ML032650867
| ML032650867 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 06/13/2003 |
| From: | Reid J Public Service Enterprise Group |
| To: | Conte R NRC/RGN-I/DRS/OSB |
| Conte R | |
| References | |
| 50-354/03-302 50-354/03-302 | |
| Download: ML032650867 (181) | |
Text
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Using the attached power to flow map, answer the following:
- The plant is operating at the 81% rodline.
v Select the MINIMUM core flow at which the plant can operate at power and still be assured of avoiding power oscillations or instabilities. .~
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I 40% - 1 45% I 48% It boundry of the EXIT or Instability Region
. 48% -Incorrect- see TS 3.4.1 .l.b and HC.OP-AB.RPV-0002
. 25% -Incorrect- see TS 3.4.1.1.b and HC.OP-AB.RPV-0002
. 40% - Incorrect- see TS 3.4.1.1.b and HC.OP-AB.RPV-0002 I
nce of all associated lines, k W theRecirculation Flow 1 Control System Lesson Plan.
I 1 HC.OP-AB.RPV-0002 Power to flow chart with region labeling__and-_line-IWlaterial Required for Examination nomenclature removed.
IQuestion ModificationMethod: Editorially Modified
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Saturday, M& IO,2003 12 08.38PM Page 1 of 161
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Given the following conditions:
- - A plant startup is in progress.
- Reactor power is 10.5 percent.
- Offgas recombiner train 0 just tripped.
- Offgas Recombiner train 1 is NOT available.
Which one of the following decribes the action required to allow use of the Mechanical Vacuum Pumps (MVP) and the bases for that power level? ___
kd Reduce-reactorpower by 5 percent; Combustible gas concentrations may cause an explosion in the SJAE Aftercondenser.
I Reduce reactorpower by 5 percent; Offsite radiological release may be above allowable limits at the South Plant Vent.
Reduce reactor power by 6 percent; Offsite radiological release may be above allowable limits I at the North Plant Vent. I Reduce reactor power by 6 percent; Combustible gas concentrations may cause an explosion 1 at the MVP suction. I L
of facility conditions and procedure limitations.
Reduce reactor power by 6 percent; Combustible gas concentrations may cause an explosion at the MVP suction. Correct. Subsequent actions of HC.OP-AB.BOP-0006 direct power reduction to less than 5 percent. A 6 percent reduction for the given conditions will result in 5 percent power. The explosion concern is the MVP inlet piping and pump.
Reduce reactor power by 5 percent; Combustible gas concentrations may cause an explosion in the SJAE Aftercondenser. Incorrect. 5 percent is the power limit. A 5 percent reduction will still result in >5 percent power. The explosion concern is the MVP inlet piping and pump.
Reduce reactor power by 5 percent; Offsite radiological release may be above allowable limits at the South Plant Vent. Incorrect. Wrong power level.
Reduce reactor power by 6 percent; Offsite radiological release may be above allowable limits at the North Plant Vent. Incorrect. Wrong bases. Wrong release path.
/HC.OP-SO.CG-0001 ~-
INOHOI
- - AIRREM i Learning Objectives
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ABBOPGE007 (R) Explain the bases fo t Actions and the information contain
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Saturday, May 10, 2003 12.08 38 PM-- Page 2 of 161
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]Material Required for Examination ] None - -
Question Modification Method:
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Saturday, May 10,
.-2003 12 08:38 PM
Given the following conditions:
- The plant is operating at I 0 0 percent power.
u - TACS is on the 'A' Loop of SACS.
- 'A' 1E 4.16 KV bus 1OA401 has de-energized due to bus fault.
Which one of the following describes a result of the bus fault and the__reason - for the result?
All RACS Pumps trip due to LO-LO Head tank level.
I'B' SACS Expansion Tank overflows due to power loss to TACS return valves. I RACS Head Tank overflows due to makeup valve power loss.
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'A& IC' SACS Pump trip due to LO-LO-LO Expansion Tank level. I I
c B Hope Creek AKI. 1 Knowledge of the operational implications of the followingconcepts as they apply to PARTIAL OR 1 I
COMPLETE LOSS OF A.C. POWER: 1 AK1.05 [Failsafe component design ~ 2.6 2.71
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'B'SACS Expansion Tank overflows due to power loss to TACS return valves. Incorrect. SACS ET Makeup uses an MOV and inverter backed instrumentation. Sluicing causes "A' tank to overflow. B Tank lowers.
All RACS Pump trip due to LO-LO Head tank level. Incorrect. Head Tank Level goes high. Pumps trip on '
bus restoration.
A & C SACS Pump trip due to LO-LO-LO Expansion Tank level. A&C Pumps trip but for the wrong reasons. A trips due to loss of power; C trips on low pump differential pressure.
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~ HC.0 P-AB.=-0 170 i
1HC.OF-G P .PB-000I 7 11 Channel, Abnormal Operating Procedure.
OABl70E001 Recoqnize abnormal indicationslalarms andlor procedural requirements for implementing, LOSS of 4.16 Kv BUS10A401 A I
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Saturday, i P a i 4 o f 161
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Given the following conditions:
- The plant is operating at 100% power.
- A LOP signal is generated due to a Loss of Off-site Power.
- Just prior to the EDG output breakers closing, a LOCA signal is generated due to a Loss of Coolant Accident.
Which of the following is the response of the LOP and LOCA sequencers for these conditions?
As soon as power is restored to the buses, the LOCA sequencer will control the restoration of all loads.
The LOCA sequencer will begin to sequence until the diesel generator output breakers close, then the LOP sequencer will complete load restoration. - .
As soon as power is restored the buses, the LOP sequencer will control the restoration of all loads.
The LOP sequencer will begin to sequence until the diesel generator output breakers close, then the LOCA sequencer will complete load restoration.
following: 2 L d
,AK2.04 1A.C. electrical loads p p
___________~_ ~ _ _ - I I 3.4 a-s soon as power is restored to the buses, the LOCA sequencer will control the restoration i of all loads. Both the LOP & LOCA sequencers start when power is available. The LOP sequencer will control until the LOCA signal is received. With a LOCA & a LOP signal present, the LOCA Sequencer 1 has priority. When the LOCA signal is received, the LOCA sequencer will control load sequencing.
INCORRECT - The LOCA sequencer will begin to sequence until the diesel generator output breakers close, then the LOP sequencer will complete load restoration. Both the LOP & LOCA sequencers start when power is available. With a LOCA & a LOP signal present, the LOCA Sequencer has priority.
INCORRECT - As soon as power is restored the buses, the LOP sequencer will control the restoration of all loads. With a LOCA & a LOP signal present, the LOCA Sequencer has priority.
.INCORRECT- The LOP sequencer will begin to sequence until the diesel generator output breakers I close, then the LOCA sequencer will complete load restoration. Both the LOP & LOCA sequencers start when Dower is available. With a LOCA & a LOP signal present, the LOCA Seauencer has Drioritv.
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Vision Bank QID#Q61290
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Saturday, May IO,2003 12:08.39PM Page 6 of. 161
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Plant conditions are as follows:
- Reactor Power is 20 %
- Control rod 30-31 is selected at position 12.
Which one of the following describes the response of RMCS if the Main Turbine were to trip?
Reactor Manual Control will: I block all
. _ _ control rod movement because the reactor has scrammed.
I allow control rod insertion using the Continuous Insert PI3 only.
i automaticallv bvbass RWM blocks due to the effects of colder feedwater.
actively enforce control rod blocks due to loss of First Stage Turbine pressure. I the upper limit of RWM Low Power Set Point or the power level that RWM enforces rod blocks. Above 20 percent, the blocks are bypassed and are indicated only. A turbine trip will cause reactor power to increase due to the positive reactivity effects of the loss of feedwater heating followng the turbine trip.
L allow control rod insertion using the Continuous insert Pb only. Incorrect. Continous Insert bypasses the Activity Control timer card, not the Rod Blocks.
block all control rod movement because reactor has scrammed. Incorrect. Reactor will not automatically scram from turbine trip at 20 percent power as a result of a turbine trip.
actively enforce control rod blocks due to loss of First Stage Turbine pressure. Incorrect First stage I
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None
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Saturday, May IO,2003 12:08:39PM
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- The plant is operating at 87% power
- It is near the end of a fuel cycle.
- Main Turbine Stop Valves (TSVs) are being tested to validate the EOC-RPT setpoints.
- Two TSVs initiate an EOC-RPT signal at 10% closed.
- Two TSVs initiate an EOC-RPT signal at 5% closed.
Which of the following is a safety implication (if any) of this condition?
There are no safety implications because the TSV EOC-RPT trip is-a 1-out-of2 logic.
There will be an excessive thermal margin upon EOC-RPT actuation if these TSVs close at power.
Reactor safetv has been enhanced by the overly conservative trip value for TSV closure.
Void reactivity feedback may exceed control rod reactivity __ if these TSVs close -at_ power. _
L-feedback due to a pressurization transient can add positive reactivity to the reactor at a faster rate than the control rods add negative scram reactivity.
Void reactivity feedback may exceed control rod reactivity if these TSVs close at power-Correct- If the TSVs generate an EOC-RPT signal at 10% closed vice the nominal f 5% closed (f 7% allowable) then an excess positive reactivity will be added upon TSV closure.
There will be an excessive thermal margin upon EOC-RPT actuation if these TSVs close at power-Incorrect- If the TSVs generate an EOC-RPT signal at 10% closed vice the nominal 5% closed (- 7% -
allowable) then an excess positive reactivity will be added upon TSV closure.
Reactor safety has been enhanced by the overly conservative trip value for TSV closure-lncorrect-setpoint is 5% +2% not 10%
There are no safety implications because the TSV EOC-RPT trip is a l-out-of-2 logic-lncorrect- The TSV closure uses various combinations, not 1 of 2 twice. I I
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l
- a. Select those sections applicable to the Main Turbine.
Evaluate Main Turbine operability and determine required action@)based upon inoperability.
es for those technical specification items associated with the Main Turbine (SROONLY)..___--J
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Material Required for Examination 1 Technical Specification 3.3.4.2 and Table 3.3.4.2-2 Page 8 of 161
__ Saturday, May I O , 2003 12 08.39
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- Saturday,
__ May 10, 2003 12.08.40
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Given the following conditions:
- - Preparations for a reactor startup from a refueling outage are in progress.
- Reactor Building ambient temperature is 74 degrees F.
- The Reactor Building Equipment Operator is charging the hydraulic control unit accumulators with nitrogen to a pressure of 590 psig
- Several days later with the Unit at 100% power, Reactor Building temperatures have stabilized at 92 degrees F Which of the following describes the impact on the Control Rod Drive Hydraulic system operations for these conditions?
(Assume NO leakage)
(Refer to attached figure.)
The individual control rod:
normal insertion seeeds will be slower and may result in control rod drift alarms.
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..I scram speeds will be slower and will result in reduced reactivity addition rates. 1 I
normal insertion speeds will be faster and may result in "double notching". 1 u
scram speeds will be faster and may result in mechanism damage. Correct. HC.OP-SO.BF-0002 Precaution 3.1.4. Interpretation of the graph places the N2 pressure too high. Warming up of the Reactor Building will cause the pressure to rise parallel to the desired precharge line and remaining too high.
normal insertion speeds will be slower and may result in control rod drift alarms. Incorrect. Drive speeds will be unchanged. Rod drifts will not occur.
scram speeds will be slower and will result in reduced reactivity addition rates. Scram speeds will be higher.
normal insertion speeds will be faster and may result in "double notching". Incorrect. Drive speeds will be unchanaed
[HC.OP-SO. BF-0002 J
-- --- - 7 CRDHYDE021 (R) Given the "Accumulator Precharge Nitrogen Pressure Versus Ambient Temperature Curve", determine the proper I accumulator precharge gas pressure, IAW HC.OP-SO.BF-0002. j 1
,-, IIWabriai Required for Examination I Accumulator charging pressure graph from HC.6P-SO.BF-0062 1 (Question Modification Method: 1 EditonallyModified
[Question Source: INPFExam Bank 1 Saturday, May 10, 2003 12:08:40 PM L._-Page 10 of 161
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IQuestionSource Comments: 1 lnpo
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Peach Bottom modified for Hopexreek
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SaturdayrMay 10, 2003 12:08:40 PM
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Given the following conditions:
- The plant is operating normally at 100 percent power.
i/
- An EHC failure causes reactor pressure to rise 10 psig in 10 seconds.
Which one of the following describes reactor power response?
(Assume NO operator action) .- I Rises initially due to lower void fraction, then lowers rapidly due to scram. 1 I
Rises initially due to lower void fraction, then stabilizes at a slightly higher power level. I I
Lowers initially due to greater feedwater heating, then-lowers rapidly due to scram.
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Lowers- initially due to greater feedwater heating, then stabilizes at a slightly higher power level.]
y due to lower void fraction, then stabilizes at a slightly higher power level. Correct. Rising pressure causes reacor power to rise. 10 psig rise above normal 1005 psig will not reach scram setpoint of 1037 or manual scram threshold of >IO30 psig of retainment override of AB-RPV-0005 Rises initially due to lower void fraction, then lowers rapidly due to scram. Incorrect. Would not reach scram threshold.
Lowers initially due to greater feedwater heating, then lowers rapidly due to scram. Incorrect. Rises initially.
,Lowers initially due to greater feedwater heating, then stabilizes at a slightly higher power level.
Incorrect. Rises initially.
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Saturday,
. May I O , 2003 1208.40 PM - Page 12 of 161 L1 __
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Given the following conditions:
- The plant is in Operational Condition 3 with RHR A in Shutdown Cooling operation.
-.- - Reactor coolant temperature and pressure is slowly rising.
- The Shutdown Cooling Inboard and Outboard Isolation valves, HV-FOO9 and F008, have now closed, and the operating RHR pump has tripped.
The reason for these automatic actions is to prevent: -
- steamvoiding in the RHR pump seals.
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overpressurizingthe RHR pump seals. 1 establishing a drain path from the RPV to the torus.
1 I
overpressurizingthe shutdown cooling suction piping. Correct. The piping isolates to minimze offsite b
radiological release to the environment if the low pressure piping fails.
overpressurizing the RHR pump seals. Incorrect. Wrong reason. Overpressurize piping establishing a drain path from the RPV to the torus. Incorrect. Reason for isolation on low RPV level.
Wrong reason..
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steam voidina in the RHR Dump seals. Incorrect. Reason for seal coolers on RHR pumps. j I
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I Saturday,-~
May
~ _ 10,
_ 2003 12.08.40 -
PM-~ L--_-Page 13 of 161
Given the following conditions:
- All 3 RFPT's are in manual speed control.
- RPV level is 35 inches.
Feedwater pump flows are as follows:
- 'A' 3.2 Mlbm/hr
- '6' 3.6 Mlbm/hr
- 'C'3.5 Mlbm/hr Main Steam flows are as follows:
- 'A' 2.6 Mlbm/hr
- 'B' 2.5 Mlbm/hr
- 'C'2.6 Mlbm/hr
- 'D'2.4 Mlbm/hr Based on these conditions, RFP speed demand must be applied to prevent the RPV Level alarm.
raise, 7 lower. 7
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ad IHope Creek
.-./ [Emergency and Abnormal Plant Evolutions 008 1 High Reactor Water Level I decrease, 7; Correct. Total Feed Flow is 10.3 Mlbm/hr. Total Steam flow is 10.1 Mlbmlhr. this mismatch Page 14 of 161:
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Page 15 of 161
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12:08:40 PM
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Using the attached transient analysis plots of a reactor scram, which one of the following f a i l u r ]
I caused the scram? ____ -2 u
EHCPressure Regulator Failure to 0%. ---I!
Master Level Control Setpoint Failure to 0 inches. __
EHC Pressure Regulator Failure to 130%.
- Master Level Control Setpoint Failure to +60 inches.
IAlsfbcerll b (Exam1S 1 - 1 Comprehension 7 -
Hope Creek 06/17/2003 Maste; Level control fails to 0 inches - correct. RPV level drops with no rpv pressure rise.
Master Level control fails to +60 inches - incorrect. RPV Level lowers EHC Press reg fails to 130 %- incorrect. RPV pressure remains steady until scram on low level.
EHC Press reg fails to 0 %- incorrect. RPV pressure remains steady until scram on low level.
Provide UFSAR fiaure 15.2-8 with title block removed 1
IUFSAR 15.2 . ,I i
I W
E0101LE006 1 (R)
, , Given any step of the procedure, describe the reason for performance of that step andlor expected system response to I I 1 J
I I
IM.f.iial Requiredfoi.~ination Provide FSAR figure 15.2-8 with title block removed.
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Saturday,
____ _ I O , 2003 May _ _ 12 08 40 PM
Given the following conditions:
- The plant is operating normally at 100 percent power.
-, - 'B' Reactor Feedwater pump trips.
- Assume NO operator actions taken.
What is the response of the plant and the reason for that response?
Full Reactor Recirc Runback. Reduce Recirc loop flow to ensure adequate NPSH-to the Recirc Pumps. - -
Intermediate Reactor Recirc Runback. Reduce power to control level.
Full Reactor Recirc Runback. Reduce core flow to ensure adequate NPSH to the Jet- Pumps. - . .-- 1
'Intermediate Reactor Recirc Runback. Reduce power to prevent power oscillations- . -.
-1 Intermediate Reactor Recirc Runback. Reduce power to prevent power oscillations. Incorrect. Wrong reason. Intermediate RB on RPT Trip is to control level.
Full Reactor Recirc Runback. Reduce core flow to ensure adequate NPSH to the Jet Pumps. Incorrect.
Wrong Runback. Wrong Reason. Bases of Total Feedwater flow RB.
Full Reactor Recirc Runback. Reduce Recirc loop flow to ensure adequate NPSH to the Recirc Pumps.
Incorrect. Wrona Runback Wrona reason. Bases for Suction valve position interlock.
Y I (NOH01RECCON-01 i II I
System Lesson Plan:
- a. Recirc M G drive motor
- b. Fluid Coupler
- c. Generator
- d. Scoop Tube Positioning Unit
- e. Exciter
- f. Tachometer
- g. Individual pump speed controllers
- h. Speed limiters #and I #2
- i. Startup signal generator
- j. Error limiting circuit
- k. Scoop tube positioning unit
- 1. Signal Failure Detector
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Saturday,
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May 10, 2003 12:08:41 PM
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Saturday, May 10, 2003 12:08:41 PM
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Given the following conditions:
- - A LOCA has occurred
- All control rods did not fully insert
- Reactor Power is 3%, and lowering
- Reactor Pressure is 1000 psig, controlled by SRV's
- Reactor Water Level is 0 inches, steady
- Drywell Temperature is 350°F, and rising
- Drywell Pressure is 33 psig, and rising
- Suppression Pool Temperature is 120°F, and rising
- Suppression Pool Level is 80 inches, steady
- Suppression Chamber Pressure is 31.5 psig, and rising
- No operator actions have been taken Which one of the following is the appropriate action for the conditions above in accordance with the
- Initiate Emergency Depressurization and-DrywellSprays.
hitiate Emergency Depressurization and Suppression Pool Cooling.
.- .- . - - I Initiate Suppression -Pool _ Cooling and D&eN Sprays. -I u
295012 j High Drywell Temperature
'2.41Emergency Procedures and Plan 2.4.6 (Knowledgesymptom based EOP mitigation strategies. I 13.1 14.01 Justification: SRO 55.43(4)
CORRECT - Initiate Emergency Depressurization and suppression pool cooling. With SP temperature above 95°F and with SRVs controlling RPV pressure, Suppression Pool Cooling is required. At a Drywell pressure of 33 psig and temperature of 350°F, Drywell Spray is precluded and ED is required IAW DWIT-
,4 thru DWIT-6.
INCORRECT - Initiate drywell sprays ONLY. At a Drywell pressure of 33 psig and temperature of 350"F, Drywell Spray is precluded.
INCORRECT - Initiate Emergency Depressurization and drywell sprays. At a Drywell pressure of 33 psig and temperature of 350"F, Drywell Spray is precluded.
INCORRECT- Initiate suppression pool cooling and drywell sprays. At a Drywell pressure of 33 psig and temperature of 350"F,'Drywell Spray is preclude . -
IHC.OP-EO.=-0102, Step DWIT-3 thru DWK-6 & DW/P-6 .- -2 i -J
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EOP flowcharts
/Editorially Modified 06 editorially modi II
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~ May _ 2003 12:08-41PM L
Page 20 of 161
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Given the following conditions:
- The plant is operating at 100 percent power.
b
- Turbine Building Chiller A K I 11 suffers an evaporator tube break.
- All Turbine Building Chilled Water pumps trip on low flow from Freon in the pump casings.
- Attempts to crosstie Chilled Water have failed.
- Drywell temperature and pressure are rising.
Which one of the following actions is required? _ _ _ ~
Manually scram the reactor at I.5 psig Drywell pressure.
Manually scram the reactor at 135 F Drywell temperature. 1 Shutdown Recirc pumps and Drywell Coolers and initiate Drywell Sprays at 1.5 psig Drywell pressure.
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I Shutdown Recirc pumps and Drywell Coolers and initiate Drywell Sprays at 135 F Drywell temperature.
AKI. I Knowledge of the operational implications of the following concepts as they apply to HIGH DRYWELL TEMPERATURE: -
1-3.1 1-34
- ,AK1.02 /Reactor power level control ____ -
Manually scram the reactor at 1.5 psig Drywell pressure. Correct. Retainment override step La. of HC.OP-AB.CONT-0001 Drywell Pressure.
--_?
Manually scram the reactor at 135 F Drywell temperature. HCTS LCO 3.6.1.7 Limit. EOP-102 decision point DW/T-2 which directs SD of RRPs and DW Coolers to spray DW IF under curve DWP-T limit. DW pressure is still too low.
Shutdown Recirc pumps and Drywell Coolers and initiate Drywell Sprays at 1.5 psig Drywell Pressure.lncorrect. DW Pressure too low. Sprays can be inservice at this pressure only if initiated at I
higher pressure then pressure subsequently lowers.
Shutdown Recirc pumps and Drywell Coolers and initiate Drywell Sprays at 135 F Drywell temperature.
Incorrect. Saturation temperature for DWT too low. Step DWT-3 allows spraying the DW between 135 and 340 degF only if DWP falls in SAFE area of curve DWT-P. This corresponds to about 3.2 psig. DWT rising due to a loss of DW cooling, not a LOCA. _____
I I
I E0102PE007
_ - (R) Given any step of the procedure, determine the reason for performance of that step and/or predict expected system response to control manipulations prescribed by that step IAW the Primary Containment Control - Drywell Lesson Plan ,
I
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R) Explain purpose of and the bases for Retainment Overr
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Saturday, May.10, 2003 12:0841 PM
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The following plantconditions exist:
- The reactor is at full power.
v - Torus Cooling is in operation and average temperature is increasing.
- HPCl testing is in progress.
The required action is to immediately stop HPCl testing if torus temperature exceeds (I), or immediately place the mode switch in shutdown if torus temperature exceeds (2) .
(I) 95 F; (2) 110 F ._
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( I ) 105- F;-(2) 110 F (1) 105 F; (2) 120 F
- - -I (I) 110 F; (2) 120 F -. I degF, place the mode switch to shutdown.
(1) 95; (2) 110 Incorrect. 95 is LCO for normal / non-heat adding conditions.
( I ) 105; (2) 120 Incorrect.
(1) 110; (2) 120 Incorrect.
IHCGS TS 3.6.2.1 ___
/HC.OP-IS.BJ-0001 I
__ I TECSPCE008 Given specific plant operating conditions, and a copy of the Hope Creek Generating Station Technical Specifications, determine 1 the following: I Saturday, May 10,2003- 12.08.41
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A startup is in progress with the following conditions:
- Reactor pressure is stable at 170 psig.
u
- Two bypass valves are full open.
- Control rods are being withdrawn.
- lRMs are between 30 and 70 on range 8 Which of the following would occur if all bypass valves were to fail closed?
(Assume NO operator action)
'The reactor would scram due to high flux. 1I The reactor would scram due to high pressure.
1 Reactor power would increase and stabilize due to the change in void fraction.-
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%Reactor power would- decrease.and _ _ stabilize due to the change in void fraction.
I The reactor would scram due to high pressure. Incorrect. HI-HI IRM flux will trp first.
Reactor power would increase and stabilize due to the change in void fraction. Incorrect. Would not L stabilize with TBVs failed closed.
Reactor power would decrease and stabilize due to the change in void fraction. Incorrect. Reactor power I would increase. I JHCOP-AB.RPV-0005 I
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I Bank Editorially Modified I INPO BANK QID #21232 Dres
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Saturday, May I O , 2003.12:08 41 PM Page 24 of 161
Given the following conditions:
- The plant was operating at 100% power when the reactor scrammed.
v The operator observes the following indications:
- Reactor Pressure: 900 psig
- All Scram valves open
- RWM: All Rods In: NO Shutdown: YES Rods Not Full In: 040 I
The reactor is:
in a cold shutdown rod configuration with forty control rods at position 02. 1 in a cold shutdown rod configuration with forty control rods out further than 02.
onlysubcritical at the present reactor temperature with forty rods at position 02. 1
.J only subcritical at the present reactor-temperaturewith forty rods out further than 02. - - - I SCRAM: I Justification:
in a cold shutdown rod configuration with forty control rods at position 02. Correct. Shutdown will be yes if all rods are at 02 or less.
in a cold shutdown rod configuration with forty control rods out further than 02. Incorrect. 02 or less only subcritical at the present reactor temperature with forty rods at position 02. Incorrect. SDM is
,assured.
only subcritical at the present reactor temperature with forty rods out further than 02.lncorrect. SDM is assured at 02 or less.
I RODMINE003 1 Computer Room):
- a. Explain the function of each indicator.
1 b. Assess plant conditions, which will cause the indicator to light or extinguish.
- c. Determine the effect of each control on the Rod Worth Minimizer.
heir
__- intended functions. .-
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I IMateriai Required for Examination I None INPO Exam Bank Direct From Source I
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Saturday, May IO,2003 12:08:42 .
PM -
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Page 25 of 161 __
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7937 Hatch 03/14/1997.
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7 Given the following conditions: I
- - Security reports that a tanker in the river has run aground and is leaking a large cloud of green gas vapor.
- The wind is carrying this gas towards the plant.
- Security officers report a strong Chlorine odor outside.
Based on these conditions, place CREF in service with CREF boundry dampers in I Mode and operators must I
,OA; ISOLATE-remain
_ _ in the Control Room. _- I OA: NORMAL: establish control at the RSP. i RECIRC; ISOLATE; remain in the Control Room.
- 1 RECIRC;
- . _ _ NORMAL; establish
- __ control at the RSP. - - . - . . - - -___ -1
-C a 295016K203
/Emergency and-Abnormal Plant Evolutions 1-' Control Room Abandonment 18 AB.HVAC-0002 Control Room Environment OA; ISOLATE; remain in the Control Room. Incorrect. Allows gas to enter Control Room Envelope.
,Mode for High Radiation response.
OA; NORMA; establish control at the RSP Incorrect. Allows gas to enter Control Room Envelope.
RECIRC; NORMAL; establish control at the RSP. Incorrect. Closes off outside air flowpath but dampers are in wrong alignment. .- -.
/HC.OP-AB.HVAC-0002 1
1 iI ...
ABHVCI E004 Explain the reasons for how planffsystem parameters respond when implementing HVAC. I I
Saturday, May ~-
10, 2003 12:08:42 PM
Due to a fire in the Control Room console, the Control Room Supervisor orders the Control Room ~
immediately evacuated. The reactor was scrammed remotely.
Which of the following statements describes how a scram is verified in accordance with Shutdown v I from Outside the Control Room, HC.OP-IO.ZZ-008? ~
I HCU accumulator pressure verified to be 950 - 1000 psig at each HCU.
SPDS display terminal "Rods Full In" in the TSC.
Reactor vessel pressure stable at 920 psig.
~
RPS Backup Scram Air Solenoids verifed de-energized.
b Memory [Hope Creek 295016K301
- 19 AK3. ] Knowledge of the reasons for the following responses as they apply to CONTROL ROOM ABANDONMENT:
-~
AK3.01 [ReactorSCkM 14.1 [4.2]
Justification:
IAW 10-0008 step 5.1.3 "If the Rx scram was not verified prior to evacuating the Control Room, then verify Rods Full In (SPDS/CRIDS (TSC) or Activity Control Cards OR Other)
SPDS display terminal "Rods Full In" in the TSC. Correct. IAW 10-0008 step 5.1.4 HCU accumulator pressure verified to be 950 - 1000 psig at each HCU. Incorrect. 950 - 1000 psig is still within the normal charged range of an HCU. The USFAR states "Observing the local nitrogen side pressure indicator for each hydraulic control unit scram accumulator for a low (post scram) pressure indication."
'v RPS Backup Scram Air Solenoids verifed de-energized. Incorrect. Rx pressure at 920 indicates the reactor is at low thermal power level but not necessarily scrammed. The USFAR states "By manually cycling a safetylrelief valve from the RSP (after RSP takeover) and observing an appropriate cooldown as indicated by a reduction in steady state reactor pressure following the steam discharge. Pressure indication can be used since pressure and temperature are directly related in a saturated system. If the reactor were critical, pressure and, correspondingly, temperature, would return to approximately their initial values since the reactor would see this evolution as a power transient."
RPS Backup Scram Air Solenoids verifed de-energized. Incorrect BU scram valves are de-energized with the scram reset. If thev were eneraized. that would indicate a reactor scram.
C.0P-I0.Z-0008 I IUFSAR 7.4.1.4
__ I I
lWlaterial Required for Examination " , 1 None IQuestiopSource: New Question Source Comments:
Saturday, May 10, 2003 12:08:42 PM
~-
Page 28 of 161
Given the following conditions:
- A plant startup is in progress following a forced outage.
- The plant has been operating with a known fuel leak.
- The plant scrammed 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> ago.
- 'A' MechanicalVacuum Pump (MVP) is placed in service with the suction valve throttled.
- The Main Condenser Vacuum Breakers are closed.
Which one of the following actions is required if the South Plant Vent (SPV) RMS Effluent reaches the HIGH level.
Throttle
- _ _ MVP- Suction valve further closed _ _ to _ reduce effluent levels in the SPV.
___ -1 a Swap to the standby MVP to reduce effluent levels-inthe SPV.
I Stop the MVP to stop release to the SPV.
Open the-Main Condenser Vacuum Breakers to stop release to the SPV. I Throttle MVP Suction valve further closed to reduce effluent levels in the SPV. Incorrect. HC.OP-SO.CG-0001 Caution 5.8.13 states the MVP does not need to be stopped if the MVP suction is throttled until the HIGH alarm setpoint is reached.
Swap to the standby MVP to reduce effluent levels in the SPV. Incorrect. Swapping MVPs will not lower release rate.
Open the Main Condenser Vacuum Breakers to stop release to the SPV. Incorrect. Not required. Would increase effluent flow.
IHC.OP-SO.CG-OOOI 1
I 1
[HC.OP-AB.CONT-0004 -- j I
, Condenser Vacuum. I I
I . .. ... .. .. . . . ... .
I
-I..-
~ ~~ ~
1
~~~
Saturday, May IO,2003 1208:42 PM Page 29
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of 161
Given the following conditions:
- The plant is operating at 100 percent power.
-, - A large leak has occurred on the Instrument Air header supplying the CRD Scram Air Header.
- The header pressure is lowering rapidly.
At what point is a reactor manual scram required and why?
When the first control rod drifts due to Low Accumulator Pressure IAW HC.OP-SO.BF-0002 Individual HCU Operation. - . - - -.
-I When a second control rod drifts due to the Cooling Water Flow ControlValve failing open IAW HC.OP-SO.BF-0001 CRD System Operation. . - _ _ . _ -
When the first control-rod drifts due to its Scram Inlet Valve opening IAW HC.OP-AR.ZZ-0011 Attachment E3 for Control Rod Drift.
- When a second control rod drifts due to its Scram Outlet Valve opening IAW HC.OP-
- -- - --1I AB.COMP-0001 _ - Instrument and/or Service Air.
/Hope Creek I Partial or Complete Loss of InstrumentAir c 295019&01 i
21
,295019 AA2: Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF
,INSTRUMENTAIR:
.AA~.OI Instrument air system pressure l3.513.4 u SRO UNIQUE - RO LEVEL QUESTION.
Justification:
When a second control rod drifts due to its Scram Outlet Valve opening IAW HC.OP-AB.COMP-0001 Instrument and/or Service Air. Correct. On loss of header air pressure, the Scram Outlet valves open.
This causes the rods to drift inward. More than one rod drifitng in requires the Mode Switch locked in Shutdown.
When a second control rod drifts due to the Cooling Water Flow Control Valve failing open IAW HC.OP-SO.BF-0001 CRD System Operation. Incorrect. Wrong reason. The Cooling Water Flow control valve fails closed on a loss of air.
When the first control rod drifts due to its Scram Inlet Valve opening IAW HC.OP-AR.ZZ-0011 Attachment E3 for Control Rod Drift. Incorrect. Wrong action. Need more than one rod drifting/scrammed.
When the first control rod drifts due to Low Accumulator Pressure IAW HCOP-SO.BF-0002 Individual HCU Operation. Incorrect. Wrong reason. Low accumulator pressure is a result of a rod scram, not the cause.
~ HC.OP-AB.COMP-000I L Material R&julred for Examidion I None Question Source: New Question ModificationMethod:
L Page 30
__ of 161
~..
~ _ _ ~ _ _May Saturday, _ _I O , 2003 12:08:43PM ~-
KPage 31 2 161-
- ~~
Given the following conditions:
- - The plant is operating at I 0 0 percent power.
- A complete loss of the service and instrument air systems occurs.
- Operators are trying to restart a compressor as air header pressure lowers.
$Whichone of the following is the effect on the Condensate/FeedwaterSystem? A Secondary -Condensate pump Min Flow valves fail closed. - .
1 I
SJAWSPE Bypass - _ Valve PDV-I719 fails open. __ ~
-1 Primary Condensate pump Min Flow valve HV-1710 fails open.
'Feedwater heater dump valves fail closed.
SJAE/SPE Bypass Valve PDV-1719 fails open. Correct. The bypass valve fails open on loss of air.
Secondary Condensate pump Min Flow valves fail closed. Incorrect. SCP min flows fail open.
iPrimary Condensate pump Min Flow valve HV-I710 fails open. Incorrect. Motor Operated Valve.
- Feedwaterheater dump valves fail closed. Cascading drain valves fail close, but high level dump valves v 'fail m e n
[ HC.OP-AB.COMP-000 1 I 1
- _________~ - ~~
Saturday, May I O , 2003 12.0843~-PM ~
_ _ Page 32 of 161
Given the following conditions:
- An ATWS occurs from 100 percent power.
-,. - As corrective actions are being taken, the MSlVs isolate on low RPV level.
- Other MSlV closure interlocks are clear.
- Main condenser vacuum is 3 InHgA.
- Reactor coolant activity levels are normal.
Based on these conditions, which one of the following will allow the MSlVs to be re-opened?
Reactor power is 10 percent. ___ - . - __ - -. -
Suppression pool temperature is rising towards HCTL.
- . I RPV Level is less than -129 inches.
I
'Emergency- _depressurization is anticipated.- i power) and Main con'denser available; (3 InHgA), and No indication of gross fuel failure or steam line break; (normal activity; NSSSS conditions clear)
Suppression pool temperature is rising towards HCTL. Incorrect. Drives pressure reduction, but does not
'give permission to open MSIVS.
RPV Level is less than -129 inches. Incorrect. EOP 301 must be performed.
Emergency depressurization is anticipated. - Depressurizationthrough TBVs is not allowed due to the
- reactoris not shutdown under all conditions without boron.
,HC.0P-EO.=-0 101A _. .-3I
_____i Saturday, May 10,2003 12:08:43PM
Given the following conditions:
-- -- The plant is operating normally at 100 percent power.
An inadvertent High Drywell Pressure Core Spray Manual Channel D initiation signal occurs.
'What effect does this have on Drywell Cooling? .___
Drywell Cooler fans A I through H I trip but can be restarted. __
Drywell Cooler fans A2 through H2 trip and CANNOT- be_--__ --.
restarted.
___ - . -.- - --.-- - - .-I Drywell Chilled Water Isolation valves close but all
- _ Drywell Cooling fans remain running. _ _
I Drywell Cooler fans A2 through H2 trip and Chilled Water Isolation . - - valves close.
-1 Drywell Cooler fans A2 through H2 trip and Chilled Water Isolation valves close. Incorrect. CS Manual D closes Isolation valves and A2 through H2 fans trip.
Drywell Cooler fans A I through H I trip but can be restarted. Incorrect. A2 through H2 fans trip.
Drywell Cooler fans A2 through H2 trip and CANNOT be restarted. Incorrect. Can be restarted if Load '
-.-/ Shed breaker to MCC re-closed.
I L - _ I D & ~ I I Ventilation System Lesson Plan:
- a. Chilled Water System
- b. Reactor Auxiliaries Cooling System (RACS)
- c. Electrical Power Supply
- d. Plant Leak Detection System
- e. Process Computer
___ -~
Saturday,
~- - May-
~
I O -, 2003 12:08:43 PM
__ ___- -_ l_._
Page 34 of 161
---_______?
Given the following conditions: II I
- The plant is in Operational Condition 4 at 180 degF.
i/ - The reactor scrammed on startup from a refueling outage.
- RHR Loop A operating in Shutdown Cooling.
- The 9 RHR pump is Cleared & Tagged for motor replacement.
- The A RHR pump develops a high vibration and trips on overcurrent.
- BC-HV-F008 has spuriously closed and will NOT reopen.
- HC.OP-AB.RPV-0009, Shutdown Cooling, is entered.
Which of the following will be adequate as an Alternate Decay Heat Removal method for the conditions above? I I Crosstie C or D RHR pump for-heat removal. .
_ - _ - 1 Inject with Condensate Transfer _- and reject with RWCU.
Use natural circulation and Drywell coolers for heat removal. ._ . - I Inject with one Core Spray pump - _ from
Inject with Condensate Transfer and reject with RWCU. Correct. Feed and bleed will ma temperatures at low decay heat loads such as at BOL.
Crosstie C or ID RHR pump for heat removal. Incorrect. Need F008 open.
Use natural circulation and Drywell coolers for heat removal. Incorrect. Will remove some decay heat but not enough to be included if RPV-0009 actions.
Inject with one Core Spray pump from the CST to the RPV. Incorrect. Need - one- loop of Core Spray.
I I
___ I Saturday, May I O , 2003 12:08:43 PM Page 35 of 161
- _ I - -
~
i, The plant is operating at 100% power when a CRD Temperature High alarm is received.
Which one of the following could have caused the CRD high temperature condition?
Eroded CRD cooling water orifice. 1 17 Stabilizing valve failed fully open.
I CRD pump low discharge pressure.
CRD flow control valve failed fully open.
be caused by abnormal drive pressure and any of the following:
- 1. leaking scram discharge valve
- 2. low cooling water flow
- 3. defective thermocouple circuit
- 4. plugged CRD cooling water orifice Justification:
CRD pump low discharge pressure.-correct per HC.OP-BD.IC-0001. Would cause low cooling water flow.
Stabilizing valve failed fully open-incorrect- this is the normal position, failure to close would possibly cause hunting of the FCV during rod motion iEroded CRD cooling water orifice-incorrect opposite effect of plugged orifice low temperatures CRD flow control valve failed fully open-incorrect high cooling water flow opposite effect lower temPeratures
/HC.OP-BD.
I IC-0001 I 1
~ ~ - PM Saturday,%y-. 10, 2003~120843 ~-
Given the following conditions:
- The plant is in OPCON 5.
u
- Core offload is in progress.
,- A spent fuel bundle is full up on the main hoist over the core.
- The refuel bridge spotter notices the fuel bundle has become unlatched and )as fallen into the vesseI.
- The water clarity has degraded significantly.
- A short time later, the following Refuel Floor Rad Monitors alarm:
- Spent Fuel Pool ARM.
- New Fuel Criticality ARM.
Based on these conditions, what operator action is required?
Suspend all refueling - __ - operations.
Remove the-Fuel Pool Cooling System com service to reduce Reactor Building- __radiation levels.!
Initiate action- to_establish-Secondary Containment within Ihour. __ - -1 Determine the location of the dropped bundle, inform theCRS, and evacuate the Refuel Floor.- I
~
a -s Memory \Hope Creek
!Emergency and Abnormal Plant Evolutions c3 2950236449 1 3 1 RefuelingAccidents - - - I - 27' 2.4.49 /Abilityto perform without reference to procedures those actions that require immediate operation of 14.0 14.01 L lsystem components and controls.
Justification:
SRO 10CFR55.43 (7) Fuel handling facilities and procedures.
SRO 55.43 (6) Procedures and limitations involved in alterations in core configuration.
6RO 55.43 (4) Radiation hazards that may arise during normal and abnormal situations.
Correct- Suspend all refueling operations. AB.CONT-0005 is entered due to the New Fuel Criticality ARM alarm. This is the Immediate Operator action.
Incorrect - Remove the Fuel Pool Cooling System from service to reduce Reactor Building radiation levels. CONT-0005 subsequent action C discusses the increased rad levels in the FPCC System but no direction to remove the system from service is provided.
Incorrect - Initiate action to establish Secondary Containment within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> - Secondary containment is required to be in place during all fuel moves there is no time limit if lost. Step 4.3 requires Secondary Containment to be verified.
Incorrect - Determine the location of the dropped bundle, inform the CRS and evacuate the Refuel Floor.
The onerator should not wait to determine the location of the dromed bundle..
ABCNT5EOO3 ' (R)From memory, recall the Immediate Operator Actions for Irradiated Fuel Damage. I one Saturdav. Mav I O . 2003 12:08:44PM
__ -~
Saturday, May [Page 38 of 161
-~I O-, 2003 1208 44 PM
~
~
Which of the following indications will positively identify a criticality event in .boqress - whileafuel 1 bundle is being lowered - into the core during refueling operations? i Source range monitor spiking repeatedly. -
v 1 A sustained upward trend on the nearest source range instrument to the fuel bundle location. J The high refuel floor radiation alarm sounds. __1 Refuel bridge hoist motion interlock activates.
1 AKl. Knowledge ofthe operational implications of the following concepts as theyapply to REFUELING I
'ACCIDENTS:
Source range monitor spiking repeatedly. Incorrect. Indications of a detector failure.
IThe high refuel floor radiation alarm sounds. Incorrect. Would be correct if in New Fuel Vault.
Refuel bridge hoist motion interlock activates. Incorrect. No hoist interlocks will activate as a result of a criticalitv in the core. . . . .. . .. . . .. .. . I jHC.RE-AP.ZZ-0049 3.3.1.D r
- b. Refueling Bridge Operator
- c. Control Room Refuel Monitor IAW NC.NA-AP.ZZ-0049.
' Modified -- I Saturday, May 10,2003 12:08:44 PM
__I____ -- 1 Page39of161 ~
Given the following conditions:
- The reactor has scrammed (all control rods are at position 00) on high drywell pressure.
- Reactor pressure is 35 psig.
i/
- Reactor level is -120 inches rising.
- Suppression pool level is 75 inches.
- Suppression pool temp is 120°F.
- Suppression chamber temp is 100°F.
- Suppression chamber press is 15 psig.
- Drywell temp is 280°F.
- Drywell pressure is 17 psig.
To control the primary containment under these conditions the operator should monitor and control hydrogen concentration in the Supp Chamber and
- placetwo loops of RHR in drywell _-
spray.
_ _ the Drywell and: 1 __
-1
-. J place one loop of RHR in suppression pool cooling and the other loop of RHR in drywell and suppression chamber spray.
-. - ___ 1 place one loop of RHR in suppression pool cooling and suppression chamber spray and the other loop of RHR in drywell - spray.
place one loop of RHR in suppression pool cooling and the other loop of RHR in d&vell spray,
__ __ -- --I 1
,and vent the suDDression chamber.
~
I .
~ .. ... . . ~ ... . ... ~ . . . .. .. . .. . .. . ... . . . .. .. .. - - ... .. ~ .. . .. .-..... i 1
W [Emergency and Abnormal Plant Evolutions m-4 _HighDrywell Pressure - 1
=I' -
Ability to determine and/or interpret the following as they apply to HIGH DRYWELL PRESSURE:
- - - r 1 4.2 14.41 I - 1 EA&O1 /Drywellpressure Place one loop of RHR in Suppression pool cooling and suppression chamber spray and the other loop of RHR in drywell spray.-Correct-adequate core cooling is assured, SP temp requires SP coolinglspray, DWSIL curve is satisfied DW Spray is appropriate.
Place one loop of RHR in Suppression pool cooling and the other loop of RHR in drywell and suppression chamber spray-Incorrect- DW spray is always on a loop by itself, if available the second loop is placed in SP cooling/spray to prevent inadvertent bypass of the containment on the operating pump trip.
Place two loops of RHR in drywell spray.-incorrect- never place both loops in DW spray this may exceed the makeup capacity of the vacuum breakers and draw the containment negative.
Place one loop of RHR in Suppression pool cooling and the other loop of RHR in drywell spray, and vent the suppression chamber-incorrect-venting the containment is only requires if pressure cannot be i_ E0102PE007
__ 1 (R)Given any step of the procedure, determine the reason for performance of that step andlor predict expected system
,.~response to control manipulations prescribed
-~ - by that step IAW the Primary Containment Control - Drywell _Lesson_Plan. _-
Saturday, M a G O , 2003 12:08:44 _
. - ~-
PM_ - of 161 Page 40 -__
/Direct From Source I
~
Saturday, May 10, 2003 12:08:44 PM---
Given the following conditions:
- A LOCA has resulted from a seismic event.
- Reactor water level is -20 inches and rising.
- Reactor pressure is 850 psig and slowly lowering.
- Drywell Pressure is 31 psig and slowly rising.
- Drywell temp is 275 O F and slowly rising.
- Suppression Chamber pressure is 30 psig and slowly rising.
- Suppression Pool water level is 77 inches.
- 1 BD417 1E 125 VDC distribution panel is de-energized.
- "A'RHR Loop is in Drywell Spray.
,- The Main Condenser is NOT available.
- All control rods are full in.
Based on the above conditions, when is Emergency Depressurization of the reactor required?
Immediately using all Turbine Bypass Valves. - - . - - -
Immediately using 5 ADS valves. . -
.- I When __ Drywell_ Temp
_ reaches 310F using 5 ADS valves.
- .. -1 When
__ Drywell Press reaches 35 psig using all Turbine Bypass ... - Valves.
- -1 L'
EM.-1Ability to determine andlor interpret the following as they apply to HIGH DRYWELL PRESSURE:-
Ee.04 puppression chamber pressure: Plant-Specific [ 3.9 [ 3.9J I
JUSTIFICATION: i SRO 55.43(5)Assessment of facility conditions and selection of appropriate procedures during normal, ~
abnormal and emergengcy situations.
CORRECT - Immediately using 5 ADS valves. Emergency de-pressurization must occur now for exceeding the PSP curve.
INCORRECT- Immediately using Turbine Bypass Valves. Emergency de-pressurization can no longer be anticipated. Emergency de-pressurization must occur now for exceeding the PSP curve. I INCORRECT - When Drywell Temp reaches 310°F using 5 ADS valves. Emergency de-pressurization must occur now before exceeding the PSP curve. Emergency de-pressurizationfor high Drywell temperature does not occur until 340°F.
INCORRECT- When Drywell Press reaches 35 psig using Turbine Bypass Valves. Emergency de-pressurization can no longer be anticipated. Emergency de-pressurization must occur now for exceeding l the PSP curve I I
I E 0 1 0 2 P E O 0 6 ~(R) Given plantnditions and access to the following curves determine the region of acceptable operation and explain the '
1 bases for the curve IAW the Primary Containment Control - Drywell Lesson Plan: 1
- a. Drywell Spray Initiation Limit W
EOPl02E009 Saturdav..~.
Mav- .10,2003
. 12:08:44 PM
manipulations prescribed by that steGAW the Primary Containment Control -
Plan.
~ ~ - -
Saturday, May 10, 2003 12:08:45
~ -.
1
_. - p
_ _ _ _ _ _ . ~ .
Which of the following components allows for absorption of energy released from a Reactor RecG I
line ___-
rupture?
~- -A Primary Containment isolation valves.
-__ -1
_ -A L ~- __
Torus to Drywell vacuum breakers. - -7 Vent header downcomer pipes. - -
Reactor Buildinn to Torus vacuum breakers. -7 295024G127-
-1 2.1 .-]:Conductof Operations - - J CORRECT: Vent header downcomer pipes direct LOCA steam from drywell to Torus to condense steam. Directs LOCA steam from drywell to the suppression chamber for condensing.
I INCORRECT: Torus to DW Vacuum Breakers do not absorb LOCA energy, they protect DW from t release of fission products orus from External Pressure/
r i
I PRICONE003 Summarize the basic construction and function of the following Primary Containment components IAW the Student Handout:
- a. Drywell Steel Shell
- b. Biological Shield Wall
- c. Drywell Head
- d. Drywell Water Seal Plate
- e. RPV Pedestal
- f. Ring Girder
- g. Torus
- h. Vent Pipes
- i. Vent Header
- j. Oowncomers
- k. Torus-to-Drywell Vacuum Breakers I. Rx. Bldg.-to-Torus Vacuum Breakers
- m. Torus-to-Drywell Expansion Bellows
- n. Personnel Access Hatch __-
_ L I 1 Page 44 of 161 i_-._p_ .
Following a reactor scram and Main Steam Isolation Valve closure, reactor pressure reaches 1050 psig.
- Which of the following describes the response of the "H" and "P" Safety Relief Valves (SRV) in the Low-Low Set mode of operation for the given conditions?
The "P" SRV opens, which actuates low-low-set causing the "H" SRV to open and both valves will control pressure at new, lower operating setpoints.
_ ~~
I The "H" and "P" SRVs both open and both valves will control pressure at the same opening setpoints and new, lower closing . __- setpoints.
The "H" SRV opens and the "H" and "Pl' SRVs control pressure together at new operating setpoints. - - _ -
The "H" and "PI' SRVS both open and the "H" SRV will control pressure at new operating - 1 setpoints with the "P" SRV operating as needed at slightly higher than "H" operating
. - - . - setpoints.
__ 1 d B
! y c n e g r emE [ 1295025A103 High Reactor Pressure 32 orrect answer: The "H" and "P" SRVs both open and the "H"SRV will control pressure at new operating
'setpoints with the "P" SRV operating as needed at slightly higher than "H"operating setpoints.
The Low-Low set function arms at 1047 psig. Once armed the H will lift at 1017 for all subsequent lifts, and the P will remain the same at 1047 psig. The H will close at 905 psig and the P will reclose at 935 psig.
The following distractors are incorrect:
The "H" and "P" SRVs both open and both valves will control pressure at the same opening setpoints and new, lower closing setpoints. Incorrect. The Low-Low set function arms at 1047 psig. Once armed the H will lift at 1017 for all subsequent lifts, and the P will remain the same at 1047 psig. The H will close at 905 psig and the P will reclose at 935 psig.
The "H" SRV opens and the "H" and "P" SRVs control pressure together at new operating setpoints.
Incorrect. The Low-Low set function arms at 1047 psig. Once armed the H will lift at 1017 for all subsequent lifts, and the P will remain the same at 1047 psig. The H will close at 905 psig and the P will reclose at 935 psig.
The "P" SRV opens, which actuates low-low set causing the "H"SRV to open and both valves will control pressure at new, lower operating setpoints. Incorrect. The Low-Low set function arms at 1047 psig.
Once armed the H will lift at 1017 for all subsequent lifts, and the P will remain the same at 1047 psig.
The H will close at 905 psig and the P will reclose at 935 psi.
/HC.OP-SO.SN-0001 I I
7 ::I The number and type of SRVs at Hope Creek.
Which SRVs have an ADS function.
7 Saturday, May 10,2003 12:08:45
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PM'
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I Page 45 of- 161
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1
- c. Power supplies to the SRV solenoids.
- d. Which SRVs can be operated remotely and the location from which each of these valves can be operated.
i I
I L
Vision Bank QID# Q53505
~
Saturday, May 10, 2003 12:08:45 PM
Given the following conditions:
- The plant is in an ATWS condition.
.- '- MSIV's are closed, pressure control band is 800 to 900 psig using SRV's.
- APRM's read 10%.
- Manual rod insertion is in progress.
- Suppression Chamber pressure is 2.8 psig.
- Reactor pressure is 850 psig.
- Suppression pool water temperature is 195 degrees F.
Based on these conditions, which of the following require an immediate reactor pressure reduction? I i
Reactor pressure stabilizes at 900 psig.
. I SuDpression Pool level reaches I 2 0 inches and is rising.
Suppression Pool water temperature reaches 205 degrees __ F. ._ 1 Suppression-Pool
. - ___ level reaches 70 inches and is lowering. - __ - - - - I
-C #Application IHope Creek
'd Pool water temperature reaches 205 degrees F. Correct. IAW EOP 102 Curve SPT-P with RPV pressure at the upper end of the band at 900 psig, you would be in the action required area of the curve. The required action is to reduce pressure to get below the curve IAW step SP/T-7.
Suppression Pool level reaches 120 inches and is rising. Incorrect. Still below the action level of 124 ~
inches.
Reactor pressure stabilizes at 900 psig. Incorrect. Pressure reductionnot required at 900 psig and 195
'DegF SPT.
Suppression Pool level reaches 70 inches and is lowering. Not required - . . until 38.5 inches SPL IEOP 102 SteD SP/T-7 EOPI 02E009 1
1 EOP Flowcharts Saturday, May 10, 2003 12:08:45 PM j P a g e 4 7 0 f 161
~
~
Given the following conditions:
- - Drywell pressure is 7.3 psig and rising.
- Rx water level is -40 inches.
- RX pressure 1008 psig and lowering.
- 67 control rods NOT Full-In.
- Suppression pool temperature 129 deg F and increasing.
- MSlVs are closed.
- Suppression pool water level is 84 inches.
Which one of the following EOP actions is required for these conditions and why?
Inject SBLC to prevent Suppression Chamber water temperature from exceeding the Heat Capacity Temerature Limit. _ _
Spray Drywell to prevent SRVs from exceeding the Suppression Chamber Dynamic
- _-- Load Limit.]
I Maintain water level +12.5"to +54" __ to prevent large _power
_. - swings on the reactor core. - .
-I Emergency Depressurize to prevent Dr&ell pressure from exceeding Primary-Containment -1 l- /
Inject SBLC to prevent Suppression Chamber water temperature from exceeding the Heat Capacity Temerature Limit. Correct. Bases of Boron injection Initiation Temperature limit.
Spray Drywell to prevent SRVs from exceeding the Suppression Chamber Dynamic Load Limit.
incorrect. Wrong limit.
Maintain water level +12.5'to + 5 4 to prevent large power swings on the reactor core. Incorrect. Level must be lowered.
Emergency Depressurizeto prevent Drywell pressure from exceeding Primary Containment Pressure Limit.hco;rect. Drywell
. . should be sprayed. ED _____ action
- ___ to .be taken if sprays
- -not effective.
J iEOP 101A bases BllT 1
I I
EOP Flowcharts
_ _ _ ~ ~ ~ - -
__ Saturday, May IO, .____
2003 12:08:45
_____-PM
~
i Page 48 of 161
~ _ _ _ _
r--
Page 49 of 161
~
Saturday,-May 10, 2003 12:08:46 PM I Page 51 of 161
____-___ ~. ~- -.
Given the following conditions:
- The plant is several hours into a LOCA.
- - HPCl automatically initiated and then subsequently tripped on low oil pressure.
- A & B RHR loops are NOT available.
- All other available ECCS are injecting.
- Drywell pressure is 64.4 psig and rising.
- HPCl Pump suction pressure is 73 psig.
- SP level indication is failed.
- SP temperature is 175°F.
What is containment water level and based on that level, which of the following actions are required?
21.2 ft; Vent the Suppression Pool. _ -
21.2.ft; Vent the Drywell.
Li procedures during normal, abnormal and emergency conditions.
Containment Level = [(HPCI Press - DW Press) 2.3 Wpsi] + 2.2ft.
CORRECT - 22.0'. [(73.0 - 64.4) 2.3) + 2.2. 22.0 ft containment level ( 263.8 inches) or equivalent to 169.8 inches indicated torus level. EOP-102 Step DWP-12 requires Venting the Suppression Pool if
-480 inches in the torus.
INCORRECT 21.2'. Vent the Suppression Pool. I(73.0 - 64.4)2.2]+2.3 (2.2 and 2.3 transposed in formula) Wrong value; correct action.
INCORRECT - 21.2'; Vent the Drywell. (2.2 and 2.3 transposed in formula) Wrong value; incorrect action based on 254.6 inches (94 inches must be subtracted to provided equivalent Suppression Pool water level.)
,INCORRECT - 22.0 ft; Vent the Drywell. Correct value; incorrect action based on 94 inches must be EOPl02E011 I (R)Given the formula for calculating containment water level and corresponding values, calculate containment water level IAW I
I EOPI 02E009 i response to ckntrol manipulations prescribed by that step IAW the Primary Containment Control - Suppression Pool Lesson I
-1 1
Saturday, _~ May I O , 2003_ 12:08:46 ! Page 50 of 161
_ _ _ _ _ PM
~ ~
__ ___ .~
Given the following conditions:
- - A LOCA has occurred.
- RPV water level is stabilized above TAF.
- Suppression Chamber water level is 60 inches and lowering.
Which one of the following correctly fills in the blanks describing the ALTERNATE Core Spray Loop to be used for Suppression Chamber Makeup and the prerequisite Suppression Chamber pressure?
Core Spray Loop is the ALTERNATE Makeup path to be used only if Suppression
__ ~, . .. ..~
. .-I
... . ... ... ~ .. ~ . .. . ~ ~ ~
6...; .below
. . . . . .. .~ .. . . . ~~. .. .. ..... ~ ~ .... ~ ... --
- to implement M/U from the CST via Core Spray. A CS Loop is the Alternate loop.
1 IHC.OP-EO.=-0315 2.2.2
- 1 I
EOP30OE002
- - - I 1
c Saturday, May 10,. 2003 12:08:46 PM of 161 Page 52 _ _
~- ~
L Which one of the following correctly describes the Technical Specification bases for the Suppression Pool low water level limit?
This limit ensures adequate SRV T-Quencher submergance during Emergency Depressurization.
=
This limit ensure adequate water volume is available based on NPSH and vortex prevention. 1 This limit prevents exceeding the Suppression Pool design temperature limit during a DBA LOCA.
10CFR55.43 (2) Facility operating limitations in the Technical Specifications and their bases.
This limit prevents exceeding the Suppression Pool design pressure limit during a DBA LOCA. Correct.
Bases for SP lower water level limit IAW HCGS TS 3.6.
This limit ensures adequate SRV T-Quencher submergance during Emergency Depressurization.
Incorrect.
This limit ensure adequate water volume is available based on NPSH and vortex prevention. Incorrect.
This limit prevents exceeding the Suppression Pool design temperature limit during a DBA LOCA.
Incorrect.
b IHCGS TS Bases 314.6.2
- - 1 II - 'I I I
PRlCONEOO9 (R) Given a Scenario of applicable operating conditions and access to technical specifications: I
- a. Select those sections which are applicable to the Primary Containment Structure IAW HCGS technical specifications.
- b. Evaluate Primary Containment Structure operability and determine required actions based upon system operability IAW HCGS technical specifications. (SROISTA ONLY)
~ - ~~
_~ Saturday, May 10,2003
~ ___12:08:46PM -
Given the following conditions:
- A Station Blackout occurred.
v - B EDG is running.
- B RHR is in Suppression Pool Cooling at 10,000 gpm.
- HPCl and RClC have tripped on Low Steam Inlet pressure.
- Suppression Pool Temperature is 225F rising slowly.
- RPV water Level is -150 inches and lowering slowly.
- Suppression Chamber pressure is 5 psig.
- Drywell temperature is 310F rising slowly.
- - Drywell pressure is 5 psig.
- Suppression Pool water level just reached 0 inches.
- NO other ECCS is running.
Which one of the following actions is required?
Lower B RHR SP Cooling flow to 9000 gpm. -__ . _ _ - - __ - _ --_ --1 u
EAI~I (LOW pressure coolantiijectiin (RHR): Plant-Specific } 4.4 14.41
-2 requires LPCl started and maximize injection flow to the RPV and ignore NPSH limits.
Lower B RHR SP Cooling flow to 5000 gpm. Incorrect. Would be correct if remaining in SPC.
Trip B RHR pump immediately. Incorrect- proper action for non ECCS pump on inadequate NPSH.
Open 5 ADS SRVs to emergency - - depressurize the RPV. Incorrect. EOP-202 step RF-2 requires SP level above 0 inches, other means
- are used to____depressurize.
(HC.OP-EO.ZZ-010 ALC-2
- . - __ - .. ._- 1 I
7 I
I
) Given any step of the procedure, describe the reason for performance of that step and/or expected system response to I 1 control manipulation prescribed by that step.
E0101LE008 (R)Explain the significance of "Minimum Steam Cooling RPV Water Level " and state its~. value.
I EOP Flowcharts
_ ~ _ -- -
Saturday, May IO, 2003 12:08:46
~ _ _ _ PM _ ~
Given the following conditions:
- Drywell pressure is 4.5 psig.
u - RPV level is -45 inches and is being intentionally lowered.
- Many control rods remain at their original positions.
- SLC, CRD, and RClC are injecting.
- Reactor power is 7 percent.
- SRVs are cycling on Low Low Set.
'- A and B RHR Loops are in Suppression Pool Cooling.
- Suppression Pool temperature is 112 F.
Which one of the following conditions permits RPV water level to be stabilized between -190 and the current RPV level when that condition is achieved?
__ 1 Reactor power reaches 3 percent.
- --1 RPV level reaches
-50 inches.
i/
abnormal, and emergency situations. This item test SRO ability to resolve the question by first choosing
'the applicable EOP Flowchart and correctly applying the stem conditions to that flowchart.
Correct. Reactor power reaches 3 percent. The conditions provided in the stem require terminate and prevent injection and RPV level reduction to lower power. The conditions require EOP-IOIA Step LP-14
- implementedfor level reduction. If reactor power reaches 3 percent from 7percent, EOP-IOIA Step LP-14 allows level reduction to be stopped. Step LP-15 then allows RPV level to be maintained between that level and -1 90 inches.
Incorrect- All SRVs remain closed. Would be correct if Drywell Pressure was below 1.68 psig.
Incorrect- Suppression Pool Temperature lowers to 108 F. With Stem conditions of 7 percent power, Step LP-I 1 is answered YES. This requires lowering level until power is less than 4 percent. You can not iback up and change the answer to LP-11 using the retainment step LP-6. Plausible misconception.
Incorrect. RPV level reaches -50 inches. Based on stem conditions, Steps LP-11 and LP-12 must be answered YES. This bypasses lowerin I .-
on for performance of that step and/or evaluate th I 1
~.
Page 55 of 161 -
EOP 101 Flowchart
-~
Saturday, May 10, 2003 12:08:47 PM - rPage 5 6 of 161
Given the following conditions:
- The plant is operating at rated power.
i,
- Control Room Overhead alarms are received:
- BI-A4 HPCl TURBINE TRIP
- D3-AI HPCVRHR A LEAK TEMP HI
- When the operator checks the HPCl panel, HPCI inboard and outboard steam line isolation valves are stroking closed.
- HPCl turbine was NOT running at the time.
- HPCI Steam line pressure is 900 psig and lowering.
Which one of the following would cause this isolation?
HPCI Pipe Chase High Temperature. __ - -
- HPCI Pump Room High _ - Temperature.
HPCI Steam-Line High - - Differential Pressure.
~
C a - . -
(Hope Creek
/Emergency and Abnormal __
- . Plant Evolutions 3 295032AlOl 2 1 High Secondary Containment Area Temperatu I 40 v
m1.] Ability to operate and/or monitor the following as they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURE:
m ? , - _- - __ -
1-
'3.6]q
,HPCI EquipmentRoom High Temp. Correct. Isolates immediately. Causes D3-A1 annunciator when
'tripped which in turn causes B l -A4 annunciator.
HPCl Steam Line Low Supply Pressure. Incorrect. Isolates at 100 psig. Pressure is 900 psig.
HPCl Pipe Chase High Temp. Incorrect. Isolates after 15 minute time delay. Would also cause annuciator D3-B1 HPCl STM LK ISLN TIMER INITIATED.
,HPCI Steam Line High Differential Pressure. Incorrect. Would cause B1-A5 annunciator, HPCl STEAM LINE DlFF PRESSURE HI.
pC.OP-AR.ZZ-0014 Attachment A I I I I EOP103E006
~
(R) Given any step in the procedure, describe the reason for performance of that step and/or expected system response to
- control manipulations prescribed by the step.
~___
__ _- 7 1
. . I
~ Paae 57 of 161
Given the following conditions:
- The plant is operating at 100 percent power.
- RACS RMS levels have begun to rise.
- Reactor Building backround radiation levels in the vicinity of RACS piping are also rising.
Which of the following would be the cause of rising RACS RMS readings?
__ - - __ - - __ - - - - ____- __- .- J Tube _rupture
_ _ - - in a Reactor Recirc Pump Seal Cooler Heat Exchanger.
- -1
-1 c -
Drywell Equipment Drain Sump Cooler leak.
EAI: ]Ability to operate and/or monitor the following as they apply to HIGH SECONDARY CONTAINMENT AREA RADIATION LEVELS:
iEJI-.O2-, [Process radiation monitoring system 3.7 13.81 Drywell Equipment Drain Sump Cooler leak. Incorrect- DWEDS Cooler cooled by Chilled Water. Can be cooled by RACS if cross-tied but RACS would then leak into the sump.
- Reactor Building Equipment Drain Sump Cooler leak.-Incorrect- Rx Bldg. Equipment Drain Sump Cooler at lower pressure than RACS. RACS would leak into the sump.
Tube rupture in RWCU Regenerative Heat Exchanger.-Incorrect- The RWCU regenerative heat
'exchanger is not cooled by RACS/only NRHX.
Tube rupture Reactor Recirc Pump Seal Cooler Heat Exchanger. - Correct. Reactor coolant from the seal area will leak into the RACS system and cause RMS to rise.
~
1 Saturday, May 10,2003 12:08:47 PM
~
1 -Page 58 of 161
Given the following conditions:
- The plant is operating at 100 percent power.
- Overhead annunciator E6-C5 "RBVS &WING AREA HVAC PNL 10C382" alarms.
- The Reactor Operator reports Reactor Building Differential Pressure is negative at 0.25 inches water gauge.
Which one of the following actions is required?
- - ___ - , __ - .- __ .- - - I Start another Reactor Building Supplyfan IAWHC.OP-SO.GR-0001 Reactor Building IVenti lation.
- __ . __ - - -- 1 Place
_ - FRVS in service IAW HC.OP-AB.=-0001 Transient Plant_Conditions: - - -__- - -
-- - - - -- -1 Isolate
__ - RBVS Isolation Dampers__IAW - HClOP-S0.SM-0001 Isolation - --
System Operation._ _ --_ _- _- 1-piace FRVS in service IAW Hc.oP-sO.GU-OOOI-FRVS - __ Operation.
- -- - I SRO 1OCRF 55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Place FRVS in service IAW HC.OP-SO.GU-0001 FRVS Operation. Correct. FRVS is required by AB-
'CONT-0003 because RB DP is less than the required 0.30 inches WC.
Place FRVS in service using HC.OP-AB.ZZ-0001 Transient Conditions. Incorrect. AB-ZZ-0001 does not provide direction for starting FRVS.
Isolate RBVS Isolation Dampers IAW HCOP-SOSM-0001 Isolation System Operation. Incorrect.
Isolations performed with this procedure are performed to isolate system breaches.
Start another Reactor Buildin fan IAW HC.OP-SO.GR-0001 Reactor Building Ventilation.
1
[HC.OP~AB.CONT-0003 __ -_ - ___ - - JI
~HC.OP-SO.GU-OOOOI I
ABCNT3E001 anize abnormal indicationslalarmsandlor procedural requirementsfor implementing Reactor Building Integrity
_ _Saturday,
_ _ ~ May - 10, 2003 12:08:47 PM
Saturday, May 10, 2003 12:08:47 PM
_ ~ _ _
1 Page 61 of 161
Given the following conditions:
- The reactor is operating at 100% power.
- , - Annunciator BI-B3 ( RClC PUMP ROOM FLOODED ) alarms with the following alarm message presented on the CRIDS display: D2887 RClC PUMP RM 41 10 LSH 4151-1 HI.
- An investigation reveals that Reactor Building Floor Drain Sump pumps have been running continuously for 20 minutes.
- The Reactor Building Operator reports the RCIC, B and D RHR Pump rooms have about 6 inches of water on the floor when he checked the elevation.
- CST level is lowering.
In addition to running the sump pumps, which of the following action(s), if any, is required by EOP 103/4?
I --- Isolate RClC II -- Immediately commence a normal reactor shutdown 111 -- Runback reactor recirculation and manually scram the reactor IV - Emergency depressurize the reactor
-I
'I, III, and IV i,
=]:Knowledge of the reasons for the following responses as they apply to SECONDARY CONTAINMENT
/HIGH SUMPlAREA WATER - - . LEVEL: . - . .
3.51361 -
I, Ill, and IV. Incorrect. Normal shutdown not scram and ED. Would apply if reactor coolant leak.
I and II. Correct. EOP-103/4 Step RB-15 The leak is not reactor coolant discharging into the area. Isolate RClC and commence a normal shutdown.
II - ONLY. Incorrect. Requires RClC Isolation.
I - ONLY. Incorrect. Reauired normal shutdown.
IHC.OP-EO.ZZ-O103/4 I
.... . . .... ......... - . -.I EOP103E006 1 (R)Given any step in the procedure, describe the reason for performance of that step and/or expected system response to control maninulations nrescribed bv the sten.
1 I
I - _____.
EOPI 03E002 I, Given a set of plant conditions, analyze and determine if entry conditions into -
-~ - ._
HC.OP-EO.ZZ-O103/4
~ exists._ _ 1 EOP103E005 (R) Define theterm "Maximum Safe Floor Level".
r -
Saturday, May -- 10, 2003 12:08'47
_ _PM- -_ - Page 60 of 161
~ __
Given the following conditions:
- An ATWS occurred from 100 percent power.
.- - All immediate operator actions have been completed.
- At 1420 hrs, both SLC pumps are started and the SLC Tank Low Level Computer point alarm is received.
- B SLC Pump has tripped immediately after start.
'Assuming the remaining SLC pump delivers Tech Spec minimum flow rate for the next 90 minutes, which one of the following actions are required?
Continue SLC pump-operation and raise reactor water level.
~- -... - . . - -
1
- Continue SLC pump operation and begin reactor cooldown.
I
'Verifv the SLC bumr, is tripbed and continue rod insertion-. - 1 Verify the SLC
- pump
__ - is
- -tripped and exit EOP-IOIA.
_ _ J b Application [HopeCreek 295037A203 44-I EA2. ] Ability to determine and/or interpret the following as they apply to SCRAM CONDITION PRESENT-AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN:
.EA2.03
- ~SBLCtanklevel I
14.3
. . 14.41 Justification:
Continue SLC pump operation and begin reactor cooldown. Correct. After 90 minutes operation with only 1 SLC pump running, less then 1100 gallons remain in th SLC Tank, but level is above the 325 gallon Low Level Pump trip setpoint. 4640 Gallons at the low level alarm point with 90 minutes runtime at 41.2 gpm 14640 - (41.2 x go)] = 932 gallons remaining. Step RC/Q-19 directs continuation at step RC/P-20 for depressurizationand cooldown.
Continue SLC pump operation and raise reactor water level. Incorrect. RPV Level cannot be raised until the reactor is shutdown under all conditions without boron.
Verify the 'A' SLC pump is tripped and continue rod insertion. Incorrect. 'A' pump will still be running.
Verify the 'A' SLC pump is tripped and exit EOP-IOIA. Incorrect. 'A' pump will still be running. EOP 101A is not exited until the reactor is shutdown under all conditions without boron.
/HC.OP-EO.Z~I II 01A r
EOP Flowcharts, HCGS TS section 3 1.5 _ - . __ .-
C Saturda7,May 10,2003 12:08:47 PM 1 Page 62 of 161
~- - _ . _ _ _ _ _ _ ~ _ - ..~
Given the following conditions:
- A turbine trip and hydraulic ATWS occur from 65 percent power.
- - EOP-IOIA and I 0 2 are currently being executed.
Current plant conditions:
RPV Parameters
- Pressure 950 psig with TBVs controlling.
- Level -80" with RClC and CRD injecting.
- Power IRM Range 8 @ 50 / 125 decreasing.
- SLC is injecting.
Containment Parameters
- Suppression Pool water temperature is 192 F steady
- Suppression Pool level 76.8 inches rising slowly Which of the following actions will improve the required margin of safety?
1 HI Lowersuppression POOI level. -
- -1 i/
Open additional Turbine Bypass Valves.
- 1 Rapidlv Depressurize the RPV. I
- Application a . [Hope Creek 295037G406 45 Emergency Procedures and Plan 2.4.6 , - 2 3 . 1 m,
(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Reduce Suppression Pool water temperature. Correct. SPT reduction allowed and required.
Lower Suppression Pool level. Incorrect. Level within allowable band.
Open additional Turbine Bypass Valves. Incorrect. Would cause cooldown with ATWS in progress. Not at SPT-P limit.
Rapidly Depressurizethe RPV. Incorrect. Not permitted at this time.
- ~~
i r
response to control manipulations prescribed_by
_ that step
_-_ -~ - - - __-_ _--
I
~~~ - ~ --- -___-
Saturdav. Mav 10. 2003 12:08:47 PM Page 63 of 161
I
.. . .. 1 ... .~ ... . . ..
Saturday, May io, 2003 1%08:48
__-__.__. ~ __ PM Page 64 of 161
Given the following conditions:
- An Unusual Event is declared due to a radiological release.
- The Meterological Tower link to Hope Creek is malfunctioning.
- The link to Salem Generating Station is working properly.
Which one of the follow sets of data must be requested from Salem Station to be communicated to the States of New Jersey and Delaware with the Initial Contact Message Form (ICMF)?
Wind- Direction TO; Wind Speed 300 fi elevation.
Wind
- _ _ Direction
_ _ - TO; Wind Speed 33 ft elevation.
- - - - - 1 Wind Direction FROM: Wind-Speed 33 ftelevation.
d B
Emergency and Abnormal Plant Evolutions
,2950381:High Off-Site Release Rate Wind Direction TO; Wind Speed 300 ft elevation. Incorrect. Wrong elevation. Wind direction is FROM Wind Direction FROM- Wind Soeed 300 ft elevation. Incorrect. Wrona elevation. Wind weed is 33 ft elev.
V c h m e n tI I
I. . . . .. .
Saturday, May 10, 2003 12:08:48 PM
-- ~ ~~~~
I Pase 65 of 161
Given the following conditions:
- A severe accident has occurred.
- - A radiological release is in progress.
Which one of the following choice correctly fills in the blanks of this statement?
The Emergency Plan prevents from receiving radiation doses of Rem Whole Body and Rem to the Thyroid.
members
__ _ _ of the _public;
_ 25; 75
- - - -.- - ____ -1 station
___ personnel; 75;200
- -- - - -7 station- dersonnel; 25;75
- 1 members of- the public; 25; 200 - - _ _ ---- - 1 members of the public; 25; 200
,members of the public; 25; 75 station personnel; 25; 75 h station personnel: 75: 200
, conditions IAW the Student Handout.
1 ...... . . ... I
~... -. . . . . . .
I None
- ___._ ~~~
Saturday, May 10,2003 12:08:48 PM Page 66 of 161
- -~ -
Which one of the following describes potential consequences of the failure to place the Containment Hydrogen Recombiners in service at the proper hydrogen concentration?
Increases threat to containment integrity caused by high temperature. 7
- high internal pressure. I ~~
high drywell to suppression chamber differential pressure.
low internal pressure.
1 EKI. Knowledge of the operational implications of the following concepts as they apply to HIGH CONTAINMENT HYDROGEN CONCENTRATIONS:
will lead to containment failure by high internal pressure.
high temperature. Incorrect. Will damage internals but not design basis high drywell to suppression chamber differential pressure. Incorrect. Function of the vent header and downcomer pipes.
low internal pressure. Incorrect. Internal Dressure will increase.
I ~ - ~ _ _
i- Saturday, May 10, 2003 i2:08:48 PM i Page 67 of 161
Given the following conditions:
- The plant is operating at 25% power performing a startup.
- Control rod 18-23 has been determined to be stuck at position 00.
- While attempting to withdraw the control rod, indicated drive water flow is reading "0"gpm.
Which of the following is the cause of this indication?
Hydraulic__Control
- - - Unit- Directional
- ___ Control
___ Valve- (I 22) has failed to reposition.
The 2 gpm Stabilizing Valve has failed to reposition. I Both Cooling Water Header to Exhaust Header-PressureEqualizing Valves have faied open. 1 The Drive Water Header Pressure Control Valve has failed closed. 1 K3. ] Knowledge of the effect that a loss or malfunction of the CONTROLROD DRIVE HYDRAULIC SYSTEM will have on following:
1 X3.03
[Control rod drive mechanisms 13.1 [ 3.21 Hydraulic Control Unit Directional Control Valve (122) has failed to reposition.- Correct- IAW M-47-1 SV-122 is the withdrawal solenoid and SV-120 is the exhaust solenoid The 2 gpm Stabilizing Valve has failed to reposition.-Incorrect- this would effect total system flow not withdrawal flow Both Cooling Water Header to Exhaust Header Pressure Equalizing Valves have failed open-Incorrect-this would allow the exhaust header pressure to rise above cooling water pressure but not effect flow to the drive The Drive Water Header Pressure Control Valve has failed closed.-Incorrect- this would remove all flow to the system (R) Given P&IDs M-46-1 and M-47-1, determine the flowpath of the Control Rod Drive Hydraulic System, including the following flowpaths:
- a. Drive Water
- b. Cooling Water
- c. Exhaust Water
- d. Charging Water
- e. Scram
- f. Seal Purge for Recirculation Pumps
- g. RPV Level Reference Leg Backfill i..
Saturday, May 10, 2003 12:08:48 PM
~~~ ~ ~~
7 I
A LOCA has occurred and reactor vessel water level is -140 inches. I I
I Which of the following describes the steps necessary to restart a CRD pump? - i v
Close the non-I E circuit breaker by depressing the CLOSE pushbutton and close the 1E circuit breaker by depressing the CRD pump START pushbutton.
Depress the LOCA OVERRIDE pushbutton, close the non-I E circuit breaker by depressing the CLOSE pushbutton, and close the 1E circuit breaker by depressing the CRD pump START pushbutton.
Depress the LOCA OVERRIDE pushbutton, close the 1E circuit breaker by depressing the CLOSE pushbutton, and close the non-I E circuit breaker by depressing the CRD pump START pushbutton.
Close the IE circuit breaker by depressing the CLOSE pushbutton and close the non-I E circuit breaker by depressing the CRD pump START pushbutton.
C 8 [Hope Creek 201001K605 ,
50:
K6. 1 Knowledge of the effect that a loss or malfunction of the following will have on the CONTROL ROD DRIVE HYDRAULIC System:
~-K6.05:. 1A.C. power 1 3.3 [ 3.31 Depress the LOCA override push-button, close the IE circuit breaker by depressing the close push-button, and close the non-I E circuit breaker by depressing the CRD pump START push-button. -Correct-
'V CRD pumps are load shed on a LOCA signal and requires both the 1E and Non 1E breakers to be closed Close the non-I E circuit breaker by depressing the close push-button and close the 1E circuit breaker by depressing the CRD pump START push-button.-Incorrect- has the power supplies titles backwards, also requires LOCA override Close the 1E circuit breaker by depressing the close push-button and close the non-I E circuit breaker by depressing the CRD pump START push-button.-Incorrect- LOCA signal is present and will require LOCA override Depress the LOCA override push-button, close the non-I E circuit breaker by depressing the close push-button, and close the 1E circuit breaker by depressing the CRD pump START push-button.-Incorrect-has titles of breakers backwards I
I HC.OP-SO. BF-0001 1 1
and P&ID M-46-1, explain the actions necessary to place in service, or shift, the following components, including specific plant locations, IAW HC.OP-SO.BF-0001:
CRD Pump Suction Filter CRD Drive Water Filter CRD Flow Control Station
- e. Stabilizing Valves _ _ ._
~-__
~~
OABI35E003 (R) Discuss the operational implications of the abnormal indications/alarms for system operating parameters Blackout/Loss Of Offsite Power Diesel Generator Malfunction, Abnormal Operating Procedure. -~ ~ ~~~ ~
None I
Saturday, May I O , 2003 12,08:48
~ ~~ ~~ ___PM
~~~~ ~-
Saturday, May I O , 2003 12:08.48 PM
Given the following conditions:
- A central Local Power Range Monitor (LPRM) detector at "C" elevation is providing signals to an
- Average Power Range Monitor (APRM) Channel and a Rod Block Monitor (RBM) Channel.
- The LPRM has just failed downscale with an adjacent rod selected.
Which one of the following describes the effect of the failure on the associated APRM and RBM channels?
The LPRM input:
will be automatically bypassed and removed from the RBM only. The APRM and the RBM readings will be lower than actual. - - ___ - --. - - -_ -
-I will be automatically bypassed and removed from both the APRM and RBM. The APRM-and --
_RBM
_ readings will NOT - be affected.
- - - -- - 1 will be automatically bypassed and-removed from the APRM only. The APRM reading will NOT 1
be affected and _ the
_ _ RBM reading will be - lower
- than actual. - - ___ - - -. J will NOT be automatically bypassed 6 the APRM or the RBM. The APRM and RBM readings -1
__ - - - -J
.will be lower than- actual.
d a -.
Comprehension [Hope Creek
[Plant Systems
_ _ E 201002 1 Reactor-ManualControl System 1 L'
AI. ] Ability to predict-andlormonitor changes in parameters associated with operating the REACTOR MANUAL CONTROL SYSTEM controls including:
A10.5- ___
13.4 The average including the failed LPRM will be lower than actual. Once the LPRM card-is bypassed, then the LPRMs are averaaed and will read normallv.
~
Rod Worth Minimizer Neutron Monitoring System Rod Block Monitor System Mode Switch Refueling System i Refueling Bridge Refueling GrapplelHoists PGificantly Modified___ __
I 1
Saturday, May _ _ 10, 2003 12:08:49 PM -
. .___~____
Saturday, May 10, 2003 12:08:49 PM
~~
Page 72 of 161
~
Given the following:
- ,- Reactor power is 83%.
'- Neither RBM is bypassed with the joystick.
- Rod 30-31 has just been selected.
Use the attached figure of the 4-Rod Display for LPRM indications (Ribbon readings are approximates)
,Assuming all other LPRMs are operable, which of the following describes the operability status of the RBM CHANNEL A and CHANNEL B?
A- Operable;__ B- Operable . - _ _ - ___ __ -
A- ODerableT B-Inoperable --- 1 A-- Inoperable;
. - -- - - B- Inoperable d SRO 55.43 (2) Facility operating limitations in the Technical specifications and their bases.
Technical Specification interpretation IAW SH.OP-AP.ZZ-0108 Exhibit 3 which requires at least 50%
LPRM inputs for each level operable.
A- Operable; B- Inoperable Correct- Only 1 of 4 LPRM for D LPRM Level makes RBM B Administratively inoperable.
A- Operable; B- Operable Incorrect- RBM B is inoperable.
A- Inoperable; B- Operable Incorrect- RBM A is operable; B is inoperable.
A- Inoperable; B- Inoperable Incorrect- RBM A is operable.
\SH.OP-AP.ZZ-0108 Exhibit 3
- . 1 cable to the Rod Block Monitor (RBM) System.
ystem operability and determine required actions based upon system inoperability.
I Specification items associated with the Rod Block Monitor (RBM) System. (SRO Only)
~ ~ -- ____
Saturday, May I O , 2003 12:08:49PM 1 Page 73 of 161
_______ -~
Given the following conditions:
- The plant is operating on the 100% Rod Line at 100% power.
u - 'B' Recirculation Pump trip occurs due to inadvertant bump of the Drive Motor Breaker local trip switch.
,- Reactor settles at 56% power.
/- Recirc Loop A Flow (FLR611A) is 38 Mlbmlhr.
- Recirc Loop B Flow (FI-R611B) is 3 Mlbm/hr.
,- Core Flow (FR-R613) is 35 Mlbm/hr.
I- There are NO indications of thermal hydraulic instability.
- - HC.OP-AB.RPV-0003, Recirculation System is entered.
!- HC.OP-AB.RPV-0002, Reactor Power Oscillations is entered.
I
'Which of the following action(s) is(are) required per HC.OP-AB.RPV-0002, Reactor Power
'Oscillations?
Raise 'A' Recirculation Pump speed until total core flow is above 45%. I Reduce 'A' Recirc flow andinsert control rods IAW Stuff Sheet: 1 Insertcontrol rods IAW Stuff Sheet.
i/
those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
Raise 'A Recirculation Pump speed until total core flow is above 45%. Inorrect. Would be response if in the Exit region. Response to operating in the exit region is either inserting rods OR raising core flow with the running recirc pump.
Lock the Mode Switch in the Shutdown position. Correct. In the Scram Region of new power to flow map.
Reduce 'A Recirc flow and insert control rods IAW Stuff Sheet.. Incorrect. Insert rods by scram.
Insert control rods IAW Stuff Sheet. Would be correct if in the exit reqion.
~- - ___
New Power to Flow Ma W
Saturday, May 10, 2003 12-08:49PM
Saturday, May IO,2003 12.08-49PM Given the following conditions:
1
- The plant is operating at 100 percent power with all systems normal.
+ '- LPCl Channel B receives a LOCA Level 1 initiation signal.
i
- BVH-210 RHR Room Cooler is in Auto Lead.
8- FVH-210 RHR Room Cooler is in Auto.
,- CRIDS Point D3122 RHR Room Cooler Low Flow alarm occurs 5 minutes later.
- B RHR Room temperature is 1I O degF.
Which one of the followina describes the status of the RHR Room Coolers in B RHR Room?
- Application
- C IHope Creek 203000K110
- -54' KI, ] Knowledge of the physical connecfions and/or cause- effect relationships between RHFULPCI: INJECTION MODE and the following:
IECCS KI .IO- room coolers 1-32 [ 3.21
- B running; F running. Incorrect. B cooler would be tripped.
B running; F NOT running. Incorrect. B cooler would be tripped. F cooler would be running.
B NOT runnina: F NOT runnina. Incorrect. F cooler would be runnina.
I_HC.OP-SO.GR-0001
_ - 3.3.5
_ . _ _ . - J
/H-83 Sheets 5 and 11 1 therein, explain the function of the supporting system, IAW the RHR System Lesson Plan. I RHRSYSE012 I the status of the Residual Heat Removal System or its components by evaluation of the controlslinstrumentationlalarmsIAW the RHR Svstem Lesson Plan H-83 sheets 5 and II .-
~
Saturday, _May
_ 10,2003 12:08:49 PM -
1 Page 76 of 161
~-
Given the following current conditions:
- The plant is operating at 95% power.
-, - The RWCU system has just been returned to service.
- The "A' RWCU pump is running.
- The "B" RWCU pump is C/T.
- RWCU return to Feedwater temp CRIDS pt A215 is reading 410F.
Based on the above conditions, what flow is the maximum allowable flow for long term operation on RWCU return to Feedwater flow CRIDS pt A2856?
___ - -- - - - - - __ - - -- _._-_ - ~ -- - ..
173 !JPm a Application IHope Creek
,204000K103 ,
55' K1. ]'Knowledgeof the physical connections and/or cause- effect relationships between REACTOR WATER CLEANUP SYSTEM and the following:
K1.03 [Reactor feedwater system 13.1 13-11
\,
1HC.OP-SO.BG-0001 Attachment 1 & 2 for ILOT use I RWCUOOE003 ~
(R) Given the necessa6 sheets of P&IDs M-44-1 and M 1:
~-
Saturday, -~
~
Given the following conditions:
- The plant is operating at 100% reactor power.
- HPCl Pump Inservice test is in progress at rated flow.
- HPCl discharge pressure is 1150 psig.
- While attempting to adjust pump flow, the flow controller setpoint remains stationary at 4000 gpm in AUTO.
- The PO reports the HPCl flow controller works in MANUAL and develops rated flow.
What effect does this have on HPCl Operability at the PRESENT time?
HPCl is operable because it can develop rated flow. -1 HPCl is "operable but non-conforming". because it is NOT capable of meeting allsurveillance requirements.
HPCl is "operable but degraded" because it has lost testing capability.
HPCl-is inoperable because it is NOT cap-able of meeting all surveillance reauirements.
/d s_i iant Systems 206000A402 1- High Pressure Coolant Injection System 1 56 Justification:
b SRO 55.43 (2) Facility operating limitations in the Technical specifications and their bases.
HPCl is inoperable because it is NOT capable of meeting all surveillance requirements. Correct. HPCl must be in AUTO with a setpoint of 5600 gpm and capable of rated flow and discharge pressure.
HPCl is operable because it can develop rated flow. Incorrect. HPCl must be in AUTO with a setpoint of 5600 gpm.
HPCl is "operable but degraded" because it has lost testing capability. An Operable but degraded case could be made if the setpoint was stuck at 5600 gpm.
lHPCl is "operable but non-conforming" because it is NOT capable of meeting all surveillance
'requirements.
- - . _ _ but non conforming Operable is not
___ - _._.-. applicable.
IHC.OP-SO.BJ-0001 I Select those sections which are applicable to the HPCl System IAW HCGS technical specifications.
Evaluate HPCl System operability and required actions based upon system operability IAW HCGS technical specifications.
(SRO Only)
I Explain the bases for those technical specification items associated with the HPCl System IAW HCGS technical specifications.
- (SRO Only) 1 v -~ -
__ _. ~
'VISION QlMC Q55949 significantly modified
~-
- _Saturday,-May
_ IO,2003 12:08:49 PM 1 Page 78 of 161 L--_____
~~ Saturday, May 10, 2003 12:08:50 PM - ' Paqe 79 of 161
From the list below, select the choice which is LOWEST in priority for use as reactor pressure control as described in HC.OP-IO.ZZ-0007 OPERATIONS FROM HOT STANDBY section Maintaining Hot Standby (MSIV's Open).
v I. Main Steam Line Drains.
II. RClC or HPCl Steam Line Drains.
Ill. RClC or HPCl in Full Flow test.
IV. RFPT's min flow oDeration.
Ill. -1
-0007 Note 5.2.5 directs the descending order of priority as follows.
Main Steam Line Drains; RFPT's min flow operation IAW system operating procedure, RClC or HPCl
,Steam Line Drains, RClC or HPCl in Full Flow test.
I l l - Correct. RClC or HPCl in Full Flow test is the last of the listed systems.
I - Incorrect.
II - Incorrect.
IV - Incorrect.
JHC.OP-1O.Z-0007Note 5.2.5 I-- . - - - --.
. - I 1
PI Operating Procedure, IAW this Lesson Plan I
/None Saturday, May 2003 12'08:50 PM I Page 80 of 161
-~ ~
Given the following conditions:
- A LOCA occurred.
u
- RPV level has been stabilized above TAF.
- Drywell sprays are in service.
- Core Spray pump 'A' amp indicator begins to fluctuate.
Which one of the following would cause the fluctuation and what action is permitted that would remedy the condition IAW HC.OP-AB.ZZ-0155 Degraded ECCS Performance?
Clogging
_ _ ofthe __ suction
- strainer; reduce
- loop
- flow.
- - 1 Partial closure of the
__ injection valve; manually open the injection valve
__ - fully. __
- -1 Clogging of the suction strainer; throttle 'A' Core Spray Pump manual discharge valve.
I Partial closure _ _ _ _ of _the_ injection
- _ _ valve; stop the- 'CCore
_ . _ - Spray pump. .- - - - - . -- ~ I 1 LUYUU-IALUD
~.
209001 I 'Low Pressure Core _ _
Spray System
-1
@._ Ability to (a)-predict the impacts of the following on the LOW PRESSURE CORE SPR (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
A206 ,
1Inadequatesystem flow _ _ __ - -
[3.2132!
Justification:
Clogging of the suction strainer; reduce loop flow. Correct. Fluctuating amps is a symptom of clogged suction strainers. Throttling loop flow directed by Attachment 2
/Partialclosure of the injection valve; manually open the injection valve fully. Incorrect. Partial closure is
'the remedy, not the cause.
Clogging of the suction strainer; throttle 'A' Core Spray manual discharge valve. Incorrect. Would reduce pump flow but not in accordance with AB.=-01 55.
Partial closure of the injection valve; stop the 'C' Core Spray pump. Incorrect. Stopping C pump would increase flow throuah A DumD and make the conditions worse.
I . ...
0AB155E001
~-
OAB155EO03- T i
1 degraded ECCS Performance/Loss Of NPSH, Abnormal Operating Procedure.
____ 7---
Saturday, May 10, 2003--12.0850 PM _.-_ Page 81 of 161_ _
~_
~
~~~
~~~
Given the following conditions:
- Drywell pressure increased to 2 psig, v
- Off-site power is lost.
i Which of the following describes the start sequence for the core spray systems after off-site power i1 was lost?
~
Core Spray pumps "A"and "B" start immediately after the diesel generator output breakers are closed. Core Spray pumps "C" and "D" start six seconds after the diesel generator output breakers are closed.
Core Spray pumps "A" and "C" start immediately after the diesel generator output breaker is closed. Core Spray pumps "B" and "D" start six seconds after the diesel generator output breakers are- closed.
Core Spray pumps "A", "B", "C", and "D" start immediately after the diesel generator output breakers are closed.
Core Spray pumps "A',"B", "C", and "D" start six seconds aker the diesel generator output breaker is closed.
d i Comprehension [Hope Creek
[Plant Systems 1 ! c3 209001K201 2 209001 I Low Pressure Core Spray System 1 _ _ - 59 L/
Core Spray pumps "A", "B', "C", and "D'start six seconds after the diesel generator output breaker is closed. Correct. With a LOP, all pumps start 6 seconds after the edg output breaker closed.
Core Spray pumps "A" and "C" start immediately after the diesel generator output breaker is closed.
Core Spray pumps " B and " D start six seconds after the diesel generator output breaker is closed.
Core Spray pumps "A", "B', 'IC', and "Dl start immediately after the diesel generator output breaker closes.
I and any of the following, IAW the Core Spray System Lesson Plan:
- a. Residual Heat Removal (RHR) System
- b. Torus Compartment
- c. 4160 VAC Class 1E Distribution System u
- d. 480 VAC Class 1E Distribution System
- e. 125 VDC Class 1E Distribution System
- f. Nuclear Boiler
- 9. Liquid Radwaste System I
- h. Condensate Storage and Transfer System 1.
i.
Primary Containment Instrument Gas (PCIG) System High Pressure Coolant Injection (HPCI) System i
- k. Condensate Storage Tank I
E 1
Emergency Diesel Generators (EDGs) I W
Nuclear Boiler InstrumentationSystem I Standby Liquid Control (SLC) System -2 I
Saturday, May 10, 2003 12:08:50 PM 6 - 1
~~~ ._._
Given the following conditions:
- A Loss of Offsite Power (LOP) concurrent with an ATWS has occurred.
u
- The "A',"B", & "C" Emergency Diesel Generators are supplying their 4kv buses.
- Emergency Diesel Generator "D" will NOT start.
- The CRS has ordered that the Standby Liquid Control System be initiated.
'Which one of the following describes the components of the Standby Liquid Control System that
_ ~~
-1
_ _ . - - - --I -
pumps and both squib valves. Correct. The "A" pump and squib valve are powered from "A" EDG. The "B" pump and squib valve are powered from the "B' EDG. The "Dl DG supplies power to the SLC isolation valve F006B which is normally open and remains open on a loss of power.
NEITHER SLC pump nor associated squib valve. Incorrect.
SLC pump A and squib valve F004A ONLY. Incorrect.
SLC pump. . B and squib valve F004B ONLY. Incorrect.
HC.OP-SO. BH-0001 sLcsysEY 1 I
(R) From memory, summarize/identify the impact that a loss or malfunctionof each of the following would have on the Standby Liquid Control System I.A.W. the Lesson Plan.
a.
b.
c.
Standby Liquid Control Squib Valve Standby Liquid Control Storage Tank Level Redundant Reactivity Control System I
_~
_ May I O , 2003 12:08:50 PM Saturday, ~ . _ _ _ _ ~
~
I Page84of161 - _ _
-~
Manipulating which one of the following components ensures the Squib valves do NOT fire and the RWCU system remains in operation during testing of the Standby Liquid Control pumps?
Closing SLC Iniection valves F006A and B.
u Opening the breakers for Fool and F004.
- Bezel keyswitches on 1OC651C console.
Local panel pump start switches.
211OOOK402 61 K4. 1 Knowledge of STANDBY LIQUID CONTROL SYSTEM design feature(s) and/or interlocks which provide for the following.
K4:02__ [Component __ and system testing _ _ -
I-3.0;~3.23 IJustification:
/Using local panel pump start switches. Correct. Local start switches prevent firing of the squib valves and lautomatic closure of the RWCU Fool and F004.
Opening the breakers for Fool and F004. Not driven by any procedure. Does not stop firing of the squib valves.
Bezel keyswitches on 10C651C console. Keyswitches function as a permissive for manual SLC initiation.
Closing SLC Injection valves F006A and B. Prevents injection to the RPV. Will not prevent RWCU isolation or Sauib valve firincl.
U r'
L .... ..... ..........
1 Page 85 of 161 Saturday, May__ IO, 2003_ _12:08:50PM a
I
During a failure-to-scram condition, which of the following indications would confirm that HC.OP-AB.ZZ-OOOO(Q), "Reactor Scram", should be exited and HC.OP-EO.=-01 OIA(Q) entered?
All Blue lights are lit on the Full Core Display.
L All HCU Accumulator Trouble liqhts are lit on the Full Core Display.
Computer Pt in Alarm Overhead Annuciator C&F5 extinguished.
All APRM "downscale" lights are illuminated.
A3. Ability
_ _ to
_ monitor
- automatic operacons of the
- REACTOR
__ - PROTECTION-SYSTEMincluding: - -!
A3.07 ISCRAM air header Dressure 13.6 t Air All Blue lights are lit on the Full Core Display. Indicates all scram valves open. This is the response for a normal scram.
[All HCU Accumulator Trouble lights are lit on the Full Core Display. This is the response for a normal kcram.
All APRM "downscale" liahts are illuminated. This is the resDonse for a normal scram.
[HC.0P-AB.- ZZ-000
- 1 Attachment
__ 1 __
IHC.OP-AR.ZZ-0011 Attachment F5 Overhead annunciator window C6 figure from HC.OP-AR.ZZ-0011 L-Saturday, May 10, 2003 12:08:51
- PM r-1:-_-
Page 86 of 161
Given the following conditions:
- The reactor has scrammed and the mode switch is in SHUTDOWN.
- The problem that caused the scram has been identified and corrected.
b
- Annunciator "CRD SCRAM DISCH VOL VVTR LVL HI" is sealed in.
Which one of the following describes the reactor protection system (RPS) response when you place the Scram Discharge Volume High Level Keylock switch in BYPASS, followed by taking the scram reset switches to RESET, and finally placing the mode switch in STARTUP for NI testing?
The RPS will- reset and remain reset.
The RPS- will -
reset and again
-- - - scram. ____
- -i Nothing will occur -_ due to the present plant conditions.
- 1 The RPS will
_ _ reset
- - when - the scram discharge volume drains.___ - - - _ _ - -1
- b Comprehension I \Hope Creek J 212000A404 I -- . - - __
Justification:
The RPS will reset and again scram. Correct. Placing the mode switch to Startup unbypasses the Bypass interlock and a scram occurs.
The RPS will reset and remain reset. Incorrect. Placing the mode switch to Startup unbypasses the Bypass interlock and a scram occurs.
Nothing will occur due to the present plant conditions. Incorrect. A scram occurs.
The RPS will reset when the scram discharge volume drains. Incorrect. The SDV will not drain due to the scram sianal oressent I .... ...... 1 I
I. . . . . . . . . .... ..... - .. . . . . . . . . . . . . ......... . .. i RPSOOOE017 Given a labeled diagramldrawing of, or access to, the Reactor Protection System controls, andlor alarms located in the Control Room:
Explain the function of each indicator.
Assess plant conditions that will cause the indicationsto light or extinguish.
None ,
1
- 7 .... 1
_. - ~~
~
Saturday, May 10, -_ 2003-12.08:51- PM
- The unit is at 100% power when the BD483 inverter output momentarily spikes to 137 volts and immediately returns to a normal regulated output of 119.5 VAC.
- The cause of this spike is unknown at this time.
I How will this transient affect the unit?
. .. .. .. ~~ .. ... . . . ..
%. scram from B APRM and a rod block from B RBM; NO other effects
- 1/2 scram and rod block from 6D and F APRMS and a- rod block from BRBM; NO other effects
_ _ - - _-. __ - .- - - -- - 1 Correct answer: % scram and rod block from B, D and FAPRMs and a rod block from 6 RBM; no I
other effects. High input voltage trips the EPA breaker on overvoltage above 132 volts.
The following distractors are incorrect as follows:
e % scram and rod block from B and D APRMs and a rod block from A and B RBM; no other effects Incorrect - BD483 UPS powers B, D, and F APRMs, and BRBM, no mention of F APRM and mention of ARBM
.% scram from B APRM and a rod block from B RBM; no other effects Incorrect- BD483 UPS powers B, D, and F APRMs, no mention of D or F
% scram and rod block from D and FAPRMs; no other effects
- Incorrect- BD483 UPS Dowers B. D. and F APRMs. and B RBM, no mention of BAPRM or B RBM IHCGS Tech Spec 3.8.4.6 I I I Saturday, May I O , 2003 12.08:51 PM -61
.~
Given the following conditions: -
- IRM C reads 5 on Range 6.
- The range switch was placed to Range 5 and then Range 4.
- Which of the following describes the resulting IRM C indication?
'(Refer to the attached figures of the IRM controls and indications.)
5-on range 5, and off-scale on range 4 - -1 C Comprehension ,
i215003 1 Intermediate Range Monitor (IRM) System re arranged with the odd scale ranges are from 0-40, and the even scale range switches are 0-125. Going from the odd ranges to the even ranges expands the scale from 0-40 to 0-
,125. Then going from the even to the odd ranges will increase the magnitude of the scale by a factor of IO.
/NOH01IRMSYS-00 I IRMSYSE009 I (R) Given a set of conditions and a drawing of the controls, instrumentation and/or alarms, or access to, the control room, evaluate the status of the IRM controlslinsthmentationlalarms IAW control room procedures.
IRMSYSE004 (R) Given a labeled diagram of, or access to, the IRM controls/indication bezel:
Explain the function of each indicator Assess the plant conditions that will cause the indicator to light or extinguish Predict the effect of each control on the IRM System Select the condition or permissives required for the control switches to perform their intended function.
IAW control room procedures
. - ~ ~ - -
Saturday,
_ _ May 10, 2003 12:08:51 PM 1 Page 89 of 161
Given the following conditions:
- Reactor startup in progress.
- IRMs are on range I O and reading 25.
- Reactor Mode switch is in STARTUP/HOT STANDBY.
- IRM C fails downscale.
I Which of the following lists rod block status with the present condition AND if the Reactor Mode switch is placed in RUN?
A Rod Bbck exists. The - - __Rod
___ __ B1ock will clear -- afterplacing the Reactor Mode switch -
- -in
- RUN.
NO Rod Block exists. Placing Reactor Mode switch in RUN will NOT result in a Rod Block.
NO Rod Block exists.-Placing the Reactor-Mode switch in RUN will result in a Rod Block. 1 A Rod Block exists. The Rod Block will NOT clearafter placing the Reactor Mode switch in RUN. _ _ _ _ .__ _ _ - . - __
'a Comprehension IHope Creek K1.
__ 1 Knowledge of the physical connections andlor cause- effect relationships between INTERMEDIATE RANGE MONITOR (IRM) SYSTEM and the following:
KI .O2 , /keactor manual control 1 3.613.61 i/
Downscale Rod Block clears after Reactor mode Switch is placed in Run.
NO Rod Block exists. Placing Reactor Mode switch in RUN will NOT result in a Rod Block. Incorrect. A rod block will exist based on the IRM downscale.
NO Rod Block exists. Placing the Reactor Mode switch in RUN will result in a Rod Block. Incorrect. A rod block will exist based on the IRM downscale.
A Rod Block exists. The Rod Block will NOT clear after placing the Reactor Mode switch in RUN. The rod block will clear after placing Mode switch to run. ... . . .
IHC.OP-SO.SE-0001
___ - . - __ - I
, evaluate the status
.-..of -the IRM
. controlslinstkmentationlalakis IAW control room procedures.
- - - ___ _ _ _ . - . ' I 1 -
'None Saturday, MayfO, 2003 12:08:51 PM i__~ ~
' Page 90 of 161
Given the following conditions:
- The Mode Switch is in the STARTUP/HOT STANDBY position.
i, '- APRM "D" is indicating 1%.
- Reactor power is approximately midscale on Range 7 of the IRMs.
- Recirc Pump speeds are at minimum.
- APRM "C" begins to fail upscale.
Predict the RPS response.
~ ___
when "C" APRM reads 51%.
- - -- 1 Half scram when "C" APRM reads 15%. I 1
'Full scram when "C" APRM reads 51%.
-AI=- ] Ability to predict andlor monitor changes in parameters associated with operating theAPRM/LPRM controls including:
/ 379 [ 4.01 AI .02 [RPS status Half scram when "C" APRM reads 15%. Correct. Half scram on RPS A when C APRM reaches 15 percent with the mode switch in Startup.
Half scram when "C" APRM reads 51% Incorrect. 15 percent with the mode switch in Startup.
Full scram when "C" APRM reads 15% Incorrect. Half scram only.
Full scram when "C" APRM reads 51% Incorrect. Half scram only. 15 percent with the mode switch in Startup.
i HC.0P-SO.SE-0001 I J APRMOOE009 the Student Handout.
- - _ _ - _ _ - 1 (R) From memory, IAW Technical Specifications,determine the rod blocks andlor scrams initiated by the APRM System, IAW
'None
~~~~~ ~
Saturday. May 10, 2003 12:08:51 PM j m -
Given the following conditions:
- A plant startup is in progress with power at 20%.
- Recirculation flow is 30%.
- The "A'APRM Flow Unit output remains at 30% as recirculation flow is raised.
As the plant startup continues, what will be the FIRST protective action to occur and the reason for that action?
A full scram will occur due to flow-biased neutron flux uescale.
-A -
control
- __ rod- block -
will occur due to flow biased __ neutron flux upscale. - - ___ - --- -- -- .I
- -1 :
!A-half scram will occur due to a flow unit "inop" signal.
~ - -- - - - - - - -
A control rod block will occur due to a flow unit comparator trip. I Average Power Range Monitor/Local Power Range Monitor System IJustification:
A control rod block will occur due to a flow unit comparator trip. Correct. The comparator trip will alarm at
,I 0 percent difference between A and B or A and C flow units.
,A control rod block will occur due to flow biased neutron flux upscale.lncorrect. FBNF RB trip is set at 42 percent APRM power with the mode switch in Run.
v A half scram will occur due to a flow unit "inop" signal. Incorrect. Flow unit inop trip by itself does not cause half scram.
A full scram -will- occur ___ due to flow biased - _ _neutron flux upscale.
- - Incorrect
- I IHC. OP-SOSE-0001 Explain the function of each indicator, IAW the Student Handout.
Assess the plant conditions that cause each indicator to light or extinguish, IAW the Student Handout.
I I
Predict the effect of each control switch on the APRMSlFlow Units, IAW the Student Handout. I Select the conditions or permissives required for the control switches to perform their intended function, IAW the Student Handout. . . ___ ~ ___ - - . . . . . . .
I I
Saturday, May 10, 2003 120852 PM Paqe 92 of 161 i~~ ~~~
- .-- - ~ _ _ _ ~. -
Given the following conditions:
- The plant has been manually scrammed.
,-, - A normal reactor cooldown is in progress.
- The reference leg backfill system is out of service.
Then, annunciator (A7-C5) RPV LEVEL 4 is received. The operator investigates and observes that reactor water level notching is occurring.
Which of the following is the most accurate indicated water level from the indicator that is experiencing notching?
An average ofthe water levels from the top AND bottom of the notch.
The water level at the bottom of the 'botch'.
- -1 The water level at the top of the notch: - 1
~
C Memory IHope Creek AI. 1 Ability to predict andlor monitor changes in parameters associated with operating the NUCLEAR BOILER INSTRUMENTATION controls including:
,AI.04 [System venting 12.6-/2.8]
b
[NOH01RXINST-01
. . - , Industry events. 1 i 1 Discuss the root cause of the plant problem/industry event IAW the associated plant problemfindustryevent document.
Discuss the HCGS design andlor procedural guidelines that mitigatelreduce the likelihood of the problemlindustry event at HCGS IAW the associated plant problemlindustry event document.
Discuss the
__ lessons- learned
__ from
_ _ the problem/event IAW the associated plant _problemhndustry
_ event document.
c
-_ _ _ _ ~ _- ~
___ Saturday,
_ _ _ _May
~ 10, 2003
_ _120852 PM Page 93 of 161
Given the following conditions:
- The plant is operating at 30 percent power.
- - The I&C department reports that reactor pressure transmitter SA-PT-N403B on instrument rack C027 has failed it's sensor calibration.
- I&C also states that the pressure transmitter must be replaced.
Based on these conditions, declare that the associated:
channel must be Dlaced-inthe trimed condition within one hour. --- -1
-- -- - - - _-. ___. - ~
channel must be placed in the tripped condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, --1 system is inoperable and returned tooperable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
channel must be placed in the tripped condition within twelve hours. - - 1 a s
~
12160006222 1216oOo]Nuclear Boiler Instrumentation - - .
70.
55.43(2) Facility operating limitations in the Technical Specification and their bases.
channel must be placed in the tripped condition within one hour. CORRECT. Operator must determine the transmitter feeds RRCS Logic from M-42-1 Sht 1 & 2. The operator then determines LCO 3.3.4.1 ATWS RPT action b. is applicable.
channel must be placed in the tripped condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. INCORRECT. Action for one NSSSS transmitter not common to RPS.
system inoperable and returned to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. INCORRECT. Action for one trip system inoperable.
channel must be placed in the tripped condition within twelve hours. INCORRECT. Action for RPS pressure transmitter.
Requires TS section 3.3.4 M-42-1SHT 2 REV 14 TECH SPEC 3.3.4.1.B
!Tech SDec 3.3.4
' a. Choose those sectionswhich are applicable to the Redundant Reactibity Control System (ATWS Circuitry), IAW HCGS 1 Technical Specifications.
- b. Evaluate Redundant Reactivity Control System operability and determine required actions based on system inoperability (ATWS Circuitry), IAW HCGS Technical Specifications. SRO/STA ONLY
- c. Explain the bases for those technical specifications associated with the Redundant Reactivity Control System (ATWS IAW HCGS Technical Specifications..
-~ _ _ _ 2 _
I i ~
I ~ - -
._QUestSOn %Urce: Facility Exam Bank Tech
.~
Spec sections_ .
3.3.1 .through 3.3.4;~-P&ID M-42-1 Sheet __
1 Significantly Modified Saturday, May -
10, 2003
~
120852 PM
~
Vision Bank Q I W Q54823 Significantly modified.
~ ____ .~ _ _ _
Saturday, May I O , 2003 12:08:52 PM -.
Page 95 of 161
Given the following conditions:
- The plant has scrammed following extended operation at 100% power.
- - The MSlVs are closed.
- The RClC system is being operated to maintain RPV water level constant.
- RClC flow controller BD-FIC-R600 is in AUTO with flow set at 600 gpm.
- Reactor water level is steady when the control oil supply line to RClC Turbine Governor Valve BD-HV-4283 ruptures.
Which one of the following describes the initial plantlsystem response to this line break?
RClC- pump
_ _ _ - discharge-flow indication__ (BD-FIC-R600)
__ - - decreases, RPV water level decreases.
RClC turbine steam inlet pressure indication (BD-PbR602) increases, RPV water level increases.
RClC pump discharge pressure indication (BD-PI-RGOI) decreases, RPV water level
. __ - - _ -- - -i decreases.
malfunction of the REACTOR CORE ISOLATION COOLING
,SYSTEM (RCIC) will have on following:
c K3.01 (Reactorwater level Justification:
'RCIC turbine speed indication (BD-SI-4280-1) increases, RPV water level increases. Correct. Loss of oil I
pressure causes the governor valve to fail full open. Turbine speed rises, flow increases, RPV level rises.
RClC pump discharge flow indication (BD-FIC-R600) decreases, RPV water level decreases. Incorrect.
Flow and RPV level rises.
,RCIC turbine steam inlet pressure indication (BD-PI-R602) increases, RPV water level increases.
Incorrect. Steam pressure will stay the same or lower due to increased steam flow.
RClC pump discharge pressure indication (BD-PI-RGOI) decreases, RPV water level decreases.
hcorrect Discharae Dressure increases. Level increases. I I
system, IAW the RClC System Lesson Plan:
- a. A given valve opening or closure
- c. Inadequate system flow
- d. An oil system malfunction
- e. Failure of the RClC Gland Seal Condenser Vacuum Pump
- f. Loss of room cooling
- g. Rupture disc failure on the RCIC exhaust
- h. Steam line break
-. ~- -
Saturday, May Page 96____
of 161
- ____ ~-
Saturday, May 10, 2003 12:08:52 PM Page
~- _ of 161 97 ~ _
'Given the following conditions:
- The Reactor Core Isolation Cooling (RCIC) system flow controller has failed full downscale demanding a "0"gpm flowrate.
- The controller is in AUTO.
Which of the following is the RClC turbine response upon receipt of a valid initiation signal for the given conditions?
I
'RCIC will start, accelerate.. .
to and run continuously at_ approximately 4000 rpm. _ -
and
- trip- on mechanical _ _ _ overspeed.
- -1
$hen
__ slow to a stop. -.
- -_ _ _- 1 then will slow to and_____
~ ._
run__ at a __ low- speed.
I d Comprehension [Hope Creek
~
The ramp generator runs RClC to about 4000 rpm until the flow signal comes out of saturatio
'v time the low sianal will control -> RClC runnina at min meed. I I. ..
J r-(R) Given RClC turbine control system failures, evaluate and determine the effect on the RCI I
_ _ ~ ~
Saturday,-May 10, 2003 120852 PM-. L P a g e 98 of 161
A plant transient is in progress with current plant conditions as follows:
- - Drywell Pressure is 3.6 psig and rising at 0.2 psi/min.
- Reactor Level is -35" and lowering at 1.5 in./min.
- Reactor Pressure is 810 psig and lowering at 10 psi/min.
- HPCI Pump is tagged for maintenance.
- All other ECCS systems have performed as expected.
'Assuming NO operator action, ADS SRVs will open immediately when:
I
~
- Level 1 is reached. ......... .___............ ............ _.___ ......... -- .....
........ ...... .-.~ _ .- .... .......... .- ...... ............. ..........
Level Iis reached and the 105 second timer times out.
Tor, of Active-Fuel (TAF) is reached and the IOSsecond tiher times out. -- I will wait until Level Iis reached to start the 105 second timer. After 105 seconds, ADS will open the ADS SRVs L
Level 1 is reached. Incorrect. Starts the 105 sec timer.
Top of Active Fuel (TAF) is reached. Incorrect. Level at which ADS is manually actuated if ECCS is available and running.
__ of Top - Active - -Fuel
- - (TAF) is _-- - and the 105 second
- -reached - __ timer
_ _ _ times out. Incorrect. Wrong
- ___ _- -setpoint.
ADSSYSE007 -- -- ~
(R) Given a set of conditions and a drawing of the controls, instrumentationand/or alarms located in the Control Room, identify the status of the Automatic Depressurization System by evaluation of the controlslinstrurnentationlalarms,IAW the Automatic Depressurization System Lesson Plan.
- ~ ~ _ _ -
___ Saturday, May 10, 2003 12:08:52 PM -- E99 of 161
Given the following conditions:
- The plant has scrammed due to a loss of offsite power.
L - HPCl and RClC fail to start both automatically and manually.
- RPV water level lowers below -129.
- The ADS CHANNEL INITIATION PENDING annunciators for both logic channels are received.
- The RO is directed to Inhibit ADS.
- The operator inadvertently arms and depresses the LOGIC B MAN INIT and LOGIC F MAN INIT pushbuttons.
Select i,ADS the statement below which describes the resDonse of ADS.
-will
- initiate immediately, _ _ regardless of core spray -
- and RHR status.
-1 ADS-will initiate in 105 seconds, only if Core Spray pumps A & C or RHR pump A or C are running. _ _ - -
_ I ADS willinitiate immediately, only if Core Spray pumps B & D or RHR pump B or D are
_running.
ADS will initiate-in 105 seconds, reaardless of Core Spray and RHR status. I a
~
/HopeCreek IPlant S y
s t e m s E- .
D 218000K402 1-0- - 1 Automatic Depres&$ation System 1 74
. ADS will initiate in 105 seconds, regardless of core spray and RHR status.- Incorrect, Manual initiation bypasses the 105 second timer
. ADS will initiate in 105 seconds, if Core Spray pumps A & C - or RHR pump A or C are running -
Incorrect, Manual initiation bypasses the 105 second timer I
IHC.OP-SO.SN-0001 section 3.3.1 i,
. - . - - - - -- _I Automatic Depressurization System Lesson Plan:
- a. Explain the function of each indicator.
- b. Assess plant conditions which will cause the indicator to light or extinguish.
i
_____ -~ _--
- Saturday,
-May ~ 10,2003
~-
12:08:53PM j-~
Page 100 of 161
Vision Bank QID# Q54168 editiorially modified.
_ _ _ ~ ~
Saturday, May 10, 2003 12:08:53 PM ' Page 101 of 161
Given the following conditions:
- B RHR pump is running in Suppression Pool Cooling at rated flow.
L, - B RHR pump trips on an electrical fault in the motor.
'Which one of the following describes the response of BC-HV-F007B Minimum Flow valve?
Automatically closes after a 10 second delay.
Automatically
__ __ - __-_ opens
- - . immediately
- - - when - the
- _ pump -
trips. - ___ __ - -- - - - - -- - -I Remains closed after the pump trip.
- -1 Remains oben after the pump trip.
-- I
~ C ,Comprehension Hope Creek
/Plant Systems E 219000A404 1- ' RHR/LPCI: Torus/Suppression Pool Cooling Mode - .___ ..
75 Remains closed after the pump trip. Correct. With the RHR pump breaker open, the Min Flow Valve HV F007B will remain closed. The RHR pump must be running for the valve to open on low flow.
Automatically closes after a 10 second delay. Incorrect. Sequence on pump startup with flow above 1250 gpm.
Automatically opens immediately when the pump trips. Incorrect. The RHR pump must be running for the
,valve to open on low flow.
'Remains open after the pump trip. Incorrect. Valve will be initially closed due to rated flow. Valve will
'remain closed.
IHC.OP-SO.BC-0001 3.3.6 I
/-
RHRSYSEOII I Given a labeled drawing of, or access to the Residual Heat Removal System controls/indication on 10C650:
7 a.
b.
c.
d.
Explain the function of each indicator IAW the RHR System Lesson Plan.
Assess plant conditions which will cause the indicators to light or extinguish IAW the RHR System Lesson Plan.
Determine the effect of each control on the RHR System IAW the RHR System Lesson Plan.
Assess plant conditions or permissives required for the control switcheslpushbuttons to perform their intended functions Saturday, May 10, 2003 12:08:53 PM
While performing a RHR system test, the breaker for Torus Spray Isolation Valve I
trips.
- Which one of the following Motor Control Centers (MCC) powers this valve?
1OB222 10B323 n
10B242 I
219000K201 e-. 1 Knowledge of electrical power supplies to the following:
12.4 10B222 Correct. Powered from 52-222083 which is a B channel 1E MCC.
10B323 Incorrect. B channel Non I E MCC powered from 1E power which is shed on a LOCA Level 1 signal.
108242 Incorrect. D Channel IE MCC with similar valve load to 10B222.
106563 Incorrect. B channel 1E MCC which Dowers Station Service Water components.
1- therein, explain the fundion o i the supporting system, IAW the RHR System Lesson Plan. ~
None
~~~-
Saturday, May 10, 2003 12:08:53PM
- ~
/0103of--
The Containment Hydrogen Recombiners aredirected to be placed in service following a L O C A f l ,
the drywell. LOCA conditions still are present.
- Which one of the following actions is the MINIMUM required to accomplish this task?
'Override NSSSS isolation and reset PClS isolation Override PClS isolation only.
Reset NSSSS isolation only.
Reset NSSSS isolation and override PClS isolation.
Override PClS isolation only. Correct. PClS isolation override is necessary to open H2 Recombiner flowpath Override NSSSS isolation and reset PClS isolation. Incorrect. NSSSS does not have override capability.
NSSSS provides isolation input to PClS isolation Reset NSSSS isolation only. Incorrect. Only half of input required to open valves.
Reset NSSSS isolation and override PClS isolation. Not minimum. NSSSs isolation reset is not necessarv
- a. Determine the source of electrical power for the NSSSS logic channels IAW the NSSSS Lesson Plan.
- b. Predict plant response to a loss of power to the NSSSS power supplies IAW the NSSSS Lesson Plan.
Saturday,
_ _ May 10,2003 12:08:53 PM -
1 Page 104 of76:
Which one of the following conditions would cause the alarm?
The A RHR...LOOPTestReturn Line manual isolation valve closed.
.- .... ..... .~.- - -1I a 226001A404 m' - . -
RHWLPCI: Containment Spray System Mode
__ 78 The CP228 ECCS Jockey Pump suction strainer clogged. Correct. A clogged suction strainer will cause jockey pump discharge pressure to drop. Low discharge pressure causes low alarm on PSL - N654A Low Discharge Pressure which in turn causes the A6-B1 alarm.
,The AP228 ECCS Jockey Pump tripped on overload. Incorrect. AP228 feeds the HPCI system.
'The A RHR Pump room is flooded. Incorrect. Causes RHR Pump Room Flooded alarm.
The A RHR Loop Test Return Line manual isolation valve closed. Incorrect. Would cause a high discharge pressure of the Jockey Pump but below the high pressure setpoint of 380 psig for detection of leakaae Dast LOOPIsolation valves.
1HC.OP-AR.ZZ-0004 Attachment B I L
-- 1 I
I ...
the status of the Residual Heat Removal System or its components by evaluation of the controls/instrumentation/alams IAW the RHR System Lesson Plan. L II .... ..... .
I
_~
Saturday, May 10,2003 12:08:53 PM j- Page 105 of 161
Given the following condkons:
- A LOCA occurred in the drywell.
- Drywell pressure is 10 psig.
- One of two Drywell pressure transmitters associated with this loop subsequently failed to zero psig.
- B RHR Injection Valve F017B is closed by the operator in preparation to spray.
Which one of the following describes the Drywell Spray Isolation Valve response when the operator is directed to place Drywell Spray in service?
~-- __ __ - _____- . -_ - ____ - - - _
Drywell
__ - __ spray - - inboard isolation - valve
__ - only
__ opens. _ - _ _ - - ~ ____
__ - __ - ___ _ _ _ ~ __ -
K11- ] Knowledge of the physical connections and/or cause- effect relationships between RHWLPCI:
CONTAINMENT SPRAY SYSTEM MODE and the following:
_Kl:O8 -1 (Nuclear boiler instrumentation 1 3.2[ 3.4 Justification:
d Both Inboard and Outboard Drywell spray isolation valves open. Correct. Only one DW pressure transmitter above setpoint is required to open both valves once the High Drywell pressure initially sealed in.
NEITHER Drywell spray valves will open. Incorrect. Both will open.
Drywell spray outboard isolation valve only opens. Incorrect. Both valves will open.
Drvwell smav inboard isolation valve onlv onens. Incorrect. Both valves will ooen.
\HC.OP-SO.BC-OOOI 3.3.5 ~
I L
I I . -. .... . . ... . ... . . ... . . . .. . ..
I precautions, and/or limitations during operation IAW HC.0P-S0.BC-0001.
~ _ -_ 7--
PM Page 106 of 161
~ ~ _ May IO, 2003 12:08:53
_ _ Saturday, ____
Given the following conditions:
- Fuel Pool level is at the normal water level.
L Which of the following describes the change in skimmer surge tank level if the first fuel pool cooling pump is started with the discharge valve full open AND the weir gate set at its lowest position?
Skimmer surge tank level . . .
'increases and then. .returns to the level that existed prior to starting z-..
the pump.
I 1- J
'increases to a level higher than existed prior to starting the pump.
~
~ _
..... ...... ~ ........ 1
[decreasesuntil
__ the pump trips on low tank level.
~-
~
.- ... __ ..... ..... ............... i
/decreases and then increases to the level lower than existed prior to. .starting ... __
the ... -.
pump. _ .-.
AI,_- ] Ability to predict and/or monitor changes in parameters associated with operating the FUEL POOL COOLING AND CLEAN-UP controls including:
__ __ 7 -------- -- - -_ - ___ - - - - __ - __ -- __ - __ .- -
A I .Ol- __ [Surge tank level 12.6j2.91
/Justification:
decreases until the pump trips on low tank level. Incorrect. The pump will not trip on low SST level
- Because the level reduction will be slight as long as pool level is at the normal level.
increases to a level higher than existed prior to starting the pump. Incorrect. SST level will lower.
increases and then returns to the level that existed prior to starting the pump. Incorrect. Decreases.
ldecreases and then increases to the level lower than existed prior to starting the pump. Correct..Lowers ~
'sliahtlv then stabilizes sliahtlv lower due to the restriction at the overflow DiDe.. I
. . . __........... .................. 1 I
r
&anup (FPcCsjSystem Lesson Plan:
- a. How normal level is controlled b Sources of makeup to the spent fuel storage pool
- - __ - - - - - __ - __ __ - - __ 71
..~
Saturday, May 10, 2003 12:08:54 PM Page 107 of 161
- _ _ i
Which one of the following describes the interfaces of and Station Service Water (SSW) for the Fuel Pool
.~ ........ ............ _ .. ............ - .....
'1, II and 111 .....only.
. _..._ ....... - - .. 1
- t. 111, and IV only.
- I,................
g-Memory
[Hope Creek
. ._.__- . .. --- 1 233000K109
_K1.- ] Knowledge of the physicacconnectionsand/or cause- effect relationships between FUEL POOL COOLING 1
AND CLEAN-UP and the following:
,KI.O9
/Componentcooling water systems 12.6 [2.61 Justification: Both Loops of SSW can be valved in for makeup source in an emergency. SACS is a cooling water source. SACS is NOT a makeup source.
I, Ill, and IV only. Correct.
I and Ill only. Incorrect. SSW B also MU source Ill and IV only. Correct. SSW Loop A also MU source.
i/ .
- I,. II and Ill only. Correct. .......................Sacs is not a MU Source. ... . ....... ~ . .. ~ J I~...............................
.. . -- .. I i . . . . . . . . . . . . . .... . . . . . . . . . . . . . . .
~ - .. -
1 cleanup (FPCCS) System Lesson Pian:
- a. How normal level is controlled
- b. Sources of makeup to the spent fuel storage pool Saturday, May IO, 2003 12:08:54 PM LPage 108 of 161
Use the attached figure of Interlock Status Display Module to answer the following -
The four indicating lights listed below on the Interlock Status Display Module are illuminated (Assume NO other lights are lit):
u
- FUEL HOIST INTERLOCK
- ROD BLOCK INTERLOCK NO. 1
- ROD BLOCK INTERLOCK NO.2
- BRIDGE REVERSE STOP NO. I Select the answer from below which describes actual plant conditions.
The Reactor Mode Switch is in STARTUP. One control rod is withdrawn. The RefuelBridge is over the core. The Refuel Bridge operator has grappled a fuel bundle and has begun to lift the bundle (i.e., - - the
._ - Main
__ _ -Hoist -
load cell senses a bundle load on the - grapple).
-1 I
The Reactor Mode Switch is in REFUEL. One control rod is withdrawn. The Refuel Bridge is over the core. The Refuel Bridge operator has grappled a fuel bundle and has begun to lift the bundle (i.e., _the _ - Main - -Hoist load cell senses a bundle __ load on the
- grapple).
The ReactorMode Switch is in REFUEL. One control rodis withdrawn. The Main Hoist is retracted. No Fuel Bundle is on the grapple. The Refuel Bridge has been moved towards the core
_ _ and the __over-core proximity switches are made - __ up._ _ __ -
The Reactor Mode Switch is in STARTUP. One control rod is withdrawn. A Fuel Bundle is grappled on the Main Hoist. The Refuel Bridge has been moved towards the core and the over-core proximity switches __ are made- up. - - - -- _ -
~
b sb Comprehension [HopeCreek
~ T I E m e r g e n c yProcedures and Plan 3.5 I
Justification:
SRO 10CFR55.43 (7) Fuel handling facilities and procedures.
,Requiresfigure 14 of LP -000226 The Reactor Mode Switch is in STARTUP. One control rod is withdrawn. A Fuel Bundle is grappled on ithe Main Hoist. The Refuel Bridge has been moved towards the core and the over-core proximity switches are made up- Incorrect-Mode Switch in STARTUP would cause Bridge Reverse Stop No. 2 light to illuminate also IAW HC.OP-SO.KE-0001 .
. The Reactor Mode Switch is in REFUEL. One control rod is withdrawn. The Refuel Bridge is over the core. The Refuel Bridge operator has grappled a fuel bundle and has begun to lift the bundle (Le., the Main Hoist load cell senses a bundle load on the grapple-correct- IAW HC.OP-SO.KE-0001
,. The Reactor Mode Switch is in STARTUP. One control rod is withdrawn. The Refuel Bridge is over the core. The Refuel Bridge operator has grappled a fuel bundle and has begun to lift the bundle (Le,, the Main Hoist load cell senses a bundle load on the grapple- Incorrect -Mode Switch in STARTUP would cause Bridge Reverse Stop No. 2 light to illuminate also. IAW HC.OP-SO.KE-0001
. The Reactor Mode Switch is in REFUEL. One control rod is withdrawn. The Main Hoist is fully retracted. No Fuel Bundle is on the grapple. The Refuel Bridge has been moved towards the core and the over-core proximity switches are made up-. Incorrect -As the Main Hoist is unloaded, no interlock L lights would be lit IAW HC.OP-SO.KE-0001 . .- ~. - .
Saturday, May___ 10, 2003 - 12:08:54 PM m g e 1 0 9 of 1 6 -
1
-~
I Saturday, May
_ _ 10,
_ _ _ _2003 12:08:54
~-
PM --
Given the following conditions:
- The plant scrammed from an inadvertent MSlV closure.
\--I 1- EOP-101 has been entered.
- LO-LO Set is controlling pressure between 905 and 1017 psig.
- The STA reports Drywell pressure has risen from 0.5 psig to I.4 psig.
Based on these indications, what is causing the drywell pressure to rise and what action is required to mitigate the event?
.~~.
. .................. .___ .. .. . .............. ... - ~.-. . . . -.
iH SRV is stuck closed. Cycle H SRV. ... ..... ............ ..-...
........... .-~..-. . . . . . ~ _ ._ . . . . . . ....
H SRV tailpipe vacuum breaker i s stuck open. Place H SRV handswitch to closed position. I P- --SRV tailpipe - - ___ vacuum
- breaker is stuck - _ _ closed.
- - -Place P SRV handswitch - - to open --_ position.
-- __ __ ]
- e. Ability to (a) predict the impacts of the following on the RELlEF/SAFET/ VALVES; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
A2.01 ! (Stuckopen vacuum breakers 13.0 Justification:
H SRV tailpipe vacuum breaker is stuck open. Place H SRV handswitch to closed position. Correct. RPV v
Pressure is within the range for H SRV operation. Rising DW pressure is caused by the H SRV tailpipe vacuum break open and discharging to the drywell airspace.
ZHSRV is stuck closed. Cycle H SRV. Incorrect. Pressure range indicates H SRV is operating on LO-LO
,P SRV is stuck open. Cycle P SRV. Incorrect. RPV pressure would continue to lower. Wrong action.
P SRV tailpipe vacuum breaker is stuck closed. Place P SRV handswitch to open position. Incorrect. A i
stuck closed~.- . ..
SRV tailDiDe vacuum breaker can .......
not be.diagnosed from .the ...
info given.
.7...
Wrong
...... .--7 action. ----
- .. .- ............. i (R) Given any step of the procedure, describe the reason for performance of that step andlor expected system response to
- 1 EOP- I O I flowchart
._. ~ - -~ r - - ~ ~
Saturday, May 10, 2003 12:08:54 PM I Page 111
-_ of 161
Given the following conditions:
- T = 0 sec LOCA occurs.
- T = 2 sec High Drywell pressure signal is generated and all equipment responds as required.
- T = 20 sec ADS CH D INITIATION PENDING (RPV Level I) annunciators alarm.
- T = 24 sec ADS CH B INITIATION PENDING (RPV Level 1) annunciators alarm.
- T = 48 sec Drywell pressure drops to 0 psig due to a large pipe break at a penetration.
- All control rods are full in.
- EOP execution is in progress.
[Plant Systems i239002A204- -
1239002 1 RelieflSafetyValves __ __ .- - - .
84
'A21 ] Ability to (a)pGdict the impacts of the following on the RELlEF/SAFEl% VALVES; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or .operations: - - _ _ - -
b & &
~ 4.1 a
Justification:
Before RPV level reaches -1 90 inches; manually. Correct. EOP 101 Step ALC-9.ADS is manually inhibited at -129 inches, ADS blowdown before -190" by manual operator action.
When T = 425 seconds; automatically. Incorrect. ADS will be inhibited at -129 inches.
When T = 302 seconds; automatically. Incorrect. ADS will be inhibited at -129 inches.
After RPV level reaches -200 inches; manually. Incorrect. All equipment responds as required. Therefore steam coolina is not reauired.
I 1 HC.0P-EO.ZZ-0101 I
. I zatioi System by evaluation of the controlslinstrumentationlalarms, IAW the Automatic
-Saturday, _ _ _IO,
~ _ _ _ _ _May _ - 12:08'54 PM
_ _2003
Given the following conditions:
- Reactor power is 50%.
ALL Turbine Control Valves fail OPEN.
The MSlVs fail to automatically close.
The reactor was scrammed and MSlVs are closed manually.
Determine which of the following combinations of reactor power and eactor pressure would
- indicatethat a Safety Limit violation had occurred?
Reactor RPV Power Pressure
. ~ ~
I 1
~
115% - 750 psig
...... .-.-.... . .. . .... ~ .-. . .-........ ..
1
~
.24% 770 psig
................. -........___ ..... ___ ................... ... -. ........ ............. ____ .... . .-.--~ . . .- .....
28%
775psig .. .... - _
................ 1
- -_ _ - - -I Comprehension b JUSTIFICATION:
55.43(2) Facility operating limitations in the Technical Specifications and their bases.
/CorrectAnswer: 28% 775 psig Power is greater than 25% with pressure less than 785 psig. -
IThe following distractors are incorrect as follows:
32% 810 psig - Pressure is greater than 785 psig so no indication of a safety limit violation.
24% 770 psig - Power is less than 25% with pressure less than 785 psig.
15%
. - .- 750 mia
. - ., .~.Q - .Power - ~ is - less than 25% with Dressure less than 785 Dsia.
~~~ ~
. w
[HCGS Tech Specs - 2.1.1 I
1 ............ ._ .. ...... -- ........ I TECSPCEOI0
~
Saturday, May IO, 2003 12:oa:55 PM -
) Page 113 of 161
Which one of the following decribes the pressure rise at the Main Turbine inlet pressure and reactor steam dome pressure as power is increased from syncronization to rated thermal power? Assume reactor power change at a constant ramp rate.
u Main Turbine inlet pressure rise is and reactor steam dome pressure rise is
- Linear; Linear.
, -~~ ........ . ........ -.. ......... -. ... - -_
1
~~~
/Linear: Non-Linear.
"on-Linear; Linear. .
~-
-~ .
I I
~..
"on-Linear; Non-Linear. ~ .. ~
a 8241000K504 1- Reactornurbine Pressure Regulating System __ _ _ 86 K5, Knowledge of the operational Implicationsof the following concepts as they apply to REACTOR/TURBINE PRESSURE REGULATING SYSTEM:
K5.04- .Furbine inlet pressure vs. reactorpressure 3.313.3 Justification:
Linear; Non-Linear. Correct. Main turb inlet presssure rises from 920 to 950 psig at 3.33 percent steam flow to 1 psig rise. Reactor pressure rises from 920 to 1005 psig. Reactor pressure rises higher due to the differential pressure caused by steam line flow increases with increased flow.
Linear; Linear. Incorrect. Reactor pressure rise is non-linear Non-Linear; Linear. Incorrect. MT inlet pressure rise is linear. Reactor pressure rise is non-linear b iNon-Linear; I----
. ~~~
Non-Linear.
~ __._... ~~~~
Incorrect. ~.
MT inlet pressure rise is linear. .
NI OHOI EHCLOG-00 fiaure 2 I
..... ......... ............ . ................ -. .............. ~- ...... __ ........................................... ..- ...........
EHCLOGE002 I (R) Given plant conditions evaluate the cause-effect relationshipbetween the pressure regulating - - system - and the following IAW I the. Lesson' Plan: - 1 Reactor Power Reactor Pressure Steam Flow Reactor Water Level 1
Saturdav. Mav 10.2003 12:08:55 PM 1 Page 114 of 161
Given the following conditions:
- The plant is operating at 100% power.
, - Main Turbine testing is in progress.
- The LOCKED OUT pushbutton on 10C650E has been depressed, energizing the lockout valve.
'Which of the following describes the effect of energizing the Lockout Valve?
The Master Trip Solenoid is bypassed ___
.___ to--prevent
- - __ - depressurizing
- - ___ the
- - Emergency
-___ ___ -_ Trip System.
- _ - - 7
/AllTurbine trips are bypassed to allow for__testing.
~.. .- _ _-_____ __ .
1
- . !AllTurbine trim
~
1__ EXCEPT for the mechanical overspeed 1 ...................
trip are___bypassed.
1 J
'Only
. the Turbine mechanical overspeed trip...is bypassed. ............ ... ..
- ~ . . __ ...... -
!K5, ..... ].Knowledgeof the operational implications of the following concepts as they apply to MAIN TURBINE
- GENERATORAND AUXILIARY SYSTEMS:
- ._- -.- .. .. ...... .-.-...- . .... - ...... .... __ .... ... .... ____ -- - - ~
.'K5.03 --. ~ [Hydraulically operated valve operation [2.6i[
-. 2.61
!Justification: Only the Turbine mechanical overspeed trip is bypassed. Correct. The lockout valve shifts the upper lockout valve spool such that high pressure fluid from the MTV is blocked and high pressure
.fluidfrom the FAS header is supplied to the steam admission valve disc dump valves. Does not bypass ithe Master Trip Solenoid.
- The Master Trip Solenoid is bypassed to prevent depressurizing the Emergency Trip System. Incorrect.
IMTS is not bypasses.lt remains active.
- AllTurbine trips are bypassed to allow for testing. Incorrect. Only Mechanical Overspeed trip is bypassed
.to allow testing.
EHCOILE009 1 (R) Concerning the hydraulic trip system, from memory summarizelidentifythe purpose of the following components.
I None ....... - .. . -
- ~ ~-
Saturday, May 10, 2003 12:08:55 PM
Given the following conditions: i The plant is operating at 100 percent power.
The Hotwell level control is selected to 'A on 1OC651A.
A large pipe break occurs on the tube side of the in service SJAE Condenser.
/Whichone of the following describes the plant response and what operator action(s) will be
/required?
'A Primary .. Condensate Pump trips; Reduce Reactor power using the 'Stuff._
Sheet'.
, ~- ....-
!All
._ Primary Condensate Pumps trip; Trip all Secondary Condensate and Feedwater Pumps.
~~
.. .- .... ~. -_
__ I
'All Primary Condensate
_____.. . . . . . . . . . . . . . Pumps trip; Lock the reactor mode switch in Shutdown.
? A Primary Condensate Pump trips; Verify Full Reactor Recirc and Feedwater pump runbacks.-l A2 .
, , . ..--~___. . .....-. ........................
'A.bi.l.ityi.. o (a) predict the impacts of the following on the REACTOR CONDENSATE SYSTEM; and (b)
/basedon those predictions, use procedures to correct, control, or mitigate the consequences of those
/abnormalconditions or oDerations:
.- ___. ..... ____. ...~-- .- .... .____ ...
A2.06 , ILow hotwell - --
level
[Grn - __
Justification:
iAll Primary Condensate Pumps trip; Lock the reactor mode switch in Shutdown. Correct. Lowering iHotwell level will result from a condensate system pipe break. A hotwell level channel will trip all three PCPs. At 100 percent power with a loss of condensate and feedwater, RPV will drop rapidly. HC.OP-AB.RPV-0004 Immediate operator action of locking the mode switch to shutdown is required.
All Primary Condensate Pumps trip; Trip all Secondary Condensate and Feedwater Pumps. Incorrect.
Wrong action per AB.
'A'Primary Condensate Pump trips; Reduce Reactor power using the 'Stuff Sheet'. Incorrect. Wrong action per AB.
'A'Primary Condensate Pump trips; Verify Full Reactor Recirc and Feedwater pump runbacks. Incorrect.
Wrong action Der AB.
IHC.OP-AB.RPV-0004 -
I - -- ---- - I MNCONDEOl5 (R) Given initial conditions and the loss of one or more condensate pumps, explain the interlocks and automatic actuations 1
~
associated with the runback and/or trip logic of the condensate, feedwater and reactor recirculationsystems. i ABRPV4E003 (R) From memory, recall the Immediate Operator Actions for Reactor Level Control. 1
~~ .______
Saturday, May 10, -
-________ 2003 12.08:55 ~
._ _ ~ _ _
Given the following conditions:
- Plant power level is 90 percent.
v
- # 2B Feedwater Heater (FWHTR) water level is rising.
Which one of the following actions occur if the FWHTR level reaches the HI-HI level setpoint?
- 1 The Main Turbine trips. - -- ._ - ~- 1
- 2B FWHTR Bleeder Trip Valves trip closed. I
- 1, DC, & 2B FWHTR Startup and Operating Vents auto close. 1 K4. 1 Knowledge of REACTOR CONDENSATE SYSTEM design feature(s) and/or interlocks which provide for 1 the following: -- I K4.06 IControl of extraction steam 1 2.8 [ 2.81 B FWHTR Condensate inlet valve auto closure. Correct. High level in the #I, #2 or Drain Cooler automatically close the condensate inlet valves AE-HV-1633 and outlet valves 1600.
Main turbine trip. Incorrect. High RPV level trips turbine Bleeder trip valves trip closed. Incorrect. Neck heaters have internal piping within the condenser shell and do not have BTVs. Correct for other heaters except 1,2,&DC.
L
emory, describe the effects of too high or too low o I
I
~- --
Sunday, May 11, 2003 11 02 37 AM- _ _
~
' Page 117 of % I
_____- ~. __ __ -
Given the following conditions:
- Reactor power is 14 percent.
L
- Reactor pressure is 923 psig.
- 'A Reactor Feed Pump is feeding the vessel in Manual control.
- Start Up Level Control Valve (SULCV) demand is 22 percent.
Which one of the following choices describes the approximate positions of the SULCV's LV-1754 and 1785?
SULCV 1754 is percent open and SULCV 1785 is percent open. __
22; 0 0; 22 El 100: 2 a2; 100 -1 I
v 22; 0 Incorrect - 1754 is 100 percent open. 1785 is 2 percent open.
0; 22 Incorrect - 1754 is 100 percent open. 1785 is only 2 percent open.
2: 100 Incorrect. Reverse of actual positions.
I Digital Feed Drawing 1H-AE-ECS-0128-03F I LP NOH01_
_ FWCONT-00 ~ i Feedwater Control System Lesson Plan:
~
~
i Explain the function of each indicator.
Assess plant conditions that will cause the indicators to light or extinguish.
Determine the effect of each control switch on the Feedwater Control System.
I
___ switches to perform their intended functions.
or permissives required for the control I
A Sunday, May Page 118of 161
__ 11,2003 11 02 37 AM
~~~
__ ~ ____
Given the following conditions:
- The plant is operating at 16 percent power during a startup.
u - An I&C Techinician has completed channel functional testing on RPV Level 8 instrumentation to the Digital Feedwater System.
The values recorded for the trip setpoints were as follows:
A - 52.3 inches B - 54.9 inches C - 55.7 inches Based on this data, which one of the following - actions are required?
Restore the affected channel(s) to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least STARTUP within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
Restore the affected channel(s) to operable status within 7 days or be in at least STARTUP within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
I I
Place the affected channel(s) in the trippedcondition AND restore one channel to operable I status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least STARTUP within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. i
,Place the affected channel(s) in the tripped condition AND restore one channel to operable status within 7 days or be in at least STARTUP within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
L b a n t Systems c 3 02G133 259002 Reactor Water Level Control System J 91 2.1 1 Conduct of Operations _ _ _ _ _ _ _ _ _ ~ 2 2.1.33 1 /Ability to recognize indications for system operating parameters which are entry-level conditions for I 3.4 I4.01 itechnical specifications. -_ _____.A Justification:
SRO 55.43(2)Facility operating limitations in the Technical Specifications and their bases.
Correct: Place the affected channel(s) in the tripped condition AND restore one channel to operable status within 7 days or be in at least STARTUP within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Only C channel is above the TS Allowable value of 55.5. No action is required for B channel. TS 3.3.9 action a. and b. are applicable.
incorrect: Restore the affected channel(s) to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least STARTUP within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Action for 2 inoperable channels. Channel C must also be placed in trip condition.
Incorrect: Restore the affected channel(s) to operable status within 7 days or be in at least STARTUP within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Channel C must also be placed in trip condition.
Incorrect: Place the affected channel(s) in the tripped condition AND restore one channel to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least STARTUP within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Action for 2 inoperable channels.
~ - _ .~
7 a. Identify those sections which are applicable to the Feedwater Control System.
i I
Sunday, May 11,200311:0237iii-
~~ ~
Ppagellsofi61
~~ _ .. _____
- b. Evaluate Feedwater Control System Operability and determine required actions based upon system operability.
c Explain the bases for those Technical Specifications sections associated with the Feedwater Control System. (SRO Only) '
-~ -.
______ ~ - - . _ _ _ ~ .
~
Sunday,_ May 11,2003 11 02:37 AM Page 120 of 161
~- - _. - ______._ _ _ _ _ ___
Given the following conditions:
- The plant is at 100% power.
L H
- The "C" steam flow detector for the feedwater level control system fails low (its output indicates 0 Ibm/hr steam flow).
SELECT the statement which describes the automatic plant response with NO operator action. -_____
Reactor water level will remain the same. The feedwater level control system will shift to single element control.
Reactor water level will decrease and stabilize at a lower than normal value. The feedwater
- I 1 level control system will remain in three element control. -1 Reactor water level will decrease and stabilize at a lower than normal value. The feedwater I level control svstem will shift to sinale element control.
Reactor water level will increase and stabilize at a higher than normal value. The feedwater 1 level control system will remain in three element control. I will have on following:
'Engineering Drawing H-I -AE-ECS-0128-0 rning Objectives
-I__ I FWCONTE013 I R ) From memorv. descnbe the remonse of the FWLC Svstem if the total steam flow sianal were to be lost. IAW the Feedwater
-_ _ _ _ _ _ _ _ ~ _ . ___
Sunday, May 11,2003 11 02 37 AM I Page 121 of 161
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- ~~ ~
~ _ _ _ ___ ~~
~
Sunday, May 11, 2003 11.02:37 AM
~
Page 122 of 161
7
~ ~
~
Given the following conditions:
- The plant is operating at 100 percent power.
u
- AD483 Inverter output power is lost.
~- __ does the loss have on the 'A' RFPT?
What effect -.
'A'
_ _RFPT
~- will NOT trip on RPV __ ___
'A RFPT trips due to Loss of Control Oil pressure. I
'A'RFPT trips due to Loss of Speed Signal. I
'A'RFPT will NOT trip on Overspeed. I
'A' RFPT trips due to Loss of Control Oil Pressure. Incorrect. Supplied from DC power source.
'A' RFPT will NOT trip on RPV Level 8. Incorrect. 2 of 3 channels remain which would satisfy logic to trip all 3 pumps.
I -
HC.OP-AB.=-0136 - ~- __I I
I lationshipbetween the NVLC System and that System, IAW the Feedwater Control System Lesson Pian:
- a. 120 VAC Non-I E Electrical Distribution
- b. 125 VDC Non-I E Electrical Distribution
- c. Main Turbine
- d. Recirculation System
- e. Rod Worth Minimizer (RWM)
- f. Main Steam
- g. Redundant Reactivity Control (RRCS)
~ ~
Sunday, May 11,2003 11 02.38 AM Page 123 of 161
~ ~~ - ~
______~ ~ ___-_____ - -
I Given the following conditions:
- The 1AD413 125 Volt Battery Charger is in service and providing a normal charge on its battery.
L
- The 1AD414 125 Volt Battery Charger is tagged for maintenance.
- While in this lineup, AC power to the charger is lost.
- The bus supplying the charger is reenergized after 20 minutes by its associated diesel generator. ,
I Which of the following describes the response of this battery charger?
The 1AD413 Battery Charger will:
return to the "float" mode to recharge the battery. . .
1 J
trip and is interlocked "off 'I with the diesel generator powering the bus.
I reset to the "equalize" mode to recharge the battery. i trip and must be manually restored as permitted by diesel generator loading. 1 AI. ] Ability to predict and/or monitor changes in parameters associated with operating the D.C. ELECTRICAL DISTRIBUTION controls including:
A I .01 [Battery charging/dischargingrate 12.5 12.81 return to the "float!' mode to recharge the battery. Correct. Although the charging
- - rate will be higher than W prior to the charger loss, the charger will remain in the Float mode.
trip and is interlocked "off with the diesel generator powering the bus. Incorrect. The charger does not
'I
- trip.The charger is restored when the bus power is restored.
reset to the "equalize" mode to recharge the battery. Incorrect. Equalize mode must be manually initiated using the timer control on the charger.
trip and must be manually restored as permitted by diesel generator loading. Incorrect. Returns when the AC bus is repowered.
1 c
Distribution Lesson Plan.
- a. 480VDC IE/NIE Power Supply 1-Page
__ 124 of i s i
_ _ ~ - ____ ~ - - _______ ~ ~ _ _ -
Given the following conditions:
- The unit tripped from 100% due to a loss of offsite power.
- HC.OP-AB.ZZ-0135 implementation is in progress.
v
- EDG B, C, & D are loaded onto their respective busses.
- EDG A is running.
- EDG A voltage is 3500 V.
- EDG A frequency is 56 Hz.
- EDG A output breaker is open.
Which one of the following actions must be taken to mitigate this event?
Raise speed and voltage from IOC65l panel to within limits and verify output breaker-automaticaIIv closes.
Dispatch an operator to the EDG Remote Panel to raise speed and voltage to within limits and verify output breaker automatically closes.
'Shutdown EDG A with the IocaVremote panel Emergency Shutdown pushbuttons. 1 Turn on the Synchroscope key and manually close EDG A output breaker from the Control Room.
1I W
frequency cannot be adjusted with a LOP signal present. Normal stop controls in the Control Room and local/remote panels are disabled. The only option is to secure the EDG with emergency stop PBs locally or leave it run unloaded. There is a time limitation for running unloaded of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
Raise speed and voltage from 1OC651 panel to within limits and verify output breaker automatically closes. Incorrect. Speed and voltage controls are disabled with LOP signal present.
Dispatch an operator to the EDG Remote Panel to raise speed and voltage to within limits and verify output breaker automatically closes. Incorrect. Speed and voltage controls are disabled with LOP signal
'present.
Turn on the Synchroscope key and manually close EDG A output breaker from the Control Room.
Incorrect. Turning on sync scope key enable the snyc check monitor. With no infeed voltage to match, the breaker will not close.
~
EDGOOOEOI
__ 1 I (R) Given plant conditions, predict the response of Diesel Generator governor and voltage regulator control circuitry to an Emergency start (LOP/LOCA).
~
- i EDGOOOEOl7 (R) Given plant conditions, predict the response of the Diesel Generator to an emergency start signal if:
iI u
- a. A normal shutdown is in progress.
tdown is i i Sunday, May 11,2003 11:02:38 AM jPage125Of161
~ ._ - -~ - _~___----- ~-~
None
___- -1
/Significantly Modified_ _ _ _ ~ J I
INPO BANK QID # 138 Sunday, May 11.2003 11 02.38AM Page 126 of 161
-~ ~
____ - ~ _ .__-_ _ _ __ ~-
Given the following conditions:
- The Starting Air Compressor IDK402 is Safety tagged for maintenance.
- The 'D' EDG Starting Air Receivers are crosstied to the 'B' EDG. (See attached figure)
L
- While pressurizing the 'D' EDG receivers, the crosstie air hose splits open.
Assuming NO operator action, which one of the following describes the effect on the associated EDGs to respond to a LOP?
Only B EDG will respond.
Only D EDG will respond. I B and D EDGs will respond. i Neither B NOR D EDG will respond.
diesels to start on the LOP Only B EDG will respond. Incorrect. Both diesels will start.
1 , ,
- a. A normai shutdown is in progress.
II I
1 1
~- -- ~ __
Sunday, May II,_2003
_ ~
I 1 02 38 AM ~
- The reactor core has been operating with one or more known fuel pin leaks.
- A reactor scram occurred from 100 percent power.
- Both Scram Discharge Volume Drain Valves did NOT go full closed.
Which one of the _- following rooms would become the most significant - radiological hazard?
Reactor Building North Equipment Sump Room.
HPCl Pump and Turbine Room. - . . I Reactor Building South Equipment Sump Room.
RClC Pump and Turbine Room.
268000 j Radwaste 97 Emergency Procedures and Plan SRO 55.43(4) Radiation hazards that may arise during normal and abnormal situations.
Correct: Reactor Building South Equipment Sump Room. The North and South Scram discharge volumes drain through a common line to the Reactor Building Equipment Drain Sump 1BT266 located in the South Reactor Building Sump Room on 54' elevation. If the drain valves did not close as stated in the stem, a LOCA would exist discharging into this room. The leaking fuel would severely raise radiation L levels in that room as well. Entry into 103/4 for any room Rad monitor alarm. It is not limited to only rooms of Table 1 and 2.
Incorrect: Reactor Building North Equipment Sump Room. North SDV does not drain to this sump.
Common misconception.
Incorrect: HPCl Pump and Turbine Room. Rad levels would increase slightly from steam line drains unless HPCl was placed I/S. Stem does not support HPCl operation.
Incorrect: RClC Pump and Turbine Room. Rad levels would increase slightly from steam line drains unless RCIC was Dlaced IIS. Stem does not suo~ortRClC ooeration.
/EOP-103/4
/EOP Conversion documents - _ ~ _ - ~- ~
-~
I None Sunday, May. 11,2003 11:02:38 AM
~ ___ - . _ _ _ _ ~
Given the following conditions:
- A discharge of the Equipment Drain Sample Tank is in progress to the river.
- The Liquid Radwaste Discharge Isolation Valve to the Cooling Tower Blowdown automatically W
closes.
Which one of the following conditions would cause this termination?
(Assume no operator action) __ ___ -
Liquid Radwaste-Effluent radiation element fails low.
Cooling Tower Blowdown weir flow rate fails low.
Liquid Radwaste Effluent sample flow rate fails high. 1 Cooling Tower Blowdown RMS radiation element fails high. 1 272000K3011 98
-K3. 1 Knowledge of the effect that a loss or malfunctionof the RADIATION MONITORING System will have on followina:
1I
" I K3.01 /Station liquid effluent release monitoring 13.2 [ 3.81 Coolinq Tower Blowdown weir flow rate fails low. Correct. Of choices given, only Cooling Tower weir flow (Dilution Flow) low will cause a release isolation and termination.
Liquid Radwaste Effluent radiation element fails low. Incorrect. Cause alarms but not isolation.
Liquid Radwaste Effluent sample flow rate fails high. Incorrect. Cause alarms but not isolation.
IHC.OP-AR.SP-OOOI Attachment 5 ~
1
___-______ ____- I RWOVEREOOS ~ (R)From memory listlidentify the five conditions that will cause a liquid release to be automatically terminated.
I ______
I None Sunday, May 11,2003 11:02 39 AM L--
Page 129 of 161
- The plant is operating at 100 percent power during hot summer conditions.
- CRIDS page 105 indicates Reactor Building Backdraft Damper PD-9438C1 is closed.
v
- All room temperature points are reading NHI.
- NO other Backdraft Dampers are closed.
What impact will this closure have on the plant? __ -. .
MSlV closure is imminent. A Reactor scram is necessary. - ---
- 1 IH Reactor Water Clean UD will isolate.
FRVS must be placed in service to maintain room temperatures. 1 Associated room temps will rise. Reset the damper. I room temps will rise. Reset the damper. Correct. IAW Subequent action F2, re-open the backdraft damper. If a high temperature isolation does not exist.
MSlV closure is imminent. A Reactor scram is necessary. Incorrect. Misinterpretationof table of L
Attachment 2.
Reactor Water cleanup will isolate. Incorrect. Damper is one of several that feed into RWCU pipechase rnnm I
FRVS must be placed in service to maintain room temperatures. Incorrect. FRVS uses the same ductwork. FRVS uses SACS instead of Chilled water. It also uses the same flowpath as RBVS.
/HC.OP-AB.CONT-0003 __-__ --
r 1 I __-- ___-
HC.OP-AB.CONT-0003Attachment 2
___ -_ _ . ~ _
Sunday, May 11,2003
___.. ~ ~-
11:02:39AM ~
Page
~_._
130 of 161
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- 7
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Given the following conditions:
- - "B" Channel Reactor Building Refuel Floor Exhaust Radiation monitor is in the trip condition for I&C surveillance testing.
- Power is lost to the "B" Channel Reactor Building Refuel Floor Exhaust Radiation monitor.
Which one of the following describes the plant response, if any?
Neither Reactor Building Ventilation Inboard and Outboard Dampers HD-9414A & B or HD- 1 I 9370A & B close. 1 I
Reactor Building Ventilation Inboard Dampers HD-9414A and HD-9370A only close. -1 Reactor Building Ventilation Outboard Dampers HD-94148 and HD-9370B only close. 1 Both Reactor Building Ventilation Inboard and Outboard Dampers HD-9414A & B and HD-9370A & B close.
1I A3. 1 Abilitv to monitor automatic oPeraions of the SECONDARY CONTAINMENT includinq: I close. Correct - Loss of-power to the same channel that is tripped only results in 1/3 trip and no dampers change position.
Reactor Building Ventilation Inboard Dampers HD-9414A and HD-9370A only close. Incorrect. 2 of 3
,logic.
,Reactor Building Ventilation Outboard Dampers HD-94148 and HD-9370B only close. Incorrect. 2 of 3 logic.
Both Reactor Building Ventilation Inboard and Outboard Dampers HD-9414A & B and HD-9370A & B close. Incorrect. 2 of 3 logic. I I
I 1- Ventilation and s t a i the Filtration Recirculation and Ventilation System (FRVS) IAW the Secondary Containment Lesson Plan. 1
- Editorially
- Modified I
Vision Bank QID# Q76880 Modified for different channel. I I
~~ ~ ~ ~ -
Sunday, May 11,2003 11'02:39- AM
~
Page 131 of 161
Given the following conditions:
- The plant is operating at 100 percent power.
- The Main Steam Tunnel (MST) Ventilation Barrier (Panel) 10S203 indicates open on the RM-I 1.
L Which one of the following describes the operational impact? ~~ ____
Turbine Building Exhaust RMS levels will rise to alert levels. ____
Degraded cooling capability for the MST Coolers. I Main Steam Line RMS detectors will read non-conservative. I Loss of secondary containment integrity.
I
~ d Memory IHope Creek 1 ~
290001K107 1
290001 Secondary Containment CONTAINMENT and the following: I Loss of secondary containment integrity. Correct. The MST Ventilation Barrier is part of secondary containment and are verifed in place using the RM-11 and surveilance HC.OP-ST.ZZ-0003
,Degraded cooling capability for the MST Coolers. Incorrect. Plausible misconception.
Main Steam Line RMS detectors will read non-conservative. MSL RMS are ion chambers and are not affected by ventialtion changes.
Turbine Building Exhaust RMS levels will rise to alert levels. Incorrect. Opening a ventilation path to the Turbine Buildina wit not drive RMS values UD unless there is a system breach in the MST.
1 1 HC.OP-ST.ZZ-0003 I
I I
1- Lesson Plan.
SECCONE005 1 (R) From memow describe the purpose and locations of the steam vent to atmosphere blowout panels, and the differential 1 Containment Lesson Plan. ~
. - ~- - -~
Sunday, May I I , 2003 II 02 39 AM Page 132 of 161-
Following sustained steady state operation a<lOO% power for 103 hours0.00119 days <br />0.0286 hours <br />1.703042e-4 weeks <br />3.91915e-5 months <br />, indicated reactor power drops to 97% without operator action. Recirculation flow and rod positions have NOT changed.
- Which of the following is the explanation Core shroud
_ _ ~ cracking has occurred.
for this change in power?
_____ -J 1
I I
I Feedwater flow to the reactor has risen. -__
I
- 2
~
1 Steam quality exiting the steam dryers - has been reduced.
I I
An Electro-Hydraulic Control (EHC) system change has caused reactor pressure to rise.
290002K303 K3. ] Knowledge of the effect that a loss or malfunction of the REACTOR VESSEL INTERNALSwill have on 1 Discuss the HCGS design and/or procedural guidelines that mitigatelreduce the likelihood of the problernlindustry event at HCGS IAW the plantl industry event.
-~ ______ ~
Sunday, May 11, 2003 11:02:39AM
Given the following conditions:
- The Control Room Ventilation System " A is operating normally.
L - The B Train is in a normal, standby lineup.
SELECT the system flow response to a Control Room Ventilation High Radiation Isolation at the points marked on the attached Figure.
A-30OOscfm: B-I 000 scfm; C-4000 scfm; D-0 scfm; E-I 8500 scfm 1 A-4000 scfm; B-0 scfm; C-0 scfm; D-4000 scfm; E-I8500 scfm I
!f A-0scfm; B-4000 scfm; C-4000 scfm; D-0 scfm; E-I4500 scfm - 1 _ _
A-0 scfm; B-I 000 scfm; C-4000 scfm; D-3000 scfm; E-I4500 scfm 1 A3. ]:Abilityto monitor automatic operations of the CONTROL ROOM HVAC including: I 13.51 I JUST1FICATION:
In the OA Mode following a high Rad signal:
- The "A'flowpath isolates; The "B"flowpath opens supplying 1000 scfm to the CREF fan; The "C" flowpath is always 4000 scfm when the CREF fan is running; The " D flowpath supplies 3000 scfm to the CFEF fans 4000 scfm total; The "E' flowpath combines 14,500 scfm with the CREF fans 4000 scfm for a total of 18,500 scfm through the CRS fan.
CORRECT - A-0 scfm, 8-1000 scfm, C-4000 scfm, D-3000 scfm, E-I4500 scfm.
INCORRECT - A-4000 scfm, B-0 scfm, C-0 scfm, D-4000 scfm, E-I8500 scfm. "A" is never 4000. "B' &
"C" are 0 during normal operation. "E' is 18,500 during normal operation.
INCORRECT - A-0 scfm, B-4000 scfm, C-4000 scfm, D-0 scfm, E-I4500 scfm. "B" is either 0 or 1000.
" D is never 0.
INCORRECT - A-3000 scfm, B-I 000 scfm, C-4000 scfm, D-0 scfm, E-I 8500 scfm. "A"is only 3000 during normal operation. I ' D is never 0. "E" is only 18,500 during normal operation.
IM-78 & M-89 1 i I I
the lesson plan:
Normal operation Isolate: Outside Air Mode I Isolate: Recirc Mode 1
.~ - -- - . .
Qusstion Source: Facility Exam Bank
~ . - - - \QuestionModificationMethod: I 'Editorially
-_ .- Modified -
I VISION BANK QID# Q53946.
I
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Sunday, May - 11,200311:02.40AM-- I L
Page 134 of 161--
~
Given the following conditions:
- A loss of coolant accident has occurred.
b
- The Reactor Auxiliaries Cooling System (RACS) has been restored.
Which of the following describes the availabilityhesponse of the Emergency Instrument Air Compressor (EIAC) for these conditions should instrument air header pressure begin lowering?
.~
The ElAC will automatically start on instrument air header pressure less than 85 psig if the LOCA signal is cleared.
II The ElAC will NOT automatically start but can be started locally after relieving intercooler Dressure. ~. . . . .- . . . .. . . . .I The ElAC is NOT available until the LOCA signal is cleared, PClS reset, and the 1E breaker is closed.
The EIAC is NOT available until the Non-lE breaker is closed and instrument air pressure is less than 85 psig.
I AIR SYSTEM: I K5.01 [Air compressors p p i 2.5 12.51 i/
CORRECT: The EIAC is not available until the LOCA signal is cleared, PClS reset, and the 1E breaker INCORRECT: The ElAC will automatically start on instrument air header pressure less than 85 psig if the 1
LOCA signal is cleared. Breaker not reset INCORRECT: The ElAC is not available until the Non-I E breaker is closed and instrument air pressure is less than 85 psig. 1E breaker that needs resseting INCORRECT: The ElAC will not automatically start but may be started manually from the Control Room or locallv. Not until 1E breaker is reset.
o the following conditions, IAW the Instrument Air System Lesson Plan:
- a. Loss of Offsite Power (LOP)
, b. Loss of Coolant Accident (LOCA)
___._ - -~ . - ~-
[Wterial M u i r e d for Examination ' . 1 None huestion Source: I Facilitv Exam Bank Question ModificationMethod: Direct From Source Questkm SOUtCe Comments: Vision Bank QID# (256532
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Sunday, May 11,2003 11:02 40 AM
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Page 135 of 161
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Given the following conditions:
- The plant is operating at 100 percent power when a Loss of Offsite Power (LOP) occurs.
L
- B Emergency Diesel Generator (EDG) trips due to electrical fault.
- D SACS Pump trips on overload.
- All other equipment functions properly.
Which of the following actions is required?
Open SsW-to-RACScrosstie valves.
1 Attempt one restart- of- B EDG.
-1 Emergency stop D EDG. -1 Start B SACS Pump. _ - - 1 400000K301 105
__K3.01 !Loads cooled by CCVVS I - -
j 2.9 13.31 Emergency stop D EDG. Correct. D EDG is running without cooling. It must be emergency stopped because the LOP signal disables the normal stop controls.
i,
/HC.OP-AB.ZZ-OI 35 - I
_- I
-. EDGOOOEOlO 1
,.... . . .~
j i
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Page 136 of 161__
Given the following conditions:
-- -1 I
Station Service Water (SSW) pump status:
- 'A' SSW pump I/S in AUTO. 1 L i
- '6'SSW pump I/S in AUTO.
- 'D' SSW pump O/S in AUTO.
Which one of the following will result in the automatic start of the ID' SSW Pump?
'A' ssw LOOP
'A' low flow..
ssw Pump low-flow.
1 I I
'6'SSW Loop __ low flow. 1
'13'ssw Pump-Low
- - flow.
I standby pump (D).
' B SSW Loop low flow. Incorrect. Low pump flow.
'A SSW Pump low flow. Incorrect. Low flow of the opposite pump within the affected loop starts the standby pump.
'A SSW LOODlow flow. Incorrect. Low DumD flow of the omosite Dump..
,HC.OP-SO.--0001 r
2 1
7
- ~ - ~- - 7-Sunday, May 11,2003 11:02.40 AM I- - Page 137 of 161
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Given the following conditions:
- The plant is at 80 percent power with a power ascension in progress.
- A Reactor Recirc Pump scoop tube is tripped.
- Local adjustment of Reactor Recirculation pump A speed is required.
Which of the following describes the MINIMUM - ~- requirements to perform this evolution?
Communications via page announcements to the operator on the scoop tube; the operator must be RO licensed.
Communications via radio the operator on the scoop tube; the operator must be RO licensed. - . 1 Communications via telephone to the operator on the scooptube; the operator must be SRO licensed.
1I Communicationsvia sound powered phone to the operator on the scoop tube; the operator
'must be SRO licensed.
The radio allows 3 way communications from the scoop tube positioner. RO license is required since moving the scoop tube directly changes reactivity.
1 Communications via page to the operator on the scoop tube; the operator must be SRO licensed.
Incorrect. Only RO license required.
Communications via radio to the operator on the scoop tube; the operator must be SRO licensed.
I
'HC.OP-SO.BB-0002 3.1.6 J
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Sunday,
- May 11,2003 11:02:40 AM 1
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Page 138 of 161
Given the following conditions:
- The plant is operating at 29 percent power.
- i- Overhead Annunciator C5C2 TCV FAST CLOSURE & MSV TRIP BYP is ILLUMINATED.
,Thenthe Main Turbine Generator trips.
- All Turbine Bypass valves responded full open.
- Overhead Annunciator B3-E5 RPV PRESSURE HI is ILLUMINATED.
- RPV pressure stabilizes at 1030 psig.
Which one of the following correctly describes the time limit required by Tech Specs to clear the high pressure alarm?
2 minutes.
I5-minutes. 1 i30 minutes.
1
'One hour. _-
29400161 11 b
15 minutes. Correct TS 3.4.6.2; The high pressure alarm comes in at the LCO limit of 1020 psig. LCO action time limit is 15 minutes.
2 minutes. Incorrect. LCO for Stuck open SRV.
30 minutes. Incorrect. Plausible but wrong.
One hour. Incorrect. Plausible but wrong.
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/
7
/HC.OP-AB.RPV-0005 E.5 -~
I None
__ c Sunday, May- 11,2003 11:02.41 AM--- Page 139 of 161-
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Given the following conditions:
- HPCl is removed from standby to perform HC.OP.IS.BJ-0101 HPCl System Valves Inservice W Test.
- Valve BJ-HV-F042 Suppression Pool Suction Valve stoke time is 2 seconds longer than the "TECH SPECS OR DESIGN LIMITS1value.
- Valve BJ-HV-F004 CST Suction Valve strokes satisfactory and was returned to open position.
Which one of the following actions is required? - _ -
Deactivate F004 open; HPCl remains operable. .- - . _ _
Deactivate F004 open; declare HPCl inoperable. - -
Deactivate F042 closed; HPCl remains operable. _-
Deactivate F042 closed; declare HPCl inoperable.
SRO 55.43 (2) Facility operating limitations in the Technical specifications and their bases.
Correct: Deactivate F042 closed; declare HPCl inoperable, Tech Spec 3.6.3 requires inoperable Primary containment Isolation Valves deactivated closed with 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. HPCl operability requires suction source L from the Suppression Pool, therefore HPCl is inoperable.
Incorrect: Deactivate F004 open; HPCl remains operable. F042 must be deactivated closed. HPCl operability requires suction source from the Suppression Pool, therefore HPCl is inoperable.
/Incorrect: Deactivate F004 open; declare HPCl inoperable. F042 must be deactivated closed. Lining up HPCl to the CST will not meet operability requirements.
Incorrect: Deactivate F042 closed; HPCl remains operable. HPCl operability requires suction source from the Sumression Pool, therefore HPCl is inoperable.
IHCTS 3.6.3 and 3.5.1.c.2 . _ _ J I
i P
r I
Evaluate HPCl System operability and required actions based upon system operability IAW HCGS technical specifications. ,
~ Sunday, May 11,2003
~ - 11:02.41AM Page 140
____ of 161
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Select the statement that satisfies _~
10CFR50.46 Acceptance ___
Criteria for ECCS.
1 Long - Term Cooling - after any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long lived radioactivity remaining in the core. - ~- I Peak Cladding Temperature -calculated maximum fuel element cladding temperature shall NOT exceed 2100°F. -- -1 Maximum Cladding Oxidation - calculated total oxidation of the cladding shall nowhere exceed 21% times the total cladding thickness before oxidation.
Maximum Hydrogen Generation - calculated total amount of H2 generated from the chemical reaction of the cladding with water or steam shall NOT exceed 17% times the hypothetical amount that would be generated if all the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.
294001G128 surrounding the plenum volume, were to react.
Correct- Long - Term Cooling -after any calculated successful initial operation of the ECCS, the 1
OCFR50.46 Acceptance
__ Criteria I 7-2 I . ... .. . . . . . . .. . . . . .. . .- .. . - - . - .- ..
From rnemoty, describe the 5 NRC ECCS acceptance criteria as they apply to the design of the Emergency Core Cooling Systems, IAW the Introduction to ECCS Student Handout.
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SundayMay I I , 2003 11:02:41 AM Page 141 of 161
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Given the following conditions:
- A reactor shutdown is in progress.
L - Power is currently 20%.
- Hydrogen Water Chemistry Injection (HWCI) is out of service.
- Main Steam Line RMS Setpoints are set High.
- 2 Condensate Demineralizers are in service at 3000 gpm each.
- Plant chemistry parameters are as follows:
- Condensate demin influent conductivity - 0.21 umhokm
- Condensate demin effluent conductivity - 0.08 umhokm
- Reactor Water Cleanup conductivity - 0.07 umho/cm
- Reactor coolant sample conductivity - 0.07 umhokm Based on these conditions, which one of the following would cause these indications and what procedure actions must be taken?
Condensate Demineralizer channeling; remove one demineralizer from service.
Crud burst from removing HWCl from sekice; restore HWCI to service. _ - 1 Main Condenser tube leak; isolate the affected condenser waterbox.
I Reactor fuel pin cladding leak; continue power reduction at normal rate.
-1 C IHope Creek
/GenericKnowledge and Abilities 294001G I 34 L
1 111 2.1 1 Conduct of Operations -.
2.1.34 /Ability to maintain primary and secondary plant chemistry within allowable limits. [ 2.3 12.91 JUSTIFICATION:
- I OCFR55.43 (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Correct. Main Condenser tube leak; isolate the affected condenser waterbox. Conductivity into the Cond Demins is high. This is a symptom of a Condenser tube Leak. Required action would be to remove the waterbox IAW AB-RPV-0008.
ilncorrect. Condensate Demineralizer channeling due to low flow; remove one demineralizer from service.
'Demineralizer outlet conductivity is normal. Would have low inlet and high outlet conductivity.
incorrect. Crud burst from removing HWCl from service; restore HWCl to service. RWCU and Reactor coolant conductivity levels are normal.
Incorrect. Reactor fuel pin cladding leak; continue power reduction at normal rate. Power reduction at normal rate not permitted due to MSL RMS setpoints are set high. Indications are not cause for
'emergency power reduction.
L,
Sunday, May I I, 2003 11:02:41AM
Given the following conditions:
- The plant is in Operational Condition 3.
- - A new system engineer has requested that the B Core Spray Pump be started with the discharge valve throttled to 75% open to determine starting current.
The Operations Superintendent . . .
mayallow the test if the STA or another SRO with an engineering degree concurs. 1 may conduct the evolution without restrictions.
_. I must withhold conducting the test until a-iPTE package has been approved. I must NOT allow the test under any conditions. i i3.21 Irenort. !
Justification:
SRO 10CFR55.43 (3) Facility license procedures required to obtain authority for the design and operating changes in the facility.
,mustwithhold conducting the test until a IPTE package has been approved. Correct. The evolution is an
'IPTE and requires a package with Test Engineer and Test Managers designated.
may conduct the evolution without restrictions. Incorrect. Requires IPTE package.
may allow the test if the STA or another SRO with an engineering degree concurs. Needs Test engineer and Test Manager approval.
must NOT allow the test under any conditions. Incorrect. May be performed if IPTE package approved.
ADMPROE082 Given access to Control Room References Determine if an activity meets the criteria for an Infrequently Performed Test or Evolution. IAW NC.NA-AP.ZZ-0084. I INPO Exam Bank INPO Bank QlD# 19840
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May I I,2003 I- Page 144 of 161
_ _ _Sunday,
~ I:02:41 AM ~
~
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Given the following conditions:
- The plant experienced a Failure to Scram from a turbine trip at 98% power.
v
- The operator successfully initiated a manual scram approximately 4 seconds following the turbine trip.
- Post trip analysis revealed the following data for the event:
- Peak Reactor Power: 116% Thermal
- Peak Reactor Pressure: 1205 psig
- Minimum Vessel Level: -1 51 inches (Fuel Zone A and B)
- Most limiting MCPR: 1.09 Which one of the following states which Safety Limit was violated?
ThermalPower Low - - Pressure - Low Flow.
Thermal Power High Pressure - High Flow. - I RPV-LevelSafety Limit. 1 H i~~~ Pressure Safety Limit. I b
~
Comprehension [HopeCreek 294001G222 113
[ 4.11 1
,Thermal Power High Pressure - High Flow Correct. Minimum MCPR is 1.10 i/
Thermal Power Low Pressure - Low Flow Incorrect. Initial power pre transient was above 25 percent with j
,pressureabove 785psig.
RPV Level Safety Limit. Incorrect. Level maintained above -161 i RPV Pressure... ..
Safetv Limit. Incorrect. Pressure below 1325 Dsia. I
=
IHCGS Tech Specs Safety Limits I
J Sunday, May 11, 2003 11:02:42 AM Page 145 of 161
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The core has been off-loaded to the fuel pool. Per HC.RE-AP.ZZ-0049, Hope Creek Conduct of Fuel Handling, what is the MINIMUM permissible complement of personnel in the crew involved in fuel movement NOT involving core alterations?
u Fuel Handling Operator Radiation Protection Technician Reactor Engineer, acting as spotter Fuel Handling Operator Refueling Bridge Operator as spotter Radiation Protection Technician Reactor Engineer Fuel Handling Operator Refueling Bridge Operator SRO acting as spotter Radiation Protection
. _ _ Technician - . . .
Fuel Handling Operator
'Refueling Bridge Operator
'Radiation Protection Technician Reactor Engineer
- Control Room Refuel Monitor . -
Memory /Hope Creek
[Generic Knowledge and Abilities 294001G226
- RIC I 2.2 i Equipment Control 2.2.26 [Knowledge of refueling administrative requirements.
I 114 2.5 3.7'
-~
Justification : IAW Technical Specifications 1.7, and HC.RE-AP.ZZ-0049 sections 5.3.2.C.7. I The procedure stipulates that the minimum crew for non-core alteration fuel handling activities in the spent fuel pool includes the Fuel Handling Operator, Radiation Protection Technician, Reactor Engineer I 1 l
and Spotter. The Reactor Engineer may fulfill the duties of the spotter; hence the minimum permissible complement of people is three. i
, H.C.R E-AP.ZZ-0049
~
I sections 5.3.2.C...7.
' I ADMPROE073 From Memorv State the minimum fuel handlrna crew reauirement for non-core alteration non irradiated fuel handling IAW 1- NC.NA-AP.Zi-0049. I
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Sunday, May 11,2003 11:02:42__
AM Page 146 of 161
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Given the following conditions:
- The plant is in Operational Condition 4.
- The Reactor Head detensioning machine is being lowered inplace to detension the reactor head.
W Which one of the following personnel must be notified prior to the beginning the detensioning wocess?
Refuelina .-Floor SRO. -1J Reactor Engineer. - __ - !
Control Room Supervisor.
Refueling Outage Manager.
GENERIC 1 115 2.2 ] Equipment Control I (2) Facility operating limitations in the Technical specifications and their bases.
Correct. Control Room Supervisor. Required signoff for HC.OP-IOZ-0005. Changes Operational Condition to OC 5.
Incorrect. Reactor Engineer. Not required.
Incorrect. Refueling Floor SRO. Not required.
4ncorrect. Refueling Outage Manager. Not required. I iHC.OP-1O.Z-0005 steo 5.2.33 I
~ I 1
I I i
.-i Page 147 of __
161
Given the following conditions:
- The plant is in Operational Condition 5.
- There are 16 fuel bundles remaining in the Reactor vessel.
L Which of the following evolutions would be considered a "Core Alteration" by Technical Specifications?
Transferring a control ___- rod from the Reactor vessel to the Spent Fuel Pool. - . __ _ _
Removing an LPRM string from the Reactor vessel.
-~
. . - I Removing an IRM _ _ detector
- from undervessel. - . 1 Transferinn a control rod blade guide from the Spent Fuei Pool to the Reactor vessel.
-1 a Memory \Hope Creek
~
-~
294001G227 2.2 ],Equipment Control 1
- r I ledge
. of the. .refueling
. process. - _ .
I-[ 2.6 13.51 Justification: TS Definitons 1.7 Core Alteration shall be the movement of any fuel, sources, or reactivity control components within the reactor vessel with fuel in the vessel. Movement of SRMs, IRMs, LPRMs, TIP, or special movable detectors (including undervessel replacement) are not considered to be core alterations H C G T G c h Specs 1.ODefinitions i / .
I 1
TECSPCEOOS Define or discuss the terms contained in Section 1.Oof Hope Creek Generating Station Technical Specifications.
- -~
None
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Sunday, May 11,2003 11:02.42 AM Page 148 of 161
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Given the following conditions:
- An initial startup is in progress.
u
- Threshold power is I 1 W/ft.
- A 20 MWe/hr ramp is established during the last rod adjustment.
- The 20 MWe/hr ramp continues until nodal power reaches 12.5 W / F t .
- Then PCIOMR preconditioning is begun by ramping power at 10 MWe/hr.
Which of the following is the result of these actions?
Fuel claddins failure is probable. 1 Fuel cladding stress is maintained within the vendor specifications.
.I The helium gas volume betweenthe fuel pelletsand fuel rod cladding will be larger than expected.
1I Fuel Dellet densification will cause hiah fuel temperature.
.91 JUST1FlCATlON:
Correct answer: "Fuel cladding failure is probable." PCIMOR Ramp rate of .I lKW/Ft/Hr or approximately 10 Mwelhr exceeded above Threshold power level of 11 KW/ft.
Fuel pellet densification will cause high fuel temperature. Incorrect. Fuel Densification is not of concern.
"The helium gas volume between the fuel pellets and fuel rod claddingwill be larger than expected."
Incorrect. The helium gas volume will be reduced by normal power operations, violating PCIOMR rules will further reduce the gap between the fuel pellets and cladding.
"Fuel cladding stress is maintained within the vendor specifications." Incorrect. By violating PCIOMR rules, the fuel cladding will be stressed beyond vendor recommended limits.
[LP NOH01RXFUEL 1 I
i
~
a.
b.
c.
' The acronym and purpose of PCIOMR.
The definition of threshold power.
The definition of enveloDe Dower.
1 ~-
Vision Bank QID##Q57174
. -~ - -
Sunday, May 11,200311:02:42AM
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-- Page 149 of 161
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Per NC.NA-APZ-0024, Radiation Protection Program, a 21 year old worker with 11 Rem Lifetime dose from the previous 3 years working at Hope Creek will have an administrative exposure control level of ( I ) mrem TEDE per year. This can be raised to a maximum of (2) mrem TEDE by the Radiation Protection Manager.
u (Assume NO delegation of authority)
___ ~
I (1) 2000 (2) 3000 (I) 2000 I (2)4000 I kd (1)3000 (2) 4500 (I) 3000 (2) 4750
~
b a
[Generic Knowledge and Abilities GENERlCj ___
) 3000 Incorrect. RP Supervisor approves extension to 3000 (1) 2000 (2) 4000 Correct per NC.NA-AP.ZZ-0024Attachment 1. Worker does not meet 5(N-17) threshold. 5(21-17) =20 Rem.
(I) 3000 (2) 4500 Incorrect. Automatic authorization during Emergency Plan implementation. II (1) 3000 (2) 4750 Incorrect. VP Operations approves extension to 4750. I 1 NC.NA-AP.ZZ-0024Attachment 1 -
I I
1 E: j ADMPROE059 Given a set of exposure conditions Identify the personnel responsiblefor approval of the following dose extension:
Yearly Dose Extension I Declared Pregnant Women Dose Extension I
- c. Lifetime Dose Extension IAW NC.NA-AP.ZZ-0024:
i I
____ ~- --J 7-I p i . -
Editorially Modified 1
i INPO Bank QlD# 21202 ?J6/
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-~ -
Sunday, May 11, 2003 p a g e 150 of-__
161
-~ _ _ _AM
_11.02:42 _ ..
~ _..___ _ _ ~ -- _ - __ - _ _
Given the following conditions:
- The plant is operating at 100 percent power.
L
- A steam leak is present on a manual valve packing in Main Steam Tunnel room.
- The work will take approximately 30 minutes.
- The RWP for the area is NOT current.
- The general area dose rates are estimated at I.5Whr.
Which one of the following will allow the maintenance work to be authorized?
One of the maintenance personnel is self monitor qualified. - -
All personnd involved in performing the work are volunteers and have been fully briefed on the hazards involved.
All work is documented in the Control Room OS/CRS Narrative log with the total dose receivedj The- job is provided with continuous Radiation
_ _ - . __ Protection coverage.
SRO 55.43(4)Radiation hazards that may arise during normal and abnormal situations.
i/
Correct: The job is provided with continuous Radiation Protection coverage. HCGS TS 6.12.l.c. allows use of continous RP coverage in place of RWP and specific dose rate info.
Incorrect: All personnel involved in performing the work are volunteers and have been fully briefed on the hazards involved. Requirement for Personnel Emergency Exposure Limit.
Incorrect: All work is documented in the Control Room OSCRS Narrative log with the total dose
,received. Not required.
,Incorrect: One of the maintenance personnel is self monitor qualified. Self monitor can not replace RP coverme
[HCGSTS 6.12.l.c. -4 II -
1 I
None Sunday, May.11, 2003_11:02:43 AM Page 151 of _161
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The moisture content of charcoaladsorber bed of the Which of the following parameter changes will occur L Rising GRW post-treatment radiation levels due to an increase in Krypton. Lower Cooler Condenser
___ - temperature. __-
Rising GRW post-treatment radiation levels due to an increase in Iodine. Raise Cooler Condenser temDerature. i Lowering GRW charcoal adsorber bed temperature. Lower offgas Dilution flow. i
'Lowering GRW charcoal adsorber bed hydrogen concentration. Raise offgas Dilution flow.
t-treatment radiation levels due to an increase in Krypton. Correct, Water on charcoal reduces adsorption process -> rad levels increase.
Rising GRW post-treatment radiation levels due to an increase in Iodine. Incorrect, iodine is soluble and should remain in the main condenser.
Lowering GRW charcoal adsorber bed temperature. Incorrect, water makes bed temperature rise due to the decay of already captured radioactive gases.
Lowering GRW charcoal adsorber bed hydrogen concentration. Incorrect, adsorber bed does not adsorb hvdroaen FP"0 1GASRWO-0 1 i i I
GASRWOEOOB ~ (R) Explainlidentify the effect of moisture in the process gas stream on the following components IAW available control room '
I references:
I I a. Recombiner I 1 b. Charcoal Beds
' J 1 ..
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i-p Sunday, May 11, 2003 11.02:43 AM ~ __ Page 152 of_-
161
___~ -.
Plant conditions are as follows:
- Reactor Power is at 70%.
u - Condenser Vacuum is 5.5" Hg absolute and degrading.
Which one of the following states immediate operator actions required?
Ensure turbine
___ sealing steam pressureis normal.
Trip the Main-Turbinewhen
- . .. - 350 Mwe is-reached.
. _-.-_ - - - - .. .. .- - .-- - i Place the standbv SJAE in-service.
Reduce reactor power. - . ___ . - - !
\Generic Knowledge and Abilities 294001G404
,GENERIC .-- 121
,Condition: Degraded Main Condenser Vacuum Action: Reduce Reactor Power as necessary to maintain Condenser vacuum < 5.0 " Hg Abs Justification:
Reduce reactor power.-Correct- See HC.OP-AB.BOP-0006 Ensure turbine sealing steam pressure is normal.-Incorrect- subsequent action A.2 Trip the Main Turbine when 350 Mwe is reached. -Incorrect- Retainment Override states < 300 Mwe vice 350 Mwe Place the standby SJAE in-service.-Incorrect- Subsequent action B Modified from 29320 Closed Reference Last used LOR 0006-05 Question Topic:Immediate Actions for Loss of Vacuum KA: 295002K3.09 [3.2/3.2] LOK F, LOD 2 Material Reauired for Examination: None
- HC.OP-AB.BOP-0006 None Sunday, May 11,2003 11.02.43 AM
~ _ _.
~~
Sunday, May 11,2003 11:02.43 AM
~
_ ~ _ __ ._ -_
I
~
Select the definition of the term, Minimum Alternate RPV Flooding Pressure. -.. ._
The lowest differential pressure between the RPV and the suppression chamber at steam flow through the minimum number of SRVs required for Emergency Depressurization is b sufficient to remove all decay heat from the core.
portion of the reactor core will generate sufficient steam flow through the specified number of open SRVs to prevent any clad temperature in the uncovered part of the core from exceeding 1800 F. __ ..
-b Memory IHope Creek 294001G417 i lowest RPV pressure at which steam flow through open SRVs is sufficient to preclude any clad temperature from exceeding 1500°F even if the reactor core is not completely covered. HC.OP-W EO.ZZ-LIMITS-CONV.
INCORRECT - The lowest differential pressure between the RPV and the suppression chamber at which steam flow through the minimum number of SRVs required for Emergency Depressurizationis sufficient to remove all decay heat from the core. Definition for Minimum RPV Flodding Pressure.
INCORRECT - The lowest differential pressure between the RPV and the suppression chamber at which the least number of SRVs can be opened, and will remove all decay heat from the core at a pressure sufficiently low that the ECCS with the lowest head will be capable of making up the SRV steam flow.
Definition for Minimum number of SRVs required for emergency de-pressurization.
INCORRECT - The lowest RPV pressure if at which Emergency Depressurization is commenced, the covered portion of the reactor core will generate sufficient steam flow through the specified number of open SRVs to prevent any clad temperature in the uncovered part of the core from exceeding 1800 F.
Definition for Minimum Zero lniection RPV Water Level.
/HC.OP-EO.ZZ-O206A
~HC.OP-EO.ZZ-LIMITS-CONVpg 27 of 60 1
__ -~ I EOP206E004 I Define the term Minimum Alternate RPV Flooding Pressure. I I !
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Sunday, May I I , 2003 I I :02:43 AM . -
Sunday, May 11,20031 1 02.43AM Page 156 of 161
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Given the following conditions:
- A reactor scram occurred due to a level transient where RPV level reached -60 inches.
- HPCl Aux Oil Pump failed to start for an unknown reason.
- RClC automatically initiated and restored level.
Which one of the following describes the reporting requirement, if any, via ENS line? I No report.
__ I One Hour report.
Four Hour repot% 1 Eight Hour report. - - 1 2.4 ' Emergency Procedures and Plan I SRO 55.43 (1) Conditions and limitations in the facility license.
10CFR55.43 (5)Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations..
Four Hour report. Correct R.A.L 11.3.1 HPCl should have actuated and injected to the vessel but did not.
Also 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> report on scram.
No report. Incorrect. Would be correct if this was preplanned sequence or test or if HPCl was out for scheduled maintenance.
One Hour report. Incorrect. Would be correct if RPV Level Safety Limit reached or Emergency Classification of UE, Alert, SAE, or GE reached.
Eight Hour report.
__ Would be correct_ if_ malfunctioning Aux Oil Pump was _ - -found prior to the event.
/HCGS ECG RAL 11.3.1 I
1 c
__ _ _ - ~ -
Given the following:
- A severe accident has occurred.
i-. - You have declared a General Emergency at 0337 hrs for loss of all three fission product barriers.
- The weather conditions are as follows:
- Clear skies
- Ambient temp = 35 degrees F Which of the following is the correct Protective Action Recomendation for the above conditions?
Evacuate all sectors 0 - 5 miles and Evacuate downwind sector +/- 1 sector 5 - I O miles.
Shelter all remaining sectors 5 - 10 miles.
Shelter all sectors.
Evacuate all sectors 0 - 5 miles and Shelter downwind sector +/- 1 sector 5 - 10 miles. Shelter all remaining sectors 5 - 10 miles.
Evacuate all sectors 0 - 5 miles.
1
~
a [Hope Creek 2.4 1 Emergency Procedures and Plan 2 12.1 14.01 L SRO 55.43 (1) Conditions and limitations in the facility license.
SRO 55.43(4)Radiation hazards that may arise during normal and abnormal situations.
LP NEPECDTYSC rev 00 Obj 5.0 ECG Attachment 4 Appendix 1 Evacuate all sectors 0 - 5 miles and Evacuate downwind sector +/- 1 sector 5 - 10 miles. Shelter all
,remaining sectors 5 - 10 miles. Correct answer. Loss of all barriers = I O pts. Question of Appendix 1 is answered yes. Weather conditions are not severe enough to warrant shelter instead of evacuation.
Shelter all sectors. Incorrect. Weather conditions are not severe enough to warrant shelter instead of ~
I evacuation.
Evacuate all sectors 0 - 5 miles and Shelter downwind sector +/- 1 sector 5 - 10 miles. Shelter all remaining sectors 5 - 10 miles. Incorrect. Evacuate downwind sectors.
Evacuate all sectors 0 - 5 miles. Incorrect. Would be correct if onlv 9 Doint GE.
IECG Attachment 4 Appendix 1 I
II
_-- I I -- __ - ~ ~ -
Appendix 1 from ECG Attachment 4 L Vision Bank QID# Q56439 Sunday, May 11,2003 11:02:44-~
AM ' Page 158 of 161
- __ -~
Following a reactor scram and loss of feedwater, the plant is being cooled down using HPCl in f u r flow recirculation. A review of the operating logs indicates that reactor pressure for the past two hours is as follows:
u Time Reactor Pressure (psig)
-------------------I------------
0000 950 0015 925
'0030 900 0045 850 0100 700 0115 650 0130 550 0145 300 0200 250 Based on these conditions, the cool down rate is administrative limits and Technical Specification limits.
within; within.
within; outside. . -
- - _ _ __ I outside; within.
- I outside; outside.
I 06/17/2003/
Application -eek L 294001G447 1 GENERIC I . _ _ I - -- 125, 2.4.47 /Ability to diagnose and recognize trends in an accurate and timely manner-utilizingthe appropriate ' 3.4
[control room reference material. -i Justification:
outside; outside. Correct. Between 0045 and 0145, cooldown reached 105 degrees within a one hour period.
outside; within. Incorrect. Exceeds both TC and Admin limits.
within; outside. Incorrect. Exceeds both TC and Admin limits.
within: within. Incorrect. Exceeds both TC and Admin limits.
jHC.OP-IO.ZZ-0004 I-HCGS TS 3.4.6.1 . .
__-___ I
-__ ____ - I RATED POWER TO COLD SHUTDOWN integrated Operating Procedure, supporting System Operating Procedures and I I Technical Specifications.
. __ -~
__ ~-
~-
__ 1 1
~~ ~
I I
L
-~
IMf.riat Required foPExaminaflon * '1
. HC.OP-IO.ZZ-0004Attachment 4. Steam Table form HC.OP-1O.Z-0008 Sunday: May 11,2003 11:02:44 AM
~ __-
iPage159of-
Significantly Modified 9557 12/1SA995 Hope Cree i
- Sunday,
- May 11,2003
~ ~ _ _ AM 11'02:44 - -
! Paae 160 of 161
Given the following conditions:
- - Suppression Pool Narrow Range Level instrument is removed from service for calibration.
- Wide Range Level instrument Channel C is reading 75 inches.
- Wde Range Level instrument Channel A is reading 73 inches.
Which one of -the -
following describes the action required, if any, and bases -. --
for your answer? I NO action ___ is required because the level would have been within- limits
_.__ - at the-time of removal. -- ]
NO action is required . __ because RCIC Suction swap would occur on an actual low level. . _
1 Makeup to Suppression Poollevel is required because the average level is belowthe limit.
-1 Makeup to SuDDression Pool level is required because level outside allowable limits. - - - I Makeup to Suppression Pool level is required because level is outside allowable limits. Correct. IAW NCNA-AP.ZZ-0005states "Station technicians and operators shall believe instrument readings and treat them as accurate unless proven otherwise."
NO action is required because the level would have been within limits at the time of failure. Incorrect.
Plausible misconception.
NO action is required because RClC Suction swap would occur on an actual low level. Incorrect. RClC does not swap on SP level.
Makeup to Suppression Pool level is required because the average level is below the limit. Incorrect.
I hC.NA-APIZZ-0005
_ _ 5.12.1 I I
- i I_-. . ........... ~ .....
~ .. -
0005. I Sunday, May 11,2003 5 0 2 4 4 AM Page 161 of 161
RO Q 1 / SRO Q 1 POWER TO FLOW 80 70 60 50 40 30 30 40 50 60 70
RO Q 1 / SRO Q 1 POWER TO FLOW c
110 100 90 80 70 60 50 40 30 20 10 0
0 IO 20 30 40 50 60 70 80 90 I00 110
ATTACHMENT 2 Page 1 of 1 ACCUMULATOR PRECHARGE NITROGEN PRESSURE VERSUS AMBIENT TEMPERATURE W
NITROGEN PRECHARCE PRESSURE (psig!
Figure 3-2. Accumulator Precharge Nitrogen Pressure Versus Amb'ient Temperature
( i RO Q 42 / SRO Q 50 ATTACHMENT 1 MAXIMUM RWCU RETURN TO FEEDWATER FLOW - 2 PUMP OPERATION Maximum RWCU Return to Feed Water Flow equivalent to Total RWCU F/D Bed Flow of 300 gpm @ 120°F with 2 pumps running 370.0 365.0 360.0
/
(0 355.0 -
350.0 - -
2 345.0 UJ -
0 340.0 _ ~ ~ ~ _
-2 335.0 If, the RWCU Return to FW Flow at the current -~
3 > 300 gpm at 120 degF.
325.0 ~ - -
Q, u.
8 320.0 ~ ~~~ ~
E
$ 315.0 -~ - - ~- ~ -I Good Region for RWCU FID Flow ~
p: If, the RWCU Return to FW Flow at the current 2 310.0 - - - Retum to FW temp. ISin this region, 3 Then, the Total RWCU Flow through FID beds is p: 305.0 or = 300 gpm at 120 degF.
300.0 295.0 -~ - -
70 90 110 130 150 170 190 210 230 250 270 290 310 330 350 370 390 410 430 450 470 RWCU Return to Feedwater Flow Temp. Crids A215 (degF)
(
RO Q 42 / SRO Q 50 ATTACHMENT 2 MAXIMUM RWCU RETURN TO FEEDWATER FLOW - 1 PUMP OPERATION Maximum RWCU Return to Feed Water Flow equivalent to Total RWCU F/D Bed Flow of 150 gpm @ 120°F with one pump running 185.0 h
E 181.0 -
-$ P cn 177.0 RWCU F/D Bed Flow Too High Region If, the RWCU Return to FW Flow at the current ~ ~
co e-4 , Return to FW temp. is in this region, I I a Then, the Total RWCU Flow through F/D beds is I a 173.0
'13 > 150 gpm at 120 degF.
(3 ' 1 1
@ 169.0
- 165.0 I~
~
I 0 j c, I 2maa 161.0 ~
0 Good Region for RWCU FID Flow LL If, the RWCU Return to FW Flow at the current 8 157.0 - ~ -~
Return to FW temp. is in this region, E
3 I
Then, the Total RWCU Flow through F/D beds is 153.0 c or = 150 gpm at 120 degF.
Lt 3
$ 149.0 Lt 145.0 70 90 110 130 150 170 190 210 230 250 270 290 310 330 350 370 390 410 430 450 470 RWCU Return to Feedwater Flow Temp. @ Crids A215 (degF)
RO Q 78 / SRO Q 77 L-AIR COMPRESSOR 0
RECEIVER COMPRESSOR ISW VLV HOSE ISW M Y 9
RECENER ISLNVLV HffiH PRESS CONNECTKlN
RO Q 99 SRO Q 99 ATTACHMENT 4 SHUTDOWN FROM RATED POWER TO COLD SHUTDOWN REACTOR COOLANT SYSTEM TEMPERATUREPRESSURE DATA (Page 3 of 3)
DATE R e a c t o r s t e a m D o m e R e a c t o r C o o l a n t S y s t e m T e m p e r a t u r e P r e s s u re c o n v e rte d to S a t u r a t e d T e m p .
R P V P r e s s + 14.7 = P S I A 5 0 0 P S I A I S t e a m Table Saturation Tern perature 4 0 0 3 0 0 2 12'F 2 0 0 H i g h e s t Recirc Suction Tern p .
o r R H R H x Inlet 0' 1 0 0 R W C U Bottom H e a d Drain 0 0.5 1 1 . 5 2 2.5 3 3.5 4 4.5 5 5.5 6 6.5 7 7.5 8 8.5 9 Note: 1. RETAIN completed Attachment 4 sheets with the on going procedure HC.OP-IO.ZZ-O004(Q).
- 2. RECORD temperatures in conjunction with HC.OP-DL.ZZ-O026(Q), Attachment 3s AND ENSURE operation to the right of the applicable curve in Tech Spec 3.4.6.1 as well as HC.OP-DL.ZZ-O026(Q), Attachment 3s.
- 3. Below 212°F water temperature must be read directly. The points are listed in order of preference (highest Recirc suction temperature, RHR Hx Inlet, RWCU Bottom Head Drain).
- 4. There must be forced flow past the temperature element in order to obtain a valid temperature reading.
- 5. Above 212°F Reactor Steam Dome pressure should be used to obtain the saturation temperature from the Steam Tables. This temperature should then be plotted.
RO Q 99 / SRO Q 99 ATTACHMENT 5 SHUTDOWN FROM OUTSIDE CONTROL ROOM REACTOR COOLANT SYSTEM TEMPERATURE/PRESSURE DATA (Page 4 of 4)
SATURATED STEAM TABLES TEMP O F ABS PRESS (PSIA) TEMP O F ABS PRESS (PSIA) 200 11.526 388 2 15.220 212 14.696 396 236.193 220 17.186 404 258.725 228 20.015 412 282.894 23 6 23.216 420 308.780 244 26.826 428 336.463 252 30.883 43 6 366.03 260 35.427 444 397.56 268 40.500 452 431.14 276 46.147 460 466.87 284 52.4 14 46 8 504.83 292 59.350 476 545.1 1 300 67.005 484 587.81 308 75.433 492 633.03 316 84.688 500 680.86 324 94.826 508 73 1.40 332 105.907 516 784.76 340 1 17.992 524 84 1.04 348 131.142 532 900.34 356 145.424 540 962.79 364 160.903 548 1028.49 372 177.648 556 1097.55 380 195.729 564 1170.10
RO Q 63 OVERHEAD ANNUNCIATOR WINDOW BOX A6 TABLE OF ATTACHMENTS
- ., 1 RHR LOGIC A RHR LOGIC C RHR LPCI RHR LPCI RHR A- OUT OF OUT OF LOOP A LOOP c PUMP ROOM SERVICE SERVICE INITIATED INITIATED FLOODED RHR RHR RHR PUMP A RHR PUMP C RHR MANUAL B- LOOP A LOOP c AUTO AUTO INIT SWITCH TROUBLE TROUBLE START START ARMED SACS SUPPLY SACS SUPPLY ECCS JOCKEY RHR A/C RHR S/D CLG C- RHR PUMP A RHR PUMP C PUMP lCP228 LPCI LINE SUCTION HDR TROUBLE TROUBLE TROUBLE BREAK PRESSURE HI ECCS A/HPCI RHR C RHR A S/D RHR HX CLG D- TRIP UNIT TRIP UNIT CLG & MIN WTR OUTLET
-. - TEST/INOP TEST/INOP FL VLV OPEN TEMP HI I I MAIN MAIN MAIN E- CONDENSER A CONDENSER B CONDENSER C VACUUM LO VACUUM LO VACUUM LO CONDENSATE CONDENSATE CONDENSATE CNDS RECIRC CONDENSATE F- TRAIN A TRAIN B TRAIN C CAPACITY DRAIN TANK TROUBLE TROUBLE TROUBLE EXCEEDED LEVEL HI/LO
i (
RO Q 80 ATTACHMENT 2 (Page 1 of 2)
SECONDARY CONTAINMENT BACKDRAFT DAMPERS HIGH TEMPERATURE ROOM TRIP SUPPLY RETURN SWITCH TRIP DAMPERS / POWER DAMPERS / POWER LOCATION LOCATION RWCU Hx's TSH-9429-1 135 PD-9429A1 I Rm. 4504 BC281 FU 8,9 None (4506) TSH-9429-2 135 PD-9429A2 I Rm. 4504 DC281 FU 8,9 (Room exhausts to RWCU Pipe Chase 4505)
RWCU Pipe Chase TSH-9438-1 135 None (4505) I I (Room supplied from RWCU Pipe Chase 4402 TSH-9438-2 135 and RWCU Hold Pump I CC281 FU6,7 I
Room 4502)
RWCU Hold Pump TSH-9457-1 135 PD-9457A1 I Rm. 4504 _
(4503) TSH-9457-2 135 PD-9457A2 I Rm. 4504 1,15 (Room exhausts to RWCU I Hold Pump Room 4502)
RWCU Pump (4405) TSH-9438-1 135 PD-9438F1 I Rm. 4404 I.............................................................................................
AC281 FU 6,7 I (RoomNone exhausts to RW( 2U TSH-9438-2 135 PD-9438F2 I Rm. 4404 CC281 FU 6,7 Pipe Chase 4402) c RWCU Pump (4403) TS H-9438-1 135 PD-9438EI/ Rm. 4404 AC281 FU 6,7 None
.................................................................................... (Room exhausts to RWCU TSH-9438-2 135 PD-9438E2 I Rm. 4404 CC281 FU 6,7 n:-- r - L - - -
I I r i p e cliiast: 44UZ)
RWCU Pipe Chase TSH-9438-1 135 PD-9438D1 I Rm. 4401 I AC281 FU6.7 1 None (4402) - ........................ (Room exhausts to RWCU TSH-9438-2 135 PD-9438D2 I Rm. 4401 CC281 FU 6,7 Pipe Chase 4505)
HPCl PiDe Chase TSH-9437-1 135 PD-9437A1 I Rm. 4328 AC281 FU 12,13 PD-9437B1 I Rm. 4326 (4327) TSH-9437-2 135 PD-9437A2 I Rm. 4328 CC281 FU 12,13 PD-943782 I Rm ..4326 South Pipe Chase TSH-9438-1 135 PD-9438C1 I Rm. 4320 AC281 FU 6,7 None (4321) - ............ ........................................................ (Room exhausts to RWCU TSH-9438-2 135 PD-9438C2 I Rm. 4320 CC28l FU 6,7 Pipe Chase 4402)
North Pipe Chase TSH-9439-1 135 PD-9439AIl Rm. 4328 AC281 FU 14,15 PD-9439B1 I Rm. 4328 I I ................................................................................................................................................................................................................................................................
(4329)
RClC Pipe Chase (4319) Room exhausts to RWCU
RO Q 80 ATTACHMENT 2 (Page 2 of 2)
SECONDARY CONTAINMENT BACKDRAFT DAMPERS HIGH TEMPERATURE ROOM TRIP SUPPLY RETURN SWITCH TRIP DAMPERS / POWER DAMPERS / POWER LOCATION LOCATION Steam Tunnel TSH-9428-1 150 PD-9428A1 I Rm 4317 AC281 FU 8,9 PD-9428B1 I Rrn. 441 1 AC281 FU 8,9 (4316)
TSH-9428-2 150 PD-9428A2 I Rrn. 4317 CC281 FU 8,9 PD-9428B2 I Rm. 441 I CC281 FU 8,9 RHR Hx TSH-9432-1 135 PD-9432A1 I Rrn. 4215 AC281 FU 1 0 , l l PD-943261 / Rm 41 12 AC281 FU 1 0 , l l (4113) TSH-9432-2 135 PD-9432A2 I Rrn. 4215 CC281 FU 1 0 , l l PD-9432B2 I Rrn. 41 12 CC281 FU 1 0 , l l RHR Hx (4109)
HPCl TSH-9433-1 TSH-9433-2 TSH-9434- 1 135 135 135 PD-9433A1 I Rrn. 4108 PD-9433A2 / Rrn. 4108 PD-9434A1 I Rm. 41 12 BC281 FU 10,11
............ ~ .............
DC281 FU 1 0 , l l AC281 FU 16,17 I PD-9433B1 I Rrn. 4205 PD-9433B2 I Rm. 4205 PD-9434B1 I Rm. 421 1 I BC281 FU 1 0 , l l DC281 FU 1 0 , l l AC281 FU 16,17
~ ............. ~ -.........................................................
(41 11) TSH-9434-2 135 PD-9434A2 I Rrn. 41 12 CC281 FU 16,17 PD-9434B2 I Rrn. 421 1 CC281 FU 16,17 RClC TSH-9435-1 135 PD-9435A1 I Rrn. 4108 BC281 FU 16,17 PD-9435B1 I Rm. 4209 BC281 FU 16,17 (41 IO) I TSH-9435-2 1 135 I PD-9435A21 Rrn. 4108 1 DC281 FU 16,17 1 PD-9435B21 Rrn. 4209 1 DC281 FU 16,17 I Torus TSH-9436-1 135 PD-9436A1 I Rrn. 4201 BC281 FU 12,13 PD-943661 I Rrn. 4215 BC281 FU 12,13 (4102) TSH-9436-2 135 PD-9436A2 I Rm. 4201 DC281 FU 12,13 PD-9436B2 I Rm. 4215 DC281 FU 12,13
'm 'I
'L 1
A 4
M ts 3
L
-d 00 CJ 0
RO Q 59 / SRO Q 61 CHANNEL B CHANNEL D I LOGICB I MAN INlT LOGIC D I
I LOGIC B/F INlT I 1
1 LOGlCD/H INlT I I
I RESET '
I I RESET HI DRYWELL I HI DRYWELL I PRESS LOGIC PRESS LOGIC RESET I
SRV I STEAM LINED STEAM LINE B 1108 PSlG SRV 1120 PSlG SRV PSV-FO13H CHANNEL B PSV-FO13P CHANNEL D SET OUT OF SVCE SET OUT OF SVCE RELIEF AUTO ALARMS AUTO ALARMS RELIEF
-0GIC LOGIC 0 0 P P E E N N ADS ACTUPiTlON TIMER 7 7 I I N N R H H M B B
RO Q 73 START UP LEVEL CONTROL I
' IU AND BYPA5 STOP VALVE:
SIU LEVEL VALVE BYPASS STOP VALVE VALVE TL D/P I'L VALVE D/P STPT DEMAND
SRO Q 67 I "INTERLOCK STATUS DISPLAY" I
("I .
-- I" lo 0
I TROLLEY POS I I1*1I I -55555 I I 1I I -55555 I I I D
0 lo1 INTERLK 00 BRIDGE REVERSE STOP #2 INTERLK
SRO Q 98 APPENDIX 1 PREDETERMINED PROTECTIVE ACTION RECOMMENDATIONS 1
1 PAR REQUIRED FOR GENERAL EMERGENCY
/'GE BASED '\\,
/ ON 10 >YES i ~ O R S 0-5 MILES 71
<i POINTS ON EVACUATE D O W N W I N D 2 I S E C T O R 5-10 MILES' SHELTER ALL REMAINING S E C T O R S 5-10 MILES 1
\ BARR1ER TABLE ,/ (See next page to determine downwind sectors)
\ ? /
NO EVACUATE ALL SECTORS 0-5 MILES DEFAULT PAR (any other GE)
CAUTION:
IF TRAVEL CONDITIONS PRESENT AN EXTREME HAZARD (SEVERE ICE, SNOW, WIND, FLOOD, QUAKE DAMAGE, ETC. ), CONSIDER SHELTER INSTEAD OF EVACUATE IN THE ABOVE SELECTED PAR I' I1
RO Q 48 PSEG Internal Use Only HC.OP-AR.ZZ-O011(Q)
OVERHEAD ANNUNCIATOR WINDOW BOX C6 TABLE OF ATTACHMENTS 1
I 2
1 I RSP/RSS RAD MONITOR MN STM LINE WMOTE RPV LOOSE .
A- TAKEOVER COMPUTER RADIATION TROUBLE Hr TROUBLE PNL 10C675 Page 2 Page 16 Page 17 DLD MN STM LINE MN STM LINE VIB MONITOR BOP B- SYSTEM RAD HI HI RADIATION PNL B/C COMPUTER ALAWTRBL OR INOP DOWNSCALE c3 74 TROUBLE Page 30 Page 42 Pane 47 Page 48 Page 57 RADIATION REACTOR CRD SEISMIC MON C- MONITORING CONTROL SYS HYDR UNIT PNL C673 SYSTEM ALARM/TRE3L INOPERATIVE TEMP HI TROUBLE C' Page 58 Page 81 Page 84 ARPM/RBM RPIS ROD OUT CRD THERMAL D- FLOW REF INOPERATIVE MOTION ACCUM MONITOR OFF NORMAL BLOCK TROUBLE ALAWRBL Page 98 Page 103 Page 105 RBM ROD CRD SCRAM RMCS E- UPSCALEOR DRIFT DISCH VOL DISPLAYS 2 INOPERATIVE NOT DRAINED NOP I 1 I
Page 120 Page 123 Page 127 I Page 132 ROD COMPUTER PT COMPUTER PT F- DO$$AL,E SYSTEM OVERTRAVEL RETURN I N '
TROUBLE TO NORMAL ALARM Page 134 Page 136 Page 146 Page 148 Page 149