ML20206R979

Exam Rept 50-368/OL-86-02 During Wk of 860523.Exam Results: One Reactor Operator Failed Written Exam.All Other Candidates Passed All Portions of Exam.Requalification Program Audit Satisfactory

ML20206R979
Person / Time
Site:	Arkansas Nuclear
Issue date:	06/25/1986
From:	Cooley R, Mccrory S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20206R940	List: ML20206R937 ML20206R979
References
50-368-OL-86-02, 50-368-OL-86-2, NUDOCS 8607070272
Download: ML20206R979 (61)

Text

.

r APPENDIX U. S. NUCLEAR REGULATORY COMMISSION REGION IV NRC Examination Report:

50-368-O L-86.02 Docket:

50-368 Licensee:

Arkansas Power & Light (AP&L)

P. O. Box 551 Little Rock, Arkansas 72203 Facility Name:

Arkansas Nuclear One, Unit 2 (ANO-2)

(x/

/o M C /h Chief Examiner:

@ephend../NcTrory,ChiefExamine Date /

b!Mh Approved:

Ralph L. Cooley, Chief, 0 $rator Licensing D4te Section Summary An audit of the ANO-2 Requalification Program was conducted by Regian IV personnel and:f und to be satisfactory.

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1 8607070272 860702 D

PDR ADOCK 05000368 8'

V PDR 14

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DETAILS 1.

Persons Examined Seven SR0s and two R0s were administered NRC generated written and orll operating requali.fication examinations. One instant SR0 candidate wa.;

i administered an oral operating examination'. One SR0 retake of a sing e 3

section of a written examination was also administered.

3 2.

Results of Examination One R0 failed the written examination. All other examinees passed all portions of the examination.

3.

Examiners i

Stephen L. McCrory, Chief Examiner j

David N. Graves j

4.

Examination Review Coments and Resolutions Facility comments are listed by. section question number and followed by the examiner's resolution.

4 a.

Questions 1.01(d)

Should also accept increase if assumption is and 5.02(d) stated that increasing pressure decreases nucleate boiling.

Resolution:

Agree.

b.

Question 2.02(b)

Following a diesel start due to loss of offsite power, all normal engine trips are functional.

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Resolution:

Agree. The only trip not listed on the key, failure to start, was added to the key.

c.

Question 2.07 See attachment 2.07 - Condition 3 is a

)

Question 6.03 restatement of of your second answer. The valves i

are comon to LTOP and ECCS.

(Attachment 2.07 is l

from AN02 procedure 2203.12J, Revision 8, pg. 18 of74)

Resolution:

Agree. Answer key modified to reflect the provided answer.

d.

Question 3.02(b)

The approximate valve positions for a 5% flow Question 6.04(b) demand is 11% and 15% open.

3 Resolution:

Agree. Answer key modified.

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.m m.-

m_..

,m.,

...m m,,

,-_,.__.,._.,,.m

p-

}'

3 J

e.

Question 3.05(c)

Containment spray is activated on a high containment pressure (23.3 psia) and on SIAS (low pressurizer or high containment pressure of 18.4 psia).

Acceptable answers would be:

l 1.

2 or 5 and 8 2.

5 and 8 l

3.

23.3 psia containment pressure and SIAS Resolution:

Agree.

Answer key modified, f.

Question 3.06(a)

Should also accept: (alternate names but same answer).

1.

PPS cabinets (or back panel) 2.

Front panel Resolution:

Agree.

Other names for some panels accepted.

I g.

Question 4.07 See attachment for new revision.

and 7.06 Resolution:

Answer key modified to the new procedure revision.

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h.

Question 5.05(a)

Should also accept " critical" based on the thumb

-rule used in your answer.

Resolution:

Agree.

Critical accepted.

i.

Question 6.06 Primary calorimer.ric = Delta T power Resolution:

Agree.

j.

Question 7.08(b)

The answer is 2 (not B).

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Resolution:

Answer key modified.

k.

Requal Question Should include EOFD per E-Plan, pg. J-1.

8.05 a & b and j

Replacement Question 8.08 a & b

]

Resolution:

E0FD acceptable and added to answer key.

1.

Requal Question 15 minutes - Arkansas Department of Health j

8.08(b) and Re-I hour - NRC l

placement Question l

8.14(b)

Resolution:

Agree.

Answer key modified.

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4 5.

Exit Meeting Summary At the conclusion of the site visit, the examiners met with utility representatives to discuss the results of the examinations.

The following personnel ~ were present for the exit meeting.

NRC AP&L S. McCrory E. Ewing

' r D.~ - Graves, -

E. Force i.

G.-Harbuck,

L. McClure W. Perks s

Duri' g the meeting, Mr. McCrory informed the attendees that all n

candidates appeared'to be " clear passes" on the oral portion of the examination and that written examination results would be forthcoming as soon as we could grade them.

The utility was also reminded of the recent revision to NUREG 1021, Operator Licensing Examiner Standards.

It was recommended tha+ the revision be reviewed since some major changes had taken place, including the exit meeting.

6.

Requalification Program Evaluation The Requalifications Program Evaluation Report is presented on the next page.

7.

Examination Copies Copies of the R0 and SR0 requalification examinations and answer keys and the SRO Section 8 replacement examination and answer keys are included in this report.

I

f REQUALIFICATION PROGRAM EVALUATION REPORT Facility:

Arkansas Nuclear One Unit 2 Examiner:

D. N. Graves Date(s) of Evaluation:

5/13-15/d6 Areas Evaluated:

XX Written XX Oral Simulator Examination Results:

)

R0 SR0 Total Evaluation Pass / Fail Pass / Fail Pass / Fail (S. M or U) i Written Examination 1/1 7/0 8/1 S

Operating Examination Oral 7/0 7/0 9/0 s

1 Simulator NA NA NA NA Evaluation of facility written examination grading NA Overall Program Evaluation Satisfactory xx Marginal Unsatisfactory (List major defi-ciency areas with brief descriptive comments) i Submitted:

Forwarded:

Approved:

V f/

N v uExaminer Seetion Chief

' Branch Chief

f 1

U.S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR REQUALIFICATION EXAMINATION Facility:

ANO - 2 Reactor Type:

PWR-CE Date Administered:

86/05/15 Examiner:

GRAVES, D.

Examinee: _

INSTRUCTIONS TO EXAMINEE:

THIS EXAMINATION REFLACES THE CURRENT CYCLE FACILITY ADMINISTERED EXAMINATION FOR DEMONSTRATION OF OPERATOR PROFICIENCY AND LICENSE RENEWAL.

FAILURE OF THIS EXAMINATION WILL REQUIRE RETRAINING UNDER THE CURRENT FACILITY REQUALIFICATION TRAINING PROGRAM AND MAY REQUIRE RE-EXAMINATION BY NRC.

Points for each question are indicated in parentheses after the question number.

The passing grade requires at least 70% in each category and a final grade of at least 80%.

Examination papers will be picked up FOUR (4) hours after the examination starts.

% of Category

% of Candidates's Category Value Total Score Value Category 15.0 25.00 1.

Principles of Nuclear l

Power Plant Operations, Fluids, and Thermodynamic 15.0 25.00 2.

Plant Design Including Safety and Emergency Systems 15.0 25.00 3.

Instruments and Controls 15.0 25.00 4.

Procedures - Normal, Abnormal, Emergency, and Radiological Control 60.0 TOTALS Final Grade All work done on this examination is my own.

I have neither given nor roceived aid.

Examinee's Signature

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.1 It__EBINQIELES_QE_SUQLE88_EQWEB_EL6NI_QEEBoIIQNi PAGE 2

IBEBdQQ1Ned101t_dE61_IB8NSEEB_aNQ_ELVIQ_ELQW QUESTION 1.01 (2.00)

Would fuel center line temperature INCREASE, DECREASE, or REMAIN THE SAME in each of the following situations?

o.

Power decreases with constant Tave.

(0.5) i b.

Tave increases with constant power.

(0.5) j c.

Core age increases with constant power.

(0.5) d.

Pressurizer pressure increases with constant power.

(0.5)

QUESTION 1.02 (1.50)

A given change in boron concentration will cause a larger reactivity change at (HIGHER / LOWER) boron concentrations?

Choose the correct answer

)

in parentheses and JUSTIFY OR EXPLAIN why your choice is correct.

(1.5) 1 QUESTION 1.03 (1.50)

The heat output of the reactor can continue to cause the fuel to heatup following a reactor shutdown.

WHAT is(are) the source (s) of this thermal power, and WHY does it NOT indicate on the nuclear instrumentation?

(1.5)

QUESTION 1.04 (3.00)

Reactor power has been at 50% for the past 3 days.

Reactor power is then raised to 100% from 1200 to 1400.

Assuming no operator action, what will be the change in each of the following parameters by 1800 of the same day?

EXPLAIN each answer.

NOTE:

Assume middle of cycle and all control and protective systems era operable and selected normally.

No values are required.

(3.0) a.

Reactor power b.

T avg c.

Pressurizer level 1

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(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)

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Iz__EBINQIELES_QE_NUQLE68_EQasB_EL8MI_QEE86IIQNt PAGE 3

i ISEBdQQ1N651Q1t_BEaI_IB8NSEE8_oND_ELVIQ_ELQW QUESTION 1.05 (1.50)

For the same input of reactivity, one positive and one negative, the response time of the reactor will be different.

State which reactivity insertion will provide a faster response time and explain why the response times for the two reactivity insertions are different.

(1.5)

QUESTION 1.06 (1.5r)

The reactor is operating at 80% steady state.

The operator increases power to 90% by increasing the turbine load.

No other actions are taken.

How will final Tave compere to the initial Tave?

Explain or account for any difference in the two Tave's.

(1.5)

QUESTION 1.07 (2.00) n.

Explain how a loss of heat sink disrupts natural circulation flow.(1.0) b.

Explain how depressurizing too rapidly disrupts natural circulation flow.

(1.0)

QUESTION 1.08 (2.00)

Indicate whether each of the following will INCREASE, DECREASE, or REMAIN UNCHANGED as the discharge valve of a running, motor operated, centrifugal pump is throttled (valve moved in the shut direction):

(2.0) n.

Pump motor amps b.

Pump discharge flow c.

Pump discharge pressure d.

Actual NPSH available at the pump (Assume fluid temperature remains constant) l

(*****

END OF CATEGORY 01 *****)

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2t__ELeNI_QE11GN_INGLUQ1NQ_SeEEI1_eND_EMEBGENQ1_11SIEMS PAGE 4

QUESTION 2.01 (2.00)

The Post Accident Sample System (PASS) can take a hard-piped liquid sample from what four (4) locations?

(2.0) i i

QUESTION 2.07 (2.50) j o.

What are three (3) possible sources of initial excitation to the j

field on a start of the emergency diesel generators?

(1.5) i i

b.

What two (2) normal diesel engine trips are still functional i

following a diesel start due to a loss of offsite power?

(1.0) 1

)

QUESTION 2.03 (2.50) f a.

What are two (2) ways to override or block HPSI flow from the control l

room following receipt of a SIAS?

(2.0) b.

TRUE or FALSE.

With no flow in the HPSI pumps except for suction recirc, the HPSI pumps will not sustain any damage for a minimum of 45 minutes.

(0.5) l I

QUESTION 2.04 (2.00)

The two 4160V ESF buses, 2A3 and 2A4, are capable of being supplied i

from four (4) sources.

What are these sources?

(2.0) l QUESTION 2.05 (2.50) a.

What conditions are necessary to cause en automatic shift of Emergency Feedwater System suction sources?

(1.0) r, i

b.

What are the suction sources involved in the above shift (normal j

and alternate)?

(1.0) i c.

Once an EFAS has been received, what automatic action prevents l

overfilling the steam generator?

(0.5) i i

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(*****

CATEGORY 02 CONTINUED ON NEXT PAGE *****)

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i 2z__ELoNI_RESIGU_INCLUQlNG_S8EEIl_aND_EMEBQENQl_S11IEd3 PAGE 5

i QUESTION 2.06 (2.00) o.

What is the normal mode or lineup of the Steam Dumps' and Dypass Control Valves' Permissive Switches (OFF, MANUAL, or AUT0)?

(1.0) b.

List three (3) locations for SDBCS EMERGENCY OFF pushbuttons.

(1.0)

QUESTION 2.07 (1.50)

What three (3) conditions will cause a valve misalignment alarm for the low temperature overpressure protection system?

Include any applicable temperature or pressure values.

(1.5) 1

(*****

END OF CATEGORY 02 *****)

i 2t__lNSIBudENIS_8NQ_GQUIBQL3 PAGE 6

QUESTION 3.01 (2.00)

What three (3) separate sources of power feed into a typical static inverter power supply?

Include the order of preference for the inputs.

Specific circuit, bus, or panel numbers are NOT required.

(2.0)

QUESTION 3.02 (2.50) c.

Describe and explain the response of the Feedwater Control System to a steam flow detector failing full scale.

Assume no reactor trip occurs.

Specific component response (i.e. Main Feed Valve, Bypass Feedwater Valve position) is not required.

(1.5) b.

What automatic actions should occur directly as a result of a Reactor Trip Override signal?

(1.0)

QUESTION 3.03 (3.00) o.

The Core Protection Calculators (CPC) will generate CEA withdrawal prohibits based on what three (3) pre-trips?

(1.0) b.

The CPC's will generate " auxiliary" trips with no pre-trips.

List five (5) conditions that will cause a CPC auxiliary trip.

INCLUDE SETPOINTS.

(2.0)

QUESTION 3.04 (2.00)

List the methods or interlocks used in the Control Element Drive Control System to limit CEA travel at the top and bottom of the core.

The list should include which CEA's are effected, method of position detection, and approximate setpoint.

(2.0)

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(*****

CATEGORY 03 CONTINUED ON NEXT PAGE *****)

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2t__INSIBudENIS_8NQ_GQNIBQLS PAGE 7

QUESTION 3.05 (3.00)

Match each of the ESF systems below (a -

) with the condition (s) that actuate (s) the ESFAS logic for that ESF system.

Numbered items may be used more than once or not at all.

Include logic (AND or OR) for systems with multiple inputs.

(3.0) n.

CIS 1.

High PZR pressure 6.

High S/G pressure b.

MSIS 2.

Low PZR pressure.

7.

Low S/G pressure c.

CSS 3.

High PZR level 8.

Another ESF system d.

SIS 4.

Low PZR level actuation signal o.

CCS 5.

High containment pressure QUESTION 3.06 (2.50)

]

The low Steam Generator pressure and low PZR pressure trip setpoints in the Reactor Protection System are variable.

a.

The variable setpoints can be reset from what three (3) locations?(1.5) b.

What action is necessary to ensure actuation does not occur during plant cooldown?

(0.5) c.

What action is necessary to raise the setpoints as plant tempera-ture and pressure are increased and why?

(0.5)

(*****

END OF CATEGORY 03 *****)

t e

4 __EBQQEQUBE1_:_UQBd6Lt_6ENQBd6Lt_EdEBGENQ1_6NQ PAGE 8

t B6R10LQQIQ6L_QQUIBQL QUESTION 4.01 (1.50)

What three (3) conditions must be met in order to start up the reactor following a reactor trip and use Plant Operating Procedure 2102.06, Reactor Trip Recovery, instead of Plant Operating Procedure 2102.01, Plant Preheatup and Precritical Checklist?

(1.5)

QUESTION 4.02 (1.50)

During a plant heatup, the operator observes that RCS temperature goes from 350 deg F to 362 deg F over a six minute period.

Explain why this does or does not violate the Technical Specification limit on heatup rate.

(1.5)

QUESTION 4.03 (3.00) n.

During a reactor startup, criticality should be achieved within

________ of the estimated critical position.

(0.5) b.

What actions should be taken, if any, if criticality occurs outside the range specified in part "a" above?

(2.0) c.

How many attempts at achieving criticality within the desired range may be made prior to consulting Nuclear Engineering for recommendations?

(0.5)

QUESTION 4.04 (1.50)

Indicate whether each of the following is TRUE or FALSE:

c.

Any RCP seal stage that has approximately equal pressure on both sides shall be considered failed.

(0,5) i b.

The affected RCP should be tripped if two (2) of its seal stages j

have failed.

(0.5)

)

c.

If component cooling water is lost to the RCP's and cannot be restored within one (1) minute, manually trip the reactor and stop all RCP's.

(0.5) l

(*****

CATEGORY 04 CONTINUED ON NEXT PAGE *****)

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st__EBQQEQUBEl_=_NQBdekt_aBNQBdeLt_EMEBGENQ1_8NQ PAGE 9

BeQIQLQQ1QeL_QQNIBQL I

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QUESTION 4.05 (1.00) f How is a vibration alarm on a RCP determined to be valid?

(1.0)

I

)

QUESTION 4.06 (1.50)

What three (3) conditions will cause a CEA to be declared inoperable for Technical Specification LCO purposes?

(1.5)

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QUESTION 4.07 (3.00)

What immediate actions should be performed if a control room evacuation I

is deemed necessary?

Component numbers are not required.

(3.0) i QUESTION 4.08 (2.00)

I a.

How long may a radiation worker stay in a 60 mrem /hr radiation field before he will exceed an ANO exposure limit?

SHOW how your answer was obtained.

Assume a completed radiation history i

is on file for the radiation worker.

(1.0) i b.

How long (total accumulated time) could the worker work in the i'

60 mrem /hr area before exceeding a 100FR20 exposure limit?

(1.0) i l

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(***** END OF CATEGORY 04 *****)

(************* END OF EXAMINATION ***************)

F NRC LICENSE EXAMINATION HANDOUT EQUATIONS, CONSTANTS. AllD CONVERSIONS 6=m*C*deltaT 6=U*A*deltaT p

P = Po*10sur*(t) p,po e /T SUR = 26/T t

l T=1*/p+(p-p)/Xp T=1/(p-5)

T = ($-p)/Xp p = (Keff-1)/Keff = deltaKeff/Keff p=1*/Reff+jeff/(1+%.T)

A = in2/tg = 0.693/tg X = 0.1 seconds-I I = Io*e "*

~

CR = S/(1-Keff) 2 R/hr = 6*CE/d feet Water Parameters 1 gallon = 8.345 lbm = 3.87 liters 1 ft3 = 7.48 gallons Density 9 STP = 62.4 lbm/ft3 = 1 gm/cm3 Heat of vaporization = 970 Btu /lbm Heat of fusion = 144 Btu /lbm i

1 atmosphere = 14.7 psia = 29.9 inches Hg.

I Miscellaneous Conversions 1 curie = 3.7 x 10iu disintegrations per second i

1 kilogram = 2.21 lb 1 horsepower =254$103 Btu /hr 1 mw = 3.41 x 105 8tu/hr i

1 inch = 2.54 centimeters degrees F = 9/5 degrees C + 32 degrees C = 5/9 (degrees F - 32) 1 Stu = 778 ft-1bf

11__EBINQ1 ELE 1_QE_NVQLE68_EQWEB_EL6NI_QEEBal10Nt PAGE 10 i

IbEBdQQ1Ned1 Gat _bE8I_IB8M1EEB_aND_ELU10_ELQW j

ANSWERS -- ARKANSAS NUCLEAR ONE-2

-86/05/15-GRAVES, D.

1 1

ANSWER 1.01 (2.00) a.

Decrease (0.5) l b.

Increase (0.5) c.

Decrease (0.5) d.

No chango (0.5).

Accept increase if the assumption is stated that increasing pressure decreases nucleate boiling.

REFERENCE Thermal Sciences FND-121, Lesson d-6, pg 15-17 I

J j

ANSWER 1.02 (1.50)

Lower (0.5).

At lower boron concentrations, a small change in boron concentration causes a larger fractional change in the number of thermal neutrons absorbed in the moderator than does the same change in boron j

concentration at higher boron concentrations (1.0).

j REFERENCE ANO2 Plant Specific Reactor Theory, pg 198 l

ANSWER 1.03 (1.50)

This thermal power is mainly from the decay of fission products (0.5).

It does not indicate on the nuclear instrumentation because it consists mainly of alpha, beta, and gamma radiation from the decay of the fission products, which are not detected on the nuclear instrumentation (1.01.

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REFERENCE AN02 Plant Specific Reactor Theory, pg 218 a

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k Iz__EBINQlELE1_QE_ NUCLE 88_EQWEB_EL8HI_QEEB8IIQNt PAGE 11 IBEBdQQYNedlCSt_BE81_IB8NSEEB_6NQ_ELMID_ELQW ANSWERS -- ARKANSAS NUCLEAR ONE-2

-86/05/15-GRAVES, D.

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ANSWER 1.04 (3.00) e.

No Change (0.25).

Reactor power will be a function of steam demand which did not change (0.75).

b.

Increase (0.25).

Xenon burnout will add positive reactivity which must be compensated for by an increase in temperature (0.75).

c.

No change (0.25).

Pressurizer level is at the upper program limit so the increase in temperature will not result in an increase in program level (0.75).

REFERENCE AN02 Plant Specific Reactor Theory, pgs 187, 206 Lesson Plan AA-52002-001, Figure 1.19 l

ANSWER 1.05 (1.50)

The positive reactivity insertion will have a faster response time (0.5).

The average neutron generation time is shorter for the positive reactivity insertion thus providing a faster response time (1.0).

REFERENCE ANO2 Plant Specific Reactor Theory, pg 148 ANSWER 1.06 (1.50)

The final Tave will be lower than the initial Tave (0.5) due to the i

negative reactivity added by the Doppler Coefficient (or Power Defect) on the power increase (1.0).

e REFERENCE AN02 Plant Specific Reactor Theory, pg 221

Iz__EB16CIELES_QE_UUQLE68_EQWEB_ELaNI_QEEB611QUt PAGE 12 IBEBdQQIUed1GSt_ME61_IB86SEEB_600_ELU10_ELQW ANSWERS -- ARKANSAS NUCLEAR ONE-2

-86/05/15-GRAVES, D.

ANSWER 1.07 (2.00) n.

With no heat removed from the primary coolant, no density difference l

will be established, thus no driving force for the fluid and no flow will occur (1.0 for concept).

b.

Natural circulation may be disrupted by violent boiling in the core which disrupts the thermal forces causing the driving force (1.0).

REFERENCE Thermal Sciences, FND-121, lesson 0-3, pg 30 of 36 ANSWER 1.08 (2.00) c.

decrease (0,5) b, decrease (0.5) c.

increase (0.5) d.

increase (0,5)

REFERENCE AP&L Thermodynamics, Fluid Flow and Heat Transfer for Nuclear Power Plants, pgs 149-150 4

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l 2t__EL8NI_ DESIGN _INCLVQINE_S8EEII_6NQ_EdEBGENQ1_SISIEd1 PAGE 13 1

ANSWERS -- ARKANSAS NUCLEAR ONE-2

-86/05/15-GRAVES, D.

l i,

ANSWER 2.01 (2.00)

- RCS hot leg j

- PZR surge line (or water space)

- PZR steam space l

- Containment building sump j

(0.5 each)

)

i REFERENCE j

ANO Procedure 2104.15, Post Accident Sampling System, Rev 3, pg 1 l

]

ANSWER 2.02 (2.50)

I j

a.

Residual magnetism i

125 VDC from the station batteries j

24 VDC from the backup flashing circuit (0.5 each) i b.

Overspeed (0.5) l Low lube oil pressure (0.5)

{

Accept Fail to Start trip also as one of the two required.

REFERENCE j

ANO Procedure 2104.36, Emergency Diesel Generator Operations, Rev 19, 1

pgs 12-13 I

ANSWER 2.03 (2.50) o.

Go to " PULL TO LOCK" on the punp handswitches (1.0) or to "0 PEN" j

then to the desired position on the injection flow control valves (1.0)

{

b.

False (0,5) 30 mirutes REFERENCE

}

ANO Procedure 2104.39, HPSI System Operation, Rev 15, pg 2 j

AN02 Lesson Plan AA-52002-004, Emergency Core Cooling, Rev 2, pg 4 4

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r 2t__EL8NI_QE110N_ INCLUDING _18EEI1_8NQ_EMEBQENQ1_11SIEd1 PAGE 14 ANSWERS -- ARKANSAS NUCLEAR ONE-2

-86/05/15-GRAVES, D.

ANSWER 2.04 (2.00)

- Unit Auxiliary Transformer

- Startup Transformer #2

- Startup Transformer #3

- Emergency Diesel Generators (0.5 each)

REFERENCE AN02 Lesson Plan AA-52002-007, Electrical Distribution, Rev 2, pg 15 ANSWER 2.05 (2.50) a.

EFAS present (0.5) and Emergency Feedwater Pump suction is

<5 psig (0.5).

b.

Normal - Condensate Storage Tank (0.5)

Alternate - Service Water (0.5) c.

Once adequate level has been restored in the steam generator (s),

The EFAS signal will close the EFP's discharge valves (0.5).

REFERENCE AN02 Lesson Plan AA-42002-021, Emergency Feedwater System, Rev 1, pgs 2-4, 16 ANSWER 2.06 (2.001 k

e.

Bypass valves in AUTO (0.5)

Atmospheric dump valves in 0FF (0.5) b.

Control Room (2C-02)

Remote Shutdown Panel (2C-80)

SDBCS Panel (0.33 each)

REFERENCE ANO Procedure 2105.08, Steam Dump and Bypass Control System Operations, j

Rev 3, pg 2, 3

2 1

1 b

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i 2t__EL8MI_QESIGN_INCLUDINQ_S8EEII_880_EMEEGENGI_11SIEMS PAGE 15 i

ANSWERS -- ARKANSAS NUCLEAR ONE-2

-86/05/15-GRAVES, D.

1

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ANSWER 2.07 (1.50)

- Closed during heatup with temperature <275 deg F.

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- Open during heatup or normal plant operations with temperature >280 deg.

Open during cooldown with temperature >275 deg F.

i Closed during cooldown with temperature <270 deg F.

A

{

(3 required at 0.5 each) j REFERENCE

[

AN02 Procedure 2203.12J, LTOP Valve Alignment Incorrect, Rev 8, pg 18 of 74 4

i

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j ANSWERS -- ARKANSAS NUCLEAR ONE-2

-86/05/15-GRAVES, D.

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j ANSWER 3.01 (2.00) i

- 480 VAC (0.5) to inverter is normal pref. erred supply (0.17) 125 VDC (0.5) is the normal backup supply (0.17) 480 VAC from a different supply (0.5) is the alternate supply (0.16)

REFERENCE AN02 Lesson Plan AA-52002-007, Electrical Distribution, Rev 2, pg 25 4

ANSWER 3.02 (2.50) s.

Feed flow will increase as the steam flow signal increases (0.3)

{

due to the steam flow / feed flow error (0.3).

Level increases as l

feed flow increases (0.3).

Once the flow error stops changing, the level error will decrease feed flow (0.3) and level should j

be restored to its original value (0.3).

b.

Feedwater pump speed decreases to minimum (0.33)

Main feedwater control valve shuts (0.33)

Bypass feedwater control valve positions to 5% flow or 11% to 15%

j open (0.33).

REFERENCE

}

ANO2 Lesson Plan AA-52002-015, Feedwater Control System, Rev 2, pgs 3-6 1

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j 21__INSIBudENI1_8NQ_GQNIBQL3 PAGE 17 i

ANSWERS -- ARKANSAS NUCLEAR ONE-2

-86/05/15-GRAVES, D.

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l ANSWER 3.03 (3.00) i s.

- Low DNBR

- High LPD

- High PZR pressure l

(0.33 each)

Internal computer processor faults i

b.

Saturation conditions for the existing PZR pressure Tc > or = 600 deg F Tc < or = 470 dag F PZR pressure < or = 1790 psia j

j PZR pressure

>or= 2360 psia

- ASI < or = -0.50 ASI > or = +0.50 One pin peaking factor < or = 1.28 One pin peaking factor > or =

4.28 Less than 2 RCP's running 1

(5 required at 0.4 each, 0.2 for the parameter and 0.2 for the j

setpoint if applicable)

REFERENCE l

AN02 Lesson Plan AA-52002-024, Core Protection Calculators, Rev 2, pg 14 i

3 ANSWER 3.04 (2.00) l INTLK STPNT POSITION-RODS AFFECTED 1

UEL 150" Reed switch All l

LEL 0.75" Reed switch All I

UGS 146" Pulse counter Full Length UGS 153" Pulse counter Part Length l

LGS 5"

Pulse counter Full Length i

LGS 18" Pulse counter Part Length (24 answers at 0.0833 each) i i

REFERENCE d

AN02 Lesson Plan AA-52002-012, Control Element Drive Control System, Rev 2, pg 3, 13 J

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21__INSIBudENI1_6NQ_GQNIBQL1 PAGE 18 ANSWERS -- ARKANSAS NUCLEAR ONE-2

-86/05/15-GRAVES, D.

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l ANSWER 3.05 (3.00)

)

c.

5 b.

7 j

c.

5 AND 8 d.

2 OR 5 I

o.

2 OR 5 i

(0.3 for each answer, 0.2 for each logic)

REFERENCE AN02 Lesson Plan AA-52002-013, Engineered Safety Feature Actuation System, j

Rev 2, pg 7-11 i

i' ANSWER 3.06 (2.50) a.

Bistable control panel (0.5)

PPS Remote Control Modules (0.5)

)

Remote Shutdown Panel (0.5) 1 9

b.

The RESET buttons.on all 4 PPS Operator Consoles must be depressed for each setpoint (0.5).

c.

None (0.25).

The setpoints automatically follow the system pressures 4

1 as they increase (0.25).

i REFERENCE j

ANO Procedure 2102.10, Plant Shutdown and Cooldown, Rev 10, pg 9 AN02 Lesson Plan AA-52002-006, Reactor Protection System, Rev 2, pg 7-8 i

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dz__EBQCEQQBES_ _NQBd8Lt_8BNQBd8Lt_EMEBGENGl_8NQ PAGE 19 l

88DIQLQQ1G8L_GQNIBQL I

ANSWERS -- ARKANSAS NUCLEAR ONE-2

-86/05/15-GRAVES, D.

f I

ANSWER 4.01 (1.50) i

- The cause of the trip is known and is or shall be corrected (0.5).

- The trip did not result from failure (s) of any safety related system (s) j such that a related reportable occurrence occurred (0.5).

A cooldown is not required (0.5).

j REFERENCE i

ANO Procedure 2102.1, Plant Preheatup and Precritical Checklist, Rev 19, i

pg i and ANO Procedure 2102.06, Reactor Trip Recovery, Rev 9, pg 1 I

j ANSWER 4.02 (1.50)

I j

The actual heatuo rate is 120 deg/hr (0.5), but only over that six j

minute period.

The actual limit is 100 deg F (0.5) over any one hour j

period (0.5).

Thus no limit was exceeded.

l REFERENCE AN02 Technical Specification 3.4.9 I

i i

ANSWER 4.03 (3.00) j o.

0.5% delta k/K (0.5) b.

Insert CEA's (0.5) to at least -2.5% subcritical (0.25) with at i

least -5.5% available shutdown margin (0.25).

Verify boron j

concentration (0.5) and recalculate the ECP (0.5).

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c.

2 (0.5) l REFERENCE f

l ANO Procedure 2102.08, Approach to Criticality, Rev 5, pg 4

)

4 1

ANSWER 4.04 (1.50) t o.

True l

b.

F s~1s e i

c.

False

{

(0.5 each) 1 i

dz__EBQQEQUBE1_:_NQBd8Lt_8HNQBd8Lt_EdEBGENQY_8ND PAGE 20 88DIQLQQ198L_QQNIBQL ANSWERS -- ARKANSAS NUCLEAR ONE-2

-86/05/15-GRAVES, D.

REFERENCE AND AOP 2203.25, RCP Emergencies, Rev 1, pg 1-2 r

ANSWER 4.05 (1.00)

If the vibration alarm clears with the vibration reset pushbutton, then realarms when the pushbutton is released, the alarm is valid (1.0).

REFERENCE ANO AOP 2203.25, RCP Emergencies, Rev 1, pg 3 1

ANSWER 4.06 (1.50)

- CEA is known to be untrippable (0,5)

- CEA is known to be immovable (mechanically bound) (0.5)

>7" deviation from other CEA,s in its group (0.5)

REFERENCE ANO AOP 2203.03, CEA Malfunctions, Rev 3, pg 3 l

ANSWER 4.07 (3.00)

- Manually trip the reactor (0.4)

Secure letdown (0.3)

- Open BAMT gravity feed valves (0.2) and close VCT outlet (0.2)

- Manually initiate EFAS C0.4)

- Trip MFW pumps (0.3)

- Shut MSIV's (0.3)

- Trip:

All four RCP handswitches (0.3) 2A1 to 2A3 tie breaker (0.3) 2A2 to 2A4 tie breaker (0.3) l REFERENCE ANO AOP 2203.14, Alternate Shutdown, Rev 2, pg 7 i

it__EBQQEQUBER_:_NQBd8Lt_eHNQBdeL&_EMEBGENQ1_8NQ PAGE 21 BeQIQLQQIQ8L_QQNIBQL ANSWERS -- ARKANSAS NUCLEAR ONE-2

-86/05/15-GRAVES, D.

J

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ANSWER 4.08 (2.00) a.

300 mrem /wk is the limit the worker would reach first (0.5).

300/60 mrem /hr 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> (0.5).

=

1 b.

3000 mrem /qtr (0.5) 10 times the length of time in part "a" = 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> (0.5) i REFERENCE ANO Red Protection Procedure 1622.011, Exposure Limits and Monitoring j

Techniques, Rev 3, pg 3 1

10CFR20 i

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U.S.

NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR REQUALIFICATION EXAMINATION l

Facility:

ANO - 2 Reactor Type: __

PWR-CE Date Administered:

86/05/15 Examiner:

GRAVES, D.

Examinee:

INSTRUCTIONS TO EXAMINEE:

THIS EXAMINATION REPLACES TLE CURRENT CYCLE FACILITY ADMINISTERED EXAMINATION FOR DEMONSTRATION OF OPERATOR PROFICIENCY AND LICENSE RENEWAL.

FAILURE OF THIS EXAMINATION WILL REQUIRE RETRAINING UNDER THE CURRENT FACILITY REQUALIFICATION TRAINING PROGRAM AND MAY REQUIRE RE-EXAMINATION BY NRC.

Points for each question are indicated in parentheses after the question number.

The passing grade requires at least 70% in each category and a f inal grade of at leeat 80%.

Examination paperc will be picked up FOUR (4) hours after the examination starts.

% of Category

% of Candidates's Category Value Total Score Value-Category 15.0 25.00 5.

Theory of Nuclear Power Plant Operations, Fluids, and Thermodynamics 15.0 25.00 6.

Plant Systems Design, Control and Instruments 15.0 25.00 7.

Procedures - Normal, Abnormal, Emergency, and Radiological Control 15.0 25.00 8.

Administrative Procedures Conditions, and Limits 60.0 TOTALS Final Grade All work done on this examination is my own.

I have neither given nor received aid.

Examinee's Signature u

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Et__IBEQBl_QE_NVQLEaB_EQWEB_ELaNI_QEEBaI10Nt_ELul0S _6NQ PAGE 2

IBEBdQQ1Nad101 QUESTION 5.01 (1.00)

Which of the following statements BEST describes what happens to a fluid as it passes through a venturi?

(1.0) n.

Pressure remains constant, but the velocity increases as the diameter of the venturi decreases.

b.

Pressure increases and velocity decreases as the diameter of the venturi decreases.

c.

Pressure increases and velocity remains constant as the diameter of the venturi increases.

d.

Pressure increases, but the velocity decreases as the diameter of the venturi increases.

i QUESTION 5.02 (3.00)

Would fuel center line temperature INCREASE, DECREASE, or REMAIN THE SAME in each of the following situations?

BRIEFLY EXPLAIN.

a.

Power decreasec with constant Tave.

(.75) b.

Tave increases with constant power.

(.75) c.

Core age increases with constant power.

(.75) d.

Pressurizer pressure increases with constant power.

(.75)

QUESTION 5.03 (3.00)

Reactor power has been at 50% for the past 3 days.

Reactor power is then raised to 100% from 1200 to 1400.

Assuming no operator action, what will be the change in each of the following parameters by 1800 of the same day?

EXPLAIN each answer.

NOTE:

Assume middle of cycle and all control and prot,ective systems are operable and selected normally.

No values are required.

(3.0) n.

Reactor power b.

T avg c.

Pressurizer level

(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)

,.m

,_____m

._m,

Ez__IBEQBl_QE_NyQLE68_EQWEB_EL6NI_QEE8611QNt_ELylD$t_6NQ PAGE 3

IBEBdQQ1Nad1QS QUESTION 5.04 (2.00)

Indicate whether each of the following will INCREASE, DECREASE, or NOT AFFECT the Departure from Nucleate Boiling Ratio r. DNB R).

Assume all other parameters remain constant.

(2.0) a.

Primary coolant temperature decreases b.

Primary coolant pressure decreases c.

Primary coolant flow decreases d.

Reactor power decreases QUESTION 5.05 (2.00) e.

With an initial count rate of 20 cps, enough reactivity is added to increase the count rate to 40 cps.

What would be the effect of adding the same amount of reactivity again?

Explain your answer. (1.0) b.

With an initial power level of 19%, enough reactivity is added to raise reactor power to 20%.

What would be the effect on reactor power if that same amount of reactivity was added again?

Explain your answer.

(1.0)

QUESTION 5.06 (2.00)

Indicate whether each of the following are TRUE or FALSE:

(2.0) n.

Decay heat is primarily from the spontaneous fission of plutonium, americium, and curium, b.

After initially building into the reactor, samarium worth (reactivity) does not vary.

c.

Equilibrium xenon worth (reactivity) at 100% power is less than twice the 50% equilibrium xenon reactivity worth.

d.

Control rod worth increases as the core ages.

(*****

CATEGORY 05 CONTINUED ON NEXT PAGE *****)

L

Et__IBEQBl_QE_NUQLEeB_EQWEB_ELeNI_QEEB6110Nt_ELu1QSt_8NQ PAGE 4

IBEBdQQ1NedIQS QUESTION 5.07 (2.00)

Indicate whether each of the following will INCREASE, DECREASE, or REMAIN UNCHANGED as the discharge valve of a running, motor operated, centrifugal pump is throttled (valve moved in the shut direction):

(2.0) a.

Pump motor amps b.

Pump discharge flow c.

Pump discharge pressure d.

Actual NPSH available at the pump (Assume fluid temperature remains constant)

(***** END OF CATEGORY 05 *****)

Ez__EL6NI_111IEd1_DE11GNt_GQNIBQLt_6NQ_INSIBudENI8IlQN PAGE 5

QUESTION 6.01 (2.00)

What three (3) separate sources of power feed into a typical static inverter power supply?

Include the order of preference for the. inputs.

Specific circuit, bus, or panel numbers are NOT required.

(2.0)

QUESTION. 6.02 (2.50) a.

What conditions are necessary to cause en automatic shift of Emergency Feedwater System suction sources?

(1.0) b.

What are the suction sources involved in the above shift (normal and alternate)?

(1.0) c.

Once an EFAS has been received, what automatic action prevents overfilling the steam generator?

(0.5)

QUESTION 6.03 (1.50)

What three (3) conditions will cause a valve misalignment alarm for the low temperature overpressure protection system?

Include any cpplicable temperature or pressure values.

(1.5)

QUESTION 6.04 (2.50) a.

Describe and explain the response of the Feedwater Control System to a steam flow detector failing full scale.

Assume no reactor trip occurs.

Specific component response (i.e. Main Feed Valve, Bypass Feedwster Valve position) is not required.

(1.5) b.

What automatic actions should occur directly as a result of a Reactor Trip Override signal?

(1.0) i 9

?

)

2

(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)

+

61__EL6NI_SX1IEdS_DE110Nt_QQNIBQLt_6MD_INSIBUMEN18119N PAGE 6

QUESTION 6.05 (3.00) a.

The Core Protection Calculators (CPC) will generate CEA withdrawal prohibits based on what three (3) pre-trips?

(1.0) b.

The CPC's will generate " auxiliary" trips with no pre-trips.

I List five (5) conditions that will cause a CPC auxiliary trip.

INCLUDE SETPOINTS.

(2.0)

QUESTION 6.06 (2.00)

COLSS calculates reactor power using three methods.

List the three methods, including when each of the three would be selected as the plant power for use by COLSS.

(2.0)

QUESTION 6.07 (1.50)

The pressurizer level transmitter selectea for level control is ined-vertently isolated at the transmitter.

The variable leg (low pressure) 4 side of the trnnsmitter is depressurized.

What automatic actions will occur as a direct result of this indicated level change?

Assume the reactor is operating at 100% and actual pressurizer level equals programmed pressurizer level initially.

Exclude alarms.

(1.5) t l

I r

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(***** END OF CATEGORY 06 *****)

i

r Zt__BBQQEQUBE1_:_NQBd8Lt_6BNQBd8Lt_EMEBGENQ1_6NQ PAGE 7

88Q10LQQ1 gal _QQNIBQL QUESTION 7.01 (1.50)

What three (3) conditions must be met in order to start up the reactor following a reactor trip and use Plant Operating Procedure 2102.06, Reactor Trip Recovery, instead of Plant Operating Procedure 2102.01, Plant Preheatup and Precritical Checklist?

(1.5)

QUESTION 7.02 (3.00) a.

During a reactor startup, criticality should be achieved within of the estimated critical position.

(0.5) b.

What actions should be taken, if any, if criticality occurs outside the range specified in part "a" above?

(2.0) c.

How many attempts at achieving criticality within the desired range may be made prior to consulting Nuclear Engineering for recommendations?

(0.5)

QUESTION 7.03 (1.50)

Indicate whether each of the following is TRUE or FALSE:

a.

Any RCP seal stage that has approximately equal pressure on both sides shall be considered failed.

(0.5) b.

The affected RCP should be tripped if two (2) of its seal stages have failed.

(0.5) c.

If component cooling water is lost to the RCP's and cannot be restored within one (1) minute, manually trip the reactor and stop all RCP's.

(0.5)

QUESTION 7.04 (1.00)

How is a vibration alarm on a RCP determined to be valid?

(1,0)

(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)

F.

Zr__88QQEDM8El_:_dQ808Lt_8ENQ8d8kA_EdE8QENQ1_8NQ PAGE 8

88DIQLQQ1Q8L_QQNIBQL QUESTION 7.05 (1.50)

What three (3) conditions will cause a CEA to be declared inoperable f or Technical Specification LCO purposes?

(1.5)

QUESTION 7.06 (3.00)

What immediate actions should be performed if a control room evacuation is deemed necessary?

Component numbers are not required.

(3.0)

QUESTION 7.07 (2.00) a.

What observation would be indicative of voiding in the reactor head while performing a natural circulation cooldown?

(1.0) b.

Briefly describe the method by which the bubble in the head region would be collapsed.

Specific procedure steps are not required for full credit.

Note the indication that would verify collapse of the bubble.

QUESTION 7.08 (1.50) n.

What two items or actions raust be verified or performed if cocked CEA protection is not established during a plant heetup?

(1.0) b.

Once cocked CEA protection is established, the remaining CEA's are withdrawn j ust enough to clear the LEL lamps.

This is performed primarily to (CHOOSE THE CORRECT ANSWER):

(0.5) 1.

Verify the CEA indication system is operable.

2.

Allow for thermal expansion of the CEA's.

3.

Verify the CEA drive system is functional for all groups.

4.

Places the CEA's in an area where reactivity response is more significant (CEA's have more " bite").

(*****

END OF CATEGORY 07 *****)

~

Sz__aQUINISIBoIIVE_EBQQEQUBESt_QQNDII10NSt_eNQ_LldlI6110NS PAGE 9

QUESTION 8.01

(.50)

Fill in the blank.

In accordance with 10 CFR 55, "if a licensee has not been actively performing the functions of an operator or senior operator for a period of..____ months or longer, he shall, prior to resuming activities licensed pursuant to this part, demonstrate to the Commission that his knowledge and understanding of facility operation and administration are satisfactory."

(0.5)

QUESTION 8.02 (2.00) a.

HOLD card tagging shall not result in intentionally entering an ACTION statement of an LCO without the prior approval of whom, in addition to the shift supervisor (TWO REQUIRED)?

(1.J) b.

An operator is preparing a tagout that affects the operation of both units.

Describe how authorization is obtained to hang the tags.

Include documentation of the authorization.

(1.0) l 4

i QUESTION 8.03 (2.00)

The Shift Supervisor has specific authority to order power reduction i

or shutdown if continued operation of the unit will result in one of four (4) undesirable conditions or events.

What are these four

]

conditions or events?

(2.0)

QUESTION 8.04 (2.00)

Describe the minimum Fire Brigade manning at ANO.

Include which assignments are made by the Unit 2 Shift Supervisor at the beginning of each shift.

i l

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1

(

(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)

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Hz__6DdlNISIB8IIVE_EBQQEQUBEft_GQNQlIl0 Nit _6NQ_L1dII6IIQNS PAGE 10 QUESTION 8.05 (3.00)

For each of the following types of evacuations, describe the area involved and indicate who is responsible for declaring an evacuation of that area.

(3.0) l o.

Plant b.

Exclusion Area c.

Area QUESTION 8.06 (2.00)

The plant is operating when a CEA drops into the core.

a.

How low may temperature drop before the LCO on minimum temperature for criticality is exceeded?

(0.5) b.

Which temperature indication is referenced in the LCO?

(0.5) i c.

How long does the operator have to restore temperature to an allowable value?

(0.5) i d.

What action must be taken per the Technical Specifications if j

temperature is not restored within the allowable time?

(0,5) i QUESTION 8.07 (2.00)

Indicate whether each of the following are TRUE or FALSE:

(2.0) e.

Control room operators involved in fuel handling shall be limited to a maximum of 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shifts with no limit on the number of days this can continue.

b.

To perform core alterations, at least two neutron flux monitors must have continuous visual and audible indication in the control room.

I c.

Core alterations will be carried out only by licensed operators, or those DIRECTLY supervised by licensed operators.

4 d.

Refueling can commence if the reactor has been shutdown for greater than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

I

(*****

CATEGORY 08 CONTINUED ON NEXT PAGE *****)

at__8Dd1NISIB611YE_EBQQEQUBES _QQNQ111QN$t_6NQ_LldlI611QNS PAGE 11 t

QUESTION 8.08 (1.50)

A licensee may take reasonable action that departs from a condition or a Technical Specification in an emergency when this action is immediately needed to protect the public health and safety and no action consistent with license conditions and Technical-Specifications that can provide adequate or equivalent protection is immediately opparent.

i a.

Such an action shall be approved, as a minimum, by whom?

(0.5) b.

What two (2) notifications must be made within 15 minutes and one hour, respectively?

(1.0) i 4

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(*****

END OF CATEGORY 08 *****)

(*************

END OF EXAMINATION ***************)

i t

NRC LICENSE EXNtIHATION HNIDOUT EQUATICIIS, C0!!STNiTS, N!D C0!!VERSI0!tS l

6 = nI*C *deltaT 6=U*A*deltaT p

I P = Po*10sur*(t)

P = P *e /T SUR = 26/T t

I T=1*/p+(p-p)/Xp T=1/(p-p)

T = (p-p)/X p p = (Keff-1)/Keff = deltaKeff/Keff p = 1*/1Keff+jeff/(1+I.T)

A = in2/tg = 0.693/tg X = 0.1 seconds-1 I = Io*e-"*

CR = S/(1-Keff) 2 R/hr = 6*CE/d feet Water Parameters 1 gallon = 8.345 lbm = 3.87 liters 1 ft3 = 7.48 gallons Density @ STP = 62.4 lb /ft3 = 1 gm/cm3 m

Heat of vaporization = 970 Btu /lbm Heat of fusion = 144 Btu /lbm 1 atmosphere = 14.7 psia = 29.9 inches Hg.

Miscellaneous Conversions I curie = 3.7 x 101U disintegrations per second 1 kilogram = 2.21 lb 1 horsepower =2.54$103 Btu /hr 1 mw = 3.41 x 106 Btu /hr 1 inch = 2.54 centimeters degrees F = 9/5 degrees C + 32 degrees C = 5/9 (degrees F - 32) 1 Stu = 778 ft-lbf

St__IBEQBl_QE_NUQLE88_EQWEB_EL6NI_QEEBoIIONt_ELVIDSt_8NQ PAGE 12 IBEBdQQ1Nad103 ANSWERS -- ARKANSAS NUCLEAR ONE-2

-86/05/15-GRAVES, D.

ANSWER 5.01 (1.00) d (1.0)

REFERENCE AP&L Thermodynamics, Fluid Flow and Heat Transfer for Nuclear Power Plants, pg 146 ANSWER 5.02 (3.00) a.

Decrease (0.25), smaller delta T required to transfer more energy from RCS (0,5).

b.

Increase (0.25), center line temperature responds to RCS temperature in order to maintain constant delta T across cladding (0.5).

c.

Decrease (0.25), fuel swelling and clad creep reduce clad gap which reduces delta T across gap and lowers center line temp (0.5).

d.

No change (0.25), pressure has little effect on heat transfer in subcooled fluids (0.5).

Accept increase if the assumption is stated that increasing pressure decreases nucleate boiling.

REFERENCE Thermal Sciences FND-121, Lesson d-6, pg 15-17 ANSWER 5.03 (3.00) c.

No Change (0.25).

Reactor power will be a function of uteam demand which did not change (0.75).

i b.

Increase (0.25).

Xenon burnout will add positive reactivity which must be compensated for by an increase in temperature (0.75).

c.

No change (0.25).

Pressurizer level is at the upper program limit so the increase in temperature will not result in an increase in program level (0.75).

REFERENCE AN02 Plant Specific Reactor Theory, pgs 187, 206 Lesson Plan AA-52002-001, Figure 1.19 2

e

-... m

_ _ _.. _... _,.. ~.. _ _,.

St__IBEQBX_QE_NUGLE88_EQWEB_ELaNI_QEEB8110Nt_ELUIDSt_aNQ PAGE 13 IBEBMQQ1NadICS ANSWERS -- ARKANSAS NUCLEAR ONE-2

-86/05/15-GRAVES, D.

i n

ANSWER 5.04 (2.00) a.

Increase (0.5) b.

Decrease (0.5) c.

Decrease (0,5) d.

Increase (0.5)

REFERENCE Thermal Sciences, Lesson D-6, pgs 14-15 I

ANSWER 5.05 (2.00) a.

The reactor would be supercritical (0.5).

Doubling the count rate approximately halves the reactivity required to reach criticality (0.5).

4 b.

Power would increase to 30% (0.5).

Power defect is linear (0.5).

1 REFERENCE i

ANO2 Plant Specific Reactor Theory, pg 140-141, 194 ANSWER 5.06 (2.00) a.

False (0.5) b.

False (0.5) c.

True (0.5) d.

True (0.5)

REFERENCE AN02 Plant Specific Reactor Theory, pgs 195, 199-210, 218 ANSWER 5.07 (2.00)

\\

I c.

dec re ase (0.5) l b.

decrease (0.5) c.

increase (0.5) i d.

increase (0.5)

Sz__IBEQBl_QE_NVQLE6B_EQWEB_EL6NI_QEEB6IIQNt_ELUIQSt_6NQ PAGE 14 IBEBdQQ1NedIQS ANSWERS -- ARKANSAS NUCLEAR ONE-2

-86/05/15-GRAVES, D.

REFERENCE AP&L Thermodynamics, Fluid Flow and Heat Transfer for Nuclear Power Plants, pgs 149-150 1

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ht__PL6HI_1111Ed3_QE11GNt_QQNIBQLt_8NQ_INSIBVMENI6IlQN PAGE 15 ANSWERS -- ARKANSAS NUCLEAR ONE-2

-86/05/15-GRAVES, D.

ANSWER 6.01 (2.00)

- 480 VAC (0.5) to inverter is normal preferred supply (0.17)

- 125 VDC (0.5) is the normal backup supply (0.17)

- 480 VAC from a different supply (0.5) is the alternate supply (C.16)

REFERENCE ANO2 Lesson Plan AA-52002-007, Electrical Distribution, Rev 2, pg 25 ANSWER 6.02 (2.50) a.

EFAS present (0.5) and Emergency Feedwater Pump suction is

<5 psig (0.5).

b.

Normal - Condensate Storage Tank (0,5)

Alternate - Service Water (0.5) c.

Once adequate level has been restored in the steam generator (s),

The EFAS signal will close the EFP's discharge valves (0.5).

REFERENCE AN02 Lesson Plan AA-42002-021, Emergency Feedwater System, Rev 1, pgs 2-4, 16 ANSWER 6.03 (1.50)

- Closed during heatup with temperature <275 deg F.

- Open during heatup or normal operations with temperature >280 deg F.

- Open during cooldown with temperature >275 deg F.

Closed during cooldown with temperature <270 deg F.

(3 required at 0.5 each)

REFERENCE AN02 Procedure 2203.12J, LTOP Valve Alignment Incorrect, Rev 8, pg 18 of 74 1

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kz__EL6HI_1111Ed1_DE11GNt_QQNIBQLt_6NQ_IN11BudENI6IIQN PAGE 16 i

i ANSWERS -- ARKANSAS NUCLEAR ONE-2

-86/05/15-GRAVES, D.

}

ANSWER 6.04 (2.50) 1

)

a.

Feed flow will increase as the steam flow signal increases (0.3) d due to the steam flow / feed flow error (0.3).

Level increases an j

feed flow increasen (0.3).

Once the flow error stops _ changing, j

the level error will decrease feed flow (0.3) and level should j

be restored to its original value (0.3).

1.

b.

Feedwater pump speed decreases to minimum (0.33)

Main feedwater control valve shuts (0.33) i Bypass feedwater control valve positions to 5% flow or 11% to 15%

open (0.33).

I i

REFERENCE AN02 Lesson Plan AA-52002-015,. Feedwater Control System, Rev 2, pgs 3-6 ANSWER 6.05 (3.00) 1 a.

- Low DNBR l

- High LPD j

- High PZR pressure

]

(0.33 each) i b.

- Internal computer processor faults

- Saturation conditions for the existing PZR pressure l

- Tc > or = 600 deg F l

- Tc < or = 470 deg F

- PZR pressure < or =

1790 psia

- PZR pressure > or = 2360 psia

- ASI < or = -0.50 l

- ASI > or = +0.50 One pin peaking factor < or = 1.28 i

- One pin peaking factor > or = 4.28 4

- Less than 2 RCP's running j

(5 required at 0.4 each, 0.2 for the parameter and 0.2 for the setpoint if applicable)

}

REFERENCE c

}

AN02 Lesson Plan AA-52002-024, Core Protection Calculators, Rev 2, pg 14 i

i 4

}

}

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_ ~ _ -

l Et__EL6NI_SISIEdS_DESIGNt_GQNIBQLt_6NQ_INSIBudENI811QN PbGE 17 i

ANSWERS -- ARKANSAS NUCLEAR ONE-2

-86/05/15-GRAVES, D.

ANSWER 6.06 (2.00)

Primary calorimetric (0.4) is used below 15% rated thermal power (0.4).

When > 15% rated thermal power (0.4), the higher of secondary calori-metric power (0.4) and calibrated turbine power (0.4) is used.

REFERENCE ANO Procedure 2105.13, COLSS Operations, Rev 8, pg 4-5 ANSWER 6.07 (1.50)

The letdown flow control valve (s) go to minimum position (0.5).

The two backup charging pumps start (0.5).

All pressurizer heaters deneergize.(0.5).

REFERENCE l

ANO Operating Procedure 2103.05, PZR Operations, Rev 6, pgs 5, 10-11 Thermal Sciences FND-21,, Lesson D-7, pg 20-21 I

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I 7 __E8QQEQyBEg_ _NQBd6(t_6BNQ8dA(t_EdEBQENQ1_6NQ PAGE 18 t

BaQ10LQQ10aL_QQUIBQL ANSWERS -- ARKANSAS NUCLEAR ONE-2

-86/05/15-GRAVES, D.

t 1

ANSWER 7.01 (1.50)

- The cause of the trip is known and is or shall be corrected (0.5).

- The trip did not result from failure (s) of any safety related system (s) such that a related reportable occurrence occurred (0.5).

- A cooldown is not required (0.5).

REFERENCE ANO Procedure 2102.1, Plant Preheatup and Precritical Checklist, Rev 19, pg 1 and ANO Procedure 2102.06, Reactor Trip Recovery, Rev 9, og 1 ANSWER 7.02 (3.00) o.

0.5% delta k/K (0.5) b.

Insert CEA's (0,5) to at least -2.5% suberitical (0.25) with at least -5.5% available shutdown margin (0.25).

Verify boron concentration (0.5) and recalculate the ECP (0.5).

I c.

2 (0.5)

REFERENCE j

ANO Procedure 2102.08, Approach to Criticality, Rev 5, pg 4 4

d ANSWER 7.03 (1.50)

I a.

True b.

False j

c.

False C0.5 each)

REFERENCE j

ANO AOP 2203.25, RCP Emergencies, Rev 1, pg 1-2 1

ANSWER 7.04 (1.00) 4 If the vibration alarm clears with the vibration reset pushbutton, then realarms when the pushbutton is released, the alarm is valid (1.0).

Zt__EBQQEQUBES_:_NQBdelt_8BNQBdokt_EMEBGEUQ1_66Q PAGE 19 BSDIQLQQIQaL_QQNIBQL ANSWERS -- ARKANSAS NUCLEAR ONE-2

-86/05/15-GRAVES, D.

REFERENCE ANO AOP 2203.25, RCP Emergencies, Rev 1, pg 3 ANSWER 7.05 (1.50)

- CEA is known to be untrippable (0.5)

- CEA is known to be immovable (mechanically bound) (0.5)

>7" deviation from other CEA,s in its group (0.5)

}

REFERENCE AND AOP 2203.03, CEA Malfunctions, Rev 3, pg 3 ANSWER 7.06 (3.00) k

- Manually trip the reactor (0.4) 3

- Secure letdown (0.3)

- Open BAMT gravity feed valves (0.2) and close VCT outlet (0.2)

- Manually initiate EFAS (0.4)

- Trip MFW pumps (0.3)

- Shut MSIV's (0.3)

- Trip:

All four RCP handswitches (0.3) 2Al to 2A3 tie breaker (0.3) 2A2 to 2A4 tie breaker (0.3)

REFERENCE ANO AOP 2203.14, Alternate Shutdown, Rev 2, pg 7 ANSWER 7.07 (2.00) l a.

A rapid pressurizer level increase while depressurizing (1,0).

l b.

Increase pressurizer pressure by energizing the heaters (0.5).

An outsurge from the pressurizer, or level decrease, will occur as the l

bubble collapses (0,5).

)

REFERENCE ANO AOP 2203.13, Natural Circulation Cooldown, Rev 4, pg 2-3

I i

Zt__EBQQEQUBER_:_NQBdeL&_8HNQBdekt_EMEBQENCY_6NQ PAGE 20 B6D19LQQ1 gel _CQNIBQL ANSWERS -- ARKANSAS NUCLEAR ONE-2

-86/05/15-GRAVES, D.

ANSWER 7.08 (1.50) 4 a.

Verify at least one boron dilution monitor is operable (0.5) and leave the trip circuit breakers open (0.5) until the heatup is j

completed.

b.

2 (0.5)

REFERENCE ANO Procedure 2102.02, Plant Startup, Rev 17, pg 9, 10 i

l i

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t I

1 4

i 4,

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Az__8DMIN11IB811VE_EBQGEQMBES&_CQNDIll0 Nit _8NQ LidII6110NS PAGE 21 ANSWERS -- ARKANSAS ~ NUCLEAR ONE-2

-86/05/15-GRAVES, D.

ANSWER 8.01

(.50)

Four (0.5)

REFERENCE 10 CFR 55.31 ANSWER 8.02 (2.00) a.

The affected unit's Operations Superintendent (0.5) or the Operations Maneger (0.5).

b.

The Shift Supervisor on the initiating unit must obtain concurrence from the opposite Shift Supervisor (0.5).

Documentation of such concurrence shall be reflected by the opposite Shift Supervisor's initials on the Hold Card Authorization Sheet next to the authorizing signature (0,5).

REFERENCE ANO Administrative Procedure 1000.27, Hold and Caution Card Control, Rev 6, pg 10 ANSWER 8.03 (2.00)

_ Immediate equipment damage (0.5)

- Danger to station personnel (0.5)

- Violation of Operating Licenses or Technical Specification requirements (0.5)

- Unnecessary automatic trip (0.5)

REFERENCE ANO Procedure 1015.01, Conduct of Operation, Rev 23, pg 7 ANSWER 8.04 (2.00)

A Fire Brigade Leader assigned by the SS at the beginning of the shift (0.5).

A Fire Brigade Member assigned by the SS at the i

beginning of the shift (0.5).

Two Fire Brigade Support Personnel (0,5) l from the Security Force.

A Fire Brigade Member from the opposite unit (0.5).

l Az__8DMINISIB8I1YE_EBQQEQUBESt_QQNDIl10NSt_6NQ_ lid 1IeIIQNS PAGE 22 i

ANSWERS.-- ARKANSAS NUCLEAR ONE-2

-86/05/15-GRAVES, D.

REFERENCE ANO Procedure 1015.07, Fire Brigade Organizstion and Responsibilities, Rev 6, pg 1 ANSWER 8.05 (3.00) s.

Plant:

Evacuates the area encompassed by physical barriers and to which access is controlled (0.5).

It is declared by the affected unit's Shift Operating Supervisor (0.17),

Emergency Coordinator (0.17), or the EOF 0 (0.16).

i b.

Exclusion Area:

Evacuates the area surrounding ANO within a minimum radius of 0.65 miles of the reactor buildings, but outside the protected area (security fence) and controlled to the extent necessary by AP&L during periods of emergency (0.5).

The affected unit's i

Shift Operating Suparvisor (0.17), the Emergency l

Coordinator (0.17), or the EOFD (0.16) is responsible l

for declaring this evacuation.

c.

Area:

Evacuates that area outside the Exclusion Area (0.5) and is declared by the Arkansas Department of Health (0.5).

)

l REFERENCE i

ANO EPIP 1903.30, Evacuation, Rev 10, pg 1-2 ANO Emergency Plan, pg J-1

)

ANSWER 8.06 (2.00) a.

Temperature must be

>or= to 525 deg F when critical (0.5).

b.

The lowest reading Tave (0.5).

c.

15 minutes (0.5) d.

If temperature is not restored within 15 minutes, the reactor must be in HOT STANDBY (shutdown the reactor) within the next 15 minutes (0.5).

REFERENCE ANO2 Technical Specifications, 3.1.1.5 I

i

Rz__8QUINISIB6IlyE_E8QQEQQBEft_QQNQlIIQN1t_6NQ_(Id118IlQNg PAGE 23 ANSWERS -- ARKANSAS NUCLEAR ONE-2

-86/05/15-GRAVES, D.

4 l

ANSWER 8.07 (2.00)

J a.

False b.

False C.

True d.

False (0.5 each)

REFERENCE ANO Procedure 2502.01, Refueling Shuffle, Rev 13, pg 3-7 ANO Procedure 1015.01, Conduct of Operations, Rev 23, pg 27 ANO2 Technical Specifications 3.9.3.a ANSWER 8.08 (1.50) a.

The action shall be approved by a licensed Senior Operator as a minimum (0.5).

b.

Arkansas Department of Health (0.5)

NRC (0.5) Operations Center REFERENCE i

ANO Procedure 2203.14, Alternate Shutdown, Rev 1, pg 5 j

ANO EPIP 1903.10, Emergency Action Level Response / Notification, Rev 18, pg 12 10 CFR 50.54.x and y i

i 1

I i

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1 a

4 1

U.

S.

NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY:

_88K8N383_NQQLE88_QNE:2__

REACTOR TYPE:

_EWB:QE__________________

DATE ADMINISTERED: _g6/05/15________________

EXAMINER:

_QB8VEft_Qz______________

APPLICANT:

INSIBUQ110U1_IQ_8EELIQ8 nil Une separate paper for the answers.

Write answers on one side only.

Staple question sheet on top of the answer sheets.

Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in the category.

Examination papers will bo picked up one and one half (1.5) hours after the examination 1

i starts.

i l

% OF CATEGORY

% OF APPLICANT'S CATEGORY

__Y8LUE_ _IQI8L

___SQQBE___

_Y8LUE__ ______________Q81EQQBl_____________

_251DQ__ 1Q0z00

________ 8.

ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS

_25tQQ__ 1QQtQQ

________ TOTALS FINAL GRADE _________________%

All work done on this examination is my own. I have neither givon nor received aid.

APPLICANT'S SIGNATURE l

i

az__eDd1NISIBoIIVE_EBQGEQUBESt_GQUQ1IIQNSt_8ND_ lid 1IeI1QUS PAGE 2

QUESTION 8.01

(.50)

Fill in the blank.

In accordance with 10 CFR 55, "if a licensee has not been actively performing the functions of an operator or senior operator for a period of _____ months or longer, he shall, prior to resuming activities licensed pursuant to this part, demonstrate to the Commission that his knowledge and understanding of facility operation and administration are satisfactory."

(0.5)

QUESTION 8.02 (2.00) a.

HOLD card tagging shall not result in intentionally entering an ACTION statement of an LCO without the prior approval of whom, in addition to the shift supervisor (TWO REQUIRED)?

(1.0) b.

An operator is preparing a tegout that affects the operation of both units.

Describe how authorization is obtained to hang the tags.

Include documentation of the authorization.

(1.0)

QUESTION 8.03 (1.50)

What are the three (3) basic types of hazardous atmospheric environments / conditions that may be encountered at AN0?

(1.5)

QUESTION 8.04 (2.50) e.

The Shift Technical Advisor shall be available to the control room within _________of call by the Shift Supervisor.

(0.5) b.

The STA should be summoned by the Shift Supervisor upon detection of a " plant transient".

What events fall under reactor transients and what events fall under secondary system transients?

(2.0)

QUESTION 8.05 (2.00)

The Shift Supervisor has specific authority to order power reduction or shutdown if continued operation of the unit will result in one of four (4) undesirable conditions or events.

What are these four conditions or events?

(2.0)

(*****

CATEGORY 08 CONTINUED ON NEXT PAGE *****)

Az__8QUINISIBoIIVE_EBQQEQUBEnt_QQUQlIIQNAt_eUQ_ lid 118Il0NS PAGE 3

QUESTION 8.06 (2.00)

What are four (4) types / classes of keys (by general function) that cre required to be administrative 1y controlled'r DO NOT LIST SPECIFIC KEYS.

(2.0)

QUESTION 8.07 (2.00)

Describe the minimum Fire Brigade manning at ANO.

Include which cssignments are made by the Unit 2 Shift Supervisor at the beginning of each shift.

QUESTION 8.08 (3.00)

For each of the following types of evacuations, describe the area involved and indicate who is responsible for declaring an evacuation of that area.

(3.0) a.

Plant b.

Exclusion Area c.

Area QUESTION 8.09 (2.00)

The plant is operating when a CEA drops into the core.

a.

How low may temperature drop before the LCO on minimum temperature for criticality is exceeded?

(0,5) b.

Which temperature indication is referenced in the LCO?

C0.5) c.

How lorig does the operator have to restore temperature to an allowable value?

(0.5) d.

What action must be taken per the Technical Specifications if temperature is not restored within the allowable time?

(0.5) l l

(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)

l

i Hz__6DdlNISIB8I1YE_PRQQEQQBESt_QQUQ111QNSt_6NQ_LIdlI611QNS PAGE 4

QUESTION 8.10 (1.50)

L!st three (3) general requirements (not authorizations) for entry into the reactor building while at power (greater than 10 EE-1%)

per OP 1622.009, Requirements for Reactor Building Power Entries.

(1.5)

QUESTION 8.11 (1.50)

What are three (3) checks made to demonstrate that the RCS leakages are within the Technical Specification limits?

(1.5)

QUESTION 8.12 (2.00)

Indicate whether each of the following are TRUE or FALSE:

(2.0) a.

Control room operators involved in fuel handling shall be limited to a maximum of 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shifts with no limit on the number of days this can continue.

b.

To perform core alterations, at least two neutron flux monitors must have continuous visual and audible indication in the control room.

c.

Core alterations will be carried out only by licensed operators, or those DIRECTLY supervised by licensed operators.

d.

Refueling can commence if the reactor has been shutdown for greater than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

Q'ESTION 8.13 (1.00)

U List four conditions that would be classified as " serious inj uries" per EPIP 1903.23, Personnel Emergency.

(1.0)

(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)

Az__eQUINISIBoIIVE_EBQQEQUBESt_QQNQ1Il0N1t_8NQ_L1dII8IIQNS PAGE' 5

QUESTION 8.14 (1.50)

A licensee may take reasonable action that departs from a condition or a Technical Specification in an emergency when this action is immediately needed to protect the public health and safety and no action consistent with license conditions and Technical Specifications that can provide adequate or equivalent protection is immediately apparent.

a.

Such an action shall be approved, as a minimum, by whom?

(0.5) i b.

What two (2) notifications must be made within 15 minutes and one hour, respectively?

(1.0) i t

l

(***** END OF CATEGORY 08 *****)

(************* END OF EXAMINATION ***************)

i i

8.

_aQMIN1118611ME_EBQQEQUBEnt_QQNQ1110NSt_6NQ_ Lid 1I8I1QNS PAGE 6

ANSWERS -- ARKANSAS NUCLEAR ONE-2

-86/05/15-GRAVES, D.

ANSWER 8.01

(.50)

Four (0.5)

REFERENCE 10 CFR 55.31 ANSWER 8.02 (2.00) o.

The affected unit's Operations Superintendent (0,5) or the Operations Manager (0.5).

1 b.

The Shift Supervisor on the initiating unit must obtain concurrence from the opposite Shift Supervisor (0.5).

Documentation of such concurrence shall be reflected by the opposite Shift Supervisor's initials on the Hold Card Authorization Sheet nextEto the authorizing s ig n at u r e (0.5).

REFERENCE l

ANO Administrative Procedure 1000.27, Hold and Caution Card Control, Rev 6, pg 10 ANSWER 8.03 (1.50)

- Oxygen deficient atmosphere (0.5)

- Toxic atmosphere (0.5)

- Airborne radioactive contamination (0.5)

REFERENCE AND Procedure 1609.002, Respiratory Hazards, Rev 2, pg 2-3 ANSWER 8.04 (2.50) a.

10 minutes (0.5) b.

A reactor transient is a power change greater than 10% full power in less than one minute (0.4), any unplanned reactor trip-(0.4),

or a heatup or cooldown in excess of 100 deg F in one hour (0.4).

A secondary system transient is a power change in excess of 100 MWe in less than one minute (0.4), or any unplanned main turbine (0.2) or main feedwater pump trip (0.2).

St__aQNIN1118611YE_EBQQEQUBEnt_QQNDIIIQNat_eND_LINIIal10NS PAGE 7

ANSWERS -- ARKANSAS NUCLEAR ONE-2

-86/05/15-GRAVES, D.

REFERENCE ANO Procedure 1002.01, STA Duties and Responsibilities, Rev 11, pg 2 ANSWER 8.05 (2.00)

- Immediate equipment damage (0.5)

- Danger to station personnel (0.5)

- Violation of Operating Licenses or Technical Specification requirements (0.5)

- Unnecessary automatic trip (0.5)

REFERENCE ANO Procedure 1015.01, Conduct of Operation, Rev 23, pg 7 ANSWER 8.06 (2.00) l

- Safety interlock and bypass keys

- Equipment cabinet and console keys

- Key operated valve / damper handswitches

- Reactor building / containment personnel hatch key

- Manual valve padlock keys

- High radiation area access keys (4 required at 0.5 each)

REFERENCE ANO Procedure 1015.05, Shift Supervisor Key Control, Rev 4, pg i ANSWER 8.07 (2.00)

A Fire Brigade Leader assigned by the SS at the beginning of the shift (0.5).

A Fire Brigade Member assigned by the SS at the beginning of the shift (0.5).

Two Fire Brigade Support Personnel (0.5) 9 rom the Security Force.

A F i r e B r ig a d e Mer.ib e r from the opposite unit (0,5).

REFERENCE ANO Procedure 1015.07, Fire Brigade Organization and Responsibilities, Rev 6, pg 1 l

1 l

r

~

a 8z__aQUINISIBallVE_EBQQEQUBESt_QQUQ1Il0NSt_aNQ_ lid 118110NS PAGE 8

ANSWERS -- ARKANSAS NUCLEAR ONE-2

-86/05/15-GRAVES, D.

ANSWER 8.03 (3.00) a.

Plant:

Evacuates the area encompassed by physical barriers and to which access is controlled (0.5).

It is declared by the affected unit's Shift Operating Supervisor (0.17),

the Emergency Coordinator (0.17), or the EOFD (0.16).

b.

Exclusion Area:

Evacuates the area surrounding ANO within a minimum radius of 0.65 miles of the reactor buildings, but outside the protected area (security fence) and controlled to the extent necessary by AP&L during periods of emergency (0.5).

The affected unit's Shift Operating Supervisor (0.17), the Emergency Coordinator (0.17), or the EOFD (0.16) is responsible for declaring this evacuation.

c.

Area:

Evacuates the area outside the Exclusion Area (0.5) and is declared by the Arkansas Department of Health (0.5).

REFERENCE ANO EPIP 1903.30, Evacuation, Rev 10, pg 1-2 ANG Emergency Plan, pg J-1 ANSWER 8.09 (2.00) e.

Temperature must be > or =

to 525 deg F when critical (0.5).

b.

The lowest reading Teve (0.5).

c.

15 minutes (0.5) d.

If temperature is not restored within 15 minutes, the reactor must-be in HOT STANDBY (shutdown the reactor) within the next 15 minutes (0.5).

REFERENCE AN02 Technical Specifications, 3.1.1.5 l

St-8Dd1NISIB811YE_P8QQEQQ8Egt_QQNQlllQN$t_8NQ_LidlI8IlQNg PAGE 9

ANSWERS -- ARKANSAS NUCLEAR'ONE-2

-86/05/15-GRAVES, D.

ANSWER 8.10 (1.50)

- RWP required

- Must use " buddy" system

. Neutron badges and dosimeters (or HP specified dosimetry)

- Air sample

- Complete entry checkoff sheet

- Notify Shift Supervisor (3 required at 0.5 each)

REFERENCE ANO Procedure 1622.009, Requirements for Reactor Building-Power Entries, Rev 1, pg 1-2 ANSWER 8.11 (1.50)

- Monitoring containment atmosphere particulate radioactivity

- Monitoring containment sump inventory and discharge

- Performance of RCS water inventory balance

- Monitoring reactor head flange leskoff temperature (3 required at 0.5 each)

REFERENCE AN02 Technical Specifications 4.4.6.2.1 ANSWER 8.12 (2.00) a.

False b.

False c.

True d.

False (0.5 each)

REFERENCE AND Procedure 2502.01, Refueling Shuffle, Rev 13, pg 3-7 ANO Procedure 1015.01, Conduct of Operations, Rev 23, pg 27 ANO2 Technical Specifications 3.9.3.a m

. m.

- --a a

9'

!,,.a

..)

at__8DMINISIBoI1YE_EBQGEQUBESt_GQNQ1IIQNat_aNQ_ LIM 11al10NS PAGE 10 ANSWERS -- ARKANSAS NUCLEAR ONE-2

-86/05/15-GRAVES, D.

ANSWER 8.13 (1.00)

- More than a momentary loss of consciousness

- An actual or suspected fracture

- A head inj ury

- An injury that may have damaged internal organs

- A serious burn

- Hemorrhaging

- Receipt of a large dose of radiation (4 required at 0.25 each)

REFERENCE ANO EPIP 1903.23, Personnel Inj ury, Rev 11, pg 2 ANSWER 8.14 (1.50) a.

The action shall be approved by a licensed Senior Operator as a minimum (0,5).

b.

Arkansas Department of Health (0.5)

NRC (0.5) Operations Center REFERENCE ANO Procedure 2203.14, Alternate Shutdown, Rev 1, pg 5 ANO EPIP 1903.10, Emergency Action Level Response / Notification, Rev 18, pg 12 10 CFR 50.54.x and y l

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