ML20211H857

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Exam Rept 50-313/OL-87-01 on 870209-13.Exam Results:All Seven Senior Reactor Operators Passed Exams.Exam Encl
ML20211H857
Person / Time
Site: Arkansas Nuclear Entergy icon.png
Issue date: 02/20/1987
From: Cooley R, Pellet J
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20211H823 List:
References
50-313-OL-87-01, 50-313-OL-87-1, NUDOCS 8702260309
Download: ML20211H857 (46)
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ANO-1 OPERATOR LICENSE EXAMINATION REPORT No. 50-313/0L-87-01 Licensee: Arkansas Power & Light P. O. Box 551 Little Rock, AR 72203 Docket: 50-313 License No: DPR-51 Operator License examinations at Arkansas Nuclear One - Unit 1 (ANO-1).

Chief Examiner: _

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  • N 1*N*U John L. Fellet, Examiner Date Signed Approved By:
  • h Ralph A. Cooley, Seftion Chief

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Date Signed Summary:

NRC administered seven Senior Reactor Operator (SRO) upgrade examinations at ANO-1 during the week of February 9, 1987. All seven candidates successfully completed the examinations and have been issued the appropriate license.

0702260309 870220 PDR ADOCK 05000313 PDR

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  • 2 REPORT DETAILS
1. PERSONS EXAMINED License Examinations: TYPE: SRO Upgrade TOTAL PASS: 7 - 100% 7 - 100%

FAIL: 0- 0% 0- 0%

2. EXAMINERS J. Pellet,NRC(ChiefExaminer)

S. McCrory, NRC R. Cooley, NRC T. Morgan EG8G

3. EXAMINATION REPORT Individual performance results are not included in this report because these reports are placed in the NRC Public Document Room.
a. EXAMINATION REVIEW COMMENT RESOLUTION In general, coitorial coments or changes made during the examination, review, or subsequent grading reviews are not addressed by this resolution section. This section reflects resolution of substantive coments made in Arkansas Power & Light rompany letter ANO-87-01870. The modifications discussed below are included in the master examination key which is included elsewhere in this report (see 3.d), as are all other changes mentioned above but not discussed herein. Note that coments from the letter referenced are paraphrased below for brevity. The full text of the coments is available in the referenced letter, which is attached (see 3.e).

5.07.a. Answer is incorrect. Behavior will be as described in the answer to question 5.03.

Resp.: Accept. Key modified.

5.12.c. Reduced boron concentration will also have an effect.

Resp.: Accept. Key modified.

6.01.a Although coupling dimensions are different, interchange is prevented by administrative controls.

Resp.: Reject. Proposed answer does not answer question asked.

6.03 The top of each hot leg is also an acceptable answer.

Resp.: Accept. Key modified.

g.

3 6.04 Answers e and g are incorrect. Correct answers are 9" and 6.5", respectively.

Resp.: Accept. Key modified.

6.06.b. Administrative controls also limit overpressurization.

Resp.: Accept. Key modified.

6.08 Development of level function is not required to fully answer the question.

Resp.: Accept. Key modified.

6.11.a. Question is unclear. Consider additional answers per additional material provided.

Resp.: Accept. Key modified.

7.01 Answers given do not cover all possible cases.

Resp.: Accept. Key modified.

7.05 Answers 2, 4, 6, and 7 are either incorrect or incomplete per reference.

Resp.: Accept. Key modified.

7.07.c. Section 7.7 of reference lists another answer.

Resp.: Accept. Key modified.

b. FEBRUARY 12, 1987 EXIT MEETING

SUMMARY

At the conclusion of the site visit, the NRC examiners met with representatives of plant staff to discuss the site visit. The following personnel were present:

NRC AP&L C Cooley D aker C. Harbuck A. Cox S. McCrory E. Force T. Morgan J. Simmons J. Pellet D. Smith J. Vandergrift E. Wentz Mr. Pellet started the discussion by noting that the examiners as a group had encountered a positive, helpful attitude in everyone concerned. The following general topics were discussed.

(1) As required by NUREG-1021, the Examiner Standards, preliminary results are no longer provided in the exit meeting.

(2) NRC will attempt to return foririal, final results within 30 days.

4 (3) The following candidate or procedure problems were observed during the site visit and are presented for the use of the facility. Note that a problem does not imply unacceptable performance but is simply an area where knowledge, skill, or support is less completely developed than in others.

(a) Several procedures had undergone major revision during training, with the result that some candidates were aware of and comfortable with the new revision, and some were not.

(b) There was confusion as to individuals permitted to sign off various reviews for temporary modifications.

(c) There was confusion as to whether double isolation was required when removing a high energy line from service.

(d) There was confusion as to how major documentation, design, or license changes were brought to the attention of the operating crew between implementation and routine training.

(e) There appeared to be confusion between the Emergency Operating Procedure (1202.01), which states that the Shift Administrative Assistant snould be notified to make the proper event classification and appropriate notifications, versus the Emergency Action Level Response / Notification Procedure (1903.10), which states that classification is performed by the Shift Supervisor,

c. CENERIC COMMENTS The generic comments below were generated during grading of the written examinations, and regional review of the oral examinations.

They are provided for the benefit of the facility and no response or corrective actions are required.

(1) Understanding of the basis for license limits on secondary system activity was weak.

(2) Understanding of the overcooling effects of initiating natural circulation with main feedwater was weak.

(3) Knowledge of minimum emergency classifications requiring site protective action or evacuation was weak.

(4) Correlation of a given radiation reading to whole body dose (whether to include beta) was weak.

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d. EXAMINATION MASTER COPY Master copies (questions and answers) for the SRO license examination follow this page,
e. FACILITY EXAMINATION COMMENTS The facility examination review coments in the form of AP&L letter ANO-87-01870 follow the examination master copies.

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U. S. NUCLEAR REGULATORY CDPMISSION '

SENIOR REACTOR OPERATOR LICENSE' EXAMINATION

, FACILITY: ~+

8Bg8N183_NUGLg68_QbC-l'__

REACTOR TYPC: _ EWB:g4W12Z______. '

DATE ADMINISTERED:_DZlD2/1D____s.___1_,_____

EXAMINER: _ EELLEIt_)t______________ i

,, -e CANDIDATE: ________,,_____

__i((_____

IN11BUCIl0NS_IQ_C8NQ1Q81El . .

Uso separate paper for the answers. Writu answersononeside,'e(nly.

Staple question sheet on top of the answer shaits. . Points for each qu stion are indicated in parentheses.after the.quecti4n. Yle e. passing grcde requires at least 70% in each category and a finnt. gradt of at ,

c, locut.80%. Examination papers will be picked up six (6) hours after i tho examination starts. <

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% OF CATEGORY  % OF CANDIDATE'S CATEGORY 4

__V8LUE_ _IQI8L ___SCQBE___ _V8LUE__ _____..________G81EAQRY_____________

_2LADD__ 2LADD ___________ ________ 5. THEORY OF NUCLEAR POWER PLANT OPERATION, fLUIOS, Ar40 THERMODYNAMICS ,

_2540D__ _25100 ___________ _______'_6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION'

, _2 LADE _ _25 ADD ___________ ________ 7. PROCC00RES - NORMAL, ABNORMAL,

EMERGENCY AND RADIOLOGICAL -

CONTROL 2 _25400__ _25200 _._________ ________ 8. ADMINISTRATIVE PROCEDURES, CONDIT' IONS, AND LIMITATIONS

< 100&DQ__ ___________ ________%' Totals

Final Grade 4

s All work done on this examination is my own. I have neither giv'en

nor received aid.

Candidate's Signature l ~

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- NRC RULES AND GU,IDELINES FOR LICENSE EXAMINATIONS

t. During the admin $stratidA of this examination the following rules apply:

^ 1. Cheating on the examination,means an automatic dental of your application endgcould result in mo r e s e'vir r e {peb,e lt ies .

2. Restroom trips are to be limiteA add only one candidate at a time may lenke. You must avoid, all contacts with anyone outside .the examination ,

room to avoid even the appearance or possibility of cheating.

  1. - f e . . ,

j f3'A lso black ink or dark pencil' 201X to f acilitate leg ib le reproductions.

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4.I Print your name in 'the blank provided on the cover sheet of the -

oxamination.. ,

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5.'l Fill in the date ,'en the cover sheet of the examination (if necessary).

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6. Use caiy the parer.provided for answers.

i j 7. Print c yeur name in the upper right-hand corner of the first page of 33gh j nection of the answer sheet.

! 8. Consecutive 1hnumber each answer sheet, write "End of Category __" as appropriate, start each category on a Deu page, write 2DlX 90 SDR 11dt of the paper, and write "Last Page" on the last answer sheet.

j 9. Number each answer as to category and number, for example, 1.4, 6.3.

J 10.-Skip at least ibtas lines between each answer.

i, I Seburate answer sheets from pad and place finished answer sheets face  !

!'.11. down on your desk or table.

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12. Use abbreviations only if they are commonly used in facility 111st31yts.
13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.
14. Show all calculations, methods, or assumptions used to obtain en answer to mathematical problems whether indicated in the question or not.

{ 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE i QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.

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! 16. If parts of the examination are not clear as to intent, ask questions of j the angelDat only.

s l 17. You'must sign the statement on the cover sheet that indicates that the l work is your own and you have not received or been given assistance in l completing the examination. This meat be done after the examination has i j heen completed.

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18. When you complete your examination, you shall;
o. Assemble your examinaticn as follows:

(1) Exam questions on top.

(2) Exam sids':i- figures, tables,.etc.

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(3) Answers pagea including figures which are part of the answer.

b. Turn in your copy oflthe examination and all pages used to answer the examination questions.
c. Turn in all scrap' paper and the balance'of the paper that yourdid.

not use for answering the questions.

d. Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.

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l 52__INEQBl_QE_NUGLE88_EQWEB_EL8NI_QEEB8IIQNt_ELUIDSt_8NQ PAGE 2 IHEBBQQ1N8dICS QUESTION 5.01 (1.00)

The reactor is subcritical with a K-eff of 0.960. Source channels are ind ic at ing 5 counts per second (cos). What is K-eff when source channels chow a count rate of 60 cps after rods are withdrawn? SHOW ALL WORK FOR FULL CREDIT.

, QUESTION 5.02 (1.50)

e. HOW does equilibrium Xenon reactivity (Xe-eq) at hot full power change as a function of core age CEFPD)? (0.5) e b. WHY does Xe-eq change as a function of core age (EFPD)? (1.0) l' QL'E S T ION 5.03 (2.00)

What two factors, other than water temperature, cause an increase in heat transfer when injecting EFW instead of main feedwater? (2.0)

QUESTION 5.04 (2.50)

If a main steam safety valve leaks at high power, how will the steam perameters below change across the leak? No explanation is required.

c. Enthalpy.

b.- Entropy.

c. Pressure,
d. Temperature.
o. Density.

l QUESTION 5.05 (1.00)

What are three (3) reasons for es t ab l is h ing regulating group insertion limits?

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

St__IHE981_9E_NUGLE88_E9 WEB _EL6HI_DEEB8II9Nt_ELUIQ3t_8NQ PAGE 3 IBEBdQDINedIGS QUESTION 5.06 (2.50)

Indicate which of the following are TRUE and which are FALSE INDICATIONS that the point of adding heat CPOAH) has been reached.

(Assume normal plant operation) (2.5)

c. SUR decreases.
b. Pressurizer level decreases.
c. Turbine bypass controller station (Bailey) output decreases.
d. Pressurizer pressure decreases.
o. T-ave increases.

QUESTION 5.07 (2.50)

Explain how each of the actions below would reduce cooldown rate during a notural circulation cooldown. Include in your answer the affect on primary cnd secondary parameters.

c. EFW flow is reduced (assume no main feedwater). (1.2G)
b. Turbine bypass valves are closed (assume NOT completely). (1.25)

QUESTION 5.08 (1.00)

TRUE or FALSE? No explanation required. (1.0)

If EFW is not available, natural circulution can be initiated using main fcedwater, but it takes longer and will result in overcooling.

QUESTION 5.09 (2.50)

Given a Startup Rate of one (1) decade per minute, what is:

a. the period. (0.8)
b. the doubling time. (0.8)
c. the time required to increase power by a factor of 25. (0.9)

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

St__IHEQBY_DE_NUGLE88 EDWEB_EL8NI_9EEB8IIQNt_ELUIQ3t_8NQ PAGE 4 IHEBdQQXN8 DIGS QUESTION 5.10 (2.50)

Explain the behavior of the AN01 reactor following a normal reactor trip from power. Discuss the affect of both prompt and delayed neutrons. (2.5)

QUESTION 5.11 (3.00)

Match the labeled areas of the chart below showing countrate as a function of time during a reactor startup with the appropriate description. (3.0)

(NOTE: Several areas have the same label.)

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  • TIME I *b i
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  • 1 *e 1
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  • 1
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  • 1. Final countrate.

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  • 2. Initial countrate.

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  • subcritical multiplication I *e 4. Equilibrium value of I
  • f inal critical countrate l l *** c 5. Prompt j ump i
  • 6. Critical reactor i
  • 7. Point of adding heat

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l 1 1 1 I i 1 10 100 1,000 10,000 100,000 l

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

Sz__IHEDBl_9E_NUGLE8B_EDWEB_EL6NI_QEEB8IIDNz_ELUIDit_6ND PAGE 5 IHEBUQDIN851G1 QUESTION 5.12 (3.00)

e. Why does the power doppler deficit curve slope change at 15%

power? (1.0)

b. If the moderator temperature increases, how will the resulting change in each of the effects below affect reactivity? Assume middle of life, medium power. (1.0)
1. Density.
2. Rod Worth.
3. Voids.
4. Mass Flow Rate.
c. How is individual control rod worth affected by core age? WHY? (1.0)

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

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6t__EL8HI_111IEBl_DE119Nt_CONIBQLt_eND_INSIBWNENIeIION PAGE 6 QUESTION 6.01 (2.00)

m. What physical design f eature(s) prevents interchanging a control rod assembly (CRA) and an axial power shaping rod assembly (APSRA) during fuel handling? (1.0)
b. Fill in the blanks in the f ollowing paragraph dealing with f uel rods.

Blanks may contain one or more words or phrases. (1.0)

All fuel rods are internally pressurized with __(1)__. The fuel is in the form of pellets. Pellet ends are __(2)__ to minimize differential l thermal expansion between the fuel and cladding. Above and below each fuel column is a __C3)__.

QUESTION 6.02 (3.00)

Explain WHY each of the statements below is FALSE. (3.0)

a. A patch panel is used to reassign specif ic APSRA dr ives.

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b. CRDM transfer relays use break-before-make contacts to' transfer auxiliary power to and from the CRDM.
c. The relative rod position indication is derived from monitoring the input pulses for each of the six-phase input leads to tiie CRDM motor.
d. The APSRA do not automatically insert on an RPS signal to open the CROM breakers because their DC hold power supply is not fed from the affected breakers.

i QUESTION 6.03 (1.50)

Where are three (3) locations monitored by the vibration and loose parts i sonitoring system on the Reactor Coolant System (RCS)? (1.5) l l

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

6t__EL8HI_SISIEMS_DESIGNt_CONIBQLt_8ND_INSIBUMENI6IIQN PAGE 7 QUESTION 6.04 (3.00)

Fill in the blanks in the following Reactor Coolant Pump starting interlocks. (3.0)

e. Reactor power is less than _____ %.
b. RCP seal inj ec t ion flow is greater than _____ gpm.
c. RCP motor ICW flow is greater than _____ gpm.
d. RCP seul Cooling ICV flow greater than _____ gpm.
o. RCP motor upper bearing oil reservoir level greater than _____ inches.
f. RCP HP oil li.ft pressure greater t han _____ ps ig .
g. RCP motor lower bearing oil reservoir level greater than _____ inches.
h. RCS temperature greater than _____ deg's F to start fourth RCP.

QUESTION 6.05 (1.50)

c. What is the purpose of the 12" Expansion Tank in the Primary Makeup and i Pur if ication system? C0.5) l l b. What specific nuclide is chosen for activity monitoring by the Failed l Fuel Radiation Monitoring System in the Primary Makeup and Purification l System? (0.5)

I l c. What is one (1) reason this nuclide was selected? (0.5) l l

l l QUESTION 6.06 (3.00)

c. Why is it desirable during cooldown to use the "A" DHR loop? C0.0)
b. How is overpressurization of the Decay Heat Removal system (OHR) prevented. Setpoints are NOT required. (2.0)
c. What are two (2) of the automatic actions from the OHR pump start circuit, NOT including DHR pump start? (1.0)

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

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62__EL8HI_11SIEUS_DESIGNt_GQNIBQLt_8NQ_INSIB9dENI8IIQN PAGE 8 QUESTION 6.07 (2.00)

c. How is inadvertent draining of the Spent Fuel Pool prevented in the:
1. pump suction piping. (0.5)
2. pump discharge piping. (0.5)
b. Which of the following is the correct reason that Spent Fuel Pool boron concentration must be maintained greater than 1800 ppm? (0.5)
1. Technical Specification requirements.
2. Assure safe storage of spent fuel.
3. Prevent dilution of transfer canal water.
c. Other than boron concentration, what two (2) Spent Fuel system design feature ensures that criticality will not occur? (0.5)

QUESTION 6.08 (2.00)

c. How does ICS prevent OTSG water inventory from dropping below that corresponding to 15% load during low power operations? (1.0)
b. How does ICS prevent flooding of the aspirating ports in the OTSG7(1.0)

QUESTION 6.09 (2.00) l e. Why are GM tubes needed in addition to N16 monitors on the main steam l lines? (What functions are accomplished by the GM that could not be

! assured with the N167) (1.0) l l b. The N16 monitor can read out in the control room for either " gross" or l "N16" indication. Which is normally used, and WHY? (1.0) l i

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QUESTION 6.10 (1.00)

How does an annunciator respond to "reflash" due to another of multiple inputs when it already displays a solid window? (1.0)

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

6t__EL8HI_SISIEMS_DESIGNt_CONIB9Lt_8NQ_INSIB9dENI8IIQN PAGE 9 t

. QUESTION 6.11 (2.00)

m. Fill-in the blanks in the following paragraph about the Inadequate Core Cooling Monitor and Display (ICCMDS). Blanks may represent one or more words, numbers, or phrases.

A Reactor Vessel (RV) level sensor cons is ts o f __(1)__ Chow many) thermocouples connected internally to provide a signal proportional to

__(2)__ (what). There are __(3)__ Chow many) level probes, each having

__(4)__ Chow many) level sensors. (1.0)

b. Besides RV level, what are the four other inputs to the ICCMDS? (1.0)

QUESTION 6.12 (2.00)

Ocscribe the three (3) trip signals supplied to RPS associated with the power range NI's, NOT including any anticipatory trips. (2.0)

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

Zi__EB0GEQUBE1_:_NQBd8Lt_8BNQBd8Lt_EMEBGENGY_8NQ PAGE 10 B8DIQLQGIG8L_GQNIBQL QUESTION 7.01 (1.50)

Por 1102.08, " Approach to Criticality," if count doubling is not observed during withdrawal of the safety groups, the startup must be stopped unless one of three conditions is satisf ied along with an increasing count rate during rod withdrawal. What are two (2) of these conditions? (1.5)

QUESTION 7.02 (2.00)

What are two (2) diverse (different) symptoms of a Steam Generator Tube Rupture per 1202.01, " Emergency Operating Procedure"? (2.0)

QUESTION 7.03 (3.50)

An irradiated component measuring 2"x 2"x 2" is surveyed with a portable instrument. At a distance of 2 feet, the open window reading is 2.65 R m/hr and the closed window reading is 1.75 Rem /hr. SHOW ALL WORK!

c. What accounts for the difference in open and closed readings? (0.5)
b. How long could you remain at the survey distance without exceeding AN01 weekly and quarterly limits (2 answers req'd per 1622.011)? (1.0)
c. Who {by job title (s)) must approve exceeding the weekly control limit? (1.0)
d. What would the WHOLE BODY dose rate be if you moved back to a distance 10 ft. from the component ? (1.0) i l QUESTION 7.04 (2.00) i In what SPECIFIC location can each of the operators listed below find his critten instructions during a remoto shutdown? (2.0)
c. Auxiliary Operator,
b. Assistant Plant Operator,
c. Plant Operator, and
d. Waste Control Operator.

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

'Z4__EBQGEDUBE2_=_HQBUALt_8BNDBM8Lt_EMEBGENGY_6NQ PAGE 11 88DIDLQGIQ8L_GQN1BQL QUESTION 7.05 (3.00)

List five (5) of the seven (7) entry conditions or symptoms given in OP 1202.01, " Emergency Operating Procedures," that require the operator to canually trip the reactor. (3.0)

QUESTION 7.06 (3.00)

Answer TRUE or FALSE for each of the statements below dealing with Tochnical Specification limitations on fuel handling.

. a. At least one decay heat removat loop shall be in operation whenever j

core geometry is being changed. -

(0.5)

b. If Reactor Building refueling area radiation monitor RE-8017 becomes inoperable, all fuel handling activities must be terminated until it is repaired. (0.5)
c. Core subcriticel. neutron flux shall be continuously monitored by at least two neutron flux monitors whenever core geometry is being i

changed. (0.5)

d. Two irradiated f uel assemblies shall not be moved simultaneously by the bridges within the fuel transfer canal. (0.5)
e. Loads in excess of 2000 pounds shall be prohibited from travel over fuel assemblies in the storage pool. (0.5)
f. All fuel handling in the Reactor Building shall cease upon notification

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of the issuance of a tornado watch for Pope, Yell, Johnson, or Logan counties. (0.5) i i

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

Z2__EB9CEDUBE1_ _NQBd8Li_8BN9Bd8Li_EMEBGENGX_8NQ PAGE 12 B8DIDLOGIG8L_GQNIBQL QUESTION 7.07 (3.00)

During a reactor trip recovery, per 1102.06, the Shift Supervisor makes the dstormination of the authorization required to restart by evaluating 7 criteria.

m. If all criteria are negative, who (by job title) must authorize-a reactor restart? (2 titles required for full credit.) (1.0)
b. If one or more criteria are positive, who (by job title) must authorize a reactor restart, in addition to those in part a? C0.5)
c. List three (3) of the seven (7) criteria mentioned above. (1.5)

QUESTION 7.08 (3.50)

m. Reactor building integrity shall be maintained when what three (3) conditions exist? (1.5)
b. What are four (4) conditions which must be satisfied for reactor building integrity to exist? (2.0)

QUESTION 7.09 (2.00)

What is the dose threshold AND basis for posting as a Radiation Area BOTH inside and outside of Controlled Access Areas? (2.0) i QUESTION 7.10 (1.50)

Fill in the blanks in the following paragraph taken from 1102.04, " Power Operations. Blanks may contain one or more words or numbers. (1.5)

The Technical Specification Limits on Quadrant Power Tilt and Imbalance must be maintained. These must be monitored and recorded in the Power History, Imbalance and Quadrant Tilt Log each 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. This requirement exists above __(a)__ % power for quadrant tilt and

__(b)__ % FP for imbalance. When logging quadrant tilt, log

__Cc)__ (which) value, l

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(***** END OF CATEGORY 07 *****)

r-B2__8DMINISIB8IIVE_EB0GEQUBEft_C9NDIIIQNSt_6NQ_LIMIIeIIONS PAGE 13 QUESTION 8.01 (2.50)

Ccmplete the table below on reactor operating conditions or modes per Tcchnical Specifications. (NOTE: NAR = No Answer Required.)

OPERATING CONDITIONS Taverage REACTIVITY REACTOR POWER Rofueling Shutdown (a) (b) NAR Cold Shutdown (c) (d) NAR Hot Shutdown Ce) NAR NAR Hot Standby NAR (f) (g)

QUESTION 8.02 (3.00)

Fill in the blanks in the following paragraph dealing with reactor coolant leakage. Each blank may contain on or more words or numbers.

e. If the total reactor coolant leakage rate exceeds __(1)__, the reactor shall be __(2)__ within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of detection. (1.0)
b. If the unidentified reactor coolant leakage exceeds __(1)__, the reactor shall be __(2)__ within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of detection. (1.0)
c. If reactor coolant leakage through a non_ isolable fault in __(1)__

exceeds __(2)__,-the reactor shall be __(3)__ within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of detection. (1.0)

QUESTION 8.03 (3.00) l What are four (4) changes to procedures that are classif ied as an " Intent Change" per 1000.06, " Procedure Review, Approval, and Revision Control"?

QUESTION 8.04 (1.00)

Msnipulation of what plant equipment requires the use of BOTH AP&L Switching Orders and ANO HOLD cards?

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

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Ai__80dINISIB8IIVE_EBQGEDMBEft_GQNDIIIONSt_8NQ_LIMII8IIONS PAGE 14 QUESTION 8.05 (2.00)

Complete the following statement dealing with Shift Supervisor authority.

The Shif t Supervisor shall have specific authority to order power re-duction or shutdown if continued operation of the unit will result in:

a. Immediate _____,
b. _____ to station personnel,
c. Violation of _____, or
d. Unnecessary _____.

NOTE: Each blank may contain one or more words or phrases.

QUESTION 8.06 (1.00)

Lis t two groups which are specifically granted accwss to the Control Room by OP 1015.01 during emergencies. Do NOT include shift operating personnel in your answer.

QUESTION 8.07 (1.00)

On your shift a monthly surveillance item is discovered overdue.

Required due date was 25th of the month, assume today is tfie 31st, and the performance of the SP has begun. All previous sur/eillances were ccmpleted on time as scheduled. Which of the statements balow is correct about the surveillance (SP)?

, n. The SP has been missed and the system must be declared inoperable until the SP is completed satisf actory,

b. The system is operable as the Technical Specifications allow a monthly SP to be waived 1 month out of 3.
c. The system is inoperable because the 3.25x time interval for 3

! consecutive SPs was not met.

d. The system is operable because the Technical Specif ications allow a time extension which has not been exceeded.

, QUESTION 8.08 (1.00) j Por 1903.10, " Emergency Action Level Response /Notif ication," what is the difference in staffing between Phase 1 and Phase 2 of the Initial Response Organization (IRO-1 & IRO-2)? (1.0)

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

i

Az__8DDINISIB8IIVE_EBQCEQUBESt_CONDIIIQNSt_8NQ_LIMII8IIQNS PAGE 15 QUESTION 8.09 (3.00)

c. What is the minimum Emergency Class where a radiological release may require onsite protective actions or of f site emergency measures? (1.0)
b. What is the minimum Emergency Class where a Plant Evacuation may be required? (1.0)
c. Describe the minimum damage required for proper classification as a General Emergency. (1.0)

QUESTION 8.10 (2.50)

Por 1015.03, " Operation Log Taking:"

s. Is it permissible to correct data sheet logs with white out? (0.5)
b. What are two (2) corrective actions required if the reading exceeds the

" Operability Difference?" (1.0)

c. What are two (2) exceptions when exceeding the " Operability Dif f erence" does not require corrective action? (1.0)

QUESTION 8.11 (3.00)

Fill in the blanks in the following paragraphs dealing with Technical l

Specification minimum conditions for cr it icality. Blanks may contain one

! or more words or numbers. (3.0)

c. The reactor coolant temperature shall be above _____ deg's F. (0.5)
b. Pressurizer level must be above __(1)__ and below __(2)__ inches. (1.0)
c. At least _____ (how many) of the emergency-powered pressurizer heater

, groups must be operable. (0.5) l

d. If the limit in part a or b is violated, it must be restored within l

i __(1)__ minutes, or be in at least Hot Shutdown within the next l __(2)__ minutes. (1.0) l

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

i

Az__80dINISIBeIIVE_EBQGEQUBEft_CONDIIIQN11_eND_LIMIIeIIONS PAGE 16 QUESTION 8.12 (2.00)

Why is a gross activity lievitt NOT required for the secondary coolant per Technical Specifications (even though it is for the primary coolant)? (2.0) 1

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

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

                              • )

Si__IHEQBY_QE_NUGLE88_EQWEB_EL8HI_QEEB8IIQNt_ELUIQSt_8NQ PAGE 17 IHEBdQDIN8 DIGS ANSWERS -- ARKANSAS NUCLEAR ONE-1 -87/02/10-PELLET, J.

ANSWER 5.01 (1.00)

CR1/CR2 = (1-K2)/(1-K1) (0.5 PROPER FORMULA)

K2 = 1 - CR1/CR2 * (1-K1) (0.25 PROPER USE OF FORMULA)

= 1 - 5/60*(1-0.960)

= 1 - 0.00333 K2 = 0.9967 (0.25 MATHEMATICS)

REFERENCE ANO Sample Questions and Answer Keys, Q1.14 (OCAN068314)

ATTS Manual ANSWER 5.02 (1.50)

e. Xe-eq gets larger as a function of core age. (0.5)
b. Xe-eq is a function of flux (not power) (0.5) and flux increases as a function of core age (0.5). (1.0)

REFERENCE AN01 PLANT SPECIFIC REACTOR THEORY, F iss ion Product Poisons ANSWER 5.03 (2.00)

1. The spray of feedwater into the steam space reduces Potsg (like pressurizer spray), reducing Tsat/Tcold and dT. (1.0)

, 2. The wetting of the tube surfaces effectively replaces steam area with water area, which increases ht. (1.0)

REFERENCE AN01 B&W ATOG Part II - Volume 1, p. 20 l

i l

52__IHE981_DE_ NUCLE 88_E9 WEB _EL8HI_9EEB8IIQNt_ELUIQ2t_8NQ PAGE 18 IBEBUDDINedIGS ANSWERS -- ARKANSAS NUCLEAR ONE-1 -87/02/10-PELLET, J.

ANSWER 5.04 (2.50)

c. Constant.
b. Increase.
c. Decrease.
d. Decrease.
o. Decrease.

(5 answers e 0.5 ea.)

REFERENCE AN01 HEAT TRANSFER THERMODYNAMICS AND FLUIDS HANDBOOK ANSWER 5.05 (1.00)

Shutdown margin (0.33), ejected rod worth (0.34); ECCS pwr peaking (0.33).

REFERENCE Tochn ical Spec if icat ion 3.10. 5. a ANSWER 5.06 (2.50)

n. True
b. False
c. False
d. False
e. True (5 4 0.5 ea) (2.5)

REFERENCE B&W Operation fundamentals a

v .-. _ -

. . - - - . _ . _ _ _ , . , _ _ - _ _ _ _ , c .- _ . . - , -- . _ , _. ._ ..m. . _

St__IHEDBY_DE_NUGLE88_EQWEB_EL8HI_9EEB8IIONt_ELUIDSt_8NQ PAGE 19 IHEBdQQ1NedIGS ANSWERS -- ARKANSAS NUCLEAR ONE-1 -87/02/10-PELLET, J.

ANSWER 5.07 (2.50)

c. Throttling EFW flow reduces tube wetting, consequently reducing effective wet area in the SG, and thereby heat transfer. If the TBV's are in manual, Potsg will increase due to less quenching spray. If the TBV's are in automatic, Potsg will stay roughly constant. CONCEPTl (1.25)
b. Closing the turbine bypass valves partially will increase Potsg immediately (0.25). This increases Tsat/Tcold (0.25) which reduces differential temperature (Thot-Tcold) (0.5) which reduces heat transfer from the RCS (0.25). (1.25)

REFERENCE AN01 B&W ATOG Part II - Volume 1, p. 20 ANSWER 5.08 (1.00)

TRUE. (False OK if stated will not overcool if level incr. before trip.)

REFERENCE AN01 B&W ATOG Part II - Volume 1, p. 52 ANSWER 5.09 (2.50)

a. t=26/SUR (0.5) so t=26 seconds (0.3).
b. DT=18/SUR (0.5) so DT=18 seconds (0.3).
c. t=1og(P/Pi)/SUR (0.5) so t=1.37 min. or 83 seconds +/-10% (0.4).

REFERENCE AN01 Plant Specific Reactor Theory, p. 131-133

Si__IHEQBY_QE_N9CLE8B_EQWEB_EL8HI_QEEB6IIONi_ELUIDSt_6NQ PAGE 20 IHEBdQDIN8 DIGS ANSWERS -- ARKANSAS NUCLEAR ONE-1 -87/02/10-PELLET, J.

ANSWER 5.10 (2.50)

After a reactor trip the reactor experiences a very sharp transient drop for a few seconds, followed by a stable shutdown period (0.9). Prompt noutrons are r espons ib le for the transient prompt drop within a few esconds, but are then basically removed from the process (0.8). This etable shutdown period is based on the half-lif e of the longest-lived dolayed precursor (s), (which is 55 seconds) (0.8).

REFERENCE AN01 Plant Specific Reactor Theory, p. 135 ANSWER 5.11 (3.00)

o. 7, b. 6, c. 5, d. 2, e. 3 (5 answers e 0.6 es. - 3.0)

REFERENCE AN01 Plant Spec if ic Reactor Theory, p. 146 ANSWER 5.12 (3.00)

a. From 0-15%, deficit is combination of MTC & FTC but >15%, MTC constant so deficit due to FTC. (CONCEPT). (1.0)
b. 1. negative (density decreases).

i 2. negative (worth increases).

3. negative (voiding increases).
4. negative OR no effect (mass flow decreases).[4 answers e 0.25 ea.)

[

c. The core is less absorptive and less competitive (e.g., decr. 82 conc.)

j with incr. age (0.8), increasing individual rod worth (0.2) (1.0)

REFERENCE AN01 Plant Reference Theory, p. 172, 176, 186 i

6t__EL8HI_SYSIENS_DESIGNt_GQUIB9Lt_6ND_INSIBudENI6IION PAGE 21 ANSWERS -- ARKANSAS NUCLEAR ONE-1 -87/02/10-PELLET, J.

ANSWER 6.01 (2.00)

c. APSRA coupling dimensions are slightly dif f erent. (1.0) j
b. 1. helium
2. dished
3. spring spacer (3 answers e 0.333 ea. - 1.0)

REFERENCE AN01 STM 1-1, p. 7, 8 ANSWER 6.02 (3.00)

c. Patch panel assigns only CRA not APSRA.

(APSRA are not r e ass ig ne d, so patch panel is not required.

b. Relays use make-before-break rather than opposite.
c. RPI monitors every other phase, not all 6.
d. APSRA do not scram due to buttons on the roller nuts. (4 8 0.75 es.)

REFERENCE AN01 STM 1-2, p. 2-8 ANSWER 6.03 (1.50)

1. Lower reactor vessel region.
2. Upper reactor vessel region.
3. Top region of each OTSG/ hot leg. (3 answers e 0.5 ea. - 1.5)

REFERENCE AN01 STM 1-3, p. 1 ANSWER 6.04 (3.00)

c. 22; b. 3; c. 250; d. 30; e. 9; f. 1750; g. 6.5; h. 500 C+/-10% full credit; 8 4 0.375 eu.)

62__EL8NI_SISIEdS_DESIGNt_GQNIB9Lt_8ND_INSIBWNENI6II9N PAGE 22 ANSWERS -- ARKANSAS NUCLEAR ONE-1 -87/02/10-PELLET, J.

REFERENCE AN01 STM 1-3, p. 6, 7 ANSWER 6.05 (1.50)

c. Tank increases transport time to allow decay of N-16. (0.5)
b. 1-135. (0.5)
c. 1. Abundant in the event of fuel failure.
2. Released from fuel easily.
3. Does not plate out.
4. Energetic gamma makes detection easy.
5. Short half-life means indication current. (Any 1 0 0.5)

REFERENCE AN01 STM 1-4, p. 2, 3 ANSWER 6.06 (3.00)

o. *** delete, no longer applicable *** (0.0)
b. 1. Isolation and suction valves interlocked on RCS pressure and CFTank outlet valve position.
2. Suction and discharge relief valves installed.
3. Admin. controls limit when in service. (Any 2/3 e 1.0 es.)
c. 1. DHR cooler service water isol. valves open.
2. Room cooler fan starts.
3. Cooling water valves to OHR room coolers open (fm fan start).
4. DHR pump bearing cooling water isol. valves open.(Any 2 8 0.5 ea.)

REFERENCE AN01 STM 1-5, p. 1-3

6t__EL8HI_SISIEMS_DESIGNt_GONIBQLt_8NQ_INSIB9dENI6IIQN PAGE 23 ANSWERS -- ARKANSAS NUCLEAR ONE-1 -87/02/10-PELLET, J.

i ANSWER 6.07 (2.00)

c. 1. Suction lines are near water level. (0.5)
2. Discharge lines have siphon breakers. (0.5)
b. 3. (0.5
c. 1. Boroflex plates Crack material). (

0.25) 2. Rack spacing. (

0.25)

REFERENCE AN01 STM 1-7, p. 1, 2 ANSWER 6.08 (2.00)

c. When level drops, control of feedwater switches to low level control to maintain the minimum dp. (CONCEPT) (1.0)-
b. When increases above maximum permitted value, control of feedwater switches to the high level control circuit, which limits OTSG load. (CONCEPT) (1.0)

REFERENCE AN01 STM 1-15, p. 4, 5 ANSWER 6.09 (2.00)

c. GM tubes are high range for releases that might over-range the N16 monitors (0.5). Plus, GM are upstream of MSIV, so they can be used to monitor dump or safety valve release (0.5). (1.0)
b. Usually left in " gross" because more sens it ive. (1.0)

REFERENCE AN01 STM 1-15, p. 9, 10 i

-61__EL8HI_SISIEUS_DE119Ni_C9 NIB 9Li_8HD_INSIBMBENI8IIQN PAGE 24 ANSWERS -- ARKANSAS NUCLEAR ONE-1 -87/02/10-PELLET, J.

ANSWER 6.10 (1.00)

Alarm window will go to " fast flash" (0.5) without audible alarm (0.5).

REFERENCE AN01 AOP 1203.12E, p. 3 ANSWER 6.11 (2.00)

s. 1. 2
2. temperature (difference)
3. 2
4. 9 (4 answers e 0.25 ea. - 1.0)
b. 1. Core Exit T/C.
2. Hot Leg Level.

! 3. RCS Pressure.

4. RCP contacts. (4 answers e 0.25 ea. - 1.0)

REFERENCE AN01 IRC Procedure 1105.08, p. 2 ANSWER 6.12 (2.00)

1. Preset maximum neutron flux.
2. Variable neutron flux based on RCS flow & neutron imbalance.
3. Variable neutron flux based on # RCP operating. (3 answers e 0.666 es.)

REFERENCE AN01 STM-67, p. 2 4

e

, - -, - - - - - - - - - --- - - . - , ,- - n, ,- ,,-m -

Zi__EB9CEQUBES_:_NQBM8Lt_6BNDBM8Lt_EMERGENCX_8NQ PAGE 25 88DIDLOGIC8L_CONIBQL ANSWERS -- ARKANSAS NUCLEAR ONE-1 -87/02/10-PELLET, J.

ANSWER 7.01 (1.50)

1. If the initial saf ety rod conf iguration was partially W.D.
2. If the ECP'is >100% on group 5 C/50% on grp 6 / 100% on grp 6).
3. If going critical on boron with high initial conc. (any 1/3 e 0.75 ea.)

NOTE: Current procedure revision eliminated above 3 with table for possible values for #1 & #2. Accept concept of #1 for full credit.

REFERENCE ANot 8 oroach to Criticality, 1102.08, Rev. 7, p. 4 l

ANSWER 7.02 (2.00)

1. Increased RCS leakage.
2. Increased secondary radiation (steam line N-16 alarm & cond. vac. pump radiation alarm).

(2 answers e 1.0 ea.)

REFERENCE AN01 E0P, 1202.01, p. 3 s

ANSWER 7.03 (3.50)

m. Beta radiation. (0.5)

! b. Stay time: qtriy : 2.50 RC1mts)/1.75 R/hr = 1.5 hr = 85 minutes (0.5) weekly: 0.30 RC1mts)/1.75 R/hr = .17 hr = 10 minutes (0.5) i (Accept B-dose if explicitly stated WB if dose to eye.)

c. HP Supervisor (4 300mr; HP Superintendent e 600mr; & GM e 1000mr) (1.0)
d. 1.75 x (2 ft.)^2 = Dose x (10 ft.)^2 (relation w/ dist 0.7, math 0.3)

Dose = 1.75 x 4 / 100 = 0.07 Rem /hr.

REFERENCE i AN01 1622.011, Rev. 4, p. 3 i

l l

l

Z2__EBQCEDUBE2_ _NQB58Li_8DNQBd8Li_EUEBGENCY_8NQ PAGE 26 B6DIQLQGIC8L_CQNIBQL ANSWERS -- ARKANSAS NUCLEAR ONE-1 -87/02/10-PELLET, J.

ANSWER 7.04 (2.00)

n. Stair rail to H2/N2 storage room (SU boiler room),
b. A-1/2, H-1/2, bus area.
c. Dasey Panel.
d. MCC B61.

(4 an=wers e 0.5 ea.)

REFERENCE AN01 OP 1203.29 ANSWER 7.05 (3.00)

1. Reaching an RPS setpoint with no reactor trip.
2. Pressurizer level > 290".
3. MSLI actuation or MSIV C' closure.
4. OTSG 1evel < 15" or > 95% on the operating range (w/ no control).
5. Any sys. degrad. where operator or TS say trip ex. req'd.
6. Pressurizer level < 100" w/ no indication it is being controlled
7. RB pressure > 3 psig (during pressure increasing transient).
8. Any automatic reactor trip.

(sny 5 8 0.6 ea.)

REFERENCE AN01 OP 1202.01 ANSWER 7.06 (3.00)

a. F b. F c. T d. F e. T f. F C0.5 ea.)

REFERENCE AN01 TS, Section 3.8, p. 58, 59

_l~

Z2__EBQGEDUBES_=_NDBU8Lt_8BNQBd8Lt_EMEBGENCY_8HD '- PAGE 27 88DIQLQGIC8L_GQNIBQL ,

ANSWERS -- ARKAtlSAS NOCLEAR ONE-1 -87/02/2.0-PELLET, J. , ,

+

1 -

r j:

ANSWER 7.07 (3.00) -

, . f. ,

c. Shift Supervisor (0.5), Operations Superintendent (0.5). /.,,

o.

1

^

b. Operations Manager (0.5). J' J' r j>
c. 1. Did automatic ESAS occur?

^' ['

2. Did any.EFIC train actuate? ,
3. Did any major equipment damage occur? ,
4. Is the cause of the trip unknown?
5. Were any RAC's written due tooff-normalper[formance?
6. Were any sections of the emergency plan activated? 4
7. Did plant parameters go beyond the bounds of the normal 4

transient limit envelop of SPDS (1600 psig <' Prcs < 2450 ps ig, Trcs < 510 F, subcooling margin <. 50 F)'. / '

8. Any TS-req'd equipment 00S prior to reactor trip. />

(Any 3/8 4 0.5 ea.) ,,

REFERENCE AN01 1102.06, Rev. 7, p. 4

.\

i:

('

m r ,

ANSWER 7.08 (3.501- ', '

e. 1. Pres >/= 300 psig l (0.5)

Trcs >/= 200 F

2. .

[d.5)

3. Nuclear fuel is in the core. (0.5)
b. 1. Equipment hatch is sealed and closeJ.
2. Both doors of airlocks are sealed and closed.

. 3. During access, at least i door of airlock / hatch sealed & closed.

4. All non-automatic RB isol. valves / flanges closed as required.4

! 5. All automatic RB isol. valves operable or closed.

! 6. RB leakage is within TS limits.

i -(Any 4/6 at 0.5 ea.)

l REFERENCE AN01 TS, Section 1.7, p. 5; Section 3.6, p. 54 I

, ANSWER 7.09 (2.00) l Inside CAA: 2.5 mr/hr (0.4) based on 100 mr / 40 hrs / week (G.6).'

l Outside CAA: 0.8 mr/hr (0.4) based on 100 mr / 120 hrs / week (0.6).

l l

l

U

.; I

. ,, ~ ; ,

PAGE 28 up' Zt__EBQGEDUBES_ _NQBd8Lt_8BNQBd8Lt_EMEBfENCY_8NQ

?- y B6010LQGIG8L_CONIBQL ANSWERS ~'- ARKANSAS NUCLEAR ONE-1 -87/02/10-PELLET, J.

/

REFERENCE AN01'1000.31, Rev. 1, p. 63 ANSWER 7.10 (1.50) c'. 15, b. 40, c. highest positive (>) (3 answers e 0.5 - 1.5)

REF,ERENCE

, AN01 1102.04, Power Operations, R 18, p. 3 I

i i

I s i

i

.. . u

>\

.s . .- ,

Dt__80MINISIB8IIVE_EB0GEDUBEft_GONDIIIONat_6NQ_LIMII6IIQN2 PAGE 29

't 1 -ANSWERS -- ARKANSAS NU' CLEAR ONE-1 -87/02/10-PELLET, J.

ANSWER 8.01 (2.50) ,

o. ~140 F, b. 1% dK/K subcritical w/ all CR removed , c. </= 200 F,
d. It dK/K subcrit., e. >/= 525 F, f. Keff=1/ critical, g. <2% rx. pwr.

t s (7 answe'rs 9 0.357 ea.)

REFERENCE s .

A%01 TS, Amend. 25, p. 1, 2 ANSWER 8.02 (3.00)

a. 1. 10 gpm, 2. shutdown (2 answers e 0.5 ea.)
b. 1. 1 gpm, 2. shutdown (2 answers 8,0.5 ea.)
c. 1. RCS strength boundary

~

2. O gpm / any amount
3. shutdown and cooldown to cold S/D initiated.

(3 answers t'O.333,ea.)

REFERENCE. c o

AN01 TS 3.1.6, Amend. 57,' p. 27

( .t ANSWER ~8.03 (1.00)

1. Change in SCOPE. ,
2. Change in PURPOSE.
3. Degrade controls prescribed in admin. proc.
4. Peduce level of saf ety (ensured by IC's).

l S. Degrade acceptance criteria.

(eny 4/5 answers e 0.75 ea.)

REFERENCE n i AN01 1000.06, Rev. 19, p. 2 ANSWER 8.04 (1.00)

Manipulation of power supplies to the plant (S/U xformer & generators,

.e., switchyard bkrs/disconn) requires both Switching Orders & HOLD cards.
- 4, 4

- _ - - ~ - , -,_m --_ _ . - . _ -

, , . , ,yy y -- , ~,.s,,-m._ - , _ - , ,

L '

4 Az__8DNINISIB8IIVE_EBQQEQUBESi_QQNDIIIONSt_8NQ_LINII6IIQNg PAGE 30

-ANSWERS -- ARKANSAS NUCLEAR ONE-1 -87/02/10-PELLET, J.

REFERENCE AN01 1000.27, Rev. 2, p. 1 ANSWER 8.05 (2.00)

a. equipment damage, b. danger ( inj ury is XXX)
c. Operating License / TS, d. automatic trip 1

(4 answers 8 0.5 ea.)

REFERENCE AN01 1015.01, Rev. 13, p. 7 ANSWER 8.06 (1.00)

1) Line-of-command mgmt personnel.
2) NRC (Resident Inspector).
3) Support personnel specifically requested (for tech. support).

(cny 2/3 answers 8 0.5 ea.)

REFERENCE AN01 OP 1015.01 i ANSWER 8.07 (1.00)

D.

REFERENCE i TS ANSWER 8.08 (1.00)

Energency Operations Facility Director present. (1.0)

REFERENCE AN01 1903.10, R 17, p. 3 l

l i

i i

2

4 at__aQUINISIB8IIVE_EB9CEQUBEft_GQNDIII9Nat_8NQ_LIMII8IIQN2 PAGE 31 ANSWERS -- ARKANSAS NUCLEAR ONE-1 -87/02/10-PELLET, J.

ANSWER 8.09 (3.00)

c. Alert.
b. S. A. E.
c. Loss of 2/3 with potential loss of 3rd fpb. (3 answers e 1.0 ea. - 3.0)

REFERENCE AN01 1903.10, R 19, p. 16, 28, 40 ANSWER 8.10 (2.50)

a. Yes. (0.5)
b. 1. Declare instrument inoperable.
2. Notify SS.
3. Write a RAC.
4. Initiate any TS LCO requirements. (Any 2/4 e 0.5 ea. - 1.0)
c. 1. Operability not required for plant mode.
2. Component being checked by log entry is in test w/ approved proc.
3. RAC preexisting. (Any 2/3 e 0.5 ea. -1.0) l

! REFERENCE

! AN01 1015.03, R 12, p. 4 ANSWER 8.11 (3.00) l l a. 525 L. 1. 45

2. 305 l

l c. 2 (of 3) l

d. 1. 15
2. 15 (6 answers e 0.5 ea. - 3.0)

REFERENCE AN01 TS 3.1.3, p. 21 l

1 l

~

q At__6DdINISIB6IIVE_EBQGEDUBEft_GQNDIIIDN2t_6NQ_LIMII6IIQN1 PAGE 32 ANSWERS -- ARKANSAS NUCLEAR ONE-1 -87/02/10-PELLET, J.

I ANSWER 8.12 (2.00) i Gross activity consists primarily of noble gases (0.667), which are continuously removed f rom the secondary activity and released (0.666).

Therefore, in the event of a DBA (MSLB) there is minimal of fsite dose casociated with gross activity in the secondary coolant (0.667). CONCEPT!

REFERENCE AN01 TS 3.1.4, 3.10, p. 23, 66

.- .__ - -__ _ . _ _ - . - _ _ - - _ _ . _ _ _ . _ . ~ _ . _ . , _ .

TEST CROSS REFERENCE PAGE ~1 QUESTION VALUE REFERENCE.

05.01 1.00 JJP0001564 05.02 1.50 'JJP0001565 05.03 2.00 JJP0001566 05.04 2.50 JJP0001567 05.05 1.00 JJP0001568 05.06 2.50 JJP0001569 05.07 2.50 JJP0001570 05.08 1.00 JJP0001571 05.09 2.50 JJP0001572 05.10 2.50 'JJP0001573 05.11 3.00 JJP0001574 05.12 3.00 JJP0001575 25.00 06.01 2.00 JJP0001576' 06.02 3.00 JJP0001577 06.03 1.50 JJP0001578 06.04 3.00 JJP0001579 06.05 1.50 JJP0001580 06.06 3.00' JJP0001581 06.07 2.00 JJP0001582 06.08 2.00 JJPG001583 06.09 2.00 JJP0001584

'06.10 1.00 JJP0001585 06.11 2.00 JJP0001586 06.12 2.00 JJP0001587 25.00.

07.01 1.50 JJP0001588

, 07.02 2.00 JJP0001589

07.03 3.50 JJP0001590 l

07.04 2.00 JJP0001591 07.05 3.00 JJP0001592

07.06 3.00 JJP0001593 I

07.07 3.00 JJP0001594 07.08 3.50 JJP0001595

! 07.09 2.00 JJP0001596 07.10 1.50 JJP0001610 25.00 l

i 08.01 2.50 JJP0001598 08.02 3.00 JJP0001599 08.03 3.00 JJP0001600

, 08.04 1.00 JJP0001601 08.05 2.00 JJP0001602 08.06 1.00 JJP0001603 08.07 1.00 JJP0001604

, a .

TEST CROSS REFERENCE PAGE 2 QUESTION VALUE- REFERENCE 08.08 1.00 JJP0001605 08.09 3.00 JJP0001606 08.10 2.50 JJP0001607

~08.11 3.00 JJP0001608 08.12 2.00 JJP0001609 25.00 100.00 l

/

l l

f I

l

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ARKANSAS POWER & LIGHT COMPANY Arkansas Nuclear One Reeves E. Ritchie Training Center P. O. Box 608 Russellville, Arkansas 72801 February 12, 1987 ,

ANO-87-01870 Mr. John Pellet U. S. Nuclear Regulatory Commission 611 Ryan Plaza Drive Suite 1000 Arlington, TX 76011

SUBJECT:

Arkansas Nuclear One - Unit 1 Docket No. 50-313 License No. OPR-51 SRO License Exam Comments

Dear Mr. Pellet:

Attached are comments concerning the Senior Reactor Operator license examination give at this facility on 2/10/87.

Very truly y s, l C Ed A. Force Training Supervisor Unit 1 Operations EAF:rab Attachments cc: ANO-DCC MEMBEA MICOLE SOUTH UTILITIES GYSTEM

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5.07 a)

Reference:

B&W AT0G, Part II, Vol. I, page 20.

ANSWER Reducing EfW flow rate will reduce effects noted in' answer to question 5.03 (same reference), reducing priamry to secondary heat transfer as long as some EFW flow is maintained the decreasing level should have no effect on natural circulation.

RECOMMENDATION Accept answers in line with above comment.

5.12 c)

ANSWER Also a reduction of boron would have an effect overcore life.

RECOMMENDATION Consider comments above as an addition to the key answer.

6.01 a)

ANSWER Although the coupling dimensions are different between APSRs and CRAs the fuel bridge will still couple to both. If an interchange occurred the different dimensions would not be noted until the head is in place and coupling of drives _is in progress. Administrative control prevents interchanging control assymbles.

RECOMMENDATION l

Accept alternate answer above.

6.03 ANSWER Tho top of each hot leg could also be an acceptable answer.

1 l RECOMMENDATIONS Accept alternate answer listed above.

l ATTACHMENT TO ANO-87-01870

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.6.04 e and g)

Reference:

OP1103.06, page 14.

ANSWER

, Answers e and g are incorrect. Correct answers are:

! e. 9"

g. 6.5" RECOMMENDATION ,

Accept correct answers.

6.06 a) j ANSWER 1

This answer is no longer correct. 1R7 added spray lines from the

makeup system as well as the "B" DH loop.

I REC 0t1MENDATION Accept either key answer or alternate answer above.

6.06b

} ANSWER Adminitrative controls and procedures also restrict when Decay Heat

can be placed in service.

I 1

RECOMMENDATION

Consider alternate answer above.

6.08 a and b)

ANSWER This question can be answered without refering to the level indication being delta P signals. The answer should deal with the ICS functions only and not from where this signal is developed.

ATTACHMENT TO AN0-87-01870

RECOMENATION Accept any alterante answer which demonstrates a knowledge of ICS high and low level limits.

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6.11 a)

Reference:

1105.08, page 2 (attached).

QUESTION Question is unclear as to what is required.

RECOMMENDATION Using attached procedure and copy of training summary accept any

, correct answer which states the concept.

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l 7.01

Reference:

1102.08, page 1,2,4 (attached) i j ANSWER i

Answers given in the key do not address all possible cases.

RECOMMENDATION 1 Using procedure accept any correct alternate answer.

4 I

j 7.05

Reference:

1202.01, Rev. 8 (attached) 4 ANSWER

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Answers 2, 4, 6 and 7 are either incorrect or incomplete. See attached for correct answers.

l RECOMMENDATION I

Accept attached answers as correct.

7.07 c)

Reference:

1102.06, page 4.

ANSWER l

7.2 of the procedure list another possible answer.

4 ATTACHMENT TO AN0-87-01870

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RECOMMENDATION Using procedure accept correct alternate answer.

GENERAL COMMENT

S Examination was well written. .

ATTACHMENT TO ANO-87-01870

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