Text

Final RO & SRO Written
	ML032320112
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/OL-03-302
	Download: ML032320112 (162)
	v • d • e

Using the attached power to flow map, answer the following:

- The plant is operating at the 87% rodline.

- Select the MINIMUM core flow at which the plant can operate at power and still be assured of avoiding power oscillations or instabilities.

25%

40%

\\Comprehension 1

Hope Creek 06/17/2003 1

.45%-Correct-IAW TS 3.4.1.l.b and HC.OP-AB.RPV-0002. 81% rodline with reduced rrp speeds, the boundry of the EXIT or Instability Region I 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 1

I~aterial,F&syi&kI fc? Examinatihn (,

. 11 HCOP-AB.RPV-0002 Power to flow chart with-reion labeling and line nomenclature removed.

Editorially Modified Monday, June 23,2003 7:23:55 AM 1

Page 1 of 162

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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?

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Reduce reactor power by 5 percent; Combustible gas concentrations may cause an explosion in the SJAE Aftercondenser.

Reduce reactor power by 5 percent; Offsite radiological r e l e a s e b e o v e allowable l%s at the South Plant Vent.

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Reduce reactor power by 6 percent; Offsite radiological release may be above allowable limits at the North Plant Vent.

Reduce reactor power by 6 percent; Combustible gas concentrations may cause an explosion at the MVP suction.

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Hope Creek 0611 712003 d

IS

,Application Emergency and Abnormal Plant EvolutionL AA2. i Ability to determine and/or interpret the following as they apply to LOSS OF MAIN CONDENSER

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VACUUM:

3.2 3.3

~ ~ 2. 0 2

/Reactorpower: Plant-Specific JUSTIFICATION:55.43(4) & (5) SRO Radiation hazards and procedure actions required by assessment 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.

Monday, June 23, 2003 7,2355 AM

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Page 2 of 162

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Monday, June 23,2003 7:23:55 AM

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Page 3 of 162

Given the following conditions:

- The plant is operating at 100 percent power.

- TACS is on the 'A' Loop of SACS.

- 'A' 1 E 4.16 KV bus 1 OA401 has de-energized due to bus fault.

y.i...

'4 Which one of the followi All RACS Pumps tri

'B' SACS Expansio RACS Head Tank

~~ overflows

- due to makeup valve power loss.

'A' & IC' SACS ~-

Pump trip due to

~~ LO-LO-LO Expansion-Tank.

level.

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B lMemory Hope C$

06/17/2003 C

I 295003K105 3

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apply to PARTIAL OR

.6 2.7 The makeup valve solenoid to RACS Head tank fails open on a lo

'B' SACS Expansion Tank overflow Makeup uses an MOV and inverter lowers.

bus restoration.

A & C SACS Pump trip due to L reasons. A trins due to loss of D loss to TACS return valves. Incorrect. SACS ET "A" tank to overflow. B Tank el goes high. Pumps trip on Expansion Tank level. A&C Pumps trip but for the wrong iDs on low pump differential pressure.

/

Reference Title I

HC.OP-AB.=-0170 Hi50 P-G P. P B-o o o I -.__-

Channel, Abnor I Operating Procedure.

r"

/

I

[Material Required for Ex$ination 1 None Question Modification Method:

Page 4 of 162

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 Coo la nt Accident.

Which of the following is the response of the LOP and LOCA sequencers for these conditions?

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As soon as power is restored to the buses, the LOCA sequencer will control e 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.

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The LOP sequencer will begin to sequence until the diesel generator output breakers close, then the LOCA sequencer will complete load restoration.

a

[Comprehension I Hope Creek 0611 712003 295003K204 4

I OMPLETE LOSS OF A.C. PO 3.4 3.51 er will control the restoration 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 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 close, then the LOCA 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.

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I Reference Title HC.OP-AB.=-0135 Learning Objectives 0 3 - (R) Discussthe operational implications of the abnormal indicationslalarms for system operating parameters related to Station BlackouULoss Of Offsite Power Diesel Generator Malfunction, Abnormal Operating Procedure.

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IMatehal Required for Examination I None L

Facility Exam Bank Monday, June 23, 2003 7 23% AM Direct From Source Page 5 of 162

Vision Bank QID#Q61290 Monday, June 23,2003 7 23.55 AM Page 6 of 162

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Plant co nd it ions are as

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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:

block all control rod movement because the reactor has scrammed.

allow control rod insertion using the Continuous Insert PB only.

automatically bypass RWM blocks due to the effects of colder feedwater actively enforce control

~~ rod

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blocks

~ due

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to loss of First Stage Turbine pressure.

I 0611 712003 B

IComprehension I Hope Creek C

I 5

utomatically bypass RWM blocks due to the effects of colder feedwater. Correct. 20 percent power is 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 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 turbine pressure no longer input to cause rod blocks. Input removed by DCP.

I Reference Title H C. 0 P-SO. S F-0003

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RODMINE005 (R) Give con orth Minimizer to initiate a rod block signal IMateriaI Required for Examination 1 None Significantly Modified INPO BANK QID# 861 Significantly rnodif Monday, June 23, 2003 7.2355 AM Page 7 of 162

Conditions are as follows:

- 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.

d Which of the following is a safety implication (if any) of this condition?

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There are

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no safety implications

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because the TSV EOC-RPT trip is a I-out-of-2 logic.

There will be an excessive thermal margin upon EOC-RPT actuation if these TSVs close at

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power

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Reactor safetv has been enhanced by the overly conservative trip value for TSV closure.

Void reactivity

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feedback may exceed control rod reactivity if these TSVs

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close at power.

Hope Creek 06/17/2003 d

S komprehension 1

Emergency and Abnormal Plant Evolutions 1

2 295005G225 6

5 Main Turbine Generator Trip I

I 2 2 Equipment Control 2.2.25 Knowledge of bases in technical specifications for limiting conditions for operations and safety limits.

2 5 3 7 Justification 55.43(2) IAW TS Bases 3/4.3.4, TS 3.3.4.2 & Table 3.3.4.2 The purpose of the EOC-RPT system is to recover the loss of thermal margin that occurs at the end of a cycle. Void reactivity 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 5% closed ( 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 I-out-of-2 logic-Incorrect-The TSV closure uses various combinations, not 1 of 2 twice.

Reference Title 1

HCTS Bases.3.3.4.2

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I - Settings. (SRO/STA

~~ Only)

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MNTURBE022 (R) Given a scenario ofapplicable operating conditions and access to Technical Specifications:

a.

b.

C

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Select those sections applicable to the Main Turbine Evaluate Main Turbine operability and determine required action(s) based upon inoperability ical specification items associated with the Main Turbine (SRO ONLY).

Technical Specification 3.3.4.2 and Table 3.3.4.2-2 Monday, June 23, 2003 7:23:56 AM Page 8 of 162

Monday, June 23, 2003 7-2356 AM Page 9 of 162

Given the following conditions:

Question Source:

INPO Exam Bank

- 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 Question Modification Method:

Editorially Modified 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 scram speed will be:

da slower

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and

~ will result in MCPR LCO penalties.

slower and will result in reduced reactivity addition rates.

faster and may result in MCPR LCO penalties.

B Comprehension 06/17/2003 normal Plant Evolutions 1

295006K205 295006 7

AK2. Knowledge of the interrelations between SCRAM and the following:

AK2.05 CRD mechanism 3.1 3 3 L

Justification:

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.

scram speeds will be slower and will result in MCPR LCO penalties. Incorrect. Scram speeds will be faster.

scram speed will be slower and will result in reduced reactivity addition rates. Scram speeds will be higher.

scram speed will be faster and may result in MCPR LCO penalties. Incorrect. MCPR LCO penalties result from slow scram speeds Reference Title HC.OP-SO.BF-0002 CRDHYDE021 IR) Given the "Accumulator Precharge Nitroqen Pressure Versus Ambient Temperature Curve", determine the proper accumulator precharge gas pressure: IAW HC.OP-SO.BF-0002.

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Monday, June 23,2003 7.2356 AM Page 10 of 162

lnpo Bank QID# 14129 03/26/2001 Peach Bottom modified for Hope Creek Monday, June 23, 2003 7:23:56 AM Page 11 of 162

Given the following conditions:

- The plant is operating normally at 100 percent power

..- - 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) tially due to lower void fraction, then lowers rapidly due to scram.

Rises initially

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due to lower void fraction, then stabilizes at a slightly higher power level.

Lowers initially due to greater feedwater heating, then lowers rapidly due to scram.

Lowers initially due to greater feedwater heating, then stabilizes at a slightly higher power level.

b B

IComprehension 06/17/2003 Emergency and Abnormal Plant Evolutions 1

07K202 295007 High Reactor Pressure a

AK2. Knowledge of the interrelations between HIGH REACTOR PRESSURE and the following:

AK2.02 Reactor Dower 3 8 3 8 tion:

Rises initially 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 >I030 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.

Reference Title I

HC.OP-AB. RPV-0005 Learning Objectives ABRPVSE004 Explain the reasons for how planthystem parameters respond when implementing Reactor Pressure

[Material Required for Examination 1 None Question Modification Method:

d Monday, June 23, 2003 7:23:56 AM Page 12 of 162

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.

J The reason for these automatic actions is to prevent:

steam voiding in the RHR pump seals.

overpressurizing the RHR pump seals.

establishing a drain path from the RPV to the torus.

overpressurizing the shutdown cooling suction piping.

reek 06/17/2003 Memory Abnormal Plant Evolutions i

1 9

the following responses as they apply to HIGH REACT0 3.0 3.2 hutdown cooling suction piping. Correct. The piping isolates to minimze offsite 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..

steam voiding in the RHR pump seals. Incorrect. Reason for seal coolers on RHR pumps.

I Reference Title 1

HCGS TS Bases 3.3.2 NSSSSOE007 Given the design bases of the NSSSS, state the purpose of that design function IAW the NSSSS Lesson Plan

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ne Editorially Modified 8896 04/06/1998 Fermi 2 Modified for Hope Creek Monday, June 23, 2003 7.23.56 AM Page 13 of 162

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

- 'B' 3.6 Mlbmlhr

- IC' 3.5 Mlbm/hr

.LJ Main Steam flows are as follows:

- 'A' 2.6 Mlbm/hr

- 'B' 2.5 Mlbm/hr

- 'C' 2.6 Mlbm/hr

- ID' 2.4 Mlbm/hr Based on these conditions, RFP speed demand RPV Level alarm.

must be applied to prevent the 06/17/2003 2

08K103 10 oncepts as they apply to HI WATER LEVEL:

tion:

decrease, 7; Correct. Total Feed Flow is 10.3 Mlbm/hr. Total Steam flow is 10.1 Mlbm/hr. this mismatch will result in an increasing RPV water level. RPV speed demand must lower to prevent the Level 7 alarm The Level 7 alarm will occur if no action is taken.

increase, 4; Incorrect. Wrong action; wrong alarm.

decrease, 4; Incorrect. Correct action, wrong alarm.

increase, 7; Incorrect. Wrong action; correct alarm.

Reference Title I

HC. OP-SO.AE-~OO I

the Feedwater Control System Lesson Plan

]Material Required for Examination I None Monday, June 23, 2003 7:23:56 AM Page 14 of 162

Question Source:

New Question Source Comments:

Monday, June 23, 2003 7:23:57 AM Question Modification Method:

Page 15 of 162

Using the attached transient analysis plots of a reactor scram, which one of the following failures caused the scram?

EHC Pressure Regulator Failure to 0%.

Master Level Control Setpoint Failure to 0 inches.

EHC Pressure Regulator Failure to 130%.

Master Level Control Setpoint Failure to +60 inches.

0611 712003 1

2 9 500 9A20 1 Low Reactor Water Level 11 determine and/or interpret the following as they apply to LOW REACTOR WATER LEVEL.

4.2 4.2 55.43(1) Conditions and limitation of the HC UFSAR Master 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 figure 15.2-8 with title block removed I

Reference Title 1

UFSAR 15.2 Learning Objectives EO101 LE006 (R) Given any step of the procedure, describe the reason for performance of that step andlor expected system response to control manipulation prescribed by that step

]Material Required for Examination I Provide FSAR figure 15.2-8 with title block removed.

Question Source:

Facility Exam Bank Question Source Comments:

Vision bank QID# Q56686 Monday, June 23, 2003 7:23:57 AM Direct From Source Page 16 of 162

Given the following conditions:

- The plant is operating normally at 100 percent power.

- 'B' Reactor Feedwater pump trips.

- Assume NO operator actions taken.

Material Required for Examination Question Source:

New 1

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.

Intermediate Reactor Recirc Runback. Reduce power to prevent power oscillations.

b R

Memory Hope Creek 06/17/2003 Emergency and Abnormal Plant Evolutions 1

1 09K301 Knowledge of the reasons for the following responses as they apply to LOW REACTOR WATER LEVEL 295009 Low Reactor Water Level 12 AK3 AK3 01 Recirculation pump run back. Plant-Specific 3.2 3 3 Correct. Intermediate Reactor Recirc Runback. Reduce power to control level. The RFP trip runback reduces reactor power and demand on the Feedwater system to within the capability of 2 RFP operation.

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. Wrong Runback Wrong reason. Bases for Suction valve position interlock.

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I Reference Title I

NOH01 RECCON-01 RECCONE002 From memory, state the purpose of the following Recirculation Flow Control components IAW the Recirculation Flow Control System Lesson Plan a

Recirc MG drive motor b

Fluid Coupler c

Generator

d.

Scoop Tube Positioning Unit e

Exciter f

Tachometer g

Individual pump speed controllers

h.

i.

Startup signal generator

j.

Error limiting circuit

k.

Scoop tube positioning unit I.

Signal Failure Detector

m.

Master Speed Controller Speed limiters #1 and #2 Monday, June 23, 2003 7:23:57 AM Page 17 of 162

Monday, June 23, 2003 7:23:57 AM Page 18 of 162

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 SRVs

- 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 Emergency Operating Procedures?

Initiate drywell sprays ONLY.

Initiate Emergency Depressurization and Drywell Sprays.

Initiate Emergency Depressurization and Suppression Pool Cooling.

Initiate Suppression Pool Cooling and Drywell Sprays.

06/17/2003 C

S Application reek 29501 26406 Emergency and Abnormal Plant Evolutions 2

29501 2 High Drywell Temperature 13 2 4 2 4.6 Knowledge symptom based EOP mitigation strategies.

3.1 4.0 Emergency Procedures and Plan 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 DWTT-4 thru DW/T-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 temDerature of 350°F. Drvwell SDrav is preclude Reference Title 1

HC.OP-EO.ZZ-0102, Step DW/T-3 thru DW/T-6 & DW/P-6 Monday, June 23, 2003 7.23.57 AM Page 19 of 162

Material Required for Examination Question Source: 1 Facilitv Exam Bank I EOP flowcharts Question Source Comments:

Vision Bank QID Q56006 editorially Editorially Modified modified.

Monday, June 23, 2003 7:23:57 AM Page 20 of 162

Question Source:

Given the following conditions:

New

- The plant is operating at 100 percent power.

- Turbine Building Chiller AKI 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.

u Which one of the following actions is required?

ly scram the reactor at 1.5 psig Drywell pressure.

Manually scram the reactor at 135 F Drywell temperature.

Shutdown Recirc pumps and Drywell Coolers and initiate Drywell Sprays at 1.5 psig Drywell pressure.

Shutdown Recirc pumps and Drywell Coolers and initiate Drywell Sprays at 135 F Drywell temperature.

a B

Memory 0611 712003 Emergency and Abnormal Plant Evolutions 2

14 295012 High Drywell Temperature AKI Knowledge of the operational implications of the following concepts as they apply to HIGH DRYWELL 3.1 3.2 Manually scram the reactor at 1.5 psig Drywell pressure. Correct. Retainment override step 1.a. 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 DWTT-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 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 D W - 3 allows spraying the DW between 135 and 340 degF only if DWP falls in SAFE area of curve DW-P. This corresponds to about 3 2 pslg DWT risina due to a loss of DW coolina not a LOCA.

u I

Reference Title HC.OP-AB.CONT-0001 ABCNTSEOOG (R) Explain purpose of and the bases for Retainment Override Steps in Irradiated Fuel Damage

[Material Required for Examination 1 None Monday, June 23,2003 7:23:57 AM Page 21 of 162

Monday, June 23,2003 7:23:58 AM Page 22 of 162

The following plant conditions exist:

Question Source Comments:

- The reactor is at full power.

- Torus Cooling is in operation and average temperature is increasing

- HPCl testing is in progress.

\\

19639 06/14/2001 Fermi 2 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).

(1) 95 F; (2) 110 F (I) 105 F; (2) 110 F (1) 105 F; (2) 120 F (1) 110 F; (2) 120 F b

R Application Hope Creek 06/17/2003 Abnormal Plant Evolutions 1

1 29501 3A102 pression Pool Temperature 15 AAI. Ability to operate and/or monitor the following as they apply to HIGH SUPPRESSION POOL TEMPERATURE:

ms that add heat to the suppression pool 3.9 3.9 Justification:

(1) 105, (2) 110 Correct. HPCI testing is required to be terminated at 105 degF. If Temp exceeds 110 degF, place the mode switch to shutdown.

(1) 95; (2) 110 Incorrect. 95 is LCO for normal / non-heat adding conditions.

(1) 105; (2) 120 Incorrect.

(1) 110; (2) 120 Incorrect I

Reference Title 1

HCGS TS 3.6.2.1 subsequent actions IAW the Primary Containment Control - Suppression Pool Lesson Plan Given specific plant operating conditions, and a copy of the Hope Creek Generating Station Technical Specifications, determine the following a If a Limiting Condition for Operation has been exceeded b If a Limiting Safety System Setting has been reached andlor exceeded c If a Safety Limit has been violated TECSPCEOOB Ilklaterial Required for Examination 1 HCGS Tech Specs section 3.6; EOP Flow charts Monday, June 23, 2003 7:23:58 AM Direct From Source Page 23 of 162

A startup is in progress with the following conditions:

- Reactor pressure is stable at 170 psig.

- Two turbine bypass valves are full open.

- Control rods are being withdrawn.

- IRMs are between 30 and 70 on range 8 Which of the following would occur if all turbine bypass valves were to fail closed and why?

(Assume NO operator action)

The reactor would scram due to high flux.

The

~ reactor would scram due to high pressure.

Reactor power would increase and stabilize due to the change in coolant temperature.

Reactor power would decrease and stabilize due to the change in void fraction Hope Creek 06/17/2003 a

B Comprehension 29501 4K201 Emergency and Abnormal Plant Evolutions 1

1 16 Inadvertent Reactivity Addition AK2. Knowledge of the interrelations between INADVERTENT REACTIVITY ADDITION and the following:

AK2.01 RPS 3.9 4.1 The reactor would scram due to high flux. Correct. TBV closure will cause pressure rise and void collapse. Power will rise. With IRMs not ranged, a reactor scram would result.

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 coolant temperature. Incorrect. Would not stabilize with TBVs failed closed.

Reactor power would decrease and stabilize due to the change in void fraction. Incorrect. Reactor power would increase.

r Reference Title HC.OP-AB. RPV-0005 Reactor Pressure Drywell Temperature Steam Flow

[Material Required for Examination I None Bank Editorially Modified INPO BANK QID #21232 Dresden 06/14/2002 Mo Monday, June 23,2003 7.23:58 AM Page 24 of 162

Given the following conditions:

- The plant was operating at 100% power when the reactor scrammed.

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 The reactor is:

in a cold shutdown rod configuration with forty control rods at position 02.

in a cold shutdown rod configuration with forty control rods out further than 02.

only subcritical at the present reactor temperature with forty rods at position 02.

only subcritical at the present reactor temperature with forty rods out further than 02.

OW1 7/2003 1

01 5K101 17 J

AKI. Knowledge of the operational implications of the following concepts as they apply to INCOMPLETE SCRAM:

3.6 3.9 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.

1 Reference Title 1

'EOP 102 HC.OP-SO.SF-0003 5.4.3 Learning Objectives E01O~AE004~ Exilain The significance of'Maximum Subcritical Banked Withdrawal Position" and state its value RODMINE003 (R) Given a labeled drawing of, or access to, the RWM Operator Display on 10C651, or the RWM Computer Display (in the Computer Room) a Explain the function of each indicator 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 d Assess plant conditions or permissives required for the control switches/pushbuttons to perform their intended functions Material Required for Examination I None Question Source:

INPO Exam Bank Monday, June 23,2003 7:23.58 AM Question Modification Method:

Direct From Source Page 25 of 162

7937 Hatch 03/14/1997 Monday, June 23, 2003 7:23:58 AM Page 26 of 162

Given the following conditions:

- 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.

L, Based on these conditions, place CREF in service with CREF boundry dampers in I

Mode and operators must OA; ISOLATE; remain in the Control Room.

OA; NORMAL; establish control at the RSP.

RECIRC; ISOLATE; remain in the Control Room.

RECIRC;

~~

NORMAL; establish control at the RSP.

B Memo 06/17/2003 C

Emergency and Abnormal Plant Evolutions 1

Control Room Abandonment 18 AK2 AK2 03 Control room HVAC 2.9 3 1 Knowledge of the interrelations between CONTROL ROOM ABANDONMENT and the Justification:

RECIRC; ISOLATE; remain in the Control Room. Correct. Retainment override step I of HC.OP-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 ABHVCI E004 Explain the reasons for how plantlsystem parameters respond when implementing HVAC (Material Required for Examination I None Question Source:

New Question Source Comments:

Question Modification Method:

Monday, June 23,2003 7:23:58 AM Page 27 of 162

Due to a fire in the Control Room console, the Control Room Supervisor orders the Control Room immediately evacuated. The reactor was scrammed remotely.

Question Source:

Which of the following statements describes how a scram is verified in accordance with Shutdown from Outside the Control Room, HC.OP-IO.ZZ-008?

New HCU accumulator pressure verified to be 950 - 1000 psig at each HCU.

~ SPDS display terminal "Rods Full In" in the TSC.

Reactorvessel pressure stable at 920 psig.

RPS Backup Scram Air Solenoids verifed de-energized.

Memo 0611 712003 S

1 295016K301 295016 Control Room Abandonment 19 AK3 Knowledge of the reasons for the following responses as they apply to CONTROL ROOM ABANDONMENT:

tor SCRAM 4 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 (SPDSKRIDS (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."

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 safetykelief 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 they were energized, that would indicate a reactor scram.

Reference Title I

HC.OP-IO.ZZ-0008 UFSAR 7.4.1.4 Learning Objectives iOP008E002 Determine if all Prereauisites have been met mor to imDlementation of the SHUTDOWN FROM OUTSIDE THE CONTROL ROOM Integrated Opiating Procedure.

[Material Required for Examination I None Monday, June 23, 2003 7:23:59 AM Page 28 of 162

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 0.0111 hours 6.613757e-5 weeks 1.522e-5 months ago.

- 'A' Mechanical Vacuum 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 by HC.OP-SO.CG-0001, Condenser Air Removal System Operation, 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.

Swap to the standby MVP to reduce effluent levels in the SPV.

Stop the MVP to stop release to the SPV Open

~

the Main Condenser Vacuum Breakers to stop release to the SPV.

0611 7/2003 1

29501 7G132 igh Off-Site Release Rate 20 2.1 Conduct of Operations 2 1 32 Ability to explain and apply system limits and precautions.

3.4 3.8 Justification:

Stop the MVP to stop release to the SPV. Correct. Required because the HIGH Setpoint is reached HC.OP-SO CG-0001 Caution 5.8.13.

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.

I Reference Title 1

HC.OP-SO.CG-0001 Caution 5.8 13 HC. OP-AB. CONT-0004 ABBOPGE007 (R) Explain the bases Condenser Vacuum.

From memory. summarize/identify the purpose of Condenser Leak Detection IAW the Lesson Plan CNDLEKEOOI IMaterial Required for Examination I None 1

1 New Question Source Comments:

Monday, June 23, 2003 712359 AM Question Modification Method:

Page 29 of 162

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.

When a second control rod drifts due to the Cooling Water Flow Control Valve failing open IAW HC.OP-SO. BF-000 1 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-AB.COMP-0001 Instrument and/or Service Air.

d S

Memory 0611 712003 mergency and Abnormal Plant Evolutions 2

29501 9A201 29501 9 Partial or Complete Loss of Instrument Air 21 AA2 AA2.01 llnstrument air system pressure 3 5 3 6 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMP INSTRUMENT AIR.

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 HC.OP-SO.BF-0002 Individual HCU Operation. Incorrect. Wrong reason. Low accumulator pressure is a result of a rod scram, not the cause.

I Reference Title H C. 0 P-AB. CO M P-000 1 Learning Objectives ABCMPlE007 (R) Explain the bases for Subsequent Actions and the information contained in the Discussion Bases Section of Instrument ahdlor Service Air (Materiai Required for Examination Question Source:

New I None Monday, June 23, 2003 7:23:59 AM Page 30 of 162

Question Source Comments:

Monday, June 23, 2003 7:23:59 AM Page 31 of 162

Given the following conditions:

- The plant is operating at 100 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.

Which one of the following is the effect on the Condensate/Feedwater System?

Secondary Condensate pump Min Flow valves fail closed.

SJAE/SPE Bypass Valve PDV-1719 fails open.

Primary Condensate pump Min Flow valve HV-1710 fails open.

Feedwater heater dump valves fail closed.

B IMemory 06/17/2003 ncy and Abnormal Plant Evolutions 2

295019K207 Partial or Complete Loss of Instrument Air 22 AK2. Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR and the following:

3.2 3.2 E/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.

Primary Condensate pump Min Flow valve HV-1710 fails open. Incorrect. Motor Operated Valve Feedwater heater dump valves fail closed. Cascading drain valves fail close, but high level dump valves fail open.

Reference Title H C. 0 P-AB. C 0 M P-0 00 1 Learning Objectives ABCMPI E004 Explain the reasons for how plantlsystem parameters respond when implementing Instrument andlor Service Air llaterial Required for Examination I None Question Source:

New Question Source Comments:

Monday, June 23,2003 7 23 59 AM Question Modification Method:

Page 32 of 162

Given the following conditions:

- An A-TWS occurs from 100 percent power.

- As corrective actions are being taken, the MSIVs inadvertantly isolate from a spurious high steam

- Other MSlV closure interlocks are clear.

- Visual inspection of the steam tunnel show NO abnormalities.

- Main condenser vacuum is 3 InHgA.

- Reactor coolant activity levels are normal.

- tunnel temperature signal.

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.

RPV Level is less than -129 inches.

Emeraency depressurization is anticipated.

a '

B Application 06/17/2003 295020K102 8

Emergency and Abnormal Plant Evolutions 2

Inadvertent Containment Isolation 23 AKI Knowledge of the operational implications of the following concepts as they apply to INADVERTENT CONTAINMENT ISOLATION:

AKI.02 lPower/reactivity control 3.5 3.8 Justification:

Reactor power is 10 percent. Correct Step RC/P-17 allows reopening if boron injection is required, (>4%

power) and Main condenser 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. - Depressurization through TBV's is not allowed due to the reactor is not shutdown under all conditions without boron.

HC. OP-EO.ZZ-010 1 A Learning Objectives

~~p E0101AE008p (R) Given any step of the procedure, explain the reason for performance of that step and/or evaluate the expected system response to control manipulations prescribed by that step

/Material Required for Examination I EOP Flowcharts Monday, June 23, 2003 7:23:59 AM Page 33 of 162

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.

Reference Title L

Question Source:

What effect does this have on Drywell Cooling?

~~

~

Drywell Cooler fans A I through H I trip but can be restarted New

~~ D G e l l Cooler fans A2 through H2 trip and CANNOT be restarted.

Drywell Chilled Water Isolation valves close but all Drywell Cooling fans remain running.

well Cooler fans A2 through H2 trip and Chilled Water Isolation valves close.

B Memory 06/17/2003 mergency and Abnormal Plant Evolutions 2

Inadvertent Containment Isolation 24 AK2. Knowledge of the interrelations between INADVERTENT CONTAINMENT ISOLATION and the following:

AK2.03 Drywellkontainment ventilationkooling: Plant-Specific 3.1 3.3 Justification:

Drywell Chilled Water Isolation valves close but all Drywell Cooling fans remain running. Correct Cooler fans powered from A and B channels.

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.

HC. OP-SO.SM-0001 Learning Objectives DWVENTE008 (R) From memorv. summarize the interrelationshit3 between the Drvwell Ventilation System and the followinq systems IAW the

~I Drywell 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 IMaterial Required for Examination 1 None Monday, June 23, 2003 7:24:00 AM Question Modification Method:

Page 34 of 162

Given the following conditions:

The plant is in Operational Condition 4 at 180 degF.

The reactor scrammed on startup from a refueling outage.

RHR Loop A operating in Shutdown Cooling.

The B 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.

H C. 0 P-AB. R PV-00 09, S h utd ow n Cooling, is entered.

Which of the following will be adequate as an Alternate Decay Heat Removal method for the conditions above?

Crosstie C or D RHR pump for heat removal.

Inject

~

with Condensate Transfer and reject with RWCU.

Use natural circulation and Drywell coolers for heat removal.

Inject with one Core Spray pump from the CST to the RPV.

0611 712003 ormal Plant Evolutions 2

295021 K302 25 AK3 Knowledge of the reasons for the following responses as they apply to LOSS OF SHU ing and bleeding reactor vessel 3 3 3 4 Justification Inject with Condensate Transfer and reject with RWCU. Correct. Feed and bleed will maintain temperatures at low decay heat loads such as at BOL.

Crosstie C or D 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.

HC.OP-AB.RPV-0009 Learning Objectives ABRPVSE007 (R) Explain the bases for Subsequent Actions and the information contained in the Discussion Bases Section of Shutdown Cooling.

[Material Required for Examination 1 None Facility Exam Bank Significantly Modified Vision Bank QID # (261332 significantly Monday, June 23, 2003 7:24:00 AM Page 35 of 162

The plant is operating at 100% power when a CRD Temperature High alarm is received.

Question Source:

Which one of the following could have caused the CRD high temperature condition?

Eroded CRD cooling water orifice.

Stabilizing valve failed fully open.

CRD pump low discharge pressure.

06/17/2003 2

295022K302 26 nses as they apply to LOSS OF CRD 2.9 3.1 HC.OP-BD.IC-0001 Effects of Loss of CRD regulating Function section 3 CRD High temperatures can 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 Eroded 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 Facility Exam Bank Reference Title HC.OP-BD.IC-0001 Learning Objectives ABlCOl E007 (R) Explain the bases for Subsequent Actions and the information contained in the Discussion Bases Section of Control Rod IMaterial Required for Examination I None Direct From Source L

Monday, June 23,2003 7:24:00 AM Page 36 of 162

Given the following conditions:

- The plant is in OPCON 5.

- Core offload is in progress.

- A spent fuel bundle is full up on the main hoist over the core.

- Subsequently, the refuel bridge spotter notices the fuel bundle has become unlatched and has fallen into the vessel.

- The water clarity has degraded significantly.

- A short time later, the following Refuel Floor Rad Monitors alarm:

1

- Spent Fuel Pool ARM.

- New Fuel Criticality ARM.

I Pu tHebc4-7e Based on these conditions, what operator action is required?

E\\

Suspend all refueling operations.

Remove the Fuel Pool Cooling System from service to reduce Reactor Building radiation levels.

Initiate action to establish Secondary Containment within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

Determine the location of the dropped bundle, inform the CRS, and evacuate the Refuel Floor Hope Creek 06/17/2003 a

S 1Memot-y i

Eme and Abnormal Plant Evolutions 3

1 2950236449 295023 Refueling Accidents 27 2.4 Emergency Procedures and Plan 2.4.49 Ability to perform without reference to procedures those actions that require immediate operation of 4 0 4 0 v

system components and controls.

Justification:

SRO 1 OCFR55.43 (7) Fuel handling facilities and procedures.

SRO 55.43 (6) Procedures and limitations involved in alterations in core configuration.

SRO 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 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months - 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 operator should not wait to determine the location of the dropped bundle..

I Reference Title 1

HC.OP-AB.CONT-0005 Learning Objectives ABCNTSE003 (R) From memory, recall the Immediate Operator Actions for Irradiated Fuel Damage

[Material Required for Examination I None Monday, June 23, 2003 7:24.00 AM Page 37 of 162

1 Editorially Modified Bank QID# (255912 Monday, June 23,2003 7:24:00 AM Page 38 of 162

Which of the following indications will positively identify a criticality event in progress while a fuel bundle is being lowered into the core during refueling operations?

Source range monitor spiking repeatedly A sustained upward trend on the nearest source range instrument to the fuel bundle location.

The

~~ high refuel floor radiation alarm sounds.

Refuel bridge hoist motion interlock activates.

~

Memory Hope Creek 0611 712003 b

IB and Abnormal Plant Evolutions 3

1 23K103 eling Accidents 28 AKI. Knowledge of the operational implications of the following concepts as they apply to REFUELING ACC I DENTS:

AKI.03 Inadvertent criticality 3.7 4.0 Justification.

A sustained upward trend on the nearest source range instrument to the fuel bundle location. Correct Responsibility of the Control Room refueling monitor is to monitor for unexpected increasing count rate.

Source range monitor spiking repeatedly. Incorrect. Indications of a detector failure.

The 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 criticality in the core.

Reference Title 1

HC.RE-AP.ZZ-0049 3.3.1.D Learning Objectives ADMPROE071 From memory State the responsibilities of the following personnel:

a.

Refueling SRO.(SRO ONLY)

b.

Refueling Bridge Operator

c.

Control Room Refuel Monitor IAW NC.NA-AP.U-0049.

IMatetial Required for Examination I None Bank Significantly Modified INPO Bank QID # 2147 Quad Cities 10/11 1-Monday, June 23,2003 7:24:00 AM Page 39 of 162

Given the following conditions:

The reactor has scrammed (all control rods are at position 00) on high drywell pressure.

Reactor pressure is 35 psig.

Reactor level is -1 20 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 the Drywell and:

place one loop of RHR in drywell spray and the other loop of RHR in drywell and suppression chamber spray.

place one loop of RHR in suppression pool cooling and the other loop of RHR in drywell and suppression chamber spray.

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 drywell spray, and vent the suppression chamber.

C B

Application Hope Creek 06/17/2003 24A201 Emergency and Abnormal Plant Evolutions 1

1 295024 High Drywell Pressure 29 EA2. Ability to determine and/or interpret the following as they apply to HIGH DRYWELL PRESSURE:

EA2.01 Drywell pressure 4 2 4.4 stification:

lace 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 cooling/spray, DWSlL 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 coolinglspray to prevent inadvertent bypass of the containment on the operating pump trip.

place one loop of RHR in drywell spray and the other loop of RHR in drywell and suppression chamber 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 maintained below the design limit of 65 psig, and only after attempts top lower pressure with DW spray.

Reference Title HC.OP-EO-ZZ-0102 Monday, June 23, 2003 7.24 00 AM

-~

Page 40 of 162

Learning Objectives 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 lMaterial Required for Examination

] EOP Flowcharts Bank VISION Bank QID# Q56010 Monday, June 23, 2003 7:24:00 AM Page 41 of 162 Direct From Source

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.

When Drywell Temp reaches 310F using 5 ADS valves.

When Drywell Press reaches 35 psig using all Turbine Bypass Valves.

Hope Creek 06/17/2003 b

IS Application 024A204 nd Abnormal Plant Evolutions 1

1 30 rywell Pressure J

EAZ. Ability

~~

todetermine and/or interpret the following as they apply to HIGH DRYWELL PRESSURE:

EA2.04 Suppression chamber pressure: Plant-Spe 3 9 3 9

~

JUSTIFICATION:

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.

INCORRECT - When Drywell Temp reaches 310°F using 5 ADS valves. Emergency de-pressurization must occur now before exceeding the PSP curve. Emergency de-pressurization for 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 the PSP curve.

1 Reference Title EOP Flow chart 102 bases for the curve IAW the Primary Containment Control - Drywell Lesson Plan a Drywell Spray Initiation Limit b Pressure Suppression Pressure (R) Given any step of the procedure, determine the reason for performance of that step and/or predict expected system I

EOP102E009 Monday, June 23, 2003 7 24 01 AM Page 42 of 162

~

response to control manipulations prescribed by that step IAW the Primary Containment Control - Suppression Pool Lesson Plan Question Source Comments:

Material Required for Examination Question Source: I Facilitv Exam Bank

] EOP Flow chart 102 VISION Bank QID# Q56157 Monday, June 23,2003 7 24:Ol AM Page 43 of 162 Editorially Modified

EOP 102 PRIMARY CONTAINMENT CONTROL, has an override that states:

IF:

Terminate drwl sprays.

Drwl sprays have been initiated THEN:

Before Drwl press reaches 0 psig.

Supp chamber sprays have been initiated Before suppression chamber press reaches 0 psig. Terminate suppression chamber sprays.

Which of the following statements describes the reason for this requirement?

0 psig drywell pressure ensures a drywell temperature below 212F, therefore there is NO need to continue drywell sprays.

It makes one more RHR loop available as soon as possible for injection into the reactor pressure vessel.

It prevents drawing a negative pressure in the containment, which would open the vacuum breakers and draw air into the containment.

This action ensures that the drywell structure will NOT endure excessive thermal stresses due to rapid cooldown.

C R

Memo reek Emergency and Abnormal Plant Evolutions I

295024 High Drywell Pressure 2.4 Emergency Procedures and Plan ge of the specific bases for EOPs.

t prevents drawing a negative pressure in the containment, which would open draw air into the containment. Correct. IAW bases for ste p PCC-1, a negative to RB vacuum breakers and de-inert containment.

06/17/2003 2950246418 31 2.7 3.6 the vacuum breakers and pressure will open the SC 0 psig drywell pressure ensures a drywell temperature below 212F, therefore there is NO need to continue drywell sprays. Incorrect.

It makes one more RHR loop available as soon as possible for injection into the reactor pressure vessel.

Incorrect. Concern is de-inerting Containment.

This action ensures that the drywell structure will NOT endure excessive thermal stresses due to rapid cooldown. Incorrect. Ensures containment is not de-inerted.

I Reference Title EOP 102 Bases for step PCC-1 Learning Objectives EO101 PE008 1

(R) Given any step of the procedure, describe the reason for performance of that step and/or expected system response to control manipulations prescribed by that step

]Material Required for Examination I None Bank Editorially Modified INPO BANK QID # 20450 Quad Cities 08/

1 Monday, June 23, 2003 7:24:01 AM Page 44 of 162

Monday, June 23, 2003 7:24:01 AM Page 45 of 162

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?

RV opens, which actuates low-low set causing the "H" SRV to open and both valves T

will control pressure at new, lower operating setpoints.

The "H" and "Pff 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 "P" SRVs control pressure together at new operating setpoints.

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.

d B

IComprehension Hope Creek 06/17/2003 cy and Abnormal Plant Evolutions 1

1 295025A103 32 igh Reactor Pressure I

EAI.

1 Ability to operate and/or monitor the following as they apply to HIGH REACTOR PRESSURE:

~~~

EA1.03 Safetyhelief valves: Plant-Specific 4.4 4 4

~~

Just i fi ca t i on :

Correct 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 "HI' 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.

Reference Title HC.OP-SO.SN-0001 Learning Objectives MSTEAME003 (R) Concerning the safety relief valves, summarize, list or identify the following a

b The number and type of SRV's at Hope Creek Which SRV's have an ADS function Monday, June 23, 2003 7 24 01 AM Page 46 of 162

c d

e f

Power supplies to the SRV solenoids Which SRV's can be operated remotely and the location from which each of these valves can be operated Purpose of the low-low set function and determine which SRV's are used for this function Determine the sequence of operation of the low-low set SRV's including arming setpoints, lift points and reclose setpoints

/Material Required for Examination 1 None Question Source:

Facility Exam Bank Question Source Comments:

Vision Bank QID# a53505 Monday, June 23, 2003 7:24:01 AM Direct From Source Page 47 of 162

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.

I Based on these conditions, which of the following require an immediate reacdr pressure reduction?

Reactor pressure stabilizes at 900 psig.

Suppression Pool level reaches 120 inches and is rising.

Suppression Pool water temperature reaches 205 degrees F.

Suppression Pool level reaches 70 inches and is lowering.

~

C B

IApplication 06/17/2003 Emergency and Abnormal Plant Evolutions 1

295026 Suppression Pool High Water Temperature 33 EKI. Knowledge of the operational implications of the following concepts as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE:

EKILO2 Steam condensation 3 5 3 8

~

Justification:

Suppression 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 EOP 102 Step SP/T-7 o'peration and explain the bases for the curve IAW the Primary Containment Control - Suppression Pool Lesson Pian EOP102E009 (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 - Suppression Pool Lesson Plan

]Material Required for Examination I EOP Flowcharts 7

1 New Question Source Comments:

Monday, June 23,2003 7:24:01 AM Question Modification Method:

Page 48 of 162

~

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.

Maintain water level +12.5" to +54" to prevent large power swings on the reactor core.

Emergency Depressurize to prevent Drywell pressure from exceeding Primary Containment Pressure Limit.

~~

~

reek 06/17/2003 B

Comprehension Pool High Water Temperature d Abnormal Plant Evolutions I

1 34 EK3 EK3 04 SBLC injection Knowledge of the reasons for the following responses as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE:

3 7 4 1 Justification:

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 +54" to prevent large power swings on the reactor core. Incorrect Level must be lowered.

Emergency Depressurize to prevent Drywell pressure from exceeding Primary Containment Pressure Limit. Incorrect. Drvwell should be sprayed. ED action to be taken if sprays not effective.

Reference Title EOP 101A bases BllT Plan EOP Flowcharts INPO Bank QID # 15341 08/23/1999 Mont Editorially Modified Monday, June 23, 2003 7 24 01 AM Page 49 of 162

Monday, June 23,2003 7 24.02 AM Page 50 of 162

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.

~~

___~---

~-

~

22.0 ft; Vent

- - the Drywell.

~~~

~- -~

~

06/17/2003 JUSTIFICATION: SRO 55.

procedures during normal, t of facility conditions and selection of appropriate 169.8 inches indicated t

. EOP-102 Step DWP-12 requires Venting the Suppression Pool if INCORRECT - 21.

based on 254.

I the Primary Containment Control - suppression Pool Lesson Plan.

response to control manipulations prescribed by that step IAW the Primary Containment Control - Suppression Pool Lesson

~

_._~______

~

-~

_ _ _ _ _ ~

~

(R) Given any step of the procedure, determine the reason for performance of that step and/or predict expected system Pian.

/

-~.

Monday,-June23, 2003 7:24:02 AM Page 51 of 162

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Monday. June 23, 2003 7:24:02 AM Paae 52 of 162

~

Given the following conditions:

- A LOCA has occurred.

- RPV water level is stabilized above TAF.

- Suppression Chamber water level is 60 inches and lowering.

J 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?i Core Spray Loop Chamber pressure is 20 psig.

is the ALTERNATE Makeup path to be used only if Suppression

-~

~

295030A106 LEVEL:

te storage and transfer (make-up to the suppressio

________-__.___~-

L A; Below. Correct IAW HC.OP-EO.ZZ-0315, Suppression Chamber pressure must be less than 20 psig to implement MIU from the CST via Core Spray. A CS Loop is the Alternate loop.

A; Above. Incorrect. A is prefered loop.

B; Above. Incorrect. Only used below 20 psig.

B: Below. Incorrect. B is Prefered.

EOP300E004 FGmernory, describe any/all flow paths established by the performance of each of the 300 series Emergency Operating Procedures.

L Mondav. June 23,2003 7:24:02 AM Paqe 53 of 162

Which one of the following correctly describes the Technical Specification bases for the Suppression Pool low water level limit?

, _ ~ _ _ _ _ _ _ _ _ ~ ~ -

dequate SRV T-Quencher submergance during Eme i-Depressurization.

This limit ensure adequate water volume is available based on NPSH

~ _ _ _ _ _ _ _ _ _ _ _

and vortexprevention. - -

This limit prevents exceeding the Suppression Pool design temperature limit during a DBA LOCA.

This limit prevents exceeding the Suppression Pool design pressure limit during a DBA LOCA.

0611 7/2003 2.2 Eauioment Control 2.2.25

/Knowledge of bases in technical specifications for limiting conditions for operations and safety limits.

' 2.5 13.7

~

3 (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. Plausible but incorrect bases.

This limit ensure adequate water volume is available based on NPSH and vortex prevention. Incorrect.

Bases for EOP Caution 2 Limits at 0 inches of SPL.

This limit prevents exceeding the Suppression Pool design temperature limit during a DBA LOCA.

Incorrect Plausible but incorrect bases.

IHCGS TS Bases 3/4.6.2 PRICONE009 '

(R) Given a Scenario of applicable operating conditions and access to technical specifications:

a.

b.

HCGS technical specifications. (SROISTA ONLY)

c.

Select those sections which are applicable to the Primary Containment Structure IAW HCGS technical specifications.

Evaluate Primary Containment Structure operability and determine required actions based upon system operability IAW Explain the bases for those technical specification items associated with the Primary Containment Structure IAW HCGS technical specifications.

I

Monday,

- June 23, 2003 7:24:02AM Page 54 of 162

~

~ _ _

_~

Given the following conditions:

A Station Blackout occurred.

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 -1 50 inches and lowering slowly.

Suppression Chamber pressure is 5 psig.

Drywell temperature is 3 1 OF 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?

~

ressurize the RPV.

-~

Open 5 ADS me Lower B RHR SP Cooling flow to 9000 gpm.

Trip B RHR pump immediately.

_ _ _ _ _ _ _ _ ~ _ _ _ _

_ _ _ _ _ ~. _ ~

~_

- ~_

~

~~

~

_ _ _ _ _ _ _ _ _ _ _ _ _ ~

~

-~

EA1. Ability to operate and/or monitor the following as they apply to REACTOR L

_~

essure coolant injection (RHR): Plant-Specific

~

Re-align B RHR for LPCl mode at rated flow. Correct. EOP - 101 step ALC-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 E0101LE006 (R) Given any step of the procedure, describe the reason for performance of that step andlor expected system response to control manipulation prescribed by that step.

E01 01 LE008

~

Monday, June 23,2003 7:24:02 AM Page 55 of 162

Given the following conditions:

Drywell pressure is 4.5 psig.

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.

response to control manipulations

~ _ _ _ _ _ _ _ ~

prescribed by that step. ---. _.

~~

- _ _ _ ~

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?

_ _ _ ~

_ _ _ _ _ _ ~ _ ~ _ -

- -~

All SRVs remain closed.

- - ~-~

Reactor power reaches 3 percent.

_ _ _ ~

Suppression Pool Temperature lowers to 108 F.

_ _ ~ _ _ _ _

7

_ _ _ - ~

RPV level reaches -50 inches.

.~

~-

_ _ - _ -~

interpr following a

w WATER LEVEL:

~

______~__--_______

~

L, 55.43(5) Assessment of facility conditions and selection of appropriate procedures during normal, 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 implemented for 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 -190 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-11 is answered YES. This requires lowering level until power is less than 4 percent. You can not '

back 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 lowering level to -50 inches.

-1 L

I Monday, June 23, 2003 7:24:03 AM Page 56 of 162

EOP 101 Flowchart Monday, June 23, 2003 7:24:03 AM -

Page 57 of 162

Given the following conditions:

- The plant is operating at rated power.

- Control Room Overhead alarms are received:

L

- BI-A4 HPCl TURBINE TRIP

- D3-A1 HPCVRHR A LEAK TEMP HI

- When the operator checks the HPCl panel, HPCl inboard and outboard steam line isolation valves are stroking closed.

- HPCl turbine was NOT running at the time.

- HPCl Steam line pressure is 900 psig and lowering.

Which one of the following would cause this isolation?

hase High Temperature.

~

-~

_ _ _ _ -~

_ _ _ - - - ~

HPCl Steam Line Low Supply Pressure.

HPCl Pump Room High Temperature.

~

~~

~

-~ _ -

~

06/17/2003

'ic tripped which in turn causes BI-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.

HPCl Steam Line High Differential Pressure. Incorrect. Would cause BI-A5 annunciator, HPCl STEAM LINE DlFF PRESSURE HI.

~._ -

EOP103E003 I

(R) Define the term "Maximum Safe Operating Temperature".

ny step in the procedure, describe the reason for performance of that step and/or expected system response to control manipulations prescribed by the step. - -.-

EOP Flowcharts

~

-~

~-

Mondav. June 23.2003 7:24:03 AM Page 58 of 162

_ _ _.. _ _ _ _ _ ~

~

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.

NDARY CONTAINMENT

_______- 13.713.8, m

~

i Justification 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 RACSlonly 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.

I iHC.OP-AP.SP-OOO1 Attachment 15 I

~~~

~

EOP103E004 1 (R) Define the term "Maximum Safe Operating Radiation Level". __.____

EOPI 03E006

~

(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.

L Monday, June 23,2003 7:24:03 AM I

Page 59 of 162

Given the following conditions:

- The plant is operating at I00 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.

b-i-

Which one of the following actions is required?

Start another Reactor Building Supply fan IAW HC.OP-SO.GR-0001 Reactor Building Ventilation.

Place FRVS in service IAW HC.OP-AB.ZZ-0001 Transient Plant Conditions.

Isolate RBVS Isolation Dampers IAW HC.OP-SO.SM-0001 Isolation System Operation.

~

-.- -~

_~~

~~

_ _ ~ _

I I Place FRVS in service IAW HC.OP-SO.GU-0001 FRVS Operation.

~ _ _ _ _ _ _ _

SRO IOCRF 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 HCOP-AB.ZZ-0001 Transient Conditions. Incorrect. AB-ZZ-0001 does not provide direction for starting FRVS.

Isolate RBVS isolation Dampers IAW HC.OP-SO.SM-0001 Isolation System Operation. Incorrect.

Isolations performed with this procedure are performed to isolate system breaches.

Start another Reactor Building Supply fan IAW HC.OP-SO.GR-0001 Reactor Building Ventilation.

~

Monday, June 23, 2003 7.24:03 A M

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Page 60 of 162

Given the following conditions:

- The reactor is operating at 100% power.

- Annunciator B1-B3 ( RClC PUMP ROOM FLOODED ) alarms with the following alarm message presented on the CRlDS 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 10314?

I --- Isolate RClC II -- Immediately commence a normal reactor shutdown Ill -- Runback reactor recirculation and manually scram the reactor IV - Emergency depressurize the reactor

~

I - ONLY

~

0611 712003 36K303 v

43

~~

~

-~

EK3.

_ _ Knowledge of the reasons for the following responses as they apply to SECONDARY CONTAINMENT

-~

~~~

HIGH SUMP/AREA WATER LEVEL:

3.5 3.6 I, Ill, and IV. Incorrect. Normal shutdown not scram and ED. Would apply if reactor coolant leak.

I and II. Correct. EOP-10314 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.

~~

EOP103E006 (R) Given any stepin the procedure, describe the reason for performance of that step andlor expected system response to Page 61 of 162

Monday, June 23,2003 7:24:04 AM Page 62 of 162

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.

~

Continue SLC pump operation and begin reactor cooldown.

~-

Verify the "Af SLC pump is tripped and continue rod insertion. _ _ _ _ _ _ ~ _

Verify

-._ the

~

- "Af

-~ SLC pump is tripped and exit EOP-IOIA.

~

_ _ ~ _ _ _ _. _ _ _ _ ~ -

~

B Application Hope Creek 06/17/2003 L

~

-~

EA2.7

._ Ability to determine and/or interpret the following as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR U

~

4.3 '4.4,

-~

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 [4640 - (41.2 x 90)] = 932 gallons remaining. Step RC/Q-19 directs continuation at step RC/P-20 for depressurization and 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 EOP Flowcharts; HCGS TS section 3 1.5 Monday, June 23,2003 7:24:04 AM Page 63 of 162

- A turbine trip and hydraulic ATWS occur from 65 percent power.

- EOP-IOIA and 102 are currently being executed.

Current plant conditions:

u 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?

Reduce Suppression Pool water temperature.

Lower Suppression Pool level.

Open additional Turbine Bypass Valves.

Hope Creek 0611 712003 a

]Application I

295037G406 45 Un 2.4 1 Emergency Procedures and Plan 3.1 4.01 I

OCFR55.43 (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.

RaDidIv DeDressurize the RPV. Incorrect. Not permitted at this time.

~~

- -~~

IEOP-IOIA and 102 Monday, June 23, 2003 7:24:04 AM 7 Page 64 of 162

EOP Flowcharts INPO Exam Bank Editorially Modified

_ _ ~-

Monday, June 23, 2003 7:24:04 AM Paae 65 of 162

~-

-~

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 following 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 ft elevation.

~

I Wind Direction TO; Wind Speed 33 ft elevation.

Wind Direction FROM; Wind Speed 300 ft elevation.

~

__.- - -~

_ _ _ ~ _ _ _

~

e following as they apply to HIGH OFF-SITE RELEASE RATE:

Wind Direction TO; Wind Speed 300 ft elevation. Incorrect. Wrong elevation. Wind direction is FROM Wind Direction FROM: Wind SDeed 300 ft elevation. Incorrect. Wrong elevation. Wind speed is 33 ft elev.

Page 66 of 162

~

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 to the Thyroid.

~~

~

Rem Whole Body and Hope Creek

[Memory 295038K301 3.6 4.5

~ _ _ _ _ _

~-

members of the public; 25; 200 295038K301 47 TE RELEASE RATE:

1 3.6 4.5

~

~-

members of the public; 25; 75 station personnel; 25; 75

~UFSAR Chapter 15

_ ~-

~-

conditions IAW the Student Handout.

Monday, June 23, 2003 7:24:04 AM 1

Page 67 of 162

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

~-

hi

~~

low internal pressure.

Hope Creek 06/17/2003 b

i iB j

IMemory J

500000K101 4a EKI. 1 Knowledge of the operational implications of the following concepts as they apply to HIGH EKlPl Containme high i 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 pressure will increase.

CONTAINMENT HYDROGEN CONCENTRATIONS:

I

~

ration s may cause a

~~

~

ILP NOH01 H2RECM-00 IEOP 102 PC/H-1 None

_ _ _ ~ -

~

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INPO Exam Bank Significantly

-~..

Modified INPO BAN Monday, June 23, 2003 7:24:05 AM Page 68 of 162

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 failedxreposition.

~~

The 2 gpm

_ _ Stabilizing

- Valve has

- failed to reposition.

Both Cooling Water Header to Exhaust Header Pressure Equalizing Valves have failed open.

The Drive Water Header Pressure Control Valve has failed closed.

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

~~

~

v 06/1 7/2003 01 K303 49 K3.

' Knowledge of the effect that a loss or malfunction of the CONTROL ROD DRIVE HYDRAULIC SYSTEM--

will have on following:

-3.132i I

I 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-lncorrect-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 svstem flowpaths:

a.

Drive Water

b.

Cooling Water

c.

Exhaust Water

d.

Charging Water

f.

, e.

Scram Seal Purge for Recirculation Pumps RPV Level Reference Leg Backfill

~

1

-.I Direct From

~.

Source Monday. June23.2003 7:24:05 AM Page 69 of 162

~

A LOCA has occurred and reactor vessel water level is -140 inches.

I Which of the following describes the steps necessary to restart a CRD pump?

I

~~

Close the non-I E circuit breaker by depressing the CLOSE pushbutton and close the 1 E circuit '

Depress the LOCA OVERRIDE pushbutton, close the non-1 E circuit breaker by depressing the CLOSE pushbutton, and close the 1 E circuit breaker by depressing the CRD pump START pushbutton.

~

Depress the LOCA OVERRIDE pushbutton, close the 1 E circuit breaker by depressing the CLOSE pushbutton, and close the non-1 E circuit breaker by depressing the CRD pump Close the 1 E circuit breaker by depressing the CLOSE pushbutton and close the non-I E circuit 1

I breaker

~- -

by depressing the CRD pump START pushbutton.

~

- -~

_ _ _ ~ _ _ -

~

_- _ ~ - _ -

~-

~

START pushbutton.

breaker by depressing the CRD pump START pushbutton.

[Comprehension 1 Hope Creek 06/17/2003 201 001 K605 50 K6. - Knowledge of the effect that a loss or malfunction of the following will have on the CONTROL ROD DRIVE HYDRAULIC System:

power Justification:

Depress the LOCA override push-button, close the 1 E 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-CRD pumps are load shed on a LOCA signal and requires both the 1 E and Non I E breakers to be closed Close the non-I E circuit breaker by depressing the close push-button and close the 1 E circuit breaker by depressing the CRD pump START push-button.-Incorrect-has the power supplies titles backwards, also requires LOCA override Close the 1 E 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 1 E circuit breaker by depressing the CRD pump START push-button.-Incorrect-has titles of breakers backwards L

I CRDHYDE015 (R) Given a CODY of HC.OP-SO.BF-0001 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:

a.

CRDPumps

b.

CRD Pump Suction Filter

c.

CRD Drive Water Filter I

d.

CRD Flow Control Station OABI 35E003 Monday, June 23,2003 7:24:05 AM Page 70 of 162

Mondav. June 23,2003 7:24:05 AM Paae 71 of 162

Given the following conditions:

- Reactor power is 85 percent during a plant start-up.

- A control rod is selected for withdrawal.

- An adjacent "C" level LPRM providing signals to an Average Power Range Monitor (APRM)

Channel and a Rod Block Monitor (RBM) Channel fails downscale once the rod is in motion.

L-Which one of the following describes the effect of the failure on the Reactor Manual Control System (RMCS) and the reason why?

RMCS will initiate control rod blocks:

~

at a lower actual local bypassed and removed from the RBM only. The APRM and the RBM readings will be lower than actual.

,at a lower actual local power level because the LPRM will be automatically bypassed and removed from both the APRM and RBM. The APRM and RBM readings will NOT be affected.

~-

at a higher actual local power level because the LPRM will be automatically bypassed and

~

- - ~-

~

~_

~

_~

~

removed from the APRM only. The APRM reading will NOT be affected and the RBM reading i

I

~

will be lower than actual.

at a higher actual local power level because the LPRM will NOT be automatically bypassed to-'

the APRM or the RBM. The APRM and RBM readinns will be lower than actual.

0611 712003 201 002A105 51

- - ~

~

~-

~

~~~~

AI.

Ability to predictand/or monitor changes in parameters associated with operating the REACTOR MANUAL CONTROL SYSTEM controls including:

~~

A I.05

, /Local reactor power

~_ 3.4 3 6 iled LPRM will be I oca1 power to cause a rod block. Once the LPRM card is bypassed, then the LPRMs are averaged and will read normallv.

~-

HC.OP-S0.S F-0002 MANCONE008

[RI From memory. exDlain the interrelationshiDs between the Reactor Manual Control Svstem and the followinq.

1 Rod Worth Minimizer Neutron Monitoring System Rod Block Monitor System Mode Switch Refueling System

~

Refueling Bridge

~

Refueling GrapplelHoists

f.

120 VAC Uninterruptible Power Supply RBMSYSEOOZ cations and alarms in the Control Room, determine the status of the Rod Mondav, June 23,2003 7:24:05 AM Page 72 of 162

Page 73 of 162

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

_ ~ _

-. ~

A-Inoperable: B-Inoperable 03 52

~

_ _ _ - ~

Conduct of Operations

_ _ _ ~

3.4 13.8 2.1.32 bbility to explain and apply system limits and precautions.

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; Ad mi nistratively 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.

u B-Inoperable Correct-Only 1 of 4 LPRM for D LPRM Level makes RBM B

__ ~.

SH.OP-AP.ZZ-0108 Exhibit 3 I

1 Select those sections which are applicable to the Rod Block Monitor (RBM) System.

Evaluate Rod Block Monitor (RBM) System operability and determine required actions based upon system inoperability.

Explain the bases for those Technical Specification items associated with the Rod Block Monitor (RBM) System. (SRO Only)

-- -~ -

I

~-

Monday, June 23,2003 7:24:06 AM Page 74 of 162

.~_.______

-~

Given the following conditions:

- The plant is operating on the 100% Rod Line at 100% power.

- 'B' Recirculation Pump trip occurs due to inadvertant bump of the Drive Motor Breaker local trip

- Reactor settles at 56% power.

- Recirc Loop A Flow (FI-R61 IA) is 38 Mlbm/hr.

- Recirc Loop B Flow (FI-R61 IB) is 3 Mlbm/hr.

- Core Flow (FR-R613) is 35 Mlbmlhr.

- 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.

- switch.

Which of the following action(s) is(are) required per HC.OP-AB.RPV-0002, Reactor Power Oscillations?

~~

Lock the Mode Switch in the Shutdown position.

~~

Raise 'A' Recirculation Pump speed until total core flow is above-45%.

Reduce 'A' Recirc flow and insert control rods IAW Stuff Sheet.

Insert control rods IAW Stuff Sheet.

~

~~~~

- ~-

A2.

1 Ability to (a) predict the impacts of the following on the RECIRCULATION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

~

3.6 3.7 A2.03 iSingle recirculation pump trip

~

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 reaion.

~~~-

~~

~

HC.OP-AB. RPV-0002 Oscillations.

~~~~~~

-~

~

-~

~

-~

~

L

-~

ABRPV2E001 1

Recognize abnormal indications/alarms andlor procedural requirements for implementing Reactor Power Oscillations.

- _ _ _ _ ~ _

~~~

ABRPV2E007 (R) Explain the bases for Subsequent Actions and the information contained in the Discussion Bases Section of Reactor Power Oscillations.

~~

-~

Mondav. June 23.2003 7:24:06 AM 1 Page 75 of 162

Monday. June 23,2003 7:24:06 AM Paae 76 of 162

The plant is operating at 100 percent power with all systems normal.

LPCl Channel B receives a LOCA Level 1 initiation signal.

BVH-210 RHR Room Cooler is in Auto Lead.

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 1 I O degF.

Which one of the following describes the status of the RHR Room Coolers in B RHR Room?

B runnina: F runnina.

B running; F NOT running.

I!!

B NOT running;

~-

F running.

_ _ _ _ ~ ~ _ _ _ _ _ ~.

~~

- _ ~ __

~

~-

~~~~

~

-~~

~~~

B NOT running; F NOT running.

I 1

__ __.~

MODE and the following:

3.2 3.21 B NOT running; F running. Correct-B cooler will start first on temperature. A low flow trip of the B cooler will cause the computer trouble alarm. F cooler will start because the cooler is in Auto.

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.

L

- therein, explak the function of the 4pporting system, IAW the RHR System Lesson Plan.

H-83 sheets 5 and 1 I

~

~~

Monday, June 23,2003 7:24:06 AM 1

~~

- ~~~_ ---

Page 77 of 162

The plant is operating at 95% power.

The RWCU system has just been returned to service.

The "AI RWCU pump is running.

The "B" RWCU pump is C/T.

RWCU return to Feedwater temp CRIDS pt A21 5 is reading 41 OF.

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?

~~

~-~

~ _ _

'Hope Creek 06/17/2003 a

[Application J

v K1.

K1.03 IReactor feedwater system 1 Knowledge of the physical connections and/or cause-effect relationships between REACTOR WATER CLEANUP SYSTEM and the following:

' 3.1 131

__-A-uires "2. OP-SO. BG-0001 Attach men ect answer. One pump operation at 41

__ ~

~

-~

~~

~-

_ _ _ ~

~~

Given any system that interrelates with the RWCU System, explain the purpose of that interface IAW the RWCU System RWCUOOE003

b. Determine the blowdown RWCU System flowpath(s) IAW the RWCU System Lesson Plan.

c. Determine the recirculation RWCU System flowpath IAW the RWCU System Lesson Plan.

I Monday. June 23.2003 7:24:06 AM Page 78 of 162

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.

L I

. -~

What effect does this have on HPCl Operability at the PRESENT time?

HPCl is operable because it can develop rated flow.

~

requirements.

HPCl is "operable but non-conforming" because it is NOT capable of meeting all surveillance HPCl is "operable but degraded" because it has lost testing capability.

HPCl is inoperable because it is NOT capable of meeting all surveillance

~~

requirements.

~-

Hope Creek 0611 712003 d

[Application I

-- _ - _ _ _ _ _ _ ~

~-

-~

~- -

- ~~~

~-

A4.

Ability to manually operate and/or monitor

- __ in the control room:

A4.02 3.8

-~

~

IFlow controller: BWR-2, 3, 4 Justification:

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.

HPCl is "operable but non-conforming" because it is NOT capable of meeting all surveillance requirements. Operable but non conforming is not applicable.

HPClOOEOl8 (R)

Given plant conditions and access to Technical Specifications:

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.

Explain the bases for those technical specification items associated with the HPCl System IAW HCGS technical specifications.

+

(SRO Only)

Tech Spec 3.5.1 Monday,June 23,2003 7:24 06AM

- ~ _ _ - - -. ---J Page 79 of 162

~

~~

Monday, June 23, 2003 7:24:06 AM

~~

I Page 80 of 162

~~

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).

I.

Main Steam Line Drains.

II.

Ill.

IV. RFPTs min flow operation.

RClC or HPCl Steam Line Drains.

RClC or HPCl in Full Flow test.

II.

712003 23 57 2.11

~

Conduct of Operations 2.1.23 'Ability to perform specific system and integrated plant procedures during different modes of plant

/operation.

3.9 4.0 J USTl Fl CAT1 ON :

IAW HC.OP-I0.Z-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, RCIC or HPCI Steam Line Drains, RClC or HPCl in Full Flow test.

Ill - Correct. RCIC or HPCl in Full Flow test is the last of the listed systems.

I - Incorrect.

I1 - Incorrect.

IV - Incorrect.

lHC.OP-IO.ZZ-0007 Note 5.2.5

~~

. _ _ ~

-~~

~

1 Operating Procedure, IAW this Lesson Plan Monday, June 23,2003 7:24:06 AM Page 81 of 162

Given the following conditions:

- A LOCA occurred.

- RPV level has been stabilized above TAF.

- Drywell sprays are in service.

- Core Spray pump 'A amp indicator begins to fluctuate.

C 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 of the suction strainer; reduce loop flow.

Partial closure of the injection valve; manually open the injection valve fully.

-~

~

~- -~

~~

Clogging of the suction strainer; throttle 'A' Core Spray Pump manual discharge valve.

~_ -~

Partial closure of the injection valve; stop the IC' Core Spray pump.

06/17/2003 A&

Ability to (a) predict the impacts of the following on the LOW PRESSURE CORE SPRAY SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those 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 Partial closure 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.ZZ-0155.

Partial closure of the injection valve; stop the 'C' Core Spray pump. Incorrect. Stopping C pump would increase flow through A pump and make the conditions worse.

L I

OABl55E005 (R) Interpret and apply charts, graphs and tables contained within the Degraded ECCS PerformancelLoss Of NPSH, Abnormal ODeratina Procedure.

~~

OABI 55E001 Recognize abnormal indications/alarms and/or pro implementing, Degra OABI 55E003

_.... Of NPSH, Abnormal Operating Procedure.

Degraded ECCS PerformancelLoss Of NPSH, Abnormal Operating Procedure.

ating parameters related to

~

None

~_

L-Monday, June 23,2003 7:24:07 AM A

_ _ ~ _ _ _ _ _ _ _

Page 82 of 162

~

~~

Given the following conditions:

- Drywell pressure increased to 2 psig,

- Off-site power is lost.

Which of the following describes the start sequence for the core spray systems after off-site power was lost?

L

~

Core Spray pumps "A" and "B" start immediately after the diesel generator out closed. Core Spray pumps "C" and "D" start six seconds after the diesel generator output breakers are closed.

Core Spray pumps "At 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 'ID" start immediately after the diesel generator output breakers are closed.

Core Spray pumps "A", "B", "C", and "D" start six seconds after the diesel generator output breaker is closed.

- - ~

~

-____~___.

_~

~

__ - __ _~

~~~_ -

~

!R

/Comprehension 1

Hope Creek 209001 K201 59 Low Pressure Core Spray System

- ~ _ _

K2.1 Knowledge of electrical power supplies to the following:

.013.11

~~

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 "At 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", "C", and 'ID" start immediately after the diesel generator output breaker closes.

Core Spray pumps "A' and "B" start immediately after the diesel generator output breaker is closed. Core Spray pumps "C" and "D' start six seconds after the diesel generator output breaker is closed.

_ _. ~ _ _ ~ _ _ _

~

~~

~

. OP-SO. BE-0001 CSSYSOE007 1 (R) Given plant problemslindustry events associated with the Core Spray System:

1

a.

b.

~

problem/industry event at HCGS, IAW the Core Spray System Lesson Plan.

Summarizelldentify the root cause of the plant problem/industry event, IAW the Core Spray System Lesson Plan.

Summarizelldentify the HCGS design and/or procedural guidelines that mitigatelreduce the likelihood of the plant Summarizelldentify the "lessons learned" from the plant problem/industry event, IAW the Core Spray System Lesson

-. ~ _ _ _ _ ~ _ _ _ _ _ _ _ ~ _ _ _

_ ___ -~ ~_

~ - _

~ __.__

set of plant conditions, from memory, summarizelidentify the interrelationship between the Core Spray System IAW the Core Spray System Lesson Plan:

Residual Heat Removal (RHR) System

~ t:

Torus Compartment

c.

4160 VAC Class 1 E Distribution System

d.

e.

480 VAC Class 1E Distribution System 125 VDC Class 1 E Distribution System

~

Page a3 of1 62

~~~~~~

~~

~

~~~

~

~-~

f.

Nuclear Boiler

g.

Liquid Radwaste System

h.

i.

j

k.

Condensate Storage Tank

1.

Automatic Depressurization System (ADS)

m.

Emergency Diesel Generators (EDGs)

n.

Nuclear Boiler Instrumentation System

0.

Condensate Storage and Transfer System Primary Containment Instrument Gas (PCIG) System High Pressure Coolant Injection (HPCI) System Standby Liquid Control (SLC) System c

Monday. June 23, 2003 7:24:07 AM Paae 84 of 162

Given the following conditions:

- A Loss of Offsite Power (LOP) concurrent with an ATWS has occurred.

- 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.

>L, Which one of the following describes the components of the Standby Liquid Control System that are available for injection?

SLC pump A and squib valve _

~

F004A ONLY.

_~ --.

-~

~~

~

NEITHER SLC pump NOR associated squib valve.

BOTH SLC pumps and both squib valves.

SLC Dump B and squib valve F004B ONLY.

~

C i

[Memory Hope Creek 21 1000~5202 60

~_

~

- ~

~

K2.

Knowledge of electrical power supplies to the following:

~

3.1 3.2 H SLC 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 DumD B and sauib valve F004B ONLY. Incorrect.

a 1

rom memory I

___ a.

Standby LiquidControl Pumps.

b.

c.

Liquid Control System I.A.W. the Lesson Plan.

a.

b.

d.

ACPower Standby Liquid Control System Squib valves.

Standby Liquid Control System Storage Tank Heaters.

SLCSYSEOI 5 Standby Liquid Control Squib Valve Standby Liquid Control Storage Tank Level

c.

Redundant Reactivity Control System Facility Exam Bank Vision Bank QID # (254183

~.

Monday, June 23,2003 7:24:07 AM 1

2 Page 85 of 162

~-

~

- _- ~-

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 Injection valves F006A and B.

_ _ ~ _ _ _

~

~~~

~

Opening the

~

breakers for Fool and F004. ~-~

_~

v

~~~

-- - - - - ~

- ~

_~

~

-~

Bezel keyswitches on 1 OC651 C console.

Local panel Dump start switches.

- _ _ ~ _ -~

~

21 1 OOOK402 61 K4.

Knowledge of STANDBY LIQUID CONTROL SYSTEM design feature(s) and/or interlocks which provide for the following:

~

K4.02 Component

~ _ _ _ _ _ _ ~ -

and system testing

-~

~-

Justification:

Using local panel pump start switches. Correct. Local start switches prevent firing of the squib valves and automatic closure of the RWCU Fool and F004.

Opening the breakers for FOOI and F004. Not driven by any procedure. Does not stop firing of the squib valves.

Bezel keyswitches on 1 OC651C 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 Squib valve firing.

system initiation I.A.W. the Lesson Plan.

~~

~

.. -~

~

~~

~~~

~.

~ ~ _ _ _ _ ~

Monday, June 23, 2003 7:24:07 AM Page 86 of 162

~

During a failure-to-scram condition, which of the following indications would confirm that HC.OP-AB.ZZ-0000(Q), "Reactor Scram", should be exited and HC.OP-EO.ZZ-01 OIA(Q) entered?

All Blue liahts are lit on the Full Core Disdav.

v All HCU Accumulator Trouble liahts are lit on the Full Core Disdav.

Computer Pt in Alarm Overhead Annuciator C6-F5 extinguished.

All APRM "downscale" lights are illuminated.

_ _ _ _ _ ~ _ _ _ _ _ _ _ _ _ _ _ _ _ _

___~_________

Hope Creek 06/17/2003 C

/Memory I

21 2000A307 62

~

-~

A3.

1 Ability to monitor automatic operations of the REACTOR PROTECTION SYSTEM including:

A3.07

/SCRAM air header pressure 3.6 3.6

~~

~

~~~

~~~~

Computer Pt in Alarm Overhead Annuciator C6-F5 extinguished. Correct. indication that CRD Pilot Air Pressure is NOT depressurized as expected following scram.

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 scram.

All APRM "downscale" lights are illuminated. This is the response for a normal scram.

IHC.OP-AB.ZZ-0001 Attachment 1 L,

ABOOOOE003 State four (4) methods by which the operator can verify a successful scram action.

1 8 n a ~ e e r i ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ x a m i n a t i o n,

I Overhead annunciator window C6 figure from HC OP-AR.ZZ-0011 Monday, June 23,2003 7:24:07 AM

~-

Page 87 of 162

~

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.

- Annunciator "CRD SCRAM DISCH VOL VVTR LVL HI" is sealed in.

L, 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 one minute later placing the mode switch in STARTUP for NI testing?

~

~~

The RPS will reset and remain reset.

The RPS will reset and again scram.

Nothing will occur due to the present plant conditions.

b i

iB

]Comprehension

]

Hope Creek 0611 7/2003 OOA404 63

'v

~~_

A4.

Ability to manually operate and/or monitor in the control room:

~~

3.9 3.91

~

~~~~~

e 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 siqnal Dressent RPSOOOE016 I

From memory, explain how to reset a scram IAW

-~

the RPS System Operating Procedure.

~~

RPSOOOE017 1

Given a labeled diagramldrawing of, or access to, the Reactor Protection System controls, and/or alarms located in the Control 1

Room:

Explain the function of each indicator.

Assess plant conditions that will cause the indications to light or extinguish.

Determine the effect of each control switch on the Reactor Protection System.

~

~~

~

Monday, June 23,2003 7:24:07xM Page 88 of 162

~

~-

~

~~

~-

Given the folioing conditions:

- 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.

W

_ _ ~

How will this transient affect the unit?

1/2 scram and rod block from D and F APRMs; NO other effects I -

1/2 scram and rod block from B and D APRMs and a rod block from A and B RBM; NO other effects

~

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

!!! 1/2 scram from B APRM and a rod block from B RBM; NO other effects

~

~_.

~~

~

-~

_ - -~.

1/2 scram and rod block from B D and F APRMs and a rod block from B RBM; NO other effects 0611 712003 21 5002K203 Rod Block Monitor System v

A K2.

Knowledge of electrical power supplies to the following:

~-

.8 2.9

~

.. ---J

~~~

K2.03

[APRM channels: BWR-3, 4, 5 JUSTIFICATION:

Correct answer: % scram and rod block from B, D and F APRMs and a rod block from B RBM; no other effects. High input voltage trips the EPA breaker on overvoltage above 132 volts.

The following distractors are incorrect as follows:

.I % 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 8, D, and F APRMs, and B RBM, no mention of F APRM and mention of A RBM

.% 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 F APRMs; no other effects

- Incorrect-BD483 UPS powers B, D, and F APRMs, and B RBM, no mention of B APRM or B RBM APRMOOE004

~

Given a system which connects to or is required for the support of the APRMS, explain the function of the system 1 interrelationship, IAW the Student Handout.

1 Mondav. June 23.2003 7:24:07 AM

~

Page 89 of 162

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?

L' 5 on range 5, and off-scale on range 4

~

___ - - -~ ~

Off-scale on range 5, and off-scale on range 4.

5 on range 5, and 50 on range 4.

~

-~~

_ _ ~

-~

~

~ _.

~~

~

_ _ _ ~ _ _ _ _ _ _ _.

- __ - ~

~_

_~

~

A4.03

/IRM range switches

-13.6 13.41 Requires attached figures.

The IRM range switches are 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 I O.

I

, evaluate the status of the IRM controls/inst;umentation/alarms IAW control room procedures. ~-

IRMSYSE004 (R) Given a labeled diagram of, or access to, the IRM controlslindication 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.

Monday, June 23,2003 7:24:08 AM Page 90 of 162

Given the following conditions:

- Reactor startup in progress.

- IRMs are on range 10 and reading 25.

- Reactor Mode switch is in STARTUP/HOT STANDBY.

- IRM C fails downscale.

Which of the following lists rod block status with the present condition AND if the Reactor Mode switch is placed in RUN?

A Rod Block exists. The Rod Block will clear after placing the Reactor Mode switch

_ _ _ _ - in R NO Rod Block exists. Placing Reactor Mode switch in RUN will NOT result in a Rod Block.

I NO Rod Block exists. Placing the Reactor Mode switch in RUN will result in a Rod Block.

A Rod Block exists. The Rod Block will NOT clear after placingthe Reactor Mode switch in RUN.

~~

~-

\\Comprehension

~

Hope Creek 06/17/2003 21 5003K102 66

~-

__..-~-

_~__.____

-____ ~

_ _ ~

~ K1. -- Knowledge of the physical connections and/or cause-effect relationships between INTERMEDIATE K1.02

, /Reactor manual control RANGE MONITOR (IRM) SYSTEM and the following:

3.6 13.6 A Rod Block exists. The Rod Block will clear after placing the Reactor Mode switch in RUN. Correct.

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.

IRMSYSE009 1

(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 controlslinstrumentationlalarrns IAW control room procedures.

~-

INPO Exam Bank INPO BANK QID# 6330 09/26/1998 Dre Monday, June 23,2003 7.24:08 AM-

-~. I Page 91 of 162

Given the following conditions:

- The Mode Switch is in the STARTUP/HOT STANDBY position.

- 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 correct RPS response.

-~ -

Half scram when "C" APRM reads 51 %.

~

Half

~ scram

_~ when

~~

"C" APRM reads

~

15%.

~

. - ~ _ _ _ _ _

~

~~

-~

~

~_

. ~_ -

_ _ -. ~

~

Full -. scram when "C" APRM reads 15%.

Full scram when "C" APRM reads 51 %.

~

~~

~

~~

_ _ _ _ _ _ _ ~ _

_ ~ _ _

-~

~~

-- ~-

- - ~

~

0611 712003 102 67 Al.1 Ability to predict and/or monitor changes in parameters associated with operating the APRMILPRM

-~

controls including:

1 3.9

~ 4.0 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

Startur, Monday, June 23,2003 7:24:08 AM

~

Page 92 of 162

_ _ _ _ _ ~ _ _ _

Given the following conditions:

- A plant startup is in progress with power at 20%.

- Recirculation flow is 30%.

- The " A I APRM Flow Unit output remains at 30% as recirculation flow is raised.

'v As the plant startup continues, what will be the FIRST protective action to occur and the reason for 7

~

that action?

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

I scram will occur due to flow biased neutron flux

~

upscale. ~-

- _ _ ~ _

~

A control rod block will occur due to flow biasedneutron flux upscale.

_ _ _ _ ~ _ _ _ _

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.

.~

~

~-

~~

K6.

I Knowledge of the effect that a loss or malfunction of the following will have on the APRMILPRM:

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

A control rod block will occur due to a flow unit comparator trip. Correct. The comparator trip will alarm at 10 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.Incorrect. FBNF RB trip is set at 42 percent APRM power with the mode switch in Run.

A half scram will occur due to a flow unit "inop" signal. Incorrect. Flow unit hop trip by itself does not cause half scram.

A full scram will occur due to flow biased neutron flux utxcale. Incorrect

_ _ ~ _ _ _ _ _ _

iHC.OP-SO. SE-000 I 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 Predict the effect of each control switch on the APRMSlFlow Units, IAW the Student Handout.

Select the conditions or permissives required for the control switches to perform their intended function, IAW the Student Handout

~

Monday, June 23,2003 7:24:08 AM Page 93 of 162

- ~~

~~~

~

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 of the water levels from the top AND bottom of the notch.

~

An average of the water levels from all indicators

~~~

- that

- - are

- - notching.

~

c l

iB IMemory Hope Creek x

21 6000A104 69 Nuclear Boiler Instrumentation

- -~

~

-~

,Al.

~ Ability to predict andlor monitor changes in parameters associated with operating the NUCLEAR BOILER INSTRUMENTATION controls including:

b RXINSTEOI I I

(R) Given plant problemslindustry events associated with the Nuclear Boiler Instrumentation System:

Discuss the root cause of the plant problemlindustry event IAW the associated plant problemlindustry event document.

Discuss the HCGS design andlor procedural guidelines that mitigatelreduce the likelihood of the problemlindustry event at HCGS IAW the associated plant problem/industry event document.

_ ~ _ _ _ _

Monday, June 23,20037,24008 AM

- -~

Page 94 of 162

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.

b' Based on these conditions, declare that the associated:

channel must be placed in the tripped condition within one hour.

~

~_

_~

_ _ _ _ ~

~

channel must be placed in the tripped condition

_ _ _ - ~ _ _

within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

_~

- -~~

~

system is-inoperable and returned to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

~

channel must be placed in the tripped condition within twelve hours.

Hope Creek 06/17/2003

[Application J

I a

i 006222 70 Nuclear Boiler Instrumentation 2.2 1 Equipment Control 4.11 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . INCORRECT. Action for one NSSSS transmitter not common to RPS.

system inoperable and returned to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . 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-42-1 SHT 2 REV 14 TECH SPEC 3.3.4.1.B

'Tech Spec 3.3.4

a. Choose those sections which are applicable to the Redundant Reactivity Control System (ATWS Circuitry), IAW HCGS

b. Evaluate Redundant Reactivity Control System operability and determine required actions based on system inoperability (ANVS Circuitry), IAW HCGS Technical Specifications. SROlSTA ONLY

c. Explain the bases for those technical specifications associated with the Redundant Reactivity Control System (ATWS

~

Technical Specifications.

-~

~

Tech Spec sections 3.3.1 through3.3.4; P&ID M-42-1 Sheet I Significant1 ed

-2

~_

_ _ _ - _ _ _ ~ -

Facility Exam Bank Monday, June 23,2003 7:24:08 AM Page 95 of 162

.~

~

-~

~~~~~

Vision Bank QID# Q54823 Significantly modified.

~

Monday, June 23,2003 7:24:09 AM Page 96 of 162

-~

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 plantkystem response to this line break?

RClC pump discharge flow indication (BD-FIC-R600) decreases, RPV water level decreases.

RClC turbine speed indication (BD-SI-4280-1) increases, RPV -

water level increases.

~.

RClC turbine steam inlet pressure indication (BD-PI-R602) increases, RPV water level increases.

RClC pump discharge pressure indication (BD-PI-R601) decreases, RPV water level

~-

-~

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

Knowledge of the effect that a loss or malfunction of the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) will have on following:

K3.01

'Reactor water level L/

creases, RPV water level increases. Correct. Loss of oil 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.

RClC turbine steam inlet pressure indication (BD-PLR602) 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-R601 ) decreases, RPV water level decreases.

NOH01 RCICOO-00 system, IAW the RClC System Lesson Plan:

a.

A given valve opening or closure

b.

Loss of DC or AC power supply I c.

Inadequate system flow 1

d.

An oil system malfunction

' e.

1

f.

LOSS of room cooling 1

g.

Rupture disc failure on the RClC exhaust

h.

Steam line break Failure of the RClC Gland Seal Condenser Vacuum Pump Monday, June 23,20037.24

~

09 AM I

Page 97 of 162

~

~~

Monday, June 23,2003 7:24:09 AM Page 98 of 162

Given the following conditions:

- The Reactor Core Isolation Cooling (RCIC) system flow controller has failed full downscale demanding a "0" gpm flowrate.

L-

- 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?

RClC will start, accelerate...

to co y at approximately 4000 rpm.

~-

~

then will slow to and run at a low speed.

_- ~-~

~

-~

~

K5 K5 02 now indication 3.1 3.1 Knowledge of the operational implications of the following concepts as they apply to REACTOR CORE ISOLATION COOLING SYSTEM (RCIC):

The ramp generator runs RClC to about 4000 rpm until the flow siqnal comes out of saturation at which I

L time the low signal will control -> RClC running at min speed.

Monday, June 23,2003 7:24:09 AM 1

Page 99 of 162

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 psilmin.

- HPCl Pump is tagged for maintenance.

- All other ECCS systems have performed as expected.

c Assuming NO operator action, ADS SRVs will open immediately when:

_ ~ _ _ _ - _._

Level 1 is reached.

~_

Level I is reached and the 105 second timer times out.

~

Top of Active Fuel (TAF) is reached.

Top of Active Fuel (TAF) is reached and the 105 second timer times out.

I A4.

Ability to manually operate and/or monitor in the control room:

_ _ _ _ _ _ ~ ~ -

~

,4.2 4.2 Level 1 is reached and the 105 second timer times out. Correct. High DW pressure is present. ADS logic

-_.. _ _ ~

~

_ _ _ _ _ _ ~ _ _ -

will wait until Level 1 is reached to start the 105 second timer. After 105 seconds, ADS will open the ADS SRVs 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.

TOD of Active Fuel (TAF) is reached and the 105 second timer times out. Incorrect. Wrong setpoint.

~HC.OP-SO.SN-OOOI L-

---___~-

ADSSYSE007 (R) Given a set of conditions and a drawing of the controls, instrumentation and/or alarms located in the Control Room, identify I the status of the Automatic Depressurization System by evaluation of the controlslinstrumentationlalarms, IAW the Automatic

_ _ _. _ _ _ ~

~~

~

_ ~ _ _ _ _ _ _ _ _ _ _

_ _ _ _ _ ~ - - -..

.~

Page 100 of 162

~

Given the following conditions:

- The plant has scrammed due to a loss of offsite power.

- 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.

L Select the statement below which describes the response of ADS.

~-

ADS will initiateimmediately, regardless of Core

__ Spray and RHR status.

ADS will initiate in 105 seconds, only if Core Spray pumps A & C or RHR pump A or C are ADS will initiate immediately, only if Core Spray pumps B & D or RHR pump B or D are running.

- _ _ _ _. ~

_ ~

running.

_ _ _ _ ~

~

_~

_~~

~

_~

~

ADS will initiate in 105 seconds, regardless-of Core Spray and RHR status.

/Comprehension I Hope Creek 21 8000K402 74 K4.

I Knowledge of AUTOMATIC DEPRESSURIZATION SYSTEM design feature(s) and/or interlocks which I

~-

provide for the following:

ustification: IAW

. ADS will initiate immediately, regardless of core spray and RHR status.- Correct ADS will initiate immediately, if Core Spray pumps A & C or RHR pump A or C are running - Incorrect, the status of core spray/RHR does not effect manual initiation

. 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 -

~~

\\

~ 3.8 4.01 HC.OP-SO.SN-0001 section 3.3.1 I

ADSSYSEOOG, (R) Given a labeled diagram/drawing of, or access to, the Automatic Depressurization System controlslindication bezel, IAW the 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

c.

Determine the effect of each control on the Automatic Depressurization System t conditions or perrnissives required for the control switches to perform their intended function.

~

~_

igure of ADS __

logic _ _

pushbuttons.

c ed Monday, June 23, 2003 7:24:09 AM Page 101 of 162

~-

~

~ - -

Vision Bank QID# Q54168 editiorially modified.

~~

Monday, June 23,2003 7:24:10 AM Page 102 of 162

Given the following conditions:

- B RHR pump is running in Suppression Pool Cooling at rated flow.

- 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?

~_

- _~

oses after a 10 second delay.

~

-~

~

Automatically opens immediately

__ when the pump trips.

~~

Remains open after the pump trip.

~

~-

~

- -~~

Remains closed after the pump trip.

_ _ _ _ _ ~

~-

~

- _ _ _ ~ _

~

~~

06/17/2003 L

A4.

1 Abilitv to manuallv ooerate and/or monitor in the control room:

I

_ _ _ ~ _ _

A4.04

[Minimum flow valves mp 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.

~

_ - _ ~

~

I RHRSYSEOI 1 Given a labeled drawinq of, or access to the Residual Heat Removal System controlslindication on 1 OC650:

a.

b.

c.

d.

IAW the RHR System Lesson Plan.

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 switches/pushbuttons to perform their intended functions Monday, June 23,2003 7:24:10 AM Paae 103 of 162

- _~

~~~

~

While performing a RHR system test, the breaker for Torus Spray Isolation Valve BC-HV-F027B trips.

~~

e of electrical power supplies to the following:

P

~

P

~

~ _

1 OB222 Correct. Powered from 52-222083 which is a B channel 1 E MCC.

108323 Incorrect. B channel Non 1E MCC powered from 1E power which is shed on a LOCA Level 1 signal.

10B242 Incorrect. D Channel 1E MCC with similar valve load to 10B222.

1 OB563 Incorrect. B channel 1 E MCC which powers Station Service Water components.

RHRSYSEOOB (R) Given a system which physically connects to or is required to support the operation of the RHR System or components therein, explain the function of the supporting system, -

IAW the RHR System Lesson

~~ Plan.

~

~~~

~

Monday. June 23,2003 7:24.10 AM I

Page 104 of 162

_ _ _ _ _ ~~

~

The Containment Hydrogen Recombiners are directed to be placed in service following a LOCA in 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.

c

_ _ _ ~ _ _ _ _ _ _ ~

~~~

~

~~

~

Reset NSSSS isolation only.

~

B 223002 Primary Containment Isolation System/Nuclear Steam Supply Shut-Off

-1 06/17/2003 223002K408 77

-~

~

~ _ _

K4. 1 Knowledge of PCIS/NSSSS design feature(s) and/or interlocks which provide for the following:

itions 3.3 3.7

~-

~ _ _ _ _ _ _ _ _.... __-

~

IS 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 NSSSSOE003 '

(R) Provided access to control room references:

a.

Determine the source of electrical power for the NSSSS logic channels IAW the NSSSS Lesson Plan.

~

~~~~

~

_ ~_

~

i I

Monday, June 23,2003 7:24:10 AM Page 105 of 162

Given the following conditions:

- The plant is operating normally at 100 percent power.

- Overhead annunciation "RHR LOOP A TROUBLE A6-B1" alarms.

Which one of the following conditions would cause the alarm?

i

~-

The AP228 ECCS Jockey Pump tripped on overload. -~

~~-

_~

~

_ ~ _ _

~

The CP228 ECCS Jockey Pump suction

~- strainer clogged. - __

~~

~

The A RHR Pump room has a high water IeveT 0611 712003 L

~

A4.

' Ability to manually operate and/or monitor in the control room:

A4.04 IKeep fill system 2.8 2.7

~~

~

..._ ~-~

- ~

~

Justification:

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 HPCl 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

~ ~ _ _ _ _ _

~~

leakage past Loop Isolation valves.

I I HC.OP-AR.ZZ-0004 Attachment B 1 RHRSYSEOIP Given a set of conditions and a drawing of the controls, instrumentation andlor alarms located in the main control room, assess '

the status of the Residual Heat Removal System or its components by evaluation of the controlslinstrurnentationlalarrns IAW the RHR System Lesson Plan.

~-

- - ~-

~~_~

~

~~

~~~

~~

~~~

1 Page 106 of 162

- 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.

L-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.

Drywell spray outboard isolation valve only opens.

NEITHER Drywell spray valves will open.

Both Inboard and Outboard Drywell spray isolation valves open.

~-

~

~~~~~

~_

~

~ - _ _

~_

~~

~

_~

~- - _

~~

712003 08 79

~-

K1.

Knowledge of the physical connections and/or cause-effect relationships between RHR/LPCI:

CONTAINMENT SPRAY SYSTEM MODE and the following:

' 3 2 r3.4

~

A Justification:

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 mrav inboard isolation valve onlv oDens. Incorrect. Both valves will oDen.

L HC.OP-SO.BC-0001 3.3.5 I

None

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Monday. June 23,2003 7:24:10 AM 1

Page 107 of 162

Given the following conditions:

- Fuel Pool level is at the normal water level.

I

\\, 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?

I Skimmer surge tank level...

_ _ _ _ _ _ _ ~ -

increases and then returns to the level that exzed prior to starting the pump.

increases to a level higher than existed prior to starting the pump.

~

~~

~

'L

~ - - - _ _ _. - _ _ ~ _ _ _ _ _

~_

decreases until the pump trips on low tank level.

decreases and then increases to tslevel lower than existed Drior to startina the DumD.

_ _ _ _ _ _ ~ _ _

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~-

1 Ability to predict and/or monitor changes in parameters associated with operating the FUEL POOL I

,2.6 COOLING AND CLEAN-UP controls including:

1

~

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.

decreases and then increases to the level lower than existed prior to starting the pump. Correct..Lowers slightly then stabilizes slightly lower due to the restriction at the overflow pipe..

I FPCCOOEOOS (R) Concernina went fuel storaae DOOI water level, summarize. from memow. the followina IAW the Fuel Pool Coolino and None

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Monday, June 23,2003 7:24:10 AM Page 108 of 162

Which one of the following describes the interfaces of Station Auxiliaries Cooling System (SACS) and Station Service Water (SSW) for the Fuel Pool Cooling and Cleanup System?

I

- SSW A Loop as a level makeup source.

- II

- SACS A Loop as a level makeup source.

I l l

- SACS B Loop as a cooling water source.

IV - SSW B Loop as a level makeup source.

I and Ill onlv.

I, II and Ill only.

I, Ill, and IV only.

Hope Creek 06/17/2003

[Memory I

233000K109

~

physical connections and/or cause-effect relationships between FUEL POOL COOLING 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. II and I l l onlv. Correct. Sacs is not a MU Source.

(FPCCS) Sysiem Lesson-plan.

Cleanup (FPCCS) System Lesson Plan:

a.

FPCCOOE008 How normal level is controlled L

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Monday, June 23,2003 7 2430 AM

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Page 109 of 162

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Given the following conditions:

- A new fuel bundle is grappled and lifted in the Spent Fuel Pool for placement in the RPV.

- The "Slack Cable" light remains lit.

- The refueling platform main hoist load cell indicates 0 pounds.

- The bundle is lowered, reseated, and released.

L Which of the following operations can continue?

Control rod removal from the RPV using the refueling bridge main hoist.

Control rod removal from the RPV using the frame mounted auxiliarv hoist.

I

~~~

~

~___________.

Fuel transfer within the Spent Fuel Pool using the monorail auxiliary hoist.

~~

~

~_

I

___._.--A 3.4 4.1~

-~

2.2.22 /Knowledge of limiting conditions for operations and safety limits.

Justification:

SRO 1 OCFR55.43 (7) Fuel handling facilities and procedures.

SRO 1 OCFR55.43 (2) Facility operating limitations in the Technical Specifications and their bases.

Justification:

Control rod removal from the RPV using the frame mounted auxiliary hoist. -Correct. The auxiliary hoist can be used for control rod removal IAW AP-108. SH.OP 108 action requires suspension of inoperable equipment use involving control rod or fuel assembly movement. Continued control rod movement with aux hoist is allowed.

Control rod removal from the RPV using the refueling bridge main hoist.. -Incorrect-The main mast cannot be used for core alts. See SH.OP 108 required action Fuel transfer within the Spent Fuel Pool using the monorail auxiliary hoist. -Incorrect-Not allowed to move fuel with the Aux hoists. Fuel bundle is too heavy for aux hoists.

Fuel transfer within the Spent Fuel Pool using the frame mounted auxiliary hoist. -Incorrect-Not allowed to move fuel with the aux hoists. Fuel bundle is too heavv for aux hoists.

1 SH.OP-AP.ZZ-0108 Exhibit 3

~~

~

REFUELEOI 2 (R) Given a scenario of applicable operating conditions and access to Technical Specifications:

a.

Specifications.

b.

Specifications.

c.

(SRO only)

Choose those sections which are applicable to the refueling platform and associated equipment IAW HCGS Technical Evaluate Refuel Platform operability and determine required actions based upon system operability IAW HCGS Technical Explain the basis for those Tech Spec items associated with the refuel platform IAW HCGS Technical Specifications.

_~

'HC Tech Spec section

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Monday, June 23, 2003 7:24:11 AM 1

Page 110 of 162

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Vision Bank QID# Q56966 Significantly modified.

>w Mondav. June 23,2003 7:24:11 AM Page 111 of 162

Given the following conditions :

- The plant scrammed from an inadvertent MSlV closure.

- 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 1.4 psig.

Based on these indications, what is causing the drywell pressure to rise and what action is required to mitigate the event?

~

~-

- ~

~

H SRV is stuck closed. Cycle H S H SRV tailpipe vacuum breaker is stuck open. Place H SRV handswitch to closed position.

P SRV is stuck open. Cycle P SRV.

~~

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~-

P SRV tailpipe vacuum breaker is stuck closed. Place P SRV handswitch to open position.

06/17/2003

~~

_ _ ~

~- - ~~

A2.

Ability to (a) predict the impacts of the following on the RETEFKAFETY VALVES; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

SRV tailpipe vacuum breaker is stuck open. Place H SRV handswitch to closed position. Correct. RPV 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.

H SRV is stuck closed. Cycle H SRV. Incorrect. Pressure range indicates H SRV is operating on LO-LO Set setpoints. If SRV Instrument gas supply line ruptured, DW pressure would rise. Wrong action.

P SRV is stuck open. Cycle P SRV. Incorrect. RPV pressure would continue to lower. Wrong action.

8P SRV tailpipe vacuum breaker is stuck closed. Place P SRV handswitch to open position. Incorrect. A stuck closed SRV tailpiDe vacuum breaker can not be diaanosed from the info aiven. Wrona action.

_~

IEOP-101

~~

~

~~

~

~ -

ifin E0101PE008 (R) Given any step of the procedure, describe the reason for performance of that step and/or expected system response to control manipulations prescribed by that step.

___ ~ _ _ _

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EOP-101 flowchart

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Monday, June 23, 2003 7:24:11 AM

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Page 112 of 162

T = 0 sec 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 1) annunciators alarm.

T = 24 sec ADS CH B INITIATION PENDING (RPV Level I) 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.

LOCA occurs.

When will ADS be initiated manually or automatically?

Before RPV level reaches -1 90 inches: manuallv.

~

When T = 425 seconds; automatically.

When T = 302 seconds; automatically.

~

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_~~

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After RPV level reaches -200 inches; manually.

06/17/2003 239002

] ReliefEafety Valves 84 LVES; and (b) based on thoscl I

predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

1 -

-~

~-

4.1 4ra

\\./

'Before RPV level reaches -190 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 ADSSYSE007

~

(R) Given a set of conditions and a drawing of the controls, instrumentation and/or alarms located in the Control Room, identify luation of the controlslinstrumentationlalarms, IAW the Automatic

~

Monday, June 23,2003 7:24:11 AM 1

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Paoe 113 of 162

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 reactor pressure would indicate that a Safety Limit violation had occurred?

Reactor RPV Power Pressure 1

750 psig

_ ~~

~

28% 775 psig 32%

810 psig 2.2 Equipment Control 2.2.22 3.4 4.1

~. _

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~~_

IKnowledge of limiting conditions for operations and safety limits.

J USTl FICATION:

55.43(2) Facility operating limitations in the Technical Specifications and their bases.

Correct Answer: 28% 775 psig - Power is greater than 25% with pressure less than 785 psig The 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 psig - Power is less than 25% with pressure less than 785 psig.

c TECSPCEOI 0 (R) Given specific plant operating conditions and a copy of the Hope Creek Generating Station Technical Specifications,

_ _ ~ -

Significantly Modified Facility Exam Bank 56697 Significantly Monday, June 23,2003 7:24:11 AM 1

Paae 114 of 162

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.

- Main Turbine inlet pressure rise is and reactor steam dome pressure rise is 06/17/2003 241 000K504 241 000 I Reactormurbine Pressure Regulating System K5,Knowledge of the operational Implications of the following concepts as they apply to REACTORlTURBlNE '

-~

_ _ _ _ ~-

3.3

~ 3.31 b 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 Non-Linear; Non-Linear. Incorrect. MT inlet pressure rise is linear.

L NOH01 EHCLOG-00 figure 2 EHCLOGEOOP

~

(R) Given plant conditions evaluate the cause-effect relationship between the pressure regulating system and the following IAW the Lesson Plan:

Reactor Power Reactor Pressure Steam Flow Reactor Water Level

________________-~____~__

_ _ ~ _ _ _ _

None Monday, June 23,2003 7:24:11 AM Page 115 of 162

- The plant is operating at 100% power.

- Main Turbine testing is in progress.

- The LOCKED OUT pushbutton on lOC650E has been depressed, energizing the lockout valve.

I L,

_~

Which of the following describes the effect of energizing the Lockout Valve?

The Master Trip Solenoid is bypassed to prevent depressurizing the Emergenc All Turbine trips are bypassed to allow for testing.

All Turbine trips EXCEPT for the mechanical overspeed trip are bypassed.

stem.

~

_ _ _ ~ _ _

~-

~~

Only the Turbine mechanical

~

overspeed trip is bypassed. -

~

245000K503 87

~

-~

K 5. 7 Knowledge of the operational implications of the following concepts as they apply to MAIN TURBINE

-~

GENERATOR AND AUXILIARY SYSTEMS:

2.6 2.6 E

~

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 fluid from the FAS header is supplied to the steam admission valve disc dump valves. Does not bypass the Master Trip Solenoid.

The Master Trip Solenoid is bypassed to prevent depressurizing the Emergency Trip System. Incorrect.

MTS is not bypasses.lt remains active.

All Turbine trips are bypassed to allow for testing. Incorrect. Only Mechanical Overspeed trip is bypassed to allow testing.

All Turbine trips EXCEPT for the mechanical overspeed trip are bypassed. Incorrect.

~

Reverse of actual.

Mechanical Trip Valve Lockout Valve

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~

Master Trip Solenoid Valve Monday,June 23, 2003 724:Il AM

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Page 116 of 162

- The plant is operating at 100 percent power.

- The Hotwell level control is selected to 'A on 10C651A.

- A large pipe break occurs on the tube side of the in service SJAE Condenser.

Which one 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'.

..~

- _ _ _ ~ -

_ ~ _ _ _ _ _ ~ ____

~-

c All Primary Condensate Pumps trip; Trip all Secondary Condensate and Feedwater Pumps.

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.

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!Comprehension 1

Hope Creek 06/17/2003 256000A206 aa Ability to (a) predict the impacts of the following on the REACTOR CONDENSATE SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

1 I

Hotwell 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.

Wrona action per AB.

M 1

associated with the runback and/or trip logic of the condensate, feedwater and reactor recirculation systems.

_ _ _ _ ~

~

ABRPV4EOO3

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c Monday, June 23,2003 7:24:1 I AM Page 117 of 162

Given the following conditions:

- Plant power level is 90 percent.

- # 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?

L

_ _ _ _ _ ~

~

The Main Turbine trips.

2B FWHTR Bleeder Tripvalves trip closed.

1, DC, & 2B FWHTR Condensate inlet and outlet valves auto close.

1, DC, & 2B FWHTR Startup and Operating Vents auto close.

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K4.

Knowledge of REACTOR CONDENSATE SYSTEM design feature(s) and/or interlocks which provide for

_ _ _ ~

K4.06 Control of extraction steam

1, DC, & 2B 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 BWs. Correct for other heaters except 1,2,&DC.

1. DC, & 2B FWHTR Startun and Oneratina Vents auto close. Incorrect. Vents auto open.

Y c

_ _ - ~

HC.OP-SO.AE-0001

_ _ _ _ _ _ _ ~

FWHEATEOOS 1 (R) From memory, describe the effects of too high or too low of a feedwater heater water level.

~

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I I

Monday, June 23,2003 7:24:12 AM Page 118 of 162

- Reactor power is 14 percent.

- 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.

L Which one of the following choices describes the approximate positions of the SULCV's LV-1754 and 1785?

percent open.

~

percent open and SULCV 1785 is SULCV 1754 is 22; 0

A3.07 IFWRV position 3.2 3.21 his valve is full open. 1785 is the 12

-L 22; 0;

2:

0 Incorrect - 1754 is 100 percent open. 1785 is 2 percent open.

22 Incorrect - 1754 is 100 percent open. 1785 is only 2 percent open..

100 Incorrect. Reverse of actual positions.

Digital Feed Drawing 1 H-AE-ECS-0128-03F LP NOH01

~

~-

~

Feedwater Control System Lesson Plan:

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.

Figure of SULCV control 1

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Monday, June 23,2003 7:24:12 AM Page 119 of 162

~ _ _

Given the following conditions:

2.1.33

- The plant is operating at 16 percent power during a startup.

- An I&C Techinician has completed channel functional testing on RPV Level 8 instrumentation to L.-

the Digital Feedwater System.

Ability to recognize indications for system operating parameters which are entry-level conditions for technical specifications.

3.4 4.01 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 ho within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

~

Place the affected channel(s) in the tripped condition AND restore one channel to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least STARTUP within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

I Phce 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

~

Hope Creek 06/17/2003

/Application I

25900261 33 c

91 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . 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 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least STARTUP within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . 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 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least STARTUP within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Action far 2 inoperable channels.

_~

FWCONTEOI 8 (R) Given a scenario of applicable operating conditions and access to Technical Specifications.

a. Identify those sections which are applicable to the Feedwater Control System.

_ _ _ _ _ ~ _ _ _ _

Page 120 of 162 1

\\--

Monday. June 23,2003 7:24:12 AM

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) ~_

~

-1 Monday, June 23,2003.

7:24:12 AM

~ 1 Page 121 of 162

- The plant is at 100% power.

- The "C" steam flow detector for the feedwater level control system fails low (its output indicates 0 L-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 level control system will remain in three element control.

Reactor water level will decrease and stabilize at a lower than normal value. The feedwater level control system will shift to single element control.

Reactor water level will increase and stabilize at a higher than normal value. The feedwater level control system will remain in three element control.

~

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06/17/2003 K3.I Knowledge of the effect that a loss or malfunction of the REACTOR WATER LEVEL CONTROL SYSTEM K3.07 /Reactor water level indication will have on following:

~

3.4 3.4 J USTl Fl CAT1 ON Reactor water level will remain the same. The feedwater level control system will shift to single element control.

Correct - the system transfers to Single Element control. With stable conditions single and three element should control at the same setpoint.

L Reactor water level will decrease and stabilize at a lower than normal value. The feedwater level control system will remain in three element control.

Incorrect - level should not decrease; the system will shift to single element control.

Reactor water level will decrease and stabilize at a lower than normal value. The feedwater level control system will shift to single element control.

Incorrect - level should not decrease.

Reactor water level will increase and stabilize at a higher than normal value. The feedwater level control system will remain in three element control.

Incorrect - level should not increase: the svstem will shift to sinale element control.

~

IEngineering

~- Drawing H-I-AE-ECS-0128-0 FWCONTE013 (R) From memory, describe the response of the FWLC System if the total steam flow signal were to be lost, IAW the Feedwater

~ -.

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Monday, June 2 3 x 0 3 7:24:12 AM Page 122 of 162

None c

Monday, June 23,2003 7:24:13 AM 1

Page 123 of 162

Given the following conditions:

- The plant is operating at 100 percent power.

- AD483 Inverter output power is lost.

What effect does the loss have on the 'A' RFPT?

'A RFPT will NOT trip on RPV Level 8.

'A RFPT trips due to Loss of Control Oil pressure.

~-

'A' RFPT trim due to Loss of SDeed Sianal.

'A' RFPT will NOT trip on Overspeed.

[Memory j

Hope Creek 06/17/2003 i

/B 262002K101 93

~. __-___ -

~

K1.

Knowledge of the physical connections and/or cause-effect relationships between UNINTERRUPTABLE

~~

./D.C.) and the following:

- ~ _ _. _ _ _

I

~

_-__-pppp2 ntrol: Plant-Specific due to Loss of Speed Signal. Correct. ( ~ 1 0 0 rpm with 60 sec time delay with the turbine control valves open.)

'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.

'A RFPT will NOT trip on Overspeed. Incorrect. Overspeed is mechcanical and will trip if needed.

L' I HC. OP-AB.=-01 36 FWCONTEOIG (R) Given any of the following systems, state the interrelationship between the FWLC System and that System, IAW the Feedwater Control System Lesson Plan:

a. 120 VAC Non-lE Electrical Distribution

b. 125 VDC Non-1 E Electrical Distribution

c. Main Turbine

d. Recirculation System

e. Rod Worth Minimizer (RWM)

f. Main Steam

g. Redundant Reactivity Control (RRCS)

h. Hydrogen Water Chemical Injection System Monday, June 23, 2003 7:24:13 AM

-- -_---A Page 124 of 162

~

Given the following conditions:

DCELECE005

- The lAD413 125 Volt Battery Charger is in service and providing a normal charge on its battery.

- 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.

L

~

(R) Summarize the interrelationship(s) between 125VDC 1 ElNlE Power Systems and the following IAW the DC Electrical Distribution Lesson Plan.

a.

48OVDC 1 E/N1 E Power Supply xilia Which of the following describes the response of this battery charger?

[Comprehension i Hope Creek 06/17/2003 AI.

DISTRIBUTION controls including:

AI.01 IBattery chargingldischarging rate Justification:

return to the "float" mode to recharge the battery. Correct. Although the charging rate will be higher than prior to the charger loss, the charger will remain in the Float mode.

trip and is interlocked "off I' with the diesel generator powering the bus. Incorrect. The charger does not 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.

' Ability to predict and/or monitor changes in parameters associated with operating the D.C. ELECTRICAL-

~~

2.5 2.8

~~

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INPO Exam Bank INPO Bank QID # 14168 03/26/2001 Pe Monday, June 23.2003 7:24:13 AM I

Page 125 of 162

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.

- 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.

L Which one of the following actions must be taken to mitigate this event?

~

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

Raise speed and voltage from 1OC651 panel to within limits and verify o automatically 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 local/remote panel Emergency Shutdown pushbuttons. -~

Turn on the Synchroscope key and manually close EDG A output breakerfrom the Control Room.

/Comprehension ]

Hope Creek 06/1 7/2003 264000A304 95

~

Ability to monitor automatic operations of the EMERGENCY GENERATORS (DIESELIJET) including:

rol 3.1 3.1

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_~

L y Shutdown pushbuttons. Correct.-Voltage and

. 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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Raise speed and voltage from 1 OC651 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.

I HC.OP-SO. KJ-0001 NOH01 EDG000-00 I

EDGOOOEOl7

~~

~

Monday, June 23,2003 7:24:13 AM

_~

Page 126 of 162

Monday, June 23,2003 7:24:13 AM

~~~

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Page 127 of 162

Given the following conditions:

- The Starting Air Compressor I DK402 is Safety tagged for maintenance.

- The 'D' EDG Starting Air Receivers are crosstied to the 'B' EDG. (See attached figure)

- While pressurizing the 'D' EDG receivers, the crosstie air hose splits open.

W 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.

B

. _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ~

and D EDGs will respond.

-~

-. -~

~- -

~

Only

- D EDG _ -

will. respond.

~

~~

Neither B NOR D EDG will respond.

~

0611 7/2003 264000K601 96 264000 I Emergency Generators (DieseVJet) 7

,K6.

Knowledge of the effect that a loss or malfunction of the following will have on GENERATORS (DI ESEUJET):

I 3.8 13.9

~

B and D EDGs will respond. Correct. Check valves in the line maintain receiver pressure to allow both diesels to start on the LOP Only B EDG will respond. Incorrect. Both diesels will start.

Only D EDG will respond. Incorrect. Both diesels will start.

Neither B NOR D EDG will respond. Incorrect. Both diesels will start.

a. A normai shutdown is in progress.

Figure from Section 5.11

-~

- 7 u

Monday. June 23,2003 7:24:14 AM Page 128 of 162

Given the following conditions:

- 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?

-~

__ ~

eactor Buildin;-North Equipment Sumr, Room.

L-

_ _ ~ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

2.4 Emergency Procedures and Plan abnormal indications for system operating parameters which are entry abnormal operating procedures. _

~

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 1 BT266 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 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: RCIC Pump and Turbine Room. Rad levels would increase slightly from steam line drains unless RClC was placed I/S. Stem does not sumor? RClC operation.

EOP-103/4 Monday, June 23,2003 7:24:14 AM 1

~

Page 129 of 162

~

~~_

-~ -

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

\\.

closes.

Which one of the following conditions would cause this termination?

(Assume no oDerator action) v Liquid Radwaste Effluent radiation element fails low.

Cooling Tower Blowdown weir flow rate fails low.

~~~

~

Liquid Radwaste Effluent sample flow rate fails high.

~

~_

Cooling Tower Blowdown RMS radiation element fails high.

272000 Radiation Monitoring System 06/17/2003 K3.

1 Knowledge of the effect that a loss or malfunction of the RADIATION MONITORING System will have on following:

3.8 Cooling 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.

Coolina Tower Blowdown RMS radiation element fails high. Incorrect. Cause alarms but not isolation.

(R) From memoly lisffidentify the five 1

~

~~~

Mondav. June 23.2003 7:24:14 AM 1

Page 130 of 162

Given the following conditions:

- The plant is operating at 100 percent power during hot summer conditions.

- CRIDS page 105 indicates Reactor Building Backdraft Damper PD-9438Cl is closed.

- All room temperature points are reading NHI.

- NO other Backdraft Dampers are closed.

L d

What impact will this closure have on the plant?

~

MSlV closure is imminent. A Reactor scram is necessary.

Reactor Water Clean Up will isolate.

FRVS must be placed in service to maintain room temperatures.

~

Associated room temps will rise, Reset the damper.

~~

-~

~

_~

~

06/17/2003 0

99 2.4 I Emergency Procedures and Plan 1

2.4.50

[Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.

____ 3.3 Justification:

Associated 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. Misinterpretation of table of.

Reactor Water cleanup will isolate. Incorrect. Damper is one of several that feed into RWCU pipechase room.

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 Attachment 2

_ _ _ -~

Monday, June 23, 2003 7:24:14 AM Page 131 of 162

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.

c Which one of the following describes the plant response, if any?

Neither Reactor Building Ventilation Inboard and Outboard Dampers HD-9414 Reactor Building Ventilation Inboard Dampers HD-9414A and HD-9370A onlv close.

9370A & B close.

~

~-

~~~~

~.

Reactor Building Ventilation Outboard Dampers HD-94148 and HD-9370B onlv close.

~

_~~

~

Both Reactor Building Ventilation Inboard and Outboard Dampers HD-9414A & B and HD-9370A & B close.

I

c.

06/17/2003

- ~-

of the SECONDARY CONTAINMENT including:

~ _ _ _ _ _ _

_ ~ _. _ _ _ _. _ _ ~

_ _ ~ _ _ _

~

Neither Reactor Building Ventilation Inboard and Outboard Dampers HD-9414A & B or HD-9370A & B 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-94146 and HD-93706 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.

HC.OP-AR.ZZ-O019(Q) Attachment A3 SECCONEOO8 (R) Given a list of plant conditions, select the four automatic signals which will shutdown and isolate normal Reactor Building Ventilation and start the Filtration Recirculation and Ventilation System (FRVS) IAW the Secondary Containment Lesson Plan.

~

-~

Monday, June 23, 2003 7:24:14 AM Page 132 of 162

Given the following conditions:

SECCONEOOZ

- The plant is operating at 100 percent power.

- The Main Steam Tunnel (MST) Ventilation Barrier (Panel) 10S203 indicates open on the RM-11.

Which one of the following describes the operational impact?

L Turbine Building Exhaust RMS levels will rise to alert levels.

Degraded cooling capability for the MST Coolers.

Main Steam Line RMS detectors will read non-conservative.

Loss of secondary containment integrity.

~

~-

~~~~~

~

_~

From memory describe the components (structures) that make up the Secondary Containment IAW the Secondary Containment Lesson Plan 0611 712003 1

001 K107 290001 Secondary Containment 101 K1.

1 Knowledge of the physical connections and/or cause-effect relationships between SECONDARY CONTAINMENT and the following Justification:

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 Buildinq wil not drive RMS values up unless there is a system breach in the MST.

Monday, June 23,2003 7:24:14 AM Page 133 of 162

Following sustained steady state operation at 100% power for 103 hours0.00119 days 0.0286 hours 1.703042e-4 weeks 3.91915e-5 months , indicated reactor power drops to 97% without operator action. Recirculation flow and rod positions have NOT changed.

Which of the following is the explanation for this change in power?

c

-~

- -~

Core shroud cracking has occurred.

Feedwater flow

~

to the reactor has risen.

Steam quality exiting

~~

the steam dryers has been reduced.

- _ _~

~ _ _ _ _

~

_~

~_

~

-~

~

~- ~-

~

~~~

-~

-~ --

_ - ~~-

~

_ _ ~ _ -

An Electro-Hydraulic Control (EHC) system chanae has caused reactor pressure to rise.

06/17/2003 102

~

3.3 3.4 JUST1 FICATION:

Correct answer: "Core shroud cracking has occurred." BWR OE. Power drops due to steam bypassing the moisture seperator and heating the feedwater in the annulus area.

Feedwater flow to the reactor has risen. Incorrect. A rise in feedwater flow will lower the inlet enthalpy and temperature, increasing coolant density and lowering void fraction and thereby causing power to rise.

Steam quality exiting the steam dryers has been reduced. Incorrect. Lowered steam quality has no direct bearing on reactor power.

An Electro-Hydraulic Control (EHC) system change has caused reactor pressure to rise. Incorrect. A higher reactor pressure suppresses the void fraction, causing power to rise.

IGE-NE-523-148-1093 DRF 137-001 0 GE BWR Core Shroud Evaluation IBWROG Letter dated 7/13/94 I

, a.

b.

HCGS IAW the plant/ industry event.

c.

Discuss the root cause of the plant problemlindustry event IAW the plant/industry event.

Discuss the HCGS design andlor procedural guidelines that mitigatelreduce the likelihood of the problemlindustry event at Discuss the "lessons learned" from this problemlevent IAW the planthndustry event.

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

~

_ _ _ ~ _ _ _ _ _ ~

None Mondav, June 23, 2003 7:24:14 AM 1

Page 134 of 162

Given the following conditions:

- The Control Room Ventilation System "A" is operating normally.

- 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-3000 scfm; B-I 000 scfm; C-4000 scfm; D-0 scfm; E-I 8500 scfm

_~ --__

A-4000 scfm; B-0 scfm; C-0 scfm; D-4000 scfm; E-I8500 scfm

~-

~~

-~ - _._

A-0 scfm; B-4000 scfm; C-4000 scfm; D-0 scfm; E-I4500 scfm

_~

~-

A-0 scfm; B-I 000 scfm; C-4000 scfm; D-3000 scfm; E-I4500 scfm 290003A301 103

~

~ _ _ _ _ _ _ _

~

_ _ - -~

ONTROL ROOM HVAC including:

J USTl FI CAT1 ON :

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, B-I 000 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" &

'ICt are 0 during normal operation. "E8 is 18,500 during normal operation.

INCORRECT - A-0 scfm, 6-4000 scfm, (2-4000 scfm, D-0 scfm, E-I4500 scfm. "Ba 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. "AI is only 3000 during normal operation. I'D" is never 0. "E" is only 18,500 during

____-- normal operation.

~-

' the lesson plan:-

Normal operation Isolate: Outside Air Mode

, Isolate: Recirc Mode I

~~

I I Simplied Control Room HVAC figure Monday, June 23,2003 7 24.1 5 AM

~

Page 135 of 162

_ _ _ ~.

Given the following conditions:

- A loss of coolant accident has occurred.

- 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.

The ElAC will NOT automatically start but can be started locally after relieving intercooler pressure.

The ElAC is NOT available until the LOCA signal is cleared, PClS reset, and the 1E breaker is closed.

The ElAC is NOT available until the Non-I E breaker is closed and instrument air pressure is less than 85 psig.

'v

~-

~

- ~

~

~_ --

~

06/17/2003 K5.

K5.01 lAir compressors Knowledge of the operational implications of the following concepts as they apply to the INSTRUMENT AIR SYSTEM:

-~

2.5 ORRECT: The ElAC is not available until the LOCA signal is cleared, PClS reset, and the 1E breaker

-.-/

closed.

INCORRECT: The ElAC will automatically start on instrument air header pressure less than 85 psig if the LOCA signal is cleared. Breaker not reset INCORRECT: The ElAC is not available until the Non-1 E breaker is closed and instrument air pressure is less than 85 psig. 1 E breaker that needs resseting INCORRECT: The ElAC will not automatically start but may be started manually from the Control Room or locallv. Not until 1 E breaker is reset.

iHC.OP-SO. KB-0001 INSAIRE015

~

IR) From memorv. determine the resDonse of the emeraencv instrument air cornDressor to the followina conditions. IAW the I

a.

Loss of Offsite Power (LOP)

Inskument Air System Lesson Plan: '

b.

c Loss of Coolant Accident (LOCA)

Compressor intercooler pressure > 5 psig and a start signal received

_ _~

Mondav. June 23. 2003 7:24:15 AM Page 136 of 162

Given the following conditions:

- The plant is operating at 100 percent power when a Loss of Offsite Power (LOP) occurs.

- B Emergency Diesel Generator (EDG) trips due to electrical fault.

- D SACS Pump trips on overload.

- All other equipment functions properly.

l/

Which of the following actions is required?

Open SSW to RACS crosstie valves.

~

Attempt one restart of B EDG.

Emergency stop D EDG.

.~

Start B SACS Pump.

~~

t that a loss or malfunction of the CCWS will have on the following:

-~

~

3.31

~

- - ~

--___.___~-__

~~~

-~

Emergency stop D EDG. Correct. D EDG is running without cooling. It must be emerge because the LOP signal disables the normal stop controls.

Attempt one restart of B EDG. Incorrect. Incorrect application of AP-0109 provision for breaker re-closure. conditions given require an inspection of the electrical equipment.

Open Service Water to RACS crosstie valves. Incorrect. SSW to RACS valves remain open on a LOP.

Start B SACS Pump. Incorrect. B SACS Pump does not have power due to the B EDG trip.

IHC.OP-SO.KJ-OOOI 1

IHC.OP-AB.ZZ-01 35 COCA and/or LOP signal IAW available control room references EDGOOOEOIO Page 137 of 162

Given the following conditions:

Station Service Water (SSW) pump status:

- 'A' SSW pump I/S in AUTO.

- 'B' SSW pump I/S in AUTO.

- IC' SSW pump O/S in AUTO.

- 'D' SSW pump O/S in AUTO.

Which one of the following will result in the automatic start of the 'D' SSW Pump?

'A' ssw Loop low flow.

~-

'A' SSW Pump low flow.

_ _ ~

~

p low flow. Correct. Low flow of the opposite pump (6) within the affected loop (6) starts the 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 Loop low flow. Incorrect. Low pump flow of the opposite pump..

SERWATEOOS (R) Identify/describe the signals that auto start the Station Service Water System. IAW available control room references L

1 Monday, June 23,2003 7:24:15 AM I

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Page 138 of 162

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.

L-Which of the following describes the MINIMUM requirements to perform this evolution?

~

Communications via page announcements to the operator on the scoop tub must be RO licensed.

~

~.

_~

~

.. _ _ _ ~~

~

Communications via radio the operator on the scoop tube; the operator must be RO licensed.

Communications via telephone to the operator on the scoop tube; the operator must be SRO licensed.

Communications via sound powered phone to the operator on the scoop tube; the operator must be SRO licensed.

~~

~_~

~ _ _

~

_ _ _ ~ _ _ _ _ _ _ _ _ _ _

~

/Memory I

Hope Creek 0611 712003 294001 G I 08 107 2.1 Conduct of Operations 2.1.8 I

3.8 3.61

~

IAbility to coordinate personnel activities outside the control room.

Communications via page announcements to the operator on the scoop tube; the operator must be RO licensed. Incorrect. One way page anouncement is permitted only durin emergencies.

Communications via radio to the operator on the scoop tube; the operator must be RO licensed. Correct.

The radio allows 3 way communications from the scoop tube positioner. RO license is required since moving the scoop tube directly changes reactivity.

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.

Incorrect. Onlv RO license reauired.

/HC.OP-SO.BB-0002 3.1.6 reactivity or power level, IAW NC.NA-AP.ZZ-0005, and HC.0P-AP.ZZ-0005

~

~_

~

RECIRCEO16 (R) Given procedure HC.OP-SO.BB-0002, Reactor Recirculation System Operation, explain the bases for listed pr

~-

I INPO Bank QID # 21 193 Dresden 06/14/2002 Monday, June 23,2003 7:24.15 AM Paae 139 of 162

Given the following conditions:

- The plant is operating at 29 percent power.

- Overhead Annunciator C5-C2 TCV FAST CLOSURE & MSV TRIP BYP is ILLUMINATED.

Then the Main Turbine Generator trips.

1-

- 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.

_ _ ~ _ _

_. ~ _ _ _ _ _ _ _ _ _ ~ _ ~

~

15 minutes.

294001G111 108 2.1

' Conduct of Operations

-~

s for systems.

-~

L-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 wrona.

I Monday, June 23, 2003 7:24:15 AM

~

~_

Page 140 of 162

Given the following conditions:

- HPCl is removed from standby to perform HC.OP.IS.BJ-0101 HPCl System Valves Inservice Test.

- Valve BJ-HV-F042 Suppression Pool Suction Valve stoke time is 2 seconds longer than the "TECH SPECS OR DESIGN LIMITS" value.

- Valve BJ-HV-F004 CST Suction Valve strokes satisfactory and was returned to open position.

Which one of the following actions is required?

L, I

Deactivate F004 open; HPCl remains operable.

Deactivate F004 open; declare HPCl inoperable.

~

Deactivate F042 closed; HPCl remains operable.

~

Deactivate F042 closed: declare HPCl inoperable.

Hope Creek 06/17/2003 d

[Application 1

294001G112 109

~~

Conduct of Operations

~

~_

2.1.12

[Ability to apply technical specifications for a system.

Justification:

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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . HPCl operability requires suction source 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 Suppression Pool, therefore HPCl is inoperable.

HPCIOOE018

~

(R) Given plant conditions and access to Technical Specifications:

1 Select those sections which are applicable to the HPCl System IAW HCGS technical specifications.

I Evaluate HPCl System operability and required actions based upon system operability IAW HCGS technical specifications.

(SRO Only)

ExDlain the bases for those technical sDecification items associated with the HPCl System IAW HCGS technical specifications.

HCTS section 3.5.1 and 3.6.3

~_

~

Monday, June 23, 2003 7:24:15 AM Page 141 of 162

_~

Select the statement that satisfies 1 OCFR50.46 Acceptance Criteria for ECCS.

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.

Peak Cladding Temperature icTlculated maximum fuel element cladding temperature shall-NOT exceed 21 00°F.

Maximum Cladding Oxidation - calculated total oxidation of the cladding shall nowhere exceed Maximum Hydrogen Generation - calculated total amount of H2 generated from the chemical L

~

_ _ _~

_~

~_

_ _ _ ~ _ _

_ ~ _ _ ~ _ _ _ _ _ _ _ _ _ ~

~

21 % times the total - cladding thickness

~_ before oxidation.

~

~-

reaction of the cladding with water or steam shall NOT exceed 17Y0 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.

Hope Creek a-1

[Memory I

29400 1 G 1 28 Generic Knowledge and Abilities 110 2.1 Conduct of Operations

~~

he purpose and function of major system components and controls.

on 3.2 3.3

~

_ _ - ~ _ _

_~

Peak Cladding Temperature shall not exceed 2200 degrees F Incorrect-Maximum Cladding Oxidation shall nowhere exceed 17% times the total cladding thickness before oxidation.

Incorrect-Maximum Hydrogen Generation - shall not exceed 1% 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.

Correct-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 rewired bv the lonu lived radioactivity remaining in the core.

11 OCFR50.46 Acceptance Criteria Systems, IAW the Introduction to ECCS Student Handout.

Direct From Source Bank Vision Bank QID# Q56873 ____

Monday, June 23,2003 7:24:16 AM Page 142 of 162

-~

Given the following conditions:

A reactor shutdown is in progress.

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 umho/cm

- Condensate demin effluent conductivity - 0.08 umho/cm

- Reactor Water Cleanup conductivity - 0.07 umho/cm

- Reactor coolant sample conductivity - 0.07 umho/cm 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.

~

_ _ ~ -

~

-~

~_

kd Crudburst from removina HWCl from service: restore HWCl-to-service.

~

~_

Main Condenser tube leak; isolate the affected condenser waterbox.

Reactor fuel pin cladding leak; continue power reduction at normal rate.

_ ~~_

~

- - __ _ _ _ _ ~

-~~

003 111

-~

Lfl Conduct of Operations I 2.3 12.9 1 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.

Incorrect. 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

~

Tech SDecs Table 3.3.2-1.

HWCIOOE006 1

(R) Explain the plant operating restrictions when a power reduction event occurs that results in reactor power below 20% of rated thermal power without the required Main Steam Line Radiation Monitor setpoint change, IAW HC.OP-IO.ZZ-0004

~

Monday, June 23, 2003 143

c c

Mondav. June 23.2003 7:24:17 AM Paae 144 of 162

_ _ _ _ _ ~ - - -

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.

L, The Operations Superintendent...

- _ _ _ _ _ _. ~

the the STA or another SRO with an engine -

-~ g degree _- concurs.

m a y

~_ ~~- __

may conduct the evolution without restrictions.

must withhold conducting the test until a IPTE package has been approved.

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~~

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must NOT allow the test under any conditions.

I Hope Creek 12003 5

is 1

\\Memory J

01 G207 112 2.2 Equipment Control bed in the safety analysis 2.0 report.

~~-

Justification:

SRO 10CFR55.43 (3) Facility license procedures required to obtain authority for the design and operating changes in the facility.

must withhold 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 Manaaer amroval.

must NOT allow the test under any conditions. Incorrect. May be performed if IPTE package approved. __ -

Paae 145 of 162 Monday, June 23,2003 7:24:17 AM 1

. __~--___

___--_1

Given the following conditions:

- The plant experienced a Failure to Scram from a turbine trip at 98% power.

- The operator successfully initiated a manual scram approximately 4 seconds

- Post trip analysis revealed the following data for the event:

trip.

- Peak Reactor Power:

116% Thermal

- Peak Reactor Pressure: 1205 psig

- Minimum Vessel Level: -151 inches (Fuel Zone A and B)

- Most limiting MCPR:

1.09 Which one of the following states which Safety Limit was violated?

~

- ~_

following the turbine Thermal Power Low Pressure - Low Flow.

2.2 Equipment Control L

Thermal Power High Pressure - High Flow Correct. Minimum MCPR is 1. I O Thermal Power Low Pressure - Low Flow Incorrect. Initial power pre transient was above 25 percent with pressure above 785psig.

RPV Level Safety Limit. Incorrect. Level maintained above -161 in terms of conditions Mondav. June 23. 2003 7:24.17 AM

-1 Paae 146 of 162

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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 involv fuel movement NOT involving core alterations?

Fuel Handling Operator Radiation Protection Technician Reactor Engineer, acting as spotter Fuel Handling Operator Refueling Bridge Operator as spotter Radiation Protection Technician 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 L

-~

Reactor

-~ _

~_ Engineer

~

_~

~

~~

~

.7, and HC.RE-AP.ZZ-0049 sections 5.3.

inimum crew for non-core alteration fuel handling activities in the

, Reactor Engineer irnum permissible None Monday, June 23,2003 7:24:18 AM I

- /

Page 147 of 162

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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.

Which one of the following personnel must be notified prior to the beginning the detensioning process IAW HC.OP-IO.ZZ-0005 Cold Shutdown to Refueling?

i-

~

- _ _ ~

Refueling Floor SRO..

Reactor Engineer.

Control Room Supervisor.

_ _ _ _ _ _ _ _ _ _ _ ~ _ _

_ _ _ _ _ ~

-~ -

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~

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___ ~. __ _

Refueling Outage Manager.

Hope Creek c

IS 1

[Memory J

-2 2.2 I Equipment Control dge of the refueling process.

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~

_ _ ~ _ _ _ _ _

SRO 55.43 (2) Facility operating limitations in the Technical specifications and their bases.

Correct. Control Room Supervisor. Required signoff for HC.OP-IO.ZZ-0005. Changes Operational Condition to OC 5.

Incorrect. Reactor Engineer. Not required.

Incorrect. Refueling Floor SRO. Not required.

L

_ _ _ - ~

I

~_

I

-. ~-

Monday, June 23, 2003 7:24:18 AM Paae 148 of 162

- ~ _ _ ~-

- ~-

Given the following conditions:

- The plant is in Operational Condition 5.

- There are 16 fuel bundles remaining in the Reactor vessel.

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.

Removing an IRM detector from undervessel.

Transfering a control rod blade guide from the Spent Fuel Pool ____-._

to the Reactor vessel.

_ _ ~ _ _ _ _ _ - -

2.2 Equipment Control

-~

~

2.2.27

/Knowledge of the refueling process.

3.51 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.

IHCGS Tech SDecs 1.0 Definitions Define or discuss the terms contained in Section 1.O of Hope Creek Generating Station Technical Spe

~

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____ ~

_~

TECSP

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Monday, June 23, 2003 7:24:18 AM Paae 149 of 162

Given the following conditions:

- An initial startup is in progress.

- Threshold power is I 1 KW/ft.

- A 20 MWe/hr ramp is established during the last rod adjustment.

- The 20 MWelhr ramp continues until nodal power reaches 12.5 KW/Ft.

- Then PCIOMR preconditioning is begun by ramping power at 10 MWe/hr.

c Which of the following is the result of these actions?

_ _ _ - _ ~. _ _ _

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~~

_ _ _ _ _ _ - - ~

~_________

Fuel cladding stress is maintained within the vendor specifications.

~

~-

____________~

The helium gas volume between the fuel pellets and fuel rod cladding will be larger than expected.

~

_ _ _ ~ _ _ _ _ _ _

~

Fuel pellet densification will cause high fuel temperature.

a 1 1

/Comprehension Hope Creek 06/1 712003 I

2.5 2.9

-~

rate of.I 1 KW/Ft/Hr or approximately 10 Mwe/hr 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 NOH01 RXFUEL RXFUELEOl3 '

Explain the following:

b.

c.

a.

The acronym and purpose of PCIOMR.

The definition of threshold power.

The definition of envelope power.

1 Monday, June 23,2003 7:24:19 AM 1

Page 150 of 162

Per NC.NA-AP.ZZ-0024, Radiation Protection Program, a 21 year old worker with 1 I T e m Lifetime dose from the previous 3 years working at Hope Creek will have an administrative exposure control level of (1) mrem TEDE by the Radiation Protection Manager.

(Assume NO delegation of authority) mrem TEDE per year. This can be raised to a maximum of (2)

El (1)2000-(2) 3000 (1) 2000 G9m=&-rn5CuKp_ cLi!mjd

?&[/ 0-

~

(2) 4000 pk (1) 3000 (2) 4500 0611 7/2003

~

2.3 Radiological Controls

~

p osure limits and contamination control, including permissible levels in 2.5 3.11 (I) 2000 (2) 3000 Incorrect. RP Supervisor approves extension to 3000 (1) 2000 (2) 4000 Correct per NC.NA-AP.ZZ-0024 Attachment 1. Worker does not meet 5(N-17) threshold. 5(21-17) =20 Rem.

(1) 3000 (2) 4500 Incorrect. Automatic authorization during Emergency Plan implementation.

~ _ _ _ _ _ _ _ _ _ _

1 NC.NA-AP.ZZ-0024 Attachment 1

~-.

a.

Yearly Dose Extension

b.

Declared Pregnant Women Dose Extension Monday, June 23,2003 7r2479AM Page 151 of 162

Given the following conditions:

- The plant is operating at 100 percent power.

- 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 1.5 R/hr.

Which one of the following is required to allow the maintenance work to be authorized in accordance with Technical Specifications?

One of the maintenance personnel is self monitor qualified.

hazards

- involved.

_~

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_ ~ _ _ _ _ _

All personnel involved in performing the work are volunteers and have been fully briefed on the

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All work is documented in the Control Room OS/CRS Narrative log with the total dose

-_ -- received.

_~

ation Protection coverage.

06/17/2003 I

I 294001G310 119 2.3.10

[Ability to perform procedures to reduce excessive levels of radiation and guard against personnel 2.9 3.31 jexposure.

JUSTIFICATION:

SRO 55.43(4) Radiation hazards that may arise during normal and abnormal situations.

Correct: The job is provided with continuous Radiation Protection coverage. HCGS TS 6.12.1.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 OS/CRS 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 coveraae.

L Y

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Monday, June 23, 2003 7:24:19 AM

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Page 152 of 162

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The moisture content of charcoal adsorber bed of the Gaseous Radwaste System (GRW) is rising.

Which of the following parameter changes will occur and what actions would mitigate the effects?

Rising GRW post-treatment radiation levels due to an increase in Krypton. Lower Cooler k i n g GRW post-treatment radiation levels due to an increase in Iodine. Raise Cooler Condenser

_ temperature.

Lowering GRW charcoal adsorber bed temperature. Lower offgas Dilution flow.

Lowering GRW charcoal adsorber bed hydrogen concentration. Raise offgas Dilution

~~- flow.

Condenser temperature.

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06/17/2003 294001G31 1 120

_ _ _ _ _ _ ~ ~ ~ _ _ _

2.3 1 Radiological Controls 2.3.11 I 2.7 13.2

-1

[Ability to control radiation releases.

J USTl Fl CAT1 ON Rising GRW post-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.

GASRWOEO08 1 (R) Explainlidentifv the effect of moisture in the process gas stream on the following components IAW available control room 1 rifirences:

a.

Recombiner I

I

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b.

Charcoal Beds L

Monday, June 23,2003 7:24:20 AM

~-

Page 153 of 162

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Plant conditions are asfollows:

- Reactor Power is at 70%.

- Condenser Vacuum is 5.5" Hg absolute and degrading.

L-Which one of the following states immediate operator actions required?

Ensure turbine sealing steam pressure is normal.

~~

Trip the Main Turbine when 350 Mwe is reached.

Place the standby SJAE in-service.

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0611 712003 294001 G404 121

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2.4 Emergency Procedures and Plan 2.4.4 IAbility to recognize abnormal indications for system operati rs which are entry-level 4.0 14.31

~~

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conditions for emergency and abnormal operating procedures.

Reference:

HC.OP-AB.BOP-0006 Main Condenser Vacuum Immediate Action 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 350 Mwe Place the standby SJAE in-sewice.-Incorrect-Subsequent action B 300 Mwe vice 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 ABBOP6E003 (R) From mernorv. recall the Immediate Operator Actions for Main Condenser Vacuum Monday, June 23, 2003 7:24:20 AM Page 154 of 162

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Monday, June 23, 2003 7:24.20 AM -

1 Paae 155 of 162

Select the definition of the term, Minimum Alternate RPV Flooding Pressure.

The lowest differential pressure between the RPV and the suppression chamber at which I

steam flow through the minimum number of SRVs required for Emergency Depressurization is

~~~~

sufficient to remove all decay heat from the core.

The 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.

W 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.

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.

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I 0611 712003 IB 1

\\Memory 1

Hope Creek 294001 G417 122

__i 2.4 Emergency Procedures and Plan 3.1 j 3.81 2.4.17

[Knowledge of EOP terms and definitions.

-2

~

J USTl FICATION:

CORRECT - The 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-EO.ZZ-LI MITS-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 Depressurization is 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.

IHC.OP-EO.ZZ-O~O~A EOP206E004 Define the term Minimum Alternate RPV Floodinq Pressure.

Page 156 of 162 Monday, June 23, 2003 7:24:21 AM

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Direct From Source VISION Bank QID# Q56150

\\u Mondav. June 23,2003 7:24:21 AM Page 157 of 162

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.

L, i,

~p Which one of the following describes the reporting requirement, if any, via ENS line?

~

~-

-~

~

p

~

No report.

~

_~

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p-p

~

p One Hour report.

Four Hour report.

A Eight Hour report.

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1 Hope Creek c 1

/Application I

01 G430 123 2.4 Emergency Procedures and Plan 2.4.30 Knowledge of which events related to system operationslstatus should be reported 3.61 SRO 55.43 (I) Conditions and limitations in the facility license.

1 OCFR55.43 (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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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.

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Monday, June 23, 2003 7:24:21 AM

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Page 158 of 162

Given the following :

- A severe accident has occurred.

- 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 i~

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 -

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.

~_

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~

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~

-.. _ _ _ _ _ ~

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Evacuate all sectors 0 - 5 miles.

1 I

Hope Creek 0611 712003 a J

[Application I

294001 G444 124

_ _ _ _ ~

2.1 4.01

-~

(I)

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 +I-I sector 5 - 10 miles. Shelter all remaining sectors 5 - 10 miles. Correct answer. Loss of all barriers = 10 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 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 only 9 point GE ECG Attachment 4 Appendix 1 ntly Modified Monday, June 23,2003 7:24:21 AM Page 159 of 162

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Following a reactor scram and loss of feedwater, the plant is being cooled down using HPCl in full flow recirculation. A review of the operating logs indicates that reactor pressure for the past two hours is as follows:

- Time 0000 001 5 0030 0045 0100 0115 01 30 01 45 0200 Reactor Pressure (psig) 950 925 900 850 700 650 550 300 250 icontrol room reference material.

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.

I K T E D 6OWER TO COLD SHUTDOWN Integrated Operating Procedure, supporting System Operating Procedures and I

HC.OP-IO.ZZ-0004 Attachment 4. Steam Table form HC.OP-I0.Z-0008 b

Monday, June 23,2003 7:24:22 AM

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Page 160 of 162

INPO Exam Bank Significantly Modified Monday, June 23, 2003 7.24:22 AM

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Page 161 of 162

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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.

- Wide 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?

NO action is required because the level would have been within limits at the time of removal.

NO action is required because RClC Suction swap would occur on an actual low level.

Makeur, to Sumression Pool level is required because the average level is below the limit.

_ _ _ ~ _ _ _ _ _ _

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Makeur, to Sutmression Pool level is rewired because level is outside allowable limits.

Hope Creek 06/17/2003

/Memory I

294001 G448 126

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24 Emergency Procedures and Plan 2.4.48 [Ability to interpret control room indications to verify the status and operation of system, and junderstand how operator action s and directives affect plant and system conditions.

3.5 3.8 Justification:

iMakeup to Suppression Pool level is required because level is outside allowable limits. Correct. IAW NCNA-AP.ZZ-0005 states "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 RCIC 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.

Levels are not averaaed to obtain actual level.

I ic-Monday, June 23,2003 7:24:22 AM Page 162 of 162

ML032320112

Text

Reference:

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