IR 05000424/1998301

NRC Operator Licensing Exam Repts 50-424/98-301 & 50-425/98-301 on 980413-15.Exams Results:All Candidates Passed Tests Except for One SRO

ML20216D277
Person / Time
Site:	Vogtle
Issue date:	05/07/1998
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20216D083	List: IR 05000424/1998301 ML20216D079
References
50-424-98-301, 50-425-98-301, NUDOCS 9805220172
Download: ML20216D277 (116)

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EXECUTIVE SUMMARY l

Vogtle Electric Generating Plant Units 1 & 2 l NRC Examination Report No. 50-424/98-301 and 50-425/98-301 l l

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During the period April 13 - 15, 1998. NRC examiners conducted an announced operator. licensing initial examination in accordance with the guidance of Examiner Standards, NUREG-1021. Interim Revision 8. This examination implemented the operator licensing requirements of 10 CFR S55.41, S55.43, and S55.4 Ooerations

. Four Senior Reactor Operator (SRO) Candidates received written examinations and operating tests. All operating tests were administered by NRC operator licensing examiners. The written examination was ,

administered on April 10, 1998, and the operating tests were -

administered April 13-15, 1998. Three candidates passed the examination. One candidate failed the administrative portion of the operating test (Section'05.1).

. Candidate Pass / Fail SRO R0 Total Percent Pass 3 0 3 75%

Fail 1 0 1 25%

. The examiners concluded that candidate performance on the written examination was satisfactory. Overall performance on the operating test was satisfactory with some weaknesses noted in the areas of recognizing adverse plant parameters, and understanding system response (Section 05.3).

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The licensee's cooperation and assistance during the examination

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development and administration added value to the examination (Section 05.2).

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9905220172 990507 PDR ADOCK 05000424 Y PDR

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. Summary of Plant Status-During the period of the examinations Unit I was at 100 percent power and Unit 2 was in'an outag I. Ooerations 05 Operator Training and Qualifications 05.1 General Comments NRC examiners conducted regular, announced operator licensing initial examinations period April 13-15, 1998. Two Senior Reactor Operator (SRO) instant and two SRO upgrade a)plicants received written examinations and operating tests. 4RC examiners developed and administered examinations in accordance with the guidelines of the

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Examiner Standards (ES). NUREG-1021. Interim Revision The licensee administered the written examination on April 10, 1998 in accordance with action ES-402 of NUREG-1021. " Administering Initial Written Examinations." Three candidates passed the examination. One candidate failed the overall examination by receiving an unsatisfactory grade on the administrative portion of the operating test. Detailed candidate performance comments have been transmitted under separate cover for management review and to allow appropriate candidate remediatio .2 Pre-Examination Activities The NRC developed the written examinations and opereting tests. A draft copy of the written examination was forwarded to the licensee for review anj comment. The licensee conducted a thorough review and provided

- valuable feedback to the NRC examiners which was incorporated into the examination. An examination preparation visit was conducted during the week of March 30, 1998. During'this visit the examination team validated the. simulator scenarios and finished the development and-validation of'the Job Performance Measures (JPMs) that were to be used during the operating test. The licensee's cooperation and assistance during this process added value to the examination and was helpful to the proces .3 Examination Results and Related Findinas. Observations. and Conclusions Scope The examiners reviewed the results of the written examination and evaluated the candidates' compliance with and use of plant procedures during the. simulator scenarios and JPMs. The guidelines of NUREG-102 Forms ES-303-3 and ES-303-4 " Competency Grading Worksheets for Integrated Plant Operations." were used as a basis for the operating test evaluation . .

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. Observations and Findinas.

I" The examiners reviewed the results of the written examination and found

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-that_ performance of the candidates was satisfactory. The mean score on the exam was 87.5. ~All candidates. passed the written examination but o demonstrated some generic' weaknesse Five common questions were missed

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by three candidates ( Nos.: 6. 50. 75. 80. 100) and one question (No. 6)-

l was missed by all four candidates. This latter question involved l recognizing the first immediate action in procedure A0P-18003-C. " Rod i

Control System Malfunction." These items as well as others that were missed on the examination should be used for review with the candidates and to provide feedback into the training progra Ovsrall, the candidate's performance on the operating test was satisfactory. However, examiners identified several weaknesses in candidate performance. Several candidates demonstrated a lack of awareness of plant conditions and understanding integrated plant

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response. On one scenario, candidates failed to recognize that the cause for decreasing plant pressure incurred from a stuck open spray valve and consequently took no compensatory measures to prevent a plant trip as a result of a low pressure condition. The malfunction was l removed prior to reaching the trip setpoint. On another scenario, the L crew was performing the required actions of procedure 19000-C. " Reactor Trip or safety Injection", during a feedwater line braak inside-

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containment scenari Reactor coolant system pressure _had decreased to 600 psig before the candidates took action to trip the reactor coolant l Jumps (RCP) as required by the RCP trip criteria of the " FOLDOUT PAGE."

l Juring the performance of a JPM where candidates were recuired to

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initiate backfill of a ruptured steam generator, two cancidates were not completely aware of the effects on steam generator pressure of

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overfeeding the steam generator and commenced feeding it at too high a i

rate of flow. Details of these and other discrepancies are described in each individual's examination report. Form ES-303-1. " Operator Licensing Examination Report." Conclusion The examiners concluded that candidate performance on the written

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examination was satisfactory. Overall performance on the operating test I

was satisfactory with some weaknesses noted in the areas of recognizing adverse plant parameters and understanding system respons V. Manaaement Meetinas X1. Exit Meeting Summary At the conclusion of the site visit.. the examiners met with representatives of the plant staff listed on the following page to discuss the results of the examination None of the material provided to the examiners was identified by the licensee as proprietary.

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PARTIAL LIST OF PERSONS CONTACTED

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Licensee

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R. Brown. Training and Emergency Preparedness Manager

} S. Chestnut. Operations Manager J. Gasser. Assistant General Manager - Operations F. Howard. Training Instructor C. Tippins Nuclear Specialist D. Vineyard. Independent Safety Evaluation Group Supervisor D. Scukanec. Operations Training Supervisor NRC J. Zeiler. Senior Resident Inspector K. O'Donohue. Resident Inspector ITEMS OPENED. CLOSED, AND DISCUSSED Closed None

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SIMULATION FACILITY REPORT Facility Licensee: Southern Nuclear Operating Company - Vogtle Plant Facility Docket Nos.- 50-424 and 50-425 l

Operating Tests Administered on: April 13-15. 1998

. This form is to.be used only to report observations. These observations do ,

not constitute audit or inspection findings and are not, without further l verification and review, indicative of noncompliance with 10 CFR 55.45(b).

These observations do not affect NRC certification or a) proval of the simulation facility other than to provide information tlat may be used in future evaluations. No licensee action is required in response to these observation While conducting the simulator portion of the operating tests the following items were observed:

ITEM DESCRIPTION j NONE

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Enclosure 2

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WRITTEN EXAMINATION (S) AND ANSWER XEY(S) :SRO/RO)

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. NRC Official Use Only

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l Nuclear Regulatory Commission Operator Licensing Examination l

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eg46 This document is removed from Official Use Only category on date of examinatio NRC Official Use Only l

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U. S. NUCLEAR REGULATORY COMMISSION

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SITE-SPECIFIC

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WRITTEN EXAMINATION

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APPLICANT INFORMATION Name: Region: II Date: 4/10/98 Facility / Unit: Vogtle Units 1 & 2 License Level: SRO Reactor Type: W Start Time: Finish Time: )

INSTR,UCTIONS Use the answer sheets provided to document your answer 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 percent. Examination papers will be picked up 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />

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after the examination start All work done on this examination is my ow I have neither given nor received ai Applicant's Signature RESULTS Examination Value 100 Points Applicant's Score Points Applicant's Grade Percent

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ANSWER SHEET

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l , ., Multiple Choice (Circle or X your choice)

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l l MULTIPLE CHOICE 023 a -b c d

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002 a b c d 025 a b c d I

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l 003 a b c d 026 a b c d j 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 i 007 a b c d 030 a b c d 008 a b c d 031 a b c d 009 a b c d .

032 a b c d l 010 a b c d 033 a b c d 011 a b c d 034 a b c d 1 012 a b c d 035 a b c d 013 a b c d 036 a b c d l 014 a b c d 037 a b c d j 015 a .b c d 038 a b c d i *

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

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ANSWER SHEET l

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j ., Multiple Choice (Circle or X your choice)

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046 a b c d 069 a b c d 047 a b c d 070 a b c d I

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048 a b c d 071 a b c d j 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 053 a b c d 076 a b c d 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

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057 a b c d 080 a 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 068 a b c d 091 a b c d

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,. ANSWER SHEET

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4'. 3 If you change your answer, write your selection in the blan a b c d

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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 i

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

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[. NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS . . _ . . .

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During the administration of this examination the following rules apply:.

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i 2,[ PART A - GENERAL GUIDELINES Cheating on any part of the examination will result in a-denial of you j application and/or action against your license.

L 'If you have any questions concerning the administration of any part of- !

l the examination, do not hesitate asking them before starting that part' {

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l l SR0 applicants will be tested at the level of; responsibility of the l senior licensed shift position (1.e., shift supervisor, senior shift !

supervisor. or whatever the title of the position may be). You'must pass every art of the examination to receive a-license or to J continue performing icense duties, f The NRC examiner'is not allowed.to reveal the results of any part of the I examination until they have been reviewed and approved by NRC managemen Grades provided by the facility licensee are preliminary

.until approved by the NRC. You will be informed of the' official- 4 examination results about 30 days after all the examinations are complet PART B - WRITTEN EXAMINATION GUIDELINES' After-you complete the examination, sign the statement on the cover

_ sheet indicating that the work is your own and you-have not received or given assistance in completing the: examinatio . To pass.the examination. you must achieve a grade of 80~ percent or greater. Every question is. worth one poin ' For an initial examination. the time. limit: for completing the examination is FIVE hour . You may bring pens, calculators. or slide rules into the. examination

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roo Use'only black ink to ensure legible copie ' Print your name in the blank-provided on the examination cover sheet and the answer sheet..You may be. asked to provide the examiner with some form of positive identificatio . Mark your answers on the answer sheet provided and do not leave any question blank. Use only the paper provided and do not write on'the.

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back side of the pages. If you decide to change yow original answe I L draw a single line through 'the- error. enter the der, ired answer, and initial the chang . If the intent of a question is unclear. ask.questicas of the NRC

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examinerorthedesignatedfacilityinstructoronJ .._ _ -

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• Restroom trips are permitted, but only one applicant at a time will be l

,s allowed to leave. Avoid all contact with anyone outside the examination !

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room to eliminate even the appearance or possibility of cheatin I

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~ When ycu complete the examination, assemble a package including the examination questions examination aids, answer sheets, and scrap paper and give it to the NRC examiner or proctor. Remember to sign the statement on the examination cover sheet indicating that the work is your own and that you have neither given nor received assistance in completing the examination. The scrap paper will be disposed of immediately'after the examinatio . After you have turned in your examination, leave the examination area as defined by the proctor or NRC examiner. If you are found in this area j while the examination is still in progress. your license may be denied i or revoke . Do you have any questions?

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.,. QUESTION: 001 (1.00)

'Given the following conditions:

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- RCS Wide Range pressure 1620 psig

- Pressurizer Pressure 1725 psig

- RCS Hot Leg temperatures 566 degrees F j

- RCS Cold Leg temperatures 560 degrees F l

- Core Exit Thermocouple temperatures 568 degrees F Which of the following is the correct amount of subcooling for the above listed conditions ? degrees F degrees F 1 degrees F

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During power escalation, the P-10 bistable has just ener ized. An

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operator adjusts Tavg and the resulting transient slight y reduces reactor powe Concerning power range neutron flux-low reactor trip, which of the following is the correct system response ? When power level falls below 10% RTP on 2 of 4 channels the nuclear instrument trips will automatically unbloc When power level falls below.10% RTP on 3 of 4 channels the nuclear instrument trips will automatically unblock, I When power level falls below 10% RTP on 2 of 4 channels the nuclear instrument trip bistable de-energize.

When power level falls below 10% RTP on 3 of 4 channels the

! nuclear instrument trip bistables re-energiz i l

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.. O'JESTION: 003 (1.00)

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'Following a Safe Shutdown Earthquake what source of water would be available to fight a fire in the Diesel Generator building ?

l Condensate system Auxiliary Component Cooling Water system Nuclear Services Cooling Water system Component Cooling Water system

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l l ,, , . QUESTION: 004 '(1.00)

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l' You are assuming the SS duties after a one week absenco.

l Procedure 10004-C provides instructions for shift relief. Which.

i one of the following is required by 10004-C during the shift l relief process.?

I l Each on-going operator shall review the r,arrative 109 l rounds sheets and checklists for all station Immediately after' shift relief, completed individual

position checklists are forwarded to Document Control l through the Unit Superintendent. Each on-going operator shall review the narrative 109 rounds sheets and checklists for his station. The review l shall include narrative _ logs since the last shift worked.

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L Non licensed operators shall make a report to the Control Room when they have assumed shift.

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l . QUESTION: 005 (1.00)

Procecare 00152-C provides a definition of a credible threat to i Vogtle Ch.ctric Generating Plant. Which of the following is NOT DEFINED as a credible threat ?

Physical evidence supporting a verbal threa ' A written threat that is received in the mai l 1 A specific group or organization claims responsibility for a written threa A verbal threat that contains specific information about plant locations or system I l

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l,. QUESTION: 006 (1.00) i Y 'A clearance has been issued and is being installed. The unit L 09erator reports that he has a throttle valve handswitch in t1e OPEN position, but the valve is frozen in the closed positio Which of the following actions would be correct in accordance with

[ procedure 00304-C " Equipment Clearance and Tagging" ? Attach a CAUTION TAG to the handswitch and to the CLEARANCE l stating the conditions, Attach a FUNCTIONAL REl. EASE TAG to the handswitch and reposition the handswitch to the same position as thm controlled componen Attach an additional HOLD TAG to the nandswitch and to the CLEARANCE stating the condition Attach an additional HOLD TAG to the handswitch and reposition the handswitch to the same position as the controlled componen .

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. QUESTION: 007 (1,00)

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Which of the followina is the most preferabie manual method of i

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determining temperature for calculating subcooling margin if the plant process computer is unavailable ?

. The average of all core exit thermocouple readings.

l l The average of the five highest core exit thermocouple reading The average of all active loop wide range hot leg temperature indicators, The single highest reading active loop wide range hot leg temperature indication.

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. QUESTION: 008 (1.00)

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Unit 2 is at 300 degrees F and is in shutdown cooling. An Auxiliary Operator performing a valve lineup verification reports that 2-HV-0606 is found to be 60 percent open. Upon further investigation both 2-FV-0618 and

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'2-FV-0610 are found to be 10 % open. Which of the following actions is required ? (drawing attached)

' HV-0610 should be close HV-0606 should be repositioned to 55 % ope FV-0618 should be close HV-0606 should be repositioned to full ope .

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r ; QUESTION: 009 (1.00)

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Which of the following will result from an Urgent Alarm generated by a rod control power cabinet failure ?

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l " locks up" the affected power cabinet by demanding zero l current to its lift coils and increased current to it stationary and movable coils.

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' " locks up" the affected power cabinet by demanding zero current to its lift coils and decreased current to its stationary and movable coils.

, " locks up" the affected power cabinet by demanding zero current to its movable coils and increased current to its stationary and lift coil , " locks up" the affected power cabinet by demanding zero current to its movable coils and decreased current to its stationary and lift coil ,

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l QUESTION: 010 (1.00)

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Y A plant startup is in progress on Unit 2. The RO is withdrawing control rods to keep RCS Tavg on program during the power ascension. At 90% power, the lift coil fuse for i

Control Rod H8 on Control Bank D blows. What effect will this have on the rod control l system?- Automatic and Manual control rod motion will be'available, but ONLY in the

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inward direction.

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'~ Control rods can ONLY be moved in Manual, Control rods CANNOT be moved from their present position.

DRPI indication will not agree with step counter demand indication for one control rod.

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QUESTION: 011 (1.00)

' 'The RCPs are tripped during a transient that has caused the distribution of voids in the vessel to change. Immediately following the RCPs trip, which of the following will occur with respect to RVLIS indications ? The Dynamic Head Range will increase with reduced flow induced pressure drop and the Full Range will come on scal The Dynamic Head Range will decay with reduced flow induced pressure drop and the Full Range will come on scal The Dynamic Head Range will increase with reduced flow induced pressure drop and the Full Range will not come on scale, The Dynamic Head Range will decay with reduced flow induced l pressure drop and the Full Range will not come on scal l

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QUESTION: 012 (1.00)

' The unit is at 220 degrees F. The secondary temperature is 225 degrees F. The RHR suction isolation valve on train A is close One PORV isolation valve is tagged and deenergized in the closed l position. Which of the following are required to ensure COPS is l operable ? Both Safety Injection pumps must be incapable of injectio l l RCS must be Depressurized and vente j The accumulator isolation valves indicate close No RCP can be restarted.

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.: . QUESTION: 013 (1.00) ,

'" ' Which of the following conditions will trip the Rotary Air Compressors ? Low-Low oil pressure

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b High lube oil temp Low TPCCW pressure  ; High-High intercooler condenser level

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, .3 QUESTION: 014 (1.00)

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'An accident is in progress and the Containment H2 monitor indicates 3.5%.

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Which of the following is NOT a potential source of hydrogen gas in the containment atmosphere.? RCDT SI Accumulators Fuel Claddin; PZR-

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l . QUESTION: 015 (1.00)

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'The following plant ccnditions exist:

- SGTR has occurred on one SG

- 19031-C. ES 3.1 Post SGTR Cooldown Using Backfill is in progress

- Ruptured SG NR level is 25 % 1

- RCS is at 390 degrees F  :

- RCS is at 400 Psig

- Cooling using steam dumps

- RCP #4 in service 19031-C, ES 3.1 Post SGTR Cooldown Using Backfill step 13 requires a return to step 3 if the RCS WR hot leg temperature is not less than 200 F. Step 3 requires the operator reverify adecuate shutdown margin. Why is it necessary to reverify shutcown margin during this procedure ? The RCS temperature change during cooldown will cause significant boron concentration changes due to PZR outsurge, i Charging to maintain PZR level during cooldown will cause significant boron concentration change The feed flow to the ruptured S'G will cause signific.nt boron concentration change _ The auxiliary saray will cause significant boron concentration clange ..

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l , .3 OUESTION: 016 (1.00)

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The plant is presently at 60% power with a )ower ascension in progres Control rods are in AUTO with control bank ) at 165 steps. Shortly after ,

you take the watch, the following alarms and indications occur: l

- POWER RANGE HI NEUTRON FLUX RATE ALERT

- TAVG/ TREF DEVIATION

- POWER RANGE CHANNEL DEVIATION

- Control rods stepping out In accordance with A0P 18003-C " Rod Control System Malfunction" the first action required to be taken is: ? Place rod control in manual and return T-ave to progra Stop any turbine load change Verify motion not caused by T-ave /T-ref deviatio Initiate emergency boration to compensate reactivity additio .

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.. QUESTION: 017 (1.00)

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i While performing 19020-C E-2. Faulted Steam Generator Isolation, you determine that the Critical Safety Function Status is: Containment Red 19251-C, FR- . Core Cooling Red 19221-C. FP . Heat Sink Orange 19235-C, FR- . Integrity Orange 19241-C, FR- . Inventory Yellow 19261-C. FR- . Subcriticality Yellow 19212-C. FR- Which of the following is current order in which the Functional Restoration Procedures must be implemented ?

a, 2.1.4.3. . ,1.3,4. .2, ~

5 .2,3.4. .

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l QUESTION: 016 (1.00)

'Y' 'Given the following information:

-It is April 15, 1998, @ 1400- l-Unit 2 is in Mode RCS drained to 188 feet 6 i '

i l -RCS temperature is 110 degrees l-RCS pressure - atmospheri Reactor was shutdown March 31, 1998 0 180 l-1 fuel assembly remains to be reload /3 of core is new fue I-A total loss of RHR cooling has occurre ;

i Which of the following is correct concerning the amount of -

time it will take to reach core uncoyery ?

Use attached figures from A0P-18019-C. " Loss of RHR"

_ minutes minutes- minutes minutes

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. QUESTION: 019 (1.00)

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Today is March 31, 1998. your shift begins at 7:00 a You are the Reactor Operator. Your recent work history is as follows:

March 26 7:00 am to 3:00 pm (normal shift)

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iarch 27 11:0 to 7:00 am (March 28)(normal shift)

iarch 28 7:0 .

o 10:00 am (held over)

March 29 10:4 am to 11:00 am (relief for operator on shift to

/ I have random drug test)

March 30 10:30 am to 11:00 am (relief for operator on shift to attend required training)

Under these circumstances, which of the following describes the minimum required review for applicable log sheets, round sheets and check lists ? hours hours hours

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. QUESTION: 020 (1.00)

'Given the following:

A system operating procedure (50P) is being performed to place a system in service following maintenance. An error is discovered in the sequence of ste)s in the SOP, which if performed, would result in starting a pump witlout the required seal water. The Jump handswitch is in AUTO.

l Which of the following actions should 3e taken ?

l l Stop performance of the procedure and implement a procedure chang Stop performance of the procedure and place the pump handswitch in the PTL positio Valve in the seal water prior to starting the pump: then submit a l procedure change, Start the pum) with SS approval; then valve in seal water and submit a procedure clange request.

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QUESTION: 021 (1.00)

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[ ' Who, by title, is the minimum authority that can authorize actions

to be taken in accordance with 18015-C SECONDARY PLANT CHEMISTRY,

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upon confirmation of one or mcre chemistry parameters outside normal operating range while in Mode 1 ? General Manager i Unit Superintendent Shift Superintendent l Chemistry Duty Foreman l

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I- 1 j ,.. QUESTION: 022 (1.00) i

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Which of the following is a NON-DELEGABLE duty of the Emergency Director ?

I Deploying radiological emergency team ; Request OSC support for emergency maintenanc Deciding to request assistance.from federal support group Coordinating VEGP Emergency operation :

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QUESTION: 023 (1.00)

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A manual reactor trip was initiated at 5% reactor power and a feedwater isolation signal was generated. Which of the j following must be performed to open the bypass feed reg valves ? i

! Momentarily close the reactor trip breakers onl ] No action required the FWI automatically reset ; Place both train A and train B FWI Reset Switches momentarily to the Reset positio i The Rx Trip Breakers must be closed and then the FWI Reset !

Switches held in the RESET position, t

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, . . , . QUESTION

024 (1.00)

• ' Which of the following is correct concerning Steam Generator I l Water Level Control ? Each steam generator's Steam Flow / Feed Flow mismatch is the i L only signal controlling its Main Feed Regulation Valve positio ] Each steam generator's Steam Flow / Feed Flow mismatch and level i- signals are used to control its Bypass Feed Regulation Valve !

l positio j ' Total. Steam Flow / Feed Flow mismatch controls Main Feed Pump speed.

r Total Steam Flow / Delta P Program controls Main Feed Pump delta P setpoin '

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, .3 . QUESTION: 025 (1.00)

D Given the following conditions:

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A large break LOCA has occurred 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> ago on Unit 1 .,

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Containment pressure is 3 PSIG l

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Containment H2 concentration is 6.3% per the H2 l monitors

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DG1A is supplying 1AA02 I Which of the following is correct concerning Post Accident Hydrogen control using the attached procedure 13130-C ? Dilute the containment hydrogen concentration using the Service Air Syste ~

) The "A" train Post LOCA Electric Hydrogen Recombiner can be '

placed in service if 1AA02 bus loading is monitore The "A" train Post LOCA Electric Recombiner can NOT be placed in service due to the DGIA carrying the 1AA02 bu The hydrogen monitors are unreliable at this poin Three more hours must pass and another hydrogen sample take l-

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-QUESTION

026 (1.00)

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Given the following data: {

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A instrument tech inadvertently deenergized 120 VAC l Vital' Bus 1AY1 I i

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-An Si actuation from low PZR pressure occur j

. i Which of the following describes the expected response of the 1 Unit 1, Train A Diesel Generator (DG) and Lne Train A SI Loads !

with 1AY1A deenergized ? 1 i .The A Train DG will start and the Train A SI Loads will Le sequenced o l

' The A Train DG will not start nor will the Train'A SI Loads !

~be sequenced o The A Train DG wi11 start, however the Train A SI Loads will l

not be sequenced o The A Train DG will not start, however the Train A SI Loads will be sequenced o .

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i QUESTION: 027 (1.00) j

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Given the following conditions- -

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- Reactor Power is 74% j I

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Pressurizer pressure is 2250 psig

- Charging flow is being controlled in MANUAL

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The BACKUP. HEATERS have just ENERGIZE 0 l i

Which of the following is the actual pressurizer level ? l

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,y_ . QUESTION: 028 -(1.00)

Y Which of the following signals will cause HV-9378. " Instrument Air to Containment Isolation Valve", to CLOSE ? Containment. Pressure at 4 psig Containment Radiation Monitor RE-003 in high alarm.

' Instrument Air header pressure of 70 psi Containment Atmosphere Radiation Monitor RE-2562 in high alar .

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. QUESTION

029 (1.00) .

l A large break LOCA has occurred. The control room operators have transitioned from 19000-C to 19010-C. " Loss of Reactor or l Secondary Coolant". The RWST Lo-Lo Level annunciator (ALB06 C4) i

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sounds and RWST level indicates 38%. The USS directs the corn.rol !

l room operators to initiate 19013-C, " Transfer to Cold Leg l Recirculation". The extra o]erator, monitoring CSFST's, then l reports that a valid red pati condition exists on Core Coolin l i The USS should direct the operators to: Perform the first six steps of 19013-C. then transfer to 1 19221-C, " Response to Inadequate Core Cooling". I I Perfonn 19013-C to completion, then transfer to 19221- !

" Response to Inadequate Core Cooling". i Immediately perform the actions of 19221-C, " Response to Inadequate Core Cooling". Immediately perform the actions of 19221-C " Response.to Inadequate Core Cooling", while the extra operator concurrently performs the actions of 19013-C.

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.,. QUESTION: 030 (1.00)

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'Which of the following conditions require an emergency '

- boration to be started ? Shutdown Margin found to be 1.3% delta k' per k in Mode' Boron concentration decreases to 2200 PPM during refueling operation ! One control rod fails to insert on a reactor tri Rod Bank Lo-lo Limit alarms during rapid power decrease, l

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. QUESTION: 031 (1.00)

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Given the following conditions:

- RCS pressure - 2335 psi l l

- RCS Tave - 588.3 deg. J

- The reactor is not trippe I

- The crew is currently in 19211-C. "FR-S.I. Response to l Nuclear Power Generation / ATWT". step i

I Which of the following describes the reason why RCS pressure j should be maintained less than 2335 psig ?  !

I Prevents the Pressurizer Relief Tank from going solid, due '

to an open PORV or PRZR Code Safety, and blowing the rupture

. disc causing a LOCA inside containmen To prevent the Reactor from tripping on high RCS pressure.

To ensure a sufficient amount of boric acid is injected into }

the core to reduce reactor powe i To ensure the Pressurizer Spray valves don't short cycle when the PORV's open to lower RCS pressure, l

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. . QUESTION: 032 (1.00)

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A loss of all AC has occurred. The control room operators have completed the innediate operator actions of 19100-C " Loss of All

.AC Power." and have attempted without success, to restore powe Per procedure 19100-C. the control switches for ESF 4160V loads are placed in the Pull-To-Lock position. The defeat of the auto start for this equipment is designed to prevent which of the following actions ?

- The unnecessary use of RWST and CST inventory that may be l needed for long term cooldown.

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" Overloading of a bus that may not be capable of handling

automatic load sequencing of large electrical loads.

l l An uncontrolled overpressurization of the RCS upon l restoration of vital powe An uncontrolled cooldown of the RCS and possible reactor re-criticality upon restoration of vital power.

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..,. QUESTION:~033 (1.00)

?,'a 1 Given the following conditions: i

- Reactor power is 6% )

- Main Feed pump "B" is in service l

- Main Feed Pump "A" is still tripped )

- AFW is in standby readiness

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- PZR Pressure 2235 l

- Pressurizer Pressure Control select switch is in the 455/456 position o j t

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- IBY1B is deenergized for 2 seconds by an inadvertent I operator action

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Which of the following actions will occur ? (Assume no o)erator I

. action is taken and no instruments remain failed after t1e bus is restored.)

' Both MDAFW pumps would start.

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) PZR Proportional heaters would deenergize and then reenergiz l

~ PZR spray valve would close then ope A reactor trip would occu . I l

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, .3 OUESTION: 034 (1.00) '

Given the following conditions:

- RX Power is 30% at 0126 EST

- A total loss of ACCW has occurred at 0115 EST

- Procedure 18022-C (Loss of ACCW) is being imp'emented l

- The RCP temperatures are being monitored on the IPC

- The RCP vibration is being monitored

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i Which of the following would be the required operator  !

response ? '\ Trip the reactor then trip all RCP' If #1 seal leakoff temperature exceeds 195 degrees F, trip that RCP.

, Trip any RCP that has its thermal barrier isolation valve shu . Trip any RCP with shaft vibration in exces.s'of 5 mil )

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,. QUESTION: 035 (1.00)

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'If all three expected res)onses for _ step 1 of 19000-C "Rx trip or SI" cannot be met, then t1e actions in the RNO must be performe (

Which of the following describes the RNO actions ? I Immediately go to 19211-C " Response to Nuclear Power Generation /ATWT." Manually trip the Rx from the redundant switch at the remote l SD panel: if Rx not tripped, then manually open the supply ;

breakers to NB08 and NB0 I 1 Manually trip the Rx: if riot tripped, then trip using the !

redundant switch on the OMCB. If still not tripped, go to 19211- !

Manually trip the Rx; if not tripped, then trip using the redundant switch on the OMC8 If still not tripped, then Emergency Borate the RCS, l

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.,. QUESTION: 036 (1.00)

'9 ' ' A containment pressure relief is in progress when a containment '

ventilation isolation (CVI) actuation occurs. Which of the following is the cause of the CVI ? Containment area high range monitor RE-005 in High alar Containment area seal table monitor RE-011 in High alar ! Plant vent effluent monitor RE-12442. in High alar <

l Containment area low range monitor RE-002 in High alar j I

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.,_ QUESTION: 037 (1.00)

' '# ' Giver, the following conditions: '

- Unit 2 is in Mode 3

- Tave - 557. degrees F

- A loss of all instrument air has occurred

- The crew enters 18028-C. LOSS OF lilSTRUMENT AIR i

- The crew proceeds to Attachment A. Establishing Charging without Instrument Air

- Charging flow is observed to be 150 gpm l

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Which of.the following is correct concerning Attachment A in j this case ? (See attached Attachment A).  !

1 It will restore VCT leve It places the Positive Displacement Pump in service to control charging flo j i It is designed to reduce and control charging flo It is necessary in order to re-establish RCP seal injection ,

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.., QUESTION: 038 (1.00)

'Given the following conditions:

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- Reactor was at 100% RTP

- All control rods fall into the core

- PZR pressure is now 1900 psig

- PRNIs indicate 4% and decreasing

- 19000-C is implemented

- Operators actuate both manual trip handswitches

- Both RTBs remain closed Which of the following states what the USS should do ? Go to step 2 of 19000- Hold on step 1 of 19000-C until the RTB's are opened

' locall Perform 19211-C. ATWT. and complete all the steps of 19211 before transitioning back to 19000- Perform to 19211. ATWT. and return to 19000-C. after completing step 4 of 19211. Check Rx power <5% and IR SUR not positiv !

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. QUESTION: 039 (1.00)

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'The reactor has just trippe Primary and secondary parameters are:

- RCS Tavg 532 degrees F

- RCS Pressure 1720 psig

- PZR Level 20%

- SG NR Levels 25 % each

- SG Pressure #1 875 psig

- Other SG pressures 890 psig

- BIT flow 190 gpm

- AFW flow 625 gpm What transient is in progress ? Steam generator tube rupture Small break LOCA Steam line break Feed line break

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_ , QUESTION: 040 (1.00)

'# ' Unit 1 has experienced a LOC Containment conditions are as follows:

TIME FOLLOWING LOCA CONTAINMENT PRESS CONTAINMENT RAD LEVEL Hrs psi rad /hr

.03 3 .000.000

.5 1 .500.000 .3 1.000.000 .1 800.000 .3 600.000 .0 400.000 .8 200.000 .6 150,000 .5 100.000 .1 75.000

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1 .5 60.000 l l

Assume linear behavior between tabular points. Which of the following describes the E0P Contain'nent values to be used at 10 minutes after 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after, and 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the LOCA began ? normal; adverse; normal

- adverse: normal: adverse adverse: adverse: normal adverse; adverse: adverse

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., OVEST10N: 041 (1.00) l

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19121-C. "ECA2.1, Uncontrolled Depressurization of All Steam Generators", cautions that a Minimum feed flow of 30 gpm must be maintained to each S/G with a narrow range level of less than 10%.

The basis for this requirement is to minimize ? Additional overcooling caused by feedwater addition, The magnitude of SG level overshoo Thermal shock to AFW component l i Thermal shock to S/G component j t

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,_ OUESTION: 042 (1.00)

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'For which of the falloving events would you NOT expect a process and effluent radiation 'onitoring system oiarm ?

_ Lors-of-Coolant accident, outside of containment Main steam line break RCS to CCWS leak Steam generator tube rupture

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Which of the following describes an event which the ECCS is !

designed to mitigate ? I Main Steam Line Break ATWT )

l Loss of Offsite Powpr Cable Spreading Room Fire  :

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OUESTION: 044 (1.00)

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Heatup to NOT/NOP per 12002 C is in progress and the RO is about to begin a !

dilution- 1

- RCS boron is 2200 ppm

- MTC is slightly positive I

- SRNIs indicate 32 CPS

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Which of the following actions is correct? Stop the dilution if SR counts increase to 64 cps while heating up the RCS l Stop the RCS heatup and the dilution if SR counts increase to 64 cp ,

i Stop the RCS heatup if SR counts increase to 64 cps while diluting the RCS l Stop the RCS heatu Start the dilution and stop it if SR counts increase to 64 cp e f

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,,.. . 0VESTION: 045 (1.00) )

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Which of the following is the basis for allowing two hours to reduce the OPTR to within its limit with a tilt condition of greater than 1.02 ? Allows corrective action in the event of a xenon redistribution follow;ng power changes, b .' Allows for identification and repositioning of a dropped or misaligned ro '

1 Allows boron concentration changes to restore OPTR to less than 1.0 d Allows for identification and correction of a failed excore detecto .

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, .: QUESTION: 046 (1.00)

'9 ' The following plant conditions exist:

- A valid reactor trip signal has been receive Rod bottom lights are NOT lit and flux is NOT decreasin The main turbine is trippe The AFW pumps are runnin Emergency boration is in progres RCS pressure is 2350 psi Which of the following represents the action to be taken and the basis for that action to reduce RCS pressure per FR- " Response to Nuclear Power Generation" ? Depressurize the RCS with pressurizer sprays to allow enough borated water flow to ensure addition of negative reactivity to the cor Depressurize the RCS with pressurizer s3 rays to prevent a rapid over cooldown. pressurization transient and aegin a controlled Depressurize the RCS with pressurizer PORVs to allow enough borated water flow to ensure addition of negative reactivity to the cor Depressurize the RCS with pressurizer PORVs to prevent a rapid overpressurization transient and begin a controlled cooldow .

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QUESTION: 047 (1.00)

' 'The following plant conditions exists:

- A reactor trip and SI have occurred

- RCS Subcooling is 45 degrees F

- Containment 3ressure is 5.0 and decreasing

- Containment Rad level is 75.000 rad /hr Which of the following represents conditions that would allow termination of Safety Injection ? RCS pressure is 1710 psig and decreasing Pressurizer level is 40%

Narrow range steam generator levels are 34%

Total AFW flow is 500 gpm RCS pressure is 1710 psig and increasing Pressurizer level is 39%

Narrow range steam generator levels are 36%

Total AFW flow is 600 gpm RCS pressure is 1710 psig and increasing Pressurizer level is 41% .

Narrow range steam generator levels are 29%

Total AFW flow is 500 gpm RCS pressure is 1710 psig and increasing Pressurizer level is 9%

Narrow range steam generator levels are 11%

Total AFW flow is 600 gpm

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. g .. QUESTION: 048 (1.00)

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'Which of the following is the basis for stopping the RCPs upon entering FR-H.1, " Response to Loss of Secondary Heat Sink" ? Allows the operator time to establish a higher flow rate for high pressure SI thus increasing the RCS cooldown rat Allows for a more controlled cooldown via natural circulation when feedwater is establishe Allows the operator time to dearessurize the .utact steam generators in order to reduce ?.CS pressure and inject the accumulator Allows the operator to reduce heat addition to the RCS and extend the inventory in the steam generator ,

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. QUESTION: 049 (1.00)

When responding to a loss of all AC power in accordance with 19100-C. "ECA-0.0 Loss Of All AC Power". the intact Steam Generators are depressurized at the maximum rate (within the capability of the TDAFW). Which of the following is the reason for depressurizing at the maximum rate ? To minimize secondary inventory los To minimize RCS inventory los To minimize RCP seal damag To minimize reactor vessel upper head voidin l

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., QUESTION: 050 (1.00)

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A loss of all .AC power occurs with a resulting reactor trip and transition to procedure 19100-C "ECA-0.0. Loss Of All AC Power".

The following conditions exist:

- Narrow range level in all SGs --- 35% and decreasing I

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Containment pressure - O psig l

- RCS pressure is 2320 psig and decreasing

- RCS subcooling margin - 40 degrees F and decreasing )

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- PZR level - 19% and decreasing

- Turbine stop valves - closed

- PORV 455 - opened on high pressure

- PORV 456 - closed

- HV-8149A. Letdown 45 gpm isolation valve - open j

- HV-8149B. Letdown 75 gpm isolation valve - closed l

- HV-8149C. Letdown 75 gpm isolation valve - closed l You have verified a reactor trip and turbine trip. which of the I following should be the next action ? '

l Go to 19231-C. FRH-1 and increase level in at least one SG i to > 50%. l i Close Valve PORV 455 and continue in ECA- ! Manually initiate Safety Injection and go to 19000-C. E- Close HV-8149A and continue in ECA- .

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f l,.. QUESTION: 051 (1.00)

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Match the following with their definition in 10012-C. E0P and A0P Writers Guid ;

A Check B Ensure C Perform D Verify l'

1 To carry out an actio To make arrangements for a stated conditio Take necessary actions to guarantee conditions are as speci fie ,

4 To observe an expected characteristic or condition exist Typically the expectation comes from some previous automatic or operator action. The appropriate contingency, either stated or implied, is to establish the expected conditio To perform a comparison with a procedural requiremen Which of the following groups are correctly matched ? A-1 . B-3 . C-2 . D-4

' A-1 . B-4 . C-2 . D-3 A-5 . B-4 . C-1 . D-3

, A-5 . B-3 . C-1 . D-4

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.. QUESTION

052 (1.00) I

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Given the following:

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i - Committed Dose Equivalent (CDE)is 2525_mr- )

l - Deep Dose Equivalent (DDE) is 2335 mr

- Lens Dose Equivalent (LDE) is 744 mr

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- Committed Effective Dose Equivalent (CEDE) 405 mr l

- Total Organ Dose Equivalent (TODE) 4865 mr l - Shallow Dose Equivalent (SDE) 435 mr

- Maximum Extremity (ME) 6565 mr What is the Total Effective Dose Equivalent (TEDE)? mr ,

1 mr mr mr

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,., OUESTION: 053 (1.00) l

% 18 days following a design basis LOCA the following conditions I exist:

- Hydrogen concentration is 4.3%

- Containment pressure is 3.1 psig ,

- RCS temperature is 168 degrees F

- NSCW Basin temperatur" is 78 degrees F )

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Which of the following are correct actions in accordance with SOP-13130, !

" Post Accident H2 Control" ? ,

! Initiate the Containment Electric Hydrogen Recombiners and I do not initiate Post LOCA Hydrogen purge syste Initiate the Containment Electric Hydrogen Recombiners and initiate Post LOCA Hydrogen purge syste Do not initiate the Coritainment Electric Hydrogen Recombiners and do not initiate Post LOCA Hydrogen purge syste Do not initiate the Containment. Electric Hydrogen Recombiners and initiate Post LOCA Hydrogen purge syste .

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... QUESTION: 054 (1.00)

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Which of these accurately describe the color coding of the sound-powered phone system loops in various plant locations ?

a. Orange jacks used for startup and maintenance testing. Red jacks used for maintaining cold shutdown condition following control room evacuation. Grey jacks used for refueling, fuel handling building, diesel generator building, and containmen b. Orange jacks used for refueling. services control room, fuel handling building, and containment. Red jacks used for startu Grey jacks used for maintaining cold shutdown condition following control room evacuatio f I

c. Orange jacks used for control room, and containment. Red jacks used j for maintaining cold shutdown condition following control room

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evacuation. Grey Jacks used for startup and maintenance testin d. Orange jacks used for maintaining cold shutdowri condition following control room evacuation. Red jacks used for refueling, services control 1 room, fuel handling building, and containment. Grey Jacks used for !

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. QUESTION: 055 (1.00)

b' 'The following is a partial list of unit differences ?

(1) When in local. Diesel Generator Train 8 Fuel 011 Transfer pumps will run in auto ~onl (2) DG Day Tanks cannot be drained back to the Fuel Oil Storage Tan (3) RHR high-point vents (HV-10465 and HV-10466) have drain lines routed outside the cubicle (4) Aux Containment Cooler isolation valves are interlocked so that the outlet valves must be opened firs (5) High point vacuum breakers are located at the ESF chillers and the CCW heat exchanger (6) NSCW motor coolers do not have orifice (7) The setpoint of the Tavg / Auctioneered Tavg alarm is 4 degrees (8) ACCW pump Low pressure auto start @ 130 psi Which of the following are correct: Applies to unit 1 only - 2. 5. and 6 AND applies to unit 2 only 1. 3. and 7 Applies to unit 1 only - 1. 3. and 6 AND

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applies to unit 2 only - 2. 4. and 8 Applies to unit 1 only - 2. 6. and 7 AND ,

applies to unit 2 only - 3. 5. and 8

- Applies to unit 1 only - 1. 4. and 7 AND applies to unit 2 only - 3. 5. and 6 I

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.- . . ,, . OUESTION: 056 (1.00)

/ ' 'Which of the following actions will occur automatically if a high level radiation alarm is actuated on the associated monitor ? Fuel Handling Building Effluent (ARE-2532A or 8: ARE 2533A or B) - Isolates gas discharges if a gas release is in progres Turbine Building Drain (RE-0848) - Re-align TB drains to dirty drain tan Mainsteam Line (RE-13120) - Aligns SJAE discharge from direct i

discharge to environment to discharge through HEPA filte Containment Low Range (RE-0002 or RE-0003) - Isolates SG blowdow . ;

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.,, QUESTION: 057 (1.00)

'Following a loss of a string of feed water heaters, the secondary plant undergoes a transient. During trouble shooting. Unit 1 J trips from 87 % powe All systems perform normall The generator does not trip for 30 seconds. What is one of the reasons for this delay ?

] Ensure RCPs seals flow can be isolated in 30 second RCPs won't overspeed on SGT Maintains RCS flow during transients where Heat Flux Hot Channel J Factor is a concer Prevents main turbine overspee ;

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! .. QUESTION: 058 (1.00)

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Condenser vacuum is DECREASING: operators decreasing

. turbine load in attempt to maintain vacuum

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Condenser vacuum - 25.8 in of Hg

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Main turbine load - 43%

Which of the following would be required or occur FIRST if condenser vacuum continued to decrease ? Auto main turbine trip on low vacuu Loss of steam dump capability, Manual reactor trip, Auto MFPT trip on low vacuu l

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QUESTION: 059 (1.00)  !

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l Which of the following explains the reason for the caution in

! 19251-C, "FR-Z.1 Response to High Containment Pressure". giving 19111-C "ECA-1.1 Loss of Emergency Coolant Recirculation" priority over FR-Z.1 for directing containment : pray operation ? The caution is a reminder that the rules of usage gives all ECA'S priarity over FR's.

L ECA-1.1 is trying to conserve RWST inventory to be utilized for core coolin FR-Z. 1 could cause the spray pumps to run without adequate NPSH if ECA-1.1 was in effect before transitioning to the F To prevent the spray pumps from being removed from service before adequate mixing of TSP has occurre .

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. QUESTION: 060 (1.00) )

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'Which of the following positions do NOT have the authority to suspend refueling operations if, in their judgement, any conditions exist which threaten personnel safety or safe handling of fuel ? Fuel Handling Coordinator Reactor Engineer Health Physics Technician Chemistry Technician l

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. QUESTION: 061 (1.00)

Which of the following is dependent on Tavg to determine it's value ? overtemperature Delta T Trip Comparator overtemperature Delta T Trip Setpoint Calculator

. auctioneering Unit Delta T Deviation Alarm f overpower Delta T Trip Comparator

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. QUESTION: 062 (1.00) i f'. ~ i One DRPI is inoperable in each of two different group You are l recaired to verify the position of the rods with inoperable position

incicators by using movable incore detectors. TS 3.1.7 Action Statement A.1 states that the required completion time is once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. What is the TS basis for this requirement ? The probability of simultaneously having a rod out of position and an event sensitive to that position is small, Power Peaking limits cannot be violated if only one DRPI per group is inoperable for one or more group Ejected rod worth limits cannot be violated if only one DRPI per group is inoperable for one or more group SDM limits cannot be violated if only one DRPI per group is inoperable for one or more group .

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,.. QUESTION: 063 (1.00)

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Which of the following i.s the correct sequence that describes a safeguards actuation signal's progression through SSPS ? Logic card, Driver Card, Master Relay, Slave Relay, Input

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Relay Input Relay, Master Relay Slave Relay, Logic card, Driver Card Input Relay, Logic card, Master Relay, Slave Relay, Driver Card Input Relay, Logic card, Driver Card, Master Relay, Slave Relay

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. QUESTION: 064 (1.00)

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Concerning the basis for. restrictions on DNBR in the core: i

.The average enthalpy in the hot leg should be to the )

enthalpy of saturated liquid. This ensures that the Delta T measured )

by incore instrumentation is to core powe I greater than or equal : proportional I less than or equal : inversely proportional I less than or equal : proportional I greater than or equal : inversely proportional l

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.. QUESTION: 065 (1.00)

Which of the following conditions can cause "AMSAC Trouble" at 45% power ? C-20 timer runs causing AMSAC to be inhibite ; First Stage Impulse Pressure channel 505 has failed downscal It has been 260 seconds since impulse pressure exceeded the arming setpoin The operator failed to press the "AMSAC OPERABLE" push button after reaching 40 percent power, i

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l . QUESTION: 066 (1.00)

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D Following the initiation of the Loss of Offsite Power Sequencer, i which of the following are true for the containment cooling fans ?

l l After 30.5 seconds fans 3. 4, 5 and 6 are running in high i

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speed AND After 30.5 seconds fans 1, 2, 7 and 8 are running in high speed.

l l- After 30.5 seconds fans 3, 4. 5 and 6 are running in slow speed AND After 30.5 seconds fans 1, 2. 7 and 8 are running in slow spee After 50.5 seconds fans 3, 4, 5 and 6 are running in high .

speed AND After 50.5 ceconds fans 1. 2, 7 and 8 are running !

in high spee i

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' After 50.5 seconds fans 3. 4, 5 and 6 are running in slow speed AND After 50.5 seconds fans 1, 2, 7 and 8 are running in slow speed.

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. QUESTION: 067 (1.00)

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A HIGH alarm received by which of the following radiation detectors will result in an Automatic Action ? Steam Line Rad Monitor RE-1312 Steam Generator Blowdown Rad Monitor RE-01 Primary-to-Secondary Leakage Rad Monitor RE-072 Condenser Air Ejector and Steam Packing Exhauster~ Rad Monitor RE-12839 .,

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... QUESTION: 068 (1.00)

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During a surveillance, RE-0020A NSCW Effluent Line is determined to have a non-conservative alarm setpoint. Which of the following is one of the immediate actions listed in the ODCM ? Verify automatic isolation has occurred Bypass the affected channel Change the setpoint to a conservative value Notify Chemistry to immediately sample the effluent l

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. QUESTION: 069 _(1.00) l Procedure 13427-1 contains the following instructions for Diesel '

Generator 1A normal start. These instructions state (for paralleling the DG to it's vital bus with the synchroscope rotating slowly in the clockwise direction):

4.2.1.13 When the Sync Scope needle reaches the 11 o' clock '

position. DEPRESS and HOLD the Diesel Generator 1A AUTO SYNC PERMISSIVE PUSHBUTTON PB-DG1 .2.1.14 VERIFY that the DG1A OUTPUT BRKR 1AA02-19 closes when the Sync Scope reaches the 12 o' clock position and RELEASE the Auto Sync Permissive Pushbutto What is the specific purpose of these steps ? To minimize frequency and phase difference To minimize frequency and current difference To minimize current and voltage difference To minimize voltage and phase d.ifference l l

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.s GUEST 10N: 070 (1.00)

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Given the following conditions:

- reactor power 100%

- containment ventilation isolation actuation

- fuel handling post accident ventilation actuation

- control room emergency filtration actuation Given there was a failure of a single 120 VAC vital instrument panel, which of the following has failed ? LAY 2A BY1B CY1A DY1B l

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QUESTION: 071 (1.00)

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'The source of power to the solid state protection system CHANNEL I

, is and CHANNEL IV is . LAY 1A. 10Y1B BY1B ICY 1B l ICY 1B. LAY 1B Y1B. 1BY1B

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. QUESTION: 072 (1.00)

'The unit 2 main generator has just been synchronized to the grid and power has been raised to 19% power. The BOP was preparing to 4 swap feedwater flow from the Bypass Fe?d Regulation Valves to the Main Feed Regulation Valves when condenser vacuum decreased to 21.5 inches of Hg, generating a turbine tri )

Which of the following are the correct actions the crew should take in response to the turbine trip ? Enter 18011-C. Turbine Trip below P-9. and reduce reactor power below 5% and control Tave using steam dump Trip the reactor and go to 19000-C. Reactor Trip or Safety Injectio ,

l Enter'18016-C, Condensate and Feedwater Malfunctions, start all available AFW pumps, and reduce reactor power to 10%. Enter 18011-C, Turbine Trip below P-9. reduce reactor power below 5%. and control Tave using atmospheric relief valve .

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QUESTION: 073 (1.00)

'At 10:58 you received the.following annunciators:

- ACCW RCP 1 L0 CLP. LOW FLOW

- ACCW RCP 1 CLR OUTLET HI TEMP At 11:01 you determine the RCP 1 ACCW inlet line is leaking. You !

are informed that the repairs will take about 15 minutes. You  !

enter both 18022-C " Loss of Auxiliary Cooling Water" and 13003-1

" Reactor Coolant Pump operations". Using the plant computer you determine the following RCP 1 criteria exist:

- Motor Bearing Temperature is 190 degrees increasing at 1 degree per minute

- Stator winding temperature is 305 degrees and steady

- RCP Pump lower seal bearing is 225 and decreasing at .5 degrees per' minute

- Seal water outlet temperat. ire is 255 and increasing at 1 degree per minute Assuming that the ACCW is not returned to service, what is the earliest time you are REQUIRED to trip RCP 1 ? :01 :06 :08 :11

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QUESTION: 074 (1.00)

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'An operator at the Diesel Generator 1A local panel has verified that Diesel Generator 1A has lost DC control power. If the diesel I is currently not running, which of the following describes how l this loss of DC power would affect Diesel Generator operation ? ) The diesel would start in response to an automatic or manual start signal, The diesel cannot be started by automatic signals or manual actio The diesel can only be started in manual at the local panel in response to an operator initiated local star The diesel can only be started manually from the control room and the local control pane ,

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QUESTION: 075 (1.00)

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iliththereactorat90percentpower,howwilltheplantrespondtoaloss of two feedwater heater strings once steady state conditions are reached ?

(Assume no operator actions and rods are in automatic) T-ref Decreases, GENERATOR OUTPUT (MW) Increases, J RCS Tavg decreases, RCS T-cold Decreases. AND RCS T-hot Increase T-ref Increases, GENERATOR OUTPUT (MW) constant, RCS Tavg is Constant, RCS T-cold Decreases, AND RCS T-hot Increase T-ref Decreases. GENERATOR OUTPUT (MW)is Constant, RCS Tavg Decreases RCS T-cold Decreases AND RCS T-hot Decrease T-ref is Constant, GENERATOR OUTPUT (MW) Increase RCS Tavg is Constant, RCS T-cold Decreases, AND RCS T-hot Increase !

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,. QUESTION: 076 (1.00)

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Given the following:

- Plant in Mode 6 with vessel head installed

- Midloop operations in progress

- S/G hot and cold leg manways removed

- S/Gnczzledamsinstalledonhotlegs .

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- S/G nozzle dams NOT installed on cold legs

- No other vents are open in the RCS

- Loss of RHR cooling occurs Which of the following will occur as a long term result of this event :

i Steam formation in upper head will increase pressure and cause the PZP, to refill rapidl ) Steam formation in u)per head will increase pressure and !

displace water out tie S/G cold leg manway Steam formation in u)per head will increase pressure ]

enough to blow out t1e hot leg nozzle dam Steam formation in upper head and resultant steam bubble expansion will displace water out the hot leg manway .-

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.., QUESTION: 077 (1.00)

b 'Which of the following 1.s true concerning normal power supplied to a 125 VDC ESF Bus ? The bus is normally supplied from 2 battery chargers. A battery bank will supply power to the bus in the event both of the battery chargers fail. The batteries are kept charged by " floating" on the bus, The bus is normally supplied from 2 battery chargers. Vital A/C aower will supply )ower to the bus through inverters in tie event both of t1e battery chargers fail. The bus is kept charged by " floating" on the batterie The bus is normally supplied from 2 battery banks. The bus ,

will be supplied directly from the chargers if both battery ,

banks fail. The bus is kept charged by " floating" on the l batterie I l The bus is normally sup Vital l A/C power will supply )ower plied to from the2bus battery bank through inverters in the event both of the )attery banks fail. The batteries are kept charged by " floating" on the bu :

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. QUESTION: 078 (1.00)

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Which of the following is NOT accomplished by 19132 "ECA SGTR with Loss of Reactor Coolant: Saturated Recovery Desired" ? { Depressurize RCS to minimum RCS pressur l' Depressurize RCS to establish pressurizer leve { l Cooldown T-cold at maximum rate using S/G PORV I Reduce ECCS flow while maintaining pressurizer level and decrease reactor coolant core exit temperatur )

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,, QUESTION: 079 (1.00)

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Procedure 19241."FR-P.l. " Response'to imminent Pressurized Thergel Shock Condition", r,tep 5.0. directs you to attempt to restart a RCP if the SI termination criteria cannat be satisfied. What is the basis

?or this step ? Restorcs PZP spray to allow RCS depressurization in subsequent steps with ECCS stili in service, Equalizes SG pressures-to allow simultaneous cooldown of all four loops in subsequent steps-. Mires ECCS injection water and RCS water to raise the. fluid temoerature entering the vessel downcomer Transfers Core Cooling to forced flow allowing the operators to terminate ECCS when the criteria are NOT satisfie ,

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,,v QUESTION:. 080 (1.00) I

'While recovering from a total loss of all AC poaer in 19100-C, '

step 30 direct,s you to 19101-C "ECA 0.1 Loss of All AC Power Recovery Without SI Required". The first step directs the operator to close the ACCW Supply Header ORC Isolation. Valve HV-197 What is-tne reason for this actio'1 ?

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a'. To prevent exceeding the cooling capacity of the RCP thermal barrier heat exchange To reduce .ACCW heat loads prior to restarting a ACCW pum l To prevent thermal shock of the RCP seals when starting a ACCW pump, To reduce the possibility of steam formation and circulation within the ACCW syste .

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.,. QUESTION: 081.-(1.00)

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Unit 1 is at 75% power with rods in Automatic conducting a power ascent in accordance with VEGP 12004-C, Power Operation. When the following indications are observed:

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Tavg - 570 degrees F and slowly loweri Pressurizer Level - 42 percent and slowly lowering j Pressurizer Pressure - 2100 ps and slowly lowering )

Reactor Power - at approximate 75% and slowly lowering t Rods - stepping in slowly l

Which of the IMMEDIATE ACTION (S) must be performed ?

I Manually trip the reacto I 1 Place the bank selector switch in MANUA Decrease turbine load to restore Tavg to Tre ., Switch Tavg input to alternate channe .

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QUESTION: 082 (1.00)

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'Given the following conditions:

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reactor power 95% 1

- 1 control rod falls into core as indicated by ORPI I

- no automatic reactor trip occurs You enter AOP 18003-C " Rod Control Malfunction". Which of the following is a required action ? j Reduce reactor power to less than 70% for rod retrieva l Emergency borate 115 pp Verify adequate shutdown margi Manually trip the reactor, i

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. QUESTION: 083 (1.00)

'During a controlled shutdown from 100% rated thermal power the -

operatcrs note a rod that is stuck more than 12 steps from the -

group step counte" demand and are unable to verify that it can

.be tri] ped. Whicn of the following identifies the MAXIMUM time availa)le before any Technical Specification actions are required to be taken ? Minutes )

1 Minutes Minutes hours

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.. QUESTION: 084 (1.00)

Given the following conditions:

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Power is 100% RTP

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'RCP SEAL OUTLET TEMP HI alarm actuated, then cleared

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RCP SEAL LEAK 0FF FLOW HI alarm actuated, then cleared

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No.1 RCP seal outlet temperature is 222 deg F and increasing

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No.1 RCP seal leak-off flow is 4.9 gpm and stable

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No. 2 RCP seal outiet temperature is 218 deg F and stable

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No. 2 RCP seal-leak-off flow is 1.8 gpm and stable

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Seal injection flow is 8.5 gpm and stable

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Seal injection temperature is 134 degrees and stable

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Stand 3ipe fill frequency is normal

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RCP viaration is 6 mils

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RCDT Level is stable and controlled

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No.1 Seal Differential pressure is 2000 psig You enter procedure 13003-1. " Reactor Coolant Pump Operation." Step 4.2.3 directs you to Figure 1. "RCP Seal Abnormalities Decision Tree." Based on the conditions noted, which of the following actions is required ? (Figures are attached)

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,- GO TO 4.2.1.4 TO STOP RCP IF REQUIRE 0 STOP RCP BY INITIATING 4.2. REPAIR AT NEXT OUTAGE FAILURE OF NO.3 OUTER DAM

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.., OUESTION: 085 (1.00)

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While operating at 100% power you receive a simultaneous loss of Source rangechannel N-31. Intermediate range channel N-35 and .

Power range channel N-41. Which of the following has happened ?

l l Vital Instrument Panel 18Y2B deenergize Vital Instrument Panel 1AY2A deenergized, Vital Instrument Panel 1CY1A deenergize Vital Instrument Panel 1AY1A deenergize I

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..- QUESTION: 086 (1.00)

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Which of the following conditions does NOT support or indicate

.that natural circulation flow is occurring ?

. SG pressure increasing core exit thermocouples - stable i RCS subcooling monitor indication - greater than 38 deg F

" RCS cold leg temperature - at saturation temperature for SG ,

pressure ,

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s QUESTION: 087 (1.00)

There is a fire in the main control room. You are the Unit Shift Superviso You direct entry into 18038-1. " Operation From Remote Shutdown Panels." You determine that is safe to perform all the actions that can be performed in the main control room (steps 1 and 2). The reactor is trippe Which of the following actions will NOT be completed prior to leaving the main control room ?

" Trip 1 & 4 RCP's i Shut Pressurizer PORV Block Valves Shut MSIVs and Bypass Valves Trip all Running ACCW pumps

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.. QUESTION: 088 (1.00)

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Chemistry is performing Technical Specification SR 3.4.16. " Reactor Coolant System Activity." Gross specific activity of the reactor-coolant is reported as 100 picocuries/ liter. Based on your ,

current power level you have 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to be within the acceptable i operating' band of Figure 3.4.16-1 or Tavg must be below 500 '

The BASIS for the Technical Specification action of reducing Tavg

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to less than 500 degrees F (assuming a SGTR exists) is to: ? j

! prevent exceedina the release of 99% of assumed iodine gap j activit j minimize the iodine spikin I stay within the abe.,orption limits of the sodium hydroxide in containment sr. ra prevent openi.g the main steam safety valve i l

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. QUESliON: 089 (1.00)

l' 'Which of the following is the primary method of determining RHR leakage into the CCW system ? NSCW differential flos alarms Leak status light on OPCP Auxiliary Building leak detection system Radiation monitor

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QUESTION: 090 (1.00)

'A HIGH alarm on which of the following radiation monitors will result in an automatic actuation of ESFAS equipment ? Outside Area monitor. RE-0069 Control Room Area monitor, RE-0001 CVCS Letdown Process monitor, RE-48000 Control Room Air Intake monitor, RE-12116 i

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. .0VESTION: 091 (1.00) '

Which of the -following parameters discriminates between a steamline leak inside containment and a small break LOCA ? Containment radiation level, Pressurizer leve Steam generator level, Containment pressur .

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~ Which of the following sets of signals BOTH actuate Containment

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Spray ? of 2 Containment Spray handswitches taken to ACTUATE or 2 of 4 Containment Pressure Channels > 21.5 #. of 2 Containment Spray handswitches taken to ACTUATE or 2 of 4 Containment Pressure Channels > 21.5 #. of 2 Containment Spray handswitches taken to ACTUATE or 2 of 3 Containment Pressure Channels > 14.5 #. of 2 Containment Spray handswitches taken to ACTUATE or 2 of 3 Containment Pressure Channels > 14.5 #. .

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QUESTION: 093 (1.09)

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Pressure control is selected to the 455/456 position with all pressure control devices in the auto aosition when PT-456 fails high. Assuming No Operator Actions, which of tie following describes the plant response ? Both spray valves open, one PORV opens, all heaters deenergize. Pressure j decreases until a reactor trip and safety injection occur. Pressure will stabilize when the pressurizer goes solid at about 2300 psi Both spray valves shut, one PORV opens, all heaters energize. Pressure decreases until the heaters can provide sufficient heat to overcome energy lost out of the PORV. Pressure will stabilize at about 2210 psi )

! Both spray valves shut, one PORV opens, all heater.s energize. Pressure !

decreases until a reactor trip and safety injection occur. Pressure I will stabilize when the pressurizer goes solid at about 2300 psig, Both spray valves shut, one PORV opens, all heaters energize. Pressure decreases until the open PORV receives a shut signal. Pressure will fluctuate around 2185 psig with no reactor trip or safety injection occurrin .

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. QUESTION: 094 (1.00) -

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Unit 1 is on RHR cooling when there is a sudden loss of instrument ai Assuming no operator actions, what are the initial effects on the RCS and RHR systems ? RCS Pressure decreases quickly. The RHR Heat Exchanger outlet valve opens and the RHR Heat exchanger bypass' valve close RCS Pressure decreases quickly. The RHR Heat Exchanger outlet valve closes and the RHR Heat exchanger bypass valve open RCS Pressure increases quickly. The RHR Heat Exchanger outlet valve closes and the RHR Heat exchanger bypass valve open RCS Pressure increases quickly. The RHR Heat Exchanger outlet valve opens and the RHR Heat exchanger bypass valve close .

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., OUESTION: 095 (1.00)

With the Unit initially operating at 100% power with all control systems i in automatic. which of the following is NOT an indication of a 50 gpm !

Steam Generator Tube Leak ? VCT level decrease i i' Condenser vacuum exhaust radiation monitor increas Steam generator level increase Increase in CVCS charging. flow

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QUESTION: 096 (1.00)

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'Which of the following will cause an isolation of the containment ventilation system ? Loss of NSCW while at 8% powe Pressurizer low pressure (2/4 channels below 1870 psig) Instrument Air header low pressure (75 psig) Containment Atmosphere Radiation Monitor RE-2562 in HIGH alar .

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.. QUESTION: 097 (1.00)

1 You Receive the PRZR HI PRESSURE Annunciato No other alarms are present. What are the Immediate Operator Actions required for a pressurizer pressure instrument failure ? Verify RCS pressure stable: if lowering, close spray valves, close affected PRZR PORV, and operate PRZR heaters as necessary Verify RCS pressure stable: if increasing, open spray valves, close affected PRZR PORV. and operate PRZR spray as necessary Verify RCS pressure stable: if lowering, open spray valves, close affected PRZR PORV, and operate PRZR spray as necessary l Verify RCS presst.re stable: if increasing, close spray valves, close {

affected PRZR POPV, and operate PRZR heaters as necessary l

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. QUESTION: 098 (1 00)

k" 'Given the following conditions: ,

- nofmal 100% power plant lineup

- decreasing pressurizer level

- increasing VCT level

- "RCP SEAL WATER INJECTION LO FLOW" annunciator

- " REGEN HX LTDN HI TEMP" annunciator l

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given conditions ? pressurizar PORV open small break LOCA letdown isolation loss of charging

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..., QUESTION: 099 (1.00)

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' Step 2 of E-0 " Reactor Trip or Safety Injection", requires plant operators to " Verify Turbine Trip."

Which of the following is the BASIS for performing this action? Prevent return to criticality in the event of a steam line brea Prevent loss of shutdown margin due to overcoolin Prevent unnecessary safety injection due to low RCS pressur Prevent unnecessary depletion of S/G heat removal capacit _

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.. OUESTION: 100 (1.00) (

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'Which of the following is one of the BASIS for Technical S)ecification l 3.4.9.b which requires at least two groups of pressurizer 1 eaters i each having a capacity of greater than or equal to 150 KW ?

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, To prevent loss of single phase flow in natural circulation l

condition To maintain sufficient Reactor Coolant System pressure to

establish letdown and CVCS degassificatio To sustain a steam bubble in the pressurizer during an insurge following a steam line brea To maintain sufficient Reactor Coolant System pressure to i prevent reactor coolant pump cavitation during normal operations.

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ANSWER KEY

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MULTIPLE CHOICE 023 c 001 d 024 d

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002 b 025 a 003 c 026 b

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004 c 005 b 028 a 006 a 029 a

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007 b 030 d 008 a 031 c 009 b 032 b 010 d 033 _

011 b 034 a

012 a 035 c 013 c 036 d 014 b 037 c 015 c 038 a

. 016 b 039 b 040 d 041 d-019 c 042 b 020 b 043 a 021 a -

044 d 022 c 045 b.

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ANSWER KEY l

046 c 069 a 047 b' 070 a 048- d 071 a

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049 b 072 d 050 d 073 a 051 d 074 b 052 a 075 d, 053 a 076 b-054 c 077 .c 078 g 056 b 079 c 057 d 080 d'

058 b 081 b 059 b 082 c 060 d 083 c 061 b 084 b 062 a 085 d 063 d 086 a 064 c 087 d 065 b 088 d-066 c 089 d 067 d 090 d 068 c 091 a L . .

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,, SENIOR REACTOR OPERATOR Page 3 ANSWER KEY i

092 b 093 d l

094 d 095 c 096 b 097 a 098 d 099 b .

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