ML071490433: Difference between revisions

Time:

Source:

Cognitive Level:

Job Position:

Dale:

Points:

NEW

HIGHER

4/2007

1.00

Version:

Answer.

o 1 23 4 5 6 7 8 9

CCCCACC BAB

Source If Bank:

Difficulty:

Plant:

Last 2 NRC?:

Scramble: Range: A - 0

SEQUOYAII

NO

Monday, May 07, 20073:22:57 PM

2

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

2 . 001 AAl .04 001

Given the following plant conditions:

Unit 1 has just reached 90% power after a linear power ascension from 50%

over the last four hours.

Control Bank D continues to withdraw with no motion demanded.

Which ONE (1) of the following describes the initial effect on the unit?

A. Reactor Power decreases AND OT Delta T setpoint decreases

B. Reactor Power decreases AND OT Delta T setpoint increases

C~ Reactor Power increases AND OT Delta T setpoint decreases

D. Reactor Power increases AND OT Delta T setpoint increases

A-Incorrect. Reactor Power response is incorrect; Control Rod withdrawal results in

Hot leg temperature rise, reactor power increase due to the positive reactivity being

added with no additional RCS heat removal . OTdeitaT setpoint response is correct;

OT and OP delta T setpoints will lower as power and temperature rise.

B-Incorrect. Reactor Power response is incorrect; Control Rod withdrawal results in

Hot leg temperature rise, reactor power increase due to the posttive reactivity being

added with no addftional RCS heat remo val . OTdeitaT setpoint response is

incorrect; OT and OP delta T setpoints will lower as power and temperature rise.

C-Correct. Control Rod withdrawal results in Hot leg temperature rise and reactor

power increase due to the positive reactivity being added with no additional RCS

heat removal. OTdeltaT setpoint response is correct; OT and OP delta T setpoints

will /ower as power and temperature rise.

D-Incorrect. Reactro Power response is correct; Control Rod wfthdrawal results in Hot

leg temperature rise and reactor power increase due to the positive reactivity being

added with no addftional RCS heat removal. OTdeitaT setpoint response is

incorrect; OT and OP delta T setpoints will/ower as power and temperature rise.

Monday. May 07,2007 3:22:57 PM

3

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

Ability 10detennine and interpret !he folowing as It1eyapply to ee Continuous Rod Withdrawal: Reactlf power and its lJend

(

Question No.

Tier 1 Group 2

Importance Rating :

Technical Refere nce:

57

R04.2

T&AA

Proposed references to be provided to applicants during examination:

None

Learning Objective:

OPL271CGFES Attachment 16 Objective 20

Question Source:

Bank

Question History:

Robinson 2007 Editorially Modified

Question Cognitive Level:

Higher

10 CFR Part 55 Content:

41.5

Comments: modified bullet in stem

Time:

I

Points:

1.00

Version:

Answer:

SEQUOYAJI

NO

(

Source:

Cognitive Level:

Job Position:

Date:

BANK

HIGHER

412007

o I 2 3 4 5 6 7 8 9

C C C A D DA D DC

Scramble Range:A- D

Source If Bank:

ROBINSON 2007 NRC

Difficulty:

Plant:

Last 2 NRC!:

Monday, May 07, 2007 3:22:57 PM

4

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

3 . 001K2.03 001

Which ONE (1) of the following is the primary source of power to the Rod Control Logic

Cabinets?

A~ Rectified AC from the Rod Drive MG output.

B. Rectified AC from 120 VAC Instrument Bus

C. 125 VDC stepped down from station battery boards

D. 250 VDC stepped down from station battery boards

A. Correct.

B. Incorrect. Backup source is this supply

C. Incorrect. Correct that it DC; incorrect supply. Plausible since 125v is supplied to

the Rod Control system via the shunt trip coil on the RTBs and as control power to

the MG set output breakers that are opened locally following an ATWS condition.

D. Incorrect. Correct that it is DC; incorrect supply. Plausible since 250v DC is used as

control power to the MG set motor breakers that are opened locally following an

A7WS condition.

KncM1edge d bus po.ver suppliestothe loIIcM1ng:One-line diag'a'nof pclWEf supplies to Ioojc circuits.

Que stion No.

Tier 2 Group 2

Importance Rating:

Technical Reference:

29

R0 2.7

Rod Control SD. Pg 32; 1.2-45N777-3 r17

Proposed references to be provided to applicants during examination:

None

Learning Objective:

Question Source:

Question History:

OPL200RDCNT Objective 4

New

Question Cognitive Level:

Lower

10 CFR Part 55 Content:

41 .6

(

Comments:

Time:

Points:

1.00

Version:

0 1 2 3 4 5 6 7 8 9

Answer: ACA B CC D D BC

Scramble Range: A - D

Monday, May 07,20073:22:57 PM

5

SEQUOYAH

NO

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

NEW

Source If Bank:

LOWER

Difficulty:

Plant:

412007

Last2 NRC?:

Source:

Cognitive Level :

Job Position:

Date:

(

Monday, May 07, 2007 3:22:57 PM

6

(

QUESTIONS REPORT

for SEOUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

4 . 003 K4.04 001

On a loss of CCS to an RCP thermal barrier heat exchanger, which ONE (1) of the

following choices contains parameters that indicate proper seal injection in accordance

with AOP-M.03, Loss of Component Cooling Water?

A.

B.

D.

Seal Injection Flow

6GPM

9 GPM

12 GPM

14 GPM

veT Outlet Temperature

126* F

135*F

128* F

125*F

(

A. Incorrect. Seal Injection flow too low per the referenced procedure.

B. Incorrect. VCT temperature too high per the referenced procedure.

C. Correct. 8-13 GPM flow to seals and less than 130 degrees F from VCT is the

requirement if TBHX cooling is lost per the referenced procedure.

D. Incorrect. Seal Injection flow too high per the referenced procedure.

Knowledge of Reps designleature(s ) and/or interlock(s) which proWje for the folICM'ing:Adequate cooling ofRep mob" andseals.

Question No.

Tier 2 Group 1

Importance Rating:

Technical Reference:

1

R0 2.8

AOP-M.03

Proposed references to be provided to applicants during exa mination:

None

Learning Obje ctive:

Question Source:

Question History:

OPL271AOP-M.03 Objective 6

New

(

Question Cognitive Level:

Lower

10 CFRPart 55 Content:

41 .3

Comments: Modified wording in stem to include AOP-M.03 and changed 0 temperature from

125 to 132 to ensure it is incorrect.

Monday, May 07,20073:22:57 PM

7

Scramble Range: A * D

SEQUOYAII

NO

NEW

LOWER

4/2007

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

Time:

I

Points:

1.00

Version:

0 1 2 3 4 5 6 7 8 9

Answer.

C AC BCA D BA B

Source IfBank:

Difficulty:

Plant:

Last 2 NRC'?:

Mes

Source:

Cognitive Level:

Job Position:

Date:

(

Monday, May 07, 2007 3:22:57 PM

8

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

5. 004 A2.03 001

Given the following plant conditions:

Unit 1 is at 100% power.

All control systems are in normai alignment.

Charging fiow is 92 GPM and rising.

1-TI-62-71, Regen HX Outlet Temp - Letdown, has decreased 5°F from its

steady state value.

VCT level is 33% and lowering.

PZR level is 59% and lowering slowly.

RCS temperature is 578°F and stable.

Which ONE (1) of the following describes the effect on the unit and the action required

in accordance with AOP R.05, RCS Leak and Leak Source Identification?

A. RCS leakage is from the charging line downstream of the regenerative heat

exchanger. Make a containment entry to determine if the leak is isolable.

B~ RCS leakage is from the charging line downstream of the regenerative heat

exchanger. Isolate Charging and Letdown.

C. RCS leakage is from the charging line upstream of the regenerative heat

exchanger. Initiate an investigation to determine if the leak is isolable.

D. RCS leakage is from the charging line upstream of the regenerative heat

exchanger. Isolate Charging and Letdown.

A. Incorrect. wrong action per AOP R.D5. Plausible because this would be required

per tech specs if leak cannot be isolated.

B. Correct. If RHX QuIet temperature is lowering, then more charging ffow is going

through the RHX. This means the leak is downstream of the RHX in containment.

Since Charging ffow increased from a normal value of 87 GPM to 92GPM with no

other changes, leakage is greater than TS limits which would require shutdown.

C. tncorrect. Wrong location ofleak, wrong action

D. Incorrect. Wrong location, correct action

Monday. May 07, 2007 3:22:57 PM

9

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

Ability 10(a) predict the impacts of the following malfunctions or operntioos m the eves; and (b) based on 1tlose pndctions, use

procedures tocorrect,control. or mitigatethe ccosecoeocee of eeee ~ or operations: 8ourIda'y isoIaticnvat.oe leak.

(

Question No.

2

Tier 2 Group 1

Importance Rating:

RO 3.6

Technical Reference:

R.05

Proposed references to be provided to applicants during examination:

None

Leaming Objective :

OPL271AOP-R.05 Objective 4

Question Source :

New

Question History:

Question Cognitive Level:

Higher

10 CFR Part 55 Content:

41.10

Scramble Range: A * D

SEQUOYAII

NO

o I 2 3 4 5 6 7 8 9

BDDA CB A D A A

Source If Bank:

Difficulty:

Plant:

Lasl2 NRC?:

Answer:

NEW

HIGHER

4/2007

Source:

Cognitive Level:

Job Position:

Date:

Comments: Correct typo in stem "downstream". Added values to PZR and VCT level, ReS

temp. Changed charging flow to 92 gpm. Added TI information to letdown temp indication in

stem. Removed the Tech Spec from each answer. This will cause applicant to differentiate

between break location and correct proceduralffech Spec action. Added in accordance with

AOP to stem to rule out A being correct.

Time:

I

Points:

1.00

Version:

(

Monday, May 07, 20073:22:58 PM

10

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

6 . 005 AI.OI 001

Given the following plant conditions:

Unit 1 cooldown in progress.

RCS temperature is 225°F.

Current RCS cooldown rate is 25°F per hour.

The OATC is placing 1A-A RHR Pump in service in accordance with

0-SO-74-1 ,

Section 5.5.2, Placing RHR in service for normal shutdown cooling.

1B-B RHR is out of service.

Which ONE (1) of the following actions is required to control RHR HX temperature in

accordance with 0-SO-74-1, Residual Heat Removal System, and the reason?

(

B.

C.

D.

RHR HX Flow Control

Throttle open RHR HX A CCS

Outlet Valve FCV-70-156

Throttle open RHR HX A CCS

Outlet Valve FCV-70-156

Throttle open RHR HX A Outlet

Valve FCV-74-16

Throttle open RHR HX A Outlet

Valve FCV-74-16

Reason

To maintain RHR HX CCS

outlet temperature <145°F.

To maintain RHR HX outlet

to RCS temperature <235°F.

To maintain RHR HX outlet

to RCS temperature <235°F.

To maintain RHR HX CCS

outlet temperature <145°F.

(

A. correct: SO-74-1 throttles the RHR HX CCS outlet valve to control CCS outlet

temperature <145F.

B. Incorrect: correct valve manipulation but the CCS valve is not used to control RHR

tempera ture. Applicant may confuse with a different precaution on placing RHR in

service at >235F which may cause flashing in the RHR system. 0-50 -74-1 , section 5.5.1

step 17 caution.

C.I ncorrect: Incorrect valve, applicant could confuse the precaution since throttling this

valve would accomplish the same result.

D. Incorrect: Wrong valve but correct reason.

Monday. May 07, 2007 3:22:58 PM

11

QUESTIONS REPORT

for SEQUO YAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

Abi~ty to pe:Ict and/ormonitro changes in ~ (to perventexceeding design ~mits) associated with op8'ating the RHRS cootrols

including: Heatup/cooIdoMl rates .

(

Question Number:

Tier 2 Group 1

Importance Rating: RO 3.5

4

Technical Reference: 0-50 -74-1, section 5.5.2 step 8 caution. page 71 of 204.

Proposed references to be provided to applicants durinq examination:

None

Learning Objective:

10 CFR Part 55 content:

Comments:

OPL200.RHR Objective 5.b

41.10

Time:

Points:

1.00

Version: o I 2 3 4 5 6 7 8 9

Answer: ABC DCA DCC B

Scram ble Range: A * 0

(

Source:

NEW

Source IfBank:

Cognitive Level:

HIG HER

Difficulty:

Job Position:

Plant:

SEQUOYAH

Date:

412007

Last 2 NRC?:

NO

(

Monday. May 07. 2007 3:22:58 PM

12

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

7. 005 AA2.02 00 I

Given the following plant conditions:

During a power decrease, ONE (1) Control Bank D rod becomes

mispositioned 14 steps out from its group .

The crew is attempting to realign the rod in accordance with the appropriate

procedure.

The rod has been misaligned for 40 minutes.

Assuming the rod can be moved, which ONE (1) of the following is the speed at which

the rod will move when it is realigned?

A. 32 steps per minute

B~ 48 steps per minute

C. 64 steps per minute

D. 72 steps per minute

A. Incorrect. 32 SPM is part of the variable speed for auto rod control 3-5 deg F

mismatch.

(

B. Correct. Control Bank in Manual or Bank Select will initiate motion at 48 SPM

C. Incorrect. 64 SPM is for Shutdown Banks

D. Incorrect. 72 SPM is maximum variable speed

(

Monday, May 07, 2007 3:22:58 PM

13

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

Ability todetermineand interpret the following as they apply to the Inoperable f Stuck Conlrol Rex!: Difference betweeo jog and run rod

speeds , effect on CRDM of stuck rod

(

Question No.

Tier 1 Group 2

Importance Rating:

Technical Reference:

58

R0 2.5

Rod Control SD, pg 38

Proposed references to be provided to applicants during examination:

None

Learning Objective:

Question Source:

Question History:

OPL271AOP-C.Ol Objective 5

OPT200RDCNT Objective 4.e

New

Question Cognitive Level:

Lower

10 CFR Part 55 Content:

41.6

(

Comments:

Time:

Source:

Cognitive Level:

Job Position:

Dale:

Points:

NEW

LOWER

412007

1.00

Version:

Answer.

o t 2 3 4 5 6 7 8 9

BDAB CB CADB

Source If Bank:

Difficuhy:

Plant :

Last 2 NRC?:

Scramble Range: A - D

SEQUOYAH

NO

Monday, May 07, 2007 3:22:58 PM

14

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

8 . 005 KI.QI 00 1

Given the following plant conditions:

The Unit was in Mode 1, at 100% load.

A LOCA is in progress.

Step 11 (DETERMINE if RHR pumps should be stopped) of E-1, Loss of Reactor

or Secondary Coolant, is in progress.

RCS Pressure is 350 psig and dropping.

Both RHR Pumps are running.

Wh ich ONE (1) of the following procedural actions is required?

A. Immediately stop both pumps and place handswitches in pull-to-lock .

B~ Open 1-FCV-70-153 and 156, CCS to the RHR Heat Exchangers.

C. Transition to ECA-1.1, "Loss of RHR Recirc".

D. Reset the Safety Injection Signal, stop both RHR Pumps, and place their

Handswitches in A-Auto.

A. Incorrect. With pressure dropping, pumps will not be stopped.

(

8. Correct. Open valves to ensure pump cooling. [EPM-3-E-l Step 10 ERG statement:

On low-head systems where the pump recirculates on a small volume circuit. there is concernfor

pump and motor overheating. Shutdown of the pump and placement in the standby mode. when

the ReS pressure meets the criteria outlined in thisstep. allows for future pump operability.)

G. Incorrect. Would not go to EGA-I .1 based upon indications given. Gredible

because loss of both RHR pumps is an entry to EGA-I.1.

D. Incorrect. With pressure still dropping, pumps will remain running.

(

Monday, May 07, 2007 3:22:58 PM

15

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

Knowledge of the ~ieaI connectionsMd/Ol" cause-effect r~ationshi ps betweeI'l the RHRS and the following systems: ccws.

(

Question No.

Tier 2 Group 1

Importance Rating:

Techn ical Reference:

3

R0 3.2

E-l ; EPM-3-E-1

Proposed references to be provided to applicants during examination:

None

Learn ing Objective:

Question Source:

Question History:

OPT200RHR.R2 Objeclive 4.d

New

Question Cognitive Level:

Higher

10 CFR Part 55 Content:

41.10

Comments: Add ed proce dure step in stem. Removed last part of each answer that was not

required information to differentiate between answers. Removed RHR flow indications from

stem to prevent overlap with question 006A3.01.

SEQUOYAH

NO

(

Time:

Source:

Cognitive Level:

Job Position:

Date :

3

Points:

BANK

HIGHER

3

1.00

Version:

Answer:

o I 2 3 4 5 6 7 8 9

BAB CD AB CDA

Scramble Range: A - D

Source If Bank:

WBN BANK

Difficulty.

Plant:

Last 2 NRC?:

Monday, May 07,20073:22:58 PM

16

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

9. 006 AJ.O! 001

Given the following plant conditions:

A plant cooldown was in progress in accordance with 0-GO-7, Unit Shutdown

from Hot Standby to Cold Shutdown.

RCS pressure was at 800 psig when a LOCA occurred.

RCS pressure is currently 450 psig and lowering slowly.

Which ONE (1) of the following describes the status of the ECCS equipment identified

below?

A. Cold Leg Accumulator level is stable;

RHR flow is rising

B. Cold Leg Accumulator level is lowering;

RHR flow is zero

C~ Cold Leg Accumulator level is stable;

RHR flow is zero

D. Cold Leg Accumulator level is lowering;

RHR flow is rising

(

A. Incorrect. Accumulators would be stable but RHR would not be injecting.

B. Incorrect. With RCS pressure at 450 psig, Cold Leg Accumulators would be

injecting if not isolated. Plausible because ifpower were not removed, the SI signal

would automatically open the valves. RHR flow will not be indicated untif RCS

pressure drops to approximately 180 psig.

C. Correct. CLAs would be not be lowering since they would be isolated with power

removed in accordance with GO-7. RHR flow will not be indicated untif RCS

pressure drops to approximately 180 psig.

D. Incorrect. Accumulator level is stable and RHR is not injecting.

(

Monday, May 07, 2007 3:22:58 PM

17

QUESTIONS REPORT

for SEaUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7 -07

Ability to monitor automatic operaliOn of the ECCS. inCluding: Accumulators.

(

Question No.

Tier 2 Group 1

Importance Rating:

Technical Reference:

6

R04.0

ECCS SD; 0-GO-7 Sect. 5.2 Step 18 incl;

Proposed references to be provide d to applicants during examination:

None

Learning Objective:

Question Source:

Question History:

OPT200.ECCS Objective 3

New

Question Cognitive Level:

Higher

10 CFR Part 55 Conlent:

41.4

Comments: Removed 51 flow from each answer based on NRC comments. Modified stem to

have accumulators isolated based on NRC comments. This chang ed correct answer to C.

Minor editorial change to stem wording.

(

Time:

Source:

Cognitive Level :

Job Position:

Dale:

Points:

NEW

HIGHER

412007

1.00

Version:

Answer:

o 1 2 3 4 5 6 7 8 9

CDDAADDBBA

Source If Bank:

Difficulty:

Plant:

Last 2 NRC ?:

Scramble Range: A - D

SEQUOYAH

NO

Monday, May 07,2007 3:22:58 PM

18

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

10. 006 K3.0\\ 00 \\

Given the following plant conditions:

A LOCA has occurred.

RCS pressure is 100 psig.

All CETC's are reading approximately 810°F and rising slowly.

Centrifugal Charging Pumps are running.

Both SI pumps are tripped .

RHR Pump 1A-A is tripped.

RHR Pump 1B-B is running.

RVLlS Lower Range Level is 55% and decreasing slowly.

In accordance with FR-O, Status Trees, which ONE (1) the following describes the

current status of core cooling and the reason why?

A. Core cooling is inadeq uate because both SI pumps are required to sustain core

cooling.

B. Core cooling is inadequate because RVLlS Lower Range level is too low to sustain

core cooling .

C. Core cooling is degraded because both SI pumps are required to sustain core

cooling .

D~ Core cooling is degraded because RVLlS Lower Range level is too low to sustain

core cooling.

A. Incorrect. Inadequate core cooling does not exist with the given CETC values.

Safety analysis assumes 1 CCP, 1 SI, 1 RHR, and 2 SI Accumulators.

B. Incorrect. Inadequate core cooling does not exist with the given CETC values and

stated RVLlS level.

C. Incorrect. Safety analysis assumes 1 CCP, 1 St, 1 RHR, and 2 SI Accumulators.

Without an SI pump running, safety analysis assumptions are not met. Applicant

may confuse the safety analysis with the TIS LCO entry conditions.

D. Correct. RVLlS Lower Range is above the FR-C.l minimum of 42% for inadequate

core cooling but is too low to sustain core cooling.

Monday, May 07, 200 7 3:22:58 PM

19

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

KnO'Nledge ollhe effect thai a loss or malfunction of the ECCS will have on the following: ReS.

(

Question No.

Tier 2 Group 1

Importance Rating:

5

R04.1

(

Technical Reference: TS 3.5.2; 1-FR-O, EPM-3-FR-O

Proposed references to be provided to applicants during examination:

None

Learning Objective:

OPL200.ECCS Objective 6

Question Source:

Bank

Question History:

WTSI (Various Exams)

q uestion Cognitive Level:

Higher

10 CFR Part 55 Content:

41,8

Comments: Modified A and B distracto rs to change adequate to inadequate core cooling

based on NRC comments and changed B distractor to read similar to D. Modified stem to

reference FR-O.

SEQUOYAH

NO

(

Time:

Source:

Cognitive Level:

Job Position:

Date:

Points:

BANK

HIGHER

412007

1.00

Version:

Answer:

o J 2 34 5 6 7 8 9

DD C B B A C C A A

Scramble Range:A- D

Source If Bank:

ROBINSON 2007 NRC

Difficulty:

Plant:

Last 2 NRC?:

Monday, May 07, 2007 3:22:58 PM

20

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

11. 007 EAI.08 001

Given the following plant conditions:

Unit 1 was operating at 40% power.

1A-A EDG was running in parallel for monthly surveillance testing .

A reactor trip occurred due to a loss of offsite power.

Tavg is 549' F and decreasing.

All equipment is operating as designed.

SG levels dropped to the following values:

SG 1 - 10% NR

SG2- 14%NR

SG 3 - 17% NR

SG 4 - 12% NR

Which ONE (1) of the following describes the operation of the 1A-A MDAFW and

Turbine Driven AFW pumps?

1A-A MDAFW Pump

TDAFWPump

A.

OFF

B~

RUNNING

(

C.

RUNNING

OFF

D.

OFF

RUNNING

A. Incorrect. Indications are that they may not have starte d on SG level, but they do start on

FWI whe n the reactor is tripped and Tavg is less than 550°F. Plausible if applicant

mistakes the SG level AFWP start with the AMSAC setpoint of 8% and is confused with

what starts the AFW pumps (0110wing a Rx trip (i.e. low level vs. trip of both MFPs)

B. Correct. Both pumps will auto start from the feedwater isolation signal. DG 1A-A breaker

will open on emergency start and the 1A-A AFW pump will tie onto the SO BD following

return of voltage when the DG ties back on.

C. Incorrect. If 1 SG was less than 10.7%, this could be true if the reactor was not tripped.

D. Incorrect. Applicant may think that the board wi/I not reenergize following the LOOP since

the DG is paralleled to the board. This condition would only have the TDAFW pump

running.

(

NOTE:

TI-28 Art 9 pg 10/ 16

SG level <10.7% decreasing on 1 out of 4 SGs will initiate MDAFW;

SG level <10.7% decreasing on 2 out of 4 SGs will initiate TDAFW.

Monday, May 07,2007 3:22:59 PM

21

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXA M REV FINAL AS GIVEN 5-7-0 7

Ability tooperate and monitor the following as they applytoa reactor trip: AFW System

(

Question No.

Tier 1 Group 1

Importance Rating:

Technical Reference:

39

RO 4.4

TI-28 Att 9 [selpoints]

Proposed references to be provided to applicants during examination:

None

Learning Objective:

Question Source:

Question History:

OPT200 .AFW Objective 12.b

New

Question Cognitive Level:

Higher

10 CFR Part 55 Conlent:

41.7

Comments: Modified wording on C and 0 based on NRC com ments.

Time:

Points:

1.00

Version: o I 23 4 5 6 7 8 9

(

Answer: BB CD CADA CA

Scramble Range: A * D

Source:

NEW

Source If Bank:

Cognitive Level :

HIGH ER

Difficulty:

Job Position:

Plant:

SEQUOYAH

Date:

4/2007

Last 2 NRC?:

NO

Monday. May 07, 2007 3:22:59 PM

22

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

12. 007EA2.06 ool

(

Given the following plant conditions:

A manual reactor trip was attempted.

The following conditions exist:

Reactor Trip Breaker 'A' Red indication exists.

Reactor Trip Breaker 'B' Green indication exists.

Reactor Power is 3% and lowering.

Rod bottom lights are lit with the exception of FOUR control bank D rods.

Their positions are as follows:

H-8 - 16 steps

K-2 - 220 steps

M-12 - 8 steps

M-8 - 20 steps

Which ONE (1) of the following describes the condition of the reactor, and the action

that will be required?

A. The reactor is tripped ; perform nonnal RCS boration for the stuck rods as directed

in ES-0.1, Reactor Trip Response.

(

B~ The reactor is tripped; initiate emergency boration for the stuck rods in accordance

with EA 68-4, Emergency Boration, as directed in ES-0.1, Reactor Trip Response.

C. The reactor is not tripped; manually insert control rods as directed in FR-S.1,

Nuclear Power Generation/ATWS.

D. The reactor is not tripped; initiate emergency boration for the stuck rods in

accordance with EA 68-4, Emergency Boration, as directed in FR-S.1, Nuclear

Power Generation/ATWS.

A. Incorrect. Boration will be fhrough EA procedure, not normal borafion.

B. Correct. Power decrea sing and 1 RTB open means the reactor is tripped. ES-O. l

will direcf emergency borafion.

C. Incorrect. Indication is that reactor is tripped. If if was not, this would be correct.

D. Incorrect. Indication is that reactor is tripped. If not, these actions would occur, but

rods would also be inserted

Monday, May 07, 20073:22:59 PM

23

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

Abi~1y to detennine or interpret the following as theyapptyto a reactor tri9: OCCurrence d a reactor trip.

(

Question No.

Tier 1 Group 1

Importance Rating :

Technical Reference:

76

SRO 4.5

E-O. ES-o.1

Proposed references to be provided to applicants during examination:

None

Learnin g Objective:

Question Source:

Question History:

OPL271E-OObjective 4

New

Question Cognitive Level:

Higher

10 CFR Part 55 Content:

43.5

Comments: modified rx power to 3% to be less than FR-S.1 setpoint.

(

Time:

Points:

1.00

Version: o I 2 3 4 5 6 7 8 9

Answer: BADACADDCB

Scramble Range: A - 0

Source:

NEW

Source If Bank

Cognitive Level :

HIGH ER

Difficully:

Job Position:

Plant:

SEQUOYAH

Date:

4/2007

Last 2 NRC?:

NO

(

Monday, May 07, 2007 3:22:59 PM

24

(

QUESTION S REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

13. 007G2.1.I2001

Given the following plant conditions:

The Control Room was evacuated due to a fire.

The crew is preparing to cool down the RCS from the Aux Control Panel.

All Pressurizer Relief Tank indications were lost during the fire.

Which ONE (1) of the following describes the impact, if any, on Technical Specification

3.3.3.5, Remote Shutdown Instrumentation, of the loss of PRT indication?

A ~ Action will be entered for loss of PRT pressure indication ONLY.

B. Action will be entered for loss of PRT level indication ONLY.

C. Action will be entered for loss of PRT pressure and level indication.

D. Action entry is NOT required.

A. Correct. TS 3.3.3.5 Table 3.3-9, Remote Shutdown Monitoring Instrumentation,

Instrument PRT Pressure is required.

B. Incorrect. Plausible because TS requires PRT instrumentation but not level.

(

C. Incorrect. Only PRT pressure is included in TS 3.3.3.5

D. Incorrect. PRT pressure is included in TS 3.3.3.5

Conduct of Operations: Ability to apply tectncal specifications for 8 system.

Question No.

Tier 2 Group 1

Importance Rating :

Technical Reference:

86

SR0 4.0

TS 3.3.3.5

Proposed references to be provided to applicants during examination:

None

Learning Objective:

Question Source:

Question History:

OPT200PZRPCS Objective 6

New

(

Question Cognitive Level:

Higher

' 0 CFR Part 55 Content:

43.2

Comments:

Monday, May 07, 2007 3:22:59 PM

25

Scramble Range: A - 0

SEQUOYAH

NO

NEW

HIGHER

412007

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

Time:

I

Points:

1.00

Version:

0 I 2 3 4 5 6 7 8 9

Answer:

A D D D DA A B B D

Source If Bank:

Difficulty:

Plant:

Last 2 NRC?:

Source:

Cognitive Level:

Job Position:

Date:

(

Monday, May 07, 2007 3:22:59 PM

26

QUESTIO NS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

14. 007Kl.03 001

(

Given the following plant conditions:

A reactor trip has occurred.

RCS pressure is 1810 psig and lowering.

Containment Pressure is 1.5 psig and rising.

Whi ch ONE (1) of the following describes the status of RCP #1 seal leakofl?

Directed to...

A. VCT

B~ PRT

C. RCDT

D. Reactor Building Floor and Equipment Sump

A. Incorrecf. Normal f1owpath.

B. Correct.

(

C. Incorrect. Flow path of #2 seal.

D. Incorrect. The seal bypass is isolated but applicant may mistake this alignment

under these conditions.

(

Monday, May 07, 2007 3:22:59 PM

27

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

KnoM edge of Ihe physical connections and/or cause-effect relationshlps between the PRT5 and the followingsystems: Res.

(

Question No.

Tier 2 Group 1

Importance Rating:

Technical Reference:

7

R03.0

OPT200.CVCS, Slide 71

Proposed references to be provided to applicants during exam ination:

None

Learning Objective:

OPT200RCP Objective 3

Question Source:

Bank

Question History:

WTSI

Question Cognitive Level:

Higher

'0 CFR Part 55 Content:

41 .7

Comments: changed D distractor to seal bypa ss based on NRC com ents.

(

MeS

Time:

Points:

1.00

Version: o I 2 3 4 5 6 7 8 9

Answer: BACDACCDCD

Scramble Range: A - 0

Source:

BANK

Source If Bank:

Cognitive Level:

HIGHER

Difficulty:

Job Position:

Plant:

SEQUOYAH

Date:

412007

Last 2 NRC?:

NO

(

Monday, May 07, 2007 3:22:59 PM

28

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

15. 008 A4.09 00 1

Given the following plant conditions:

Unit 1 is operating at 100% RTP

0-XA-55-27B-B, Window A-5, "LETDOWN HX OUTLET FLOWfTEMP

ABNORMAL" alarms.

The RO determines that Letdown HX Outlet temperature is 132°F and rising

slowly.

Which ONE (1) of the following describes the event that has occurred, and the effect

on the unit prior to any action by the crew?

1-TCV-70-192, Letdown Heat Exchanger Temperature Control Valve......

A. temperature input is failing high. Mixed Bed demins have automatically bypassed.

B. temperature input is failing high. Letdown is aligned to the Mixed Bed demins.

C. temperature input is failing low. Mixed Bed demins have automatically bypassed.

D~ temperature input is failing low. Letdown is aligned to the Mixed Bed demins.

A. Incorrect. Incorrect input faifure, incorrect MB status- MB bypass occurrs at 137"F.

If MB bypassed at alarm SP, alignment would be this way.

B. Incorrect. Incorrect input faifure,correct MB status.

C. Incorrect. Correct input faifure, incorrect MB status.

D. Correct. if temperature input fails iow, the rev wiil close to raise temperature. If

temperature from the letdown HX rises, the mixed bed demins wiil auto bypass when

letdown temperature reaches 137"F.

Monday, May 07,20073:22:59 PM

29

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

Abi lity 10rnanuaIIy operate and/or monitel' in !he etrllrOl room: CON Temperalure controlwive.

(

Question No.

Tier 2 Group 1

Importance Rating:

Technical Reference:

8

R0 3.0

OPT200

Proposed references to be provided to applicants during examination:

None

Leaming Objective:

Question Source:

Question History:

OPT200.CVCS Objective 8

New

Question Cognitive Level:

Higher

10 CFR Part 55 Content:

41 .7

Comments: modified each answer to reference alarm setpoint vs. mixed bed bypass setpoint

based on NRC comments and made it a new question.

(

Time:

Source:

Cognitive Level:

Job Position:

Date:

Points:

NEW

HIGHER

4/2007

1.00

Version:

Answer:

o I 2 3 4 5 6 7 8 9

DBBDBB ACCD

Source If Bank:

Difficulty:

Plant:

Last 2 NRC?:

Scramble Range: A - D

SEQUOYAH

NO

Monday. May 07.20073:22:59 PM

30

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

16. 008 G2.4.4 001

Given the followi ng plant conditions:

- A LOCA has occurred.

- The crew has just entered E-1, Loss of Reactor or Secondary Coolant.

- The following parameters exist:

- All SG pressures - 730 psig and slowly trending down.

All SG levels - being controlled at 40% NR.

PRZR level - off-scale high.

- RVLlS Lower Range indicates 50%.

- Containment Pressure - 2.5 psig.

RWST level- 74% and decreasing slowly.

- RCS pressure - 875 psig and decreasing slowly.

- Highest CET - 500°F.

Based on these indications, which ONE (1) of the following procedures will the crew

enter next?

A. ES-1.1, SI Termination

B~ ES-1.2, Post LOCA Cooldown and Depressurization

(

C. ES-1.3, Transfer to RHR Containment Sump

D. E-2, Faulted Steam Generator Isolation

A-Incorrect. (see B- SI termination criteria not met).

B-Correct. RCS Pressure not stable, and low RCS inventory (low reactor vessel level

and high PRZR level indicates a large head bubble).

C-Incorrect. RCS pressure and RWST level are high. Entry to ES-l.3 on low RWST

level.

D-Incorrect. SG pressures are trending down because RCS temperature is trending

down and AFW flows maintaining levels.

(

Monday. May 07, 2007 3:22:59 PM

31

QUESTIONS REPORT

for SEOUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

EmergencyProceduresI Plan Abiity to rec::oglil:e abnormal indications lor system operating parc.neterr> whiChare entry-leve cen:1itions

for emergencyand abnormal operaling prooedlXes.

(

Question No.

Tier 1 Group 1

Importance Rating:

Technical Reference:

77

SR0 4.3

E-1

Proposed references to be provided to applicants during examination:

None

Leaming Objective:

OPL271 E-1 Objeclive 5

Question Source:

Bank

Question History:

North Anna 2006

Question Cognitive Level:

Higher

10 CFR Part 55 Content:

43.5

Comments:

SEQUOYAH

NO

(

Time:

Source:

Cognitive Level:

Job Position:

Date:

Points:

BANK

HIGHER

412007

1.00

Version:

Answer:

o I 2 3 4 5 6 7 8 9

BDDD C CDBD C

ScrambleRange: A- D

Source IfBank:

NORTH ANNA 2006 NRC

Difficulty:

Plant:

Last 2 NRC?:

Monday, May 07, 2007 3:22:59 PM

32

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

17. 010 K2.03 001

Which ONE (1) of the following describes the power supplies to the PZR PORVs and

their associated MCR position indicating lights?

PORVs

Indicating Lights

A.

120 VAC Vital Instrument Power

125 VDC Vital Battery Boards

B.

125 VDC Vital Battery Boards

120 VAC Vital Instrument Power

C~

125 VDC Vital Battery Boards

D.

120 VAC Vital Instrument Power

A. Incorrect. Credible due to low voltage source and120 v AC is used for control of the

SG PORVs and for indication of the PZR spray valves. If applicant confuses this with

the PORVs or the spray valves then he may select 120 v AC as correct answer.

B. Incorrect. Credible due to low voltage source, and partially correct.

C. Correct. Ul PORVs and indications are from 125V DC Battery Boards I and II.

D. Incorrect. Credible due to low voltage source

KnoMedge of bus powersupplies t) thefoloNing: Indicator for PORV position.

Question No.

Tier 2 Group 1

Importance Rating:

Technical Reference:

9

R0 2.8

45N601-4

Proposed references to be provided to applicants during examination:

None

Learning Objective:

Question Source:

Question History:

OPT200.PZRPCS Objective 4.b

New

(

Question Cognitive Level:

Lower

10 CFR Part 55 Content:

41.7

Comments:

Monday, May 07, 2007 3:23:00 PM

33

Scramble Range : A * D

SEQUOYAII

NO

NEW

LOWER

4/2007

QUESTIONS REPORT

for SEOUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

Time:

I

Points:

1.00

Version:

0 I 2 3 4 5 6 7 8 9

Answer:

C ACCCA C B BC

Source IfBank:

Difficulty:

Plant:

Last 2 NRC?:

Source:

Cognitive Level:

Job Position:

Date:

(

Monday, May 07, 20073:23:00 PM

34

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

18. 011 EA2.10 001

Given the following plant conditions:

A Reactor Trip and Safety Injection have occurred.

Containment pressure is 5.6 psig and stable.

RCPs have been stopped.

RVLlS Lower Range is indicating 35%.

Core Exit Thermocouples are indicating 710°F.

PZR level is off scale low.

PZR pressure is 400 psig.

RCS Wide Range Hot Leg Temperatures are indicati ng 680°F.

Which ONE (1) of the following conditions could currently exist, and which procedure

would be required?

A. A PZR vapor space break has occurred and a transition to FR-C.l , Response to

Inadequate Core Cooling, is required.

B. A PZR vapor space break has occurred and a transition to FR-C.2, Response to

Degraded Core Cooling , is required.

C~ An RCS hot or cold leg break has occurred and a transition to FR-C.l , Response to

Inadequate Core Cooling , is required.

D. An RCS hot or cold leg break has occurred and a transition to FR-C.2, Response to

Degraded Core Cooling , is required.

A. Incorrect. Incorrect event, correct procedure entry.

B. Incorrect. Incorrect event, incorrectprocedure entry.

C. Correct. Requires applicant to disfinguish between Orange and Red Path on Core

Cooling CSF. Also requires determination ofbreak location by determining that PZR

level is not off-scale high.

D. Incorrect. Correct event, incorrect procedure entry

Monday, May 07, 2007 3:23:00 PM

35

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

Ability10 detetmine or interpret Ihe followingas theyapplytoa Large Bfeak l OCA: Vefifteation of adequate core cooling

(

Question No.

Tier 1 Group 1

Importance Rating:

Technical Reference:

78

SRO 4.7

CSFSTs Core Cooling F.02

Proposed referen ces to be provided to applicants during examination:

None

Learning Objective:

OPL271FR-C.1 Objeclive 2.b

Question Source:

Modified

Question History:

WTSI Various (Harris)

Question Cognitive Level:

Higher

10 CFR Part 55 Content:

43.5

Comments:

(

Time:

Source:

Cognitive Level:

Job Position:

Date:

Points:

MODIFIED

HIGH ER

4/2007

1.00

Version:

Answer:

o I 2 3 4 5 6 7 8 9

CAA BCC DDDD

Source If Bank:

Difficulty:

Plant:

Last 2 NRC?:

Scramble Range: A - D

SEQUOYAH

NO

Monday, May 07,2007 3:23:00 PM

36

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

19. 011 EK3.14 001

Given the following plant conditions:

A LOCA has occurred.

Safety Injection has failed to actuate.

The crew is performing E-O, Reactor Trip or Safety Injection.

RCS pressure is 100 psig and lowering.

Containment pressure is 5.5 psig and rising.

The crew is attempting to initiate ECCS flow.

Which ONE (1) of the following describes the operation of RCPs for this event?

A. Maintain RCPs running to provide core cooling until SI is actuated.

B. Maintain RCPs running to prevent phase separation of RCS liquid.

C. Trip all RCPs to minimize mass flow out of the RCS break.

D~ Trip all RCPs to prevent damage due to loss of cooling water flow.

A. Incorrect. Plausible because Rep Trip criteria for Westinghouse SBLOCA requires SI or

Charging Pumps running to allowtrip of Reps

B. Incorrect. Reps will be tripped, butplausible because the Westinghouse analysis for Rep

trip on SBLOCA indicates that phase separation is a concern ifReps are not tripped in a

timely manner, and are subsequently tripped later in the event due to loss of power or

otherreason

C. Incorrect. Correct operation of pumps, buta LBLOGA is in progress. SBLOCA analysis is

concerned with mass beingpumped out of the break, consistent with B above.

D. Correct. Phase B has occurred and CCS to Reps is lost.

Reps must be stopped to

prevent running with no cooling water.

EOP Basis: Item #1: RCP Trip Criteria *...RCPs must also be stopped upon actuationof a

Phase B isolation due to loss of CCS cooling water. Continued pump operation without

cooling water willresult in pump damage.

Monday, May 07, 2007 3:23:00 PM

37

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

Kna.Medge of the reasons lorthe following responses as the apply to the LargeBreak l OC A; Re p tripping requirement

(

Question No.

Tier 1 Group 1

Importance Rating:

Technical Reference:

40

R04.1

E-O: EPM-3-E-0 Foldout Page Ilem: ERG Executive Volume.

RCP Trip/Restart

Proposed references to be provided to applicants during examination:

None

Learning Objective:

Question Source :

Bank

Question History:

WTSI Bank

Question Cognitive Level:

Higher

10 CFR Part 55 Conlent:

41.5

Comments:

(

Mes

Time:

Source:

Cognitive Level:

Job Position:

Date:

Points:

BANK

HIGHER

412007

1.00

Version:

0 I 2 3 4 5 6 7 8 9

Answer:

D O C A A C D A D A

Scramble Range: A - 0

Source If Bank:

WTSI

Difficulty:

Plant:

SEQUOYAJ-I

Last 2 NRC?:

NO

Monday. May 07, 2007 3:23:00 PM

38

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

20. 012KI.04 001

Given the following plant conditions:

A secondary steam leak has occurred with Control Bank D initially at 220 steps.

Over the past 5 minutes, reactor power has been slowly rising, followed by a

sharp spike from 104% to approximately 109%.

All equipment has ope rated as designed in response to the transient.

W hich ONE (1) of the following describes the Control Bank D Rod Position and Group

Demand indications following the termination of this event?

A. Rod Bottom Lights extinguished; Group Demand indication at 220 steps.

B. Rod Bottom Lights extinguished; Group Demand indication less than 220 steps.

C~ Rod Bottom Lights illuminated; Group Demand indication less than 220 steps.

D. Rod Bottom Lights illuminated; Group Demand indicatio n at 220 steps.

A. Incorrect. Rod bottom lights will be on. Runback and reactor trip should occur {see

C].

B. Incorrect. Rod bottom lights will be on. Runback and reactor trip should occur {see

C].

C. Correct. Rod Bottom lights would all be illuminated following the reactor trip. OPDT

runback would have demanded rod insertion {setpoint: 1% below the reactor trip

setpoint], prior to the reactor trip resulting in Control Bank 0 {CB 'D' Groups 1 & 2

indicators l -XI-85-5105Dl & l-XI-85-5105D2} demand being lower. Also, bank

demand is lower because control rods insert in response to the turbine runback load

reduction, Tave above Tref. The power spike to 109% would cause a reactor trip

{setpoint: 2/4 PRNls ~ 1 09% reactor trip setpoint}.

D. Incorrect. Rod Bottom lights would all be illuminated following the reactor trip.

Although CB 'D' rod stop is at 220 steps, it is an outward motion stop ONLY. CB 'D'

Groups 1 & 2 indicators will be below 220 steps indicated from the runback prior to

the reactor trip.

Monday. May 07,20073:23:00 PM

39

QUESTIONS REPORT

for SEQ UOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

Knowledge of tile physiCal connectiMs andJOfcause effect relationships between the RP$ and the following systems: RPIS.

(

Question No.

,,

Tier 2 Group 1

Importance Rating:

R0 3.2

Technical Reference:

(TI-28)

Proposed references to be provided to applicants during examination :

None

Learning Objective:

OPT200 .RPS. B.3 & 4

Question Source:

New

Question History:

Question Cognitive Level:

Higher

10 CFR Part 55 Content:

41.6

Scramble Range: A - D

SEQUOYAH

NO

NEW

HtGHER

4/2007

Source:

Cognitive Level:

Job Position:

Date:

Comments: Modified B and C wording based on NRC comment. Removed "Rod Position

Indication below 220 steps" from A and B based on NRC comment.

Time:

I

Points:

1.00

Version:

0 1 2 3 4 5 6 7 8 9

Answer:

C B B B B AC DC D

Source IfBank:

Difficulty:

Plant:

Last 2 NRC?:

(

Monday, May 07, 2007 3:23:00 PM

40

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

21 . 012 K2.01001

Given the following plant conditions:

Unit 1 is at 100% power.

A loss of 125 VDC Vital Battery Board I occurs.

Which ONE (1) of the following describes the effect on the associated Reactor Trip

Breaker?

A~ Reactor Trip Breaker indication on the MCB is losl and RTA is NOT capable of

tripping on a SHUNT lrip.

B. Reaclor Trip Breaker indication on the MCB is lost and RTA lrips OPEN.

C. Reactor Trip Breaker indication on the MCB remains IiI and RTA trips OPEN.

D. Reactor Trip Breaker indication on the MCB remains IiI and RTA is NOT capable of

tripping on a SHUNT trip.

A. Correct. 125 VDC Vital Battery Board I supplies control power, and indication power,

to RTA and also shunt trip mechanism. Without this power, M-4 indication is lost and

the shunt trip feature of RTA is disabled- the RTA will not open from the loss of 125

VDC control power.

B. tncorrect. Correct indication, but loss of power to shunt coil will not cause RTA to

trip. Plausible since UV coil losing power [from 120 VAC] will cause the reactor trip

breakers to open.

C. Incorrect. RTA indication is not available at M-4 or at other locations in the MCR or

from other sources orpower supplies but applicant may assume indication is from a

different source. Loss of power to shunt coil will not cause RTA to trip. Plausible

since UV coil losing power [from 120 VAC] will cause the reactor trip breakers to

open.

D. Incorrect. Indication is not available on MCB; losing power to shunt coil will not

cause RTA to open.

Monday, May 07. 2007 3:23:00 PM

41

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

KnoMedgeof bus powersupplies tothe following: RPS chMnels. components. and Inlercormections.

(

Question No.

10

Tier 2 Group 1

Importance Rating:

RO 3.3

Technical Reference:

0-45N706*3

Proposed references to be provided to applicants during examination:

None

Learning Objective:

OPL271AOP-P.02. Obj 2.d

Question Source:

Bank

Question History:

Question Cognitive Level:

Lower

10 CFR Part 55 Content:

41.7

SEQUOYAH

NO

BANK

LOWER

4/2007

Source:

Cognitive Level:

Job Position:

Date:

Comments: modified B and C based on NRC comments.

Time:

I

Points:

1.00

Version:

0 I 2 3 4 5 6 7 8 9

Answer:

A A B A CAB D A B

Scramble Range: A * D

Source If Bank:

WTSI

Difficulty:

Plant:

Las12 NRC! :

(

Monday, May 07, 2007 3:23:00 PM

42

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

22. 013 K3.03 001

Given the following plant conditions:

A Loss of Off-Site Power has occurred.

A steam line break generated a safety injection signal but only Train B of Safety

Injection actuated.

Containment pressure indicates 2.9 psig and rising.

Assuming no manual actions have been initiated, which ONE (1) of the following

correctly describes a consequence of the failure of Train A of Safety Injection to

actuate?

A. 1A-A CCP will NOT be sequenced onto the Shutdown Board.

B. Only two Main Steam Isolation Valves (MSIV's) will be closed.

C. Only one Phase B Containment Isolation Valve in each penetration will be closed.

D~ Only one Phase A Containment Isolation Valve in each penetration will be closed.

A. Incorrect. 'A'train loads willbe loaded in blackout mode, including 1A-A CCP . Applicant

may recognize that the EDG will automatically start but incorrectly think a loss of the A train

ESFAS willprevent the load sequencing and not load/start 1A-A CCP.

B. Incorrect. 2 MSIVs get signals from 1 of 2 slave or 'K' relays from each train; both trains

function this way; so only 2 closing is incorrect. Plausible since a failure of 1 of 2 'A I train K

relays for the MSIVs will prevent two valves from closing on an actuation signal. This is

consistent with what the applicantshave seen on the simulator.

c. Incorrect. Both trains of phase B CIVs will isolate, Applicant may think the ESFAS failure

will will affect the phase B isolation.

D. Correct. Failure of 1 train of ESFAS will result in 1 group of CIA valves closing.

Monday, May 07,20073:23:00 PM

43

QUESTIONS REPORT

for SEQUO YAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

Kno.vIedgeof the effect that a loss or malfunction d ~e ESFAS winhave on lhe foIloMng : Containmenl

(

Question No.

Tier 2 Group 1

Importance Rating:

Technical Reference:

12

R04.3

47W611-88-1, 63-1; TI-28 AUg BfO Load Sequence, page 5

Proposed references to be provided to applicants during examination:

None

Learning Objective:

OPT200.CtmtStruclure

Question Source:

Bank

Question History:

Salem 2002

Question Cognitive Level:

Higher

10 CFR Part 55 Content:

41,7

Comments:

SEQUOYAII

NO

(

Time:

Source:

Cognitive Level:

Job Position:

Date:

Points:

BANK

HIGHER

4/2007

1.00

Version:

Answer:

o I 2 3 4 5 6 7 8 9

D C CBD C CCB A

Scramble Range: A* O

Source If Bank:

SALEM 2002

Difficulty:

Plant:

Last 2 NRC?:

Monday, May 07, 2007 3:23:01 PM

44

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

23. 01S AAl.1O00 1

Given the following plant conditions:

A malfunction has occurred that resulted in a reduction of Component

Cooling Water (CCS) flow to the RCPs.

The US has entered AOP-M.03, Loss of Component Cooling Water.

The OATC is referring to AOP-R.04, Reactor Coolant Pump Malfunctions.

Current parameters for ALL RCPs are as follows:

RCP Lower Bearing temperature is 222°F and rising.

RCP Seal Wate r Outlet temperature is 228°F and rising.

Based on current conditions, which ONE (1) of the following describes the action

required in relation to the RCPs?

A. Initiate a plant shutdown to Mode 3.

Trip RCPs based upon RCP Lower Bearing temperature exceedi ng an operating

limit.

B. Trip the reactor and enter E-O, Reactor Trip or Safety Injection.

Trip RCPs based upon RCP Lower Bearing temperature exceeding an operating

limit.

C. Initiate a plant shutdown to Mode 3.

Trip RCPs based upon RCP Seal Water Outlet temperature exceeding an

operating limit.

D~ Trip the reactor and enter E-O, Reactor Trip or Safety Injection.

Trip RCPs based upon RCP Seal Water Outlet temperature exceeding an

operating limit.

A. Incorrect. Bearing is not exceeding a limit. Plausible since if the applicant applies

the tech spec methodoiogy of an inoperable RCP, then the plant would have to be

in HSB in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

B. Incorrect. Wrong bearing exceeding limit, but correct actions.

C. Incorrect. Correctlimit, incorrect actions.

D. Correct. RCP seal water outlet limit is 225°F.

Monday, May 07, 2007 3:23:01 PM

45

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

Abilitytodetennine and interpret the following as they applyto the Reactor CooI¥It PlJTlp Malhn:tions (Loss d RC Flow): When kl

secze Reps on loss of cooling or seal if1ec:tioo

(

Question No.

Tier 1 Group 1

Importance Rating:

Technical Reference:

41

R0 3.7

AOP-M.03 R11. AOP-R.04 section 2.1.

Proposed references to be provided to applicants during examination:

None

Learning Objective:

Question Source:

Question History:

OPL271AOI-M.03 Objective 8

Modified

Question Cognitive Level:

Higher

10 CFR Part 55 Content:

41.10

Comments : Modified stem based on NRC comments. Added R.04 to stem for Re p trip

criteria monitoring. AOP-M.03 does not contain the Rep trip criteria.

(

Time:

Source:

Cognitive Level:

Job Position:

Date:

Points:

NEW

LOWER

412007

I.()()

Version:

Answer:

o I 234 5 6 7 8 9

DCDABD ACC B

Source If Bank:

Difficulty:

Plant:

Last 2 NRC?:

Scramble Range: A - 0

SEQUOYAH

NO

Monday, May 07, 2007 3:23:01 PM

46

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

24. 0' 5 Kl.08 001

Given the following plant conditions:

Unit 1 is in Mode 5

Operators are preparing for a plant heatup.

Which ONE (1) of the following describes the MiNIMUM requirement for monitoring

Source Range indications while starting RCPs, and the reason why, in accordance

with the precautions and limitations of 1-50-68-2, RCP Operations?

A. Monitor indications during startup of the first RCP to monitor for indications of a

cold water transient.

B~ Monitor indications during startup of the first RCP to monitor for indications of a

boron dilution event.

C. Monitor indications during startup of all RCPs to monitor for indications of a cold

water transient.

D. Monitor indications during startup of all RCPs to monitor for indications of a boron

dilution event.

A. Incorrect. Correct action butreason isincorrect. Applicantmay mistake this

precaution with a different precaution in 50-68-2 for circulating cold water on the

start of the first RCP but this precaution is tied to a pressure transient vice a

reactivityaddftion. Both would cause a positive reactivity addition.

B. Correct. Per 50-68-2 P&L L: 'The Source Range Monitor instrumentation must be monitored

during startup of the first Rep. This precaution is necessaryin order to detect rapid boron dilution

from unborated water that could be flushed into the cere."

Per SO-68-2 P&L M: "Ifaf! Rep's are shutdown, do not start the first Rep if it is known that any

volume of dilute water has been introduced into any reactor coolant loop untitan actionplan for

recovery is developed as described in Appendix F Prevention of Rapid Boron Dilution...

C. Incorrect.

P er 50-68-2 P &L F : *Only performed for first RCP start and is concerned with

boron dilution. Regarding an RCP start-eold water concern, 50-68-2 When RCS temperature is

greater than chargingand seal injection water temperature and RCPs have been idle >5 minutes, a

steam bubble is needed in the pressurizer priorto starting an RCP. This precaution willminimize the

pressure transient when the cold water injected by the chargingpump is circulated in the RCS."

D. Incorrect. Only performed for first RCP start per the SO.

Monday, May 07,20073:23:01 PM

47

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

Know1edgeof !he phr-; icaI connections andIor ceose-ettect relationships bet\\oieen!he NIS and the folloMng systems: RCS (pump start).

(

Question No.

TIer 2 Group 2

Importance Rating:

Technical Reference:

30

R02.6

1-50-68-2

Proposed references to be provided to applicants during examination:

Learning Objective:

Question Source:

Question History:

OPT200,RCP Objective 5a

New

Question Cognitive Level:

Lower

10 CFR Part 55 Content:

41.5

Comments: Editorially mod ified A and C wording transient vs. accident.

Mes

Time:

Points:

1.00

Version: o I 2 3 4 5 6 7 8 9

(

Answer:

BBB CDADBCB

Scramble Range: A - 0

Source:

NEW

Source If Bank:

Cognitive Level:

LOWER

Difficulty:

Job Position:

Plant:

SEQUOYAil

Date:

412007

Last 2 NRC? :

NO

(

Monday, May 07, 2007 3:23:01 PM

48

(

QUESTIONS REPORT

for SEOUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

25. 017A4.02 001

Given the following plant conditions:

A LOCA has occurred.

RCS pressure is 200 psig.

Due to equipment failures, the crew has transitioned to FR-C.1, Response to

Inadequate Core Cooling.

Which ONE (1) of the following describes operation of the RCPs for this event?

A. RCPs are only started following transition to SACRG-1.

B. All available RCPs are started prior to performing secondary depressurization.

C~ If secondary depressurization is ineffective as determined by Core Exit

Thermocouple temperature greater than 1200'F, then start one RCP at a time, if

available.

D. Following secondary depressurization and when its associated SG NR level is

greater than 10% and support conditions are established per EA-68-2, Establishing

RCP Start Conditions, then start each available RCP.

A. Incorrect. Reps are started one at a time but the start criteria is 1200 vs 700 of . 700 OF

is the temperature to enter FR-G.1 coincident with loss of RVLlS level <42% and is also

used in FR-C.1 to determine if actions are effective.

B. Incorrect. While forced coofing is desirable, Reps are no t running when secondary

depressurization is initiated.

C. Correct. One Rep at a time is started if GETs remain >1200 OF following secondary

depressurization until all available Re ps are runn ing. To temporarily restore core cooling,

the opera tor is instructed to start Re ps one at a time until core exit res < 1200°F. The

Reps should force two phase flow through the core, temporarily keeping it cool. Even

single phase forced steam flow will coo l the core for some time provided the RCPs can be

kep t running and a heat sink is available.

D. Incorrect. early in FR-C.1, operators are directed to attempt to *establish RCP start

conditions, but RCPs are not started until secondary depressurization is determined to be

unsuccessful much later in the C-1. Subsequantiy, when C-1 directs RCP start, SG(s)

availablility is met if secondary side >10%; then RCP started regardless of EA-68

conditions . RCP damage due to absence or loss of normal support conditions is an

acceptable consequence in this procedure.

Monday, May 07, 2007 3:23:01 PM

49

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5*7*07

Abi~ty to manually operate and/ormonitar in the controlroom: Temperature vaIoesusedtodetermineRCSJRCP operation durng

inadequate core cx:ding (i.e., if applieable. average d fi..-e hiItJe5t values ).

(

Question No.

Tier 2 Group 2

Importance Rating :

Technical Reference:

31

R0 3.8

FR-C.1; EPM-3-FR-C.1

Proposed references to be provided to applicants during examination:

None

Learning Objective:

Que stion Source:

Question History:

OPL271FR-C.1 Objective 3

New

Question Cognitive Level:

Lower

10 CFR Part 55 Content

41.10

Comments: Editorial change to D Distractor and C answer wording.

(

Time:

Points:

1.00

Version: o I 23 4 5 6 7 8 9

Answer:

CCAABCABAD

Scramble Range: A - D

Source:

NEW

Source If Bank:

Cognitive Level:

WWER

Difficulty:

Job Position:

Plant:

SEQUOYAII

Date:

412007

Last 2 NRC?:

NO

(

Monday. May 07, 2007 3:23:01 PM

50

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

2 6 . 022 A2.03 001

Given the following plant conditions:

Unit 1 is at 100% power.

The followi ng alarm is received:

1-AR-M6E-E5. MOTOR TRIP OUT PNL 1-M-9

Lower Compartment Cooling Fan 1B-B control switch has a white light

illuminated.

Fans 1A-A and 1C-A are running.

Which ONE (1) of the following describes the primary concem related to this failure,

and the procedural action(s) required?

A. Lower Ice Condenser doors may open due to high DP: Stop one Upper

Compartment Cooling fan to equalize pressure in accordance with 1-AR-M6E-E5.

B. Lower Ice Condenser doors may open due to high DP: Start any Lower

Compartment Cooling fan in standby using the alarm response and 0-SO-30-5,

Lower Compartment Cooling.

C. PZR Enclosure may heat up and cause PZR Safety Valves to leak; Isolate the

tripped cooler TCV and bypass the in-service Lower Compartment Cooler TCVs to

increase cooling to the PZR Enclosure in accordance with 1-AR-M6E-E5.

D~ PZR Enclosure may heat up and cause PZR Safety Valves to leak; Start any Lower

Compartment Cooling fan in standby using the alarm response and 0-SO-30-5,

Lower Compartment Cooling.

A. Incorrect. Containment DP is a concern forother ventilation systems such as purge, or an

event such as a LOGA.

B. Incorrect. Containment DP is a concern for other ventilation systems such as purge, or an

event such as a LOCA. Action to start a fan is correct

C. Incorrect. rev for the tripped fan is allowed to go open. Not isolated

D. Correct.

Three fans should be running. With one in standby, it may be started.

Monday, May 07, 2007 3:23:01 PM

51

(

QUESTIONS REPORT

for SEQ UOYAH 2007 - NRC EXAM REV FINAL AS G IVEN 5-7-07

Abil ity to (a) predict the impacts 01 the followingmatfunclicns or opemtio1s on Ihe CCS: and (b) based on those predictions . use

procedures tocorrect. control, or:mitiga te the coosecoe-ces of thase malfunctionsor operations: Fan motor Ihemlal ovef1oadIhigl-speed

_.

Question No.

Tier 2 Group 1

Importance Rating:

Technica l Reference:

87

SRO 3.0

AR-M6E, E5; 0-50-30-5

Proposed references to be provided to applicants during examination:

None

Learning Objective:

Question Source:

Question History:

OPT200 .CONTCOOLING Objective 7

New

Question Cognitive Level:

Higher

10 CFR Part 55 Content:

43.5

(

Comments:

Time:

Source:

Cognitive Level:

Job Position:

Date:

Points:

NEW

HIGHER

412007

1.00

Version:

Answer:

o I 23 4 5 6 7 8 9

DDCDDDAAB D

Source If Bank:

Difficulty:

Plant:

Last 2 NRC?:

Scramble Range: A - 0

SEQUOYAH

NO

Monday, May 07, 2007 3:23:01 PM

52

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

27. 022M .OI 001

Given the following plant conditions:

Reactor Trip & Safety Injection occurred due to Large Break LOCA.

Containment Phase B isolation has occurred

All systems responded normally

Which ONE (1) of the following describes the response of the Lower

Compartment Coolers whe n Phase B is reset?

A. Incorrect. Fans will start. Valves will remain closed

(

Fans in A*P AUTO

A. remain off

B~ start

C. remain off

D. start

Cooler ERCW Valves

Cooler ERCW containment

isolation valves open.

Cooler ERCW containment

isolation valves remain closed.

Cooler ERCW containment isolation

valves remain closed.

Cooler ERCW containment isolation

valves open.

(

B. Correct.

C. Incorrect. Fans will start but valve position is correct.

D. Incorrect. Correct fan operation but valves will not open

ERCW does not realign on Phase B reset. Fans in A-P AUTO will start as the Phase B

is cleared

Monday. May 07, 2007 3:23:01 PM

53

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

Ability to manuallyoperate andIa monitorirl lhe rontrol room: ccs fans.

(

Question No.

Tier 2 Group 1

Importance Rating:

Technical Reference:

13

R03.6

1-47W611-67-3,1-47W61 1-30-4

Proposed references to be provided to applicants during examination:

None

Learning Objective:

OPT200.CONTCOOLlNG, Obj. B.4

OPT200.ERCW, Obj B.4

Question Source:

Bank

Question History:

Sequoayah Bank

Question Cognitive Level:

Lower

10 CFR Part 55 Conlent:

41.4

Comments:

(

Time:

Source:

Cognitive Level:

Job Position:

Date:

Points:

BANK

LOWER

4/2007

1.00

Version:

0 I 2 3 4 5 6 7 8 9

Answer.

B A C B A BC CAe

Scramble Range: A - 0

Source If Bank:

SEQUOYAll BANK

Difficulty:

Plant:

SEQUOYAH

Lasl2 NRC?:

NO

Monday. May 07, 2007 3:23:01 PM

54

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

28. 022 AKJ.03 001

Given the following plant conditions:

Unit 1 is at 100% RTP whe n a smail RCS leak develops.

The operating crew takes the necessary actions to stabilize PZR level.

Subsequently, the crew attempts to determine the leakage source.

After isolating charging and letdown, the crew observes the foilowing

parameters/indications:

Letdown flow 0 gpm;

Cntmt pressure approximately 0.3 and decreasing;

RCS pressure approximately 2235 psig and stable;

pzr level approximately 82% and increasing;

Charging flow 40 gpm;

Cntmt radiation monitors trending downward slowly.

Which ONE (1) of the following actions will the operating crew perform next?

A. Immediately trip the reactor and enter E-O, Reactor Trip or Safety Injection because

PZR level exceeds the maximum limit.

B~ Place excess letdown in service and adjust seal injection flow to maintain PZR level

within limits.

C. Initiate Emergency Shutdown per AOP-C.03, Emergency Shutdown due to RCS

leakage exceeding limits.

D. Re-establish normal letdown and dispatch leak search team to identify leakage

source.

A. Incorrect. Applicant may mistake 82% Pzr level for the reactor trip value of 92%; however,

the initial conditions do not require an immediate reactor trip.

B. Correct. The ReS leak was determined to be on norma/letdown based on containment

pressure decreasing and radiation monitor readings stabilized. Therefore to maintain

pressurizer level and ReS chemistry (chemical feed and boretion), excess letdown is

required.

C. Incorrect. Since pressurizer level is stabilized and is controllable by placing Excess

Letdown in service, the other initialconditions indicated the RCS leak has been isolated a

rapid shutdown is not requiredper AOP-R.05.

D. Incorrect. The leak has been isolatedto the normal letdown line. Therefore, a leak search

team is unnecessary.

The student mustunderstand from the initiaf conditions the RCS leak was on the letdown fine

and has been isolated. In order to maintainpressurizer fevef and chemistryexcess

letdown is required.

Monday. May 07. 2007 3:23:02 PM

55

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

l<ncI>Medge of the reasonsfor the IoIIoMng responses as they app/yto the Loss of Reaclor Coolant Pump Makeup: Perfoimanceof lineup

to establish excess IeldoMl after determining need

(

Question No.

Tier 1 Group 1

Importance Rating:

Technical Reference:

42

R0 3.1

AOP-R.05

EA-62-3

Proposed refe rences to be provided to applicants during examination:

None

Learning Objective:

OPL271C367, Objectives Obj 2, 4

Question Source:

Bank

Question History:

AOP R.05-B.2-6

Question Cognitive Level:

Higher

10 CFR Part55 Content:

41 .10

Comments: Modified 0 distractor.

SEQUOYAH

NO

(

Time:

Source:

Cognitive Level:

Job Position:

Dale:

Points:

BANK

IllGHER

4/2007

1.00

Version:

Answer:

o 1 2 3 4 5 6 7 8 9

B A A C A AA DB C

Scramble Range:A - D

Source If Bank:

SEQUOYAH BANK

Difficulty:

Plant:

Lasl2 NRC?:

Monday. May 07. 2007 3:23:02 PM

56

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

29. 025 A2.01 00 1

Given the followi ng plant conditions:

Unit 1 is at 100% power.

The following alarm is received:

1-AR-M6E, C7, FLOOR COOLANT DP LO

Which ONE (1) of the following describes the potential effect of this failure, and the

action required to mitigate the effect?

M

1-AR-M6E, 66, FLOOR COOLANT TEMP HI, will alarm: Ensure at least 1 Glycol

Floor Circ Pump is running in accordance with 1-AR-M6E, C7.

6. 1-AR-M6E, 66, FLOOR COOLANT TEMP HI, will alarm: Swap Glycol Circ Pumps

and chiller packages in accordance with 1-S0-61-1, Ice Condenser.

C. 1-AR-M6E, 67, FLOOR COOLANT TEMP LO, will alarm: Ensure at least 1 Glycol

Floor Circ Pump is running in accordance with 1-AR-M6E, C7.

D. 1-AR-M6E, 67, FLOOR COOLANT TEMP LO, will alarm: Swap Glycol Circ Pumps

and chiller packages in accordance with 1-S0-61-1, Ice Condenser.

A. Correct: If DP is low, then the floor coolingpumpmusthave tripped, and temperature will

rise

B. Incorrect. Would swap Glycol Pumps and chillers ifthe Glycolpump tripped.

C. Incorrect. Wrong direction on temperature. Action is correct.

D. Incorrect.

Wrong direction on temperature Action is incorrect. Would be taken if a Glycol

e ire Pump tripped.

Monday, May 07. 2007 3:23:02 PM

57

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

Abilityto (a) predict the impactS of the fo/lolMng malfunctions or operalions on the ice condenser system; correct, contrct or mitigate the

ccoseoeences of ecee malfunctions or operations: Trip of g/)<:oI eircuIatial pumps .

(

Question No.

Tier 2 Group 1

Importance Rating:

Technical Reference:

88

SRO 2.7

AR-M6 -E, C-7, B-6:

Proposed references to be provided to applicants during examination:

None

Learning Objective:

Question Source:

Question History:

OPT200.ICE Objective 5.d

New

Question Cognitive Level:

Higher

10 CFR Part 55 Content:

43.5

Comments:

(

Time:

Source:

Cognitive Level:

Job Position:

Date:

Points:

NEW

HIGHER

412007

1.00

Version:

Answer:

o 1 2 34 5 6 7 8 9

ADA CB CCAD A

Source If Bank;

Difficulty:

Plant:

Last 2 NRC?:

Scramble Range: A - 0

SEQUOYAII

NO

Monday, May 07, 2007 3:23:02 PM

58

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5*7*0 7

30. 025 A2.06 001

Given the followi ng plant conditions:

Unit 1 is at 100% power.

Ice Condenser Floor Glycol Temperature Control Valve 1-TCV*61-71 setpoint

fails.

Ice Condenser floor temperature is 12°F and lowering.

Which ONE (1) of the following describes the potential effect of this failure, and the

action(s) required in accordance with 50-61-1, Ice Condenser?

A. Excessive sublimation of the ice bed; throttle closed 1-TCV-61-71 Bypass valve

1-61-751, Temperature Control Valve Bypass Valve.

B. Excessive sublimation of the ice bed; close 1-TCV-61-71 isolation valve and

operate 1-61-751 , Temperature Control Valve Bypass Valve, to control

temperature.

C. Ice Condenser door binding; throttle closed 1-TCV-61-71 Bypass valve 1-61-751,

Temperature Control Vaive Bypass Valve.

D~ Ice Condenser door binding; close 1-TCV-61-71 isolation valve and operate

1-61-751, Temperature Control Valve Bypass Valve, to control temperature.

A. Incorrect. Wrong effect, wrong direction for the valve under the conditions

presented. Would control temperature if valve was open, but valve is normally closed.

B. Incorrect. Wrong effect. Correct actions

C. Incorrect. Correct effect. Wrong direction for the valve under the conditions

presented

D. Correct. Temperature swings can cause buckling of the Ice Condenser wear slab,

causing doors to jam. Sublimation happens when temperature is too high. High glycol

flow would cause temperature to be reduced. By isolating the TCV and throttling open

on the bypass valve glycol flow is reduced, allOWing the floor temperature to rise.

Monday, May 07, 20073:23:02 PM

59

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

Ability to (a) predict the impacts d the following malfunctions or operations on the ice condenser system; COITect. con lrol. or mitigate the

consequences 01 thosemafunctions orcceaecos: Decreasingice eeooeose- temperature.

(

Question No.

Tier 2 Group 1

Importance Rating :

Technical Reference:

14

R02.5

50-61-1 , AR-M6E-B7

Proposed references to be provided to applicants during examination:

None

Learning Objective:

Question Source:

Question History:

OPT200.lCE Objective 5

New

Question Cognitive Level:

Higher

10 CFR Part 55 Content:

41 .10

Comments: Editorial change to answers based on NRC comment

(

Mes

Time:

Source:

Cognitive Level:

Job Position:

Date:

Points:

NEW

HIGHER

412007

1.00

Version:

Answer:

o I 2 34 5 6 7 8 9

DDDACDD ACC

Source If Bank:

Difficulty:

Plant:

Last 2 NRC?:

Scramble Range: A * D

SEQUOYAH

NO

Monday, May 07, 2007 3:23:02 PM

60

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

31. 025AKI.Ol 001

Given the following plant conditions:

Unit 2 is in Mode 6.

RCS drained to elevation 701 feet.

Upper internals are installed .

RCS temperature is 110°F.

Core decay heat is approximately 5 MW.

A total loss of RHR cooling has occurred.

Which ONE (1) of the following is correct concerning the amcunt of time it will take to

reach 200°F in the RCS using 0-GO-15. App. U. Time for RCS to reach 200°F?

Reference Provided

A. 26 minutes

B~ 32 minutes

C. 43 minutes

D. 306 minutes

A. Incorrect. Equation 1 was incorrectly applied to determine value.

Time = 1460[(200-110)/5000KW] = - 26 min

B. Correct. Apply equation 2 of 0-GO-15. Appendix U.

Time = 1780[(200-110)/5000KW] = - 32 min

C. Incorrect. Equation 3 was incorrectly applied to determine value.

Time =2390[(200-110)/5000KW] =- 43 min

D. tncottect. Equation 4 was used to determine value.

Time =17000[(200-110)/5000KW] =306 min

Monday, May 07, 2007 3:23:02 PM

61

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

Knowledge of the operatiMaI implicationsd the foIlONing concepts as they apply to loss of Residual Heat Removal S~Ie-n : loss of

RHRS during all modes d opEl'"ation

(

Question No.

Tier 1 Group 1

Importance Rating:

Technical Reference:

43

R03.9

D-GO-15, App. U

Proposed references to be provided to applicants during exami nation :

O*GO*15, App U

Learning Objective:

OPL271AOP-R.03, Obj 1

OPL271GO* 15, Obj 16

Question Sou rce:

Bank

Question History:

AOP-RD3.B.1-6

Question Cognitive Level:

Higher

10 CFR Part 55 Content:

41.5

Comments: editorial change to stem based on NRC comment.

SEQUOYAH

NO

(

Time:

Source:

Cognitive Level:

Job Position:

Date:

Points:

BANK

HIGHER

4/2007

1.00

Version:

Answer:

o I 2 3 4 5 6 7 8 9

B D C CACA B B B

Scramble Range: A - D

Source If Bank:

SEQUOYAH BANK

Difficulty:

Plant:

Last 2 NRC?:

Monday, May 07, 2007 3:23:02 PM

62

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

32. 026 G2.2.25 001

Given the following plant conditions:

A loss of Component Cooling Wate r has occurred on Unit 1.

The crew is performing actions of AOP M.03, Loss of Component Cooling

Water.

1A Surge Tank level indicates 0%.

1B Surge Tank level indicates 57% and stable.

Assuming no other failures exist, which ONE (1) of the following describes the effect of

this condition, and the action required?

A~ Component Cooling Water System operability meets safety analysis assumptions.

Trip the reactor.

B. Component Cooling Water System operability meets safety analysis assumptions.

Reactor trip is NOT required.

C. Component Cooling Water operability does NOT meet safety analysis assumptions.

Trip the reactor.

D. Component Cooling Water operability does NOT meet safety analysis assumptions.

Reactor trip is NOT required.

A

Correct. Single f ailure loss of 1 train, still have B train available. Trip Rx due to loss

of surge tank level

B

Incorrect. Reactor trip is required.

C and 0 are incorrect because safety function is still met for these conditions. C

contains the correct actions, 0 contains plausible actions for loss of train with TS

implications

Monc:lay. May 07. 2007 3:23:02 PM

63

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

EQuipment Control KncM1edgea bases... tecMical specifiea~sfor limiting cooditions lor operations andsafetyNmits.

(

Question No.

Tier 1 Group 1

Importance Rating:

Technical Reference:

79

SRO 3.7

TS 3.7.3, M.03

Proposed references to be provided to applicants during examination:

None

Learning Objective:

Que stion Source:

Question History:

OPL271AOP-M.03 Objective 5, 6, 9

New

Question Cognitive Level:

Higher

10 CFR Part 55 Content:

43.5

Comments: Editorially modified actions based on NRC comments

(

Time:

Source:

Cognitive Level:

Job Position:

Dale:

Points:

NEW

HIGHER

412007

1.00

. Version:

Answer.

o I 2 3 4 5 6 7 8 9

AADDDCC DBD

Source If Bank:

Difficulty:

Plant:

Last 2 NRC?:

Scramble Range: A - D

SEQUOYAH

NO

Monday, May 07. 2007 3:23:02 PM

64

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

33. 026 G2.4.6001

The following plant conditions exist on Units 1 and 2:

Unit 1 is at 100% RTP.

1 B-B Containment Spray Pump is tagged for maintenance.

Unit 1 CCS is supplying the Spent Fuel Pool Cooling System.

C-S CCS pump suffers catastrophic bearing failure.

Which ONE of the following describes the actions required, if any, to provide B train

cooling for the given conditions in accordance with AOP-M.03, Loss of Component

Cooling Wate r?

A. No required actions.

B~ Only 2B-B CCS pump can be realigned for B train cooli ng.

C. Oniy 1B-B CCS pump can be realigned for B train cooling.

D. Either the 1B-B or 2B-B CCS pump can be realigned for B train cooling.

Justification:

A.

Incorrect. The 18-8 and 28-8 CCS pumps are not alignedfor B train cooling and these pumps will

not auto start on a loss of the C-S pump.

B.

Correct. Align the unit's B eeSpump that is not alignedto SFP cooling.

C.

Incorrect. Align the unit's B eeS pump that is notaligned to SFP coofing.

D.

Incorrect. Align the unit's BeeSpump that is not aligned to SFP coofing.

Knowledge of symptom based EO? mitigation strategies.

Noles:

KIA {CFR}:

026AK3.04

008A2.01

008G2.4.11

[3.513.71

[3.313.6)

[3.4/3.6)

{41.S. 41.10}

{41.S, 43.S}

{41.10. 43.S}

(

References:

LP/Objectives:

History:

Level:

Comments:

0-AR-M27-B-B. E-4; AOP-M.03; 0-50-70-1. Sects S.l & 8.4;

OPL271C42S. Obj. 3.

SON Bank

Higher

Monday, May 07, 2007 3:23:02 PM

65

SEQUOYAH

IL1'2006 PRAC 2

BANK

HIGHER

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

Time:

3

Points:

1.00

Version:

0 1 2 3 4 S 6 7 8 9

Answer:

B B B B B B B B B B

Items Not Scrambled

Source If Bank:

SQN BANK

Difficulty:

Plant:

Last 2 NRC?:

Source:

Cognitive Level:

Job Position:

Date:

(

Monday, May 07, 2007 3:23:02 PM

66

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

3 4 . 026 K4.01 001

Which ONE (1) of the following states the interlock(s) that must be satisfied before

FCV-72-22 (Containment Spray Suction from RWST) ("B" train) can be opened?

A. FCV-63-73 (RHR Suction from Containment Sump) closed.

B~ FCV-72-23 (Containment Spray Suction from Containment Sump) closed.

C. FCV-74-21 or FCV-74-3 (RHR Pump Iniet from RWST) control valves closed.

D. FCV-74-21 and FCV-74-3 (RHR Pump Inlet from RWST) control valves closed.

A. Incorrect. 63-73 is not required to be open, but 72-21 is required to be closed

B. Correct. Interlock designed to ensure sump and RWST are not cross-tied

C. Incorrect. 74-21 and 3 is not required to be closed.

D. Incorrect. 63-73 is not required to be open

Knowledge of CSS design leature(s) cndIor interiock(s) which proIIide foree following: Source d water lorCSS. includingrecirculatioo

phase after LOCA.

Question No.

Tier 2 Group 1

Importance Rating:

Technical Reference:

15

R04.2

1-47W611-72-1; 1-47W611-74-1; 1-47W611-6:l-2. 5; 0-45N779-B. l0.

25,26;

Proposed references to be provided to applicants durinq examination:

None

Learning Objective:

OPT200.CS, Obj 12.f

Question Source:

Bank

Queslion History;

SQN CSS B.12.F-2

Question Cognitive Level:

Lower

10 CFR Part 55 Content:

41.7

Comments:

(

Time;

Points:

1.00

Version:

0 I 2 3 4 5 6 7 8 9

Answer:

BBBB A A CDD B

Scramble Range: A - D

Monday, May 07, 2007 3:23:03 PM

67

SEQUOYAII

NO

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

BANK

Source If Bank:

SEQUOYAH BANK

LOWER

Difficulty:

Plant:

412007

Last 2 NRC?:

Source:

Cognitive Level:

Job Position:

Date:

(

Monday, May 07, 2007 3:23:03 PM

68

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

35. 027 G2.4.49 001

Given the following plant conditions:

Unit 1 is operating at 100% RTP.

Pressure Control Channel Selector Switch, 1-XS-68-340D is selected to

PT-68-340 & 334 position.

PT-68-340 fails LOW.

All equipment functions as designed.

Which ONE (1) of the following statements describes (1) the effect on the plant, and

(2) the required actio n?

A. ONE (1) PZR PORV will be inoperable; direct a manual reactor trip and enter E-O,

Reactor Trip or Safety Injection.

B~ PZR PORVs remain operable; direct manual control of PZR pressure and enter

AOP 1.04, Pressurizer Instrument Malfunctions.

C. ONE (1) PZR PORV will be inoperable; direct manual control of PZR pressure and

enter AOP 1.04, Pressurizer Instrument Malfunctions.

D. PZR PORVs remain operable; direct a manual reactor trip and enter E-O, Reactor

Trip or Safety Injection .

Monday, May 07,20073:23:03 PM

69

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

A. Incorrect. Wrong challenge. PORVs should not be inop as a result of the instrument

failure, as they are capable ofbeing cycled manually. Incorrect procedure, reactor trip

would not be required. Credible if the applicant thinks the PORV will not operate for this

condition.

B. Correct.TS 3.4.3 bases states that the PORV operability is determined by being

capable of manual control of PORVs to control reactor coolant system pressure. With

the controlling PZR pressure channel failing low, pressure will rise and be controlled by

PZR PORV 68-334. AOP-I-04, will be used to mitigate the effects ofthe failure.

C. Incorrect. PORV operability will not be affected, correct procedure.

D. Incorrect. Correct that PORVs do remain operable but procedure is incorrect.

Credible if applicant thinks he must rely on manual control of the PORVs to control

pressure and takes conservative actions per EPM-4 to trip reactor.

B.

When a parameter is approaching a protective action setpoint in an

uncontrolled manner, the operator is expected to:

1.

Evaluate the parameter magnitude and trend.

2.

Take actions as necessary to place the plant in a safe condition without

relying solely on automatic actions.

Emergency Procedures I PIa-1 Ability 10perioml without referencekI procedures !hoseactions thai require immediate operntion 01

syslem components Mel ccntrols.

Question No.

Tier 1 Group 1

Importance Rating:

Technical Reference:

80

SRO 4.0

AOP-1.04, Technical Specification Bases 3.4.3, page B 314 4-2

Proposed references to be provided to applicants durinq examination:

None

Learning Objective:

Question Source:

Question History:

OPL271AOP-1.04 Objective 3

New

(

Question Cognitive Level:

Higher

10 CFR Part 55 Content:

43.5

Comments :

Monday, May 07, 20073:23:03 PM

70

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

Time:

I

Points:

1.00

Version:

0 1 2 3 4 5 6 7 8 9

Answer:

B A C B OD DO C 0

Scramble Range: A - D

(

Source:

Cognitive Level :

Job Position:

Date:

NEW

HIGHER

412007

Source If Bank:

Difficulty

Plant:

Last 2 NRC?:

SEQUOYAH

NO

Monday, May 07,20073:23:04 PM

71

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

36. 027 G2.4.6 00 1

Given the following plant conditions:

A reactor trip and safety injection has occurred on Low PZR pressure.

The crew is performing E-O, Reactor Trip or Safety Injection.

All Train A ECCS equipment is operating as required.

Train 8 ECCS equipment has failed to start.

All RCPs are running.

While performing E*Ostep 10, Check pressurizer PORVs, safety valves and

spray valves, the following indications are observed by the OATC:

RCS pressure is 1200 psig.

PZR level is 100%.

PORV 68-340 red position indicator is lit.

Both PZR Spray Vaive indicators are not lit.

Which ONE (1) of the following describes the actions that are required in accordance

with E-O?

A. Leave RCPs running, attempt to close the failed open PORV and/or block valve.

8 . Leave RCPs running, allempt to close the failed open PZR Spray Valve.

e

C~ Trip RCPs; allemptto close the failed open PORV and/or block valve.

D. Trip RCPs; attempt to close the failed open PZR Spray Valve.

A. Incorrect. RCPs must be tripped, and E-Owill not be continued if this condition

exists.

B. Incorrect. Correct actions with exception of tripping RCPs.

C. Correct. With RCS pressure <1250 psig and 1 SI or CCP running, trip RCPs, E-O

step 10 RNO directs operators to attempt to close the PORVls), associated Block

Va/vels) and transffion to E-l if the failed valves cannot be isolated.

D. Incorrect. Correct actions for RCPs. Applicant may confuse spray valves, they

should not be open in the stated conditions.

(

Monday, May 07, 2007 3:23:04 PM

72

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

EmergencyProcedures ' Plan Knowledge S)'1TlPtcmbased EOP mitigalion strategies.

(

Question No.

45

Tier 1 Group 1

Importance Rating:

R0 3.1

Technical Reference:

E-O. step 10 RNO, E-OFoldout/Handout Page; EPM-3-E-0 pgs

18&22/64

Proposed references to be provided to applicants during exami nation:

None

Learning Objective:

OPL271E-Q Objective 5

Question Source:

New

Question History:

Question Cognitive Level:

Higher

10 CFR Part 55 Content:

41.10

Scramble Range: A * D

Removed all from stem.

SEQUOYAH

NO

NEW

HIGHER

412007

Cognitive Level;

Job Position:

Date:

Comments: editorial change to A and B based on NRC comment.

Time:

I

Points:

1.00

Version:

0 1 2 3 4 5 6 7 8 9

Answer:

CADA DC BB C D

Source If Bank:

Difficulty:

Plant:

last 2 NRC? :

Source :

(

Monday, May 07. 2007 3:23:04 PM

73

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

37. 028AlO.05 001

(

Given the following plant conditions:

Unit 1 is at 100% power.

Load is being reduced to 50% to remove a Main Feedwater Pump from service.

Pressurizer Level Control is selected to L1-68-339.

The Pressurizer Level Master Controller setpoint fails at its current value.

The load reduction is initiated.

Which ONE (1) of the following describes the effect on the unit and the action(s)

required?

A. Actual pressurizer level will be higher than program level for the actual power level

as load is decreased;

Change the pressurizer level channel input to the master controller and restore

Letdown.

B~ Actual pressurizer level will be higher than program level for the actual power level

as load is decreased;

Take manual control of Charging fiow.

C. Actual pressurizer level will be lower than program level for the actual power level

as load is decreased;

Take manual control of Charging fiow.

D. Actual pressurizer level will be lower than program level for the actual power level

as load is decreased;

Change the pressurizer level channel input to the master controller and restore

Letdown.

A. Incorrect. Changing inpul will not affect the controller because the setpoint has

failed, not the input

B. Correct.

C. tncorrect. Charging flow must be reduces because level will be artificially high

D. Incorrect. Changing input will not affect the controller because the setpoint has

failed. Letdown will not isolate because actual level will indicate the same on all

channels

(

Monday, May 07, 2007 3:23:04 PM

74

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

Knowtedge 01the reascns for the following responses as Ihey apply 10the PressLrizer level Conlrol ~nctions : Actions contained in

EOP for PZR levelmetfunction .

KA Match Jl,I$tification: KA is matched since Ihe actions for this failure are included in ARP's vs. EOP's. Actions would be the same if

faiue occured during performance of EOP's.

Question No.

Tier 1 Group 2

Importance Rating:

Technical Reference:

59

1-AR-M5-A E-4

Proposed references to be provided to applicants dunnq examination:

None

Learning Objective:

Question Sou rce:

Question History:

OPT200.PZRLCS Objective 5.d

New

Question Cognitive Level:

Higher

10 CFR Part 55 Content:

41.10

Comments:

(

Time:

Source:

Cognitive Level:

Job Position:

Date:

Points:

NEW

HIGHER

4/2007

1.00

Version:

Answer:

o I 23 4 5 6 7 8 9

BBD AAAABAA

Source If Bank

Difficulty:

Plant:

Last 2 NRC?:

Scramble Range : A * 0

SEQUOYAH

NO

Monday, May 07,20073:23:04 PM

75

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

38. 028 K6.01 001

Given the following plant conditions:

Unit 1 was at 100% power

A LOCA has occurred

Which ONE (1) of the following describes the effect on Containment if B train

Hydrogen Recombiner is lost after SI initiates?

Containment hydrogen concentration will

_

A ~ remain below 4%.

B. increase to greater than 4% but tess than 8%.

C. increase to greater than 8% UNLESS Containment Purge is placed in service.

D. increase to greater than 8% UNLESS Hydrogen Igniters are placed in service.

A. Correct. A single recombiner is designed to maintain H2 < 4% during a design basis

LOCA.

B. Incorrect. 4% is as high as concentration will get with only 1 recombiner in service

C. Incorrect. Purge not required to maintain design basis

D. Incorrect. Ignfters not required to maintain design basis

Monday, May 07. 2007 3:23:04 PM

76

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

Knowledge of the effect of a loss or m~bon m the ft*lwing win have on the HRPS: H)(Jrogen reccnbnes.

(

Question No.

Tier 2 Group 2

Importance Rating:

Technical Reference:

32

R0 2.6

etmt Combustible Gas Control Handout

Proposed references to be provided to applicants during examination:

None

Learning Objective:

OPT200.CGCS. Obj. 5.d

Question Source:

Bank

Question History:

SQNCTMT COM GAS B.15.A-1

Question Cognitive Level:

Lower

10 CFR Part 55 Content:

41.8

Comments: Removed 1A-A DG being OOS from stem based on NRC comments

(

Mes

Time:

Source:

Cognitive Level:

Job Position:

Date:

Points:

BANK

WWER

412007

1.00

Version:

0 I 2 3 4 5 6 7 8 9

Answer:

A D B D D DCA A D

Scramble Range: A - D

Source If Bank:

SEQUOYAll BANK

Difficulty:

Plant:

SEQUOYAll

Last 2 NRC? :

NO

Monday. May 07, 2007 3:23:04 PM

77

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

39. 029 A4.01 00 1

Given the following plant conditions:

Unit 1 is in Mode 6.

A Containment Purge is planned.

Which ONE (1) of the following conditions requires a limit on the maximum fiow for

Containment Purge, and what isthe maximum flow for the condition in accordance with

SO-30-3, Containment Purge System Operation?

A. Fuel Handling activities inside containment are in progress; 8,000 CFM

B~ Fuel Handling activities inside containment are in progress; 16,000 CFM.

C. SFP Fuel Transfer Tube is open; 8,000 CFM

D. SFP Fuel Transfer Tube is open; 16,000 CFM.

A. Incorrect. Flow limit is too low by half

B. Correct. If Fuel Handling activities are in progress, 16,000 CFM is the limit.

C. Incorrect.

SFP level may be affected on initial startup of Purge. Limit is too low

D. Incorrect. No limit on Purge flow based on Transfer Tube position. There are

concerns about level changes in the SFP when operating ventilation in containment,

but no minimum or maximum purge flow restrictions

Monday, May 07, 2007 3:23:04 PM

78

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

Abilitytomanually operate aldIormonitor in thecontrol room:Contairmentpurgeflow rate.

(

Question No.

Tier 2 Group 2

Importance Rating:

Technical Reference:

33

R0 2.5

0-SO-30-3

Proposed references to be provided to applicants during examination:

None

Leaming Objective:

Question Source:

Question History:

OPT200.CONTPURGE Objective 5

New

Question Cognitive Level:

Lower

10 CFR Part 55 Content:

41.10

(

Comments:

Time:

Source:

Cognitive Level :

JobPosition:

Date:

Points:

NEW

LOWER

412007

1.00

Version:

Answer:

o 1 23 4 5 6 7 8 9

BCC BD DACBC

Source If Bank:

Difficulty;

Plant:

Last 2 NRC?:

Scramble Range : A - D

SEQUOYAH

NO

Monday, May 07.20073:23:04 PM

79

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

40. 02902.2.25001

Which ONE (1) of the following describes the purpose of the AMSAC system during an

ATWS?

A. ONLY prevents exceeding the Reactor Core Safety Limit .

B. ONLY prevents exceeding the RCS Pressure Safety Limit .

C. ONLY maintains Fuel Integrity.

D~ Maintains Fuel Integrity AND prevents exceeding the RCS Pressure Safety Limits.

A. Incorrect. Maintains Fuel Integrity AND prevents exceeding the RCS Pressure

Safety Limits

B. Incorrect. Also maintains fuel integrity.

C. Inorrect. AMSAC operation also prevents exceeding pressure safety limits.

D. Correct. AMSAC operates to prevent exceeding the overpressurization and fuel

integrity safety limits.

FSAR 7.7.1.12

The design of the AMSAC system shall comply with 10 CFR 50.62 "Requirements for Reduction of

Risk from ATWS Events for light-Water-Cooled Nuclear Power Plants: The function of AMSAC is to

mitigate the effects of an ATWS by providing alternate means of trippingthe main turbine and

actuating auxiliary feedwater (AFW) flow independent from the reactor protection system (RPS).

AMSAC actuation will mitigate a common mode failure within RPS upon detection of low-low steam

generator levels existing in three-out-of-four steam generators coincident with plant power levels above

approximatel y 40 percent. This actuation will prevent reactor coolant system (Re S) overpressurization,

ma intain fuel integrity, and meet 10 CFR 100 radiation release requirements.

Monday, May 07, 2007 3:23:04 PM

80

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

Equipment Control Knc:JwIedge of basesirl teetv1iCaI specifica\\ic:l'ls for limiting conditions lor operations end safety limits.

(

KA justificatia'l: The KA is met by the fact eat the AMSAC circuitry is relied on in Ted1 Specs to pl'eva'lt exceeding the TIS safety limits.

Question No.

Tier 1 Group 1

Importance Rating:

Technical Reference:

46

R0 2.5

T5 2.1.2

Proposed references to be provided to applicants during examination:

None

Learning Objective:

Question Source:

Question History:

New

Question Cognitive Level:

Lower

10 CFR Part 55 Content:

43.2

(

Comments:

Mes

Time:

Source:

Cognitive level:

Job Position:

Date:

Points:

NEW

LOWER

4/2007

1.00

Version:

Answer:

o I 23 4 S 6 7 8 9

DCABBCDAB C

Source If Bank:

Difficulty:

Plant:

Last 2 NRC?:

Scramble Range: A - 0

SEQUOYAII

NO

Monday. May 07, 2007 3:23:04 PM

81

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

41. 034 A3.02001

Which ONE (1) of the following describes the function of the Unit 1 Manipulator Crane

SENSOTEC load system interlock, SEN-8H, Light Assembly Overload Limit Switch?

A~ Stops upward motion of the hoist and illuminates the overload light.

B. Prevents gripper from disengaging.

c. Stops hoist travel in the down direction, if over the reactor core and the gripper is

not disengaged.

D. Stops bridge travel at the edge of the reactor until the trolley is aligned with the

transfer system centerline.

A.

Correct:

B.

Incorrect: explanalion for SEN 6H gripper safely. If the readout shows 1200# or

more, fhe gripper air solenoid valve will nof operate fa allow gripper to

disengage.

C.

Incorrect. Lima switch LS-9 is a GEMCO geared limit switch (GL-2). This limit

switch stops fhe hoist in the down direction, if over the reactgor core and the

gripper is not disengaged.

D.

Incorrect - LS-6 operates to open a contact to stop bridge travel at the edge of

the reactor until the trolley is aligned with the transfer system centerline.

Monday, May 07, 2007 3:23:04 PM

82

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

Ability to monitor automatic operation of the Fuel Handling System, including: load limit s

(

Question No.

Tier 2 Grou p 2

Importance Rating:

Technical Reference:

36

RO 2.5

Proposed references to be provided to applicants during examination:

None

Learning Objective:

OPT200.FH objective 8 .12.0

Question Source:

Bank

Question History:

SON Bank

Question Cognitive Level:

Lower

10 CFR Part55 Content:

41.11

(

Comments:

Time:

Source:

Cognitive Level:

Job Position:

Dale:

Points:

BANK

LOWER

4/2007

1.00

Version:

0 I 2 3 4 5 6 7 8 9

Answer:

A B D e C A D D 0 C

Scramble Range: A - 0

Source If Bank:

SQN BANK

Difficulty:

Plant:

SEQUOYAH

Last 2 NRC?:

NO

Monday, May 07, 2007 3:23:05 PM

83

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

42. 035 K5.03 001

Which ONE (1) of the following describes the basis for programmed SG level

setpoi nts?

A. 33% level at low power is low enough to preve nt fiooding of the moisture

separators from a 10% step load increase.

B. 44% above 20% power is lowenough to minimize watercarryover from a

postulated steam line break.

C~ 44% above 20% power is high eno ugh to minimize the probability of a reactor trip

caused by shrink due to a load rejection at high power.

D. 33% at low power is high enough to keep the SG feed ring cove red in the event of

an Inadvertent MSIV closure.

C is correct.

A is incorrect because it is part of the basis for 44% at higher power

B is incorrect because it is part of the basis for 33%

D is incorrect but credible because an MS/V closure will cause shrink. Just not part of

the consideration for programmed level

Knowledgeof operational implicationsof the following concepts as the applytotheSlGS: Shrinkand swell concept.

(

Question No.

Tier 2 Group 2

Importance Rating:

Technical Reference:

34

R0 2.8

SQN.OO' D.SG P9 69-71

Proposed references to be provided to applicants during examination:

None

Learning Objective:

Question Source:

Question History:

OPT200 .SG. Obj 13.c. 15.8

New

Question Cognitive Level:

Lower

10 CFR Part 55 Content:

41.4

(

Comments:

Time:

Source :

Cognitive Level:

Job Position:

Date:

Points:

NEW

LOWER

4/2007

1.00

Version:

Answer:

o I 2 3 4 S 6 7 8 9

C A A BCAC B A A

Source If Bank:

Difficulty:

Plant:

Last 2 NRC?:

Scramble Range: A - D

SEQUOYAH

NO

Monday, May 07, 2007 3:23:05 PM

84

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

43 . 038 EKJ.06 001

When performing E-3, Steam Generator Tube Rupture, why is it important to isolate

the ruptured steam generator from the intact steam generators?

A. Ensures that RCS pressure, during the RCS depressurization following the

cooldown, does not reach the RCP trip criteria.

B ~ Ensures RCS SUbcooling is maintained when primary to secondary leakage is

terminated in subsequent steps.

C. Ensures that the subsequent cooldown will NOT result in a challenge to the PTS

Safety Function.

D. Ensures that the differential pressure between the intact and ruptured SGs remains

high enough to ensure early detection of subsequent failures.

A. Incorrect. RCP trip criteria is not applicable following operator initiated cooldown.

B. Correct. Cooling down the ruptured SG by depressurizing it will cause a higher OP,

and more flow, from the RCS to the SG.

C. Incorrect. Challenge s to Integrity are controlled by C/O rate.

O. Incorrect. Having a OP between the ruptured and intact SGs does not ensure early

detection of additional failures. Applicant may confuse this with the step following

cooldown that checks delta P between ruptured and intact SGs to diagnose a faulted

and ruptured SG.

Monday, May 07, 2007 3:23:05 PM

85

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5*7*07

Knowledge of the reasons for the following respcn ses as the apply to the SGTR; Actions contained in EOP for RCS water ~tory

balance, SIGtube rupture, ald pIalt shut<bNn pceeeoee

(

Question No.

Tier 1 Group 1

Importance Rating:

Technical Reference:

47

R04.2

E-3 Basis

Proposed references to be provided to applicants during examination:

None

Learning Objective:

OPL271 E-3 Objective 4

Question Source:

Bank

Question History:

Callaway 2005

Question Cognitive Level:

Lower

10 CFR Part 55 Content:

41.10

Comments: modified A distractor based on NRC comme nt.

(

Time:

Source:

Cognitive Level:

Job Position:

Date:

Points:

BANK

LOWER

412007

1.00

Version:

0 I 2 3 4 5 6 7 8 9

Answer:

B CD 0 A C e B O D

Scramble Range: A - D

Source If Bank:

CALLAWAY 200S NRC

Difficulty:

Plant:

SEQUOYAll

Last 2 NRC?:

NO

Monday. May 07, 2007 3:23:05 PM

86

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

44. 039 A2.04 OOJ

Given the following plant conditions:

Unit 2 is at 15% power during startup.

Steam Dumps are operating in the Steam Pressure mode.

Main Steam Header pressure instrument PT-1-33 fails high.

Which ONE (1) of the following describes the effect on the unit, and the actions

required to control RCS temperature?

A. Steam Dumps will fail closed.

Place and maintain the Steam Dump mode selector in OFF/RESET, then place the

controller in MANUAL and throttle open the steam dumps.

B. Steam Dumps will fail open.

Place and maintain the Steam Dump mode selector in OFF/RESET, then place the

controller in MANUAL and throttle closed the steam dumps.

C. Steam Dumps will fail closed.

Place and maintain the steam dump controller in MANUAL and throttle open the

steam dumps.

D~ Steam Dumps will fail open.

Place and maintain the steam dump controller in MANUAL and throttle closed the

steam dumps.

A. tncotrect. Dumps will open because they are already armed due to being in steam

presure mode. Credible because steam dump has arming and modulation signals in

Tave mode that could require this action

B. Incorrect. Dumps will open because they are armed from being in steam pressure

mode. Credible because steam dump has arming and modulation signals in Tave mode

that could require this action

C. Incorrect. Failure of dumps is opposite from what will actually occur

D. Correct. Operator will have to take manual control of the steam dumps in

accordance wah procedure GO-4 and AOP-S.05 to close the steam dump valves and

then control RCS temperature.

In Steam Pressure mode, the steam dumps have an arming signal already in.

Dumps must be placed in manual when the presure input fails, so they can be operated

Monday. May O?, 2007 3:23:05 PM

87

QUESTIONS REPORT

for SEOUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

Ability to (a) predict the impacts of !he fol\\olNing mal-func\\ic:l'ls Of operations on the MRSS; and (b) based al pre:lictiol'ls, use procedures

to correct, CXlfltrol, or mitigate !he cons equences of those malfunctions or operations: Malfunctioning steam dump .

(

Question No.

Tier 2 Group 1

Importance Rating:

Technical Referen ce:

17

R03.4

OPT200.S0S

Proposed references to be provided to applicants during exa mination:

None

Learning Objective:

Question Source:

Question History:

OPTZOO.SOCS Objective 5.d

New

Question Cognitive Level:

Higher

10 CFR Part 55 Conlent:

41.4

Comments: Mod ified A and B distractors to read similar to C and D based on NRC comment.

Time:

Source:

Cognitive Level:

Job Position:

Date:

Points:

NEW

HIGH ER

4/2007

1.00

Version:

Answer:

o I 23 4 567 8 9

DB ADDCACAC

Source If Bank:

Difficulty:

Plant:

Las12 NRC?:

Scramble Range: A - D

SEQUOYAI I

NO

Monday, May 07, 2007 3:23:05 PM

88

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

45. 0)9 AJ.02 001

(

Given the following plant conditions:

A plant cooldown is in progress.

RCS pressure is 1850 psig.

RCS temperature is SOS* F.

All required actions have been taken for the cooldown in accordance with plant

procedures.

An event occurs:

RCS pressure is 1700 psig and lowering at 10 psi per second.

SG pressures are 700 psig and lowering at2S psi per second.

Containment pressure is 1.2 psig and rising.

Assuming all equipment operates as designed, which ONE (1) of the following

describes the ESF actuation status?

A. Safety Injection has occurred; Main Steam Line Isolation has occurred.

B. Safety Injection has occurred; Main Steam Line Isolation has NOT occurred.

C~ Safety Injection has NOT occurred; Main Steam Line lsolatlon has occurred.

D. Safety Injection has NOT occurred; Main Steam Isolation has NOT occurred.

A. Incorrect. 51is blocked and has not occurred

B. Incorrect. Actuations are reversed from actual

C. Correct.

D. Incorrect. MSLI would actuate on rate. Si is blocked

Below P-l1, Low Steam pressure 51is blocked. High negative rate MSLI at 100 psi per

50 seconds is reinstated. CNMT pressure is not yet high enough to cause 51

(

Monday, May 07,20073:23:05 PM

89

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

Ability to monitor automaticoperationcI the MRSS, including: Isolation d the MRSS.

(

Question No.

Tier 2 Group 1

Importance Rating:

Technical Reference:

16

R0 3.1

ESF SO

Proposed references to be provided to applicants during examination:

Non e

Learning Objective:

Question Source:

Question History:

OPT200.RPS. B.4.g & h

New

Question Cognitive Level:

Higher

10 CFR Part 55 Content:

41.7

Comments: editorial revision of answers to make them read the same based on NRC

comment.

(

Time:

Source :

Cognitive Level:

Job Position:

Date:

Points:

NEW

HIGHER

4/2007

1.00

Version:

0 I 2 3 4 5 6 7 8 9

Answer.

C C A B A B BAD C

Scramble Range: A - 0

Source If Bank:

Difficulty:

Plant:

SEQUOYAll

Last 2 NRC?:

Monday. May 07, 2007 3:23:05 PM

gO

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

46 . 039 G2.1.32 001

Given the following plant conditions:

A reactor trip and safety injection has occurred .

Prior to the trip, #1 Main Steam Line radiation was in alarm.

The crew is preparing to transition from E*O, Reactor Trip or Safety Injection.

The currentconditions exist:

RCS pressure is 1550 psig and lowering slowly.

SG 1 pressure is 300 psig and lowering slowly.

SG 2, 3, and 4 pressures are 650 psig and stable .

Containment Pressure is 2.3 psig and rising slowly.

Which ONE (1) of the following describes the procedure transition required from E-O,

and appropriate actions to be taken after the transition to mitigate the event?

A ~ E-2, Faulted SG Isolation; SG 1 will NOT be used for a controlled RCS cooldown.

B. E-2, Faulted SG Isolation; SG 1 MAY be used dUring a controlled RCS cooldown

until no longer required for secondary heat sink.

C. E-3, Steam Generator Tube Rupture; TDAFW Pump steam suppiy will be isolated

from SG 1.

D. E-3, Steam Generator Tube Rupture; TDAFW Pump steam supply may remain

aligned to SG 1 until no longer required for secondary heat sink.

A. Correct:

Caution in E-2 alerts SRO to not unisolate a faulted SIG unless needed

for RCS cooldown. With 3 S/Gs intact, the faulted S/G should remain

isolated.

B. Incorrect: the faulted S/G should not be considered for cooldown under given

conditions. This is the reverse of the caution in E-2, which state the

faulted SG should not be unisolated unless needed for RCS cooldown.

Applicant may may misinterpret the caution.

C. Incorrect

E-2 transition is prior to the E-3 transition in E-O. E-3 will be entered after

addressing E-2, Faulted SI G Isolation, if conditions still exist. Could

potentialy enter ECA-3.1 if the faulted SG is also ruptured. Correct

action for isolating the TDAFWP.

D. Incorrect: E-2 transition is prior to the E-3 transition in E-O. E-3 will be entered after

addressing E-2, Faulted S/G Isolation. Incorrect E-3 caution.

Monday, May 07, 2007 3:23:05 PM

91

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

Conduct aI Operations: Ability to ID:Plain and applyall systemlimits and precautions.

(

Question No.

Tier 2 Group 1

Importance Rating:

Technical Reference:

89

SRO 3.8

E-O, E-2, E-3

Proposed references to be provided to applicants during examination:

None

Learning Objective:

Question Source:

New

Question History:

New

Question Cognitive Level:

Higher

10 CFR Part 55 Conlent:

43.5

Comments:

(

Mes

Time:

Source:

Cognitive Level:

Job Position:

Date:

Points:

NEW

HIGHER

4/2007

1.00

Version:

Answer:

o 1 2 3 4 5 6 7 8 9

ACDCDC CCAB

Source If Bank:

Difficulty.

Plant:

Last 2 NRC! :

Scramble Range: A - D

SEQUOYAH

NO

Monday, May 07, 2007 3:23:05 PM

92

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

47. 041A2.03 001

Given the following plant conditions:

Unit 1 is at 12% power during a plant startup.

A loss of Essential and Non-Essential Control Air has occurred.

The crew is attempting to restore Control Air in accordance with AOP-M.02,

Loss of Control Air.

RCS temperature is 559°F and rising.

PZR level is 91% and rising.

Whi ch ONE (1) of the following describes the effect on the unit, and identifies the

procedure that will be used to control RCS temperature while performing EOPs?

Trip the reactor due to...

A~ pressurizer level approaching an RPS setpoint.

Manual actions to control RCS temperature are contained in EA-1-2, Local Control

of SIG PORVs.

B. pressurizer level approaching an RPS setpoint.

Manual actions to control RCS temperature are contained in AOP-M.02.

C. OT Delta T approaching an RPS setpoint.

Manual actions to control RCS temperature are contained in EA-1-2, Local Control

of SIG PORVs.

D. OT Delta T approaching an RPS setpoint.

Manual actions to control RCS temperatu re are contained in AOP-M.02.

A. Correct. Trip based on loss of PZR level control, with trip occurring automatically at

92%. SG PORV 1 and 4 have reached rods and may be controlled manually, but

procedure calls for any or all PORVs.

.

B. Incorrect. PORVs 1 and 4 have reach rods, but all 4 PORVs may be used

C. Incorrect. Loss of RCS temperature control is not the reason for the trip.

Subsequenttemperature control is correct

D. Incorrect. PORVs 1 and 4 have reach rods, but any PORVs are used. RCS

temperature control is not the reason for the trip, although at 12% power wah steam

dumps open, it does cause problems

92% PZR level is trip criteria in M.02

Monday, May 07, 2007 3:23:05 PM

93

QUESTIONS REPORT

far SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

Ability 10 (a) predict the impacts of the foll(;:,Mng matfunclions or operations on the 5DS; aoo (b) based on those predictions or mitigate

the consequences cA those ~nctions or cceceecos: Loss of lAS.

(

Question No.

Tier 2 Group 2

Importance Rating:

Techn ical Reference:

92

SRO 3.1

M.02

Proposed references to be provided to applicants during examination:

None

Learning Objective:

Question Source:

Question History:

OPL271AOP-M.02, B.8.b

New

Question Cognitive Level:

Higher

10CFR Par1 55 Content:

41 .7, 41.10

Comments: Modified answers to eliminate 1 and 4 ONLY based on NRC comment. Also took

out concurrent use statement at end of each option, as it was redu ndant

(

Time:

Source:

Cognitive Level:

Job Position:

Dale:

Points :

NEW

HIGHER

412007

1.00

Version:

Answer:

o 1 2 3 4 5 6 7 8 9

ADBA CDCCAC

Source IfBank:

Difficulty:

Plant

Lasl 2 NRC?:

Scramble Range: A * D

SEQUOYAH

NO

Monday. May 07,2007 3:23:06 PM

94

(

c

(

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

48. 045 G2.4.30 001

Power is at 56% when a Main Generator Differential relay actuation occurs.

Which ONE (1) of the following describes the MOST LIMITING NRC reportability

requirem ent, if any, for this event?

Reference Provided

A. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months notification

Bl" 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months notification

C. 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months notification

D. No reportability required

A. Incorrect. Generator trip does not require a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months report

B. Correct.

4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months report is most limiting notification requirement per 50.72(b)(2)(iv)(B)

C. Incorrect.

8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months report also required per 50.72(b)(3)(iv)(A), but is not the most

. limiting

D. Incrrect.

Notification is required, but applicant may misinterpret reactor power and

not think a Main Genera tor differential will result in a reactor trip.

Monday, May 07, 2007 3:23:06 PM

95

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

Emergency Procedures I Plan KnO'Nledge of which events related to systemoperations/status should be reported tooutside agencies.

(

Question No.

Tier 2 Group 2

Importance Rating:

Technical Reference :

93

SRO 3.6

SPP-3.5

Proposed references to be provided to applicants during examination:

SPP-3.5. Appendix A

Learning Objective:

Question Source:

Question History:

OPL271C168 Objective 2

New

Question Cognitive Level:

Higher

10 CFR Part 55 Content:

43.5

Comments: Modified answer to ask the most limiting notification requirement based on NRC

comment.

(

Time:

Source:

Cognitive Level:

Job Position:

Date:

Points:

NEW

HIGH ER

412007

1.00

Version:

Answer:

o I 2 3 4 5 6 7 8 9

B D DB A A A C D C

Source IfBank:

Difficulty:

Plant:

Last 2 NRC?:

Scramble Range: A - D

SEQUOYAII

NO

Monday. May 07. 2007 3:23:06 PM

96

(

c

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

49. 055 EK3.01 001

Which ONE (1) of the following describes the minimum amount of time and reason why

the station batteries (and load shedding instructions) are designed to ensure DC

power is available after a loss of all AC power?

Time

Reason

A.

2 Hours

station blackout rule

B.

2 Hours

technical specification requirement

C~

4 Hours

station blackout rule

D.

4 Hours

technical specification requirement

a. Incorrect, batteries are rated (with load shedding) for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , without chargers to

provide necessary DC power to maintain BOTH reactors at hot shutdown after a

loss of ALL AC sources. Applicant may cofuse time with Tech Spec action time.

b. Incorrect, batteries are rated (with load shedding) for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , without chargers to

provide necessary DC power to maintain BOTH reactors at hot shutdown after a

loss of ALL AC sources

c. Correct.

d. Incorrect, batteries are rated (with load shedding) for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , without chargers to

provide necessary DC power to maintain BOTH reactors at hot shutdown after a loss of

ALL AC sources

Monday, May 07, 2007 3:23:06 PM

97

QUESTIONS REPORT

for SEQUOYAH 2007 - NRC EXAM REV FINAL AS GIVEN 5-7-07

Knowledge of the reasons for the following responses as theapply to the Station Blackout: Length of timefor whichbatterycapacity is

designed

(

Question No.

Tier 1 Group 1

Importance Rating:

Technical Reference:

48

R02.7

FSAR 8.0 Electric Power, 8.1.4 Design Basis

Proposed references to be provided to applicants during examination:

Learning Objective:

OPL271ECA-Q.O, Obj 3.a

OPL273C0527, Obj 5

Question Source:

Bank

Question History:

2004 NRC Exam

Question Cognitive Level:

Lower

10 CFR Part 55 Content:

41.8

Comments:

SEQUOYAll

YES

(

Time:

So urce:

Cognitive Level:

Job Position:

Date:

Points:

BANK

LOWER