Exam repts 50-327/95-300 & 50-328/95-300 on 950508-12.Exam ...

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p2 nag UNITED STATESp

'o'4NUCLEAR REGULATORY COMMISSION

* REGION 11

(' o 101 MARIETTA STREET, N.W.. SUITE 2900.$ 8 ATLANTA, GEORGIA 3G3234199

\.....JReport Nos.: 50-327/95-300 AND 50-328/95-300

Licensee: Tennessee Valley Authority6N 38A Lookout Place1101 Market StreetChattanooga, TN 37402-2801

Docket Nos.: 50-327 and 50-328 License Nos.: DPR-77 and DPR-79

Facility Name: Sequoyah Nuclear Plant Units 1 and 2

Examination Conducted: May 8-12, 1995

Chief Examiner: /Q/d 6/WM5JonganH.Bartl Date ' Signed

Accompanying Personnel: Ronald Aiello, RIIPaul Steiner, RII

b JI f[Approved by:Thomas A. Peebles, Acting Chief __Date Signed

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Operatar Licensing SectionOperations BranchDivision of Reactor Safety

SUMMARY

Scope:

NRC examiners conducted regular, announced operator licensing initialexaminations and associated inspection activities during the period ofMay 8-12, 1995. Examiners administered examinations under the guidelines ofthe Examiner Standards, NUREG-1021, Revision 7. Six Senior Reactor Operatorcandidates received written and operating examinations.

"9506200016 950530PDR ADOCK 05000327V PDR'

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Results:

Candidate Pass / Fail:

SRO R0 Total Percent

Pass 6 0 6 100 %

Fail 0 0 0 0%

Examiners identified an unresolved item regarding a declining trend insecurity performance (paragraph 2.b).

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Enclosure 1

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REPORT DETAILS

1. Persons Contacted

Licensee Employees

*B. Adney, Site Vice President*H. Birch, Operations Training Instructor*R. Driscoll, Nuclear Assurance and Licensing Manager*G. Enterline, Operations Manager*T. Flippo, Site Support Manager*W. Hunt, Operations Training Instructor*R. King, Operations Training Manager*K. Meade, Acting Compliance Licensing Manager*R. Proffitt, Licensing Engineer*J. Setliffe, Site Security Manager '

*R. Shell, Site Licensing Manager*M. Shepherd, Training Manager

Other licensee employees contacted included instructors, engineers,technicians, operators, and office personnel. '

NRC Personnel

*W. Holland, Senior Resident Inspector - Sequoyah IS. Schaeffer, Resident Inspector - Sequoyah '

*R. Starkey, Resident Inspector - Sequoyah

* Attended exit interview

Acronyms and initialisms used in this report are listed in the last !

paragraph.

2. Discussion

a. Summary

NRC examiners conducted regular, announced operator licensing initialexaminations during the period of May 8-12, 1995. Examinersadministered examinations under the guidelines of the ExaminerStandards, NUREG-1021, Revision 7. Six SRO-Instant license applicantsreceived written examinations and operating tests. The applicants'written examination scores ranged between 87.6 percent and 93.8percent with an average of 91.2 percent. Although all of thecandidates passed the operating examination, four of the six weremarginal passes. The licensee will be sent complete examinationpackages for the four marginal passes so the candidates can beremediated on the identified weaknesses. The examination teamidentified a URI concerning a declining trend in security performanceand three significant performance deficiencies during the operatingexaminations.

Enclosure 1

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Report Details 2

b. Security Performance

The examiners observed several problems with site security durine thepreparation and administration of the examinations. The examinersobserved an inattentive guard, two instances of failing to follow thesite physical security plan, and a lack of knowledge of how to use thenew card readers.

(1) An SR0 candidate, while being administered a walkthroughexamination, observed a security guard relaxed in a reclinedposition with head back and eyes closed.

(2) The examiners observed that various security guards were notchecking personnel against an access list. One guard checked theaccess list on the last day of the examinations, found that theexaminer was not on the list, and let the examiner pass becausehe was an NRC employee.

The examiners observed that the security guards were not issuingsite visitor badges in accordance with the site physical securityplan.

(3) The examiners observed that some security guards were notfamiliar with processing an escorted visitor through the new card

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readers.

These problems, in conjunction with VIO 50-327,328/94-301-01,| " Unattended vehicle in protected area," and IFI 50-327,328/95-03-07, ;

| " Follow-up to Inattentive Guard," indicate a declining trend inI security performance which could have regulatory significance. These

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items will be followed up by a security inspector as URI 50-327,328/95-300-01, " Declining trend in security performance."

c. Examination Development

The examiners identified problems with the written examination bank,the written examination pre-review, and the simulator passwordprotection feature during the development of the examination. *

(1) The NRC examiners conducted a review of the Sequoyah writtenexamination bank. The examiners identified three areas ofconcern with the bank.

(a) There were systems with no questions written for them. A KAsearch was performed on the exam bank which showedapproximately eight systems or KA topics with no questionsidentified.

(b) There was a non-proportional number of questions on RodControl. Approximately 20 percent of the exam bank dealtwith rod control or reactivity manipulations.

Enclosure 1

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Report Details 3

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(c) There were many questions repeated in the bank making theexam bank seem larger than its actual size.

(2) The NRC identified three invalid questions on the writtenexamination during the post-examination review. Two of thesequestions were recommended as substitute questions by thelicensee and the third was reviewed and approved by the licenseeduring the pre-review. These three questions had no correctanswer and were deleted. Refer to Attachment A for the detailedquestion analysis.

(3) The licensee identified on April 25, 1995, that the passwordprotection feature for scenario IC sets was not working. Thefailure was discovered after two examination scenarios werevalidated on April 24, 1995, and left on the simulator computerovernight. The simulator was used part of that night fortraining the examination candidates. The licensee and examinersreviewed the simulator operation record and determined that noone had accessed the scenario ICs during the night. The licenseetook prompt action to fix the password protection feature so thatit was available during the week of the examinations.

d. Operator Performance

The examination team evaluated the candidates' performance duringsimulator scenarios and JPMs using the guidelines of NUREG-1021," Examiner Standards," Revision 7, Supplement 1, and concluded thatapplicant performance was satisfactory. All candidates passed thesimulator and walkthrough examinations. The candidates passed 56 ofthe 60 JPMs administered. The examiners noted three significantperformance deficiencies during the administration of the operatingexaminations which resulted in four of the six candidates beingmarginal passes.

(1) One JPM required starting an RCP and responding to a highvibration (20 mils) in the RCP. The indication given to thecandidates was annunciator 1-AR-MS-A Window D3, " Vibration &Loose Parts Monitoring ALM," which alarmed when the pump wasstarted and stayed in alarm. Two of the three candidates statedthis was an expected alarm and never referred to the ARI.

(2) Scenario 1 consisted of an increasing severity CCS leak (affectsRCPs and containment spray pumps) with a subsequent large breakloss of coolant accident. Both crews properly transitioned toFR-P.1, " Pressurized Thermal Shock," due to an orange path. FR-P.1, Step 6 RNO, directed starting an RCP if RCS subcoolingmargin was greater than 40 degrees F and all RCPs were stopped.One of the two crews incorrectly implemented this step andattempted to start an RCP even though the RCP did not have CCSand the RCS was at saturation.

Enclosure 1

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Report Details 4

(3) Scenario 1 alp tested the crews' ability to recognize andimplement the transition to ECA-1.1, " Loss of RHR SumpRecirculation," while performing the actions of ES-1.3, " Transferto RHR Containment Sump." This transition was required due to 1ARHR pump being out of service and the failure of 1-FCV-63-73(Train B RHR containment sump valve) to open. One of the twocrews did not recognize the need to transition to ECA-1.1 until16 minutes after both trains of RHR were lost.

3. Action on Previous Inspection Findings (92701)

(Closed) IFI 50-327, 328/95-03-07, " Follow-up to Inattentive Guard."This item concerned a security guard who was observed relaxed in areclined position with head back and eyes closed on February 15, 1995.This issue will be included in the follow-up for URI 50-327,328/95-300-01, identified in paragraph 2.b of this report, and this IFI isclosed.

(Closed) IFI-50-327, 328/94-301-02, This item concerned errors in A0I-29," Dropped or Damaged Fuel Assembly or loss of Reactor Cavity Water Level,"and 0-TI-CEM-030-030.0, " Manual Calculation of Plant Gas, Iodine, andParticuate Release Rates for Offsite Dose Calculation Manual (0DCH)Compliance."

a. A01-29 required tne operator to obtain a gas release rate inaccordance with TI-030 or the P-250 computer. Current plant guidanceissued in a standing order was to perform the TI-030 calculation andnot use the P-250 computer. Also, TI-030 had been renumbered to0-TI-CEM-030-030.0. The examiners reviewed the draft of A0P-M.07," Refueling Malfunctions," Revision 0, which will replace A0I-29 anddetermined it corrected the above errors,

b. Appendix A of 0-TI-CEM-030-030.0, Rev 1, step [2] used the incorrectradiation monitor for Service Building radiation. The procedurerequired 0-RM-90-101B for the calculation of the Service Buildingradiation when it should have been 0-RM-90-1328. The examinersreviewed 0-TI-CEM-030-030.0, Rev 3, and determined that step [2] wascorrected to indicate the correct radiation monitor.

The inspectors considered the licensee's corrective action to be adequate,and this IFI is closed.

4. Exit Interview

At the conclusion of the site visit, the examiners met withrepresentatives of the plant staff listed in paragraph I to discuss theresults of the examinations and inspection findings. The licensee did notidentify as proprietary any material provided to, or reviewed by theexaminers. The examiners further discussed in detail the inspectionfinding listed below. Dissenting comments were not received from thelicensee.

Enclosure 1

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Report Details 5

Type item Number Status Description

URI 50-327,328/95-300-01 Open Declining trend in securityperformance (paragraph 2.b).

5. List of Acronyms and Initialisms

ARI Annunciator Response InstructionCCS Component Cooling SystemIC Initial ConditionIFI Inspector Follow-up ItemJPM Job Performance MeasureKA Knowledge and Ability

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NRC Nuclear Regulatory CommissionRCP Reactor Coolant PumpRCS Reactor Coolant SystemRHR Residual Heat RemovalRNO Response not obtainedSR0 Senior Reactor OperatorURI Unresolved itemVID Violation ,

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Enclosure 1

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NRC WRITTEN EXAMINATION ANALYSIS

Question #8:

This question was recommended by the facility licensee as a replacementquestion for a question they deemed beyond the scope of an SR0 task. Thereplacement question was obtained from the facility examination bank and hasno correct answer.

Which ONE of the following radiation exposures describes aTVA-Sequoyah Administrative Radiation Exposure Limit?

a. 3000 m' ems / quarter gamma exposure to the whole body.b. 5000 mRems/ year gamma exposure to the skin.c. 52 MPC-hours internal radiation exposure in any seven

days.d. 1500 MPC-hours internal radiation exposure per quarter.

The answer listed in the examination bank was b. The licensee explained that5 rems / year gamma to the skin was equivalent to their TEDE limit of 5rems / year for submitting a Planned Special Exposure. 5 rems / year is the NRC10 CFR 20 TEDE dose and does not represent a TVA administrative limit. Thecorrect answer per RCI #3 is 1 Rem / year TEDE. This is the point at which theRadcon Manager would have to grant approval to allow further exposure.

Question 76

This question was modified from the facility examination bank which wasincorrect and the error carried through. The error was not caught by thelicensee examination review team.

Which ONE of the following events / conditions will cause the valueof the estimated critical control rod position to INCREASE?

a. Estimated startup time after a trip from 100% powerincreases from 10 hours to 15 hours.

b. Desired critical baron concentration changes from 1100ppm to 950 ppm.

c. The anticipated Tavg at startup changes from 549F to544F.

d. Reactor Coolant System pressure is increased from 2150psig to 2250 psig.

The answer in the key was listed as d. No correct answer exists.

Question 91: , , .,

Question #91 was contains no correct answer. The examination review team madea point that "c" was the correct answer not "d" during the examination review.

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Enclosure 1, Attachment A

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Given the following:

- Condenser pressure increasing- Condenser vacuum is 2.6 psia- Turbine load is 75 percent

Which ONE of the following would be the FIRST to automatically occur or beprocedurally required?

a. " CONDENSER VACUUM LOW" annunciator lit, requiring reduction inturbine load.

b. Loss of Steam dump capability,c. " CONDENSER VACUUM LOW" annunciator lit, requiring

manual reactor trip.d. Automatic trip of the main turbine.

Per A0I-ll no action will occur or be required until 2.7 psia is reached inthe main condenser. The licensee stated during the examination review that areactor trip is required. The turbine will not automatically trip until 3.9 -5.4 psia, and at 75 percent the procedure requires the reactor to be trippedat 2.7 psia, this would then require a turbine trip.

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Enclosure 1, Attachment A

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SIMULATOR FACILITY REPORT

Facility Licensee: DPR-77 and DPR-79

Facility Docket Nos.: 50-327 and 50-328

Operating Tests Administered On: December 12-16, 1994

This form is to be used only to report observations. These observations donot constitute, in and of themselves, audit or inspection findings and arenot, without further verification and review, indicative of noncompliance with10 CFR 55.45(b). These observations do not affect NRC certification orapproval of the simulation facility other than to provide information that maybe used in future evaluations. No licensee action is required solely inresponse to these observations.

No configuration or fidelity items were observed while conducting thesimulator portion of the operating tests .

Enclosure 2

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NRC Official Use Only - -

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Nuclear Regulatory CommissionOperator Licensing

Examination

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This document is removed fromOfficial Use Only category on

date of examination.

NRC Official Use Only

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OT -|U. S. NUCLEAR REGULATORY COMMISSION . M a.J o *~

SITE SPECIFIC EXAMINATIONSENIOR OPERATOR LICENSE ' !

REGION 2 (!

CANDIDATE'S NAME::!

FACILITY: Sequoyah 1 & 2 ;[

REACTOR TYPE: PWR-WEC4

l.DATE ADMINISTERED: 95/05/08 |

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INSTRUCTIONS TO CANDIDATE:,

Use the answer sheets provided to document your answers. Staple this cover !sheet on top of the answer sheets. Points for each question are indicated in '

parentheses after the question. The passing grade requires a final grade of :at least 80%. Examination papers will be picked up four (4) hours after theexamination starts.

CANDIDATE'STEST VALUE SCORE % !

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NO % TOTALS

FINAL GRADE |

All work done on this examination is my own. I have neither given nor freceived aid. !

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Candidate's Signature j

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,eL,j\ (m< SENIOR REACTOR OPERATOR i,._) Page 1

ANSWER KEY *-

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t 1 3 w s c,MULT1PLE CHOICE 023 dg q

001 d 024 c

002 a 025 d

003 b 026 d

004 d 027 c

005 b 028 c

006 c 029 a

007 b 030 b

iDh008 b 031 b

009 d b 032 b

010 d g 033 a C C Cc011 b 034 b C CC

012 c 035 b

013 d 026 C

014 d 037 c

015 c 038 b

016 b 039 D

017 c 040 c q018 a 041 d

h *D019 c 042 c

020 b 043 a D

021 c 044 c

022 a 045 d

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| 1 3 45 G tt 3 MSGb 069 c046 c

047 d 070 b

[ d C\p 048 b C r 071 a

049 b 072 b

050 b 073 ag051 d 074 b

052 c 075 d

% 076 d C. 9 q053 c

054 c 077 d

055 a 078 b

056 a 079 c

057 b 080 d

058 d ( 081 b

g 082 d D b]_059 a

060 a 083 d b061 b 084 a

(k c, b C, b 085 bh062 a c,

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064 b 087 d

,p065 CC C 088 da

cz b 089 b066 dl

067 b D 090 d q l

068 a 091 c q g

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* ANSWER KEY "

$ b .b H b092 b d

093 c b094 d

095 c

096 b

097 a

098 d

099 c b

100 c

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O OSENIOR REACTOR OPERATOR Page 2w w

ANSWER SHEET

Multiple Choice (Circle or X your choice)

If you change your answer, write your selection in the blank.

MULTIPLE CHOICE 023 a b c d

001 a b c d 024 a b c d

002 a b c d 025 a b c d

003 a b c d 026 a b c d

004 a b c d 027 a b c d

005 a b c d 028 a b c d

006 a b c d 029 a b c d

007 a b c d 030 a b c d

d 4 [*-000 s b c 031 a b c d

009 a b c d 032 a b c d

010 a b c d 033 a b c d

011 a b c d 034 a b c d

012 a b c d 035 a b c d

013 a b c d 036 a b c d

014 a b c d 037 a b c d

015 a b c d 038 a b c d

016 a b c d 039 a b c d

017 a b c d 040 a b c d

018 a b c d 041 a b c d

019 a b c d 042 a b c d

020 a b c d 043 a b c d

021 a b c d 044 a b c d

022 a b c d 045 a b c d

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ANSWER SHEET



046 a b c d 069 a b c d

047 a b c d 070 a b c d

048 a b c d 071 a b c d

049 a b c d 072 a b c d

050 a b c d 073 a b c d

051 a b c d 074 a b c d

052 a b c d 075 a b c d

okkJ053 a b c d 076 a b c d 4pg

054 a b c d 077 a b c d

055 a b c d 078 a b c d

056 a b c d 079 a b c d

057 a b c d 080 a b c d

058 a b c d 081 a b c d

059 a b c d 082 a b c d

060 a b c d 083 a b c d

061 a b c d 084 a b c d

062 a b c d 085 a b c d

063 a b c d 086 a b c d

064 a b c d 087 a b c d

065 a b c d 088 a b c d

066 a b c d 089 a b c d

067 a b c d 090 a b c d

kk W068 a b c d 491 a b c d gg

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ANSWER SHEET



092 a b c d

093 a b c d

094 a b c d

095 a b c d

096 a b c d

097 a b c d

098 a b c d

099 a b c d

100 a b c d

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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS

During the administration of this examination the following rules apply:

1. . Cheating on the examination means an automatic denial of your applicationand could result in more severe penalties.

2. After the examination has been completed, you must sign the statement onthe cover sheet indicating that the work is your own and you have notreceived or given assistance in completing the examination. This must bedone after you complete the examination.

3. Restroom trips are to be limited and only one applicant at a time mayleave. You must avoid all contacts with anyone outside the examinationroom to avoid even the appearance or possibility of cheating.

4. Use black ink or dark pencil ONLY to facilitate legible reproductions.5. Print your name in the blank provided in the upper right-hand corner of

the examination cover sheet and each answer sheet.

6. Mark your answers on the answer sheet provided. USE ONLY THE PAPER PROVIDEDAND DO NOT WRITE ON THE BACK SIDE OF THE PAGE.

7. Before you turn in your examination, consecutively number each answer sheet,including any additional pages inserted when writing your answers on theexamination question page.

8. Use abbreviations only if they are commonly used in facility literature.Avoid using symbols such as e or > signs to avoid a simple transpositionerror resulting in an incorrect answer. Write it out.

9. The point value for each question is indicated in parentheses after thequestion.

10. Show all calculations, methods, or assumptions used to obtain an answer toany short answer questions.

11. Partial credit may be given except on multiple choice questions. Therefore,<

ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.

12. Proportional grading will be applied. Any additional wrong informationthat is provided may count against you. For example, if a question isworth one point and asks for four responses, each of which is worth 0.25points, and you give five responses, each of your responses will be worth0.20 points. If one of your five responses is incorrect, 0.20 will bededucted and your total credit for that question will be 0.80 instead of1.00 even though you got the four correct answers.

13. If the intant of a question is unclear, ask questions of the examiner only.

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14. When turning in your examination, assemble the completed examination withexamination questions, examination aids and answer sheets. In addition,turn in all scrap paper.

15. Ensure all information you wish to have evaluated as part of your answer ison your answer sheet. Scrap paper will be disposed of immediately followingthe examination.

16. To pass the examination, you must achieve a grade of 80% or greater.

17. There is a time limit of four (4) hours for completion of the examination.

18. When you are done and have turned in your examination, leave the examinationarea (EXAMINER WILL DEFINE THE AREA). If you are found in this area whilethe examination is still in progress, your license mLy be denied or revoked.

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QUESTION: 001 (1.00)

In the CVCS Makeup Control System, a blockage of the entire line betweenFCV-62-140, " Boric Acid Flow Control Valve," and FCV-62-144, " Blenderdischarge to charging pump suction header," has occurred.

Which ONE of the following paths to the charging pump suction header isthe only path available to add a combination of boric acid and primarywater?

a. Via FCV-62-128, Isolation to VCT Spray valve flowpath.b. Via HCV-62-929, Manual Emergency Boration valve flowpath. ;

c. Via FCV-62-143, Primary water makeup valve flowpath.

d. Via FCV-62-138, Emergency Boration valve flowpath.|

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GOI-1, " Plant Startup from Cold Shutdown to Hot Standby," contains aprecaution that states, " Avoid changing the RCS temperature by greaterthan 50 degrees F whenever the reactor trip breakers are closed and anycontrol rod bank or shutdown bank is not withdrawn at least 5 steps."

Which ONE of the following is limited or prevented by observing thisprecaution?

The possibility of a stuck rod due to thermal expansion.a.

b. The possibility of a rod ejection accident.

c. Jamming of the movable gripper due to thermal growth.d. Reactivity changes from moderator temperature coefficient when

adequate rod trip reactivity is unavailable.

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Which ONE of the follewing describes an effect on the rod control systemupon receiving a Rod Control System Power Cabinet " Urgent Failure"alanm?

Stationary and movable grippers de-energize to prevent roda.

motion.

b. Non-affected banks can be moved in bank select.

c. Only automatic rod motion la prevented.

d. The lift coil energizes to prevent rod motion.


Which ONE of the following describes the MINIMUM actions necessary toreset a main steam isolation signal?

a. Ensure all MSIV's indicate CLOSED, then pull all MSIV hand-switches to RESET.

b. Ensure all MSIV hand-switches are in the CLOSE position, thenpull all MSIV hand-switches to RESET.

c. Ensure all MSIV's indicate CLOSED, then pull only the loop 1MSIV hand-switch to RESET.

d. Ensure all MSIV hand-switches are in the CLOSE position, thenpull only the loop 1 MSIV hand-switch to RESET.

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Which of the following is a true statement regarding theresulting adjustment of the power range channels based on acalculated calorimetric?

a. If the feedwater temperature used in the calorimetriccalculation was lower than actual feedwatertemperature, actual power will be higher than indicatedpower.

b. If the reactor coolant pump heat input used in thecalorimetric calculation was neglected, actual powerwill be less than indicated power.

c. If the steam flow used in the calorimetric calculationwas lower than actual steam flow, actual power will beless than indicated power.

d. Caution must be taken in adjusting the power rangechannel gamma compensating voltage becauseovercompensating will cause actual power to be higherthan indicated power.


Which ONE of the following is assured by limiting the quantity ofradioactive material in the Liquid Holdup Tanks (Condensate StorageTank, Steam Generator Layup Tank, Outside Temporary Tanks for

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Radioactive Liquid) to within the Technical Specification limits? '

a. Assures that in the event of an uncontrolled release theresultant off-site radiation dose exposures will be limited to asmall fraction of the 10 CFR 100 limits.

b. Assures that during any uncontrolled release of the tankcontents the total body exposure of an individual at the nearestexclusion area boundary will NOT exceed 0.5 rem.

c. Assures an uncontrolled release of tank contents will NOT exceedthe 10 CFR 20 limits of the nearest potable water or surfacewater supply in an unrestricted area.

1d. Assures that any liquid radwaste releases of tank contents will

NOT exceed the 10 CFR 50 limits for radioactive effluent in theprotected area.

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Which CNE of the following is the minimum required by SSP-12.3," Equipment Clearance Procedure" when preparing a clearance requiring useof an air-operated valve that fails open as a boundary valve?

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a. The valve shall be closed with air, and its air supply solenoid'

shall be electrically isolated and appropriately tagged.

b. The valve shall be closed and blocked with an installed jackingdevice which is appropriately tagged.

c. The valve shall be closed manually, with its air supply isolatedand appropriately tagged both mechanically and electrically.

d. The Maintenance Supervisor's specific approval shall beobtained.

,,/' L WGQUESTION: 008 (1.00)

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Which ONE of the following radiation exposures describes aTVA-Sequoyah Administrative Radiation Expf3sure Limit?

3000 mRems/ quarter g/amma posure to the whole body.a.

5000 mRems/ year,ganIna exposure to the skin.b.

52 MPC-hours' internal radiation exposure in any sevenc.

days. //

d. 1,50D MPC-hours internal radiation exposure per quarter.

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Which ONE of the following is the maximum expected dose at the exclusionboundary that should be received by a person following an inadvertentrelease from a Waste Gas Decay tank?

a. A total body exposure of 0.002 rem.

b. Not more than the 10 CFR 100 icdine limit over a one hourperiod.

c. A small fraction of the 10 CFR 100 iodine limit over a two hourperiod.

d. A total body exposure of 0.5 rem.


Which CNE of the following describes how the Spent Fuel Pool CoolingSystem (SFPCS) is connected to the RHR System for core cooling during" flood mode" operations with the vessel head removed for refueling?

a. Normally locked closed isolation valves on the outlet of theSFPCS heat exchangers are used to provide a path to the outletof the RHR heat exchangers.

b. Normally locked closed isolation valves on the inlet to the ,

SFPCS heat exchanger are used to provide a path to the inlet of !the RHR heat exchangers. '

Ic. A spool piece, which connects to the outlet of the SFPCS heatexchangers, provides a path to the outlet of the RHR heatexchangers.

d. A spool piece, which connects to the outlet of the SFPCS heatexchangers, provides a path to the inlet of the RHR heatexchangers.

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The following indications are received in the Unit 1 control room whileat full power:

- Loss of Power Range Channel N-42- Loss of Auto Makeup to VCT.

Loss of which ONE of the following 120 VAC vital instrument power boardswould cause these indications?

a. 1-I

b. 1-II

c. 1-III

d. 1-IV


After a reactor trip and safety injection on Unit 1, the followingconditions are observed for Emergency Diesel Generator 1A-A:

- Diesel Generator Fail to Run alarm - not energizedGreen engine running light above panel 1-M-1 - energized

- Red engine running light above panel 1-M-1 - notenergized

- " Diesel Gen 1A-A Running > 40 RPM", annunciator LIT- Diesel generator 6.9KV breaker - open- Diesel generator voltage - zero

Which ONE of the following correctly describes the response of EDG 1A-A?

a. EDG 1A-A started and then shutdown after 10 seconds.b. EDG 1A-A did not receive a start signal.

EDG 1A-A started but engine speed did not increase abovec.

550 rpm.

d. EDG 1A-A started, engine speed exceeded 550 rpm but did not increaseabove 850 rpm.

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SENIOR REACTOR OPERATOR Page 13--


Which ONE of the following explains why loads in excess of 2100 poundsshould NOT be moved over the spent fuel pool?

a. Ensure a radiation release from dropping a heavy load into theSFP is less than 10% of the total gaseous activity from any twoirradiated fuel assemblies.

b. Prevent the SFP Bridge Crane from operating with loads in excessof the monorail or manual hoist ratings.

c. Ensure that the activity released from the worst case droppedload accident would not exceed that of two fuel assemblies.

d. Ensures that any possible danage to the fuel storage racks willnot result in a critical array.


Plant conditions:

- At 0810 a small break LOCA occurs requiring a reactor trip andsafety injection.

- At 0829 the break was isolated terminating the LOCA, RCS pressureand inventory begin to recover.

- At 0832, the SI signal is reset and a recovery in progress.- At 0845 the LOCA reinitiated and a loss of all off-site poweroccurred resulting in the loss of power to all SI equipment.

Which ONE of the following describes how REINITIATION of SI will occurin response to the loss of off-site power?

SI will occur automatically upon recovery from the loss ofa.

offsite power.

b. SI will occur automatically when required conditions exist.

SI will occur automatically following sequencing of Emergencyc.

Diesel generators.

d. SI will only occur following manual action by the operator.

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O CJSENIOR REACTOR OPERATOR Page 14-


T<he unit is being runback f rom 85% power due to the loss of a feedwaterpump. Control Bank D group 1 Rod M-12 position remains at 185 steps.Control Bank D group 2 Rod D-12 position is indicating 150 steps and thebank demand is 135 steps.

Which ONE of the following actions is required?

a. Trip the reactor and implement E-0, " Reactor Trip or SafetyInjection".

b. Stabilize the plant and align the mispositioned rods per AOI-2,Malfunction of the Reactor Control System.

c. Shutdown the unit in a controlled manner within 6 hours.

d. Initiate immediate boration.


Unit 2 is operating at 40 % thermal power, Loop 2 RCP trips.

Which ONE of the following describes the initial unit response? l

(NOTE: Assume no operator action.)

a. A reactor trip occurs, loop 2 S/G narrow range water levelswells, and auctioneered Tavg decreases.

b. A reactor trip occurs, loop 2 S/G narrow range water levelshrinks, and auctioneered Tavg increases.

c. A reactor trip will NOT occur, loop 2 S/G narrow range waterlevel swells, and auctioneered Tavg decreases.

d. A reactor trip will NOT occur, loop 2 S/G narrow range water :

level shrinks, and auctioneered Tavg increases.

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Immediately following a reactor trip without a safety injection thefollowing conditions exist:

- Reactor trip breakers open- Reactor bypass breakers open- NIS power is 4% and decreasing- Control rod H-10 indicates 228 steps, with no rod bottom light- Control rod H-6 indicates 228 steps, with no rod bottom light

Which ONE of the following identifies the procedure flow path for thissituation?

a. Immediately enter both E-0, "Rx Trip or Safety Injection" andAOI-34, " Emergency Boration" and initiate emergency boration tothe specified volume while performing immediate actions of E-O.

b. Immediately enter procedure FR-S.1, " Nuclear PowerGeneration /ATWS", and initiate emergency boration per AOI-34," Emergency Boration".

c. Enter procedure E-0, "Rx Trip or Safety Injection" and at step 4(SI verification) go to ES-0.1, " Reactor Trip Response", andinitiate emergency boration per step 5 RNO.

d. Enter procedure E-0, "Rx Trip or Safety Injection" and at step 1go to FR-S.1 and initiate emergency boration per AOI-34," Emergency Boration".

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Given the following:- Loss of power on 120V AC vital instrument board 1-II.- Feedwater flow and steam flow selected to channel II.- Reactor trip did NOT occur- All automatic control of the control rods has been lost.- Automatic control of steam generator water level on the affectedsteam generator (s) has been lost.

- Feedwater pump speed must be controlled manually.

Which ONE of the following describes the cause for these conditions?

a. High flux rod stop due to NI instrument failure, loss of steamflow input to the feed water flow control system, loss of steamflow signal to feed pump speed control circuit.

b. High flux rod stop due to NI instrument failure, loss offeedwater flow input to the feedwater flow control system, lossof feed flow signal to feed pump speed control circuit.

c. Rod stop due to C-5 interlock, loss of input signal to steamgenerator program level, loss of steam header pressure tofeedwater pump speed controller.

d. Rod stop due to C-5 interlock, loss of steam header pressureinput to steam generator level control, loss of steam flowsignal to feed pump speed control circuit.

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Following a control room evacuation due to a fire, control of allcomponents has been transferred to the backup controls. The shutdownboard hand-switch for the 1B-B SI pump is "OFF" when a valid SI signalis received.

Which ONE of the following describes the response of the 1B-B SI pump?

a. Pump will start and remain running until SI is reset.

|b. Pump will start and remain running until its associated breaker

is opened locally.

c. Pump will NOT start but the operator may start the pump usingthe shutdown panel switch without first resetting SI.

d. Pump will NOT start but the operator may start the pump by firstresetting SI and then using the shutdown panel switch.

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The unit is in the process of ramping to full power at 10% per hour withthe following conditions as read on the control boards:

- NIS power 80%- Turbine load 960 MWE- Tavg 585 degrees F- Pressurizer pressure 2225 psig

Which ONE of the following describes plant status with respect toTechnical Specification's limitations?

a. Plant conditions exceed DNB parameter limits; no action isrequired, since cause is due to the power ramp.

b. Plant conditions exceed DNB parameter limits; the requiredaction is to restore temperature to within the specified limitsfor this pressure or reduce thermal power.

c. Plant conditions within DNB parameter limits; no action isrequired.

d. Plant conditions exceed DNB parameter limits; the required |

action is to restore pressure to within the specified limits orreduce thermal power.

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\SENIOR REACTOR OPERATOR Page 19--


A loss of all off site power has occurred. Both emergency diesel I

generators are running carrying 1500 KW each. The plant is being cooleddown per ES-0.2, " Natural Circulation Cooldown."

Which ONE of the following concerns should be addressed before 2 hourshave passed since the loss of off-site power?

a. Boration to cold shutdown must be completed to prevent a restartaccident due to the cooldown.

b. Diesel generator loads must be reduced to assure continuedavailability.

c. Diesel generators loads must be increased to minimize theaccumulation of turbo-charger and exhaust combustibles.

d. SI must be manually actuated to maintain pressurizer level dueto cooldown rate increase.


Which ONE of the following describes the type of leakage and the actionrequired by Technical Specifications if a pressurizer PORV is leaking tothe PRT at a rate of 15 gpm with all other systems operating normally?(Assume no other RCS leakage)

a. Identified leakage that requires shutdown.

b. Unidentified leakage that requires shutdown.l

c. Identified leakage, but does not require shutdown.|

d. Unidentified leakage, but does not require shutdown. !

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SENIOR REACTOR OPERATOR Page 20--


Which ONE of the following is NOT an assumption on which SequoyahNuclear Plant's ALARA program is based?

Radiation easily disrup's the molecular balance in living cells.a.

b. Any amount of radiation exposure is a potential hazard.

Any change due to ionizing radiation has the potential to be ac.

detriment.

d. Internal exposure to radiation is the most detrLnental andminimizing it must be the OVERRIDING goal.


During a refueling outage the NIS intermediate range channels were notadjusted to account for the changes in neutron leakage resulting fromthe core reload. All other NIS channels were properly adjusted asrequired per the Restart Test Program Procedure.

Which ONE of the following will occur during the startup following therefueling outage?

Source and intermediate range overlap will be excessive.a.

b. A rod stop will occur due to the significantly lowerintermediate range high flux setpoints.

c. The intermediate range high flux trips will occur at asignificantly higher power.

d. There will be no source and intermediate range overlap.

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SENIOR REACTOR OPERATOR'

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Which ONE of the following describes the process used to trip the MFPs when aFeedwater Isolation signal is initiated?

The overspeed trip valve is electrically operated to dump the tripa.oil pressure.

b. The overspeed trip plunger is electrically operated to dump tripoil pressure.

A 250V DC trip solenoid is energized to dump trip oil pressure.c.

d. Two separate 125V DC solenoid valves are energized to dump tripoil pressure.


Unit 2 is in Mode 5 with core cooling supplied by the RHR system. If both RHRpumps are lost, which ONE of the following groups of actions should beinitiated in accordance with AOI-14, " Loss of RHR Shutdown Cooling"?

a. Flood the refueling canal, check open the Transfer Tube gatevalve, verify the blind flange removed, adjust spent fuel coolingsystem for maximum cooling.

b. Start both CCPs and open the RHR letdown to CVCS isolation valveFCV-62-83.

c. Establish RCS feed and bleed.

d. Establish a heat sink usiig the S/Gs.

O OSEMIOR REACTOR OPERATOR Page 22w w


Which ONE of the following will occur if 125V DC bus Vital Battery Board IIwas lost?

a. MDAFW pump 1B-B starts on LOW-LOW Steam Generator Level.

b. All trip status lights malfunction.

c. All control air compressors will unload.

d. Normal letdown isolation valves FCV-62-69 and FVC-62-70 failCLOSED.


Given the following conditions:

- Reactor power 90%.

- Rods in AUTOMATIC.

- Power Range channel N41 fails HIGH,

Which ONE of the following will be the INITIAL response of the rod controlsystem with no operator action?

a. All rods will be inserted by a reactor trip.,

b. Rods will step in at 48 steps per minute.

c. Rods will step in at 72 steps per minute.

d. Rods will not move due to the C-2 rod stop being in effect.

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Which ONE of the following is NOT a reason for the Main Generator PCBopening time delay of 30 seconds?

To allow for Emergency Diesel Generators to reach rated speeda.

and voltage following a loss of offsite power.

b. To provide power to the Reactor Coolant Pumps to keep themrunning for at least 30 sec during Loss of Offsite Power.

c. To prevent turbine overspeed.

d. To ensure the start of Natural Circulation and prevent a DNBRviolation.


An emergency event occurs at 1200. Which ONE of the following is the lowestposition in the chain of command with the authority to approve actions thatviolate technical specifications or license conditions in order to protect thesafety and health of the public?

a. Licensed Reactor Operator,

b. Licensed Senior Reactor Operator.

c. Operations Supervisor.

d. Plant Manager.

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All RCP's are secured and the plant is in natural circulation. Per ES-0.2, " Natural Circulation Cooldown," which ONE of the following conditionswould require that subcooling be maintained greater than 100 degreesFahrenheit instead of greater than 50 degrees?

a. All four S/G narrow range levels less than 10%.

b. Less than 3 CRDM f ans running.

c. Pressurizer level less than 15%.

d. Pressurizer backup heaters not available.


Which ONE of the following is the reason that ES-1.3, " Transfer To RHRContainment Sump," contains the caution to restart the SI pumps only ifRCS pressure is less than 1500 psig?

a. Limits amount of time the SI pumps have to run above shutoffhead with the lower NPSH supplied by the containment sump.

b. The SI pump miniflow valves are shut during the recirc mode.

Conserves pump lif e for use during post-LOCA long term cooling,c.

d. Minimize depletion of RWST inventory while transferring torecirc mode.

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- Reactor Trip and SI have occurred.

- Operators are performing a controlled cooldown using E-3, " SteamGenerator Tube Rupture" to mitigate the event.

- A Steam Generator atmospheric dump valve fails OPEN.

Which ONE of the following describes the applicability of the RCP tripcriteria on the foldout page of E-3?

a. RCP's should be tripped during performance of E-3 any time thefoldout RCP trip criteria are met.

b. RCP's should be tripped during E-3 ONLY if the RCP trip criteriaare met at Step 5 of E-3, when the operator is specificallyrequired to check the criteria.

c. RCP's should be tripped during E-3 ONLY if the RCP trip criteriaare met before beginning the cooldown and depressurization.

d. RCP trip criteria are not applicable in E-3 EXCEPT as an RNO ifi

voiding in the reactor head is experienced.

QUESTION: 034 (1.00) |

Which ONE of the following is the basis for the immediate action step to" Verity Turbine Trip" in FR-S.1, " Response to Nuclear Power Generation/ATWS?"

a. Limits the mass lost from the RCS during the transient.

b. Limits the mass lost from the S/G during the transient.

c. Limits the pressure excursion in the RCS during the transient.

d. Limits the pressure differential across the S/G U-tubes during,

the transient. |

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Which One of the following is the basis for reducing Tave to less than 500 Ffollowing a shutdown required by a high Dose Equivalent I-131 level?

a. Slows coolant / fuel reaction rate, immediately reducing thesource term of the activity.

b. Reduces the potential release of activity following a steamgenerator tube rupture.

c. Ensures doses at the Low Population Zone boundary are a smallfraction of Part 100 limits with a 1 gpm primary to secondary leak.

d. Minimizes the iodine spiking phenomena which occurs due to thelarge change in power level from the unit shutdown.



An ATWS has occurred- FR-S.1, " Response to Nuclear Power Generatlon/ATWS" is in progress.- While implementing step #2 " Verify turbine trip":

1) An SI signal is received2) A pressurizer PORV opens2) All rods insert.

Which ONE of the following actions should De performed?

a. Immediately exit FR-S.1, and Perform E-0, " Reactor Trip orSafety Injection.

b. Exit FR-S.1 when the RED path clears, and perform the steps ofE-1, " Loss of Reactor or Secondary Coolant", that close theblock valve to the open PORV.

c. Remain in FR-S.1 until completed or directed to transition toanother procedure.

id. Shut the block valve for the open PORV while '

simultanecusly performing FR-S.1 until completion.

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Which ONE of the following correctly describes the operation of the "RCS toRHR System Supply Valves" FCV-74-1/2?

a. RCS pressure greater than 687 psig causes auto-closure.

b. PHST suction valve 63-1 SHUT prevents opening.

c. RCS pressure greater than 380 psig prevents opening.

d. RCS temperature less than 350 degrees causes auto-opening.



- The Unit is at 50% power.

- I&C is performing a monthly surveillance on Pressurizer Levelchannel II and its associated B/S's are in trip.

- Loss of 120V AC Vital Instrument Power Board 1-I occurs.

Which ONE of the following is a consequence of the loss of power?

a. Reactor Trip and Safety Injection occurs.

b. Reactor Trip occurs and S/G PORVs will control secondary pressure.

All diesel generators start but do not power safety boards.c.

d. Channel I of pressurizer level fails low and no SI or Reactortrip signals are generated.

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|Which ONE of the following describes why the ERCW System is designed suchthat only one pump in each pair (JA-QA, PB-MB, NB-LB, KA-RA) should beoperated at any cae time?

a. Prevents exceeding design flow through any one supply header ofthe system with accident loads being supplied.

b. Prevents dead-heading a weak pump under low flow conditions andpossible overheating of pump bearings.

c. Prevents high flow in CCS and DG heat exchangers that wouldcause baffle damage and water hammer.

d. Prevents any DG from supplying more than one ERCW pump whichwould result in an overload of the DG if loaded with ESF loads.


Which ONE of the following describes the reason for RCP restart in FR-P.1," Pressurized Thermal Shock," if the SI termination criteria cannot besatisfied?

a. Restores PZR spray to allow RCS depressurization in subsequentsteps with ECCS still in service.

b. Equalizes SG pressures to allow simultaneous cooldown of allfour loops in subsequent steps.

c. Mixes ECCS injection water and RCS water to raise the fluidtemperature entering the vessel downcomer.

d. Transfers Core Cooling to forced flow allowing the operators toterminate ECCS when the criteria are NOT satisfied.

O OSENIOR REACTOR OPERATOR Page 29, ,,


Given that the pressurizer pressure channel selector switch (1-XS-68-340) is in NORMAL. Below are listed two (2) possible failures to pressurizerpressure instrumentation:

- Channel I Pressurizer Pressure detector (1-PT-68-340) failed HIGH

- Channel II Pressurizer Pressure detector (1-PT-68-334) failed HIGH

Assume that the failures occur separetely, with the plant stable at fullpower. No operator action is taken. Which ONE of the following statementsbest compares the severity of the plant pressure transient produced by eachfailure.

a. The pressure drop RATE will be smaller for the PT-340 failure.

b. The pressure drop RATE will be the same for the two failures,but the PT-334 failure will result in a lower final plantpressure.

c. The pressure drop RATE and overall plant pressure drop will bethe same for the two failures.

d. The pressure drop RATE and overall plant pressure drop w.ll begreater for the PT-340 failure.

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e ASENIOR REACTOR OPERRTOR Page 30m- =-



- Unit is at 50% power.

- All automatic control systems are in their normal lineup.

- Controlling pressurizer level transmitter LT-339 (Ch.I) sticks atthe program level corresponding to 50% load.

- Assume no operator action is taken.

Which ONE of the following describes the effect on charging flow and PZRlevel when the plant load is increased to 100%?

a. Charging flow increases and actual PZR level remain constant.

b. Charging flow decreases and actual PZR level decreases.

c. Charging flow increases and actual PZR level increases.

d. Charging flow remains constant and actual PZR level decreases.


Which ONE of the following is a design function of the steam generatorpower operated atmospheric relief valves?

Provide cooldown to RHR operating temperatures and pressuresa.

when a loss of steam dump capability exists.

b. Provide for overpressure protection for the worst case loadreduction from 102% to 0% when steam dumps are not available.

c. Provide absorption of reactor heat on a turbine runback when,

steam dumps are not available.|.

d. Provide for cooldown during Steam Generator Tube Rupture withoffsite power available.

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|QUESTION: 044 (1.00)|

Which One of the following is the relative proportion in flow capacity between |

one of the Steam Generator safety valves and one of the atmospheric reliefvalves lifting at 1085 psig?

a. 20

b. 10

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A large break LOCA has resulted in automatic actuation of containmentspray. FR-Z.1 requires that the operators ensure that the ReactorCoolant Pumps are tripped. Which ONE of the following is the basis forensuring RCP's are tripped?

a. Delays the onset of two phase flow.

b. Prevents excessive RCS inventory loss, and precludes coreuncovery if the RCP's were inadvertently tripped later.

c. Reduces the containment high pressure transient by lowering theenergy release rate to containment from forced flow.

d. Prevents RCP motor bearings from overheating on loss ofComponent Cooling Water.

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Which ONE of the following events requires notification of the NRC within 1hour?

a. Event or condition which results in manual or automatic reactortrip without actuation of ESF system.

b. Any liquid effluent release that exceeds 20 times the minimumpermissible concentration at the point of discharge.

Any loss of Special Nuclear Material other than nornal operatingc.

loss.

d. A worker is transported to a local hospital with LiOH burns to skin.


While acting as safety observer for a confined space entry that is determinedto be Immediately Dangerous to Life or Health (IDLH), you notice the worker isslumped over in the area and apparently unconscious. Your required action isto perform which ONE of the following:

a. Don Self Contained Breathing Apparatus and retrieve theindividual.

b. Perform an emergency entry and retrieve the individual.

c. Sample the environment and if conditions permit retrieve the Iindividual. |

|d. Call for emergency rescue personnel and wait for them to

retrieve the individual.

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Given the following conditions:- The Unit is in Mode 4.- Shift turnover is in progress.- You are the oncoming SOS.- A radiological technician (RT), has called and will be one hour

late.- The offgoing crew have all worked 16 hours.

Which ONE of the following is the necessary course of action for staffing theoncoming shift in accordance with Technical Specifications (TS)?

a. No action required because an RT is not required for Mode 4.

b. No action required because TS allow the crew to be short an RT forup to two hours to account for unexpected absences.

Have the Plant Manager authorize performance of RT duties by ac.

licensed SRO until the late individual arrives.d. Obtain the Operations Superintendent's authorization to hold the RT

over from the offgoing crew.


Which ONE of the following is NOT an allowable method of performing thesecond check of a valve position for Independent Verification purposesafter a valve is initially positioned OPEN.

Verify the position of a motor operated throttle valve in a verya.

high radiation area by verifying a specified flow rate in the maincontrol board and at an accessible local flow meter.

b. Verify the position by operating the valve in the open direction todislodge the valve from its seat until no further counter-clockwise

,

|rotation occurs.

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c. Use of remote position indicator (provided that periodic testing (has shown the indicator to be consistently accurate).

d. Visually verify stem height position if the stem is physicallymarked or other positive verification method is available.

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SENIOR REACTOR OPERATOR Page 34-- '"


Which ONE of the following is a POTENTIAL result of placing a new and UNUSEDMixed Bed Demineralizer in service?

LiOH concentration decreases in the primary.a.

b. OTdeltaT Turbine Runback.

c. Automatic outward rod motion occurs.1

d. Fuel cladding corrc7 ion increases.

QUESTION: 051 (1.00)l

Which ONE of the following describes the plant response to a rising vaccuum !condition, from 25" Hg vaccuum to 30" Hg vaccuum while operating at 100%? ji

a. Condenser pressure increasingi

b. Reactor thermal power increasing

c. Hotwell temperature increasing

d. Generator megawatt meter increasing

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- The unit is operating at 30% steady state reactorpower.

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- An Instrument Maintenance (IM) mechanic receives permission toperform a calibration on PR N-41.

- The IM mechanic mistakenly pulls the Instrument and ControlPower fuses on N-42. Realizing his mistake, he reinserts thefuses for N-42 and pulls the fuses for the correct channel, N-41, causing a reactor trip.

Which CNE of the following statements describes the reason for the reactortrip?

a. PR neutron flux low setpoint trip.

b. SSPS General Warning.

c. PR rate trip.

d. PR neutron flux high setpoint trip.


Which ONE of the following is the reason for removing the Ice Condenser |

Units air handling system from service whi]a in a RED PATH condition per '

FR-Z.1 "High Containment Pressure?"

a. Operating fan units are susceptible to failure at high pressurecreating a projectile hazard.

1

b. The fan units cannot move compressed air at high pressure. |c. Operating fan units may cause hydrogen accumulation to occur in the

ice condenser ductwork.

d. The fan units are no longer needed for containment pressurereduction.

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Which ONE of the following will NOT result in a feedwater isolation signal?a. Containment pressure of 2.71 psid.

b. Pressur.1zer pressure of 1850 psig.

c. Pressure drop of 100 psig on 1 of 4 S/G's.

d. Tave of 540 F coincident with P-4.


Which ONE of the following describes the position of Unit l's AuxiliaryFeedwater LCV's on a loss of air?

a. MDAFW LCV's fail OPEN, TDAFW LCV's fail CLOSED.

b. TDAFW LCV's fail OPEN, MDAFW LCV's fail CLOSED.

c. MDAFW bypass LCV's fail OPEN.

d. Both MD & TD AFW LCV's fail OPEN.


During normal CVCS operation at full power, which ONE of thefollowing is an abnormal condition that would require operatoraction to correct?

A. VCT pressure at 13 psig

B. Letdown heat exchanger exit temperature is 126.7 degrees F.

C. The RCP seal injection water temperature is 120 degrees F andflows to the seals are 8 gpm/ pump

D. RCP seal differential pressure is 500 psid.

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Which ONE of the following would be the correct use of E-0," Reactor Trip or Safety Injection" with a loss of off-site powerand failure of the Emergency Diesels to recover the vital buses?

a. Go immediately to ECA-0.0, Loss of all A/C Power priorto entering E-0.

b. Ensure reactor trip and turbine trip in E-0, and thenGo to ECA-0.0.

c. Go immediately to ECA-0.1, Loss of All A/C PowerRecovery Without SI Required.

d. Complete E-0 IMMEDIATE ACTIONS then go immediately toECA-0.0, Loss of All A/C Power.


Which ONE of the following is addressed in FR-Z.2, " ContainmentFlooding," as a potential source of excessively highcontainment sump levels?

a. Condensed steam from a steam break

b. RWST

c. Raw Cooling Water

d. DI water

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Which ONE of the following is NOT a major action associated withES-0.2, " Natural Circulation Cooldown"?

a. Borate the RCS to hot shutdown concentration

b. Cooldown and depressurize the RCS

c. Monitor for upper reactor vessel head voiding

d. Isolate cold leg accumulators


Which ONE of the following is a basis for maintaining the lowestoperating-loop Tavg equal to or greater than 541 degrees whileachieving Reactor criticality?

a. The Reactor vessel temperature will remain above itsminimum reference temperature (RT) for nil-ductilitytransition (NDT)

b. The Moderator-Temperature Reactivity Coefficient willbe more negative than -4 pcm/ degree.

c. The Control Rod Drive Mechanisms are maintained at atemperature which will prevent rod binding.

d. The differential temperature between the RCS cold legsand the Pressurizer is within limits for sprayinitiation.

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On a unit startup, when reactor power reaches 10%, severalindications must be verified. Without any operator actions,which ONE of the following is NOT observed at 10%.

1a. "P-7 Lo Pwr Rx Trip Blocked" light off. |

b. "I/R Train A trip Blocked" light illuminated.,

1

I c. "P-10 Nuclear At Power" permissive light illuminated.||

d. "P-13 Turbine Not At Power " permissive light off. |



a) Reactor Trip Breakers CLOSEDb) Steam Generators (SG) under nitrogen pressure

Which ONE of the following is the reason that the nitrogen must be vented toatmosphere prior to opening the Main Steam isolation valves (MSIV's)?

a. To prevent an ESF actuation on LOW-LOW SG level due to shrink causedby a differential pressure across the MSIV's.

.

Ib. To prevent nitrogen binding of the MSIV actuators.

c. To prevent the Reactor Protection System from initiating an ESFactuation on high Main-Steam flow.

d. To prevent an ESF actuation on low Main Steam pressure.

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e &SENIOR REACTOR OPERATOR Page 40- --


Which ONE of the following conditions requires action in less than 1hour according to Tech Specs if in Mode 2 on Unit 1?

a. The shutdown margin is 1.6% delta k/k.

b. The boric acid storage tank temperature is 60 degrees F.

c. One Shutdown rod not fully withdrawn.

| d. Primary Containment average air temperature is 140*F inthe lower compartment.


Which ONE of the following is NOT correct concerning the useconventions associated with Emergency Operating Procedures(EOP's)?

Even after a transition to another procedure, the stepsa.

started BEFORE the transition was made must still becompleted, but not to delay the transition.

b. Continuous warnings contained in a caution (step) areNOT in effect when an operator is referenced to aprocedure to be performed concurrently with the EOP ineffect.

|1

c. If a caution statement occurs before step one of an EOPit may apply either to the whole procedure or just tothe first step.

d. Unless otherwise specified, a required task need not befully completed before proceeding to the nextinstruction; it is enough to begin the task and havesome assurance that it is progressing satisfactorily.

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O OSENIOR REACTOR OPERATOR Page 41- -


With a fire in the auxiliary building, according to AOI-30, " PlantFire," which ONE of the following is the reason for transferringcontrol of the fire pumps to the shutdown board by placing thetransfer switches to the auxiliary position?

a. to regain control of the pumps if the fire is aroundthe Aux Lighcing Board #2 on Unit 2.

b. the shutdown board is the desired location for controlof the pumps.

c. to prevent a simultaneous loss of all three pumps if a fire occursaround the RCP relay board on the Unit 2 side.

d. to place fire pumps in a continuous operating mode thusensuring adequate fire water pressure.



a) Fuel loading operations in progressb) Initial nucleus of twelve assemblies are loadedc) Exclude the ANTICIPATED change in count rates due to detector

and/or source movement.

Which ONE of the following conditions would REQUIRE fuel shuffle operations tobe immediately stopped?

a. An unanticipated increase in the neutron count rate by a factor of2 on ANY responding nuclear channel during a single loading step.

b. An unanticipated increase in the neutron count rates by a factorof 2 occurs on all responding nuclear channels after loadingtwelve additional assemblies.

c. An increase in c.he neutron count rate by a factor of 10 on ANYresponding nuclear channel after loading twelve additionalassemblies.

d. The neutron count rate on an individual nuclear channel increasesby a factor of 6 during a single loading step.

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O OSENIOR REACTOR OPERATOR Page 42we =-


Which ONE of the choices below best completes the followingstatement?

The Turbine Driven Auxiliary Feedwater Pump.

a. can trip on thermal overload of its trip / throttle valveafter emergency start on 2/4 SG lo-lo levels.

b. can operate properly on 100 psig steam.

will automatically start on LOW-LOW level in any steam generator.c.

d. suction valves will transfer to ERCW when the suctionpressure drops to 16 psig for 6 seconds.


The Tech. Spec. for Containment Vacuum Relief requires that theContainment Vacuum Relief Valves be OPERABLE with an actuationsetpoint of less than or equal to psid. Pick ONE below.

a. .1 psid

b. .15 psid

c. .31 psid

d. 1.5 psid

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O bSENIOR REACTOR OPERATOR Page 43-- --


Which ONE of the following is NOT required by SSP-12.1 to regain ActiveStatus of their NRO license?

a. Have a current / valid NRC license.

b. Complete 40 hours on shift under the direction of andin the position to which the individual will beassigned.

Complete an interview with the Operations Manager orc.

Operations Superintendent.

d. Maintain a personal journal, perform a complete planttour and complete all required shift turnoverprocedures.


Which ONE of the following Reactor trip signals provides theleast protection against violating the DNB-Ratio limit?

Overtemperature Delta-T Reactor trip,a.

b. Overpower Delta-T Reactor trip.

c. NIS Power-Range negative-rate Reactor trip.

1

d. Undervoltage on a Reactor Coolant Pump power bus. ;.

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O O ISEMIOR REACTOR OPERATOR Page 44- --

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Which ONE of the following best describes the result of motoringthe Turbine / Generator?

Overheating of turbine blading because of windage ifa.

the Turbine / Generator is motored.

b. The Reactor Coolant System will go below minimumtemperature for criticality.

c. Excessive current flow in the Generator Statorwindings.

i

d. Environmental limit on Condenser Circulating Waterdischarge temperature will be exceeded.


Which ONE of the following statements below is correct concerning the onsiteelectrical system?

a. The blackout relays (BOX & BOY'S) are placed in the non |

blackout position immediately prior to paralleling the |D/G to the offsite power during recovery from the |blackout.

!1

b. Placing a 43TL switch in test prevents the D/G's fromstarting from a blackout on its associated shutdown !

board. |1

c. An 86 relay actuation will lockout the normal feeder |breaker, diesel generator breaker, utility breaker andselect the alternate feeder to close.

d. The D/G electrical trips are bypassed during an.

|emergency start.

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O OSENIOR REACTOR OPERATOR Page 45- =-

.


Which ONE of the following statements below best describes the Steam DumpSystem C-9 interlock, which is required before the system can be armed withone (or more) Condenser Circulating Water pump is running?

a. The Steam Dump System will arm only after| Main-Condenser pressure is 3.4 psia or less (2/2l pressure switches).

b. The Steam Dump System will arm only afterMain-Condenser vacuum is >23-inches-mercury or more(1/2 pressure switches).

The Steam Dump System will arm only afterc.

Main-Condenser pressure is 2.7 psia or less (2/2pressure switches).

d. The Steam Dump System will arm only afterMain-Condenser vacuum is >30-inches-mercury or more(2/2 pressure switches).


Which ONE of the following would require the trip of theaffected reactor coolant pump?

a. A vibration of two (2) mils.

b. Motor amperes of 610 amps (cold loop minimum volts).

Motor winding temperature of 220 deg F with the reactorc.

coolant system at operating temperature.

d. Upper motor bearing cooling water temperature of 120deg F for more than a two-hour period.

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O OSENIOR REACTOR OPERATOR page 46- -


Which ONE of the following correctly describes the initiation ofan automatic start of the Component Cooling System Pump C-S,while it is aligned normal to supply B-train cooling water?

a. SIS on Unit 1 or Unit 2, after a 20-second time delay.

b. Low B-train discharge header pressure <40 psig.c. A Station Blackout, after a 20-second time delay.

d. 30 seconds after power is restored to "B-Train".

! $

$C9%QUESTION: 076 (1.00)

G IS 95--

Which ONE of the following events / conditions will'bause the valueof the estimated critical control rod position'to INCREASE?

'

Estimated startup time after,a'$ rip from 100% powera.

increases from 10 hours to-15 hours.-

b. Desired critical boron' concentration changes from 1100''ppm to 950 ppm.

.

c. The anticipated avg at startup changes from 549F to544F. ,/

d. Reacpaf Coolant System pressure is increased from 2150/psig to 2250 psig.

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O OSENIOR REACTOR OPERATOR Page 47=- --


While draining the RCS to MID-LOOP, only ONE RHR pump operationis allowed. Which ONE of the following is the basis for thisrequirement?

a. Prevent uncontrolled RCS drainage by limiting maximum flow.

b. A lower flow rate reduces the probability of thermal shock.

c. Prevent one pump dead heading the other pump.

d. Minimize the effects of vortexing in RHR suction line.


Which ONE ot the following separate events will cause Channel IIOT delta T setpoint to decrease? Assume init ially at 100% power,all control systems in auto except control reis which are inmanual.

Auctioneered high Tavg fails low.a.

b. N42 power range lower detector fails low.

Controlling pressurizer pressure channel fails high.c.

d. Reduce power to 50% with normal pressure andtemperature.

.

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m aSENIOR REACTOR OPERATOR =- "" Pago 48


Which ONE of the following statements describes the purpose of thereactor vessel head vent system:

To depressurize the RCS below SI pump shutoff heada.

during certain types of small LOCAs.

! b. To provide a means of sampling the upper head for'

fission product gases in or to quantify fuel damage inan accident.

To allow the removal of non-condensible gases in orderc.

to enhance natural circulation flow when required.d. To allow the removal of gases that came out of solution

during RCS draindown to mid-loop operations.


What condition will cause Auxiliary Air Compressor A-A to shutdownif automatically started due to low air pressure?

a. Oil Level high

b. Jacket Water temperature 180 deg F.;

|c. Discharge temperature 430 deg F.|

d. Reciever pressure >100 psig for 5 minutes.

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O bSEMIOR REACTOR OPERATOR Page 49=r '-



a) Loss of all AC power in progressb) Excess Letdown was in service prior to the event and not isolated

during verification of RCS isolation per ECA 0.0, " Loss of all AC"

Which ONE of the following leak paths could this Excess Letdown valvemisalignment create:

a. PRT via the letdown line relief valve. RCS inventoryloss could increase, reducing the time to PRToverfill / rupture.

b. PRT via the RCP seal return relief valve. RCSinventory loss could increase, reducing the time tocore uncovery.

RCDT via the RCP controlled leakage seals. RCSc.

inventory loss could increase, possibly leading to coreuncovery.

d. RCDT, increasing the RCS leakage rate until it isterminated by automatic isolation upon PZR low level.


Unit 1 was operating at 100% power when a control rod dropped, without causing ;

the Reactor to trip. The Quadrant Power Tilt Ratio is determined to be 1.05.Which ONE of the following statements correctly describes a Tech-Specrestriction placed on power operation over the next several hours?

a. Reactor thermal power must be lowered to less than 50%within 30 minutes.

b. The NIS Power-Range high-flux trip setpoints must belowered to 85% within 2 hours.

c. The Reactor must be in hot shutdown within 4 hours.d. Reactor thermal power must be lowered to less than 75%

within one hour.

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a eSEWIOR REACTOR OPERATOR Page 50- -



- A loss of offsite power has occurred on Unit 1.- All EDGs have started and tied on to the vital buses.- All vital loads have sequenced on.- E-0, " Reactor Trip or Safety Injection", has been entered.- While the immediate actions are geing completed, a SafetyInjection (SI) is received.

Which ONE of the following describes the response of the ReactorLower Containment Coolers from the time the SI signal isreceived?

The coolers load shed on the SI signal and sequencea.

back on the vital bus.

b. The coolers load shed on the SI signal and must berestarted manually by the operator.

c. The coolers do not load shed on the SI signal and mustbe secured by the operator to prevent overloading theEDG.

d. The coolers do not load shed on the SI signal andremain energized and running throuhout the transient.


FR-H.1, " Loss of Secondary Heat Sink," has been entered duringresponse to a small-break loss-of-coolant accident. Which ONE ofthe following statements explains why the Reactor Coolant Pumpsare being stopped by the operators?

a. Auxiliary Feedwater cannot be restored. Stopping theRCP's will remove an unnecessary heat-sink load,

b. To prevent RCP seal damage from occurring, due toformation of steam in the No.-1 shaft seal.

c. Auxiliary Feedwater has been restored. Forced coolingis no longer required in the Reactor Coolant System.

d. The RCP's can not be operated during the feed and bleed due toinadequate system pressure.

O Oi SENIOR REACTOR OPERATOR page 51i - ,,


Containment pressure is presently 3.5 psig, following aloss-of-coolant accident. As a MINIMUM, which ONE of thefollowing describes the action necessary to permit CS Pump 1B-Bto be stopped and its handswitch returned to A-Auto?

a. Depress train-A and train-B Containment Spray Pumpreset pushbuttons on Control Panel M-6.

b. Depress the train-B Containment Spray Pump resetpushbutton on Control Panel M-6.

c. Block the train-B Safety Injection Signal and thendepress the train-B Containment Spray Pump resetpushbutton.

d. Reset train-B Phase-B Containment Isolation Signal andthen depress the train-B Containment Spray Pump resetpushbutton.


Which ONE of the following describes why Sodium Tetraborate is added to theice:

a. In conjuction with the Boric Acid in the RCS, createsan acidic solution which promotes iodine hydrolisis tonon-volatile forms.

b. Ensures boric acid does not plate out on fuel rods.

Minimizes hydrogen production in the Zirc-Water reaction.c

d. Maintains Containment Spray and ECCS pH 9.0-9.5.

O bSENIOR REACTOR OPERATOR Dage 52- --


Which ONE of the following corresponds to a RVLIS lower rangelevel of 40% when all RCPs are out of service?

a. Bottom of the reactor vessel head.

b. Top of the hot leg.

c. Top of the core.

d. 3.5 feet above the bottom of the core.


A Waste Gas Decay Tank is being vented to atmosphere using thenormal gaseous waste discharge path. If high gaseous activityexists in this tank, which ONE of the following radiationmonitors would alarm?

a. Auxiliary Building Ventilation Monitor [RE-90-101]

b. Containment Purge Air Exhaust Monitor [RE-90-130]

c. Service Building Ventilation Monitor [RE-90-132]

d. Shield Building Ventilation Monitor [RE-90-400]

O OI| SENIOR REACTOR OPERATOR page 53- -


The plant is operating at 92% power with rod control in AUTO.Which ONE of the following IMMEDIATE actions must be taken per AOI-2," Malfunction of Reactor Control System," if continuous outwardrod motion is observed on control bank D?

a. Shift rod control out of AUTO, insert rods, then verify outward rodmotion stops.

b. Check Tave/ Tref mismatch, if outward rod motion unwarranted,shift rod control out of AUTO and verify rod motior. stops.

c. Check Tave/ Tref mismatch, check for possible ins rument failure(Power Range NIS, T-avg channel, Turbine impulse pressure), ifoutward rod motion unwarranted, shift rod control cit of AUTO andverify rod motion stops.

d. Check Tave/ Tref mismatch, if outward rod motion unwarranted, placerod control to manual or bank select, insert rods to restore Tave toTref, if rod motion does not stop, trip the reactor and enter E-0," Reactor Trip or Safety Injection"


Which ONE of the following would result from a loss of Train "B"auxiliary air?

a. Pressurizer spray valve supplied from loop 1 failsclosed.

h. The atmospheric relief valve for No. 1 S/G fails closed

c. CCS valves to No. 1 RCP and No. 3 RCP bearing oilcooliers fail closed.

d. "B" train Main Control Room ventilation dampers fail tothe emergency pressurization position.

a ASENIOR REACTOR OPERATOR Page 54=- --

QUESTION: 091 (1.00) ~' ' '

/g 6 i'U|)Given the following: |

5 /S/9s- Condenser pressure increasing /- Condenser vacuum is 2.6 psia ,,-- Turbine load is 75% ,e

Which ONE of the following would be the F RST to automatically occur or beprocedurally required?

a. " CONDENSER VACUUM LOW" dnunciatorlit, requiring reduction inturbine load.

b. Loss of Steam dump capability./

c. " CONDENSER VACUUM LOW" annunciator lit, requiring manual reactorftrip. /

,/'d. Aut tic trip of the main turbine.


Given the following plant conditions:

- Pressurizer temperature is 540 F- Pressurizer relief tank (PRT) pressure 14.7 psig- PRT temperature 90 degrees F- Reactor is shutdown

Assume ambient heat losses are negligible and the steam qualityin the pressurizer bubble is 100%. Also assume pressurizer andPRT conditions do NOT change.

Using the provided Mollier diagram which ONE of the following PORV downstreamtemperatures would be caused by a leaking pressurizer PORV?

a. 290 F

b. 310 F;

c. 330 F

d. 350 F

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e &SENIOR REACTOR OPERATOR Page 55-- --


Whicil CNE of the following events does the Sequoyah Probalistic SafetyAssessment identify as the MOST Dominant Sequence?

a. Loss of Emergency Raw Cooling Water System.

b. Loss of 6.9 kv Shutdown Board 1A-A.

c. Total Loss of Component Cooling System.

d. Loss of Offsite Power with Small break LOCA.


Which ONE of the following statements is FALSE?

A. CONCURRENT Verification shall be performed, in lieu ofindependent verification for locked / sealed throttledvalves.

B. INDEPENDENT Verification requirements may be waived forcases of significant radiation exposure, personnelsafety, plant emergency consideration, or otheroperational considerations.

C. CONCURRENT Verification shall be performed, in lieu ofindependent verification, for actions or activitieswhose incorrect performance would immediatelyjeopardize the operation of safety systems.

D. CONCURRENT Verification shall be performed, in lieu ofindependent verification, for che verification ofadequacy and accuracy of clearances to ensure workerand equipment safety.

,

4 ASENIOR REACTOR OPERATOR Paga 56- -


Per Function Restoration Guideline FR-Z.1, the HydrogenRecombiners should not be placed in service above which ONEof the following minimum hydrogen concentrations?

a. 2%

b. 4%

c. 6%

d. 8%


Which ONE of the following indicates the personnel who must comply with the" Fitness for Duty" program at Sequoyah Nuclear Station.

a. All persons who must have escorted access to theProtected Area.

b. Personnel who are notified to report to the CentralEmergency Control Center during a radiologicalemergency.

NRC inspectors who are visiting Sequoyah.c.

d. Offsite emergency Fire response personnel.

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O O|

SENIOR REACTOR OPERATOR Page 57- -

QUESTION: 097 (1.00) l

i

IWhich ONE of the following statements concerning verification isCORRECT?

la. When releasing a clearance INDEPENDENT verification is

NOT required on a device that is not returned to normaldue to another hold order being issued on the samedevice.

b. INDEPENDENT verification is NOT required for removaland return to normal of equipment related to certainsafety-related systems.

c. CONCURRENT verification is required for removal and isrequired for return to normal of equipment related toALL safety-related systems.

d. INDEPENDENT verification is required for removal but isNOT required for return to normal of equipment relatedto certain safety-related systems.


Which ONE of the following CAN NOT be held responsible fordeclaring an Emergency based on the guidance within EPIP-1?

a. Shift Operations Supervisor.

b. Assistant Shift Operations Superviser.

c Site Emergency Director.

d. Site Vice President

O OSENIOR REACTOR OPERATOR Page 58we =r


Which ONE of the following statements does NOT correctly describethe Safety Parameters Display System (SPDS) ?

a. The SPDS displays plant parameters that indicatesafety status of the plant, during all modes ofoperation.

b. The SPDS is designed with consideration for thebarriers against release of radioactivity resultingfrom Reactor fission products.

c. The SPDS Bypassed-and-Inoperable Status Indication(BISI) provides verification that a safety system hasfailed and determines the appropriate functionrestoration guideline that must be implemented.

d. The SPDS provides current plant status to theTechnical Support Center at Sequoyah and to the TVACentral Emergency Control Center in Chattanooga duringa radiological emergency.


Which ONE of the following statements below explains why the CAS must benotified prior to recovering from an Auxiliary Building Isolation(ABI), per SOI 30.5, " Recovery From ABI."

a. To unlock padlocked doors allowing operators access toessential components.

b. To perform an accountability check for all site i

personnel.

c. To prevent traffic flow through the auxiliary buildingdoors to avoid personnel injury.

d. To block access into the main control room in the event i

of a possible Extreme Emergency Mode of Isolation.

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

1

SENIOR REACTOR OPERATOR Page 59

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ANSWER: 001 (1.00) i1

d.

REFERENCE: |

P&ID 2 - 4 */W8 0 9 - 2, Flow diagram CVCS Chemical Control.OPL-271 2022, Objective B.394 SRO eram #1004010A201 [3.0/3.7]

| 004010A201 ..(KA's)|

|

ANSWER: 002 (1.00)

a.

REFERENCE: |)

GOI-1, Rev. 98, page 5 of 50, para L.OPL-271-C049, Rev. 4, page 13 of 34 iOPL-271-C049, Objective: B.3 and B.10 '

94 SRO exam #3001000K104 [3.2/3.4] {

f1

001000K104 ..(KA's)

ANSWER: 003 (1.00)

b.

REFERENCE:

OPL-271-C046, Rev. 4, page 18 of 30, para. DOPL-271-C046, Objective: B.3.94 SEO exam #4001050A201 [3.7/3.9]

001050A201 ..(KA's)

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

- - ___ ____ ________ - ___ ___

SENIOR REACTOR OPERATOR || I Page 60

ANSWER: 004 (1.00)

d.

REFERENCE:

OPL-271-C222, Rev. 2 page 20 of 20P&ID 2-47W611.1, Note .

OPL-271-C222, Objective: B.1494 SRO exam #7

| 013000A402 [4.3/4.4]|

013000A402 ..(KA's)

|

ANSWER: 005 (1.00)

b.

REFERENCE:

0-SI-OPS-092-078.0, EGT206.001 / B.2, Seq Exam Bank #PL-0319(2.6/3.1)

015000K504 ..(KA's)

ANSWER: 006 (1.00)

c.

REFERENCE:

Technical Specifications Bases, 3/4.11.1.4, page B 3/4 11-1OPL-271-C102, Objective: B.294 SRO exam #17068000K401 [3.4/4.1) |

1

068000K401 ..(KA's)

1ANSWER: 007 (1.00)

b.

1

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4 ASENIOR REACTOR OPERATOR Page 61

~ "

REFERENCE:

SSP-12.3, Rev. 8, page 17 of 77, para. 3.1.7.B.1. |

Objective: NONE Found94 SRO exam #15194001K102 [3.7/4.1]

I194001K102 ..(KA's) |

ANSWER: 008 (1.00) ,,/' gpb '

, /'/ $ ) 8 0.5REFERENCE: y/

/ 1RCI-03, OPL271C260-Obj. #B.5, Seq exam bank #PL-0854. |

(2.8/3.4P' )/ ID01K103 ..(KA's)

ANSWER: 009 (1.00)

d.

REFERENCE:

Technical Specification 3.11.2.6 Bases, page 3/4 11-2OPL-271-C102, Objective: B.294 SRO #26000060G007 [3.1/3.4]

000060G007 ..(KA's)

ANSWER: 010 (1.00)

d.

SET 410R REACTOR OPERATOR ||I d|b Page 62w w

REFERENCE:

PLID 2-47W810-1 (C-6), and P&ID 2-47W855-1 (A-7/8)AOI-7, Rev. 21, page 33 of 150, para. V.B.3 & V.B.4Objective: NONE FOUND94 SRO #39033000K102 [2.5/2.7]

033000K102 ..(KA's)

ANSWER: 011 (1.00)

b.

REFERENCE:

AOI-25-2, Rev. 9, page 2 of 6, para. I.B.1 and I.B.7Facility Exam Bank: PL-0328, Obtained from PL-0749Lesson Plan / Objective: OPN-220.02194 SRO #43062000A305 [3.5/3.6]

062000A305 ..(KA's)

ANSWER: 012 (1.00)

c

REFERENCE:

OPL-271-C065, Rev. 5, page 26 of 48, para. C.4.e.OPL-271-C065, Objective: B.1094 SRO exam #44064000A303 [3.4/3.3]

064000A303 ..(KA's)

ANSWER: 013 (1.00)

d.

-._

-_. __ _

SENIOR REACTOR OPERATOR ||b iib Page 63 I. -

REFERENCE:

Technical Specifications, 3/4.9.7, page B 3/4 9-2 )OPL-271-C0137, Objective: B.6 j

94 SRO exam #60000036G004 [2.6/3.8]

000036G004 ..(KA's)

ANSWER: 014 (1.00)I d.

REFERENCE:

ES-1.1, page 3 of 21, Step 1 cautionOPL-271-C257, Objective: B.794 SRO exam #61000056K302 [4.4/4.7]

000056K302 ..(KA's)

ANSWER: 015 (1.00)

c.

REFERENCE:

AOI-2, Rev 16, page 13 of 20, para. IV, CAUTIONOPL-271-C085, Objective: B.4 & B.594 SRO exam #64000005G012 [3.1/3.3]

000005G012 ..(KA's)

ANSWER: 016 (1.00)

b.

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

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

SENIOR REACTOR OPERATOR deb d|b Page 64- -

REFERENCE:

OPL-271-C090, Rev. 2, page 8 of 12, Para. E (instructor notes)SON P&ID 47611-99-6OPL-271-C090, Objective: B.794 SRO exam #67

000015K102 [ 3 . 7 /4 .1]

000015K102 ..(KA's)

ANSWER: 017 (1.00)

c.

REFERENCE:

AOI-34, Rev. 8, page 2 of 9, para. 2.C.OPL-271-C070, Objective: B.194 SRO exam #68000024G011 [3.8/3.9]

000024G011 ..(KA's)

ANSWER: 018 (1.00)

a.

REFERENCE:

AOI-25.2, Rev. 9, page 2 cf 6, para. I.B.OPL-271-C028, Rev 5. page 15 of 40.)

,

OPL-271-C098, Objective: B.5 I

94 SRO exam #75000057A217 [3.1/3.4]

000057A217 ..(KA's)

ANSWER: 019 (1.00)ljc.

|

1

|1

_ _ _ _ _ _ _ _ _ _ _ _

SENIOR REACTOR OPERATOR ||b d|I Page 65;

- -

REFERENCE:

AOI-27, SON EQB: PL-0279OPL271C103, Objective: B.894 SRO exam #77000068A121 [3.9/4.1]

|

| 000068A121 ..(KA's)

| ANSWER: 020 (1.00)

b.

REFERENCE:

Technical Specifications 3.2.5, page 3/4 2-15 and 16OPL-271-C180, Objective: B.2

000027G008 [ 3 .1/ 3 . 6 ]

016000G011 ..(KA's)

ANSWER: 021 '' 00)

c.

REFERENCE:

AOI-35, Rev. 15, page 3 of 21, para. V CAUTIONOPL-271-C257, Objective: B.794 SRO exam #87000055A203 [3.9/4.7]

000055A203 ..(KA's)1

iANSWER: 022 (1.00) li

a. ;

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

SENIOR REACTOR OPERATOR ||b ||h Page 66- -

REFERENCE:

Technical Specification 3/4.6.2.dOPL-271-C072, Objective: B.1

000009G008 [3.2/3.9)

000009G008 ..(KA's)

ANSWER: 023 (1.00)

d.

REFERENCE:

OPL-271-C260, Rev. 4, page 27 of 36.OPL-271-C260, Objective: B.894 SRO exam #99194001K104 [3.3/3.5]

194001K104 ..(KA's)

ANSWER: 024 (1.00)

c

REFERENCE:

OPL-271-C045, Rev. 4, Attachment 1, page 3 of 13, para. SUMMARYOPL-271-C045, Objectives: B.2.b and B.2.c.94 SRO exam #95000033A202 [3.3/3.6]

000033A202 ..(KA's)

ANSWER: 025 (1.00)

d

i

!

)I

l

i

|

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

SENIOR REACTOR OPERATOR O O Page 67- -

REFERENCE:

SON OPL 271C034 page 28, Obj. #B.6.(3.0/3.1)

059000A207 ..(KA's)

ANSWER: 026 (1.00)

d

REFERENCE:

SQN AOI-14, REV 15, page 6.(3.9/4.3)

|

000025K101 ..(KA's)

ANSWER: 027 (1.00)

C

REFERENCE:

SQN AOI-21.2, page 2, rev 13.(3.7/3.9)

000058G005 ..(KA's)

ANSWER: 028 (1.00)

c

REFERENCE:

1) OPL271C046, Obj. B.3, page 212) [4. 5/4.4] .

001000K105 ..(KA's)

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

deb iib Page 68SENIOR REACTOR OPERATOR-

ANSWER: 029 (1.00)

a

REFERENCE:

1) SQN OPL271C047 Obj. 6.a, page 302) [3.4/3.6] .

045010K423 ..(KA's)

ANSWER: 030 (1.00)

b.

REFERENCE

1) SSP-12.01, page 46,2) [3.1/4.4]194001A116 ..(KA's)

ANSWER: 031 (1.00)

b |

|REFERENCE:

EGT206.004 page 6, Obj. B.3(3.9/4.2)

:000074A102 .(KA's) 1

ANSWER: 032 (1.00) |

b |

.

SEMIOR REACTOR OPERATOR deb deb Page 69- -

REFERENCE: I

ES-1.3, page 6(Z.9/4.1)

000011K308 ..(KA's)

ANSWER: 033 (1.00)

a

REFERENCE:

E-3, caution on page 13.( 4 .1/ 4 . 2 )

000038K308 ..(KA's)

ANSWER: 034 (1.00)

b

REFERENCE:

OPL271C143 page 15, Obj. #4(4.4/4.7)

000029K312 ..(KA's)

ANSWER: 035 (1.00)

b

REFERENCE:

Technical Specification Bases B 3/4.8 page B 3/4-6.(2.6/3.5)

000076G003 ..(KA's)

|

|

SENIOR REACTOR OPERATOR ||b d|b Page 70- -

ANSWER: 036 (1.00)

c

REFERENCE:

EPM-4, page 46.( 4 .1/ 4 . 2 )

000029G012 ..(KA's)

ANSWER: 037 (1.00)

c

REFERENCE:

OPL271CO23 page 12, Obj. #4.(3.2/3.4)

000025A206 ..(KA's)

ANSWER: 038 (1.00)

b

REFERENCE:

AOI-25.1, page 3.(4.0/4.3) j

000057A219 ..(KA's)

ANSWER: 039 (1.00)

D|

|;

!

I

|

\

SENIOR REACTOR OPERATOR ||b ||I Page 71- -

REFERENCE:

OPL271C027 page 15.(3.1/3.6)

064000K102 ..(KA's)

ANSWFR: 040 (1.00)

c

REFERENCE:

OPL271C152 page 5 of attachment A.(4.5/4.7)

000040K304 ..(KA's)

ANSWER: 041 (1.00)

d

REFERENCE:

OPL271C019 page 20.(3.6/3.5)

000027A103 ..(KA's)

ANSWER: 042 (1.00)

c

REFERENCE:

OPL271C019 page 34.(3.4/3.8)

000028A202 (KA's)

SENZOR REACTOR OPERATOR ||b d|b Page 72- -

ANSWER: 043 (1.00)

a

REFERENCE:

OPL271C029 page 11(3.3/3.3)

039000K102 ..(KA's)

ANSWER: 044 (1.00)

C

REFERENCE:

OPL271CO29 page 11, (890,000 pph vs 890,000 ppb).(2.1/2.4)

039000K601 ..(KA's)

ANSWER: 045 (1.00)

d

REFERENCE:

OPL271C154 page 16(4.1/4.2)

000011K314 ..(KA's)

ANSWER: 046 (1.00)

c

SENIOR REACTOR OPERATOR iib d|b Page 73- -

REFERENCE:

SSP-4.5, pages 10-13.(3.6/3.8)

194001A105 ..(KA's)

ANSWER: 047 (1.00)

d.

REFERENCE:

SIIM VI, page VI-28.(3.3/3.6)

194001K114 ..(KA's)

ANSWER: 048 (1.00)

b.

REFERENCE:

Technical Specifications Administrative section page 6-2.(3.4)

194001A103 ..(KA's)

ANSWER: 049 (1.00)

b

REFERENCE:

SSP-12.6 pages 10-11(3.6/3.7)

194001K101 ..(KA's)

l

i

I

SENIOR REACTOR OPERATOR dIb d|b Page 74w w

ANSWER: 050 (1.00)

b

REFERENCE:

SOI-62.4 page 37 Caution.(2.8/3.2)

004020A109 ..(KA's)

ANSWER: 051 (1.00)

d

REFERENCE:

AOI-11, I.B. 1 & 2.(2.7/2.9)

000051G011 ..(KA's)

ANSWER: 052 (1.00)

c

REFERENCE:

OPL271C048 App. 1,

(3.9/4.3)

012000K402 ..(KA's)

ANSWER: 053 (1.00)

c.

1j;

|\

|

l

SENIOR REACTOR OPERATOR ( )i Page 75

REFERENCE:

OPL271C154 step 8, page 18.(3.8/4.2)

000069K301 ..(KA's)

ANSWER: 054 (1.00)

C

REFERENCE:

OPL271CO25 page 10 Obj. B.2, OPL271C222 page 19 Obj #2.(3.2/3.4)

059000K419 ..(KA's)

ANSWER: 055 (1.00)

a

REFERENCE:

OPL271C035 pages 12 and 20, Obj. B.4, and B.S.c(3.4/3.5)

061000A207 ..(KA's)

ANSWER: 056 (1.00)

a

REFERENCE:

1-SO-62-1.(3.4/3.3)

000022A101 ..(KA's)

i;

!

I

__

SENIOR REACTOR OPERATOR Page 76

ANSWER: 057 (1.00)

b

REFERENCE:

E-0, page 3(4.0/4.6)

000007K301 ..(KA's)

ANSWER: 058 (1.00)

d

REFERENCE:

FR-Z.2, page 3(3.8/4.2)

000069K301 ..(KA's)

ANSWER: 059 (1.00)

a

REFERENCE:

ES-0.2(4.1/4.4)

000055K102 .(KA's)

ANSWER: 060 (1.00)

a

|1

||

I

I

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

f Ih Page */7SENIOR REACTOR OPERATOR

REFERENCE:

TS 3.1.1.4, page B 3/4 1-2.(2.9/3.8)

001010K106 ..(KA's)

1

AMSWER: 061 (1.00)

b

REFERENCE:

0-GO-2-3, page 17(3.6/3.6)

012000A403 ..(KA's)

ANSWER: 062 (1.00)

a

REFERENCE:

GOI-01, precaution W, page 6.(2.8/2.8)

039000G010 ..(KA's)

ANSWER: 063 (1.00)

c

REFERENCE:

)TS 3.1.3.5, page 3/4 1-20.|(3.7 / 4.1) i

001000G005 ..(KA's)

..

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

SENIOR REACTOR OPERATOR ( ) Page 78

ANSWER: 064 (1.00)

b

REFERENCE:

EPM-04, SEQ exam bank #PL-0229(3.9)

194001A102 ..(KA's)

ANSWER: 065 (1.00)

a

REFERENCE:

AOI-30, IV.G, page 5.(3.3/4.1)

000067K304 ..(KA's)

ANSWER: 066 (1.00)

d

REFERENCE:

FHI-03, precaution AH.2, page 10.(2.5-3.4)

034000K105 ..(KA's)

ANSWER: 067 (1.00)

b

..

.. __

SENIOR REACTOR OPERATOR |h Page 79

REFERENCE:

OPL271C035, page18, Obj. #5.a(3.5-3.7)

061000K405 ..(KA's)

ANbWER: 068 (1.00)

a.

REFERENCE:

TS 3.6.6.1, page 3/4 6-38, OPL271C083 OBJ. B.4(3.5/3.7)

103000K107 ..(KA's)

ANSWER: 069 (1.00)

c

REFERENCE:

SSP-12.1, page 77 OPL271C209 Obj. #B.10(2.5, 3.4)

||

194001A103 ..(KA's)

ANSWER: 070 (1.00)

b

REFERENCE: )

OPL271C048, Appendix 1, page 64, Obj. #B.10.d(3.1 / 3.5)

012000K603 .(KA's)

|

F

SENIOR REACTOR OPERATOR ||b ||h Page 80--

ANSWER: 071 (1.00)

a

REFERENCE:

0-GO-2-3, page 6, precaution 3.2.C, OPL271C324 Obj. #B.2(3.3/3.6)

045050K101 ..(KA's)

ANSWER: 072 (1.00)

b

REFERENCE:

0-45N767-5, OPL271C065, Obj. #b.4(2.2-2.5)

064000K606 ..(KA's)

ANSWER: 073 (1.00)

a

REFERENCE:

1-AR-M4-A, OPL271C030 Obj. # B.10, SEQ exam bank #PL-0451(2.6/2.9) i

1

041020K106 ..(KA's)

ANSWER: 074 (1.00)

b

||

!

|

|

|

|

f'

! SENXOR REACTOR OPERATOR deb ||h Page 81- -

REFERENCE:

1-SO-68-2, Appendix D page 2 of 3, Seq Exam bank, OPL271C018(1.9-2.4)

003000K106 ..(KA's)

ANSWER: 075 (1.00)

d

REFERENCE:

0-45N779-2, OPL271C026 Obj. #B.3.(3.2/3.2)

008010A301 ..(KA's)

ANSWER: 076 (1.00)

d M'

C/YMSREFERENCE:

NuPOP (Nuclear Parameters d Operations Package), SEQ exam bank #PL-0616EGT206.001, Obj. #B.1

(3.9/4.0)

001010 3 ..(KA's)

,

ANSWER: 077 (1.00)

d

REFERENCE:

0-SO-74-1, page 75, section 8.4.1 caution 1. OPL271C141 Obj. B.3(3.4/3.6)

000025G007 ..(KA's)

r-

SENZOR REACTOR OPERATOR iib ||I Page 82- -

ANSWER: 078 (1.00)

b

REFERENCE:

OPL271C020 page 14, Obj. #B.6(3.1/3.2)

016000A205 ..(KA's)

ANSWER: 079 (1.00)

c

REFERENCE:

FR-I.3, OPL271C157 page 10.(2.9/3.2)

002000K403 ..(KA's)

ANSWER. 080 (1.00)

d

REFERENCE:

OPL271C038 page 37 Obj. #B.4(2.2/2.7)

000065A204 ..(KA's)

ANSWER: 081 (1.00)

b.

REFERENCE: |;

ECA-0.0, OPL271C149 page 9 #3, Obj. B.2|

(3.9-4.0) I1

000055G012 ..(KA's)

l

1

I

ISENIOR REACTOR OPERATOR |h Page 83

ANSWER: 082 (1.00)

d

REFERENCE:

TS 3.1.3 action c.3.d page 3/4 1-15, OPL271C180 Obj. # B.2.(3.3-3.8)

000003G003 ..(KA's)

ANSWER: 083 (1.00)

d

REFERENCE:

OPL271CO24 Obj. #b.18( 4 .1/ 4 . 3 )

022000A301 ..(KA's)

ANSWER: 084 (1.00)

a

REFERENCE:

FR-H.1, OPL271C147 page 23 step 7, Obj. B.4 |(4.4-4.6) !

|000054K304 ..(KA's) i

!

|

ANSWER: 085 (1.00)

b

|

I

||

. - - - _

_

SENIOR REACTOR OPERATOR dh Page 84

REFERENCE:,

1-47W611-72-1, SEQ exam bank #PL-09151-47W611-88-10PL271C024 Obj. #B.5

(3.3/3.7)

013000K410 ..(KA's)

ANSWER: 086 (1.00)

d

REFERENCE:1

OPL271C024 page 9, Obj. #B.3(3.1/3.6)

026000K402 ..(KA's)

ANSWER: 087 (1.00)

d

REFERENCE:

OPL271C021 page 14 item 6, Obj. #b.6(3.9/4.3)

000009A238 ..(KA's)

ANSWER: 088 (1.00)

d

REFERENCE:

OPL271C013 page 18, Obj. #B.3(3.6/3.9)

073000K101 ..(KA's)

ANSWER: 089 (1.00)

b

SENIOR REACTOR OPERATOR h Page 85- -

REFERENCE:

AOI-02, section B, OPL271C085 Obj. # B.1(4.5/4.8)

000001A203 ..(KA's)

ANSWER: 090 (1.00)

d

REFERENCE:

AOI-10, page 17, item M OPL271C089 Obj. #B.5(3.1/3.1)

000065G009 ..(KA's)

ANSWER: 091 (1.00)

c

t '

REFERENCE:{j / d$ OCd ,

AOI-11, page 2(3.9/4.1) { |g q3

00,0031A202 ..(KA's)

ANSWER: 092 (1.00)

b

REFERENCE:

WOG ERG, Seq Exam bank PL-1341, EGT206.011 Obj. #B.4(3.2/3.7)

000008K101 ..(KA's)

_ _ _ .

r

SENIOR REACTOR OPERATOR deb I|b Page 86- - |

ANSWER: 093 (1.00)

C.

REFERENCE:

EGT206.010 page 9, Obj. #B.S.(3.5/3.7)

000026K304 ..(KA's)

ANSWER: 094 (1.00)

d

REFERENCE:

SSP-12.6, OPL271C210 Obj. #B.6, Seq exam bank #PL-0404(3.6\3.7)

194001K101 ..(KA's)

ANSWER: 095 (1.00)

c.

REFERENCE:

FR-Z.1 page 8, OPL271C010 Obj. #B.4(3.8)

194001K115 ..(KA's)

ANSWER: 096 (1. 0 0 )

b

(

SENIOR REACTOR OPERATOR ||b d|I Page 87- -

REFERENCE:

SSP-1.6, Seq Exam Bank #PL-675, OPL271C198 Obj. #B.2(4.4)

194001A116 ..(KA's)

ANSWER: 097 (1.00)

a

REFERENCE:

SSP-12.3, Seq Exam Bank #PL-0988, OPL271C328 Obj. #b.11(3.7\4.1)

194001K102 ..(KA's)

ANSWER: 098 (1.00)

d

REFERENCE:

SQN-EPIP-06, OPL271C198 Obj. #B.2, Seq Exam Bank #PL-1031(3.1/4.4)

194001A116 ..(KA's)

ANSWER: 099 (1.00)

c

REFERENCE:

EPM-04, OPL271C043 Obj. #B.1, Seq Exam Bank #PL-1137(3.1/3.4)

194001A115 ..(KA's)

I

l

:

_

r

SENIOR REACTOR OPERATOR O M Page G8- -

ANSWER: 100 (1.00)

c

REFERENCE:

SOI-30.5 page 4 precaution D, Seq Exam Bank #PL-1396, OPL271C222 Obj. #B.2(3.4/3.4)

194000K109 .(KA's)


p- -

,

iib TEST CROSS REFERENCE d|b Page 4- -

SRO Exam PWR Reactor

Organized by KA Group

PLANT WIDE GENERICS

QUESTION VALUE KA

064 1.00 194001A102048 1.00 194001A103069 1.00 194001A103046 1.00 194001A105099 1.00 194001A115098 1.00 194001A116030 1.00 194001A116096 1.00 194001A116094 1.00 194001K101049 1.00 194001K101097 1.00 194001K102

Iv00-194001K103- hG & * /007 1.00 194001K102. E/9/{J0 08~ --

023 1.00 194001K104047 1.00 194001K114095 1.00 194001K115

______

PWG Total 16.00

PLANT SYSTEMS

Group I

QUESTION VALUE KA

063 1.00 001000G005002 1.00 001000K104 ;

028 1.00 001000K105 !'

0010LUA303__)p,bT17 6 ---l-rOO t 'f /P r060 1.00 001J10K106 J

003 1.00 001050A201074 1.00 003000K106001 1.00 004010A201050 1.00 004020A109004 1.00 013000A402

:

085 1.00 013000K410 l005 1.00 015000K504 I

083 1.00 022000A301086 1.00 026000K402 I025 1.00 059000A207 |054 1.00 059000K419055 1.00 061000A207067 1.00 061000K405 l006 1.00 068000K401 |

PS-I Total 19 00

t"~

iib TEST CROSS REFERENCE d|I Page 5w -



PLANT SYSTEMS

Group II

QUESTION VALUE KA

079 1.00 002000K403061 1.00 012000A403052 1.00 012000K402070 1.00 012000K603078 1.00 016000A205020 1.00 016000G011010 1.00 033000K102066 1.00 034000K105062 1.00 039000G010043 1.00 039000K102044 1.00 039000K601011 1.00 062000A305012 1.00 064000A303039 1.00 064000K102072 1.00 064000K606088 1.00 073000K101068 1.00 103000K107

......

PS-II Total 17.00

Group III

QUESTION VALUE KA

075 1.00 008010A301073 1.00 041020K106029 1.00 045010K423071 1.00 045050K101

bbPS-III Total 4______

PS Total 4bbb

EMERGENCY PLANT EVOLUTIONS

Group I

QUESTION VALUE KA

089 1.00 000001A203082 1.00 000003G003015 J.00 000005G012032 1.00 000011K308

(.

||b TEST CROSS REFERENCE ||I Page 6- -



EMERGENCY PLANT EVOLUTIONS

Group I

QUESTION VALUE KA

.

045 1.00 000011K314016 1.00 000015K102017 1.00 000024G011093 1.00 000026K304036 1.00 000029G012034 1.00 000029K312

ew} 519 ?S040 1.00 000040K304'091 E 00 000 0 51 A2 02--- ;

051 1.00 000051G011j

021 1.00 000055A203081 1.00 000055G012059 1.00 000055K102018 1.00 000057A217038 1.00 000057A219065 1.00 000067K304019 1.00 000068A121058 1.00 000069K301053 1.00 000069K301 l

031 1.00 000074A102 1

035 1.00 000076G003 |......

EPE-I Total 24.001;

Group II|

QUESTION VALUE KA

057 1.00 000007K301092 1.00 000008K101087 1.00 000009A238022 1.00 000009G008056 1.00 000022A101037 1.00 000025A206 j

077 1.00 000025G007'

026 1.00 000025K101041 1.00 000027A103024 1.00 000033A202 1

033 1.00 000038K308084 1.00 000054K304027 1.00 000058G005009 1.00 000060G007080 1.00 000065A204090 1.00 000065G009

i ......

EPE-II Total 16.00

1, 1

,-_.

||b TEST CROSS REFERENCE d|I Page 7 1

- -

i

SRO Exam PWR Reaetor

Organized by KA Group!

EMERGENCY PLANT EVOLUTIONS ;

Group III

QUESTION VALUE KA

042 1.00 000028A202 ;

013 1.00 000036G004 )014 1.00 000056K302 !

......

EPE-III Total 3.00......

......

EPE Total 43.00......

......

......

Test Total -10 0 . 00-T7. 0

.

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