ML041070222

Feb-March 2004 Exam 50-280,50-281/2004-301 Draft SRO Written Exam

ML041070222
Person / Time
Site:	Surry
Issue date:	02/19/2004
From:	NRC/RGN-II
To:
References
50-280/04-301, 50-281/04-301
Download: ML041070222 (138)

See also: IR 05000281/2004301

Text

SURRY EXAM

50-280, 50-281/2004-301

RUARY 24 - MARCH 2

& MARCH 4,2004 (WRITTEN)

U.S. Nuclear Regulatory Commission

Site-Specific

DRAFT SRO Written Examination

Applicant Information

Instructions

Use the answer sheets provided to document your answers. Staple this cover sheet on

top of the answer sheets. Po pass the examination you must achieve a final grade of at

least 80.88 percent overall, with a 70.80 percent or better on the SWB-only items if given

in conjunction with the RO exam; SWO-only exams given atone require an 80.00 percent

to pass. You have eight hours lo complete the combined examination, and three hours if

you are only taking the SWO portion.

I

Applicant Certification

All work done on this examination is my own. I have neither given nor received aid.

I

I RO / SRB-Bnty / Examination Values:

Applicant's Scores:

-I-/-

Points

Appiicant's Grades:

I

Surry Nuclear Plant 2804-381

DRAFT SRO lnital Exam

1. 003K4.03 001/2/1/wCP LUBRICATIQNMEM 2 5/2 8/N!SR0430lNMAESDR

.

Which ONE of the following correctly describes the Reactor Coolant Pump (RCP)

bearing oil lift system?

A! The oil lift pump discharge pressure must be greater than 350 psig prior to RCP

start. Once the RCP reaches operating speed the thrust runner circulates oil in the

upper and lower bearing assemblies.

E3. The oil lift pump discharge pressure must be greater than 3QQ psig prior to RCP

start. Once the RCP reaches operating speed the RCP Oil Lift System supplies the

bearing lubrication.

C. The oil lift pump discharge pressure must be greater than 350 psig prior to RCP

start. Once the WCP reaches Operating speed the RCP Oil Lift System supplies the

bearing lubrication.

D. The oil lift pump discharge pressure must be greater than 300 psig prior to RCP

start. Once the RCP reaches operating speed the thrust runner circulates oil in the

upper and lower bearing assemblies.

References:

MD-88.1 -bP-6, Reactor Coolant Pumps, Rev. 16

Elistractor Analysis:

A. Correct because there is a 350 psig discharge interlock with respective RCP. The

Oil Lift Pump ensures adequate lubrication upon RCP start, but once the pump

reaches operating speed, the thrust runner acts as an oil pump and circulates oil in

the upper and lower bearing assemblies.

B. Incorrect because pressure interlock is at 350 psig, not 308 p i g .

C. Incorrect because thrust runner circulates oil in upper and lower reservoir, not the

5. Incorrect because pressure interlock is at 350 psig, not 300 psig.

Oil Lift System.

003 Reactor Coolant Pumps

M4.63: Knowledge of RCPs design feature($) and / or interlock(s) which provide for the

following: Adequate lubrication of the RCP.

Surry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

The following Unit 1 conditions exist:

- Operating at 85% power

- Pressurizer pressure control is in its normal configuration

- A Pressurizer Safety Valve is leaking

- IC-BB, PWZR LO PRESS, annunciates

- 1-AP-31.00, increasing or Decreasing WCS Pressure, has been entered

Which ONE of the following correctly describes the affect on charging flow and an

appropriate mitigating action in accordance with 1 -AP-31 . O W

A. Charging flow initially increases. Place the PRZR PRESS MASTER CNTRL in

MANUAL and increase the demand to try to stop the pressure decrease.

MANUAL and increase the demand to try to stop the pressure decrease.

&. Charging flow initially decreases. Place the PWZR PRESS MASTER CNTRL in

CY Charging flow initially increases. Place the PRZR PRESS MASTER CNTRL in

MANUAL and decrease the demand to try to stop the pressure decrease.

D. Charging flow initially decreases. Place the PRZR PRESS MASTER CNTRL in

MANUAL and decrease the demand to try to stop the pressure decrease.

Surry

References:

ND-93.3-LP-5, Pressurizer Pressure Control, Rev. 9

ND-$8.3-LP-2, Charging and Letdown, Rev. 10

1 -AP-31 .00, Increasing or Decreasing WCS Pressure, Rev. 4

Distractor Analysis:

A. Incorrect because increasing the demand will lower pressure, not increase it.

B. Incorrect because charging flow will not initially decrease and increasing the

6. Correct because charging flow will initially increase due to the sudden pressure drop

demand will lower pressure, not increase it.

in the 86s. Also, decreasing the demand on the controller while in manual will act

to try to raise pressure.

D. Incorrect because charging flow will not initially decrease.

(404 Chemical and Volume Control

A2.17: Ability to (a) predict the impacts of the following malfunctions or operations on

the CVCS; and (b) based on those predictions use procedures to correct, control, or

mitigate the consequences of those malfunctions or operations: Low PZR pressure.

Surry Nuclear Plant 2004-301

DRAFT SRB lnital Exam

The following Unit 1 conditions exist:

- RCS level is 12.5 feet on I-RC-LI-10OA

- RCS level is 12 feet 5 inches on 1-RC-LR-105

- A loss of decay heat removal has occurred and 1 -AP-27.00, boss of Decay Heat

Removal Capability, has been entered.

- The RHW system has just been made available.

Which ONE of the following methods per 1 -AQ-27.00 should be used to sweep air from

the RHR lines during a loss of decay heat removal capability if inadequate time exists

to completely vent the RHW System prior to boiling in the core?

A:' Refill the RCS Bo 13.5 feet, verify 10 O F subcooling, and run an RHR pump at a flow

rate of > 2950 gprn.

3. Maintain RCS level at 12.5 feet, verify subcooling, and run an WHR pump at a flow

of > 2950 gprn.

C. Maintain RCS level at 12.5 feet, verify subcooling, and run an RHR pump at a flow

sf < 2950 gpm.

B. Close WH-MOV-1720A and B, RWR Outlets, then open "A" and "C" Safety Injection

Accumulator Isolation MBVs.

Surry

References:

ND-88.2-LP-1, Residual Heat Removal System Description, Rev. 8

NB-88.2-LP-2, Operation of Residual Heat Removal System, Rev. 15

ND-88.2-LP-3, Draindown and Midloop Operations, Rev. 12

1 -AP-27.00, Loss of Decay Heat Removal Capability, Rev. 10

Distractor Analysis:

A. Correct because based on procedural Note in 1-AP-29.00, Page 16 or 19, Rev. 10.

B. Incorrect because WCS needs to be filled to 13.5 feet.

C. Incorrect because WCS needs to be filled to 13.5 Beet. Also flow needs to be

greater than 2950 gpm.

D. incorrect because no procedural guidance exists to support the actions.

Suri-y ILT Exam Bank Question #275

005 Residual Heat Removal

K5.02: Knowledge of the operational implications of the following concepts as they

apply to the RHRS: Need for adequate subcooling.

Sur9 Nuclear Plant 2004-301

DRAFT SWO lnital Exam

- Steam Generator levels are 20% and rising

- Subcooling based on CETCs is 0 O F

- E-Q, Reactor Trip or Safety Injection, has been exited and Safety Function Status

- WCP Seal Injection flow is 3 gpm to all WCPs

- RCP Seal delta-Ps are all approximately 200 psid

- Source Range Startup Rate is zero

- Attempts to establish HHSI flow have failed

Trees are being monitored

Surry Nuclear Plant 26304-301

DRAFT SRQ lnital Exam

References:

MD-95.3-LP-38, FW-6.1 Response to Inadequate Core Cooling, Rev. 8

FR-C. 1, Response to Inadequate Core Coding, Rev. 18

Distractor Analysis:

A. Incorrect because 1 .O x IO5 PPH is well below the MSBV closure setpoirat and does

not even approach the maximum rate (an entire order of magnitude low}.

B. Incorrect because 1 .O x 10' PPH is well below the MSlV closure setpoint and does

not even approach the maximum rate (an entire order of magnitude low).

C. Incorrect because RCPs should be started even when normal conditions not met.

D. Correct because procedural guidance exists to supporl the actions. MSlV closure

will occur if flow is greater than 1 .O x 1 Q6 PPH. The purpose for the actions is to

establish low head flow from accumulators and LHSI. RCP support criteria is

desirable, but not a prerequisite for starting RCPs.

006 Emergency Core Cooling

K.6.03: Knowledge of the effect of a loss or malfunction on the following will have on

ECCS: Safety Injection Pumps.

Surry Nuclear Piant 2084-381

DRAFT SWO lnital Exam

5. 007EK2 02 001/l/ilBEAKER REACTOR TRIP/C/A 2 6/2.gW/SR04301I~ARISDR

..

~

The following conditions exist:

- Unit 1 is at 90% power

- Reactor protection testing is in progress

- Reactor Trip Breaker "A" is closed

- Reactor Trip Breaker "B" is open

- Reactor Trip Bypass Breaker "B" is racked in and closed

Which ONE of the following describes the plant response if reactor trip bypass breaker

"A" is racked in and closed?

A. Both reactor trip bypass breakers "A" and "B"

and reactor trip breaker "A" will trip

open and the reactor will trip.

B:' Only reactor trip bypass breakers "A" and "B" will trip open and the r@actos will trip.

C. Reactor trip breaker "A" will trip open and the plant will remain at 90% power.

D. Reactor trip bypass breaker "A" will trip open and the plant will remain at 90%

power.

Surry

References:

ND-93.3-LP-17, AMSAC, Rev. IO

ND-93.3-LP-18, Reactor Protection

~ General, Rev. 5

Distractor Anaysis:

A. Incorrect because reactor trip breaker "A" will not open.

B. Correct because this is the correct response per ND-93.3-LP-10.

C. Incorrect because reactor trip breaker "A" will not open and plant will trip.

D. lncorrect because the plant will trip.

Ssrrsy ILT Bank Question #I 667

009 Reactor Trip Stabilization

EK2.02: Knowledge of the interrelationships between a reactor trip and the following:

Breakers, relays, and disconnects.

Surry Nuclear Plant 2004-301

DRAFT SRO M a l Exam

Given the following Unit 1 conditions:

- A heatup is in progress to return to power from a cold shutdown condition

- RCS is filled and vented

- Pressurizer is solid

- A nitrogen blanket has been established on the PWT

- PRP Level = 95%

- Pressurizer Heaters are energized

Which ONE of the following must be accomplished prior to drawing a bubble in the

Pressurizer?

A? Drain the PRT to 68 - 80%.

B. Verify VCT oxygen concentration less than 3%.

C. Drain the Pressurizer to 22.2%.

D. Pressurize the WCS to 200 - 270 psig on $1-1 -403, Nar Range.

Surry

References:

1-GOP-1 .I

~ Unit Startup, RCS Heatup from Ambient to 195 Degrees F,, Rev. 25

1 -0P-RC-011, Pressurizer Relief Tank Operations, Rev. 13

Distractsr Analysis:

A. Correct bemuse GOP-1 .I Step 5.5.4 directs establishment ob normal PWT level

prior to drawing a bubble. OP-WC-011 Step 5.1.1 states the normal PRT level to be

60 - 80%.

B. Incorrect because GQP-1.1 Step 5.5.6 requirement is to verify VCT oxygen < 2%.

C. Incorrect because this is an action following establishment of drawing a bubble

6). Incorrect because RCS should be between 300 and 390 psig on $1-1-403.

(GOP-1.1, Step 5.5.13).

009 Pressurizer Relief / QuenchTank

6.02: Knowledge of the operational implications of the following concepts as they

amlv to PRTS: Method of forrnina a steam bubble in the PZR.

Surty Nuclear Plant 2804-301

DRAFT SRO lnital Exam

9. D08AA2.06 001/1/1/PRESS~TRE

TR.4NSMITTEWUA 3.3/3.4/NiSR04301/1R/MhR/SDR

~~

___

7

- -

Given the following Unit 1 conditions:

- Reactor power = 180%

= All other parameters are at normal steady state vaiues

- Subsequently PT-444 fails high

Assuming no operator action is taken, which ONE of the following is correct?

A. POWV-1455C opens, pressure decreases to 2000 psig, PORV-1455C closes, and

B. PORV-1456 opens, pressure decreases to 2000 psig, PORV-1456 closes, and

pressure stabilizes around 2000 psig.

pressure stabilizes around 2080 psig.

CY POW-145% opens, at 2000 psig PORV-I 455C closes; however, pressure will

continue to decrease causing a reactor trip and safety injection.

D. PQRV-I456 opens, at 2000 psig PQRV-1456 closes; however, pressure will

continue to decrease causing a reactor trip and safety injection.

References:

ND-93.3-LP-5, Pressurizer Pressure ControlI Rev. 9

Distractor Analysis:

A. incorrect because both spray valves also open, which causes pressure to continue

to decrease.

B. Incorrect because both spray valves open, which causes pressure to continue to

decrease. Also incorrect because PORV-1456 does not open.

6. Correct because both spray valves open causing a reactor trip on QT-delta-T or

Low Pressurizer Pressure, followed by SI.

D. Incorrect because PQRV-I456 does not open.

008 Pressurizer Pressure Control

AA2.03: Ability to determine and interpret the following as they apply to the pressurizer

vapor space accident: PORV logic control under Iow-pressure conditions.

Surry Nuclear Plant 2004-301

DRAFT Sf30 lnital Exam

Which ONE of the following correctly describes loads cooled by the Component

Cooling Water (CCW) System or subsystem of CCW?

A. RCP bearing lube oil coolers, neutron shield tank coolers, RCP seal water return

cooler, outside recirc spray pump seals.

8. HHSI pump seals, LHSl pump seals, RHW pump seals, RCP motor air coolers.

CY RHR pump seals, WCP bearing lube oil coolers, neutron shield tank coolers, HHSI

pump seals.

D. LHSl pump seals, RHW pump seals, RCP motor air cosiers, neutron shield tank

cmlers.

Surry

Reference:

ND-88.5-LP-1, Component Cooling, Rev. '89

ND-88.3-LP-5, Charging System, Rev. 16

Distractor Analysis:

A. lncorrect because outside recirc spray pump seals are not cookd by CC.

B. Incorrect because LHSl pump seals are not cooled by CC.

C. Correct because all are cooled by CC or a subsystem.

D. Incorrect because LHSl pump seals are not cooled by CC.

Requal Bank Question #527

088 Component Cooling

K1 .Q2: Knowledge of the physical connections and / or cause-effect relationships

between the CCWS and the following systems: Loads cooled by CCWS.

Surry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

9. W8K4 01 001/2/1/COMPONEiNT COOLINGEM 3.1133/B/SR04301/SDR-

-- 1

~

__

1

__ T e L c o n d i t e r a Safety Injection occurs, the "A"Ccmponent Cooling Pump trips.

Which ONE of the following describes the operation of the CC pumps?

A! The "B" CC pump will not auto start without a required operator action.

B. The "B"

CC pump will auto start 68 seconds after the "A" CC pump trips.

C. The "B" CC pump will auto start as soon as the "A" CC pump trips.

B. The "B" CC pump will auto start 50 seconds after the "A" CC pump trips.

~

i

~-

l - - -

_

_

_

~

~

_

_

_

_

~

-

Swry

References:

ND-88.5-CP-1

I Component Cooling Water System, Rev. 19.

Distractor Analysis:

A. Correct because Auto Start Inhibit due to SI will prevent auto starl of the CC pump,

but the pump may be manually started at any time.

B. Incorrect because the Auto Start Inhibit will block the auto start.

C. Incorrect because the Auto Start Inhibit will block the auto start.

D. Incorrect because the Auto Start Inhibit will block the auto start.

ILT Bank Question ## 537

008 Component Cooling Water System

K4.81: Knowledge of CCWS design feature(s) and/or interlock(s) which provide for the

following: Automatic start of standby pump.

Surry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

10. 010A1.01

~

00lI2IlIBORON

_

_

~

I53 SpK YiC&2.8/29/N/SKO

r - The followina plant conditions exist:

- Dilution tociiticality has just been completed

- Operators note that inadequate proportional heaters appear to be energized

- Pressurizer Pressure is 2230 gsig.

Which ONE of the following could result from inadequate Pressurizer Heater output

during a dilution to criticality?

(Assume ail other controls and cornpoments working properly in their normal

configuration .)

~

A:' Boron concentration will be higher in the Pressurizer than in the WCS.

B. Boron concentration will be lower in the Pressurizer than in the RCS.

C. Pressurizer and RCS boron concentration will be approximately equal.

D. The Pressurizer Spray Nozzle will be susceptible Io thermal shock.

L

References:

1 -GOP-I . I , Unit Startup, RCS Heatup From Ambient to 195 Degrees F, Rev. 25

Distaactor Analysis:

A. Correct because WCS boron will be IOWI&F

due to the dilution. The Pzr will still be at

a higher boron concentration untif spray flow has created enough out-surge to

adequately equalize the boron with the RCS. (Lack of heaters creates lack of

sprays.)

8. Incorrect because boron concentration will be higher in the Pmr.

C. Incorrect because the lack of heater output will not allow for adequate mixing.

D. Incorrect because the bypass spray valves are normally open, which is sufficient to

prevent thermal sh5ck. (Have utility verify that this is in fact the normal

configuration.)

01 0 Pressurizer Pressure Control

AI .01: Ability to predict and / or monitor changes in parameters (to prevent exceeding

design limits) associated with operating the Pzr PCS controls including: PZR and RCS

boron concentration.

Surry Nuclear Plant 2004-301

DRAFT §BO initas Exam

Given the following conditions:

- LQCA has occurred

- RCS subcooling is 63 *F

- RWST Level = 15% and slowly decreasing

- Containment Pressure = 9 psig and decreasing

- Safety Injection Actuation has been reset

Which ONE of the following is the correct action to be taken?

A. Close Charging Pump Miniflow Recirc Valves. With RWST level at 15%, push both

RMT pushbuttons fear each train if automatic transfer does not occur.

BY Close Charging Pump Miniflow Reeire Valves. When RWST level reaches 1374,

push both RMT pushbuttons for each train if automatic transfer does not occur.

C. With RWST level at Is%, push both WMT pushbuttons for each train if automatic

transfer does not occur. Secure Containment Spray Pumps immediately following

verification of Phase 1 and 2 RMT.

B. With RWST level at 13%, push both RMT pushbuttons for each train if automatic

transfer does not occur. Secure Containment Spray Pumps immediately following

verification of Phase 1 and 2 WMT.

Surry Nuclear Plant 2004-301

DRAFT Sa0 lnital Exam

-

12. 01 1K6.06 001/2/2/CHAKGIN;G PRESSIrRIZER/MEM

~~~~

2 5/2.8/N/SR0430I/lUtvI.4B/SJK ~

~

- ._

__ -

Due to a controller faituse, the Unit 1 Operator places the Charging Flow Controller to

MANUAL to control charging flow. A high Pressurizer Level causes the Operator to try

to reduce charging flow to 20 gpm.

Which ONE of the following correctly describes the behavior sf FCV-1122 when the

Operator attempts to reduce charging flow to 20 gprn?

A! The Flow Limit Summator no longer limits flow and FCV-1122 can be manually

closed to allow 20 gpm flow.

B. The Flow Limit Summator no longer limits flow, however, FCV-I122 can only be

manually closed to allow 25 gprn flow.

6. The Flow Limit Summator will prevent FCV-1122 from being closed past 25 gpm

flow.

D. The Flow Limit Summator will prevent FCV-I 122 from being closed past 30 gpm

flow.

Surry Nuclear Plant 2004-306

DRAFT SRO lnital Exam

Surry

References:

ND-93.3-LP-7, Pressurizer Level Control System, Rev. 6

Distractor Analysis:

A. Correct because when the Charging Flow Controller is in MANUAL, the Flow Limit

Summator no longer limits the maximum and minimum values of charging.

Therefore FCV-I 122 can be closed manually to any value.

B. Incorrect because when the Charging Flow Controller is in MANUAL, the Flow Limit

Summator no longer limits the maximum and minimum values of charging.

Distractor is incorrect because FCV-1122 may be manually closed to any value,

even below 25 gpm flow. Distractor is plausibe because candidate may not know

that FCV-1122 may be throttled to any value with controller in MANUAL.

6. Incorrect because when the Charging Flow Controller is in MANUAL, the Flow Limit

Summator no longer limits the maximum and minimum values of charging. The

distractor states that the Flow Limit Summator will limit flow, which is contrary to the

fact that it will not limit flow. Distractor is plausible because candidate may not

know that the Flow Limit Summator does not function with controller in MANUAL.

D. Incorrect because when the Charging Flow Controller is in MANUAL, the Flow Limit

Summator no longer limits the maximum and minimum values of charging. The

distractor states that the Flow Limit Summator will limit flow, which is contrary to the

fact that it will not limit flow. Distractor is plausible because candidate may not

know that the Flow Limit Summator does not function with controller in MANUAL.

01 1 Pressurizer Level Control

K6.06: Knowledge of the effect of a loss or malfunction on the following will have on

the PZW LCS: Correlation of demand signal indication on charging pump flow valve

controller to the valve position.

Surry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

~ 13. 012A.104

__ 001/2/l/RPS

~

T E D TEI\\.IP/('/A 3 . ~ B l S R C H H " M A B ~ S I X <

~

- __

___

"A" Loop Narrow Range Pcold fails low while the reactor is at 100%.

Which ONE of the following will occur?

A. Rod Insertion Limit Low and Extra Low alarms will be received.

B. Ch 1 OTBT setpoint will decrease.

CY

"A" Loop Delta T Protection Bistable will trip.

D. The Tavg / Pref Deviation alarm will be received

References:

ND-93.3-LP-2, DeHa T / Tavg Instrumentation System, Rev. 9

NB-93.3-LP-3, Rod Control System, Rev. 14

Distractor Analysis:

A. Incorrect because failed Tcold is filtered out by Median Signal Selector.

B. Incorrect because OTDT setpoint will actually increase.

C. Correct because Teald is fed directly to the RPS even when failed low.

D. Incorrect because failed Teold is filtered out by Median Signal Selector.

012 Reactor Protection System

84.84: Ability to manually operate and / or monitor in the control room: Bistables, trips,

resets, and test switches.

Surry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

14. 012K1.05 00ll2l1/hl/SAC'/MEM

3.8/3 9lRiSR04301/RIMAB/SDR

Which ONE of the following lists the method by which AMSAC causes a reactor trip?

r-

A. Tripping the reactor trip and bypass breaker shunt coils.

B. Tripping the reactor trip and bypass breakers IBV coils.

I

.

C. Tripping the rod drive MG set output breakers.

DY Tripping the rod drive MBG set supply breakers.

__

~.

Surry

References:

ND-93.3-LP-17, Anticipatory Mitigating System Actuating Circuitry, Rev. 10.

Distractor Analysis:

A. Incorrect because this does not occur.

B. Incorrect because this does not occur.

C. Incorrect because this does not occur.

D. Correct becasue this is as stated in ND-93.3-LP-17, Rev. IO.

01 2 Reactor Protection System

K1 .05: Knowledge of the physical connections and / or cause-effect relationships

between the RPS and the following systems: ESFAS

Surry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

15. 013A3

~- 02 001/2/1/SAWIY

-~

HNECHON/MEM 4 114 2/W/SKO4301/WiVlAB/SDR ~

~

-

Which ONE of the following correctly states automatic actions that would occur given a

Unit 1 how Pressurizer Pressure Safety Injection Signal being present for 5 minutes?

A. Hydrogen Analyzer Heat Tracing energizes AND Containment Vacuum Pumps trip.

B. Pressurizer Liquid Sample (SS-TV-180A) receives a close signal AND Motor Driven

ARM Pumps start after a 45 second time delay.

62:' Accumulator Nitrogen Relief Lines (SI-TV-101 A,B) receive a eisse signal AND

Primary Drain Transfer Tank Vents (VG-TV-109NB) receive a close signal.

B. Main Steam Trip Valves (MS-TV-IOlNJBIC) receive a close signal AND Seal Water

I

Return Valve (MQV-3819 receive a close signal.

References:

ND-9b 4P-2, Safety Injection System Description, Rev. 16

ND-91 -kP-2, Safety Injection System Operations, Rev. 15

P&ID 1 1448-FM-0684, FlowNalve Operating Numbers Diagram Feedwater System

Sur9 Power Station Unit 1, Rev. 57

Distractor Analysis:

A. Incorrect because SI signal must be present B Q ~ 8 minutes for heat trace to

B. Incorrect because MDAFW Pump starts after 50 sec delay.

6.

Correct because both get a close signal on any SI Signal.

D. Incorrect because MSTVs only get a close signal on a High Steam Flow SI Signal.

01 3 Engineered Safety Features Actuation

A3.82: Ability to monitor automatic operation of the ESFAS including: Operation of

actuated equipment.

energize.

Surry Nuclear Plant 2004-301

DRAFT SWO lnital Exam

~

~

~

~.

16. 013K3.01 001/2/I/HOT ~.

IEG RECIRChQiM

-__. 4.414.7/WYSR043UI~~B/SDR

Which ONE of the following could mcur if ES-1.4,

Transfer to Hot Leg Recirculation, is

performed 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> after the start of a Large Cold Leg Break LOCA?

A. Debris from the ln-Core sump could block coolant Blow by blocking the lower core

plate.

B. Reflux cooling could be lost due to boron precipitation in the hot leg nozzles.

C. Fouling of core heat transfer surfaces due to the dilution of boric acid.

D I Reduction in size of the incore coolant flow channels due to boron precipitation.

Surry

Wefe re nces :

ND-95.3-LP-11, ES-1.4, Transfer To Hot beg Recirculation, Rev. 8

ES-1.4, Transfer To Hot beg Recirculationl Kev. 4

Distractor Analysis:

A. Incorrect because debris in the sump will not block water discharged from the SI

B. Incorrect because boron precipitation is a concern in the core, not the hot legs.

C. Incorrect because fouling of core heat transfer surfaces is a result of boron

precipitation, not dilution.

D. Correct because boron precipitation is a concern when bsil-off continues and when

core temperature decreases. The standard time for transfer to hot leg recirc is 8

hours, not 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br />, as stated in the stem.

pumps.

01 3 Engineered Safety Features Actuation

M3.01: Knowledge of the effect that a loss or malfunction of the ESFAS will have on

the following: Fuel

Surry Requal Bank Question #299

Surry Nuclear Plant 2004-301

DRAFT SRO Bnitai Exam

d 7.

014A2 05 001/2/2RPIS ROD POSITIONiMEM

.~

3.9/4. I~/SR04301/R/MAW/SDR

.~

~-

I

The following Unit 1 conditions exist:

- A Small Break LOCA has occured

- Automatic Safety Injection has occurred

- I-E-Q, Reactor Trip or Safety Injection, has been implemented

- The CWO observes the Rod Position Indication as displaying Control Rods on the

bottom of the reactor core, with the exception of three Control Rods.

Which ONE of the following actions is procedurally required as a result of this finding

by the CRO?

A." Continue with 1 -Ea>, Reactor Trip or Safety Injection.

3. Emergency borate while proceeding through 1 -E-$), Reactor Trip or Safety Injection

C. Manually insert control rods while proceeding through I-E-0, Reactor Trip or Safety

Injection.

D. Go directly to I-FW-S.1, Response to Nuclear Power Generation / ATWS, Step 1.

References:

1 -FR-S.l~ Response to Nuclear Power Generation / ATWS, Rev. 18

1-E-0, Reactor Trip or Safety Injection, Rev. 46

Distractor Analysis:

A. Correct because E-0 should be entered upon Reactor Trip per the rules of EOP

usage.

B. lncorrect because if emergency boration is needed, it will be directed by FR-S.I.

C. Incorrect because if manual sod insertion is needed, it wilt be directed by FR-S.1.

B. Incorrect because FR-S.l should only be entered as directed by E-Q (or if E-8 has

been completed then an Orange or Red path).

Qf4 Rod Position Indication

A2.05: Ability to (a) predict the impacts of the following malfunctions OF operations on

the RPIS; and (b) based on those predictions, use procedures to correct, control, or

mitigate the consequences of those malfunctions or operations: Reactor Trip.

Surry ILT Bank Question #lo37

Surly Nuclear Plant 2004-301

DRAFT SRO lnital Exam

The following Unit 1 conditions exist:

- Reactor Power is 5%

- Turbine First Stage Impulse Pressure PT-446 is selected

~

Power Range Nuclear Instrument N-41 fails high

- PT-446 fails high

Which ONE of the following correctly describes the impacts of the failures?

A. Control Rods do not move. The Reactor Protection System At-Power Trips are

enabled due to the N-41 failure.

B. Control Rods step out at 72 steps per minute. The Reactor Protection System

At-Power Trips are enabled due to the N-41 failure.

6. Control Rods do not move. The Reactor Protection System At-Bower Trips are

enabled due to the PT-446 failure.

B:' Control Rods step out at 72 steps per minute. The Reactor Protection System

At-Power Trips are enabled due to the PP-446 failure.

Surly

References:

ND-93.3-LP-16, Permissive/Bgipass/rip Status Lights, Rev. 8

Surly Simulator Malfunction Cause and Effects, Rev. 6, Malfunction MMS-14

Distractor Analysis:

A. lncorrect because Tref will go to 574 O F , which will cause rods to step out at max

rate of 72 steps/rnin. Also incorrect because 2/4 PR Nls must be a b o ~ e 10% to

enable At Power Trips.

B. lncorrect because 2/4 PR Nls must be above 10% to enable At Power Trips.

6. Incorrect because Tref will go to 574 O F , which will cause sods to step out at max

rate of 72 stepshin.

B. Correct because Tref will go to 574 O F , which will cause rods to step out at max rate

of 72 steps/rnin and only 1/2 Turbine First Stage PTs need to be above 18% to

enable At Power Trips.

015 Nuclear Instrumentation

K4.07: Knowledge of NES design feature(s) and / or interlock(s) provide for the

following: Permissives.

Surry Nuclear Plant 2084-301

DRAFT SRO lnital Exam

The following condition exists:

- Unit 1 at 100% reactor power

- All systems and equipment functions as designed

- All protection channel 111's are selected

- First stage impulse pressure channel IV fails low

Which ONE of the following would occur initially without operator action?

A. AMSAC would be operationally disabled after 60 seconds.

B. Steam Bumps would all open.

C. FRVs would control SG level at no load Bevel.

Df MOV-CP-100, Condensate Polishing Building Bypass Valve, would open.

~-

~

-

~

~

__

.~

~

~

~

SUt'W

Wef?EXlCe§:

ND-93.3-LP-17, Anticipatory Mitigating System Actuating Circuitry (AMSAC), Rev. 10

ND-93.3-LP-9, Steam Dump Control System, Rev. 10

ND-93.3-LP-8, SG Water bevel ConttQl System, Rev. 6

Distractor Analysis:

A. incorrect because this would occur after 360 seconds.

B. Incorrect because Channel III is selected.

C. Incorrect because Channel III is selected.

D. Correct because, as stated in ND-93.3-LP-9, CP-100 will open in anticipation of the

upcoming increase in feedwater flow that will occur during load rejection.

01 6 Non-Nuclear instrumentation

A 4 0 1  : Ability to manually operate and / or monitor in the control room: NNl channel

select controls.

Surry Requal Bank Question #279

Surry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

_ _

20. 022AK1

~

.OI OQl/l/I/RCP

-

~ SEALWU-A

2.8/3.2/N/SK0430l/RMMABIST)K _.

~

-

I

The following Unit 1 conditions exist:

- Reactor trip has occurred due to a loss of all AC power

- Power has been restored

- The following Reactor Coolant Pump parameters are present for all RCPs:

I

- No. 1 Seal Water Outlet Temperatures are 225 "F

~

Lower Seal Water Bearing Temperatures are 220 O F

1-AP-9.02, Loss of WCP Seal Cooling.

- The Shift Supervisor directs the operators to restore cooling to the RCP seals per

Which ONE of the following correctly states the requirements for restoring cooling to

the RCP seals and why?

A. Do not establish seal injection flow or component cooling flow to the thermal barrier

heat exchanger because the No. I Seal Water Outlet Temperatures are too high.

Seal cooling should be restored by cooling the RCS using natural circulation.

1

3. Do not establish seal injection flow or component cooling flow to the thermal barrier

heat exchanger because the Lower Seal Water Bearing Temperatures are too high.

Seal cooling should be restored by cooling the RCS using natural circulation.

C. Slowly establish seal injection flow to minimize RCP thermal stresses, followed by

slowly introducing component cooling flow to the thermal barrier heat exchanger to

limit introduction of steam into the CC system.

I

I

D:' Slowly establish component cooling flow to the thermal barrier heat exchanger to

limit introduction of steam into the CC system, followed by slowly introducing seal

injection flow to minimize the RCP thermal stresses.

-1

~

~

-

-

~

_

_

-

-

~

~

Surly Nuclear Plant 2884-301

DRAFT SRO initas Exam

References:

1 -AP-9.02, boss of WCP Seal Cooling, Rev. 8.

ND-88.1-LP-6, Reactor Coolant Pumps, Rev. 16.

Distractor Analysis:

A. Incorrect because AP-9.02 (Caution page 7) states if No. 1 Seal Water Outlet Temp

is > 235 O F then Seal Inj and CCW to Thermal Barrier H.X. should not be restored.

instead N.C. should be used to cool the seals.

Temperature is > 225 O F then Seal Inj and CCW to Thermal Barrier H.X. should not

be restored. instead N.C. should be used to cool the seals.

C. Incorrect because CC flow should be established prior to seal injection flow.

D. Correct as stated in 1 -AP-9.02 NOTE prior to step 7 and CAUTIONS prior to steps 9

3. Incorrect because A$-9.02 (Caution page 7) states if Lower Seal Water Bearing

and 15.

822 Loss of Wx Coolant Makeup

AK1 .Qf : Knowledge of the operational implications of the following concepts as they

apply to Loss of Reactor Coolant Pump Makeup: Consequences of thermal shock to

RCP seals.

Surry Nuclear Plant 2004-301

DRAFT SRQ M a l Exam

Unit 1 has tripped and Safety Injection has actuated due to a Large Break Loss of

Coolant Accident (LOCA).

Many complications have occurred.

The crew has exited E-0, Reactor Trip OF Safety Injection. The Shift Technical Advisor

has started to monitor Critical Safety Function Status Trees and reports:

- Subcriticality - Orange Path

- Heat Sink - Yellow Path

- Core Cooling - Orange Path

~ Containment - Red Path

Which ONE of the following states the correct procedure transition?

A. FR-SI, Response to Nuclear Power GeneratiodATWS, based on Subcriticality

Orange Path.

B. FR-H.1, Response to Secondary Heat Sink, based on Heat Sink Yellow Path.

C. FR-6.1, Response to Inadequate Core Cooling, based on Core Cooling Orange

Path.

B:' FR-Z.1, Response to High Containment Pressure, based on Containment Red

Path.

Surry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

Sur9

References:

NB-95.3-LP-26, Critical Safety Function Status Trees, Rev. 5

Distractor Analysis:

A. Incorrect based on the rules of use for safety function status trees (ND-95.3-LP-26

Page 15). The Subcriticality Orange Path does not take priority over any Red Path.

B. Incorrect based on the rules of use for safety function status trees (ND-95.3-LP-26

Page 15). The Heat Sink Yellow Path does not lake priority over Containment Red

Path.

6. Incorrect based on the rules of use for safety function status trees (ND-95.3-LP-26

Page 15). Core Cooling Orange Path does not take priority over Containment Red

Path.

D. Correct based on the rules of use for safety function status trees (ND-95.3-LP-26

Page 15). The Containment Red Path takes priority over the other paths. Only

knowledge of safety function priority rules are needed to answer this question.

022 Containment Cooling

(32.422: KnowBedge of the bases for prioritizing safety functions during abnormal and

emergency operations.

Turkey Point Bank Question TP03301

Surry Nuclear Plant 2QO4-301

DRAFT SWO Bnital Exam

i

Unit 2 is operating at 100% power with Chilled CC in service to containment.

I

I

2-CD-REF-IA trips due to a fault.

1

Which ONE of the following describes the effect on Unit 2 containment parameters?

A. Indicated partial pressure will increase. Containment temperature will decrease.

B. Indicated partial pressure will increase. Containment temperature will increase.

6. Indicated partial pressure will decrease. Containment temperature will decrease.

D! Indicated partial pressure will decrease. Containment temperature will increase.

I

Surry (Utility should add noun names to equipment in the stern.)

References:

ND-88.5-LP-1, Component Cooling, Rev. 99

Distractor Analysis:

A. incorrect because partial pressure will decrease due to loss of chilled CC.

B. incorrect because partial pressure will decrease due to loss of chilled CC.

6. Incorrect because containment temperature will increase due to a loss of chilled

D. Correct because partial pressure will decrease and containment temperature will

cc .

increase due to a loss of chilled CC.

Bamk Question # 544

022 Containment Cooling

K3.02: Knowledge of the effect that a loss or malfunction of the CCS will have on the

following: Containment Instrument Readings.

Surly Nuclear Plant 2004-301

DRAFT SRO lnital Exam

23. 026A2.07 00 1/21 l/CONTAINMENT S P K A E A 3.6/3 .9/1V/SK042~h/R/MM/SDK

~

~.

~

~

~

__

-

The following Unit 1 conditions exist:

~ Safety Injection has actuated

- Containment Pressure peaked at 28 psia

- Current Containment Pressure is 15.8 psia

- "IA', "2A" and "2B" Recirculation Spray Pumps are operating

- "1 B" Recirculation Spray Pump tripped on Overload (OL)

- 1A-E7, RS PP l A VIB, annunciates and the alarm cannot be cleared

r

A Large Break LQCA has occurred inside containment

Which ONE of the following states the correct operator action for these conditions?

i

A. Secure Recirculation Spray Pump "IA" using the handswitch in the control room.

B:' Place the Recirc Spray Pump 1A in PTL, then secure Recirculation Spray Pump

C. Reset CLS, then place the handswitch for Recirculation Spray Pump "1A" in PTL.

D. Allow Recirculation Spray Pump "1A" to operate, but monitor vibrations closely.

"1A" locally at the breaker (14H4).

~

~.

~

_. -

~~

~

_

_

_

.

~

-

Surry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

Surry

References:

ND-91 -LP-5, Containment Spray System, Rev. 13

NB-91-LP-6, Recirculation Spray System, Rev. 9

1 -RM-A7, RSISW WX A ALERT/FAItlBWE, Rev. 5

Distractor Analysis:

A. Incorrect because with CLS present, the handswitch in the control F Q O ~

cannot be

used to secure the pump. Containment pressure must be less than 12 psia Bo reset

CLS. Pressure currently is 15.8 psia.

ARP will have the operator place the handswitch in PTL, but the lesson plan

(ND-91-LP-6 Page 6) states that the pump cannot be secured from the control

room with CLS present. Furthermore, the ARP gives guidance to secure the

distressed pump as long as two other WS Pumps are operating. The stem states

that two other pumps are operating ("2A" and "2B").

6. Bncorrect because the CLS cannot be reset until containment pressure is less than

12 psia.

B. Incorrect because the ARP gives guidance to secu~e the distressed pump as long

as two other RS Pumps are operating. The stem states that two other pumps are

operating ("2A" and 2B)~

3. Correct because local operation of the breaker will stop the pump. In addition, the

026 Containment Spray

142.87: Ability to (a) predict the impacts of the following malfunctions or operations on

the CSS; and (b) based or! those predictions, use procedures to correct, control, or

mitigate the consequnces sf those malfunctions or operations: Loss of containment

spray suction when in recirculation mode, possibly caused by clogged sump screen,

pump inlet high temperature (exceeded cavitation, voiding), or sump level below cutoff

(interlock) limit.

Note:

The ARP states that high vibration alarms may be caused by cavitation of the pump.

Cavitation could be caused by high water temp, low water level, etc.

Surry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

-

24. 02GAK3.02

~

00 1/ 1IlICCW ~-

SAFETY INJECTIONIME~5/3.4/M/SK(P130

~.

IIBUMAB/SDR -

~.

A High Steam Flow Safety Injection Signal is received.

Which ONE of the following correctly describes the response of the Component

Cooling Water System components?

A! TV-CC-I 09A and B (CC Isolation Valves from RHF?) close and TV-CC-11 OA, B, and

C (Reactor Cont Air Recirc Cooler CC Outlet Flow Outside Trip Valve) remain as-is.

B. TV-CC-IO9A and B (CC Isohtion Valves from RHR) remain as-is and PV-CC-1 IOA,

B, and 6 (Reactor Cont Air Recirc Cooler CC Outlet Flow Outside Trig Valve)

remain as-is.

C (Reactor Cont Ais Recirc Cooler CC Outlet Flow Outside Trip Valve) close.

D. TV-CC-109A and B (CC Isolation Valves from RHR) remain as-is and TV-CG-l10A,

B, and C (Reactor Cont Air Recirc Cooler CC Outlet Flow Outside Trip Valve) close.

References:

88-05-01

~ Component Cooling Water System, Rev. 19

Distractor Analysis:

A. Correct because lesson plan states CC-I 09 closes on Phase 6 and 1 10 closes ora

Phase HI isolation.

B. Incorrect because lesson plan states CC-189 closes on Phase I and 1 10 only closes

on Phase I l l isolation.

closes on Phase Ill isolation.

Phase Ill isolate.

C. Incorrect because lesson plan states CC-109 closes on Phase

D. Incorrect because lesson plan states CC-109 closes on Phase

and 110 only

and I10 closes on

026 Loss of Component Cooling

AK3.02: Knowledge of the reasons tor the following responses as they apply to Loss of

Cooling Water: The automatic actions (alignments) within the CCWS resulting from the

actuation of the ESFAS.

The loss of CCW occurs in part of the system due to the ESFAS isolation of

TC-CC-109NB.

Serrry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

The following Unit 1 conditions exist:

~

A Large Break LOCA has occurred

- Safety Injection has actuated

~ Containment Pressure peaked at 27 psia

- RCS subcooling is 0 OF

~ Steam Generator bevels are 22% and slowly rising

- RWST emptied while performing E%-1 3,

Transfer to Cold Leg Recirculation

- ES-1.3, Transfer to Cold Leg Recirculation, has been completed and the crew has

- Alf equipment operated normally

transitisned back to E-1 , Loss of Reactor or Secondary Coolant

Which ONE of the following alarms is consistent given the above plant conditions?

A. 1 E-A1 , HI-HI GTMT PRESS CLS cn-1

B! IB-BI, CS PP la LOCKOUT OW OL TRIP

C. 1A-147, RS PP 1A LOCKOUT OW OL TRIP

B. 1B-F6, CTMT INST AIR HBR LO PRESS

Surry Nuclear Plant 2084-301

DRAFT SRO M a l Exam

References:

1 -E-1

~ Loss of Reactor or Secondary Coolant, Rev. 21

1 -ES-t.3, Transfer to Cold Leg Recirculation, Rev. 12

1 E-81, HI-HI CTMT PRESS ChS CH-7, Rev. 0

15-F6,

CTMT INST AIR HDR LO PRESS, Rev. 1

NB-91 -LP-5, Containment Spray System, Rev. 13

NB-91 -LP-6, Recirculation Spray System, Rev. 9

IB-Bf, CS PP 1A LOCKOUT OR OL TRIP, Rev. 0

1 AD7, RS PP 1 A LOCKOUT OR Ob TRIP, Rev. 0

Distractor Analysis:

A. Incorrect because containment pressure is now less than the setpoint, which is

known by CLS having been reset. As a pala of going to Cold Leg Wecirc, CLS and

SI must be reset.

B. Correct because 1 -ES-1.3 has been completed and the RWST has been emptied;

therefore, the CS Pumps would be placed in PPL due to the lack of a suction

source (cavitation). Placing the CS Pumps in PTL yields 1 B-B1 for CS Pump 1 A.

AUTO when stopped.

and instrument air would have been restored to containment.

C. Incorrect because Outside Recirc Spray Pump 1A would be placed in

D. Incorrect because CLS and SI must have been reset prior to completion of 1 -ES-f.3

026 Containment Spray

G2.4.46: Ability t~ verify that alarms are consistent with plant conditions.

Surry Nuclear Plant 2004-307

DRAFT SRO lnital Exam

Which ONE of the fsilowing describes the operation of the lodime Filtration Fans

(1 -VS-F-3A/33)?

B. Automatically stad on a containment gas high alarm.

6. AutomaticalEy stop on a Hi-Hi CLS signal.

D:' Must be manually started under all conditions.

Surry

References:

ND-88.4-LP-6, Containment Ventilation, Rev. 5

Distractor Analysis:

A. Incorrect because fans are only manually operated.

B. Incorrect because fans are only manually operated.

C. incorrect because fans are only manually operated.

B. Correct because fans are only manually operated.

(327 Containment lodine Removal

A4.83: Ability to manually sperate and / of monitor in the control r ~ o m : Cl WS fans.

Question Status:

Surry Bank ILT Question #741

Surry Nuclear Plant 2804-301

DRAFT SRO lnital Exam

27. 027AK3 01

__ 001/1/1/PFG3SUKILBK

___

SPRAY/C/A 3 513 8 ~ / S R 0 4 3 0 1 ~ h . I ~ ~ / S L ) I K

~.

The following Unit I conditions exist:

The Reactor is at 100% Power.

- A malfunction in the Pressurizer Heater Control Circuit has resulted in Proportional

Heaters being de-energized.

~

A small amount of leakage in the Pressurizer Auxiliary Spray Valve is occurring.

- Pressurizer Pressure is 2215 psig and slowly lowering.

1-AP-31 .OQ, Increasing or Decreasing RCS Pressure, has beers entered.

Which ONE of the following states the correct position of the normal sprays and

backup heaters?

I

A. Normal sprays are OFF (valves closed) and backup heaters are ON.

B. Normal sprays are ON (valves open) and backup heaters are OFF.

CY Normal sprays are OFF (valves closed) and backup heaters are OFF.

D. Normal sprays are ON (valves open) and backup heaters are ON.

I

-

Surry

Ref@ re nces:

ND-93.3-LP-5, Pressurizer Pressure Control, Rev. 9

1C-B8, PRZR LO PRESS, Rev. 1

1 4P-31 .QO,

Increasing or Decreasing RCS Pressure, Rev. 6

Distractor Analysis:

A. Incorrect because backup heaters do not energize until 2210 psig.

B. Incorrect because spray valves do not start to open until 2255 psig.

C. Correct because backup heaters do not energize until 221 Q psig and spray valves

D. Incorrect because backup heaters do not energize until 221 0 psig.

027 Pressurizer Pressure Control System Malfunction

AK3.01: Knowledge of the reasons for the following responses as they apply to

pressurizer pressure control malfunctions: Isolation of PZR spray following loss of PZR

heaters.

do not open until 2255 psig.

Sury Nuclear Plant 2004-301

DRAFT SRO lnital Exam

28. 02SG2.2 -_

12 001/212/HYDRQGEN

-

IPECOMBIWWCIA

-_

~

3 O/3.6/N/SR0430IIRIhlABISDK

-

-

~

.~

The following Unit 1 conditions exist:

- The plant is at 50% power

- 1 -PT-37.2, Electric Hydrogen Recombiner, is about to be performed to determine the

reference power that would be used in the event that the Recombiners are used

following a LOCA.

Which ONE of the following correctly states 1 -PT-37.2 limitations that are applicable

during the performance of this test?

A:' At no time should the heater temperature be allowed to exceed 1400 O F as

monitored by the highest thermocouple reading AND containment hydrogen

concentration must be verified to be less than 0.75%.

B. At no time should the heater temperature be allowed to exceed 14QQ

O F as

monitored by the highest thermocouple reading AND containment hydrogen

concentration must be verified to be less than 1 .OQ%.

C. At no time should the heater temperature be allowed to exceed 1300 'F as

monitored by the highest therm~couple reading AND containment hydrogen

concentration must be verified to be less than 0.95%.

D. At no time should the heater temperature be allowed to exceed 1300 O F as

monitored by the highest thermocouple reading AND containment hydrogen

concentration must be verified to be less than 1 .OO%.

Surry Nuclear Plant 2004-381

DRAFT SRO lnital Exam

References:

1 -PB-37.2, Electric Hydrogen Recombiner, Rev. 9

Distractor Analysis:

A. Correct because these are both requirements listed on Section 4.0 sf 1 -PT-37.2.

The unit is at power, therefore 4.3 states that containment hydrogen concentration

must be verified less than 0.75% (being at power and making the operator

determine if 4.3 applies is part of what makes the question C/A). Section 4.2 states

that heater temperature must remain less than 1400 O F at all times.

B. Incorrect because verifying containment hydrogen less than I % is not the correct

requirement. Plausible because applicant may not know that the requirement is

0.75%, vice 1.0%.

6. Incorrect because verifying the highest temperature less than 1300 O F is not the

correct requirement. True ~ if the operator ensures temperature is less than

1300 O F , then he has also ensured that it is less than 1400 O F , but this question

tests the knowledge of the requirement, net simply a method for meeting the

requirement. Plausible because applicant may not know the temperature

requirement.

D. Incorrect because of reasons in C and 63 distractor analysis.

028 Hydrogen Recombiner and Purge Control

G2.2.12: Knowledge of surveillance procedures.

Surly Nuclear Plant 2004-301

DRAFT SRO inital Exam

Which ONE of the following describes the reason why charging pump suctions are

manually aligned to the WWST during an ATWS vice manually initiating a Safety

Injection?

A. Prompt operator action will ensure the most direct method of bosating into the WCS

and manual alignment of charging pump suction to the RWST prevents

compounding the problem by charging the RCS solid via Safety Injection.

and initiation of SI would reduce the possible paths for emergency boration and add

to an RCS overpressure condition if one exists.

B. Prompt operator action will ensure the most direct method of bosating into the WCS

C. Manual initiation of Safety Injection would delay the addition of borated water to the

RCS and complicate the recovery actions. Alignment of charging pump suction to

the RWST is the most direct method of borating the RCS.

DY Manual initiation of SI would result in the undesirable trip of Main Feedwater Pumps

and alignment of Charging Pump suction to the RWST is the most direct method of

borating the RCS.

Surry Nuclear Plant 2004-301

DRAW SRQ lnital Exam

References:

ND-95.3-LP-36-DRR, FR-S.1 Response to Nuclear Power Generation / AtWS, Rev. 10

FR-S.1, Response to Nuclear Power Generation / ATWS, Rev. 15

Distractor Analysis:

A. Incorrect because the concern with initiating SI is not creating a solid plant

condition, but with reducing the probability sf maintaining a secondary heat sink

because MFW pumps will trip upon Si initiation.

B. Incorrect because the concern with initiating SI is not creating 8 high WCS pressure

condition, but with reducing the probability of maintaining a secondary heat sink

because MFW pumps will trip upon SI initiation.

C. Incorrect because manual initiation wouDd not delay addition of borated water. The

concern is with reducing the probability of maintaining a seondary heat sink

because MFW pumps will trip upon SI initiation.

D. Correct because per NB-95.3-kP-36-DRIRt FR-S.1 Response to Nuclear Power

Generation / ATWS, both of these statements accurately reflect the basis for Step

4.

029 ATWS

EK3.09: Knowledge of the reasons for the following responses as they apply to the

ATWS: Opening centrifugal charging pump suction valves from RWST.

Modified CLT Bank Question # 3390

Surry Nuclear Plant 2004-381

DRAFT SRO lnital Exam

30.

032,441.01 001/1/2/SOIRCE INTI:KMEDIAT\\TE/C/A 3.1/3 4/B/SRO430l/R/MAB/SDR

-

~

__

~

-~

-

The following conditions exists:

- Present time is 1428 hours0.0165 days <br />0.397 hours <br />0.00236 weeks <br />5.43354e-4 months <br />

- Reactor tripped at 1405 hours0.0163 days <br />0.39 hours <br />0.00232 weeks <br />5.346025e-4 months <br />

~

All Rod Bottom Lights are lit

- N-35 reading is 2 x 10-l' amps

- N-36 reading is 4 x IO-" amps

- Source Range is not energized

- Power level prior to trip was 98%

Which ONE of the following describes the correct actions given the above parameters?

A. When both IR channels read K 5 x lUro amps, verify source range channels

energized.

B. Place the source range trip bypass switches in the NORMAL position.

6:'

Energize the source range channels by depressing the source range manual reset

pushbuttons.

D. Transfer NR-45 to one S Q U K ~ range and one intermediate range channel.

~

~

~

~

~

~

~

~.

Ref e re nces :

ND-93.2-LP-3, Intermediate Range Nls, Rev. 10.

Distraetsr Analysis:

A. Incorrect because SR energizes at 2/2 IR

B. Incorrect because SR should already be energized in the NORMAL position and this

C. Correct because IR are under-compensated and SR must be manually energized.

8. Incorrect because SI3 should both be energized.

5 x

amps.

action would not energize the SR.

032 Loss of Source Range NI

AAI .01 : Ability to operate and / or monitor the following as they apply to loss of source

range nuclear instrumentation: Manual restoration of power.

Surty Nuclear Plant 2004-301

DRAFT SRO lnital Exam

The following Unit 1 conditions exist:

- Critical approach has just been completed.

- Reactor is stable at the Point of Adding Heat.

One Intermediate Range (IR) Nuciear Instrument (Nl) is suspected of displaying

inaccurate indications.

Which ONE of the following correctly describes the expected Power Range (PR) NI

and the known operable IR NI indications for the above conditions to verify that the

suspect IF? NI is in fact falsely indicating?

A. IR = 2.5 x lo-* Amps; PW between 0.2 and 1 Yo

B!' IR = 2.0 x 10" Amps; PR between 0.2 and 1 %

D. IR = 1 .O x IC5 Amps; PR 6 0.2 %

References:

NB-93.2-LP-4, Power Range Nls, Rev. 16

1 -GOPI

.4, Unit Startup, HSD to 2% Reactor Power, Rev. 29

Distractor Analysis:

A. Incorrect because 2.5 x IO-* Amps is about where critical data is taken (too low).

B. Correct based on above two references: NB-93.2-LP-4 (HTT-4.3) & 1-GOP-1.4

6.

Incorrect because 1 .0 x 1 0-* Amps is about where critical data is taken (too low).

D. Incorrect because 1.6 x ID5 Amps is above the POAH and should correspond to

(Page 29 CAUTION).

about 2% power.

033 boss of Intermediate Range NI

AA2.04: Ability to determine and interpret the following as they apply to the loss of

intermediate range nuclear instrumentation: Satisfactory overlap between

source-range, intermediate-range, and power-range instrumentation.

Surgy Nuclear Plant 2004-301

DRAFT SRQ lnital Exam

32. 6334A4.01

~

001/2/2/RADIATION

~

MC"OW~3.3/3.4/~/SiSR043Ul/~~rZRiSDR

_

_

~

-

_

_

_

~

~-

Unit 1 is in a refueling outage when the following events occur:

- Purge Isolation Valves (MOV-VS-I OOA, B, C. and D) Close

.. Unit Purge Supply Fans (4A and 4B) Trip

~ Containment Instrument Air Suction Valves (PV-IA-101 N B ) Close

Which ONE of the following radiation monitors could have caused these actions?

i

A. Process Vent Particulate and Gas Monitors (RM-WI-IO1 / 102)

5. RM-I 61 (Containment High Range Gamma Monitor)

C:' WM-162 (Manipulator Crane Monitor)

D. RM-I 63 (Reactor Containment Area Monitor)

~-

~

Surry

References:

ND-93.5-LP-1, Pre-TMl Radiation Monitoring System, Rev. 5

Distractor Analysis:

A. incorrect because RM-RI-101 / 102 do not cause these actions.

B. Incorrect because RM-I 61 does not cause these actions.

C. Correct per ND-93.5-LP-1.

5. incorrect because RM-163 does not cause these actions.

834 Fuel Handling Equipment

A4.01: Ability to manually operate and / or monitor in the control room: Radiation

Levels.

Surry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

33. 03SA3.01

.~

001/2/2/STEAM GENERATOWCIA

- _ _ _

4.01~.3/N/SR04jOl/K/MAR/SDM

~-

__

___

.-

~

-~

r -~

I

I

The following Unit 1 conditions exist:

- Plant is stable at 75% Power

~

Ab SG Steam Line FT-MS-475 is selected for Steam Generator bevel control

- A SG Steam kine PT-MS-475 fails high

I

Which ONE of the following correctly describes the impact on the A Steam Generator

j

Level CQntrQl?

A. Feedwater Regulating Valve opens because indicated steam flow is greater than

indicated feedwater flow.

1

B. Feedwater Regulating Valve does not move as a result sf the failure.

6. Feedwater Regulating Valve closes because the pressure transmitter is

overcompensating for density.

D. Feedwater Regulating Valve opens to reduce the level error created by the failure.

References:

ND-93.3-LP-8, SG Water Level Control System, Rev. 6

Distractor Analysis:

A. Incorrect because PT-MS-475 does not cornpensate steam flow for FT-MS-475.

B. Correct because PV-MS-475 does not compensate steam flow for W-MS-475.

C. Incorrect because PT-MS-475 does not compensate steam flow for FT-MS-475.

D. Incorrect because $8-MS-475 does not compensate steam flow for FT-MS-475.

035 Steam Generator

A3.81: Ability to monitor automatic operation of the S/G including: S/G water level

control.

Surry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

Which ONE of the following is correct regarding safety injection termination during a

steam generator tube rupture event?

Safety Injection termination ...

A:' may occur with total AFW flow less than 350 gpm as long as 350 gpm is available.

B. may occur with Pressurizer level less than 22% as long as level is increasing.

6. may not not occur with a void in the reactor head due to presenting RCS pressure

control problems.

control problems.

D. may not not occur with a void in the reactor head due to presenting RCS leV@l

Sur9

References:

ND-95.3-LP-13, E-3 Steam Generator Tube Rupture, Rev. I 1

I-E-3,

Steam Generator Tube Rupture, Rev. 19

Distractor Analysis:

A. Correct because if no intact SG is available the ruptured SG will be used to cool the

RCS. In this instance the AFW flow may be less than 950 gprn, but 350 gpm must

still be available to that SG. If sufficient flow is available, then SI termination criteria

is considered to be met (MD-95.3-LP-13).

B. Incorrect because pressurizer level must be greater than 22% to meet the SI

termination criteria.

C. Incorrect because safety injection may be terminated when there is a void in the

reactor head. This will present some challenges with RCS pressure and level

control, but it is not a large enough concern to prevent SI termination if the specified

criteria are met (ND-95.3-LP-13).

D. Incorrect because safety injection may be terminated when there is a void in the

reactor head. This will present some challenges with RCS pressure and level

control, but it is not a large enough concern to prevent SI termination if the specified

criteria are met (ND-95.3-LP-13).

038 Steam Gen. Tube Rupture

EK3.09: Knowledge of the reasons for the following responses as they apply to the

SGTR: Criteria for securing / throttling ECCS.

c

Surry Nuclear Plant 2004-381

DRAFT SRO lnital Exam

With Unit 1 at 100% power, the Condenser Air Ejector and Main Steam bine Radiation

Monitor alarms are recieved. The Condenser Air Ejector Radiation Monitor reads 700

cpm (ALERT and HIGH alarms are in) while tocal Main Steam NWC Radiation Monitors

read "A" .03 rnr/hr, and "B" -01 mr/hr, and "C"

.01 mdhr. The Team has implemented

1 -AP-l8.00, Excessive RCS Leakage, and the WCS leak rate is determined to be 60

gpm.

Which ONE of the following describes the actions required?

A. Verify automatic Condenser Air Ejector divert to Containment, intiate 1 -AP-24.00

(Minor SG Tube Leak), manually trip the reactor and go to 8-E-0 (Reactor Trip or

Safety injection).

B. Verify automatic SGBD w/ trip isolation and Condenser Air Ejector divert to

Containment, manually trip the reactor and initiate SI, Go to 1 -E-0.

e. Verify automatic Condenser Air Ejector divert to Containment, initiate 1 AP-24.01

(Large Steam Generator Tube Leak), verify letdown isolated, and commence a

normal Unit shutdown lAW GOPs.

D I Verify automatic Condenser Air Ejector divert to Containment, initiate 1 -AP-24.01

(Large Steam Generator Tube Leak), and manually trip the reactor and go to 1 - L O .

Surry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

References:

MD-89.3-LP-2, Main Condensate System, Rev. I6

ND-93.5-LP-3, Post-TMI Radiation Monitoring System, Rev. 6

1 -AP-I 6.60, Excessive RCS Leakage, Rev. 1 1

1 -AP-24.08, Minor SG Tube beak, Rev. 8

1 -AP-24.01, Large Steam Generator Tube Leak, Rev. 11

Distractor Analysis:

A. Incorrect because 60 ggm leakage is more than minor. AP-24.01 should be entered

for a large steam generator tube leak.

B. Incorrect because St should not be initiated.

C. hxrrect because the reactor must be manually tripped with leakage greater than

B. Correct because air ejectors will divert to containment on an air ejector high

50 gpm.

radiation, AP-24.01 should be entered due to 60 gpm leak rate with air ejector high

radiation, and E-0 should be entered following a manual reactor trip.

039 Main and Reheat Steam

A I .09: Ability to predict and / or monitor changes in parameters (to prevent exceeding

design limits) assosiated with operating the MRSS controls including: Main stearn line

radiation monitors.

SUFY Nuclear Plant 2004-301

DRAFT SRO lnital Exam

-~ _ _

__

36. 054G2.4.31 001/lillALARMS RODSiCYA 3.313.4~iSR04301~ARISDIP

_ _

_ _ _ _ _ _

The following Unit 1 conditions exist:

- The Reactor was operating at 78% power when a loss of the "A" Feedwater Pump

- The Team is taking the required immediate actions in acordance with 1 -AP-21 .QO,

- The Reactor Operator is driving rods in manual to lower Tavg

- Tavg is within 3 O F of Tref

- Annunciator 1 G-68,

ROD BANK D LO LIMIT, has annunciated

occurred

"Lsss of Main Feedwater Flow"

Which ONE of the following is the correct response to the given plant conditions?

A. Shutdown margin is not sufficient for the given plant conditions and operators

should emergency borate to regain the required shutdown margin.

BJ The operator has driven rods in too Bar for the existing boron concenration and

should borate from the Boric Acid Tanks.

C. Shutdown margin is not sufficient for the given plant conditions and operators

should trip the Reactor and go to 1-E-8, Reactor Trip or Safety Injection.

B. The turbine load has decreased too far and the operator should raise turbine load.

__

~

_

_

~

-

~

_

_

~

Surry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

Ref e rences :

ND-89.3-LP-3, Main Feedwater System, Rev. 12

NB-95.1-LP-4, Loss of Feedwater, Rev. 3

1 -AP-21 .OO, Loss of Main Feedwater Flow, Rev. 5

1G-G8, ROD BANK B LO LIMIT, Rev. 0

Distractor Analysis:

A. Incorrect because (1) not enough information is given to make the determimation

that SBM is insufficient, and (2) even if SBM is not above that which is required,

emergency boration would not be the preferred method for regaining the required

SDM. This is clearly the wrong method for boration because xenon is building in

and only small borations would be desired to withdraw rods to clear the alarm.

alarm. Bsration from the Boric Acid Tanks would be the correct mitigation strategy

and as such, is directed by the ARP. Operators would only borate the necessary

amount to clear the alarm.

designed to handle this magnitude of transient. Furthermore, the plant does not

need to be tripped with rods approaching or below insertion limits. Rod positions

just have to be restored to within limits.

D. Incorrect because turbine load should not be raised. Immediate actions have the

operators reduce turbine load to match steam flow and feed flow. Raising turbine

load under these conditions would not be the correct action. It would also be

nonconservtaive to add positive reactivity via the turbine during a transient condition

such as described in the stern.

B. Correct because rods being within 10 steps of its insertion limit would cause the

G. hcorrect because the initial power level was less than 85% and the plant is

054 Loss of Main Feedwater

G2.4.31: Knowledge of annunciators and indications and use of response instructions.

Bank Question from 2003 Farley Exam (Farley WA was 05462.2.20).

Surry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

37. 055EK1.01 001/1/1/BATTERY/C/A 3.3/3.7/B/SWU4301/IUP/I..\\B/~~

~

r--

~

~

___

The following plant conditions exist:

- A loss of all AC power has occurred.

~ Operators have implemented ECA-0.0, Loss of All AC Power.

- Attempts to regain A@ power have failed.

- Operators are performing ECA-0.0, Step 28, "Check DC Bus toads"

Which ONE of the following should be performed to lower the Black Battery discharge

rate by the largest amount per ECA-Q.O?

A. Secure Air Side Seal Oil Pump only.

B. Secure Air Side Seal Oil Pump and Emergency Turbine Lube Oil Pump.

C. Secure Air Side Seal Oil Backup Pump only.

D:' Secure Air Side Seal Oil Backup Pump and Emergency Turbine Lube Oil Pump.

1

Surry Nuclear Plant 2004-301

DRAFT SRB lnital Exam

References:

NB-90.3-LP-6, 125 Vdc Distribution, Rev. 10

ECA-0.0, LOSS of All AC Power, Rev. 21

Distractor Analysis:

A. Incorrect because the Air Side Seal Oil Pump is not a DC load, as is

the Air Side Seal Oil Backup Pump. Plausible because the candidate may not

know major Black Battery DC Loads, or may not know what actions are permitted

by ECA-0.0.

B. Incorrect because the Air Side Seal Oil Pump is not a DC load, as is

the Air Side Seal Oil Backup Pump. Plausible because the candidate may not

know major Black Battery DC Loads, or may not know what actions are permitted

(ASSOBUP) and Emergency Turbine Lube Oil Pump, not just ASSOBUP.

Plausible because the applicant may not know that there is more than one pump to

secure to conserve Black Batteries.

B. Correct because per ECA-0.0 step 28 and Basis for this step in NB-95.03-LP-l?,

the purpose is to secure both pumps, which are large Black Battery DC loads, to

conserve the batteries (reducing battery discharge rate, thus prolonging battery life).

by ECA-0.0.

C. Incorrect ECA-0.0 will direct the securing of both Air Side Seal Oil Backup Pump

Surry ILT Bank Question #724

055 Station Blackout

EK1.01: Knowledge of the operational implications of the following concepts as they

apply to the Station Blackout: Effect of battery discharge rate on capacity.

Surry Nuclear Plant 2004-301

DRAFT SRB lnital Exam

The following Unit 1 conditions exist:

- Two Main Feedwater Pumps are operating

- Reactor Power = 85%

- Condensate Pumps 1 -6N-P-1 A and B are operating

- Condensate Pump 1 -CN-P-1 C is Tagged Out of Service

- Condensate Pump 1 -6N-PIA trips and camnot be restarted

- Main Feedwater Pump Suction Pressure = 105 psig and slowly lowering

- Stearn Generator Levels are slowly lowering

- 1H-F8, FW PP SUCT HER LO PRESS, is in alarm

W hich ONE of the following is the correct operator action?

Pa! Enter 1 -AP-21 .QO, Loss of Main Feedwater Flow, and reduce turbine load to match

steam flow and feedwater flow.

B. Manually trip the Reactor and enter E-0, Reactor Trip or Safety injection.

@. Secure one of the operating Main Feedwater Pumps and monitor the operating

Main Feedwater Pump Suction Pressure.

D. Enter 1-AP-21 .00, Loss of Main Feedwater Flow, and start a second HP Drain

Pump.

Surry Nuclear Plant 2884-301

DRAFT SRO lnital Exam

Surry

References:

ND-89.3-LP-2, Main Condensate System, Rev. 16

NB-89.3-LP-3, Main Feedwater System, Rev. 12

ND-95.1-LP-4, Loss of Feedwater, Rev. 3

1-AQ-21 .00, Loss of Main Feedwater Flow, Rev. 5

1 H-F8, FW PP SUCP HDW LO PRESS, Rev. 0

1J-G4, CN PPS DISCH HDR LO PRESS, Rev. 0

1 H-G8, FW PP DISCH HBR LO PRESS, Rev. 0

Distractor Analysis:

A. Correct because MFW Pump Low Suction Pressure and Discharge Pressure

Alarms are entry conditions into A$-21 .00. Furthermore, with power at 65% the

direction is to reduce turbine load to match steam and feed flows. This will also

help to recover MWM Pump suctisnldischarge pressure.

B. Incorrect because no trip criteria are met and AP-21 .OO directs power reduction.

C. Incorrect because tripping a MFW Pump will not alleviate the issue and there is no

procedural guidance to trip a M W

Pump. Typically a M

W

Pump will be secured

at about 40% power.

D. lncorret because there is no guidance to start a second heater drain pump. The

correct response is to lower turbine load.

056 Condensate

A2.04: Ability to (a) predict the impacts of the following malfunctions or operations on

the Condensate System; and ($1 based on t h ~ s e

predictions, use procedures to

correct, control, or mitigate the consequences of those malfunctions or operations:

Loss of condensate pumps.

Surry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

The following conditions exist:

- A Loss of 08-Site Power has occurred

- #I Emergency Diesel Generator has started but failed to auto load

~

It has been determined that the auto-closure circuit for 75H3, #I EDG Output

~ When the operator places the sync switch for 15H3 to "ON" he observes 120 volts 01

Breaker, is inoperable and that 15H3 can be manually closed

the "incoming" meter, 0 volts on the "running" meter, and the synchroscope is

stationary at "3-0'cl~cSc"

I

Which ONE of the following actions is necessary prior to closing 15613?

A. Raise EDG speed until the synchroscope is turning slowly in the fast direction, then

close 15H3 at "1 1 o'clock".

I

B. Momentarily press the "field flash" pushbutton, then sync and close 15H3.

C. Raise EDG voltage until the running meter indicates 120 volts, then sync and close

15H3.

I

B7 No additional action is necessary. Close 15H3.

References:

ND-90.3-LP-I , Emergency Diesel Generator, Rev. I4

ND-98.3-LP-4, Station Service and Emergency Distribution Protection and Control,

Rev. 17

Distractor Analysis:

A. incorrect because the bus is dead. Raising EDG speed will not synchronize the

phases.

B. incorrect because it will not be possible to synchronize (nor is it necessary because

the bus is dead). Also, field flash PB does not need to be pushed.

C. Incorrect because raising the EDG voltage will not raise running voltage. Incoming

voltage is the EDG voltage (not running voltage).

D. Correct because the synchroscope has been turned on, there is no over-current or

differential and the aux trip relay does not need to be reset (ND-90.3-LP-7 pg. 18).

Therefore, all criteria for manually closing the breaker are met.

056 Loss of Off-Site Power

AAI .26: Ability to operate and / or monitor the following as they apply to the Loss of

Off-Site Power: Circuit 5reakers

Surry Nuclear Plant 2004-301

DRAFT SWO lnitai Exam

~

-

40.

-__

053AK3.01 001/1/1/VIThT,

__ -

AC/C/A

~

4.1/4

._

4/F3ISM04301/R/MAB/SDR

--

-

___

Which ONE of the following reasons correctly states why the reactor would be tripped

for a sustained loss of Vital Bus II?

A. Power to the Reactor Protection System is lost.

B. Pressurizer pressure control is Isst.

C. Control of Steam Generator Feed Regulating Valve is lost.

Surry

References:

1 -AP-lQ.O2, Loss of Vital Bus II, Rev. 9

ND-90.3-LP-5, Vital and Serni-Vital Bus Distribution, Rev. 1 I

Bistractsr Analysis:

A. Incorrect because WPS is de-energize to trip. If due to other channel failures, etc.,

the loss of VB Sl will not preclude a trip if one is needed.

B. Incorrect because Pzr P Controller will transfer to AUTO-HOLD, but MANUAL

control is still possible, thus precluding the need for rx trip.

C. Incorrect because FW-FCV-1488 Flow Controller will transfer to AUTO-HOLD, but

MANUAL control is still possible, thus precluding the need for rx trip.

D. Correct because Component Cooling is lost to the 'Wb RCP Lube Oil Cooler. RCP

Parameters will eventually exceed limits (1-AP-10.02 Att. I), requiring that the RCP

be secured following a manual IX trip.

057 Loss of Vital AC Inst. Bus

AK3.61: Knowledge of the reasons for the follo~tng responses as they apply to the

Loss of Vital Ai2 Instrument Bus: Actions contained in EOP for loss of vital ac electrical

bus.

Surry IhT Bank Question #223

Surry Nuclear Plant 2084-301

DRAFT SRO M a l Exam

44. 058M2.01 ~

~

~

001/1/1A,OSS OF DC POWER/C'/A 3.714 l/N/SR~Ol/pvMAu/SH>w - - .___

r-

~

The following Unit 1 conditons exist:

- 1 K-88, UPS SYSTEM TROUBLE, annunciates

- 4 K-A7, BATT SYSTEM 1A TROUBLE, annunciates

- An operator reports that Battery Charger DC Output for UPS 1A-1 reads 0 amps

Which ONE of the following correctly describes the power supply to the associated BC

and Vital AC buses?

A. BC Bus 1A-1 will be supplied by only Battery 1A as indicated by DC Bus voltage

slowly trending down over time and Vital AC Buses 1 and 7A will be supplied by

Bus 1H1-1.

B. DC Bus 1A-I will be supplied by only Battery 1A as indicated by DC Bus voltage

slowly trending down over time and Vital AC Buses 1 and 1A will be supplied by

Bus 1 HI -2.

@. DC Bus 1A-1 will be supplied by UPS $A-2 as indicated by Df.2 Bus Voltage

remaining stable at 125 VBC and the Vital AC Buses 1 and 1A will be supplied by

1Ht-1.

BY 56 Bus 1 A-1 will be supplied by UPS 1 A-2 as indicated by BC Bus Voltage

remaining stable at 125 VBC and the Vital AC Buses 1 and 1 A will be supplied by

1H1-2.

Sur9 Nuclear Plant 2064-381

DRAFT SRO lnital Exam

Surly

References:

ND-90.3-LP-5, Vital and Semi-vital Bus Distribution, Rev. 11

ND-90.3-LP-6, 125 VBC Distribution, Rev. 18

4 K-A7, BATT SYSTEM 1A TROUBLE, Rev. 5

1 K-A8, UPS SYSTEM TROUBLE, Rev. I

1 1448-FE-1 G, Sheet 1 of 1, 125V DC One Line Diagram Surty Power Station Unit 1,

Rev. 33

Distractor Analysis:

A. Incorrect because the battery should not be supplying the BC Bus alone. The DC

Bus is being supplied by the other UPS from 1 HI -2. Also, vital AC Buses 1 and 1 A

are being supplied by Bus 1 HI -2, which is the alternate AC source.

B. Incorrect because the battery should not be supplying the BC Bus alone. The DC

Bus is being supplied by the other UPS from 1 HI-2.

C. Incorrect because the Vital AC Buses 1 and 1A are being supplied by Bus 1 HI -2,

which is the alternate AC source.

B. Correct because the other UPS will still be supplying DC Bus IA-1 and the Alternate

AC Source 1 Hl-2 will supply Vital AC Buses I and 1 A.

058 boss of DC Power

AA2.81: Ability to determine and interpret the following ais they apply to the loss of DC

Power: That a loss of DC Power has occurred; verification that substitute power

sources have come on line.

Surry Nuclear Plant 2004-301

DRAFT SRO M a l Exam

42. -____

05 9.4 1.03 00 1/21 I /MAIN FEEDWATEMEM

.

2.7/2.9/l?ISR044301

..

.___

/TP/Mm/SDR

___

Which ONE of the following set of practices should be observed by operators for

starting the second Main Feedwater Pump per GOP-1.5 (Unit Startup, 2% Reactor

Power to Max Allowable Power) and OP-RM-004 (Main Feedwater System Operation)?

A. The second Main Feedwater Pump should be started prior to exceeding 50% power

to preclude problems with main feedwater flow capability. Following pump start, if

the Main Feedwater Pump Reciculation Valve is in AUTO, the operator should

observe that valve closure will occur as the feed flow rises above 3000 gpm.

B. The second Main Feedwater Pump should be started between 50% power and 65%

power to preclude problems with main feedwater flow capability. Following pump

start, if the Main Feedwater Pump Recirculation Valve is in AUTO, the operator

should observe that valve closure will occur as the feed flow rises above 3286 gpm.

CY The second Main Feedwater Pump should be started prior to exceeding 50% power

to preclude problems with main feedwater flow capability. Operating the second

Main Feedwater Pump on recirculation with the discharge MOV closed should be

minimized to prevent overpressurization of the piping between the discharge cheek

valve and the MOV as the system heats.

D. The second Main Feedwater Pump should be started between 50% power and 65%

power to preclude problems with main feedwater flow capability. Operating the

second Main Feedwater Pump on recirculation with the discharge MOV closed

should be minimized to prevent overpressurization of the piping between the

discharge check valve and the MOV as the system heats.

Surry Nuclear Plant 2084-301

DRAFT SRO lnital Exam

References:

4 -GOPI

5,

Unit Starhp, 2% Reactor Power to Max Allowable Power, Rev. 32

1 -OP-FW-(404, Main Feedwater System Operation, Rev. 8

MB-89.3-LP-3, Main Feedwater System, Rev. 12

Distractor Analysis:

A. Incorrect because reeirc should modulate closed at 4000 gpm.

B. Incorrect because recirc should modulate closed at 4008 gpm.

C. Correct because of NOTE on Pg. 34 of 44 of GOP-1.5 and CAUTION on Pg 12 of

B. Incorrect because second feedwater pump should be started prior to 50% power.

059 Main Feedwater

A I .03 Ability to predict and / or monitor changes in parameters (to prevent exceeding

design limits) associated with operating the MFW controls including: Power level

restrictions for operation of MFW pumps and valves.

34 of OP-FW-004.

SUFV

Nuclear Plant 2004-301

DRAFT SRO lnital Exam

-

43. O59AAI

~

.01 001/1/ULIQT.JlD

-~

-

RAD _.

RELEASElCIA

~

3 3 3

~.

5/M/SR04~M/MARIST)R __

~.

The following Unit 1 conditions exists:

~ The "B" Train of Recirc Spray (RS), the only available train, is in service

- 1 -RM-G7, DlSCH TNL ALERT / FAILURE, annunciates

- I-RM-AB, RSISW HX B ALERT/FAILURE, annunciates

~

Reactor Operator notes the RS/SW HX B Monitor is trending up, but the Discharge

Tunnel Rad Monitor is indicating all EEEEEs with Red and Yellow Lights Lit and

Green Light out.

- A Large Break LOSS sf Coolant Accident has QCCUrWd

Which ONE of the following is the mrrect operator response?

A. Ensure no additional releases are in progress and secure RS.

8.3 Ensure no additional releases are in progress, and increase radiation monitoring.

C. Verify all automatic actions have occurred and reset the Discharge Tunnel Digital

Rate Meter and perform a source check.

D. Verify all automatic actions have occurred and raise the Discharge Tunnel Monitor

set point.

Surry Nuclear Plant 2604-301

DRAW SRQ lnitaE Exam

Surry

References:

ND-93.5-LP-I

~ Pve-TMI Radiation Monitoring System, Rev. 8

I-RM-G7, DlSCH TNL ALERT / FAILURE, Rev. 4

1-WM-A8, RSISW HX B ALEWTFAILUBE, Rev. 3

Distractor Analysis:

A. Incorrect because the last available RS train should not be secured, as stated in

RM-GS and RM-A8 Caution Statements. Plausible because this is the correct

course of action if the other train was available.

B. Correct because the last train of WS should not be secured. Other rad monitors

should be checked to see if blowdowns have been diverted, to verify that there is no

CCW/SW HX leak, and to verify that no CP Bid Liquid releases are occurring.

Additional monitoring is called for by the ARPs due to the fact that the last train of

WS should not be secured.

C. Incorrect because there are no automatic actions to verify. Plausible because the

applicant may not know that there are no auto actions associated with these

particular monitors. With a failed monitor, ARPs will direct 8 reset and source

check, which adds to the plausibility.

D. Incorrect because there are no automatic actions to verify. Plausible because the

applicant may not know that there are not auto actions associated with these

particular monitors and it is not uncommon for an alarm setpeint to be raised to

sled operators of worsening conditions. The Discharge Tunnel Monitor has the

indications of being failed, therefore adjusting the setpoint is not a success path.

(359 Accidental Liquid Radwaste Release

,441.01 : Ability to operate and / or monitor the following as they apply to the Accidental

Liquid Wadwaste Releases: Radioactive-liquid monitor

Modified Surry ILT Bank Question #I977

Surry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

I-CN-TM-1, Emergency Condensate Storage Tank (ECST), is supplying AFW Pumps

for Residual Heat Removal via Steam Generators. 1J-F4, CST 110,000 GAL LO LVL,

has annunciated. ECSP level is 90% and lowering.

Which ONE of the following is correct regarding refilling of the ECST?

A. Filling shall commence prior to the ECST level reaching 54%. AFW pumps must be

secured prior to commencing the fill.

B. Filling may commence after the ECST level drops below 60,000 gaElons as long as

refill begins within two hours of securing the AFW pumps.

6.

AFW Pumps must be secured prior to commencing the fill and the ECST must be

filled within two hours.

D!' Filling of the ECST shall commence prior to the ECST level reaching 54%. AFW

pumps may continue ts operate during the refill.

References:

NB-89.3-LP-4, Auxiliary Feedwater System, Rev. 19

f J-F4, CST 1 10,000 GAL LO LVL, Rev. 3

tech Spec 3.6-1,

Amendment No. 224 and 220

Distractor Analysis:

A. Incorrect because AFW pumps may continue to run during refill based on AWP

1 J-F4 Note.

B. Incorrect because volume must remain above 60,000 gal (54%).

C. ln~o~rect

because AFW pumps do not need to be secured for refill.

D. Correct based on all three of the above references.

(461 Auxiliary Feedwintea

Al.04: Ability to predict and / or monitor changes in parameters (to prevent exceeding

design limits) associated with operating the AFW controls ineluding: AFW source tank

level.

Surry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

I

I

I

6:' Spent Fuel Pit Bridge Crane Radiation Monitor and Ventilation Vent Gaseous

Kadiatisn Monitor

I

45. 061 A M 0 1 00111R/ARM RAD MONITOWMEM

~_

3.5/3.7/W/SR04301/IUMAR/ST)R

-___

If a spent fuel assembly is damaged by being dropped in the spent fuel pool, which

ONE of the following pairs of radiation monitors would indicate an increase in radiation

level?

r

A. Spent Fuel Pit Bridge Crane Radiation Monitor and Auxiliary Building Control

Victoreen Area Radiation Monitor

B. Ventilation Vent Particulate Radiation Monitor and Auxiliary Building Control

Victoreen Area Radiation Monitor

Sur9 (Utility needs to add correct RM equipment numbers.)

References:

ND-93.5-LP-1, Pre-TMI Radiation Monitoring System, Rev. 8

0-AP-22.00, Fuel Handling Abnormal Conditions, Rev. 18

Q-WM-B3, 1-RM-RI-153 HIGH, Rev. 4

0-RM-B4, 1-RM-RI-152 HIGH, Rev 8

Distractor Analysis: (maybe get Some help to provide a little better distractor analysis?)

A. Incorrect because the Aux Bld Control Victoreen Area Radiation Monitor would not

show an increased indication.

B. Incorrect because the Aux Bld Control Victoreen Area Radiation Monitor would not

show an increased indication.

C. Correct because both monitors would show an increased indication.

D. incorrect because a liquid waste process effluent monitor would not see the results

of the failed fuel.

061 ARM System Alarms

AA2.QI : Ability to determine and interpret the following as they apply to the Area

Radiation Monitoring (ARM) System Alarms: ARM panel displays.

Surry Wequal Bank Question #I 18

Surry Nuclear Plant 2004-301

DRAFT Sa0 lnital Exam

46. __

06241.01

~ 001/1/2/EDG 1)IESELMEhI 3.413 8NSR04301IWMABISDR

~

~

~

I

~

r

The following conditions weie noted during the performance of 1 -OPT-EG-001,

Number f Emergency Diesel Generator Monthly Start Exercise Test:

- The EBG was loaded at a rate sf 550 KW/MIN

- The Maximum load attained was 2650 KW

.. The Maximum KVAR was 508 KVAR out

- The output voltage was stable at 4300 VAC

Which ONE of the following was in violation of the EBG Precautions and Limitations

per 1 -0QV-EG-OOI ?

A Load Rate

8. Maximum Load

C. Maximum KVAR out

D. Output voltage

I

I

surry

References:

1 -OPT-EG-001

I Number 1 Emergency Diesel Generator Monthly Start Exercise Test,

Rev. 24

1-QP-EG-001, Number 1 Emergency Diesel Generator, Rev. 17

Distractor Analysis:

A. Correct because the loading rate should not exceed 500 KW/MIN during normal

operations.

B. Incorrect because rnax load rating is 2750 KW.

6. Incorrect because rnax KVAR out is 500 KVAR.

D. Incorrect because output voltage shsuld be maintained between 4800 and 4400

VAC .

062 A 6 Electrical Distribution

Al.01: Ability to predict and / or monitor changes in parameters (to prevent exceeding

design limits) associated with operating the ac distribution system controls including:

Significant D/G Isad limits.

-

-

Scerry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

47. 062AA1 06 001/1/1/SERVICE WATER/MEM 2.9/2.9/U/SR~3301IRIMAU/SDR

_ _

-

-~ _

~

-

-

_

_

.

_

_

-

~

The following Unit 1 conditions exist:

- During testing, an Intake Canal Low Level Isolation Signal is inadvertently actuated

Which ONE of the following correctly states the plant response caused by the Low

Level lsolatisn Signal?

~

Power = 100%

A. 1 -SW-MOV-l02A and B (CCHX and SW-P-4 Supply) will close and can only be

reopened after 5 minutes.

B. l-SW-MQV-102A and B (CCHX and SW-P-4 Supply) will go to 25% open and can

be fully opened after 5 minutes.

CY 1 -SW-MOV-lWA and B (CCHX and SW-P-4 Supply) will close and can be

reopened when the low level signal is reset.

D. 1 -SW-MOV-l02A and B (CCHX and SW-P-4 Supply) will go 25% open and can be

fully opened when the low level signal is reset.

Suvy (Utility needs to verify technical accuracy and provide any additional reference

material (electrical print?)).

References:

ND-89.5-LP-2, Sewice Water System, Rev. 20

Distractor Analysis:

A. Incorrect because the valves will C ~ Q S ~ ,

but cannot be re-opened until Canal Low

bevel Isolation Signal is cleared. If the valves would have been closed due to a

CLS, then they could have been re-opened after 5 minutes even without the CLS

cleared.

B. Incorrect because the valves will go fully closed.

C. Correct because the valves will close, but cannot be re-opened until Calaal Low

Level Isolation Signal is cleared. If the valves would have been closed due to a

CLS, then they could be opened after five minutes without resetting CLS.

B. Incorrect because, as states above, the valves will close.

062 Loss of Nuclear Svc Water

AAl .06: Ability to operate and / or monitor the following as they apply lo the Loss of

Nuclear Sewice Water (SWS): Control of flow rates to components cooled by the

SWS.

Surry Nuclear Plant 2004-381

DRAFT SRO Inital Exam

48. 063A4.01 00112f

~

1IEREAKERSICIA

-

~

2 8 / 3 . l C N i S R ~ O l ~ Y I l S U R _

- __ - __

Unit 1 was operating at 68% power when the following plant conditions developed:

- I K-A7, BATB SYSTEM I A TROUBLE, alarm annunciates

- "A" SG PORV Indicating tights are not lit

- MSTV Indicating bights are not lit

- POWV 1455C/1456 Indicating Lights are not lit

- "A", "D", and "H" Breaker Indicating Lights are not lit

- There is no indicated letdown flow

- The Turbine Driven AFW Pump is running

-

~

Which ONE of the following describes the plant conditions assuming no other failures

in addition to the cause of the above conditions?

A.' The reactor will automatically trip. The turbine will automatically trip when the

reactor is manually tripped.

B. The turbine will automatically trip. The reactor will automatically trip due to the

C. The reactor must be manually tripped. The turbine must also be manually tripped.

automatic turbine trip.

3. The reactor will automatically trip. The turbine will not automatically trip and must

be manually tripped.

Surly

ReBe rences :

ND-90.3-LP-6, 125 Vdc Distribution, Rev. 18

Distractor Analysis:

A. Correct because the reactor will automatically trip on loss of voltage to the "A" RTB

UV coil to a loss of the "A" DC Bus (see ND-90.3-LP-6). The turbine will not trip

until the reactor is manually tripped in accordance with E-0.

B. Incorrect because the reactor will automatically trip due to loss of voltage to the "A"

RTB UV coil due to the loss of the "A" dc Bus.

e. lmcorrect because the reactor does not need to be manually tripped to trip the

reactor and the turbine will automatically trip when the reactor is tripped per E-0.

D. lncorrect because the turbine does not need lo be manually tripped. The turbine

will trip when the reactor is manually tripped in E-0 or when the other train of RPS

occurs due to low SG levels.

063 DC Electrical Distribution

A4.01: Ability to manually operate and / or monitor in the control room: Major breakers

and control power fuses.

Surry Nuclear Plant 2004-301

DRAFT SRO M a l Exam

The following plant conditions exist:

~ Bus 1J1 voltage drops to 407 volts and returns to 480 volts seven seconds later and

- Bus 2J1-1 voltage is 441 volts and stable

Which ONE of the following correctly states the source of power for Diesel Generator

1. 3's Air Compressors?

remains stable

A. Bus 1 J1 remained the power supply throughout the seven second voltage drop.

B. Six seconds after the voltage dropped on Bus tJ1, Bus 2J1-1 became the power

supply. Bus 2J1-1 will remain the power supply until manually transferred back to

Bus 1J1

CI Six seconds after the voltage dropped an Bus 4J1, Bus 2J1-1 became the power

supply. Bus 2J1-1 will remain the power supply for 30 minutes with Bus 1J1 greater

than 440 volts, at which time it will automatically return to Bus 1 J1.

5. Six seconds after the voltage dropped on Bus 1 J1, Bus 2J1-1 became the power

supply. Bus 2J1-I will remain the power supply for six seconds with Bus 1 J1

greater than 440 volts, at which time it will automatically return to Bus fJ1.

Surry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

Surry

References:

ND-90.3-LP-1, Emergency Diesel Generator, Rev. 14

P&IB 11448-FE-IAA, Appendix W Evaluation Electrical One Line Diagram Surry Power

P&lB 11448-FE-1 P1,480V One Line Diagram MCC 1 J1 -lA Surry Power Station Unit 1,

Station Unit 1, Rev. 23

Rev. 4

Distractor Ana%ysis:

A. Incorrect because 1 J1 voltage was less than 41 Ov for greater than 6 seconds.

Therefore, 2J1-1 became the power supply after 6 seconds. The ABT will check for

2J1-1 voltage greater than 440v prior to swapping to the alternate power supply.

normal-seeking ABT. Therefore, at the beginning of this sequence, the power

supply would have been 1 J1.

Therefore, 2J1-1 became the power supply after six seconds. The ABB will check

for 2J1-I voltage greater than 44th prior to swapping to the alternate power supply.

When the normal power supply voltage is restored to > 44Ov, a 30 minute time delay

is started. If the vottage remains above 44Qv for 30 minutes, then it transfers back

to the normal power supply (131).

B. Incorrect because this is the alternate power supply and the ABT is a

C. Correct because 1J1 voltage was less than 41 Ov for greater than 6 seconds.

D. Incorrect because of the 30 minute time delay mentioned above.

064 Emergency Diesel Generator

K2.01: Knowledge of bus power supplies to the following: Air Gompessors.

Sur9 Nuclear Plant 2004-301

DRAFT SRO lnital Exam

The following plant conditions exist:

- Unit 2 /s in intermediate shutdown

~ Operators are attempting to warm the RHR system

- An instrument air leak has developed, but the location is yet to be determined

- An Operator reports the sound of compressed air leaking in the area of the RHR

pump platform.

- 1B-E6, IA LOW HDR PRESSAA COMPR 1 TRBL, has annunciated

- Instrument air pressure is approximately stable at 60 psig

Which ONE of the following correctly explains the potential effect on warming the RHR

system?

A. If the air leak is a rupture upstream of the isolation valve for the air supply Bo

HCV-1758 (RHR Heat Exchanger Outlet Valve), the valve will fail closed. The line

may be crimped if the leak will not affect vital control instruments. Operators shoulc

use the portable air bottle, via quick disconnect, to operate the valve.

B. If the air leak is a rupture upstream of the isolation valve for the air supply to

HCV-I 758 (RHR Heat Exchanger Outlet Vaive), the valve will fail open. The line

may be crimped if the leak will not affect vital control instruments. Operators shoulc

use the portable air bottle, via quick disconnect, to operate the valve.

CI' If the air Beak is a rupture upstream of the isolation valve for the air supply to

HCV-1642 (CVCS Flow Regulator Control Valve), the valve will fail closed. The line

may be crimped if the leak will not affect vital control instruments.

D. If the air leak is a rupture upstream of the isolation valve for the air supply to

HCV-1142 (CVCS Flow Regulator Control Valve), the valve will fail open. The line

may be crimped if the leak will not affect vital control instruments.

Surry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

Susry

Ref e rences :

ND-88.2-LP-1, Residual Heat Removal System Description, Rev. 8 (Pages 9, 10, 11)

NB-88.2-LP-2, Operation of Residual Heat Removal System, Rev 15

P&ID 11448-FM-087A Sh 2 of 2, Residual Heat Removal System, Rev. 26

P&IB 11 448-FM-075E Sh 1 of 2, Compressed Air System, Rev. 43

1B-E6, IA LOW HDR PRESS/IA COMPR 1 PRBL, Rev. 9

Distractor Analysis:

A. Incorrect because HCV-1758 fails open and cannot be operated with a portable air

bottle. Plausible because the applicant may get consfused on which valve in this

flowpath has the portable air bottle feature.

B. Incorrect because HCV-1758 cannot be operated with a portable air bottle.

Plausible because the applicant may get consfused OR which valve in this

flowpath has the portable air bottle feature.

C. Correct because HCV-1142 is fail closed and this is the flow path for system

warmup. ARP states that leaks may be stopped via crimping if the leak will not

affect vital instrumentation.

D. Incorrect because HCV-1142 fails closed. Plausible because the applicant may get

confused on failure modes of HCV-1142,

especially since it does have a backup air

bottle feature for App. R purposes.

065 Loss of Instrument Air

AA2.01: Ability to determine and interpret the following as they apply to the loss of

instrument air: Cause and effect of low pressure instrument air alarm.

Surry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

51. 069G2.4.18 0 0 1 / 1 / 2 & ~ KWST CChIEM

~.

2 7/3 6/B/SR04301/IP/MAB/SDR

~

r

In FCA-8.00, Limiting Auxiliary Building Fire, if Charging Pump CC Pumps are not

running, the operator is directed to shift charging pump suction to the WWST. Which

ONE of the following describes the basis for this step?

A. Suction is shifted to the WWST to maximize boron injection before the charging

pumps overheat and are lost due to a time-overcurrent breaker trip.

B. Suction is shifted to the RWST to maximize boron injection before the charging

pumps overheat and are lost due to a instantanesus-overcur~en~

breaker trip.

6.

The loss of Charging Pump CC will eventually result in a loss of VCT level due to a

loss of makeup; therefore suction is shifted to the RWST.

D? The RWST supplies cooler water to the Charging Pumps; thereby minimizing the

cooling requirements for the Charging Pumps.

~

~

.~

~~~

~~~

Surry (Utility needs to verify technical acuracy and supply additional supporting material

if any is availble.)

Refernces:

ND-95.6-LP-3, Safe Shutdown Fire FCAs, Rev. 5

0-FCA-8.00, Limiting Auxiliary Building Fire, Rev. 13

Distractor Analysis:

A. Incorrect because the concern is with overheating the pump, not maximizing boron

injection prior to the pump overheating. Supplying cooler RWST water will reduce

the pump temperatures.

B. Incorrect because the concern is with overheating the pump, not maximizing boron

injection prior to the pump overheating. Supplying cooler RWST water will reduce

the pump temperatures.

C. lncorrect because VCT level will not be reduced as a result of no E.

B. Correct because cooler RWST water will help reduce pump temps when CC is lost.

069 Plant Fire On-Site

G2.4.18: Knowledge of specific bases for EOPs.

Surry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

52. 068134 01 001/212/RADIh?'ION MONITOR/C/A 3.4/4.I/R/SR093C)l~ABISDR

~

~

_

_

-__

~

_

_

_

_

- _ _

The following Conditions exist:

- Both Units are at 180% P0wer

- Unit I Operators have discovered indication of a small tube leak in the "A" Steam

Generator for their Unit

- Spent Fuel is being moved in the Spent Fuel Storage Pool to facilitate rack

inspections

- O-RM-M4, 1 -VG-Wl-104 HIGH, alarms

- All Radiation Monitors appear to be operating satisfactorily

- Ventilation and Radiation Monitors are in their normal alignment

Which ONE of the following could cause RM-VG-104 (#I Vent Stack RM) to detect

higher than normal activity?

I

A. A Steam Generator Tube Leak on Unit 1.

BY A spill of high activity coolant in the Chemistry Hot Lab.

C. A spill of high activity coolant in the High Rad Sample System Room.

B. A dropped fuel assembly in the fuel building.

~

~~~

-~

-

-

~

_ _ _ _ _

~

_

_

_

~

Surry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

Sur9

References:

O-RM-M4, I -VG-Rl-I 04 HIGH? Rev. 2

0-AP-22.00, Fuel Handling Abnormat Conditions, Rev. 18

ND-95.3-LP-1, $re-TMI Radiation Monitoring System, Rev. 8

Distractor Analysis:

A. Incorrect because the normal configuration for the ventilation system would not

have Main Condenser Air Ejector aligned to discharge to the Number 1 Vent Stack

upstream of Radiation Monitor 1 -VG-RM-I 04.

B. Correct because O-RM-ld14 alarming could be caused by a coolant spill in the Chem

Hot Lab according to the ARP.

C. Incorrect a spill in the High Radiation Sample System Room would not cause this

alarm according to the ARP.

D. Incorrect because fuel clad damage would not be detected by RM-VG-104 when in

its normal configuration. 0-AP-22.80 does not list WM-VG-104 as a potential means

of indication for damaged fuel clad.

068 Liquid Wadwaste

K4.01: Knowledge of design feature@) and / or interlock(sj which provide for the

following: Safety and environmental precautions for handling hot, acidic, and

radioactive liquids.

Sur9 Wequal Exam Bank Question #462 (ID:ARP0076)

Suny Nuclear Plant 2064-301

DRAFT SRO lnital Exam

53.

~ 07 1K4.06 OOlI2I2iWASTE

~

GAS GASEOUSiMEM

~

2.7/3.5:N/SRO4301/R/MA13/SDR

~

-_

___

~

A discharge of a waste gas decay tank is in progress when RM-GW-101 reaches the

high alarm setpoint and alarm O-RM-M3, 1 -GW-RI-I 01 HIGH, annunciates. Which

ONE of the following is NOT an automatic action initiated by the high radiation fevets

from the waste gas decay tank release?

r

t

A. 1-GW-FCV-101, Decay Tank Bleed Isolation Valve, closes.

B. 1 -GW-FCV-I 60,

CTMT Vacuum Pump Discharge Isolation Valve closes.

C. 1 -GW-FCV-260, CTMP Vacuum Pump Discharge Isolation Valve, closes.

D l Associated vacuum pumps trip.

Surry (Utility needs to verify technical accuracy)

References:

ND-92.4-LP-1, Gaseous and Liquid Waste Processing Systems, Rev. 8

ND-93.5-LP-1, Pre-TMI Radiation Monitoring System, Rev. 8

0-RM-KS, 1-GW-81-181 HIGH, Rev. 0

Distractor Analysis:

A. Incorrect because according to ARP, this valve will close on reaching the high alarm

B. Incorrect because according to AWP, this valve will close QIB reaching the high alarm

C. Incorrect because acording to ARP, this valve will close on reaching the high alarm

B. Correct because the pumps must be manually secured if GW-160 or GW-260 are

setpoint.

setpoint.

setpoint.

closed. This info is in a CAUTION in the AWP and a step is provided in the ARP to

secure the pumps following the closure of GW-I 60 / 260.

071 Gaseous and Liquid Waste Processing Systems

K4.86: Knowledge of design(s) features and / 01 interlocks which provide for the

following: Sampling and monitoring of waste gas release tanks.

-

Surry Nuclear Plant 2004-381

DRAFT SRO lnital Exam

Which ONE of the following is sufficient conclusive indication to warrant a correct entry

into AP-16.08, Excessive RCS Leakage?

A. Rising containment humidity, rising containment temperature, and rising

containment pressure.

B." Rising steam generator water level, rising charging flow, and rising Condenser Air

Ejector Radiation Monitor reading.

!

C. Rising Condenser Air Ejector Radiation Monitor reading, rising steam generator

blowdown radiation monitor reading, and stable containment pressure.

D. Rising containment sump level, lowering pressurizer pressure, and rising

containment pressure.

Surry

References:

1 -A$-1 6.00, Excessive RCS Leakage, Rev. 11

1 -A$-24.0(4, Minor SG Tube Leak, Rev. 8

1 - E O , Reactor Trip or Safety Injection, Rev. 46

ND-93.5-LP-1, Pre-TMI Radiation Monitoring System, Rev. 8

Distractor Analysis:

A. Incorrect because a steam line break can cause containment humidity, temperature,

and pressure to rise. Distractor is plausible because these are all possible for an

RCS leak.

steam/feed mismatch. The charging flow and Air Ejector Wad monitor corroborates

that the problem is tube leakage.

C. Incorrect because these parameters are indications that there may be a tube

leak; however, these same indications may present themselves with a fuel failure or

crud burst. Distractor is plausible because these parameters may indicate as

stated if RCS leakage actually exists. Distractor is incorrect because these

parameter trends may be caused by increased RCS activity.

D. Incorrect because the Combination of these parameters may be caused by a

steam leaidbreak. Distractor is plausible because these parameters may actually

change as indicated during an 86% leak.

B. Correct because SG water level is indication that there may be a tube leak or

073 Process Radiation Monitoring

62.1 23: Ability to perform specific system and integrated plant procedures during all

modes of plant operation.

Surry Nuclear Plant 2004-301

5 R A R SRO Inital Exam

Hydrogen peroxide has just been added to Unit 2 RCS resulting in an increase in the

primary coolant activity. The first indication that the activity level has increased will be

seen on the

and the team should

A. Containment particulate radiation monitor; increase flow through the letdown cation

bed.

BY Letdown radiation monitor; monitor letdown filter differential pressure.

C. Letdown radiation monitor; monitor seal return filter differential pressure.

5. Containment particulate radiation monitor; decrease flow through the letdown cation

bed.

Surry

Ref@ re nces :

ND-93.05-LP-1, Pre-TMI Radiation Monitoring System

ND-88.3-LP-3, Seal Injection, Rev. 6

Distractor Analysis:

A. Incorrect because containment particulate radiation monitor would not change

significantly.

B. Correct because letdown radiation monitors would indicate quickly due to hydrogen

peroxide increasing reactor coolant activity and letdown filter dP would also rise.

C. Incorrect because the hydrogen peroxide should not affect the seal return dP, at

least not as readily or as SOOR as the letdown filter dP. There is 8 gal of CVCS

water that goes to each RCP for seal injection. Five ot these gallons flows down

the shaft past the thermal barrier and ends up in the WCS. The other three gallons

eventually passes through the seal return filter. The CVCS water that enters the

RCP seal area has already been filtered prior to getting to the RCP seals. This

prefiltesing is designed to protect the seals. The water corning from the

RCP seal area should be relatively clean CVCS water. not RCS water; therefore

making the seal return filter a relatively poor indicator of a crud burst.

significantly.

D. Incorrect because containment particulate radiation monitor Would not change

Surry Bank 1L.T Exam Question #I 606

076 High Reactor Coolant

AK2.01: Knowledge of the interrelations between the High Reactor Coolant Activity

and the following: Process radiation monitors.

Sur9 Nuclear Plant 2004-301

DRAFT SRO lnital Exam

56. 046K2.04 001/211/SERVICE

~

WATERICI.4 2 5/2.6/NlSR04301/R/1C1.413/SDK

~

.~

___

~

The following Unit 1 conditions exist:

- A Large Break LOCA occurred 45 minutes ags

- Recirculation Spray is operating

- MCC 11-11 -2 de-energizes

Which ONE of the following correctly describes the impact on Sewice Water to and

from the Wecirc Spray Heat Exchangers?

r-

A! Recirc Spray Heat Exchanger I-RS-E-1A Sewice Water Inlet (MQV-SW-104A) t

Outlet (MOV-SW-105641) Valves de-energize.

B. Wecirc Spray Heat Exchanger 1 -RS-E-l B Service Water Inlet (MOV-SW-184B) i

Outlet (MOV-SW-1058) Valves de-energize.

C. Recirc Spray Heat Exchanger Service Water lnlet (MOV-SW-103.4) and Recirc

Spray Heat Exchanger 1 -RS-E-I A Sewice Water lnlet (MOV-SW-104A) Valves

de-energize.

D. Recirc Spray Heat Exchanger Service Water Inlet (MOV-SW-1033) and Recirc

Spray Heat Exchanger 1 -RS--I B Service Water Inlet (MOV-SW-IQ4B) Valves

de-energize.

S U P 9

References:

ND-91 -LP-6, Recirculation Spray System, Rev. 9

ND-89.5-LP-2, Service Water System, Rev. 20

P&ID 1 1448-FE-1 M, Sh 1 of 1, 48OV Qne Line Diagram Surry Power Station - Unit 1

$&ID 1 1448-FE-1 L, Sk 1 of 1 I 480V One Line Diagram Surry Power Station - Unit 1,

Rev. 59

Rev. 52

Bistractor Analysis:

A. Correct because MBV-SW-104A and 1Q5A are both powered from 1 HI-2.

3. Incorrect because MOV-SW-IQ4B and 185B are both powered from 1J1-2.

C. Incorrect because MOV-SW-103A is powered from 1 H1-1.

5. Encorrect because MOV-SW-103B is powered from 1J1-1 and MOV-SW-I045 is

powered from 1J1-2.

676 Service Water

K2.04: Knowledge of bus power supplies to the following: Reactor building closed

coo I i n g water .

Surry Nuclear Plant 2004-301

DRAFT SRO inital Exam

57.

__

075.44.01

__ 001/2/1/INSTRUhIENT

ALIP/C/A 3 1/3.1/B/SW04301RMBB/SDR

~

~.

- -

~

__

- --

Unit 1 is at 50% power and the team is experiencing problems controlling feedwater

flow. An Instrument Air Low Pressure Alarm is received in the Control Room. White

monitoring Instrument Air pressure, the 80 notes pressure is 50 psig and slowly

Iswering.

Which ONE of the following actions should be taken?

A. Commence a SIOW

power reduction to Hot Shutdown.

3. Commence a fast power reduction to Cold Shutdown.

Ca' Trip the Reactor and go to 1 -E-O, Reactor Trip or Safety Injection.

B. Isolate Sewice Air frsm instrument Air and start the Scrllair Diesel.

References :

ND-92.1-LQ-1, Station Air Systems, Rev. 13

f B-E6,

IA LOW HBR PRESS / IA COMPR 1 TRBL, Rev. 9

0-AP-40.00, Non-recoverable Loss of instrument Air, Rev. 17

Distractor Analysis:

A. Incorrect because 1B-E6 and AP-40.00 directs rx trip, not power reduction.

3. Incorrect because 1 B-E6 and AP-40.00 directs rx trip, mot power reduction. (Initial

distractor from exam bank was changed because it may have been a second

correct answer).

C. Correct because this is the guidance provided by 1 B-E6 and AQ-40.00.

B. Incorrect because 1 B E 6 and AP-40.00 directs rx trip, not power reduction when

pressure reaches 50 psig.

078 Instrument Air

A4.01 : Ability to manually operate and / 08 monitor in the control room: Pressure

gauges.

Surry Requal Exam Bank Question 428

Sw-y Nuclear Plant 2004-301

DRAFT SRO lnital Exam

I

A. Instrument Air is normally supplied by the Service Air System and the system is

backed up by IA when IA pressure reaches 95 psig.

B. Instrument Air is normally supplied by IA Compressors and the system is manually

backed up by the Sullair Diesel.

With ALL air systems aligned in the automatic mode, which ONE of the following

describes the operation of the Station instrument Air (IA) System for Unit I ?

(Assume no operator action is taken.)

CY Instrument Air is normally supplied by the Service Air System and is backed up by

the IA System when IA pressure reaches 90 psig.

D. Instrument Air is normally supplied by the Service Air System and is backed up by

the Condensate Polishing Instrument Air System when IA pressure reaches 98

psig.

References

ND-92.1-LP-I , Station Air Systems, Rev. 13

Distraetsr Analysis:

A. Incorrect because pressure must drop below 98 psig for IA to backup Service Air.

B. Incorrect because the IA System is normally supplied by Sewice Air.

C. Correct because Service Air is the normal supply and IA is the backup when

pressure drops below 90 psig.

B. Incorrect because IA is not backed up by the Condensate Polishing Instrument Air

System when pressure drops to 98 psig. It is backed up by the IA System when

pressure drops below 90 psig.

078 Instrument Air

K4.02: Knowledge of the IAS design feature(s) and or interlock(s) which provide for the

following: Cross-over to other air systems.

Surry Requal Bank Question #512

Surry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

I

i

The following Unit 1 conditions exists:

- A steam line rupture in Containment occurred several minutes ago

~ Maximum Containment Pressure reached 24 psia

- Containment Pressure Transmitters now read:

~

PT-LM-I OOA = 17.7 psia

- PT-LM-10OB = 17.8 psia

- PT-LM-1 OOD = 17.9 psia

- PT-LM-1006 = 17.6 psis

Which ONE of the following correctly describes resetting of Consequence Limiting

Safeguards (CLS) given the above conditions?

A. The CLS TWAlN A/B) RESET PERMISSIVE annunciator is lit. CLS HI and CLS

HI-HI may be reset at this time. Upon reset, the multiplying relays will energize.

B! Neither CLS Hl or CLS HI-HI may be reset at this time. The multiplying relays are

de-energized.

C. The CLS Hl-HI RESET PERMISSIVE annunciator is lit. CLS HI-HI may be reset at

this time. Upon reset the multiplying relays will de-energize.

D. Neither CLS HI or CLS Hl-HI may be reset at this time. The multiplying relays are

energized.

Surry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

Surry

References:

NB-88.4-LP-2, Containment Vessel, Rev. 8

ND-91 -LP-5, Containment Spray System, Rev. 13

Distractor Analysis:

A. Incorrect because pressure must be reduced to less than 14.2 psia on 2/4 channels

B. Correct because pressure must be reduced to less than 14.2 psia on 2/4 channels

to reset both Hi and Hi-Hi subsystems.

to reset both Hi and Hi-Hi subsystems. Also, when CLS is actuated, the multiplying

relays are de-energized.

C. Incorrect because pressure must be reduced to less than 14.2 psia on 2/4 channels

to reset both Hi and Hi-Hi subsystems. Also, when CLS is actuated, the multiplying

relays are de-energized.

B. Incorrect because when CLS is actuated, the multiplying relays art? de-energized.

103 Containment

A4.04: Ability to manually operate and / or monitor in the control room: Phase A and

Phase B resets.

Surry Nuclear Plant 2004-381

DRAFT SRO lnital Exam

68. (32.1 11 001/3NTbCH SPECSICIA -~

3 0/3.8/N/SR0~301~?UM~/SDK

,

The following Unit 1 conditions exist:

- Plant is at 74% power after just completing a rapid power reduction due to Heater

- Axial Flux Difference was outside of the Parget Band on 11/03/2803 from 0800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br />

- Axial Flux Difference was outside of the Target Band on 11/04/2003 from 0940 hours0.0109 days <br />0.261 hours <br />0.00155 weeks <br />3.5767e-4 months <br />

to 8840 hours0.102 days <br />2.456 hours <br />0.0146 weeks <br />0.00336 months <br />

- The Axial Flux Difference has remained within the Technical Specification Limits sf

Figure 3.12-3, Axial Flux Difference Limits As A Function Of Rated Power: for the

entire time

Brain Pump problems

to 0845 hours0.00978 days <br />0.235 hours <br />0.0014 weeks <br />3.215225e-4 months <br />

Which ONE of the following actions are required by Technical Specifications?

W Reactor power was required to be less than 58% by 0825 hours0.00955 days <br />0.229 hours <br />0.00136 weeks <br />3.139125e-4 months <br /> on

B. Reactor power was required to be less than 58% by 8855 hours0.102 days <br />2.46 hours <br />0.0146 weeks <br />0.00337 months <br /> on

1 /04/2003.

1 /04/2008.

C. Reactor power was required to be less than 58% by 091 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> on 1 1/04/2003

D. No power reduction was required, but power should not have been raised above

75% until Axial Flux Difference was within the Target Band.

Surry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

Reference:

Technical Specification 3.1 2.B.4.b.(1), Amendment No. 186

Technical Specification 3.12.5.4.b.(2), Amendment No. 186

D ist racto I

Analysis:

A. Correct because AFD may deviate from its target band for one hour within a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

period. When this is violated, then power must be reduced to less than 50% within

30 minutes. From 11/03 8 (4880 hrs to 11/04 Q 0755 hrs a total of ome hour

outside of target band was accumulated. Therefore, by 0825 hrs (30 minutes later)

power must be less than 50%.

B. Incorrect because because the correct answer is as described in above analysis.

Plausible because 0855 hours0.0099 days <br />0.238 hours <br />0.00141 weeks <br />3.253275e-4 months <br /> is 60 minutes after 0755 hrs, which is when the 30

minute clock starts to have power less than 50%.

C. Incorrect because the correct answer is as described in above analysis. Plausible

because 091 0 hrs is 30 minutes after 0840 hrs, which was given as the second time

frame where AFD was outside of its target band.

D. incorrect because because the correct answer is as described in above analysis.

Plausible because candidate may confuse 50% and 75% power restrictions.

G2.1.11

Knowledge of less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> technical specification action statements for systems.

Surry Nuclear Plant 2004-301

DRAFT S R 6 Bnital Exam

The following conditions exist:

- Unit 1 has been shutdown for 10 days for SG tube plugging

- RCS water level is being maintained at 12.4 feet as indicated on 1 -RC-hl-? 00A

- The "B" and "C" loops are isolated with the primary and secondary SG manways

removed for SG tube plugging

The reactor vessel head is tensioned

= The "A" RWW pump is in operation with oscilrating amperage indications

= Flow indication 1 -RH-FI-I 605 is oscillating between 2500 and 2760 gpm.

Which ONE of the following actions is appropriate for the SRO to direct in accordance

with AP-27.08, boss of Decay Heat Removal Capability?

(A$-29.00 Attachments 1 and 2 provided)

A. Raise RCS level to 12.5 feet as indicated on 1 -RC-Ll-l OOA and stabilize flow at

2600 gprn.

B. Throttle open 1 -RH-HCV-175% and throttle close 1 -RH-FCV-I 605 to reduce RHR

flow to 2200 gprn.

C. Throttle close 1 -RH-HCV-I 758 and throttk open I -RH-FCV-I 605 to reduce WHR

flow to 1200 gpm.

DY Throttle close 1-RH-FCV-1605 to reduce WHR Blow to 2200 gpm and raise level to

12.5 feet as indicated un 7 -RC-LI-1 OOA.

Sur9 Nuclear Plant 2084-301

DRAFT SRO lnitai Exam

sur9

References:

1 -AP-27.00, Loss of Decay Heal Removal Capability, Rev. 10

ND-88.2-LP-1

I Residual Heat Removal System Description, Rev. 8

ND-88.2-LP-02, Operation of Residual Heat Removal System, Rev. 15

NB-95.2-LP-12, Loss of RHW Events, Rev. 9

Distractor Analysis:

A. Incorrect because AP-27 Att. 2 indicates that 12.5 Beet is in the unacceptable region

of operation for 2600 gpm RHR flow rate.

B. Incorrect because AP-27 Att. 2 indicates that 2200 gpm RHR flow rate is in the

unacceptable region of operation for 12.4 feet.

C. incorrect because AB-27 Att. 1 indicates that 12W gpm WHW Row rate is less than

the required Blow rate of 2200 gpm.

D. Correct because these actions place the plant in an acceptable region of AP-27 AH.

1 and 2 for required flow rate for 10 days after shutdown.

AP-27 Att. 1 and 2 will need to be provided to the applicant.

G2.1.25: Ability to obtain and interpret station reference materials such as graphs,

monographs, and tables which contain performance data.

Surly Nuclear Plant 2884-301

DRAFT SRO hital Exam

Which ONE of the following is correct with respect to Technical Specifications?

A. The Safety Limit for core thermal power is 109% sf Rated Thermal Power and tht

RCS pressure limit is 2735 pig.

3. The Safety Limit for core thermal power is 109% of Rated Thermal Power and the

single loop loss of flow reactor trip shall be unblocked when power range nuclear

flux is greater than or equal to 50% of Rated Thermal Power.

C. The reactor trip on low pressurizer pressure, high pressurizer level, turbine trip, ai

low reactor coolant flow for two or more loops shall be unblocked when power is

greater than or equal to 10% or Rated Thermal Power.

D. The source range high flux, high setpoint trip shall be unblocked when the

intermediate range nuclear flux is less than or equal to ~ X I U ~

amperes.

Surly

References:

Technical Specification 2.1 (Amendments 11 6); 2.2 (Amendments 203); 2.3

(Amendments 175, 176,206)

Distractor Analysis:

A. Incorrect because the safety limit for core thermal power is I t %%.

B. Incorrect because the safety limit tor core thermal power is 11 %%.

C. Correct because this is the correct statement taken from Tech Specs.

8. Incorrect because source range high flux, high setpoint trip shall be unblocked when

the intermediate range nuclear flux is less than or equal to 5x10- amperes.

Generic K/A 2.2.22

Knowledge of limiting conditions for operations and safety limits.

Surry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

63.

62.2 -

27 001/3/~F;trELINC;lMEM

-~

2.6/3 5/NISR04301iWMARISUK

~

___

~

- ~-

1 - Which O N E f the following c o L l y states the level of authorization needed for

bypassing the Manipulator Crane Overload Interlock?

A. Refueling SRO or Fuel Handling Supervisor

B. Refueling SRQ and Shift Supervisor

CY SNSOC and Refueling S R 8

5. SNSBConly

k f e leRCBS:

VPAP-1401, Conduct of Operations, Rev. 11 (Section 6.5)

Distractor Analysis:

A. Incorrect because SNSOC pre-approval is needed per l-OP-Ft-i-Od5 Step 4.62.

B. incorrect because SNSOC pre-approval is needed per 1-OP-FH-015 Step 4.12.

C. Correct because SRO approval is needed per 1 -OQ-FH-015 Step 4.1 0 AND

El. Incorrect because SWO approval is needed per 1 -OQ-FH-015 Step 4.1 0.

SNSOC pre-approval is needed per 1-QP-FH-Oi5 Step 4.12.

G2.2.27

Knowledge of the refueling process.

Surrgr Nuclear Plant 2004-301

DRAFT SRO lnital Exam

guard against personnel exposure.

The following conditions exists:

- Unit 2 is at full power

~ Unit 1 is in refueling

~

Fuel repair is being performed

- A damaged fuel rod is raised loo close to the surface of the water

- Area radiation monitors alarm in the vicinity of the fuel movements

- Operators enter Q-AP-22.00, Fuel Handling Abnormal Conditions

All components operate as designed

Which ONE of the following are immediate actions of AQ-22.00?

A! Stop fuel handling operations, Secure Normal MCR Ventilation by closing

1-VS-MOD-103C and 1 -VS-MOD-l83D, Dump Cable Vault Air Bottles by closing

1 -VS-MOD-l03B.

E. Stop fuel handling operations, Secure Normal MCR Ventilation by closing

l-VS-MBD-103C and I-VS-MOD-l03D, Bump MER 3 Air Bottles by closing

1 -VS-MBD-I Q3A.

C. Evacuate the affected areas, Secure Normal MCR Ventilation by closing

1 -VS-MOD-1 Q3C and 1 -VS-MQB-I 03D, Dump MEW 3 Air Bottles by closing

1 -VS-MOD-l038.

D. Stop fuel handling operations, Evacuate the affected areas, Stop Main Control

-~

R O O ~

Fans

~-

1 -VS-F-Xand -~

1 -VS-AC-4.-

-

-

~

-

-

~

SLsrry

References:

0-AP-22.00, Fuel Handling Abnormal Conditions, Rev. 18

Distractor Analysis:

A. Correct these are all listed as immediate actions of AQ-22.00.

B. Incorrect because I -VS-MOD1 Q3A is in the RNO column to be performed if 103B

does not CIOS~.

However, the stem states that all equipment operates as designed,

so the operator would not go to the RNO column.

does not close.

e. Incorrect because 1 -VS-MOB-I 0314 is in the RNO column to be performed if 1038

S. Incorrect because stopping MCR Ventilation Fans is not an immediate action.

(32.3.1 0: Ability to perform procedures to reduce excessive levels of radiation and

Surry Nuclear Plant 2004-301

DRAFT SWO inital Exam

65. 6 2 3 2 001/3//RADIATIe)N RESPIR TORfC/A 2

~

_

_

I

Work in a radiation area must be performed. The following conditions exist:

~ A point source is present and emits 50 rnrem/hour at 1 foot

- The air has a Derived Air Concentration (DAC) of 10

Which ONE of the following methods will result in the lowest amount of awumulated

dose?

A. Two workers using hand tools can perform the work in one hour at a distance of two

feet wearing no respirator.

5. Three workers using remote tools perform the work in two hours at a distance of six

feet wearing no respirator.

I

C. Two workers using hand tools perform the work in four hours at a distance of two

Beet wearing a respirator with a protection factor of 50.

I

D I Three workers using remote tools perform the work in 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> at a distance of six

feet wearing a respirator with a protection factor of 50.

References:

Dominion Nuclear Employee Training Manual Volume Ii BWWT, RPT, CSET, SCAT,

FWT, Rev. 11, January, 2003.

Distractor Analysis:

A. Incorrect: 75 mrem z 56.7 rnrem. {[(2 men)(l hr)(58mrem/hr)(d/2)2J+[(10 DAG)

5. Incorrect: 158.3 mrem 1 56.7 rnrem. {[(3 men)(2 kr)(50rrtrem/hr)(1/6)2]+[(10 DAC)

C. incorrect: 104 mrem 1 56.7 rnrern. {[(2 menl(4 hr)(50mrern/hr)(l/6)*]+[(le) DAC)

B. Correct: [(3 men)(lO hrs)(50 rnrem/hr)(1/6)*] + [(IO BAC)(I/50)(3 menj(l0 hrs)(2.5

(2 men)(l hrl(2.5 mremlDAC-HR)] = 75 rnrern}

(3 rnen)(2 hrI(2.5 mrem/DAC-HRj] = 158.3 rnrem)

(1/50)(2 rnen)(4 hr)(2.5 mrem/BAC-HW)] = 104 rnrem}

mrern/l BAC-HR)] = 41.7 + 15 = 56.7 mrem.

G2.3.2

Knowledge of facility ALARA program.

Surry Nuclear Plant 2804-301

DRAFT SRO lnital Exam

66. 62.3.9 00 1 / 3 N c ' O ~ " H . ~ ~ ~ , ~ ~ ~ ~ ~ ~ ~ / A

2&33.4/N/SKO43Ol /R/MAR/SDR

1

- -

~-

- - -

The following Unit 1 conditions exist:

- The WCS temperature is 190 O F .

- Operators are performing Section 5.2 of 1 -8P-VS-001, Containment Ventilation, to

place the Containment Purge System in service using 1 -VS-F-58A or 1 -VS-F-58B,

Filter Exhaust Fans.

- The Containment Purge Form requires 5080 cfm purge flow.

Which ONE of the following correctly states selection criteria, in accordance with

1-OP-VS-001, for choosing which valve to use for obtaining the correct purge flow

rate?

A!

1 -VS-MBV-I OOD (Ctrnt Purge Exh) should be throttled instead of 1 -VS-MOV-1 01

(Ctmt Purge B/P) due to the high flow rate required by the Containment Purge

Form.

B. 1 -VS-MOV-l 01 (Ctmt Purge B/P) should be throttled instead of 1 -VS-MOV-I 80D

(Ctmt Purge Exh). This is due to the need to open the supply breaker to

1 -VS-MOV-l OOD in order to throttle it. Opening the breaker will prevent automatic

CTMT Purge isolation.

C. 1 -VS-MOV-I 01 (Ctrnt Purge B/P) should be throttled instead of 1-VS-MOV-I OOD

(Ctmt Purge Exh) due to the low flow rate required by the Containment Purge

Form.

B. I-VS-MOV-IOOB (Ctrnt Purge Exh) should be throttled instead of I-VS-MOV-181

(Ctrnt Purge BP). This is due to the need to open the supply breaker to

1 -VS-MOV-I 01 in order to throttle it. Opening the breaker will prevent automatic

CTMT Purge isolation.

i

- - - - - - -

Surry Nuclear Plant 2004-301

DRAFT SRQ lnital Exam

Surry

References:

1 -0P-VS-801, Containment Ventilation, Rev. 20

Distractor Anaiysis:

A. Correct because 1008 should be throttled due to the Containment Purge Form

allowing more than 3000 cfm. The bypass will not have enough capacity at this

flow rate.

B. Incorrect because even though auto containment purge isolation will not occur with

the breaker open, the procedure still directs the use of 180D due to the high flow

rate. Plausible because applicant may think it logical to not intentionally

incapacitate auto containment isolation.

6. !ncorrect because with the Blow rate greater than 3000 gpm, 1 00D should be used.

Plausible because 3000 gpm is not a very high flow rate.

D. Incorrect because the bkr does not need to be opened and at 5000 gpm, the

procedure directs 101 to be used for fine tuning the flow rate. Plausible because

preventing auto ctmt purge isolation is a concern when using 100D.

G2.3.9: Knowledge of the process for performing a containment purge.

Surry Nuclear Plant 2884-301

DRAFT SRO lnital Exam

B. Immediate action steps may be performed in any order, except for the first four

immediate action steps of E-0, Reactor Trip or Safety Injection. which must be

performed in the order in which they appear in the procedure.

CY Immediate action steps may be performed in any order except for the first four

immediate action steps of E-Q, Reactor Trip or Safety Injection, and the immediate

action steps of FR-S.1, Response to Nuclear Generation / ATWS, which must be

performed in the order in which they appear in the procedure.

D. immediate action steps may be performed in any order except for the immediate

~

References:

ND-95.3-LP-2, Emergency Procedure Writer's Format, Rev. 8

(Have Utility add any addional references that may support answer.)

Distractor Analysis:

A. Incorrect because only immediate actions of E-0 and FR-§.I must be performed in

B. lncurrect because only immediate actions of E-8 and FR-S.1 must be performed in

C. Correct because immediate actions of E-0 and FR-S. 1 must be perfurmed in the

the order in which they appear in the procedure.

the order in which they appear in the procedure.

order in which they appear in the procedure. This requirement / expectation is

stated in ND-95.3-LP-2 Page 12.

D. Incorrect because ECA-0.0 are not required to be performed in any specific order.

G2.4.11: Knowledge of abnormal condition procedures.

Surry Nuclear Plant 2004-301

DRAFT SRO M a l Exam

A situation presents itself that requires a Reactor Operator (BO)

to take quick decisive

action to ensure Station Safety. Personnel are not in immediate danger and the action

requires no reactivity manipulations.

Which ONE of the following correctly describes the requirements for performing the

actions?

A.' The 80 may take necessary action without prior approval from another licensed

operator.

B. The WO must immediately request approval from the Unit SRQ to perform the

action and only take action a b r approval is granted.

C. The RO may take action only after another licensed operator has been notified and

concurs with the action.

D. The RO may take action only after obtaining a peer check to concur with the action.

Surly

References:

OPAP-0006, Shift Operating Practices, Rev. 4

Bistractor Analysis:

A. Correct because OPAP-0806 Step 6.1 0.3 states, "During emergencies, Shift Team

members may take necessary immediate actions required to ensure personnel and

Station safety without prior approval. The Shift Supervisor shall be promptly

informed of these actions."

B. Incorrect because action may be taken prior to obtaining permission.

6. Incorrect because action may be taken prior to notifying or obtaining permission

D. Incorrect because immediate action is authorized to protect the Station.

from another Team Member.

G2.4.12: Knowledge of general operating crew responsibilities during emergency

QperatiQnS.

Surry Nuclear Plant 2004-301

5RAFT SRO lnital Exam

69. G2.4.49

-~

O O Z i 3 / l K O ~ ~ ~ N T R ~ ~ / A ~ / 4 . ~ ~ ~ R ~ ~ O ~ ~ ~ ~ ~ ~ R

__ - -

~

Given the following conditions:

~

Reactor Power = 85%

- Control Rods are in automatic

- Control Bank D begins to insert without a turbine runback

- Pave and Tref are matched within 0.5 O F

r

-

Which ONE of the following describes the correct immediate operator response to

these conditions?

A. Verify quadrant power tilt and axial flux difference within limits.

B! Place ROD CONT MODE SEL switch in MANUAL.

I

C. Manually trip the reactor.

1

D. Verify lWPl operating properly.

Surry

References:

0-AP-1 .OO, Rod Control System Malfunction, Rev. 9.

Distractor Analysis:

A. Incorrect because the initial response is to place ROD CONP MODE SEL switch in

MANUAL.

B. Correct per AP-1 .QO.

6. lncorrect because this would not be performed until 805 CQNT MODE SEL switch

S. Incorrect because AP-I .00 directs placing ROD CONT MODE SEL switch in

was placed to MANUAL and rod motion had slopped.

MANUAL as an immediate action.

G2.4.49

Ability to perform without reference to procedures those actions that require immediate

operation of system components and controls.

Surry Nuclear Plant 2004-301

DRAFT SRB lnital Exam

90. WEWEK3.2

__

00111iULOCA

~

~ OUTSIDEIC'IA 3.W4.OIMISK0430 1I7UIVIABISDR

I

.__

~~

I

Which ONE of the following correctly states actions contained in I-ECA-1.2, LOCA

Outside Containment, and the reasons for those actions?

A. Open l-SI-MOV-189BA (LHSl to Hot Leg) or l-SI-MOV-1890B (LHSI to Hot Leg) to

provide a flow path for Low Head Safety Injection. Then close 1 -SI-MOV-I 8906

(LHSI to Cold Legs) and monitor RCS pressure.

I

t

B. If closing 1 -SI-MOV-1 890C (LHSI to Cold Legs) does not result in an RCS pressure

rise then allow it to remain closed because this will give operators time to check AUX

Building alarms while the flow path is isolated.

C. If the leak is not identified and isolated then transition to 1451, boss of Reactor or

Secondary Coolant, because RCS inventory is continued to be lost outside of

containment.

j

83: If closing 1-SI-MOV-1890C (LHSI to Cold Legs) results in an RCS pressure rise,

then place the LHSl pumps in PTL because their suction valves from the RWST will

be closed to isolate potential leak paths.

Surry

References:

ND-95.3-LP-21, ECA-I .2 LOCA Outside Containment, Rev. 7

ECA-I 2,

LOCA Outside Containment, Rev. 5

Bistracto r Analysis:

A. incorrect because ECA-I .2 does not give any direction to open I-SI-MOV-189OA &

B. These valves should be left in the closed position. This distractor is plausible

because ECA-1.2 does give guidance to close 189OC.

decreasing, then the leak was not isolated and the valve needs to be re-opened.

This is the normal SI flow path and it is important to re-establish this path if closing

the valve did not isolate the leak.

C. hmrrect because if the leak is not iso(ated, then the correct transition would be to

go to 1 -ECA-l. 1, boss of Emergency Coolant Recirculation.

D. Correct because if WCS pressure rises upon closure of 1 -Sl-MOV-l89OC, then the

leak was isolated and 1 -ECA-1.2 directs the LHSB pumps to be placed it7 PTL and

the suction valves from the RWST to be closed.

B. incorrect because if 1 -SI-MOV-1890@ is closed and RCS pressure is still

w EO4

EK3.2: Knowiedge of the reasons for the following responses as they apply to the

(LOCA Outside Containment): Normal, abnormal, and emergency operating

procedures associated with &OCA Outside Containment).

Surry Nuclear Plant 2004-301

DRAFT SRO Inital Exam

~

~

~

.~

~

71

~ WE06LK3.1 001/1/2/CORE COOLINGblEM 3 W3,8/5/SR04301ITQIM~HISDR

-~

~-

- -~

I-

1-FR-6.1, Response to Inadequate Core Cooling, is being performed. Which ONE of

the following is the reason RCPs are stopped prior to depressurizing the SGs to less

than 150 psig during an inadequate core cooling event?

A. RCP operation with the SGs at atmospheric pressure is prohibited due to excessive

hydraulic stress on the SG kl-tubes.

B. The SGs will depressurize more quickly if no Forced Circulation RCS flow exists.

C. To minimize heat input to the RCS.

D:' The SG depressurization will lead to a loss of RCP suppod conditions.

Sursy

References:

ND-95.3-LP-38, Response to Inadequate Core Cooling, Rev. 8

FR-C. 1

~ Response to Inadequate Core Cooling, Rev. 18

Distractor Analysis:

A. Incorrect because securing RCPs is necessary because the depressurization will

result in losing the RCP seal support conditions, which could damage the RCPs.

B. Incorrect because the basis for securing RCPs is not associated with heat input into

the wcs or forced Blow.

C. Incorrect because the basis for securing RCPs is not associated with heat input into

the 86s.

D. Correct because this is the stated reason in NB-95.3-LP-38. Losing #1 Seal

s~app~rt

conditions could result in damage to the RCPs.

074 Inad. Core Cooling

E06EK3.1: Knowledge of the reasons for the following responses as they apply to

(Degraded Core Cooling): Facility operating characteristics during transient conditions,

including coolant chemistry and the effects of temperature, pressure, and reactivity

changes and operating limitations and reasons for these operating characteristics.

Surry Wequal Exam Bank Question #467

Surry Nuclear Plant 2884-301

DRAFT SWO lnital Exam

The following Unit 1 conditions exist:

- Reactor power is 58% and rising

- RCS pressure is at 221 0 psig and slowly lowering

- Tavg is 557 O F and slowly lowering

- Pressurizer level is slowly lowering

~ Turbine load is stable at 400 MW

- SG levels are at 46% NR

- SG pressures are at 970 psig and slowly lowering

- Containment pressure is 9.5 psia and slowly rising

- Condenser Air Ejector RM reads 113 cpm

Which ONE of the following correctly diagnoses the event?

A. Ruptured and faulted steam line break inside containment.

B:' Steam line break inside containment.

C. LOCA inside containment.

3. Steam line break outside containment.

Surry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

References:

General operator knowledge.

Distractor Analysis:

A. Incorrect because although there are parameters to support the steam line break,

there are no parameters to support a SGTR. Plausible because Condenser Air

Ejector RM reading is given, but the value is not representative of a SGTR.

B. Correct because reactor power and ctmt pressure are rising; RCS pressure, Tavg,

and SG pressures are lowering. These are all indicative of a steam line break

inside ctrnt.

C. Incorrect because reactor power would not be rising during a h0CA as it would

during a steam line break. Plausible because many of the parameters coincide with

a LOGA.

B. Incorrect because ctmt pressure is rising. Plausible because of the aforementioned

parameters that are indicative of a steam line break.

WE08 RCS Overcooling

G2.1.7: Ability to evaluate plant performance and make operational judgements based

on operating characteristics, reactor behavior, and instrument interpretation.

Surry Recgual Bank Question #I 77 (ID: EOP6076)

Surry Nuclear Plant 2004-301

DRAFT SRO Cnital Exam

73.

__

WE1 lEhl

~

2 001/1/2LWSI

~

LOCA

~ RWST' IIISVC/A 3,5/3mSK1)430I/R/MAB/SDR

I

The following conditions exist:

- LOCA has occurred.

- RWST level = 13% and decreasing.

- Recirculation Mode Transfer (RMT) keyswitch is in WMT Mode.

- White RMT Status Light is lit.

- Amber RMT Status Light is lit.

1 -SI-MOV-l868,4 (LHSI Suction from Sump) opens fully and 1 -SI-MOV-I 8608 (LHS

Suction from Sump) strokes to 50% open where it trips on thermal overload. Which

ONE of the following gives the correct status of Safety Injection?

A." LHSl from the RWST is injecting into the cold legs and HHSl from LHSl pump

discharge is injecting into the cold legs.

B. No Safety Injection is injecting water to the cold legs.

C. HHSl directly from the RWST (not from LHSl discharge) is injecting into the cold

legs, but no LHSl is injecting into the cold legs.

D. hHSl from the RWST and HHSl directly from the RWST (not from LHSl dischargi

is being injected into the cold legs.

Sur9

Wefe re nces:

ND-91.3-LP-3, Safety Injection System Operations, Rev. 15

1 -ES-I 3,

Transfer to Cold Leg Recirculation, Rev. 1 1

Distractor Analysis:

A. Correct because 1 -SI-MBV-l862A&B will not close until 1 -Sl-MOV-l860A&B open

due to an interlock.

B. Incorrect because RWST is still the suction source to the LHSl pumps.

6. Incorrect because LHSl Pumps are taking suction from the RWST and injecting into

the cold legs and HHSl is not taking suction directly from the WWSP.

3 ~ Incorrect because HHSl is not taking suction directly from the RWST. HHSl is

taking suction on the discharge of the LHSl Pumps.

WE1 t

EA1 2: Ability to operate and / or monitor the following as they apply to the (Loss of

Emergency Coolant Recirculation): Operating behavior characteristics of the facility.

Surry Nuclear Plant 2004-301

DRAW SR8 lnital Exam

74. WE12EK2.2 001/l/I/A1W&EM

._ ~-

3.6/?.9Ml4301/R/MAU!SL&

~

- __

A steam break has occurred and all Steam Generators are faulted.

Which ONE of the follo~ing is the basis for maintaining a minimum of 60 gpm AFW

flow to each Steam Generator per ECA-2.1, Uncontrolled Depressurization of All

Steam Generators?

A. 60 gpm is needed to meet minimum heat sink flow requirements.

5. 60 gpm to each Steam Generator will ensure even thermal hydraulic distribution

across the core.

6:' 60 gpm is the minimum indicated flow rate to prevent Steam Generator dryout.

D. 68 gpm is the minimum indicated flow that will ensure the feed lines stay warm to

prevent excessive thermal shock to the feed lines during recovery actions.

References:

ND-95.3-LP-22, ECA-2.1 Uncontrolled Depressurization of All Steam Generators,

1 -E-3, ECA-2.1, Uncontrolled Depressurization of All Steam Generators, Rev. 16

Rev. 9

Distractor Analysis:

A. lncorrect because this requirement is not based on minimum heat sink flow

B. incorrect because this requirement is not based on thermal hydraulic distribution

C. Correct because 66 ggm is the minimum verifiable flow rate to a steam generator.

requirements, it is based on SG dryout.

across the core. It is based on S/G dayout.

This ensures 8 nominal flow rate of 25 gpm to the S/G, considering detector

uncertainties, to prevent dryout and thermal shock to the S/G.

D. Incorrect because the concern is with thermal shock to the SG if AFW flow rates are

rasied.

840 (W/E12) Steam tine Rupture

~ Excessive Heat Transfer

EK2.2: Knowledge of the interrelations between the (Uncontrolled Depressurization of

All Steam Generators) and the following: Facility's heat removal systems, including

primary coolant, emergency coolant, the decay heat removal systems, and relations

between the proper operation of these systems to the operation of the facility.

Modified Surry ILT Bank Question #1 Of 0

Sur9 Nuclear Plant 2004-3Qf

DRAFT SRO lnital Exam

~

__

~

1 -E-3, Steam Generator Tube Rupture, has been entered due to a ruptured tube in the

"A" Steam Generator. The Team is performing Step 4, which directs "A" Steam

Generator Narrow Range SG Level to be greater than 12% prior to stopping feed flow.

I -

Surly

References:

1453, Steam Generator Tube Rupture, Rev. 25

ND-95.3-LP-13, E-3 Steam Generator Tube Rupture, Rev. 11

Distractor Analysis:

A. Correct because this is the basis as stated in NB-95.3-LP-13.

3. Incorrect because the concern is not thermal gradients across the tubes. The

concern is to cover the tubes for thermal stratification and then stop AFW flow as

soon as the tubes are covered to give margin to overfili, while mitigating release to

the public.

C. Incorrect because this SG will not be used for the RCS cooldown.

8. Incorrect because the dP is still going to induce leakage even at 12% SG level.

WE1 3 Steam Generator Over-pressure

EK2.1: Knowledge of the interrelations between the (Steam Generator Overpressure)

and the following: Components, and functions of control and safety systems, including

instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Question is modified from a Braidwood Question.

Surry Nuclear Plant 2004301

DRAFT SRO lnital Exam

~

~

~

76. 00 1 G2.4 30

~- 00 1 /2/2/REPBKTABII .ITY&IEM

____

2 2/3 .AIB/SR0430 I/S/MABISLlR

-

- _ _ _

--

Which ONE of the following states an event that is required to be reported to the NRC

within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of discovery?

A. An inadvertant Safety injection due to an instrument surveillance error.

B: The Shift Supervisor authorizes the individual insertion of control rods into the core

without bank overlap to shutdown the reactor in an emergency.

C. A hypochlorite spill outside the Polishing Building of which the EPA has been

notified.

D. A radioactive release such that if an individual had been present for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, they

could have received an intake in excess of one occupational annual limit on intake.

I

References:

VPAP-2802, Notifications and Reports, Rev. 17.

Distractor Analysis:

A. incorrect because this is a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> reportable event. Plausible because the applicant

may think that inadvertant safety injection is important enough to require reporting

to the NRC within one hour.

B. Correct per VPAP-2802 Section 6.3.3 for deviation from Tech Specs. (VPAP-2802

C. incorrect because this is a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> reportable event. Plausible because the applicant

Page 77.)

may think that a hypochlorite spill with EPA motification is imporitant enough to

require reporting to the NRC within one hour.

applicant may think that a large radioactive release is important enough to require

reporting to the NRC within one hour.

B. incorrect because this is a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> reportable event. Plausible because the

001 Control Rod Drive

G2.4.38 Knowledge of which events related to system operations / status should be

reported to outside agencies.

Surry Nuclear Plant 2004-301

DRAFT SWO lnital Exam

- 77.

-_ 004G2

~

1.32 002/211iC~VCS/bZEM 3 3/3.61N/SRC)1301ISIMABISDR

~

-.

- _

_

~

-

During Unit 1 REFUELING SHUTDOWN and COLD SHUTDOWN operations, the

following valves shall be locked, sealed, or otherwise secured in the closed position

except during planned dilution or makeup activities.

- 1-CH-223, or

- 1-CH-212, 1-CH-215,

and 1-CH-218

Which ONE of the following correctly describes the time requirement and reason for

locking, sealing, or othetwise securing these valves following a planned dilution or

makeup activity in accordance with Technical Specifications?

A:' 15 minutes to prevent inadvertant boron dilution of the RCS.

B. 68 minutes to ensure the proper safety system alignment.

6. 15 minutes to ensure the proper safety system alignment.

D. 60 minutes to prevent inadvertant boron dilution of the RCS.

References:

Technical Specification 3.2.E.3, Amendment 199

Bistractor Analysis:

A. Correct per Technical Specifications and Basis.

B. Incorrect because Technical Specifications require within 15 minutes.

C. Incorrect because Technical Specifications Basis states that these valves shall be

closed to provide assurance that an inadvertant boron dilution will not occur.

D. Incorrect because Technical Specifications require within 15 minutes.

004 Chemical and Volume Control

G2.1.32: Ability to explain and apply all system limits and precautions.

Surry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

98. 009Eh2 39 001/l/~iNATUKA%.

CIRCUI,ATION/C/A 4.33 7/MiSR04301/SIR1ARISDR

- .

Given the following Unit 1 conditions:

- A small break LOCA has occurred

= As directed by the EOPs, the RCPs have been tripped

- 1 -ES-1.2, Post-LOCA Cooldown and Depressurization, Step 20, "Verify Natural

- RCS pressure is 1490 p i g

~ Wide Range T-Coid indications are 505 O F and slowly decreasing

- Wide Range T-Hot indications are 515 O F and slowly decreasing

- CETCs are 581 O F and stable

~ Containment Pressure is 18 psia

- Containment Radiation Levels are: 5.0 x 1 O5 Whr

I SG Narrow Range Levels are: A=22%, B=24%, C=22%, and slowly decreasing

- SG Pressures are 715 psig and stable

- RVLlS Full Range = 50%

According to 1-ES-1.2,

which ONE of the following correctly states the status of Natural

Circulation and the correct operator actions?

Circulation?" is being performed

A. Natural Circulation criteria are met. Begin depressurizing when subcooling is

9 85 O F .

B. Natural Circulation criteria are not met due to CETCs not decreasing. Depressurize

the SGs by raising steam flow rate through the steam dumps. Then depressurize

when subcooling is > 95 O F .

C. Natural Circulation criteria are not met due to SG pressure parameters not satisfied.

Depressurize the SGs by raising steam flow rate through the steam dumps. Then

depressurize when subcooling is 9 $5 O F .

WCS by raising steam flow rate through the steam dumps. Then depressurize wher

subcooling is > 95 O F .

D:' Natural Circulation criteria are not met due to inadequate subcooling. Cool the

Surry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

References:

1 -ES-I 2, Post LOCA C o o I d ~ ~ n

and Depressurization, Rev. 21

Distractor Analysis:

A. Incorrect because there is not adequate subcooling.

B. Incorrect because CETCs do not need to be decreasing.

C. Incorrect because SG parameters are satisfied.

B. Correct because there is inadequate subcooling (16 O F e 85 OF). ES-1.2 Step 20

RN8 directs dumping of more steam. The basis for Step 21 of dumping steam until

subcooling is < 95 O F is to ensure that the 85 O F natural circ criteria is not violated.

The Degraded Containment numbers were used due to the

CETC = 581'F; P = 1490 psig = 1505 psia; Psat(l505 psia) = 597

O F ;

Subcsoling = 59% - 581 = 16 O F

Surry IbT Bank Exam Question #I 869

009 Small Break LQCA

EA2.39: Ability to determine or interpret the following as they apply to a small break

LOCA: Adequate core cooling.

Surry Nuclear Plant 2004-301

DRAFT SRO M a l Exam

The following Unit 1 conditions exist:

RCS is not pressurized

- RCS level is 16.00 feet as read on 4 -RC-LI-1 OOA

Which ONE of the following specifies the mimimum mandatory backup cooling

method@) required to be available before entering the above plant conditions, in

accordance with OSP-ZZ-804, Unit 1 Safety Systems Status List For Cold Shutdown /

Refueling Conditions?

A. Reflux Boiling AND Gravity Feed and Bleed.

B. Gravity Feed and Bleed ONLY.

C. Forced Feed and Bleed AND Gravity Feed and Bleed.

D:' F ~ r ~ e d

Feed and Bleed ONLY.

References:

1 -0SP-ZZ-0644, Unit 1 Safety Systems Status List For Cold Shutdown / RefUelit7g

I -AP-27.00, Loss of Decay Heat Removal Capability, Rev. 10

ND-95.2-LP-12, Loss of WHR Events, Rev. 9

Conditions, Rev. 27

Distractor Analysis:

A. Incorrect: Per 4 -OS$-ZZ-004, Step 6.1 2, Forced Feed and Bleed is the only

B. Incorrect: Per 1 -0SP-ZZ-004, Step 6.1 2,

Forced Feed and Bleed is the only

C. Incorrect: Per 1-OSP-ZZ-0434, Step 6.1.2, Forced Feed and Bleed is the only

D. Correct: Per 1 -OS$-ZZ-604, Step 6.1 2,

Forced Feed and Bleed is the only

Mandatory Backup method required.

Mandatory Backup method required.

Mandatory Backup method required.

Mandatory Backup method required.

025 LO§§ of RHB

G2.4.7: Knowledge of event based EQP mitigation strategy

Surry Nuclear Plant 2004-301

DRAFT SRO lnitai Exam

The following conditions exist:

- A loss of all AC power has occurred.

- The STA reports the status of the CSFs are as follows:

- Subcriticality - RED

- Core Cooling - RED

- Heat Sink - RED

- Integrity - GREEN

- Containment - GREEN

~

Inventory - YELLOW

Which ONE of the following proceures should be used to mitigate these conditions?

A. 1 -FW-S.1 , Response to Nuclear Power Generation / ATWS

B.' 1-ECA-8.0, LOSS of All A 6 Power

6. 1-FR-H.1~

Response to Loss of Secondary Heat Sink

D. 1-FR-@.I, Response to Inadequate Core Cooling

Refe re nees:

1 -ECA-0.0, Loss of All AC Power, Rev. 21

Distractor Analysis:

A. Incorrect because FR's should not be implemented while in ECA-0.0. (see NOTE

B. Correct because this is the correct procedure to mitigate the loss of ac power.

6. incorrect because FR's should not be implemenkd while in ECA-0.63.

3. Incorrect because FR's should not be implemented while in ECA-0.0.

Susry IbT Exam Bank Question #899

prior to step 1 of ECA-0.0)

855 Station Blackout

EA2.03: Ability to determine or interpret the following as they apply to Station Blackout:

Actions necessary to restore power.

Surry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

81 . 056G2.4 45 001/2/1/CONDENSATEi(3/A

3 ~ M 2 ~ 3 0 l / S l M A B / s I I R

~

-

~

- - _ _ _

~

-

-

The following Unit 1 conditions exist:

- Condenser vacuum is lowering slowly.

- Steam Generator Bevels are 45% and lowering.

- Several alarms have annunciated, including:

- PQWW = 180%

~

1 H-G8, FW PP DISCH HDW LO PRESS

- 1J-G4, CN PPS DISCH HDR LO PRESS

- 1 C-AI, RCP 1 A CC RETURN LO FLOW

1C-B1, we19 1B cc RETURN LO FLOW

- 1 C-CI

I RCP 1 C CC RETURN LO FLOW

-

Which ONE of the following states the SWO's correct prioritization of the above

conditions as indicated by the procedures and actions chosen to mitigate or correct the

conditions?

A. Trip the Reactor followed by tripping the Reactor Coolant Pumps. Enter 4 ,

Reactor Trip or Safety Injection.

B.J Enter AP-10.05, LOSS of Semi-vital Bus. Verify that the standby condensate pump

has started and reduce turbine load.

C. Enter AP-21 .00,

Loss of Main Feedwater Flow. Maintain full power operation and

manually control Steam Generator levels by placing Feedwater Regulating Valves

in MANUAL control.

D. Enter Ab)-23.00, Rapid Load Reduction, to bring the unit offline, followed by tripping

the Reactor Coolant Pumps.

Sur9 Nuclear Plant 2004-301

DRAFT SRO lnital Exam

Sur9

References:

ND-90.3-LP-5, Vital and Semi-Vital Bus Bistribution, Rev. 11

1 -AP-l0.05, Loss of Semi-Vital Bus, Rev. I 6

I -A$-21 .00, Loss of Main Feedwater Flow, Rev. 5

1 -AP-23.00, Rapid Load Reduction, Rev. 15

1 H-G8, FW PP DlSCH HDW LO PRESS, Rev. 6

1 J-G4, CN PPS DISCH HBW LO PRESS, Rev. 0 16-141, RCP I A CC RETURN LO FLOW, Rev. 2

1 C-B1, RCP 1 B CC RETURN LO FLOW, Rev. 2

IC-C1,

RCP 1C CC RETURN LO FLOW, Rev. 2

Distractor Analysis:

A. Incorrect because loss of SVB causes indication to be lost for RCP CC Flow

Indication. RCPs should not be tripped. Plausible because if RCPs actually had no

cooling, the Rx should be tripped and RCPs should be secured.

B. Correct because a11 indications in the stern are caused by a loss of SVB. Verifying

S/B Condensate Pump starts and turbine load reduction are correct per AP-I 0.05.

@. Bncorrect because maintaining load at 100% will cause SG levels to continue to go

down. The F\\M and Condensate Recircs have Bailed open on the loss of the SVB,

thus making a load reduction a necesity. Plausible because SG levels are lowering

and an Applicant may think that opening a FRV may help to mitigate the condition.

B. lncsrrect because the unit should not be taken off line using AP-23.00 and RCPs

should not be tripped due to the loss of the SVB. Plausible because rapidly

bringing the unit off line and securing RCPs, given the stated conditions, may

appear logical to the applicant.

Modified Sur9 ILT Exam Bank Question #224 (maybe it could be considered a new

question?)

856 Condensate

G2.4.45: Ability to prioritize and interpret the significance of each annunciator or alarm.

Surry Nuclear Plant 2084-301

DRAFT SRO M a l Exam

The following Unit 1 conditions exist:

- Reactor Power = 30%

- Plant is in a Chemistry hold during a power ascension

- A loss of Vital Bus Ill occurs and operators enter I-AP-10.03, Loss of Vital Bus Ill

- Electricians quickly find a fault on Vital Bus 1-118 and believe that it will take 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />

- 1 -CC-TV-I 154, CCW TV for the A Reactor Coolant Pump (RCP), has closed and

- WCP temperatures are starting to slowly rise.

to repair.

cannot be reopened.

Which ONE of the following set of actions should the Senior Reactor Operator (SRO)

direct given the above conditions?

A. The SRO should direct the securing of the A RGP. Reactor power may be

maintained at 38% for the duration of the 18 hour2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> pepair to re-energize Vital Bus

1-111.

B. The SRO should direct the securing of the A RCP. Reactor power may be

maintained at 30% for two hours, at which time the SRO should direct preparation

to bring the unit to hot shutdown within the following six hours.

C: The SRO should direct a Reactor Trip, followed by the securing of the A RCP.

The SWO should then direct performance of 1 -E-Q Reactor Trip or Safety Injection,

and continue with applicable actions of I -AP-I 8.03.

B. The SRO should direct a controlled plant shutdown. If RCP temperatures exceed

action level limits, the pump should be secured and the SWO should direct

continuation of the controlled plant shutdown.

Surry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

Surry

References:

NB-93.3-LP-16, Permissiv~//Bypass?Vrip

Staters bights, Rev. 8

NB-93.3-LP-10, Reactor Protection - General, Rev. 5

MD-90.3-LP-5, Vital and Semi-vital Bus Distribution, Rev. I1

1 -AP-f0.03, Loss of Vital Bus IIi, Rev. 8

Distractor Analysis:

A. Incorrect because TS 3.16 and commitments made in GL-91-11 (also located in

Note prior to Step 17 in AP-I 0.83). The VB must be re-powered within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or

the unit must be in HSB within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Also incorrect because AP-1 0.03

will require a reactor trip. Plausible because the loss of VB causes a loss sf cooling

to "A" RCP. It may appear OK to continue operation because the power is K P-8.

3. Incorrect because AQ-10.03 requires a reactor trip and securing of RCP if CCW will

not be restored prior to RCP temperatures reaching action level limits. Plausible

because of the NOTE mentioned in the previous distractor analysis.

C. Correct because AP-10.83 directs Wx Trip and securing of RCP if CCW will not be

restored prior to getting cooling back Bo that pump. The stem states that the TV is

closed and cannot be re-opened, thus preventing cooling to be restored to the

RCP.

D. Incorrect because AP-I 0.03 directs Wx Trip, not a controlled shutdown. Plausible

because power is < $43, which may allow the applicant to incorrectly believe that a

shutdown is accepatble.

057 Loss of Vital AC lnst Bus

G2.1.6: Ability to supervise and assume a management role during plant transients

and upset conditions.

Surry Nuclear Plant 2804-301

DRAFT §BO lnital Exam

The following Unit 1 conditions exist:

- Unit 1 power is 100%

- No annunciators are lit

- Annunciator 1 K-H1 has just extinguished

Which ONE of the foilswing is the correct Abnormal Procedure to enter and correct

Event Classification?

(Reference provided)

A. Enter 8-AP-10.13, Loss of Main Control Room Annunciators, due to the loss of OR@

of the power supplies to Unit 1 annunciators. Enter the Emergency Plan and

deciare a Notification sf Unusual Event if the loss of annunciators lasts for greater

than 15 minutes.

B:' Enter O-AP-I 0.1 3: Loss of Main Control Room Annunciators, due to the loss of both

power supplies to Unit 1 annunciators. Enter the Emergency Plan and declare a

Notification of Unusual Event if the loss of annunciators lasts for greater than 15

minutes.

C. Enter 1 -AP-l8.@6, Loss of BC Power, and O-AP-10.13, Loss of Main Control Room

Annunciators, due to a loss of DC power and loss of one of the power supplies to

Unit 1 annunciators. Enter the Emergency Pian and declare an Alert if the loss of

annunciators lasts for greater than 15 minutes.

D. Enter 1 -AP-l0.06, Loss OB BC Power, and O-AP-10.13, Loss of Main Control W Q O ~

Annunciators, due to a loss of B@ power and loss of both power supplies to Unit 1

annunciators.

annunciators lasts for greater than 15 minutes.

Enter the Emergency Plan and declare an Alert if the loss of

Sway Nuclear Plant 2004-301

DRAFT SRO lnital Exam

References:

0-AP-18.13, Loss of Main Control Board Room Annunciators, Rev. 4

EPIP-I .01, Emergency Manager Controliing Procedure, Rev. 43

Distractor Analysis:

A. Incorrect because 1 K-HI not lit is indication of both power supplies to Unit 1

annunciator Panels having been lost.

3. Correct because 1 K-H1 not lit is indication of both power supplies to Unit 1

annunciator Panels having been lost. EPIP-I .01 Page 6 states that if safety system

annunciators are lost for greater than 15 minutes while above CSD, then a NOUE

shall be declared.

annunciator Panels having been lost. Since the plant is still at 100% power, there is

ne, indication that any BC Bus has been Isst; therefore 1 -AP-lO.86 should not be

entered. An Alert classification based on the loss of DC would be incorrect. As

stated above, a NQUE is the correct cfassificatisn.

Bus has been Isst; therefore 1-AP-10.06 should not be entered. An Alert

classification based on the loss of DC would be incorrect. As Stated abOVe, a

NOW is the correct classification.

C. Incorrect because 1 K-HI not lit is iRdkatiQR of both power supplies to Unit 1

D. Incorrect because the plant is still at 188% power, there is no indication that any DC

Provide EPIP-1.01 Pages 6 and 27

058 Loss of BC Power

G2.4.32: Knowledge of operator response to a loss of all annunciators

Surry Nuclear Plant 2084-301

DRAFT Sa0 lnital Exam

84.

062A2.12 001/2/1/VITAL

~. AC RUS/C/A 3.2/3.b/N/SR0330I/S/M~/SI>K

-~

___

~

1 I

~

Uni;

at 100% p ~ ~ e J t

exzences a loss of V i z u s I at 1200 hours0.0139 days <br />0.333 hours <br />0.00198 weeks <br />4.566e-4 months <br /> on Monday.

1

Operators enter 1-AP-10.01, Loss of Vital Bus I, and re-energize the Vital Bus from its

alternate source at f 21 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> on Monday.

Which ONE of the foliowing correctly states the required actions based on the above

condition?

I

I

A. In accordance with 1-AP-10.01, Vital Bus I must be reenergized from its primary

source by 1400 hours0.0162 days <br />0.389 hours <br />0.00231 weeks <br />5.327e-4 months <br /> on Monday, or be in Hot Shutdown by 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> on

Monday.

B. In accordance with 1 -AP-10.01, Vital Bus I must be re-energized from its primary

source by 1415 hours0.0164 days <br />0.393 hours <br />0.00234 weeks <br />5.384075e-4 months <br /> 081 Monday, Or be in HQt ShU~dOWn by 2815 hQUrS Qn

Monday.

C.J In accordance with 1-AP-10.01, Vital Bus I must be re-energized from its primary

source by 1200 hours0.0139 days <br />0.333 hours <br />0.00198 weeks <br />4.566e-4 months <br /> on Tuesday, or be in Hot Shutdown by 1880 hours0.0218 days <br />0.522 hours <br />0.00311 weeks <br />7.1534e-4 months <br /> on

Tuesday.

D. No shutdown requirements are in effect as long as Vital Bus B is energized.

i

I

Surry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

References:

1 -AP-l0.01, Loss of Vital Bus I , Rev. 13

NB-90.3-LP-5, Vital and Semi-vital Bus Distribution, Rev. 11

Distractor Analysis:

A. Incorrect because per AP-10.01 Step 16 c, the VB must be powered from its

normal source within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or the unit must be placed in HSD within the next 6

hours (also see MD-90.3-LP-5 Page 15). Plausible because if the bus is not

energeized, it must be repowered within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 1400 hours0.0162 days <br />0.389 hours <br />0.00231 weeks <br />5.327e-4 months <br /> is 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after

12010 hours.

B. Incorrect because per AP-10.01 Step 16 c, the VB must be powered from its

normal source within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or the unit must be plamd in HSD within the next 6

hours (also see ND-90.3-LP-5 Page 15). Plausible because if the bus is not

energeized, it must be repowered within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and f 41 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> is 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after

121 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.

normal source within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or the unit must be placed in MSD within the rIext 6

hours (also see ND-96.3-LP-5 Page 15). The consequences sf having VB-I not

energized by it5 primary source are mitigated, or corrected, by ensuring that it is

energized from its primary source within the specified time requirement.

B. Incorrect because per AP-10.01 Step 16 c, the VB must be powered B~om its

normal source within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or the unit must be placed in HSD within the next 6

hours (also see ND-98.3-LP-5 Page 15). Plausible because the Vital Bus is

energized and the plant would be operating satisfactorily.

C. Correct because per AP-10.01 Step 16 c, the VB must be powered from its

062 AC Electrical Distribution

A2.12: Ability to (a) predict the impacts of the following malfunctions or operations on

the ac distribution system; and (b) based on those predictions, use procedures to

correct, control, or mitigate the consequences of those malfunctions or operations:

Restoration of power to 8 system with a fault on it.

Surry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

~ 85. 062AA2 04 00111/1/SERVICE -~

WA1'EWC/A 2 512

~.

9/N/SR04301fSMABISDR

~

-

The following Unit 1 conditions exist:

- 1 -CH-P-1 A Charging Pump is operating

- 1 -SW-P-1 0A Charging Pump Service Water Pump is operating

- 1 -SW-P-1 OB Charging Pump Service Water Pump is in standby

- 1 D-G5, SW OW CC PPS DISCH TO CHG BPS LO PRESS, alarms

~

1 -CH-P-1 A Charging Pump Bearing Temperature = 175 O F

- 1 -CH-P-1 A Charging Pump Oil Cooler Outlet Temperature = 150 O F

Power = 100%

The Pressure Indication on the discharge of f-SW-P-10A Charging Pump Sewice

Water Pump (SW-Pl-26) reached a minimum value of 10 psig where it remains

stable.

- The Operator in the field reports back to the Control Room that 1-SW-P-10A

Charging Pump Service Water Pump is noisy and has high vibrations.

Which ONE of the following correctly states the appropriate assessment of the above

conditions and appropriate operator action based on that assessment?

A. Bearing Temperature is not within limits. The "A" Charging Bump is INOPERABLE.

Direct starting standby Charging Pump Service Water Pump, direct securing the

"A" Charging Pump Service Water Pump, and notify the System Engineer.

B. Bearing Temperature is not within limits. 1-CH-P-IA Charging Pump is

INOPERABLE. Verify auto start of 1 -SW-P-1 OB Charging Pump Sewice Water

Pump, and notify the System Engineer.

C:' Oil Cooler Outlet Temperature is not within normal operating band. 1 -CH-P-1 A

Charging Pump is OPERABLE. Direct starting standby Charging Pump Service

Water Pump, direct securing the "A" Charging Pump Service Water Pump, and

notify the System Engineer.

D. Oil Cooler Outlet Temperature is not within normal operating band. Performance of

Charging Pump Operability and Performance Test for 1 -CH-P-1 A Charging Pump

must be directed to determine OPERABILITY.

Surry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

Surry

References :

I D-G5, SW OR CC PPS DISCH PO CHG PPS LO PRESS, Rev. 3

11448-FM-071 B, Sh. 1 of 2, Flow / Valve Operating Numbers Diagram, Circulating and

ND-89.5-LP-2, Service Water System, Rev. 20

1 -0P-CH-002, Charging Pump A Operations, Rev. 13

1 -OPT-CH-001, Charging Pump Operability and Performance Test For 1 -CH-P-IA,

Service Water System, Surry Power Station Unit 1, Virginia Power, Rev. 50.

Rev. 33

Distractor Analysis:

A. Incorrect because Bearing Temperature is less than 180 OF. OPT-CH-001 Pg 9

B. Incorrect because Bearing Temperature is less than 180 O F . OPT-CH-OB1 Pg 9

states that the upper adrnin limit is 180 O F . The Charging Pump is still OPERABLE.

states that the upper admin limit is 180 O F . The standby pump will not start until 8

@. Correct because Oil Cooler Outlet Temperature is not within the normal operating

band (80 - 120 OF) as states in OPT-CH-001. However, the problem is not with the

Charging Pump, but with the Service Water flow, so swapping Charging Pump

Service Water Pumps is the correct initial action based on the ARP.

5. Incorrect because there is no indication that the Charging Pump has a problem.

Given the above alarm, all indications suggest that the problem is with the Service

Water flow. Therefore, performance of the Operability and Performance Test for

the Charging Pump would serve 690 purpose.

psig.

062 boss of Svc Water

AA2.04: Ability to determine and interpret the following as they apply to the Loss of

Nuclear Service Water: The normal values and upper limits for the temperatures of the

components cooled by SWS.

Surry Nuclear Plant 2004-301

DRAFT SWO lnital Exam

The following Unit 1 conditions exist:

- Reactor Power = 188%

- A loss of Containment Instrument Air has occurred

- 1 B-F6, CPMT INSP AIR HDR LO PRESSURE, annunciates

~

1 B-C6, PRZR PWR RELIEF VV LO AlR PRESS, annunciates

- Containment Instrument Air Pressure = 75 psig

~ Containment instrument Air was crosstied with Instrument Air

Which ONE of the following operator actions is required?

A. Both Pressurizer PQWVs are operable following the crosstie. Verify the operability

by closing POWV Block Valves, stroking PORVs, then re-opening the PORV Block

Valves.

5. Both Pressurizer PBWVs are operable following the crosstie. No further action

associated with the PORVs is required.

C:' Declare both Pressurizer PORVs inoperable. Close and remove power from both

PQRV block valves within one hour and be in HSD within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

D. Declare both Pressurizer PORVs inoperable. Close, but leave energized, both

POWV block valves within one hour and be in HSB within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Surry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

Surry

References:

ND-92.1-LP-1, Station Air Systems, Rev. 13

ND-88.1-LP-3, Pressurizer and Pressure Relief, Rev. 12

1 3433, CTMT lNSB AIR HDR LO PRESS, Rev. 1

1 D-C6, PRZW PWR RELIEF VV LO AIR PRESS, Rev. 4

Technical Specification 3.1 .A.6.c, Reactor Coolant System / Relief Valves

Distractor Analysis:

A. Incorrect because (per 1 3-CGJ with CTMT lnst Air P e 80 psig, the PORVs are

inoperable.

B. Incorrect because (per 1 D-@Gj with CTMT lnst Air P 6 80 psig, the PORVs ale

inoperable.

C. Correct because POWVs are not capable of being m ~ ~ u a l l y

cycled with CTMT lnst

Air P < 80 pig. Therefore, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, the CTMT lnst Air pressure must be > 80

psig or the block valves must be closed and de-energized. Furthermore, the plant

must be in HSD within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

D. Incorrect because, as stated in "C"

above, power must be removed from the block

valves.

Surry Requal Bank Question #394 (ARP0001)

079 Station Air

G2.4.48: Ability to interpret control room indications to verify the status and operation

of system, and understand how operator actions and directives affect plant and system

conditions.

Sur9 Nuclear Plant 2004-301

DRAFT SRO lnital Exam

The following Unit 1 conditions exist:

- Plant is in Mode 1

- Personnel Airlock Seal Leakage Testing has just been completed

- The Personnel Airlock lnner Boor Seal exceeded Technical Specifications leakage

- Earlier in the year the Personnel Airlock lnner Door exceeded Technical

limits

Specifications leakage limits and the Personnel Airlock Outer Door was opened for a

total of 59 minutes during the inoperability of the Personnel Airlock Inner Door

Which ONE of the following actions would satisfy required Technical Specification

Actions for the Personnel Airlock Doors?

A. The Personnel Airlock Outer Door may not be opened to pursue the repair and

retest. The plant must be shutdown and cooled down per Plant General Operating

Procedures. The plant must be in Hot Shutdown within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Cold Shutdown

Within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

B:' The Personnel Airlock Outer Door may be opened for 10 minutes to pursue the

repair and retest of the Personnel Airlock lwner Door Seal. Per VPAP-0106,

Subatmospheric Containment Entry, the Shift Supervisor shall supervise the

containment ent9 and exit process.

C. The Personnel Airlock Outer Door may be opened for 15 minutes to pursue the

repair and retest of the Personnel Airlock lnner Boor Seal. Per VPAP-0106,

Subatmospheric Containment Entry, the Unit SRO shall supervise the containment

entry and exit process.

D. The Personnel Airlock Outer Door may be opened for 1 how to pursue the repair

and retest of the Personnel Airlock Inner Door Seal. Per VPAP-0106,

Subatmospheric Containment Entry, the Unit SRO shall supervise the containment

ent9 and exit process.

Surry Nuclear Plant 2804-301

DRAFT SRO lnital Exam

References:

VPAP-0106, Subatmospheric Containment Entry, Rev. 5

Technical Specifications 3.8, Containment (Amendments 172 and 171 1; 1.6.G,

Definitions (Amendment 180)

Distractor Analysis:

A. Incorrect because the Outer Door may be opened for 10 minutes since it has

already been opened 50 minutes this year while the inner door was inoperable.

B. Correct because per Tech Specs, the Outer Door may be opened for I5 minutes or

60 minutes for the year (which leaves 10 more minutes for this instance).

Furthermore, the SS must supervise the containment entry and exit process pes

VPAP-0106 Section 5.6.

C. Incorrect because the Outer Door may be opened for 10 minutes and the SS must

supervise the containment entry and exit per VPAP-0106 Section 5.1.

B. Incorrect because the Outer Door may be opened for 10 minutes.

103 Containment

A2.01: Ability to (a) predict the impacts of the following malfunctions or operations on

the containment system; and (b) based on those predictions, use procedures to correct,

control, or mitigate the consequences of those malfunctions or operations: Integrated

Leak Rate Tests.

Surry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

The following Unit 1 conditions exist:

- Chemistry has just provided the following results from a Reactor Coolant System

The Unit has been operating at 100% power for the past two weeks

sample that was taken 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> ago:

- RCS Chloride = 0. I 5 ppm

- RCS Fluoride = 0.1 5 ppm

- RCS Oxygen = 0.1 5 ppm

Which ONE of the following describes the above conditions and appropriate operator

action?

A. Oxygen concentration is above the allowable Technical Specification limit. Per

Technical Specifications, corrective action must be taken immediately to bring the

plant to cold shutdown conditions.

B:' Oxygen concentration is above the allowable Technical Specification limit. Per

Technical Specifications, corrective action must be taken immediately to bring the

oxygen concentration within limits. If the oxygen concentration is outside of the limil

after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, then the plant must be taken to cold shutdown.

C. Chloride concentration is above the allowable Technical Specification limit. Per

Technical Specifications, corrective action must be taken immediately to bring the

plant to cold shutdown conditions.

D. Chloride concentration is above the allowable Technical Specification limit. Per

Technical Specifications, corrective action must be taken immediately to bring the

chloride concentration within limits. If the chloride concentration is outside of the

limit after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, then the plant must be taken to cold shutdown.

Surry Nuclear Plant 2084-301

DRAFT SRO M a l Exam

Surry

References:

Technical Specifications 3.1 F.1 and Basis

Distractor Analysis:

A. Incorrect because, according to the Tech Spec Basis, the plant has 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to see

if their corrective actions will bring the parameter within spec. If after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> the

parameter is not within spec, then the plant must be taken to cold shutdown using

normal plant procedures.

B. Correct because, according to the Tech Spec Basis, the plant has 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to see if

their corrective actions will bring the parameter within spec. If after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> the

parameter is not within spec, then the plant mush be taken to cold shutdown using

normal plant procedures.

6. incorrect because Chloride concentration is within limits.

D. Incorrect because Chloride concentration is within limits.

G2.1.34: Ability to maintain primary and secondary plant chemistry within allowabk

limits.

Surry Nuclear Plant 2804-301

DRAFT SRO lnital Exam

~

-

89. G2.1.4 001/3NTECH

~. SPEC STAIjl:ING/C/A 2.3/3.4/NISR~301/S/M[AB/SDIP_

~

~

~

~

-

The following plant conditions exist:

~

Unit 1 is shutdown and subcritical by 5.35% delta k / k

- Unit 1 Tavg is 100 O F

~

Unit 2 is shutdown and subcritical by 2.35% delta k / k

~

Unit 2 Tavg is 198 O F

Which ONE of the following correctly states the MINIMUM shift crew composition per

Technical Specifications?

A. I SS, 1 Unit SWO,

3 ROs, 4 AOs, and 1 STA.

B. 1 SS, 2 Unit SROs, 3 ROs, 4 AOs, and no STA.

C. 1 SS, 1 Unit SRO, 3 ROs, 4 AQs, and no STA.

D?' 1 SS, no Unit SRO, 2 ROs, 4. AOs, and no STA.

Surty

References:

Technical Specification Table 6.1 -1 (Minimum Shift Crew Composition), Amendrnemt

No. 123.

Distractor Analysis:

A. Incorrect because it does not match the minimum requirements for one unit in Cold

B. Incorrect because it does not match the minimum requirements for one unit in Cold

C. Incorrect because it does not match the minimum requirements for one unit in Cold

D. Correct because it matches the requirement for one unit in Cold Shutdown and one

Shutdown and one unit in Refueling Shutdown.

Shutdown and one unit in Refueling Shutdown.

Shutdown and one unit in Refueling Shutdown.

unit in Refueling Shutdown.

G2.1.4

Knowledge of shift staffing requirements.

Surry Nuclear Plant 2004-301

DRAFT SWO lnital Exam

The following conditions exist:

- Unit 1 is at 58% power

- Unit 2 is in startup mode with Tavg = 41 0°F

- Unit 2 Steam Driven AFW Pump and Motor Driven AFW Pump are declared to be

= Unit 2 Motor Driven AFW Pump is restored to operable status at 1 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> and Unii

inoperable at 0800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> on August 11 (all other AFW equipment is operable)

2 Tavg = 41 8°F

Which ONE of the following set of Technical Specification actions is correct?

(Reference provided)

A. Initially (with both pumps inoperable) both AFVV Pumps must be restored or Unit 2

must not enter Hot Shutdown. All Unit 1 Technical Specification Actions will be less

restrictive than the Unit 2 Technical Specification Actions.

B. Unit 2 A W actions do not apply. lnitialty (with both pumps inoperable) Unit 1 must

be in Hot Shutdown by 88/25 at 1400 hours0.0162 days <br />0.389 hours <br />0.00231 weeks <br />5.327e-4 months <br /> and Cold Shutdown by 08/26 at 2000

hours. After the Motor Driven A W Pump is operable no Unit 4 actions would be in

effect.

6:' Initially (with both pumps inoperable) Unit 2 must be in Cold Shutdown by 08/12 at

2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> and restore either AFW pump by 08/25 at 0800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> or Unit 1 must be

placed in Hot Shutdown by 08/25 at 1400 hours0.0162 days <br />0.389 hours <br />0.00231 weeks <br />5.327e-4 months <br />. After the Motor Driven AFW

Pump us restored, the Steam Driven AFW Pump must be restored by 88/14 at

0800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> or Unit 2 must be in Hot Shutdswn by 08/14 at 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br />.

D. Initially (with both pumps inoperable) Unit 2 shall not enter Hot Shutdown and must

be in Cold Shutdown by 88/12 at 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> and restore either A W pump within

I4 days or Unit 1 must be placed in Hot Shutdown by 08/25 at 0800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br />. After

the Motor Driven AFW Pump is restored, the Steam Driven AFW Pump must also

be restored by 08/14 at 0800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> or Unit 1 must be in Hot Shutdown by OW1 4 at

2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br />.

§wry Nuclear Plant 2004-301

DRAFT SRO InitaC Exam

Surry

Ref e rences:

Technical Specifications 3.6.C, 3.6.F, 3.6.G, and 3.01

Distractor Analysis:

A. Incorrect because Unit 2 does not need to be placed in HSD until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> following

B. Incorrect because Unit 2 Technical Specification Actions do apply above 350 O F and

C. Correct because LCO 3.0.1 is entered with both pumps inoperable because there is

not a tech spec condition that covers this situation. Once the MBAFW Pump is

operable, LCO 3.0.1 is exited, but 3.6.F and 3.6.C

still applies for Unit 2.

D. Bncorrect because Unit 1 does not need to be shutdown with only the Unit 2 Steam

Drivem AFW Pump inoperable.

08/14 at 0800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br />.

458 psig.

G2.2.23

Ability to track limiting conditions for operations.

Surry Nuclear Plant 2004-301

DRAFT SRO Bnital Exam

__

~

I

9 1 . G2.2.3 I OOl/3//REFUELINGK~A

~

7.2/2.9/N/SROJ301/SIM~UISDR_

_.

-~~

r

Unit 1 has been shut dawn for 21 days and fuel movement has just commenced.

Which ONE of the following is correct with regard to Fuel Building Exhaust and

Containment Purge Exhaust?

A:' Fuel Building Exhaust and Containment Purge Exhaust must be manually aligned

to continuously pass through CAT1 filters during fuel movements.

3. Fuel Building Exhaust and Containment Purge Exhaust will automatically align to

the CATl filters if a fuel handling accident occurred at this time.

I

6. There is no need to manually align Fuel Building Exhaust or Containment Purge

Exhaust to the CATl filters because the fuel has decayed for a sufficient period of

time such that radiofogical cansequences from a fuel handling accident would be

acceptable without iodine filtration.

I

D. Fuel Building Exhaust and Containment Purge Exhaust must be secured during fuel

movements to prevent automatically tripping the purge.

Surry

References:

ND-92.5-LP-7, Refueling Abnormal Procedures, Rev. 10

Distractor Analysis:

A. Correct because the automatic alignment feature is bypassed when fuel has

decayed for less than 30 days. Therefore, it must be manually aligned prior to

moving fuel.

decayed for less than 30 days.

B. Incorrect because the automatic alignment feature is bypassed when fuel has

6.

Incorrect because 30 days is considered sufficient decay time, not 21 days.

D. Incorrect because this in only a requirement during movement of the upper

internals.

G2.2.31

Knowledge of procedures and limitations involved in initial core loading.

Surry Nuclear Plant 2004-301

DRAFT SRQ lnital Exam

92. G2.2.4 -

001/3//PROCEDUKE CHAKGE/MkiM 2 . 3 / 3 . 3 M I ~ 3 0 1 / ~ A H / S ~

~

-

r

-

___

Which ONE of the following correctly states items that require a Regulatory Screen to

be performed in accordance with VPAP-300lt Station and Regulatory Reviews?

A. Emergency Action Level Change AND Station Curve Changes

B." Seismic Analyses AND Heating-Ventilation and Air Conditioning Analyses

C. Fire Protection Plan Changes AND Plant Flood Analyses

D. Oftsite Dose Calculation Manual Changes AND Equipment Qualification Analyses

Surry

-

~

~

~

_

_

_

_

~

_

_

_

_

-

.-

References:

VPAP-3001, Station and Regulatory Reviews, Rev. 9

Bistractor Analysis:

A. Incorrect because Emergency Action Level Changes are to be processed IAW

VPAP-0502 (see VPAP-3001 Page 2 of Att. 3), a Regulatory Screen is not required.

Plausible because both items are listed on VPAP-3001 Att. 3 Page 2.

B. Correct per VPAP-3001 Page 2 of Att.3.

C. Incorrect because Fire Protection Plan Changes are to be performed MW

VPAP-2481 (see VPAP-SO81 Page 2 of Att. 3), a Regulatory Screen is not required.

Plausible because both items are listed on VPAP-3001 Att. 3 Page 2.

Regulatory Screen is not required. Plausible because both items

are listed on VPAP-36301 Att. 3 Page 2.

D. lncorrect because ODCM changes are to be performed IAW VPAP-2103N, a

G2.2.6: Knowledge of the process for making changes in procedures as described in

the safety analysis report.

Surry Nuclear Plant 2004-301

DRAFT SRO M a l Exam

Which ONE of the following are all responsibilities that shall E be delegated by the

Station Emergency Manages?

A. Ordering Site Evacuation, Authorizing Emergency Exposure Limits.

BY Authorizing Notifications of NRC, State and Local Agencies of the Emergency

Status, Authorizing Emergency Exposure Limits.

i

1

C. Authorizing Notifications of NRC, State and Local Agencies of the Emergency

Status, Restricting Access to the Site.

8. Authorizing Emergency Exposure Limits, Restricting Aecess to the Site.

WefW@nC@§:

ND-95.5-LP-2, Station Emergency Manager, Rev. 8

Site Emergency Plan, Rev. 46

Distractor Analysis:

A. Incorrect because ordering a site evacuation may be delegated.

5. Correct because the answer is clearly stated in both sf the references.

C. Incorrect because restricting access to the site may be delegated.

D. l~correct because restricting access to the site may be delegated.

G2.4.29: Knowledge of the emergency plan.

Surry Nuclear Plant 2804-301

DRAFT SRO lnital Exam

94. G2.4.38 C01I3NSEMIMEM -

2.2/4OiNISR04301/S~~/SDR - - - - __ 1

- - -

1 -

Which ONE of the following correctly states the preferred order for assuming the

Station Emergency Manager responsibilities from the Shift Supervisor once the

Technical Support Center is activated?

A. Manger Nuclear Operations, Director Nuclear Station Safety and Licensing, Director

Nuclear Station Operations and Maintenance, Another Qualified SRQ

B. Site Vice-President, Director Nuclear Station Safety and Licensing, Director Nuclear

Station Operations and Maintenance, Manger Nuclear Operations

CJ Site Vice-president, Director Nuclear Station Operations and Maintenance, Director

Nuclear Station Safety and Licensing, Manger Nuclear Operations

D. Site Vim-President, Director Nuclear Station Operations and Maintenance, Manger

Nuclear Operations, Director Nuclear Station Safety and Licensing

Surry

References:

ND-95.5-LP-2, Station Emergency Manager, Rev. 8

Distractor Analysis:

A. Incorrect because this is not the preferred order as specified in ND-95.5-LP-2 Pg 3.

B. Incorrect because this is not the preferred order as specified in ND-95.5-LP-2 Pg 3.

C. Correct because this is the preferred order as specified in ND-95.5-LP-2 Pg 3.

D. incorrect because this is not the preferred order as specified in ND-95.5-LP-2 Pg 3.

G2.4.38: Ability to take actions called for in the facility emergency plan, including (if

required) supporting or acting as emergency coordinator.

Surry Nuclear Plant 2084-301

DRAFT SRO Onital Exam

Given the following plant conditions following an automatic reactor trip:

- RCS has been verified to be intact per 1 -E-8, Reactor Trip OF Safety Injection

- AFW Flow to "A" SG = 125 gpm

- AFW Flow to "B" SG = 4 10 gpm

- AFW Flow to "C" SG = I30 gpm

- NR "A" SG Level = 10%

NR "B" SG Level = 8%

NR "C" SG Level = 9%

~

-

~

RCS Pressure = 1750 psig and slowly rising

- PRZR heveh = 24% and slowly rising

- WCS subcooling based on CETCs is 8O'F

Operators have reached the point in 1 -E-0 where the:

reduced.

are to check if SI flow should be

Which ONE of the following would be the next series of operator actions?

A: Direct STA to begin monitoring Critical Safety Function Status Trees, Reset SI and

CLS, verify Instrument Air available, then stop all but one Charging Pump, followed

by isolating High Head SI to the Cold begs.

3. Transition to 1-ES-I .1, SI Termination, establish letdown, followed by raising

Pressurizer level to > 35%, then secure all but one Charging Pump.

C. Establish letdown, followed by raising Pressurizer level to > 3%%,

transition to

I-ES-1 .I, SI Termination, then secure all but one Charging Pump.

D. Direct STA to begin monitoring Critical Safety Function Status Trees, Reset SI and

CLS, verify Instrument Air available, align Charging Pump suction to the VCT, then

stop all but one Charging Pump.

Surv Nuclear Plant 2004-301

DRAFT SWO lnital Exam

Surry

References:

I-E-8, Reactor Trip or Safety Injection: Rev. 46

1 -ES-1.1, SI Termination, Rev. 29

ND-95-03-03, E-0, Reactor Trip of Safety Injection, Rev. 14

Distractor Analysis:

A. Correct because these actions are directed by 1 -E-0 Steps 26 through 32.

B. Incorrect because letdown would not be established prior to Pnr L > 35%. Plausible

because transition to ES-1~

1 is logical and distractor states that the goal is to get

Pnr L > 35%.

C. Incorrect because letdown would not be established prior to Pzs L > 35%. Plausible

because transition to ES-1 .I is logical and distractor states that the goal is to get

Bzr b z 35%.

D. Incorrect because Charging Pump suction would not be aligned to VCT until after all

but one Charging Pump is secured. Plausible because all actions are directed by

procedure, except that the order of the suction swap and pump stopping is

reversed.

WE01 Wediagnosis and SI Termination

G2.1.20: Ability to execute procedures.

Surry Nuclear Plant 2004-301

BRAFT SRO lnital Exam

-

96.

._

Wb03Eh3.1

__ O O l I ~ Q Q L ~ W N ~ C N ~ M

3SiJ.2/13/§~430l~ABISDR __

__

I

Operators are responding to a LOCA outside of containment using 1 -ECA-1.2, LOCA

Outside Containment. The crew efforts to isolate the break are unsuccessful.

Which ONE of the following identifies the procedure ECA-I .2 will direct the operators

to in order to cool and depressurize the reactor coolant system?

A. 1 -E-1 , Loss of Reactor or Secondary Coolant

B. 1 -ES-I 2,

Post LOCA Cooldown and Depressurization

6. I-ES-1.3, Transfer to Cold Leg Recirculation

BI' 1 -ECA-1.1 I Loss of Emergency Coolant Recirculation

Surry

References:

1-E-1

I Loss of Reactor or Secondary Coolant, Rev. 21

I-ES-I

.2: Post LOCA Cooldown and Depressurization, Rev. 21

1 -ES-1.3, Transfer to Cold Leg Recirculation, Rev. 12

1 -ECA-l.l~ Loss of Emergency Coolant Recirculation, Rev. 17

Distractor Analysis:

A. Incorrect as stated in Distractor D Analysis. Plausible because there is a Loss of

B. Incorrect as stated in Distractor D Analysis. Plausible because the goal is to cool

C. Incorrect as stated in Distractor D Analysis. Plausible because this is a noma!

D. Correct because Step 2 RNO of ECA-1.2 directs operators to ECA-1.1 if efforts to

Reactor Coolant in progress.

and depressurize the RCS.

transition for long term cooling during a LOCA.

isolate the leak are not successful.

W/E03 LOCA C O O ~ ~ O W ~

- Depress.

EA2.1: Ability to determine and interpret the following as they apply to the (LOCA

Cooldown and Depressurization): Facility conditions and selection of appropriate

procedures during abnormal and emergency operations.

Bank Question TPQ2301.

Surry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

The following conditions exist:

- A manual Safety Injection was initiated due to a Steam Break in Safeguards

- All SG pressures are steadily decreasing

- All SG NR levels are off-scale low and WR levels are steadily decreasing

- Pressurizer level is off-scale low

- Pressurizer pressure is steadily decreasing

- RCS temperature is decreasing uncontrollably

- Adequate Auxiliary Feedwater flow exists

Which ONE of the following is the correct procedure transitions for the event in

progress?

A.@ E-0 BO E-% to ECA-2.1

B. E-0 to E-1 to E-2 to ECA-2.1

c. E-O to E-1 to ECA9.1

8. E-O to E-2 to E-1

Surry

References:

1 - E O , Reactor Trip or Safety Injection, Rev. 46

1 -E-2, Faulted Steam Generator Isolation, Rev. 9

I -ECA-2.1, Uncontrolled Depressurization of All Steam Generators, Rev. 19

Distractor Analysis:

A. Incorrect because E-0 would be entered upon Rx Prig. Step 21 of E-0 sends the

team to E-2. Step 2 of E-2 sends the team to ECA-2.1.

5. Incorrect because E-0 Step 21 directs performance of E-2. E-1 is not directed until

C. Incorrect because E-0 Step 21 directs performance of E-2. E-1 is not directed untii

D. Incorrect because E-2 would not be entered until after -I.

E-0 Step 23.

E-O Step 23.

WE05 Inadequate Heat Transfer - Loss of Secondary Heat Sink

EA21 : Ability to determine and interpret the following as they apply to the (Loss of

Secondary Heat Sink): Facility conditions and selection of appropriate procedures

during abnormal and emergency operations.

Surry ILT Bank Question #1342

Surry Nuclear Plant 2004-301

DRAFT SWQ lnitai Exam

9%.

WEIOEA2.1

~ 001/1/2/NATURAL

__ --

CITPCUl~A~IO~/C/A~~.9/R/SROL93~SiM~SI)R-

- ~

-

-1

During a Natural Circulation Cooldown IAW ES-0.3, Natural Circulation Cook?own with

Steam Void in Rx Vessel, a steam bubble forms in the vessel head. The STA

recommends transition to FR-1.3, Response to Voids in Reactor Vessel, to vent the

head.

Which ONE of the following courses of action is appropriate?

A. Initiate FR-1.3 since E%-0.3 assumes FR-1.3 is in effect to eliminate the steam void.

B. initiate SI and go to FR-1.3 to vent the head.

C. The NC Cooldown should be stopped and a transition to FR-1.3 should be made.

D! Stay in ES-0.3. Void growth is expected and ES-0.3 provides guidance to control

the void growth.

Slarry

References:

I -FR-1.3, Reponse To Voids In Reactor Vessel, Rev. 16

1 -ES-0.3, Natural Circulation Cooldown With Steam Void in Rx Vessel, Rev. 12

Distractor Analysis:

A. Incorrect because ES-0.3 does not assume that FR-1.3 is being used.

B. Incorrect because SI should not be initiated and there is n~ need lo vent the head.

C. Incorrect because ES-0.3 does provide guidance for managing void growth.

B. Correct because ES-0.3 does provide guidance for managing void growth.

WE4 0 Natural Circ.

Ea2.1: Ability to determine and interpret the following as they apply to the (Natural

Circulation With Steam Void in Vessl with / without RVLlS): Facility conditions and

selection of appropriate procedures during abnormal and emergency operations.

Surry Requal Bank Question #247

Surry Nuclear Plant 2004-301

DRAFT SRO lnital Exam

1

99. WE13EA2.1001/1R/SG

- __ - -

STEAM -

GENEKATOR/C/A

~

~

2 ~)134iRI~0430~ISIMA~SDR- - .~

I -

During performance of E-3, Steam Generator Tube Rupture, the operating team is

directed to adjust the SG POWV setpoint on the ruptured SG to 1035 psig. The

Reactor Operator observes ruptured SG pressure to be f07Q psig afld the P6RV

cycling.

Which ONE of the following is the appropriate course of action and reason for the

action?

A. Transition to FR-H.2, "Response to Steam Generator Overpressure" to prevent an

overpressure condition in the ruptured SG.

I

B. Increase feed flow to the ruptured SG to stop the release and remain in E-3.

C. Increase the setpoint above 1070 psig to prevent release to the public and

transition to ECA-3.1, SGTW With Loss of Reactor Coolant - Subcooled Receovery.

D:' Leave the PBRV setpoint at 1035 psig to minimize challenges to the SG Code

Safeties and remain in E-3.

Surry Nuclear Plant 2004-301

DRAFT SRO M a l Exam

References:

1 -E-3, Steam Generator Tube Rupture, Rev. 25

ND-95.3-LP-13, E-3 Steam Generator Tube Rupture, Rev. 11

Distractor Analysis:

A. Incorrect because the correct response is simply to verify that the PORV seats

when pressure drops below 1835 gsig. Furthermore, FR-H.2 is not associated with

any Red or Orange paths.

when pressure drops below I035 psig. Furthermore, feeding a ruptured SG will not

limit the exposure to the public.

when pressure drops below 1035 psig. Furthermore, this action may challenge the

code safeties, which is not desirable.

for the step.

5. Incorrect because the correct response is simply to verify that the PORV seats

C. Incorrect because the correct response is simply to verify that the PORV seats

&. Correct because this is the correct direction in the procedure and the correct basis

WE13 Steam Generator Overpressure

EA2.1: Ability to determine and interpret the following as they apply to the (Steam

Generator Overpressure): Facility conditions and selection of appropriate procedures

during abnormal and emergency operations.

Scerry Requal Exam Bank Question #324

Surey Nuclear Plant 2004-301

DRAFT SRO lnital Exam

f - - - - -

00. WEISEA2.1 0 2 / 1 / 2 / C O N T A I N ~ ~ N T ~ ~ ~ ~ ~ E ~ 2 . 7 / 3 . 2 ~ ~ 0 4 3 ~ I I S I M A ~ R I S D ~

__

-

~

-

-

The Control Room Operators are performing FR-S.2, Response to Loss of Core

Shutdown, in response to a yellow path condition shown on the Critical Safety Function

(CSF) status tree.

Which ONE of the following is correct with regard to transitions out of this procedure?

A. The operators must leave this procedure at any step as soon as the Loss of Core

Shutdown CSF adverse condition has cleared. (Green path established)

13. The operators must leave this procedure before completion and ge, to FR-H.1,

Response to Loss of Secondary Heat Sink, if the heat sink CSF status tree

indicates a yellow path condition.

C. The operators must leave this procedure before completion and go to FR-C.3,

Response to Saturated Core Cooling, if the Core Cooling status tree indicates a

yellow path condition.

D:' The operators must leave this procedure before completion and go to FW-Z.2,

Response to Containment Flooding, if the containment CSF status tree indicates an

orange path condition.

Surry

Refernces:

NB-95.3-LP-26, Critical Safety Function Status Trees, Rev. 5

Distractor Analysis:

A. Incorrect because the operator does mot have to immediately leave FR if it is not

5. Incorrect because yellow path does not warrant this action.

C. Incorrect because yellow path does not warrant this action.

B. Correct because orange path takes priority.

completed.

WE1 5 Containment Flooding

EA2.1: Ability to determine and interpret the following as they apply to the

(Containment Flooding): Facility conditions and selection of appropriate procedures

during abnormal and emergency operations.

Surry ILT Bank Question # 1350