTP DPUPUU RRDPRPDPU PTDDDDTUP L_f!_ RPRRRP URRTDTD

. l \y a a as atm ke5 QtD a a a a LZ r

HL-15R RO NRC Exam

35. 05 1AA1 .04 002/1/2/LOSS VACUUM-ROD POST/C/A 2.5 / 2.5/NEW/HL-1 5R NRC/RO/DS/TNT The plant is at 100% power in when the following valid annunciator is received following a circulating water pump trip:

TURB CNDSR LO VAC

- The SS enters AOP 18013-C, Rapid Power Reduction.

- The Unit Operator rapidly reduces turbine load.

- Tave is 3.8°F higher than Tref.

Wwhich one of the following is correct regarding:

1) AOP-1 8013-C, power reduction target, and

2) preferred operation of the control rods in accordance with 18013-C?

Target Control Rods A. low vacuum alarm clear manually insert at 48 steps per minute B low vacuum alarm clear automatically insert based on Tavg-Tref deviation C. 20% rated thermal power manually insert at 48 steps per minute D. 20% rated thermal power automatically insert based on Tavg-Tref deviation K/A 051 Loss of Condenser Vacuum AAI.04 Ability operate and I or monitor the following as they apply to the Loss of Condenser Vacuum.

Rod position.

K/A MATCH ANALYSIS The question presents a plausible scenario where a low condenser vacuum alarm is received. The candidate must know the effect of this alarm on steam dump operation when C-7 is received and whether or not control rods are operated in auto versus manual.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Per 18013-C, the target should be a power reduction until the low vacuum 73

HL-15R RO NRC Exam plausible if auto does not work and the AOP has been recently changed to make automatic operation the desired method.

B. Correct. The target per 18013-C is to reduce power until low vacuum alarm clears.

Rods are desired to be operated in auto per 18013-C which is a recent change.

C. Incorrect. 20% is used as a target since this is the power level 18013-C directs to reduce power and trip the reactor, therefore it is plausible. Manual is plausible if auto does not work and the AOP has been recently changed to make automatic operation the desired method.

D. Incorrect. 20% is used as a target since this is the power level 18013-C directs to reduce power and trip the reactor, therefore it is plausible. Rods are desired to be operated in auto per 18013-C which is a recent change.

REFERENCES AOP-1 8013-C, Rapid Power Reduction V-LO-PP-27101, Rod Control System, slides 40 and 41 (included).

VEGP learning objectives:

LO-PP-60331-01, Describe the entry and exit conditions associated with using the Rapid Power Reduction AOP.

LO-PP-60331-03, Describe the type conditions that might warrant the use of the Rapid Power Reduction AOP.

V-LO-PP-27101-07, Describe the operation of the Rod Control System with the bank selector switch in the manual or automatic position. Include the following.

a. Rod Speed

c. Input signals 74

Approved By Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant 18013-C 7 DateApproved PaqeNumber 3/24/09 RAPID POWER REDUCTION 1 of 11 ABNORMAL OPERATING PROCEDURE CONTINUOUS USE PURPOSE This procedure provides instructions when plant conditions require a rapid load reduction or plant shutdown in a controlled manner in the judgment of the SS.

Entry Condition Target Approx. Time @ 3-5%Imin 1701 5-D05 MFPT High Vibrations <70% RTP 5-8 minutes 1701 5-E01 1 701 9-B04 Condenser Low Vacuum Vacuum >22.42 Hg and STABLE 18025-C or Circ Water Pump Trip or RISING or Loss of Utility Water 18009-C SG Tube Leak (75 gpd with an <50% RTP within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 10-17 minutes ROC 30 gpdlhr) 18009-C SG Tube Leak (5 gpm) 20% RTP within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> & trip 16-27 minutes reactor 18039-C Confirmed Loose Part 20% RTP quickly 16-27 minutes SS determination based on As determined by the SS plant conditions MAJOR ACTIONS

• Perform Pre-job Brief.

• Perform rapid power reduction.

Printed October 8, 2009 at 16:26

Rod Control System TAVG CONTROL UNIT V-LO-PP.27101 Rev-2.O 40 Jef?es The average reactor coolant temperature mismatch channel develops a temperature error signal by comparing actual auctioneered high reactor coolant average temperature, T-avg, to a reference reactor I T 5Aurfr /

coolant average temperature, T-ref. T-ref represents the reactor coolant average temperature demanded by the turbine. T-ref is developed by sensing turbine first-stage pressure. The pressure signal, which represents the turbine load, is sent to a T-ref programmer. The T-ref programmer generates a linearly increasing T-ref signal of 557 deg F at 0% turbine load up to about 587.5 deg F at 100% turbine load. Notice that the rod control Tref program is different from the normal Tavg program. It is adjusted to account for the auctioneered Hi Tavg reading being higher than the average of all loops Tavg. This makes Tref and auctioneered Hi Tavg match when the average of Tavg is on program. The rod control Tref is adjusted as needed to maintain a match with the auctioneered Hi Tavg average temperature signals. The lead-lag unit accomplishes two purposes; the lag portion of the circuit filters noise and the lead portion boosts the rate of change of the temperature signal to compensate for the delay of the RTDs in seeing an actual core temperature change due to the transit time of the RCS coolant as well as the delay associated with RTD response time. This output signal is then sent to the summing unit for comparison to the T-ref signal. The T avg signal and the T-ref signal are also directed to a two-pen meter on the main control board (QMCB) for indication to the operator and to a comparator (Bistable) for a T-avg-T-ref deviation alarm if their difference is greater than 3 deg F.

V-LO-PP-27101 Rev2.O 40

Rod Control System TAVG CONTROL UNIT I PecA

/

5 S V-LO-PP-27101 Rv-2.O 41 /)/\

( 5 i_ fc_ ç To improve the T-avg control units response during transients, a rate of power mismatch channel also develops a temperature input signal to the summing unit. This circuit compares the rate of change of the difference between reactor power and turbine power. When the difference is constant, the circuit provides no input signal to the summing unit so that the average reactor coolant temperature channel makes the fine adjustments to T-avg. The input signals to this channel are reactor power and turbine load Reactor power is developed by auctioneering the actual nuclear power signals from the four nuclear power channels so that the highest signal is used as the input to a rate comparator. Turbine power is developed from turbine first-stage pressure. The rates of change between reactor power and turbine power are compared in the rate comparator circuit. When the rates of change differ, an error signal is produced and sent to a nonlinear gain unit.

The nonlinear gain unit converts the power mismatch signal to a temperature signal. When the power mismatch signal is greater than 2 percent, it is multiplied by 1.5 deg F/percent to produce the temperature signal. When the power mismatch signal is equal to or less than 2 percent, it is multiplied by O.3F deg/percent to produce the temperature signal. The output of this unit is then sent to the variable gain unit.

The variable gain unit compensates for reactor gain. This is needed because the reactivity changes at low power levels have smaller effects on nuclear power than at higher levels. The temperature signal input is multiplied by 2 when turbine load is equal to or less than 50 percent, Between 50 percent and 100 percent, the temperature signal input is multiplied by one divided by percent turbine power (gain is inversely proportional to turbine power). This adjusted temperature signal is then sent to the summing unit.

V-LO-PP-27101 Rev-2.O 41

Approved By I Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant 1801 3-C 7 Date Approved Page Number RAPID POWER REDUCTION 3/24/09 3 of 11 SHUTDOWN BRIEFING METHOD

• Auto rod control should be used.

• Reduce Turbine Load at approximately 3% RTP per minute (approx 36 MWe) up to 5% RTP (approx 60 MWe).

• Borate considering the calculations from the reactivity briefing sheet and BEACON.

• Maintain AFD within the doghouse.

• SS (or SRO designee) - Maintain supervisory oversight.

• AW rod withdrawals will be approved by the SS.

• Approval for each reactivity manipulation is not necessary as long as manipulations are made within the boundaries established in this briefing (i.e. turbine load adjustment up to 60 MWe, etc.).

• A crew update should be performed at approximately every 100 MWe power change.

• If manpower is available, peer checks should be used for all reactivity changes.

OPERATIONAL LIMITS

• Maintain TAVG within +/-6°F of TREE. If TAVGITREF mismatch >6°F and not trending toward a matched condition or if TAVG 551°F, then trip the reactor.

• If load reduction due to a loss of vacuum, every effort should be made to maintain the steam dumps closed (Permissive C-9 24.92 Hg).

INDUSTRY OE

• Shift supervision must maintain effective oversight and exercise conservative decision making.

• Correction of significant RCS TAVG deviations should only be via secondary plant control manipulations and not prim ary plant control manipulations (i.e., do not withdraw control rods or dilute).

Printed January 5, 2010 at 19:37

Approved By Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant 1 8013-C 7 Date Approved Page Number RAPID POWER REDUCTION 3/24/09 4 of 11 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

1. Perform SHUTDOWN BRIEFING.

_2. Verify rods in AUTO.

_3. Reduce Turbine Load at the desired rate up to 5%/mm (60 MWE/min).

4. Borate as necessary by initiating 13009, CVCS REACTOR MAKEUP CONTROL SYSTEM.

5. Initiate the Continuous Actions Page.

_*6. Check desired ramp rate LESS-

• 6. IF conditions warrant a turbine load THAN OR EQUAL TO 5%IMIN. rate greater than 5%/mm, THEN perform the following:

a. Trip the reactor.

b. Go to 19000-C, E-0 REACTOR TRIP OR SAFETY INJECTION.

• 7 Maintain Tavg within 6°F of Tref:

a. Monitor Tavg/Tref deviation (UT-0495).

b. Verify rods inserting as required. _b. Manual rod control should be used with insertions of up to 5 steps at a time.

c. Energize Pressurizer back-up heaters as necessary.

Printed January 5, 2010 at 19:37

HL15R RO NRC Exam

36. 054AK1.O1 001/1/1/LOSS FW-MF LINE BRK/C/A -4.1 / 4.3/NEW/HL-15RNRC/RO/TNT/DS Given the following plant conditions:

- The unit is at 100% power.

- A Main Feedwater line break occurs at the piping connection to SG # 3.

RCS temperature will prior to the reactor trip and SG # 3 pressure will RCS temperature response SG # 3 pressure response A. rise stabilize when an SLI occurs B rise continue to depressurize after a FWI occurs C. lower stabilize when an SLI occurs D. lower continue to depressurize after a FWI occurs 75

HL-15R RO NRC Exam K/A 054 Loss of Main Feedwater (MFW)

AKI.01 Knowledge of the operational implications of the following concepts as they apply to the Loss of Main Feedwater (MFW):

MFW line break depressurizes the SIG (similar to a steam line break).

K/A MATCH ANALYSIS The question presents a plausible scenario where a Feedwater line break occurs at the connection to SG # 3. The student must be able to differentiate the RCS temperature response prior to a reactor trip and whether or not a FWI or SLI will isolate the break.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Classic FW line break response is for RCS temperature to be stable or rise in response. An SLI will not isolate the break and SG pressure would continue to depressurize following an SLI as inventory is lost via the feedwater break.

B. Correct. Classic FW line break response is for RCS temperature to be stable or rise in response. A FWI will not isolate the break and SG pressure would continue to depressurize following a FWI as inventory is lost via the feedwater break.

C. Incorrect. RCS temperature should stabilize or rise in response to a FW line break.

RCS temperature lowering is the response to a SL break. An SLI will not isolate the break and SG pressure would continue to depressurize following an SLI as inventory is lost via the feedwater break.

D. Incorrect. RCS temperature should stabilize or rise in response to a FW line break.

RCS temperature lowering is the response to a SL break. A FWI will not isolate the break and SG pressure would continue to depressurize following a FWI as inventory is lost via the feedwater break.

REFERENCES V-LO-HO-37121-OO1, Faulted Steam Generator Isolation, page # 8 (included).

VEGP learning objectives:

LO-LP-317121-05, Describe the plant response to the following conditions.

a. Steam line break versus feed line break.

d. Feed break inside last check valve versus feed break outside last check valve.

76

Feedline Breaks For an intermediate feedline break in which the control systems are incapable of compensating for the loss of flow, the secondary side would experience a slowly decreasing steam generator water level in at least one steam generator. A slowly increasing primary average temperature prior to reactor trip may occur due to the loss of main feedwater and degraded steam generator heat transfer. The transient is eventually terminated by manual reactor trip or when the low low level trip setpoint is reached in any one steam generator. This results in a reactor trip and auxiliary feedwater initiation. A subsequent turbine trip occurs due to reactor trip.

If the break occurs downstream of the main feedline check valves, all steam generators continue to experience a reverse blow down through the steam generator associated with the faulted loop until a low steamline pressure setpoint is attained resulting in a safety injection and steamline and feedline isolation.

The faulted steam generator will then blow down until atmospheric pressure is reached.

If the break occurs upstream of the feedline check valves, the feedwater spillage is terminated and the auxiliary feedwater system is sufficient to mitigate the consequences of the resultant loss of normal feedwater transient.

The system parameter trends that are used to identify a faulted SG are an uncontrolled pressure decrease in at least one steamline or a SG that is completely depressurized. Other symptoms include decreasing water level in at least one steam generator and slowly rising primary system average temperature prior to reactor trip.

For either of the above transients, if the break occurs inside containment, an increasing containment temperature and/or pressure indication could be observed. If the break occurs outside containment, audible or visual indications may assist the operator in diagnosing the transient.

Large Secondary Break The least likely and most severe of the postulated loss of secondary coolant events is the double ended break.

Main Steamline Break For the double ended main steamline break, an immediate decrease in pressure in at least one steamline occurs depending upon the location of the break. The low steam line pressure setpoint is reached which VLOfr°I 8

HL-15R RO NRC Exam

37. 055EG2.4.02 001/1/1/LOSS ALL AC-FOP ENTR/C/A -4.5 / 4.6/NEW/HL-15R NRC/RO/DS/TNT Which one of the following choices list conditions which require entry into EOP 19100-C, ECA-0.0 Loss of All AC Power with Unit 1 at 100% power?

A 2 of the 3 white potential lights for both 1AAO2 and 1BAO3 extinguish.

1AAO2 and 1BAO3 breaker position indication lights remain lit.

B. I of the 3 white potential lights for both 1AAO2 and 1BAO3 extinguish.

1AAO2 and 1BAO3 breaker position indication lights extinguish.

C. 2 of the 3 white potential lights for both 1 NAO1, 1 NAO4 and 1 NAO5 extinguish.

1 NAO1, 1 NAO4 and 1 NAO5 breaker position indication lights remain lit.

D. 1 of the 3 white potential lights for 1 NAO1, 1 NAO4 and 1 NAO5 extinguish.

1 NAO1, 1 NAO4 and 1 NAO5 breaker position indication lights extinguish.

K/A 055 Station Blackout EG2A.02 Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions:

K/A MATCH ANALYSIS The question requires the student to interpret the electrical board indications to determine when a station blackout has occurred and entry into ECA-0.0 is required.

ANSWER I DISTRACTOR ANALYSIS A. Correct. 2 of the white lights are fed by the 41 60V AC busses through stepdown transformers. The 3rd light indicates DC control power for the switchgear. DC control power also provides breaker position indications. For the indications given, this would be a loss of AC power to both AC emergency busses meeting the ECA-0.0 symptoms.

B. Incorrect. The data given for this choice indicates a loss of DC control power to the emergency busses. This would not require entry into ECA-0.0 since all breakers would remain in their previous positions.

77

HL-15R RO NRC Exam C. Incorrect. This choice provides indications for a loss of AC power to the non-i E 41 60V AC busses. This does not meet ECA-0.0 entry conditions.

D. Incorrect. This choice provides indications of a loss of DC control power to the non-i E 41 60V AC switchgear. This does not meet ECA-0.0 entry conditions.

REFERENCES 19100-C, ECA-O.O Loss of All AC Power page 1 18034-1, Loss of Class 1 E 1 25V DC Power page 13 LO-LP-60329, Loss of Class ie 125V DC page #4 One Line 1X3D-AA-DO3A, 1BAO3 light indications.

VEGP learning objectives:

V-LO-LP-37031 -01:

Define loss of all AC power condition. Explain its immediate implications for operation of plant equipment.

V-LO-PP-Oi 101-05:

Describe how a failure of DC control power affects the electrical distribution system and its components.

78

Approved By Procedure Number Rev C. S. Waidrup Vogtle Electric Generating Plant 19100-C 33.2 DateApproved Page Number 2/27/09 ECA-O.O LOSS OF ALL AC POWER 1 of 51 EMERGENCY OPERATING PROCEDURE CONTINUOUS USE PURPOSE This procedure provides actions to respond to a loss of all AC power. (Applicable in Modes 1, 2,3,4)

SYMPTOMS/ENTRY CONDITIONS The symptoms are:

• Both emergency AC buses are de-energized.

The entry conditions are:

• 19000-C, E-0 REACTOR TRIP OR SAFETY INJECTION MAJOR ACTIONS

• Check Plant Conditions

• Restore AC Power

• Maintain Plant Conditions for Optimal Recovery

• Evaluate Energized AC Emergency Bus

• Select Recovery Guideline After AC Power Restoration Printed November 12, 2009 at 13:02

Approved By Procedure Number Rev S. A. Phillips Vogtle Electric Generating Plant 18034-1 10 DateApproved Page Number LOSS OF CLASS IE 125V DC POWER 8/2/08 13 of 82 ATTACHMENT A Sheet 1 of 23 LOSS OF 125V DC BUS 1AD1 EQUIPMENT RESPONSE DUE TO LOSS OF TRAIN A 125V DC POWER NOTE Feeder Breakers must be locally controlled in the event the transfer to an alternate power supply is required. IF DG1A is not running, it may not be selected as an alternate power source.

• Main Feedwater Isolation, Bypass Feedwater Isolation, and Bypass Feedwater Regulation Valves close resulting in Feedwater Isolation.

• Main Steam Isolation and Bypass Steam Isolation Train A Valves close resulting in steamline isolation.

• Above PlO, Reactor and Turbine trip occurs from loss of main feedwater.

• Below P10, Reactor trip occurs from Intermediate Range Instrumentation.

• Control Power is lost to 1AAO2, 1ABO4, 1ABO5, and 1AB15 SWGR Breakers.

• DG1A control power to Generator Control Panel PDG1 and Engine Control Panel PDG2 is lost rendering the DG inoperable; if running, it will fail as is with a loss of electrical protective trips, frequency, and voltage control. Due to loss of power to the Low Speed Relay, the generator space, Engine Lube Oil and Jacket Water Heaters and Lube Oil and Jacket Water Keep-Warm Pumps will come on.

• Loss of Train A DG AUTO sequencer reset.

• Power to Inverters 1AD1I1 and 1AD1 Ill is lost causing 120V AC Vital Busses 1AY1A and 1AY2A to de-energize.

• Instrument Air Containment Isolation Valve 1-HV-9378 closes resulting in loss of instrument air inside Containment.

• Power To Isolation Panel 1ACQIP1 is lost rendering the annunciators in Train A inoperable.

• Pressurizer PORV 1-PV-455A fails closed.

• TDAFW Steam Supply 1-HV-3019 fails as is.

Printed November 12, 2009 at 13:27

LO-LP-60329-06 Ill. LESSON OUTLINE: NOTES INTRODUCTION A. This procedure provides the actions to be followed in the event that power is lost to one of the 125V DC vital busses B. Present lesson objectives PRESENTATION A. Present this lesson using the latest revision of AOP 18034 B. Procedure consists of four subsections: Objective 4

1. Loss of 125V DC Bus IADI

a. Symptoms

1) Loss of 1AY1A and 1AY2A will be obvious to the operator due to all Channel I bistable lights lite up and sequencer trouble.

2) Loss of control power lights on 1AAO2, 1ABO4, 1ABO5, and 1AB15.

3) Train ASLI

4) TrainAFWl

5) Loss of voltage on 1AD1

2. Loss of 125V DC Bus IBDI

a. Symptoms

1) Loss of 1 BY1 B and 1 BY2B causing all Channel II bistable lights to lite and sequencer trouble alarm.

2) Loss of control power lights on 1 BAO3, 1BBO6, 1BBO7, and 1BB16.

3) Train B SLI

4) Train B FWI

5) Loss of Voltage on IBD1 4

H HEY ARE IERELY LOANED AND ON THE BORROWER EXPRESS AGREEMENT THAT THEYI WILL NOT BE REPRODUCED, COPIED, LOANED, EXHIBITED, OR USED EXCEPT IN THE LIMITED WAY AND PRIVATE USE PERMITTED BY ANY W[TTEN CONSENT GIVE!

TO SYNCHRONIZING OIL 1X30-AA-BO5A TO DIESEL GEN WHM OWO 1X30-AA-003B INCOMING FROM RESERVE AUXILIARY XFMR 1NXR8 1 -1801 -T3-ORB O/L IX3O-AA-O2A TO SWGR 1AAOZ INCOMING US

HL-15R RO NRC Exam

38. 055K3.01 001/2/2/CARS-MAIN CONDENSER/C/A -2.5 / 2.5/NEW/HL-15RNRC/RO/TNT/DS Initial conditions:

- The unit is at full power

- The time is 1200.

Current conditions:

- The time is 2400.

- The UO notes Circulating Water temperature has lowered by 6°F since 1200.

Which ONE of the following is CORRECT regarding the effect of lowering Circulating Water temperature on Main Condenser pressure and Main Turbine MW output?

Main Condenser Pressure (psia) Main Turbine MW Output A. Rises Rises Lowers Rises C. Rises Lowers D. Lowers Lowers 79

HL-15R RO NRC Exam KIA 055 Condenser Air Removal System (CARS):

K3.01 Knowledge of the effect that a loss or malfunction of the CARS will have on the following:

Main Condenser K/A MATCH ANALYSIS The question presents a plausible scenario where Circulating Water pump temperature lowers from day shift to night shift. The student must determine the effect on Main Condenser Vacuum in psia and Main Turbine MW output.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Colder CW temperature would cause vacuum to get better (psia lowers) resulting in more MW output. Plausible the student could invert either parameter.

B. Correct. Colder CW temperature would cause vacuum to get better (psia lowers) resulting in more MW output.

C. Incorrect. Colder CW temperature would cause vacuum to get better (psia lowers) resulting in more MW output. Plausible the student could invert either parameter.

D. Incorrect. Colder CW temperature would cause vacuum to get better (psia lowers) resulting in more MW output. Plausible the student could invert either parameter.

REFERENCES V-LO-PP-18101, Condensate and Feedwater, slide # 12 VEGP learning objectives:

LO-LP-18101-02, Discuss how the following will impact Main Condenser vacuum.

a. Circulating Water pump temperature.

80

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HL-15R RO NRC Exam

39. 056AK3.01 002/1/1/LOSS OFFSITE PWR-SEQ/C/A 3.5/3.9/NEWIHL-15R NRC/RO/TNT/DS Initial conditions:

- ACCW pump # 1 running A plant event results in the both ESF Sequencers running, final plant conditions are:

- Both ACCW pumps running.

- The last load starts at 50.5 seconds.

Which ONE of the following is the CORRECT initiating event?

A.SI BUN C. SI followed by a U/V D. U/V concurrent with an SI KIA 056 Loss of Offsite Power AK3.01 Knowledge of the reasons for the following responses as they apply to the Loss of Offsite Power:

Order and time to intiation of power for the load sequencer.

K/A MATCH ANALYSIS The question requires the student to correctly identify from given plant conditions the type of ESF sequence which initiated.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. An SI only sequence does not start an ACCW pump but the pump that is running would remain running. Two ACCW pumps running would rule out the SI sequence. In addition, the SI sequence completes in 30.5 seconds where a UV sequence runs for 50.5 seconds to allow the Containment Cooler starts to be staggered.

B. Correct. U/V sequence runs in 50.5 seconds and a single train UV or dual train UV will end up with both ACCW pumps running.

C. Incorrect. An SI sequence would not start the second ACCW pump, when the 81

HL-15R RO NRC Exam subsequent UV occurs, following load shed, the SI sequence prevails and would block start of either or both ACCW pumps whether, no ACCW pumps would be running. The SI sequence would also be the predominant sequence and takes 30.5 seconds to complete.

D. Incorrect. A simultaneous SI and UV sequence occurring would result in the SI sequence running as it is the predominant sequence. no ACCW pumps would be running. The SI sequence also takes 30.5 seconds to complete.

REFERENCES V-LO-TX-28201, Sequencer pages 13, 14, 16, 17, and 18.

V-LO-PP-04101, ACCW slides 15, 17, and 19.

VEGP learning objectives:

LO-PP-28201 -03:

Describe sequencer operation, including load shedding, load sequencing, and diesel generator operation under the following conditions:

a. Undervoltage (UV)

b. Safety Injection

c. UVfollowedbySl

d. SI followed by UV 82

V- 2 LO-T

° ) )-

The U/v Schemes are as follows:

First level voltage- (INSTANTANEOUS Trip) <71.5 %(2975 VAC) for >0.8 sec.

Coincidence is 2/4.

Second level voltage- DEGRADED(Trip) <90  %- (3746 VAC) for >20 sec.

Coincidence is 2/4.

Third level voltage- (Alarm only) <93.1 % (3873 VAC) for >10 sec.

Coincidence is 2/4.

The operation of the Vogtle Safety Features Sequencer System is automatic. Upon receiving a bus undervoltage (U/V) signal, the sequencer will automatically shed loads from the power bus, provide a start signal to the diesel generator, and, when the diesel generator is on line, sequentially return selected loads to the bus.

On an undervoltage actuation (First or second Level) the following actions are performed by the sequencer: (There is figure 1, Sequencer Manual Test Panel at the end of this section that can be used with the discussions of Sequencer operation)

Time=0 Emergency start signal sent to Diesel Generator.

sec Load shed occurs:

1 sec trip signal sent to the Normal and Emergency feeder breakers to the 4160 VAC bus.

1 sec trip signal sent to pump breakers to the 4160 VAC bus.

1 sec trip signal sent to 480 VAC secondary side (low side)

Switchgear breakers for all 1E and Non 1E loads.

Auto/Manual Block circuit is enabled.

u/v non sequenced load relays are energized to actuate those loads. This same signal generates the UNDERVOLTAGE light on the sequencer panel.

Signal sent to Loss of Power (LOP) monitor circuit.

Reset and stop signal sent to ATI subsystem. ATI step counter is reset to 00 and stopped.

Reset and inhibit signal sent to Manual test circuits.

Lights U/V SIGNAL (red), U/V RELAYS ACTUATED (red), BLOCK AUTO/MNL SIG (red), SEQ LOGIC FAILURE (amber), TINDERVOLTAGE (amber) generated on Sequencer panel.

Sequencer Trouble and AO2 (BAO3)SWGR Trouble alarms received in the control room on the QEAB.

Time=0.5 Sends DO Breaker Auto Closure permissive to DG output sec breaker closure circuit if not blocked by LOP monitor circuit.

13 Revision 2.0

The U/V Schemes are as follows:

First level voltage- (INSTANTANEOUS Trip) <71.5 %(2975 VAC) for >0.8 sec.

Coincidence is 2/4.

Second level voltage- DEGRDED(Trip) <90 % (3746 VAC) for >20 sec.

Coincidence is 2/4.

Third level voltage- (Alarm only) <93.1 % (3873 VAC) for >10 sec.

Coincidence is 2/4.

The operation of the Vogtle Safety Features Sequencer System is automatic. Upon receiving a bus undervoltage (U/V) signal, the sequencer will automatically shed loads from the power bus, provide a start signal to the diesel generator, and, when the diesel generator is on line, sequentially return selected loads to the bus.

On an undervoltage actuation (First or second Level) the following actions are performed by the sequencer: (There is figure 1, Sequencer Manual Test Panel at the end of this section that can be used with the discussions of Sequencer operation)

Time=0 Emergency start signal sent to Diesel Generator.

sec Load shed occurs:

1 sec trip signal sent to the Normal and Emergency feeder breakers to the 4160 VAC bus.

1 sec trip signal sent to pump breakers to the 4160 VAC bus.

1 sec trip signal sent to 480 VAC secondary side (low side)

Switchgear breakers for all 1E and Non 1E loads.

Auto/Manual Block circuit is enabled.

U/v non sequenced load relays are energized to actuate those loads. This same signal generates the UNDERVOLTAGE light on the sequencer panel.

Signal sent to Loss of Power (LOP) monitor circuit.

Reset and stop signal sent to ATI subsystem. ATI step counter is reset to 00 and stopped.

Reset and inhibit signal sent to Manual test circuits.

Lights U/V SIGNAL (red), U/V RELAYS ACTUATED (red), BLOCK AUTO/MNL SIG (red), SEQ LOGIC FAILURE (amber), TflDERVOLTAGE (amber) generated on Sequencer panel.

Sequencer Trouble and AAO2 (BAO3)SWGR Trouble alarms received in the control room on the QEAB.

Time=0.5 Sends DG Breaker Auto Closure permissive to DG output sec breaker closure circuit if not blocked by LOP monitor circuit.

13 Revision 2.0

Time=6.0 to When DG ready to load, DO output breaker closes. D-G BRKR 11.5 secs CLOSED (red) light generated on Sequencer panel. Sequencer elapsed time display begins running.

Brkr CL SEQ STEPS INDICATION (red) for steps 1A-9A and lC-9C will

+0.5 to begin flashing in the intervals specified in the list 30.5 secs below as the components are sequence on.

UV LOAD SEQUENCE Train A only (Train B loads are similar)

TIME LOAD 0.5 secs CCP A, 480 VAC Secondary side feeder breakers 5.5 sec NONE 10.5 sec NBO1 (Stub bus secondary side feeder breaker closes) 15.5 sec ACCW Pump 1 20.5 sec CCW Pumps 1 and 3 MDAFW Pump A 25.5 sec NSCW Pumps 1 and 3 CCW Pump 5 (if CCW Pumps 1 or 3 breaker did not close) 30.5 sec CTMT Cooling Units 5 and 6 (Fast Speed)

NSCW pump 5 (if NSCW Pumps 1 or 3 breaker did not close)

CTMT Cooling Units 1 and 2 start contact closed (Fast Speed) 5O L 5ecvd CTMT cooling units must not all be started at the same acL5 10 time to prevent bus voltage transients. Analysis has shown that if all four were allowed to be simult us DG voltage could dro below 80 percent. The sequencer es a CTMT cooling unit start signal at the 30.5 second step. Coolers 1, 2, 7, and 8 start at 50.5 seconds due to an additional time delay of 20 seconds by an agastat time delay relay in the auto-start circuit.

This is for a UV condition only.

Time=32 SEQ STEPS INDICATIO re flashing lights extinguish.

secs SAF EQPT FAIL TO START (amber) light to indicate that Cnmt Coolers 1 and 2 have not started. Audible alarm is sounded on the sequencer panel. Alarm generated on QEAB.

ATI stop removed and ATI restarted Time =36 BLOCK AUTO/MNL SIC (red) light extinguishes.

CTMT Cooling Units 1 and 2 start (Fast Speed) This is not a sequencer function but internal to the start logic of the cooler high speed motors SAF EQPT FAIL TO START (amber) light extinguishes. QEAB alarm clears.

14 Revision 2.0

Sequencer Operation in Response to an SI Condition.

The Sequencer is also awaiting an SI signal. The following diagram shows how the SI signal is processed from the SSPS to the plant loads.

SI Signal Processing Block Diagram Trip Stub Bus On receipt of an SI signal, the following actions are performed by the sequencer:

Time=0 sec Emergency start signal sent to Diesel Generator.

Auto/Manual Block circuit is enabled.

ATI step counter is reset to 00 and stopped.

Reset and inhibit signal sent to Manual test circuits.

Non sequenced SI maintained relays actuated.

Non sequenced SI Momentary relays actuated for 1 sec Lights SI SIGNAL (red), SI MAIN RELAYS ACTUATED (amber),

BLOCK AUTO/MNL SIG (red), SI MOM RELAYS ACTUATED (amber) generated on Sequencer panel.

Time0 .1 SI sequence timing starts. Sequencer elapsed time display secs begins running. This is set at 90 millisecs.

Time=1.0 Light SI MOM RELAYS ACTUATED (amber) extinguishes.

sec Time=0 .9 Lights SEQ STEPS INDICATION (red) will begin flashing in to 30.5 the intervals specified in the list below. Steps 1-9, 1A-secs 9A, 1B-9B, and 1C-9C steps are sequenced on.

16 Revision 2.0

SI LOAD SEQUENCE Train A only (Train B loads are similar)

TIME LOAD 0.5 sec CCP A 5.5 sec SIP A

)

10.5 sec RHR pump A CC rCC\

15.5 sec Containment Spray Pump A (W/ CSAS) 20.5 sec CCW pumps 1 and 3 /1 /

MDAFW Pump A

/V (i/I 25.5 sec NSCW pumps 1 and 3 CCW 5 (if CCW Pumps 1 or 3 breaker did not close) 9 C

30.5 sec CTMT Cooling Units 1, 2, 5 and 6 (Slow Speed)

NSCW 5 (if NSCW Pumps 1 or 3 breaker did not close)

Note that the Containment spray pump sequence is unique. At 15.5 secs the sequencer will send a 1 sec start signal. IF the CNMT Spray Actuation slave relay is energized the pump will start. This one sec start signal is sent on all the following steps to start CNMT spray if the actuation signal occurs. Therefore it is possible for the CNMT Spray pump to start at 15.5, 20.5, 25.5, or 30.5 secs depending on when a CNMT Spray actuation signal is received.

Time=6.0 When DG ready to load, DG READY FOR LOADING (red) light to 11.5 generated on Sequencer panel.

secs Time=32 SEQ STEPS INDICATION (red) for lB-9B and lC-9C steps secs flashing lights extinguish. The 1-9 and lA-9A steps will continue to flash until the SI signal is reset.

Time=36 BLOCK AUTO/MNL SIG (red) light extinguishes.

secs Note that if CCW pumps 5 or 6 or NSCW pumps 5 or 6 are in service before an SI signal occurs, then all three pumps will be running in those trains after the SI sequence is complete.

ATI must be manually reset after the SI signal is reset.

17 Revision 2.0

Sequencer Operation in Response to an SI Condition with a U/V Condition.

There are five separate com)Dinations to consider on operation with an SI and an u/v condition. They are:

SI signal and u/v simultaneously SI signal following u/v (before sequencing is complete)

SI signal following u/V (after sequencing is complete) u/v following SI signal (before SI is reset) u/v following SI signal (after SI is reset)

The general rule of operation in these conditions is the sequencer operation will be a combination of the U/v and SI sequences. By understanding priority system of the sequencer, each of the above combinations can be evaluated. The U/V sequence will predominate until the ESF bus is energized. With the bus energized, the SI signal will predominate.

If a SI and u/v signal are received simultaneously, the SI sequence will be initiated after the completion of the load shed and subsequent 9(t,IeoA3re-energization of the lE bus by the EDG.

If a u/v signal is received after SI actuation, the sequencer will initiate a load shed and generate the permissive for the E]JG to re energize the 1E bus (the EDG would have previously been started by the initiating SI signal) . After the lE bus is re-energized, the loading sequence will be a function of the status of the SI signal. If the SI signal is still present, the SI sequence will be initiated at step 1.

If the SI signal is no longer present (i.e. SI has been reset) when the EDG re-energizes the lE bus, then the u/V sequence will be initiated at step 1. After completion of the loading sequence, any SI loads required to be in service that were not started during the U/v sequence will have to be manually placed in service (i.e. SI pumps, RHR pumps, Containment Cooler Low Speed motors, ESF Chilled Water pumps and ESF Chillers)

If an SI signal is received after the u/v sequence has been initiated, the sequencer will suspend the u/v sequence upon receipt of the SI signal and restart at step 1 of the SI sequence. If the SI signal is received after completion of the u/v sequence, then the sequencer will begin at step 1 of the SI sequence. For these conditions, since the 1E bus was energized at the time the SI signal was received, no additional load shed will occur. Any u/v loads started from the initiating u/v signal will remain in operation.

18 Revision 2.0

AUTO START LOGIC SEQ BLOCK AUTO START OPENS ON SEQ START SI OR LOSP MOMENTARY I LOSp LOSES ON Low Pressure CLOSE AUTO START ON LOP 5

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ACCW pump response to LOSP (A train ONLY), with no SI Initial conditions: 1AAO2 and 1BAO3 energized from the RATs, ACCW pump

1. 1 in service with ACCW pump #2 in standby.

Click #1: shows circuit energized up to first contacts all the time.

Click #2: Loss of 1AAO2, pump #1 trips (no amber light).

Click #3:

For A Train: Contact for LOSP and contact for Low Header Pressure closes, contact for Sequencer Block AUTO Start on SI or LOSP opens.

For B Train: Sequencer BLOCK AUTO Start is normally closed. Upon reaching the setpoint for low header pressure, the contact for low hdr pressure closes thus completing the circuit for a start of the standby pump.

Click #4: DGs start and re-energize 1AAO2 and IBAO3. Sequencer closes in the last contact to make up the circuit for a pump start. A pump now running from LOSP and B pump running from low header pressure.

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START SEQUENCE SI, no LOSP SEQ BLOCK AUTO START OPENS ON SI OR LOSP LOSES ON Low Pressure LAUTO START I eve,v4- frcc

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ACCW pump response to SI with no LOSP Initial conditions: ACCW pump #1 in service with ACCW pump #2 in standby.

Click #1: shows circuit energized up to first contacts all the time.

Click #2: Low pressurizer pressure SI signal received Click #3: Contact for Sequencer Block AUTO Start on SI or LOSP opens, SI.

contact opens, contact for Sequencer Start Momentary SI or LOSP closes.

Pump #1 remains running, Pump #2 remains in standby. Auto start signals for both pumps BLOCKED by SI signal.

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Lo 19

HL-15RRO NRC Exam

40. 057AA2.05 001/1/1/LOSS VIT AC-SGMETER/C/A -3.5/3.8/NEW/HL-15R NRC/RO/DS / TNT The following indications occur with the unit at full power:

All four SGs channel 1 NR levels go off-scale low All four SGs channel 1 pressures go off-scale low This is a loss of and the correct action to take is to...

A 1AY1A place all 4 MFRVs and MFPT SPEED CONTROL MASTER in manual and match channel 2 feed flows to channel 2 steam flows while maintaining SG NR levels.

B. 1AY2A place all 4 MFRVs and MFPT SPEED CONTROL MASTER in manual and match channel 2 feed flows to channel 2 steam flows while maintaining SG NR levels.

C. 1AY1A verify reactor trip and initiate 19000-C, E-O Reactor Trip or Safety Injection.

D. 1AY2A verify reactor trip and initiate 19000-C, E-O Reactor Trip or Safety Injection.

K/A 057 Loss of Vital AC Inst. Bus AA2.05 Ability to determine and interpret the following as they apply to the Loss of Vital AC Instrument Bus:

SIG pressure and level meters.

KIA MATCH ANALYSIS The question presents the indications of a loss of vital AC instrument bus 1AY1A for the SG levels and pressure instruments. The student is required to correctly diagnose the failure and take the proper actions to stabilize the plant and prevent a reactor trip matching the K/A topic.

ANSWER I DISTRACTOR ANALYSIS 83

HL-15R RO NRC Exam A. Correct. A loss of 1AY1A will result in all channel 1 SG level and pressure instruments failing low, refer to AOP 18032-1 Attachment A Table for the l&C loads for 1AY1A. The actions to stabllze the plant with power above P10 (10%) are to control SG NR levels between 60% and 70% with the MFRVs in manual and the MFPT SPEED CONTROL MASTER in manual. Since SG channel 1 pressures input into the density compensation circuit the channel 1 steam flow instruments will also be reading down scale requiring the UO to use the channel 2 steam flow indications.

B. Incorrect. A loss AC vital instrument bus 1AY2A will not affect any SG level or pressure instruments, so the there is no need to manually control SG NR levels.

IAY2A is a train A vital instrument bus with different actions contained in a different section (B) of the same abnormal operating procedure (18032-1) used to address the loss of 1AY1A.

C. Incorrect. The diagnosis for the vital AC instrument bus is correct. A loss of 1AY1A will result in all channel I SG level and pressure instruments failing low, refer to AOP 18032-1 Attachment A Table for the I&C loads for 1AY1A. The actions for this choice are incorrect since reactor power is above the PlO setpoint of 10%. The actions list for this choice are those specificed by the AOP for a loss of 1AY1A with power below PlO (10% reactor power).

D. Incorrect. A loss AC vital instrument bus 1AY2A will not affect any SG level or pressure instruments, so the there is no need to perfrom the actions specified for section A of AOP 18032-1.

REFERENCES AOP 18032-1, Loss of 12OVAC Instrument Power Section A Loss of 1AY1A, pages 10 and 25 Section B Loss of 1AY2A, pages 28 and 30 VEGP learning objectives:

LO-LP-60324-01:

Given the appropriate plant drawings, logics, and/or procedures, describe how the plant will respond to a loss of the following 1 2OVAC instrument panels:

a. 1AY1A

b. 1AY2A

c. 1BY1B

d. 1BY2B

e. 1CY1A

f. 1DY1B

g. 1NY1N

h. 1NY2N

i. 1NY3N

j. 1NY4N

k. 1NYC2 84

HL-15R RO NRC Exam I. 1NYJ m.1NYR n.INYS o.1NYRS

p. 1NYO1 LO-LP-60324-02:

Given that a loss of 12OVAC instrument power has occurred to any of the following panels, and given the appropriate plant procedures, describe the operator actions required and why these actions are taken.

a. 1AY1A

b. 1AY2A

c. 1BY1B

d. 1BY2B

e. 1CY1A

f. 1DY1B

g. 1NY1N

h. 1NY2N

i. 1NY3N

j. 1NY4N

k. 1NYC2 I. 1NYJ

m. 1NYR

n. 1NYS

o. 1NYRS

p. 1NYO1 85

Approved By Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant 18032-1 27 Date Approved Page Number LOSS OF I2OVAC INSTRUMENT POWER I 3/22/09 10 of 100 A. LOSS OF VITAL INSTRUMENT PANEL 1AY1A (CB-B52)

ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED IMMEDIATE OPERATOR ACTIONS

_A1. Check reactor power GREATER

- Al. Perform the following:

THAN P-l0 SETPOINT.

_a. Verify reactor trip.

_b. Initiate 19000-C, E-0 REACTOR TRIP OR SAFETY INJECTION.

_A2. Verify ROD BANK SELECTOR SWITCH in manual.

• A3 Control SG NR levels BETWEEN-60% AND 70%:

• MFRVs in manual.

• MFPT SPEED CONTROL MASTER in manual.

SUBSEQUENT OPERATOR ACTIONS

_A4. Initiate the Continuous Actions Page.

• A5 Control charging to:

. Maintain seal injection flow to all RCPs -8 TO 13 GPM.

• IF letdown isolated, THEN adjust charging flow to approximately 10 gpm greater than total seal injection flow.

Printed October 27, 2009 at 15:57

Approved By I IProcedure Number Rev J. B. Stanley Vogtle Eectric Generating Pant 1180321 27 Date Approved Page Number LOSS OF I2OVAC INSTRUMENT POWER 3/22/09 25 of 100 ATTACHMENT A Sheet 2 of 4 TABLE 1 I&C LOADS PANEL 1AY1A INST. NO. DESCRIPTION ALTERNATE STEAM GENERATOR INSTRUMENTATION LR-501-P1 SG 1 WR Level Rec LI-501 SG 1 WR Level (I)

LI-529 SG 2 NR Level (I) LI-528 (III)

Ll-539 SG 3 NR Level (I) Ll-538 (Ill)

LI-551 SG 1 NR Level (I) LI-517 (IV)

Ll-554 SG 4 NR Level (I) LI-547 (IV)

PR-514 SG 1 & 2 Press Rec PI-514A SG 1 Press (I) PI-516A (IV)

PI-524A SG 2 Press (I) Pl-526A (Ill)

PI-534A SG 3 Press (I) PI-536A (III)

PI-544A SG 4 Press (I) PI-546A (IV)

AUXILIARY FEEDWATER INSTRUMENTATION Ll-5111A CST 1 Level (I) LI-5101 (II)

Ll-5116A CST 2 Level (I) LI-5104 (II)

FI-51 50A SG 4 AFW Flow Fl-5152A SG 1 AFW Flow MISCELLANEOUS SYSTEMS INSTRUMENTATION LR-990-P1 RWST Level Rec LR-990-P2 Ll-990 RWST Level (I) Ll-991A (II)

LI-i 02A Boric Acid Tk Level (I) LI-i 04A (IV)

P1-937 CNMT NR Press (I) P1-936 (II)

P1-505 Turbine Impulse Press (I) P1-506 (II)

P1-i 636 NSCW Train A Supply Press P1-i 637 FT-22425 Train A Essential Chiller Evaporator Flow PT-6161 Turbine ETS_pressure PT-I 0942 Containment Wide Range Pressure (Computer input)

QT-2791 Diesel Generator Train A Power Output (Computer input)

Fl-12191 CB Control Room Exhaust Flow PDIS-12136 CB Control Room HEPA Filter Differential Pressure Fl-i 2542 Aux Build Piping Penetration Room Exhaust Flow AFI-1 2551 FHB Post Accident Exhaust Flow PDI-2550 Aux Bldg Piping Penetration Room Differential Pressure APDIS-2524 FHB Post Accident HEPA Filter Differential Pressure PDIS-2546 Aux Bldg Piping Penetration Room HEPA Filter Differential Pressure APDT-12567 FHB Differential Pressure Printed October 27, 2009 at 15:58

Approved By Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant 18032-1 27 DateApproved Page Number LOSS OF 120V AC INSTRUMENT POWER 3/22/09 28 of 100 B. LOSS OF VITAL INSTRUMENT PANEL 1AY2A (AB-1 18)

ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTES

• Train A ESF sequencer will not operate following loss of Panel 1AY2A.

• Loss of power to Panel 1AY2A will result in a Containment Ventilation Isolation.

• SG-1 and SG-4 ARVs will NOT operate from QMCB following loss of 1AY2A.

BI. Notify Chemistry that the following radiation monitors will be out of service and will need to be reset when power is restored:

e 1 RE-0002 (CVI)

. 1 RE-0005 e 1 RE-2532A (FHBI) 1RE-2532B (FHBI) 1RE-12116 (CR1)

. 1RE-13119

. 1RE-13120 B2. Dispatch an operator to restore Panel 1AY2A by initiating 13431, 12OVAC 1 E VITAL INSTRUMENT DISTRIBUTION SYSTEM.

_B3. Refer to ATTACHMENT B to determine affected instrumentation.

_B4. Refer to Technical Specifications and complete any applicable action statements.

Printed October 27, 2009 at 15:59

Approved By Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant 18032-1 27 Date Approved Page Number LOSS OF 120V AC INSTRUMENT POWER 3122109 30 of 100 ATTACHMENT B Sheet 1 of 1 TABLE 1 PANEL 1AY2A LOAD LIST BREAKER LOAD 03 SAFETY RELATED DISPLAY CONSOLE DRMS 04 DATA MODULE 1 RX0005 - CNMT AREA MONITOR 05 DATA MODULE 1RX0002 - CNMT AREA MONITOR 06 DATA MODULE ARX-2532 - FUEL HANDLING BLDG HVAC MONITOR 07 DATA MODULE 1RX-12116 - CR AIR INTAKE MONITOR 08 SEQUENCER BOARD 1-1821 -U3-001 09 BOP SAFETY ACTUATION CABINET 11CQESF 10 TRAIN A SYSTEM STATUS MONITORING PANEL 11 PREAMP1RT-005 12 DATA MODULE 1RX-1 3119- MAIN STEAMLINE MONITOR 13 DATA MODULE 1RX-13120 - MAIN STEAMLINE MONITOR 14 DISPLAY PROCESSING UNIT (DPU-A) 15 REMOTE PROCESSING UNIT Al, CHANNEL I 16 REMOTE PROCESSING UNITA2, CHANNEL I 17 SERVO-AMP FOR ATM DUMP VALVE 1ATPY3000 18 SERVO-AMP FOR ATM DUMP VALVE 1ATPY3O3O 19 SPARE 20 SPARE 21 SPARE 22 SPARE 23 SPARE 24 SPARE END OF ATTACHMENT B Printed October 27, 2009 at 15:59