Text

Draft - Lsro Written
	ML031130141
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: ML031130141 (56)
	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 10Y412 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 ml 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 .

may I- may continue because the panel feeds Non-1 E loads.

El must be suspended because all SRM drives lose power.

must be suspended because the panel is required by Tech Specs.

AnswHpd [Application l Hope Creek emi[B 0 B p : 03/10/2003 ier Emergency and Abnormal Plant Evolutions ] 2' l rl 1 295003A101 295003 'Partial or Complete Loss of A.C. Power = -

1 AA1. Abilitytooperateand/ormonitorthefollowingastheyapplytoPARTIALOCOMPLETE LOSSOFA.C.

POWER:

AA1.01 A.C. electrical distribution system 3.7 3.8

... ....... ... . .. . . ........... .... ... ... ... I... ... .

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

A- incorrect; electrical spec in OP Cond 1, 2, & 3 1B- incorrect; Panel contains IE loads C- incorrect; 10Y208 powers SRM Drive Cabinet. SRMs not listed in 10Y412 Load list.

TS 3.8.3.2 4 IIMMIUM q !gir 1EACOOE028 Given a scenario of applicable conditions and access to Technical Specifications:

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

b. Evaluate 1E AC power operability and de Mta M Load Reference table for 10Y412; Tech Specs without Definitions, Bases, or

.Admin Section.

MM. New B y dI tX9 0 II - . I-- I - -- I--- 1- I ... I II ... II Wednesday, January 22, _2003 2:56:20 PM I Page 1 of 56

Given the following conditions:

- The plant is in Operational Condition 5.

- Core off-load is in progress.

- All control rods are fully inserted.

- 1BJ483 Inverter to the Main Control Annunciation cabinet malfuctioned during an electrical transient.

- The inverter output was restored via manual transfer.

- The Control Room reports that most Control Room Annunciation is no longer working.

Which one of the following Control Room overhead alarms, if inoperable, would require suspension of Core Alterations?

d SRM UPSCALE.

Il FUEL POOL COOLING SYS TROUBLE._

Ei SLCTANK TROUBLE.

2ii TIP SHEAR VALVE CLOSED/INOP.

a x L e B Ae

.Application 00 i' Hope Creek 02/24/2003 ex Emergency and Abnormal Plant Evolutions J l 21 I0 ! 1 295003G432 295003 Partial or Complete Loss of A.C. Power 2 2.4 Emergency Procedures and Plan 2.4.32 Knowledge of operator response to loss of all annunciators. 3.3 3.5 E*10fahlbfl Justification:

I 'fl t'J IdCorrect: SRM UPSCALE. Tech Spec 3.9.2.a requires SRM annunciation in the control room. With the requirements not met, suspend all operations involving core alterations.

Incorrect: FUEL POOL COOLING SYS TROUBLE. Fuel Pool Cooling system malfunctions listed under

.this annunciator would not require suspension of core alterations.

Incorrect: SLC TANK TROUBLE.SLC malfunctions listed under this annunciator would not require suspension of core alterations.

Incorrect: TIP SHEAR VALVE CLOSED/INOP. TIP malfunctions listed under this annunciator would not require suspension of core alterations.

Tech Spec 3.9.2.a

.......... a

- _ A, _ t t AiH

.'21.d o 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.

6'teria M Requ~d Ž 0 g KJ Tech Specs without Definitions, Bases, or Admin Section.

A li0~d~ New = 00 = l pft.On Ifi 6 y ho l ie Wednesd, J y 2, 23.2 0 PM g 2 o 56 Wednesday, January 22, 2003 2:56:20 PM Page 2 of 560

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:

SRM A 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 of loading the remaining bundles?

(Assume all 10 bundles have equal reactivity worth.)

One more bundle will cause a local criticality.

I Inserting all remaining bundles will cause a local criticality.

E SRM-i "B" will indicate 500 cps when the core is fully loaded.

El SRM "B" will indicate 800 cps when the core is fully loaded.

Arser b RExm B JWMPiee Comriprehension FaiyHope Creek I 02/24/2003 Tier, Emergency and Abnormal Plant Evolutions ! p 1 l ol 295014A201 295014 Inadvertent Reactivity Addition L r u. 3 AA2. Ability to determine and/or interpret the following as they apply to INADVERTENT REACTIVITY ADDITION:

AA2.01 Reactor power 4.1 4.2 lanaioof Justification:

Correct: Inserting all remaining bundles will cause a local criticality. The first doubling was reached with 2 bundles. Step 1154 would be 400; Step 1156 would be 800; Step 1158 would be 1600; Step 1160 would be 3200 or 5 doublings at which time the reactor would be critical.

Incorrect: One more bundle will cause a local criticality. One more bundle would be < 400 cps or less than 2 doublings; Not critical.

Incorrect: SRM "B" will indicate 500 cps when the core is fully loaded. Value if counts added linearly. At least 3200 cps by calculation.

Incorrect: SRM "B" will indicate 800 cps when the core is fully loaded. Value at 3 doublings. At least 3200 cps by calculation.

FRO a.ftI Reactor Theory RXOPEREOO9 Describe how to determine if a reactor is critical.

I XW94"e~&EitW', ;None Wednesday, January 22, 2003 2:56:20 PM Page 3 of 56

sINPO Exam Bank ___t_ __ 1____ :Significantly Modified Sor V INPO Bank QID 14536 09/13/1996 Peach Bottom Wednesday, January 22, 2003 2:56:20 PM Page 4 of 56

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?

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

pal Non-Coincident UPSCALE scram from only 1 SRM channel. __ _

El Coincident UPSCALE scram from at least 2 SRM channels.

L1 Coincident UPSCALE scram from only 1 SRM channel.

7 Anwe b _L'B . Comprehension l Hope creek a pate: 02/24/2003 T Emergency and Abnormal Plant Evolutions 0 90 bl 1 $ oPR 1 295014K205 295014 Inadvertent Reactivity Addition R 4rd..... 4 AK2. Knowledge of the interrelations between INADVERTENT REACTIVITY ADDITION and the following:

AK2.05 Neutron monitoring system 4.0 4.1

[iati Justification- Correct- Non-Coincident Scram from only 1 SRM channel. RPS Shorting links must be removed for a Non-Coincident scram from any 1 NI Channel to withdraw a control rod with SDM less tthan 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 l.=':S.~~*~~.~7 MEO

'L E.Q".&.,'&i'&

c

. _IUN,

_ _m MNCAIN

.1 .. ...

Tech Specs 3.9.2 /4.9.2.d Tech Specs Table 3.3.1-1 Footnote ( C)

Tech Spec Bases 3/4/3.1 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.

MaeralRquiedfo Eahlii at Tech Specs without Definitions, Bases, or Admin Section.

I E~sti~~Iure:&INew _________ uslnjfctdKehd 19 I Wednesday, January 22, 2003 2:56:21 PM Page 5 of 56

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?

( Use attached Table Q 5 for valve nomenclature)

I. KA-PI- 4610A

11. KA-PI- 4610B Ill. KA-PI- 4610C IV. KA-PI-4610D I and I._

Iland Ill.

i II and Ill.

101 Ill and IV.

RAnse a B I {Application a Hope Creek Cr.ee Ex>mPit 02/24/2003 Tite Emergency and Abnormal Plant Evolutions X 9 U 3 Surpl 1 295023A204 295023 Refueling Accidents Rebr Numbe 5 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 lExilnation ofg.l 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- 461 OC and KA-PI- 461 OD 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-0013 Attachment B-5 FPCCOOE004 (R) From memory, describe/explain how leakage is detected from the spent fuel storage pool, dryer-separator storage pit, reactor well and fuel shipping cask pit liners, lAW the Fuel Pool Cooling and Cleanup System (FPCCS) Lesson Plan.

Maeiai"0 iequlrd fP&lD M-53 sheet 2; Provide SAP valve list from H1KA KA-V300 to H1 KA tKA-V387 1io til Facility Exam Bank j Mttd:< Direct From Source Wednesday, January 22, 2003 2:56:21 PM Page 6 of 56

S2002 Quesio LSRO Requal exam.

Wednesday, January 22, 2003 2:56:21 PM Page 7 of 56

The unit is in OPCON 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.

Which of the following contains a makeup source (OTHER THAN hoses on 201' Reactor Building) to the Fuel Pool/Refueling Cavity that requires component manipulation from outside the Control Room ONLY?

EliIFire Water.

El Service Water.

Hl Demineralized Water.

E~Condensate Storage & Transfer.

Anwrd xmylB ContveL~lMemory ~ cIt:Hope creek ExmDt: 02/24/2003 Te:Emergency and Abnormal Plant Evolutions3 1253G5 295023 Refueling Accidents 6 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. ______________

Answer 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 - Demnineralized Water - can oniy 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 FPCCOOEOO(R) Concerning spent fuel storage pool water level, summarize, from memory, the following lAW the Fuel Pool Cool ing and Cleanup (FPCCS) System Lesson Plan:

a. How normal level is controlled

b. Sources of makeup to the spent fuel storage poo IMatra 40MN"a00M $i77 None Facility Exam Bank t~ MiMaifi"atoi A~i~ Editorially Modified QuesionSour~C~ment;~< vision Bank QlD# Q53948 Modified to disallow hoses on 201 el.

Wednesday, January 22, 2003 2:56:21 PM Page 8 of 56

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 from the core?

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

EIopened because RHR pump min flow will be open.

closed because the RHRflow will be at pump runout.

El closed because the RHR flow will be bypassing the core.-

ne d E el B ill ! V I4 Comprehension ll Hope reek lExi ae 02/24/2003 Tier: Emergency and Abnormal Plant Evolutions . IPI 2 SRGr 2 295001A101 295001 Partial or Complete Loss of Forced Core Flow Circulation R r e 7 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 oIf Justification:

Correct answer: closed because the RHR flow will be bypassing the core. The 'B' Recirculation Pump 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 Wednesday, January 22, 2003 2:56:21 PM P Page 9 of 56

Matril EqUaminatiDn8 None-I New Q~st~n~oifi4ionM~th~I:-

Qi Sucommo"nts Wednesday, January 22, 2003 2:56:21 PM Page 10 of 56

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?

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

El1 SSW Loop B Yard Dump valve EA-HV-2356B spuriously opens.

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

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

Answerg d EPamR Levl B 0 0 l lComprehension 0al llHope Creek Exam Dae: 02/24/2003 TEmergency and Abnormal Plant Evolutions 09 A 2 loll 2 295018A201 295018 Partial or Complete Loss of Component Cooling Water Reord Number 8 AA2. Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER:

AA2.01 Component temperatures _ ___ 3.3 3.4 1'toi "

fIai~#0"' Justification:

1#44 M:J 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.

P&ID M-1 1 sheet 1 and M-1 0 sheet 2

::: : :~ :: Lea : :: : : :::

FPCCOOE01 5 (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 HP&ID M-1l sheet 1 and M-10 sheet 2 IQtn$ Ic :S

__ - _: t in :New Qusto SouroeCm ents:00 -

Wednesday, January 22, 2003 2:56:21 PM ; P Page II of 56

Given the following conditions:

The plant is in Operational Condition 4, when the following alarms actuate in the Control Room:

- RACS REMOTE CONTROL PNL 0C202 for RACS HEAD TANK HI-HI LEVEL.

- RACS RMS ALARM.

(Alarms have been verified to be valid.)

Based on these conditions:

Ii RACS is leaking into RWCU.

El RWC i-s leaking into RACS.

El RACS is leaking into Service Water.

2I Service Water is leaking into RACS.

ArR b E1al IB Xi Comprehension 'Hope Creek Exam.ata: 02/I /4/2003 er:Emergency and Abnormal Plant Evolutions l . pl 21 l rl 2 295018K1101 295018 Partial or Complete Loss of Component Cooling Water _ 9 AK1. Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER:

AK1i01 Effects on component/system operations 3. 5 3.6 Eaplaofs, IJustification:

MOwrjjj 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.

HC.OP-AR.ZZ-001 1 Attachment C1-8 RAGSOOE013 (R) Given M-13-0 and M-13-1 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 Mtei.,l~qirdoErnatn J'g& M-13-0 =f fL- f-f fIl iNPO Exam Bank _"Mifi M Mth Editorially Modified l; SINPO Bank QID# 7975 03/14/1997 Hatch. Modified for Hope Creek Wednesday, January 22, 2003 2:56:21 PM Page 12 of 56

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?

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

Eil Maximize RWCU bottom heard drain flow._

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

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

Answer a Exa B 'eIt Application a Hope Creek Exm 02/24/2003 Tier; Emergency and Abnormal Plant Evolutions ROGr 3 S[.Gru 2 295021A104 295021 Loss of Shutdown Cooling Reodume 10 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 e Unw=owJustification A040 Adz Crosstie "C" or "D" RHR pump for heat removal. correct - RPV-0009 subsequent action E, RHR Pumps C & D may be realigned to provide alternative decay heat removal.

Restart the Reactor Recirculation System for heat removal. -incorrect- pe4 subsequent action C RRPs do not provide heat removal to maintain less than 200 degrees, only circulation.

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.

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

HC.OP-AB.RPV-0009

~~ ~ ~ -- - --: --

D - 7 DAT ---- TTT -D ;T i,,~il>E¢, ggEl MI 02;SSsj00 ME anmkt~ i .c zR .;

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

Mtrarniato None Facility Exam Bank to ibif etbd:- Significantly Modified S< VISION Bank QID# Q61332 Sig Mod Wednesday, January 22, 2003 2:56:21 PM Page 13 of 56

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.

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

Rod 30-31 must be...

IE uncoupled.

[I withdrawn.

1 scrammed.

I1 disarmed.

Aser. c IEa I B ltyel Comprehension Fa 1l Hopecreek O 02/24/2003 TKer Emergency and Abnormal Plant Evolutions ROZr 2 R p' 2 295022K102 295022 Loss of CRD Pumps Number 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 Etlihatione Justification:

Correct: Scrammed. Any control rod withdrawn unless under TS 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- uncoupled. Rod needs to be at notch 48.

Incorrect- withdrawn. Can not withdraw without a CRD pump running.

Incorrect- disarmed. Action for Op Con 1 and 2.

Tech Spec 3.1.3.5 CRDHYDEO33 (R) Given ascenario 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.

lMateriaRequfredfo ~piTech Specs without Definitions, Bases, or Admin Section.

lCU QbO6e: New _oI _ethe Qu on- ,

uc 1ueCon IRi=---ts:

Wednesday, January 22, 2003 2:56:22 PM Page 14 of 56

Given the following:

- 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 K New Fuel Vault; line-of-sight to the refueling cavity. _

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

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

W1 Reactor Building Exhaust; theducts above the refueling cavity.

RAnswe b I B . kL Y Comprehension tHope Creek am Date 02/24/2003 Tieir Emergency and Abnormal Plant Evolutions I2 SOFufl 2 295033A101 295033 High Secondary Containment Area Radiation Levels rN 12 EA1. 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 Expla Atipp7 Justification: Spent Fuel Pool Area rad monitor activates the evacuation siren on the wall opposite the elevator. 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 IAW the Radiation Monitoring System Lesson Plan if a high radiation level is indicated for:

a. Main Steam Lines

b. Liquid Radwaste Monitoring c.

lMatrilReured#fr quoiElN None 0X0 0 00 00X0 IFacility l : ' Exam Bank I n : . iDirect From Source QuesionSource me VISION Bank QID# 056244 Wednesday, January 22, 2003 2:56:22 PM Page 15 of 56

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

El An LPRM removal using the LPRM removal tool.

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

lilA TIP detector withdrawn into the TIP Drive Mechanism.

IEl A Control Rod Blade unlatched by theCRB renmoval tool.

e c gI -Comprehension lExmB F0a IHope Creek 5 0a. ae: 02/24/2003 Tier: Emergency and Abnormal Plant Evolutions lo ll 2 Rr 2 295033K304 295033 High Secondary Containment Area Radiation Levels Recor Number 13 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 4.4

[Eiaptionpf,1l Justification: Correct: A TIP detector withdrawn into the TIP Drive Mechanism. High radiation source is outside 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

- 4 RRceIIHRe 1 , F&. I TlPS00E09 - (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 M0ena1 frEx~tin1 None ElQions~urce: New - Q~ o' Moifa o_ Mthd:

Questn# Sori CoInm Wednesday, January 22, 2003 2:56:22 PM Page P 16 of 56

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?

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

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

(II To prevent contamination of normal ventilation ductwork.

[I To allow accurate monoitgoroingeofareleatseetoetheenvironment.

Anser b Ea l B Ql GP lMemory F Hope Creek a .ae 02/24/2003 i Emergency and Abnormal Plant Evolutions U r Pl U 2 OGo 2 295034K301 295034 Secondary Containment Ventilation High Radiation Rc Number14 X

EK3. Knowledge of the reasons for the following responses as they apply to SECONDARY CONTAINMENT VENTILATION HIGH RADIATION:

EK3.01 isolating secondary containment ventilation 3.8 4.1 of Justification:

Answer 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.

Tech Spec Bases 3/4 6.5 j~if~vggl}U^Sggi, gF

..i.00t'-'b'll =- iearn Objetvs dd v - . ........

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 None MateiaReird for E~ainai :17.> lNodne o INPO Exam Bank =W Qdest M etNW Editorially Modified QuestionSot& nt; ,.INPO BANK QID#18083 10/16/1998 Pilgrim Wednesday, January 22, 2003 2:56:22 PM Page 17 of 56

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:

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

must be approved by the Operations Manager prior to the action taking place.

Elmust be approved by -a icensed SRO on the operating shift prior to the action taking place.

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

Anwr c 1 B l Memory lciI3tHope Creek Eag o: 02/24/200 3 V Emergency and Abnormal Plant Evolutions Rr 3 SRO o 2 295035G101 295035 Secondary Containment High Differential Pressure eor Nmr 5 2.1 Conduct of Operations 2.1.1 Knowledge of conduct of operations requirements. 3.7 3.8 Expanationwofl 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) 7 - T7X

--- V a DD- 7;: - - -- -n------- T : _ = X T 7 ADMPROE007 From Memory Explain the circumstances and approval required for Licensed Operators to deviate from Technical Specifications or license conditions. IAW NC.NA-AP.ZZ-0005 and 10CFR50.54(x),and SH.OP-AP.ZZ-01 02 MaeallReqired for Exainato 77I. None l lFacility Exam Bank Qustoo Meathod:Direct From Source i

Qeto SQice ti 557018 Wednesday, January 22, 2003 2:56:22 PMP Page 18 of 56

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?

EIl CRS.

~O. . _. _ _ _ _ _. .. __.. _.

ES RO/PO.

M1 STA.

Answer b Exa veI B 900git* L Memory ltYHope Creek Exam 02/24/2003 Emergency and Abnormal Plant Evolutions 33 R r 2 295035G104 295035 Secondary Containment High Differential Pressure e Nubr16 2.1 Conduct of Operations 2.1.4 Knowledge of shift staffing requirements. 2.3 3.4 4glnatoneof Generic KA FORCED to Procedures section on purpose IGNORE 295035 K/A Title.

1*,weyr ~M']-1Justification:

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 NC.NA-AP.ZZ-0005, rev 1 Section 5.14, attachment 9 Par 5 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 MateiallReui rTech Specs without Definitions, Bases, or Admin Section.

FacilityExamBank = 0 Direct From Source Qtuestioii Sour~ce Corn ~it~jVISION Bank QID# Q54277 Wednesday, January 22, 2003 2:56:22 PM Page 19 of 56

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 would be a violation of the Hope Creek Operating License?

1-i44 full crates stored on top of each other.

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

i1 2 new bundles in the inspection stand and one suspended from the polar crane.

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

Ase a la'mF-'L B C i y Application Creek rlty0Hope , Date 02/24/2003 Emergency and Abnormal Plant Evolutions Rfru 3. SRr 2 295035G110 295035 Secondary Containment High Differential Pressure R0ec rdNumber 17 2.1 Conduct of Operations 2.1.10 Knowledge of conditions and limitations in the facility license. 2.7 3.9

[xiplatof Generic KA FORCED to Procedures section on purpose. IGNORE 295035 K/A Title.

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 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. (SRO/STA Only)

Z Tech Specs without Definitions, Bases, or Admin Section.

t o New = ~ _,_-

' ' -- - -- ---- ~'- -- l l -,9 ^n h osi Quetit~on^Sour~ce .cmmets^. el Wednesday, January 22, 2003 2:56:23 PM P 20 of 56 Page

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?

iE Red Blocki ng Tag.

E]I White Caution Tag.

Yellow Permissive Tag.

Y1 EIl Administrative Tag. 0 f DD0 bAnswer

. b Exm B 9j Q I Memory IF S Hope Creek E-, ate: 02/24/2003 T Emergency and Abnormal Plant Evolutions G o 3 SQ .l 2 295035G220 Record Numbe- 1.

295035 Secondary Containment High Differential Pressure Wqm#~~1 8 2.2_ Equipment Control 2.2.20 Knowledge of the process for managing troubleshooting activities. 2.2 3.3 Justification:

of0lna Answer Correct: White Caution Tag. lAW NC.NA-AP.ZZ-001 5 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

R 771 NC.NA-AP.ZZ-0015 5.4.4 NA0015EO04 Identify the kinds of tags and their purpose IAW the Safety Tagging Procedure, NC.NA-AP.ZZ-0015(Q) and the SAPIWCM Tagging Operations Procedure, SH.OP-AP.ZZ-0015(Q).

[Ma~teral Requred f G;i None E $u e Z Other Facility -_-- 1°sWI iat M Editoriaiiy Modified Qu(E.s: ti Gl. -l2. Peach Bottom 2002 LSRO question 3-8 modified for Hope Creek.

Wednesday, January 22, 2003 2:56:23 PM Page 21 of 56

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?

E 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

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

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

li3 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ._ _

Aser b ae B '1iK" Le Application FilEll HopeCreek E Oaf 02/2412003 Tier: Emergency and Abnormal Plant Evolutions I 3 SRO r 2 295035G430 295035 Secondary Containment High Differential Pressure imber 0 19 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.

lXtWn, Generic KA FORCED to Procedures section on purpose. IGNORE 295035 K/A Title.

1 e2 J Justification:

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

' ,1 i' - 1' , '!" ii,, " j g !-""" " 4 191AZE'MV ONE 'I, 1 2001212 29 MMMMMMI__

M,INg MM 1I !5'11111'7

_T'z Is J.E]jj mm! 21.10 -, SmithMwi' b" I - -EiI LL

, o Materil ReHC ECG without Introduction and Usage section tIQ S . ... New Q sto [ a M-thod; Quest'on r tsmentl: E Plan for Operations Duties NEPLICOPSHCC Obj 6.1 Wednesday, January 22, 2003 2:56:23 PM Page 22 of 56

The Operational Fire Protection Program requires special precautions for metals that may be combustible in an oxygenated atmosphere.

Which of the following equipment would require such precautions?

B Refueling platform main hoist.

I FueI assembly upper tepplates.

El Fuel assembly channels.

M Control rod blades. _ _ _

Answr c a el B i l l Memory FIl Hope Creek Em Datel 02/24/2003 Emergency and Abnormal Plant Evolutions 2 60000OK302 600000 Plant Fire On Site 20 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 r-xnati~onol:Justification:

Answer Correct: Fuel assembly channels. As specified in NC.NA-AP.ZZ-0025, fuel assembly channels are made of Zircronium which is an ignitable metal.

Incorrect: Refueling platform main hoist. Made of stainless steel.

Incorrect: Fuel assembly upper tie plates. Made of stainless steel.

Incorrect: Control rod blades. CRBs are made from stainless steel, boron carbide, and/or halfnium NC. NA-AP.ZZ-0025 R .c Detr mine where i a lr - .

ADMPROEO64 Given access to Control Room References Determine where ignitable metals are used at HCGS. lAW NC.NA-AP.ZZ-0025.

MIla~terial Required for Ea i ' & None 1tion lour~e::

Facility Exam Bank Questi0

.r Moddl

- a ton e '. v Editorially Modified Question 8our mnts~ fQ57138 Modified Wednesday, January 22, 2003 2:56:23 PM Page 23 of 56

Given the following conditions:

- The plant is in Operational Condition 5.

- 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)?

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

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

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

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

gAnwr d Ex L IB Itveev Comprehension Faflity' *Hope Creek ExamR.ate:1 02/24/2003 W Plant Systems Iroup 1 SRG pl I 20300OK301 203000 RHR/LPCI: Injection Mode (Plant Specific) RcrNb er 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 Flixnaiio f 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-IO.ZZ-0005 step 5.2.46 IOPO05EO06 (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.

jIM90equird 0*amiWiMfl INone E~tion~u~e'New ____, ___ jpMthd Question ce non=tsI:-

....1our - II ..

Wednesday, January 22, 2003 2:56:23 PM Page 24 of 56

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.

Ed if NO fuel movement occurs.

Eli ondy in the quadrants monitored by SRMs A and D only in the quadrants monitored by SRMs B and C.

r CAn Enevel C l Bo iy l Application Hope Creek Exam 02/24/2003 ier:. Plant Systems 0 o lou

, $RA 1 215004G232 215004 Source Range Monitor (SRM) System e Numbrx 22 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.

[EkonatiOndof Justification: IAW Tech Spec 3.9.2.b - One operable SRM must be in the quadrant where the core

[erU . 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.

"can continue if no fuel movement occurs" - Incorrect- core Alts in only A/D quadrants.

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

Tech Spec 3.9.2.b SRMSYSEO07 (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.

Mater ia lu r9 . Tech Specs without Definitions, Bases, or Admin Section.

Q Facility Exam Bank l esti onMtodm Significantly Modified uesinSot' VISION QID# 60987. Significantly modified.

Wednesday, January 22, 2003 2:56:23 PM Page 25 of 56

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)

El 1of 4 "B" level LPRMs assigned to "A" APRM is placed in BYP.

[a 3 of 4 "B" leve LPRMs assigned to "1A" APRM are placed In CAL.

[i 8 of 21 LPRMs assigned to "A AP RM are placed in BYP.

IiI 1of 21 LPRMs assigned to "A" APRM is placed in CAL.

Anwr c . l IB Comprehension oi'Ie R-RiHope Creek Eia 02/24/2003 Tier- Plant Systems RO: 0u 1 9ru 1 215005A103 215005 Average Power Range Monitor/Local Power Range Monitor System R~co . 23 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==M 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 APRMOOE01 0 (R) From memory, explain why an inoperable trip of an APRM channel is initiated if there are less than 14 LPRM inputs, lAW the Student Handout.

lMatera RhtiOn None jqo6R! New NuW e~id~ M-oftaio QueItionIIource C t __ ,_ X Wednesday, January 22, 2003 2:56:23 PM Page 26 of i56

Given the following:

- The plant is in OPCON 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)

El 0cfm.

El 9.0 000 cfm.. .

2I 10,000 cfm.

I1 180,000 cfm.

d B logv,,QelComprehension Lell 1Anwer F Hope creek e 02/24/2003

.i7 Plant Systems I Groufr 1 ROG 1l 261000A301 261000 Standby Gas Treatment System

___..__ 0e0o 0 Number 24

. .M_

A3. Ability to monitor automatic operations of the STANDBY GAS TREATMENT SYSTEM including:

A3.01 System flow 3.2 3.3 Exnton.o.f JUSTIFICATION:

Correct Answer:"180,000 cfm" All six FRVS fans will automatically start on the high Refuel Floor Exhaust radiation signal. The 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***

HC.OP-SO.GU-0001

. :L=>N" >N WM__ LM-9 _-2i. , A RBVENTE006 (R) Given plant conditions, distinguish between the automatic starts and stops associated with the Filtration Recirculation Ventilation System (FRVS) Recirc Fans.

MaterialRe1 qui oErnISn l None l Facility Exam Bank V S.n. . gdif iI ieto Editorially Modified Comnt OQutionouroe Vision Bank QID # Q60662.

Wednesday, January 22, 2003 2:56:24 PM Page 27 of 56

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?

_ _ .,_,.__ I _1 ..., _

El The diesel generator will NOT automatically start.

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

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

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

a Exa iel B Lel I Memory FwiIg Hope Creek E lDe 02/24/2003 T

,:4- Plant Systems 1 S9oP 1 264000K609 i I. I~- ... . ............

-- =... .... ...

264000 Emergency Generators (Diesel/Jet) [Recoumber : 25 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 Explanation ofj 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

-M .- .............. .................

EDGOOOE01 1 (R) Given plant conditions, predict the response of Diesel Generator governor and voltage regulator control circuitry to an Emergency start (LOP/LOCA).

lMate*ria _ l None 1Question'urce,::;l Facility Exam Bank l gtoidim!alo Mod: I Editorially Modified Qustr c t Vision 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.

Wednesday, January 22, 2003 2:56:24 PM Page 28 of 56

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

1 Jet pump inlet riser.

1 Jet pump diffuser section.

E1Shutdown Cooling suction line.

IiII Low pressure LPCI injection line. _ _

Answer b elB !Ea l b ~e Memory l Hope creek [God 02/24/2003 Tier Plant Systems o P 2 lr 2 202001K401 202001 Recirculation System r 26 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

[~xpianation~o] Justification:

Answer 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. ------- _-_-_

l110 7 . 4000 . US I 7i77777777 Tech Spec Bases 3/4.4.1 RECIRCEOO7 Explain/Discuss how the Recirculation System is designed to ensure a 2/3 core height refloodable volume is maintained. AW available control room references:

Materia r Examinatn Geuired lNone o u eI.INPO Exam Bank uet ifii itd Direct From Source IQuestion So deCo'lmmen INPO Bank QID# 12456 07/02/1 999 Limerick Wednesday, January 22, 2003 2:56:24 PM Page 29 of 56

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?

Hi Alternate Shutdown cooling using "D" to "B" RHR pump cross-tie.

Ei Manually operate SDC valves and restart B RHR pump.

El RHR RPV head spray with all RPV head vent valves open. 0

[E] Maximizing Fuel Pool cooling with both Fuel Pool heat exchangers.

b xael B Application FailitlHope Creek Exam 02/24/2003 Tier.' Plant Systems R G u 2i R94 2 205000A205 205000 Shutdown Cooling System (RHR Shutdown Cooling Mode) R Number 27 A2. Ability to (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 E3xlfnatioin'of Justification:

1 J

.lCorrect: 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.

HC.OP-AB.RPV-0009 subsequent actions "A3 and A4" eari~~~ing~ . Obetvs.0".

ABRPV9E07 (R) Explain the bases for Subsequent Actions and the information contained in the Discussion Bases Section of Shutdown Cooling.

lMaterial Required for Eain0 , None p0i00ek # Facility Exam Bank - _ _ QuosnMo del igfc itly Modified ueo r mVISION Bank QlD #Q61331 Stem modified to provide a correct answer. Answers modified.

Wednesday, January 22, 2003 2:56:24 PM Page 30 of 56

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?

El RHR SHUTDOWN CLG INBD ISLN(F009) closes; RHR LOOP B RET TO RECIRC(F015B) opens; RHR PMP B SUCT FROM RECIRC(FO06B) closes.

Eil RHR SHUTDOWN CLG OUTBD ISLN(F008) closes; RHR PMP B SUCT FROM RECIRC(FO06B) closes; RHR LOOP B RET TO RECIRC(F015B) closes.

El RHR SHUTDOWN CLG OUTBD ISLN(F008) closes; RHR LOOP B RET TO RECIRC(F015B) closes; RHR PMP B MIN FLOW MOV(FO07B) opens.

.RHR S UTDOWN .LG O UT..B_ D_ .I -SI-L-N .(F-00__8_)_ccloses; RHR SHUTDOWN CLG INBD ISLN(FO09) closes; RHR LOOP B RET TO RECIRC(F015B) closes.

d Exa I B JAnswerPowe Comprehension liiltR4 Hope Creek EaDae 02/24/2003 Tier: Plant Systems Rro l 2 SO u 2 205000K402 205000 Shutdown Cooling System (RHR Shutdown Cooling Mode) er 28 K4. 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:Ejianti6* fJustification:

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(F006B) 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(FO07B) 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 RHRSYSE011 Given a labeled drawing of, or access to the Residual Heat Removal System controls/indication on 10C650:

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

b. Assess plant conditions which will cause the indicators to Wednesday, January 22, 2003 2:56:24 PM Page 31 of 56

Maeia~l Require o xadlinition 4~gNone IFacility Exam Bank Q Mop on M : Significantly Modified esionS-ouii e aommentl:l I Q55155 Significantly Modified Wednesday, January 22, 2003 2:56:24 PM Page 32 of 56

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

tiE 24 Volt Non-i E DC batteries.

El 125 Volt_N...on--..l.E-. D--.C. -batte-ries..

El 125 Volt 1E DC batteries.

El 250 Volt 1E DC batteries.

Answera EiTI .B 0 Memory l Hope Creek Examte 02/24/2003 Tir Plant Systems Q Ip 1 2 215003K201 215003 Intermediate Range Monitor (IRM) System R Nb 29 K2. Knowledge of electrical power supplies to the following:

K2.01 IRM channels/detectors 2.5 2.7 Jstcatio-n:

Epunation Answer ; 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 1:If001 o = IN"0 S 0- Mm NO EaO II I ii II E-001 0 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 IMaterialRqiu~i re' I forgExamiatwip None 19 New 1Ques M9oIifitin metho IQussio;$5ir C<-- --

Wednesday, January 22, 2003 2:56:24 PM Page 33 of 56

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?

Ell The CRB is still coupled to the drive.

Ell The ORB bail handle has broken free.

U The CRB removal tool air hoses are slack.

E The CRB exceeds the setpoint of the hoist cutoff.

Answer a I e B Ceith Lv[ Comprehension l : Hope creek Ex Date 02/24/2003 Plant Systems 323400K105 234000 Fuel Handling Equipment e m 30 K1. Knowledge of the physical connections and/or cause- effect relationships between FUEL HANDLING EQUIPMENT and the following:

K1.05 Reactor vessel components: Plant-Specific 2.9 3.3 nTEXTonX12o] Justification:

e 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 exceeds the setpoint of the hoist cutoff.. Monorail hoist load cell cutoff is set at 500

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

HC.RE-FR.ZZ-0002 Caution 5.6.11 itn 0 M~a00iSR A 'aWv h10t0500::0j0000 REFUELE010 Given a drawing of, or access to, the frame mounted hoist or monorail hoist control pendant, explain the controls and indications lAW the Student Handout.

Mateialqir fNone IINPO Exam Bank e Modifon Method: Significantly S Modified lQueionSoure C : INPO Bank QID# 14062 07/02/1999 Peach Bottom Wednesday, January 22, 2003 2:56:25 PM Page 34 of 56

Given the following conditions:

- The plant is in Operational Condition 5.

- The Reactor Mode switch is in the REFUEL position.

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

Which one of the following would cause a Rod Block under these conditions?

IfThe Fuel Grapple is loaded with fuel.

iI The Fuel Grapple is in the FULL UP position.

i The Frame Mounted Auxiliary Hoist is loaded with fuel.

Eil All rods are Full-In, except for a selected rod at position 02.

nswerla E i_ IB Ilgny------ lComprehension FlIt Hope Creek Exa Date:DA 02/24/2003 Tier Plant Systems R rp 3 $ Gup 2 234000K502 234000 Fuel Handling Equipment Re Number K5. Knowledge of the operational implications of the following concepts as they apply to FUEL HANDLING EQUI PMENT:

K5.02 Fuel handling equipment interlocks 3.1 3.7 goki!*1600B Justification:

war. - 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

_~i

_11_l<R. i~ ____ __,__

. i W_ _,__ . __. _.

HC.OP-SO. KE-0001

, Learin"n ObNetivs >

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 ff6 'iia _' . None l9j arcei:ll Facility Exam Bank '9. M.o I Editorially Modified QuestinSurc Yt~ :Vision bank QID #Q56552 Wednesday, January 22, 2003 2:56:25 PM Page 35 of 56

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?

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

Etl Reactor Building Exihaust Radiation monitors will alarm first and RBVS will start. Evacuate the refueling floor.

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

Refuel F-ioor Exhaust Radiation monitors will alarm first and RBVS will trip. Evacuate the refueling floor.

Answer c x el B eLevel; lMemory F 0 Hope Creek Exam Date: 02/24/2003 Tier:f. Plant Systems .l u 2 UPl 2 272000A201 272000 Radiation Monitoring System eorNmr 32 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 Exphinationbf Justification:

nwrd J 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.

Incorrect: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 IRADIATED FUEL DAMAGE HC.OP-AB.CONT-0005 [RADIATED FUEL DAMAGE M-76, M-84 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.

IMatnlRaurado iti<

Wens _ . January 22, 23 PM Pg 36 of 5 Wednesday, January 22, 2003 2:56:25 PM Page 36 of 56

None Oue&tien Mddiflb4ti&n M9hot tn QuestionSource l New~

Sour~e Comens:

Wednesday, January 22, 2003 2:56:25 PM AXPage 37 of 56

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?

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

Eil Low-Low-Low Level in the A SACS Expansion Tank.

El Low differential pressure on the B SACS pump.

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

Anser b evelB IExam I le m emory It Hope Creek Fiji 02/24/2003 Tije: Plant Systems 0 Up 2 G il 2 400000K401 400000 Component Cooling Water System (CCWS) 33 K4. Knowledge of COWS design feature(s) and or interlocks which provide for the following:

K4.01 Automatic start of standby pump 3.4 3.9 jipnationof& Justification:

[00L J ICorrect: 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.

HC.OP-SO. EG-0001 ii :t~

g LovmnOjcies~je- r0;f 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 Material equired rE tOn 1 None lQion5og l New I9 ^.l Modifin Quesion S§ r w Wednesday, January 22, 2003 2:56:25 PM Page 38 of 56

Why does HC.RE-FR.ZZ-0002 require that upon completion of uncoupling a CRD from it's blade, that the associated HCU is valved out by closing the HCU insert riser valve 101 before closing the withdraw riser 102 valve?

iI To prevent equipment damage in the event the rod is scrammed.

To prepare the CRD drive unit for replacement.

H To ensure the rod operability Tech Spec is satisfied.

I1 To prevent inadvertent draining of the reactor vessel if the CRD drive unit is removed.

Answe a l. L"0B .t eL Memory NWi Hope Creek 9aWp+te 02/24/2003 T Plant Systems i / 9U 2 S 3 201003K101 201003 Control Rod and Drive Mechanism ecrNumber 34 Ki. 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 l o l Justification:

F Correct: To prevent equipment damage in the event the rod is scrammed. Bases for procedure Caution 5.5.4.F contained in the Caution.

Other answer choices are plausible misconceptions.

HC.RE-FR.ZZ-0002 5.4.4.F HC.OP-SO.BF-0002 limitation 3.2.2.

,. _ _~ i&_~ _<~

. _.~. e_

CRDHYDEO16 Given procedure HC.OP-SO.BF-0002, explain why the CRD HCU Insert Riser Valve (1BF-V101) must be closed prior to closing either the Withdraw Riser Valve (1BF-V1 02) or the Scram Discharge Riser Valve (1BF-V1 12) when isolating a CRD HCU.

Material R#quir rxmin.on- - None lQ3qio or: INPO Exam Bank e's0i Widz onl Moietion-,Mh Editorially Modified Quesion m en INPO Bank QID# 16849 03/19/i1998 Quad Cities modified for Hope Creek.

Wednesday, January 22, 2003 2:56:25 PM Page 39 of 56

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)?

EI No automatic actions occur when onlyone NSSSS channel manual isolation switch is actuated.

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

i The TIP Guide Tube Ball Valve automatically closes: cutting the detector cable and sealing the guide tube._

The T TIP detector wilI withdraw to its "in-shield" position and the TIP Guide Tube Bali Valves automatically close.

Answer d xa Le B CognoyemLvp Memory Tl t Hope Creek IEx~am Date:l 02/24/2003 Tier: Plant Systems i i 3 $S P 3 2150011K604 I. ..A i gz ......

. , vyac :. . A,,

215001 Traversing In-Core Probe J 3_

.. _ i.i __ . I__ Reor Number . 35 K6. 0 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&lI(Not- BWR1) 3.1 3.4 fExplaation6fl Justification:

AI ML '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-01 7 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.

HC.RE-SO.SE-0001, Section 3.1, Precautions and Limitations HC.OP-SO.SM-0001, Table SM-017; 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.

M ~at~erial Required r m ati > . None Ewon Srce Facility Exam Bank 0 _ o t e Editorially Modified Qution S rcC m t Vision Bank QID# Q53710 editorially modified due to correct answer was longest and most detailed answer.

Wednesday, January 22, 2003 2:56:26 PM Page 40 of 56

Given the following conditions:

- A core reload is in progress.

- 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.

- The refuel floor ARM is NOT alarming.

What immediate operator action is required by HC.OP-AB.COOL-0004 FUEL POOL COOLING and why?

E Isolate Fuel Pool Cooling because that is a potential source of the leak.

Ei Move the bridge over the reactor cavity because it is further away from the fuel pool.

E Place the fueI assembly in the designated open rack location in the fuel pool because it is a safe location.

d Suspend movement of the fuel assembly at its present condition because Core Alterations must be suspended.

lAnswe c EXa eB Cont1 Memory FacWWI Hope Creek E 02/24/2003 Tier:' Plant Systems 0 Gy.lp 3 $RO 3 233000A202 233000 Fuel Pool Cooling and Clean-up Ro 36 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 Ej40666611of Justification:

.Aswer -Correct: Place the fuel assembly in the nearest open rack location in the fuel pool because it is a safe location. Immmediate 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: Isolate Fuel Pool Cooling because that is a potential source of the leak. Subsequent action.

Incorrect: Move the bridge over the reactor cavity because it is further away from the fuel pool. 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 10A. 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.

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

Wednesday, January 22, 2003 2:56:26 PM Page 41 of 56

jaeriIRqfrdfrEamnt None__ .

I n INPO Exam Bank 0 i0stion lid Editorially Modified Question Seurc ~ei:1 tg INPO Bank QIO# 16907 03/19/1 998 Quad Cities Wednesday, January 22, 2003 2:56:26 PM Page 42 of 56

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?

HI No action is required.

El Return CRB 1l5 to the control cell within one hour.

tij Return SLC Tank within specification within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

El Insert control rod 30-31 within one hour.

.r x ma...S ..

Answer d lExamL B . g0 v Ln Application Faciyi.y Hope Creek at: 02/24/2003 IT*# Plant Systems I Ar up 31 $RO ru 3 290002K605 M: '9'1 "".'7'.-"? =_"'- 'i: :'" } _

290002 Reactor Vessel Internals Record Number 37 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 E[iali7ifdi Justification:

Anwer~SCorrect: 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.

7I,77 7.~ ~~ ~ ~ ~~~1 'y~2t-0 __ iW4 1 @

it

!Ca ar 110";R.Y}'b}kYYY .~.0ile ; : XA 00 7777M I

Tech Spec 3.1.5 and 3.9.10.2 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 IMaea qR i. in l Tech Specs without Definitions, Bases, or Admin Section.

I:'ins6i1 New l i M ctMthod:

Questio _ I . . 0SouM#---------

Wednesday, January 22, 2003 2:56:26 PM Page 43 of 56

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?

EIl 1000 Mrem/yr.

2000 Mrem/yr.

-Ei

[Il 3000 Mrem/yr.

A 4000 Mrem/yr.

Answer b I *B B -_ G v Leve Memory 'F§IyHope Creek Ea ae 02/24/2003 Generic Knowledge and Abilities dTer: I"Oll 11 S P p 1 294001 G301 GENERICR 38 2.3 Radiological Controls 2.3.1 Knowledge of 10 CFR 20 and related facility radiation control requirements. 2.6 3.0

.ExthaOf, Justification:

Correct: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.

N:.NA-AP.ZZ-02

Atahe ::

<~~~~~

Rev. ' _

, $ ferideTi $09 i i 0dRata-. 0. 000t ^t;tle00 77

=' _^ _ 1I 'I I ' -S'- -_

I } I NC. NA-AP.ZZ-0024 Rev. 1 1 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:

Mt*Rqltidn :None 1tiontou.ciZFacility Exam Bank uEsn :::c : Meth: Direct From Source Ql o. Sourge C : SVISION Bank QID# Q60666 Wednesday, January 22, 2003 2:56:26 PM- Page 44 of 56

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.

II 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.

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

El 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 Exam B G3 teLep Application l>^iyHope Creek E 02/24/2003 Tier: Generic Knowledge and Abilities A _ 1 8R l 1 294001G302 GENERIC RecordNumber : 39 2.3 Radiological Controls 2.3.2 Knowledge of facility ALARA program. 2.5 2.9 Expti;Nd Justification 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 6OmR/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.

lMaterial euired fr ~ l~None 46n0M INPO Exam Bank . .l Direct From Source estiol INPO Exam Bank QID# 7593 11/04/1997 FitzPatrick Wednesday, January 22, 2003 2:56:26 PM Page 45 of 56

Given the following conditions:

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

- The 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) l VP-Operations; to 4000_mr/yr TEDE.__

Radiation Protection Supervisor; to 3000 mr/yr TEDE.

II Operations Manager; to 3000 mr/yr TEDE.

@ Radiation Protection Manager; to 4000 mr/yr TEDE.

Answer a xm B gN W Comprehension FcIl Hope Creek Exam 02/24/2003 Tier: Generic Knowledge and Abilities ! 1 1 SR r 1 294001G304 GENERIC ReordNumberI 40 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.

Explanationtof Justification:

LnsL........j Correct: VP-Operations, to 4000 mr/yr TEDE. VP needed for 4000 mr/yr extension because > 2(N-17) 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-17) 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.

e I1 Tate 'p-NC.NA-AP.ZZ-0024 Attachment 1 2(N-17) 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

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

IMaterial =ecuired{&~i~L0I~ None Facility Exam Bank u Metho l Editorially Modified QuiestionSourc co $ VIION Bank QID# Q55936 Modified to remove delegation to other approvers and to make 3000 limits plausible.

Wednesday, January 22, 2003 2:56:26 PM Page 46 of 56

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

[I 100 cpm 100 cpm above backround 100 cpm 300 cpm above backround I 300 cpm 100 cpm above backround 300 cpm 300 cpm above backround Answer c Exam Leve B MOPt e Memory i Hope Creek Exam Date: 02/24/2003 Ter Generic Knowledge and Abilities R 1 r 1 SRO Re ord.

.. Num.

294001G305 b

.41.......

GENERIC ff~ber:'41 2.3 Radiological Controls 2.3.5 Knowledge of use and function of personnel monitoring equipment. 2.3 2.5 tn If greater than 300 cps, find another frisker or notify RP tech. Indication of contamination is 100 cps above initial backround reading.

777777I gg AR. i Radiation worker training handout material

_, _ .r , , n.

. e . _v Material Requiedfor Ex amj in 7 None O6Po>

er eFacitity S -1 'et'1 Qiestiqndificatioi Direct From Source lQuestion St . Peach Bottom 2002 LSRO Exam question 4-5 unmodified. Hope Creek has the same limits.

Wednesday, January 22, 2003 2:56:27 PM Page 47 of 56

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?

IAJ Aux Hoist uptravel with a fuel bundle is stopped 6 feet below the water surface.

i The Main Fuel Grapple must be used to remove a fuel bundle.

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

I3 Minmize the time the fuel bundle in the Drywell Bellows Area.

Answer b Exam B Co e Memory Wa] tyo HopeCreek l at 02/24/2003 Generic Knowledge and Abilities l _ SR1l A l 1 294001G310 GENERIC r b 42 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.

E[014atoln>, 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.

HC.OP-ST.KE-0001 HC. OP-FT. KE-0002 UFSAR 9.1.4.1 REFUELE011 From memory, identify the only grapple which may be used to move fuel in the reactor vessel or spent fuel pool lAW Technical Specifications.

Material Requiredor ExlamX ~ l None l Qion 0eurcelj New X = A n tipnthod:

Question Soure 06i: lm.ents:l Wednesday, January 22, 2003 2:56:27 PM Page 48 of 56

Delayed neutrons are neutrons that:

EiI have reached thermal equilibrium with the surrounding medium.

El are born within 1OE-14 seconds of the fission event.

El are produced from the radioactive decay of specific fission fragments.

are responsible for the majority of U-235 fissions.

CAnr C Levl B . ve IMemory a HopeCreek EaDe 02/24/2003 Tier: Fundamentals R u 0 RGo 0 2920011K102 292001 Neutrons Reo.r...Number.......

....1.-l- 43 K1.0 K1.02 Define prompt and delayed neutrons. .0 3.1 Anawr;IVo.of<

NON E ference;~ 1Tit 777

...... i S U iS i- .L XM KINETIE003 - __

6 . on - - --- -

Describe the production of delayed neutrons.

_ WE --

-'WW .

-- i-

_-.___-_.11 ..........._- -_ ..

lMatrlRquiredor Ex ~None L NRC Exam Bank = = M-hl Direct From Source iuetng;<^SBWR GFE BANK QUESTION IDQ: B1 945 (P845)

Wednesday, January 22, 2003 2:56:27 PM P Page 49 of 56

Refer to the reactor response curve attached (Q 44) 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 El reactor period; subcritical.

Eil reactor period, critical.

reactor fission rate; subcritical.

I-i reactor fission rate; critical.

Answer C Ex e B WOitiV Comprehension F1MiIlt Hope Creek ExamPate: l 02/24/2003 Tioer: Fundamentals 0° )S O1r'u D 292003K107 292003 Reactor Kinetics and Neutron Sources Reor Nmber 44 wEi...'.....

K1.0 K1 .07 Explain prompt critical, prompt jump, and prompt drop. .3 3.3 Explanation of Ansawer:SSX:i

,~ .- fer nce :T:tLeS OMINN . ~ ~ ~ ~ ~~A MIME eOkjective' KINETIE007 Explain prompt critical, prompt jump, and prompt drop.

MatriRequire foAttached figure from GFE Question B3250 ion rf NRC Exam Bank_ _ M iationMetod Direct From Source IQuj un NRC BWR GFE Bank Question ID: B3250 (P3249)

Wednesday, January 22, 2003 2:56:27 PM Page 50 of 56

Compared to beginning of core life, the Doppler coefficient of reactivity is negative at end of core life due to (Assume the same initial fuel temperature.)

HI less; depletion of U-238.

K1more; burnout of gadolinium.

l ess; buildup of fission products.

more; bui dup of Pu-240. _ _ _

Ad xIBe B Hope lH _v-P-Iemor Creek Eaial 02/24/2003 em. Fundamentals O4 GU ok 292004K105 292004 Reactivity Coefficients Reactivity Coefficients t 0 . 1 45 K1.0 _ -

K1i.05 Define the doppler coefficient of reactivity. .9 2.9 Eilapat..W==

i hpc of reactivity.

..1a BWRTHRE01 1 Explain the doppler coefficient of reactivity.

hNone R8 IMOterl

,NRC Exam Bank M

' oI ~--

co Direct From Source

_... ....... B E E. m n.BQ ....o D 1 . ..

I M.NRC BWR GFE Exam Bank Question ID: B1353 Wednesday, January 22, 2003 2:56:27 PM Page 51 of 56

Rod position indication shows that a control rod is at position 22. If the control rod is then moved to position 12, it is being: ..........

H inserted 30 inches.

Ii withdrawn 30 inches.

IAl inserted 60 inches.

By withdrawn 60 inches.

IAnsn A ll B A wl na comprehension l Hope Creek EIm l 02/24/2003 Fundamentals ...... . ...

0 0O = 292005K101 292005 Control Rods [ 0 N0 be 46 K1.0 ---- =--=____

K1.01 Relate notch and rod position. .2 3.3

.. AETNA11' Cm!

I

'i I-M CONTROE001 Relate notch and rod position 0j;00001.............. ....... .......... ....

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

RAW None l NRC Exam Bank . Direct From Source lQiiestio'iSeulae I~~m' lNRC BWR GFE Exam Bank Question ID: B3054 WJ7 y 0 52 of 56 Wednesday, January 22, 2003 2:56:27 PM Page 52 of 56

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

How much time Lwill be= required ~~~~~. for core heat production ... ........=:=L==L to decrease to 1%following the scram?

I 1to-8 Days. y ____

i1to 8 Hours.

El 1to 8 Minutes.__ __ ___

Li 1 to 8 Seconds.

lAnse b lxa e B I iMemory la . Hope Creek l m al 02/24/2003 e Fundamentals ll _ oj $ I 292008K130 292008 Reactor Operations . 47 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.

Mt ......... .... .. ....

RXOPERE031 Explain the relationship between decay heat generation and

a. power level history

b. power production

c. time since reactor shutdown Im _None NRC Exam Bank lt _ tibd: Direct From Source E NRC BWR GFE Exam Bank Question ID: B2272 (P572)

Wednesday, January 22, 2003 2:56:28 PM Page 53 of 56

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 the "before" and "after" conditions described above?

i12. _- --=- ----- ;

Anwer a A dz 1B I eC [Cmprehension ] i Hope Creek EAm g 22/24/2003 Fundamentals ] [ 0 293006 K1 13 293006 Fluid . ~~~

Statics ~ ~ .. .. .... .. .. _ _

48 K1.1 Pumps and Pump Characteristics K1.13 --Explain the results of putting centrifugal pumps in parallel or series combinations. .6 2.7 p_-E----.......

a-:ipo -- ..-

4 esM~ Nm 0,

I 27777 IN-,"

I IU1~~II, 11I1 .. ~-I1 IIII~~

.111... -- 1--. .~ 11.-I1 ...... I. .1..I.~ II-~ - --- L'Il.

PUMPSOE01: 3 t Describe the operation of centrifugal pumps in series and in parallel arrangements.

I . 7. - 77

- : :: : 7 I rlphat9 I Figure of pump curves for GFE Bank question B2279 NRC Exam Bank Qti i ' Di rect From Source b yllM NRC BWR GFE Exam Bank Question ID: B2279 (P1524)

Wednesday, January 22, 2003 2:56:28 PM P Page 54 of 56

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

I Subcooling of the fluid.

M1 A phase change in the fluid.

i An enthalpy change in the fluid.

El Radiative heat transfer to the fluid.

F-e c lEa v 1B IlMemory ItHope Creek l D 02/24/2003 Fundamentals ] JA 0 9iouP 0 2 93008K1<06 293008 Thermal Hydraulics R... . . Ir 49 K1.0 Pool Boiling Curve (T vs. Q/A)

K1.06 'Define a natural convection heat transfer. .5 2.6

.........cT...

.hea D ta 6 WI St THRMHYE008 Define natural convection heat transfer.

IMat r ~ %~ ~ i e a INone NRC Exam Bank 4p ioz MMM Direct From Source NRC BWR GFE Exam Bank Question ID:B387 Wednesday, January 22, 2003 2:56:28 PM t.- ----

Page

.........55 of 56

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).

Ii-greater than; greater than Ed greater than; less than El less than; greater than less than; less than Aer d le ]-IqI'emory mB XilH Hope creek iE 02/24/2003 Fundamentals *.. .. o] 0 29301 OK101 293010 I-Brittle Fracture and Vessel Thermal Stress

e RWi 50 K1.0 K1.01 State the brittle fracture mode of failure. .4 2.8 E_ - - =-- L ==n'7f l s:

Wt':5 0 A' _ g0E-L IMMMM MR%

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

BRITTLE005 State the brittle fracture mode of failure.

- E -::: -T: =:::::::::::: :: :: : :: ::: :::::::::::::::::: : :: :::::: :: ::::::: : :::: :::::::: ::::::: : :::::::::: ::: ::: :: :: :::::::::::::.::.::.:.

II NRa - ~None..

NRC Exam Bank _____r e = Direct From Source

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

1 1'. . - -

NRC BWR GFE Exam Bank Question ID:B2499 (P2496)

.. 1 1-. - - . 1. '. ... ~ l , 1. 1 .. ~~ ., .....I . ... .11. . '.. - 1. .. -

Wednesday, January 22, 2003 2:56:28 PM -......... ..- I I - --.. I -

Page 56 of 56

ML031130141

Text

Reference:

Navigation menu

Search