ML091730153
| ML091730153 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 06/22/2009 |
| From: | NRC/RGN-II |
| To: | Southern Nuclear Operating Co |
| References | |
| 50-321/09-301, 50-366/09-301 50-321/09-301, 50-366/09-301 | |
| Download: ML091730153 (161) | |
Text
Name: ______________________________ __
HLT 4 NRC Exam Form: 0 Version: 0
- 1. 2010011<5.02 001/2/2/CRDINEWIHIGHERlHT2009-301IRO/EUICMEI Following a reactor scram in which the Control Rod Drive (CRD) pump remains running, the 2CII-R600 "CRD Flow Controller" flow INDICATION indicates significantly (1) than normal.
The implication of this flow indication is that (2) will occur if the reactor scram is NOT reset.
A':" (1) higher (2) excessive reactor vessel bottom head cooldown B. (1) higher (2) CRD pump runout.
C. (1) lower (2) excessive reactor vessel bottom head cooldown D. (1) lower (2) CRD pump runout Thursday, January 15, 200910:26:42 AM 1
HL T 4 NRC Exam Link to KA: Tests whether the candidate understands how CRD flow indication is affected by a scram, and then asks the implications of this flow rate indication.
A. Correct The charging water header taps off downstream of the CRD flow element (upstream of the CRD flow control valves, 2Cll-F002A/B). When the SDV and the reactor pressures are equalized with the scram not reset, the charging water header will add to RPV inventory utilizing the cooling water flow path and leaking past the CRDM seals. This cooler water will cool the bottom head area and will eventually result in violation of TS cooldown rate limits.
B. Incorrect 1st part is correct (see "A" description) 2nd part is incorrect. Plausible since high flowrates are indications of pump runout. Pump runout is only prevented by use of orifices and a throttled valve in the RCU accumulator charging water header.
C. Incorrect 1 st part is incorrect. Plausible since this response requires the candidate to realize which side of 2Cll-F002 the flow element is on. Up stream side (with charging water header downstream of the flow element) results in a high flow indication, downstream side would result in a low flow indication.
2nd part is correct (see "A" description)
D. Incorrect 1st part is incorrect. (see "c" description).
2nd part is incorrect. Plausible because the candidate must distinguish between actual flow and indicated flow, then determine whether actual flow and indicated flow are the same. If the candidate assumes the flow element is downstream of2Cll-F002 and the charging header upstream, then a low flow indication would be expected with an actual high flow condition in the charging water header.
Thursday, January 15,2009 10:26:43 AM 2
HL T 4 NRC Exam SYSTEM: 201001 Control Rod Drive Hydraulic System K5. Knowledge of the operational implications of the following concepts as they apply to CONTROL ROD DRIVEHYDRAULIC SYSTEM: (CFR: : : 41.5/45.3)
K5.02 Flow indication..................................... RO 2.6 / SRO 2.6 Reference(s) used to develop this question:
CII-CRD-LP-00I01 Control Rod Drive System lesson plan (section II.a.d, and Fig 01)
B31-RRS-LP-00401, "Reactor ReciTe Lesson" (section ill.K.4) 34AB-C71-001-2, "Scram Procedure" Obj 18 (EO 001.013.a.07)
Reference(s) provided to the student:
None Thursday, January 15, 2009 10:26:43 AM 3
HL T 4 NRC Exam
- 2. 202001K4.07 001l2/2IRECIRCINEWIHIGHERJHT2009-301IRO/EUICMEI Unit 1 is operating at 80% power.
o Both reactor recirculation pumps are running at 78% speed o All Condensate and Condensate Booster Pumps are in service
- 4160 V AC Bus" 1 CIt is inadvertently de-energized.
Which ONE of the following describes how the Reactor Recirculation system will be affected?
A'!' Both Reactor Recirc pumps will automatically trip.
B. Both Reactor Recirc pumps will be running at 61 % speed.
C. Both Reactor Recirc pumps will be running at 78% speed.
D. The "lA" Reactor Recirc pump will automatically trip.
The "lB" Reactor Recirc pump will be running at 61 % speed.
Description of event.
A loss of 4160 "lC" will result in a loss of the Reactor Recirc AC oil pumps which causes a loss of oil pressure to the Recirc MG sets. This results in a trip of BOTH Recirc MG Sets. A unit difference exists, V2 has Adjusttable Speed Drives (ASDs) installed, VI uses Recirc MG sets.
The "lC" Condensate and "lC" Condensate Booster Pumps (CBP) will lose power when this bus is lost.
A. Correct; see description above.
B. Incorrect; see description above.
Plausible since low condensate discharge pressure will result in a #3 speed limiter. It makes sense that when a condensate pump is lost (" 1 CIt 4160 V AC Bus supplies power to the "1 CIt Condensate Pump and the "lC" CBP at rated power, low CBP suction pressure will occur, which results a #3 runback signal. If the candidate does not remember that a loss of" 1 CIt 4160 V AC bus results in a loss of the Recirc MG set AC oil pumps, this will appear to be the correct answer.
C. Incorrect; see description above.
Plausible since 2 Condensate and 2 CBPs are adequate to maintain RPV water level at this reactor power. The candidate may focus on the loss of the "lC" Condensate Pump and the "lC" CBP (which will actually be lost) without considering the impact of a loss of oil to the Recirc MG sets.
D. Incorrect; see description above Plausible if the candidate thinks the AC oil pumps have independent power supplies and that the loss of the Condensate pumps will result in a runback on the remaining Recirc Pump.
Thursday, January 15, 200910:26:45 AM 4
HL T 4 NRC Exam SYSTEM: 202001 Recirculation System K4. Knowledge of RECIRCULATION System design feature(s) and/or interlocks which provide for the following:(CFR: 41.7)
K4.07 Motor generator set trips: Plant-Specific.................... RO 2.8 / SRO 2.9 Reference(s) used to develop this question:
34AB-R23-004-l, Loss of 4l60V bus lA, lB, lC, or lD 34AB-R23-002-l, Loss of600V bus lA, lAA, lBB, or lB 34AR-602-l0l-l, Drive Motor A Trip B3l-RRS-LP-0040 1, Reactor Recirculation System lesson plan Reference(s) provided to the student:
None Thursday, January 15, 200910:26:45 AM 5
HL T 4 NRC Exam
- 3. 203000G2.4.8 00 1I2/11RHRJNEW/HIGHERJHT2009-30 llRO/EUICMEI Unit 2 was operating at 100% power when a Loss Of Coolant Accident occurred.
These conditions exist:
o The reactor has scrammed and all control rods fully inserted o RPV pressure.................. 960 psig, increasing at 4 psig/minute o Drywell CDW) pressure... 7 psig o BOTH RHR loops............ DW spray mode o RPV level........................ -145 inches, decreasing at 2"/minute o 4160 V AC bus "2E" has de-energized and cannot be re-energized lAW 34AB-R23-001-2, "Loss of 600 Volt Emergency Bus", energizing 600 VAC bus "2C" using the 4160/600V "2CD" Transformer is ________ _
A. NOT allowed until the reactor is in Mode 4 S,.. allowed since it supports the Emergency Operating Procedures C. NOT allowed because the "IB" Emergency Diesel Generator would be overloaded D. allowed, but is NOT required since all low pressure Emergency Core Cooling Systems will inject at rated flow, when required Thursday, January 15,200910:26:46 AM 6
HL T 4 NRC Exam
==
Description:==
The CD Transformer is only allowed to be used when in Mode 4 or 5 OR as required to support the EOPs. "IB" EDG loading must be considered when the "IBEDG is supplying the bus.
In this case, the CD transformer is needed to support the EOPs. DIW spray valves must be closed to ensure all ECCS water is injected to the RPV rather than being diverted to the DW.
The "2A" RHR loop will not inject at rated flow to the reactor because the DW spray valves are de-energized in the open position. This will allow water to be diverted from the core to the D/W.
In this case, the "2F" bus is being supplied from offsite power rather than the "IB" EDG, so the "IB" EDG is not running and therefore, not loaded so the consideration (at this time) of diesel loading is not an issue.
A. Incorrect; see description above.
Plausible since Mode 4 is a minimum requirement when not in the EOPs.
B. Correct; see description above.
C. Incorrect; see description above.
Plausible since the "IB" EDG loading is a limiting criteria for use of the CD Transformer. If the "IB" EDG were running (instead of the "IA" EDG) the candidate would have to take EDG loading into the decision making process for use of the CD Transformer.
D. Incorrect; see description above.
Plausible since the "2C" RHR pump will run and the injection valve will open; however, water will be diverted to the DW (see description above)
SYSTEM: 203000 RHRlLPCI: Injection Mode (Plant Specific) 2,4 Emergency Procedures 1 Plan 2.4,8 Knowledge of how abnormal operating procedures are used in conjunction with EOPs, (CFR: 41.10 1 43.5/45.13) IMPORTANCE RO 3.81 SRO 4.5 Reference(sl used to develop this question:
CP-I, Alternate Level Control, Steam Cooling & Emergency RPV Depress EOP flowchart DI-OPS-59, Operations management Expectations 34AB-R23-001-2 "Loss of 600 Volt Emergency Bus" (Pg 2 of7, step 4.3) 34S0-EII-OIO-02, "RHR System" Attach 2, RHR system electrical lineup Reference(sl provided to the student:
None Thursday, January 15,2009 10:26:48 AM 7
HL T 4 NRC Exam
- 4. 204000A4.09 001 12/21R WCUINEW /FUNDIHT2009-30 llRO/EUICMEI Unit 2 is at 2% power performing a reactor heatup and pressurization.
The Reactor Water Cleanup (RWCU) system is aligned for vessel level control (Blowdown mode to the Main Condenser).
o Reactor Water Level (RWL) begins to increase o Non-Regenerative Heat Exchanger (NRHX) Effluent temperature......... 133°F Which ONE of the following completes both of these statements?
NRHX Effluent temperature is monitored at 2HII-(1)
IA W 34S0-G31-003-2, "Reactor Water Cleanup System", 2G31-R606, Manual Controller for 2G31-F033, "RWCU Blowdown Flow Control Valve" is required to be throttled (2)
A. (1) P602 (2) open to lower RWL B~ (1) P602 (2) closed to lower NRHX effluent temperature C. (1) P614 (2) open to lower RWL D. (1) P614 (2) closed to lower NRHX effluent temperature Thursday, January 15, 2009 10:26:48 AM 8
HLT 4 NRC Exam Description The RWeU temperatures are monitored at the 2H11-P602 paneL 34S0-G31-001-2 directs the operator to throttle 2G31-F033 to maintain Non-Regenerative Heat Exchanger (NRHX) Effluent temperature below 125°F, and 34AR-602-427-2 "RWeU FILTER INLET TEMP HIGH" requires dump flow be reduced to clear the high temp alarm (130°F). The operator is using dump flow to maintain RWL during the RPV heatup. As level goes up, normally the dump flow is raised to maintain RWL within a band. In this case the temperature limits are being exceeded and dump flow must be reduced A. Incorrect; see description above 1 st part is correct 2nd part is incorrect since procedure temperature limits are being exceeded. Plausible since RWeU is currently lined up for dump flow and RPV level is increasing.
B. Correct; see description above C. Incorrect; see description above 1st part is incorrect; plausible since 2H11-P614 contains 5 digital multipoint temperature recorders and 2 sec temperature chart recorders (many temperature points recorded), including RWeU Room Temperatures, Bottom head drain (606 recorder).
2nd part is incorrect since procedure temperature limits are being exceeded. Plausible since RWeU is currently lined up for dump flow and RPV level is increasing.
D. Incorrect; see description above 1 st part is incorrect; plausible since 2H11-P614 contains 5 digital multipoint temperature recorders and 2 sec temperature chart recorders (many temperature points recorded), including RWeU Room Temperatures, Bottom head drain (606 recorder).
2nd part is correct; see description above KA:
SYSTEM: 204000 Reactor Water Cleanup System A4. Ability to manually operate andlor monitor in the control room:
(eFR: 41.7 145.5 to 45.8)
A4.09 Reactor water temperature................ RO 2.9 1 SRO 2.9 Reference(s) used to develop this question:
34AR-602-427-2, "RWeU Filter Inlet Temp High" Annunciator Response Reference(s) provided to the student:
None Thursday, January 15, 200910:26:50 AM 9
HL T 4 NRC Exam
- 5. 205000K2.01 001/2/1IRHRlBANK MOD/FUNDIHT2009-301IROIEUICMEI Unit 1 is in Mode 4 with "IC" and "ID" Residual Heat Removal (RHR) pumps running in Shutdown Cooling.
o An over-current condition develops on Emergency Bus 4160VAC "IF" which causes it to de-energize.
Which ONE of the following indicates the status of Shutdown Cooling?
A~ "1 COl RHR Pump has stopped "ID" RHR Pump has stopped B. "IC" RHR Pump is running "ID" RHR Pump has stopped C. "IC" RHR Pump has stopped "ID" RHR Pump is running D. "IC" RHR Pump is running "ID" RHR Pump is running
==
Description:==
The "IF" bus supplies power to the "IC" and "ID" RHR pumps. When this bus is lost, those pumps will no longer be operating.
Plausibility:
. The KA for this question requires the candidate to know the power supplies to the RHR pumps when RHR is in SDC mode of operation. Plausibility for distracters is based on whether or not the candidate actually does remember the power supply lists, and whether the candidate confuses the various pumps supplied by the buses (i.e. RHRSW pump "ID" is powered by 4160 VAC bus "IG" while RHR pump "ID" is powered from 4160 VAC bus "IF")
A. Correct; see description and plausibility statement above B. Incorrect; see description and plausibility statement above C. Incorrect; see description and plausibility statement above D. Incorrect; see description and plausibility statement above Thursday, January 15, 2009 10:26:51 AM 10
HL T 4 NRC Exam SYSTEM: 205000 Shutdown Cooling System (RHR Shutdown Cooling Mode)
K2. Knowledge of electrical power supplies to the following: (CFR: 41.7)
K2.01 Pump motors................... RO 3.1 * /3.1
- Reference(sl used to develop this question:
E11-RHR-LP-00701, Pg 81 EO 006.001.a.02 34S0-E11-01O-1, Residual Heat Removal System section (7.4.4)
Reference(sl provided to the student:
None Thursday, January 15, 200910:26:51 AM 11
HL T 4 NRC Exam
- 6. 206000KS.08 00 1 1211/HPCI VACUUM BREAKERJNEW/HIGHERlHT2009-30 1 IRO/ELJ/CMEI Unit 2 is operating at 100% power.
o 34SV -E41-002-2, "HPCI PUMP OPERABILITY" has just been completed.
o When the High Pressure Coolant Injection (HPCI) system is secured, BOTH HPCI Turbine Exhaust Vacuum Breaker check valves (2E41-F102 and F103) fail in the OPEN position
- If the HPCI system receives an automatic start signal, HPCI will (1) and (2)
A. (1) experience a significant water hammer event in its exhaust line (2) this event can be diagnosed from control room indications B:" (1) discharge steam directly to the Suppression Chamber air space (2) this steam discharge can be isolated by closing 2E41-F104 C. (1) experience a significant water hammer event in its exhaust line (2) this event can NOT be diagnosed from control room indications D. (1) discharge steam directly to the Suppression Chamber air space (2) this steam discharge can ONLY be isolated by tripping HPCI Thursday, January 15, 200910:26:52 AM 12
HL T 4 NRC Exam Description of this condition The HPCI exhaust line vacuum breakers are installed to prevent siphoning water into the HPCI exhaust line. The line connects to the Torus air space. Both check valves failed in the open position will allow steam to pass through the line directly into the suppression chamber air space.
The HPCI exhaust line has two motor operated isolation valves (2E41-FI04 and FIll) which can be operated from the Main Control Room (P601lP602).
Plausibility - Water hammer will occur if these valves failed in the closed position rather than the open position. This requires the candidate to demonstrate understanding of how the Vacuum Breakers function. If the candidates reverse their thinking about how this vacuum breaker line functions, and considering that there are no indications of the vacuum breaker check valves in the Main Control Room, some candidates may/will assume that the event can NOT be diagnosed from there. Others may assume that indication of exhaust pressure fluctuations or torus pressure increase will give evidence of the water hammer event.
A. Incorrect; see description and plausibility statement above B. Correct; see description above C. Incorrect; see description and plausibility statement above D. Incorrect; see description and plausibility statement above SYSTEM: 206000 High Pressure Coolant Injection System K5. Knowledge of the operational implications of the following concepts as they apply to HIGH PRESSURE COOLANT INJECTION SYSTEM: (CFR: 41.5/45.3)
K5.08 Vacuum breaker operation: BWR-2,3,4.................. RO 3.0/ SRO 3.2 Reference(s) used to develop this question:
E41-HPCI-LP-00501, HPCI system lesson plan/simplified one-line diagram.
EO 005.001.a.04 Reference(s) provided to the student:
None Thursday, January 15, 200910:26:53 AM 13
HL T 4 NRC Exam
- 7. 206000K6.02 00 1 1211 !HPCI!NEW !HIGHERlHT2009-30 llRO/EUICMEI Unit 2 is operating at 100%.
o 2R24-S022, "250 VDC RX BLDG ESSEN. 2B" de-energizes (remains de-energized) o A transient results in a scram and a High Pressure Coolant Injection (HPCI) automatic initiation signal.
Which ONE of the following describes how the HPCI system will respond to the automatic initiation signal?
HPCI will ______ _
A. automatically start and operate normally B. automatically start; however, only local control will be possible C. NOT automatically start, but can be manually started D" NOT automatically start and can NOT be manually started Description of how HPCI is impacted by this loss of DC HPCI can NOT be started since the power supply to several HPCI components has been lost, including power to the Aux Oil pump. Oil pressure supplied by the Aux Oil pump is required to open the HPCI governor valve. If the governor valve is not opened, the turbine will not tum.
HPCI cannot be started in this condition. If HPCI had been running prior to the loss of 2R24-S022, it would continue to run.
A. Incorrect; see description above.
Plausible if the candidate does not know which components are supplied by this DC bus.
If HPCI were running before the DC bus loss, the second part of this distractor would be a correct answer.
B. Incorrect; see description above.
Plausible if the candidate does not know which components are supplied by this DC bus.
Guidance for local operation of HPCI is provided in 31 RS-E41-00 1-2, "HPCI Operation From Outside The Control Room" If HPCI were already operating when the DC power was lost, HPCI operation would continue, however, local manual control would be the only way to control the system valves.
C. Incorrect; see description above.
Plausible if the candidate does not know which components are supplied by this DC bus or if the HPCI and RCIC systems are confused (RCIC can be manually started without DC power).
D. Correct; see description above.
Thursday, January 15, 200910:26:55 AM 14
HL T 4 NRC Exam SYSTEM: 206000 High Pressure Coolant Injection System K6. Knowledge of the effect that a loss or malfunction of the following will have on the HIGH PRESSURE COOLANT INJECTION SYSTEM: (CFR: 41.7/45.7)
K6.02 D.C. power: BWR-2,3,4................... RO 3.3 / SRO 3.7*
Reference(s) used to develop this question:
31 RS-E41-00 1-2, HPCI Operation From Outside the Control Room 34AB-R22-001-2, "Loss of DC Buses" E41-HPCI-LP-00501, "HPCI Injection System" lesson plan R42-ELECT-LP-02704 "DC Electrical" EO 200.018.a.01 Reference(s) provided to the student:
None Thursday, January 15,200910:26:55 AM 15
HL T 4 NRC Exam
- 8. 209001A4.04 OOl/2/l/CORE SPRAY CSINEWIFUNDIHT2009-301IRO/EUICMEI Unit 1 is at 100% power.
o Both loops of Core Spray (CS) are in their normal standby lineup.
o CS receives an automatic start signal Which ONE of the following completes BOTH of these statements?
In their normal standby lineup, the CS Minimum Flow valves (2E21-F031A1B) are normally (1)
The CS Minimum Flow valves (2E21-F031A1B) will remain in their standby lineup until (2)
A't (1) open (2) discharge flow exceeds 950 gpm B. (1) open (2) the CS pump motor breaker is closed C. (1) closed (2) discharge flow exceeds 950 gpm D. (1) closed (2) the CS pump motor breaker is closed Thursday, January 15, 200910:26:56 AM 16
HL T 4 NRC Exam
==
Description:==
The CS minimum flow valves are nOmially in the OPEN position.
The min flow valves close when system flow goes above 950 gpm.
A. Correct; see description above B. Incorrect; see description above 1 st part is correct 2nd part is plausible since power available (pump breaker closed) to the pump motor is an input to the discharge valve logic. Additionally, the RHR minimum flow valves open when a low flow condition <<2310 gpm) exists as soon as the pump breaker is closed.
C. Incorrect; see description above 1 st is plausible since HPCI and RCIC min flow valves are normally closed and the candidate may confuse which ECCS systems normally have open vs closed min flow valves.
2nd part is correct D. Incorrect, see description above 1 st is plausible since HPCI and RCIC min flow valves are normally closed and the candidate may confuse which ECCS systems normally have open vs closed min flow valves.
2nd part is plausible since power available (pump breaker closed) to the pump motor is an input to the discharge valve logic. Additionally, the RHR minimum flow valves open when a low flow condition <<2310 gpm) exists as soon as the pump breaker is closed.
SYSTEM: 209001 Low Pressure Core Spray System A4 Ability to manually operate and/or monitor in the control room:
(CFR: 41.7/45.5 to 45.8)
A4.04 Minimal flow valves............ RO 2.9/ SRO 2.9 Reference(s) used to develop this question:
E21-CS-LP-00801 Core Spray lesson plan E11-RHR-LP-00701, "RHR System" lesson plan 34SV-SUV-018-1, ECCS Status Check EO 008.002.a.05 State the interlocks associated with the following CS valves; a. Min Flow Valve F031 Reference(s) provided to the student:
None Thursday, January 15, 2009 10:26:57 AM 17
HL T 4 NRC Exam
- 9. 209001A4.05 001l2/1/CORE SPRAY CSINEWIHIGHERlHT2009-301IROIEU/CMEI Unit 2 is operating at 100% power, with HPCI tagged out for maintenance.
A trip of both reactor feedwater pumps results in a low reactor water level with conditions occurring as indicated in the following time line:
Time (minutes)
+0 min.
Reactor scram, all rods fully insert
+S mm.
Drywell pressure:
0.7 psig.
+S min.
Reactor water level: -110 inches, decreasing at O.S inch/minute.
+S min.
Reactor pressure:
960 psig.
+6 min.
Both Core Spray (CS) systems are in their normal standby lineups.
lAW 30AC-OPS-003-0, "Plant Operations", which ONE of the following describes the MINIMUM required operator actions (if any)?
At manually start both CS pumps (only)
B. manually start both CS pumps and manually open discharge valves 2E21-F004A & F004B (only)
C. manually start both CS pumps and manually open discharge valves 2E21-FOOSA & FOOSB (only)
D. no manual operation ofCS is required at this time, continue to monitor until system operation is required Thursday, January 15,200910:26:58 AM 18
HL T 4 NRC Exam
==
Description:==
Core Spray pumps should have automatically started at -101 inches, but failed to do so. NMP-OS-007 requires the starting ofESF systems when they fail to auto start.
2E21-F004 is normally open in a standby lineup, so the action of manually opening the valve is not required. Interlocks prevent F005 from being manually opened at this pressure when 2E21-F004 valve is already opened. Starting the pump(s) is the MINIMUM action required.
A. Correct; see description B. Incorrect; see description 1 st part is correct.
2nd part is not correct since 2E21-F004 is normally open in a standby lineup, so the action of manually opening the valve is not required. Plausibility stems from the fact that the candidate must know which discharge valve is normally closed and that manual start section of the system operating procedure directs the operator to open F005 after starting theCS pump. If the candidate thinks the F004 valve is normally closed in standby, it is reasonable for them to think opening F004 is required (i.e. F004 confused with F005) by procedure.
C. Incorrect; see description 1 st part is correct, NMP-OS-007 requires the starting of ESF systems when they fail to auto start.
2nd part is not correct since interlocks prevent F005 from being manually opened at this pressure when 2E21-F004 valve is already opened. Plausible since the manual start ofCS section of the system operating procedure directs the operator to open 2E21-F005 after starting the CS pump.
D. Incorrect; see description NMP-OS-007 requires the starting ofESF systems when they fail to auto start. Plausibility is that the candidate may not have recognized the system has exceeded a start signal and failed to auto start, and in this case, differentiating between the procedure setpoint of -101 inches and the TS setpoint of -113 inches.
SYSTEM: 209001 Low Pressure Core Spray System A4 Ability to manually operate and/or monitor in the control room:
(CFR: 41.7/45.5 to 45.8)
A4.05 Manual initiation controls.................. RO 3.8 / SRO 3.6 Reference(s) used to develop this question:
34S0-E2l-001-2 Core Spray System procedure Unit 2 TS E21-CS-LP-00801 Core Spray Lesson Plan EO 008.002.a.Ol From a list, select the actions necessary to manually start the CS system per NMP-OS-007, section 5.9.3.3 Reference(s) provided to the student:
None Thursday, January 15,2009 10:27:00 AM 19
HL T 4 NRC Exam
- 10. 211000K4.03 001l2/1/SLCINEWIFUND/HT2009-301IRO/EUICMEI Concerning the Unit 1 Standby Liquid Control (SLC) system, which ONE of the following completes both of these statements?
During normal operation, the primary means of ensuring that sodium pentaborate remains in solution inside the SLC tank is by the use of (1)
This ensures that the reactor will remain in at least (2) shutdown condition when SLC tank level has decreased to 32%, following SLC initiation during an Anticipated Transient Without a Scram (ATWS), regardless of control rod position.
A"!' (1) tank heaters (2) Hot B. (1) an air sparger (2) Hot C. (1) tank heaters (2) Cold D. (1) an arr sparger (2) Cold Thursday, January 15,2009 10:27:00 AM 20
HL T 4 NRC Exam
==
Description:==
The SLC system uses tank heaters, boron concentration and water volume to maintain boron in solution. A bubbler level detector is used to determine level in the boron tank.
An air sparger inside the boron tank is used to mix boron during boron addition and prior to taking samples. A caution tag is used to ensure the SLC pump is not operated when the air sparger is in use since it will cause pump cavitation.
35% SLC tank level corresponds to Hot Shutdown boron weight 14% SLC tank level corresponds to Cold Shutdown boron weight.
A. Correct; see description.
B. Incorrect; see description.
The 1 st part is plausible since an air sparger inside the boron tank is used to mix boron during boron addition and prior to taking samples.
The 2nd part is correct.
C. Incorrect; see description The 1 st part is correct.
The 2nd part is plausible since the candidate must recall from memory the SLC tank level that corresponds to Hot vs. Cold Shutdown Boron Weight.
D. Incorrect; see description.
The 1 st part is plausible since an air sparger inside the boron tank is used to mix boron during boron addition and prior to taking samples.
The 2nd part is plausible since the candidate must recall from memory the SLC tank level that corresponds to Hot vs. Cold Shutdown Boron Weight.
SYSTEM: 211000 Standby Liquid Control System K4. Knowledge of STANDBY LIQUID CONTROL SYSTEM design feature(s) and/or interlocks which provide for the following: (CFR: 41.7)
K4.03 Keeping sodium pentaborate in solution............. RO 3.8 / SRO 3.9 Reference(s) used to develop this question:
C41-SBLC-LP-01101 Standby Liquid Control lesson plan (EO 011.001.a.lO)
EOP-CURVES-LP-2306, "EOP Curves and Limits" EOP-CP3-LP-20327, "Level/Power Control (CP-3)" lesson plan (EO 201.092.a.02)
EOP-RCA-LP-20328, "RPV Control - ATWS (RCA) lesson plan (EOs 201.070.a.04 &
201.071.a.18)
Reference(s) provided to the student:
None Thursday, January 15, 2009 10:27:02 AM 21
HL T 4 NRC Exam
- 11. 212000Al.07 OOl/2/1IRPS CRDINEWIFUNDIHT2009-301IROIEUICMEI Unit 2 was operating at 100% power o Control rod 26-35 is at notch position 48 o The Rod Scram Switch for control rod 26-35, at 2HII-P610, is placed into the "DOWN (scram)" position o 2 seconds later, the switch is placed in the "UP (normal)" position.
Which ONE of the following is the expected Rod Position Indication at 2HII-P603 for control rod 26-35 after these actions are performed?
A. blank (over-travel)
B. 00 C,- 24 D.46 Thursday, January 15, 200910:27:02 AM 22
HL T 4 NRC Exam
==
Description:==
Toggling this switch will de-energize the scram pilot solenoid valves for this rod only, and only when the switch is in the "down (scram)" position. Reactor pressure is adequate to cause rods to insert without a CRD pump running.
The rods will only move as long as the scram pilot solenoid valves are de-energized and 2 seconds is not long enough to fully insert control rods in this condition. When the switch is returned to the normal (up) position, the scram pilot solenoid valves will re-energize and the control rod will settle at the next notch position. There is not a seal in feature when individually scramming control rods. In this case, 2 seconds is not sufficient to move the control to a full in position.
Max. insertion time for a control rod is 7 seconds. 31EO-EOP-103-2 requires the switch be maintained in the "Up (scram)" position until rod movement stops to ensure the rod is fully inserted. Per TS, it takes up to 3.35 seconds to cause a rod that is at position48 to insert to position 06.
A. Incorrect; see description above.
Plausible since a scram does normally seal in and a scrammed rod window indication is initially blank until the scram is reset, or the SDV fills and CRDM piston dp equalizes.
B. Incorrect; see description above.
Plausible since 00 is the normal position the candidates see as rod position for a fully inserted control rod.
C. Correct; see description above.
D. Incorrect; see description above.
Plausible since this position corresponds to the rod position the rod would settle at if it were moved at normal drive speed.
Thursday, January 15, 200910:27:03 AM 23
HL T 4 NRC Exam SYSTEM: 212000 Reactor Protection System At. Ability to predict and/or monitor changes in parameters associated with operating the REACTOR PROTECTION SYSTEM controls including: (CFR: 41.5 /45.5)
A1.07 Rod position information........... RO 3.4 / SRO 3.4 Reference(sl used to develop this question:
C7l-RPS-LP-OlOOl RPS lesson plan EO 010.020.a.02 Given a list, identify the statement that describes the effect on individual CRDMs of placing the individual control rod scram test switches in scram.
Cll-CRDM-LP-00102, CRDM lesson plan U2 TS table 3.1.4-1 3lEO-EOP-l03-2, "EOP Control Rod Insertion Methods" Reference(sl provided to the student:
None Thursday, January 15,200910:27:04 AM 24
HLT 4 NRC Exam
- 12. 215001A3.03 001l2/2ffIP/BANKMODIHIGHERJHT2009-301IRO/ELJ/CMEI Unit 1 is operating at 100% power with the "A" Traversing In-Core Probe (TIP) inserted in the core to perform 57CP-C51-01O-0, "TIP Flux Probing Monitor".
A transient occurs on Unit 1 with the following plant conditions:
Reactor pressure..................... 900 psig and stable Reactor level (lowest)............ -20 inches and slowly increasing Drywell pressure.................... 1.5 psig and stable Drywell temperature.............. 129°F Which ONE of the following completes the statement below?
With NO operator action, the "A" TIP will ______ _
A. remain in the core with its Ball valve OPEN and its Shear valve ACTUATED.
B. remain in the core with its Ball valve OPEN and its Shear valve NOT ACTUATED..
C. withdraw to the "in-shield" position with its Ball valve CLOSED and its Shear valve ACTUATED..
D~ withdraw to the "in-shield" position with its Ball valve CLOSED and its Shear valve NOT ACTUATED.
==
Description:==
The TIP receives a signal to withdraw to the "in-shield" position upon receipt of a group 2 signal (1.85 psig DW press or +3" RPV water level. The ball valve auto closes when the probe is fully withdrawn.
A. Incorrect; see description above Plausible since the TIP shear valve is actuated if the TIP were stuck in the core with evidence of a leak via that pathway..
B. Incorrect; see description above Plausible since the candidate must know that the withdrawal occurs at + 3" rather than -35",
-60" or -101" inches C. Incorrect; see description above Plausible since the TIP shear valve is actuated if the TIP were stuck in the core with evidence of a leak via that pathway..
D. Correct; see description above Thursday, January 15, 2009 10:27:05 AM 25
HL T 4 NRC Exam SYSTEM: 215001 Traversing In-Core Probe A3. Ability to monitor automatic operations ofthe TRAVERSING IN-CORE PROBE including: (CFR: 41.7/45.7)
A3.03 Valve operation: Not-BWRI...... RO 2.5* / SRO 2.6*
Referencefsl used to develop this question:
T23-PC-LP-01301 Primary Containment lesson plan EO (013.036.A.02, 013.039.a.04)
EO (013.007.a.03, 013.012.a.02, 013.019.a.02, 013.020.a.02)
Referencefsl provided to the student:
None Thursday, January 15, 2009 10:27:06 AM 26
HL T 4 NRC Exam
- 13. 215002K6.04 001l2/2IRBMlBANKIFUNDIHT2009-301IRO/EUICMEI Unit 2 APRM channel "B" was bypassed with the reactor at rated conditions.
What effect will this action have on the RBM system?
- The "B"RBM ------------------
A. will be INOP due to a loss of its reference APRM.
B1 reference APRM will automatically shift to APRM channel D C. reference APRM will automatically shift to APRM channel C D. reference APRM will automatically shift to APRM channel A
==
Description:==
When the RBM loses its primary reference APRM it automatically selects the first alternate APRM. D APRM is the fIrst alternate, second alternate is C APRM.
A. Incorrect The RBM automatically selects the first alternate APRM B. Correct D APRM is the first alternate, second alternate is C APRM.
C. Incorrect C APRM is the second alternate, not the first alternate D. Incorrect A APRM is only used as a reference for the A RBM channel SYSTEM: 215002 Rod Block Monitor System K6. Knowledge of the effect that a loss or malfunction of the following will have on the ROD BLOCK MONITOR SYSTEM: (CFR: 41.7 / 45.7)
K6.04 APRM reference channel: BWR-3,4,5........ RO 2.8 / SRO 3.0 Referencefsl used to develop this question:
C51-PRNM-LP-01203, "Power Range Neutron Monitoring System" Referencefsl provided to the student:
None Thursday, January 15,2009 10:27:07 AM 27
HL T 4 NRC Exam
- 14. 215003A1.03 001l2/1lIRM RPSIBANKMODIHIGHERlHT2009-301IRO/EUICMEI A Unit 1 Reactor Startup is in progress.
IRMs read as follows:
IRM Reading Range Reading Range A
38/125 8
B 35/125 8
Which ONE of the following predicts how the plant will respond to the following switch manipulations?
o IRM "A" range switch is placed in the RANGE 7 position.
o IRM "B" range switch is placed in the RANGE 7 position.
The IRMs will _____ _
A. NOT initiate any automatic action B. initiate a rod block (only)
C'!" initiate a rod block and a half scram (only)
D. initiate a full scram Thursday, January 15,2009 10:27:08 AM 28
HL T 4 NRC Exam
==
Description:==
The candidate must know or correctly calculate the following setpoints to correctly answer the question:
Rod block on the odd numbered range switches: 25.6/40 Rod block on the even numbered range switches: 80/125 Reactor Scram signal on the odd numbered range switches: 36.8/40 Reactor Scram signal on the even numbered range switches: 115/125 The candidate must know that the reading on an odd numbered range is numerically the same as the next higher range.
In this question, the "A" channel is above the scram setpoint on range 7 (reading 38/40) and the "B" channel is above the rod block setpoint, but below the scram setpoint on range 7 (reading 35/40)
Rod block setpoint is added to provide plausibility for the 3rd distractor A. Incorrect; see description above Plausibility is based on the candidates knowledge of how the indications and setpoints are impacted by changing ranges on the IRMs. The candidate must perform the correct calculation of the setpoints candidates often have trouble transitioning from a 0 to 40 scale from a 0 to 125 scale).
B. Incorrect; see description above Plausibility is based on the candidates knowledge of how the indications and setpoints are impacted by changing ranges on the IRMs. The candidate must perform the correct calculation of the setpoints candidates often have trouble transitioning from a 0 to 40 scale from a 0 to 125 scale).
C. Correct; see description above D. Incorrect; see description above Plausibility is based on the candidate knowledge of how the indications and setpoints are impacted by changing ranges on the IRMs. The candidate must perform the correct calculation of the setpoints (i.e. 40 on a 125 scale rather than 40 on a 100 scale which would indicate a scram setpoint of 32/40 rather than 36.8/40) (Note, a rod block would be in if a full scram did occur (SDV level), but not be due to the IRMs).
Thursday, January 15, 2009 10:27:09 AM 29
HL T 4 NRC Exam SYSTEM: 215003 Intermediate Range Monitor (IRM) System At. Ability to predict and/or monitor changes in parameters associated with operating the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM controls including:
(CFR: 41.5/45.5)
A1.03 RPS status............. RO 3.6/ SRO 3.7 Reference(s) used to develop this question:
34AR-603-221-1 IRM Bus A Upscale Trip or Inop ARP 34AR-603-221-1 IRM Upscale ARP C51-lRM-LP-01202, Intermediate range Monitors EO (012.003.C.09, 012.007.A.02)
EO (012.003.C.OlO, 012.007.B.02)
Reference(s) provided to the student:
None Thursday, January 15, 2009 10:27:09 AM 30
HL T 4 NRC Exam
- 15. 215004K6.01 OO1l2/11SRM RPSINEWIHIGHERlHT2009-301IRO/EUICMEI Unit 1 is starting up.
o All Intermediate Range Monitors (IRM) are on RANGE 4 (reading on-scale) o All Source Range Monitors (SRM) are all partially withdrawn and indicating 5 x 104 counts per second ( cps) o The" IA" SRM High Voltage power supply experiences a failure that causes its voltage decrease from 350 VDC to 20 VDC, resulting in the "IA" SRM count rate decreasing to 4 cps A Control Rod Block _
....... (,.".,1).<--_ occur because _
....... (,=,2).<--_
A. (1) will (2) the "IA" SRM is partially withdrawn S,.. (1) will (2) of an INOP trip on the "IA" SRM C. (1) will NOT (2) of the IRM range switch positions D. (1) will NOT (2) the "IA" SRM is reading 4 cps Thursday, January 15, 200910:27:10 AM 31
HL T 4 NRC Exam
==
Description:==
SRM high voltage low will result in an Inop rod withdraw block.
This rod block is bypassed when IRMs are on range 8 or higher.
The partially withdrawn SRM detector rod block is bypassed because the IRM range switches positioned on a range are above range 3.
The SRM downscale is bypassed when the IRM range switches positioned on a range are above Range 3.
TS operability requirements is 3 cps for the downscale rod block setpoint.
A. Incorrect; see description above 1 st part is correct.
2nd part is plausible since a partially withdrawn SRM detector will result in a rodblock if the rRM range switches are on range 1 or 2.
B. Correct; see description above C. Incorrect; see description above Both parts are incorrect.
Plausible since the candidate may focus on the 4 cps (downscale) SRM indication (overlook the Inop due to the high voltage low condition). The downscale rod block is bypassed since the range switches are on range 4.
D. Incorrect; see description above Both parts are incorrect.
Plausible since the indication is above the required TS limit of 3 cps for the downscale rod block.
SYSTEM: 215004 Source Range Monitor (SRM) System K6. Knowledge of the effect that a loss or malfunction of the following will have on the SOURCE RANGE MONITOR (SRH) SYSTEM: (CFR: 41.7 / 45.7)
K6.05 Trip Units........................... RO 2.6 / SRO 2.8 Reference(sl used to develop this question:
C51-SRM-LP-01201 SRM lesson plan Unit 1 TS Surveillance Requirement SR 3.3.1.2.4 Reference(sl provided to the student:
None Thursday, January 15, 2009 10:27:11 AM 32
HL T 4 NRC Exam
- 16. 215005G2.1.20 00 1/21 1 IOPRMlNEW IHIGHERlHT 2009-3011ROIEU/CMEI Unit 1 is operating at 40% power in Single Recirc Loop Operation (SLO).
o The "lB" Reactor Recirculation Pump is in service o The team is performing actions to re-start the "lA" Reactor Recirculation pump o Both Circulating Water pumps are in service o All Oscillation Power Range Monitors (OPRMs) are inoperable due to a calibration error o The breaker supplying power to "lB" 4160 V AC bus trips.
Which ONE of the following describes the required procedural actions?
A":! Manually scram the reactor.
B. Maintain current power level and re-start the "lA" Recirculation Pump.
C. Reduce reactor power by inserting control rods in reverse order of the Pull Sheets to stabilize Main Condenser Vacuum >25" Hg Vac.
D. Reduce reactor power by inserting control rods in reverse order of the Pull Sheets as required to remain in the allowable region of the Power to Flow map.
==
Description:==
When 4160 "B" supply breaker trips, the only operating Reactor Recirc pump trips.
A condition which requires a manual scram then exists because all OPRMs are inoperable. A shutdown is not immediately required when OPRMs are operable (allowed to operate for 12 hrs per TS). One Circ Water pump continues to operate and will be adequate to maintain vacuum for this power level.
A. Correct; see description above.
B. Incorrect; see description above.
Plausible since restarting a Recirc Pump is a correct action if the OPRMs were operable and a scram is not inserted.
C. Incorrect; see description above.
Plausible if the candidate does not know of the requirement to scram with the OPRMs inop without forced circulation and assumes that one Circ Water pump is not adequate to maintain condenser vacuum at this power leveL D. Incorrect; see description above.
Plausible since this is a required action if the unit were not scrammed (i.e. the OPRMs were operable).
Thursday, January 15, 200910:27:13 AM 33
HL T 4 NRC Exam SYSTEM: 215005 Average Power Range Monitor/Local Power Range Monitor System 2.1 Conduct of Operations 2.1.20 Ability to interpret and execute procedure steps. (CFR: 41.10/43.5/45.12) I IMPORTANCE RO 4.61 SRO 4.6 Reference(s) used to develop this question:
B31-RRS-LP-00401, "Reactor Recirculation System" EO 200.037.a.02 34AB,..B31-001-1 "Trip of One or Both Reactor Recirculation Pumps" 34AB-R22-004-1 "Loss of 4160V Bus lA, 1B, 1C or 1D" Reference(s) provided to the student:
None Thursday, January 15,200910:27:13 AM 34
HL T 4 NRC Exam
- 17. 215005G2.4.l1 OOll2l1/APRM EOPINEWIHIGHERlHT2009-301IRO/EU/CMEI Unit 1 is operating at 40% power when a scram signal is received.
o Some control rods do NOT fully insert o APRMs are all indicating 3% (steady) o Both Reactor Recirculation pumps are running at minimum speed o No entry conditions for the EOP flowcharts have been met or exceeded IA W 34AB-C11-005-1, "Control Rod Insertion Methods", which ONE of the following completes both of these statements?
o Overriding of all automatic scram signals to insert control rods is (1) o Procedurally, tripping of the Reactor Recirculation pumps is (2)
A. (1) allowed (2) required B. (1) allowed (2) NOT required C. (1) NOT allowed (2) required Dt (1) NOT allowed (2) NOT required Thursday, January 15, 200910:27:14 AM 35
HLT 4 NRC Exam
==
Description:==
Without EOP entry conditions being exceeded, with power less than 5%, a scram is controlled by use of abnormal procedures, even if all rods do not insert. If power were above 5%, 31EO-EOP-103-1 would be used to insert control rods. In this case 34AB-C71-001-1 directs operators to insert rods per 34AB-C 11-005-1. This procedure does NOT allow overriding of "all" automatic scram signals (the SDV high level trip is allowed to be overridden).
The scram procedure (34AB-C71-001-1) directs Recirc Pumps to be started if they are tripped.
A. Incorrect 1st part is not correct; this action is not allowed when using 34AB-CII-005-l to insert control rods. Plausible since this is a required action for repeating a manual scram when using 3IEO-EOP-I03-1 to insert control rods.
2nd part is not correct, 34AB-C71-001-1 actually requires the Recirc Pumps to be restarted if tripped, this will minimize thermal stratification within the RPV. Plausible since the Recirc pumps would be manually tripped if RCA were in use with power above 5% in order to reduce reactor power.
B. Incorrect 1st part is not correct; this action is not allowed when using 34AB-CI1-005-1 to insert control rods. Plausible since this is a required action for repeating a manual scram when using 3IEO-EOP-I03-1 to insert control rods.
2nd part is correct.
C. Incorrect 1 st part is correct 2nd part is not correct, 34AB-C71-001-1 actually requires the Recirc Pumps to be restarted if tripped, this will minimize thermal stratification within the RPY. Plausible since the Recirc pumps would be manually tripped if RCA were in use with power above 5% in order to reduce reactor power.
D. Correct Thursday, January 15, 2009 10:27:15 AM 36
HL T 4 NRC Exam SYSTEM: 215005 Average Power Range Monitor/Local Power Range Monitor System 2.4 Emergency Procedures / Plan 2.4.11 Knowledge of abnormal condition procedures. (CFR: 41.10/43.5/45.13)
IMPORTANCE RO 4.0 SRO 4.2 Reference(s) used to develop this question:
34AB-C11-005-1, Control Rod Insertion Methods 34AB-C71-001-1, Scram Procedure Reference(s) provided to the student:
None Thursday, January 15,200910:27:16 AM 37
HL T 4 NRC Exam
- 18. 217000K2.03 001l2/1IRCICINEWIFUND/HT2009-301lRO/EUICMEI Which ONE of the following is the power supply to the Unit 2 RCIC flow controller?
S,.. DC Cabinet "2A" (2R2S-S001)
C. Instrument Bus "2A" (2R2S-S064)
D. Reactor Bldg ESS MCC "2B" (2R24-S022)
==
Description:==
The RCIC controller is supplied AC power via an inverter from 12SV DC Cabinet "2A".
A. Incorrect; see description above Plausible since Vital AC supplies power so many Control Room instruments. Battery provides backup.
B. Correct, see description above C. Incorrect, see description above Plausible since Instrument Bus "2A" supplies logic power for RCIC at the RSDP.
D. Incorrect, see description above Plausible since this bus supplies HPCI components (including the HPCI flow controller via 2R24-S002) and can be confused by the candidate. If this were changed to Reactor Bldg ESS MCC "2A" (2R24-S021) it would be a correct answer.
SYSTEM: 217000 Reactor Core Isolation Cooling System (RCIC)
K2. Knowledge of electrical power supplies to the following: (CFR: 41.7)
K2.03 RCIC flow controller......................... RO 2.7* / SRO 2.8*
Reference(sl used to develop this question:
34AR-602-319-2 RCIC Inverter K603 Power Failure 34AB-R2S-002-2 Loss of Instrument Buses 34AB-R2S-001-2 Loss of Vital AC Bus E41-HPCI-LP-OOSOI HPCI lesson plan Reference(sl provided to the student:
None Thursday, January 15,200910:27:17 AM 38
HL T 4 NRC Exam
- 19. 218000Kl.06 00 1121 lIAD SIBANK MOD/HIGHERfHT2009-30 llRO/EUICMEI Unit 2 has experienced a LOSP.
The following conditions exist:
o Reactor........................ All rods in o All 4160 VAC buses.......... De-energized o RPV Pressure...................... 700 psig o RWL............................ -135 inches o Drywell Pressure............. 3 psig o ADS Inhibit Switches.......... "Normal" position o These conditions are maintained for 5 minutes Immediately after the 5 minutes has elapsed, an operator starts the "2A" Emergency Diesel.
- "2A" Core Spray (CS) pump starts.
- "2A" CS develops normal discharge pressure for shut off head operation.
Which ONE ofthe following identifies when all the ADS valves will automatically open?
A"! Immediately B. 102.5 seconds later C. 11 minutes later D. 12 minutes and 42.5 seconds later Thursday, January 15,2009 10:27:18 AM 39
HL T 4 NRC Exam Refer to logic drawing provided as a reference for developing this test item.
==
Description:==
The 102.5 second timer began timing as soon as the following conditions existed concurrently: D/W pressure> 1.85 psig and RPV level <-101". As a result, with the exception of low pressure ECCS confirmatory pressure, all other necessary logic connections are made to auto open the ADS valves without a time delay.
A. Correct; see description above B. Incorrect; see description above Plausible if the candidate assumes the CS permissive is part of the 102.5 sec timer. The low pressure ECCS pressure permissive must be present after the 102.5 sec timer times out in order for the ADS valves to auto open.
C. Incorrect; see description above The 11 minute timer initiated and sealed in as soon as RPV level dropped below -101". CS is not part of the 11 minute timer initiation logic. Plausible if the candidate assumes the CS permissive is part of the 11 min. timer.
The low pressure ECCS pressure permissive must be present after the 11 minute and 102.5 sec timers time out in order for the ADS valves to auto open.
D. Incorrect; see description above The 102.5 second timer began timing as soon as the following conditions existed concurrently:
DIW pressure> 1. 85 psig and RPV level <-101". The 11 minute timer initiated and sealed in as soon as RPV level dropped below -101".
Plausible if the candidate assumes that all conditions must be met for the timers begin to time out or if the candidate assumes the CS permissive is part of the 11 min. timer. The low pressure ECCS pressure permissive must be present after the 102.5 sec timer times out in order for the ADS valves to auto open.
SYSTEM: 218000 Automatic Depressurization System Kl. Knowledge of the physical connections and/or cause/effect relationships between AUTOMATIC DEPRESSURIZATION SYSTEM and the following:
(CFR: 41.2 to 41.9 / 45.7 to 45.8)
K1.06 Safety/relief valves.................................. RO 3.9* / SRO 3.9*
Reference(s) used to develop this question:
B21-ADS-LP-0381 "Auto Depress System" EO 038.004.a.02 B21-ADS-03801 Fig 2 & Fig 4 34S0-B21-001-2, "ADS and LLS System" procedure Reference(s) provided to the student:
None Thursday, January 15,2009 10:27:20 AM 40
HL T 4 NRC Exam
- 20. 223001K.3.09 001l212IRWLINDICATIONINEWIFUNDIHT2009-301IRO/EUICMEI Unit 2 is operating at 100% power.
o A complete loss of Drywell (D/W) Cooling results in D/W temperature increasing from 130°F to 285°F If all other parameters remain the same, which ONE of the following describes the relationship between indicated and actual reactor water level (RWL) for the 2B2l-R604 and 2C32-R606 instruments when monitoring RWL at the 2Hll-P603 panel?
2B2l-R604A and B 2C32-R606A, Band C A.
indicated = actual indicated> actual S,-
indicated> actual indicated> actual C.
indicated < actual indicated < actual D.
indicated = actual indicated < actual
==
Description:==
DW temp increasing will result in a decreasing density in the reference leg. As reference leg density decreases due to heating, with the variable leg remaining relatively constant, dp will decrease. Decreasing dp will result in a higher indicated level, even with no change in actual leveL Plausibility: Candidates historically have difficulty remembering (i.e. common misconception) the relationship between heating reference legs, dp and indicated level vs actual leveL A. Incorrect; see description above.
I st part is not correct. Plausible since these reference legs are heated.
2nd part is correct.
B. Correct; see description above C. Incorrect; see description above Both parts are incorrect.
Both parts are plausible since candidates historically have difficulty remembering (i.e. common misconception) the relationship between heating reference legs, dp and indicated level vs actual level.
D. Incorrect; see description above Both parts are incorrect.
I st part is plausible since these reference legs are heated.
2nd part is plausible since candidates historically have difficulty remembering (i.e. common misconception) the relationship between heating reference legs, dp and indicated level vs actual level.
Thursday, January 15, 200910:27:21 AM 41
HL T 4 NRC Exam SYSTEM: 223001 Primary Containment System and Auxiliaries K3. Knowledge of the effect that a loss or malfunction of the PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES will have on following: (CFR: 41.7/45.4)
K3.09 Nuclear boiler instrumentation.... RO 2.8/ SRO 3.1 Reference(sl used to develop this question:
B 11-RXINS-LP-04404, "Reactor Vessel Instrumentation" EO 200.002.a.014 Reference(sl provided to the student:
None Thursday, January 15, 200910:27:22 AM 42
HL T 4 NRC Exam
- 21. 223002K3.11 00 1I2/1IHPCIINEWIHIGHERlHT2009-30 1IROIEUICMEI Unit 2 is at 100% power with the Main Control Room Environmental Control (MCREC)
System aligned in the NORMAL mode.
The following annunciators are received:
o MAIN STEAM LINE FLOW A HIGH (603-214-2) o MAIN STEAM LINE FLOW B HIGH (603-215-2) o GROUP I SYSTEM A TRIP (603-208-2) o GROUP I SYSTEM B TRIP (603-209-2)
IAW 34S0-Z41-001-1, "Control Room Ventilation System" the MCREC System __ _
A. must be aligned to the Purge mode B. must be aligned to the Isolation mode C. may remain aligned in the Normal mode 0" must be aligned to the Pressurization mode Description; High Main Steam flow (of 169 psid) is one of the setpoints used to automatically shift the MCREC system into Pressurization mode.
A. Incorrect; see description above.
Plausible since purge mode is a valid mode ofMCREC and the candidate must know when this mode of operation is required to be used.
B. Incorrect; see description above.
Plausible since isolation mode is a valid mode ofMCREC and the candidate must know when this mode of operation is required to be used.
C. Incorrect; see description above.
Plausible since the candidate must know that high steam flow setpoint for pressurization mode is the same as the setpoint for the high steam flow annunciators to eliminate this distractor.
D. Correct; see description above.
Thursday, January 15, 2009 10:27:23 AM 43
HL T 4 NRC Exam SYSTEM: 223002 Primary Containment Isolation SystemJNuclear Steam Supply Shut-Off K3. Knowledge of the effect that a loss or malfunction of the PRIMARY CONTAINMENT ISOLATION SYSTEMINUCLEAR STEAM SUPPLY SHUT -OFF will have on following:
(CFR: 41.7 I 45.4)
K3.11 Plant ventilation................... RO 2.8 I SRO 2.9 Reference(sl used to develop this question:
T41-SCHVAC-LP-01303, SECONDARY CONTAINMENT VENTILATION, 037.008.a.02 34S0-Z41-001-1, "Control Room Ventilation System" 34AR-603-214/215-2, "MAIN STEAM LINE FLOW AlB HIGH" ARPs Reference(sl provided to the student:
None Thursday, January 15,2009 10:27:23 AM 44
HLT 4 NRC Exam
- 22. 226001K2.02 OOl/2/2/RHR E1IIBANKlHIGHERlHT2009-301IRO/EUICMEI Unit 1 was operating at 100% power.
o The Alternate Supply Breaker to 4160VAC bus "IE" is tagged out.
A loss of Startup Transformer (SAT) "ID" occurred.
o Torus Pressure reaches 3 psig during the transient.
o "IA" and "IB" RHR pumps running, both being used to spray the Torus.
The power supply for the 4160 V AC bus to the:
- "IB" RHR Pump is (2)
A. (1) SAT "IC" (2) SAT "IC" B~ (1) its associated EDG (2) SAT "IC" C. (1) SAT "IC"
- 0) its associated EDG D. (1) its associated EDG (2) its associated EDG Thursday, January 15, 2009 10:27:24 AM 45
HL T 4 NRC Exam
==
Description:==
Startup auxiliary transformer (SAT) "ID" is the normal supply to all VI emergency buses. SAT" 1 C is the alternate supply to the emergency buses. EDGs are the emergency supply in case there is a loss of both normal and alternate supplies. Ordinarily the emergency buses automatically transfer to alternate, however in this case the "IE" (power to the "IA" RHRpump) bus alternate supply breaker is tagged out, so when the "ID" SAT is lost, the "IA" EDG will start and supply the "IE" emergency bus. The "IG" emergency bus (power supply to the "1 B" RHR pump) will be supplied power via the alternate supply breaker (SAT "IC").
A. Incorrect; see description above.
1 st part is not correct. Plausible if the candidate thinks that SAT" 1 C" is the normal supply to the emergency buses.
2nd part is correct.
B. Correct; see description above.
1 st part is correct.
2nd part is correct C. Incorrect; see description above.
1 st part is not correct; plausible since the examinee must know which bus supplies power to the pump to eliminate this distractor.
2nd part is not correct; plausible since the examinee must know which bus supplies power to the pump to eliminate this distractor.
D. Incorrect; see description above.
1 st part is correct.
2nd part is not correct; plausible since all VI EDGs start when SAT "ID" is lost. The candidate may assume they start and tie to the emergency buses.
SYSTEM: 226001 RHRlLPCI: Containment Spray System Mode K2. Knowledge of electrical power supplies to the following: (CFR: 41.7)
K2.02 Pumps.............................. RO 2.9* / SRO 2.9*
Reference(s) used to develop this question:
R22-4160 VAC-LP-02702 4160 VAC lesson plan, EO 200.017.aa.Ol 34S0-R22-001-1, "4160 VAC System" procedure Reference(s) provided to the student:
None Thursday, January 15,2009 10:27:25 AM 46
HL T 4 NRC Exam
- 23. 233000G2.2.25 OOl/2/2/FPCINEWIFUNDIHT2009-301IROIEUICMEI Unit 1 Fuel Pool level is required to be lowered, while Fuel movement is in progress.
lAW with Tech Spec Limiting Condition for Operation (LCO) 3.7.8, Spent Fuel Storage Pool Water Level, which ONE of the following is the LOWEST level that will still meet LCO 3.7.8 WITHOUT entering a Required Action Statement (RAS), when lowering the fuel pool water level?
A. 23.1 feet B. 22.1 feet C!' 21.1 feet D. 20.1 feet
==
Description:==
During movement of irradiated fuel in the spent fuel pools, Ul TS requirement for fuel pool level is >= 21 ft over top of irradiated assemblies seated in the spent fuel storage racks.
A. Incorrect; see description above.
Plausible since 23 feet is the limit for minimum water level maintained above irradiated fuel seated within the RPV (TS 3.9.6).
B. Incorrect; see description above.
Plausible since RHR operability requirements change abovelbelow 22ft 118 inch (TS 3.9.7).
C. Correct; see description above.
(TS 3.7.8)
D. Incorrect; see description above.
Plausible since this is the lowest level that does require a RAS.
(TS 3.7.8)
Thursday, January 15, 2009 10:27:26 AM 47
HL T 4 NRC Exam SYSTEM: 233000 Fuel Pool Cooling and Clean-up 2.2 Equipment Control 2.2.25 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits. (CFR: 41.5/41.7/43.2) IMPORTANCE RO 3.2/ SRO 4.2 Reference(sl used to develop this question:
TS section 3.7.8 TS section 3.9.6 TS section 3.9.7 Reference(sl provided to the student:
None Thursday, January 15, 2009 10:27:27 AM 48
HL T 4 NRC Exam
- 24. 239001K5.09 001l2/2IDECAY HEATINEWIFUNDIHT2009-301IROIEUICMEI Unit 2 was at 100% power when a reactor scram and a Main Turbine trip occurred.
The operators complete all steps on the following placards:
- RC-l
- RC-2
- RC-3
- TC-I SELECT the component that will be contributing to RPV decay heat removal?
A. Reactor feed pump low pressure steam supply.
S,.. Reactor feed pump high pressure steam supply.
C. Moisture Separator Reheater first stage steam supply.
D. Moisture Separator Reheater second stage steam supply.
==
Description:==
The RFP low pressure steam is isolated when the main turbine is tripped.
TC-l isolates the MSR 2nd stage by closing the RSSVs. The MSR 1st stage steam supply is isolated by the turbine stop valve closure, which is completed per TC-l. The RFP high pressure steam supply (main steam) is still available.
A. Incorrect, see description above.
(LP supply) Plausible since the RFPT does have two sources of steam. The low pressure supply is designed to be the primary steam source to the RFP at rated power. The LP control valve fully opens before the HP control valve opens.
B. Correct; see description above.
C. Incorrect, see description above.
(1 st stage MSRs) Plausible since the RSLL V s are still be open and had the operator not remembered that the RSSVs were on TC-l then this area of the MSRs would still be using steam.
D. Incorrect, see description above.
(2nd stage MSRs) Plausible because it does use main steam from the high pressure turbine.
Had the main turbine not been tripped manually or automatically it would be using steam. It also requires the candidate to remember the source ofMSR first stage heating steam.
Thursday, January 15, 2009 10:27:28 AM 49
HL T 4 NRC Exam SYSTEM: 239001 Main and Reheat Steam System K5. Knowledge ofthe operational implications of the following concepts as they apply to MAIN AND REHEAT STEAM SYSTEM: (CFR: 41.5 /45.3)
K5.09 Decay heat removal....................... RO 3.4 / SRO 3.5 Reference(s) used to develop this question:
34AB-C71-001-2, "Scram Procedure" N21-CNDFW -LP-0020 1, "Condensate and F eedwater System" lesson plan Reference(s) provided to the student:
None Thursday, January 15,2009 10:27:29 AM 50
HL T 4 NRC Exam
- 25. 239002A3.01 00 112/1 lADS LIGHTSIBANKMOD/FUNDIHT2009-301IRO/ELJ/CMEI Unit 1 has experienced a transient and the Automatic Depressurization System (ADS) automatically actuated.
o Reactor pressure decreases from 1000 psig to 50 psig.
o No ADS switches have been manipulated.
Which ONE of the following describes ALL ofthe lights that are illuminated for each ADS valve?
The Red illuminated.
A. light ONLY is B. and Green lights are C,- and Amber lights are D.,Amber and Green lights are Thursday, January 15,200910:27:29 AM 51
HL T 4 NRC Exam
==
Description:==
When ADS actuates, the associated red light illuminates indicating power is sent to the solenoid valve. The red light remains energized until operator action is taken to reset the ADS logic.
The SRV Amber light is actuated by a SRV tail pipe pressure switch, and this light is sealed in until reset by a switch on the IHll-P602 panel.
The green light is automatically extinguished by the ADS logic when ADS actuates.
Plausibility; candidates historically confuse ADS light indications vs LLS light indications. The amber light will not illuminate if reactor pressure decreases to below the mid 200s pressure range when an SRV is opened..
A. Incorrect, see description above.
Plausible if the candidate assumes the amber light automatically extinguishes when tailpipe pressure is decreases to below 85 psig. This would be a correct answer if ADS had initiated with low reactor pressure (below mid 200 psig range) since the amber light does not illuminate at RPV pressures below this pressure range.
B. Incorrect, see description above.
Plausible if the candidate confuses the difference between ADS and LLS indications and assumes the amber light automatically extinguishes when tailpipe pressure decreases to below 85 psig. The green light remains illuminated when LLS actuates.
C. Correct, see description above.
D. Incorrect, see description above.
Plausible if the candidate confuses the difference between ADS and LLS indications for ADS and sensor actuated indications, but remembers that the amber light seals in. The green light remains illuminated when LLS actuates.
Thursday, January 15,200910:27:30 AM 52
HL T 4 NRC Exam SYSTEM: 239002 Relief/Safety Valves A3. Ability to monitor automatic operations of the RELIEF/SAFETY VALVES including:
(CFR: 41.7/45.7)
A3.01 SRV operation after ADS actuation.............. RO 3.8* / SRO 3.9*
Reference(s) used to develop this question:
B21-ADS-LP-03801, "Automatic Depressurization System" lesson plan EO 038-002.a.02 34S0-B21-001-1, "ADS and LLS Systems" procedure Reference(s) provided to the student:
None Thursday, January 15, 200910:27:31 AM 53
HL T 4 NRC Exam
- 26. 241 000K1.25 001l2/2!TURB RUNBACKIBANK MODIHIGHERlHT2009-301IRO/EUICMEI Unit 1 is starting up, currently at 35% power.
The Main Generator system parameters are listed below:
o Power......... 300 MW e o Voltage...... 24,000 volts o Current....... 7,000 amps Stator Cooling system parameters associated with the Main Generator are listed below:
o Generator Inlet Flow................... 500 gpm o Generator Inlet Pressure.............. 34 psig o Generator Outlet Temperature.... 84°C These conditions will result in the ------------------
A. bypass valves remaining closed with the generator remaining on line.
B,.. bypass valves being throttled partially open with the generator remaining on line.
C. turbine control valves throttling closed, followed by a high neutron flux Reactor Scram.
D. turbine control valves throttling closed, followed by a high reactor pressure Reactor Scram.
Description; The runback will cause the Main Turbine control valves to close to 24% main generator load. The impact on the plant is determined by whether the sufficient pressure control margin exists to maintain pressure, rather than resulting in a significant pressure increase. If power is above the capacity of the system to maintain pressure (approx 50%), pressure will increase and the plant will either scram due to high pressure psig or high neutron flux.
The limits for this question include the runback setpoint of24% load, 21 % bypass valve capacity(BPV), and 5%steam loads for RFPTs, SJAE, Gland seal condenser realizing this value will be between 2-5% at less than 100% power. Therefore; 35% power is within the capacity of the pressure control system following the runback.
A. Incorrect, see description above.
Plausible if the candidate assumes the turbine control valves and existing steam loads can maintain system pressure without the assistance of the Bypass Valves.
B. Correct, see description above.
C. incorrect, see description above.
Phlusible since pressure does increase as a result of the runback.
D. Incorrect, see description above.
Plausible since neutron flux does increase as pressure increases.
Thursday, January 15,2009 10:27:33 AM 54
HL T 4 NRC Exam SYSTEM: 241000 Reactor/Turbine Pressure Regulating System Kl. Knowledge of the physical connections and/or cause/effect relationships between REACTOR/TURBINE PRESSURE REGULATING SYSTEM and the following:
(CFR: 41.2 to 41.9/45.7 to 45.8)
Kl.25 Stator water cooling: Plant-Specific.................... RO 2.8 1 SRO 2.8 Reference(s) used to develop this question:
B21-SLLS-LP-01401, N43-SWC-LP-02301, "Stator Water Cooling" EO 200.089.a.Ol N32 DEHC-LP-01902 34S0-N43-003-1, "Stator Water Cooling System" Reference(s) provided to the student:
None Thursday, January 15,2009 10:27:33 AM 55
HL T 4 NRC Exam
- 27. 259002A4.01 001l211/RFPTINEWIHIGHERlHT2009-301IROIEUICMEI Unit 2 is in Mode 2, performing a Startup.
o Reactor pressure.......................... 900psig o Reactor power............................. 3 %
o Reactor Water Level (RWL)....... 37 inches o RWL control................................ Reactor Feed Pump Turbine (RFPT) "2A", in automatic at 2800 rpm.
o RFPT "2B"................................... Tagged out of service Due to intermittent problems with the "2A" RFPT MIA station, 2C32-R601A, Maintenance has requested permission to take 2C32-R601A to the maintenance shop for repair.
lAW 34S0-N21-007-2, "Condensate and Feedwater System", which ONE of the following describes the affect on plant startup if the MIA Station is removed for repair?
A. Plant startup CANNOT continue. Reactor pressure must be reduced to 400 psig.
B. Plant startup CANNOT continue. RFPT "2A" speed must be reduced below 2100 rpm.
C'f Plant startup CAN continue up to the point where RFPT "2A" flow reaches 7.0 Mlbmlhr.
D. Plant startup CAN continue, up to the point where RFPT "2A" reaches 6300 rpm
==
Description:==
Operation of the RFP on the Speed Setter will allow RFP operation up to the high speed stop of 5800 RPM. Overspeed trip occurs at a point between 5900 - 6300 RPM.
34S0-N21-007 contains a section, "RFPT Alternate Startup" which allows the startup to continue with RFP operation controlled by the speed setter.
A. Incorrect, see description above.
Plausible since 400 psig will restore RPV pressure to below the discharge head of a Condensate Booster pump.
B. Incorrect, see description above.
Plausible since 2100 rpm is the speed referenced within the Condensate and Feedwater procedure for transferring control of the RFP from the speed setter to the MIA station.
C. Correct, see description above.
D. Incorrect, see description above.
Plausible if the candidate assumes the speed may be increased to the overspeed trip setpoint using the speed setter controls. This is the switch used to set the overspeed trip setpoint.
Thursday, January 15,200910:27:34 AM 56
HL T 4 NRC Exam SYSTEM: 259002 Reactor Water Level Control System A4. Ability to manually operate and/or monitor in the control room:
(CFR: 41.7/45.5 to 45.8)
A4.01 All individual component controllers in the manual mode....... RO 3.8 / SRO 3.6 Reference(s) used to develop this question:
34S0-N21-007-2, Attachment 8 refers to the high speed stop of5800 rpm.
N21-CNDFW-LP-00201, Condensate and Feedwater System" EO 002.027.a.Ol Reference(s) provided to the student:
None Thursday, January 15,2009 10:27:35 AM 57
HL T 4 NRC Exam
- 28. 2610{)0A2.12 OO1l2111SBGT RADIBANKMODIHIGHERJHT2009-301IRO/EU/CMEI Unit 2 is in a Refueling Outage with fuel movement in progress.
Minor damage to a fuel bundle results in an increase in Refueling Floor radiation levels.
The "Refueling Floor Vent Exhaust" monitors are reading:
o 2Dll-K634A - 19 mRem/hr o 2Dll-K634B - 18 mRem/hr o 2D11-K634C - 15 mRemlhr o 2D11-K634D - 14 mRemlhr Which ONE of the following predicts BOTH the impact of these conditions on the Unit 1 and Unit 2 Standby Gas Treatment Systems (SBGT) and any procedural actions required for mitigation of the event.
The total number (Unit 1 and Unit 2 combined) of SBGTs that will start is (1)
(2) ventilation systems will be verified tripped and isolated.
A. (1) two (ONLY)
(2) ONLY the Unit 1 and Unit 2 Refueling Floor B. (1) four (2) ONLY the Unit 1 and Unit 2 Refueling Floor C. (1) two (ONLY)
(2) BOTH the Unit 1 and Unit 2 Refueling Floor and Reactor Building D" (1) four (2) BOTH the Unit 1 and Unit 2 Refueling Floor and Reactor Building Thursday, January 15, 2009 10:27:36 AM 58
HL T 4 NRC Exam
==
Description:==
There are 12 Refuel Floor Process Radiation monitors.
Setpoints:
o 2D11-K634A-D rad monitors for V2.... 5.7 mRlhr o 2D11-K611A-D rad monitors for V2.... 18 mRIhr o 2D11-K635A-D rad monitors for V2.... 6.9 mRlhr With all 4 of the 2D11-K634 rad monitors above the setpoint, all 4 SBGT systems will auto start.
For plausibility, the following system operation was considered. Exceeding the setpoint on the "A" and "B" monitors auto start only SBGT "A." The system requires monitors "C" and "D" to reach their setpoints to auto start the "B" fan of SBGT. VI only has 4 process rad monitors for the Refuel Floor with a setpoint of 18 mRlhr A. Incorrect, see description above.
Plausible since the candidate may confuse the rad monitors and their setpoints. If the candidate remembers the 18mR/hr setpoint (or if the candidate confuses VI and V2 setpoints) and then only 2 SBGT systems are expected to start since 2 monitors are above 18 mR/hr and 2 monitors are below 18 mRlhr.
B. Incorrect, see description above.
Plausible if the candidate assumes that Refuel Floor Ventilation radiation only aligns SBGT to the Refuel Floor.
C. Incorrect, see description above.
Plausible since the candidate may confuse the rad monitors and their setpoints., If the candidate remembers the 18mR/hr setpoint (or if the candidate confuses VI and V2 setpoints) and then only 2 SBGT systems are expected to start since 2 monitors are above 18 mRlhr and 2 monitors are below 18 mRlhr.
D. Correct, see description above.
Thursday, January 15,2009 10:27:37 AM 59
HL T 4 NRC Exam SYSTEM: 261000 Standby Gas Treatment System A2. Ability to (a) predict the impacts of the following on the STANDBY GAS TREATMENT SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
(CFR: 41.5/45.6)
A2.12 High fuel pool ventilation radiation: Plant-Specific........... RO 3.2 / SRO 3.4 Reference(s) used to develop this question:
34AB-T22-003-2, "Secondary Containment Control" T46-SBGT-LP-03001, "Standby Gas Treatment System" 030.006.a.Ol Reference(s) provided to the student:
None Thursday, January 15, 200910:27:38 AM 60
HL T 4 NRC Exam
- 29. 262001K4.02 001/2111R22 BREAKER TRlPINEWIHIGHERlHT2009-301!RO/EUICMEI Unit 2 is at 50% power.
o 4160 VAC buses "2A" and "2B" are all being powered from their alternate source due to maintenance activities.
- The "20" 4160V AC normal supply breaker trips due to an internal fault.
Which ONE of the following predicts the status of the plant TWO MINUTES after the supply breaker trIps?
The Reactor will have a ------------------
A"! full scram due to loss of Main Condenser Vacuum B. full scram due to loss of Condensate and Condensate Booster Pumps C. half scram (only); however, the plant will continue to operate at 50% power with no significant threats to plant operation D. half scram (only); however, manual action will be required to prevent Turbine Building loads, cooled by PSW, from over-heating Thursday, January 15,2009 10:27:38 AM 61
HL T 4 NRC Exam A. Correct When the alternate breaker closes in for the Emergency Bus, 4160 VAC bus 2A and 2B de-energize. This results in a loss of Main Circ Water pumps, which results in very rapid loss of main condenser vacuum. The loss of vacuum results in a main turbine trip which causes a full reactor scram since power is> 27.6%.
B. Incorrect 4160 C&D will remain energized following the turbine trip since they are already on alternate.
Plausible since the student must know the correct power supplies to the Condensate and Condensate Booster Pumps in order to eliminate this distract or. The candidate(s) will choose this distractor if they think 4160V AC "A" and "B" supply power to these pumps, C. Incorrect A full reactor scram will occur due to loss of condenser vacuum. Plausible; if the candidate does not know the interlock which de-energizes 4160V AC bus A&B if an emergency bus is on alternate, this will look like the correct answer.
D. Incorrect A full reactor scram will occur due to loss of condenser vacuum. Plausible; if the candidate does not know the interlock which de-energizes 4160V AC bus A&B if an emergency is on alternate, this wi11look like a possible correct answer. The PSW loads overheating is a distraction implying the Turbine Building isolation valves close due to a temporary loss of power to the 4160 VAC bus (if the bus swaps to alternate, the valves do not actually go closed. If the bus de-energizes and remains de-energized, the valves actually remain open since they are MOVs)
SYSTEM: 262001 A.C. Electrical Distribution K4. Knowledge of A.C. ELECTRICAL DISTRIBUTION design feature(s) and/or interlocks which provide for the following: (CFR: 41.7)
K4.02 Circuit breaker automatic trips....................... RO 2.9 / SRO 3.3 Reference(s) used to develop this question:
34AB-R22-002-2, "Loss of 4160V Bus Interlocks" R22-4160 VAC-LP-02702, "4160 VAC" EO 027.009.a.03 34S0-R22-001-2, "4160 VAC System" Reference(s) provided to the student:
None Thursday, January 15, 2009 10:27:40 AM 62
HL T 4 NRC Exam
. 30. 262001K4.05 001l2/1/EDG, SYNCHROSCOPEIBANKMOD/FUNDIHT2009-301IRO/EUICMEI An operator is preparing to close the "IA" Emergency Diesel Generator (EDG) output breaker during the performance of34SV-R43-00I-I, "Diesel Generator IA Monthly Te:st",
lAW 34SV-R43-00I-I, which synchroscope (1, 2, 3 or 4) indicates that the required conditions are met to manually close the breaker based on both of the following?
o Direction of rotation o Pointer placement At 1 B.
2 C.
3 D.
4 Thursday, January 15,200910:27:40 AM 63
HL T 4 NRC Exam A. Correct The procedure requirements are: rotation in the clockwise (fast) direction and the synchroscope indicates 2 minutes to 12.
B. Incorrect Direction of rotation is not correct. The pointer is on the wrong side of the 12:00 position.
Plausible because these are the correct conditions for closing the normal or alternate supply breaker when the EDG is supplying power to the emergency bus.
C. Incorrect because the direction of rotation is incorrect. Direction of rotation is plausible because it is the required direction for closing the normal or alternate supply breaker when the EDG is supplying power to the emergency bus. The pointer placement is in the correct position.
D. Incorrect because the pointer is in the wrong position (past the 12:00 position).. Direction of rotation is correct. Pointer indication is plausible because it is the correct position for closing the normal or alternate supply breaker when the EDG is supplying power to the emergency bus.
SYSTEM: 262001 A.C. Electrical Distribution K4. Knowledge of A.C. ELECTRICAL DISTRIBUTION design feature(s) and/or interlocks which provide for the following: (CFR: 41.7)
K4.05 Paralleling of A.c. sources (synchroscope)........... RO 3.4 / SRO 3.6 Referencefs) used to develop this question:
34SV-R43-001-1, Diesel Generator lA Monthly Test R43-EDG-LP-02801, "Emergency Diesel Generators" 028.023.a.11 Referencefs) provided to the student:
None Thursday, January 15,200910:27:42 AM 64
HL T 4 NRC Exam
- 31. 262002A3.01 00 112/1 NITAL ACINEWIHIGHERJHT2009-301IROIEUICMEI Unit 2 is operating at 86%ofrated power.
An alarm is received in the main Control Room. As you scan the control panels you notice the following alarm and indications:
(Alarming)
(Amber Light Illuminated)
Using these indications, determine the power source which is supplying the Vital AC Bus:
A. 600 V AC Bus 2C through the Vital AC Static Inverter.
B. 600 V AC Bus 2D through the Vital AC Static Inverter..
Ct 600 V AC Bus 2C, through the Vital AC Essential Transformer.
D. 600 VAC Bus 2D, through the Vital AC Essential Transformer Thursday, January 15, 200910:27:42 AM 65
HL T 4 NRC Exam A. Incorrect 600V AC bus does not supply power to the Vital AC Battery Charger that provides power to the inverter.
Plausible; candidate must know which bus supplies the battery charger to eliminate this distracter. 600V AC bus 2C does supply power to Vital AC in this instance, but not through a battery charger. It supplies power via an essential transformer and static transfer switch.
B. Incorrect Normal supply is 600 VAC "2D", which is not available, per the amber and white light status.
Plausible; the candidate must know which bus is the normal supply and which is alternate. The fact that 600V AC bus "2D" is actually the normal supply through the battery chargers, static inverter, etc.
C. Correct D. Incorrect The 600V AC bus "2D" does not supply the Essential transformer.
Plausible the candidate must know which bus is the normal supply and which is alternate.
600V AC bus "2D" is actually the normal supply which does not go through the ess. xfrmr.
SYSTEM: 262002 Uninterruptable Power Supply (A.C.ID.C.)
A3. Ability to monitor automatic operations of the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) including: (CFR: 41.7/45.7)
A3.01 Transfer from preferred to alternate source.......... RO 2.8 / SRO 3.1 Reference(s) used to develop this question:
34AR-651-150-2, Vital AC Sys on Alt Supply ARP R25-ELECT -LP-02705 Vital AC lesson plan EO 200.020.A.05 Given a set of conditions, select the correct power supply (if any) to the Vital AC bus.
Reference(s) provided to the student:
None Thursday, January 15, 200910:27:43 AM 66
HLT 4 NRC Exam
- 32. 263000A2.02 001l211IDC VENTILA TIONIBANKiFUNDIHT2009-30 llRO/EUICMEI Unit 2 is operating at 100% power.
o The Control Building Ventilation System normal ventilation is lost.
IAW 34AB-T41-001-2, "Loss OfECCS, MCREC Or Area Ventilation System(s)", which ONE of the following actions is required to protect the Station Service DC Electrical Distribution System?
A't Start the emergency battery room exhaust fans to reduce the potential for buildup of combustible gases.
B. Add electrolyte to all 125/250 VDC battery cells due to the expected increase in the rate of electrolyte evaporation.
C. Rotate the battery chargers so that in-service time for each battery charger is minimized, thereby reducing battery charger heatup.
D. Reduce DC loads in the plant to minimize the current being drawn from the 125/250 VDC batteries, thereby reducing battery heatup.
A. Correct B. Incorrect No guidance exists in the procedure to do add electrolyte; however, the concept that higher temperatures will result in accelerated evaporation of water in the battery is plausible/true, so the requirement to add electrolyte is plausible.
C. Incorrect No guidance exists in the procedure to rotate battery chargers; however, it is plausible to think:
that electrical equipment will heat up over time and require down time to cool off. We do rotate some equipment when cooling is lost (loss ofPSW procedure we rotate CondiCond Booster Pumps based on high temperature alarms)
D. Incorrect Normally, DC is supplied by the battery chargers rather than the batteries. Nothing in conditions given suggest a change from this situation. Plausible if the candidate assumes the batteries, rather than the battery chargers, normally supply DC loads.
Thursday, January 15,2009 10:27:45 AM 67
HL T 4 NRC Exam SYSTEM: 263000 D.C. Electrical Distribution A2. Ability to (a) predict the impacts of the following on the D.C. ELECTRICAL DISTRIBUTION; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 /45.6)
A2.02 Loss of ventilation during charging........ RO 2.6 / SRO 2.9 Reference(sl used to develop this question:
34AB-T41-001-2, "Loss OfECCS, MCREC Or Area Ventilation System(s)"
Z41-CBHVAC-LP-03703, "Control Building HVAC" 34AB-P41-001-2, "Loss of Plant Service Water" Reference(sl provided to the student:
None Thursday, January 15, 2009 10:27:45 AM 68
HL T 4 NRC Exam
- 33. 264000K4.06 OO1l2/11EDG WAD LIMITIBANKMODIFUND/HT2009-301IRO/EUICMEI Which ONE of the following governor control panel settings will shut off the fuel supply to a running Plant Hatch Emergency Diesel Generator?
A't Load Limit set at "0" B. Speed Droop set at "0" C. Load Limit set at "10" D. Speed Droop set at "100" A. Correct B. Incorrect, adjusting the speed droop to 0 results in the diesel increasing the position of the fuel racks when additional load has been placed on the EDG, thereby not affecting the shutdown of the diesel.
Plausible as this is one of the actual controls on the EDG ands is adjusted during surveillances from 0 to 50 but an infrequent manipulation.
C. Incorrect, adjusting this knob to 10 allows the diesel to operate at 100% fuel rack position.
Plausible since it is the correct knob for shutting down the EDG, but not using terms like fully upscale. The candidate must remember that 10 equals full scale and 100% speed. Also this knob is seldom operated so familiarity with how it works requires significant knowledge of the EDG controls. Additionally, max load limit may be interpreted as the most limiting or zero fuel.
D. Incorrect, adjusting this knob to 100 allows the diesel to slow by 7% speed before changing fuel supply for added electrical loading.
Plausible since it is a controller on the governor and is periodically adjusted during EDG runs but not to this value. Also this knob is seldom operated so familiarity with how it works requires significant knowledge of the EDG controls.
Thursday, January 15, 200910:27:46 AM 69
HL T 4 NRC Exam SYSTEM: 264000 Emergency Generators (Diesel/Jet)
K4. Knowledge of EMERGENCY GENERATORS (DIESEL/JET) design feature(s) and/or interlocks which provide for the following: (CFR: 41.7)
K4.06 Governor control.............. RO 2.6 / SRO 2.7 Reference(sl used to develop this question:
R43-EDG-LP-02801, "Emergency Diesel Generators" 028.022.a.03,Section V, Control features and interlocks Reference(sl provided to the student:
None Thursday, January 15, 2009 10:27:47 AM 70
HLT 4 NRC Exam
- 34. 264000K5.06 001l2/1lEDG LOAD SEQUENCEINEWIHIGHERlHT2009-301IRO/EUICMEI Following a LOCAILOSP on Unit 1, the following sequence occurs:
o The "IA" and "IB" EDGs start o The" 1 C" EDG fails to start When the EDG output breakers close, pumps start in this sequence:
o Time = 0 sec o Time = 0.5 sec o Time = 10 sec o Time = 12 sec o Time = 22 sec "IA" Core Spray pump starts "IC" RHR pump starts "ID" PSW pump starts "IA" and "ID" RHRpumps start "IA" PSW pump starts Which ONE of the following states the status of the load sequence and operational implication for both of the statements below?
o The "IA" EDG load sequence was (1) o The "IB" EDG load sequence was (2)
A. (1) correct, operation is as designed.
(2) correct, operation is as designed.
Bi (1) correct, operation is as designed.
(2) NOT correct and there is a high probability of experiencing an overload condition.
C. (1) NOT correct and there is a high probability of experiencing an overload condition.
(2) correct, operation is as designed.
D. (1) NOT correct; however, there should be no adverse consequences to EDG operation.
(2) NOT correct; however, there should be no adverse consequences to EDG operation.
Thursday, January 15, 200910:27:48 AM 71
Actual correct loading sequence:
Time = 0 sec Both CS pumps Time = 0.5 sec RHRpump "IC" HL T 4 NRC Exam Time = 12 sec RHR pumps "IA", "IB" and "ID" RHR start Time = 22 sec PSW Pumps "IA"and "IB" Time = 32 sec PSW pump "IC" starts if"IA" did not Time = 37 sec PSW pump "ID" starts if "IB" did not The above sequence assumes all 3 EDGs start. In this question the "IC" EDG did not start, so the pumps on "IG" 4160VAC bus will not run ("IB" RHR, "IB" CS & "IB" PSW)
A. Incorrect; See correct loading sequence above.
1 st part is correct.
2nd part is incorrect (loading sequence is not correct)
Plausible if the candidate does not know the EDG loading sequence timers. Changing times associated with the pump starts will make this a correct answer B. Correct; See correct loading sequence above.
C. Incorrect; See correct loading sequence above.
1 st part is incorrect (loading sequence is correct) 2nd part is incorrect (loading sequence is not correct)
Plausible if the candidate does not know the EDG loading sequence timers. Changing times associated with the pump starts will make this a correct answer.
D. Incorrect; See correct loading sequence above.
1 st part is incorrect (loading sequence is correct) 2nd part is incorrect (loading sequence is not correct)
Plausible if the candidate does not know the EDG loading sequence timers. Changing times associated with the pump start will make this a correct answer, i.e. making the times longer than designed would make the sequence still incorrect, yet allow additional time for motor starting currents to subside.
Thursday, January 15,200910:27:49 AM 72
HL T 4 NRC Exam SYSTEM: 264000 Emergency Generators (Diesel/Jet)
KS. Knowledge of the operational implications of the following concepts as they apply to EMERGENCY GENERATORS (DIESEL/JET) : (CFR: 41.5/45.3)
K5.06 Load sequencing...... RO 3.4 / SRO 3.5 Reference(sl used to develop this question:
34AB-R22-002-1, Loss of 4160V Emergency Bus, Attachment 3 R43-EDG-LP-02801, Emergency Diesel Generators lesson plan, Table 3 EO 028.025.A.02 Given a list, select the proper loading sequence for any 4160 VAC bus after the D/G starts and the following conditions exist; a. LOSP b. LOCAILOSP Reference(sl provided to the student:
None Thursday, January 15, 200910:27:50 AM 73
HLT 4 NRC Exam
- 35. 268000A2.01 001l2/2IDW LEAK TSINEWIHIGHERJHT2009-301IROJEUICMEI Unit 2 experiences a rupture on the "2A" Core Spray (CS) suction piping in the Torus Area of the Reactor Building.
The following annunciators are alarming (not a complete list of all alarming annunciators) o TORUS S-E AREA INSTR SUMP LVL HIGH (657-089-2) o RB S-E DIAGONAL FLOOR DRN SUMP LEVEL HIGH-HIGH (657-034-2)
The following valve alignment exists:
o 2T45-F002 "Torus N-E and S-E Outboard Sump Iso1 Valve" is open o 2T45-F003 "Torus N-E and S-E Inboard Sump Isol Valve" is open Which ONE of the following completes BOTH of these statements?
A.
B.
C.
If NO operator action is taken, the flow of water to the Radwaste facility (1) automatically isolate as Torus area water levels continue to increase.
IA W 34AR-657-034-2, "RB S-E DIAGONAL FLOOR DRN SUMP LEVEL HIGH-HIGH", 2T45-F002, "Torus N-E and S-E Outboard Sump Isol Valve" is required to be (2)
(1) will (2) open (1). will (2) closed (1) will NOT (2) open D!' (1) will NOT (2) closed Thursday, January 15, 2009 10:27:51 AM 74
HL T 4 NRC Exam
==
Description:==
As water leaks into the Torus room area, it enters instrument sumps and then flows into the diagonal floor drain sump. There are isolation valves located between the diagonal floor drain sumps and the torus room sumps. These valves automatically close if level in the floor drain sumps reaches the High-High level setpoint, the isolation valves automatically close to prevent flooding the diagonaL When floor drain sump level decreases, the valves automatically open to process more water. If the diagonal floor drain sump pumps can keep up with the drainage coming from the Torus room area, the isolation valves between the diagonal and the torus room area remain open. If the Torus area instrument sump High level alarm is received, its ARP requires that the isolation valves be verified to be open.
In the case of this question, the isolation valves fail to auto close. The Diagonal Floor Drain sump High-High level ARP requires the valves be confirmed closed.
A. Incorrect, see description above 1 st part is plausible since the candidate is required to know which sump level corresponds to the demand to close the isolation valve. If the candidate assumes the isolation occurs at high-high-high, then this will appear to be correct..
2nd part is plausible since the torus high level instrument sumps ARP requires the valves to be open to allow drainage of water from the torus area to the diagonal floor drain sumps.
B. Incorrect; see description above 1 st part is plausible since the candidate is required to know which sump level corresponds to the demand to close the isolation valve. If the candidate assumes the isolation occurs at high-high-high, then this will appear to be correct.
2nd part is correct C. Incorrect; see description above 1 st part is correct 2nd part is plausible since the torus high level instrument sump ARP requires the valves to be open to allow drainage of water from the torus area to the diagonal floor drain sumps.
D. Correct; see description above Thursday, January 15, 200910:27:52 AM 75
HL T 4 NRC Exam SYSTEM: 268000 Radwaste A2. Ability to (a) predict the impacts of the following on the RADWASTE; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5/45.6)
A2.01 System rupture................... RO 2.9 / SRO 3.5 Referencefsl used to develop this question:
T22-SC-LP-01301 "Secondary Containment" Simplified diagrams (Fig 2 & 4) and Terminal Objective 201.078.b TORUS S-E AREA INSTR SUMP LVL HIGH (657-089-2)
RB S-E DIAGONAL FLOOR DRN SUMP LEVEL HIGH-HIGH (657-034-2)
Referencefsl provided to the student:
None Thursday, January 15,2009 10:27:52 AM 76
HL T 4 NRC Exam
- 36. 286000Al.0 1 001 1212IFIRE PROTECTIONIBANKIFUND/HT2009-30 llRO/EUICMEI A plant fire protection sprinkler system actuation results in a fire main header pressure of98 psig.
Which ONE of the following predicts tll,e fire pump response?
A. Only the electric fire pump starts.
B. Only the "A" and "B" diesel fire pumps start.
C!' Only the electric fire pump and the "A" diesel fire pump starts.
D. "A" and "B" diesel fire pumps and the electric fire pump starts.
Actual Pump start pressures:
o Electrical fire pump IX43-C001 starts at 110 psig o Diesel fire pump IX43-C002A starts at 100 psig o Diesel fire pump 1X43-C002B starts at 90 psig A. Incorrect; see actual pump start pressures listed above.
Plausibility, changing the system pressure to> 100 psig and <110 psig would make this a correct answer.
B. Incorrect; see actual Pump start pressures listed above Plausibility, if the candidate assumes the diesel fire pumps start before the electric fire pump.
C. Correct; see Actual Pump start pressures listed above D. Incorrect; see actual pump start pressures listed above.
Plausibility, changing the system pressure to less than 90 psig would make this a correct answer.
Thursday, January 15,200910:27:53 AM 77
HL T 4 NRC Exam SYSTEM: 286000 Fire Protection System AI. Ability to predict and/or monitor changes in parameters associated with operating the FlREPROTECTION SYSTEM controls including: (CFR: 41.5/45.5)
A1.01 System pressure......... RO 2.9 / SRO 2.9 References used to develop this question:
34S0-X43-001-1, Fire Pumps Operating Procedure X43-FPS-LP-03601, Fire Protection lesson plan EO 036.020.B.13 Given a fire has started, describe the starting sequence for the three fire pumps and the pressure when they start Reference(s) provided to the student:
None Thursday, January 15,200910:27:54 AM 78
HLT 4 NRC Exam
- 37. 295001 AK2.03 00 1I111lRECIRCINEW IHIGHERlHT2009-30 llROIEUICMEI Unit 1 is operating at 63% power with both Reactor Feedwater pumps in service.
o Recirculation pump speed is at 63%
o Steam flow is 63%
o The "lA" Condensate Booster pump (CBP) trips o The" 1 C" CBP automatically starts.
o The "lA" Reactor Feedwater pump trips due to low suction pressure.
o The" 1B" Reactor Feedwater pump continues to operate.
o +23 inches is the lowest Reactor Water Level during the transient With no operator action, the final speed of the Reactor Recirculation pumps will be __.
At 33%
B.
53%
C.
61%
D.
63%
==
Description:==
The Reactor Recirc pumps run back to 33% if the following conditions are met, (steam flow is above 65% or RWL < 32") AND the RFPs receive a TMR trip AND at least one RFP has <20% flow. In this case, steam flow is initially below 65%; therefore the 33% runback will not occur until the RFP trips and RWL goes below 32 inches.
The Recirc Pumps have a variable runback feature, when RWL goes below 30", Recirc Pump speed is reduced by 6.7%linch, down to a low of33%.
A. Correct, see description above B. Incorrect, see description above Plausible if the candidate calculates the runback based on the variable Recirc Runback 23" RWL is a 7 inch drop from +30" therefore a runback signal of6.7%/inch*7 inches=47%
(100%-47%=53% speed.) This runback did actually occur, but is not the most limiting.
C. Incorrect, see description above Plausible since the Recirc Pumps do run back to 61 % following a CBP or RFP low suction pressure conditions (following a 5 or 10 sec time delay, respectively). This runback may occur, however it will not be the most limiting runback.
D. Incorrect, see description above Plausible if the candidate considers that steam flow below 65% is required for this runback to occur, without taking low RWL (below 32") into account. This is a popular misconception.
Thursday, January 15, 2009 10:27:55 AM 79
HL T 4 NRC Exam APE: 295001 Partial or Complete Loss of Forced Core Flow Circulation AK2. Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION and the following: (CFR: 41.7/45.8)
AK2.03 Reactor water leveL................................ RO 3.6/ SRO 3.7 Reference(s) used to develop this question:
B31-RRS-LP-00401, "N2l-CNDFW-LP-0020I "EO 004.001.a.07 34S0-B31-001-1, "Reactor Recirculation System" step 5.2.17 "Recirc Speed Limiter Summary" Reference(s) provided to the student:
None Thursday, January 15, 200910:27:56 AM 80
HL T 4 NRC Exam
- 38. 295003G2.4.34 00 l/1/1/EDGINEWIFUNDIHT2009-30 lIRO/EUICMEI Unit 1 is experiencing a Loss of Offsite Power (LOSP).
The following conditions exist:
o The" IA" Emergency Diesel Generator (ED G) failed to start o An operator is dispatched to locally start the "IA" EDG o All "Control Power On" lights on IR43-POOIA are illuminated IA W with 34AB-R43-001-1, Diesel Generator Recovery, which ONE of the following describes the requirements for starting the "IA" EDG locally AND flashing the generator field?
The "IA" EDG Control Switch must be placed in the (1) position prior to depressing the "START" push-button.
The Control Switch must be in the circuit.
A. (1) "REMOTE" (2) "AT ENG" B. (1) "REMOTE" (2) "REMOTE" C. (1) "AT ENG" (2) "AT ENG" D,- (1) "AT ENG" (2) "REMOTE" Thursday, January 15,2009 10:27:57 AM (2) position to enable the Generator Field Flash 81
HL T 4 NRC Exam
==
Description:==
The EDG start logic requires the EDG Control Switch to be placed in the "At Eng" position for the local start switch to be used to start the EDG. In order to "flash" the generator field, the switch must be placed in the "Remote" position.
Plausibility is based on the candidate recalling the design features associated with the EDG start logic.
A. Incorrect; see description above Both parts are incorrect 1 st part is plausible if the candidate assumes the correct position for this task in "Remote" 2nd part is plausible if the candidate assumes the correct position for this task in "At Eng" B. Incorrect; see description above 1st part is plausible if the candidate assumes the correct position for this task in "Remote" 2nd part is correct C. Incorrect; see description above 1 st part is correct 2nd part is plausible if the candidate assumes the correct position for this task in "At Eng".
The switch is placed back in the At-Eng position after the field has been flashed D. Correct; see description above APE: 295003 Partial or Complete Loss of A.C. Power 2.4 Emergency Procedures / Plan 2.4.34 Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects. (CFR: 41.10 / 43.5/45.13)
IM:PORTANCE RO 4.2 / SRO 4.1 Reference(s) used to develop this question:
34AB-R43-00 1-1 R43-EDG-LP-02801 EDG lesson plan EO 027.051.a.Ol EO 028.006.a.02, 028.024.a.Ol Reference(s) provided to the student:
None Thursday. January 15. 2009 10:27:58 AM 82
HL T 4 NRC Exam
- 39. 295004AAl.03 OOll1l1/DC ELECTRICAUNEWIHIGHERJHT2009-301IRO/EUICMEI Unit 2 was operating at 100% power.
o A loss of2R22-S016 (125/250VDC "A") occurs.
o Subsequently a Main Turbine trip occurs Which ONE of the following completes both of these statements?
- The Main Generator PCBs
- 0)
- When the PCBs open/are opened all station service 4160 VAC buses (2) automatically transfer to their alternate source.
A. (1) will automatically open (2) will B. (1) must be manually opened lAW 34AB-R22-001-2, "Loss of DC Buses" (2) will C. (1) will automatically open (2) will NOT D'!" (1) must be manually opened lAW 34AB-R22-001-2, "Loss of DC Buses" (2) will NOT Thursday, January 15, 2009 10:27:59 AM 83
HL T 4 NRC Exam
==
Description:==
If a Main Turbine trip occurs, the Main Generator output breakers will NOT auto open. 4160v buses will NOT transfer to startup supply.
Plausibility is based on whether the candidate remembers the impact of a loss of DC on both the main generator PCBs for a turbine trip and whether the station service buses will auto transfer in this condition.
A. Incorrect; see description above.
Plausible since, normally, this is the correct operation of these breakers. This would also be the correct answer if the SO 17 had been lost rather than SO 16 B. Incorrect; see description above Plausible since the 1 st part is correct for this condition and the 2nd part is the normal operation of the station service buses.
C. Incorrect; see description above Plausible since the 1 st part is the normal operation of these breakers, and it would be correct if SO 17 had been lost rather than SO 16. The 2nd part is correct for this condition.
D. Correct; see description above APE: 295004 Partial or Complete Loss of D.C. Power AAI. Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER: (CFR: 41.7 / 45.6)
AA1.03 A.C. electrical distribution...... RO 3.4 / SRO 3.6 References used to develop this question:
34AB-R22-001-2, "Loss of DC Buses" EO 200.018.a.01 R42-ELECT-LP-02704 DC Electrical Lesson Plan Reference(sl provided to the student:
None Thursday, January 15, 2009 10:28:00 AM 84
HL T 4 NRC Exam
- 40. 295005AK3.04 00ll1l1IMT PCBINEWIFUNDIHT2009-301IRO/ELJ/CMEI Unit 1 is at 40% power.
o The Main Turbine automatically trips due to a problem with the lubricating oil system.
When the Main Turbine trips, the Main Generator will trip (1)
This trip arrangement protects the Turbine from (2)
A. (1) immediately (2) overspeeding B. (1) immediately (2) overheating C~ (1) seconds later (2) overspeeding D. (1) seconds later (2) overheating
==
Description:==
The generator output breakers have two types of trips, sequential and simultaneous. The sequential are associated with turbine trips and occur 3.75 seconds after the turbine trip in order to use all the steam left in the turbine after the stop valves are shut to avoid turbine damage from over-speeding. The simultaneous trips are for problems associated with the generator itself, Overcurrent, etc.
A. Incorrect, this does NOT result in a simultaneous trip, the concern is to prevent overspeed.
Plausible because the concern is overspeed and since the candidate must remember the type trips that cause simultaneous and sequential trip (this is confusing to most candidates).
B. Incorrect, this does NOT result in a simultaneous trip, it is a sequential trip to ensure an overspeed condition does not occur.
Plausible since reverse power can damage the turbine through overheating; however, overspeed protection is the over-riding concern in this condition and since the candidate must remember the type trips that cause simultaneous and sequential trip (this is confusing to most candidates).
C. Correct, see description above D. Incorrect, because the concern is to prevent overspeed Plausible because it is a sequential trip and since motoring of the main generator will occur for a short period of time.
Thursday, January 15, 200910:28:01 AM 85
HL T 4 NRC Exam APE: 295005 Main Turbine Generator Trip AK3. Knowledge of the reasons for the following responses as they apply to MAIN TURBINE GENERATOR TRIP: (CFR: 41.5 /45.6)
AK3.04 Main generator trip........ RO 3.2 / SRO 3.2 Reference(s) used to develop this question:
N30-MTA-LP-01701 N40-MG-LP-lO002, "Main Generator" 017.006.A.09 Reference(s) provided to the student:
None Thursday, January 15,2009 10:28:02 AM 86
HL T 4 NRC Exam
- 41. 295006AKl.Ol OOllllllDECAYHEATIBANKNRC 07IHIGHERlHT2009-301IRO/EUICMEI Unit 2 was operating at 100% for one year when a spurious scram occurred due to surveillance testing.
The following conditions exist five minutes after the scram:
o All rods fully inserted o MSIVs open o Auxiliary steam loads still in service Which ONE of the following is the expected main turbine bypass valve position and the corresponding inventory makeup that is required to maintain level constant within the normal level band?
A I bypass valve will be fully open.
The required makeup is within the capacity of one CRD pump.
B. I bypass valve will be fully open.
The required makeup exceeds the capacity of one CRD pump.
C. I bypass valve will be controlling, varying between 0 - 50% open.
The required makeup is within the capacity of one CRD pump.
D'l" I bypass valve will be controlling, varying between 0 - 50% open.
The required makeup exceeds the capacity of one CRD pump.
Thursday, January 15,2009 10:28:03 AM 87
HL T 4 NRC Exam
==
Description:==
Simulator results are 0 - 50% bypass valve cycling. CRD pump capacity is ~ 120 gpm whereas actual demand is ~ 0.8 Mlbmlhr, A. Incorrect; see description above.
Plausible if applicant thinks that auxiliary steam loads are sufficient to keep the bypass valves closed and does not know the post-scram steaming rate.
B. Incorrect, see description above.
Plausible if applicant does not understand that stem states the auxiliary steam loads are still in service.
C. Incorrect, see description above.
Plausible if applicant does not know the post-scram steaming rate.
D. Correct see description above..
APE: 295006 SCRAM AKI. Knowledge of the operational implications of the following concepts as they apply to SCRAM: (CFR: 41.8 to 41.10)
AK1.01 Decay heat generation and removaL....... RO 3.7 / SRO 3.9 References used to develop this question:
Plant specific stand-alone simulator: Mode switch to shutdown scram from 100% power resulted in the following:
1.5 minutes after scram........ main turbine trips 1.75 minutes after scram....... 2% - 50% one bypass valve cycling 3 minutes after scram........ 9% - 50% one bypass valve cycling 5 minutes after scram........ 23 - 34% one bypass valve cycling Reference(s) provided to the student:
None Thursday, January 15,2009 10:28:04 AM 88
HL T 4 NRC Exam
- 42. 29501 OG2.2.22 001/1/2/TSINEWIFUNDIHT2009-30 lIRO/EUICMEI Unit 2 is operating at 90% power o Unit 2 Drywell (DW) pressure is 0.5 psig.
o At 10:00 DW pressure begins going up at 0.05 psig/minute IA W Tech Spec Limiting Condition for Operation (LCO) 3.6.1.4, Drywell Pressure, the EARLIEST time listed that a entry into a Required Action Statement (RAS) based on DW pressure is (1) and DW pressure is required to be restored to within limit no later than (2) from entering the RAS.
A. (1) 10:03 (2) 15 minutes B. (1) 10:03 (2) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> C. (1) 10:26 (2) 15 minutes D!' (1) 10:26 (2) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />
==
Description:==
The TS LCO is DW pressure shall be <= 1.75 psig, and if it is not, restore it to below 1.75 psig within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. (TS 3.6.1.4) 15 minutes is the RAS for Steam Dome Pressure (TS 3.4.10) 10:03 corresponds to 0.65 psig, 10:26 corresponds to 1.8 psig which exceeds the LCO of <=1.75 psig A. Incorrect; see description above.
Plausible since pressure at this time corresponds to the alarm setpoint for the Primary Containment Pressure High annunciator (34AR-603-115-2). 15 minutes is the TS RAS for steam dome pressure.
B. Incorrect; see description above. Second part is correct.
Plausible since pressure at this time corresponds to the alarm setpoint for the Primary Containment Pressure High annunciator (34AR-603-115-2).
C. Incorrect; see description above. (First part is correct).
Plausible since 15 minutes isthe TS RAS for steam dome pressure.
D. Correct, see description above.
Thursday, January 15,200910:28:05 AM 89
HL T 4 NRC Exam APE: 295010 High Drywell Pressure 2.2 Equipment Control 2.2.22 Knowledge of limiting conditions for operations and safety limits.
(CFR: 41.5/43.2 /45.2) IMPORTANCE RO 4.0/ SRO 4.7 Reference(s) used to develop this question:
Unit 2 Tech Specs (TS 3.6.1.4 and TS 3.4.10)
T23-PC-LP-01301, Primary Containment lesson plan 300.010.a.12 Reference(s) provided to the student:
None Thursday, January 15, 200910:28:06 AM 90
HL T 4 NRC Exam
- 43. 295013AK2.01 001l1l2IRHRlNEW/HIGHERlHT2009-301IRO/EUICMEI Unit 2 Bulk Torus water temperature is 103°F.
lAW the 34AB-T23-003-2, "Torus Temperature Above 95°F" and 34S0-EII-01O-2, "Residual Heat Removal (RHR) System", if ALL REQUIRED RHR systems are placed in Torus Cooling Mode, the TOTAL RHR cooling flow to the Torus is ___ _
A. 7,700 gpm.
B. 11,500 gpm.
C. 17,000 gpm.
D~ 23,000 gpm.
==
Description:==
IA W with 34AB-T23-001-2 requires All available torus cooling be started. The system operating procedure 34S0-EII-0I0-2 states that with 2 pumps operating in a loop flow should be 11, 500 gpm. With 2 loops ofRHR in this mode, the max total flow is 23,000 gpm.
A. Incorrect, see description above.
Plausible since 7700gpm is the max torus cooling flow for one RHR pump. This would be the correct answer if torus temperature were between 95F and 100F.
B. Incorrect, see description above.
Plausible since 11,500 gpm is the flow for torus cooling with only one loop ofRHR. The candidates will pick this answer if they only take into consideration only one loop of RHR.
C. Incorrect, see description above.
Plausible since 17,000 is the max RHR flow for containment spray mode per loop.
D. Correct, see description above.
Thursday, January 15,2009 10:28:07 AM 91
HL T 4 NRC Exam APE: 295013 High Suppression Pool Temperature AK2. Knowledge of the interrelations between HIGH SUPPRESSION POOL TEMPERATURE and the following: (CFR: 41.7 /45.8)
AK2.01 Suppression pool cooling.............................. RO 3.6/ SRO 3.7 Reference(sl used to develop this question:
34S0-E 11-0 1 0-2, "RHR System" 34AB-T23-003-2, "Torus Temperature Above 95F" T23-PC-LP-01301, "Primary Containment" lesson plan EO 200.040.A.Ol Reference(sl provided to the student:
None Thursday, January 15, 200910:28:07 AM 92
HL T 4 NRC Exam
- 44. 2950 15G2.2.42 001/1/2/SBLC SLCIBANK MODIFUND/HT2009-301/RO/CMEIEUI Unit 2 is in an Anticipated Transient Without a Scram (ATWS) that required a reactor water level (RWL) band of -155 inches to -185 inches.
While RWL is in this band, a Tech Spec Safety Limit Violation (1) occurred and adequate core cooling (2) exist.
A't (1) has (2) does B. (1) has (2) does NOT C. (1) has NOT (2) does D. (1) has NOT (2) does NOT Thursday, January 15, 2009 10:28:08 AM 93
HL T 4 NRC Exam
==
Description:==
IA W TS,' a safety limit is violated when RPV level goes below -155 inches regardless of plant conditions (ATWS/Non-ATWS). Adequate Core Cooling exists for both until RPV level goes below -185 inches (Minimum Steam Cooling RWL, i.e. steam cooling with injection) or -200 inches ifthere is NO injection (Minimum Zero-Injection RWL, steam cooling without injection).
A. Correct, see description above.
B. Incorrect, see description above.
1 st part is correct.
2nd part incorrect; however, it is plausible since RWL is below TAF it may be assumed that uncovered fuel is not adequately cooled. Adequate core cooling is assured when level is above T AF, by definition.
C. Incorrect, see description above.
1 st part is not correct; however it is plausible since this condition is directed by the EOPs it may be assumed the EOPs will not direct you to violate a safety limit.
2nd part is correct D. Incorrect, see description above.
1 st part is not correct; however it is plausible since this condition is directed by the EOPs it may be assumed the EOPs will not direct you to violate a safety limit.
2nd part incorrect; however it is plausible since RWL is below TAF it may be assumed that uncovered fuel is not adequately cooled.
APE: 295015 Incomplete SCRAM 2.2 Equipment Control 2.2.42 Ability to recognize system parameters that are entry-level conditions for Technical Specifications. (CFR: 41.7 /41.10 /43.2 / 43.3 /45.3)
IMPORTANCE RO 3.9 SRO 4.6 Reference(s) used to develop this question:
Unit 2 Tech Specs and Bases LP-30005, Tech Specs EO 300.003.a.Ol EOP-TERMS-LP-20304, EOP Terminology and Definitions EO 201.093.A.08 Reference(s) provided to the student:
None Thursday, January 15,2009 10:28:09 AM 94
HL T 4 NRC Exam
- 45. 295016AA2.01 00llll1l31RS EVACUATIONINEW/HIGHERlHT2009-301IROIEUICMEI An evacuation of the Main Control room is in progress.
o Prior to leaving the Main Control room, RC-l actions were performed on both units.
o All rods did NOT fully insert on Unit 1.
All necessary actions of31RS-OPS-OOl-l, "Shutdown From Outside Control Room" have been taken to establish control of Unit 1 at the Remote Shutdown Panel.
At this time, reactor power __________ _
A. can be determined by counting how many scram inlet and outlet valves are open B. can be determined by using SPDS, in the TSC, but ONLY if power level is in the Power Range (APRMs)
C"!" can be determined by using SPDS, in the TSC, if power level is in EITHER the Source Range (SRMs) or the Power Range (APRMs)
D. can NOT be determined due to lack of neutron monitoring indications when operating outside in the Main Control Room Thursday, January 15, 2009 10:28:10 AM 95
HL T 4 NRC Exam
==
Description:==
SPDS can be accessed in the TSC. Immediate action for 31RS-OPS-00l-l is for the SS to send someone, preferably the STA, to the TSC to operate SPDS. RC-l actions have been performed, SRM detectors and IRM detectors have been inserted, so power level will be indicated on the SRM's and/or APRMs.
A. Incorrect; Checking how many scram valves that are open is, at best, a gross estimate of control rod status. This method of checking rod position will not be accurate if there is a hydraulic lock on the SDV. The stem states that all rods did not insert, and that all the blue lights are illuminated (indicates a probable hydraulic lock ofthe SDV). An accurate power calculation is not possible using this method.
Plausible since 31RS-OPS-00l-l does direct operators to confirm the reactor is shutdown by visually confirming that each scram inlet and outlet valve is open, but only if SPDS is NOT available.
C. Incorrect; 1st part is correct (power level can still be determined using SPDS in the TSC).
The second part is not correct since SRM indication is available on SPDS.
Plausible if the candidate does not consider that RC-l being completed means that SRM detectors have been fully inserted or does not recall that SRM indication is available on SPDS.
C. Correct, see description above D. Incorrect; SPDS can still be used to determine reactor power level (SPDS can be accessed in the TSC).
Plausible since, if SPDS were not available, this would be a correct answer (there are no SRM, IRM or APRM indications at the RSDP).
APE: 295016 Control Room Abandonment AA2. Ability to determine andlor interpret the following as they apply to CONTROL ROOM ABANDONMENT: (CFR: 41.10/43.5/45.13)
AA2.01 Reactor power......................................... RO 4.1
- 1 SRO 4.1
- References used to develop this question:
31 RS-O PS-OO 1-1, Shutdown from outside control room 34AB-C71-001-1, Scram Procedure X75-SPDS-LP-05601, SPDS System 056.001.A.Ol Given a list of plant locations, select the locations where SPDS consoles are found Reference(s) provided to the student:
None Thursday, January 15, 200910:28:11 AM 96
HL T 4 NRC Exam
- 46. 295018AK3.03 OOl/1/1IRBCCWINEWIHIGHERlHT2009-301IROIEUICMEI Unit 2 is at 0% power after being manually scrammed from 100% power.
o The "2A" Reactor Building Closed Cooling Water (RBCCW) pump is running o "2B" and "2C" RBCCW pumps can NOT be started All actions associated with 34AB-P42-001-2, "Loss of Reactor Building Closed Cooling Water" for plant conditions in which only ONE RBCCW pump running have been performed.
The actions will ensure which ONE of the following?
A. A Control Rod Drive pump is running and is adequately cooled.
B. The Reactor Recirculation pumps are running and are adequately cooled.
C. A Fuel Pool Cooling pump is running and is protected from cavitation damage.
D'!' The "2A" RBCCW pump is running and is protected from pump run-out conditions.
Thursday, January 15,200910:28:12 AM 97
HL T 4 NRC Exam
==
Description:==
IA W 34AB-P42-001-2, Loss of Reactor Building Closed Cooling Water, NOTE: lAW the FSAR, the following step is performed to ensure adequate RBCCW flow inside containment with a single operating RBCCW pump AND will prevent possible pump runout by isolating RBCCW to Reactor Building equipment on elevations 158 FT AND above.
This is done by closing isolation valves 2P42-F033 and 2P42-F034. This action isolates RBCCW flow to the Recirc MG sets and diverts any cooling to the Reactor Recirc components in the DW through an orifice (true for the CRD pumps too). This section of procedure requires the following to be tripped; RWCU (tripped and isolated), Reactor Recirc Pumps, CRD Pumps (UNLESS they are needed for Control Rod Insertion AND their pumping medium is less than 250°F)
The FPC pump will remain running. The procedure does not require securing the FPC pump, it does require the fuel pool temps be monitored. As Fuel Pool water temp goes up, the FPC system will get closer and closer to cavitation conditions. Fuel Pool saturation conditions can be reached in about 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
A. Incorrect, see description above Plausible since some RBCCW flow still goes to the CRD pumps through an orifice B. Incorrect; see description above Plausible since some RBCCW flow still goes to the Reactor Recirc pumps through an orifice C. Incorrect; see description above Plausible if the candidate thinks that RBCCW flow to the FPC heat exchanger is aligned.
Also plausible because the pump will be running (procedure does not require securing this pump, it does require the fuel pool temps be monitored)
D. Correct; see description above Thursday, January 15, 200910:28:13 AM 98
HL T 4 NRC Exam APE: 295018 Partial or Complete Loss of Component Cooling Water AK3. Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER: (CFR: 41.5/45.6)
AK3.03 Securing individual components (prevent equipment damage)....... RO 3.1/ SRO 3.3 Reference(sl used to develop this question:
34AB-P42-001-2, Loss of Reactor Building Closed Cooling Water LT-LP-20201, "Introduction to Abnormal Procedures Lesson Plan LT-LP-20201.019 P42-RBCCW-LP-00901, "RBCCW" Lesson Plan FSAR Reference(sl provided to the student:
None Thursday, January 15, 200910:28:14 AM 99
HL T 4 NRC Exam
- 47. 295019AA2.02 00 111/1 lAIR MSIV/BANKlHIGHERlHT2009-301IRO/EUICMEI Unit 2 is operating at 100% power.
o 2P52-F565, "Rx Bldg Inst N2 To Non-Int Air Bl185 Isol Vlv", has been tagged in the closed position.
o Unit 2 experiences a loss of all Unit 2 Station Service Air Compressors.
o The air cross-tie valve between Unit 1 and Unit 2 cannot be opened due to a bent stem.
Which one of the following predicts how the MSNs will respond?
A. Inboard and Outboard MSNs will remain OPEN B. Inboard and Outboard MSN s will eventually drift CLOSED C'!" The Inboard MSIVs will remain OPEN The Outboard MSN s will eventually drift CLOSED D. The Inboard MSIV s will eventually drift CLOSED The Outboard MSN s will remain OPEN
==
Description:==
The Inboard MSIV s receive pneumatic supply pressure from nitrogen, via the drywell pneumatic system. The instrument air system, normally, has no connection with the Drywell Pneumatics system. The outboard MSN s normally receive pneumatic supply pressure from the Instrument Air system which is supplied from the Station Service Air Compressors.
The backup nitrogen supply valve (2P52-F565) opens when Instrument Air pressure is lost.
MSIV s drift closed on a loss of pneumatic supply pressure.
A. Incorrect The outboard MSNs will go closed. Plausible if the candidate thinks nitrogen is the normal supply for all MSN s.
B. Incorrect Only the outboard MSN s will go closed based on the conditions given. Plausibility; the candidate assumes that air is normal pneumatic supply for all MSIV s and that nitrogen backup is not available any of the MSNs since 2P52-F565 is tagged closed.
C. Correct; see description above.
D. Incorrect The opposite situation is true, the inboard valves will remain open and the outboard valves will drift closed. Plausibility; a candidate can easily confuse which valves use nitrogen as the normal pneumatic supply and those which use nitrogen as a backup pneumatic supply.
Thursday, January 15,200910:28:15 AM 100
HL T 4 NRC Exam APE: 295019 Partial or Complete Loss of Instrument Air AA2. Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR:
(CFR: 41.10/ 43.S /4S.13)
AA2.02 Status of safety-related instrument air system loads (see AK2.1 - AK2.19).................................. RO 3.6/ SRO 3.7 AK2.0S Main steam system.....................................
Reference(sl used to develop this question:
34AB-PS1-00 1-2, "Loss of Instrument and Service Air System or Water Intrusion Into The Service Air System" PS1-PS2-P70-Plant Air-LP-03S01, "Plant Air lesson Plan" EO 200.02S.a.OS Reference(sl provided to the student:
None Thursday, January 15,200910:28:16 AM 101
HL T 4 NRC Exam
- 48. 295020AKl.05 OOl/1/2IDW COOLINGIBANK MODIHIGHERlHT2009-301IRO/EUICMEI Unit 2 is operating at 100% power.
o Nitrogen (N2) is being added to the Unit 2 Drywell (DW) from the Unit 1 N2 Storage Tank 34S0-T48-002-2, "Containment Atmosphere Control and Dilution System" section 7.3.1, "Alternate Primary Containment Nitrogen Makeup From CAD loop A, Unit 1 or Unit 2 N2 Storage Tank" is being used to add the N2.
o 2T48-Fl13, "Nitrogen to DW isolation valve" is OPEN o 2T48-Fl14, "Nitrogen to DW isolation valve" is OPEN o DW venting using Standby Gas Treatment is in progress o A fault in 2C71-P001, "RPS Power Dist Panel" results in a loss ofthe"2A" Reactor Protection System (RPS) bus o It will take 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to restore the "2A" RPS bus If the operator stationed at the 2H11-P657 panel does NOT take any action, which ONE of the following describes the operational implications for the DW?
At N2 addition to the DW will continue and a loss ofDW cooling will eventually occur due to high DW pressure.
B. N2 addition to the DW will continue and DW cooling will remain in operation indefinitely.
C. N2 addition to the DW will automatically isolate and DW cooling will remain in operation indefinitely.
D. N2 addition to the DW will automatically isolate and a simultaneous loss ofDW cooling will occur.
Thursday, January 15, 200910:28:17 AM 102
HL T 4 NRC Exam
==
Description:==
The procedure (34S0-T48-002-2) explains that 2T48-FI13, 2T48-F114 will not close on a group 2 isolation and that an operator must be stationed to close the valves in the event that an isolation signal/condition actually occurs.
As a result of the loss ofRPS, a group 2 signal will be generated; however, nitrogen will continue to be added to the DW. The group 2 signal will isolate the DW vent line up. As a result, DW pressure will increase and eventually result in a DW LOCA pressure signal (assuming no operator actions are taken, which the question stem states).
Typically nitrogen addition to the DW automatically isolates on a Group 2 isolation signal; however, when performing Alternate Nitrogen makeup (as in this question), it does not auto isolate.
A. Correct; See description above.
B. Incorrect, See description above. DW cooling will be lost eventually when a DW pressure LOCA signal occurs. 1st part is correct. 2nd part is plausible if the candidate assumes that DW venting remains in service.
C. Incorrect, See description above. Nitrogen addition will not auto isolate and DW cooling will eventually be lost when a DW pressure LOCA signal occurs. Plausible if the candidate assumes nitrogen addition auto isolates due to the Gr 2 signal generated by the loss ofRPS.
D. Incorrect, See description above. 1st part is not correct, nitrogen addition does not auto isolate. 2nd part is not correct, DW cooling is not lost until the DW pressure signal is received (in this case). Plausible if the candidate assumes that DW cooling is lost as a result of a group 2 signal which is generated due to the RPS loss. DW cooling loss would actually occur if a Gr 2 signal due to DW pressure (1.85 psig) occurs.
APE: 295020 Inadvertent Containment Isolation AKI. Knowledge of the operational implications of the following concepts as they apply to INADVERTENT CONTAINMENT ISOLATION: (CFR: 41.8 to 41.10)
AK1.05 Loss of drywell/containment cooling................... RO 3.3 /3.6 Reference(sl used to develop this question:
34S0-T48-002-2, "Containment Atmospheric Control and Dilution Systems" 34AB-C71-002-2, "Loss ofRPS" T23-PC-LP-01301, "Primary Containment" EO 013.009.A.Ol Reference(sl provided to the student:
None Thursday, January 15, 200910:28:19 AM 103
HL T 4 NRC Exam
- 49. 29502102.4.35 0011111/RHR SDCINEWIHIGHERJHT2009-301IRO/EUICMEI Unit 2 was in Mode 5 when a Loss of Shutdown Cooling (SDC) occurred.
An operator is dispatched to perform the following three steps from Attachment 4, "Shutdown Cooling Loop Preparation for Immediate Service", of34AB-E11-001-2, "Loss of Shutdown Cooling".
3.28 At 130RJR19, CONFIRM 2P11-F020B OPEN, THEN OPEN Shutdown Cooling Condensate Flush Supply, 2E 11-F084 AN D 2E 11-F083.
3.29 Monitor flow noise through the Shutdown Cooling Condensate Flush Supply, 2E11-F084 AND 2E11-F083 by listening to the pipe.
3.30 WHEN flow into the Shutdown Cooling suction pipe stops, CLOSE the Shutdown Cooling Condensate Flush Supply, 2E11-F084 AND 2E11-F083.
lAW 34AB-Ell-001-2, which ONE of the following describes the purpose of steps "3.28 and 3.29" and the operational effects of continuing to step "3.30" prior to completing step "3.29".
Steps "3.28 and 3.29" ensure that (1)
The consequences of closing 2E11-F084 with flow noise still present is (2)
A. (1) the SDC suction line is completely filled with water (2) air binding the RHR Pump when it is started B" (1) the SDC suction line is completely filled with water (2) decreasing RPV water level during SDC valve alignment C. (1) adequate flow is available for flushing the line oflow quality water (2) exceeding Tech Spec water quality limits for conductivity D. (1) adequate flow is available for flushing the line oflow quality water (2) exceeding Tech Spec water quality limits for chlorides Thursday, January 15,200910:28:20 AM 104
HL T 4 NRC Exam Description; The procedure directs the operator to close the valves 2E 11-F084 AN D 2E11-F083) when flow noise stops. Next, the RPV suction from recirc line is opened.
Plausibility; Plant Hatch had this very event in 1987. An upstream valve in the line-up to fill the RHR suction piping was closed. The operators did not hear any flow noise and believed the RHR piping was full. This was reported to the Control Room. When the SDC suction from the RPV was opened, RPV level dropped. This allowed an air bubble to migrate up, into to the RPV.
A. Incorrect; see description above. 1st part is correct. 2nd part incorrect (air bubble would migrate back to the RPV rather than the pump. Plausible if the candidate thinks the air bubble will be drawn to the pump when the pump is started, thereby, air binding the pump.
B. Correct, See description above.
. C. Incorrect; Neither part is correct. Plausible since this water is relatively stagnant and there are Tech Spec limits for water quality. The Chemistry department does draw samples for water quality when normally placing the system in service.
D. Incorrect; Neither part is correct. Plausible since this water is relatively stagnant and there are Tech Spec limits for water quality. The Chemistry department does draw samples for water quality when normally placing the system in service.
APE: 295021 Loss of Shutdown Cooling 2.4 Emergency Procedures / Plan 2.4.35 Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects. (CFR: 41.10/43.5/45.13) IMPORTANCE RO 3.8 / SRO 4.0 Reference(s) used to develop this question:
LT-LP-10012, "ECCS Industry Events" 34AB-Ell-001-2, "Loss of Shutdown Cooling" EII-RHR-LP-00701, "RHR System" lesson plan, EO 200.049.A.Ol Reference(s) provided to the student:
None Thursday, January 15,200910:28:21 AM 105
HL T 4 NRC Exam
- 50. 295023AAl.04 00 1/1/1/RADINEWIHIGHERlHT2009-30 llRO/EUICMEI An irradiated fuel bundle is on the Unit 2 Refueling Bridge Main Grapple, which is in the "FULL-UP" position, and can NOT be lowered due to an equipment malfunction.
o The Fuel Pool Transfer Canal seals deflate which causes Fuel Pool water level to decrease to its lowest possible level Which ONE of the following describes the status of the fuel bundle that is on the Refueling Bridge Main Grapple and the Area Radiation Monitor (ARM) radiation level indications that will be observed in the Main Control Room?
Fuel Pool Water level will ------------------
A. remain above the top of the Fuel Bundle and the highest indication expected on the Refuel Floor ARMs will be 4 mR/Hr B. remain above the top of the Fuel Bundle and the highest indication expected on the Refuel Floor ARMs will be 50 mR/Hr C. go below the top of the fuel bundle and the highest indication expected on the Refuel Floor ARMs will be 1,000 mRlHr D'l" go below the top of the fuel bundle and the highest indication expected on the Refuel Floor ARMs will be upscale Thursday, January 15, 2009 10:28:22 AM 106
HL T 4 NRC Exam Description; An exposed, irradiate fuel bundle will cause radiation levels of 3,000 RlHr at 30 ft (per LT-LP-10015). A seal leak on the transfer canal seals which allows water to drop to the lowest level possible will completely uncover the fuel bundle (fuel bundles are periodically transferred through the canal to move bundles from one units pool to another). The stem states the Fuel Pool level drops to the lowest possible level.
Plausibility; a significant loss of Fuel Pool water due to deflated seals did occur at Hatch (1986) and Connecticut Yankee. There was no fuel bundle on the grapple at the time. A plant recently left a bundle hanging on a grapple for an extended period of time (days) during a holiday season.
A. Incorrect; see description above. Plausible if the candidate assumes the bundle remains covered and the rad levels on the Refuel Floor remain normal.
B. Incorrect; see description above. Plausible ifthe candidate assumes the bundle remains covered but some increase in rad levels occur due to lowered Fuel Pool water level. 50 mRlHr is max normal rad levels on the Refuel Floor.
C. Incorrect; see description above. Plausible if the candidate assumes the bundle becomes partially uncovered with a significant increase in rad levels due to lowered Fuel Pool water level. 1000 mRlHr is max safe rad levels on the Refuel Floor.
D. Correct; see description above.
APE: 295023 Refueling Accidents AAI. Ability to operate and/or monitor the following as they apply to REFUELING ACCIDENTS: (CFR: 41.7/45.6)
AA1.04 Radiation monitoring equipment........................ RO 3.4 / SRO 3.7 Reference(sl used to develop this question:
LT-LP-10015, "Refueling Industry Events" LO LT-10015.001 Reference(sl provided to the student:
None Thursday, January 15, 200910:28:23 AM 107
HLT 4 NRC Exam
- 51. 295024EA2.06 OOl/I/l/TORUS HEATUPIBANK NRC 05IFUNDIHT2009-30IIROIEUICMEI A steam line break inside containment has occurred on Unit 1.
o Drywell pressure is steady at + 10.5 psig.
o Drywell or Torus sprays have NOT yet been initiated.
Which ONE of the following best describes the effect on Torus water temperature?
A. The saturation temperature of the Torus water will be lower than at normal operating parameters due to the non-condensable gases discharged to the Torus.
B~ The Torus water temperature will initially heat up evenly throughout the Torus due to the design of the downcomers.
C. The Torus water temperature will heat up more quickly below the area of the leak in the drywell due to more energy being distributed to the Torus in that area.
D. The Torus water average temperature is unreliable until suppression pool cooling is established to provide even mixing of the water.
==
Description:==
The steam will enter the torus via a ring header and downcomers. The ring header helps to ensure steam distribution is approximately equal throughout the torus.
Plausibility for local area heating of the torus is SRVs leaking.
A. Incorrect; see description above. The saturation temperature is higher due to the higher pressure. Plausible if the candidate confuses fundamental principles of Psat-Tsat relationship.
B. Correct; see description above.
C. Incorrect; see description above. With a steam leak the steam is distributed evenly through the downcomers. Plausible if the candidate does not understand the purpose/design of the ring header/ downcomers.
D. Incorrect; see description above. The temperature monitors of the torus still work and the average temp is just an average of all the monitors. Plausible if the candidate is not familiar with how water temperature is measured and monitored in the suppression chamber.
Thursday, January 15,200910:28:25 AM 108
HL T 4 NRC Exam EPE: 295024 High Drywell Pressure EA2. Ability to determine and/or interpret the following as they apply to HIGH DRYWELL PRESSURE: (CFR: 41.10/43.5/45.13)
EA2.06 Suppression pool temperature.......................... RO 4.1 / SRO 4.1 Reference(sl used to develop this question:
Steam Tables T23-PC-LP-01301, "Primary Containment" lesson plan EO 200.004.A.03 Reference(sl provided to the student:
None Thursday, January 15,200910:28:25 AM 109
HL T 4 NRC Exam
- 52. 295025G2.1.23 OOl/llllHPCI EOP-I07IBANKMODIHIGHERlHT2009-301IROIEUICMEI 31EO-EOP-107-2, "ALTERNATE RPV PRESSURE CONTROL" is in progress.
o The HPCI system is being used to control reactor pressure.
o The HPCI flow controller setpoint is at its normal standby value lAW 31EO-EOP-107-2, with the HPCI Turbine controller in automatic mode, to INCREASE the reactor cooldown rate (CDR), the operator is directed to throttle
- 0) in the (2) direction.
A"! (1) 2E41-F008, "Test to CST VLV" (2) close B. (1) 2E41-F008, "Test to CST VLV" (2) open C. (1) 2E41-FOll, "Test to CST VLV" (2) close D. (1) 2E41-FOll, "Test to CST VLV" (2) open Thursday, January 15,200910:28:26 AM 110
HL T 4 NRC Exam Description; While HPCI is in pressure control mode with the controller in automatic, per procedure the cooldown rate (CDR) is controlled by throttling 2E41-F008, "Test to CST VLV".
31EO-EOP-107-2 specifies that throttling F008 in the closed direction will increase the CDR if the controller is in auto. If the controller is in Manual, throttling F008 will have minimal effect on CDR. In Manual the CDR is increased by increasing the controller output and decreased by reducing the controller output.
This concept has been difficult for some students to master (which direction to throttle the valve to increase CDR).
A. Correct; see description above.
B. Incorrect, 1st part is correct, 2nd is not correct, opening the valve will reduce the CDR.
Plausible if the candidate assumes that opening the valve results in more water flow, which would require more steam flow.
C. Incorrect, 1st part is not correct (wrong valve). 2nd part is correct. Plausible if the student does not remember which valve is throttled to control CDR. The valves (F008 & FOIl) are in series and have the same name.
D. Incorrect, 1st part is not correct (wrong valve). 2nd part is not correct. Plausible if the candidate assumes that opening the valve results in more water flow, which would require more steam flow.
EPE: 295025 High Reactor Pressure 2.1 Conduct of Operations 2.1.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation. (CFR: 41.10/43.5/45.2 / 45.6) IMPORTANCE RO 4.3 / SRO 4.4 Reference(s) used to develop this question:
31EO-EOP-107-2, "Alternate RPV Pressure Control" EOP-107-LP-20318, "Alter Press Control" EO 005.015.A.04 Reference(s) provided to the student:
None Thursday, January 15, 200910:28:27 AM 111
HL T 4 NRC Exam
- 53. 295026EK2.01 OOlI1I1IRHRlNEW/HlGHERlHT2009-301lRO/EUICMEI Unit 1 is operating at 90% power.
o A Safety Relief Valve (SRV) inadvertently opened causing Suppression Pool water temperature to increase o Suppression Pool water temperature reaches 97°F before operators are able to close the SRV Which ONE of the following describes the required Residual Hear Removal (RHR)
Suppression Pool Cooling alignment?
A"! Place only one loop ofRHR in Suppression Pool cooling, and the RHR heat exchanger is required to be isolated prior to starting the RHR pump.
B. Place only one loop ofRHR in Suppression Pool cooling, and the RHR heat exchanger is NOT required to be isolated prior to starting the RHR pump.
C. Place all available RHR loops in Suppression Pool cooling, and the RHR heat exchanger is required to be isolated prior to starting the RHR pumps.
D. Place all available RHR loops in Suppression Pool cooling, and the RHR heat exchanger is NOT required to be isolated prior to starting the RHR pump.
Description; Between 95F and lOOF, 34AB-T23-003-1 requires one loop ofRHR to be placed in service. When below 100F the RHR HX is required to be isolated. Above lOOF, all available RHR pumps except for pumps required for adequate core cooling. Above lOOF, the RHR heat exchanger is not required to be isolated prior to starting RHR pumps since it is being placed in service to support the EOPs.
A. Correct; see description above B. Incorrect; see description above. 1 st part is correct. 2nd part is not correct since conditions to enter the EOPs have not yet been reached. Plausible of the candidate does not recall the difference between RHR alignment requirements above and below 1 OOF.
C. Incorrect; see description above. 1 st part is not correct. This is only required if operation is being directed by the EOPs. Plausible ofthe candidate does not recall the difference between RHR alignment requirements above and below 1 OOF. 2nd part is correct.
D. Incorrect; 1st part is not correct. This is only required if operation is being directed by the EOPs. Plausible of the candidate does not recall the difference between RHR alignment requirements above and below 100F.
Thursday, January 15, 2009 10:28:29 AM 112
HL T 4 NRC Exam EPE: 295026 Suppression Pool High Water Temperature EK2. Knowledge of the interrelations between SUPPRESSION POOL HIGH WATER TEMPERATURE and the following: (CFR: 41.7/45.8)
EK2.01 Suppression pool cooling..... RO 3.9 / SRO 4.0 Reference(s) used to develop this question:
31EO-EOP-012-1, Primary Containment Control EOP flowchart 34AB-T23-003-1, Torus Temperature Above 95°F 34S0-El1-01O-1, RHR System Reference(s) provided to the student:
None Thursday, January 15,2009 10:28:29 AM 113
HL T 4 NRC Exam
- 54. 295028EAl.03 00 1/111/DW COOLINGIBANK MODIHIGHERlHT2009-30 IJROIEUICMEI Unit 2 was at 100% power when a pipe break occurred inside the Drywell (DW).
The following conditions now exist:
o Drywell Pressure:.................................. 8 psig o Bulk Average Drywell Temperature:.... 245°F lAW 31EO-EOP-100-2, "Miscellaneous Emergency Overrides", the "2A" DW Chiller -----
A. is NOT allowed to be restarted due to high DW pressure B. is NOT allowed to be restarted due to high DW temperature C'!' is allowed to be restarted. The operator must fIrst place the LOCA override switch to "BYPASS" and then reset the 86 Lockout relay at the DW Chiller breaker.
D. is allowed to be restarted. The operator must fIrst reset the 86 Lockout relay at the DW Chiller breaker and then place the LOCA override switch to "BYPASS".
==
Description:==
31EO-EOP-100 prohibits performing action to restart DW chillers and coolers when DW temp >250F if the LOCA signal is due to a LOCA (i.e. can only do it if the loss of DW cooling is due to a loss ofDW cooling rather than a leak if temp is above 250F). In this question temperature is below 250F, so the actions may be performed even though a pipe break inside containment has occurred.
The sequence of component manipulation is critical to restart the chiller. The LOCA override/bypass must be performed fIrst, or the lock out relay will not reset.
A. Incorrect; see description above (LOCA override is allowed). Plausible since DW pressure is above 1.85 psig due to an actual pipe break inside the DW.
B. Incorrect; see description above (temp is below 250F). Plausible since the candidate must recall from memory the temperature limits for restarting the DW chillers during a LOCA.
C. Correct; see description above. Temperature is below 250F, the LOCA is bypassed, the lock out is reset.
D. Incorrect; see description above (sequence is incorrect). Plausible because the actions taken are the correct actions; however, the sequence is reversed. Temperature conditions allow the restart.
Thursday, January 15,200910:28:30 AM 114
HL T 4 NRC Exam EPE: 295028 High Drywell Temperature EAt. Ability to operate and/or monitor the following as they apply to mGH DRYWELL TEMPERATURE: (CFR: 41.7/45.6)
EA1.03 Drywell cooling system...... RO 3.91 SRO 3.9 Reference(sl used to develop this question:
31EO-EOP-IOO-2, "Miscellaneous Emergency Overrides" P64-PCCCW-LP-01304, "Primary Containment Cooling and Chilled Water System" EO 013.059.A.06 Reference(sl provided to the student:
None Thursday, January 15,2009 10:28:31 AM 115
HL T 4 NRC Exam
- 55. 295030EKl.02 00 111 11 INPSHINEWIHIGHERlHT2009-3 0 lIROIEUICMEI Unit 2 is operating at 100% power when a Loss of Coolant Accident occurs.
o The High Pressure Coolant Injection system is being used to control RPV water level.
o "2A" loop of Residual Heat Removal(RHR) is placed in Suppression Pool Cooling o Torus level..................................... 135 inches o "2A" RHR flow.............................. 7,700 gpm o Torus temperature.......................... 225°P o Suppression Chamber Pressure...... 8 psig Which ONE of the following completes both of these statements?
- "2A" RHR pump operation is in the _->(,."'-1.1-) _region of the RHR NPSH Limit graph.
- "2A" RHR flow (2)
Reference provided A. (1) safe (2) must be maintained at or below its current flow rate B. (1) safe (2) should be increased to maximize suppression pool cooling C. (1) unsafe (2) changes cannot restore the pump to the safe area of the graph D'!' (1) unsafe (2) reduction to 5,000 gpm will restore operation to the safe area of the graph Thursday, January 15, 200910:28:32 AM 116
HL T 4 NRC Exam
==
Description:==
The correct answer to this question is dependent on analyzing the correct graph.
Graph selection is detennined by whether suppression pool water level is At or Above 146 inches, or Below 146 inches. Common misconception among candidates on how to use the graphs, hard copy, due to having the safe region changing as torus pressure changes.
A. Incorrect 1st part is incorrect (is correct on >146" graph). Plausible if the candidate refers to Graph 12A.
2nd part is incorrect. Actually, flow must be reduced to restore safe operation. Plausible since 7700 gpm is the procedure limit for flow on one pump. This option is dependent on the candidate referring to the wrong graph plotting operation in the safe area of the NPSH graph in the first place.
B. Incorrect 1st part is incorrect (is correct on >146" graph). Plausible if the candidate refers to Graph 12A.
2nd part is incorrect. Actually, flow must be reduced to restore safe operation. Plausible since the pump is capable of more flow and the candidate may assume this is allowed during implementation of the EOPs in which the EOPs direct all available Torus Cooling be place in service (There are conditions in the EOPs where LPCI is allowed to inject at max possible flows, without requiring flow being throttled to below 7700 GPM, e.g. adequate core cooling is not assured).
C. Incorrect 1st part is correct. This is only detennined by analyzing EOP graph 12B 2nd part is incorrect. Plausible if the candidate interpolates or uses the wrong line on the graph (below 5 psig limit), this will appear to be a correct option.
D. Correct 1st part is correct. This is only detennined by analyzing EOP graph 12B, rather than 12A.
2nd part is correct. This is only detennined by analyzing EOP graph 12B, rather than 12A Thursday, January 15, 2009 10:28:33 AM 117
HL T 4 NRC Exam Reference(sl provided to the student:
Unit 2 EOP Graph 12A, RHR Pump NPSH Limit, (Suppression Pool Water Level Below 146")
Unit 2 EOP Graph 12B, RHR Pump NPSH Limit, (Suppression Pool Water Level At or Above 146")
EPE: 295030 Low Suppression Pool Water Level EKI. Knowledge of the operational implications of the following concepts as they apply to LOW SUPPRESSION POOL WATER LEVEL: (CFR: 41.8 to 41.10)
EK1.02 Pump NPSH............ RO 3.5 / SRO 3.8 Reference(sl used to develop this question:
Unit 2 EOP Graph 12A, RHR Pump NPSH Limit, (Suppression Pool Water Level Below 146")
Unit 2 EOP Graph 12B, RHR Pump NPSH Limit, (Suppression Pool Water Level At or Above 146")
EOP-CURVES-LP-20306, "EOP Curves and Limits", EO 201.065.A.27 Reference(sl provided to the student:
Unit 2 EOP Graph 12A, RHR Pump NPSH Limit, (Suppression Pool Water Level Below 146")
Unit 2 EOP Graph 12B, RHR Pump NPSH Limit, (Suppression Pool Water Level At or Above 146")
Thursday, January 15,200910:28:34 AM 118
HL T 4 NRC Exam
- 56. 29503lEAl.OI OOII1I1ILP ECCS RHRCSINEWIFUNDIHT2009-301IROIEUICMEI Unit 2 is operating at 100% power.
o A Loss of Coolant Accident occurs which results in a reactor scram.
o Drywell pressure: 3.7 psig.
o Reactor pressure has decreased to 285 psig.
With no additional operator action, ___ will be injecting at this time.
A. NEITHER Core Spray nor RHR B. BOTH Core Spray and RHR C.,.- ONLY Core Spray D. ONLYRHR VIRHR VI CS V2RHR V2CS Shutoff Head (Referenced on the RC & RCA EOP flowcharts) 220 psig 280 psig 220 psig 383 psig Hatch V2 simulator indicates approximately 1500 gpm per CS pump at 285 psig A. Incorrect; see information above. Plausible if candidate believes shutoff head for both is less than 285 psig.
B. Incorrect; See information above. Plausible if candidate believes shutoff head is greater than 285 psig for both C. Correct; see information above.
D. Incorrect; See information above. Plausible if candidate believes shutoff head is greater than 285 psig for RHR but not CS.
Thursday, January 15, 2009 10:28:35 AM 119
HL T 4 NRC Exam EPE: 295031 Reactor Low Water Level EA1. Ability to operate and/or monitor the following as they apply to REACTOR LOW WATER LEVEL: (CFR: 41.7/45.6)
EA1.01 Low pressure coolant injection (RHR): Plant-Specific.RO 4.4* / SRO 4.4*
Reference(sl used to develop this question:
Hatch Unit 2 simulator 31EO-EOP-01O-l RC RPV Control (NonATWS) Flowchart 3 1 EO-EOP-Ol 0-2 RC RPV Control (NonATWS) Flowchart EOP-RC-LP-20308, "RPV Control (Non-ATWS)lesson plan EO 201.065.A.16 Reference(sl provided to the student:
None Thursday, January 15, 200910:28:35 AM 120
HL T 4 NRC Exam 57. 295032EAl.03 00 11 1 121RB VENTINEWIFUND/HT2009-301/RO/EUICMEI Unit 2 is operating at 100% power.
The "A" Reactor Core Isolation Cooling (RCIC) Pump Room Cooler (2T41-B004A) is operating with the control switch in the "RUN" position.
o A Plant Service Water (PSW) system problem results in "0" PSW flow to 2T41-B004A.
When will the "2T41-B004B" RCIC Pump Room Cooler automatically start?
A. ONLY when RCIC is started.
B. ONLY due to high RCIC room temperature.
C'!' It requires EITHER, a high RCIC room temperature OR a RCIC start.
D. It requires BOTH, a high RCIC room temperature AND a RCIC start.
Description; The standby RCIC room cooler for RCIC will auto start when room temperature exceeds its high temperature setpoint -Of-when the RCIC system is started.
A. Incorrect; high room temperature will also start the standby cooler. Plausible since this is one of the signals that causes the standby cooler to start and tests the candidates knowledge concerning these signals.
B. Incorrect, a RCIC start will also start the standby cooler. Plausible since this is one of the signals that causes the standby cooler to start and tests the candidates knowledge concerning these signals.
C. Correct; see description above D. Incorrect, it doesn't take both signals, either one alone will start the standby cooler.
Plausible since either of these signals will cause the standby cooler to start and tests the candidates knowledge concerning these signals.
Thursday, January 15, 200910:28:36 AM 121
HL T 4 NRC Exam EPE: 295032 High Secondary Containment Area Temperature EAt. Ability to operate and/or monitor the following as they apply to mGH SECONDARY CONTAINMENT AREA TEMPERATURE: (CFR: 41.7/45.6)
EA1.03 Secondary containment ventilation..........,.......... RO 3.7/ SRO 3.7 Reference(s) used to develop this question:
T41-SC HV AC-LP-01303, "Secondary Containment HVAC Systems" EO 037.003.A.06 34AR-650-234-2, "SEC System Auto Initiation Signal Present" Reference(s) provided to the student:
None Thursday, January 15, 2009 10:28:37 AM 122
HL T 4 NRC Exam
- 58. 295035EA2.02 00ll1l2/SC DPINEWIFUND/HT2009-301IROIEUICMEI Unit 1 is operating at 100% power.
A.
B.
o A Main Steam Line (MSL) rupture inside the Reactor Building occurs o The leak can NOT be isolated o The rupture results in a pressure spike that "blows out" the Reactor Building to Turbine Building blowout panels As compared to the same MSL rupture in which the blowout panels remained intact, this event will result in OFF SITE release rates near the site boundary being
- 0)
The release will be a(an)
(2) release.
(1) higher
. (2) elevated (1) the same (2) elevated C!' (1) higher (2) ground level D. (1) the same (2) ground level Thursday, January 15,200910:28:37 AM 123
HL T 4 NRC Exam Description; The blowout panels will allow radioactive steam to enter into the Turbine Building. The Turbine Building ventilation system discharges into the Reactor Building Ventilation system and results in a ground level type of release. A ground level release will result in a higher concentration of radiation in the vicinity of the Plant Hatch site. If the blowout
. panels had remained in tact, the radioactivity would have been processed via SBGT filters, with SBGT flow being sent to the Main Stack. For a release to considered as an elevated release, the release must occur via the Main Stack.
A. Incorrect, see description above (release is not elevated). The 1st part is correct. The 2nd part is incorrect. Plausible since the candidate may assume the release is elevated if it goes out the Reactor Building Ventilation Stack - or - if the candidate thinks the TB exhaust goes to the Main Stack.
B. Incorrect, see description above (some of the radioactive steam will not be filtered via the SBGT system filter and it is a ground level release, so rad levels near the plant site will go up). The 1 st part is NOT correct. Plausible if the candidate assumes the release is elevated and that an elevated release will carry any radioactivity away from the plant. The 2nd part is incorrect. Plausible since the candidate may assume the release is elevated if it goes out the Reactor Building Ventilation Stack - or - if the candidate mistakes the TB exhaust as going to the Main Stack.
D. Correct; see description above.
C. Incorrect; see description above. 1 st part is NOT correct. Plausible if the candidate assumes the TB ventilation flow is adequately treated, to remove radioactivity, when it passes through the RB ventilation system. 2nd part is correct.
EPE: 295035 Secondary Containment High Differential Pressure EA2. Ability to determine and/or interpret the following as they apply to SECONDARY CONTAINMENT HIGH DIFFERENTIAL PRESSURE: (CFR: 41.8 to 41.10)
EA2.02 tOff-site release rate: Plant-Specific................ RO 2.8* / SRO 4.1 Reference(sl used to develop this question:
EOP-SCRR-LP-20325, "Secondary Containment Radioactive Release Control" EO 201.082.A.Ol T22-SC-LP-01302, "Secondary Containment" lesson plan EO 200.023.B.07 Reference(sl provided to the student:
None Thursday, January 15,2009 10:28:39 AM 124
HL T 4 NRC Exam
- 59. 295037EK2.12 OOlllllIRWMlNEWIFUNDIHT2009-301IROIEUICMEI Unit 1 was operating at 100% power when a reactor scram occurred.
After the scram, the following conditions exist:
o Reactor power..... 6%
o Control rod insertion is in progress IA W 31EO-EOP-103-2 "Control Rod Insertion Methods" Without further Reactor Engineering evaluation, the operating crew can determine that the reactor will remain shutdown under ALL conditions when -------
A':I all control rods are at position "02" B. all IRM indications are stable on Range 5 C. Standby Liquid Control(SLC) tank level is at 20%
D. two control rods at position "04" and all other control rods at position "00" Description; The reactor is shutdown under all conditions when:
o All control rods are position "02" -or-o All rods are at "00" with the exception of the highest worth control rod.-or-o Cold Shutdown Boron Weight has been injected into the RPV (14% tank level)
Hot Shutdown Boron Weight = 35% boron tank level.
IRMs below range 6 with no boron injection is an EOP definition for "subcritical"; however, this is not a definition for "shutdown under all conditions".
A. Correct; see description above.
B. Incorrect; see description above. (it is the definition subcritical, not shutdown under all conditions). Plausible if the candidate confuses the definitions for the conditions.
C. Incorrect; see description above, (not Cold Shutdown Boron Weight level yet). Plausible because the candidate must know, from memory, the value for Cold Shutdown Boron Weight to eliminate this distractor. Tank level below 14% would make this a correct answer.
D. Incorrect; see description above, (does not meet a definition of shutdown under all conditions). Plausible because this represents a much higher rod density than either all rods at position "02" or all rods at position"OO" with one rod fully withdrawn.
Thursday, January 15, 200910:28:41 AM 125
HL T 4 NRC Exam EPE: 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown EK2. Knowledge ofthe interrelations between SCRAMCONDITION PRESENT AND REACTOR,POWER ABOVE APRM DOWNSCALE OR UNKNOWN and the following:
(CFR: 41.7/45.8)
EK2.12 Rod control and information system: Plant-Specific... RO 3.6/ SRO 3.8 Reference(sl used to develop this question:
CII-RWM-LP-05403 EOP-CVRVES-LP-20306, "EOP Curves and Limits" EO 201.070.A.03 (CSBW)
EOP-RCA-LP-20328, "RCA Flowchart", EO 201.070.A.06 (Subcritcal)
VI TS definitions SDM (One rod full out)
Reference(sl provided to the student:
None Thursday, January 15, 200910:28:41 AM 126
HL T 4 NRC Exam
- 60. 295038EKl.02 00 1/1/11MSIV CWSING TIMEIBANK MODIFUND/HT2009-301IROIEUICMEI Unit 1 is operating at 18% power.
The following alarms are received soon after a cold water injection into the reactor:
o MAIN STEAM LINE RADIATION HIGH (601-425) o MAIN STEAM LINE RADIATION HIGH-HIGH/INOP (603-125) o Main Steam Line radiation levels are 6,000 mRemlhr (increasing) o Main Steam Isolation Valves are OPEN o IB31-FOI9 and IB31-F020, "Reactor Recirculation Sample Valves" are CLOSED IA W 34AB-B21-001-1, "Main Steam Line High Radiation or Suspected Fuel Element Failure",
expected automatic actions (1) occurred.
lAW 34AB-B21-00 1-1, the required action( s) to protect the public from radioactive release include (2)
A't (1) have (2) scramming the reactor and closing the MSIVs B. (1) have (2) performing a fast reactor shutdown IA W 34GO-OPS-014, "Fast Reactor Shutdown" C. (1) have NOT (2) scramming the reactor and closing the MSIVs D. (1) have NOT (2) performing a fast reactor shutdown lAW 34GO-OPS-014, "Fast Reactor Shutdown" Thursday, January 15, 200910:28:42 AM 127
HL T 4 NRC Exam
==
Description:==
The Recirc Sample valves auto close when MSL rad monitors exceed 2.5 x normal. The setpoint also results in MAIN STEAM LINE RADIATION HIGH-HIGH/INOP (603-125). The MSIV auto closure at this setpoint is no longer applicable at Plant Hatch. If it is suspected that the cause of this alarm is Fuel Element Failure (FEF), 34AB-B21-001-1 requires the reactor be scrammed and MSIVs closed to limit rad release to the public.
A. Correct; see description above.
B. Incorrect; see description above. 1st part is correct. 2nd part is not correct (requirement is to scram rather than perform a Fast Reactor Shutdown). Plausible because a Fast Reactor Shutdown is the required action if the cause ofthe alarms is NOT suspected to be FEF -or-only if the MSL Rad High alarm is received when FEF is suspected..
C. Incorrect; see description above. 1st part is not correct (Recirc Sample Valves isolated as required) plausible if the candidate thinks the MSIV s auto close on high MSL radiation (this used to be an auto isolation at Hatch). 2nd part is correct.
D. Incorrect; see description above. 1st part is not correct (Recirc Sample Valves isolated as required) plausible if the candidate thinks the MSIVs auto close on high MSL radiation (this used to be an auto isolation at Hatch). 2nd part is not correct (requirement is to scram rather than perform a Fast Reactor Shutdown). Plausible because a Fast Reactor Shutdown is the required action if the cause of the alarms is NOT suspected to be FEF -or-only if the MSL Rad High alarm is received when FEF is suspected..
EPE: 295038 High Off-Site Release Rate EKI. Knowledge of the operational implications of the following concepts as they apply to HIGH OFF-SITE RELEASE RATE: (CFR: 41.8 to 41.10)
EK1.02 tProtection of the general public...................... RO 4.2* / SRO 4.4*
Reference(s) used to develop this question:
B21-SLLS LP-01401, "Main Steam and Low Low Set" lesson plan EO 200.098.A.Ol 34AB-B21-001-1, "Main Steam Line High Radiation or Suspected Fuel Element Failure" MAIN STEAM LINE RADIATION HIGH-HIGH/INOP (603-125)
Reference(s) provided to the student:
None Thursday, January 15, 200910:28:43 AM 128
HL T 4 NRC Exam
- 61. 300000Kl.03 OOl/211/DW PNEUMATICSINEWIFUND/HT2009-301IROIEUICMEI Unit 2 is shutting down for a refueling outage.
o Instrument Air has been aligned to the Drywell (DW) Pneumatic system o At 10:00 an airline break inside the DW results in an air flow rate of 50 SCFM Air being supplied to the DW through this DW Pneumatics header (1) automatically isolate (2)
A. (1) will (2) immediately B~ (1) will (2) following a 10 minute delay C. (1) will NOT (2) since the Instrument Air flow is below the DW Pneumatics isolation setpoint D. (1) will NOT (2) since the Instrument Air to DW Pneumatics connection point is downstream of the DW Pneumatics automatic isolation valves
==
Description:==
Air is aligned upstream of the flow sensor which causes the isolation at 30 CFM.
There is a 10 minute time delay associated with the isolation.
A. Incorrect; see description above. 1st part is correct. 2nd part is NOT correct (there is a 10 minute time delay)
B. Correct; see description above.
C. Incorrect; see description above.lst part is not correct. 2nd part is not correct (An isolation will occur). Plausible since the candidate must recall the setpoint for the isolation from memory to eliminate this distractor.
D. Incorrect; see description above. 1st part is not correct. 2nd part is not correct. Plausible since the candidate must recall the piping arrangement from memory to eliminate this distractor. The candidate must also differentiate between this connection and the Emergency Nitrogen (N2 Bottles) supply for SRV operation which taps in downstream of the flow sensors which cause this isolation.
Thursday, January 15, 2009 10:28:45 AM 129
HL T 4 NRC Exam SYSTEM: 300000 Instrument Air System (lAS)
Kl Knowledge of the connections and / or cause effect relationships between INSTRUMENT AIR SYSTEM and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8)
Kl.03 Containment air............ RO 2.8 / SRO 2.9 Reference(sl used to develop this question:
P51-P52-P70-Plant Air-LP-03501, "Plant Air Systems" lesson plan EO 042.001.A.Ol &
EO 042.001.A.03 34AR-700-120-2 "DRWL Pneu Sys Supply Line Press Low MSIV" 34AR-700-133-2 "DRWL Pneu Sys Supply Line Press Low" Reference(sl provided to the student:
None Thursday, January 15,2009 10:28:45 AM 130
HL T 4 NRC Exam
- 62. 400000K3.0 1 00 1I2/1/PSW /NEW /FUNDIHT2009-30 lIROIEU/CME/
Unit 1 was operating at 100% power when a scram occurred.
o After the scram, the "IE" 4160 VAC bus de-energized and was re-energized from the lA" Emergency Diesel Generator (EDG) o A total loss of Plant Service Water occurs With NO operator action, which ONE of the following describes how the "IA" EDG will respond?
The "lA" EDG will --------------------
A"!' continue to run until major component damage occurs B. trip due to high jacket coolant temperature C. trip due to high lube oil temperature D. trip due to high crank case pressure
==
Description:==
The EDGs have several automatic trips that only function when the EDG is in Test Mode, including high jacket coolant temperature, high lube oil temperature, high crank case pressure. When the EDG is NOT in test, as is the case in this question, those automatic trips are disabled. The only trips that are in effect in this case are overspeed. low lube oil pressure, start failure and differential current.
A. Correct; see description above.
B. Incorrect; see description above. Only functions in test. Plausible since this is an actual EDG trip when the EDG is in test mode. The loss ofPSW cooling will result in temperatures increasing. The candidate must recall from memory when the high temperature trip is in effect.
C. Incorrect; see description above. Only functions in test. Plausible since this is an actual EDG trip when the EDG is in test mode. The loss of PSW cooling will result in temperatures increasing. The candidate must recall from memory when the high temperature trip is in effect.
D. Incorrect; see description above. Only functions in test. Plausible since this is an actual EDG trip when the EDG is in test mode. The loss ofPSW cooling will result in this pressure increasing. The candidate must recall from memory when the high pressure trip is in effect.
Thursday, January 15,200910:28:47 AM 131
HLT 4 NRC Exam SYSTEM: 400000 Component Cooling Water System (CCWS)
K3. Knowledge of the effect that a loss or malfunction of the CCWS will have on the following: (CFR: 41.7 / 45.6)
K3.01 Loads cooled by CCWS.............. RO 2.9 / SRO 3.3 Reference(s) used to develop this question:
34AB-P41-001-1, "Loss of Plant Service Water" 34S0-R43-001-1, "Diesel Generator Standby AC System" R43-EDG-LP-02801, "Emergency Diesel Generators" lesson plan EO 028.023.A.02, P41-PSW-LP-03301 "PSW System" lesson plan EO 200.013.A.05 34AR-R43-104-1, "Jacket Coolant Temp High" 34AR-R43-102-1, "Lube Oil Temp High" 34AR-R43-111-1, "Crank Case Pressure" Reference(s) provided to the student:
None Thursday, January 15, 2009 10:28:47 AM 132
HLT 4 NRC Exam
- 63. 500000EIG.06 OOll1/2/TORUS VENTINEWIHIGHERlHT2009-301IROIEUICMEI Unit 1 has experienced a transient that results in these Primary Containment parameters:
o Hydrogen concentration.... 8%
o Oxygen concentration....... 7%
o Drywell (DW) pressure.... 14 psig and slowly increasing o Torus level........................ 280 inches (steady)
Which ONE of the following describes the Emergency Operating Procedures (EOPs) strategy and reason(s) for the strategy for these conditions?
The EOP strategy is to (1)
The reason( s) for taking this action is ___
-'.!(2:.,r.) ____ reduce the flammability of the Primary Containment atmosphere.
A. (1) spray the DW (2) scrub radionuc1ides AND B. (1) spray the DW (2) ONLYto C'!" (1) vent the Suppression Chamber (2) scrub radionuc1ides AND D. (1) vent the Suppression Chamber (2) ONLYto Thursday, January 15, 2009 10:28:48 AM 133
HL T 4 NRC Exam
==
Description:==
With the concentrations ofH2 and 02 given in this question, the EOPs direct the operators to:
o Vent the Suppression Chamber if Suppression Pool water level is below 300" o Vent the DW if Suppression Pool water level is > 300" o Spray the DW and Suppression Chamber Spraying containment will cause some of the radionuclides to be entrained in the water (scrubbing) and reduce the flammability of the Primary Containment atmosphere by adding water vapor to the atmosphere.
A. Incorrect; see description above. 1st part is incorrect (DW spray is not permitted when torus level is > 197.5 inches). Plausible because this is a required action for high H2 and 02 concentrations and the only reason it cant be performed is due to Suppression Chamber level.
The candidate must be aware of this EOP flowchart limitation to eliminate this distractor. 2nd part is a correct statement.
B. Incorrect; see description above. *1 st part is incorrect (DW spray is not permitted when torus level is > 197.5 inches). Plausible because this is a required action for high H2 and 02 concentrations and the only reason it cant be performed is due to Torus level. 2nd part is a partially correct statement.
C. Correct; see description above.
D. Incorrect; see description above. 1st part is correct (venting the Torus is required). 2nd part is only partially correct and states this is the ONLY reason.
Plausible because this is a required action for high H2 and 02 concentrations and since the candidate must know that there are two reasons for performing the action.
EPE: 500000 High Containment Hydrogen Concentration EK3. Knowledge of the reasons for the following responses as they apply to HIGH PRIMARY CONTAINMENT HYDROGEN CONCENTRATIONS: (CFR: 41.5/45.6)
EK3.06 Operation of wet well vent........... RO 3.1 / SRO 3.7 Reference(sl used to develop this question:
EOP-PC-LP-2031O, "Primary Containment Control" lesson plan EO 201.072.A.27, EO 201.072.A.25, EO 201.072.A.19 3IEO-PCG-001-1, "Primary Containment Gas Control" EOP Flowchart Reference(sl provided to the student:
None Thursday, January 15, 2009 10:28:49 AM 134
HL T 4 NRC Exam
- 64. 600000AKl.Ol 001l1l11FlREINEWIFUND/HT2009-301IROIEUICMEI IA W 34AB-X43-001-2, "Fire Procedure", if a class __ fire exists, the affected component must be de-energized.
A.A B. B Ct C D. D
==
Description:==
Fire classifications:
Class A - Fire that leaves an ash (wood, paper or trash)
Class B - Flammable or combustible liquids or gases)
Class C - Energized Electrical equipment fire Class D - Combustible metals A. Incorrect; see description above. It is actually an electrical fire Plausible since the candidate must recall from memory the correct fire classification B. Incorrect; see description above. It is actually an electrical fire Plausible since the candidate must recall from memory the correct fire classification C. Correct; see description above.
D. Incorrect; see description above. It is actually an electrical fire Plausible since the candidate must recall from memory the correct fire classification Thursday, January 15, 2009 10:28:50 AM 135
HL T 4 NRC Exam APE: 600000 Plant Fire On Site AKl Knowledge of the operation applications of the following concepts as they apply to Plant Fire On Site:
AK1.01 Fire Classifications by type........ RO 2.5 / SRO 2.8 Reference(s) used to develop this question:
Hatch GET handbook X43-FPS-LP-03601, "Fire Protection System", lesson plan EO 200.092.A.Ol 34AB-X43-001-2, "Fire Procedure" Reference(s) provided to the student:
None Thursday, January 15, 2009 10:28:51 AM 136
HL T 4 NRC Exam
- 65. 700000G2.4.4 OOl/1/1IDEGRADED GRIDINEWIHIGHERlHT2009-301IRO/EUICMEI Unit 1 and Unit 2 are operating at 100% power.
Over the past hour the following has occurred:
o Voltage in the 230 KV switchyard has decreased from 238KV to 234KV.
o 4160 VAC bus "IE"............................................ De-energized due to a ground fault o 4160 VAC bus "IF" and "lG" bus voltage......... Decreased to 3840 VAC The following Unit 1 annunciators have been received:
o (4160V BUS 1E) Loss Of Offsite Power (652-102-1) o 4160V BUS 1E VOLTAGE LOW (652-122-1) o 4160V BUS 1F VOLTAGE LOW (652-222-1) o 4160V BUS 1G VOLTAGE LOW (652-322-1)
. Select the procedure that is required to be entered and also contains actions that are required to be performed.
A. 34AB-R22-003-1, "Station Blackout" due to the Loss of Offsite Power annunciator.
B. 34AB-R22-003-1, "Station Blackout" due to the 4160 VAC bus low voltage annunciators.
C~ 34AB-S 11-001-0, "Operation With Degraded System Voltage" due to the 4160 V AC bus low voltage annunciators.
D. 34AB-S 11-00 1-0, "Operation With Degraded System Voltage" due to the Loss of Offsite Power annunciator.
Thursday, January 15,2009 10:28:52 AM 137
HLT 4 NRC Exam
==
Description:==
Conditions given in the question require entry into 34AB-S11-001-0, Degraded System Voltage. 4160 V AC bus low voltage annunciator setpoint = 3867 V AC. These annunciators are entry conditions for 34AB-S11-001-0.
Entry conditions have been met for a station blackout procedure; however, since 2 Emergency Buses are energized, the procedure is immediately exited without any actions required to be performed. Requirements for entering station blackout abnormal: At least two of the three Unit 1 4160V AC Emergency busses IE, IF, AND IG are de-energized, as indicated by TRIPPED breaker indications AND extinguished bus pot lights on IHll-P652. Requirements to exit this procedure:
4.4.2 IF power is restored to more than one of the 4160V AC Emergency busses, THEN exit this procedure.
A. Incorrect; see description above (more than 1 Emergency Bus is energized). Plausible because the candidate must know the procedure requirements for exiting 34AB-R22-003-1 from memory in order to eliminate this distractor. Additionally the LOSP annunciator is an entry condition for 34AB-R22-003-1.
B. Incorrect; see description above (more than 1 Emergency Bus is energized). Plausible because the candidate must know the procedure requirements for exiting 34AB-R22-003-1 from memory in order to eliminate this distractor. Additionally the low voltage annunciators are entry conditions for 34AB-R22-003-1.
C. Correct; see description above.
D. Incorrect; see description above. 1 st part is correct. 2nd part is not correct since a LOSP annunciator is not an entry condition for 34AB-S 11-001-0. Plausible because a Loss of Site Power annunciator implies a degraded power condition. This alarm will be received if bus voltages continue to degrade (3,255 V AC).
Thursday, January 15, 200910:28:53 AM 138
HL T 4 NRC Exam APE: 700000 Generator Voltage and Electric Grid Disturbances 2.4 Emergency Procedures / Plan 2.4.4 Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.
(CFR: 41.10 / 43.2 /45.6) IMPORTANCE RO 4.5 / SRO 4.7 Reference(sl used to develop this question:
34AB-S 11-00 1-0, "Operation With Degraded System Voltage" 34AB-R22-003-1, " Station Blackout" Sll-LP-02706, "Basic Grid Operating Concepts" EO 200.116.A.Ol ARP 34AR-652-122-1 "4160V Bus IE Voltage Low" ARP 34AR-652-102-1 "Loss of Off site Power" Reference(sl provided to the student:
None Thursday, January 15,200910:28:54 AM 139
HL T 4 NRC Exam
- 66. G2.1.27 OOl/3/1IRWMlNEWIFUNDIHT2009-301IRO/EUICMEI Select the ONE choice which completes the following statement to defme the primary purpose of the Rod Worth Minimizer (RWM).
- The primary purpose of the RWM is to ensure that _________ _
A. fuel enthalpy does not exceed 280 cal/gm during a control rod drop accident when power is ~29%
B'" fuel enthalpy does not exceed 280 cal/gm during a control rod drop accident when reactor power is < 10%
C. the Minimum Critical Power Ratio remains greater than 1.08, while withdrawing control rods, when power is ~ 29%
D. the Minimum Critical Power Ratio remains greater than 1.08, while withdrawing control rods, when reactor power is < 10%
Thursday, January 15,200910:28:54 AM 140
HL T 4 NRC Exam
==
Description:==
The purpose of the RWM system is to limit control rod worth such that the fuel enthalpy limit of280 callgm will not be exceeded during a Control Rod Drop Accident (CRDA).
TS Table 3.3.2.1-1 requires the RWM to be operable in modes 1 and 2 with thermal power
<lO%RTP.
This question tests the purpose of the RWM by using aspects of the RBM purpose as distracters.
The purpose of the RBM is to limit control rod withdrawal iflocalized neutron flux exceeds a predetermined setpoint during control rod manipulations. It is assumed to function to block further control withdrawal to preclude a MCPR Safety Limit (SL) violation.
The RBM is required to be operable when power is >=29% power On Unit 2, MCPR limit is ::::1.08 for two Recirc Pump operation.
A. Incorrect; see description above. 1st part is correct. 2nd part is incorrect. Plausible because
>= 29% is the TS requirement for the RBM, and the candidate may confuse the requirements between the RBM and RWM.
B. Correct; see description above.
C. Incorrect; see description above. 1 st part is incorrect. Plausible because the RBM does provide rod blocks to prevent MCPR from being exceeded. 2nd part is incorrect. Plausible because >= 29% is the TS requirement for the RBM, and the candidate may confuse the requirements between the RBM and RWM.
D. Incorrect; see description above. 1st part is incorrect. Plausible because the RBM does provide rod blocks to prevent MCPR from being exceeded due to additional rod withdrawal.
2nd part is incorrect. 2nd part is incorrect. Plausible because >= 29% is the TS requirement for the RBM, and the candidate may confuse the requirements between the RBM and RWM.
2.1 Conduct of Operations 2.1.27 Knowledge of system purpose and/or function. (CFR: 41.7)
IMPORTANCE RO 3.9/ SRO 4.0 55.41 b.6 Reference(sl used to develop this question:
CII-RWM-LP-05403, Rod Worth Minimizer lesson plan. EO 001.010.D.Ol C51-PRNM-LP-01203, Power Range Neutron Monitoring TS 2.1 Safety limits TS Table 3.3.2.1-1, Control Rod Block Instrumentation Reference(sl provided to the student:
None Thursday, January 15, 2009 10:28:56 AM 141
HL T 4 NRC Exam
- 67. G2.1.31 001l311IRCIC RESTARTIBANKMODIHIGHERlHT2009-301IROIEUICMEI Unit 2 RCIC has tripped with an Initiation Signal present. The trip condition has been corrected and RCIC is needed for" NOTE: The light next to the "M" on the above controller is ILLUMINATED.
IA W 34S0-E51-001-2, "Reactor Core Isolation Cooling (RCIC) System", the controller (1) setup for RCIC start with an initiation signal present.
This controller is located on panel (2)
A. (1) IS (2) 2Hll-P601 B. (1) IS NOT (2) 2Hll-P601 C. (1) IS (2) 2HI1-P602 O'!' (1) IS NOT (2) 2HI1-P602 Thursday, January 15, 2009 10:28:56 AM 142
HL T 4 NRC Exam
- 67. G2.1.31 00 1/3/1IRCIC RESTARTIBANK MODIHIGHERlHT2009-30 lIROIEUICMEI Description; Following a RCIC trip, to start RCIC for injection to the RPV, the procedure requITes:
7.1.4.2 IF RCIC Initiation Signal CAN NOT BE RESET/Flow Controller Failure, THEN perform the following:
7.1.4.2.1 Transfer 2E51-R612, Turbine Controller, to MANUAL and adjust output to 50%.
7.1.4.2.2 IF necessary, RESET the Mechanical overspeed device per section 7.3.4, Mechanical Overspeed Trip Reset.
A. Incorrect, see description above. 1st part is incorrect (not at 50% output). Plausible since indicator is typically full upscale when the controller is in a standby alignment or for a restart without an initiation signal present. 2nd part is not correct (controller is on 2HII-P602).
Plausible because the procedure states "Unless indicated otherwise, the following steps will be performed at panels 2H11-P601 and 2H11-P602" B. Incorrect, see description above. 1st part is correct. 2nd part is incorrect (not at 50%
output). Plausible since indicator is typically full upscale when the controller is in a standby alignment or for a restart without an initiation signal present.
C. Incorrect, see description above. 1st part is incorrect (not at 50% output). Plausible since indicator is typically full upscale when the controller is in a standby alignment or for a restart without an initiation signal present. 2nd part is correct (controller is on 2HII-P602). Plausible because the procedure states "Unless indicated otherwise, the following steps will be performed at panels 2H11-P601 and 2H11-P602" D. Correct, see description above.
2.1 Conduct of Operations Thursday, January 15, 200910:28:58 AM 143
HL T 4 NRC Exam
- 68. G2.1.39 OO1l3111CONSERV ATIVE DECINEWIHIGHERlHT2009-30 llRO/EUICMEI Unit 1 is operating at 60% power.
o The Shift Supervisor (SS) and Shift Technical Advisor (STA) are currently on tour outside the Main Control Room o All Oscillation Power Range Monitors (OPRM) are INOPERABLE A Reactor Recirculation transient results in the plant operating in the Region Of Potential Instabilities.
o The Average Power Range Monitors (APRMs) are oscillating at 6%
o The oscillations are slowly increasing As the Control Board Operator (CBO) in the Main Control Room, your first action lAW 34AB-C51-001-1, "Reactor Operations With fuoperable OPRM System" will be to ___ _
A't scram the reactor B. begin continuous monitoring of the APRMs C. insert control rods after receiving STA recommendation D. raise Reactor Recirculation flow after receiving SS permission Thursday, January 15, 2009 10:28:58 AM 144
HL T 4 NRC Exam Description; When oscillations occur, with all OPRMs inoperable, the procedural requirement to scram lAW 34AB-C51-00 1-1 is:
4.1 IF at ANY time while operating in the Region of Potential Instabilities (RPI), ANY of the following conditions exist, enter 34AB-C71-001-1, Scram Procedure, AND SCRAM the Reactor:
4.1.1 Any APRM is observed to have bandwidth oscillations ~ 5% peak to peak oscillations AND INCREASING.
Not taking action because the candidate is waiting on SS or STA permission/recommendation is an example of a decision that is not conservative (SER 23-93).
A. Correct. See description above B. Incorrect. See description above. Plausible because the procedure provides additional guidance when a condition requiring a scram is not present; Step 4.4 IF the APRM OR LPRM flux noise increases noticeably over current OR normal values OR IF LPRM alarms are received, begin continuous monitoring of the APRMs AND LPRMs (as practical), concentrating particularly in areas where alarms are received.
C. Incorrect. See description above. Plausible because the procedure provides additional guidance when a condition requiring a scram is not present, step 4.6.2 Insert ** groups of rods to their RWM insert limit using the continuous in switch per the STA's recommendation to reduce power below the 61 % load line on appropriate Attachment 1 OR 2 Power/Flow Map.
D. Incorrect. See description above. Plausible because the procedure provides additional guidance when a condition requiring a scram is not present, 4.6.1 Increase Core Flow to greater than 50% (54% if using Attachment 2).IF possible.
2.1 Conduct of Operations 2.1.39 Knowledge of conservative decision making practices. (CFR: 41.10 143.5/45.12)
IMPORTANCE RO 3.61 SRO 4.3 Reference(s) used to develop this question:
SER 23-93, "Delayed manual scram following a core flow reduction event and entry into a region of core instability" 34AB-C51-001-1, "Reactor Operations With Inoperable OPRM System" LT-10023 IE Conservative Decision Making LO LT-10023.001 Reference(s) provided to the student:
None Thursday, January 15, 200910:29:00 AM 145
HL T 4 NRC Exam
- 69. G2.2.1S 001l3/2NENT & DRAlNINEWIFUNDIHT2009-301IRO/EUICMEI The operating team is performing a tagout of the Unit 2 Core Spray system.
o 2E21-F019A, "Torus Suction Valve" is closed o 10 seconds later the 2E21-F019A's solenoid valve fuse blows (refer to drawing)
I.
I I
A/S I
N.D.
~
I_REF.19 I
. SHT.2
. T23-)(20BA.1 P601' I KEY LOCK AO
. I ~r~--o~
F019 A L -c=r~?::(C Which ONE of the following completes both of these statements?
. 2E21-F019A will (1)
Air will be (2) the 2E21-F019A actuator.
A"! (1) stroke open (2) vented off of B. (1) stroke open (2) supplied to C. (1) remain closed (2) vented off of D. (1) remain closed (2) supplied to Thursday. January 15, 200910:29:01 AM 146
HL T 4 NRC Exam Description; 2E21-FO 19A fails open on a loss of air pressure. When the solenoid is de-energized, air is vented off of the actuator ofFOI9A, which results in the valve stroking open.
The candidate must use the drawing provided to determine the impact of the fuse blowing on the operation of2E21-F019A (i.e. is the solenoid energized when the valve is ope~ or closed, does the solenoid vent air off the actuator when the solenoid is energized or de-energized))
A. Correct; see description above.
B. Incorrect; see description above. 1st part is correct. 2nd part is not correct (air is vented off the actuator). Plausible because the candidate must use print reading skills to determine the impact and configuration of the current lineup for air and position ofF019.
C. Incorrect; see description above. 1 st part is NOT correct. Plausible because the candidate must use print reading skills to determine the impact and configuration of the current lineup for air and position of FO 19. 2nd part is correct D. Incorrect; see description above. 1 st part is NOT correct. 2nd part is NOT correct. Both are plausible because the candidate must use print reading skills to determine the impact and configuration of the current lineup for air and position ofF019.
2.2 Equipment Control 2.2.15 Ability to determine the expected plant configuration using design and configuration control documentation, such as drawings, line-ups, tag-outs, etc. (CFR: 41.10/43.3/45.13)
IMPORTANCE RO 3.9/ SRO 4.3 Reference(sl used to develop this question:
H26018, "Core Spray System P&ID" H27661, "Core Spray System Elementary Diagram, sh 4 of 6" SO-LP-OOOll, "Valves and Valve Operators" LO SO-00011.035 Reference(sl provided to the student:
None Thursday, January 15, 200910:29:02 AM 147
HL T 4 NRC Exam
- 70. 02.2.25 001l3/2IMCPRJNEW/FUNDIHT2009-30 llROIEUICMEI Unit 1 is operating at 75% power with a Minimum Critical Power Ratio of 1.03.
lfthe Main Turbine were to trip AND the Main Turbine Bypass Valves fail to open, the risk of fuel damage is significantly increased due to which ONE ofthe following:
At Transition Boiling B. Stored Decay Heat C. Zirc-Water Reaction D. Pellet Clad Interaction This question tests the bases for the MCPR safety limit by use of distracters composed from the bases/failure mechanism from other thermal limits. Candidates often confuse the mechanisms associated with safety/thermal limits.
A. Correct B. Incorrect; this is the failure mechanism due to APLHGR following a LOCA. Nothing stated in the stem indicates that APLHGR is being violated. Plausible si~ce it is one of the safety limit failure mechanisms C. Incorrect, this one of the effects from elevated clad temperatures (APLHGR). Zirc Water reactor occurs when clad temperatures reach 2200°F, which is the temperature associated with APLHGR limit violation. Nothing stated in the stem indicates that APLHGR is being violated.
Plausible since it is a condition associated with one of the safety limit violations.
D. Incorrect; this is the failure mechanism for LHGR violation and nothing stated in the stem indicates that LHGR is being violated.
2.2 Equipment Control 2.2.25 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits. (CFR: 41.5/41.7 /43.2) IMPORTANCE RO 3.2/ SRO 4.2 Reference(s) used to develop this question:
LT-LP-00305, "Technical Specifications" LO 300.003.A.04 TS 2.0 Safety Limits TS Bases for safety limits B2.1.1 Reference(s) provided to the student:
None Thursday, January 15, 200910:29:03 AM 148
HL T 4 NRC Exam
- 71. 02.2.39 001l3/2/TSINEWIHIGHERJHT2009-301IROIEUICMEI Unit 1 is at 100% power, late in core life.
o The "lB" Control Rod Drive (CRD) pump is tagged out o All control rods are withdrawn The following sequence of events occur:
o 10:00 the" 1A" CRD pump trips o 10:05 an operator attempts to restart the "lA" CRD pump and is NOT successful o 10: 10 the Shift Supervisor declares two CRD scram accumulators inoperable lAW with Tech Specs 3.1.5, which ONE ofthe following is the EARLIEST time that the Reactor Mode Switch MUST be in the Shutdown position?
A. 10:06 B. 10:11 C"!' 10:31 D. 11:01 Description; When 2 or more accumulators on withdrawn control rods are inoperable with
. steam dome pressure greater than 900 psig and charging water header pressure cannot be
. restored to >940 psig within 20 minutes (TS 3.1,5) the reactor mode switch must be placed in the shutdown position immediately. The 20 minute clock for this question starts at the time when 2 accumulators are declared inoperable.
Additional plausibility and difficulty for this question comes from the fact that the candidate must answer TS RAS's from memory.
A. Incorrect; see description above. Plausible if the candidate determines that an immediate scram is required when it becomes obvious that charging water header pressure cannot be restored (cant restart the CRD pump).
B. Incorrect; see description above. Plausible because this would be a correct answer if reactor pressure were <900 psig.
C. Correct; see description above.
D. Incorrect; see description above. Plausible because the control rod is required to be declared inoperable within one hour with charging water pressure is >940 psig and the accumulators inoperable.
Thursday, January 15, 2009 10:29:05 AM 149
HL T 4 NRC Exam 2.2 Equipment Control 2.2.39 Knowledge of less than or equal to one hour Technical Specification action statements for systems. (CFR: 41.7/41.10 /43.2/45.13) IMPORTANCE RO 3.9/ SRO 4.5 Reference(sl used to develop this question:
Unit Two Tech Spec (TS 3.1.5)
Cll-CRD-LP-OOlOl, "CRD System" lesson plan EO 200.045.A.02 Reference(sl provided to the student:
None Thursday, January 15, 2009 10:29:05 AM 150
HL T 4 NRC Exam
- 72. G2.3.14 00l/3/3/EOP 103INEW/FUNDIHT2009-301IRO/EU/CME/
Unit 1 is at 35% rated o The Hydrogen Injection System is placed in service in AUTOMATIC - EXTERNAL mode o Power is raised from 35% power to 100% power o At 100% power hydrogen flow rate indicates 40 SCFM Which ONE of the following answers both of these statements?
IA W 34S0-P73-001-1, "Hydrogen and Oxygen injection and Control for HWC", hydrogen injection flow rate is (1) the normal 100% power flow rate.
Radiation levels in the Condenser Bay will stabilize (2) expected normal full power radiation levels.
A. (1) above (2) at B. (1) below (2) at C. (1) below (2) below D~ (1) above*
(2) above Thursday, January 15, 2009 10:29:06 AM 151
HL T 4 NRC Exam Description; In the Automatic - External Mode the hydrogen injection system is load following. It is normally placed in service at just above 30% Rx power. As Rx power is increased, the hydrogen flow rate is increased to the maximum amount that the controller is set for. The amount of hydrogen has varied over time from as low as 8 SCFM to as high as 45 SCFM. The plant is currently operating at 10 SCFM, but this amount is determined by the Chemistry Dept. and is not specifically listed in a procedure. There is a system shutdown on hydrogen flow of25 SCFM at this time. There was an event at Plant Hatch: Hydrogen flow was manually lowered in Internal mode, the controller was then placed in External mode and hydrogen gas flow ramped back up to the original setpoint. This resulted in a radiation exposure to several individuals that was above the expected dose for the task. As hydrogen injection rate is increased, radiation levels increase around any area where Main Steam is piped.
A. Incorrect, radiation levels are expected to increase to above normal levels..
Plausible because flow is higher than normal, but the candidate must understand the operation of the Automatic-External mode and the effects ofH2 injection on Radiation levels.
B. Incorrect, H2 flow rate is NOT below normal.
Plausible, if candidate does not understand the operations of the Automatic-External mode and the effects of H2 inj ection on Radiation levels or if candidate incorrectly recalls that normal flow is above 40 SCFM.
C. Incorrect, H2 flow rate is NOT below normal.
Plausible, if candidate does not understand the operations of the Automatic-External mode and the effects ofH2 injection on Radiation levels or if candidate incorrectly recalls that normal flow is above 40 SCFM.
D. Correct, see above.
2.3 Radiation Control 2.3.14 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities. (CFR: 4l.12 /43.4/45.10)
IMPORTANCE RO 3.4 / SRO 3.8 Reference(sl used to develop this question:
34S0-P73-001-1, "Hydrogen and Oxygen injection and Control for HWC" P73-HWCI-LP-07301, "Reactor Water Chemical Injection System", EO 026.034.A.05 and 026.034.A.04 Reference(sl provided to the student:
None Thursday, January 15,2009 10:29:07 AM 152
HL T 4 NRC Exam
- 73. G2.3.7 OOl/3/3IRWPINEWIFUNDIHT2009-301IROIEUICMEI Which ONE of the following completes this statement?
lAW 60AC-HPX-004-0, "Radiation and Contamination Control", while signed onto the Operations GENERAL Radioactive Work Permit (RWP 09-0004), you are allowed to enter an area where WITHOUT being required to sign on to a SPECIFIC RWP.
A. Very High Radiation levels exist B. a highly contaminated system being breached C!' airborne radioactivity concentration is 0.2 DAC D. loose surface contamination level = 300,000 dpm/lOOcm2 Thursday, January 15, 200910:29:08 AM 153
HL T 4 NRC Exam 1 ________________________________________________________ __
Description; lAW 60AC-HPX -004-0:
8.2.4.1 General RWPs are used for inspection, plant tours and general area walkdowns in areas other than Highly Contaminated, High Radiation, Very High Radiation or Airborne Radioactivity Areas.
Step 8.2.5 from 60AC-HPX-004, Specific RWPs are used for controlling work that is not controlled by a General RWP and occurs in a High Radiation Area, High Contamination Area or areas with changing radiological conditions. The Specific RWP will be modified on the basis of surveys. Radiological surveys SHALL be performed as required to support RWP authorized work.
8.2.5.3 A Specific RWP will be considered if any of the following conditions exist:
o General area radiation levels >= 100 mRem/hr o Airborne radioactivity concentration >30% of Derived Air Concentrations (DACs) specified in 10CFR20.
o Loose surface contamination levels >200,000 DPM/100 cm2 o Breach of a highly contaminated system, OR in a High or Very High Radiation Area A. Incorrect, see description above. Plausible because the candidate must recall the requirement for General RWPs vs. Specific RWPs from memory to eliminate this distractor.
Very High Rad areas require a Specific R WP.
B. Incorrect, see description above. Plausible because the candidate must recall the requirement for General RWPs vs. Specific RWPs from memory to eliminate this distractor.
Areas where highly contaminated systems are being breached require a Specific RWP.
C. Correct, see description above.
D. Incorrect, see description above. Plausible because the candidate must recall the requirement for General RWPs vs. Specific RWPs from memory to eliminate this distractor. Areas with loose surface contamination level >200,000 DPM/100cm2 require a Specific RWP.
Thursday, January 15,200910:29:09 AM 154
HLT 4 NRC Exam 2.3 Radiation Control 2.3.7 Ability to comply with radiation work permit requirements during normal or abnormal conditions. (CFR: 41.12/45.10)
IMPORTANCE RO 3.5 / SRO 3.6 Reference(sl used to develop this question:
60AC-HPX-004-0, "Radiation and Contamination Control" LT-LP-30008, "Rad Control Admin Procedure" LO LT-30008.003 Reference(sl provided to the student:
None Thursday, January 15,2009 10:29:09 AM 155
HL T 4 NRC Exam
- 74. G2.4.1 001/3/4/EOP ENTRYINEWIHIGHERlHT2009-301IRO/EUICMEI Unit 1 is operating at 70% power when a transient occurs.
Current plant conditions:
o Reactor water level............ +5 inches (lowest level reached) o Reactor pressure................ 1080 psig (highest pressure reached) o Drywell pressure................ 1.90 psig (highest pressure reached)
Which ONE of the following completes both of these statements?
Entry conditions have been met or exceeded _--->("""1.1-) __ Emergency Operating Procedure (EOP) flow chart(s).
lAW 34AB-C71-001-2, "Scram Procedure", performance of the RC-l, RC-2 and
. (2) placards are required IMMEDIATE actions.
A. (1 ) ONLY for the Reactor Controls (RC)
(2) TC-l B. (1) ONLY for the Primary Containment (PC)
(2) RC-3 C'!' (1) for BOTH the RC and PC (2) RC-3 D. (1) for BOTH the RC and PC (2) TC-l Thursday, January 15, 2009 10:29:10 AM 156
HL T 4 NRC Exam Description; EOP entry conditions are as follows:
RC EOP flowcharts are:
o RWL<+3" o RPV pressure> 1074 psig o DW pressure> 1.85 psig o A conditions that requires reactor scram and reactor power >5%
o A conditions that requires reactor scram and reactor power is unknown PCC EOP flowchart:
o Torus water level> 150" o Torus water level <146" o DW pressure> 1.85 psig o Torus water temp> 1 OOF o DW temp> 150F o PC H2 concentration> 1.5%
Upon entry of Scram Procedure, 34AB-C71-00l, the RC-1, RC-2 and RC-3 placards are performed. TC-1 is directed to be performed as the first subsequent action step ofthe scram procedure.
A. Incorrect, see description above. 1 st part is partially correct (entry into both RC-1 and PCC is required). 2nd part is not correct because TC-1 is not directed from an immediate action step. Plausible because the candidate must recall from memory the EOP entry condition for the EOP flowcharts in addition to recalling the Immediate Action steps from the scram procedure.
B. Incorrect, see description above. 1 st part is partially correct (entry into both RC-1 and PCC is required). Plausible because the candidate must recall from memory the EOP entry condition for the EOP flowcharts in addition to recalling the Immediate Action steps from the scram procedure. 2nd part is correct.
C. Correct, see description above.
D. Incorrect, see description above. 1st part is correct. 2nd part is not correct because TC-1 is not directed from an immediate action step. Plausible because the cantdidate must recall from memory the EOP entry condition for the EOP flowcharts in addiion to recalling the Immediate Action steps from the scram procedure.
Thursday, January 15, 200910:29:11 AM 157
HL T 4 NRC Exam 2.4 Emergency Procedures / Plan 2.4.1 Knowledge of EOP entry conditions and immediate action steps.
(CFR: 41.10/43.5/45.13) IMPORTANCE RO 4.6 / SRO 4.8 Reference(sl used to develop this question:
31EO-EOP-012-1 31EO-EOP-01O-l 34AB-C71-00 1-1 Reference(sl provided to the student:
None Thursday, January 15, 200910:29:12 AM 158
HL T 4 NRC Exam
- 75. G2.4.9 001l3/4/LOW POWER INJECTIONINEWIHIGHERlHT2009-301IRO/CME/EUI Unit 2 is shutdown with aU rods fully inserted.
o Reactor pressure................... 35 psig o Main Condenser Vacuum.... "0" in. Hg. Vac o The "2B" loop of Residual Heat Removal (RHR) is operating in Shutdown Cooling (SDC) o The "2A" loop ofRHR is tagged out A Logic System Functional Test results in an inadvertent DryweU pressure Loss of Coolant Accident (LOCA) signal.
Which ONE of the following predicts the status ofRHR SDC system and how reactor pressure will be controlled?
Shutdown Cooling (1)
IAW 34S0-Ell-OI0-001-2, "RHR System" and/or 34AB-Ell-001-2, "Loss ofSDC",
reactor pressure will be controlled by (2)
A. (1) remains in service (2) throttling 2EII-F048B, "RHR Heat Exchanger Bypass Valve" B. (1) remains in service (2) throttling 2EII-F003B, "RHR Heat Exchanger Outlet valve" C. (1) has been lost (2) opening the MSIV s and using the Main Turbine Bypass Valves to send steam to the Main Condenser, defeating the low vacuum trips as necessary.
D!' (1) has been lost (2) raising RWL to the Main Steam Lines (MSL) and opening the MSL drain valves, defeating the low vacuum trips as necessary.
Thursday, January 15,200910:29:13 AM 159
HL T 4 NRC Exam Description; With SDC in service, when a high DW pressure signal is received SDC is lost because the RHR injection valve (2EII-FOI5B) will auto close; however the RHR pump remains running. Low water level (+3 inches) or high reactor pressure (138 psig) will also take RHR out of service, the difference in this case is that the F008 and F009 valves close, which causes an RHR pump trip.
With RHR in service, cooldown is controlled, per 34S0-EII-01O-2, by:
7.2.3.93 WHILE maintaining RHR Loop A(B) flow rate of 6200 to 8200 GPM, establish an increase OR decrease in the cooling rate as required to obtain/maintain the desired coolant AND/OR RPV temperature(s) as specified by Shift Supervisor, by manipulating one OR a combination of the following MOVs:
7.2.3.93.1 2E11-F003A(B), Hx Outlet VI v, (OPEN to increase cooling; CLOSE to decrease cooling).
7.2.3.93.2 2E11-F048A(B), Hx Bypass Vlv, (CLOSE to increase cooling; OPEN to decrease cooling).
7.2.3.93.3 2E11-F068A(B), Hx A(B) RHRSW Disch VI v, (OPEN to increase cooling; CLOSE to decrease cooling).
Note: Steam pressure of 35 psig is not adequate to provide steam seals for establishing Main Condenser vacuum.
A. Incorrect, see description above. 1 st part is not correct. Plausible since the RHR pump continues to run in this case. 2nd part is not correct (2EII-FOI5 closing will stop flow to the RPV). Plausible since this is a normal method of controlling cooldown (see above). If the candidate considers the loop to be in service, this will look to be a correct answer.
B. Incorrect, see description above. 1 st part is not correct. Plausible since the RHR pump continues to run in this case. 2nd part is not correct (2E II-FO 15 closing will stop flow to the RPV). Plausible since this is a normal method of controlling cooldown (see above). If the candidate considers the loop to be in service, this will look to be a correct answer.
C. Incorrect, see description above. 1 st part is correct. 2nd part is not correct (Bypass valves close at 7" Hg-Vac.) Plausible because using bypass to send steam to the main condenser is an alternate pressure control path. Additionally, the auto closure of the MSIVs can be overridden, per procedure, and this provides a distractor in that the candidate may confuse which low vacuum isolation is allowed to be overridden.
D. Correct, see description above. Steam seals cannot be formed on the Main Condenser so the next method in 34AB-EII-00l is to raise level and establish flow through the MSL drain valves.
Thursday, January 15,200910:29:14 AM 160
HL T 4 NRC Exam 2.4 Emergency Procedures / Plan 2.4.9 Knowledge oflow power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.
(CFR: 41.10/43.5/45.13) IMPORTANCE RO 3.8 / SRO 4.2 Reference(s) used to develop this question:
34AB-E11-001-2 34AB-E11-010-2 Reference(s) provided to the student:
None Thursday, January 15,200910:29:15 AM 161