IR 05000254/1996302
| ML20129K449 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 11/07/1996 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20129K448 | List: |
| References | |
| 50-254-96-302OL, 50-265-96-302OL, NUDOCS 9611250216 | |
| Download: ML20129K449 (161) | |
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p U. S. NUCLEAR REGULATORY COMMISSION I REGION lli 1 !
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Reports No: 50-254/96302(OL); 50-265/96302(OL)
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Licensee: Commonwealth Edison Company (Comed)
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Facility: Quad Cities Nuclear Power Station, Units 1 & 2
! Location: 22710 206th Avenue North Cordova, IL 61242 ,
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Dates: October 7 - October 11,1996 i
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i Examiners: D. McNeil, Chief Examiner
, R. Doornbos, Rlli Examiner :
1 T. Bettendorf, Pacific Northwest Laboratories i
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Approved by: Melvyn N. Leach, Chief, Operator Licensing Branch Division of Reactor Safety
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9611250216 961107 PDR ADOCK 05000254 V PDR
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4 EXECUTIVE SUMMARY Ouad Cities Nuclear Power Station, Units 1 and 2 Examination Reports No. 50-254/96302; 50-265/96302 An NRC initial license examination was administered to eight indi/iduals; three who had applied for Reactor Operator licenses and five who had applied for Senior Reactor Operator license Results: One Reactor Operator applicant failed the written portion of his examination and -
was denied an operating license. All other applicants passed all portions of their examinations and were issued Reactor Operator or Senior Reactor Operator license Summarv: The examination material was prepared by the Quad Cities Nuuoar Power Station training staff in accordance with the NRC operator licensing pilot program guidance, Operator Licensing Examiner Standards (NUREG 1021), and 10 CFR 55. The NRC chief examiner determined that some examination material did not meet all reqWements of the examiner's standards and 10 CFR 55. Examination material not meeting minimum requirements was modified in accordance with the NRC chief examiner's request. (05.1 )
One individual signed the examination security agreement and subsequently participated in a training session with license applicants in the simulator. The Quad Cities Station Training Department Training Instruction No.103, Examination Security, specifically prohibited participation in instruction involving the applicants after signing the security agreement. (05.2)
The examination was conducted in accordance with NRC requirements. Facility personnel were responsive to the needs of the NRC examiners and their applicants. The examination was well controlled by facility personnel with only minor errors. Two generic weaknesses were noted during the administration of the in plant job performance measures. (05.3)
The OCNPS Operations Training Supervisor stated there would be no post examination comments from the facility. During the post examination review conducted by Rlll NRC examiners, one answer key error was corrected and two questions were discovered to be unacceptable for use on an NRC examination. The two questions were subsequently deleted. (05.4)
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The simulated plant process computer failed during the examination week. This required the applicants to operate the simulator during the dynamic simulator scenarios without the benefits of many of the video monitor displays the candidates were trained to use. The impact on applicant performance was minimal. The process computer was made available later in the week during administration of the job performance measures. (05.5)
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- Reoort Details Summarv of Examination l An NRC initiallicense 6xamination for eight applicants was conducted at the Quad Cities
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Nuclear Power Station (OCNPS) during the week of October 7,1996. OCNPS training i department management agreed to participate in a program in which the licensee developed the examination material. NRC examiners reviewed and approved the OCNPS i developed examination. NRC examiners administered the operating examination; QCNPS
-facility instructors administered the written examination. Examination preparation and
! administration was prescribed by NUREG-1021, " Operator Licensing Examiner Standards,*
3 Revision 7, and superseded in part by Interim Pilot Examination Guidance approved by
[ Nuclear Reactor Regulation, Headquarters Operator Licensing Branc i l
1. Ooerations l l 05 - Operator Training and Qualification
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05,1 Examination Develooment j
! yVritten Examination The submitted written examination met the minimum standards of NUREG-1021, l Operator Licensing Examiner Standards. The chief examiner requested several replacement questions for the reactor operator (RO) examination to bring the examination into conformence with 10 CFR 55.41, Written examination: operator Those questions required knowledge from reactor operators beyond that which is required for an RO by 10 CFR 55.41. The chief examiner also requested that several other questions be modified. The difficulty level of those questions was not correct for ROs. Some questions were simple memory questions that all employees at OCNPS should be able to answer by attending General Employee Traisiing. Other ;
questions required Senior Reactor Operator (SRO) knowledge to correctly answer !
the question. Those questions were deleted and replaced with questions applicable ;
to the RO job classification. Some questions were acceptable as submitted, but were modified to bring the examination difficulty levelinto conformance with examinations recently administered by NRC examiners at other facilitie Administrative Job Performance Measures One of the submitted administrative job performance measures (JPMs) for the ROs was not accepted. The JPM required the ROs to perform at the Senior Reactor Operator (SRO) level. The JPM was replaced by another JPM that provided an I evaluation at the correct licensing leve Ooeratina Job Performance Measures The operating JPMs submitted by the facility instructors were acceptable, with only minor modifications being necessary to make them ready for the examination. The
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operating JPP,1 fuuow up quevions did not meet the guidelines established ca l
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NUREG 1021, form ES-602-1. Mcat of the submitted questions were easy memory questions or could be quickly found ! direct look-up) in an allowed reference.
' Dynamic Simulator Scenarios Of the five dynamic simulator scenarios submitted, none met all the guidance of NUREG-1021, Form ES-301-5. This form outlines the evolutio .s (reactivity, instrument failure, component failure, etc) and the number of evolutions each applicant is expected to perform during the dynamic simulator sessions. OCNPS instructors assigned to construct the scenarios mis-interpreted NUREG 1021 and the interim guidance instructions and believed that a normal evolution and a reactivity evolution were the same thing. This led to all scenarios needing a normal evolution such as a surveillance or significant, controlled evolution. Two of the l submitted scenarios lacked sufficient instrument failures. The scenarios were l modified to meet NRC requirement Conclusion The facility instructors constructed an examination that was of good quality, but required many minor changes. The majority of the changes made to the operating examination were caused by a lack of understanding the requirements of NUREG 1021 and 10 CFR 55. After the requested modificatior.s were made, the NRC chief examiner concluded the examination material was adequate to administer the examinatio '
06.2 Examination Security Insoection Scoce OCNPS license applicants in the third of three simulator crews being trained on September 30,1996, discovered a simulator hardware technician that had signed a security agreement participating in their training scenario and alerted OCNPS training department instructors and management. The NRC Chief Examiner reviewed the circumstances surrounding the event and the investigation conducted by facility personhs!. Observations and Findinas During development of the examination by facility and contractor personnel, a security agreement was initiated to keep track of personnel who had been exposed to examination specific material. OCNPS Training Department Training Instruction (OTl) No.103, Section Ill.D.b, states, "At the point when any person gains specific knowledge of the content of an exam, that person shall read and sign the security agreement (NUREG-1021, Form ES-201-2/601-1, or equivalent) for that exam."
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J Contrary to this statement, simulator support personnel were directed to sign the security agreement by training department personnel without being exposed to examination specific materia OTI No.103, flection Ill.D.c, states, "Once a person has signed the security agreement, he/she shall no longer be involved in the training of individuals who will be taking the exam." Contrary to this requirement, a simulator support technician participated in a simulator training session for license applicants as the simulator operator subsequent to signing the security agreemen Immediately upon being notified of the participation by the technician in the training session, QCNPS training department management initiated an investigation to determine if the examination had been compromised. Investigators interviewed approximately 40 station personnel and were able to determine that the technician:
(1) failed to read and comprehend the security agreement prior to signing the agreement, (2) participated in the instruction of 2 simulator crews as the simulator operator, (3) had not been exposed to examination specific material wt;sn he signed the security agreement, (4) had not been exposed to examination maidrial between the time he had signed the security agreement and the time he was discovered in the simulator, (5) had not participated on the simulator floor as an instructor, (6)
was not allowed to select or direct the simulatn. scenario in any way, and therefore, (7) did not compromise the examinatio The investigators found that '.ne simulator instructor and the applicants on the first two crews that attended the training should have identified the technician as someone that had signed the security agreement. They also found that no one had informed the simulator scheduler that the technician had signed the security ,
agreement, thus allowing the scheduler to place the technician's name on the schedule to participate in a training scenario with the applicant l The Chief Examiner reviewed the results of the investigation, discussed the event I with several personnel at the facility and concluded that the investigation was thorough, rapid, and correct in its assessment. Based on that conclusion, the Chief {
Examiner determined that the examination had not been compromised and allowed the examination to be given as schedule c. Conclusion I
Multiple opportunities to prevent this occurrence were missed by facility personne !
Examination security was understood by first line supervision in the training department, but supervisor expectations of examination security were not fully understood nor implemented by some facility employees. Greater emphasis needs J to be placed on the mechanics of how to implement the examination security I agreement by first line supervisors to prevent reoccurrence of examination security i problem ~
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05.3 Examination Administration Observations l l
The examination was conducted in accordance with NRC requirements. The j examination was well controlled except for the operating JPMs. Some operating l JPMs were long and required a simulator setup unique to that operating JPM. This I caused multiple resets of the simulator resulting in significant time delays with applicants waiting for their turn to perform certain operating JPMs. The NRC Chief .
Examiner discussed methods of controlling JPMs to expedite the examination I process with the Operations Training Supervisor. The written examination, administrative JPMs and dynamic simulator were well controlled by facility i personnel with only minor errors in examination contro Job Performance Measures Two generic weaknesses were noted during the administration of the in plant JPMs. (1) During the performance of a JPM to shift steam jet air ejectors, some i candidates became confused while using the procedure. One candidate was unable to cr'mplete the procedure correctly and two candidates improperly isolated the in-service steam jet air ejector before placing the standby steam jet air ejector in service. (2) There are personnel monitoring portals located outside the common emergency diesel generator room. Some candidates used the portal monitor and some indicated they did not need to use the monitor. This indicates a lack of training or ineffective training concerning the use of portal monitors other than the portal monitors used for exiting the radiological controlled area of the plan Dynamic Simulator Scenarios Candidates appeared well prepared for the dynamic simulator scenarios. OCNPS training personnel stationed at the n.7tructor's console provided timely, accurate feedback to applicants during the scenarios. The feed back was provided in such a way that no prompting of applicants occurre Conclusion OCNPS instructors need to pay more attention to detail to complete an examination l
without errors. The candidates were well prepared for the operating examinatio .4 Post Examination Activities !
Written Examination i i
QCNPS training department personnel were invited to submit post examination comments for the written examination. The Operations Training Supervisor contacted the NRC Chief Examiner on October 18,1996, and indicated there would be no post examination comments from the facilit j l
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. During the post examination review conducted by Rill NRC examiners, one answer i key error was corrected and two questions were discovered to be unacceptable for use on an NRC examination. The two questions were subsequently deleted.
f RO question #49 (SRO question # _13) was deleted after it was determined there =
was no correct answer. One of the initial conditions of the question states there is 1 10,000 gallons of water in the condensate storage tank (CST). The question then i asks what should occur based on the given plant conditions. The expected answer
- was that the suction for the High Pressure Core Injection (HPCI) system should' shift
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from the CST to the torus.- The actual setpoint for the shift is less than 10,000 gallons in the CST Since the setpoint had not been exceeded in the initial
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conditions, the correct answer would have been: no change is expected. Since that
- answer was not provided, there is no correct answer to the question. RO question j #49 (SRO question #13) was deleted from the written examination.
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RO question #78 (SRO question #42) was deleted aftar it was discovered the
- question had no correct answer. The question asked for the immediate actions an operators would perform upon discovering a blown fuse. The expected answer was
- found in the supplementary actions of the procedure. Since no immediate actions
were provided to answer the question, there was no correct answer for the
{ question. RO question #78 (SRO question #42) was deleted from the written i examination.
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l It was determined the answer key for RO question #94 (SRO question #58) was in
. error. The answer key indicated distractor a. was the correct answer. The answer -
l key has been amended to accept only c. as the correct answer.
i Several questions were found to be missed by more than 50% of the applicants. A i- list of these questions with a brief description of the training item missed is provided in Enclosure 3. The list should be reviewed by the OCNPS training
- department for feedback into the systematic approach to training (SAT) based j program used by OCNPS in the initiallicense training program.
i 05.5 Simulator Observations
) Plant Process Comouter i
i During the week of the examination the simulated the plant process computer i failed. This required the candidates to operate the simulator without the benefits of
. many of the video monitor displays the candidates were trained to use. Candidates j were well trained and able to control plant processes without the benefit of the
- simulated plant process computer. The impact was minimal during the dynamic
- simulator examination process. The process computer was made available later in
!- the week during administration of the JPM (-
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V. Manaaement Meetinas
'X* Exit Meeting Summary The NRC examiners conducted an exit meeting with members of licensee i management on October 11,1996. The licensee acknowledged the generic observations presented and indicated that materials reviewed were not considered proprietar I I
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c PARTIAL LIST OF PERSONS CONTACTED Licensee D. Cook, Operations Manager -
J. Kudalls, Support Services Director A. Chernick, Training Supervisor R. Armitage, License Training Supervisor D. Bowman, Lead Examination Developer NRC C. Miller, Senior Resident Inspector L. Collins, Resident inspector
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O Enclosure 2 SIMULATION FACILITY REPORT Facility Licensee: Quad Cities Nuclear Power Station Facility Licensee Docket Nos: 50-254; 50-265 Operating Tests Administered: October 7 -11,1996 The following documents observations made by the NRC examination team during the August 1996 initiallicense examination. These observations do not constitute audit or inspection findings and are not, without further verification and review, indicative of noncompliance with 10 CFR 55.45(b). These observations do not affect 4!RC certification or approval of the simulation facility other than to provide information which may be used l in future evaluations. No licensee action is required in response to these observation During the conduct of the simulator portion of the operating tests, the following item was observed:
ITEM DESCRIPTION Plant Process Computer The simulated plant process computer failed prior to the dynamic scenarios and was not available for applicant us This did not impact the examination as operators were able to use instruments to obtain all necessary data to respond to simulated event . .- . _ -. - - - . . . ._
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Enclosure 3 ;
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Written Examination Weaknesses
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Question # Description
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RO #3 Switch manipulations necessary to place torus cooling in service during a LOCA.
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RO #13 Automatic actions that take place on an automatic I
voltage regulator failur RO #14 Reactor building vent fan response to a trip signal.
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RO #16 Knowledge of the interlocks on the 1/2 250 VDC battery charger supply breakers.
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RO #35 Fuel clad temperatures after an RPV emergency depressurizatio RO #36 RPV levelinstrument response under specified
RO #42/SRO #6 Actions to take when a limiting rod pattern exists with !
a failure of the rod block monito !
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RO #46/SRO #10 Effect of placing the ADS drywell pressure reset switch
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- RO #62/SRO #26 SBGT response to a loss of RPS bus power supplie RO #97/SRO #61 Indications of core instabilities.
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SRO #81 Drywell pneumatic local alarm indications and meanings.
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1 l U. S. NUCLEAR REGULATORY COMMISSION QUAD CITIES NUCLEAR POWER STATION WRITTEN EXAMINATION APPLICANT INFORMATION Name: MASTER EXAMINATION Region: lll ,
Date: 10/11/96 Facility / Unit: QUAD CITIES UNITS I & 11 License Level: RO Reactor Type: GE
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INSTRUCTIONS 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 percen '
Examination papers will be picked up 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after the examination start All work done on this examination is my own. I have neither given nor received ai Applicant's Signature
. RESULTS Examination Value 9PMPoints Applicant's Score Points Applicant's Grade Percent
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ES-402 Policies and Guidelines Attachment 2 i for Taking NRC Written Examinations i
l Cheating on the examination will result in a denial of your application and could result i in more severe penalties.
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- 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 examination.
! To pass the examination, you must achieve a grade of 80 percent or greater.
, The point value for each question is indicated in parentheses after the question i number.
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! There is a time limit of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for completing the examination.
! 6'. Use only black ink or dark pencil to ensure legible copies.
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[ Print your name in the blank provided on the examination cover sheet and the answer
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l Mark your answers on the answer sheet provided and do not leave any question blank.
! If the intent of a question is unclear, ask questions of the examiner only.
l 1 Restroom trips are permitted, but only one applicant at a time will be allowed to leave.
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Avoid all c.ontact with anyone outside the examination room to eliminate even the
! appearance or possibility of cheating, i
! 1 When you complete the examination, assemble a package including the examination i questions, examination aids, and answer sheets and give it to the examiner or proctor.
- Remember to sign the statement on the examination cover sheet.
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1 After you have turned in your examination, leave the examination area as defined by the examiner.
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, REACTOR OPERATOR Page 3 i
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OUESTION: 001 (1.dO) .
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l The mechanisms to ensure. adequate core cooling are only core submergence,
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a. and Spray Coolin b. Steam Cooling with injection, and Spray Cooling, c. Steam Cooling without injection, and Spray Coolin0 i
d. Steam Cooling with injection, Steam Cooling without injectio '
UUESTION: 002 (1.00)
Rod withdrawal is in progress with rod F-8 selected. The step has a withdraw limit of 1 i The operator mistakenly moves the selected rod to position 14. All other rods are in ste Which of the following statements best describes the RWM response to this event? !
a. Selected rods except F-8 can either be inserted or withdraw b. No other rod but F-8 can be moved in either direction until F-8 is inserted to within the limits of the current ste c. As long as the operator remains in the same step, the selected rods can be inserted only. No rods can be withdrawn until the withdraw error is correcte d. No rod other than F-8 can be moved in either direction. Rod F-8 can be withdrawn and/or inserted. Withdraw blocks are not applied until a rod is 2 notches past its withdraw limi _ _ __ . . . _ _ _ _ . . . . _ _ . _ . _ _ . . . _ _ . _ . . . . . _ , _ . _ . . _ _ -
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QUESTION: 003 (1.00) .
You have the following plant conditions:
- - Drywell pressure - 3.2 psig l
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Drywell temperature - 170*F j - Torus pressure - 1.8 psig
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Torus temperature - 96*F l
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Reactor water level .+ 25 inches !
1 The plant has scrammed on high Drywell pressure and OCGP 2-3 is being carried out. The 'i
! RHR system was in a normal lineup at the beginning of the transient and all automatic actions !
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occurred as designed. The Shift Engineer orders Torus Cooling started on the "A"RHR Loop.
- Which of the following switch manipulations will have to be performed in order to start Torus l C*ooling On the "A" RHR Loop LAW QCOP 1000-307 l
l Place RHR Loop "B" CONTAINMENT CLG PERMISSIVE SWITCH 17 to ON -!
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. Place RHR Loop "B" RHR SW START PERMISSIVE SWITCH 19 to MANUAL l OVERRD position.
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, MANUAL OVERRD positio j
- Place either RHR Loop "A" or Loop "B" CONTAINMENT CLG PERMISSIVE l j SWITCH 17 to ON position.
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REACTOR OPERATOR Page 5
QUESTION: 004 (1.00) .
- Unit 2 was operating at approximately 65% power when the "A" Recirc Pump tripped. Prior
, to ANY operator actions, a LOCA occurred. Plant conditions are as follows:
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Reactor Power: 0%
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Drywell Pressure: 7 psig and rising
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Reactor Pressure: 890 psig and decreasing
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"A" Recire Loop Riser Pressure: 880 psig and decreasing l
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"B" Recire Loop Riser Pressure: 875 psig and decreasing
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Which ONE of the following AUTOMATIC actions should occur? The "A" recirc loop discharge valve will clos ' ,
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, The "A" LPCI injection valves will be interlocked open only if reactor level drops below -59".
j- ' The "B" LPCI injection valves will be interlocked closed only if reactor pressure b drops below 325 psig.
l QUESTION: 005 (1.00)
I l During power operations on Unit 1, the following annunciators are received:
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901-8 E-9 120/240V ESS SERVICE BUS ON EMERG SUPPLY
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901-8 F-8 ESS SERV UPS TROUBLE Which ONE of the following sources is powering the Essential Service Bus (ESS)? MCC 18-2
, Bus 18 l Bus 17 3 VDC
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REACTOR OPERATOR Page 6 OUESTION: 006 (1.00) -
Which of the following conditions, in and of itself, would require the plant to be in HOT !
SHUTDOWN in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />? gpm pressure boundary leakage for 1 hou . gpm identified Reactor Coolant system ;eakage for greater than 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> gpm unidentified Reactor Coolant system leakage for greater thaii 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> gpm increase in unidentified Reactor Coolant system leakage over a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, with the mode Switch in STARTU ..
QUESTION: 007 (1.00)
The RWCU system in operation with "A" pump running and "B" pump available and in standby. The "A" RWCU pump trips. The system will: remain on line, with the Filter Domins on hold,'when "B" pump auto start , remain on line with the "B" pump auto starting, no low flow is experience ' trip on low flow causing the resin to slough off the Filter Domins, due to a loss of flo remain in that lineup, with the Filter Demins going on hold. "B" pump has no auto start capabilit ,
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OUESTION: 008 (1.00) -
j The Augmented Primary Containment . Vent (" hardened vent") accepts flow from the i and provides a(n) discharge to the .
i" drywell only, treated; Chimney i drywell or torus, untreated; Chimney I torus only, untreated; Standby Gas Treatment System
- drywell of torus, treated; Standby Gas Treatment System i
e dUESTION: 009 (1.00)
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j The Unit One Standby Coolant Valves are located . They are controlled by
{ switch (es) on the panel.
i in the Low Pressure Heater Bay, a common,912-1
' in the High Pressure heater bay, individual,901-5 j in the Low Pressure Heater Bay, individual,901-6 1 outside the Rad Waste Doorway, a common,912-5
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, QUEST!ON: 010 (1.00) -
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- With Unit One at full power, annunciator 901-5-A-1, SCRAM VALVE AIR SUPPLY LOW
- PRESSURE, alarms. What is this alarm actually measuring and what i.s expected to happen if ;
- the condition continues to degrade? H
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, Scram air header pressure is less than 75 psig. If not corrected, the scram i
valves will begin to open, scramming in rod The 1 A Instrument Air Receiver pressure is less than 75 psig. If not corrected,
. the CRD Flow Control Valves will fail open, putting the pump in runout.
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' Scram air header pressure is less than 75 psig. If not corrected, when scram air
, header pressure reaches 50 psig, an automatic scram signal will be generate A Instrument Air Receiver pressure is less than 75 psig. If not corrected, the I
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CRD Flow Control Valves will fait closed, which will prevent the ab,ility to drive rods.
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- QUESTION
- 011 (1.00)
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l Unit 2 is operating at 100% power when all Unit 2 Generator Hydrogen Seal Oil pumps are
. lost. Hydrogen pressure drops and finally stabilizes at 25 psig. Given the resulting hydrogen l pressure for this condition, which of the following represents an allowed generator output?
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l MW,300 MVAR's laggin MW,200 MVAR's leading.
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3 ' REACTOR OPERATOR Page 9 t
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, OUESTION: 012 (1.bO) -
j Unit 1 is shutting down and reactor water temperature is 200 *F, with "A" loop RHR operating
in the SDC mode. The NSO begins to throttle closed MO-1-1001-28A, Outbd LPCI Inj. Vi Which of the following would be an expected response to this manipulation?
, RHR Flow recorder 1-1040-7 indicates that "A" loop flow has decreased and
"B" loop flow has increased.
- RHR Flow recorder 1-1040-7 indicates that "A" loop flow has decreased to less
, than 1500 gpm; NO valves open to provide minimum flow protection.
i RHR Flow recorder 1-1040-7 indicates that "A" loop flow has decreased to less
, than 1500 gpm; MO-1-1001-47, SDC HDR DOWNSTREAM SV, closes if it's 1 breaker is closed.
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i RHR Flow recorder 1-1040-7 indicates that "A" loop flow has decreased: Flow i Indicator 1-1040-11 A " Containment Spray Flow indicates "A" loop flow has l increased.
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! QUESTION: 013 (1.00) -
- Which of the following describes the automatic actions that will occur if the Unit One Main i
Generator Automatic Voltage Regulator fails high while in control?
. An annunciator will alarm immediately.10 seconds later, if the overexcitation
] condition still exists, the Main Generator will trip.
- The Manual Voltage Regulator will take control immediately. 5 seconds later, if the overexcitation condition still exists, the Main Generator will trip.
i An annunciator will alarm and the Manual Voltage Regulator will take control immediately. If the overexcitation condition still exists 10 seconds later, the
, Main Generator will trip, An annunciator will alarm immediately.10 seconds later, the Manual Voltage Regulator will take control. If the overexcitation condition still exists, 5 seconds later, the Main Generator will tri i
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- REACTOR OPERATOR Page 11
j - QUESTION: 014 (1.00) -
Which of the following describes the way Reactor Building Vent Fans respond to an isolation
- - signal? I Both the Reactor Building Supply Fans and Exhaust Fans receive a trip signal
- .
directly from the isolation signa I
! The Reactor Building Supply Fans receive a trip signal directly from the isolation
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signal. The Reactor Building Exhaust Fans only receive a trip signal from the low i- flow condition resulting from the Isolation Dampers closin i.
- The Reactor Building Exhaust Fans receive a trip signal directly from the j- , isolation signal. The Reactor Building Supply Fans only receive a trip signal from
- the low flow condition resulting from the isolation Dampers closing.
- Both the Reactor Building Supply Fans and Exhaust Fans only receive a trip
. signal from the low flow condition resulting from the Isolation Dampers closin )
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REACTOR OPERATOR Page 12
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. QUESTION: 015 (1.00) -
Why must the Priming Line Valve be opened prior to opening the System isolation Valve when resetting a Multimatic Fire Valve after an actuation? The priming line will overpressurize if the System Isolation Valve is opened first.
2 The Main Valve will open and water will flow out of the system if the priming i line is not opened first.
. The priming line will drain water to the drain funnel if it is not opened before the System Isolation Valve.
j . The Main Valve will not allow water to flow on a subsequent actuation if the
. nriming line is not opened first.
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] REACTOR OPERATOR Page 13
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QUESTION: 016 (1.00) -
Which of the following below describes the key interlock for the 1/2 250 VDC Battery Charger I Supply Breakers?
i
! There is a single key which is required to close either of the Supply Breakers located at the 250 VDC Distribution Panels. This prevents closing both at the same time and cross connecting Busses 18 and 28.
- There is a single key which is required to close either of the Supply Breakers i located at the 480 VAC MCCs. This prevents closing both at the same time and cross connecting the Unit One and Unit Two 250 VDC Batterie , There is a single key which is required to close either of the Supply Breakers
,
located at the 250 VDC Distribution Panels. This prevents closing both at the same time and cross connecting the Unit One and Unit Two 250 VDC Batteries.
j There is a single key which is required to close the Supply Bresker at MCC
- 28-2. This administratively prevents closing both at the same time and cross connecting Busses 18 and 28.
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QUESTION: 017 (1.00) -
A steam leak in the MSIV room has caused r om temperature to reach 250*F. This is below
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the Maximum Safe Area Temperature of 304*F. Considering the definition of Maximum Safe Area Temperature, which one of the following stttements is TRUE?
,
l i Equiprnent in the MSIV room necessary for the safe shutdown of the plant is still operabl The temperature is below the room high temperature setpoint but personnel
.
access is precluded.
- The temperature has exceeded the temperature for equipment operability but j personnel access is still allowed.
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! The reactor building ventilation system should be capable of reducing the area
, temperature even with the leak unisolate ,
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REACTOR OPERATOR Page 15
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OUESTION: 018 (1.00) -
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- Given the following plant conditions
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Drywell Flooding is in progress.
- -
Torus level is + 1 inch.
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HPCI suction has automatically swapped to it's alternate suppl All low pressure ECCS systems are operabl i
-
The Unit Supervisor has diremtod injection with systems which take a suction from outside containmen *
Which one of the below listed systems can be used before any further alignment?
i
, RCIC.-
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' LPCI pumps.
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, Standby Coolan I
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I HPCI cooling water pum l l
l QUESTION: 019 (1.00)
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Unit 1 is in Mode 5 and core offload is in progress when the Refuel Foreman calls the control
{ room to report all normal lighting has been lost on the refuel floor. Which one of the following is an effect? .
l
) Loss of all fuel pool cooling pump '
l J Reactor building ventilation isolatio Reactor building door interlock mechanisms fai Refuel floor ARM local auxiliary Unit will not function.
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a REACTOR OPERATOR Page 16 QUESTION: 020 (1.00) -
! The operator was using the " continuous **Mdraw" mode of operation during a reactor
! startup. During rod withdrawal the operatoc cherved indications that the RMCS timer had restarted, the settle function had been completed and the rod was automatically de-selected.
I
The condition that led to this sequence of events was:
4 RPIS Inop.
- A Rod Worth Minimizer rod bloc i l j RMCS Rod Select Relay Maifunction.
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, The operator released the Rod Notch Out Override switc '
4-QUESTION: 021 (1.00)
? DG 1 has just received a valid auto start signal. Which one of the conditions below will start ,
Diesel Generator Cooling Water Pump? The pump will start: I When DG-1 speed reaches 800 RP I
' When DG-1 Start Relay (STR) is energize seconds after DG-1 speed reaches 200 RP seconds after DG-1 Start Relay (STR) is energize l
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! REACTOR OPERATOR Page 17 l
OUESTION: 022 (1.00) -
!
Which item below describes the status of Backup Scram Solenoids 0302-19 A and 19B j following receipt of a scram signal on RPS "A"? i l Both 19A and 19B are energize '
' Both 19A and 19B are de-energize A is energized,19B is de-energize A is de-energized,19B is energize *
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OUESTION: 023 (1.00)
{
The following conditions exist 4 minutes after a LOCA occurred on unit ;
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Reactor level -70" l
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Drywell Pressure 3.5 psig What is the expected response of the Core Spray System if PS-1-0263-52A, ECCS RPV low pressure permissive, fails high and RPV pressure subsequently drop below 325 psig? Both "A" and "B" LPCS pumps will inject into the RP Neither pump injects into the RPV, both remain on minimum flow, "B" CS pump injects into the RPV, "A" CS pump remains on minimum flo "A" CS pump injects into the RPV, "B" CS pump remains on minimum flo ;
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REACTOR OPERATOR Page 18
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OUESTION: 024 (1.00) -
l The following indications are observed on the CRD system following a reactor scra l I
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Amps on the running CRD pump increase from 27 to 3 FCVs 6A and 6B closed.
I -
CRD system flow off scale high, greater than 100 gpm.
- -
Charging header pressure low, less than 1300 psig.
'
Which statement describes the proper action to be taken by the unit operator?
- Start the second CRD pump to reduce load on the operating pump.
'
i 4 , Monitor system operation, indications are normal for the described conditions.
' Place the CRD system flow controller in manual and attempt to reduce system flow, Throttle the drive header pressure control valve to restore normal system l parameter '
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REACTOR OPERATOR Page 19 OUESTION: 025 (1.00) -
Which statement describes how CRD drive header differential pressure is maintained as !
reactor pressure increases from 0 to 1000 psig during a reactor startup? !
1 The operator adjusts both the flow control and the drive pressure contol valve to maintain drive header differential pressur i The operator periodically throttles the drive pressure control valve in the open
,
direction to maintain drive header differential pressure.
l l The operator periodically throttles the drive pressure control valve in the closed l direction to maintain drive header differential pressur . The CRD flow control valve automatically opens to maintain system flow, thus automatically maintaining drive header differential pressure.
,
OUESTION: 026 (1.00)
l l
Assume the plant is operating normally at 100% steady state power. How will the following
! parameters be affected by a loss of air to the reactor feed pump minimum flow valves?
_
indicated Total Answer Feed Flow RFP Disch Press RPV Level RFP Minimum Flow Decrease Decrease Decrease Increase Remain 'as is Remain as is Remain as is Remain as is Increase Remain as is Remain as is increase l Remain as is Remain as is Decrease Remain as is
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REACTOR OPERATOR Page 20 QUESTION: 027 (1.00) -
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Which of the following causes the Main Turbine Bypass Valves to open if the main turbine should trip with the reactor at 60% power?
- Steam line pressure will be greater than the EHC pressure regulator signal.
, The bypass jack signal will be greater than the control valve demand signal.
. The EHC pressure regulator pressure signal will be less than the load limit setting.
l Turbine control valve demand signal will be less than the maximum combined
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flow limiter signal, i 1 i
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REACTOR OPERATOR Page 21 QUESTION: 028 (1.00) -
A full reactor scram has occurred due to a large LOCA coupled with a loss of all Off Site power. All low pressure ECCS has started and is injecting water to the vessel. Reactor water level is -180 inches and rising steadily. You are directed to place all available torus cooling in service. Torus temperature is currently 105*F. Which of the following statements best describes the actions you should take in placing torus cooling in service?
, When torus temperature reaches the Heat Capacity Limit, divert all RHR pumps to establish torus coolin When reactor water level is greater than -142 inches, divert RHR pumps as needed to establish torus coolin ~ When reactor water level is greater than -172 inches, RHR pumps may be diverted as needed to establish torus coolin When torus temperature reaches 1100F, RHR pumps may be diverted as needed to establish torus cooling since reactor water level is greater than -184 inche QUESTION: 029 (1.00)
A total loss of Unit Two 24V DC has occured due to a short in the system. Which of the following best describes the effect on Unit 2 operation? A full reactor scram will occu Main Chimney isolation capability is los Control Room annunciator power will be los RCIC will automatically start, but not injec ._ __ _ _ . _ . _ _ _ _ . _ . _ _ _ . _ - _ . . _ . _ _ . . . . _ _ _ . . _ . . _ . _ . _ _ . _ _ _ . . _ -
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' REACTOR OPERATOR Page 22
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! OUESTION: 030 (1.00)
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Unit 1 is shutdown and midway through a refueling outage. Unit 2 is operating at full power
! and in the third hour of LCO 3/4.9.3 due to the Unit 2 D/G out for planned maintenance and
- TR-22 OOS due to a cooling problem. It is anticipated that TR-22 will be returned to service in four (4) hours. Suddenly, a turbine trip occurs on Unit 2. The 1/2 D/G fails to start. Neither
- the 14-1 to 24-1 nor 13-1 to 23-1 crossties are closed. Which of the following best describes
! the status of the station?
i l Both units have experienced a station blackout.
j A loss of off-site power is in progress on Unit 2.
l , Both units have experienced a loss of off-site power.
- A loss of off-site power and station blackout is in progress on Unit 2.
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[ OUESTION: 031 (1.00)
- A reactor cooldown is in progress with RHR in the shutdown cooling mode of operation.
j Reactor pressure is 50 psig. Fifteen minutes ago, reactor pressure was 90 psig. Which of the i following best describes the overall cooldown rate (degrees F/hr) using the_ past fifteen
- minutes data?
!
i 'F/hr i
j *F/hr.
i *F/hr
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! F/hr
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, OUESTION: 032 (1.00) .
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i The following conditions exist following a LOCA on Unit 1: !
-
The Unit Supervisor has entered OGA 500-3, "Drywell Flooding". i
- -
RPV water level is unknow l
- -
' Core cooling could not be established following entry into the RPV flooding
. procedur A Core Spray pump is pumping water from the torus to the RPV.
1 -
1 B Core Spray pump and all RHR pumps are pumping water from the CCSTs to the RPV.
d -
Additional available systems are being started as directed by procedure.
1 -
Drywell and torus pressures are about 7.5 psig and steady.
l *
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Drywell temperature is 225'F and risin Containment water level is now 28 ft. and rising slowly.
- 4
+
Which action listed below should be taken' next? l
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)
' Spray the drywell using one loop of RH !
t i '
4 Prevent injection from outside containment.
f Install jumpers to defeat the Group 1 isolation, and open the MSIVs.
' Control RPV pressure to prevent exceeding the ADS Valve Tailpipe Limit (OGA
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QUESTION: 033 (1.00) .
i After a transient initiates on Unit 2, the following parameter values are noted on the 902-3 panel:
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Drywell pressure 4.5 psig risin Drywell air temperature 140'F rising.
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Torus pressure 4.8 psig rising.
,
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Torus water temperature 82*F stable.
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Which of the following is indicated? l
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- A safety valve has opened and close l
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- l l A high pressure discharge into the torus airspac !
- The containment is functioning normally following a water break LOCA.
l A high pressure discharge into the drywell and at least one torus to drywell
- ! vacuum breaker is open.
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I OUESTION: 034 (1.00)
,
- With the unit operating normally at 75% power, a. transient causes the turbine generator to i trip. The Unit NSO places the Mode Switch to SHUTDOWN. All systems respond normall Assuming no operator action, reactor pressure vessel level
) Controls at + 15 inches.
j Controls at + 25 inche '
} Drops to the ECCS initiation setpoin Controls at + 18 inches, since level control is in 3 element.
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l QUESTION: 035 (1.00) .
- With all control rods fully inserted, OGA 500-2, STEAM COOLING, directs RPV Blowdown l
, when RPV water level reaches -184 inches and no injection source is availabl '
How will the plant respond because of this action? :
i Fuel temperatures will drop, giving some additional time to regain injectio Reactor core differential pressure will drop, increasing natural circulation.
j Reactor power will drop into the Source Range, lowering the fuel heat up rat Torus radiation levels will rise significantly as steam is discharged from the relief
- valves.
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REACTOR OPERATOR Page 26 i
OUESTION: 036 (1.00) .
Given the following plant conditions:
l -
A LOCA occurred 20 minutes ago.
1 -
Reactor pressure is 150 psig.
' -
- Drywell temperature is 260*F
! -
Reactor building temperature at 198'F l -
RVWLIS backfill is in operation.
The following reactor water levels are noted:
'
l Yarway wide range . -110 inches
- GEMAC lower 400 -140 inches
i Yarway narrow range -50 inches GEMAC upper 400 + 80 inches The Narrow Range Yarway level indicators are indicating higher than actual level because: density has dropped in the instrument reference leg boiling has occurred in the instrument reference leg c, gasses have come out of solution in the instrument reference legs, the reference leg condensing pot ooes not function properly at these temperature i I
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REACTOR OPERATOR Page 27 i
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. QUESTION: 037 (1.00) ' .
!
! Given the following:
!
! - HEATER 2B2 HIGH LEVEL alarm (901-6, E-2) actuates and will not reset.
'
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Feedwater temperature drops 30* Recirculation pumps are in Individual Manual, t - Reactor power has risen 3% and continues to rise.
- -
Reactor power is 55% and core flow is 55% (80% FCL).
You should: manually scram the reacto * reduce recirculation flow by 10%. insert Cram Array control rods to 00.
i drive the selected rod group to its target in position.
'
OUESTION: 038 (1.00)
!
You are reviewing a surveillance completed from the Control Room. One step required an
<
operator to open a valve on an instrument rack in the Reactor Building. How should completion of this step be recorded on the surveillance sheet?
' The local operator places his initials next to the completed step, followed by the
- NSOs initials.
4-l The local operator's initials, placed by the NSO, and the US's initials should be next to the completed step.
- The local operator's initials, placed by the NSO, and the NSO's initials should
! be next to the completed ste The US's initials, the NSO's initials, and the local operator's initials should be j next to the completed step.
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REACTOR OPERATOR Page 28
- QUESTION
- 039 (1.00) .
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Given the following: l
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A reactor startup is in progress.
i -
Reactor pressure is 920 psig.
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Reactor power is 3%.
-
Core flow is 30%.
j Which ONE of the following will directly result in a reactor scram?
I Main Turbine Trip l
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!_ Reactor power of 16%
!
' Full Group 1 isolation i
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Main Condenser vaccuum at 20" Hg i
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QUESTION: 040 (1.00)
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LAW QCAP 230-5, INDEPENDENT VERIFICATION, which of thi following is the PREFFERED 4 method for verifying the position of locked closed valve, 1-1001-127A,1 A RHR LOOP TO RB I FLR DRN SUMP DRN VLV?
l
' Both verifiers check that the valve is locked in the S-lock log.
, Each verifier should remove the lock and attempt to close the valv Both verifiers check that parameters downstream of the valve are correc Turn the valve in the closed direction without removing the locking devic :
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l QUESTION: 041 (1.00) .
'
LAW OOA 4700-6, TOTAL LOSS OF INSTRUMENT AIR,1(2)-3207A, the Low Flow Feedwater Regulator is closed in order to:
i j prevent RFP runout.
. prevent a HPCl/RCIC isolation.
j prevent exceeding +48 inches RPV water leve isolate non-essential loads to preserve the air system as long as possible.
!
!
! QUESTION: 042 (1.00)
Annunciator 901-5 A-7, RBM HIGH OR INOP, has just been received. From QOS 0005-03, i " UNIT OPERATOR'S DAILY SURVEILLANCE OF NUCLEAR LIMITS", it has been determined I
- that a Limiting Control Rod Pattern DOES exis .j You should-3 Leave the failed RBM channel in the tripped conditio j i
i Verify that one RBM channel is operable, then bypass the failed RBM channe Bypass the failed RBM channel and establish a non-limiting control rod pattern.
!
! Bypass the failed RBM and station a second qualified person to verify rod
] movements.
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REACTOR OPERATOR Page 30 QUESTION: 043 (1.00) .
A temporary modification to the Reactor Building Closed Cooling Water System requires that the intent of a step be changed in OCOP 3700-02, RBCCW System Startup and Operatio One interim procedure that affects QCOP 370042 is in effect but it affects a different ste What type of procedure change is required? A Procedure Change Request is required but only to review writing styl A Procedure Change Request is required with independent technical revie Write an additional interim procedure to supplement the existing OCOP 3700-02 and address the new concer * Conduct a 50.59 evaluation for the new concem and attach it to the current interim procedure to address the new concer QUESTION: 044 (1.00)
Annunciator 912-1 D-1, " REACTOR BLDG COOLING WATER LOW PRESSURE", is receive In accordance with QCOA 3700-1, "RBCCW LOW PRESSURE", within one minute you should shutdown the: Drywell Coolers, Fuel Pool Cooling Syste Reactor recirculation pump Reactor Water Cleanup Syste .w-
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I a i OUESTION: 045 (1.00) .
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l Which of the following is a symptom which could indicate a situation requiring use of the l
,
QARPs to bring the affected unit to a safe shutdown condition?
j Severe flooding in the RHR Room
.
l l Severe, uncontrolled fire in any one plant are ;
j Complete loss of AC power to the unit for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
4 Inability to scram the reactor from the Control Room.
- QUESTION: 046 (1.00)
i
. Which of the following correctly describes the effect of placing the Drywell Pressure Reset 4
.
keylock in RESET?
j~ Placing the switch in RESET ivill.
i
- reset the ADS 110 second timer, if it has started.
l reset the ADS 8.5 minute timer, if it has starte '
} . b.
- cause ADS valves to close if they have opened on a valid initiation signal.
i prevent ADS initiation on LO-LO level combined with high drywell pressure.
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REACTOR OPERATOR Page 32
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- QUESTION
- 047 (1.00) .
A high drywell pressure signal (2.5 psig) coincident with a low low reactor level signal ( 59")
j is received but only one Core Spray pump and NO RHR pumps start. ADS will initiate:
. In 110 seconds.
- in 8.5 minute seconds after an RHR pump is started.
i when the second CS pump or any RHR pump is starte !
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QUESTION: 048 (1.00) 2 i
RCIC is in Standby when a valid initiation signal occur *
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The minimum flow valve 1(2)-1301-60 will: ! A'utomatically open and remain ope Remain closed regardless of changes in system flo Remain open, then close when system flow reaches 40 gp Automatically open, then close when system flow reaches 80 gp .. . . . . . . . - . . . - - - . . - . . - . - . -. . -_ - ... ..-.-- - . . - - - . .
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QUESTION: 049 (1.00) MJ p,v Chid EG*si~<
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Given the following conditions: ho coracl 2 me,
- HPCI room temperature 150 CCST level .7 feet (10,000 gals.)
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Reactor Pressure 150 psi Pump Suction Pressure 10" Hg.
!
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Drywell Pressure 2.0 psig How will the HPCI system respond? HPCI turbine sh d tri l
- HPCIpu suction should transfer to the Toru '
- H steam supply isolation valves should clos d HPCI turbine exhaust vacuum breakers should clos ,
QUESTION: 050 (1.00)
Given:
-
Unit Two is at 100% powe Reactor recirculation flow control is in EGC local manua Which of the following describes the unit's response to reducing LOAD SET? Reactor pressure and reactor power will be increase Recirculation flow and reactor power will be reduce Recirculation flow will drop and reactor pressure will increas Turbine megawatts will be reduced and reactor power will remain constan ,
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l OUESTION: 051 (1.00) ,
l The Unit 1 Diesci Generator started on undervoltage and closed in to Bus 14-1. Annunciator 901-8 G8, " DIESEL GENERATOR 1 RELAY TRIP" was then received. The diesel generator j
breaker opened and the diesel shutdown. What is the likely cause of the diesel generator breaker trip and subsequent diesel shutdown?
l A generator overcurrent condition.
! A generator differential overcurrent.
! An under frequency condition on Bus 14-1.
' A generator neutral winding overvoltage faul ,
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l QUESTION: 052 (1.00)
{
Unit 1 and Unit 2 are in COLD SHUTDOWN with one train of SBGTS out of service. Which ONE of the following changes in plant conditions would require BOTH trains of SBGTS to be operable? Unit 1 reactor water temperature rises above 212 Unit 2 Operational Mode is changed to MODE 5 (Refueling). Unit 1 has a loss Secondary Containment Differential pressur Unit 2 Reactor Building Outlet Damper AO 2-5742A is failed shu __ - . _ __ ._ __
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REACTOR OPERATOR Page 35
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l l e l QUESTION: 053 (1.00) .
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Which ONE of the following practices should be used to ensure personnel safety when closing in disconnects to place a 345KV line in service?
' Wear a helmet with a face shield and 500 volt rubber gloves.
4 Verify that any temporary ground connecteo to the line is properly attached.
, Verify both disconnect circuit breakers are open from the control room and i locally, i l
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- Leave protective cards hanging until the disconnect is closed and the breakers l '
are shut.
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j OUESTION: 054 (1.00)
! A caution in OCOP 2300-6, HPCI SYSTEM MANUAL START-UP, states that HPCI system operation should be avoided with torus temperature above 140*F. What is the reason for this caution?
' Pump damage due to cavitation.
. Turbine damage due to high exhaust temperatur '
, Equipment damage due to inadequate lube oil cooling.
S Inadequate condensing in the Barometric Condenser will cause high airborne radiation problem _ ._ _ _ . _ . _ _ . _ . . _ - _ _ . _ . _ _ _ _ . _ _ _ . . _ . . _ . _ _ _ _ _ . _ . _ _ _ . _ _ _ _ _ _ - _ _ _ . - _ _ _ _
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l REACTOR OPERATOR Page 36 l
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QUESTION: 055 (1.00) .
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With the reactor operating normally at full power, RWCU valve MO-2-1201-133, DEMIN BYPASS,is being throttled to maintain RWCU pump discharge pressure while one RWCU filter
domineralizer is removed from service. During the evolution the bypass valve is opened too
far, causing RWCU pump discharge pressure to drop below 1050 psig. Which of the following {'
! actions may occur? ,
- I i- RWCU pumps will experience runout.
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- RWCU pumps will trip on low pump flow.
{ RWCU system will isolate on high temperature.
i The remaining on-service RWCU demin will isolate on low flow.
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OUESTION: 056 (1.00)
.
This year you have accumulated 10 REM Shallow Dose exposure. How much more external l
f dose whole body skin exposure can you receive before you exceed the Legal Annuallimit? '
- Rem J
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{ Rem
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REACTOR OPERATOR Page 37
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I QUESTION: 057 (1.00) .
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You are directed to enter a LOCKED HIGH RAD AREA to perform a valve lineup. You will need:
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- ' ' a Timekeeper to control acces 'i l to wear Class 4 Protective Clothin l l
. to be accompanied by an RP Supervisor.
! to carry a copy of the latest survey map with yo i i
f ObESTION: 058 (1.00)
i Unit 2 is operating at 100% power when Annunciator 902-3 C-2, "OFFGAS HIGH HIGH
.
RADIATION",is received. Which ONE of the following AUTOMATIC actions will occur? i
i AO 2-5406, "OFFGAS TO STACK OR VENT" will close 15 mins after the alarm '
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is received.
' AO 5401 A and B ana 5402A and B, "SJAE SUCTION VLVS", close 15 min after the alarm is received.
I SO 2-5437, " PRESS DRN TK OUTLET", closes immediately, followed by AO
- 2-5406, "OFFGAS TO STACK OR VENT" closing 15 mins. later.
- automatically placed into service 15 mins, after the alarm is received.
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- REACTOR OPERATOR Page 38
f OUESTION: 059 (1.00) .
Given the followng conditions on Unit 1:
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Reactor power is 40%
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APRM "3" fails "Downscale" l
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APRM "3" has not been bypassed -
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! Due to this, Rod Block Monitor (RBM) Channel "7":
1 is not affecte is automatically bypasse * generates a rod withdrawal bloc ' shifts to an alternate reference APR QUESTION: 060 (1.00)
q Given:
-
Core thermal power is 37% of rate Total core flow is 42 M lb/h Which ONE of the following transients, or evolutions, will drive the plant closer to the i instability region? Assume NO. operator action. (Refer to figure O202-24 as necessary.) Increasing recire, flo Lowering Main Condenser Vaccuu A Continuous Control Rod Withdrawa Placing a "D" Feedwater Heater is servic . . . . _. . .. _ _ _ _ , . _ . _ _ _ . _ _ _ _ . . . _ . - . _ _ _ _ _ _ - - -
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REACTOR OPERATOR Page 39
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OUESTION: 061 (1.00) .
Which of the following Reactor recirculation conditions could interfere with proper LPCI Loop
,
Selection? -
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Reactor Recirc Pump A Recirc Pump B
Power Speed Speed j % 100 % 92%
, :
- % 102% 96% % 37 % 32%
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- % 93% 77 %
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i I OUESTION: 062 (1.00)
Both Units are in Mode 5, no fuel moves are in progress. The 1/2 A SBGTS is in PRIMARY and the 1/2 B SBGTS is in STANDBY for upcoming System Engineering testing. RPS "B" is momentarily de-energized while swapping RPS power supplies from RESERVE to NORMA PCIS Group 2 Trip B trips on the loss of RP How will the SBGTS system be affected? ! /2 SBGTS "B" will automatically start 25 seconds after the loss of RP ; /2 SBGTS "A" will automatically start 25 seconds after the loss of RP Both SBGTS trains will remain off since only 1/2 of an initiation signal is presen /2 SBGTS "A" will automatically start,1/2 SBGTS "B" will automatically start 25 seconds later cau. sing "A" to tri .. __ - _ _ . .-__ _ _ _ _ _ _ _ _ _ _ _ ._. . __ _ _ _ . . . _ . . __ . _ . . . _ . _ _ _ _ _ . _
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} REACTOR OPERATOR Page 40 i
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OUESTION: 063 (1.00) .
- Which one of the following would qualify as a " Temporary Alteration" as defined in OAP 300-12, " Temporary Alterations"?
l A circuit card is pulled to disable an annunciato A hose is installed to drain a heat exchanger under an OOS.
. Installation of an electrical jumper for testing under an approved work procedure
- which is to be completed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
. An electrical lead is lifted in accordance with a surveillance procedure which is
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to be completed by the end of shif ,
QUESTION: 064 (1.00)
- .
Which of the following alarms is REQUIRED to be logged in the Nuclear Station Operator
. (NSO) Unit 2 Log?
! All unexpected alarms.
!
i An alarm that comes in and clears within 5 minute ; An unexpected alarm received during plant shutdown that clears after 8 -
,
minutes, t An expected alarm received during a surveillance that is in for the duration of the surveillanc i
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REACTOR OPERATOR Page 41
QUESTION: 065 (1.50) .
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Using the attached P&lD (M-38), determine how FCV 1-1901-58 will be affected by a loss of i MCC 18-3. The valve will: fail in the open positio fail in the closed positio lock up in its current positio I remain in its current position, due to the power feed swapping to MCC 19- .
QUESTION: 066 (1.00) l Which ONE of the following describes the operation of the PCIS Status Box on the Safety Parameter Display System (SPDS)? i A 1/2 Group Isolation signal with no valve movement will cause a RED ligh A full Group Isolation signal with no valve movement will cause a GREEN ligh A full Group Isolation signal with 2 out of 2 isolation valves in a line closed with cause a RED ligh A ful! Group Isolation signal with 1 out of 2 isolation valves in a line closed will cause a GREEN ligh _ _ _ _ _ . _ . . . _ _ _ .. _ . _ . _ _ . . _ . . _ . ._.__._ _ _._ . . _ . . . ._
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l REACTOR OPERATOR ' age 42 ,
- QUESTION
- 067 (1.00) .
! A reactor scram occurs. QGA 100 has been entered. Shortly afte the scram RPS "A" is los l
- What ONE of the following describes the effect on the operators ability to control RPV j pressure? RWCU is isolated and unavaileble for pressure control.
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1 The inboard RCIC isloation valve is closed, making RCIC unavailablefor pressure control.
l The inboard MSIVs are closed, making the Main Condenser unavailable for
- pressure control.
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QUESTION: 068 (1.00)
- Maximum torus cooling is in service on Unit 2 when an electrical fault causes Bus 23-1 to
! deenergize. All other buses remain energized. Which one of the following describes how the i torus cooling lineup will be affected by this bus loss?
i
,. The 2C and 2D RHR Pumps and the 2C and 2D RHRSW Pumps will be
! deenergized.
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$ All RHR Pumps will be operable; the 2A and 2B RHRSW Pumps will be deenergized.
, The 2A and 2B RHR Pumps and the 2A snd 2B RHRSW Pumps will be
- deenergized.
) The 2A and 28 RHR Pumps will be deenergized; all RHRSW Pumps will be j operabl . _ _ _ - - _ _ _ _ _ . _ . _ _ _ _ . ._. . _ . . . _ . _ . _ _ _ . _ _ _ . _ _ _ _ _ _ . . _ -
REACTOR OPERATOR Page 43 i
OUESTION: 069 (1.00) ;
Unit 2 is at 100% power when the RPIS SYS INOP annunciator is received. Subsequently, a recirc. pump seal fails on the "B" pump. The RPIS failure is preventing insertion of the selected rod group. The NSO should: Scram the reacto ! Insert the CRAM arra l l Reduce recirculation flow to 44 Mibm/h Drive rods in reverse sequence to 00 using Emergency i QUESTION: 070 (1.00)
i While moving a spent fuel bundle in the Fuel Pool, a Fuel Pool Storage Low Level Alarm is '
received and Fuel Pool Level is confirmed to be decreasing. WHICH ONE of the following is the expected immediate operator action? Hold the bundle at its current height and stop further bridge movemen b.- Determine the nearest open storage location and place the bundle in that ,
locatio '
I Return the bundle to its original location (assume that it is not the nearest open !
location). l l Complete the move by placing the bundle in its required new position (assume l that it is not the nearest open location).
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I REACTOR OPERATOR Page 44 i
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OUESTION: 071 (1.00) .
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Unit 2 is in Cold Shutdown with both recirc pumps secured. RHR loop B is aligned for SDC but secured. Moderator temperature is 145'F and stable. Vessel level is 96 inches.
j inadvertent cycling of MO-2-1001-43A, RHR PUMP SDC SUCT VLV with the "A" RHR pump i drains open resulting in a vessel level drop of 8 inches.
.
i Which of the following conditions now exist? RPV level is too low to ensure NPSH for restart of RHR pump RPV level is too low to support natural circulation in the vessel.
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i RPV level is too low to maintain thn upper 400 GEMAC reference leg ful *
l j Level is low enough that restart of a recirc pump may cause an inadvertent low i level scram signa I
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! OUESTION: 072 (1.00)
l
- In addition to a sudden rise in indicated total core flow, which one of the following indications
!
would be indicative of a jet pump failure?
. A rise in core thermal powe ,
! '
{ A drop in core plate differential pressur ;
- A rise in main generator electrical output.
- -
j A drop in individual recirc pump flow for a given speed.
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REACTOR OPERATOR Page 45 QUESTION: 073 (1.00) .
Unit 2 is operating at 60% power with a HPCI surveillance in progress when the following annunciator is received:
-
3A TARGET ROCK RELIEF VLV OPEN Assuming reactor pressure is normal, when would the crew be REQUIRED to initiate a manual reactor scram? l When Torus Bulk Water Temperature reaches 95' When Torus Bulk Water Temperature reaches 110* * Immediately AFTER verifying the safety relief valve (SRV) is actually ope Immediately, IF the SRV is actually open and it cannot be closed with the keylock switc l
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I REACTOR OPERATOR Page 46
QUESTION: 074 (1.00) .
Unit 2 is operating at approximately 90% rated power when a voltage transient in the EHC electrical system causes reactor steam dome pressure to increase to 1065 psig. Which one -
of the following statements describes the final plant conditions following this transient? 4 i The reactor scrams and the reactor recirculation pumps tri ' The reactor scrams and the reactor recirculation pumps run back to minimum .;
l speed.
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l i The main turbine control valves open slightly to lower reactor pressure, then l return to their original positio l
[ The main turbine bypass valves open to lower reactor pressure, then close after reactor pressure returns to norma .
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OUESTION: 075 (1.00)
An ATWS has occurred as the reactor failed to scram cn high drywell pressure. Which of the .
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following could occur if ADS initiated when level is intentionally lowered below the ADS !
initiation setpoint? A rapid cooldown of the moderator in the core, leading to a substantial power increas A large reduction in control rod worth because of voiding, leading to a power l
l excursion.
t An increase in injection of relatively cold, unborated water resulting in a rapid power rise.
i Loss of boron from the RPV to the torus, resulting in additional time required i to shut down the reactor.
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REACTOR OPERATOR Page 47 OUESTION: 076 (1.00)
Complete the following statement:
- The basis for the RBM rod block function is to prevent exceeding the
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! MCPR Safety Limit for a single rod withdrawal error from a limiting control rod patter ! LHGR Limit for a fuel node during a single rod withdrawal error from any control I l
rod pattern.
(
l MCPR Operating Umit for multiple rod withdrawal errors from a limiting control l rod patter , )
! APLHGR Limit for a fuel bundle during multiple rod withdrawal errors from any l l control rod pattern.
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QUESTION: 077 (1.00)
i When can a check valve be used as an isolation boundary for an Out of Service? It is not allowe l If both upstream ad downstream sides of the valve are depressurize If the valve is gagged closed with an OOS card on the gagging device.
l If the valve is in series with a gate, globe, or ball valve that is closed on the OOS.
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l QUESTION: 078 (1.00) C0(/d h6# $5 30 *
i Wwliafe achin
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Which of the following immediate actions is ppropriate virhen replacing a blown fuse? (No
- Emergency Exists.)
' Ensure a PlF is writt Contact engi ing prior to replacing the fuse, if it is like for like.
, Immedia y replace the fuse with one of the same amperage rating, until a like for lik n be foun mediately replace the fuse with one of the same voltage rating, until a like
.
for like can be foun I l
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REACTOR OPERATOR Page 49
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QUESTION: 079 (1.00)
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Drywell temperature is 200'F. The RPV was RAPIDLY depressurized from 1000 psig to 50
- psig. Which ONE of the following correctly describes the effect of the RVWLIS backfill modification on RPV level indication during this transient? Wide and Narrow range RPV level indication remains reliable throughout the I transient since no gasses come out of solution.
, Narrow range level indication remains accurate but Wide range level is unreliable since RVWLIS is not connected to Wide range instruments, Wide range levelindication remains accurate but Narrow range levelis unreliable
,
since RVWLIS is not connected to Narrow range instrument Wide and Narrow range RPV level indication is unreliable since RVWLIS is designed to counteract the effects of vaporization of reference leg water.
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I REACTOR OPERATOR Page 50 OUESTION: 080 (1.00) .
With Unit 1 at fullload, how will the 1B Feed Water Regulating Valve respond to a trip of the valve's hydraulic pumps? .
.. The vaive will continue to operate in AUTOMATIC using the standby pump from the 1 A ski The valve immediately locks in place with the loss of power to both hydraulic pumps, i The valve will operate using accumulator pressure until low oil pressure causes a locku * The valve will automatically close and the 1 A valve will open as needed to provide feed flow to the reactor.
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l QUESTION: 081 (1.00)
- Which of the following describes the response of the CAM system when reactor water level i
reaches -59 inches? Either the Torus sample valves or the Drywell sample valves will open immediately on each subsyste Both the Torus sample valves and the Drywell sample valves will open immediately on the selected subsystem.
, Either the torus or drywell sample valves will open when 81/2 minutes has expired or RPV pressure drops to 325 psi Both subsystems will receive a permissive signal; their selected valves will not open until Drywell pressure reaches 2.5 psig.
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REACTOR OPERATOR Page 51 l QUESTION: 082 (1.00) .
i With the 1/2C RBCCW Pump running on Bus 19, what actions must be taken to restart the
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pump on Bus 19 if a LOCA on Unit One causes Drywell pressure to increase > 2.5 psig?
, Place BOTH of the U1 DIV I and DIV ll DW CLR/RBCCW/FPC TRIP BYPASS Switches in the BYPASS position and manually restart the pum Place BOTH of the U1 DIV i and DIV 11 DW CLR/RBCCW/FPC TRIP BYPASS Switches in the BYPASS position and the pump will restart automaticall Place EITHER the U1 DIV 1 or the U2 DIV 1 DW CLR/RBCCW/FPC TRIP BYPASS Switches in the BYPASS position and manually restart the pump.
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- Place EITHER the U1 DIV l or the U2 DIV 1 DW CLR/RBCCW/FPC TRIP BYPASS
- Switches in the BYPASS position and the pump will restart automatically.
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- QUESTION
- 083 (1.00) .
With Unit One at full power, annunciator 912-1-D-2, TURB BLDG COOLING WATER LOW
- PRESS, alarms. The SER and panel indications show that the problem is with the Unit One
.
TBCCW system. Which of the following describes what the alarm tile is telling the operator j and what are the possible consequences if the condition is not corrected?
i There_ is low pressure at the inlet of the Unit One TBCCW Pumps. If not
- corrected, the loads cooled be TBCCW will overheat.
i l There is low pressure at the outlet of the Unit One TBCCW Pumps. If not
- - corrected, the loads cooled by TBCCW will overheat.
!
! There is a low levelin the Unit One TBCCW expansion tank. If not corrected, I *
the TBCCW pumps will trip on low suction pressure.
i I There is a low levelin the Unit One TBCCW expansion tank. If not corrected, l
selected loads cooled by TBCC5N will auto trip on low cooling flow signals.
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- QUESTION
- 084 (1.00)
l Given a copy of OCOP 5400-7 Att. A., determine the required Preheater Outlet Pressure to
- decrease 1 B Recombiner Outlet temperature from 575'F to 550*F. Preheater outlet pressure '
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is currently 250 psig, i psig psig psig psig
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l l OUESTION: 085 (1.00)
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OGA Detail D-1 tells you that an RPV water level instrument may not be used if Drywell
temperature is at or above RPV Saturation Temperature.
This is because:
.
! the variable leg is assumed to have flashed, causing level to read falsely low.
. the reference leg is assumed to have flashed, causing level to read falsely high.
. bothe the variable and reference legs are assumed to have flashed, causing level l to read falsely lo *
l outgassing of non-condensibles is assumed to have occurred, causing level to
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read falsely high.
k I i-4 OUESTION: 086 (1.00)
. An ATWS has occurred, only one quarter of the control rods are inserted. RPV water level
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is being maintained between -120 and -80 inches Reactor pressure is 850 psig. Hot .
j Shutdown Boron Weight has just been injected. Under which condition below would you l expect the reactor to go critical again?
' Cooldown of the reactor.
I Placing RCIC in service to maintain vessel leve Placing RWCU on service to stabilize reactor pressure.
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- Decay of Xenon concentration over the next several hours.
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, OUESTION. 087 0.00) .
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! According to OCFHP 110-2, inadvertent Criticality During Fuel Moves, which of the following
- indications will positively identify a criticality event in progress while a fuel bundle is being l lowered into the core during refueling operations?
! Source range monitor spiking repeatedly.
i
' A high refuel floor radiation alarm sounds.
f Refuel bridge reverse motion interlock activates, A sustained upward trend on the nearest source range instrument to the fuel
bundle location.
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, QUESTION: 088 (1.00)
i h Modification M-4-2-94-OO7, performed during the Spring 1995 Outage on Unit 2, installed four shroud stabilizers on the Core Shroud. What is the purpose of this modification? Minimize the possibility of shroud access cover failure.
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- To add seismic restraints to limit vertical motion of the shroud.
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! Prevent flow paths from inside the shroud to the annulus regio Prevent existing cracks from providing a flow path to drywell atmosphere.
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QUESTION: 089 (1.00) .
It has been determined that LPRM 24-57A must be bypassed. When the LPRM m;xle switch (S1) is placed in BYPASS the following annunciators are received at the 90X-5 panel.
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CHANNEL A/B NEUTRON MONITOR
-
CHANNEL 1-3 APRM Hi-HI OR INOP j
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CHANNEL A REACTOR SCRAM I
Which of the following is correct with regard to the LPRM system?
' The selected LPRM was the ninth detector bypassed in APRM 1.
, The seiseted LPRM was the eighth detector bypassed in APRM , The selected LPRM was the second A level detector bypassed in APRM .
. The selected LPRM was the third A level detector bypassed in APRM . _ . . .. . . . _ _ _ _ - - . . ___ ____ _ _ . - __ , _. . . ._
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QUESTION: 090 (1.00)
RCIC is operating in RPV pressure control mode following a transient from full power operations. RPV level was never lower than -35". The annunciator which indicates RCIC
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pump suction auto transfer actuates. Which one of the following describes the expected RCIC j system response to these conditions?
i Torus suction valves, MO-25 & 26 open. CCST suction valve MO 22 remains
- ope . Torus suction valves, MO-25 & 26 remain closed. CCST suction valve MO-22 remains ope ' Torus suction valves, MO-25 & 26 open. CCST suction valve MO-22 closes when MO-25 & 26 are full ope Torus suction valves, MO-25 & 26 open. CCST suction valve MO-22 closes as soon as MO-25 & 26 begin to ope _ _-
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OUESTION: 091 (1.00) .
- A plant startup is in progress. Assume all rods are fully inserted and RPV pressure is zero when rod pull to criticality is commenced. Which one of the following startup/heatup scenarios meets ALL of the Tech. Spec. requirements for a reactor startup?
i -F prior to -F when -F at -F at POAH -F at POAH 4 pulling rods critical POAH plus 30 min plus 60 min
! F 95F 102F 142F 182F I
! F 105F 108F 160F 200F i
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i l * F 111F 115F 165F 210F i
i F 118F 120F 162F 222F i
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OUEST!ON: 092 (1.00)
The plant is in Operational Mode 5. All shorting links have been removed to comply with Tec j Specs. A scram signalis received on both divisions of RP '
Which of the following could have generated the trip signal? SRM down scale, <3 cp ! SRM detector not full i ! SRM high count rate,10E6 cp SRM function switch not in OPERAT .- . . - . - - -- .. .. .-. _.. .- -.- - . ..
REACTOR OPERATOR Page 58 i
OUESTION: 093 (1.00) .
A plant transient has occurred which requires the injection of boron into the RPV via Standby Liquid Control. The operator places the SBLC system control switch to the 2+1 positio What is the expected response of the SBLC pumps to the initiation if the B SQUlB valve fails to fire? Both SBLC pumps should start and supply a total of 80 gpm to the RP Both SBLC pumps should start and supply a total of 40 gpm to the RP A SBLC pump should start and supply 40 gpm to the RPV, 8 SBLC pump remains of ' A SBLC pump should start and supply 80 gpm to the RPV, 8 SBLC pump remains of . _ _ . . _ _ _ . _ _ _ . ._. . _ . _ _ _ _ _ _ . _ . . . _ _ . . _ _ _ ___ _ .. _ . . .
i REACTOR OPERATOR Page 59
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i OUESTION: 094 (1.00)
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The plant is operating at 100% steady state power. The operator notes the "A" steam line flow indicator,640-23A, on 90X-5 is trending toward zero. Assume no operator action is taken. Which statement below describes the rssponse of RPV level if this indicator continues
'
to drop to zero?
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RPV level will:
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1 docrease to the low level trip setpoin increase to the high level trip setpoint.
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, decrease and be maintained at a new, lower, steady state water level.
l increase and be maintained at a new, higher, steady state water level.
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OUESTION: 095 (1.00) .
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A LOCA is in progress. RHR Loop Select Logic has determined that injection into the RPV will
, be through the B recirculation loop. MO-1000-298 fails to automatically open when the
! reactor low pressure premissive is satisfied. Select the statement below that describes the action (s) necessary to initiate RHR LPCI flow into the RPV. Assume that RPV level remains below -59" and injection valve 29B cannot be opene I
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RHR injection valves 28A and 29A must be manually opened because of loop select interlock l j, Reset the LPCl Loop Select Logic, RHR injection valves 28A and 29A' will j automatically ope * Reset the LPCI Loop Select Logic, then open RHR injection valves 28A and 29A using the control switches on 90X- l Wait for the 5 minute Loop Select Timer to time out, then open RHR injection valves 28A and 29A using the control switches on 90X- !
I
m REACTOR OPERATOR Page 61 QUESTION: 096 (1.00)
j Unit 1 is operating at full power when a transient occurs. During the transient, alarms 901-5-l F-8, RX VESSEL LOW LEVEL, and 901-5 G-8, FW PUMP MAXIMUM CAPACITY, actuate, and j RPV level is noted to be +5 inches and rising rapidly. The.NSO attempts to control level I
{ ' manually. Which of the following is TRUE regarding manual RPV level control? I The operator will have control over all feed reg valves, since RPV level is >-20 inche The operator will have control over the low flow valve, but not the main feed regulating valve Placing feed regulating valve controllers in manual will have no effect on feed
flow to the reacto Depressing the 1 A(B) VLV RESET pushbuttons on the 901-5 panel will allow the operator to control level manuall l I
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j REACTOR OPERATOR Page 62
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CUESTION: 097 (1.00) .
, The plant was operating at 100% rated power when the "A" recirculation pump tripped. As 1 an immediate action of OCOA 0202-4, the operator is directed to monitor for oscillations.-
- indicating core instability. Which of the following is an indication that core wide oscillations are occurring?
l i Excessive core plate Dp noise exceeding a value of 0.5 psi peak to peak.
" Regular oscillations of reactor water level with a 2-3 second periodicity, i
j Noise signal on the LPRMs with a characteristic periodicity of 1.5 to 2.5 j second *
1 High values of APRM noise that occur with no regutar frequency and are l random in magnitude, i
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l- . QUESTION: 098.(1.00)
!
, . Unit 2 is operating at 80% power when the gland seal and gland steam 55 foot loop seal is i blown; condenser vacuum begins to decrease. Which of the following IMMEDIATE actions 4 should be taken?
i
- Trip the main turbine at 5 in. Hg back pressure.
- -
l Continuously insert all CRAM rods to position 00.
I 4 Reduce recirc flow and insert control rods as necessar .
!
1 Send Rad Protection to monitor the turbine building for high airborne activity.
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. REACTOR OPERATOR Page 63
QUESTION: 099 (1.00) .
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Unit 1 is in cold shutdown with RHR loop A in shutdown cooling when RHRSW is lost. The l
following plant conditions exist:
- -
Reactor water temperature is 190*F and rising at 2'F every 10 minutes.
'
-
Reactor water level is + 40 inche The drywell airlock is open.
j -
Recirc loop A is in operation, loop B is secure Assuming that shutdown cooling cannot be restored, which of the following actions should l
,
be considered within the next 1 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />? l 2 Re-establish primary containment integrit .
J Increase CRD flow to promote shutdown coolin Place the B recirculation loop back in servic ! Shut the MSIVs to prevent flooding the main steam line QUESTION: 100 (1.00)
Work in the area of the Unit 1 Reactor Building ventilation radiation monitors has resulted ir-the loss of signals from both the A and B radiation monitor channels. The crew should verify: Unit 1 Reactor Building ventilation has isolate Reactor Building ventilation continues to operate normall Only the inboard Reactor Building ventilation dampers have shu Only the outboard Reactor Building ventilation dampers have shu (* * * * * * * * * * END OF EXAMINATION * * * * * * * * * *)
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REACTOR OPERATOR Page 1
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ANSWER: 001 ANSWER: 007 REFERENCE: REFERENCE:
OGA INTRO LP page 32,34 LIC-1200 page 12 295031K101 ..(KA's) NLO-12OO 204000A204 ..(KA's)
ANSWER: 002 ANSWER: 008 REFERENCE: i LIC-0207 page 16,18 REFERENCE:
OCOP 207-1 page 2 LIC-1602 Figure 1 ,
201006K511 ..(KA's) 295010K301 ..(KA's)
ANSWER: 003 ANSWER: 009 REFERENCE: REFERENCE:
LIC-1000 page 38 LIC-3900 page 12,26 j
~219000K403 ..(KA's) 295031A108- ..(KA's) "
- ANSWER: 004 ANSWER: 010 REFERENCE: REFERENCE:
LIC-1000 L C-470 OCOA 1000 page 1 QCAN 901(2) 5-A-1 1 203OOOK411 ..(KA's) 295019 GOO 5 ..(KA's)
ANSWER: 011 ANSWER: 005 REFERENCE: -1 REFERENCE: LIC 5300 page 50 I OCAN 901-8 E-9 2 QCOP 5100-5 page 3 6000-1 Electrical Distribution OCGP 1-1 Attachment A page 2 262002K401 ..(KA's) 245000K603 ..(KA's)
i ANSWER': 006 ANSWER: 012 ' REFERENCE: REFERENCE:
LIC-0201 LIC-1000 TSUP 3.6.H page 11 QCOP 1000-05 page 9 290002G005 ..(K A's) 205000K502 ..(KA's)
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REACTOR OPERATOR Page 2 l ANSWER: 013 b f ^#MSWER: 019 l WO C0"*0 3*S # ' d.
REFERENCE: REFERENCE:
LIC- 00 ILF-1800 page 30 AN-901(2)-8-G-10 295034K101 ..(KA's) ,
245000K609 ..(KA's)
l ANSWER: 020 ANSWER: 014 i REFERENCE: l REFERENCE: LIC-0281 LP page 36 4E-1400A 201002K301 ..(KA's)
4E-13878 LIC-5751 page 32 l 288000K402 ..(KA's) ANSWER: 021 REFERENCE:
ANSWER: 015 LIC 6600 K104 ..(KA's) ,
REFERENCE:
LO/NLO-4100 page 30 QCOP 4100-4 ANSWER: 022 286000K504 ..(KA's) REFERENCE:
LIC-0500 AN3WER: 016 212000K408 ..(KA's)
d.
REFERENCE:
QOP 6900-1 page 6,7 ANSWER: 023 263000K402 ..(KA's) REFERENCE:
4E-1431 ANSWER: 017 4E-1430 Sheet 2 LIC 1400 page 50 REFERENCE: 4E-1430 Sheet 1 OGA 300 LP page 10 216000K306 ..(KA's)
295032K104 ..(KA's)
ANSWER: 024 ANSWER: 018 REFERENCE:
REFERENCE:
OGA 500-3 IC 0300-2 295024K101 ..(KA's) 201001A204 ..(KA's)
-. - - - _ _ .
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REACTOR OPERATOR Page 3-f ANSWER: 025 ANSWER: 031 .
' REFERENCE: REFERENCE:
LIC 300 2 QCOS 0201-02 201001A308 ..(KA's) Steam Tables 295021A201 ..(KA's)
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ANSWER: 026 ' ANSWER: 032 REFERENCE: LIC-3200 page 80 REFERENCE:
259001K601 ..(KA's) QGA 600-3 LP page 10,12 296029G012 ..(KA's)
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ANSWER: 027 ANSWER: 033 REFERENCE: REFERENCE:
LIC-5652 Figure 6 LIC 1600-1 page 82,84 i 295005K307 ..(KA's) 295024A204 ..(KA's)
!
ANSWER: 028 ANSWER: 034
, REFERENCE: REFERENCE:
) QGA 200 LP page 28 QCOP 0600-02 page 3 OGA Intro LP, page 33,34 295006K202 ..(KA's)
295013A101 ..(KA's)
.
ANSWER: 035
>
ANSWER: 029 REFERENCE:
. REFERENCE: QGA 500-2 LP page 6 OOA 6900-3 page 2 295031K305 ..(KA's)
295004K203 ..(K A's)
ANSWER: 036 ANSWER: 030 REFERENCE:
REFERENCE: QGA Details LP, pages 4,6, and QCOA 6100-3 295028K203 ..(KA's)
295003K106 ..(KA's)
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REACTOR OPERATOR Page 4
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, ANSWER: 037 ANSWER: 043 REFERENCE: REFERENCE:
- QCOA 3500-1 page 2 LER 1-93-017 !
'295014A102 ..(KA's) QCAP 1100-5 page 5 QCAP 1100-4 l
'
ANSWER: 038 294001A103 ..(KA's) REFERENCE:
QCAP 211-2 page 4 ANSWER: 044 1
.294001A106 ..(KA's) !
REFERENCE:
OCOA 3700-1 page 2 )
ANSWER: 039 295018K201 ..(KA's) REFERENCE:
ILT 0500 page 44 ANSWER: 045 212000K412 ..(KA's) REFERENCE:
QARP 000-2 page 1 ANSWER: 040 QARP LP G011 ..(KA's)
REFERENCE:
OCAP 230-5 page 10 294001K101 ..(KA's) ANSWER: 046 REFERENCE:
ANSWER: 041 .LIC-0203 page 46 K501 ..(KA's)
REFERENCE:
QOA 4700-6 page 3 295019K203 ..(KA's) ANSWER: 047 REFERENCE:
ANSWER: 042 LIC-0203 page 38 K102 ..(KA's)
REFERENCE:
QCOS 0700-07 page 5
.215002A205 ..(KA's) ANSWER: 048 REFERENCE:
LIC-1300 page 20 217000A201 ..(KA's)
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REACTOR OPERATOR Page 5 J
ANSWER: 049
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ANSWER: 055 deleted REFERENCE: REFERENCE:
LIC-2300 page 70 QCOP 120011 page 2
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QCOP 2300-01, page 3 and 4 LIC-1200 QCAN 901(2)-3 A-1 204000G010 ..(KA's)
, 206000K419 ..(KA's)
ANSWER: 056
-
ANSWER: 050 REFERENCE: REFERENCE:
ILT-HP-CH2 IC-5652 page 70 TH-HP-Chapter 2 page 6
'
241000A112 ..(KA's) QCAP 630-06, page 17 294001K103 ..(KA's)
ANSWER: 051 ANSWER: 057 REFERENCE: LO/NLO-6600 page 44 REFERENCE:
, 264000K401 ..(KA's) ILT-HP-CH2 TH-HP-Chapter 2 page 20 QCAP 0620-01, page 8 ANSWER: 052 294001K104 ..(KA's)
a.
-
REFERENCE:
LIC-7500 page 50 ANSWER: 058 TSUP 3. a.
'
261000G005 ..(KA's) REFERENCE:
LIC-5450 page 44 QCAN 901(2)-3 C-2
- ANSWER: 053 271000K408 ..(KA's)
i c.
- REFERENCE
QOP 6400-2 pago 1 ANSWER: 059 294001K107 ..(KA's) b.
"
REFERENCE:
l LIC-0705 page 32
. ANSWER: 054 215002K604 .. (K A's) REFERENCE:
LO/NLO-1601 QCOP 2300-06 page 12 295026K101 ..(KA's)
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- l 3.
l REACTOR OPERATOR Page 6
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- ANSWER
- 060 ANSWER: 066 i REFERENCE: REFERENCE:
LIC-0202 page 94 Plant Process Computer
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- ' ANSWER
- 061 ANSWER: 067 l
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i- REFERENCE: REFERENCE:
- LIC-02021 page 110 QOA 7000-01 page 2,3,6 202001K116 ..(KA's) 1600-2 Containment Auxiliaries
'
223002K608 ..(KA's)
t l ANSWER: 062 '
i ANSWER: 068
'
REFERENCE: d.
. LIC-7500 page 22 REFERENCE:
I 261000A201 ..(KA's) LIC-1000 page 88
, 219000K202 ..(KA's)
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ANSWER: 063.
i a .- ANSWER: 069-REFERENCE: a.
. QAP 0300-12 page 2,3 REFERENCE:
- Temp Alts OJT/OJE LIC-0280 294001K102 ..(KA's) QCOA 0280-01 page 1
- 214000K303 ..(KA's)
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ANSWER: 064 ANSWER: 070
- REFERENCE: b.
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QCAP 0211-02 page 11 REFERENCE:
CREW-OJT QCOA 1900-1 page 2
,
294001A106 ..(KA's) .
OCFHP 0110-05 page 1
. 295023G010 ..(KA's)
,
, ANSWER: 065
! ANSWER: 071 REFERENCE: '
P&lD M38 AF REFERENCE:
- LO/NLO-1900 LIC-1000 i 294001A107 ..(KA's) QCOA 1000-02 page 5 j OCOP 1000-05 page 7
- 205000K303 ..(KA's)
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REACTOR OPERATOR Page 7 l
ANSWER: 072 ANSWER: 078 & <
REFERENCE: REFERENCE:
LIC-0800 QCAP 0400-13 page 2
QCOA 0202-01 page 1 294001K107 ..(KA's)
295001A205 ..(KA's)
ANSWER: 079 ANSWER: 073 REFERENCE:
REFERENCE: QCOP 0201-11 page 3,4,5 QCOA 0203-01 page 2 216000K506 ..(KA's)
- LIC-0203 295013A201 ..(KA's)
ANSWER: 080 ANSWER: 074 REFERENCE: LIC-0600 page 54 REFERENCE: 259002K413 ..(KA's)
LIC-0500 page 42 295025K201 ..(KA's)
'
ANSWER: 081 ANSWER: 075 REFERENCE:
_ LIC-2400 page 38
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REFERENCE: 223001K404 . .(KA's)
OGA 101 LP page 6 l
295037G007 ..(KA's)
. ANSWER: 082 ANSWER: 076 REFERENCE: LIC-3700 page 22 REFERENCE: 295018G006 ..(KA's)
j LIC-0705 page 2 215002K401 ..(KA's)
ANSWER: 083 b.
l ANSWER: 077 REFERENCE: OCAN 912-1-D-2
- REFERENCE
- LIC-3800 QCAP 230-4 page 12 295018G005 ..(KA's)
4 294001K102 ..(KA's)
. . .. _. . - - . ..
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REACTOR OPERATOR Page 8
ANSWER: 084 ANSWER: 090 REFERENCE: REFERENCE: i
QCOP 5400-7 Att. LIC-1300
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ANSWER: 085 ANSWER: 091 !
- 295028K203 ..(KA's) QCGP 1-1 page 2 )
0201-1 page 38 290002G005 ..(KA's) l ANSWER: 086 - REFERENCE: ANSWER: 092 QGA 101 LP page 34,36 c.
- 295037K104 ..(KA's) REFERENCE:
LIC 0700-1 215004K402 ..(KA's)
ANSWER: 087 .
- '
REFERENCE: ANSWER: 093 LIC/FH-0805 QCFHP 110-2 page 1 REFERENCE:
295023K103 ..(KA's)
LIC-1100 page 8,18 211000A202 ..(KA's)
ANSWER: 088
LIC 0600 259002K603 ..(KA's)
ANSWER: 089 REFERENCE: ANSWER: 095 LIC-0703 page 62 A308 ..(KA's) REFERENCE:
LIC-1000 page 24 203000A203 ..(KA's)
. . . . - ~ _ . . . - . _ . . ._. _ _ . . . . . . . . - . -.
REACTOR OPERATOR Page 9
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ANSWER: 096 REFERENCE:
OCAN 901(2)-5 G-8 page 1 I QCOA-0201-9 page 3 295009A102 ..(KA's)
ANSWER: 097 REFERENCE:
'
QCOA-0202-4 page 6,7 295001G010 ..(KA's)
ANSWER: 098 REFERENCE:
QOA 3300-02 295002G010 .. (KA's)
ANSWER: 099 REFERENCE:
QCOA 1000-2 295021A201 ..(KA's)
ANSWER: 100 REFERENCE: l LIC 1701 page 84 272000A309 ..(KA's) l l
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(" ' * * * ' * * END OF E.(AMINATION * ' " * * " * * )
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..
REACTOR OPERATOR Page 1 l ANSWER KEY '
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MULTIPLE CHOICE O23 a 001 d 024- b i 002 b 025 d i
003 b - 026 a I
004 '. a 027 a 005 a ~028 b 3 006 ' ' a 029 a
,
007 d 030 d
!
i- - 008 b 031 b 009 c- 032 c 010 a 033 b 011 d 034 b
'012 b 035 a l l
013 d 036 a J 014 a 037 b I i
015 b 038 c l i
016 d- 039 b l
017 a 049 d 01 c 019 d 042 a l 020 a 043 b 021 b 044 c 022 b 045 b i
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Page 2 {
ANSWER KEY l MULTIPLE CHOICE- 068 d 046 d 069 a 047 a 070 b !
048 d 071 b 049 deleted 072 b 050 b 073 d 051' b 074 b 052 a 075 c 053 c 076 a 054 c 077 c 055 a -070 : b 056 b 079 a 057 a 080 c 058 a 081 c 059 b 082 a-060 c 083 b 061 d 084 c 062 a 085 b 063 a 086 a 064 c 087 d 065 ' b 088 c 066 d 089 a 067 a 090 c
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- REACTOR OPERATOR' Page 3 i ANSWER KEY
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MULTIPLE CHOICE i 091 e s-
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092 c-
,
a 093 a i
094 e a
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095 ' a -
v
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096 c
, 097 c
098 c-n
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- 099 a i
100 a i
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(* ' * * " * * * END OF EXAMINATION * * ' * * * * ")
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! U. S. NUCLEAR REGULATORY COMMISSION
,i QUAD CITIES NUCLEAR POWER STATION .
l WRITTEN EXAMINATION l
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APPLICANT INFORMATION !
.
Name: MASTER EXAMINATION Region: lil
!
j Date: 10/11/96 Facility / Unit: QUAD CITIES UNITS I & 11 l
,
! License Level: SRO Reector Type: GE i
l
- *
! INSTRUCTIONS i 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 j the question. The passing grade requires a final grade of at least 80 percent.
- Examination papers will be picked up 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after the examination starts.
.
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All work done on this examination is my own. I have neither given nor received aid.
!
l Applicant's Signature 5 _
j . RESULTS l Examination Value 98h Points
!
! Applicant's Score Points a
i Applicant's Grade
.
_ Percent W
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i i
l ES-402 Policies and Guidelines Attachment 2
! for Taking NRC Written Examinations
!
I Cheating on the examination will result in a denial of your application and could result j in more severe penalties.
I i After you complete the examination, sign the statement on the cover sheet indicating i
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.
j The point value for each question is indicated in parentheses after the question j number.
i j There is a time limit of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for completing the examination.
- 6'. Use only black ink or dark pencil to ensure legible copies.
i
) Print your name in the blank provided on the examination cover sheet and the answer
! sheet.
!
! Mark your answers on the answer sheet provided and do not leave any question blank.
t l If the intent of a question is unclear, ask questions of the examiner onl <
! 1 Restroom trips are permitted, but only one applicant at a time will be allowed to leave.
j
)
Avoid all contact with anyone outside the examination room to eliminate even the !
- appearance or possibility of cheating.
!
l 1 When you complete the examination, assemble a package including the examination ,
i questions, examination aids, and answer sheets and give it to the examiner or procto )
l Remember to sign the statement on the examination cover sheet.
(
, 1 After you have turned in your examination, leave the examination area as defined by
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the examine .
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s SEN;'R REACTOR OPERATOR Page 3 i
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- - QUESTION
- 001 (1.00) .
I i Given the following:
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HEATER 2B2 HIGH LEVEL alarm (901-6, E-2) actuates and will not rese Feedwater temperature drops 30*F.
,
- Recirculation pumps are in Individual Manual.
p -
Reactor power has risen 3% and continues to rise.
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Reactor power is 55% and core flow is 55% (80% FCL). j You should:
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- manually scram the reacto j
- l reduce recirculation flow by 10%.
l 1 insert Crsm Array control rods to 0 i
! drive the selected rod group to its target in position.
. I j QUESTION: 002 (1.00)
l You are reviewing a surveillance completed from the Control Room. One step required an
- operator to open a valve on an instrument rack in the Reactor. Building. How should
- completion of this step be recorded on the surveillance sheet?
, The local operator places his initials ne'x t to the completed step, followed by the ' NSOs initials.
!
, The local operator's initials, placed by the NSO, and the US's initials should be l next to the completed step.
i
! The local operator's initials, placed by the NSO, and the NSO's initials should l
be next to the completed ste l l
l The US's initials, the NSO's initials, and the local operator's initials should be j next to the completed ste !
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SENIOR REACTOR OPERATOR Page 4 i
QUESTION: 003 (1.00) .
Given the following: I
- -
A reactor startup is in progress.
- -
Reactor pressure is 920 psig.
- - -
Reactor power is 3%.
.
-
Core flow is 30%.
Which ONE of the following will directly result in a reactor scram?
i
- Main Turbine Trip
!
l Reactor power of 16%
,
.
1 Full Group i isolation i Main Condenser vaccuum at 20" Hg
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QUESTION: 004 (1.00)
a
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LAW QCAP 230-5, INDEPENDENT VERIFICATION, which of the following is the PREFFERED method for verifying the position of locked closed valve, 1-1001-127A,1 A RHR LOOP TO RB
- FLR DRN SUMP DRN VLV? Both verifiers check that the valve is locked in the S-lock lo Each verifier should remove the lock and attempt to close the valve.
- Both verifiers check that parameters downstream of the valve are correct.
- Turn the valve in the closed direction without removing the locking device.
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SENIOR REACTOR OPERATOR Page 5 l
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OUESTION: 005 (1.00) .
i IAW OOA 4700-6, TOTAL LOSS OF INSTRUMENT AIR,1(2)-3207A, the Low Flow Feedwater
. Regulator is closed in order to:
I
- prevent RFP runou ' prevent a HPCl/RCIC isolation.
! prevent exceeding +48 inches RPV water leve isolate non-essential loads to preserve the air system as long as possible.
,i .
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OUESTION: 006 (1.00)
i l _ Annunciator 901-5 A-7, RBM HIGH OR INOP, has just been received. From OOS 0005-03,
" UNIT OPERATOR'S DAILY SURVEILLANCE OF NUCLEAR LIMITS", it has been determined
! that a Limiting Control Rod Pattern DOES exist.
i 1 You should:
i Leave the failed RBM channel in the tripped condition.
!
,' Verify that one RBM channel is operable, then bypass the failed RBM channel.
' Bypass the failed RBM channel and establish a non-limiting control rod patter Bypass the failed RBM and station a second qualified person to verify rod
{. movements.
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- SENIOR REACTOR OPERATOR Page 6
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OUESTION: 007 (1.00) .
- A temporary modification to the Reactor Building Closed Cooling Water System requires that j the intent of a step be changed in OCOP 3700-02, RBCCW System Startup and Operation.
One interim procedure that affects CCOP 3700-02 is in effect but it affects a different ste What type of procedure change is required?
i A Procedure Change Request is required but only to review writing styl A Procedure Change Request is required with independent technical review.
- Write an additionalinterim procedure to supplement the existing OCOP 3700-02
- and address the new concer *
, Conduct a 50.59 evaluation for the new concern and attach it to the current
,
interim procedure to address the new concern.
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OUESTION: 008 (1.00)
Annunciator 912-1 D-1, "80 ACTOR BLOG COOLING WATER LOW PRESSURE", is receive In accordance with OCOA ISOO-1, "RBCCW LOW PRESSURE", within one minute you should shutdown the: Drywell Cooler Fuel Pool Cooling Syste Reactor recirculation pump Reactor Water Cleanup Syste _ _ _ . _ . , _ . . _ . _ _ _ . . . . _ _ _ _ . _ _ . . _ . _ . . . _ _ _ _ . . . _ _ _ . . _ . _ _ _ _ _ _ _ . _ _ . _ _ . . . . . . _ _ - _ ..
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QUESTION: 009 (1.00) .
Which of the following is a symptom which could indicate a situation requiring use of the
!. QARPs to bring the affected unit to a safe shutdown condition?
i Severe flooding in the RHR Room i Severe, uncontrolled fire in any one plant area.
- Complete loss of AC power to the unit for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
- d.- Inability to scram the reactor from the Control Room.
j dUESTION: 010 (1.00)
Which of the following correctly describes the effect of placing the Drywell Pressure Reset keylock in RESET?
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Placing the switch in RESET will:
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- reset the ADS 110 second timer, if it has started.
) reset the ADS 8.5 minute timer;if it has started.
i cause ADS valvos to close if they have opened on a valid initiation signal.
f prevent ADS initiation or, i.O i.0 level combined with high drywell p.' essur .
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SENIOR REACTOR OPERATOR Page 8 OUESTION: 011 (1.00) .
j A high drywell pressure signal (2.5 psig) coincident with a low low reactor level signal (-59") l is received but only one Core Spray pump and NO RHR pumps start. ADS will initiate:
a. In 110 second b. in 8.5 minute c. 110 seconds after an RHR pump is starte d. when the second CS pump or any RHR pump is starte .
QUESTION: 012 (1.00)
RCIC is in Standby when a valid initiation signal occurs.
The minimum flow valve 1(2)-1301-60 will: Automatically open and remain ope b. Remain closed regardless of changes in system flow, Remain open, then close when system flow reaches 40 gp Automatically open, then close when system flow reaches 80 gp . _ _ _ _ _ _ . _ . . _ . _ _ . _ _ _ _ _ - . _ . _ _ _ _ _ . _ . _ _ _ _ . _ _ _ _ . _ _ . . . _ . . . . _ . . . . _ . . . . _ _ _ _ .
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QUESTION: 013 (1.00)
J .pgr hubww Given the following conditions: go comd Sed 8^
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- HPCI room temperature 150 '
- CCST level .7 feet (10,000 Is.)
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Reactor Pressure 150 psi .
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- Pump Suction Pressur 0" Hg.
4 - Drywell Pressure 2. psig s
How will the HPCI system re nd?
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, HPCI turb' should trip.
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- HP pump suction should transfer to the Torus.
I PCI steam supply isolation valves should close.
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. HPCI turbine exhaust vacuum breakers should clos )
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- OUESTION
- 014 (1.00)
Given:
- Unit Two is at 100% powe Reactor recirculation flow control is in EGC local manua Which of the following describes the unit's response to reducing LOAD SET? Reactor pressure and reactor power will be increase Recirculation flow and reactor power will be reduce Recirculation flow will drop and reactor pressure will increase, Turbine megawatts will be reduced and reactor power will remain constan .. - - _ . _ _ . . _ _ . . . . _ - . _ _ . . _ _ . _ . _ _ . - ._ . _ _ . _ _ . _ ._ _._ _ _ __ __ .-
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! QUESTION: 015 (1.00) .
The Unit 1 Diesel Generator started on undervoltage and closed in to Bus 14-1. Annunciator
901-8 G8, " DIESEL GENERATOR 1 RELAY TRIP", was then received. The diesel generator
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breaker opened and the diesel shutdown.
- What is the likely cause of the diesel generator breaker trip and subsequent diesel shutdown?
l a. A generator overcurrent conditio b. A generator differential overcurrent, c. An under frequency condition on Bus 14- *
d. A generator neutral winding overvoltage faul QUESTION: 016 (1.00)
Unit 1 and Unit 2 are in COLD SHUTDOWN with one train of SBGTS out of service. Which ONE of the following changes in plant conditions would require BOTH trains of SBGTS to be operable?
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i a. Unit 1 reactor water temperature rises above 212* j b. Unit 2 Operational Mode is changed to MODE 5 (Refueling).
c. Unit I has a loss Secondary Containment Differential pressur !
d. Unit 2 Reactor Building Outlet Damper AO 2-5742A is failed shu l i
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l, QUESTION: 017 (1.00) ,
i Which ONE of the following practices should be used to ensure personnel safety when closing
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in disconnects to place a 345KV line in service? Wear a helmet with a face shield and 500 volt rubber glove Verify that any temporary ground connected to the line is properly attached.
l Verify both disconnect circuit breakers are open from the control room and i locally.
l Leave protective cards hanging until the disconnect is closed and the breakers
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are shu QUESTION: 018 (1.00)-
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A caution in OCOP 2300-6, HPCI SYSTEM MANUAL START-UP, states that HPCI system operation should be avoided with torus temperature above 140*F. What is the reason for this caution? Pump damage due to cavitatio Turbine damage due to high exhaust temperatur Equipment damage due to inadequate lube oil coolin Inadequate condensing in the Barometric Condenser will cause high airborne radiation problem ;
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1 QUESTION: 019 (1.00) .
i l With the reactor operating normally at full power, RWCU valve MO-2-1201-133, DEMIN i BYPASS, is being throttled to maintain RWCU pump discharge pressure while one RWCU filter i
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domineralizer is removed from service. During the evolution the bypass valve is opened too 1
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far, causing RWCU pump discharge pressure to drop below 1050 psig. Which of the following l
actions may occur?
- RWCU pumps will experience runout.
i RWCU pumps will trip on low pump flow.
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- RWCU system will isolate on high temperature.
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- The remaining on-service RWCU domin will isolate on low flow.
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i' QUESTION: 020 (1.00)
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This year you have accumulated 10 REM Shallow Dose exposure. How much more external dose whole body skin exposure can you receive before you exceed the Legal Annuallimit? Rem
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j- OUESTION: 021 (1.00) .
- You are directed to enter a LOCKED HIGH RAD AREA to perform a valve lineup. You will need:
i a. a Timekeeper to control access.
- b. to wear Class 4 Protective Clothing.
_ c. to be accompanied by an RP Supervisor.
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j d. to carry a copy of the latest survey map with you.
ODESTION: 022 (1.00)
i Unit 2 is operating at 100% power when Annunciator 902-3 C-2, "OFFGAS HIGH-HIGH RADIATION",is received. Which ONE of the following AUTOMATIC actions will occur?
. a. AO 2-5406, "OFFGAS TO STACK OR VENT" will close 15 mins, after the alarm j is received.
l b. AO 5401 A and B and 5402A and B, "SJAE SUCTION VLVS", close 15 min after the alarm is receive c. SO 2-5437, " PRESS DRN TK OUTLET", closes immediately, followed by AO 2 5406, "OFFGAS TO STACK OR VENT" closing 15 mins. late d. AO 2-5408, " HOLDUP PIPE DRN", valve closes and the Charcoal Adsorber is automatically placed into service 15 mins. after the alarm is receive SENIOR REACTOR OPERATOR Page 14 QUESTION: 023 (1.00) .
Given the followng conditions on Unit 1: l
- Reactor power is 40%
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APRM "3" fails "Downscale"
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APRM "3" has not been bypassed Due to this, Rod Block Monitor (RBM) Channel "7": is not affecte is automatically bypasse * generates a rod withdrawal bloc shifts to an alternate reference APR l OUESTION: 024 (1.00)
i Given:
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Core thermal power is 37% of rate Total core flow is 42 M lb/hr.
Which ONE of the following transients, or evolutions, will drive the plant closer to the instability region? Assume NO operator action. (Refer to figure O202-24 as necessary.) Increasing recirc. flo Lowering Main Condenser Vaccuu A Continuous Control Rod Withdrawa Placing a "D" Feedwater Heater is servic ._ . . _ . . _ _ . . - . . _ . _ _ . _ _ . . _ _ _ . _ . _ _ . . _ _ _ _ _ _ . _ . _ . . _ . . _ _ _ . _ _ .
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j QUESTION: 025 (1.00) .
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l l Which of the following Reactor recirculation conditions could interfere with proper LPCI Loop I
, Selection?
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Reactor Recirc Pump A Recirc Pump B
! Power Speed Speed
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- % 100 % 92%
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- % 102% 96%
l % 37 % 32%
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j % 93 % 77%
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4 QUESTION: 026 (1,00)
. Both Units are in Mode 5, no fuel moves are in progress. The 1/2 A SBGTS is in PRIMARY i
and the 1/2 B SBGTS is in STANDBY for upcoming System Engineering testing. RPS "B" is
[ momentarily de-energized while swapping RPS power supplies from RESERVE to NORMAL.
I /2 SBGTS *B" will automatically start 25 seconds after the loss of RP i I
, /2 SBGTS "A" will automatically start 25 seconds after the loss of RP i
- Both SBGTS trains will remain off since only 1/2 of an initiation signal is
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present.
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i /2 SBGTS "A" will automatically start,1/2 SBGTS~ "B" will automatically start 25 seconds later causing "A" to tri . _ - . . . - . . - . - - . . - . - . . - - . - . . - - . . . . - . . - . . . - . . - . - - . -
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QUESTION: 027 (1.00) .
Which one of the following would qualify as a " Temporary Alteration" as defined in OAP l 300-12, " Temporary Alterations"?
j- a. A circuit card is pulled to disable an annunciator.
] b. A hose is installed to drain a heat exchanger under an OOS.
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c. Installation of an electrical jumper for testing under an approved work procedure which is to be completed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
T d. An electrical lead is lifted in accordance with a surveillance procedure which is
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to be completed by the end of shift.
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1 QUESTION: 028 (1.00)
I Which of the following alarms is REQUIRED to be logged in the Nuclear Station Operator
(NSO) Unit 2 Log?
a. All unexpected alarm b. An alarm that comes in and clears within 5 minute c. An unexpected alarm received during plant shutdown that clears after 8 minutes, d. An expected alarm received during a surveillance that is in for the duration of the surveillanc .- - -= . _ - . - _ _- --- - -..._....-..-- . - _ . . . - - _ . - . - .
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OUESTION: 029 (1.00) .
Using the attached P&lD (M-38), determine how FCV 1-1901-58 will b( iffected by a loss of j MCC 18-3. The valve will:
i j fail in the open positio fail in the closed position.
4 lock up in its current position, remain in its current position, due to the power feed swapping to MCC 19- ;
QUESTION: 030 (1.00)
- Which ONE of the following describes the operation of the PCIS Status Box on the Safety
- Parameter Display System (SPDS)?
' A 1/2 Group Isolation signal with no valve movement will cause a RED light.
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- A full Group Isolation signal with no valve movement will cause a GREEN light.
t i A full Group Isolation signal with 2 out of 2 isolation valves in a line closed with cause a RED ligh A full Group Isolation signal with 1 out of 2 isolation valves in a line closed will cause a GREEN light.
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. QUESTION: 031 (1.00)
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l A reactor scram occurs. OGA 100 has been entered. Shortly after the scram RPS "A" is lost.
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, . What ONE of the following describes the effect on the operators ability to control RPV pressure?
) RWCU is isolated and unavailable for pressure contro i l The Inboard RCIC isloation valve is closed, making RCIC unavailablefor pressure l contro j The inboard MSIVs are closed, making the Main Condenser unavailable for
- pressure contro t
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j The inboard MSL Drain Valve is closed, making the MSL Drains unavailable for l pressure control.
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QUESTION: 032 (1.00)
f Maximum torus cooling is in service on Unit 2 when an electrical fault causes Bus 23-1 to
, deenergize. All other buses remain energized. Which one of the following describes how the
- torus cooling lineup will be affected by this bus loss?
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' The 2C and 2D RHR Pumps and the 2C and 2D RHRSW Pumps will be deenergize All RHR Pumps will be operable; the 2A and 28 RHRSW Pumps will be deenergize The 2A and 2B RHR Pumps and the 2A and 2B RHRSW Pumps will be deenergize The 2A and 2B RHR Pumps will be deenergized; all RHRSW Pumps will be operabl . _ _ _ _
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SENIOR REACTOR OPERATOR Page 19 i OUESTION: 033 (1.00) .
i j Unit 2 is at 100% power when the RPIS SYS INOP annunciator is receiverd. Subsequently, a 1 reciro. pump seal fails on the "B" pump. The RPIS failure is preventing insertion of the ;
l selected rod grou The NSO should: Scram the reactor, i Insert the CRAM arra Reduce recirculation flow to 44 Mlbm/h , Drive rods in reverse sequence to 00 using Emergency i '
l GUESTION: 034 (1.00)
i While moving a spent fue. bundle in the Fuel Pool, a Fuel Pool Storage Low Level Alarm is received and Fuel Pool Le /el is confirmed to be decreasin !
WHICH ONE of the follo wing is the expected immediate operator action? Hold the bundle at its current height and stop further bridge movemen Determine the nearest open storage location and place the bundle in that locatio Return the bundle to its original location (assume that it is not the nearest open location), Complete the move by placing the bundle in its required new position (assume that it is not the nearest open location).
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OUESTION: 035 (1.00)
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i Unit 2 is in Cold Shutdown with both recirc pumps secured. RHR loop B is aligned for SDC but secured. Moderator temperature is 145'F and stable. Vessel level is 96 inches.
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inadvertent cycling of MO-2-100143A, RHR PUMP SDC SUCT VLV with the "A" RHR pump i drains open resulting in a vessel level drop of 8 inches.
Which of the following conditions now exist?
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" RPV levelis too low to ensure NPSH for restart of RHR pumps.
i RPV level is too low to support natural circulation in the vessel.
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' RPV level is too low to maintain the upper 400 GEMAC reference leg full.
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i Level is low enouJh that restart of a recire pump may cause an inadvertent low l t
level scram signa l
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1 QUESTION. 036 (1.00)
in addition to a sudden rise in indicated total core flow, which one of the following indications would be indicative of a jet pump failure? A rise in core thermal powe A drop in core plate differential pressur A rise in main generator electrical outpu A drop in individual recirc pump flow for a given spee i l
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OUESTION: 037 (1.00) .
i Unit 2 is operating at 60% power with a HPCI surveillance in progress when the following l annunciator is received:
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3A TARGET ROCK RELIEF VLV OPEN
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Assuming reactor pressure is normal, when would the crew be REQUIRED to initiate a manual reactor scram?
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. When Torus Bulk Water Temperature reaches 95' ,
, When Torus Bulk Water Temperature reaches 110* l
i Immediately AFTER verifying the safety relief valve (SRV) is actually open.
3 Immediately, IF the SRV is actually open and it cannot be closed with the keylock switch.
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1 QUESTION: 038 (1.00)
- Unit 2 is operating at approximately 90% rated power when a voltage transient in the EHC l j electrical system causes reactor steam dome pressure to increase to 1065 psig. Which one of the following statements describes the final plant conditions following this transient? i i
- The reactor scrams and the reactor recirculation pumps tri The reactor scrams and the reactor recirculation pumps run back to minimum j spee The main turbine control valves open slightly to lower reactor pressure, then j return to their original positio I
- The main turbine bypass valves open to lower reactor pressure, then close after reactor pressure returns to normal.
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OUESTION: 039 (1.00)
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An ATWS has occurred as the reactor failed to scram on high drywell pressure. Which of the i
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following could occur if ADS initiated when level is intentionally lowered below the ADS j
initiation setpoint? A rapid cooldown of the moderator in the core, leading to a substantial power increas A large reduction in control rod worth because of voiding, leading to a power excursio An increase in injection of relatively cold, unborated water resulting in a rapid power ris Loss of boron from the RPV to the torus, resulting in additional time required to shut down the reacto !
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i j QUESTION: 040 (1.00)
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Complete the following statement:
The basis for the RBM rod block function is to prevent exceeding the:
' MCPR Safety Umit for a single rod withdrawal error from a limiting control rod pattern.
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- LHGR Umit for a fuel node during a single rod withdrawal error from any control rod pattern.
, MCPR Operating Umit for multiple rod withdrawal errors from a limiting control rod pattern.
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I APLHGR Umit for a fuel bundle during multiple rod withdrawal errors from any
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control rod pattern.
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!. OUESTION: 041 (1.00)
When can a check valve be used as an isolation boundary for an Out of Service?
! It is not allowed.
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f If both ups+ ream ad downstream sides of the valve are depressurize If the valve is gagged closed with an OOS card on the gagging device.
- If the valve is in series with a gate, globe, or ball valve that is closed on the
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QUESTION: 041 (1.00)
hk g h 'g ( (*?vw #
yg 4.ju Q d l
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Which of the following immediate actions is appropria when replacing a blown fuse ? (No ack Emergency Exists.)
l Ensure a PlF is written
, Contact engineering prior o replacing the fuse, if it is like for like.
! Immediately replac he fuse with one of the same amperage rating, until a like
. for like can be f nd.
i Immediat replac:, the fuse with one of the same voltage rating, until a like i
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for like n be found.
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l SENIOR REACTOR OPERATOR Page 25 OUESTION: 043 (1.00) .
Drywell temperature is 200'F. The RPV was RAPIDLY depressurized from 1000 psig to 50 psig. Which ONE of the following correctly describes the effect of the RVWLIS backfill modification on RPV level indication during this trarsient? Wide and Narrow range RPV level indication remains reliable throughout the tiensient since no gasses come out of solutio Narrow range level indication remains accurate but Wide range level is unreliable since RVWLIS is not connected to Wide range instrument Wide range level indication remains accurate but Narrow range level is unreliable
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since RVWLIS is not connected to Narrow range instrument Wide and Narrow range RPV level indication is unreliable since RVWLIS is designed to counteract the effects of vapcrization of reference leg wate .
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i SENIOR REACTOR OPERATOR Page 26 QUESTION: 044 (1.00) .
With Unit 1 at fullload, how will the 1B Feed Water Regulating Valve respond to a trip of the valve's hydraulic pumps? The valve will continue to operate in AUTOMATIC using the standby pump from the 1 A skid, i The valve immediately locks in place with the loss of power to both hydraulic pump !
. The valve will operate using accumulator pressure until low oil pressure causes 4 a locku ;
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' ' The valve will automatically close and the 1 A valve will open as needed to ;
provide feed flow to the reacto l
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. QUESTION: 045 (1.00)
- l l Which of the following describes the response of the CAM system when reactor water level
reaches -59 inches?
3 Either the Torus sample valves or the Drywell sample valvss will open immediately on each subsystem.
4 Both the Torus sample valves and the Drywell sample valves will open immediately on the selected subsystem.
- Either the torus or drywell sample valves will open when 81/2 minutes has expired or RPV pressure drops to 325 psi Both subsystems will receive a permissive signal; their selected valves will not open until Drywell pressure reaches 2.5 psi _ _ _ _ .__ _ _ . _ _ _ ._ . . . _ _ _--__ . _ .
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, SENIOR REACTOR OPERATOR Page 27 QUESTION: 046 (1.d0) .
j With the 1/2C RBCCW Pump running on Bus 19, what actions must be taken to restart the pump on Bus 19 if a LOCA on Unit One causes Drywell pressure to increase > 2.5 psig? Place BOTH of the U1 DIV I and DIV 11 DW CLR/RBCCW/FPC TRIP BYPASS
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Switches in the BYPASS position and manually restart the pum Place BOTH of the U1 DIV i and DIV 11 DW CLR/RBCCW/FPC TRIP BYPASS Switches in the BYPASS position and the pump will restart automatically.
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' Place EITHER the U1 DIV i or the U2 DIV 1 DW CLR/RBCCW/FPC TRIP BYPASS Switches in the BYPASS position and manually restart the pump.
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Place EITHER the U1 DIV I or the U2 DIV 1 DW CLR/RBCCW/FPC TRIP BYPASS j Switches in the BYPASS position and the pump will restart automaticall l
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- l QUESTION: 047 (1.00) .
With Unit One at full power, annunciator 912-1-D-2, TURB BLDG COOLING WATER LOW
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PRESS, alarms. The SER and panel indications show that the problem is with the Unit One TBCCW system. Which of the following describes what the alarm tile is telling the operator and what are the possible consequences if the condition is not corrected? There is low pressure at the inlet of the Unit One TBCCW Pumps. If not
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corrected, the loads cooled be TBCCW will overheat.
- There is low pressure at the outlet of the Unit One TBCCW Pumps. If not
corrected, the loads cooled by TBCCW will overhea * There is a low level in the Unit One TBCCW expansion tank. If not corrected,
, the TBCCW pumps will trip on low suction pressure.
- There is a low level in the Unit One TBCCW expansion tank. If not corrected, selected loads cooled by TBCCW will auto trip on low cooling flow signals.
QUESTION: 048 (1.00)
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J Given a copy of GCOP 5400-7 Att. A., determine the required Preheater Outlet Pressure to decrease 1B Recombiner Outlet temperature from 575 oF to 550*F. Preheater outlet pressure is currently 250 psi psig psig a psig psig
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. QUESTION: 049 (1.00)
! OGA Detail D-1 tells you that an RPV water level instrument may not be used if Drywell temperature is at or above RPV Saturation Temperature. .
l This is because:
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1 the variable leg is assumed to have flashed, causing level to read falsely low.
i j the reference leg is assumed to have flashed, causing level to read falsely high.
l bothe the variable and reference legs are assumed to have flashed, causing level i to read falsely low.
j' * outgassing of non-condensibles is assumed to have occurred, causing level to'.
read falsely high.
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! OUESTION: 050 (1.00)
l An' ATWS has occurred, only one quarter of the control rods are inserted. RPV water level
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is being maintained between -120 and -80 inches. Reactor pressure is 850 psig.- Hot Shutdown Boron Weight has just beein injected. Under which condition below would you i expect the reactor to go critical again?
, Cooldown of the reactor.
j Placing RCIC in service to maintain vessel level.
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l Placing RWCU on service to stabilize reactor pressure.
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) Decay of Xenon concentration over the next several hours.
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QUESTION: 051 (1.00) .
According to OCFHP 110-2, inadvertent Criticality During Fuel Moves, which of the following indications will positively identify a criticality event in progress while a fuel bundle is being
- lowered into the core during refueling operations? Source range monitor spiking repeatedl A high refuel floor radiation alarm sounds.
. Refuel bridge reverse motion interlock activates.
l j A sustained upward trend on the nearest source range instrument to the fuel bundle locatio ,
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l QUESTION: 052 (1.00)
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Medification M-4-2-94-OO7, performed during the Spring 1995 Outage on Unit 2, installed j four abroud stabilizers on the Core Shroud. What is the purpose of this modification?
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- , Minimize the possibility of shroud access cover failure.
-: To add seismic restraints to limit vertical motion of the shrou Prevent flow paths from inside the shroud to the annulus regio Prevent existing cracks from providing a flow path to drywell atmosphere.
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SENIOR REACTOR OPERATOR Page 31 ,
QUESTION: 053 (1.00) .
It has been determined that LPRM 24-57A must be bypassed. When the LPRM mode switch *
l (S1) is placed in BYPASS the following annunciators are received at the 90X-5 pane ;
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CHANNEL A/B NEUTRON MONITOR
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CHANNEL 1-3 APRM HI HI OR INOP l -
CHANNEL A REACTOR SCRAM l
l Which of the following is correct with regard to the LPRM system?
l l The selected LPRM was the ninth detector bypassed in APRM 1.
l l The selected LPRM was the eighth detector bypassed in APRM '
! The selected LPRM was the second A level detector bypassed in APRM 2.
l The selected LPRM was the third A level detector bypassed in APRM l l
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SENIOR REACTOR OPERATOR Page 32 i
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QUESTION: 054 (1.00) .
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t i RCIC is operating in RPV pressure control mode following a transient from full power l operations. RPV level was never lower than -35". The annunciator which indicates RCIC
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pump suction auto transfer actuates. Which one of the following describes the expected RCIC system response to these conditions?
l a. Torus suction valves, MO-25 & 26 open. CCST suction valve MO-22 remains j open.
l l b. Torus suction valves, MO-25 & 26 remain closed. CCST suction valve MO-22
! remains open.
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c. Torus suction valves, MO-25 & 26 open. CCST suction valve MO-22 closes when MO-25 & 26 are full open.
l d. Torus suction valves, MO-25 & 26 open. CCST suction valve MO-22 closes as l soon as MO-25 & 26 begin to open.
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SENIOR REACTOR OPERATOR Page 33 l
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QUESTION: 055 (1.00) .
A plant startup is in progress. Assume all rods are fully inserted and RPV pressure is zero l
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when rod pull to criticality is commenced. Which one of the following startup/heatup scenarios meets ALL of the Tech. Spec. requirements for a reactor startup?
(Use the attached Minimum Reactor Vessel Metal Temperature vs. Reactor Vessel Pressure curve as necessary)
-F prior to -F when -F at -F at POAH -F st POAH pulling rods critical POAH plus 30 min plus 60 min F 95F 102F 142F 182F l l F 105F 108F 160F 200F
. F 111F 115F 165F 210F F 118F 120F 162F 222F QUESTION: 056 (1.00)
The plant is in Operational Mode 5. All shorting links have been removed to comply with Tec Specs. A scram signalis received on both divisions of RP Which of the follov.ing could have generated the trip signal? SRM down scale, <3 cp SRM detector not full i SRM high count rate,10E6 cp SRM function switch not in OPERATE.
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l SENIOR REACTOR OPERATOR Page 34 OUESTION: 057 (1.00) .
A plant transient has occurred which requires the injection of boron into the RPV via Standby Liquid Control. The operator places the SBLC system control switch to the 2+ 1 positio What is the expected response of the SBLC pamps to the initiation if the B SOUlB valve fails l
to fire? Both SBLC pumps should start and supply a total t,f 00 gpm to the RP I Both SBLC pumps should start and supply a total of 40 gpm to the RP l A SBLC pump should start and supply 40 gpm to the RPV, B SBLC pump remains of * A SBLC pump should start and supply 80 gpm to the RPV, B SBLC pump l remains of .
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i SENIOR REACTOR OPERATOR Page 35
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i OUESTION: 058 (1.00)
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The plant is operating at 100% steady state power. The operator notes the "A" steam line flow indicator, 640-23A, on 90X-5 is trending toward zero. Assume no operator action is taken. Which statement below describes the response of RPV level if this indicator continues l to drop to zero?
RPV level will:
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! decrease to the low level trip setpoint.
1 increase to the high level trip setpoin '
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- , decrease and be maintained at a new, lower, steady state water level.
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! increase and be maintained at a new, higher, steady state water level.
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, SENIOR REACTOR OPERATOR Page 36 i
!
OUESTION: 059 (1.00) .
j A LOCA is in progress. RHR Loop Select Logic has determined that injection into the RPV will
- be through the B recirculation loop. MO-1000-29B fails to automatically open when the
- reactor low pressure premissive is satisfied. Select the statement below that describes the i action (s) necessary to initiate RHR LPCI flow into the RPV. Assume that RPV level remains below -59" and injection valve 298 cannot be opened.
+ RHR injection valves 28A and 29A must be manually opened because of loop
. select interlock I
' Reset the LPCI Loop Select Logic, RHR injection valves 28A and 29A will
] automatically open.
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4 Reset the LPCI Loop Select Logic, then open RHR injection valves 28A and 29A
,
using the control switches on 90X-3.
. Wait for the 5 minute Loop Select Timer to time out, then open RHR injection valves 28A and 29A using the control switches on 90X-3.
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- SENIOR REACTOR OPERATOR Page 37 l
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OUESTION: 060 (1.00)
Unit 1 is operating at full power when a transient occurs. During the transient, alarms 901-5-F-8, RX VESSEL LOW LEVEL, and 901-5 G-8, FW PUMP MAXIMUM CAPACITY, actuate, and
! RPV level is noted to be +5 inches and rising rapidly. The.NSO attempts to control level i manually. Which of the following is TRUE regarding manual RPV level control?
4 The operator will have control over all feed reg valves, since RPV level >-20 inches.
- The operator will have control over the low flow valve, but not the main feed 3 regulating valves.
!
j Placing feed regulating valve controllers in manual will have no effect on feed
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, flow to the reactor.
i Depressing the 1 A(B) VLV RESET pushbuttons on the 901-5 panel will allow l I the operator to control level manuall l
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SENIOR REACTOR OPERATOR Page 38
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! OUESTION: 061 (1.00) .
!
' The plant was operating at 100% rated power when the "A" recirculation pump tripped. As ,
! .an immediate action of OCOA 0202-4, the operator is directed to monitor for oscillations
- indicating core instability. Which of the following is an indication that core wide oscillations i i are occurring?
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{ Excessive core plate Dp noise exceeding a value of 0.5 psi peak to peak.
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l Regular oscillations of reactor water level with a 2-3 second periodicity.
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' Noise signal on the LPRMs with a characteristic periodicity of 1.5 to 2.5 i second ' High values of APRM noise that occur with no regular frequency and are random in magnitude.
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! OUESTION: 062 (1.00)
j- Unit 2 is operating at 80% power when the gland seal and gland steam 55 foot loop seal is ,
j blown: condenser vacuum begins to decrease. Which of the following IMMEDIATE actions l
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should be taken?
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! Trip the main turbine at 5 in. Hg back pressure.
l lf - Continuously insert all CRAM rods to position 00.
I Reduce recirc flow and insert control rods as necessary.
i Send P.ad Protection to monitor the turbine building for high airborne activity.
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SENIOR REACTOR OPERATOR Page 39 i
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OUESTION: 063 (1.00) .
l' Unit 1 is in cold shutdown with RHR loop A in shutdown cooling when RHRSW is lost. The following plant conditions exist:
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Reactor water temperature is 190*F and rising at 2*F every 10 minute Reactor water level is +40 inche The drywell airlock is ope Recirc loop A is in operation, loop B is secure Assuming that shutdown cooling cannot be restored, which of the following actions should be considered within tne next 1 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />? Re-establish primary containment integrit ,
1 Increase CRD flow to promote shutdown coolin ; Place the B recirculation loop back in servic I Shut the MSIVs to prevent flooding the main steam line '
)
OUESTION: 064 (1.00) l i
Work in the area of the Unit 1 Reactor Building ventilation radiation monitors has resulted in the loss of signals from both the A and B radiation monitor channels. The crew should verify: l l Unit 1 Reactor Building ventilation has isolate ' Reactor Building ventilation continues to operate normall Only the inboard Reactor Building ventilation dampers have shu Only the outboard Reactor Building ventilation dampers have shu . . . _ . - . . ._ . .. .. . _ - . _ . _ _ - -
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SENIOR REACTOR OPERATOR Page 40
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- QUESTION
- 065 (1.00) .
j Which of the following is a responsibility of the Acting Station Director / Station Director? Require the general public to shelter or evacuate as neede l Distribute potassium iodide tablets to the general populace.
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- Provide for access control to the Control Room, TSC, and OSC, as appropriat l 2 l 4 Authorization of personnel exposure beyond 10CFR20 limits under emergency l
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condition j I
- i QUESTION
- 066 (1.00) l l
During a GSEP condition, stet? 3gency updates are required. If an ALERT were* declared at 0945 the initial state agency notification should be done by , and updates should j j be done by and .
.
l 2 ,1000,1200 ,1200,1300 f
i ,1145,1245 i
l ,1100,1200
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SENIOR REACTOR OPERATOR Page 41
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OUESTION: 067 (1.O0) .
Given the following initial conditions:
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Unit 1 is in cold shutdown during a maintenance outage.
- - Shutdown Cooling is in service on the "A" RHR Loo Bus 14-1 is de-energized for maintenanc All other buses are energized.
.
A transient results in a loss of Bus 13-1. Reactor water temperature is slowly rising and is i now 215*F. Your required actions are:
i Trip the "A" Recire. Pump.
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- Confirm LPCI has initiated.
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> Evacuate the Reactor Building.
- Confirm a Group 2 isolation has occurred.
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~ QUESTION: 068 (1.00)
i OGA 500-2, STEAM COOLING, directs RPV Blowdown when RPV water level reaches -184 -
i inches and no injection source is available. Why is this action taken?
. Blowdown results in significant void ' formation which reduces reactor power production, Blowdown increases steam flow up through the core improving heat transfer from the fue At lower pressures, less enthalpy is required to create steam, thus more steam is available for coolin RPV Blowdown dumps any radioactivity resulting from fuel failure into the torus, preventing uncontrolled release late . . _ _ . - - . . . _ . - . . ~ - . . . - . - _ _ - = . _ . . . . . . . _._ .- . - . - . - - . .
SENIOR REACTOR OPERATOR Page 42
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OUESTION: 069 (1.00) .
A LOCA has occurred. Reactor pressure is 150 psig. Drywell temperature is 260*F, and i Reactor building temperature at 1980F. The following reactor water levels are noted:
Yarway wide range -110 inches
, GEMAC lower 400 -140 inches
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Yarway narrow range -50 inches
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GEMAC upper 400 + 80 inches
- Which of the above level indicators cannot be used in these plant conditions?
i i GEMAC upper 400 i '
l GEMAC lower 400
! Yarway wide range l Yarway narrow range
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SENIOR REACTOR OPERATOR Page 43 '
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OUESTION: 070 (1.00) .
PRIMARY CONTAINMENT CONTROL, OGA 200, has an override that states: j IF: Drywell pressure drops BELOW 2.5 psig THEN: Stop Drywell sprays.
l Which of the following statements describes the reason for this step?
j .5 psig corresponds to 180*F, so there is no need to continue drywell sprays.
i
] It makes one more RHR loop available as soon as possible for injection into the l
l reactor pressure vesse * This action ensures that the drywell structure doesn't endure excessive thermal
- stresses due to rapid cooldow .I
! It prevents drawing a negative pressure in the containment, which would open
- the vacuum breakers and draw air into the containment.
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SENIOR REACTOR OPERATOR Page 44
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- QUESTION
- 071 (1.00)
l Unit 1 has experienced a LOCA in the drywell as the "A" FW line failed at the vessel nozzle.
!, HPCI is being used to maintain RPV level. With HPCI flow at 2500 gpm and reactor pressure i at 250 psig and dropping slowly, level is constant at -80 inches CRD, CS and RHR are
operable. A HPCI Room high area temperature alarm is received and HPCI Room temperature l Is 152*F and rising. An EA confirms that there is a steam leak on the HPCI governor valvo.
l Which action below should be taken?
!
, Isolate HPC Conduct an RPV Blowdown.
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i Anticipate RPV Blowdown and rapidly depressurize.
l Evacuate the HPCI room and install jumpers to defeat the automatic isolatio ;
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{ OUESTION: 072 (1.00)
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- An entry into the Unit One LP Heater Bay at power is scheduled to repair a heater LC Hydrogen addition is on at normal flow rates. Admission to the area may be granted by:
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- The Operations Shift Supervisor, i
l The Radiation Protection Shift Supervisor.
4 Operations and Radiation Protection Supervision.
! The Radiation Protection Technician and Work Supervisor in charge of the job.
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SENIOR REACTOR OPERATOR Page 45
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QUESTION: 073 (1.00) .
l The plant is operating at 100% power and has been for the last two weeks. The chemist calls the control room and informs the operator:
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- Cl is 0.6 ppb
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COND is 0.066 F/T
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l-131 is 6.5 uCi/ SO4 is 1.2 ppb
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You should:
- after 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, reactor water must be sampled every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> until within the limit *
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< Isolate all main steam lines within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or be in at least HOT SHUTDOWN within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
< immediately begin sampling reactor water every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. After 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />,if the
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condition is not corrected, close the MSIVs.
I enter a 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> LCO has been entered, if the problem can not be fixed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> the plant must be in HOT SHUTDOWN with the main steam lines t- closed in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
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i SENIOR REACTOR OPERATOR Page 46 QUESTION: 074 (1.00) .
OGA 200, " Primary Containment Control," primary containment pressure control path, directs the primary containment to be vented. Why does the procedure direct the operator to vent i via the torus as the preferred method vice via the drywell? Venting the primary containment ~
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via the torus will: Reduce the levels of radioactivity released as it passes through the water in the l toru I Allow better control of the release rate due to the sizing of the path's piping and valve : Allow a more rapid reduction in primary conteinment pressure than venting from
the drywell Minimize chugging due to loss of non condensibles from the drywell I atmospher I
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j SENIOR REACTOR OPERATOR Page 47 l
l OUESTION: 075 (1.bO)
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Complete the following statement: l
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- Three fission product barriers would be considered lost if
)
i A casualty occurs which leads to a TAF blowdown. RPV flooding is subsequently entered and level is restored on the narrow range level
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instruments.
- A fuel failure results in a sustained 120 R/ hour on tha DW radiation monitors
, and a Group i isolation occurs due to sustained Hi-HI radiation on the MSL
- radiation monitors.
! A large LOCA results in DW pressure peaking at 12.5# and vessel level below j TAF All isolations and auto starts occur as designed. Vessel level drops to 1
- -134" for several minutes and is subsequently restored.
$ A steam leak occurs in the MSIV Room resulting in room temperatures of
- 320*F. The Group i isolation fails; however, ECCS Systems operate correctly
[ and restore vessel level prior to reaching TAF.
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SENIOR REACTOR OPERATOR Page 48
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l OUESTION: 076 (1.00) . l Given the following plant conditions:
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Rx pressure is 900 psig
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Rx scram and all rods are in to 00
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MSIVs are closed
- -
RPV level -142" and lowering I! -
No high pressure injection systems are ava;lable
'
The Unit Supervisor directs RPV blowdown per QGA 500-1. The operator reports only four ADS valves opened. Which of the following actions is required?
l* * Enter QGA 500-2 to steam cool the core since four ADS valves will not depressurize the RP b.- Allow the vessel to depressurize through the open ADS valve Re-open MSIVs to establish condenser as a heat sink and open the turbine bypass valve Enter OGA 500-4 to flood the RPV since the condenser is unavailabl l
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i SENIOR REACTOR OPERATOR Page 49
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i OUESTION: 077 (1.00) .
Following a major transient, the following conditions exist:
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Torus Prer.sure 15 psig
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Torus Water Level 14 ft.
- -
Torus Water Temperature 145'F i
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DW Pressure 4 psig i
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DW Temperature 275'F and INCREASING j -
Reactor Pressure 1000 psig and STABLE
- -
RPV Water Level +9 inches and INCREASING I-
] What action should be taken to control containment parameters?
{ Conduct an RPV Blowdown.
!-
l Initiate Drywell spray l Vent the Torus using SBG I
- Reduce RPV pressure to at least below 900 psig, t
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! QUESTION: 078 (1.00)
! Given the following information:
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Torus pressure is O psig
-
Torus level is 10 feet i - Torus temperature is 150*F.
- -
RHR Pump "A" is the only available RHR pump.
!
What is the approximate MAXIMUM flow that can be achieved with NO LPCI pump cavitation?
i l gpm gpm
l gpm i
! gpm J
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QUESTION: 079 (1.00) .
Control Rod J-1 was selected and was given a notch-out withdraw signal to move it from 08
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- to 10.
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Alarm 901-5 A-3, Control Rod Drift, annunciate Alarm 901-5 B-3, RWM Block, annuciate Alarm 9015 C-3, Control Rod Block, annunciate The NSO notes that position indication is lost for rod J-1. According to Tech Specs, what actions if any, are required for control rod J-17 Substitute position 10 in the rod worth minimizer for control rod J- ' Move control rod J-1 to a known position, or fully insert rod J-1 and disarm the directional control valve Commence insertion of all operable control rods immediately and have all control rods fully inserted within 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Immediately fully insert rod J-1. Electrically disarm it's directional control valves if the position indication cannot be repaired within 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> .- . . _ . . _ - - -- . .. - . - - - - _ _ - . . - - -__ .- -.
i SENIOR REACTOR OPERATOR Page 51
i l QUESTION: 080 (1.00) l
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- The plant is operating in Mode Three. The IMs report that LT 57A on Unit One has a trip setpoint of +4". With regards to the containment isolation function, what actions are required by Tech. Specs.? l
, Establish Secondary Containment with SBGT system operating within one hour.
- Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN i within the next 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Close the affected system isolation valves within one hour and declare the j affected system inoperable.
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- Declare LT 57A INOP, and place a trip on the channel of the affected Group Isolation circuits within one hou QUESTION: 081 (1.00)
Unit Two has just raised load from 50% to 100% power. The next Core Monitoring Code indicates that MFLPD at location 27-32-4 is 1.002. Which of the action (s) is(are) required by Tech Specs? Reduce thermal power to less than 25% of rated within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Be in HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with Tech Spec Initiate corrective action within 15 minutes and have MFLPD less than 1.0000 within 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and COLD SHUTDOWN within the following 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> . _ . _ _ _ . . _ __...__ __ ._ . . _
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i SENIOR REACTOR OPERATOR Page 52 r
QUESTION: 082 (1.00) .
Unit Two is operating at 75% power when 2B Main Steam Line Radiation Monitor fails
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downscale. Which one of the following will satisfy the LCO?
, Be in at least STARTUP with the associated isolation valves closed within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, Place the B MSL Radiation Monitor Trip System in a tripped condition within one hou Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and COLD SHUTDOWN within
,
the next 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> *
j Close the affected system isolation valves within one hour and declare the
affected system inoperabl !
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SENIOR REACTOR OPERATOR Page 53 j
- OUESTION
- 083 (1.00)
i What does the red Dryer Failure Indicating Light on the in service Dryer for the Drywell j Pneumatic system being lit mean and what is a possible co7 sequence if it is not corrected?
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, The temperature at the outlet of the Dryer is above the alarm setpoint. If not
, corrected, the efficiency of the downstream filter may be affecte The moisture content at the outlet of the Dryer is above the alarm setpoint. If
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not corrected, the efficiency of the downstream filter may be affected.
, The temperature at tha outlet of the Dryer is above the alarm setpoint. If not
{ corrected, the Compressor will trip and pressure will be maintained by either
,
Nitrogen Makeup or Instrument Air.
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- The moisture content at the outlet of the Dryer is above the alarm setpoint. If
- not corrected, the Compressor will trip and pressure will be maintained by either the Nitrogen Makeup or Instrument Air.
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SENIOR REACTOR OPERATOR Page 54
QUESTION: 084 (1.00) .
.; The Unit 1 125VDC battery has been determined INOPERABLE during maintenance and j requires replacement. Both Units are operating at full power. How long can the Alternate
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125VDC Batteries, with a full capacity charger, be used in place of the Normal Unit 1 125VDC Batteries, before the LCO requiring a shutdov/n is entered for either Unit?
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i hours i days
j , days
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4 OUESTION: 085 (1.00)
!
d4 A reactor startup is in progress on Unit One with reactor power on Range 8 of the IRMs. The
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125 VDC Logic power is lost to the Unit One "B" Core Spray system. What actions are now
!
required?
!
i Maintain the reactor in Mode 2 until logic power is restored.
! Enter a 30 day LCO because the B Core Spray pump is inoperable.
l I Enter a 7 day LCO for Core Spray subsystem B and the B Diesel Generator.
' Be in HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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SENIOR REACTOR OPERATOR Page 55
i QUESTION: 086 (1.00) .
, OGA 500-3, "Drywell Flooding" directs the RPV vented when the Primary Coi;tainment water level reaches 28 ft. This action ensures which ONE of the following?
L .
i a.- The core will be submerged when drywell level reaches e'. ft.
i
- . The drywell will be vented via the reactor vessel vent path.
}- RPV level instrument reference legs will be free of non-condensables.
! Vessel level will have reached the main steam lines when drywell flooding is j complete.
i QUESTION: 087 (1.00)
i While operating a 100% steady state power, unit 1 experiences a loss of TR-12. A
-
recommendation is made to start and load EDG 1 and 1/2 to their respective buses in _
! anticipation of a loss of off-site power. What effect would these actions have if a loss of off-
,
site power were to occur on unit 17 1-1 EDGs may overload due to failure of loads to shed when power is lost.
!
j Reliability of power will be enhanced because the buses are already energized from the EDGs.
- ;
i Power to loads on the buses will be lost momentarily as the loads are shed and I sequenced back on.
) Power to the buses will be lost momentarily as the EDG output breakers automatically trip open and reclose.
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SENIOR REACTOR OPERATOR Page 56 QUESTION: 088 (1.00)
The plant is in MODE 5 with a core offload in progress. All IRM indications are lost. Based on these conditions, which one of the following describes the actions that should be taken by the NSO? Halt CORE ALTERATIONS until IRM indication is restore Halt CORE ALTERATIONS uniti the Shorting Links are installe No action is needed; IRM indication is not required in MODE Dispatch an operator to the Essential Services Bus to investigate the loss of power to the IRM .
QUESTION: 089 (1.00)
A reactor scram has occurred due to a main turbine trip. The ADS valves are cycling to maintain reactor pressure. Which one of the following statements describes the actions that should be taken for this condition and the basis for those actions? Open ADS valves to reduce pressure to 940 psig and minimize heat load on the condense Open ADS valves manually to reduce pressure to 940 psig and minimize dynamic stresses on the tailpipe No action is necessary, the relief valves are left in automatic to allow time for the bypass valves to take control of pressur The NSO should operate Relief valves manually only if needed to prevent safety valve actuation, to minimize heat input to the containmen l
!
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i SENIOR REACTOR OPERATOR Page 57
i
QUESTION: 090 (1.00) .
!
The plant is operating at 100% power. All components and systems are aligned for normal
!
'
power operation. The "A" reactor food pump auto trips due to an oil leak. RPV level begins to trend down. No other automatic actions have occurred. Which statement below describes
- the action (s) required to maintain RPV level within normal operating bands?
i j Drive cram arrays to lower power and maintain level.
! Manually start the standby reactor feed pump IAW OOP 3200-3.
. Trip one recirc pump and run the other to minimum to maintain level.
l ' Take manual control of FWRVs and open the FWRVs as needed to maintain j level.
i OUESTION: 091 (1.00)
A valid reactor scram signal is received and the NSO takes immediate actions IAW OCGP 2-3 and reports no control rod movement. The NSO notes that all " blue" scram lights are li Which one of the following statements best describes the actions that must be taken in order to attempt control rod insertion? !
l i Bypass the SDV high level scram, reset RPS, reset ARI from 90X- ' Reset RPS, disarm the ARI initiation switches, initiate a manual scram, Bypass reactor scram signals, reset RPS, manually de-energize the ARI solenoid Reset RPS after a 10 second time delay, manually override ARI to energize the ,
solenoid _ _ _ _ . . . .. _
. - . _ . _ . . - - . - - . . . .- .. . . ..- - . . . ..- -. .-.- -. --
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SENIOR REACTOR OPERATOR Page 58
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QUESTION: 092 (1.00) .
J Both units are operating at 100% power. Alarm 912-1 F3, STACK GAS HI RAD, is received
'
and confirmed to be valid. Which statement below describes the immediate actions that
'
should be taken by the crew for the described conditions?
1 Hold reactor power constant on both units.
l
! Significantly reduce power on both units.
s
- c; Scram the unit that is believed to be the cause of the increas i l
l l Significantly reduce power on the unit believed to be the cause of the increas '
i I
- . <
$
i OUESTION: 093 (1.00)
.,
Both units were aligned for normal full power operation. A grid disturbance caused a loss of j off-site power to unit 1. This was followed shortly by a loss of off-site power to unit 2. All l AC systems have responded as designed. Assume no operator ections have been taken to l
- . this point. Which statement below describes expected status of the EDG 1/2 if a LOCA signal l j were now to be received on unit 2?
EDG 1/2:
i l output breaker will OPEN, and then close to bus 23-1.
i
- will remain running and its output breaker will remain closed to bus 13- c, will remain running and its output breaker will remain closed to bus 23-1.
7 l
! l
- will remain running, its output breaker will trip, and the diesel will run unloaded.
! ;
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SENIOR REACTOR OPERATOR Page 59 l
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! .
i OLESTION: 094 (1.00) .
< An ATWS has occured coupled with a loss of both CRD pumps. OGA 101, RPV Control i (ATWS), has been entered. Emergency Depressurization may be required in approximately 15
- to 20 minutes due to low reactor water level. Which of the following operator actions will be
, affected by the emergency depressurization?
!
, The ability to scram the reactor.
i l The ability to bypass and reset AR The ability to restart the CRD pumps.
i
, The ability to inject SBLC at the proper flowrat QUESTION: 095 (1.00)
A primary system leak has resulted in radiation levels exceeding Maximem Safe Operating Levels in more than one area. Which of the following will be accomplished by' emergency depressurizing the RPV? Depressurization reduces the driving head for flow from the lea The depressurization will rapidly void the core, reducing power faster than a scram, Cooldown of the coolant willincrease saturability, keeping radioactive gases in solutio Emergency depressurization rejects heat to the drywell vice outside the primary containmen .
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! SENIOR REACTOR OPERATOR Page 60 i
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j QUESTION: 096 (1.00) .
!
4 Over the last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the Unit 1250 VDC battery voltage has been degrading slowly and has now reached 258 VDC on float charge. Which of the following actions is required?
, Commence a Unit I reactor shutdown immediatel ;
'
, Restore the battery to operable status in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
i
!
I Place HPCI INOP and test RCIC and ADS for operabilit !
' Perform a capacity test of the Unit 2 250 VDC battery within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
-
i i
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QUESTION: 097 (1.00)
4 A severe, uncontrolled fire has developed in the central turbine building mezzanine level. What !
! immediate action must be taken to control reactor pressure before leaving the control room? ' Place RCIC in a full flow test lineu Open one ADS valve and close the MSIV Inhibit ADS and place the ADS valve control switches in OF Reduce pressure set to 850 psig and verify bypass valves ope :
-. . .,. -
. . . _ . - . ._- _ _ _ _ _ _ . . . . _ _ . . - . . _ . . _ . _ _ _ - . _ . _ - . _ . ._ _ . . . __
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SENIOR REACTOR OPERATOR Page 61 J
QUESTION: 098 (1.00) .
, Given the following plant conditions, determine the maximum Unit 1 electrical output:
-
Date is October 7.
. - . Unit 2 is operating at 600 MWe, holding load.'
i
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Upstream River temperature obtained from Chemistry is 72*F.
l
- River flow is 12,000 cf MWe.
- MWe.
'
.. MWe.
.
- MWe, i
.
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4 OUESTION: 099 (1.00)
! The NSO is conducting CRD Weekly Exercising.
i
,
-
Annunciator 902-5 A-2, " ROD OVTRVL"_ alarms for rod A- l
- Which ONE of the following actions would NOT satisfy the requirements for the plant l conditions? SCRAM rod A-8 and give it an insert signal every shif ,
l Place the Unit in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
. Declare rod A-8 inoperable, fully insert it and electrically disarm the directional j control valve ' Attempt to recouple rod A-8 by fully inserting it into the core and then demonstrate that it has been recoupled by demonstrating that the rod will not go to the overtravel positio I j;
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t j SENIOR REACTOR OPERATOR Page 62
!
,
OUESTION: 100 (1.00) ,
l OGA 100 has been and Alternate injection Systems are needed to restore and maintain RPV l water level greater than +8". RPV pressure is 700 psig and steady. Which ONE of the l Alternate injection paths listed below will provide injection flow for the given plant conditions? Condensate cross tie Fire System through RHR.
)
l HPCI Cooling Water Pump.
,
i SSMP from the Fire System.
! .
,
i
9
!
!
,
t
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(* * * * * * * * * * END OF EXAMINATION * * * * * * * * * *)
.-
- SENIOR REACTOR OPERATOR Page 1 ANSWER
- 007 ANSWER: 001 b.
- REFERENCE
REFERENCE: LER 1-93-017 QCOA 3500-1 page 2 QCAP 1100-5 page 5 295014A102 ..(KA's) QCAP 1100-4 294001A103 ..(KA's)
ANSWER: 002 (1.00) ANSWER: 008 REFERENCE: OCAP 211-2 page 4 REFERENCE:
ANSWER: 003 4 ANSWER: 009 REFERENCE: ILT 0500 page 44 REFERENCE:
212000K412 ..(KA's) QARP 000-2 page 1 QARP LP 295016G011 ..(KA's)
ANSWER: 004 d.
, REFERENCE: ANSWER: 010 QCAP 230-5 page 10 K101 ..(KA's) REFERENCE:
LIC-0203 page 46 218000K501 .. (KA's)
ANSWER: 005 ANSWER: 011 REEFERENCE: QOA 4700-6 page 3 REFERENCE:
295019K203 ..(KA's) LIC-0203 page 38 218000K102 ..(KA's)
ANSWER: 006 ANSWER: 012 REFERENCE: OCOS 0700-07 page 5 REFERENCE:
215002A205 ..(KA's) LIC-1300 page 20 217000A201 ..(KA's)
!
!
,. _ _ . - . . _ _ . . - . _ . - . _ . . - _ __ _ .-.___ _ .. _. _ _.._.._ - _ ..... . _ __- _ . . _ . - .
!
!,
}- . SENIOR REACTOR OPERATOR Page 2 ANSWER: 013 & F ##' O ANSWER: 019
. U a.
'
REFERENCE: REFERENCE:
- LI 300 page 70 QCOP 1200-11 page 2 j OP 2300-01, page 3 and 4 LIC-12OO
OCAN 901(2) 3 A 1 204000G010 ..(KA's)
{. 206000K419 ..(KA's)
I ANSWER: 014 ANSWER: 020 b.
i REFERENCE:
l ILT-HP-CH2 ;
i REEFERENCE: TH.HP-Chapter 2 page 6 .l I LIC-5652 page 70 - QCAP 630-06, page 17
) 241000A112 ..(KA's) 294001K103 ..(KA's)
i I j ANSWER: 015 ANSWER: 021 I s a.
I REFERENCE: REFERENCE:
. LO/NLO-6600 page 44 ILT-HP-CH2 j l- 264000K401 ..(KA's) TH-HP-Chapter 2 page 20 '
{ QCAP 0620-01, page 8 ;
{^
294001K104 ..(KA's) =
ANSWER: 016 l l REFERENCE: ANSWER: 022 l-l LIC-7500 page 50 '
i TSUP 3. REFERENCE:
. 261000 GOO 5 ..(KA's) LIC-5450 page 44
QCAN 901(2)-3 C-2
. 271000K408 ..(KA's)
1 ' ANSWER: 017 l'
. REFERENCE: ANSWER: 023
! OOP 6400-2 page 1 lK107 ..(KA's) REFERENCE:
LIC-0705 page 32 215002K604 ..(KA's)
ANSWER: 018 REFERENCE: ANSWER: 024 LO/NLO-1601 QCOP 2300-06 page 12 REFERENCE:
295026K101 ..(KA's) LIC-0202 page 94 202OO2K102 ..(KA's)
-
_, _ _ _ _ _ . _ . _ _ . . . _ _ . _ . _ _ _ . _ - _ . _ . . . _ . . _ _ _ _ . _ . _ . . _ _ _ . _ _ _
l
>
SENIOR REACTOR OPERATOR Page 3
[ ANSWER: 025 ANSWER: 031 REFERENCE:
- LIC-02021 page 110 REEFERENCE:
4 202OO1K116 ..(KA's) QOA 7000-01 page 2,3,6 l 1600 2 Containment Auxiliaries j -223OO2K608 ..(KA's)
] ANSWER: 026 4 a.
"
REFERENCE: ANSWER: 032 l LIC-7500 page 22 d.
- '261000A201 ..(KA's) REFERENCE
j LIC-1000 page 88 q 219000K202 ..(KA's)
,
ANSWER: 027 3 ANSWER: 033 REEFERENCE: a.
i QAP 0300-12 page 2,3 REFERENCE:
l Temp Alts OJT/OJE LIC-0280 294001K102 ..(KA's) QCOA 0280-01 page 1 214000K303 ..(KA's)
!-
!- ANSWER: 028
, ANSWER: 034
- REFERENCE: b.
i QCAP O211-02 page 11 REFERENCE:
l CREW-OJT QCOA 1900-1 page 2
. 294001A106 ..(KA's) QCFHP 0110-05 page 1 j 295023G010 ..(KA's)
!
- - ANSWER: 029 l
} . ANSWER: 035
-
REFERENCE: b.
- ' P&lD M38 AF REFERENCE:
LO/NLO-1900 LIC-1000 a 294001A107 ..(KA's) QCOA 1000-02 page 5
- OCOP 1000-05 page 7 205000K303 ..(KA's)
ANSWER: 030- REFERENCE: ANSWER: 036 Plant Process Computer QOP 9900-102 page 4 REFERENCE:
294001A115 ..(KA's) LIC-0800 QCOA 0202-01 page 1 295001A205 ..(KA's)
i SENIOR REACTOR OPERATOR Page 4
.
ANSWER: 043 ANSWER: 037 , REFERENCE: i REFERENCE: OCOP O201-11 page 3,4,5
.QCOA 0203 01 page 2 216000K506 ..(KA's)
'
LIC 0203 295013A201 ..(KA's) l ANSWER: 044 c.
ANSWER: 038 REFERENCE: LIC-0600 page 54 I REFERENCE: 259002K413 ..(KA's)
LIC-0500 page 42 295025K201 ..(KA's)
ANSWER: 045 c.
ANSWER: 039 REFERENCE: LIC-2400 page 38 223001K404 ..(KA's)
REEFERENCE:
QGA 101 LP page 6 295037G007 ..(KA's) ANSWER: 046 l REFERENCE:
ANSWER: 040 LIC-3700 page 22 G006 ..(KA's)
REFERENCE:
LIC-0705 page 2 215002K401 ..(KA's) ANSWER: 047 REFERENCE:
ANSWER: 041 QCAN 912-1-D 2 LIC-3800 REFERENCE: 295018G005 ..(KA's)
QCAP 230-4 page 12 294001K102 ..(KA's)
ANSWER: 048 ANSWER: 042 Nd REFEFERENCE:
4 RO QCOP 5400-7 Att. A.
REFERENCE: 294001A108 ..(KA's)
QC 0400-13 page 2 4001K107 ..(KA's)
_
. . - . . . -. -. - - - . . _ - . - . - - .~ . _ . _ - _-
)
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SENIOR REACTOR OPERATOR Page 5
- ANSWER: 049 (1.00) ANSWER: 055 (1.00) c.
REFERENCE: REFERENCE:
,
-
QGA DETAILS LP page 6 ILT 201-1 Figure 21
,
'295028K203
.
..(KA's) QCGP 1-1 page 2 0201-1 page 38
, 290002G005 ..(KA's)
.
ANSWER: 050 REFERENCE: ANSWER: 056 QGA 101 LP page 34,36 c.
- 295037K104 ..(KA's) REFERENCE
- LIC 0700-1 ,
"
215004K402 ..(KA's) l l ANSWER: '051 I )
i - REFERENCE: ANSWER: 057 LIC/FH-0805 a,
'
QCFHP 110-2 page 1 REFERENCE: I l 295023K103 ..(KA's) LIC-1100 page 8,18 211000A202 ..(KA's) ! ]
'
ANSWER: 052 .i s ANSWER: .058 i REFERENCE: c.
-
lLT 201-1 page 14 REFERENCE:
i 290002K402 ..(KA's) LIC 0600
259002K603 ..(KA's)
ANSWER: 053
"
, ANSWER: 059 l REFERENCE: LIC-0703 page 62 REFERENCE:
215005A308 ..(KA's) LIC-1000 page 24 i 203OOOA203 ..(KA's)
ANSWER: 054 ANSWER: 060 REFERENCE: LIC-1300 REFERENCE:
. 217000A403 ..(KA's) QCAN 901(2)-5 G-8 page 1
. - OCOA-0201-9 page 3
- 295009A102 ..(KA's)
e d
.
e
, - -
. _ - - . . . . . - . .
SENIOR REACTOR OPERATOR Page 6 ANSWER: 061 ANSWER: 067 (1.00) ;
REFERENCE: REFERENCE: i QCOA-0202-4 page 6,7 QCOA 1000-02 page 4
.295001G010 ..(KA's) 295021G011 ..(KA's) )
l i
ANSWER: 062 ANSWER: 068 REFERENCE: REFERENCE:
QOA 3300-02 QGA 500-2 LP page 6 295002G010 ..(KA's) 295031K305 ..(KA's)
ANSWER: 063 ANSWER: 069 REFERENCE: REFERENCE:
QCOA 1000-2 QGA DETAILS LP page 4,6,8 295021A201 ..(KA's) 295028K203 ..(KA's)
i l
ANSWER: 064 ANSWER: 070
. REFERENCE: REFERENCE:
LIC 1701 page 84 QGA 200 LP page 12 272OOOA309 . .(KA's) 295024G012 ..(KA's) i ANSWER: 065' ANSWER: 071
. a .-
REFERENCE: REFERENCE:
QEP 105-1 QGA 300 LP page 8,10 GSEP Manual page 4-105 295032G012 ..(KA's)
294001A116 ..(KA's)
ANSWER: 072 ANSWER: 066 REFERENCE:
REFERENCE: QCAP 0620-02 page 2 QEP 0300-01 page 2 294001K104 ..(KA's)
294001A116' ..(KA's)
ANSWER: 073 (1.00) REFERENCE:
TSUP 3.6.l. page 13 204000G005 ..(KA's)
.
SENIOR REACTOR OPERATOR Page 7 ANSWER: 074 ANSWER: 080 REFERENCE: REFERENCE:
QCOP 1600-13 TSUP Table 3.2.A-1 QGA 200 page 50 216000G005 ..(KA's)
295010K301 ..(KA's)
ANSWER: 081 ANSWER: 075 REFERENCE:
REFERENCE: TSUP 3.1 QEP 02OO-T1 page 4 LIC-0800 295038K102 ..(KA's) 290002K501 ..(KA's)
ANSWER: 076 ANSWER: 082 REFERENCE: REFERENCE:
QGA 500-1 LP page 12 TSUP Table 3.1.A-1 295031G012 ..(K A's) TSUP Table 3. OO2 GOO 5 ..(KA's)
ANSWER: 077 ANSWER: 083 REFERENCE: QGA 200 LP page 22 REFERENCE:
295024G012 ..(KA's) LIC-4700 P&lD M-24 Sheet 2 295019 GOO 5 ..(KA's)
ANSWER: 078 REFERENCE: ANSWER: 084 QGA DETAILS LP figure ll-D K102 ..(KA's) REFERENCE: l TSUP 3.9.C page 13 LO/NLO-6900 ANSWER: 079 263000 GOO 5 ..(KA's) REFERENCE:
LIC-0280 ANSWER: 085 TSUP 3.3.1. page 14 GOO 5 ..(KA's) REFERENCE:
TSUP 3/4.5-2 TSUP 3/4.9-2,3 4E 1430 sheet 2 209001G005 ..(KA's)
- -
- ____ __ ._ _ __ - ..
. _ . . ._ . .. . ._ _ . . . __ _ _ _ __ -
l SENIOR REACTOR OPERATOR Page 8 ANSWER: 086 ANSWER: 092
- !
REFERENCE: REFERENCE:
. QGA 500-3 LP page 10 QCOA 1700-01
,
295029G012 ..(KA's) 295017G012 ..(KA's)
ANSWER: 087 ANSWER: 093 3 REFERENCE: REFERENCE: 4 QOA 6100-1 ILT-6600 page 94,96 l
,
295003A101 ..(KA's) - '295003K202 ..(K A's)
'
i ANSWER: 088 ANSWER: 094 a.
. REFERENCE: REFERENCE:
QCOA 0700-03 QCOA 300-1 LIC-0702 ILT 0300-2
ANSWER: 089 ANSWER: 095 I
,
REFERENCE: REFERENCE:
QGA-100 LP page 46,48 QGA 300 l 295007A104 ..(KA's) 295033K301 ..(KA's)
i ANSWER: 090 ANSWER: 096 b.
i REFERENCE: REFERENCE:
"
LIC-3200 TSUP 3/4.9-12 QCOA 0201-9, page 3 295004G003 ..(KA's)
QOA 3200-1, page 1 295009G012 ..(KA's)
ANSWER: 097 c.
ANSWER: 091 REFERENCE: QARP 700-01 page 1 REFERENCE: 295016A108 ..(KA's)
QCOP 300-28 295015A202 .. (KA's)
i i
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.. . - -- - . ..- . . ~ . - . . .-. . . . - . . . .
SENIOR REACTOR' OPERATOR Page 9
- ANSWER: '098'(1.00) REFERENCE:
QAP 3OO-T24 QAP 300-32 page 1,2 294001A108 ..(KA's) i i
l ANSWER: 099 ; l J
REFERENCE:
TSUP 3.3.H page 3/4.3-12 1 201003 GOO 5 ..(KA's) i i
l
ANSWER: 100 I REFERENCE: l QCOP 4100-11, page 1 QCOP 2300-10, page 1 QCOS 2900-01, page 5 QCOP 3300-12, page 3 l 295031G012 ..(KA's)
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(* ' * " " * * END OF EXAMINATION * ' " * * " *)
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SENIOR REACTOR OPERATOR Page 1 3 ANSWER KEY a ;
,
MULTIPLE CHOICE O23 b
,
001 b 024 c 002 c 025 d t !
. 003 b 026 a l t
,' 004 d 027 a
'
005 c 028 c
!
006 a 029 b
- -
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- . 007 b 030 d
,
008 c 031 a
,
009 b 032 d
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010 d 033 a
- 011 a 034 b i
!
012 d 035 b
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013 deleted 036 b 014 b 037 d
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i
015 b 038 b I
- 016 a' 039 c
.
017 c 040 a i
- - 018 c 041 c 019 M kkW O20 b 043 a I O21 ' a ~ 044 c !
022 a 045 e i
. . . . . . . . - . . _ ~ . . . . - . . . _ _ - . . . - . . . . - . . . _ . - . . ~ - . . . . . . . . _ . - ._. ....- . ._
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. SENIOR REACTOR OPERATOR Page 2 ANSWER.. KEY t
MULTIPLE CHOICE 068 b i 046 a 069 d l 047 b I 070 d i l
048 c 071 a
.
.
049 ' b - 072 c- i i
050 a 073 b
-
,
051 d '074 a 1
- 052 J c - 075 b I
l: 053 a : 076 c-054 c 077 a i 055.. c 078 b
!
056 c 079 b 057 a' 080 d 058 c 081 c
059 a- 082 b !l 060 c 083 a
- 061 c- 084 d 062. c 085 c 063 a. . a ~ 087 a
!'
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[ - 065. d 088 a
,
066 b 089 b I
- 067 a 090 b
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SENIOR REACTOR OPERATOR Page 3 ANSWER KEY MULTIPLE CHOICE i
091 e i
.092 a l 093 a i
094 a l
095 a j 096 b 097 c 098 b 099 a
,
.
100 d i
l (" * " * " * * END OF EXAMINATION * * * ' * * * * *)
_ _