ML051240488

Initial Submittal of the Written Examination for the LaSalle Initial Examination - March 2005

ML051240488
Person / Time
Site:	LaSalle
Issue date:	03/07/2005
From:	Ross J Exelon Generation Co
To:	Office of Nuclear Reactor Regulation
Shared Package
ML050900103	List: ML051240350 ML051240371 ML051240372 ML051240377 ML051240396 ML051240410 ML051240468 ML051240473 ML051240488
References
50-373/05-301, 50-374/05-301 50-373/05-301, 50-374/05-301
Download: ML051240488 (102)

Text

INITIAL SUBMITTAL OF THE WRITTEN EXAMINATION FOR THE LASALLE INITIAL EXAMINATION - MARCH 2005

LaSalle 2005 Initial Examination Proposed Facility-Developed Written Examination RO Questions 1-75 SRO Questions 76-1 00

v 1

I I

~

L

1. 295001 AKl.O 1 00 1/29500 l/AK1.O 1/3.5/3.6/RO/MEMORY//

~

~

(1.OO Point)A differential temperature between the bottom of the RPV and the RPV flange that is greater than 80°F on LOA-RR-l Ol, Attachment B, Monitoring of the Reactor Water Temperature / Pressure During Loss of Both RR Pumps is indicative of...

A? inadequate natural circulation flow.

B. inadequate RT bottom head drain flow.

C. inadequate control rod cooling water flow.

D. excessive RT non-regenerative heat exchanger WR flow.

~~~

~~~

~

~~

Answer A is correct The referencedprocedure states Yfdelta Tper Attachment B is >80%bt the direction of the Shrfi Manager RAISE Rx Level to at least +50 inches on Shutdown Range / Upset Range Rx Level using available systems, as necessary to promote natural circulation and prevent temperature stratlfcation while continuing with subsequent actions "

Reference:

LOA-RR-101, Revision 16, page 8, Step B.2.1.7 v

2003-01 ILT NRC Exam, Version: 5 Page 3 of 105

2. 295003 AK2.06 00

~~~ 1/295003/AK2.06/3.4/3.5/RO~EMORYl006.00.18/

(1.OO Point) Assume that AC Electrical Power is lost to the Division 1, 125 vdc Battery Chargers.

Continuing to operate without restoring AC Electrical Power to the Division 1, 125 vdc Battery Chargers will result in...

A!' inoperability of RCIC due to loss of power to the governor electronics and several RCIC instruments.

1 B. inoperability of the 1 A Diesel Generator due to loss of field flash and control power C. failure of the turbine lube oil to automatically start Emergency Bearing Oil Pump when required.

I D. automatic closure and isolation of many inboard primary containment isolation valves.

~-

-. ~-

~-

Answer A is correct. Without a battery charger, the Division1 Battery will continue to discharge resulting in inoperability of loads supplied by I I IX and I I I Y. Circuit Breaker 8 on I I I Y feeds RCIC and FW Interlocks.

Specifically, the RCIC governor electronics, flow controller, steam supply pressure meter, exhaust pressure meter, pump discharge pressure meter, and pump suction pressure meter.

The other answers are incorrect because they are NOT supplied by Division I 125 vdc batteries I ) the turbine lube oil Emergency Bearing Oil Pump is supplied by 250 vdc

2) P U S Inboard isolation logic for Groups 2, 4, 5, 6, 7, and 10 is supplied by 112Y (Division 2)

3) IA DGfieldflash and control power is supplied by I12Y (Division 2)

Reference:

LOA-DC-IOI, Revision 07, pages 184, 192, 193, and204 2003-01 ILT NRC Exam, Version: 5 Page 4 of 105

3. 295004 AK3.03 001/295004/AK3.03/3.1/3.5/RO/MEMORY/049

00. IO/

(1.OO Point)Why will the Unit-1 reactor automatically scram as a result of simultaneous lossof 11 1Y and 112Y?

v A. De-energizes both divisions of ARI initiation logic.

B. De-energizes both backup scram valves.

I C. De-energizes control power to both TDRFPs.

I D1/ De-energizes both divisions of isolation logic.

L--

-~

~.

Answer D IS correct. Loss of I I I Y causes the Division I hayof the Golation logic to de-energize. The logic is fail-safe and initiates a half-isolation. Loss of I i2Y de-energizes the Division 2 half of the logic resulting in a PCIS Group I isolation (MSIVs and Drain Valves).

The other answers are incorrect. The TDRFPs continue to operate as-is without tripping capability when control power is lost. Both ARI and Backup Scram valves are energize to actuate NOTde-energize.

Reference:

LOA-DC-101, Revision 07, pages 184 and 193 2003-01 ILT NRC Exam, Version: 5 Page 5 of 105

4. 295005 AA 1.02 00 11295005/AAI.02/3.2/3.6/RO/MEMORY/O49.OO.10/

~

I (1.OO Point) Given the following initial conditions:

- Reactor Power at 20% of RTP

- Normal electrical lineup for Mode 1 Considering the above initial conditions, which one of the following states the status of the eighi white RPS Group Scram lights immediately following a Main Turbine trip when Turbine Stop Valve #1 sticks OPEN?

A? Top row LIT, and bottom row LIT B. Top row OFF, and bottom row OFF C. Top row LIT, and bottom row OFF D. Top row OFF, and bottom row LIT Answer A IS correct When above 25% power first stage pressure) the reactor will scram follow a turbine trip and TSV closure With power below 25% this scram is automatically bypassed than therefore NO scram signal IS received

Reference:

LOS-RP-Q2, Revision 14, page 3, Step A.2.1, andpage 4, Step C. I LOR-IH13-P603-A21 I 2003-01 ILT NRC Exam, Version: 5 Page 6 of 105

5. 295006 AK3.O 1 001 l295006IAK3.O 1 /3.8/3.9/RO/MEMORY//

(1.OO Point) Following a manual reactor scram from rated conditions, RPVwater level decreased to -25 inches then started to recover.

+

The RPV water level response is...

A. abnormal, the TDRFPs should respond quicker to minimize the transient.

B. normal, due to RR pumps downshifting to slow speed.

C? normal, due to void collapse causing shrink.

~

~

~

I D. abnormal, the MDRFP should have started to minimize the transient.

~

~

Answer C is correct. The rapid reduction in thermal power following a scram causes a significanldecrease in voids. This void collapse causes level to shrink, giving this normal level response to a scram from full power.

The other answers are incorrect because the response is normal not abnormal and Predefined FW Profile mitigates the response but does not prevent it.

i

Reference:

LOP-RL-01, Revision 18, page and 26, Attachment A System Description 03 I, page 15 2003-01 ILT NRC Exam, Version: 5 Page 7 of 105

v I

I I

6. 295006 AA2.02 001/295006/AA2.02/4.3/4.4/RO~IGW~

- ~-

1 (1.OO Point) LGA-001 entry conditions have been met.

Which one of the following conditions would require entering LGA-OlO?

A. one control rod is at position 04 and one at 02; all other control rods are at position 00 BY two control rods are at position 04; all other control rods are at position 00 C. twenty-five control rods are at position 02; all other control rods are at position 00 D. one control rod is a position 48; all other control rods are at position 00

~

Remove references to position 02 from LCA-001 and LGA-010prior to using as handouts.

Answer B is correct. LGA-001 requires Exiting LGA-001 andEntering LGA-010 ifthe answer to the quesrion '211 rods except one in to at least 02?" is answered "No" or "unknown" because the scram has NOT been successful.

The other answers are incorrect because with the control rods as described in the distracter, the answer would be

'yes, I' and you would continue in LGA-001 instead of exiting to LGA-010 because the reactor scram has been successful.

Determined to be higher cognitive level question because the RO has to interpret the control rodpositions in order to determine the answer to the question "11 rod except one in to at least 02."

Reference:

LGA-001, Revision 06 L

2003-01 ILT NRC Exam, Version. 5 Page 8 of 105

7. 295016 G2.1.32 001/295016/2.1.32/3.4/3.8/RO/MEMORY/054.00.05/

~

(1.OO Point) You have been directed to transfer RHR from Suppression Pool Cooling to Shutdown Cooling at the Remote Shutdown Panel (RSP).

While performing the above task, there are NO interlocks that would prevent you from opening 1 E12-F006B, Shutdown Cooling Suction if

~

W is already OPEN.

A? 1 El 2-FO 16B, 1 B RHR Upstream Drywell Spray Isolation I

B. 1 El 2-F004B, 1 B RHR Suppression Pool Suction I

C. lE12-F024B, 1B RHR Full Flow Test I

D. 1 El 2-F027B, 1 B RHR Suppression Chamber Spray I

.~

Answer A is correct. RHR System isolation, automatic initiations, realignment and protective features are bypassed when control isf/om the RSP with the following exceptions:

I) Reactor Drain Interlocks; IE12-FO06B will NOT open unless IEI2-F027B, -F024B, and F004B are all closed

2) Pump electrical interlocks

3) IE12-FO04B will NOT open unless IE12-FO06B is fill closed

Reference:

LOP-RX-06, Revision 06, page 2 and 3, Step D. I. I, D. 1.2 and D. 1.3.

2003-01 ILT NRC Exam, Version: 5 Page 9 of 105

8. 29501 8 AKl.01 00 1/2950

- 18/AK 1.O 1/3.5/3.6/RO/MEMORY/O 1 1.00.2 1/

( 1.OO Point) The 1 A Diesel Generator (DG) Cooling Water Pump Tripped while the 1 A DG was -'

running under load during surveillance testing.

Assuming no operator action, which one of the following statements describes the expected impact on continued 1 A DG operation?

~

A. The 1A DG governor will runback the load limiter to 10% to prevent damaging the 1A DG.

B. The pump trip will directly actuate the DG lockout which will trip the 1A DG.

I CY The 1A DG will eventually trip on high cooling water temperature.

D. The 1A DG will eventually fail with possible damage due to lack of cooling.

Answer C is correct. With no ECCS signal present, the DG will heat up and trip on high cooling water temperature (208F) before the DG is damaged.

Reference:

LOP-DG-01. Revision 30, page 5, step 0.2.2.2 System Description 01 I, page 45, item 2 on table v

2003-01 ILT NRC Exam, Version: 5 Page 10 of 105

9.

~

295019 AK2.06 001/295019/AK2.06/2.8/2.9/RO/MORY/080.00.16/

I (1.OO Point) With the unit operating at near full power andthe Off-Gas system in a normal operating configuration, Instrument Air (IA) pressure to the Off-Gas system components starts to decrease and can NOT be corrected.

Which one of the following describes the effect on the 2N62-F300A and 2N62-F300B Condenser Off-Gas Outlet valves as IA pressure decreases?

A!' They will automatically CLOSE when IA pressure decreases to the closure trip setpoint.

B. They will automatically OPEN when IA pressure decreases to the opening trip setpoint.

C. They will fail CLOSED when IA pressure decreases to less than spring pressure.

D. They will fail OPEN when IA pressure decreases to less than spring pressure.

u

~

I

-~

~~

L-Answer A is correct. The Condenser Off-Gas Outlet Valves have accumulators that provide enougkstored energy to ensure one positive closing of the valves. When IA pressure decreases below 75 psig, air stored in the accumulator will force the valve operator closed.

The other answers are incorrect. They fail closed, NOT open. They do not drift in the closed direction as IA pressure decreases, they get a close signal when IA pressure is less than the trip setpoint and accumulator pressure is used to close them.

Reference:

System Description 080, page 5, Section IYB, page 18,Section V.A.2, andpage 20, Table 2003-01 ILT NRC Exam, Version 5 Page 11 of 105

10. 29502 1 AK3 02 00 1/29502 l/AK3.02/3.3/3.4/RO/MEMORY//

The unit is operating in Mode 3 when it is discovered that lEl2-FO09, Shutdown Cooling Suction Inboard Isolation Valve will NOT open.

The Unit Supervisor directs the following:

- using 1 A RHR to establish reactor water level +190 to +260 inches

- opening 3 SRVs

- throttling 1A RHR Full Flow Test valve (1E12-FO24A) to maintain reactor water level in the I

- 1 band and a flow rate of 6500 gpm through the 1A RHR pump I

Following the Unit Supervisor's instructions will...

i A. provide adequate flow to remove the decay heat load without damaging the SRVs.

I B. cause runout and subsequent damage to the RHR system.

C. cause cavitation and subsequent damage to the 1 E 12-FO24A.

D!' provide adequate flow to remove the decay heat load but may damage the SRVs.

~

Answer D is correct.--This method establishes predetermined (ay engineering evaluation) conditions for draining and makeup to removed the decqv heat and bring the unit to Mode 4from Mode 3.

The other answer are incorrect. Ifuiedfor waterflow the SRVs cannot be considered operable without an engineering evaluation. Each SRV will only pass 2150 gpm water flow, the sum of which is less than pump runout conditions. The RHR Full Flow Test valve is throttled to maintain level in the RPV and a caution in the procedure sqvs to throttle the Full Flow Test valve to maintain less than 7200 gpm so the pump does not go into runout.

Reference:

LOP-RH-17, Revision 18, page 4, Step B.1.3, andpages 60 to 63, Steps E.25 and E.26 L'

2003-01 ILT NRC Exam, Version: 5 Page 12 of 14

11. 295023 AA1.02 001/295023/AA1.02/2 9/3.1/RO/MEMORY/029.00.05J/

~

-~

(1;OO Point) The Fuel Pool Cooling and Cleanup (FC) system was operating with the 1 A FC Pump running and both Filter Demineralizers (F/Ds) were in service. The 1B FC Pump was in standby.

Subsequently a large leak develops in the refueling bellows.

Which one of the following correctly states the expected response assuming NO mitigating operator actions are taken?

u A.

I B.

CY D.

4 An FC trouble alarm due to 1A FC Pump trip on low Spent Fuel Pool level and both F/D Hold Pumps start.

Condensate Storage Tank (CST) level decreasing due to Cycled Condensate (CY) automatically aligning to maintain minimum Fuel Pool level.

An FC trouble alarm due to 1A FC Pump trip on low Suction Pressure and both F/D Hold Pumps start.

Makeup Condensate Storage Tank level decreasing due Clean Condensate (MC) automatically aligning to maintain minimum Fuel Pool level.

~

~

~

~

i Answer C is correct The FC Pumpswill automatically trip on I ) LGSuction Pressure (2-second time delay), 2)

Under Voltage, and 3) Overload The other answers are incorrect. The F/D Hold Pumps will automatically start, however the standby FC Pump has NO airtomatic start feature.

Reference:

LOP-FC-03, Revision 23, page 14, Steps 0.2.2 and 0.2.4 System Descriplion 029, page 18,Section III.F.3.

2003-01 ILT NRC Exam, Version 5 Page 13 of 105

~~

L-I i

I

12. 295024 EA2.01 00 1/295024/EA2.01/4.2/4.4/RO/HIGW/

( 1.OO Point) An event occurred resulting in the following conditions:-

- Reactor scrammed, all control rods are full-in

- HPCS, LPCS and LPCI automatically initiated

- All three DGs are operating

- RCIC is in standby

- All plant equipment operated as designed The above response is expected when is the only initiating parameter.

A. high RPV pressure B. low RPV level (Level-2)

C. low RPV level (Level-1)

D!' high Drywell pressure Answer D is correct. All of the indications have high Dtywellpressure as an initiation signal.

The other answers are incorrect. Level did not drop to level-2 or level-] because RCIC never received an initiation signal. An if RPV Pressure was high, ECCS systems and the DGs would not have automatically started.

Reference:

LGA-001, Revision 06, Detail L LOA-PC-101, Revision 07, pages 16-18 System Description 060, page 6,Section II. E.

System Description 01 I, page 44,Section V.A. 1 System Description 049, page 26, Table of scram setpoints System Description 032, page 2,Section II. A.

2003-01 ILT NRC Exam, Version 5 Page 14 of 105

- 13. 295025 G2.2.22 001/295025/2.2.22/3.4/. IIROIMEMORYN -~

(1.OO Point) Per the Technical Specification LCO, Drywell and Suppzssion Chamber pressure shall be maintained 2-0.5 psig and in MODES 1,2, and 3.

v A. 10.60 psig BY 10.75 psig C. 51.00 psig D. 11.77 psig I

~~

L -

~

Answer B IS correct. TS 3.6 1 4 states, "Drywell and suppression chamber pressure shall be 2-0 5 psig and SO 75 ps1g "

Reference:

Technical Specrficalion 3.6.1.4, page 3.1.6.4-1 v

2003-01 ILT NRC Exam, Version 5 Page 15 of 105

~~ 14. 295026 EKl.01 001/295026/EK1.01/3.0/3.4/RO/HIGH//

(1.OO Point) Due to LOCA conditions in the Drywell, the following conditions ex:;:

- Suppression Pool Temperature is 195°F

- Suppression Pool Level is +6 inches

- Suppression Chamber Pressure is +0.5 psig

- Drywell temperature is 195°F

- One Containment Vacuum Breaker is stuck OPEN Which one of the following would DECREASE the probability of damaging the LPCS pump if operated under the above conditions?

A. Starting all available Drywell Cooling per LGA-VP-0 1, Primary Containment Temperature Reduction.

B. Starting one loop of Drywell Spray per LGA-RH-103, Unit 1 AB RHR Operations in the LGAs/LS AMGs.

C. Returning Suppression Pool level to within limits per LOP-RH-16, Raising and Lowering of Suppression Pool Level.

D!' Starting two loops of Suppression Pool Cooling per LGA-RH-103, Unit 1 AB RHR Operations in the LGAsLSAMGs.

~~~~~

~~

Answer D IS correct With suppression Pool (SP) temperature high, starting SP Cooling would increase NPSH to the LPCS Pump The other answers are incorrect. Lowering SP Level would reduce NPSH. Starting Drywell Spray (or SC Spray) would reduce static pressure in the containment and therefore reduce NPSH. Starting Drywell Cooling will have little or no effect on NPSH.

Reference:

LGA-003, Revision 05, figure NL 2003-01 ILT NRC Exam, Version 5 Page 16 of 105

15. 295028 EK2.03 00 1/295028/EK2.03/3.6/3.8/ROlHIGW4 14.00.02l

~~

(1.OO Point) A LOCA has occurred resulting in the following plant conditions:

- Drywell Temperature indicates 290°F

- RPV Pressure indicates 50 psig

- Reactor Building Temperature indicates 155°F

- Narrow Range Level indicates 0 inches

- Wide Range Level indicates -90 inches

- Upset Range Level indicates +20 inches

- Shutdown Range Level indicates +20 inches

- Fuel Zone Level indicates - 120 inches Based on the above conditions, which one of the following statements about RPV level indication is true?

I A? Both Fuel Zone and Wide Range instruments are usable.

B. None of the level instruments are usable.

C. Fuel Zone is the ONLY instrument that is usable.

D. Wide Range is the ONLY instrument that is usable.

~

~

Provide LCA-001 lo the examinee as a handout.

Answer A is correct. With the given RPVpressure and Drywell temperature, the RPV Saturation Temperature curve (Figure J ) has NOT been exceeded and reference legs most likely have NOTJashed Therefore, analysis of individual instruments based on building temperatures per RPV Level instrument Criteria (Table K) must be made.

Fuel Zone level is on-scale as long as level in the indicating range of the FZ instrument (-31 I to - I I I inches). This gives FZ an on-scale reading. Wide Range (WR) is on-scale Vindicating greater than or equal to -97 inches regardless of Reactor Building temperature. This gives WR an on-scale reading. Therefore both WR and FZ are indicating on-scale and the other answers are incorrect.

Reference:

LGA-001, Revision 06, Detail I LGA-010, Revision 06, Detail I 2003-01 ILT NRC Exam, Version 5 Page 17 of 105

16. 295030 EK3.07 00 1/295030/EK3.07/3.5/3.8/RO/HIGW/

(1.OO Point) Given the following initial plant conditions:

- Suppression Pool Level is -3 feet

- Suppression Pool Temperature is 2 10°F

- Drywell Pressure is 8 psig

- Suppression Chamber Pressure is 3 psig Which one of the following would most likely cause system damage while operating the 1 B RHR pump in the LPCI mode?

AJ Suppression Pool Level decreases by 18 inches.

B. Drywell Pressure increases by 4 psig.

I C. Suppression Pool Temperature increases by 5°F.

I D. Suppression Chamber Pressure decreases by 2 psig.

~~~

~

~~

Answer A is correct because decreasing SP Level by 18 inches would make SP L i e 1 -4.5feet and the -I3 foot curve becomes applicable (vice the -4 foot curve) and this increases the probability of operating with inadequate Net Positive Suction Head (NPSH).

The oiher answers are incorrect because I ) increasing SP Temperature by 5°F does not cross the -?foot curve, 2) decreasing Suppression Chamber Pressure by 2psig does not cross the -4 foot curve, and 3) increasing D W Pressure by 4 psig would increase Suppression Chamber pressure, moving you further awqfrom the -4 foot curve

Reference:

LGA-001, Revision 06 LGA-003, Revision 05 LGA-005, Revision 06 LGA-010, Revision 06 v

2003-01 ILT NRC Exam, Version: 5 Page 18 of 105

~

17. 29503 1 EAI.08 001129503 1/EA1.08/3.8/3.9/RO/MEMORY/400.00.02/

-~

(1;OO Point) When using the Standby Liquid Control System injection line as an alternate injection flowpath, the NSO can monitor proper operation by observing that running with discharge pressure greater than reactor pressure and SBLC squib valve continuity L

lights out.

A. TWO Makeup Condensate (MC) pumps are B. ONE Diesel Fire Pump (DFP) is C. TWO SBLC pumps are D!' ONE SBLC pump is

~

i~

Answer D is correct The referenced procedure directs starting one SBLC pump andpumprng water from the SBLC Tank or the Test Tank and injecting into the vessel through the normal SBLCflow path The other answers are incorrect. FP and MC are viable distracters because the procedure directs these systems as makeup to the SBLC Storage Tank and to the SBLC Test Tank when SBLC is lined up for alternate injection. It the examinee knows the source of water but doesn't know the lineup then they wouldpick either of the two answers.

Reference:

LGA-SC-102, Revision 01, page 4, Step C.5.4 andpage 5, Step C.5.5.c 2003-01 ILT NRC Exam, Version: 5 Page 19 of 105

- - 18. 295037 EA2.0 1 00 ll295037EA2.0 1

~ /4.2/4.3/RO/MEMORY//

~~~

-~

I (1.OO Point) While operating at full power in a normal electrical 1ineup;an event has occurred on Assuming no operator action, how would you determine if Reactor Power is above 3%?

~

D. White APRM Downscale lights are OUT.

~

~~~

~

~

~

Remove all occurrences of the parenthetical statement "(75 on IRMrange 8)" from the LCA-010 handout.

Answer C is correct. Per LGA-010power is less than 3% iflRMs are reading 75 on Range 8 The other answers are incorrect. With both RPS busses de-energized the APRMs are de-energized and the white downscale lights are expected to be out. Also, both RPS busses de-energies will isolate the MSIVs and therefore any Turbine Bypass Valve that is open is either stuck or has been manually opened using the jack. Finally, the SPDS displqv does no/ indicate reactor power in the IRA4 or SRMranges.

Reference:

LGA-010, Revision 06, Override under step 6, and level bandstep 9 2003-01 ILT NRC Exam, Version 5 Page 20 of 105

19. 295038

~

~~

G2.1.14 001/295038/2.1.14/2.5/3.3/ROIMEMORYN

~-

(1.OO Point) The lowest E-Plan event classification level that REQUIRES activation of the i

assembly siren based on off-site release rate with no danger to on-site personnel is...

1 A. an Unusual Event.

B. an Alert.

C!' a Site Emergency.

D. a General Emergency.

Answer C is correct. An assembly is required after a Site Area Emergency or General Emergency has been declared The other answers are incorrect. Assembly should be considered when an Unusual Event or Alert has been declared.

Reference:

EP-MW-I 13-100, Revision 01, page 2, Note at Step 4.1, andpage 3, Note at top ofpage 2003-01 ILT NRC Exam, Version 5 Page 21 of 105

20. 600000 AKI.01 001/600000/AK1.01/2 5/2.8/RO/MEMORY/125.00.08/

~

( 1.OO Point) The control roomreceived a report of an electrical fire in the 1 B Diesel Geneitor i-Room.

The fire should be classified as a ( 1 )

consider when entering the room is (2) type fire, and the immediate hazard to

~

in the area.

A? (1) Class C; (2) suffocation due to C02 B. (1) Class B; (2) suffocation due to C02 C. (1) Class C; D. (1) Class B; (2) electrical shock due to water (2) electrical shock due to water

~~~

~-

- ~~~

Answer A is correct. An electricalfire is classified as a Class C type fire (Class B is flammable Guids) and the Diesel Generator Rooms are protected by C02flood systems that automatically initiate.

Reference:

LOA-FP-101, Revision 06, page 25, warning at top ofpage System Description 125, page 20,Section III.M. 4.a. I) 2003-01 ILT NRC Exam, Version: 5 Page 22 of 105

W

21. 2950 14 AA 1.07 001/295014/AA 1.07/4.0/4.1/RO/MEMORY/434.00.01/

~

~

~

(1.OO Point) Which one of the following actions are taken by the control room operators, to PREVENT cold water injection during an ATWS?

I A. Rapidly lowering level to lower reactor power during an anticipated transient without scram event.

BY Holding the hand switches for l(2)A and l(2)B RHR injection valves in CLOSE while RPV pressure decreases through the automatic open permissive setpoints.

C. Using only systems that inject inside the core shroud during an anticipated transient without scram event.

D. Injecting with Alternate ATWS Systems to restore and maintain RFV level on the Wide Range level instruments.

L-Answer B is correct. Thejrst action step in LGA-010 directs inhibiting ADS andpreventing injection fiom HPCS,-

LPCS, and LPCI. When the 505psig interlocks clear, the l(2)A and l(2)B LPCI injection valves will realign for injection. You can prevent the injection valves from opening by holding the injection valve control switches in CLOSE as you pass through the 505 psig interlocks.

The other answers are incorrect. Onb systems that inject OUTSIDE the core shroud should be used (FW, SC, CRD, RCIC, etc.). When using Alternate A TWS Systems you ARE intentionally injecting cold water, therefore the use of Alternate A TWS Systems does not prevent cold water injection. Step 9 of LGA-010 starts with a caution that warns you that injecting too fart while in Step 9 can damage the core.

Reference:

LGA-010 Lesson Plan, pages 4 and 5,Section I V. B. I

~ 2, and 3 L

2003-01 ILT NRC Exam, Version. 5 Page 23 of 105

22. 295020 AA2.05 001/295020/AA2.05/3.6!3.6/RO/HIGW/

(1.OO Point) Following a reactor water level transient on Unit-2 the following conditions exist=

- Reactor automatically scrammed on low reactor water level

- All control rod are full-in

- RCIC is in standby

- The Emergency Diesel Generators are in standby The following valves indicate CLOSED on both the PCIS Status CRTs and the control panels:

- 2B33-FO 19, RR Sample Inboard Isolation

- 2B33-FO20, RR Sample Outboard Isolation

- 2E12-F040A, 2A RHR Heat Exchanger Blowdown Downstream Isolation

- 2E12-F040B, 2B RHR Heat Exchanger Blowdown Downstream Isolation

- 2E 1 2-F049A, 2A RHR Heat Exchanger Blowdown Upstream Isolation

- 2E12-F049B, 2B RHR Heat Exchanger Blowdown Upstream Isolation 7-All other PCIS valves are in their normal, expected positions based on the transient.

Based on the valve positions, a spurious PCIS isolation has occurred.

A. Group2 BY Group 3 C. Group4 D. Group7

~~~

~

~

~ _ _

Answer B is correct. Reactor water level did NOT drop below the Level-2 setPoin[ because RCIC and the EDGs are-still in standby. PCIS Group 7 isolates on Level-3 and should be isolated because the reactor scrammed on low Iwel (Level-3). Therefore it can be determined that level dropped below Level-3 but did NOT go below Level-2. Any isolation that occured based on RPV level dropping below Level-2 would be spurious. PCIS Groups 2, 3. and 4 all isolate on Lwel-2. Valves 2B33-FO20 and 2B33-FOI9 are PCIS Group 3 valves and have spuriously isolated.

\\-

Reference:

LOA-PC-201. Revision 09, page I6 2003-01 ILT NRC Exam, Version 5 Page 24 of 105

23. 295022 G2 1.23 001/295022/2.1.23/3 9/4.O/RO/HIGW049.00.14/

~~~

~

(1.OO Point) Unit-1 is pulling rods forastartup. The reactor is critical below the Point of Adding Heat (POAH) when the following annunciators are received:

- 1 H 13-P603-A303, CRD Feed Pump Suction Filter DP Hi

- 1H 13-P603-A103, 1A CRD Feed Pump Auto Trip

- 1 H 13-P603-A204, CRD Charging Wtr Press Lo The NSO observes CRD Charging Header pressure at 1000 psig and pressure has decreased 30 1

psig over the last minute.

I Based on the above information, the NSO's first priority is to...

I I

1 A. dispatch an NLO to determine the cause of 1A CRD pump trip per LOR-lH13-P603-A102, 1A CRD Feed Pmp Auto Trip.

BY insert a manual scram per the scram hardcard and LGP-3-2, Reactor Scram.

C. dispatch an NLO to swap CRD Suction Filters per LOP-RD-14, CRD System Pump Suctior Filter Replacement.

D. start 1B CRD pump per LOA-RD-101, Control Rod Drive Abnormal.

Answer B is correct. Scram setpoint for low CRD pressure is 1124psig with the mode switch in either REFUEL or STARTUP. Reactor is operating in Mode 2 with the mode switch in STARTUP because control r o h are being withdrawn andpower less than that required to transfer the mode switch to RUN. Based on mode switch position, the scram setpoint for CRD pressure has been exceeded and there are no indications that the reactor has automatical[y scrammed.

The other answers are incorrect. LOA-RD-IO1 does not cover a tripped CRDpump. Although suction filters mqV need swapped and the cause of the pump trip need to be determined, the highest priority is the scram setpoint being exceeded and failure to scram

~~~

~-~.~

Reference:

LOP-AA-03, Revision 19, page I3 2003-01 ILT NRC Exam, Version: 5 Page 25 of 105

u

24. 295032 EK1.04 001/295032/EK1.04/3.1/3.6/RO/HIGW/

(1.OO Point) Which one of the following would explain an increase in HCUyiessure with no accumulator trouble alarm?

A. The nitrogen side of the accumulator is leaking through the fill connection.

B. CRD Pressure Control Valve was throttled closed.

C. Charging water leaked past the accumulator seals.

D? Reactor Building Area temperatures are approaching Max Normal.

~

~~~

~~

Answer D is correct. As Reactor Building (RB) temperature increases above normal, the nitrogen in the accumulator expands at a predetermined rate (-25psig for evety 20°F increase in RB Temperature).

The other answers are incorrect. If charging water leakedpast the accumulator seal you would get an accumulator alarm on high level. The other two choices are incorrect because throttling the CRD Pressure control valve closed will increase Charging Water pressure but the accumulator piston is already at its full chargedposition and will not move any more therefore pressure will not change while a4usting the FCV. Leaking nitrogen from an accumulator would cause a low accumulator pressure alarm and would not cause HCU pressure to increase.

Reference:

LOP-RD-IO, Revision 16, page II, Attachment A LGA-002, Revision 03 2003-01 ILT NRC Exam, Version: 5 Page 26 of 105

25. 295002 EK2.06 001/295002/EK2.06/2.6/2.7/RO/HIGH/O80.00.18/

_ ~.

(1.OO Point) Unit-1 is operating-near full power with two Circulating Water (CW) Pumps operating. Condensate Polisher (CP) inlet temperature is 128°F.

Differential pressure across one of the operating traveling screens increases to 16 inches of water.

This would cause CP inlet temperature to...

- 1 I

A. remain at 128°F because Turbine Hood Spray would automatically initiate.

BY increase above 128°F as condenser vacuum decreases.

I C. remain at 128°F because Total Steam Flow would NOT change.

D. decrease below 128°F as condensate depression decreases.

I-Answer B ;correct.

Vacuum decreases when a CWpumps trips causing a decrease in condensate subcooling (condensate depression). Less subcooling of the condensate would cause an increase in condensate temperature.

The other answers are incorrect. Turbine Hoodspray has no eflect on condensate depression.

Reference:

LOP-CW-05, Revision I I, page 2, step D. I System Description I I I, page 31, Section B.2.

LGP-2-1, Revision 64, page 45, note at top ofpage

-v-2003-01 ILT NRC Exam, Version: 5 Page 27 of 105

- 1 i

\\

26. 295035 EK3.02 001/295035/EK3.02/3.3/3.5/RO/HIGW118.00.05N

~

~-

(1.OO Point) Unit-1 is operating at full power with Reactor Building Ventilation (VR) inthe following lineup:

- 1B VR Exhaust Fan OFF

- Two Unit-1 VR Supply and two Unit-1 VR Exhaust Fans are running

- Unit-2 VR is shutdown and no Unit-2 VR fans are running

- 1C VR Supply Fan OFF Assuming NO operator actions are taken, how will the VR System respond to a trip of the 1 A VR Exhaust Fan?

A. All running VR Supply and Exhaust Fans will trip to OFF on high building differential pressure.

BY 1 A and 1 B VR Supply Fans will cycle on high reactor building differential pressure.

C. 1 C VR Exhaust Fan will cycle on low reactor building differential pressure.

D. 1A VR Supply Fan will cycle on high reactor building differential pressure and 1 B VR Supply Fan will continue to run.

~

Answer B is correct. The fan trip unbalances the supply versus exhaust airflow. With an extra supply fan running, the building will pressurize until pressure reaches the high drflerential pressure trip (+2.0 inches H20), then the supply fans will trip. Without operator intervention (Supply Fan hanakwitchs remain in normal-after-start) when building pressure decreases, then the VR Supply Fans will restart. The building will re-pressurize and the Supply Fans will trip. The cycle will repeat until an additional exhaust fan is started or the supply fan handswitches are removed from normal-afier-start.

Reference:

LOP-VR-02, Revision 25, step 0. 4 System Description 118, page 21, 26 and 28 L

Thursday, January 27, 2005 7:46:06 AM 28

27. 295036 EA 1.02 001/295036EA 1.02/3.5/3.6/RO/MEMORY/418.00.0 11 Point) Therounds NLO reports that the Unit-2 Northwest Reactor Building corner room sump is overflowing due to a blown pump shaft seal in the northwest corner room.

Closing the suction isolation valve on which one of the following pumps would most likely isolate the leak?

I 1

A!' Division 1 Residual Heat Removal (RHR) B. Reactor Core Isolation Cooling (RCIC)

I I

C. Low Pressure Core Spray (LPCS)

D. High Pressure Core Spray (HPCS)

-~

~~~~-

~-

Answer A iscorrect The following pumps are located in the reactor building corner rooms:

Northwest = A RHR Northeast = LPCS and RCIC Southeast = B and C RHR Southwest = HPCS and both CRD

Reference:

LOP-RE-OIT, Revision 12, page 5 System Description 064, page 16, Section F.

2003-01 ILT NRC Exam, Version. 5 Page 29 of 105

28. 203000 K2.0 I..

00.

1 /203000/K2.0 113.5/3.5/RO/MEMORY/OO5.00.05E/

~

(1.OO Point) The following conditions exist-on Unit-1 :

- All ECCS systems are lined up for STANDBY An ECCS initiation signal occurs concurrently with Bus 142Y undervoltage.

When Bus 142Y re-energizes, the (1) approximately 5-seconds later.

I o

YO p ower, steady state I

~

starts immediately and (2) starts A!' (1) 1C RHR Pump; B. (1) LPCS Pump; C. (1) 1B RHR Pump; D. (1) 1A RHR Pump; (2) 1B RHR Pump (2) 1A RHR Pump (2) 1C RHR Pump (2) LPCS Pump I

Answer A is correct When Bus 142Y is re-energized, the IC RHR Pump will start ~mmed~a~ely, and then I B RHR Pump will start following a 5-second time delay to prevent over loading the 1A Diesel Generator The other answers are incorrect. When Bus 141 Y is re-energized, the LPCS will start immediately, and then IA RHR Pump will start following a 5-second time delay to prevent over loading the Common Diesel Generator.

Reference:

LOS-DG-110, Revision 01, System Description 005, page IS, paragraph 0.2

u.

2003-01 ILT NRC Exam, Version: 5 Page 30 of 105

~

29. 205000 K3.03 00 1/205000/K3.03/3.8/3.9/SRO/HIGH/023.00.05/

~~

~

~~~~

(1.OO Point) Unit-1 has been in Cold Shutdownfor 7 days following an extended run:

- 1 B RHR is operating in Shutdown Cooling with a stable suction temperature of 190°F

- 1A RHR Pump is 00s

- RPV water level is +145 inches

- 1A RR Pump is in slow speed

- 1B RR Pump is 00s Which one of the following describes the response of the Reactor Recirculation pump suction temperatures 15 minutes after a trip of the 1 B RHR pump (assume NO operator action)?

I 1A RR Pump 1B RR Pump Suction Temperature Suction Temperature A.

Remains relatively stable Remains relatively stable B.

Increase Increase CY Increase Remains relatively stable Remains relatively stable Increase I

I L..

~

~

~-

-.~

~

~~~~~

~~

~

Answer C is correct. The I A RR Loop would warm up with the increasing RPV Temperature due to decay heat. The I B Loop temperature should be at near drywell temperature and will remain relatively constant since its suction or discharge isolation valve is closed while operating in SDC to prevent short cycling the SDCJow. Therefore there is no flow in the I B RR Loop.

Reference:

LOP-RH-07, Revision 50, page 8 b

2003-01 ILT NRC Exam, Version: 5 Page 31 of 105

30. 209001 K2.03 001/209001/K2.03/2.9/3.1/RO/HIGH/063.00.14/

~~~~

I (1.OO Point) Given the following initial plant conditions:

- Low Pressure Core Spray (LPCS) was running for a surveillance.

- Full Flow Test Valve, lE21-FO12 was throttled to achieve rated pump discharge pressure

- Injection Valve, 1 E2 1 -F005 was closed

- Minimum Flow Valve 1 E2 1 -FO 1 1 was closed Subsequently a loss of Division 1, 125 vdc occurs, followed by Drywell pressure increasing to 2.0 psig.

Assuming no operator actions have been taken, what is the final status of the following LPCS components?

Full Flow Test Minimum Flow Valve 1 E2 1 -FO 1 1 Valve 1 E2 1 -FO 12 A.

Closed Open BY Throttled Closed C.

Closed Closed D.

Throttled Open

~-

~~~

~

~~

Answer B is correctxoss of Division I, 125 vdc (I I I Flow Test valve (IE21-FO12) will not receive an automatic close signal and will remain open (in its currently throttledposition). With adequate flow passing through the fuilflow test valve r> 750 gpm) the Minimum Flow valve (IEZI-FOI I ) will remain closed.

will result in loss ofpower to the LPCS logic. The Full

Reference:

System Description 063, page 8, section F. andpage 20. power supply table W

2003-01 ILT NRC Exam, Version: 5 Page 32 of 34

-I

31. 209001 K6.03 001/217000/K6.04/3.5/3.5/RO/HIGH/032.00.05D/

The Reactor Core Isolation Cooling (RCIC) system is running for a surveirance in the Condensate Storage Tank (CST) to CST test mode when a vehicle drives into the side of the CST leaving only 2 feet of water in the tank. Suppression Pool (SP) level is normal.

The RCIC turbine will continue to run...

I I

A. in the CST to CST test mode.

BY however the RCIC minimum flow valve will automatically open.

C. however RCIC will re-align to the SP to SP test mode.

D. however RCIC will re-align and pump the SP to the CST.

~

Answer B is correct. When CSTlevel drops below 3 feet, the Suppression Pool Suction valve (IES~F031) will automatically OPEN and then the CSTSuction valve (IESI-FOIO), the CST Full Flow Test valves (IESI-F022 and IESI-FO59) will CLOSE. When IESI-FO22 and IE51-FO59 close, RCIC will not have afloupath and the Minimum Flow valve (IESI-F019) will automatically OPEN.

Reference:

System Description 032, page 9, Section C.4.a and C.4.c.

System Description 032, pages 16 and 17, Section 3.a.2)

IE-]-4226AE, AS, and A W W

2003-01 ILT NRC Exam, Version: 5 Page 33 of 35

~~

32. 209002 K1.03 001/209002K1.03/3.0/3.O/RO/MEMORY/061.00.05G/

-~

~~

(1.OO Point) DEGRADATION of the HPCS Water Leg pump will...

I I

I A? increase the probability of water hammer.

B. degrade HPCS Pump seal cooling flow.

C. decrease the HPCS Pump suction pressure.

1 D. reduce the response time for HPCS injection.

I

~~~

~

~-

-~

Answer A is correct The HPCS water leg pump takes a suction from the Suppression 601 anddrscharges into the HPCS Pump discharge header. The discharge header needs to remain full to allow HPCSflow to quickly reach the reactor. A drained header delays HPCS delivery and may cause water hammer when the HPCS pump starts.

Reference:

System Description 061, pages 4 and 8 2003-01 ILT NRC Exam, Version: 5 Page 34 of 105

33. 209002 AI.08 001/209002/A1.O8/3.1/3.3/RO/HIGH/061.00.05/

~~

~

(1.OO Pornt) An event has occurred on Unit-1 resulting in the following conditions:

I

- Drywell pressure has increased to 1.O psig

- All ECCS pumps have automatically started Subsequently RPV level reached +66 inches and then decreased to -1 0 inches.

Which one of the following actions would cause the HPCS system to inject water into the RPV?

A. Turn the injection valve handswitch to OPEN and THEN turn the minimum flow valve hand switch to CLOSE.

BY Depress the HPCS High Water Level reset pushbutton, then turn the injection valve handswitch to OPEN.

C. Depress the HPCS High Drywell Pressure reset pushbutton, then turn the injection valve handswitch to OPEN.

D. Arm and depress the HPCS System Manual Initiation pushbutton.

~~

Answer B is correct. After automatic initiationTthe injection valve will automatically close when RPV Level-8 is reached (59.5 inches). The high level (Level-8) signal seals itselfin until either the low level initiation is reached (Level-2) or the High Water Level reset pushbutton is depressed. gthepushbutton is used. the injection valve will not aiitomatically re-open until level reaches the low level (Level-2) setpoint, therefore the operator will have to manually open the injection valve (after resetting the high level trip). The other answers are, therefore, incorrect.

Reference:

System Description, page 2 5 Section E.

L 2003-01 ILT NRC Exam, Version: 5 Page 35 of 105

~

34. 2 1 1000 A2.0 1 001/211000/A2.01/3.5/3.8/RO/HIGH/028.00.2 I /

(1.OO Point) During an ATWS, the NSO started the 1 A Standby Liquid Control (SBLC) system per the Scram Hardcard and verified proper operation of SBLC.

After pumping 1000 gallons of boron solution, Bus 14 1 Y tripped on overcurrent.

Which one of the following actions should the NSO take, and why?

A. Verify both SBLC system handswitches are in STOP per LGA-010, Failure to Scram

~

sv because cold shutdown boron has injected.

BY Inject boron using 1 B SBLC system per LGA-SC-10 1, Initiation of Standby Liquid Control, because the 1A SBLC pump is lost.

C. Insert rods per LGA-NB-01, Alternate Rod Insertion, because neither SBLC system is I

available.

I I D. Lineup and inject boron per LGA-RT-103, Alternate Boron Injection Using RWCU, because neither SBLC system is available.

~~

~

Answer B IS correct. Loss of Bus I41 Y will cause the IA SBLC Pump to lose power and the operator should continue to inject boron per LGA-SC-IO1 because LGA-010 has not been exited.

The other answer are incorrect. The IB SBLC System (andpump) are available. The SBLC Tank must be pumped down below a level of 3050 gallons in order to say cold shutdown boron has been injected. There are approximately 4950 gallons in the tank when it is&ll. 4950 minus 1000 gallons that were pumped, equals 3950 gallons left in the tank. You need to pump a minimum of 900 more gallons of solution to reach cold shutdown boron L

injection.

Reference:

LGA-SC-101, Revision 01, page 3, Srep C.3.4 LGA-010, Revision 06 2003-01 ILT NRC Exam, Version: 5 Page 36 of 105

Based on the above MSIV closures, RPS relays in...

A. both channels should have de-energized causing a full scram.

B. both channels should remain energized.

L-Answer D G correct Using the BADC-CADB logic the following combinations of Inboard OROutboard MSIV closed will cause a halfscram BA, DC, CA or DB LOS-RP-Q3 explains the logic in a note as follows:

NOTE: Valve limit switches are connected to the RPS logic as per the folIowing table. "AD" and "BC" combinations will not give a halfscram.

Valves I B2 I -F022x and IB2I-FO28x A

B C

D X

X X

X X

X X

X

Reference:

LOS-RP-Q3, Revision 13, Attachment IA.

IE-I-421jAA, AC, AD, AE, AF, AH, andAL v

2003-01 ILT NRC Exam, Version: 5 RPS Su bchannels A-I A-2 B-I B-2 (A+B = channel A-I tripped)

(C+D = channel A-2 tripped)

(A+C = channel B-1 tripped)

(B+D = channel B-2 tripped)

Page 37 of 39

36. 215003 A4.04 00112 15003/A4.04/3.1/3.3/RO/HIGH/042.00.14/

(1.OO Point) Unit-1 is performing a startup. The reactor is critical below the point of adding heat.

Using the figure below, which one of the following lights are expected to be lit for the conditions provided?

IRMCHANNELB 1

I lC51-K601B 1

A. BOTH Upscale Alarm AND Upscale High B. Inop ONLY CY BOTH Inop AND Upscale Alarm D. Upscale Alarm ONLY

~~

~

~

~

Answer CTs correct The STANDBYposition of the IRMSelector switch has the samefirnctions as when; OPERATE with the exception of inserting an INOP trip Therefore the INOP light is lit With power approximately 90/125s of scale, the UPSCALE ALARM is lit and not the UPSCALE HIGH trip should be in until 120/12js of scale.

Reference:

System Description 042, paze 13, first paragraph 2003-01 ILT NRC Exam, Version: 5 Page 38 of 105

37. 215004 G2.1.28 001/215004/2.1.28/3.2/3.3/RO/MEMORY/300.010/

~.

(1.OO Point) During a reactor startup, the NSO can more closely monitor the SRMs by...

A. selecting the SPDS computer screen.

B. change the indicating range.

I CY changing the recorder chart speed.

D. adjusting the discriminator voltage.

I

~

- ~. -~

Answer C is correct. During a Reactor Startup, on the SRM dual pen recorders, SELECT the highest indicating SRM for each operable channel A or C and B or D on recorder 1(2)C51-R602.... Place the SRM recorder in fast speed and LABEL charts with the TimdDate of going to fast speed, and Channels selected The other answers are incorrect. Only lRMs have selectable ranges. Operators do not adjust the discriminator voltage. There are no SRM indications associated with the SPDS computer screen.

Reference:

LOP-NR-01, Revision 12, page 5, Step E.2.

L 2003-01 ILT NRC Exam, Version: 5 Page 39 of 105

38. 2 15005 K1.07 00 1/2 l5005/K1.07/2.6/2.9/RO/HIGH/044.00.

I8C/

~-

(1.00 Point) As the Local Power Range Monitor (LPMYdetectors ag<the output of the Average Power Range Monitors (APRMs) will

[Assume initial APRM Gain Adjustment Factors (AGAFs) were equal to 1.OO and actual power remains the same as the LPRM ages.]

A. INCREASE and the AGAFs will be LESS than 1.OO I

BY DECREASE and the AGAFs will be GREATER than 1.OO C. DECREASE and the AGAFs will be LESS than 1.OO i

I D. INCREASE and the AGAFs will be GREATER than 1.OO L

-~

~.

Answer B is correct. As U-235 depletes in the detector, there are fewerfissions for a given neutron fluxfield, resulting in a decreased detector output. This decreased detector output will result in a lower indicatedpower.

AGAF = (Actual Power)/(indicated Power). Therefore, if indicatedpower decreased and actual power stayed the same, the AGAF values will be increasing.

Reference:

System Description 043, page 7 System Description 044, page 27 W

2003-01 ILT NRC Exam, Version: 5 Page 40 of 105

39.215005 A4.04 00 112 15005/A4.04/3.2/3.2/RO/MEMORY/043.00.071

~

(I.OO Point) Unit-1 is operating at 15% RTP following a reactor startup. During performance of the control room rounds, it is discovered that the AMBER light for the LPRM labeled 5D-40-25 is lit on back panel 1 H 13-P608.

Which one of the following would cause the above indication?

A. The output signal from the LPRM reading zero.

B. The LPRM mode switch in the BY (bypass) position.

C. The LPRM mode switch in the CA (calibrate) position.

DY The output signal from the LPRM reading upscale.

~

~

~

~~

Answer% is correct The amber Iight associated with each LPRM indicates an upscale condition Highjlwc is the only one of the conditions that causes an upscale trip There are three lights associated with each L P M. From left to right, they are:

I ) White - Bypassed

2) White - Downscale

3) Amber - Upscale Therefore, the other answers are incorrect.

Reference:

System Description 043, page 16, Section B. 1.a.

L-2003-01 ILT NRC Exam, Version: 5 Page 41 of 105

40. 2 17000 K2.0 1 00 1 /2 1 7000/K2.0 1/2.8/2.8/RO/MEMORY/032

.OO I 6/

~ _

~

(1.OO Point) How would a loss of 250 vdc bus 12 1 Y affect the Reactor Core Isolation Cooling I

(RCIC) system?

RCIC (1) isolation signal.

A. (1) would B. (1) would NOT C!' (1) would NOT; D. (1) would automatically start following an automatic initiation signal, and u

~

(2) automatically isolate following a PCIS high RCIC room temperature I

I (2) would (2) would NOT (2) would (2) would NOT i-

~-

Answer C is correct. 121 Y supplies the components that may need to automatically reposition for automatic initiation function to work. The PCIS valves arepoweredfiom 136Y-2 (IE5l-FO63) and 135X-1 (lE51-FO08).

Therefore the automatic initiation finction will NOT work however the PCIS function is still operable.

Reference:

LOA-DC-101, Revision 07, page 202, Attachment i21Y Thursday, January 27, 2005 I I.48:35 AM 42

41. 2 18000 K3.01 001/2 18000/K3.0 1 /4.4/4.4/RO/HIGW062.00.08/

(1.OO Point) An event has occurred on Unit-1 resulting in the following conditions:

- Bus 112Y is de-energized

- Division 1 ADS High Drywell pressure transmitters are inoperable and transmitting 0 psig

- Actual Drywell pressure is 2.0 psig

- RFV pressure is approximately 450 psig and steady RPV Level has just reached the Level-I setpoint.

Low pressure ECCS systems will start to inject...

A. in less than 1 minute.

B. in 9 to 10 minutes.

C. in 2 to 3 minutes.

D!' in 11 to 12 minutes.

~

~~~~

Answer D is correct. Division 2 ADS will not initiate because Bus I I2Y is dead. Division I ADS should initiate in IO5 seconds however the Drywell pressure transmitter is inoperable so the 9 minute timer starts, when it times out the 105 second timer will start. This give a total time delay of 645 secondr (IO. 75 minutes). Then after -I I minutes Division l ADS will initiate. Shortly afler, UPVpressure will drop from 45Opsig to less than LPCSshutofJhead

(-440 psi@ at which time LPCS will start to inject. The correct answer is just under I I minutes for ADS to initiate and then less than one minute for RPVpressure to drop below 440psig (I 1-12 minutes).

Reference:

LOP-MS-03, Revision 06, page 4, Step D.5

+.

2003-01 ILT NRC Exam, Version: 5 Page 43 of 45

42. 218000 A3.02 001J2 18000/A3.02/3.6/3.7/RO/HIGW070.00.05D/

LGA entry conditions exist on Unit-1 with the following conditions:

I

- reactor pressure = 790 psig

- Drywell pressure = 5 psig

- Bus 1 1 1 Y is de-energized After depressing all four ADS Manual Initiation pushbuttons, you take the following SRV Tailpipe temperature readings from the appropriate back panel recorders:

SRV Temperature SRV Temperature SRV Temperature C

3 19°F K

260°F R

260°F D

3 17°F L

258°F S

3 18°F E

315°F M

265°F U

3 17°F F

255°F P

350°F V

272°F H

350°F Based on the above tailpipe temperatures, how many ADS valves are open?

A. 7 ADS valves are open.

B. 6 ADS valves are open.

CY 5 ADS valves are open.

D. 4 ADS valves are open.

~

~~

~~~

Provide Steam Tables to the examinee as a handout.

Answer C is correct. Expansion through the SRVs is isenthalpic. Using the steam tables, f y o u start at 805 psia (1200 BTUsAbm) then side horizontally (constant enthalpy line) to the right, stopping at 20 psia. You will stop on just below the 320°F line. Therefore any tailpipe near 320°F represents an open SRV. There are 5 SRVs open using the above readings, all 5 are ADS SRVs.

The other answers are incorrect because their temperatures are too low for an SRV that is full open. Two non-ADS SR Vs (H and P) have failed readings, they are reading too high for isenthalpic expansion of saturated steam at 800 psig.

Reference:

Steam Tables L-2003-01 ILT NRC Exam, Version: 5 Page 44 of 46

43.

223002 K4.01 001/223002/K4

~ ~- 01/3.0/3.2/RO/HIGW091

.OO 01/

(1.OO Point) While operating-at 1 OO%, level indicating switch 1B2 1 -N704C which supports PCIS Group 2 and 3,-has failed downscale, This particular switch is powered from RPS A.

L Subsequently a leak in the Drywell has resulted in RPV level dropping to -1 00 inches. Level then returned to the normal control band.

Based on the above event, isolate.

PCIS Group 2 and 3 valves will A. NEITHER Inboard NOR Outboard B. ONLY the Outboard C. ONLY the Inboard DY BOTH Inboard AND Outboard

~

Answer D is correct. Redundancy and physical separation are required in the electrical and mechanical design to ensure that no single failure in the system prevents the system fiom performing its safety function. PCIS Groups 2 and 3 initiates at Level-2 (approximately -48 inches).

Reference:

LOA-PC-IOI, Revision 07, pages 16, 17 and IS, Attachment A (hardcard)

LSCS-UFSAR page 6.2-54 Technical Specifications Appendix G L-2003-01 ILT NRC Exam, Version: 5 Page 45 of 105

44.

239002 K5.05 00 1/239002/K5.05/2.6/2.9/RO/MEMORY/O7O.OO.O5C/

(1.OO Point) Which one of the following would be an indication that an SRV 21 pipe has sheared at the T-quencher and the T-quencher is no longer performing as designed?

A? Greater than normal localized heating at the SRV tail pipe discharge.

B. Containment Vacuum breaker cycles each time the SRV opens.

C. Greater than normal SRV tail pipe temperature each time the SRV opens.

I I

D. Large spike in Drywell pressure each time the SRV opens.

i

~

~

-~

Answer A is correct. One end of the T-quencher has drilled holes that allow the steam to escape and are positioned so the discharge creates a swirling efect in the suppression pool. The swirling eflect promotes better mixing ofthe steam and pool water. The arrangement minimizes localized thermal stress from a single valve discharge.

The other answers are incorrect because the steam would still condense and a pressure spike would be NOT be created in either the Suppression Chamber (SC) or Drywell (OW). Because there would be no pressure spike, SC pressure would remain less than D Wpressure and the Containment Vacuum Breakers would NOT open.

Reference:

System Description 070, pages 7 and 8,Section III.C.

2003-01 ILT NRC Exam, Version: 5 Page 46 of 105

45. 259002 K6.02 00 1/2590021K6.02/3.3/3.4/ROIMEMORYN (1.OO Point) Unit-] is at 100% power with both TDRFPs in 3-element control. The level s e t p z is set at +36 inches.

L A fuse blows de-energizing 1 A Narrow Range (NR) Level transmitter, 1 C34-NO04A.

Assuming NO operator action, the reactor...

~

I I

A. will scram because the main turbine will trip on high level.

BY will NOT scram because the RWLC System will ignore the 1A NR level indication.

I C. will scram because level will decrease to the low level scram setpoint.

I D. will NOT scram but power will be reduced due to an automatic RR Pump downshift.

I i Answer B is correct Ifany level signal deviatesfiorn the majority value it will automatically be drsconnected and an event message will be displayed on the IDSOOI Operator Station. Disconnection will be unnoticeable

Reference:

System Description 040, page 34, Table System Description 03/, page 12,Section III.A.

LOP-FW-I6, Revision IS, page 34, Attachment 2 u

2003-01 ILT NRC Exam, Version 5 Page 47 of 105

46. 261000 A1.02 001/261000/A1.02/3.1/3.2/RO/HIGW095.00.20/ ~

(1.OO Point) Standby Gas Treatment (VG) is being used to vent the containment. Drywell pressure is 1.4 psig and steady when the VG Train Flow Transmitter fails to 5000 cfm.

Based on the VG train automatic response to this failure, Drywell pressure will...

A. decrease because the VG Flow will stabilize at 5000 cfm.

BY increase because the VG Flow Control Damper will close.

C. increase because the VG Primary Fan will trip.

D. remain constant and VG Flow will remain at 4000 cfm.

j I

~~~~~~

~

~

i

~~~

~~~

Answer B is correct. 5OOOcfm is IO00 cfm greater than normal. The Flow Control Damper will close trying to controlflow at 4000 cfm. This will cause Dlywellpressure to start to increase because no flow will be leaving the Drywell. The other answers are incorrect.

Reference:

LGA-VQ-01, Revision 08, page 15, step E.3.d 6)d)

LOS-VG-MI. Revision 29, page I I, Attachment IA, step 2.5 System Description 095, page 7, top ofpage, page 17, table givesflow setpoint

.b 2003-01 ILT NRC Exam, Version 5 Page 48 of 105

47. 261000 A4.07 001/261000/A4.07/3.1/3.2/RO/HIGW095.00.05E/

~

~~

_ _ _ ~

(1.OO Point) Standby Gas Treatment (VG) has automatically initiated. Ten minutes later the NSO recognizes that the VG Outlet Isolation Damper (1 VG003) indicates intermediate.

Which one of the following would confirm to the NSO that the 1VG003 is NOT full open?

A. Reactor Building differential pressure is -0.5 inches H20.

B. The VG Flow Control Damper, 1VG002 indicates intermediate.

C. VG Train Flow indicates 4000 c h.

D!' The electric heater is de-energized.

..~.

Answer D is correct. rj-train airflow is NOTgreater than 3200 scfm then the electric heater will NOT start. r f f r w drops below 3200 cjn then the electric heater will trip. Therefore ifthe electric heater is off it is a good indication that there is inadequateflow through the VG train.

The other answers are incorrect. VG train normalflow is 4000 cfm. The I VG002 throttles to maintain 4000 cfm therefore dual indication on the I VG002 is normal. VG is designed to maintain less than -0.2.5 inches H 2 0 differential reactor building pressure. Differential pressure at -0.5 inches H 2 0 would indicate adequate SBGTJow.

Reference:

LOR-ZPM07-A401. Revision 02, setpoint System Description 09.5, page 7, Section C., page 17, Table b

2003-01 ILT NRC Exam, Version 5 Page 49 of 105

48. 262001 A2.02 001/262001/A2.02/3.6/3.9/RO/HIGWO11.OO. 14/

~

(1.OO Point) Unit-1 is shutdown with the UAT unavailable. Unit-2 is in a normal full power electrical lineup.

Subsequently a spurious trip of the SAT Feed to 14 1 Y (breaker 14 12) occurred.

The 0 DG is running loaded with the following breaker lineup per the AC Power Abnormal procedure LOA-AP-1 0 1 :

- SAT is available and carrying 142X, 142Y, and 143

- ACB 141 1, UAT Feed to 141X is OPEN (and NOT available)

- ACB 14 12, SAT Feed to 14 1 Y is OPEN (and NOT available)

- ACB 14 13,O DG Output is CLOSED

- ACB 14 14, Unit Tie Breaker is OPEN (and available)

- ACB 141 5,14lX/Y Tie Breaker is CLOSED 1

1 While operating in the above lineup, an ECCS initiation signal is received on Unit-1.

The operator should expect (1)

LOA-AP-101, Unit-1 AC Power System Abnormal, to (2)

A. (1) ACB 1415 to OPEN, de-energizing bus 141X and ACB 1413 to REMAIN CLOSED; and then enter the appropriate section of (2) RE-ENERGIZE 141X if Service Water pumps are required to supply fire protection.

B. (1) ACBs 1413 and 1415 to REMAIN CLOSED, there is NO undervoltage on Unit-2; (2) VERIFY all electrical buses are energized and the 0 DG is NOT overloaded C!' (1) ACBs 1413 and 1415 to OPEN, and then 1413 to RECLOSE on undervoltage; (2) RE-ENERGIZE 141X if Service Water pumps are required to supply fire protection.

D. (1) ACB 1413 to OPEN and ACB 1414 to CLOSE, fast transferring to the Unit -2 SAT; (2) VERIFY all electrical buses are energized and return the 0 DG to Standby

~- - ~ _ _ _ _ _ _ ~ ~ _ _ _ _ _

~

Answer C is correct. It is the onb choice which correctly describes breaker response and expectedprocedural actions The other answers are incorrect. An ECCSsignal will trip the DG output breaker (1413) ifclosed. When 1413 opens, the X-Y tie breaker (1415) will trip on undervoltage. There is NO fast transfer feature to off-site power through the unit tie breaker (1414).

Reference:

LOP-DG-02. Revision 37, pages 7 and S, Steps 0.4.3.1 and 4.3.2 System Description 01 I, page 49,Section V.C.2. b LOA-AP-101, Revision 19, page 2, TOC L'

2003-01 ILT NRC Exam, Version: 5 Page 50 of 105

49. 262002 A3.01 00 1/262002/A3.O 1/2.8/~.1/RO/MEMORY/012.00.05 J/

(1.OO Point) Which one ofthe following would indicate that the Plant Process Computer (PPC) 1 Uninterruptible Power Supply (UPS) should have automatically transferred to the Alternate Source?

L i

A!' INVERTER FAILURE light is ON B. 250 vdc battery supply voltage reads zero C. SYNC FAILURE light is ON

~

D. Static Bypass Switch INVERTER light is ON

~~~~~

.. ~ ~

~

~~

~

~~

~~

~

Answer A is correct. Loss of both the Normal AC Source and the Backup DC Source [or failure of the 25 KVA Inverter as indicated by the INVERTER FAILURE white light] will cause the Static Transfer Switch to swap to the Alternate AC Source.

The other answers are incorrect. The SYNC FAILURE light indicates that the UPS is unable to synchronize between the Alternate Source and the Inverter Output for Static Bypass Switch operation. The INVERTER light on the Static Bypass Switch indicates that the UPS is on the inverter (Normal or Backup supplies). On loss of 250 vdc Backup Supply the UPS will remain on the Normal Supply.

Reference:

LOP-CX-08, Revision 04, page 8, step E 2 System Description 050, page 37 System Description 012, Figure 12-03 Photo of PPC UPS 2003-01 ILT NRC Exam, Version. 5 Page 51 of 105

50.

- 263 000 A4.03 00 1 /263000/A4.03/2.7/2.8ROMIGW006.OO. 141 (1.OO Point) A loss ofALL 125 vdc Battery Chargers has occurred on Unit-1.

Assume equal loads of 25 amperes are being supplied by each battery, which one of the following lists the expected relationship between the Unit-1 battery voltages after 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />?

1

-i A. Division 1,2 and 3 battery voltages approximately equal.

B. Division 2 and 3 battery voltage approximately equal and greater than Division 1.

C!' Division 1 and 2 battery voltage approximately equal and greater than Division 3.

D. Division 1 and 3 battery voltage approximately equal and greater than Division 2.

I-

~

_ _ ~

~~~

Answer C is correct. The Division I and Division 2 125 vdc Batteries are rated at I128 amp hours. The smafier Division 3 125 vdc Battery is rated at 308 amp hours. Therefore it is expected that the Division 3 Battery voltage will decrease at a faster rate than the other two batterips.

At 25 amps per hour, the Division I and 2 batteries will last 45 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br />. Division 3 battery will last only 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Therefore the Division 3 battery voltage is dropping at almost 4 time as fast as the other 2 batteries.

Reference:

System Description 006, page 24, Section 1.a and 1.b u

2003-01 ILT NRC Exam, Version 5 Page 52 of 105

51. 264000 G2.1.30 001/264000/2.

I.30/3.9/3.4iRO/MEMORY/Ol l.OO.OS/_

r I

I (1.OO Point) You have been directed by the Unit Supervisor to place the 1B DG Maintenance-Auto Transfer Switch in the MAINT position.

1 Where is the switch located?

I A? In the 1B DG Room on Local Control Panel (1E22-P301B).

~

B. In 143 Switchgear room in the auxiliary compartment above the 1B DG Output Breaker.

I C. In the Main control room on panel lH13-P601.

I D. In the 1 B DG Room next to the 1 B DG Governor.

L ~-

~

~~~

Answer A is correct. The Maintenance-Auto Switch is located on the local control panel in the IBDG room (IE22-P301 B).

Reference:

LOP-DG-01, Revision 30, page 22, Step E.3.2 L

2003-01 ILT NRC Exam, Version. 5 Page 53 of 105

52. 300000 K1.05 001/300000/K1.05/3.1/3.2/ROfMEMORY/097.00.05W

~

~

~

(1.OO Point) With Unit-1 operating at-full power, a spurious PCIS Group 10 isolation occurred.

to be replaced.

Based on the above event, which one of the following describes the minimum actions required to ensure the Inboard MSIVs will remain open during the replacement of 1 IN22M?

L i

n g

the recovery, the NLO reports that the IN Rupture Disc, 1 IN22M has ruptured and needs L-I 1 NOTE:

1) 1INO59 - DW Nitrogen Inst Air Downstream Stop Valve

2) 1INO60 - DW Nitrogen Inst Air Upstream Stop Valve

3) 1INO17 - DW Inst N2 Regulated Hdr Drywell Supply Valve A. START 1 A IN Compressor and OPEN 1 IN0 17 ONLY B. OPEN valve 1 IN0 17 ONLY C. OPEN valves 1IN0.59 and 1IN060 ONLY D? OPEN valves 1IN059, 1IN060, and 1IN017 I

~~~~~~~~

~~

Answer D is correct. Following a PCIS Group I O isolation, the IN Suction Isolation valves close and the IN Compressors trip on low suction pressure. lIN059 and llN060 are opened to allow SA to supply pressure to the pneumatically operated IN isolation valves. To allow SA to supply the regulated IN header, llNOl7 must also be opened.

The other answers are incorrect. Without opening all three valves, SA can NOT supply IN to the MSIVs. You can NOT replace IIN22M with the IN Compressor in operation.

Reference:

System Description 097, page 28,Section X A. 3 and X A. 4 LOA-IN-101, Revision 05, page 13, Attachment I A

+

2003-01 ILT NRC Exam, Version: 5 Page 54 of 56

53. 400000 K2.01 001/400000/K2.01/2.9/2.9/RO/MEMORY/114.00.16/

(1.OO Point) Loss of bus 24 lYwill render inoperable due to loss of power.

A. both 0 and 2B RBCCW Pumps C!' ONLY the 2A RBCCW Pump B. ONLY the 2B RBCCW Pump D. both 2A and 2B RBCCW Pumps L - -

Answer C is correct. The RBCCW Pumps are powered as follows:

Power Suuulv Comuonents 241 Y via 233 242Y via 234X 142Y via 134X 2A RBCC W Pump 2B RBCC W Pump 0 RBCCW Pump

Reference:

System Description 114, page I I, Power Supply Table System Description 005, Figure 05-07 2003-01 ILT NRC Exam, Version 5 Page 55 of 105

-~

54. 201 003 K3.01 001/201003/K3.01/3.2/3.4/RO/HIGH/208.00.02/ -

~~

( 1.OO Point) Unit-1 icoperating at 60% reactor power on a 80% Flow Control Line.

I I I

A control rod is being positioned from position 10 to 08. While settling, it is discovered that a complete failure of the collet fingers has occurred.

I Based on the above failure and NO operator action, total core flow will...

A? decrease due to an increase in two-phase flow resistance.

B. increase due to a decrease reactor recirculation ratio.

C. remain constant since rod position does NOT affect total core flow.

D. increase due to a decrease in two-phase flow resistance.

L-

~~~

-.- ~~~~

i Answer A is correct. A colletjinger failure would result in the control rod drifting out @I1 out with NO operator action). Reactor Power will increase as result of the rod withdrawal. The increase power would cause increased boiling and a corresponding increase inflow resistance due to two-phase flow.

b 2003-01 ILT NRC Exam, Version 5

Reference:

LOA-RD-101, Revision 07, page 8, step B. 1.12 Page 56 of 105

55. 201006 K5.1 1 001/201006/K5.1 1/3.213 3/RO/MEMORY/048.00.14/

~

( 1.OO Point) A rod position being displayed in magenta on the RWM indicates that...

A. that control rod has a substituted value.

BY only that control rod can be withdrawn.

r-

- 1 I

C. only that control rod can be inserted.

D. that control rod is electrically out of service.

I.

...~. -~ -

~~~

~~

~~~~

~

Answer B is correct. An insert error is indicated by Magenta font used to indicate rod position on the R WM displqv.

When an insert error exists, an insert rod block is generated f t h e RWM is enforcing block (less than the LPSP or Blocks to Full). At LSCS R WM is normally operated in Blocks to Full mode. Therefore when control rod position is indicated in Magenta, the R WM will be generating an Insert Block.

Reference:

LOP-RW-OI, Revision 14, page 4, Step 0.2. I. ]

System Description 048, page 7, Section Ill. C.Z.b.4) 2003-01 ILT NRC Exam, Version 5 Page 57 of 105

-~

~~

-~~

56. 202002 K6.03 00 1 1202002/K6.03/2.8/2.8/R0/H1GW023.00.05 JI

~

( 1.OO Point) A seal failure on the 1 A Reactor Recirculation (RR) pumpdevelops while operating in Mode 1 with power <25%. This causes the following:

- Drywell pressure increases to 2.0 psig

- 1 A RR pump trips on overcurrent

- Narrow Range level indication remains on-scale during the event Which one of the following correctly states the expected loop flow controller indications for the RR Flow Control Valves (FCVs) following the event?

1A RR-FCV 1B RR-FCV Position Position A.

15% open Full-open B.

Full-open 1 5% open CY Full-open Full-open D.

15% open 1 5% open I

I I

I

~

~~

Answer C is correct While operating at 25%power, RR is in slow speed with the FCVs full-open The RR-FCVs will lockup on high drywell pressure so they would remain at their position PRIOR to the event Cfull-open) The other answers are therefore incorrect

Reference:

LOR-IH13-P602-A301, Revision 3 System Description 023, page 16, section V.A.4 w

2003-01 ILT NRC Exam, Version 5 Page 58 of 105

57. 2 I4000 K3.0 1 00 1/2 14000K3.0 1/3.0/3.2/RO/MEMORY/048.00.14/

~~

(1.OO Point) Following a reactor scram from full power, when the RWM Scram Mode is enabled, IF multiple rods are at unknown positions due to Rod Position Information System (RPIS j failures, and all other control rods indicate full-in, the box will say...

A? ROD UNKNOWN in a yellow box B. ROD OUT in a red box I

D. ROD UNKNOWN in a red box C. ROD OUT in a yellow box I-Answer A iscorrect. Ifrods are at unknown positions, but none are known to be beyond 00, the box will be colored yellow and contain the words "ROD UNIWOW. "

Reference:

LOP-RW-01, Revision 14, page 19, secondparagraph of note at tope ofpage.

\\-

2003-01 ILT NRC Exam, Version. 5 Page 59 of 105

~

-- I I

58.2 16000 ~ 1. 0 4 oo1~21~ooo~~1.o4/3.~~~.0/~0m1~~o4o.oo.14~

(1.OO PointjThe Unit-1 is at 100% power.

An NLO reports that an RPV Instrument Line is cracked and is slowly venting to the Unit-1 Reactor Building atmosphere.

The NSO scanned RPV pressure and RPV level instruments associated with the cracked instrument line.

Which one of the following sets of readings would indicate that the leak is associated with an instrument reference leg?

1 I

I I

Indicated Indicated RPV Pressure:

RPV Level:

AJ 980 psig 40 inches B.

1019 psig 34 inches C.

960 psig 32 inches D.

1015 psig 39 inches L

..~...

Answer A is correct. RJ'V Pressure instruments are connected to instrument reference legs (vice variable legs) and therefore would indicate lower than actual pressure ifthe reference leg is vented. Differential pressure across level instruments would decrease (or even reverse) causing level to indicate abnormally high. Therefore the correct answer is the only choice that has pressure less than normal (< IO05 psig) and level greater than normal (>36 inches). The other answers are therefore incorrect.

u

Reference:

System Description 040, page 42, Section C. 1. (2ndparagraph) 2003-01 ILT NRC Exam, Version 5 Page 60 of 105

59. 21 9000 A2.03 001/219000/A2.03/3.1/3.2/RO/HIGW/

~~

~

(1.OO Point) Unit-2, Division 2 RHR is running in Suppression Pool Cooling with flow established at 5000 gpm. 2E1 2-F048B7 RHR Heat Exchanger Bypass and 2E12-F003B7 RHR Heat Exchanger Outlet valves have been adjusted to mid-position in order to establish the required cooldown rate. The Suppression Pool cooldown rate has NOT required adjustment for approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Subsequently a short causes the 2E12-FO48B valve to travel to the full CLOSED position.

I j

\\

As a result of the above event, the Suppression Pool cooldown rate will (1) and the NSO must throttle the 2E12-F003B, RHR Heat Exchanger Outlet valve in the direction in order to re-establish the previous cooldown rate.

I (2)

A!' (1) increase; B. (1) decrease; C. (1) increase; D. (1) decrease; (2) CLOSED; (2) OPEN (2) OPEN (2) CLOSED

~~

~.

~

- 2 Answer A IS correct Per LOP-RH-I 3,Suppression Pool Cooling Operation, you are directed to ESTABLISH desired cooling by THROTTLING 2E12-FO48B and THROTTLING 2E12-FO03B as required When 2EI 2-FO48B closes, more flow is forced through the Heat Exchanger increasing the cooldown rate The operator will have to close the 2ElZ-FO03B to reduce the cooldown rate to the previous rate

Reference:

LOP-RH-13, Revision 26, pages 6 and 7, Step E.1.5 2003-01 ILT NRC Exam, Version: 5 Page 61 of 105

BY Initiation of Suppression Chamber Spray during a LOCA.

C.

D.

Initiation of Drywell Spray during a LOCA.

Opening an SRV with a tail pipe broken in the Suppression Chamber air space.

~

~

~

~-

Answer B is correct. The function ofthe Suppression Chamber-to-Drywell Vacuum Breakers is to relieve vacuum in the Drywell. They allow air and steam to flowfi-om the Suppression Chamber to the Drywell when the Drywell is at a negative pressure with respect to the Suppression Chamber. A negative diflerential pressure across the Drywell wall is caused by rapid depressurization of the drywell. Events that cause this rapid depressurization are cooling cycles, inadvertent Drywell Spray actuation, and steam condensationfiom sprays or subcooled water reflood of a break in the event of a primary system rupture.

The other answers are incorrect. Drywell Spray will rapidly condense steam in the drywell causing negative pressure on opening the vacuum breakers. Isolating a steam leak in the Drywell will allow steam to condense on the cooler drywell wall and other components causing drywell pressure to decrease (although not as rapidly as caused by dtywell spray). Opening an SRV with a broken tailpipe in the Suppression Chamber will cause the Suppression Chamber to pressurize without pressurizing the Drywell. This will also result in drywell pressure being negative with respect to the Suppression Chamber and the vacuum breakers will open.

Reference:

System Description 090, page 21,Section VIII.E.5 and VIII.E. 7, page 23, Section VIIi.G

'u 2003-01 ILT NRC Exam, Version. 5 Page 62 of 105

61. 23900 1 A4.02 001/239001/A4.02/3.2/3.2/RO/HIGW091

~~~

.OO. -

16/

~

(1.OO Point) Unit-1 is operating in Mode 1, After receiving annunciator 1 H 13-P603-A508, 1A RPS MG Set Trouble, the following valves indicate OPEN in the main control room:

- 1 E5 1 -F086, RCIC Turbine Exhaust Vacuum Breaker Upstream Isolation Valve

- 1 B2 1 -F019, MSIV Drain Header Outboard Stop Valve

- 1 CM022A, 1A Post LOCA H2/02 Monitor Drywell Suction Valve

- 1 WR040, RBCCW Return Header Outboard Isolation Valve Which one of the listed valves is NOT in its expected position?

I :

A? lB21-FO19 B. 1WR040 C. lE51-FO86 D. 1CM022A j

~.~~

~~~~~~

~~ i Answer A is correct. The I A RPS MG Trouble Alarm is caused by a trip of the MG set which de-energizes its associated RPS Bus (in this case A RPS bus is de-energized).

A loss of RPS-A will cause the following:

- Deenergizes Logic A(A I ) and C(A2)

- 1/2 isolation of MSIV (all MSIVs remain open)

- Deenergizes Outboard isolation Logic to all OUTBOARD Isolation valves for PCIS Groups I, 2, 3, 5, 6, 7 and IO, with the exception of MSIVs, VP and RBCCW. Additionally, Division I Post LOCA Monitor will automatically START PCIS Group 8 is not affected, so the IE5I-FO86 should remain open. The I A Post LOCA monitor will start, so the I CM022A will reposition open (PCIS Group 2). WR is exempted, and will remain open on a loss of RPS A. The MSIV drain line isolation valves are not exempted and therefore should have isolated

Reference:

LOA-RP-101, Revision 07, page 6 System Description 091, page 41, Section Vii.A u

2003-01 ILT NRC Exam, Version: 5 Page 63 of 65

62. 256000 G2 4.6 001/256000/2.4.6/3.1/4.O/RO/H1GH/400.00.19/

i-( 1.OO Point) An LGA event occurred on Unit-1 :

- the reactor scrammed following a trip of both TDWPs

- MSIVs automatically closed on low level

- RCIC is out of service for emergent maintenance

- All control rods are full-in

- Bus 152 tripped on overcurrent

- HPCS has tripped on overcurrent

- RPV level is -130 inches on WR and decreasing consistent with decay heat load Based on the above conditions, which one of the following level control strategies would be appropriate?

A. Lineup but don't start Alternate Injection Systems, wait until level is below -1 50 inches then start and control level with Alternate Injection Systems.

B. Start the Motor Driven Reactor Feed Pump, then control level using the FRVs.

CY Reduce RPV pressure using Safety Relief Valves, then control level using the Condensate Pumps and the FRVs.

D. Initiate ADS per the LGAs, then control level with low pressure ECCS systems.

~

I Answer C is correct. Per the strategies document, "Establish a reactor pressure band of 450 to 650 psig. This pressure band will be suficiently below the CondensateLondensate Booster pump discharge pressure such that the MDRFP can be secured and condensate can be used for level control. This band would only be established ajier ver9cation that an RPVIeak was not present which would result in a cool down rate in excess of IOO'Fper hour."

The other answers are incorrect. The MDRFP has no power available without the SAT. You don't do a blowdown for level control unless level cannot be restored and maintained above -150 inches and this hasn't been proven yet becairse pressure is still high and all sources of makeup haven't been tried (also level is still above -150 inches).

Although lining up alternate injection systems is a good idea, just sitting back and waiting for level to drop below

-150 inches is definitely not appropriate.

Reference:

Strategies for Successful Transient Mitigation, Revision 01, page 7, j h t bullet under step 3.6 LGA-001, Revision 06 2003-01 ILT NRC Exam, Version: 5 Page 64 of 105

63. 268000 K 1.05 00 1/268000/K1.05/2.9/3.2/RO/MEMORY/121.00.02/

~-

(1.OO Point) Which one of the following tank levels will increase as a result of the Unit-1 Drywell Equipment Drain Sump (1 RE02) pumping down?

A. Reactor Building Equipment Drain Tank (IREOIT) v i

B. Chem Waste Collector Tank (1 WZOlT)

C. Waste Surge Tank (1WE02T)

D!' Waste Collector Tank (1 WE0 IT)

I

- __ _ _ _ - -~

Answer Dis correct One of the inputs to the I WEOlT is the Drywell Equipment Drain Sump

Reference:

LOP-RE-01 T, Revision 12, page 2 System Description 121, page IO,Section IV.A h-2003-01 ILT NRC Exam, Version: 5 Page 65 of 105

L A!' Loss of air to the VR Exhaust Differential Pressure Control dampers 1 VR07YNB/C/D.

~

~

~~

~ ~. - ~

~

~~~~~~~

~.

~

~~

I Answer A is correct. VR Exhaust Differential Pressure Control Dampers (I VRO7YA/B/C/D) fail open on loss of air and therefore exhaust flow would increase. This would cause Reactor Building differential pressure to increase (become MORE negative).

The other answers are incorrect. Tripping the Blast Coils would cause colder (more dense) air to enter the building.

As the air heats it will expand causing differential pressure to become less negative. The VR Supply Flow Control Damper (I VR06y) fail open on loss ofair, causing more supply flow than exhaustflow, this would cause differential pressure to be less negative. Tripping a Station Heat Recovery pump would have the same effect as tripping the Blast Coils.

Reference:

LOA-PC-101, Revision 07, pages 5 and 6, Section B. I, Steps 3 and 4 LOA-PC-101, Revision 07, page 14, Discussion step C.1 System Description I 18, page 09, and page I4 b

2003-01 ILT NRC Exam, Version: 5 Page 66 of 105

65. 290002 K4.03 00 1/290002/K4.03/3.2/3.2/ROfMEMORY/020.OO.O5P/

Pressure Vessel Internal design provides for installation ofcore Orifices in the...

I A. Lower Fuel Tie Plate.

B. Baffle Plate.

C. Core Bottom Plate.

D? Fuel Support Pieces.

-~ _ _ _ -. -

.~_.

Answer D is correct. There are two types offuel support pieces; four-lobed andperipheral. The four-lobedsupporr pieces are used in the central core region while the peripheral support pieces are located at the perimeter of the core. One of the primary functions of the fuel support pieces is to control the flow through each fuel bundle by use of replaceable orifice plates.

Reference:

System Description 020, pages 23 and 24, Section Q.

c 2003-01 ILT NRC Exam, Version: 5 Page 67 of 105

66. GENERIC 2 I I OOl/GENERIC/2.1.1/3.7/3.8/RO/HIGH/790.020/

- ~~

i

~

I b-

~

I (1.OO Point) A licensed operator worked on the following safety related hours during a seven-day period:

Monday Tuesday Wednesday Thursday Friday Saturday Sunday 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 12 hours 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 0 hours 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> 12 hours 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> All work periods began at 08:OO on each work day. Upon review, it was determined that this employee exceeded the overtime guidelines stated in LS-AA-I 19, Overtime Controls.

When did the employee exceed the overtime guidelines?

A. Both Wednesday and Saturday BY Saturday ONLY C. Sunday ONLY D. Both Saturday and Sunday

~~~~~

Answer B is correct. The employee's hours for Friday and Saturday exceeded the guideline for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in a 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period (guideline was exceeded starting at 16:OO on Saturday). All other work periods met the guidelines, therefore the other answers are incorrect.

The guidelines stated in LS-AA-I I9 allow the following:

( I ) I6 hours in a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period (2) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in a 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period (3) 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in a 7 day (I 68 hour7.87037e-4 days <br />0.0189 hours <br />1.124339e-4 weeks <br />2.5874e-5 months <br />) period (4) must have 8 how break between workperiods

Reference:

LS-AA-119, Revision 02, page 3, step 2.5. I, 2.5.2, and2.5.3

\\-

2003-01 ILT NRC Exam, Version: 5 Page 68 of 105

67. GENERIC 2 1 3 1 001/GENERIC/2.1.3 1/4.2/3.9/RO/HIGH/125.00.06/

(1.OO Point) Unit-1 and Unit-2 are both operating in Mode 1 with no equipment out of service. A valving error caused Fire Protection Header Pressure to drop to 116 psig for 30 seconds.

Based on the above event, the control room NSOs should verify proper operation of the Diesel Fire Pumps (DFPs) by observing on...

A. lPM09J that OA DFP red light is lit and on 2PM09J that OB DFP green light is lit.

B. 1 PM09J that OA DFP red light is lit and on 2PM09J that OB DFP red light is lit.

C. 1PM09J that OA DFP red light is lit and OB DFP green light is lit.

D?' 1PM09J that both OA and OB DFP red lights are lit.

L

~

~~~

Answer D is correct. The OA DFP automatically starts at 124 psig and the OB DFP starts at I20 psig. Therefore if FP Header Pressure drops to I I6psig, both OA and OB DFPs should automatically start Controls and indication for the DFPs and the Fire Jockey Pumps are located on IPM09J There is one fire siren pirshbutton are located on each unit (IPMO9J and 2PM09J).

The other answers are therefore incorrect.

Reference:

LOP-FP-02, Revision 16, page 5, step 0.2 and 0. 3 System Description 125, page 36, Table L

2003-01 ILT NRC Exam, Version: 5 Page 69 of 105

~ -

68. GENERIC 2 2 1 1 001/GENERIC/2.2.11/2.5/3.4/RO/HIGW/

~

I (1.OO Point) Which one of the following is considered a formal Temporary Configuration I

r Change that would require an entry in the Operations Temporary Change Tracking Log as described in CC-AA-1 12, Temporary Configuration Changes?

A!' A temporary strip chart recorder connected to read the points normally monitored by a L

1 I

recorder that has failed.

B. A temporary pressure gauge installed on an engineered test point being used for troubleshooting.

C. Removal of RHR Pump motor control power fuses as part of a Clearance Order to allc repair of the pump motor.

~

~~~

Answer A is correct It is the only choice that requires a formal TCCP per CC-AA-I I2 u a temporary configuration change... is reviewed and controlled in other processes, or controlled by other procedures._. then a formal TCCP with content described in this procedure is NOTrequired.

A TCCP is not required for MT&E installed for troubleshooting eflorts on equipment withcut engineering test points that meet the... requirements.

Reference:

CC-AA-I 12, Revision 08, page 22, first paragraph, page 25, step 5. b Y

D. A charging hose with a pressure gauge attached being used to charge an HCU accumulator per LOP-RD-20.

2003-01 ILT NRC Exam, Version: 5 Page 70 of 105

69. GENERIC 2.2.12 002/GENERIC/2.2.12/3.0/3.4/RO/HIGW400.00.15/

~

(1.OOPoint) Referto the ACTIONS Table provided in Technical Specification Example 1.3-6 below to answer the following question.

~ ' 1 > s ?

1. 3 4 ACTIOXS C O W 1 TI OS RiQLlIRED ACTION A. I Perform SR 3. x. x. x.

OR A. 2 Reducc THERMAL POKER to I 50% RTP.

B. 1 Be in YODE 3.

COYPLETIOS TJUE

~~

~

Once per-8 hours 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> I2 ll0Ul.S One channel is declared inoperable at 08:OO on June 1. The designated Surveillance Requirement (SR 3.x.x.x) is completed at 12:OO on June 1.

Including any extensions permitted by Technical Specifications, which one of the following describes the LATEST time and date to perform the Surveillance next without requiring entry into Condition B?

A. 20:OO on June 1 BY 22:OO on June 1 C. 0O:OO on June 2 D. 02:OO on June 2 L-

~

~

Answer B is correct. Action A. I is a "once per... " completion time which qualifies for the 25% extension, per SR 3.0.2, to each performance after the initialperformance [12:00 + (8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />sx 1.25) = 22:OO on June I ]. VRequired Action A. I is followed and the Required Action is NOT met within the Completion Time (plus the extension aiiowed by SR 3.0.2). Condition B is entered The other choices are incorrect because they represent typical calculational errors (using incorrect start time, etc.):

I ) without an extension; 12:OO + 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> = 20:OO on June I

2) without an extension starting at time of discovery; 08:OO + 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> + 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> = 0O:OO on June 2

3) starting at time of discovery, applying an extension; 08:OO + 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> + 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> = 02:OO on June 2

Reference:

Technical Specifjcation 1.3, pages 1.3-10 and 1.3-11 Technical Specification SR 3.0.2. page 3.0-4 2003-01 ILT NRC Exam, Version: 5 Page 71 of 105

70. GENERIC 2.2.28 001/GENERIC/2.2.28/2.6/3.5/RO/MEMORY//

~

(1.OO Point) Who maintains the master copy of the Nuclear Component Transfer List (NCTL) and is responsible for signing off the completed NCTL sheets while performing component movements within the Unit-2 reactor or spent fuel storage pool.

~

5-.

A. Refuel Hoist Operator BY Fuel Handling Supervisor C. Unit-2 Unit Supervisor D. Unit-2NSO Answer B is correct. All fuel movements shall be performed in accordance with an approved Nuclear Component Transfer List (NCTL). The Fuel Handling Supervisor shall initial each step completed and sign off the completed sheet.

Reference:

LFP-400-1, Revision 24, page 8, Step D.2 LFP-100-2, Revision I I, page 8, Step E.3

'-c 2003-01 ILT NRC Exam, Version: 5 Page 72 of 105

71. GENERIC 2.3. I OOI/GENERIC/2.3.1/2.6/3.0/RO/HIGH//

(1.OO Point) You are a 25 year old Oc&pational Radiation Worker and have NOTTxceeded any legal or administrative exposure limits through the year 2003.

In 2004 you received 3.0 Rem of Routine Exposure and 4.0 Rem of authorized Planned Special Exposure (PSE).

What is your 10 CFR 20 NRC Exposure Limit, for Routine Occupational Radiation Work, during the year 2005?

B. 1.0 rem C. 3.0rem Dr' 5.0 rem A. 0.0 rem

~~

~

~~~~

Answer D is correct. Including his 2004 exposure, the employee in question has NOT exceeded any exposure limit.

The IO CFR 20 limit for ROUTINE exposure is 5.0 Rem and the limitfor PSE is an additional 5.0 Rem (up to a lifetime dose of 25 Rem PSE).

The other answers are incorrect because they contain math errors based on common misconceptions.

Reference:

RP-AA-203, Revision 2, page I, step 2.6, andpage 2, Table I.

L 2003-01 ILT NRC Exam, Version: 5 Page 73 of 105

72. GENERIC 2.3.4 OOI/GENERIC/2.3.4/2.5/3.1/RO/MEMORY//

~

The Station Emergency Director is assigned the non-delegable responsibilsy for authorizing personnel exposure under emergency conditions recommended NOT to exceed Rem TEDE for protecting valuable property.

A? 10 B. 15

- 1

c. 20 D. 25 I

Answer A is correct. The Station Emergency Director musrauthorize exceeding 10 CFR 20 guidelines. Up to IO Rem TEDE to protect valuable property and up to 25 Rem TEDE for lifesaving or protectron of large populations.

Reference:

EP-AA-1000, Revision 16, page K-I, Section K, step I EP-AA-113. Revision 05, page 6, Step 4.3.3

b.

2003-01 ILT NRC Exam, Version: 5 Page 74 of 105

~~

73. GENERIC 2 3.10 002/GENERiC/2.3.10/2.9/3.3/RO/HIGW/

~-

_ _ _ ~

(1.Ob Point) Per HU-AA-101, Human Performance Tools and Verification Practices, which one of the following correctly states the philosophy to be applied while performing Independent in Mode l ?

I A. Request that the Shift Manager waive the IV due to the low frequency of personnel traffic in I

Verification (IV) of safety related valves located in the Outboard MSIV Room while operating the areas and complete the IV for the rest of the checklist.

B. Request that the WEC SRO waive the IV due to the low frequency of personnel traffic in the '

areas and complete the IV for the rest of the checklist.

C!' Request that the Shift Manager waive the IV for these valves to reduce personnel exposure and complete the IV for the rest of the checklist.

D. Perform the IV after receiving High Radiation and ALARA briefs from the WEC SRO.

~~~~~

Answer C is correct because the Shift Manager may waive the verification requirementsfor ALARA concerns.

Alternative verification techniques shall be considered. Outboard MSIV room is an infrequently accessed Locked High Rndiation Area.

Reference:

HU-AA-101, Revision 2, page 6, Step 4.3.1.1

--..e 2003-01 ILT NRC Exam, Version: 5 Page 75 of 77

I

74. GENERIC 2.4.1 00 l/GENERIC/2.4.1/4.3/4.6/RO/MEMORY/304.010/

~~

(1.OO Point) During a ATWS with power greater than 3%, which one of the following actions can the panel operators take without Unit Supervisor direction?

A!' Initiate Standby Liquid Control B. TripRRpumps 1

C. Inhibit ADS 1

D. Prevent injection from ECCS

~~~~

~~~~~~~~~~

~

~~~~~

Answer A is correct. The scram hardcard directs the operator to 1) Initiate SBLCper LGA-010; 2) Initiate ARI; and

3) Insert Control Rodper LGA-NB-01 or LOA-RD-101 IF control r o b remain out following a reactor scram. These actions are taken regardless of the Unit Supervisors pIace in the LGA flow charts.

The other answers are incorrect because direction to perform those steps is directed by the LGAJow charts onIy.

Reference:

LGP-3-2, Revision 50, page 18, Attachment E (Scram Hardcard) 2003-01 ILT NRC Exam, Version: 5 Page 76 of 105

75. GENERIC 2.4.34 OOl/GENERIC/2.4.34/3.8/3.6/RO/MEMORY//

(1.OO Point) L O A - E - 10 1, Unit-1 Control Room Evacuation Abnormal procedure directs you ta OPEN RPS System A and B feed breakers (CB2A and CB2B) located at the (1) left open until recovery to (2)

A. (1) RPS MG Sets on the 749 foot elevation of the Auxiliary Building. These breakers are (2) prevent resetting the reactor scram BY (1) RPS Distribution Panel (2) ensure against inadvertent isolation reset and valve opening caused by shorts C. (1) RPS MG Sets (2) ensure against inadvertent isolation reset and valve opening caused by shorts D. (1) RPS Distribution Panel (2) prevent resetting the reactor scram Answer B is correct. Per UFSAR discussion, it is assumed that the operator can manually scram the reactor b e f o y leaving the Main Control Room. However, the RPS System A and B feed breakers are opened as a backup means of scramming the reactor AND to close the containment and reactor vessel isolation valves. The breakers are le8 open until recovery to ensure against inadvertent isolation reset and valve opening caused by shorts.

The other answers are incorrect. RPS scram cannot be remotely reset,

Reference:

LOA-RX-101, Revision 04, pages 5 and 15 L

2003-01 ILT NRC Exam, Version: 5 Page 77 of 105

76. 295003 G2.4.4 001/295003/2 4.4/2.913.6/SRO/HIGH/400.00.15/

(1.OO Point) A breaker trip event occurred on Unit-2, some of the equipment de-energized is I

listed below:

- 2A Diesel Cooling Water Pump

- 2C and 2D RHR Service Water Pump

- 2B Fuel Pool Emergency Makeup Pump

- 2B Primary Containment Vent Fan

- 2B Primary Containment Chiller Technical Specification 3.6.3.1 (Primary Containment Hydrogen Recombiner) is not being met because (1)

Supervisor should direct actions to recover power per LOA-AP-201, Unit-2 AC Power System Abnormal for (2)

Hydrogen Recombiner is OPERABLE for Unit-2, and the Unit A. (1) ONLY Unit-1 B. (1) neither (2) Loss of Bus 235WY; (2) Loss of Bus 236X/Y; CY (1) ONLY Unit-1 (2) Loss of Bus 236WY; D. (1) neither (2) Loss of Bus 235X/Y;

~

~

~

~

~

~

Answer C is correct. All items listed as de-energized are poweredfiom 236X and 236Y. Therefore LOA-AP-201 entry for Loss of Bus 236X/Y is appropriate. These busses are required by Technical Specijications.

Two primary containment hydrogen recombiners, including the associated Residual Heat Removal (RHR) pumps, piping and valves necessary to provide recombiner cooling, must be OPERABLE. Attachment K of LOP-AP-242Y lists the Division 2 power supplies to the following HG system valves:

2HG001A, Unit-2 HG Unit-2 Drywell Suction - 236Y-1 compartment GI 2HG002A, Unit-2 HG Unit-2 Drywell Suction - 236Y-1 compartment G2 2HG009, Unit-2 HG Unit-I Cross Over, 236Y-1 compartment G5

Reference:

Technical Specification B 3.8. I, pages B 3.8.1-4 and B 3.8.1-7 Technical Specification B 3.6.3. I, page B 3.6.3.1-2 LOP-AP-242X page 204, Attachment K System Description 094, Figure 094-01 u

2003-01 ILT NRC Exam, Version: 5 Page 80 of 105

- Division 1 Post LOCA monitor automatically started

- All SRV OPEN and CLOSED position indicating lights are out Dr' LOA-DC-10 1, Unit 1 DC Power System Failure due to loss of 1 1 1 Y.

~

i

~~

Answer D is correct Per the referenced procedure, loss of 125 vdc Bus I I 1 Y would cause all ofthe listed indications Entering LOA-DC-IO I is the highest priority because step I on loss of I I I Y is to manually scram the reactor The other answers are incorrect because loss of all of the listed symptoms are Division I and therefore I12Y is incorrect. Although there are indications of both a PCIS Group 2 and Group 4 isolation, entering LOA-PC-101 before entering LOA-DC-IO I would be imprudent becuase LOA-PC-IOI does NOT require a manual scram.

Reference:

LOA-DC-IO], Revision 07, page 52 LOA-PC-IOI, Revision 07, pages 16 to 18, Attachment A, hard card v

2003-01 ILT NRC Exam, Version. 5 Page 81 of 83

78. 29501 6 (32.2.25 00 112950 16/2.2.25/2.5/3.7ISRO/MEMORY//

(1.OO Point) The Remote Shutdown MonitoringSystem LCO provides for operability of instrumentation that is required for all of the following EXCEPT?

L A. RPV pressure control using SRVs B. RPV inventory control using RCIC C. Decay heat removal using Shutdown Cooling DY Containment control using Drywell Spray

~

-~

~- -

~~

Answer D is correct. Although control of IE12-FOldB is located on the Remote Shutdown Panel (RSP), you can NOT initiate Drywell Spray because controls for 1E12-FOI 7B are NOT located on the RSP. Control for IEI2-FOl6B are only provided to STOP DWSprays ifinitiatedprior to exiting the control room or ifthe fire caused a short and opened the valve before control was transfered to the RSP.

The other answers are incorrect. The Remote Shutdown Monitoring System LCO provides the requirements for the OPERABILITY of the instrumentation... that is required for:

1) Reactor pressure vessel (Rp v) pressure control

2) Decay heat removal; and

3) RPV inventory control.

Reference:

Technical Specification B 3.3.3.2, page B 3.3.3.2-2 UFSAR, pages 7.4-18 through 7.4-21, Section 7.4.4 LOP-RX-01 T, Rwision IO, entire list (doesn't have control switch for drywell spray valves)

W 2003-01 ILT NRC Exam, Version: 5 Page 82 of 105

79. 295021 AA2.05 001/29502 I/AA2.05/3.4/3.5/SRO/MEMORY//

(1.OO Point) Unit-1 is in Mode 4 with reactor coolant temperature at 190°F. The operating loop of Shutdown Cooling trips and can NOT be restarted. Both Reactor Recirculation pumps are out of service and Reactor Water Cleanup (RWCU) is isolated.

It has been determined that it will take 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to align and start the standby Shutdown

\\-

I,

Cooling loop.

As the Shift Manager you should...

A. authorize raising RPV level to between 50 and 60 inches in order to provide the extra Net Positive Suction Head (NPSH) needed to start the standby Shutdown Cooling pump when coolant temperature is above 180°F.

BY authorize raising RPV level to between 220 and 260 inches in order to more accurately I

determine vessel metal temperatures and demonstrate that stratification is NOT occurring.

C. authorize raising RPV level to between 220 and 260 inches in order to increase the water volume available for removal of decay heat which increases heat transfer rate to the containment atmosphere.

D. direct maintaining RPV level between 30 and 40 inches in order to increase the Time to Boil allowing extra time to start the standby Shutdown Cooling loop.

-~

~

~~

-~

Answer B is correct. The Shifi Manager has to direct increasing reactor water level to a range of +220 inches to

+260 inches enhances vessel metal temperature determination to demonstrate that stratijcation is not occurring.

Additionally, this aids natural circulation, which starts to occur at +50 inches (bottom of skirt).

Reference:

LOA-RH-101, Revision 07, page 20

\\

2003-01 ILT NRC Exam, Version 5 Page 83 of 105

80. 295023 G2.4.4 00 1 /295023/2.4.4/4.014.3~SROhlEMORY/O9 1.00.08l

- (1.OO Point) An event occurs on Unit-2 that results in the following alarms:

- 2H13-P601-E306, Fuel Pool Vent Rad Hi

- lH13-P601-E306, Fuel Pool Vent Rad Hi

- 2H13-P601-F205, Div 1 Fuel Pool Rad Hi-Hi An Extra NSO verified that Fuel Pool Monitors 2A and 2B have tripped.

1 The crew should be directed to enter (1) prevent (2) to verify isolations are complete to A?' (1) both Unit-1 and Unit-2 LGAs; (2) off-site doses from exceeding 10 CFR limits B. (1) only the Unit-2 LGA; (2) over exposure of personnel on the refuel floor C. (1) both Unit-1 and Unit-2 LGAs; (2) over exposure of personnel on the refuel floor D. (1) only the Unit-2 LGA; (2) off-site doses from exceeding 10 CFR limits

~

~~

-~

Answer A is correct. The operators should be directed to enter LGAs for both units because annunciator 2HI 3-P60l-F205 indicates that the high radiation trips set point has been exceeded (the other two annunciators are only alarms). When the trip setpoint is exceeded on either unit, it causes a PCIS Group 4 isolation on BOTH units. The crew should therefore verifj, isolations are complete because the Ts Bases states that the VR and Fuel Pool Radiation Monitor trip setpoints are based on NOT exceeding 10CFR20 and IOCFRIOO limits. The other answers are therefore incorrect.

Reference:

2H13-P601-F205, Revision 02 Technical Specrfication Bases 3.3.6. I, pages B 3.3.6.1-14 and 15 Technical Specrfication Bases 3.3.6.2, pages B 3.3.6.2-6 W

2003-01 ILT NRC Exam, Version. 5 Page 84 of 105

81. 295028 G2.1.32 001/295028/2.1 32/3.4/3.8/SRO/HIGW400.00.12/

~

(1.OO Point) Unit-1 was manually scrammed due to a small leak in the primary containment. The following conditions exist:

- Drywell pressure is 2.0 psig and slowly rising

- Drywell temperature is 2 15°F and slowly rising Which one of the following identifies whether LGA-VP-0 1, Primary Containment Temperature Reduction can be used to reduce Drywell temperature and the basis for that decision?

A. No, because neither LGA-001, RPV Control nor LGA-003, Primary Containment Control have entry conditions that are met.

BY No, because even though LGA-003, Primary Containment Control has entry conditions that are met, Drywell temperature is greater than 2 12°F with a LOCA in the containment.

C. Yes, because LGA-001, RPV Control and LGA-003, Primary Containment Control both have LOCA entry conditions that are met.

1 I

D. Yes, because LGA-003, Primary Containment Control has entry conditions met, and I

I temperature in the drywell is greater than 135°F.

~~

-i Answer B is correct because LGA-003 should be entered when drywell temperature is above 135"Fand LGA-VP-01, Prerequisites, Entry Conditions state "There has NOT been a LOCA (large or small) on the Unir which has raised DW temperature above 212°F I' The other answers are incorrect because LGA-001 does not have any entry conditions that are met.

Reference:

LGA-VP-01, Revision 08, page I, Step B. 1.a.

LGA-001, Revision 06 LGA-003, Revision 05 v

2003-01 ILT NRC Exam, Version 5 Page 85 of 105

82. 29503 1 EA2.02 00 1129503 1/EA2.02/4.0/4.2lSRO/HIGWl (1.OO Point) An ATWS-event has occurred on Unit-1.

- Drywell pressure is +0.5 psig

- A PCIS Group 1 Isolation has occurred

- One SRV was open and maintaining reactor pressure in LLS You are maintaining RPV level per the level band prescribed by step 9 of LGA-010 when the Unit NSO reports that at -100 inches on the Wide Range the SRV went closed. Reactor pressure I is approximately 900 psig and steady.

As the Unit Supervisor, you should..,

A. exit level band 9 and control level per level band 8 because power is still ABOVE 3%.

B. continue to control level in the currently prescribed level band because power is still ABOVE 3%.

I 1

C. exit level band 9 and control level per level band 7 because power has decreased BELOW 3%.

~

DY continue to control level in the currently prescribed level band even though power has decreased BELOW 3%.

I L---

~

~~~

Provide LGA-010 to the examinee as a handout.

Answer D is correct. Once you enter level band 9 you remain there until LGA-010 is exited because it is the actions of Step 9 that will reduce power until cold shutdown boron is injected or all control rod are inserted or the QNE determines that the reactor will remain shutdown under all conditions.

The other choices are incorrect because the erroneously state that power is above 3% or have you change level bands.

Reference:

LGA-010, Revision 06 LGA Flowchart Use Lesson Plan, page 21, Steps I and J v

2003-01 ILT NRC Exam, Version: 5 Page 86 of 105

83. 295009 AA2.03 00 1/295009/AA2.03/2.9/2.9/SRO/HIGW027.00.05H/

~~

~

(1.OO Point) Unit-2 was operating in Mode 2. RWCU Reject to the Main Condenser is being used to control RPV level per LOP-RT-09, RWCU Sytem - Coolant Rejection.

- RPV Water Level was +34 inches and steady The NSO made an adjustment to the blowdown flow rate, and 20 minutes later the RPV Low Water Level alarm (Level-4) was received.

Based on the above information, the NSO increased the reject flow rate by approximately (1) to the band directed by (2) and the Unit Supervisor should then direct the NSO to restore RPV level A. (I) 50 gallons per minute; (2) LOP-RL-01, Operation of the Reactor Water Level Control System BY (1) 25 gallons per minute; (2) LGP 1 Normal Unit Startup I

C. (1) 25 gallons per minute; I

j (2) LOP-RL-0 1, Operation of the Reactor Water Level Control System D. (1) 50 gallons per minute; (2) LGP-1-1 Normal Unit Startup

~~~

~

Answer B is correct. Per the reference procedure, there are approximately 200 gallons per inch in the RPV. The RPV low level (level-4) alarm is set at 31.5 inches.

34 - 31.5 = 2.5 inch level decrease 200 gallons/inch x 2.5 inches = 500 gallons druinedfiom the vessel 500/20 minutes = 25 gallon per minute change in the reject flow rate To cause level to decrease, the NSO must have opened the blowdown flow control valve, or increased the reject rate by 25 gallons per minute. LGP-1-1 is the correct procedure based on the unit operating in Mode 2 (startup) without LGA-001 entry conditions met. The low level alarm is at +31.5 inches and entry condition for LGA-001 is + I I inches. Therefore the other answers are incorrect.

Reference:

LOP-NB-02, Revision 08, page 5, Step D. I. 7 LGP-1-1, Revision 72, page 22, Step E.3.12 LOP-SF-06, Revision 12, page 72, Attachment A4 u

2003-01 ILT NRC Exam, Version: 5 Page 87 of 105

84. 295007

~

G2.1.7 001/295010/2.

I.7/3.7/4.4/SRO/HIGW070.00.14/

(1.OO Point) During a reactor pressure transient, 4 Safety Relief Valves (SRVs) aGornatically opened when RPV pressure reached 1090 psig. All other SRVs have remained closed during the transient.

Which one of the following correctly states the status of the SRVs and the actions that should be directed by the Unit Supervisor?

A. Four SRVs have opened before their relief setpoint, close the SRVs per LOA-SRV-101 Stuck Open Safety Relief Vales.

B. All thirteen SRVs have functioned as designed, control pressure below 1059 psig per LGA-00 1, RPV Control.

C!' Two SRVs have failed to open at their relief setpoint, control pressure below 1059 psig per LGA-001 RPV Control D. One SRV has opened before its relief setpoint, close the SRV per LOA-SRV-101, Stuck Open Safety Relief Valve.

- ~~~

Answer C is correct. The SR V setpoints are as follows:

Relief Relief Low-Low Valve Valve Set K V OPEN CLOSE OPEN s/u I076 976 1076 C/D/E/K I086 986 IO86 F/P I096 996

- 41096 H, L,M 1106 1006 R, V 1116 1016

Reference:

LOA-SRV-101, Revision 04, page 11, Table 2 2003-01 ILT NRC Exam, Version. 5 Lo w-L 0 w Set Re-Ouen l046/1 006

~~~~

Low-Low Set CLOSE 926/896 946

- 4946 S a ! w Valve OPEN I I50 1175 I185 1195 1205 Safety Valve CLOSE 1115 I I40 I I49 I159 I169 Page 88 of 105

85. 295033 G2.4.6 001/295033/2.4.6/3 I/4.0/SRO/HIGW/

I (1.OO Point) Following an event on Unit-1, the control room operators gather the following information:

- RE3 RCIC Room Area Radiation Monitor is pegged high

- RE3 North HCU Area Radiation Monitor is pegged high

- SBGT Area Radiation Monitor is reading 100 mrem/hr

- RCIC has received an isolation signal however the isolation valves have failed to close

- Unit-1 continues to operate in MODE 1 I

The above conditions require a (1) because two (2) values for the same parameter have been exceeded.

A!' (1) manual scram (LGP-3-2) and blowdown (LGA-004 or LGA-006);

(2) Max Safe B. (1) unit shutdown (LGP-2-1);

(2) Max Normal C. (1) manual scram (LGP-3-2) and blowdown (LGA-004 or LGA-006);

(2) Max Normal D. (1) unit shutdown (LGP-2-1);

(2) Max Safe i -.

~

~

~

Answer A is correct. Whezhe Reactor Building A M s in Table R of LGA-002 are pegged high, theirkax Safe values have been exceeded. With the unit in MODE I, the mode switch is in run and the reactor is pressurized. The same parameter (radiation) in two or more areas above Max Safe, indicates that the problem is wide spread. With RCIC NOT isolated, the LGAs will direct a manual scram and two or more areas above Max Safe for the same parameter, a blowdown is required to reduce the leak rate and spread of contamination.

The other answers are incorrect. Without aprimary system discharging into the reactor building the LGAs only require a unit shutdown when two Max Safe values are reached for the same parameter.

Reference:

LGA-002 Lesson Plan, page 21.Section X.A LGA-002, Revision 03, Table R

.--/

2003-01 ILT NRC Exam, Version: 5 Page 89 of 105

86. 209001 A2.03 001/209001/A2.03/3.4/3.6/SRO/HIGH/063.00.14/

(1.OO Point) A transienthas occurred on Unit-1 resulting in the following conditions:

- 142Y is de-energized

- RPV pressure is 290 psig and slowly decreasing

- RPV level is -30 inches and rising at 50 inches per minute Subsequently, annunciator 1 PMO 1 J-A3 13 Feed to 1 3 5 W 133 Auto Trip (SER R-point R02 1 8 Bus 141Y Fd Bkr to 135WY A-Trip) is received.

Based on the above conditions, the Unit Supervisor should direct action to...

A!' trip the LPCS pump and locally close the LPCS Injection valve per LOP-AA-04, Operation of valves.

I I

B. locally close the LPCS Injection valve per LOP-AA-04, Operation of Valves, but leave the LPCS Pump running until plant conditions are stabilized.

I C. return the LPCS system to standby when level is above the scram setpoint per LOP-LP-03, Shutdown of LPCS After an Automatic Initiation.

D. return the LPCS system to standby when level is above the LPCS initiation setpoint per LOP-LP-03, Shutdown of LPCS After an Automatic Initiation.

~~~~~~~

~

~-

Answer A is correct. LPCS is NOT needed for level control because level is rising rapidly. Therefore LPCS injection should be stoppedper LGA-001, RPV Control. Tripping thepumpfiom the control room will stop injection, however, the injection valve is de-energized (loss of 135m7 and must be manually closed per LOP-AA-04.

LOP-LP-03 is not applicable because it assumes that power is available to the LPCS components. LPCS can NOT be placed in standby when a component that is required to automatically re-align for injection is de-energized and not available.

Reference:

LGA-001. Revision 06, LOP-AA-04, Revision 21, pages I 1 through 14 LOP-AP-I41 Y, Revision 02, page 81, Attachment G

-v 2003-01 ILT NRC Exam, Version: 5 Page 90 of 105

_ _ _ ~

~~~

87. 2 12000 (32.4.30 00 1/212000/2.4.30/2.2/3.6/SRO/HIGH/7

- 15.001/

Unit-2 was operating-;

full rated conditions when an event occurred requiring a manual scram:

1) The NSO armed and depressed all manual scram pushbuttons

2) All group scram lights remained lit and control rods did NOT move

3) The NSO manually initiated both divisions of AM

4) All control rods fully inserted following ARI initiation Based on the above event, (1) and therefore, (2) are required to be notified per Nuclear Accident Reporting System (NARS)

A. (1) NONE of the EALs are applicable; (2) NO outside agencies BY (1) EAL MA3 is the highest applicable EAL; (2) ONLY Illinois EMA and REAC C. (1) EAL MS3 is the highest applicable EAL; (2) Illinois EMA and REAC, the local county Sheriff and EMA, and LaSalle County ESDA D. (1) EAL MS3 is the highest applicable EAL; (2) ONLY Illinois EMA and REAC

~~

Provide EP-AA-1005, pages LS 3-11, LS 3-61, and LS 3-62 lo the examinee as a handout.

Answer B is correct. The above went is classrfied as an Alert per EAL MA3 based on manual initiation ofARIprior to automatic initiation as described in EAL MS3. Per the NARS Form, when the went is classified as a Unusual Event, Alert, or Site Emergency use NARS (dial) Code 20 to notifi: I ) Illinois EMA, and 2) Illinois REAC. The onk time state agencies are contacted using the initial NARS transmittal form, is when the initiating event is classrfied as a General Emergency.

Reference:

EP-AA-1005, Revision 17, pages LS 3-1 I, LS 3-61, and LS 3-62 EP-W-114-100-F-01, Revision A, page 2

-r_-

2003-01 ILT NRC Exam, Version: 5 Page 91 of 93

I i

i i

1 i

~

88. 261 000 A2.05 001/26 1 OOO/A2.05/3.0/3.1/SRO/HIGW095.00.05B/

-~

~~

~~

(1.O-oint)

Several hours after an automatic initiation during a LOCA, and with the initiation signal still present, the Unit-1 Standby Gas Treatment Primary Fan trips.

After the trip, the Unit-1 Assist NSO observed the following at panel 1 PM07J:

- SBGT Primary Fan - OFF

- SBGT Cooling Fan - OFF

- SBGT Heater - OFF

- SBGT Flow - 0 scfm

- 1 VGOO 1, SBGT Inlet Isolation damper - CLOSED

- 1VG003, SBGT Outlet Isolation damper - OPEN The SBGT train (1) direct the operator to (2) to the Primary Fan trip and the Unit Supervisor should A. (1) responded as expected; (2) SHUTDOWN the SBGT train per LOP-VG-02, Shutdown of the SBGT System B. (1) responded as expected;

~

(2) START the standby SBGT per LGA-VG-101, Secondary Containment Pressure Control C!' (1) did NOT respond correctly; (2) manually START the Cooling Fan per LOR-1 PM07J-A502, SBGT Pri Fan Auto Trip D. (1) did NOT respond correctly; (2) CLOSE the Outlet Isolation damper per LOR-1 PM07J-A502, SBGT Pri Fan Auto I

Trip

Reference:

LOR-lPMO7J-A502, Revision 02 2003-01 ILT NRC Exam, Version 5 Page 92 of 105

(1.OO Point) Unit-1 is-being shutdown for emergent maintenance. The Shutdown Cooling Isolation Interlocks have just cleared and Main Turbine Bypass valves are being used to control reactor pressure.

While troubleshooting ADS Accmulator alarms, the FIN Team reports that the ADS A!' Direct actions per annunciator procedures ONLY because ECCS subsystems are capable of providing flow into the RPV.

B. Direct actions per the IN abnormal procedure AND take actions per TS 3.5.1 required action G.l AND G.2 because ADS does not meet the single failure criteria used in the accident analysis.

C. Direct actions per the IN abnormal procedure AND take actions per TS 3.5.1 required action G.2 ONLY because low pressure ECCS subsystems are currently incapable of providing flow into the RPV.

D. Direct actions per annunciator procedures AND take actions per TS 3.5.1 required action F.l ONLY because even though portions of ADS are inoperable, ADS still meets the single failure criteria used in the accident analysis.

~

-~.

- ~-

~-

Provide Technical Specification3.5.1, pages 3.5.1-1, -2, -3, -4 and -5 to the examinee asahandout.

Answer A is correct. Annunciator IH13-P601-F102 will alarm when any low accumulator is received (pressure less than 151 psig) and therefore actions should be taken per the LOR. The abnormalprocedure (LOR-IN-101) does NOT cover ADS accumulator pressure events. The ADSportion of TS 3.5.1 is applicable in MODE I, and in MODES 2 and 3 when reactor steam dome pressure is >I 50 psig. TS 3.5.

I is NOT applicable per the bases because when S150psig the low pressure ECCS systems are capable ofprovidingflow to the RPV. The other answers are incorrect because they assume that TS 3.5.1 pressure is greater than I50 psig.

Reference:

LOR-lHl3-P601-M02, Revision 03 Steam Tables Technical Specrfication 3.5.

I pages 3.5.1-1, through 3.5.1-5 2003-01 ILT NRC Exam, Version: 5 Page 93 of 95

I I

I I

i I

90. 400000 A2.03 001/400000/A2.03/2.913.O/SRO/HIGW/

~~

~

~

(1 rO0 Point) Unit-1 is operating at full power and Unit-2 is in an extended maintenance outage.

The Ice Melt Line was just taken out of service to support emergent maintenance activities on the Unit-2 Circulating Water Piping. Maintenance is expected to last another 4 days.

I The following conditions exit:

- Circulating Water Inlet temperature is 55°F

- Outside Air Temperature is 10°F and forecasted to drop over the coming week

- The OA Service Water Jockey Pump is the only Service Water Pump running

- Service Water Pressure is 107 psig Over the next week, Service Water pressure is expected to (1)

Supervisor should direct system pressure be controlled by (2)

, and the Unit A. (1) decrease; (2) isolating unnecessary Service Water loads per LOA-WS-101, Service Water System Abnormal.

B. (1) decrease; (2) start an additional Service Water Pump per LOP-WS-02, Service Water Pump and Service Water Jockey Pump Startup and Operation.

CY (1) increase; (2) throttling flow through an idle RBCCW Heat Exchanger Outlet Valve per LOP-WS-02, Service Water Pump and Service Water Jockey Pump Startup and Operation.

D. (1) increase; (2) placing a Service Water Strainer on continuous backwash per LOP-WS-05, Service Water Strainer Operations.

~ -..

~

~~

.~. ~ ~ _ _ _

Answer C is correct. Without Ice Melt in service, Circulating Wier Inlet temperature will decrease, loads will automatically throttle back and Service Water (WS) pressure will increase. With on& one pump running, pumps can NOT be turned offto control pressure, therefore, LOP-WS-02 directs increasing flow through an idle RBCCW heat exchanger to control pressure.

Reference:

LOP-WS-02, Revision 14, pages 8 and 9 LOP-WS-02, Revision 14, page 13, NOTE ai top ofpage 2003-01 ILT NRC Exam, Version: 5 Page 94 of 105

91. 202002 G2.1.14 001/202002/2.1.14/2.5/3.3/SRO/MEMORY/f

~~

~~~

(1.OO Point) The Shift Manager is REQUIRED to notify the Duty Station Managerper OP-AA-106-101, Significant Event Reporting, for which one of the following events?

A!' Entering a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shutdown time clock due to an inoperable RR Flow Control Valve.

B. Entering a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> time clock requiring restoration of ECCS due to a scheduled surveillance.

C. Personnel injury that was treated and released by the site nurse.

D. Load reduction of 20% per a pre-approved Electric Power Operations request.

Answer A is correct. OP-AA-lO6-IOI, Signrjcant Event Reporting requires the Shifr Manager to notfi the Duty Station Manager to make notrjcations for the following Technical Specification reasons:

I ) forced entry into a 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or less shutdown LCO, or

2) $a LCO will not be met within the time requirement.

Reference:

OP-AA-106-101, Revision 05, pages 3, Step 4.2.1, andpages 6 and 7, Attachment I I

2003-01 ILT NRC Exam, Version: 5 Page 95 of 105

i I

I I

I

~

i

92. 2 16000 A2.07 001/2 16000/A2.07/3.4/3.5/SRO/HIGH/400.00.0l/

~~

( 1.OO Point) An ATWs and LOCA have occurred on Unit-1 and the appropriate LGAs have been entered. Power was 6% and the NSO was trying to maintain level at -130 inches, however level continued to drop slowly.

The following conditions exist now:

- Power is 5% and decreasing slowly

- Seven ADS valves are open

- RPV Pressure is 40 psig and decreasing slowly

- Drywell pressure is 22 psig and steady

- Suppression Pool temperature is 102°F and rising slowly

- Drywell temperature is 290°F and steady

- Fuel Zone is reading upscale

- All other RPV water level instruments indicate on-scale and rising quickly Which one of the following represents the next required LGA action and why?

A. Wait until RPV pressure is below that indicated in Table G to take advantage of steam cooling before re-injecting into the vessel.

B. Cool down to Cold Shutdown using Shutdown Cooling per LOP-RH-07 to allow entering the recovery phase of EOP actions.

C!' Enter the EOP for RPV Flooding because reference legs have flashed.

D. Bypass isolations per LGA-MS-01 to maintain the condenser as a heat sink.

~

~~

~~

Provide LGA-003, LGA-005 and LGA-010 to the examinee as a handout.

Answer C is correct. Based on RPVpressure and Drywell temperature, the RPV Saturation Temperature limit has been exceeded. This does not automatically mean that reference legs have flashed, however the NSO should be trying to maintain level between - I50 and -60 inches per LGA-010 step 8 and with level rising rapidly during an ATWS when attempting to maintain level low you have indications that reference legflashing has taken place.

Therefore per the overriding step, Enter LGA-005 at step 13, where thejirst action block directs preventing all injection except Boron, CRD and RCIC.

The other answers are incorrect. Waiting for RPVpressure to drop or waiting for Cold Shutdown Boron injection are both appropriate per LGA-010 however when level is unknown the override directs you to LGA-005. If LGA-MS-01 was appropriate it should have been done prior to establishing a level bandper LGA-010 step 8.

Reference:

LGA-003, Revision 05 LGA-005, Revision 06 LGA-010. Revision 06 2003-01 ILT NRC Exam, Version: 5 Page 96 of 105

93. 234000

~~

- A4.01 001/234000/A4.0 113 713 9/SRO/HIGW/

(1.OO Point) During a refueling outage the reactor core is beingioaded.

- All SRMs are fully inserted

- All control rods are fully inserted

- The signal to noise ratio for all SRMs is 15:l

- There are at least 6 fuel assemblies adjacent to each SRM

- Fuel assemblies are being loaded into the same quadrant where SRM A is located While lowering a Fuel Assembly into the same quadrant as SRM A is located, the NSO takes the following SRM readings:

1 1

- There are more than 10 fuel assemblies loaded in each quadrant I

1 SRM A = 2.5 SRM B = 2.0 SRM C = 4.0 SRM D = 3.0 Based on the above SRM indications core alterations should be...

~

A. allowed to continue because SRM D is operable and is located in an adjacent quadrant to where core alterations are taking place.

I B. suspended immediately because the capability to detect local reactivity changes in the core is degraded with SRM B inoperable.

C. allowed to continue because SRMs D and A are operable and are located in an adjacent quadrant and a quadrant where core alterations are taking place.

DY suspended immediately because the capability to detect local reactivity changes in the core is I

I I

degraded with SRM A inoperable.

~ ~ _ _ _ _ _ _ _

Provide Technical 3.3.1.2-1 through 3.3.1.2-6 to the examinee as a handout.

.J Answer D is correct. With one or more required SRMs inoperable (signal to noise ratio is <20: l so must read 3.0 to be operable) in MODE 5. the capability to detect local reactivity changes in the core during refueling is degraded. CORE ALTERA TIONS must be immediate& suspended, and action must be immediately initiated tojilly insert all insertable control rods in core cells containing one or more fuel assemblies.

Reference:

Technical Specification 3.3.1.2 and bases 2003-01 ILT NRC Exam, Version 5 Page 97 of 105

94. GENERIC 2.1.1 1 001/GENERIC/2.

I.I 113.0/3.8/SRO/MEMORY/400.00.15/

(1.OO Point) Unit-1 is performing a normal unit shutdown. The containment is being de-inerted when annunciator lH13-P603-B501, Pri Cnmt Pressure Hiko is received. Drywell pressure indication is verified to read negative 0.6 psig (-0.6 psig).

be exceeded if Drywell Spray is initiated.

D. restore pressure to within the limits in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> so that containment pressure remains within 1

i The NEXT action required by Technical Specifications is to.:.

A!' restore pressure to within the limits in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> so that containment pressure remains within design values if Drywell Spray is initiated.

B. place the mode switch in SHUTDOWN immediately because containment design limits will be exceeded if Suppression Chamber Spray is initiated.

~

_~__.

~

Answer A is correct. TS 3.6.1.4 LCO requires DryweN and Suppression Chamber pressure be maintained >-0.5 psig and

+O. 75 psig. Ifpressure is outside of these bounds then Required Action A. I states "Restore drywell and suppression chamber pressure to within limits" in I hour. So that containment pressure remains within the design values, the basis for the LCO sqvs that containment pressure htust be below the upper limit for the LOCA analysis and must be above the lower limit during inadvertent operation of dFywell sprays.

Reference:

Technical Specification 3.6.1.4, page 3.6.1.4-1 Technical Specification B 3.6. I. 4, page B 3.6.1.4-1 and B 3.6.1.4-2 L

2003-01 ILT NRC Exam, Version 5 Page 98 of 105

-7

95. GENERIC 2.1.13 OOl/GENERIC/2.1.1/2.0/2.9/SRO/h4EMORY//

(1.OO Point) A casualty has occurred resulting in the following:

- A leak has developed in the Reactor Water Cleanup (RWCU) Heat Exchanger room and

- RWCU has failed to automatically isolate and can NOT be isolated from the main control off-site release rates are increasing room I

Ten minutes following the initial transient, it was deemed necessary to dispatch an NLO to manually close the IG33-FO04, RWCU Suction Outboard Isolation valve. It is estimated that the NLO will exceed his federal exposure limits.

The Technical Support Center (TSC) is NOT ready to assume command and control.

i As the Unit Supervisor, you should...

A. Obtain RP Manager AND Station Manager authorization AND then direct the NLO to enter the room and close the valve per EP-AA-113, Personnel Protective Actions.

BY Obtain Emergency Director authorization AND then allow the NLO to enter the room to 1

close the valve per EP-AA-113, Personnel Protective Actions.

C. Authorize the NLO to enter the room AND close the valve per W-AA-460, Controls for High and Very High Radiation Areas.

D. Obtain RP Manager authorization AND then direct the NLO to enter the room and close the valve RP-AA-460, Controls for High and Very High Radiation Areas.

.~

~-~

~~~

Answer B is correct. R WCU outboard room is a Locked High Radiation Area (LHRA) during power operations.

Failure RWCU (a PCIS line) to isolate is a GSEP condition requiring implementation of the E-Plan. This requires the Shft Manager to become the Emergency Director who has the authority to authorize emergency exposure (can NOT be delegated).

Reference:

EP-AA-I 13, Revision 05, page 5, step 4.3.3 2003-01 ILT NRC Exam, Version: 5 Page 99 of 105

96. GENERIC

-~

2.2.23 001/GENERIC/2.2.23/2.6/3.8/SRO/HIGH/4OO.OO.I

-.~

5 /

(1.OO Point) Unit-1 is in Mode 1 at 100% power:

- at 12:OO on June 1, the 1A RHR WS pump is declared inoperable

- at 20:OO on June 1, the 1 B RHR WS pump is declared inoperable

- at 15:OO on June 2, the 1 A RHR WS pump is restored to OPERABLE I

I Including any extensions permitted by Technical Specifications, which one of the following is entering a shutdown timeclock?

the latest time and date allowed to restore the 1B RHR WS pump to OPERABLE status without

~

~,

A. 20:OO on June 1 B. 04:OO on June 2 C!' 12:OO on June 8 D. 20:OO on June 8 1

- ~ ~ _ _ _ _

~~

-~

~~

1 I

Provide Technical Specifications 3.7.1, pages 3.7.1-1 and 3.7.1-2 to the examinee as a handout.

Answer C is correct. Examinee must recognize that IA and 1 B RHR WSpumps are in the same subsystem (pump numbering does NOTfollow the normal numbering scheme) and therefore NO extensions apply. Per the Technical Specijication Bases, both pumps in the same subsystem must be operable for the subsystem to be operable. When the first pump goes inoperable, start a 7-day clock per Required Action A. I. The second pump going inoperable does NOT aflect the time clock. When the first pump is returned to operable, the subsystem remains inoperable because of the secondpump and therefore the clock can NOT be reset. The clock will expire 7 days after the first pump goes inoperable (12:OO on June 8).

The other answers are incorrect because they assume that IA and IB pumps are in opposite subsystems and therefore incorrectly applies the 8-hour time clock per Required Action B. 1.

NOTE: Technical Speclfication 3.4.9, Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown is NOT applicable in Mode I, it is ONLY in Mode 3.

Reference:

Technical Specrjication 3.7. I, pages 3.7.1-1 and 3.7.1-2 and bases page B 3.7.1-1.

5 _ -

2003-01 ILT NRC Exam, Version: 5 Page 100 of 105

.- 97. GENERIC 2.2.20 00 1 IG ENERICR.2.20/2.2/3.3/SRO/MEMORY//

(1.OO Point) You are the Unit Supervisor when-&

IMD Technician requests to troubleshoot sluggish response of the selected CRD Flow Valve. The troubleshooting activity would involve the lowering CRD Flow Controller Setpoint by 5 gpm, then returning the controller to its original setpoint while IMD observes the CRD Flow Control Valve in the field.

I As the Unit Supervisor, you should...

A. require the IMD Technicians to produce a work package with a troubleshooting plan previously approved by the Work Week Manager per MA-AA-716-004, Conduct of Troubleshooting.

L 55, B. permit the troubleshooting activity, provided the Licensed NSO makes all adjustments to the controller setpoint per LOP-RD-29, Determination of CRD System Problems.

CY require the IMD Technician to produce a work package with a troubleshooting plan prior to approving the troubleshooting activity per MA-AA-7 16-004, Conduct of Troubleshooting.

D. permit the IMD Technicians to perform the troubleshooting activity ONLY under the direct supervision of a Licensed NSO per LOP-RD-29, Determination of CRD System Problems.

~~~

~~

~

~ ~ _ _ _ _ _ _ _ _ _ ~ ~

~

~~~

~~~

~~~~~

Answer C is correct. For simple cases of unknown repair activity designated as "troubleshooting" the Troubleshooting Log, Attachment I (MA-AA-716-004, Conduct of Troubleshooting), should be used with the troubleshooting WWAR or Work Order task. Specific Operations authorization is required for the troubleshooting to commence in thefield. Simple Troubleshooting Plans - the WWAR or Work Order task and Attachment I constitutes the Troubleshooting Plan. Beyond the WWAR or Work Order approval, the First Line Supervisor/Project Manager and Shifr Supervisor approval is required on Attachment I.

Reference:

hL4-AA-716-004, Revision 02, step 1.3.2 andstep 4.10.2 2003-01 ILT NRC Exam, Version: 5 Page 101 of 105

'v minutes, then ONE operator completes the task in a 80 mremhour in 30 minutes, then the mechanic returns to remove the shielding in a 300 mrem/hour field for 20 minutes.

~-

~

98. GENERIC 2 3.2 002/GENERIC/2.3.2/2.5/2.9/SRO/HIGW648.1 O/

~

(1.OO Point) Rad Protection (l@)

has surveyed a work site anddetennined that personnel working in the area would be exposed to 300 mrem/hr gamma radiation. During an ALARA review, four options were discussed.

Which one of the following four options listed would be in alignment with ALARA principles?

A. Option 1 : TWO mechanics install 2 layers of lead shielding in a 300 mremhour field for 20 minutes, then ONE operator completes the task in a 60 mremhour field in 30 minutes, then the TWO mechanics return to remove the shielding in a 300 mremhour field for 20 minutes.

D. Option 4: ONE mechanic installs 1 layer of lead shielding in a 300 mremhour field for 20 minutes, then TWO operators completes the task in a 100 mremhour field in 30 minutes, then the mechanic returns to remove the shielding in a 300 mremhour field for 20 minutes.

Answer B is correct because Option 2 would on& expose the crew to 240 mrem.

Option 1:

Option 2:

2 x (300i60) x 20 = 200 mrem to hang shielding I x (60i60) x 30 = 30 mrem to complete task 2 x (300/60) x 20 = 200 mrem to remove shielding Total exposure to crew = 430. mrem I x (300160) x 20 = IO0 mrem to hang shielding I x (80/60) x 30 = 40 mrem to complete task 1 x (300/60) x 20 = I00 mrem to remove shielding Total exposure to crew = 240 mrem Option 3:

Option 4:

2 x (300i60) x 20 = 200 mrem to hang shielding 2 x (50160) x 30 = 50 mrem to complete task 2 x (300/60) x 20 = 200 mrem to remove shielding Total exposure to crew = 450 mrem I x (300/60) x 20 = 100 mrem to hang shielding 2 x (100/60) x 30 = 100 mrem to complete task I x (300/60) x 20 = 100 mrem to remove shielding Total exposure to crew = 300 mrem

Reference:

RP-AA-401, Revision 04, page I, step 2. I, page 4, step 5.F andpage 17, ALARA Briejkg Checklist 2003-01 ILT NRC Exam, Version: 5 Page 102 of 104

99. GENERIC 2.4.22 00 I/GENERIC/2.4.22/3.0/4.O/SROMIGH/400.00.02/

~~

(1.OO Point) Given the following plant conditions:

- Suppression Pool level is +15 inches and slowly increasing

- Suppression Chamber pressure is 11 psig and slowly increasing

- Drywell pressure is 15 psig and slowly increasing

- Suppression Pool temperature is 190°F and steady

- RPV level is -1 65 inches and decreasing

- RPV pressure is 200 psig and steady

- All control rods are full-in

- RCIC is injecting into the RPV

- RR Pumps are in Pull-To-Lock

- MSIVs are isolated When returned to service, the Unit Supervisor should direct the 2B RHR Pump to be used...

I A. Per LGA-003, start lowering Suppression Pool level to prevent exceeding the SRV Tail Pipe,

Level Limit because of the high Suppression Pool level.

B. Per LGA-003, start Drywell Spray to prevent exceeding the PSP limit because of high Suppression Chamber pressure.

C!' Per LGA-00 1, start injecting into the RPV to prevent loosing adequate core cooling because

~

of low RPV water level.

D. Per LGA-003, start Suppression Pool Cooling to prevent exceeding the HCTL because of the increasing Suppression Pool temperature.

~

Provide LGA-001 and LCA-003 to the examinee as a handout.

Answer C is correct. Possible loss of adequate core cooling exists with the given conditions and therefore RPV level is the highest priority.

The other answers are incorrect. Drywell Spray is only initiated when it is determined that you cannot restore and hold SC pressure and SP level below the PSP. Although Suppression Pool temperature is high enough that the EOP directs starting all available pool cooling, it also says do NOT use pumps needed for core cooling. Lowering pool level can wait because the SRV Tail Pipe Level Limit will not be exceeded with out level increasing several feet (not several inches).

Reference:

LGA-001, Revision 06 LGA-003, Revision 05 2003-01 ILT NRC Exam, Version: 5 Page 103 of 105

100. GENERIC 2.4.49

~~

001/GENERIC/2.4.49/4.0/4.0/SRO/MEMORYII

~~

~

(1.OO Point) Unit-1 is at 100% power and Unit-;! is in Cold Shutdown. The OA-VCNE Train i s out-of-service for scheduled maintenance and the appropriate Technical Specification time clocks per LCO 3.7.4 have been entered.

Subsequently the OB VC Supply Fan trips and can NOT be restarted.

Which one of the following actions is directed by Technical Specifications?

A!' Immediately enter and take actions per LCO 3.0.3 since two Control Room Air Filtration (CRAF) subsystems are inoperable and may be incapable of performing their intended function.

B. Immediately place the Unit-1 Reactor Mode Switch to shutdown since both Control Room Air Filtration (CRAF) subsystems are inoperable and may be incapable of performing their intended function.

C. Start the OB Emergency Makeup Unit Fan within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> since both Control Room Air Filtration (CRAF) subsystems are inoperable and may be incapable of performing their intended function.

D. No additional time clocks are required since the OB Emergency Makeup Fan will still operate without the OB VC Supply Fan and it is capable of performing its intended function.

~~~~

-~

.~ -~

~~

Answer A is correct. Per LCO 3.7.4, two CRAF systems shall be operable. With both subsystems inoperable, Condition D is entered and Required Action D. I requires immediately entering LCO 3.0.3. The other answers are therefore incorrect.

Reference:

Technical Specrfication 3.74, pages 3.7.4-1, 3.7.4-2 and 3.7.4-3 and bases page B 3.7.4-5 i

2003-01 ILT NRC Exam, Version: 5 Page 104 of 105