ML20207K671

Forwards NRC Approved Operator Licensing Exam (Facility Outline & Initial Exam Submittal & as-given Operating Exam) for Tests Administered on 980626

ML20207K671
Person / Time
Site:	Salem
Issue date:	03/12/1999
From:	Curley V NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9903170267
Download: ML20207K671 (201)

Text

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NOTE T0: NRC Document Control Desk Hail Stop 0-5-D-24 FROM: kos'l boaleir .L Opert;ingLicens1yTranch,RpsingAssistant )

SUBJECT:

ERATOR LICENSING EXAMINATION ADMINISTERED ON  ;

st.17 AT ._S;La o , .

ET 950.it22 end Ji-3lJ use.7% R.*G l

Ams: Ib l Exass Level: lS l Connitive level: l Memory l l F ," - "-- Process for revising OTSC: The Sponsor Organization shall revise an OTSC (including those that have been ofAnswer approved but not implemented) as follows: A. Prepare a new OTSC with the next sequential change number. B.

• Include the following in the Description of Change on Form-4: "OTSC [old number] superseded by OTSC [new number]". C. Perform remaining steps for initiating an OTSC.

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T Tier: l Generic Knowledge and Abilities l RO Gree, . l1 l SRO Group: ll_

Systesm/ Evolution Number: l_ l System / Evolution

Title:

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Catesory: l2.2 l Equipment Control KA: l2.2.6 l Knowledge of the process for making changes in procedures as described in the safety analysis report.

F O Value: l2.3 l SRO Value: l3.3 l CFR: l 43.3 / 45.13 j

Refenince Refenwace Number Reference Section Page Number (s) Revision Isarn. Ob]

Use and Control of 0300-000.00S- IV.O 57 10M)SS6 6 Procedures PROCED-01 Nuclear Procedure System NC.NA-AP.ZZ- 5.12.5 21 12 0001(Q)

, Question Source New Question Modification Method Question Source Comments: l l Mrterial Required for Emandantion:

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Question Topic: l Surveillance time interval applicability Given the following conditions for Unit 1:

o Reactor pow 2- 100%

o I A Emergency Diesel Generator was declared inoperable at 0800.

o S1.OP-ST.500-0001(Q)," ELECTRICAL POWER SYSTEMS AC SOURCES AllGNMENT" was last performed 7 days ago (Sunday) at 1300.

Which one of the following correctly identifies the latest time for completion of this surveillance today?

a. 0900

b. 0915

c. 1300

d. No later than 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br /> from 1300.

Ans: la l Exam Level: lS l Cognitive Level: l Comprehension l Explanation T.S action a. requires surveillance be performed within ONE hour. For T.S. surveillances Spec 4.0.2 allows a ofAnswer surveillance interval extension is allowed based on engineeringjudgment and recognition that the most probable result of any particular surveillance test will demonstrate operability. 4.0.2 can be applied to any Surveillance Requirement in the Technical Specifications (by definition). However, when a Technical Specification Action Statement refers to a Surveillance Requirement to compensate for an LCO not satisfied, the initial surveillance is being performed to prove the operability of that system, as is this case; and, Specification 4.0.2 cannot be applied to periodic compensatory actions.

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Tier: l Generic Knowledge and Abilities l RO Group: l1 l SRO Group: I1 Systesn/ Evolution Nunkr: l l Systene/ Evolution

Title:

l Catesory: l 2.2 l Equipment Control -

KA: l 2.2.12 l Knowledge of surveillance procedures.

RO Value: l3.0 l SRO Value: l3.4 l CFR: l 41.10 / 45.13 Reference Reference Number Reference Section Pane Nunnber(s) Revision I >=m. Obj Surveillances and Testing 0300 400.00S- III.C.6 12 11/14/97 3 SURV00-00 ELECTRICAL POWER St.OP-ST.5004001(Q) 1.3 2 7 SYS'IEMS AC SOURCES ALIGNMENT Salem - Unit 2 Technical 3.8.1.1 ACTION a 3/48-1 Specifications Question Source New Question Modificatior - I Method Question Source Conunents: l Material Required for Exasnimation:

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1 Ousselse Topice l Independent verificanon acuans Given the following conditions for Unit I-  !

o A Tagout is hanging on a valve located in a high radiation area a

All work associated with the clearance has been completed and all i individuals have signed off _

o The tag must be removed from the valve and the valve must be re-positioned to be IAW the required  !

lineup 1 o Due to the location of the valve, RadPro estimates that the NEO and independent verifier would each receive 225 mrem performing the operation Which one of the following correctly describes the requirement for independent verification during this r, volution? .

a. " Hands-on" independent verification must be performed unless waived by the Operations Manager f

b. Operator self-check is substituted for independent verification whenever an accumulated dose of 210 mrem is involved

c. " Hands-on" independent verification is not required but an alternative means such as system flow or pressure must be used to verify conect positioning of the valve i

d. From a position allowing the best vantage point but lowest dose, the verifier should observe the first  ;

operator self-check and manipulate the valve. " Hands-on" independent verification is not required Amst Ie l Esaan Level: lS l Commitive Imel: l Mamary l Esplanation Selection b. is correct as far as 10 mrem is concerned. However, the actions of c. are required to verify the correct ofAnswer position. '

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l Tier: l Generic Knowledge and Abilities l RO Group: l1 l SRO Group: 11 Systese/ Evolution Number: l l System / Evolution

Title:

l Catgef: l2.2 I Equipment Control KA: l 2.2.13 l Knowledge of tagging and clearance preedures.

RO Value: l3.6 l SRO Value: l3.8 l CFR: l 41.10 / d' (i Refenece Reference Number Refe e.ce Section Pase Number (s) Revision 12arn. Ob]

Conduct Of Operations 0300 @ 0.00S- Independent 24 01/07/97 10 CONDOP-00 Verification STATION OPERATINO NC.NA-AP.ZZ- Attachment 6,1.4 1 9 PRACTICES 000$(Q) i l

Question Source New Question Modification '

Method ,

Question Source Comments: l i i

Material Required for Exaimination:

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Question Topic: l Configuration Change Process Which one of the following ccrrectly identifies the type of design change used to correct existing configuration documents so that they agree with actual plant design but do NOT necessarily require a 10CFR 50.59 review?

a. Standard Design Change

b. As-Built Design Change

c. Engineering Change Authorization Change

d. Generic Equivr. lent Replacement Configuration Change Anst Ib l Exam Level: lS I Cognitive Level: l Memory l Explanation Salem identifies two basic design changes: 1) Standard Design Change - implements a new design for such reasons ofAnswer as increased capacity, improved safety or to meet new regulatory requirements. It requires a 10CFR 50.59 review and any necessary Safety Evaluation to determine if the change affects the UFSAR, requires a revision to Technical Specifications, or results in a unreviewed safety question; and 2) As-Built Design Change - corrects existing configuration documents including the UFSAR to agree with actual plant design after documented design analysis.

Choices c. and d are types of Configuration Change - a change to the fundamental configuration documentation database.

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Tier l Generic Knowledge and Abilities l RO Group: l1 l SRO Group: l1 ,

SysteaWEvolution Nummber: l l Systeni/ Evolution

Title:

l Catenory: l 2.2 l Equipment Control i l

4 KA: l 2.2.14 l Knowledge of the proce s for making configuration changes. 1 RO Value: l2.1 l SRO Value: l3.0 l CFR: l 43.3 /45.13  !

Reference Reference Nunnber Reference Section Pane Numiber(s) Revision Imrn. Obj l Work Control And DCR 0300-000.00S- III.E.2.b 9 2/5/97 2,3 4 Process WC"I'T 30 '

Question Source New Question Modification i Method j @* Source Cosaments: l l Material Required for

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Question Topic: l Determination of acceptable risk Which one of the following correctly identifies an outage condition / action that requires Station General Manager approval?

a. Tagging OOS an EDG that provides power to a RHR pump, when RHR is providing core cooling

b. Going to RHR mid-loop operation, with fuel in the reactor vessel

c. Entering a TSAS for electrical power sources to perform emergent, corrective maintenance

d. Removal of one SW loop from service while on RHR cooling Ans: lb l Ezasn Level: lS l Cognitive Level: l Memory l l Explanation Per NC.NA.AP.ZZA)055(Q) Outage Management Planning, Station GM approval is required prior to going to mid-ofAmerc loop, with fuelin the vessel.

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Tiers l Generic Knowlea,e and Abilities l RO Group: l1 l SRO Group: l1 Systesn/ Evolution Number: l l System / Evolution

Title:

l Category: l 2.2 l Equipment Control KAt l 2.2.18 l Knowledge of the process for managing maintenance activities during shutdown operations.

RO Value: l2.3 l SRO Value: l3.6 l CFR: l 43.5 / 45.13 Reference Reference Number Reference Section Pane Nuamber(s) Revision Learn. Ob]

Outage Risk Management 0300-000.00S- V.C.2.b 47 12/20/96 4&6 OUTAGE-00 Outage Management Program NC.NA-AP.2Z- 5.3 1 0005(Q) l Question Source New Question Modification Method Question Source Comments: l Matedal Required for Frasminarian:

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o Quesdom Topic: AFD Given the following conditions for Unit 2:

4 o - Reactor poweris currently 40%

1 o Cunent AFD is -1.0%

o _100% power AFD target is -6.7%

o AFD history (past 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />):

POWER LEVEL TIME OUTSIDE TARGET BAND

, 80 % 20 minutes

{ 65 % 15 aninutes 1 55 % 10 minutes 40% 20 minutes j Assuming that AFD remains at its current value, which one of the following correctly identifies the

manmum pernussible power level and restrictions (if applicable) during a power ascension?

a. 50 %

f

b. 90%

c. 90%, as long as AFD remains within "the doghouse"

d. 100 %

Aas: Id l Exam Level: lS l Cognitive Level: l Application l R-,* * * - -

The T.S. requirement limits time outside Target Band to 60 minutes within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. At >50% power, count ONE ofAnswer MINUTE for each minute outside band; at <50% power, count 1/2 MINUTE for each minute outside band. In this case total time outside band = 55 minutes. Any power increase above 50% must meet TIME requirement; above 90% power it must meet the TIME requirement and be within the 44,-9 Target Band.

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Tiers l Generic Knowledge and Abilities l RO Group: l1 l SRO Group: l1 S,.^ tvolution Number: l l SysteamfEvolution

Title:

l Cac.y: l2.2 l Equipment Control KAt l 2.2.22 l Knowledge of limiting conditions for operations and safety limits.

RO Value: l3.4 l SRO Value: l4.1 l CFR: l 43.2 / 45.2 Reference Referwace Number Reference Section Page Number (s) Revision Imrn. Obj Power Distrihtion Limits 0300 400.00S- II.A.2.a.6) & 7) 16-17 11/22/96 5 POWER 0-00 ONE HOUR TECHNICAL 0300 000.00S- II.E.1 16-17 8/3096 1  ;

SPECIFICATIONS TSPIHR-00 '

Salem - Unit 2 Technical 3.2.1, Actions a, b & c 3/42-1&2-2 Specifications Question Source New Question Modification Method Question Source Comments: l Material Required for Exaemination:

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Quessies Tepic: l Refueling level basis Wluch one of the following correctly describes the primary reason for maintaining Refueling Cavity water 1:v:1 greatu than 23 feet (127'1%" elevation) over the top of the reactor pressure vessel flange?

a. Sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly,

b. Sufficient water volume is available to provide time for the operator to recognize the indications of a dilution accident before Keff can exceed 95% delta-k/k.

c. Sufficient water volume is available to provide adequate time to implement procedures for cooling the cote, in the event that RHR flow is lost

d. Sufficient water is maintained above the top of the fuel assembly during movement to ensure that the radiation levels at the operating elevation for fuel handling equipment remain below 5 mR/hr.

Ams: Ia l Exam Level: lS l Connitive Level: l Memory l Explanaties De restnctions on minimum water level ensure that sufficient water depth is available to remove 99% of the ofAnswer assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly. The basis for maintaining ONE RHR loop operable (T.S. 3.9.8) with the reactor vessel head removed and 23 feet of water above the reactor pressure vessel flange, is that a large heat sink is available for core cooling. Rus, in the event of a failure of the operating RHR loop, adequate time is provided to initiate emergency procedums to cool the core.

Water does have the function of reducing the radiation levels, but this is NOT the bases for the 23 ft. requirement.

The required boron concentration has a basis for maintaining suberiticality.

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ner: l Generic Knowledge and Abilities l RO Group: l1 l SRO Group: l1 4

Systena/ Evolution Nunnber: l l Systeni/ Evolution Mtle: l Catesory: l2.2 l Equipment Control KAt l 2.2.25 l Knowledge of bases in technical specifications for limiting conditions for operations and safety limits.

21 Value: l2.5 l SRO Value: l3.7 l CFR: l 43.2 Refeewace Reference Number Reference Section Pase Numiber(s) Revision Learn. Obj REFUELING SYSTEM 0300-000.00S- VIII.A.11.d 58 10/30/96 10.b e

REFUELING S2.OP-SO.SF-0009(Q) step 5.1.2.a 4 3 l OPERATIONS Salem - Unit 2 Technical Bases - 3/4.9.10 & B 3/4 9-3 I Specifications 3/4.9.11 Question Source NRC Exam Bank Question Modification Method Question Source Conunents: l Material Requirtd for Er==dantion:

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. Questies Topic: l Fuel handling Which one of the following correctly describes a prerequisite for closing the fuel transfer tube gate valve?

i a.

An interlock prevents valve closure unless the fuel transfer cart is located on the Fuel Handling l Building side

b. An interlock prevents valve closure unless the fuel transfer cart is located on the Containment side ,

c. The fuel transfer cast must be located at or transferred to the Fuel Handling Building side in order to i prevent damage to the fuel handling system ,

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d. The fuel transfer cart must be located at or transferred to the Containment side in order to prevent damage to the fuel handling system I

l Ams: Id l Esasalevel: lS l cm.ative Level: 1 Memory (

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A precaution in S2.OP-SO.SF 0009 and one or more abnormal procedures requires locating the transfer cart on the

of Answer containment side before closing the gate valve, to prevent damage to the transfer system

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Tiers l Generic Knowledge and Abilities l RO Group: l1 l SRO Group: l1 SystemafEvolution Number: l l SystennfEvolution"Iltle: l Catesory: l 2.2 l Equipment Control KA: l 2.2.26 l Knowledge of refueling administrative requirements.

RO Value: l2.5 l SRO Value: l3.7 l CFR: l 43.5 /45.13 '

1 Reference Reference Nunnber Reference Section Pane Number (s) Revision Learn. Obj REFUELING SYSTEM 0300-000.00S- IX.A.4.e 78 10/30/96 12 l REFUEL 00 REFUELING S1.OP-SO.SF-0009(Q) 3.4 2 J OPERATIONS Quesden Source New Question Modi 5 cation Method

@A Source Conunents: l  !

i Material Required for Framminadan:  !

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Quessies Topic: l Exwe Limit As of May 1,1998, an operator assigned to your shift has the following exposure history:

o Cunent Year to date- 1450 mrem TEDE o Cunent Quarter to date- 104 mrem TEDE In the week following the above reported information, the operator records the following exposure:

o Gamma - 50 miem o Neutron-27 mrem Which one of the following correctly states the remaining TEDE dose margin before the administrative l

control level for this individual is exceeded?

a. 550 mrem for the year.

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b. 473 mrem for the year.~ j

c. 3% mrem foithe quarter.

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d. 319 mrem for the quarter. i Ams: lb l Examlevel: lS l Connitivelevel: l Application I l

i NWata= Limit = 2000 mrem /yr. Total TEDE = 1450 + 50 + 27 = 1527 mrem. Margin = 2000 1527 =473 mrem. }

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Tier: l Generic Knowledge and Abilities l RO Group: l1 l SRO Group: l1 Sy.".nTvolution Number: l l System / Evolution

Title:

l Cate;;.it l 2.3 l Radiation Control KA: l2.3.1 l Knowledge of 10 CFR: 20 and related facility radiation control requirements.

m Value: l2.6 l SRO Value: l3.0 l CFR: l 41.12 / 43.4. 45.9 / 45.10 Reference Reference Number Reference Section Revision Learn. Obj 1

Pase Number (s)

RADIATION 0300400.00S- IV.B.5.a 13 12/1096 2.b PROTECTION PROGRAM RADCON40 RADIATION NC.NA-AP.ZZ- Attachment i I 8 '

PROTECTION PROGRAM 0024(Q) l i

Question Source New Question Modification Method Question Source Comments: l M:terial Required for Emandnation:

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Question Topic: l R/A release -

l Given the following conditions for Unit 2: ,

o L The plant is in MODE 5 j

o : A Containment Purge is planned  !

o . System alignment has been checked but the purge has NOT been initiated Which one of the following cormctly identifies a condition that would preclude initiation of a Unit 2 Containment Purge in MODE 57

a. Unit 2 Plant Vent Flow Monitor is inoperable ,

b. Unit 2 Auxiliary Building pressure is slightly negative I

c. Unit 2 VCT is being purged to the plant vent I

d. 2R11 A, Containment Air Particulate Monitor, is inoperable Aw lc l Exam Level: lS l Commitive Level: l Comprehension l '

j Fwplanation Precaution in both S2.OP.WG-0005 and 0006: Do not purge / release from Unit 2 VCr to the Plant Vent during a tofAnswer Containment Purge.

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Tier: l Generic Knowledge and Abilities l RO Grwp: l1 l SRO Group: l1 SystemafEvolution Number: l l System /Evolutiot

Title:

l Category: l2.3 l Radiation Control 4, -

, KAt l2.3.6 l Knowledge of the requirements for reviewing and approving release permits.

RO Value: l2.1 l SRO Value: l3.1 l CFR: l 43.4 /45.10 Referomee Reference Nummber Reference Section Page Nunser(s) Revision Learn. Obj RADIOACTIVE WASTE 0300-000.00S- IX 54,55 9/25/96 12 GAS SYSTEM '

WASGA3-00 CONTAINMiiNT PURGE S2.OP-SO.WG- Precaution 3.3 3 12 TO THE PLANT VENT 0006(Q)

Question Source Facility Exam Bank Question Modification Method Question Source Comunents: l l Matedal Required for Er==al== tion:

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i Quesdom Topic: l ProceduralTransition  !

Unit I has tripped and experienced a Safety Injection.

While performing 1-EOP-LOCA-1 " Loss Of Reactor Coolant", a PURPG path condition was noted for the Core Cooling Critical Safety Function and 1-EOP-FRCC-2, " Response To Degraded Core Cooling", was entered. While performing the steps of tMs procedure, the STA reports concurrent RED path condition for i the Heat Sink Critical Safety Function and for the Containment Environment Critical Safety Function. No i other abnormal condition was noted.

Which one of the following correctly identifies the appropriate action for the above conditions?

a. Complete the actions of 1-EOP-FRCC-2, and then transition to 1-EOP-FRHS-1. .

b. Complete the actions of 1-EOP-FRCC-2, and then transition to 1-EOP-FRCE-1.

c. Stop performing 1-EOP-FRCC-2, and immediately transition to 1-EOP-FRHS-1.

d. Stop performing 1-EOP-FRCC-2, and immediately transition to 1-EOP-FRCE-1.

Aus: lc j Exasa Level: lS l Cognitive Level: l Comprehension l i Explanation Once A FRP is entered, that FRP is performed until the point of the defined transition, unless pre-empted by a i  !

ofAnswer higher psiority critical safety function condition. The operator will suspend a FRP it. use prior to completion if a RED or PURPLE path of higher priority exists ( the operator transitions to that FRP). All patia of RED priority should be completed in the t.equence indicated by the CFST Hierarchy prior to addressing any other path. A RED

PATH on heat sink would have priority over a RED PATH on containment environment.

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Tier: l Generic Knowledge and Abilities a Group: l1 l SRO Group: l1 Systene/ Evolution Nasaber: l l SydMvolution

Title:

l Cat.y: l2.4 l Emergency Procedures / Plan KA: ,1 2.4.8 Knowledge of how the event-based emergency / abnormal operating procedures are used in conjunction I with the symptom-based EOPs.

ROIalue: l3.0 l SRO Value: l3.7 l CFR: l 41.10/43.5 /45.13 Reference Reference Nuenber Referwece Section Page Nussber(s) Revision Learn. Ob]

EOP-TRIP-1, REAuUR 0300400.00S-TRP001- 4.2-4.3 31 05/10/96 9 4 TRIP OR SAFETY 01 INJECTIONAND INTRODUCTION TO THE USE OFEOPs USE OF PROCEDURES SC.OP-AP.Z7,0102(Q) 5.3.12.E 19 5 Question Source NRC Exam Bank , Question Modifica. tion . . - - - -. > - - + <

Method l Question Source Conuments: l Material Required for 2

Fummination:

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Quaseles Topic: Guadeline implementation eva'uation i i

Given the following conditions for Unit 2.

o A reactor trip and Safety Injection has occurred from 100% power o l 2-EOP-TRIP-1 " Reactor Trip Or Safety Injection" has just been entered i o 21 SG pressure is 500 psig below the other THREE SG pressures  !

o 2C 4KV Vital Bus did not energize and 21 CS Pump failed to start  ;

o Containment pressure is 48 psig i Which one of the choices correctly completes the following statement? ,

I

'Ihe 21 AFI1 and 21AF21 valves should be closed... i

a. within 10 minutes.The actions of 2-EOP-FRCE-1 should be performed following verification of the  ;

--. 'immediate actions of 2-SOP-TRIP-1  ;

b. humediately. The actions of 2-EOP-FRCE-1 should be performed following verification of the immediate actions of 2-EOP-TRIP-1  !

c. as soon as transition is made from EOP-TRIP-1. The actions of 2-EOP-FRCE-1 should be performed when directed by step (s) of 2-EOP-TRIP-1

d. following verification of the inunediate actions of 2-EOP-TRIP-1. The actions of 2-EOP-FRCE-1 should be performed following transition from 2-EOP-TRIP-1 Ams: Id l Essen Level: lS I Commitive Level: l Comprehension l  !

t Wwya===*Aa= In EOP-TRIP-1, the CAS is organized into two sections, one of which is a summary of RED Path transitions to the ofAnswer function restoration procedures. 'Ihe NON-RED Path CAS's apply as swn as the immediate actions are verified; the RED Path transitions are effective when TRIP-1 specifically directs the operator to begin monitoring the critical .

function status trees.

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T l Generic Knowledge and Abilities l RO Group: l1 l SRO Group: l1 Systemn/ Evolution Nussber: l l Systeni/ Evolution

Title:

l Ca' . y: l2.4 l Emergency Procedures / Plan KAt l 2.4.16 l Knowledge of EOP implementation hierarchy and coordination with other support procedures.

Rt9 Value: l3.0 l SRO Value: l4.0 l CFR: l 41.10/43.5 /45.13 Reference Reference Nuniber Reference Section Pase Nunnber(s) Revision Lears. Obj EOP-TRIP-1, REACTOR 0300-000.00S-TRP001- 2.3.2 17 05/10/96 1.B TRIP OR SAFETY 01 INJECTION AND INTRODUC110N TO THE USE OF EOPs ReactorTrip Or Safety 1-EOP-TRIP-1 CAS Sheet i 20 Injecuon Qasstion Source New Question Modification .

Method ' 1 Question Source Coenments: l Material Required for Fumpdmation:

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r Ques 6es Topict l Status Tree monitoring Given the following conditions for Unit 1: -

- Reactor trip due to loss of coolant accident and partial loss of feedwater.

- All RCPs have been tripped.

- Containment Spray has actuated.

The STA notes the following:

?

o Intermediate Range NIs - 10E-03 Amps,-0.3 dpm o . CSTs-580 *F, rising slowly  ;

o RVLIS Full Range - 88% level o All SG NR Ievels-Off-scale Low o - Aux FeedwaterFlow-24E041bm/hrtotal o Containment Pressure - 16 psig, slowly lowering o Pressurizer level- Off-:cale Low Which one of the following correctly identifies the monitoring frequency required for the Critical Safety Function Status Trees?

a. Continuous.

b. Every 5 minutes.

c. Every 15 minutes.

d. Every 30 minutes.

Ans: Ia l Exam Level: lS l Cognidve Level: l Comprehension l Fvpla== tion Once EOP-CFST-1 has been initiated, the fo!!owing rules apply to usage of the EOP r-twork: 1) Monitoring ofAnswer shocid be continuous if any RED or PURPLE condition found to exist, or 2) Monitora g irequency may be reduced to 15 minutes if all CFST indicate YELLOW or GREEN Containment pressure is sti"? in a PURPLE Path.

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Tier: l Generic Knowledge and Abilities l RO Grr.up: l1 l SRO Group: l1 Systesm/ Evolution Nunnber: l l Systenn/ Evolution

Title:

l Catemory: l 2.4 l Emergency Procedures / Plan KA: 2.4.21 Knowledge of the parameters and logic used to assess the status of safety functions including: 1.

Reactivity control 2. Core cooling and heat removal 3. Reactor coolant system integrity 4. Containment condidons 5. Radioactivity release control.

R3D Value: l3.7 l SRO Value: l4.3 l CFR: l 43.5 /45.12

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Reference Reference Number Reference Sectica Fase NuWs) Revision Learn. Ob]

EOP-TRIP-1, REACTOR 0300-000.00S-TRP001- 3.3.4 28 05/10/96 7,10 TRIP OR SAFETY 01 INJECTION AND INTRODUCTION TO THE USE OF EOPs USE OF PROCEDURES SC.OP-AP.ZZ 0102(Q) 5.3.12.G 20 5 Question Source ' NeO Question Modification Method Question Source Connnents: l Material Required for 1-EOP-CFST-1, if required Exannnation:

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Queseles Tesict l Misoperation of task Selector Switch

• Given the following conditions for Unit 2:

o Reactor Power- 25%-

o Control Rods are at step 150 on Control Bank D o The Bank Selector Switch is currently in MANUAL '

When at step 140 on Control Bank D, the operator felt these may be a problem with rod control. As part of  ;

troubleshooting, the operator moved Control Bank D out 10 steps with the Bank Se ector Switch in CONTROL BANK D position. The operator then placed the Bank Selector Switch in MANUAL The  !

troubleshooting showed that no problem existed with rod control. ,

Which one of the following conectly identifies the :ffect of this action on the rod control system if operation continues? I

a. On a controlled shutdown, the ROD BOTTOM alarm (OHA E-48) will actuate 10 steps sooner thaa expected.

b. On a controlled shutdown, the overlap between Control Bank D and Control Bank C will be 10 steps less than normal.

c. While operating, the Bank Demand Position Indication will read 10 steps lower than the Bank Individual Rod Position Indication. ,

d. While operating, the Rod Insertion limit alarms (OHA E-8 and E 'i6) for Control Bank D would actuate 10 steps lower than the actual alarm setpoint positions.

Aas: lb l Exass Level: lS l Cognitive level: 1 Comprehension l

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Re * - ^' E If a Control bank is MOYed individually it muSt be returned to Original poSiliOG prior 10 going 10 AUTO or i ofAnswer MANUAL, Otherwise with bank overlap counter inhibited, bank overlap unit will lose track of bank position. In l this case a 10 step difference will exist from the " normal" overlap program.

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Tiers l Plant Systems l RO Group: l1 l SRO Group: l1 Systema / Evolution Number: l001 l Systenn/ Evolution

Title:

l Control Rod Drive System Cab..- j: K4 Knowledge of Control Rod Drive System design feature (s) and or interlock (s) which provide for the i following:

l KAt l K4.02 l Control rod mode seleck control (movement control)

R7 Value: l3.8 l SRO Value: l3.8 l CFR: l 41.7 Reference Reference Number Reference Section Pate Number (s) Revision Learn. Obj ,

ROD CONTROL AND 0300-(XX).00S- IV.B.8.f.7) 39 12/11/96 6.g.iv '

POSITIONINDICATION RODS 00-00 SYSTEMS l

Question Source New Question Modification Method Questior. Saurce Comment: l-Material Required for Exandnation:

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Quaselen Topic: l RCS cooldown Given the following conditions for Unit 2:

o RCS cooldown and depressurization in progress in accordance with o ' S2.OP-lO.ZZ-0006 " Hot Standby To Cold Shutdown" i

o Recent RCS Cooldown data: '

o Imag. RCS Temo RCS oressure -

0130 430*F 1225 psig l 0200 400*F 1200 psig 0230 370*F 1200 psig Which one of the following correctly identifies the lowest allowable temperature for the RCS at 0300 if RCS

. pressure remains at 1200 psig?

., 1 a.' 340*F. > I l

l b. 313*F. I

c. 300*F.

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d. 238'F.

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Ib l Exasa I4 vel: lS l Cosaitive leva: l Application l Explanation See curve. Limiting curve is the POPS curve which has to be placed in service when any RCS cold leg temp = 312 l ofAmewer F. Otherwise maximum allowed cooldown - 100*F in any I hour; max heatup - 60*F in any one hour and cooldown curve operational limit at 238*F.

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Tier: l Plant Systems l RO Group: l2 l SRO Group: 12 S,.' Tvolution Number: l 002 l Systena/ Evolution Tide: l Reactor Coolant System Ca'.m i: l A3 l Ability to monitor automatic operations of the Reactor Coolant System including:

KA: l A3.03 l Pressure, temperatures, and flows R@ Value: l 4.4 l SRO Value: l4.6 l CFR: l 41.7 /45.5 Referwace Reference Number Reference Section Page Number (s) Revision Learn. Obj REAuun COOLA!G 0300-000.00S.RCS000- Vill.L 55 05/07/96 10 SYSTEM 02 REACTOR COOLANT S2.OP-TM.ZZ-0001(Q) 2.2.4 & RCS Pressure / 2&4 2 SYSTEM PRESSURE- Temperature Curves TEMPERATURE CURVES AND RCS/ PRESSURIZER HEATUP/COOLDOWN PLO13 Salem - Unit 2 Technical 3/4.4.10;3.4.10.3 3/4 4 4-29; Specifications 3/4 4-31 Question Source New Question Modification Method Question Source Comments: l Material Required for S2.OP-TM.ZZ-0001(Q): RCS Pressure / Temperature Curves, page 6 Exandmation:

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W Topic l Effect of starting an RCP The following conditions exist:

o RCS pressum- 335 psig o RCS temperature - 340 'F o Steam Generator pressure - 50 psig o A bubble exists in the Pressurizer o No RCP's are running Which one of the following cormctly describes the initial msponse of RCS temperatum and pressure if a Reactor Coolant Pump is started?

RCS RCS TEMPERATURE PRESSURE

a. RISE RISE

b. RISE LOWER

c. LOWER RISE

d. LOWER IDWER Ans! ld l Exam Level: lS l Cognitive Level: l Comprehension l Explanation SG temperature equivalent 298 F. RCS would cooldown resulting in a pressure drop when forced flow initiated.

of Answer Page 39

l Tier: l Plant Systems l RO Group: l1 l SRO Group: l1 Systesn/ Evolution Number: l 003 l System / Evolution

Title:

l Reactor Coolant Pump System l Category: Al Ability to prediu and/or monitor changes in parameters associated with operating the Reactor Coolant Pump System controls including: 1 1

KA: lA1.07 l RCS temperature and pressure l

RO Value: l 3.4* l SRO Value: l3.4 lCFR: l 41.5 /45.5 l Reference Reference Number Reference Section Page Number (s) Revision Learn. Obj ,

REACTOR COOLANT 0300-000.00S- IX.A.3.e 50 06/10/96 10.d,12 l i PUMP RCPUMP-01 REACTOR COOLANT S2.OP-SO.RC-0001(Q) 3.2.8 4 14 PUMP OPERATION l l

Question Source NRC Exam Bank Question Modification ,

Method i Question Source Comments: l l Matedal Required for Steam Tables Exandnation:

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l Questies Tende; l BATTS Operability i l

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Given the following conditions for Unit 2:

o RCS temperature 400 *F )

c o RWST concentration -2350 ppm  ;

o 21 BAT concentration-6750 ppm ,

i o 22 BAT concentration-6750 ppm j o 22 BATlevel-37%  ;

J Which one of the following correctly describes the MINIMUM level for 21 B AT that meets or exceeds the operability requirements for the BAT? i

a. 55%.

b. 62%. ,

c. 92%.

d. 95%. ,

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Amst lb l Exasa level: lS l Cognitive level: l App'ication l Explaantion Per Figure 3.1-2, the BAT system (21 + 22 BATS) level must be equal to or greater than 94.25% @ RWST boron  !

of Answer conc. of 2350 ppm & BAT boron conc. of 6750 ppm. BAT system req = 21 BAT level + 22 BAT level; 94.25 -

37 = 57.25. 62% is closest that meets req. l l

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Tiers l Plant Systems l RO Group: l1 l SRO Group: l1 Systenn/ Evolution Number: l004 l System / Evolution

Title:

l Chemical and Volume Control System Catemory: l2.1 l Conduct Of Operations KA: l 2.1.12 l Ability to apply technical specifications for a system.

RO Value: l2.9 l SRO Value: l4.0 l CFR: l 43.2 /43.5 /45.3 Reference Reference Number Reference Section Paac Number (s) Revision Learn. Obj CHEMICAL AND 0300-000.00S- VIII.A.2.c 20 1004/96 10 VOLUME CONTROL CVCS00-00 SYSTEM Salem-Unit 2 Technical 3.1.2.6.a 3/41-12 & 12 (a) Amend.

Specifications No.151 Question Source New Question Modification Method Question Source Comunents: l Material Required for Technical Specification Figure 3.1-2, pg. 3/4 1 12(a)

Exandmation:

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Quaselos Tosiez l Charging Pump auto start Given the following conditions for Unit 2:

o Unit 2 tripped from 100% power due to a loss of offsite power o 23 Charging Pump was running prior to the event.

o Following the reactor trip, an automatic Safety Injection signal was, generated on SSPS Train A only Which one of the following correctly describes the status of the CVCS Charging Pumps and supporting ,

ung.ts following completion of the SEC Mode Op and before SEC reset?

a. All THREE Charging Pumps will be running, and the aux. oil pumps for the 21 and 22 Charging Pumps will be running.

b. Only 21 and 23 Charging Pumps will be running, and the aux. oil pump for the 21 Charging Pump  ;

will be running.

c. ' Only 21 and 22 Charging Pumps will be running with their aux, oil pumps de-energized.

d. Only 21 Charging Pump will be running with its aux. oil pump de-energized.  ;

Amst lc l Exam Level: lS l Cognitive level: l Comprehension l R ," " Mode I, SI Only: both trains of SSPS send a SI signal to each SEC Cabinet. Derefore, each SEC will operate in ofAnswer h

MODE I. SEC Mode 2, Blackout: if at least 2 of 3 Vital Buses indicate a UV/ Blackout condition exists, either by  ?

operation of the specific sustained degraded voltage relays or the UV relays, all three SEC Cabinets will initiate a Mode II Loading Sequence. Herefore, with Mode I and Mode II conditions existing simultaneously, SEC will initiate a Mode III Logic Sequence. For CVCS: 23 Charging Pump motor is tripped by SEC tripping signal in e Modes 2,3 & 4. When pumps are started on a SEC signal, manual operation (start / stopping) is locked out until SEC is reset. Although Charging Pump start /stop signals also supply start signals to auxiliary lube oil pump, the SEC blocks aux. oil pumps operation since 230VAC bus is load shed.

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Tier: l Plant Systems l RO Group: 11 l SRO Group: 11 Systeun/ Evolution Nuanber: l 004 l System / Evolution

Title:

l Chemical and Volume Control System Cat-:--ri: l K2 l Knowledge of electrical power supplies to the following:

KA: l K2.03 l Charging pumps RO Value: l3.3 l SRO Value: l3.5 l CFR: l 41.7 Reference Reference Number R ferenceSection Paac Number (s) Revision Learn. Obj CHEMICAL AND 0300-000.00S- T v.C.20.c.6); 44;46-47 10/04/96 6 VOLUME CONTROL CVCS00-00 IV.C.20.d.2) & 8).e)

SYSTEM SAFEGUARDS 0300-000.00S-SEC000- IV.C.2.a. IV.C.3.b, 16-18 5/29/96 4,5 EQUIPMENT CONTROL 00 IV.C.S.a SYSTEM Q ;A Source

. New Question Modification Method Question F~srce Comments: l Materia. Required for  ;

Examination:

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Questies Topic: l Start of 2nd RHR Pump Given the following conditions for Unit 2:

o~ RCS temperature 185 *F o RCS pressure -250 psig o 21 RCP is running o 21 RHR pump is in service o 22 RHR pump is to be placed in service Which one of the following correctly describes why 2CV18, Ietdown Heat Exchanger Outlet, should be f adjusted immediately following the start of 22 RHR pump?

a. Prevent overpressurization of the letdown line and demineralizers.

b.' Prevent fill of the pressurizer and subsequent pressurization.

c. Reduce the RHR discharge pressure to prevent lifting of the SJ48 relief valves.  :

d. Prevent RCP seal differential pressure from dropping below the minimum required for operation.

) Ams: Id l Exam Level: lS l Commitive Level: l Comprehension l F ,"- "'-

When second pump is started, an increase in flow causes an initial drop in RCS pressure, affecting the minimum >

of Answer pressure /Ap for RCP #1 seal ops. CV18 is operated to control RCS pressure.

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i Tier: l Plant Systems l RO Group: l3 l SRO Group: l3 j Sy t. Tvolution Nunnber: l 005 l Systenn/ Evolution

Title:

l Residual Heat Removal System I Cckgrf: Al Ability to predict and/or monitor changes in parameters associated with operating the Residual Heat Removal System controls including:

KA: l A1.02 l RHR flow rate

, RO Value: l3.3 l SRO Value: l3.4 l CFR: l 41.5 /45.5 Reference Reference Nuniber Reference Section Pcae Nunnber(s) Revision Learn. Ob]

, RESIDUAL HEAT 0300 000.00S- IX.A.I 53 4/23/96 12 d

REMOVAL SYSTEM RHR000-01 INITIATING RHR S2.OP-SO.RHR- 3.7; 3.8 3 9 0001(Q)

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l Question Source NRC Exam Bank Question Modification 1

Method i Question Source Comunents: l 2

Material Required for Ew==I== tion:

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Quessies Tonde: l Accumulator A reactor shutdown is in progress due to an RCS leak into 23 SI Accumulator. The following conditions exist on the Unit:

o . RCS Tavg - 525 'F o RCS piessure- 1950 psig o SI Accumulator Isolation Valve 23SJ54 is closed and energized with VALVE OPERABIE position selected on Panel 2RP4 In accordance with the UFSAR accident analysis, which one of the following states the response of the ECCS Accumulators if a Design Basis IDCA occurs on the 22 Imop Cold Ieg, at this time?

a. 23SJ54 automatically opens and all FOUR Accumulators will inject into the RCS and provide core cooling

b. 23SJ54 automatically opens and THREE Accumulators,21,'23 and 24, will inject into the RCS and pmvide core cooling

c. TWO Accumulators,21 and 24, will inject into the RCS and provide core cooling, unless the operator depresses the OPEN pushbutton for 23SJ54

d. THREE Accumulators,21,22 and 24, will inject into the RCS and provide core cooling unless the operator depresses the OPEN pushbutton for 23SJ54 Ams: lb l Examlevel: lS l Commitive level: l Comprehension l

^ ' -

F_; " -

23SJ54 will open on SI signal since it is energized. 22 Accumulator will spill to Containment floor per design ofAnswer basis.

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Tier: l Plant Systems l RO Gic,;. : l2 l SRO Giwi,: l2 SystenvEvolu8 ion Nenhber: l 006 l SystenvEvolution

Title:

l Emergency Core Cooling System Category: K6 Knowledge of the of the effect of a loss or malfunction on the following will have on the Emergency Core Cooling System:

KA: l K6.02 l Core flood tanks (accumulators)

RO Value: l3.4 l SRO Value: l3.9 l CFR: l 41.7 / 45.7 Reference Reference Number Reference Section Revision Pase Numbeds) Learn. Ob]  ;

EMERGENCY CORE 0300-000.00S- IV.E.3.c; IV.E.3.e.1); 35-38 10/09/96 4.e; 8 COOLING SYSTEM ECCS00-00 IV.E.4.c.1)

No. 2 Unit Safety injection 205334 D-7 sh.4 41 P&ID 23SJ54 Logic Diagram 239938

, Question Source New Question Modification Method Question Source Comments: l Material Required for '

Eumadnation:

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Questies Topic: l Evaluation of PRTinleakage '

During performance of the procedure for forming a bubble in the pressurizer, the pressurizer is filled as indicated on the cold calibrated channel. Heaters are then energized, pressure is controlled at approximately i

65 psig, and when pressurizer temperature reaches approximately 300 F, PR1 and PR2 are manually opened .

for 10-15 minutes.

• W .ach one of the following correctly describes the reason for opening PRI and PR2?  !

a. . Establishes flow from the RCS into the pressurizer to ensure boron concentrations are equalized j b. Provides a path for venting non-condensible gases out of the pressurizerduring bubble formation

c. Verification that the PORV tailpipe temperature device will respond to changes in temperature j d. Provides assurance that pressure-temperature limits will not be exceeded during bubble formation Ams: lb l Exam level: IS l Camative Level: l Memory l j F ;"- " --

The procedure states that the purpose of the step is to vent gases. The RCS and PZR are filled from the same  !

, ofAnswer source so there should be little difference in boron concentration and the plant has not reached a condition where boron concentration differences are a concern. Tailpipe temperature response can be verified by means other than

fully opening both PORV's. Automatic overpressure protection could be provided by POPS without manually

opening both PORV's.

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Tier: l Plant Systems l RO Group: l3 l SRO Group: l3 Systean/ Evolution Number: l007 l System / Evolution

Title:

l Pressurizer Relief Tank / Quench Tank System Cc y .y: K5 Knowledge of the operational implications of the following concepts as they apply to the Pressurizer Relief Tank / Quench Tank System:

KA: lK5.02 l Method of forming a steam bubble in the PZR R3 Value: l3.1 l SRO Value: l3.4 l CFR: l 41.5 /45.7 Reference Reference Number Reference Section Pase Number (s) Revision Learn. Obj PRESSURIZER AND 0300-000.00S- IV.B.8.a.2) 27 4/16/96 12 PRESSURIZER RELIEF PZRPRT-00 TANK REACiOR COOLANT 0300-000.00S-RCS000- IX.A. I .d 57 5 0/96 12.c SYSTEM 02 FILLING AND VENTING S2.OP-SO.RC-0003(Q) 5.6.9-5.6.10 17-18 THE REACTOR COOLANT SYSTEM Questien Source New Question Modificatidn- - ~ ~

Method Question Source Comments: l l

Mitedal Required for Exandnation: ,

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Ques 6es Topic: l Power supply indication Which one of the following correctly describes how the Component Cooling Water Pumps respond to the indicated SEC Mode operation?

a. In Mode II, all pumps start; including any selected to LOCAL at the HSD Panel

b. In Mode III, all pumps start; except any selected to LOCAL at the HSD Panel

c. In Mode I, running pumps remain in service. Any non-running pump shifts to AUTO unless it i::

selected to LOCAL at the HSD Panel i

d. In Mode II*, the pump on the affected bus shifts to AUTO unless it is selected to LOCAL at the HSD Panel  !

Ans: la l ExamiIevel: IS l Cognitive Level: l Comprehension l F=7' - Safeguards start signals are independent of the LOCAL / REMOTE switch position ofAnswer _

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Tier: l Plant Systems l RO Group: l3 l SRO Group: l3 Systema / Evolution Nuanber: l 008 l Systema / Evolution

Title:

l Component Cooling Water System Category: l K2 l Knowledge of electrical power supplies to the following:

KAt l K2.02 l CCW pump, including emergency backup RO Value: l 3.0* l SRO Value: l 3.2* l CFR: l 41.7 Refermice Reference Nunnber Reference Section Page Nunnber(s) Revision Imrn. Obj COMPONENTCOOLING 0300-000.00S- V.A.I.f; V.D.7.a & c 32 1/15/97 7.c; 8.a & d WA'll:.R CCW0004) 23 CCW Pp logic Diagram 224399 Question Source 1/97 NRC Exam Question Modi 5 cation Omnged correct answer and modified (Modified) Method distractors by adding LOCAL switch position Question Source Conaments: l Macedal Required for Evandmation:

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Questies Topic: 1 Spray valve open Given the following conditions for Unit 2:

i o Reactor power 24%

o Pressurizer pressuin - 2150 psig and lowering very slowly o The MASTER Pressure Controlleris functioning properly ,

o Pressurizer spray valve 2PS3 is partially open and will NOT I respondin AUTO orMANUAL i

Assuming a containment entry is not possible, which one of the following correctly describes the general path of the procedural actions?

5

a. Raise charging and reduce letdown; enter TSAS 3.2.5-restore the parameter within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be

<5% power within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> l

, b. , M,anually block the lo,w pressure ,SI and stop ,23 {tCP , , , ,

c. Stop 23 RCP and initiate a plant shutdown

d. Manually block the low pressure SI, stop 23 RCP and initiate a plant shutdown  ;

Anst lc l Fu== Level: IS l Cognitive Level: l Comprehension l Fwpt=== tion 'Ihe 23 RCP provides flow through 2PS3 spray valve. If power level allows, the RCP should be tripped and a i ofAnswer shutdown initiated within one hour once SG Icvels stabilized. Choice a. is not aligned with conservative action philosophy. Choices b. and d. involve blocking a valid SI signal.  ;

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Tier: l Plant Systems l RO Genup: 12 l SRO Group: l2 Systesm/ Evolution Nunnber: l010 l System / Evolution

Title:

l Pressurizer Pressure Control System Cctegory: A2 Ability to (a) predict the impacts of the following on the Pressurizer Pressure Control System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:

KAg l A2.02 l Spray valve failures RO Value: l3.9 l SRO Value: l3.9 l CFR: l 41.5 /43.5/45.3/45.13 Reference Reference Number Reference Section Pase Number (s) Revision Learn. Ob]

PRESSURIZER PRESSURE 0300-000.005- IV.B.I.k; IX.B.1 23-24;40 3/20/97 4.e; 12 AND LEVEL CONTROL PZRP&I 00 PRESSURIZER PRESSURE 0300-000-00S- Steps 16,20-24,32-38 15-17;19-20 7/15/97 3.c MALFUNCI1ON ABPZR1-00 PRESSURIZER PRESSURE S2.OP-AB.PZR- 3.28 - 3.32 5 MALFUNCTION 0001(Q)

Question Source New Question Modification Method Question Source Comments: l Material Required for Examination:

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Onesties Topic: l Przrlevelcontrol Given the following conditions for Unit 2:

o Reactor power-40% ,

o Pressunzerlevel- 33%

o- Pressurizer master level controller in MANUAL o Charging and letdown flow am balanced Which one of the following correctly identifies what will happen to Pressurizer level if reactor power is raised to 100% with NO operator adjustment to the Master 12 vel Controller?

a._ Rises to approximately 50%.

b. Remains at appmximately 33%.

. c. Rises to the high level reactor trip setpoint.

d. Iowers to 17%, then rises to the high level reactor trip setpoint.

Amst - l a l Exam level: lS l Connitive Level: l Comprehension l Explanation RCS volume increases proportionally from 0% - 100% power. With balanced letdown & charging, Pzr level of Answer , should nearly follow the normal level program.

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A A - - , - s e .- A., w: a e _ km .s .a4_., ..,.a _aw . a_a_ ..

'Ders l Plant Systems l RO Group: l2 l SRO Group: l2 Systeam/ Evolution Nuseber: l 01I l Systena/Erolution

Title:

l Pressurizer Level Control System Category: ' K5 Knowledge of the operationalimplications of the following concepts as they apply to the Pressurizec Level Control System: I KAt lK5.12 l Criteria and purpose of PZR level program n3 Value: l2.7 l SRO Value: l3.3 l CFR: l 41.5 / 45.7 Rderence Reference Number Reference Section Pase Number (s) Revision Learn. Obj PRESSURIZER PRESSURE 030(M)00.00S- II.B 14 3/26/97 1 AND!EVEL CONTROL PZRP&lAk)  !

REACmR COOLANT 0300-000.00S-RCS000- IV.C.1; IV.C.3.b 18-19 5 0/96 4.c SYS7EM 02 Q waa= Source NRC Exam Bank Question Modification '

Method OueMien Source Comuments: l Material Required for i

E===d== tion:

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Quesdes Tende: l Delta T mution change Which one of the following correctly describes how the steady-state overpower delta-T (OPAT) and the overte'.sperature delta-T (OTAT) setpoints change while raising power from 50% to 100%?

a. ' OTAT setpoint rises and OPAT setpoint lowers.

b. OTAT setpoint lowers and OPAT setpoint rises.

c. OTAT setpoint remains constant and OPAT setpoint lowers. '

d. OTAT setpoint lowers and OPAT setpoint remains constant.

Auss Id l Exasm Level: lS l Connitive Level: l Comprehension l l

FvpI=== tion OPAT sp = Dit) [ K4 - K5 [(t3 s / ( I + 13 s )] T - K6 ( T - TM )] & 100% OPAT is 108% of full power DT ,

ofAnswer and CANNOT be raised above this value; OTAT sp = DTO { K1 - K2 I(1 + 11s)/ (1 + 12s)](T-T') + K3 (P -

4 P')- f(I)) & 100% power setpoint for OTAT is 116.4% of full power AT. At lower AT setpoint is higher 4

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l Tier: l Plant Systems l RO Group: l2 l SRO Group: l2 Systema / Evolution Nussber: l 012 l Systena/ Evolution

Title:

l Reactor Protection System Category: l2.2 l Equipment Control KAt l 2.2.22 l Knowledge of limiting conditions for operation.t and safety limits.

RO Value: 13.4 l SRG Value: l4.1 l CFR: l 43.2 / 45.2 Referwace Referesree Number Reference Section Page Number (s) Revision Imrn. Obj REACIOR COOLANT 0300-000.00S- IV.C.1.h-i; IV.C.2.d-f 21 22:24 $n/96 7 TEMPERATURE RCIT.MP-00 INS'IltUMENTATION ,

Salem. Unit 2 Technical 2.2.1. Table 2.2-1, FU 7 2-4 9 Amend. l Specifications &8 No.147 Question Source NRC Exam Bank Question Modification Method Question Source Comuments: l Material Requ!W for Evand== tion:

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ousselee Tesic l Operation orseactor trip bypass breakers l

Given the following conditions for Unit 2: '

i o Mode 3 ,

o Testing on Reactor Trip Breaker A (RTA) is in progress  :

o RTA is racked in and closed {

o Reactor Trip Breaker B (RTB) is racked in and open l o Reactor Bypass Breaker A (BYA) is racked in and closed Which one of the following correctly describes what will occur if Reactor Bypass Breaker B (BYB) is rackedin with the above conditions?

a. Only BYB will trip open due to a GENERAL WARNING on both trains.

b. BYA will open due to breaker electrical interlock when BYB is closed.

)

c. BYA, BYB and RTA will trip open due to a GENERAL WARNING on both trains.

d. Both BYA AND BYB will open due to breaker electrical interlock, when BYB is closed. i Amst lc l Exasm Level: lS l Commitive level: Comprehension l l E ,* ^* --

General Warning Alarm System: GW active on 1) Reactor trip bypass breaker racked in; 2) Testing. Reactor trip is ofAnswer accomplished via an alarm relay. Deenergizes when a problem exists, opening a contact to UV coil. Alarm relay ,

from both trains must de-energize for circuit to open and trip reactor

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Tier: l Plant Systems l RO Group: l2 l SRO Group: l2 Systems / Evolution Number: l 012 l System / Evolution

Title:

l Reactor Protection System Category: K1 Knowledge of the physical connections and/or cause-effect relationships between Reactor Protection System and the following:

KAt lKl.03 lCRDS R3 Value: l3.7 l SRO Value: ltJ l CFR: l 41.2 to 41.9 / 45.7 to 45.8 Reference Reference Number Reference Section Pase Number (s) Revision Learn. Obj REACTOR PROTECTION 0300-000.00S- IV.C.4.d.4); VII.D.7 29;53 5/15/96 10;20 SYSTEM RXPROT-00 NA 1 & 2 Units Reactor 221051 sh.2 11 Protection System - Reactor Trip Signals logic Question Source New Question Modification Method Question Source Comments: l Material Required for Exandmation:

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Onassion Tesic l Seeandine signals Given the following conditions for Unit 2:

o Mode 3 with RCS cooldown and depressurization underway in i.cs -di-ce with procedural duection -

o RCS pressure- 1900 psig .

l c' RCS temperature-525 'F '

o SG pressures - 830 psig l

1 Which one of the following correctly describes the current status ESF actuation system if a Main Steam Safety Valve fails open, resulting in the following SG prr,rures: 820 psig (21); 780 psig (22); 700 psig

(23); 810 psig(24)?

a. No ESF signal has been generated.

b. Only a Safety Injection signal has been generated.

c. Only a Main Ste m Line Isolation signal has been generated.

d. A Safety injection signal and a Main Steam Line Isolation signal have been generated.

Ans: lb l Fma-level: lS 1 Cognitive Level: l Camarehension l Explomation A SI signal is generated for Steam Line High Differential Pressure when the setpoint of 100 psid is reached. A SI ofAnswer signal due to High Steam Line Flow Coincident with either Low Steam Line Pressure or Low-Low Tavg are normally manually blocked to allow for controlled depressurization/cooldown of the SGs (at P-12). The Steam Line Isohtion is NOT generated because High Steam Flow coincident with low steam header pressure is well above the required setpoint of 600 psig.

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1 Tier: l Plant Systems l RO Group: l1 l SRO Group: l1 l

Systeam/ Evolution Number: l 013 l System / Evolution

Title:

l Engineered Safety Features Actuation System )

Category: K4 Knowledge of Engineered Safety Features Actuation System design feature (s) and or interlock (s) which i provide for the following:

KA: l K4.03 l Main Steam Isolation System

)

R3 Value: l3.9 l SRO Value: l4.4 l CFR: l 41.7 Reference Reference Number Reference Section Pate Number (s) Revision Learn. Obj i INTRODUCTION TO 0300-000.00S-ESF000- VII.B.I. B & c 50 6/25/96 21 ENGINEERED SAFETY 00 FEATURES AND DESIGN CRITERIA RPS - Steam Generator Trip 221056 sh.7 7 Signals Logic RPS-Safeguards Actuation 221057 sh.8 l Signals '

Question Sourre New Question Modification Method .

Question Source Comments: l l Mate ial Required for E===dnation: ,

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omoseles Topic: l ESF actuanon logic with channel OOS i

! Given the following conditions for Unit 2: {

3 o ' Containrnant pressure transmitter (f'%aa-1 H) has been

. declaredinoperable

{

o (%nnel II feeds both HI and HI-HI Containment Pressure e o Actions are complete for placing the channel in a tripped condition 'per i S2.OP-SO.RPSJ)005(Q), " Placing Containment Pressure (%nnel In Tripped Condition" Which one of the following correctly describes the effect this action has on the ESF/ Safeguards actuation coincidence associated with containment pressure?

a. Both Spray Actuation and Safety Injection become 1/3  !

b. Spray Actuation.becomes 2/3; Safety Injection becomes 1/2

c. Both Spray Actuation and Safety Injection become 2/3

d. Spray Actuation becomes 1/3; Safety Injection becomes 1/2 Ams2 lb l Rummi tmel: lS l Cognitive Level' l Memory l Explanation *Ihe normal coincidence is 2 out of 4 channels :.or CNMT Spray and 2 out of 3 for SI. When the channel is ofAnswer removM from service, the associated Si bistab e is tripped reducing the coincidence for operable channels.

However, the CNMT Spray bistable is bypasse.d (one is allowed to be bypassed per TS).

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Tier: l Plant Systems ._

i RO Group: l1 l SRO Group: l1 Systean/ Evolution Number: l 013 l Systent/ Evolution

Title:

l Engineered Safety Features Actuation System Cat.g, y: K6 Knowledge of the of the efrect of a loss or malfunction on the following will have on the Engineered Safety Features Actuation System:

KA: l K6.01 l Sensors and detectors ,

RO Value: l 2.7* l SRO Value: l 3.l* l CFR l 41.7 /45.7 Reference Reference Number Reference Section Page Number (s) RaMon Learn. Obj REACTOR PROTECTION 0300 @ 0.00S- XI.A.5 58-59 5/1566 12,15 SYSTEM RXPROT-00 INTRODUCTION 7D 0300000.00S-ESF000 VI.B.I .d; VI.B.3.b; 51-52 6/2566 21 ENGINEERED SAFETY 00 VI.B.4.b FEATURES AND DESIGN I

CRITERIA PLACING CONTAINMENT S2.OP-SO.RPS- 5.2 3 PRESSURE CHANNELIN 0005(Q)

TRIPPED CONDITION Question Source New Question Modification Method Question Source Conunents: l Material Required for Exandnation:

P. age 64

.a Question Topic: l RCD BOTivM Which one of the following choices correctly completes the following statement?

During a reactor startup, ROD B01 TOM (OHA E-48) will actuate for a dropped rod in each control bank...

a. when Control Bank A demand position is above 20 steps.

b. only after Control Bank D demand position is above 35 steps.

c. when Control Banks A, B and C demand position is above 20 steps, and for Control Bank D when .

its demand position is above 35 steps.

d. when Control Bank A demand position is above 20 steps, and for Control Banks B, C and D when ,

their demand position is above 35 steps.

Ans; ld l Exam Level: lS l Cognitive level: l Memory l Explanation Rod Bottom Bistable Modules provide output signal with a setpoint of 20 steps. OHA FA8 ROD BOTTOM will ofAnswer be received if any control bank A rod drops from a position >20 steps (from demand position indicator) or if any control bank B C or D bank is dropped from a position >35 steps (from demand position indicator).

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I Ber: l Plant Systems _ l RO Group: l2 l SRO Group: l1 SystemvEvolution Number: l 014 l System / Evolution Male: l Rod Position Indication System Category: l 2.4 l Emergency Procedures / Plan KAt l2.4.31 l Knowledge of annunciators alarms and indications. and use of the response instructions.

RO Value: l3.3 l SRO Value: l3.4 l CFR: l 41.10 / 45.3

~Refersace Reference Number Reference Section Pase Number (s) Revision Learn. Obj ROD CONTROL AND 0300-000.00S- IV.B.11.a.2).c) 43 12/11/96 6.1; I 1.e POSITIONINDICATION RODS 00-00 SYSTEMS l OVERHEAD S2.OP-AR.ZZ-0005(Q) Window E-48 67 12  !

ANNUNCIATORS WINDOW E l

l Question Source NRC Exam Bank Question Modification l M etteod

,Q uestion Source Comuments: l Matedal Required for Er==d=mtion:

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Quaosies Topic: l PRNIS Channel Comparator operatian l

j Load is being raised from 40% power to 100% power on Unit 2.

l Which one of the following choices conectly completes the following statement?

l l The Power Range NIS Channel Comparator Circuit will provide an UPPER SECT. ION DEVIATION (OHA j E-38) alarm when a 2% deviation exists between...

I l

l a. the highest and the lowest normalized upper sectiou current on 1 out of 4 channels.  :

b the highest and the lowest normali=d upper section curtent on 1 out of 4 channels, efter power is  :

greater than or equal to 50%.

1

c. the hig!est and the average normalized upper section current on 1 out of 4 channels.

l

d. the highest and the average normalized upper section cunent on 1 out of 4 channels, after power is greater than or equal to 50%.

Ans: ld l Erman Level: IS l Conn'tivs Level: l Comprehension I _

Frphmation The Detector Current Comparatokare used to monitor the fluk distribution in the core durmg ogwrations at greater ofAaswer than 50% power, to alert the operator of a Radial flux tilt. 'Ihis is accomplished by comparing the output current from each detector to the average output cunent from all four detectors. The alarm furstion is automatically defeated below 50% power on all four detectors.

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Tier: l Plant Syttems l RO Group: l1 l SRO Group: l1 Systan/ Evolution Nun 6er: l 015 l System / Evolution 'Dtle: l Nuclear Instrumentation System Cr tesory: l A3 l NUity to monitor automatic operations of the Nuclear Instrumentation System including:

KA: l A3.03 l Verification of proper functioning / operability 60 Value: l3.9 l SRO Value: l3.9 l CFR: l 41.7 /45.5 Reference Reference Number Reference Section Pase Number (s) Revision EXCORE NUCLEAR 0300-000.00S-Learn. Ob[

IV.E.3.c. 35-36 9/4/96 5.g INSTRUMENTATION EXCORE-00 SYS'E M OVERHEAD S2.OP-AR.Z7 0005(Q) Window E-38 56 12 ANNUNCIATORS WINDOW E Question Source l NRC Exam Bank Question Modification i Metbod Question Source Comments: l Material Required for Examination:

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Question Topic: l Actual vs. calorimetric power _

During performance of S2.RE-ST.Z7e0001(Q) " Calorimetric Calculation", The feedwater temperature points utilized were reading 10 *F higher than actual feedwater temperature. Power range NI's were adjusted in accortlance with the directions of the calorimetric procedure.

Which orae of the following correctly describes the effect of the NIS adjustment?

E

a. Indicated power is less than actual power, therefore, power range instruments are set conservatively,

b. Indicated power is less than actual power, thwefore, power range 'nstruments are set non-conservatively,

c. Indicated power is greater than actual power, therefore, power range instruments are set comervatively.

d. Indicated power is greater than actual power, therefore, power range instruments are set non-conservatively.

Ams- lb l Eumm level: lS l Cognitive Level: l Application l E" ^':: Qsec = m (hs-hf)- mhs(SGBD). Qprim = Qcore - QRCPs. At equilibrium Qprim = Qsec. If feedwater temp is of N wer read higher, then hfis higher and a lower power level than actual is calculated. 'Iherefore, the NI's were set at a lower than actual value. Since trip setpoints do NOT change, the reactor would trip with actual power greater than the trip setpoint, which is non-conservative. _

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Tier: l Plant Systems l RO Group: l1 l SRO Group: l1 Systesa/ Evolution Number: l 015 l System / Evolution

Title:

l Nuclear Instrumentation System Category: K5 Knowledge of the operational implications of the following concepts as they apply to the Nuclear Instrumentation Synem:

KA: lK5.04 l Factors affecting accuracy and reliability of calorimetric calibrations RO Value: l2.6 l SRO Value: l3.1 l CFR: l 41.5 / 45.7 Referwece Reference Number Reference Section Pate Number (s) Revision I4arn. Ob]

EXCORE NUCLEAR 0300-000.00S- X.A.I 78 9/4/96 14 INSTRUMENTATION EXCORE-00 SYSTFJ4 f

Question Source NRC Exam Bank Question Modification Method Question Source Comme.its: l Material Required for Exasmination:

i Page 70

Question Topic: l temperature rise Unit 2 is operating at 100% power with CET Display for the hottest in-core thermocouple reading 610 F.

Temperature in the area of the Reference Junction boxes for the thermocouples rises 30 *F over the shift.

Reactor power level remains constant over the shift. '

Which one of the following correctly describes how core exit thermocouple (CET Display) readings are affected by the temperature chenge in the area of the reference junction' boxes?

a. Will read higher due to higher voltage differential between the metals at the cold junction.

b. Will read lower due to lowered voltage differential between the metals at the hot junction.

c. Will remain the same because the temperature error is compensated for by the CET Processor System.

l d. Will remain the same since temperature change does NOT affect the signal from the reference junction.

Ans- le l ExamI4 vel: lS l Cognitive Level: l Memory l

, Explanation There are 2, Class IE Reference Junction boxes. Reference Junction boxes are located in Panels 1020-1 and ofAnswer 1021-1 in the Process and Equipment room. One function of each Reference Junction Box (and associated components) is to measure the cold junction temperature with a semiconductor temperature probe, and provide that temperature signal to the CETPS so that the temperature error introduced by the cold jura: tion may be compensated.

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'Ilers l Plant Systems l RO Group: l1 l SRO Group: l1 SystemvEvolution Number: l 017 l SystenvEvolution'Iltle: l In-Core Temperature Monitor System Category: Al Ability to predict and/or monitor changes in parameters associated with operating the In-Core Temperature Monitor System controls including:

KA: lAl.01 l Core exit temperature RO Value: l3.7 l SRO Value: l3.9 l CFR: l 41.5 /45.5 Refenece Reference Number Reference Section Page Number (s) Revision I4arn. Obj INCORE NUCLEAR 0300-000.00S- IV.D.2.c.2) 30 11/13/96 7.b INSTRUMENTATION INCORE-00 SYSTEM INCLUDING THERMOCOUPLES AND SATURATION MARGIN MONITOR i

Question Source NRC Exam Bank Question Modification Method Question Source Comments: l Matedal Required for <

Exandmation: '

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Queseles Topic: l CICU operational difTerences Procedums allow a Containment Fan Coil Unit (CFCU) to be run without Service Water (SW) as long as certain motor winding and bearing temperatures are not exceeded.

1 4

Which one of the following cormctly describes the speed of operation and purpose of mnning a CFCU without SW7 1

a. SLOW speed for containment air mixing following recovery from a LOPA and only one SW Pump is mnning

b. SLOW speed for containment air mixing if at least thme other CFCU's are in FAST speed, with one I SW header OOS and the unit at full power

c. FAST speed for containment atmosphere hydrogen mixing while the remaining CFCU's are in

' SLOW speed, following a LOCA 1 1

d. FAST speed when high airborne contamination levels or adverse atmospheric conditions are determined (by HP) to exist inside containment in Modes 4,5 or 6 Anst ld l Exam level: lS l Cognitive level: l Memory l i

Expleantion A CFCU may be run without SW only when high sirborne contamination levels or adverse atmospheric conditions cfAnswer are determined by HP personnel. This is limited to Mode 4 and below. There are no conditions in the EOP's where CFCU's are run w/o SW. There is no point to running a CFCU in SLOW w/o SW, with the unit at power.

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Der: l Plant Systems l RO Group: l1 l SRO Group: l1 Systeen46veludon Number: l 022 l System / Evolution "Iltle: l Containment Cooling System C' ._--,: l A4 l Ability to manually operate and/or monitor in the control room:

KAt l A4.01 l CCS fans R'0) Value: l3.6 l SRO Value: l3.6 l CFR: l 41.7 /45.5 to 45.8 Rderence Reference Number Reference Section Pase Number (s) Revision IAern. Obj CONTAINMENT AND 0300-000.00S. III.H.I.g 70 5/13/96 8.c & 12 CONTAINMENT CONTMT-00 SUPPORT SYSTEMS CONTAINMENT S2.OP-SO.CBV-0001 Section 5.2, Note 8 9 '

VENTILATION OPERATION _

Question Soorte New Question Modification l Method Question Source Conunents: l Material Required for Ew=d== tion:

Page 74

r QA Topic: Securing containment spray The following conditions exist on Unit 2:

o A WCA is in progress and transition has been made to 2-EOP-LOCA-3 " Transfer To Cold leg Recirculation" o Containment Spray actuated due to high containment pressure and is currently in operation Which one of the following correctly identifies the conditions that allow stopping Containment Spray Pumps and containment spray flow?

a. BOTH pumps are stopped when RCS subcooling is above 0*F and containment pressure falls below 13 psig. This stops all containment s, ray flow. I

b. ONE pump is stopped when RCS subcooling is above oaf and containment pressure falls below 13 psig. The remaining pump is stopped when the RWST level reaches LO-LO level. This stops all ,

containment spray flow.

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c. BOTH pumps are stopped when RHR suction alignment to the sump is complete. Containment spray flow is stopped when RHR spray flow is stopped during system realignment per 2-EOP- '

WCA-4 " Transfer To Hot leg Recirculation".

d. ONE pump is stopped when RHR suction alignment to the sump is complete. The remaining pump is stopped when RWST level renches the LO-LO level. Containment spray flow is stopped when RHR spray flow is stopped during system realignment per 2-EOP-LOCA-4 " Transfer To Hot Leg Recirculation".

Anst Id l Exam Level: IS l Cognitive Level: l Comprehension I Expleantion During the Recirculation Phase of SI, the Containment Spray system is operated as so: 1) At RWST Lo-Lo LVL ofAnswer alarm stop remaining running CS pump & 2) Align for recirculation using RHR System and CS-36s to containment i spray headers. l I

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Tier: l Plant Systems l RO Group: l2 l SRO Group: lI Systean/ Evolution Number: l 026 l System / Evolution

Title:

l Containment Spray System Category: A2 Ability to (a) predict the impacts of the following on the Containment Spray System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:

KA: l A2.08 l Safe securing of containment spray when it can be done)

RO Value: l3.2 l SRO Value: l3.7 l CFR: l 41.5 /43.5/45.3/45.13 Refenace Reference Number Reference Section Pate Number (s) Revision Learn. Obj r CONTAINMENTSPRAY 0300-000.00S- IX.B.3 45 7/31/96 12 SYSTEM CSPRAY-00 EMERGENCY CORE 0300 000.00S- VIII.C.3.b; C.4; D.4.c 74-76 10/09/96 3.b COOLING SYSTEM ECCS00-00 Quesdon Source New I Question Modification '

Method j

Question Source Comments: l Matedal Required for Exandmation:

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_Questies Topic: l RWST/ Containment Spray relationship Given the following conditions for Unit 2:

{

o A LOCA has occurred o The Injection Phase of SIis in progress j

o 22 Containment Spray Pump failed to start t

Which one of the following correctly describes the plant response to this event? l I

a. The Safeguards Pumps will operate for a longer time with suction from the RWST before swapover  ;

to the containment sump. I

b. The higher pressure in containment will result in overpressurizing the RHR suction piping when swapover to the containment sump occurs.

c. A portion of the 22 RHR pump discharge flow from the heat exchanger must be diverted to provide flow through the affected spray header.  ;

l

d. Water level in the containment sump will be insufficient to supply the ECCS pumps when alignment >

of the systerns for cold leg recirculation is complete.

Aas- la l Exam Level: lS l Cognitive Level: l Comprehension l Explanation During the injection pnase of SI, the CS pumps take suction from the RWST and deliver borated water (mixed with ofAnswer sodium hydroxide) to the Containment atmosphere. As for the other ECCS pumps, cold water from the RWST is injected into the RCS Cold legs via Centrifugal Charging Pumps and Intermediate Head SI pumps, and Low Head RHR pumps will inject on a large break LOCA.

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Tier: l Plant Systems l RO Group: l2 l SRO Group: l1 Systesm/ Evolution Number: 1026 l Systean/ Evolution

Title:

l Containment Spray System Category: KI Knowledge of the physical connections and/or cause-effect relationships between Containment Spray System and the following:

KA: lKl.01 l FICS RO Value: l4.2 l SRO Value: l4.2 l CFR: l 41.2 to 41.9 / 45.7 to 45.8 Reference Reference Number Reference Section Page Number (s) Revision Learn. Obj CONTAINMENT SPRAY 0300-000.00S- IV.B.2.a.2) 18 7/31/96 3.b.iii SYSTEM CSPRAY-00 EMERGENCY CORE 0300-000.00S- IV.A.4.a 20 104)9/96 4.a COOLING SYSTEM ECCS00-00 l

Question Source New Question Modification Method J Question Source Comments: l

)

Matedal Required for E===Imation: 1 i

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Quesdes Tesde: l IRU wation The 21 Containment Iodine Removal Unit (RU) was started in preparation for a planned containment entry.

Which one of the following comctly describes the status of the RUs if a Safety Injection signal is actuated?

a. 21 RU continues to run and 22 RU is locked out.

b. 21 RU is tripped and locked out, and 22 IRU is locked out.

c. 21 RU continues to run and 22 RU starts on SEC Mode Operation.

d. 21 IRU trips then restarts on SEC Mode Operation, and 22 IRU starts on SEC Mode Operation.

Ams: Ib l Exame Level: lS l CWImel: l Memory l WM Fans are non-safety related and power is locked out by SEC Mode Ops.

ofAnswer Page 79

11ers l Plant Systems l RO Group: l3 l SRO Group: l2 SysteaWEvolution Number: l 027 l SystenvEvolution 'ntle: l Containment Iodine Removal System Category: l A4 l Ability to manually operate and/or monitor in the control room:

KA: l A4.01 l CIRS controls RS Vaine: l 3.3* l SRO Value: l 3.3* l CFR: l 41.7 / 45.5 to 45.8 Reference Reference Nunnber Reference Section Pase Number (s) Revision learn. Obj CONTAINMENTAND 0300-000.00S- IV.D.2.a.1).a) 78-79 5/13/96 8 '

CONTAINMENT CONTMT-00 SUPPORT SYSTEMS Question Source New Question Modi 5 cation Method Question Source Comments: l 1

. Material Required for l Frandantina:

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1 Quaseles Tosic: l Hyt.a;x EGon and control I A Q-basis LOCA has occuned on Unit 2. 1 l

Which one of the following correctly describes the major source of hydmgen in containment and when the system to control hydrogen is placed in service? l 1

a. h radiolytic decomposition of water is the major source of hydrogen early in the accident. Both ~

hydrogen recombiners are required to be placed in service when hydrogen levels exceed 4%

l

b. N zirconium-water reaction is the major source of hydrogen early in the accident. Both hydrogen  !

recombiners are required to be placed in service when hydrogen levels exceed 4% J

c. The radiolytic decomposition of water is the major source of hydrogen early in the accident. At least  !

one hydrogen recombiner is required to be placed in service within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the LOCA if  :

hydrogenlevels are belew 4% '

d. The zirconium-water reaction is the major source of hydrogen early in the accident. At least one hydrogen recombiner is required to be placed in service within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the LOCA if hydrogen levels att below 4%

c Ams: Id = l Exams tevel: IS l C-:_ i^!ve Level: I en=nrehension I

. a M Per CBV CBD Section 4.0 one hydrogen recombiner is place in service within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following LOCA. Per SO l ofAnswer seconduner is placed in service with hydrogen > 2%, but not to exceed 4% indicated. Design Basis also states single most important source of hydrogen in the containment at the time of and immediately after a LOCA is the zirconium-water reaction; radiolytic decomna=ition is an important factor in the long term buildup of hydrogen.

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Tirr: l Plant Svstams l RO Group: l3 l SRO Group: l2 Systems /Evolutica Nasaber: l 028 l System / Evolution 'I1tle: l Hydrogen Recombiner and Purge Control System Category: A2 l Ability to (a) predict the impacts of the following on the Hydrogen Recombiner and Purge Control System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:

KAt l A2.02 l LOCA condition and related concern over hydrogen EO Value: l3.5 l SRO Value: l3.9 l CFR: l 41.5 /43.5/ 45.3/45.13 Reference Reference Number Reference Section Pase Nuseber(s) Revision learn. Obj CONTAINMENT AND 0300-000.00S- IX.B.2; IX.D.I.a i16;118 5/13/96 1.j CONTAINMENT CONTMT-00 SUPPORT SYSTEMS INTRODUCTION 'ID 0300-000.00S-ESF000- VID.3.d 39 6/25/96 17,18 ENGINEERED SAFETY 00 FEATURES AND DESIGN CRITERIA HYDROGEN S2.OP-SO.CAN- 2.3 2 RECOMBINER 0001(Q)

OPERATION Question Source New Question Modification Method Question Source Comments: l Material Required for Ex==dantion:

Page 82

Question Topica l Design SFP drain protection 22 Spent Fuel Pool (SFP) Pump is running providing SFP cooling.

Which one of the following correctly describes when 22 SFP Pump will lose suction if a leak develops in the pump suction line?

a. Six feet above the fuel, the level corresponding to the minimuin NPSH for the pump

b. Any running pump trips when SFP LVL LO alarm (OHA C-35) actuates

c. When the pump suction intake is uncovered,4 feet below normal SFP level

d. Three feet below normal SFP level, when the anti-siphon hole in the suction pipe uncovers Ans; Ic l En= Level: lS l Cognitivelevel: l Memory l Explanation Suction piping is located approximately 4 feet below normal pool level. Discharge piping has the anti-siphon hole, of Answer not the suction piping. The law level alarm is set at 0.5 ft. below normal and there are no automatic interloc ts with the pump. The discharge piping terminates six feet above the fuel.

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Tier: l Plant Systemr l RO Group: l2 l SRO Group: l2 Systesm/ Evolution Number: l 033 l Systesn/ Evolution

Title:

l Spent Fuel Pool Cooling System Category: K4 Knowledge of Spent Fuel Pool Cooling System design feature (s) and or interlock (s) which pro >ide for the following:

KA: l K4.01 l Maintenance of spent fuellevel FO Valuet l2.9 l SRO Value: l3.2 l CFR: l 41.7 Reference Reference Nunnber Reference Section Pate Nunnber(s) ' Revision 12arn. Obi SPENT FUEL POOL 0300-000.00S-SFP000- IILH.1 14 8/l4/96 2; 4.b COOLING SYSTEM 01 Question Source NRC Exam Bank Question ModiNation Method Question Source Comunents: l Matesial Required for Exanninstion:

o Prge 84 w_ _

p_

p Topic: l Fuel Transfer Cart interlocks Given the following fuel handling associated interlocks:

1)-Transfer Tube Gate valve open 2)-Fuel Handling Crane hoist full up

3) - Reactor side (Containment) lifting frame down

4) - Pit ride (Fuel Handling Buildine lifting frame down

5) - Reactor Containment Manipmator Crane gripper tube up (LS-3)

Which one of the following correctly list those interlocks that must be satisfied to enable initiation of the Fuel Transfer System Conveyor Car from the Reactor side to the Fuel Pit side?

a. 1,2 and 3.

b. 1,3 and 5.

c. 1,4 and 5

d. 1,3 and 4.

Anse ld l Exam Level: lS l Cognitive Level: l Memory l Explanation As giwn: 3 interlocks for movement of car. The other two interlocks are valid for crancs, and/or movement into the ofAnswer area of the upenders but do not affect car motion.

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Tier: l Plant Systems l RO Group: l3 l SRO Group: l2 .

1 Systena/ Evolution Number: l 034 l System / Evolution

Title:

1 Fuel Handling Equipment System I Category: K4 l Knowledge of Fuel Handling Equipment System design feature (s) and or interic :k(s) which provide for i the following:

KAt l K4.02 l Fuel movement RO Value: l2.5 l SRO Value: l3.3 l CFR: l 41.7 Reference Reference Number Reference Section Page Number (s) Revision Learn. Ob] j REFUELING SYSTEM 0300-000.00S- IV .C.6.J.1 33 10/30/96 6.c REFUEL 00 j Question Source NRC Exam Bank Question Modification Method i Question Source Comments: l Matejal Required for Exands.ation:

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Page 86

Questles Tende: l "New" SGs  :

Wluch one of the following choices correctly completes a description of the difference in operating characteristics for the new Unit 1 Model F SG's (RSG's) as compared to the removed Model 51 SGs j (OSG's)? l The heat transfer area and heat transfer capability of the RSG's is...

i

a. less; therefore 100% Tavg has been raised 2 'F in order to maintain design steam pressure at 100% j power. It is expected that SG level response during transients will be toimilar to the previous design.  !

b. less. Since the Tavg program has not changed, steam pressure at 100% power is lower.

l It is expected that SG level response during transients will be more pronounced.

. c. i greater; therefore 100% Tavg has been lowered 2 *F and steam pressure is still somewhat higher. It is expected that SG level response during transients will be similar.

d. greater. Since the Tavg program has not changed, steam pressure at 100% power is higher. i It is expected that the level response during transients will be more pronounced.

Ams: lc lEmmaslevd: IS l Ceamitive level: l Memory l F ," ^ ' _ Unit I has Replacement SGs (RSG) with different operating characteristics from Unit 2 (or the Old SGs, OSG).

etAnswer 'the full power operating pressure for the RSG in the Salem-1 loops will exceed that of the OSG. Therefore, the l

plant could be operated at full power with a greater stearr. pressure or a lower average primary system temperature (Tave). Tavg was lowered approx. 2 'F and steam pressure is approx. 25 psi higher. Concerning NR level, the cross  :

sectional area of the RSG drum, outside of the first stage separators is greater than the cross sectional area on the {

OSG. "Ihe larger cross sectional area of the RSG drum offsets the reduction in narrow rang; span of the RSG.

Therefore, it is expected that the level response during transients of the SGs in percent of narrow range span will be similar.

Page 87

Tier: l Plant Systems j RO Group: l2 l SRO Group: l2 Systene/ Evolution Museber: l 035 l System / Evolution

Title:

l Steam Generator System Catenory: 12.2 l Equipment Control KA: l2.2.3 l (multi-unit) Knowledge of the design,. procedural, and operational differences between units.

RO Value: l3.1 l SRO Value: l3.3 l CFR: l 41/43 /45 Reference Reference Number Reference Section Page Nuanber(s) Redulon I4arn. Obj STEAM GENERATOR. SG 0300-000.00S- VII.B.I .a; VII.B.S.a.3) 50;53 7/15//97 II BLOWDOWN AND DRAIN STMGEN-01 SYSTEMS Question Source New Question Modification Method Question Source Comments: l Matedal Required for Exannimation:

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l Page 88

Quesden Topic l SG pressure channel failure At 50% power on Unit 2, the steam pressure instrument for the 21SG (PT-516A) fails high with 21MS10, 21 SG Atmospheric Relief valve, selected to AUTO.

Which one of the following describes the effect that the Irr-516A failure will have on 21MS107

a. 21MS10 will go to the full open position.

b. 21MS 10 control will automatically switch to MANUAL

c. 21MS10 remains in AUTO with a pressure instrument on the same DPU, Irr-516D (24MS10),

providing theinput signal.

d. 21MS10 remains in AUTO with the ADFWCS providing an average stearr. pressure signal developed from the otherTHREE SGs.

Amst Ib l Exam level: lS l Connitive level: 1 Memory l E=; " When any one pressure channel fails: a) Switches MS10 control for that loop to MANUAL; t' Actuates "ADFCS ofAnswer SWAP TO MANUAL", OHA G7; c) Gives ADFCS TRBL OHl In other ADFWCS circuit: , alternative and/or averaging circuits are used for norrnal operations and/or as automatic replacements for failed , asignals. Controllers for 21MS10 and 24MS10 are from same DPU - 2/52.

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Page 89

Tier: l Plant Systems l RO Group: l2 l SRO Group: l2 SystenvEvolution Nussber: l 039 l Systeen/ Evolution

• Dele: l Main and Reheat Steam System Category: K1 Kncwledge of the physical connections and/or cause-effect relationships between Main and Reheat Steam <

System and the following:

KA: lKl.02 l Atmospheric relief dump valves

!"O Value: l3.3 l SRO Value: l3.3 l CFR: l 41.2 to 41.9 / 45.7 to 45.8 '

Reference Reference Nuneber Reference Section Pase Number (s) Revision Imrn. Ob]

ADVANCED DIGITAL 0300-000.00S- V.A.7.d.2) 33 6/27/96 8.g.12.c FEEDWATER CONTROL ADFWCS-00 SYSTEM MAIN SEAM SYSTEM 0300-000.00S- IV.B.4.g.2).a); X.A.4 21;44 5/6/96 4.c;13 MSTEAM-00 Question Source New Question Modification Method Qi==*Ian Source Conuments: l Matedal Required for Exaniination:

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Page 90 l

W Topic: l Stemn Dump failure Given the following conditions for Unit 2:

o Reactor power- 100%

o 2A 115 VAC Vital Bus poweris lost Which one of the following correctly describes the reason the operator'is directed to shift the Steam Dump controller from TAVG to MS PRESS CONT 7

a. The steam dumps are armed. The steam dump valves will open due to the signal from the load rejection controller if Tavg exceeds Tref by 5 F.

b. A steam dump demand signal is generated from the plant trip controller. If an arming signal is geaerated, the st.; un dump valves will open to the dem= dad position. 1

c. A steam dump demand signal is generated from the load rejection controller. ' If an arming signal is

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generated, the steam dump valves will open to the demanded position.  !

d. The steam dumps CANNOT be armed from the turbine first stage pressure signal. If ONE reactor i

. trip breaker fails to open on a trip, the steam dumps would be inoperable in TAVG Mode.

i Aas: lc l Fum- Level: lS l Cognitive Level: l Comprehension l j FM Channel 1 First Stage impulse Pressure (IT-505) is lost which generates a demand signal in Tavg mode on the

{

etAnswer load Rejection circuit. ~

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I Page 91 -

Tier: l Plant Systems l RO Group: l3 l SRO Group: l3 i

Systemn/ Evolution Number: l 041 l System / Evolution'ntle: l Steam Dump System and Turbine Bypass Control  !

Category: K3 Knowledge of the effect that a loss or malfunction of the Steam Dump System and Turbine B.ypass l Control will have on the following: 1 i -

KA: l K3.02 l RCS I

' 4 RO Value: l3.8 l SRO Value: l3.9 l CFR: l 41.7 / 45.6 Reference Reference Number Reference Section Pane Number (s) Revision Learn. Obj STEAM DUMP SYSTEM 0300-000.00S- V.A.S.a); V.A.8.c 20;26 9/6/96 10 STDUMP-00 LOSS OF 2A,2B,2C AND 0300 000.00S-AB il51- III.C.6 10 8/22,97 3 2D 115V VITAL 00 INSTRUMENT BUS Steam Dump Logic 221059 l

Question Source Facility Exam Bank Question Modification

Method Question Source Comments

l l Material Required for Exandmation:

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Onassisa Telet l Fw Isolanon Given the following conditions for Unit 2:

)~ o Reactor power- 14%

o Pressurizer y.hu- 2225 psig o RCS Tavg -551 *F d

c Main feed in operation using BF40s ; ,

o SG levels stable at 35%

o : Steam y.A. - 1000 psig 2

o Steam dumps maintaining Tavg with STEAM PRESSURE in MANUAL o One Steam Dump valve fails open Following isolation of steam flow:

o Pressurizer pressure - 1825 psig and beginning to rise o RCS Tavg - 542 *F o SG levels at 48%-

! o Steam pressure- 870 psig i

i F

Which one of the following correctly explains why feedwater was isolated to the SG? (Assume all systems respond normally and NO operator action is taken.)

a. The pressurizer pressure drop resulted in an SI.

b. - High steam flow coincident with the low Tavg resulted in an SI.

c. A reactor trip coincident with low Tavg res :lted in actuation of the feedwater interlock.

d. The opening of the steam dump caused water level in the SGs to swell above the P-14 setpoint.

Ams: Ie l Examilevel: IS I Commitive Level: l Comprehension l 1 Explanation SI signal on PZR Pressure low at <t765 psig on 2/3 PZR pressure detectors when NOT blocked. Feedwater ofAnswer Interlock is generated by a Reactor' nip (P-4) coincident with a low Tavg (<554'F auctioneered high Tavg), P-14 occurs when SG level >67% on 2/3 levels on 1/4 S/G's. High Steam Flow with Low Tavg or Iow Steam Line Pressure SI occurs >40% steam flow (at this power level) coincident with RCS Low Tavg <543 F, on 2/4 Tavgs.

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i Page 93 l

Tier: l Plant Systems l RO Group: l1 l SRO Group: l1 Systems / Evolution Nusaber: l059 l Systein/ Evolution

Title:

l Main Feedwater System Category: l A3 l Ability to monitor automatie operations of the Main Feedwater System including:

KAt l A3.06 l Feedwaterisolation RO Vahee: l 3.2* l SRO Value: {3.3 l CFR: l 41.7 / 45.5 Reference Reference Number Reference Section Page Nunnber(s) Revision Learn. Ob]

CONDENSATE AND 0300-000.00S- VII.A.6 76 5/23/96 9.d FEEDWATER SYSTEM CN&FDW-00 RPS Feedwater Control & 221062;221063 1E-2E 6;5 Isolation Logic - sht.13 & 14 Question Source New Question Modification Method Question Source Comments: l Material Requisul for E===d== tion:

Page 94

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Qusseles Tesic: APW Pump runout protection 1

j Given the following conditions for Unit 2: I i

o . Unit is in Mode 3 with Tavg at 547 *F o : The 23 AFW pump is NOT available '

o 'Ibe 21 AFW pump has been just stopped due to unusual motor noises t o The 22 AFW pump is running with normal parameters 4 r o An operator has been dispatched to open 21 AF923 and 22AF923 to allow I j' cross-tie of the AFW headers from the motor-driven AFW pumps.

o The CRS wants to maintain all SG levels within normal operating band I

Which one of the choices cormctly completes the following statement?

Once the AF923 valves are open, the PO... 1 I

i a. only needs to throttle the AF21 valve to each SG.

9

b. must depress the PRESS OVERRIDE DEFEAT for each AF21 valve, and then throttle the AF21 valve to each SG 1

c. must depress the PRESS OVERRIDE DEFEAT for 21 AF21 and 22AF21 valves, and then throttle 4

the AF21 valve to each SG.

d. must depress the PRESS OVERRIDE DEFEAT for 23AF21 and 24AF21 valves, and then throttle

^

the AF21 valve to each SG Aas: ld l Exam Level: IS l Cognitive Level: l Comprehension l Fwgdammelam Runout protection for the MD AFW pumps is provided by respective AF21 valve With AFW pump disch header j ofAnswer pressure < 1150 psig, the associated AF 21 valves remain closei Should it become necessary to feed tb '"Gs immediately and pump discharge is < 1350 psig, the operator will depress PRESS OVERRIDE DEFFA '

pushbutton to remove runout protection for respective pump and allows operator control of AF21 valves. Since the headers are tied and the 21 AFW Pump is not running, the operator must override 23 and 24AF21.

l Page 95 .l

Tier: l Plant Systems l RO Group:

^

l1 l SRO Group: ll_

SystenvEvolution Number: l 061 l SystenvEvolution

Title:

l Auxiliary / Emergency Feedwater System Category: Al Ability to predict'and/or monitor changes in parameters associated with operating the Auxiliary /

Emergency Feedwater System controls including:

l KA: lAl.01 l S/G level RO Value: l3.9 l SRO Value: l4.2 l CFR: l 41.5 /45.5 i l

Refereace Reference Number Reference Section Pase Number (s) Revision Imrn. Obj AUXILIAttY FEEDWATER 0300-000.00S- IV.B.3.g.2).d); 27 5/15/97 4.h;8.c SYSTEM AFWOOO-01 IV.B.3.g.3).a) l Question Souste New Question Modification Method ,

Question Source Comments: l l

Material Required for l Exandmation:

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Questian Topic: l Minimum rent for RCS heat renoval

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Which one of the following correctly dederibes the basis for the Auxiliary 161 water System technical  !

specification?

a. Maintain the capability to cooldown the RCS to RHR initiation conditions following a complete loss of off-site power

b. Ensure the capability to cooldown and maintain the RCS at <500 *F for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> in the event of a complete loss of off-site power, assuming failed fuel

c. Remove decay heat and maintain the RCS at HSB conditions for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following a complete loss of off-site power i

d. . Provi:le the RCS heat removal capability necessary to prevent a challenge to the pressurizer safety valves during a full power ATWT, followed by a complete loss of off-site power Ans: la l Exam Level: IS l Coanitivelevel: l Memory l l Explanation Pe T9 B:.is 3.7.1.2 ofAnswer i

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Page 97 I

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• m,I : l Plant Sy,sjems _ , _ _

l RO Group: II I SRO Group: l1

,ldann/Eiolution Numiber: J061 ] System / Evolution

Title:

l Auxilitry / Emergency Redwater System lUegory: K3 Know'A'ge of the ellect .that a loss or malfunction of the / cxiliary/ Emergency Feedwater System will have ce the followit L _

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_KA: l K3.(I'~ l RCS  ;

RO Value: 14.4 l SRO Value: 14.6 l CFR.: l 41.7 /45.6 Reference Reference Nand$er Reference Section ' Paac Number (s) Revision Learn. Obj AUXILLARY FEEDWATER 0300-000.00S- VIII.C 56 5/15/97 2 l SYSTEM AFW000-01 TS Basis 3.7.1.2 l Question Source New Question Modincation Method Question Source Cosiments: l M:*edal Required for Exm=d== tion:

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O Page 98

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l Quessise Topic: l toss 500 KV breaker Given the following conditions:

o 500KV Switchyard is in a normal lineup w' en a Unit 2 main generator fault occurs at 100% power I

l o 13KV breakers 2-3 and 5-6 are open (no cal condition)  :

o 500KV l-9 (32X) Breaker failed to open l o 500KV BKR FAIL, OHA K-14,in alarm  ;

Which one of the following describes the effect that the 1-9 Bieaker fa3ure will have on the electrical

., system?

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a. The Hope Creek 500KV tic line will be deenergized.  !

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b. All 4KV vital buses will be fed from either the No.13 or No. 24 Station Power Transformers. )

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c. 4KV group buses fed from the No.12. and No. 22 Station Power Transformers will be deenergized.

d. 4KV vital buses previously aligned to No.14 or No. 23 Station Power Transformers will be powered from their respective EDGs.

Amst lb l h== level: lS l Cognitive Level: l Comprehension l Frps *iam See K-6 OHA sheet (S2.OP-AR.ZZ-0010(Q)). Auto action for 32X Breaker Failure / Ground Protection trips the ofAnswer following: Unit 2 Turbine (Backup); Unit 2 Generator Excitation Field Breaker,500KV BS 1-5 breaker (12X);

500KV BS 1-8 breaker (20X); 500KV BS 9-10 breaker (30X). This removes power to 14 STP and '!3 SPT infeed )i bus section D.

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Page 99

Tier: l Plant Systems l RO Group: l2 l SRO Group: l2 Systemi/ Evolution Numiber: l062 l Systeni/ Evolution

Title:

l A.C. Electrical Distribution Category: K3 Knowledge of the effect that a loss or malfunction of the A.C. Electrical Distribution will have on the following:  !

KA: l K3.01 l Major systern loads '

RO Value: l3.5 l SRO Value: l3.9 lCFR: l 41.7 /45.6 Reference Reference Nunnber Refe:ence Section Page Numiber(s) Revision Learn. Ob]

500KV ELECTRICAL 0300-000.00S- V.B.3.e 42 1/17/97 3,6 SYSTEM 500KV0-00 13KV ELECTRICAL 0300-000.00S- VI.A.2; IX.C.4 79;83 4/11/96 3.1 SYSTEM (EXCLUDES 13KVAC-00 UNIT NO. 3)

OVERHEAD S2.OP-AR.ZZ-0010(Q) Windows K-14 & K-6 68;34 (page 15 i ANNUNCIATORS (CRT Point 675) of 16)

WINPOW K Q :baSource New Question Modification

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Method Question Source Comments: l Mcterial Required for Switchyard Simplified Drawing with no breaker positions indicated Frandmation:

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I Tier: 1 Plant Systems l RO Group: l2 l SRO Group: l1 l Systena/ Evolution Noseber: l 063 l System / Evolution

Title:

l D.C. Electrical Distribution C " ;..y: A2 Ability to (a) predict the impacts of the following on the D.C. Electrical Distribution and (b) based on

{

those predictions, use procedores to correct, control, or mitigate the consequences of those abnormal 1 operation:

I KA: l A2.01 l Grounds l 4 i RO Value: l2.5 l SRO Value: l 3.2* l CFR: l 41.5 /43.5/45.3/45.13 Reference Reference Number Reference Section Pase Number (s) Revision Learn. Obj

< DC ELECTRICAL 0300-000.00S- IV.B.4.f.5), VIII.A.I.b 21:30-31 8/21/96 10,12

, SYS'IEMS DFFT FC-00 & 3.e 28VDC GROUND S2.OP-SO.284)003(Q) Precaution 3.3 2 2 DETECTION 2B 28VDC BAITexY S2.OP-SO.28-0002(Q) 2.3; 3.3 2 ]

, CHARGER OPERATION 1 Q#=-2 Source New Question Modification

' Method C-- "' -a Source Comments:

1 Material Required for Exannination:

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i Questles Topict l Exciter controls i

Given the following:

l c  !

Emergency Diesel Generator (EDG) 2A was started locally and is running unloaded  !

o & NEO mistakenly places the MANUAI1 AUTOMATIC Voltage Regulator Switch (RMAS) in the  !

MANUAL position '

Which one of the following conectly describes what will happen if a Mode III SEC actuation occurs before the RMAS Switch is re-positioned?

a. 2A EDG continues to run, load control shifts to ISOCHRONOUS and 2A EDG breaker closes

b. 2A EDG continues to mn while load control shifts to ISOCHRONOUS. However,2A EDG breaker l will not close with the RMAS switch in MANUAL

c. 2A EDG stops while load control shifts to ISOCHRONOUS and the Voltage Regulator shifts to AUTO. It will then restart and the breaker will close l

d. 2A EDG stops while speed controls shift to ISOCHRONOUF but the Voltage Regulator will remain  ;

in MANUAL. 2A EDG then restarts and the breaker will close but only if the MANUAL setting  !

allows the generator to achieve the minimum required voltage l

Amst Ia l Exann Level: lS l Cognitive Level: I Comprehension l Fwpl===h if an EDG is running when a SEC start occurs, it continues to run. However, controls are automatically realigned ofAnswer for auto loading and running isolated before the breaker closes /re-closes.

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Page 103

Tier: I Plant Systems l RO Group: l2 l SRO Group: l2 Systesm/ Evolution Number: l 064 l System / Evolution

Title:

l Emergercy Diesel Generators Ccte p ry: A2 Ability to (a) predict the impacts of the following on the Emergency Diesel Generators and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:

KA: l A2.03 l Parallel operation of ED/Gs RO Value: l3.1 l SRO Value: l3.1 l CFR: l 41.5 /43.5/ 45.3/45.13 Reference Reference Number Reference Section Pase Number (s) Revision Learn. Obj EMERGENCY DIESEL 0300-000.00S- IV.B.8.b.5).b); 59-60;63 6/3/96 4.g; 4.h; 7 GENERATORS EDG000-00 IV.B.8.b.5).f)

EMERGENCY DIESEL 0300-000.00S- V.A.3..d.4).c) 84 6/3/96 6.f i GENERATORS EDG000-00 Quen Source New Question Modification Method Question Source Comments: l Mate:421 Required for Examination:

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Page 104 I

s Questles Topict l DegrM bus following SIh set Given the following conditions for Unit 2:

o A LOCA has occurred .

o Safeguards Actuation signals and SECS have been reset i

o Emergency Diesels Generators (ElXis) 2A,2B and 2C are running unloaded i o

Vital bus 2A is aligned to 23 Station Power Transformer (SM)  ;

o Vital buses 2B and 2C are aligned to 24 SPT o 23 SPT ==dag voltage falls to 3800 voks Which one of the following conectly dscribes the consequence to 2A 4KV Vital Bus as a result of the 23 SPT problem?  !

I a.. Auto transfer to the 24 Sirr. I

[

b. Remams energized from the 23 SPT.

+ 1

,. c. Energized from 2A EDG on a MODE II* SEC operation. '

[ d. Energized from 2A EDG on a MODE IV SEC operation.

O Ans: Ic l Rwam Level: IS l Cognitive level: l Comprehension l F "-- " ' _ Bus is less than 95.1% nominal voltage (for > 13 sec): Degraded voltage blocks auto transfer to alternate offsite ofAnswer source. DG ties to bus and SEC!onds in MODE II*: Single Bus Blackout, because both SI and SEC have been i reset.

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L. . l

Tier: l Plant Systems l RO Group: l2 l SRO Group: l2 Systenn/ Evolution Nusuber: l064 l Systesn/ Evolution

Title:

l Emergency Diesel Generators Category: l A3 l Ability to monitor autonatic operations of the Emergency Diesel Generators including:

KA: l A3.07 l Imad sequencing RO Value: l 3.6* l SRO Value: l3.7 l CFR: l 41.7 /45.5 Reference Reference Number Reference Section Page Number (s) Revision Learn. Obj 4160 ELECTRICAL 0300-000.00S- V.C.I.b.3); V.C.I.c.2) 49;51 6/4/96 4.e SYSTEM 4KVACO-00 I

SAFEGUARDS 0300-000.00S-SEC000- IV.C.3.a: IV.C.4; 17-18:24 5/29/96 3.b; 4 l EQUIPMENT CONTROL 00 IV.F.3 l i

SYSTEM '

Question Source NRC Exam Bank Question Modification Method Question Source Comments: 1 Material Required for Frandmation:

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Page 106

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Quessies Topic: l RCDT outlet isolation valves Given the following conditions for Uni: 2- '

l o A Pressurizer Safety Valve has been leaking i o Procedural actions being taken to prevent overpressurizing the PRT

}

are generating liquid waste l

1 Assuming the containment isolation valves are open, which one of the following correctly describes the flow l path of the water after the operator opens 2PR14, PRT Drain valve?

l

a. The PRT gravity drains to the in-service CVC HUT .

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b. The PRT gravity drains to the RCDT. The RCDT pumps automatically cycle on RCDT level, l pumping to the in-service CVC HUT.

i

c. The RCDT Pump in AUTO cycles to control PRT level whenever 2PR14 is open

d. RCDT pumps start on interlock with 2PR14, directing flow to the in-service CVC HUT.

Amst ld l Fumm Level: lS l Commitive Level: l Memory l Explanation Opening the PRT drain valve, PR14, will cause both RCDT pumps to stan if WL12 and WL13 are open. PRl4 ofAnswer aligns the PRT to the suction of the RCDT pumps, which are normally aligned to the CVCS Holdup Tanks.

Page 107

Tiers l Plant Systems l RO Group: l1 l SRO Group: l1 Systesm/Rvolution Nuanber: l 068 l Systema / Evolution

Title:

l Liquid Radwaste System  !

l Category: K1 Knowledge of the physical connections and/or cause-effect relationships between Liquid Radwaste System and the following:

KAt lKl.07 l Sources ofliquid wastes for LRS I'G Value: i2.7 l SKO Value: l2.9 l CFR: l 41.2 to 41.9 / 45.7 to 45 8 R derence Refennee Noenber Reference Section Pane Number (s) Revision Learn. Ob] I RADIOACTIVE LIQUID 0300-000.00S- IV.C.13.d.2).e); IV.D.1 36;43 8/20/96 3.b.I; 6.c WASTE SYSTEM WASLIQ-00 N2. 2 Unit Waste Disposal 205339 C-2 Sh.3 28 Liquid Question Source Facility Exam Bank Question Modification Method Question Source cosaments: l Material Requiral for Frandmadam:

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Page 108

Questies Topic: l CR radiation detectors Which one of the following correctly describes the functions of the Control Room (CR) area radiation monitors?

a. - 1RI A, Unit 1 CR area monitor, alarm only.

- 2RI A, Unit 2 CR area monitor, alarm only. .

b.~ - 1RI A, Unit 1 CR area monitor, alarm only.

- 2RI A, Unit 2 CR area monitor, ac:uates Unit I and Unit 2 CR ventilation in ACCIDENT Mode.

c. - 1RIA, Unit 1 CR area monitor, actuates Unit I and Unit 2 CR ventiladon in ACCIDENT Mode.

- 2RI A, Unit 2 CR area monitor, alarm only,

d. - 1RI A, Unit 1 CR area monitor, actuates Unit 1 CR ventilation in ACCIDENT Mode.

- 2RI A, Unit 2 CR area monitor, actuates Unit 2 CR ventilation in ACCIDENT Mode.

Amst la l Exana Level: IS l Cognitive level: l Memory l Explanation These area monitors provide alarm function only. Duct monitor will provide for actuation of CR Vent.

of Answer-Page 109

Tier: l Plant Systems l RO Group: l1 l SRO Group: l1 Systeam/ Evolution Nonsber: l 072 l Systeni/ Evolution

Title:

l Area Radiation Monitoring System Category: K4 Knowledge of Area Radiation Monitoring System design feature (s) and or interlock (s) which provide for the following:

KA: l K4.03 l Plant ventilation systems

[ID Value: l 3.2* l SRO Value: l 3.6* l CFR: l 41.7 Reference Reference Number Reference Section Page Number (s) Revision Learn. Ob]

RADIATION 0300-000.00S- IV.B.3.f; Attachment 1 35 10/01/96 4.a MONITORING SYSTEM RMS00(M)0 &2 Question Source New Question Modification Method Question Source Comments: l Matedal Required for E===d== tion:

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Question Topic: l$GBD Which one of the following cenectly describes the effect on the GB4 valves (SG Blowdown Outlet Isolations) to a rising radiation condition on Unit I and Unit 2 R19D, Steam Generator Blowdown Liquid Monitor (14,24 SG)?

a. On Unit 1, all GB4 valves will close on high alarm condition..

On Unit 2, only 24GB4 will close on high alarm condition.

b. On Unit 1, only 14GB4 will close on warning alarm condition.

On Unit 2, all GB4 valves will close on warning alarm condition.

c. On Unit 1, only 14GB4 will close on warning alarm condition.

On Unit 2, all GB4 valves will close on high alarm condition.

d. On Unit 1, all GB4 valses will close on warning alarm condition.

On Unit 2, only 24GB4 will close on high alarm condition.

Amst la l Exaan Ievel: IS l Cognitive Level: I Comprehension l l F.* --- % n Unit I has NO warning alarm actions; Unit 2 does have warning alarm actiot.s for other SGBD valves. On Unit 1, of Answer any R19 alarm closes all GB4 valves; On Unit 2 only the affected SG isolation valve is closed.

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Tiers l Plant Systems l RO Group: l2 l SRO Group: l2 SysM-Wution Number: l073 l System / Evolution

Title:

l Process Radiation Monitoring System Cr.tesory: l A4 l Ability to manually operate and/or monitor in the control room:

KA: l A4.01 l Effluent release RO Value: l3.9 l SRO Value: l3.9 l CFR: l 41.7 / 45.5 to 45.8 Reference Reference Number Reference Section Pase Number (s) Revision Learn. Ob[

STEAM GENERATOR, SO 0300-000.00S- IV.B.10.g 31 ~1/15/97 9 BLOWDOWN AND DRAIN STMGEN-01 SYSTEMS RADIATION 0300-000.00S- IV.B.1.1 24 10/1/96 6.k; 11 MONITORING SYSTEM RMS000-00 Questson Source New Question Modification Method Question Source Comments: l Matedal Required for Framination:

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Question Topic: l EDO CO2 4

Wluch cae of the following correctly desc' fxs the differewe in response for an AUTO as compared to a MANU/.L actuation of the DG Area CO2 system?

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a. AUTO CO2 actuation is bloch! e a SEC start.

b. If the associated EDG is running, it will trip following a MANUAL CO2 actuation.

c. If the associated EDG is in LOCKOUT, the AUTO CO2 actuation is blocked. ,

d. There is NO 13 second CO2 discliarge delay on a MANUAL actuation.

Amst lc l Emass Level: lS l Cosaltive Level: l Comprehension - l  ;

s' , "

When a CO2 system actuates, the system performs a planned sequence of events as follows: 1)Fust, a system timer ofAnswer is energized to start timing a predischarge period; 2) Once the predischarge period is timed out the CO2 discharge i

starts. The CO2 System may be discharged on a loss of power by using the operating lever installed on the electro-manual pilot operating cabinet. This method of operation bypases the timer functions. There are no interlocks between EDO switch positions and CO2 actuation.

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Tier: l Plant Systems l RO Group: l2 l SRO Group: l2 SysteavEvolution Nuanber: l 086 l SystenvEvolution 11 tie: l Fire Protection System Ca'. ;.y: K4 Knowledge of Fire Protection System design feature (s) and or interlock (s) which provide for the following:

KA: l K4.06 lCO2 RO Value: l3.0 l SRO Value: l3.3 l CFR: l 41.7 i

Reference Reference Number Reference Section Pase Nuniber(s) Revision Learn. Obj I FIRE PRuitCTION 0300-000.00S- IV.B.2.p.15).g) 55,58 11/21/96 4.c.v,7.b &

SYSTEM FIRPRO4X) e i

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Method l Question Source Comunents: l Material Required for

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l Quesnem Tepic l Effect of rod enouon on power ~

t Given the following conditions for Unit 2:

l o Peacrar power - 50%

o . Control Rod position - 170 steps on Control Bank D o_ Control Rods began to withdraw due to an IT-505, Turbine First Stage Pressure, fails high '

Which one of the following correctly indicates when rod motion will stop and the steady state values of f

reactor power and Tavg after rod motion has stopped? Assume rod motion it no more than 10 steps.  !

a. Rod motion continues until rod contiel is placed in MANUAL; Reactor power rises slightly; Tavg [

rises i

b. Rod motion continues until rod control is placed in MANUAL; Reactor power rises slightly; Tavg

. remains the same i

c. Rod motion continues until the Tavg-Tref mismatch signal decays out of the circuit; Reactor power  !

rises slightly;Tavg rises  !

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d. Rod motion continues until the Tavg-Tref mismatch signal decays out of the circuit; Reactor power rises slightly; Tavg remains the same Ams: Ia l Rum == level: IS l Cm== hive level: l Cn==Ansion l 5' f

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At 50% power with rods below 225 steps (C-II), rods will move out when PT-505 fails high. Rod motion without ofAnswer significant change in turbine load will result in an increase in Tavg. A small rise in reactor power will be observed at steady state due to the rise in turbine first stage pressure.

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Tier: l Emerger.cy and Abnormal Plant Evolutionn l RO Group: l2 l SRO Group: l1 SyhTvolution Nunser: l 001 l Systesn/ Evolution

Title:

l Continuous Rod Withdrawal Ca'+ j: AKI Knowledge of the operational implications of the following concepts as they apply to Continuous Rod Withdrawal:

KA: l AKl.03 l Relationship of reactivity and reactor power to rod movement RO Value: l3.9 l SRO Value: l4.0 lCFR: l 41.8 / 41.10 / 45.3 Reference Reference Numiber Reference Section Revision Pase Nunser(s) learn. Ob]

CONTINUOUS ROD 0300-000.00S- III.C. Step 37,38,39, 17 6/4/97 4 MOTIO i ABROD3-00 40, & 4i ROD CONTROL AND 0300-000.00S- V.C. I .b.2) 63-64 12/11/96 6.e POSITIONINDICATION RODS 00-00 SYSTEMS INSTRUMENT FAILURE 0300-000.00S-ICFAlle V.Ill.D.1.a 31-32 5.a REVIEW 00 Question Source New Question Modification Method Question Source Conuments: l Materfal Required for Exandnation:

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Quesdom Topic: l Reactor trip Given the following conditions for Unit 2:

o Reactor power- 25%

o Tavg - 553 'F o Control Bank D Group Step Counters - 150

. o ROD BO1 TOM, OHA B-48 is in alarm for a dropped rod in Shutdown Bank B o Recovery of the dropped rod is in progmss when a rod in Control Bank D drops Which one of the following correctly describes the required operator action (s) and the basis for the action (s)?

c. The reactor should be tripped because the safety analysis shows that DNBR will exceed its limit,

b. The reactor should be tripped because this could be indication of multiple common mode malfunctions.

. c. Continue recovery of the dropped Shutdown Bank rod if Tavg has remained above the minimum

temperature for criticality.

d. Continue recovery of the dropped Shutdown Bank rod and initiate actions for a flux map to confirm the dropped Control Bank D rod.

Ans: lb ] Exam level: 1S l Connitive Level: l Comprehension l F-; " "': a 'Ihe basis for tripping the reactor is ine pctential for common mode failures. The AB basis discusses the fact that ofAnswer DNBR will not be exceeded.

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Der: l Emergency and Abnormal Plant Evolutions l RO Grusp: l2 l SRO Group: l1  ;

S, ^ _Tvolution Nunaer: j003 l Systenn/ Evolution Dtle: l Dropped Control Rod Ca',;.j: l AK3 l Knowledge of the reasons for the following responses as they apply to Dropped Control Rod:

KAt l AK3.04 l Actions contained in EOP for dropped control rod P3 Value: l 3.8* l SRO Value: l 4.1* l CFR: l 41.5,41.10 / 45.6 / 45.13 Reference Reference Nunser Reference Section Page Namiber(s) Revision Learn. Obj DROPPED ROD 0300-000.00S- IV.A.4 14 12/23/96 6.A ABROD2-00 DROPPED ROD S2.OP-AB. ROD- 2.3 3 5 TECHNICAL BASES 0002(Q)

DOCUMENT Q-A Souvre New Question Modification Method C;.A Source Conunents: l Material Required for F.unaminatian: '

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l Quase6em Tesic: l Trip breaker failure l

Given the following conditions for Unit 2:

{

s o A reactor trip has occuned from 100% power i

o . Train A reactor trip breaker failed to open .

o IAC has NOT installed jumpers for the breaker P-4 output

'!r Which one of the following correctly describes the consequences of the Train A reactor trip breaker  :

remaining closed?  !

f

a. Ced-r steam dumps will NOT function in Average Temperature mode.  !

b. . Main Steamline Isolation will NOT function on Steam Flow High coincident with low Tavg. i t

c. Control room operators will have lost the ability to reset and block a Train A automatic safety injection signal. (

d. Control room operators will have to manually close the Feedwater Regulating Valves (BF19's) and l Feedwater Bypass Valves (BF40's).-

l Ams: Ie lExamsLevd: lS l C#ve Level: I Linprehension l .

p.ya ma= In order to RESET SI, both trains must be reset. This allows operators to control safeguards components and it I

etAnswer j blocks other AUTV SI signals. If P-4 not present, the Auto SI signal is NOT blocked. Other components are '

affected but NOT in way listed: Steam Dump arming signal from Train A P-4 will NOT occur but SDs will get arming signal due turbine loss ofload (C-7) ar.d will function. FW Interlock affects only FRVs & BPRVs and is NOT affected by single train. One train of steam flow instruments are affected on each SO such that Turbine i impulse pressure signal from PT-505 provides Steam flow setpoint (which is ZERO load)instead of rx trip fixed no-load signal. l I

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Tier: l Emergency and Abnormal Plant Evolutions l RO Group: l2 l SRO Group: l2 Systesm/ Evolution Number: l 007 l System / Evolution

Title:

l ReactorTrip Cat;;.f:lEA2 l Ability to determine and interpret the following as they apply to Reactor Trip:

KA: l EA2.03 l Reactor trip breaker position F3 Value: l4.2 l SRO Value: l4.4 l CFR: l 43.5 /45.13 Reference Reference Number Reference Section Pane Numeer(s) Revision Learn. Obj EOP-TRIP-2, REAC10R 0300-000.00S-TRP002- 3.3.6.1 14-15 5/3/96 2 TRIP RESPONSE 01 REACTOR PROTECTION 0300-000.00S- VII.B.6.b-d 50 5/15/96 10 SYSTEM RXPROT-00 RPS-Safeguards Actuation 221057 2-D sh.8 Signals Question Source NRC Exam Bank Question Modincation Method C;A Source Conunents: l Material Required for Ex==d== tion:

Page 120

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Q uaselos T opic: l Vapor space leak conditions l i

l Given the following conditions for Unit 2:

o Reactor power- 100%  !

o Pressurizer pressureinstmments:

PI-456 - offscale low Other channels - 2185 psig and dropping slowly o Pressurizerlevelinstrurnents:

U-460-75% rising rapidly '

Other channels -50% stable )

Which one of the following correctly identifies the event that is occurring?

a. A Pressurizer PORV has failed open.

i i

b. Pressurizer pressure channel II(PT-456) has failed low. l

c. Controlling Pressurizer level channel II(LT-460) has failed high.

d. A leak has developed at the upper tap to level channel II (LT-460). I Ass

• ld l Exami Level: 1S l Ca==8tive Level: l Comprehension l Fwydammelam.

We tap for Pzr level instrument LT-460 and pressure instrument IT-456 is shared. When the upper tap breaks, tie j ofAnswer reference leg for LT will flash, resulting in indicated increasing Pzr level. Pzr pressure will drop due to the vapor j l space leak (the pressure instrument on tap would rapidly indicate depressurization). PRT conditions are within i

normal range, NOT as expected for open PORV. Channel II (LI-460) is NOT available as a CONTROL channel i for Pzr level. Channel 11 (PI-456) failure would NOT result in any actuations.

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3 Tier: l Emergency and Abnormal Plant Evolutions l RO Group: l2 l SRO Group: j2 Systems / Evolution Number: l 008 l Systein/ Evolution"Iltle: l PressurizerVaporSpace Accident 2.'J3

_ j '

C@ri: l AA2 l Ability to determine and ir.terpret the following as they apply to Pressurizer Vapor Space Accident:

KA: l AA2.12 l PZR level indicators RO Value: l?.4 l SRO Value: l3.7 l CFR: l 43.5 /45.13 Reference Reference Nunnber Reference See*ka Pane Number (s) Revision learn. Obj PRESSURITER PRESSURE 0300-000.00S- IV.A.2.a; IV.B.2.b & 16;25-26 2/20/97 3,12 AND LEVEL CONTROL PZRPale00 c.1) '

PRESSURIZER PRESSURE 0300-000.00S- IX.B.2.f 44 3/20/97 AND LEVEL CONTROL PZRP&I 00 EOP-LOCA-01, Loss of 0300-000.005- 2.1 20-22 3.A Reactor Coolant and Loss of LOCA01-00 Coolant Accident Analysis Qth Source New Question Modification Method Question Source Comments: l Material Required for E==md== tion:

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Quescem Top!a l SMM inputs Which one of the following correctly lists the parameters used by the Subcooling Margin Monitor to calculate subcooling margin?

a. RCS pressure and CET temperatures.

b. Pressurizer pressure and CET temperatures. ,

c. RCS pressure and RCS hot leg temperatures.

d. Pressurizer pressure and RCS hot leg temperatures.  :

Aus: la l Examslevel: lS l Cognitive Level: l Memory l Explanation The temperature margin to saturation is calculated by the CETPS. The inputs to this calculation are the ofAnswer iepresentative CET temperature, RCS pressure Containment pressure, and Containment radiation level. When Containment pressure exceeds 4 psig or Containment Radiation exceeds IES R/hr, the SMM automatically transfers to " ADVERSE" and uses the ADVERSE steam table values.

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'Her: l Emergency and Abnormal Plant Evolutions l RO Group:

I2 l SRO Group: l2 SystemafEvolution Number: l 009 l SysteavEvolution

Title:

l Small Break LOCA Category: l EAl l Ability to operate and / or monitor the following as they apply to Small Break LOCA:

KAt l EAl.16 l Subccoling margin monitors

! M Value: l4.2 l SRO Value: l4.2 l CFR: l 41.7 / 45.5 / 45.6 Reference Refenwace Number Reference Section

" Paac Number (s) Revision Imrn. Obj INCORE NUCLEAR 0300-000.00S- IV.E.2 33 1I/13/96 5.e; 7.f

, INSTRUMENTATION INCORB-00 SYSTEM, INCLUDING THERMOCOUPLES AND SATURATION MARGIN MONITOR 4

Y e

Question Souste New Question Modification i Method Question Source Comunents: l Material Required for Fwandandam:

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Page 124

J Questies Topic: l IDCA during shutdown Given the following conditions for Unit 2:

o _ RCS temperature- 325 'F o RCS pressure- 340 psig o - 21 RHR loop has been providing shutdown cooling for the past 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> o 21 Charging Pump is aligned for normal charging and seal injection o S2.OP-AB.14CA-0001(Q) " Shutdown LOCA" has been entered i due to lowering Pressurizerlevel o Pressurizer level - 9%, lowering -apidly o' letdown and Excess Letdown valves have been closed i

Based on these conditions, which one of following correctly describes the next operator actica that should be '

performed?

1 a.- Initiate a MANUAL Safety Injection

b. Start all available ECCS pumps I

c. Start 22 RHR Pump

d. Stop 21 RHR pump i

Anst Id l Exame Level: IS l Cognitive Level: l Comprehension l  ;

Fwgdemana= 1AW S2.OP-AB.LOCA-0001, Continuous Action Summary, if at any time Pzr level is < 11.0%, then stop the ofAnswer operating RHR pumps aligned for shutdown cooling. Another CAS action with the same conditions calls for raising Charging flow OR starting ECCS pumps one at a time, but this is only applicable after first pass through the ;

procedure (to step 3.124/3.125). CAUTION about starting second SI Pump. SI should NOT be actuated.

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Tier: l Emergency and Abnormal Plant Evolutions l RO Group: l2 RO Group: l2 SysteasfEvolution Number: l009 l Systena/ Evolution 'Iltle: l Small Bicak LOCA

~

Catenery: l2.4 { Emergency Procedures / Plan KA: 2.4.9 h5owledge oflow power / shutdown implications in accident (e.g. LOCA or loss of RHR mitigation strategies).

RO Value: l3.3 l SRO Value: I3.9 lCFR: l 41.10 / 43.5 / 45.13 Referisce Reference Number Reference Section Pane Nunnber(s) Re M en I4arn. Obj SHUTDOWN LOCA 0300-000.00S- II.C.1 1I 4/3@7 6,7 ABLOCA-01 Shutdown LOCA St.OP-AB.LOCA- Continuous Action 1 2 fMl(Q) Sununary. 5 0 Question Source New Question Modi 5 cation Method Question Source Conmsents: l Material Requiswd for E===dantion: '

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Question TopkE l RHR valveInEriocks Given the following condilicIs for Unit '2:

o SI has actualed due to a large break LOCA j

o RWSTLEVELLOW alarm has actuated o 22SJ44 RHR Pump Suction Valve to Containment Sump did , 1 NOT open automatically. I

\

-\

Which one .of the foHowing conectly identifies an operation that requires stopping 22 RHR pump in order to i complete the switchover to Cold leg Recirculation? j

a. Closing 22RH4 RWST to RHR Pump Suction Valve

b. Closing 22SI45 RHR to Charging /SI Pump Suction Valve l

c. Closing 22RH19 RHR Hx Discharge Cross-connect Velve

d. Closing 22CS36 RHR System to CS System Isolation Valve  ;

Ams: la l Exam Leel: lS l Consitive Level: l Cm. A ion l l F W ea= With interlock between Sump suction (to open) and RWST suction (to close), the RHR pump must & stopped to ofAnswer manually swap suction sources. The other valves are in required position (RH19) but have NO inteilock, have  ;

inverse interlocks (CS36) and (SJ45) , and do not require stopping RHR pump.  !

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'Iler: l Err.ergney and Abin.. l Plant Evolutions l RO Group: l2 l SRO Group: l1 SysteenfEvolution Namiber: l 011 l Systeni/ Evolution

Title:

l Large Break LOCA j

Ca" y: l EAl l Ability to operate and / or monitor the following as they apply to imge Break LOCA-KAt l EA1.13 l Safety injection components R'2 Value: l 4.l* l SRO Value: l4.2 lCFR: l 41.7 / 45.5 / 45.6 Reference Reference Nunnber Reference Section Pane Nuenber(s) Revision Learn. Obj EOP-IACA-03, TRANSFER 0300-000.00S- 3.3.4.1 14 4/6/97 4 TO COLD LEG LOCA0342 RECIRCULATION RESIDUAL HEAT 0300-000.00S- IV.B.4.c.9); IV.B.4.f; 23;25;26 04/23/96 6.c REMOVAL SYSTEM RHR00041 IV.B.4.h CONTAINMENT SPRAY 0300-000.00S- IV.B.9.f 26 6 SYSTEM CSPRAY-00 Question Source New Question Modification Method Question Source Comunents: l Material Required for Exannimation:

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Questles Topic: l LOCA eval l Given the following conditions for Unit 2: l i

o A SI han -----

o RCS Pressure - 1200 psig and decreasing i

o PZRIevel-offscalelow o Core Exit Thermocouple temperatures -. 535 'F o Containnent pressure - 9 psig and rising o Containment Rad monitors are in ALARM  !

o SG Pressures -775 nsig and decreasing slowly 1 Which one of the following correctly identifies the event that is occurring?

a. A pressurizer PORV has fully opened

b. A LOCA has occurred on the 21 loop cold leg

c. ONE SG has a steamline mpture inside containment  ;

d. A feedwater line to ONE SG has mptured and MSIVs have failed to close  !

i Ass: lb l Exam Level: !S l Commitive Level: l Co. Adon l I Explanation The conditions with all SG nearly the same pressure and lowr than normal, with RCS temperatures remaining high  ;

cfAnswer (W/I 107 of expected post trip semps) is indicative of a LOCA. Rad monitors would NOT be expected to increase i on secondary problem. PZR level going offscale low with SG pressure lowering is NOT expected for PORV I failure.

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Tiers l Emergency and Abnormal Plant Evolutions l RO Group: l2 l SRO Group: lI SystemvEvolution Nusuber: l 011 l SystenvEvolution

Title:

l Large Break LOCA Category: l EA2 l Ability to determine and interpret the following as they apply to Imge Break LOCA KAt l EA2.13 l Difference between overcooling and LOCA indications r:0 Value: l 3.7* l SRO Value: l 3.7* l CFR: l 43.5 / 45.13 Refersace Reference Number Reference Se Pare Number (s) Revision Learn. Ob]

EOP-1.OCA-01, Loss of 0300-000.00S. 1.6.9 16-17 7/26/96 3 Reactor Coolant and Loss of LOCA0100 Coolant Accident Analysis EOP-TRIP-1, REACIOR 0300-000.00S-TRP001- 7.3.26;7.3.28 68;70 SI10/96 23,25 TRIP OR SANTY 01 INJECTION AND INTRODUCTION TO THE USE OF EOPs Reactor Trip Or Safety 1-EOP-TRIP-1 steps 26; 28 sheet 4 Injection Question Source New Question Modification Method  !

Question Source Conansents: l Material Required for Evandmadon:

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0uesties Tosder l RCP Seal Failure eval Given the following conditions for Unit 2:

o Reactor power- 100%

o The following console alarms have actuated:

SEAL EAKOFF FLOW HI-LO STANDPIPE EVEL HIGH i

1. 1 SEALLOW D/P o l Seal 12akoff Flow Recorder pens indicates one reading high 4 o Charging flow rose by 40 gpm to maintain pressurizer level l

l Which one of the following correctly identifies the RCP seal failure that resulted in those indications, assuming NO operator action?

1

a. No. I seal on 21 RCP has failed and the RCS pressure drop is across the No. 2 seal.

b. No. I and No. 2 seals on 22 RCP have failed and all seal injection flow is directed to the RCDT. j

c. No. 2 and No. 3 seals on 23 RCP have failed and the RCS pressure drop is across the No. I seal,

d. No. 2 seal on 24 RCP has failed and the additional real injection flow has filled the RCP standpipe.

Ams: la l Exam Level: IS l Commitive Level: l Comprehensior. I Explaantion With No. I seal failed, full RCS pressure goes across No. 2 seal. Indications of failure include: increased No.1 seal l ofAnswer leakoff flow (until isolated), increased No.2 seal leakoff flow (increase in standpipe level), and low seal D/P. If j only No. 2 seal failed, any condition that filled the standpipe would NOT have rcsulted in increased No. I seal leakoff flow indication. With No. 2 & No.3 seal failures, standpipe low level alarms would come ia without the No.

I seal alarms. If No.1 and No. 2 seals have failed nos sll leakoff flow goes to the RCDT, since No. I seal leakoff continues to the VCT.

. Page 131

i Tier: l Emergency and Abnormal Plant Evolutions l RO Group: lI l SRO Group: l1 i Systeen/ Evolution Nuc. . l 015 l Systena/ Evolution

Title:

l Reactor Coolant Pump Malfunctions Cater. y: l AA2 l Ability to determine and interpret the following as they apply to Reactor Cnalant Pump Malfunctions:

KA: l AA2.01 l Cause of RCP failure RO Value: l3.0 l SRO Value: l 3.5* l CFR: l 43.5 /45.13 Reference '

Reference Number Reference Section Page Nusnber(s) Revision Learn. Obj REACTOR COOLANT 0300 400.00S- II.A.I.b; III.A.2.i & k 9;12-13 10/28/96 1.B PUMP ABNORM ALITY ABRCPI-00 l REACTOR COOLANT 0300400.00S- IV.B.8.g.6) 21 6/10/96 8

PUMP RCPUMP-01 j REACTOR COOLANT S2.OP-AB.RCP- 3.4; Tech Basis Doc 3;6

PUMP ABNORMALITY 0001(Q) 2.4.D Question Source New Question Modification '

l, Method j Question Source Comments: l l M terial Required for l Exannuation:

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Questise Topic: I VCT level control Given the following conditions for Unit 2:

o VCTis at 16%

o An Instrument technician mistakenly cracks open the transmitter equalizing valve forir.strument LTI12.

Which one of the choices correctly completes the following statement conceming the meter (U112) my and the affect on the system?

l Indicated VCr level will...

i

a. lower and .aitomatic makeup will start filling the VCT.-

b. rise and automatic makeup will start shortly thereafter.

c. rise and 2CV35, Diversion Valve, diverts to the CVC HUT.

j

d. lower and 2CV40 and 2CV41, VCT Discharge Stop valves, close.

Ans: lc l Exasa Level: lS l CWe Level: l C= sehension l Nwyammesia= LT112 controls: 1) Auto makeup stop and start based on VCT level, and 2) Full diversion ofletdown from the j etAnswer VCT to the HITr. Opening the equalizauon valve results in level indicating 300% instrument span. At 87% on LI- i 112. the full diversion will occur. Makeup to VCr is blocked since LT 112 bitiates auto M/U at 14%. Eventually  !

actual VCTlevel will go offscale low and Charging pumps suction may be lost. Auto swap to RWST(isolation of VCT Discharge Stop valves) requires BOTH VCr level channels at low-low setpoint (3.57%) and 2SJ1 & 2 fully open. t 4

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l Tier: l Emergency and Abnormal Plant Evolutions l RO Group: l2 l SRO Group: l2

'Sy=Tvolution Namiber: l 022 l Systent/ Evolution 'ntie: l Loss of Reactor Coolant Makeup Catesory: l AAl l Ability to operate and / or monitor the following as they apply to Loss of Reactor Coolant Makeup:

KA: l AAl.08 l VCTlevel RO Value: l3.4 l SRO Value: l3.3 l CFR: l 41.7 / 45.5 / 45.6 Reference Reference N=Ar Reference Section Pase Number (s) Rev'8- Learn. Obj CHEMICAL AND 0300400.00S- IV.C.15; IV.G.1 38;59 10/4/96 4.a.xiv VOLUME CONTROL CVCS0040 SYSTFM CHEMICAL AND 0200-000.00S. V.b.2.o.2) 77 10/4/96 8 VOLUME CONTROL CVCS00-00 SYSTEM j Question Source New Question Modification Question Source Conunents: l

Material Required for Examination

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W Topic: l Rapid boration Given the following conditions for Unit 2:

o An ATWS event is m progress o RCS presstue-2250 psig o Main turbine has been tripped manually o SIhas NOT actuated o 2-EOP-FRSM ," Response to Nuclear Power Generation", has been implemented Which one of the following correctly describes the preferred boration flowpath?

a. Aligning the charging pumps suction to the RWST by opening 2SJ1 and 2SJ2, and fully opening the normal charging path flow control valve 2CV55.

b. Aligning the charging pumps suction to the RWST by opening 2SJ1 and 2SJ2, and opening the BIT Inlet and Outlet valves 2SJ4,2SJ5,2SJ12 and 2SJ13.

c. Running at least one Boric Acid Transfer Pump in FAST, opening the Rapid Borate Stop Valve 2CV175, and opening the normal charging path flow control valve 2CV55.

d. Running at least one Boric Acid Transfer Pumps in FAST, opening the Rapid Borate Stop Valve 2CV175, and opening the BIT Inlet and Outlet valves 2SJ4,2SJ5,2SJ12 and 2SJ13.

1 Anst lc l Exam Level: lS l Coanitive Level: l Comprehension l Fwgdom. tion The intended buration path here is the most direct one available, using the normal charging pump (s). Several ,

of Answer means are usually available for emergency boration and are specified in order of preference: (1) Rapid Boration with the Charging Pumps taking a suction directly from the Boric Acid Pumps, (2) Alignment to RWST without SI, (3) Alignment to RWST with SI actuated.

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Tier: } Emergency and Abnormal Plant Evolutions l RO Group: I1 l SRO Group: l1 Systems / Evolution Number: l 024 l Systein/ Evolution

Title:

l Emergency Boration Catesory: l 2.1 l Conduct Of Operations KA: 2.1.7 Ability to evaluate plant performance and snake operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

R7 Value: l3.7 l SRO Value: l4.4 l CFR: l 43.5 / 45.12 / 45.13 Referwece Reference Number Reference Section Page Number (s) Revision Learn. Obj I EOP-FRSM-1 and 2 0300-000.00S- 3.2.3.1 19-20 4 0/97 2.A; 5.A

, Response to Nuclear Power FRSM00-02 l

Generation '

CHEMICAL AND 0300-000.00S- V.B.2.ee.1); IX.A.8.b 97;133 10/4/96 8;12 VOLUME CONTROL CVCS00-00 <

SYS7EM '

l Response To Nuclear Power 1-EOP-FRSM-I step 3.1 sheeti Generation Question Source New Question Modi 5 cation M&M '

Question Source Comuments: l Matedal Required for  ;

Exannimation.

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Quessies Tepic: l Ahernatecooling Abnonnal procedure S2.OP-AB.RHR-0001 "less of RHR", states that hot leg injection is the preferred l altemate method for removing decay heat if RHR is lost and core exit thermocouples are reading 2200 *F.

l Which one of the following is NJ.I a valid trason for why hot leg injection is the preferred alternate method, as compared to other procedural means? '

a. Reflux boiling requires a minimum number of steam generators to be available as a heat sink t

b. Both natural circulation and forced flow (RCP) cooling require one or more steam generators to be available as a heat sink l

c. Pressurization of the RCS may inhibit and/or lower flow to the core if cold leg injection is used

d. Cold leg injection requires opening / ensuring a RCS vent path. A RCS vent path is not required for ,

hot leg injection '

Ams: Id l Exasslevel: lS l Commitive level: l C==archension l

• F ." Both Cold leg and hot leg injection are feed and bleed-type operations until RHR cooling is re-established.

i ofAnswer According to the AB Basis, all of the other statements are correct.

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Tiers l Emergency and Abnormal Plant Evolutions l RO Group: l2 l SR9 Group: l2 Systion/ Evolution Number: l 025 l SystenvEvolution

Title:

l less of Residual Heat Removal System Catge.f: AK3 Knowledge of the reasons for the following recoonses as they apply to Loss of Residual Heat Removal System:

KAt l AK3.01 l Shift to alternate flowpath RO Value: l3.1 l SRO Value: l3.4 l CFR: l 41.5,41.10 / 45.6 / 45.13 Reference Reference Number Reference Section Revision Learn. Obj Pane Number (s)

LOSS OF RHR 0300-000.00S- V.A.3 12 10/29/96 4,5 ABRHRl-00 loss of RHR S2.OP-AB.RHR-0001 step 3.28 and 8; I 8 Attachment 7,1.0 F.

LOSS OF RHR S2.uP-AB.RHR-0001 Attachment 7 9 TECHNICAL BASES DOCUMENT Question Source New Question Modification Method '

Question Source Comments: l Material Required for Examination:

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1 Page 138 l

W T opic: l CCW leak

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Given the following conditions for Unit 2:

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o A LOCA has occurred o Actions of 2-EOP-LOCA-3 " Transfer To Cold I.eg Recirculation" have been completed o Two CCW Pumps are running Which one of the following correctly identifies a consequence resulting from a tube leak in the CVCS letdown HX?

a. A high level alarm will actuate for the CCW Surge tank

b. The CCW Pump supplying the non-safety related header will eventually trip due to loss of NPSH

c. The CCW Pump supplying the safety related header will eventually trip due to loss of NPSH

d. Both CCW Pumps will eventually trip on loss of NPSH Ans: lb l Exam level: lS l Cognitivelevel: l Comprehension i Frplammelaa At the completion of EOP-LOCA-3, the CCW systems are aligned as independent loops with 22 CC loop supplying ofAnswer the non-safety loads. The letdown HX is NOT isolated (unlike SFP HX outlet and excess Itdn HX) and, since letdown is isolated by the CIA signal (or low Pu level), CCW pressure will allow in-leakage to CVCS. The CCW Surge Tank internal baffle ( top at 37.5 % (24")) indicated level will prevent loss of suction (NPSG) to the 21 CCW Pump but the side the 22 CC Pump takes suction from will fall and NPSH lost at 20' head.

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Page 139 t_

Tier: l Emergency and Abnormal Plant Evolutions l RO Group: l1 l SRO Group: lI System / Evolution Nunnber: l 026 l Systena/ Evolution'ntle: l Loss of Component Cooling Water Catesory: l AA2 l Ability to determine and interpret the follcwing as they apply to Ims of Component Cooling Water:

d KAt l AA2.02 l The cause of possible CCW loss RO Value: l2.9 l SRO Value: l3.6 l CFR: l 43.5 /45.13 Reference Reference Nunnber Reference Section Pane Number (s) Revision Learn. Obj COMPONENTCOOLING 0300 400.00S- IV.B.I.b.3) & 4) 19 1/15/97 3.c; 4.a WATER CCW000-01

., EOP-LOCA-03, TRANSFER 0300-000.00S- 3.3.39 41 4/6/97 2 TO COLD LEG LOCA03-02 RECIRCULATION No. 2 Unit Componera 205331 Sheet 1 & 2  ;

Cooling

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Question Source New Question Modification j Question Source Comments: l l

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l Quesdom Tosde: l Pzr pressure channels i Given the following conditions for Unit 2:

o Pzr pressure-2235 psig o RCS temperature - 547 'F o Channel IU (PT-457) hr.s been : elected as controlling channel  !

Which one of the following conectly describes Pressurizer pressure response ifIrr-457 fails LOW and NO t operator action is taken?

a. Rise until Pressurizer Spray valves open.

f

b. Rise until both "rs-uk PORVs open.  ;

c. Rise until ONE Pressurizer PORV opens. I

d. Rise until ONE Pressurizer Code Safety lifts.

i Ans: lc l Examalevel: lS l Commidve tevel: l Comprehension l I F=; " "' 7 Failure low will cause all pressurizer heaters to energize. Pressure will rise and the spray valves will not open.  ;

ofAnswer Pressure will continue to rise until the PORV w/o the failed channel opens. '

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Tier: l Emergency and Abnormal Plant Evolutions l RO Group: lI l SRO Group: l2 Systema / Evolution Neanber: l 027 l Systent/ Evolution

Title:

l Pressurizer Pressure Control Malfunction Catemory: l AA2 l Ability to determine and interpret the following as they apply to Pressurizer Pressure Control Malfunction:

KAt l AA2.03 l Effects of RCS pressure changes on key components in plant 17 O Vaine: l3.3 l SRO Value: l3.4 l CFR: l 43.5 /45.13 Referumee Reference Nassber Reference Section Pase Nusabeds) Revision Learn. Obj Pressurizer Pressure OVXMX)0-00S- II.C.1 step 2 12 7/15/97 1, 3.B Malfunction ABPZRI-00 PRESSURIZER PRESSURE 03004)00.00S- IV.B.I.f & I;IX.B.I.a.2) 18 & 24-25;40 3/20/97 3.a.I. 8,12 AND LEVEL CONTROL PZRP&I 00 RPS - Pressurizer Pressure & 221060 sh.I1 Level Control Logic Diagram Question Source New Question Modnication Method Question Sourre Comunents: l Matedal Required for Emandmation:

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.0 Page 142

Q-=da= Tople: l Pzr level alarms Given the following conditions for Unit 2:

o. Powerlevel-75%

o An electronic circuit probleni causes 2CV55, Charging Flow Control Valve, to go full open Assuming no operator action, which one of the following is the closest to what actual Pressurizer level indication will be reading when the annunciator PZR HTR ON LEVEL HIGH (OuA E-20), actuates?

L 44 %

b. 50%

c. 55 %

d. 64%

Amst Ib l E=== level: lS l C==itive level: I Comprehension l

, Expleastion Control channel will turn on BU heaters when levelis 5% above setpoint and cause OHA E-20 actuation. Pzr level ofAnswer changes from 22.3% @ 0% power to 50.7% @ 100% power (if Tavg is on program). At 75% power, Pzr level would read 44%, so the alarm is at 49%.

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Tier: l Emergency and Abnormal Plant Evolutions l RO Group: l3 l SRO Group: l3 l SystenvEvolution Number: l 028 l SystenvEvolution

Title:

l Pressurizer Level Control Malfunction Catesory: l AA2 l Ability to determine and interpret the following as they apply to Pressurizer Level Control Malfunction:

KA: l AA2.01 l PZR levelindicators and alarms RO Value: l3.4 l SRO Value: l3.6 l CFR: l 43.5 / 45.13 I

Reference Reference Nussber Reference Section Pene Number (s) Revision Iearn. Ob] {

PRESSURIZER PRESSURE 0300-000.00S- IV.B.2.c.1).a); 26 3/20/9i 4.j; 6.g AND LEVEL CONTROL PZRP&I 00 IV.B.2.c.2)

PRESSURIZER PRESSURE 0300-000.00S- V.D.7 33 3/20/97 8 AND IEVEL CONTROL PZRP&I 00 l OVERHEAD S2.OP-AR.Z7 0005(Q) window E-20 32-33 ANNUNCIATORS i

WINDOW E I l

Question Source NRC Exam Bank Question Modification Method Question Source Conunents: l Material Required for Exandmation:

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Page 144

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j Oueseles Topic: l AMSAC Given the following conditions for Unit 2: l o Reactor power- 50%

o The onlyrunning SGFP trippw o j o ~SGlhenarrow reactor and turbine failed to manuaHy trip range levels are now offscale low o Control rods are being manually inserted per 2-EOP-TRIP-1 o Unloading rate -20%/ minute )

Which one of the following correctly describes of the expected response of the ATWS Mitigation System Actuation Circuitry (AMSAC)?

a. Trips the reactor through diverse circuits, indirectly causing a turbine trip 5

b. Trips the turbine through redundant circuits and starts all AFW pumps

c. Trips the turbine through redundrat circuits and starts only the motor driven AFW pumps _

i

d. AMSAC will NOT actuate since power level will be less than the C-20 setpoint before the actuation delay expires i

Ams: lb l Exnas Level: lS l Ca==ative Imel: l Cn=anhension I l F"

"Ihe AMSAC circuit will actuate due to > 3/4 SG levels < 5% foi 25 seconds. With (turbine) power > 40% (C-20),

ofAnswer the AMSAC

• active" condition will remain in effect for 260 seconds after power decreases < 40%. AMSAC trips the turbine via the 20/Jr and 20 ET trip solenoids (different trains) and independently swts all AFW pumps.  ;

Additional action is idation of SGBD & SG camaling via AFW circuits. I i'

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Tier: l Emergency and Abnormal Plant Evolutions l RO Group: l2 l SRO Group: l1 System / Evolution Number: l 029 l Systena/ Evolution

Title:

l Anticipated Transient Without Scram Catemory: lAAI l Ability to operate and / or monitor the following as they apply to Anticipated Transient Without Scram:

KA: l AA1.15 l AFW System RO Value: l4.1 l SRO Value: l3.9 l CFR: l 41.7 / 45.5 / 45.6 Refenece Reference Number Reference Section Pate Number (s) Revision Learn. Obj ATWS MITIGATION 0300-000.00S- IV.B.2.a.3).b) & .5).b) 15;16-17 9/6/96 4 SYSTEM ACTUATION AMSA N CIRCUITRY (AMSAC)

ATWS MITIGATION 0300-000.005- IV.C.2.c; IV.C.2.d 23 9/6/96 7 SYSTEM ACTUATION AMSACO-00 CIRCUITRY (AMSAC) _

Question Source New Question Modification Method Question Source Conunents: l Matedal Required for Exandnation:

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l Quesdes Topic: l ATWS conditions '

Wluch one of the following correc J.y describes the reason why it is worse for a full power ATWS event to l occur at the Beginning-of-life O'JL) as compared to the End-of-Life (EOL)?

l

a. The additional burnable poisons provide less heat conduction; therefore, the fuel pin outer clad temperatures are higher.  ;

b. The effective delayed neutron fraction is higher; therefore, the rate of power reduction is slower.

c. The Moderator Temperature Coefficient (MTC) is less negative; therefore, the reactor power reduction due to heat addition is less.

d. The higher boron concentration in the RCS causes the emergency boration to be less effective; therefore, it takes longer to achieve adequate Shutdown Margin (SDM).

Amst lc l Exam Level: lS l Cognitive Level: l Memory l

^'-

F=;' An A'IWS event is more severe early in core life (least negative MTC) due to RCS pressure response; pressurc  ;

of Answer could exceed safety setpoints ,

f Page 147

Tier: 1 Emergency and Abnormal Plant Evolutions l RO Group: l2 l SRO Grr,sp: l1 S3." Tvolution Nasaber: l 029 l Systesm/ Evolution

Title:

l Anticipated Transient Without Scram Ca'.,,sy: AKI Knowledge of the operational implications of the following concepts as they apply to Anticipated Transient Without Scram:

KAt l AKl.05 l definition cf negative temperature coefficient as applied to large PWR coolant systems FO Value: l2.8 l SRO Value: l3.2 l CFR: l 41.8 / 41.10 / 45.3 _

Reference Reference Number Reference Section Page Nusnber(s) Revision Learn. Ob]

EOP-FRSM-1 and 2 0300-000.00S- 1.3.4; 1.4.5 8-9 4987 2.A; 5.A Response to Nuclear Power FRSM00 02 Generation Question Source NRC Exam Bank Question Modification Method i

Question Source Conuments: l Material Required for F.ramination:

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Page 148

Querden Topic: l SR failure at power Given the following conditions for Unit 2:

o Reactor power- 100%

o SR DET VOLT TREL(OHA E-5)is in alarm o Source Range readings - offscale high (N31); 0 cps (N32)

Which one of the following correctly describes the cause of the event in progress?

a. The block of the high voltage power supply to N31 has failed.

b. The instrument power fuse for source range N32 channel has blown.

c. An I&C technician has placed source range N31 level Trip switch in BYPASS.

d. The operator has inadvertently depressed the Reset Source Range A pushbutton.

Aus: la l Examalevel: lS l Commitive level: l Comprehension l Explanation The only condition that would give the indications listed is inadvertent repowering of SR NIS channel. Instrument ofAnswer power fuse would haver no impact, as would operation of only one pushbutton. If both pushbuttons depressed and P-10 auto blocked failed, BOTH channels energize, and reactor trips. Placing the Ixvel Trip in BYPASS causes OHA A-5, NIS CH ON TEST and 10) E-29, SR & IR TRIP BYP to alarm.

Page 149

"Iler: l Emergency and Abnormal Plant Evolutions l RO Group: l2 l SRO Gewap: l2  ;

SysteaWEvolution Number: l 032 l System / Evolution

Title:

l Loss of Source Range Nuclear Instrumentation Category: AA2 Ability to determine and interpret the following as : hey apply to Loss cf Source Rang- Nuclear Instrumentation:

KA: l AA2.05 l Nature of abnormality, from rapid survey of control room data RO Value: l 2.9* l SRO Value: l 3.2* ! CFR: l 43.5 /45.13 Reference Reference Number Reference Section Paac Nuanber(s) Revision Leers. Obj_

NUCLEAR 0300-000.00S- IILC step 18 13 6/5/97 I INSTRUMENTATION ABNISI-00 SYS1EM MALFUNCTION EXCORE NUCLEAR 0300-000.00S- IV.C.3.j.2).f); 27;70 9/4/96 7.c & f; 9.c INSTRUMENTATION EXCORFA0 V.B.3.b.1)

SYSTEM NUCLEAR S2.OP-AB.NIS- step 3.18 5 INSTRUMENTATION 0001(Q)

SYSTEM MALFUNCTION Question Source New Question Modification Method Question Source Comments: l Material Required for Exasmination:

l Page 150

l W Topic: l IR failure shutdown  !

Given the following conditions for Unit 2:

o A reactor startup is in progress o Source range counts - 8200 cps (N31) and 8300 cps (N32) o Intermediate range amps - 8E-11 amps (N35) and 9E-11 amps (N36) o Control Bank D- 136 steps  !

)

Which one of the following correctly identifies the power limits imposed ifIntermediate Channel N35 fails i LOW? i

a. Reactor power may NOT exceed IE-10 amps until N35 is repaired.

b. Reactor power may be raised to, but CANNOT exceed 5% until N35 is repaired.

c. Reactor power may be raised to, but CANNOT exceed,10% until N35 is repaired.

d. Reactor power must be maintained below 10E5 cps to avoid an automatic reactor trip  !

Ans: la l Exam Level: lS l Cognitive Level: l Comprehension l )

Frplanation Per SO and Technical Specification, with power below P-6 (IE-10 ICA) the inoperable channel must be restored to ofAnswer OPERABLE prior to increasing power above P-6 setpoint.

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Tier: l Emergency and Abnormal Plant Evolutions l RO Group: l2 l SRO Group: l2 SystenvEvolution Number: l 033 l System / Evolution

Title:

l Loss ofIntermediate Range Nuclear Instrumentation Category: AK3 Knowledge of the rec.ons for the following responses as they apply to Loss of Intermediate Range Nuclear Instrumentation:

KAt l AK3.01 l Termination of startup following loss of intermediate range instrumentation

_ R3 Value: l3.2 l SRO Value: l3.6 l CFR: 'l 41.5,41.10/45.6/45.13 Reference Reference Number Reference Section Page Number (s) Revision Learn. Obj EXCORE NUCLEAR 0300-000.00S- VIII.B; IX.A.1 75-77 9/4/96 11 INSTRUMENTATION EXCORE-00 SYS11EM 3 NUCLEAR S2.OP-SO.RPS- steps 5.5.9-5.5.11 11 2 2

INSTRUMENTATION 0001(Q)

CHANNELTRIP /

RESTORATION SJem - Unit 2 Technical 33.1.1 - Table 33-1 3/43-6 Specifications FU 5, ACTION 3.a Question Source NRC Exam Bank Question Modification Method Question Source Comments: l Material Required for Exandnation:

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Page 152 l

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I Quesenen Topic: l So leak avaluation Given the following conditions for Unit 2:  !

o Reactor power- 100%

o identified leakage - 9.7 gpm o A SG tube leak has been verified for 21 SG o Trending shows: ,

- 1000- 30 spa

- 1030- 45 gpd

- 1100 - 70 gpd  !

- 1130 - 100 gpd

- 1200- 120 gpd

- 1230 - 125 gpd

- 1300 - 145 gpd

- 1330 - 160 gpd

- 1400 - 180 gpd Based on indications and trending, which one of the following correctly desuibes the required action?

a. A unit shutdown is required due to conditions as of 1130.

b. A Unit shutdown is required due to conditions as of 1300.

c. 100% power operation may continue until SG tube leakage exceeds 432 gpd.

1

d. 100% power operation may continue until SG tube leakage exceeds 500 gpd.

l Ans: lb l Exam Level: lS l Cognitive Level: l Comprehension l Explanation in the cases where leakage is projected to or actually exceeds 140 gpd, the unit is removed from service based on ofAnswer the rapid propagation of tube leaks. The other finnting value is a 60 gpd increase per hour. The Technical Spec I limit is 500 gpd per any SG, but the AOP requires action to remove unit prior to exceeding this limit. The other limit for Identified Leakage does NOT apply to SG tube leakage limits.

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4 Tier: l Emergency and Abnormal Plant Evolutions l RO Group: 'l2 l SRO Group: l2 .

Systems / Evolution Number: l 037 l SystenvEvolutionTitle: l Steam Genceator Tube Leak Catesory: l2.1 l Conduct Of Operations KAt l 2.1.32 l Ability to explain and apply all system limits and precautions.

l'O Value: l3.4 l SRO Value: l3.8 l CFR: l 41.10 / 43.2 / 45.12 Reference Reference Number Reference Section Pase Number (s) Revision Learn. Obj

STEAM GENERATOR 0300-000.00S- III.C.I.a 11 10/24/96 3,4 1 TUBE LEAK ABSG01-00 STEAM GENERA 70R S2.OP-AB.SG-0001(Q) step 3.17 7 10 1

TUBE LEAK

STEAM GENERATOR S2.OP-AB.SG-0001(Q) step 3.16 8-9 l

TUBEIEAK TECHNICAL BASES DOCUMENT

j. Question Source New Question Modification Method Question Source Comments: l l

Material Required for Evad=ation:

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j onese6en T wic: ISolevel l Which one of the following correctly describes the basis for maintaining greater than 9% narrow range level '

4 in the ruptured steam generator as directed by 2-EOP-SGTR-1" Steam Generator Tube Rupture" ? i i

t a. Assures adequate heat sink maintained in the ruptured SG in the event it is required for cooldown

] b. Maintains a thermal stratification layer above the top of the tubes to prevent depressurization of the ruptured SG a

c. Maintains releases from the ruptured SG less than 10CFR100 limits by absorbing short-lived ,

radioactive gases i

d. Prevents excessive thermal stress across the ruptured SG tubes reducing the potential for increasing j the severity of the rupture ,

)

Ams lb l Exass Level: lS l Cosmitive level: I Memory l l

! F. Maintaining level in the narrow range in the ruptured SG is important to the overall recovery strategy. When the l ofAnswer ruptured SG is isolated and steaming stops, the liquid inventory becomes stagnant. A hot thermal layer will form at

the SG free liquid surface, insulating the steam space from the cooler liquid and tubes below. This pressure ,

difference is needed to maintain RCS subcooling. 'Ihe purpose is to establish and maintain a water level in the t j ruptured SGs above the top of the U-tubes in order to promote thermal stratification to prevent ruptured SG J depressurization.

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Tier: l Emergency and Abnormal Plant Evolutions l RO Group: l2 l SRO Group: l2 Systasm/ Evolution Number: l 038 l System / Evolution

Title:

l Steam Generator Tube Rupture Catemory: lEK3 l Knowledge of the reasons for the following responses as they apply to Steam Generator Tube Rupture:

KA: EK3.06 Actions contained in EOP for RCS water inventory balance, S/G tube rupture, and plant shutdown procedures i RO Value: l4.2 l SRO Value: l4.5 l CFR: l 41.5,41.10 / 45.6 / 45.13 s

Refersuce Reference Number Reference Section Paze Number (s) Revision learn. Obj 4 EOP-SGTR 1. STEAM 0300-000.00S- 4.3.6 40-41 4/4/97 7, 8 l GENERATOR TUBE SGTR0141 RUPTURE STEAM GENERATOR l-EOP-SGTR-1 6 sheet 2 20 l TUBE RUPTURE Question Source NRC Exam Bank Question Modification ,

Method j Question Source Comments: l ,

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A Quessies Topic: l Steam line leak i

Given the following conditions for Unit 2:

o Reactor power- 25%, rising ,

o RCS Tavg- 549 'F, lowering  :

o Pressurizer pressure - 2150 psig, loweiing  !

o Pressurizerlevel- 22%, lowering  !

o Turbine load is stable I o - SG pressures are - 950 psig (21SG); 900 psig (22SG); 950 psig (23SG); -

950 psig(24SG), alllowering l o MS los, atmospheric relief valves, indicate closed l o- Steam Dumps indicate closed ,

o Containment temperature and pressure are normal o Sound heard in the control room indicates that a MSSV may be open Which one of the following correctly describes the action to be taken for the above conditions? I

c. ImmeAiately close 22MS167 and initiate a rapid unit shutdown i

b. Immediately close 22MS167; initiate a reactor trip; initiate MANUAL SI(if necessary) j

c. Trip the reactor; close 22MS 167; initiate MANUAL SI (if necessary)  !

d. Trip the reactor; initiate Main Steamline Isolation (all loops); initiate MANUAL SI (if necessary) l Arm Id l Examslevel: lS l Cosaitive Iavel: l Comprehension i Explanation S2.OP-AB.STM-000), Attachment I (Continuous Action Summary) requires MANUAL reactor trip, MSLI, and SI of Answer (if necessary) if reactor power is rising uncontrollably i

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Tier: l Emergency and Abnormal Plant Evolutions l RO Group: lI l SRO Group: l1 Systein/ Evolution Number: l 040 l System / Evolution

Title:

l Steam Line Rupture Catesory: l AA2 l Ability to determine and interpret the following at they apply to Steam Line Rupture:

KA: l AA2.04 l Conditions requiring ESFAS initiation RO Value: l4.5 l SRO Value: l4.7 l CFR: l 43.5 / 45.13 Reference Reference Number Reference Section Page Number (s) Revision Learn. Ob]

EXCESSIVE STEAM 0300-000.00S- III.D.3.a 10 10/30/96 4.c FLOW ABSTMI-00 EXCESSIVE STEAM S2.OP-AB.STM- Attachment I 6 FLOW 0001(Q) )

Question Source New Question Modification Method

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Question Source Conunents: l Material Required for Examination:

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Quanden Taoic: l loss of vacuum eval Unit 2 is at 65% power when a loss of condenser vacuum occurs and the following parameters are observed:

Time - Power Generator MW Cond. Pressure (in. Hg) 0800 65 % 753 2.8 0801 60 % 698 3.7 0802 55 % 620 4.6

'0803 49% 555 5.2 0804 44 % 483 5.2 -

0805 35 % 433 5.4 0806 30% 380 5.4 0807 25 % 323 5.2 For the above conditions, which one of the following correctly describes the action (s) to be taken?

c. A turbine trip is required by 0808.

b. A reactor trip was required at 0803.

c. Turbine load should be stabilized at 25% and reset of the Load Rejection arming signal will be required

d. Turbine unloading should be continued and reset of the lead Rejection arming signal will be required Ams - I d l Rum == Level: lS l Cognitive Leveli l Application l Fwgdammela" Initiate a load reduction at $5% hr minute, which is required if backpressure exceeds ATTACH 4 limit (at 0803).

ofAnswer The final value given at 25% exceeds ATTACH 4 limit of 5.0 inches Hg for load reduction, even though backpressure has stabilized. Turbine load reduction ramp rates $5%/ minute are desirable to prevent operation of Steam Dumps, which could degrade the low vacuum condition. Since the rate did exceed 5%, steam dumps would have armed and could open on temperature difference. If the Steam Dump Anning signal is present due to the rapid load reduction and the Steam Dumps are no longer needed to help reduce load, the operator is directed to reset the load rejection signal.

L Page 159

Tier: l Emergency and Abnorreal Plant Evolutions l RO Group: l1 l SRO Group: l1 Systema / Evolution Number: l 051 l Systent/ Evolution

Title:

1 Loss of Condeaser Vacuum 3 Catesory: l AA2 l Ability to determine and interpret the following as they apply to loss of Condenser Vacuum:

_KA: l AA2.02 l Conditions requiring reactor and/or turbirm trip RO Value: l3.9 l SRO Value: l4.1 l CFR: l 43.5 / 45.13 Reference Reference Nunnber Reference Section Page Number (s) Reviskm learn. Ob]

IDSS OFCONDENSER 0300-000.00S- III.C.I .b i1/18/96 5 4 VACUIR1 ABCOND-00 STEAM DUMP SYSTEM 03004)00.00S- V.A.5.d 20 9/6/96 12 STDUMP-00 LOSS OFCONDENSER S2.OP-AB.COND- 3.13-3.24; Attachumnt 34 VACUUM 0001(Q) 4 Question Source New Question Modification Method '

Question Source Conenents

l Material Required for S2.OP-AB.COND-0001(Q), A'ITACHMENT 4 " CONDENSER.3ACKPRESSURE LIMITS" Ermannation:

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Page 160

Question Tonde: l SG depressurization Given the following conditions for Unit 2:

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o A loss of all AC has occurred. )

o Containment preasure is 2.5 psig i

o Steam geaerator depressurization has been initiated '

l 2-EO.P-LOPA-1, Loss of all AC Power, specifies that depressurization of the intact steam gent rators be -

terminated when RCS cold leg temperatures are <310 'F OR SG pressure is <235 psig.

Which one of die following correctly describes the basis for terminating the depressurization wl en RCS cold leg ternperatures are <310 *F7

a. To preve nt a chtllenge to the THERMAL SHOCK critical safety function status tree

b. Rapid cooldown below that SG piessure leads to voiding in the RV head, invalidating PZR level indication l

l c. Further coaldown leads to reactor criticality concerns since there is no boron injection flowpath asallable j d. To prevent RCS pmssum from lowering to a point where nitrogen from the ECCS accumuh tors is J

[ injected into the RCS Aas:,Ia f Exam Leveh l 5 l Cognitire Level: l Cootprehension l

~

Exdmotion Per th712PA Basin Dccument: Selectionit. is correct for 235 psig, Selection a. is correct for 310 *F. The others of An swer are not sugwtedh Basis Document and/or are not a concern at this point in the evolution.

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Page 161 W.

i lies: ) Eimgency and Abnormal Plant Evolutions l RO Group: l1 l SRO Group: l1 SystiserEvolution Nandwr: l 055 l Systesm/ Evolution litle: l Station Blackout Cates #,si: l 2.4 l Energency Procedures / Plan KAt l 2.4.18 l Knowledge of the specific bases for EOPs.

I O Vahns: l2.7 l SRO Value: l3.6 l CFR: l 41.10/45.13 Referesee Reference Nunnber Reference Section Page Nunnber(s) Revision f.nrn. Ob] l EOP-1DPA 1,2,3; LOSS 03004)00.00S- 43.373 49 4/15/97 5 i OF ALL AC POWER AND LOPA0042 l RECOVERY  !

Loss Of All AC l'ower and 1-EOP-LOPA-1 Step 37 (37.5) sheet 3 Basis Docunent i

I Question Source New Question Modification Method Quest $a Source Comunients: 1 Materfat Required for Fxmedstilon:

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P Page 162

Questies Topic: l Natural Cire Cooldown Which one of the following correctly describes the purpose of operating THREE Rod Drive Vent Fans when perfonning a natural circulation cooldown? 3

a. It prevents the local air temperature from exceeding the specifications for IRPI accuracy

b. It provides the major heat removal mechanism for the vessel head area.

c. It ensures the heat generated by the CRDM units is NOT added to the RCS.

d. It prevents the local air temperature from exceeding the specifications for NIS accuracy.

Ans: lb l Eram level: lS l CognitiveIevel: l Memory l Fugda==Ma= Direction is provided to start cooling fans. 'Ihree CRDM vent fans are staned for Reactor Vessel head heat ofAnswer removal. At least two CRDM vent fans are required for Reactor head cooling for prevention of void formatios.

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Page 163

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, Tier: l Emergency and Abnorrnal Plant Evolutions l RO Group: l3 l SRO Group: l3 l Systema / Evolution Nuneber: l 056 l Systesm/ Evolution 'Htle: l less of Off-Site Power l Category: AKI Knowledge of the operational implications of the following concepts as they apply to loss of Off-Site

Power

[ KA: { AKl.01 l Principle of cooling by natural convection RO Value: l3.7 l SRO Value: l4.2 l CFR: l 41.8 / 41.10 / 45.3

Reference Reference Nonnber Reference Stdion Pase Nusnber(s) Revision Learn. Obj NATURAL CIRCULATION 0300-000.00S- III.C.1 8 10/25/96 3

, ABRC04-00 NATURAL CIRCULATION S2.OP-AB.RC-0004(Q) 2.4 5 3 TECHNICAL BASIS DOCUMENT

EOP TRI' l,5,6; 0300-000.00S-TRP004- 3.3.3 13 2.A j NATURAL CIRCULATION 01 j COOLDOWN

Questica Source New Question Modification l Method j _W Source Conunents

l Material Required for E===d== tion:

Page 164

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Ques 6es Topic: l Loss of one vital bus Given the following conditions for Unit 2:

o A loss of 2A 115V Vital Instrument Bus has occurmd i o *Ihe mactor was manually tripped from 100% power due to decreasing Pressurizer pressum Which one of the following correctly describes the effect on the AFW system?

a. 21 AFW Pump is locked out by a low suction pressure trip signal.

b. 21 AF52, AFW Pump Backup Suction valve fails open to allemate supply piping.

c. 23AF21 and 24AF21, SG Inlet valves fail closed due to pmssum override interlock.

d. The 23 AFW Pump automatically starts when 2MS132, Turbine Start-Stop Valve, fails open.

A rs: lc l Exna Imel: lS { Cognidvelevel: l Comprehension l F_",

": Ioss of 115VAC, causes failure low for PA-3450,21 AFW Pump Disch Pitssure, causing 23AF21 and 24AF21 to of Answer ' fait closed due to Pressure Override Interlock. The low suction pressure trip is in effect only ifit is armed; only l when a tornado is likely.

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Page 165

I Tier: l Emergency and Abnormal Plant Evolutions l RO Group: l1 l SRO Group: l1 Systemn/ Evolution Number: l 057 l Systern/Evolu* ion

Title:

l less of Vital AC Instrument Bus Category: l AK2 l Knowledge of the interrelations between Loss of Vital AC Instrument Bus t vi the following:  ;

KA: l AK2.01 l Valves RO Value: l1.9 l SRO Value: l *2.1 l_ CFR: l 41.7 /45.7 Reference Reference Number Reference Section Pate Numeer(s) Revision Iaarn. Obj )

LOSS OF 2A. 2B,2C AND 030M)00.00S-ABil51- III.C.10 10 8/22S 7 4 1

2D 1ISV VITAL 00 TNSTRUMENT BUS AUXILIARY FEEDWATER 0300-00J.00S. IV.A.7.b; IV.C.4.f.3).b) 20;35 5/15 S 7 6.c; 8 l SYSTEM AFWCJ0-01 1 LOSS OF 2A i15V VITAL S2.OP-AB.115- 3.10; Attachment 2,1.0 4;I  !

INSTRUMENT BUS 0001(Q)

Question Source New Question Modification Method  ;

Question Source Comments: l Material Required for Examination:

NOTE: KA <2.5 justified by relationship to 061 A3.01, Ability to monitor automatic operation of the AFW, including: AFW startup and flows -4.1/4.2. j l

i Page 166

Quanden Topict l Effect of loss of DC to 4KV bus Given the following conditions for Unit 2:

o Reactor power- 100%

o 2C Vital 4KV Bus is aligned to 24SPT (breaker 24CSD closed) e Poweris lost to 2C Vital 125 VDC Bus o Prior to restoring power to the 2C DC Bus, 24 SFT is deenergized' Which one of the following correctly describes the status of 2C 4KV Vital Bus for these conditions?

a. Powered from the 2C EDG.

b. Energized from 23SPT(23CSD).

c. De nergized with all in-feed breakers tripped.

d. Deenergized with in-feed breaker 24CSD closed.

Ans: Id l Eun= Level: IS l Cognitive Level: I Comprehension l Explanation If power is lost from 24 Sirr before control power is restored,2C Vital Bus is lost (the SI'T infeed will NOT open, ofAnswer the alternate power infeed breaker and the 2C EDG output breaker cannot close) and the DC trip circuit is deenergized. Also,2C EDG may not be available because ofloss of DC to exciter flashing circuit.

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Tier: l Emergency and Abnormal Plant Evolutions l RO Group: l2 l SRO Group: l2 Systema / Evolution Nameer: l058 l Systenn/ Evolution

Title:

l Imss of DC Power Cateaory: l AA2 l Ability to determine and interpret the following as they apply to Loss of DC Power:

KAt l AA2.03 l DC loads lost: impact on to operate and monitor plant systems RO Value: l3.5 l SRO Value: l3.9 lCFR: l 43.5 / 45.13 Refenece Reference Nuanber Reference Section Pate Number (s) Revision Learn. Obj 4160 ELECTRICAL 0300-000.00S- V.B.4.a.1).b) & 2); 42;76 6/4/96 6, 8.d SYS"IEM 4KVACO-00 IX.C.4.d DC ELECTRICAL 0300-000.00S- IV.B.2.1.5); V.B.2.a.4) 20;28 8/21/96 3.b; 12 SYSTEMS DCELEC-00 OVERHEAD S2.OP-AR21e0002(Q) window 18 - NOTE 34 ANNUNCIATORS WINDOW B Question Source New Question Modification Method Question Source Comments: l Macedal Required for F.nmanation:

l Page 168

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Questies Tepic: l WGDT relief path Which one of the following correctly identifies the radiation monitor first affected if the inservice Waste Gas -

Decay Tank (WGDT) relief valve lifts?  ;

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a. 2R41, Plant Vent Noble Gas.

b. 2R34, South Pipe Penetration.

c. 2R42, Waste Gas Decay Tanks.

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d. 2R5, Fuel Handling Building - SFP.

Ams: Ia l Exams lavel: lS I th.nitive Level: l Memory l Explanatin High activity is sensed in the Punt Vent during a release by noble gas monitor R41C. The line from WGDT ties ofAnswer into the Fuel Hamiling Building exhaust line (downstream of any rad monitors that would sense release). R42 is an j area monitor in the vicinity of the WGDTs sensing gamma radiation.

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Page 169

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'Ilers l Emergency and Abnormal Plant Evolutions l RO Group: l2 l SRO Group: l2

$denWEvolution Nusnber: !060 l System / Evolution

Title:

l Accidental Gaseous Radwaste Release Cateaory: l AK2 l Knowledge of the interrelations between Accidental Gaseous Radwaste Release and the following:

KAt l AK2.01 l ARM system, including the normal radiation-level indications and the operability status T"*4 Value: l2.6 l SRO Value: l 2.9* l CFR: l 41.7 /45.7 Reference Reference Nun 6er Reference Section Page Number (s) Revision learn. Obj )

RADIOACITVE WAS111 0300-000.00S- IV.B.2.f.6); IV.B.3.g 24;25 9/25/96 3.b.ii,4.g I GAS SYSTEM WASGAS-00 RADIATION 0300-000f>0S- VI.A.I 64 10/1/96 4.u MONITORING SYSTEM RMS000-00 Auxiliary Building - 205337 G-1I sh. I Ventilation Question Source New Question Modification Method Question Source Comments: l Material Required for Examination:

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Page 170

E 1 Quaseine Tople: I SW Tech Spec i

Given the following conditions for Unit 2:

i o 23 SW Pump has been tagged out for the past 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for motor l

bearing replacement  ;

o A Nuclear Equipment Operator (NEO) on rounds reported flooding in SW Bay 4  ;

I Cy direction of S2.OP-AB.SW-0003(Q) " Service Water Bay leak", the control ro~:n crew started 22 SW l' Pump and stopped the 24 SW Pump. Additionahy contml power to the 24,25 and 26 SW Pumps was h.s4 zed. 'Ihe leak is now isolated on the SW discharge header in SW Bay 4.

Which one of the following cortectly describes the required Technical Specification action (s) for this situation?

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a. 21SW23 and 22SW23 Nuclear Hdr Cross-Over MOVs must be opened within the next hour in order to maintain both headers operable. l

b. The action statemcat entered when the 23 SW Pump was taken OOS still applies. Two SW loops ,

must be made operable within the next 66 hours7.638889e-4 days <br />0.0183 hours <br />1.09127e-4 weeks <br />2.5113e-5 months <br />. I

c. Tech. Spec. 3.0.3 applied when control power was removed from 24,25 and 26 SW Pumps. A unit l shutdown must be initiated within the next hour.

d. The action statement for less than two operable SW loops should be entered when control power is .

removed from the 24,25 and 26 SW Pumps. Two SW loops must be made operable within the next 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Amst lc l Exam Level: IS l Cosastive Level: 1 Comprehension l Espinnation When a Service Water Bay is removed from service, and the Service Water Pump fed from *B" bus in the ofAnswer OPERABLE Service Water Bay is unavailable (23 SWP), then LCO 3.0.3 is applicable. This is due to the potential for a runout flow demand that would be placed on the two running Service Water pumps immediately following the automatic SEC alignment.

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i Page 171

Tiers l Emergency and Abnormal Plant Evolutions l RO Group: II l SRO Group: l1 Systems / Evolution Nussber: l 062 l System / Evolution

Title:

l Loss of Nuclear Service Water Cateaory: l2.1 l Conduct Of Operations KAt l 2.1.12 l Ability to apply technical specifications for a system.

RO Value: l2.9 l SRO Value: l4.0 l CFR: l 43.2 / 43.5 / 45.3 Reference Reference Number Reference Section Page Number (s) Revision Iaara. Ob]

SERVICE WATER BAY 0300-000.00S- II.B.1;III.C 9;11 10/28/96 2.A; 4 LEAK ABSWO3-00 SERVICEWATER 0300-000.00S- IV.B.3.m; VII.C; IX.B 23;39-40;54-56 1/16/97 5;10;13 SYSTEM -INTAKE BAYS SWBAYS-02 SERVICE WATER BAY S2.OP-AB.SW. NOTE at step 3.22; 3;4 LEAK 0003(Q) Tech Basis 2.4.1 Question Source New Question Modification Method Question Source Connments: l Matedal Required for Er==d== tion:

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Page 172 a______

W Topic: l SD outside CR with fire The Control Room has been evacuated due to a fire in the Unit 2 Control Room. ,

t Which one of the choices correctly completes the following statement regarding the reason a NEO starts and aligns the SBO Air compressor?

These actions assure air is available for...

a. opening the MS10 Atmospheric Relief valves to initiate RCS cooldown

b. operating one or more Pressurizer PORVs as needed to control RCS pressure

c. opening the AFI1 Steam Generator AFW Inlet valves to supply flow from TDAFW Pump

d. operating 2CV55 Charging Flow Control valve to establish seal injection flow to the RCPs Ans: Id l Exaa 14 vel: lS l Commitive level: l Comprehension l Explanation Seal Flow must be established to RCPs to eliminate the poseibility of a LOCA To accomplish this, air is necessary ofAnswer for the operation of the CV55 valve and AC pawer, CCW and SW are necessary for the charging pump. The SBO Air Compressc7 is started to supply air to CV55.

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l Page 173

Ber: l Emergency and AbnEnai Plant Evolutions l RO Groep: l1 l SRO Group: 11 4

Systesm/ Evolution Nunnber: l 067 l Systean/ Evolution Utle: I Plant Fire on Site Category: l AA2 l Ability to determine and interpret the following as they apply to Plant Fire on Site:

KAt l AA2.04 l The firei extent of potential operational damage to plant equipment j R4 Value: l3.1 l SRO Value: l4.3 l CFR: l 43.5 /45.13 Reference Reference Nunnber Reference Section Pase Nuniber(s) Revision Learn. Ob]

4 CONTROL ROOM 030(M)00.00S- III.B.4.a; III.D.8.c 11;20 9/15/97 2,3 EVACUATION DUE TO ABCR02-00 FIRE

, CONTROL ROOM S2.OP-AB.CR-0002(Q) Tech Design Basis, 18 8 EVACUATION DUETO ATTACHMENT 8 FIREIN CONTROL ROOM, REIAY ROOM, OR

, CEILING OFTHE 460/230V SWITCHGEAR ROOM Question Source New Question Modifkation Method Question Source Comments: l Material Required for Exandantion:

l Page 174

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Question Topic: l RCPs on Control Room evacuation A Control Room evacuation has been initiated due to a fire in the relay room. The actions of S2.OP-AB.CR-000(Q) " Control Room Evacuation Due to a Fue in the Control Room, Relay Room or Ceiling of the 460/230V Switchgear Room" are being performed.

Which one of the following describes operation of the Reactor Coolant Pumps under these conditions?

a. All RCP's are stopped prior to leaving the control room, as part of the immediate actions

b. RCP's are run in accordance with the EOP network, after 2-EOP-TRIP-1 is entered

c. All RCP's are stopped by operation of switches at the HSD Panel, when directed by AB.CR-0002

d. RCP's are stopped shortly after the control room is evacuated, when the group bus in-feed breakers are opened locally Ans ld l Exann Level: lS l Connitive level: l Memory l Explanation RCP's are stopped when NEO#3 opens the group bus in-feed breakers. Stopping the RCP's are not part of the ofAnswer immediate actions for the procedure. A caution in the procedure indicates that the EOP's are not to be implemented when this procedure is in effect. There are no controls for the RCP's at the HSD Panel.

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Page 175

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l Der: l Emergency and Abnormal Plant Evolutions l RO Group: lI l SRO Group: I1 SystemafEvolution Number: l 068 l System / Evolution'Iltle: l Control Room Evacuation Category: l AK2 l Knowledge of the interrelations between Control Room Evacuation and the following: l 1

KAt l AK2.01 l Auxiliary shutdown panellayout '

F O Vaine: l3.9 l SRO Value: l4.0 l CFR: l 41.7 / 45.7 Reference Reference Number Reference Section Page Numbeds) Revision 12arn. Obj CONTROL ROOM 03004)00.00S- 9/15/97 3 EVACUATION DUETO ABCR02-00 FIRE CONIROL ROOM S2.OP-AB.CR-0001(Q) ATTACHMENT 3, 8 7 EVACUATION step 25.0 Question Source NRC Exam Bank Question Modi 6 cation Method Question Source Comments: l Material Required for Exasmination:

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Page 176

Quesnes Topic: l Containment closure Unit 2 is in MODE 5 with preparations in progress to lower RCS level to the 100 ft elevation, with fuel in the reactor vessel.

Which one of the following correctly identifies a requirement that ensures Containment Closure can be accomplished prior to the onset of fuel damage, if a problem develops?

a. All containment isolation valves must be closed prior to opening the RCS

b. The containment hatch must be installed prior to commencing reduced inventory operations

c. All RCS openings / vent pattuacept the one on which maintenance is being performed, must be closed

d. In order to lower the decay heat load, reduced inventory operations are never permitted until at least 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> have elapsed since shutdown Amst Ib l Exame Level: IS l ComaitiveI4 vel: l Comprehension l Explomation Per several procedures, since the containment latch takes more than two hours to install, it must be installed prior

~o f Answer to reduced inventory operations. Since reduced inventory operations cannot commence until 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> have elapsed since shutdown, adequate time exists for AB. CONT-0001 to establish containment isolation in the event of a problem. Certain procedurally directed RCS vent paths must be available prior to redte~t inventory operations.

Page 177

1)er: - l Emergency and Abnormal Plant Evolutions l RO Group: l1 l SRO Group: l1 System /Evoluties Number: l 069 l SystenvEvolution

Title:

l Loss of Containment Integrity Catemory: l 2.4 l Emergency Procedures / Plan KA: 2.4.9 Knowledge of low power / shutdown implications in accident (e.g. LOCA rs loss of RHR mitigation strategies).

RO Value: l3.3 l SRO Value: l3.9 l CFR: l 41.10/43.5 /45.13 Refessace Refenace Number Reference Section Pase Nursber(s) Revision Learn. Obj CONTAINMENT 03004X)0.00S- III.D.2. NOTE I 10 7/15/97 2 CLOSURE / INTEGRITY ABCONT-00 CONTAINMENT S2.OP-AB. CONT- Tech Design Basis 2.4 5 4 CIDSURE/ INTEGRITY 0001(Q)

DRAINING THE S2.OP-SO.RC-0006(Q) 2.18.3.A 4 REACIOR COOLANT SYSTEM < 101 FT ELEVATION WITH FUEL IN THE VESSEL Question Source New Question Modification Method Question Source Comunents: l Matedal Required for E=-d== tion:

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Page 178

l cr':n T:g= l Actions for suspected high activity l Given the following conditions for Unit 2:

i o Reactorpower 60%

o RMS TRBL(OHA A-6) has alarmed e

Letdown Line Failed Fuel monitor showed rising indication, is presently in alarm, but is now stable e

S2.OP-AB.RC-0002(Q) "High Activity In Reactor Coolant System" has been entered o

Chemistry has initiated sampling of the RCS but results are NOT yet available Which one of the following correctly idemifies the action to be taken until Chemistry reports the sa results?

a. Initiate a power reduction to less than 50% power.

b. Request Reactor Engineering perform Core Flux Map.

c. Notify Primary NEOs that high activity in RCS is suspected.

d. Direct the Primary NEO to place the standby Mixed Bed Demin in service.

Ans: Ie l Eman: Level: }S l Coanitive Level: l Memory l Explanation ofAnswer Operators need to be aware of personnel notification requirements for increased radiation conditions.

Building areas are of special concern because CVCS piping goes throughout the Aux. Bldg. This, along w Radiation Protection surveys, are the only actions required until RCS sampling confirms high rad conditions.

i Page 179

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'Iler: l Emergency and Abnormal Plant Evolutions l RO Group: l1 l SRO Group: lI Systena/ Evolution Nusaber: l 076 l Systean/ Evolution

Title:

l High Reactor Coolant Activity C A . y: l2.1 l Conduct Of Operations

, KA: l 2.1.14 l Knowledge of system status criteria which require the notification of plant personnel.

RO Value: l2.5 l SRO Value: l3.3 l CFR: l 43.5 /45.12 Reference Reference Nunnber Reference Section Pase Nuneber(s) Revision Learn. Ob]

HIGH ACTIVITY IN 0300-000.00S- D.I .c 9 10/25/96 3 REACTOR COOLANT ABRC02-00 SYSTEM HIGH ACTIVITY IN S2.OP-AB.RC-0002(Q step 3.6 2 2 REACTOR COOLANT SYSTEM Question Source New Question Modification Method Question Source Comunents: l M:terial Required for Emannimation:

Page 180

(FA Topic: l SI Termination Criteria  !

Which one of the following identifies the minimum parameters that are always evaluated to determine that SI  ;

i may be safely terminated?

a. RCS subcooling and RCS inventory ,

b. RCS subcooling and SG level ,

c. RCS inventory and SG level

d. RCS subcooling, RCS inventory and SG level J

Auss la l Enm Level: lS l Cognitive level: l Memory l Fugd. tion RCS subcooling is used for all Si termination checks in the EOPs. RCS inventory is always evaluated before ofAnswer terminating safety injection. Either pressurizer level or reactor vessel level is checked.

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-Tier: l Emergency and Abnormal Plant Evolutions l RO Group: l1 l SRO Group: lI i SystenvEvolutiori Nuseber: l E02 l SystenvEvolution

Title:

l SI Termination Catemory: l EK3 l Knowledge of the reasons for the following responses as they apply to SI Termination: I KA: l EK3.2 l Normal, abnormal and emergency operating procedures associated with (SI Termination).

1 RO Value: l3.3 l SRO Value: l3.8 !CFR: l 41.5,41.10 / 45.6 / 45.13 Refeanece Reference Nunnber Reference Section Revision Pase Number (s) Imrn. Ob.i EOP Generi: Issues 0300-000.00S- 2.2.3 19-20 3/4/96 4,, 5 GENISS-00 EOI 4DCA-5, . LOSS OF 0300-000.00S- 3.3.18 24-25 4/15/96 6 i l EMERGENCY LOCA05-01 RECIRCULATION 4

EOP-FRTS-1 and 2, 0300-000.00S-FRTS00- 3.2.9 22;25 8.A Response To Pressurized 00 Therrnal Shock Conditions i

Question Sourre New Question Modification Method O.A Source Comuments: l

, Mcterial Required for

' 1 Evandmation:

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i Question Topic: l Effect ofincreasing subcooling i

A small break LOCA has occurred on Unit 2. The actions of 2-EOP-LOCA-2 " Post-LOCA Cooldown and Depressurization" are being performed. With ONE Charging Pump and ONE SI Pump running, Pzr level is stabilized at RCS pressure of 900 psig. With RCS subcooling at 9 'F, the operator tums on a set of backup heaters.  !

Which one of the following describes the result of turning on the backup heaters?

a. Break flow falls; SI flow rises '

b. Break flow rises; Pressurizer level falls

c. Break flow falls; Pressurizer level rises 1

d. Break flow and Pressurizer level remain constant Ans: lb l Run= Level: lS l Cognitive level: l Comprehension l l E-;"

^'-_ Raising RCS pressure via the Pu heaters raises RCS pressure; raising break flow and lowering makeup flow. )

of Answer l

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

Tier: l Emergen:[aird Abnormal Plant Evolutions l RO Group: l2 l SRO Group: l2 SystemvEvolution Numiber: l E03 l Systern/ Evolution

Title:

l LOCA Cooldown and Depressurization Category: EK3 Knoveledge of the reasons for the following responses as they apply to LOCA Cooldown and Depressurization:

KA: EK3.3 Manipulation of controls required to obtain desired operating results during abnormal, and emergency situations.

RO Value: l3.9 l SRO Value: l3.9 l CFR: l 41.5,41.10 / 45.6 / 45.13 Reference Reference Number Referrace Section Pate Number (s) Revision Learn. Obj  ;

EOP-LOCA-02, POST 0300-000.00S- 1.33,1.3.4;1.4.1 6-7 2/29/96 1,4 l LOCA COOLDOWN AND LOCA02-01 DEPRESSURIZATION j Question Source New Question Modification Method Question Source Comunents: l Material Required for i Evand== tion:

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W Topic: l Core c#;ng evaluation l l

Given the following conditions for Unit 2:

o A loss of main feedwater has occuned at 100% power operation  !

l o AFW flow was NOT established o Actions of 2-EOP-FRHS-1" Response To Loss Of Secondary Heat Sink" i i are being performed i l

e The crew delayed in initiating bleed and feed and the RCS reached saturated conditions I with pressure at 2350 psig before injection flow was established l'

Which one of the following correctly describes the minimum actions necessary to prevent core uncovery?

a. Ensure that both centrifugal charging pumps are operating  !

b. Establish feed flow to at least one SG

c. At least one centrifugal charging pump must be operating and either PRI or PR2 fully open j i

d. Both centrifugal charging pumps must be operating and PRI and PR2 fully open Ams: lb l Exam Level: lS l Cognitive Level: l Memory l Explanation Boiling begins when reactor coolanIreaches saturation temperature. RCS steam generation results in large ofAnswer volumetric increase and PZR PORVs may not be able to compensate for this. If manually initiated, SI would not be effective in achieving recovery, because SI system has limited injection capacity when RCS pressure is at/above j PZR PORV setpoint. Values given - PZR PORV: -50 to -100 LB/sec; Charging /SI pumps: +40 lb/sec (290 gpm)

(both trains), so RCS will lose inventory and core will be uncovered. 'Ihe only possible means for preventing core uncovery once the transient enters this phase is to restore feedwater to the SG.

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Tier: l Emergency and Abnormal Plant Evolutions l RO Group: l2 l SRO Group: l2 Systenn/ Evolution Nunnber: l EOS l System / Evolution

Title:

l Loss of Secondary Heat Sink Category: EKI Knowledge of the operational implications of the following concepts as they apply to Loss of Secondary Heat Sink:

KA: l EKl.2 l Normal, abnormal and emergency operating procedures associated with (Loss of Secondary Heat Sink).

l RO Value: l3.9 l SRO Value: l4.5 l CFR: l 41.8 / 41.10 / 45.3 Reference Reference Number Reference Section Pase Namiber(s) Revision f earn. Obj EOP-FRHS-1,2, 3,4, and 5 0300-000.00S- 1.2.9; 2.7.6 9;13 4/10/97 2,4 HEATSINK FUNCTIONAL FRHS00-02 RESTORATION Question Source New Question Modification Method Question Source Conunents: l Material Required for Exandmation:

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Quesdes Tesde: l RCP ops during inadequate core cooling 2

Given the following conditions for Unit 2:

i o Actions of 2-EOP-FRCC-2 " Response To Degraded Core Cooling"

" l are being performed I o The SI accumulators are being isolated in preparation for depressurizing the SGs from 130 psig to atmospheric pressure o' RHR injection flow has been noted on cold leg injection e _During the crew brief prior to SG depressurization to atmospheric pressure, the RO asks if the running RCPs should be stopped Which one of the following correctly describes the required action and basis, relative to the RCP's?

a. RCPs should be stopped because the drop in #1 seal delta-P during depressurization may cause pump damage

b. RCPs should be stopped to avoid mechanical damage that can occur due to cavitation when the RCS reaches ssturation conditions

c. RCPs should NOT be stopped Wam flow through the core must be maintained when inadequate core cooling exists

d. RCPs should NOT be stopped because stagnating flow during depressurization could cause a bubble to form in the tractor vessel head Amst Ia l Eumm Level: lS- l Cosmidve Level: l Comprehension l Fwpaamaela= in preparation for the subsequent depressurization of the SGs to atmospheric pressure, the RCPs are stopped due to ofAnswer the anticipated loss of No. I seal operating requirements. Continued operation may result in damage to the RCPs.

RHR flow indicates RCS piessure has dropped to point of concern.

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Tier: l Emergency and Abnormal Plant Evolutions l RO Group: l1 l SRO Group: l1 l

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SystenWEvolution Number: l E06 l Systesn/ Evolution

Title:

l Degraded Core Cooling Category: l EA1 l Ability to operate and / or monitor the following as they apply to Degraded Core Cooling:

KAt l EAl.2 l Operating behavior characteristics of the facility.

R3 Value: l3.5 l SRO Value: l3.8 l CFR: l 41.7 / 45.5 / 45.6 Reference Reference Number Reference Section Pase Number (s) Revision Imrn. Obj EOP-PRCC 1,2, and 3 0300-000.00S- Vll.B.21 81 5/10/96 3b ACCIDENT MITIGATION FRCCX)0-01 STRATEGY Response To Degraded Core 1-EOP-FRCC-2 step 21 sheet 2 20 Cooling i

Question Source NRC Exam Bank Question Modification i Method Question Source Comments: l M:terial Required for Exandmation i

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I Questies Tosde: l RCP operation with ITS l t

- Given the following conditions for Unit 2:

o ' A rupture has occuned in the steam line from 21 SG to 23 AFW Pump o ' 2-EOP-FRTS-1 " Response To hnminent Pressurized Thermal Shock Conditions" has been entered due to a PURPI.E path condition o SIhas actuated and is reset o All RCPs are stopped o ECCS flow CANNOT be terminated 4 o Conditions required to support an RCP start are met.

s Which one of the following correctly describes operator actions relative to staning a RCP and the basis for the decision? i 1

a. 23 RCP should be started to establish PZR pressure control j

b. 23 RCP should be started to provide mixing of the ECCS injection flow

c. A RCP should NOT be started because the pressure surge will aggravate the PTS condition

d. A RCP should NOT be started in order to allow natural circulation to slowly remove thermal gradients Amst lb l Exam Level: lS I Cognitive I2 vel: l Comprehension l W ," " ' - -

An analysis of the effect of an RCP restart has been made to ensure the safety of this action relative to vessel  !

ofAnswer integrity, in order to mix the cold incoming SI water and the warm reactor coolant water, and thereby decrease the  !

likelihood of a PTS condition, an RCP restart is attempted. I i

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Tier: l Emergency and Abnormal _ Plant Evolutions l RO Group: ll l SRO Group: l1 Systan/ Evolution Number: l E08 l SystenvEvolution

Title:

l Pressurized Thermal Shock Category: l EK2 l Knowledge of the interrelations between Pressurized Thermal Shock and the following:

KA: EK2.2 Facilityi heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility. I l

RO Value: l3.6 l SRO Value: l4.0 lCFR: l 41.7 / 45.7  ;

l Referwece Reference Number Reference Section Page Number (s) Revision Learn. Obj '

EOP-FRTS-1 ar.d 2, 0300-000.00S-FRTS00- 3.2.9.4; 3.2.9.5 23-24 3/5/96 3.A; 8.A l Response To Pressurized 00 .

Thermal Shock Conditions Response To imminent 1-EOP-FRTS-1 step 9.4 stmet 2 20 Pressurized Tiermal Shock Condition Question Source New Question Modification Method Question Source Comments: l Material Required for Exandmation:

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Quesden Topic: l PTS operations l Given the following conditions:  !

o A main steam line break occuned inside containment.

e The actions of 2-EOP-FRTS-1, " Response To Imminent Pressurized Thennal Shock Conditions" are being perfonned  ;

e A sequired one hour soak hasjust been initiated l o RCP's are not running j

Which one of the following correctly identifies an evolution that can be performed during the soak?  !

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a. Place PZR auxiliary sprayin service

b. Lower the setpoints of the MS10's by 25 psi

c. Raise SG waterlevels fmm 25% to 50% 1

d. Energize all Pressurizer backup heaters l Ams: la l Examslevel: lS l Considve Level: l Coinprehension l Explomados During the L..w.4 e soak, actions of procedures shall neither: 1) Cool down RCS, NOR 2) Raise RCS pressure.

ofAnswer Only placing Aux Spray in service qualifies since this would act only to lower pressure.

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I Tiers l Emergency and Abnormal Plant Evolutions l RO Group: lI I SRO Group: l1 Wyolution Nusnber: l E08 l Systeni/ Evolution

Title:

{ Pressurized Thermal Shock Catesory: l EK3 l Kno'<.icdge of the reasons for the following responses as they apply to Pressurized Thermal Shock:

KA: EK3.3 Manipulation of controls required to obtain desired operating results during abnormal, and emergency situations.

RO Value: l3.7 l SRO Value: l3.8 l CFR: l 41.5,41.10 / 45.6 / 45.13 Reference Reference Numiber Reference Section Page Number (s) Revision Imrn. Obj EOP-FRTS-1 and 2, 0300-000.00S-FRTS00- 3.2.32 39 3/5/96 7 Response To Pressurized 00 i

Thermal Shock Conditions Response To Imminent 1-EOP-FRTS-1 step 29 4 20 Pressurized Thermal Shock Conditions i

Question Sourre NRC Exam Bank Question Modification Method Question Source Comuments: l Matedal Required for Exandantion:

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W Topic: I Eval of Natural Circ Given the following conditions for Unit 2:

o Actions of 2-EOP-SGTR-1 " Steam Generator Tube Rupture" are being performed due to a rupture in 24 SG o RCS pressure- 1120 psig ,

o Intact Steam Generator pressures - 610 psig o Ruptured SG Pressure - 925 psig o All RCPs are stopped Which one of the following correctly indicates the existence of natural circulation flow?

a. RCS cold leg temperature 490 *F and stable.

b. RCS hot leg temperature 500 *F and increasing.

c. RCS cold leg temperature 535 *F and decreasing.

d. Core Exit Thermocouple temperature 520 *F and increasing.

Ans: la l Exam Level: IS l Coanitive Level: l Application l E=; ' - tion 'Ihe evaluation of natural circulation is determined by CET and/or Hot leg temperatures stable or decreasing and ofAnswer RCS cold leg temperatures at Tsat for intact SG pressures. At 610 psig, Tsat = 490.5 F Page 193

Tier: l Emergency and Abnormal Plant Evolutions l RO Group: l1 l SRO Group: l1 Systeam/ Evolution Nunnber: l E09 l System / Evolution

Title:

l Natural Circulation Operations Catesory: l EAl l Ability to operate and / or monitor the following as they apply to Natural Circulation Operations:

KA: l EAl.3 l Desired operating results during abnormal and emergency situations.

RO Value: l3.5 l SRO Value: l3.8 l CFR: l 41.7 /45.5 /45.6 Refenece Reference Number Reference Section Page Number (s) Revision Learn. Obj EOP GenericIssues 0300-000.00S- 2.4.9 31-33 3/4/96 9 GENISS-00 EOP-SGTR-1, STEAM 0300-000.00S- 4.3.50 105-106 4/4/97 6 GENERATOR TUBE SGTR01-01 RUPIURE Steam Generator Tube I-EOP-Suix-1 step 50 sheet 7 Rupture Question Source New Question Modification Method Question Source Comments: l Material Requind for Steam Tables Exammination:

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(>==d== Topic l Reducing voids in Nat Cire Given the following conditions for Unit 2:

i o A natural circulation cooldown is in progress in accordance with o 2-EOP-TRIP-6 " Natural Circulation Rapid Cooldown With RVLIS" ,

o RCS pressure 1000 psig o Pressurizerlevel- 92% ~

o RVLIS Full Range-68%

Which one of the following correctly describes the method to be used to collapse the apparent void? 1

a. Maintain RCS pressure stable and then initiate venting of the reactor vessel head.

b. Raise RCS pressure using Pressurizer heaters and then lower charging flow.

c. Start a Charging pump to raise RCS pressure while maintaining RCS temperature constant.

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d. IAwer RCS temperature by raising the steaming rate from the SGs and maintain RCS pressure constant.

Ams- lb l Rumm Level: lS l Cognitive Level: l Comprehension l Explomation If the PZR level increases to greater than 90% during the depressurization, the PZR heaters are energized to ofAnswer maintain pressure stable. Adjust charging and letdown to reduce system inventory (preferred method). Step 10.1 calls for raising RCS pressure until RVLIS full range >74% to limit the void growth to the top of the hot legs, minimizing the potential for introducing voids into the SO U-tubes and affecting natural cire. At no time in this procedure is it appropriate to make a transition to EOP-FRCI-3 RESPONSE TO VOID IN REACTOR VESSEL, 1 and perform a head venting procedure. Since the intent is to allow a void to exist under controlled conditions, l venting the head would not climinate the void and just result in a loss ofinventory due to water flashing to steam as pressure Odiv,ssed.

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1 Mers l Emergency and Abnormal Plant Evolutions l RO Group: l1 l SRO Group: l1 SystessfEvolution Number: E10 SystenvEvolution

Title:

Natural Circulation with Steam Void in Vessel

! with/without RVLIS Category: EK1 Knowledge of the operational implications of the following concepts as they apply to Natural Circulation with Steam Void in Vessel with/without RVLIS:

KA: EKl.2 Normal, abnormal and emergency operating procedures associated with (Natural Circulation with Steam Void in Vessel with/without RVLIS).

E3 Value: l3.4 l SRO Value: l3.6 l CFR: l 41.8 / 41.10 / 45.3 Reference Reference Number Reference Section Pase Number (s) Revision Learn. Obj EOP-TRIP- 4,5,6; 03M-000.00S-TRP004- 6.0;7.3.9;7.3.10.1 64-65;74-75 5/17/96 6.B; 8 NATURAL CIRCULATION 01 COOLDOWN NARJRAL CIRCULATION 1-EOP-TRIP-6 step 9 & 10 (10.1) sheet 1 1 RAPID COOLDOWN WITH RVLIS Question Source New Question Modification Method Question Source Comments: l Material Required for Fxandantion:

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i Quasesem Topic: l Eval of Decay Heat removal l l

Given the following conditions for Unit 1:  ;

o A teactor trip and SIoccuned at 0700 o RHR system problems resulted in a loss of recirculation capability o Current time is 1350 hours0.0156 days <br />0.375 hours <br />0.00223 weeks <br />5.13675e-4 months <br /> ,

o RCS subcooling - 10 *F o 11 and 12 Charging Pumps are mnning  !

o BIT flow - 350 gpm I o 11 SI Pump flow -90 gpm ,

o 12 SI Pump flow - 100 gpm. j i

Which one of the following identifies the ECCS pumps that should remain running following determination '

cf minimum SI flow for decay heat removal? (Assume equal flow from each Charging Pump.) )

a. BOTH SI Pumps.

b. BOTH Charging Pumps.

c. I1 Charging Pump and 11 SI Pump.

d. 12 Charging Pump and 12 Si himp.

Amst le l Exam level: lS l Coanitive level: l Application l  ;

W,"--^' Based on 1-EOP-LOCA-5, at 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 50 min, following the trip (410 minutes), the required ECCS flow is approx. l ofAnswer 210 gpm. With Chg flow per pump of 175 gpm (350/2), the additional 90 gpm flow from 12 SI pump repree..cs the ECCS flow situation as close as possible but remaining above minimum (265 gpm) reauirement.

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Tier: l Emergency and Abnormal Plant Evolutions l RO Group: l2 l SRO Group: l2 SihWolution Nuneer: l ElI l Systesn/ Evolution

Title:

l Loss of Emergency Coolant Recirculation Ca". ;i: l EK2 l Knowledge of the interrelations between loss of Emergency Coolant Recirculation and the following:

KA: EK2.2 Facility 1s heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the opere. tion of the facility.

I'O Value: l3.9 l SRO Value: l4.3 l CFR: l 41.7 / 45.7 Reference Reference Number Reference Section Page Number (s) Revision Learn. Obj EOP-LOCA 5. - LOSS OF 0300-000.00S- 3.3.18.2 26 4/15/96 2, 7.A EMERGENCY LOCA05-01 RECIRCULATION I LOSS OF EMERGENCY l EOP-LOCA-5 Figure A (step I8.2) sheet 2 20 RECIRCULATION Question Source New Question Modification Method 1 Question Source Conunents: l 1

Matedal Required for 1-EOP-LOCA-5 Figure A Minimum ECCS Flow Versus Time After Trip I Exandmation:

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& Topic: 50 overfill Given the following conditions for Unit 2: i o Reactor trip a7.1 SI occuned from 100%

o RCPs are stopped c MSIVs have closed o ADVERSE containment conditions exirts o 32 SG pressure is 1130 psig o 22 SG narrow ran3e level peaked at 93% but has since fallen to S5%

o The CRS has elected to implement 2-EOP-FRHS-2 " Response To Steam Generator Overpressure" I o The CRS is considering opening 22MSI8 (Warmup Valve) to relieve SG pressure Which one of the following describes the correct action relative to steam ielease from 22 SG7

a. Steam release should NOT occur since natural circulation flow in other loops may be dismpted.

b. Steam may be released without restriction since narrow range level has been adequately established.

c. Steam may be released at less than 50 psi per hour to remain within SG tube difterential pressure limits.

d. Steam release should be evaluated by the TSC since damage may occur due to water trapped in the steamline. '

Amst ld l Exam Level: lS l Cognitive Level: l Memory l Explanation If the affected SG te el has reached the upper tap, then the SG may be Pled to the steamline. Decreasing the ofAnswer affected SG level into the narrow rage does not ense.e that water is reno ved from the affected steamline.

Steamline conditions should be evaluated prior to relensing steam from any SG with level above 92% (90%

Adverse) to prevent potential damage to downstream components.

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Tier: l Emergency and Abnormal Plant Evolutions l RO Group: l3 l SRO Group: l3 Systesn/ Evolution Numsber: l E13 l Systein/ Evolution

Title:

l Steam Generator Overpressure ,

i Catesory: l EK3 l Knowledge of the reasons for the following responses as they apply to Steam Generator Overtryssure:

KA: EK3.3 Manipulation of controls required to obtain desired operating results during abnormal, and emergency

, situations.

R3D Value: l3.2 l SRO Value: l3.4 l CFR: l 41.5,41.10 / 45.6 / 45.13 Reference Reference Nunnber Reference Section Pase Number (s) ReWien learn. Obj EOP-FRHS-1,2,3,4, and 5 0300-000.005- 8.2.4 85-86 4/10/97 6,10 HEATSINK FUNCTIONAL FRHS00-02  ;

RESTORATION l EOP-PRHS-1,2, 3,4, and 5 0300-000.00S- 11.2.1 93-94 4/10/97 HEAT SINK FUNCTIONAL FRHS00-02 RESTORATION Response To Steam 1-EOP-FRSH-2 step 5 sheet 1 Generator Overpressure Question Source New Question Modification Method Question Source Comments: l l Mate ial Required for Exandmation:

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