ML081370256

PUMP MIN FLOW

VALVE.

d.

IF ..... 3-FCV-71-34,

PUMP MIN FLOW VALVE,

CANNOT be opened from control room,

THEN ... DISPATCH personnel to 250V

DC RMOV Board 3B,

Compartment 50, to perform the following:

1)

PLACE 3-XS-071-0034,

RCTC

PUMP MIN FLOW

VALVE EMER TRANS SWITCH, to EMERG.

2)

OPEN 3-FCV-71-34,

PUMP MIN FLOW

VALVE.

7 .

WHEN

Suppression Pool level reaches -5.5 in.,

THEN

VERIFY CLOSED the following valves:

3-FCV-73-30,

HPCI

PUMP MIN FLOW VALVE

3-FCV-71-34,

PUMP MIN FLOW VALVE.

8.

DISPATCH personnel to 250V DC RMOV Board 3B,

Compartment 50, to VERIFY 3-XS-071-0034,

Intermediate Range Monitors

PUMP MIN

FLOW VALVE EMER TRANS SWITCH, in NORMAL.

55. RO 295031G2.4.6 OOl/C/A/TlGl/Cl//295031G2.4.6//RO/SRO/NO

Given the following plant conditions:

Unit 2 was operating at 98% power when an automatic scram occurred due to a Group I

isolation.

All control rods fully insert as reactor water level immediately drops below Level 2.

The Recirc pumps trip.

HPCI automatically initiates but immediately isolates due to a blown inner turbine exhaust

rupture diaphragm.

RCIC had been tagged out of service previously to repair an oil leak.

All other systems are operable.

EOI-1, RPV Control, is entered.

Pressure control was established with SRVs.

The remaining high pressure injection systems are unable to maintain reactor water level which is

currently at -150 inches and lowering.

Which ONE of the following contingency procedures would be appropriate to execute?

A. to!

C1, Alternate Level Control

B.

C2, Emergency RPV Depressurization

C.

C4, RPV Flooding

D.

C5, Level/Power Control

KIA Statement:

295031 Reactor Low Water Level / 2

2.4.6 - Emergency Procedures / Plan Knowledge symptom based EOP mitigation strategies

KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific

plant conditions to determine the appropriate Emergency Procedure used to mitigate a low reactor water

level condition.

References:

2-EOI-1, EOIPM Sections O-V-C and O-V-G

Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble,

sort, and integrate the parts of the question to predict an outcome. This requires mentally using this

knowledge and its meaning to predict the correct outcome.

0610 NRC Exam

(

REFERENCE PROVIDED: None

Plausibility Analysis:

In order to answer this question correctly the candidate must determine the following:

1. Whether the given conditions are indicative of a 1055 of HP injection.

2. Based on Item #1 above , which EOI Contingency is appropriate to mitigate that condition.

A is correct.

B is incorrect. This is plausible since ED will eventually become necessary following the initial actions of

EOI-C1. However, additional actions are required before EOI-C2 is appropriate.

C is incorrect. This is plausible since OW temperature may be high enough following ED to create a

condition where RPV level instruments become unavailable. However, additional actions are required

before EOI-C4 is appropriate.

D is incorrect. This is plausible since the only given condition which contradicts the use of EOI-C5 is the

current rod pattern. However, with all rods inserted, EOI-C5 is not appropriate.

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56. RO 295037EK2.l 1 OOl/C/A/TlGI/RMCS//295037EK2.Il//RO/SROINO

A hydraulic ATWS has occurred on Unit 2 and the Unit Operator is inserting control rods in accordance

with the EOI appendices 10, 1F, &2.

(

With these plant conditions...

A. >I all insert blocks are bypassed.

B.

rod drift indication is received as soon as rod motion begins.

C.

stabilizing valves open to provide increased drive pressure.

D.

all RMCS timer functions are bypassed except for the settle bus.

KIA Statement:

295037 SCRAM Condition Present and Power Above APRM Downscale or Unknown 11

EK2.11 - Knowledge of the interrelations between SCRAM CONDITION PRESENT AND REACTOR

POWER ABOVE APRM DOWNSCALE OR UNKNOWN and the following : RMCS : Plant-Specific

KIA Justification: This question satisfies the KIA statement by requiring the cand idate to use specific

plant conditions to determine the status of the RMCS while executing procedures to mitigate an ATWS

condition .

References: 2-EOI Appendicies 1D, 1F, and 2

Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble ,

sort, and integrate the parts of the question to predict an outcome. This requires mentally using this

knowledge and its mean ing to predict the correct outcome.

0610 NRC Exam

REFERENCE PROVIDED: None

Plausibility Analysis:

In order to answer this question correctly the candidate must determine the following:

1. What affect the actions performed by EOI Append ix implementation have on the RMCS system.

2. What affect the RMCS manipulations required by implementation of the EOI Appendicis have on plant

indications.

A is correct.

8 is incorrect. This is plausible since a Rod Drift indication will occur for each inserted control rod,

however the indication does not occur until the rod is fully inserted and the CRD NOTCH OVERRIDE

switch is released.

C is incorrect. This is plausible because CRD stabilizing valves DO have an effect on drive water

pressure, but the efffect is to prevent oscilations while moving control rods, NOT increase pressure.

(

D is incorrect. This is plausible since RMCS timers are bypassed by using the CRD NOTCH

OVERRIDE switch in accordance with EOI Appendix 1D. However, the Settle Bus timer is also bypassed.

(

2-EOI APPENDIX-1D

Rev.

6

Page 1 of 3

...

2-EOI APPENDIX-1D

INSERT CONTROL RODS USING REACTOR MANUAL CONTROL

SYSTEM

LOCATION:

Unit 2 Control Room,

Panel 9-5

ATTACHMENTS:

1. Tools and Equipment

2. Core Position Map

NOTE:

This EOI Appendix may be executed concurrently with

EOI Appendix 1A or IB at SRO's discretion when time

and manpower permit.

1.

VERIFY at least one CRD pump in service.

NOTE:

Closing 2-85-586,

CHARGING WATER ISOL, valve may

reduce the effectiveness of EOI Appendix 1A or lB.

2.

IF

Reactor Scram or ARI

CANNOT be reset,

THEN

DISPATCH personnel to close 2-SHV-85-586,

CHARGING WATER SHUTOFF (RB NE, El 565 ft).

3.

VERIFY REACTOR MODE SWITCH in SHUTDOWN.

4.

BYPASS Rod Worth Mi ni mi z e r .

5.

REFER TO Attachment 2 and INSERT control rods in the

area of highest power as follows:

a.

SELECT control rod.

b.

PLACE CRD NOTCH OVERRIDE switch in EMERG ROD IN

position UNTIL control rod is NOT moving inward.

c.

REPEAT Steps 5.a and 5.b for each control rod to be

inserted.

NOTE:

A ladder may be required to perform the following

step.

REFER TO Tools and Equipment, Attachment 1.

IF necessary,

an alternate ladder is available at,

the HCU Modules,

EAST and West banks. It is stored

by the CRD Charging Cart.

6.

WHEN ... NO further control rod movement is possible or

desired,

THEN ... DISPATCH personnel to verify open 2-SHV-85-586,

CHARGING WATER SHUTOFF (RB NE,

El 565 ft).

END OF TEXT

(

2-EOI APPENDIX-IF

Rev.

4

Page I of 7

...

2-EOI APPENDIX-1F

MANUAL SCRAM

LOCATION:

Unit 2 Control Room

ATTACHMENTS:

1. Tools and Equipment

2.

Panel 2-9-15,

Rear

3.

Panel 2-9-17,

Rear

1.

VERIFY Reactor Scram and ARI reset.

a.

IF ...*. ARI

CANNOT be reset,

THEN .*. EXECUTE EOI Appendix 2 concurrently with

Step 1.b of this procedure.

b.

IF

Reactor Scram CANNOT be reset,

THEN

DISPATCH personnel to Unit 2 Auxiliary

Instrument Room to defeat ALL RPS logic

trips as follows:

1)

REFER to Attachment 1 and OBTAIN four 3-ft banana

jack jumpers from EOI Equipment Storage Box.

2)

REFER to Attachment 2 and JUMPER the following

relay terminals in Panel 2-9-15,

Rear:

a)

Relay 5A-K10A

(DQ) Terminal 2 to Relay

5A-K12E

(ED) Terminal 4,

Bay 1.

b)

Relay 5A-K10C

(AT) Terminal 2 to Relay

5A-K12G

(BH) Terminal 4,

Bay 3.

3)

REFER to Attachment 3 and JUMPER the following

relay terminals in Panel 2-9-17,

Rear:

a)

Relay 5A-K10B

(DQ) Terminal 2 to Relay

5A-K12F

(ED) Terminal 4,

Bay 1.

b)

Relay 5A-K10D

(AT) Terminal 2 to Relay

5A-K12H

(BH) Terminal 4,

Bay 3.

2.

WHEN

RPS Logic has been defeated,

THEN

RESET Reactor Scram.

3.

VERIFY OPEN Scram Discharge Volume vent and drain valves.

(

57. RO 295038EKl.Ol OOlIMEMlTlGlINEWI1295038EKl.Ol//RO/SRO/RWM

Given the following plant conditions:

Unit 2 has experienced a LOCA with a loss of Primary Containment.

You have volunteered for a team dispatched from the OSC to enter the Reactor Building and

attempt to energize 20 480v RMOV board.

Due to environmental and radiological conditions present in the Reactor BUilding, Radcon

provides you with a Sodium Chloride and Potassium Iodine tablet during the prejob briefing.

Which ONE of the following describes the benefit of ingesting Potassium Iodine prior to the Reactor

Building entry?

A.

It will reduce the risk of dehydration and heat stress.

B. It will reduce the absorption of radioactive Iodine by the lungs.

C~ It will reduce the absorption of radioactive Iodine by the thyroid.

D.

It will reduce the absorption of radioactive Potassium in the blood stream.

KIA Statement:

295038 High Off-site Release Rate I 9

EK1.01 - Knowledge of the operational implications of the following concepts as they apply to HIGH

OFF-SITE RELEASE RATE : Biological effects of radioisotope ingestion

KIA Justification: This question satisfies the KIA statement by requiring the candidate to correctly

identify the pathway and adverse effect of iodine ingestion.

Reference: EPIP -14 Revision 18, page 4

Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall

or recognize discrete bits of information.

0610 NRC Exam

REFERENCE PROVIDED: None

Plausibility Analysis:

A is incorrect. The sodium chloride tablets would be used for this purpose. It is plausible if the

candidate is unsure of the purpose of KI tablets.

B is incorrect. Only the thyroid is the organ at risk, but it is plausible if the candidate assumes that

airborne ingestion is limited to absorption by the lungs.

C is correct.

D is incorrect. Iodine is the element that is absorbed. Potassium becomes a plausible answer due to

recent media coverage regarding health risks related to low potassium levels in the blood stream.

(

BROWNS FERRY

RADIOLOGICAL CONTROL PROCEDURES

EPIP-14

3.6

Issuing Potassium Iodide (KI)

3.6.1

If the TSC RP Manager has reason to believe that a person's projected

cumulative dose to the thyroid from inhalation of radioactive iodine might exceed

10 rems (see Appendix A), the exposed person should be started immediately on

a dose regimen of KI. This decision shall be immediately communicated to the

SED.

3.6.1.1

If the TSC is not staffed or the RP Manager position has not been filled,

then the senior onsite RP Supervisor has the authority to issue KI

utilizing the bases described in step 3.6.1.

3.6.1.2 The initial dose of KI should be not delayed since thyroid blockage

requires 30 to 60 minutes. Anyone authorized to initiate KI shall be

familiar with the Food and Drug Administration (FDA) patient package

insert and be sure that each recipient is similarly informed .

3.6.1.3 Prior to issuing KI to an individual , the person should be asked if he/she

is allergic to iodine. If the person indicates a possible sensitivity to iodine

they should not be issued KI.

3.6.2

KI is stored in the plant RP supply cage and the REP Van instrument kits.

3.6.3

RP normally will not dispense a container or package of KI to TVA Personnel

involved in activities to support a radiological emergency. RP will however

dispense a single individual dose of KI to team members dispatched from the

OSC.

3.6.4

Follow the dosage outlined on the FDA patient package insert (Appendix B). A

copy of the FDA approved patient package insert shall accompany the issuance

of KI. If KI is distributed in individual doses then verbal instructions of the

significant information on the patient package insert by a knowledgeable

individual is sufficient.

3.6.5

Complete the KI Issue Report (Appendix C) or document on an RWP time sheet

as appropriate for issuance of KI. If the RWP time sheet is used to document

distribution of the KI, note the time of KI distribution on the back of the time

sheet.

PAGE 4 OF 9

REVISION 0018

(

58. RO 600000AA1.08 00 l/MEM/Tl GlIRSWl1600000AA 1.081IRO/SRO/I1120107 RMS

Which ONE of the following describes the appropriate fire extinguishing agent for the specific class of

fire?

A.

Water used on Class "B" fires.

B."

Low pressure CO2 used on Class "C" fires.

C.

Dry Chemical (PKP) used on Class "c" fires.

D.

Aqueous Film Forming Foam (AFFF) used on Class "A" fires.

KIA Statement:

600000 Plant Fire On-site I 8

AA1.08 - Ability to operate and I or monitor the following as they apply to PLANT FIRE ON SITE: Fire

fighting equipment used on each class of fire

KIA Justification: This question satisfies the KIA statement by requiring the candidate to identify the

correct fire fighting agent for a specific class of fire.

References: TVA Safety Manual

Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall

or recognize discrete bits of information.

0610 NRC Exam

REFERENCE PROVIDED: None

Plausibility Analysis:

In order to answer this question correctly the candidate must determine the following :

1. Which flammable material is of concern based on Fire Class A, Band C.

2. Which extinguishing agent is appropriate for each class of fire.

3. Which extinguishing agent is inappropriate for a given class of fire.

A is incorrect. Class "B" fires are flammable liquids. Using water could cause serious damage by

allowing the liquid to splatter and spread.

B is correct.

C is incorrect. Dry chemical agents are extremely corrosive to electrical components and insulation

typical of Class "B" electrical fires .

D is incorrect. AFFF is designed as a flooding and diluting agent for Class "B" flammable liquid fires.

Application on a Class "A" fire is not effective in extinguishing flammable materials such as wood and

paper.

59. RO 295009AK2.01 OOl/C/A/TlG2/PR.INSTRl13/295009AK2.0l/9619/RO/SRO/ll/20/07 RMS

Given the following Unit 1 plant conditions:

Due to multiple high pressure injection system failures, 1-EOI-C1, Alternate Level Control has

been entered.

RHR Pump 1A is running and lined up for LPCI injection.

Core Spray Pumps 1Band 1D are running and lined up for injection.

Drywell Temperature is 240 OF and rising slowly.

Which ONE of the following conditions describes the appropriate point where Emergency

Depressurization may be performed in accordance with 1-EOI-C1, Alternate Level Control?

Post Accident Flooding Range level instrument 3-L1-3-S2 is reading __inches with reactor pressure at

___psig.

REFERENCES PROVIDED

KIA Statement:

29S009 Low Reactor Water Level I 2

AK2.01 - Knowledge of the interrelations between LOW REACTOR WATER LEVEL and the following:

Reactor water level indication

KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific

plant conditions to determine actual reactor water level under conditions of low reactor water level.

References:

1-EOI-C1 Flowchart, PIP-9S-64 Rev 12

Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble,

sort, and integrate the parts of the question to predict an outcome. This requires mentally using this

knowledge and its meaning to predict the correct outcome.

0610 NRC Exam

REFERENCE PROVIDED: 1-EOI-C1 Flowchart, PIP-95-64 Rev 12.

Plausibility Analysis:

(

In order to answer this question correctly the candidate must determine the following:

1. Recognize the requirement that RPV level must be less than -162 inches before Emergency

Depressurization is appropriate.

2. Recognize that the indicated RPV level must be corrected for pressure using PIP-95-64.

3. Recognize that two or more injection systems must be lined up with pumps running to meet the

requirement to Emergency Depressurize.

4. Recognize that only one RHR pump is required to qualify as an injection subsystem since each RHR

pump is rated for 100% capacity.

NOTE: Each distractor is plausible because the conditions specified are possible given the current plant

conditions.

A is correct.

B is incorrect. Level is 5 inches too high or pressure is 100 psig too high.

C is incorrect. Level is -4 inches too high or pressure is 240 psig too high.

D is incorrect. Level is -4 inches too high or pressure is 100 psig too high.

(

-150"

TAF

-162"

-175"

....J

W>

W

-200"

....J

0

W....-co

-225

0

Z

-250"

-268"

3-LI-3-52 & 62 CORRECTION CURVES

-

-162" =TAF (RED LINE)

-185" =MSCRWL (GREEN LINE)

-200" =MZIRWL (BLUE LINE)

-215" =TWO-THIRDS CORE HEIGHT (BLACK UNE)

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......

o

100

200

300

400

500

600

700

800

900 1000 1100

REACTOR PRESSURE (PSIG)

ACTUAL

LEVEL

-162"

-185"

-200"

-215"

PIP-95-64

REV. 12

OPL171.003

Revision 17

Page 22 of 54

INSTRUCTOR NOTES

Since no trips or alarms are

associated with this range, this

level signal is not directed

through the Analog Trip System.

(d)

One MCR indicator on Panel 9-3

monitors this range of level

indication .

(4)

Post-accident Flood Range

(a)

-268" to +32" range covering

active core area and overlapping

the lower portion of the Normal

Control Range .

(b)

Referenced to instrument zero

(c)

Intended for use only under

accident conditions with reactor

at 0 psig and recirculation pumps

tripped.

(d)

Variable leg tap is from diffuser of

jet pumps 1 and 6 (or 11 and 16).

(e)

Per Safety Analysis on water

level instruments the conclusion

Injecting with RHR

is that the accident range

L1-3-52 and 62

instruments adequately indicate

(Accident Range)

water level--provided they are

Technical Support

corrected for off-calibration

letter dated 9/13/95

conditions of RPV pressure

(See LP Folder)

utilizing the operator aid on Panel

Use Conservative

9-3 for level correction.

Decision Making

Obj. V.8.15.

(f)

An interlock associated with this

Obj. V.8 .11.

range will prevent using the RHR

System for containment de

pressurization when it is needed

to flood the core region .

(g)

The -68" to -168" portion of this

range is recorded in the MCR on

Unit 3 Recorder

2-L1-3-62 Recorder and two

displays a scale of

indicators monitor the full range

+32" to -268"

of these instruments.

E

MINATION

REFERENCE

(PROVIDED TO

CANDIDATE

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-162"

-175"

....Jw>w

-200"

....J

0

W

'-<<o

-225

0

Z

-250"

-268"

3-LI-3-52 & 62 CORRECTION CURVES

-

-162" =TAF (RED LINE )

-185" =MSCRWL (GREEN LINE)

-200" =MZIRWL (BLUE LINE)

-215" =TWO-THIRDS CORE HEIGHT (BLACK LINE)

,

~

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ACTUAL

LEVEL

-162"

-185"

-200"

-215"

I

o

100

200

300

400

500 600

700

800

900 1000 1100

REACTOR PRESSURE (PSIG)

PIP-95-64

REV. 12

60. RO 295012G2.2.22 00 l/C/A/Tl G2/64/12/2950 12G2.2.22/IRO/SRO/0606S NEW6/28/2007

Given the following plant conditions:

You are the oncoming Unit 3 Unit Supervisor.

During turnover the onshift Unit Supervisor informs you that 2 Drywell Coolers had been

secured during his shift while performing ground isolation on 3C 480v RMOV board.

Drywell Average Temperature is 152°F and stable.

Which ONE of the following describes the appropriate condition and required action?

A'! Exceeded 3-SR-2, Instrument Checks and Observations, Drywell temperature limit. Address Tech Spec section 3.6.

B. Exceeded the normal operating Drywell temperature limit. Drywell temperature must be logged

hourly until below the limit.

C. Exceeded the normal operating Drywell temperature limit. Restore Drywell average air temperature

below the limit in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

D.

Exceeded 3-EOI-2, Primary Containment Control entry condition. Enter and execute 3-EOI-2,

Primary Containment Control.

KIA Statement:

295012 High Drywell Temperature 15

2.2.22 - Equipment Control Knowledge of limiting conditions for operations and safety limits

KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific

plant conditions to determine that Technical Specification limits have been exceeded.

References:

Unit 3 Tech Specs Section 3.6.1.4

Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble,

sort, and integrate the parts of the question to predict an outcome. This requires mentally using this

knowledge and its meaning to predict the correct outcome.

0610 NRC Exam

REFERENCE PROVIDED: 1-EOI-C1 Flowchart, PIP-95-64 Rev 12.

Plausibility Analysis:

(

In order to answer this question correctly the candidate must determine the following:

1. The appropriate entry condition for U3 Tech Spec Section 3.6.1.4.

2. The appropriate entry condition for 3-EOI-2, Primary Containment Control.

3. The appropriate action based on the given condition.

A is correct.

B is incorrect. This is plausible because the Tech Spec Iimt was exceeded, however the required action

is to restore the Orywell Temperature within the limit in 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . There is no requirement for hourly logging

of OW temperature.

C is incorrect. This is plausible because the Tech Spec limt was exceeded, however the required action

is to restore the Orywell Temperature within the limit in 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months limit is based on performing

the surveillance on Orywell Temperature.

o is incorrect. This is plausible because the entry condition for 3-EOI-2 is only 8 of above the given

temperature. However, the entry condition has not been met and OW temperature was reported as

"stable".

Drywell Air Temperature

3.6.1.4

3.6

CONTAINMENT SYSTEMS

3.6.1.4

Drywell Air Temperature

LCO 3.6.1.4

Drywell average air temperature shall be s 150°F.

APPLICABILITY:

MODES 1, 2, and 3.

ACTIONS

CONDITION

REQUIRED ACTION

COMPLETION

TIME

A. Drywell average air

A.1

Restore drywell average

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

temperature not within

air temperature to within

limit.

B. Required Action and

B.1

Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

associated Completion

Time not met.

AND

B.2

Be in MODE 4.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months

BFN-UNIT 3

3.6-17

Amendment No. 212

(

SURVEILLANCE REQUIREMENTS

SURVEILLANCE

Drywell Air Temperature

3.6.1.4

FREQUENCY

(

SR 3.6.1.4.1

BFN-UNIT 3

Verify drywell average air temperature is

within limit.

3.6-18

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

Amendment No. 212

(

61. RO 295015AKl.02 OOlIMEMlTlG2IBASISI1295015AKl.02///lll21107 RMS

EOI-1 flowchart path RC/Q directs the operator to inhibit the ADS auto blowdown function once Standby

Liquid Control injection has begun.

Which ONE of the following describes why ADS is inhibited under these conditions?

A.

ADS actuation would impose a severe pressure and temperature transient on the reactor vessel.

B.

The operator can control pressure better than an automatic system like ADS.

C. otI Severe core damage from a large power excursion could result, if low pressure systems

automatically injected on depressurization.

D.

If only steam driven high pressure injection systems are available an ADS actuation could lead to a

loss of adequate core cooling.

KIA Statement:

295015 Incomplete SCRAM I 1

AK1.02 - Knowledge of the operational implications of the following concepts as they apply to

INCOMPLETE SCRAM : (CFR 41.8 to 41.10) Cooldown effects on reactor power

KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific

plant conditions to determine the effect of a significant cooldown when an incomplete scram has

occurred.

References:

Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall

or recognize discrete bits of information.

0610 NRC Exam

REFERENCE PROVIDED: None

Plausibility Analysis:

In order to answer this question correctly the candidate must determine the follow ing:

1. The basis for inhibiting ADS under the specific conditions of boron injection.

NOTE: Each of the three distractors are plausible based on their relationsh ip to the bases for inhibiting

ADS under circumstances OTHER than boron injection. Specifically, Alternate RPV Level Control

actions. Refer to the attached excerpt from EOIPM Section O-V-G.

A is incorrect. This applies whenever ADS actuates , but is only the precursor to the issue related to

boron injection.

B is incorrect. This statement is true, but is not addressed in the basis for boron injection.

C is correct.

D is incorrect. This statement applies particularly to a low RPV level condition.

I

(

EOI PROGRAM MANUAL

SECTION O-V-C

STEP:

RC/Q-14 and RC/Q-15

EOI-1, RPV CONTROL BASES

EJlECUlE RC/Q- , 2 "",0 RC/Q-22 CONCVRRENllY

Re/Q-'1

BORON INJ IS REQUIRED

(£01- I. RC/P-8)

L

Re/Q-13

'---

-r-e-

' L

Rc/Q-14

INHlBIl NlS

RC/Q-15

L-

---,

-.JL

VERIFY RWCU SYSTEI.I ISOLATION

L

RC/Q-1S

WHILE EXECUTING THE FOLLOWING STEPS:

1£

Ii:!lli

SLC tNlK LVl DROPS TO c....3>.

tRIP THE SLC PUMPS

RC/Q-I7

~~-__-

...-----... L

034

SECTION o-v-c

PAGE 116 OF 127

REVISION 1

EOI-1, RPV CONTROL BASES

EOI PROGRAM MANUAL

SECTiON o-v-c

RC/Q-14 and RC/Q-15

STEP

DISCUSSION:

1

1

The RC/Q-14 action step directs the operator to manually initiate the SLC System. Because this step is prioritized

with the miniature before decision step RClQ-12 symbol, this action should be performed before suppression pool

temperature reaches <A.64>, Boron Injection Initiation Temperature. EOI Appendix 3A provides step-by-step

guidance for manual initiation of the SLC System. Boron in solution absorbs neutrons, providing negative

reactivity to achieve reactor subcriticality, since the reactor is not yet subcritical on control rod insertion alone.

The RC/Q-15 action step directs the operator to defeat automatic ADS function by placing the ADS inhibit

switches in the inhibit position. Because this step is prioritized with the miniature before decision step RC/Q-12

symbol, this action should be performed before suppression pool temperature reaches.<A.64>, Boron Injection

Initiation Temperature.

ADS initiation may result in the injection of large amounts ofrelatively cold, unborated water from low pressure

injection systems. With the reactor still critical or subcritica1on boron, the positive reactivity addition due to boron

dilution and temperature reduction from injection of cold water may result in a reactor power excursion large

enough to cause substantial core damage. Defeating ADS is, therefore, appropriate whenever boron injection is

required. If emergency depressurization of the RPV is subsequently required, explicit direction is provided in the

appropriate EOL Therefore, the ability to maintain automatic initiation capability of ADS is not required.

( *

REVISION 1

PAGE 117 OF 127

SECTION O-V-C

EOI PROGRAM MANUAL

SECTION O-V-G

STEP: Cl-l

C1, ALTERNATE LEVEL CONTROL BASES

.

IfOl-l

RPv CbNTROL

1

RC/L-12

~~>::;~;F(./'~~;:* *~'CAut(O~- : *

,.

...

.

.,." fH::

~ #1

AMBiENT TEMP MAY AffECT RPv WATER LVl

t:

INDICATION AND TREND

i

. ," '

' ~

. ,....., ~ "..

.-. "

....'.

"; '

.' ~' .

~

L

I

INHIBIT ADS

I L

Cl **1

WHILE EXECUTING THIS PROCEDURE:

lE

.I1::f.£.N

-.. ' .-

-

ALL CONTROL RODS ARE l\\lOI

ExIT THIS PROCE~RE AND

INSERTED TOPc¥: BFYOND

ENTER C5. LEVEL POWER CONTROL

POSITION <A. 0>,

RPV WATfR L~ CMINQI

[XIT THIS PROCEDURf AND

BE DETERMINE ,

ENTER C4, RPV flOODING

RPV WATER LV!. IS RISING.

EXITTHIS PROCEDURE AND

LNT£R £01-1, RPV CONTROL,

AT STEP RC/L-l

CI-2

L

SECTION O-V-G

PAGE 8 OF 50

REVISION 0

EOI PROGRAM MANUAL

.

SECTION O-V-G .

DISCUSSION: STEP Cl-l

cr, ALTERNATE LEVEL CONTROL BASES

1

Thisactionstep directsthe operatorto defeatautomatic ADS function. An ADSactuationwith

the RPVat pressureimposesa severethermaltransientonthe RPVand may significantly

complicate effortsto restoreand maintainRPVwaterlevelas specifiedin this procedure.

.

Because ADSinitiationlogicreceiveslimitedinputsignals,a varietyofplant conditions mayexist

whereautomatic depressurization of the RPV is not appropriate. In certain cases (e.g., RCIC

available but LPCI/CSinjectionvalvesclosedandcontrolpowerfor their operationnot available)

ADSactuation may directly leadto loss of adequate corecoolingand core damage, conditions

thatmightotherwisehave been avoided. Further,conditions assumedin the designof ADS

actuation logic (e.g., no operatoractionforten minutes) do not exist when actionsspecifiedin

thisprocedure are being carriedout.

.

Finally, an operatorcan draw on muchmoreplantinformation than is availableto ADSlogic

(e.g., equipmentout ofservicefor maintenance, operating experiencewith certainsystems,

probability of restorationofoffsitepower,etc.)andthus can betterjudge, based on logicspecified

inthisprocedure, when and how to depressurize the RPV. For all ofthese reasons,it is

appropriate to preventautomaticinitiationofADSas specified.

REVISION 0

PAGE 9 OF 50

SECTION O-V-G

62. RO 295020AK3.08 OOl/MEMlEOI/BASIS//295020AK3.08///l1/21/07 RMS

Unit-2 was at 100% rated power when a spurious Group I isolation occurred. The pressure transient

caused a small LOCA to occur inside the drywell.

EOI-2, section PC/P requires certain actions before and after reaching 12 psig Suppression Chamber

pressure.

Which of the following is the reason that 12 psig in the Suppression Chamber was selected?

A.

Drywell sprays must be initiated prior to this pressure to prevent opening the Suppression Chamber

to Reactor Building vacuum breakers and de-inerting the containment.

B.

Above this pressure indicates that almost all of the nitrogen and other non-condensible gases in the

drywell have been transferred to the torus so initiating Drywell Sprays will not result in containment

failure.

C.oI Above this pressure indicates that almost all of the nitrogen and other non-condensible gases in the

drywell have been transferred to the torus and chugging is possible .

D.

Above this pressure indicates that almost all of the nitrogen and other non-condensible gases in the

torus have been transferred to the drywell air space and Suppression Chamber Sprays will be

ineffective.

KIA Statement:

295020 Inadvertent Cont. Isolation I 5 & 7

AK3.08 - Knowledge of the reasons for the following responses as they apply to INADVERTENT

CONTAINMENT ISOLATION: Suppression chamber pressure response

KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific

plant conditions to determine the effect on Suppression Chamber pressure due to an inadvertent

containment isolation and the basis for that response .

References: EOIPM Section O-V-D

Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall

or recognize discrete bits of information.

0610 NRC Exam

REFERENCE PROVIDED: None

j

Plausibility Analysis:

(

In order to answer this question correctly the candidate must determine the following :

1. The basis for the Pressure Supression Pressure Limit of 12 psig Suppression Chamber pressure.

A is incorrect. This is plausible because initiation of OW sprays at high SC pressure could reduce

pressure low enough to open the SC to RB vacuum breakers. However, this is part of the bases for the

Orywell Spray Initiation Pressure Limit Curve #5.

B is incorrect. This is plausible because initiating SC sprays with high temperature non-condensible

gases in the SC will result in evaporative cooling and a rapid pressure drop. However, the SC to OW

vacuum relief system is capable of compensating for this pressure drop. This is also part of the bases for

the Orywell Spray Initiation Pressure Limit Curve #5.

C is correct.

D is incorrect. This is plausible if the LOCA occurred inside the Suppression Chamber and NOT the

Orywell as given in the stem.

(

EOI PROGRAM MANUAL

SECTION O-V-O

STEP:

PC/P-6

.

NO L

YES

L

o

INITIATE

SUPPR CHMBR SPRAYS USII\\C w..x RHR PUMPS WI

REQI)RED TO ASSURE ADEOUA.TE CORE COOLING BY" COO11NlJ()J$

IN! (N)PX 17C)

PC/P-5

EOI-2, PRIMARY CONTAINMENT CONTROL BASES

L

SUPPR CHI.4BR PRESS EXCEEDS <A.65>

CONTINUE

IN THIS PROCEDURE

PC/P-6

L

i(

SECTION O-V-O

TO CURVE 5

(Refer to EOI Program "'onuot "-

Section IV, Appendi. A. Curves

and Tables Used in the EOls)

PAGE 42 OF 244

NO L

TO PC/P-ll

REVISION 0 *

EOI-2. PRIMARY CONTAINMENT CONTROL BASES

EOI PROGRAM MANUAL

SECTION o-v-n

DISCUSSION: STEP PCIP-6

1

This contingent action step requires the operator to wait until the stated condition has been met

before continuing in EOI-2. Performance ofsubsequent actions in this section ofEOI-2 will not

be performed until suppression chamberpressure exceeds Suppression Chamber Spray Initiation

Pressure.

Engineering calculations have determined that ifsuppression chamber pressure exceeds <A.65>,

Suppression Chamber Spray Initiation Pressure, there is no assurance that chugging will be

prevented at downcomer openings ofthe drywell vents. This value is rounded-off in the EOI to

use the closest, most conservative value that can be accurately determined on available

instrumentation.

Suppression Chamber Spray Initiation Pressure is defined to be the lowest suppression chamber

pressure that can occur when 95% ofnoncondensables in the drywell have been transferred to

airspace ofthe suppression chamber. Scale model tests have demonstrated that chugging will not

occur so long as the drywell atmosphere contains at least 1% noncondensables. To prevent the

occurrence ofconditions under which chugging may happen, Suppression Chamber Spray

Initiation Pressure is conservatively defined by specifying 5% noncondensables.

Chugging is the cyclic condensation ofsteam at downcomer openings ofthe drywell vents.

Chugging occurs when steam bubbles collapse at the exit ofdowncomers. The rush ofwater that

fills the void (some ofwhich is drawn up into the downcomer pipe) induces a severe stress at the

junction ofthe downcomer and vent header. Repeated application ofthis stress can cause these

joints to experience fatigue failure (cracks), thereby creating a pathway that bypasses the pressure

suppression function ofprimary containment. Subsequent steam that discharges through

downcomers would then exit through the fatigued cracks, and directly pressurize suppression

chamber air space, rather than discharging to and condensing in the suppression pool.

Although operation ofsuppression chamber sprays by itselfwill not prevent chugging, the

requirement to wait to initiate drywell sprays until reaching Suppression Chamber Spray Initiation

Pressure assures that suppression chamber spray operation is attempted before operation of

drywell sprays. Therefore, actions to initiate drywell sprays need to be directed only ifsuppression

chamber sprays were unable to reduce primary containment pressure or they could not be

initiated.

REVISION 0

PAGE 43 OF 244

SECTION O-V-O

EOI-2, PRIMARY CONTAINMENT CONTROL BASES

EOI PROGRAM MANUAL

SECTION O-V-O

DISCUSSION:

STEP PCIP-8

I

~

Thisdecision stephas the operatorevaluatethe presentstatusofdrywellpressureand drywell

temperature to determineif conditionsare favorable fordrywellsprayoperation.

Drywell sprayoperationreducesdrywellpressureandtemperature through the combinedeffects

of evaporative and convectivecooling.Duringevaporative cooling,water spray undergoesa

changeof state,liquid to vapor,whereasconvective coolinginvolvesno changeofstate.

Evaporative coolingoccurs when wateris sprayedinto a superheatedatmosphere.Waterat the

surfaceof eachdroplet is heated and flashesto steam,absorbingheat energy fromthe drywell

atmosphere until the atmospherereachessaturated conditions. In the drywell,with a typical

drywell sprayflowrate, the evaporativecoolingprocessresults in an immediate,rapid, large

reduction in pressure. This pressurereductionoccursat a rate much faster than can be

compensated for by the primarycontainment vacuumreliefsystem.Unrestrictedoperationof

drywell sprayscouldcause an excessivenegativedifferential pressureto occurbetweenthe

drywell andsuppressionchamber,largeenoughto causea lossofprimarycontainmentintegrity.

Convective coolingoccurs when water is sprayedinto a saturatedatmosphere. Sprayedwater

droplets absorbheat from the surrounding atmosphere throughconvectiveheat transfer(sensible

heatfromthe atmosphereis transferredto the waterdroplets). This effect reducesdrywell

ambienttemperature andpressureuntilequilibrium conditionsare established.Theconvective

coolingprocessoccurs at a rate much slowerthan the evaporative cooling process. An operator

caneffectively controlthe magnitudeofa containment temperature/pressure reductionfrom

convective coolingby terminatingoperationof drywellsprays.

Considering the pressuredrop concernsdescribed above,engineeringcalculationshave

determined that primarycontainmentintegrityis assuredwhendrywellsprays areoperatedin the

safeareaofDrywell SprayInitiationLimit Curve(Curve5).DrywellSprayInitiationLimit is

defined to bethe highestdrywelltemperature at whichinitiationofdrywellsprayswill not result

in an evaporative cooling pressuredropto beloweither:1)drywell-below-suppression chamber

differential pressurecapability,or 2) high drywell pressurescramsetpoint,

If drywell temperature and pressureare withinthe safe areaofCurve 5, the operatorcontinuesat

StepPCIP-9.

Ifdrywell temperature and pressureare not withinthe safeareaofCurve 5, then drywell spray

operation is not permitted, and the operatoris directed to Step PCIP-l1.

REVISION 0

PAGE 47 OF 244

SECTION O-V-D

c'

63 . RO 295032EAl.OI OOl/C/A/TlGI/E0I-3//295032EAl.Ol//RO/SR0/1l/20/07 RMS

Given the follow ing plant conditions:

Unit 2 experienced a MSL break from full power.

Both inboard and outboard MSIVs on the "B" steam line fail to isolate however, the reactor

scrams and all rods insert.

Steam Leak Detection panel 9-21 indications are as follows :

- 2-TI-1-60A

- 2-TI-1-60B

- 2-TI-1-60C

- 2-TI-1-60D

320°F

323°F

33rF

318°F

(

No other temperature indications are alarming at this time.

Which ONE of the following describes the appropriate operator actions and the reasons for those

actions?

REFERENCE PROVIDED

A.

Emergency depressurize the reactor due to two EOI-3 areas being above Max Safe.

B.

Rapidly depressurize the reactor due to one EOI-3 areas above Max Safe and one area

approaching Max Safe.

c.'; Enter 2-EOI-1, RPV Control and initiate a Reactor Scram due to one EOI-3 area being above Max

Safe.

D.

Enter 2-GOI-100-12A, Unit Shutdown and commence a normal shutdown and cooldown due to a

primary system dscharging outside Primary Containment.

KIA Statement:

295032 High Secondary Containment Area Temperature / 5

EA1.01 - Ability to operate and/or monitor the follow ing as they apply to HIGH SECONDARY

CONTAINMENT AREA TEMPERATURE : Area temperature monitoring system

KIA Justification: This question satisfies the KiA statement by requiring the candidate to use specific

plant conditions to determine the required actions which result from high secondary containment

temperatures as indicated by Area Temperature Monitoring instrumentation.

References:

2-EOI-3 Flowchart, EOIPM Section O-V-E

Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble,

sort, and integrate the parts of the question to predict an outcome. This requires mentally using this

knowledge and its meaning to predict the correct outcome.

0610 NRC Exam

(

REFERENCE PROVIDED: 2-EOI-3 Flowchart

Plausibility Analysis:

In order to answer this question correctly the candidate must determine the following :

1. Which area(s) are above or approaching Max Safe

2. Based on Item #1 above, determine the appropriate action and the basis for that action.

A is incorrect. This is plausible because all four temperatures provided are greater than 3150F as

indicated on Table 3. However, only one indicator applies to an EOI 3 area, therefore only ONE area is

above Max Safe.

B is incorrect. This is plausible because one area is above Max Safe and given conditions indicate an

un-isolable leak exists which implies conditions are degrading. However, with no other temerature

indications in alarm, anticipating the requirement to Emergency Depressurize is NOT ppropriate.

C is correct.

D is incorrect. This is plausible because all four temperatures provided are greater than 3150F as

indicated on Table 3. However, only one indicator applies to an EOI 3 area, therefore only ONE area is

above Max Safe. In addition, this step is only addressed if Emergency Depressurization will not reduce the

discharge into Secondary Containment. In this case, it would.

EOI-3, SECONDARY CONTAINMENT CONTROL BASES

EOI PROGRAM MANUAL

SECTION O-V-E

ENTRY CONDITIONS:

EOI-3

DISCUSSION:

~--

I

Entry conditions for this procedure are symptomatic ofconditions which, if not corrected, could degrade into an

emergency. Adverse affects on equipment operability and conditions that directly challenge secondary containment

integrity were specifically considered in the selection ofthese entry conditions. Following is a description of each

entry condition:

Area temperature above the maximum normal operating value of Table 3

A secondary containment area temperature above the maximum normal operating value of Table 3, Secondary

Containment Area Temperature, is an indication that steam from a primary system may be discharging into

secondary containment. As temperatures continue to increase, continued operability ofequipment needed to carry

out EOI actions may be compromised. High area temperatures also present a danger to personnel since access to

secondary containment may be required by actions specified by EOls.

Maximum normal operating temperature is defined to be the highest value ofa secondary containment area

temperature expected to occur during normal plant operating conditions with all directly associated support and

control systems functioning properly.

Differential pressure at or above <A.38> inches of water

High secondary containment differential pressure is indicative ofa potential loss of secondary containment

structural integrity, and could result in uncontrolled release of radioactivity to the environment.

Reactor Zone Ventilation exhaust radiation level above <A.39>

High Reactor Zone Ventilation exhaust radiation levels may indicate that radioactivity is being released to the

environment when the system should have automatically isolated.

Refuel Zone Ventilation exhaust radiation level above <A.40>

High Refuel Zone Ventilation exhaust radiation levels may indicate radioactivity is being released to the

environment when the system should have automatically isolated.

Floor drain sump water levcl above <A.41>

A secondary containment floor drain sump water level above maximum normal operating level is an indication that

steam or water may be discharging into secondary containment.

Maximum normal operating floor drain sump water level is defined to be the highest value of secondary

containment floor drain sump water level expected to occur during normal plant operating conditions with all

directly associated support and control systems functioning properly.

Area watcr level above <A.42>

Secondary containment area water level above maximum normal operating level is an indication that steam or water

may be discharging into secondary containment.

Maximum normal operating secondary containment area water level is defined to be the highest value of secondary

containment area water level expected to occur during normal plant operating conditions with all directly associated

support and control systems functioning properly.

REVISION 1

PAGE 9 OF 73

SECTION O-V-E

.... .... ..

...._._ - - - - - - - - - - - - - - - - - - - - - - - - - - - - _ ....

EOI PROGRAM MANUAL

SECTION O-V-E

scrr-6

and

SCrr-7

DISCUSSION:

EOI-3, SECONDARY CONTAINMENT CONTROL BASES

el

1

(

Step SCrr-6 is a before decision step that has the operator evaluate current and future efforts to lower secondary

containment area temperatures, in relation to the current value and trend ofsecondary containment area

temperatures, to determine if a reactor scram is necessary. The before decision step requires that this determination

and subsequent actions be performed before any secondary containment area temperature reaches its respective

maximum safe operating temperature value provided in Table 3.

Maximum safe operating temperature is defined to be the highest temperature at which neither: I) equipment

necessary for the safe shutdown of the plant will fail, nor 2) personnel access necessary for safe shutdown ofthe

plant will be prevented. The maximum safe operating temperature value for all secondary containment areas is

provided in Table 3, Secondary Containment Area Temperature.

This step is reached only when additional actions have been required to reverse an increasing secondary

containment area temperature trend. If all secondary containment area temperatures can be maintained below their

respective maximum safe operating values, the operator returns to Step SCrr*I . Ifit is determined that all

secondary containment area temperatures cannot be maintained below their respective maximum safe operating

values, the operator continues at Step scrr-7.

Step SCrr-7 is an enter and execute concurrently step that requires the operator to enter

EOI-I, RPV Control, at Step RC-I , and to perform the actions concurrently with this procedure. Because this step

is prioritized with the miniature before decision step symbol relating to SCrr-6, this action should be performed

before any secondary containment area temperature reaches its respective maximum safe operating value.

Initiation of reactor scram (Step RC-I) before any secondary containment area temperature reaches its respective

maximum safe operating value may halt the increase in secondary containment area temperature(s), since the RPV

is the only significant source of heat, other than a fire, that could cause secondary containment area temperatures to

exceed their respective maximum safe operating values.

REVISION 1

PAGE 27 OF 73

SECTION O-V-E

TABLE 3

SECONDARY CNTMT AREA TEMP

PANEL~

PANEL9-21

MAX

POTENTi6,L

AREA

AlARM WINDOW

TEMPEleMENT

NORMA\\.

SI.,FE

ISOlATION

(UNLESS NOTED)

(UNLESSNOTED)

VALUEof

SOURCES

RHRSYS I PUMPS

XA-6!>-36-4

74-95.1.,

A\\..I.,RMED

160

FCV-7447,48

RHRSYS II PUMPS

X6,-65-3E-4

74-958

ALARMED

210

FC\\L1441,48

HPCI ROQM

X"I-65-3F-l0

73-65.1.,

ALI.,RMED

270

FCV-73-2, 3,44, 81

CSSVS I PUMPS

XA-.55-30-10

11-41A

ALARMED

190

FCV-1'-2, 3. 39

RCICRooM

CSSYSIlPUMPS

X4-.55-3E-29

7lH>9B (pANEL~)

ALARMED

150

NONE

X6,-.55-3D-10

71041S,C, D

AlARMED

200

FC\\L1'-2,3

TO? OFTORUS

X6,-55-3F-10

73-55B.C,D

Al.6,RMED

240

FCV-13-2, 3, 81

XA-65-3E-4

74-9SG, H

AlARMED

240

FCV-1404i.48

STEAMTUNNEL(RB)

XA-.5!:--3D-24

1-00.4 (PANEL9-3)

Al.A,RMED

315

MSIVa

FCV-71-2, 3, FCV-'S9-1, 2, 12

QWACCESS

X4-55-3E-4

74-95E

Al.6,RMED

170

FC\\L74047.48

R8 EL565W

X4-55-.5B-32 (PANEL9-5)

69-S3SA.B,C.D

ALARMED

170

FCV~-l, 2, 12

(RWCUPIPETRENCH)

XI.,-.55-.5B-33 (PANEL9-5)

{AUXINST ROOM}

ALARMED

RWCUH. X. ROOM

X4-5'".>-3D-t1

G9-29F,G, H

AL4.RMED

220

FCV~-1. 2, 12

RWCUPUMPA

X4.-.55-3D-17

G9-29D

ALARMED

215

FCV-&.:!-1.2. 12

~"'CUPUMP8

XA-55-3D-17

G9-29E

ALARMED

215

FCV.(;'9-1, 2.12

RBE L593

X4-.55-3E-4

74-95C,D

Al.6,RMED

195

FCV-14047,48

RBEL621

XI.,-.55-3E-4

74-95F

ALl.,RMED

155

FCV-43-13, 14

/"""'\\

...J

.J

...J

I:l

...J

!

l! ~

I~

..

~ i

~

~~

lZ ~

i '

....

S~

.. ~Ii

SIC.

all

.. Il.! I

U

~~

I:

)o\\!.l.

~ ~ !

JI

m

~I

~

..

i

~'"

L

...

WHILE EXECUTING THE FOLLOWING STEPS:

IF

THEN

EMERGENCY RPV DEPRESSURIZATION

IS ANTICIPATED

AND

RAPIDLY DEPRESSURIZE THE RPV

WITH THE MAIN TURB BYPASS VLVs

THE REACTOR WILL REMAIN SUBCRITICA,L

IRRESPECTIVE OF COOLDOWN RATE

WITHOUT BORON UNDER ALL CONDITIONS

(SEE NOTE)

RCIP-3

)--

E

MINATION *

REFERENCE

PROVIDED TO

( CANDIDATE

il"u

hl~

3

i

!I

=I

J

j.

~

C")*-ow

(If)*-fa

1 ,L.a~.....,..,..

il'

W-l-\\++-t+tT1

~I

SIU':"'H.++-t+t

lill * .. " ! i

i

I H II '

[n '*11

1' . .

lit!II!! !l

II

.!Ii!

. i :1*

.

illili II

!

iit

11 I I

I

it

I I~ I il ill Ii I

fl .

1 I

I

I I

II: I .

111"1

I 'II

II I,,,

t

LI

C")*-ow

64. RO 295033EA2.01 OOl/C/A/TlG2/SC/R//295033EA2.0l//RO/SR0/1l/20/07 RMS

Given the following plant conditions:

Unit-2 is at 100% rated power.

A RWCU drain line cracked and is spilling into the Reactor Building .

Area Radiation Monitors in the Reactor Building read as follows:

Reactor Building Elevation 593

Reactor Building Elevation 565 West

Reactor Building Elevation 565 East

Reactor Building Elevation 565 Northeast

All other Reactor Building areas

RWCU has been successfully isolated

1100 mR/hr

800 mR/hr

850 mR/hr

1050 mR/hr

NOT ALARMED

Which ONE of the following describes the required action that MUST be directed by the Unit Supervisor

and/or Shift Manager?

REFERENCE PROVIDED

A.

Enter 2-EOI-1, RPV Control and initiate a Reactor Scram due to two EOI-3 areas being above Max

Safe.

B.oI

Enter 2-GOI-100-12A, Unit Shutdown and commence a normal shutdown and cooldown due to two

areas above Max Safe with the source of the leak isolated .

C.

Scram the reactor and Emergency depressurize the reactor due to two EOI-3 areas being above

Max Safe.

D.

Rapidly depressurize the reactor with Bypass Valves due to Emergency Depressurization being

anticipated.

KIA Statement:

295033 High Secondarv Containment Area Radiation Levels / 9

EA2.01 - Ability to determine and/or interpret the following as they apply to HIGH SECONDARY

CONTAINMENT AREA RADIATION LEVELS: Area radiation levels

KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific

plant conditions to determine the required actions which result from high secondary containment radiation

levels as indicated by Area Radiation Monitoring instrumentation.

References:

2-EOI-3 Flowchart

Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble,

sort, and integrate the parts of the question to predict an outcome. This requires mentally using this

knowledge and its meaning to predict the correct outcome.

0610 NRC Exam

(

REFERENCE PROVIDED: 2-EOI-3 Flowchart

Plausibility Analysis:

In order to answer this question correctly the candidate must determine the following:

1. Which area(s) are above or approachin,g Max Safe

2. Based on Item #1 above, determine the appropriate action.

A is incorrect. This is plausible because this action requires at least one area greater than Max Safe.

However, this is not appropriate since the source of the leak is isolated.

B is correct.

C is incorrect. This is plausible because this action requires two areas greater than Max Safe. However,

this is not appropriate since the source of the leak is isolated.

D is incorrect. This is plausible because this action requires at least one area greater than Max Safe and

another area approaching Max Safe. However, this is not appropriate since the source of the leak is

isolated.

TABLE 4

-

SECONDARY CNTMTAREA RADIATION

APPUCABLE

MAX

P01ENTIAL

AREA

RADIATION

NORMAL

SAFE

ISOLATION

INDICATORS

VALUEMRlHR

VALUEMR/HR

SOURCES

RHR SYSJPUMPS

9o-2!iA

ALARMED

1000

FCV-74-41*.1a

RHRSYSII PUMPS

9O-2M

ALARMED

1000

FCV-74-41*.1a

HPClROOM

9O-2AA

AlARMED

1000

FC~~2.3.44.81

CSSYSI PUMPS

9O-26A

ALAfUAED

1000

FCV-71-2.3.~

RClCROOM

CS SYSIIPUMPS

<X>-27A

ALARMED

1000

NONE

TQPOFTORUS

FCV-13-2. 3. 81

~29A

ALARMED

1000

FCV-74-47.48

GENERAL AREA

FCV-71-Z,3

RBEI.$5W

~20A

ALARMED

1000

FCV~1.2.12

SDVVENfS & DRAINS

RBEl.56SE

~21A

ALARMED

1000

SDVVENfS 8; DRAINS

RBEI.!05NE

~~A

AlARMED

1000

NONE

TIPROOM

~22A

ALARMED

100.000

TIP8All VALVE

R8E1.593

~13A.14A

ALARMED

1@

FCV-74-47.48

R8 El.G21

9O-9A

ALARMED

1000

FCV-43-13,14

RECIRC tAG SETS

00-4A

ALARMED

1000

NONE

REFUEl. FLOOR

00-1A. 2A. 3A

AI.A.QMED

1000

NONE

NO L

L

\\

~

_..**...

.'

.;'

..-

/.'.'

.'...

"

~..~...

SC/M

NO

L

l0

,,I'

............................................_\\

.'

..\\

<<,,#.

.'

I'

..-

..,~4r

.'

1"

YES L

SC/R-4

\\ ....

L

.~

_._ -.__ _._-

-

\\

L

WHEN

AADlATJONlVlslN20R MOREAAEAS

--

AAEA80VE MAXSAFE{rABLE4}

ll1EN

CONnNUE

SClR-7

EMERGENCY RPV DEPRESSURIZAnON IS REQUIRED

{EOI-1,RClN; C1-2; C1-21;C5-1)

SCJR-8

L

.,

~

WHILE EXECUTING THE FOLLOWING STEPS:

IF

THEN

EMERGENCY RPV DEPRESSURIZATION

IS ANTICIPATED

AND

RAPIDLY DEPRESSURIZE THE RPV

WITH THE MAIN TURB BYPASS VLVs

.

THE REACTOR WILL REMAIN SUBCRITICAL

IRRESPECTIVE OF COOLDOWN RATE

WITHOUT BORON UNDER ALL CONDITIONS

(SEE NOTE)

E

MINATION

REFERENCE

PROVIDED TO

( CANDIDATE

(

C"')*-oIII

(If)*

-

rl

o

W

J

11HT-tTTi":rn

I

il'ltTt1lnTTm

S *"

iI

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IJ .

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-

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(If)*-oII

65. RO 295035EA2.02 OOl/C/A/EOI/EOI-3/S85/295035EA2 .02//RO/SRO/

Given the following plant conditions:

Unit 2 is at 100% power.

During the backwash of a RWCU demineralizer the backwash receiving tank ruptured.

The RWCU system has been isolated.

Secondary containment conditions are as follows:

- All Reactor and Refuel Zone radiation monitors trip on high radiation.

- ONLY SGT train "C" can be started.

- It is operating at 10000 scfm and taking suction on the refuel and reactor zones .

- Refuel zone pressure :

- Reactor zone pressure:

- AREA RADIATION LEVELS

RB EL 565 W, 565 E, 565 NE:

RB EL 593

RB EL 621

-0.12 inches of water.

+0.02 inches of water.

250 mr/hr

upscale

Which ONE of the following describes the required action and the type of radioactive release in progress?

REFERENCE PROVIDED

A.

Initiate a shutdown per 2-GOI-100-12A. Elevated radiation release.

B~ Initiate a shutdown per 2-GOI-100-12A Ground level radiation release.

C. Scram the reactor, emergency depressurize the RPV. Elevated radiation release .

D.

Scram the reactor, emergency depressurize the RPV. Ground level radiation release.

KIA Statement:

295035 Secondary Containment High Differential Pressure

EA2.02 - Ability to determine and/or interpret the following as they apply to SECONDARY

CONTAINMENT HIGH DIFFERENTIAL PRESSURE: Off-site release rate: Plant-Specific

KIA Justification: This question satisfies the KIA statement by requiring the candidate to correctly

identify the type of off-site release and required actions due to high differential pressure in the secondary

containment.

References: 2-EOI-3 Flowchart

Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble,

sort, and integrate the parts of the question to solve a problem. This requires mentally using this

knowledge and its meaning to resolve the problem .

0610 NRC Exam

REFERENCE PROVIDED: 2-EOI-3 flowchart

Plausibility Analysis:

In order to answer this question correctly the candidate must determine the following :

1. Which area(s) are above or approaching Max Safe

2. Based on Item #1 above, determine the appropriate action.

3. Whether plant conditions indicate an elevated or ground level release.

NOTE: EOI-3 steps SC/R-8 and SC/R-9 apply, requiring shutdown per 2-GOI-100-12A because 2 or

more areas are above max safe rad levels but a primary system is not discharging to the RB. Insufficient

RB to atmosphere dp (greater than -0.25 inches of water) indicates loss of secondary containment

integrity. The positive reactor zone pressure is causing an unmonitored and uncontrolled ground level

release of radioactive contaminants.

A is incorrect. The release from the Reactor Building is not elevated . This is plausible because the

required actions are correct except the differential pressure results in a ground level release.

B is correct.

C is incorrect. Conditions do not warrant a scram at this point. In addition, the release from the Reactor

Building is not elevated. This is plausible if the candidate fails to recognize that a primary system is not

discharging to the Reactor Building.

D is incorrect. Conditions do not warrant a scram at this point. This is plausible if the candidate fails to

recognize that a primary system is not discharging to the Reactor Building.

TABLE 4

-

SECONDARY CNTMT AREA RADIATION

APPLICABLE

MAX

POTENTIAL

AREA

RADIATION

NORMAL

SAFE

ISOLATION

INDICATORS

VALUEMR/HR

VALUEMRIHR

SOURCES

RHR SYS I PUMPS

9o-~

ALARMED

1000

FCV-74-47.48

RHR SYS II PU}.1PS

gO-2M

ALARt.1ED

1000

FCV-14-47.48

HPCIROOM

9O-2AA

ALARMED

1000

FCV-73-2. ~. 44. 81

CSSYS I PUMPS

9O-2GA

AlARMED

1000

FCV-11 -2.~ . 39

RCICROOM

CS SYS II PUMPS

9O-27A

ALARMED

1000

NONE

TOP OF TORUS

FCV-73-2.3, 81

9O-29A

AlARMED

1000

FCV-74-41.48

GENERAL AREA

FCV-11-2,3

RBEl&S5W

9O-2OA

AlARt.IED

1000

FCV~1 .2.12

SOVVENrS &. DRAINS

RB El!i65 E

90-21A

AlARMED

1000

SOVVENrS 8.DRAINS

RBEl!i65 NE

9O-23A

AlARMED

1000

NONE

TIP ROOM

9O-22A

AlARMED

100.000

nPBAtt VALVE

RBEl593

90-13A., 14A

AlARMED

100:>

FCV-14-47.48

RBEl621

9O-9A

AlARMED

100:>

FCV-43-13.14

RECIRC MG SETS

9iJ.4A

AlARMED

1000

NONE

REFUel A.OOR

9O-1A. 2A, 3A

AlARMED

1(XX)

NONE

t~

ISOIoATe&.SYSTEMS THATAREDISCHARGING

INTOTHEAREAEXCEPT SYSTEMSREQUIRED TO;

8EOPEAAlEO 8YEO's

QR

..

SUPPRESSA FIRE

L

SCIR-3

NO L

L

\\

,..:

~

..-

/.'

./.'

"

....*/ .

SCfR-9

L

(0

/'

.............................................\\

.,.-

".'

/

,..".-.

" IJ "

<<,r,ff

YES L

SCIRr4

f)

\\

NO L

L

80M

L

._ _---_._--_ _._

_

\\

L

VllHEN

RADIATION LVls IN:2OR MOREAREAS

AREABOVEMAXSAFE{TABLE4)

D:!.Sti

CONl1NUE

SCJR-7

EMERGENCY RPV DEPRESSURlZAll0N IS REQUIRED

(EOI-1,RClP-4: C1-2; C1-21;C5-1)

L

SCIR-8

E

MINATION

REFERENCE

..

PROVIDED TO

CANDIDATE

CW)

J

- I

ri

. . . .

0

f

(

W

..

m* :!

e , i!

Jii II

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I~~~-.L_------,--_

(

66. RO GENERIC 2.1.33 OOl/C/A/T3///GENERIC 2.1.33//RO/SRO/ll/25/07 RMS

Which ONE of the following describes the protective function(s) required to be Operable for the specified

mode and/or condition?

A.

Starting up in Mode 2 with IRM's on range 1 to 2:

IRM Hi Scram function

BPWS

RBM

APRM Hi (setdown - 15%).

B.

Starting up in Mode 2 with APRM downscales clear:

APRM Hi (setdown - 15%)

APRM Hi (120%)

Mode switch - Shutdown position

RWM.

C'!" Shutting down in Mode 2 with IRM's on range 1 to 2:

IRM Hi Scram function

BPWS

Manual Scram pushbuttons

RWM.

D.

Shutting down in Mode 2 with average SRM readings at -5x101 cps:

IRM Hi Scram function

APRM Hi (setdown - 15%)

OPRM upscale trip

RWM.

KIA Statement:

Conduct of Operations

2.1.33 Ability to recognize indications for system operating parameters which are entry-level conditions for

technical specifications

KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific

plant conditions to determine when entry into Technical Specifications is required .

References:

Technical Specifications

Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble,

sort, and integrate the parts of the question to predict an outcome. This requires mentally using this

knowledqe and its meaning to predict the correct outcome.

0610 NRC Exam

(

REFERENCE PROVIDED: None

Plausibility Analysis:

In order to answer this question correctly the candidate must determine the following:

1. Tech Spec applicability for the listed systems with the given plant condition.

NOTE: The distractors are all plausibe since only one system or function is incorrect in each distractor.

A is incorrect. The RBM is not required until >27% rated power.

S is incorrect. The APRM Hi (120%) is not required until Mode 1

C is correct.

D is incorrect. the OPRMs are not required until Mode 1 and >25% rated power.

Control Rod Block Instrumentation

3.3.2.1

Table 3.3.2.1-1 (page 1 of 1)

Control Rod Block Instrumentation

FUNCTION

APPLICABLE

MODES OR

OTHER

SPECIFIED

CONDITIONS

REQUIRED

CHANNELS

SURVEILLANCE

ALLOWABLE

REQUIREMENTS

VALUE

1.

Rod Block Monitor

a.

Low Power Range - Upscale

(a)

2

b.

Intermediate Power Range - Upscale

(b)

2

c.

High Power Range - Upscale

(f),(g)

2

d.

Inop

(g),(h)

2

e.

Downscale

(g),(h)

2

2.

1(c),2(c)

(e)

SR 3.3.2.1.8

(e)

SR 3.3.2.1.8

(e)

SR 3.3.2.1.8

NA

(i)

SR 3.3.2.1.2

NA

SR 3.3.2.1.3

SR 3.3.2.1.5

SR 3.3.2.1.7

3.

Reactor Mode Switch -

Shutdown Position

(d)

2

SR 3.3.2.1.6

NA

(a)

THERMAL POWER ~ 27% and s 62% RTP and MCPR less than the value specified in the COLR.

(b)

THERMAL POWER> 62% and s 82% RTP and MCPR less than the value specified in the COLR.

(c)

With THERMAL POWER s 10% RTP.

(d)

Reactor mode switch in the shutdown position.

(e)

Less than or equal to the Allowable Value specified in the COLR.

(f)

THERMAL POWER> 82% and < 90% RTP and MCPR less than the value specified in the COLR.

(g)

THERMAL POWER ~ 90% RTP and MCPR less than the value specified in the COLR.

(h)

THERMAL POWER ~ 27% and < 90% RTP and MCPR less than the value specified in the COLR.

(i)

Greater than or equal to the Allowable Value specified in the COLR.

(

BFN-UNIT 1

3.3-20

Amendment No. 2-M,262

September 27, 2006

RPS Instrumentation

3.3.1 .1

Table 3.3.1.1-1 (page 1 of 3)

Reactor Protection System Instrumentation

FUNCTION

APPLICABLE

MODES OR

OTHER

SPECIFIED

CONDITIONS

REQUIRED

CHANNELS

PER TRIP

SYSTEM

CONDITIONS

REFERENCED

FROM

REQUIRED

ACTION 0 .1

SURVEILLANCE

ALLOWABLE

REQUIREMENTS

VALUE

1.

a.

Neutron Flux - High

2

3

G

s 120/125

divisions of full

scale

SR 3.3.1.1.9

5(a)

3

H

~ 120/125

SR 3.3.1.1.4

divisions of full

SR 3.3.1.1.9

scale

b.

Inop

2

3

G

SR 3.3.1.1.3

NA

5(a)

3

H

SR 3.3.1.1.4

NA

SR 3.3.1.1 .14

2.

Average Power Range Monitors

a.

Neutron Flux * High,

2

3(b)

G

~ 15% RTP

Setdown

SR 3.3.1.1.6

SR 3.3.1.1.7

b.

Flow Biased Simulated

3(b)

F

s 0.66 W

Therma l Power - High

SR 3.3.1.1.2

+ 66% RTP

SR 3.3.1.1.7

and s 120%

RTP(c)

c.

Neutron Flux - High

3(b)

F

~ 120% RTP

SR 3.3.1.1.2

SR 3.3.1.1.7

Reactor Vessel Water Level -

(continued)

(a)

With any control rod withdrawn from a core cell containing one or more fuel assemblies.

(b)

Each APRM channel provides inputs to both trip systems.

(c)

[0.66 W + 66% - 0.66 t>W] RTP when reset for single loop operation per LCO 3.4.1, "Recirculation Loops Operating :

BFN-UNIT 1

3.3-6

Amendment No. 236, 262, 269

March 06, 2007

(

RPS Instrumentation

3.3.1 .1

Table 3.3.1.1-1 (page 2 of 3)

Reactor Protection System Instrumentation

FUNCTION

2.

Average Power Range

Monitors (continued)

d.

Inop

e.

2-0ut-Of-4 Voter

f.

OPRM Upscale

3.

Reactor Vessel Steam Dome

Pressure - High(d)

4.

Low, LeveI3(d)

5.

Main Steam Isolation Valve -

Closure

6.

Drywell Pressure - High

7.

Scram Discharge Volume

Water Level - High

APPLICABLE

MODES OR

OTHER

SPECIFIED

CONDITIONS

1,2

REQUIRED

CHANNELS

PER TRIP

SYSTEM

2

8

2

CONDITIONS

REFERENCED

FROM

REQUIRED

ACTION D.1

G

F

G

SURVEILLANCE

REQUIREMENTS

SR 3.3.1.1.14

SR 3.3.1.1.7

SR 3.3.1.1.17

SR 3.3.1.1.8

SR 3.3.1.1.10

SR 3.3.1.1.8

SR 3.3.1.1.13

SR 3.3.1.1.13

ALLOWABLE

VALUE

NA

S 1090 psig

~ 528 inches

abovevessel

zero

S 10% closed

s 2.5 psig

a.

Resistance Temperature

Detector

1,2

2

G

s 50 gallons

5(a)

2

H

s 50 gallons