Text

Final - Written (RO & SRO) W/Answer Key
	ML040970260
Person / Time
Site:	Calvert Cliffs
Issue date:	03/05/2004
From:	Sickle J; Constellation Nuclear
To:	Conte R; NRC/RGN-I/DRS/OSB
	Conte R
References
	50-317/04-301, 50-318/04-301
	Download: ML040970260 (141)
	v • d • e

i

)

ES-40 1 Site-Specific RO Written Examination Form ES-401-87 Cover Sheet U.S. Nuclear Regulatory Commission Site-Specific RO Written Examination Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets.

achieve a final grade of at least 80.00 percent. Examination papers will be collected six hours after the examination starts.

To pass the examination you must Applicant Certification Ail work done on this examination is my own. I have neither given nor received aid.

_ _ _ _ - _ _ ~ - - -

Applicant's Signature Results Examination Value Points Applicant's Score

-~

Points Applicant's Grade Percen 31 of 34 NUREG-1021, Draft Revision 9

Name:

2.

U-2 is operating at 100% when a reactor trip occurs.

The RO observes the following indications on the CEA mimic:

4 CEAs do not have the amber lights energized 2 of the above CEAs have green lights energized What must be performed when performing the Reactivity Control Safety Function?

A. Take the alternate actions to deenergize CEDM MG sets and verify all CEAs are B. Depress the Reactor Trip pushbuttons on 2C15, verify reactor power is C. Verify reactor power is lowering, check that no CEA deviation alarms are present, inserted.

lowering, and verify a negative startup rate exists.

verify a negative startup rate exists, check that RCS makeup is secured and inform CRS that Reactivity Control is complete.

pump using all available charging pumps.

D. Commence RCS boration to at least 2300 ppm via gravity feed or a Boric Acid Given the following conditions:

RCS pressure:

1600 PSlA PZR level:

360" T h:

532.5"F Tc:

532.2"F Containment Pressure:

2.4 PSlA Containment temp:

115°F S/G pressures:

880/885 PSlA EOP-0 is being implemented What is the most likely cause of these conditions?

A. RCS cold leg break B. RCS leak at the top of the PZR C. S/G tube leak D. Main Steam line break in containment 2004RO.TST Version: 0 Page: 1

3. The CRS ordered Unit-I manually tripped due to rapidly loweing PZR level and RCS pressure. EOP-0 is being implemented and the following conditions exist:

RCS pressure:

1900 PSlA Tc:

532.5"F Containment pressure 0.5 PSlA Containment temperature 98" F What is the status of the Containment Air Coolers? (Assume no operator action)

A. 4 Coolers in slow speed with maximum SRW flow B. 4 Coolers in fast speed with normal SRW flow C. 3 Coolers in slow speed with maximum SRW flow D. 3 Coolers in fast speed with normal SRW flow

4.

2004RO.TST Version: 0 Page: 2 Which one of the following describes the credited RCS inventory and core heat removal processes during a large break LOCA?

A. HPSl injection provides makeup and heat is removed via natural circulation flow to B. HPSl pumps, LPSl pumps and the SlTs provide makeup and heat is removed via C. LPSl pumps and the SlTs provide makeup and heat is removed via forced flow to D. Charging pumps provide makeup and heat is removed via flow out the break.

the S/Gs.

flow out the break.

the S/Gs.

5. Given the following conditions:

-1 1A RCP tripped due to a breaker fault

-EOP-0 has been completed, no alternate actions were required How will the RCS and Steam Generators have responded?

A. 11 and 12 loop differential temperatures will be equal and 11 and 12 S/G pressures B. 11 loop will have an inverted differential temperature and 1 I S/G pressure will be C. 12 loop differential temperature will be greater than 1 1 loop differential temperature D. 12 loop will have a smaller differential temperature than 11 and 12 S/G pressure will will be equal.

lower than 12 S/G pressure.

and 11 and 12 S/G pressures will be equal.

be lower than 1 1 S/G pressure.

6.

Why does the Component Cooling system realign on a SIAS?

A reactor trip has occurred and the following conditions exist:

-- Pressurizer level is 140 inches and stable

-- One Charging Pump is available

-- Pressurizer pressure is 1900 psia and rising

-- RCS Subcooling is 65°F and steady After performing the immediate actions for PIC, the Reactor Operator reports "Pressure and Inventory Control cannot be met" to the CRS.

What is the reason for this report?

A. Letdown has been isolated.

6. RCS subcooling is not in band.

C. All Charging Pumps are not in operation.

D. Pressurizer level is not trending toward setpoint.

A. Minimize dose rates due to contamination of Component Cooling system B. Provide long term cooling to containment after RAS

c. Minimize load on the Saltwater system to ensure containment cooling via Service D. Provide continuous cooling to LPSl pump seals Water

8. Unit-I is in Mode 3 with the following conditions:

RCS pressure is 2150 PSlA and lowering Pressurizer Spray Valves, 1 -RC-1 OOE and F are fully open PIC-I OOX is indicating 2400 PSIA, controller output is 100%

PIC-IOOY is indicating 2150 PSlA output is 0%

What action is required?

A. Stop 1 I A and 11 B Reactor Coolant Pumps.

B. Energize all Pressurizer Heaters.

c. Place PZR PRESS CH SEL switch, 1-HS-100 in 'Y'.

D. Place PRESSURIZER SPRAY VALVE CONTROLLER, 1 -HIC-100 in manual with a 100% output.

2004RO.TST Version: 0 Page: 3

9. EOP-8 has been implemented because the Reactivity Safety Function was not met in EOP-0. What indications are used to verify that boration is successfully meeting the acceptance criteria?

A. WRNl power is less than IO4% and SUR is negative or zero B. LRNI power is less than IO4% or SUR is negative

c. A boric acid pump is running and charging header flow is 40 GPM or greater D. SUR is zero and the CHG HDR FLOW LO PRESS LO alam is clear 1 C.

2004RO.TST Version: 0 Which one of the following is the reason for equalizing the pressure on the primary and secondary sides of a ruptured Steam Generator per the applicable EOP?

A. Lowering the RCS pressure allows HPSI flow to restore Pressurizer level.

B. Reducing the differential pressure lowers the RCS leak rate.

C. Reducing RCS pressure and temperature aids initiation of natural circulation.

D. Equalizing RCS and S/G secondary side pressures initiates backflow to control affected S/G level.

Page: 4

11. Emergency Operating Procedures provide specific guidance for feeding a dry S/G to restore RCS heat removal.

This guidance is based on (Select the phrase that correctly completes the above statement)

A. minimizing S/G tube voiding, which would inhibit natural circulation B. preventing a rapid RCS cooldown, avoiding a pressurized thermal shock to the C. preventing uneven cooling of the RCS, which may result in a localized reactivity D. minimizing the probability of creating a waterhammer, and damaging S/G Reactor Vessel excursion internals

1 4 EOP-0 was completed and the following conditions exist:

I 11 S/G pressure 725 PSlA and lowering 12 S/G pressure 840 PSlA and stable 11 S/G level

-260" and lowering 12 S/G level

- 80l and rising slowly Tc 521 "F Pzr pressure 1830 PSlA MSlVs and S/G Blowdown Isolation Valves are shut

13.

Which event would cause these indications?

A. A Feedwater line rupture inside Containment B. An RCS leak inside Containment C. A Main Steam line rupture in the Turbine Building D. A rupture of the S/G Blowdown Tank EOP-7 (Station Blackout) has been initiated on U-I and the CRS has directed the CRO to restore power to 11 4KV bus using the OC Emergency Diesel Generator.

Which Control Room annunciator condition reflects that the bus has been re-energized?

A. "ACTUATION SYS LOSS OF POWER alarm clears B. "ACTUATION SYS U N RELAY TRIP" alarm clears C. "SEQUENCER INITIATED" alarm actuates D. "OC DG CONTR BOARD 1 C19c" alarm actuates

14. Unit-1 has experienced a Loss of Offsite Power from 100% power.

11 and 14 4 KV buses have been re-energized by their associated Diesel Generators.

EOP-0 is being implemented.

What action must the CRO take for step B, "ENSURE TURBINE TRIP" that would NOT be expected on a reactor trip with offsite power available?

A. Depressing the Turbine TRIP button.

B. Opening I 1 GEN FIELD BKR, 1-CS-41.

C. Shutting the MSIVs due to not being able to verify Turbine speed dropping.

D. Dispatching an operator to shut the MSR 2nd stage bypass valves.

2004RO.TST Version: 0 Page: 5

How is 4 KV bus 1 I affected?

A. 1A DG will start but not load because ESFAS logic cabinet ZA remains deenergized, maintaining a UV (load shed signal) to 4 W bus 11 loads.

6. 4 W bus 11 will be re-energized by manually starting and loading OC Diesel Generator.

c. 1A DG will automatically start and load to energize 4W bus 11 after the 1 B DG starts and energizes 4KV bus 14.

D. 4 W bus 11 cannot be re-energized until power is restored to 1YO1 via DC bus

16. A plant transient has ocurred and the following conditions exist:

--All Unit-1 annunciator lights are deenergized

--CC CNTMT RETURN, 1 -CC-3833-CV has failed shut How will SPDS indicate the cause of this event?

A. All Safety Function boxes turn red, and a "Loss of AC bus alarm appears on the Vital Auxiliaries" Safety Function screen.

B. The 'Vital Auxiliaries" Safety Function box turns red, and the indicator for the affected AC bus on the electrical systems mimic flashes.

c. The "Vital Auxiliaries" Safety Function box turns yellow, and the indicator for the affected DC bus on the electrical systems mimic changes color.

D. All Safety Function boxes turn magenta and a small red box appears next to the indicator for the affected DC bus on the electrical systems mimic.

2004RO.TST Version: 0 I?.

Page: 6 2-HS-5155,2W22B SRW HXR EMERGENCY OUTLET VLVS handswitch is inadvertantly placed in 'OPEN'.

How are the Service Water Heat Exchangers affected?

A. 22N22B SRW HXR emergency outlets valves open, but normal SW flow is B. 22N22B SRW heat exchangers are removed from service because the heat

c. 21N21 B SRW heat exchangers lose SW flow because the emergency overboard maintained because the emergency overboard valve is normally gagged shut.

exchangers' SW inlet valves will also shut.

valve automatically opens, and 22N22B SRW heat exchangers SW outlets shift to 21 SW supply header.

D. 21N21 B SRW heat exchangers' SW inlet and outlet valves automatically shut, and 22N22B SRW heat exchangers will be supplied by 21 SW header.

18.

1 Group 5 CEAs were being withdrawn from 120.5" to 128.0" using Manual Sequential.

CEA-1 dropped to 124.25" by secondary indication after CEDS was turned off.

Primary and secondary position indication for all other CEAs in the group is 127.25" to 128.0".

What is the expected primary position indication for CEA-I?

A. 127.25l to 128.OI

3. 124.25

c. 120.5" D. 0" 1

Unit-2 is at full power with 21 Plant Air Compressor in Standby when a leak causes Instrument Air header pressure to decrease to 85 psig. Plant Air header pressure is 95 psig. No operator actions have been performed.

Which list is composed of all the air compressors that will be running?

A. 21 and 22 Instrument Air Compressors, 11 and 21 Plant Air Compressors, 21 and B. 21 and 22 Instrument Air Compressors, 11 Plant Air Compressor, 21 and 22 C. 21 Instrument Air Compressor and 11 Plant Air Compressor D. 21 and 22 Instrument Air Compressors and I 1 Plant Air Compressor 22 Saltwater Air Compressors Saltwater Air Compressors

20.

2004RO.TST Version: 0 Given the following conditions:

Unit-2 is on Shutdown Cooling, RCS temperature is 120°F RCS pressure is 14.7 PSlA The reactor vessel head is fully tensioned Reactor Trip Circuit Breakers are open One of two operable WRNI channels has failed low What action is required immediately?

A. Commence boration of at least 40 GPM until RCS boron is 2300 PPM or greater.

B. Suspend all operations involving positive reactivity additions.

C. Commence actions to restore two WRNI channels to operable status.

D. Perform SDM verification per Surveillance requirement 3.1.I.I.

Page: 7

2 4hich of the following would be classified as a fuel handling incident per AOP-6D?

I.4. A large object was dropped in the Spent Fuel Pool and is laying on top of a spent i

fuel assembly.

i 3. During refueling of the core, a fuel assembly was placed in an incorrect core 1

location.

' C. A new fuel assembly was dropped when being moved from the New Fuel Storage i

1 I

1 D. A portable light pole hanging off the refueling machine bridge was damaged when i

1 I

I I

I Area to the New Fuel Inspection Platform.

performing the refueling machine operational checks per 01-25C.

23.

2 f "RMS PANEL 1 C26" alarm at 1 C18 has annunciated.

2-RI-7010, Unit-2 BAST room Area Radiation Monitor is reading off-scale, high.

No other indications of abnormal conditions are present.

What action is directed by plant procedures?

A. Contact Chemistry to obtain samples of the BASTS, VCT and RCS.

6. Recommend Radiation Safety Supervision post the area.

c. Obtain CRS permission to bypass the alarm to clear the alarm at 1 C18.

D. Sound the emergency alarm, evacuate the immediate area and declare a Radiological Event per ERPIP 3.0.

During a severe fire in the Control Room, (AOP-SA), why are the Fairbanks Diesel Generators shutdown?

A. To prevent overloading the Diesel Generators when equipment starts, as the

6. To ensure fuel is conserved for continued extended operation of the OC Diesel C. To protect the engine from damage due to loss of cooling D. To ensure MCCs 104 and 114 are de-energized to keep PORVs from failing open sequencers may not be operable.

Generator

24. On Unit-2 PAMS, what does a single "?" next to a CET temperature indication signify?

A. The indication is outside the quality check parameters.

6. The CET is the highest reading CET in it's quadrant.

c. The CET has been "bypassed", and the value is an old, non-updated indication.

D. The indication is a calculated value, not an actual temperature measurement.

004RO.TST Version: 0 Page: 8

25.

2r: Using provided references, given the following Unit-2 information:

Reactor Power:

100%

1 Tc:

547.7"F and steady i

Letdown flow:

30 GPM Charging flow:

135 GPM Lowering at 2.5 inchedminute I

RCS pressure:

221 0 PSlA and slowly lowering i

Total CBO flow:

6 GPM 1 A. 135 What is the approximate RCS leak rate, in GPM?

, D. 172 Which phrase describes the relationship of RCS activity to the Process Rad Monitor?

A. detects increases in specific isotopes due to fuel failures The Process Radiation Monitor:

2 2004RO.TST Version: 0 Page: 9

{-Which one of the following conditions would allow you to exit EOP-8?

A. A plant cooldown has been completed, shutdown cooling flow has been established, and Core/RCS Heat Removal and Pressure/lnventory safety function status checks for EOP-8 are met.

B. All the safety function acceptance criteria for success paths implemented are being met, a single event diagnosis can be made and intermediate safety function status checks for single event are being met.

C. The CRS or STA has analyzed plant conditions and has verified that steps in an optimal recovery procedure, or an Operating Procedure, will address the safety functions such that EOP-8 final acceptance criteria for all the safety functions will be met.

D. In the case of multiple events, one event has been terminated, (such as a when the affected S/G goes dry during an ESDE) and all intermediate safety function status checks for EOP-8 are being satisified.

' A. To prevent the 4KV degraded voltage relays from actuating upon RCP start.

B. To prevent tripping the oil lift pumps on low voltage when the first RCP is started.

C. Ensures that the running component cooling pump will operate within its design

0. Ensures that excessive starting current is not developed which could damage RCP voltage limits.

windings.

23. Given the following plant conditions:

-- Unit One has tripped due to a Loss of Offsite Power

-- 11 and I 4 4KV busses are energized from the EDGs

-- Pzr level is 100' and slowly lowering How does this affect charging pump operation to restore Pzr level?

A. One charging pump starts automatically, the other charging pumps must be manually started and will stop automatically when Pzr level reaches +I 3 inches above program.

stop on Pzr level deviations from program.

automatically when Pzr level reaches +I 3 inches above program.

operated manually to control pressurizer level.

B. All 3 charging pumps must be started manually and will receive no signals to

c. All 3 charging pumps must be started manually and the backup pumps will stop D. One charging pump starts automatically the other charging pumps must be Which of the following is a possible cause when the following alarm has actuated?

3c.

--On panel 1 C19 "U-I 4KV Eng SF Motor Overload" A. 152-1 204 (1 1 Condensate Booster Pump breaker) tripped B. 152-1 1 14 (U-440-I 1 A high side Feeder) tripped

c. 152-1 104 (1 1 LPSl Pump breaker) tripped D. 152-21 07 (21 Containment Spray Pump breaker) tripped 2004RO.TST Version: 0 Page: 10

. During recovery from a LOCA on U-2, you are directed by the U-2 CRS to reset SIAS from the control room using the implemented EOP. Containment pressure is 2.0 psig and PZR pressure is 800 psia. What sequence of actions must occur to complete this action?

32.

A. Match required handswitches per the EOP attachment, block PZR pressure SIAS, B. Block Pzr pressure SlAS and depress either SlAS channel reset pushbutton.

C. Match required handswitches and depress both SlAS channel reset pushbuttons.

D. Block the Pzr pressure SlAS and depress both SIAS channel reset and depress both SlAS channel reset pushbuttons.

pushbuttons.

Unit 2 is in Mode 1 at 100% power when a loss of Component Cooling occurs.

Which condition from this event alone would require a manual Reactor trip?

A. Main Generator gas temperature of greater than 48°C for at least 15 minutes.

B. RCP bleed off temperature of 200°F.

c. Component Cooling heat exchanger outlet temperature of 175°F.

D. Letdown is automatically isolated due to high temperature.

33. Unit 1 is in Mode 5, preparing for a plant heatup.

E01, QUENCH TK TEMP LVL PRESS is in alarm on 1 C06.

Given the following Quench Tank parameters:

1)

Pressure is 12 PSlG

2)

Temperature is 105°F

3)

Level is 29 inches What action is required?

A. Open WGS CNTMT ISOL valves, WGS-2180,218I-CVs and open QT VENT, B. Place PORV handswitches, 1-HS-1402 and 1-1404 in "OVERRIDE"

c. Open Quench Tank Drain, 1-RC-401-CV D. Open Containment Nitrogen Supply Valve, 0-N2-238.

1 -RC-400-CV.

2004RO.TST Version: 0 Page: 11

RCS pressure is 1600 PSlA Tc is 532.4"F Contaiment pressure is 2.2 PSIG No other malfunctions occurred.

How many Component Cooling Pumps would be running, assuming no operator actions?

A. 0

35. RCS pressure is initially 2250 PSIG.

Spray Valve Controller, 1-HIC-100 fails to a 0% output.

What is a direct result of this failure?

A. All Backup Heaters will energize if in "Auto".

B. Spray Valves 1 -RC-1 OOE and F will fully open.

c. All Backup heaters will deenergize.

D. Proportional heaters will receive full power 2004RO.TST Version: 0

36.

Page: 12 Unit-2 is at 16% power, with the Turbine Generator having just been paralleled with the grid.

A malfunction in RPS channel B causes the Power Trip Test Interlock (PTTI) to actuate.

How is the Turbine Generator affected?

A. A turbine trip will result due to ESFAS B logic cabinet initiating a turbine trip signal.

B. Trip logic will be reduced to 1 out of 3, since channel B Loss of Load trip unit will C. The Turbine Generator will not be affected since the Loss of Load Trip is disabled.

D. RPS will initiate a Turbine Trip signal, but the signal is bypassed at ESFAS due to actuate.

low reactor power.

31. Using provided references:

If 1Y03 were de-energized, which RPS matrix power supply lights at I C15 would be extinguised?

A. 5and15 B. 5,9and7 C. 8,12and 15 D. 9 and 10 38 2004RO.TST Version: 0 A S/G tube rupture has been diagnosed, the correct EOP has been implemented and the following conditions exist:

RCS pressure is 1280 PSlA RCS temperature is 485°F PZR level is 85l 11A and 12B RCPs are running Cooldown rate is 95"F/hr, using TBVs The affected S/G has been isolated and pressure is 700 PSlG What action is required?

A. Secure the remaining RCPs to prevent exceeding pump curve limits.

B. Throttle HPSl flow to allow for backflow from the affected S/G as RCS C. Lower TBV controller output to avoid exceeding cooldown rate limits when HPSl D. Increase RCS depressurization using Main Spray to lower the leak rate into the depressurization continues.

injection begins.

affected S/G.

Page: 13

39. Part of the 2003 modification to the LOCI sequencer advanced the start of the Service Water pumps from step 4 to step 0.

Why was this modification made?

A. Prevents overloading the Emergency Diesel Generators B. Prevents tripping the supply breakers for the safety related 4 KV buses C. Prevents damage to the Service Water Pump motors caused by excessively high D. Prevents a rupture of the Service Water system caused by water hammer in the starting currents Containment Air Coolers

4(

41.

2A Diesel Generator is being taken out of service for routine maintenance.

Which component is also potentially affected? ASSUME NORMAL SYSTEM LINEUPS A LOCA has occurred, SIAS initiated and RWT level is 7 feet and lowering.

A. 21 Containment Spray Pump

8. 22 Charging Pump C. 12 SFP Cooling Pump D. 23 Saltwater Pump

42. With reactor power at 25%, what indication is available to the operator to monitor a S/G tube leak of 5 GPD?

2004RO.TST Version: 0 Page: 14

The following conditions exist on Unit 2:

Reactor/Turbine trip has just occurred (Power prior to trip--1 00%)

S/G pressures are currently 850 psig What operator action (in the Control Room) must initially be taken to prevent an overcooling of the RCS per EOP-O?

A. Press "Close Valves" button on the turbine control panel.

B. Press "Reset" button on the MSR control panel.

C. Shut the MSIVs.

D. Press the BFV "reset" buttons.

4d.

4. Unit 1 is operating at 50% power.

An electrical system malfunction occurs resulting in the loss of 12 and 13 Condensate Pumps.

What is the effect of this transient, and what action must be taken?

A. Reduced feed flow to the S/Gs and lowering levels will result. Bias feed pumps as required to maintain S/G levels.

B. Lower SGFP suction pressure will exist. Verify a Condensate Booster Pump automatically starts.

C. Reduced feed flow to the S/Gs and lowering levels will result. Trip the reactor and implement EOP-0.

D. Low suction pressure to the SGFPs and runout of the operating Condensate Pump will result. Reduce power to maintain condensate header flow less than 8,000 GPM.

4 Unit-1 is at 100% power. Both feedwater flow transmitter signals from 12 S/G to DFWCS fail low (out of range).

How is 12 FRV, 1 -FW-1 121-CV, affected?

A. The last good feed flow input is used and 12 FRV control is shifted to the Backup B. Both CPUs fail and 12 FRV controller is shifted to "MANUAL".

C. An "1 1/12 S/G FW CONTR XFER INHIBIT" alarm is received, a shift from high CPU.

power to low power control mode will not occur and 12 FRV will be controlled by the Backup CPU.

in single element control.

D. Steam flow/feed flow error signal is not used and the Main CPU operates the FRV 2004RO.TST Version: 0 Page: 15

Unit-2 was initially at 100% power when a major plant transient occurred. The following conditions exist:

RCS pressure is 1800 PSlA Containment pressure is 0.4 PSIG 21 S/G pressure is 865 PSlG 22 S/G pressure is 680 PSlG

47.

Which list correctly identifies Main Feedwater/Condensate system automatic actions?

A. Both SGFPs trip, all Condensate Pumps trip, both Heater Drain Pumps trip, only 22 Main Feed MOV and 22 MSIVs shut.

B. Both SGFPs trip, all Condensate Booster Pumps trip, both Heater Drain Pumps trip, both Main feed MOVs shut and both MSlVs shut.

c. Both SGFPs trip, all Condensate Booster Pumps trip, both Heater Drain Pumps trip, only 22 Main Feed MOV and 22 MSlV shut.

D. Both SGFPs trip, all Condensate Pumps trip, all Condensate Booster Pumps trip, both Heater Drain Pumps trip. both Main Feed MOVs shut and both MSlVs shut.

What is the basis for the AFW flow controller automatic setpoints of 150 GPM?

A. S/G levels will be restored to EOP-1 limits within I O minutes of AFAS actuation with MFW isolated, and AFW suction piping flow limits are not exceeded.

B. EDG ratings are not exceeded on SlAS with a Loss of Offsite Power, and S/G inventory is adequate for worst case decay heat with 2 trains of AFW operating.

C. AFW flow will be adequate with one AFW train to remove highest decay heat, but low enough to prevent initiating SIAS due to RCS overcooling with 2 trains operating.

of AFAS Block to the affected S/G with no operator action, yet low enough to prevent RCS cooldown to less than 525°F with one train operating.

D. AFW flow will be adequate to maintain S/G level in the unaffected S/G in the event 2004RO.TST Version: 0 Page: 16

A. 275 GPM B. 300GPM C. 600 GPM D. 900 GPM

49. Unit-I is at 100% power when 13B 480 Volt Bus is lost.

What is the major affect to the plant, and what action must be taken?

A. Boration via the R W from all operable Charging Pumps causes power to decrease. Place 2 Charging Pumps in Pull-To-Lock and shift suction back to the VCT.

B. All Circulating Water Pumps lose excitation. Trip the Reactor and implement C. Feedwater Heater Level Dump Valves fail open, reactor power increases. Reduce D. 12 and 13 Condensate Pumps' bearing temperatures rise due to loss of lube oil EOP-0.

Reactor power, match HLDV handswitches and tie lYO9 and IYIO.

cooling. Tie MCCs 106T and 116T.

2004RO.TST Version: 0

50.

Page: 17 The 13 HPSl pump breaker charging spring has failed to charge after securing the pump for an STP. How will this condition be detected in the control room?

A. 13 HPSl PP BKR L/U IMPR alarm

6. 13 HPSl PP SlAS BLOCKED AUTO START alarm

c. U-I 4KV ESF MOTOR OVERLOAD alarm D. ACTUATION SIGNAL BLOCKED alarm

f 1A Diesel Generator is out of service for maintenance when a Loss of Offsite Power I occurs.

j 2B Diesel Generator did not load due to a faulted 4 KV bus.

1 What affect does this have on the DC electrical distribution system as indicated at I

I I 1c24A7 A. 11 DC bus will be supplied only by I 1 battery.

6. 21 DC bus will be supplied by 21 battery charger.

C. 12 DC bus will be supplied by 24 battery charger.

D. 22 DC bus will be supplied by 22 battery charger.

A. Alpha

6. Beta C. Gamma What type of radiation do the Component Cooling, Service Water and S/G Blowdown Recovery (process rad. monitors) detect?

53. During normal operation at 100% power, what is the largest heat load on the Service Water system?

A. Main Generator Hydrogen Coolers.

B. Hydrogen Seal Oil Coolers C. Containment Air Coolers D. 16 Diesel Generator

. After a SlAS actuation, what is the source of Instrument Air supplied to the AFW flow control valves?

A. Saltwater Air Compressors B. The opposite unit's Plant Air Compressor C. Auxiliary Feedwater system air accumulators D. Nitrogen backup to Instrument Air 2004RO.TST Version: 0 Page: 18

f i \\/Vhich of the following is a requirement for a containment entry at power?

I (without using CMI bypass features)

A. Tavg-Tref deviation alarm.

8. Group 5 CEAs below the PDIL.

C. 2 out of 4 TM/LP channel pretrips at RPS.

D. A misaligned CEA 7.5 inches from its group.

57. The reactor is at steady state conditions and turbine load has been adjusted to maintain Tc on program.

Given the following:

T cold is 538°F T hot is 556°F What is reactor power?

A. 18%

6. 34.5%

c. 37.5%

D. 40.5%

2004RO.TST Version: 0 Page: 19

Which statement satisfies the requirements for minimum operable position indication channels for a CEA?

59.

A. CEA voltage divider reed switch position indicator channel capable of determining and the absolute CEA position within S inches CEA pulse counting position indicator channel.

the absolute CEA position within k7.5 inches CEA "Full Out" reed switch position indicator channel only if the CEA is fully withdrawn as verified by actuation of the applicable position indicator.

B. CEA voltage divider reed switch position indicator channel capable of determining or C. CEA voltage divider reed switch position indicator channel and CEA pulse counting position indicator channel in agreement within 4.5 inches.

D. CEA voltage divider reed switch position indicator channel capable of determining the absolute CEA position within &I

.75 inches of absolute position CEA "Full Out" reed switch position indicator channel only if the CEA is fully withdrawn as verified by actuation of the applicable position indicator.

Or Which condition would cause audible WRNl count rate to rise?

A. Pulling CEAs to criticality when performing the first reactor startup following a

6. Reinserting a once-burned fuel assembly in a new core location C. During RCS drain down to reduced inventory for RCP seal replacement D. Withdrawing CEA # I from a fuel assembly while swapping CEAs refueling outage 6C. How are the sample locations indicated on the Hydrogen Analyzer recorders on I

(2)ClO selected?

A. Manually at the recorder

3. Automatically or manually by the plant computer

c. Automatically or manually from sample panels in the Aux. Building D. Automatically at the recorder 2004RO.TST Version: 0 Page: 20

61. Refueling operations are in progress and Containment Purge is in operation. While taking logs in the Cable Spreading Room, the CRO notices that channel ZF of CRS is bypassed.

How does this affect Containment Purge?

A. Containment Purge will be automatically secured if any other channel of CRS

6. In the event of a valid CRS signal, one Containment Purge CV will remain C. Containment Purge must be secured (or fuel movement suspended), per Technical D. The remaining channels of CRS must be verified operable to allow Containment actuates.

open.

Specification requirements.

Purge to remain in operation

62. High Spent Fuel Pool temperature is corrected by what action?

A. Adjusting spent fuel pool temperature controller setpoint.

B. Throttling 1 I N 6 SRW heat exchanger Saltwater outlet valves open.

c. Adjusting SFP CLR OUT THROTTLE valve to obtain a discharge pressure of D. Throttling open SFP CLR DISCH HDR stop valve.

greater than 120 psig.

2004RO.TST Version: 0

63.

Page: 21 Performing which evolution poses the highest radiological risk to the operator?

A. Discharging the contents of 12 RCW Monitor Tank

6. Filling 21 RCW Degassifier Vacuum Pump reference leg C. Filling I 1 RCW Ion Exchanger with resin D. Recirculating 11 RCWMT through a MWS prefilter

64. Control Room Vent RMS, 0-Rl-5350 is in alarm.

How is the Control Room HVAC system affected?

A. Outside air dampers open to purge the Control Room, and the air conditioning unit B. Control Room ventilation is in recirculation with Post-LOCI filter fans in C. The Control Room HVAC shifts to winter mode of operation with Post-LOCI filter D. Control Room air handling unit is secured. Only outside air dampers open.

is shutdown operation and the kitchen exhaust fan secured.

fans in operation.

65 What condition will start the diesel fire pump?

i i

I A. Fire main header pressure less than 105 PSIG 1

1 B. A smoke detector or temperature detector actuation I C. Both electric fire pump feeder breakers being open 1 D. Preaction solenoid valve or sprinkler alarm check valve actuation I

i Which category of deficient equipment status should be annotated on the Shift Turnover Information Sheet to communicate the status of 21 Condensate Pump which has a broken lube oil pump?

A. (00s)

Out Of Service B. (VF) Inoperable But Functional C. (D) Degraded D. (0) Inoperable 6. What is the condenser differential temperature (condenser delta T) limit, as stated in the facility license?

1 A. The calculated flow weighted hourly average of the temperature rise across both B. The calculated flow weighted hourly average of the temperature rise across each C. The calculated average of the 24 flow weighted hourly readings of both units for a D. The calculated average of the 24 flow weighted hourly readings of each unit for a condensers is limited to 12°F condenser is limited to 12°F.

calendar day is limited to 12°F.

2004RO.TST Version: 0 Page: 22

Which method of informing the GS-NPO is required per administrative procedures?

A. Voicemail B. Alpha-page

c. Alpha-page and detailed voicemail D. Talk directly

69.

r Which condition requires that the Spent Fuel Pool Ventilation charcoal filters be placed in service?

A. Spent fuel is being loaded into an ISFSI storage cask.

6. New fuel is being loaded into the Spent Fuel Pool.

C. A dummy fuel assembly is being transferred from the Spent Fuel Pool to the D. Refueling is in progress which does not include a complete core offload.

Refueling Pool for RFM testing.

7C.

2004RO.TST Version: 0 Whereisthe regulatinggroup CEA "All Rods Out" (ARO) position stated?

A. NEOP-13 (23)

6. COLR figure 3.1.6 C. System 55 (CEDS) setpoint manual D. 01-42, CEDM System Operation Page: 23

71. What documents, used by Operations personnel to run the plant, are updated to communicate the core reactivity effect changes due to core age or fuel composition?

A. USFSAR and NFM Operator Surveillance Procedures (NEOP-301/302)

B. TRM and Offsite Dose Calculation Manual

c. COLR and Technical Data Book (NEOPs)

D. Calvert Cliffs Operating Manual and Technical Specification LCOs

72. The Shift Manager has declared an Alert per ERPIP 3.0 The Operational Support Center is not yet staffed.

A plant operator is required to perform a task in the Auxiliary Building where dose rates are unknown.

What is required prior to the operator being sent to perform the task?

73.

2004RO.TST Version: 0 Which operation always requires an approved Discharge Permit?

A. Dumping Condensate to the Circulating Water System after system cleanup B. Discharging S/G sludge lancing water

c. Dewatering the Saltwater side of a Component Cooling Heat exchanger D. Initiating S/G Blowdown to Circulating Water Page: 24

74. Which one of the following defines the term "success path" as it applies to EOP-8?

A. A course of action based on plant conditions used to address a safety function.

B. A series of actions which, if performed correctly, will allow the CRS to make a single

c. A table which directs the operator to a set of actions to assess a safety function.

D. A form which provides criteria for the STA to use in evaluating safety function event diagnosis.

status.

t 1 Saltwater pump tripped due to a motor overload and reactor trip criteria were reached before the system could be recovered.

I The RO manually tripped the reactor from 100% and all systems responded normally.

I Which Control Room panel would have no alarms annunciated?

I j A. 1C18 1 6. IC13 1 C. IC08 1 D. 1C03 2004RO.TST Version: 0 Page: 25

AOP-2A Rev 18/Unit 1 AlTACHMENT (1)

Page 1 of 2 ESTIMATE GROSS LEAK RATE

cord the following information:

3. Initial PZR Level

3. Initial RCS Temp., TAM

a.

inches

b.

"F

. Initial Time

d. Final PZR Level

e. Final RCS Temp., TAM

f. Final Time

g. RCS Pressure

h. Charging Flow

i. Letdown Flow

j. Total CBO Flow letermine Factors:

C.

d.

inches

e.

"F f,

9-psia

h.

gpm

k. Estimate' PZR volume factor based on RCS Pressure Step g.

2200 psia = 18.9 gallonshnch 1500 psia = 21.5 gallonshnch 1000 psia = 23.3 gallonsfinch 500 psia = 25.5 gallonshnch 200 psia = 27.4 gallonshnch

k.

1. Estimate RCS expansion factor based on RCS Temp., TAM Step e.

550°F = 78.8 gallons/* F 500" F = 63.1 gallons/" F 450" F = 55.7 gallons/" F 400" F = 50.4 gallons/" F 350" F = 43.1 gallons/" F 300" F = 38.2 gallons/" F 250" F = 32.1 gallons/" F

1.

Number:

Effective Datenime:

Expiration DateTTime:

03-53, Rev. 0 08-01-2003 / 1200 Page 1 of 4 r

i B

P This Standing Order is intended to provide basic guidance for Operations to ensure consistent response at varying levels of unidentified RCS leakage. This Standing Order is not intended to change any responses or actions dictated by the CCOM, the Tech Specs, or any other Operational guidance.

Definitions:

Unidentified RCS Leakage - Leakage from the RCS that has not been determined to be from a specific source. For example, if total RCS leakage has been determined to be 0.6 GPM, but 0.4 GPM has been detmmined to be from 12 Charging Pump primary packing leakage, then the unidentified RCS leakage, as referred to in this Standing Order, would be 0.2 GPM.

onsiderations:

Historical baseline RCS leakage for both units has typically been in the range of 0.1 GPM to 0.15 GPM following a refueling outage. This value tends to increase over the fuel cycle due to minor degradations of RCS sealing interfaces (e.$,

packing, etc...). Larger leakrates are typically seen very near the end of a fuel cycle (during times of increased CVCS diversion) due to inaccuracies in the diversion integrator, Calculated leakrate values will be greatly impacted by non-steady-state operation.

Consideration of minor changes in RCS leakrates should be given only when the RCS has been in steady-state conditions.

0 The sensitivity of the Containment Gaseous and Particulate detectors is based on a source term with 1%-failed fuel. Therefore, these detectors will be essentially blind to leakage within the range of this Standing Order.

0 The values presented in this Standing Order are to be considered general guidance. Plant conditions may dictate that actions be taken prior to these values being reached.

Any actions taken to attempt to identify sources of unidentified RCS leakage should be documented in the CRO logs (e.g., Quantified charging pump primary leakage per OMA. No primary feakage detected.) This will ensure efficiency in the search, should it go over several shifts.

Small changes in RCS leakage may need to be trended over several shifts before actions to find leakage need to be taken.

Single evolutions that cause the planned loss of RCS inventory (but ate not leakage) should be annotated in the CRO logs. Examples include large quantities of charging pump venting (such as restoring from maintenance), large diversion activities (such as rinsing a CVCS IX), etc,..

I 0

iuidance I

i NFOSSO 03-03 08/01/03 Page 2 of 4 I. RCS Leakage Condition I Definition:

A) Unidentified RCS leakage r0.2 GPM B) Unexplained increase of 0.1 GPM t

Actions: -

Notify the GS-NPO (voicernail).

__ Evaluate Charging Pumps for increased primary packing leakage.

Consider performing the Miscellaneous portions of STP 0-27 (e.g.,

RCDT leakage).

Consider performing the Leak Identification attachment of AQP-2A.

If potential leak sources are addressed, start a Supplemental 3TP-0-27 to veriv the effect.

Jf, RCS leakage Condition 2 Definition:

A) Unidentified RCS leakage >0.4 GPM B) Unexplained increase of 0.3 GPM C) Unexplained Containment Sump Frequency of <8 Hours concurrent with increased RCS Leakage.

NOTE A leakrate of 0.1 GPM into a completely empty containment 49-gallon (44-gallon) sump will cause the afarm to come in every 8.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (7.3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months ).

i

~~

?

WOSSO 03-03 08101103 Page 3 of 4

)

Actions:

Initiate an Issue Report per QL-2-100.

Notiry the GS-NPO and PE-PSE {alpha-page).

Leave detailed Vdcemail for Site Managers per the Notification Matrix.

Perform the Miscellaneous portions of STP 0-27.

Perform the Leak Identification attachment of AOP-2A.

If the increased RCS leakrate is indicated in the Containment:

9 Begin planning a Containment entry while carrying out other actions.

After planning is complete, the decision to make the entry will be made by the GS-NPO.

Request Chemistry obtain a fresh sample of the 12/22 ECCS pump room sump fur Boric Acid and hydrazine content. Chemistry should grab the sample while the containment sump is being drained, EValuat8 SRW and CC system leakrates for changes.

Request Health Physics obtain a sample of the Containment atmosphere for indications of RCS leakage.

If potential leak sources are addressed, start a Supplemental STP-0-27 to verify the effect.

111. RCS Leakage Condition 3 e

Definition:

A. Unidentified RCS leakage >0.5 GPM with all potential corrective actions taken.

B. Unexplained Containment Sump Frequency of c4 hours concurrent with increased RCS leakage.

~~

~

_. -~

WOSSO 03-03 08/01/03 Page 4 of 4 Act ions: -

Alert Site Management with alpha-page and detailed voicemail per Notification Matrix. Ensure you talk directly to the GS-NPO and PE-PSE.

c _ Begin planning a controlled unit shutdown. Activate the Forced Outage Protocol Checklist per OAP 01 -03.

NOTE:

A leakrate of 0.2 GPM into a completely empty containment 49-gallon (44-gallon) sump will cause the alarm to come In every 4.1 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months (3.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ).

If the increased RCS leakage is indicated in the containment:

B Implement the Rapid Containmont entry procedure. Consideiatiur I for personnel safety must be applied. If the RCS leakrate is degrading a containment entry may no be advisable.

Review RCS Leakage Condition 2 checklist for appropriate actions.

Approved by:

GS-NPO Original signed by J. K. Mills Printed Name and Signature Canceled by:

GS-NPO Date I

Printed Name and Signature

ES-401 Site-Specific SRO Written Examination Form ES-401-8 Cover Sheet U.S. Nuclear Regulatory Commission Site-Specific SRO Written Examination Start Time:

I Finish Time:

II 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 RO 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 portion.

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

Applicant's Signature II Results RO I SRO-Only I Total Examination Values I-Points Applicant's Scores 1 -

Points Applicant's Grade 1 -

Percen

/--

I

\\<

NUREG-1021, Draft Revision 9 32 of 34

Name:

1. U-2 is at 100% power when the "ACTUATION SYSTEM CIS TRIPPED" alarm is received on 2C08. CIS has been determined to be invalid by the crew, but all attempts to reset CIS "A" from the control room have failed. The Reactor Operator tripped the reactor, as directed by the CRS.

I When are the RCPs tripped?

2.

A. Immediately after the CRS completes the mid-EOP-0 brief B. When the RO has reported that controlled bleedoff temperatures exceed 200"F, or C. After the RO has reported the reactivity safety function is complete D. After an attempt has been made to reset CIS from the cable spreading room bearing temperatures exceed 195°F Per Technical Specifications, under what conditions can a spent fuel pool cooling loop replace a Shutdown Cooling loop?

A. There is less than 23 feet of water above the fuel in the reactor vessel, the spent fuel pool cooling loop is aligned to provide flow to the core, and the heat generation rate is less than the heat removal capacity of the spent fuel pool cooling loop.

B. There is greater than 23 feet of water above the fuel in the reactor vessel, the heat generation rate is less than the heat removal capacity of the spent fuel pool cooling loop, and no operations are permitted that would cause a reduction of the Reactor Coolant System boron concentration.

c. There is greater than 23 feet of water above the fuel in the reactor vessel, the heat generation rate is less than the heat removal capacity of the spent fuel pool cooling loop, and both spent fuel pool cooling loops are available.

fuel pool cooling loop is aligned to provide flow to the core, and no draining operations to further reduce the RCS water volume are permitted.

D. There is less than 23 feet of water above the fuel in the reactor vessel, the spent 04SRO.TST Version: 0 Page: 1

3. Unit-I was at 100% power when the following alarms annunciated:

PZR CH 100 PRESS 04SRO.TST Version: 0 ACTUATION SYS SENSOR CH ZF TRIP 1 -PIC-I OOX indicates 221 0 PSlA 1-PIC-IOOY indicates 1400 PSIA 1 -LI-I ?OX indicates 208 inches 1 -LI-I 1 OY indicates 360 inches What additional alarm would be expected with these indications and what action will be taken?

A. PZR PRESS BLOCK A PERMITTED, block SIAS

8. ACTUATION SYS SlAS TRIP, verify SIAS actuation

c. CNTMT NORMAL SUMP LVL HI, implement AOP-2A D. PORVISAFETY VLV ACOUSTIC MON, implement EOP-0 Page: 2

4. Unit-2 has experienced a loss of 2Y09. Reactor trip criteria was reached and the RO depressed the reactor trip buttons on 2C05.

Approximately 10 seconds later the RO reported WRNl power indications on 2C05 are reading approximately 100% power and startup rate is 0.

What is the cause of these indications and what actions are required?

A. The reactor failed to trip when the trip buttons were depressed. The electrical buses feeding the CEDM MG sets are deenergized.

6. Normal reactor response immediately following a reactor trip from 100% power.

After 30 seconds, verify a prompt drop in reactor power and a negative SUR exists.

C. Loss of power to NI instrumentation due to loss of 2Y09. Verify all CEAs are inserted and delta-T power is lowering.

D. An overpower condition occurred due to feedwater heater high level dump valves opening on the loss of 2Y09. Verify all CEAs are inserted and the turbine has tripped.

5. A feed system malfunction has occurred and the following indications exist:

Reactor power is 100%

SGFPs speed is 4550-4650 RPM SGFP suction pressure is 380 PSIG I 1 S/G level is -2 12 S/G level is -25 TRIO1 I/?

11 1 recorder blue pen (feed flow) is slightly greater than the red pen (steam TRl021/1121 recorder red pen (steam flow) is greater than blue pen (feed flow) flow)

What is the proper action for the CRS to direct the panel operators to perform?

A. Trip the reactor and implement EOP-0.

B. Start the standby Condensate Booster pump.

c. Place 12 S/G FRV controller in manual and open 12 FRV.

D. Bias SGFP speed to maintain FRV D/Ps greater than 75 PSIG.

. Given an electrical system malfunction, what Control Room panel indications will reflect the status of the 125 VDC buses to allow selection of the correct section of the AOP to implement?

A. Reactor Protective System cabinets at 1 C15 B. Steam Generator Feed Pump emergency lube oil pump lights on IC03 C. AFW pump controls on IC04 D. Containment pressure transmitter isolation valves on 1 C10

6.

04SRO.TST Version: 0 1 B Diesel Generator was out of service for maintnenance when a loss of offsite power occurred. 1 A Diesel Generator tripped shortly after starting. Per the appropriate procedure, in what order are the following actions taken?

I.

Minimize 250 VDC battery discharge and restoration of forced circulation if desired.

2. Establish RCS Heat Removal and protect the condenser from overpressure.

3. Attempt to regain either an onsite or an offsite power source.

4. Evaluate the need for a plant cooldown via either forced or natural circulation.

A. I, 2, 3, 4 B. 3, 2,4, 1 C. 2, 1, 3, 4 D. 2, 3, 1, 4 Page: 3

8. Upon a loss of MCC-114, what Technical Requirements Manual credited boration flow path would be available?

A. RWT to RWT charging pump suction valve (CVC-504) to charging pump suction B. 12 BA pump to BA direct M/U valve (CVC-514) to charging pump suction C. 12 BA pump to BA flow control valve (CVC-21 OY) to VCT M/U valve (CVC-512) to D. 11 or 12 BAST gravity drain valves (CVC-508 or 509) to charging pump suction VCT outlet (CVC-501) to charging pump suction

9.

4SRO.TST Version: 0 U-1 is in Mode 1. The latest leakage reports are:

-- 0.6 gpm from Pressurizer Safety Valve leakage

-- 1.8 gpm from leakage past check valves from the RCS to the SI system

-- 0.15 gprn from primary to secondary leakage (12 S/G)

-- 0.5 gpm reactor vessel head seal leakage

-- 4.8 gpm total RCS leakage Based upon these known leak rates, which of the following Technical Specification RCS leakage limits are being exceeded?

A. Pressure Boundary leakage and Identified leakage.

B. Primary to Secondary leakage and Unidentified leakage.

C. Identified leakage and Unidentified leakage.

D. Primary to Secondary leakage and Pressure Boundary leakage.

Page: 4 1 C. What assumptions are made for the implementation of AOP-SA in addition to a major fire in the Control Room?

A. Station Blackout, LOCA

6. Loss of Offsite power, no other accidents C. Station Blackout, no other accidents D. Loss of Offsite power, LOCA

1 I.

1 An RCS leak has resulted in implementing the applicable plant procedure.

Using provided references:

Functional Recovery Procedure, EOP-8, has been implemented and the following plant conditions exist:

--4 CEAs indicate fully withdrawn

--SUR is 0

--All charging pumps are inoperable

--RVVT is available and operable

--SIAS has actuated and 2 HPSl pumps are running

--One 500 KV bus is energized

--Both SG levels indicate at -1 00' and constant and AFW flow is available

--Containment pressure is 0.4 psig and lowering Which one of the following groups of Success Paths is implemented to assess and restore safety functions?

A. VA-1, PIC-3, HR-2, CE-2, RLEC-2 B. VA-1, PIC-4, HR-3, CE-2, RLEC-1 C. VA-1, PIC-3, HR-3, CE-3, RLEC-2 D. VA-1, PIC-4, HR-2, CE-2, RLEC-1 04SRO.TST Version: 0 Page: 5

13. 13 Component Cooling Pump is to be run 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for PMT after bearing replacement.

What action is required and why?

A. 12 Component Cooling Heat Exchanger must be placed in service to ensure a B. 13 Component Cooling Pump must be powered from 11 4KV bus to ensure both C. 12 Component Cooling Pump must be placed in PTL to prevent damage to a SDC D. IX BYPASS I

-CVC-520 must be placed in BYPASS to prevent a reactivity event Component Cooling loop remains in operation.

loops remain operable.

Heat Exchanger due to high flow if a SlAS occurs.

due to lowering letdown temperature.

1 q. Using provided references:

14' After investigating an alarm at 1 C33, the CRO returns from the cable spreading room and reports that #23 Battery Charger output voltage is 120 VDC.

How is the 125 VDC system affected and what action is required?

21 B SRW Heat Exchanger is to be removed from service for cleaning today.

How is the Containment Cooling System affected?

A. The manual SRW inlet isolation valve on 21 or 22 Containment Cooler is shut to B. One train of Containment Cooling is declared inoperable because 2A Diesel C. The Containment Cooling System is degraded but remains operable and functional.

D. 21 and 22 Containment Coolers must be declared inoperable because 21 maintain 2A Diesel Generator operable.

Generator is inoperable.

Component Cooling Heat exchanger must be taken out of service.

A. One DC channel is inoperable. Restore to operable status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

B. #23 battery charger remains operable as long as it's offsite power source remains C. #23 battery charger is inoperable, verify an operable battery charger is suppling 11 D. The battery charger is inoperable, 11 125 VDC bus is inoperable and 1YO1 must be operable, perform a breaker lineup per STP-0-90.

125 VDC bus.

placed on the Inverter Backup Bus.

04SRO.TST Version: 0 Page: 6

1' Which list reperesents plant personnel that must be notified in the event of a Containment entry at power?

A. Rad Con ALARA, Nuclear Security, Mechanical Maintenance Supervisor B. Nuclear Security, Nuclear Training, Operations Work Control C. Mechanical Maintenance Supervisor, Instrument and Controls Maintenance D. Instrument and Controls Maintenance Supervisor, Rad Con ALARA, Nuclear Supervisor, Control Room Supervisor Security Given the following plant conditions:

-- Unit One has tripped due to a loss P-13000-1

-- 11 4KV bus is energized from 1A Diesel Generator

-- Pzr level is 100" and slowly lowering

-- RCS pressure is 1920 PSIA and slowly lowering The RO reports that only 12 Charging Pump is running and Pressure and Inventory is being monitored for positive trends.

What alternate actions must the CRS direct or verify?

A. Verify SIAS actuation when RCS pressure reaches 1725 PSIA.

B. Manually start 1 1 and 13 charging pumps to restore pressurizer level to greater C. Isolate letdown, check that charging pumps automatically start to restore than 1Ol"and locally reset 1 1 pressurizer backup heater breaker.

pressurizer level and reset pressurizer proportional heaters by momentarily placing the handswitches to OFF.

D. Verify charging pumps start automatically to restore Pressurizer level to greater than I O I ", verify 12 and 14 pressurizer backup heaters start to restore RCS pressure.

04SRO.TST Version: 0 Page: 7

Id A core shuffle is in progress and the refueling machine is indexed over a core location with a fuel assembly grappled in the hoist box. What condition would require the Fuel Handling Supervisor to stop core alterations?

A. Count rate increases from 10 CPS to 12 CPS when the fuel bundle is inserted into E. Communications between fuel handling stations is lost.

C. One channel of 3 available nuclear instrumentation channels is declared out of D. The personnel airlock doors are both open.

the core.

service.

19.

A. Approval--GS-NPO, cancellation-- Shift Managers B. Approval--Manager-Nuclear Operations, cancellation--Manager-Nuclear C. Approval -GS-NPO, cancellation--GS-NPO D. Approval--Shift Manager, cancellation--GS-NPO Operations Which selection is the requirement for notification of plant management in the event that a deviation to a Controlling Technical Procedure was approved by the CRS and performed? (Assume no Technical Specification deviation was required)

A. Shift Manager E. Shift Manager, GS-NPO or M-NO

c. Shift Manager, GS-NPO, M-NO and NRC resident D. Shift Manager, GS-NPO, M-NRM

20.

04SRO.TST Version: 0 Who has approval authority for Nuclear Plant Operations Section Standing Orders ant who can cancel them?

Page: 8

21. What systemskomponents are credited for protection of the RCS Pressure Safety Limit?

2

23.

Which evolution requires direct supervision by a Senior Reactor Operator?

Who is responsible for writing an SWP for an Operations evolution when the task is noi covered by an exisiting permit?

A. Bypassing an RAS sensor module B. Discharging a RCWMT C. Performance of any "trip sensitive" PE D. Placing the SFP ion exchanger in service Unit-2 has been stable for the past 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months with RCS pressure 100 PSIA.

RCS temperature is 11 0°F.

Containment closure deviations exist.

Pressurizer level starts rapidly lowering from 160".

What procedure provides the required actions for these conditions?

A. AOP-2A, Excessive RCS Leakage B. EOP-5, Loss of Coolant Accident C. AOP-3B, Abnormal Shutdown Cooling Conditions D. AOP-4A, Loss of Containment Integrity/Closure 04SRO.TST Version: 0 Page: 9

2!( The RCP Trip Strategy in EOP-0 and any of the succeeding EOPs will:

A. Ensure that during a cold leg break LOCA, RCPs remain running as long as possible; during non-LOCA conditions, pressurizer spray is maintained.

B. Ensure during a hot leg small break LOCA, RCPs are secured early enough to prevent a deep uncovery of the core; during non-LOCA conditions pressurizer spray is maintained and minimizes voiding in the RV upper head during cooldown.

C. Ensure that during LOCAs, sufficient flow is maintained to aid heat removal from the reactor vessel head; during non-LOCA events, pressurizer spray heat losses are minimized.

of an uncovery and that the use of aux spray is minimized.

D. Ensure that during all events (LOCA and non-LOCA) the core is kept from too deep 04SRO.TST Version: 0 Page: 10

VI. RESOURCE ASSESSMENT TABLE EOP-8 Rev 26/Unit 1 Page 28 of 54

~~

SAFETY FUNCTION SUCCESS PATH DETERMINATION RESOURCE ACCEPTANCE REACTIVITY CONTROL SUCCESS PATH CONDITIONS CRiTERlA

C-I : GEA Insertion

a. CEAs are able to be inserted, and SUR is negative OR

b. A loss of ALL Vital 4KV Buses may have occurred tC-2: Boratim Using e. Charging pump cvcs is available for boron addition b; Boric acid source is available:

0 BAST RVVT

c. Charging path is available via norinal flow path or SIS flow path RC-3: Boration Using a. HPSl pump is available SIS for boron addition

b. RWT is available as boric acid source

c. Aflow path is available

1. NO more than ONE.

-CEA NOT fully inserted, WRNl power is lowering, OR

2. WRNl power below 104%

and SUR is negative or zera I.

Boration rate greater than or equal to 40 GPM, WRNi power is lowering, and SUR is negative OR 0

2. WRNl power below 1.04%

and SUR is negative or zerc I.

Boration rate greater than or equal to 40 GPM, WRNi power is lowering, and SUR is negative OR

2. WRNl.power below 104%

and SUR is negative or zerl

EOP-8 Rev 26Nnit I Page 29 of 54 It. RESOURCE ASSESSMENT TABLE SAFETY FUNCTION SUCCESS PATH DETERMINATION RESOURCE ACCEPTANCE

a. At least ONE 500KV Bus IITAL UJXILIARIES SUCCESS PATH CONDITIONS CRITERIA i-I:

500KV Offsite I.

At least ONE 4KV vital Power is available bus is energized

2. lly 12,21 and 22 125V DC Buses, ALL greater than 105 volts

3. At least THREE 120V AC Vial Buses are energized:

11 12 13 14 24-2: Diesel

a. IA, PB OROC is available Generators Diesel Generator (continue)

4. EITHER iYO9 or 1Y?O is energized I.

At least ONE 4KV vital bus is energized 2.11y12, 21 and 22 125V DC Buses, ALL greater than 105 volts

3. At least THREE.

120V AC Vital Buses

' are energized:

I1 12 013 14 is energized

4. EITHER lYO9 or 1Y10

Vl. RESOURCE ASSESSMENT TABLE EOP-8 Rev26Nnit 1 Page 30 of 54 ACCEPTANCE CRITERIA SAFETY FUNCTION SUCCESS PATH DETERAIIINATION VITAL RUXlLlARlES (continued)

RESOURCE SUCCESS PATH CONDITIONS VA-3: SMECO

a. SMECO Power Supply System is available I.

At least ONE 4KV vital bus is energized 2.11,12,21 and 22 125V DC Buses. ALL greater than 105 volts

3. At least THREE 120V AC Vital Buses are energized:

I 1 12,

13 14

4. EITHER 1YO9 or IYIO is energized

EOP-8 Rev 26Rlnit I Page 31 of 54 tCS PRESSURE MD INVENTORY

oNTRoL SUCCESS PATH CONDITIONS CRITERIA RESOURCE ACCEPTANCE rl. RESOURCE ASSESSMENT TABLE

'IC-I : cvcs

'IC-2: PORVs or Pressurizer Vent

a. Charging pump is available

b. Charging path is available via normal flow path or SIS flow path

c. A charging source is available:

VCT BAST RWT

d. A method of pressurizer pressure control is available:

0 Pressurizer heaters Main Spray 0 AuxSpray Controlled Steaming

e. SlAS has NOT actuated OR has been reset

a. PORV or Pressurizer Vent required to reduce pressure

b. PORV or Pressufuer Vent available to control pressure

c. Charging and letdown and/or SIS is-available to control pressurizer level

d. Once-Through-Cooling is NOT in progress

1. Pressurizer pressure less than the upper limits of AB. (I)

2. Pressurizer level greater than 30 inches

3. RCS subcooling is between 25°F and 140"F:based on CET temperatures

4. RVLMS indicates level above the top of the hot leg I. Pressurizer pressure less than 2400 PSlA

2. Pressurizer pressure less than the upper limits of Att. (1)

3. RCS subcpoling is betweei 25°F and 140°F based on CET temperatures

4. Pressurizer level greater

5. RVLMS indicates level than 30 inches (90) above the top of the hot le!

?

(continue)

EOP-8 Rev 26/Unit I Page 32 of 54 rl. RESOURCE ASSESSMENT TABLE

~

- ~

SAFETY FUNCTION SUCCESS PATH DETERMINATION kCS PRESSURE AND INVENTORY FONTROL continued)

RESOURCE ACCEPTANCE iUCCESS PATH CONDITIONS CRITERIA Att. (1)

'IC-3: Loss Of

a. A loss of ALL 4KV Vital

b. SIAS has'NOT actuated I.

Pressurizer pressure less Vital AC Buses has occurred than the upper limits of OR has been reset

2. RCS subcooling greater than 25°F based on CET temperatures (1)

OR CET temperatures less than 50°F superheated (I)

3. RVLMS indicates the core is covered

'IC4 SLS

a. SIAS has actuated

1. IF RAS has NOT occurred, AND pressurizer pressure i:

greater than 1270 PSIA, THEN at least ONE Charging Pump operating OR SIS is able to be used to supply RCS makeup

2. HPSi and LPSl Pumps are injecting water into the RCS PER Atts. (I

0) and (I I)

(2) (:

is covered

3. RVLMS indicates the core (1) Refer to Attachment (12) to read CETs white vital AC buses are de-energized.

(2) Limits in Attachments (IO) and (1 I) are not required to be met if SIS throttle criieria (3) LPSl Pumps are NOT required post-RAS.

are met.

Rev 26Wnit I Page 33 of 54 VI: ESOURCE ASSESSMENT TABLE SAFEN FUNCTION SUCCESS PATH'DETERM1NATION RESOURCE ACCEPTANCE CORE AND RCS HEAT REMOVAL JCCESS PATH CONDlTlONS CRITERIA

-I

SIG Heat Sink a. At least ONE S/G level Wfi NO SIS Operation

b. Feedwater is available:

greater than (-)350 inches 0 Main Feedwater O A F W 0 Booster Pump Injection OR has been reset

c. SlAS has NOT actuated

d. SIS operation NOT required I.

At least ONE S/G has level between (-)24 inches and (+)30 inches OR S/G level is being restored by feedwater flow THENTHoT minusTcW is less than 1 q"F

3. IF RCPs are NOT' operating THNTHoT minusTcW is less than 50°F

4. RCS subcooiing greater than 25°F based on CET temperatures (1)

5. RVLMS indicates level above the top of the hot leg

2. IF RCPs are operating, I

-~

~

1) Refer to Attachment (12) to read CETs while vital AC buses are de-energized.

(continue)

EOP-8 Rev 26/Unit 1 Page 34 of 54

1. RESOURCE ASSESSMENT TABLE

~

3.-

SAFEW FUNCTlON SUCCESS PATH DETERMINATION ORE AND RCS.

EAT REMOVAL

ontinued)

RESOURCE ACCEPTANCE UCCSS PATH CONDITIONS CRITERIA R-2: SG Heat Sink Wrth SIS

. greater than (-)350 inches level between Operation

a. At least ONE S/G level

b. Feedwater is available:

I.

At least ONE S/G has 0 inches and (+)38 inches OR SIG level is being restored by feedwater flow Main Feedwater AFW Booster Pump injection

c. SIAS has actuated or SlS operation required

2. CET temperatures less than 50°F superheated (I)

3. IF RAS has NOT occurred; AND pressurizer pressure. i:

greater than 1270 PSIA, THEN at least ONE Charging Pump,operating

4. HPSI and LPSl Pumps are injecting water into the RCI PER Atts. (IO) and (1 I)

(2) (.

I) Refe; to Attachment (12) to read CETs while vital AC buses are de-energized.

2) Limits in Attachments (IO) and (I

1) are not required to be met if SIS throttle criteria

3) LPSI Pumps are NOT required post-RAS.

are met.

(continue)

II. RESOURCE ASSESSMENT TABLE

3. HPSl and LPSl Pumps are injecting water into the RCS PER Atts. (1 0) and (1 I)

(2) (3 EOP-8 Rev 26Nnit 1 Page 35 of 54 continued)

RESOURCE ACCEPTANCE SUCCESS PATH CONDITIONS CRITERfA

+ ~ - 3 :

Shutdown

a. CET temperatures less

b. Radiation levels are low enough to allow valve repositioning Cooling than 300" F System 4R-4: Once-Through-a. HPSl pumps are available Cooling

b. BOTH PORVs are available

c. Flow path is available

d. R W is availabie as a makeup source

. I.

CET temperatures le'ss.

than 300°F and less than 50°F superheated (1)

2. HPSl Pumps are injecting water into the RCS PER Att. (10) (2)

3. Pressurizer pressure less than 270 PSlA(245).

4. RVLMS indicates the core is covered I.

CET temperatures less than 50°F superheated (I)

2. IF RAS has NOT occurred, AND HPSl throttle criteria are NOT met, THEN ALL available Charging Pumps operating

4. Pressurizer pressure less than 1270 PSIA8 OR is lowering (1) Refer to Attachment (12) to read CETs while vital AC buses are de-energized.

(2) Limits in Attachments (I

0) and (1 I) are not required to be met if SIS throttle criteria (3) LPSl Pumps are NOT required post-MS.

are met.

/I. RESOURCE ASSESSMENT TABLE EOP-8 Rev 26/Unit 1 Page 36 of 54 SAFETY FUNCTION SUCCESS PATH DETERMINATION.

RESOURCE ACCEPTANCE

ONTAINMENT ZWRONMENT SUCCESS PATH CONDITIONS CRITERIA

E-I: NO CIS

a. Containment pressure I.

Containment pressure less than 2.8 PSlG less than 2.8 PSlG

b. CIS has NOT actuated OR has been reset less than 220°F (11

c. Containment radiation alarms are clear with NO unexplained rise (2)

.2 Containment temperature

3. Containment radiation alarms are clear with NO unexplained rise (2) 11 NOT available if IYIO is de-energized.

2) NOT applicable if 00s due to loss of power.

(continue)

EOP-8 Rev 26RInit I Page 37 of 54 VI. RESOURCE ASSESSMENT TABLE SAFETY FUNCTiON SUCCESS PATH DETERMINATION CONTAINMENT ENVIRONMENT (continued)

RESOURCE ACCEPTANCE SUCCESS PATH CONDITIONS CRITERIA CE-2: Containment

a. Containment pressure I.

Containment pressure Isolation less than 4.25 PSlG

less than 4.25 PSlG

b. CSAS has NOT actuated

2. ALL available Containment Air Coolers are operating with maximum SRW flow OR has been reset

. 3. ALL containment penetrations required to be shut have a n isolation valve shut

4. Hydrogen concentration less than 05% (1)

OR ALL available hydrogen recombiners are energized with Hydrogen concentratio less than 4.0% (I)

OR Hydrogen purge operation per Tech Support recommendation (I) i) Hydrogen concentration acceptance criteria may be omitted until Chemistry has been able to place hydrogen monitors in service.

(continue)

I

EOP-8 Rev 26fUnit 1 Page 38 of 54 VI. RESOURCE ASSESSMENT TABLE

'SAFETY FUNCTION SUCCESS PATH DETERMINATION

ONTAINMENT IWIRONMENT continued)

RESOURCE ACCEPTANCE iUCCESS PATH CONDITIONS

. CRITERIA

>E-3: Containment

a. Containment pressure I.

Containment pressure spray greater than 4.25 PSlG less than 50 PSlG

2. ALL available Containment Air Coolers are operating with maximum SRW flow.

3. Containment spray flow is greater than 1350 GPM per pump, if operating

4. ALL containment penetrgtions required to be shut have an isolation valve shut less than 0.5% (I)

5. Hydrogen concentration OR ALL available hydrogen recombiners are energized with Hydrogen concentratioi less thari 4.0% (I)

OR Hydrogen purge operation

.per Tech Support recommendation (1)

(1) Hydrogen concentration acceptance criteria may be omitted until Chemistry has been able to place hydrogen monitors in service.

EOP-8 Rev 26RJnit 1 Page 39 of 54

/I. RESOURCE ASSESSMENT TABLE SAFETY FUNCTION SUCCESS PATH DETERMINATION 9DIATION WELS EXTERNAL 3 CONTAINMENT UCCESS PATH CONDITIONS CRlTERlA RESOURCE

. ACCEPTANCE LEG1 :Normal Levels a. Normal Radiation levels exist I.

Noble Gas Monitor outside of containment

b. Containment pressure less than 2.8 PSlG

c. A loss of ALL Vital 4KV Buses may have occurred (I

-RIC-5415) alarm clear with NO unexplained rise

2. Condenser Off-Gas RMS (I

-Ri-l752) alarm clear with NO unexplained iise (1)

(I-Rl-4014) alarm clear with NO unexplained rise (I)

3. S/G BID RMS

4. Main Vent Gaseous RMS (I-R1-5415) alarm clear with NO unexplained rise (11 I) NOT applicable if 00s due to lossaf power.

(continue)

I

I EOP-8 Rev 26/Unit I Page4Oof54 Vi. RESOURCE ASSESSMENT TABLE

. SAFETY'FUNCTION SUCCESS PATH DETERMINATION WlATION WELS EXTERNAL

> CONTAINMENT ontinued)

JCCESS PATH CONDITIONS RESOURCE

. ACCEPTANCE CRITERIA LEC-2:Containment a. Radiation detected

.. isolated outside containment OR Containment pressure greater than, 2.8 PSlG I.

ALL of the following alarms are clear with NO unexplained rise:

Noble Gas Monitor 0 Condenser Off-Gas RMS SIG B/D RMS Main Vent Gaseous RMS (1-RIC-5415)

(I

-RI-I,752)

. (I-R1-4014)

(I -R1-5415)

OR

2. ALL containment.

penetrations required to be shut have an isolation valve shut IF a tube rupture is identified in a S/G, ALL release paths from the affected S/G to the environment are isolated Affected SIG pressure less than 920 PSlA

EOP-8 Rev 26/Uhii I Page4Oof54 fl. RESOURCE ASSESSMENT TABLE

. SAFETY 'FUNCTION SUCCESS PATH DETERMINAT1ON

%.ADlATION

,VELS EXTERNAL kJCCESS PATH CONDITIONS CRITERIA ro CONTAINMENT

'continued)

RESOURCE

. ACCEPTANCE

,EC-2:Containment a. Radiation detected I.

ALL of the following

. tsolated outside containment alarms are clear with NO unexplained n'se:

OR greater than, 2.8 PSlG m Noble Gas Monitor Condenser Off-Gas RMS 0 SIG 510 RMS Main Vent Gaseous RMS Containment pressure (I-RIC-5415)

(I -R1-1752)

. (I-RI-4014)

(1 -RIM541 5)

OR

2. ALL containment penetrations required to be shut have an isolation valve shut 1F a tube rupture is identified in a S/G, ALL release aths from environment are isolated the affected 8 /G to the Affected SfG pressure less than 920 PSlA

DC 'Sources-Operati ng

.. -308.4 CONDITION REQUIRED ACTION

.'3.8 ELECTRICAL POWER SYSTEMS COMPLETION TIME

'.3.8.'4' DC Sources-Operat3 ng A.

One DC channel inoperable due t o an inoperable battery and the reserve battery avai 1 able.

I B.

One DC channel inoperable for reasons other than Condition A.

C.

Required Action and associ ated Compl eti on lime not met.

LCO 3.8.4 Four channels o f DC electrical sources shall be OPERABLE.

A. 1 Rep1 ace i noperabl e 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months battery w i t h reserve battery.

B. l Restore DC channel t o 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months OPERABLE status.

C.l Be i n MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months AND c.2 -

Be i n MODE 5.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months APPLICABILITY:

MODES 1, 2, 3, md 4.

CALVERT CLIFFS - UNIT 1 CALVERT CLIFFS - UNIT 2 3.8.4-1 Amendment No. 227 Amendment NO. 201

'DC-Sources-Operating.

. 3.8.4 SURVEILLANCE REQUJ REMENTS

. SURVEILLANCE SR 3.8.4.1 Verify battery terminal voltage is 2 125 V on float charge.

SR 3.8.4.2 Verify no visible corrosion a t battery termi nal s and connectors.

Verify battery connection resistance is w i t h i n 1 imi t s.

SR 3.8.4.3 Verify battery cells, cell plates, and racks show no visual indication of physical damage or abnormal deteri orati on that degrades performance SR 3.8.4.4 Remove visible terminal corrosion and verify battery cell tu cell and terminal connections are coated with anti -corrosion material.

SR 3.8.4.5 Verify battery connection resistance is w i t h i n limits.

SR 3.8.4.6 Verify each battery charger supplies 2 400 amps a t 2 125 V for 2 30 minutes.

FREQUENCY 7 days 32 days' 18 months 18 months 18:month.s 24 months CALVERT CLIFFS - UNIT 1 CALVERT CLIFFS - UNIT 2 3.8.4-2 Amendment No. 227 Amendment No. 201

I

.DC.. Sources-Operating

.3..a. 4 SURVEILLANCE REQUIREMENTS (conti nued)

SURVEILLANCE I

FREQUENCY SR 3.8.4.7

NOTE-------------------

The modified performance discharge t e s t i n SR 3.8.4.8 may be.performed i n l i e u o f the service t e s t i n SR 3.8.4.7.

Verify battery capacity i s adequate t o supply, and maintain i n OPERABLE status, the required emergency loads f o r the design duty cycle when subjected t o a battery service test.

CALVERT CLIFFS - UNIT 1 CALVERT CLIFFS - UNIT 2 24 months 3.8.4-3 Amendment No. 227 Amendment No. 201

DC Sources-Operati ng a

3.8.4 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE SR 3.8.4.8 Verify battery capacity i s 2 80% o f the manufacturer's rating when subjected t o a performance discharge t e s t o r a modified performance discharge test.

CALVERT CLIFFS - UNIT 1 CALVERT CLSFFS - UNIT 2 3.8.4-4 FREQUENCY 50 months 1ND 12 months when battery shows degradation o r has reached 85%

o f the expected l i f e with capacity 100% o f manufacturer I s rating.

24 months. when battery has reached 85% o f the expected l i f e with.

capacity 2 100% o f manufacturer's r a t i n g Amendment No. 227 Amendment No. 201 i

c DC Sources-Qperati ng B 3.8.4 B 3.8 ELECTRICAL POWER SYSTEMS

.I B 3.8.4 DC Sources-Operating.

BASES t

BACKGROUND The station DC sources provide the AC emergency power system with control power.

It also provides both motive and

. control power t o selected safety related equipment and preferred AC vital bus power (via inverters).

As required by Reference 1, Appendix IC, Criterion 39, the DC electrical 1 power sources are designed to have sufficient independence, redundancy, and testabil i ty t o perform their safety functions, assuming a single failure. The DC sources also conform t o the recommendations of References 2 and 3.

I i

The 125 VDC electrical power sources consist O f.four independent and redundant safety related Class 1E DC channels.

associated battery charger for each battery, and all the associated control equipment and intercomecti ng cab1 i ng.

Each channel consists of one 125 VDC battery, the During nor&l operation, 'the 125 VDC load is powered'.from the battery chargers w i t h the batteries floating on the system.

In cases where momentary loads are greater than 'the charger capability, or a loss o f normal power t o the battery charger,,the DC load is. automatically powered from the station batteries.

.The DC channels provide the control power for i t s associated Class 1E AC power 'load group, 4.16 kV switchgear, and 480 V load centers.

The DC channels also provide a DC source t o the i'nverters, which i n turn power the AC v'i.ta1 b.uses.

The DC sources.are described i n mo.re detail in' the Bases for LCO 3.8.9 and for LCO 3.8.10.

I Each battery has adequate storage capacity t o carry the

required load continuously for a t least 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and to carry Each 125 VDC battery is separately housed i n a ventilated room apart from its charger and distribution centers.

Each channel i s separated physical 1 y and el ectrical ly from the other channel t o ensure that a single failure i n one channel does not cause a failure i n a redundant channel.

There is load duty cycle as discussed i n Reference 1, Chapter 8.

I '

CALVERT CLIFFS - UNITS 1 & z B 3.864-1

-.Revi sion 2

'DC 'Sources-Operati ng B' 3.8.4 bus, are required to be OPERABLE t o ensure the availability of the required power t o shut.down the.reactor and maintain Loss of any DC channel does n.ot.prevent the minimum safety function from being performed (Reference 1, Chapter 8).

. I i t i n a safe condition after-an A00 0r.a postulated DBA.

1 An OPERABLE DC channel requires the battery and one OPERABLE charger t o be operating and connected t o the associated DC bus(es)

A battery charger i s. considered OPERABLE as. long as -it i s.

receiving power from its normal offsite source and can be connected t o a DG within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months followlng an event.

The DC sources are required t o be OPERABLE i n MODEs 1, 2, 3, 1 and 4 t o ensure safe u n i t operation and t o ensure that:

APPLICABILITY

a.

Acceptable fuel design limits and reactor coolant pressure boundary limits are not exceeded as a result of AOOs or abnormal transients; and integrity and other vital functions are maintained i n the event of a postulated DBA.

b.'

Adequate core cooling is provided, and containment I

The DC sources requirement for MODEs 5 and 6 are addressed

.in the Bases for LCO 3.8,5.

ACTIONS A. 1 '

Required Action A. l requires the inoperable battery t o be replaced by the reserve battery within four hours when one DC channel 'is inoperable due t o an.inoperable battery and the reserve battery is available. The reserve battery is a qualified battery that can rep1 ace and. perform the required function o f any inoperable battery.. The four hour Completion'Time is acceptable based on the capability o f the reserve battery and the time i t takes t o replace the Inoperable battery with the reserve battelsy wki 1 e minimizing '

the time i n this degraded condition..'

B. l Condition B represents one channel with a loss of a b i l i t y t o completely respond t o an event, and a potential.loss o f PA1 VCRT f'l TFFT - IINTTT 1 R 7 R ?.R.d-3 Revisinn 7

DC Sources-Operati ng B 3.8.4 BASES

'I j

. charge required t o overcome the internal losses of a battery and maintain the battery i n a fully charged state. The voltage requirements are based on the nominal design voltage of the battery (2.13 V per cell average) and are consistent

.with Reference 6 and the initial state o f charge conditions assumed i n the battery sizing calculations.

The. 7 day

' Frequency i s conservative when compared wi t h manufacturer recommendations and Reference 6 SR 3.8.4.2 V i sual inspection t o detect corrosion o f the battery 'cel I s and connections, or measurement of the resistance of each cell t o cell and terminal connection, provides an i ndi cation of physi cal damage or abnormal deterioration that coul d potenti a1 l y degrade battery performance.

I I

The limits established for this SR must be no.more than 20%

above the resistance as measured during installation or not above the ceiling value established.by the manufacturer..

The SR Frequency for these inspections, which can detect I

conditions that can cause power 1 osses due to resi stance heating, is 92 days.

This Frequency is considered acceptable based on operating experi ence re1 ated t o

. detecting corrosion trends..

SR 3.8.4.3.

Vi sual i nspecti on of the battery cell s, cef l',pl ates, and

. battery racks provides an indicati.on o f physical damage or abnormal deterioration that could potenti a1 ly.degrade battery performance...

The presence of physi cal damage' or deterioration does not necessarily represent a failure of this SR,.provided an evaluation determi neS that the' physi cal damage or deterioration does not affect the OPERABILITY of the battery ( i t s ability to perform i ts design function).

The 18 month Frequency is based on engineering.judgment.

I Operating experience has shown that these components usually pass the SR when performed a t the I 8 month Frequency.

CALVERT CLIFFS - UNITS 1 & 2 B 3.8.4-5 Revisinn 7

J DC Sources-Operati ng B 3i8.4 BASES I

The SR Frequency is acceptable, given the u n i t conditions required t o perform the test and the.other administrative controls existing t o ensure adequate charger performanc-e during these 24 month intervals.

Frequency i s i ntended to. be consistent w i t h expected fuel cycle 1 engths.

In addition, this SR 3.8.4.7 A battery service test is a special test of battery capabi 15 ty, as found and with the associated battery charger disconnected, t o satisfy the design requirements (battery duty cycle) of the DC source.

The test duration must be 2 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and battery terminal voltage must. be maintained 2 105 volts during the test. The discharge rate and test

length should correspond t o the design accident load (duty) cycle requirements as specified in Reference 1, Chapter 8.

. A dummy load simulating the emergency loads of the design duty cycle may be used i n lieu o f the actual emergency-1 dads.

The SR Frequency of 24 months is consistent with expected fuel cycle lengths.

T h i s SR is modified by a Note.

The Note a1,lows the 1

I I

performance of a modified performance discharge t e s t in 1 i eu of a service test. This substitution i s acceptable because a modified performance discharge test represents a more severe t e s t of battery capacity than SR 3.8.4.7.

SR 3..8.4.8 A battery performance discharge test is a test o f constant current capacity of a battery after having been in service, t o detect any change i n the capaci'ty determined by the acceptance test. The' test i s intended t o determine overall battery degradation due t o.age a.nd usage.

A. battery modified performance discharge test i s a simulated duty cycle consisting o f just two rates; the one minute rate published for the battery 'or the largest current load of the duty cycle, followed by the test rate employed for the performance discharge test, both of which envelope the duty

1 DC Sources-Operati ng B 3.8.4 BASES the previous performance test or when it is 2 10% below the manufacturer's rating. These Frequencies,are consistent

with the recommendations i n Reference 6.

I,

REF.ERENCES

1.

. 2.

3.

4.

. 5.

6.

7.

UFSAR Safety Guide 6, Revision 0, "Assumptions Used for Evaluating the Potenti al Radiological Consequences of a Steam Line Break Accident for Boi I i ng Water Reactors, I1 March 1971 IEEE Standard -308-1978, "IEEE Standard Criteria for.

Class 1E Power Systems for Nuclear Power Generating Stat i on s 'I IEEE Standard -485-1983, "Recommended Practice for Si zing Large Lead Storage Batteries for Generating Stations and Substations (ANSI),I1 June 1983 Regul atory Gui de 1.93, "Avai 1 abi 1 i ty. of E7 ectri c Power Sources, 'I December 1974 IEEE Standard -450-1995, "IEEE Recommended Practi ce for Maintenance, Testing, and Replacement o f Vented Lead-Acid Batteries for Stationary Appl i cations,I1 May 1995 Regulatory Guide 1.32, Revision 2, "Criteria for Safety-Re1 ated El ectri c' Power Systems for Nuclear Power PI ants," February 1977

I 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months 'from discovery o f failure to meet Lirni ti ng Condition for Operation.

Di.stri buti on.Systems-Operatfng

. '3.8.9 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months from discovery of failure t o meet Limiting Condition for 1 Operation i

3.8 ELECTRICAL. POWER SYSTEMS r t '

3i8.9 'Distribution Systems-Operating LCO 3.8.9 The AC, DC, and AC vital bus electrical power distribution subsystems shall be OPERABLE APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION A.

One or more AC el ectri cal power di stri buti on subsystems i noperabl e..

B.

One or more AC vital bus subsystem( s) i noperabl e.

REQUIRED ACTION

\\. 1 Restore AC electrical power distribution subsystems to OPERABLE status 6.1 Restore AC vital bus s u bsy s terns to OPERABLE status.

~~

COMPLETION TIME 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months 2 hours AND CALVERT CLIFFS.. UNIT 1 CALVERT CLIFFS - UNIT 2 3.8.9-1 Amendment No. 227 Amendment No. 201

...-Di stri bution Systems-Operati ng SURVEILLANCE SR 3.8.9.1 Verify correct breaker a1 i gnments and voltage to AC, DC, and AC vital bus

- electrical power distribution subsystems.

-3.8.9 FREQUENCY 7 days ACTIONS (continued)

CONDITION C.

' One DC electrical.

power di stri buti on subsystem inoperable, D.. Required Action and associ ated Completion Time not met, E,

Two or more electrical power distribution subsystems inoperable that result i n a loss of function.

REQUIRED ACTION C.1 Restore DC electrical power di stri buti on subsystem to OPERABLE status.

D. 1 Be i n MODE 3.

AND D.2 Be i n MODE 5.

E.l Enter LCO 3.0.3.

COMPLETION TIME 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months AND 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months from discovery o f fai 1 ure t o meet Limiting Condi ti on for Operation 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 36 hours

(

Immediately CALVERT CLIFFS - UNIT 1 CALVERT CLIFFS - UNIT 2 3.8.9-2 Amendment No. 227 Amendment No. 201

Distribution -:SyStems-.O.perati ng

..B 3.8.9 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.9 Distribution Systems-Operating BASES r

1 BACKGROUND

. The onsite Class 1E AC, DC, and AC vital bus Electrical Power Di stri buti on Systems are divided 5 nto two redundant and independent AC electrical power dS stri bution subsystems and four independent and redundant DC and AC vital bus electrical power distribution subsystems (Reference I, Chapter 8).

The AC primary Electrical Power Distribution System consists of two 4.16 kV ESF buses, each having a t least one separate and independent offsite source of power as well as a dedicated onsite DG source.

Each 4.16 kV ESF bus i s normally connected t o a preferred offsite source.

After a loss of the preferred offsite power source t o a 4.16 kV ESF bus, the onsite emergency DG supplies power t o the 4.16 kV ESF bus.

Control power for the 4.16 kV breakers i s supplied from the Class 1E batteries.

Additional description of this system may be found in the Bases f o r LCOs 3.8.1 and 3.8,4.

The 480 V system include the safety-re1 ated I oad centers, motor control centers, and d.i stri buti on panel s shown i n Table B.3.8,9-1.

The 120 VAC vital buses are divided into four independent and isolated subsystems and are normally supplied from an inverter.

The alternate power supply for the vital buses are non-Class 1 E 120 VAC Buses fed from a Class 1E ESF motor control center through the regulating transformer, and i t s use i s governed by LCO 3.8.7.

Each constant voltage source transformer i s powered from a Class 1E AC bus.

There are four independent 125 VDC electrical power distribution subsystems The l i s t of all required Distribution Systems-Operating i s presented i n Table B 3.8.9-1.

The initial conditions of DBA and transient analyses in Reference 1, Chapters 6 and 14, assume ESF systems are OPERABLE.

The AC, DC, and AC vital bus Electrical Power Distribution Systems are designed t o provide sufficient I

APPLICABLE SAFETY ANALYSES t

CALVERT CLIFFS - UNITS I & 2 B 3.8.9-1 Revision 5

iDi.str3 burion Systems-.Operating

. B..3.81.9 BASES capacity, capabi 1 i ty, redundancy, and re1 i abi 1 i ty to ensure the availability of necessary power to ESF systems so that the fuel, RCS, and containment design limits are not exceeded. These limits are discussed i n more detail i n 'the Bases for Sections 3.2, 3.4, and 3.6.

The OPERABILITY of the AC, DC, and AC vital bus Electrical Power Distribution Systems is consistent with the initial assumptions of the accident analyses and is based upon meeting the design basis of the u n i t.

T h i s includes maintaining power distribution systems OPERABLE during accident conditions i n the event of:

a.

b.

I I

An assumed loss of all offsite power or all onsite AC electrical power; and A worst case single failure.

-The distribution systems satisfy 10 CFR 50.36(c) (2.) (ii),

Criterion 3.

LCO i

The requi red el ectri cal power di stri b u t i on subsystems 1 i sted i n Table B 3.8.9-1 ensure the availability o f AC, DC, and AC vital bus electrical supply for the systems required to s h u t down the reactor and maintain it i n a safe condition after an A00 or a postulated DBA.

The AC, DC, and AC vital bus electrical power di stri bution subsystems are required to be OPERABLE.

I Maintaining the AC, DC, and AC vital bus electrical power distribution subsystems OPERABLE ensures that the redundancy incorporated into the design of ESF is not defeated.

Therefore, a single failure within any system or within the electrical power distribution subsystems wi 11 not prevent safe shutdown o f the reactor.

OPERABLE AC el ectri cal power di stri bution subsystems require the associated buses, load centers, motor control centers, and di stri b u t i on panel s t o be energized to their proper voltages.

OPERABLE DC el ectri cal power distribution subsystems require the associated buses to be energized t o their proper voltage from either the associated battery or charger.

OPERABLE vi tal bus electrical di stri buti on CALVERT CLIFFS - UNITS 1 & 2 B 3.8.9-2 Revision 2

'Di stri:butYon.; Systems--.Operat.i.ng B.3.8.9

BASES 1

I subsystems require the associ ated buses t o be energized t o thei r proper vol tage.

In addition, t i e breakers between redundant safety-related AC, DC, and AC vital bus distribution subsystems, i f they exist, must be open.

This prevents any electrical malfunction i n any distribution subsystem from propagating t o the redundant subsystem, which could cause the failure of a redundant subsystem and a loss of essential safety function(s).

If any t j e breakers are closed, the affected redundant electrical distribution subsystems are considered redundant el ectri cal power di stri buti on subsystems.

I inoperable.

Thi s appl i es t o the onsi te, safety-re1 ated I

APPL I CABILITY The electrical distribution subsystems are requi red to be OPERABLE i n MODES 1, 2, 3, and 4 t o ensure that:

a.

Acceptable fuel design limits and reactor coolant pressure boundary 1 imi ts are not exceeded as a result of AOOs or abnormal transients; and Adequate core cooling is provided, and Containment OPERABILITY and other vital functions are maintained i n the event of a postulated DBA.

b.

Electrical distribution subsystem requirements for MODES 5 and 6 are covered i n the Bases for LCO 3.8.10.

ACTIONS A. l With one or more required AC buses, load centers, motor control centers, or di stri b u t i on panel s, except AC vi tal buses, inoperable and a loss o f function has not yet occurred, the remaining AC electrical power distribution subsystems are capable of supporting the minimum safety functions necessary t o shut down the reactor and maintain i t i n a safe shutdown condition, assuming no single failure.

The overall reliability i s reduced, however, because a single failure i n the remaining power distribution subsystems could result i n the minimum required ESF functions not being supported. Therefore, the required AC buses, load centers, motor control centers, and distribution panels must be restored to OPERABLE status within eight hours.

CALVERT CLIFFS - UNITS 1 & 2 B 3.8.9-3 Revision 2

Di-stri.but i on:..Syst;ems:-Opera't-i ng B '.3..8.9-

' BASES' Condition A worst scenario is one train without AC power (i.e., no offsite power t o the train and the associated DG inoperable).

In this condition, the u n i t is more vulnerable t o a complete loss of AC power.

I t is, therefore, imperative that the u n i t operator's attention be focused on minimizing the potential for loss of power to the remaining train by stabilizing the unit, and on restoring power t o the affected train. The eight hour time limit before requiring I a u n i t shutdown in this condition is acceptable because of:

a, The potential for decreased safety i f the u n i t operator's attention i s diverted from the evaluations and actions necessary to restore power to the affected train, t o the actions associated w i t h taking the u n i t to shutdown within this time limit; and The potential for an event i n conjunction with a single failure o f a redundant component i n the train w i t h AC power.

b.

i The second Completion Time for Required Action A.l establishes a limit on the maximum time allowed for any combination of requi red distribution subsystems to be inoperable during any single contiguous occurrence of failing t o meet the LCO.

for instance, a DC bus is inoperable and subsequently restored OPERABLE, the LCO may already have been not met for up to two hours. This could lead t o a total of ten hours, since initial failure of the LCO, t o restore the AC distribution system.

A t this time, a DC circuit could again become i noperabl e, and AC di stri buti on restored OPERABLE.

This could continue indefinitely.

If Condition A is entered while, 1

The Completion Time allows for an exception t o the normal "time zero" for beginning the allowed outage time "clock.lB This will result i n establishing the "time zero" a t the time the LCO was initially not met, instead of the time Condition A was entered.

The 16 hour1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months Completion Time is an acceptable limitation on this potential t o fail t o meet the LCO indefinitely.

CALVERT CLIFFS - UNITS 1 81 2 B 3.8.9-4 Revision 2

Di stri:buti..on SystemsLOperating.

. '.B 3.8.9 B.l With one or more AC vital buses inoperable and a loss of Function has not yet occurred, the remaining OPERABLE AC vital buses are capable o f supporting the minimum safety functions necessary to s h u t down the u n i t and maintain i t i n the safe shutdown condition.

Overall re1 iabi 1 i t y is reduced, however, since an additional single failure could result i n the minimum required ESF functions not being supported. Therefore, the AC vital bus must be restored t o OPERABLE status within two hours by powering the bus from an I associated inverter via DC or the non-Class 1E 120 VAC bus powered by an ESF motor control center through a regulating I transformer.

Condition B represents one or more AC vital buses without power; potentially both the DC source and the associated AC source are non-functioning.

I n this situation, the u n i t is [

significantly more vulnerable to a complete loss of all noninterruptible power.

the operator's attention focus on stabilizing the unit, minimizing the potential for loss of power to the remaining vital buses, and restoring power to the affected vital bus.

I t is, therefore, imperative that T h i s two hour 1 imi t is more conservative than Completion I

Times allowed for the vast majority of components that are without adequate vital AC power.

Taking exception t o LCO 3.0.2 for components without adequate vital AC power, which would have the Required Action Completion Times shorter than two hours if declared inoperable, is acceptable I because o f :

a.

The potential for decreased safety by requiring a change i n u n i t conditions (i.e., requiring a shutdown) and not allowing stable operations to continue; The potential f o r decreased safety by requiring entry into numerous Appl i cab1 e Conditions and Requi red Actions for components without adequate vital AC power and not providing sufficient time for the operators t o perform the necessary eval uati ons and actions for restoring power to the affected train; and The potential for an event i n conjunction w i t h a single fai 1 ure o f a redundant component.

b.

c.

CALVERT CLIFFS - UNITS 1 & 2 B 3.8.9-5 Revision 2

~

~~

-~

~

Distribution Systems-Operating B. 3.8.9 BASES

~

I The two hour Completion Time takes into account the importance t o safety of restoring the AC vital bus t o OPERABLE status, the redundant capabi 1 i ty afforded by the other OPERABLE vital buses, and the low probability of a DBA occurri ng during thi s peri od.

The second Completion Time for Required Action B.l establishes a limit on the maximum allowed for any combination of required distribution subsystems t o be i noperable during any single contiguous occurrence o f failing t o meet the LCO.

If Condition B i s entered while, for instance, an AC bus i s inoperable and subsequently returned OPERABLE, the LCO may already have been not met for up t o eight hours. This could lead t o a total of ten hours, I since initial failure of the LCO, t o restore the vital bus distribution system.

A t this time, an AC train could again become inoperable, and vi tal bus distribution restored OPERABLE.

This could continue indefinitely.

This Completion Time allows for an exception to the normal "time zero" for beginning the allowed outage time atclock.ll This will result in establishing the "time zero" a t the time the LCO was initially not met, instead of the time Condition B was entered. The 16 hour1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months Completion Time i s an acceptable limitation on this potential t o fail t o meet the LCO i ndef i ni tel y.

c. 1 With one DC bus inoperable, the remaining DC electrical power distribution subsystems are capable o f supporting the minimum safety functions necessary to shut down the reactor and maintain i t i n a safe shutdown condition, assuming no single failure. The overall reliability is reduced, however, because a single failure in the remaining DC electri cal power di stri buti on subsystem could result i n the minimum required ESF functions not being supported.

Therefore, the DC bus must be restored t o OPERABLE status w i t h i n two hours by powering the bus from the associated battery or charger.

1 CALVERT CLIFFS - UNITS 1 & 2 B 3.8.9-6 Revision 2

',." D.i:strJ:bu t ton :Systems -operat i ng B.3....a A BASES Condition C represents one DC bus without adequate DC power; potentially both with the battery significantly degraded and the associated charger nonfunctioning. In this situation, the u n i t I s significantly more vulnerable to a complete-,loss of all DC power.

operator's attention focus on stabilizing the u n i t,

minimizing the potential for loss of power to the remaining trains and restoring power to the affected train.

It is, therefore, imperative that the This two hour limit is more conservative than Completion I

Times allowed for the vast majority of components which would be without power.

Taking exception t o LCO 3.0.2 for components without adequate DC power, which would have Required Action Completion Times shorter than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , is acceptable because of:

a.

The potential for decreased safety by requiring a change i n u n i t conditions (i.e., requiring a shutdown) while allowing stable operations to continue; The potential for decreased safety by requiring entry i nto numerous appl i cab1 e Conditions and Requi red Actions for components without DC power and not providing sufficient time for the operators to perform the necessary evaluations and actions for restoring power to the affected train; and The potential for an event i n conjunction w i t h a single fai '1 ure of a redundant component.

b.

c.

I The two hour Completion Time for DC buses is consistent w i t h Reference 2.

The second Completion Time for Required Action C. l establishes a limit on the maximum time allowed for any combination of required distribution subsystems t o be i noperabl e during any sing1 e contiguous occurrence of failing to meet the LCO.

for instance, an AC bus is inoperable and subsequently returned OPERABLE, the LCO may already have not been met for up t o eight hours, This could lead to a total of ten hours, since initial failure of the LCO, to restore the DC distribution system.

A t this time, an AC train could again become i noperabl e, and DC distribution restored OPERABLE.

This could continue indefinitely.

If Condition C is entered while, I

CALVERT CLIFFS - UNITS 1 81 2 B 3.8.9-7 Revision 2

'.' :D i: s t r i. b.ut:i on.; Sy s t ems-Ope rdt i ng B :3'. 8:.'9 I

. EASES This Completion Time allows for an exception t o the normal "time zero" for beginning the allowed outage time This will result i n establishing the "time zero" a t the time the LCO was initially not met, instead of the time Condition C was entered. The 16 hour1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months Completion Time is an acceptable limitation on this potential t o fail t o meet the LCO indefinitely.

D. l and D.2 If the inoperable distribution subsystem cannot be restored t o OPERABLE status within the required Completion Time, the unit must be brought t o a MODE i n which the LCO does not apply.

To achieve this status, the u n i t must be brought t o a t least MODE 3 within six hours and to MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

The allowed Completion Times are reasonable, based on operating experience, to reach the required u n i t conditions from full power conditions i n an orderly manner and without challenging u n i t systems.

I E. l Condition E corresponds t o a level of degradation in the electrical distribution system that causes a required safety function to be lost. When more than one inoperable el ectri cal power di stri buti on subsystem results in the loss of a required function, the plant is i n a condition outside the accident analysis. Therefore, no additional time is justified for continued operation. Limiting Condition for Operation 3.0.3 must be entered immediately t o commence a control 1 ed shutdown.

I SURVEILLANCE SR 3.8.9.1 REQUIREMENTS This SR verifies t h a t the AC, DC, and AC vital bus El ectri cal Power Distribution Systems are functi oni ng properly, wi t h the correct ci rcui t breaker a1 i gnment. The correct breaker a1 ignment ensures t h e appropriate separation and independence of the electrical divisions is mai ntai ned, and the appropri ate voltage i s avai 1 ab1 e t o each requi red bus.

The verification o f proper voltage availability on the buses ensures that the requi red voltage i s readi ly avai 1 ab1 e for motive as well as control functions f o r critical system I

CALVERT CLIFFS - UNITS 1 81 2 B 3.8.9-8 Revision 2

BASES I

1 oads connected t o these buses.

-The

seven day Frequency takes into account the redundant capability o f the AC, DC, and AC vi tal bus electrical power di stri buti on subsystems, and other indications available i n the Control Room that alert the operator t o subsystem malfunctions.

I REFERENCES

1.

UFSAR

2.

Regul atory Gui de 1.93, "Avai l abi 1 i t y o f El ectri c Power Sources, December 1974 CAI VERT CLIFFS - UNITS 1 & 2 B 3.8.9-9 Revision 2

Distribution Systems-Operating B 3.8.9 BASES 1

Table B 3.8.9-1 (page 1 o f 1)

AC and DC E l ectri cal Power Di stri buti on Systems)

u I

4160 Volt Emergency Bus No. 11 (Unit l), No. 21 (Unit 2) 1 4160 Volt Emergency Bus No. 14 (Unit 1); No. 24 (Unit 2) 480 Volt Emergency Bus No. 11A (Unit l), No. 21A (Unit 2) 480 Volt Emergency Bus No. 1IB (Unit l), No. 21B (Unit 2) 1 1

480 Volt Emergency Bus No. 14A (Unit l), No. 24A (Unit 2) 480 Volt Emergency Bus No. 14B (Unit l), No. 24B (Unit 2) 480 Volt Emergency Bus No. 104R (Unit 1). No. 204R (Unit 2) 480 Volt Emergency Bus No. 114R (Unit 1), No. 214R (Unit 2) 1 1

I 1

120 Volt AC Vital Sus No. 11 (Unit l), No. 21 (Unit 2) 120 Volt AC V i tal Bus No. 12 (Unit 1), No. 22 (Unit 2) 120 Volt AC Vital Bus No. 13 (Unit 1), No. 23 ( U n i t 2) 120 Volt AC Vi tal Bus No. 14 (Unit 1), No. 24 (Unit 2) 125 Volt DC Bus No. 11 (Unit 1 and Unit 2) 125 Volt DC Bus No. 12 (Unit 1 and Unit 2) 125 Volt DC Bus No. 21 (Unit 1 and Unit 2) 1 1

I,

1 1

1 125 Volt DC Bus No. 22 (Unit 1 and Unit 2)

(l)

Each bus of the AC and DC Electrical Power Distribution System i s a subsystem.

CALVERT CLIFFS - UNITS 1'& 2 B 3.8.9-10 Revision 2

Name:

I.

RCP MALFUNCTIONS 002//BANK-I/CRO-63-1-3/ 1.7E/020630311/015AA2.10/3.7/3.7/SD 63 - ES U-2 is at 100% power when the "ACTUATION SYSTEM CIS TRIPPED" alarm is received on 2C08. CIS has been determined to be invalid by the crew, but all attempts to reset CIS "A" from the control room have failed. The Reactor Operator tripped the reactor, as directed by the CRS.

When are the RCPs tripped?

A. Immediately after the CRS completes the mid-EOP-0 brief B. When the RO has reported that controlled bleedoff temperatures exceed 2OO0F, or bearing temperatures exceed 195°F

4. After the RO has reported the reactivity safety'function is complete D. After an attempt has been made to reset CIS from the cable spreading room A and B are incorrect, the reactor is first tripped manually to avoid an automatic trip.

C is correct per the Alarm Response Manual--(3-06 D is incorrect, there is no requirement to attempt resetting form the ESFAS panels, and this action is not directed by a procedure.

Basis: "ACTUATION SYSTEM CIS TRIPPED" Alarm

References:

55.41 : 10 55.43:5 /

ARP 1 C08KA1: 063A7.02KA2: 01 3000GEN8 04SRO.TST Version: 0 Page: 1

2. LOSS OF RHR 001// NEW-1/0524017.0/203.011/025 2.1.12/2.9/4.0/

Per Technical Specifications, under what conditions can a spent fuel pool cooling loop replace a Shutdown Cooling loop?

./A. There is less than 23 feet of water above the fuel in the reactor vessel, the spent fuel pool cooling loop is aligned to provide flow to the core, and the heat generation rate is less than the heat removal capacity of the spent fuel pool cooling loop.

8. There is greater than 23 feet of water above the fuel in the reactor vessel, the heat generation rate is less than the heat removal capacity of the spent fuel pool cooling loop, and no operations are permitted that would cause a reduction of the Reactor Coolant System boron concentration.

C. There is greater than 23 feet of water above the fuel in the reactor vessel, the heat generation rate is less than the heat removal capacity of the spent fuel pool cooling loop, and both spent fuel pool cooling loops are available.

fuel pool cooling loop is aligned to provide flow to the core, and no draining operations to further reduce the RCS water volume are permitted.

D. There is less than 23 feet of water above the fuel in the reactor vessel, the spent A is correct per TS 3.9.5, the first note.

B and C are incorrect. SFP cooling may only be used to replace a SDC loops with RCS at low water levels. At greater than 23, only I SDC loop is required.

D is incorrect. Per the TS, the heat generation rate must be less than the SFP heat removal capacity. The requirement for not being permitted to lower RCS level is for de-energizing all pumps for less than 15 minutes to shift trains.

References:

43.2 04SRO.TST Version: 0 Page: 2

PZR CH 100 PRESS PZR CH Y LVL ACTUATION SYS SENSOR CH ZF TRIP 1-PIC-IOOX indicates 2210 PSlA 1 -PIC-I OOY indicates 1400 PSIA 1 -LI-I 1 OX indicates 208 inches 1 -LI-I 1 OY indicates 360 inches What additional alarm would be expected with these indications and what action will be taken?

A. PZR PRESS BLOCK A PERMITTED, block SlAS B. ACTUATION SYS SIAS TRIP, verify SlAS actuation

4. CNTMT NORMAL SUMP LVL HI, implement AOP-2A D. PORV/SAFETY VLV ACOUSTIC MON, implement EOP-0 04SRO.TST Version: 0 Page: 3

I Unit-2 has experienced a loss of 2Y09. Reactor trip criteria was reached and the RO depressed the reactor trip buttons on 2C05.

Approximately 10 seconds later the RO reported WRNI power indications on 2C05 are reading approximately 100% power and startup rate is 0.

What is the cause of these indications and what actions are required?

4. The reactor failed to trip when the trip buttons were depressed. The electrical buses feeding the CEDM MG sets are deenergized.

B. Normal reactor response immediately following a reactor trip from 100% power.

After 30 seconds, verify a prompt drop in reactor power and a negative SUR exists.

C. Loss of power to NI instrumentation due to loss of 2Y09. Verify all CEAs are inserted and delta-T power is lowering.

D. An overpower condition occurred due to feedwater heater high level dump valves opening on the loss of 2Y09. Verify all CEAs are inserted and the turbine has tripped.

A is correct, SUR would be negative if CEAS inserted and an initial decrease in NI power is expected.

B is incorrect, per EOP-0.

C is incorrect, NI indications are powered from vital buses.

D is incorrect, although a high power condition may have resulted, the normal SUR and NI indications would still be evident if the reactor had tripped.

Reference:

43.5 04SRO.TST Version: 0 Page: 4

5 LOSS OF RN SRO 0031 I NEW-31 202-3G-714.01202.0401 054AA2.0812.913.31 A feed system malfunction has occurred and the following indications exist:

Reactor power is 100%

SGFPs speed is 45504650 RPM SGFP suction pressure is 380 PSIG 1 I S/G level is -2" 12 S/G level is -25" TRIO1 1/1111 recorder blue pen (feed flow) is slightly greater than the red pen (steam TRI 021/1121 recorder red pen (steam flow) is greater than blue pen (feed flow) flow)

What is the proper action for the CRS to direct the panel operators to perform?

A. Trip the reactor and implement EOP-0.

B. Start the standby Condensate Booster pump.

0. Bias SGFP speed to maintain FRV D/Ps greater than 75 PSIG.

A is incorrrect, trip criteria is not being challenged yet (-50')

B is incorrect, suction pressure is sufficient.

C is correct, given that a feed system malfunction exists (no steam leak), with feed flow less than steam flow, a FRV problem is most likely.

D is incorrect, SGFP speed is in the normal range and the pumps will respond to adjustment of the FRVs automatically.

References:

41.10, 43.5

4. Place 12 S/G FRV controller in manual and open 12 FRV.

6. SRO-201-7-1-03 0031 NONE/BANK-2/SRO-201-7-/ 1.0/201.004/055 2.1.20/4.314.2/201 - EMER 1 B Diesel Generator was out of service for maintnenance when a loss of offsite power occurred. 1 A Diesel Generator tripped shortly after starting. Per the appropriate procedure, in what order are the following actions taken?

1. Minimize 250 VDC battery discharge and restoration of forced circulation if desired.

2. Establish RCS Heat Removal and protect the condenser from overpressure.

3. Attempt to regain either an onsite or an offsite power source.

4. Evaluate the need for a plant cooldown via either forced or natural circulation.

A. 1, 2, 3, 4

8. 3, 2,4, 1

c. 2, 1, 3, 4

43. 2, 3, 1,4 D is correct per EOP-7 Basis: Major Actions of EOP-7

References:

EOP-7 Rev. 4 U-I placekeeper

References:

41. IO, 43.5 04SRO.TST Version: 0 Page: 5

7. CRO-54-1-1-25 001// BANK-I/ CRO-54-1-1/10.3/202.101/058AA2.01/ 3.7/4.1/

Given an electrical system malfunction, what Control Room panel indications will reflect the status of the 125 VDC buses to allow selection of the correct section of the AOP to implement?

A. Reactor Protective System cabinets at 1 C15 B. Steam Generator Feed Pump emergency lube oil pump lights on I C03 C. AFW pump controls on IC04

43. Containment pressure transmitter isolation valves on 1 C10 A is incorrect, RPS channels are powered by Vital AC buses and have been used to give improper reports of DC bus losses in the past.

B is incorrect, these are listed indications for losses of turbine building MCCs C is incorrect, AFW pump controls are Vital Instrument bus power supplies, and can be fed from the opposite unit.

D is correct, these indications are located close together, each is powered by a different DC bus and provide the CRS with a quick diagnostic tool for losses of DC bus reports.

References:

43.5 CRO-107-1-3-28 029// MOD-21 CRO-107-1-/2.14.1/204.094/024 2.2.22/3.4/4.1/ BORATION Upon a loss of MCC-114, what Technical Requirements Manual credited boration flow path would be available?

A. RWT to RWT charging pump suction valve (CVC-504) to charging pump suction C. 12 BA pump to BA flow control valve (CVC-21OY) to VCT M/U valve (CVC-512) to D. I 1 or 12 BAST gravity drain valves (CVC-508 or 509) to charging pump suction A is incorrect, 504 is powered from MCC 114 B is correct, all listed components are powered from MCC I04 C is incorrect, this is not a flowpath taken credit for meeting the TRM D is incorrect, both gravity feed valves are powered from MCCI 14 Basis: Boration Flow Path Availability

References:

55.43.2 55.45.13 KAI :

&. 12 BA pump to BA direct M/U valve (CVC-514) to charging pump suction VCT outlet (CVC-501) to charging pump suction 006K2.02-3 KA2: 004000K2.0 1 04SRO.TST Version: 0 Page: 6

U-1 is in Mode 1. The latest leakage reports are:

-- 0.6 gpm from Pressurizer Safety Valve leakage

-- 1.8 gpm from leakage past check valves from the RCS to the SI system

-- 0.15 gpm from primary to secondary leakage (12 S/G)

-- 0.5 gpm reactor vessel head seal leakage

-- 4.8 gpm total RCS leakage Based upon these known leak rates, which of the following Technical Specification RCS leakage limits are being exceeded?

A. Pressure Boundary leakage and Identified leakage.

C. Identified leakage and Unidentified leakage.

D. Primary to Secondary leakage and Pressure Boundary leakage.

A is incorrect, no pressure boundary leakage is evident, identified leakage limit is 10 GPM B is correct, primary to secondary is 216 GPD, limit is 100 GPD, and unidentified leakage is 2.25 GPM, limit is 1 GPM per 3.4.13 C is incorrect, identified limit is 10 GPM D is incorrect, no pressure boundary leakage is evident.

Basis: RCS leak rate limits

References:

LCO TS 3.4.13 43.2, 43.3 KAI:

02005A8.02KA2:

43. Primary to Secondary leakage and Unidentified leakage.

10. LOR-202 0251 AOP-SA/ BANK-11202-9A-02/2.01201.001/ 068 2.2.25/2.513.7/AOP-SA What assumptions are made for the implementation of AOP-9A in addition to a major fire in the Control Room?

A. Station Blackout, LOCA

43. Loss of Offsite power, no other accidents C. Station Blackout, no other accidents

0.

Loss of Offsite power, LOCA A. is incorrect because a loss of offsite is assumed not a SBO and a LOCA is not B. is correct per AOP-SA notes.

C. is incorrect because a loss of offsite is assumed not a SBO.

0.

is incorrect because a LOCA is not considered.

References; AOP-SA, Basis, 43.1, 43.5, 43.2 considered.

4SRO.TST Version: 0 Page: 7

LOCATE SOURCE OF LEAKAGE HAVE BEEN COMPLETED)

A. Evaluate operation of the plant with letdown isolated and align charging pumps to operate as needed to prevent exceeding a pressurizer level of 225 inches.

4 3. Commence a plant shutdown to COLD SHUTDOWN per OP-3, OP-4, and OP-5.

C. Trip the reactor and implement EOP-0 when PZR level deviates from program by

15.

D. Perform a rapid power reduction and trip the reactor when Tave is less than 537°F and implement EOP-0.

A is incorrect, but is close to steps in AOP-71 where letdown is lost.

B is correct per AOP-2A C is incorrect, but similar to steps if the leak is greater than the capacity of one charging pump.

D is incorrect, but reflects actions for a S/G tube leak in excess of 1 charging pump

.Basis: REQUIREMENTS TO TRIP REACTOR

References:

AOP-2AI 43.5 KAI:

000037EK3.07KA2:

04SRO.TST Version: 0 Page: 8

Using provided references:

Functional Recovery Procedure, EOP-8, has been implemented and the following plant conditions exist:

--4 CEAs indicate fully withdrawn

--SUR is 0

--All charging pumps are inoperable

--RWT is available and operable

--SIAS has actuated and 2 HPSl pumps are running

--One 500 KV bus is energized

--Both SG levels indicate at -1 00 and constant and AFW flow is available

--Containment pressure is 0.4 psig and lowering Which one of the following groups of Success Paths is implemented to assess and restore safety functions?

A. VA-1, PIC-3, HR-2, CE-2, RLEC-2 B. VA-1, PIC-4, HR-3, CE-2, RLEC-1 C. VA-1, PIC-3, HR-3, CE-3, RLEC-2

43. VA-1, PIC-4, HR-2, CE-2, RLEC-1 A is incorrect, PIC-3 would be used if no 4 KV bus was available and SlAS had not actuated.

B is incorrect, HR-3 requires SDC initiation to satisfy heat removal.

C is incorrect-see above for PIC-3 and HR-3 D is correct per the resource assessment table.

RC-3 is used for more than one stuck CEA, SUR is not negative and CVCS not available, PIC-4 uses SIS for pressure/inventory control, S/G and HPSl are used for heat removal.

Basis: Success path determination via R.A.T. EOP-8

References:

EOP-8, 43.5 KAI: KA2:

04SRO.TST Version: 0 Page: 9

I 3. COMP CLG SRO 001//NEW-2/ L0I-15-1-0/7.0/015.001/ 008A2.07/2.5/2.8/

13 Component Cooling Pump is to be run 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for PMT after bearing replacement.

What action is required and why?

A. 12 Component Cooling Heat Exchanger must be placed in service to ensure a B. 13 Component Cooling Pump must be powered from 11 4W bus to ensure both

4. 12 Component Cooling Pump must be placed in PTL to prevent damage to a SDC D. IX BYPASS I-CVC-520 must be placed in BYPASS to prevent a reactivity event Component Cooling loop remains in operation.

loops remain operable.

Heat Exchanger due to high flow if a SlAS occurs.

due to lowering letdown temperature.

A is incorrect, either heat exchanger can be inservice.

6 is incorrect, 11 Component Cooling Pump remains operable.

C is correct, per the note for in 01-16.

D is incorrect, procedures for shifting pumps does not require bypassing the ion exchanger. For the short period of time 2 pumps are running, temperature does not change appreciably.

Refernces: 01-16, 41 S, 43.5

14. CONTAINMENT CLG SRO 001// NEW-2/01 1-1-0/2.0. 15.0/032.006/022A2.0412.9/3.21 21 B SRW Heat Exchanger is to be removed from service for cleaning today.

How is the Containment Cooling System affected?

-'A. The manual SRW inlet isolation valve on 21 or 22 Containment Cooler is shut to maintain 2A Diesel Generator operable.

Generator is inoperable.

6. One train of Containment Cooling is declared inoperable because 2A Diesel C. The Containment Cooling System is degraded but remains operable and functional.

D. 21 and 22 Containment Coolers must be declared inoperable because 21 Component Cooling Heat exchanger must be taken out of service.

A is correct, per 01-29 and Technical Specification LCO 3.7.6 basis.

B is incorrect, one train of air coolers is declared inoperable, but it is because one cooler is isolated, not due to the EDG.

C is incorrect per TS.

D is incorrect, only one heat exchanger is out of service, but that makes the train out of service.

References:

01-29, 43.5 04SRO.TST Version: 0 Page: 10

15. CRO-54-1-1-11 001/TS 3.8.41 BANK-O2/CRO-54-1-1/ 10.31020540502/ 063 2.1.321 3.4t3.81 Using provided references:

After investigating an alarm at 1 C33, the CRO returns from the cable spreading room and reports that #23 Battery Charger output voltage is 120 VDC.

How is the 125 VDC system affected and what action is required?

A. One DC channel is inoperable. Restore to operable status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

B. #23 battery charger remains operable as long as it's offsite power source remains

4. #23 battery charger is inoperable, verify an operable battery charger is suppling 11 D. The battery charger is inoperable, 11 125 VDC bus is inoperable and 1YO1 must be A is incorrect, as long as one charger is operable, the bus is operable.

B is incorrect, the chargers must maintain greater than or equal to 125V at 400 amps or greater to maintain operability.

C is correct per TS 3.8.4 basis.

D incorrect, one battery charger is available to maintain the bus operable.

References:

TS LCO 3.8.4 and basis, 41.10 43.2 operable, perform a breaker lineup per STP-0-90.

125 VDC bus.

placed on the Inverter Backup Bus.

16. CONTAINMENT SRO 003//

NEW-1/204-104-I/ 1.0/204.076/ 103 2.1.14/2.5/3.3/

Which list reperesents plant personnel that must be notified in the event of a Containment entry at power?

A. Rad Con ALARA, Nuclear Security, Mechanical Maintenance Supervisor B. Nuclear Security, Nuclear Training, Operations Work Control C. Mechanical Maintenance Supervisor, Instrument and Controls Maintenance

43. instrument and Controls Maintenance Supervisor, Rad Con ALARA, Nuclear A is incorrect, Mechanical Maintenance Supervisor is not required to be notified.

B is incorrect, Nuclear training is not required to be notified.

C is incorrect, Mechanical Maintenance Supervisor is not required to be notified.

D is correct. Per NO-I -1 04, 1&C is notified to perform airlock door testing after entry is complete, Rad Con is notified prior to entry, and Security must be notified and to verify the EAL outer hatch woodruff key is installed.

References:

NO4 -1 04, 43.5 Supervisor, Control Room Supervisor Security 04SRO.TST Version: 0 Page: 11

1 7. PZR LVL CONT SRO 001// NEW-2/ CRO-107-1-/8.3/201.085/ 01 IA2.05/3.3/3.7/

Given the following plant conditions:

-- Unit One has tripped due to a loss P-13000-1

-- 11 4KV bus is energized from 1 A Diesel Generator

-- Pzr level is I O O I and slowly lowering

-- RCS pressure is 1920 PSIA and slowly lowering The RO reports that only 12 Charging Pump is running and Pressure and Inventory is being monitored for positive trends.

What alternate actions must the CRS direct or verify?

A. Verify SIAS actuation when RCS pressure reaches 1725 PSIA.

43. Manually start 11 and I 3 charging pumps to restore pressurizer level to greater C. Isolate letdown, check that charging pumps automatically start to restore than 1Ol"and locally reset 1 1 pressurizer backup heater breaker.

pressurizer level and reset pressurizer proportional heaters by momentarily placing the handswitches to OFF.

D. Verify charging pumps start automatically to restore Pressurizer level to greater than I O I ", verify 12 and 14 pressurizer backup heaters start to restore RCS pressure.

A is incorrect, action should be taken so that SIAS does not actuate.

B is correct, charging pumps must be manually started, and heaters must be reset, and they will not operate below I01

'I in the pressurizer.

C is incorrect, letdown will automatically isolate, 11 and 13 charging pumps will not automatically start, and the proportional heaters will take a long time to restore RCS pressure.

D is incorrect,l 1 and 13 charging pumps will not automatically start with the normal and alternate 4 KV bus feeder bkrs open and 12 and 14 heaters will not have power available.

References:

EOP basis docs. 41 -5, 43.5 04SRO.TST Version: 0 Page: 12

18.

CRO-113-6-4-20 001//

BANK-1/CRO-113-6-/2.0.C/203.011/034 2.4.49/4.0/4.W A core shuffle is in progress and the refueling machine is indexed over a core location with a fuel assembly grappled in the hoist box. What condition would require the Fuel Handling Supervisor to stop core alterations?

A. Count rate increases from 10 CPS to 12 CPS when the fuel bundle is inserted into the core.

43. Communications between fuel handling stations is lost.

C. One channel of 3 available nuclear instrumentation channels is declared out of D. The personnel airlock doors are both open.

A is incorrect, a small increase in countrate may be normal, depending on age of the fuel assembly and proximity of excore NI.

B is correct, per TRM TNC 15.9.2 C is incorrect, TS LCO only requires 2 Nls operable D is incorrect, TS LCO 3.9.3 allows both doors to be open, as long as one is capable of being shut.

References:

TRM, 41.

I O, 43.2, 43.6 service.

19. SRO-204-1-013-002 002//

NEW-1/204-300/2.0/ 204.008/2.1.2/ 3.014.01 Which selection is the requirement for notification of plant management in the event that a deviation to a Controlling Technical Procedure was approved by the CRS and performed? (Assume no Technical Specification deviation was required)

A. Shift Manager

43. Shift Manager, GS-NPO or M-NO C. Shift Manager, GS-NPO, M-NO and NRC resident D. Shift Manager, GS-NPO, M-NRM A is incorrect, additonal notification is required B is correct, per NO-I -200, 5.1.C C is incorrect, NRC resident is not part of plant staff, and is not required to be notified per RM-1-101.

D is incorrect, Manager-Nuclear Regulatory Matters notification is not a requirement per NO-I -200.

References:

NO-I -200, 43.3, 43.5 04SRO.TST Version: 0 Page: 13

20. NlTElSTANDlNG ORDRS 00111 NEW-1///204.037/2.1.15/ 2.313.W I Who has approval authority for Nuclear Plant Operations Section Standing Orders and

' who can cancel them?

I A. Approval--GS-NPO, cancellation-- Shift Managers

6. Approval--Manager-Nuclear Operations, cancellation--Manager-Nuclear Operations

4. Approva I --GS-N P 0, cancel I at i on--GS-N PO D. Approval-Shift Manager, cancellation--GS-NPO C is correct, per the forms in the NPO Section Standing Orders book.

Distractions are possible Operations Management personnel.

References:

41. I O, 43.3, 2 1. CRO-212-1-1-02 00311 NEW-l/TECH.SPECW25,26/ 204.094/ 2.2.22/ 3.4/4.1/

What systemskomponents are credited for protection of the RCS Pressure Safety Limit?

A. All systems listed in the Limiting Conditions for Operations of Technical

6. PORVs, Steam Bypass Control System (ADVs and TBVs), Pressurizer Pressure Specifications.

Control system valves

4. RPS high RCS pressure trip, Pressurizer Safety Valves and main steam safety D. Auxiliary Feedwater system, ESFAS system and RPS actuation of PORVs C is correct per B 2.1.1, Applicable Safety Analyses Distractors are systems of components not listed in TS Basis.

References:

TS Safety Limits, basis, 43.1, 43.2

22. SRO RESPONSIBILITIES 00111 NEW-2/ 048-1-01 1.I 11 048.0071 2.3.31 1.8R.91 Which evolution requires direct supervision by a Senior Reactor Operator?

A. Bypassing an RAS sensor module B. Discharging a RCWMT C. Performance of any "trip sensitive" PE D. Placing the SFP ion exchanger in service A is correct per 01-34, 5.0.8 B is incorrect, a signed permit is required, but not direct supervision of the lineup.

C is incorrect, only requires notification and approval to perform.

D is incorrect, requires PWS supervision, which can be a non-licensed operator per

References:

01-34, 43.4,43.5 NO-I -200 04SRO.TST Version: 0 Page: 14

23 9ADWORK PERMIT 00111 NEW-1111204.0071 2.3.712.013.31 Who is responsible for writing an SWP for an Operations evolution when the task is not covered by an exisiting permit?

A. The person who is in charge of performing the task B. Operations ALARA Coordinator C. Operations Work Control

4. An ALARA Dlanner D is correct per RSP-1-200 Distractors are Operations personnel or shift personnel without this responsibility or authority.

References:

41.I 0, 43.3, 43.4

24. EMER PRO 00111 NEW-2I 052-4-015.01204.09312.4.414.014.31 Unit-2 has been stable for the past 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months with RCS pressure 100 PSIA.

RCS temperature is 11 0°F.

Containment closure deviations exist.

Pressurizer level starts rapidly lowering from 160'.

What procedure provides the required actions for these conditions?

A. AOP-2AI Excessive RCS Leakage B. EOP-5, Loss of Coolant Accident

4. AOP3B, Abnormal Shutdown Cooling Conditions D. AOP-4A1 Loss of Containment Integrity/Closure A is incorrect, this provides guidance only when RCS is at a higher pressure.

B is incorrect, EOP-5 assumes a trip and EOP-0 have been completed C is correct.Section V is specifically for a loss of inventory.

D is incorrect. Deviations are allowed in Mode 5, there is no guidance in AOP-4A for these indications, and AOP3B will address containment closure.

References:

43.2, 43.5 04SRO.TST Version: 0 Page: 15

25. SRO-201-0-3-08 0081 NONEIBANK-2/ SRO-201-04 3.1/201.001/2.4.18/2.7/3.6/ EOP - EMER The RCP Trip Strategy in EOP-0 and any of the succeeding EOPs will:

A. Ensure that during a cold leg break LOCA, RCPs remain running as long as possible; during non-LOCA conditions, pressurizer spray is maintained.

43. Ensure during a hot leg small break LOCA, RCPs are secured early enough to prevent a deep uncovery of the core; during non-LOCA conditions pressurizer spray is maintained and minimizes voiding in the RV upper head during cooldown.

C. Ensure that during LOCAs, sufficient flow is maintained to aid heat removal from the reactor vessel head; during non-LOCA events, pressurizer spray heat losses are minimized.

of an uncovery and that the use of aux spray is minimized.

D. Ensure that during all events (LOCA and non-LOCA) the core is kept from too deep A is incorrect, maintaining spray flow is not a requirement-aux spray is available.

B is correct per EOP-0 basis IV.D.l, rev 14 Distractors are not referenced in the Basis

References:

EOP Rev. 3 Basis Document 41.10, 43.5 04SRO.TST Version: 0 Page: 16

Name:

SRO-201-0-2-05 00511 NEW -21 SRO-201-04 3.01 0205904081 007EA1.06/4.414.51 EOP-0, STU U-2 is operating at 100% when a reactor trip occurs.

The RO observes the following indications on the CEA mimic:

4 CEAs do not have the amber lights energized 2 of the above CEAs have green lights energized What must be performed when performing the Reactivity Control Safety Function?

A. Take the alternate actions to deenergize CEDM MG sets and verify all CEAs are

6. Depress the Reactor Trip pushbuttons on 2C15, verify reactor power is C. Verify reactor power is lowering, check that no CEA deviation alarms are present, inserted.

lowering, and verify a negative startup rate exists.

verify a negative startup rate exists, check that RCS makeup is secured and inform CRS that Reactivity Control is complete.

pump using all available charging pumps.

0.

Commence RCS boration to at least 2300 ppm via gravity feed or a Boric Acid A is incorrect, the reactor has tripped, de-energizing the MG sets is not required B is incorrect, the reactor has tripped and depressing the trip buttons on 2C15 is not required.

C is incorrect, these are normal trip actions, but do not include the actions required for more than one CEA failing to insert.

D is correct per EOP-0 and bases, page1 3, rev 14 Basis: EOP-0 reactivity control question

References:

EOP-0 Rev. 3KA1: KA2:

2004RO.TST Version: 0 Page: 1

Given the following conditions:

RCS pressure:

1600 PSlA PZR level:

360" T h:

532.5" F Tc:

532.2"F Containment Pressure:

2.4 PSlA Containment temp:

115°F S/G pressures:

8801885 PSIA EOP-0 is being implemented What is the most likely cause of these conditions?

A. RCS cold leg break C. S/G tube leak D. Main Steam line break in containment

43. RCS leak at the top of the PZR A is incorrect, a differnece in Tc and T h would be more evident, also, pressurizer level would not be high during EOP-0 with RCS pressure above the shutoff head of the HPSl pumps B is correct--classic indications C is incorrect, pressurizer level again does not support this D is incorrect, Tc and S/G pressures don't support this 2004RO.TST Version: 0 Page: 2

3. CONTAINMENT COOLING 001//NEW-2/LOl-052-3/5.0/201.093/009A1.07/3.7/3.9/

The CRS ordered Unit-I manually tripped due to rapidly loweing PZR level and RCS pressure. EOP-0 is being implemented and the following conditions exist:

RCS pressure:

1900 PSlA Tc:

532.5"F Containment pressure 0.5 PS Containment temperature 98°F A

What is the status of the Containment Air Coolers? (Assume no operator action)

A. 4 Coolers in slow speed with maximum SRW flow

6. 4 Coolers in fast speed with normal SRW flow C. 3 Coolers in slow speed with maximum SRW flow

4. 3 Coolers in fast speed with normal SRW flow A is incorrect, but describes operation during SIAS. SIAS setpoints have not been reached.

B and C are incorrect, but could be options with manual actions by the operator D is correct per OMA, normal system lineup.

4. SRO-201-5-1-06 006// BANK-Ol/SRO-201-5-/ 1.1.3/033480602/ 01 lEK2.02/2.6/2.7/ EOP - EMER Which one of the following describes the credited RCS inventory and core heat removal processes during a large break LOCA?

A. HPSl injection provides makeup and heat is removed via natural circulation flow to the S/Gs.

flow out the break.

the SIGs.

43. HPSl pumps, LPSl pumps and the SITS provide makeup and heat is removed via C. LPSl pumps and the SITS provide makeup and heat is removed via forced flow to D. Charging pumps provide makeup and heat is removed via flow out the break.

A is incorrect, for Large break LOCAs, heat removal is via flow out the break and inventory is established by SITS and LPSls B is correct per EOP-5 basis pages 10-1 1 C is incorrect, S/Gs are not providing heat removal for DBA LOCA D is incorrect, but this is correct for small RCS leaks Basis: Core Heat Removaf

References:

EOP-5 Rev. 3 Basis DocumentKAl :

02007K3.01 KA2:

2004RO.TST Version: 0 Page: 3

5. RCP MALFUNCTIONS 00111 NEW-3/201-2-8/ 1.0/201.024/015AK1.04/ 2.9/3.1/

Given the following conditions:

-1 1A RCP tripped due to a breaker fault

-EOP-0 has been completed, no alternate actions were required How will the RCS and Steam Generators have responded?

A. 11 and 12 loop differential temperatures will be equal and 11 and 12 S/G pressures B. 11 loop will have an inverted differential temperature and 11 S/G pressure will be

4. 12 loop differential temperature will be greater than 11 loop differential temperature D. 12 loop will have a smaller differential temperature than 11 and 12 S/G pressure will A is incorrect, this reflects equal flow conditions.

B is incorrect, 11 loop will still have forward flow with one RCP in operation, and S/G pressures will be equal.

C is correct, validated with simulator response. 12 loop differential temp. will be about 2"F, 1 1 loop differential temperature will be approximately 1 O F. S/G pressures are essentially equal due to operation of the TBVs.

D is incorrect, 12 loop differential temperature will be approximately twice 11 and S/G pressures will be equal.

will be equal.

lower than 12 S/G pressure.

and 11 and 12 S/G pressures will be equal.

be lower than 11 S/G pressure.

2004RO.TST Version: 0 Page: 4

A reactor trip has occurred and the following conditions exist:

-- Pressurizer level is 140 inches and stable

-- One Charging Pump is available

-- Pressurizer pressure is 1900 psia and rising

-- RCS Subcooling is 65°F and steady After performing the immediate actions for PIC, the Reactor Operator reports "Pressure and Inventory Control cannot be met" to the CRS.

What is the reason for this report?

A. Letdown has been isolated.

6. RCS subcooling is not in band.

C. All Charging Pumps are not in operation.

A is incorrect, letdown status is not a basis for meeting Pressure and Inventory Control B is incorrect, Subcooling band is 30 to 140 O F.

C is incorrect, charging pump status is not a basis for Pressure and Inventory D is correct per EOP-0.

ControlBasis: Proper Report from RO to CRS

References:

EOP-0 Rev. 3 and

43. Pressurizer level is not trending toward setpoint.

NO-1-201 KA1: 03PA3.01 KA2: 03PA3.03 Why does the Component Cooling system realign on a SIAS?

A. Minimize dose rates due to contamination of Component Cooling system

43. Provide long term cooling to containment after RAS C. Minimize load on the Saltwater system to ensure containment cooling via Service D. Provide continuous cooling to LPSl pump seals Water 2004RO.TST Version: 0 Page: 5

8. PRESSURIZER PCS MALF 0011 ARM1 BANK-21 CRO-62-1-314.11064.0361 027AK3.0313.714.11 Unit-1 is in Mode 3 with the following conditions:

RCS pressure is 2150 PSlA and lowering Pressurizer Spray Valves, 1 -RC-1 OOE and F are fully open PIC-IOOX is indicating 2400 PSIA, controller output is 100%

PIC-I OOY is indicating 21 50 PSlA output is 0%

What action is required?

A. Stop 1 1 A and 11 B Reactor Coolant Pumps.

B. Energize all Pressurizer Heaters.

D. Place PRESSURIZER SPRAY VALVE CONTROLLER, 1-HIC-100 in manual with a A is incorrect, would stop depressurization, but indications are of a failed instrument channel.

B is incorrect, with spray valves failed open, RCS would still depressurize.

C is correct, per 1 C07 ARM, E-29.

D is incorrect, 100% ouput would keep spray valves open.

4. Place PZR PRESS CH SEL switch, 1-HS-100 in "Y".

100% output.

9.

SRO-201-0-3-29 02911 NEW-11 SRO-201-04 3.11 0220106011 029EK1.0313.613.91 EOP - EMER EOP-8 has been implemented because the Reactivity Safety Function was not met in EOP-0. What indications are used to verify that boration is successfully meeting the acceptance criteria?

4. WRNI power is less than IO4% and SUR is negative or zero B. LRNl power is less than IO4% or SUR is negative

c. A boric acid pump is running and charging header flow is 40 GPM or greater D. SUR is zero and the CHG HDR FLOW LO PRESS LO alam is clear A is corrrect per EOPT8,appendix 1,rev 26.

B is incorrect, both conditions are required C is incorrect, boration at the given rate in addition to lowering power and negative SUR indications D is incorrect, negative SUR and boration rate of at least 40 GPM is specified.

2004RO.TST Version: 0 Page: 6

secondary sides of a ruptured Steam Generator per the applicable EOP?

A. Lowering the RCS pressure allows HPSl flow to restore Pressurizer level.

C. Reducing RCS pressure and temperature aids initiation of natural circulation.

D. Equalizing RCS and S/G secondary side pressures initiates backflow to control

4.

Reducing the differential pressure lowers the RCS leak rate.

affected S/G level.

1 1. STEAM LINE RUPTURE 001//NEW-1/201-3-7/1.0/201.038/040AK1.07/3.4/4.2/

Emergency Operating Procedures provide specific guidance for feeding a dry S/G to restore RCS heat removal.

This guidance is based on (Select the phrase that correctly completes the above statement)

A. minimizing S/G tube voiding, which would inhibit natural circulation

6. preventing a rapid RCS cooldown, avoiding a pressurized thermal shock to the C. preventing uneven cooling of the RCS, which may result in a localized reactivity Reactor Vessel excursion internals

4. minimizing the probability of creating a waterhammer, and damaging S/G A is incorrect, the steps to slowly introduce water into the feedring is not based on voiding.

B is incorrect, under the conditions outlined in the procedure, thermal shock is not an issue.

C is incorrect, reactivity addition is not a concern for this method of RCS heat removal.

D is correct per EOP-3 Basis,step IV.K.2. page 37, rev.20.

2004RO.TST Version: 0 Page: 7

12. LOSS OF FEEDWATER 001//NEW-3/201-4-6/3.0/201.015/054AK1.01/4.1/4.3/

EOP-0 was completed and the following conditions exist:

Which event would cause these indications?

dA. A Feedwater line rupture inside Containment B. An RCS leak inside Containment C. A Main Steam line rupture in the Turbine Building D. A rupture of the S/G Blowdown Tank 2004RO.TST Version: 0 Page: 8

13. STATION BLACKOUT 002/NONE/MOD-1/SRO-201-7J9.0/201.077/055EA1.06/4.1/4.5/201 - EMER EOP-7 (Station Blackout) has been initiated on U-I and the CRS has directed the CRO to restore power to 11 4KV bus using the OC Emergency Diesel Generator.

Which Control Room annunciator condition reflects that the bus has been re-energ ized?

A. "ACTUATION SYS LOSS OF POWER! alarm clears B. "ACTUATION SYS U N RELAY TRIP" alarm clears

4. "SEQUENCER INITIATED" alarm actuates D. "OC DG CONTR BOARD 1C19C" alarm actuates A is incorrect. This will not be in alarm, power to the actuation systems is maintained by station batteries, through inverters.

B is incorrect. A U N signal will still exist on 14 bus, not allowing the alarm to clear.

C is correct per 01-21C.

D is incorrect. Closing the output and bus feeder breakers will not cause this alarm.

Basis: Restore Vital buses.

References:

EOP-7 Rev. 4 step R,01-21C. 41.7

14. LOSS OF OFFSITE 001//

NEW-Z201-0-9/ 1.8/202.071/056AA2.43/3.9/4.1/

Unit-I has experienced a Loss of Offsite Power from 100% power.

11 and 14 4 KV buses have been re-energized by their associated Diesel Generators.

EOP-0 is being implemented.

What action must the CRO take for step B, "ENSURE TURBINE TRIP" that would NOT be expected on a reactor trip with offsite power available?

A. Depressing the Turbine TRIP button.

B. Opening 11 GEN FIELD BKR, 1 -CS-41.

C. Shutting the MSIVs due to not being able to verify Turbine speed dropping.

43. Dispatching an operator to shut the MSR 2nd stage bypass valves.

A is incorrect, this must be performed for all turbine trips.

B is incorrect, DC power is available, the Generator Field Breaker will automatically open.

C is incorrect, Turbine speed indication is available D is correct per EOP-0, alternate action 3.1 2004RO.TST Version: 0 Page: 9

deenergized, maintaining a UV (load shed signal) to 4 KV bus 11 loads.

B. 4 KV bus 11 will be re-energized by manually starting and loading OC Diesel Generator.

<. 1A DG will automatically start and load to energize 4 W bus 11 after the 1 B DG starts and energizes 4KV bus 14.

D. 4 KV bus 11 cannot be re-energized until power is restored to 1YO1 via DC bus 11.

16. LOSS OF DC POWER 001//NEW-1/69-5-4/1.0/094.013/058-2.1.19/3.0/3.0/

A plant transient has ocurred and the following conditions exist:

--All Unit-I annunciator lights are deenergized

--CC CNTMT RETURN, 1 -CC-3833-CV has failed shut How will SPDS indicate the cause of this event?

A. All Safety Function boxes turn red, and a "Loss of AC bus I' alarm appears on the "Vital Auxiliaries" Safety Function screen.

B. The "Vital Auxiliaries" Safety Function box turns red, and the indicator for the affected AC bus on the electrical systems mimic flashes.

vC.

The "Vital Auxiliaries" Safety Function box turns yellow, and the indicator for the affected DC bus on the electrical systems mimic changes color.

D. All Safety Function boxes turn magenta and a small red box appears next to the indicator for the affected DC bus on the electrical systems mimic.

A is incorrect, indications are for a loss of 21 DC bus, listed indications are not supported by SPDS B is incorrect, indications are for a loss of 21 DC bus, listed indications are not supported by SPDS C is correct, per AOP-7J and SPDS alarm response manual.

D is incorrect, indications are for a loss of 21 DC bus, listed indications are not supported by SPDS 2004RO.TST Version: 0 Page: 10

2-HS-5155,22N22B SRW HXR EMERGENCY OUTLET VLVS handswitch is inadvertantly placed in 'OPEN'.

How are the Service Water Heat Exchangers affected?

A. 22N22B SRW HXR emergency outlets valves open, but normal SW flow is B. 22N22B SRW heat exchangers are removed from service because the heat maintained because the emergency overboard valve is normally gagged shut.

exchangers' SW inlet valves will also shut.

valve automatically opens, and 22N22B SRW heat exchangers SW outlets shift to 21 SW supply header.

D. 21N21 B SRW heat exchangers' SW inlet and outlet valves automatically shut, and 22N22B SRW heat exchangers will be supplied by 21 SW header.

<. 21N21 B SRW heat exchangers lose SW flow because the emergency overboard

18. AOP-7D-11 01 I//

NEW-ZCRO-202-7D/ 1.1/202.069/065AA1.04/3.5/3.4/AOP-7D, IN Unit-2 is at full power with 21 Plant Air Compressor in Standby when a leak causes Instrument Air header pressure to decrease to 85 psig. Plant Air header pressure is 95 psig. No operator actions have been performed.

Which list is composed of all the air compressors that will be running?

A. 21 and 22 Instrument Air Compressors, 11 and 21 Plant Air Compressors, 21 and B. 21 and 22 Instrument Air Compressors, 11 Plant Air Compressor, 21 and 22 C. 21 Instrument Air Compressor and 11 Plant Air Compressor 22 Saltwater Air Compressors Saltwater Air Compressors

4. 21 and 22 Instrument Air Compressors and 11 Plant Air Compressor A is incorrect, 21 Plant Air Compressor will not start until PA header pressure is less than 91 psig,the SW air compressors do not start automaticially on low air pressure B is incorrect, the SW air compressors do not start on low air pressure.

C is incorrect, 22 Instrument Air Compressor will start at 93 psig.

D is correct for the given pressures per AOP-7D and the ARM.

Basis: AUTO BACKUP FEATURES FOR INSTRUMENT AIR SYSTEM

References:

AOP-7D 41.7 2004RO.TST Version: 0 Page: 11

13. DROPPED CEA002//NEW-2/CRO-60-1/11.5/202.008/003AA2.01/3.7/3.9/SD 60 - CE Group 5 CEAs were being withdrawn from 120.5" to 128.0" using Manual Sequential.

CEA-1 dropped to 124.25" by secondary indication after CEDS was turned off.

Primary and secondary position indication for all other CEAs in the group is 127.25" to 1 28.0'.

What is the expected primary position indication for CEA-I?

./A. 127.25" to 128.OI

6. 124.25

c. 120.5" D. 0" A is correct. The primary (computer) indication is a pulse counting system, CEA-1 woulc get the same signals as the other CEAs within the group inManual Sequential.

B is incorrect, the stem of the question stated that the CEA dropped, not that it did not move above 124.25".

C is incorrect. This indication would be correct if the CEA did not move at all. If this were the case, the operator would have received alarms by 125" and stopped CEA motion.

D is incorrect, primary position is reset to zero by the rod bottom reedswitch.

Basis: CEA Indicator Post-trip Reset

References:

55.41 :2,6 55.43KAl: 060K4.08KA2: 01 4000K1.01 2004RO.TST Version: 0 Page: 12

20. LOSS OF WRNl 001// NEW-1/CRO-57-1-5/6.1.2/020570201/032 2.2.22/3.4/4.1/SD 57 - NU Given the following conditions:

Unit-2 is on Shutdown Cooling, RCS temperature is 120°F RCS pressure is 14.7 PSlA The reactor vessel head is fully tensioned Reactor Trip Circuit Breakers are open One of two operable WRNl channels has failed low What action is required immediately?

A. Commence boration of at least 40 GPM until RCS boron is 2300 PPM or greater.

C. Commence actions to restore two WRNl channels to operable status.

D. Perform SDM verification per Surveillance requirement 3.1.1.1.

A is incorrect, SDM margin verification is not required for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , which would leak to the boration,if required.

B is correct per Technical Specifications ( 3.3.12 action A) referenced in ARM D-41.

C is incorrect, but is a specific requirement if 2 Channels are 00s in Mode 6.

D is incorrect. SDM margin verification is not required for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months per TS 3.3.12 Basis: WR NI Requirements

References:

55.41:lO 55.43:2 I Tech Spec 3.3.1.I KAI: 057K8.1 KA2: K8.2,K8.4

4. Suspend all operations involving positive reactivity additions.

21. FUEL HANDLING ACCIDE 001/1 NEW-2 L0I-113-6/1.8/202.052/036AA2.0213.4/4.1/

Which of the following would be classified as a fuel handling incident per AOP-GD?

4. A large object was dropped in the Spent Fuel Pool and is laying on top of a spent fuel assembly.

location.

Area to the New Fuel Inspection Platform.

performing the refueling machine operational checks per 01-25C.

B. During refueling of the core, a fuel assembly was placed in an incorrect core C. A new fuel assembly was dropped when being moved from the New Fuel Storage D. A portable light pole hanging off the refueling machine bridge was damaged when A is correct per AOP-GD, rev. 15, I.B.2.

B-D are incorrect, the procedure is written to address an incident where there is a possibility of damage to an irradiated fuel assembly which could result in damage to a fuel pin.

2004RO.TST Version: 0 Page: 13

22. AREA RAD MON 0011 I NEW-2/77-1-012.0/ 079.008/

061 2.1.I/ 3.7B.W "RMS PANEL 1 C26" alarm at 1 C18 has annunciated.

2-RI-7010, Unit-2 BAST room Area Radiation Monitor is reading off-scale, high.

No other indications of abnormal conditions are present.

What action is directed by plant procedures?

A. Contact Chemistry to obtain samples of the BASTS, VCT and RCS.

6. Recommend Radiation Safety Supervision post the area.

D. Sound the emergency alarm, evacuate the immediate area and declare a

4. Obtain CRS permission to bypass the alarm to clear the alarm at 1 C18.

Radiological Event per ERPIP 3.0.

A is incorrect, there is no direction nor need to have chemistry take samples on an Area RMS alarm.

B is incorrect, Rad Safety should be called to take surveys.

C is correct per 01-35, rev.26 D is incorrect. The indications stated are indications of an instrument failure, a Rad Event should not be declared unless actual rad levels are high.

23. CRO-202-9A-249 0491 NONE/ NEW-11 CRO-202-9AJ 1.2,1.2.1/ 0220203011 067AK3.04/3.3/4.11AOP-9A During a severe fire in the Control Room, (AOP-SA), why are the Fairbanks Diesel Generators shutdown?

A. To prevent overloading the Diesel Generators when equipment starts, as the B. To ensure fuel is conserved for continued extended operation of the OC Diesel sequencers may not be operable.

Generator

4. To protect the engine from damage due to loss of cooling

0. To ensure MCCs 104 and 114 are de-energized to keep PORVs from failing open A is incorrect, Diesel loading is not a concern in this condition.

B is incorrect, enough fuel is available for the all diesels to operate within the bounds of this procedure.

C is correct per AOP-SA Unit 1 Bases, rev.8 page 2 of 11.

D is incorrect, but is the basis for opening the Load Center breakers for the MCCs.

Basis: AOP-SA Control Room evacuation positions

References:

NO-I -200 Att. 2 41.5, 41.10 2004RO.TST Version: 0 Page: 14

24. 1-OR-I 14-1-03-08 00811 BANK-11 114-1-0312.41201.0731074EA1.1614.414.61 CET On Unit-2 PAMS, what does a single "?' next to a CET temperature indication signify?

4. The indication is outside the quality check parameters.

B. The CET is the highest reading CET in it's quadrant.

C. The CET has been "bypassed", and the value is an old, non-updated indication.

0.

The indication is a calculated value, not an actual temperature measurement.

A is correct, per design documents, CCNPP-PAMS-0003-03.

B is incorrect, this would be indicated by ?? in the CET number and 0 for value on the C is incorrect, bypassed CET indications are preceded by a 'B' and have a blue background and does not revert to an old indication.

D is incorrect, the calculated value associated with the CET temperature is Tcrep.

C05 default screen.

Ref. CCNPP-PAMS-0003-03; CFR 41.7

25. CRO-107-1-3-55 05511 BANK-1/CRO-107-1-12.51202.0451076AK2.0112.6/3.0/ PROCES RAD Which phrase describes the relationship of RCS activity to the Process Rad Monitor'?

The Process Radiation Monitor:

4A. detects increases in specific isotopes due to fuel failures B. detects only increases in RCS activity specifically related to CRUD bursts C. measures RCS activity changes associated with Severe Accident Mitigation D. measures dose rates in the Letdown HX room at power due to CRUD bursts or A is correct per system description #41 B is incorrect, the PRM will detect increases in RCS activity, the specific isotope (I) distinguishes fuel failures from crud burst activity C is incorrect, letdown would be isolated in a SAM condition.

D is incorrect, the PRM does not measure dose rates.

Basis: Boron Concentration High Alarm

References:

55.41 : 10 55.435 / ARP C07 F-19KA1: 006K5.16KA2: 004000GEN8 scenarios fuel failures.

2004RO.TST Version: 0 Page: 15

26. EXCESS RCS LEAKAGE 001 / AOP-2A ATT/ BANK-2/ 202-2A-6/1.1/ 064.01 71 CEAl6AA1.1/3.4/3.6/

Using provided references, given the following Unit-2 information:

Reactor Power:

100%

Tc:

547.7"F and steady Letdown flow:

30 GPM Charging flow:

135 GPM PZR level:

RCS pressure:

Total CBO flow:

6 GPM Lowering at 2.5 inches/minute 2210 PSlA and slowly lowering What is the approximate RCS leak rate, in GPM?

A. 135

43. 146 C. 152 D. 172 6 is correct, 2.5 inches/minute( 18.9 GPM)= 47.25 +( 135-36)=146.25

References:

AOP-2A attachment 1. 41.7 Which one of the following conditions would allow you to exit EOP-8?

A. A plant cooldown has been completed, shutdown cooling flow has been established, and Core/RCS Heat Removal and Pressure/lnventory safety function status checks for EOP-8 are met.

4. All the safety function acceptance criteria for success paths implemented are being met, a single event diagnosis can be made and intermediate safety function status checks for single event are being met.

optimal recovery procedure, or an Operating Procedure, will address the safety functions such that EOP-8 final acceptance criteria for all the safety functions will be met.

D. In the case of multiple events, one event has been terminated, (such as a when the affected S/G goes dry during an ESDE) and all intermediate safety function status checks for EOP-8 are being satisified.

A is incorrect,EOP-8 will direct SDC operations, all safety functions must be met for the procedure you are tansiting to, or all EOP-8 criteria are satisfied.

B is correct per EOP-8,V.B. rev.26 C and D are incorrect, conflict with EOP-8 requirements.

Basis: EOP-8 exit conditions

References:

EOP-8 Rev. 3 Step F 41 5, 41.1 0 C. The CRS or STA has analyzed plant conditions and has verified that steps in an 2004RO.TST Version: 0 Page: 16

26. AOP-3F-06 00611 BANK-If CRO-202-3Ff 10.2/020050427/ 003 2.1.32f 3.4/3.8/ AOP3F. RC When restoring forced circulation it is necessary to verify the 4KV bus voltage greater than 41 00 volts prior to starting the RCPs.

What is the basis for this requirement?

./A. To prevent the 4 W degraded voltage relays from actuating upon RCP start.

8. To prevent tripping the oil lift pumps on low voltage when the first RCP is started.

C. Ensures that the running component cooling pump will operate within its design D. Ensures that excessive starting current is not developed which could damage RCP voltage limits.

windings.

A is correct per 01-1A Precaution L, rev.26 B is incorrect, lift pumps do not have undervoltage protection, and this is not listed as a concern.

C is incorrect, not part of the design limitations.

D is incorrect, not listed as a basis for the limit.

Basis: RCP RESTART CRITERINBASIS FOR ENSURING 4KV VOLTAGE > 3950 VOLTS

References:

AOP-3F KA1 : 02005A6.1 O W : 41.10, 43.2 2004RO.TST Version: 0 Page: 17

Given the following plant conditions:

-- Unit One has tripped due to a Loss of Offsite Power

-- 11 and 14 4KV busses are energized from the EDGs

-- Pzr level is 100" and slowly lowering How does this affect charging pump operation to restore Pzr level?

A. One charging pump starts automatically, the other charging pumps must be manually started and will stop automatically when Pzr level reaches + I 3 inches above program.

stop on Pzr level deviations from program.

automatically when Pzr level reaches +I 3 inches above program.

operated manually to control pressurizer level.

B. All 3 charging pumps must be started manually and will receive no signals to C.

All 3 charging pumps must be started manually and the backup pumps will stop D. One charging pump starts automatically the other charging pumps must be A is incorrect, none of the charging pumps will automatically start with the normal and alternate 4 KV bus feeder bkrs open.

B is incorrect, backup pumps will stop at +13" from program.

C is correct per Lesson Plan LOI-107-1 and electrical print 61 075, sh23 D is incorrect, none of the charging pumps will automatically start with the normal and alternate 4 KV bus feeder bkrs open.

References:

41.7

30. CRO-48-3-0-09 001// BANK-I/ CRO-48-3/3.4/004.010/ 005K2.01/ 3.01324 KV Which of the following is a possible cause when the following alarm has actuated?

--On panel 1 C19 "U44KV Eng SF Motor Overload" A. 152-1 204 (1 1 Condensate Booster Pump breaker) tripped B. 152-1 1 14 (U-440-11 A high side Feeder) tripped D. 152-21 07 (21 Containment Spray Pump breaker) tripped A is incorrect, Condensate Booster pumps are not ESF loads B is incorrect, this is a service transformer feeder breaker and does not supply an ESF motor C is correct per 1 C18 ARM M-04 D is incorrect, this is a Unit-2 load.

Basis: "U-I 4KV Eng. Sf. Fdr. Bkr Trip" Alarm on 1 C19

References:

41.7

4. 152-1 104 ( I 1 LPSl Pump breaker) tripped 2004RO.TST Version: 0 Page: 18

31. CRO-63-1-3-18 01811BANK-2CRO-63-1-3/1.7B.

1.7C1201.0431006K4.11/3.9/4.2/SD 63 - ES 1 During recovery from a LOCA on U-2, you are directed by the U-2 CRS to reset SIAS from the control room using the implemented EOP. Containment pressure is 2.0 psig and PZR pressure is 800 psia. What sequence of actions must occur to complete this action?

./A. Match required handswitches per the EOP attachment, block PZR pressure SIAS, and depress both SIAS channel reset pushbuttons.

B. Block Pzr pressure SIAS and depress either SIAS channel reset pushbutton.

C. Match required handswitches and depress both SIAS channel reset pushbuttons.

D. Block the Pzr pressure SIAS and depress both SIAS channel reset A is correct per EOP-5 and basis B is incorrect, without matching handswitches, SIAS cannot be reset from the Control Room, also, both reset pushbuttons must be depressed.

C is incorrect, without blocking Pzr pressure signals, SIAS will not stay reset.

D is incorrect, without matching handswitches, SIAS cannot be reset from the Control Room.

Basis: Steps for Evolution Requirement for Resetting SIAS

References:

55.41 :7,10 55.43:

5KA1: 063K4.03W: A4.02 pushbuttons.

32. CRO-113-5-5-19 01911 BANK-1ICR0-113-5-112.01202.0671008K3.0113.413.5/113 - SERV Unit 2 is in Mode 1 at 100% power when a loss of Component Cooling occurs.

Which condition from this event alone would require a manual Reactor trip?

A. Main Generator gas temperature of greater than 48°C for at least 15 minutes.

C. Component Cooling heat exchanger outlet temperature of 175°F.

D. Letdown is automatically isolated due to high temperature.

A is incorrect, this system is cooled by SRW and 48°C is not a trip criteria for Unit-2.

B is correct per 2C07 A&B ARM C and D are incorrect, these are not trip criteria in any procedure.

Basis: Loss of CC with Unit 2 in Mode 1 at 100% Power

References:

41.4, 41.7

43. RCP bleed off temperature of 200°F.

2004RO.TST Version: 0 Page: I 9

33. PZR QUENCH TNK 00111 MOD-2/ CRO-5-2-10/ 16.V 064.005/007A2.02/ 2.6/3.2/ SD 5 - RCS Unit 1 is in Mode 5, preparing for a plant heatup.

E01, QUENCH TK TEMP LVL PRESS is in alarm on 1 C06.

Given the following Quench Tank parameters:

1)

Pressure is 12 PSlG

2)

Temperature is 105°F

3)

Level is 29 inches What action is required?

4A. Open WGS CNTMT SOL valves, WGS-2180,218I-CVs and open QT VENT, 1 -RC-400-CV.

B. Place PORV handswitches, 1 -HS-1402 and 1-1404 in "OVERRIDE" C. Open Quench Tank Drain, 1 -RC-401 -CV D. Open Containment Nitrogen Supply Valve, 0-N2-238.

A is correct per 01-1 B and 1 C06 ARM, E-01 B is incorrect, in Mode 5, no PORV leakage would go to the QT C is incorrect, level is normal D is incorrect, this would pressurize the QT even more.

Basis: Data on Quench Tank

References:

55.41:lO 55.43 / 01-IBKAI:

005K5.08KA2: 007000K1.03 2004RO.TST Version: 0 Page: 20

34. COMPONENT CLG 00111 NEW-2 L0I-15-1-0/6.01202.0291008K4.09/2.7/2.9/

Unit-I was initially at 100% power when a major plant transient ocurred. The following conditions exist:

The 500 KV Red Bus was lost (P-I 3000-2 is de-energized)

RCS pressure is 1600 PSlA Tc is 532.4"F Contaiment pressure is 2.2 PSIG No other malfunctions occurred.

How many Component Cooling Pumps would be running, assuming no operator act ions?

A. 0 B. 1 4. 2 D. 3 A is incorrect, but would be true if only a loss of offsite power existed B is incorrect, but reflects normal conditions with no SlAS condition present C is correct, the SlAS will start 11 and 12 CC pumps, 1 B diesel will pick up 14 4KV bus. LD-58A and 61080 sh 15 D is incorrect, 13 CC pump would only start if the pump aligned to the same electrical bus failed to start.

References:

41.7 I

c RCS pressure is initially 2250 PSIG.

Spray Valve Controller, 1 -HIC-1 00 fails to a 0% output.

What is a direct result of this failure?

A. All Backup Heaters will energize if in "Auto".

B. Spray Valves I

-RC-1 OOE and F will fully open.

C. All Backup heaters will deenergize.

43. Proportional heaters will receive full power 2004RO.TST Version: 0 Page: 21

- 3' I RPS MALF 001llNEW-2158-1-0116.2,10.51058.0041012K3.02/3.2/3.3/

Unit-2 is at 16% power, with the Turbine Generator having just been paralleled with the grid.

A malfunction in RPS channel B causes the Power Trip Test Interlock (PTTI) to actuate.

How is the Turbine Generator affected?

A. A turbine trip will result due to ESFAS B logic cabinet initiating a turbine trip signal.

43. Trip logic will be reduced to 1 out of 3, since channel B Loss of Load trip unit will actuate.

C. The Turbine Generator will not be affected since the Loss of Load Trip is disabled.

D. RPS will initiate a Turbine Trip signal, but the signal is bypassed at ESFAS due to low reactor power.

A is incorrect, only one input is satisifed, so no trip signal is generated.

B is correct, power is greater than 15%, so loss of load is enabled, and PTTI will trip channel B loss of load trip unit. FSAR chapter 7 and lesson plan LOI-058-1 C is incorrect, Loss of Load is enabled at 15% or greater.

D is incorrrect, RPS is in 1/3 logic, ESFAS signal to trip the turbine is not affected by power level.

references: 4 1.7

37. RPS POWER SUPPLIES 001//NEW-1/58-1-01/8.3/058.007/012 2.4.47/3.4/3.7/

I Using provided references:

i If 1Y03 were de-energized, which RPS matrix power supply lights at 1 C15 would be ext i ng ui sed?

A. 5and15 B. 5,9and7

4. 8,12and 15 D. 9 and 10 C is correct per FSAR figure 7-2 All distractors are other power supplies but not affected by loss of 1Y03

References:

41.10, 43.2 2004RO.TST Version: 0 Page: 22

38 ESFAS 00111 NEW-3/201-6-712.01202.120/013Al.01l4.014.2/

A S/G tube rupture has been diagnosed, the correct EOP has been implemented and the following conditions exist:

RCS pressure is 1280 PSlA RCS temperature is 485°F PZR level is 85' 11A and 12B RCPs are running Cooldown rate is 95"F/hr, using TBVs The affected S/G has been isolated and pressure is 700 PSlG What action is required?

A. Secure the remaining RCPs to prevent exceeding pump curve limits.

B. Throttle HPSl flow to allow for backflow from the affected S/G as RCS depressurization continues.

injection begins.

affected S/G.

4. Lower TBV controller output to avoid exceeding cooldown rate limits when HPSl

0. Increase RCS depressurization using Main Spray to lower the leak rate into the A is incorrect, at 485"F, pumps can be run to 850 PSIA.

B is incorrect, HPSl throttling is not permitted below 101" PZR level.

C is correct, HPSl injection will start at about 1270 PSIA and will increase RCS cooldown rate as Cooler RWT water is injected.

D is incorrect, although depressurization is a main goal of the procedure, with these conditions, the RCS cooldown rate would be exceeded if injection occurs before steaming rate is reduced. Also, with 2 RCPs running Main Spray is not effective, EOP-6 directs the use of Aux. spray if depressurization is desired.

References:

EOP-6, 41.5 2004RO.TST Version: 0 Page: 23

Part of the 2003 modification to the LOCI sequencer advanced the start of the Service Water pumps from step 4 to step 0.

Why was this modification made?

A. Prevents overloading the Emergency Diesel Generators B. Prevents tripping the supply breakers for the safety related 4 KV buses C. Prevents damage to the Service Water Pump motors caused by excessively high starting currents Containment Air Coolers

43. Prevents a rupture of the Service Water system caused by water hammer in the

40. CONTAINMENT SPRAY 001/lNEW-1/052-3-1/13.0//026K2.01/3.4/3.6/

2A Diesel Generator is being taken out of service for routine maintenance.

Which component is also potentially affected? ASSUME NORMAL SYSTEM LINEUPS 4A. 21 Containment Spray Pump B. 22 Charging Pump C. 12 SFP Cooling Pump D. 23 Saltwater Pump A is correct, powered from 2A EDG fed 21 4KV bus, per 01-27C B is incorrect, powered from 24 480 volt bus, fed by 28 EDG C is incorrect, powered from 24 480 volt bus, fed by 28 EDG D is incorrect, powered from 24 4KV bus, fed by 28 EDG

References:

41.7 2004RO.TST Version: 0 Page: 24

4 1 ;ONTAINMENT SPRAY 0021/ NEW-11201-5-6/ 1.21 201.0631 026A2.07/3.6/3.9/

A LOCA has occurred, SIAS initiated and RWT level is 7 feet and lowering.

What actions are directed by the applicable EOP to prevent or mitigate cavitation of the Containment Spray Pumps?

A. Align a HPSl pump to the suction of a CS PP if discharge pressure lowers and

43. Verify Containment Sump level rises as RWT level lowers and, after RAS has I

amps fluctuate.

initiated, place a second CC HX in service.

than 28". When RAS actuates, place a second CC HX is in service.

attachments. After RAS has initiated, verify flow less than 1300 GPM.

c. Prior to RAS, place both CS PPs in PULL TO LOCK. Verify sump level is greater D. When RWT level lowers to 4 feet, throttle CS PP discharge valves per EOP A is incorrect, but aligning a CS PP to the suction of a HPSl PP is directed in the EOP if a HPSl is cavitating.

B is correct per EOP 5, Rev. 19 Section IV, steps P and S.

C is incorrect, but there is direction to verify the sump has at least 28l of water in it.

D is incorrect, but the procedure directs CS PPs to be placed in PTL if RWT level is 4 feet and CSAS has not initiated. It also directs throttling HPSl valves per an attachment.

References:

EOP-5, 41.5, 43.5

42. MN STM RMS 0011lNEW-21 L0(_77-1-011,2/202.0161039A1.09/2.5n.7/

With reactor power at 25%, what indication is available to the operator to monitor a S/G tube leak of 5 GPD?

A. Only S/G sample results reported by Chemistry B. N-16 monitors and Condenser Off Gas RMS

4. Condenser Off Gas RMS and Main Steam Line Radiation Monitors D. N-16 monitors only A is incorrect, newer instrumentation has allowed CCNPP to detect S/G tube leakage on the order of a few GPD.

B is incorrect, the N-16 indication is only accurate above 50% power, note in 01-35 states that below 50%, N-16 may indicate 0.00 GPD leak rate.

C is correct, the condenser off gas RMS has a reactor power input that is used in calculating the leakrate. The main steam line RMS will indicate a difference in the affected steam header vs. non-affected header, since it indicates actual steam line radiation levels. The Recorder would also be used to show trends.

D is incorrect, the N-16 indication is only accurate above 50% power, note in 01-35 states that below 50%, N-I6 may indicate 0.00 GPD leak rate.

References:

41.5 2004RO.TST Version: 0 Page: 25

4,.,

. -,RO-102-2-16 016// BANK-2/CRO-102-8/20.0/093.015/ 039A4.01/2.9R.W 102 - 1/2C

-The following conditions exist on Unit 2:

Reactornurbine trip has just occurred (Power prior to trip-I 00%)

S/G pressures are currently 850 psig What operator action (in the Control Room) must initially be taken to prevent an overcooling of the RCS per EOP-O?

A. Press "Close Valves" button on the turbine control panel.

4 3. Press "Reset" button on the MSR control panel.

C. Shut the MSIVs.

0. Press the BFV "reset" buttons.

6 is correct per EOP-0, Unit-2 basis.

Distractors are actions which are not directed by the procedures Basis: Reactor Trip With Unit 2 at 800 MWE and MSRs in Service

References:

41.7

44. CRO-103-2-4-82 08ZAOP-3GI MOD-2/ CRO-103-2-/3.0/ 202.0361 056A2.04/2.6/2.8/ T2G1 Unit 1 is operating at 50% power.

An electrical system malfunction occurs resulting in the loss of 12 and 13 Condensate Pumps.

What is the effect of this transient, and what action must be taken?

A. Reduced feed flow to the S/Gs and lowering levels will result. Bias feed pumps as required to maintain S/G levels.

B. Lower SGFP suction pressure will exist. Verify a Condensate Booster Pump automat i cal I y starts.

C. Reduced feed flow to the S/Gs and lowering levels will result. Trip the reactor and implement EOP-0.

43. Low suction pressure to the SGFPs and runout of the operating Condensate Pump will result. Reduce power to maintain condensate header flow less than 8,000 GPM.

A is incorrect, biasing feed pumps will cause additonal loss of NPSH to the SGFPs.

B is incorrect, Per AOPs-3G and 71, a main concern is runout of the condensate pump and increased wear and cavitiation, so power must be reduced.

C is incorrect, a trip should only be required if greater than 70% per AOP-71.

D is correct. This guidance is available in both AOP-71 and AOP3G

References:

55.41 :4 55.43.5 /

2004RO.TST Version: 0 Page: 26

45.

MAIN FEED 0 4 0011 NONE1 MOD-2/ L0145B-119,0,14.01 202.0401 059K4.0812.512.7/ FRDFWCS MA Unit-I is at 100% power. Both feedwater flow transmitter signals from 12 S/G to DFWCS fail low (out of range).

How is 12 FRV, l-FW-1121-CV, affected?

A. The last good feed flow input is used and 12 FRV control is shifted to the Backup B. Both CPUs fail and 12 FRV controller is shifted to "MANUAL".

C. An "1 1/12 S/G FW CONTR XFER INHIBIT" alarm is received, a shift from high CPU.

power to low power control mode will not occur and 12 FRV will be controlled by the Backup CPU.

in single element control.

43. Steam flow/feed flow error signal is not used and the Main CPU operates the FRV A is incorrect, with both signals out of range, a deviation alarm is sent, and the main CPU continues to operate the system in single element control.

B is incorrect, main continues to operate in Auto.

C is incorrect, the system would not shift to High Power Mode if in low power mode, but low power mode is unaffected.

D is correct, single element control is used with the loss of both steam flow channels or both feed flow channels.

See LOR LP 301-1-98, or ES-I99602497 Basis: Failure of a Feed Flow input signal offscale LOW

References:

01-12A 41.7 2004RO.TST Version: 0 Page: 27

46. UFW ISOL 001//NEW-2/048-1-0/ 1.7.D/201.055/059A3.06/ 3.2/3.3/

Unit-2 was initially at 100% power when a major plant transient occurred. The following conditions exist:

RCS pressure is 1800 PSIA Containment pressure is 0.4 PSlG 21 S/G pressure is 865 PSlG 22 S/G pressure is.680 PSlG Which list correctly identifies Main Feedwater/Condensate system automatic actions?

A. Both SGFPs trip, all Condensate Pumps trip, both Heater Drain Pumps trip, only 22 Main Feed MOV and 22 MSIVs shut.

4. Both SGFPs trip, all Condensate Booster Pumps trip, both Heater Drain Pumps trip, both Main feed MOVs shut and both MSIVs shut.

C. Both SGFPs trip, all Condensate Booster Pumps trip, both Heater Drain Pumps trip, only 22 Main Feed MOV and 22 MSlV shut.

0. Both SGFPs trip, all Condensate Pumps trip, all Condensate Booster Pumps trip, both Heater Drain Pumps trip, both Main Feed MOVs shut and both MSlVs shut.

A is incorrect, Condensate Pumps do not receive SGIS B is correct per 2C03 ARM and 2-LD58A.

C is incorrect, both MSIVs and MOVs shut.

D is incorrect, Condensate Pumps do not receive SGIS.

References:

41 5, 43.5 SCO4RO.TST Version: 0 Page: 28

What is the basis for the AFW flow controller automatic setpoints of 150 GPM?

A. S/G levels will be restored to EOP-I limits within I O minutes of AFAS actuation with MFW isolated, and AFW suction piping flow limits are not exceeded.

B. EDG ratings are not exceeded on SIAS with a Loss of Offsite Power, and S/G inventory is adequate for worst case decay heat with 2 trains of AFW operating.

low enough to prevent initiating SIAS due to RCS overcooling with 2 trains operating.

of AFAS Block to the affected S/G with no operator action, yet low enough to prevent RCS cooldown to less than 525°F with one train operating.

4. AFW flow will be adequate with one AFW train to remove highest decay heat, but D. AFW flow will be adequate to maintain S/G level in the unaffected S/G in the event

48. AFW XCONN 001//

MOD-21 CRO-34-2-7/9.0/201.031/061A1.03/3.1/3.6/ LIMITS A Loss of Offsite Power exists with Unit-I previously at 100% power and Unit-2 in Mode 5. Unit-2 has been unable to restore Shutdown Cooling and is using 13 AFW Pump to restore S/G levels.

Unit-I is using 11 AFW Pump to feed I 1 and 12 S/Gs at 150 GPM per S/G.

What is the flow limit for 13 AFW pump to supply Unit-2?

A. 275 GPM 143.300 GPM 1 C. 600 GPM D. 900 GPM Per OI-32A, Rev.19, Unit-I 6.3.C.1, total AFW flow should be less than 600 GPM when feeding both units from a single AFW System.

600-300 =300, so B is correct..

A is incorrect, but represents the Motor limit of 575 GPM, minus the 300 GPM to Unit-I carried by the steam pump.

C is the total flow limit for one system.

D is the 1200 two system limit minus, 300 GPM used on Unit-1 Basis: AFW flow Iimits

References:

10CFR55.41: 7 - SD 01-32KA1:

02034K6.02 KA2: 06 1 OOOK4.04 2004RO.TST Version: 0 Page: 29

49.

AC DISTRIBUTION 00111 NEW-2/ 202-71-41 3.01 202.097/ 062A2.01/ 3.413.91 Unit-1 is at 100% power when 13B 480 Volt Bus is lost.

What is the major affect to the plant, and what action must be taken?

A. Boration via the RWT from all operable Charging Pumps causes power to decrease. Place 2 Charging Pumps in Pull-To-Lock and shift suction back to the VCT.

B. All Circulating Water Pumps lose excitation. Trip the Reactor and implement C. Feedwater Heater Level Dump Valves fail open, reactor power increases. Reduce EOP-0.

Reactor power, match HLDV handswitches and tie 1YO9 and IYIO.

cooling. Tie MCCs 106T and 116T.

4. 12 and 13 Condensate Pumps' bearing temperatures rise due to loss of lube oil A is incorrect, this describes a loss of MCC-104 or 14A 480 Volt Bus.

3 is incorrect, this describes a loss of 15 480 Volt Bus.

C is incorrect, this describes a loss of MCC-114 or 11A 480 Volt Bus.

D is correct per AOP-71.

References:

AOP-71, 41.5, 43.5

50. CRO-7-1-5-134 134// BANK-1ICRO-7-1-5/31.2/201.050/063A4.01/2.8/3.1/ SD 7 - SAF The 13 HPSl pump breaker charging spring has failed to charge after securing the pump for an STP. How will this condition be detected in the control room?

A. 13 HPSl PP BKR UU IMPR alarm

43. 13 HPSl PP SlAS BLOCKED AUTO START alarm C. U-I 4KV ESF MOTOR OVERLOAD alarm D. ACTUATION SIGNAL BLOCKED alarm A is incorrect. This alarm is caused by any combination other than its disconnect shut with the breaker racked in. Chraging spring will not affect it.

B is correct per ARM H-I 9 C is incorrrect, but would actuate if 13 HPSl supply breaker tripped.

D is incorrect, this alarm is caused by vital 4 KV feeders being open witha UV on the bus.

Basis: 13 HPSl Pump Breakder Charging Spring Failed to Charge

References:

55.41 :7 55.43 / CO9 Alarm ManualKAl : 07K4.01 KA2: 006000K6.03 2004RO.TST Version: 0 Page: 30

5 1 iDG SYS 001//NEW-2/LOl-02-1-0/3.01002.010/064K1.04/3.6/3.9/

1A Diesel Generator is out of service for maintenance when a Loss of Offsite Power occurs.

2B Diesel Generator did not load due to a faulted 4 KV bus.

What affect does this have on the DC electrical distribution system as indicated at 1 C24A?

A. 11 DC bus will be supplied only by 11 battery.

B. 21 DC bus will be supplied by 21 battery charger.

c. 12 DC bus will be supplied by 24 battery charger.

4. 22 DC bus will be supplied by 22 battery charger.

A is incorrect, 11 bus will have power from 23 battery charger B is incorrect, 21 battery charger is powered from 24A 480 volt bus, which remained deene rg ized.

C is incorrect, 24 battery charger is powered from 24B 480 volt bus which remained deenereg ized D is correct, 22 battery charger is powered from 21 B 480 volt bus, carried by 2A Diesel Generator

References:

41.2-41.9

52. CRO-122-1-3-07 0091 I BANK-I1 LOI-77-7 -01 1.01 079.0081 073K5.01 I 2.513.01 What type of radiation do the Component Cooling, Service Water and S/G Blowdown Recovery (process rad. monitors) detect?

A. Alpha

3. Beta

4. Gamma D. Neutron C is correct per SD-77.

References:

41.5 2004RO.TST Version: 0 Page: 31

During normal operation at 100% power, what is the largest heat load on the Service Water system?

./A. Main Generator Hydrogen Coolers.

6. Hydrogen Seal Oil Coolers C. Containment Air Coolers D. 16 Diesel Generator A is correct per AOP-7B, loss of SRW.

r After a SIAS actuation, what is the source of Instrument Air supplied to the AFW flow control valves?

Basis: Largest Heat Load(s) on Service Water System During A Trip of SW

References:

41.5 A. Saltwater Air Compressors

43. The opposite unit's Plant Air Compressor C. Auxiliary Feedwater system air accumulators D. Nitrogen backup to instrument Air A is incorrect, a manual valve, 1-IA-728 or 2-IA-314 (or 317) must be opened. and header pressure must be less than 85 PSlG B is correct, the PNIA cross connect opens at 88 PSlG and will return pressure to normal.

C is incorrect. The accumulators will supply air if the header pressure is less than 85 PSlG D is incorrect, nitrogen backup requires manual valve operation and lower instrument air header pressure.

References:

SD. 41.7 2004RO.TST Version: 0 Page: 32

5 5 CONTAINMENT 04 012// NEW-2/212-1-411.3/204.0761103A2.05/2.9/3.9f

[-Which of the following is a requirement for a containment entry at power?

44. Containment airlock door seals must be tested within 7 days after the last entry I

B. Someone must be stationed outside the airlock door while it is open

c. A containment vent must be performed prior to containment entry 1 D. An FME log (MN-1-109 att.5) is required A is correct per EN-4-105 and NO-I -1 04 B is incorrect, there is no requirement to have anyone at the containment door while open.

C is incorrect, there is no requirement to vent containment prior to entry.

D is incorrect, MN 1-1 09 does not require the log. FME concerns are adressed by NO-1 04 checklists.

References:

EN-4-105, NO-I -1 04 41 5, 43.5

56. CRO-60-1-04 004/1 BANK--UCRO-60-11 10.1,2, 11 055.0031 001A4.0813.713.4/ SD 60 - CE Under which condition can CEAs be WITHDRAWN in the manual sequential mode?

(without using CMI bypass features)

./A. Tavg-Tref deviation alarm.

B. Group 5 CEAs below the PDIL.

C. 2 out of 4 TM/LP channel pretrips at RPS.

D. A misaligned CEA 7.5 inches from its group.

A is correct, per 1 C05 ARM, all other conditions result in a CWP I

Basis: Withdrawing of CEAs in Manual Sequential Mode

References:

55.41 :2,6 55.43KA1: 060K4.06KA2: K11.01 2004RO.TST Version: 0 Page: 33

5 I. CRO-5-2-10-06 001//

BANK-2/CRO-5-2-3118.1/078.010/002A1.0614.0/4.0/

The reactor is at steady state conditions and turbine load has been adjusted to maintain Tc on program.

Given the following:

T cold is 538°F T hot is 556°F What is reactor power?

A. 18%

B. 34.5%

4. 37.5%

D. 40.5%

Tc @O%

is 532, @ I 00% is 548, 1611 00=6/x so, x=37.5%

Or delta T at 0% is 0, delta T at 100% is 48, existing delta T is 18, 18/48=xl100=37.5

References:

41.5 2004RO.TST Version: 0 Page: 34

Which statement satisfies the requirements for minimum operable position indication channels for a CEA?

A. CEA voltage divider reed switch position indicator channel capable of determining the absolute CEA position within i 6 inches CEA pulse counting position indicator channel.

the absolute CEA position within k7.5 inches or CEA "Full Out" reed switch position indicator channel only if the CEA is fully withdrawn as verified by actuation of the applicable position indicator.

and

8. CEA voltage divider reed switch position indicator channel capable of determining

4. CEA voltage divider reed switch position indicator channel and CEA pulse counting position indicator channel in agreement within 4.5 inches.

D. CEA voltage divider reed switch position indicator channel capable of determining the absolute CEA position within i1.75 inches of absolute position or CEA "Full Out" reed switch position indicator channel only if the CEA is fully withdrawn as verified by actuation of the applicable position indicator.

A is incorrect, no requirement exists for the 6l limit.

B is incorrect, 7.5 inches is the T.S. limit for deviation between CEAs within a group, and 2 position indicator channels are required.

C is correct, reflects wording in TNC 15.1.5 D is incorrect, the statement involving 1.75 inches was deleted, and 2 channels are required, not one.

Basis: CEA Position Channels

References:

55.41 :2,6,10 55.43:2 2004RO.TST Version: 0 Page: 35

Which condition would cause audible WRNI count rate to rise?

A. Pulling CEAs to criticality when performing the first reactor startup following a refueling outage

43. Reinserting a once-burned fuel assembly in a new core location C. During RCS drain down to reduced inventory for RCP seal replacement D. Withdrawing CEA #I from a fuel assembly while swapping CEAs I

60. H2 ANALYZER 001// NEW-I/ LOl-052-5/ 12.81 073.0031 028A4.0313.1/3.3/

How are the sample locations indicated on the Hydrogen Analyzer recorders on 1 (2)ClO selected?

A. Manually at the recorder B. Automatically or manually by the plant computer D. Automatically at the recorder C is correct per SD-38B and Chemistry Procedures

References:

41.7

4. Automatically or manually from sample panels in the Aux. Building 2004RO.TST Version: 0 Page: 36

6 i CONTAINMENT PURGE 001// NEW-2/ CRO-122-112.0, 3.01 0201 5020W 029K4.03/3.2/3.5/

Refueling operations are in progress and Containment Purge is in operation. While taking logs in the Cable Spreading Room, the CRO notices that channel ZF of CRS is bypassed.

How does this affect Containment Purge?

A. Containment Purge will be automatically secured if any other channel of CRS B. In the event of a valid CRS signal, one Containment Purge CV will remain actuates.

open.

Specification requirements.

Purge to remain in operation

4. Containment Purge must be secured (or fuel movement suspended), per Technical D. The remaining channels of CRS must be verified operable to allow Containment A is incorrect, with one sensor bypassed, it requires 2 channels to trip.

B is incorrect, bypassing a sensor will not effect how the components reposition on a valid signal.

C is correct. TS 3.3.7 requires all sensors to be operable during fuel movement.

D is incorrect per TS 3.3.7

References:

55.41.7, 55.43.2. Related to Calvert Cliffs LER involving refueling with one channel of CRS inoperable during 2001 refueling outage

62. CRO-113-4-3-06 001// BANK-ZCRO-113-4/ 3.5/202.057/033K3.03/3.0/3.3/

High Spent Fuel Pool temperature is corrected by what action?

A. Adjusting spent fuel pool temperature controller setpoint.

B. Throttling 1 I N B SRW heat exchanger Saltwater outlet valves open.

C. Adjusting SFP CLR OUT THROTTLE valve to obtain a discharge pressure of greater than 120 psig.

43. Throttling open SFP CLR DISCH HDR stop valve.

A is incorrect, there is no controller for SFP cooling.

B is incorrect, no spent fuel pool cooler is cooled by 1 I SRW header.

C is incorrect, but is the method for controlling SFP cooling system flowrate to prevent pump runout.

D is correct per 1 C13 ARM and 01-24A

References:

41.7 2004RO.TST Version: 0 Page: 37

63. LIQUID RADWASTE 002//

N E W 4 LOI-32-lI16.0,17.0/ 064.0411 068K5.041 3.23.51 Performing which evolution poses the highest radiological risk to the operator?

A. Discharging the contents of 12 RCW Monitor Tank C. Filling I 1 RCW Ion Exchanger with resin D. Recirculating 11 RCWMT through a MWS prefilter A is incorrect, no breach of the system is required, no entry into a contaminated area is required.

B is correct, 01-1 7C-1 contains cautions for the operator to help avoid a radiogas contamination.

C is incorrect, the areas requiring access to fill the ion exchangers are radioactively clean, and do not handling of contaminated equipment for the most part.

D is incorrect, no breach of the system is required, no entry into a contaminated area is required.

References:

41.5

43. Filling 21 RCW Degassifier Vacuum Pump reference leg

64. CRO-134-1-5-45 04511 BANK-1lCRO-134-1-12.5.Cl0243B05031072K1.0413.313.51134 - HVAC Control Room Vent RMS, 0-Rl-5350 is in alarm.

How is the Control Room HVAC system affected?

A. Outside air dampers open to purge the Control Room, and the air conditioning unit is shutdown operation and the kitchen exhaust fan secured.

fans in operation.

43. Control Room ventilation is in recirculation with Post-LOCI filter fans in

c. The Control Room HVAC shifts to winter mode of operation with Post-LOCI filter D. Control Room air handling unit is secured. Only outside air dampers open.

A is incorrect, outside flow paths are secured.

B is correct per LP 134 and system description.

C is incorrect, outside air dampers remain open in winter mode.

D is incorrect, outside air dampers remain shut.

Basis: Automatic Action of The Control Room HVAC on High Rad Signal

References:

55.41 :4 55.43 / SD 43BKA1: 43BK4.09KA2: 072000K4.03 2004RO.TST Version: 0 Page: 38

65 tIRE SYSTEM04 001//BANK-l///013.024/086A3.01/2.9/3.3/

What condition will start the diesel fire pump?

A. Fire main header pressure less than 105 PSlG

8. A smoke detector or temperature detector actuation

4. Both electric fire pump feeder breakers being open D. Preaction solenoid valve or sprinkler alarm check valve actuation A is an incorrect setpoint, pressure must be less than 85 PSlG B and D are incorrect, but will cause "FIRE PROT PANEL 1C24B" alarm C is correct, per 1 C17 ARM L-06 References; 41.7

66. SHIFT TURNOVER04 OOl//MOD-I/ SRO-204-1J4.0/204.043/2.1.3/3.0/3.4/204 - SHlF Which category of deficient equipment status should be annotated on the Shift Turnover 1nformation.Sheet to communicate the status of 21 Condensate Pump which has a broken lube oil pump?

./A. (00s)

Out Of Service B. (VF) Inoperable But Functional C. (D) Degraded D. (0) Inoperable A is correct, although the pump could be run until bearing temperatures rise, this wouic not be prudent and AOP-71 directs a power reduction when power is lost to these pumps. NO-1-207 rev. 34 page 13 B is incorrect, this designation is for TS equipment only.

C is incorrect, degraded implies that the equipment can perform it's designed function.

D is incorrect, this desigination is for TS equipment only.

Basis: Complete the Shift Turnover Information Sheet

References:

NO-1-207 41. I O 2004RO.TST Version: 0 Page: 39

E. ' CONDITIONSLIMITS 001/INEW-1/042-119.0/042.002/2.1.1012.7/3.9/

What is the condenser differential temperature (condenser delta T) limit, as stated in the facility license?

A. The calculated flow weighted hourly average of the temperature rise across both

6. The calculated flow weighted hourly average of the temperature rise across each

4. The calculated average of the 24 flow weighted hourly readings of both units for a D. The calculated average of the 24 flow weighted hourly readings of each unit for a condensers is limited to 12°F condenser is limited to 12°F.

calendar day is limited to 12°F.

A is incorrect, the limit is not an hourly average.

B is incorrect, the limit is not an hourly average, and is for both units.

C is correct per 01-14A, rev 15, precaution A.

D is incorrect, the limit is for both units, not each unit.

References:

43.1

68. SYSTEM STATUS COMM 001// NEW-2/ AOP-W /204.013/ 2.1.14/2.5/3.3/

Given Nuclear Plant Operations Section Standing Order 03-03:

A known Component Cooling system leak is causing a Unit-2 sump frequency of 3.4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Sump frequency changes to 95 minutes with a corresponding increase in unidentified RCS leak rate.

Which method of informing the GS-NPO is required per administrative procedures?

A. Voicemail B. Alpha-page C. Alpha-page and detailed voicemail

43. Talk directly A is incorrect, this is only allowed for RCS Leakage condition 1.

6 is incorrect, this is for RCS Leakage Condition 2 C is incorrect, this is the method to be used to contact OTHER site management.

D is correct, with containment sump frequency less than 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , Condition 3 exists, per the first action, direct communication with GS-NPO and PE-PSE is required.

References:

41.10, 43.5 2004RO.TST Version: 0 Page: 40

69. FUEL MOVES 00111 NEW-2/209-0-08lA-41032.00212.2.28/2.613.5/

Which condition requires that the Spent Fuel Pool Ventilation charcoal filters be placed in service?

A. Spent fuel is being loaded into an ISFSI storage cask.

B. New fuel is being loaded into the Spent Fuel Pool.

C. A dummy fuel assembly is being transferred from the Spent Fuel Pool to the Refueling Pool for RFM testing.

43. Refueling is in progress which does not include a complete core offload.

A is incorrect, fuel loaded into the cask has not been in a critical core within the previous 32 days.

B is incorrect, new fuel has not been irradiated C is incorrect, the dummy is not recently irradiated fuel as defined in 01-22D (and in A, above).

D is correct, refueling involves the transport of "recently irradiated fuel"--part of a critical core within the last 32 days-- unless in an extended outage.

References:

OI-22DI 41.7, 41.8, 41.I 0, 43.7 70 CFA PROGRAM O O l l l NFW-1f 60-1-1.31 I 2 3 O!%.O12/ 2.2.33 2.5/2,%/

I Where is the regulating group CEA "All Rods Out" (ARO) position stated?

./A. NEOP-13 (23)

B. COLR figure 3.1.6 C. System 55 (CEDS) setpoint manual D. 01-42, CEDM System Operation A is correct. Figure IV. 6.1 B is incorrect, not in COLR C is incorrect, setpoint manual has CEDS setpoints, but not CEA position D is incorrect, the 01 is not updated each RFO to reflect CEA ARO position.

References:

41. IO, 43.6 2004RO.TST Version: 0 Page: 41

' What documents, used by Operations personnel to run the plant, are updated to communicate the core reactivity effect changes due to core age or fuel composition?

A. USFSAR and NFM Operator Surveillance Procedures (NEOP-3011302)

B. TRM and Offsite Dose Calculation Manual

4. COLR and Technical Data Book (NEOPs)

0.

Calvert Cliffs Operating Manual and Technical Specification LCOs

72. RADIATION CONTROL04 001// NEW-1/217-30-9/ 1.7.1 1204.1 39/2.3.10/ 2.9/3.3/

The Shift Manager has declared an Alert per ERPIP 3.0 The Operational Support Center is not yet staffed.

A plant operator is required to perform a task in the Auxiliary Building where dose rates are unknown.

What is required prior to the operator being sent to perform the task?

A. The 2-person rule must be invoked, the operator is not allowed to work alone.

B. The Shift Manager must approve the action and the selection of personnel to

c.

The Shift Radiation Technician must be contacted to assess radiological conditions D. A pre-evolution brief must be held with the Interim Radiation Protection perform the task.

and preferred access and egress routes.

Director and the CRS in attendance.

A is incorrect. 2-person rule is for a security threat, not an unknown radiation hazard.

B is incorrect. There is no requirement for the SM to provide this oversight, and it is not a task for the SEC.

C is correct per ERPIP-108 and ERPIP-103 6.2.B.2.c D is incorrect, Security and CRS attendance is not a requirement.

References:

41. IO, 41.12 43.4 2004RO.TST Version: 0 Page: 42

7 3. RADIATION RELEASE 001// NEW-1/CRO-134-1-/2.4.F/093.080/2.3.11/2.7/3.2/

Which operation always requires an approved Discharge Permit?

A. Dumping Condensate to the Circulating Water System after system cleanup

8. Discharging S/G sludge lancing water C. Dewatering the Saltwater side of a Component Cooling Heat exchanger AI. Initiating S/G Blowdown to Circulating Water A is incorrect, a sample is required, but not a permit B is incorrect, a sample is required, but not a permit.

C is incorrect, no permits are required.

D is correct, per 01-8A.

References:

4.1. I O, 41.12

74. SRO-201-8-1-15 01 5/ / BANK-I/ SRO-201-84 1.3/ /2.4.17/

3.1B.W 201 - EMER Which one of the following defines the term "success path" as it applies to EOP-8?

+'A. A course of action based on plant conditions used to address a safety function.

B. A series of actions which, if performed correctly, will allow the CRS to make a single C. A table which directs the operator to a set of actions to assess a safety function.

D. A form which provides criteria for the STA to use in evaluating safety function status.

Basis: Definition of success paths

References:

EOP-8 Rev. 3 Basis document page 3KA1: KA2:

event diagnosis.

2004RO.TST Version: 0 Page: 43

i VERIFY ALARMS 00111 NEW-2/ 101201-0-91 1.61 204.1 381 2.4.461 3.513.61 1 I Saltwater pump tripped due to a motor overload and reactor trip criteria were reached before the system could be recovered.

The RO manually tripped the reactor from 100% and all systems responded normally.

Which Control Room panel would have no alarms annunciated?

A. IC18

6. IC13 v/c. IC08

0. IC03 A is incorrect, U-I ESF MOTOR OVERLOAD would be on given 11 SW pump tripped.

B is incorrect, 1 I SW HDR PRESS LO at a minimum, would be on.

C is correct, any alarm on this panel indicates something more severe than an uncomplicated trip is happening and some ESFAS actuation is ocurring or required.

D is incorrect, lowering S/G levels due to the trip will cause alarms.

References:

41 5, 41.7, 43.5 2004RO.TST Version: 0 Page: 44

7 4ADIATION RELEASE 00111 NEW-1/CRO-134-1-/2.4.F/093.080/2.3.1112.7/3.2/

[ What operation requires an approved Discharge Permit?

\\/A. Initiating S/G Blowdown to Circulating Water

8. Pumping the Containment Sump

c. Dewatering the Saltwater side of a Component Cooling Heat exchanger

0.

Placing 12 RCWMT in recirculation

74. ERPIP DEFINITIONS 003//

NEW-I/ 21730-8/ 1.lW 204.137/ 2.4.17/3.1/3.81 Per the ERPIP, what is the area that should be considered when applying the Severe Weather Conditions criteria for procedure implementation?

A. CCNPP Protected Area

6. CEG service territory within the state of Maryland

4. CCNPP or any of the 500 KV tie lines rights of way

0.

Within the ten mile radius of CCNPP B is correct per ERPIP 3.0, rev. 31 Attachment 20

References:

41.10

75. VERIFY ALARMS 0011l NEW-2/ L01201-0-9/ 1.6/204.13812.4.46/ 3.5l3.61 11 Saltwater pump tripped due to a motor overload and reactor trip criteria were reached before the system could be recovered.

The RO manually tripped the reactor from 100% and all systems responded normally.

Which Control Room panel would have no alarms annunciated?

A. IC18 B. 1C13

4. IC08

0.

1C03 A is incorrect, U-1 ESF MOTOR OVERLOAD would be on given 11 SW pump tripped.

B is incorrect, 11 SW HDR PRESS LO at a minimum, would be on.

C is correct, any alarm on this panel indicates something more severe than an uncomplicated trip is happening and some ESFAS actuation is ocurring or required.

D is incorrect, lowering S/G levels due to the trip will cause alarms.

References:

41 5, 41.7, 43.5 2004RO.TST Version: 0 Page: 42

Fi-iday, February 20,2004 @ 09:29 AM Answer Key Page: 1 Test Name: 2004RO.TST

'l'cst Date:

Thursday, February 19,2004 L

Answer(s)

Question ID Type Pts 0

1 2

3 4

5 6

7 8

9 1:

1 SRO-201-0-2-05 005 MC-SR 1

D A

B C

D A

B C

D A

1:

2 VAPORSPACEACC 001 MC-SR 1

B C

D A

B C

D A

B C

1 :

3 CONTAINMENTCOOLING 001 MC-SR 1

D A

B C

D A

B C

D A

1:

4 SRO-201-5-1-06 006 MC-SR 1

- B C D A

B C

D A

B C

nn 1 n

1 r n A n r n A D r ~~~~~

"V 1 I.

v Y

L A.

Y

1.

Y w

1 D

A B

C D

A B

C D

A y/TQL.

i 1:

6 LOSSOFRCSMAKEUP 001 MC-SR 1:

7 CRO-113-5-5-25 025 MC-SR 1

B C

D A

B C

D A

B C

1:

8 PRESSURTZERPCSMAZF 001 MC-SR 1

C D

A B

C D

A B

C D

1:

9 SRO-201-0-3-29 029 MC-SR 1

A B

C D

A B

C D

A B

1:

11 STEAMLINERUPTURE 001 MC-SR 1

D A

B C

D A

B C

D A

1 :

12 LOSS OF FEEDWATER 001 MC-SR 1

A B

C D

A B

C D

A B

1:

13 STATIONBLACKOUT 002 MC-SR 1

C D

A B

C D

A B

C D

1:

14 LOSSOFOFFSITE 001 MC-SR 1

D A

B C

D A

B C

D A

1:

15 LOSSOFVITALAC 001 MC-SR 1

C D

A B

C D

A B

C D

1:

16 LOSSOFDCPOWER 001 MC-SR 1

C D

A B

C D

A B

C D

1:

17 LOSSOFSRW 001 MC-SR 1

C D

A B

C D

A B

C D

I :

18 AOP-7Dl1 011 MC-SR 1

D A

B C

D A

B C

D A

1:

19 DROPPEDCEA 002 MC-SR 1

A B

C D

A B

C D

A B

1:

20 LOSSOFWRNI 001 MC-SR 1

B C

D A

B C

D A

B C

1:

21 FUELHANDLINGACCIDE 001 MC-SR 1

A B

C D

A B

C D

A B

1:

22 AREARADMON 001 MC-SR 1

C D

A B

C D

A B

C D

1 23 CRO-202-9A-2-49 049 MC-SR 1

C D

A B

C D

A B

C D

1 : Di LOR-1 14-1-03-08 008 MC-SR 1

A B

C D

A B

C D

A B

1 : 25 CRO-107-1-3-55 055 MC-SR 1

A B

C D

A B

C D

A B

1:

26 EXCESSRCSLEAKAGE 001 MC-SR 1

B C

D A

B C

D A

B C

1 : 27 SRO-201-8-1-18 018 MC-SR 1

B C

D A

B C

D A

B C

1:

28 AOP-3F-06 006 MC-SR 1

A B

C D

A B

C D

A B

1 :

29 CRO-107-1-9-01 001 MC-SR 1

C D

A B

C D

A B

C D

1:

30 CRO-48-3-0-09 001 MC-SR 1

C D

A B

C D

A B

C D

1:

31 CRO-63-1-3-18 018 MC-SR 1

- A B C D A B C D A B 1 : 32 CRO-113-5-5-19 019 MC-SR 1

B C

D A

B C

D A

B C

1:

33 PZRQUENCHTNK 001 MC-SR 1

P B

C D

A B

C D

A B

1.

1 4 I d 1-c I nn 1 h

m 1

1 v

n

~

a r

n

~

r n

n 1:

34 COMPONENTCLG 001 MC-SR 1

C D

A B

C D

A B

C D

1:

36 RPSMALF 001 MC-SR 1

B C

D A

B C

D A

B C

1:

37 RPSPOWERSUPPLES 001 MC-SR 1

C D

A B

C D

A B

C D

1:

38 ESFAS 001 MC-SR 1

C D

A B

C D

A B

C D

1:

39 CONTAINMENTCOOLING 003 MC-SR 1

D A

B C

D A

B C

D A

1:

40 CONTAINMENTSPRAY 001 MC-SR 1

A B

C D

A B

C D

A B

1:

41 CONTAJNMENTSPRAY 002 MC-SR 1

B C

D A

B C

D A

B C

1:

42 MNSTMRMS 001 MC-SR 1

C D

A B

C D

A B

C D

1 :

43 CRO-102-2-16 016 MC-SR 1

B C

D A

B C

D A

B C

1:

44 CRO-103-2-4-82 082 MC-SR 1

D A

B C

D A

B C

D A

1 :

45 MAINFEEDO4 001 MC-SR 1

D A

B C

D A

B C

D A

1:

10 SRO-201-6-1-30 030 MC-SR 1

. B C D A

B C

D A B C

Friday, February 20,2004 @ 09:30 AM Answer Key Page: 2 Test Name: 2004RO.TST 1 cct Date.

Answer($

Thursday, February 19,2004 Question ID Type Pts 0

1 2

3 4

5 6

7 8

9 I

46 MFWISOL 001 MC-SR 1

B C

D A

B C

D A

B C

I 47AFW04 001 MC-SR 1

C D

A B

C D

A B

C D

1 48 AFWXCONN 001 MC-SR 1

B C

D A

B C

D A

B C

1 49 ACDISTRIBUTION 001 MC-SR 1

D A

B C

D A

B C

D A

I SO CRO-7-1-5-134 134 MC-SR 1

B C

D A

B C

D A

B C

1 51 EDGSYS 001 MC-SR 1

D A

B C

D A

B C

D A

1 52 CRO-122-1-347 009 MC-SR 1

C D

A B

C D

A B

C D

1 53 CRO-113-2-5-24 024 MC-SR 1

A B

C D

A B

C D

A B

I 54 INSTRUMENTAIR04 002 MC-SR 1

B C

D A

B C

D A

B C

1 55 CONTAINMENT04 012 MC-SR 1

A B

C D

A B

C D

A B

I 56 CRO-60-1-04 004 MC-SR 1

A B

C D

A B

C D

A B

I 57 CRO-5-2-1046 001 MC-SR 1

C D

A B

C D

A B

C D

1 58 CRO-60-1-45 045 MC-SR 1

C D

A B

C D

A B

C D

1 59 NUCLEARINSTRUMENTS 001 MC-SR 1

B C

D A

B C

D A

B C

1 60 H2ANALYZER 001 MC-SR 1

C D

A B

C D

A B

C D

1 61 CONTAINMENTPURGE 001 MC-SR 1

C D

A B

C D

A B

C D

1 62 CRO-113-4-3-08 008 MC-SR 1

C D

A B

C D

A B

C D

1 63 LIQUIDRADWASTE 002 MC-SR 1

B C

D A

B C

D A

B C

1 64 CRO-134-1-5-45 045 MC-SR 1

'B C D A B C D A B C I

65 FIRESYSTEM04 001 MC-SR 1

C D

A B

C D

A B

C D

I 66 SHIFTTURNOVER04 001 MC-SR 1

A B

C D

A B

C D

A B

1 67 CONDITIONSLIMITS 001 MC-SR D

A B

C D

A B

C D

1 68 SYQTFM-

A J h 1

69 FUELMOVES "o"0: EE $1 X

X A

1 70 CEAPROGRAM 001 MC-SR 1

A B

C D

A B

C D

A B

1 71 COREREACTMTY 001 MC-SR 1

C D

A B

C D

A B

C D

1 72 RADIATIONCONTROL04 001 MC-SR 1

C D

A B

C D

A B

C D

1 73 RADIATIONRXLEASE 001 MC-SR 1

D A

B C

D A

B C

D A

1 74 SRO-201-8-1-15 015 MC-SR 1

A B

C D

A B

C D

A B

I 75 VERIFYALARMS 001 MC-SR 1

C D

A B

C D

A B

C D

Friday, February 20,2004 @ 09:27 AM Answer Key Page: 1 Test Name: 04SRO.TST Test Date:

Thursday, February 19,2004 Answer@)

Question ID Type Pts 0

1 2

3 4

5 6

7 8

9 7-h 1:

1 RCPhL4LF"NCTIONS 002 MC-SR 1

C D

A B

C D

A B

C D

1:

2 LOSSOFRHR 001 MC-SR 1

A B

C D

A B

C D

A B

1.

3 PZRCONTROLMALFSRO 001 MC-SR 1

C. D A B C D A B C D 1:

4 ATWSSRO 001 MC-SR 1

A B

C D

A B

C D

A B

1:

5 LOSSOFFWSRO 003 MC-SR 1

C D

A B

C D

A B

C D

1:

6 SRO-201-7-1-03 003 MC-SR 1

D A

B C

D A

B C

D A

1:

7 CRO-54-1-1-25 001 MC-SR 1

D A

B C

D A

B C

D A

1:

8 CRO-107-1-3-28 029 MC-SR 1

B C

D A

B C

D A

B C

1:

9 CRO-202-2A-0-04 004 MC-SR 1

B C

D A

B C

D A

B C

G! 1: Di LOR-202 025 MC-SR 1

B C

D A

B C

D A

B C

1:

11 AOP-2A-03 003 MC-SR 1

B C

D A

B C

D A

B C

1:

12 SRO-201-8-1-19 019 MC-SR 1

D A

B C

D A

B C

D A

1.

13 COMPCLGSRO 001 MC-SR 1

C D

A B

C D

A B

C D

1 14 CONTAlNhENTCLGSRO 001 MC-SR 1

A B

C D

A B

C D

A B

001 MC-SR 1

C D

A B

C D

A B

C D

1:

16 CONTAINMENTSRO 003 MC-SR 1

D A

B C

D A

B C

D A

1.

17 PZRLVLCONTSRO 001 MC-SR 1

B C

D A

B C

D A

B C

1:

18 CRO-11364-20 001 MC-SR 1

B C

D A

B C

D A

B C

~

1.

19 SR0-204-1-0/3-002 002 MC-SR 1

B C

D A

B C

D A

B C

1.

20 NITE/STANDINGORDRS 001 MC-SR 1

C D

A B

C D

A B

C D

1:

21 CRO-212-1-1-02 003 MC-SR 1

C D

A B

C D

A B

C D

1 : 22 SRO RESPONSIBILITIES 001 MC-SR 1

A B

C D

A B

C D

A B

1:

23 RADWORKPERMIT 001 MC-SR 1

D A

B C

D A

B C

D A

1 :

24 EMERPRO 001 MC-SR 1

C D

A B

C D

A B

C D

1' 25 SRO-201-0-3-08 008 MC-SR 1

B C

D A

B C

D A

B C

I 1:

15 CRO-54-1-1-11

<a, I'

ML040970260

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