Text

Initial Exam 2009-301 Final SRO Written Exam
	ML100470549
Person / Time
Site:	Catawba
Issue date:	02/02/2010
From:	; NRC/RGN-II
To:	;
References
	Download: ML100470549 (113)
	v • d • e

FACILITY NAME: ______ C;.;:;;ao;;:ta;;.;;;w;;..;;;b~a____- _

_ _ _----.;C;:.::a::,:t:aw::.:..b::.:a=--_ --_-

~2=0~0~9-~3~01~----

______~2~0~0~9-~3~01~------

REPORTNUMBER: ________

FINAL SRO 2009 WRITTEN EXAM CONTENTS:

o Final SRO Written Exam (100 'as given' questions with changes made D

during administration annotated) o Reference Handouts Provided To Applicants D

o Answer Key D

Location of Electronic Files:

Not applicable.

CATAWBA NUCLEAR NU-CLEAR STATION Question: 1 2009 SRO NRC Examination (1 point)

Unit 1 is operating at 75% power and increasing power following a refueling outage.

The following occurs at 1015:

1AO-1, 1AD-1, E/8 "TURB GEN HI VIBRATION" actuates

Turbine bearing vibration readings on the OAC indicate:

o Bearing #9 at 11 mils and stable o Bearing #10 at 13 mils and stable What response is required for these conditions per OP/1/B/61001010B (Alarm Response for Panel 1AD-1)?

AO-1)?

A. Immediately trip the turbine only B. Immediately trip the reactor and verify the turbine trips automatically C. Monitor bearing vibrations and if vibration is unchanged at 1017, trip the turbine only D.

O. Monitor bearing vibrations and if vibration is unchanged at 1017, trip the reactor and verify the turbine automatically trips Page 1 of 100

CATAWBA NUCLEAR STATION Question: 2 2009 SRO NRC Examination (1 point)

Unit 1 is operating at 100% power. Given the following conditions and sequence of events:

Initial Conditions

1NV-294 (NV Pmps A&B Disch Flow Ctrl) is in manual

PRT pressure is 6 psig and increasing

PRT level is 67% and increasing

The PRT has a small external leak (0.05 gpm)

Pressurizer pressure is 2235 psig Final Conditions 10 minutes later

The crew notes pressurizer pressure is 2197 psig and slowly decreasing

PRT pressure is 15 psig and increasing

PRT level is 68% and increasing Based on the conditions stated above, and assuming no operator action:

1. How (if at all) is charging flow affected?

2. How (if at all) is the external leakage rate from the PRT affected?

A. Charging flow is stable PRT external leakage rate will increase B. Charging flow is stable PRT external leakage rate is stable C. Charging flow is increasing PRT external leakage rate is stable D. Charging flow is increasing PRT external leakage rate will increase Page 2 oflOO of tOO

CATAWBA NUCLEAR STATION Question: 3 2009 SRO NRC Examination (1 point)

Unit 1 was operating at 100% power when a LOCA occurred. Given the following conditions:

The crew is in EP/1/A/5000/E-1 (Loss of Reactor or Secondary Coolant)

NC pressure is 1200 psig and stable

Containment pressure peaked at 2.8 psig and is slowly decreasing

NC subcooling is 2°F and stable

Pressurizer level is 18% and slowly increasing

Total CA flow is 400 gpm and stable

All S/G NR levels are approximately 21% and slowly increasing Can safety injection be terminated at this time and why or why not?

A. Yes, safety injection termination criteria are met B. No, CA flow is too low C. No, NC subcooling is insufficient D. No, pressurizer level is too low Page 3 of 100

CATAWBA NUCLEAR STATION Question: 4 2009 SRO NRC Examination (1 point)

Unit 1 was operating at 100% power when a LOCA occurred. Given the following conditions and sequence of events:

The crew is performing EP/11A15000/E-1 EP/1/A15000/E-1 (Loss of Reactor or Secondary Coolant)

1A train and 1 1B B train ECCS are RESET

1A 1A DIG load sequencer is RESET

The crew can not reset 1 B DIG load sequencer

An NLO completes the local action(s) required due to the inability to reset 1 1BB DIG load sequencer from the control room

A complete loss of offsite power occurs

FWST level is 57% and decreasing Assuming no further operator action, which ECCS pumps are running 1 minute after the loss of offsite power?

A. 1A NVonly B. 1A NV and 1B NVonly C. 1A NV, 1A NI and 1A NDNO only D. 1A NV, 1B NV, 1B NI and 1B NOND Page 4 ofl00 oflOO

CATAWBA NUCLEAR STATION Question: 5 2009 SRO NRC Examination (1 point)

Unit 1 is operating in Mode 3. Given the following temperatures on the 1A NCP at 0200 hrs:

Motor bearing temperature is 187.2°F

Lower pump bearing temperature is 211.2°F

#1 seal outlet temperature is 223.2°F

Motor stator winding temperature is 301.2°F 301 .2°F

All temperatures are increasing at 1°F per minute What is the earliest time that the 1A NCP must be secured per OP/1/A/61501002A OP/1/A/6150/002A (Reactor Coolant Pump Operation)?

A. 0208 B. 0210 C. 0212 D. 0214 Page 5 ofl00 of 100

CATAWBA NUCLEAR STATION Question: 6 2009 SRO NRC Examination (1 point)

Unit 1 is in Mode 5 preparing to enter Mode 6 after being shutdown 10 days ago. Given the following conditions and sequence of events:

NC wide range level is 18% and stable

NO loop is in service 1A ND

NO Pump A Discharge Temperature is 120°F and stable ND

NO Hx A Outlet Temperature is 100°F and stable ND

NO pump trips The 1A ND

NO pump can not be started The 1B ND Based on current conditions, what is the minimum time before core boiling occurs?

Reference provided A. 31-32 minutes B. 36-37 minutes C. 47 -48 minutes D. 55-56 minutes Page 6 oflOO of 100

CATAWBA NUCLEAR STATION Question: 7 2009 SRO NRC Examination (1 point)

Unit 1 was operating at 100%. Given the following conditions and sequence of events:

S/G PORV opened momentarily and caused NC pressure to decrease to A SIG 2205 psig

With NC pressure at 2207 psig, 1AD-2, FIg Fig "DeS "DCS Alternate Action" actuates

NC pressure is at 2250 psig and slowly increasing when the crew enters AP/11A155001011 AP/1 IA/5500/o11 (Pressurizer Pressure Anomalies), Case 2, Pressurizer Pressure Increasing Why will manual action be required to control Pzr pressure?

A. All Pzr heaters are on B. To prevent lifting a Pzr PORV C. Automatic Pzr spray actuation only is blocked D. Automatic Pzr spray and automatic PORV actuation are blocked Page 7 oflOO of 100

CATAWBA NUCLEAR STATION Question: 8 2009 SRO NRC Examination (1 point)

Unit 1 is operating at 100% power. The basis for manually inserting control rods during an ATWS A TWS event is to reduce reactor power to:

A. prevent exceeding rated thermal power limits defined in Technical S pecifi catio ns B. prevent rapid heatup of the NC system and potential overfill of the pressurizer C. ensure the only heat being added to the NC system is from core decay heat and NC pump heat D. ensure there is sufficient steam dump capacity to prevent opening the steam line code safeties Page 8oflOO 8 of 100

CATAWBA NUCLEAR STATION Question: 9 2009 SRO NRC Examination (1 point)

Given the following:

1A CF pump is out of service for maintenance

1B CF pump trips on low lube oil pressure

The crew enters AP/1/A/5500/006 AP/1/A155001006 (Loss of S/G Feedwater).

Feedwater), Case I (Loss of CF Supply To S/Gs).

What is the minimum power level which requires a manual reactor trip and what is the basis for manually tripping the reactor above that power level?

A. 5%; to ensure CA can maintain adequate S/G levels B. 10%; to ensure CA can maintain adequate S/G levels C. 5%; to ensure an inadvertent mode change does not occur D. 10%; to ensure an inadvertent mode change does not occur Page 9 oflOO of 100

CATAWBA NUCLEAR STATION Question: 10 2009 SRO NRC Examination (1 point)

Regarding the actions of EP/1/A/5000/ECA 0.0 (Loss of All AC Power):

1. What is the basis for rapidly depressurizing intact S/Gs to 165 psig?

2. What is the basis for stopping the depressurization at 165 psig?

A. 1. To enhance natural circulation until forced cooling can be restored

2. To prevent injecting N2 from the CLAs B. 1. To enhance natural circulation until forced cooling can be restored

2. To prevent voiding in the reactor vessel upper head C. 1. To reduce NC temperature and pressure to minimize NC system inventory loss

2. To prevent injecting N2 from the CLAs D. 1. To reduce NC temperature and pressure to minimize NC system inventory loss

2. To prevent voiding in the reactor vessel upper head Page 10 of 100 oflOO

CATAWBA NUCLEAR STATION Question: 11 2009 SRO NRC Examination (1 point)

Unit 1 is operating at 75% power and Unit 2 is operating at 100% power when the following switchyard PCBs open:

PCB 17

PCB 18

PCB 19 Which one of the following statements correctly describes the effect on Units 1 and 2?

A. Unit 1 remains at 75% power and Unit 2 runs back to 48% power at 18% Imin B. Unit 1 runs back to 48% power at 18% 1m in and Unit 2 remains at 100% power Imin C. Unit 1 remains at 75% power and Unit 2 runs back to 23% power at 25% Imin D. Unit 1 runs back to 23% power at 25% Imin and Unit 2 remains at 100% power Page 11 of 100

CATAWBA NUCLEAR STATION Question: 12 2009 SRO NRC Examination (1 point)

Unit 1 is operating at 30% power when 1ERPA de-energizes.

Based on the above conditions, which one of the following lists contains operations would.§!!

which would all require manual action? (Evaluate each condition separately)

1. 1A DIG start on a 1 ETA blackout

2. 1A DIG FD Day Tank makeup on low level

3. Swap RN to the pond on a Train A Emergency Lo Pit Level

4. Rod withdrawal on a -1.5°F combined error signal

5. Makeup to VCT A. 1 and 2 B. 3 and 4 C. 1,3, and 5 D. 2,4, and 5 Page 12 of 100

CATAWBA NUCLEAR STATION Question: 13 2009 SRO NRC Examination (1 point)

Unit 1 NC system temperature is at 335°F and decreasing for a refueling outage.

Given the following:

Vital charger 1 ECC fails

1EMXA is unavailable to power spare charger 1 ECS 1ECS

1. What other MCC can provide an alternate supply to 1ECS?

2. Does OP/1/A/63501008 OP/1/A/6350/008 (125VDC/120VAC Vital Instrument and Control Power System) allow alignment of the alternate supply to 1 ECSAbased ECS"based on current Unit 1 conditions?

~"PPL't fl>wEe.

....0 1wPPL't

"(0 pt>wEe TO IE-be.

IE-be..

A. 1. 1EMXC

2. Yes B. 1. 1EMXC

2. No C. 1. 1EMXJ

2. Yes D. 1. 1EMXJ

2. No Page 13 of 100

CATAWBA NUCLEAR STATION Question: 14 2009 SRO NRC Examination (1 point)

Units 1 and 2 are operating at 100% power with RN pump 2A running. Given the following conditions and indications:

An Emergency Lo Pit Level actuation occurs

The crew is performing AP/0IAl55001020 AP/O/A/55001020 (Loss of Nuclear Service Water),

Case 2 (Loss of RN Pit Level)

During a board walkdown, the following are noted:

o 1RN-538 (Station RN Disch Hdr X-Over) is open o 1 RN-54A (Station RN Disch Hdr X-Over) is closed

1. Where is RN discharge flow from the DIGs directed based on current conditions?

2. Which of the above RN valves must be realigned per AP/O/A/55001020?

AP/01Al55001020?

A. 1. Standby Nuclear Service Water Pond

2. 1RN-538 B. 1. RL discharge piping to Lake Wylie

2. 1RN-538 C. 1. Standby Nuclear Service Water Pond

2. 1RN-54A D. 1. RL discharge piping to Lake Wylie

2. 1RN-54A Page 14 of 100

CATAWBA NUCLEAR STATION Question: 15 2009 SRO NRC Examination (1 point)

Unit 1 is in Mode 3. Given the following initial valves and positionslfunctions:

positions/functions:

VALVE POSITION/FUNCTION POS IT 10 N/FUNCTIO N 1 NV-1A (NC Letdn to Regen Hx Isol) I 1 Open 1 CA-149 (S/G 1A CF Byp To CA Nozzle) Open 1NV-153A (Letdn Hx Otlt 3-Way Vlv) "DEMIN" 1 RN-291 (1A KC HX Outlet Throttle) "MINIFLOW" The Instrument Air (VI) system completely depressurizes.

When the VI system is re-pressurized, which one of the following valves will return to its initial position/function with no operator action?

A. 1NV-1A B. 1CA-149 C. 1NV-153A D. 1RN-291 Page 15 of 100

CATAWBA NUCLEAR STATION Question: 16 2009 SRO NRC Examination (1 point)

Which one of the following indications is used in EP/1 EP/1/A/5000/ECA-1.2 IA/5000/ECA-1.2 (LOCA Outside Containment) to verify that a leak outside containment has been successfully isolated?

A. NO Pump discharge pressure increasing B. Pressurizer level increasing C. NO/NS room sump levels decreasing O. Reactor coolant pressure increasing Page 16 of 100

CATAWBA NUCLEAR STATION Question: 17 2009 SRO NRC Examination (1 point)

Unit 1 is operating at 100%. Given the following conditions and sequence of events:

Initial Conditions

The crew tripped the reactor due to a loss of all feedwater

EP/11A1S0001E-O (Reactor Trip or Safety Injection) was completed EP/1/A/5000/E-O

The crew was depressurizing the S/Gs to feed from the condensate booster pumps per EP/11A1S000/FR-H.1 EP/1/A/5000/FR-H.1 (Response to Loss of Secondary Heat Sink)

Current Conditions

Bleed and feed criteria have been met and the crew is establishing bleed and feed per EP/11A1S000/FR-H.1 EP/1/A/5000/FR-H.1

1. Which cold leg accumulators provide backup motive force to open the PZR PORVs?

2. What is the function of the NC PORVs when establishing bleed and feed in EP/11A1S000/FR-H.1 ?

EP/1/A/5000/FR-H.1 A. 1. A and B

2. To provide for adequate NC system heat removal until secondary heat sink can be restored B. 1. A and B

2. To depressurize the NC system to protect the S/G tubes from creep failure C. 1. C and D

2. To provide for adequate NC system heat removal until secondary heat sink can be restored D. 1. C and D

2. To depressurize the NC system to protect the S/G tubes from creep failure Page 17 of 100

CATAWBA NUCLEAR STATION Question: 18 2009 SRO NRC Examination (1 point)

Unit 1 was operating at 100% power. Given the following sequence of events:

0300 Reactor trip occurred due to a general warning 0320 Safety injection actuated due to a large break LOCA 0322 1B NI pump trips due to a faulty breaker 0350 Loss of emergency coolant recirculation occurred and the crew entered EP/1/A/5000/ECA-1.1 (Loss of Emergency Coolant Recirculation) 0351 NV and NI pump status are determined as follows:

1A1A NV pump flow is 180 gpm and stable

1 B NV pump flow is 170 gpm and stable

1A NI pump flow is 340 gpm and stable

1 B NI pump is tagged out 0420 The crew is performing step 19b RNO of EP/11A15000IECA-1.1 to establish the appropriate safety injection flow per Enclosure 5 When the actions required per Step 19b RNO are complete, which ECCS pump(s) are in service?

Reference provided A. 1A NI pump B. 1A NV pump and 1 1A B NV pump 1B C. 1A NV pump and 1A NI pump 1A D. 1 B NV pump and 1A NI pump Page 18 of 100

CATAWBA NUCLEAR STATION Question: 19 2009 SRO NRC Examination (1 point)

Unit 1 is operating at 100% power. Given the following:

While performing the RCCA movement test, control rod H-8 slips into the core to 200 steps withdrawn

All other Bank D control rods are at 216 steps withdrawn as indicated on DRPI and step demand counters

perfonning AP/1/A155001014 The crew is performing AP/1/A/55001014 (Control Rod Misalignment) and currently referring to OP/1/A161501008 OP/1/A/6150/o08 (Rod Control)

1. What is the maximum time allowed to restore rod H-8 to within limits per Technical Specification 3.1.4 (Rod Group Alignment Limits)?

2. Which control rod lift coil(s) will be disconnected per OP/1/A161501008 OP/1/A/61501008 (Rod Control), when initially attempting to recover control rod H-8?

A. 1. 30 minutes

2. Rod H-8 B. 1. 30 minutes

2. All rods in the affected bank except H-8 C. 1. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months

2. Rod H-8 D. 1. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months

2. All rods in the affected bank except H-8 Page 19 of 100

CATAWBA NUCLEAR STATION Question: 20 2009 SRO NRC Examination (1 point)

Unit 1 is increasing power vvhen when the following OAC indications are noted:

OAC reading Point Color N 3 .12E-5 amps GREEN N 2.75E-6 amps GREEN N-41 -11.3% BLUE N-42 -11.2% GREEN N-43 --0.0%

0.0% MAGENTA N-44 -1 0.8%

-10.8% GREEN

1. Which intermediate range instrument is reading correctly?

2. What does the BLUE color of the OAC point for N-41 indicate?

A. 1. N-35

2. The OAC point quality is "suspect" B. 1. N-35

2. The OAC point quality is "bad" C. 1. N-36

2. The OAC point quality is "suspect" D. 1. N-36

2. The OAC point quality is "bad" Page 20 of 100

CATAWBA NUCLEAR STATION Question: 21 2009 SRO NRC Examination (1 point)

Unit 1 was operating at 50% power. Given the following:

A steam generator tube leak occurred on 1A S/G AP/1 IAl550010 10 (Reactor Coolant

The crew depressurized the NC system per AP/1/A155001010 Leak), Case I (Steam Generator Tube Leak)

Both ICC monitors are unavailable

Current NC pressure is 665 psig

Core exit thermocouple temperatures are 488°F

T-Colds are 487.rF

1. Based on current conditions, what subcooling value would the ICC monitors be reading if they were available?

2. Based on current conditions, what is steam header pressure?

Reference provided A. 1. -8 °0 F

2. 608 psig B. 1. -8 of

2. 593 psig C. +120 F

1. +12°F

2. 608 psig D. +120 F

1. +12°F

2. 593 psig Page 21 of 100

CATAWBA NUCLEAR STATION Question: 22 2009 SRO NRC Examination (1 point)

The Spent Fuel Pool Ventilation (VF) system is operating in unfiltered mode when a recycle paper bin near the Spent Fuel Pool truck bay doors catches fire and smoke is drawn into the VF system supply ductwork.

1. What affect will this condition have on the operation of the VF system?

2. How is this fire classified?

A. 1. The VF Supply fan trips only

2. Class A B. 1. The VF Supply fan trips only

2. Class B C. 1. The VF Supply fan trips and the VF Filtered Exhaust fans will swap to filter mode Class A

2. ClassA D. 1. The VF Supply fan trips and the VF Filtered Exhaust fans will swap to filter mode

2. Class B Page 22 of 100

CATAWBA NUCLEAR STATION Question: 23 2009 SRO NRC Examination (1 point)

Unit 1 is operating at 100% power. Given the following:

A small break LOCA occurs

EP/1/A15000/ES-1 .2 (Post LOCA Cooldown and The crew is performing EP/1/A/5000/ES-1.2 Depress Dep ress urization)

All NC pumps are secured

1. What is the NC cool cooldown down rate specified in this procedure?

2. Which method is the first choice for the NC system depressurization for the given conditions?

A. 1. As close as possible without exceeding 100°F in an hour

2. NV auxiliary spray B. 1. At the maximum rate

2. NV auxiliary spray C. 1. As close as possible without exceeding 100°F in an hour

2. NC PORV D. 1. At the maximum rate

2. NC PORV Page 23 of 100

CATAWBA NUCLEAR STATION Question: 24 2009 SRO NRC Examination (1 point)

Unit 1 was operating at 100% power. Given the following:

A large steam line break occurred 20 minutes ago

All MSIVs fail to close by any means

NC cold leg temperatures are 248°F and slowly decreasing

NC pressure is 520 psig and slowly increasing

The crew has completed EP/1/A15000/E-0 EP/1/A/5000/E-0 (Reactor Trip or Safety Injection)

1. Based on current conditions, which procedure will the crew transition to when EP/1/A15000/E-0 EP/1/A/5000/E-0 is exited?

2. At what point during this steam line break accident is the reactor vessel most susceptible to failure and why?

A. 1. EP/1/A/5000/FR-P.1 (Response to Imminent Pressurized Thermal Shock)

2. During the initial cooldown because total stress on the inner wall of the vessel is higher at this time B. EP/1 IA/5000/E-2 (Faulted Steam Generator Isolation)

1. EP/1/A/5000/E-2

2. During the initial cooldown because total stress on the inner wall of the vessel is higher at this time C. 1. EP/1/A/5000/FR-P.1 (Response to Imminent Pressurized Thermal Shock)

2. During the subsequent heatup after the S/Gs depressurize because total stress on the outer wall of the vessel is higher at this time D. 1. EP/1/A/5000/E-2 (Faulted Steam Generator Isolation)

2. During the subsequent heatup after the S/Gs depressurize because total stress on the outer wall of the vessel is higher at this time Page 24 of 100

CATAWBA NUCLEAR STATION Question: 25 2009 SRO NRC Examination (1 point)

Given the following:

1A S/G pressure is 1212 psig and increasing

1A S/G level is approximately 18% N/R level and stable

The crew is performing EP/1/A/5000/FR-HA (Response to Loss of Normal Steam Release Capabilities)

Operators have been dispatched to manually operate the 1A S/G PORV

1. Howwill1A S/G indicated level initially respond when the PORV is opened?

2. What is the reason for this response?

A. 1. Indicated level will initially decrease

2. Increased voiding in the S/G downcomer region B. 1. Indicated level will initially increase

2. Increased voiding in the S/G downcomer region C. 1. Indicated level will initially decrease

2. Increased voiding in the S/G tube bundle area D. 1. Indicated level will initially increase

2. Increased voiding in the S/G tube bundle area Page 25 of 100

CATAWBA NUCLEAR STATION Question: 26 2009 SRO NRC Examination (1 point)

Unit 1 is operating at 100% power. Given the following:

A main steam line ruptures in containment

NC Tcold temperatures are 278°F and slowly decreasing

Containment pressure is 3.2 psig and increasing

No NS pump flow is indicated

All S/G NR levels are offscale low

CA flow is 640 gpm

EP/1/A/SOOO/E-0 (Reactor Trip or Safety Injection)

The crew is preparing to exit EP/1/A/5000/E-0 Which one of the following procedures is the crew required to enter next?

A. EP/11A1S000/E-2 (Faulted Steam Generator Isolation)

EP/1/A/5000/E-2 B. EP/1/A/5000/FR-H.1 (Response to Loss of Secondary Heat Sink)

EP/11A1S000/FR-H.1 C. EP/1/A/5000/FR-P.2 (Response to Anticipated Pressurized EP/11A1S000/FR-P.2 Thermal Shock Condition)

D. EP/11A1S000/FR-Z.1 (Response to High Containment Pressure)

EP/1/A/5000/FR-Z.1 Page 26 of 100

CATAWBA NUCLEAR STATION Question: 27 2009 SRO NRC Examination (1 point)

Unit 1 is operating at 100% power. Given the following:

At 1100 a large break LOCA occurs

At 1215, when the crew is in EP/1/A/5000/E-1 (Loss of Reactor or Secondary Coolant), waiting for the time to swap to hot leg recirc, containment sump level is 13 feet and slowly increasing

1. If containment sump level is increasing at a constant rate of 0.25 feet per minute, at what time is entry into EP/1/A/5000/FR-Z.2 (Response to Containment Flooding) required?

2. What signal closes valves inside containment that are not qualified for submergence?

A. 1. 1225

2. St B. 1. 1245

2. St C. 1. 1225

2. Sp D. 1. 1245

2. Sp Page 27 of 100

CATAWBA NUCLEAR STATION Question: 28 2009 SRO NRC Examination (1 point)

Regarding the reactor coolant pump motor stator:

1. What system provides cooling water?

2. What signal will isolate that cooling water source?

A. 1. KG

2. Phase A B. 1. KG

2. Phase B G. 1. YV

2. PhaseA Phase A D. 1. YV

2. Phase B Page 28 of 100

CATAWBA NUCLEAR STATION Question: 29 2009 SRO NRC Examination (1 point)

Unit 1 was decreasing power from 100%. Given the following:

Pressurizer level was 53%

Reactor power was 85%

The power decrease was halted when the crew noted that 1NV-294 (NV Pumps A & B Disch Flow Ctrl) was not maintaining pressurizer level at program level

1NV-294 was placed in manual and adjusted to restore pressurizer level to program level

1. When the crew discovered 1NV-294 was not maintaining program level, how did actual level compare to program level?

2. When 1NV-294 is repaired, what operator action(s) will be required to restore automatic pressurizer level control?

A. 1. Actual level was lower than program level

2. Place 1NV-294 in automatic only B. 1. Actual level was lower than program level

2. Place 1NV-294 in automatic and restore the Pressurizer Level Master to automatic using the DCS soft controls C. 1. Actual level was higher than program level

2. Place 1NV-294 in automatic only D. 1. Actual level was higher than program level

2. Place 1NV-294 in automatic and restore the Pressurizer Level Master to automatic using the DCS soft controls Page 29 of 100

CATAWBA NUCLEAR STATION Question: 30 2009 SRO NRC Examination (1 point)

Unit 1 is operating at 100% power.

For current conditions, what is the normal position of 1N0-33 1NO-33 (NO Sys Rtn To FWST) and how is this position administratively controlled?

A. Closed; white tagged in position B. Closed; locked in position C. Open; white tagged in position O. Open; locked in position Page 30 of 100

CATAWBA NUCLEAR STATION Question: 31 2009 SRO NRC Examination (1 point)

Unit 1 is operating at 100% power when a large break LOCA occurs.

The crew has just transitioned to EP/1 /A/5000/ES-1 .3 (Transfer to Cold Leg EP/1/A/5000/ES-1.3 Recirculation) and no manual actions have been taken

1. Based on the above conditions, which ECCS pump(s) are currently taking suction from the containment sump?

2. If NO is required to be aligned for hot km kill recirculation, to which NC hot legs will NO inject?

A. 1. NO pumps only

2. Band C NC hot legs only B. 1. NO pumps only

2. All 4 hot legs C. 1. NV pumps, NI pumps, NO pumps

2. Band C NC hot legs only O. 1. NV pumps, NI pumps, NO pumps

2. All 4hotlegs 4 hot legs Page 31 of 100

CATAWBA NUCLEAR STATION Question: 32 2009 SRO NRC Examination (1 point)

Unit 1 was operating at 100% power when it experienced a steam line break accident resulting in a safety injection and the crew has entered EP/1/A/5000/E-O EP/11A15000/E-O (Reactor Trip or Safety Injection).

Which one of the following completes the statement below?

Per Enclosure 1 of EP/1/A/5000/E-0, when NC pressure decreases to less than

_ _ _ the crew will close 1NV-202B and 1 NV-203A (NV Pumps A&B Recirc lsol) and when NC pressure increases to greater than the crew will re-open these valves.

Closed Re-opened A. 1500 psig 2000 psig B. 1500 psig 1955 psig C. 1620 psig 2000 psig D. 1620 psig 1955 psig Page 32 of 100

CATAWBA NUCLEAR STATION Question: 33 2009 SRO NRC Examination (1 point)

Unit 1 was operating at 100% power when a LOCA occurred at Time O. Given the NC System pressure trend shown below:

2500 2000

...::ru

r

...u D-A.

1500 E

u t: 1000 1:

g; ii u

z 500 o

o 1 2 3 4 5 6 7 8 .9 9 10 11 Time (minutes) lime Which of the following describes when each of the ECCS pumps listed begins to inject flow into the NC system?

NI Pumps ND Pumps A. Between 0 and 1 minute Between 8 and 9 minutes B. Between 1 and 2 minutes Between 8 and 9 minutes C. Between 0 and 1 minute Between 9 and 10 minutes D. Between 1 and 2 minutes Between 9 and 10 minutes Page 33 of ortOO 100

CATAWBA NUCLEAR STATION Question: 34 2009 SRO NRC Examination (1 point)

Unit 1 completed NC fill and vent activities following a forced outage. Given the following:

A nitrogen blanket was placed on the pressurizer during the outage

NC level is 85% and stable

The crew is in the process of drawing a bubble in the pressurizer

1. Per Selected Licensee Commitment 16.5-4 (Pressurizer), what is the limit on pressurizer heat up rate?

2. Per OP/1/A/61 OP/11A161 001001 (Controlling Procedure for Unit Startup), when a pressurizer PORV is opened, what indication verifies nitrogen venting is complete?

A. 1. 100°F in anyone hour period

2. PRT temperature equalizes with Pzr steam space temperature B. 1. 100°F in anyone hour period

2. PRT level increases without a corresponding PRT pressure increase C. 1. 80°F in anyone hour period

2. PRT temperature equalizes with Pzr steam space temperature D. 1. 80°F in anyone hour period

2. PRT level increases without a corresponding PRT pressure increase Page 34 of 100

CATAWBA NUCLEAR STATION Question: 35 2009 SRO NRC Examination (1 point)

Unit 1 is in Mode 3. Given the following:

1AD-6, E/1 (NGP (NCP A Thermal Barrier KC KG Outlet Hi/lo Flow) actuates

The flow trend below:

100 T--~~--'---T--'~--**-'-~~-

90 .+-~....;.---+

80 +-~--+--

70 +--+------

60 .

.~ E

.S!

50 50*

t: a..

0.. .

~co ~ ~ 40 40 . ll.....;;;;;;;;;/.;;;;;;;;;;;;;;;

CQ - i""""

co 30 .

E OJ 20

.c

!-- 10*

10 u

.~

o o 10 20 30 40 50 60 70 80 90 100 110 120 130 140 150 Time (seconds)

1. At what time on the above graph should 1 KG-394A KC-394A (NC(NG Pump 1A Therm Bar Otlt) have automatically closed?

2. If the NGP NCP 1A thermal barrier can not be isolated from the KC KG System by any means, how is over pressurization of the KC KG surge tanks prevented?

A. 1. 100 seconds

2. The KC KG surge tanks vent line is large enough to prevent over pressurization B. 1. 130 seconds

2. The KC KG surge tanks vent line is large enough to prevent over pressurization C.

G. 1. 100 seconds

2. The KC KG surge tanks relief valve to the KC KG drain sump is large enough to prevent over pressurization D. 1. 130 seconds

2. The KC KG surge tanks relief valve to the KC KG drain sump is large enough to prevent over pressurization Page 35 of 100

CATAWBA NUCLEAR STATION Question: 36 2009 SRO NRC Examination (1 point)

Unit 1 was operating at 100%. Given the following conditions and sequence of events:

A slight cooldown of the NC system causes the "C" PZR heaters to be full "on"

A single channel of pressurizer pressure failed offscale high

A second channel of pressurizer pressure failed offscale low Assuming no operator actions, what affect does this combination of channel failures have on the Pressurizer Pressure Control system?

A. The Pressurizer Pressure Master swaps to manual and pressure will increase to the reactor trip setpoint B. The Pressurizer Pressure Master swaps to manual and pressure will increase until 1 NC-32B and 1 NC-36B open C. The Pressurizer Pressure Master remains in automatic and pressure is controlled based on the average of the 2 unaffected channels D. The Pressurizer Pressure Master remains in automatic and pressure is controlled based on the highest of the 2 unaffected channels Page 36 of 100

CATAWBA NUCLEAR STATION Question: 37 2009 SRO NRC Examination (1 point)

Unit 1 is increasing power following a refueling outage. Given the following:

At 45% reactor power, the P-8 permissive failed "as-is"

Reactor power is currently at 50%

Based on the current conditions:

1. What is the condition of the P-8 "Hi Pwr Lo Flo Rx Trip Block" status light on 181-18?

2. If one NC pump were to trip, would an automatic reactor trip occur?

A. 1. Lit

2. Yes B. 1. Lit

2. No C. 1. Dark

2. Yes D. 1. Dark

2. No Page 37 of 100

CATAWBA NUCLEAR STATION Question: 38 2009 SRO NRC Examination (1 point)

Unit 1 is in Mode 3 and reducing NC pressure and temperature for a refueling outage.

Given the following:

Pressurizer pressure channel 2 fails "as-is" at 1985 psig

follows:

Other pressurizer pressure channels indicate as follo\t'\lS:

o Channel 1 is 1972 psig and slowly decreasing o Channel 3 is 1965 psig and slowly decreasing o Channel 4 is 1951 psig and slowly decreasing

No operator actions have been taken Of the remaining channels that input into the P-11 circuit, how many of these must be below the P-11 setpoint before "ECCS TRN AlB PZR PRESS" can be blocked?

A. 1 of the other 2 B. 2 of the other 2 C. 1 of the other 3 D. 2 of the other 3 Page 38 of 100

CATAWBA NUCLEAR STATION Question: 39 2009 SRO NRC Examination (1 point)

Unit 1 was operating at 100% power with lower containment temperature stable at 1OrF. Given the following events:

10rF.

1A,1 B, and 1 1A, 18, 1C C Lower Containment Ventilation Units (LCVU's) are in service

LCVU 18 1B fan trips on over-current Based on the conditions above, what is the first action OP/1/A164501001 OP/1/A/6450/001 (Containment (V V) Systems) specifies to provide additional cooling to lower containment?

Ventilation (VV)

A. Start 1D LCVU in LOW speed B. Place 1A and 1C LCVUs in "MAX" cooling only C. Place 1A and 1C LCVUs in "MAX" cooling and HIGH speed D. Start 1D LCVU in HIGH speed and place 1A and 1 C LCVUs in HIGH speed Page 39 of 100

CATAWBA NUCLEAR STATION Question: 40 2009 SRO NRC Examination (1 point)

Unit 1 is operating at 100% power. Given the following:

Initial Conditions

Lower Containment Vent Unit (LCVU) status and outlet temperatures per the OAC are:

LCVU Status Outlet Temperature readings 1A Running 1 -108.1°F - 107.5°F 2 -1 07.5°F 1B Running 1 - 109.9°F 2 --1108.9°F 08.9°F 1C Running 1 -112.4°F 1-112.4°F 2 --111.2°F 111.2°F 1D 10 Secured 1 - 105.6°F 2 -1

-104.4°F 04.4°F Current Conditions

The crew notes that 1B LCVU has justtripped just tripped

OAC temperature readings as stated above have not begun to change yet Based on current plant conditions, and per PT/1/A/46001002A PT/1/Al4600/002A (Mode 1 Periodic Surveillance Items):

1. What temperature would be recorded for "Primary Containment Lower Compartment Average Air Temp"?

2. What are the Technical Specification temperature range limits for lower containment temperature?

A. 1. 108.5°F

2. 60°F -120°F

- 120°F B. 1. 109.8°F

2. 60°F -120°F C. 1. 108.5°F

2. 100°F -120°F o.

D. 1. 109.8°F

2. 100°F -120°F Page 40 of 100

CATAWBA NUCLEAR STATION Question: 41 2009 SRO NRC Examination (1 point)

Which one of the following lists of conditions would all result in 1NF-228A (NF Supply Containment Isolation Valve) closing? (Evaluate each condition separately)

1. Manual Phase A

2. Manual Phase B

3. Loss of instrument air

4. Glycol Expansion Tank Lo Lo Level

5. Trip of all running Ice Condenser Glycol Pumps A. 1 and 5 B. 2 and 5 C. 1,3 and 4 D. 2,3 and 4 Page 41 of 100

CATAWBA NUCLEAR STATION Question: 42 2009 SRO NRC Examination (1 point)

One hour ago, Unit 1 was operating at 100% power when a LOCA occurred requiring a reactor trip and safety injection. Given the following:

NO The automatic and manual swap to cold leg recirculation failed and both ND pumps have been secured

The crew implemented EP/1/A/5000/ECA-1.1 (Loss of Emergency Coolant Recirculation)

Current conditions are:

o Containment pressure is 11 .2 psig and slowly increasing o FWST level is 4.9% and decreasing For the current conditions, how many of each type of pump listed is required to be running?

NV NS A. 1 1 B. 1 o C. o 1 D.

O. o o Page 42 of 100

CATAWBA NUCLEAR STATION Question: 43 2009 SRO NRC Examination (1 point)

Unit 1 was operating at 100% power when a 6 gpm tube leak occurred in 1A S/G.

Given the following:

The crew is operating the turbine controls in manual to quickly reduce load per AP/11A155001009 AP/1/A155001009 (Rapid Downpower)

Current power level is 30% and decreasing

1. Assuming a stable tube leakage rate, how does 1EMF-71 1EMF-71 (S/G A Leakage) count rate for current conditions compare to the 100% power readings?

2. Per NSD 513 (Primary to Secondary Leak Monitoring Program), which method(s) is/are islare used to monitor S/G leak rate based on the current power level?

A. 1. Lower counts now than when power was at 100%

1EMF-33 (Condenser Air Ejector Exhaust) only

2. 1EMF-33 B. 1. The same counts now as when power was at 100%

2. 1EMF-33 1EMF-33 (Condenser Air Ejector Exhaust) only C. 1. Lower counts now than when power was at 100%

1EMF-33 (Condenser Air Ejector Exhaust) and 1 EMF-71

2. 1EMF-33 D. 1. The same counts now as when power was at 100%

2. 1EMF-33 (Condenser Air Ejector Exhaust) and 1 EMF-71 Page 43 of 100

CATAWBA NUCLEAR STATION Question: 44 2009 SRO NRC Examination (1 point)

Unit 1 is in Mode 3 preparing for reactor startup. Given the following 1SV-19 (S/G 1A PORV) nitrogen bottle pressures:

Bottle #1 - 1970 psig

Bottle #2 - 2165 psig required for 1SV-19 per Selected Licensee Commitment What action (if any) is reguired 16.10-1 (Steam Ventto Vent to Atmosphere)?

A. No action is required B. Immediately take action to restore the nitrogen supply to operable status only C. Immediately enter the applicable conditions of Technical Specification 3.7.4 (S/G PORVs) only D. Immediately take action to restore the nitrogen supply to operable status and immediately enter the applicable conditions of Technical Specification 3.7.4 (S/G PORVs)

Page 44 of 100

CATAWBA NUCLEAR STATION Question: 45 2009 SRO NRC Examination (1 point)

Unit 1 is operating at 100% power. Given the following conditions:

Initial Conditions

1CF-37 (8/G (S/G 1 B CF Ctrl) is in manual and throttled

1CF-39 (S/G 1 B CF Byp Ctrl) is in manual and full open 1CF-39 (8/G 8/G 1B Level is 66% and stable

S/G Current Conditions

1 1CF-37 CF-37 has just been placed in automatic Based on current conditions:

How will 1CF-37

1. Howwill 1CF-37 respond?

2. What is the status of 1B S/G 8/G level control?

A. 1. Throttle further open

2. Automatic B.

B. 1. Throttle further open

2. Manual C. 1. Throttle further closed

2. Automatic D. 1. Throttle further closed

2. Manual Page 45 of 100

CATAWBA NUCLEAR STATION Question: 46 2009 SRO NRC Examination (1 point)

Initial conditions:

Unit 1 was operating at 100% power

A loss of offsite power occurred

The reaetor reactor tripped 30 minutes later the following sequence of events occurred:

All CA pumps are running

Train "A" CA is RESET

Train "B" "8" CA is not RESET

The CA common suction piping from condensate grade sources ruptures

1AD-5, E/1 "CA Pumps Train A Loss of Norm Suct" Suet" actuates

1AD-5, E/2 "CA Pumps Train 8 B Loss of Norm Suct" actuates One minute later, what is the status of the CA system assuming no operator action?

A. 1A CA pump is tripped CAPT #1 is tripped 1 B CA pump suction is shifted to the RN system 18 B.

B. 1A CA pump suction is shifted to the RN system 1A CAPT #1 suction is shifted to the RN system 1 B CA pump is tripped 18 C. 1A CA pump suction is shifted to the RN system 1A CAPT #1 is tripped 1 B CA pump is tripped 18 D. 1A 1A CA pump is tripped CAPT #1 suction is shifted to the RN system 1 B CA pump suction is shifted to the RN system 18 Page 46 of 100

CATAWBA NUCLEAR STATION Question: 47 2009 SRO NRC Examination (1 point)

Given the following:

The incoming breaker to 1ETB spuriously opened

11B B DIG automatically started and loaded

While monitoring DIG operating parameters, the crew noted that DIG 1B "VOLTS" was 4000 V

Voltage was adjusted to normal How did 1 Howdid 1BB DIG output "AMPS" and "P/F" indications respond to this adjustment?

AMPS P/F PIF A. increase more lagging B. increase same stay the sam e c.

C. decrease more lagging D. decrease stay the sam e Page 47 of 100

CATAWBA NUCLEAR STATION Question: 48 2009 SRO NRC Examination (1 point)

Unit 1 is operating at 100% power.

povver. Given the following:

A complete loss of offsite power occurs

All DIG load sequencing is complete Which one of the following Fire Protection (RF/RY) system pumps does not have an available power supply?

A. RYpump A RYpumpA B. RF Jockey Pump A c.

C. RY Pump C D. RF Jockey Pump C Page 48 of 100

CATAWBA NUCLEAR STATION Question: 49 2009 SRO NRC Examination (1 point)

Which one of the following sets of equipment receives power from the 250VDC Auxiliary Power System?

A. CF Pump Control Power and Turbine Emergency Bearing Oil Pump B. Deadlight Panels and Turbine Emergency Bearing Oil Pump C. CF Pump Control Power and 6.9 KV Switchgear Control Power D. Deadlight Panels and 6.9 KV Switchgear Control Power Page 49 of 100

CATAWBA NUCLEAR STATION Question: 50 2009 SRO NRC Examination (1 point)

Unit 1 is operating at 100% power. Given the following conditions and sequence of events:

Initial Conditions

A blackout occurs on 1ETA

All 1A D/G AII1A DIG sequencing is complete

1AD-11, A/B AlB "125VDC Diesel Gen A Control Pwr Sys Trbl" actuates

1AD-11, B/7 "DIG "D/G 1A Panel Trouble" actuates

1A D/G DIG Panel alarm E/5 "Loss of DC Control Power" actuates Final Conditions

Normal power is restored to 1ETA

1A D/G DIG is secured

The annunciators listed above are still lit Based on the final conditions:

1. Which of the following methods could have been used to secure the 1A D/G DIG locally?

2. Will1A Will 1A D/G DIG start on a subsequent 1ETA blackout?

A. 1. The RUNISTOP RUN/STOP knob only

2. Yes B. 1. The RUNISTOP RUN/STOP knob only

2. No C. 1. The RUNISTOP RUN/STOP knob or the STOP pushbutton

2. Yes D. 1. The RUNISTOP RUN/STOP knob or the STOP pushbutton

2. No Page 50 of 100

CATAWBA NUCLEAR STATION Question: 51 2009 SRO NRC Examination (1 point)

Unit 1 is in Mode 3 preparing for a reactor startup. Given the following:

1ETA experiences a blackout

The 1A DIG starts and is sequencing on loads per the accelerated sequence

Half way through the accelerated sequence, the fuel racks stick in an open position causing DIG speed to rapidly increase

1. What is the setpoint for the DIG overspeed trip?

2. For the conditions stated above will the DIG trip on overspeed if the setpoint is exceeded?

A. 1. 495 rpm

2. No B. 1. 495 rpm

2. Yes c.

C. 1. 517.5 rpm

2. No D.

D. 1. 517.5 rpm

2. Yes Page 51 of 100

CATAWBA NUCLEAR STATION Question: 52 2009 SRO NRC Examination (1 point)

Unit 1 was operating at 50% power when a power spike causes a 1EMF-33 (CSAE EXHAUST) Trip 2 alarm.

What manual actions (if any) does OP/1/B/61 00101 OX (Annunciator Response to Radiation Monitoring Panel 1RAD-1) direct the operator to take to place systems affected by this EMF alarm in their post alarm condition?

A. Swap "Unit 1 CSAE EXH" discharge from "UNIT VENT" to "AUTO" B. Close 1 BB-48 (BB Pumps Disch to TB Sump)

C. Close S/G Sample valves D. None Page 52 of 100

CATAWBA NUCLEAR STATION Question: 53 2009 SRO NRC Examination (1 point)

Both units are operating at 100% power. Given the following:

Only 1A 1A RN pump in service

An 86N relay actuates on 1ETB Based on the above conditions, which RN pumps are operating one minute later?

A. 1A only B. 1A and 1B only c.

C. 1A and 2B only D. 1A, 1A, 1B, 1B, and 2B only Page 53 of 100

CATAWBA NUCLEAR STATION Question: 54 2009 SRO NRC Examination (1 point)

Unit 1 automatically trips from 100% power and the crew entered EP/1/A/5000/E-O (Reactor Trip or Safety Injection). Given the following conditions and sequence of events:

Both trains of CA SYS VL V RESET are "RESET"

CA flow is throttled to maintain proper S/G levels

A complete loss of VI pressure occurs How is manual control of the CA flow control valves and CA flow affected by the loss of VI pressure?

A. Control of the CA flow control valves is immediately lost and CA flow will immediately decrease B. Control of the CA flow control valves is immediately im mediately lost and CA flow will immediately increase C. Control of the CA flow control valves is maintained for at least 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and then CA flow will decrease D. Control of the CA flow control valves is maintained for at least 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and then CA flow will increase Page 54 of 100

CATAWBA NUCLEAR STATION Question: 55 2009 SRO NRC Examination (1 point)

Unit 1 was operating at 100% power. At 1000, a secondary steam leak occurred in the Unit 1 containment. The crew has started a controlled shutdown using AP/1/A/55001009 AP/1 IA/55001009 (Rapid Downpower). Containment pressure and temperature trends indicate:

Time Temperatu re Temperature Pressure 1000 113°F 0.15 psig 1005 11rF 0.28 psig 1010 119°F 0.42 psig 1015 122°F 0.48 psig 1020 126°F 0.65 psig 1025 129°F 0.91 psig No manual operator actions have been taken related to LCVU operation.

During which time period did the Lower Containment Ventilation Units (LCVUs) cooling water bypass valves (full flow valves) automatically open?

A. Between 1005 and 1010 B. Between 1010 and 1015 C. Between 1015 and 1020 D. Between 1020 and 1025 Page 55 of 100

CATAWBA NUCLEAR STATION Question: 56 2009 SRO NRC Examination (1 point)

Unit 1 was in Mode 3 at 55rF and 2235 psig with shutdown banks withdrawn in preparation for startup. Given the following conditions and sequence of events:

1TXS is red tagged

A fault results in a loss of all loads on the short side of 1TC se~s) has/have a power supply available and what procedure is entered to Which MG set(s) address the above conditions?

A. Only 18 MG set; EP/1/A/5000/E-O (Reactor Trip or Safety Injection)

B. Only 18 MG set; AP/1/A/5500/004 (Loss of Reactor Coolant Pump)

C. 1A and 18 MG sets; EP 1A /1/A/5000/E-O (Reactor Trip or Safety Injection)

EP/1/A/5000/E-O D. 1A and 18 MG sets; AP/1/A/5500/004 (Loss of Reactor Coolant Pump)

Page 56 of 100

CATAWBA NUCLEAR STATION Question: 57 2009 SRO NRC Examination (1 point)

Unit 1 is in a mid loop condition conducting a vacuum refill of the NC system at the end of a refueling outage.

Which one of the following reactor vessel level instruments will provide the greatest accuracy for this condition?

A. NC level loop indication B. Sight glass loop indication C. Ultrasonic level indication D. RVLlS lower range indication Page 57 of 100

CATAWBA NUCLEAR STATION Question: 58 2009 SRO NRC Examination (1 point)

Unit 1 is operating at 100%. Given the following indications:

Pressurizer level is 61% and increasing

Spray valves have throttled open and pressurizer pressure is slowly decreasing towards normal pressure

Charging flow initially decreased and has stabilized at a lower flow rate

Letdown flow is 0 gpm

All pressurizer heaters are off

1. Which one of the following failures would account for all these indications?

2. What procedure will be used to address the results of this failure?

A. 1. Selected Pressurizer Level (SPL-1) "PZR LVL LOW" output signal is present

2. AP/1/A155001012 (Loss of Charging or Letdown)

B. 1. Channel 1 pressurizer level/pressure common reference leg is leaking

2. AP/1/A/55001012 (Loss of Charging or Letdown)

C. 1. Selected Pressurizer Level (SPL-2) "PZR LVL LOW" output signal is present

2. AP/1/A/55001011 (Pressurizer Pressure Anomalies)

D. 1. Channel 1 pressurizer level/pressure common reference leg is leaking

2. AP/1/A155001011 (Pressurizer Pressure Anomalies)

Page S8 of 100

CATAWBA NUCLEAR STATION Question: 59 2009 SRO NRC Examination (1 point)

Unit 1 is operating at 100% power. Assuming intermediate range N-35 channel compensation voltage is lost and a reactor trip occurs:

1. How will N-35 indication compare to N-36 when actual reactor power has decreased into the source range?

2. What is the minimum number of intermediate range instruments required to be below the P6 setpoint to automatically re-energize the source range instruments N-31 and N-32?

A. 1. N-35 will read higher than N-36

2. One B. 1. N-35 will read higher than N-36

2. Two c.

C. 1. N-35 will read lower than N-36

2. One D. 1. N-35 will read lower than N-36

2. Two Page 59 of 100

CATAWBA NUCLEAR STATION Question: 60 2009 SRO NRC Examination (1 point)

Unit 1 is operating at 100% power. Given the following:

Containment Pressure Channel I failed high

The failed channel is bypassed Which of the following lists the logic for a Containment Spray actuation based on the conditions above?

A. 1/3 B. 2/3 C. 1/2 D. 2/2 Page 60 of 100

CATAWBA NUCLEAR STATION Question: 61 2009 SRO NRC Examination (1 point)

Regarding hydrogen production during a design basis large break LOCA:

1. Which one of the following is a larger contributor to post LOCA hydrogen buildup in containment?

contain ment?

2. Per OP/1/A16450101 OP/1/A/6450101 0 (Containment Hydrogen Control Systems), what minimum containment hydrogen concentration requires Technical Support Center (TSC) approval prior to placing the Hydrogen Recombiners in service?

A. 1. Zirc-Water reaction in core region 2.4%

2. 4%

B. 1. Zirc-Water reaction in core region

2. 6%

C. 1. Dissolved hydrogen in the NC System 2.4%

2. 4%

D. 1. Dissolved hydrogen in the NC System 2.6%

2. 6%

Page 61 of 100

CATAWBA NUCLEAR STATION Question: 62 2009 SRO NRC Examination (1 point)

Given the following:

A transfer of control to the Standby Shutdown Shutdovvn Facility (SSF) has been completed

Natural circulation is being verified per OP/0/B/61 001013 (Standby Shutdown Facility Operations)

system pressure is 1700 psig NC s%tem

In-core thermocouple readings are:

o 582°F o 610°F o 588°F o 600°F o 585°F

1. Per OP/0/B/61001013, OP/0/B/61 001013, what NC system pressure indication is used for the purpose of determining the status of natural circulation?

2. For the given conditions, does natural circulation currently exist?

Reference provided A

A. 1. Pressurizer pressure

2. Yes B. 1. Pressurizer pressure

2. No C. 1. NC Loop B WR pressure

2. Yes D. 1. NC Loop B WR pressure

2. No Page 62 of 100

CATAWBA NUCLEAR STATION Question: 63 2009 SRO NRC Examination (1 point)

Unit 1 was operating at 100% power.

Initial Conditions

Both CF pumps tripped

No CA pumps could be started

WR level in all S/Gs is 0%

The crew established bleed and feed per EP/1 /A/5000/FR-H.1 (Response to EP/1/A/5000/FR-H.1 Loss of Secondary Heat Sink)

Current Conditions

CA is available

Core exit thermocouple temperatures are decreasing Based on the current conditions, how does EP/1/A/5000/FR-H.1 initially direct CA flow to be established?

A. Feed 1 S/G at greater than 450 gpm B. Feed 4 S/Gs at a total flow of greater than 450 gpm c.

C. Feed 1 S/G at less than or equal to 100 gpm D. Feed 4 S/Gs at less than or equal to 100 gpm each Page 63 of 100

CATAWBA NUCLEAR STATION Question: 64 2009 SRO NRC Examination (1 point)

Unit 1 is at 14% power synchronizing the turbine to the grid.

1. What is the normal position of the "STEAM DUMP SELECT" switch for current conditions?

2. How will steam dump demand respond to a turbine trip under current conditions?

A. 1. "PRESS"

2. decrease B. 1. "PRESS"

2. increase C. 1. "T-AVG"

2. decrease D. 1. "T-AVG"

2. increase Page 64 of 100

CATAWBA NUCLEAR STATION Question: 65 2009 SRO NRC Examination (1 point)

Unit 1 was operating at 80% power when the turbine tripped due to an inadvertent generator over frequency signal.

Which of the following lists automatic actions that will all occur as a result of this turbine trip?

1. Steam dumps actuate in Plant Trip mode

2. Zone 1A and Zone 1 B Lockouts occur

3. C-Heater Drain Pumps trip

4. Feedwater heater extraction steam check valves close A. 1,2 and 3 B. 1,3 and 4 C. 2,3 and 4 D. 1,2 and 4 Page 65 of 100

CATAWBA NUCLEAR STATION Question: 66 2009 SRO NRC Examination (1 point)

Per NSD 512 (Maintenance of RO/SRO Licenses), which of the fol/owing following lists two requirements to maintain an Active RO License?

A. 1. Serve in the RO or BOP position for 5 complete 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months shifts per quarter

2. Complete a plant tour in the areas listed in NSD 512 once per quarter B. 1. Serve in the RO or BOP position for 5 complete 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months shifts per quarter

2. Maintain respirator qualifications current C. 1. Serve in the RO or BOP position for 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months per quarter including shift turnovers

2. Complete a plant tour in the areas listed in NSD 512 once per quarter.

D. 1. Serve in the RO or BOP position for 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months per quarter including shift turnovers

2. Maintain respirator qualifications current Page 66 of 100

CATAWBA NUCLEAR STATION Question: 67 2009 SRO NRC Examination (1 point)

Per OMP 2-31 (Control Room Instrumentation Status):

1. Where is a list of active Increased Surveillances required to be maintained?

2. Who (by title) determines detennines the monitoring frequency of an Increased Surveillance?

A. 1. OPS Shift Routine Logbook 2.0SM

2. OSM B. 1. OPS Shift Routine Logbook

2. Unit Supervisor C. 1. AUTOLOG

2. OSM 2.0SM D. 1. AUTOLOG

2. Unit Supervisor Page 67 of 100

CATAWBA NUCLEAR STATION Question: 68 2009 SRO NRC Examination (1 point)

The crane operator has determined that a fuel handling interlock must be bypassed that is not addressed by a procedure.

Assume approval has been obtained to bypass the interlock.

In accordance with OP/1/A/6550/007 (Reactor Building Manipulator Crane Operation) which of the following, by title, must be notified when any "Interlock Bypass" is placed into and removed from bypass when handling new fuel assemblies?

A. WCCSRO B. Control Room Supervisor C. Reactor Systems Engineer D. Fuel Handling Equipment Engineer Page 68 of 100

CATAWBA NUCLEAR STATION Question: 69 2009 SRO NRC Examination (1 point)

Which one of the following completes the statement:

At power and increasing, the flow restricting orifice on the _ _ _ S/Gs will begin forcing flow through the CA nozzle.

A. 15%; Unit 1 B. 40%; Unit 1 C. 15%; Unit 2 D. 40%; Unit 2 Page 69 of 100

CATAWBA NUCLEAR STATION Question: 70 2009 SRO NRC Examination (1 point)

Unit 1 NC system conditions are 248°F and 600 psig.

1. For the stated plant conditions, is the FWST required to be operable per Technical Specification 3.5.4 (Refueling Water Storage Tank)?

2. When the FWST is required to be operable, and the FWST is below the borated water volume required by SR 3.5.4.2, what is the required action?

A. 1. No

2. Immediately take action to restore the FWST to operable status B. 1. No

2. Restore the FWST to operable status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months C. 1. Yes

2. Immediately take action to restore the FWST to operable status D. 1. Yes

2. Restore the FWST to operable status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Page 70 of 100

CATAWBA NUCLEAR STATION Question: 71 2009 SRO NRC Examination (1 point)

The NV system is being aligned for startup. The procedure being used calls for independent verification of a single valve located in a room with a general dose rate of 130mRlhr.

130mR/hr.

Given the following conditions:

Estimated time to independently verify the valve's position is 10 minutes

There are no known hot spots in the area

There is no airbome activity in this room

The room has no surface contamination areas

Assume any necessary approvals are obtained Per NSD 700 (Verification Techniques), can independent verification of the valve above be waived and why or why not?

A. Yes, because the general area dose rate is greater than 100mRlhr B. No, because the general area dose rate is less than 1 Rlhr C. Yes, because the radiation exposure for a single verification would exceed the allowable limit D. No, because radiation exposure for a single verification is within the allowable limit Page 71 of 100

CATAWBA NUCLEAR STATION Question: 72 2009 SRO NRC Examination (1 point)

Unit 1 is operating at 75% power. An engineer needs to enter the Reactor Building pipechase area to perform some scoping work for a future modification.

Per Site Directive 3.1.2 (Access to Reactor Building and Areas Having High Pressure Steam Relief Devices):

1. For current plant conditions, whose permission is required to allow entry into this area?

2. What operational modes require use of the "buddy system" for entry into the Reactor Building?

A. 1. Radiation Protection and the WCC SRO

2. Modes 1 and 2 only B. 1. Radiation Protection and the WCC SRO

2. Modes 1,2,3 and 4 C. 1. Radiation Protection only

2. Modes 1 and 2 only D. 1. Radiation Protection only

2. Modes 1,2,3 and 4 n of Page 72 100 oflOO

CATAWBA NUCLEAR STATION Question: 73 2009 SRO NRC Examination (1 point)

An RWP that you are preparing to work under states that the highest dose rate in a particular area (at 30 cm) is 325 mRlhr. When you arrive at the work site, a flashing yellow light is noted in the entry path to the area.

1. How would an area with the dose rate specified in the RWP be designated?

2. What is the significance of the flashing yellow light?

A. 1. Radiation Area

2. Radiography is in progress B. 1. High Radiation Area

2. Radiography is in progress C. 1. Radiation Area

2. The area has been designated as a "Locked" High Radiation Area D. 1. High Radiation Area

2. The area has been designated as a "Locked" High Radiation Area Page 73 of 100

CATAWBA NUCLEAR STATION Question: 74 2009 SRO NRC Examination (1 point)

Unit 1 was operating at 100% power when a complete loss of offsite power occurred.

A LOCA occurred as a result of the transient. Given the following conditions:

1A DIG did not start and manual start attempts have failed

Containment pressure is 3.4 psig

NC subcooling is -8°F

Core exit thermocouples are 642°F

RVLlS level is 37%

SIG NR levels are all 13%

CA has not been reset

CA flow is 470 gpm

IIR SUR is 0 DPM

SIR SUR is 0.3 DPM Which one of the following is the highest priority Critical Safety Function for the given plant conditions?

A. SUBCRITICALITY B. CO RE COOLI CORE COOLINGNG C. CONTAINMENT D. HEAT SINK Page 74 of 100

CATAWBA NUCLEAR STATION Question: 75 2009 SRO NRC Examination (1 point)

During an emergency event, which one of the following identifies:

1. The on-site emergency facility that assumes responsibility for communications with offsite agencies including the NRC once it is activated?

2. What minimum level of emergency classification always requires an evacuation of all non-essential personnel?

A. 1. Technical Support Center (TSC)

2. Site Area Emergency B. 1. Technical Support Center (TSC)

2. General Emergency C. 1. Operations Support Center (OSC)

2. Site Area Emergency D. 1. Operations Support Center (OSC)

2. General Emergency Page 75 of 100

CATAWBA NUCLEAR STATION Question: 76 2009 SRO NRC Examination (1 point)

Unit 1 was operating at 100%. Given the following conditions:

A pressurizer safety valve opened and did not fully reclose

Current Conditions per the OAe OAC are shown to the right:

1. Based on the current conditions what is the temperature downstream of the partially open pressurizer safety valve?

2. Which NC overpressure transient establishes the required relief capacity of the pressurizer safety valves per the Technical Specification Bases?

Reference provided A. 1. 550°F

2. NCP locked rotor B. 1. 550°F

2. Main turbine trip C. 1. 320°F

2. NCP locked rotor D. 1. 320°F

2. Main turbine trip Page 76 of 100

CATAWBA NUCLEAR STATION Question: 77 2009 SRO NRC Examination (1 point)

Unit 1 was heating up after a refueling outage. Given the following conditions and sequence of events:

NC temperature is 225°F and stable

1A NV pump was in service

An oil leak resulted in damage to the pump rotating element and work is currently in progress to replace the damaged rotating element

1. Which one of the following Selected Licensee Commitments (SLCs) applies based on current conditions?

2. What post maintenance requirement(s) must be completed prior to restoring 1A NV pump to OPERABLE status per the SLCs?

A. (Soration Systems Charging Pump-Shutdown)

1. SLC 16.9-9 (Boration

2. Post maintenance work order functional only B. 1. SLC 16.9-9 (Soration Systems Charging Pump-Shutdown)

2. Post maintenance work order functional and in-service testing iC.

C. 1. SLC 16.9-10 (Boration Systems Charging Pump-Operating)

2. Post maintenance work order functional only D. 1. SLC 16.9-10 (Boration Systems Charging Pump-Operating)

2. Post maintenance work order functional and in-service testing Page 77 of 100

CATAWBA NUCLEAR STATION Question: 78 2009 SRO NRC Examination (1 point)

Unit 1 was operating at 100% power. Given the following conditions and sequence of events:

Manual reactor trip and safety injection occurred due to a tube rupture on 1A S/G

A complete loss of offsite power occurs

All S/G pressures have been maintained stable by automatic operation of the S/G PORVs

EP/1/A/SOOO/E-3 (Steam Generator Tube Rupture) cooling down The crew is in EP/1/A/5000/E-3 the NC system to the required core exit temperature based on 1A S/G pressure

CUrrent NC loop Tcolds are as follows and decreasing:

o A Loop - 245°F o BLoop - 530°F o C Loop - 526°F o 0 D Loop - 528°F

1. What is the basis for the required core exit temperature selection?

2. What is the correct procedure flowpath for this situation?

A. 1. To ensure intact S/G pressures will remain above the main steam isolation setpoint during the NC system cooldown and depressurization

2. Transition to EP/1/A/5000/FR-P.1 (Response to Imminent Pressurized Thermal Shock Condition)

B. 1. To ensure intact S/G pressures will remain above the main steam isolation setpoint during the NC system cooldown and depressurization

2. Continue in EP/1/A/5000/E-3 (Steam Generator Tube Rupture)

C. 1. To establish and maintain subcooling for the NC system cooldown and depressurization

2. Transition to EP/1/A/5000/FR-P.1 (Response to Imminent Pressurized Thermal Shock Condition)

D. 1. To establish and maintain subcooling for the NC system cooldown and depressurization

2. Continue in EP/1/A/5000/E-3 (Steam Generator Tube Rupture)

Page 78 of 100

CATAWBA NUCLEAR STATION Question: 79 2009 SRO NRC Examination (1 point)

Unit 1 was operating at 100% power and Unit 2 was in No Mode. Given the following events and conditions:

Terrorist attack resulted in a total loss of switchyard

1A DIG failed to start

The OSM orders an evacuation of the control room due to the ongoing security threat per AP/1/A/55001017 (Loss of Control Room)

1. Which Unit 1 CA pump(s) is/are currently available and can be controlled from the location to which the crew is evacuated?

2. How is secondary side pressure control maintained automatically from the location to which the crew is evacuated?

A. 1. CAPT #1 only

2. By the steam line code safety valves B. 1. CAPT #1 only

2. By the S/G PORVs C. 1. CAPT #1 and 1 B motor driven CA Pump

2. By the steam line code safety valves D. 1. CAPT #1 and 1 B motor driven CA Pump

2. By the S/G PORVs Page 79 of 100

CATAWBA NUCLEAR STATION Question: 80 2009 SRO NRC Examination (1 point)

Unit 1 was operating at 100% power and Unit 2 was in No Mode. Given the following:

1100 The TCC reported that RTCA (Real Time Contingency Analysis) indicates CNS switchyard voltage would not be adequate if Unit 1 should trip

1102 The crew entered AP/1 /A/5500/037 (Generator Voltage and Electric Grid Disturbances), Case I (Abnormal Generator or Grid Voltage)

1300 The crew notified SPOC to install jumpers per AM/1/A/51 00/008 (4Kv Essential Power (EPC) System Degraded Voltage Logic)

1. Which one of the following describes the operability of the VCIYC system per Technical Specifications at 11 DO?

1100?

2. What is the basis for installing these jumpers under the current conditions?

A. 1. VC/YC VC/yC is operable

2. To bypass the 8 second time delay for load shed/sequencer actuation on a blackout signal B. 1. VC/yC VC/YC is operable

2. To prevent the D/G sequencer from double sequencing if a LOCA occurs c.

C. 1. VC/YC is not operable

2. To bypass the 8 second time delay for load shed/sequencer actuation on a blackout signal D. 1. VC/YC is not operable

2. To prevent the D/G sequencer from double sequencing if a LOCA occurs Page 80 of 100

CATAWBA NUCLEAR STATION Question: 81 2009 SRO NRC Examination (1 point)

Unit 1 was operating at 50% power. Given the following:

Initial Conditions

1A 1A CA pump red tagged for PMs

UST level is 104%

During a board walkdown, 1CA-6 (CA Pmps Suct From CA CST) was noted to be closed Current Conditions

The reactor was manually tripped due to a main steam equalization header break

All power is lost to 1ETB

1SM-5 (S/G 1B SM Isol) failed to close

1SM-3 (S/G 1C SM Isol) failed to close

1A S/G conditions - NR level is 10%; pressure is 650 psig and stable

1B1B S/G conditions - NR level is 0% ; pressure is 5 psig and stable

1C S/G conditions - NR level is 0% ; pressure is 7 psig and stable

1D S/G conditions - NR level is 8% ; pressure is 675 psig and stable

1. Based on the Initial Conditions, which CA pump(s) was/were required to be declared inoperable per Technical Specifications?

2. Based on the Current Conditions, what procedure will the crew transition to upon exit of EP/1 /A/5000/E-0 (Reactor Trip or Safety Injection)?

A. 1. 1A CA Pump only

2. EP/1/A/5000/E-2 ( Faulted Steam Generator Isolation)

B. 1. All three CA pumps

2. EP/1/A/5000/FR-H.1 (Response to Loss of Secondary Heat Sink)

C. 1. 1A CA Pump only

2. EP/1/A/5000/FR-H.1 (Response to Loss of Secondary Heat Sink)

D. 1. All three CA pumps

2. EP/1/A/5000/E-2 ( Faulted Steam Generator Isolation)

Page 81 of 100

CATAWBA NUCLEAR STATION Question: 82 2009 SRO NRC Examination (1 point)

Unit 1 is operating at 98% power. A surveillance test per PT/1/A/46001001 (RCCA Movement Test) is in progress. As Control Bank 0 was being moved, one control rod in Control Bank 0 slipped to 120 steps withdrawn and stopped. This is an incore thermocouple map one minute later.

R P N M L K J H G FED C B A 576 572 576 605 546 599 611 603 2

3 605 624 BAD 569 4 597 602 609 617 5 602 625 616 614 6 563 611 611 626 623 568 7 602 606 617 620 8 561 612 634 572 626 9 619 611 613 614 10 548 I 610 I 606 613 574 11 627 BAD 610 614 12 612 599 608 622 13 620 619 576 626 562 14 540 628 I 613 608 614 15 BAD I 622 580

1. Which single rod has slipped into the core to 120 steps withdrawn?

2. In addition to reducing power to less than 75% and verifying SOM, what other surveillances (if any) are required per Technical Specification 3.1.4 (Rod Group Alignment Limits) to allow continued power operation in Mode 1?

A. Rod 0-12; no additional surveillances are required B. Rod M-4; no additional surveillances are required C. Rod 0-12; FNt.H(X,Y)

FNilH(X,Y) and Fo(X,Y,Z)

O. Rod M-4; ~t.H(X,Y)

~ilH(X,Y) and Fo(X,Y,Z)

Page 82 of 100

CATAWBA NUCLEAR STATION Question: 83 2009 SRO NRC Examination (1 point)

Unit 1 was in Mode 6 performing core alterations. Given the following conditions and sequence of events:

N-31 was removed from service 3 days ago 2

N-32 indicates 1.1 x1 02 cps

~HUTDOWN MARGIN ALARM) actuates 1AD-2, E/2 (TRAIN A ?HUTDOWN

BDMS Train A indicates IE1" "E1" 2

BDMS Train B indicates 1 .4 x1 0 2 cps 1.4

1. What procedure is entered to address the stated conditions?

2. Are core alterations required to be stopped based on the above conditions?

A. AP/1 IA/55001013 (Boron Dilution)

1. AP/1/A/55001013

2. No B. AP/1 IA/55001013 (Boron Dilution)

1. AP/1/A/55001013

2. Yes c.

C. AP/1 IA/55001016 (Malfunction of Nuclear Instrumentation System)

1. AP/1/A/55001016

2. No D. AP/1 IA/55001016 (Malfunction of Nuclear Instrumentation System)

1. AP/1/A/55001016

2. Yes Page 83 of 100

CATAWBA NUCLEAR STATION Question: 84 2009 SRO NRC Examination /

(1 point) fOIlOW7-Unit 1 was at 3% power performing a startup. Given the followin Initial conditions /

Intermediate Range channel N-35 failed low

The crew removed N-35 from service per the pplicable abnormal procedure Current conditions

IAE retumed the channel to service nd reported that the SUR circuitry for N-35 had to be disabled as part the channel repair

Engineering has evaluated the epair and determined that all other functions will operate as designed wit he SUR circuitry disabled

1. When the crew removed N- from service per the abnormal procedure, which fuses (if any) were remov (I?

CI?

2. For the conditions abov ,is N-35 operable?

A. 1.

2.

B.

B. control power fuses

2. Y s C. No fuses were removed No

1. No fuses were removed

2. Yes Qv~S\~olJ Q V~s.\~OIJ f/t Page 84 of 100

CATAWBA NUCLEAR STATION Question: 85 2009 SRO NRC Examination (1 point)

Unit 1 was in Mode 3 and Unit 2 was operating at 100% power when a fire occurred on site. Given the following:

A large amount of smoke entered the Auxiliary Building Ventilation (VA) system ductwork

1A main transformer was de-energized due to the fire Based on the above conditions:

1. How will the VA system respond?

2. What is the status of the "A" train VA system per Technical Specifications?

A. 1. VA Supply fans (ABSU) will trip and then the Unfiltered Exhaust Fans will trip

2. Only 1A train is not operable B. 1. VA Supply fans (ABSU) will trip and then the Unfiltered Exhaust Fans will trip

2. Both 1A and 2A trains are not operable

c. 1. VA Supply fans (ABSU) will trip and then the ABSU inlet dampers will close 1A train is not operable

2. Only 1A D. 1. VA Supply fans (ABSU) will trip and then the ABSU inlet dampers will close

2. Both 1A 1A and 2A trains are not operable Page 85 8S of 100

CATAWBA NUCLEAR STATION Question: 86 2009 SRO NRC Examination (1 point)

Unit 1 is at 3% power and increasing. Given the following has just occurred:

1AD-7 C/1 NCP #1 "SEAL LEAKOFF HI FLOW" actuates

1B NCP sealleakoff is 6.5 gpm and stable

1B NCP Seal Outlet temperature is 165°F and stable

1B NCP Lower Bearing temperature is 140°F and stable

AP/11A155001008 (Malfunction of Reactor Coolant Pump)

The crew enters AP/1/A/5500/008 Per AP/1/A/5500/008, AP/1/A/55001008, what is the maximum time 1B NCP can remain in service and what procedure will the crew be directed to enter when the pump is tripped?

A. 5 minutes; EP/1/A/5000/E-0 EP/1/A/5000/E-O (Reactor Trip or Safety Injection)

B. 5 minutes; AP/1/A/55001004 (Loss of Reactor Coolant Pump)

C. 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; OP/11A161001002 OP/1 /A/61 00/002 (Controlling Procedure For Unit Shutdown)

D. 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; AP/1/A/55001004 AP/1/A/5500/004 (Loss of Reactor Coolant Pump)

Page 86 of 100

CATAWBA NUCLEAR STATION Question: 87 2009 SRO NRC Examination (1 point)

Unit 1 is at 450°F and 1860 psig heating up with shutdown banks withdrawn following a refueling outage. Given the following conditions and sequence of events:

Initial Conditions

YO has been manually aligned to provide cooling to 1A NV pump per AP/1/A/55001021 (Loss of Component Cooling)

Current Conditions

KC has not been restored

The crew trips the reactor and all reactor coolant pumps

A 100 gpm NC system leak occurs on an NC loop crossover pipe AP/1/A/55001021

1. If a safety injection becomes necessary, what guidance does AP/1 IA/55001021 provide for maintaining Unit 1 NC pump seal injection?

2. What procedure will AP/1/A/55001021 direct the crew to implement next?

A. 1. Maintain YO aligned to 1A NV pump and continuously monitor motor parameters

2. AP/1/A/55001005 (Reactor Trip or Inadvertent S/I Below P-11)

B. 1. Maintain YO aligned to 1A NV pump and continuously monitor motor parameters

2. AP/1/A/55001027 (Shutdown LOCA)

C. 1. Realign KC to the 1A NV pump and start the Standby Makeup Pump

2. AP/1/A/55001005 (Reactor Trip or Inadvertent S/I Below P-11)

P-11 )

O. 1. Realign KC to the 1A NV pump and start the Standby Makeup Pump

2. AP/1/A/55001027 (Shutdown LOCA)

Page 87 of 100

CATAWBA NUCLEAR STATION Question: 88 2009 SRO NRC Examination (1 point)

Given the following conditions on Unit 1:

Initial Conditions

Reactor power was 100% and stable

Turbine impulse pressure was 693 psig and stable Final Conditions

Security has reported significant steam escaping from the interior doghouse

AP/1/A155001028 (Secondary Steam Leak)

The crew is performing actions in AP/1/A/55001028

Reactor power is 99.9% and stable

Turbine impulse pressure is 657 psig and stable

1. What single steam relief valve passing full flow produced the conditions noted?

AP/1/A155001028 based on the final conditions?

2. What actions will be directed by AP/1/A/55001028 A. 1. A steam line safety EP/1/A15000/E-0 (Reactor Trip or Safety

2. Trip the reactor and go to EP/1/A/5000/E-0 Injection B. 1. A steam line safety AP/1/A155001009 (Rapid Downpower)

2. Initiate a unit shutdown per AP/1/A/55001009 C. 1. A S/G PORV EP/1/A15000/E-0 (Reactor Trip or Safety

2. Trip the reactor and go to EP/1/A/5000/E-0 Injection D. 1. AS/G PORV AP/1/A155001009 (Rapid Downpower)

2. Initiate a unit shutdown per AP/1/A/55001009 Page 88 of 100

CATAWBA NUCLEAR STATION Question: 89 2009 SRO NRC Examination (1 point)

Unit 1 is operating at 100% power. Given the following timeline:

12/15/09 1500 During the review of completed procedures it was noted that 12/07109 at 0300 was the last time the 7 day surveillance was performed on the 1A DIG battery cells.

12/15/09 1800 IAE completed the 1A DIG battery cell surveillance and noted one bad battery cell.

Based on the above conditions and in accordance with Technical Specification 3.8.4 (DC Sources - Operating):

BA?t:D oI-.J ~t>J:-r.:roN$

BA?£"D oIV ~D:t.-r.rt;:>N$ A-r ,500:

,5DO: CAA(..J) 1fA"~ SEEN CAA. c.. j) FfIl..\{e:SEN

1. What is the latest time that this surveiliance.eaR l3e-.completed before the LCO for Technical Specification 3.8.4..wi1l 3.8.4.wi1l nm be met? vJouc...D.J;::!!!:C B~ ME-rr IHtV-- BE.tW vJouc...D.J;::!!!:r IJ.lWe-. ME.--rf>

2. When is the cascade to Technical Specification 3.8.1 (AC Sources - Operating) reauired to be made based on the bad battery cell?

A. 1. 12/15/09 at 2100

2. Immediately B. 1. 12/15/09 at 2100

2. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months C. 1. 12/22/09 at 1500

2. Immediately D. 1. 12/22/09 at 1500

2. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Page 89 of 100

CATAWBA NUCLEAR STATION Question: 90 2009 SRO NRC Examination (1 point)

Unit 1 and 2 are operating at 100% power with 2A RN pump in service. Given the following conditions and sequence of events:

1A DIG is tagged for PMs

The following annunciators actuate:

o 1AD-12, Al2 - "RN Essential Hdr A Pressure - Lo" o 2AD-12, Al2 - "RN Essential Hdr A Pressure - Lo" o 1AD-12, Al5 - "RN Essential Hdr B Pressure - Lo" o 2AD-12, Al5 - "RN Essential Hdr B Pressure - Lo"

The 2A RN pump trips

The 1A RN pump is started and the alarms listed above clear

1. Is the 2A DIG operable per Technical Specifications for the conditions stated above?

2. What actions are required per OP/O/A/64001006C OP/0/A/64001006C (Nuclear Service Water System)?

A. 1. No

2. Ensure the RN essential header supply crossover valves are open B. 1. No

2. Rackout and tag the "A" VC Compressor breaker C. 1. Yes

2. Ensure the RN essential header supply crossover valves are open D. 1. Yes

2. Rackout and tag the "A" VC Compressor breaker Page 90 of 100

CATAWBA NUCLEAR STATION Question: 91 2009 SRO NRC Examination (1 point)

Given the following:

Unit 1 is performing a power decrease required by Technical Specifications

At 45% power, 1SMPS5219 (Turbine Impulse Pressure 1 ATWSI AMSAC Related) stopped decreasing

The power supply to 1SMPS5218 (Turbine Impulse Pressure 1 ATWSI AMSAC Related) has failed

IAE has been notified to repair failed channels

Current reactor power is 28.5%

1. What guidance does OP/1/A/61 001003 (Controlling Procedure for Unit Operation) provide for the AMSAC system for this situation?

2. Per the Selected Licensee Commitment (SLC), the basis for AMSAC, is to provide independent means of initiating a turbine trip signal and a under ATWS conditions.

A. 1. Con tin ue the power decrease Continue

2. reactor trip signal B. 1. Continue the power decrease

2. CA autostart signal C. 1. Stop the power decrease

2. reactor trip signal D. 1. Stop the power decrease

2. CA autostart signal Page 91 of 100

CATAWBA NUCLEAR STATION Question: 92 2009 SRO NRC Examination (1 point)

Unit 1 entered Mode 6 three days ago. Given the following conditions and sequence of events:

The head crew is currently lifting the reactor vessel upper internals

The B train Containment Purge Exhaust System (VP) duct heaters lose power

The B train VP system fan status! system flow is unchanged

1. For the conditions above, is the VP system required to be operable per Technical Specification 3.9.3 (Containment Penetrations)?

2. Is proper functioning of the VP duct heaters required for operability of the VP system per the bases of Technical Specification 3.9.3?

A. 1. Yes

2. Yes B. 1. No

2. Yes C. 1. Yes

2. No D. 1. No

2. No Page 92 of 100

CATAWBA NUCLEAR STATION Question: 93 2009 SRO NRC Examination (1 point)

Unit 1 is in Mode 6 with core unload in progress. Given the following:

An RP technician in the Spent Fuel Pool (SFP) building calls the control room and reports that level in the SFP is decreasing

The fuel assembly most recently removed from the core is currently in the reactor building manipulator crane

1. Based on current conditions, what procedure will direct movement of the fuel assembly being handled to a safe location?

2. Where is the fuel assembly in the reactor building manipulator crane required to be placed?

A. 1. AP/1/A/55001026 (Loss of Refueling Canal Level)

2. IInto nto the upender and lowered to the fully down position B. 1. AP/1/A/55001026 (Loss of Refueling Canal Level)

2. Fully down in the core or the deep end of the canal C. 1. AP/1/A/55001041 (Loss of Spent Fuel Cooling or Level)

2. Into the upender and lowered to the fully down position D. 1. AP/1/A/55001041 (Loss of Spent Fuel Cooling or Level)

2. Fully down in the core or the deep end of the canal Page 93 of 100

CATAWBA NUCLEAR STATION Question: 94 2009 SRO NRC Examination (1 point)

Unit 1 entered a forced outage 3 days ago to replace an NC pump motor.

2 work crews are involved in replacing the NC pump motor, an NC pump crew and an IAE crew. The tagout boundary is being shared by both crews

The NC pump crew has the new motor in place and has outstanding Work Order Task Assignments (WOTAs) for wiring completion

The IAE crew has asked the Unit 1 SRO for clearance of their WOT A to allow electrical testing of the new motor which requires connecting an outside power source Per NSD 500 (Red Tags/Configuration Control Tags), in addition to ensuring the work can be performed safely, what other condition is required to be met by the Unit 1 SRO prior to granting clearance to the IAE crew WOTA?

A. The Unit 1 SRO must obtain OSM concurrence B. The Unit 1 SRO must ensure the two crews have established constant communication C. The Unit 1 SRO must ensure all affected crews have been informed of the intent to apply an outside power source D. The Unit 1 SRO must ensure a Safety Group representative has walked down the area where the outside power source will be applied Page 94 of 100

CATAWBA NUCLEAR STATION Question: 95 2009 SRO NRC Examination (1 point)

Unit 1 was operating at 100% power. Given the following:

1N0-64 (1 B NO Train Gold Leg Injection Return Safety Relief) lifted and did not reseat

The crew used the closest valves in the piping available to isolate the leak

Due Oue to parts availability it will be 4-5 days before repairs can be completed

The crew began a Technical Specification shutdown

1. Which trains of NO, if any, are operable?

2. When shutting down, what is the next operational mode in which Technical Specification 3.5.2 (EGGS - Operating) will no longer apply?

Reference provided A. 1. No trains of NO are operable

2. Mode 4 B. 1. No trains of NO are operable

2. Mode 5 G. 1. 1A Train of NO only

2. Mode 4 O. 1. 1A Train of NO only

2. Mode 5 Page 95 of 100

CATAWBA NUCLEAR STATION Question: 96 2009 SRO NRC Examination (1 point)

Unit 1 is in Mode 3. Given the following:

An operator is bleeding off nitrogen pressure from the 1A NW surge tank

At the completion of venting, the crew notes NW surge tank pressure is 21 psig Per Technical Specification 3.6.17 (Containment Valve Injection Water System):

1. What are the modes of applicability for this Technical Specification?

2. What is the current status of 1A Train of NW?

A. 1. Modes 1, 2 and 3 only

2. Operable B. 1. Modes 1, 2 and 3 only

2. Not operable C. 1, 2, 3 and 4

1. Modes 1,2,3

2. Operable D. 1. Modes 1,2,3 and 4

2. Not operable Page 96 of 100

CATAWBA NUCLEAR STATION Question: 97 2009 SRO NRC Examination (1 point)

Given the following:

A LOCA has occurred in Unit 1

The TSC has been fully staffed and activated

An individual is needed for lifesaving activities during which 27 Rem of TEDE exposure is expected to be received

1. Is the individual required to be a volunteer?

2. Who is required to approve the exposure for this lifesaving activity?

A. 1.Yes

2. The TSC Emergency Coordinator B. 1. Yes

2. The Operations Shift Manager C. 1.No

2. The TSC Emergency Coordinator D. 1. No

2. The Operations Shift Manager Page 97 of 100

CATAWBA NUCLEAR STATION Question: 98 2009 SRO NRC Examination (1 point)

An individual with a current year Total Effective Dose Equivalent (TEDE) of 1500 mR is preparing to perform a job in the RCA. The RWP states that the dose rate in the area is 200 mRlhr.

mR/hr.

Per NSD 507 (Radiation Protection):

1. What is the maximum time he can work in this area until he accumulates enough dose to require an extension prior to the next entry into the RCA?

2. What is the minimum level RP position that could provide the approval of a dose extension if needed?

A. 1. 0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months

2. An RP Supervisor B. 1. 0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months

2. The RP Manager C. 1. 1.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months

2. An RP Supervisor D. 1. 1.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months

2. The RP Manager Page 98 of 100

CATAWBA NUCLEAR STATION Question: 99 2009 SRO NRC Examination (1 point)

Unit 1 was at 25% and increasing following a refueling outage. Given the following conditions and sequence of events:

Unit 1 experiences a large break LOCA

Containment sump level is 5.2 ft and increasing

recirculation, 1NI-136B (ND During the swap to cold leg r.ecirculation, (NO Supply To NI Pump 1B) could not be opened

S/I flow per EP/1/A15000/ES-1.3 The crew verified SI/ EP/1/A/5000/ES-1.3 (Transfer to Cold Leg Reci rculation)

Recirculation)

The crew is preparing to transition out of EP/1/A15000/ES-1.3 EP/1/A/5000/ES-1.3 when all running ECCS pumps begin showing signs of cavitation

1. What action, if any, was required per EP/1/A15000/ES-1.3 EP/1/A/5000/ES-1.3 when 1 1NI-136B NI-136B would not open?

2. What procedure will EP/1/A15000/ES-1.3 EP/1/A/5000/ES-1.3 direct the crew to enter to address cavitation of the ECCS pumps?

A. 1. 11B BNNII pump was secured

2. EP/1/A/5000/ECA-1.1 (Loss of Emergency Coolant Recirculation)

B. 1. 1B NI pump was secured

2. EP/1/A/5000/ECA-1.3 (Containment Sump Blockage)

C. 1. No action required

2. EP/1/A/5000/ECA-1.1 (Loss of Emergency Coolant Recirculation)

D. 1. No action required

2. EP/1/A/5000/ECA-1.3 (Containment Sump Blockage)

Page 99 of 100

CATAWBA NUCLEAR STATION Question: 100 2009 SRO NRC Examination (1 point)

Given the following two events:

Earthquake oo Seismic instrumentation indicates .04 g vertical acceleration and .05 g horizontal acceleration o 1AD-4, B/8 "OBE Exceeded" is not actuated

High Auxiliary Building radiation o 1RAD-1, B/3 "EMF41 Aux Bldg Vent Hi Rad" is actuated 6

o EMF41 is reading 1.2 x 106 counts per minute and slowly increasing Which one of the above events requires activation of the site assembly alarm per RP/O/A/50001010 (Conducting a Site Assembly or Preparing the Site for an Evacuation) and from what location(s) is this alarm activated?

A. Earthquake Event; Control Room only B. Earthquake Event; Control Room and Secondary Alarm Station (SAS)

C. High Auxiliary Building radiation; Control Room only D. High Auxiliary Building radiation; Control Room and Secondary Alarm Station (SAS)

Page 100 of 100

Reference Listfor:

List for: 2009 SRO NRC Exam CN-1561-1.1 CN-1562-1.3 Databook Figure 148 Databook Figure 158 Databook Figure 57 EP/1/A/5000/ECA-1.1 Enclosure 5 Steam Tables

10 12

...x>~~t::;1Va?~

t..X>l:~:c;fVO~\S uhU- ubu.... G-el C-cl Fuu-<;t~£ cal'\~

Fuu.....4l'C £ c.oP\~

(} r'N'\.<

o ~(Ll<'1 'f p-~, CJ(LV'/ 1"" .

I- ~

I t ~I~

l

~T-@

.1

~t....

. I d

1!i!'1... @

l

. IGtt "G) 1I 1110,,1 1

~,

I

~

L ..__ wj

'1-GZE!l ,~.

24

.. 1'It=z..*r~

QA CONDITION 2

". OA QA CONOrT{ON.

eOND IT ION r1

~.............., ~

QUI(E POWER DUKE POWEA COfo'IPANY' COMPANY CATAWBA NUCLEAR STATrDN STATtON UNIT It

...". -;-"~~==,~~~ .. !Ill! FLO\(

ROW O(AORAM OIAORAI'! OFOf"

! Iti ~!fi If i g it iSU ilafi iii H -

COA -

PFS-Pf'8 0131 RESIDIJAL RESIDUAL HEAT R£MOVAt.

0/1 ( ..

SVSl'EM REMOVAL SYSTEM (t.(j)

I NO)

DIbWN~OIto'tlblt:lL. ""

_.~DIITt ERN. CN000HL V OWO. NO. CN-1561-1.1I CN-.!561-1.

I IT I 14 14

a Unit 1 Data Book DPC*1552.08-00*0014, Rev. 8 Source: DPC*1552.0S-00-0014, Figure 16*2 16-2 95 1I 90 Initial NONC ~----~----+----~4------r-----+----~~~

~

85 80 75 +------ System Temperature 70°F a F

~---+-~__+ _____+-_

~

", ~ --- -

~

70

- w 90 of I

~ ~

70 ---1 -- -w -- 100 100 of OF ~----j-------j----

~ ..- .. ..-

65 -t--------i -- -120 of

---120 of ------I-----1---:::7'..e::.--I----::;;--F-- "

-);-140 of

-*-140 .,.......V" ...... ..

U!

(/)

~~-

£~

I 60 "

l C

55

-+-160

--*-160 of

---+-180 of of

°F~

J- . ----- I

~ -- .. ..

..-."- J- ..

" - - ' .. '" ~ -

I ..- ....

. .. .. - I

~ 55 +-------il --+-180

1E

~200QFI ~

~I .- ........L ......

........... " , " \....

,-200°F

.S 01 50 c:

~ 45 V,.../_-'.... " r:

1--'-

~

l...----"-'"

-=

co CIl

=.8 40 Q)

(!) z~

/

/1""'

/

~ ~

..~ ~ ...

1--""

... ... -" ~~-I-'...;+t-

~'

~-*""

~~ -- ~

E 35 ,.... ....

.- I _.""~

i=

30

.L't ~

/ .""..:'___"-_:_ ~"

.. ~~ ~ ~~ ~ ~

25 .'~~-~

c ~" . ~

.......- ..a--" ~

i-~

20 ~ /" ....-.---~

15 t,; ~ ~:-.-~

_+-

-+-- ~-

,--+--

~

10 ~~

5

~

a0 ~

oa 5 10 15 20 25 30 35 40 45 Days After Shutdown Rev. 0,11/19/2003 Figure 148 14B - Loss of Decay Heat Removal, 18% NO Level, High Decay Heat NC level, Page 15 ~

of

(.~ '\

~. .

Unit 1 Data Book DPC*1552.0a*OO*0014, Rev. 8a Source: DPC*1552.0a-00-0014, 16-8 Figure 16*8 140 l 1I :~

-r--.----.--

130 1--_._ ..__ .. _--

Initial NC Initial NC ~

..... ~ ~

System i Temperature 120 .~I

~

~ ~ . .

110

--70°F

- 70°F

- - - 90 of --

~

- -100 of

~ .

-- -- ~-"'

~...,. i I

100 ...

- --120 OF of

~ ..

-c:c 80

(/)

(f) 90 -_.----

~140*F

-*-140 of r--.-~ ~

./

V I. . ........

t .. .- .. ........ ..

OJ Q)

-+-160 of

/ .......... ..

I

~,....

~~

l

-+-180 of

~180°F

~,....

.. I-

./

~

iE

-- - ~....

~

- ~

L / . -

--200

- - 2 0of 00F'/

~

§

.f

'0 co cl Oi 0

70 60 I

.j. I.I*, /

[- .~ ..........

..,,~~

~7'.........

. .. 1- - . -- - ..

~

~.

~~

- V4" IlJ 0

V ..... ...

7-~--- ---..........

t-ff I

L -,,' ,. ,*___

..... ~ t~ ~-

I~

~

,~ ~ ~ ~~

. ","'" ~~ .

IOIIJ E

i= 50 i . / ." /

/

~

",.... ,.... ' ,._J-''-~- *___ ~e---

. ~ -~

40 30 10

~~'

20 ~V-- ~

20 .

?:,,~

~~ /"/

~ /:/~* ,

~ ~-

~----

~ ~l~

r'"

~

t~

~- -- ,---

10 ~~

10 I~

~ - ,

-I-' --.

o0 II __ __ **___

___ .L. .*... _."_~-'-_ ~

'._.,._~ ..... _ _ **

~ .**

... __.j __ .. _1. _.,

..... _.l .* _

_~.."-i--'--',. . . .-_.- .__'-_. .-'---'_ . . . . . , II o 5 10 15 20 25 30 35 40 45 Days After Shutdown 11/19/2003 0,11/19/2003 Rev. 0, Figure 15B 158 - Loss of Decay Heat Removal. Removal, 18% NC Level, Low Decay Heat Page 21 I

i_.

) .~ ... -.,'

')

Data Book Unit 1 Data Source: OAC Reactor Coolant Saturation Calculations 2500~------~-------r------~------~--------~------~--~--~-------r------~/~----~

~ V 2400

~~

2300 +------l--------r---'----i-----+--------I------i-------j-----r--,,___

~

2200+------+--------1*-------------*-----+--------+--~,----~-------~r_-----~----~T-~----

~

,, I j

°° 2 1~O f-.----'-'-

21 I- -.-.---+------,-,---,----,-.--'-'----"--------+-------+------t,----/-/--+------1

.:-.---- '----,-r---+-----;

~

20 00 ---+-------+-----r-*'. - . *---,-------*----t-----+----*-----*------I----/---t------l 1900 t-----.----t.--------------+---.------.------------,--.------l--- ..- --~- ----1------+-.,.1------+-,/

I ,--~

1800 -.-.,,------ - - - - . -- - - - " .. _._-- ...... _............- .-_ .. - _ ......... __ . __ ...* -.----

1800 ----...... - - - , . , - - - - . - - - . - *. _--'- - - - - - - - - . -...-...........- ............... -.. - ..- ...---- ...- .... - ..-....- ....----.... -- ..- - - -_ _---j_ _ _ ,_____ V Vi

_ 1700 - " - _.---:-. ..---- --... _-_._....._-. _.. --+----/-I-----+------i

.~ 1600

.~ ~ ~~~ ,=--.:=---~-

- -.._- .--~=I*---+--**-*~=-=r~~~~=~=:~~,~-=--~:~--*-*-----+---*

.-.._- ._--- ._-_._.._..- ----.-- .-!---l------I-----l

/

~

-; 1500 .. -- ....-----...:-----.--.---r----..

.--...... -.---'------- ----j

.------/ Subcooled Region ,.....----------.-.. _- ---"-""--'---"-'-- -_.....---,--/-......,i-----+-----~

--..-------.--'---' -"---./-/-~l-----I-------t ie~~~~. =._.--_--_-'-_.

~ 1400* -"-"--"-'- ,....----.-

1300 -I--'

-_--t-_--

___-=_~=~--------r---_=_-_--r_-._==-__-..I.-_=='=~~,~~_-~~~~=,~!---/-,~/--+.---

J..

- - - - - --'-" /

/

-*--+-----1 ct

~ 1200 ---.----1-----,-. ---t------- ---._-----*--*--~--------4-------~------_r------~-----~

1200 ..- .------- --.----I----1,L.---+-- .-.---I------l

'*E

(\) 1100 ------+----.-j-----------_.-.. + --.._ -----..

. --.--.- - -.-------I---------I----/-/-'-I_-------+-----+------l

-----I-'------i---L..........,.,o;....r-----~----+------j

/

~ 1000

~z>- 000 ----------j-----+------I-------- -----.-.-- -------t-_-L-_""'_/-_--. . . '-----I--*-----I------I W

(,)

19 900 ---.---.-_j_-,---+-------r------f-----.....- ---- ****-----.;-7--,.*************

...- - ' - - - - + -

L

. " c - --f------i------t-'----f

~-.-_L~'/-L:.....----I--.---r----+------j Z 800 1-.

~~

~~ I----=:=---~=_==_-=-==--=+~

500 t1------1--------

"-I--' -=~=--==~~--~E-'

Z~-=~.- / --------

_~_

1

- c---

rsup~rh~~i;d-Regio~}-*-+----,...

Superheated Region ..-

400 400 - - - ~~ - . ..,----- I i I -- ---.--......,

300 - - - - - - + - - - - .

300 ~~

200 200+-------I~-----_j_--- __

~ __ ~-=~*---+-------4*---------I-* -----t-------+------.----.-----I 100 200 100L------~=_--~~~~-~-------+------~r------ ---+_------j---------r-----.---

250 o0+-~~--_r--------_I_--------+--------+---------l-------__tr__-----_+--------~--------r-------_4 300 350 400 450 500 550 600 650 700 NC System Temperature (Oeg (Deg F) 4/5r ')

Rev. 2, 4/51" Figure 57 - Reactor Coolant Satur Satur ' Curve PagE' Pagf

CNS LOSS OF EMERGENCY COOLANT RECIRCULATION PAGE NO.

EP/1/A15000/ECA-1.1 EP/1/A/SOOO/ECA-1.1 74 of 83 Enclosure 5 S - Page 1 of 1 Rev 31 DeS DCS Minimum 511 S/I Flowrate Versus Time After Trip sir SII FLOW REQUIRED TO MATCH DECAY HEAT FLOW RATE CGPM) (GPM) 600 ~----~~~-r~~~--~--~~~~~~--~--T-~~~-n I\,

\. - - -.-- ......-.I-f--H--H---+----l

--\.

\.

'\.

\.

'\.

\.

'\.

500 ~--~~~_+_+_+~~+---~--~~~r+++~--_+--+_~~~_H

'\.

\.

'---+f\---t-I--t---t--H-t--t---- -- .--!---I--I-I j - H - - I - - * - t - - f - - -"

I---}--",,\.--! \. - - - - -+-++-++-1-----1"---- .----f-- -

\.'\ -- - - -I-+-t--!-}-----f--_t__ -I-}--.j-++

'\.

I\,

\.

400

\.

400~--~--+-~~rHH----+--~~-rHH+---~--~r+~~

I\,

-**---*-I-------!--t-H-t--H------*-+-+-l-++++i---*- - - . . .- -.- --

-- -1\- - ------1---+-+-+--+-1-+++------ --- . --+--1-++.+-

300 300-tr---!-~~~~~~-r__l___l__~~H+--+-_+_+_l__++~

f-----I-I----I---- +~--.

'""-"- - . --..-. -.--.1-+--1-+-1-- - . _ - - -.--...-1----- .-

'""' I"\..

---+----I~- -... r-- -_OJ--t++-----+--+. ,,-~----_I.--+-I-t+--I---- __ ---+-+-+++-1 200 200~---+--+_~~~H_--_+--~~~HH+---~--~~44+H '"

- - - - . - -- -+- - .--..-.- .---- - - - - ........

-+-1--........

-" - ......d---*** - - - - - .-. -

- ..- - f - - - - __ r-

__+.+ -t-t-*,-_-*==__

__ . -+-*_- __ -+**_---t-..::--tt_*--j}---!-j--j-j.---__ ~ ::........~...... f---+---1-+-1-- ....... - -"

100

- - - - - - - --+----I-l-l-H-----f--~-+__l- -------.- ---i-t-- ..- - -----

o 3456789 ".:;3 44 5 5 607 -78 89 9 3456789 10 100 1000 10000 TIME (MINUTES)

, ~' "

EAN CLEAN COpy OF TEAM STEAM T BLES TABLES

Examination KEY for: 2009 SRO NRC Examination Question Answer Number 1 B 2 0 D

3 A A

4 A A

5 A A

6 A A

7 0 D

8 C 9 A A

10 C 11 B 12 0 D

13 0 D

14 A A

15 0 D

16 D 0

17 A A

18 A A

19 D 0

20 A A

21 B 22 A A

23 C 24 A A

25 0 D

12/17120095:18:37 PM Printed 12117120095:18:37 0/4 Page 10/4 Pagel

Examination KEY for: 2009 SRO NRC Examination Ouestion Question Answer Number 26 0 27 A 28 0 29 0 30 B 31 A 32 A 33 0 34 B 35 B 36 A 37 B 38 B 39 A 40 0 41 C 42 0 43 A 44 B 45 C 46 0 47 0 48 A 49 B 50 B Printed 12117120095:18:42 PM Page2of4

Examination KEY for: 2009 SRO NRC Examination Question Answer Number 51 D 52 A A

53 A A

54 D 55 C 56 B 57 C 58 A A

59 B 60 B 61 B 62 D 63 C 64 B 65 B 66 B 67 B 68 B 69 D 70 D 71 C 72 B 73 D 74 B 75 B Printed 12117120095:18:42 PM Page30f4 Page 3 0/4

,t ,,

1 Examination KEY for: 2009 SRO NRC Examination Question Answer Number 76 C 77 D 78 D 79 A 80 D 81 C 82 D 83 A 84 D D bELs-\E~ yt:

85 D 86 D 87 C 88 A 89 A 90 A 91 B 92 D 93 B 94 C 95 A 96 D 97 A 98 D 99 D 100 C Printed 1211712009 5:18:42 PM Page4of4

ML100470549

Text

Navigation menu

Search