Text

301 Draft RO Written Exam
	ML112270225
Person / Time
Site:	Vogtle
Issue date:	08/08/2011
From:	; NRC/RGN-II/DRS/OLB
To:	; Southern Nuclear Operating Co
References
	05-424/11-301, 05-425/11-301
	Download: ML112270225 (516)
	v • d • e

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HL-16 NRC Written Examination KEY

1. 001G2.4.9 001/1/2/CONT ROD W/D/3.8/4.2 C/A/NEW/RO/SRO/NRC/TNT/GCW Reactor power is stable at 2 X 10 -

% with critical data being taken when the following indications occur.

- Red RODS OUT light illuminates on QMCB vertical panel C.

- Control Bank D begins stepping at 72 steps per minute.

- SR and IR SUR meters indicate positive SURs.

Which ONE of the following Reactor Trips would:

(1) automatically function to mitigate this event, and (2) what is the FIRST action(s) the OATC should take for these indications lAW AOP-1 8003-C, Rod Control System Malfunction?

A. (1) Source Range Hi Flux Trip (2) Place rods in MANUAL and Rod Motion Switch in hold, verify rod motion stops.

B (1) Intermediate Range Hi Flux Trip (2) Place rods in MANUAL and Rod Motion Switch in hold, verify rod motion stops.

C. (1) Source Range Hi Flux Trip (2) Manually trip the reactor, enter 19000-C, E-O, Reactor Trip or Safety Injection.

D. (1) Intermediate Range Hi Flux Trip (2) Manually trip the reactor, enter 19000-C, E-O, Reactor Trip or Safety Injection.

Feedback 001 Continuous Rod Withdrawal 2.4 Emergency Procedures I Plan 2.4.9 Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.

(CFR: 41.10 I 43.5 / 45.13)

K/A MATCH ANALYSIS A plausible scenario is given during a reactor startup where indications of uncontrolled Page lot 208

HL-16 NRC Written Examination KEY outward rod motion are given. The candidate must determine which reactor trip would auto trip the reactor with the given conditions and first actions the OATC should take to mitigate the event.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. The Tech Spec basis for the Source Range Hi Flux Trip identifies it as mitigating an uncontrolled outward rod motion event at low power. However, at the given power level, the source range hi flux trip should be blocked. Placing rods in MANUAL and the Rod Motion Switch in hold and verifying rod motion stops is the OATC first IQA and is part is correct.

B. Correct. The Tech Spec Basis for the Intermediate range Hi Flux Trip identifies it as mitigating an uncontrolled outward rod motion event at low power. Placing rods in MANUAL and the Rod Motion Switch in hold and verifying rod motion stops is the OATC first IOA and is part is correct.

C. Incorrect. The Tech Spec basis for the Source Range Hi Flux Trip identifies it as mitigating an uncontrolled outward rod motion event at low power. However, at the given power level, the source range hi flux trip should be blocked. Manually tripping the reactor and entering E-0 is the RNO if rod motion does not stop but is not the next action the OATC should take.

D. Incorrect. The Tech Spec Basis for the Intermediate range Hi Flux Trip identifies it as mitigating an uncontrolled outward rod motion event at low power, this part is correct. Manually tripping the reactor and entering E-0 is the RNO if rod motion does not stop but is not the next action the OATC should take.

REFERENCES 18003-C, Section B for Uncontrolled Continuous Rod Motion (page 7)

Technical Specification 3.3.1, FU 4 and FU 5 for SR and IR Neutron Flux Tech Spec Bases 3.3.1, FU 4 and FU 5 for SR and IR Neutron Flux Procedure 12003-C reactor Startup (Mode 3 to Mode 2) (pages 27-28)

VEGP learning objectives:

LO-LP-60303-09 Describe how and why the following will change during an uncontrolled continuous rod withdrawal:

a. Tavg

b. pressurizer level

c. pressurizer pressure

d. delta flux Page 2 of 208

HL-16 NRC Written Examination KEY

e. reactor power LO-LP-60303-12 State the immediate operator actions required for an uncontrolled continuous rod motion. Include RNO and substeps of the immediate action.

Page 3 of 208

RTS Instrumentation 3.3.1 Table 3.31-1 (page 1 of 9)

Reactor Trip System Instrumentation I

APPLICABLE MODES OR NOMINAL OTHER TRIP SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE REQUIREMENTS VALUE SETPOINT n FUNCTION CONDITIONS CHANNELS CONDITIONS Manual Reactor 1,2 2 B SR 3.3.1.13 NA NA Trip (3a) (4a) 5 (a) 2 C SR 3.3.1.13 NA NA

2. Power Range Neutron Flux

a. High 1,2 4 D SR 3.3.1.1 111.3% RTP 109% RTP SR 3.3.1.2 SR 3.3.1.7 SR 3.3.1.11 SR 3.3.1.15

b. Low 2

(

1 b) 4 E SR 3.3.1.1 27.3% RTP 25% RTP SR3.3.1.8 SR 3.3.1.11 SR 3.3.1.15

3. Power Range 1,2 4 E SR 3.3.1.7 6.3% RTP 5% RTP Neutron Flux High SR 3.3.1.11 with time with time Positive Rate constant constant 2sec 2sec

4. Intermediate b) (2 (1 c) 2 E.G SR 3.3.1.1 41.9% RTP 25% RTP Range Neutron , SR 3.3.1.8 Flux SR 3.3.1.11 2 H SR 3.3.1.1 41.9% RTP 25% RTP 2 (d) SR3.3.1.8 SR 3.31.11 (continued)

(a) With Reactor Trip Breakers (RTB5) closed and Rod Control System capable of rod withdrawal.

(b) Below the P-10(Power Range Neutron Flux) interlocks.

(c) Above the P-6 (Intermediate Range Neutron Flux) interlocks.

(d) Below the P-6 (Intermediate Range Neutron Flux) interlocks.

(n) A channel is OPERABLE with an actual Trip Setpoint value outside its calibration tolerance band provided the Trip Setpoint value is conservative with respect to its associated Allowable Value and the channel is readjusted to within the established calibration tolerance band of the Nominal Trip Setpoint. A Trip Setpoint may be set more conservative than the Nominal Trip Setpoint as necessary in response to plant conditions.

Vogtle Units 1 and 2 3.3.1-14 Amendment No. 128 (Unit 1)

Amendment No. 106 (Unit 2)

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 2 of 9)

Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER NOMINAL TRIP SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS SETPOINT(n)

VALUE

5. Source Range (d) 2 2 I,J SR 3.3.1.1 1.7 E5 1.0 E5 Neutron Flux SR 3.3.1.8 cps cps SR 3.3.1.11 (a) 4 3 (a) 5(8) 2 J,K SR 3.3.1.1 1.7 E5 1.0 E5 SR 3.3.1.7 cps cps SR 3.3.1.11 1 L SR 3.3.1.1 4

3

( e) 5 (e) SR 3.3.1.11 NA NA

6. Overtemperature T 1,2 4 E SR 3.3.1.1 Refer to Note 1 Refer to Note 1 SR 3.3.1.3 (Page 3.3.1-20) (Page 3.3.1-20)

SR 3.3.1.6 SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1,15

7. Overpower T 1,2 4 E SR 3.3.1.1 Refer to Note 2 Refer to Note 2 SR 3.3.1.7 (Page 3.3.1-21) (Page 3.3.1-21)

SR 3.3.1.10 SR 3.3.1.15 (continued)

(a) With RTBs closed and Rod Control System capable of rod withdrawal.

(d) Below the P-6 (Intermediate Range Neutron Flux) interlocks.

(e) With the RTBs open. In this condition, source range Function does not provide reactor trip but does prxwide input to the High Flux at Shutdown Alarm System (LCO 3.3.8) and indication.

(n) A channel is OPERABLE with an actual Trip Setpoint value outside its calibration tolerance band provided the Trip Setpoint value is conservative with respect to its associated Allowable Value and the channel is readjusted to within the established calibration tolerance band of the Nominal Trip Setpoint. A Trip Setpoint may be set more conservative than the Nominal Trip Setpoint as necessary in response to plant conditions.

Vogtle Units 1 and 2 3.3.1-15 Amendment No. 128 (Unit 1)

Amendment No. 106 (Unit 2)

Approved By Procedure Number Rev A. S. Parton Vogtle Electric Generating Plant 8003-C 24.1 Date Approved Page Number ROD CONTROL SYSTEM MALFUNCTION C128120b0 2of23 MAJOR ACTIONS

Respond to Uncontrolled Continuous Rod Motion.

Respond to Dropped Rods in Mode 1.

Respond to Misaligned Rods in Mode 1.

Respond to Dropped or Misaligned Rods in Modes 2 through 5.

Printed January 17, 2011 at 14:07

Approved By I Procedure Number Rev Vogtie Electric Generating Plant A. S. Parton e APproved t

c I 18003-C 24.1 Page Number ROD CONTROL SYSTEM MALFUNCTION 23 B. UNCONTROLLED CONTINUOUS ROD MOTION ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED IMMEDIATE OPERATOR ACTIONS Bi. Stop uncontrolled Rod motion by performing the following:

_a. Place ROD BANK SELECTOR SWITCH in MAN position.

_b. Place the Rod Motion Switch in hold.

_B2. Check Rod motion STOPPED. B2. Perform the following:

_a. if in Modes 1, 2, or 3, trip the reactor, THEN go to 19000-C, E-0 REACTOR TRIP OR SAFETY INJECTION.

_b. jEinModes4or5, THEN open the reactor trip breakers and go to Step 86.

SUBSEQUENT OPERATOR ACTIONS B3. Check the following alarms B3. if in Mode 1 or 2, EXTINGUISHED: THEN perform the following:

ALB1O-C4 ROD BANK LO LIMIT . Borate as necessary to restore Rod height.

. ALB1O-D4 ROD BANK LO-LO

Initiate the following:

LIMIT TS3.1.5 TS 3.1.6 B4. Restore Tavg to program by adjusting B4. Adjust RCS boron concentration to turbine load, restore Tavg to program.

Printed January 17, 2011 at 14:07

RTS Instrumentation B 3.3.1 BASES APPLICABLE 3. Power Range Neutron Flux High Positive Rate (continued)

SAFETY ANALYSES, LCO, and OPERABLE because other RTS trip Functions and APPLICABILITY administrative controls will provide protection against positive reactivity additions. In MODE 6, no rods are withdrawn and the SDM is increased during refueling operations. The reactor vessel head is also removed or the closure bolts are detensioned preventing any pressure buildup. In addition, the NIS power range detectors cannot detect neutron levels present in this mode.

4. Intermediate Range Neutron Flux The Intermediate Range Neutron Flux (NI-0358, D, & E, NI-036B, D, & G) trip Function ensures that protection is provided against an uncontrolled RCCA bank rod withdrawal accident from a subcritical condition during startup. This trip Function provides redundant protection to the Power Range Neutron Flux Low Setpoint trip Function. The NIS intermediate range detectors are located external to the reactor vessel and measure neutrons leaking from the core. The NIS intermediate range detectors do not provide any input to control systems. Note that this Function also provides a signal to prevent automatic and manual rod withdrawal prior to initiating a reactor trip.

The LCO requires two channels of Intermediate Range Neutron Flux to be OPERABLE. Two OPERABLE channels are sufficient to ensure no single random failure will disable this trip Function.

Because this trip Function is important only during startup, there is generally no need to disable channels for testing while the Function is required to be OPERABLE. Therefore, a third channel is unnecessary.

In MODE 1 below the P-1O setpoint, and in MODE 2, when there is a potential for an uncontrolled RCCA bank rod withdrawal accident during reactor startup, the Intermediate Range Neutron Flux trip must be OPERABLE.

(continued)

Vogtle Units 1 and 2 B 3.3.1-14 Revision No. 0

RTS Instrumentation B 3.3.1 BASES APPLICABLE 4. Intermediate Range Neutron Flux (continued)

SAFETY ANALYSES, LCO, and Above the P-1O setpoint, the Power Range Neutron APPLICABILITY FluxHigh Setpoint trip and the Power Range Neutron Flux High Positive Rate trip provide core protection for a rod withdrawal accident. In MODE 3, 4, or 5, the Intermediate Range Neutron Flux trip does not have to be OPERABLE because the reactor cannot be started up in this condition. The core also has the required SDM to mitigate the consequences of a positive reactivity addition accident. In MODE 6, all rods are fully inserted and the core has a required increased SDM. Also, the NIS intermediate range indication is typically low off-scale in this MODE.

5. Source Range Neutron Flux The LCO requirement for the Source Range Neutron Flux trip (NI-0031 B, D, & E, NI-0032B, D, & G) Function ensures that protection is provided against an uncontrolled RCCA bank rod withdrawal accident from a subcritical condition during startup.

This trip Function provides redundant protection to the Power Range Neutron Flux Low Setpoint and Intermediate Range Neutron Flux trip Functions. In MODES 3, 4, and 5, administrative controls also prevent the uncontrolled withdrawal of rods. The NIS source range detectors are located external to the reactor vessel and measure neutrons leaking from the core.

The NIS source range detectors do not provide any inputs to control systems. The source range trip is the only RTS automatic protection function required in MODES 3, 4, and 5.

Therefore, the functional capability at the specified Trip Setpoint is assumed to be available.

The LCO requires two channels of Source Range Neutron Flux to be OPERABLE. Two OPERABLE channels are sufficient to ensure no single random failure will disable this trip Function.

The LCO also requires two channels of the Source Range Neutron Flux to be OPERABLE in MODE 3, 4, or 5 with RTBs closed.

The Source Range Neutron Flux Function provides protection for control rod withdrawal from (continued)

Vogtle Units 1 and 2 B 3.3.1-15 Rev. 1-3/99

RTS Instrumentation B 3.3.1 BASES APPLICABLE 5. Source Rance Neutron Flux (continued)

SAFETY ANALYSES, LCO, and subcritical, boron dHution (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 1T The Overtemperature T trip Function (TDI-041 1C, TDI-0421C, TDI-0431C, TDI-0441C, TDI-0411A, 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 T trip Function must provide protection. The inputs to the Overtemperature T trip include pressure, coolant temperature, axial power distribution, and reactor power as indicated by loop T 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 1 and 2 B 3.3.1-16 Rev. 1-3/99

Approved By J. B. Stanley Date Approved i Vogtle Electric Generating Plant A Procedure Number Rev 12003-C 49 Page Number 10/24/2010 i REACTOR STARTUP (MODE 3 TO MODE 2) 27 of 39 INITIALS CAUTION Source Range high flux reactor trip will occur at 1 cps.

4.2.24 At an IR indication of approximately 2 X 10%, perform the following: (1984300048, 1984301809)

a. Observe the following status lights illuminated:

(1985303275, 1985303276)

(1) IR P6 NC35D (TSLB-4, 3.1), or (2) lR P6 NC36D (TSLB-4, 3.2)

(3) SOURCE RANGE BLOCK PERMISSIVE P6

b. Block the source range hi flux reactor trip by placing both SR BLOCK RESET A/B switches HS-40030/HS-40031 to the BLOCK position,

c. Observe the following BPLP status lights illuminated:

(1) SR TRAIN A TRIP BLKD (2-1)

(2) SR TRAIN B TRIP BLKD (2-2)

CAUTION If the startup is stopped after blocking the SR Hi flux trip and power level is allowed to decrease, the SR trip may become unblocked. There is no audible indication to alert the operator if this occurs. The SR Hi Flux trip should be verified blocked and, if necessary, the above steps repeated to re-block the trip prior to reaching 1 E+5 cps.

d. Configure Nuclear Instrumentation for continuing power increase:

(1) Verify both channels of Intermediate Range Nis indicating properly on recorder NR-45.

(2) Remove Audible Count Rate monitor from service per 13501, Nuclear Instrumentation System.

Printed January 17, 2011 at 14:09

Approved By J. B. Staney Vogtle Electric Generating Plant Procedure Number Rev Date Approved 12003-C 49 Page Number 10/24/2010 REACTOR STARTUP (MODE 3 TO MODE 2) 28 of 39 INITIALS 4.2.25 Raise power to 2 X 1 0% in the Intermediate Range by adjusti ng control rods as necessary to establish a SUR of approximately 0.5 dpm.

4.2.26 Stabilize power at an Intermediate Range indication of approximately 2 X 1 0% and complete:

a. OSP 14940, Estimated Critical Condition Calculation Data Sheet 1, Actual Critical Data, OR

b. 88010-C, Computer Calculation Of Estimated Critical Conditions Data Sheet 3.

c. Place a copy of the above Data Sheet(s) in the Start-up Log tab of the Reactor Trip Log.

d. IF this reactor startup IS NOT a dilution to criticality for LPPT, Tavg recording per Data Sheet 2 can be terminate d

if ALB12AO5 is NOT illuminated.

NOTE The Avg/Tref Deviation alarm, ALB12AO5, provides actio ns to maintain Tavg above 551°F, the minimum temperature for criticality.

4.2.27 Monitor Tavg/Tref Deviation alarm, ALB12AO5, during the remainder of the startup and take corrective action as direc ted to maintain Tavg at 557°F +/-2°F. (TS SR 3.4.2.1) (1996232 496) 4.2.28 Unblock both Source Range channels HFASA circuits per 13501, Nuclear Instrumentation System.

(1) Source Range Channel N31 (2) Source Range Channel N32 4.2.29 IF this reactor startup was a dilution to criticality for LPPT, perform LPPT-GAEIGBE-01.

Printed January 17, 2011 at 14:09

HL-16 NRC Written Examination KEY 001 K2.05 001 /2/2/CONTROL RODS-MG SETS/3. 1/3.5 MEM/lIL- I SR NRC/RO/SRO/NRC/TNT/GCW Which ONE of the following describes the power supplies to the Rod Drive Motor Generator (MG) sets and the breakers that are required to be locally tripped during an ATWT per 19211-C, FR-S.l, Response to Nuclear Power Generation / ATWT, if the Reactor Trip breakers will NOT open?

MG Set Power Supplies Breakers to Trip A. 1NBO9 and 1NB1O M/G set motor breakers B. 1NBO9 and 1NB1O M/G set output breakers C. 1 NBO8 and 1 NBO9 M/G set motor breakers D 1 NBO8 and 1 NBO9 M/G set output breakers Feedback 001 Control Rod Drive System Knowledge of bus power supplies to the following:

(CFR: 41.7)

K2.05 M/G sets K/A MATCH ANALYSIS The question presents a plausible scenario where an ATWT is in progress, the student must know which non-i E switchgear powers the MG set and to open both MG sets output breakers during an ATWT.

ANSWER / DISTRACTOR ANALYSIS A. Incorrect. It is plausible that the MG set motor breakers could be opened as this would interupt power and cause the rods to fall in. It is plausible thati NBO9 and 1 NB1 0 power supplies to the MG sets as they are both non-i E switchgear and 1 NBO9 is one of the two power supplies to the MG sets. Both MG sets output breakers must be opened per 19211-C to trip the reactor. The output breakers are to be opened versus the motor breakers. Opening the output breakers immediately cuts power to the rod control power cabinets and all rods immediately drop in.

Opening the motor 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. Incorrect. It is plausible that the MG set motor breakers could be opened as this would interupt power and cause the rods to fall in. It is plausible thati NBO9 and 1 NB1O power supplies to the MG sets as they are both non-i E switchgear and Page 4 of 208

HL-16 NRC Written Examination KEY 1 NBO9 is one of the two power supplies to the MG sets. Both MG sets output breakers must be opened per 19211-C to trip the reactor. The output breakers are to be opened versus the motor breakers. Opening the output breakers immediately cuts power to the rod control power cabinets and all rods immediately drop in.

Opening the motor 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. It is plausible that the MG set motor breakers could be opened as this would interupt power and cause the rods to fall ml NBO8 and 1 NBO9 are the switchgear which power the MB sets. Both MG sets output breakers must be opened per 19211-C to trip the reactor. The output breakers are to be opened versus the motor breakers. Opening the output breakers immediately cuts power to the rod control power cabinets and all rods immediately drop in. Opening the motor 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. 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 motor breakers. Opening the output breakers immediately cuts power to the rod control power cabinets and all rods immediately drop in. Opening the motor 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 V-LO-PP-28101, Solid State Protection System, slide #59.

19211-C, FR-S. 1, Reponse to Nuclear Power Generation/ATWT, step 8 RNO Re-use from Vogtle March 2010 NRC RO exam.

VEGP learning obiectives:

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

Page 5 of 208

Approved By J. D. WilHams Vogtle Electric Generating Plant Procedure Number Rev

[Date Approved 1921 1 -c 20.3 FR-S.1 RESPONSE TO NUCLEAR POWER Page Number

()-23-2007 GENERATION/ATWT 6 of 20 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

7 Check for SI:

_a. SI signal EXISTS Q -

_a. jf an SI signal is actuated ACTUATED.

during this procedure, THEN initiate ATTACHMENT A.

GotoStep8

b. Initiate ATTACHMENT A.

8. Check the following trips have occurred:

_a. Reactor trip.

_a. Locally trip the Reactor trip and Bypass breakers.

I.E the trip breakers will NOT open, THEN trip the Control Rod Drive MG Set output breakers at the Reactor Trip Swltchgear.

_b. Turbine trip.

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

9 Check Reactor power:

_a. LESS THAN 5%.

_a. Go to Step 10.

_b. IR SUR LESS THAN 0 DPM.

_b. Go to Step 10.

_c. GotoStep24.

Printed January 17, 2011 at 14:14

Reactor Trip Reactor Trip Bypass Breaker Bypass Breaker MIG Set Output Breaker Rod Control Reactor Trip Reactor Trip rn rn Breaker A Breaker B Cabinets M/G Set Output Breaker V-LO-PP-281 01-04.1 59

1. 001K2.05 002/2/2/RODS-MG SETS/MEM- 2.9 / 3.2/BANK HARRIS 2009/HL-15R NRC/RO/TN T/DS Which ONE of the following describes the power supplies to the Rod Drive Motor Generator (MG) sets and the breakers that are required to be locally tripped during an ATWT per 1921 1-C, FR-S.1, Response to Nuclear Power Generation/ATWT, if the Reactor Trip breakers will not open?

MG Set Power Supplies Breakers to Trip A. 1NBO9 and 1NB1O MG set motor breakers B. 1NBO9 and 1NB1O MG set output breakers C. 1 NBO8 and 1 NBO9 MG set motor breakers D 1 NBO8 and 1 NBO9 MG set output breakers

{6 V6fz(

7/0/0 VAJTZC

HL-16 NRC Written Examination KEY

3. 002K5. 11 OO1/2/2/RCS-SECONDARY/4.O/4.2 C/AIBANK/RO/SROINRC/TNT/GCW Which ONE of the following symptoms can be used to differentiate between a primary leak and a secondary steam leak and why?

A. Reactor Power rising due to positive MTC.

B Reactor Power rising due to negative MTC.

C. Pressurizer Level lowering due to inventory loss.

D. Pressurizer Level lowering due to excessive cooldown.

Page 6 of 208

HL-16 NRC Written Examination KEY Feedback 002 Reactor Coolant System (RCS)

Knowledge of the operational implications of the following concepts as they apply to the RCS:

(C FR: 41.5/45.7)

K5.1 1 Relationship between effects of the primary coolant system and the secondary coolant system K/A MATCH ANALYSIS This question test the knowledge of the candidate of how a secondary system leak effects the primary system and how to differentiate between the two.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect but plausible because Reactor Power will increase but not because of positive MTC. With a Postive MTC, Reactor Power would lower due to the RCS cooling effects of a steam leak.

B. Correct a secondary steam leak will cause Reactor Power to increase during the core life when MTC is negative. Primary leak will never cause Reactor Power to increase.

C. Incorrect but plausible because Pressurizer Level will lower due to due to the cooling effect caused by the secondary steam Leak but can not be soley because Pressurizer Level will lower due to inventory losses also.

D. Incorrect but plausible because Pressurizer Level will lower due to inventory losses but can not be soley because secondary steam Leak will also cause Pressurizer Level to lower due to the cooling effect.

REFERENCES 18004-C, RCS Leakage.

18008-C, Secondary Coolant Leakage.

LO-LP-60308-02-2 LORQ Bank.

COLR VEGP Ui Cycle 162.3 Moderator Temperature Coefficient VEGP learning objectives:

LO-LP-60308-04 Discuss the parameters that distinguish primary coolant leakage from secondary coolant leakage.

Page 7of 208

Approved By Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant 18004-C 27.2 Date Approved REACTOR COOLANT SYSTEM LEAKAGE Paqe Number

/18/10 82 ABNORMAL OPERATING PROCEDURE CONTINUOUS USE PURPOSE Section A specifies the actions to be taken for Reactor Coolant System leakage during Modes 1, 2, and 3 with RCS pressure greater than 1000 psig.

Section B specifies the actions to be taken for Reactor Coolant System leakage during Mode with RCS pressure less than 1000 psig, and Mode 4. 3 Section C specifies the action to be taken for Reactor Coolant System Leakage during Mode 5.

SYMPTOMS

Unexplained change in charging flow.

A rise in VCT makeup frequency.

Unexplained lowering of PRZR level and pressure.

PRT temperature, pressure or level rising.

CNMT moisture alarm or activity rising.

CNMT sump level rising.

CNMT Air Cooler condensate flow rising alarm.

Transition from 18019-C, LOSS OF RESIDUAL HEAT REMOVAL.

Printed January 17, 2011 at 14:22

Approved By SA. PhiUps Vogtle Electric Generating Plant Procedure Number Rev 18008-C 9.1 SECONDARY COOLANT LEAKAGE ci Page Number 6

ABNORMAL OPERATING PROCEDURE CONTINUOUS USE PURPOSE This procedure specifies operator actions for secondar y leaks which do NOT actuate Engineered Safeguards Features.

SYMPTOMS

ALB13-AO1 (BOl, COl, DOl) STM GEN 1(2,3,4)

FLOW MISMATCH

ALB13-A06 (B06, C06, D06) STM GEN 1 (2, 3, 4) HI/L O LVL DEVIATION

High containment pressure, temperature, moisture, and sump levels WITHOUT radiation.

High condenser hotwell makeup rates.

High CST makeup rates.

Observed secondary leakage.

Unexplained rise in reactor power.

Reactor power significantly higher than turbine power.

MAJOR ACTIONS

Evaluate and stabilize plant conditions.

Locate and isolate leakage.

Determine whether to continue operation or initiate plan t shutdown.

Printed January 17, 2011 at 14:22

COLR for VEGP UNIT 1 CYCLE 16 2.0 OPERATING LIMITS The cycle-specific parameter limits for the specifications listed in section 1 .0 are presented in the following subsections. These limits have been developed using NRC-approved methodologies, including those specified in Technical Specification 5.6.5.

2.1 SHUTDOWN MARGIN - MODES 1 AND 2 (Technical Requirement 13.1.1) 2.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.30% tik/k.

2.2 SHUTDOWN MARGIN - MODES 3,4 AND 5 (Specification 3.1.1) 2.2.1 The SHUTDOWN MARGIN shall be greater than or equal to the limits shown in Figures 1 and 2.

2.3 Moderator Temperature Coefficient (Specification 3.1.3) 2.3.1 The Moderator Temperature Coefficient (MTC) limits are:

The BOLJARO/HZP MTC shall be less positive than +0.7 x io kIkI°F for power levels up to 70% RTP with a linear ramp to 0 k/k/°F at 100% RTP.

The EOLJARO/RTP-MTC shall be less negative than

-5.50 x 10 1k/k/°F.

1 2.3.2 The MTC Surveillance limits are:

The 300 ppm/ARO/RTP-MTC should be less negative than or equal to -4.75 x io kIk/°F.

1 The 60 ppm/ARO/RTP-MTC should be less negative than

-5.35 x 10 ik/k/°F.

1 where: BOL stands for Beginning of Cycle Life ARO stands for All Rods Out HZP stands for Hot Zero THERMAL POWER EOL stands for End of Cycle Life RTP stands for RATED THERMAL POWER 2.4 Shutdown Bank Insertion Limits (Specification 3.1.5) 2.4.1 The shutdown banks shall be withdrawn to a position greater than or equal to 225 steps.

Applicable for full-power T-average of 584.4°F to 587.4°F.

1 PAGE 2 of 11

1. LO-LP-60308-02_0021L0LP60308/LO-TA-60037/000EA2.02/////

Which of the following parameters can be used to distingui sh between a primary leak and a secondary steam leak inside containment?

A. Pressurizer level B. Containment sump level C. Containment pressure D Reactor power

HL-16 NRC Written Examination KEY

4. 003K5.03 OO1/2/1/RCP-TAVE INDICATION/3. 1/3.5 C/A!LOIT BANK/RO/SRO/NRC/TNT/GCW Unit 1 reactor power is 6% Rated Thermal Power with all four RCPs running. The following Loop 2 and 3 RCP indications are noted by the Control Room Staff.

Loop 2 RCP: Motor Bearing Temperature = 195 °F Motor Stator Winding Temperature = 312 °F Seal Water Inlet Temperature = 224 °F RCP Shaft Vibration = 14 mils RCP Frame Vibration = 3 mils Loop 3 RCP: Motor Bearing Temperature = 175 °F Motor Stator Winding Temperature = 310 °F Seal Water Inlet Temperature 226 °F RCP Shaft Vibration = 16 mils RCP Frame Vibration = 4 mils Based on the above indications, assuming the required operator actions are taken, which ONE of the following describes the response of the affected loop Tave and the reason for the affect on Tave?

A Loop 2 Tave will initially decrease due to securing Loop 2 RCP.

B. Loop 2 Tave will initially increase due to securing Loop 2 RCP.

C. Loop 3 Tave will initially decrease due to securing Loop 3 RCP.

D. Loop 3 Tave will initially increase due to securing Loop 3 RCP.

Feedback 003 Reactor Coolant Pump System (RCPS)

Knowledge of the operational implications of the following concepts as they apply to the RCPS:

(C FR: 41.5/45.7)

K5.03 Effects of RCP shutdown on T-ave., including the reason for the unreliability of T-ave. in the shutdown loop K/A MATCH ANALYSIS The K/A is matched for the following reasons:

- Knowledge of operational implications of an RCP being tripped is matched by testing the directional trend in Tave when the RCP is initially tripped. The operational implication of tripping an RCP is that Tave initially trends down.

- The reason for the unreliability of Tave is matched because Loop 2 Tave becomes Page 8 of 208

HL-16 NRC Written Examination KEY unrepresentative due to the Loop 2 RCP being tripped. Therefore, the Loop 2 Tave indication is unreliable due to the Loop 2 RCP being tripped.

ANSWER I DISTRACTOR ANALYSIS A. Correct because Loop 2 Tave is required to be tripped (Ref. 2) due to high motor stator winding temperature (>311°F) and Tave will initially decrease due to reverse flow, which occurs when the Loop 2 RCP is tripped.

B. Incorrect because Loop 2 Tave will not increase due to a reduction in hot leg temperature due to the reverse flow, which occurs when Loop 2 RCP is tripped.

Plausible because a misconception could exist where the applicant may believe that the coolant may have a longer transit time through the core in the forward direction, which would cause Thot to increase, thus increasing Tave.

C. Incorrect because Loop 3 RCP is not required to be tripped. Plausible because motor stator winding temperature is close to the value that requires the RCP to be tripped. Also, RCP Shaft vibration is at a level that creates an alarm in the control room, but below the value that requires the pump to be tripped (20 mils).

D. Incorrect because Loop 3 RCP is not required to be tripped. Plausible because motor stator winding temperature is close to the value that requires the RCP to be tripped. Also, RCP Shaft vibration is at a level that creates an alarm in the control room, but below the value that requires the pump to be tripped (20 mils).

REFERENCES System Operating Procedure 13003-1, Rev. 42, Step 2.2.10 Vogtle LOIT Bank 003K5.03-01 VEGP learning objectives:

LO-PP-16101-01 Describe the RCS narrow range temperature instrumentation.

Page 9of 208

Approved By S. E. Prewitt Vogtle Electric Generating Plant Procedure Number Rev Date Approved 13003-1 42 Page Number 3/5/l0 REACTOR COOLANT PUMP OPERATION 5 of 35 2.2.9 During RCS filling and venting, RCS pressure must be grea ter than 325 psig prior to starting an RCP to verify adequate seal DIP is main tained 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 befo re restarting the RCP.

2.2.10 An RCP shall be stopped if any of the following condition s 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 prev ent seal damage, RCP shaft bowing, ACCW System damage, etc. due to excessive thermal stresses.

Printed January 17, 2011 at 14:31

HL-16 NRC Written Examination KEY

5. 003K6.02 OO1/2/IIRCP SEALS/2.7/3. I MEMILOIT BANK/RO/SRO/NRC/GCW An Loss of All AC Power has taken place. The operators are recovering the plant in accordance with 19102-C, Loss of All AC Power Recovery With SI Required.

- The crew is preparing to start a CCP and an ACCW pump.

1) HV-8103A, B, C, D the RCP Seal Injection Isolation valves are all CLOSED.

2) HV-1 979 ACCW Supply Header CRC Isolation valve is CLOSED.

Which ONE of the following is the CORRECT basis for closing the above listed valves?

A. 1) prevents seal leakage to the VCT which could relieve to the Auxiliary Building.

2) minimizes steam formation in ACCW system due to Thermal Barrier heating.

B 1) prevents potential damage to the RCP seals and shaft on CCP restarts.

2) minimizes steam formation in ACCW system due to Thermal Barrier heating.

C. 1) prevents seal leakage to the VCT which could relieve to the Auxiliary Building.

2) prevents runout of the ACCW pump as voids collapse in Thermal Barrier piping.

D. 1) prevents potential damage to the RCP seals and shaft on CCP restarts.

2) prevents runout of the ACCW pump as voids collapse in Thermal Barrier piping.

Page lOot 208

HL-16 NRC Written Examination KEY Feedback 003 Reactor Coolant Pump System (RCPS)

Knowledge of the effect of a loss or malfunction on the following will have on the RCPS:

(CFR: 41 .7/45/5)

K6.02 RCP seals and seal water supply K/A MATCH ANALYSIS Question gives a plausible scenario where a Loss of All AC Power occurs. Candidates have to know the basis for isolating ACCW valves to RCP seals on power restoration.

ANSWER / DISTRACTOR ANALYSIS A. Incorrect, a is basis for closing HV-8100. b basis is correct.

B. Correct.

C. Incorrect. a is basis for closing HV-8100. b is plausible due to heating of the ACCW system creating voids and the effect on the ACCW pump upon collapsing.

D. Incorrect. a basis is correct. b same as C REFERENCES V-LO-HO-37031 -001 -C, Loss of All AC Power.

HL-15 Audit LOIT Bank 003K6.02-01 VEGP learning objectives:

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

LO-PP-1 6401-05 Given a loss of RCP Seal Injection, describe the indications that would be monitored and impact to continued operation of the RCP.

Page 11 of 208

LO-HO-37031-OO1-04-C: Loss Of AU AC Power number 1 seal chamber and shaft area which has the potential for therma l shock and subsequent damage to the RCP seals and shaft.

Defeating automatic loading of other major equipment in this step also functions to protect the respective equipment. This permits the operator to verify valve alignments prior to starting pumps.

STEP: Check if AC buses should be energized locally using 13427, 4160 AC ELECTRICAL DIST.

PURPOSE: To initiate local actions to restore ac power BASIS:

If the operator cannot restore ac power from the control room, it must be restored by local actions. This step instructs the operator to dispatch personnel to locally restore power using a plant specific procedure for ac power restoration.

STEP: Dispatch Personnel To Locally Close Valves To Isolate RCP Seals Per Attachment E And Place Valve Switches In CLOSED Position PURPOSE: To isolate the RCP seals BASIS:

This step groups three actions, with different purposes, aimed at isolatin g the RCP seals. The actions are grouped since all require an operator, dispatched from the control room, to locally close containment isolation valves. Concurrent with dispatching the operator, the control room operator should place the valve switches for the motor operated valves in the closed position so that the valves remain closed when ac power is restored.

Isolating the seal return line prevents seal leakage from filling the volum e control tank (VCT) (via seal return relief valve outside containment) and subsequent transfer to other auxilia ry building holdup tanks (via VCT relief valve) with the potential for radioactive release within the auxiliary buildin

g. Such a release, without auxiliary building ventilation available, could limit personnel access for local operati ons.

isolating the RCP seal injection lines prepares the plant for recovery while protecting the RCPs from seal and shaft damage that may occur when a CCP is started as part of the recove ry With the RCP seal injection lines Isolated, a CCP can be started in the normal charging mode without concern for cold seal injection flow thermally shocking the RCPs Seal injection can subsequently by established to the RCP consistent with appropriate plant specific procedures Isolating the RCP thermal barrier ACCW return outside containment isolation valve prepares the plant for recovery while protecting the ACCW system from steam formation due to RCP thermal barrier heating. Following the loss of all ac power, hot reactor coolant will gradually replace the normally cool seal injection water in the RCP seal 1-22

LO-HO-37031-001 C: Loss Of All AC Power area. As the hot reactor coolant leaks up the shaft, the water in the thermal barrier will heat up and potentially form steam in the thermal barrier and in the ACCW lines adjacent to the thermal barrier. Subsequent automatic start of the ACCW pump would deliver ACCW flow to the thermal barrier, flushing the steam into the ACCW system. it abnormal RCP seal leakage had developed in a pump, the abnormally high leakage rate could exceed the cooling capacity of the ACCW flow to that pump therma l barrier and tend to generate more steam in the RCP thermal barrier ACCW return lines Isolating these lines prevents the potential introduction of this steam into the main portion of the ACCW system upon ACCW pump start This keeps the main portion of the ACCW system available for cooling equipment necessary for recovering the plant when ac power Is restored.

STEP: Dispatch Operator To Check Hot Well Level Valves CLOSED.

PURPOSE: To ensure that CST inventory is conserved for makeu p to the steam generators BASIS:

The operator should verify that a dedicated supply of AFW exists for delivery to the steam generators.

The source of AFW is normally the condensate storage tank. Howev er, following loss of all ac power, the CST may remain unisolated from the condenser hot well, or other water source, and be capable of draining through unisolated lines to the condenser hot well, or other water source. This step isolates the CST to preserve AFW inventory for continued steam generator heat removal.

STEP: Dispatch Operators to prepare for local operation of SG ARVs per 13610, Steam Generator and Main Steam Operation.

PURPOSE: To ensure operators are stationed to perform a second ary depressurization in a timely manner.

BASIS:

This step was added by Plant Vogtle to remind the crew to get a step ahead of this action which require considerable time for breaker operation and setup s of headsets for local communications.

1-23

HL-16 NRC Written Examination KEY

6. 004A3 .06 001/2/1 /CVCS-TAVE/TREF/3 .9/3.8 C/A/NEW/RO/SROINRC/GCW Unit 1 is at 50% power with PRZR level control system in auto.

- RCS Loop 3 NR Tcold fails high.

Which ONE of the following describes the effect on Charging Flow and RCP Seal Injection flows, with NO operator actions.

Charging Flow Seal Iniection Flows A. Go up Go down B Goup Goup C. Go down Go down D. Godown Goup Feedback 004 Chemical and Volume Control System Ability to monitor automatic operation of the CVCS, including:

(C FR: 41.7/45.5)

A306 T-ave. and T-ref KIA MATCH ANALYSIS Question presents a plausible scenario where a RCS NR Tcold (Tave) fails high.

The candidate must determine the effect of the failure on the CVCS charging system and RCP seal injection flows.

ANSWER / DISTRACTOR ANALYSIS A. FV-0121 would modulate open due to PRZR level program would raise to 60% due to the NR TI failure. This would cause charging flow to increase which is correct.

Seal injection valve HC-182 would not change position and since downstream of FV-01 21 as charging flow raises, seal injection flows would also raise.

B. Correct.

C. FV-0121 would modulate open due to PRZR level program would raise to 60% due to the NR TI failure. This would cause charging flow to increase which is incorrect first half of this choice. Seal injection valve HC-1 82 would not change position and since downstream of FV-01 21 as charging flow raises, seal injection flows would also Page 12 of 208

HL-16 NRC Written Examination KEY raise which is correct in 2nd part of this choice.

D. FV-0121 would modulate open due to PRZR level program would raise to 60% due to the NR TI failure. First part of this choice is incorrect. Seal injection valve HC-182 would not change position and since downstream of FV-0121 as charging flow raises, seal injection flows would also raise which is also incorrect.

REFERENCES Farley March 2010 NRC RO exam question # 5 use as a base for inspiration.

V-LO-PP- 16101 slides#20 and 31.

V-LO-PP-16302 slides #43 thru 51.

V-LO-PP-09200 slides # 77 and 78.

VEGP learning objectives:

LO-PP-16302-01 Describe the Pressurizer level program and explain why level is programmed with Auctioneered Tavg.

LO-PP-16302-03 Describe how Pressurizer level control maintains level on program.

LO-PP-09200-01 State the purpose and describe the control signals, setpoints, and any interlocks for the following:

a. VCT outlet valves, LV-1 12B, LV-1 12C

b. RWST suction supply valves to charging pumps, HV-1 12D, HV-1 12E

c. charging flow control valve, FV-121

d. seal injection flow control valve, HV-1 82

e. pressurizer auxiliary spray valve, HV-8145

f. RCS normal charging valve, HV-8146 including the check valve around the normal charging valve

g. RCS alternate charging valve, HV-8147 Page 13 of 208

0 Auctioneered HIGH Tavq

- Rod Control (temperature deviation)

- Steam Dumps (temperature deviation)

- Pressurizer Level Control (determines level setpoint)

- Rod Insertion Limit Computer (zeroed out and not usually listed)

Auctioneered LOW Tavq

- C-16: Stop Turbine Loading (RCS C/D)

553O OR 200 below Tref (Tref Tavg)

Auctioneered HIGH AT

- RIL Computer

- Compared to COLR, figure 3 V-LO-PP-1 6101-04.1 20

C Q Question:

Unit 1 is at 100% Power. Loop 4 Thot RTD fails HIGH.

How do the following respond?

A. Loop 4 Tavg? Increase B. Loop 4zT? Increase C. Loop 4 OT T setpoint? Decrease D. Loop 4 OPAT setpoint? Decrease E. Rod Control in AUTO? Insert F. Steam Dumps in Tave Mode? Demand will increase G. Pressurizer Level Control? Charging flow will increase V-LO-PP-1 6101-04.1 31

CVCS Charging HV-182 RCP Seal Injection F V.121 Charging Pump FIou Normal Charging rio RCS V-LO-PPO92OO-O2. 1 77 Operatedfrom the QMCB.

Air operated valve that fails OPEN on loss of air.

Used during norm al plant operations to maintain RCP seal injection flow rates between 8 to 13 gpm.

When F V.121 is opened; charging flow is directed toward the RCP seal injection line and up to HV-182 which is at maximum output, i.e., HV.182 fully SHUT. (NOTE that HC-182 is a MANUAL Controller only.)

The UP and DOWN arrows on controller HC.182 are controlling the process parameter Seal injection flow!

Depressing the UP pushbutton shuts HV.182 and raises seal injection flow.

Depressing the DOWN pushbutton opens HV-182 and lowers seal injection flow.

Some procedure sections of SOP 13006.1/2 may have HC-182 at minimum output, in which case the valve will be full OPEN.

Now when HC.182 is opened to dire Ct flow toward the normal charging header, seal injection flow will lower.

HOW DO YOU PREVENT LOSING ALL OF THE SEAL INJECTION FLOW This is where the SOP directs you on how to OPERATE!

SIMULTANEOUSLY, PERFORM the following:

(1) ADJUST Seal Flow Control 1-HC-0182 to obtain between 8 and 13 gpm.

(2) ADJUST Charging Flow Control 1.FIC-0121 to obtain the desired charging flow.

This procedure direction allows you to incrementally open HC-182 by depressing the DOWN ARROW.

When you do that, total charging flow will remain essentially the same, but what youve done is change the direction of some of the charging flow. A parallel path is now open the normal charging header and the seal injection header. The effect of opening HV.182 will be lower seal injection flows. This parameter and total charging flow are values to be monitored when performing this evolution.

To maintain total seal injection flow between 8.13 gpm, F V.121 output will have to be increased as well. That is total flow in FI.121 has to increase. If charging flow needed to be raised to 87 gpm as indicated on FI-121, then approximately 32 gpm would be directed toward the seal injection header with the remaining V-LO-PP-O2(u1l the normal charging header. 77

CVCS Charging HV-182 RCP Seal Injection 182 32 gpm F V.121 AI$r1i

(:;i)

S7gpm Charging Pump Flow hh[4 ..._..Norrnalcharging 4

V-LO-PP-09200-02. 78 Some procedure sections of SOP 13006-1/2 may have HC-182 at minimum output, in which case the valve will be full OPEN.

Now when HC-182 is opened to dire Ct flow toward the normal charging header, seal injection flow will lower.

HOWDO YOUPRE VENT LOSING ALL OF THE SEAL INJECT1ONFLOW?

This is where the SOP directs you on how to OPERATE!

SIMULTANEOUSLY, PERFORM the following:

(1) ADJUST Seal Flow Control 1-HC-0182 to obtain between 8 and 13 gpm.

(2) ADJUST Charging Flow Control 1-FIC-0121 to obtain the desired charging flow.

This procedure direction allows you to incrementally open HC-182 by depressing the DOWN ARROW.

When you do that, total charging flow will remain essentially the same, but what youve done is change the direction of some of the charging flow. A parallel path is now open the normal charging header and the seal injection header. The effect of opening HV.182 will be lower seal injection flows. This parameter and total charging flow are values to be monitored when performing this evolution.

Solo maintain total seal injection flow between 8-13 gpm, F V.121 output will have to be increased as welL That is total flow in FI-121 has to increase. If charging flow needed to be raised to 87 gpm as indicated on FI-121, then approximately 32 gpm would be directed toward the seal injection header with the remaining directed toward the normal charging header.

3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

LCO 3.5.5 Seal Injection Flow LCO: Reactor coolant pump seal injection flow shall be within limits.

APPLICABILITY: MODES 1, 2, and 3.

SR 3.5.5.1 Not required to be performed until 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after the Reactor Coolant System pressure stabilizes at 22lSpsig and 2255 psig.

V-LO-PP-09200-02. 1 78

Pressurizer Level Control Auct High Tavg r

L Setpoint r

Level Setpoint 43 V-LO-PP- 16302 Rev-02V-LO-PP- 16302-02

A I-

.I[1I]

ICHI1 ii ,

rdD

-J 19 ] J iIIii:lJj

.flHhI C 0 a

Pressurizer Level Control LT-460 (II) LT-461 (III) s.p.

Setpoint 2014016018011j Pressurizer Al High Level PZR B/U Low Level Level Deviation Heaters Deviation Recorder Close Close LV-460 LV-459 Heaters Off High Flow Low Flow Flow Indicator FI-121A,B,C V-LO-16302-02

Pressurizer Level Control (Auto Control)

1) The setpoint that is developedfrom the Auct.

high Tavg card in the NSSS (7300) control cabinet is sent to the Pressurizer Master controller (LIC-This is also a card in the NSSS control cabinet.

The Pressurizer Master controller compares the setpoint to the selected pP\$Ø:qtpiip Pressurizer LevEl Transmitter and develops a charging flow 46 demand

p C

((11E]I I EtOi HDffEi.

Wc C)

S 1

4 LI

Pressurizer Level Control (Auto Control)

FIC-121, which is just another card in the NSSS control cabinet, compares the demandedflow signalfre master ontroller to actual charging flow and position FV-121 as necessary to achieve the desiredflowrate.

48 V-LO-PP- 16302 Rev-02V-LO-PP- 16302-02

Pressurizer Level Control LT-460 (II) LT-461 (III) LT (I) Auct High Tavg 1

II 4

201401601801100 I

Pressurizer High Level PZR BIU Low Level Level Alarm Deviation Heaters Deviation Alarm on Recorder Setpoint Close Close LV-460 LV-459 Interlock Heaters Off Close Orifice High Isolation Valves Peff3nded floii Low Flow signal Flow U Indicator Fl- 121A,B,C FV- 121 V-LO-PP- 16302-02

C.1 iIJ:

Iia 0

Pressurizer Level Control LT-460 (II) LT-461 (III) Auct High Tavg s.P.

High Level PZR B/U Low Level Level Deviation Heaters Deviation Alarm on Alarm Recorder Setpoint Close Close LV-460 LV-459 Interlock Heaters Off High Flow Low Flow FI w

0 U Indicator Fl- 12 1A,B,C FV-121 V-LO-PP-16302-02

5. 004K4.13 001 Given the following conditions:

- The reactor is at 60% power.

- Pressurizer level is on program.

- 120 gpm letdown is in service.

- FV-01 21, Charging Flow Controller is in Auto.

The OATC inadvertently places letdown_(1 )hands witch to the CLOSED position.

This will result in FV-0121, Charging Flow Controller initiall y modulating in the

_(2)_ direction.

(1) (2)

A. orifice HV-8149A OPEN B orifice HV-8149A CLOSED C. isolation LV-459 OPEN D. isolation LV-459 CLOSED Page: 3 of 84 1/3/2011

HL-16 NRC Written Examination KEY

7. 005A1 .07 OOII2IIIRHR-TS SURVI2.513. I CIA/BANK FARLEY 20101R0/SROINRCIGCW A stroke time test of 1-HV-8811A RHR Sump Suction in the open direction has been performed per OSP 14825-1, Quarterly Inservice Valve Test.

Open direction HIGH LIMIT is 14.8 seconds.

Open direction MAX STROKE TIME is 20 seconds.

- Stroke time obtained is 15.9 seconds.

The 1ST Engineer was notified and the valve was retested.

- Stroke time obtained is 15.7 seconds.

Which ONE of the following describes 1-HV-881 1A OPERABILITY lAW Technical Specifications and actions required per OSP 14825-1?

Ab 1-HV-8811A is OPERABLE Analyze stroke time data results to determine if new stroke time represents acceptable valve operation.

B. 1-HV-8811A is INOPERABLE Analyze stroke time data results to determine if new stroke time represents acceptable valve operation.

C. 1-HV-881 1A is OPERABLE Repair or replace 1 HV-8811A.

D. 1-HV-8811A is INOPERABLE Repair or replace 1 HV-881 1 A.

Page 14 of 208

HL-16 NRC Written Examination KEY Feedback 005 Residual Heat Removal System (RHRS)

Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RHRS controls including:

(C FR: 41.5/45.5)

Al .07 Determination of test acceptability by comparison of recorded valve response times with Tech-Spec requirements K/A MATCH ANALYSIS The candidate is given a plausible scenario where an RHR valve exceeds the High Limit during stroke time testing but is within the Max Stroke Time limit. The candidate must determine if the valve is OPERABLE and correct actions to perform.

ANSWER / DISTRACTOR ANALYSIS A. Correct. Valve is operable and 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months to determine if new stroke times acceptable.

B. Incorrect. Valve is operable until 96 hour0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months evaluation is performed.

C. Incorrect. Valve does not need to be replaced until 96 hour0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months evaluation is performed.

D. Incorrect. Valve does not need to be replaced until 96 hour0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months evaluation is performed.

REFERENCES OSP-1 4825-1, Quarterly Inservice Valve Test pages 88, 89, and 106.

Farley 2010 NRC RO Exam questions # 6 VEGP learning obiectives:

LO-PP-12101-04 State the Design temperature and pressure rating for the RHR system Page 15 of 208

I Approved By J. B. Stanley I Vogtle Electric Generating Plant Procedure Number Rev 14825-1 90.1 Date Approved I Page Number 0

/

1 (j

2 010 4 I QUARTERLY INSERVICE VALVE TEST 88 of 125 6.0 ACCEPTANCE CRITERIA NOTE The Acceptance Criteria is generic to all Data Sheets.

6.1 The Stroke Time Test results are recorded on the applicable data sheet as SAT indicating that the Actual Stroke Time of each valve tested did NOT exceed the low limit QE the high limit.

62 The Exercise, Fail Safe AND Position Indication Verification Test results are recorded on the appUcable data sheet as SAT IndIcating that each valve tested exercised satisfactorily, failed CLOSED on loss of power, AND the position indication corresponded to actual valve position (recorded in Step 5.2.3).

Printed January 17, 2011 at 15:05

I Approved By Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant 14825-1 90.1 I Date Approved Page Number 1/04/2010 QUARTERLY INSERVICE VALVE TEST 89of 125 7.0 EVALUATION AND REVIEW Actions required jf Acceptance Criteria are NOT met QE valve exceeds High OR Low Acceptance Time limits:

a. Notify SS.

b. if in applicable mode and a valve exceeds the High QE Low Acceptance Time limits:

(1) The 1ST Engineer SHALL be notified AND the valve retested.

IF the second set of data does NOT meet acceptance criteria, the data will be analyzed WITHIN 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months to verify that the new stroke time represents acceptable valve operation QE the valve will be declared INOPERABLE.

IE second set of data does meet acceptance criteria, the cause of the initial deviation SHALL be analyzed AND the results documented in the record of tests.

-OR-(2) The 1ST Engineer SHALL be notified AND the valve will be declared INOPERABLE UNTIL such time that the valve is repaired, replaced, Q the data analyzed to determine the cause of the deviation AND the valve shown to be operating acceptably.

c. IF in Applicable Mode AND valve exercised unsatisfactorily, did NOT fail CLOSED on loss of power, position indication did NOT correspond to actual valve position, QE maximum allowed stroke time was exceeded; corrective action should be initiated immediately. The valve SHALL be declared INOPERABLE at the time of discovery AND the ACTION statement of applicable Technical Specification(s) entered.

d. IF NOT in Applicable Mode AND a valve requires corrective action, the condition SHALL be corrected PRIOR to entry into the mode requiring the valve to be operable.

e. A retest showing acceptable operation SHALL be completed for each valve requiring corrective action PRIOR to the valve being returned to service AND the plant entering a mode requiring the valve to be operable.

Printed January 17, 2011 at 15:05

Proced4umber Rev J. B. Stanley Vogtle Electric Grating Plant 14825-1 90.1 Date Approved Page Number 11/04/2010 I QUARTERLY INSERVICE VALVE TEST 106 of 125 Sheet 1 of 1 DATA SHEET 6.1 - RESIDUEL HEAT REMOVAL TRAIN A TECHNICAL SPECIFICATIONS APPLICABILITY MODES 1 2, 3, 4 (TEST SECTION 5.3.6)

TEST PURPOSE: PLANT MODE D Surveillance TEST STARTED D Maintenance Retest-MWO#__________

TIME DATE C Other VALVE RESTORED VALVE FAIL POS REF MAX ACTUAL STROKE VALVE(TRAIN) VALVE TO INITIAL TEST INIT SAFETY EXER SAFE IND STROKE LOW HIGH STROKE STROKE DESCRIPTION TYPE TIME POSITION DATE POST POS TEST TEST TEST TIME UMIT LIMIT TIME TIME TEST (INITIALS) 1-HV-8812A(A) MOV 0 0/C Sat N/A Sat O 0 0 20 0 Sat RWST TO RHR C Unsat Unsat C C

1HV51 1A(A) C C Unsat IV MOV 0 0/C Sat N/A Sat 0 0 O 20** 0 Sat RHR SUMP C Unsat Unsat C C

1FVl10(A) C C Unsat IV MOV 0 0/C Sat N/A Sat 0 0 0 10 0 Sat RHR MINIFLOW C Unsat Unsat C C

1HV8716A C C Unsat IV MOV 0 0/C Sat N/A Sat 0 0 0 30 RHR TO HOT LEG 0 Sat C Unsat Unsat C C C C Unsat See Test Section 5.3.6 IV prior to testing.

RWST to CNMT Sump swap-over response time requirement.

Results obtained through performance of this procedure meet ACCEPTANCE CRITERIA of Section 6.0. DYES DNO W NO was checked, refer to Section 7.0, EVALUATION AND REVIEW.

Comments (include any abnormai conditions and corrective actions taken):

Completed By:

Signature Date/Time Supervisory Review:

Signature Date/Time 1ST Review:

Signature Date/Time Printed January 17, 2011 at 15:05

6.005 A 1.07001 1 Z9f0 r- A time stroke of Qi Eli M0V8889, RHR TO RCS HOT LEGS ISO, in the open direction has been performed per STP-1 1.6, Residual Heat Removal Valves Inservice Test.

Open direction ACCEPTABLE STROKE TIME RANGE is 9.96 to 13.47 Sec.

Open direction MAXIMUM ALLOWABLE TIME is 16 Sec.

Stroke times obtained were as follows:

At 1000 First time stroke: 15.35 Secs

At 1005 Second time stroke: 15.52 Secs Which one of the following describes MOV-8889 OPERABILITY lAW Technical Specifications and the actions that are required to be placed in a CR for these results lAW STP-11.6?

A

MOV-8889 is OPERABLE

Analysis of the time stroke results within 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months to determine if new stroke time is acceptable.

B.

MOV-8889 is INOPERABLE

Analysis of the time stroke results within 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months to determine if new stroke time is acceptable.

C.

MOV-8889 is OPERABLE

Repair or replacement of M0V8889.

D.

MOV-8889 is INOPERABLE

Repair or replacement of MOV8889.

Page: 5 of 88 1/3/2011

HL-16 NRC Written Examination KEY L 8. 005A4.03 001/2/1 /RHR-TEMP-HTRS-FLOW/2. 8/2.7 MEMJLOIT BANK/RO/SRO/NRC/GCW Given the following Unit 1 conditions:

- Two RCPs are in service.

- Both Trains of RHR are in service.

- HV-0606 and HV-0607 (RHR Hx outlets) are 50% open.

- RHR flow is throttled to 3000 gpm per train (FV-061 8 I FV-061 9 RHR Hx Bypasses).

- RCS is stable at 180°F and 340 psig.

The Shift Supervisor has directed an RCS heatup to 330°F to begin.

Which ONE of the following:

1) describes the CORRECT actions to take to commence the heatup, and

2) what is the MAXIMUM heatup rate allowed in accordance with UOP 12001-C Unit Heatup to Hot Shutdown (Mode 5 to Mode 4)?

A. 1) Adjust FV-0618 / FV-0619 to establish a heatup rate.

2) 50°F per hour.

B 1) Adjust HV-0606 / HV-0607 to establish a heatup rate.

2)100°F per hour.

C. 1) Adjust FV-061 8 I FV-061 9 to establish a heatup rate.

2)100°F per hour.

D. 1) Adjust HV-0606 / HV-0607 to establish a heatup rate.

2) 50°F per hour.

Page 16 of 208

HL-16 NRC Written Examination KEY Feedback 005 Residual Heat Removal System (RHRS)

Ability to manually operate and/or monitor in the control room:

(C FR: 41.7/ 45.5 to 45.8)

A4.03 RHR temperature, PZR heaters and flow, and nitrogen KIA MATCH ANALYSIS Question gives a plausible scenario where a heatup to Mode 4 from 180 degrees F has been initiated. The candidate must determine the proper heatup rate to establish using RHR and the correct RHR valves to use.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Per Tech Specs and procedure there is a 100 F heatup rate limit.

Plausbile the candidate may confuse heatup rate limit or confuse the valves used to establish the heatup rate. HV-606 / 607 are the valve to use to throttle cooling.

B. Correct. Heatup limited to 100 F per hour and HV-606 / 607 are the correct valves.

C. Incorrect. Per Tech Specs and procedure there is a 100 F hour heatup rate limit.

This is the correct valves to use but the wrong heatup rate.

D. Incorrect. Per Tech Specs and procedure the heatup limit is 100 F per hour and this part is correct. However, this is the wrong RHR valves to use for starting heatup.

Valve to use should be HV-606 / 607.

REFERENCES UOP-12001-C, Steps A4.5.1 and A4.5.2 LOIT Bank 005A4.03-001 VEGP learning obiectives:

LO-PP-1 2101-11 Describe how the RCS cooldown rate is controlled while in shutdown cooling alignment.

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

LO-LP-61209-12, Describe basically how an RCS heatup is performed.

Page l7of 208

Approved By Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant Date Approved 12001-C 71 Page Number 10/24/2010 UNIT HEATUP TO HOT SHUTDOWN (MODE 5 TO MODE 4) 44 of 83 UNIT INITIALS A45 MODE 4 ENTRY AND HEATUP TO 330°F A4.5. 1 Commence RCS/Pressurizer pressure AND temperature trending at 30 minute intervals using Data Sheet 1 AND Plant Computer.

(SR 3.4.3.1 AND TRS 13.4.2.1) (1984300518, 1984300519, 1995230447, 1995230631, 1996332008, 1996232012, 1987212942, 1987212943)

Data taking AND plotting may be suspended during holds in the heatup IF the duration is expected to exceed one hour.

CAUTiON If ITC is positive, RCS heatup can be initiated with dilution in progress provid ed Shutdown Margin is monitored and maintained.

A4.5.2 Commence the RCS heatup to 330°F at a rate NOT greater than 100°F in any one hour (TS Limit), by performing the following:

4 NOTE See Figure 1 for recommended heatup path.

a. Hold RCS pressure at 340 psig +/-25 psig (as measured by P1-0403 AND/OR P1-0405) by energizing Pressurizer heaters as necessary.

b. Reduce RHR cooling to establish heatup by adjusting RHR HX Outlet Valves HV-0606 AND HV-0607.

Printed January 17, 2011 at 15:10

1. 005A4.03 OO1/2/1/RHR- TEMP, HTRS, FLO/MEM 2.8/NEWIRINRC RO/TNT I RLM Given the following Unit 1 conditions:

- Two RCPs are in service

- Both Trains of RHR are in service

- HV-606 and HV-607 (RHR Hx outlets) are 50% open.

- RHR flow is 3000 gpm per train.

- PRZR level is being maintained at 25%.

- RCS is stable at 180 degrees F and 340 psig.

- All mode 4 entry requirements have been completed.

The Shift Supervisor has directed an RCS heatup to 330 degrees F to begin.

Which ONE of the following describes the CORRECT actions to take to commence the MAXIMUM heatup rate allowed in accordance with UOP 12001-C Unit Heatup to Hot Shutdown (Mode 5 to Mode 4)?

A. Adjust FV-061 8/ FV-61 9 to establish a heatup rate not to exceed the procedural limit of 50 F per hour.

B Adjust HV-606 / HV-607 to establish a heatup rate not to exceed the procedural limit of 100 F per hour.

C. Adjust HV-606 / HV-607 to establish a heatup rate not to exceed the procedural limit of 50 F per hour.

D. Adjust FV-61 8/ FV-61 9 to establish a heatup rate not to exceed the procedural limit of 100 F per hour.

Page: 1 of 2 12/6/2010

K/A 005 Residual Heat Removal System (RHR):

A4.03 Ability to manually operate and / or monitor in the control room.

RHR temperature, PRZR heaters and flow, and nitrogen.

KIA MATCH ANALYSIS Question gives a plausible scenario where a heatup to Mode 4 from 180 degrees F has been initiated. The candidate must determine the proper heatup rate to establish using RHR and the correct RHR valves to use.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Per Tech Specs and procedure there is a 100 F hour heatup rate limit.

Plausible the candidate may confuse heatup rate limit or confuse the valves used to establish the heatup rate. HV-606 / 607 are the valves to use to throttle cooling.

B. Correct. Heatup limited to 100 F per hour and HV-606 / 607 are the correct valves.

C. Incorrect. Per Tech Specs and procedure there is a 100 F hour heatup rate limit.

This is the correct valves to use but the wrong heatup rate.

D. Incorrect. Per Tech Specs and procedure the heatup limit is 100 F per hour and this part is correct. However, this is the wrong RHR valves to use for starting heatup.

Valves should be HV-606 / 607..

REFERENCES UOP-12001-C, Steps 4.5.1, 4.5.2, and 4.5.3 for Mode 4 entry and heatup to 330 F.

VEGP learning objectives:

LO-LP-61209-12, Describe basically how an RCS heatup is performed.

Page: 2 of 2 12/6/2010

HL-16 NRC Written Examination KEY

9. 006A1 .07 001/2/IIECCS-PRESS HI-LOW/3.3/3.6 MEMJNEW/RO/SRO/NRC/GCW Unit 1 is at 100% power.

The following is the status of each ECCS accumulator:

- Accumulator # 1 2 pressure 631 psig N - Cb - 1911 ppm Level 30%

- Accumulator # 2 N 2 pressure 651 psig

- Gb 2100 ppm

- Level 69%

- Accumulator # 3 N 2 pressure 615 psig

- Cb 2000 ppm

- Level 51%

- Accumulator # 4 N 2 pressure 674 psig

- Cb 2590 ppm

- Level 59%

Per Tech Spec 3.5.1, Accumulators, the ECCS accumulator parameters above are A. all within Tech Spec limits.

B. not within Tech Spec limits due to level.

C not within Tech Spec limits due to N 2 pressure.

D. not within Tech Spec limits due to boron concentration.

Page l8of 208

HL-16 NRC Written Examination KEY Feedback 006 Emergency Core Cooling System (ECCS)

Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ECCS controls including:

(CFR: 41 .5/45.5)

Al .07 Pressure, high and low K/A MATCH ANALYSIS The candidate is given ECCS accumulator parameters to monitor and has to determine if the given parameters are within limits per 14000-1, Tech Spec Rounds.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect, accumulator # 3 pressure is too low. All others are within tolerances.

B. Incorrect, accumulator # 3 pressure is too low. All others are within tolerances.

C. Correct. Accumulator # 3 pressure is too low, this could result in an insufficient injection volume too late.

D. Incorrect, accumulator # 3 pressure is too low. All others are within tolerances.

REFERENCES OSP-1 4000-1, Operations Shift and Daily Surveillance Logs.

Technical Specification 3.5.1, Accumulators Tech Spec Bases 3.5.1, Accumulators VEGP learning obiectives:

LO-PP-1 3101-06 State the following parameters for the ECCS pumps:

a. shutoff head (pressure)

b. rated flow (gpm)

c. design features to prevent runout and provide pump miniflow LO-PP-13101-13 Given that accumulator pressure and/or level is abnormally high or low, assess the impact on plant safety that the abnormality could cause following a large LOCA.

Page 19 of 208

Approved By S. E. Prewitt Date Approved VogtIe Electric Generating Plant A Procedure Number Rev 14000-1 87.2 Page Number 6/21/2010 OPERATIONS SHIFT AND DAILY SURVEILLANCE LOGS 5 of 34 DATA SHEET 1 - MODE 1 & 2 MODE Sheet 1 of 10 DATE J TECHSPEC INDICATION METHOD OF VERIFICATION SURV REQ PARAMETER INSTRUMENT DAY NIGHT TOLERANCE CONTAINMENT PRESSURE SR 3.6.4.1 CONTAINMENT LCOIPROC 1P14935 SHALL BE MAINTAINED PRESSURE -0.3 PSIG AND 3 64 WITHIN LIWTS 1.8 PSIG AND (PSIG) 1PI-0937 VERIFY PRESSURE rHANNEL CHECK 1 P1-0934 REQUIRED 3 3.3.2(0)

ESFAS INSTRUMENTATION SR 3.3.2.1 REQUIRED 4 3.3 2(E) 1PI-0936 SHALL BE OPERABLE FCN IC,2C,4C CHANNEL CHECK 1PI-1 0945 ACCIDENT MONITORING SR 33 3 1 9 P1-0937 OR 1 P1-10945 CANNOT BE USED TO SATISFY REQUIRED CHANNELS INSTRUMENT SHALL BE FON 7 . REQUIRED 2 3.3.3 NOTE: PlO ON OMCI HAVE POSITIVE RANGE ONLY.

OPERABLE COMPUTER POINT (B,G,H.J)

P9871 (1)

CHANNEL CHECK (PSIG)

.PONM EACH ACCUbEJLATOR SHALL IPISNIA El OFERASLE VINFY PRESSURE MD SR 3.5.12 535 p545 CCLMJLATOR , P1-3552*

WATER LEVEL NITROGEN 1 AND PRESSURE 1PI-3N3A PING) 873 P540 I_______

IP1-3553A IP1-0557A 12.1 1u-3G5O 1U551 SR 3.5.12 ACCUSSJLATOR 2S2%

U4532 WATER LEVEL NO IL)4558 11-)

73.7%

U4 1L1-CNS

- 1uu IU4SST SACN AOCUS*JLATOR SHALL III LIlA U-CSW BEOPERAILI .3.1 VERIFY BORON 1U4551 CONCENTRA11ON AFYIR -

V0U INCREASE ICCUSULATOR U431r (ADDON (QI PROW THE WATER LEVEL RWS1) MCREISE U-oGSr wir -

REQUEST 1LO3W IU-3Sfl*

- IU-3SW iur 0$TMI ACCUESAATOR WATER LEVELS AT LAST S*111 FOR EACH INDICATOR FROII CURRENT PERFORMANCE OF 14225.1, OPERAT1ONS MONTtR.Y SURVU.*14C1 LOGS. OR AT POINT WI OF MOST RECENT CNRMSTRT SA1#1.E IF LATER. COIARE PRESENT LEVEL ACCUNJLATOR LEVEL INDICATOR TO PREVIOUSLY RECORDED LEVEL FOR FOR 74*

(1) IF CONTAINMENT PRESSURE IS GREATER THAN 0.60 PSIG, INITIATE ACTION THE INDICATOR.

TO PERFORM PRESSURE RELIEF PER 13125-1 COMPLETED BY: DAY: TIME: NIGHT: TIME:

SSREVIEW: DAY: TIME: NIGHT: TIME:

Printed January 17, 2011 at 15:15

Accumulators 3.5.1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.1 Accumulators LCO 3.5.1 Four ECCS accumulators shall be OPERABLE.

APPLICABILITY: MODES 1 and 2, MODE 3 with pressurizer pressure> 1000 psig.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One accumulator A.1 Restore boron 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable due to boron concentration to within concentration not within limits.

limits.

B. One accumulator 8.1 Restore accumulator to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months inoperable for reasons OPERABLE status.

other than Condition A.

C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A or B not met.

C.2 Reduce pressurizer 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months pressure to 1000 psig.

D. Two or more D.1 Enter LCO 3.0.3. Immediately accumulators inoperable.

Vogtle Units 1 and 2 3.5.1-1 Amendment No. 129 (Unit 1)

Amendment No. 107 (Unit 2)

Accumulators 3.5.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.1 Verify each accumulator isolation valve is fully open. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.5.1.2 Verify borated water volume in each accu mulator 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is 6555 gallons and 6909 gallons.

SR 3.5.1.3 Verify nitrogen cover pressure in each 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months accumulator is 617 psig and 678 psig.

SR 3.5.1.4 Verify boron concentration in each accumula tor is 31 days 1900 ppm and 2600 ppm.

AND For each affected accumulator, once within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after each solution volume increase of 67 gallons, that is not the result of addition from the refueling water storage tank SR 3.5.1.5 Verify power is removed from each accumula tor 31 days isolation valve operator when pressurizer pressure is> 1000 psig.

Vogtle Units 1 and 2 3.5.1-2 Amendment No. 96 (Unit 1)

Amendment No. 74 (Unit 2)

Accumulators B 3.5.1 BASES ACTIONS C.1 and C.2 (continued) 1000 psig within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Tim es are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orde rly manner and without challenging plant systems.

D. 1 If more than one accumulator is inoperable, the plant is in a condition outside the accident analyses; therefore, LCO 3.0.3 mus t be entered immediately.

SURVEILLANCE SR 3.5.1.1 REQUIREMENTS Each accumulator valve (HV-8808A, B, C, D) should be verified to be fully open every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This verification ensures that the accumulators are available for injection and ensures timely discovery if a valve should be less than fully open. If an isolation valve is not fully open, the rate of injection to the RCS would be redu ced.

Although a motor operated valve position should not chan ge with power removed, a closed valve could result in not meet ing accident analyses assumptions. This Frequency is considere d reasonable in view of other administrative controls that ensure a misp ositioned isolation valve is unlikely.

SR 3.5.1.2andSR 3.5.1.3 Every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , borated water volume (Ll-0950, 0951

, 0952, 0953, 0954, 0955, 0956, 0957) and nitrogen cover pressure (PI-0960A&B, 0961A&B, 0962A&B, 0963A&B, 0964A&B, 0965A&B, 0966A&B, 0967A&B) are verified for each accumulator. This Freq uency is sufficient to ensure adequate injection during a LOCA

. Because of the static design of the accumulator, a 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Freq uency usually allows the operator to identify changes before limits are reached.

Operating experience has shown this Frequency to be appropriate for early detection and correction of off normal trends.

(continued)

Vogtle Units 1 and 2 B 3.5.1-7 Revision No. 0

Accumulators B 3.5.1 BASES SURVEILLANCE SR 3.5.1.4 REQUIREMENTS (continued) The boron concentration should be verified to be within required limits for each accumulator every 31 days since the static design of the accumulators limits the ways in which the concentration can be changed. The 31 day Frequency is adequate to identify changes that could occur from mechanisms such as stratification or inleakage.

Sampling the affected accumulator within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a 1% volume increase (7% of indicated level) will identify whether inleakage has caused a reduction in boron concentration to below the required limit.

It is not necessary to verify boron concentration if the added water inventory is from the refueling water storage tank (RWST), because the water contained in the RWST is within the accumulator boron concentration requirements. This is consistent with the recommendation of NUREG-1366 (Ref. 6).

SR 3.5.1.5 Verification every 31 days that power is removed from each accumulator isolation valve operator when the pressurizer pressure is

> 1000 psig ensures that an active failure could not result in the undetected closure of an accumulator motor operated isolation valve.

If this were to occur, only two accumulators would be available for injection given a single failure coincident with a LOCA. Since power is removed under administrative control, the 31 day Frequency will provide adequate assurance that power is removed.

This SR allows power to be supplied to the motor operated isolation valves when pressurizer pressure is 1000 psig, thus allowing operational flexibility by avoiding unnecessary delays to manipulate the breakers during plant startups or shutdowns.

(continued)

Vogtle Units 1 and 2 B 3.5.1-8 Rev. 2-11/03

HL-16 NRC Written Examination KEY

10. 006G2. 1.30 OOII2IIIECCS CONTROLS/4.4/4.0 CIAJMOD BANK/RO/SRO/NRC/GCW A Control Room evacuation has taken place.

- All Shutdown Panel B (PSDB) handswitches have been taken to LOCAL.

- All Shutdown Panel A (PSDA) handswitches are still in REMOTE.

- A Safety Injection occurs.

Which ONE of the following is CORRECT regarding:

1) the status of the SI pumps, and

2) when SI Termination criteria is met, WHERE the SI pump(s) should be stopped.

SI Pumr status Location to shutdown the SI pumps A. Train A only running Locally at 41 60V Swgr 1 AAO2 Train A only running Locally at PSDA C. Both Train A and B running Locally at 4160 Swgrs 1 AAO2 and 1 BAO3 D. Both Train A and B running Locally at PSDA and PSDB Feedback 006 Emergency Core Cooling System (ECCS) 2.1.3 Conduct of Operations 2.1.30 Ability to locate and operate components, including local controls.

(C FR: 41.7/45.7)

K/A MATCH ANALYSIS The question gives a plausible scenario where an SI actuation occurs with only one Remote Shutdown Panel transfered to LOCAL control. The candidate must determine the status of the SI pumps and where the pump handswitches are located.

ANSWER I DISTRACTOR ANALYSIS A. Pump status is correct but stopping location is incorrect. SIP A has a locally handswitch on PSDA and can be stopped there versus the local switchgear.

B. Correct answer.

Page 20 of 208

HL-16 NRC Written Examination KEY C. Pump status is wrong as only SIP A would start with the handswitch for SIP B in LOCAL control which would prevent auto start of SIP B. Pumps could be stopped with local handswitch.

D. Pump status is wrong as only SIP A would start with the handswitch for SIP B in LOCAL control which would prevent auto start of SIP B. Both pumps could be stopped with local handswitches.

REFERENCES AOP 18038-1 Rev 32 CAUTION prior to step 11, step # 36.

LO-LP-60327-02-0 1 VEGP learning objectives:

LO-PP-60328-07: 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 LO-PP-60328-09: Describe the actions required if a Safety Injection occurs after transferring component controls to the remote shutdown panel.

LO-PP-60327-06: Describe how the following equipment interlocks are affected after transfer to local control.

a. NSCW Pumps and associated valves

b. PORVs / COPs

c. MDAFW Mini-flow valves

d. Pressurizer heater LO-LO level protection

e. VCT low level protection

f. RHR pump miniflow protection LO-PP-60327-07 Describe the response of ECCS equipment to Safety Injection signal after the Local/Remote transfer switch has been taken to Local position.

Page 21 of 208

Approved By J. B. Stanley Vogtle Electric Generating Plant Procedure Number Rev t

J e Aro a ved 18038-1 32 OPERATION FROM REMOTE SHUTDOWN Page Number

( I27I20l0 PANELS 12 of 123 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED CAUTION If an SI actuation occurs, any components previously transferred to the Shutdown Panels will only actuate if manually actuated by the Operator Diesel Generator sequencing is also disabled for any load previously transferred to the Shut down Panels Components not transferred to the Shutdown Panel should realign to the safety configuration

11. Check Reactor TRIPPED:

11. Trip the Reactor:

Reactor Trip and Bypass . 1 -HS-40002A RX TRIP Breakers OPEN.

(Shutdown Panel A)

Neutron flux LOWERING.

. 1-HS-40002B RX TRIP (Shutdown Panel B)

-OR-Locally open Reactor Trip and Bypass Breakers (CB-B71).

-OR-At the MG Set Control Panels (CB-B71), perform the following:

a. Place MG Sets Generator Circuit Breaker Control handswitches in TRIP position.

_b. Place MG Sets Motor Circuit Breaker Control handswitches in TRIP position.

Printed January 17, 2011 at 15:18

Approved By Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant 18038-1 32 Date Approved OPERATION FROM REMOTE SHUTDOWN Page Number

/27/201 0 PANELS 32 of 123 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

35. Check if ECCS flow should be 35. Do not reduce ECCS flow.

reduced:

. RCS subcooling GREATER _Consult TSC when it is staffed.

THAN 24°F (using Core Exit Go to Step 37.

Temperature and RCS WR pressure)

. RCS pressure STABLE OR RISING

PRZR level GREATER THAN no,

/0

Secondary heat sink:

Total feed flow to SGs -

GREATER THAN 570 gpm.

-OR-WR level in at least one SG

- GREATER THAN 65%.

NOTE Train B is the preferred charging train for a Control Room fire when operating from Remote Shutdown Panels.

36. Reduce ECCS flow by stopping the following equipment:

All but 1 CCP

SI Pumps

RHR Pumps Printed January 17, 2011 at 15:19

1. LO-LP-60327-02 OO1/L0LP60327/LO-TA-600 09/0 00AI .22/I//I The following sequence of events occurs on Unit 1:

Control Room evacuation due to fire.

Control of all components has been transferred to the shutdown panels.

A valid Safety Injection signal is received.

Which ONE of the following describes the resp onse of CCP-A due to the SI signal?

A. CCP-A will automatically start, and can be placed back in standby only after SI is reset.

B CCP-A will not automatically start, but may be manually started from the PSDA at any time.

C. CCP-A will automatically start, and can only be stopped from the PSDA by placing the hand switch in Pull-To-Lock.

D. CCP-A will not automatically start, but the operator may start the pump using the PSDA switch only after the sequencer has com pleted the SI load sequencing.

HL-16 NRC Written Examination KEY 007A 1.03001/2/I /PRT TEMP MONITORING/2.6/2.7 MEMINEW/RO/SROINRC/GCW The following alarm has illuminated:

PRZR REL TANK HI TEMP Which of the following would have caused the PRT Hi Temperature and what would be the FASTEST method to restore the tank temperature to normal?

A CVCS Letdown Relief Valve (downstream of orifice isolation valves).

Cooldown using RMWST fill valve and Recirc valve to the RCDT.

B. CVCS Letdown Relief Valve (downstream of Letdown Hx).

Cooldown using RMWST fill valve and Recirc valve to the RCDT.

C. RHR Pump Suction Relief Valves.

Cooldown using recirculation method through the RCDT Hx.

D. RHR Pump Discharge Relief Valves.

Cooldown using recirculation method through the RCDT Hx.

Page 22 of 208

HL-16 NRC Written Examination KEY Feedback 007 Pressurizer Relief Tank/Quench Tank System (PRTS)

Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PRTS controls including:

(C FR: 41.5/45.5)

Al .03 Monitoring quench tank temperature K/A MATCH ANALYSIS The question require the student to diagnose what was the cause of the PRT temperature rise and also determine one of two plausible methods of cooling down the PRT.

ANSWER / DISTRACTOR ANALYSIS A. Correct-PSV-Si 17 (downstream of CVCS orifice isolation valves) relieves to the PRT. Cooling down using the Spray and Drain method in 13004 is a one hour evolution B. Incorrect-PSV-81 19 (downstream of the CVCS VCT) relieves to the VCT. The cooldown mentioned in A is correct.

C. lncorrect-RHR suction relief valves (PSV-8708A & B) do relieve to the PRT. The cooldown method using the RCDT Hx is an eight hour evolution.

D. lncorrect-RHR discharge relief valves (PSV-8842 & 8856) relieve to the Recycle Holup Tank (RHT). The cooldown method using the RCDT Hx is an eight hour evolution.

REFERENCES Procedures 13004-1, Pressurizer Relief Tank Operation and 17012-i, Annunciator Response for PRT Hi Temperature P&ID Drawing 1X4DB114, 1X4DB115, 1X4DB121, 1X4DB122 VEGP learning objectives:

LO-PP-16301-01 List the sources of input into the PRT.

LO-PP-i 6301-09 Describe the methods for cooling the PAT.

Page 23 of 208

Approved By JB Stanley Procedure Number Rev Vogtle Electric Generating Plant Date Approved 13004-1 19.2 I Page Number 09/21/09 I PRESSURIZER RELIEF TANK OPERATION 27 of 41 INITIALS 4.4.3 PRT Cooldown Using Spray And Drain (One Hour Cooldown)

NOTE Two methods for cooling the. ooling the drain is designed to cool the PRT In 11 nethc nd drains to the Waste Processing ugh the RCDT Hx is designed to c

. izes makeup water use and V *ired to cool the PRT and water usage shoL.., which method to uses 4.4.3.1 Establish communications between the Liquid Waste Processing System Panel (PLPP) 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. (1AB-RA27) 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 1HS-1003A (PLPP)

2 1HS-1003B (PLPP)

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

b. Place REACTOR COOLANT DRAIN TANK LEVEL 1-LC-1003 in MANUAL and open the valve (PLPP).

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

Printed January 17, 2011 at 15:28

WINDOW E03 ORIGIN SETPOINT PRZR REL TANK 1-TE-0468 115°F HI TEMP 1.0 PROBABLE CAUSE

1. One or more of the following valves has lifted or is leaking to the Pressurizer Relief Tank:

a. Pressurizer Safety Valve,

b. Pressurizer (PRZR) Power Operated Relief Valves (PORV)s,

c. Chemical and Volume Control System (CVCS) Letdown Relief Valve 1-PSV-8117,

d. CVCS Seal Return Relief Valve 1-PSV-8121,

e. Residual Heat Removal (RHR) Relief Valves 1-PSV-8708A and B

during shutdown conditions.

2. Safety grade letdown in use and aligned to the Pressurizer Relief Tank.

2.0 AUTOMATIC ACTIONS NONE 3.0 INITIAL OPERATOR ACTIONS NONE 4.0 SUBSEQUENT OPERATOR ACTIONS

1. Determine the actual temperature of the Pressurizer Relief Tank using 1 -Tl-0468 on the QMCB.

2. Check tailpipe temperatures for the Pressurizer Safety Valves

, PRZR PORVs, and CVCS Letdown Relief Valve.

Printed January 17, 2011 at 15:29

Approved By S. A. Phillips Vogtle Electric Generating Plant Procedure Number Rev Date Approved 1701 2-1 19.2 ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 12 ON Page Number 10/5I07 PANEL 1C1 ON MCB 37 of 51 WINDOW E03 (Continued)

3. IF a Pressurizer Safety OR PRZR PORV Valve has actuate d, check valve closes when system pressure is reduced.

4. jf a Pressurizer Safety Valve is open Q fails to close following an actuation, initiate 18004-C, Reactor Coolant System Leakag e.

5. if a PRZR PORV 455A1456A is open QE fails to close following an actuation:

a. Place the Control Switch for the affected valve to the closed position,

b. if the affected valve will NOT close, close the associated Block Valve,

c. Refer to Technical Specification LCO 3.4.11.

6. if the temperature rise is due to the CVCS Letdown Relief Valve being open, isolate letdown and initiate 18007-C, Chemical And Volum e

Control System Malfunction.

7. IF the temperature rise is due to a failed RHR Relief Valve, isolate the affected Train of RHR and Initiate 18019-C, Loss Of Residu al Heat Removal

8. if the temperature rise is due to a failed Seal Return Relief Valve, attempt to Isolate the leak.

9. Restore the Pressurizer Relief Tank temperature to norma l per 13004-1, Pressurizer Relief Tank Operation.

10. IF equipment failure is indicated, initiate maintenance as required.

5.0 COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE

/

REFERENCES:

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HL-16 NRC Written Examination KEY i 2. OO8AK 1.01 001 / I / I /PZR VAPOR SPACE ACC/3 .2/3.7 C/AJLOIT BANK/ROIS RO/NRC/GCW With the Unit operating at 100% power, the Reactor Trips on low Pressu rizer pressure.

Pressurizer Relief Tank (PRT) pressure indicates 35 psig. The crew suspects that a PORV opened inadvertently and is now stuck partially open.

Which ONE of the following confirming indications could be expected if a PORV is stuck partially open?

A PORV relief line temperature stabilized at 281 °F. The opposite PORV and the Code Safety Valves tail pipe temperatures slowly rising.

B. PORV relief line temperature stabilized at 281 °F. The opposite PORV and the Code Safety Valves tail pipe temperatures indicate ambient temperature and stable.

C. PORV relief line temperature stabilized at 259 °F. The opposite PORV and the Code Safety Valves tail pipe temperatures slowly rising.

D. PORV relief line temperature stabilized at 259 °F. The opposite PORV and the Code Safety Valves tail pipe temperatures indicate ambient temperature and stable.

Page 24 of 208

HL-16 NRC Written Examination KEY Feedback 008 Pressurizer (PZR) Vapor Space Accident (Relief Valve Stuck Open)

AK1 .01 Thermodynamics and flow characteristics of open or leaking valves K/A MATCH ANALYSIS Questions gives a plausible scenario where a PORV tail pipe is leaking. The candidate must use steam tables to determine the correct tailpipe temperature and use system knowledge to determine which tail pipes should be stable and/or rising.

ANSWER / DISTRACTOR ANALYSIS A. Correct. All tail pipes connected and all will eventually rise.

B. Incorrect. Temperature part is correct, part about tail pipes is wrong. Plausib le if candidate thinks seperate tail pipes for each valves.

C. Incorrect. Temperature would be correct if student forgets to convert to PSIA.

Portion about tail pipes is correct.

D. Incorrect. Temperature would be correct if student forgets to convert to PSIA.

Part about tail pipes is wrong. Plausible if candidate thinks seperate tail pipes for each valves.

REFERENCES LOIT Bank 008AA2.20 Steam Tables VEGP learning obiectives:

LO-PP-16301-06 Determine the expected tail pipe temperature for a leaking or open PORV.

Page 25 of 208

HL-16 NRC Written Examination KEY

13. 009G2.4.8 001/1/1/SB LOCAJ3.8/4.5C/A/NEW/RO/SRO/NRC/GCW I A Reactor trip is performed from full power per AOP direction.

Which ONE of the following AOPs would no longer be performed upon entering the EOP network?

A 18004-C, Reactor Coolant System Leakage due to inability to maintain PRZR level.

B. 18028-C, Loss of Instrument Air with Instrument Air pressure less than 70 psig.

C. 18034-C, Loss of Class 1 E 1 25V DC Power following a loss of DC bus 1 AD1.

D. 18040-C, Partial Loss of Condenser Vacuum due to vacuum <23 inches Hg.

Page 26 of 208

HL-16 NRC Written Examination KEY Feedback 009 Small Break LOCA 2.4 Emergency Procedures I Plan G24.8 Knowledge of how abnormal operating procedures are used in conjunction with EOPs.

(CFR: 41.10/43.5/45.13)1 KIA MATCH ANALYSIS The questions gives plausible scenarios and associated AOP s that could force a reactor trip / shutdown. The candidate must determine which AQP is not performed in conjunction with the EOP network.

ANSWER / DISTRACTOR ANALYSIS A. Correct. Inability to maintain PRZR level would require a reactor trip and SI per 18004-C and this procedure would not be continued, as RNO steps direct go to E-0.

B. Incorrect. A loss of instrument air with pressure < 70 psig requires a reactor trip and continued performance of 18028-C.

C. Incorrect. A loss of DC bus AD1 or BD1 requires a plant trip and performance of 18034-C in conjunction with the EOPs.

D. Incorrect. A loss of condenser vacuum requires a plant trip and performance of 18040-C in conjuction with the EOPs.

REFERENCES 18004-C, Reactor Coolant System Leakage 18028-C, Loss of Instrument Air 18034-1, Loss of 1 25V Class 1 E DC Electrical Power.

18040-C, Partial Loss of Condenser Vacuum 19001-C, Reactor Trip Response VEGP learning objectives:

LQ-LP-37002-03 State how transitions are made to other proc edures.

Page 27 of 208

Approved By I C. S. Waidrup Vogtle Electric Generating Plant Procedure Number Rev Date Approved 18O341 11 7/4/10 LOSS OF CLASS 1E 125V DC POWE Page Number R

4 of 84 A. LOSS OF 125V DC BUS 1AD1 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTES

This procedure should be performed con current with 19000-C, E-0 REACTOR TR SAFETY INJECTION. IP OR

RCP 1 under-voltage and underfrequency trips will NOT actuate.

See ATTACHMENT A for equipment resp onses, breaker and valve control loss, valv failures from loss of instrument air, and e annunciator failures.

Al. Verify reactor trip.

Al. Perform the following:

_a. Trip the reactor.

_b. Initiate 19000-C, E-0 REACTOR TRIP OR SAFETY INJECTION.

_A2. Initiate the Continuous Actions Page.

_A3. Dispatch an operator to 1AAO2 SWGR Room (CB-A48).

NOTE IF DG1A is NOT running, it can NOT be started.

A4. Check DG1A RUNNING.

_A4. Go to Step A7.

Printed January 17, 2011 at 15:54

Approved By JB Stanley Vogtle Electric Generating Plant Procedure Number Rev 18028-C 26.2 LOSS OF INSTRUMENT AIR PageNumber ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED Check instrument Air header *19.

pressure REMAINS GREATER Perform the following:

THAN 70 PSIG..

_a. Trip the reactor.

..b. Initiate 19000-C, E-O REACTOR TRIP OR SAFETY INJECTION.

_c. Initiate ATTACHMENT A, LOSS OF INSTRUMENT AIR IN MODE 3.

20. Check header pressure STABLE

20. if leakage source can NOT be OR RISING.

isolated, THEN restore/isolate UNAFFECTED unit Instrument Air as follows:

a. Perform one of the following:

IF Unit 1 is selected for the swing compressor, THEN close 2-240 1 -U4-51 0.

-OR-jf Unit 2 is selected for the swing compressor, THEN close 1-2401 -U4-51 0.

_b. Verify swing compressor is running (TB-A-TC1 1).

21. Check Instrument Air header

_21. Go to Step 24.

pressure on P1-9361 GREATER THAN 100 PSIG.

Printed January 17, 2011 at 15:55

Approved By J. B. Stanley Vogtle Electric Generating Plant.

Procedure Number Rev Date Approved 18040-C PARTIAL LOSS OF CONDENSER VACU Page Number UM ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTES

A Main turbine trip will occur at 22.42 Hg.

vacuum lowering [2 out of 2 inst. in any one exhaust hood (IPC group 229)]. An automat ic Rx trip will occur if at or above 40% reac power. tor

MCB recorder PR 6292 displays condenser pressure inches of Hg Vacuum.

SGFP trip will occur at 5.9 psia (12 in Hga

).

Steam Dump operation is blocked when con denser vacuum is less than 24.92 Hg.

lowering, (2 out of 2 inst. in any one exhaus t hood), due to Permissive C-9 lost or blocked.

1. Initiate Continuous Actions page.

2 MonItor Condenser pressure.

a. ffl condenser pressure is at or approaching 25 Hg Vac and lowering, THEN increased monitoring of condenser pressure is required.

3 Monitor turbine trip criteria.

_a. Check condenser vacuum greater than 23 Hg vac. a. Perform the following:

1) jE reactor power equal to or greater than 5%,

THEN trip the reactor:

_a) Initiate 19000-C, E-0 REACTOR TRIP OR SAFETY INJECTION.

Step 3 continued on next page

Approved By J. B. Stanley Vogtle Electric Generating Plant Procedure Number Rev Date Approved 18040-C 1

-3-2010 PARTIAL LOSS OF CONDENSER VACUUM Page Number 4of 7 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

_b) Perform the remainder applicable step of this procedure in conjunction with 19000-C, E-0 REACTOR TRIP OR SAFETY INJECTION.

2) Proceed to Step 4.

Approved By J. B. Stanley Vogtle Electric GeneratIng Plant Procedure Number Rev Date Approved 18004-C 27.2 REACTOR COOLANT SYSTEM LEAKAGE Page Number 82 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 Model or2.

_SECTION B. RCS LEAKAGE (MODE 3 <1000 PSIG AND 4)

-OR-

-OR

_ln 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 QE

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 January 17, 2011 at 15:52

Approved By LB. Staney Vogtle Electric Generating Plant Procedure Number Aev Date Approved 118004-c 27.2 7

REACTOR COOLANT SYSTEM LEAKAGE Page Number 82 A. RCS LEAKAGE (MODE 1, 2, AND 3 WITH RCS PRESSURE >1000 PSIG)

ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

3) IF PRZR level can NOT be maintained greater than 9%,

THEN perform the following:

_a) Trip the Reactor.

_b) WHEN Reactor trip verified, THEN actuate SI.

_c) Go to 19000-C, E-0 REACTOR TRIP OR SAFETY INJECTION.

CAUTION The NCP will NOT have miniflow when the CCP normal miniflow valves are closed.

A4 MaintaIn VCT level usIng automatic *A4. Shift charging suction to the or manual makeup control.

RWST:

a. Open RWST TO CCP A&B SUCTION valves:

LV-0112D

LV-0112E

b. Close VCT OUTLET ISOLATION valves:

LV-0112B LV-0112C Step 4 continued on next page Printed January 17, 2011 at 15:52

Approved By I J.B. Staney Vogtle Electric Generating Plant Procedure Number Rev I Date Approved 19001C 31.1

(/22/2OO8 ES 0.1 REACTOR TRIP RESPONSE Page Number 8 of 25 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED IF output breakers can NOT be opened, THEN perform the following:

. Open the following switchyard breakers for the AFFECTED unit bus:

UNIT 1 UNIT 2 BUS-i: BUS-2: BUS-i: BUS-2:

161720 161910 161540 161660 161730 161720(U1) 161740 161820(U1) 161760

Notify Augusta TMC duty personnel referenced on the POD to verify AFFECTED unit Sub-Station Breaker open:

UNIT 1:GOS-WHT (165058)

UNIT 2: WARTHEN (505220 and 505110)

9. Perform the following:

_a. Check 18009-C, STEAM

_a. Go to Step 9.d.

GENERATOR TUBE LEAK IN -

EFFECT.

b. Perform steps 13, 15, 16, and 17 of this procedure as appropriate.

_c. Go to 18009-C, STEAM GENERATOR TUBE LEAK, step in effect.

_d. Check other AOPs IN EFFECT.

_d. Go to Step 10.

Step 9 continued on next page Printed January 17, 2011 at 15:51

Approved By J.B. Stanley Vogtle Electric Generating Plant Procedure Number Rev 19001-c 31.1

/22/2008 ES -0.1 REACTOR TRIP RESPONS Page Number E

ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

_e. Initiate actions of AOPs in conjunction with remainir

,$ actions of this procedure

10. Check PRZR level control:

_a. Instrument Air AVAILABLE,

a. Perform the following:

1) Establish Safety Grade Charging by initiating 13006, CHEMICAL AND VOLUME CONTROL SYSTEM.

2) Establish Safety Grade Letdown:

a) Open RX HEAD VENT TO LETDOWN ISOLATION VLVs:

. HV-8095A

. HV-8096A HV-8095B

HV-8096B b) Open REACTOR HEAD VENT TO PRT flow control valves as necessary:

HV-0442A

HV-0442B

3) GotoSteplo.d.

Step 10 continued on next page Printed January 17, 2011 at 15:51

HL-16 NRC Written Examination KEY 001/2/1 /PRZR PRESS-PRT/19/3.2 C/A!LOIT BANKJRO/SRO/NRC/GCW 1 Given the following:

HV-8000B is closed and de-engergized to isolate PORV 456 which is stuck open.

- A total loss of 125V DC bus AD1 results in a Reactor trip.

- SI actuates due to inadequate throttling of AFW flow post trip.

- 15 minutes later, PRZR level is now rapidly rising.

- Assuming no further operator actions are taken.

Which ONE of the following would prevent RCS pressure from exceed ing the RCS Pressure Safety Limit?

A. Both PRZR sprays open to control RCS pressure.

B. Both PRZR sprays and 1 PRZR PORV open to control RCS pressu re.

C. Neither PRZR spray opens, 1 PRZR PORV opens to control RCS pressu re.

D Neither PRZR spray opens, PRZR Code Safeties open to control RCS pressure.

Feedback 010 Pressurizer Pressure Control System (PZR PCS)

Knowledge of the effect of a loss or malfunction of the following will have on the PZR PCS:

K6.04 PRT K/A MATCH ANALYSIS Question gives a plausible scenario where a transient is in progress requiri ng a vent path to the PRT. Vent path is currently unavailable. Candidate will have to choose the correct alternate vent path.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. PRZR sprays wont open due to loss of air on SI and loss of AD1 de-energizes the Instrument Air isolations to containment.

B. Incorrect. PRZR sprays wont open due to loss of air on SI and loss of AD1 de-energizes the Instrument Air isolations to containment. Neither PORV will open due to the loss of AD1 and Train B Block valve being shut.

Page 28 of 208

HL-16 NRC Written Examination KEY C. Incorrect. PRZR sprays wont open due to loss of air on SI and loss of AD1 de-energizes the Instrument Air isolations to containment. Neither PORV will open due to the loss of AD1 and Train B Block valv e being shut.

D. Correct. PRZR Safeties are the only relie f flow path. PAZR sprays wont open due to loss of air on SI and loss of AD1 de-energiz es the Instrument Air isolations to containment. Neither PORV will open due to the loss of AD1 and Train B Block valve being shut.

REFERENCES LOIT Bank 010K6.04-O1(HL-15 Audit Ques tion #14)

VEGP learning objectives:

LO-PP-1 6301 Describe the purpose of the follo wing pressurizer components or auxiliaries:

a. Variable Heaters

b. Backup Heaters

c. Spray Valves

d. Bypass Spray Valves

e. PORVs

f. PORV Block Valves

g. Code Safety Valves

h. PRT

i. Surge Line Page 29 of 208

Approved By C. S. Waidrup Vogtle Electric Generating Plant Procedure Number Rev 4OateAProved 18034-1 11 LOSS OF CLASS 1 E 1 25V DC POWER Page Number 84 ATTACHMENT A Sheet 1 of 23 LOSS OF 125V DC BUS 1AD1 EQUIPMENT RESPONSE DUE TO LOSS OF TRA IN A 125V DC POWER NOTE Feeder Breakers must be locally controlled in the event the transfer to an alternate power supply is required. !E DG1A is not running, it may not be selected as an alternate power source.

Main Feedwater Isolation, Bypass Feedwate r Isolation, and Bypass Feedwater Regulation Valves close resulting in Feedwater Isolation.

Main Steam Isolation and Bypass Steam Isola tion Train A Valves close resulting in steam line isolation.

Above P10, Reactor and Turbine trip occurs from loss of main feedwater.

Below P10, Reactor trip occurs from Intermed iate Range Instrumentation.

Control Power is lost to 1AAO2, 1ABO4, 1AB O5, and 1AB15 SWGR Breakers.

DG1A control power to Generator Control Panel PDG1 and Engine Control Panel PDG2 lost rendering the DG inoperable; if running, it is 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 Wate Keep-Warm Pumps will come on. r

Loss of Train A DG AUTO sequencer reset.

Power to Inverters 1AD1I1 and 1AD1I1 1 is lost causing 120V AC Vital Busses 1AY1A 1AY2A to de-energize. and

Instrument Air Containment Isolation Valve 1 -HV-9378 closes resulting in loss of instrume air inside Containment. nt

Power To Isolation Panel 1 ACQIP1 is lost rendering the annunciators in Train A inoperab le.

Pressurizer PORV 1 -PV-455A fails closed

TDAFW Steam Supply 1 -HV-301 9 fails as is.

Printed January 17, 2011 at 16:01

HL-16 NRC Written Examination KEY

15. 011 EA 1.03002/1 / 1/LB LOCA-RCP TRIP/0/4.0 MEMI MOD BANK/RO/SROINRC/GCW Based on the following events:

- Safety Injection has been manually actuated.

- No CCPs are currently running or available.

- SIP A is running, SIP B is tripped.

- RCS pressure has lowered to 1350 psig.

- CETCs indicate 560°F and are slowly rising.

- 19010-C, Loss of Reactor or Secondary Coolant has just been entered.

Which of the following is the CORRECT action to take at this time and why?

A. RCPs should remain running, one SIP running is NOT adequate.

B. RCPs should remain running, at least one CCP running is required.

C RCPs should be tripped, to prevent excessive

) depletion of RCS inventory.

D. RCPs should be tripped, the pump heat may lead to uncovering the core earlier.

Page 30 of 208

HL-16 NRC Written Examination KEY Feedback 011 Large Break LOCA EA1 Ability to operate and monitor the following as they apply to a Large Break LOCA:

(CFR 41.7! 45.5 / 45.6)

EA1.03 Securing of RCPs K/A MATCH ANALYSIS Question presents a plausible scenario where RCS pressure is lower than RCP Trip Criteria but only one SIP is available. Candidate has to determine whether or not to trip RCPs and basis.

ANSWER! DISTRACTOR ANALYSIS A. Incorrect. One SIP running IS adequate for RCP trip criteria.

B. Incorrect. One SIP running IS adequate for RCP trip criteria, a CCP is not required.

C. Correct. This is the basis for RCP Trip Criteria.

D. Incorrect. Basis is to prevent RCPs forcing more water out of core leading to uncovering the core earlier, not from RCP heat inpu t.

REFERENCES 19010-C LORQ Bank RQ-SG-97300-37-01 VEGP learning objectives:

LO-LP-371 11-06 State the RCP trip criteria. Tell why it is especially important in the case of a small break loca.

Page 31 of 208

Approved By J. B. Staney Vogtle Electric Generating Plant Procedure Number Rev Date Approved 19010-C 33.1 El LOSS OF REACTOR OR SECONDARY Page Number J18/10 COOLANT 26 of 26 FOLDOUT PAGE 1 RCP TRIP CRITERIA Trip all RCPs if BOTH conditions listed belo w occur a CCPs or SI pumps AT LEAST ONE RUN NING b RCP Tnp Parameter RCS PRESSURE LESS THAN 1375 psig

2. SI REINITIATION CRITERIA Operate ECCS pumps as necessary if EITHER condition listed below occurs.

RCS subcooling LESS THAN 24°F [38° F ADVERSE].

PRZR level CANNOT BE MAINTAINED GREATER THAN 9% [37% ADVERSE].

Initiate ATTACHMENT A if it is necessary to re-establish CCP Cold Leg Injection.

3. SECONDARY INTEGRITY CRITERIA Go to 19020-C, E-2 FAULTED STEAM GENERA TOR ISOLATION, if any SG pressure is lowering in an uncontrolled manner or has bee n completely depressurized, and has not been isolated.

4. E-3 TRANSITION CRITERIA Go to 19030-C, E-3 STEAM GENERATOR TUB E RUPTURE, if any SG level rises in an uncontrolled manner or any SG has abnormal radiation.

5. COLD LEG RECIRCULATION SWITCHOVER CRITERION Go to 19013-C, ES-1.3 TRANSFER TO COLD LEG RECIRCULATION, if RWST level lowers to less than 29%.

6. AFW SUPPLY SWITCHOVER CRITERION Switch to alternate CST by initiating 13610, AUXILIARY FEEDWATER SYSTEM when CST level lowers to less than 15%.

Printed January 17, 2011 at 16:05

trip times. The change of slope of each curve in Figure 1 represents a shift in break flow quality from nearly zero to one, as the RCS drains to the break elevation. Case A is the Final Safety Ana lysis Report (FSAR) 3-inch diameter break calculation for a 3-loop plant des ign, in which RCP trip at the time of reactor trip was assumed. For Case A, the RCS drains to the break elevation and the break flow changes to all steam flow at approximately 575 seconds after break initiation. Cases B and C represent analyses in which RCP trip occurs prior to the time of RCS drain to the break elevation for Case A.

Figure 1 illustrates that the differen ce in the integrated break discharge is insignificant for these cases. The refore, the liquid mass inventory remaining in the RCS is also comparable, yieldin g peak clad temperatures (PCTs) similar to the FSAR case results, which are belo w the regulatory limit of 2200°F.

Cases D through G represent transients in which the RCPs remain operational for times equal to or greater than the time of RCS drain to the break elevation for Case A, and demonstrate significant differences in the integrated break discharge mass. Forc ed loop flowrates induced by RCP operation maintain the inner vessel mixt ure level above the hot leg nozzle elevation. This allows for continued circulation of liquid around the loo providing a source of liquid to the brea ps, k region. Therefore, continued RCP operation prolongs the period of liqu id break discharge as the RCS drains.

The difference in time of the slope chan ge for the delayed RCP trip cases is additional time of liquid break disc harge. The prolonging of the liquid brea discharge further depletes the liquid k mass inventory remaining In the RCS.

Immediately following RCP trip for thes e cases, loop flowrate decreases and steam-water phase separation occurs. A rapid reduction in RCS mixture level results, which may partially uncover the fuel. Prolonged RCP operation and the resultant additional liquid mass dep letion can greatly affect the degree and duration of core uncovery. Dependin g on plant type and break size, a range of RCP trip times may yield PCTs greater than the FSAR case result. The effect of RCP trip time on calculated PCTs is illustrated in Figure 2.

If RCPs remain operational throughout the transient (Case H of Figure 1) depletion of primary liquid mass is max imized. Nevertheless, PCTs remain below FSAR case results due to enhanced well core cooling caused by the high core steam flowrates indicative of RCP operati on. However, continuous operation RCP TRIP 6 HP/LP Rev. 2, 4/30/2005 RCPTRIP .doc

1.RQ-SG-97300-37 OO1/LO-LP-371 I I-IOILO-TA-37015/009EA2.231B/1.O/3 Based on the following events: MIN /19010-C/GENERIC BKGROIJND Safety Injection has been manually

actuated Due to complications, no CCPs or SIPs are currently running RCS pressure has lowered to 1350

psig CETCs indicate 560°F and are slow ly rising 19010-C, Loss of Reactor or Sec ondary Coolant has just been entered Which of the following is the correct action to take at this time and why?

A RCPs should remain running unt il at least one CCP or SIP can be provide needed core heat removal. started to B. RCPs should remain running unt il at least one of the CCPs can be star because of its higher head injection ted pressure.

C. RCPs should be tripped immediately because they are causing excessive dep of RCS inventory which might lead to letion core uncovery.

D. RCPs should be tripped becaus e the extra pump heat input along wit head pumps will lead to core uncove h no high ry sooner than postulated.

Page: 1 1/17/2011

HL-16 NRC Written Examination KEY

16. 01 1K4.03 003/2/2/PZR LEVEL DENSITY CO/2.6/2.9 MEM/BANK CALLAWAY 07/RO/SRO/NRC/GCW Which ONE of the following Pressurizer Level instruments is COLD calibrated, and how does its indication differ from the other pressurizer level instruments at normal operating temperature and pressure?

Av LT-462; indicates lower.

B. LT-462; indicates higher.

C. LT-459; indicates lower.

D. LT-459; indicates higher.

Page 32 of 208

HL-16 NRC Written Examination KEY Feedback 011 Pressurizer Level Control System (PZR LCS)

Knowledge of PZR LCS design feature(s) and/or interlock(s) which provid e for the following:

(CFR: 41.7)

K4.03 Density compensation of PZR level K/A MATCH ANALYSIS Question ask the distinction between Pressurizer Cold and Hot Calibrated level instrumentation. Matches the KA due to the density compensation of the Cold Cal.

ANSWER I DISTRACTOR ANALYSIS A. Correct B. Incorrect-Plausible because the instrument is correct. LT-462 is calibrated for cold plant conditions and used for heating up or cooling down.

C. Incorrect-Plausible because of the direction of the reading but it is the wrong instrument. LT-459 is Hot calibrated.

D. Incorrect-Wrong instrument and wrong direction of reading.

REFERENCES Callaway NRC 2007 Q # 31 Plant Technical Data Book Tabs 3.2.1 and 3.2.2 (Pressurizer Level Instruments Hot and Cold Cals)

V-LO-TX-1 6001 pages 72-73 VEGP learning obiectives:

LO-PP-1 6302-08 Compare and contrast hot and cold calibrated Pressurizer level indications under various PZR operating conditions using the PTDB.

Page 33 of 208

Approved Vogtle Electric Generating Plant TAO. iv Date Approved Page Number Plant Technical Data Book Unit 1 3 of 14 MISCELLANEOUS CURVES AND CONVERSIONS TAB 3.2.2 PRESSURIZER LEVEL COLD CALIBRATION C

C I

C

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0

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SECTION H PRESSURIZER LEVEL CONTROL 16-56 CONTROL FUNCTIONS AND INTERLOCKS The function of the Pressurizer Level Control System is to maintain a constant mass in the RCS I or all operating conditions (Tavg 557°F to 586.4°F). Since the volume of coolant increases as Tavg increases, the programmed level set point rises from 25%

at no load Tavg of 557° F, to 60% for Tavg of 588.4°F.

Since program level is based on Auctioneer high Tavg, at 100% power, program level is set to 57.8%, due to full load Tavg being 586.4°F. The Pressurizer Level Protection System protects the pressurizer from becoming water solid or from completely draining during plant operation.

16-57 Pressurizer Level Instrumentation The Pressurizer System uses four differential ?

pressure (P) transmitters to sense water level In the .

pressurizer. The transmitters send a level signal to control room indicators, alarms, and protection and control circuits.

All four pressurizer level detectors use the same principle of operation.

Each level instrument is made up of a closed reference leg, a differential pressure (AP) transmitter, and a condensing pot. The P transmitter compares water level pressure of the reference leg to the water level pressure in the pressurizer The condensing pot Is located on top of the 4 reference leg and acts as a collection point for condensation. The condensing pot, which is located on the top of the reference leg, ensures that under normal steady state conditions the reference leg remains full of water In DRAIN VALV order to correctly Indicate pressurizer level. The steam EQUALIZIWG from the pressurizer enters the reference leg condensing VALVE pot. The condensing pot is not insulated which causes the steam to condense when cooled by containment ambient temperature. The condensate pot also seals the reference leg from the hydrogen gas In the pressurizer vapor space, which might give erroneous level Indications if allowed to mix with water in the reference leg. To minimize the penetrations made in the pressurizer, the reference the legs are shared by other instruments. Unwanted actuations could occur if proper planning is not performed draining the common reference legs. (Reference drawing IX4DB112) before During normal operation and when the plant is in hot standby condition, Level Transmitters LT-459, LT 460 and LT-461 are used to provide pressurizer level information. These three instruments are known hot calibrated instruments. These transmitters are calibrated for the normal operational conditions as take into consideration the fact that the reference leg water temperature is close to the containment and ambient temperature of 90°F, and the pressurizer liquid temperature of 653°F. These temperatures produce a relatively constant difference in density, and therefore different head pressures felt by the detectors. As the plant is cooled down, however, the level indicated by these instruments becomes accurate and appears to be higher than the actual level. Level Transmitter LT-462 is calibrated less for cold plant conditions and is used when heating up or cooling down the plant. Unlike the hot calibrated level instruments, the cold calibrated level instrument is calIbrated assuming that the water in the pressurizer is at a lower temperature and more dense. When the water is cooler, it exerts more force (pressure) level transmitter than an equal level of water in the pressurizer under high temperature conditions. on the 72 Revision 7.3

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

aP type level transmitters that are calibrated for normal operating condit ions may be inaccurate under.

abnormal conditions such as a LOCA or steam fine break in the containment. Specifically, the referince leg piping and condensing pots are exposed to the containment atmosp here. 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 th6 reference leg. The pressure that the reference exerts on the level transm itter is less than the pressure tha*1 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 pr rmly present inaccurate ifldlt1onSand under such conditions, must be considered unreli. - ns that may affect presunzer level indicat ion are reference leg leaks or partial draining c nent calibration activities, and a phenomena caused by hydrogen gas coming or oth&tj.LIØnanGe out of solution in the refer flci1e She. .

the reference leg temperature is cooler than the pressurizer, it has a higher affinity for absgjng hydroge gas The hydrogen gas could come out of solution during transie nts The results from aVitie above mention would be reduction in the AP This reduction in zP would cause the pressurizer level lndle.tlon being higher than actual. ....

16-58 Pressurizer Level Control System 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 I .

461 and LT-460, or Channels LT-459 and LT-461. Only

on channels LT-459 and LT-461 can be selected for primary level 40 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 0 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-loa d 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 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 7.3

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Reactor Operator Question #31 Examination Outline Cross- Level RO SRO reference:

Tier# 2 Group# 2 KIA# 011 K4.03 Importance Rating 2.6 Knowledge of PZR LCS design feature(s) and/or interlock(s) which provide for the following: Density compensation of PZR level Proposed Question: Common 31 Which ONE (1) of the following Pressurizer Level instruments is COLD calibrated, and how does its indication differ from the other pressurizer level instruments at normal operating temperature and pressure?

A. LT-461; indicates higher.

B. LT-461; indicates lower.

C. LT-462; indicates higher.

D. LT-462; indicates lower.

Proposed Answer: D Explanation (Optional):

A. Incorrect. Wrong transmitter, and cold cal indicates lower due to lack of density compensation B. Incorrect. Wrong transmitter C. Incorrect. Wrong level indication D. Correct.

Technical Reference(s) OTG-ZZ-00006, pg 8 & 54 (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective:

(As available)

Page 61 of 200

HL-16 NRC Written Examination KEY

17. 01 2A4.07 001/2/1/RPS-MG SET BRKRSI3.913.9 CIAJLOIT BANKJRO/SROINRC/GCW C

Given the following conditions at 100% power:

- All systems in normal alignment except for the following.

- Reactor Trip Breaker testing is in progress.

- RTB A QMCB light indication is green.

- RTB Bypass Breaker A QMCB light indication is red.

During testing, the following QMCB annunciator illuminates.

- ROD DRIVE M-G SET TROUBLE

- Motor Generator (MG) Set # 1 Output Breaker has tripped open.

Which ONE of the following is CORRECT regarding the MG Set Output Breakers and the Reactor Trip Breakers?

A. Reactor trip occurs.

Reactor Trip Breakers are in series, power is lost to the Rod Control Power Cabinets.

B. Reactor trip occurs.

MG Set Output breakers are in series, power is lost to the Rod Control Power Cabinets.

C. Reactor trip does NOT occur.

Reactor Trip Breakers are in parallel, power is maintained to the Rod Control Power Cabinets.

D Reactor trip does NOT occur.

MG Set Output breakers are in parallel, power is maintained to Rod Control Power Cabinets.

Page 34 of 208

HL-16 NRC Written Examination KEY Feedback 012 Reactor Protection System (RPS)

Ability to manually operate and/or monitor in the control room:

(C FR: 41.7/45.5 to 45.8)

A4.07 M/G set breakers K/A MATCH ANALYSIS Question asks about MG set power supply to RTB and RTB Bypass Breakers while RTB testing is in progress. Candidate must choose whether reactor trips and why or why not.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. No Rx trip, MG Set Output Breakers are in parallel.

B. Incorrect. No Rx trip, MG Set Output Breakers are in parallel.

C.lncorrect. MG Set # 1 does not specifically power RTB A.

D. Correct..

REFERENCES V-LO-PP-281 01, slide # 59 LOIT Bank 01 2A4.07 VEGP learning objectives:

LO-PP-271 01-02 State the power supplies for the Rod Control System Page 35 of 208

Reactor Trip Reactor Trip Bypass Breaker Bypass Breaker A B Output Breaker j]

4 j.

Reactor Trip 1

[°g Reactor Trip Rod Control Cabinets Breaker A Breaker B

@WGGutput Breaker V-LO-PP-281 01 -04.1 59 0

1. 012A4.07 OO1/2/1/RPS-MG SET BRKRS/MEM-3.9INEWIROIHL-15 AUDIT/TNTIDS Given the following conditions at 100% power:

- All systems in normal alignment except for the following.

- Reactor Trip Breaker (RTB) testing is in progress.

- RTB A QMCB light indication is green.

- RTB Bypass Breaker A QMCB light indication is red.

During testing, the following QMCB annunciator illuminates.

- ROD DRIVE MG SET TROUBLE

- Motor Generator (MG) Set # 1 Output Breaker has tripped open.

Which ONE of the following is CORRECT regarding the MG Set Output Breakers and the Reactor Trip Breakers?

A. Reactor trip occurs.

Reactor Trip Breakers are in series, power is lost to the Rod Control Power Cabinets.

B. Reactor trip occurs.

MG Set Output breakers are in series, power is lost to the Rod Control Power Cabinets.

C. Reactor trip does NOT occur.

Reactor Trip Breakers are in parallel, power is maintained to the Rod Control Power Cabinets.

Dv Reactor trip does NOT occur.

MG Set Output breakers are in parallel, power is maintained to Rod Control Power Cabinets.

Page: 1 of 2 12/7/2010

KIA 012 Reactor Protection System:

A4.07 Ability to manually operate and/or monitor in the control room:

M/G set breakers KIA MATCH ANALYSIS Question asks about MG set power supply to RTB and RTB Bypass Breakers while RTB testing is in progress. Candidate must choose whether reactor trips and why or why not.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. No Rx trip, MG Set Output Breakers are in parallel.

B. Incorrect. No Rx trip, MG Set Output Breakers are in parallel.

C.lncorrect. MG Set # 1 does not specifically power RTB A.

D. Correct..

REFERENCES V-LO-PP-28101, slide #67.

VEGP learninci objectives:

Page: 2 of 2 12fl/2010

RD Portion of Exam

2. Unit 1 plant conditions: 7t A7Z Reactor power = 50% power Reactor Protection System testing in progress A Reactor Trip Breaker is CLOSED A Reactor Trip BYPASS Breaker is CLOSED B Reactor Trip Breaker is CLOSED B Reactor Trip BYPASS Breaker is OPEN Based on the above conditions, if an operator closes (attempts to close) B Reactor Trip BYPASS Breaker which ONE of the following states (1) the status of the Reactor Trip BYPASS Breakers and (2) the status of the Reactor?

A. (1) Both Reactor Trip BYPASS Breakers will open (2) The reactor will trip

/B. (1) Both Reactor Trip BYPASS Breakers will open (2) The reactor will NOT trip C. (1) Only B Reactor Trip BYPASS Breaker will open (2) The reactor will trip D. (1) Only B Reactor Trip BYPASS Breaker will open (2) The reactor will NOT trip 2

HL-16 NRC Written Examination KEY

18. 01 3K5.01 001/2/1/ESFAS-SAFETY TRAIN/2.8/3.2 MEMJLORQ BANKJRO/SRO/NRC/GCW I is divided into 2 distinct input, logic, and output bay __b______

with 3 or 4 of process control equipment used for the signal processing of unit parameters measured by the field instruments.

Which ONE of the following CORRECTLY describes the above mentioned instrumentation?

A. a- ATWT Mitigation System Actuation Circuit (AMSAC) b-Trains c- Channels Bw a- Solid State Protection System (SSPS) b- Trains c- Channels C. a- ATWT Mitigation System Actuation Circuit (AMSAC) b- Channels c- Trains D. a- Solid State Protection System (SSPS) b- Channels c- Trains Page 36 of 208

HL-16 NRC Written Examination KEY Feedback 013 Engineered Safety Features Actuation System (ESFAS)

Knowi edge of the operational implications of the foil owing concepts as they apply to the ESFAS:

(C FR: 41.5/45.7)

K5.01 Definitions of safety train and ESF channel K/A MATCH ANALYSIS Question is a fill in the blank question where the candidate has to choose the correct description of ESFAS, trains, and channels. Two logic bays, trains, and 3 or 4 input channels. CIA, SLI, SI given to differentiate between RPS and ESFAS.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. AMSAC is for turbine trip and AFW actuation on loss of FW.

B. Correct. SSPS (ESFAS) actuates CIA, SLI, SI, etc.

C. Incorrect. AM SAC is for turbine trip and AFW actuation on loss of FW.

D. Incorrect. SSPS (ESFAS) actuates CIA, SLI, SI, etc. but testing circuitry divided into 2 logic trains with 3 or 4 channels of input instrumentation.

REFERENCES Vogtle Text Chapter 28 for SSPS, RPS, and ESFAS selected material.

VEGP learning obiectives:

LO-PP-281 01-03 Describe how a signal is processed through SSPS. Include the progression of the signal from the input relay to a reactor trip/ESF actuation.

Page 37of 208

COMMITMENTS:

FF 86.008 FF 89.026 FF 89.02 1 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 dMded 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 Inout 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) Muclear jnstrumentation ystem (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 souie.

Input Relay Bay Source Channel I 1AY1A Channel II 1BY1B Channel III 1CY1A Channel IV 1DY1B V-LO-TX-281 01-08.1 42

totally independent actuating system had to be developed to minimize the possibility of this event.

Westinghouse produced the AMSAC system to meet the criteria set by the Nuclear Regulatory Commission (NRC), and the Institute of Electrical and Electronics Engineers (IEEE). The final rule requiring AMSAC was made in 1983 as 10CFR5O.62.

The AMSAC system is designed to be highly reliable, resistant to inadvertent actuation, and easy to maintain. The design functions are to initiate a turbine trip, close the steam generator sample and blowdown isolation valves, and to actuate the Auxiliary Feed Water System independently of the RPS. /

28.24 System Description The AMSAC system is a seismically qualified, digital, microprocessor-based system with the exception of its analog inputs. The RPS system uses all analog circuits and Its components are provided by a different manufacture. The only thing that is common about AMSAC and the Reactor Protection System is their inputs, which are separated by Isolation devices. The AMSAC system is designed to prevent a common mode failure from reducing the protection of the reactor pressure vessel. The AMSAC system Is a non-safety related, non-tech spec related, and therefore non-train related system. Its single cabinet located in the control room provides reliability and redundancy by Its three independent circuits called ALPS.

The actuation ogIc Erocessors (ALPs) are used In a majority voting system The reliability of this majority voting system prevents the failure of a single circuit from causing an actuation. In addition, three-out-of-four low feed water flow coincidence logic and a time delay (that is dependent on turbine load) have been selected to further minimize the potential for inadvertent actuatlons.

28.25 SYSTEM OPERATION The AMSAC system is activated when Turbine Power is AMSAC VARIABLE TIMER raised above 40/o power. 250 Inputs from PT-200 =

i i 505 and PT-506 =

I I I urbine Impulse z io Pressure are usedto 100 determine S7sec(e turbine power -.* 50 level. Both 0 == 2iaJ channels must 40 50 80 100 indicate 40%

power for C-20 % TURBINE LOAD to be enabled. A timer set at 260 second maintains AMSAC active until timed out. The timer (TDD) starts as soon as 1 out of the 2 turbine impulse pressures drop below 40% power. After the timer times out, AMSAC is blocked.

To avoid an inadvertent actuation, another timer is used to allow the feed water system time to recover from the transient. The second timer (TDE) is variable based on auctioneered high turbine power. This timer is in line with the 3 out 4 low feed water flow logic. At higher power levels the AMSAC system is designed to respond faster than at lower power levels. The time delay allows the system time to either control or gives the reactor protection system time to V-LO-TX-281 01-08.1 65

HL-16 NRC Written Examination KEY 1& 014A1 .02 001/2/2/DRPI INDICATION/3.2/3.6 C/AILORQ BANKJRO/SRO/NRC/GCW While performing 12003-C, Reactor Startup (Mode 3 to Mode 2), the OATC withdraws Control Bank C rods and observes the following:

Prior to the rod withdrawal the followinci conditions existed:

DRPI: 42 Steps Group 1 Step Counter: 42 Steps Group 2 Step Counter: 41 Steps After the rod withdrawal the following conditions exist:

DRPI: 42 Steps Group 1 Step Counter: 56 Steps Group 2 Step Counter: 55 Steps The Rod Dev annunciator (1) and the Rod Bank LO-LO-Limit annunciator (2)

A (1) illuminated (2) resets B. (1) illuminated (2) remains illuminated C. (1) does not alarm (2) remains illuminated D. (1) does not alarm (2) resets Page 38 of 208

HL-16 NRC Written Examination KEY Feedback 014 Rod Position Indication System (RPIS)

Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RPIS controls, including:

(C FR: 41.5/45.5)

Al .02 Control rod position indication on control room panels KIA MATCH ANALYSIS The question presents a plausible scenario during a reactor startup where DRPI and demand position do not match within + or 11 steps after a rod withdrawal. The candidate must determine whether the Rod Bank Lo-Lo Limit alarm and the Rod Dev alarms are illuminated or clear.

ANSWER / DISTRACTOR ANALYSIS A. Correct. Rod Bank Lo-Lo Limit clears at 48 steps + or 2 steps. Rod Dev alarm would be illuminated since DRPI and demand> 11 steps difference.

B. Incorrect. Rod Bank Lo-Lo Limit illuminated part is wrong. The part for Rod Dev alarm illuminated is correct.

C. Incorrect. Rod Dev not alarming part is wrong. Rod Bank Lo-Lo Limit remaining illuminated part is also wrong.

D. Incorrect. Rod Dev not alarming part is wrong. Rod Bank Lo-Lo Limit resetting part is correct.

REFERENCES ARP-17010, window D06 for ROD DEV UOP-1 2003-C, Reactor Startup VEGP learning objectives:

LO-PP-27101-21 State the alarms associated with the rod insertion limits; include setpoints and the source of the set points.

LO-PP-27201 -06 State the conditions which will cause the following:

h. Rod Deviation Annunciator Page 39 of 208

Approved By Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant 12003-C 49 Date Approved Page Number 10/24/2010 REACTOR STARTUP (MODE 3 TO MODE 2) 26 of 39 INITIALS NOTE ALB1ODO4 ROD BANK LO-LO LIMIT and ALB1OCO4 ROD BANK LO LIMIT should normally reset when Control Bank C exceeds 48+/-2 steps and 58+/-2 respectively. The calculation that determines the reset point for these alarms has an input from % Delta T. If this input is greater than 0%, the reset may be several steps greater than the 2 step tolerance. Ref CR 2005100263 4.2.18 WHEN Control Bank C exceeds 48+/-2 steps, verify annunciator ROD BANK LO-LO LIMIT (ALB1ODO4) resets.

4.2.19 WHEN Control Bank C exceeds 58+/-2 steps, verIfy annunciator ROD BANK LO LIMIT (ALB1OCO4) resets.

4.2.20 WHEN Control Bank C reaches 115 steps, verify Control Bank D begins withdrawing. (TS SR 3.1.6.3) (1995330520) 4.2.21 IF this startup is a dilution to criticality for LPPT, perform the following:

a. Withdraw Control Bank D to the ARC position in 50 step increments or less as recommended by reactor engineering and approved by the SS.

b. Commence dilution to criticality at ARC, per LPPT-GAEJGBE-01.

c. Log Mode 2 entry time and update IPC per Step 4.2.14.

4.2.22 Verify IR indication comes on scale as source range count rate rises. (SR indication 3 X 102 cps) 4.2.23 WHEN criticality is reached, Log the reactor is critical in the Control Room Log.

Printed January 17, 2011 at 16:51

Approved By Procedure Number Rev S. E. Prewitt Vogtle Electric Generating Plant 17010-1 49 Date Approved ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 10 ON PANEL 1C1 Page Number 10/18/2010 ONMCB 46 of 64 WINDOW D06 SETPOINT ORIGIN ROD DEV 1 -YC-0400 Rod Position Supervision Program 1.0 PROBABLE CAUSE

1. Rod Position Supervision Program:

NOTES

Rod Deviation Computer point values may be found on the IPC by typing ALLRODS on IPC keyboard and pressing ENTER.

IF reactor trip breaker position (UD0006) is invalid, THEN the last known valid position is used.

On a reactor trip (UD0006) all of the plant computer demand positions are set to 0.

Plant computer points K0016 and K0017 are periodically set by Reactor Engineering using Procedure 87046-C.

The Rod Position Supervision Program satisfies the requirement for a Rod Position Deviation Monitor in Technical Specifications SR 3.1.4.1.

a. STUCK ROD (UD0358) Individual DRPI >5 steps (K0018) withdrawn at 30 seconds (K0256) after the reactor trip breakers open (UD0006).

b. SHUTDOWN ROD OFF TOP (UD0359, UC0453 UC0456, -

UC0460) Shutdown Bank demand is typically set at <210 steps (K0016). Alarm is enabled for each bank when bank demand typically >222 steps (KOOl 7) and defeated if the reactor trip breakers are open (UD0006).

c. ROD vs. ROD DEVIATION (UD0360) Shutdown or Control Bank Individual DRPI greater than +/-11 steps (K0015) from DRPI average. Alarm is defeated if the average wide range cold leg temperature (UV0406) is less than 350°F (Ki 003) or if the reactor trip breakers are open (UD0006). (jf wide range Tc is bad or questionable, the alarm is not defeated.)

Printed January 17, 2011 at 16:52

Approved By Procedure Number Rev S. E. Prewtt Vogtle Electric Generating Plant 1701 0-1 49 Date Approved ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 10 ON PANEL 1C1 Page Number 10/18/2O1O ONMCB 47 of 64 WINDOW D06 (Continued) d ROD vs DJATION (UD0361) Shutdown or Control Bank demand grethi +/-11 steps (KOOl 5) from DRPI bank averagei

e. ROD WITHDRAWAL SEQUENCE ERROR (UD0362, UC0149 -

UC0152, UC0249 LJC0252) Control Bank demand positions indicate bank overlap is not equal to 115 step +/-0 step tip to tip or any control bank demand >10 steps (KOOl 3) before all shutdown bank demands are typically >225 steps (K0017). This alarm is defeated following a reactor trip (UD0006).

INDIVIDUAL DRPI AVG DRPI DEMAND SDA GR 1 C0009 C0012

- UCOO1O UC0053 SDA GR 2 C0014 C0017 SDB GR 1 C0019 C0022

- UCCOO1 1 UC0054 SDB GR 2 C0023 C0026 SDC C0028 C0031

- UCOO12 UC0055 SDD C0032 C0035

- UCOO13 UC0056 SDE C0036 C0039

- UCOO14 UCOO6O CBA GR 1 C0041 C0042

- UC0001 UC0049 CBA GR 2 C0045 C0046 CBB GR 1 C0050 C0053

- UC0002 UCOO5O CBB GR 2 C0054 C0057 CBC GR 1 C0059 C0062

- UC0003 UCOO51 CBC GR 2 C0063 C0066 CBD GR 1 C0068 C0069

- UC0004 UC0052 CBD GR 2 C0072 C0074

f. Demand position bad. This occurs if an attempt is made to withdraw shutdown or control banks past their physical limit of 231 steps.

2.0 AUTOMATIC ACTIONS NONE 3.0 INITIAL OPERATOR ACTIONS NONE Printed January 17, 2011 at 16:52

HL-16 NRC Written Examination KEY

20. 017K6.O1 001/2/2/IN-CORE DETECTORS/2.7/3.0 C/AIBANKIRO/SROINRC/GCW Given the following:

- The crew is implementing 19001-C, Reactor Trip Response.

- Natural circulation verification is in progress.

- 2 Core Exit Thermocouples are failed due to open circuits.

The input from these CETCs to the Subcooling Monitor are (1) and the calculated subcooling margin will be (2)

A. (1)failed low (2) higher B (1) failed low (2) unaffected C. (1) failed high (2) lower D. (1) failed high (2) unaffected Page 40 of 208

HL-16 NRC Written Examination KEY Feedback 017 In-Core Temperature Monitor System (ITM)

Knowledge of the effect of a loss or malfunction of the following ITM system components:

(C FR: 41.7/45.7)

K6.01 Sensors and detectors K/A MATCH ANALYSIS The question presents a plausible scenario where a couple of CETs have failed due to an open circuit following a reactor trip. The crew must determine direction the CETs will be failed and the effect on the RCS subcooling calculation.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Indication fails low but the subcooling monitor only uses the 5 highest reading CETs for the calculation, subcooling monitor would be unchanged.

B. Correct.

C. Incorrect. Indication fails low but the subcooling monitor only uses the 5 highest reading CETs for the calculation, subcooling monitor would be unchanged.

D. Incorrect. Indication fails low but the subcooling monitor only uses the 5 highest reading CETs for the calculation, subcooling monitor would be unchanged.

REFERENCES V-LO-PP-17301, Core Exit Thermocouples, slide # 10, slides # 19 and 20 VEGP learning objectives:

LO-PP-1 7301-01 Discuss the Operation of the Core Exit Thermocouples to include:

a. How the operator can determine when they fail

b. Range of operation Page 41 of 208

In-core Thermocuoules and MIDS F

INCOfl TIIEMUOCOUPU oet**TIOH V-LP-PP-17301 Rev-03 10 Objective #1 Thermocouple modes of failure:

a. An open circuit results in a low or zero reading

b. A short would result in a reading of the temperature in the vicinity of the short

c. If water comes in contact with the thermocouple, the water will act as a electrolyte and induces a voltage, which results in a higher reading.

V-LP-PP-17301 Rev-03 10

In-core Thermocuoules and MIDS 1J.LZJ. Auctioneer LOW Wide Rang. RCS Presaure Opu t

k 1 ran Auction..r HIGH Quadrant r+/-i Jr

.4.1 Subcooling Avereg.

at Two Quadrant

4

\.QuadraNrant Data Processing Unit V-LP-PP-17301 Rev-03 Indication Subcooling Margin = TSAT TMQA

TSAT = Standard steam table value based on RCS WR pressure (auctionee red low).

TMQA = Maximum quadrant average of Core Exit Thermocouples.

Both the PSMS and IPC computers display RCS subcooling however actual subcooling calculation is performed by the DPUs.

Note temperature indicating range of the PSMS is 200°F to 2300°F V-LP-PP-1 7301 Rev-03 19

In-core Thermocuoules and MIDS C)

.-*. Ipc I

DPUA CH. I Quadrants _.. Ipc 1,23&4 APU A3 Th.rmocouples CH. III IPc

-1

}J3/W36R!P Quadrants RPU 83 Ipc N32Nt I 1,23&4 Th.rmocoupl.s I CH.IV Sb Pb*oor at V-LP-PP-1 20 CORE EXIT TICs indication on IPC Average of the 5 highest Core Thermocouples Note temperature indicating range of the IPC is 0°F to 2300°F V-LP-PP-1 7301 Rev-03 20

1. 017K6.01 00 1/2/2/IN CORE TEMP-SNS/DET/MEM-2.71B-HARRIS 09/HL-15R AUDIT/RO/TNT / DS Given the following:

- The crew is implementing 19001-C, Reactor Trip Response.

- Natural circulation verification is in progress.

- 2 Core Exit Thermocouples are failed due to open circuits.

The input from these CETs to the Subcooling Monitor are (1) and the calculated subcooling margin will be (2)

A. (1) failed low (2) higher B (1) failed low (2) unaffected C. (1) failed high (2) lower D. (1) failed high (2) unaffected Page: 1 of 2 12/6/2010

HL-16 NRC Written Examination KEY NRC/GCW

21. 022A4.05 OO1/2/I/CC-TEMP/PRESSJHUMID/3.8/3.8 MEM/BANK SUMMER2009/RO/SRO/

ratures Which ONE of the following describes the limitation on Containment Air Tempe in Technical Specification 3.6.5, Containment Air Temperature?

A. The HIGHEST temperature reading cannot exceed 120°F.

B. The HIGHEST temperature reading cannot exceed 135°F.

C The AVERAGE temperature reading cannot exceed 120°F.

D. The AVERAGE temperature reading cannot exceed 135°F.

Page 42 of 208

HL-16 NRC Written Examination KEY Feedback 022 Containment Cooling System (CCS)

Ability to manually operate and/or monitor in the control room:

(C FR: 41.7/45.5 to 45.8) ity system A4.05 Containment readings of temperature, pressure, and humid K/A MATCH ANALYSIS The OATC is Question asks Tech Specs limits on Containment Temperture monitoring.

daily rounds.

required to monitor Containment Average Temperature per Tech Specs ANSWER / DISTRACTOR ANALYSIS s.

A. Incorrect. 120°F is correct but the AVERAGE is what Tech Spec require Specs.

B. Incorrect. 135°F is incorrect and it should be the AVERAGE per Tech ARP for 135°F is plausible as a significant temperuture mentioned in the HVAC NIS conduit temperature must be maintained less than 135°F.

rature to not C. Correct. Tech Spec 3.6.5 requires Containment Average Air Tempe exceed 120°F.

D. Incorrect. 135°F is incorrect and it should be the AVERAGE per Tech Specs.

ARP for 135°F is plausible as a significant temperuture mentioned in the HVAC NIS conduit temperature must be maintained less than 135°F.

REFERENCES Techincal Specification 3.6.5, Containment Air Temperature V. C. Summer 2009 NRC RO Exam question # 20 VEGP learning obiectives:

re and LO-PP-291 01-08 Describe routine actions taken to adjust Containment pressu temperature.

e during LO-PP-29101-09 State the likely sources of Containment pressure increas normal operations.

Page 43 of 208

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 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met.

B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.5.1 Verify containment average air temperature is 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months within limit.

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

Amendment No. 74 (Unit 2)

1. 022 A4.05 003/MODIFIED/ILOWERJ/RO/SUMMERJ2/2009/NO Which ONE (1) of the following describes the limitation on Containment Air

Temperature in Technical Specification 3.6.1.5, Containment Air Temperature?

A. The HIGHEST temperature reading cannot exceed 120°F.

B. The HIGHEST temperature reading cannot exceed 135°F.

C The AVERAGE temperature reading cannot exceed 120°F.

D. The AVERAGE temperature reading cannot exceed 135°F.

Page: 1 of 4 12/6/2010

A. 1 St part plausible because this is the correct T.S. temperature limit for containment. (Ref T.S. 3.6.1.5).

2nd part plausible because there are nine temperature elements that are utilized for determining the average containment temperature. (Ref CAP -106.1 Attachment Ill ppg 10-13). It would be conservative to use a peak/highest temperature.

Incorrect because this is not the method for determining the containment temperature defined in CAP-i 06.1 attachment III ppg 10-13. Surveillance requirement 4.6.1.5 states: The primary containment average air temperature shall be the arithmetical average of the temperature at or above the following locations. 1) Elevation 412 3 locations. 2) Elevation 436 3 locations. 3)

Elevation 463 3 locations. This does not, however, completely describe the method of averaging these temperatures as detailed in OAP-106.1 B. 1St part plausible because this is the High Temperature Alarm for the discharge of the Reactor Building Cooling Units. (Ref XCP-606 1-3). Also plausible because 135°F is the alarm setpoint for every HI TEMP alarm in the RB on the HVAC panel (see XCP-6210-LCB1).2nd part plausible because there are nine temperature elements that are utilized for determining the average containment temperature. (Ref CAP -106.1 Attachment Ill ppg 10-13). It would be conservative to use a peak/highest temperature.

Incorrect 1 st part because this is not the correct T.S. temperature limit for containment. (Ref T.S. 3.6.1.5). Also incorrect 2nd part because this is not the method for determining the arithmetical average for the T.S. reading of average containment temperature defined in CAP-i 06.1 Attachment Ill ppg 10-13.

Surveillance requirement 4.6.1.5 states: The primary containment average air temperature shall bee the arithmetical average of the temperature at or above the following locations. 1) Elevation 412- 3 locations. 2) Elevation 436- 3 locations. 3) Elevation 463 3 locations. This does not, however, completely describe the method of averaging these temperatures as detailed in OAP-106.1 C. CCRRECT: This is the correct T.S. temperature limit for containment. (Ref.

T.S. 3.6.1.5), and this represents the correct method for determining the T.S.

average containment temperature reading. (Ref. OAP-106.1 Attachment Ill ppg 10-13).

D. 1st part plausible because this is the High Temperature Alarm for the discharge of the Reactor Building Cooling Units. (Ref XCP-606 1-3). Also plausible because 135°F is the alarm setpoint for every HI TEMP alarm in the RB on the HVAC panel (see XCP-6210-LCB1).

2nd part plausible because it is correct and represents the correct method for determining the T.S. average containment temperature reading. (Ref.

OAP-106.1 Attachment Ill ppg 10-13).

Incorrect 1st part because this is not the correct T.S. temperature limit for containment. (Ref T.S. 3.6.1.5).

Page: 2 of 4 12/6/2010

K/A 022 A4.05:

(Containment Cooling System) Ability to manually operate and/or monitor in the control room:

Containment readings of temperature, pressure, and humidity system.

Tier:

Group:

Importance Rating: RO 3.8 Technical

References:

OAP-106.1, Attachment Ill, Pages 10-13 of 21

T.S. 3.6.1.5 Proposed references to be provided to applicants during examination:

None Learning Objective: AB-13-17, 18, & 19 Question History:

MODIFIED (Although written from scratch, this question is similar enough to Closed Reference questions TECH SPEC 93 & 94 to be classified as MODIFIED) 10 CFR Part 55 Content: 41(b)(7)

K/A Match:

The KA is matched because the operator must monitor RB temperature (via the Control Building Operator T.S. logs)

Page: 3 of 4 12/6/2010

NRC Form ES-401-9 Comments (2009 NRC Exam):

022A4.05 In distracters C and D, by telling the applicant that you average the 3 Temp elements, you are A and B are questionable because of the complexity teaching.

of attachment Ill of OAP-106.1.

All of the s 1 t parts are OK Suggestion for 2 nd part.:

A. The sum of all the temperature elements divided by the averages calculated at each elevation.

B. The sum of all the temperature elements divided by the averages calculated at each elevation.

C. The average of the averages calculated at each elevation D. The average of the averages calculated at each elevation You do NOT need to say the average of the 3 temp second part erature elements at each elevation because it is implied in the 12/31/08 Facility Response:

Version 001 Deleted nine from Choices A & B. Changed the 2 nd part of Choices C &

suggested above. D to read as Feedback from validation indicated that we are testing ROs on T.S. surveillance requ be reserved for SROs. Additionally, as written, irements, which should this is one of 2 questions on the RO exam whic requires knowledge of manual log-taking h specifically (see 02.1.18). With Auto-Log, all of the calcu transparent to the operator. Since operators lations are done rarely take manual logs, we feel that two man over emphasize this topic. With these issue ual log questions s in mind, we revised second part of all choi of a conceptual approach versus a specific ces to make it more distinction in how the surveillance is met.

Duplicate Question. We propose using versi on 003.

Atlanta Review (RFA approved): 06/03/09:

Version 003 OK as-is Page: 4 of 4 12/6/2010

HL-16 NRC Written Examination KEY

22. 022AA2.03 002/I/I/LOSS RX MIUI3. 1/3.6 CIAILORQ BANKIRO/SRO/NRC/GCW

(

With the unit at 100% power and all control systems in automatic, the QATC observes the following:

- ALBO7BO6 Charging Line Hi/Lo Flow is alarming

- ALBO7AO5 Regen HX LTDN Hi Temp is alarming

- Fl-132 letdown flow is oscillating between 0 & 100 gpm

- RCP seal injection flowmeters are off scale high Using the above indications the correct diagnosis and action to take is:

A. Controlling PRZR level channel Ll-459 has failed high.

B. Charging Flow control valve FV-121 has failed closed.

C Seal Flow Control control valve HV-182 has failed closed.

D. Letdown Temperature control valve TV-130 has failed closed.

Feedback 022 Loss of Reactor Coolant Makeup Ability to determine and interpret the following as they apply to the Loss of Reactor Coolant Makeup:

(CFR 43.5/ 45.13)

AA2.03 Failures of flow control valve or controller K/A MATCH ANALYSIS Question presents a plausible scenario giving the candidate various charging flow indications and annunciators. The candidate must determine the correct failure which would cause the indications.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. While LT-459 failing high would reduce charging flow to the point where letdown could flash and the charging Hi/Lo alarm is received, it would not cause seal injection flow to be high, it would be low instead. Therefore, this choice is incorrect.

B. Incorrect. While FV-121 failing shut would reduce charging flow to the point where letdown could flash and the charging Hi/Lo alarm is received, it would not cause seal Page 44 of 208

HL-16 NRC Written Examination KEY injection flow to be high, it would be low instead. Therefore, this choice is incorrect.

C. Correct. HV-1 82 failing shut would starve charging flow to the point letdown would flash and the charging Hi/Lo alrams would be received. HV-182 failing shut would cause seal injection flow to raise to maximum.

D. Incorrect. TV-130 would cause letdown to the VCT temperature to raise but would not affect charging flow and seal injection.

REFERENCES LORQ Bank 022AA2.03-02 P&ID 1X4DB116-1 P&ID 1X4DB139 VEGP learning objectives:

LO-LP-60321 -02 State the fail position of the following valves on loss of instrument air:

a. extraction steam non-return valves

b. feedwater heater high level dump valves

c. HV-182 (charging flow control valve)

d. containment instrument air header isolation valves

e. MSIVs

f. SGFP mini-flow valves

g. FRV

h. FRV bypass

i. RHR heat exchanger outlet valve

j. RHR heat exchanger bypass valve

k. CVCS letdown isolation valve I. Containment isolation valves

m. CVI valves

n. HV-128 (RHR to Letdown valve)

o. FV-121 (Charging Flow Control valve)

LO-PP-09200-06 Given that a partial or complete loss of instrument air has occurred, determine how the CVCS charging system will respond and describe the steps that are required to control RCS inventory and seal injection.

Page 45 of 208

Date: 1/17/2011 Time: 05:05:43 PM

--4 FROM DRAIN SEAL INJECTION VALVE #010 BACKFLUSHABLE FiLTER L. PROJECT CLASS 212 NOTE l8 112O8F40O4

20. HEAT TRACE CABLE H ONLY TEMPERATURE A HEAT TRACE CABLE C

21. VALVE 1HV81O9 MM NCP FLOW RATES CR1

22. FOR PUMP MOTOR CC CONNECTIONS. SEE F!

23. UNES 5031/2 & E NRC Exam\HL-16 REFERENCES FROZEN 12-03-10\P&IDs INFO ONLY\P&tDs hr -

Date: 1/17/2011 Time: 05:10:58 PF) 1 424 415 CD V Cr,

(-)

C-)

U)

NOTE 6 C 12D D

U)

CvCS LETDOWN HEAT EXCHANGER (SEE DWG 1X4DB115) I 0 PROJECT CLASS TUBE SIDE 212 SHELL SIDE 415 11 20AE60O3 /

L__ _J CVcS..

V AC CWS V

________IEThE HEAT I

6 NRC Exam\HL-16 REFERENCES FROZEN 12-03-10\P&IDs INFO ONLY\P&IDs - U

Pressurizer Level Control LT-460 (II) LT-461 (III) LT-459 (I) Auct High Tavg 1

s.P.

20 140 I oat 80 I 100 II Pressurizer High Level High Level PZR B/U Low Level Level Deviation Heaters Deviation Alarm on Alarm Recorder Setpoint 1

Close Close LV-460 Interlock Heaters Off High Flow Indicator FI-121A,B,C V-LO-PP-16302-02

HL-16 NRC Written Examination KEY

23. 022K3.02 OO1/2!1/CCS-CNMT INST/3.O/3.3 MEMJMOD BANK SANOFRE/RO/SRO/NRC/GCW A LOCA resulting in an SI actuation has occured. Containment parameters have increased to the following values:

- Containment Temperature 243°F

- Containment Pressure = 12 psig Which ONE of the following describes effects of the containment environment on the Steam Generator level instrumentation?

Indicated level is...

A. LOWER than actual level due to the high containment TEMPERATURE.

B. LOWER than actual level due to the high containment PRESSURE.

C HIGHER than actual level due to the high containment TEMPERATURE.

D. HIGHER than actual level due to the high containment PRESSURE.

Feedback 022 Containment Cooling System (CCS)

Knowledge of the effect that a loss or malfunction of the CCS will have on the following:

(C FR: 41.7/45.6)

K302 Containment instrumentation readings K/A MATCH ANALYSIS Question presents a plausible scenario where a LOCA has occured affecting containment temperature and pressure. Candidate has to determine the cause and affect on SG level instrumenation.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. SG level would read higher, this is plausible if candidate inverts the affect.

Temperature affect is the correct reason.

B. Incorrect. SG level would read higher, this is plausible if candidate inverts the affect.

SG level would read higher, this is plausible if candidate inverts the affect.

Temperature is the correct reason but pressure is plausible due to adverse contaiment figures used in EOPs are based on pressure (which is indicative of high containment pressure).

Page 46 of 208

HL-16 NRC Written Examination KEY C. Correct.

D. Incorrect. SG level would read higher. Temperature is the correct reason but pressure is plausible due to adverse contaiment figures used in EOPs are based on pressure (which is indicative of high containment pressure).

REFERENCES San Onofre 2006 NRC RO exam question # 21.

WOG Background Documents for Adverse Containment VEGP learning objectives:

LO-PP-29101-08 Describe routine actions taken to adjust Containment pressure and temperature.

LO-PP-29101-09 State the likely sources of Containment pressure increase during normal operations.

LO-LP-361 04-01 List and describe four adverse environmental conditions that affect the reliability of instrumentation associated with critical plant parameters.

Page 47 of 208

2.7.3 Steam Generator Level Tap Relocation Below the Top of the U-Tubes To increase the steam generator level operating band) a number of utilities have lowered the narrow range level tap to below the top of the SG U-tubes or purchased replacement SGs with this feature already in place. With this modification, a typical narrow range level span (e.g.

, for Model F or 05 SGs, or Delta 75 replacement SGs) increases from 128 (10.7 ft) to 233 (19.4 ft).

Although the wider operating band allows for easier contro l without trip during low power operation, the level must be increased to about 15% of the new span to cover the top of the longest U-tubes in these SGs (note: for Model 51 Steam Generators, the required level to cover the top of the SG U tubes increases to 20%; for Model 44/44F SGs or Delta 47 replace ment SGs, this level increases to approximately 25% of the new span). In order to ensure that the top of the SG U-tubes are covered, the EOP setpoints for steam generator level just on span in the narrow range should be increased to the levels noted above plus uncertainties (normal channel accuracies plus refer ence leg process errors up to the transition point for normal/adverse contain ment). A setpoint value of 30% (versus less than 10% previously) would be typical for the modified span. A similar setpoint is expected for the AFW actuation setpoint, an ERG setpoint considered for control of AFW to the rupture d steam generator in the SGTR recovery procedures. Based on a normal containment value of around 30%, and with added reference leg heatup and environmental allowances, it is understandable that the corresponding low steam generator narrow range level setpoints for adverse containment conditions could reach or exceed the recommended maximum value of 50% specified for these setpo ints.

Because of over-fill considerations, the operator is expect ed to not allow the steam generator levels to exceed a certain maximum value, either the upper level tap (minus uncertainties) or the high-high level for feedwater isolation. A somewhat more restrictive limit is used for control of AFW to the ruptured steam generator in the post-SGTR recovery guidelines (ES-3.1, -

3.2. -3.3 and ECA-3.1, -3.2, and 3.3). In these case s, an additional 5%

margin is included to the high-high level setpoint for feedwater isolation to avoid this automatic signal, since receipt of this signal could possibly complicate recovery for some plants. With a typical high-high SQ level setpoint of 80%, the operator should have a reasonable operating band for normal containment conditions (e.g., 30% to 75%). Howev er, when the associated adverse containment uncertainties are subt racted from the high-high GENERIC INSTR 34 HP/LP-Rev. 2, 4/30/2005 GEN-INSTR.doc

r C

mm i m Figure 5. STEAM GENERATOR LEVEL o

MEASUREMENT SYSTEM LAYOUT Steam Nozzle Moisture Separator Swirl Vane Cylinder Narrow Range Noimal Water Level Tube Wrapper U-Tubes Transmitter Trans4tlon Cone

-o Wide I

Range CD r\)

LJ 4

P Transmitter U

Tap 5576 0W06848 054

/LS

ES-401 Sample Written Examination 1-s;i iioef &

Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO Tier#

2 Group# 1 KJA# 022 K3.02 Importance Rating 3.0 Containment equipment subject to damage by high or low temperature, humidity, and pressure readings. Containment instrumentation Proposed Question: Common 13 An ESDE resulting in a SIAS has occurred. Con tainment parameters have increased to the following values:

Containment Temperature 243° F.

Containment Pressure = 12 psig.

Which ONE (1) of the following describes effec ts of the containment environment on the Pressurizer level instruments?

Indicated level is...

A. LOWER than Actual level due to the high containm ent TEMPERATURE.

B. LOWER than Actual level due to the high containm ent PRESSURE.

C. HIGHER than Actual level due to the high containm ent TEMPERATURE.

D. HIGHER than Actual level due to the high cont ainment PRESSURE.

Proposed Answer: C Explanation (Optional):

A. Incorrect. Correct reason but opposite effect of reference leg heating B. Incorrect. Reading will be higher. Pressure effects are minimal, other than that the pressure is elevated due to high temperature C. Correct.

D. Incorrect. Temperature is the cause, pressure in this case is a byproduct of temperature Technical Reference(s) J-BBB-021, TLU Calc. and (Attach if not previously provided)

Setpoint verification for Pressurizer Level Page 25 of 200

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Proposed references to be provided to applicants during examination:

None Learning Objective:

(As available)

Question Source: Bank # A67922 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:

Page 26 of 200

HL-16 NRC Written Examination KEY

24. 025AK 1.01 001 / I / 1/LOSS OF RHRI3.9/4.3 C/A/BANK/RO/SRO/NRC/GCW The following conditions exist on Unit 1:

- The plant is 37 days and 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months into a refueling outage.

- The unit is in Mode 5 at mid-loop conditions.

- RHR 1A heat exchanger inlet temperature is stable at 100 °F.

- Core offload has NOT yet commenced.

- Subsequently the lA RHR Pump trips.

- 18019-C, Loss of Residual Heat Removal is entered.

Which ONE of the following is correct if the loss of RHR continues without mitigation?

REFERENCE PROVIDED A The time to boiling is less than 50 minutes. Promptly initiate actions to protect personnel inside containment and establish containment closure.

B. The time to boiling is more than 60 minutes. Promptly initiate actions to protect personnel inside containment and establish containment closure.

C. The time to boiling is less than 50 minutes. Promptly align SI pumps for Hot Leg Injection and establish Charging flow needed to maintain RCS level.

D. The time to boiling is more than 60 minutes. Promptly align SI pumps for Hot Leg Injection and establish Charging flow needed to maintain RCS level.

Feedback 025 Loss of Residual Heat Removal System (RHRS)

Knowledge of the operational implications of the following concepts as they apply to Loss of Residual Heat Removal System:

(CFR 41.8/41.10/45.3)

AK1 .01 Loss of RHRS during all modes of operation KIA MATCH ANALYSIS An implication of loss of RHR at mid-loops is that there is a small amount of time until the core reaches saturation conditions. This question tests the ability of the applicant to use tools to calculate the time to boil and based on this time, determine required operator actions. Therefore, operational implications of loss of RHR are being tested.

Page 48 of 208

HL-16 NRC Written Examination KEY ANSWER I DISTRACTOR ANALYSIS A. Correct. Step B5 of 18019 gives direction to calculate heatup rate. If the time to boil is less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , then do steps B6 and B7. Steps B6 and B7 contain actions to evacuate personnel and close containment and start fans in low speed.

In this question the time to boil is less than one hour, per Figure 3 (900 hours0.0104 days 0.25 hours 0.00149 weeks 3.4245e-4 months at 100 F initial temp = < 45 minutes).

B. Incorrect. If Figure 4 is incorrectly used, then time to boil is about 67 minutes, leaving 7 minutes until steps B6 and B7 need to be performed. Plausible due to possibility of using the wrong figure. Actions listed to take are correct.

C. Incorrect. If Figure 4 is incorrectly used, then time to boil is about 67 minutes, leaving 7 minutes until steps B6 and B7 need to be performed. Plausible due to possibility of using the wrong figure. Actions listed are from step BlO RNO if CETs reach > 200°F. RHR Hx inlet temperature at 100°F implies CETs still well below 200°F. Action listed is correct for high CET temperature but not for listed conditions.

D. Incorrect. If Figure 4 is incorrectly used, then time to boil is about 67 minutes, leaving 7 minutes until steps B6 and B7 need to be performed. Actions listed are from step BlO RNO if CETs reach > 200°F. RHR Hx inlet temperature at 100°F implies CETs still well below 200°F. Action listed is correct for high CET temperature but not for listed conditions.

REFERENCE 18019-C, Loss of Residual Heat Removal, Rev. 27.1 Figures 3 and 4 of 18019-C, should be provided to the candidates.

VEGP learning obiectives:

LO-LP-60315-03 Given fiqures 1 -5 of AOP 18019-C, determine minimum ECCS flow, time to saturation, time to core uncovery and heatup rate.

Page 49 of 208

Approved By Procedure Number Rev C. S. Waidrup Vogtle Electric Generating Plant 18019-C 27.1

ate Approved Page Number

/3I10 LOSS OF RESIDUAL HEAT REMOVAL 46 of 70 FIGURE 1 RCS HEAT-UP RATE 9

8 Lu I

D 7 z

6 Lu cL.

U 5 CD Lu 4

0 3

2 0

0 50 0 150 200 250 300 350 400 450 500 550 600 650 700 750 800 850 900 950 00 TIME AFTER SHUTDOWN (HAS)

Assumptions:

1) Mid Loop Conditions

2) RCS Vented To Atmosphere With or Without Loop Dams Printed January 17, 2011 at 10:49

Approved By Procedure Number Rev C. S. Waidrup Vogtle Electric Generating Plant 18019-C 27.1

, ate Approved Page Number 3/3/10 LOSS OF RESIDUAL HEAT REMOVAL 47 of 70 FIGURE 2 CORE FLOW TO MAINTAIN 195 DEG F vs TIME AFTER REACTOR SHUTDOWN 1440 1360 1280 w 1200 1120 0

0 1040 w 960 z

I 880 C 800 D 720 0

640 I

I 560 I 0 C

480 400 320 240 160 0 40 80 120 160 200 240 280 320 360 400 440 480 520 560 600 640 680 720 760 800 TIME AFTER SHUTDOWN (HRS)

Assumptions:

1) Mid Loop Conditions

2) RCS Vented To Atmosphere

3) Injection Flow Assumed a 100 Degrees F From RWST Pnnted January 17, 2011 at 10:49

Approved By Procedure Number Rev C. S. Waldrup Vogtle Electric Generating Plant 18019-C 27.1

--4ate Approved LOSS OF RESIDUAL HEAT REMOVAL Page Number

.3/3/10 48 of 70 FIGURE 3 RCS TIME TO BOILING (FULL SPENT CORE) 90

.HLIH!.

I H I 80 -

-H-- i I,

-L 70 H III i L - -.- -

I I H

0 60 --------

I 4 4- - I -

I INITL4LTEMP75 50 I I INITIAL TEMP 100 ci) 40 w H -

I

_ I I - -

I I I INITIAL TEMP 125 z

30 -

INITIAL TEMP 150 20 ftr+h 10 H-H. - -

I

-- H- I L

II 0

0 50 100 150 200 250 300 350 400 450 500 550 600 650 700 750 800 850 900 950 1000 TIME AFTER SHUTDOWN (HRS)

Assumptions:

1) Full Spent Core Heat Load

2) Mid Loop Conditions

3) RCS Vented To Atmosphere With or Without Loop Dams Printed January 17, 2011 at 10:49

Approved By Procedure Number Rev C. S. Waidrup Vogtle Electric Generating Plant 18019-C 27.1

- Jate Approved Page Number 3I3/10 LOSS OF RESIDUAL HEAT REMOVAL 49 of 70 FIGURE 4 RCS TIME TO BOILING (RELOADED CORE) 90 80 70 INITIAL TEMP 100 z I I 0 60 1 I

INITIAL TEMP 125 D 50 I I I

U) 40 :1NITIALTEMP 150 0

I C,) 30 w -

z 20 bi J 13 0

+/-E H 4 JF LL j 0 50 130 0 200 250 300 350 400 450 500 550 600 650 700 750 800 850 900 950 1300 TiME AFTER SHUTDOWN (HRS)

Assumptions:

1) Reloaded Core Heat Load

2) Mid Loop Conditions

3) RCS Vented To Atmosphere With or Without Loop Dams Printed January 17, 2011 at 10:49

Approved By C. S. Waidrup Procedure Number Rev Vogtle Electric GeneratIng Plant 18019-C 271 ate Approved Page Number

/3/10 LOSS OF RESIDUAL HEAT REMOVAL 50 of 70 FIGURE 5 TIME TO CORE UNCOVERY (RCS TEMPERATURE AT SATURATION) 360 340 320 300

>- 280 260 O 240 C)

Z 220 D

iii 200 o 180 160 140 120 D

100 80 60 40 20 0

0 50 100 150 200 250 300 350 400 450 500 550 600 650 700 750 800 850 900 950 1000 TIME AFTER SHUTDOWN (HRS)

Assumptions:

1) Initial RCS Temperature is 212 Degrees F

2) Initial RCS Level at Mid-Loop

3) RCS Vented to Atmosphere With or Without Loop Dams Printed January 17, 2011 at 10:49

HL-16 NRC Written Examination KEY

25. 026K3 02 001/2/1 /CSS-RECIRC/4.2/4.3 C/A/NEW/RO/SROINRC/GCW Containment Spray has actuated due to a Large Break LOCA.

Containment Spray pump A shaft shears upon startup.

Both Containment Emergency Sump Levels read 0%.

RWST level currently reading 5%.

19111-C, Loss of Emergency Coolant Recirculation has been entered froml9Ol3-C, Cold Leg Recirculation.

Which ONE of the following is CORRECT regarding:

1) the Containment Spray system status, and

2) the actions the crew should take regarding the Containment Spray pump?

A 1) Only one Containment Spray ring is receiving flow.

2) Stop Containment Spray pump B.

B. 1) Both Containment Spray rings are receiving flow.

2) Leave Containment Spray pump B running.

C. 1) Only one Containment Spray ring is receiving flow.

2) Leave Containment Spray pump B running.

D. 1) Both Containment Spray rings are receiving flow.

2) Stop Containment Spray pump B.

Feedback 026 Containment Spray System (CSS)

Knowledge of the effect that a loss or malfunction of the CSS will have on the following:

(C FR: 41.7/45.6)

K3.02 Recirculation spray system Page 50 of 208

HL-16 NRC Written Examination KEY K/A MATCH ANALYSIS Question presents a plausible scenario where the crew is in 19111-C, Loss of Emergency Coolant Recirculation with Containment Emergency Sump levels at 0%.

The candidate has to determine the proper spray system operation and action to take regarding the remaining spray pump.

ANSWER I DISTRACTOR ANALYSIS A. Correct, Containment Spray ring headers are separated into 2 distinct trains.

Per note at beginning of 19111-C, any ECCS or CS pump that loses suction should be stopped per CAUTION prior to step 1 of 19111-C.

B. Incorrect. Plausible candidates could think either spray pump would pump to both the ring headers (crosstie), not something you routinely consider. Leaving pump B running is also incorrect as 19013-C does not shut down spray pumps on RWST level and only mention ECCS pumps Spray pumps are stopped only if sump suctions cant be opened on alignment to recirculation.19111-C also has a table that stops spray pumps based on Containment pressure, RWST level, and containment coolers. Candidates may not be aware of CAUTION prior to first step of 19111-C.

C. Incorrect. Containment Spray ring headers are separated into 2 distinct trains.

Leaving pump B running is incorrect as 19013-C does not shut down spray pumps on RWST level and only mention ECCS pumps Spray pumps are stopped only if sump suctions cant be opened on alignment to recirculation.19111-C also has a table that stops spray pumps based on Containment pressure, RWST level, and containment coolers. Candidates may not be aware of CAUTION prior to first step of 19111-C.

D. Incorrect. Plausible candidates could think either spray pump would pump to both the ring headers (crosstie), not something you routinely consider. Per note at beginning of 19111-C, any ECCS or CS pump that loses suction should be stopped per CAUTION prior to step 1 of 19111-C.

REFERENCES 19013-C, Cold Leg Recirculation 19111-C, Loss of Emergency Coolant Recirculation V-LO-PP-15101, Containment Spray System, slide #7 VEGP learning objectives:

LO-PP-1 5101-04 List all components that receive a Containment Spray Actuation signal and their change in status.

Page 51 of 208

HL-16 NRC Written Examination KEY recirculation flow path is performed.

Page 52 of 208

Approved By Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant 1901 3-C 28 Date Approved ES-i .3 TRANSFER TO COLD LEG Page Number

/11/08 RECIRCULATION 4of 19 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTES

FRPs should not be implemented until at least one flow path exists from the CNMT Sump to the RCS Cold Legs and the completion of Step 12.

Steps 1 through 12 should be performed without delay.

The RWST inventory between the RWST LO-LO and Empty alarms is suffici ent for a minimum of approximately 11 minutes of ECCS injection flow assuming that the RHR pumps are isolated from the RWST or stopped within the first 6 minutes after the RWST LO-LO alarm is received.

CAUTION 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).

1. Verify SI Reset.

_2. G.T Emergency Sump level _2.. IE CNMT Sump level IndicatorS EATER THAN Q EQUAL TO LI-764 and Ll-765 are both les 13.5 INCHES than 13.5 INCHES.

THEN stop RHR Pumps A and 0 Ll-764 and go to 19111-C, ECA-1.t Ll-765 LOSS OF EMERGENCY COOLANT RECIRCULATION Printed January 17, 2011 at 17:31

Approved By Procedure Number Rev C. S. Waidrup Vogtle Electric Generating Plant 1911 1 -C 33 e Approved t

JDa ECA-1 .1 LOSS OF EMERGENCY COOLANT Page Number 4/25/10 RECIRCULATION of 48 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED CAUTION If suction source is lost to any ECCS or CS Pump, the pump should be stoppe d

_1. Initiate the Continuous Actions Page.

2. Check emergency coolant *2. Continue attempts to restore at least recirculation capability -

one train of recirculation.

RESTORED.

a. Power available and operable: _WHEN recirculation capability is restored, TRAIN A THEN return to procedure and step in effect.

HV-881 1A CNMT SUMP

- Continue with Step 3.

TO RHR PMP-A SUCTION

RHRPumpA

HV-8809A RHR PMP-A TO COLD LEG 1&2 ISO VLV

. RHR Train A Hx TRAIN B

. HV-881 lB CNMT SUMP TO RHR PMP-B SUCTION

. RHRPumpB

. HV-8809B RHR PMP-B TO COLD LEG 3&4 ISO VLV

RHR Train B Hx Step 2 continued on next page Printed January 17, 2011 at 17:32

Containment Spray V-LO-PP-15101 Rev-3.O 7

V-LO-PP-15101-03, Describe the Con tainment Spray systems normal standby alignment.

In Modes 1, 2, 3, and 4 the Containmen t Spray pumps are in auto standby with the discharge valves shut. Suction valves from the RWST are open.

Spray Components:

Two Containment Spray Pumps (Tr ain A and B) 2600 gpm each Three Spray Headers per train

Containment sump and encapsulation vess els

Three trisodium phosphate storage bas kets Valves and piping V-LO-PP-15101 Rev-3.0 7

HL-16 NRC Written Examination KEY

26. 027AK2.03 OO1/J/1/PZR PRESS CNTRL/2.6J2.8 CIAJLORQ BANKJRO/SRO/NRC/GCW Given the following:

- The unit is at 100% power.

- All control systems are in their normal alignments.

- The Pressurizer Master Pressure Controller output demand fails LOW.

Assuming no action has been taken by the crew, which ONE of the following describes the effect on the Pressurizer heaters, and the resulting effect on the plant?

A PZR heaters energize.

PZR pressure rise is controlled by PZR PORV operation.

B. PZR heaters energize.

PZR pressure rise is controlled by PZR spray valve operation.

C. PZR heaters de-energize.

ONLY the PZR spray valves open, reactor trips on low PZR pressure.

D. PZR heaters de-energize.

PZR spray valves and one PZR PORV open, reactor trips on low PZR pressure.

Page 53 of 208

HL-16 NRC Written Examination KEY Feedback (PZR PCS) Malfunction 027 Pressurizer Pressure Control System the Pressurizer Pressure Control Knowledge of the interrelations between Malfunctions and the following:

(CFR 41.7 I 45.7)

AK2.03 Controllers and positioners K/A MATCH ANALYSIS e the PZR Master Controller fails LOW.

Question presents a plausible scenario wher the plant and the PZR heaters.

The candidate has to determine the effects on ANSWER I DISTRACTOR ANALYSIS A. Correct.

output low, the spray valves will get a B. Incorrect. PZR heaters will energize but with closed signal versus an open.

sible if the candidate inverts the system C. Incorrect. PZR will energize but this is plau ing causing a reactor trip on low response to the demand. The spray valves open t is inverted.

pressure would be the correct response if effec sible if the candidate inverts the system D. Incorrect. PZR will energize but this is plau V would have been the old response if response to the demand. PZR and PZR POR ted by the master controller.

the effect is inverted. PORVs no longer are affec REFERENCES HL-1 5 Audit Exam question # 28 V-LO-PP-16303, PZR Pressure Control VEGP learning objectives:

tor switch is in the NORMAL position LO-LP-60301 -10 Given that the channel selec t will respond to the following pressurizer (455/456), describe how and why the plan separately and include effects on the pressure instrument failures. Consider each alarms, RPS, and ESF actuations.

Pressurizer Pressure Control System response,

a. 455 fails high

b. 455 fails low

c. 456 fails high Page 54 of 208

r I

/ *11 1;

e Pressurizer Master Controller 100%

75% 25 3O % is approximately the normal output with 50% the proportional heaters controlling Pressurizer pressure.

25%4 0%

LO-PP- 16303-03 54

0 0 2 Pressurizer Master Controller 7300 Protection Cabinets PT-458 Isolation Device 55 Controller Output 25% (EY*°&nal 0 psig)

PlC 455A Przr Press Master Controller 9

2310 psi Variable 2345 psig (opens) Heater Przr Hi 2325 psig (closes) Pressure PORV 455 Backup Heaters Przr Control Lo Pressure

& Htrs on 1 63O 56 Pressurizer Controlling Pressure Control PV-455B PV-455C

C) oo rn -D 0 Cl)

CI)

CCD

-I a .aa aa a a a a a.aaaaaaaaaa.aaa.aalaaaaaa iaa .aa a a a al

.aaaaaa.aa Co

a. a a . *

.a a a a a a a a

a a .a a ia a a . a a a a a a a 0

4 a a a a a a a a a a N 4 4

L1 0

C 0 Pressurizer Pressure 58

C 0 Pressurizer Master Controller 100%

75% 34.4% the proportional heaters are fully off 50%

25%

0%

LO-PP- 16303-03 59

C, 0 .

C Cl) C

-CI)

-D - .zo C C) C.)

50 4

I..a..a......a.a.................. I.

I b

N I

Co a.....

I Si (31 0

0 CA) 4

. a a a a. a a.. a . .. . . . I **a ci (3 C

a a....

+

ci

+

C

Pressurizer Master Controller 100%

40. 6%ProportiOflal 75%

Spray Valves start to open.

50%

25%

0% 61 LO-PP- 16303-03

Q P re ss ur iz er M as te r C ontr ol le r 7300 Protection Cabinets PT-458 Isolation Isolation Device F Device 62 Controller Output 40.6% r11°Wgnal 25psig)

PlC 455A Przr Press Master Controller P&l I23lOpsig

>2345 psig (opens) Przr Hi 2325 psig (closes) Pressure PORV 455 Backup Heaters Przr Control La Pressure

& Htrs on Pressurizer Controlling -PP- 16301 63 Pressure Control

i-Th ED Backup Heaters Proportional a I a a I I a a I Heaters I I I

a a a a a a a I a I a ProportionaI I a I a a a I a a a a I

a a a Spr4s a a I a a I a a a a I I a a a a a a a a

. a a a I a I a a a a a a a a I a I a I a a a I a a a a a a a a a a a a a a a a a I a

a I a a I a a a a a a a a I a a a a a a I I a I a a a I a

a a a a I I a I a I a

a a a a I a a PSIG 2195 2210 2220 2235 2250 2355 2218 Error Signal -40 17 -15 0 ÷15 ÷5 ÷75 ÷120 (psig) a a

I a

a Master Controller 0 9.4 14,4 15.6 25 34.4 46 71.9 100 LO-PP- 16303-333 64 Output a

a I

C Pressurizer Master Controller 100%

75% 71.9% Proportional Spray Valves are fully open.

50%

25%

0%

LO-PP- 16303-03 65

C Pressurizer Master Controller 7300 Protection Cabinets PT-458 Isolation Isolation Device Device 66 Controller Output 71.9% Yr1°Sfgnal 75psig)

C PlC 455A Przr Press Master Controller P&l I 2310 psig Variable 2345 psig (opens) Heater Przr Hi 2335 psig (closes) Pressure PORV 455 Backup Heaters ig Przr Control Lo Pressure

& Htrs on Pressurizer Controlling 4PP163O 67 Pressure Control PV-455B PV-455C

C, oo cn Cl)

C) CI) cCD .

1%)

*Baaai 1**.. a.. a.. a a a .a.aa a a ..aaaa ala a. .aa...a

$1

.a.aaa...

a a a a a I i..... a... a a a .a..aa a a a a.. .a aa a I

a.aai r

0 0 4 a.aaaal 0

a a a a a a cI 0

0 4

..a..* .ia a aaa a... a. a... aI.

a;i.

(0

+

00 di

HL-1 6 NRC WLEG ritten ExaminO/SR ation KEY O/HL-16 NRC/GCW

/2.8/3. I C/AILORQ BANKIR

27. 028AK1 .01 001/I/2IPZR LEVEL REF

, except Rod h all control systems in automatic The Unit is operating at 100% wit Control which is in manual.

following occurs:

rizer leve l con trol is sel ect ed to the 461/460 position when the Pressu channel.

ll Pre ssu riz er leve l ref ere nce leg leak occurs on the controlling

- A sma ponse of the ON E of the foll owi ng wou ld be CORRECT regarding the res Which

?

Pressurizer level control system increase.

than actual, charging flow would A. LT-460 level would read lower l wou ld rea d low er tha n act ual , charging flow would increase.

B. LT-461 leve se.

leve l wou ld rea d hig her tha n actual, charging flow would decrea C. LT-460 se.

leve l wou ld rea d hig her tha n actual, charging flow would decrea Dw LT-461 Page 55 of 208

KEY HL-16 NRC Written Examination Feedback ntrol Malfunction 028 Pressurizer (PZR) Level Co ts as they ow led ge of the ope rat ion al im plications of the following concep Kn Malfunctions:

apply to Pressurizer Level Control (CFR 41.8/41.10145.3) ormalities AK1 .01 PZR reference leak abn K/A MATCH ANALYSIS the 46 1/460 PZR level control selector switch in Question gives plant status with Candidate must LT- 461 wou ld be the con trol ing channel in this configuration. ak positio n.

the con trol ling cha nne l plu s the effects of a small reference leg bre be able to identify density of the the con trolling cha nne l. A refe rence leg break would decrease the on on rela tive to the imp ulse leg resu lting in an artificially high level sensed reference leg in response to con trolling cha nne l. PZR leve l control would reduce charging flow the the artificially sensed high level.

YSIS ANSWER I DISTRACTOR ANAL s is ct. Lev el wou ld read hig her than actual if the controlling channel. Thi A. Incorre not the controlling channel.

se.

actual. Charging flow would decrea B. Incorrect. Level reads higher than el wou ld read hig her than actu al if the controlling channel. This is C. Incorrect. Lev not the controlling channel.

rease.

actual and charging flow would dec D. Correct. Level reads higher than REFERENCES Control V-LO-PP- 16302 Pressurizer Level VEGP learning objectives:

of the common sequence of a leak on or isolation LO-PP-1 6302-05 Describe the con reference leg.

Page 56 of 208

Leak on the level transmitter reference leg Missile Shield I

If I

75%

\50%

25% Variable Leg 0%

I 101 lOOP 4 V-LO-PP-1 Hot Leg LO-PP- 16302-02

Leak on the level tr a n s m it te r r e f e r e n c e leg

A leak on the reference leg will reduce the weight of the column of water (assuming the condensing pot can.

maintain the level).

AP4 icaI .,d le v el w il l re a d h ig h er In,ø

$hi,/aczk,dl Pressurizer level

102 V-LO-PP- 16302 Rev-02V-LO-PP- 16302-02

Leak on the variable leg of the level transmitter I

100%

75%

0%

25 Variable jr 4103 LOOP V-LO-PP-163 Hot Leg LO-PP- 16302-02

Leak on the variable leg of the level transmitter

This will act to increase the zIP between the reference and variable leg of the level transmitter.

APf Iz.:.clicated level will read lower thanacrual Pressurizer level.

104 V-LO-PP- 16302 Rev-02V-LO-PP- 16302-02

16 N R C W ri tt en E xa m in at io n KEY HL- MEM/B ANK/RO/SRO/NRC/GCW

28. 029EK3.02 001 / 1/1 /ATWT/3. 1/3.1 Given the following:

- ATWT in progress.

ct.

Power Generation! ATWT in effe

- 19221-C, Response to Nuclear

- SI has NOT actuated at this time.

- The NCP is running.

ng pump is running.

Step 4.b states verify a chargi p?

foll owi ng is CO RR EC T reg arding the performance of this ste Which ONE of the flow.

can be left run nin g, pur pos e is to provide emergency boration A The NCP n flow.

req uir ed to be sta rte d, pur pos e is to provide emergency boratio B. A CCP is control.

P can be left runnin g, pur pos e is to provide for RCS inventory C. The NC ry control.

P is req uir ed to be sta rted, pur pose is to provide for RCS invento D. A CC Page 57 of 208

KEY HL-16 NRC Written Examination Feedback t Scram (ATWS) 029 Anticipated Transient Withou the led ge of the rea son s for the foll owing responses as the apply to Know ATWS:

(CFR 41.5/41.10/456/45.13) ng pump EK3.02 Starting a specific chargi K/A MATCH ANALYSIS of 19211-C, nario while performing the actions Question presents a plausible sce present Pow er Gen erat ion I AT WT . The candidate has to determine if the Nuclear ty related quate or does he need to start a safe CVCS charging pump running is ade reason.

Centrifugal Charging Pump and the YSIS ANSWER I DISTRACTOR ANAL y boration is the Cor rect. The NC P can pro vid e emergency boration flow. Emergenc A.

g.

action the procedure step is directin ncy A CC P is not req uire d to be star ted, the NCP can provide the emerge B. Incorrect.

boration flow.

rgency running. The reason is to provide eme C. Incorrect. The NCP can be left boration flow, not inventory control e emergency orre ct. A CC P is not req uire d to be started. The reason is to provid D. Inc boration flow, not inventory control REFERENCES I ATWT, step # 4.b 19211-C, Nuclear Power Generation HL-14 Audit VEGP learning objectives:

lear Power of EOP 19211-C, Response to Nuc LO-LP-3737041 -06 State the intent Generation ATWT.

Page 58 of 208

Approved By Procedure Number Rev J. D. Williams Vogtle Electric Generating Plant 1921 1 -c 20.3 Page Number Date Approved FR-S.1 RESPONSE TO NUCLEAR POWER 1-23 2007 4of 20 GENERATION/ATWT ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED SUBSEQUENT OPERATOR ACTIONS

3. Check AFW Pumps RUNNING:-

MDAFW Pumps _Start Pumps.

TDAFW Pump, if required _Open Steam Supply valve HV-51 06.

4. Emergency borate the RCS:

_a. Start at least one Boric Acid Transfer Pump.

_b. Verify a Charging Pump is running.

_c. Open EMERGENCY BORATE c. IF HV-8104 will NOT open, valve HV-81 04. THEN open the following:

. FV-11OA, BATO BLENDER.

. FV-11OB, BLENDER OUTLET TO CHARGING PUMPS SUCT.

_d. Verify charging flow GREATER THAN 42 GPM.

e. Verify boric acid flow GREATER

- e. Initiate 13009, CVCS THAN 30 GPM. REACTOR MAKEUP CONTROL SYSTEM as necessary to establish an alternate emergency boration flowpath.

Step 4 continued on next page Printed January 17, 2011 at 17:54

HL-16 NRC Writ ten ExaminatioGCW n KEY AD MONIT/3.3/3 .6 CIAJNEWIROISROJNRCI 29.029KLO2001/2/2/CNMT PURGE-R gress at full power.

A containment pressure relief is in pro

- The following alarm illuminates:

TION ALARM ALBO5 window B03 for INTMD RADIA d:

- The following alarm remains extinguishe TION ALARM ALBO5 window 003 for HIGH RADIA An investigation has revealed:

2, CN MT Low Ran ge Are a mo nito r has failed to the Intermediate level.

- RE-00 nitor indicates normal levels.

- RE-003, CNMT Low Range Area mo r indicates normal levels.

- RE-2562A1C CNMT Air Process monito nitor indicates normal levels.

- RE-2565A/B/C, CNMT Vent Rad mo re relief?

ECT regarding the containment pressu Which ONE of the following is CORR d in service and can remain in service.

A Pressure relief path has remaine d in service but is required to be isolated B. Pressure relief path has remaine CIA signal.

ically isolated due to the receipt of a C. Pressure relief path has automat CVI signal.

ically isolated due to the receipt of a D. Pressure relief path has automat Feedback S) 029 Containment Purge System (CP s

tions and/or cause effect relationship Knowledge of the physical connec tem and the following systems:

between the Containment Purge Sys (CFR: 41.2 to 41.9/45.7 to 45.8) r Ki .02 Containment radiation monito K/A MATCH ANALYSIS ef is in io where a containment pressure reli Question presents a plausible scenar DIATE.

ss whe n a Con tain men t Are a Low Range Rad monitor fails to INTERME progre relief there is an auto effect on the pressure The candidate must determine whether ain in service.

and if pressure relief can or cannot rem Page 59 of 208

n KEY HL-16 NRC Written Examinatio YSIS ANSWER I DISTRACTOR ANAL ain in auto acti on on INT ER ME DIA TE radiation. The release may rem A. Correct. No 3 / RE-2565) monitors are still available (RE-00 progress since two Tech Spec Rad candidates may INTERMEDIATE radiation. Plausible B. Incorrect. No auto action on monitor.

k Tec h Sp ecs req uire s term inat ion of the release on failed rad thin reach HIGH, ct. Flo w path wou ld NO T auto isolate as radiation level did not C. Inc orre I, not CIA.

if it did reach high it would be CV h HIGH, Flo w path wou ld NO T auto isol ate as radiation level did not reac D. Incorrect.

I, not CIA.

if it did reach high it would be CV REFERENCES Purge System Procedure 131 25-1 Containment M2 ced ure 171 02- 1 AR P for the Saf ety Related Display Console QR Pro Ventilation Isolation Instrumentation TECH SPEC 3.3.6 Containment VEGP learning objectives:

t Enviroment

-29 101 -01 Sta te the pur pos e of each of the following Containmen LO-PP Control Systems:

a. CONTAINMENT PURGE

b. CONTAINMENT MINI-PURGE

c. PRE-ACCESS FILTER

d. CRDM COOLING FANS NS

e. REACTOR CAVITY COOLING FA OLING

f. AUXILIARY CONTAINMENT CO LA TO RS

g. LOWER LEVEL AIR CIRCU

h. CONTAINMENT COOLERS signal 06 Sta te whi ch sys tem s hav e auto actions from a high radiation LO-PP-291 01-

-12 Des crib e the imp orta nce of RE-12442C and RE-i 2444C on LO-PP-29101 Containment Mini-Purge.

Page 60 of 208

. Procedure Number Rev Approved By S. E. Prewitt Vogtle Electric Generating Plant 13125-1 52 Page Number Date Approved

/20/2O10 CONTAINMENT PURGE SYSTEM 22 of 36 INITIALS 4.4 NON PERIODIC OPERATION NOTE When monitoring and changing containment pressure during this procedure, computer point P-9871 OR 1-PI-1 0945 (QHVC) should be used. These are the only containment pressure instruments that will indicate a negative pressure.

4.4.1 ContaInment Pressure Relief 4.4.1.1 IF the Unit is in MODE 1,2,3 or4:

a. Review Limitations 2.2.5c, 2.2.7, 2.2.8, and 2.2.10.

b. Place additional containment cooling units in service if required, to correct the high pressure condition.

4.4.1.2 NotIfy Chemistry of the upcoming Mini-Purge operation QE Pressure Relief operation:

a. Obtain the current approved Containment Gaseous Release Permit.

OR

b. jf an updated permit is unavailable, request that Chemistry sample the containment atmosphere and prepare for the gaseous release.

Printed January 17, 2011 at 18:01

Approved By . . Procedure Number Fev S. E. Prewitt Vogtle Electric Generating Plant 13125-1 52 Date Approved Page Number

/2O/2O1O CONTAINMENT PURGE SYSTEM 23 of 36 INITIALS 4.4.1.3 WHEN a current approved Containment Gaseous Release Permit is obtained, perform the following:

a. Verity at least TWO of the following radiation monitors are operable for CVI purposes (TS 3.3.6):

I -RE-2565A&B QE 1 -RE-2565C

1 -RE-002

1-RE-003

b. Verify at least ONE of the following radiation monitors is operable for ODCM purposes:

1-RE-12442C

1-RE-12444C CAUTION The pressure relief should NOT be initiated until the current approved Containment Gaseous Release Permit is obtained.

4.4.1.4 Releases may not continue beyond the date /time on the Release may not continue beyond (Date/Time) block indicated on 36022-C Data Sheet 1.

Printed January 17, 2011 at 18:03

RELATED DISPLAY CONSOLE QRM2 ANNUNCIATOR AND DETECTOR LOCATIONS

), Radiation indicators and Radiation Detectors, Data Processing Modules (DPM procedures.

Alarms are located as indicated on the individual sub-They will also be Any RMS alarm will annunciate on ALBO5 on the MCB.

CRT of Panel QRM1.

indicated on the plant computer and the monochrome TYPES OF ALARMS on the ERF Color CRT by Each channel may display several conditions indicated the channel identifier displayed in colors as:

Top of Scale Red High Alarm Red Alert Alarm Yellow Equipment Trouble Magenta Test Mode Magenta Out of Service Magenta Normal Green may be damaged, and The Top of Scale alarm is a latching alarm. The detector should be recalibrated by Chemistry prior to continued use.

A High Alarm indication will remain active until reset.

nated hnnell Instructions In this procedure are for a High Alarm on the desig on any channó s

The following generic actions should be taken for other alarm Alert Alerffl

1. NotIfy the Shift Superviso Printed January 17, 2011 at 18:04

entation Containment Ventilation Isolation Instrum 3.3.6 3.3 INSTRUMENTATION Instrumentation 3.3.6 Containment Ventilation Isolation rumentation for each Function in LCD 3.3.6 The Containment Ventilation Isolation inst Table 3.3.6-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.6-1.

ACTIONS

NOTE.

h Function.

Separate Condition entry is allowed for eac REQUIRED ACTION COMPLETION TIME CONDITION 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months A. Only one radiation A. 1 Restore at least two channels monitoring channel to OPERABLE status.

OPERABLE.

(continued) 3.3.6-1 Amendment No. 105 (Unit 1)

Vogtle Units 1 and 2 Amendment No. 83 (Unit 2)

Containment Ventilation Isolation Instrumentation 3.3.6 Table 33.6-1 (page 1 of 1)

Containment Ventilation Isolation Instrumentation APPLICABLE MODES OR OTHER REQUIRED SURVEILLANCE FUNCTION SPECIFIED CHANNELS REQUIREMENTS TRIP SETPOINT CONDITIONS Manual Initiation 12,3,4 2 SR 3.3.6.6 NA

2. Automatic Actuation Logic SR 3.3.6.2 NA and Actuation Relays 12,3,4 2 SR 3.3.6.3 SR 3.3.6.5

3. Containment Radiation SR 3.3.6.1 SR 3.3.6.4 1,2,3,4,6 SR 3.3.6.7 SR 3.3.6.8 (b)

a. Gaseous (RE-2565C)

(b)

b. Particulate (RE-2565A)

(b)

c. Iodine (RE-2565B) l5mr/h(C)

d. Area Low Range 50x background (d)

(RE-0002, RE-0003) 1,2,3,4 Refer to LCO 3.3.2, ESFAS lnstrumentation,Function 1, for all

4. Safety Injection(d) initiation functions and requirements.

(a) Containment ventilation radiation (RE-2565) is treated as one channel and is considered OPERABLE if the particulate (RE-2565A) and iodine monitors (RE-2565B) are OPERABLE or the noble gas monitor (RE-2565C) is OPERABLE.

(b) Setpoints will not exceed the limits of Specifications 5.5.4.h and 5.5.4.i of the Radioactive Effluent Controls Program.

(c) During CORE ALTERATIONS and movement of irradiated fuel assemblies within containment.

(d) During MODES 1,2, 3, and 4.

Vogtle Units 1 and 2 3.3.6-5 Amendment No. 105 (Unit 1)

Amendment No. 83 (Unit 2)

HL-16 NRC Writ ten Examination KEY C/A/WIT BANK/RO/SRO/NRC/GCW 6

30. 033AK3 .01 001 / 1/2/LOSS OFIRJ3 .2/3.

plant to tor star tup is in pro gre ss on Uni t 1, the operators have stab ihzed the A reac nts occur:

take critical data when the following eve ates.

S 1BD1I2 1BD1I12 TROUBLE illumin

- Annunciator ALB34, C02 INVERTER ates.

iato r AL B34 , CO l 120 V AC PA NE LS 1BY1B 1BY2B TROUBLE illumin

- Annunc ate.

- All channel Il trip status lights illumin power level for taking critical data.

- Reactor power remains stable at the rect action(s) for the crew to take?

Which ONE of the following is the cor rod s in MAN UAL , con trol SG NR levels in MANUAL from 60 70% NR A. Pla ce 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

Place rod s in MAN UAL , red uce reactor power to < P-6 within the next B.

ety Injection.

Ma nua lly trip the reac tor and go to 19000-C, E-O Reactor Trip or Saf C

power stable.

positive reactivity changes, maintain D. Suspend all operations involving Feedback Nuclear Instrumentation 033 Loss of Intermediate Range Loss dge of the rea son s for the foll owi ng responses as they apply to the Knowle rumentation:

of Intermediate Range Nuclear Inst (CFR 41.5,41.10/ 45.6 I 45.13) owing loss of intermediate range AK3.01 Termination of startup foll instrumentation K/A MATCH ANALYSIS to a giv es a plau sibl e sce nar io whe re an Intermediate Range NIS is lost due Question choose the does not trip. The candidate must 120V Vital bus failure and the reactor ices.

correct answer from 4 plausible cho IS ANSWER I DISTRACTOR ANALYS an auto Inc orre ct. Firs t two acti ons from AO P-18032 if candidate does not realize A.

e occurred.

reactor trip on IR NIS hi flux should hav does not Tech Spec action to take if candidate B. Incorrect. First action of AOP and hi flux should have occurred.

realize an auto reactor trip on IR NIS e occurred on lR HI flux.

C. Correct. A reactor trip should hav Page 61 of 208

HL-16 NRC Written Examination KEY r,

ct. Thi s is the pro per IOA for IR NIS malfunction from AOP-18002. Howeve D. Incorre d.

an auto reactor trip should have occurre REFERENCES 1AY1A 18032-1 Loss of Vital Instrument Panel 18002-C NIS Malfunction Tech Spec 3.3.1 RTS Instrumentation VEGP learning objectives:

onse, if any, to an intermediate range LO-LP-60302-08 Describe the reactor resp s.

ditions. Include effects on affected rod stop instrument failure under the following con NOTE: If no response, so indicate.

a. power level below P-6 failed high

b. power level below P-6 failed low failed high power level above P-6 but less than P-b c.

0 failed low power level above P-6 but less than P-i d.

-60 302 -09 Des crib e how and why an intermediate range instrument failure will LO-LP :

occurs) during the following conditions affect reactor operation (assuming no trip

a. reactor startup with power below P-6 and 5%

b. reactor startup with power between P-6

c. reactor power above 5%+

e rator action required for an intermediat LO-LP-60302-i0 State the immediate ope range channel N35 or N36 malfunction.

Page 62 of 208

Procedure Number Rev Approved By C. S. Waidrup I

Vogtle Electric Generating Plant 18032-1 28.2 Page Number R

Date Approved LOSS OF 120V AC INSTRUMENT POWE 1AY1A (CB-B52)

A. LOSS OF VITAL INSTRUMENT PANEL RESPONSE NOT OBTAINED ACTION/EXPECTED RESPONSE IMMEDIATE OPERATOR ACTIONS Al. Perform the following:

_A1. Check reactor power GREATER-THAN P-1O 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 January 17, 2011 at 18:09

Procedure Number Rev Approved By J. B. Stanley Vogtle Electric Generating Plant 1 8002-C 19 PageNumber M

)ateApproved NUCLEAR INSTRUMENTATION SYSTE 3 of 10 h3/22J09 MALFUNCTION CHANNEL MALFUNCTION A. SOURCE / INTERMEDIATE RANGE RESPONSE NOT OBTAINED ACTION/EXPECTED RESPONSE Al. - I operations involving

.a reactivity changes çrS 33 1 J.9.3, and TR 13.9.6)

_A2. Stabilize count rate.

_A3. Go to Step A7.

A3. Check MODE MODE 6. -

A4. Suspend core alterations. (TS 3.9.3 and TR 13.9.6)

_A5. GotoStepA7.

_A5. Check Source Range Channels -

TWO INOPERABLE.

A6. Initiate action to restore at least one Source Range Channel to OPERABLE status. (TS 3.9.3)

_A7. GotoStepAl5.

_A7. Check power LESS THAN P-6.

A8. GotoStepAl2.

_A8. Check Source Range Channels -

TWO INOPERABLE.

A9. Open Reactor Trip Breakers.

(TS 3.3.1)

A10. Verify adequate shutdown margin by initiating 14915, SPECIAL CONDITIONS SURVEILLANCE LOGS. (TS 3.3.8)

_A1 1. Verify unborated water sources isolated by initiating 14915, SPECIAL CONDITIONS SURVEILLANCE LOGS. (TS 3.3.8)

Printee January 17, 2011 at 18:11

RTS Instrumentation 3.3.1 ACTIONS (continued)

REQUIRED ACTION COMPLETION TIME CONDITION E. One channel inoperable. ------------------NOTE A channel may be bypassed for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for surveillance testing.

Place channel in trip. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months El OR Be in MODE 3. 78 hours9.027778e-4 days 0.0217 hours 1.289683e-4 weeks 2.9679e-5 months E.2 Reduce THERMAL 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months F. THERMAL POWER F.1 POWER to < P-6.

> P-6 and < P-b, one Intermediate Range Neutron Flux channel OR inoperable. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months F.2 Increase THERMAL POWER to> P.10.

Suspend operations Immediately G. THERMAL POWER G.1 involving positive reactivity

> P-6 and < P-b, two additions.

Intermediate Range Neutron Flux channels inoperable. AND Reduce THERMAL 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months G.2 POWER to < P-6.

Restore channel(s) to Prior to increasing THERMAL POWER H.1 H. OPERABLE status. THERMAL POWER

< P-6, one or two to> P-6 Intermediate Range Neutron Flux channels inoperable.

(continued) 3.3.1-3 Amendment No. 116 (Unit 1)

Vogtle Units 1 and 2 Amendment No. 94 (Unit 2)

RTS Instrumentation 3.3.1 I

Table 3.3.1-1 (page 1 of 9) rumentation Reactor Trip System Inst NOMINAL APPLICABLE TRIP MODES OR OTHER SURVEILLANCE ALLOWABLE SET IN I PO REQUIRED VALUE SPECIFIED CONDITIONS REQUIREMENTS CONDITIONS CHANNELS FUNCTION NA NA B SR 3.3.1.13 1,2 2 NA Manual Reactor SR 3.3.1.13 NA Trip 2 C 5

3 4

( a)

2. Power Range Neutron Flux 109%RTP SR 3.3.1.1 111.3%RTP 4 D 1,2 SR 3.3.1.2

a. High SR 3.3.1.7 SR 3.3.1.11 SR 3.3.1.15 25% RTP SR 3.3.1.1 27.3% RTP E

(1b) 2 SR 3.3.1.8

b. Low SR 3.3.1.11 SR 3.3.1.15 6.3% RTP 5% RTP E SR 3.3.1.7 with time 1,2 4 SR 3.3.1.11 with time constant

3. Power Range constant Neutron Flux High 2sec 2sec PositIve Rate 25% RTP SR 3.3.1.1 41.9% RTP 2 F,G (1b) (2 c) SR 3.3.1.8

4. IntermedIate , SR3.3.1.11 Range Neutron 25% RTP Flux SR 3.3.1.1 41.9% RTP 2 H SR 3.3.1.8 2 (d) SR 3.3.1.11 (continued) withdrawal.

trol System capable of rod Trip Bre ake rs (RT Bs) closed and Rod Con (a) With Reactor ks.

ge Neutron Flux) interloc (b) Below the P-iC (Power Ran interlocks.

te Range Neutron Flux)

(c) Above the P-6 (Intermedia interlocks.

te Range Neutron Flux)

(d) Below the P6 (intermedia d provided the Trip Setpol nt Set pol nt valu e out side Its calibration tolerance ban to within the established LE with an actual Trip channel is readjusted A channel Is OPERAB Allowable Value and the than the Nominal Trip (n) e is con serv ativ e wit h respect to its associated A Trip Set pol nt may be set more conservative valu Nom inal Trip Set poi nt.

d of the calibration tolerance ban ditions.

in response to plant con Setpoint as necessary 1) 3.3.1-14 Amendment No. 128 (Unit

. 106 (U nit 2)

Vogtle Units I and 2 Amendment No

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 2 of 9)

I Reactor Trip System Instrumentation APPLICABLE MODES OR NOMINAL TRIP OTHER SURVEILLANCE ALLOWABLE SPECIFIED REQUIRED SETPOINT()

REQUIREMENTS VALUE CONDITIONS CHANNELS CONDITIONS FUNCTION 1.7 E5 1.0 E5 2 I,J SR 3.3.1.1 cps Source Range d)

(2 SR 3.3.1.8 cps 5.

Neutron Flux SR 3.3.1.11 1.7E5 1.0 ES 2 J,K SR 3.3.1.1 cps 5

3

(

4

( a) a) , SR 3.3.1.7 cpa SR 3.3.1.11 I L SR 3.3.1.1 NA

, SR 3.3.1.11 NA 5

3

(

4

( e) e)

Refer to Note I Refer to Note 1 4 E SR 3,3.1.1 1,2 (Page 3.3.1-20) (Page 3.3.1-20)

6. OvertemperatureT SR 3.3.1.3 SR 3.3.1.6 SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.15 Refer to Note 2 Refer to Note 2 4 E SR 3.3.1.1 1,2 (Page 3.3.1-21) (Page 3.3.1-21)

7. Overpower T SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.15 (continued) em capable of rod withdrawal.

(a) With RTBs closed and Rod Control Syst n Flux) interlocks.

(d) Below the P-S (IntermedIate Range Neutro Flux or trip but does provide input to the High source range FunctIon does not provide react (a) With the RTBs open. In this condItIon, and Indication.

at Shutdown Alarm System (LCO 3.3.8) tolerance band provided the Trip Setpolnt chan nel Is OPER ABLE with an actua l Trip Setpoint value outside Its calibration readjusted to within the established (n) A ciated Allowable Value and the channel Is value is conservative with respect to Its asso Setpoint. A Trip Setpoint may be set more conservatIve than the Nominal Trip al Trip calibration tolerance band of the NomIn conditions.

Setpoint as necessary in response to plant 3.3.1-15 Amendment No. 128 (Unit 1)

Vogtle Units 1 and 2 Amendment No. 106 (Unit 2)

HL-16 NRC W14.4 ritten Examinat ion KEY MEMJNEW/RO/SROJNRC/GCW

31. 033G2.4.49 001 1212/SFPC ACTIONS/4.6 ent of irradiated fuel assemblies occuring A refueling outage is in progress with movem in the Spent Fuel Pools.

ld ALLOW movement of irradiated Which ONE of the following conditions wou tinue?

assemblies in the Spent Fuel Pool to con is 1969 ppm.

A. Spent Fuel Pool boron concentration inches on local reading.

B. Spent Fuel Pool level is 216 feet, 4 ls have been lost in the control room.

C Both SR NIS audio count rate channe E.

systems have been declared INOPERABL D. Both FHB Post Accident Ventilation Feedback PCS) 033 Spent Fuel Pool Cooling System (SF Emergency Procedures I Plan ce to procedures those actions that 2.4.49 Ability to perform without referen components and controls.

require immediate operation of system (CFR: 41.10/43.2/45.6)

K/A MATCH ANALYSIS stio n asks can didate whic h cond ition would all continued movement of irradiated Que ire immediate suspension of actitivities per assemblies in the SFP area. 3 choices requ actions.

Tech Specs and one would allow continued ANSWER I DISTRACTOR ANALYSIS ired due to SEP Gb <2000 ppm which is A. Incorrect. Suspension of movement requ below Tech Spec requirements.

ired due to SEP level below the Tech B. Incorrect. Suspension of movement requ Spec requirements.

for movement in SEP. They only apply to C. Correct. Audio count rates not required CORE ALTERATIONS in containment.

ired due to neither FHB Post Accident D. Incorrect. Suspension of movement requ Ventilation System available.

REFERENCES ter Level Tech Spec 3.7.15, Fuel Storage Pool Wa Page 63 of 208

HL-16 NRC Written Examination KEY Concentration Tech Spec 3.7.17, Fuel Storage Pool Boron dent Ventilation System (common TRM 13.9.5, Fuel Handling Building Post Acci system)

Indication TRM 13.9.6, Source Range Monitor Audible Logs Procedure 1 4225-C Ops Weekly Surveillance Leve l Procedure 17005-1 E02 Spent Fuel Pit Lo VEGP learning objectives:

3.9 of Tech Specs, be able to:

LO-LP-39213-01 For any given item in section

a. State the LCO.

ns.

b. State any one hour or less required actio 13.9 of the TRM, be able to:

LO-LP-39213-03 For any given item in section

a. State the TR for operation.

ns.

b. State any one hour or less required actio Page 64 of 208

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 I and 2 3.7.15-1 Amendment No. 96 (Unit 1)

Amendment No. 74 (Unit 2)

Fuel Storage Pool Boron Concentration 3.7.17 3.7 PLANT SYSTEMS 3.7.17 Fuel Storage Pool Boron Concentration The fuel storage pool boron concentration shall be 2000 ppm.

LCO 3.7.17 APPLICABILITY: When fuel assemblies are stored in the fuel storage pool.

ACTIONS REQUIRED ACTION COMPLETION TIME CONDITION A. Fuel storage pool boron NOTE concentration not within LCO 3.0.3 is not applicable.

limit.

A.1 Suspend movement of Immediately fuel assemblies in the fuel storage pool.

AND A.2.1 Initiate action to restore Immediately fuel storage pool boron concentration to within limit.

SURVEILLANCE_REQUIREMENTS FREQUENCY SURVEILLANCE Verify the fuel storage pool boron concentration is 7 days SR 3.7.17.1 within limit.

Vogtle Units 1 and 2 3.7.17-1 Amendment No. 99 (Unit 1)

Amendment No. 77 (Unit 2)

Fuel Handling Building Post Accident Ventilation System TR 13.9.5 13.9 Ref ueling Operations TR 13.9.5 Fuel Handling Building Post Accident Ventilation System (common system)

TR 13.9.5 Two independent Fuel Handling Building Post Accident Ventilation Systems shall be OPERABLE.

APPLICABILITY: Whenever irradiated fuel is in either storage pool.

TR 13.0.3 is not applicable.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One Fuel Handling A.1 Place the remaining Fuel 7 days Building Post Accident Handling Building Post Ventilation System Accident Ventilation System inoperable, in operation discharging through at least one train of HEPA filters and charcoal adsorbers.

28 (continued)

Vogtle Units I and 2 13.9-6 Rev.3 Technical Requirement 9/18/03

Fuel Handling Building Post Accident Ventilation System TR 13.9.5 ACTIONS (continued)

REQUIRED ACTION COMPLETION CONDITION TiME A.2 Suspend all operations 7 days involving movement of irradiated fuel in the fuel handling building, movement of new fuel over irradiated fuel in the fuel handling building, or crane operation with loads over irradiated fuel in the fuel handling building until Required Action A.1 above is met.

Two Fuel Handling B.1 Suspend all operations Immediately B.

Building Post Accident involving movement of Ventilation Systems irradiated fuel in the fuel inoperable, handling building, movement of new fuel over irradiated fuel in the fuel handling building, or crane operation with loads over irradiated fuel in the fuel handling building until Required Action A.1 above is met.

TECHNICAL REQUIREMENT SURVEILLANCES SURVEILLANCE FREQUENCY Verify system operation by initiating each system from 15 days on a TRS 13.9.5.1 the control room with flow through the HEPA filters and STAGGERED TEST charcoal adsorbers and operating each system for 10 BASIS continuous hours with the heater circuit energized.

(continued) 13.9 7 Rev. 3 Vogtle Units 1 and 2 -

9/18/03 Technical Requirement

Source Range Monitor Audible Indication TR 13.9.6 13.9 Refueling Operations TR 13.9.6 Source Range Monitor Audible Indication TR 13.9.6 At least one source range monitor shall provide audible indication in the containment and control room.

APPLICABILITY: M ODE 6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Required source range A.1 ----------- N OTE---

monitor audible indication Makeup to the reactor inoperable or not coolant system (RCS) is operating. allowed, provided the makeup source has been verified to be greater than the required refueling boron concentration (reference Technical Specifications Paragraph 3.9.1)

Suspend all operations Immediately involving CORE ALTERATIONS or positive reactivity changes.

TECHNICAL REQUIREMENT SURVEILLANCES SURVEILLANCE FREQUENCY TRS 13.9.6.1 Perform CHANNEL CHECK 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months TRS 13.9.6.2 NOTE Neutron detectors are excluded from CHANNEL CALIBRATION.

Perform CHANNEL CALIBRATION 18 months Vogtle Units I and 2 13.9 10 Rev. 2 Technical Requirement 4/21/04

Procedure Number Rev Approved By S. E. Prewitt I Vogtle Electric Generating Plant A 14225-1 Page Number 35 Date Approved 7 of 14 VEILLANCE LOGS OPERATIONS WEEKLY SUR io I Sheet 2 of 3 MODE DATA SHEET 1 DATE MODE 1 & 2 WEEKLY SURVEILLANCES LIMIT(Si LCO INDICATION TOLERANCE LCO/PROC PARAMETER INSTRUMENT METHOD OF VERIFICATION TECH SPEC SURV REQ Boric Acid Flow Path Boric acid flowpath ))L Temp ( F) TR 13.1.3 operable Verify fiowpath temp Cl-i TRS 13.1.3.1 Panel 0018 Aux BId RCI21 Cl-S CI -9 C3-i C3-S 65F C3-8 C3-1 I C4-7 C4-8 C4-10 C4-11 C4-12 At standby HEAT TRACE PANEL 1-1 81 7-U3-OO1B the CALIREAD Switch to LOCATE the card specified in the 1nstrument Co umn, ROTATE the common DIGITAL READ position and RECORD TEMPERATURE indicated on DISPLAY. TR 13.1.7 Chemical Analysis >7000 ppm TRS 13.1.7.3 BAST Boron conc. (ppm)

Borated water sources <7700 ppm ooerable I Ll-0102A IRS 13.1.7.4 BAST level (%)

Verity BAST boron concentration 83%

and level 1LI-0104A and

>2400 ppm TR 13.1.7 RWST Boron conc, Chemical Analysis RWST boron concentration and IRS 13.1 .7.1 <2600 ppm SR 3.5.4.3 (ppm) level 3.5.4 lLl-099iA TRS 13.1.7.1 SR 3.5.4.2 RWST level (%) 1 Li-0993A

>94%

1 Ll.0990A I LI-0992A ALSO6EOØ ALARMASSEN 7S 2*of Waler &ialibi SR3.1i& Fu&?OLe4(tø a

h atoraept I Rdrnodeodrdlaledfu,nnme VfrfriirIQ. po tevd The Axial Fiux Difference shalt be SR 3.2.3.1 PTDB TAB 6.0 to maintained within the limits of the (50% Power)

Axial Flux Difference (%)

Q) 1NI-41C Verify AFD 3.2.3 1NI-42C 14915-1 1 NI-43C INI-44C AFD per 1491 5-1.

NOTE: With the AFD monitor alarm inoperable commence recording temperature for operability. A on the opposite panel can be used to determine the minimum If either panel 001 A or 0018 circuit indicator is tailed the same circuit for any failed circuit indications.

per 17104-1 real or info LCO and corrective actions should be taken Printed January 17, 2011 at 18:26

Procedure Number Rev Approved By C. S. Walcirup Vogtle Electric Generating Plant 17005-1 32.1 Page Number Date Approved ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 05 ON 43 of 66 4/25/10 PANEL 1A2 ON MCB WINDOW E02 ORIGIN SETPOINT SPENT FUEL PIT 1 -LSHL-625 217 feet elevation LO LEVEL 1.0 PROBABLE CAUSE

1. Insufficient inventory during filling or refueling operation.

2. Normal evaporation.

3. System leak.

4. Loss of air to the Fuel Transfer Canal and/or Cask Loading Pit Gate Seals.

2.0 AUTOMATIC ACTIONS NONE 3.0 INITIAL OPERATOR ACTIONS NONE Printed January 17, 2011 at 18:28

Approved By Procedure Number Rev C. S. Waidrup Vogtle Electric Generating Plant 17005-1 32.1 Page Number Date Approved ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 05 ON 4/25/10 PANEL 1A2 ON MCB 44 of 66 WINDOW E02 (Continued) 4.0 SUBSEQUENT OPERATOR ACTIONS

1. Dispatch an operator to determine actual level locally.

(see Figure 1 in this procedure).

2. Notify the Security Alarm Station (CAS) to dispatch a security patrol to check for any indications of sabotage.

3. Refer to 13719-1, Spent Fuel Pool Cooling And Purification and return the Spent Fuel Pit to normal level (218.5 feet).

4. IF level cannot be maintained greater than 217 feet with fuel movement in containment in progress or 216.5 feet with the Spent Fuel Pool Gate Valve closed, THEN suspend movement of irradiated fuel assemblies in the Spent Fuel Pool and all crane operations over the Spent Fuel Pool.

Initiate 18030-C, Loss Of Spent Fuel Pool Level Or Cooling and 18006-C Fuel Handling Event.

5. Check service air to gate seals and refer to 13710-1, Service Air System to restore service air if lost.

6. Refer to Technical Specification LCO 3.7.15.

5.0 COMPENSATORY OPERATOR ACTIONS NOTE If the East and West pools are connected through the cask loading pit, Unit 1 annunciator ALBO5EO2 will detect a low level condition for both pools.

Verify Spent Fuel Pool Level every 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months per 11883-1, Radwaste Rounds Sheets.

END OF SUB-PROCEDURE

(

REFERENCES:

1X4DB13O, PLS, 1X5DT0037, Technical Specifications LCO 3.7.15 Printed January 17, 2011 at 18:28

HL-16 NRC W/4.6riC/AttJLO en Examination KEY IT BANK!RO/SROINRC/GCW

32. 035A4.06 OO1/2/2/SG-ISOL SGTL/4.5 Given the foNowing:

ed on Steam Generator # 3.

- A Steam Line break has occurr Steam contained in 19020-C, Faulted

- The crew has performed actions Generator Isolation.

50 psig and lowering.

- Steam Generator # 3 pressure is level is off-scale low.

- Steam Generator # 3 Wide Range led at 25%.

Narrow Range levels are control

- Steam Generators # 1, 2, and 4 and pressures are approximately 900 psig

- Steam Generators # 1, 2, and 4 stable.

and the reason ON E of the foll owi ng des cri bes the action required 1 9020-C Which for this action?

Generator # 3.

A Isolate AFW flow to Steam and mass energy release.

Prevent excessive RCS cooldown Generator # 3.

B. Isolate AFW flow to Steam ck.

ven t fail ure of SG com pon ent s that may occur due to thermal sho Pre Generator # 3 at 30 GPM.

C. Control AFW flow to Steam release.

RCS cooldown and mass energy Prevent SG dryout and excessive Generator # 3 at 30 GPM.

D. Control AFW flow to Steam ck.

out and fail ure of com pon ent s that may occur due to thermal sho Prevent SG dry Page 65 of 208

KEY HL-16 NRC Written Examination Feedback S) 035 Steam Generator System (S/G monitor in the control room:

Ability to manually operate and/or (C FR: 41.7/ 45.5 to 45.8) or tube rupture/leak A4.06 S/G isolation on steam leak K/A MATCH ANALYSIS and all n pre sen ts a scenar io wit h a fau lted SG and parameters for the faulted Questio g AFW flow rmine the correct actions for throttlin intact SGs. The candidate must dete and the basis for the actions.

IS ANSWER I DISTRACTOR ANALYS mass energy Cor rect. Isolates AF W flow to fau lted SG to limit RCS cooldown and A.

release from the SG.

venting lted SG part is correct. Basis for pre B. Incorrect. Isolates AFW flow to fau thermal shock to faulted SG is wrong.

iting cooldown basis for all SGs faulted and the lim C. Incorrect. Preventing dryout is the and mass enery part is correct.

ting thermal Inc orre ct. Pre venting dry out is the basis for all SGs faulted and preven D.

from all SGs faulted.

shock on re-initiation is also a basis REFERENCES Isolation 1 9020-C, Faulted Steam Generator atio n of all SGs 19121-C Uncontrolled Depressuriz 1 .07) used as base for modification.

Harris 2008 NRC RO Exam (040AK ound Documents Vogtle E-2 and ECA-2.1-BG Backgr VEGP learning objectives:

EOP 19020-C, Faulted SG Isolation LO-LP-37121-01 State the intent of each step is C as a guide, briefly descried how LO-LP-37121-02 Using EOP 19020-accomplished.

Page 66 of 208

. . Procedure Number Rev I Approved By Vogtle Electric Generating Plant 19020-C 18 J. B. Stanley Page Number E-2 FAULTED STEAM GENERATOR ISOLATION jDateAroved 4of9 F /29/2010 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED Isolate Auxiliary Feedwater to the _6. j[ valves can NOT be closed, 6.

faulted SG(s): THEN dispatch operator to locally isolate faulted SG(s).

Close affected MDAFW Pump Throttle Valves:

HV-5139 SG 1 FROM MDAFW PMP-A HV-5132 SG 2 FROM MDAFW PMP-B HV-5134 SG 3 FROM MDAFW PMP-B HV-5137 SO 4 FROM MDAFW PMP-A

Close affected TDAFW Pump

) Throttle Valves:

HV-5122 SG 1 FROM

TDAFW HV-5125 SG 2 FROM TDAFW HV-5127 SG 3 FROM TDAFW HV-5120 SO 4 FROM TDAFW

STEP DESCRIPTION TABLE FOR E-2 Step 4 STEP: Isolate Faulted SG(sj PURPOSE To isolate all feedwater to and steam flow from the faulted SG( s BASIS:

wn Isolation of the feedwater to the faulted SG maximizes the cooldo capability of the nonfaulted loops following a feedline break and minimizes the RCS cooldown and mass and energy release following a also steamline break. Isolation of steam paths from the faulted SG ment.

minimizes the RCS cooldown and mass and energy release to contain In addition, isolation of these steam paths could isolate the break.

ACTIONS:

o Determine if SG PORV5 are closed o Determine if appropriate valves cannot be closed o Isolate main feedline o Isolate AFW flow o Close steam supply valves to turbine-driven AFW pump o Determine if SG blowdown isolation valves are closed o Isolate other secondary piping o Close appropriate valves block o Dispatch operator to locally close appropriate valves or valves INSTRUMENTATION:

o SG PORV5 position indication and o Position for plant specific valves associated with feedwater to steam flow from SGs E-2 Background 39 HP-Rev. 2, 4/30/2085 HE2BG.doc

Approved By Procedure Number Rev C. S. Walclrup Vogtle Electric Generating Plant 19121-C 26 PageNumber DateApproved ECA-2.1 UNCONTROLLED DEPRESSURIZATION

/16/1O OF ALL STEAM GENERATORS of 41 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

4 Control feed flow to minimize RCS cooldown:

_a. Monitor shutdown margin by initiating 14005, SHUTDOWN MARGIN AND KEFF CALCULATIONS.

_b. Check cooldown rate in RCS _b. Lower feedflow to 30 gpm to Cold Legs LESS THAN

- each SG.

100°F/HR.

Go to Step 4.d.

c. Check NR level in all SGs LESS - _c. Control feed flow to maintain NR THAN 65%. level in all SGs less than 65%.

_d. Check NR level in all SGs -

d. Maintain a minimum feed flow of GREATER THAN 10% [32% 30 gpm to each SG with less ADVERSE], than 10% [32% ADVERSE] NR level.

_e. Check RCS WR Hot Leg _e. Control feed flow or dump steam temperatures STABLE QE

- to stabilize RCS WR Hot Leg LOWERING. temperatures.

5 Check if RCPs should be stopped:

a. ECCS Pumps AT LEAST ONE

- _a. Go to Step 6.

RUNNiNG:

. CCP or Si Pump Step 5 continued on next page Printed January 17, 2011 at 18:44

STEP DESCRIPTION TABLE FOR ECA-2.l Step 2 CAUTION CAUTION: A minimum feed flow of (S.04) gpm must be maintained to each SG with a narrow range. level less than (M.02)% [(M.83)% for adverse containment].

PURPOSE To alert the operator to maintain a minimum feed flow to minimize any. subsequent thermal shock to SG components BASIS:

If feed flow to a SG is isolated and the SG is allowed to dry out, subsequent reinitiation of feed flow to the SG could create significant thermal stress conditions on SG components. Maintaining a minimum verifiable feed flow to the SG allows the components to remain in a wet condition, thereby minimizing anythermal shock effects if feed flow is increased.

ACTIONS:

Determine if SG narrow range level is less than (M.02)% [(M.03)% for adverse containment]

INSTRUMENTATION:

SG narrow range level indication CONTROL/EQUIPMENT:

N/A KNOWLEDGE:

N/A PLANT-SPECIFIC INFORMATION:

o (S.04) Feed flow value in plant specific units corresponding to 25 gpm.

o (M.02) SG level just in the narrow range) including allowances -For normal channel accuracy and reference leg process errors.

o (M.03) SG level just in the narrow range, including allowances for normal channel accuracy) post accident transmitter errors, and reference leg process errors, not to exceed 50%.

ECA-2.1 Background 24 HP-Rev. 2, 4/30/2005 HECA21BG.doc

HL-16 NRC Written Exam ination KEY

/RO/SRO/NRC/GCW

33. 038EK3.0 1 001/1 / I /SGTRJ4. 1/4.3 MEM!BANK HARRIS2007 Given the following.

Rupture is in effect.

- A SGTR has occurred and 19030-C, Steam Generator Tube completed.

- RCS cooldown to 518°F and RCS depressurization have been RCS pressure to match Which ONE of the following is the correct reason for reducing ruptured SG pressure at this point in the procedure?

to SI termination.

A To restore RCS inventory and reduce break flow prior due to SG overfill.

B. To prevent damage to the secondary side steam piping cooldown is recommenced.

C. To minimize the possibility of a PTS event when RCS ospheric Relief Valves.

D. To prevent any release of radioactivity through the SG Atm Page 67of 208

HL-16 NRC Written Examination KEY Feedback R) 038 Steam Generator Tube Rupture (SGT ng responses as the apply to the Knowledge of the reasons for the followi SGTR:

(CFR 41.5 /41.10/45.6/45.13) and secondary sides of ruptured S/G EK3.01 Equalizing pressure on primary K/A MATCH ANALYSIS depressurization after max rate cooldown in Question asked the purpose of the RCS correct basis.

19030-C, SGTR. Candidate must choose the ANSWER / DISTRACTOR ANALYSIS A. Correct.

R procedure but not the basis for the step B. Incorrect. This is a concern in the SGT R procedure but not the basis for the step C. Incorrect. This is a concern in the SGT R procedure but not the basis for the step.

D. Incorrect. This is a concern in the SGT REFERENCES 19030-C, Steam Generator Tube Rupture Harris 2002 and 2007 NRC RO Exams.

Watts Bar 2004 NRC RO Exam.

VEGP learning objectives:

of 19030-C, Steam Generator Tube LO-LP-3731 1-06 State the major objectives Rupture.

Page 68 of 208

Procedure Number Rev

[Approved Sy J. B. Stanley Vogtle Electric Generating Plant 19030-C 37.1 PageNumb RE E-3 STEAM GENERATOR TUBE RUPTU RESPONSE NOT OBTAINED ACTION/EXPECTED RESPONSE

_31. Depressurize RCS using Normal PRZR Spray to refill PRZR

_a Spray PRZR with maximum available spray b Close Normal PRZR Spray

_b Normal PRZR Spray -

Valves EFFECTIVE AT REDUCING RCS PRESSURE GotoStep32.

c. GotoStep37.

32. Go to Step 35.

_32. Check at least one PRZR PORV -

AVAI LABLE.

_33. IF neither PRZR PORV Block

33. Check at least one PRZR PORV Valve available, Block Valve AVAILABLE.

THEN Go to Step 35.

Printed January 17, 2011 at 18:51

generators, the RCS is cooled to After isolation of the ruptured steam d steam generator pressure by less than saturation at the rupture steam generators. This ensures dumping steam from only the intact r depressurization to the ruptured adequate subcooling in the RCS afte uent actions. With of-Fsite power steam generator pressure in subseq m to the condenser provides available, the normal steam dump syste cooldown rapidly, as demonstrated sufficient capacity to perform this rease during this cooldown as in Figure 9. RCS pressure will dec nds the steam bubble in the shrinkage of the reactor coolant expa le tube failures, RCS pressure pressurizer (Figure 10). For multip the ruptured steam generator (Figure 11) may decrease to less than generated in the RCS during the pressure as steam voids, which were Reverse flow, i.e.,

initial depressurization, condense.

this time would reduce the secondary-to-primary leakage, during erators and delay steam generator inventory in the ruptured steam gen overfill, as shown in Figure 12.

Since leakage from the primary tured steam generator pressures side will continue until RCS and rup entory is required before stopping equalize, an excess amount of inv entory required depends on the RCS SI flow. The excess amount of inv pressure equals the pressure in pressure and reduces to zero when RCS necessary to accommodate the the ruptured steam generator. It is SI flow is stopped. To establish decrease in pressurizer level after decreased by condensing steam in sufficient inventory, RCS pressure is This increases SI flow and the pressurizer using normal spray.

pressurizer, as demonstrated in reduces break flow which refills the h the cooldown of the primary Figures 13 and 14. Note that althoug pressurizer did not refill since side also decreased RCS pressure, the me. Similarly, spraying the the net effect reduced coolant volu concurrently with the primary pressurizer to decrease RCS pressure refilling the pressurizer, as side cooldown is not as effective in 5, FREQUENT QUESTIONS, for shown in Figure 15 (refer to Section 22 HP-Rev. 2, 4/30/2005 E-3 Background HE3BG.doc

1. 038 EK3.01 O23IBANKJIP3 2002/LOWERJfS/7/20070 Initial Conditions:

A SGTR has occurred and Path-2 actions are in progress.

The RCS cooldown to target temperature has just been completed.

Which ONE (1) of the following describes the reason(s) for reducing RCS pressure to match ruptured SG pressure at this point?

A To restore RCS inventory and reduce break flow prior to SI termination.

B. To prevent damage to secondary side steam piping due to overfill.

C. To minimize the probability of a Pressurized Thermal Shock event when RCS cooldown is recommenced.

D. To prevent any release of radioactivity through the SG Atmospheric Dump valves.

A. Correct.

B. Incorrect. May not be able to stop overfill if release is large enough.

C. Incorrect. PTS is only a concern if ruptured SG pressure is low. (Also faulted)

D. Incorrect. May not be able to stop release if SGTR is large enough.

pressure on Knowledge of the reasons for the following responses as they apply to the Steam Generator Tube Rupture: Equalizing primary and secondary sides of SG Question Number: 44 Tier 1 Group 1 Importance Rating: RO 4.1 Technical

Reference:

Path 2 BD Proposed references to be provided to applicants during examination: None Learning Objective:

10 CFR Part 55 Content: 41.10 Comments:

Page: 1 12/9/2010

HL-16 NRC Written ExNEW/ amination KEY RO/SROINRC/GCW

34. 039G2.2.39 OO1/2/IIMAIN/REHEAT STM -TSI3.9J4.5 MEM/

Given the following:

- The unit is at full power.

t reactor trip) occurs.

- A secondary transient (loss of external load without a direc

- An RCS Safety Limit (SL) is exceeded.

one of the following are, In accordance with Technical Specifications and Bases, which eding the (SL), and

1) Main Steam components that should have prevented exce pliance with the limit.

2) the time limit Tech Specs requires for restoration of com Main Steam Component Restore Compliance Within A. Atmospherice Relief Valves one hour 5 minutes B. Atmospherice Relief Valves one hour C Main Steam Safety Valves 5 minutes D. Main Steam Safety Valves Feedback 039 Main and Reheat Steam System (MRSS)

Equipment Control Technical Specification G2.2.39 Knowledge of less than or equal to one hour action statements for systems.

(CFR: 41.7/41.10/43.2/45.13)

K/A MATCH ANALYSIS Main Steam Safety Valves The questions presents the design bases event in which idate must choose the Main should prevent exceeding the RCS Safety Limits. The cand (MSSV5) and the time Tech Steam components designed to prevent exceeding the SL Specs requires restoration of the SL.

ANSWER I DISTRACTOR ANALYSIS ling this event is specifically A. Incorrect. MS Atmospheric Relief Valve credit for hand exceeding either the core excluded in the Bases for Tech Specs. 2nd half is correct, of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

or pressure safety limit in Mode 1 requires restoration time Page 69 of 208

HL-16 NRC Written Examination KEY event is specifically B. Incorrect. MS Atmospheric Relief Valve credit for handling this eding either the excluded in the Bases for Tech Specs. 2nd half is incorrect, exce n of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . In Modes core or pressure safety limit in Mode 1 requires restoratio time 3, 4, or 5 this limit would be 5 minutes so it is plausible.

s to handle this event C. Correct.Main Steam Safety Valves are specified in the Base and the restoration time limit is 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months if in Mode 1 or 2.

s to handle this event D. Incorrect. Main Steam Safety Valves are specified in the Base part is wrong but and the restoration time limit is 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months if in Mode 1 or 2. 2nd plausible as 5 minutes is the time limit if in Modes 3, 4, or 5.

REFERENCES Tech Spec 2.0, Safety Limits (SL) and Bases VEGP learning objectives:

ty Limit is exceeded.

LO-LP-39203-01, State the required action in the event a Safe LO-LP-39203-03, State the Safety Limit for RCS pressure.

ty Limit Curve and the LO-LP-39203-04, Describe the bases for the Reactor Core Safe RCS Pressure Safety Limit.

ty Limits are designed to LO-LP-39203-07, Describe Safety Limit and state what Safe protect.

Page 70 of 208

S Ls 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs R, Reactor Coolant In MODES I and 2, the combination of THERMAL POWE urizer pressure shall System (RCS) highest loop average temperature, and press not exceed the SLs specified in Figure 2.1.1-1.

2.1.2 RCS Pressure SL tained 2735 psig.

In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be main 2.2 SL Violations within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 2.2.2 If SL 2.1.2 is violated:

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2.1 In MODE I or 2, restore compliance and be in MODE 3 within 2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.

ce with 2.2.3 Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months notify the NRC Operations Center, in accordan 10 CFR 50.72.

t Vogtle.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , notify the Plant Manager and Vice Presiden 2.2.4 ared and submitted 2.2.5 Within 30 days a Licensee Event Report (LER) shall be prep to the NRC pursuant to 10 CFR 50.73.

d by the NRC.

2.2.6 Operation of the unit shall not be resumed until authorize 2.0-1 Amendment No. 148 (Unit 1)

Vogtle Units I and 2 Amendment No. 128 (Unit 2)

Reactor Core SLs B 2.1.1 BASES n System BACKGROUND The proper functioning of the Reactor Protectio tion of the ents viola (continued) (RPS) and main steam safety valves prev reactor core SLs.

a result of normal APPLICABLE The fuel cladding must not sustain damage as or core SLs are established to SAFETY ANALYSES operation and AOOs. The react ria:

wing fuel desig n crite preclude violation of the follo rience centerline fuel

a. The hot fuel pellet in the core must not expe melting; and a 95% confidence level

b. There must be at least 95% probability at (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience DNB.

ties in plant operating In meeting the DNB design criterion, uncertain fuel fabrication parameters, nuclear and thermal parameters, cons idered. As described parameters, and computer codes must be been statistically rtain ties have in the FSAR, the effects of these unce mine design limit combined with the correlation unce rtain ty to deter DNB desig n crite rion. The Vantage 5 DNBR values that satisfy the with design limit DNBR fuel is analyzed using the WRB-2 correlation ble cells, respectively.

values of 1.24 and 1.23 for the typical and thim lation with design The Lopar fuel is analyzed using the WRB-1 corre al and thimble cells, limit DNBR values of 1.23 and 1.22 for the typic respectively.

rming the safety Additional DNBR margin is maintained by perfo een the design and analyses to a higher DNB limit. This margin betwt known DNBR safety analysis limit DNBR values is used to offse provide DNBR to penalties (e.g., rod bow and transition core) and margin for operating and design flexibility .

bination with The Reactor Trip System setpoints (Ref. 2), in com d to prev ent any antic ipate all the LCOs, are designed Coolant combination of transient conditions for Reactor L POWER eratu re, press ure, and THE RMA System (RCS) temp boili ng ratio eate level that would result in a departure from nucl (continued)

B 2.1.1-2 Revision No.0 Vogtle Units I and 2

Reactor Core SLs B 2.1.1 BASES the average SAFETY LIMITS centerline temperature remains below melting, thatenthalpy of enthalpy in the hot leg is less than or equa l to the (continued) limits defined by saturated liquid, or that the exit quality is within the the DNBR correlation.

factor limits provided in The curves are based on enthalpy hot channel 2.1.1 -1 shows an example of a the COLR. The dashed line of Figure B the various RPS it illus trate s limit curve at 2235 psig. In addition, reaching the limit.

functions that are designed to prev ent the unit from the AFD is within the The SL is higher than the limit calculated when or trip.

(AI) function of the overtemperature AT react limits of the F 1

AFD effec t on the When the AFD is not within the tolerance, the ints to provide ce the setpo overtemperature AT reactor trips will redu s. 3 and 4).

protection consistent with the reactor core SLs (Ref these are the only APPLICABILITY SL 2.1.1 only applies in MODES 1 and 2 because protection is critic al. Auto mati c MODES in which the reactor ES I and 2 to functions are required to be OPERABLE during MOD main steam safety ensure operation within the reactor core SLs. The ent RCS heatup to prev valves or automatic protection actions serve to or trip function, which the reactor core SL conditions or to initiate a react or trip functions are forces the unit into MODE 3. Setpoints for the react

) Instrumentation. In specified in LCO 3.3.1, Reactor Trip System (RTS since the reactor is ired MODES 3, 4, 5, and 6, Applicability is not requ not generating sign ifica nt THE RMA L POW ER.

Actions to be taken SAFETY LIMIT Section 2.2, SL Violations, provides the Required base s for the Required VIOLATIONS in response to a violation of Safety Limits. The reactor core SLs Actions of Section 2.2 applicable to a viola tion of the are discussed below.

(continued)

B 2.1.1-4 Revision No.0 Vogtle Units 1 and 2

RCS Pressure SL B 2.1.2 BASES (continued)

APPLICABLE The RCS pressurizer safety valves, the main steam safety ngs setti SAFETY ANALYSES valves (MSSVs), and the reactor high pressure trip have not be exceeded.

established to ensure that the RCS pressure SL will t system The RCS pressurizer safety valves are sized to preven more than 10%, as pressure from exceeding the design pressure by r Power Plant specified in Section III of the ASME Code for Nuclea lishe s the require d Components (Ref. 2). The transient that estab s, is a and lift setting relief capacity, and hence valve size requirements the react or trip. During complete loss of external load without a direct y pt that the safet transient, no control actions are assumed, exce the steam valves on the secondary plant are assumed to open when valve setting s.

pressure reaches the secondary plant safety the settings The Reactor Trip System setpoints (Ref. 5), together with for norma l operati on and of the MSSVs, provide pressure protection to t is specifi cally set AOOs. The reactor high pressure trip setpoin (Ref. 5). The safet y provide protection against overpressurization r trip and the RCS press urize analyses for both the high pressure mptions relative safety valves are performed using conservative assu to pressure control devices.

following:

More specifically, no credit is taken for operation of the

a. Pressurizer power operated relief valves (PORVs);

b. Main steam atmospheric relief valves;

c. Steam Dump System;

d. Rod Control System;

e. Pressurizer Level Control System; or

f. Pressurizer Spray System.

8, SAFETY LIMITS The maximum transient pressure (P1-0408, P1-04 18, P1-042 P1-0438) allowed in the RCS press ure vess el unde r the ASME (continued)

B 2.1.2-2 Revision No. 0 Vogtle Units 1 and 2

HL-16 NRC Writ ten Examination KEY

/4.61 4.4 C/A/NEW/ROISRO/NRC/GCW 35._054G2.4.49 001/Ill/LOSS OF MFW Given the following:

unciators/indications occur.

- The Plant is at 67% when the following ann

- ALB15 D03, MFPT A TRIPPED SS

- ALB15 B05, MFPT DISCH HDR LO PRE all loops.

- ALB13 STM GEN FLOW MISMATCH on pressure at 650 p51g.

- MFPT B is at 5900 rpms with discharge ring.

- S/G levels are 46% NR and rapidly lowe Which ONE of the following is the CORRECT action(s) for the crew to perform Water Malfuction?

per AOP 1801 6-C, Condensate and Feed S/G levels.

A. Trip the Reactor and go to E-0 due to power level.

B. Trip the Reactor and go to E-O due to MFP indications.

C Trip the Reactor and go to E-0 due to ed to 5300 rpm.

D. Reduce Turbine load to lower MFP spe Page 71 of 208

HL-16 NRC Written Examination KEY Feedback 054 Loss of Main Feedwater (MFW) 2.4 Emergency Procedures / Plan 2.4.49 Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.

(CFR: 41.10/ 43.2 I 45.6)

K/A MATCH ANALYSIS Question gives a plausible scenario with a MFPT trip and the other running MFPT has apparently sheared a shaft or has some other malfunction. The candidate has to pick the correct IOA to perform in accordance with AOP-l 8016-C, section A.

ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Reactor trip not required due to SG levels which are > 40%.

B. Incorrect. Reactor trip not required due to power which is < 70%

C. Correct. Reactor trip required due to not MFPT running and feeding forward.

D. Incorrect. Reactor trip is required, this action would be correct if trip not required.

REFERENCES AOP-18016-C, Condensate and Feedwater Malfunction section A for MFP Malfunction IQAs steps Al thorugh A7.

17015-1 ALB 15, MFPT Disch Hdr Lo Press alarm procedure 17015-1 ALB 15, MFPT Tripped alarm procedure 17013-1 ALB 13, Stm Gen Flow mismatch alarms VEGP learning obiectives:

LO-LP-60314-02, Describe the operator actions required if during the performance of AOP-18016-C, Condensate and Feedwater Malfunction a loss of SG level is imminent.

LO-LP-6031 4-03, State actions required on loss of both SGFPTs with power> 20%

and power < 20%

Page 72 of 208

Approved By Procedure Number Rev J.B. Stanley Vogtle Electric Generating Plant 18016-C 23.2 Date Approved Page Number CONDENSATE AND FEEDWATER 5I14/09 MALFUNCTION 6 of 22 A. LOSS OF MAIN FEED PUMP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED IMMEDIATE OPERATOR ACTIONS

._A1. Check reactor power LESS THAN

- _A1. Perform the following; QE EQUAL TO 70%.

_a. Trip the reactor.

_b. Go to 1 9000-C, E-0 REACTOR TRIP OR SAFETY INJECTION.

_A2. Check at least one MFP RUNNING

- A2. Perform the following:

AND PROVIDING FLOW.

_a. Trip the reactor.

_b. Go to 19000-C, E-0 REACTOR TRIP OR SAFETY INJECTION.

SUBSEQUENT OPERATOR ACTIONS

._A3. Initiate the Continuous Actions Page.

_*A4. Check SG NR levels GREATER

A4. Perform the following:

THAN 40%.

_a. Trip the reactor.

_b. Go to 19000-C, E-0 REACTOR TRIP OR SAFETY INJECTION.

_A5. Operate MFRVs as required to stabilize SG NR levels at program.

A6. Reduce turbine load to lower the running MFP(s) speed to s5300 rpm.

A7 Adjust control rods AS-NECESSARY TO MATCH TAVG -

TREF.

Printed January 17, 2011 at 19:07

Approved By Procedure Number Rev S. E. Prewitt Vogtle Electric Generating Plant 17015-1 40 Date Approved ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 15 ON PANEL 1 Bi Page Number (10/29/2010 ONMCB 49of 89 WINDOW D03 ORIGIN. SETPOINT MFPT A 1-PS-5346 75 psig TRIPPED 1.0 PROBABLE CAUSE Main Feedwater Pump Turbine (MFPT) A tripped due to one or more of the following:

a. Manual trip using 1-HS-3169 on the QMCB.

b. Low oil pressure to MFPT Bearings.

c. Low oil pressure to MFP Bearings.

d. Low suction pressure to MFP.

e. Low vacuum in Condenser.

f. MFPT Steam Exhaust Valve has closed.

g. MFPT Thrust Bearing wear.

h. Steam Generator HI-HI level or Safety Injection actuation.

Manual operation of Local Trip Lever.

j. Turbine overspeed.

2.0 AUTOMATIC ACTIONS NONE 3.0 INITIAL OPERATOR ACTIONS Initiate 18016-C, Condensate And Feedwater Malfunction.

4.0 SUBSEQUENT OPERATOR ACTIONS NONE Printed January 17, 2011 at 19:07

Procedure Number Rev Approved By S. E. Prewitt Vogtle Electric Generating Pant 1701 5-1 40 15 ON PANEL 1B1 Page Number Date Approved ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 26 of 89 0/29/2010 ON MCB WINDOW B05 T

POIN SET MFPT GIN ORI 1000 psig DISCH HDR 1-PSL-5226 LO PRESS 1.0 PROBABLE CAUSE control malfunction.

1. Main Feedwater Pump Turbine (MFPT) failure or suction pressure to

2. Failure of Condensate System to maintain adequate feed pumps.

3. Feedwater line rupture.

4. SJAE Miniflow Valve 1 -FV-4486 failed open.

5. Condensate Filter/Demineralizer high D/P.

2.0 AUTOMATIC ACTIONS NONE 3.0 INITIAL OPERATOR ACTIONS rol 1 -FIC-4486 to maintain IF 1 -FV-4486 malfunction is indicated, manually cont 1 -FI-4486) greater MFPT sucti on pressure greater than 300 psig and flow (on than 7400 gpm.

Printed January 17, 2011 at 19:09

Procedure Number Rev Vogtle Electric Generating Plant

[Approved By 40 17015-1 S. E. Prew,tt Page Number PANEL 1 Bi Date Approved ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 15 ON 27 of 89 l0/29/2010 ONMCB WINDOW B05 (Continued) 4.0 SUBSEQUENT OPERATOR ACTIONS

1. Verify correct setpoint on the MFPT Speed Controller.

in

2. an MFPT Speed Controller has failed, place the Speed Controller d to MANUAL and raise speed to restore feedwater pressure as require maintain the Feedwater Control Valve in the controlling band.

for

3. Verify MFPT Discharge Valves 1-HV-5208 and 1-HV-5209 are open operating pumps by observing indicating lights on the QMCB.

4. IF level cannot be maintained, initiate 18016-C, Condensate And Feedwater Malfunction.

40 psid, perform the

5. if Condensate Filter/Demineralizer dP is greater than following:

a. Verify Condensate Demin Bypass Valve 1 -PV-30223 begins to open.

b. Implement procedure 13616-1 Condensate Demineralizer System Section 4.4.3 to place the standby vessel in service.

e is indicated, Go To 18008-C

6. jf Main Feedwater or Main Steam leakag Secondary Coolant Leakage.

7. equipment failure is indicated, initiate maintenance.

5.0 COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE

REFERENCES:

1X4DB168-2, 1X3D-BC-C51A, CX5DT11O1-91B Printed January 17, 2011 at 19:09

HL-16 NRC Written Examination KEY

36. 055EA 1.0! 002/1 / 1/STATION B/O/3 .7/3.9 MEM/BANK SUMMER2009/RO/SRO/HL I 6/EMT

- /GCW Which ONE of the following describes a condition that will require the crew to transition out of 19100-C,Loss Of All AC Power prior to restoring power to at least one AC ESF B us?

A CETCs indicate greater than 1200°F and rising.

B. The Tcolds for all RCS loops lower to less than 280°F.

C. Both Intermediate and Source Range startup rates are positive.

D. NR levels in ALL S/Gs drop below 10% with no AFW flow available.

Feedback 055 Loss of Offsite and Onsite Power (Station Blackout)

Ability to operate and monitor the following as they apply to a Station Blackout:

(CFR 41.7/45.5/45.6)

EA1 .01 In-core thermocouple temperatures K/A MATCH ANALYSIS Question places candidate in 19100-C, Loss of All AC Power and he has to determine which of the choices would allow an exit from the procedure prior to power restoration.

ANSWER! DISTRACTOR ANALYSIS A. Correct. With T/Cs> 1 200 F, an exit to the SAMGs procedures is required per step 0

38.

B. Incorrect-Choice B would have have Tcolds lowering to a possible PTS condition challenging the Integrity CSFST. This could occur during a rapid cooldo wn of the primary as directed by the EOP. However, although you would stop the colldow n per procedure, you would not exit at this point due to Tcolds <280°F. The choice is plausible as a major procedure action.

C. Incorrect. During RCS depressurization, IF SR or IR SURs become positiv e, the procedure directs a halt to the depressurization but does not direct exiting the procedure.The choice is plausible as a major procedure action.

D. Incorrect. Plausible because conditions listed indicate a red path on the CSFST safety functions. The procedure states that the CSFSTs are to be monito red for information only and are not to be performed.

REFERENCES Page 73 of 208

HL-16 NRC Written Examination KEY 19100-C, Loss of All AC Power Summer 2009 NRC RO Exam question # 38 VEGP learning objectives:

LO-LP-37031 -08, Using EOP 19100 as a guide, briefly describe how each step is accomplished.

LO-LP-37031 -09, Given a NOTE or CAUTION state ment from the EOP, state the bases for that NOTE or CAUTION statement.Incorrec

t. Choice A would have Tcolds lowering to a possible PTS condition challenging the Integrity CSFST. This could occur during a rapid cooldown of the primary as direc ted by the EOP. However, although you would stop the cooldown per procedure, you wou ld not exit at this point due to Tcolds <

280°F. The choice is plausible as a major proc edure action.

Page 74 of 208

I Approved By Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant 19100-C 35.3 Date Approved Page Numb&

ECA-O.O LOSS OF ALL AC POWER ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

36. Check Containment Ventilation 36. Close dampers and valves.

Isolation using ATTACHMENT C:

IF dampers and valves can NOT Computer Points closed, THEN locally close.

-OR Handswitch Indication

37. Check Containment radiation - LESS _37. Close Containment isolation THAN mR/HR. valves as necessary using ATTACHMENT D.

IF valves can NOT be closed, THEN locally close at least one valve at each penetration as time permits.

_38 Check Core Exit TCs LESS THAN

- _38 if Core Exit TCs greater than 1200°F 1200°F and rising, THEN go to SACRG-1, SEVERE ACCIDENT CONTROL ROOM GUIDELINE INITIAL

RESPONSE

39. Check if AC emergency power is restored:

a. AC Emergency Busses AT -

_a. Go to Step 40.

LEAST ONE ENERGIZED.

_b. GotoStep42.

Printed January 18, 2011 at 07:30

38. 055 EAt .01 001/NEW/IHIGHERJIRO/SIJMMERJ2/20091N0 Which ONE (1) of the following describes a condition that will require the crew to transition out of EOP-6.O, Loss of All ESFAC Power prior to restoring power to at least one AC ESF Bus?

A. The Tcolds for all three RCS loops decrease to less than 280°F.

B. Both Intermediate and Source Range startup rates are positive.

C. Wide range levels in ALL 3 S/Gs drop below 30%.

D Core Exit T/Cs indicate greater than 1200°F and rising.

V

HL-16 NRC Written Examination KEY

37. 055K3.0 1 001 /2/2/COND AIR REMOVAL/2.5/2.7 IvEEM/NEW/RO/SRO/NRC/GCW Given the following:

- Reactor power is 100%.

- The Instrument Air supply is lost to the Condenser Vacuum Breakers.

Which ONE of the following is CORRECT?

A Condenser pressure is unaffected, Vacuum breakers can be opened.

B. Condenser pressure rises rapidly and Main Turbine Mw output lowers.

C. Condenser pressure is unaffected, Vacuum breakers cannot be opened.

D. Condenser pressure rises rapidly and Main Turbine Mw output rises.

Page 75 of 208

HL-16 NRC Written Examination KEY Feedback 055 Condenser Air Removal System (CARS)

Knowledge of the effect that a loss or malfunction of the CARS will have on the following:

(C FR: 41.7/45.6)

K3.01 Main condenser K/A MATCH ANALYSIS Questions presents a plausible scenario where instrument air supply is lost to the condenser vacuum breakers. The candidate has to determine the effect on condenser pressure and if vacuum breakers can still be opened.

ANSWER I DISTRACTOR ANALYSIS A. Correct. Without instrument air, a local air accumulator will allow the vacuum breakers to be opened. Loss of air will not cause vacuum breaker status to change.

B. Incorrect. Loss of instrument air will not cause vacuum breakers to fail open, condenser pressure will remain stable.

C. Incorrect. Without instrument air, a local air accumulator will allow the vacuum breakers to be opened. Loss of air will not cause vacuum breaker status to change.

D. Incorrect. Loss of instrument air will not cause vacuum breakers to fail open, condenser pressure will remain stable.

REFERENCES V-LO-PP-26201 Condenser Air Removal slides 138,139,142,143 Procedure 17019-1 ARP for ALB1 9 on Panel 1 B2 on MCB (Window B04)

VEGP learning obiectives:

N/A Page 76 of 208

c e cal Manual Operation Of The ondenser Vacuum Breakers One per condenser section.

If instrument air is lost, the vacuum breakers can be manually overridden locally by actuating a switch device on the solenoid actuator.

Locally mounted air accumulator tanks provide air necessary to open the valves.

2O1-O2 138

J

-I

o Vacuum Breakers

Air accumulator attached to air supply for each valve (backup air supply 30 gal)

Normal air supply to valve turbine building instrument air 142

Instrumentation and Controls

MCB Control Switch One switch controls all three valves Local Manual Operation Vacuum breakers can be manually overridden by operating the local solenoid operator Manual operation only under

,zOl-02 direction from control room 143

Approved By Procedure Number Rev S. E. Prewiti Vogtle ElectrIc Generating Plant 17019-1 25 Date Approved ANNUNCIATOR RESPONSE PROCEDURES FOR ALS 19 ON PANEL 1 B2 Page Number 8/1O/201O ON MCB 20 of 42 WINDOW B04 ORIGIN SETPOINT TURB CNDSR 1-PISH-6292A and B 24.92° Hg. Vacuum LO VAC 1-PISH-6293A and B Lowering 1-PISH-6294A and B (1 out 2 inst. in any one exhaust hood) 1.0 PROBABLE CAUSE

1. Condenser Air Ejector Q. Gland Seal Steam System fault.

2. Excessive air leakage into the condenser

3. Insufficient circulating water flow for the existing turbine load.

4. Loss of power to instruments listed above.

5. High ambient outside air temperature.

2.0 AUTOMATIC ACTIONS WARNING if Condenser vacuum continues to lower, a turbine trip will occur at 22.42 Hg.

vacuum lowering (2 out of 2 inst. in any one exhaust hood (IPC group 229)).

An automatic Rx trip will occur IF above P9.

1. Mechanical Vacuum Pumps will auto start WHEN condenser vacuum is less than 25in Hg. (Vacuum lowering) if in AUTO PULL TO LOCK.

2. Steam Dump operation is blocked WHEN condenser vacuum is less than 24.92 Hg. (Vacuum lowering), (2 out of 2 inst. in any one exhaust hood),

(Permissive C-9 lost OR blocked).

Printed January 18, 2011 at 7:49

Date: 1/6/2011 Time: 09:32:19 AM e1 I

1-u LUND N L/

1 U SH(L B COND. VAU UMf fiAX R S//( s TO CON&7VL9ER L4ILI C

Title:

C:\DATA\HL-15 Recovery References\P&lDs Unit 1\1X4DB182.cal

HL-16 NRC Written Examination KEY

38. 056AK3 .01 001/1/I ILOSP/3 .5/3.9 MEM/MOD BANKJRO/SRO/NRC/GCW Which ONE of the following is correct regarding the response of the Containment Cooling fans following a simultaneous UV on both trains of SSPS?

Upon receipt of the sequencer start signal...

A. eight Containment Cooling fans start in low speed at 30.5 seconds.

Reason: High moisture and density air could damage motor windings.

B. eight Containment Cooling fans start in high speed at 30.5 seconds.

Reason: High moisture and density air could damage motor windings.

C. four Containment Cooling fans start in low speed at 30.5 seconds, then four more start in low speed at 50.5 seconds.

Reason: To prevent overloading the DG during loading.

D four Containment Cooling fans start in high speed at 30.5 seconds, then four more start in high speed at 50.5 seconds.

Reason: To prevent overloading the DG during loading.

Page 77 of 208

HL-16 NRC Written Examination KEY Feedback 056 Loss of Offsite Power Knowledge of the reasons for the following responses as they apply to the Loss of Offsite Power:

(CFR 41.5,41.10/ 45.6 I 45.13)

AK3.01 Order and time to initiation of power for the load sequencer K/A MATCH ANALYSIS Questions asks the candidate the proper sequence of the containment coolers following a simultaneous LOSP (UV) on both SSPS trains.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. This is the correct times and sequence for SI not UV. However, this is the correct basis for for an SI sequence.

B. Incorrect. This is the correct times and sequence for SI not UV. The fan speeds are wrong for a UV. This is the correct basis for an SI sequence.

C. Incorrect. This is the correct times and sequence for a UV sequence but the fan speeds are wrong. This is the correct basis for a UV sequence.

D. Correct. This is the correct times and sequence for a UV sequence with the correct fan speeds. This is the correct basis for a UV sequence.

REFERENCES V-LO-PP-29101, Containment HVAC Systems, slides # 22 and 23.

LO-PP-291 01-13-03 VEGP learning obiectives:

LO-PP-29101-13, State why two speeds are used 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 setpoints where applicable.

Page 78 of 208

0 C Containment Cooling Units Emergency 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 windings are energized.

Four fans (single Train) in slow speed provide adequate heat removal

Containment Cooling Units Emergency 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

1. LO-PP-29101- 13 OO3ILOPP291OI/LO.TA-37004/022A3.O1/////

Which ONE of the following best describes the response of the Containment Cooling fans following an SI actuation on both trains of SSPS. Assume all control switches in auto and transfer switches set to Control Room.

Upon receipt of the sequencer start signal A eight Containment Cooling Unit fans start in low speed at 30.5 seconds.

B. eight Containment Cooling Unit fans start in high speed at 30.5 seconds.

C. four Containment Cooling Unit fans start in low speed at 30.5 seconds, then four more start in low speed at 50.5 seconds.

D. four Containment Cooling Unit fans start in high speed at 30.5 seconds, then four more start in high speed at 50.5 seconds.

HL-16 NRC Written Examination KEY

39. 057AA2.02 001/1/1/LOSS OF VITAL AC/3.7/3.8 MEMJNEWIRO/SRO/NRC/GCW The following indications occur with the unit at full power:

- Accumulator # 1 level and pressure indications both fail due to a loss of power.

Which ONE of the following would be correct regarding:

1) power supply to accumulator level and pressure indicators, and

2) the direction the instruments would fail?

power supply failure direction A. 125V DC fails low B. 125V DC fails high C 120 VAC fails low D. 120 VAC fails high Feedback 057 Loss of Vital AC Electrical Instrument Bus Ability to determine and interpret the following as they apply to the Loss of Vital AC Instrument Bus:

(CFR: 43.5/45.13)

AA2.02 Core flood tank pressure and level indicators K/A MATCH ANALYSIS Question asks the basic power supplies to the ECCS accumulator pressure and level indicators and which direction the isntruments fail on loss of power.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Control board instruments are not power from 1 25V DC. Failure direction part is correct.

B. Incorrect. Control board instruments are not power from 1 25V DC. Failure direction part is incorrect.

C. Correct. Control board instruments are from 120V AC and fail low on loss of power.

D. Incorrect. Control board instruments are from 120 V AC but the failure mode is low on loss of power.

Page 79 of 208

HL-16 NRC Written Examination KEY REFERENCES 18032-1 Loss of 120 vac Instrument Power, Attachment J, 1 NY4N Load List Drawing 1X3D-AA-GO5B for Panel 1 NY4N Drawing 1 X3D-AA-GO2B for Panel 1 NY2N 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 I. 1NYJ

m. 1NYR

n. 1NYS

o. 1NYRS

p. 1NYO1 Page 80 of 208

Approved By Procedure Number Rev C. S. Waidrup Vogtle Electric Generating Plant 18032-1 28.2 Date Approved Pagf LOSS OF 120V AC INSTRUMENT POWER ATTACHMENT J Sheet 1 of 1 TABLE 1 - PANEL 1NY4N LOAD LIST BREAKER LOAD 03 SPARE 04 ç PROCESS RACK CONTROL PNL GROUP 1 (Redundant Power Supply 1 NYSO4) 05 SPARE 06 AUX RELAY RACK3 07 CONTROL BOARD DEMULTIPLEXER 08 SPARE 09 PLANT VENT RAD MON SMPLG VLVS 1 HV-1 2835A,B 10 DATA MODULE 1 RX-1 2444 PLANT VENT POST ACCIDENT MONITOR 11 PLANT VENT RAD MON SMPLG FLOW XMTR 1 -FT-i 2835 12 SPARE 13 SPARE 14 AMSAC CABINET 1-1626-Q5-AMS 15 SPARE 16 SPARE 17 SPARE 18 SPARE 19 SPARE 20 SPARE 21 SPARE 22 BOP CONTROL PANEL 2 23 SPARE 24 SPARE 25 SPARE 26 SPARE 27 SPARE 28 SPARE END OF ATTACHMENT J Printed January 18, 2011 at 08:45

Date: 1I16111 Time: P38:27 AM w

8 7 6 SHEET REVISION CARD FRAME CHANNEL NO. CHANNEL DESCRIPTION (1 1A>

2 09 POWER SUPPLY FAILURE 3 02 F145, P-153 RCP I SEAL WATER INJECTION FLOW HO*

RCP 1 SEAL AP STA RCP 1 SEAL LEAK OFF CONTROL - HI RANGE TSE 4 7 02 F-161, F157 RCP 1 SEAL LEAK OFF CONTROL - [0 RANGE 5 2 02 L- 102 BORIC ACID TANK 1 LEVEL 6 7 02 T- 126 REG HX CHARGING TEMPERATURE F-I 10 BORIC ACID BLEND SYSTEM - BORIC ACID FLOW 7 1 OA 02 F-Ill BORIC ACID BLEND SYSTEM - TOTAL MAKE UP WATER FLOW 8 1 OA 02 S 7 02 L-1 12 VOLUME CONTROL TANK LEVEL SPARE 9 SPARE SPARE 13 7 03 F-406. F-138 RV VENT LETDOWN FLOW TO PRT 14 7 03 H -606 RHR HX 1 OUT 15 7 *D3 L-95O, P-96D ACCUMULATOR TANK 1 LEVEL ACCUMULATOR TANK 1 PRESSURC 16 7 03 L-952. P-962 ACCUMULATOR TANK 2 LEVEL ACCUMULATOR TANK 2 PRESSURE 7 4 03 L-99O REFUELING WATER STORAGE TANK LEVEL 18 9 03 F183, T449 EMERGENCY BORATION FLOW, PRZR SAFE REL. VLV DISCH TEMP.

9 7 03 T-604 RHR 1 RETURN TEMP 20 8 03 F-61 8 RHR HX 1 BYPASS FLOW CONTROL 21 9A 03 F -61 8 LOW HEAD SAFETY INJECTION FLOW 22 SPARE 23 SPARE 24 SPARE 25 SPARE 26 SPARE 27 SPARE STEAM GENERATOR LEVEL CONTROL - LOOP 28 6 04 L-519, F-510 04 F-51O STEAM GENERATOR LEVEL CONTROL - LOOP 1 28 A F-Sb, F-512 STEAM GENERATOR LEVEL CONTROL - LOOP 29 04 30 SPARE 31 7 04 T-401 REACTOR VESSEL FLANGE LEAK OFF TEMPERATURE 32 7 04 P-469, L470 PRESSURIZER RELIEF TANK PRESSURE.

PRESSURIZER RELIEF TANK LEVEL 33 7 04 T-463 PRESSURIZER RELIEF DISCHARGE TEMPERATURE 34 6 04 [-550 FEEDWATER CONTROL 35 SPARE 36 SPARE 37 2 05 -460, P-455 PRFScIIRI7FR FVF[ frnNTRfll PRESSURIZER PRESSURE CONTROL 38 8 05 P-155 PRESSURIZER PRESSURE CONTROL 38A 8 05 P-455 PRESSURIZER PRESSURE CONTROL 39 8 07 P-455 PRESSURIZER PRESSURE CONTROL 39A 8 07 P-455 PRESSURIZER PRESSURE CONTROL 40 6 05 P-455 PRESSURIZER PRESSURE CONTROL 41 1 1A 05 P-506, P-507 STEAM DUMP STEAM GENERATOR HEADER PRESSURE 42 6 05 P-5D8, P-509 FEEDPUMP SPEED CONTROL 43 6 05 F-5D9, 5-509, P-509 FEEDPUMP SPEED CONTROL 44 3A 05 5-509 FEEDPUMP SPEED CONTROL 45 2A 05 S FEEDPUMP SPEED CONTROL 46 SPARE 47 SPARE 48 8 06 P-405, T-454 RCS WIDE RANGE PRESS.

PRZR. VAPOR TEMP.

49 6 06 T-500 STEAM DUMP CONTROL 50 6 06 U-500, T-500 STEAM DUMP CONTROL 51 SPARE 52 7 06 F-Sb, F-512 STEAM GENERATOR LEVEL CONTROL - LOOP I 53 7 06 F-511, E-513 STEAM GENERATOR LEVEL CONTROL - LOOP I PRINTED CIRCUIT CARD INDEX

Title:

C:\DATA\Documentum\Viewed\1 X6AUOI -00369.cal

Date: 1/18/911 Time: 39:O5 AM O6 I flNNECTOR 003 0 TWISTED WIRE CONNECTED TO

/ CONNECTOR NUMBER EPOINT CONNECTOR 067 i /

4MBER CARD CONTINUATION SYMBOL I 26-OO3-O1 02-001-25 26-067-25 LEGEND 02-00 1 -26\2b-O67-26 26-003-02 118 VAC SOiRCE >> -

CONTROL BOARD MOUNTED CABLE BETWEEN FIELD MOUNTED DIFFERENT ITEM NO 5.

CAPABLE OF PROCESSING MULTIPLE CHANNELS.

VI . REFERENCES ELECTRICAL SIGNAL LINE THIS SYMBOL SHOULD BE USED IN CONJUNCTION 4.

WILH OTHER PLANT DOCUMENTATION, SUCH AS A. CHANNEL TEST SWITCH LOGIC DWG. 8817060 SHEET 3 TERMINAL NUMBER THE ASSOCIATED CARD LIST. TO DETERMINE IF ALL ALARM CONTACTS FOR ANNUNCIATION OPEN TO ALARM UNLESS VII.

MULTIPLE SIGNALS ARE PROCESSED BY THE CARD. TNDIFATFD OT-IFRWISF.

CRCUND SYMBOL VIII. PIN 33 OF ALL CONNECTORS MOUNTED ON TERMINATION PANEL (26)

CARD CONNECTOR (PIN) NUMBER ARE JUMPERED USING 22 BLUE TEFLON WIRE 19022.

CONNECTOR PIN X CARD MNEMONIC 18.0 CONFIGURATION GROUP NUMBER CROSS REFERENCE LETTER SHEET NO.

INDICATES ASTEC/7300 INTERFACE MONITOR POINTS 4 LOOP PLANT SYSTEM FORM DWG 8799055 SH. 01 CABINET 05 CONTROL 1 WESTINGHOUSE ELECTRIC CORPORATION TITLE INTERCONNECTING WIRING DIAGRAM-CABINET OS GEORGIA PWR. CO. ALVIN W. VOGTLE UNITS 1 2 DINSIOS IS SCALE

[.

Z ociR. i%.. 4A4 Ui

. 8 8 1 7 D 60 aD SHEET I EGUIPI4ENT TAG 1-1604-05-PCi Si,,. I U 0 PJHSG. PR USA INDUSTRY SYSTEMS DIVISION 1 3 2 I 1 I I

Title:

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Date: 1/18(111 Time: >4:59 AM 7

0 6 5 8 I I CHANNEL NO. CHANNEL DESCRIPTION IX. CARD SHEET REVISION CARD FRAME WHEN THE (9A) BE USED T 2 09 POWER SUPPLY FAILURE F-143, P-151 RCP 3 SEAL WATER INJECTION FLOW, HOWEVER, 3 02 STANDARD RCP 3 SEAL P F-159, F-155 RCP 3 SEAL LEAK OFF CONTROL - HI RANGE, THEN THE 4 6 02

- LO RANGE RCP 3 SEAL LEAK OFF CONTROL 5 6 02 T-452 SPRAY LINE TEMPERATURE 6 6 02 T- 1 03 BORIC ACID TANK 1 TEMPERATURE 7 6 02 SPARE 8 6 02 P-614. T-612 RHR PUMP 1 DISCHARGE PRESSURE, RHR UI DISCHARGE TEMPERATURE 9 (7A) 02 P -131 LOW PRESS. LETDOWN PRESSURE 10 SPARE 11 SPARE 12 SPARE 13 8 03 H-123, H-182 LETDOWN HX CONTROL, CHARGING HEADER CONTROL 14 3 03 H-i 28 LETDOWN RHR HX CONTROL 15 6 03 L-954, P-964 ACCUMULATOR TANK 3 LEVEL, ACCUMULATOR TANK 3 PRESSURE 16 6 03 L-956, P-966 ACCUMULATOR TANK 4 LEVEL.

ACCUMULATOR TANK 4 PRESSURE 17 03 L-992 REFUELING WATER STORAGE TANK LEVEL 18 (7A) 03 P -923 SAFETY INJECTION PUMP 2 DISCHARGE PRESSURE 19 SPARE 20 SPARE

- LOOP 3 21 6 03 F-53O, F-532 STE AM GENERATOR LEVEL CONTROL

- LOOP 3 22 6 03 F-531. F-533 STE AM GENERATOR LEVEL CONTROL 23 SPARE 24 SPARE 25 SPARE 26 SPARE 27 SPARE 28 5 04 L-539. F-530 STEAM GENERATOR LEVEL CONTROL LOOP 3 28A 0 04 F -530 STEAM GENERATOR LEVEL CONTROL LOOP 3 29 9 04 F-530. F-532 STEAM GENERATOR LEVEL CONTROL LOOP 3 30 SPARE 31 SPARE 32 6 04 T -464 PRESSURIZER SAFETY VALVE DISCHARGE TEMPERATURE 33 5 04 L-570 FEEDWATER CONTROL

Title:

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Date: 1/18P11 Time: (>45:34 AM I

CABLE BETWEEN DIFFERENT ITEM NOS.

VI. REFERENCES A. CHANNEL TEST SWITCH LOGIC DWG. 8817D62 SHEET 3 8 4 VII. ALL ALARM CONTACTS FOR ANNUNCIATION OPEN TO ALARM UNLESS INDICATED OTHERWISE.

VIII. PIN 33 OF ALL CONNECTORS MOUNTED ON TERMINATION PANEL (26)

ARE JUMFERED USING Z2 BLUE TEFLON WIRE IO22.

LOOP PLANT SYSTEM FORM DWG SH.

CABINET 07 CONTROL 3 A

WESTINGHOUSE ELECTRIC CORPORATION TITLE INTERCONNECTING WIRING DIAGRAM-CABINET GEORGIA PWR. CO. ALVIN W. VOGTLE UNITS 1 & 2 f%_ OlI6IOP6 N INcIS SCALE OIVFCDLHAUE1 D.

OlW V4 881 7 D 62

- SHEET 1 EQUIPMENT TAG 1-16O-Q5-PC3 st,v. I L L w INDUSTRY SYSTEMS DIVISION PISaQPA L$A 3 2 MICROSTAT THIS DWG. IS REFERENCED IN VENDOR MANUAL: Souther N /A TA8/SECT. N/A PAGE N/A VOGILE FIGURE N/A VERSION 15.0 IDATE 081506 TITLE:

Title:

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HL-16 NRC Written Examination KEY

40. 058G2.4.49 001/1/I/LOSS OF DC/4.6/4.4 MEMINEW/RO/SRO/NRC/GCW The plant is in Mode 6 with core offload in progress.

- A fault occurs on BD1, 125V DC Vital Bus.

- The required DC sources for the present mode are NOT met.

Considering the following Technical Specification actions:

1) Suspend CORE ALTERATIONS.

2) Suspend movement of irradiated fuel assemblies.

3) Initiate actions to suspend operations involving positive reactivity additions.

Which of these actions would have to all be IMMEDIATELY performed for the present Mode and DC source status?

A. # 1 only B. #land#2only C. # 1 and # 3 only D #1, #2, and #3 Page 81 of 208

HL-16 NRC Written Examination KEY Feedback 058 Loss of DC Power 2.4 Emergency Procedures I Plan G2.4.49 Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.

(CFR: 41.10/43.2/45.6)

K/A MATCH ANALYSIS The question presents a scenario during mode 6 where the required DC sources are lost. The candidate has to choose determine which of several listed IMMEDIATE LCO action apply. All 3 actions apply for the LCO.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect. All 3 actions must be performed.

B. Incorrect. All 3 actions must be performed.

C. Incorrect. All 3 actions must be performed.

D. Correct.

REFERENCES Tech Spec 3.8.5, DC Sources Shutdown Tech Spec 3.8.10, Distribution Systems Shutdown VEGP learning obiectives:

LO-LP-3921 2-01, For any given item in section 3.8 of Tech Specs, be able to;

a. State the LCQ

b. State any one hour or less required actions.

Page 82 of 208

DC Sources Shutdown 3.8.5 3.8 ELECTRICAL POWER SYSTEMS 3.8.5 DC Sources Shutdown LCO 3.8.5 DC electrical power sources shall be OPERABLE to support the DC electrical power distribution subsystem(s) required by LCO 3.8.10, Distribution Systems Shutdown.

APPLICABILITY: MODES 5 and 6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required DC A.l.l Declare affected required Immediately electrical power sources feature(s) inoperable.

inoperable.

OR A.2.1 Suspend CORE Immediately ALTERATIONS.

AND A.2.2 Suspend movement of Immediately irradiated fuel assemblies.

AND A.2.3 Initiate action to suspend Immediately operations involving positive reactivity additions.

AND A.2.4 Initiate action to restore Immediately required DC electrical power subsystems to OPERABLE status.

Vogtle Units I and 2 3.8.5-1 Amendment No. 96 (Unit 1)

Amendment No. 74 (Unit 2)

Distribution Systems Shutdown 3.8.10 3.8 ELECTRICAL POWER SYSTEMS 3.8.10 Distribution Systems Shutdown LCO 3.8.10 The necessary portion of AC, DC, and AC vital bus electrical power distribution subsystems shall be OPERABLE to support equipment required to be OPERABLE.

APPLICABILITY: MODES 5 and 6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Declare associated Immediately AC, DC, or AC vital bus supported required electrical power feature(s) inoperable.

distribution subsystems inoperable.

A.2.1 Suspend CORE Immediately ALTERATIONS.

AND A.2.2 Suspend movement of Immediately irradiated fuel assemblies.

AND A.2.3 Initiate action to suspend Immediately operations involving positive reactivity additions.

AND (continued)

Vogtle Units I and 2 3.8.10-1 Amendment No. 96 (Unit 1)

Amendment No. 74 (Unit 2)

HL-16 NRC Written Examination KEY

41. 059K4.18 OO1/2/1/MFW-FW REDUCTION/2.8/3.O C/A/MOD BANK/RO/SRO/NRC/GCW Reactor trip has occurred from 100% power with the following conditions:

- Reactor Trip Breaker B has remained closed.

- Loop 1 Tave = 563°F

- Loop 2 Tave = 563°F

- Loop 3 Tave = 566°F

- Loop 4 Tave = 564° F The crew is performing 19001-C, Reactor Trip Response step for Check FW Status.

Which ONE of the following is CORRECT regarding the expected status of FWI?

A. MFRVs-OPEN, BFRVs-OPEN B. MFRVs-SHUT, BFRV5-OPEN C MFRVs OPEN,

- BFRVs SHUT D. MFRVs-SHUT, BFRVs-SHUT Page 83 of 208

HL-16 NRC Written Examination KEY Feedback 059 Main Feedwater (MFW) System Knowledge of MFW design feature(s) and/or interlock(s) which provide for the following:

(CFR: 41.7)

K41 8 Automatic feedwater reduction on plant trip K/A MATCH ANALYSIS Question gives a plausible scenario where a Reactor Trip Breaker fails to open after a reactor trip. Candidate must determine the effects on a plant control system (FWI) to determine actions to take to stabilize the plant. In this instance, FWI has occurred but due to Rx Trip Breaker B closed, the MFRVs will not go shut. UO will have to manually close the valves.

ANSWER / DISTRACTOR ANALYSIS A. Incorrect. MFRVs would be OPEN and BFRVs would be SHUT.

B. Incorrect. MFRVs would not shut in this condition with RTB B closed.

C. Correct. Only MFRVs would be open, BFRVs would be SHUT.

D. Incorrect. MFRVs would be open and BFRVs would be SHUT.

REFERENCES 19001-C, Reactor Trip Response V-LO-PP-281 03, Reactor Trip and ESFAS Signals (slides 11-13, 16, 18-20, 22, 137-139)

V-LO-PP-1 8101 Cond and Feedwater (slides 157-1 65)

LOIT Bank 007EA2.02-01 VEGP learning obiectives:

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.

Page 84 of 208

Approved 8y Procedure Number Rev J.B. Stanley Vogtle Electric Generating Plant 19001-C 31.1 Page Number ES -0.1 REACTOR TRIP RESPONSE ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

5 Check FW status:

_a. Average RCS temperature - _a. WHEN average RCS LESS THAN 564°F. temperature is less than 564°F, THEN perform Step 5.b.

GotoStep6

_b Verify FW isolation valves closed

. MFIVs

. BFIVs

. MFRVs

. BFRVs

6. Check total feed flow capability to 6. Establish feed flow to the SGs from SGs GREATER THAN 570 GPM

- the following:

AVAILABLE.

_AFW System

-OR-Main Feedwater:

a. jf required, start a MFP by performing the following:

_1) Check Condenser vacuum established.

_2) Lower GE pot setting to zero.

_3) Reset MFP.

Step 6 continued on next page Printed January 18, 2011 at 08:23

What gives you Train A P-4?

RTA and its BYA both open gives you Train A P-4

c o What does Train A P-4 Do?

-Generates a Main Turbine Trip (mechanical)

V-LO-PP-281 03-6.2 12

Train WAW FWI 564°F V-LO-PP-281 03-6.2 13

C) C)

P-4 Train A Seals in the Train A Logic for Feed Water Isolation due to either a Safety Injection or Hi-Hi Steam Generator Water Level V-LO-PP-281 03-6.2 16

C) 0 PEA,CTC:IR TRIP E V.P :2 B 1 ZL-4[1C144 What gives you Train B P-4?

RTB and its BYB both open gives you Train B P-4 V-LO-PP-281 03-6.2 18

0 What does Train B P-4 Do?

-Generates a Main Turbine Trip (electrical)

V-LO-PP-28103-6.2 19

c What does Train B P- 4 Do?

Lo-Tavg I P-4 FWI V-LO-PP-281 03-6.2 20

C) 0 What does Train B P-4 Do?

Seals in Train B FWI signal on Hi-Hi Steam Generator Level and SI Train B V-LO-PP-281 03-6.2 22

C What is the function of the FWI signal generated by P-14?

Prevents water intrusion into the steam lines (main turbine and MFP turbines)

Trips the Main Turbine and the Main Feed Pump Turbines in addition to the isolation of the Feed Water Valves V-LO-PP-281 03-6.2 137

Lo Tavg / P-4 FWI Reset

Place both FWI Isolation switches to reset position.

V-LO-PP-281 03-6.2 138

C 0 P-14 or SI FWI Reset

Actuation signal has to be cleared

The P-4 seal-in is removed by cycling the reactor trip breakers.

Reset FWI by placing both FWI Isolation switches to reset position.

V-LO-PP-28103-6.2 139

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1. 007EA2.02 OO1/1/IIRX TRIP NO AUTO/C/A-4.3INEWIROIHL-15 AUDIT/TNTIDS trip has occurred with the following conditions:

Reactor

- Reactor Trip Breaker B has remained closed.

- Loop 1 Tave 563°F

-Loop2Tave=563°F

-Loop3Tave=566°F

- Loop 4 Tave = 564°F The crew is performing 19001 -C, Reactor Trip Response step for Check FW Status.

Which ONE of the following is CORRECT regarding the expected status of FWI for the given plant conditions?

MFIVs BFIVs MFRVs BFRVs A closed closed open closed Manually close all the MFRVs.

B. closed closed closed closed No actions are required to be taken.

C. closed closed open open Manually close all the MFRVs and BFRVs.

D. open open open open Verify valves close when Lo Tave FWI coincidence is met.

Page: 1 of 2 1/18/2011

Feedback K/A 007 Reactor Trip Stabilization:

EA2.02 Ability to determine and interpret the following as they apply to a reactor trip:

Proper actions to be taken if the automatic safety functions have not taken place K/A MATCH ANALYSIS Question gives a plausible scenario where a Reactor Trip Breaker fails to open after a reactor trip. Candidate must determine the effects on a plant control system (FWI) to determine actions to take to stabilize the plant. In this instance, FWI has occurred but due to Rx Trip Breaker B closed, the MFRVs will not go shut. UO will have to manually close the valves.

ANSWER I DISTRACTOR ANALYSIS A. Correct.

B. Incorrect. MFRVs would not shut in this condition with RTB B closed.

C. Incorrect. Only MFRVs would be open and need closing, not the BFRVs.

D. Incorrect. FWI coincidence of 2/4 < or = to 564oF is met. All valves should be shut except for the MFRVs.

REFERENCES 19001-C, Reactor Trip Response V-LO-PP-281 03, Reactor Trip and ESFAS Signals VEGP learning obiectives:

Categories Task Number (LO-TA): 1 Objective/Source:

K/A: RX TRIP NO AUTO

- Exam/Question Type: C/A-4.3 Cognitive Level: NEW Origin/Rev Info: RO Reference I: HL- 15 AUDIT Reference 2: TNT/DS Page: 2 of 2 1/18/2011

HL-16 NRC Written Examination KEY

42. 059K4.19 OO1/2/1/MFW-FWIJ3.2/3.4 MEMJNEWIRO/SROINRC/GCW A plant feedwater transient at 100% power has resulted in the following annunciator:

FW I SI or P-14 SG HI-HI LVL Which ONE of the following is correct regarding:

1) Impact on MEW system, and

2) actions necessary to restore MFW capability?

A. 1) only closes all EWI valves.

2) reactor trip breakers are required to be cycled to reset FWI.

B. 1) only closes all FWI valves.

2) reactor trip breakers are NOT required to be cycled to reset FWI.

C 1) closes all FWI valves and trips both MFPTs.

2) reactor trip breakers are required to be cycled to reset FWI.

D. 1) closes all FWI valves and trips both MFPTs.

2) reactor trip breakers are NOT required to be cycled to reset FWI.

Page 85 of 208

HL-16 NRC Written Examination KEY Feedback 059 Main Feedwater (MFW) System Knowledge of MFW design feature(s) and/or interlock(s) which provide for the following:

(CFR: 41.7)

K4.1 9 Automatic feedwater isolation of MFW K/A MATCH ANALYSIS Question presents a scenario where a FWI annunciator is received. The candidate has to choose the correct impact on the MEW system and actions to restore MEW.

ANSWER! DISTRACTOR ANALYSIS A. Incorrect. FWI valves shut and MFPTs trip. Trip breakers are required to be cycled to reset FWI due to this seals in on either SI or SG Hi-Hi level.

B. Incorrect. EWI valves shut and MEPTs trip. Trip breakers are required to be cycled to reset FWI due to this seals in on either SI or SG Hi-Hi level.

C. Correct. FWI valves shut and MFPTs trip. Trip breakers are required to be cycled to reset FWI due to this seals in on either SI or SG Hi-Hi level.

D. Incorrect. FWI valves shut and MFPTs trip. Trip breakers are required to be cycled to reset FWI due to this seals in on either SI or SG Hi-Hi level.

REFERENCES V-LO-PP-28103 Slides 137-1 39 VEGP learning obiectives:

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

Page 86 of 208

C 0 Q What is the function of the FWI signal generated by P-14?

Prevents water intrusion into the steam lines (main turbine and MFP turbines)

Trips the Main Turbine and the Main Feed Pump Turbines in addition to the isolation of the Feed Water Valves V-LO-PP-281 03-6.2 137

C Q Lo Tavg / P-A FWI Reset

Place both FWI Isolation switches to reset position.

V-LO-PP-281 03-6.2 138

P-14 or SI FWI Reset

Actuation signal has to be cleared

The P-4 seal-in is removed by cycling the reactor trip breakers.

Reset FWI by placing both FWI Isolation switches to reset position.

V-LO-PP-281 03-6.2 139

HL-16 NRC Written Examination KEY

43. 060AK2.0 1 001 / 1/2/GASEOUS RELEASE/2.6/2.9 MEMINEWIRO/SRO/NRC/GCW During core offload, a dropped assembly results in a gas bubble rising from the cavity into Containment atmosphere.

- RE-2562C, Containment Air Radiogas goes into HIGH alarm.

- RE-002 and RE-003, Containment Low Range monitors are

- in alarm.

1) How can the operators check the High and Alert setpoints for the Containment -

Low Range monitors, and

2) what indication would inform the operators if RE-002 and RE-003 have failed?

A 1) may be checked on the SRDC in Control Room

2) the amber Equipment Trouble light would be illuminated.

B. 1) may be checked on the SRDC in Control Room

2) the blue Bypass light would be illuminated.

C. 1) would have to check locally on the DPM

2) the amber Equipment Trouble light would be illuminated.

D. 1) would have to check locally on the DPM

2) the blue Bypass light would be illuminated.

Page 87 of 208

HL-16 NRC Written Examination KEY Feedback 060 Accidental Gaseous Radwaste Release Knowledge of the interrelations between the Accidental Gaseous Radwaste Release and the following:

(CFR 41.7/45.7)

AK2.01 ARM system, including the normal radiation-level K/A MATCH ANALYSIS The question presents a plausible scenario where a dropped irradiated results in an accidental gaseous release in containment with conflicting radiation monitor indications.

The candidate has to determine where he can check the high and alert level setpoints for the containment area low range monitors and indications of possible trouble with the rad monitors.

ANSWER / DISTRACTOR ANALYSIS A. Correct. The setpoints can be checked on the SRDC in control room. The amber equipment trouble light would be lit if the rad monitor has failed.

B. Incorrect. The setpoints can be checked on the SRDC in control room. The amber equipment trouble light would be lit not the blue bypass light. Bypass is plausible that the monitor has bypassed itself on trouble but requires a local action by chemistry.

C. Incorrect. The setpoints can be checked locally but do no have to be checked locally. The amber equipment trouble light would be lit if the rad monitor has failed.

D. Incorrect. The setpoints can be checked locally but do no have to be checked locally. The amber equipment trouble light would be lit not the blue bypass light.

Bypass is plausible that the monitor has bypassed itself on trouble but requires a local action by chemistry.

REFERENCES ARP-1 7102-C for the Safety Related Display Console QRM2 13508-1, Radiation Monitoring Systems VEGP learning objectives:

LO-PP-321 01-06, State the purpose of the safety-related display console.

LO-PP-321 01-07, Explain the meaning of each of the five color-coded lamps on the date processing modules.

Page 88 of 208

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Approved By . . Procedure Number Rev S. E. Prewitt Vogtle Electric Generating Plant Date Approved 13508-1 20.1 Page Number 1/29/2010 RADIATION MONITORING SYSTEMS 15 of 32 INITIALS 4.2.3 SRDC Controls and Indicators Cohtrols and Indicators differ between channels, and may include the following:

4.2.3.1 Colored press-to-test lights that, WHEN lit, indicate:

a. Trouble (Amber)

Equipment trouble.

b Bypass (Blue)

DPM is in local control mode

c. Alert (Yellow)

Radiation level at detector has exceeded alert set point.

d HIGH (Red)

High radiation level at detector $

e AC Power Available (Green)

Power (1 20V AC) Is being supplied to the unit. (Should be lit for normal operation.)

4.2.3.2 Amber LED which indicate, WHEN lit:

a. PURGE The monitor is isolated from the process sample line and is being purged (purge valve is open).

b. SHUTDOWN The monitor is isolated from the process sample line with both inlet and outlet valves closed.

Printed January 18, 2011 at 9:31

Types of DPMs

Safety Related DPM Same size as expanded DPM

Secured with clamps around edges of door

Has a Readout and Control box(RCB) on inside of front door. Can change setpoints in local control using remote/bypass key switch

- RCB indicates current rad level and setpoints

Communications with Corn Console is isolated from DPM. Sent directly to the SRDC in Control Room

Seismic DPM Special non-safety unit assembled in same reinforced enclosure as SR DPM.

V-LO-PP-32101-O1 Rev-02.1 40

HL-16 NRC Written Examination KEY

44. 061 K2.0 1 001/2/1 /AFW/3 .2/3.3 MEM/LOIT BANK/RO/SRO/NRC/GCW The UO notices the following indications for AFW valves:

- HV-5106, TDAFW Steam Admission Valve, red and green lights are out.

- FV-5132 & 5134, MDAFW SG 2 & 3 Discharge Throttle Valves, red, green, and white lights are out.

Which of the following is the correct interpretation of these conditions?

A. Loss of C train 1 E 480 VAC and B train 1 E 125 VDC power.

B. Loss of C train 1 E 125 VDC and B train 1 E 125 VDC power.

C. Loss of C train 1 E 120 VAC and B train 1 E 480 VAC power.

D Loss of C train 1 E 125 VDC and B train 1 E 480 VAC power.

Feedback 061 Auxiliary I Emergency Feedwater (AFW) System K2 Knowledge of bus power supplies to the following:

(CFR: 41.7)

K2.01 AFW system MOVs K/A MATCH ANALYSIS AFW is divided between three Trains of Safety Related power. Valves listed in stem are from two of the three trains. Knowledge of loss of power is applicable to determine what equipment would still be available.

ANSWER I DISTRACTOR ANALYSIS A. INCORRECT-HV-5106, TDAFW Pump Steam Admission Valve is powered from C Train 125 VDC power 1(2)CD1. FV-5132, 5134 (B Train pump discharge valves) are 480 v AC 1 E power from B Train 1 (2)BBB.

B. INCORRECT-i (2)CD1. FV-5132, 5134 (B Train pump discharge valves) are 480 v AC 1 E power from B Train 1 (2)BBB. No AFW valves are powered from 125 vdc, all are 480 vac.

C. INCORRECT-HV-5106, TDAFW Pump Steam Admission Valve is powered from C Train 125 VDC power 1 (2)CD1. Only instrumentation is powered from 120 vac and not valves.

Page 89 of 208

HL-16 NRC Written Examination KEY D. CORRECT REFERENCES 18034-1, Loss of Class 1 E 1 25V DC Power, Attachment Cl, Loss of 1 25V DC to Bus 1CD1 (Rev 10)

V-LO-PP-20101 AFW System Slide 18 Drawing 1X3D-AA-F17A VEGP learning objectives:

LO-PP-201 01-09 Determine the impact to AFW system operation and the overall integrated plant operations to the following types of power supply failures:

a. U/V condition on either AAO2 or BAO3 with the bus being re-energized from the EDG while at 100% power.

b. UN condition on either AAO2 or BAO3 with the bus remaining de-energized while at 100% power.

c. Loss of a 120 VAC 1 E vital instrument bus.

d. Loss of a 125 VDC 1 E bus

e. Loss of All AC Power Page 90 of 208

Approved By Procedure Number Rev C. S. Waidrup Vogtle Electric Generating Plant 18034-1 11 Page Number LOSS OF CLASS 1 E 1 25V DC POWER 84 ATTACHMENT Cl Sheet 1 of 3 LOSS OF 125V DC TO BUS 1CD1 EQUIPMENT RESPONSE DUE TO LOSS OF TRAIN C 125V DC POWER

Power to Inverter 1CD1I3 is lost causing 120V AC Vital Bus 1CY1A to de-energize.

TDAFW Pump Mechanical Trip and Throttle Valve 1-PV-1 5129 will fail as is with no control capability.

TDAFW Pump Speed Governor Valve 1-SV-15133 will fail full open.

Power to Isolation Panel 1 CCQIP3 is lost rendering the annunciators in Train C inoperable.

RHR PMP A UPSTREAM SUCTION FROM HOT LEG 1-HV-8701B.

Loss of RCP 3 Class 1 E breaker control including RCP underfrequency trip and undervoltage input to SSPS.

Turbine Driven AFW PMP SUCTION HDR l-HV-51 13.

Turbifl Drivén AFW PMP STEAM ADMISSION VALVE t-HSi6

Turbine Driven AFW PMP DISCHARGE HDR 1-HV-5127.

Turbine Driven AFW PMP DISCHARGE HDR 1-HV-5125.

Turbine Driven AFW PMP DISCHARGE HDR 1-HV-5122.

Turbine Driven AFW PMP DISCHARGE HDR 1-HV-5120.

Printed January 18, 2011 at 09:37

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HL-16 NRC Written Examination KEY

45. 061 K6.02 001/2/1 /AFW PUMPS-LOSS/2.6/2.7 C/AJNEW/RO/SRO/NRC/TLH While the TDAFW pump was tagged out for governor replacement, a spurious Reactor Trip from 100% occurred.

The Unit Operator monitoring AFW flow reports the following:

- Steam Generator #1 = 0 gpm.

- Steam Generator #2 = 300 gpm.

- Steam Generator #3 300 gpm.

- Steam Generator #4 = 0 gpm.

Which ONE of the following is CORRECT concerning the MDAFW pumps status and their associated mini-flow valve position?

MDAFW Pump Status Mini-flow Valve Status A. MDAFW pump train A RUNNING, OPEN MDAFW pump train B STOPPED, CLOSED B. MDAFW pump train A STOPPED, CLOSED MDAFW pump train B RUNNING, OPEN C. MDAFW pump train A RUNNING, CLOSED MDAFW pump train B STOPPED, OPEN D MDAFW pump train A STOPPED, OPEN MDAFW pump train B RUNNING, CLOSED Feedback 061 Auxiliary I Emergency Feedwater (AFW) System Knowledge of the effect of a loss or malfunction of the following will have on the AFW components:

(C FR: 41.7/45.7)

K6.02 Pumps K/A MATCH ANALYSIS Page 91 of 208

HL-16 NRC Written Examination KEY This question test the candidates knowledge of which MDAFW pumps supply what S/G and the affect of the loss of the MDAFW pump will have on the status of their mini-flow valves (AFW corn ponents).

ANSWER I DISTRACTOR ANALYSIS A. Incorrect MDAFW pump A is not running which is evident by no flow indicated to its associated SIG5. MDAFW pump B supplies S/Gs 2 and 3 only. This answer is a plausible distractor if the student was under the mindset of MDAFW pump A supplies the North S/Gs (2 and 3) and the MDAFW pump B supplies the south SIGs (1 and 4).

B. Incorrect The MDAFW pumps status are correct but the mini-flow status are incorrect. The status of the mini-flow valves are plausible if the candidate thought the valves were interlocked with the pump breakers.

C. Incorrect MDAFW pump A is not running which is evident by no flow indicated to its associated S/Gs. MDAFW pump B supplies SIGs 2 and 3 only and the mini-flow valves status are incorrect also. This answer is a plausible distractor if the student was under the mindset of MDAFW pump A supplies the North SIGs (2 and 3) and the MDAFW pump B supplies the south S/Gs (1 and 4). The status of the mini-flow valves are incorrect but plausible because if MDAFW pump A total flow is greater than 150 gpm its mini-flow valve would be closed.

D. Correct MDAFW pump B supplies SIGs 2 and 3 also its mini-flow valve closes on high flow, MDAFW pump A mini-flow will be open because its total flow is less than 150 gpm.

REFERENCES 13610-1/2 Auxiliary Feed Water System section 4.2.

Drawing 1X4DB161-2 AFW V-LO-PP-20101 Slides 15 and 20 VEGP learning objectives:

Not applicable Page 92 of 208

Approved By S. E. Prewitt Date Approved i

I Vogtle Electric Generating Plant A Procedure Number Rev 13610-1 48.1 Page Number 8/18/2010 j AUXILIARY FEEDWATER SYSTEM 16 of 98 INITIALS 4.2 OPERATION 4.2.1 Feeding Steam Generators from the Auxiliary Feedwater Syst em NOTES

. The Steam Generators may be gravity filled using 13601-1 Main Steam System section for raising Steam Generator level.

. If in Modes 4, 5, or 6 with the A and B AFW systems cross-tied, either MDAFW pump may be used to provide fill to all 4 steam genera tors, as long as total flow does not exceed 600 gpm.

Cautions

. When feeding Steam Generators, the flow rate to a Steam Genera tor should be maintained greater than 30 gpm to avoid stratification in the feedwater piping. Also during heatup, cooldown, and hot standb y

operation, when feedwater requirements are small, flow should be maintained at consistent lower fill rates rather than larger interm ittent ones.

. When feeding forward below 150 gpm, frequent checks of mini-f low vatvØ position must be performed to ensure corre Operatlorn 4.2.1.1 To supply water from the Motor-Driven Auxiliary. Feedwate Pumps, perform the follow1ng

a. Start the Motor-Driven Auxiliary Feedwater Pump(s) for feeding the applicable steam generator(s) per Section 4.1.3:

To feed Steam Generators 1 and/or 4, start MDAFW Pump A..

To feed Steam Generators 2 and/or 3, start MDAFW Pump B IF in Modes 4, 5, OR 6 with the A and B AFW systems cross--tied, start the required MDAFW pump to feed all 4 steam generators.

Printed January 12, 2011 at 12:26

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Pump Miniflow Protection VI Allows limited pump operation under low flow conditions without excessive heating iMotor pump miniflows

Motor-operated miniflow control valve

Senses pump discharge flow

Auto-closes once main flowpath flow is established Open at 280 gpm Close at 330 gpm

Orifice in line limits flow to approximately 175 gpm per pump

Must be aligned to the same CST providing suction to the MDAFW Pump.

V-LO-PP-20101 Rev-03.2 20

HL-16 NRC Written Examination KEY

46. 062K4.02 001/2/I /AC-BREAKER TRIPS/2.512.7 MEM/MOD BANK/RO/SRO/NRC/TLH The Maintenance switch located on the QEAB for a 4160V AC bus has been placed in the MAINT position to support bus inspection.

Regarding the bus supply breaker(s), when the switch is returned to the NORMAL position, this switch will...

A. enable the instaneous overcurrent trips.

B. bypass the instaneous overcurrent trips.

C enable the time delay on the overcurrent trips.

D. bypass the time delay on the overcurrent trips.

Page 93 of 208

HL-16 NRC Written Examination KEY Feedback 062 AC Electrical Distribution System Knowledge of ac distribution system design feature(s) and/or interlock(s) which provide for the following:

(CFR: 41.7)

K4.02 Circuit breaker automatic trips K/A MATCH ANALYSIS Question specifically asks the purpose of the Maintenance I Normal switch on a 41 60V electrical switchgear which is a design feature that provides circuit breaker trips.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect instanteous overcurrent trips are always enabled. Plausible because taken the switch to the maintenance position makes the timed delay trip an instantenous overcurrent trip.

B. Incorrect the instantenous overcurrent trips are never bypassed by this switch.

Plausible because taken the switch to the maintenance position bypasses the timed delay which makes it instantenous overcurrent trip too.

C. Correct this switch when positioned to normal reinstates the timed delay portion of the overcurrent trip.

D. Incorrect but plausible because this is what happens when the switch is placed in the maintenance position.

REFERENCES 13435-C Circuit Breaker Racking Procedure.

062K4.02-1 from LOIT bank (HL-15 NRC exam)

VEGP learninci obiectives:

V-LO-PP-01 101-03: Describe the need for the Maintenance / Normal switch on the 13.8 KV and 4.16 KV buses.

Page 94 of 208

Approved By . Procedure Number Rev S. E. Prewitt Vogtle Electric Generating Plant 13435-C 39 Date Approved Page Number 1/ii/2010 CIRCUIT BREAKER RACKING PROCEDURE 8 of 140 INITIALS 4.1.1 Removal from CONNECT to DISCONNECT or REMOVED Position (1 3.8KV/4.1 6KV Bkr) 4.1.1.1 Prerequisite 3.3 has been reviewed AND Attachment 3 of NMP-SH-003 has been completed.

Critical 4.1.1.2 Identify the applicable breaker to be racked:

(breaker number)

Cv 4.1.1.3 Check breaker OPEN light and the mechanical breaker position indicator indicates OPEN; IF NOT, notify SS.

NOTE 4.1.1.4 PRIOR to racking any breaker in the following switchgear, place the Maintenance Control Switch on QEAB to MAINT:

13.8kV Switchgear 1 (2)NAA BUS-i (2)NAA 13.8kV Switchgear 1 (2)NAB BUS-i (2)NAB 4.16kV Switchgear i(2)NAO1:

IF in Normal Alignment BUS-i (2)NAO1 jf being fed from ANAO2 BUS-ANAO2 4.16kV Switchgear i(2)AAO2 BUS-i (2)AAO2 4.16kV Switchgear i(2)BAO3 BUS-i (2)BAO3 4.16kV Switchgear i(2)NAO4:

jf in Normal Alignment BUS-i(2)NAO4 if being fed from ANAO3 BUS-ANAO3 Printed January 12, 2011 at 13:18

Page 1 of I

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1. 062K4.02 001/2/1/AC ELECTICAL DISTRJMEM - 2.5 I 2.7/BANK LZITNTfDS

/3y,4c7 The Maintenance / Normal switch located on the QEAB for a 41 60V AC bus has been placed in the Maintenance position to support bus inspection.

When placed in the Maintenance position, this switch...

A bypasses the time delay on the overcurrent trips on the bus supply breakers.

B. bypasses the time delay on the overcurrent trips on the bus load breakers.

C. bypasses the instantaneous overcurrent trips on the bus supply breakers.

D. bypasses the instantaneous overcurrent trips on the bus load breakers.

Page: 1 of 2 12/13/2010

Feedback K/A 062 A.C. Electrical Distribution System K4.02 Knowledge of A.C. Distribution System design feature(s) and/or interlock(s) which provide for the following:

Circuit breaker automatic trips K/A MATCH ANALYSIS Question specifically asks the purpose of the Maintenance / Normal switch on a 41 60V electrical switchgear.

ANSWER I DISTRACTOR ANALYSIS A. Correct.

B. Incorrect. Bypasses time delay (to instantaneous) is correct. Supply breaker is the one that trips to protect personnel. Load is where problem detected, tripping open the supply protects personnel.

C. Incorrect. Bypasses time delay (to instantaneous) is correct. Supply breaker is the one that trips to protect personnel. Load is where problem detected, tripping open the supply protects personnel.

D. Incorrect. Bypasses time delay (to instantaneous) is correct. Supply breaker is the one that trips to protect personnel. Load is where problem detected, tripping open the supply protects personnel.

REFERENCES 13435-C, Circuit Breaker Racking Procedure, NOTE prior to step #4.1.1.2 VEGP learning obiectives:

LO-LP-1 1101-03 Describe the need for the maintenance/normal switch on the 13.8kV/4160V kV buses.

Notes Response form Page: 2 of 2 12113/2010

HL-16 NRC Written Examination KEY

47. 063A2.0 1 001/2/1 /DC GROUNDS/2.5/3 .2 C/AJBANKJRO/SRO/NRC/GCW While at 100% power, the following annunciator illuminates:

- ALB34, window BOl for 125V DC SWGR BD1 TROUBLE

- The Control Building Operator (CBO) has been dispatched to investigate.

Which ONE of the following is CORRECT regarding indications / actions to take if a GROUND has occurred?

A. There are no local bus ground detection targets associated with BD1, the CBO would have no indication of a ground, maintenance would determine the cause.

B. A bus ground detection target would be dropped on BD1, de-energize panels BD1 M, BD1 1, and BD12 one at a time to locate the panel with the ground.

C. There are no bus ground detection targets associated with BD1, de-energize all loads on BD1 M, BD1 1, and BD12 one at a time until the alarm clears.

D A bus ground detection target would be dropped on BD1, de-energize selected loads on BD1 M, BD1 1, and BD12 one at a time to locate the source of the ground.

Page 95 of 208

HL-16 NRC Written Examination KEY Feedback 063 D.C. Electrical Distribution Ability to (a) predict the impacts of the following malfunctions or operations on the DC electrical systems; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

(CFR: 41.5/43.5/45.3/45.13)

A2.01 Grounds K/A MATCH ANALYSIS Question asks plausible scenario where a ground has appeared on a 125V DC electrical bus. The operator must choose which of the indications would be available to determine the source of the ground and corrective action to take.Ground fault indication is available on BD1 but not the panels fed from BD1. The panels fed from BD1 should have loads selectively de-energized one at a time to determine the source of the ground. Not all loads can be de-energized. MSIVs and MFIVs in particular are powered from BD11 / BD12 and would result in a plant trip if de-energized.

ANSWER / DISTRACTOR ANALYSIS A. Inorrect. There is ground fault indication available on BD1. Even though maintenance may have to be eventually called, selectively de-energizing loads on BD1 and buses fed from BD1 may find the ground.

B. Incorrect. Ground fault indication is available for BD1. However, de-energzing the BD11 or BD12 panel would result in a reactor trip due to MSIVs / MEl Vs shutting when their solenoids de-energized.

C. Incorrect. There is ground fault indication available for BD1. However, ALL loads on BD11 or BD12 cannot be de-energized without resulting in a reactor trip. See B above.

D. Correct. Ground fault indication is available for BD1 and BD1 M, BD1 1, and BD1 2 should have loads selectively de-enrgized to try to find the ground.

REFERENCES ARP-1 7034, window BOl for 125V DC SWGR 1BD1 TROUBLE Previously given on HL-1 5 Question # 45 VEGP learning obiectives:

Not applicable.

Page 96 of 208

Approved By Procedure Number Rev S. E. Prewitt Vogtle Electric Generating Plant 17034-1 23 Page Number ANNUNCIATOR RESPONSE PROCEDURE FOR ALB 34 ON EAB PANEL C. 92 WINDOW BOl ORIGIN SETPOINT 125V DC SWGR 1 BD1 Prot RIy Not Applicable 1 BD1 TROUBLE 1.0 PROBABLE CAUSE

1. One of the breakers on Switchgear 1 BD1 tripped.

2.: Bus ground fault.

3. Bus undervoltage or loss of voltage.

4. Loss of control power.

2.0 AUTOMATIC ACTIONS NONE 3.0 INITIAL OPERATOR ACTIONS NONE 4.0 SUBSEQUENT OPERATOR ACTIONS

1. Check for associated alarms and indications.

2. Wnecessary, dispatch: an operator to the switchgearto check for.

a. GrOund faultindlcatiOns.

b. Loss of control power, (Battery Charger shutdown with corresponding switchgear breaker open, or control power breaker open)

c. Tripped breakers.

d. Bus undervoltage or loss of voltage.

e. Other abnormal conditions.

Printed January 18, 2011 at 10:23

Approved By I Procedure Number Rev S. E. Prewitt I Vogtle Electric Generating Plant 117034..1 23 Date Approved I ANNUNCIATOR RESPONSE PROCEDURE FOR ALB 34 ON EAB PANEL I Page Number 0-5-2009 I 27 of 92 WINDOW BOl (Continued)

3. alarm is due to a breaker tripping on fault:

a. Determine what loads are affected.

b. If necessary, dispatch an operator to the affected switchgears to manually operate the breakers, under the direction of Control Room.

c. Notify Maintenance and return affected equipment to service once the cause has been corrected.

NOTE The DC distribution panels do not have ground detection relays.

4. if. a bus ground protection alarm is indicated, selectively de-energiza components on 1 BD1, 1BD1 M, 1 BD1 1, and 1 BDI 2 to locate the ground:

a. Using 120V AC/125V DC Panel Load Data Base 1X3D-AA-MO1C, determine which circuits can be momentarily de-energized by evaluating effect on plant systems.

b. Momentarily de-energize selected circuits to locate the source of the ground.

c. Initiate maintenance to clear the ground.

5. f the alarm is due to an undervoltage condition or loss of voltage:

a. Return battery and/or charger to service if possible.

b. IF the chargers CANNOT be returned to service, monitor battery voltage and selectively strip the bus loads.

c. jf battery 1 BD1 B voltage drops to 109.7, request Maintenance to perform 28912-C, Class 1E Quarterly Battery Inspection And Maintenance.

Printed January 18, 2011 at 10:23

Approved By Procedure Number Rev S. E. Prewitt Vogtle Electric Generating Plant 17034-1 23 Do Page Number ANNUNCIATOR RESPONSE PROCEDURE FOR ALB 34 ON EAB PANEL 92 WINDOW C03 ORIGIN SETPOINT 125V DC PNL 1BD12 Prot RIy Not Applicable 1BD12 TROUBLE 1.0 PROBABLE CAUSE

1. One of the breakers on Panel 1 BD1 2 tripped.

2. Panel undervoltage or loss of voltage.

2.0 AUTOMATIC ACTIONS NONE 3.0 INiTIAL OPERATOR ACTIONS NONE 4.0 SUBSEQUENT OPERATOR ACTIONS

1. Check for associated alarms and indications.

2. If necessary, dispatch an operator to Panel IBDI2 to check for:

a. Undervoltage indication.

b. Tripped breakers.

c. Other abnormal conditions.

3. IF alarm is due to a breaker tripping on fault or loss of voltage:

a. Determine what loads are affected.

b. If necessary, dispatch an operator to the affected switchgears or panels to operate the breakers manually, under the direction of Control Room personnel.

c. Notify Maintenance and return affected equipment to service once the cause has been corrected.

Printed January 18, 2011 at 10:26

Approved By Procedure Number Rev S. E. Prewitt Vogue Electric Generating Plant 17034-1 23 Date Approved Page Number ANNUNCIATOR RESPONSE PROCEDURE FOR ALB 34 ON EAB PANEL 31 of 92 10-5-2009 WINDOW 603 SETPOINT ORIGIN 125V DC PNL 1 BD1 1 Prot Rly Not Applicable 1 BD1 1 TROUBLE 1.0 PROBABLE CAUSE

1. One of the breakers on Panel 1 BD1 1 tripped.

2. Panel undervoitage or loss of voltage.

2.0 AUTOMATIC ACTIONS NONE 3.0 INITIAL OPERATOR ACTIONS NONE 4.0 SUBSEQUENT OPERATOR ACTIONS

1. Check for associated alarms and indications.

2. If necessary, dispatch an operator to Panel IBDI1 to check for:

a. Undervoltage indication.

b. Tripped breakers.

c. Other abnormal conditions.

3. if alarm is due to a breaker tripping on fault or loss of voltage:

a. Determine what loads are affected.

b. If necessary, dispatch an operator to the affected switchgears or panels to operate the breakers manually under the direction of Control Room personnel.

c. Notify Maintenance and return affected equipment to service once the cause has been corrected.

Printed January 18, 2011 at 10:25

HL-16 NRC Written ExaFARL mination KEY EY 201 OIRO/SRO/NRC/GCW

48. 064G2.2. 12001/2/1 IDG SURVEILLANCES/3 .7/4.1 C/A/BANK prior to closing the DG1 B is being paralleled with the grid for surveillance testing, and on.

output breaker, the synchroscope is turning fast in the FAST directi ONE of the Per OSP-14980B-1, Diesel Generator lB Operability Test, which following should be:

41 60V bus)

1) the component with the highest frequency (DG1 B output or 1 BAO3 and closing the

2) which push button must be depressed to adjust frequency prior to output breaker?

(1) (2)

A DG1B frequency Speed B. DG1B frequency Voltage C. 1 BAO3 bus frequency Speed D. 1 BAO3 bus frequency Voltage Feedback 064 Emergency Diesel Generators (EDIG)

Equipment Control G2.2.12 Knowledge of surveillance procedures.

(CFR: 41.10145.13)

KIA MATCH ANALYSIS the surveillance Question presents a scenario where DG1 B is being paralled per 1BAO3 should procedure. The candidate must determine whether DG1B or 4160 bus push button have the highest frequency before closing the output breaker and which r.

should be used to adjust frequency prior to closing the output breake ANSWER I DISTRACTOR ANALYSIS is used to adjust A. Correct. DG1 B frequency should be higher and the speed button frequency prior to closing the output breaker.

button, not the B. Incorrect. First part DG1 B is correct, 2nd part should be speed voltage pushbutton.

Page 97of 208

HL-16 NRC Written Examination KEY button is C. Incorrect. 1 BAO3 at higher frequency is incorrect. 2nd part about speed correct.

button is D. Incorrect. 1 BAO3 at higher frequency is incorrect. 2nd part about speed also incorrect.

REFERENCES 1 4980B-1, Diesel Generator 1 B Operability Test Farley 2010 NRC RO Exam question # 45 VEGP learning objectives:

and/or the LO-PP-1 1101-49 For the following controls on the Generator Control Panel on of each position:

QEAB, describe the response of the diesel generator to the selecti

a. Local/Remote switch

b. Speed RAISE/LOWER switch (pushbutton)

c. Exciter Enable pushbutton

d. Emergency Shutdown pushbutton

e. Delete

f. Field Flash pushbutton

g. Voltage Control RAISE/LOWER switch (pushbutton)

h. Unit/Parallel switch

i. Exciter Permissive Switch

j. Load Pot Page 98 of 208

Approved By . Procedure Number Rev J.B. Stanley Vogtle Electric Generating Plant 14980B-1 23.3 Date Approved Page Number

,.5/8/09 DIESEL GENERATOR 1 B OPERABILITY TEST 28 of 75 INITIALS NOTE A Synchroscope Meter indication of 12 oclock may indicate that another breaker synchronization switch is ON.

CAUTION Never place two sync-switches to the ON position at the same time. A blown PT fuse may result.

5.1 .26 Place the BRKR 1 BAO31 9 Synchronizing Switch to ON.

5.1.27 Set 1BAO3 4160V bus phase voltage to the highest value on the QEAB Voltmeter by moving the BUS 1 BAO3 NORM INCM VM SW through all positions.

5.1.28 Verify 1 BAO3 Bus Voltage is between 4025V and 4250V prior to paralleling DG.

5.1.29 i.E 1 BAO3 Bus Voltage is NOT between 4025V and 4250V, contact Northern Transmission Control Center, (8-506-6965) to establish these conditions 5.1.30 Set DG1 B voltage to lowest value on QEAB Voltmeter by movina the DSL GEN 1 B VM SW through all positions.

5.1.31 Adjust DG speed using the DSL Gen 1 B Speed Control Raise or Lower pushbuttons, until the Synchroscope needle is rotating 8 to 10 seconds per rotation in clockwise (fast) direction.

5.1.32 Verify that synchronizing lights are bright at the 6 oclock position and dark at the 12 oclock position and that the AUTO SYNC PERMISSIVE red light comes on near the 12 oclock position.

5.1.33 Adjust generator voltage to approximately 50V above the highest phase of 1 BAO3 bus voltage using DSL Gen 1 B Voltage Control Raise or Lower pushbuttons.

5.1.34 Adjust DSL GEN lB LOADING SET PT CONTROL, 1SE-4916 to 1.00(10% DIG LOAD).

Printed January 18, 2011 at 10:30

The 1 B DG is being paralleled with the grid for surveillanLe testing, and prior to closing the 1 B DG output breaker, the Synchroscope is turning fast in the FAST direction.

Which one of the following states:

1) the component with the highest frequency (the 1 B DG output or 1 G 41 60V Bus),

and

2) which switch must be turned to adjust frequency prior to closing the output breaker?

(1) (2)

A. 1G Bus Frequency VOLTAGE ADJUST VOLTS/MVARS B. 1G Bus Frequency GOVERNOR MOTOR SPEED/MW C. 1 B DG Frequency VOLTAGE ADJUST VOLTS/MVARS D lB DG Frequency GOVERNOR MOTOR SPEED/MW Page: 39 of 88 1/3/2011

HL-16 NRC Written Ex amination KEY EW/RO/SRO/NRC/TLH

49. 064K3.03 OO1/211/DG-MANUAL LOADS/1613.9 C/A/N e at 100% RTP.

Unit 1 has experienced a loss of power to 1AAO2 whil ed on Overspeed when the D/G-lA automatically started but immediately tripp CCW pumps started.

the D/G-lA, the Control Room While the investigation and repairs are in progress on Crew was able to re-energize 1 AAO2 from the SAT encer (1) The reason When 1AAO2 was re-energized from the SAT, the Sequ is to (2) the Sequencer Reset must be performed in this case A. (1) will start its UV loads.

pment powered by 1AAO2.

(2) allow automatic and manual operation of equi ing required).

B (1) will NOT start its UV loads. (Manual load nt powered by 1AAO2.

(2) allow automatic and manual operation of equipme C. (1) will start its UV loads.

(2) reset the Loss of Offsite Power (LOP) Monitor.

ing required).

D. (1) will NOT start its UV loads. (Manual load (2) reset the Loss of Offsite Power (LOP) Monitor.

Feedback 064 Emergency Diesel Generators (EDIG) of the EDIG system will have Knowledge of the effect that a loss or malfunction on the following:

(C FR: 41.7/45.6)

K3.03 ED/G (manual loads)

K/A MATCH ANALYSIS of the 4160 VAC bus will be Question matches the K/A because manual loading source other than its associated required when the IE bus is repowered by another Diesel after a UV.

ANSWER I DISTRACTOR ANALYSIS Page 99 of 208

HL-16 NRC Written Examination KEY A. Incorrect (1) The UV sequence will not occur because the sequencer is waiting for D/G-lA output breaker to close before the UV sequence to occur. (2) Incorrect but plausiable because this is the second function of the sequencer reset, but since the Diesel Generator Output breaker closed and sequenced started the auto sequencer reset after 60 seconds occurred. Distractor is plausible because it is reasonable to believe that the sequencer would start its sequence when the UV condition cleared and the second part of the answer is true if the output breaker never closed.

B. Correct (1) The UV sequence will not occur because the sequencer is waiting for D/G-lA output breaker to close before the UV sequence to occur. (2) Since the sequencer is locked up, one of the reason for the sequencer reset is to allow for auto starts of equipment and allow the operators to stop equipment that had been manually loaded.

C. Incorrect- (1) The UV sequence will not occur because the sequencer is waiting for D/G-lA output breaker to close before the UV sequence to occur. (2) Plausible because the LOP (Loss of Power) indicator is would be lit and the option is given to reset with permission from the SS.

D. Incorrect (1) The UV sequence will not occur because the sequencer is waiting for DiG-i A output breaker to close before the UV sequence to occur is correct. (2)

Plausible because the LOP (Loss of Power) indicator is would be lit and the option is given to reset with permission from the SS.

REFERENCES 1 3427A-1 41 60V AC Bus 1 AAO2 1 E Electrical Distribution System Rev 6.2 Step 2.1.6 and 4.4.1.21.

LO-TX-28201 Safety Features Sequencer System (Page 22-23)

VEGP learning obiectives:

LO-PP-28201 -04 Describe the function of the Sequencer Reset pushbutton following:

a. A loss of offsite power with the failure of the D/G breaker to close.

b. D/G breaker trip during load sequencing and subsequent reenergization from the D/G.

Page 100 of 208

Approved By Procedure Number Rev Vogtle Electric Generating Plant J. B. Stanley Date Approved I 41 60V AC BUS 1 AAO2 1 E ELECTRICAL DISTRIBUTION SYSTEM I

13427A-1 Page Number 6.2 10/22/09 4 of 60 2.1.5 When the SAT is connected to both 1 E 41 60V AC Buses, there exists a potential breaker coordination problem that could result in the 41 60V bus incoming breaker not detecting a fault and the SAT breaker opening to clear the fault.

This would result in an LOSP. If an LOSP occurs with the SAT connected to both 1 E 41 60V Buses, consideration of this scenario should be addressed in recovery actions.

2.1.6 2.1.7 Prior to removing 1AAO2 from service, the required boration flowpath and shutdown cooling should be reviewed for impact.

Printed January 7, 2011 at 15:06

Approved By Procedure Number Pev J. B. Stanley Vogtle Electric Generating Plant 13427A-1 6.2 Date Approved Page Number 4160V AC BUS 1AAO2 1E ELECTRICAL DISTRIBUTION SYSTEM 39 of 60

10/22109 INITIALS 4.4.1 .15 Check Train A 480V AC Motor Control Centers energized by checking j4Q MCC TROUBLE alarms on QEAB Annunciator Panel ALB36.

a. jf any Train A 480V AC 1 E Motor Control Center NOT energized, obtain SM permission and energize per 13429-1, 480V AC 1 E Electrical Distribution System.

b. E anyjrain A 480V AC Non 1 E Motor Control Center NOT energized, obtain SM permission and energize per 13430-1, 480V AC Non 1 E Electrical Distribution System.

4.4.1 .16 Notify SM that loads listed in Table 1 may be restored.

4.4.1 .17 At 1 AAO2, reset relay targets that are dropped.

4.4.1.18 Generate caution tags to alert personnel that Bus 1AAO2 is energized from its alternate incoming source.

4.4.1 .19 Perform the following steps only if in MODE 5 OR 6, OTHERWISE, exit this procedure.

NOTE Keys to the Sequencer door may be obtained from C&T and are identified as 1OP3-41 or 1OP3-42.

4.4.1.20 IF required, energize Sequencer A per 13540A-1,Safety Features Sequencer System.

4.4.1.21 IE DG Output Breaker did NOT close on loss of power, reset the Sequencer by performing the following at the Sequencer Test Panel:

a. On the NORMAL OPERATION OVERVIEW screen, depress the SEQUENCER RESET button.

b. Check UN red indicator light is not lit.

c. With the SS concurrence, reset the LOP counter using the LOP RESET button, if required.

d. Reset alarms as necessary. ;*.

Printed January 7, 2011 at 15:05

ELAPSED TIME j7 SECONDS STEPS ANNUNCIATION UNDERVOLTAGE SEQUENCER LOGIC FAILURE SAFETY EQUIPMENT FAIL TO START SEQUENCER RESET ALARM ACKNOWLEDGE SEQUENCER NORMAL OPERATION OVERVIEW SCREEN FIGURE 3 Printed January 24, 2011 at 12:12

Sequencer U/V Reset Circuit From the preceding discussions of Sequencer operation, the importance of the DO output breaker closure on sequencer operation should be evident. On a U/V condition the sequencer is waiting for the DG output breaker closure with the Auto/Manual Block enabled to begin load sequencing. If power cannot be restored from the OG due to a problem with the DG or its breaker, but power is restored from an offsite source, the Sequencer will not sequence loads and manual stopping of loads on the bus is blocked. This is referred to as Sequencer lockup. A design change was implemented to install an auto reset circuit and U/V reset pushbutton to address this issue. The figure below describes the U/V reset logic.

UN SEQ COMPLETE UN SIGNAL DG BRKR CLOSED DG SEQUENCER UN RESET LOGIC There are two parts to this logic, the Auto reset and the U/V reset pushbutton enable. The requirements for the Auto reset circuit are:

U/V signal present DG breaker trip alarm relay actuated Sequencer loading sequence started DG output breaker open If conditions are present, a 60 sec timer is enabled. When the timer times out, a reset signal is sent to the U/V retentive memory. This will cause a momentary reset of the U/V signal which will reset the sequencer and process another load shed signal. This is necessary to remove any loads previously sequenced to prevent overloading the power 22 Revision 2.0

source by energizing the bus already partially loaded and prevent the cumulative starting currents from challenging overcurrent relay settings.

The requirements for enabling the U/V reset pushbutton are:

U/V Signal present DG output breaker open Auto U/V reset initiated If conditions are present, a 60 sec timer is enabled when the DG output breaker trips. When the timer times out, the sequencer is reset to Time

) thereby allowing a subsequent U/V signal to be processed.

Zero (T 0 60 seconds after the subsequent U/V, the U/V reset pushbutton illuminates.

Depressing the pushbutton will do the following:

Reset the Auto/Manual Block signal Reset the U/V signal retentive memory Reset the sequencer timing bus to T 0 (normal standby condition)

The sequencer can also be locked up due to LOP monitor receiving three U/V signals in a two hour period. This will require a manual reset on the LOP monitor card in the sequencer.

23 Revision 2.0

HL-16 NRC Written ExaRO/SR mination KEY O/NRC/GCW

50. 067AA1.07 001/1/2/FIRE PANEL RESETI2.913.O MEMINEW/

the alarm overview The Fire Alarm Computer (FAC) has a SOLID icon displayed on page.

and

1) The computer for primary Fire Alarm response location is_____

2) the SOLID icon means the message is_____

A 1) Main Control Room

2) acknowledged B. 1) C & T (EBI workstation)

2) acknowledged C. 1) Main Control Room

2) reset or cleared D. 1) C & T (EBI workstation)

2) reset or cleared Page 101 of 208

HL-16 NRC Written Examination KEY Feedback 067: Plant fire on site Ability to operate and I or monitor the following as they apply to the Plant Fire on Site:

(CFR 41.7! 45.5 I 45.6)

AA1 .07 Fire alarm reset panel K/A MATCH ANALYSIS The question asked the primary location of the Fire Alarm Computure (FAC) panel and the icon which symbolizes an outstanding reset or cleared message.

ANSWER I DISTRACTOR ANALYSIS A. Correct. Main Control Room is the primary location, solid icon means the message is acknowledged.

B. Incorrect. C & T is plausible since it is the backup location and at one time was the primary location. Hollow icon means the message is reset or cleared.

C. Inorrect. Main Control Room is the primary location, solid icon means the message is acknowledged.

D. Incorrect. C & T is plausible since it is the backup location and at one time was the primary location. Hollow icon means the message is reset or cleared.

REFERENCES ARP-1 71 03-B, Annunciator Response Procedure for Fire Alarm Computer VEGP learning objectives:

Not applicable.

Page 102 of 208

Approved By S.A. Phillips Date Approved i Vogtle Electric Generating Plant A ANNUNCIATOR RESPONSE PROCEDURES FOR FIRE ALARM Procedure Number Rev 171038-C 13 Page Number 0/22/08 COMPUTER 48 of 74 APPENDIX 1 (Continued)

FIRE ALARM COMPUTER (FAC)

OPERATING INSTRUCTIONS AND ATTACHMENTS The C: ie Main ry Fire alarm response locat4 The Q _}C&t i Fire alarm response locaj Printed January 18, 2011 at 10:35

HL-16 NRC Written Examination KEY 51.068A3.02 001 /2/2/LIQUID RW-ISOLATION/3 .6/3.6 C/A/LOIT BANKJRO/SRO/NRC/GCW Prior to liquid radwaste release, all associated equipment is verified to be operable and all normal sample and approval processes have been completed.

Waste Monitor Tank 009 is in the process of being released on Unit 1 when power to the DPM for RE-Ol 8 (Liquid Waste Monitor) is temporarily cycled due to an error made by an electrician.

Which ONE of the following CORRECTLY explains the affect on the liquid radwaste release?

A. Liquid Radwaste Isolation Valve RV-01 8 closes. RV-01 8 may be immediately re-opened to allow the release to continue.

B Liquid Radwaste Isolation Valve RV-018 closes. RV-018 may NOT be immediately re-opened to allow the release to continue.

C. Liquid Radwaste Isolation Valve RV-018 remains open. Immediately close the valve from the control room to stop the release.

D. Liquid Radwaste Isolation Valve RV-018 remains open. Immediately instruct an operator in the field to close the valve to stop the release.

Page 103 of 208

HL-16 NRC Written Examination KEY Feedback 068 Liquid Radwaste System (LRS)

Ability to monitor automatic operation of the Liquid Radwaste System including:

(C FR: 41.7/45.5)

A3.02 Automatic isolation KIA MATCH ANALYSIS Question asks what will happen during a Liquid Waste Release when the power to the effluent rad monitor loses power. Automatic isolation occurs.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect-Plausible because the isolation valve will close, but cannot be re-opened with no power.

B. Correct-Valve will close and cannot be re-opened with no power.

C. Incorrect-Isolation valve will close. Valve cannot be operated from the control room.

D. Incorrect-Isolation valve will close. Plausible because the valve can be isolated locally.

REFERENCES LOIT Bank 068K6.1O-02 13216-1, Liquid Waste Release Rev 43.1 VEGP learning objectives:

LO-PP-47101-09 State the conditions that require immediate termination of a Liquid waste release.

Page 104 of 208

Approved By . . Procedure Number Rev S. E. Prewitt Vogue Electric Generating Plant 13216-1 44 Date Approved Page Number

,/28/2O1O LIQUID WASTE RELEASE 3 of 86 2.0 PRECAUTIONS AND LIMITATIONS 2.1 PRECAUTIONS 2.1 .1 The Liquid Waste Processing System is potentially radioactive. Caution should be exercised to avoid spillage and to minimize exposure.

2.1.2 Once a Waste Monitor Tank (WMT) has been placed on recirculation for sampling, the tank shall remain isolated to prevent introduction of liquids that could alter the concentration of the contained volume.

2.1.3 Radiation Monitor 1 -RE-OO1 8 reading should be observed at least once every 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months during the release to assure that the activity does not exceed the setpoint on the Batch Liquid Release Permit.

2.1.4 If a high alarm is received from 1 -RE-OO1 8 while releasing a tank, the release shall be stopped immediately and the Shift Supervisor and Chemistry notified.

21 5 If I -RE-0O1 8 reads less than expected, release can continue provded Chemistry is notifledandiRX0OI8 does ntshow a tróubló ccndftin 2.1.6 Q NOT release more than one Waste Monitor Tank per plant site at the same time, unless authorized by the Chemistry Manager.

2.1 .7 If a high alarm is received from 1 -RE-OO1 8 while flushing with tank water, flush with demin water per Section 4.8.

2.1.8 If required to reset Dilution Flow Totalizer A-FQI-7620 prior to starting a release, Chemistry should be notified and Dilution Flow Totalizer A-FQI-7620 value recorded in Auto Log for the purpose of tracking tritium.

Printed January 18, 2011 at 10:40

C Q Loss of Power to Rad Monitors

A loss of power to certain monitors will cause ESFAS actuations

Those monitors are:

RE-002/003 RE-12116/12117 RE-2532A1B and RE-2533A1B RE-2565A,B,C Note these are energize to actuate. A loss of power to the DPM results in a relay race between DPM output relay and the SSPS input relay. May or may not get an actuation.

VLO-PP-321O1-O1 Rev-02.1 56

HL-16 NRC Written Examination KEY

52. 068AK3 .08 001/1/2/CONTROL RM EVACUATIO/3 .4/3.9 MEM!NEW/RO/SRO/NRC/GCW The control room has been evacuated due to a fire which posed a personnel hazard and the operators did not trip the MFPTs.

Later, the Shift Supervisor has opted to locally trip the MFPTs. Regarding the Condensate pumps, which of the following is correct regarding the required final condensate pump status and the reason?

Pump final status Reason for pump status A. all pumps stopped prevents overfilling the steam generators.

B. all pumps stopped prevents an excessive cooldown of the RCS.

C. one pump running maintains hotwell reject capability to the CSTs.

D one pump running cools equipment needed to maintain vacuum.

Page 1 05 of 208

HL-16 NRC Written Examination KEY Feedback 068 Control Room Evacuation Knowledge of the reasons for the following responses as they apply to the Control Room Evacuation:

(CFR 41.5,41.10/45.6/4513)

AK3.08 Trip of the MFW and necessary Condensate pumps K/A MATCH ANALYSIS Question gives a plausible scenario where the control room has been evacuated and the SS is directing shutdown of both MFPTs and appropriate condensate pumps. The candidate must choose how many condensate pumps must be shutdown and why.

ANSWER! DISTRACTOR ANALYSIS A. Incorrect. 18038 directs to leave one pump running. With MFPT running at a higher pressure than SG pressure, overfill of SGs is possible but not with condensate pumps which have a lower discharge pressure than SG pressure.

B. Incorrect. 18038 directs to leave one pump running. This is a reason for tripping the MFTPs to prevent cooling down the RCS due to steam draw when shutdown but not for the condensate pumps.

C. Incorrect. One pump running part is correct. Hotwell reject capability is maintained but is not the reason we leave a condensate pump running.

D. Correct. One pump should be left running post trip to ensure cooling to equipment necessary for condenser vacuum. (SJAE and SPE)

REFERENCES 18038, Operation from Remote Shutdown Panels steps # 42 and 43.

LO-PP-1 8101, Condensate and Feewater slides 58 through 67.

VEGP learning obiectives:

LO-PP-1 8101-06, Describe the operation of the Steam Jet Air Ejector Miniflow to include:

a. Why do we need this flow path Page 106 of 208

Approved Sy Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant 18038-1 32 DateApproved OPERATION FROM REMOTE SHUTDOWN PageNumber

/27/2010 PANELS 4 of 123 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED At the dJhift Supervisor, I6I each MFPT at its front standard NOTE Breaker for standby Condensate Pump should be rackOdtOd1Scai.ect before tripping running pumps breakers to prevent unwanted auto sta.

43. . At the discretion -

Supervisor, shut c

Both Heater Drain Pumps (TB Switchgear Room)

. lNAO4Brkrl3

1NAO1 Brkr 12

All but one Condensate Pump (TB Switchgear Room)

I NM Brkr 6

.--.. 1NAABrkr5

,NABBrkr4

All but one Circulating Water Pump (TB Switchgear Room):

. 1NAABrkr4 1NABBrkr5

All River Water Makeup Pumps not required considering requirements of both units (river intake):

ANAO1A Brkr 2

ANAO1A Brkr 3

ANAO1B Brkr 11

ANAO1B Brkr 12 Printed January 18, 2011 at 10:54

V-L .2 58

C 0 0 60

MIrJIFL:

1 FI-4486 1 FIC::-4466 1

8JAE COND MIN FLOW MINIFLOW RECIRC Fl-4486 indicates the total water being supplied by the Condensate Pumps I

Zi4.S Al 62

C) 7,000 gpm @119°F 12,000 gpm @131°F MIN FLOW RECIRC MFP FLOW FT CARD ENABLE\

0 DISABLE CARD SWITCH REDUCE FLOW TO FUNCTION GENERATOR SETPOINT 64 V-LO-PP-1 81 ALVES 1

° FV-4486

p1Fw rrn* r A 1* 1 r 4

r e .* jt.

r

CD CD ifl r

0 co cL 0

a

-J

C 0

.:(:rJ[. r,lIrJIFL::,

1 FI-44E 1 FIC:-44Ell At 15% Reactor Power the Unit Operating Procedure will have the operator reduce FIC 4486 to 0 output or 12,000 gpm total flow closing the mini-flow valve.

The potentiomete for FIC-4486 is set at 0.0 per the Unit Operating Procedure V-LO-PP-18101 Rev-8.2 guidance

7

HL-16 NRC Written Examination KEY

53. 073K1 .01 001/2/IIPRM SYSTEMS/3.6/3.9 MEM/LOIT BANK/RO/SROINRC/GCW A small RCS leak in Containment has illuminated the INTMD and HIGH RAD alarms.

Regarding the following valves:

- HV-12975, CNMT ISO SUPPLY TO RAD MON

- HV-1 2976, CNMT ISO SUPPLY TO RAD MON

- HV-12977, CNMT ISO RETURN FROM RAD MON

- HV-12978, CNMT ISO RETURN FROM RAD MON These valves would close due to a direct (1) signal from (2)

A. (1)CIA (2) RE-2562 B. (1)CIA (2) RE-2565 C. (1)CVI (2) RE-2562 D (1) CVI (2) RE-2565 Page 107 of 208

HL-16 NRC Written Examination KEY Feedback 073 Process Radiation Monitoring (PRM) System -

Knowledge of the physical connections and/or cause effect relationships between the PRM system and the following systems:

(CFR: 41.2 to 41.9145.7 to 45.8)

Ki .01 Those systems served by PRMs K/A MATCH ANALYSIS Direct match because the Containment Effluent Rad Monitor (RE-2565) will cause a Cnmt Ventilation Isolation.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect-CIA only happens Manually or from an SI. Plausible because it is Cnmt related for valve isolation but not the valves listed. RE-2562 is a Cnmt Rad Monitor for Cnmt Atmosphere and is alarm only.

B. Incorrect-C IA above in A. Plausible because RE-2565 is the Cnmt Vent Effluent Rad Monitor and will initiate a Cnmt Ventilation Isolation.

C. lncorrect-CVI will close the listed valves but not from RE-2562. RE-2562 is a Cnmt Rad Monitor for Cnmt Atmosphere and is alarm only.

D. Correct REFERENCES V-LO-TX-32101 Digital Rad Monitoring System ARP 17102-1 for 1-RE-2562A and C ARP 171 00-1 for 1-RE-2565 A, B, C VEGP learning obiectives:

LO-PP-321 01-09 Describe the automatic actions that occur for each of the following non-safety monitors when its high alarm setpoint is exceeded:

a. Containment Vent Effluent (RE-2665 A, B, C)

Page 108 of 208

Procedure Number Rev

.17100-1 26.1 umber 54 of 88 ORIGIN SETPOINT 1 -RE-2565A Moving Paper As determined by (High)

Airborne Chemistry Department Particulate Effluent Monitor NOTES

This Moving Paper Particulate Monitor has 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months time constant so indicator changes should be slow.

For other than HIGH conditions see Pages 4 and 5.

1.0 PROBABLE CAUSE High level of radiation from airborne Particulates in Containment Purge Vent.

2.0 AUTOMATIC ACTIONS Initiates Containment Ventilation Isolation.

3.0 INITIAL OPERATOR ACTIONS

1. InitIate evacuation of Containment jf the alarm is due to unexpected or unexplained radiation increases, Q jf appropriate HP controls are NOT in place for the radiological conditions indicated.

2. IF the alarm is due to expected radiation increases from preplanned evolutions AND appropriate HP controls are in place, THEN request HP and Chemistry to investigate the cause of alarm. jf required, initiate evacuation of Containment.

4.0 SUBSEQUENT OPERATOR ACTIONS NOTE Exhaust gasses are monitored at the plant vent by 1-RE-i 2442 A, B and C.

1. Verify Containment Ventilation Isolation by observing MLB Lights.

2. Account for all personnel in the Containment.

3. Notify Health Physics to survey and determine the source of radioactivity.

Pnnted January 9, 2011 at 16:39

Srvitt Vogtle Electric Generating Plant A Date Approved ANNUNCIATOR RESPONSE PROCEDURE FOR THE PROCESS AND Page Number 3/14/2010 EFFLUENT RADIATION MONITORING SYSTEM (RMS) 55 of 88 1 -RE-2565A (Continued)

4. Check for increased level of radioactivity indicated on 1 -RE-i 2442A.

5. Refer to NMP-EP-110, Emergency Classification And Implementing Instructions.

6. Obtain detector trend data per 13508-1, Radiation Monitoring Systems.

7. Monitor the channel for further changes.

8. if the cause was a spurious alarm AND WHEN conditions permit, have Chemistry reset and return the monitor to normal.

9. if sampling and analysis determine the channel has malfunctioned:

a. Comply with Technical Specification LCO 3.3.6.

Critical

b. Unlock CVI BLOCK PANEL 1-1609-P5-CB3 (Equipment Bldg R-117), and place 1-HS-13261 to BLOCK, Initial CV Initial

c. Request Chemistry to investigate and take corrective action.

10. WHEN conditions permit, reset CVI per 11886-1, Recovery From ESF Actuations.

5.0 COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE

REFERENCES:

1 X4DB21 3-1, 1 X5DX41 51 Printed January 9, 2011 at 16:39

Approved By I S. E. Prewitt I Vogtle Electric Generating Plant Procedure Number Rev 17100-1 26.1 Date Approved ANNUNCIATOR RESPONSE PROCEDURE FOR THE PROCESS AND Page Number 3/14/2010 EFFLUENT RADIATION MONITORING SYSTEM (RMS) 56 of 88 ORIGIN SETPOINT 1 -RE-2565B Effluent Iodine As determined by (High)

Monitor Chemistry Department NOTE For other than HIGH conditions see Pages 4 and 5.

1.0 PROBABLE CAUSE High concentration of gaseous radioactive iodine in the Containment Purge Vent.

2.0 AUTOMATIC ACTIONS Initiates Containment Ventilation Isolation.

3.0 INITIAL OPERATOR ACTIONS

1. InitIate evacuation of Containment if the alarm is due to unexpected or unexplained radiation increases, if appropriate HP controls are NOT in place for the radiological conditions indicated.

2. IF the alarm is due to expected radiation increases from preplanned evolutions AND appropriate HP controls are in place, THEN request HP and Chemistry to investigate the cause of alarm. jf required, initiate evacuation of Containment.

4.0 SUBSEQUENT OPERATOR ACTIONS NOTE Exhaust gasses are monitored at the plant vent by 1-RE-12442A, B and C.

1. Verify Containment Ventilation Isolation by observing MLB Lights.

2. Account for all personnel in the containment.

3. Notify Health Physics to survey and determine the source of radioactivity.

4. Check for increased level of radioactivity indicated on 1-RE-i 2442B.

Printed January 9, 2011 at 16:39

Eritt Vogtie Electric Generating Plant eeumb2eR1ev Date Approved ANNUNCIATOR RESPONSE PROCEDURE FOR THE PROCESS AND Page Number 3/14/2010 EFFLUENT RADIATION MONITORING SYSTEM (RMS) 57 of 88 1 -RE-2565B (Continued)

5. Refer to NMP-EP-1 10, Emergency Classification And Implementing Instructions.

6. Obtain detector trend data per 13508-1, Radiation Monitoring Systems.

7. Monitor the channel for further changes.

8. if the cause was a spurious alarm AND WHEN conditions permit, have Chemistry reset and return the monitor to normal.

9. if sampling and analysis determine the channel has malfunctioned:

a. Comply with Technical Specifications LCO 3.3.6.

Critical

b. Unlock CVI BLOCK PANEL 1-1609-P5-CB3 (Equipment Bldg R-1 17), and place 1 -HS-1 3261 to BLOCK, Initial CV Initial

c. Request Chemistry to investigate and take corrective action.

10. WHEN conditions permit, reset CVI per 11886-1, Recovery From ESF Actuations.

5.0 COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE

REFERENCES:

1X4DB213-1, 1X5DX4151 Printed January 9, 2011 at 16:39

I Approved By S. E. Prewitt I Vogtle Electric Generating Plant A Procedure Number Rev 17100-1 26.1 Date Approved ANNUNCIATOR RESPONSE PROCEDURE FOR THE PROCESS AND Page Number

.3/14/2010 EFFLUENT RADIATION MONITORING SYSTEM (RMS) 58 of 88 ORIGIN SETPOINT 1 -RE-2565C Effluent Radiogas As determined by (High)

Monitor Chemistry Department NOTE For other than HIGH conditions see Pages 4 and 5.

1.0 PROBABLE CAUSE Increase in concentration of radioactive gas in the Containment Purge Vent.

2.0 AUTOMATIC ACTIONS Initiates Containment Ventilation Isolation.

3.0 INITIAL OPERATOR ACTIONS

1. InitIate evacuation of Containment if the alarm is due to unexpected or unexplained radiation increases, QE if appropriate HP controls are NOT in place for the radiological conditions indicated.

2. IF the alarm is due to expected radiation increases from preplanned evolutions AND appropriate HP controls are in place, THEN request HP and Chemistry to investigate the cause of alarm. jf required, initiate evacuation of Containment.

4.0 SUBSEQUENT OPERATOR ACTIONS NOTE Exhaust gasses are monitored at the plant vent by 1-RE-i 2442A, B and C.

1. VerIfy Containment Ventilation Isolation by observing MLB Lights.

2. Account for all personnel in the containment.

3. Notify Health Physics to survey and determine the source of radioactivity.

4. Check for increased level of radioactivity indicated on 1-RE-i 2442C.

Printed January 9, 2011 at 16:39

Srvitt Date Approved Vogtle Electric Generating Plant A eeumb2eR;v ANNUNCIATOR RESPONSE PROCEDURE FOR THE PROCESS AND Page Number

3/14/2010 EFFLUENT RADIATION MONITORING SYSTEM (RMS) 59 of 88 1 -RE-2565C (Continued)

5. Refer to NMP-EP-110, Emergency Classification And Implementing Instructions.

6. Obtain detector trend data per 13508-1, Radiation Monitoring Systems.

7. Monitor the channel for further changes.

8. if the cause was a spurious alarm AND WHEN conditions permit, have Chemistry reset and return the monitor to normal.

9. jf sampling and analysis determine the channel has malfunctioned:

a. Comply with Technical Specifications LCO 3.3.6.

Critical

b. Unlock CVI BLOCK PANEL 1-1609-P5-CB3 (Equipment Bldg R-1 17), and place 1-HS-13261 to BLOCK, Initial CV Initial

c. Request Chemistry to investigate and take corrective action.

10. WHEN conditions permit, reset CVI per 11886-1, Recovery From ESF Actuations.

5.0 COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE

REFERENCES:

1X4DB213-1, 1X5DX4151 Printed January 9, 2011 at 16:39

PLANT VENT AIR MONITOR (LOW RANGE)

RE- 124424 8, C These monitors continuously monitor activity as It Is discharged to the environment through the main plant vent and provide redundant data necessary to ensure that activity releases from the plant vent are below specified limits. The plant vent monitors provide the final monitoring of airborne discharges to the environment. Because the main plant vent discharges directly to the atmosphere, the data from these monitors are most representativ e of activity releases to th. plant environs. The specific source of release within the plant may be determined through analysis of the various process monitors upstream of the vent monitor and/or with portable monitoring devices.

RE-2585A Containment Vent Effluent Air Particulate Monitor.

RE-25655 Containment Vent Effluent Iodine Monitor.

RE-2SEEC Containment Vent Effluent Radlogas Monitor.

These monitors measure radioactivity in the containment purge vent and provide the first Indication necessary to ensure that the release rate through the containment vent during purging Is below specified limits. The containment purge flow is filtered and muted through the main plant vent, where activity Is again monitored by the plant vent monitor RE-i 2442A,B,C and then discharged to the atmosphere.

RE-2562A Cont.IflmentAtmo.pher. Proceea Air Particulate Monitor R&2562B- Containment Atmoaphe. Iodine Cartridge RE-2562C Containment Atmosphere Procue Radlogu Monitor.

II Revision 0

PERMS MONITORS Gas Effluent Monitor Type Monitor Auto Actions Plant Vent BC-400 X RE-i 2442A Airborne Particulate Plant Vent Na I X RE-i2442B Airborne Iodine Plant Vent BC-400 X RE-i 2442C Airborne Gas Cnrnt Vent BC-400 X CVI RE-2565A Airborne Particulate Cnmt Vent Na I X CVI RE-2565B Airborne Iodine Cnmt Vent BC-400 X CVI RE-2565C AbornaGas RE-2562A Airborne Particulate Cnmt Atrnos Passive (filter)

RE-25628 Cnmt Atrnoa BC-400 RE-2562C Airborne Gas CR Intake GM Inline CR!

RE-12116 Vent CR Intake GM Inline CR!

RE-12117 Vent Waste Gas GM Inline ARE-013 Gas Waste Gas GM Inline X Auto closes RV-14 to isolate Waste Gas release ARE-Oi 4 Gas WG Decay GM Inline X Tank exh Vent Gas AE-39AB FHB Vent GM Inline X FHB Isolation RE-2532A1B Vent Gas FHB Vent GM Inline X FHB Isolation RE-2533NB Vent Gas RPF Vent Beta Scint. X ARE-i 6980 Steamline N16 RE-0724 SJAE Noble Gas RE-081 0 23 Revision 0

NORMAL OPERATIONS The Digital Radiation Monitoring system is primarily a data acquisition system. It gathers information about plant radiation levels and presents this information in an understandable format to those who must make the decisions based on this data. DRMS allows operators in the main control room to continually monitor the general radiation environment of the plant site and make adjustments to plant operation when needed. Under accident conditions, the radiation data aids the operator in determining the type and extent of the accident. The system also provides similar information to technical support personnel in the Technical Support Center (TSC) and in the Emergency Offsite Facility (EOF) to aid in reducing the radiological consequences of the event.

Although the DRMS is normally operated in the data-acquisition mode (non-interactive mode), the system is also provided with an interactive controls. All interactive tasks (such as testing and recalibrations) are performed by chemistry and l&C personnel.

Chemistry is responsible for performing interactive SRDC and communications console operations, including testing and changing settings. The SRDC and communications console keyboard are normally operated in the disabled (non-interactive) mode: operation of the keyboard will not affect parameters or data currentiy stored in system memory. To enable the interactive functions of the keyboard, passwords/keys must be used. Chemistry is contacted to perform any keyboard enabled function, such as resetting a high radiation alarm, suppressing an alarm, or setting the time.

The following actions will occur automatically if a high level radiation alarm is actuated on the associated monitor:

4$ fltalnment Vent Effluent (RE-25e5A, 8 or C Containment Ver olation (CVI) Isolation of IcóntainrnentpurgeandtheSamplethiasforUu1smonltór..

Control Room Intake (RE-12116 or 12117)- Control Room Isolation (CR1); places the control room and TSC HVAC in the emergency recirculation mode of operation.

Waste Gas Processing System Effluent (ARE-0014) isolates gas discharges if a gas release is In progress Fuel Handling Building Effluent (ARE-2532A or B; ARE 2533A or B) FHB isolation; places the FHB HVAC in the reclrculation mode of operation.

Waste Liquid Effluent (RE-0018) Isolates liquid waste release if in progress.

Steam Generator Blowdown (RE-0021) - isolates biowdown Turbine Building Drain (RE-0848) Re-align TB drains to dirty drain tank.

Steam Jet Air Ejector (RE-i 2839C, D, or E) Align discharge from direct discharge to environment to discharge through HEPA filter Containment Low Range (RE-0002 or RE-0003) - CVI 28 Revision 0

Several of the radiation monitors are required to be operable by the Technical Specifications. These are primarily monitors with automatic actuatlons; Enoperability of the monitor requires declaring the affected system inoperable.

Plant operation is then limited by the LCO for the affected system. The following monitors have tech spec requirements:

LCO 3.3.3, Post-Accident Monitoring RE-005, 006 Containment High Range RE-13119, 13120, 13121, 13122 Main Steamline Monitors LCO 3.3.8, CW lnsttum.ntatlon RE-002, 003, and 2565 ContaInment Low Range LCO 3.3.7, CREFS Instrumentation RE-12116, 12117CR Intake LCO 3.4.18, RCS Leakage Detection Systems RE-2562 Containment Atmosphere TR-13.4.6, FHB Post-Accident Actuation Instrumentation RE-2532 A/B FHB Atmosphere RE-2533 A/B FHB Atmosphere Liquid and osseous effluent radiation monitors requirements are located in the Qifsite Dose Calculation Manual (00CM). The ODCM should be referenced to determine individual rad monitor requirements.

STUDENT LEARNING ACTIVITES

1. Locally block and unblock CVI radiation monitor

2. Locally block and unblock FHBI radiation monitor

3. Verify proper automatic equipment responses and control room indications to the following high radiation alarms:

RE-12116,RE-12117

RE-0018

RE-002, RE-003

ARE-2532NB, ARE-2533A/B

ARE-0014

RE-2565A, 2565B, 2565C

RE-0021

RE-0848

4. Observe a DRMS source check

5. Trend specific radiation monitor readings on IPC

6. Identify DPMs that remain powered following an SI/LOSP by using IPC indications

7. Note radiation monitor responses to small RCS and SG tube leaks.

8. Using information from IPC determine when RE-724 will show bad data

9. Unit is in mode 1 at 100% power, RE-i 2442C has just failed. What actions are required?

10. IRE-0018 failed low during a liquid release. What actions must be taken and can the release be resumed with the monitor out of service?

11. The unit is in mode 1 at 100% power and RE-i 950 is showing an increasing trend. What other plant parameters could you look at to validate the responseof RE-i 950?

33 Revision 0

HL-16 NRC Written Examination KEY

54. 073K5.02 OO1/2/I/PRM-SOURCE DISTANCE/2.5/3. I C/A/BANK SURRY 09/RO/SRO/NRC/GCW Initial conditions:

- Radiography is in progress on a section of main steam piping.

- The radiographers want to verify the correct position of the camera by using a main steam line radiation monitor located on the same elevation and close to the area where the radiography needs to take place.

- To obtain a baseline reading, the camera source was placed 3.21 feet away from the radiation monitor detector.

- The radiation monitor read 5.92 Rlhr.

Current conditions:

- The camera has been moved into position to image the piping section.

- Engineering calculations show that the camera should be placed 17.46 feet away from the radiation monitor detector.

The distances listed above inlcude the difference in height from the camera to the radiation monitor detector. Consider the radiography camera as a radiation point source. Carry all calculations to three (3) decimal places.

Based on current conditions, which ONE of the following correctly identifies the expected reading on the radiation monitor, if the camera was positioned correctly?

A. 0.037 RIhr By 0.200 RIhr C. 1.O88RIhr D. 2.538 A/hr Page 109 of 208

HL-16 NRC Written Examination KEY Feedback 073 Process Radiation Monitoring (PRM) System Knowledge of the operational implications as they apply to concepts as they apply to the PRM system:

(C FR: 41.5/45.7)

K5.02 Radiation intensity changes with source distance K/A MATCH ANALYSIS Questions presents a plausible scenario where a point source calculation should be made. The candidate is presented initial dose at a given distance. The candidate has to make a dose calculation for a new given distance.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect-Plausible if a math error is made due to the reverse progression of the other numbers in the distractors.

B. Correct-I x (D1) 1 2 = 12 x (D2). Therefore, 5.92 RIhr x (3.21 ft) 2 2 / (17.46 ft) 2 = 12 Therefore, 12 = 0.200 r/hr.

C. Incorrect-Plausible if candidate does not square the distances in the equation.

D. Incorrect-Plausible if candidate uses the square root of the distances, instead of properly squaring the distances.

REFERENCES Surry 2009 NRC RO Exam question #51 V-LO-LP-38000-C Health Physics Fundamentals (point source formula)

VEGP learning objectives:

LO-LP-38000-14 Calculate the dose rate at a distance from a gamma point source or line source when given the dose rate at some other distance.

Page 110 of 208

II. LESSON OUTLINE: V-LO-LP-38000-04-C NOTES 2.5 rem

3. Minimize Time of Exposure:

Total Dose = Dose Rate X Time

a. Exercise Calculate the total dose in REM to a person who stays 30 minutes in a field of 5 mR/hr gamma and 2 mrad/hr neutron Solution Dose = DR x time Dose = 5 mR/hr x 30 mm x hrI6O mm Dose = 2.5 mrem (QF = 1 for gamma)

DR =DxQF

=2 mrad/hrxl0

= 20 mrem/hr

= 20 mrem/hr x 30 mm x 1 hr/60 mm

= 10 mrem Total Dose = 2.5 mrem + 10 mrem Total Dose = 12.5 mrem

4. Mkiçnizbistanóe frniSojic a Foutypéof sóürc geometry 1).. Point soUrce tsity.:of rdiatipn decrsesithqure Qf distance Iryerse square ia Eciithih,thtjeIatesdse ratea,id iénsitofj,rná radiatio DR= E r

C = activjtyircLries E = energy of ganima in,MeV d=distance from source in feet DR = dose rate in R/h a) Exérci.

Calpulate the distanceéquired to receive a dose rte of 100 rnRlhrif a point sóu yields an intensity of 75 RIhr at 1 foot Sohitioj

= (75E3 mrern/hr)(1 ft) 51

V-LO-LP-38000-04-C II. LESSON OUTLINE: NOTES 2

d .= (750 by;* Exercise Calculate the activity of a coblt-6O soUrée a détet reads 100 mR/hr 15 inches fforii the,. sour Soluti&i 0.0104 Ci

2) Line source radiation emitted Objective 14 from a uniform line in space.

Intensity falls off inversely with distance up to 1/2 the length of the line source. Beyond that distance the source acts like a point source. At 1/2 the length of the source the intensity is I d., = 12 2 d

1 ifd > 2 and d L > l..2, use 2

a) Exercise A line source if 10 feet long.

Three feet from the line source, the dose rate is 50 mR/hr. At what distance is the dose rate 2 mRlhr?

Solution

[ d1 = 12 d 2

12 (50 mr/hr)(3 ft) 5ft 12 = 30 mr/hr next use point source equation 52

RO Portion of Exam

51. Unit 1 Initial Conditions:

Radiography operations are in progress on a section fltlT o9 of main steam piping.

The radiographers want to verify the correct position of the camera by using a main steam line radiation monitor located on the same elevation and close to the area where the radiography needs to take place

To obtain a baseline reading, the camera source was placed 3.21 feet away from the radiation monitor detector. The radiation monito r read 5.92 R/hr.

Current conditions:

The camera has been moved into position to image the piping section.

Engineering calculations show that the camera shou ld be placed 17.46 feet away from the radiation monitor detector.

The distances listed above include the difference in height from the camera to the radiation monitor detector. Consider the radiography camera as a radiation point source. Carry all calculations to three (3) decimal place s.

Based on the current conditions, which ONE of the follow ing correctly identifies the expected reading on the radiation monitor, if the came ra was positioned correctly?

0.037 R/hr 0.200 R/hr C. 1.088R1hr D. 2.538 R/hr 51

HL-16 NRC Written Examination KEY

55. 076A3.02 OO1/2/IINSCW HEAT LOADS/3.7/3.7 MEMIMOD BANK/RO/SRO/NRC/GCW Given the following:

- A Reactor Trip and Safety Injection (SI) occur.

Which ONE of the following is CORRECT regarding NSCW cooling water status to the Containment Cooling system?

A. Auxiliary Coolers open-Containment Coolers open B. Auxiliary Coolers closed Containment Coolers closed C Auxiliary Coolers close Containment Coolers open-D. Auxiliary Coolers open Containment Coolers closed Feedback 076 Service Water System (SWS)

Ability to monitor automatic operation of the SWS, including:

(C FR: 41.7/45.5)

A3.02 Emergency heat loads K/A MATCH ANALYSIS The KA matches because the student must determine which Cnmt Cooling Systems are Safety Related and require NSCW cooling upon receipt of a Safety Injection.

ANSWER I DISTRACTOR ANALYSIS A. lncorrect-Cnmt Cooler valves are normally open but do AUTO open upon an SI signal if they are closed. Aux Cavity Coolers are not required after Rx trip and SI.

Their supply and return NSCW valves auto close on the SI.

B. Incorrect-Cnmt Cooler valves will AUTO open with the SI if they are closed.

Plausible because the Aux Cooler valves do AUTO close.

C. Correct Page 111 of 208

HL-16 NRC Written Examination KEY D. lncorrect-Cnmt Cooler valves will AUTO open with the SI if they are closed.

Plausible because the Aux Cooler are normally open but will close on the SI.

REFERENCES P&IDs 1X4DB135-1, 135-2 V-LO-TX-061 01 NSCW V-LO-TX-291 01 Containment HVAC Systems LOIT Bank LO-PP-29101-01-006 VEGP learning objectives:

LO-PP-06101-09 Describe the NSCW system response to an SI or LOSP signal.

LO-PP-29101-21 State any auto actions that occur in the systems listed as a result of the following signals: SI, High Rad, and CVI.

a. Containment Cooling Coolers

b. Preaccess filter

c. Preaccess (normal) Purge

d. Mini Purge

e. CRDM Cooling

f. Reactor Support Cooling

g. Reactor Cavity Cooling

h. Post-LOCA Cavity Purge Page 112 of 208

Chapter 06 NUCLEAR SERVICE COOLING WATER SYSTEM TABLE OF CONTENTS

6.1 INTRODUCTION

6.1.1 SYSTEM DESCRIPTION 6.1.2 DESIGN

SUMMARY

6.1.3 SYSTEM INTERFACES A. Fire Protection System B. Well Water System C. River Water System 6.2 MAJOR FLOW PATHS 6.3 COMPONENTS 6.3.1 NSCW Pumps & Motors 6.3.2 NSCW Cooling Towers 6.3.3 NSCW Transfer Pumps 6.3.4 System Valves A. NSCW Pump Discharge B. NSCW Temperature Control C. NSCW Tower Level Control D. Containment Cooler Isolation B* NSCW Tower Blow down 6.3.5 NSCW Tower Fans 6.3.6 Radiation Monitor 6.4 OPERATIONS 6.4.1 NORMAl. OPERATIONS A. System Startup from Drain Down Conditions B. Shifting Pumps C. System Shutdown 6.4.2 NON-PERIODIC OPERATIONS A. Cold Weather Operations

1. Deicing an Operating Train

2. Deicing an Idle Train B. Outage Operations Revision 7.1

0. Containmnt Coel*r Z.oltio There is a supply valve and a return valve for each pair of containment fan coolers for a total of 8 valves

. In addition, each train has supply and return valves for the flow to an auxiliary containment cooler and a reactor cavity cooler for an additional four valves for the two trains. Each of these 12 valves has a handswitch controller on the main control panel in the NSCW section. The switch has three positio ns:

open, close, and auto, with a spring return to auto.

In the AUTO position, if the valves isolating the main containment coolers are closed when a safety injection actuati on occurs, they will receive an automatic open signal. The operato r will also be prevented from manually closing these valves with the SI signal ptesent. tn contrast, the auxiliary cooler and the cavity cooler are not needed with the reactor shutdo wn so the supply and return valves to those units will receive an automatic closing command on the SI signal. Likew ise, the operator will not be able to manually open them before resetting the SI actuation signal.

B. NSCW Tower Blow down The blow down valve has a 3 position switch on the main control panel: close, auto, and modulate on. The switch controls the position of a blocking soleno id valve in the air line to the blow down valve. The auto blow down feature is not used at Vogtle: the valve is modul ated manually by chemistry from the NSCW chemical control buildin

g. A Safety Injection closes the air solenoid and the blowdown valve closes.

19 Revision 7.1

Additional potential paths could include containment building HVAC loads cooled by NSCW. The feature that makes all these scenarios unlikely is the operating pressure of NSCW being above the design pressures expected from each of these potential sources, including peak containment building pressure after a design basis loss of reactor coolant accident.

One scenario could be in-leakage from one of these sources with NSCW shutdown and depressurized followed by a subsequent startup of NSCW.

The other potential source that might be overlooked is an ongoing release from another path to the environment (plant stack, steamline break, TDAFW pump discharge, etc.) being carried into the NSCW system via the cooling towers and fans drawing in outside air.

6.4.4 SYSTEM RZSPONSZ TO St & LOSP A. SI When a SI signal is generated the NSCW system rej starting 2 pumps in each train (1 & 3 in train A, Train Br., opening the containment cooler isolation valj shutting the auxiliary and reactor cavity cooler isola valves, and shutting the NSCW tower blow down valve.

NSCW is normally in service with 2 pumps running, all containment cooler isolation valves open, and tower blow down in service as needed for chemistry control.

The NSCW pumps receive their start signal via sequencer contacts that close at a specified time to prevent overloading the electrical power source supplying AAO2/BAO3. NSCW pumps are one of the last loads to be sequenced on.

If the normal standby pump (5 or 6) is in service prior to the SI then after load sequencing is complete, you will have all 3

pumps running. This will require the operator to secure the standby pump (5 or 6) to return NSCW back to its design configuration.

The normal pump startup interlocks (discharge valve fully shut) is still active during the SI sequence.

B. LOSP When a tjv condition occurs on AAO2 or BAO3 all load feeder breakers for pumps are opened, then after the bus is re energized from the emergency diesel generator, the tnT load sequence is run. Immediately after the bus is re-energized the NSCW pump discharge valves for the 2 pumps that were in service prior to the tn start to close.

30 Revision 7.1

Chapter 29 CONTAINNT HVAC SYSTEMS INTRODUCTION 29.1 SYSTEM FLOWPATHS/COMPONENTS 29.2 SYSTEM INTERFACES 29.3 INSTRUMENTATION AND CONTROLS 29.4 SYSTEM OPERATIONS 29.5 TECHNICAL SPECIFICATIONS Revision 2.0

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 Coolin g 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 interlo cks 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 Instrumentation The Containment Coolers have hand switch controls on the QHVC panel and their respective Remote shutdown panels. There are separa te 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 MOVs on the QMCB. There are status lights on the monito r light boxes (MLB5) for the Containment Cooler low speed operati on, and for the NSCW MOV5.

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 interlo cked 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 offsite 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

On an SI signal, SSPS will send an open signal to the NSCW supply and return MOVs The NSCW supply and return MOVs are normally maintained Oefl.

Containment Aux Coolers System 1515 Controls and Instriimentation The Containment Aux Coolers have hand switch controls on the QHVC panel. There Hand switches on the QMCB for the NSCW supply and return MOVs. There is no instrument indicators provided for this system. There are individual fan low flow alarms on ALB 052.

Control Functions and Interlocks The CNNT Aux coolers will auto trip on low flow after a time delay. The same flow switch that trips the fan also drives the alarm. All other operations are manual.

The MOVS provided on the supply and return lines to the Aux cooler/Reactor Cavity cooler automatically isolate NSCW on an SI signaL Containment Lower Level Circulation System 1503 Controls and Instrumentation The Containment lower level circulators have hand switch controls on the QHVC panel. These hand switches are stop-start, maintained switches. There is no instrument indicators provided for this system.

Control Functions and Interlock.

There are no interlocks. All operations with this system are manual.

Containment Mini-Purge System and Main Purge System System 1505 and System 1506 These systems will be discussed together as they share common flow paths and are interlocked with each other.

Controls and Instrumentation The Containment Main and Mini Purge supply and exhaust fans have hand switch controls on the QHVC panel. There is no instrument indicators provided for this system. There is a narrow range CT pressure indicator P1-10945 on the QHVC panel with a range of -2.0 to ÷4.0 psig that is used to monitor containment pressure during containment purge operations. There are individual fan low flow alarms on ALB 052. There are High and High-high temperature alarms for the exhaust unit carbon filter on ALBO52.

16 Revision 2.0

Control PunctionB and Interlocks These systems are interlocked to prevent simultaneous operati on, because parts of their flow paths are shared. Each system has identical interlocks.

The heater in the Supply filter banks is interlocked so it can be energized only when a minimum measured air flow is presen t and either systems supply fan energizes. On Low supply air temperature, the running Purge Supply Fan will trip after a five minute delay.

Parallel supply isolation valves (HV-2627A and HV-2627B) outside of containment are interlocked so only one can be open at a time, as are the parallel supply isolation valves inside of containment (HV-2626A and HV-2626B).

Parallel exhaust isolation valves (HV-2629A and HV-26293) outside of containment are interlocked so only one can be open at a time, as are the parallel supply isolation valves inside of containment (HV-2628A and FIV-2628B).

The Mini-purge filter inlet isolation valve HV-12592 has a Close-Auto-open maintained hand switch on the QHVC This damper is downs tream of the ORC mini-purge containment isolation valve and provid ed with bypass flow orifice (FO-12593) around the damper to provide a contro lled vent path when containment pressure is greater than 0.30 psig to prevent damage to ductwork downstream. In Auto position, I-IV 12592 is interlocked to open upon mini-purge exhaust fan start. The main purge fans discharge damper (HV-2632A) has a Close-Auto-open mainta ined hand switch on the QHVC is interlocked to open upon fan start in Auto. The mini-purge exhaust f an discharge damper (HV-2632B) has all Auto functions deleted It is normally open and fails open on loss of power or air. It is only there for maintenance use.

All isolation valves close upon receipt of a containment ventilation isolation signal. The mini-purge isolation valves are air-ope rated and all fail closed.

Supply fans are interlocked so that only one may operate at a time. The exhaust fans are similarly interlocked. All fans stop on a loss of power and do not restart upon restoration of power.

Reactor Cavity Cooling Units System 1511 Controls and Instrumentation The Reactor Cavity Cooling Units have hand switch contro ls on the QHVC panel. There are no instrument indicators provided for this system. Low air flow alarms are provided on the QHVC. RX cavity cooler low flow (NSCW) alarms are provided on ALB 002 and ALB 003 on the QMCB Control Functions and Interlocks The Reactor Cavity Cooling units are automatically started on a loss of power by the Sequencer. This protects the Nuclear Instrum entation from high temperature on a loss of offsite power.

17 Revision 2.0

The MOVs provided on the supply and return lines to the Aux.

cooler/Reactor Cavity cooler automatically isolate NSCW on an St signal Unit 1/Unit 2 Differences On Unit 2, the MOVs provided on the supply and return lines to the Aux cooler/Reactor Cavity cooler automatically isolate NSCW on a Loss of offsite power. This is done by the Sequencer UV non sequenced relays as part of a NSCW water hammer prevention modification. This modification was not effective so it was not installed on Unit 1.

Reactor Support Cooling Fans System 1512 The Reactor Support Cooling Fans have hand switch controls on the QHVC panel. These hand switches are stop-start, maintained switches. There are no instrument indicators provided for this system. Low air flow alarms are provided on the QI-IVC.

Control Functions and Interlocks The fans are interlocked with their respective discharge dampers, causing the dampers to open upon fan start Control Rod Drive Mechanism Fans System 1509 Controls and Instrumentation The Fans have hand switch controls on the QHVC panel and on the Remote Shutdown panels. There are fan current meters supplied on the QHVC.

Unit 1/Unit 2 Differences On Unit 2, there is a with less than two CRDM fans running, a low flow alarm will annunciate in the control room. This design prevents damage to the CRDMs from overheating. Operation of the CRDMs without the fans running for greater than 30 minutes can cause rod drops due to failures or cause failure of the rods to operate. Unit 1 does not have this alarm.

Control Functions and Interlocks There are no interlocks. All operations with this system are manual.

Containment Preaccess Filter Units System 1504 Controls and Instrumentation The Containment Aux Coolers have hand switch controls on the QHVC panel. There is no instrument indicators provided for this system.

There are individual fan low flow alarms and filter High DP, High moisture, and high charcoal bed temperature alarms on ALS 052.

18 Revision 2.0

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1. LO-PP-29101-01 0061L0PP29101///////

An SI has occurred and has NOT been reset.

Which ONE of the following is CORRECT regarding NSCW cooling water status to the Containment Cooling system?

COOLERS NSCW VALVES A. Containment Cooler OPEN Auxiliary (Aux) Cooler OPEN Reactor Cavity Cooler OPEN B. Containment Cooler OPEN Auxiliary (Aux) Cooler SHUT Reactor Cavity Cooler OPEN C Containment Cooler OPEN Auxiliary (Aux) Cooler SHUT Reactor Cavity Cooler SHUT D. Containment Cooler SHUT Auxiliary (Aux) Cooler SHUT Reactor Cavity Cooler SHUT Feedback Categories Task Number (LO-TA): LOPP29 101 Objective/Source:

K/A: Exam/Question Type:

Cognitive Level: Origin/Rev Info:

Reference 1: Reference 2:

Page: 1 1/9/2011

1. LO-PP-29 101-10 002//022A2.04//C/A 3.1 / 3.5/LO-PP-2910 1-10//!

Given the following sequence of events:

The plant is at 100% power, all systems in normal configuration.

A reactor trip and safety injection (SI) occur.

Which ONE of the following would be a CORRECT response regarding NSCW cooling water status to the containment cooling systems and action(s) to take, if any?

A. Containment Cooler supply and return MOVs would be open.

Auxiliary Cooler and Rx. Cavity Cooler supply and return MOVs would be open.

No operator action is necessary, the alignment is proper.

B Containment Cooler NSCW supply and return MOVs would be open.

Auxiliary Cooler and Rx. Cavity Cooler supply and return MOVs would be shut.

No operator action is necessary, the alignment is proper.

C. Containment Cooler supply and return MOVs would be shut.

Auxiliary Cooler and Rx. Cavity Cooler supply and return MOVs would be open.

Shut the Auxiliary Cooler and Rx. Cavity Cooler supply and return MOVs D. Containment Cooler supply and return MOVs would be shut.

Auxiliary Cooler and Rx. Cavity Cooler supply and return MOVs would be shut.

Open the Containment Cooler supply and return MOVs.

Feedback Categories Task Number (LO-TA): Objective/Source: 022A2.04 K/A: Exam/Question Type: C/A 3.1 / 3.5 Cognitive Level: LO-PP-29101-10 Origin/Rev Info:

Reference 1: Reference 2:

Page: 1 1/9/2011

HL-16 NRC Written Examination KEY

56. 076AA2.03 001/1/2/HIGH RCS ACTIVITY/2.5/3.0 MEM/LOIT BANKIRO/SRO/NRC/GCW Initial conditions:

- Unit at full power with all systems in automatic.

Current conditions:

- CVCS letdown radiation monitor RE-48000 is off-scale high.

- Chemistry has validated the alarm with a sample.

Which of the following choices correctly describes the possible cause for this indication and the appropriate actions to take?

A. CVCS letdown mixed bed demineralizer over temperature.

Bypass the CVCS mixed bed demineralizer and divert letdown to the RHT.

B RCS fuel rod cladding leak.

Dispatch HP personnel to measure dose rates in the penetration room area and determine RCS Activity levels.

C. RCS fuel rod cladding leak.

Maximize CVCS and excess letdown flows and limit any future power changes until RE-48000 reading lowers to normal background.

D. CVCS letdown mixed bed demineralizer over temperature.

Isolate CVCS normal letdown and then place excess letdown in service.

Page 113 of 208

HL-16 NRC Written Examination KEY Feedback 076: High Reactor Coolant Activity Ability to determine and interpret the following as they apply to the High Reactor Coolant Activity:

(CFR: 43.5 /45.13)

AA2.03 RCS radioactivity level meter K/A MATCH ANALYSIS A plausible scenario is given where the High Radiation alarm is received on RE-48000, (CVCS Letdown) The student must determine the cause and appropriate actions to take.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect-A mixed-bed demin overtemperature would cause a release of their contaminants. Plausible, but RE-48000 is prior to the demins in the letdown flowpath.

B. Correct-Per the ARP, Fuel Rod Cladding leak or a Crud Burst are the probable causes of RE-48000 reading high. HP will measure dose rates in the area.

C. Incorrect-Fuel Rod Cladding leak correct. Plausible because for other RCS chemistry issues (i.e. Chlorides and Fluorides) letdown flow would be raised.

D. Incorrect-Same as A for the demin. Plausible for isolating letdown to prevent the hazard from exiting containment.

REFERENCES ARP-1 7005-1, window C03 for High Radiation.

ARP-1 7100-1, RE-48000 (High), CVCS Letdown Radiation Monitor.

LOIT Bank 076AA2.03-01 (Previously HL-15 Audit SRO-Removed SRO parameters)

V-LO-TX-09101 Chemical Volume and Control System VEGP learning obiectives:

Not applicable.

Page 114 of 208

Approved By I Procedure Number Rev S. E. Prewitt Vogtle Electric Generating Plant 17100-1 26.1 Date Approved ANNUNCIATOR RESPONSE PROCEDURE FOR THE PROCESS AND Page Number 3/14/2010 EFFLUENT RADIATION MONITORING SYSTEM (RMS) 87 of 88 ORIGIN SETPOINT 1 -RE-48000 Process Liquid As determined by (High)

Monitor Chemistry Department NOTE For other than HIGH conditions see Pages 4 and 5.

1.0 PROBABLE CAUSE High radiation level in the CVCS letdown due to:

a. Fuel rod cladding leak.

b. Crud burst.

2.0 AUTOMATIC ACTIONS NONE 3.0 INITIAL OPERATOR ACTIONS NONE Printed January 9, 2011 at 13:29

Eritt Vogtle Electric Generating Plant eeumb2eR;v Date Approved ANNUNCIATOR RESPONSE PROCEDURE FOR THE PROCESS AND Page Number 3/14/2010 EFFLUENT RADIATION MONITORING SYSTEM (RMS) 88 of 88 1 -RE-48000 (Continued) 4.0 SUBSEQUENT OPERATOR ACTIONS

1. Notify Health Physics to measure dose rates in Auxiliary Building RA-lO and determine RCS activity levels per 43014-C.

2. Request Chemistry to obtain and analyze RCS sample to confirm alarm.

3. Obtain detector trend data per 13508-1, Radiation Monitoring Systems.

4. Refer to 18014-C, Primary Plant Chemistry.

5. Refer to NMP-EP-1 10, Emergency Classification And Implementing Instructions.

6. Monitor the channel for further changes.

7. if sampling and analysis determine the channel has malfunctioned, request Chemistry to deactivate the channel.

5.0 COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE

REFERENCES:

1X4DB115, 1X5DS3GO4 Printed January 9, 2011 at 13:29

Approved By Procedure Number Rev C. S. Waidrup Vogtle Electric Generating Plant 17005-1 32.1 Date Approved ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 05 ON Page Number PANEL 1A2ON MCB 27 of 66 WINDOW C03 ORIGIN SETPOINT HIGH K-i Not Applicable RADIATION ALARM 1.0 PROBABLE CAUSE A high alarm on one or more of the Radiation Monitor Channels.

2.0 AUTOMATIC ACTIONS The following actions will occur if a High Level Radiation Alarm is actuated on the associated monitor:

1. 1 -RE-0002 or 1 -RE-0003, Containment Low Range Area Monitor:

Containment Ventilation Isolation (CVI).

2. A-RE-001 4, Waste Gas Processing System Effluent Radiogas Monitor:

Closes valve A-RV-001 4 to the Waste Gas Processing System discharge.

3. 1-RE-0018, Waste Liquid Effluent Monitor: Closes 1-RV-0018 to isolate the Liquid Waste Discharge Line.

4. 1-RE-0021, Steam Generator Blowdown Liquid Process Monitor: Isolates Steam Generator Blowdown Processing System.

5. 1 -RE-0848, Turbine Building Drain Effluent Monitor: Diverts Turbine Building Drains to Dirty Drains Tank.

6. A-RE-2532 A or B or A-RE-2533 A or B, Fuel Handling Building Effluent Radiogas Monitors: Fuel Handling Building Isolation (FHBI).

7. 1-RE-2565 A, B or C, Containment Ventilation Effluent Monitors:

Containment Ventilation Isolation (CVI).

8. 1-RE-i 2116 and 1-RE-i 2117, Control Room Intake Airborne Monitors:

Control Room Ventilation Isolation (CR1).

Printed January 18, 2011 at 11:35

Approved By Procedure Number Rev C. S. Waidrup Vogtle Electric Generating Plant 17005-1 32.1 Date Approved ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 05 ON Page Number 1./25/10 PANEL1A2ONMCB 28 of 66 WINDOW C03 (Continued)

9. 1-RE-12839 C, Condenser Air Ejector and Steam Packing Exhauster Effluent Monitor: Diverts air ejector discharge to filtration.

10. A-RE-50003, Technical Support Center Air Intake Monitor: Technical Support Center Ventilation Isolation.

3.0 iNITIAL OPERATOR ACTIONS NONE 4.0 SUBSEQUENT OPERATOR ACTIONS

1. Check the Safety Related Display Console (QRM2), the RMS Communications Console (QRM1) and the Plant Computer to determine the monitor in alarm and Go To 17100-1, Annunciator Response Procedure For The Process And Effluent Radiation Monitor System (RMS) or 17102-1, Annunciator Response Procedure For The Safety Related Display Console QRM2 as appropriate,

2. Initiate a CR documenting Alarm condition.

5.0 COMPENSATORY OPERATOR ACTIONS Monitor Plant Computer for radiation alarms if annunciator is inoperable or in solid.

END OF SUB-PROCEDURE

REFERENCES:

AX4DB1 23-2, AX4DB1 29, AX4DB1 81, AX4DB2O4-2, AX4DB235, 1X4DB126, 1X4DB142-1, 1X4DB179-2, 1X4DB18O-1, 1X4DB206-1, 1 X4DB21 3-1, 1 X4DB21 3-2, 1 X4DB229-3 Printed January 18, 2011 at 11:34

CHAPTER 9 CHEMICAL AND VOLU) CONTROL SYSTEM TABLE OF CONTENTS INTRODUCTION SECTION A CVCS LETDOWN SYSTEM 9.1 GENERAL LETDOWN SYSTEM FLOW PATH 9.2 COMPONENT DESCRIPTIONS 9.3 CVCS CHEMICAL ADDITION SYSTEMS 9.4 TECHNICAL SPECIFICATIONS AND TECHNICAL REQUIREMENT MANUAL SECTION B CVCS CHARGING SYSTEM 9.5 GENERAL CHARGING SYSTEM FLOW PATH 9.6 MAJOR COMPONENTS OF THE NORMAL CHARGING FLOW PATH 9.7 MAJOR COMPONENTS OF THE SEAL INJECTION FLOW PATH 9.8 CVCS LETDOWN AND CHARGING SYSTEM OPERATION 9.9 TECHNICAL SPECIFICATIONS AND TECHNICAL REQUIREMENT MANUAL SECTION C BORON THERMAL REGENERATION SYSTEM 9.10 OVERVIEW 9.11 OPERATION SECTION D EXCESS LETDOWN SYSTEM 9.12 EXCESS LETDOWN SYSTEM GENERAL FLOW PATH 9.13 EXCESS LETDOWN SYSTEM OPERATION SECTION E CVCS REACTOR MAKEUP CONTROL SYSTEM 9.14 INTRODUCTION 9.15 REACTOR MAKEUP CONTROL SYSTEM OPERATION SECTION F BORIC ACID STORAGE TANK 9.16 INTRODUCTION 9.17 BAST COMPONENTS 9.18 EMERGENCY BORATION Revision 4.2

to automatically close on low Pressurizer Level. The Letdown system is protected form overpressure by three Relief Valves. The first relief valve )PSV-8117) is O located downstream of the Letdown Orifice Isolation valves. The relief set point is set at 600 psig and its discharge point is to the Pressurizer Relief Tank.

relief valve discharge line, a tail pipe temperature detector that provides indication and an alarm which will warn the operator when the relief valve is On the lifting or if leak by is occurring. Downstream of the relief valve are three air operated fail close valves. Two of the three are Inside (IRC)and Outside(ORC)

Containment Isolation valves which are train related. Two of the three are also High Energy Line break actuation valves. The inside Containment Isolation valve (HV-8160) serves two functions as both Train A CIA and also Train A HELBA isolation. The CIA portion isolates the nonessential process in order to minimize leakage of fission products to the environment. Any Safety Injection Actuation generates a Containment Isolation signal. The High Energy Line Break actuation prevents damage to safety related systems and structures in the Auxiliary Building.

The penetration rooms in which the CVCS letdown line exits Containment and enter the Auxiliary Building have two trains of temperature transmitters. If the room temperature reaches the high set point in which you would expect if its line Letdown line ruptured, the HELBA valves will automatically close. If any of the three isolation valves (CIA/HELBA valves) shut for any reason CVCS Letdown should be removed from service as soon as possible. This is to minimize relief valve (PSV 8117) operation. This can be a hidden problem because the Letdown Flow indicator is downstream of the Containment isolation valves. The Operators may not think that Letdown is still in service based on flow indication showing zero flow meanwhile Letdown is continuously filling the PRT. Further downstream is the Temperature Control Valve TV-381B which regulates the amount flow through the Boron Thermal Regenerative System )BTRS) Letdown Reheat Heat Exchanger. This system will be discussed later in the Student Text. When TV-381B throttles close this of course increases the differential pressure which forces more flow through the Letdown Reheat Heat Exchanger. The diverted Letdown flow rejoins downstream of TV-381B which is used during the BTRS Boration Mode of operation. Normally isolated during at power operation the RHR system interfaces as what known as Low Pressure Letdown penetrates the Normal Letdown downstream of the TV-381B. During these modes operations while the RHR system is in service, the RCS pressure is not great enough to provide the driving force through the Letdown Orifices for adequate flow.

More than enough Letdown flow can be delivered from the discharge of the RHR pumps.

The Letdown temperature is further reduced by the Letdowu Heat Exchanger

$ch is the neict component inline. The cooling medium i iliary Componen dUng Watex system. The ACCW flow is throttled.uiqto maintain t Letdown temperáue downstream to approximately 1G4% asons for maintaini4 this temperature is: i to prevent the CVCS DemineraZere front rejecting t1era ctfliflaflts), and 2) Provide the RCP seals with a cool water supply. TV-130 is an air operated valve which fails open for maximum cooling. The next component in line is the Letdown Pressure Control Valve PV-l3l which maintains the back pressure on the Letdown Orifices at approximately 350 psig maintains subcooled conditions. Both the Regenerating Heat Exchanger and the Letdown Pressure control which is a fail open air operated valve that work in conjunction to prevent flashing at the Orifices. Two things an Operator can do to cause flashing in the Letdown system; 1) reducing CVCS charging flow to much, 2) or not being able to control Letdown Pressure while placing the system in service. Operation of the system will be discussed later in the text. The system flashing can cause serious water hammering which can cause equipment damage and injury to plant personnel.

A small portion of the Letdown flow just downstream of PV-131. is continuousl sampled by Radiatioa Monitor R-48OGO for activity levels in the RCS. This is not illustrated in the diagram but is shown on print l/2X4DB115. High levels of activity in the Reactor Coolant are indicative of Fuel cladding leaks or failures.

Not only is the activity in the RCS a Technical Specification concern but also it can cause the dose rates throughout the Containment and Auxiliary Building to increase. An important note is that an interface with the Dernineralized Water system exist which could be a Reactivity Concern for a dilution event. Demin water is used to flush RE-48000 under the Chemistry departments direction. During 6

Revision 4.2

flow outlet temperature TI-130 provides a signal into the TV-130 valve controller and it provides indication of the main control board. The desired temperature is Q adjusted on the controller potentiometer. The tube side of the unit is fabricated from austenitic stainless steel, and the shell side is carbon steel.

heat exchanger is located in the auxiliary building.

The letdown Letdown temperature controller normally operates automatically with its potentiometer adjusted to control letdown temperature at ll5°F. On a loss of air or power, the temperature control valve fails open. If the temperature input to TV-130 failed high, ACCW flow would maximize. If the temperature input failed low, TV-130 would throttle shut, trying to raise temperature. This would have the opposite effect on letdown temperature, though. A letdown high temperature alarm (127.5°F) should alert the operator to this condition. The appropriate response would be to manually lower letdown temperature by using TIC-130 in the control room. Letdown high temperature could also be the result of reduced charging flow. Again, the operator should manually raise charging flow. If letdown temperature cannot be lowered, letdown flow should be isolated.

Letdown flow indicatiom is provided in the control room on panel A and C respectively by FI-l32C and F1-132A. A high flow alarm is actuated if letdown flow reaches 130 gom.

Letdown Pressure Controller, PIC-131 Another very important letdown controller is the letdown pressure controller, which is set to automatically maintain pressure between the valve and the letdown orifices between 360 and 380 psig. This ensures that letdown flow does not flash between the orifices and downstream pressure controller, PIC-l31. Just as with TV-l30, controller input failures can result in system parameters exceeding normal operating values. A letdown heat exchanger outlet high pressure alarm occurs at 425 psig. P1-131 provides letdown pressure indication in the main control room.

Downstream of PIC-131 is pressure relief valve PSV-8l19. It provides pressure relief downstream of the low pressure letdown valve and protects the low pressure piping, demineralizers and filter from overpressure when this section of the system is isolated. The capacity of the relief valve is equal to the maximum flow rate through all letdown orifices. The relief valve is set to lift at 300 psig and discharges to the VCT. This is equal to the design pressure of the demineralizers.

¶Letdown pressure control valve )PV-l3l) (fails open) is an air-operated modulating valve which can be manually controlled from an automatic/manual controller on the QMCB. The pressure upstream of the valve is maintained above saturation for the letdown fluid to prevent flashing at the letdown orifices. Flashing at the letdown orifices would cause pressure oscillations and excessive erosion of the orifices.

Flashing occurs when the letdown pressure is reduced below the saturation pressure for the temperature of the letdown water. Flashing would occur if the pressure control valve fails open, fails to control pressure properly or if pressure oscillations occur in the letdown piping. Pressure oscillations may occur due to improper operator actions when placing the Letdown System on service. Care must be exercised during manual control of system valves to prevent this occurrence.

CVCS Letdown Radiation ?-

A small porti Radiátiont4onitor F t)at moattprs for t 10* radiation 1ev-I6e1C. 0*, &etud bur &tionawii follo4 Aquiri4t6. confirm n IftbeaJ idHealq 4ouzd;beriotified It rateá in tb:Auxi i.y Building.

I0 Revision 4.2

1. 076AA2.03 001/1/2/HI RCS ACTVTY-METERJC/A 3.OINEW/SRO/HL-15 AUDITIDSITNT Initial conditions:

- Unit at full power with all systems in automatic Current conditions:

- CVCS letdown radiation monitor RE-48000 is off-scale high

- Chemistry has validated the alarm with a sample Which of the following choices correctly describes the possible cause for this indication and the appropriate actions to take?

A. CVCS letdown mixed bed demineralizer over temperature.

Bypass the CVCS mixed bed demineralizer and divert letdown to the RHT.

B RCS fuel rod cladding leak.

Dispatch HP personnel to take local area radiation reading to determine if Dose Equivalent Iodine exceeds technical specification limits.

C. RCS fuel rod cladding leak.

Maximize CVCS and excess letdown flows and limit any future power changes to a maximum 3% / hr until RE-48000 reading lowers to normal background.

D. CVCS letdown mixed bed demineralizer over temperature.

Isolate CVCS normal letdown and then place excess letdown in service.

Page: 1 of 2 1/18/2011

Feedback KIA 076 High Reactor Coolant Activity AA2.03 Ability to determine and interpret the following as they apply to the High Reactor Coolant Activity:

RCS radioactivity level meter K/A MATCH ANALYSIS A plausible scenario is given where the High Radiation alarm is received at low power due to a crud burst following chemical addition. The candidate must choose the appropriate radiation monitor (RE-48000, CVCS Letdown) and the cause (Crud burst versus failed fuel).

ANSWER I DISTRACTOR ANALYSIS A.

REFERENCES ARP-1 7005-1, window C03 for High Radiation.

ARP-1 7100-1, page 88 and 89 for RE-48000, CVCS Letdown Radiation Monitor.

Page # 6 for detectors also referenced for RE-0007A, Rad Chem Lab disabled.

VEGP learning objectives:

Not applicable.

Categories Task Number (LO-TA): I Objective/Source: 2 KJA: HI RCS ACTVTY-METER Exam/Question Type: C/A 3.0 Cognitive Level: NEW Origin/Rev Info: SRO Reference 1: I-IL- 15 AUDIT Reference 2: DS/TNT Page: 2 of 2 1/18/2011

HL-16 NRC Written Examination KEY

57. 076K 1.05001/2/1 /NSCW-DG/3 .8/4.0 C/A/NEW/RO/SRO/NRC/GCW Given the following:

- DG 1A has been started per 14980A-1 Diesel Generator 1A Operability Test.

- During the surveillance run, the following alarms illuminated:

NSCW TRAIN A DG CLR LO FLOW NSCW TRAIN A LO HDR PRESS NSCW PMP HOUSE TRN A DRN SUMP HI LVL Which of the following has occurred and what will be the effect on the 1A DG?

A. NSCW Pump trip.

DG will trip on Hi Jacket Water Temperature.

B NSCW leak.

DG will trip on Hi Jacket Water Temperature.

C. NSCW Pump trip.

DG will remain running.

D. NSCW leak.

DG will remain running.

Feedback 076 Service Water System (SWS)

Knowledge of the physical connections and/or cause- effect relationships between the SWS and the following systems:

(C FR: 41.2 to 41.9/45.7 to 45.8)

K1.05 DIG K/A MATCH ANALYSIS The question has the student determine what event has occurred and whether the DG will trip on Hi Jacket Water temperature on a Normal Start.

ANSWER / DISTRACTOR ANALYSIS A. Incorrect-The tower sump alarm indicates a leak and not a pump trip. The DG is Page 115 of 208

HL-16 NRC Written Examination KEY running under the surveillance which is a Normal start. Hi Jacket water temperature will trip the DG. Plausible because a NSCW pump trip could cause a Hi Jacket Water Temp trip.

B. Correct C. Incorrect-A pump trip is not indicated, If the DG were running under an Emergency start condition, the DG would remain running. Plausible because the DG would remain running if it were an Emergency start.

D. Incorrect-The sump alarm in the NSCW tower is an indication of a leak.

Plausible because the DG would remain running if it were an Emergency start.

REFERENCES LOIT Bank LO-PP-06101-09-02 13145A-1 Diesel Generator Train A Annunciator Reponse Procedures:

17002-1 NSCW Train A DG Cir Lo Flow 17002-1 NSCW Train A Lo Hdr Press 17035-1 DG1A Ttrip Hi Temp Jacket Water 17062-1 NSCW Pmp House Ttrn A Drn Sump Hi Level VEGP learning objectives:

LO-PP-1 1101-25 Identify Diesel Engine Trips associated with the Jacket Water System, including:

a. setpoints and coincidences

b. when applicable LO-PP-1 1101-31 List the diesel engine trips and for each:

a. State the setpoint and coincidence (if applicable).

b. Identify as emergency or non emergency trip.

Page 116 of 208

I Approved By S. E. Prewitt I I

Vogtle Electric Generating Plant A Procedure Number Rev 13145A-1 3.2 Date Approved Page Number 3/2/10 I DIESELGENERATORTRAINA 1 of 78 DIESEL GENERATOR TRAIN A PROCEDURE USAGE REQUIREMENTS SECTIONS Continuous Use: Procedure must be open and readily available ALL at the work location. Follow procedure step by step unless otherwise directed.

Reference Use: Procedure or applicable section(s) available at NONE the work location for ready reference by person performing steps.

Information Use: Available on plant site for reference as needed. NONE Printed January 9, 2011 at 11:23

Approved By . Procedure Number Rev S. E. Prewitt Vogtle Electric Generating Plant 131 45A-1 3.2 Date Approved Page Number 3/2J1O DIESEL GENERATOR TRAIN A 2 of 78 1.0 PURPOSE This procedure provides instructions for the operation of the Train A Diesel Generator. This procedure should also be used for maintenance troubleshooting or maintenance testing. Operability of the Diesel Generator is proven using 14980A-1, Diesel Generator Operability Test. Specific instructions are provided in the following sections:

4.1.1 Preparing Train A Diesel Generator For Automatic Operation 4.1.2 Local Startup Of Train A Diesel Generator 4.1.3 Fast Start Of Train A Diesel Generator From The Control Room 4.1.4 Slow Start Of Train A Diesel Generator From The Control Room 4.1.5 Local Emergency Startup Of Train A Diesel Generator 4.3.1 Stopping Train A Diesel Generator 4.4.1 Cylinder Moisture Check 4.4.2 Emergency Stopping Train A Diesel Generator 4.4.3 Diesel Generator Operation Under Emergency Conditions 4.4.4 Adding Lube Oil To The Diesel Generator Sump 4.4.5 Switching From In-Service Lube Oil Filter To Standby Filter With Diesel Generator In Operation 4.4.6 Switching From In-Service Fuel Oil Fitter To Standby Fitter With Diesel Generator In Operation 4.4.7 Switching From In-Service Fuel Oil Strainer To Standby Fuel Oil Strainer With Diesel Generator In Operation 4.4.8 Generator Failure During Loss of Offsite Power 4.4.9 Transferring The Generator To The Redundant Rectifier Bridge 4.4.10 Transferring The Generator To The Alternate Voltage Regulator 4.4.11 Disabling Train A Diesel Generator Automatic Operation Printed January 9, 2011 at 11:23

I Approved By . Procedure Number Rev S. E. Prewitt Vogtle Electric Generating Plant 13145A-1 3.2 rDate Approved Page Number 3/2I1O DIESEL GENERATOR TRAIN A 3 of 78 4.4.12 Removing Train A Diesel Generator And Auxiliary Systems From Service 4.4.13 Response To An Air Receiver With Dewpoint Outside Required Limits Or Depressurized 4.4.14 Swapping Compressors To Opposite Air Receivers 4.4.15 Making Adjustments To Jacket Water Level 4.4.16 Inspecting The Control Air System Following Maintenance That Isolates And Restores The Control Air System 2.0 PRECAUTIONS AND LIMITATIONS 2.1 PRECAUTIONS 2.1.1 A Diesel Generator must be taken out of service if any resistance to engine rotation is encountered while operating the Pneumatic Barring Device.

2.1.2 The following Diesel Generator Electrical Protection Relays are enabled when the Diesel Generator is paralleled to the offsite power grid (i.e., surveillance testing). These are normally bypassed during a normal start when NOT in Parallel Mode.

a. Reverse Power 132. (trips 186C Lockout Relay)

b. Underfrequency 181.

c. Negative Phase Sequence 146. (trips 186C Lockout Relay)

NOTE Phase Overcurrent and Loss of Field will trip engine and breaker on a normal start.

2.1.3 When operating under Emergency Start conditions, the only active Diesel Generator protective devices are a.... Ge b.:: Lo

.. En Printed January 18, 2011 at 11:40

Procedure Number Rev Approved By S. E. Prewitt Vogtle Electric Generating Plant 131 45A-1 3.2 Page Number Date Approved 3/2/1O DIESELGENERATORTRAINA 4of78

d. Loss of Field (trips 1 86B lockout relay; trips only the breaker during LOSP or Emergency Start Switch starts; NOT active on SI start).

e. Phase Overcurrent (trips 1 86B lockout relay; trips only the breaker during LOSP or Emergency Start Switch starts; NOT active on SI start).

2.1.4 The Lube Oil and Jacket Water Keep-Warm Pumps and Heaters and the Generator Space Heater should be operating whenever a Diesel Generator is aligned for automatic startup.

2.1.5 The Maintenance Department should be notified per 00350-C, Work Request Program to make any changes or corrections to the governor settings. The governor Load Limit, Speed, or Speed Droop settings should NOT be altered unless:

Required by an approved test procedure.

OR The Torque Seal has been damaged or broken.

If the Diesel Generator is in continuous operation, additional supplies of fuel oil shall be ordered on, or before, the fifth day of continuous operation.

2.1.7 Emergency Diesel Generators shall NOT be used for peaking service.

2.1.8 If the Diesel Generator is being operated in the Parallel Mode, the LOCAL-REMOTE Switch 1 HS-451 6 on PDG1 shall NOT be transferred to LOCAL, as this will take governor and voltage regulator out of Droop Mode.

2.1.9 When the Diesel Generator is paralleled to the offsite power grid, the kVAR load should be maintained OUT and less than one half of the kilowatt load as shown in Vogtle Administrative Limits in Figure 2. The System Engineer must approve operation outside Vogtle Administrative Limits.

2.1 .10 The Diesel Generators should NOT be operated in parallel with the offsite grid for prolonged periods of time. This is to keep disturbances in the grid from affecting the Diesel Generators.

2.1 .11 Only one Diesel Generator should be operated at a time EXCEPT during Emergency Conditions.

Printed January 9, 2011 at 11:23

Approved By Procedure Number Rev S. E. Prewftt Vogtle Electric Generating Plant 131 45A-1 3.2 Date Approved Page Number 3/2/10 DIESELGENERATORTRAINA 5of78 2.1 .12 Generator excitation must be shutdown prior to transferring voltage regulators.

2.1.13 When DSL GEN 1A UNIT/PARALLEL switch is placed in PARA/SLOW START, the Unit Parallel Relay will energize and latch in. This will only allow the DG to Slow Start, even if an Emergency (Fast) Start is initiated. To remove the Slow Start Signal, the UNIT/PARALLEL switch must be taken to UNIT, which will de-energize the Unit Parallel Relay, and permit a 30-second Slow Start timer (internal to the governor) to begin timing. After 30 seconds, the DG can Fast Start. Subsequently, the Diesel Generator will be INOPERABLE from the time the UNIT/PARALLEL switch is placed in PARA/SLOW START until the DG is excited. If the DG is shutdown before the UNIT/PARALLEL switch has been placed in UNIT, the Diesel Generator will be INOPERABLE from the time it is shutdown until 30 seconds after the UNIT/PARALLEL switch has been placed in UNIT.

2.1.14 To serve as a dependable backup-power source, a DG should be kept separate from the offsite source if it is the only OPERABLE diesel. The DG should remain in standby and only be paralleled with an offsite source to meet surveillance requirements. Parallel operations may be conducted as a part of a preplanned activity if a supporting risk assessment has been completed.

2.1 .15 If Control Air pressure is depleted during DG operation, the DG will continue to run.

Restoration of Control Air pressure while DG is running may impact load carrying capability of the DG. Control Air pressure shall NOT be restored until DG shutdown is desired.

2.1.16 In the event of an abnormal trip of the DG, all personnel should avoid the Left Bank Side of the Engine for at least 15 minutes due to the risk of a Crankcase Explosion.

2.1.17 Any time a DG is operated greater than six hours in any 24-hour period, Environmental Services must be notified to verify compliance with EPA Regulations. Notification should be made by calling Environmental Services on-call Southern Linc phone # 1 -205-288-2064.

2.1.18 Independent Verifications performed in this procedure should be documented on Checklist 3, Train A Independent Verification.

Printed January 9, 2011 at 11:23

Approved By Procedure Number Rev S. E. Prewitt Vogtle Electric Generating Plant 13145A-1 3.2 Date Approved Page Number DIESEL GENERATOR TRAIN A 6 of 78 C 3/2/1O 2.2 LIMITATIONS 2.2.1 A Diesel Generator will NOT accept an Emergency Start signal from the Control Room if any of the following conditions exist:

a. LOCAL-REMOTE switch 1HS-4516 at PDG1 in LOCAL.

b. Starting air pressure in both air headers less than 150 psig.

c. Engine controls in the Maintenance Mode.

d. Emergency Stop circuit energized.

e. Overspeed trip NOT reset.

NOTE A Diesel Generator Emergency Start is initiated by closure of the Train A Engineered Safety Feature Safety Injection contacts, Loss of Offsite Power, or operation of the manual Emergency Start Switch at the Engine Control Panel.

All other Diesel Generator start signals are considered to be a Normal Start.

2 22 The following Diesel Erigjne shutdown signals are bypassed during aI

a. High Crankcase Pressure.

b. High Engine/Turbocharger Vibration.

c. Low Turbocharger Oil Pressure.

d. High Engine Bearing Temperature.

e. High Engine Lube Oil Temperature.

f. Low Jacket Water Pressure.

g. Loss Of Field And Phase Overcurrent 186B (SI only).

h. High Jacket-WaterTemperattjre Printed January 18, 2011 at 11:40

Approved By Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant 17002-1 23 Date Approved ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 02 ON PANEL 1A1 Page Number 11/04/2010 ONMCB 23of42 WINDOW C03 ORIGIN SETPOINT NSCW TRAIN A 1-FSL-1 650 1550 gpm DG CLR LO FLOW 1.0 PROBABLE CAUSE

1. Diesel Generator 1 A Cooler Tube blockage or bio fouling.

2. Loss of Nuclear Service Cooling Water (NSCW).

2.0 AUTOMATIC ACTIONS NONE 3.0 INITIAL OPERATOR ACTIONS Check NSCW pressure and flow indications and initiate 18021-C, Loss Of Nuclear Service Cooling Water System if NSCW is lost.

Printed January 9, 2011 at 12:13

Approved By Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant 17002-1 23 Date Approved ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 02 ON PANEL 1A1 Page Number 11/04/2010 ONMCB 24 of 42 WINDOW C03 (Continued) 4.0 SUBSEQUENT OPERATOR ACTIONS NOTE Diesel Generator 1 A will shut down at 200°F Jacket Water temperature if operating from a Normal Start.

1. if actual flow to the DG coolers is needed, dispatch an operator to check 1-FIT-i 650B (DG1A BLDG), NSCW DG Cooler Inlet, and convert % flow to GPM by multiplying % flow of 1-FIT-1650B X 2500 gpm. (Example:

62% flow = .62 x 2500 = 1550 gpm) (same type conversions in 14830, step 5.4.3.3).

2. jf 1-FIT-i 650B transmitter failure is expected, compare 1-FIT-i 6508, NSCW DG Cooler Inlet, with 1-FIT-i 650A, NSCW DG Cooler Outlet and convert % flow to gpm as described above.

3. IF Diesel Generator 1A is running, monitor Panel PDG2.

4. IF not an emergency start and fouling is suspected or equipment failure is indicated, stop the diesel and initiate maintenance as required.

5. Refer to Technical Specification LCO 3.8.1 or LCO 3.8.2.

5.0 COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE

REFERENCES:

1 X4DB1 35-1, 1 X3D-BD-KO4A, 1 X5DV032, CX5DT1 01-135 Printed January 9,2011 at 12:13

Procedure Number Rev Approved By J. B. Stanley Vogtle Electric Generating Plant 17002-1 23 Date Approved ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 02 ON PANEL 1 Al Page Number

11/04/2010 ON MCB 16 of 42 WINDOW BOl SETPOINT ORIGIN NSCW TRAIN A 1-PSL-1602B 70 psig LO HDR PRESS 1 -PSL-l 608B 1 -PSL-l 636B 1.0 PROBABLE CAUSE

1. Pump trip.

2. Leak in Nuclear Service Cooling Water System.

2.0 AUTOMATIC ACTIONS Standby pump starts on low header pressure.

3.0 INITIAL OPERATOR ACTIONS I.E low header pressure is due to pump trip Q NSCW leakage, Go To 18021-C, Loss Of Nuclear Service Cooling Water System.

4.0 SUBSEQUENT OPERATOR ACTIONS jf alarm is due to reasons other than listed in 3.0 (pressure switch failure, etc.),

shut down an unaffected pump per 131 50A-1, Train A Nuclear Service Cooling Water System.

5.0 COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE

REFERENCES:

1 X4DB1 33-1, 1 X3D-BD-KO4A, 1 X3D-BD-K04C, 1 X3D-BD-KO4E, 1X5DN087-1, -2, -3, CX5DT1O1 -135, 137, 138, FSAR Section 9.2.1 Printed January 9,2011 at 12:13

Approved By Procedure Number Rev S. E. Prewitt Vogtle Electric Generating Plant 17035-1 33.2 Date Approved ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 35 ON EAB Page Number 10-8-2009 PANEL 57 of 116 WINDOW C05 SETPOINT ORIGIN DG1A TRIP HI 1 TSH 19112 200° FT TEMP JACKET WATER 1.0 PROBABLE CAUSE

1. Three-way Mixing Valve 1 -TCV-1 9096 to Jacket Water Cooler has failed in the BYPASS position.

2. Engine-driven Jacket Water Pump malfunction.

3. Insufficient Nuclear Service Cooling Water flow through Jacket Water Cooler.

4. Engine overloaded.

2.0 AUTOMATIC ACTIONS

1. it .operatng foin icrnai Start, tile PésEWGállhutd*

2. If operating from an Emergency Start this alarm will only annunciate, NOT trip the Diesel Generator.

CAUTION Due to risk of a crankcase explosion, if circumstances permit and the SS/SM has not directed otherwise, the left side of the engine should be avoided for at least 15 minutes following a diesel engine trip.

3.0 INITIAL OPERATOR ACTIONS

1. operating from a Normal Start, verify the Diesel Generator is shut down.

2. IF operating from an Emergency Start AND plant conditions permit, transfer loads to the opposite train to reduce engine load Q secure the Diesel Generator.

See Note on Page 4.

Printed January 18, 2011 at 11:41

Approved By Procedure Number Rev S. E. Prewitt Vogtle Electric Generating Plant 17035-1 33.2 Date Approved ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 35 ON EAB Page Number (J 10-8-2009 PANEL 58 of 116 WINDOW C05 (Continued) 4.0 SUBSEQUENT OPERATOR ACTIONS

1. Dispatch an operator to investigate the cause of the trip.

2. Refer to Technical Specifications LCO 3.8.1 or 3.8.2.

5.0 COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE

REFERENCES:

1X4AKO1-443, IX4DB17O-1, 1X3D-BH-GO3F, CX5DT1O1-40G, FSAR Section 9.5.5 Printed January 9, 2011 at 11:27

Approved By Procedure Number Rev S. E. Prewitt Vogtle Electric Generating Plant 17062-1 22 Date Approved ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 62 ON PROCESS Page Number 8/5/2010 CONTROL PANEL 6 of 41 WINDOW AOl ORIGIN SETPOINT NSCW PMP HOUSE 1-LSH-7886 10 in. from TRN A DRN SUMP top of sump HI LVL 1.0 PROBABLE CAUSE

1. Leakage from Nuclear Service Cooling Water (NSCW) line tunnel.

2. Sump Pump 1-1215-P4-011 not operating.

3. Level Switches 1 -LSH-7886 or 1 -LSHL-7878 malfunctions.

2.0 AUTOMATIC ACTIONS NONE NOTE Sump Pump should have started at 12 in. from top of sump.

3.0 INITIAL OPERATOR ACTIONS NONE 4.0 SUBSEQUENT OPERATOR ACTIONS

1. Dispatch operator to NSCW Pumphouse to:

a. Verity 1-HS-7878 is in auto and Sump Pump running.

b. IF NOT, place 1 -HS-7878 to start position.

c. Check for possible leakage.

2. E Sump Pump is inoperable, notity Maintenance to install a portable pump and conduct repairs.

Printed January 9,2011 at 12:15

HL-16 NRC Written Examination KEY

58. 077AA2.07 00 i / I / I JGRID DISTURBANCES/36/4.0 C/A/MOD BANK ROBINSON/RO/SRO/NRC/GCW Given the following:

- Plant is in Mode 1 at 100% RTP.

- The Load Dispatcher reports several major generating stations have tripped and grid voltage is degrading.

- For the last 30 seconds 1 E 41 60V bus voltage has been reading 3740 volts.

Following the above conditions, the operator would expect the emergency busses to be energized from A. offsite power, the Diesel Generators will not start until bus voltage is approximately 2900 volts.

B. offsite power, the Diesel Generators will not start until bus voltage is approximately 3740 volts for one minute.

C. the Diesel Generators, due to reaching the Loss of Voltage setpoint.

D the Diesel Generators, due to reaching the Degraded Voltage setpoint.

Page llTof 208

HL-1 6 NRC Written Examination KEY Feedback 077 Generator Voltage and Electric Grid Disturbances Ability to determine and interpret the following as they apply to Generator Voltage and Electric Grid Disturbances:

(C FR: 41.5 and 43.5 I 45.5, 45.7, and 45.8)

AA2.07 Operational status of engineered safety features K/A MATCH ANALYSIS Question tests the ability to determine the setpoint of the LOP sequencer start on loss of voltage due to Grid instability.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect-The 1 E bus voltage has dropped to the point where the RAT feeder breaker would trip open and the DG would start. Plausable because the Loss of Voltage setpoint has been reached.

B. Incorrect-Same as above for the RAT. Bus voltage only has to be at approximatley 3740 volts for 20 seconds. Plausible because this is the correct voltage for the Degraded Voltage setpoint.

C. Incorrect-DG would have started on the Degraded voltage setpoint and not the Loss of Voltage.

D. Correct REFERENCES 18017-C Abnormal Grid Disturbances/Loss of Grid Tech Spec 3.3.5 LOP Instrumentation V-LO-PP-28201 Sequencer Modified Robinson 2009 NRC VEGP learning objectives:

LO-PP-28201 -02 List the signals that will start the sequencer, with applicable setpoints and coincidences.

Page 118 of 208

LOP Instrumentation 3.3.5 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Required Actions and D.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Times not met in MODES 1,2,3, or4.

D.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months E. Required Action and E.1 Enter applicable Immediately associated Completion Condition(s) and Time not met when the Required Action(s) for the associated DG is associated DG made required OPERABLE by inoperable by LOP DG LCO 3.8.2. start instrumentation.

SURVEILLANCE_REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.5.1 Perform COT. 92 days SR 3.3.5.2 Perform CHANNEL CALIBRATION with Nominal Trip 18 months Setpoint and Allowable Value as follows:

A. Loss of voltage Allowable Value 2912 V with a time delay of 0.8 second.

Loss of voltage Nominal Trip Setpoint 2975 V with a time delay of 0.8 second.

B. Degraded voltage Allowable Value 3683 V with a time delay of 20 seconds.

Degraded voltage Nominal Trip Setpoint 3746 V with a time delay of 20 seconds.

(continued)

Vogtle Units 1 and 2 3.3.5-2 Amendment No. 111 (Unit 1)

Amendment No. 89 (Unit 2)

Procedure Number Rev Approved By J B Stanley Vogtle Electric GenertingIant 18017-C 8 Page Number DateApproved ABNORMAL GRID DISTURBANCES/LOSS OF 1 of 52 V2912010 GRID ABNORMAL OPERATING PROCEDURE CONTINUOUS USE PURPOSE Section A of this procedure provides instructions to ensure vital equipment is available for LOSP due to degraded system voltage.

Section B of this procedure provides instructions to respond to a total loss of offsite power.

SYMPTOMS Symptoms for entry into Section A. DEGRADED GRID CONDITIONS are:

Notification from the Power Control Center that the security tools will be unavailable for greater than eight hours during normal system and weather conditions or for greater than one hour under abnormal conditions. (Entry into this AOP is NOT required while PCC is performing their manual calculations.)

Notification from the Power Control Center that the distribution center is one contingency away from being unable to maintain system voltage between 230 and 242 kV.

Annunciator Response Procedures 17036 and 17037, SEQ TROUBLE windows, if a degraded grid condition exists.

Symptoms for entry into Section B. LOSS OF GRID are:

All offsite power sources are de-energized.

MAJOR ACTIONS

Respond to degraded grid conditions.

Respond to loss of grid.

Printed January 7, 2011 at 16:31

Approved By Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant 18017-C 8 DateApproved ABNORMAL GRID DISTURBANCES/LOSS OF Page Number

()129/2010 2of52 GRID CONTINUOUS ACTIONS Step Actions SECTION A. DEGRADED GRID CONDITIONS A5 Maintain Main Generator operating within the reactive capability curve of FIGURE 1.

Al 3 Monitor 41 60V 1 E busses energized by offsite power.

A14 Check 1E 4.16kV switchgear voltages once per hour.

SECTION B. LOSS OF GRID Bil Control AFW to maintain SG NR levels between 10% and 65%.

B17 Try to restore offsite power.

Printed January 7, 2011 at 16:31

Approved By . Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant 1801 7-C 8

[Date Approved Page Number ABNORMAL GRID DISTURBANCES/LOSS OF (y29/2o1 0 GRID 3 of 52 A. DEGRADED GRID CONDITIONS ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

_A1. Check Diesel Generators - IN _A1. Restore Diesel Generators to STANDBY. operable status.

_A2. Terminate maintenance or testing activities on critical electrical distribution components.

_A3. Check Main Generator Power System _A3. Perform actions of TABLE 1, as Stabilizer on CCI PSS ENABLED.

- necessary.

_A4. Initiate the Continuous Actions Page.

A5 Maintain Main Generator - _A5. H Main Generator can NOT be OPERATING WITHIN THE maintained within the capability REACTIVE CAPABILITY CURVE OF curve, FIGURE 1. THEN trip the reactor and initiate I 19000-C, E-0 REACTOR TRIP OR SAFETY INJECTION.

A6. Place the following on alternate power supply using 13800, MAIN TURBINE OPERATION:

. Main Turbine Turning Gear

. Turning Gear Oil Pump A7. Verify Turning Gear Oil Pump:

With turbine on line IN AUTO.

-OR

_With turbine on turning gear IN - _Verify Main Turbine Auxiliary OPERATION. Emergency DC Lube Oil Pump is operating.

1 Printed January 7, 2011 at 16:31

Approved By Procedure Number Rev J B Stanley Vogtle Electric Generating Plant 18017-C 8 Page Number DateApproved ABNORMAL GRID DISTURBANCES/LOSS OF (J129/2olo 4of52 GRID A. DEGRADED GRID CONDITIONS ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

_A8. Inform shift personnel including Work Week Coordinator of degraded offsite electrical system condition and potential for loss of offsite power.

_A9. Designate individuals to monitor Diesel Generators and TDAFW pumps if a loss of offsite power occurs.

_A10. Verify SAT in standby and available.

_A1 1. Check Plant Wilson MANNED AND- _A1 1. Dispatch personnel to place all OPERATING. available CTs on turning gear.

_A12. Initiate 11400-C, STATION SERVICE REDUCTION CHECKLIST.

A13 Check either 4160V IE bus - *A13 Perform the following:

ENERGIZED BY OFFSITE POWER.

_a. Trip the reactor.

b. Initiate 19000-C, E-0 REACTOR TRIP OR SAFETY INJECTION.

_c. Go to Section B. LOSS OF GRID.

A14 Perform the following once per *A14 Perform the following:

hour:

. Check 1E 4.16kV switchgear a. Initiate TS 3.8.9 or 3.8.10.

voltages BETWEEN 3873 AND 4326V. b. Notify Engineering to evaluate long term effects of motor

. Record 1E 4.16kV switchgear operation outside the range of voltages in the Unit Control Log. 3873 to 4326V.

Printed January 7, 2011 at 16:31

Approved By Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant 1801 7-C 8 DateApproved ABNORMAL GRID DISTURBANCES/LOSS OF Page Number fr2912010 5of52 GRID A. DEGRADED GRID CONDITIONS ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

_A1 5. Check if PCC has terminated Al 5. Return to Step Al.

contingencies.

_A16. Return to procedure and step in effect.

° END OF SUB-PROCEDURE TEXT Printed January 7,2011 at 16:31

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ILC-09 NRC Exam 15.

Given the following:

- Plant is in Mode I at 40% RTP.

- The Load Dispatcher reports that several major generating stations have tripped and that grid voltage is degrading.

- 2 minutes ago, 480V bus voltage reduced to 425V.

Which ONE (1) of the following describes the current status of 480V Busses E-1 and E-2?

Emergency busses are energized from A. offsite power and the EDGs will NOT start until the generator lockout separates the emergency busses from offsite power.

B. offsite power and the EDGs will NOT start until the emergency busses undervoltage setpoint is reached.

the EDGs with Blackout loads operating.

D. the EDGs with Safeguards loads operating.

15

HL-16 NRC Written Examination KEY

59. 103K 1.08001/2/I /CNMT-SI-RESET/3 .6/3.8 C/AILOIT BANK!RO/SRO/NRC/GCW Unit 1 has the following conditions:

- A LOCA has occurred and SI actuation.

- Containment Pressure is 5 psig and slowly lowering.

- SI has NOT been reset.

- Instrument Air to Containment has been isolated.

Which ONE of the following describes what actions must be done to re-establish Instrument Air to Containment?

A Reset CIA.

B. Reset SI, then reset CIA.

C. Containment pressure < 3.8 psig, then reset CIA.

D. Containment pressure <3.8 psig, then reset SI, then reset CIA.

Page 119 of 208

HL-16 NRC Written Examination KEY Feedback 103 Containment System Knowledge of the physical connections and/or cause effect relationships between the containment system and the following systems:

(C FR: 41.2 to 41.9/45.7 to 45.8)

Ki .08 SIS, including action of safety injection reset K/A MATCH ANALYSIS Question requires student to determine what interlocks are required to be cleared or reset in order to establish IA to containment during a LOCA.

ANSWER I DISTRACTOR ANALYSIS A. Correct-CIA can be reset at any time.

B. Incorrect-SI reset is not required in order to reset CIA.

C. Incorrect-Containment pressure of 3.8 psig will initiate a SI, but the parameter does not have to be cleared to reset CIA.

D. Incorrect-An SI actuation signal is not required to be cleared in order to reset SI and SI reset is not required in order to reset CIA.

REFERENCES Procedure 11886-1 Recovery From ESF Actuations P&ID 1X4DB186-4 Instrument Air to Containment Building V-LO-TX-281 01-08 Reactor Protection System LOIT Bank 065AA1 .03-03 VEGP learning objectives:

LO-PP-021 01-08 List the conditions which will isolate instrument and service air to containment.

Page 120 of 208

35) HV-8843 SI boron injection test line bypass isolatio n train B

36) HV-8871 SI check valve test containment isolation valve train A

37) HV-8964 SI check valve test containment isolation valve train B

38) HV-8881 SI hot leg 1 & 4 check valve test isolatio n valve train B

39) HV-8823 SI Cold Leg injection check valve test isolatio n train B

40) HV-8890A SI pump recirc test line isolation

41) HV-8890B SI pump recirc test line isolation ROttIg6rCiA

1) Bt reset 1 sw 4

p hV ap it ep abch Io it in esty g due 3 tbfetentive memory .rithactuatir) bocIçcircuit Tliis ins that CIA dan RESET RESET tht1 with acuatioi1 thej original tüatir signal iustfirtcIe 4 ar,.Jhis Will IHS.4rnJ aIIow*aI3yfutufe I actuationignal to cause then.$ A IKS-4O1 5 e-ajt a 2)to Thst A, boti ret switcIce on the main contró j bordrrjjs be momerifarily pIced I theésetQsitidI ?

Containment Ventilation Isolation The purpose of CVI is similar to CIA that is to isolate cont ainment penetrations to prevent or minimize the escape of radioac tivity to the general public. The CVI actuation mainly isolates or effec ts HVAC systems.

CI-AICV)

CVI Actuation Signals

1) Safety Injection Signal

2) Hi radiation signal from Containment Area Radiation Monito r RE-002 or RE-003 IHS- NB

3) Hi radiation signal from Containment air particulate monito r

RE-2565A

4) Hi radiation signal from Containment iodine monitor RE-2565B

5) Hi radiation signal from Containment gas monitor RE-25 65C

6) Manual CIAICVI Actuating switches I out of 2 *

7) Manual Containment Spray Actuation 2 out of 2 on I out of 2 stations

V-LO-TX-281 01-08.1 34

C 0 r%J 01 C) 0 xi C) 3 0 a)

NJ H I

H P1 C) 0 II

Approved By . . Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant 11886-1 26.1 Date Approved Page Number

..2IO9 RECOVERY FROM ESF ACTUATIONS 14 of 23 DESIRED SYSTEM/COMPONENT STATUS INITIALS CORRECTIVE ACTIONS 4.3 REcEJj,CONTAiNMENT ISOLATION PHASE A (CIA)

INITIALS Isolation Phase A.

4.3.2 lnstrumiiifAir Containment OPEN Open using both Isolation 1 -HV-9378 1 -HS-9378A 1 -HS-9378B (QMCB);

4.3.3 Normal letdown (preferred) Established Return to normal CVCS or excess letdown to the VCT operation per 13006-1 or initiate excess letdown to the VCT per 13008-1, THEN; f open, close all Reactor Vessel Head Vent valves:

1 -HV-0442A and B, 1 -HV-8095A and B, 1 -HV-8096A and B.

4.3.4 RCP Seal Return Isolation OPEN .jf Open valves.

1-HV-8100 and 1-HV-8112 RCS>100 psig 4.3.5 Reopen PRT Fill Isolation 1-HV-8028.

4.3.6 As directed by the Shift Supervisor, place the RCDT back in service per 13002-1.

4.3.7 Re-align Hydrogen Monitor to normal for startup per 13130-1, Post-Accident Hydrogen Control if required.

4.3.8 CNMT DRAINS RX CAVITY AND CNMT SUMPS DISCH HDR ISOLATIONS 1 -HV-0780 OPEN 1-HV-0781 OPEN Printed January 18, 2011 at 11:52

1. 065AA 1.03 003/1/1/INSTRUMENT AIRJMEM 2.9/B/VG0530 I /RJMA B/RSB Unit 1 has the following conditions:

- An inadvertent SI has occurred.

- The SI signal has NOT yet been reset.

- 1 HV-9378, instrument Air to Containment, is close d.

Which ONE of the following describes why 1-HV-937 8 closed, and necessary actions to open the valve?

A 1-HV-9378 closed due to a CIA signal generated by the SI. The CIA can be reset without resetting SI, 1-HV-9378 can then be open ed from the QMCB.

B. 1-HV-9378 closed due to a CIA signal generated by the SI. The SI signal MUST be reset to allow CIA to be reset. After CIA is reset, 1-HV

-9378 can be opened from the QMCB.

C. 1-HV-9378 closed due to a CVI signal generated by the SI. The SI signal MUST be reset to allow CVI to be reset. After CVI is reset, 1-HV

-9378 can be opened from the QMCB.

D. 1-HV-9378 closed as a direct result of the SI. 1-HV

-9378 must be opened locally using the valve handwheel Q reset SI signal and then open 1-HV-9378 from the QMCB.

Page: 1 1/7/2011

HL-16 NRC Written Examination KEY

60. WEO4EK I .1 001 / I/I /LOCA ORC/3 .5/3.9 C/A!BANKjRO/SRO/NRC/GCW Which of the following correctly states actions contained in 19112-C, LOCA Outside Containment and the reasons for those actions?

A. Open HV-8802A (SI PMP-A TO HOT LEG 1 & 4 ISO VLV) to provide a flow path for Low Head Safety Injection. Then close HV-8809A (RHR PMP-A TO COLD LEG 1 &

2 ISO VLV) and monitor RCS pressure.

B. Close HV-8809A (RHR PMP-A TO COLD LEG 1 & 2 ISO VLV). If RCS pressure does not rise then allow the valve to remain closed because this will allow time for the operators to check Aux Building leak detection alarms while the flow path is isolated.

C. If the leak is not identified and isolated then transition to 1901 0-C, Loss of Reactor or Secondary Coolant, because RCS inventory will continue to be lost outside of containment.

D Close HV-8809A (RHR PMP-A TO COLD LEG 1 & 2 ISO VLV). If this action results in an RCS pressure rise then stop the A Train RHR Pump.

Page 121 of 208

HL-16 NRC Written Examination KEY Feedback E04 LOCA Outside Containment Knowledge of the operational implications of the following concepts as they apply to the (LOCA Outside Containment)

(CFR: 41.8/41.10, 45.3)

EK1 .1 Components, capacity, and function of emergency systems.

K/A MATCH ANALYSIS This question tests operational implications of closing valves to identify the source of the leak. Therefore, the operational implications of components during a LOCA ouitside containment are being tested.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect- No procedural guidance is given to open HV-8802A. Plausible because guidance does exist to close HV-8809A.

B. Incorrect-If pressure does not rise when the valve is closed, then closing the valve did not isolate the leak and the procedure directs the valve to be re-opened.

C. Incorrect-If the leak is not isolated, then the correct action would be to transition to 19111-C ECS-1 .1 Loss of Emergency Coolant Recirculation. Plausible because a LOCA does exist.

D. Correct- 19112-C Step 2.

REFERENCES 19112-C LOCA Outside Containment Vogtle 2005-30 1 Exam VEGP learning obiectives:

LO-PP-371 16-02 Describe the steps taken to isolate a LOCA outside containment.

LO-PP-371 16-03 Describe the indications used to confirm that a LOCA outside containment was successfully isolated.

LO-PP-371 16-04 Describe the possible EOP transitions out of 19112-C based on either successful or unsuccessful isolation of a LOCA outside containment.

Page 122 of 208

Approved By Procedure No.

T E.

. Tynan Vogtle Electric Generating Plant NUCLEAR OPERATIONS A 19112-C Revision No.

Date 3172006 Unit COMMON Page No.

1 of 4 EMERGENCY OPERATING PROCEDURE ECA-1.2 LOCA OUTSIDE CONTAINMENT PURPOSE PRB REVIEW REOUIRED This procedure provides actions to identify and isolate a LOCA outside containment. (Applicable in Modes 1, 2, and 3.)

ENTRY CONDITIONS

19000-C, E-0 REACTOR TRIP OR SAFETY INJECTIO N

19010-C, E-1 LOSS OF REACTOR OR SECONDARY COOLANT

19005-C, REDIAGNOSIS MAJOR ACTIONS

Verify Proper Valve Alignment.

Identify and Isolate Break.

Check If Break Is Isolated.

PROCEDURE NO.

REVISION NO. PAGE NO.

VEGP 19112-C 5 2 of 4 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

1. Verify proper valve alignment:

a. RHR Pump suction from RCS - CLOSED:

Q

HV-8701A - RHR PMP-A DOWNSTREAM SUCTION FROM HOT LEG LOOP-i Q

HV-8701B - RHR PMP-A UPSTREAM SUCTION FROM HOT LEG LOOP-i D

HV-8702A - RHR PMP-B DOWNSTREAM SUCTION FROM HOT LEG LOOP-4 D

HV-8702B - RHR PMP-B UPSTREAM SUCTION FROM HOT LEG LOOP-4

b. RHR Pump Hot Leg b. Dispatch an Operator to injection valve CLOSED:

close affected Unit valve:

Q

HV-8840 - RHR TO HL Q l-HV-8840 ISO VLV - RHR TO HL ISO VLV (AB-A13)

Q 2-HV-8840 - RHR TO HL ISO VLV (AB-A18)

c. SI Pump Hot Leg injection

c. Dispatch an Operator to valves CLOSED:

close affected Unit valves:

D

HV-8802A - SI PMP-A TO HOT LEG 1&4 ISO VLV Q l-HV-8802A - SI PMP-A TO HOT LEG 1&4 ISO D. HV-8802B - SI PMP-B TO VLV (AB-A09)

HOT LEG 2&3 ISO VLV Q 1-HV-8802B - SI PMP-B TO HOT LEG 2&3 ISO VLV (FHB-A1O) 2HV-8802A - SI PMP-A TO HOT LEG i&4 ISO VLV (AB-A103)

Q 2-HV-8802B - SI PMP-B TO HOT LEG 2&3 ISO VLV (FHB-AO1)

PROCEDURE NO. REVISION NO. PA6E NO.

VEGP 19112-C 5 3 of 4 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

2. Try to identify and isolate RI-JR Cold Leg injection break:

Q a. Close RI-JR PMP-A TO COLD LEG 1&2 ISO VLV HV-8809A.

Db. Check RCS pressure Db. Open RI-JR PMP-A TO COLD

- RISING LEG 1&2 ISO VLV HV-8809A.

D Go to Step 2d.

D c. Go to Step 2f.

Q d. Close RI-JR PMP-B TO COLD LEG 3&4 ISO VLV HV-8809B.

D e. Check RCS pressure D e. Open RI-JR PMP-B TO COLD

- RISING LEG 3&4 ISO VLV HV-8809B.

DGo to Step 3.

D f. Stop RHR Pump in train with leak isolated.

Dg. Go to Step 4.

PROCEDURE NO. REVISION NO. PAGE NO.

VEGP 19112-C 5 4 of 4 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

3. Try to identify and isolate SI Cold Leg injection break:

Q a. Close SI PMP-A TO COLD LEG ISO VLV HV-8821A.

Db. Check RCS pressure Qb. Open SI PMP-A TO COLD LEG

- RISING ISO VLV HV-8821A.

QGo to Step 3d.

Dc. Go to Step 3k.

D d. Close SI PMP-B TO COLD LEG ISO VLV HV-8821B.

D e. Check RCS pressure D e. Open SI PMP-B TO COLD LEG

- RISING ISO VLV HV-8821B.

DG0 to Step 3g.

D f. Go to Step 3k.

Dg. Close COLD LEG INJECTION FROM SIS HV-8835.

D h. Check RCS pressure D h. Open COLD LEG INJECTION

- RISING FROM SIS HV-8835.

DG0 to Step 4.

i. Stop both SI Pumps.

Dj. GotoStep4.

D k. Stop SI Pump in train with leak isolated.

4. Check if break is isolated:

D a. Check RCS pressure D a. Go to 19111-C, ECA-l.1

- RISING LOSS OF EMERGENCY COOLANT RECIRCULATION.

Ob. Go to 19010-C, E1 LOSS OF REACTOR OR SECONDARY COOLANT.

END OF PROCEDURE TEXT

Vogtle Nuclear Plant 2005-301 SRO Inital Exam

70. WEO4EK 1.1 001/1/ 1/LOCA OUTSIDE CONTAIN/MEM 3.5/B/VG05301/R/M AI3/RSB Which ONE of the following correctly states actions contained in 191 12-C, ECA-l .2 LOCA Outside Containment, and reasons for those actions?

A. Open HV-8802A (SI PM P-A TO HOT LEG 1 & 4 ISO VLV) to provid e a flow path for Low Head Safety Injection. Then close HV-8809A (RHR PMP-A TO COLD LEG 1

& 2 ISO VLV) and moi,itor RCS pressure.

B. Close HV-8809A (RHR PMP-A TO COLD LEG 1 & 2 ISO VLV). If this action does not result in an RCS pressure rise then allow the valve to remain closed because this will allow time for the operators to check Auxiliary Building alarms while the flow path is isolated.

C. If the leak is not identified and isolated then transition to 19010-C, Loss of Reactor or Secondary Coolant, because RCS inventory will continue to be lost outside of containment.

D Close HV-8809A (RHR PMP-A TO COLD LEG 1 & 2 ISO VLV).

If this action results in an RCS pressure rise then stop the A RHR Pump.

Vogtle Nuclear Plant 2005-301 Sf0 Inital Exam K/A WEO4 LOCA Outside Containment EK1 .1 Knowledge of the operational implications of the following concepts as they apply to the (LOCA Outside Containment): Components, capaci ty, and function of emergency systems.

K/A MATCH ANALYSIS The question tests operational implications of closing valves to identify the source of the leak. Therefore, the operational implications of compone nts during a LOCA outside containment are being tested.

ANSWER / DISTRACTOR ANALYSIS A. Incorrect. No procedural direction is given to open HV880 2A. Plausible because guidance does exist to close HV-8809A.

B. Incorrect. If pressure does not rise when the valve is close d, then closing the valve did not isolate the leak and the procedure directs the valve to be re-opened.

Plausible because guidance does exist to close HV-8809A.

C. Incorrect. If the leak is not isolated, then the correct action would be to transition to ECA-1 .1, Loss of Emergency Coolant Recirculation. Plausible because a LOCA does exist.

D. Correct. 191 12-C, Step 2.

REFERENCES

1. Surry 2004-301 Exam Question WEO4EK3.2.

2. 19112-C, ECA-1 .2, LOCA Outside Containment, Rev. 4.1, 11/22/

2000.

HL-16 NRC Written Examination KEY

61. WEO5EA 1.1 001/1/1/LOSS OF HEAT SINK/4. 1/4.0 C/AILOIT BANK/RO/SRO/NRC/GCW Given the following:

- A transition from E-0 19000-C to 19231-C Response To Loss of Secondary Heat Sink has occurred.

- All Steam Generator Wide Range levels are approximately 5%.

- RCS temperature is approximately 580°F and stable.

Per 19231-C, which one of the following describes the preferred method of initiating Auxilary Feed flow for these conditions?

A Feed a selected SG at 30 - 100 gpm until WR level > 9%.

B. Feed a selected SG with no flow restrictions until NR level is> 10%.

C. Feed any or all SGs at 30 - 100 gpm until WR level > 9%.

D. Feed any or all SGs with no flow restrictions until NR level is> 10%.

Page 123 of 208

HL-16 NRC Written Examination KEY Feedback E05 Loss of Secondary Heat Sink Ability to operate and / or monitor the following as they apply to the (Loss of Secondary Heat Sink)

(C FR: 41.7/45.5/45.6)

EA1 .1 Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

K/A MATCH ANALYSIS Question presents a plausible scenario where rate of feeding a Dry SG is determined to prevent possible SG tube damage upon restoring level.

ANSWER / DISTRACTOR ANALYSIS A. Correct-A selected SG is used because of the level and stable RCS temperature.

B. Incorrect-A flow rate restriction of 30 -100 gpm is set because of SG WR level being < 9%. If RCS temperature were rising, the flow rate restriction will not be set. Plausible because of a selected SG.

C. Incorrect-Only a selected SG is fed at 30 -100 gpm. Plausible because of the feed rate.

D. Incorrect-A selected SG is used because of the level and stable RCS temperature.

REFERENCES 19231-C Response To Loss Of Secondary Heat Sink VEGP learning obiectives:

LO-LP-37051-05 State the precautions which should be taken in feeding a hot, dry SG following recovery from a loss of heat sink accident.

Page 124 of 208

Approved By Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant 19231-C 33.4 Date Approved Page Number FR-H.1 RESPONSE TO LOSS OF SECONDARY 18/10 HEAT SINK 25 of 54 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED CAUTION Feed flow rates should be controlled to prevent excessive RCS cooldown.

49 Try to establish MDAFW flow to at least one SG:

a. Check MDAFW Pump - a. Perlorm the following:

AVAI LABLE:

Power available

Initiate actions to restore

Suction pressure an MDAFW Pump:

Discharge pressure

_a) Reference 13610, AUXILIARY FEEDWATER SYSTEM

_b) Verify MDAFW Pump discharge throttle valves closed.

. WHEN MDAFW Pump is started, THEN go to Step 49.b.

Go to Step 53.

b. Select SG(s) to feed: b. Perform the following:

1) All SG WR levels LESS

Restore feed flow to THAN 9% [31% ADVERSE]. Non-Dry SG(s) by going to Step 50.

Step 49 continued on next page Printed January 7, 2011 at 12:46

Approved By Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant 19231-C 33.4 Date Approved Page Number FR-H.1 RESPONSE TO LOSS OF SECONDARY ii 8

jii 0 HEATSINK 26of54 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

_c. Check Core Exit TCs STABLE - _c. Do NOT limit feed flow to the OR LOWERING. selected SG if Core Exit TCs are rising and go to Step 49.f.

d. Restore feed flow to selected SG

- BETWEEN 30 GPM AND 100 GPM:

IPC Point UF5403.

e. Check Dry SG WR level - _e. WHEN Dry SG WR level is GREATER THAN 9% greater than 9%

[31% ADVERSE]. [31% ADVERSE]

THEN raise feed flow to restore NR level greater than 10% [32% ADVERSE].

Go to Step 70.

, Raise teed flow to restore NR level greater than 10%

[32% ADVERSE] and go to Step 70.

50. Verify MDAFW Pump throttle valves 50. Perform the following as open for selected SG(s): necessary to establish MDAFW feed flow:

HV-5139 MDAFW Pump A to SG 1 Open MDAFW Pump HV-51 37 MDAFW Pump A to SG crosstie valves:

HV-5132 MDAFW Pump B to SG 2

_1) 1 302-U4-055 HV-5134 MDAFW Pump B to SG 3

_2) 1 302-U4-056

Limit flow rate to avoid pump runout LESS THAN 600 GPM.

Printed January 7,2011 at 12:46

HL-16 NRC Written Examination KEY

62. WE08EA 1.1 001/1 /2JPTS-RCS OVERCOOLING/3 8/3.8 C/A/LOIT BANKJRO/SRO/NRC/GCW Given the following:

- Unit 1 Reactor tripped and a natural circulation cooldown was required.

- The COPs system was placed in service in accordance with 19002-C, Natural Circulation Cooldown.

- Subsequently, a faulted Steam Generator resulted into entry into 19241-C, Pressurized Thermal Shock.

- All RCS temperatures have been stablized at approximately 290°F.

- RCS pressure is 650 psig.

- Loop 3 WR Cold Leg temperature fails bottom of scale LOW.

Which ONE of the following describes the effect on the Unit?

A. PORV 455 will open and remain open unless it is manually closed.

B. PORV 456 will open and remain open unless it is manually closed.

C. PORV 455 will open until RCS pressure lowers below the minimum COPs setpoint.

D PORV 456 will open until RCS pressure lowers below the minimum COPs setpoint.

Feedback E08 Pressurized Thermal Shock Ability to operate and I or monitor the following as they apply to the (Pressurized Thermal Shock)

(C FR: 41.7/45.5/ 45.6)

EA1 .1 Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

K/A MATCH ANALYSIS Question has the student determine COPS protection and operations with temperature and pressure values and a failed temperature channel and what PORV is affected.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect-PORV 455 will be unaffected because WR Cold Leg temperatures are associated with B Train, PORV 456.

Page 125 of 208

HL-16 NRC Written Examination KEY B. Incorrect-Plausible because PORV 456 will open but will close if pressure drops below the COPS setpoint.

C. lncorrect-PORV 455 will unaffected. Plausible because 456 will close if pressure drops below the COPS setpoint.

D. Correct REFERENCES LOIT Bank WEO8EA1 .1-01 PLS Document AX6AAO4-00030 V-LO-TX-1 6501 (slides 21, 23, 29)

VEGP learning objectives:

LO-LP-37071 -03 State the difference between the following:

a. pressurized thermal shock event

b. pverpressure event

c. cold overpressure event LO-LP-37071 -04 State the actions preventing or mitigating the severity of overcooling and repressurizing LO-PP-16501-01 Describe how COPS provides overpressure protection, including instrument inputs and how the change of temperature will change the lift set point.

LO-PP-1 6501-02 Describe the operation of COPS when the actual pressure exceeds the COPS pressure set point.

Page 126 of 208

Approved By C. S. Waidrup Vogtle Electric Generating Plant Procedure Number Rev Date Approved 119002-c 21

/27/09 ES-O.2 NATURAL CIRCULATION COOLDOWN Page Number 9 of 22 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

_g. Verify closed PRZR Spray Valves:

PV-0455B PV-0455C

_h. Adjust RCP SEAL FLOW CONTROL HC-0182 as necessary to establish 8 to 13 gpm.

_i. Adjust CHARGING FLOW CONTROL FIC-0121 as necessary to establish required Aux Spray flow.

_j. GotoStepl5.

14. D RCS to 1950 psig usj

a. ReturntStep12 PRZR _b. Open PRZR PORV Block I PORY Valve

.alveOPEN.

_c. Open PRZR PORV.

_15. Check RCS pressure LESS THAN _15.

De-energize PRZR Heaters as 1950 psig. necessary.

_Return to Step 12.

_1 6. Stop RCS depressurization.

Printed January 18, 2011 at 12:10

Approved By Procedure Number Rev S. A. Phillips Vogtle Electric Generating Plant 1701 2-1 19.2 Date Approved ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 12 ON Page Number

/5/07 PANEL 1C1 ON MCB 19 of 51 WINDOW C04 ORIGIN SETPOINT A COLD OP 1-TE-0413A 220°F LOW AUCT RCS 1-TE-0423A TEMP 1 -TE-0433A 1 -TE-0443A 1.0 PROBABLE CAUSE Plant cooldown in progress.

2.0 AUTOMATIC ACTIONS NONE 3.0 INITIAL OPERATOR ACTIONS NONE SUBSEQUENT OPERATOR ACTIONS

1. Check Reactor Coolant System Wide Range Hot Leg temperatures less than 220°F.

2. Verify Cold Overpressure Protection System is in service per 1 2006-C, Unit Cooldown To Cold Shutdown.

3. Refer to Technical Specification LCO 3.4.12.

5.0 COMPENSATORY OPERATOR ACTIONS NONE END OF SUB-PROCEDURE

REFERENCES:

1X6AUO1-161, 669, PLS Printed January 18, 2011 at 12:04

Approved By Procedure Number Rev S. A. Phiflips Vogtle Electric Generating Plant 1701 2-1 Date Approved 19.2 ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 12 ON Page Number i0/5/07 PANEL 1C1 ON MCB 25 of 51 WINDOW D04 ORIGIN SETPOINT A RCS PRESS 1 -PB-0405A Pressure is 20 APPROACHES psi below the COLD OP LIMIT PORV lift setpoint 1.0 PROBABLE CAUSE Reactor Coolant System (RCS) pressure rise due to:

a. Loss of letdown,

b. Pressurizer sprays not sufficient,

c. Excessive Pressurizer Heaters in service.

2.0 AUTOMATIC ACTIONS NONE 3.0 INITIAL OPERATOR ACTIONS

1. IF RCS is water solid, check RCS temperatures stable or lowering.

2. Control RCS temperatures to maintain them stable.

4.0 SUBSEQUENT OPERATOR ACTIONS CAUTION With Residual Heat Removal in service, RCS pressure must be maintained below 425 psig.

1. Check Pressurizer Sprays and Heaters and adjust as necessary to reduce RCS pressure.

2. Check letdown and if isolated, reduce charging flow and re-establish letdown.

3. IF Pressurizer Sprays are NOT available, reduce RCS pressure using 1-PIC-0131 on the QMCB.

Printed January 18, 2011 at 12:05

4. RCS Cold Overpressure Mitigation System A. Power relief valve cold overpressure actuation

1. Comparators (PB-403B, P8-4058) PACT - PALL > 0 psi (open)e PACT - PALL -20 psi (close)

2. Pressure 1 m1t function generator (TY41314, TY-423N) (See below)

Low PORV (PCV-455A) High PORY (PCV..456)

Unit 1 Unit 1 Auctioneered Setpoint Auctionured SetDoint Low Temp. (TT-41314) Low Temp. (Tfr4.314)

PSIG PSIG 70 560 70 605 90 560 90 605 140 590 140 635 201 705 201 750 375 705 375 750 485 2335 485 2335 700 2335 700 2335 Low PORV (PCV-455A) High PORV (PCV.416)

Unit 2 Unit 2 Auctioneered Setpoint Auctioneered Setpelat Low T. (Tf .4 1314) Low Temp. (TY-42314)

PSIG PSIG 70 530 70 575 90 530 90 575 140 560 140 605 201 705 201 750 375 705 375 750 485 2335 485 2335 700 2335 700 2335

See General Nate 5 NormalOpcratmgTemperatu,s AX6AAO4-30 SNC VER 51.0

a) ci Cl) cc 0

0 a)

-c Cl)

C) 0 C 0 Cl)

-J D (I) ci)

C) b

)

1 C a)

CI)

Cl)

C C)

U a) C,,)

0 0 0 It) 0 0 0

a)

C) 0 C C Cl) 0 C-) 0.

C)

E im: G) 0 0)

1-0 C 0 0 a)

1-C-)

C,)

iI2II C 0

C,)

a) 0)

d5cI)

C) a)

-c -0J 4.-. 4.-

C) 0=

0 4-Cu 00) cc Co .-c I 0 3 00 0 IIIIU\ cc 0 cd ilifli 4 4 4 **

C U-

Train A COPS (I) P-405 (ACTUAL PRESS)

LOoP 4 PCV455A (TRAIN A)

HI PRESS DEV (ALERT)

(LO (S.,P. PRESS)

(I) LOOP 1 TH (I) LOOP 2 (I) LOOP (I) Loop 4 I S.P. LO TEMP AUCT (ARMING 10 REQUIRED)

TEMP V-LO-PP-16501 Rev 4.0 23

a Vole Un4 Pressure Temperature LAm Reçxit Southern Nuclear Tabe 3-1 cr Vogtle Unrt 1 Data Pokits for the Maximum AIIow the Noimnat COPS PORV SetpoEts (I) 860 a-F 800 z

0 (201,761 Temperature PORV Setpoint 0 (Iegf) (psig) I 750 70 612 700 0

90 612 0

50 140 642 (70,

/

bOO 201 760 550 202 760 500 350 760 460 400 0 50 100 ib) 2 250 3)0 350 AUCTIONEERED LCtØI MEASURED RCS TEMPERPS1URE

)EO F)

V-LO-PP-16501 Rev 4.0 29

HL-16 NRC Written Examination KEY

63. WEO9G2.4.9 001/1/2/NAT CIRC/3.8/4.2 C/A/LOIT BANK/RO/SRO/NRC/GCW Given the following conditions:

- A small break LOCA has occurred.

- The RCPs have been secured.

- 19012, Post LOCA Cooldown and Depressurization, is in progress.

- RCS pressure is 1490 psig.

- Wide Range T-Cold indications are 505°F and slowly lowering.

- Wide Range T-Hot indications are 515°F and slowly lowering.

- Core Exit Thermocouples (CETCs) are 581 °F and stable.

- S/G pressures are 715 psig and stable.

According to 19012-C, which ONE of the following correctly states the status of Natural Circulation and the correct operator actions?

A. Natural Circulation is occurring. Continue attempts to start a RCP.

B. Natural Circulation is occurring. Maintain Steam Dump operation.

C. Natural Circulation is NOT occurring due to stable CETCs. Raise rate of Steam Dump Operation.

D Natural Circulation is NOT occurring due to inadequate subcooling. Raise rate of Steam Dump Operation.

Page 127 of 208

HL-16 NRC Written Examination KEY Feedback E09 Natural Circulation Operations 2.4 Emergency Procedures I Plan 2.4.9 Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.

(CFR: 41.10/43.5/45.13)

K/A MATCH ANALYSIS The question has the candidate determine if Natural Circulation exists with the data given and the proper actions to take to establish it if required.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect-Indications show that Natural Circulation is NOT occurring due inadequate subcooling based on CETCs of 581°F and RCS pressure of 1490 psig (-.596°F). Actual subcooling is 17°F. Per Attachment B of 19012-C, Verification of Natural Circulation subcooling required is 24°F. Attempts to start a RCP is required for PRZR spray.

B. Incorrect-See above A for inadequate subcooling explanation. Steam Dump operation is required, but must be raised to establish Natural Circulation.

C. Incorrect-Natural Circulation is NOT occurring. CETCs per Attachment B of 19012-C, Verification of Natural Circulation must be Stable or Lowering and this is true as stated in the stem.

D. Correct-Natural Circulation is NOT occurring for inadequate Subcooling explanation in A. Per step 32c RNO, if an RCP cannot be started, the rate of Steam Dump operation must be increased.

REFERENCES 19012-C, Post LOCA Cooldown and Depressurization Modified LOIT Bank V-LO-LP-371 12-01 -01 V-LO-LP-370 12-C, Natural Circulation Cooldown VEGP learning objectives:

LO-LP-37012-03 State the operational guidelines used to establish a controlled cooldown using natural circulation.

LO-LP-371 12-01 Using EOP 19012 as a guide, breifly describe how each step is accomplished.D.

Page 128 of 208

Approved By Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant 19012-C 32 Date Approved ES - 1.2 POST-LOCA COOLDOWN AND Page Number 18/1O DEPRESSURIZATION 23of42 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

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

i.E natural circulation NOT established, THEN raise rate of dumping steam using Steam Dumps After natural circulation is verified, maintain rate of dumping steam.

33. Check PRZR Spray status:

_a. RCP4orRCP1 -RUNNING. _a. Start RCP 4 or RCP 1 or other RCP(s) as necessary to provide Normal PRZR Spray using ATTACHMENT A.

_b. Stop RCP(s) NOT required for Normal PRZR Spray.

c. Close PRZR Spray Valve(s) for stopped RCP(s):

RCP 1: PIC-0455C RCP 4: PIC-0455B

d. Open RCP breakers for RCP(s)

NOT running.

Printed January 18, 2011 at 12:34

Approved By J. B. Stanley Vogtle Electric Generating Plant Procedure Number Rev Date Approved 19012-C 32 ES 1.2 POST-LOCA COOLDOWN AND Page Number 2I18I10 DEPRESSURIZATION 37 of 42 ATTACHMENT B Sheet 1 of 1 VERIFICATION OF NATURAL CIRCULATIO N

The following conditions support or indicate natural circulation flow:

RCS subcooling GREATER THAN 24°F [38° F ADVERSE].

SG pressures STABLE OR LOWERING.

RCS WR hot leg temperatures STABLE Q LOWERING.

Core exit TCs STABLE OR LOWERING.

RCS WR cold leg temperatures AT SATURA TION TEMPERATURE FOR SG PRESSURE.

° END OF ATTACHMENT B Printed January 4, 2011 at 08:15

HL-16 NRC Written Examination KEY

64. WEI 1EK2.2 001/1/1/LOSS OF COOLANT RECIJ3.9/4.3 C/A!LOIT BANKJRO/SROINRC/GCW The following Unit 1 conditions exist:

- A primary LOCA outside containment is in progress.

- Reactor Trip and SI have occurred.

- 19111 -C, Loss of Emergency Coolant Recirculation is in effect.

Which ONE of the following choices describes the correct actions to take in 19111-C under these conditions?

A. Start makeup to the RWST from the Boric Acid System.

Terminate ECCS flow with proper RVLIS indication ONLY.

B. Start makeup to the RWST from the Spent Fuel Pool.

Establish one train of ECCS flow and maintain subcooling > 74°F.

C. Start makeup to the RWST from the Reactor Coolant Drain tank.

Terminate ECCS flow with proper RVLIS indication and subcooling > 74°F.

D Start makeup to the RWST from the Spent Fuel Pool.

Establish one train of ECCS flow and maintain subcooling > 24°F.

Feedback Eli Loss of Emergency Coolant Recirculation Knowledge of the interrelations between the (Loss of Emergency Coolant Recirculation) and the following:

(CFR: 41 .7/45.7)

EK2.2 Facilitys heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.

K/A MATCH ANALYSIS The RWST is providing inventory for heat removal and is relied upon for a longer period of time when emergency coolant recirculation is not available. The question tests knowledge that is required to keep the core covered and maintain a cooling medium for the fuel and requirements for possible termination of ECCS.

ANSWER I DISTRACTOR ANALYSIS Page 129 of 208

HL-16 NRC Written Examination KEY A. lncorrect-RWST makeup can be from either the Boric Acid System or the Spent Fuel Pool. ECCS termination requires proper RVLIS indication and subcooling of 74°F.

B. Incorrect-Spent Fuel Pool makeup is acceptable. After establishing one train of ECCS, minimum subcooling requirement is 24 °F.

C. lncorrect-RWST makeup can be from either the Boric Acid System or the Spent Fuel Pool. RCDT is a plausible source of makeup to the RWST but not a proper borated source. Termination of ECCS can only been done with proper RVLIS indication and subcooling greater than 74°F.

D. Correct-RWST makeup from the Spent Fuel Pool is correct and the minimum subcooling is 24 °F.

REFERENCES 19111-C, Loss of Emergency Coolant Recirculation 19112-C, LOCA Outside Containment LOIT Bank WE1 1 EK2.2-O1 P&ID 1X4DB127 and 1X4DB121 VEGP learning objectives:

LO-LP-371 14-12 State the intent of EOP 19111, Loss of Emergency Coolant Recirc Page 130 of 208

Procedure Number Rev Approved By C S Waidrup Vogtle ElectrtcGenerating Plant 19111-C 33 Page Number Date Approved ECA-1 .1 LOSS OF EMERGENCY COOLANT 1 of 48 i25Il0 RECIRCULATION EMERGENCY OPERATING PROCEDURE CONTINUOUS USE PURPOSE ant recirculation capability, to delay This procedure provides actions to restore emergency cool

, and to depressurize the RCS depletion of the RWST by adding makeup and reducing outflow to minimize break flow. (Applicable in Modes 1, 2, and 3.)

ENTRY CONDITIONS ANT

19010-C, E-1 LOSS OF REACTOR OR SECONDARY COOL

19013-C, ES-1.3 TRANSFER TO COLD LEG RECIRCULATION

19112-C, ECA-1.2 LOCA OUTSIDE CONTAINMENT

19005-C, ES-0.0 REDIAGNOSIS SSURE

19251-C, FR-Z.1 RESPONSE TO HIGH CONTAINMENT PRE MAJOR ACTIONS

Continue attempts to restore ECR.

Increase/Conserve RWST level.

Initiate cooldown to cold shutdown.

Depressurize RCS to Minimize RCS subcooling.

Try to add makeup to RCS from alternate source.

Depressurize SGs to cool down and depressurize RCS.

Maintain RCS heat removal.

Printed January 5, 2011 at 10:12

Approved By Procedure Number Rev

..Vogtle Electric Generating PIant.:

C. S. Waidrup 19111-C 33 Page Number Date Approved ECA-1 .1 LOSS OF EMERGENCY COOLANT 1/25/10 RECIRCULATION 8of48 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

8. Check if Containment Spray should be aligned for recirculation:

_a. CS Pumps - RUNNING. a. Go to Step 9.

b. Containment Emergency Sump b. WHEN Containment levels GREATER THAN OR Emergency Sump level EQUAL TO 13.5 INCHES: indicators Ll-764 or LI-765 greater than or equal to LI-764 13.5 INCHES, THEN perform Step 8c.

-OR-Continue with Step 9.

Ll-765

_c. Initiate ATTACHMENT B, CONTAINMENT SPRAY SWITCHOVER FROM INJECTION TO RECIRCULATION.

MRnJI Printed January 5,2011 at 10:12

Procedure Number Rev Approved By C. S. Wakirup VogtleEIec nc Generating Plan 19111-C 33 Page Number ANT ECA-1.1 LOSS OF EMERGENCY COOL 9 of 48

/25/10 RECIRCULATION RESPONSE NOT OBTAINED ACTION/EXPECTED RESPONSE

10. Check intact SG levels:

_a. H all SGs NR levels less

_a. NR level AT LEAST ONE than 10% [32% ADVERSE],

GREATER THAN 10% [32% THEN maintain total feed ADVERSE]. flow greater than 570 gpm.

_b. Maintain NR level between 10%

[32% ADVERSE] and 65%.

c. Go to Step 11.

_c. NR level ANY RISING IN AN UNCONTROLLED MANNER.

_d. Stop feed flow to that SG.

_*1 1. Swap to alternate CST by

11. Check CST level GREATER THAN initiating 13610, AUXILIARY 15%. FEEDWATER SYSTEM.

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 100°F/HR.

c. Dump steam from intact

_c. Dump steam to Condenser from SO(s) using SG ARV(s).

intact SG(s) using Steam Dumps.

IF no intact SG(s) available, THEN use faulted SG(s).

Pnnted January 5, 2011 at 10:12

Procedure Number Rev Approved By C. S. Waidrup ... Vogtle Electric Generating Plant 19111-C 33 Page Number Date Approved ECA-1 .1 LOSS OF EMERGENCY COOLANT 4/25/l0 10 of 48 RECIRCULATION ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE When the low steamline pressure SI/SLI signal is blocked, main steamline isolation will occur if the high steam pressure rate setpoint is exceeded.

13. Check if low steamline pressure SIISLI should be blocked:

_a. Steam Dumps AVAILABLE.

_a. Go to Step 14.

_b. PRZR pressure LESS THAN -

_b. WHEN PRZR pressure is 2000 PSIG. less than 2000 psig and the high steam pressure rate alarms are clear, THEN block low steamline pressure SI/SLI by

) performing step 13d.

Go to Step 14.

_c. High steam pressure rate alarms

- CLEAR.

d. Block low steamline pressure Sl/SLI using the following:

. HS-40068 HS-40069 Printed January 5, 2011 at 10:12

Procedure Number Rev 4,:

-Vogtle Electric GeneratingPlant Approved By ..

19111-C 33 C. S. Waidrup Page Number J Date Approved ECA-1.1 LOSS OF EMERGENCY COOLANT 11 of 48 125/10 RECIRCULATION ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

14. Check if ECCS is in service: _14. GotoStep24.

CCPs ANY RUNNING.

-OR-

_BIT NOT ISOLATED.

-OR

_RHR Pumps ANY RUNNING IN INJECTION MODE.

_a. StartorstopaCCPto establish only one Pump running.

_b. Start or stop an SI Pump to establish only one Pump running.

_c. Stop RHR Pumps.

Go to Step 16.

_d. Start or stop an RHR Pump to establish only one Pump running.

Printed January 5, 2011 at 10:12

Procedure Number Rev Approved By C. S. Wadrup I *VogtIe Electric Generating Plant 19111C 33 Page Number LDate Approved ECA-1 .1 LOSS OF EMERGENCY COOLANT 12 of 48 S./25/1O RECIRCULATION ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

16. Check no backflow from RWST to Containment Emergency Sumps:

a. CNMT SUMP TO RHR PMP-A _a. RHR Pump A is running SUCTION CLOSED: and Containment Emergency Sump level is less than 13.5 INCHES,

HV-8811A THEN stop RHR Pump A.

Close RWST TO RHR PMP-A SUCTION HV-8812A.

b. CNMT SUMP TO RHR PMP-B _b. H RHR Pump B is running SUCTION CLOSED: -

and Containment Emergency Sump level is HV-8811B less than 13.5 INCHES, THEN stop RHR Pump B.

Close RWST TO RHR PMP-B SUCTION HV-881 2B.

17. Check if an RCP should be started:

_a. All RCPs STOPPED. -

a. Stop all but RCP 4 or RCP 1.

Close Spray Valve for idle RCP:

RCP 1: PIC-0455C RCP 4: PIC-0455B Go to Step 18.

Step 17 continued on next page Printed January 5, 2011 at 10:12

t .

Approved By . Procedure Number Rev C S Waidrup Vogtle Electric Generating Plant 19111-C 33 Date Approved Page Number ECA-1.1 LOSS OF EMERGENCY COOLANT

ç4/25/10 RECIRCULATION 13 of 48 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

b. Close Spray Valve for idle RCP:

RCP 1: PIC-0455C RCP 4: PIC-0455B Go to Step 18.

_c. Start RCP 4 or RCP 1 or other RCP(s) as necessary to provide Normal Spray using ATTACHMENT D.

d. Close Spray Valve for idle RCP:

RCP 1: PIC-0455C RCP 4: PIC-0455B

18 Check if ECCS can be terminated:

_a. Applicable RVLIS indication: _a. GotoStep24.

RCP(s) running Required Indication 0 Full Range greater than 63%

1 Dynamic Range greater than 25%

2 Dynamic Range greater than 34%

3 Dynamic Range greater than 50%

4 Dynamic Range greater than 72%

Step 18 continued on next page Printed January 5, 2011 at 10:12

Approved By Procedure Number Rev C. S. Waidrup Vogtle Electric Generating Plant 1911 1-C 33 Page Number Date Approved ECA-1.1 LOSS OF EMERGENCY COOLANT q /25/1 0 RECI RCULATION RESPONSE NOT OBTAINED 14 of 48 ACTION/EXPECTED RESPONSE

b. RCS Subcooling basdón Core: b. Establish minimum ECCS Exit TCs GREATER THAN 74°F flow to remove decay heat

[88°F ADVERS by performing the following:

1) Determine minimum ECCS flow required using the following:

TABLE 1 or FIGURE 1

_2) Throttle ECCS flow to minimum value.

_3) Go to Step 24.

CAUTION Repositioning Phase A Isolation Valves may cause radiation problems throughout the plant.

_1 9. Reset Containment Isolation Phase A.

20. Establish Instrument Air to Containment:

_a. Instrument Air pressure - _a. Start additional Air GREATER THAN 100 PSIG. Compressors as necessary.

_b. Open INSTR AIR CNMT ISO VLV HV-9378.

_c. Verify PRZR Spray Valves operating as required.

Printed January 18, 2011 at 13:13

HL-16 NRC Written Examination KEY

65. WEI2EK2.1 OO1/1/IIMSL RUPTURE/3.4/3.7 MEM!MOD BANK WOLF CRKJRO/SRO/NRC/GCW During the performance of 19121-C, Uncontrolled Depressurization Of All Steam Generators, the following conditions exist:

- RCS cooldown rate is determined to be 125 °F/HR.

- All SG NA levels are off-scale low.

Which one of the following describes how the crew is directed to control AFW flow?

A Flow is reduced to 30 gpm to each SG and WR Hot Leg temperatures are monitored to ensure secondary heat sink Es maintained.

B. Flow is terminated to all but a single SG, which is fed at 30 gpm, and WR Cold Leg temperatures are monitored for conditions that may result in Pressurized Thermal shock.

C. Total flow is maintained > 570 gpm until ANY SG NR level is> 10%, and WR Hot Leg temperatures are monitored to ensure secondary heat sink is maintained.

D. Total flow is maintained > 570 gpm until ANY SG NR level is> 10%, and WR Cold Leg temperatures are monitored for conditions that may result in Pressurized Thermal shock.

Page 131 of 208

HL-1 6 NRC Written Examination KEY Feedback E12 Uncontrolled Depressurization of all Steam Generators Knowledge of the interrelations between the (Uncontrolled Depressurization of all Steam Generators) and the following:

(C FR: 41.7/45.7)

EK2.1 Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

K/A MATCH ANALYSIS The controlling of the Safety Related system of AFW flow is important due to the cooldown affect of the depressurization of all SGs. The student must know the value to minimize cooldown and establish a secondary heat sink.

ANSWER I DISTRACTOR ANALYSIS A. Correct-With RCS cooldown rate >100 °F/HR AFW flow is lowered to 30 gpm/SG and maintained at 30 gpm each until NR level is> 10%. WR hot Leg Temps are monitored and feed flow controlled as necessary to stabilize temperature.

B. Incorrect-Plausible because flow is initiated at 30 gpm to only one SG in 19231-C, Loss of Secondary Heat Sink.

C. Incorrect-570 gpm is plausible because this flow rate is maintained normally when NR level is < 10% is 19000-C, Reactor Trip or SI, 19001-C, Reactor Trip Response and other EOPs for normal Heat Sink parameters. WR Hot leg monitoring is correct.

D. lncorrect-570 gpm is plausible from C above. WR CL temperatures are monitored for the arming of COPS and not PTS.

REFERENCES 19121-C, Uncontrolled Depressurization Of All Steam Generators Wolf Creek-August-2009 NRC WE12EK2I VEGP learning obiectives:

N/A Page 132 of 208

Approved By .

C. S. Wadrup Vogtle Electric Generating Plant Procedure Number Rev Date Approved

.: 19121-C 26 ECA-2.1 UNCONTROLLED DEPRESSURIZAT Page Number 2/16/10 ION OF ALL STEAM GENERATORS 7 of 41 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED

4 Control feed flow to minimize RCS cooldown:

a. Monitor shutdown margin by initiating 14005, SHUTDOWN MARGIN AND KEFF CALCULATIONS.

LorTeeqwto3q:gpJj)q EHA qq*h tEfl3I IAN&

I tmprtUrs -STABLE(

LEi1J

5 Check if RCPs should be stopped:

a. ECCS Pumps AT LEAST ONE

a. GotoStep6.

RUNNING:

. CCPorSIPump Step 5 continued on next page Printed January 5, 2011 at 10:14

Wolf Creek 2009 NRC Examination Facility:

Vendor: WEC Exam Date: 08/2009 Exam Type: R Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 1 K/A# E12 EK2.1 Importance Rating 3.4 Knowledge of the interrelations between the (Uncontrolled Depressurizat ion of all Steam Generators) and the following: Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and autom atic and manual features.

Proposed Question: RO Question # 43 During the performance of EMG C-21, Uncontrolled Depressurization of All Steam Generators, the following conditions exist:

RCS cooldown rate is determined to be 165° F/Hr.

All SG NR levels are off-scale low.

Total AFW flow is 300,000 Ibm/hr.

Which ONE of the following describes how the crew is directed to contro l AFW flow?

Flow is reduced to 30,000 Ibm/hr to each SG, and Thot is monitored to ensure secondary heat sink is maintained.

B. Flow is terminated to all but a single SG, which is fed at 30,000 Ibm/hr

, and Tcold is monitored for conditions that may result in Pressurized Thermal Shock.

C. Total flow is maintained >270,000 Ibm/hr until ANY SG narrow range level is >6%, and Thot is monitored to ensure secondary heat sink is maintained.

D. Total flow is maintained >270,000 Ibm/hr until ALL SG narrow range levels are >6%,

and Tcold is monitored for conditions that may result in Pressu rized Thermal Shock.

Proposed Answer: A 113 Post NRC Review A 8/7109

Wolf Creek 2009 NRC Examination Explanation (Optional):

A. Correct. See EMG C-21 step 5 and basis B. Incorrect. Plausible because flow is initiated to only I SG in EMG FR-HI.

C. Incorrect. Plausible because this flow is maintained under these conditions in EMG E-0 or in EMG C-21 if RCS cooldown rate is <100°F/Hr.

D. Incorrect. Plausible because second half is true, but with RCS cooldown rate

>100°F/hr, AFW flow is minimized.

Technical Reference(s): EMG C-21, Rev 17, Step 5 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: L01732334 R2 R3 (As available)

Question Source: Bank #

Modified Bank # WTSI 52615 (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergenc y operating procedures for the facility.

Comments:

WTSI 52615 From VC Summer Audit 2006. There are items in our bank with similar context.

114 Post NRC Review A 8/7/09

HL-16 NRC Written Examination KEY

66. G2. 1.1 OOI/3/N/AJCOND OF OPSI3.8/4.2 MEMJNEWIRO/SRO/NRC/GCW The On-Duty Shift Manager is required to notify the Operations Manager On-Call (OMOC) promptly of significant plant events.

Which ONE of the following would (1) require the notification to the OMOC and (2) the minimum requirements of the individuals fulfilling the OMOC function?

A. (1) Apprentice Mechanic has been hurt while working in the Maintenance Shop.

(2) Must be knowledgeable at the SRO level.

B. (1) Chemistry reports Secondary Chemistry Action Level 2.

(2) Must have a current or formerly held an SRO license.

C (1) A missed Tech Spec surveillance or test.

(2) Must be knowledgeable at the SRO level.

D. (1) A reportable event per 10CFR.

(2) Must have a current or formerly held an SRO license.

Page 133 of 208

HL-16 NRC Written Examination KEY Feedback 2.1 Conduct of Operations 2.1.1 Knowledge of conduct of operations requirements.

(CFR: 41.10/45.13)

K/A MATCH ANALYSIS Knowledge of what is required to be reported to Ops Management is directed by the Conduct of OPS Admin Procedure.

ANSWER / DISTRACTOR ANALYSIS A. Incorrect-Personnel injury notification is for Operations personnel only. SRO knowledge is required.

B. Incorrect-Secondary chemistry action level 2 or greater is reportable but the requirement for the current or formerly held SRO is incorrect.

C. Correct D. Incorrect-A reportable event per 1 OCFR is correct but the requirement for the current or formerly held SRO is incorrect.

REFERENCES NMP-OS-007-OO1 Conduct of Operations Standards and Expectations (Version 6.0)

VEGP learning obiectives:

LO-LP-63404-02 Describe the following as applicable to the surveillance test program:

-failure of surveillance tests Page 134 of 208

Southern Nuclear Operating Company SOUTHERNaZS Nuclear Management i Conduct of Operations I NMP-OS-007-001 Version 8.0 COMPANY Standards and Expectations i Instruction Page 1 of 49 Instruction Owner: David Vineyard I Fleet Operations Manager / SNC Corporate (Print: Name I Title I Site)

Approved By: Original signed by David Vineyard I 12/06/2010 (Instruction Owners Approval Signature) (Approval Date)

Effective Dates: NA 12/07/2010 12/07/2010 12/07/2010 Corporate FNP HNP VEGP PRB Review Not Required PROCEDURE USAGE REQUIREMENTS SECTIONS Procedure must be open and readily available at the Continuous Use: work location. Follow procedure step by step unless otherwise_directed_by_the_procedure.

Reference Use: Procedure or applicable section(s) available at the work location for ready reference by person performing steps.

Information Use: Available on site for reference as needed. X

Southern Nuclear Operating Company I

F NMP-OS-007-001 Conduct of Operations Version 8.0 Standards and Expectations Page 5 of 49 cted or uncertain for the stored energy within the reactor core. Faced with unexpe safe condition and do not hesitate to conditions, operators place the plant in a known, reduce power or shut down the reactor.

6.1.2 Expectations 6.1.2.1 Safety First Operators do not Safe operation takes precedence over all other considerations.

r safety is challenged.

hesitate to reduce power or shut down the reactor if nuclea ge plant safety or Operators recognize degraded conditions that could challen reliability.

Licensed operators comply with Technical Specifications.

rs avoid hasty Because the need for a time-critical response is infrequent, operato decisions. When faced with time-critical decisions, operato rs:

Utilize alternate indications to validate information ise

Assume the available indications are valid until proven otherw

Develop contingency actions, if time allows

Use all available resources, including people offsite, if necessary

Do not proceed in the face of uncertainty

Do not allow production or cost to override safety

Do not challenge the safe operating envelope 6.1.2.2 Conservative Approach to clearly

Information is gathered and analyzed from relevant sources in order Short and define and provide options for resolution of operational concerns.

with long-term risks, consequences, and the aggregate impact associated decision options are critically and objecti vely consid ered.

plans that

When addressing operational concerns, operators implement enhance include contingencies and compensatory measures to mainta or in safety. Decision-makers and their roles and responsibilities are clearly identified. Command and control responsibilities are carried out in accordance with approved procedures. The bases for decisions are significant communicated throughout the organization. The effectiveness of operational decision-making is periodically evalua ted.

whenever

Human performance tools, including team involvement, are used ns.

practicable to avoid inappropriate actions when reaching operating decisio 6.2 Operations Manager On-Call 6.2.1 Standard to improve The on-duty Shift Manager consults with the management team in order the on-dut y Shift Manag er notifies the the quality of operational decisions. To this end, cant plant events .

Operations Manager on Call (OMOC) promptly of signifi

Southern Nuclear Operating Company Nuclear Management i Conduct of Operations NMP-OS-007-001 SOUTHERN COMP Instruction i Standards and Expectations Version 8.0 Page 7 of 50 6.2.2 Expectations 6.2.2.1 OMOC Responsibilities (may be coincident with EP on call functions)

An individual, assigned by the Operations Manager, is on-call to assist the duty Shift Manager in the resolution of problems.

i neinaiviauai assignea .5 ,JiVILJIJ TOr.tfle: sge;is or ablefam t!oji.ar 6.2.2.2 OMOC Notification The OMOC serves as a resource for the Shift Manager to use to discuss operational decisions and insights from an Operations Department leadership point of view. The off shift operations leadership may have insights that are not known to the Shift Manager regarding overall plant conditions. The duty Shift Manager should contact the OMOC regarding conditions or events that are deemed to be of interest to management. If in doubt, the duty Shift Manager is encouraged to notify the OMOC.

In addition to notifications required specifically by procedure, the duty Shift Manager notifies OMOC without delay if any of the following occur:

Unplanned Reactor Protection System or Engineered Safeguard Function actuation ealp

Unplanned entry into a Technical Specification LCO Action Statement of less than 7 days

Valid entry into an abnormal or emergency procedure

Unplanned plant transient or load reduction

Oufof-specificati6ij.,chemistry tLat could result in signir3jte damage (action level 2 or greater)

Significant radiological event

Unplanned condition that results in a heightened risk (orange or red) jo CFR 5072 o 5073,

Missed technical specification surveillance or tej

Valid entry into the site Emergency Plan

Reactivity incident or event

Consequential component mispositioning

Significant human performance error involving Operations personnel 6.3 Performance Monitoring 6.3.1 Standard Operations personnel reinforce desired behaviors to optimize individual and team performance.

HL-16 NRC Written Examination KEY

67. G2. 1.23 OO1/3/N/AJCOND OF OPS/SYSTEMS/4.3/4.4 MEMJLOIT BANKJRO/SRO/NRC/GCW UOP 12004-C, POWER OPERATION is in effect.

The transfer from Auxiliary Feed Water to Main Feed Water step is in progress.

The Bypass Feed Isolation Valve (BFIV) is opened while the Main Feed Isolation Valve (MFIV) is verified closed. The flowpath of main feedwater to the selected steam generator as the Bypass Feed Reg Valve (BFRV) is opened is:

A. established because the MFIV leak-by is great enough to supply SG flow needs at low power through the BFRV.

B established since the BFRV and BFIV are in series and not isolated by the MFIV.

C. not established, because the MFIV isolates flow to both the BFRV and the MFRV.

D. not established as the MFIV must be opened before flow can proceed to the SG, as the BFIV and MFIV are in series.

Page 135 of 208

HL-16 NRC Written Examination KEY Feedback 2.1 Conduct of Operations 2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation.

(C FR: 41.10/43.5/45.2/45.6)

K/A MATCH ANALYSIS The transfer of water sources to the SGs is an integrated operation involving multiple valves and various level and flow indications. The match is there due to involvement of multiple systems and the complexity of the task.

ANSWER / DISTRACTOR ANALYSIS A. INCORRECT-The valves that are known to leak by are the BFRVs when the BFIVs are opened and not the MFIVs.

B. CORRECT C. INCORRECT-The MFIVs are not in the flow path when the BFIVs and BFRVs are being used.

D. INCORRECT-The MEl Vs are not in the flow path when the BFIVs and BFRVs are being used. The MFIVs and BFIVs are in parallel.

REFERENCES 12004-C, Power Operation (Mode 1) Steps 4.1.7 and 4.1.8 (rev 89)

P&ID 1X4DB168-1 VEGP learning obiectives:

LO-LP-61202-04 Describe the steps required to be performed to shift from the AFW System to the bypass feed regulating valves.

Page 136 of 208

Procedure Number Rev Approved By J. B. Stanley I

I Vogtle Electric Generating Plant A 112004-C I Page Number 89 Date Approved I POWER OPERATION (Mode 1) 1 of 89 I24/2010 I POWER OPERATION (Model)

S- SECTIONS PROCEDURE USAGE REQUIREMENT ilable ALL except Continuous Use: Procedure must be open and readily ava sections 4.3.1, step by at the work location. Follow procedure 4.3.2, and 4.3.3 step unless otherwise directed.

ilable at Reference Use: Procedure or applicable section(s) ava 4.3.1, 4.3.2, and by person the work location for ready reference 4.3.3 performing steps.

needed. NONE Information Use: Available on plant site for reference as Printed November 30, 2010 at 11:36

rcNmber8Rv Vogtle Electric Generating Plant J( aey Date Approved 24/2010 POWER OPERATION (Mode 1)

Page Number 9 of 89 INITIALS e a Main Feed Pump in 4.1.7 At approximately 4% reactor power, plac service as follows:

(MFRV) and the

a. Verify the Main Feed Regulating Valve

) Controllers are in Bypass Feed Regulating Valve (BFRV manual with the valves closed.

NOTE Tavg with rod motion when changing Maintain Tavg at approximately no load steam demands.

ice per 13615

b. Verify two condensate pumps are in serv Condensate And Feed Water System.

15, Condensate

c. Start a Main Feed Pump (MFP), per 136 And Feed Water System.

trip to MDAFW

d. Restore AFWAS circuitry from MFPT(s) pumps by performing the following:

(1) Verify at least one MFPT reset.

owing hand (2) Remove Caution tags and align the foll switches as follows:

UNIT 1: Train A: I -HS-5344 in NORMAL IV (Aux Relay Pnl I NCPAR2)

Train B: 1-HS-5345 in NORMAL IV (Aux Relay PnI I NCPAR4)

UNIT 2: Train A: 2-HS-5344 in NORMAL IV (Aux Relay PnI 2NCPAR2)

Train B: 2-HS-5345 in NORMAL IV (Aux Relay Pnl 2NCPAR4)

TO START (3) Verify Annunciator ALB16FO5, AFW AU t.

MFPT TRIP RLY CNTRL PWR LOSS, rese ves are closed.

e. Verify the Main Feed Water Isolation Val 1)

(1985303297, 1985304988, 199132152 Printed November 30, 2010 at 11:36

aey Vogtle Electric Generating Plant A Page Number Date Approved (2/24I2olo POWER OPERATION (Mode 1) 10 of 89 INITiALS NOTES

BFRVs may leakby as the BFIV opens, SG levels should be monitored and AFW flow adjusted to maintain SG levels constant.

The following step should be completed prior to continuing with Step 4.1.8.

f. Open the Bypass Feed Isolation Valve for all SGs one at a time. (1985303297, 1985305760, 1985304988, 1991321521)

SG1 HV-15196

SG2 HV-15197

SG3 HV-15198

SG4 HV-15199 Printed November 30, 2010 at 11:36

aey Vogtle Electric Generating Plant A rNber8Rv Page Number Date Approved 11 of 89 2/24/2010 POWER OPERATION (Mode 1)

INITIALS 4.1.8 Transfer from Auxiliary Feed Water to Main Feed Water, one Steam Generator at a time, by performing the following:

(1985303296)

NOTE points UF-5404, Total feed flow (AFW plus BFRV flow) can be monitored via IPC UF-5424, UF-5444, UF-5464.

CAUTION mini-flow During AFW forward flow operations of less than 150 gpm, correct d

valve positions must be maintained. The mini-flow should be checke frequently.

a. Verify MFPT Delta P is maintained between 100 and 150 psid:

SG1

SG2

SG3

SG4

b. Stabilize the SG NR level between 60% and 70%, and note total feed flow:

SG 1 Feed Flow

SG 2 Feed Flow

SG 3 Feed Flow

SG 4 Feed Flow Printed November 30, 2010 at 11:36

aey VogUe Electric Generating Plant A rd;Nnber8Rv Page Number Date Approved

-2/24I201O POWER OPERATION (Mode 1) 12 of 89 INITIALS

c. Slowly open the BFRV and verify a slight increase in feedwater flow:

SG I LV-5243

SG 2 LV-5244

SG 3 LV-5245

SG 4 LV-5242

d. Close the Auxiliary Feed Water Supply Valve:

SG 1 HV-5139

SG2 HV-5132

SG3 HV-5134

SG4 HV-5137

e. Adjust the BFRV to re-establish total feed flow as noted in Step b.:

SG 1 LV-5243

SG 2 LV-5244

SG 3 LV-5245

SG 4 LV-5242

f. Stabilize SG level and then place the BFRV controller in automatic: (1985306829)

SG 1 LIC-550

SG 2 LIC-560

SG 3 LIC-570

SG 4 LIC-580 Printed November30, 2010 at 11:36

rd57ber8RY t

Vogtle Electric Generating Plan Page Number Laey 13of89 Oate Approved 24/2010 POWER OPERATION (Mode 1) INITIALS Initial for SG just swapped and rep eat Steps a. through f.

g. ators:

for the remaining Steam Gener

SGI

SG2

SG3

SG4 with the idle MFPT

h. Check the MFPT is feeding forward n 6d, AFW auto tripped, and exit LCO 3.3.2 Functio actuation capability from the trip of two main feed pumps.

NOTE ry Fee d Wa ter Sy ste m mu st be placed in standby readiness prior to The Auxilia wer (Mode I entry).

exceeding 5% Rated Thermal Po e placing the system in Stop the AFW pumps and initiat xiliary Feed Water Standby readiness per 13610, Au

5)

System. (1984301714, 198430171 MFP has commenced.

j. Notify Chemistry that feeding via a 4.1.9 Coordinate with Chemistry to:

space sample is in

a. Verify that the Pressurizer Steam service.

on monitor RE-0724 in

b. Initiate actions to place N-16 radiati service.

6 Printed November 30, 2010 at 11:3

Date: 1/5/2011 Time: 09:49:37 AM L_

C-)

cWW X-ioD CX-i) ujC 0570 OLa.

1A(i 34.0 D

COO4D BLEED TifI* r:\flATA\fln([Jmentum\\Jjewed\1 X4DB1 68-3.cal

HL-16 NRC Written Examination KEY

68. G2.2.13 001/3/N/A/C & T/4.I/4.3 MEM/BANK-FIL--I5RJRO/SROINRC/GCW A fail open air operated valve (AOV) with a handwheel must be tagged shut as a boundary point for a clearance.

To meet NMP-AD-003-002, Tagging Standards:

1) The handwheel is required to be

2) The air supply valve is 1 2 A. in the closed position required to be isolated and vented B in the closed position not required to be isolated and vented C. in the open position required to be isolated and vented D. in the open position not required to be isolated and vented Page 137 of 208

HL-16 NRC Written Examination KEY Feedback 2.2 Equipment Control 2.2.13 Knowledge of tagging and clearance procedures.

(CFR: 41.10/45.13)

K/A MATCH ANALYSIS The question presents a plausible scenario where a fail open AOV is to be tagged closed. The question straight forward asks the tagging requirements for this valve in accordance with NMP-AD-003.

ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Handwheel must be closed is correct, however, the air supply does required isolation and venting. The handwheel is adequate to maintain the valve closed.

B. Correct. The handwheel must be closed, the air is not required to be closed and vented.

C. Incorrect. The valve handwheel must be closed, the air does not require to be isolated and vented.

D. Incorrect. The valve handwheel must be closed, the air does ii require to be isolated and vented.

REFERENCES G2.2.13-03 from LOIT exam bank. This question was previously used on the HL-15R NRC RO exam.

NMP-AD-003-002, Tagging Standards step 6.15.8.3 VEGP learning obiectives:

LO-LP-63304-15, Given a tagout involving mechanical and electrical tagout points, describe proper tag placement requirements for the following:

Valves Page 138 of 208

Southern Nuclear Operating Company Nuclear NMP-AD-003-002 SOUTHERN Management Tagout Standards Version 6.0 COMPANY E. yr,W Instruction Page 11 of 13 6.15.8.2 When using a fail closed AOV with out a hand wheel as a boundary point the following items shall be Danger tagged when applicable:

Local and/or remote control switches in the Closed position

The air supply valve closed Air must be vented off of the valve operator.

6.15.8.4 When using a fail open AOV without a hand wheel as a boundary point the following items shall be tagged when applicable:

Local and/or remote control switches in the Closed position.

Mechanically or hydraulically (as appropriate) gag the valve in the closed position and Danger tag the gagging device.

6.15.8.5 When using a fail as is AOV in the Condensate Demineralizer System as a boundary point, perform the following:

Local and/or remote control switches in the Closed position.

6.15.8.6 When tagging an AOV for purposes other that use as a boundary valve, such as flow Scan testing, valve setup or other AOV actuator work, the air supply isolation valve need not be tagged.

6.15.9 Freeze Seals 6.15.9.1 When other means of isolation are not available, freeze seals may be used for fluid isolation provided the following requirements are met:

The use of a freeze seal as a boundary point is documented in the Holder Instructions section of the Tagout

The Tagout contains steps to install the freeze seal per the approved site process at the appropriate step in the sequence (normally before venting and draining of the system)

During the period when the freeze seal is in place, the Tagout boundary should be checked periodically for leakby out the drains/vents

The Tagout contains steps to remove the freeze seal per the approved site process at the appropriate step in the sequence (normally the last boundary point unisolated, allowing time for the freeze seal to thaw, time dependent on ambient temperatures)

During the period when the freeze seal is thawing, the piping within the original boundary should be checked periodically for leaks 6.15.10 Check Valves 6.15.10.1 Stop check valves may be used as a Tagout point by tagging its actuator in the Closed position

HL-16 NRC Written Examination KEY

69. G2.2. 14001 /3/N/AJEQUTP CONFIQ/3 .9/4.3 MEM/NEW/RO/SRO/NRC/GCW Procedure 10000-C, Operations Administrative Controls has requirements for Tracking Configuration of Repositioned Components.

A component that is procedurally manipulated from the Control Room to an Off Normal configuration is expected to remain in that position past shift turnover time.

Which of the following is CORRECT concerning how this component is tracked in order to return it to its normal alignment?

A. The only requirement is the component is Caution Tagged stating the reason for its repositioning.

B. The only requirement is the controlling procedure is placed into the Procedures in Progress book in the Control Room.

C. The component is Cautioned Tagged stating the reason for its repositioning and flagged with a Pink Off Normal Component flag.

D The component should be flagged with a Pink Off Normal Component flag and controlling procedure placed into the Procedures in Progress book in Control Room.

Page 139 of 208

HL-16 NRC Written Examination KEY Feedback 22 Equipment Control 2.2.14 Knowledge of the process for controlling equipment configuration or status.

(CFR: 41.10/43.3/45.13)

K/A MATCH ANALYSIS Mispositioned components have been a major issue on site and direct guidance has been given in the Conduct of Operations procedure for controlling this problem. Direct match for the KA.

ANSWER I DISTRACTOR ANALYSIS A. INCORRECT-Caution Tag is used when there is no existing administrative control (e.g., procedure or clearance). In this case, the component was procedurally manipulated.

B. INCORRECT- Procedures in Progress book is correct but not the only requirement. The component must be flagged with the Off Normal Component ring.

C. INCORRECT-The Off Nomal Component flag is used, but the Caution Tag is not required.

D. CORRECT REFERENCES 10000-C, Operations Administrative Controls Step 4.5 VEGP learning obiectives:

LO-LP-63500-32 State the standard for equipment status control.

Page 140 of 208

Approved By J. B. Staney Vogtle Electric Generating Plant 4 Procedure Number Rev Date Approved 10000-c 91 Page Number 10/24/2010 OPERATIONS ADMINISTRATIVE CONTROLS 31 of 49 4.5 TRACKING CONFIGURATION OF REPOSITIONED COMPONENTS 4.5.1 IF the repositioning of a component is a compensatory action for a degraded SSC, instead of an alteration in support of maintenance or performance of a procedure, then a 10 CFR 50.59 screening should be perfo rmed.

For example, if an upstream isolation valve is closed to allow a leaking pressure gauge to be repaired or replaced then closing the upst ream isolation valve would be considered an alteration in support of maintenance. How ever isolation valve is closed to maintain system functionality, then , if the same that would be considered a compensatory action for a degraded SSC and would be subject to locFR5o .59 screening.

4.5.2 IF lantcomponentisma 4%

an its jgçrit, e

nfturation wilL tZJJ!J (1) This includes systems which are intended to be continuo usly in service, such as battery room fans, even if they are shutdown per the applicable SOP. This will ensure that their status is tracked and they are returned to service in a timely manner.

(2) In most cases one flag placed on a conspicuous com ponent in the control Room is sufficient. When the system in ques tion has no components in the Control Room, the associated proc edure in progress along with turnover sheet information will be used to track the status of the system.

Printed January 5, 2011 at 10:38

HL-16 NRC Written Examination KEY

70. G2.3. 15 002/3/N/A/RMS-KNOWLEDGE OF/2.9/3. 1 MEM/NEWIRO/SRO/NRC/GCW Per 00930-C, Radiation and Contamination Control, when exiting Contaminated Areas, hand carried tools and materials (e.g., clipboards, notebooks, pencils, etc.) shall be frisked for 1 contamination.

Personnel decontamination should be performed when the results of personnel monitoring exceed 2 above background.

A 1) Beta-Gamma

2) 100 cpm B. 1)Alpha

2) 50 cpm C. 1) Beta-Gamma 2)1000 dpm/lOOcm 2

D. 1) Alpha

2) 20 dpm/lOOcm 2

Page 141 of 208

HL-16 NRC Written Examination KEY Feedback 2.3 Radiation Control 2.3.15 Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

(CFR: 41.12/43.4! 45.9)

K/A MATCH ANALYSIS A direct match for the KA requiring the knowledge to know what type of radiation is being monitored and to know what limits require personnel to be decontaminated.

ANSWER / DISTRACTOR ANALYSIS A. Correct-Beta-Gamma is correct. Gamma is the most common dose received on site. 100 cpm above background is correct per the procedure.

B. Incorrect-Alpha is incorrect. 50 cpm is plausible because this is the setpoint for Fixed Surface Contamination for Alpha.

C. Incorrect-Beta-Gamma is correct. 1000 dpm/lOOcm2 is plausible because this is the setpoint for Loose Surface Contamination for Beta-Gamma.

D. Incorrect-Alpha is incorrect. 20 dpm/lOOcm2 is plausible because this is the setpoint for Loose Surface Contamination for Alpha.

REFERENCES 00930-C Radiation and Contamination Control V-LO-LP-63930-C Radiation and Contamination Control VEGP learning obiectives:

LO-LP-63930-1 1 State the frisking requirements necessary when leaving the RCA.

Page 142 of 208

Procedure Number Rev Vogtle Electric Generating PIant Approved By 26 00930-C C.R.Dedrickson ,

Page Number RADIATION AND CONTAMINATION CONTROL Date Approved 3 of 28 IO5I2009 2.3 CONTAMINATED AREAS, MATERIALS AND EQUIPMENT following limits shall be Any area where loose surface contamination exceeds the following measurements posted as a Contaminated Area. Items exceeding the rdance with Procedure will be posted as radioactive material and handled in acco 315) 00960-C, Control of Radioactive Material. (1984301 2.3.1 2.4 CONTROLLED AREA site boundary, access to Any area outside of the restricted area but inside the which can be limited for any reason.

2.5 DERIVED AIR CONCENTRATION (DAC) h, if breathed by the The concentration of a given radionuclide in air whic in an intake of one ALl.

reference man for a working year of 2000 hours0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months , results Table 1, Column 3.

DAC values are given in 10CFR2O, Appendix B, (1993327366) 2.6 HIGH RADIATION AREA levels from radiation Any area, accessible to individuals, in which radiation ual receiving a dose sources external to the body could result in an individ es) from the radiation equivalent in excess of 100 mrem/hr at 30 cm (12 inch trates. (2000341553) source or from any surface which the radiation pene (1985306089)

Printed January 5, 2011 at 11:44

Procedure Number Rev Approved By C R Dednckson Vogtle Electric Generating Plant 00930-C 26 Page Number Date Approved RADIATION AND CONTAMINATION CONTROL 21 of 28 I05I2009 hands and feet at the 5.4.5 Personnel exiting contaminated areas should frisk their found.

nearest frisker location and notify HP if contamination is ld proceed to the nearest 5.4.6 After performing a hands and feet frisk, personnel shou ey. This survey is to be personnel contamination monitor for a whole body surv es.

performed prior to the individual putting on street cloth 5.4.7 ation is detected on any 5.4.8 HP shall be notified immediately whenever contamin individuals to perform self individual or their personal articles. At no time are decontamination without HP Personnel present.

qj N 549 HWpjpbfoj equvlert Exceptions may be warranted if isotopic (e.g., Rb-88, Cs-i 38).

identification demonstrates that the material is short lived HP Supervision shall approve the exceptions.

aminated individuals. HP 5.4.10 HP shall directly supervise the decontamination of cont and document personnel shall use approved methods of decontamination ical assistance may be contamination and decontamination effectiveness. Med of approved methods.

required when contamination cannot be removed by use ssays shall be performed in 5.4.11 Whenever internal contamination is suspected, bioa accordance with Procedure 00940-C, Bioassay Program.

aminated or radioactive 5.4.12 HP shall control, on an individual basis, the use of cont nal articles until sufficient articles and maintain possession of contaminated perso into compliance with radioactive decay or decontamination bring items Procedure 00960-C or until items are discarded.

Printed January 5, 2011 at 10:49

HL-16 NRC Written Examination KEY

71. 02.3.5 OOI/3/N/AIRMS-USE OF/2.9/2.9 MEM/NEWIRO/SROINRC/GCW The Fuel Handling Supervisor (SRO) has exited Containment and is proceeding to the HP Control point.

Which of the following is CORRECT concerning the Personnel Monitoring requirements for the FHSRO prior to leaving the RCA?

A Hands and Feet should be frisked at the nearest frisker location.

Proceed to the nearest personnel contamination monitor for a Whole Body survey.

B. Hands and Feet should be frisked at the nearest frisker location.

Proceed to the PESB personnel contamination monitor for a Whole Body survey.

C. No portion of the body is required to be frisked prior to exiting the HP Control Point.

Proceed to the nearest personnel contamination monitor for a Whole Body survey.

D. No portion of the body is required to be frisked prior to exiting the HP Control Point.

Proceed to the PESB personnel contamination monitor for a Whole Body survey.

Page 143 of 208

HL-16 NRC Written Examination KEY Feedback 2.3 Radiation Control 2.3.5 Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

(CFR: 41.11 141.12143.4145.9)

K/A MATCH ANALYSIS Question asks the requirements for the use of Portable Survey Instruments and personnel monitoring equipment required by plant Admin procedures.

ANSWER I DISTRACTOR ANALYSIS A. Correct-Hands and Feet are frisked as a minimum to prevent spreading of contamination prior to exiting the RCA. A Whole Body Survey must be performed prior to leaving the RCA.

B. Incorrect-Hands and Feet are frisked as a minimum to prevent spreading of contamination prior to exiting the RCA. Exiting to the PESB is plausible because there are personnel monitors present but the Whole Body survey is done prior to the RCA exit.

C. Incorrect-Upon leaving every contamination area, requirements are to frisk hands andd feet prior to leaving that area to prevent the spread of contamination. Whole Body survey is done prior to the RCA exit.

D. Incorrect-First part same as C above. Whole Body survey is done prior to the RCA exit.

REFERENCES 00930-C Radiation and Contramination Control VEGP learning objectives:

LO-LP-63930-1 1 State the frisking requirements necessary when leaving the RCA.

Page 144 of 208

Approved By :. - -i : :. Procedure Number Rev Vogtle Electric Generating llant C R Dedrickson 00930-C 26 Date Approved Page Number 0810512009 RADIATION AND CONTAMINATION CONTROL 20 of 28 5.3.1.4 The use of protective clothing for purposes other than contamination control is prohibited. Protective clothing should be distinctly marked or colored to facilitate control. Yellow plastics used for contamination control shall not be used as uncontrolled clothing for use in protection from the elements (e.g., raingear).

5.3.1.5 Protective clothing is stored in accordance with Procedure 92015-C, Transient Combustibles Control.

5.3.2 Control Of Radioactive Contamination (1984301315) 5.3.2.1 HP shall establish controls to ensure that the spread of contamination during work is minimized.

5.3.2.2 These controls shall include, but are not limited to consideration of the following:

a. Engineered Controls

b. Special Training

c. Job Planning

d. Monitoring of Materials and Equipment 5.3.2.3 Step-off pads will be used to control entrance and exit from a contaminated area, to minimize the spread of contamination from an area.

5.3.2.4 Multiple step-off pads (two or more) may be used in areas of high contamination.

5.4 PERSONNEL MONITORING 5.4.1 Personnel monitoring shall be performed in a manner consistent with the training each has received.

5.4.2 ..

b,ied.3 access control*.t point 4 or o 4

at ther areas designated by HP WhoIé bodmonitor

- .-j1.-* . .. .*

sbu.!.sed as primary.means,ofsurveying ,whn provided 5.4.3 The HP Manager, or designee may exempt the whole body survey requirement for exit from Radiation Controlled Areas where sealed radioactive material is stored, after surveys verify the escape of radioactive material is unlikely to occur under normal conditions.

5.4.4 Personnel exiting contaminated areas are responsible for surveying themselves when they cross a local control point equipped with either friskers or personnel contamination monitors.

Printed January 5, 2011 at 11:50

5.4.5 p

5.4.7 When exiting contaminated areas, hand carried tools and materials (e.g.,

clipboards, notebooks, pencils, etc.), shall be frisked for beta-gamma contamination by HP Personnel unless a red tool bag has been provided (outside the contaminated area in accordance with HP procedures) for their return.

5.4.8 HP shall be notified immediately whenever contamination is detected on any individual or their personal articles. At no time are individuals to perform self decontamination without HP Personnel present. requiring 5.4.9 Personnel decontamination should be performed when the results of personnel monitoring exceed 100 cpm above background per probe area when using an HP-210 probe or equivalent. Exceptions may be warranted if isotopic identification demonstrates that the material is short lived (e.g., Rb-88, Cs-138

).

HP Supervision shall approve the exceptions.

5.4.10 HP shall directly supervise the decontamination of contaminated individuals. HP shall use approved methods of decontamination and document person nel contamination and decontamination effectiveness. Medical assistance may be required when contamination cannot be removed by use of approved methods.

5.4.11 Whenever internal contamination is suspected, bioassays shall be perform ed in accordance with Procedure 00940-C, Bioassay Program.

5.4.12 HP shall control, on an individual basis, the use of contaminated or radioac tive articles and maintain possession of contaminated personal articles until suffici ent radioactive decay or decontamination bring items into compliance with Procedure 00960-C or until items are discarded.

Printed January 5, 2011 at 11:49

HL-1 6 NRC Written Examination KEY 7 G2.3 .7001 /3/N/AIRWP/3 .5/3.6 MEM/LOIT BANK/RO/SRO/NRC/GCW An individual is reviewing his RWP before beginning work.

If the dose rate in the area is 1123 mrem/hr, he would expect to find a colored RWP requiring an ALARA briefing prior to (b) entry.

A (a) RED (b) each B. a) YELLOW (b) each C. (a)RED (b) the first D. (a) YELLOW (b) the first Page 145 of 208

HL-16 NRC Written Examination KEY Feedback 2.3 Radiation Control 2.3.7 Ability to comply with radiation work permit requirements during normal or abnormal conditions.

(CFR: 41.12/45.10)

K/A MATCH ANALYSIS Questions asks what color and RWP entry requirements are needed for a specified dose rate.

ANSWER I DISTRACTOR ANALYSIS A. Correct.

B. Incorrect.> 1000 mr/hr requires RED RWP.

C. Incorrect.> 1000 mr/hr requires ALARA prior to every entry.

D. Incorrect.> 1000 mr/hr requires RED RWP and ALARA for every entry.

REFERENCES 43007, Issuance Use and Control of Radioactive Work Permits LOIT Bank G2.3.7 HL-15 Audit VEGP learning objectives:

LO-LP-63930-08 Given a requirement for entry into an area containinng radiation or contamination, state the RWP requirement for that area.

Page 146 of 208

Approved E fr l.A. Kochery Vogtle Electric Génerating Plant Procedure Number Rev Date Approved 43007-C 26.2 511712007 Page Number I

ISSUANCE USE AND CONTROL OF RADIOACTIVE WOR K PERMITS 29 of 44 RISKED BASED RWP FORMAT Color Code Radiological Types of RWPs Type of Briefing Required Significance General Radiological Conditions Dose Rate: < 100 mrem/hr All General RWPs Contamination Levels: < 200,000 dpmIlOO cm2 And No ALARA briefing required.

h2i Airborne Levels:

Specific RWPs with < 0.3 DAC low radiological risk Workers should always refer to the most recent survey information for the area(s) being worked in.

. Initial ALARA briefing required prior to first Specific RWPs that entry. Dose Rate: < 1000 mremlhr are tied to unique Work Groups . Additional ALARA briefing required when Contamination Levels: 500,000 dpmIlOO cm2 specified rad conditions are exceeded. Airborne Levels: < 0.3 DAC Specific RWPs

Yellow Moderate A pre-job ALARA briefing will be required if: Workers should always refer to the most recent covering work in . Radiological conditions that are addressed survey information for the area(s) being worked areas with in the Worker Instructions section may be in.

intermediate levels exceeded, or of radiological risk.

If the RWP default settings for the accumulated dose or dose rate alarms may be exceeded, or

. Breach of a contaminated system ALARA Briefing required prior to each entry.

Radiological conditions on the RWP will be Dose Rate: > 1000 mrem!hr Specific RWPs based on actual, projected or historical Contamination Levels: > 500,000 dpmllOO cm2 covering work in survey information.

Lq Airborne Levels:

areas with high levels Latest rad conditions and specific > 0.3 DAC of radiological risk. instructions will be covered in the pre-job

Workers should always refer to the most recent ALARA briefing survey information for the area(s) being worked in.

TABLE I Printed January 5,2011 at 14:10

HL-16 NRC Written Examination KEY

73. G2.4. 14 001/3/N/A/GENERAL EOP USAGE/3.8/4.5 C/A/LOIT BANK/RO/SRO/NRC/GCW Following a Reactor Trip, the lOAs have been completed for E-O:

The following conditions exist.

- SG # 1 has a Main Steam Safety Valve stuck open.

- SG # 2 has been identified as having a tube leak by Chemistry sample.

- Both SG levels are currently 12% in the Narrow Range.

- Containment pressure is 4.4 psig.

- The UO desires to isolate AFW flow to both SGs prior to reaching procedural direction to do so.

The UO would be allowed to isolate AFW flow to (a) , SS permission (b)______ required prior to taking the action.

Av (a) SG # 1 only, (b) is B. (a) SG # 1 only, (b) is not C. (a) both SG5 # 1 and # 2, (b) is D. (a) both SG5 # 1 and # 2, (b) is not Page 147 of 208

HL-16 NRC Written Examination KEY Feedback 2.4 Emergency Procedures I Plan 2.4.14 Knowledge of general guidelines for EOP usage.

(CFR: 41.10/45.13)

K/A MATCH ANALYSIS Question gives a plausible scenario with a safety valve stuck open on an SG and the other SG having a small SGTR. The candidate must choose which SG may have AFW flow isolated in accordance with 10020-C, EOP Users Guide.

ANSWER I DISTRACTOR ANALYSIS A. Correct.

B. Incorrect. SS permission is required to take prior actions.

C.Incorrect. SG # 2 level is below required for Adverse Containment.

D. Incorrect. SG # 2 level is below required for Adverse Containment and SS permission is required to take prior actions.

REFERENCES 10020-C, EOP and AOP Rules of Usage G2.4.14 HL-15 Audit VEGP learning objectives:

LO-LP-37121-01 State the intent of EOP 19020-C, Faulted SG Isolation Page 148 of 208

Approved By *

Procedure Number Rev J B Stanley Vogtle Electric Generating Plant Date Approved 10020-C 8 1 Page Number 1/20/10 EOP AND AOP RULES OF USAGE 4 of 26 3.1.3 Ti fólfvingacffons h&ibén analyzed ahdá b&akii jrob:edira1 diie&iifllWing th mplétin of sl1NI*BE IMLEMNE&AR. EAtIdEENVERIFIEI AND niethod frifying the tjiese adtion ill increasefR uhdueha the e/lüti 1*. -

v .

isateAw(FeecfWateronaIeaThng oru-v

çv ptureGaru4

. - Conifol hatup follwihg blowdo,n of faulted SG 3.1.4 EOPs (as well as AOPs) are entered based on the Entry Condifions or Symptoms at the beginning of the procedure. Operators are expected to be knowledgeable of these without referral.

3.1.5 Initial entry into the EOPs will be to 19000-C, E-0 REACTOR TRIP OR SAFETY INJECTION. In Mode 4 the operator may enter 19100-C, ECA-0.0 LOSS OF ALL AC POWER directly based on symptoms.

3.1.6 After verifying symptoms or entry conditions upon entering a procedure, the operator should go to Step 1 ACTION/EXPECTED RESPONSE (AER) column unless directed otherwise by the procedure just exited.

V Printed January 5, 2011 at 14:21

HL-16 NRC Written Examination KEY

74. G2.4.2 001 13/N/A!EOPS STPTS-ENTRY/4.5/4.6C/AJBANKJRO/SRO/NRC/GCW The following conditions exist:

- Unit is at 5% reactor power following a start up.

- A Pressurizer Spray valve fails open.

Which one of the following would be the first to trip the reactor? (Assume no operator action).

A. CT Delta T Reactor Trip B. Pressurizer High Level Reactor Trip C. Low Pressurizer Pressure Reactor Trip D Pressurizer Pressure Low Safety Injection Page 149 of 208

HL-16 NRC Written Examination KEY Feedback 2.4 Emergency Procedures I Plan 2.4.2 Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions.

(CFR: 41.7/45.7/ 45.8)

K/A MATCH ANALYSIS The question test the ability of the student to determine from a list of Reactor Trips the one that is applicable for the plant conditions.

ANSWER I DISTRACTOR ANALYSIS A. Incorrect-OTDeltaT is a variable setpoint with input from Pressurizer Pressure, Tavg and Neutron Flux. With Delta Tat a very low power level it would take a very long time and pressure would have to decrease significantly to reach the setpoint.

B. Incorrect-Pressurizer High Level Reactor Trip is only applicable with power above the P-7 setpoint (2/4 Power Range NI channels at 10% or 1/2 Turbine Impulse pressure channels at 10%).

C. Incorrect-Low Pressurizer Pressure Rx trip is only applicable with power above the P-7 setpoint (2/4 Power Range NI channels at 10% or 1/2 Turbine Impulse pressure channels at 10%).

D. Correct-Pressurizer Low Pressure Safety Injection is always applicable and would be the first event to trip the reactor in these conditions.

REFERENCES 17009-1 Annunciator Response for ALBO9 Panel 1 Cl on MCB-Window A04 Functional Diagrams 1 X6AAO2-00229 and 1 X6AAO2-00230 Tech Spec LCO 3.3.1 V-LO-PP-281 03-6.2 Rx Trip and ESFAS signals (Slide 31 and 33)

V-LO-PP-1 6101-04 RCS Temperature Instrumentation (Slide 28)

VEGP learning objectives:

LO-PP-28103-02 List all permisssives with appilcable set points, coincidences and functions.

LO-PP-281 03-02 List all reactor trip set points, coincidences, permissives and blocks.

Page 150 of 208

Approved By Procedure Number Rev S A Phillips Vogtle Electric Generating Plant -

- 17009-1 11 Date Approved ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 09 ON PANEL 1C1 Page Number 3/27/08 ONMCB 9of38 WINDOW A04 ORIGIN SETPOINT PRZR PRESS Any two of the 1870 psig SI following: RX TRIP

a. 1-PT-0455

b. 1 -PT-0456

c. 1-PT-0457

d. 1 -PT-0458 1.0 PROBABLE CAUSE

1. Loss of coolant accident or Steam Generator tube rupture

2. Pressurizer Pressure Control System malfunctions 2.0 AUTOMATIC ACTIONS NOTE This function is manually blocked below P-il permissive.

1. Reactor Trip

2. Turbine Trip

3. Safety Injection 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, 1X6AAO2-230, PLS, Technical Specification s LCQ 3.3.2, and 3.3.1 Printed January 5, 2011 at 15:24

RTS Instrumentation 3.3.1 Table 33.1-1 (page 4 of 9)

Reactor Trip System Instrumentation I

APPLICABLE MODES OR NOMINAL OTHER TRIP SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS CONDITIONS SEWOIrrr(n REQUIREMENTS VALUE

11. Undervoltage (f) 1 2per bus M SR 3.3.1.9 9481 V 9600V RCPs SR 3.3.1.10 SR 3.3.1.15

12. Underfrequency (f) 1 2 per bus M SR 3.3.1.9 57.1 Hz 57.3 Hz RCPs SR3.3.1.10 SR 3.3.1.15

13. Steam Generator 1,2 4perSG E SR 3.3.1.1 35.9% 37.8%

(SG)Water Level - SR 3.3.1.7 LowLow SR3.3.1.10 SR 3.3.1.15 (continued)

(f) Above the P-7 (Low Power Reactor Trips Block) interlock.

(n) A channel Is OPERABLE with an actual Trip Setpolnt value outside Its calibratIon tolerance band provided the Trip Setpolnt value Is conservative with respect to its associated Allowable Value and the channel is readjusted to wIthIn the established calibration tolerance band of the Nominal Trip Setpoint. A Trip Setpolnt may be set more conservatIve than the Nominal Trip Setpoint as necessary In response to plant conditions.

Vogtle Units I and 2 3.3.1-17 Amendment No. 128 (Unit 1)

Amendment No. 106 (Unit 2)

o C OTL\T OTAT Rx Trip: - 214 Logic

- OTAT set point AT

- inputs from Tavg I PRZR Pressure I Al

- bases: protects against DNB

- Nominal 100% power value = 114.9%

Tavg U causes set point to reduce Pressurizer Pressure J causes set point to reduce Al outside of -23 to +10 causes set point to reduce Tavg t causes set point to increase Pressurizer Pressure U causes set point to increase V-LO-PP-1 6101-04.1 28

o What Reactor trips now become active?

High pressurizer level Low pressurizer pressure RCP UV RCP UF Two loop loss of flow V-LO-PP-28103-6.2 33

What will give you P-7?

214 PR> 10% power or 1/2 Turbine power 10%

V-LO-PP-28103-6.2 31

HL-16 NRC Written Examination KEY

75. G2.4.34 OO1/3/N/AJEOP TASK OUT CR142/4. 1 MEMJMOD BANK-FARLEY/RO/SRO/NRC/GCW Given the following:

- A fire required evacuation of the Control Room.

- Procedure 18038-1 ,Operation From Remote Shutdown Panels is in progress.

- All equipment has been transferred to the appropriate Shutdown Panels.

The crew is at the step for controlling PRZR level within the required band. Charging was aligned to the VCT upon exiting the Main Control Room.

Upon receipt of a VCT Lo-Lo Level, the RWST to CCP suction valves and VCT outlet isolation valves (1) automatically re-position and 1 FV-1 21 will be controlled locally (2)

A. (1) will (2) from the B Train Shutdown Panel.

B. (1) will (2) from the control box located on C level of the Aux Building.

C. (1) will not (2) from the B Train Shutdown Panel.

D (1) will not (2) from the control box located on C level of the Aux Building.

Page 151 of 208

HL-16 NRC Written Examination KEY Feedback 2.4 Emergency Procedures I Plan 2.4.34 Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects.

(CFR: 41.10/43.5/45.13)

K/A MATCH ANALYSIS The subject tests the knowledge of where equipment is operated locally and interlocks associated with valves that have been transferred to local upon evacuation of the control room.

ANSWER / DISTRACTOR ANALYSIS A. Incorrect-Auto swap interlock associated with the RWST and VCT suction valves are defeated upon transfer to local control. 1 FV-1 21 has no local control station on the B SDP.

B. Incorrect-Interlock same as A above. 1 FV-121 is locally controlled from the Aux Bldg control box.

C. Incorrect-Interlocks are defeated while in local control and there is no control station on B SDP.

D. Correct-See A, B and C above.

REFERENCES Procedure 18038-1, Operation From Remote Shutdown Panels P & ID 1X4DB116-1 Farley NRC 2007-Dec Q #74 VEGP learning oblectives:

LO-PP-60327-06 Describe how the following equipment interlocks are affected after transfer to local control.

-VCT low level protection Page 152 of 208

Approved By : Procedure Number Rev J. B. Stanley 4Vogtle Electric Generating Plant 18038-1 32 DateApproved Page Number OPERATION FROM REMOTE SHUTDOWN 8/27I20l0 PANELS 34 of 123 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED CAUTIONS jfjred

PRZR Heaters will NOT cut off on low PRZR level after controls have been transferred to the Shutdown Panels.

When operating from the Shutdown Panels, Train B is the preferable charging train.

Closing BSIVs by opening 1AD12-03 and IBDI2-03 will isolate RMW to VCT blender.

39 Control PRZR level 50% to 70%:

a. Check charging pump suction a. Align charging pump suction aligned to VCT: to RWST:

. Letdown in service At Shutdown Panel B:

(1-Fl-0132B on Shutdown Panel A). . Open 1-LV-0112E.

. 1-LV-01 I 2B VCT OUTLET

Close 1 -LV-01 12C.

ISOLATION on Shutdown Panel A - OPEN. -OR-

. 1-LV-0112C VCT OUTLET At Shutdown Panel A:

ISOLATION on Shutdown Panel B OPEN.

- . Open 1-LV-0112D.

. 1-LV-01 12D RWST TO . Close 1-LV-O1 12B.

CCP-A&B SUCTION on Shutdown Panel A -

CLOSED.

1-LV-0112E RWSTTO CCP-A&B SUCTION on Shutdown Panel B -

CLOSED.

Step 39 continued on next page Printed January 5, 2011 at 17:56

Approved By Procedure Number Rev J B Stanley Jogtle Electric Generating Plant 18038-1 32 DateApproved Page Number OPERATION FROM REMOTE SHUTDOWN

() PANELS 36 of 123 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTES

At SS discretion, ATTACHMENT F may be used to establish Safety Grade Charging even if Instrument Air is available.

CCP motor starting limitations are two consecutive starts from ambient temperature, one start from operating temperature; subsequent start after 15 minutes running or 45 minutes standstill.

g. Maintain PRZR level between g. Maintain PRZR level 50% and 70%: between 50% and 70%

using either of the following:

. Throttle charging using 1-FHC-0121 (outside NCP With CCP B running.

Room in AB-C113).

1) Verify mini-flow path:

Control seal injection flow 8 to 13 gpm per RCP by . 1HV8110 throttling 1-1208-U6-136 and CCP-A&B closing 1-1208-U6-134 (both COMMON in NCP valve gallery MINIFLOW open AB-Ci 12). (Shutdown Panel A).

-OR-

. 1-HV-8111B CCP-B MINIFLOW open

. IF instrument air is NOT (Shutdown Panel available, B).

THEN maintain PRZR level between 50% and 70% by using ATTACHMENT F.

Step 39 continued on next page Printed January 5, 2011 at 17:55

Approved By Procedure Number Rev Vogtle Electric Generating Plant J. B. Stanley 18038-1 32 DateApproved Page Number OPERATION FROM REMOTE SHUTDOWN 8/27/2010 PANELS 105 of 123 ATTACHMENT L Sheet 1 of 2 LOCAL OPERATION OF 1-FHC-0121 NOTES

Communication equipment is to be found on the wall inside communication storage cabinet labeled 1-FV-121 (ALB Cl 13). A flat bladed screwdriver will be needed to open this box.

Sound powered phone jacks may be found next door in room C109 (1ABD room).

To establish control of 1-FV-0121 perform the following:

1. Establish communication with operator at local controller 1-FHC-0121 (AB C113).

2. Note charging flow on 1FI-0121B (AB C113).

3. Note Pressurizer level on 1LI-0460B (AB C113).

4. Note pressure of Auto gauge on left side of 1-FHC-0121 (IF no pressure is indicated skip the next step).

NOTES

Rotating the control knob clockwise raises air pressure and decreases charging flow.

Rotating control knob counter-clockwise lowers air pressure and raises charging flow.

5. Rotate control knob on right side of control box 1-FHC-0121 as required to increase or decrease pressure on right hand gauge until both pressure gauges are reading approximately the same.

ALB 17011 F02 CHG FCV-121 IN LOCAL CONTROL

6. Place switch on left side of control box 1-FHC-0121 to Manual position.

Printed January 5,2011 at 16:41

Approved By Procedure Number Rev J. B. Stanley Vogtle Electric Generating Plant 18038-1 32 jDate Approved Page Number OPERATION FROM REMOTE SHUTDOWN 8/27/2010 PANELS 106 of 123 ATTACHMENT L Sheet 2of 2 LOCAL OPERATION OF 1-FHC-0121

7. Adjust charging flow as directed by control room/shutdown panel operator (Increasing controller output decreases charging. Lowering controller output increases charging.

8. WHEN directed by operator, maintain pressurizer level (1 Ll-0460B) by slightly increasing or decreasing controller flow (in approximately 5 gpm increments) as read on I Fl-0121 B.

WHEN local control is no longer desired perform the following:

NOTE Adjusting 1FIC-0121 on QMCB will change Auto pressure (left gauge) and allow a bump less transfer to occur when the two pressures are matched.

9. On QMCB Adjust 1-FIC-0121 until Auto pressure matches pressure of Manual controller.

-OR-

10. Locally on 1-FHC-0121 select Auto mode.

11. In Main control room verify:

Annunciator 17011 F02hascleared.

1-FIC-0121 is controlling properly.

12. On 1-FHC-0121 decrease air controller setpoint to minimum by slowly turning control knob counter-clockwise until air pressure is approximately 0 psig on manual gauge.

° END OF ATTACHMENT L Printed January 5,2011 at 16:41

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1. G2.4.34 OOJINEW/ILOWERJ/RO/FARLEY Given the following:

room.

A fire required evacuation of the control

, Fire in the Control Room.

The crew is performing actions of AOP-28.2

HSD Panel A is manned and functional.

FLOW, to maintain pressurizer level The crew is at the step to adjust HIK-l 22, CHG is reported:

within the required band when the following FCV-l 22 properly.

H 1K-i 22 on the HSD panel is not controlling

Pressurizer level is 16% and trending down.

Which ONE of the following contains a cor rect method for controlling PRZR level and operated lAW AOP-28.2?

the correct location of the components to be from the HSD panel.

A. Close LCV-459 or 460, LTDN LINE ISO, nd FCV-122 locally in the 100 Piping Control charging flow using the bypasses arou Penetration Room entrance.

N ORIF ISO, at the HSD panel.

B. Close HV-8i49A and 8i49B or C, LTD nd FCV-122 locally in the 100 Control charging flow using the bypasses arou hallway BIT area.

ISO, from the HSD panel.

C. Close LCV-459 or 460, LTDN LINE B, HHSI TO RCS CL, locally in the 100 Control charging flow using MOV-8803A or Piping Penetration Room entrance.

N ORIF ISO, at the HSD panel.

D Close HV-8149A and 8149B or C, LTD HHSI TO RCS CL, locally in the 100 Control charging flow using MOV-8803A or B, hallway BIT area.

A. incorrect. LCV-459 or 460 can not isolated at the HSDP and would not be isolated procedurally per the note below. Control of charging using the bypass valves is an option, actually the first option, (step 14.6) but in this case the first part of the distracter is not correct and the location is correct.

AOP-28.2 NOTE: Isolation of letdown due to low pressurizer level (15%) will unnecessarily complicate plant recovery (LCV 459 & 460 cannot be re-opened from the HSDP, Reactor head vents must then be used for removing mass from the primary system). Therefore, emphasis should be placed on controlling charging flow to establish a stable or slowly rising pressurizer level that compensates for any effect on level due to cooldown.

B. incorrect. Placing 2 orifices on service is correct at step 25 in the procedure for controlling level, and bypassing FCV-122 is correct, but the location is not correct.

C. incorrect. LCV-459 or 460 can not isolated at the HSDP and would not be isolated procedurally per the note below.

control charging flow using M0V8803A or B, HHSI TO RCS CL, is correct but the location is not correct.

D. Correct. Placing 2 orifices on service is correct step 25 in the procedure for controlling level, and control charging flow using M0V8803A or B, HHSI TO RCS CL, is correct and the location is correct.

ML112270225

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