Text

Final - Lsro Written
	ML031130318
Person / Time
Site:	Hope Creek
Issue date:	04/01/2003
From:	Reid J; Public Service Enterprise Group
To:	Conte R; NRC/RGN-I/DRS/OSB
	Conte R
References
	50-354/03-301 50-354/03-301
	Download: ML031130318 (63)
	v • d • e

Given the following conditions:

- The plant is in Operational Condition 5.

- 10A401 A Channel 4.16 KV bus is de-energized for maintenance.

- Core Alterations are in progress.

- The infeed breaker for 1OY412 120/208 Volt AC Distribution panel trips open.

,(Use attached table Q1 for load listing)

Which one of the following correctly describes the effect on Core Alterations and reason?

Core Alterations ...................... .. ........

.=LL ai. may continue provided the panel is re-energized within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

MIKmay continueebecause the panel feeds Non-1 E loads. __ __

Eil must be suspended_ because al SRM drives lose power.

EI1 must be suspended because the panel is required.

d ,L B ntlIA Aplwca Coiin Hope creek 0 11 Ni _ 03/10/2003 0.. .... Emergency and Abnormal Plant Evolutions ..

2 Soll 6 1 295003A101 295003 Partial or Complete Loss of A.C. ..

Power IA1.7 Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C.

POWER: __ ~~~~~~~. .........

=. .........

AA1.01 !A.C. electrical distribution system 3.7 L3.8 8 ~~~~~~~~~~..............I.... ...................... ..........- .................... ..........

Ex I tio h.D- correct; B Channel Panel 10Y412 is required by 3.8.3.2 because Channel A 10A401 is de-energized.

' m._A-incorrect; electrical spec in OP Cond 1, 2, & 3 B- incorrect; Panel contains 1E loads

C- incorrect; 10Y208 powers SRM Drive Cabinet. SRMs not listed in 10Y412 Load list.

38.3.2 1T 1 EACOOE028 1 Given a scenario of applicable conditions and access to Technical Specifications:

a. Choose those sections, which are applicable to the 1E AC Power Distribution System IAW Technical Specifications.

b. Evaluate 1E AC power operability and de Load Reference table for 10Y412; Tech Specs section 3.8 WNW New I MM= X-IO -_

Tuesd-ay,@March04Q2003 11o334MPge f Tuesday, March 04, 2003 11:30:34 AM Page 1 of 58

2 Given the following condidtions:

- The plant is in Operational Condition 4.

- "B" Fuel Pool Cooling is in operation with the FP Demins bypassed.

Then the following occurs:

- All offsite power is lost.

- All other equipment functions properly.

Which one of the following describes how and when cooling to the Fuel Pool is re-established?

E1 Automatica-y-;within 2 minutes. __ _

Ei Automatically; greater than 2 minutes.

I Manually; within 2 minutes.

II Manually; greater than 2 minutes. _____

l id erd IB 3 Memory ime Eac

] i0l Hope Creek go_

E 02/24/2003 Emergency and Abnormal Plant Evolutions 0 l 2 SRO Gt 01 295003G411 295003 ;Partial or Complete Loss of A.C. Power 2.4 Emergency Procedures and Plan 2.4.11 Knowledge of abnormal condition procedures. 3.4 3.6 l140a0t0C Justification:

,,, ^Wvl Correct: Manually; greater than 2 minutes. The FPCC pumps are not sequenced back following a LOP.

The Control Room Operator must manually start the pump after automatic load sequencing is completed 95 seconds after output breaker closure or 108 seconds after LOP. Restart is accomplished.using HC.OP-SO.EC-0001. The FP Filter demin Bypass valve must be reclosed for pump start. It is also the third piece of equipment in order allowed to be restarted as load permits in accordance with HCOOP-AB.ZZ-01 35.

Incorrect: Automatically; within 2 minutes. The FPCC pumps are not automatically sequenced back following a LOP. Plausible misconception.

Incorrect: Automatically; greater than 2 minutes. The FPCC pumps are not automatically sequenced back following a LOP. Plausible misconception.

Incorrect: Manually; within 2 minutes. The FPCC pumps must be manually restarted following a LOP.

The pump is no longer inhibited from starting after 95 seconds. As a minimum, the FP Demin bypass Svalve must be manually closed for pump restart. It is also the third piece of equipment in order allowed to

be restarted as load permits in accordance with HC.OP-AB.ZZ-0135.

HC.OP-AB.ZZ-01 35 Step 4.7.9

.. 0 .. 1 i. =K 1 . . =...........................

EMU LL OAB135E04 Explain the reasons for how plant/system parameters respond when implementing, Station Blackout/Loss Of Offsite Power I Diesel Generator Malfunction, Abnormal Operating Procedure.

OAB1 35E006 (R) Explain the bases for Subsequent Actions and the information contained in the Discussion Section of Station Blackout/Loss Of Offsite Power Diesel Generator Malfunction, Abnormal Operating Procedure.

rtdI ~q~e

None Tuesday, March 04, 2003 11:30:35 AM P 2 of 58 Page

New QuE DtixMfimcIa).Mh7 Tuesday, March 04, 2003 11:30:35 AM Page 3 of 58

sr, 3 Given the following conditions:

- Core reload is in progress at fuel movement step 1150.

- Ten (10) new fuel bundles remain to be loaded into the core into the "B" quadrant.

- SRM readings are as follows:

SRMA SRM B SRM C SRM D Step 1150 75 100 75 75

- After loading two of the ten new fuel bundles, the SRMs read as follows:

SRMA SRM B SRM C SRM D Step 1152 90 200 90 90 WHICH ONE (1) of the following states the expected results during loading of the remaining bundles?

(Assume all 10 bundles have equal reactivity worth.)

I31inserting one more bundle will cause a local criticality.=

EIl Inserting two more bundles will cause a local criticality.

g SRM "B" will indicate 500 cps when the core is fully loaded.

i SRM "B" will indicate 1000 cps when the core is fully loaded.

Ase b . l I: IComprehension Eacfflty: Hope Creek 02/24/2003

Emergency and Abnormal Plant Evolutions RGrUP 1 RG i 1i 295014A201 295014 Inadvertent Reactivity Addition AA2. Ability to determine and/or interpret the following as they apply to INADVERTENT REACTIVITY ADDITION:

AA2.01 Reactor power 4.1 4.2 fla ll Justification:

{Angsti~hI.Correct: 2 more will cause a local criticality. The first doubling was reached with 2 bundles. The same amount of reactivity will be added with the next 2 bundles at which time the reactor would be critical at ileast locally.

Incorrect: One more bundle will cause a local criticality. One more bundle would be half the amount of reactivity needed to be critical at that local area.

Incorrect: SRM "B" will indicate 500 cps when the core is fully loaded. Value if 100 counts added per 2 bundles.

Incorrect: SRM "B" will indicate 1000 cps when the core is fully loaded. Value if 100 counts added per bundle.

Reactor Theory RXOPERE009 Describe how to determine if a reactor is critical.

RXOPERE005 Explain the characteristics to be observed when the reactor is very close to criticality.

Tuesday, March 04, 2003 11:30:35 AM Page a 4 of 58

None n _ M M Significantly Modified

$~ure~Com~e~s:

Quest IINPO Bank QID 14536 09/1 3/1 996 Peach Bottom Tuesday, March 04, 2003 11:30:35 AM P Page 5 of 58

~st~a umbW5 4 Given the following conditions:

- Control rod friction testing is in progress.

- Shutdown Margin is determined to be 0.25% delta k/k analytically.

- The One-Rod-Out interlock is Operable.

(Assume all SRMs are operable)

Which one of the following is the minimum required to automatically mitigate an inadvertant criticality?

Eid Non-Coincident UPSCALE scram from at least 2 SRM channels.

11 Non-CoincidentU PSCALE scram from only 1 SRM channel.

EdI Coincident UPSCALE scram from at least 2 SRM channels.

El Coincident UPSCALE scram from only 1 SRM channel.

Aw b EI [B ::::::LvIeComprehension F Hope Creek E pate: 02/24/2003

[Emergency and Abnormal Plant Evolutions lQI*

R 1i l ol 1 29501 4K205 295014 Inadvertent ._.....

Reactivity Addition AK2._ Knowledge of the interrelations between INADVERTENT REACTIVITY ADDITION and the following:

AK2.05 Neutron monitoring system 4.0 4.1

[go wQ.RJustification- Correct- Non-Coincident Scram from only 1 SRM channel. RPS Shorting links must be Emo I removed for a Non-Coincident scram from any 1 Ni Channel to withdraw a control rod with SDM less than analytical limit of .38% delta k/k.

Incorrect- Non-Coincident Scram from at least 2 SRM channels. Only 1 SRM required with shorting links removed.

Incorrect- Coincident Scram from at least 2 SRM channels. Without SDM greater than the limit, the shorting links must be removed.

Incorrect- Coincident Scram from only 1 SRM channel. With Shorting Links installed, need at least 2 channels Tech Specs 3.9.2 /4.9.2.d Tech Specs Table 3.3.1-1 Footnote ( C) lTech Spec Bases 3/4/3.1 O=.4-i1 iiii, Objectives ., 7-

SRMSYSE007 (R) Given a scenario of applicable operating conditions and access to technical specifications.

a. Choose those sections which are applicable to the SRM system.

b. Evaluate SRM operability and determine required actions based upon system operability.

Mtial Requie R lTech Spec section 3.3 with Table 3.3.1-1 removed Iti;S ii0e: New IQuesti&odewne Tuesday, March 04, 2003 11:30:35 AM Page P 6 of 58

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5 Given the following conditions:

- The plant is in OPCON 4 having just completed Refueling Operations.

- The Refueling Cavity is being decontaminated.

- Several bundles are being shuffled in the Fuel Pool in preparation for sipping operations.

- The Control Room reports the FUEL POOL COOLING SYS LEAKAGE Hi alarm has been received.

Assuming the alarm was caused by gate leakage due to low seal pressure, which of the following pressure indicators would read low?

1. KA-PI-4610A

11. KA-PI-4610B Ill. KA-PI-4610C IV. KA-PI- 4610D pI lIA and 1I.__ _ _ _ _ _ _ ___ _

I1and Ill.

Ml 1Iand l l.

Ill and IV.

xm I B Coniiv 'aApplication Igt Hope Creek xmDte: 02/2. 4/2003 Emergency and Abnormal Plant Evolutions ] G31 3 lI 295023A21J4 295023 Refueling Accidents AA2. Ability to determine and/or interpret the following as they apply to REFUELING ACCIDENTS:

AA2.04 Occurrence of fuel handling accident 3.4 4.1

'glan;a~idbf] Justification:

~ I and 11Correct - Leakage would be from the Inner fuel pool gate inner and outer seals since the Fuel Pool is full and the Refueling Cavity is drained. I and 11pressure indicators monitor the inner seals.

KA-PI- 4610C and KA-PI- 4610D are for the outer Fuel Pool Gates.

Not direct lookup because trainee must use P&ID to determine answer.

Incorrect: I and Ill Incorrect: II and Ill Incorrect: Ill and IV P&ID M-53 sheet 2 HC.OP-AR.ZZ-001 3 Attachment B-5

_Le rning Obj t MAiveP FPCCOOE004 (R) From memory, describe/explain how leakage is detected from the spent fuel storage pool, dryer-separator storage pit, I reactor well and fuel shipping cask pit liners, iAW the Fuel Pool Cooling and Cleanup System (FPCCS) Lesson Plan.

1M"W"iIjO A IPlEca

&IDM-53 sheet 2; Facility Exam Bank o Mn M lt ot Direct From Source 2002 LSRO Requal exam.

Tuesday, March 04, 2003 11:30:36 AM Page 7 of 580

Tuesday, March 04, 2003 11:30:36 AM Page 8 of 58 ber 6 The unit is in OP CON 5 with the following plant conditions:

- RHR Loop "B" is operating in Shutdown Cooling.

- Both Fuel Pool Cooling Pumps and Heat Exchangers are in service.

- The Refueling Cavity is flooded and the Fuel Pool gates are removed.

- Preparations for Core Alterations are in progress.

- A control circuit malfunction causes a vessel draining event.

- Operator actions have slowed the lowering level.

- Level is currently lowering 1 foot every 11 minutes.

- Due to high radiation conditions, the refueling floor is NOT accessible.

Based on these conditions, which of the following makeup sources is available to the Fuel Pool/Refueling Cavity that requires operator component manipulation from outside the Control Room?

(Exclude hand operation of MOVs) l Fire Water.

t~l Service Water.

il [DemineraiIzed Water.o Ail Condensate Storage & Transefer.

[gig d xanve B Le Memory IaltHope Ht Creek Ex Da 02/24/2003

-.-. Emergency and Abnormal Plant Evolutions ROGip 3 [ZIGr l 1 295023G435 295023 Refueling Accidents 2.4 Emergency Procedures and Plan 2.4.35 Knowledge of local auxiliary operator tasks during emergency operations including system geography 3.3 3.5 and system implications.

[;Inai,~ Justification:

CORRECT - Condensate Storage & Transfer - Requires manual valve manipulation only to initiate makeup water flow to the cavity/pool through RHR SDC.

INCORRECT - Fire Water - Requires opening of HV-4648 from the Control Room. Fire hose on 201 disallowed by question stem.

INCORRECT - Demineralized Water - can only be aligned with hoses on 201' RB.

INCORRECT - Service Water. Service Water requires opening valves from the Control Room.

Ref M-10-1 and M-53-1

L

li'm$w'^-V ____ i s. Leam.ir'4iO.'C'..t__....

FP6COOE008 (R) Concerning spent fuel storage pool water level, summarize, from memory, the following lAW the Fuel Pool Cooling and Cleanup (FPCCS) System Lesson Plan:

a. How normal level is controlled

b. Sources of makeup to the spent fuel storage poo IMateia uir ~at . None

'- , Facility Exam Bank u M Meth6 l < Editorially Modified Tuesday, March 04, 2003 11:30:36 AM Page 9 of 58

Q s W Mo 2 Vision Bank QID# Q53948 Modified to disallow hoses on 201 el.

Tuesday, March 04, 2003 11:30:36 AM Page 10 of 58

T e 7 Given the following conditions:

- The reactor is shutdown.

- RHR Loop "B" is in Shutdown Cooling (SDC).

- RHR Heat Exchanger Bypass valve, BC-HV-F048B, is closed.

During shift turnover, Reactor Recirculation was found in the following condition:

- 'B' Reactor Recirculation Pump suction valve is 100% open.

- 'B' Reactor Recirculation Pump discharge valve is 10% open.

Based on these conditions, which of the following describes the actions necessary to maximize SDC heat removal?

The 'B' Recirculation Pump discharge valve must be fully opened because the RHR pump will be running at shutoff head.

-i1 il openedbecause RHR pump min flow will be open.

i' closed because the RHR flow will be at pump runout.

closed because the RHR flow will be bypassing the core.

Anw d a l B t Comprehension Fit Hope Creek Exa 02/24/2003

[ Emergency and Abnormal Plant Evolutions i J~9Group 2 SRO 2 295001A101 295001 Partial or Complete Loss of Forced Core Flow Circulation AA1. Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION:

AA1.01 Recirculation system 3.5 3.6 l:Exon~ l Justification:

-i0-lCorrect answer: closed because the RHR flow will be bypassing the core. The 'B' Recirculation Pump

ej discharge valve must be fully shut. With the discharge and suction valves open, some SDC flow will bypass the core reducing heat removal from the core. Closing the B discharge valve will establish full SDC flow through the core.

Incorrect: opened because the RHR pump will be running at shutoff head. - Wrong direction, wrong reason. RHR pump will have normal SDC flowpath to and from the B Recirc loop.

Incorrect: opened because RHR pump min flow will be open. Wrong direction, wrong reason. Flow through RHR loop flow element will be adequate to close min flow valve.

Incorrect: closed because the RHR flow will be at pump runout. Correct direction but wrong reason. The valve must be closed to stop core bypass flow. Pump runout is not why the Recirc discharge valve is closed. F015B is throttled to prevent runout.

HC.OP-SO.BC-0002 Limitations 3.2.5 and 3.2.11 RHRSYSE009 (R) Given plant problems/industry events associated with the Residual Heat Removal System:

a. Discuss the root cause of the plant problem/industry event lAW the associated plant problems/industry event document.

b. Discuss the HCGS design a Tuesday, March 04, 2003 11:30:36 AM Page 11 of 58

~ait Rqu None IQ40W*00"ml New Qetion Modification MethI:

Tuesday, March 04, 2003 11:30:36 AMP Page 12 of 58

r Ques' N- - lbe'r 88 Given the following conditions:

- The plant is in Operational Condition 5.

- The reactor core has been completely off-loaded to the Fuel Pool.

- Fuel Pool Cooling Heat Exchanger is supplied by B SACS Loop.

- Prior to reload, the Control Room reports Fuel Pool temperature is increasing.

Which one of the following malfunctions would cause the rise in Fuel Pool temperature?

-SACS Loop B HX Bypass Isolation valve EG-HV-2457B air supply line breaks off.

[A SSW Loop B Yard Dump valve EA-HV-2356B spuriously opens.

EilS SSW Loop B to RACS valve EA-HV-2204 closes.

Ei SACS Loop B Temperature Control valve EG-TCV-2517B fails full open.

Anwe d RW ev B i l gpiv e Comprehension Hope Creek Ea Date: 02/24/2003 e [Emergency and Abnormal Plant Evolutions ROlPeI 2 SRO Gr i 2 295Ci 18A201 295018 Partial or Complete Loss of Component Cooling Water AA2. Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER:

.AA29 ni Clmnrennt tfimnerftiurn 3.3 3.4 r1aiamUi fl Justification.

a .Correct: SACS Loop B Temperature Control valve EG-TCV-2517B fails full open. Full open bypasses the SACS HX. SACS Loop B temp will rise as well as FP temp.

Incorrect: SSW Loop B Yard Dump valve EA-HV-2356B spuriously opens. Would increase SSW flow through the SSW side of the SACS HX. Worst case would make no change if no water came out.

Incorrect: SSW Loop B to RACS valve EA-HV-2204 closes. FP is cooled by SACS. Loop B SSW flow would increase though SACS HX, lowering SACS temp or worst case, no change.

Incorrect: SACS Loop B HX Bypass Isolation valve EG-HV-2457B air supply line breaks off. Valve fails closed on loss of air, forcing more flow through SACS HX and lowering SACS temps.

i=.R..... Title I '.i P&ID M-1 1 sheet 1 and M-10 sheet 2 A,___

mm 4Learning ObU ~~& ME M~> .

FPCCOOE015_ (R) Given any of the following systems, from memory, summarize the interrelations between the FPCCS and that system, IAW the Fuel Pool Cooling and Cleanup System (FPCCS) Lesson Plan:

a. Instrument Air System

b. Area Radiation Monitoring Sys STACSOE006 Summarize/identify how the STAcS system temperature is automatically controlled. IAW available control room references P&ID M-11 sheet 1 and M-10 sheet 2 IQuestion Source,. New UN59W0i IfiatonM'ho,'

E-=

__ - - -S C M;=L Tuesday, March 04, 2003 11:30:36 AM Page P 13 of 58

j~zsto\Mbe 9 Given the following conditions:

- The plant is in Operational Condition 4

- Control Room CRIDS computer is out of service when the following alarms actuate in the Control Room:

- RACS REMOTE CONTROL PNL 10C202 (A2-F2)

- RACS RMS is in ALARM on RM-1 1

- An operator has been dispatched to determine RACS Head Tank level.

Based on these conditions:

El RACS is leaking into RWCU.

il RWCU is leaking into RACS.

HI RACS is leakinginto Service Water.

-XService Water is leaking into RACS.

Fi5W b EFaL e eB !Rlie Comprehension Hopecreek Wbill EciRE O2/2 '4/2003 WiEr: Emergency and Abnormal Plant Evolutions G 2 l p 2 295018K1 01 295018 -Partial or Complete Loss of Component Cooling Water AK1. Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER:

AK1.01 Effects on component/system operations 3. 5 3.6

-- atit Justification:

Jf, Answer Correct: RWCU is leaking into RACS. RWCU is source of activity causing RACS Rad Monitor alarm.

Inward leakage into RACS makes Head tank level rise.

Incorrect: Service Water is leaking into RACS. Would not cause RACS Rad Monitor alarm.

Incorrect: RACS is leaking into Service Water. Head tank level would lower.

Incorrect: RACS is leaking into RWCU.. Would not cause RACS Rad Monitor alarm. Head tank level would lower.

er Title _ __ .C HC OP-AR ZZ-0O1 1 Attachment C1 -8 RACSOOE01 3 (R) Given M-1 3-0 and M-1 3-1 assess the interrelationship between RACS and any of the following for a given set of plant conditions:

a. Control Rod Drive

b. Liquid Radwaste Collection System

c. Liquid Radwaste System

d. Solid R MatGialReui fM 13-1 lQbn'SArPI iNPO Exam Bank WM I thitidn l Editorially Modified lQu wu Cm m INPO Bank QID# 7975 03/14/1997 Hatch. Modified for Hope Creek Tuesday, March 04, 2003 11:30:36 AM Page 14 of 58

10 Given the following conditions:

- The plant is in Operational Condition 4 following a forced shutdown 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months ago.

- 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.

- HC.OP-AB.RPV-0009, Shutdown Cooling, is entered.

Which of the following will be adequate to maintain Operational Condition 4?

a. Crosstie "C" or"D" RHR pump for heat removal.

I Maximize RWCU bottom head drain flow.

HJ Raise level to +80 inches using natural circulation for heat removal.

21 Inject with Core Spray from the CST to the RPV.

Aw a E11 B :1iiJeI Application FacWWl Hope Creek Exam Oatr: 02/24/2003 Emergency and Abnormal Plant Evolutions IO.Group 3 2 295021A104 295021 Loss of Shutdown Cooling AA1. Ability to operate and/or monitor the following as they apply to LOSS OF SHUTDOWN COOLING:

AA1.04 Alternate heat removal methods 3.7 3.7

[fi-inof Justification Crosstie "C"or "" RHR pump for heat removal. correct - RPV-0009 subsequent action E, RHR Pumps C & D may be realigned to provide alternative decay heat removal.

Maximize RWCU bottom head drain flow. -incorrect- Maximizing bottom head drain flow does not provide heat removal adequate to maintain less than 200 degrees.

Raise level to +80 inches using natural circulation for heat removal. -Incorrect- per subsequent action E Natural Circulation does not provide heat removal, only circulation.

Inject with Core Spray from the CST's to the RPV. -incorrect- This is not an approved method of Alternate DHR.

H-Ce, BRP70 009 te Titli f HC.OP-AB. RPV-0009 M 'it , L ml g O etie , .ji:

ABRPV9EO07 (R) Explain the bases for Subsequent Actions and the information contained in the Discussion Bases Section of Shutdown Cooling.

Matera ~ 7L None wlf~l2}l Facility Exam Bank jqtestin MkM eh ~]Significantly Modified festo lorVISION Bank QD# 0Q61332 Sig Mod Tuesday, March 04, 2003 11:30:37 AM Pag e 15 of 58

Given the following conditions:

- The plant is in Operational Condition 5.

- Control rod friction testing is in progress.

- 'B' CRD Pump is C/T for maintenance.

- Control rod 30-31 is at notch 04.

- 'A' CRD pump trips and cannot be restarted.

Which one of the following is the most limiting consequence of the pump trip?

aft Rod 30-31 must be electrically disarmed.

Rod 30-31 must be hydraulically disarmed.

IH Rod 30-31 must be scrammed.

IEl Rod 30-3 1 cell must be unloaded.

lAnwrk c l ExmLeve B Cgnt.LQ.el Comprehension Facilitfi: Hope Creek xmate: 02/24/2003 Tie Emergency and Abnormal Plant Evolutions Rr 2 $R9upd 2 295022K102 295022 Loss of CRD Pumps AK1. Knowledge of the operational implications of the following concepts as they apply to LOSS OF CRD PUMPS:

AK1.02 [Reactivity control 3.6 3.7 EX I lW,-4 Justification:

4. Correct: Scrammed. Any control rod withdrawn unless underTS 3.9.10.1 or 3.9.10.2 must have an operable scram accumulator. 3.1.4.5 In Op Con 5* With one or more control rods inoperable, upon discovery immediately initiate action to fully insert inoperable withdrawn control rods. The CRD Accumulator trouble alarm will eventually alarm requiring a reactor scram.

Incorrect- Rod 30-31 cell must be unloaded. Not required. 30-31 must be inserted immediately.

Incorrect- Rod 30-31 must be electrically disarmed. Action for Op Con 1 and 2.

Incorrect- Rod 30-31 must be hydraulically disarmed. Action for Op Con 1 and 2.

Tech Spec 3.1.3.5 I .. erig.jetvs ..

CRDHYDE033 (R) Given a scenario of applicable operating conditions and access to Technical Specifications complete each of the following lAW Technical Specifications:

Select those sections applicable to the CRDH System.

Evaluate CRDH System operability and determ ABIC01 E007 (R) Explain the bases for Subsequent Actions and the information contained in the Discussion Bases Section of Control Rod.

Mtr ea ,' Tech Spec section 3.1

[QetIojAurcMI iNew u Md i e Quesiemmur ~95nmSgl en: 0000X Tuesday, March 04, 2003 11:30:37 AM Page 16 of 58

1-W11111 IW .

Given the following:

ffn"t. 12

- LPRM changouts are being perfomed within the reactor vessel cavity.

- One of the old fission chambers is accidently lifted 1 inch clear of the water.

Which one of the choices correctly completes the following statement regarding the Refueling Floor Evacuation Alarm in the reactor building?

The radiation monitor activates the Evacuation Alarm because its detector(s) is(are) located in 1Il New Fuel Vault; line-of-sight to the refueling cavity.

ISI1 Spent Fuel Pool; line-of-sight to the refueling cavity.

LI Refuel Floor Exhaust; the ducts above the refueling cavity.

El Reactor BuildingExhaust; the ducts above the refueling cavity.

Answer b E*im IB 1 Comprehension l aci Hope creek 02/24/ 2003 Tieri, Emergency and Abnormal Plant Evolutions RQrup 2 Group 2 295033A10 1 295033 High Secondary Containment Area Radiation Levels EAi._ Ability to operate and/or monitor the following as they apply to HIGH SECONDARY CONTAINMENT AREA RADIATION LEVELS:

EA1 .01 Area radiation monitoring system 3.9 4.0

[glantonf 1 Justification: Spent Fuel Pool Area rad monitor activates the evacuation siren on the wall opposite the ZV I Ilelevator. Detector is an area radiation monitor also mounted on the wall next to the siren.

New Fuel Vault rad Monitors also activate evacuation siren but are shielded by concrete from sources outside the vault.

RFE may alarm from shine but does not activate siren.

HC.OP-AR.ZZ-001 9 Attachment A4 RMSYSOE004 (R) Given a scenario of plant operating conditions, evaluate the effect on plant operations lAW the Radiation Monitoring System Lesson Plan if a high radiation level is indicated for:

a. Main Steam Lines

b. Liquid Radwaste Monitoring c.

Iateria!...e~qu foPr E f tion w I None l FadlhtyExm Bank - - 0 na ficati Method: Direct From Source esn50 m : VISION Bank QID# Q56244 Tuesday, March 04, 2003 11:30:37 AM Page 17 of 58

_ _ _ _ _ _ _ _ _ _ _ 13 Which one of the following would require evacuation of part of the Reactor Building area to prevent possible personnel over-exposure?

E An LPRM removal using the LPRM removal tool.

- An SRM Detector driven out of the core using the SRM Drive.

i A TIP detector withdrawn into the TIP Drive Mechanism.

I- A Control Rod Blade unlatched by the CRB removal tool.

Answerl C lExa Lev B l I LeF e lComprehension lFac l Hope Creek Exam 02/24/2 2003 Emergency and Abnormal Plant Evolutions . R0G 2 l l 2 295033K304 295033 High Secondary Containment Area Radiation Levels EK3., Knowledge of the reasons for the following responses as they apply to HIGH SECONDARY CONTAINMENT AREA RADIATION LEVELS:

EK3.04 Personnel evacuation 4.0 44.4 U01ihtibn Justification: Correct: A TIP detector withdrawn into the TIP Drive Mechanism. High radiation source is

-S r j Ioutside the normal storage location and in an easily accessible unshielded location.

Incorrect: An SRM Detector driven out of the core using the SRM Drive. High radiation source detector remains inside the reactor.

Incorrect: An LPRM removal using the LPRM removal tool. Performed underwater for shielding.

Incorrect: A Control Rod Blade unlatched by the CRB removal tool. Performed underwater for shielding TIPSOOEOO9 (R) Given plant problems/industry events associated with the TIP System:

a. Discuss the root cause of the plant problem/industry events.

b. Discuss the HCGS design and/or procedural guidelines that mitigate/reduce the likelihood of the plan

[MP*O e'ur~f7 ~itoi 1 None I

I __

Tuesday, March 04, 2003 11:30:37 AM Page 18 of 58

Quetn Num 14 With a Reactor Building Exhaust Ventillation Radiation High Alarm present, EOP-1 03/4 directs the operator to verify secondary containment isolation of reactor building ventilation and the initiation of FRVS.

WHICH ONE (1) of the following is the reason for this verification?

_=.tr eated nd A r=ollr 0 gro und release Lcon of the activity is provided.

= .ed-g-

A treated and controlled elevated release of the activity is provided.

Il To prevent contamination of normal ventilation ductwork.

M1 To allow accurate monitoring of a release to the environment.

Answe b 1 < B Cognitive el Memory Facilitll Hope Creek Exam Dae 02/24/2003 T.ier Emergency and Abnormal Plant Evolutions 2 $RO Group 2 295034K301 295034 Secondary Containment Ventilation High Radiation EK3. Knowledge of the reasons for the following responses as they apply to SECONDARY CONTAINMENT VENTILATION HIGH RADIATION:

EK3. 01 Isolating secondary containmentventilation 3.8 4.1 9nation0 "Justification:

0 Correct: A treated and controlled elevated release of the activity is provided. Secondary Containment is designed to minimize any ground level release od radioactivity which may result from an accident. FRVS Fans provide charcoal filters to remove radioactive iodine Incorrect: A treated and controlled ground release of the activity is provided. Elevated release Incorrect: To prevent contamination of normal ventilation ductwork. Occurs normally Incorrect: To allow accurate monitoring of a release to the environment. Release from RB is always monitored. Not reason FRVS started.

>f:4K~~~3/4~~iii~ ~ .......... . . ~ ,..~ eeec i Tech-Spec Bases 3/4 6.5 RBVENTE036 (R) Given a scenario of applicable operating conditions and access to Technical Specifications:

a. Choose those sections which are applicable to Reactor Building Ventilation.

b. Evaluate Reactor Building Ventilation operability.

c. E Matria Reuird3fr>:i>. None ionore INPO Exam Bank Qen Mciltti,;eod:g Editorially Modified l uiCA Gbmrneiit I INPO BANK QID#18083 10/16/1998 Pilgrim Tuesday, March 04, 2003 11:30:37 AM Page 19 of 58

s b 15 10CFR 50.54(X) and NC.NA-AP.ZZ-0005 "Station Operating Practices" allow "reasonable action that departs from a license condition or a Technical Specification in an emergency when this action is immediately needed to protect the public health and safety..."

These actions:

El*must be reported to the NRC within 15 minutes of the action being taken.

El must be approved by the Operations -Mianager prior to the action taking place.

mEl be approved by a licensed SRO on the operating shift prior to the action taking place.

must El must be approved by any member of the plant staff who holds a Senior Operators License.

Answe c ERNLkel; B POP i Memory F..l .....

HopeCreek exa 02/24/2003 Xi~ Emergency and Abnormal Plant Evolutions l O p 3 SRPI u 2 295035G101 295035 Secondary Containment High Differential Pressure 2.1 Conduct of Operations 2.1.1 LKnowledge of conduct of operations requirements. 3.7 3.8 9W4a0An Generic KA FORCED to Procedures section on purpose. IGNORE 295035 K/A Title.

1 . Justification:

Correct: - must be approved by a licensed SRO on the operating shift prior to the action taking place.

Incorrect - must be approved by the Plant Manager prior to the action taking place. An SRO on the crew must approve a 50.54(x) call prior to the decision.

Incorrect - must be approved by any member of the plant staff who holds a Senior Operators License. An SRO on the crew must approve a 50.54(x) call prior to the decision.

Incorrect - must be reported to the NRC within 15 minutes of the action being taken. A 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months notification is required.

Reference:

NC.NA-AP.ZZ-0005, Rev. 11, Section 5.4.3 10CFR50.54(x)

ADMPROE007 From Memory Explain the circumstances and approval required for Licensed Operators to deviate from Technical Specifications or license conditions. lAW NC.NA-AP.ZZ-0005 and 10CFR50.54(x),and SH.OP-AP.ZZ-01 02 MateriL e for xaNone Facility Exam Bank .

lQuestion atoi, i Direct From Source QuenSur = leQ57018 I

Tuesday, March 04, 2003 11:30:37 AM Page 20 of 58

N 16 Which one of the following describes the only position on the shift complement specified in the Technical Specifications that can NOT be reduced temporarily by one less than the minimum to accomodate unexpected absence of on-duty shift crew members?

El CRS.

E OSP.

El RO/PO.

E ST A.

b xFam :I B KgI Level'Memory le Faciit Hope Creek E.RPR e 02/24/2003 Tier: Emergency and Abnormal Plant Evolutions 0 AdGr 3 $RZru 2 295035G104 295035 Secondary Containment High Differential Pressure 2.1 Conduct of Operations 2.1.4 Knowledge of shift staffing requirements. 2.3 3.4 E~nat~i&f, Generic KA FORCED to Procedures section on purpose. IGNORE 295035 K/A Title.

pWer J Justification:

Correct:OS. As stated in NC.NA.AP.ZZ-0005 Attach 9 paragraph 5 and Tech Specs Table 6.2.2-1 Incorrect: CRS. May be short up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Incorrect: RO/PO. May be short up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Incorrect: STA. May be short up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months

.I 7K M R iss WilINWr =1t11 l - ,, :m .

NC.NA-AP.ZZ-0005, rev 1 Section 5.14, attachment 9 Par 5

_ .~i _< _- n_b e ti e .~. ..

ADMPROE021 Given plant conditions and/or access to control room references Determine the following:

The level of licensing required for the OS, CRS, and RO/PO.

Minimum shift manning requirements for all plant conditions.

Normal shift staffing levels.

When a pers IMZ~ate~aI Required f&,~w* ino n. fl None. Remove Tech Spec Admin section from references.

j00iQt Source Facility Exam Bank l- sti niModifcation Me j Direct From Source Quedtl Source Go ~: l VISION Bank QID# Q54277 Tuesday, March 04, 2003 11:30:38 AM Page 21 of 58

'QUesi9n tJumber 17 L

Given the following conditions:

- New fuel is being lifted to the refueling floor from the Reactor Building Truck Bay with the polar crane.

- Workers are preparing new fuel for inspection.

Which of the following configuations of new fuel is NOT allowed?

i-I 4 full crates stored on top of each other.

1 5full crates laid sicde by side next to each other.

iI 2 new bundles in the inspection stand and one suspended from the polar crane.

M 5 new bundles in the new fuel storage rack with one in an open crate.

Anwra Emev B Contv Application Falitl Hope Creek Exa Date: 02124/2003 T Emergency and Abnormal Plant Evolutions . 3 SRO 2 295035G110 295035 Secondary Containment High Differential Pressure 2.1 Conduct of Operations 2.1.10 Knowledge of conditions and limitations in the facility license. 2.7 3.9 l a poo] Generc KA FORCED to Procedures section on purpose. IGNORE 295035 K/A Title.

An.~~rjj~j

- Justification:

4 full crates stored on top of each other. Correct. HC Operating License condition 2.C.6 states "Fresh fuel asemblies, when stored in their shipping containers shall be stacked no more than 3 containers high."

5 full crates laid side by side next to each other. Incorrect. Not prohibited by HC Operating License condition 2.C.6 2 new bundles in the inspection stand and one suspended from the polar crane. Incorrect. Not prohibited by HC Operating License condition 2.C.6.

5 new bundles in the new fuel storage rack with one in an open crate. Incorrect. Not prohibited by HC Operating License condition 2.C.6.

HC.RE-FR.ZZ-0001 P&L 3.2.5 HC Operating License condition 2.C.6

.------ OL0ectMeR:. --....

TECSPCE010 (R) Given specific plant operating conditions and a copy of the Hope Creek Generating Station Technical Specifications, evaluate plant/system operability and determine required actions (if any) to be taken. (SROISTA Only) ia;R f/;r m i HC 100 percent operating license.

~~ionrSeurl7 New Xues0 on ModificationMethod:

[Quein Source Comments;>

Tuesday, March 04, 2003 11:30:38 AM Page 22 of 58

s n00~p~ N i5X-,0r 1 18 Given the following conditions:

- Troubleshooting on the refuel platform air accumulator auto drain trap is complete.

- The clearance tags have been removed.

- The trap was still blowing air by slightly.

- The air system will be returned to service with the trap manually isolated and instructions to manually unisolate and blowdown hourly when in use.

Which one of the following tags is placed on the isolation valve to document the instructions while allowing hourly use?

El Red Blocking Tag._

[Al White Caution Tag.

'lYellow Permissive Tag.

E Administrative Tag.

MOWe b B Ggdli Lel Memory 1Fal:0llW Hope Creek Exam 02/24/2003 Ti Emergency and Abnormal Plant Evolutions ROGroup 3 Id rop 2 295035G220 295035 Secondary Containment High Differential Pressure 2.2 _, Equipment Control 2.2.20 Knowledge of the process for managing troubleshooting activities. 2.2 3.3 Eflnaitoio Generic KA FORCED to Procedures section on purpose. IGNORE 295035 K/A Title.

Justification:

Correct: White Caution Tag. lAW NC.NA-AP.ZZ-0015 5.4.4. Used for abnormal operating conditions Incorrect: Red Blocking Tag. Does not allow valve manipulation with tag present.

Incorrect: EMIS Tag. Used to identify malfunction.

Incorrect: Administrative Tag. Not a physical tag. Communicates personnel of adminstrative and safety requirements NC.NA-AP.ZZ-0015 5.4.4

__ -i -- 0-. --- s- At\ h.jL NA0015E004 Identify the kinds of tags and their purpose lAW the Safety Tagging Procedure, NC.NA-AP.ZZ-0015(Q) and the SAPANCM Tagging Operations Procedure, SH-OP-AP.ZZ-0015(Q).

Material eiwre fo None 1e stuM :jj Other Facility 9i die I etd Editorially Modified

'Peach Bottom 2002 LSRO question 3-8 modified for Hope Creek.

Tuesday, March 04, 2003 11:30:38 AM Page 23 of 58

iQabi Number 19 Given the following conditions:

- A fuel handling tool malfunctions causing high radiation conditions on the refuel floor.

- A worker receives an accidental radiation exposure on the Refueling Floor of 6.5 Rem TEDE.

Which one of the following correctly describes the time limit for reporting the event to the NRC?

Ili1 1hour.

Hi 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

EI 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .__

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

nswe b Exame B CognitvL$Vl Application lFacilty:: Hope Creek Pxai at 02/24/2003 tier: Emergency and Abnormal Plant Evolutions : Q.Group 3 SRO lGrop 2 295035G430 295035 Secondary Containment High Differential Pressure 2.4 Emergency Procedures and Plan 2.4.30 Knowledge of which events related to system operations/status should be reported to outside 2.2 3.6 agencies.

l§iaaonT] Generic KA FORCED to Procedures section on purpose. IGNORE 295035 K/A Title.

Answer IJustification:

Correct: 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . ECG Reportable Action Level # 11.4.2.a Incorrect: 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Incorrect: 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Incorrect: 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months HC ECG RAL 11.4.2.a

--.. rin I ....

[Matprla1 RequiredEal wint ;HC ECG without Introduction and Usage section IpusAOdOzR] New lig.rs'i }'Modif-cation Metf6d; llon R sous.Ga'm n7 E Plan for Operations Duties NEPLICOPSHCC Obj 6.1 Tuesday, March 04, 2003 11:30:38 AM . Page 24 of 58

m 20 New fuel is being unloaded on the refuel floor for inspection and placement in the New Fuel Vault.

Which one of the following is required by the Fire Protection Program for this work and the reason for the requirement?

tH Continous Fire Watch because the shipping crates are combustable.

El Hot Wok Permit because flammable chemicals are required.

Mi Class D fire extinguisher because ignitable metals are present.

M ALL FOG nozzles placed on fire hose stations to prevent inadvertant criticality.

WE c E vI B Clgni l Comprehension ciity: Hope Creek lmDate 02/24/2003

...... Emergency and Abnormal Plant Evolutions Id.l 2 god l 2 600000K302 600000 Plant Fire On Site EK3. Knowledge of the reasons for the following responses as they apply to PLANT FIRE ON SITE:

EK3.02 Steps called our in the site fire protection plant, fire protection system manual, and fire zone manual 2.2 2.8 lpentio. Justification:

Correct: Class D fire extinguisher because ignitable metals are present. Fuel bundles consist of fuel pins, fuel pin end plugs, channels, and spacer made from Zirconium which is an ignitable metal. A class D fire extinguisher must be readily available.

Incorrect: Continous Fire Watch because the shipping crates are combustable. Shipping crated are non-combustable metal.

Incorrect: Hot Work Permit because flammable chemicals are required. Hot work is cutting or welding.

Use of chemicals requires a chemical use permit.

Incorrect: ALL FOG nozzles placed on fire hose stations to prevent inadvertant criticality. Fog nozzles are replaced with solid stream nozzles when new fuel is present to prevent foam moderation causing criticality.

I 2. ,.,eN'.0'le'eof m" ^'"-Wtle^

NC.FP-AP.ZZ-0012 Table 2 and step 4.4.2

_.a.ni.

.. . _ Obj.e v WmG^ wo, on ADMPROE064 Given access to Control Room References Determine where ignitable metals are used at HCGS. lAW NC. NA-AP.ZZ-0025 IMat -ird~ E oW ~NC.

I NA-AP.ZZ-0025, NC. FP-AP.ZZ-0025, and NC.FP-AP.ZZ-0012

[ noueFI New Q Modification Med Quetin.Soulc Tuesday, March 04, 2003 11:30:38 AM Page 25 of 58

Ir ue 21 Given the following conditions:

- The plant is in Operational Condition 5 following shutdown using normal procedures.

- Core offload is in progress.

- Shutdown Cooling is in-service through the B Loop.

- An inadvertent full Channel C LPCI Initiation signal is received.

Which one of the following describes the response (if any)? _

RlReactor cavity water level will lower because LPCI Loop C test return valve will open to the suppression pool.

N Reactor cavity water level will lower because LPCI Loop C min-flow valve will open.

Reactor cavity water level will rise because LPCI Loop C will inject to the reactor vessel.

LI Reactor cavity water level will remain unchanged because LPCI Loop C is isolated from injection.

Aswer d REx)ejl B i Comprehension Hope Creek xa Date 02/24/2003 fr.ei Plant Systems ROGp 1 SROGrop 1 203000K301 203000 RHR/LPCI: Injection Mode (Plant Specific)

K3. Knowledge of the effect that a loss or malfunction of the RHR/LPCI: INJECTION MODE will have on following:

K3.01 Reactor water level 4.3 4.4 Justification: =

Reactor cavity water level will remain unchanged because LPCI Loop C is isolated from injection during refueling. Correct. HC.OP-IO.ZZ-0005 step 5.2.46 isolated RHR and Core Spray from injection prior to refueling.____

HC.OP-1O.ZZ-0005 step 5.2.46

- - , .......... f emig Objectives = f 1OP005E006 (R) Analyze plant conditions and parameters to determine if plant operation is in accordance with the COLD SHUTDOWN TO

- REFUELING Integrated Operating Procedure, supporting System Operating Procedures and Technical Specifications.

Material e .dP Pm Noniei f00 QusTnWre _ . New Quetifl~nMdmti tiptn Methock r Teii daiysMrchewm203113: Ag o Tuesday, March 04, 2003 11:30:38 AM Page 26 of 58

tn Nur 22 Given the following conditions:

- Refueling is in progress.

- The Reactor Mode Switch is locked in REFUEL.

- Source Range Monitors A, C, and D are operable; SRM B is inoperable.

- Shutdown margin has been verified.

- All control rods are at position 00.

- As a fuel assembly is taken to the fuel pool through the transfer canal, the RO observes that the

'C' SRM counts have dropped to zero.

- The refueling crew stops the bridge in the fuel pool.

After reviewing applicable procedures, Core Alterations can continue:

El with no restrictions.

El for Control Rod Blades only.

l only inthe quadrants monitored by SRMs A_ nd D.

i1 only in the quadrants monitored by SRMs A and C.

Answer E0 i~

c ,13B C giv L Application i Hope Creek lxamDate: 02/24/2003 Plant Systems 11 SR I 11 215004G232 215004 Source Range Monitor (SRM) System 2.2 Equipment Control 2.2.32 Knowledge of RO duties in the control room during fuel handling such as alarms from fuel handling 3.5 3.3 area, communication with fuel storage facility, systems operated from the control room in support of fueling operations, and supporting instrumentation.

9Ia0i0060 Justification: lAW Tech Spec 3.9.2.b - One operable SRM must be in the quadrant where the core alteration is taking place and one in an adjacent quadrant.

"can continue only in the quadrants monitored by SRMs A and D"- Correct "can continue with no restrictions" - Incorrect- core Alts in only A/D quadrants.

"for Control Rod Blades only." - Incorrect- core Alts in only A/D quadrants.

"can continue only in the quadrants monitored by SRMs A and C." - Incorrect- can continue in A/D quadrants only.

_ .~ _ _ _ _ _. _

Tech Spec 3.9.2.b SRMSYSE007 (R) Given a scenario of applicable operating conditions and access to technical specifications.

a. Choose those sections which are applicable to the SRM system.

b. Evaluate SRM operability and determine required actions based upon system operability.

I.NOW,. i :Tech Spec section 3.9 s0003 Fs" Facility Exam Bank Q Significantly Modified Go'iii rC ients:>VISION QiD# 60987. Significantly modified.

Wednesday, March 19, 2003 10:15:29A M Page 27 of 58

s 23 Given the following conditions:

- The reactor is defueled.

- The reactor mode switch is in locked in Shutdown.

- LPRM changeouts are in progress.

- The Control Room reports a control rod block and half scram is received from "A" APRM.

Which one of the following would cause the rod block and half scram?

(Consider each answer choice seperately and assume remaining LPRMs are working normally) assigned to "A" APRM is (are) placed in EIl 1 of 4 "B" level LPRMs; BYP.

3 of 4 "B" level LPRMs; CAL.

3I 8 of 21 LPRMs; BYP.

II1 of 21 LPRMs; CAL.

Awe c Exax B ltiveLel Comprehension lity Hope Creek Exam 02/24/2003 Tier: I Plant Systems O 1 rup 1 215005A103 215005 Average Power Range Monitor/Local Power Range Monitor System Al. Ability to predict and/or monitor changes in parameters associated with operating the APRM/LPRM controls including:

A1.03 Control rod block status 3.6 3.6 Justification:

1 -S0 ;8 of 21 LPRMs assigned to "A" APRM are placed in BYP. Correct. With only 13 LPRMs in Operate, APRM INOP occurs with Reactor Mode Switch NOT in Run.

1 of 4 "B" level LPRMs assigned to "A" APRM is placed in BYP. Incorrect. 20 LPRMS remain in Operate.

No APRM Inop trip.

3 of 4 "B" level LPRMs assigned to "A" APRM are placed in CAL. Incorrect. 18 LPRMS remain in Operate. No APRM Inop trip. Administrative Inop only.

1 of 21 LPRMs assigned to "A" APRM is placed in CAL. Incorrect. 20 LPRMS remain in Operate. No APRM Inop trip HC.OP-SO.SE-0001 APRMOOE010 (R) From memory, explain why an inoperable trip of an APRM channel is initiated if there are less than 14 LPRM inputs, 1AW the

- Student Handout.

foil:an u .i Tech spec section 3.3 with Table 3.3. 1-1 removed.

New Q M a _M l i:.%~ ... i POW..

tvrs. 7w, a:.i- ----

Tuesday, March 04, 2003 11:30:39 AM Page 28 of 58

[pues-io I 24 Given the following conditions:

- The plant is in Operational Condition 5.

- All RBVS and RBVE fans are running.

- FRVS is in a normal standby configuration.

- "B" and "D" Diesel Generators are tagged out for maintenance.

A radiological incident on the Refuel Floor causes Refuel Floor Exhaust Radiation to reach 4.5E-3 uci/ml.

Select total FRVS recirculation flow one minute after this event.

(Assume no operator actions)

EOl cfm.

Ei 90,000 cfm.

I- 120,000 cfm. ____

IM 180,00 cfrn.

I l d aLel B itvle Comprehension Fll ilylHope Creek Exam 02/24/2003 WE Plant Systems I RQlrup 1 1 261000A301 261000 Standby Gas Treatment System A3. Ability to monitor automatic operations of the STANDBY GAS TREATMENT SYSTEM including:

A3.01 System flow 3.2 3.3

[ o ofl JUSTIFICATION:

A -Correct Answer:"180,000 cfm" All six FRVS fans will automatically start on the high Refuel Floor Exhaust radiation signal because there is no loss of bus power or offsite power stated. All RBV fans running.

EDGs out for maintenance will not prevent start of all fans.

Incorrect Answers: 0 cfm. All recirc fans start and run until manually secured.

"120,000 cfm" . Incorrect. No operator actions taken to secure 2 fans.

"90,000 cfm". Incorrect. B and D EDG maintenance have no effect.

- - End of Justification***

_P_ .: .G en e_ ... _tl HC OP-SO GU-0001 Wff_NW N >j :27, [Nn RBVENTE006 (R) Given plant conditions, distinguish between the automatic starts and stops associated with the Filtration Recirculation Ventilation System (FRVS) Recirc Fans.

HN I2None Cetiop SourceFacility Exam Bank queston Moifict,{th alEditorially Modified Queton StcCn Vision Bank QID # Q60662.

Tuesday, March 04, 2003 11:30:39 AM Page 29 of 58

$uWWMONun 25 125 VDC bus 1CD417 is deenergized when an Emergency Diesel Generator start signal is received.

Which of the following describes the effect on Diesel Generator 1CG400?

EiI The diesel generator will NOT automatically start.

1El The diesel generator will start from the Main Control Room but the automatic trips will be disabled. _

Ed The diesel generator will automatically start but in the DROOP mode.

ML The diesel generator will automatically start but the output breaker can only be shut manually.

a Ea ee B nW I: Memory li Hope Creek E ~m 02/24/2003 Piant Systems If 1 ROP;ou 1 Sp 264000K609 264000 Emergency Generators (Diesel/Jet)

K6. Knowledge of the effect that a loss or malfunction of the following will have on the EMERGENCY GENERATORS (DIESEL/JET):

K6.09 D.C. power 3.3 3.5 lanatlp Justification:

Correct: The diesel generator will not automatically start. DC control power is needed to open the Air Start Solenoids. Energized to open.

Incorrect: The diesel will automatically start but the output breaker can only be shut manually. The diesel will not start manually or automatically.

Incorrect: The diesel generator will automatically start but in the DROOP mode. The diesel will not start manually or automatically.

Incorrect: The diesel generator will start from the Main Control Room but the automatic trips will be disabled. The diesel will not start from the Main Control Room or automatically.

HC.OP-SO.KJ-0001 HC OP-AR.KJ-0006 Attach 37 ff _ 0 =.. __=gXjecivs :< , of0X 0A EDGOOOE01 1 (R) Given plant conditions, predict the response of Diesel Generator governor and voltage regulator control circuitry to an Emergency start (LOP/LOCA).

lMtn1 Re wi for kamino None Facility Exam Bank Q Modifi ne Editorially Modified Qt SreVision Bank QiD# Q53558. Modified answer choice "The diesel generator will start but the automatic trips will be disabled." because it is correct. Air Start Solenoids can be positioned with manual levers. The engine will start but there are no elec trips.

Tuesday, March 04, 2003 11:30:39 AMPg0f Page 30 of 58

26 WHICH ONE of the following ruptured areas would prevent establishing two-thirds(2/3) core coverage following a Design Bases Accident LOCA?

l Jet pump inlet riser.

-Ed Jet pump dffuser section.

E1 Shutdown Cooling suction iine.

Low pressure LPC injetion line.

Li b PPRie1l B C6n Lel Memory FaciII!yl Hope Creek E ahfite 02/24/2003 iPlant Systems WRMJ 21 2 1SRGo 20200 1 K401 202001 Recirculation System K4. Knowledge of RECIRCULATION System design feature(s) and/or interlocks which provide for the following:

K4.01 2/3 core coverage: Plant-Specific 3.9 3.9 lxianatioW4 f:, Justification:

Correct: Jet pump diffuser section. Part of jet pump boundry to establish 2/3 core floodable volume Incorrect: Jet pump inlet riser. Located in the downcomer region which if ruptured, would not drain the rpv.

Incorrect: Shutdown Cooling suction line. Would only drain the downcomer annulus area.

Incorrect: Low pressure LPCI injection line. Located at the top of the core at the core plate which is above 2/3 core coverage.

Tech Spec Bases 3/4.4.1 RECIRCE007 Explain/Discuss how the Recirculation System is designed to ensure a 2/3 core height refloodable volume is maintained. lAW available control room references:

atl R e or i O None Euioi<So.0 INPO Exam Bank 1Qt Modifition Metod Direct From Source QuestionSUrctegCmmes: INPO Bank QID# 12456 07/02/1999 Lime Tuesday, March 04, 2003 11:30:39 AM Page 31 of S8

ustioi~umir 2 27 Given the following conditions:

- The plant is in Operational Condition 4 with coolant temperature at 185 0 F.

- The "B" Loop of RHR is in Shutdown Cooling.

- An RPV water level transient occurs.

- RPV level has lowered to -20 inches and is stabilized at -15".

- RPV water level cannot be raised to Level 3.

- HC.OP-AB.RPV-0009, Shutdown Cooling is entered.

Which of the following decay heat removal methods will be effective for these conditions?

[i] Alternate Shutdown cooling using "D" to "B" RHR pump cross-tie.

Manually

-ll operate SDC valves and restart B RHR pump.

El RHR RPV head spray with all RPV head vent valves open.

I Maxirmizing Fuel Pool cooling with both Fuel Pool heat exchangers.

sw b Exam L1 B 1 i0tvLe Application Hope Creek Exam Da 02/24/2003

[Plant Systems RGu 2 [#UP 2 205000A205 205000 Shutdown Cooling System (RHR Shutdown Cooling Mode)

A2. Abiltyto (a) predict the impacts of the following on the SHUTDOWN COOLING SYSTEM/MODE; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

A2.05 System isolation 3.5 3.7 hVxpnattn9ot,. Justification:

nse .Correct: Manually operate SDC valves and restart B RHR pump. Based on HC.OP-AB.RPV-0009 Shutdown Cooling Action A3 and A4.

Incorrect: Alternate Shutdown cooling using "D" to "B" RHR pump cross-tie. Still isolated at Level 3{+12.5"].

Incorrect: RHR RPV head spray with all head vent valves open. Requires 2 SRV's open with at least 50 psid RPV to SP per HC.OP-AB.RPV-0009 section H. Head spray line would be isolated with RPV level below Level 3.

Incorrect: Maximizing fuel pool cooling. In OC 4, the RPV head is still in place, FPCCU cannot be used.

ON S 1 i fl1- -A HC.OP-AB.RPV-0009 subsequent actions "A3 and A4"

_______ < LerigO Xf~ - b . * ;S^§... . a ABRPV9EO07 (R) Explain the bases for Subsequent Actions and the information contained in the Discussion Bases Section of Shutdown Cooling.

lMWAWe"~Rqie frxuiin

\ S

. Sv . S. .

\ ww .. . .... x . . . = CY4X,

- None Facility Exam Bank

QUestio M ationMe0th l Sign ificantiy Modified mRwp VISION Bank QID #Q61331 Stem modified to provide a correct answer. Answers modified.

Tuesday, March 04, 2003 11:30:39 AM Page 32 of 58

e 28 Given the following conditions:

- The plant is operating in Operational Condition 4 in preparation for refueling outage.

- Residual Heat Removal (RHR) System Loop B is operating in the Shutdown Cooling (SDC) mode.

- RPV pressure has increased from 50 to 100 psig.

In addition to B RHR PUMP(BP202) tripping, which of the following system automatic responses will occur?

[i RHR SHUTDOWN CLG INBD ISLN(F009) closes; RHR LOOP B RET TO RECIRC(F015B) opens; RHR PMP B SUCT FROM RECIRC(F006B) closes.

RHR SHUTDOWN CLG OUTBD ISLN(FO08) closes; RHR PMP B SUCT FROM RECIRC(FO06B) closes; RHR LOOP B RET TO RECIRC(F015B) closes.

MIRHR SHUTDOWN CLG OUTBD ISLN(F008) closes; RHR LOOP B RET TO RECIRC(FO15B) closes; RHR PMP B MIN FLOW MOV(FO07B) opens.

El RHR SHUTDOWN CLG OUTBD ISLN(F008) closes; RHR SHUTDOWN CLG INBD ISLN(F009) closes; RHR LOOP B RET TO RECIRC(FO15B) closes.

Alswer d lamel B 0 gnl9i&Le'e Memory Hope Creek Eam D 02/24/2003 mg Plant Systems -I el 2 ROGop 2 205000K402 205000 Shutdown Cooling System (RHR Shutdown Cooling Mode)

K4. J Knowledge of SHUTDOWN COOLING SYSTEM/MODE design feature(s) and/or interlocks which provide for the following:

K4.02 High pressure isolation: Plant-Specific 3.7 3.8 l ,a f Justification:

RHR SHUTDOWN CLG OUTBD ISLN(F008) closes, RHR SHUTDOWN CLG INBD ISLN(F009) closes, RHR LOOP B RET TO RECIRC(F015B) closes. Correct. 8,9, and 15's close on RPV Pressure isolation.

RHR SHUTDOWN CLG OUTBD ISLN(F008) closes, RHR PMP B SUCT FROM RECIRC(FO06B) closes, RHR LOOP B RET TO RECIRC(F015B) closes. Incorrect. F006 does not automatically close.

RHR SHUTDOWN CLG OUTBD ISLN(F008) closes, RHR LOOP B RET TO RECIRC(F015B) closes, RHR PMP B MIN FLOW MOV(F007B) opens. Incorrect: F007B does not automatically open on a pump trip. The pump trips prior to pump low flow 10 sec time delay.

RHR SHUTDOWN CLG INBD ISLN(F009) closes, RHR LOOP B RET TO RECIRC(FO15B) opens, RHR PMP B SUCT FROM RECIRC(F006B) closes. Incorrect. F015B does not automatically open. F006B does not automatically close.

HC.OP-SO.BC-0002 r~~~~~~ < ==-:: - Objectives

- X~~ti ::; X :

RHRSYSE01 1 Given a labeled drawing of, or access to the Residual Heat Removal System controls/indication on 1OC650:

a. Explain the function of each indicator lAW the RHR System Lesson Plan.

b. Assess plant conditions which will cause the indicators to Tuesday, March 04, 2003 11:30:40 AM Page 33 of 58

WOW i na0o Tech Spec section 3.3 with TABLE 3.3.2-1 removed; Provide panel drawing of 10C650A for RHR A and B.

99sinorIo Facility Exam Bank= uestion i Wibctio l Significantly Modified CbmQ55155 Q.esnSourcl Significantly Modified Tuesday, March 04, 2003 11:30:40 AM Page 34 of 58

o 29 Which one of the following supplies power to the Intermediate Range Monitoring System Channel drawers?

24 _i Volt Non-1 E DC batteries.

--- 125 Volt Non-1 E DC batteries.

125 Volt 1E DC batteries.

iM 250 Volt 1E DC batteries.

(a E Ldz 1B 1 lo Memory lA l Hope Creek Ea e 02/24/2003 Oeri 'Plant Systems grou 1l S5R9PGo , 2 215003K201 215003 Intermediate Range Monitor (IRM) System K2. Knowledge of electrical power supplies to the following:

K2 01 IRM channels/detectors _ 2.5 2.7 gxlaxpg" Justification:

Correct: 24 Volt Non-1 E DC batteries. Supplies all SRM and IRM drawer power Incorrect: 125 Volt Non-1 E DC batteries. No connection to IRMs Incorrect: 125 Volt 1E DC batteries. No connection to IRMs Incorrect: 250 Volt 1E DC batteries. HPCI and RCIC only. No connection to IRMs E-001 0

.:~jgOJnYS1__

___~:;2:..

DcELECE004 (R) Summarize the interrelationship(s) between 24VDC Power System and the following lAW the DC Electrical Distribution Lesson Plan.

a. Auxiliary Building Ventilation System

b. 1E AC Electrical Distribution System

c. Neutron Monitoring Syst

_riaI RK dIfr"Tech Spec section 3.8 with 3.8.3.1 removed. _ -

I oi eNew IQ i

.7 M -- ---

Thursday, March 20, 2003 11:15:57AM i Page 35 of 58

Qeon e 30 320 Plant conditions are as follows:

- In-vessel maintenance is in progress.

- Control Rod Blade (CRB) 18-15 is to be removed using the Combined Grapple.

- The Control Room does NOT have position indication.

- When the CRB is lifted using the Monorail Hoist, load indication is +400 lb.

- The 'UP' pushbutton is released.

WHICH ONE of the following explains the reason for these indications?

Ei The CRB is still coupled to the drive .-0 E]The CRB bail handleehas broken free.

El The CRB removal tool air hoses are slack.

El The CRB exceeded the setpoint of the hoist cutoff.

Anse a lExamLF B oniive Lel Comprehension Faciity:lHope Creek a t 02/24/2003 T: Plant Systems 3 SRO Grp 2 234000K105 234000 Fuel Handling Equipment K1. Knowledge of the physical connections and/or cause- effect relationships between FUEL HANDLING EQUIPMENT and the following:

KK1.05 Reactor vessel components: Plant-Specific 2.9 3.3 Atiswr Correct: The CRB is still coupled to the drive. HC.RE-FR.ZZ-0002 Caution 5.6.11 indications of a uncoupled CRB are 340 pounds; >400 coupled.

Incorrect: The CRB bail handle has broken free. The weight would be lower than a CRB.

Incorrect: The CRB removal tool air hoses are slack. Air hoses are slack deliberately, otherwise they pull up on the load.

Incorrect: The CRB exceeded the setpoint of the hoist cutoff.. Monorail hoist load cell cutoff is set at 500

+1-50 pounds HC.OP-ST.KE-0001 step 5.4.8.

ROOM= Tit}e HC.RE-FR.ZZ-0002 Caution 5.6.11 REFUELE010 Given a drawing of, or access to, the frame mounted hoist or monorail hoist control pendant, explain the controls and indications IAW the Student Handout.

eonMonorail Hoist Pendant Figure from Refueling Platform Lesson Plan a Metod: Significantly Modified

INPO Bank QID# 14062 07/02/1999 Peach Bottom Tuesday, March 04, 2003 11: 30:40 AM Page 36 of 58

Que1stionNume iM i+' ' 0 31 31 Given the following conditions:

- The plant is in Operational Condition 5.

- The Reactor Mode switch is in the REFUEL position.

- The Refueling Plafform (bridge) is over the Reactor Vessel.

A control rod block will occur when...

IN the_Fuel Grapple is loaded with fuel.

El9the FuelsGrapple is in the- FULL UP position.

{El the Frame Mounted Auxiliary Hoist is loaded with fuel.

El1 all rods are Full-In, except for a selected rod at position 02.

Ans5e a Exa LP:el B 4g ll Comprehension Faii Hope Creek Exam Date: 02/24/2003 IE Plant Systems R rp 3 SO op 2 234000K502 234000 Fuel Handling Equipment K5. Knowledge of the operational implications of the following concepts as they apply to FUEL HANDLING EQUIPMENT:

K5.02 Fuel handling equipment interlocks 3.1 3.7

[OIAKAtW~fi Justification:

- The Fuel Grapple is loaded with fuel.-Correct- IAW HC.OP-SO.KE-0001 section 3.3.1

- The Fuel Grapple is in the FULL UP position.-Incorrect- nothing associated with full up

- The frame mounted Auxiliary Hoist is loaded with fuel.-Incorrect- the Auxilairy hoist has a load cutout at 500 lbs to prevent fuel moves See TS 4.9.6.b

- All rods are Full-In, except for a selected rod at position 02.-Incorrect- 02 same as 00 so no rod block from RMCS B;^ 't..ce..rD^-

HC.OP-SO.KE-0001

'k. MManngO~eci REFUELE005 (R) Given a drawing of, or access to, the interlock status display panel, and normal Control Room references, explain the information provided by each light and any automatic actions which should occur when light is illuminated lAW the Student Handout.

Material eFigure of Interlock Display panel from Refueling platform lesson plan. Tech Spec section 3.9 with 3.9.1 removed.

Qu6es~urei Facility Exam Bank Question Mdifitin Method: Editorially Modified e I uc-- Comments: Vision bank QID #Q56552 Tuesday, March 04, 2003.11:30:40 AM Page 37 of 58

Qutir4?lne 32 Given the following conditions:

- Core reload is in progress.

- Reactor Building Ventilation is aligned for refueling.

- An irradiated fuel bundle bumps the RPV wall and falls free into the core.

Based on this observation, how will the Radiation Monitoring System respond and what immediate operator action is required IAW HC.OP-AB.CONT-0005 IRRADIATED FUEL DAMAGE?

il Reactor Building Exhaust Radiation monitors will alarm first and FRVS will trip. Suspend all refueling operations.

EliIRReactor Building Exhaust Radiation monitors will alarm first and RBVS will start. Evacuate the refueling floor.

Refuel Floor Exhaust Radiation monitors will alarm first and FRVS will start. Suspend all refueling operations.

9I Refuel Floor Exhaust Radiation monitors will alarm first and RBVS will trip. Evacuate the refueling floor.

Cc E e1 IB VIo Lee lMemory FaclitW Hope Creek l 02/24/2003 Plant Systems uP 2 S rup 2 272000A201 272000 Radiation Monitoring System A2. Ability to (d) predict the impacts of the following on the RADIATION MONITORING SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations

A2.01 Fuel element failure 3.7 4.1

[pnatfio1nol Justification:

Answer Correct:Refuel Floor Exhaust Radiation monitors will alarm first and FRVS will start. Suspend all refueling operations. Design basis of RFE RMS. All airflow off the refuel floor passes RFE RMS elements 4856A,B,C. FRVS starts on RFE Hi Rad 2.OE-3 uci/cc. IOA of HC.OP-AB.CONT-0005 IRADIATED FUEL DAMAGE Incorrect:Reactor Building Exhaust Radiation monitors will alarm first and FRVS will trip. Suspend all refueling operations. RFE RMS will alarm first. FRVS starts.

lncorrect:Reactor Building Exhaust Radiation monitors will alarm first and RBVS will start. Evacuate the refueling floor. RFE RMS will alarm first. RBVS trips and isolates.

Incorrect: Refuel Floor Exhaust Radiation monitors will alarm first and RBVS will trip. Evacuate the refueling floor. Subsequent action of HC.OP-AB.CONT-0005 RADIATED FUEL DAMAGE it'.le4'0g0 X:§0. .................. i.s'^j' ' 0'e ' !]'

7"d'"

Hi HC.OP-AB.CONT-0005 RADIATED FUEL DAMAGE M-76, M-84 1'% .Q v . Rlo........ .s d50i j0 ( .................

ABCNT5EO03 (R) From memory, recall the Immediate Operator Actions for Irradiated Fuel Damage.

ABCNT5EO04 Explain the reasons for how plant/system parameters respond when implementing Irradiated Fuel Damage.

RMSYSOE004 (R) Given a scenario of plant operating conditions, evaluate the effect on plant operations IAW the Radiation Monitoring System Lesson Plan if a high radiation level is indicated for:

a. Main Steam Lines

b. Liquid Radwaste Monitoring c.

Tuesday. March 04, 2003 11:30:40 AM Page 38 of 58

None lgtion" -New X 0 I Qu. o M diulcat n M e IQuesti SQ laneiC~ainm =s: _

Tuesday, March 04, 2003 11:30:41 AM XPage P 399 of 58

_m 33 Given the following conditions:

- The plant is in Operational Condition 4 with all systems running normally.

- A and C SACS pump are running supplying TACS.

- B SACS pump is running.

- D SACS pump is in AUTO, NOT running.

Which one of the following results in an automatic start of the non-running SACS pump?

El Low flow on the A Control Room Chilled Water pump.

El Low-Low-Low Level innthe A SACS Expansion TTank..

El Low differential pressure on the B SACS pump.

El Low-Low-Low Level in the B SACS Expansion Tank.

Answer0 b Exam=eel IB Cogtv Le Memory Facill Hope Creek l'aJDatd 02/24/2003 Plant Systems ROGoup 2 ROru 2 400000K401 400000 Component Cooling Water System (CCWS)

K4. Knowledge of CCWS design feature(s) and or interlocks which provide for the following:

K4.01 Automatic start of standby pump 3.4 3.9 lExollatib f Justification:

Correct: Low-Low-Low Level in the A SACS Expansion Tank. Low-Low-Low in the Expansion tank supplying TACS will isolate TACS valves, causing low flow in the A SACS Loop which auto starts B and D SACS Pumps. B is already running.

Incorrect: Low flow on the A Control Room Chilled Water pump. Auto starts A SACS Pump if not running.

Incorrect: Low differential pressure on the B SACS pump. Trips the B SACS pump but does not start the D.

Incorrect: Low-Low-Low Level in the B SACS Expansion Tank. Would isolate TACS if on the B loop.

[I <.f: ~ ~~~ .d, 3'i:. 2gRefrenc Tdeg~dK l-..l<°k;g tl t0\0: 0Q0000; HC.OP-SO.EG-0001 STACSOE016 Determine the following information for SACS pumps:

Time delay associated with the SACS pumps when automatically started by either the LOCA or LOP sequencer.

Automatic start signals Automatic trip signals lAW available control room references ij yfEx None Dstiuci] New Modi onM ol Tuesday, March 04, 2003 11:30:41 AM Page 40 of 58

34 During control rod scram testing, control rod 26-19 is observed to scram (fully) in 2 seconds.

Which one of the following describes the cause of the observed condition and the components that would be damaged?

IHl Excessive accumulator gas pressure. Damage to the Beliville washers.

IiN Worn lower stop piston seals. Damage to the CRD Guide tube.

E] Worn drive piston seals. Damage to the Collet Fingers.

tell Inadequate accumulator gas pressure. Damage to the Index tube.

Answer a EeveI B Cognitiv Level Comprehension Faility: Hope creek Examate 02/24/2003 Plant Systems 2 Slo ig 3 201003K101 201003 Control Rod and Drive Mechanism K1. Knowledge of the physical connections and/or cause- effect relationships between CONTROL ROD AND DRIVE MECHANISM and the following:

K1.01 Control rod drive hydraulic system 3.2 3.3 lna .f JUSTIFICATION:

I lCorrect Answer: Fast scram times can result from excessive accumulator gas pressure and can damage the Bellville washers. - Excessive accumulator gas would cause higher pressure to the P under area which would result in excessive speeds. This would cause damage to the BelIville washers.

The following distractors are incorrect as follows:

Worn lower stop piston seals. Damage to the CRD Guide tube. Incorrect. Scram speed will increase but the CRD Guide tube is not damaged.

Worn drive piston seals. Damage to the Collet Fingers. Incorrect. Worn drive piston seals result in slower scram times and the collet fingers are damaged by excessive withdrawal speeds.

Inadequate accumulator gas pressure. Damage to the index tube.- Only excessive pressures will damage the index tube.

NOH01 CRMECH-00 CRMECHE012 (R) From memory, describe the possible CRDM damage that could result from scramming a control rod too fast.

CRMECHE004 (R) Given various plant conditions, select those conditions that could potentially cause a CRDM to scram too rapidly.

Matea Re 1ton ue Facility Exam Bank j"o Modficato Mbth d: Significantly Modified o.3 Vision Bank QlD # Q54298 Tuesday, March 04, 2003 11:30:41 AM Page 41 of 58

_ m 35 A TIP machine is being retested when an instrument technician error causes actuation of the NSSSS Channel A manual isolation logic.

Which of the following describes the TIP system response (if any)?

1i No automatic actions occur when only one NSSSS channel manual isolation switch is actuated.

Ed The TIP detector will withdraw to its indexer, the TIP Shear Valve automatically fires to cut the detector cable and seal the guide tube.

T- he TIP Guide Tube Ball Valve automatically closes, cutting the detector cable and sealing the guide tube.

E] Tthe TIP detector w-il withdraw to its "in-shield" position and the TIP Guide Tube Ball Valves automatically close.

Asw d L B IE.a ] LuV1 Memory FaI Hope Creek Dael 02/24/2003 pant Systems Pe O 3 SROGroup 3 215001K604 215001 Traversing In-Core Probe K6. Knowledge of the effect that a loss or malfunction of the following will have on the TRAVERSING IN-CORE PROBE:

K6.04 Primary containment isolation system: Mark-l&ll(Not- BWR1) 3.1 3.4 rlIanat loof Justification:

The TIP detectors not in the "in-shield" position will automatically withdraw to their "in-shield" position and the TIP Guide Tube Ball Valves automatically close.

Correct- IAW HC.RE-SO.SE-0001, Section 3.1, Precautions and Limitations and HC.OP-SO.SM-0001, Table SM-017 The TIP detector will withdraw to its indexer, the TIP Shear Valve automatically fires to cut the detector cable and seal the guide tube. Incorrect - the Shear Valves must be manually initiated.

The TIP Guide Tube Ball Valve automatically closes, cutting the detector cable and sealing the guide tube. Incorrect - the Ball Valve will not close with the cable inside the valve.

No automatic actions occur when only one NSSSS channel manual isolation switch is actuated.

Incorrect - manual initiation of NSSSS Channel "A" will cause isolation of affected systems, including TIP.

NIXOre.NeN T :Sig;- <. 7 t HC.RE-SO.SE-0001, Section 3.1, Precautions and Limitations HC.OP-SO.SM-0001, Table SM-017; jiK~~i.>.i1 chtjtewnw LeMMgObet~

TIPSOOE006 (R) From memory explain the response of the TIP System following the receipt of an isolation signal from the Nuclear Steam Supply Shutoff System.

Mateijalaxaniiniodn :R -None

.Facility Exam Bank n Editorially Modified lsoSolC Vision Bank QID# Q5371 0 editorially modified due to correct answer was longest and most detailed answer.

Tuesday, March 04, 2003 11:30:41 AM Page 42 of 58

t 36 Given the following conditions:

- A core reload is in progress.

- Fuel Pool/ Reactor cavity level is steady at 1 inch above NORMAL water level.

- A fuel assembly has been grappled in the fuel pool and just raised to the NORMAL-UP position.

- The fuel bundle destination is 31-32 in the vessel.

The following occurs:

- Fuel pool level is recognized and confirmed to be LOWERING due to a leak on the out of service Fuel Pool Cooling pump discharge line.

- Normal FP makeup source is NOT available.

- The refuel floor ARM is NOT alarming.

How far will FP/Rx cavity level lower and what immediate operator action is required?

(Assume NO evaporative losses or operator actions taken)

El 6 inches; Isolate Fuel Pool Cooling because that is the source of the leak.

Ell 6 inches; Move the bridge over the reactor cavity because it is further away from the fuel pool.

Eil 9inches; Place the fuel assembly in the designated open rack location in the fuel pool because it is a safe location.

i1 9 inches; Suspend movement of the fuel assembly at its present condition because Core Alterations must be suspended.

A

__ ;c Ia L B . b aill Comprehension FiHope Creek E xia 02/24/2003 fi Plant Systems ] enili 3 SRO lT 3 233000A202 233000 Fuel Pool Cooling and Clean-up A2. Ability to (a) predict the impacts of the following on the FUEL POOL COOLING AND CLEAN-UP; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

A2.02 Low pool level 3.1 3.3

[tptiaati6ifl~ Justification:

ovCorrect: 9 inches; Place the fuel assembly in the nearest open rack location in the fuel pool because it is a safe location. 1 inch above normal water level is Elev 200' 1". The bottom of the Skimmer Surge Tank inlet pipe and FPCC Discharge pipe anti siphon holes is 199' 4". Water level will drop 9 inches and stop.

6 inches is based on 22' 2" above RPV flange from normal water level. Immediate operator action for lowering FP level are: evacuate the Refuel Floor; Return irradiated fuel assembly to the vessel or pool; Lower any bundle in the Fuel Prep Machine to the full down position. Since the IOA are bulleted, they can be performed in any order or simultaneously.

Incorrect: 6 inches; Isolate Fuel Pool Cooling because that is the source of the leak. Subsequent action.

Wrong level.

Incorrect: 6 inches; Move the bridge over the reactor cavity because it is further away from the fuel pool.

Wrong Level. Wrong reason. AB-COOL-0004 allows movement to either the FP or core, however the reason movement allowed to the core is to put the bundle down in a safe position.Tech Spec Definitions

1.7. Incorrect

Suspend movement of the fuel assembly at its present condition because Core Alterations must be suspended. Not an IOA. 10-0009 3.4.2 states "The RFS shall direct personnel performing CORE ALTERATIONS to place hoisted fuel or core components in a stable configuration and suspend subsequent CORE ALTERATIONS.

Tuesday, March 04, 2003 11:30:41 AM Page 43 of 58

HC.OP-AB.COOL-0004 HC.OP-IO-ZZ-0009 ABCOL4EO03 (R) From memory, recall the Immediate Operator Actions for Fuel Pool Cooling.

FPCCOOE005 (R) From memory, explain the methods used to preclude draining of the spent fuel storage pool, IAW the Fuel Pool Cooling and Cleanup (FPCCS) System Lesson Plan.

FPCCOOE016 (R) Summarize the immediate operator actions required for a Loss of Fuel Pool Inventory, IAW HC.OP-AB.ZZ-0144.

1Matera0iRquie , None IQonsouR;l New __ __ -=-s-- n M9dfi°tio <

IPles Qu Anew. Major revision to INPO BANK QID# 16907. Replace submitted question based on KAMM.

Tuesday, March 04, 2003 11:30:41 AM P Page 4o of 58 44

t 37 Given the following conditions:

- Core Alterations are in progress.

- The Reactor Mode Switch is in the REFUEL position.

- Control Rod Blade (CRB) 06-15 is on the Frame Mounted Aux Hoist.

- CRDM 06-15 is in the overtravel position with its position indication bypassed.

- Control Rod 30-31 is withdrawn for friction testing.

- The Standby Liquid Control Tank concentration is now reported at 13.5 percent with tank level at 4850 gallons.

- All other systems are operable.

Which one of the following actions (if any) are required?

El No - action is required.

[ Return CRB 06-15 to the control cell within one hour.

l Return SLC Tank within specification within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .I Insert control rod 30-31 within one hour.

Aw d Ex Me B Cgievl Application FacIyHope Creek Exmflatew 02/24/2003 Plant Systems 0 U G up 3. $I P, 3 290002K605 290002 Reactor Vessel Internals K6. Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR VESSEL INTERNALS:

K6.05 SBLC 3.3 3.4 Explanation of lJustification:

Correct: Insert control rod 30-31 within one hour. 3.9.10.2 is not applicable to friction testing since all 4 fuel assemblies surrounding the control rod would be in place. Therefore SLC must be operable with a rod 06-15 withdrawn.

Incorrect: No action is required. Must insert 30-31 within one hour.

Incorrect: Return CRB 06-15 to the control cell within one hour. Not required because 4 surrounding bundles are removed.

Incorrect: Return SLC Tank level within specification within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . Action time in Op Con 1 and 2.

Tech Spec 3.1.5 and 3.9.10.2

. ga . &j0 SLCSYSE025 (R) Given a scenario of applicable operating conditions and access to Technical Specifications:

a. Select those sections applicable to the Standby Liquid Control System, lAW HCGS Technical Specifications.

b. Evaluate Standby Liquid Control

[MO e ir ETech l 3r Spec sections 3.1 and 3.9 with 3.9.1 removed.

~IanQ4ceIJ New Qn io Tuesday, March 04, 2003 12:00:55 PM Page 45 of 58

De Number 0 38 Conditions are as follows:

A group of new PSE&G employees is currently at Hope Creek during your shift.

One of the group is a 36 year old who is an ex-radiation worker, and has completed an NRC-4 form with a total exposure of 20 Rem received prior to arriving at the Hope Creek Site, and no radiation exposure this calendar year.

He is badged for the site, has completed the GET and RWT courses.

For this individual, which one of the following would be the correct administrative dose limit?

Liz 1000 Mrem/yr. __

El 2000 Mrem/yr.

Li 3000 Mrem/yr.

[E14000 Mrem/yr .

.721b .i* t 'BMemory Faclitope Creek E.a 02/24/2003

'F: Generic Knowledge and Abilities l Gu 1 SGroup 1 294001G301 GENERIC 2.3 Radiological Controls 2.3.1 Knowledge of 10 CFR 20 and related facility radiation control requirements. 2.6 3.0 tanion of 2 Justification:

asw e yCorrect:Justification: IAW NC.NA-AP.ZZ-0024 Rev. 11 Attachment 1, the limit for a person with a life time dose of <2(N-17) is 2000 mrem, with the Radiation Protection Manager required to allow an increase to 3000 mrem.

NC. NA-AP.ZZ-0024 Rev. 11 Attachment 1 ADMPROE059 Given a set of exposure conditions Identify the personnel responsible for approval of the following dose extension:

a. Yearly Dose Extension

b. Declared Pregnant Women Dose Extension

c. Lifetime Dose Extension IAW NC.NA-AP.ZZ-0024:

Mattil Rquredfo Exm~atI 3/4 i~ None Facility Exam Bank Q o c l Direct From Source

= l VISION Bank QID#Q60666 Tuesday, March 04, 2003 11:30:41 AM Page 46 of 58

OUMsti~~wie 39W 39 Which one of the following meets'ALARA principles for performing a job?

El 1 man accomplishing the job in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months in a 60 mR/hr field.

l 1 man installing shielding for 30 minutes in a 60 mR/hr field and then accomplishing the job in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months in a 6 mR/hr field.

[J 2 men accomplishing the job in 25 minutes in a 60 mR/hr field.

[I 2 men installing shielding for 15 minutes in a 60 mR/hr field and then accomplishing the job in 25 minutes in a 6 mR/hr field.

Answer d EX l B on l Application FayHope Creek EmDate 02/24/2003 Tier: Generic Knowledge and Abilities ,C Group 1 wil 1 294001 G302 GENERIC 2.3 Radiological Controls 2.3.2 Knowledge of facility ALARA program. 2.5 2.9 rwinationo. Justification:

Anwe Li Correct: 2 men installing shielding for 15 minutes in a 60 mR/hr field and then accomplishing the job in 25 minutes in a 6 mR/hr field. 2(.25 hour2.893519e-4 days 0.00694 hours 4.133598e-5 weeks 9.5125e-6 months X 60mR/hr)+2(25/60 hour X 6mR/hr) = 35 mR TEDE Incorrect: 1 man accomplishing the job in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months in a 60 mR/hr field. (1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months X 60 mR/hr)= 60 mR TEDE Incorrect: 1 man installing shielding for 30 minutes in a 60 mR/hr field and then accomplishing the job in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months in a 6 mR/hr field. 1(.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months X 60 mR/hr) + 1( 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months X 6 mR/hr) = 36 mR TEDE Incorrect: 2 men accomplishing the job in 25 minutes in a 60 mR/hr field. 2( 25/60 hour X 60 mR/hr) = 50 mR/hr NC.NA-AP.ZZ-0024 HEAPHYE019 Define stay time and perform calculations to determine stay time or dose received.

.'.6_.4.

17_:'.5S.a;E: .;N'_':Wi9PoK,?:Y..._ _..bN.iR..>8H)kiK:'

t=s:.v>/z-K: a.//..; S..:=K'1A...z:fi:K.. l EXT Otln I None uel 9 l INPO Exam Bank 1,MMQ01l-to Method:

Direct From Source l INPO Exam Bank Q1D# 7593 11/04/1997 FitzPatrick

~110 __ .. ._ .-

Tuesday, March 04, 2003 11:30:42 AM Page 47 of 58

Quetin'Nrner 40 Given the following conditions:

- A worker with specific skills must enter a high radiation area to repair a leaking valve.

- This job is estimated to take a continuous exposure of 1.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months in a 200 mrem/hr gamma field.

- Current dose for the year is 2725 mrem.

- The worker is 33 years old and has received a lifetime dose of 34.4 REM.

Who must approve this dose limit extension needed to complete the task and to what new dose control level?

(Assume NO delegation of authority)

I1VP-Operations; to 4000 mr/yr TEDE.

ElI Ra-d iat-ion--P-r-o-t-ectio-n -S-u--pervisor; to 3-0060_mr/yr TEDE.

C, Op erat i o n--s- Man-a-g er-;- to- 30 00 mr/y r TE DE.

El -Rad-iation Protection Manager; -to 4000 mr/yrTTED E.

Anwra ExamLPVq 1B CgtleLvlComprehension Fiit:Hope Creek ExmDt: 02/24/2003 Te;Generic Knowledge and Abilities ROGop 1SOGop 1294001 G304 GENERIC 2.3 Radiological Controls 2.3.4 Knowledge of radiation exposure limits and contamination control, including permissible levels in 2.5 3.1 excess of those authorized.

Justification:

Correct: VP-Operations, to 4000 mr/yr TIEDE. VP needed for 4000 mr/yr extension because > 2(N-1 7) lifetime dose exceeded. Estimated dose to complete the job will exceed 3000 mrem therefore extension above 3000 is needed.

Incorrect: Radiation Protection Supervisor; to 3000 mr/yr TEDE. Incorrect approver. Incorrect limit.

Would be approver for current limit if <2(N-1 7) lifetime.

Incorrect: Operations Manager, to 3000 mr/yr TEDE. Incorrect approver. Incorrect limit.

Incorrect: Radiation Protection Manager; to 4000 mr/yr TEDE. Incorrect approver. Would be approver for needed limit if <2(N-17) lifetime.

' 'flvlj '

l : I -, , I I '-- ; , I I I 1

NC. NA-AP.ZZ-0024 Attachment 1 2(N-1 7) Lifetime Dose Action Level ADMPROE059 Given a set of exposure conditions Identify the personnel responsible for approval of the following dose extension.

a. Yearly Dose Extension

b. Declared Pregnant Women Dose Extension C. Lifetime Dose Extension lAW NC.NA-AP.ZZ-0024:

Mat I~qii~

$~r ~nii~t l.)U None QtUetioSbburce:5 Facility Exam Bank Qustl Miiiaton Method: - ditorially Modified Questionsoiia VISION Bank QID# Q55936 Modified to remove delegation to other approvers and to make 3000 limits plausible.

Tuesday, March 04, 2003 11:30:42 AMPae4of5 Page 48 of 58

m 41 Which one of the following is the maximum permitted backround count rate on a frisker prior to use, and the minimum count rate above backround that indicates the contaimination limit has been reached?

Max Backround Contamination Limit H, 100 cpm 100 cpm above backround 1100 cpm 300 cpm above backround E 300 cpm 100 cpm above backround 300 cpm 300 cpm above backround r c RxaPL _ B I l ,gRi e Memory Fclj Hope Creek lEmte:g 02/24/2003 Tir Generic Knowledge and Abilities 1 $RO 1 294001 G305 GENERIC 2.3 Radiological Controls 2.3.5 Knowledge of use and function of personnel monitoring equipment. 2.3 2.5

,,,,ai ;If greater than 300 cps, find another frisker or notify RP tech. Indication of contamination is 100 cps

,,ionf above initial backround reading.

.O lAustin' =,a, 101111 4k.WJleferene T le§g .§ W Radiation worker training handout material 5 IBM= =_ f= i ti ON

-- - - -- ... -- - - ; a-- , - 4'a a ---

ateial ei forExamination 4l None

~Queston~aure: Cth.r Pnn'ilitxY Mtf Direct From Source Peach Bottom 2002 LSRO Exam question 4-5 unmodified. Hope Creek has the same limits.

Tuesday, March 04, 2003 11:30:42 AM Page 49 of 58

42

_Q6#6iordiber Which one of the following limitations of HC.OP-ST.KE-0001 prevents overexposure to Refuel Platform workers when a fuel bundle is removed from the RPV?

Ai- Hoist uptravel with a fuel bundle is stopped 6 feet below the water surface.

Aux SThe Main Fuel Grapple must be used to remove a fuel bundle.

il Maximize the amount of water shielding between the fuel bundle and the reactor vessel wall.

E} Minmize the time the fuel bundle in the Drywell Bellows Area.

Answrb EaneeB C itv Memory FaiIHopeCreek xaD 02/24/2003 Generic Knowledge and Abilities l R rp 1 O lp 1 294001G310 GENERIC 2.3 Radiological Controls 2.3.10 Ability to perform procedures to reduce excessive levels of radiation and guard against personnel 2.9 3.3 exposure.

ln i Justification:

Incorrect: Aux Hoist uptravel with a CRB is stopped 6 feet below the water surface.Fuel bundles not allowed to be moved with the Aux hoists. Uptravel limits switches and mechanical blocks are set to prevent a Control Rod Blade from being lifted to within 6 ft of normal water level for personnel protection.

Correct: The Main Fuel Grapple must be used to remove a fuel bundle. Main fuel grapple is required to be used for fuel movement.

Incorrect: Maximize the amount of water shielding between the fuel bundle and the reactor vessel wall.

Concern for personnel in the Drywell.

Incorrect: Minmize the time the fuel bundle in the Drywell Bellows Area. Concern for personnel in the Drywell.

= ,:St i o 0 e'T '

HC.OP-ST.KE-0001 HC.OP-FT.KE-0002 UFSAR 9.1.4.1 REFUELE01 1 From memory, identify the only grapple which may be used to move fuel in the reactor vessel or spent fuel pool lAW Technical Specifications.

lM~aterial Requiti.foE:9 , : I$None lQup^l 'New Que 9!t ettI5S.

QuesionSOce~ Comns Tuesday, March 04, 2003 11:30:42 AM Page 50 of 58

Delayed neutrons are neutrons that:

II have reached thermal equilibrium with the surrounding medium.

EIi are born within 10E-14 seconds of the fission event.

are produced from the radioactive decay of specific fission fragments.

I01 are responsible for the majority of U-235 fissions.

l F *,i [B ] }. IMemory a Hope Creek §fm 02/24/2003 i Fundamentals = 01 . : 1 :°l 292001 K102

.. ........... ~. .... ..

292001 Neutrons ... ..-

K1.0 K1.02 lDefine prompt and delayed neutrons. .0 3.1

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

....=E TE D

== _1 m KINETIE003 Describe the production of delayed neutrons.

'None I. .....

.. .................... - - - .. . ~

1 I

'NRC Exam Bank Direct From Source 1MR, s BWR GFE BANK QUESTION IDQ: B1 945 (P845)

Tuesday, March 04, 2003 11:30:42 AM P Page 51 of 58 51

o 45 Refer to the reactor response curve attached (Q 45) for a reactor that was initially stable in the source range.

.A momentary control rod withdrawal occurred at time = 0 sec.

The response curve shows

~~~~~~~~~. ...........

versus time for a reactor that was initially

] reactor period; subcritical. ___ __ . __

E reactor period; critical.

El reactor fission rate; subcritical.

El reactor fission rate; critical.tica. -__

o I 4 B [b X _ y Comprehension i Hope Creek a e 02124/2003 0° a 0 9f9 0 W Fundamentals 292003K107 292003 Reactor Kinetics and Neutron Sources SK1.0

.3 3.3 iK1.07 - .. ...... prompt critical,

!Explain .

prompt jump, and prompt f....drop.

KINTIOO- Ea pomp KINETiE007 lExplain promptecritical, prompt jump, and prompt drop.

Attached figure from Question GFE..............--..

B3250 NRC Exam Bank Direct From Source

- - 6 NRC BWR GFE Bank Question ID: B3250 (P3249)

Tuesday, March 04, 2003 11:30:42 AM Page 53 of 58

~

Ih rA 46 Compared to beginning of core life, the Doppler coefficient of reactivity is egative r14 pat end of core life due to (Assume the same s.. .. ................... .......... .... .

initial .....

fuel temperature.) ..........

Wi ess; depletion of U-238.

? more; depletion of U-238.

S less; buildup of Pu-240.

Wmore; buildup of Pu-240.

MD B Hope Creek 02/24/2003 Aw d Ep.e l liMemory Fundamentais Op $R 1 p 0 292004K1 05 292004 Reactivity Coefficients

. ............... ... .................. ....................... ..... Reactivity Coefficients K1.0 KK1.05 Define the doppler coefficient of reactivity. .9 2.9 1-..... ... - - , , . , . '-. I --- ... ....... - I .- I - . -. I..I-11 I.." I I BWRTHRE01 1 Explain the doppler coefficient of reactivity.

- ~ ~ ~ ~ ~ ...:::::::: .... ::::::::::: :::::::: :::: :::: :::

.None NRC E "a Exam _ Bank N_ B R G 1IIl.1x Bak Qr IQJ MW

$Am meitCh I EditcLrially Modified gNtt 2 ,Ij.. ~.11,111I I - I111 I . . . .. . .. ... .... '

NRC BWR GFE Exam Bank Question ID: B131 33 Modified 1AW A. Blarney comments.

Tuesday, March 04, 2003 11:30:43 AM Page 54 of 58

Which one of the following, if decreased, i............ ....

will affect not................

Keff?

H Fuel enrichment.

IHi Control rod worth.

'Neutron contribution from neutron sources. ____

Ii Shutdown margin when the reactor is subcritical. _ __

c B Comprehension FI XHope r creek 02/24/2003 IR Fundamentals - I l or R Group o 292002K108 292002 Neutron Life Cycle K1.0 _I I

_1 ......._._..

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

K1.08 i Define effective multiplication factor and discuss its relationship to the state of d reactor. .7 2.8

~~~.. ..... ...A..... . ;;.,,.------------....

.1..I . 11 . - . -, . : 7:7 7- , =7 77 _.:-- - - - --

I . . I I . -. 1-1 I-- 1.-...---.... .1...-. 1-1

... II.. I I-- --- ...I.......I-. .. 1.1. .1 I . I II -- ... I.. 11 . II -

OR NEULIFE004 I Define K excess Jose!

:1:::: :::: -- .~. .. . _ 1 _

.None - -

NRC Exam Bank S mot-cm____ lDirect From Source NRC BWR GFE Exam Bank Question ID: B348

. 1.-I..."I I 11 .-I ...

Tuesday, March 04, 2003 11:30:42 AM Page 52 of 58

47 A reactor has been operating at 100% power for several weeks when a reactor scram occurs.

How much time .will be required .....

for ...core heat production to decrease to 1% following the scram?

Gi1 to 8 Days. -= :D;f .

E 1 to 8 Hours.

id 1to 8 Minutes.

WL 1 to 8 Seconds.

FARRwe b ..... RW M B I Me~Jm ory ... ........ E W Hope Creek 02/24/2003 Ter:Fundamentals 0 0 292008K1 30 292008 lReactor Operations K1.3 Normal Reactor Shutdown K1.30 Explain the relationship between decay heat generation and: a) power level history, b) power .2 3.5 production, and c) time since reaction shut down. _

g ,....... "" - -......

............ - , - . - .- I ~

I .. - - - -.....-.1-1II I . I . .....I 1I- ~~- I - .. .I

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

- II II--I -11 I I~ I~ I~ ~ 1.-.1 RXOPERE031 Explain the relationship between decay heat generation and

a. power level history

b. power production

c. time since reactor shutdown __ _ . _ _ _ _ _ _ _

I. ::.::'.:::::::::::::::::::::::::::::::::::::::

-:::I:::::

' 2None NRC Exam Bank

.... ~~~~~~~~~......... ..........

........ .Di rect From Source fNRC BWR GFE Exam Bank Question ID: B2272 (P572)

Tuesday, March 04, 2003 11:30:43 AM : Page .,.I 1.

P 55 of 58 I....I..I..

048 Refer to the attached drawing (Q 48) of four sets of centrifugal pump operating curves. Each set of curves shows the results of a change in pump/system operating conditions.

Two identical constant-speed centrifugal pumps are operating in parallel in an open system when one pump trips.

Which set of operating curves depicts both the "before" and "after" conditions described above?

E~~ ....

~2 _ ___ _ _ _ __

A a ]A ll !B 1 IZ& Comprehension '#Hope Creek 02/24/2003 T Fundamentals l 01°ol 9GloEip 0 2930 06K1 13 293006 Fluid Statics K1. 1 Pumps and Pump Characteristics K1.1l --Explain the results of putting centrifugal pumps in parallel or series combinations. .6 2.7 c = >44 'MM Hi MN I PUMPSOE013 I Describe the operation of centrifugal pumps in series and in parallel arrangements.

I! : ::::::::::::::: :: :: ::::::::::::::::::::::

T Figure of pump curves for GFE Bank question B2279 NRC Exam Bank IQ Diirect From Source Qlap$J _ D- hn . lt NRC BWR GFE Exam Bank Question ID B2279 (P1 524)

Tuesday, March 04, 2003 11:30:43 AM Pg56o5 Page 56 of 58

49 XWhich one of the following conditions must occur to sustain natural convection in a fluid system?

R Subcooling of the fluid.

'A phase change in the fluid.

enthalpy change in the fluid.

. -An I1 Radiative heat transfer to the fluid.

AnOVe c lF i B Memory 'l. Hope Creek 02/24/2003 Fundamentals 0 2930M3K1 06 293008 Thermal Hydraulics K1 0 Pool Boiling Curve (T vs. Q/A)

K1 06 Define a natural convection heat transfer. .5 2.6 101%"I"'MllC,?"-.,IlN&W III III AL THRMHYE008 Define natural convection heat transfer.

Mate~

e1ir ~

EgiitO al ,None NRC Exam Bank ld uIt 0 t od: Direct From ISource iNRC BWR GFE Exam Bank Question ID:B387

...-- ....I............11- -11I I Tuesday, March 04, 2003 11:30:43 AM I P Page 5o of 58 57

II W .50 i Brittle fracture of a low-carbon steel can only occur when the temperature of the steel is the nil ductility temperature, and will normally occur when the applied stress is the steel's yield strength (or yield stress).

I g-reater than; greater than ___ ___ ___

1 greater than; less than ____ _

Eqless than; greater than Kless than; less than nd ] _ [B: Memory ci Hope Creek 02/24/2003 E Fundamentals 293010K101 293010 Brittle Fracture and Vessel Thermal Stress iK1.0 K1.01 State the ~. brittle fracture mode of failure. .4 2.8 A. W_

MEMO-

'El"rLol W Mm HE R BRITTLE005 1 State the brittle fracture mode of failure.

= :::::::::::::::::::::::::: I : :: :

- ,- ::::::::::1--

_None NRC Exam Bank f1 tt o dM Doo From Source iirect IQ ~ l NRC BWR GFE Exam Bank Question ID:B2499 (P2496)

Tuesday, March 04, 2003 11:30:43 AM Page 58 of 58

Q1 10Y412 Panel Load List FunctLocation Description Planning plant H1BD -10Y412-07 LVL SWCST 1LISN-035A & E NNUC H1GM -10Y412-06 PNL,D/G AREA PNL RM SUP 1BC486 NNUC H1PG -10Y412-01 HTR,SPACE SUB 10B420 NNUC HIPG -10Y412-05 HTR,SPACE MCC 10B421 NNUC H1SB -10Y412-02 VERT BD,RPS LOGIC 10C611 NNUC H1Z7 -10Y412-04 SPARE NNUC H1ZZ -10Y412-08 SPARE NNUC H1ZZ -10Y412-09 SPARE NNUC H1 ZZ -1OY412-10 SPARE NNUC H1ZZ -10Y412-1 1 SPARE NNUC H1 ZZ -1OY412-12 SPARE NNUC HIZZ -10Y412-13 SPARE NNUC H1ZZ -10Y412-14 SPARE NNUC H1ZZ -10Y412-15 SPARE NNUC H1ZZ -10Y412-16 SPARE NNUC H1ZZ -10Y412-17 SPARE NNUC H1ZZ -10Y412-18 SPARE NNUC

Q0 RAISE/LOWER LEFT/RIGHT FORWARD/REVERSE

(

ENGAGE/RELEASE I

( I

(

4 I BACKUP HOIST LIMIT FUEL HOIST INTERLOCK 1

MONO AUX HOIST INTERLOCK TROLLEY AUX HOIST INTERLOCK ROD BLOCK INTERLOCK NO. I

...I.

ROD BLOCK INTERLOCK NO. 2-BRIDGE REVERSE STOP NO. I BRIDGEJ L

REVERSE STOP NO. 2

[ FAULT FAULT q f 7, INTERlLOCK SlATUS DISPLAY

-0

Q 45 I

LII1

Q 48 pimp PuW HEAD HEAD FLWjRATE F~T 2LC11RS T:<

PUMP Rmp HEAD HEAD R"#]Ws 13.1 w4 FLOW RATE CENTRIFUGAL PUMP CURVES

ML031130318

Text

Reference:

1.7. Incorrect

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