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Feb-March 2004 Exam 50-280, 50-381/2004-301FINAL SRO Written with References
	ML040890642
Person / Time
Site:	Surry, 05000381
Issue date:	02/19/2004
From:	; NRC/RGN-II
To:	;
References
	50-280/04-301, 50-281/04-301
	Download: ML040890642 (142)
	v • d • e

Final Submittal SURRY EXAM 50-280, 50-281/2004-301 FEBRUARY 24 - MARCH 2

& MARCH 4,2004 (WRITTEN)

U.S. Nuclear Regulatory Commission Site-Specific SRQ Written Examination Applicant Information Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. To pass the examination you must achieve a final grade of at least 80.00 percent overall, with a 70.00 percent or better on the SRO-only items if given in conjunction with the 80 exam; SRO-only exams given alone require an 80.00 percent to pass. You have eight hours to complete the combined examination, and three hours if you are only taking the SRO portio Applicant Certification All work done on this examination is my own. I have neither given nor received ai Applicant's Signature Results 80 / SRQ-Only / Examination Values:

-1-1-Points Applicant's Scores:

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Surry Nuclear Plant 2004-301 SRQ lnital Exam K3.03 001/2/1/RCP I.UHKICATION/ME;.M

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

bearing oil lift system?

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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 assemblie start. Once the RCP reaches operating speed, the WCP Oil Lift Pump supplies the bearing lubricatio start. Once the RCP reaches operating speed, the RCP Oil Lift Pump supplies the bearing lubricatio start. Once the RCP reaches operating speed, the thrust runner circulates oil in the upper and lower bearing assemblie B. The oil lift pump discharge pressure must be greater than 300 psig prior to RCP The oil lift pump discharge pressure must be greater than 350 psig prior to RCP D. The oil lift pump discharge pressure must be greater than 300 psig prior to RCP

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Sur9 References:

ND-88.4-LP-6, Reactor Coolant Pumps, Rev. 16 Distractor 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 assemblie. Incorrect because pressure interlock is at 350 psig, not 300 psi C. Incorrect because thrust runner circulates oil in upper and lower reservoir, not the B. Incorrect because pressure interlock is at 350 psig, not 300 psi Reactor Coolant Pumps K4.03: Knowledge of RCPs design feature@) and I or interlock(s) which provide for the following: Adequate lubrication of the RC Oil Lift Syste Surry Nuciear Plant 2004-301 SWO lnital Exam 2. 004.42 17 001/2/1/CVCSICIA 3 413 7NliSKQ4301EUkW3ISLK

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The following Unit 1 conditions exist:

- Operating at 85% power

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Pressurizer pressure control is in its normal configuration

- A Pressurizer Safety Valve is leaking

- 1 C-58, PRZR LO PRESS, annunciates

- 1-AP-31.00, lncreasing or Decreasing RCS Pressure, has been entered Which ON of the following correctly describes the affect on charging flow and an appropriate mitigating action in accordance with 1 -AP-31.OO?

A. Charging flow initially increases. Place the PRZR PRESS MASTER CNTRL in B. Charging flow initially decreases. Place the PRZR PRESS MASTER CNPWL in CY Charging flow initially increases. Place the PRZR PRESS MASTER CNTRL in B. Charging flow initially decreases. Place the PRZR PRESS MASTER CNTRL in MANUAL and increase the demand to try to stop the pressure decreas MANUAL and increase the demand to try to stop the pressure decreas MANUAL and decrease the demand to try to stop the pressure decreas MANUAL and decrease the demand to try to stop the pressure decreas Surry References:

ND-93.3-LP-5, Pressurizer Pressure Control, Rev. 9 ND-88.3-LP-2, Charging and Letdown, Rev. 70 1 -AP-31.BO, lncreasing or Decreasing RCS Pressure, Rev. 4 Distractor Analysis:

A. Incorrect because increasing the demand wil! lower pressure, not increase i B. Incorrect because charging flow will not initially decrease and increasing the C. Correct because charging flow will initially increase due to the sudden pressure drop demand will lower pressure, not increase i in the RCS. Also, decreasing the demand on the controller while in manual will act to try to raise pressur B. Incorrect because charging flow will not initially decreas Chemical and Volume Control A2.17: Ability to (a) predict the impacts of the following malfunctions or operations on the CVCS; and (6) based on those predictions use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Low PZR pressur Surry Nuclear Plant 2004-301 SRO lnital Exam 3. 005K5.02 001/2/1iRHW'EM 3.4/3.5/B/SR04301/RIMAR/SDR

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The following Unit 1 conditions exist:

- WCS level is 12.5 feet on 4 -WC-LI-l OOA

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

- A loss of decay heat removal has occurred and 1 -AP-27.00, boss of Decay Heat Removal Capability, has been entere The RHR system has just been made availabl Which ONE of the following methods per 1-AQ-27.00 should be used to sweep air fror the RHR lines during a loss of decay heat removal capabikty if inadequate time exists to completely vent the RHR System prior to boiling in the core?

A? Refill the RCS to 13.5 feet, verify 10 OF subcooling, and run an RHR pump at a flow B. Maintain RCS level at 12.5 feet, verify subcooling, and run an RHR pump ai a flaw rate of > 2950 gpr of > 2950 gp of c 2950 gp C. Maintain RCS level at 12.5 feet, verify subcooling, and run an RHR pump at a flow D. Close RH-MOV-172RA and B, RHR Outlets, then open "A" and "C" Safety Injection Accumulator Isolation MOV Surry References:

ND-88.2-LP-1, Residual Heat Removal System Description, Rev. 8 ND-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-24.00, Loss of Decay Heat Removal Capability, Rev. 10 Bistractor Analysis:

A. Correct because based on procedural Note in 1 -AP-27.00, Page 16 of 19, Rev. 1 B. Incorrect because RCS needs to be filled to 13.5 fee C. Incorrect because RCS needs to be filled to 13.5 feet. Also flow needs to be greater than 2950 gp D. Incorrect because no procedural guidance exists to support the action Surry ILT Exam Bank Question #275 OR5 Residual Heat Removal K5.02: Knowledge of the operational implications of the following concepts as they apply to the RHRS: Need for adequate subcoolin Surry Nuclear Plant 2004-301 SRO lnital Exam

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4. W6K6 03 001/2/Il/CIA 3 6'3 9/N/SR04301i?JMABISDR -

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Unit 1 tripped 30 minutes ago. The following plant conditions currently exist:

- 7 CETCs indicate between 700 O F and 750 OF and slowly rising

- No RCPs are operating

- RVblS Full Range is indicating 46% and slowly lowering

- Subcooling based on CETCs is 0 O F

- E-0, Reactor Trip or Safety Injection, has been exited and Safety Function Status

- Fa-6.1, Response to Inadequate Core Cooling, has been implemented

- WCP Seal Injection flow is 3 gpm to all RCPs

- RCP Seal delta&

are all approximately 200 psid

- Attempts to establish HHSl flow have failed Which ONE of the following describes the correct strategy for mitigating the consequences of these conditions?

Steam Generator levels are 20% and rising Trees are being monitored Source Range Startup Rate is zero A. Depressurize all intact steam generators ab maximum rate, while maintaining steam flow less than 1.O x lo5 PPH to try to establish accumulator and LHSl flow. If CETCs rise above 1200 O F, then check conditions for starting a RC B. Depressurize all intact steam generators at maximum rate, while maintaining steam flow less than 1.O x 1 O5 PPH to try to establish accumulator and LHSl flow. If CETC injection flo I I

temperatures rise above 1200 OF, RCPs should not be started due to low WCP seal C. Depressurize all intact steam generators at maximum rate, while maintaining steam flow less than 1.O x 106 PPH to try to establish accumulator and LHSl flow. If CETC temperatures rise above 1200 O F, WCPs should not be started due to low RCP seal

I injection flo I B I Depressurize all intact steam generators at maximum rate, while maintaining steam flow less than 1.0 x lo6 PPH to try to establish accumulator and LHSl flow. If CETCs rise above 1200 O F, then check conditions for starting a RCF'.

Surry Nuclear Plant 2004-301 SRO lnital Exam Surry References:

NB-95.3-LP-38, FR-C.l Response to Inadequate Core Cooling, Rev. 8 FR-C.1, Response to Inadequate Core Cooling, Rev. 18 Distractor Analysis:

A. Incorrect because 1.0 x 1 O5 PPH is well below the MSlV closure setpoint and does not even approach the maximum rate (an entire order of magnitude !ow).

3. Incorrect because 1.0 x 1 O5 PPH is well below the MSIV 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 me B. Correct because procedural guidance exists to support the actions. MSIV c/osure will occur if flow is greater than 1.O x 1 Q6 PPH. The purp~se for the actions is to establish low head flow from accumulators and LHSI. WCP support criteria is desirable, but not a prerequisite for starling RCP Emergency Core Cooling K.6.03: Knowledge of the effect of a toss or malfunction on the following will have on ECCS: Safety Injection Pump Surry Nuclear Plant 2004-301 SRO lnital Exam 5. 007EK2.02 OOlll/l/EREAKER

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The following conditions exist:

- Unit 1 is at 90% power

- Reactor protection testing is in progress

- Reactor Trip Breaker " A is closed

- fleactor Trip Breaker "B" is open

- Reactor Trip Bypass Breaker "B" is racked in and closed Which ON 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 tri BY Only reactor trip bypass breakers "A and "B" will trip open and the reactor will tri C. Reactor trip breaker "A" will trip open and the plant will remain at 90% powe D. fleactor trip bypass breaker "A" wil( trip open and the plant wilj remain at 90%

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Surry References:

NB-93.3-LP-17, AMSAC, Rev. 10 NB-93.3-LP-10, Reactor Protection - General, Rev. 5 Distractor Anaysis:

A. Incorrect because reactor trip breaker "A" will not ope. Correct because this is the correct response per ND-93.3-LP-1 C. Incorrect because reactor trip breaker "A will not open and plant will tri B. Incorrect because the plant will tri Surry ILT Bank Question #1667 007 Reactor Trig Stabilization EK2.02: Knowledge of the interrelationships between a reactor trip and the following:

Breakers, relays, and disconnect Surry Nuclear Plant 2004-301

§RQ lnital Exam 6. 007 __

02 001/2/1E'RT

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PZR BtJl3BLEICIA 3 1/3

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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 PRT

- PRT Level = 95%

- Pressurizer Heaters are energized Which ONE of the following must be accomplished prior to drawing a bubble in the Pressurizer?

A 1 Drain the PRT to 60 - 80%.

B. Verify VCT oxygen concentration less than 3%.

C. Brain the Pressurizer to 22.2%.

B. Pressurize the RCS to 200 ~ 290 psig on PI-1-403, Nar Rang Surry References:

1 -GOP-l.l, Unit Startup, RC§ Heatup from Ambient to 195 Degrees F, Rev. 25 1 -QP-RC-Oll, Pressurizer Relief Tank Operations, Rev. 13 Distractor Analysis:

A. Correct bemuse GQP-1.1 Step 5.5.4 directs establishment of normal BRT level prior to drawing a bubble. QP-WC-011 Step 5.1.1 states the normal PRT level to be 60 - 80%.

B. Incorrect because GOP-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 D. Incorrect because RCS should be between 300 and 370 psig on PI-1-40 Pressurizer Relief / BuenchTank K5.02: Knowledge of the operational implications of the following concepts as they apply to BRYS: Method of forming a steam bubble in the PZ (GQP-1.1, Step 5.5.13).

Sur9 Nuclear Plant 2004-301 SRO lnital Exam pressure stabilizes around 2000 psi CI PQRV-1455C opens, at 2000 psig PBRV-f455C closes; however, pressure will continue to decrease causing a reactor trip and safety injectio Sur9 References:

NB-93.3-LP-5, Pressurizer Pressure Control, Rev. 3 Bistractor Analysis:

A. Incorrect because both spray valves also open, which causes pressure to continue to decreas B. Incorrect because both spray valves open, which causes pressure to continue to decrease. Also incorrect because PQWV-1456 does not ope. Correct because both spray valves open causing a reactor trip on OT-delta-T or Low Pressurizer Pressure, followed by S. Incorrect because POWV-1456 does not ope Pressurizer Pressure Control AA2.03: Ability to determine and interpret the following as they apply to the pressurizer vapor space accident: PQRV logic control under low-pressure condition Surry Nuclear Plant 2004-301 SRO lnital Exam 8. 008~1 02 I~~I/~/I/CCW/MJM 3 313 4/BISR01301mhf4B/SnK

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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 seal I I

3. HHSl pump seals, LHSl pump seals, RHR pump seals, RCP motor air cooler CY RHR pump seals, RCP bearing lube oil coolers, neutron shield tank coolers, HHSl pump seat cooler B. LHSl pump seals, RHR pump seals, RCP motor air coolers, neutron shield tank SUrV Reference:

NB-88.5-LP-1, Component Cooling, Rev. 19 ND-88.3-LP-5, Charging System, Rev. 16 Bistractor Analysis:

A. Incorrect because outside recirc spray pump seals are not cooled by C B. Incorrect because LHSl pump seals are not cooled by C C. Correct because all are cooled by CC or a subsyste D. Incorrect because LHSI pump seals are not cooled by C Requal Bank Question #527 008 Component Cooling K1.02: Knowledge of the physical connections and / or cause-effect relationships between the CCWS and the following systems: Loads cooled by CCW Surry Nuclear Plant 2004-301 SRO lnital Exam 9. OORK4 01 OOl/Z/i/COMI'OhZ.NT COOI.INCi/MF~M 3 113 3/R/SRO4301/RIMAB/SDR

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Ten seconds after a Safety Injection occurs, the "A" Component Cooling Pump trip Which ONE of the following describes the operation of the CC pumps?

A: The 3 CC pump will not auto start without a required operator actio B. The "B" CC pump will auto start 60 seconds after the "A" CC pump trip C. The "B" CC pump will auto start as soon as the "A" CC pump trip D. The "B" CC pump will auto start 50 seconds after the "A" CC pump trip Surry References:

ND-88.5-LP-1, Component Cooling Water System, Rev. 1 Distractor Analysis:

A. Correct because Auto Start Inhibit due to SI will prevent auto start of the CC pump, but the pump may be manually started at any tim B. Incorrect because the Auto Start Inhibit will block the auto star C. Incorrect because the Auto Start Inhibit will block the auto star D. Incorrect because the Auto Start Inhibit will block the auto star ILT Bank Question # 537 008 Component Cooling Water System K4.01: Knowledge of CCWS design feature@) and/or interlock(s) which provide for the following: Automatic start of standby pum Surty Nuclear Plant 2004-301 SRO lnital Exam 4 0. OIOA1.01 001/2/1/BORON PZR SPRAYICIA 2,8/2.9/N/S~O43OI/WhliW/SDR

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I The following plant conditions exist:

- Dilution to criticality has just been completed

- Operators note that inadequate proportional heaters appear to be energized

- Pressurizer Pressure is 2230 psi Which ONE of the following could result from inadequate Pressurizer Heater output during a dilution to criticality?

(Assume all other controls and components working properly in their normal configuration.)

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A." Boron concentration will be higher in the Pressurizer than in the RC B. Boron concentration will be lower in the Pressurizer than in the RC C. Pressurizer and RCS boron concentration will be approxirnate8y equa B. The Pressurizer Spray Nozzle will be susceptible to thermal shoc ~

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Surry References:

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

A. Correct because RCS boron will be lower due to the dilution. The Pzr will still be at a higher boron concentration until spray flow has created enough out-surge to adequately equalize the boron with the RCS. (Lack of heaters creates lack of sprays.)

B. Incorrect because boron concentration will be higher in the Pm C. Incorrect because the lack of heater output will net allow for adequate mixin. Incorrect because the bypass spray valves are normally open, which is sufficient to prevent thermal shoc Pressurizer Pressure Control Al.O1: 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 concentratio Surry Nuclear Piant 2084-301 SFiO lnital Exam 1 1. 01 E A 1 13 00111111SN;E~Y INJEClIONIC'IA~4 114 2/N/SR04301flUh?.~WISDR Given the following conditions:

- LOCA has occurred

- RCS subcooling is 0 O F

- WWST Level = 15% and slowly decreasing

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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 for each train if automatic transfer does not occu B! Close Charging Pump Miniflow Recirc Valves. When WWST level reaches 13%

push both RMT pushbuttons for each train if automatic transfer does not occu C. With RWST level at 15%, 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 RM D. With RWST level at 13%, push both RMT pushbuttons fer each train If automatic transfer does not occur. Secure Containment Spray Pumps immediately following verification of Phase 1 and 2 WM Surry Nuclear Plant 2004-301 SRO M a l Exam Surry References:

NB-95.3-LP-7, E-1 boss of Reactor or Secondary Coolant. Rev. 14 1 -E-1, Loss of Reactor or Secondary Coolant: Rev. 21 1-ES-1.3, Transfer to Cold Leg Recirculation, Rev. 52 Distractor Analysis:

A. Bncorrect because RMT transfer should not occur at 15%. It will occur at 13.5% and procedurally should be verified at 13%.

B. Correct because verifications should be made at 13% and manually initiated if needed (directed by E§-1 3).

C. Incorrect per ES-1.3 spray pumps should not be secured until they show signs of cavitation. Also E-1 does not call for spray to be secured until containment pressure is less than 12 p i a. Also, RMB transfer should not occur at 15%.

cavitation. Also El does not car[ for spray to be secured until containment pressure is less than S 2 psi B. Incorrect per ES-1.3 spray pumps should not be secured until they show signs of Surry IkT Bank Question # 872 01 1 Large Break LOCA EAI.13: Ability to operate and monitor the following as they apply to a large break LOCA: Safety injection Component Surry Nuclear Plant 2004-301 SRO lnital Exam Due to a controller failure, 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 gp Which ONE of the following correctly describes the behavior of FCV-1122 when the Operator attempts to reduce charging flow to 20 gpm?

A!' The Flow Limit Summator no longer limits flow and FCV-1122 can be manually closed to ailow 20 gpm flo manually closed bo allow 25 gpm flo B. The Flow Limit Summator no ionger limits flow, however, FCV-1122 can only be C. The Flow Limit Summator will prevent FCV-1122 from being closed past 25 gpm flo D. The Flow Limit Summator will prevent FCV-I 122 from being closed past 30 gpm flo Surry Nuclear Plant 2004-301 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 chargin Therefore FCV-1122 can be closed manually to any valu B. Incorrect because when the Charging Flow Controller is in MANUAL, the Flow Limit Summator no longer limits the maximum and minimum values of chargin Distractor is incorrect because FCV-1122 may be manually closed to any value:

even below 25 gprn flow. Distractor is plausibe because candidate may not know that FCV-1122 may be throttled to any value with controller in MANUA C. 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 flew, 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 MANUA 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 Summalor does not function with controller in MANUA Pressurizer Level Control K6.06: Knowledge of the effect of a loss or malfunction on the following will have on the PZR IC§: Correlation of demand signal indication on charging pump flow valve controller to the valve positio Surry Nuclear Plant 2004-301 SRO lnital Exam

? 3. 012A4.04

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TCOLD TEIv!l'/'/C/A

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Which ONE of the following will occur if "A" Loop Narrow Range Tcold fails low while the reactor is at f OO%?

A. Rod Insertion Limit Low and Extra Low alarms will be receive. Ch 1 OTDT setpoint will decreas C I "A" Loop OP and OT Delta T Protection Bistables will tri D. The Tavg / Tref Deviation alarm will be receive Surry References:

NB-93.3-LP-2, Deita T / Tavg Instrumentation System, Rev. 9 ND-93.3-LP-3, Rod Control System, Rev. 14 Distractor Analysis:

A. Incorrect because failed Tcoid is filtered out by Median Signal Selecto. Incorrect because OTBT setpoint will actually increas C. Correct because Tcold is fed directly to the RPS even when failed lo D. Incorrect because failed Tcold is filtered out by Median Signal Selecto (312 Reactor Protection System A4.04: Ability to manually operate and/or monitor in the control room: Bistables, trips, resets, and test switche..

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Surry Nuclear Plant 2004-301 SRO lnital Exam 1 4. 012KI.OS 001/2/U.-WSAC/MEM

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

A. Tripping the reactor trip and bypass breaker shunt coil B, Tripping the reactor trip and bypass breakers UV coil C. Tripping the rod drive MG set output breakers D I Tripping the rod drive MG set supply breaker Slirry References:

NB-93.3-LP-17, Anticipatory Mitigating System Actuating Circuitry, Rev. 10 Distractor Analysis:

A. Incorrect because this does not occu B. Incorrect because this does not occu C. Incorrect because this does not occu D. Correct becasue this is as stated in NB-93.3-LP-17, Rev. 1 Reactor Protection System K1.05: Knowledge of the physical connections and I or cause-effect relationships between the RPS and the following systems: ESFAS

Surry Nuclear Plant 2004-301 SWO lnital Exam

~~~~~~~~~~~~~~~~~~ 15. 013A3.02 001/2/l/SAEEI'Y

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Which ONE of the following correctly states the automatic actions that would occur given a Unit 1 Low Pressurizer Pressure Safety Injection Signal being present for 5 minutes?

A. Hydrogen Analyzer Heat Tracing energizes AND Containment Vacuum Pumps tri ~

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B. Pressurizer Liquid Sample (SS-TV-10OA) receives a close signal AND Motor Driven ~

AFW Pumps start after a 45 second time dela C! Accumulator Nitrogen Relief Lines (SI-TV-1018,5) receive a close signal AND

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Primary Drain Transfer Tank Vents (VG-TV-109NB) receive a close signa D. Main Steam Trip Valves (MS-BV-1QIINB/C) receive a close signal AND Seal Water Return Valve (MOV-381) receive a close signa Sur9 References:

ND-91 -LP-2, Safety Injection System Description, Rev. 16 ND-91 -bP-2, Safety Injection System Operations, Rev. 15 P&ID 1 1448-FM-068A, FlowNalve Operating Numbers Diagram Feedwater System Surry Power Station Unit 1, Rev. 57 Distractor Analysis:

A. Incorrect because SI signal must be present for 8 minutes for heat trace to energiz.

Incorrect because MDAFW Pump starts after 50 sec dela C. Correct because both get a close signal on any SI Signa B. Incorrect because MSTVs only get a close signal on a High %team Flow SI Signa Engineered Safety Features Actuation 83.02: Ability to monitor automatic operation of the ESFAS including: Operation of actuated equipmen Surly Nuclear Plant 2004-301 SRO lnital Exam Which ONE of the following could occur if ES-1.4, Transfer to Hot Leg Recirculation, is performed 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months after the start of a Large Cold Leg Break LBCA?

A. Debris from the In-Core sump could block coolant flow by blocking the lower core plat B. Reflux cooling could be lost due to boron precipitation in the hot leg nozzle C. Fouling of core heat transfer surfaces due to the dilution of boric aci B I Reduction in size of the incore coolant flow channels due to boron precipitatio Sur9 References:

ND-95.3-LP-11, ES-1.4, Transfer To Hot Leg Recirculation, Rev. 8 E§-I.4, Transfer To Hot Leg Recirculation, Rev. 4 Bistractor Analysis:

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

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pump B. Incorrect because boron DreciDitation is a concern in the core. not the hut leq. Incorrect because foulin6 of cbre heat transfer surfaces is a result of boron precipitation, not dilutio B. Correct because boron precipitation is a concern when boil-off continues and when core temperature decreases. The standard time for transfer to hot leg recirc is 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , not 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months , as stated in the ste Engineered Safety Features Actuation K3.01: Knowledge of the effect that a loss or malfunction of the ESFAS will have on the following: Fuel Surly Requal Bank Question #299

Surry Nuclear Plant 2004-301 SRO lnital Exam

I The following Unit 1 conditions exist:

- A Small Break LOCA has occured

- Automatic Safety Injection has occurred

- 1 G O, Reactor Trip or Safety Injection, has been implemented

- The CRO observes the Rod Position Indication as displaying Control Rods on the bottom of the reactor core. with the exception of three Control Rod Which ONE of the following actions is procedurally required as a result of this finding by the CRO?

A: Continue with 1-E-0, Reactor Trip or Safety Injectio B. Emergency borate while proceeding through 1 - E O, Reactor Trip or Safety Injectio C. Manually insert control rods while proceeding through I-EO, Reactor Trip or Safety Injectio B. Go directly to 1-FR-S.1, Response to Nuclear Power Generation / ATWS, Step ~

Surry References:

1-FR-S.1, Response to Nuclear Power Generation / ATWS, Rev. 18 1 - E O, 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 B. Incorrect because if emergency boration is needed, it will be directed by FW-S C. hcorrect because if manual rod insertion is needed, it will be directed by FR-B. Incorrect because FR-S.1 should only be entered as directed by E O (or if E-0 has usag been completed then an Orange or Red path).

014 Rod Position Indication A2.05: Ability to (a) predict the impacts of the following malfunctions or 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 Tri Surry Ibb Bank Question #lo37

Sur9 Nuclear Plant 2004-301 SRO Inital 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 foliowing 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 failur B. Control Rods step out at 72 steps per minute. The Reactor Protection System At-Power Trips are enabled due to the N-41 failur CY Control Rods do not move. The Reactor Protection System At-Power Trips are enabled due Bo the PT-446 failur D. Control Rods step out at 72 steps per minute. The Reactor Protection System At-Power Trips are enabled due to the PT-446 failur Surry References:

ND-93.3-LP-16, Permissive/Bypass/Trip Status bights, Rev. 8 Sur9 Simulator Malfunction Cause and Effects, Rev. 6, Malfunction MMS-14 NB-93.2-LP-4, Power Range Nk., Rev. 16 Surry Simulator Malfunction Cause and Effects, Rev. 6, Malfunction MNL-10 Distractor Analysis:

A. Incorrect because 214 Nis must be above 10% to enable At-Power Trip B. Incorrect because 2/4 NBs must be above 10% to enable At-Power Trips and rods will not mov C. Correct because rods are ifl MANUAL at 5% power and will not move (P-2 prevents movement). Ai-Power trips are enabled when 1 of 2 Pimp channels goes above 10%.

D. Incorrect because rods are in MANUAL at 5% and will not mov Nuclear Instrumentation K4.07: Knowledge of NIS design feature(s) and / or interlock($) provide for the following: Permissive Surry Nuclear Plant 2804-301 SRO lnital Exam 19. OIGA4.01 OO1/2/2/PIMF'

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IMPEI.SE/C/A 2.9/218/s/SRo431)lflu~~~~/S1~~K~~~~~~~-.

The following condition exists:

- Unit 1 at 100% reactor power

- All systems and equipment functions as designed

- All protection channel Ill'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 Dumps would all ope C. FRVs would control SG level at no load leve D I MOV-CP-100, Condensate Polishing Building Bypass Valve, would ope Surry References:

ND-93.3-LP-17, Anticipatory Mitigating System Actuating Circuitry (AMSAC), Rev. 10 ND-93.3-LP-9, Steam Dump Control System, Rev. 10 NB-93.3-LP-8, SG Water Level Control System, Rev. 6 Distractor Analysis:

A. Incorrect because this would occur after 360 second B. Incorrect because Channel 811 is selecte C. incorrect because Channel 111 is selecte 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 rejectio Non-Nuclear Instrumentation A4.01: Ability to manually operate and / or monitor in the control room: NNI channel select control Surry Rsqual Bank Question #279

Surry Nuclear Plant 2004-301 SRQ lnital Exam

___ 20. 022AK1 01 00111111'RCP SEN SICIA 2 8/? 2INISKO430lflUMAEIISDR

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The following Unit 7 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:

- No. 1 Seal Water Outlet Temperatures are 225 OF

- Lower Seal Water Bearing Temperatures are 220 O F 1-AP-9.02, Loss of RCP Seal Coolin 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. 1 Seal Water Outlet Temperatures are too hig Seal cooling should be restored by cooling the RCS using natural circulatio B. 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 hig Seal cooling should be restored by cooling the RCS using natural circulatio C. Slowly establish seal injection flow to minimize RCP thermaB stresses, followed by slowly introducing component cooling flow to the thermai barrier heat exchanger to limit introduction of steam into the CC syste BY 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 stresse __

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Surry Nuclear Plant 2004-301 SRO lnital Exam Surry References:

1-AP-9.02, LQSS of RCP Seal Cooling, Rev. ND-88.1 -LP-6, Reactor Coolant Pumps, Rev. 1 Bistractor Analysis:

A. Incorrect because A$-9.02 (Caution page 7) states if No. 1 Seal Water Outlet Temp is > 235 T-then Seal lnj and CCW to Thermal Barrier H.X. should not be restore Instead N.C. should be used to cool the seal 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 seal C. Incorrect because CC flow should be established prior to seal injection flo D. Correct as stated in 1-AP-9.02 NOTE prior to step 4 and CAUTIONS prior to steps 9 B. Incorrect because AP-9.02 (Caution page 7) states if Lower Seal Water Bearing and 1 Loss of Rx Coolant Makeup AK1.01: Knowledge of the operational implications of the following concepts as they apply to boss of Reactor Coolant Pump Makeup: Consequences of thermal shock to RCP seal Surry Nuclear Plant 2004-301 SRO Inital Exam 21. 022W 4 22 00112/1/SAFETY

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R:N? IONS/C/A 3 014 OIB/SR04301/K/M~~~lSDR Unit 1 has tripped and Safety Injection has actuated due to a Large Break Loss of Coolant Accident (LBCA).

Many complications have occurre The crew has exited E-0, Reactor Trip or Safety Injection. The Shift Technical Advisor has started to monitor Critical Safety Function Status Trees and reports:

- Subcnticality ~ 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-S.1, Response to Nuclear Power Generation/ATWS, based on Subcriticality Orange Pat B. FR-H.1, Response to Secondary Heat Sink, based on Heat Sink Yellow Pat C. FR-6.1, Response to Inadequate Core Cooling, based on Core Cooling Orange Pat B I FR-Z.1, Response to High Containment Pressure, based on Containment Red Pat Surry Nuclear Plant 2504-306 SRO lnital Exam Surry References:

ND-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 Pat B. Incorrect based on the rules of use for safety function status trees (NB-95.3-LP-26 Page 15). The Heat Sink Yellow Path does not take priority over Containment Red Pat C. Incorrect based on the rules of use for safety function status trees (ND-95.3-LP-26 Page 15). Cere Cooling Orange Path does not take priority over Containment Red Pat D. Correct based on the rules of use for safety function status trees (NB-95.3-LP-26 Page 15). The containment Red Path takes priority over the other paths. Only knowledge of safety function priority rules arc? needed to answer this questio Containment Cooling G2.4.22: Knowledge of the bases for prioritizing safety functions during abnormal and emergency operation Turkey Point Bank Question TP03301

Surly Nuclear Plant 2004-301 SRO lnital Exam 22. 022.K3.02

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001/2/1/C:ONTAIN,ME,IIT

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3.0/3.3/Tl/SR0430lflUMAB/SDR

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Unit 2 is operating at 100% power with Chilled CC in service to containmen CD-REF-lA, Unit 2 Turbine Building Chiller Unit, trips due to a faul Which ONE of the following describes the effect on Unit 2 containment parameters?

A. Indicated partial pressure will increase. Containment temperature will decreas B. Indicated partial pressure will increase. Containment temperature will increas C. Indicated partial pressure will decrease. Containment temperature will decreas Indicated partial pressure will decrease. Containment temperature will increas ~

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Surly (Utility should add noun names to equipment in the stem.)

References:

ND-88.5-LP-1, Component Cooling, Rev. 19 Distractor Analysis:

A. Incorrect because partial pressure will decrease due to loss of chilled C B. Incorrect because partial pressure will decrease due to loss of chilled C C. Incorrect because containment temperature will increase due to a loss of chilled D. Correct because partial pressure will decrease and containment temperature will C increase due to a loss of chilled C Bank 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 Reading Surry Nuclear Plant 2004-301 SRO lnital Exam 23. 026A2.07 001/2/1/CONTAiNMENl SPRAYIC ~

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3.6!3.9/N/SR043Oi~MAB/SDR~

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The following Unit 1 conditions exist:

- A Large Break LOCA has occurred inside containment

- Safety Injection has actuated

~ Containment Pressure peaked at 28 psia

- Current Containment Pressure is 15.8 psia

- "2A" and "2B" Outside flecirculation Spray Pumps are operating

- "la" Inside Recirculation Spray Pump is operating

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

- lAE7, fl§ PP l A VIB, annunciates and the alarm cannot be cleared Which ONE of the following states the correct operator action for these conditions?

A. Secure Inside Recirculation Spray Pump "1A" using the handswitch in the control roo :

Place the Inside Recirc Spray Pump 1A in PTL, then secure Inside Recirculation Spray Pump "1A" Iocally at the breaker (14H4).

C. Reset CLS, then place the handswitch for Inside Recirculation Spray Pump "1A" in PT D. Allow Inside Recirculation Spray Pump "IA" to operate, but monitor vibrations closel Surry Nuclear Plant 2004-301 SWO lnital Exam Surry References:

ND-91 -LP-5, Containment Spray System, Rev. 13 ND-91 -LP-6, Recirculation Spray System, Rev. 9 I-RM-AS, RS/SW HX A ALEWT/FAAILURE, Rev. 5 Bistractor Analysis:

A. Incorrect because with CbS present, the handswitch in the control room cannot be used to secure the pump. Containment pressure must be less than 12 psia to reset CLS. Pressure currently is 15.8 psi. Correct because local operation of the breaker will stop the pump. In addition, the ARP will have the operator place the handswitch in FTL, but the Iesson plan (ND-91-LP-6 Page 6) states that the pump cannot be secured from the control room wit& CLS present. Furthermore, the ARP gives guidance to secure the distressed pump as long as two other RS Pumps are operating. The stem states that two other pumps are operating ("2A and "28").

C. Incorrect because the CLS cannot be reset until containment pressure is less than 12 psi D. Incorrect because 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 "28").

026 Containment Spray A2.07: Ability to (a) predict the impacts of the following malfunctions or operations on the CSS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequnces of those malfunctions or operations: boss 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) limi Note:

The AWP states that high vibration alarms may be caused by cavitation of the pum Cavitation could be caused by high water temp, low water level, et Surry Nuclear Plant 2004-301 SRO lnital Exam

~~~~~1 24. 026hK3.02 001/1/1/~:CW S.4FFI'Y

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INJECTION/MEM 3.5/3.7/hllSR04301IRIMAH/SDR

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A High Steam Flow Safety Injection Signal is receive Which ONE of the following correctly describes the response of the Component Cooling Water System components?

A! TV-CC-103A and Ei (CC Isolation Valves from RHR) close and TV-CC-1 10A, B, and C (Reactor Cont Air Recirc Cooler CC Outlet Flow Outside Trip Valve) remain as-i B. TV-CC-1 Q9A and 4 (CC Isolation Vdves from RHR) remain as-is and TV-CC-1 10A, B, and 6 (Reactor Cont Air Recirc Cooler CC Outlet Flow Outside Trip Valve)

remain as-i C. TV-CC109A and B (CC Isolation Valves from RHR) close and TV-CC-I 108, B, and C (Reactor Cont Air Recirc Cooler CC Outlet Flow Outside Trip Valve) clos D. TV-CC-1098 and E3 (CC Isolation Valves from RHR) remain as-is and TV-CC-1 1014, B, and C (Reactor Cont Air Recirc Cooler CC Outlet Flow Outside Trip Valve) clos Surry References:

88-05-01, Component Cooling Water System, Rev. 19 Distractor Analysis:

A. Correct because lesson plan states CC-IO9 closes on Phase I and 110 closes on B. Incorrect because lesson plan states CC-109 closes on Phase I and 110 only closes 6. lncorrect because lesson plan states CC-IO9 closes on Phase I and 11 0 only D. lncorrect because lesson plan states CC-109 closes on Phase I and 11 0 closes on Phase Ill isolatio on Phase Ill isolatio closes on Phase 111 isolatio Phase Ill isolat Loss of Component Cooling AK3.02: Knowledge of the reasons for the following responses as they apply to Loss of Cooling Water: The automatic actions (alignments) within the CCWS resulting from the actuation of the ESFA The loss of CCW occurs in pari of the system due to the ESFAS isolation of TC-CC-I 09A/ Surry Nuclear Plant 2004-301 SWO lnital Exam I

25. -~

026G2 4 46 0011211lC'ONTAINMkN~ SPKAYICIA 3 SI3 6/?\\rlS~/RIMAI3lSDR

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The following Unit 4 conditions exist:

- A barge Break LOCA occurred 45 minutes ago

- Safety Injection has actuated

- Containment Pressure peaked at 27 psia

- RCS subcooling is 0 OF

- Steam Generator bevels are 22% and slowly rising

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RWST emptied while performing ES-1.3, Transfer to Cold keg Recirculation

- S-1.3, Transfer to Cold beg Recirculation. has been completed and the crew has

- All equipment operated normally Which ONE of the following alarms is consistent given the above plant conditions?

transitioned back to 2-1, Loss of Reactor or Secondary Coolant A. 1 -A?, HI-HI CTMT PRESS CLS CH-1 B.' 1B-BI, CS PP 1A LOCKOUYOROLTRIP C. 1A-D7, RS PP 1A LOCKOUT OW OL TRIP D. 1 B-F6, CTMT INSP AIR HDW LO PRESS

Sur9 Nuclear Plant 2004-301 SRO lnital Exam Sur9 References:

1 -E-1, Loss of Reactor or Secondary Coolant, Rev. 21 143-1 3. Transfer to Cold Leg Recirculation, Rev. 12 1 E-AI, HI-HI CPMP PRESS CLS CH-1. Rev. 0 1 B-B1, CS PF l A LOCKOUT OR OL TRIP, Rev. 0 1 B-Ffj? CTMT INST AIR HBR LO PRESS, Rev. 1 ND-91-LP-5, Containment Spray System, Wev. 13 ND-91 -LP-6, Recirculation Spray System, Rev. 9 Distractor Analysis:

A. Incorrect because containment pressure is now less than the setpotnt, which is 1A-D7, WS PP 1A LOCKOUT OR Ob TRIP, Rev. 0 known by CLS having been reset. As a part of going to Cold Leg Recirc, CLS and SI must be rese B. Correct because I-ES-1.3 has been completed and the RWST has been emptied; therefore, the CS Pumps would be placed in PTL due to the lack of a suction source (cavitation). Placing the CS Pumps in PTL yields 1 B-I31 for CS Pump 1 C. Incorrect because Inside Recirc Spray Pump 1A would be placed in AUTO when stoppe D. Incorrect because CLS and SI must have been reset prior to completion of 1-ES-I.3 and instrument air would have been restored to containmen Containment Spray G2.4.46: Ability to verify that alarms are consistent with plant condition Surry Nuclear Plant 2004-301 SRO lnital Exam

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26. 027A4.03 00 1/2/2/IODINE/MEM

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Which ONE of the following describes the operation of the Iodine Filtration Fans A. Automatically start on a Hi-Hi CL B. Automatically start on a containment gas high alarm C. Automatica[ly stop on a Hi-Hi CLS signa D I Must be manually started under all condition Surry References:

ND-88.4-LP-6, Containment Ventilation, Rev. 5 Distractor Analysis:

A. Incorrect because fans are only manually operate B. Incorrect because fans are only manually operate C. Incorrect because fans are only manually operate B. Correct because fans are only manually operate Containment Iodine Removal A4.03: Ability to manually operate and/or monitor in the control room: ClRS fan Question Status:

Surry Bank ILT Question #741

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Surry Nuclear Plant 2004-301 SRO Inital Exam The following Unit 1 conditions exist:

- The Reactor is at 100% Powe A malfunction in the Pressurizer Heater Control Circuit has resulted in Proportional Heaters being de-enerize A small amount of leakage in the Pressurizer Auxiliary Spray Valve is occurrin Pressurizer Pressure is 2215 psig and slowly lowerin AP-31 BO, Increasing or Decreasing RCS Pressure, has been entere Which ONE of the following states the correct position of the Normal Pressurizer Sprays, Proportional Heaters, and Backup Heaters, assuming that the Proportional Heaters are now operable?

A! Normal Sprays are OFF (valves closed); Proportional Heaters are ON; Rackup Heaters are O B. Normal Sprays are OFF (valves closed); Proportional Heaters are OFF; Backup Heaters are O C. Normal Sprays are OFF (valves closed); Proportional Heaters are ON; Backup Heaters are OF D. Normal Sprays are ON (valves open); Proportional Heaters are ON; Backup Heaters are O ~ _ _ _

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Surry References:

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

A. Correct because backup heaters are always on, proportional heaters will be on below about 2220 psig, and spray valves will be closed below about 2255 psi (The correct answer was validated on the simulator at 2215 psig.)

B. Incorrect because proportional heaters would be o C. Incorrect because backup heaters would be o D. Incorrect because spray valves would be close 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 heater Surry Nuclear Plant 2004-301 SRO M a l Exam 28. 028Ci2

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2 12 001/2/2/IIYDROCiEN -~

KFCOMBINFK/C'/A 3 O/j 4/N/SKO4Xl/K/MAB/SDK The following Unit 1 conditions exist:

_. 1-PT-37.2, Electric Hydrogen Recombiner, is about to be performed to determine the The plant is at 50% power reference power that would be used in the event that the Wecombiners are used following a LOC Which ONE of the following correctly states 1 -PT-37.2 limitations that are applicable during the performance of this test?

A! Turn Power Adjust potentiometer to obtain 48 KW on the wattmeter and approximately I225 O F heater output. At no time should the heater temperature be allowed to exceed 1400 OF on the highest thermocouple readin B. Turn Power Adjust potentiometer to obtain 48 KW on the wattmeter and approximately 7 000 OF heater output. At no time should the heater temperature be allowed to exceed 1200 O F on the highest thermocoerple readin. Turn Power Adjust potentiometer to obtain 36 KW on the wattmeter and approximately 1225 O F heater output. At no time should the heater temperature be allowed to exceed 1400 O F on the highest thermocouple readin D. Turn Power Adjust potentiometer to obtain 36 KW on the wattmeter and approximately 1080 OF heater output. At no time should the heater temperature be allowed to exceed 1200 OF on the highest thermocouple readin Surry Keferences:

1 -PT-37.2, Electric Hydrogen Recombiner, Rev. 9 Distractor Analysis:

A. Correct because as stated in the procedure temperature must remain less than B. Incorrect because 48KW does not equate to 1000 OF heater output and the temp e. Incorrect because 36 KW does not equate to 1225 'F heater outpu B. Incorrect because as stated in the procedure temperature must remain less than 1400 O F at all times and the potentiometer shall be adjusted to 48 K limit is 1400 O O Hydrogen Recombiner and Purge Control G2.2.12: Knowledge of surveillance procedure Surry Nuclear Plant 2004-301 SWO tnital Exam

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29.029hK3 09 001/1/1/ATWS LWIMEM 3 714 0MSK04301RvMNilSD~

If the Emergency Boration Flowpath is not available, which ONE of the following describes the reason why charging pump suctions are manually aligned to the RWST during an ATWS vice manually initiating a Safety Injection?

A. Prompt operator action will ensure the most direct method of borating into the RCS and manual alignment of charging pump suction to the RWST prevents compounding the problem by charging the RCS solid via Safety Injectio and initiation of SI would reduce the possible paths for emergency boration and add to an RCS overpressure condition if one exist B. Prompt operator action will ensure the most direct method of borating into the RCS 6. 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 RC D I 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 RC ~

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Surry Nuclear Plant 2804-301 SRO lnital Exam Surry 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 of maintaining a secondary heal sink because MFW pumps will trip upon SI initiatio B. Incorrect because the concern with initiating SI is not creating a high WCS pressure condition, but with reducing the probability of maintaining a secondary heat sink because MFW pumps will trip upon SI initiatio C. Incorrect because mantial initiation would not delay addition of borated water. The concern is with reducing the probability of maintaining a seondary heat sink because M W pumps will drip upon SI initiatio.

Correct because per MD-95.3-bP-36-DRR, FR-S.l Response to Nuclear Power Generation / ATWS, both of these statements accurately reflect the basis for Step ATWS EK3.09: Knowledge of the reasons for the following responses as they apply to the ATWS: Opening centrifugal charging pump suction valves from RWS Modified EbT Bank Question # 3390

Surry Nuclear Plant 2004-301 SRO lnital Exam 30. 032AAI 01 001/1/2/SOI!RCE

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LVTERMEDIATEICIA 3 __

113 4/B/SK04301/IUMhn/SDR

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The following conditions exists:

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

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Reactur tripped at 1405 hours0.0163 days 0.39 hours 0.00232 weeks 5.346025e-4 months

- All Rod Bottom Lights are lit

- N-35 reading is 2 x lo-amps

- N-36 reading is 4 x 1V amps

- Source Range is not energized

- Power level prior to trip was 9Q%

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

A. When both IR channels read e 5 x lo-amps, verify source range channels energize.

Place the source range trip bypass switches in the NORMAL positio CY Energize the source range channels by depressing the source range manual reset pushbutton B. Transfer NR-45 to one suurce range and one intermediate range channe ~

Surry References:

ND-93.2-LP-3, Intermediate Range Nls, Rev. 1 Distractor Analysis:

A. incorrect because SR energizes at 2/2 IR 6 5 x lo-amp.

Incorrect because SR should already be energized in the NORMAL position and this action would not energize the S C. Correct because BR are under-compensated and SW must be manually energize B. Incorrect because SR should both be energize Loss of Source Range NI AA1.01: Ability to operate and / or monitor the following as they apply to loss of Source range nuclear instrumentation: Manual restoration of powe Surry Nuclear Plant 2004-381 SRO lnital Exam 31. 033AA2.04 001/1/ZOVERLAP NYhlEM

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The following Unit 1 conditions exist:

- Critical approach has just been complete Reactor is stable at the Point of Adding Hea One Intermediate Range (BR) Nuclear Instrument (NI) is suspected of displaying inaccurate indication 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 IR NI is in fact falsely indicating?

A. 18 = 2.5 x 10.' Amps; PR between 0.2 and 4 %

B I 18 = 2.0 x 1 0-6 Amps; PR between 0.2 and 1 %

6. IR = 1.O x Amps; PW < 0.2 %

83. iR = 1.0 x 4 U 5 Amps; PR < 0.2 %

Surry References:

ND-93.2-LP-4, Power Range Nls, Rev. 16 1 -GOP-1.4, Unit Startup, HSD to 2% Reactor Power, Rev. 29 Distractor Analysis:

A. Incorrect because 2.5 x l o 8 Amps is about where critical data is taken (too low).

E3. Correct based on above two references: ND-93.2-LP-4 (HR-4.3) & 1 -GOP- (Page 29 CAUTION).

C. Incorrect because 1.O x D. Incorrect because 1.O x about 2% powe Amps is about where critical data is taken (too low).

Amps is above the PBAH and should correspond to 033 Loss of Intermediate Range NI AA2.04: Ability to determine and interpret the following as they apply to the loss of intermediate range nuclear instrumentation: Satisfactoy overlap between source-range, intermediate-range, and power-range instrumentatio Unit 4 is in a refueling outage when the following events occur:

- Purge Isolation Valves (MOV-VS-IQOA, B, C, and D) Close

- Containment Instrument Air Suction Valves (TV-IA-101 N B ) Close Which ONE of the following radiation monitors could have caused these actions?

A. Process Vent Particulate and Gas Monitors (RM-RI-101 / 102)

B. RM-161 (Containment High Range Gamma Monitor)

6:' RM-162 (Manipulator Crane Monitor)

D. WM-163 (Reactor Containment Area Monitor)

Surry References:

NB-93.5-LP-1, Pre-TMI Radiation Monitoring System, Rev. 5 Bistractor Analysis:

A. Incorrect because WM-RI-I01 / 102 do not cause these action B. Incorrect because RM-161 does not cause these action C. Correct per ND-93.5-LP-D. Incorrect because RM-163 does not cause these action Fuel Handling Equipment A4.01: Ability to manually operate and / or monitor in the control room: Radiation Level Surry Nuclear Plant 2004-301 SRO lnital Exam The following Unit 1 conditions exist:

- Plant is stable at 75% Power

- "A" SG Steam Line FT-MS-475 (CH-IV) is selected for Steam Generator Level control

- "A" SG Steam bine FT-MS-475 (CH-Ill) fails high Which ONE of the following correctly describes the impact on the "A" Steam Generator Level control?

A. Feedwater Regulating Valve opens because indicated steam flow is greater than indicated feedwater flo B! Feedwater Regulating VaDve does not move as a result of the failur. Feedwater Regulating Valve closes because the pressure transmitter is overcompensating for densit. Feedwater Regulating Valve opens to reduce the level error created by the failur Surry References:

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

A. Incorrect because FT-MS-475 does not compensate steam flow for FT-MS-47 B. Correct because PT-MS-475 does not compensate steam flow for FT-MS-47 C. Incorrect because PT-MS-475 does not compensate steam flow for FP-MS-47 D. Incorrect because PT-MS-495 does not compensate steam flow for FT-MS-47 Steam Generator A3.01: Ability to monitor automatic operation of the S/G including: S/G water level contro Surry Nuclear Plant 2004-301 SRO M a l Exam 34. 03REK3.09 001/1/1/SAFF.TY INJECTION/MEM

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4.1/-1.5~/SK04301/K/MAH/SIIR

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Which ONE of the following is correct regarding safety injection termination during a steam generator tube rupture event?

Safety Injection termination...

I AY may occur with total AFW flow less than 350 gpm as long as 350 gpm is availabl. may occur with Pressurizer level less than 22% as long as level is increasin C. may not occur with a void in the reactor head due to presenting RCS pressure control problem. may not occur with a void in the reactor head due to presenting RCS level control problem Surry References:

ND-95.3-LP-13, E-3 Steam Generator Tube Rupture, Rev. 11 1-E-3, Steam Generator Tube Rupture, Rev. 19 Distractor Anaiysis:

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 350 gpm, but 350 gpm must still be available to that SG. If sufficient flow is available, then SI termination criteria is considered to be met (ND-95.3-LP-13).

3. Incorrect because pressurizer level must be greater than 22% to meet the SI termination criteri. 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 Ievei 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 ECC Surry Nuclear Plant 2004-301 SRO lnital Exam 35 039A1 09 001/2/1/hINN STEAM KAUIA'I IC)N/C/A 2 5/2 7/B/SR04301/RIMhR/SDR

r-With Unit 1 at 108% power, the Condenser Air jector and Main Steam Line Radiation Monitor alarms are recieved. The Condenser Air Ejector Radiation Monitor reads 900 cpm (ALERT and HIGH alarms are in) while local Main Steam NRC Radiation Monitors read "A".03 mr/hr, and "B".01 rndhr, and "C".01 mr/hr. The Team has implemented 1-AP-16.00, Excessive RCS Leakage, and the RCS leak rate is determined to be 60 gp Which ONE of the following describes the actions required?

A. Verify automatic Condenser Air Ejector divert to Containment, intiate 1 -AQ-24.00 (Minor SG Tube Leak), manually trip the reactor and go to 1-E-0 (Reactor Trip or Safety Injection).

Containment, manually trip the reactor and initiate SI, Go to 1 - L O.

C. Verify automatic Condenser Air Ejector divert to Containment, initiate 1 -AP-24.01 (barge Steam Generator Tube Leak), verify letdown isolated, and commence a normal Unit shutdown BAW GOP B. Verify automatic SGBD TV trip isolation and Condenser Air Ejector divert to BY Verify automatic Condenser Air Ejector divert to Containment, manually trip the reactor and initiate 1-E-0 (Reactor Trip or Safety Injection), and go to 1-AP-24.01 (Large Steam Generator Tube beak).

Surry Nuclear Plant 2004-301 SRO lnital Exam Surry References:

ND-89.3-LP-2, Main Condensate System, Rev. 16 ND-93.5-LP-3, Post-TMI Radiation Monitoring System, Rev. 6 1-AQ-16.00, Excessive RCS Leakage, Rev. 11 1 -AP-24.00, Minor SG Tube beak, Rev. 8 1 -AP-24.01, barge Steam Generator Tube Leak, Rev. 1 1 Distractor Analysis:

Incorrect because 60 gpm leakage is more than minor. A$-24.01 should be entered for a large steam generator tube lea Incorrect because SI should not be initiate Incorrect because the reactor must be manually tripped with leakage greater than 50 gp Correct because air electors will divert to containment on an air elector hiah radiation, AP-24.01 should be entered due to 60 gpm leak rate 4 t h air ejector high radiation, and E-0 should be entered following a manual reactor tri Main and Reheat Steam A I.09: Ability to predict and / OF monitor changes in parameters (to prevent exceeding design limits) associated with operating the MRSS controls including: Main steam kine radiation monitor Surry Nuclear Plant 2004-301 SRO lnital Exam 3 ~

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~ 054G2. ~

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1 001/1/I/ALARMS RODS/C/A 3.3/3.~B/SR0430IIRIMAB/SDR ~~~

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.OO,

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

- Tavg is within 3 O F of Tref

- Annunciator 1G-G8, ROB BANK B LO LIMIT, has annunciated Which ONE of the following is the correct response to the given plant conditions?

occurred

"LOSS of Main Feedwater Flow" A. Shutdown margin is not sufficient for the given plant conditions and operators should emergency borate to regain the required shutdown margi BY The operator has driven rods in too far for the existing boron concentration and should borate from the Boric Acid Tank. Shutdown margin is not sufficient for the given plant conditions and operators should trip the Reactor and go to I-E-0, Reactor Trip or Safety Injectio B. The turbine load has decreased too far and the operator should raise turbine loa Surry Nuclear Plant 2004-301 SRO Dnital Exam Surry References:

NB-89.3-LP-3, Main Feedwater System, Rev. 12 ND-95.1-LP-4, Loss of Feedwater; Rev. 3 1-AP-21.00, Loss of Main Feedwater Flow, Rev. 5 1G-68, ROD BANK D LO LIMIT, Rev. 0 Distractor Analysis:

A. Incorrect because (1) not enough information is given to make the determination that SBM is insufficient, and (2) even if SDM is not above that which is required, emergency boration would not be the preferred method for regaining the required SBM. This is clearly the wrong method fer boration because xenon is building in and only small borations would be desired to withdraw rods to clear the alar alarm. Boration 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 alar 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 limit B. 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 nonconsetvtaive to add positive reactivity via the turbine during a transient condition such as described in the ste B. Correct because rods being within 10 steps of its insertion limit would cause the C. Incorrect 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 instruction Bank Question from 2003 Farley Exam (Farley WA was 054612.2.20).

Sur9 Nuclear Plant 2004-301 SRO lnital Exam 37. 055EK I.O 1 09 1/ I/ l/B ATTERYlClA 3.3/3.7/B/SR0430 MUM -~

AIilSDR The following plant conditions exist:

- A loss of all AC power has occurre ~

Operators have implemented ECA-0.0, Loss of All AC Powe Attempts to regain AC power have faile Operators are performing ECA-0.0, Step 28, "Check DC Bus Loads"

- Annunciator J-F-6, TURB GEAR ZERO SPEED, is lit Which ONE of the following should be performed to lower the Black Battery discharge rate by the largest amount per ECA-O.O?

A. Secure Air Side Seal Oil Pump onl B. Secure Air Side Seal Oil Pump and Emergency Turbine Lube Oil Pum C. Secure Air Side Seal Oil Backup Pump onl B I Secure Air Side Seal Oil Backup Pump and Emergency Turbine Lube Oil Pum Surry Nuclear Plant 2004-301 SRO Inital Exam Surry References:

ND-90.3-LP-6, 125 Vdc Distribution, Rev. 10 ECA-0.0, Loss of All AC Power, Rev. 21 1 J-F6, TURB GEAR ZERO SPEED, Rev. 1 Bistractor 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-B. Incorrect bemuse 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 BC Loads, or may not know what actions are permitted by ECA-C. Incorrect EGA-0.Q will direct the securing of both Air Side Seal Oil Backup Pump (ASSOBUP) and Emergency Turbine Lube Oil Pump, not just ASSOBU Plausible because the applicant may not know that there is more than one pump to secure to conserve Black Batterie D. Correct because per ECA-0.0 step 28 and Basis for this step in NB-95.03-LP-17, 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).

Surry IbT 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 capacit Surry Nuclear Plant 2004-301 SRO M a l Exam

~~~~~~~ 38. 056.42.04

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~~ 001/2/11CONDENSATE

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PUMP/C/A 2.6/?.fi/NISK03301iK/M*~~lS~R

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The following Unit 1 conditions exist:

- Two Main Feedwater Pumps are operating Reactor Power = 65%

Condensate Pumps 1 -CN-P-lA and B are operating

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

- Condensate Pump 1-CN-P-lA trips and cannot he restarted

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

- Steam Generator bevels are slowly lowering 1 H-F8, FW PP SUCT HDR LO PRESS: is in alarm Which ONE of the following is the correct operator action?

A! Enter 1-AP-27.OO, Loss of Main Feedwater FLOW, and reduce turbine load to match steam flow and feedwater flo B! Manually trip the Reactor and enter E-0, Reactor Trip or Safety Injectio C. Secure one of the operating Main Feedwater Pumps and monitor the operating Main Feedwater Pump Suction Pressur D. Enter 1-AP-21 BO, Loss of Main Feedwater Flow, and start a second HP Brain Pum Surry References:

ND-89.3-LP-2, Main Condensate System, Rev. 16 NB-89.3-bP-3, Main Feedwater System, Rev. 12 NB-95.1 4P-4; Loss of Feedwater, Rev. 3 1-AP-21 DO, Loss of Main Feedwater Flow, Rev. 5 1 H-FB, FW PF SUCT HDR LO PRESS, Rev. 0 1 H-G8, FW PP DlSCH HBW LO PRESS, Rev. 0 1J-G4, CN PFS DlSCH HBR LO PRESS, Rev. 0 Distractor Analysis:

A. Correct because MFW Pump Low Suction Pressure and Discharge Pressure Alarms are entry conditions into AP-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 MFW Pump suctionldischarge pressur B. Incorrect because no trip criteria are met and AP-24.00 directs power reductio C. Incorrect because tripping a MFW Pump will not alleviate the issue and there is no procedural guidance to trip a MFW Pump. Typically a MFW Pump will be secured at about 40% powe Surry Nuclear Plant 2004-301 SRO lnital Exam D. lncorret because there is no guidance to start a second heater drain pump. The correct response is to lower turbine loa Condensate A2.04: Ability to (a) predict the impacts of the following malfunctions or operations on the Condensate System; and (b) based on those predictions, use procedures to correct. control, or mitigate the consequences of those malfunctions or operations:

boss of condensate pump FACILITY POST EXAM COMMENT:

Facility Comment: Most trainees applied the Management standard for conservative decision making. If reactor trip is imminent, then manually trip the reactor and perform the immediate actions of E-0. The definition of IMMINENT is within one to two hours and continuing deteriorating conditions exis faced with unexpected or uncertain conditions will place the plant in a safe condition and will not hesitate, if necessary, to reduce power or trip the reacto suction pressure at 105 psig and decreasing, can only result if the only running condensate pump is also significantly degraded. This places the plant in a condition not considered in the development of AQ-21.OO, Loss of Main Feedwater Flo The Supervisor of Shift Operations said that he would not hesitate to trip the reactor given the conditions provided in the stem of the questio We ran this scenario on the simulator and were unable to keep the unit online. It resulted in a reactor trip 100% of the tim of 10 trainees chose answer (B).

DNOS-0101, Nuclear Safety and Conservative Decision Making, states Operators The conditions provided in this question, one condensate pump running with feed pump Recommendations: Based on the above information, accept (B) as an alternate correct answe NRC Resolution: Recommendation accepted: the question has two correct answers (A and B).

The NRC concurs that it is reasonable and conservative for an operator to manually trip the reactor with Main Feedwater pump suction pressure at 105 psig and slowly lowering. The stern conditions created the sense that the system conditions were continuing to degrade therefore, a natural assumption was that a reactor trip was imminen Surry Nuciear Plant 2004-301 SRO lnital Exam 39. 056AA1.26 OOllllllDIESEL BREAKEWCIA 2.5/2.6iBISR04301/WM~IS~R-1

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The following conditions exist:

- A Loss of Off-Site Power has occurred

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

- It has been determined that the auto-closure circuit for 15H3, #1 EDG Output

~ When the operator places the sync switch for 15H3 to "QN" he observes 120 volts or 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-o'clock Which ONE of the following actions is necessary prior to closing 15H3?

A. Raise EDG speed until the synchroscope is turning slowly in the fast direction, then close 15H3 at "1 1 o'cloc B. Momentarily press the "field flash" pushbutton, then sync and close 15H C. Raise EDG voltage until the running meter indicates 120 volts, then sync and dose 154 D I No additional action is necessary. Close 15H Surry References:

ND-90.3-LP-1~ Emergency Diesel Generator, Rev. 14 ND-90.3-LP-7, 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 phase B. Incorrect because it will not be possible to synchronize (nor is it necessaty because the bus is dead). Also, field flash Pi? does not need to be pushe C. Incorrect became raising the EDG voltage will not raise running voltage. Incoming voltage is the EDG voltage (not running voltage).

5. 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-kP. pg. 18).

Therefore, all criteria for manually closing the breaker are me Loss of Off-Sits Power AAl.26: Ability to operate and / or monitor the following as they apply to the Off-Site Power: Circuit Breakers

.oss of

Surry Nuclear Plant 2004-301 SRO lnital Exam 40. 0VAK3 01 001/1/I/VITAL ACM'IAJ 114 4IB/SRM301/R.MABISDR

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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 los B. Pressurizer pressure control is los C. Control af Steam Generator Feed Regulating Valve is los BY "8" Reactor Coolant Pump must be stoppe I

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Surry References:

1-AP-10.02, Loss of Vital Bus II, Rev. 9 ND-90.3-LP-5, Vital and Semi-vital Bus Distribution, Rev. 11 Distractor Analysis:

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

the loss of VB II will not preclude a trip if one is neede. Incorrect because Pnr P Controller will transfer to AUTO-HOLD, but MANUAL control is still possible, thus precluding the need for rx tri C. Incorrect because FW-FCV-1488 Flaw Controller will transfer to AUTO-HOLD, but MANUAL control is still possible. thus precluding the need for rx tri. Correct because Component Cooling is lost to the "3" RCP Lube Oil Cooler. RCP Parameters will eventually exceed limits (1-AP-10.02 Att. l), requiring that the RCP be secured following a manual rx tri Loss of Vital AC Inst. Bus AK3.01: Knowledge of the reasons for the following responses as they apply to the LOSS of Vital AC Instrument Bus: Actions contained in EOP for loss of vital ac electrical bu Surry ILT Bank Question #223

Surry Nuclear Plant 2004-301 SRO lnital Exam The following Unit 1 conditons exist:

- 1 K-A8, UPS SYSTEM TROUBLE, annunciates

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

- An operator reports that Battery Charger BC Output for UPS 1A-1 reads 0 amps Which ONE of the following correctly describes the power supply to the associated DC and Vital AC buses?

A. (9C Bus 1A will be supplied by only Battery 1A as indicated by DC Bus voltage slowly trending down over time and Vital AC Buses 1 and I A will he supplied by BUS 1 Hl-B. BC Bus 1A will be supplied by only Battery ? A as indicated by DC Bus voltage slowly trending down over time and Vital AC Buses 1 and 1A will be supplied by BUS 1 H1-C. DC Bus 1A will be supplied by UPS 18-2 as indicated by DC Bus Voltage remaining stable at 125 VBC and the Vital AC Buses 1 and 4A will be supplied by 1H1-BY DC Bus 1 8 will be supplied by UPS 1A-2 as indicated by BC Bus Voltage remaining stable at 125 VDC and the Vital AC Buses 1 and 1A will be supplied by 1H1-Surry Nuclear Plant 2004-301 SRO lnital Exam Surry References:

ND-90.3-LP-5, Vital and Semi-vital Bus Distribution, Rev. 11 ND-98.3-LP-6, 125 VDC Distribution, Rev. 10 1 K-A7, BATT SYSTEM 1 A TROUBLE, Rev. 5 11448-FE-lG, Sheet 1 of 7, 125V BC One Line Diagram - Surry Power Station Unit 1

~

1K-A8, UPS SYSTEM TROUBLE, Rev. 1 Rev. 33 Distractor Analysis:

A. Incorrect because the battery should not be supplying the DC Bus alone. The DC Bus is being supplied by the other UPS from 1Hl-2. Also, vital AC Buses 1 and 1A are being supplied by Bus 1H1-2, which is the alternate AC sourc B. Incorrect because the battery should not be supplying the DC Bus alone. The DC Bus is being supplied by the other UPS from 1 HI - C. Incorrect because the Vital AC Buses I and 1A are being supplied by Bus 1H1-2, which is the alternate AC sourc D. Correct because the other UPS will still be supplying BC Bus 1A-1 and the Alternate AC Source 1 H1-2 will supply Vital A 6 Buses 1 and 1 boss of DC Power AA2.01: Ability to determine and interpret the following as they apply to the loss of BC Power: That a loss of DC Power has occurred; verification that substitute power sources have come on lin Surry Nuclear Plant 2004-301 SRO lnital Exam Which ONE of the following sets of practices should be observed by operators for starting the second Main Feedwater Pump per GOP-I.5 (Unit Startup, 2% Reactor Power to Max Allowable Power) and OP-W-004 (Main Feedwater System Operation)'

A. The second Main Feedwater Pump should be started prior to exceeding 50% powe 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 gp. 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 gp C I The second Main Feedwater Pump should be started prior to exceeding 50% powei 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 heat 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 MQV as the system heat ~

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Surry Nuclear Plant 2004-301 SRO lnital Exam Surry References:

1 -GOQ-1.5, Unit Startup, 2Y0 Reactor Power to Max Allowable Power, Rev. 32 1 -0P-FW-004, Main Feedwater System Operation, Rev. 8 NB-89.3-LP-3, Main Feedwater System, Rev. 12 Bistractor Analysis:

A. Incorrect because recirc should modulate closed at 4000 gpr B. Incorrect because recirc should modulate closed at 4000 gp C. Correct because of NOTE on Pg. 34 of 44 of GOP-I.5 and CAUTION on Pg 12 of 34 of OP-FW-00 D. Incorrect because second feedwater pump should be started prior to !?&)yo powe Main Feedwater A1.03 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 valve Surry Nuclear Plant 2004-301 SRO lnital Exam 4 AA1 01 001/1/2/I IQIJID RAD KEI.E4SE/C/A 3 5/3 5 / ~ l / S R O 4 3 0 l / W M ~ I <

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The following Unit 1 conditions exists:

- A Large Break Loss of Coolant Accident has occurred

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

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

- l-RM-A8, RS/SW HX B ALERWFAILUWE, annunciates

- Reactor Operator notes the RSISW HX B Monitor is trending up, but the Discharge Tunnel Rad Monitor is indicating all EEEEEs with Wed and Yellow Lights kit and Green Light ou Which ONE of the following IS the correct operator response?

A. Ensure no additional releases are in progress and secure R B:' Ensure no additional releases are in progress, and increase radiation monitorin C. Verify ail automatic actions have occurred and reset the Discharge Tunnel Digital Rate Meter and perform a source chec D. Verify all automatic actions have occurred and raise the Discharge Tunnel Monitor set poin Surry Nuclear Plant 2004-301 SRO lnital Exam Surry References:

ND-93.5-LP-1, Pre-TMI Radiation Monitoring System, Rev. 8 1-RM-G7, BISCH TNL ALERT / FAILURE, Rev. 4 I-RM-A8, RSISW HX B AhERT/FAILURE, Rev. 3 Distractor Analysis:

A. Incorrect because the last available RS train should not be secured, as stated in RM-G7 and flM-A8 Caution Statements. Plausible because this is the correct course of action if the other train was availabl B. Correct because the last train of RS 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 Bld Liquid releases are occurrin Additional monitoring is called for by the ARBS due to the fact that the last train of RS should not be secure 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 a reset and source check, which adds to the plausibilit 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 setpoint to be raised to alert operators of worsening conditions. The Discharge Tunnel Monitor has the indications of being failed, therefore adjusting the setpoint is not a success pat Accidental Liquid Radwaste Release AAI.01: Ability to operate and / or monitor the following as they apply to the Accidental Liquid Radwaste Releases: Radioactive-liquid monitor Modified Surry ILT Bank Question #1977

Surry Nuclear Plant 2004-301 SRO lnital Exam 44 061A104 -~

001/2/1/;\\FW COhUkNSAIWMDM

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7 913 9/N/SR04301/R/MAB/SI)K

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4 -CN-TK-1, Emergency Condensate Storage Tank (ECST), is supplying AFW Pumps for Residual Heat Removal via Steam Generators. 1 J-F4, CST 1 18,000 GAL LO LVL, has annunciated. ECST ievel IS 90% and lowerin Which ON of the following is correct regarding refilling of the ECST?

A. Filling shall commence prior to the ECST level reaching 54% (60,000 gallons).

8. Filling may commence after the ECST level drops below 54% (60,000 galions) as C. AFW Pumps must be secured prior to commencing the fill and the ECST must be D: Filling of the ECST shall commence prior to the ECST level reaching 54% (60,000 AFW pumps must be secured prior to commencing the fil long as refill begins within two hours of securing the AFW pump filled within two hour gallons). A W pumps may continue to operate during the refil Surry References:

NB-89.3-LP-4, Auxiliaty Feedwater System, Rev. 19 fd-F4, CST 1 f 0,000 GAL LO IVL, 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 OR ARP ld-F4 Not E3. incorrect because volume must remain above 60,000 gal (54%).

C. Incorrect because AFW pumps do not need to be secured for refil D. Correct based on all three of the above reference Auxiliary Feedwater Al.04: Ability to predict and / or monitor changes in parameters (to prevent exceeding design limits) associated with operating the AFW controls including: AFW source tank leve Surry Nuclear Plant 2004-301

§RQ lnital Exam If a spent fuel assembly is damaged by being dropped in the spent fuel pool, which i

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

A. Spent Fuel Pit Bridge Crane Radiation Monitor and Auxiliary Building Control Victoreen Area Radiation Monitor 5. Ventilation Vent Particulate Radiation Monitor and Auxiliary Building Control Victoreen Area Radiation Monitor CY Spent Fuel Pit Bridge Crane Radiation Monitor and Ventilation Vent Gaseous Radiation Monitor D. Ventilation Vent Gaseous Radiation Monitor and the Liquid Waste Effluent Process Monitor

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Surry (Utility needs to add correct WM equipment numbers.)

References:

ND-93.5-LP-1 I Pre-TMI Radiation Monitoring System, Rev. 8 0-AP-22.00, Fuel Handling Abnormal Conditions, Rev. 7 8 O-RM-54,1-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 Vicloreen Area Radiation Monitor would not show an increased indicatio. Incorrect because the Aux 5ld Control Victoreen Area Radiation Monitor would riot show an increased indicatio C. Correct because both monitors would show an increased indicatio D. Incorrect because a liquid waste process effluent monitor would not see the results of the failed fue O-RM-D3, 1-RM-RI-153 HIGH, Rev. 4 061 ARM System Alarms AA2.01: Ability to determine and interpret the following as they apply to the Area Radiation Monitoring (ARM) System Alarms: ARM panel display Surry Requal Bank Question #I 18

Surry Nuclear Plant 2004-301 SRO lnital Exam 4 ~~ 062Al 01 001!1/?/I~I)G DIESEL!MEM 3 4!3 8ITJISR01301~lAH/SDR The following conditions were noted during the performance of 1 -OPT-EG-001, Number 1 Emergency Diesel Generator Monthly Start Exercise Test:

- The EDG was loaded at a rate of 550 KWlMlN

- The Maximum load attained was 2650 KW

- The Maximum KVAR was 500 KVAR out

- The output voltage was stable at 4300 VAC Which ONE of the following was in violation of the EDG Precautions and Limitations per 1 -OPT-EG-001?

8.' Load Rate B. Maximum Load C. Maximum KVAR out D. Output voltage Surry References:

1-OPT-EG-001, Number 1 Emergency Diesel Generator Monthly Start Exercise Test, Rev. 24 1-OP-EG-001, Number 1 Emergency Diesel Generator, Rev. 17 Distractor Analysis:

A. Correct because the loading rate should not exceed 500 KWlMlN during normal operation. Incorrect because max load rating is 2750 K. Incorrect because max KVAR out is 500 WA B. Incorrect because output voltage should be maintained between 4000 and 4400 VA ~

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062 AC Electrical Distribution Af.01: Ability to predict and I or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ac distribution system controls including:

Significant B/G load limit Surry Nuclear Plant 2004-301 SRQ lnital Exam

~~ 4 A.41 SI6 00 lI1IIISEKVICE

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WATPX/MI!M

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2.9/2.9/B/SRo430lfl~/S~~K~

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The following Unit 1 conditions exist:

- Power = 100%

- 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 Isolation Signal?

A. 1-SW-MQV-102A and E3 (CCHX and SW-P-4 Supply) will close and can only be reopened after 5 minute. l-SW-MQV-102A and 5 (CCHX and SW-P-4 Supply) will go to 25% open and can be fully opened after 5 minute C I 1-SW-MQV-102A and B JCCHX and SW-P-4 Supply) will close and can be reopened when the low level signal is rese fully opened when the low level signal is rese B. 1-SW-MQV-102A and B (CCHX and SW-P-4 Supply) will go 25% open and can be

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Surry (Utility needs to verify technical accuracy and provide any additional reference material (electrical print?)).

References:

ND-89.5-LP-2, Service Water System, Rev. 20 Distractor Analysis:

A. Incorrect because the valves will close, but cannot be re-opened until Canal bow Level 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 CbS cleare B. Incorrect because the valves will go fully close C. Correct because the valves will close, but cannot be re-opened until Canal how Level Isolation Signal is cleared. If the valves would have been closed due to a CLS, then they could be opened affer five minutes without resetting U S.

D. Incorrect because, as states above, the valves will clos boss of Nuclear Svc Water AAl.06: Ability to operate and / or monitor the following as they apply to the Loss of Nuclear Service Water (SWS): Control of flow rates to components cooled by the SW Surry Nuclear Plant 2004-301 SRO lnital Exam Unit 1 was operating at 68% power when the following plant conditions developed:

- 1 K-A7, BATT SYSTEM ? A TROUBLE, alarm annunciates

- "A" SG P O W Indicating Lights are not lit

- MSTV Indicating Lights are not lit

- P O W 1455G/l456 Indicating Lights are not lit

- "A", "D", and "H" 41 60 V Bus 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 trippe B, The turbine will automatically trip. The reactor will automatically trip due to the automatic turbine tri C. The reactor must be manually tripped. The turbine must also be manually trippe D. The reactor will automatically trip. The turbine will not automatically trip and must be manually trippe Surry References:

ND-90.3-LP-6, 125 Vdc Distribution, Rev. 10 Distractor Analysis:

A. Correct because the reactor will automatically trip on loss of voltage to the "PI" 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-B. Incorrect because the reactor wilt automatically trip due to loss of voltage to the "A" RTB UV coil due to the loss of the " A dc Bu C. Incorrect 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-B. Incorrect because the turbine does not need to 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 level DC Electrical Distribution 84.01 : Ability to manually operate and / or monitor in the control room: Major breakers and control power fuse Surry Nuclear Plant 2004-301 SRO lnital Exam The following plant conditions exist:

- Bus 131 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

3's Air Compressors?

remains stable A. Bus 1J1 remained the power supply throughout the seven second voltage dro B. Six seconds after the voltage dropped on Bus 1 J1, Bus 2J1-1 became the power supply. Bus 2J1-1 will remain the power supply until manually transferred back to Bus 1J1 C! Six seconds after the voltage dropped on Bus 1J1, Bus 2J1-1 became the power supply. Bus 2J7 -1 will remain the power supply for 30 minutes with Bus 1 J l greater than 440 volts, at which time it will automatically return to Bus 7 J 1 ~

D. Six seconds after the voltage dropped on Bus 1 J1. Bus 2J1-1 became the power supply. Bus 2J1-1 will remain the power supply for six seconds with Bus 141 greater than 440 volts, at which time it will automatically return to Bus 13 Surry Nuclear Plant 2004-301 SRO lnital Exam Surry References:

ND-90.3-LP-I, Emergency Diesel Generator, Rev. 14 P&ID 11448-F-1AA, Appendix R Evaluation Electrical One Line Diagram Surry Power

..

Station Unit 1, Rev. 23 Rev. 4 P&ID 11448-E-1 P1,480V One Line Diagram MCC J1-1A Surry Power Station Unit 1, Bistractor Analysis:

A. Incorrect because 1 J l voltage was less than 410 for greater than 6 second B. Incorrect because this is the alternate power supply and the ABT is a C. Correct because 1 J1 voltage was less than 440v for greater than 6 second Therefore, 2J1-1 became the power supply after 6 seconds. The ABT will check for 2J7 -1 voltage greater than 440v prior to swapping to the alternate power suppl normal-seeking ABT. Therefore, at the beginning of this sequence, the power supply would have been 1J Therefore, 2J1-1 became the power supply after six seconds. The ABT will check for 2J1-1 voltage greater than 440v prior to swapping to the alternate power suppl When the normal power supply voltage is restored to > 440v, a 30 minute time delay is started. If the voltage remains above 440v for 30 minutes, then it transfers back to the normal power supply (18).

D. Incorrect because of the 30 minute time delay mentioned abov Emergency Diesel Generator K2.01: Knowledge of bus power supplies to the following: Air Compessor Surry Nuclear Plant 2004-301 SRO Inital Exam The following plant conditions exist:

- Unit 2 is 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 WHR

- 15-F-6, CTMT INST AIR HBR LO PRESS? has annunciated

- Instrument air pressure is approximately stable at 60 psig pump platfor Which ONE of the foiiowing correctly explains the potential effect on warming the WHW system?

A. If the air leak is a rupture upstream of the isolation valve for the air supply to 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 should use the portable air bottle, via quick disconnect, to operate the valv B. If the air leak is a rupture upstream of the isolation valve for the air supply to HCV-1758 (WHR Heat Exchanger Outlet Valve), the valve will fail open. The line may be crimped if the leak will not affect vital control instruments. Operators should use the portable air bottle, via quick disconnect, to operate the valv CY 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 closed. The line may be crimped if the leak will not affect vital control instrument D. If the air leak is a rupture upstream of the isolation valve for the air supply to HCV-1142 CVCS Flow Reaulator Control Valve). the valve will fail orsen. The line

Surry Nuclear Plant 2004-301 SRO Inital Exam Surry References:

ND-88.2-LP-1, Residual Heat Removal System Description, Rev. 8 (Pages 9, 10, 4 I)

NB-88.2-LP-2, Operation of Residual Heat Removal System, Rev 15 P&ID 11448-FM-0878 Sh 2 of 2, Residual Heat Removal System, Rev. 26 P&ID 11448-FM-075E Sh 1 of 2, Compressed Air System, Rev. 43 1 B-F6, CTMT INST AIR HDR LO PRESS, Rev. 1 Bistractor 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 featur E!.

Incorrect because HCV-1758 cannot be operated with a portable air bottl Plausible because the applicant may get consfused on which valve in this flowpath has the portable air bottle featur 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 instrumentatio D. Incorrect because HCV-1 142 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 purpose 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 alar Sur9 Nuclear Plant 2004-301 SRO lnital Exam 51. 067G2.4.18 001/1/2IFIKT~

RWST CCMEM 2.7/3.6/B/SR041011AB/SDR

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In FCA-8.08, 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 RWST to maximize boron injection before the charging pumps overheat and are iost due to a time-overcurrent breaker tri B. Suction is shifted to the RWST to maximize boron injection before the charging pumps overheat and are lost due to an instantaneous-overcurrent breaker tri C. The loss of Charging Pump CC will eventually result in a loss of VCT revel due to a loss of makeup; therefore suction is shifted to the WWS D I The RWST supplies cooler water to the Charging Pumps; thereby minimizing the cooling requirements for the Charging Pump Surry (Utility needs to verify technical acuracy and supply additional supporting material if any is availble.)

Refernces:

ND-95.6-LP-3, Safe Shutdown Fke FCAs, Rev. 5 Q-FCA-8.Q0, 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 temperature 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 temperature C. Incorrect because VCT level will not be reduced as a result of no C B. Correct because cooler WWST water will help reduce pump temps when CC is los Plant Fire On-Site (32.4.18: Knowledge of specific bases for EOP Surry Nuclear Plant 2004-301 SRO Inital Exam

_ _ ~

52. 068K4.01 001/2/2RAI)ItYI'ION

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The following Conditions exist:

- Both Units are at 100% Power

- Unit 1 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

- 0-RM-M4, 1-VG-WI-I04 HIGH, alarms

- AH 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-I04 (#I Vent Stack RM) to detect higher than normal activity?

A. A Steam Generator Tube Leak on Unit By A spill of high activity coolant in the Chemistry Hot La C. A spill of high activity coolant in the High Rad Sample System Room D. A dropped fuel assembly in the fuel buildin ~

__

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Surry Nuclear Plant 2004-301 SRO M a l Exam SLirry References:

0-RM-M4, 1-VG-WI-104 HIGH, Rev. 2 0-AP-22.00, Fuel Handling Abnormal Conditions, Rev. 18 NB-95.3-LP-1, Pre-TMI Radiation Monitoring System, Rev. 8 Bistractor Analysis:

A. Incorrect because the normai 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 I-VG-RM-I 0 B. Correct because 0-RM-M4 alarming could be caused by a coolant spill in the Chem Hot Lab according to the AR C. Incorrect a spill in the High Radiation Sample System Room would not cause this alarm according tu the AR D. Incorrect because fuel clad damage would not be detected by RM-VG-104 when in its normal Configuration. 0-AP-22.00 does not list RM-VG-104 as a potential means of indication for damaged fuel cla Liquid Radwaste K4.01: Knowledge of design feature@) and / or interbck(s) which provide for the following: Safety and environmental precautions for handling hot, acidic, and radioactive liquid Surry Requal Exam Bank Question #462 (ID:ARPOQ76)

Surry Nuclear Plant 2004-301 SRO lnital Exam A discharge of a waste gas decay tank is in progress when WM-GW-101 reaches the high alarm setpoint and alarm 0-RM-K3, 1 -GW-RI-lOl HIGH, annunciates. Which ON of the following is an automatic action initiated by the high radiation levels from the waste gas decay tank release?

A. 4-GW-FCV-101 I Decay Tank Bleed Isolation Valve, close B. 1 -GW-FCV-l60, CTMT Vacuum Pump Discharge Isolation Valve close C. 1 -GW-FCV-260, CTMT Vacuum Pump Discharge Isolation Valve, closes D' Associated vacuum pumps tri 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 Q-RM-K3, 1-GW-RI-101 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 ARP, this valve will close on reaching the high alarm 6. 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 setpoin setpoin setpoin closed. This info is in a CAUTION in the ARP and a step is provided in the ARP to secure the pumps following the closure of GW-160 I26 Gaseous and Liquid Waste Processing Systems K4.06: Knowledge of design@) features and I or interlocks which provide for the following: Sampling and monitoring of waste gas release tank Surry Nuclear Plant 2004-301 SRO M a l Exam

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Which ONE of the following valves AUTOMATICALLY closes when a HIGH alarm is received on Component Cooling Water Radiation Monitor, RM-CC-1 05?

A? Surge Tank Vent Valve, HCV-CC-10 B. Excess Letdown Heat Exchanger Outlet. HCV-CC-10 C. RCP's Thermal Barrier CC Outlet Flow Inside and Outside Trip VLV, TV-CC-14 D. Thermal Barrier Heat Exchanger Isolation Valve, TV-CC-120A. I

-

Surry References:

ND-88.5-LP-1, Component Cooling, Rev. 19 Distractor Analysis:

A. Correct as stated in ND-88.5-LP-7 Page 1 B. Incorrect, but plausible due to the outlet valve also being an HCV in the CCW Sy C. Incorrect, but plausible due to CCW cooling the RCP Th Barrier H D. Incorrect, but plausible due to CCW cooling the RCP Th Barrier H Process Radiation Monitoring 62.1 23: Ability to perform specific system and integrated plant procedures during all modes of plant operatio Surry Nuclear Plant 2004-301 SRO lnital Exam Hydrogen peroxide has just been added to Unit 2 RCS resulting in an increase in the primary cooiant 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 be :' Letdown radiation monitor; monitor letdown filter differential pressur. Letdown radiation monitor; monitor seal return filter differential pressur. Containment particulate radiation monitor; decrease flow through the letdown cation be Surry References:

NB-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 significantl. Correct because letdown radiation monitors would indicate quickly due to hydrogen peroxide increasing reactor coolant activity and letdown filter dP would also ris C. Incorrect because the hydrogen peroxide should not affect the seal return dP, at least not as readily or as soon as the letdown filter dP. There is 8 gal of CVCS water that goes to each RCP for seal injection. Five of these gallons flows down the shaft past the thermal barrier and ends up in the RCS. 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 prefiltering is designed to protect the seals. The water coming 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 burs significantl D. Incorrect because containment particulate radiation monitor would not change Surry Bank ILT Exam Question #1606 096 High Reactor Coolant AK2.01: Knowledge of the interrelations between the High Reactor Coolant Activity and the following: Process radiation monitor Surry Nuclear Plant 2004-301 SRO lnital Exam The following Unit 1 conditions exist:

- A barge Break $OCA occurred 45 minutes ago

- Recirculation Spray is operating

- 41 60V "H" bus de-energizes Which ONE of the following correctly describes the impact on Service Water to and from the Recirc Spray Heat Exchangers?

A. ONLY the Service Water Outlet Valves (MOV-SW-105s) from each Recirc Spray B. ONLY the Service Water Inlet Valves (MOV-SW-104s) from each Recirc Spray C. The Service Water Inlet Valves (MOV-SW-184s) and Service Water Outlet Valves (MOV-SW-1 05s) from both Inside Recirc Spray Heat Exchangers de-energiz DY The Seivice Water Inlet Valves (MOV-SW-184s) and Service Water Outlet Valves Heat Exchanger de-energiz Heat Exchanger de-energiz (MOV-SW-105s) from one Inside and one Outside Recirc Spray Heat Exchanger de-energiz Sury References:

NB-91 -hP-6, Recirculation Spray System, Rev. 9 ND-89.5-LQ-2. Service Water System, Rev. 20 P&ID 11448-FE-1 M, Sh 1 of 1,480V One bine Diagram Surry Power Station - Unit 1,

$&ID 11448-FE-1 L, Sh 1 of 1, 488V One Line Diagram Surry Power Station - Unit 1, Rev. 59 Rev. 52 Distractor Analysis:

A. Incorrect because the inlet and outlet valves from one inside and one outside recirc spray H.X. de-energiz B. Incorrect because the inlet and outlet valves from one inside and one outside recirc spray H.X. de-energiz C. Incorrect because valves from only one inside recirc spray H.X. de-energiz D. Correct because valves de-energize for the A (IRS H.X.) and C (0% H.X.) onl Service Water K2.04: Knowledge of bus power supplies to the following: Reactor building closed cooling wate Surry Nuclear Plant 2004-301 SRQ lnital Exam

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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. While monitoring Instrument Air pressure, the WQ notes pressure is 50 psig and slowly lowerin Which QNE of the following actions should be taken'?

A. Commence a slow power reduction to Hot Shutdown, B. Commence a fast power reduction to Cold Shutdow C! Trip the Reactor and go to 145-0, Reactor Trip or Safety Injection D. Isolate Service Air from instrument Air and start the Sullair Diese Surry References:

NB-92.1-LP-1, Station Air Systems, Rev. 13 0-AP-40.00: Non-recoverable Loss of Instrument Air, Rev. 17 Distractor Analysis:

A. Incorrect because 1 BE6 and AP-40.00 directs rx trip, not power reductio B. Incorrect because 1 B E 6 and AP-40.00 directs rx trip, not power reduction. (Initial distractor from exam bank was changed because it may have been a second correct answer).

1B-E6, lA LOW HDR PRESS / IA COMPW 1 TWBL, Rev. 9 C. Correct because this is the guidance provided by 1 B E 6 and AP-40.0 D. Incorrect because 1B-E6 and AP-40.00 directs TX trip, not power reduction when pressure reaches 50 psi Instrument Air A4.01: Ability to manually operate and / or monitor in the control room: Pressure gauge Surry !?equal Exam Bank Question 428

Surry Nuclear Plant 2004-301 SRO Inital Exam

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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 l ?

(Assume no operator action is taken.)

A. Instrument Air is normally supplied by the Service Air System and the system is backed up by IA when IA pressure reaches 95 psi. Instrument Air is normally supplied by IA Compressors and the system is manually backed up by the Sullair Diese Instrument Air is normally supplied by the Service Air System and is backed up by the IA System when IA pressure reaches 90 psi B. 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 psi ~

Sur9 References NB-92.1 -bP-l~

Station Air Systems, Rev. 13 Distractor Analysis:

A. Incorrect because pressure must drop below 90 psig for IA to backup Service Ai B. Incorrect because the IA System is normally supplied by Service Ai C. Correct because Service Air is the normal supply and IA is the backup when pressure drops below 90 psi D. 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 psi Instrument Air K4.02: Knowledge of the IAS design feature@) and or interlock(s) which provide for the following: Cross-over to other air system Surry Recgual Bank Question #512

Surry Nuclear Plant 2804-301 SRO M a l Exam

~- 59. 103.44 04 001/2/1/CLS

~- CONTAINMJ+JIC/A 3-5/3 SiN/SROS301~ABISI)K 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-1 OOA = 17.7 psia

- PT-LM-1 OOB = 17.8 psia

- PT-LM-1 OOC = 17.6 psia

- PT-LM-1OOB = 17.9 psia Which ONE of the following correctly describes resetting of Consequence Limiting Safeguards (CLS) given the above conditions?

A. The CbS TRAIN A(5) RESET PERMISSWE annunciator is lit. CLS HI and CLS HI-HI may be reset at this time. Upon reset, the HI CLS relays will energize and the HI HI CLS relays will de-energiz de-energized and the HI HI CLS realys are energize this time. Upon reset the HI Hi CLS relays will de-energiz : Neither CLS HI or CLS HI-HI may be reset at this time. The HI CLS relays are C. The CLS HI-HI RESET PERMISSIVE annunciator is lit. CLS HI-HI may be reset at B. Neither CLS HI or CkS HI-HI may be reset at this time. The HI CLS relays are energized and the HI HI CLS relays are de-energize Surry Nuclear Plant 2004-301 SRO lnital Exam Surry References:

ND-88.4-LP-2, Containment Vessel, Rev. 8 ND-91 -tP-5, Containment Spray System, Rev. 13 Distractor Analysis:

A. Incorrect because pressure must be reduced to less than 114.2 psia on 2/4 channels to reset both Hi and Hi-Hi subsystem B. Correct because pressure must be reduced to less than 14.2 p s h on 2/4 channels to reset both Hi and Hi-Hi subsystems. Also, when CLS is actuated, the HI CLS relays are de-energized and the HI HI CLS relays are energize 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 HI CLS relays are de-energized and the HI HI CbS relays are energize B. incorrect because when CLS is actuated, the HI CLS relays are de-energized and the HI HI CLS relays are energize Containment A4.04: Ability to manually operate and / or monitor in the control room: Phase A and Phase B reset Surry Nuclear Plant 2004-301 SRO Inital Exam The following Unit 1 conditions existed:

- Plant was at 74% power after just completing a rapid power reduction due to High Pressure Heater Drain Pump problems

- Axial Flux Difference was outside of the Parget Band on 11/03/2003 from 0800 hours0.00926 days 0.222 hours 0.00132 weeks 3.044e-4 months to 0845 hours0.00978 days 0.235 hours 0.0014 weeks 3.215225e-4 months

- Axial Flux Difference was outside of the Target Band on 11/04/2003 from 0740 hours0.00856 days 0.206 hours 0.00122 weeks 2.8157e-4 months to 0840 hours0.00972 days 0.233 hours 0.00139 weeks 3.1962e-4 months

- The Axial Flux Difference has remained within the Technical Specification Limits of Figure 3.12-3, Axial Flux Difference Limits As A Function Of Rated Power, for the entire time Which ONE of the following actions are required by Technical Specifications?

A:

Reactor power was required to be less than 50% by 0825 hours0.00955 days 0.229 hours 0.00136 weeks 3.139125e-4 months on 11/04/200 B. Reactor power was required to be less than 50% by 0855 hours0.0099 days 0.238 hours 0.00141 weeks 3.253275e-4 months on 11/04/200 C. Reactor power was required to be less than 50% by 0910 hours0.0105 days 0.253 hours 0.0015 weeks 3.46255e-4 months on 11/04/280. No power reduction was required, but power should not have been raised abuve 75% until Axial Flux Difference was within the Target Ban Surry Nuclear Plant 2004-301 SF40 lnital Exam Surry Reference:

Technical Specification 3.12.B.4.b.(l), Amendment No. 186 Technical Specification 3.52.5.4.b.(2), Amendment No. 186 Distractor Analysis:

A. Correct because AFD may deviate from its target band for one hour within a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period. When this is violated, then power must be reduced to less than 50% within 30 minutes. Fram 11/03 @ 0800 his to 11/04 @ 0755 hrs a total of one hour autside of target band was accumulated. Therefore, by 0825 hrs (30 minutes later)

power must be less than 50%.

Plausible because 0855 hours0.0099 days 0.238 hours 0.00141 weeks 3.253275e-4 months 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 0910 hrs is 30 minutes after 0840 hrs, which was given as the second time frame where AFD was outside of its target ban Plausible because candidate may confuse 50% and 7.5% power restriction B. Incorrect because the correct answer is as described in above analysi B. Incorrect because the correct answer is as described in above analysi G2.1.I 1 Knowledge of less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months technical specification action statements for system Surry Nuclear Plant 2004-301 SRO lnital Exam 6 1. G?. 1.25

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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-Re-LI-IOOA

- The "B" and "6" loops are isolated with the primary and secondary SG manways

- The reactor vessel head is tensioned

- The "A" RHR pump is in operation with oscillating amperage indications

- Flow indication I-RH-FI-1605 is oscillating between 2500 and 2700 gp Which ONE of the following actions is appropriate for the SRO to direct in accordance with AP-27.00, Loss of Decay Heat Removal Capability?

(AP-29.00 Attachments I and 2 provided)

removed for SG tube plugging A. Raise RCS level to 12.5 feet as indicated on 1-RC-LI-1OOA and stabilize flow at 2600 gp flow to 2200 gp flow to t 200 gp.5 feet as indicated on 1-RC-LI-IOO B. Throttle open 1-RH-HCV-1958 and throttle close 1-RH-FCV-1605 to reduce RHR C. Throttle close 1-RH-HCV-1758 and throttle open I-WH-FCV-1605 to reduce RHR D! Throttle close 1-WH-FCV-1685 to reduce RHR flow to 2200 gpm and raise level to

Surry Nuclear Plant 2004-301 SRO lnital Exam Surry References:

1 -AP-27.00, Loss of Decay Heat Removal Capability, Rev. 10 NB-88.2-LP-1, Residual Heat Removal System Description, Rev. 8 ND-88.2-LP-02, Operation of Residual Heat Removal System, Rev. 15 ND-95.2-LP-12, Loss of WHR Events, Rev. 9 Bistractor Anaiysis:

A. Incorrect because AP-27 Att. 2 indicates that 12.5 feet is in the unacceptable region of operation for 2600 gprn RHR flow rat B. Incorrect because AP-27 Att. 2 indicates that 2200 gpm RHR flow rate is in the unacceptable region of operation for 12.4 fee. Incorrect because AQ-27 Att. 1 indicates that 1200 gprn WHR flow rate is less than the required flow rate of 22QO gpr D. Correct because these actions place the plant in an acceptable region of AQ-27 At and 2 for required flow rate for 10 days after shutdow AQ-27 Att. 1 and 2 will need to be provided to the applican G2.1.25: Ability to obtain and interpret station reference materials such as graphs, monographs, and tables which contain performance dat Surry Nuclear Plant 2004-301 SRO lnital Exam

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Which ONE of the following is correct with respect to Technical Specifications?

A. The Safety Limit for core thermal power is 109% of Rated Thermal Power and the RCS pressure limit is 2935 psi B. The Safety Limit for core thermal power is 109% of Rated Tkermai 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 Powe CY The reactor trip ora low pressurizer pressure, high pressurizer level, turbine trip, and low reactor coolant flow for two or more loops shall be unblocked when power is greater than or equal to 10% of Rated Thermal Powe D. The source range high flux, high setpoint trip shall be unblocked when the intermediate range nuclear flux is less than or equal to 5x10- ampere Surry References:

Technical Specification 2.1 (Amendments 116); 2.2 (Amendments 203); (Amendments 175, 176, 206)

Distractor Analysis:

A. Incorrect because the safety limit for core thermal power is 1 18%.

B. Incorrect because the safety limit for core thermal power is 11 8%.

C. Correct because this is the correct statement taken from Tech Spec D. Incorrect because source range high flux, high setpoint trip shall be unblocked when the intermediate range nuclear flux is less than or equal to 5x7 0. ampere Generic K/A 2.2.22 Knowledge of limiting conditions for operations and safety limit Surry Nuclear Plant 2004-301

§RO lnital Exam 6 ~ Ci2.2.27

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Which ONE of the following correctly states the level of authorization needed for bypassing the Manipulator Crane Overload Interlock?

A. Refueling §a0 or Fuel Handling Supervisor B. Refueling SRO and Shift Supervisor CI SNSOC and Refueling SWO D. SNSOConly Surry References:

VPAP-1401, Conduct of Operations, Rev. 11 (Section 6.5)

Bistractor Analysis:

A. Incorrect because SNSOC pre-approval is needed per I-OP-FH-015 Step 4.1. Incorrect because SNSOC p~e-approval is needed per I-BP-FH-015 Step 4.1 C. Correct because SRO approval is needed per 1-OP-FH-015 Step 4.10 AND B. Incorrect because SRO approval is needed per 1-OQ-FH-015 Step 4.1 G2.2.27 Knowledge of the refueling proces ~

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SNSOC pre-approval is needed per 1-OP-FH-015 Step 4.1 Surry Nuclear Plant 2004-301 SRQ lnital Exam 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 too close to the surface of the water

- Area radiation monitors alarm in the vicinity of the fuel movements

- Qperators enter 0-AP-22.00, Fuel Handling Abnormal Conditions

- All components operate as designed Which ONE of the following are immediate actions of AP-22.00?

A? Stop fuel handling operations, Secure Normal MCR Ventilation by closing 1-VS-MOD-103C and 1 -VS-MOD-l03D, Bump Cable Vault Air Bottles by closing 1-VS-MOB-103 B. Stop fuel handling operations, Secure Normal MCR Ventilation by closing 1-VS-MOD-1032 and l-VS-MOD-l03D, Dump MER 3 Air Bottles by closing 1 -VS-MOD-? 03 C. Evacuate the affected areas, Secure Normal MCR Ventilation by closing l-VS-MOD-103C and 1-VS-MOD-1038, Dump MER 3 Air Bottles by closing 1 -VS-MODI 03 D. Stop fuel handling operations, Evacuate the affected areas, Stop Main Control

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Room Fans 1 -VS-F-15 andVS-AC<-

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Surry References:

0-AP-22.00, Fuel Handling Abnormal Conditions, Rev. 7 8 Distractor Analysis:

A. Correct these are all listed as immediate actions of AQ-22.0 B. Incorrect because l-VS-MOD-IO3A is in the WNQ column to be performed if 103B does not close. However, the stem states that all equipment operates as designed, so the operator would not go to the RNQ colum does not clos C. Incorrect because I-VS-MQD-103A is in the WNQ column to be performed if 103B D. Incorrect because stopping MCR Ventilation Fans is not an immediate actio (32.3.10: Ability to perform procedures to reduce excessive levels of radiation and guard against personnel exposur Surry Nuclear Plant 2004-301 SRO M a l Exam 65. G2 3 2 001/3/RADI4TION

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RESPIRAI'OIUC/A 2 5/2 9/N/SK03301/R/MAB/Sl>)K ____

Work in a radiation area must be performed. The following conditions exist:

- A point source is present and emits 50 mrem/hour at 1 foot

- The air has a Derived Air Concentration (DACJ of 10 Which ONE of the following methods will result in the lowest amount of accumulated dose?

A. Two workers using hand tools can perform the work in one hour at a distance of two feet wearing no respirato B. Three workers using remote tools perform the work in two hours at a distance of six feet wearing no respirato feet wearing a respirator with a protection factor of 5 C. Two workers using hand tools perform the work in four hours at a distance of two DI Three workers using remote tools perform the work in 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months at a distance of six feet wearing a respirator with a protection factor of 5 ~

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Surry References:

Dominion Nuclear Employee Training Manual Volume II BRWT. RPT, CSET, SCAT, FWT, Rev. 11, January, 200 Distractor Analysis:

A. Incorrect: 75 mrem > 56.7 mrem. {[(2 men)(l hr)(50mrern/hr)(l/2)2]+[(10 DAC)

B. Incorrect: 158.3 mrem > 56.7 mrem. ([(3 men)(2 hr)(50mrem/hr)(l/6)2]+[(10 DAC)

C. Incorrect: 104 mrem > 56.7 mrem. {[(2 men)(4 hr)(50mrem/hr)(l/6)2]+[(10 DAC)

D. Correct: [(3 men)(lO hrs)(5Q ~1rem/hr)(1/6)~]

+ [(lo DAC)(1/50)(3 menJ(l0 hrs)( (2 men)(l hr)(2.5 mrem/DAC-HR)] = 75 mrem)

(3 men)(2 hrJ(2.5 mrem/BAC-HR)] = 458.3 mrem}

(1/50)(% men)(4 hr)(2.5 mrem/BAC-HR)] = 104 mrem}

mrem/l BAG-HR)] = 41.7 + 15 = 56.7 mre. Knowledge of facility ALARA progra Surry Nuclear Plant 2004-301 SRO lnital Exam 6 G2. ~

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The following Unit 1 conditions exist:

- The RCS temperature is 190 O F.

- Operators are performing Section 5.2 of 1 -0P-VS-001, Containment Ventilation, to place the Containment Purge System in service using 1 -VS-FQBA or 1 -VS-F-58B, Filter Exhaust Fan The Containment Purge Form requires 10,000 cfm purge flo Which ONE of the following correctly states selection criteria, in accordance with 1-OP-VS-004

~ for choosing which valve to use for obtaining the correct purge flow rate?

A? 1-VS-MOV-1008 (Ctmt Purge Exh) should be throttled instead of 1-VS-MOV-101 (Ctmt Purge 5/P) due to the high flow rate required by the Containment Purge For B. I-VS-MOV-101 (Ctmt Purge B/P) should be throttled instead of 1-VS-MOV-IOOD (Ctmt Purge Exh). This is due to the need to open the supply breaker to 4-VS-MOV-1OOD in order to throttle It. Opening the breaker wilk prevent automatic CTMT Purge isolatio C. 1-VS-MOV-101 (Ctmt Purge B/P) should be throttled instead of I-VS-MOV-1OOB (Ctmt Purge Exh) due to the low flow rate required by the Containment Purge For D. 1 -VS-MOW1 B O 5 (Ctmt Purge Exh) should be throttled instead of 1 -VS-MOV-lOI (Ctmt Purge B/P). This is due to the need to open the supply breaker to 1-VS-MOV-101 in order to throttle it. Opening the breaker will prevent automatic CTMT Purge isolatio Surry Nuclear Plant 2004-301 SRO Inital Exam Surry References:

1 -OQ-VS-001, Containment Ventilation, Rev. 20 Distractor Analysis:

A. Correct because 100D should be throttled due to the Containment Purge Form allowing more than 3000 cfm. The bypass will not have enough capacity at this flow rat. Incorrect because even though auto containment purge isolation will not occur with the breaker open, the procedure still directs the use of 1 OOD due to the high flow rate. Plausible because applicant may think it logical to not intentionally incapacitate auto containment isolatio Plausible because 3000 gpm is not a very high flow rat procedure directs 101 to be used for fine tuning the flow rate. Plausible because preventing auto ctmt purge isolation is a concern when using 10O G2.3.9:

Knowledge of the process for petforming a containment purg C. Incorrect because with the flow rate greater than 3000 gprn, lO0D should be use D. Incorrect because the bkr does not need to be opened and at 10,000 gpm, the

Surly Nuclear Plant 2004-381 SRO lnital Exam 67. ii2.3.11 001ii//iU3NORMAL

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PROCIiI>lJRE/MEhl3.4/3.6/N/SR04~01~AB/SI)K

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Which ONE of the following correctly states the requirements for performing immediate action steps within emergency procedures?

A. Immediate action steps must be performed in the order in which they appear in any procedur B. Immediate action steps may be performed in any order, except for the first four immediate action steps of E O, Weactor Trip or Safety Injection, which must be performed in the order in which they appear in the procedur C I 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, 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 procedur D. Immediate action steps may be performed in any order except for the immediate action steps of FR-S. I, Response to Nuclear Generation / ATWS, and ECA-0.0, Loss of Ali AC Power, which must be performed in the order in which they appear in the procedur 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-S.1 must be performed in the order in which they appear in the procedur B. Incorrect because only immediate actions of E O and FR-S.1 must be performed in the order in which they appear in the procedur C. Correct because immediate actions of E-0 and FR-S.1 must be performed in the order in which they appear in the procedure. This requirement / expectation is stated in ND-95.3-LP-2 Page 1 D. Incorrect because ECA-0.0 are not required to be performed in any specific orde (32.4.1 1 : Knowledge of abnormal condition procedure Surry Nuclear Plant 2004-301 SRO Initai Exam A situation presents itself that requires a Reactor Operator (RO) to take quick decisive action to ensure Station Safety. Personnel are not in immediate danger and the action requires no reactivity manipulation Which ONE of the following correctly describes the requirements for performing the actions?

A! The WO may take necessaly action without prior apgrovat from another licensed ope rato r.

action and oniy take action after approval is grante B. The RO must immediately request approval from the Unit SWO to perform the C. The RO may take action only after another licensed operator has been notified and concurs with the actio B. The RO may take action only after obtaining a peer check Bo concur with the actio Surly References:

OPAP-0006, Shift Operating Practices, Rev. 4 Distractor Analysis:

A. Correct because OPAP-QO06 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."

5. Incorrect because action may be taken prior to obtaining permissio. Incorrect because action may be taken prior to notifying or obtaining permission B. Incorrect because immediate action is authorized to protect the Statio G2.4.12: Knowledge of general operating crew responsibilities during emergency operation from another Team Membe Surry Nuclear Plant 2004-301 SWO Bnital Exam 6 ~2 4 49 oo2/mor) CorimouciA 4 ou O/NISRW~OI/KMAH/SDR Given the following conditions:

- Reactor Power = 85%

- Control Kods are in automatic

- Control Bank D begins to insert without a turbine runback

- Tave and Tref are matched within 0.5 O F Which ONE of the following describes the correct immediate operator response to these conditions?

A. Verify quadrant power tilt and axial flux difference within limit B! Place ROD CONT MODE SEL switch in MANUA C. Manually trip the reacto B. Verify BRPB operating properl Surry References:

Q-AP-1.Q0, Rod Control System Malfunction, Rev. Distractor Analysis:

A. Incorrect because the initial response is to place ROD CONT MODE SEL switch in MANUA B. Correct per AP-1.Q C. Incorrect because this would not be performed until ROB CONT MODE §EL switch D. Incorrect because AP-1.QO directs placing ROB CONT MODE §EL switch in was placed to MANUAL and rod motion had stoppe MANUAL as an immediate actio G2.4.49 Ability to perform without reference to procedures those actions that require immediate operation of system components and control Surry Nuclear Plant 2004-301 SRO lnital Exam

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V1004EK3.2 00 I/ 1 /I /I.OCA OUTSIDEICIA. 1.4/4.OMSR0430

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Which ON of the following correctly states actions contained in 1-ECA-7.2, LOCA Outside Containment, and the reasons for those actions?

A. Open 1-SI-MOV-1890A (LHSI to Hot beg) or 1-SI-MOV-1890B (LHSB to Hot Leg) to provide a flow path for Low Head Safety Injection. Then close 1-SI-MOV-1890C (LHSI to Cold Legs) and monitor WCS pressur B. If closing 1-SB-MOV-1890C (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 isolate C. If the leak is not identified and isolated then transition to 1 - E l, Loss of Reactor or Secondary Coolant, because WCS inventory is continued to be lost outside of containmen D: 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 vaives from the RWST will be closed to isolate potential leak path ~

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Surry References:

ND-95.3-LP-21, ECA-1.2 LOCA Outside Containment, Rev. 7 ECA-1.2, LOCA Outside Containment, Rev. 5 Bistractor Analysis:

A. Incorrect because ECA-1.2 does not give any direction to open I-%I-MOV-I890A &

5. These valves should be left in the closed position. This distractor is plausible because ECA-7.2 does give guidance to close 189O decreasing, then the leak was not isolated and the valve needs to be re-opene This is the normal SI flow path and it is important to re-establish this path if closing the valve did not isolate the lea C. Incorrect because if the leak is not isolated, then the correct transition would be to go to I-ECA-1.1, Loss of Emergency Coolant Recirculatio B. Correct because if RCS pressure rises upon closure of l-SI-MOV-7890C, then the leak was isolated and 1-ECA-1.2 directs the LHSl pumps to be placed in PTL and the suction valves from the RWST to be close. Incorrect because if I-SI-MOV-1890C is closed and RCS pressure is still W E04 EK3.2: Knowledge of the reasons for the following responses as they apply to the (LQCA Outside Containment): Normal, abnormal, and emergency operating procedures associated with (LOCA Outside Containment).

Surry Nuclear Plant 2804-301 SRO lnital Exam I-FR-C.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 U-tube B. The SGs will depressurize mure quickly if no Forced Circulation RCS flow exist. To minimize heat input to the RC DI The SG depressurization will lead to a loss of RCP support condition Surry References:

ND-95.3-LP-38, Response to Inadequate Core Cooling, Rev. 8 FR-e.?, 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 RCP B. Incorrect because the basis for securing RCPs is not associated with heat input into the RCS or forced flo C. Incorrect because the basis for securing RCPs is not associated with heat input into the RC D. Correct because this is the stated reasun in ND-95.3-LP-38. Losing #1 Seal support conditions could result in damage to the RCP had. 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 characteristic Surry Requal Exam Bank Question #467

Surry Nuclear Plant 2084-301 SRO lnital Exam 72. WEOX02.1.7 001/1/2/TURE RL!ITJKR STEAM/C/A 3.7/4.4/A/SK04301/WMABISDR

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The following Unit 1 conditions exist:

- Reactor power is 58% and rising

- RCS pressure is at 2210 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 pressures are at 970 psig and slowly lowering

~ Containment pressure is 9.5 psia and slowly rising

- Condenser Air Ejector RM reads I8 cpm Which ONE of the following correctly diagnoses the event?

A. Ruptured and faulted steam line break inside containmen B:' Steam line break inside containmen. LQCA inside containmen D. Steam line break outside containmen ~

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Surry Nuclear Plant 2004-301 SRQ lnital Exam Surry References:

General operator knowledg 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 SGT 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 etm C. Incorrect because reactor power would not be rising during a LOCA as it would during a steam line break. Plausible because many of the parameters coincide with a LOC D. Incorrect because ctmt pressure is rising. Plausible because of the aforementioned parameters that are indicative of a steam line brea WE08 RCS Overcooling G2.1.7: Ability to evaluate plant performance and make operational judgements based on operating characteristics, reactor behavior, and instrument interpretatio Sur9 Wequal Bank Question #17? (ID: EOP0076)

Surry Nuclear Plant 2004-301 SRO lnital Exam

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HHSVClA 3.5/3.X/N/SRO4301XI/M~/SDK

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The following conditions exist:

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LOCA has occurre WWST level = 13% and decreasin Recirculation Mode Transfer (RMT) keyswitch is in RMT Mod White RMT Status Light is li Amber RMT Status Light is li SI-MOV-1860A (LHSl Suction from Sump) opens fully and l-SI-MOV-1860B (LHSI 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:' "B" LHSl Pump from the RWST and "A" LHSl Pump from the Containment Sump is injecting into the cold legs and HHSl from LHSI pump discharge is injecting into the cold leg B. No Safety Injection is injecting water to the cold leg C. HHSI directly from the RWST (not from LHSl discharge) is injecting into the cold B. Both LHSI Pumps from the RWST and HHSI directly from the RWST (not from Legs, but no LHSl is injecting into the cold leg LHSI discharge) is being injected into the cold leg ___

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Surry References:

NB-91.3-LP-3, Safety Injection System Operations, Rev. 15 1 -ES-l.3, Transfer to Cold Leg Recirculation, Rev. 1 1 Distractor Analysis:

A. Correct because 1-SI-MOV-18628 will not close until 181-MOV-1860B opens fully (due to an interlock).

injecting into the cold legs and HHSI is not taking suction directly from the RWS taking suction on the discharge of the LHSl Pump B. Incorrect because RWST and Sump are suction sources for LHSl pump C. Incorrect because LHSB Pumps are taking suction from RWST and Sump and D. Incorrect because HHSl is not taking suction directly from the RWST. HHSl is WE1 1 EA1 2: Ability to operate and I or monitor the following as they apply to the (Loss of Emergency Coolant Recirculation): Operating behavior characteristics of the facilit Surry Nuclear Plant 2004-301 SRQ lnital Exam A steam break has occurred and all Steam Generators are faulte Which ONE of the following is the basis for maintaining a minimum of 6Q gpm AFW flow to each Steam Generator per ECA-2.4, Uncontrolled Depressurization of All Steam Generators?

A. 60 gpm is needed to meet minimum heat sink flow requirement B. 60 gpm to each Steam Generator will ensure even thermal hydraulic distribution across the cor gpm is the minimum indicated flow rate to prevent Steam Generator dryou D. 60 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 action ~~~

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Surry References:

ND-95.3-LP-22, ECA-2.1 Uncontrolled Depressurization of All Steam Generators, 1-E-3, ECAP.l, Uncontrolled Depressurization of All Steam Generators, Rev. 16 Distractor Analysis:

A. Incorrect 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 60 gpm is the minimum verifiable flow rate to a steam generato Rev. 9 requirements, it is based on SG dyou across the core. It is based on S/G dryou This ensures a nominal flow rate of 25 gprn to the SIG, considering detector uncertainties, to prevent dryout and thermal shock to the S/ B. Incorrect because the concern is with thermal shock to the SG if AFW flow rates are rasie JW/E12) Steam Line 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 facilit Modified Surry ILT Bank Question #lo1 0

Surry Nuclear Plant 2804-301 SRO lnital Exam 75. WE13EK2 I 001/1/2/SGIK OVERPRESSUREhEM

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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 flo Which ONE of the following correctly states the basis for this step?

I A:' To ensure that the ruptured steam generator tubes are covered to promote thermal stratificatio B. To ensure thermaL gradients across the tubes of the ruptured steam generator do not exacerbate existing tube damag C. To ensure sufficient heat sink for reactor coolant system cooldow D. To prevent excessive primary to secondaty leakag Surry References:

1-E-3, Steam Generator Tube Rupture, Rev. 25 NB-95.3-LP-13, E 9 Steam Generator Tube Rupture, Rev. 11 Bistractor Analysis:

A. Correct because this is the basis as stated in NB-95.3-LP-1 B. 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 overfill, while mitigating release to the publi C. Incorrect because this SG will not be used for the RCS cooldow. Incorrect because the dP is still going to induce leakage even at 12Y0 SG leve WE13 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 feature Question is modified from a Braidwood Questio Surry Nuclear Plant 2004-3Qf Sf30 lnital Exam Which ONE of the following identifies an event that is required to be reported to the NRC within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of discovery?

(Reference provided)

A. An inadvertant Safety Injection due to an instrument surveillance erro."

The Shift Supervisor authorizes the individual insertion of control rods into the core without bank overlap to shutdown the reactor in an emergenc C. A hypochlorite spill outside the Polishing Building of which the PA has been notifie D. A radioactive release such that if an individuai had been present for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , they could have received an intake in excess of one occupational annual limit on intak Surry References:

VPAP-2802, Notifications and Reports, Rev. 1 NB-95.5-LP-2, Station Emergency Manager, Rev. 8 Bistractor Analysis:

A. Incorrect because this is a 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months reportable event. Plausible because the applicant may think that inadvertant safety injection is important enough to require reporting to the NRC within one hou B. Correct per VPAQ-2802 Section 6.3.3 for deviation from Tech Specs. (VPAQ-2802 Page 77.)

C. Incorrect because this is a 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months reportable event. Plausible because the applicant may think that a hypochlorite spill with EPA notification is important enough to require reporting to the NRC within one hou applicant may think that a large radioactive release is important enough to require reporting to the NWC within one hou B. Incorrect because this is a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months reportable event. Plausible because the 001 Control Rod Drive G2.4.30 Knowledge of which events related to system operations I status should be reported to outside agencie Surry Nuclear Plant 2004-301 SWO lnital Exam 7.1.32

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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 activitie ~

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 f 0 F locking, sealing, or otherwise 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 RC B. 68 minutes to ensure the proper safety system alignmen C. 15 minutes to ensure the proper safety system alignmen D. 60 minutes to prevent inadvertant boron dilution of the RC Surry References:

Technical Specification 3.2.E.3, Amendment 199 Distractor Analysis:

A. Correct per Technical Specifications and Basi. Incorrect because Technical Specifications require within 15 minute C. Incorrect because Technical Specifications Basis states that these valves shall be D. Incorrect because Technical Specifications require within 15 minute Chemical and Volume Control G2.1.32: Ability to explain and apply all system limits and precaution closed to provide assurance that an inadvertant boron dilution will not occu Surry Nuclear Plant 2004-301 SRO lnital Exam Given the following Unit 1 conditions:

- A small break LOCA has occurred

- As directed by the EOPs, the RCPs have been tripped

- 1-E§-1.2, Post-LOCA Cooldown and Depressurization, Step 20, "Verify Natural

- WCS pressure is 1490 psig

- Wide Range T-Cold indications are 505 O F and slowly decreasing

- Wide Range T-Hot indications are 515 O F and s.lowly decreasing

- CETCs are 581 O F and stable

- Containment Pressure is 10 psia

- Containment Radiation Levels are: 5.0 x IO5 Whr

- SG Narrow Range bevels are: A=22%, 8=24%, C=22%, and slowly decreasing

- SG Pressures are 715 p i g 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

> 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 satisfie Depressurize the SGs by raising steam flow rate through the steam dumps. Then depressurize when subcooling is > 85 O F.

RCS by raising steam flow rate through the steam dumps. Then depressurize wher subcooling is > 95 O F.

D I Natural Circulation criteria are not met due to inadequate subcooling. Cool the

Surry Nuclear Plant 2004-301 SRO lnital Exam Surly References:

f 43-1.2, Post LOCA Cooldown and Depressurization, Rev. 21 Distractor Analysis:

A. Incorrect because there is not adequate subcoolin B. Incorrect because CETCs do not need to be decreasin C. Incorrect because SG parameters are satisfie D. Correct because there is inadequate subcooling (1 6 O F e 85 OF). ES-1.2 Step 20 RNO directs dumping of more steam. The basis for Step 21 of dumping steam until subcooling is e 95 O F is to ensure that the 85 O F natural circ criteria is not violate The Degraded Containment numbers were used due to the CETC = 581 O F ; P = 1490 psig = 1505 psia; Tsat( 1505 pia) = 597 O F ;

Subcooling = 597 - 581 = 16 O F Surty ILT Bank Exam Question #I 069 009 Small Break LOCA EA2.39: Ability to determine or interpret the following as they apply tu a small break LOCA: Adequate core coolin Surry Nuclear Plant 2004-301 SRO lnital Exam 79. 025AG2. ~~

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~ 00111111RHR

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~ WED AND BLEEI~ICIA

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~ 1.113.8/nlSR0430IISIMADISDR

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The following Unit 1 conditions exist:

- RCS is not pressurized

- RCS level is 16.00 feet as read on 1 -RC-LI-I BOA Which ONE of the following specifies the MINIMUM MANDATORY backup cooling method@) required to be available before entering the above plant conditions, in accordance with OSP-ZZ-004, Unit 1 Safety §ystems Status List For Cold Shutdown I Refueling Conditions?

A. Reflux Boiling AND Gravity Feed and Blee B. Gravity Feed and Bleed ONL C. Forced Feed and Bleed AND Gravity Feed and Blee D I Forced Feed and Bleed ONL Surly References:

1 -0SP-ZZ-004, Unit 7 Safety Systems Status List For Cold Shutdown / Refueling 1 -AP-27.00, Loss of Decay Heat ReMOVal Capability, Rev. 10 ND-95.2-LP-12, Loss of RHR Events, Rev. 9 Distractor Analysis:

A. Incorrect: Per 1 -OSP-ZZ-OQ4, 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-004, Step 6.1.2, Forced Feed and Bleed is the only D. Correct: Per 1-OSP-ZZ-004, Step 6.1.2; Forced Feed and Bleed is the only Conditions, Rev. 27 Mandatory Backup method require Mandatoy Backup method require Mandatoy Backup method require Mandatory Backup method require Loss of RHR G2.4.7: Knowledge of event based EOP mitigation strategy

Surry Nuclear Plant 2804-301 SRO lnitaf Exam

- 80. 055EA2

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03 001/1/1BLACKOUT SBO/MEM 3 Y/4 7B/SR04301/S/M~/SI)K ___

The following conditions exist:

~ A loss of all AC power has occurre ~ The STA reports the status of the CSFs are as follows:

- Subcriticality ~ RED

- Core Cooling - RED

- Heat Sink - WED

- Integrity - GREEN

- Containment - GREEN

- Inventory - YELLOW Which ONE of the following procedures should be used to mitigate these conditions?

A. 4-FR-S.1, Response to Nuclear Power Generation I ATWS 3:'

1-ECA-0.0, boss of All A6 Power C. 1-FW-H.1, Response to Loss of Secondary Heat Sink D. 1-FR-C.1, Response to Inadequate Core Cooling Surry References:

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-8.0. (see NOTE B. Correct because this is the correct procedure to mitigate the loss of ac powe C. Incorrect because FR's should not be implemented while in ECA-D. Incorrect because FR's should not be implemented while in ECA-Surry ILT Exam Bank Question #839 Q55 Station Blackout EA2.03: Ability to determine or interpret the following as they apply to Station Blackout:

Actions necessary to restore powe prior to step 1 of ECA-0.0)

Surry Nuclear Plant 2004-301 SRO lnital Exam 81. 056Ci2 4 45 0011111l~ONDENSATEIC'IA 3 313 6MlSR0430l~lSlSDR

~

I The following Unit 1 conditions exist:

- Power = 100%

- Condenser vacuum is lowering slawl Steam Generator levels are 45% and lowerin Several alarms have annunciated, including:

- 1 H-G8, FW PP DlSCH HDR LO PRESS

- 1$-G4, CN PPS DlSCH HBR LO PRESS s IC-A1, RCP 1A CC RETURN LO FLOW

- lC-Bl, RCP 1B CC RETURN LO FLOW

- 1C-61, RCP IC CC RETURN LO FLOW Which ONE of the following states the SROs correct prioritization of the above conditions as indicated by the procedures and actions ChQSen to mitigate or correct the conditians?

A. Trip the Reactor followed by tripping the Reactor Coolant Pumps. Enter E-0, Reactor Trip or Safety Injectio has started and reduce turbine loa BI Enter AP-10.05, Loss of Semi-vital Bus. Verify that the standby condensate pump C. Enter AP-21.OO, bass of Main Feedwater Flow. Maintain full power operation and manually control Steam Generator levels by placing Feedwater Regulating Valves in MANUAL contro D. Enter AP-23.00, Rapid Load Reduction, to bring the unit offline, followed by tripping the Reactor Coolant Pump Surry Nuclear Plant 2004-301 SRO lnital Exam Surry References:

ND-90.3-LF-5, Vital and Semi-vital Bus Distribution, Rev. 11 1-AP-lQ.Q5, boss of Semi-Vital Bus, Rev. 16 1-AP-21.OO, boss of Main Feedwater Flow, Rev. 5 1 -AQ-23.00, Rapid Load Reduction, Rev. 15 1 H-G8, FW PF DISCH HDR LO PRESS, Rev. Q 1J-G4, CN PPS DISCH HDR LO PRESS, Rev. 0 1C-Al, RCP 1ACC RETURN LO FLOW, Rev. 2 1C-Bl, RCP 1B CC RETURN LO FLOW, Rev. 2 IC-C1, RCP l C 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 actuarly had no cooling, the Rx should be tripped and RCPs should be secure B. Correct because all 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-10.0 C. Incorrect because maintaining load at 100% will cause SG levels to continue to go down. The FW and Condensate Recircs have failed 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 FWV may help to mitigate the conditio D. Incorrect because the unit should not be taken off line using AP-23.08 and RCPs should not be tripped due to the loss of the SVB. Piausible because rapidly bringing the unit off line and securing RCPs, given the stated conditions, may appear logical to the applican Modified Sur9 ILT Exam Bank Question #224 (maybe it could be considered a new question?)

056 Condensate (32.4.45: Ability to prioritize and interpret the significance of each annunciator or alar Surry Nuclear Plant 2004-301 Sa0 lnital Exam

__ 8 G2 1 6 OOI/I/I/VITAL AC A 0 ACTIONSICIA 2 1/4 J/N/SR04301/S/hIAH/SI)K The following Unit 1 conditions exist:

- Reactor Power = 30%

- Plant is in a Chemistry hold during a power ascension

- A loss of Vital Bus 111 occurs and operators enter 1-AP-10.03, Loss of Vital Bus 111

- Electricians quickly find a fault on Vital Bus 1-111 and believe that it will take 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months

- I-CC-TV-1058, CCW TV for the "A" Reactor Coolant Pump (RCP), has closed and

- RCP temperatures are starting to slowly ris Which ONE of the foliowing sets of actions should the Senior Reactor Operator (SRO)

direct given the above conditions?

to repai cannot be reopene A. The SRO should direct the securing of the "A" RCP. Reactor power may be maintained at 30% for the duration of the 10 hour1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months repair to reenergize Vital Bus l-ll 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 SWO should direct preparation to bring the unit to hot shutdown within the following six hour CY The SR6 should direct a Reactor Trip, followed by the securing of the " A RC The SRO should then direct performance of I-E-0, Reactor Trip or Safety Injection, and continue with applicable actions of I-AP-lQ.0 D. 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 shutdow Surry Nuclear Plant 2004-301

§BO lnital Exam Surry References:

ND-93.3-LP-16, Permissive/Bypass?Trip Status Lights, Rev. 8 ND-93.3-LP-10, Reactor Protection - General, Rev. 5 ND-90.3-LP-5, Vital and Semi-vital Bus Distribution, Rev. 1 f 1-AQ-10.03, Loss of Vital Bus Ill, Rev. 8 Distractor Analysis:

A. Incorrect because fS 3.16 and commitments made in Gb-91-11 (also located in Note prior to Step 17 in AP-10.03). The VB must be re-powered within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or the unit must be in HSB within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Also incorrect because AP-10.03 will require a reactor trip. Plausible because the loss of Vi3 causes a loss of cooling to "A" RCP. It may appear OK to continue operation because the power is < P-B. incorrect because A$-t 0.03 requires a reactor trip and securing of RCP if CCW will not be restored prior to WCP temperatures reaching action level limits. Plausible because of the NOTE mentioned in the previous distractor analysi C. Correct because AP-10.03 directs Rx Trip and securing of RCP if CCW will not be restored prior to getting cooling back to that pump. The stem states that the TV is closed and cannot be re-opened, thus preventing cooling to be restored to the RC D. Incorrect because AP-10.03 directs Rx Trip, not a controlled shutdown. Plausible because power is < P-8, which may allow the applicant to incorrectly believe that a shutdown is accepatbl boss of Vital AC lnst Bus G2.f.6: Ability to supervise and assume a management role during plant transients and upset condition Surry Nuclear Plant 2004-301 SRO M a l Exam The following Unit 1 conditions exist:

- Unit 4 power is 100%

~

No annunciators are lit

- Annunciator 1 K-H1, Hathaway Power Available Lights, has just extinguished Which ONE of the following is the correct Abnormal Procedure to enter and correct Event Classification?

(Reference provided)

A. Enter 0-A$-70.13, Loss of Main Control Room Annunciators, due to the loss of one of the 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 45 minute B:' Enter 0-AP-10.13, Loss of Main Control Room Annunciators, due to the loss of both power supplies to Unit I annunciators. Enter the Emergency Plan and declare a Notification of Unusual Event if the loss of annunciators lasts for greater than 15 minute. Enter 1-AQ-10.06, Loss of BC Power, and 0-PIP-40.13, boss of Main Control Room Annunciators, due to a loss of BC power and loss of one of the power supplies to Unit 1 annunciators. Enter the Emergency Plan and declare an Alert if the loss of annunciators lasts for greater than 15 minute D. Enter 1-AP-10.06, Loss of DC Power, and 0-AP-10.19, Loss of Main Control Room Annunciators, due to a loss of BC power and loss of both power supplies to Unit 1 annunciator annunciators lasts for greater than 15 minute Enter the Emergency Plan and declare an Alert if the loss of

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Surry Nuclear Plant 2004-304 SRO lnital Exam Surry References:

0-AQ-10.13, Loss of Main Control Board Room Annunciators, Rev. 4 EPIP-1.81, Emergency Manager Controlling Procedure, Rev. 43 Distractor Analysis:

A. Incorrect because 1 K-HI not lit is indication of both power supplies to Unit 1 B. Correct because 1 K-I not lit is indication of both power supplies to Unit 1 annunciator Panels having been los annunciator Panels having been lost. EFT-1.01 Page 6 states that if safety system annunciators are lost for greater than 15 minutes while above CSB, then a NOUE shall be declare annunciator Panels having been lost. Since the plant is still at 1QO% power, there is no indication that any DC Bus has been lost; 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 NOUE is the correct classificatio B. Incorrect because the plant is still at 100% power, there is no indication that any DC Bus has been lost; therefore 1-AQ-10.06 should not be entered. An Alert classification based on the loss of DC would be incorrect. As stated above, a NOUE is the correct classificatio C. Incorrect because 1K-HI not lit is indication of both power supplies to Unit I Provide PIP-1.01 Pages 6 and 27 058 Loss of DC Power G2.4.32: Knowledge of operator response to a loss of all annunciator Surry Nuclear Plant 2004-301 SRO lnital Exam Unit 1 is at I OOo/o power. It experiences a loss of Vital Bus I at 1200 hours0.0139 days 0.333 hours 0.00198 weeks 4.566e-4 months on Monda Operators enter 1-AQ-10.01, Loss of Vital Bus I, and re-energize the Vital Bus from its alternate source at 1215 hours0.0141 days 0.338 hours 0.00201 weeks 4.623075e-4 months on Monda Which ONE of the following correctly states the required actions based on the above condition?

A. In accordance with 1-AP-18.01, Vital Bus I must be re-energized from its primary source by 1400 hours0.0162 days 0.389 hours 0.00231 weeks 5.327e-4 months on Monday, OF be in Hot Shutdown by 2000 hours0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months on Monda source by 1415 hours0.0164 days 0.393 hours 0.00234 weeks 5.384075e-4 months on Monday, OF be in Hot Shutdown by 2815 hours0.0326 days 0.782 hours 0.00465 weeks 0.00107 months on Monda C I In accordance with 1-AP-10.01, Vital Bus B must be reenergized from its primary source by 1200 hours0.0139 days 0.333 hours 0.00198 weeks 4.566e-4 months on Tuesday, or be in Hot Shutdown by 1800 hours0.0208 days 0.5 hours 0.00298 weeks 6.849e-4 months on Tuesda. In accordance with 1-AP-10.01, Vital Bus I must be reenergized from its primary D. No shutdown requirements are in effect as long as Vital Bus I is energize ~

Surry Nuclear Plant 2004-301 SWO lnital Exam Surry References:

1-AP-10.01 Loss of Vital Bus I? Rev. 43 ND-90.3-bP-5., Vital and Semi-vital Bus Distribution, Rev. 11 Distractor Analysis:

A. Incorrect because per AB-fO.O1 Step 16 c, the VB must be powered from its normal source within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or the unit must be placed in HSD within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months (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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and 1400 hours0.0162 days 0.389 hours 0.00231 weeks 5.327e-4 months is 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after 1200 hour0.0139 days 0.333 hours 0.00198 weeks 4.566e-4 months B. Incorrect because per AB-10.01 Step 16 c, the VB must be powered from its normal source within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or the unit must be placed in HSD within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months (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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and 141.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months is 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after 1215 hour0.0141 days 0.338 hours 0.00201 weeks 4.623075e-4 months normal source within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or the unit must be placed in HSD within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months (also see ND-90.3-LP-5 Page 15). The consequences of having VB-I not energized by its primary source are mitigated, or corrected, by ensuring that it is energized from its primary source within the specified time requiremen 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or the unit must be placed in HSD within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months (also see ND-90.3-LP-5 Page 15). Plausible because the Vital Bus is energized and the plant would be operating satisfactoril 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 a6 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 a system with a fault on i Surry Nuclear Plant 2004-301 SRO lnital Exam

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85. 062AA2.04 001/1/1/SEKVICE

- WATTEWCIA 2.5/2.9~/SRo4301/S/I\\IAB/SI)R

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The following Unit 1 conditions exist:

- Power = 100%

- 1-CH-P-IA Charging Pump is operating

- 1-SW-P-1OA Charging Pump Service Water Pump is operating

- l-SW-P-1OB Charging Pump Service Water Pump is in standby

- 1 D-G5, SW OW CC PPS BlSCH TO CHG PPS LO PRESS, alarms

- 1-CH-P-IA Charging Pump Bearing Temperature = 195 OF

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1-CH-P-1A Charging Pump Oil Cooler Outlet Temperature = 150 O F

- The Pressure indication on the discharge of I-SW-P-IOA Charging Pump Service Water Pump (SW-PI-26) reached a minimum value of 10 psig where it remains stabl Charging Pump Sewice Water Pump is noisy and has high vibration The Operator in the field reports back to the Control Room that I-SW-P-1OA 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 Pump is INOPERABL Direct starting standby Charging Pump Service Water Pump, direct securing the

"A" Charging Pump Service Water Pump, and notify the System Enginee INOPERABLE. Verify auto start of 1-SW-P-1OB Charging Pump Service Water Pump, and notify the System Enginee B. Bearing Temperature is not within limits. I-CH-P-1A Charging Pump is CY Oil Cooler Outlet Temperature is not within normal operating band. I-CH-P-1A 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 Enginee B. Oil Cooler Outlet Temperature is not within normal operating band. Performance of Charging Pump Operability and Performance Test for 1 -CH-P-1A Charging Pump must be directed to determine OPERABILIT Sur9 Nuclear Plant 2004-301 SRO lnital Exam Surry References:

1 DG5, SW OW CC PPS DlSCH TO CHG PPS LO PRESS, Rev. 3 11448-FM-Q71B, Sh. 1 of 2, Flow / Valve Operating Numbers Diagram, Circulating and ND-89.5-LP-2, Service Water System, Rev. 20 1-OP-CH-002, Charging Pump A Operations, Rev. 13 1-OPT-CH-001, Charging Pump Operability and Perfurmance Test For 1-CH-P-1 A, Service Water System, Sur9 Power Station Unit 1, Virginia Power, Rev. 5 Rev. 33 Distractor Analysis: I%. Incorrect because Bearing Temperature is less than 180 OF. OPT-CH-001 Pg 9 states that the upper admin limit is 180 O F. The Charging Pump is still OPERABL Incorrect because Bearing Temperature is less than 180 O F. OPT-CH-001 Pg 9 states that the upper admin limit is 180 O F. The standby pump will not start until 8 psi Correct because Oil Cooler Outlet Temperature is not within the normal operating band (80 - 120 O F ) 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 AR Incorrect because there is no indication that the Charging Pump has a proble Given the above alarm, all indications suggest that the problem is with the Service Water flow. Therefore, perfurmance of the Operability and Performance Test for the Charging Pump would serve no purpos boss of Svc Water AA2.04: Ability to determine and interpret the folluwing as they apply to the Loss of Nuclear Service Water: The normal values and upper limits for the temperatures of the components cooled by SW Surry Nuclear Plant 2004-301 SRO Inital Exam

~ 86. 07W2 4 48 OOI/ZG/INSTRUMENT AIK/C/A 3 5/3 S/B/SR04301/S/MAB/SDR The following Unit 1 conditions exist:

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Reactor Power = 7 00%

- A loss of Containment Instrument Air has occurred

- 7 5-!=6, CTMT INST AIR HBR LO PRESSURE. annunciates 1n.PC DR7R D \\ A / R PEI IFF\\/\\/ I A A I D DRFQC anniinristec

-, w - v v,, 1 1 L 1 1 I v v,, Z l

L L

l L

l

" 1 L " _,,, I,,L"", L.,,,,..,,W,L.."W

- Containment lnstrument Air was crosstied with Turbine Building Instrument Air and is reading 75 psi Which ONE of the following operator actions is required?

(Reference provided)

i A. Both Pressurizer PORVs are operable following the crosstie. Verify the operability by closing PORV Block Valves, stroking PORVs, then re-opening the P8RV Block Valve B. Bath Pressurizer PORVs are operable following the crosstie. No further action associated with the POWVs is require ' Declare both Pressurizer PORVs inoperable. Close and remove power from both PORV block valves within one hour and be in HSD within 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Sur9 Nuclear Plant 2004-301 SRO lnital Exam Sur9 References:

NB-92.1-LP-1, Station Air Systems, Rev. 13 ND-88.1 -bP-3, Pressurizer and Pressure Relief, Rev. 12 1B-F6, CTMT INST AIR HBW LO PRESS, Rev. 1 Technical Specification 3.1.A.6.c, Reactor Coolant System / Relief Valves Distractor Analysis:

A. Incorrect because (per 1 D-C6) with CTMT lnst Air P e 88 psig, the PORVs are inoperabl B. Incorrect because (per 1 D-C6) with CTMT lnst Air P < 80 psig, the PORVs are inoperabl C. Correct because (per 1 D-C6) CTMT lnst Air P must be > 80 psig for the PORVs to be considered operable and capable of being stroked. Per TS the block valves shall be closed and de-energize D. Incorrect because, as stated in "C" above, power shall be removed from the block valves because the POWVs cannot be relied upon to stroke with 75 psig air D-CS, PRZR PWR RELIEF VV LO AIR PRESS: Rev. 4 Sur9 Requal Bank Question #394 (ARPOOQl)

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 condition Surry Nuclear Plant 2004-301 SRO Inital Exam 87. 103.42.01 001/2/1/COXTAINMENT IEAKAGE/C/A 2.0/?.6/N/SR04301/S/MABISI)R

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The following Unit 1 conditions exist:

- Plant is in Mode 1

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Personnel Airlock Seal Leakage Testing has just been completed

- The Personnel Airlock Inner Door Seal exceeded Technical Specifications leakage

- Earlier in the year the Personnel Airlock Inner Boor exceeded Technical limits Specifications leakage limits and the Personnel Airlock Outer Door was opened for a total of 50 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 Boors?

A. The Personnel Airlock Outer Boor 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and Cold Shutdowr within the following 38 hour4.398148e-4 days 0.0106 hours 6.283069e-5 weeks 1.4459e-5 months :

The Personnel Airlock Outer Boor may be opened for 10 minutes to pursue the repair and retest of the Personnel Airlock Inner Boor Seal. Per VPAP-0106, Subatmospheric Containment Entry, the Shift Supervisor shall supervise the containment entry and exit proces C. The Personnel Airlock Outer Boor may be opened for 15 minutes 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 entry and exit proces and retest of the Personnel Airlock Inner Boor Seal. Per VPAP-QI06, Subatmospheric Containment Entry, the Unit SRO shall supervise the containment entry and exit proces D. The Personnel Airlock Outer Boor may be opened for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to pursue the repair

Surry Nuclear Plant 2004-301 SRO lnital Exam Surry References:

VPAP-0106, Subatmospheric Containment Entry, Rev. 5 Technical Specifications 3.8, Containment (Amendments 17% and 174); 1.O.G, Definitions (Amendment 180)

Distractor Analysis:

A. Incorrect because the Outer Boor may be opened for 10 minutes since it has already been opened 50 minutes this year while the inner door was inoperabl. Correct because per Tech Specs, the Outer Boor may be opened for 15 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 per VPAP-0106 Section. Incorrect because the Outer Door may be opened for 10 minutes and the SS must supervise the containment entry and exit per VPAQ-0106 Section B. Incorrect because the Outer Door may be opened for 10 minute 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 Test Surry Nuclear Plant 2004-301 SF30 lnital Exam The following Unit 1 conditions exist:

- The Unit has been operating at f 00% power for the past two weeks

- Chemistry has just provided the following results from a Reactor Coolant System sample that was taken 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ago:

- RCS Chloride = 0.15 ppm

- WCS Fluoride = 0.15 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 allowabbe Technical Specification limit. Per Technical Specifications, corrective action must be taken immediately tQ bring the plant to cold shutdown condition Technical Specifications, corrective action must be taken immediately to bring the oxygen concentration within limits. If the oxygen concentration is outside of the limit after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , then the plant must be taken to cold shutdow B:' Oxygen concentration is above the allowable Technical Specification limit. Per 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 condition 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , then the plant must be taken to cold shutdow D. Chloride concentration is above the allowable Technical Specification limit. Per

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Surry Nuclear Plant 2004-301 SWO lnital 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to see if their corrective actions will bring the parameter within spec. If after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months the parameter is not within spec, then the plant must be taken to cold shutdown using normal plant procedures, 8. Correct because, according to the Tech Spec Basis; the plant has 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to see if their corrective actions will bring the parameter within spec. If after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months the parameter is net within spec, then the plant must be taken to cold shutdown using normal plant procedure C. tncorrect because Chloride concentration is within limit D. Incorrect because Chloride concentration is within h i t s.

G2.1.34: Ability to maintain primary and secondary plant chemistry within allowable limit Surry Nuclear Plant 2004-301 SRO lnital Exam 89. G2. I.4 001/3/niiCH SPEC STAFFlhG/C/A 2.3/3.4/~/SR04301/S/~A~/SI)R

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The following plant conditions exist:

- Unit 1 is shutdown and subcritical by 5.35% delta k / k

- Unit 1 Tavg is 108 O F

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Unit 2 is shutdown and subcritical by 2.35% delta k / k

- Unit 2 Pavg is 190 O F Which ONE of the following correctly states the MINIMUM shift crew composition per Technical Specifications?

A. 1 SS, 1 Unit SRO, 3 ROs, 4 AOs, and 1 ST. 1 SS: 2 Unit SRQs, 3 ROs, 4 AOs, and no ST C. 1 SS, 1 Unit SWO, 3 ROs, 4 AOs, and no ST DY 1 SS: no Unit §BO, 2 ROs, 4 AOs, and no ST Surry References:

Technical Specification Table 6.1 -1 (Minimum Shift Crew Composition), Amendment No. 72 Bistractor Analysis:

A. Incorrect because it does not match the minimum requirements far 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 B. Correct because it matches the requirement for one unit in Cold Shutdown and one Shutdown and one unit in Refueling Shutdow Shutdown and one unit in Refueling Shutdow Shutdown and one unit in Refueling Shutdow unit in Refueling Shutdow G2. Knowledge of shift staffing requirement Surry Nuclear Plant 2004-301

§BO lnital Exam The following conditions exist:

- Unit I is at 50% power

- Unit 2 is in startup mode with Tavg = 41 0 O F

- Unit 2 Steam Driven AFW Pump and one Motor Driven AFW Pump are declared to be inoperable at 0800 hours0.00926 days 0.222 hours 0.00132 weeks 3.044e-4 months on August 11 (all other AFW equipment is operable)

- Unit 2 Motor Driven AFW Pump is restored to operable status at 11 00 hours0 days 0 hours 0 weeks 0 months and Unit 2Tavg=410OF Which ONE of the following sets of Technical Specification actions is correct?

(Reference provided)

A. Initially (with both pumps inoperable) Unit 2 must be in Cold Shutdown by 08/12 at 2000 hours0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months and restore either AFW pump by 08/25 at 0800 hours0.00926 days 0.222 hours 0.00132 weeks 3.044e-4 months or Unit 1 must be placed in Hot Shutdown by 08/25 at 1400 hours0.0162 days 0.389 hours 0.00231 weeks 5.327e-4 months . After the Motor Driven AFW Pump is restored, no Unit 1 or Unit 2 actions would be in effec B. Unit 2 is not required to enter any Technical Specification Action Statement Initially (with both pumps inoperable) Unit 1 must be in Hot Shutdown by 08/25 at j400 hours and Cold Shutdown by 08/26 at 2000 hours0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months . After the Motor Driven AFW Pump is operable no Unit 1 actions would be in effec CY Initially (with both pumps inoperable) Unit 2 must be in Cold Shutdown by 08/12 at 1400 hours0.0162 days 0.389 hours 0.00231 weeks 5.327e-4 months and restore either AFW pump by 08/25 at 0800 hours0.00926 days 0.222 hours 0.00132 weeks 3.044e-4 months or Unit 1 must be placed in Hot Shutdown by 08/25 at 1400 hours0.0162 days 0.389 hours 0.00231 weeks 5.327e-4 months . After the Motor Driven AFW Pump is restored, no Unit 1 or Unit 2 actions would be in effec D. Initially (with both pumps inoperable) Unit 2 shall not enter Hot Shutdown and must be in Cold Shutdown by 08/12 at 1400 hours0.0162 days 0.389 hours 0.00231 weeks 5.327e-4 months and restore either AFW pump within 14 days or Unit 1 must be placed in Hot Shutdown by 08/25 at 0800 hours0.00926 days 0.222 hours 0.00132 weeks 3.044e-4 months . 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 0.222 hours 0.00132 weeks 3.044e-4 months or Unit 1 must be in Hot Shutdown by 08/14 at 2000 hour0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months Surry Nuclear Plant 2004-301 SRQ lnital Exam Surry References:

Technical Specifications 3.6.C, 3.6.F, 3.6.6, and 3.01 Distractor Analysis:

A. Incorrect because Unit 2 must be placed in CSD by 8/12 at 1400 hrs, not 2000 hr B. Incorrect because Unit 2 Technical Specification Actions do apply above 350 O F and 450 psig. LCO 3.6.F only covers the situation where one AFW Pump is inoperable, thus requiring the use of LCO 3. C. Correct because K O 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, no action statements will be in effec. Incorrect because Unit 1 does not need to be shutdown with only the Unit 2 Steam Driven AFW Pump inoperabl (32.2.23 Ability to track limiting conditions for operation Surry Nuclear Plant 2004-301 SRO lnital Exam

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Unit 1 has been shut down for 21 days and fuel movement has just commence Which ONE of the following is correct with regard to Containment Purge Exhaust?

A. Containment Purge Exhaust is normally manually aligned to continuously pass though CAT 1 filters during fuel movement. If a containment radiation monitor high alarm isolates purge flow, fuel movement may continue after purge flow is verified isolate B. Containment Purge Exhaust is normally manually aligned to continuously pass through CAT 1 filters during fuel movement. If a containment radiation monitor failure isolates purge flow, fuel movement may continue after purge flow is verified isolated and radiation levels are verified norma C. Containment Purge Exhaust is normally manually aligned to continuously pass through CAT 2 filters during fuel movement. If a containment radiation monitor high alarm isolates purge flow, fuel movement may continue after purge fEow is verified isolate B I Containment Purge Exhaust is normally manually aligned to continuously pass through CAT 2 filters during fuel movement. If a containment radiation monitor failure isolates purge flow, fuel movement may continue after purge flow is verified isolated and radiation levels are verified norma Surry References:

ND-92.5-LP-7, Refueling Abnormal Procedures, Rev. 10 Distractor Analysis:

A. Incorrect - For fuel movement, Containment Purge is now normally aligned using B. Incorrect - For fuel movement, Containment Purge is now normally aligned using C. Incorrect - if a High Alarm is received fuel movement may not continu B. Correct - For fuel movement, Containment Purge is now normally aligned using VS-F-53, which flows through the CAT 2 filte VS-F-59, which flows through the CAT 2 filte VS-F-59, which flows through the CAT 2 filte G2.2.31 Knowledge of procedures and limitations involved in initial core loadin Surry Nuclear Plant 2004-301 SRO lnital Exam 9 (i2. ~~~~~~~~~~~~~~~~~ 001/3//PKOCEI>URE

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CKANGEEIILIEM

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2. 3 / 3. 3 N I S R 0 4 3 0 1 / S / 1 B ~ ~

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Which ONE of the following correctly states items that require a Regulatory Screen to be performed in accordance with VPAP-3001, Stafety and Regulatory Reviews?

(Reference provided)

A. Emergency Action Level Change AND Station Curve Changes BY Seismic Analyses AND Heating-Ventilation and Air Conditioning Analyses C. Fire Protection Plan Changes AND Plant Flood Analyses D. Offsite Dose Calculation Manual Changes AND Equipment Qualification Analyses Surry References:

VPAQ-3001, Station and Regulatory Reviews, Rev. 9 Distractor Analysis:

A. Incorrect because Emergency Action bevel Changes zre to be processed BAW VPAP-0502 (see VPAP-3001 Page 2 of Att. 3), a Regulatory Screen is not require Plausible because both items are listed on VPAP-3001 Att. 3 Page. Correct per VPAP-3001 Page 2 of At C. Incorrect because Fire Protection Plan Changes are to be performed IAW VPAP-2401 (see VPAP-3004 Page 2 of Att. 3), a Regulatory Screen is not require Plausibie because both items are listed on VPAP-3001 Att. 3 Page Regulatory Screen is not required. Plausible because both items are listed on VPAP-3001 Att. 3 Page ___-

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D. Incorrect because ODCM changes are to be performed IAW VPAP-2103N, a (32.2.6: Knowledge of the process for making changes in procedures as described in the safety analysis repor Suny Nuclear Plant 2004-301 SRO lnitai Exam 9 G2.4.2.Y 001/3/IEMEKGENCY P1,AVMERI ~. ~ / ~. O / N / S R ~ ~ ~ O I / S / ~ I Z H I S I ) K

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Which ONE of the following are all responsibilities that shall Station Emergency Manager?

be delegated by the A. Ordering Site Evacuation, Authorizing Emergency Exposure Limit BY Authorizing Notifications of NRC, State and Local Agencies of the Emergency Status, Authorizing Emergency Exposure Limit Status, Restricting Access to the Sit C. Authorizing Notifications of NWC, State and Local Agencies of the Emergency D. Authorizing Emergency Exposure Limits, Restricting Access to the Sit Surry References:

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 d legate B. Correct because the answer is clearly stated in both of the reference C. Incorrect because restricting access to the site may be delegate D. Incorrect because restricting access to the site may be delegate (32.4.29: Knowledge of the emergency pla Surry Nuclear Plant 2004-301 SRO lnital Exam 94. G2.4.38

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001/3//SEM/MEM 2.2/4.O/NiSR03301/S/MAm/SDR

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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 SRO 3. Site Vice-president, Director Nuclear Station Safety and Licensing, Director Nucteai Station Operations and Maintenance, Manger Nuclear Operations C I Site Vice-President, Director Nuclear Station Qperations and Maintenance, Director Nuclear Station Safety and Licensing, Manger Nuclear Operations B. Site Vice-President, Director Nuclear Station Operations and Maintenance, Manger Nuclear Operations, Director Nuclear Station Safety and Licensing

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suw References:

NB-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 B. incorrect because this is not the preferred order as specified in NB-95.5-LP-2 Pg C. Correct because this is the preferred order as specified in ND-95.5-LP-2 Pg D. Incorrect because this is not the preferred order as specified in NB-95.5-LP-2 Pg G2.4.38: Ability to take actions called for in the facility emergency plan, including (if required) supporting or acting as emergency coordinato Surry Nuclear Plant 2004-301 SWO Bnital Exam

~~ 95. WE01 G2.1.20 001/1/2/SI

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TERMINATION/C/A 4.3/4.2/N/SROJ301/S/MAR/SI~K

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Given the following plant conditions following an automatic reactor trip:

- RCS has been verified to be intact per I - E O, Reactor Trip or Safety Injection

- AFW Flow to "A" SG = 125 gpm

- AFW Flow to "B" SG = 110 gpm

- NR "A" SG bevel = 10%

- NR "5" SG Level = 8%

- RCS Pressure = 1750 psig and slowly rising

- PRZR Level = 24% and slowly rising

~ RCS subcooling based on CETCs is 80°F Operators have reached the point in f - 4 where they are to check if SI flow should be reduced.

- AFW Flow to "C" SG = 130 gpm

- NR "C" SG Level = 9%

Which ONE of the following would be the next series of operator actions?

A:' Direct SBA to begin monitoring Critical Safety Function Status Trees, Reset SI and CLS (E necessary), verify Instrument Air available, then stop all but one Charging Pump, followed by isolating High Head SI to the Cold Leg. Transition to I-ES-1.1, SI Termination, establish letdown, followed by raising Pressurizer level to > 35%, then secure all but one Charging Pum C. Establish letdown, followed by raising Pressurizer level to > 35%, transition to 1-ES-1.1, SI Termination, then secure all but one Charging Pum D. Direct STA to begin monitoring Critical Safety Function Status Trees, Reset SI and CLS (if necessary), verify Instrument Air available, align Charging Pump suction to the VCT, then stop all but one Charging Pum Sur9 Nuclear Plant 2004-301 SRO lnital Exam Sur9 References:

1 -E-0, Reactor Trip or Safety Injection, Rev. 46 1 -ES-l.l, SI Termination, Rev. 29 ND-95-03-03, E O, Reactor Trip of Safety Injection, Rev. 14 Bistractor Analysis:

A. Correct because these actions are directed by 1-E-0 Steps 26 through 3 B. Incorrect because letdown would not be established prior to Pzr L > 35%. Plausible because transition to ES-1.1 is logical and distractor states that the goal is to get Pzr L > 35%.

C. Incorrect because letdown would not be established prior Lo Pzr L > 35%. Plausible because transition to ES-1.1 is logical and distractor states that the goal is to get Pzr L > 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 reverse W/EO1 Wediagnosis and SI Termination G2.1.20: Ability to execute procedure Surry Nuclear Plant 2004-301 SRO Inital Exam

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96. WE03EA2. I 00 I/ I/Z/COOLDOU;N

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3.4/4.2/B/SR0430 l/S/MAB/SDR Operators are responding to a LOCA outside of containment using 1-ECA-I 2, LOCA Outside Containment. The crew efforts to isolate the break are unsuccessfu Which ONE of the following identifies the procedure ECA-1.2 will direct the operators to in order to coo[ and depressurize the reactor coolant system?

A. 1 -I51 I Loss of Reactor or Secondary Coolant B. 1-E%-1 2, Post LOCA Cooldown and Depressurization C. 1-§-I.3, Transfer to Cold Leg Recirculation D1 1-ECA-1.1, boss of Emergency Coolant Recirculation Surry References:

1-E-I, Loss of Reactor or Secondary Coolant, Rev. 21 1423-1 2, Post LOCA Cooldown and Depressurization, Rev. 21 I-ES-I 3, Transfer to Cold Leg Recirculation, Rev. 12 I-ECA-1.1, boss of Emergency Coolant Recirculation, Rev. 17 Distractor Analysis:

A. Incorrect as stated in Distractor 5 Analysis. Plausible because there is a Loss of Reactor Coolant in progres B. Incorrect as stated in Distractor D Analysis. Plausible because the goal is to cool and depressurize the RC C. Incorrect as stated in Distractor D Analysis. Plausible because this is a normal transition for long term cooling during a LOC D. Correct because Step 2 RNO of ECA-1.2 directs operators to ECA-1.1 if efforts to isolate the leak are not successfu WE03 LOCA Cooldown - Depres 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 operation Bank Question PP02301~

Surry Nuclear Plant 2004-301 SRO lnital Exam 97. WE05EA2. I 001i111iMAIN STEAM LINEiCiA. 3.4/4.~ilBiSRU430lIS~MABISDR The following conditions exist:

- A manual Safety Injection was initiated due to a Steam Break in Safeguards

- A11 MSBVs have been closed

- A11 SG pressures are steadily decreasing

- All SG NR levels are off-scale low and WR levels are steadily decreasing

- Pressurizer pressure is steadily decreasing

- RCS temperature is decreasing uncontrollably

- Adequate Auxiliary Feedwater flow exists Which ON of the following is the correct procedure transitions for the event in progress?

Pressurizer level is off-scale low A: E-0 to E-2 to ECA-B. E-0 to E-I to E-2 to ECA-C. E-0 to E-1 to ECA-D. E-0 to E-2 to E-1 Surry References:

1-E-0, Reactor Trip or Safety Injection, Rev. 46 1452, Faulted Steam Generator Isolation: Rev. 9 1 -ECA-2.1, Uncontrolled Depressurization of All Steam Generators, Rev. 19 Distractor Analysis:

A. Correct because E-0 would be entered upon Fix Trip. Step 21 of E-0 sends the team to 2 Step 2 of E-2 sends the team to ECA-B. Incorrect because E-0 Step 21 directs performance of E-2. E-1 is not directed until E-0 Step 2 C. Incorrect because E-0 Step 21 directs performance of E-2. E-? is not directed until B. Incorrect because E-2 would not be entered until after E-1.

WE05 Inadequate Heat Transfer - Loss of Secondary Heat Sink EA2.1: 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 operation Surry ILT Bank Question M342 E-0 Step 2 Sur9 Nuclear Plant 2004-301 SRO lnital Exam 9 WE 1 OEA2.1 00 1/1/2/NATTI RAL CIRCULATION/C/A

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During a Natural Circulation Cooidown IAW ES-0.3, Natural Circulation Cooldown 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 hea Which ON of the following courses of action is appropriate?

A. Initiate FR-1.3 since ES-0.3 assumes FR-1.3 is in effect to eliminate the steam voi B. Initiate SI and go to FW-1.3 to vent the hea C. The NC Cooldown should be stopped and a transition to FR-1.3 should be mad DI Stay in ES-0.3. Void growth is expected and ES-0.3 provides guidance to control the void growt Surry References:

1 -FR-1.3, Repanse To Voids In Reactor Vessel, Rev. 16 143-0.3, Natural Circulation Cooldown With Steam Void in Rx Vessel, Rev. 12 Bistractor Analysis:

A. Incorrect because ES-0.3 does not assume that FR-1.3 is being use. Incorrect because SI shauld not be initiated and there is no need to vent the hea. Incorrect became ES-0.3 does provide guidance for managing void growt B. Correct because ES-0.3 does provide guidance for managing void growt WE10 Natural Cir EA21 : Ability to determine and interpret the following as they apply to the (Natural Circulation With Steam Void in Vessl with / without RVLIS): Facility conditions and selection of appropriate procedures during abnormal and emergency operation Sur9 Wequal Bank Question #247

Surry Nuclear Plant 2004-301 SRO lnital Exam 99. WE 13EA2. I 001/1/2/SG STEAM GEENERATOK/CJA 2. 9 l ~. 4 / ~ / S R ~ 4 3 ~ 1 / S / ~ A ~ ~ S I ~ I ~ ~

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During performance of E-3, Steam Generator Tube Rupture, the operating team is directed to adjust the SG POW setpoint on the ruptured SG to 1035 psig. The Reactor Operator observes ruptured SG pressure to be 1070 psig and the P O W cyclin 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 S E3. Increase feed flow to the ruptured SG bo stop the release and remain in E-C. Increase the setpoint above 1090 psig to prevent release to the public and transition to ECA-3.1, SGTR With Loss of Reactor Coolant - Subcooled Receover D I Leave the P O W setpoint at 1035 psig to minimize challenges to the SG Code Safeties and remain in E-Surry Nuclear Plant 2004-301 SRQ Inital Exam Surry References:

1-E-3, Steam Generator Tube Rupture, Rev. 25 NB-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 PQRV seats when pressure drops below 1035 psig. Furthermore, FR-H.2 is not associated with any Red or Orange path when pressure drops below 1035 psig. Furthermore, feeding a ruptured SG will not limit the exposure to the publi when pressure drops below 1035 p i g. Furthermore, this action may challenge the code safeties, which is not desirabl for the ste. Incorrect because the correct response is simply to verify that the PQRV seats C. Incorrect because the correct response is simply to verify that the POWV seats D. Correct because this is the correct direction in the procedure and the correct h s i s W E 1 3 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 operation Surry Requal Exam Bank Question #324

Surry Nuclear Plant 2004-304 SRO lnital Exam The Control Room Operators are performing FR-S.2, Response to boss of Core Shutdown, in response to a yellow path condition shown on the Critical Safety Functior (CSF) status tre 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 boss of Core B. The operators must leave this procedure before completion and go to FR-H.1, Response to boss of Secondary Heat Sink, if the heat shk CSF status tree indicates a yellow path conditio 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 conditio Response to Containment Flooding, if the containment CSF status tree indicates ar orange path conditio Shutdown CSF adverse condition has cleared. (Green path established)

D The operators must leave this procedure before completion and go to FR-Z.2,

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Surry Refernces:

ND-35.3-LP-26, Critical Safety Function Status Trees, Rev. 5 Distractor Analysis:

A. Incorrect because the operator does not have to immediately leave FR if it is not complete B. Incorrect because yellow path does not warrant this actio C. Incorrect because yellow path does not warrant this actio D. Correct because orange path takes priorit WE15 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 operation Surry ILT Bank Question # 1350

SRO Reference Documents Provided for Initial Exam 2004-301 1) EPIP-1.01, Emergency Action Level Table Rev. 44 2) Annunciator Response Procedure 1 D-C6, PRZR PWR RELIEF: VV LO AIR PRESS Rev. 5 3) VPAQ-2802, Notifications and Reports Rev. 17 4) VPAP-3001, Safety and Regulatory fleviews Rev. 9

RHR FLOW REQUIREMENT VERSUS TIME AFTER SHUTDOWN RHR FLOW RATE ( G P M )

5000 4500 4000 3500 3000 2500 2000 1500 1000 500

0 100 200 300 400 500 600 700 800 900 1000 TIME AFTER SHUTDFNN (HOURS)

MINIMUM RCS LEVEL VERSUS RHR FLOW ( 1 -RC - LI - 1 OOA)

MI 1.6 1.4 1.2

' N I M U M RCS LEVEL FT (1-RC-LI-1OOA)

1

500 1000 1500 2000 2500 3000 3500 4000 4500 5000 RHR FLOW RATE (GPM)

SURRY LORP EQUATION SHEETS Q = mc,AT g=mAh Q = UAAT

2 m = - m v 2 w = V A P wvg = mAFv Pwr = Wf m Pwr=WfAh Net Work Out Energy In Cycle Effeciency =

1

s = vot +-at

v = sit

w = -

t f =ma w = m g PE = mgn F=PA th = v,Ap k = p 4 V 32.21bm - j?

lbf -secZ R, =

VCCN H, a N 2 BHP 0: N'

LV2 H, = f-2 0 1Mw = 3. 4 1 ~ 1 0 ~ ~ ~ % ~

lhp = 2. 5 4 ~ 1 0 ~ Bfgr 1 Btu = 7 7 8 j lbf OC = (5/9)("F-32)

"F = (9/5)("C) + 32 lkg = 2.21 lbrn 13' = 7.48 gal

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'TS 3.6-1 03-12-01 TURBINE CYCLE Applicability Applies to the operating status of the Main Steam and Auxiliary Feed System Obiectives To define the conditions requited in the Main Steam System and Auxiliary Feed System for protection of the steam generator and to assure the capability to remove residual heat from the core during a loss of station power/or accident situation Specification A. A unit's Reactor Coolant System kmperature or pressure shall not exceed 350°F or 450 psig, respectively, or the reactor shall not be critical unless the five niain steam line code safety valves associated with each steam generator in unisolated reactor coolant loops are OPERABLE with lift settings as specified in Table 3.6-IA and 3.6-1 B. To assure residual heat removal capabilities, the following conditions shdl be met prior to the commencement of any unit operation that would establish reactor coolant system conditions of 350°F and 450 psig which wouid preciudc operation of the Residual Beat Removal System. The following shall apply:

1. Two motor driven auxiliary feedwater pumps shall be OPERABL. A minimum of 96,000 gallons of water shall he available in the protected condensate storage tank to supply emergency water to the auxiliary feedwater pump suction. All main steam line code safety valves, associated with steam generators in unisolated reactor coolant loops, shall be OPERABLE with lift settings as specified in Table 3.6-1A and 3.6-1 Amendment Nos. 224 and 224

TS 3.6-2 06-57-99 4. The auxiliary feedwater cross-connect capability shall he available, as follows:

a. Two of the three auxiliary feedwater pumps on the opposite unit (automatic initiation instrumentation need not be OPERABLE) capable of being used with the opening of the crowconnec b. A minimum of 60,000 gallons of water available in the protected condensate storage tank of the opposite unit to supply emergency water to the auxiliary feedwater pump suction of that uni c. Emergency power supplied to the opposite unit's auxiliary feedwater pumps and to the AFW cross-connect valves, as follows:

1. Two diesel generators (the opposite unit's diesel generator and the shared backup diesel generator) OPERABLE with each generator's day tank having at least 290 gallons of fuel and with a minimum on-site supply of 35,000 gallons of fuel availabl. Two 4160V emergency buses energize. Two OPERABLE flow path5 for providing fuel to the opposite unit's diesel generator and the shared backup diesel generato. Two station batteries, two chargers and the DC distribution systems OPERABL. Emergency diesel generator battery, charger and the DC control circuitry OPERABLE for the opposite unit's diesel generator and for the shared back-up diesel generato. The 480V emergency buses energized which supply power to the auxiliary feedwater cross-connect valves:

a. For AFW from Unit 1 to Unit 2: Buses lHl and 1J b. For AFW from Unit 2 to Unit 1: Buses 2HI and 23 Amendment Nos. 220 and 220

TS 3.6-3 06-07-99 7. One ofthe two physically independent circuits from the offsite transmission network energizing the opposite units emergency buse I C. Prior to reactor power exceeding IO%, the stem driven auxiliary feedwater pump shall be OPERABL D. System piping: valves, and wntrol board indication required for operation of the components enumerated in Specifications 3.6.B and 3.6.C shall be OPERABLE (automatic initiation instrumentation associated with the opposite units auxiliary feedwater pumps need not be OPERABLE).

E. The specific activity of the secondary coolant system shall be 5 0.10 k@i/cc DOSE EQUIVALENT 1-131. If the specific activity of the secondary coolant system exceeds 0. IO pCi/cc DOSE EQUIVALENT I-13 1, the reactor shall be shut down and cooled to 500°F or less witRin 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after detection and in COLD SHUTDOWN within the following 30 hour3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months F. With one auxiIiary feedwater pump inoperable, restore at least three auxiiiay feedwater pumps (two motor driven feedwater pumps and one steam driven feedwater pump) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUTDOWN within the following 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months I G. The requirements of Specifications 3.6.B and 3.6.D above concerning the opposite units auxiliary feedwater pumps; associated piping, valves, and control board indication; and the protected condensate storage tank may be modified to allow the following componenes to be inoperable, provided immediate attention is directed to making repair I

1. One train of the opposite units piping? valves, and control board indications or two of the opposite units auxiliary feedwater pumps may be inoperable for a period not to exceed 14 day Amendment Nos. 220 and 220

TS 3.6-4 06-07-99 2. Both trains of the opposite units piping, valves, and control board indications; the opposite units protected condensate storage tank; the cross-connect piping from the opposite unit; or three of the opposite units auxiliary feedwater pumps may be inoperable for a period not to exceed 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . A train of the opposite units emergency power system as required by Section 3.6.B.4.c above may be inoperable for a period not to exceed 14 days; if this trains inoperability is related to a diesel fuel oil path, one diesel he1 oil path may be inoperable for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months provided the other flow path is proven OPERABLE; if after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the inoperable flow path cannot be restored to service, the diesel shall be considered inoperable. During this 14 day period, the following limitations apply:

a. If the offsite power source becomes unablc to energize the opposite units OPERABLE train, operation may continue provided its associated emergency diesel generator is energizing the OPERABLE trai b. If the opposite units OPEKABLE trains emergency diesel generator becomes unavailable, opcration may continue for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided the offsite power source is energizing the opposite units OPERABLE trai c. Return of the originaily inoperable train to OPERABLE status allows the second inoperable train to revert to the 14 day limitatio If the above requirements are not met, be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the next 30 hour3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . me requirements of Specification 3.6.B.2 above may be modified to allow utilization of protected condensate storage tank water with the auxiliary steam generator feed pumps provided the water level is maintained above 60,000 gallons, sufficient replenishment water is available in the 300.000 gallon condensate storage tank, and replenishment of the protected condensate storage tank is commenced within two hours after the cessation of protected condensate storage tank water consumptio Amendment Nos. 220 and 220

IR 05000381/2004301

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v t e Inspection Report - - 2004301 (Operator Exams)
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2004 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (0) Assessments Mid Cycle End of Cycle Security Operator Exams 301(0) Emergency Planning Other

IR 05000381/2004301

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