IR 05000334/1997010
| ML20198L774 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 01/09/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20198L756 | List: |
| References | |
| 50-334-97-10OL, NUDOCS 9801160116 | |
| Download: ML20198L774 (109) | |
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U. S. NUCLEAR REGULATORY COMMISSION REGION 1
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Docket No.:
50 334 l
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i Report No.1 97 10 I
License No.t -
DPR 66 l
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. Licensee:
Duquesne Light Company
- Facility:-
Beaver Valley Unit 1 Nuclear Power Plant a
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Location:-
Shippingport, Pennsylvanla l
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Dates:.
December 15 18,1997 l
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t Chief Examiner:
P. Bissett, Senior Operations Engineer / Examiner l
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Examiners: ~
T. Kenny, Senior Operations Engineer / Examiner J. Caruso, Operations Engineer / Examiner
- Approved By:
Glenn W. Meyer, Chief, Operator Licensing and Human Performance Branch i
Division of Reactor Safety
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EXECUTIVE SUMMARY
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Beaver Valley Unit 1 Nuclear Power Plant Inspection Report No. 50 334/97 10 Onorations Three Unit 1 senior reactor operator instant (SROl) candidates were administered initial
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licensing examinations. All three candidates passed all portions of the license examination.
Generic strsngths were noted during the Unit 1 examination in the area of crew communi:stions, control board awareness, and crew briefings during the simulator portion of the operating euminatica. The NRC examiners observed communications to be direct,
succinct, and that all crew members were kept wellinformed. Crew briefings were routinely held during those instances in which time permitted. The written exam was developed at the appropriate SRO knowledge level, as were the job performance measures and follow up questions. Several JPMs, in lieu of questions, were appropriately developed to test the knowledge level of the applicante in the administrative area of the exam.
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O flenort DetaRt 1. Ooerations
Operator Training and Qualifications 05.1 S.ggigr Reactor Operator Initial Examinationjl a.
Scope The examinations were prcpared by Beaver Valley Power Station (BVPS) personnel in accordance with the guidelines in interim Revision 8, of NUREG 1021," Examiner Standards," and Revision 5 of NUREG BR 0122, * Examiners' Handbook for Developing Operator Licensing Written Examinations," and Revision 1 of NUREG-1122, * Knowledge and Abilities Catalog for Nuclear Power Plant Operators:
Pressurized Water Reactors." The examiners administered initial operating licensing examinations to three Unit 1 senior reactor operator instant (SRol) candidates. The written examinations were administered by the facility's training organization, b.
Observations and Findinas The results of SRO examinations for Unit i are summarized below:
SRO Pass / Fall Written 3/0 Operating 3/0 Overall 3/0 The written examinations, job performance measures (JPMs) and simulator scenarios were developed by Beaver Valley Power Station (BVPS) representatives in accordance with NUREG 1021. The exam development team was comprised of DVPS training and operation's representatives. Allindividuals signed a security agreement once the development of the examination commenced. BVPS personnel also validated the exam prior to their submitting it to the NRC. The NRC subsequently reviewed and validated, along with DVPS personnel, all portions of the proposed examinations. Also, various changes and/or additions to the proposed examinations that were requested by the NRC following their review, were subsequently validated and approved. BVPS personnel subsequently incorporated the NRC's comments and finalized the examinations.
The written examination was administered on December 15,1997, and consisted of 100 multiple choice questions. There were no comments by either the NRC or the utility concerning the validity of questions on the written examination, however, the answer key for one question was in error due to a typographical mistake.
Based on the grading of the written examination, one question, #49, was missed by all three candidates, indicating a weakness in the general understanding of the subject area denling with the effects and indications of a dropped rod at 100%
power,
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The operating examinations were conducted from December 16 17,1997, and consisted of three simulator scenarios and ten JPMs. All JPMs were followed up with two system related questions. All candidates were also examined using JPMs and/or questions to evaluate the administrative requirement portion of the examination.
During the exam preparation week, the examiners noted that the f acility routinely developed JPMs involving f ailed instruments, transmitters, pumps, etc. with each f ailure having already occurred prior to the candidato starting performance of the JPM. Thus, the candidate would start the JPM with the simulator frozen and the
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f ailure already having occurred. The NRC examiners stated that JPMs involving instrument or component f ailures would be greatly enhanced if the individual were to assume the watchstander's position with all systems operating normally and have the failure subsequently occur. The f acility stated that failures of this type were
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routinely performed and evaluated during the simulator scenario portion of the exam. The NRC stated that the JPM setting would allow for a greater number of individuals to be evaluated in one particular area of performance and also would be more realistic for these types of failures. The facility agreed with this approach and stated that JPMs of this type would be developed for future evaluations.
Simulator performance by the Unit 1 candidates was, for the most part, very good.
Communications was also good, including the use of repeatbacks. The examiners noted that crew briefings were routinely performed by the SROs Control board awareness by the operators was evident throughout each of the three scenarios.
For the administrative segment of the operating portion of the examination, administrative job performance measures (JPMs) were used in a number of instances in lieu of administrative topic questions. The examiners determined that candidate performance was good as evaluated in this area, c.
Cpnclusions The candidates performed well on both the written and operating examinations, and thus were issued licenses. The candidates appeared to be well prepared for the examinations, indicating that the f acility thoroughly evaluated the knowledge and ability of each candidate in an effort to determine which individuals were ready to sit for an NRC exam Crew communications, control board awareness, and crew briefings were very good. As noted in the past, the BVPS training department, again, did an excellent job in adhering to the examiner standards and in developing the examination materials needod to administer the examinations.
E8 Review of UFSAR commitments A recent discovery of a licerises operating their facility in a manner contrary to the updated final safety analysis report (UFSAR) description highlighted the need for a special focused review that compares plant practices, procedures and /or l
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parameters to the UFSAR descriptions. While performing the preexamination activities discussed in this report, the inspectors reviewed applicable portions of the l
UFSAR that related to the selected examination questions or topic areas. No discrepancies were identified as a result of this review,
- V. Manaaement Meetinas Xi Exit Meeting Summary
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On December 18,1997, the NRC ex9 miners discussed their observations from the f
examinations with Beaver Valley Unit 1 operations and training management e
representatives. The examiners discussed generlo candidate performance, including l
communications and briefings, both of which were seen as being very good.
i The examiners also expressed their appreciation for the cooperation and assistance that was provided during both the preparation and examination week by licensed operator
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training personnel and operations personnel. Beaver Valley personnel contacted and/or present at the exit meeting included the following partiallisting, i
BEAVER VALLEY K. Beatty, General Manager, Nuclear Support R. Brooks, Sr. Nuclear Operations Instructor T. Burns, Director, Operator Training S. C Jain, Vice Paesident, Nuclear Services R. Hart, Senior Licensing Supervisor T. Kuhar, Licensed Operator Training Supervisor B. Tulte, General Manager, Nuclear Operations NRC
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P. Bissett, Senior Operations Engineer, Chief Examiner T. Kenny, Senior Operations Engineer D. Kern, Senior Resident inspector, Beaver Valley Units 1 & 2 Attachments:
1. Beaver Valley Unit 1 SRO Written Examination w/ Answer Key
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2. Simulation Facility Report s
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Attachment 1 BV.1 SRO WRITTEN EXAM W/ ANSWER KEY r
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RTL WA5.620.H DUQUESNE LIGHT COMPANY Volume 3 Nuclear Power Division Procedure 5-5 Training Administrative Manual Figure 5 5.1 Revision 12 Page 1 of 1
WRITTEN EXAMINATION COVER SHEET
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PROGRAM: Licensed Operator Trainina
CLASS NUMBER: 1 LOT 3 l
l SUBJECT:. SRO Initial NRC Exam
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By this signature, I state that all of the work done on this examination is my own. I have neither given nor received aid.
SIGNATURE DATE 12/15/97
. i NAME DLC EMP #
(Please Print)
COMPANY _
(if other than DLC)
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POSSIBLE POINTS '100 SCORE Instructor Initials
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i PRfiPARED BY ' A, Beckert TRAINING DIRECTOR / SUPERVISOR SIGNATURE
/2MM APPROVAL
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. NRC Exam: ILOT3, Rev 1
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Question 12 971 l
RCS cooldown rate and total flow while on RiiR is controlled by...
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A. manually adjusting R11R flow through the heat exchanger and automatically controlling CCR j
flow through the heat exchange-f f
H. automatically controlling RifR flow through the heat exchanger and manually adjusting -
s CCR flow through the heat exchanger.
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C. automatically controlling RilR flow through the heat exchanger and manually adjusting total-l
R11R flow through the system,
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D. manually adjusting R11R flow through the heat exchanger and automatically controlling total j
Rilk flow through the system.
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j NRC Exam: ILOT3 Rev I
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Question 12 97 2 l
t An Operating Manual precaution in OM 24 states that two (2) Condensate pumps must be i
running befoie starting the second Fecdwater pump. The reason for this is to..,
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A ensure adeqsate now capability to prevent a low suction pressure Feed pump trip.
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cil. ensure adequate Dow capability to prevent robbing the Steam Generators when the Feed f
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pump recirculation valves open.
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C. ensure edequate Dow capability to prevent run out on the lleater Drains pumps.
D. ensure adequate flow to prevent water hammer in the first point heaters.
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BVPS 1 NRC Exam: ILOT3, Rev i
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- Question 12 97 3
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Given the following:
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Reactor trip and Safety injection due to a Small Break LOCA.
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All ESF equipment operated as designed.-
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i Various PAB and Safeguards Area Radiation Monitors are in alarm. -
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i The procedure that should be entered based upon the above conditions is:
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A. ECA-1,1," Loss of Emergency Coolant Recirculation."
11. FR.Z.2," Response to Containment Flooding."
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- C. ECA 1,2,"LOCA Outside Containment."
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D. FR Z.1," Response to liigh Containment Pressure."
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Question 12 97 4 Which of_the following events would result in a Containment Particulate and Gas Radiation Monitor [RM.lRM.215A(ll)] alarm?
A. RCS leak at the incore Seal Table.
II. PZR Safety Valve seat leakage.
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C. Steam Generator Tube Rupture.
D. RCP fil seal failure.
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ItVPS.1 NRC Exain: ILOT3, Rev i
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Question 12 97 5 Olven the following plant conditions:
Unit 1 is in Mode 2.
Reactor power is at 10'"' arnps in the I.R. (P 6 activated).
- Preparing to block the Source Range detectors per the start up procedure.
- The detector for N31 Source Range fails low.
- Which action is required by procedure?
A. Enter E-0," Reactor Trip or Safety injection,"in response to the automatic Reactor trip.
11. Iloid power at 10'"' amps, until repairs are inade.
C. Insert Control Rods until power is less than P 6.
D. Place the N31 Level Trip Switch in the Bypass position.
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IIVPSol NRC Exam: 11.OT3, Rev i
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Question IN97 6 Given the following conditions:
A Reactor trip and Safety injection have occurred due to a SGTR in the 'C' S/G.
- E.3," Steam Generator Tube Rt.pture "is in effect with a cooldown started at
maximum rate.
The highest S/G Pressure is 900 psig.
- RCS pressure is 1000 psig and dropping.
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liigh llead Safety injection flow is approximately 300 gpm.
e Which of the following describes what should be done with the Reactor Coolant Pumps?
A. RCP's should be tripped because RCP trip criteria is currently met, 11. RCP's should be tripped because RCP trip criteria applies after an operator initiated RCS depressurization is commenced.
C. RCP's should not be tripped because RCP inp criteria does not apply once an operator initiated RCS cooldown is commenced.
D. RCP's should not be tripped because RCP trip criteria does not apply until the operator initiated RCS coold?wn is completed.
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IIVPS.1 NI(C Exam: lLOT3,1(w I
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Question 12 97 7 The following conditions exist:
A Reactor trip from 100% power has occurred.
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The Turbine has failed to trip both automatically and manually.
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Which of the following actions is/are required, by procedure, to be performed next?
A. Itunback the Turhine using the " Turbine Manual" and " Fast Down" pushbuttons.
II. Close the Main Steam Trip and Ilypass Trip Valves.
C. Close all Reheat Stop and interceptor Valves.
D. Place the nmning EllC Pump in Pull to-Lock, i
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NRC Exam: li.OT3, Rev i
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Question 12 97 8
Given the following:
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- Reactor power is 90*A I
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e PZR levelis $2% and stable.
RCS pressure is stable at 2235 psig.
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i Containment pressure and humidity are rising.
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Net Charging flow is 0 gpm.
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i Steam flow on all three loops is 3.78E6 lbm/hr.
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- Pressure is stable, i
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- NR level is dropping.
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Which of the following events is in progress'l
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A. = There is a Charging line leak inside Containment.
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- B. "C" S/O controlling Feed flow channel instrument line is ruptured.
C. There is a Feed line break inside Containment.
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- D. There is a Sinall Break RCS LOCA inside Containment.
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IIVPS.1 NRC IIxam: lLOT3 Rev i
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Question 12 97 9 Given the following conditions:
The Reactor has been shutdown for 2 days.
- RCS temperature is 150 degrees F.
- RCS pressure is Atmospheric.
- PZR is at nonnal level for shutdown cooling.
- Assume RilR cooling is lost. Which of the following describes the time available until core boiling occurs?
A. Less than 10 minutes.
II. I1 to 20 minutes.
C. 21 to 30 minutes.
D. 31 to 40 minutes.
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NRC Exam: ILOT3, Rev 1 i
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- Question 12 97 10-
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The following conditions exist:.
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i PZR Spray valves [PCWIRC-455A & B) are closed.
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PZR PORV [PCWlRC-455C] is closed.
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e l-PZR PORV [PCV lRC 456)is cycling.
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RCS pressure is approximately 2000 psig.
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Which of the following Pressurizer pressure transmitter failures has occurred? -
' A. [PT-lRC 445) failed low.
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' II. -[PT lRC-445} failed hlgh.
. C. [PT lRC-444) failed low.
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D. [PT lRC-444) failed high.
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a Question 12 97 11 Given the following:
Unit 1 is operating at 100% power with all systems in their nonnal system arrangement.
Number i DC Ilus voltage is 0 VDC.
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A Reactor trip occurs.
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The Main Feedwater pumps are secured per the Alarm Response Procedure.
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Which of the following Auxiliary Feedwater pump (s) will auto start?
A. MDAFW pump 3A and the TDAFW pump.
11. MDAFW pump 311 and the TDAFW puap.
C, TDAFW pump only.
D. MDAFW pump 31) only.
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IIVPSol NRC IIxam: 11.0T.1, Rev i
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Question 12 97 12 The following plant conditions exist:
Unit I is operating at 100% power.
- Pressuriter Spray valve [PCV lRC-45$11]is stuck OPEN.
- All efforts to close the Spray valve have failed.
- RCS pressure is dropping rapidly
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Which of the following actions should be taken?
A. Trip the 'A' Reactor Coolant Pump, then trip the Reactor.
It Trip the 'C' Reactor Coolant Pump, then trip the Reactor.
C. Trip the Reactor, then trip the 'A' Reactor Coolant Pump.
D. Trip the Reactor, then trip the 'C' Reactor Coolant Pump.
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. Question 12 9713
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Pressurier PORY failing open?.
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PRTTemperature PORV Tallpipd Temperature
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13VPSil NRC Exam: ILOT3, Rev i
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Question 12 97 14 Given the following:
A Turbine / Generator trip has caused a Reactor trip.
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The operators are in ES 0.I," Reactor Trip Response," at step 4," Check RCS
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Temperature Stable at or Trending to $47F."
RCS pressure is 1810 psig and slowly dropping.
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Pressurizer level is 22% and stable.
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Core exit T/Cs are $75'F and slowly rising.
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Containtnent pressure is 19 psia and slowly rising
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All S/G NR levels are 20% and slowly rising.
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Which of the following actions should be taken?
A Dump steam to the Condenser and proceed to step $ of ES 0.1.
11. Initiate Si and go to E 0," Reactor Trip or Safety injection," step 1.
C. Transition to FR il.1," Response to Loss of Secondary lleat Sink."
D. Transition to FR Z.1," Response to liigh Containment Pressure."
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IIVPSol NRC lixarn: ILOT3, Rev i
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Question 12 97 15 Given the following:
The Unit is in Mode I at 50% power proceeding to full power following a refueling
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outage.
All control systems are in automatic.
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All Pressurizer 11 eaters are turned on to allow for boron inixing as power is raised.
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A loss of Containment instrument Air occurs.
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Assuming no Operator actions are taken, which of the following will occur'/
1. Reactor will trip on high Pressurizer level.
2. Letdown willisolate.
3. I'ressurizer pressure will rise.
4. Charging willisolate.
A.1,2, and 3 only, 11. I and 2 only.
C. 2 and 4 only.
D. 1,2,3, and 4.
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liVPSol NRC !!xam: ILOT3 Rev 1
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Question 12 97 16 The plant is in Mode I with all systems in their normal system arrangement. Containment pressure transmitter [PT LM 101C) failed low and actions were completed per 10M 1.4.lF.
" Instrument Failure." Subsequently, Containment pressure transmitter [PT LM 10lli) fails high.
Which of the following will occur?
1. SIS 2. MSLI 3.ClA 4. Clll A. 1, 2,3, and 4.
i 11. 1,2, and 3 only.
C. I and 3 only.
D. I and 2 only.
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NRC Iham: ILOT3 Rev 1 Question 12 97 17
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The plant is preparing for a start up, The Shutdown llank tods are fully withdrawn, and all
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i Control Danks are fully inserted, in preparation for the start up the Control Room operators are performing an Operations Surveillance Test (OST) on Source Range detector N31. Which of the fellowing is correct regarding the perfomiance of this test?
A. The Shutdown llanks must be inserted and the Reactor trip breakers opened. The OST will generate a Reactor trip signal.
11. The Shutdown Danks must be inserted and the Reactor trip breakers opened. This will provide a lower baseline aource Range count to allow nll setpoints to be tested.
C. The Shutdown llanks can be left withdrawn. Ne Reactor trip signal is generated during the performance of the OST.
D. Tbc Shutdown llanks can be left withdrawn. Placing the Level Trip switch to Bypass will prevent the OST from causing a Reactor trip.
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f IIVPS.1 NRC Exam: ILOT3 Rev i
Question 12 97 18 Given the following:
A spurious Reactor trip and Safety Injection have occurred during a Reactor start up,
Tavg is stable at 547'F.
e PZR pressure is stable at 2235 psig.
- All Steam Generator water levels and pressures remained stable at 33% and 1005
psig respectively, liefore resetting S1, which of the following pumps can provide feedwater to the S/Gs?
1.
Main Feedwater pumps 2.
Auxiliary Feedwater pumps 3.
Dedicated AFW pump A. I and 2.
II. 2.
C. 2 and 3.
D. 1,2, and 3.
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- IIVPS 1 NRC Exan.: ILOT3. Rev i
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Question 12 97 19
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When detennining subcooling requirements while progressing through the EOPs, the value for -
subcooling that is read on the inadequate Core Cooling Monitor (ICCM) is the difference l
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between the saturation temperature for the...
A. Pressurint pressure.md the highest loop ilot Leg temperature.
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II. wide range RCS pressure and the average loop 110t Leg temperature.
C Pressurizer pressure and the average of the five hottest Core exit T/Cs.
D, wide range RCS pressure and the average of the five hottest Core exit '"Cs.
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NRC Exam,' l LOT.1, Rev 1
Question 12 97 20 The plant has entered hiode 5 for a refueling outage. Contal'nmer.t vacutun has been broken.
St.ortly aller placing Containment Purge and Exhaust into service. Containment Purge Exhaust Radiation hionitor [Rhi.1 VS.104 A) fails high. This will result in...
1. Containment Purge and Exhaust dampers closing.
2. Auxiliary llui' ding exhaust fans tripping,
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3. hiain Filter llank inlet dampers openin3 4. Containment Purge and Exhaust fans tripping.
A. I and 2 only, 11. 1,2, and 3.
C. 1, 3, and 4.
D. 2 and.1 only.
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Question 12 9? 21 The plant has entered r mini outage to repair Condenser tube leaks. The Reactor is stable in Mode 3. A decision is made to completely isolate the Condenser Circulathg Water system and repa.r all four Condenser sections simultaneously. This evolution should NOT be allowed to occur because this will result in a loss of...
- A. a Dowpath for the Reactor Plant River Water system for Append.i t requirements.
B. nonnal Feedwater and would require Auxiliary Feedwater to maintain Steam Generator levels which is a violation of Technical Speci6 cations, C. the ultimate heat sink requirement for Technical Specifications.
D.- Condenser Steam Dumps and there will be no way to get the plant to Mode 5 if needed.
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NRC Exam:. lLOT3,~ Rev 1 -
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- Question 12 97-22
- When fighting an' electrical fire using foam, wh'ich 'of the following are precautions to be -
- exercised by the Fire Brigsde members?
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- 1. Anticipate and avoid the run olT from the electrical equipment being sprayed.-'
-2. Always wear rubber boots for electrical insula: ion.
'
4. Maintain a minimum distance of 15 feet from tha electrical equipment being sprayed.
-.-
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1 and 3 'only/
.
- IL 2 and 4 only.
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- Question 12 97 23;-
.
- Which of the following actians are directed in AOP.I.6.5," Shutdown LOCA"?J
~
1. Manually initiate Safety injection (both pushbuttons) to recover Pressurizer level.
- i
'2, Align lillSi pumps through the Hot Leg Iq'xtion flowpathi
~ 3.) Manuallp depressurize and cooldown the RCS to place.RHR into service; 4. - Manually start 1111S1 pumps to recover Pressurizer level.
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- A.1,2; and 3.
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B.' 2 and_4 only.
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D..- 2, 3, and 4.
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'*4 lNRC Exam: lLOT3, Rev l-
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- Question 12 97 244
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.
.
.
.
What'is the expected procedure flow path for a Small Break LOCA that is too small for two1 111FISI pumps, but too large for one llHS1 pump, ii., a " smart break"?l
- 1. lOM 11.4.M," Recovery from Safet Injection."
"
2. lOM-10.4.A," Residual Heat Removal System Startup."'
3. E-0, " Reactor Trip or Safety injection." 1
- 5. ES-!.1, "Si Termination."
6. E-1, "L'oss of Reactor or Secondary Coolant."'
?A.'6,5, then 2.l
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B. 3, 6, 5, then 4.-
C. 3, 6, 5, 6, then 1.:
D. 3,6, then 1.
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(Question 12 97 25-~
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- Which'of the following actions does ECA-1,1;" Loss of Emergency Coolant Recirculation'.' '
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1. Provide guidance on aligning th'e HHSI pump suction directly to the Containment sump.
,
L2. - Terminate Cold Leg Recirculation and restore Charging and Letdown.
l t3. Cooldown'and depressurize the RCS to allow RHR to be put into service.
-
4; Provide methods'to mak'e-up to the RWST.
-
A'.
1. 2, and 3.
,
B.;1 and 4 only.
,
C. 3, and 4 only.
D. 2,3, and 4
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NRC ExamiILOT3 Rev't!.
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.: Question'12 97 26
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The unit is at 100% power with all systems in their NSA configurations for the current power
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- level. The RO inadvertently changes the Pressurizer Pressure Master Controller setpoint to 2185 :
.
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psig. Assume a step change in the setpoint and assume the controller remains in automatic.
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- Which of the following is the immediate automatic response of the system?
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- A. Spray valves open; Variable heaters go to minimum output.
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- H. L PORV 455C opens; Spray valves open; Variable heater go to maximum output.
.
i i C. Spray valves open; Variable heaters go to maximum output.
- D Spray valves close; Variable heaters go to minimum.
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DVPS.1 NRC Exam: ILOT3, Rev 1
-
Question 12 97 27 Given the following:
Control Rod D12, a control bank "C" Group 1 rod, has fallen into the core due to an
-.
equi ment failure.
t The equipment failure has been corrected and all retests are completed satisfactorily.
.
The dropped rod recovery is in progress per AOP-l.l.5," Dropped RCCA."
- All applicable switches are in their correct position for the rod recovery.
.
AOP-1,1.5, step 17.a directs the operator to " Anticipate a Rod Control System Urgent
Failure Alann."
The Rod Control System Urgent Alarm is caused by a:
A.
Logic Cabinet failure and all rod motion will be inhibited.
B.
Logic Cabinet failure and only those rods aligned to Power Cabinet 1 AC will move.
C.
Power Cabinet 1 AC failure and only those rods aligned to Power Cabinet 2AC will m o Vc.
D.
Power Cabinet 2AC failure and only these rods aligned to Power Cabinet 1 AC will move.
l
.
.
IIVPS 1 NRC Exam: ILOT3, Rev 1
' Question 12-97-28
- Technical Specifications restrict the quantity of radioactive liquids in [lLW-TK-7Il] to less than 10 curies (excluding tritium and dissolved or entrained Noble gases). The basis for this limit is
'
to...
A. prevent over-exposure to personnel who must work near or pass by the tank (Transient Pathway Radiation Levels).
II. prevent exceeding 10CFR20 Appendix illimits at the nearest surface water supply in the
'
event of an accidental release.
C. maintain activity low enough so that the tank may be discharged with minimum design dilution flow.
D prevent exceeding 10CFR100 limits for child thyroid dose at the site boundary in the event of an accidental release.
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" NRC Exam: ILOT3, Rev l-
'
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- Question 12 97 29 Which of the following does the NSS sign for when discharging a Gaseous"/aste Decay Tank? -
- 1. Verification of the proper tank and approval for the dischargec 2.z The appropriate Rad Monitor alarms have been adjusted, f
3. The opposite units NSS has been informed of the discharge.
~ 4. Only one batch discharge is being done at a time.
,
_ A.-~ 1,'and 2 only.
B. 1,2, and 4.:
- C.- 3 and 4 only.
,
- D. 2,3, and 4. -
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Question 12-97 30l
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. Which of the following must be; met in order to terminate Safety injection following a Small -
P LBicak LOCA7?
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.
ll( Adequate subcooling.'
,
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- 2; At least one Reactor Coblant Pump in operation.
-
.
.
.
i 3. Adequate secondary heat sink.
. 4. RCS pressure stable or rising.'-
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^ 5. RCS Hot Leg temperrtures stable or rising.:
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~ 6. Adequate Pressurizer level,
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' 7. Charging and Letdown available.
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Question 12 97 31:)
- LOisen thifollowing
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- - e Ths 'B' RTB has failed to open,.
.-
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' controller will maintain Tavg at.
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r Load Rejecticn -
549*F :
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1 Reactor Trip -
547 F.
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-
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{ Question 12-97-32-
-
j The plant is in Mode I with Reactor Plant River Water Pumps [ LWR P-1 A] and [1WR-P 1B]'in-
.-
-
1 service, in order to replace [1WRiP-1 A] with [1WR-P-lC], which sequence;of operations must-
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iRack on [1WR P-lC]..
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. Rack off[1WR-P-1 A).
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' Start [1WR-P-1C]._
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- Stop [1WR P_-1 A];
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C. L4,1, 3, 2.
D. L.4, 2, 1, 3.
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$ i uestion' 12-97-;3% '
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' A.? 1 AE Emergency Bus feeder breaker, [ACB-1 A10] trips on overcurrent, Blj l AH Emergency Bus reverse piiase PT has a blown fuse.
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1 C. 1 A NeMia! 4KV Bus feeder break 0r, [ ACB-41C) trips on overcurrent.
j D.11 A Normal 4KV Bus feeder breaker, [ACB-41C] trips due to the Main Generator tripping on j
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-l (Question 12 97 34
i Id ECA'-0.0J' Loss of Emergency AC Power ll prior to restoring power to an emergency bus, all:'
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majciloads are placed into Pull to-Lock with the exception of the Reactor Plant River Water ;
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Pump, This pump is left in automatic to...
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. AJ provide a load for the Diesel Generator to prevent it from tripping on overspeed when started.
in the emergency mode, l
H.^ ' provide cooling for the Diesel Generator to prevent overheating and possible fdilure of the
-
fDiesel.
C iprovide cooling for the Charging pumps oil coolers so that make-up to the RCS can begin
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immediately to replace the RCS lost through the RCP seals,
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D.l provide cooling to ths Control Room emergency back up cooling coils to maintain Control
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BVPSol NRC Exam: ILOT3, Rev i
-
Question 12 97 35 The unit is at 100% power when the following Radiation Monitors go into alarm:
[Rhi lMS 102A] N 16 SG Leak Monitor.
[RM 1BD 100] S/G Blowdown Effluent Monitor.
.
[RM ISV-100] Condenser Air Ejector Vent.
.
Within 15 minutes, analysis reveals a 0.12 gpm tube leak with 0 GPD/IIR rate of rise on the 'A'
S/G. Which of the following actions should be taken?
A. Enter TS 3.4.6.2 action for RCS leakage and restore the leak to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in llot Shutdown within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
B. Enter AOP-l.6.4,"S/G Tube Leakage," and commence an emergency shutdown to be in Mode 3 as soon as possible.
C. Enter AOP-l.6.4,"S/G Tube Leakage," and place the plant in llot Standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
D. Manually trip the Iteactor and initiate Safety injection and enter E-0.
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'i i NRC Exam: ILOT3, Rev 11
'
Obestion 12 97-36
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With the ur:it'at 100% power, a HIGil HIGH alarm is received on Condenser Air Ejector Vent
'
Radiation Monitor [RM lSV-100).This HIGH HIGH alarm _will result in the Condenser Air-
.;
Ejector exhaust...:..
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,
l-._Ai being isolated.
[
!!L diverting through the Main Filter Banks.
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- C diverting toLContainment.-
D.: diverting to the Gaseous Waste Surge Tank.
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BVPSol NRC Exam: ILOT3, Rev i
'
Question 12-97 37 There has been a Reactor trip caused by a loss of off site power. E-0, " Reactor Trip or Safety injection" has been entered. A transition to ECA-0.0," Loss of Emergency 4KV AC Power" was done at step 6, because both emergency 4KV busses were de-energized. The crew has reached the step where they are to determine which bus should be selected as the cross-tie bus when the
- 2 Diesel Generator is started locally and the IDF bus is automatically loaded. Which of the following is the appropriate response?
A. Continue on in ECA-0.0 until both eme,3ency busses are restored.
B. Skip ahead to the step which determines the appropriate recovery procedure.
C. Transition to E-0," Reactor Trip or Safety injection."
D. Concurrently perform E-0 and ECA-0.0 until the cross tie is completed for Bus 1 AE.
.
.
.,
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NRC Iharn: ILOT3. Rev 1 -
.t
- -
Question 12-97-38 Which of thE following statements is correct if the Power Range channels have been adjusted based on a calculated calorin.etric? '
.
A'. If the Blowdown How is ignored in calculating the calorimetric, then actual Reactor power would be lower than indicated Reactc,r power.
,
B. ~ If the Illowdown Dow is ignored in calculating the calorimetric, thea actual Reactor po,:r would be higher than indicated Reactor power.
C. If the Feedwater temperature used in calculating the calorimetric had been 10 degrees lower than actual Feedwater temperature, then actual Reactor power would be higher than indicated
,
Reactor power.
D. If the Feedwater temperature used in calculating the calorimetric had been 10 degrees higher -
than actual Feedwater temperature, then actual Reactor power would be lower than indicated-Reactor power.
.
,
IWPS+1 -
NRC Exam: ILOT3, Rev i
'
Question 12 97 39 Given the following:
The plantwas in Mode I when a loss of offsite power occurred.
.
Iloth Emergency Diesel Generators and the ERF D/G have started and loaded their
.
respective busses.
Control of air-operated valves outside of Containment, is...
A.~ not possible until offsite power is restored.
II. not possible until the Diesel Air Compressor is started.
C. Possible since the Station Air Compressors will auto load onto the Emergency Diesel Generators.
D. Possible since the Station Air Compressors will auto load onto the ERF Diesel Generator.
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L NRC Exam: ILOT3, Rev 1 l
. Question 12 97-40 l
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.
SVhich of the following is~ an in'dication that natural circulation exists in the RCS7"
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-
- RCS RCS
' indicaied -
S/G RCS Cold CNMT-
,
' Pressure:
_ Subcooling -Pressure'
. Leg Temp.
Pressure.
on ICCM -
1 00'psig 40 F T400 psig-
,400*F-
= 10 psia
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-A.-
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B..
I100 psig
.30*F-500 psig
'470 F~
7 psig
-
.
. C.
- 1500 psig 35'F 600 psig.
490 F 10 psia -
'
D.
- 2000 psig 40*F 700 psig 550*F 7 psig
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p-
. Question 12 97f41 :
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1The RCP Thermal Barrier Component Cooling Outlet Trip _ Valve,;[TWICC 107B], =..
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(lr Instrument Air and will auto close on high.
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- Question 12 97 42
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The location of the 18 inch escape manway associated with'the CNMT Personnel Airlock Doors
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- A.
on the inner and outer doors with no associated ' interlocks;
- B..
only on the inner door with no associated interlocks..
-
on the inner and outer doors, interlo' ked to prevent any inner and outer door from being
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. opened simultaneously.
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on the inner door, interlocked to prevent any inner door and the outer door from being -
. opened simultaneously.
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BVPS1 NRC Exam: 1 LOT 3, Rev 1
Question 12 97-43 Given the following conditions:
Reactor power level is 97% and dropping.
- Control Rods are in automatic and stepping in,
Tavg is 574'F and dropping.
e Pressurizer pressure is 2225 psig and dropping.
- Pressurizer level is 53% and dropping.
.
Which of the following actions should be taken Hrst by procedtre?
A. Check the In llold-Out lever is in the Hold position.
B. Enter AGP-l.51.1, " Emergency Shutdown."
C. Ilace the Control Rod selector switch in Manual.
D. Reduce Turbine load to stabilize primary plant parameters.
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-
l NRC Exam: ILOT3, Rev l =
"
- Question 12 97-44-
The folloiving conditions exist:
.
- e s - Loops 1 and 3 Tavg indicates 576'F. -
'
. e' ; Loop 2 Tavg indicates offscale high.
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-
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- Loops 1 and 3 Delta T indicates 100%.-
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Loop 2 Delta T indicates 0%.
.
P Which of the following is the cause of these indications? i
,
'A.; Loop 2 Tecid failed Imv.
11. Loop 2 Tcold failed high.
C. Loop 2 Thot failed lor /.
. D.-- Loop 2 Thot failed high..
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.NRC & ami ILOT3, Rev 1:
'
e - Question 12 97-45:
- .
- The following events have occurred: -
' e ;.- The unit is at 100% pc c.er with all systems in their NS A configurations for the '
i current power level when a Reactor trip and Safety Injection occur..
.
.
1 A S/G pressure is droppii.g rapi_dly. -
-
,
1 A steam line indicates 2.5E6 lbnvhr steam flow.
- e
e ~ Containment pressure is 8.5 psig and rising.
- All ESF actuations occur as designed.
-
.
>
,
EAssuming no operator actions takeni a possible c~onsequence of this accident is...
l t
Ai an acidic Containment spray solution since the steam will dilute the spray ring water.
B. insufficient Containment sump levels to support Recirculati n Spray pump operation.
'
C. - a postulated flaw in the Reactor Vessel wall propogating if RCS pressure rises, Dc a loss of the steam driven AFW pump due to all three steam lines depressurizing.
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Given the following:
' The unit is at 100% power with all systems in their NSA configurations -
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L for the cun ent power level when a Reactor Trip and Si occur.
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. The operators transition to FR il.1," Response to Loss of Secondary 11 eat Sink _
due to low S/G NR levele and low AFW flow.
.
RCS pressure is less than S/G pressure and FR-li.I d:reco a transition to E 1,:
.
" Loss of Reactor or Secondary Coolant.
.
.
l Based on this information, select the statement that corre tly summarizes plant conditions:
-
.
A._ Large Break LOCA in progress; secondary heat sink required.
,
B. Large Break LOCA in progress; secondary heat sink not required.
- C. Small Break LOCA in progress; secondary heat sink required,
,
- D. Small Break LOCA in progress; secondary heat sink not required.
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That the injected Accumuiator nitrogen volume has expanded sufficiently to maintain i
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RCS saturation temperature less than the UFSAR design basis,'
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1 Question 12-97-48_-
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l C,=: RilR return flow to the RCS dropping.
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Question 12 97-49
Which of the following is indicative of, or the result of, a dropped rod with an initial power le f
of 100%7 1. Rod insetilon Lituit drops.
,
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2. "NIS l'ower Range Coinparator Deviation" alatin.
i 3. Tavg drops rapidly, i
4. Axlal Flux Difference becoines inore negative, A. I,2, and 3, 11. I,2. and 4.
C, I and 3 o:aly.
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D. 2 and 3 only.
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Question 12 97 50
When responding to a Degraded Core Cooling condition in FR.C.2, the operator is directed to l
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.
" Verify St Valve Alignment" with the Si systent in the Cold Leg injection mode. Ech of the
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following valves should be closed for the current plant conditions?
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A.
RWST Discharge to Charging s' umps Suellen Valve [MOWICll.Il5B].
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Regen llX/Chg Itender inlet CNMT isolation Valve [MOWICli.289].
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C.
. AFW Tutbine Steam Supply 11 Trait Trip Valve (TW1MS 10$H).
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D.
lilt Outlet isolation Valve (MOWISI 867D].
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Question 12 97 51 Given the following conditions:
j The unit was rt 100% power with all systems in their NSA configurations for the I
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e current power level, A Main Steam Line break has occurred inside Containment.
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The crew is in FR P.I. " Response to Inuninent Pressurized Thennal Shock," which l
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?A.&Wann up the RilR system and commence a cooldown of the RCS.
11. - Raise the faulted S/G water level to 50% and secure the TDAFW pump.
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C. Place PZR Auxiliary Spray in service to control RCS pressure.
D.; Energize the PZR heaters to raise the saturation temperature of the PZR, l
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IIVPS 1 NRC Exam: II.0T3, Rev 1 Question 12 97 52 Given the following canditions:
The unit is at 92% power with all systems in their NS A configurations for the cunent
power level.
The Reactor did not trip, and the crew is responding with FR S.1, " Response to e
Nuclear Power Generation /ATWS."
After completing the immediate Manual Actions, the Reactor trip breakers are
opened locally and the Reactor is shutdown, No other malfunctions or actuations have occurred, e
liased on the above infonnation, the crew should...
A. perform the first 15 steps ofli 0 while continuing in FR S.I.
11. transition to !! 0, perfonn the first 15 steps, and then return to FR S.I.
C. return to procedure and step in effect.
D. continue with proceduie and step in effect.
.
IIVPS.)
NRC Exam: 11.0T3. Rev i
.
Question 12 97 53 The following conditions exist:
Unit I is in 1101 Standby following an unscheduled maintenance outage.
e The id nt Operator is performing valve stioke OST I A7.30.
- a
[TV ISS 105A2], RCL llot Leg Samples Outside Cnmt itol valve's closing time is
found to be 3.2 seconds.
The acceptable range listed in the OF. is s 2 0 seconds, the ASME limiting stroke
time is 2.0 seconds, and the Technical Specification limiting time is 21.0 seconds.
liased on this information, the OST should be marked:
A. unsat, and penetration declared inoperable, 11. unsat, and the valve declared inoperable.
C. sat, because the average stroke time was less than Technical Specification limit, but frequency of testing must be doubled.
D. sat, because the second stroke time was within 25% of the ASME limiting stroke time which is acceptable for valves with. stroke times less than 1 minute.
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-
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IIVPSol NRC Exam: ILOT3. Rev i
Question 12 97 54 The Technical Specification limit for RCS activity ensures that the dose at the site boundary will not exceed a small fraction of the Part 100 limits in the event that a occurs.
A. Steam lino rupture induced tube leak.
II. Small lireak LOCA with a stuck open Atmospheric Steam Dump Valve.
C. Rod !!jection accident.
D. Locked RCP rotor accident.
,
.
IIVPS.1 NRC Exam: 11.OT3, Rev 1
Question 12 97 55 Given the following:
A Steam Generator Tube Rupture has necurred in the til S/G.
.
The Crew is in E 3, " Steam Generator Tube Rupture," preparing to cooldown the
.
RCS.
111 narrow range level at 74% and rising.
.
PZR pressure is 1900 psig.
.
Tavg is $47*F.
.
hiain Condenser vacuum is 13"lig Absolute and stable.
.
111 and IC Cire Water Pumps are running.
.
Which of the following actlans is/are necessary to commence cooldown in accordance with E 37 A. Take the Steam Dumps to Steam Pressure mode and manually open the dumps to conunence the cooldown.
II. Take the Steam Dumps to Steam Pressure mode, take both Steam Dump Control Selector Switches momentarily to the Defeat Tavg Interlock position, and then manually open the dumps to commence the cooldown.
C. Commenec the cooldown using Steam Generator Atmospheric Dumps
[PCW1 hts 101 A,11, & C].
D. Commence the cooldown using Steam Generator Atmospheric Dumps
-
[PCWlMS 10l A & C).
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NRC Exam: ILOT3, Rev 1
'
Question 12 97 56 IIVPS 1 Steam Generator Atmospheric Steam Dump valves [PVC lMS.101 A,B,C]
.
i Auto / Manual stations are normally set at 1035 psig. Assuming a malfunction of the
- Auto / Manual station, at what steam pressure will the Atmospheric Steam Dumps open with no i
operator action?
i A. 10$0 psig.
II, 1060 psig.
C. 1070 psig.
.
D. 1080 psig.
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lIVPS 1 NRC Exam: lLOT3, Rev i
l Question 12 97-57 Given the following conditions:
Reactor tripped dua 50 a loss of off site power.
- RCS pressure is 700 psig.
.
Tcold is 370 F.
.
Core exi' T/Cs are 506 F.
.
Pressurizer level is 68% and rising.
- Operators are perfonning ES-0.2," Natural Circulation Cooldown."
e Which of the following describes the cause of the abnormally high Pressurizer level?
A. Pressurizer level instruments are inaccurate due to the loss of Containment cooling, 11. Letdown has not been placed into service due to the loss of offsite power.
C. RCS pressure has reached the injection point for the Accumulators.
D. RCS temperature and pressure are at the point where voiding is occuring in the Reactor Vessel head.
.
IIVPS-1 NRC Exam: ILOT3, Rev i
Question 12 97 58 The following conditions exist:
The Unit is at 92% power with all systems in their NS A configurations for the
.
current power level, Pressurizer level control channel selector is in the 461/460 position.
e
,
'
Pressurizer level channels indicate as follows:
e Channel 459 is 54%.
Channel 460 is 56%.
Channel 461 is 0%.
Channel 462 is 40%.
Which of the following describes the plant response with no operator intervention?
A. llackup heaters energize.
11. Charging flow drops to minimum.
C, Reactor trips on high PZR level.
D. Reactor trips on high PZR pressure.
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BVPS.1 NRC Exam: ILOT.1, Rev 1
'
Question 12 97-59 Rods are being withdrawn in manual during a Reactor start up, with all systems operable. For the Control llanks, which of the following describes the status of the Rod Bottom Lights at the moment A4126. " ROD BOTTOM ROD DROP" annunciator clears?
A. Ilanks A, B, C & D OFF.
II. Dani.s A, B, C. & D - ON.
C. Ilanks A, B, & C OFF; Bank D - ON.
D. Ilank A OFF; Danks B, C, & D. ON.
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NRC Exam: ILOT3, Rev i
Question 12 97 60 Administrative controls or interlocks provided with the Charging pumps are designed to accomplish which of the following:
l 1. If only two pumps are operable, ensure that they are not powered from the same bus.
2. Prevent operation of a Charging pump when the primary plant is water solid.
,
3. Prevent cross tic of emergency busses through the swing pump breakers.
4. Prevent RCS over pressurization in Modes 5 & 6 due to excessive flow.
A. I and 3 only.
11. 1, 2, and 4.
C. 1,3, and 4 D. 2 and 3 only.
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' Quesdon 12 97-61
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1Which of the following variables affect the Containment Spray systems capacity to depressurize
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.
the C6nialnment in the event of a Design Basis Accident (DBA)?
l 1. - Containment temperature, s
2. Containment pressure.
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3.. RWST temperature.
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NRC Exam: 11.OT3, Rev i Question 12 97 62 Which of the following describes the sources of innuent into the Primary Drains Transfer Tank (IDO TK l)?
t A. RCP seal leak off, Samplo System drains, and Si valve stem leak-off.
14. CVCS Excess Letdown divert line, RCP seals and VCT drains.
C. Si header drains, PRT, and Sample System drains.
D. Valve stem leak-off, CVCS Excess Letdown divert line, and Reactor Vessel head 0 ring leak off.
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NRC IIxam: II.OT3. Itev i
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Question 12-97 63 The following plant conditions exist:
Unit I has tripped due to a loss of off site power.
e
- 1 Diesel Generator has failed to start.
e All Steam Generator narrow range levels have remained above 21%.
Which of the following describes the status of the AFW system 60 seconds after the Reactor trip?
A. No AFW pumps running.
II. Iloth inotor driven pumps are OFF; the steam driven pump is supplying AFW flow.
C. Iloth motor driven pumps are supplying AFW flow; the sterm driven pump is OFF.
1). The ' A' motor driven pump is OFF; the 'll' motor driven and the steam driven pumps are supplying AFW flow.
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Question 12 9744 j
- Which one of the following is tru,s with regard to Axial Flux Difference?
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A. Horation will cause the Axial Flux Difference to become more negative.
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H.' When Axial Flux Difference is negative, more power is being produced in the top of thm l
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- C. For power levels greater than 50%, Axlal Flux is maintained within + or.7% of target band.
D. When power distribution is distributed equally through the core, AFD is 1.
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NRC Exam: ILOT3, Rev i j
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Question 12 97 65 While responding to inadequate core cooling, the operators are unable to estabilsh liigh if cad l
Safety injection. Core exit T/C's are 1450'F and RCS pressure is 165 psig. Which of the i
following states the reason for starting the RCPs under these conditions?
i i
A. Flush nitrogen from the S/G tubes so natural circulation flow can be established in
,
subsequent steps.
IL Provide 2. phase forced flow for temporary core cooling to reduce Core exit T/C temperatures, C. Assure the core remains shutdown by adding borated water from the loops to the voided Core.
D. Paavide forced RCS flow for heat transfer during S/G depressurization,
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Question 12 97 66 Given the following conditions,:
Make up to the RCS has increased and the following alarms are received:
.
Reactor Coolant Pump Seal Leak off Temp liigh.
.
Reactor Coolant Pump Seal Leak off Flow liigh.
.
Reactor Coolant Pump I A Seal Vent Pot Level liigh.
.
Reactor Coolant Pump No. I Seal Differential Pressure Low.
.
Which of the following has occurred to the l A RCP7 A. til seal has failed.
II. til and t/2 seals have failed.
C. All the seals have failed.
D. Sealinjection has failed.
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f, Question 12 97 67
=Olven the following:
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.;
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~ In this condition, the Containment Air Recirculation Fans...
.
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. A.; niust continue Io run to ensure Containment pressure returns to sub atmospheric within
-
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. H, will continue to run until the Contal.iment sump level reaches the high level trip retpoint.
.
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' ' C. will be tripped due to the ClH.
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- D. will be tripped due to the SIS.-
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I Question 12 97 68 Unit i Laundry and Contaminated Shower Drain Tank [lLW.TK 6A)is being discharged to the
,
Unit I Cooling Tower lilowdown. A 111G11111G11 alarm is received in the control room from the discharge Radiation Monitor [RM lLW il6). Identify the automatic actions that occur.
(FCV lLW 103] Contaminated
[TV.lLW.ll6] Contaminated hQhposal Flow Control Drains Disdjarge header Rad TrinJ_alYs A.
Closes Closes.
i 11.
Closes Opens.
C.
Opens Closes.
D.
Opens Opeas.
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IIVPS l NRC IIxam: !!.OT3, llev i
Question 12 97 69 Given the following:
Unit I has experienced a Large Ilreak LOCA.
- Containment pressure is 20 psig and dropping.
- The Crew has reached the step in the procedure where Clliis to be reset. Which of the following apply in this a ation?
A.
Clll cannot be reset until Containment pressure is below the actuation setpoint.
II.
Clll cannot be reset until Containment pressure is subatmopherie.
C.
Cill can be veset, but will re actuate as soon as the reset switches are released.
D.
Clll can be reset regardless of Containment pressure.
,
.
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NRC Exam: ILOT3 Rev 1
!
.- Question 12 97 70 l
i
. In the event of a gross fuel element failure, wlilch one of the follhwing monitors would be the
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i 13.- Containment Area Radiation Monitor..
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C. Vent stack Radiation Monitor.
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. D.- SPINO Radiation Monitor.
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NRC Exam: ILOT3, Rev i
Question 12 97 71 A plant heat up is in progress. The following RCS temperatures were recorded at the given
,
times:
IIME TEhiP 1000 362F f
1030 383F i
'
1100 412F 1130 440F
'
1200 459F Which of the following statements is correct?
A. No Administrative or Tech Spec limits were exceeded.
11. The Administrative limit was exceeded but the Tech Spec limit was not.
C. Iloth the Administrative and the Tech Spec limits were exceeded.
- D. Not enough data has been gathered to determine if any limits were exceeded.
.
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NRC Exam: ILOT3, Rev i
_ Question 12 97-72 The unit is at 100% power with all systems in their NSA configurations for the current power
' level.
Which of the following will occur if Vital Bus IV is de energized?
A. ' An OTAT Turbine runback will occur.
- H. All Condenser Steam Dumps will open,
,
C.; 'C' S/0 Feed Reg Valve [FCWIFW.498] shifts to manual.
D. The controlling PZR level channel fails low.
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NRC Exam: ILOT3, Rev i
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Question 12 97 73
.
F The following conditions exist:
i
- e z Unit I is operating at 100% power.
I Tavg is 576'F.
j e
Pressurizer pressure is 2235 psig.
l o
e: Delta 'l' is 48.
l
e - Delta T is 60'F..
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.
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cA.- Delta 'l' lowers to -2.
14. Tavg rises to 578"F.
l C. Delta T drops to 55'F with Tavg remaining constant.
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D.: Pressurizer pressure rises to 2260 psig.
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Question 12 97 74 l
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Prior to an outage, the Steam Generators.were producing steam at 780 psia with an RCS Tavg of
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575'F. During the outage 12% of the Steam Generator tubes were plugged.
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What will be the new full power steam pressure if the RCS Tavg is to remain constant at the
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- H. 725 ps!n i 10 psia.
-C:780 psia i_10 psia.
D. 825 psia i 10 psia.
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NRC Exam: II.OT3, Rev i Question 12 97 75 What condition (s) would prevent the Reretor cavity upender from being operated?
1. The transfer car is at the Fuel Pool end ofits travel.
2. The transfer tube valve is clor,ed, 3. The manipulator crane is over the upender with the mast in the full DOWN position.
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Question 12 97 76 Which of the following describes the condition associated with the starting air system for the L
Diesel Generator that would cause the Control Room Diesel Generator Not Available annunciator?
A. One of the two starting air header pt:ssures less than 165 psig.
11. Iloth of the staiting air header pressures less than 165 psig.
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C. One of the two starting air header pressures less than 200 psig.
D. Iloth of the starting air header pressures less than 200 psig.
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B. Fuel Building sump 'evel higli~alann.
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12 97-78.
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A' problem in the Instrument Air' system caused air pressure to momentarily (30' seconds) drop to
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B. It closes and remains closed.
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Given the following:
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' f Cooll'ng, Water Surge Tank?ls
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-A, Excess Letdowr Heat' Exchanger, u
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Question 12-97-811
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. Which of the following represents the condition of ths steam entering the PRT from a leaking:
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NRC Exam: IL' T3, Rev i Question 12-97-82 Given the following:
Unit 1 in Mode 3 preparing for a normal plant cooldown.
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Condenser Steam Dump system is in the Steam Pressure Mode controlling Tavg at 547F
.
in automatic.
Main Steam 11eader pressure transmitter [PT-lMS-464] fails low.
.
What manual actions will be required by the operator to continue the cooldown?
A. Manually close steam dumps by switching to the Tavg mode. Cooldown manually in the Tavg mode.
B. Manually close steam dumps by placing the controller to manual and reducing demand.
Cooldown manually in the Steam Pressure Mode.
C. Manually open the steam dumps by placing the controller to manual and raising demand.
Cooldown manually in the Steam Precure Mod.e.
D. Steam Dump control is inoperable. Cooldown manually with the Atmospheric Steam Dump Valves or the Residual f eat Release Valve.
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L Question 12 97-83:
Given the following: -
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Operations manning is at minimum shift compliment at each unit.- -
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' An oil fire breaks out in the Unit 1 Turbine building basement.
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The NSS determines a Unit I plant shutdown is required.
.
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Which of the following is correct?
A. Off site fire fighting personnel must be called in to combat the fire. Not enough personnel c
are available on-site to fight the fire and shutdown simultaneously.
B.- Off-site fire fighting personnel must be called in to combat the fire. NPDAP 3.5," Fire Protection," requires the use of off-site personnel for any oil or chemical fires.
C. The' opposite units fire brigade will fight the fire and the affected unit will perform the required shutdown duties.
D. Minimum shift compliment provides enough persona:1 to safely shutdown the plant and pei ig m the required fire lighting duties.
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. Obestion 12 97 84-
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' A Control Room fire caused evacuation to the Shutdown Panel (SDP). -
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Plant control is established at the SDP with' all equipment transferred.
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-With equipment controlled fr the SDP, RCS semperature is controlled using
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- Condenser Steam Dumps PZR IIcater group C.
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ANospheric Steam Dump valves.
PZR heater groups A and B.
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- Condenser Steam Dumps.
Auxiliary spray.
D.
Atmospheric Steam Dump valves, PZR heater group C.
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A Reactor trip and Safety Injection have occurred, along with the loss of all Feedwater.
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The steam driven AFW pump hasjust been returned to service.
.
The crew goes to step 28 of FR H.1," Response to Loss of Secondary Heat Sink,"
.
(attached)
The following conditions are noted:
.
RCS Hot Leg temperatures tre all greater than 520F.
- Al S/G wide range levels are less than 10%.
.
Core i eit TCs are stable.
.
The action that should be taken is...
A. establish AFW flow ofless than 100 gpm to one S/G until narrow range level is greater than 6%.
B. establish AFW How ofless than 100 gpm to all S/Gs until narrow range level is greater than 6%.
C establish maximum AFW Gow to one intact S/G.
D. establish maximum AFW flow to all intact S/Gs.
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-~ Questiori12 97-86 ;'
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TWhich of the following designates an immediate Manual ' Action: statement in~ Emergencp-j
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NRC Exam: ILOT3, Rev 1 Question 12 97-87 Which of the following correctly describes the monitoring of the Critical Safety Function' Status Trees? s A. If an orange terminus is encountered, the STA is expected to monitor the remaining trees. If
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a red terminus is encountered on a lower tree, the crew will continue the orange tenninus -
with the higher priority.
B. If a red terminus is encountered during the performance of the immediate action steps of E-0,
" Reactor T.-ip or Safety injection", the crew is expected to immediately perfonn the Functional Recovery Procedure required by the tenninus.
C. A yellow tenninus will require continuous monitoring until all conditions are satisified.
D. -if only green or yellow terminus exist, monitoring frequency may be relaxed to once every 10 to 20 minutes, unless a significant change in plant status occurs.
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Question 12-97-88 J
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Burnup is 8000 MWD /MTU.-
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.. Shutdown Banks are withdrawn, a
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~ Using the attached plant curves, deterniine which of the following statements is correct:
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A; Miniraum Shutdown Boron requirement is satisfied for the existing conditions.
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- B. A cooldown to Cold Shutdown can be accomplished at t e present boron concentrat on.
C; A cooldown to Cold Shutdown can be aeomplished at the present boron concentration if the Shutdown Banks are inserted.
D, The current boron concentration is adequate to maintain Hot Standby requirements if the
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Ci A visual exam is all that is necessary, a " hands-on" verification is not required.
.D.- The Independent Verification for a procedure step can be substituted with the cunt-
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A Reactor trip and Safety injection have occurred from 100% power,
E-0, " Reactor Trip or Safety injection," has been completed through step 13 (Verify e
Feedwater Isolation).
No Feedwater flow to the Steam Generators is indicated.
- AFW pumps cannot be started.
- Which of the following actions is required?
A. Transition to FR-li.1 when directed to by E-0.
B. Immediately transition to FR-11.1," Response to Loss of Secondary Heat Sink."
C. Transition to FR-il.1 as soon as a transition out of E-0 occurs.
D. Go to ES-0.0,"Rediagnosis," which will allow a transition to FR-H.l.
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(Question 12-97 91:
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f A"."A change in initial conditions.
' B. _ A modification to setpoints.
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D.- Addition of steps to return equipment to NSA.-
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{ Question.I2 97 92L-
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A.! Up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> aner the date on the copy.=
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i C. Up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after the date on tl': copy,
' D. Must be checked each shift.
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[ Question 12-97 93'
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..Whicif of the following are considered."extremitles" when considering radiatiori de2e limits?
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- Question 12-97-94-
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Operators havejust stabilized the_ Unit from an unexplained 15% load rejection from full pt iver <
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iwhen another, also unexplained.ll5% load rejection occurs._Which of ths following actions.
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A. Re-stabilize the Unit in accordance_with the Load Rejection AOP -
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- the operator is another valve that emits 100 mR/hr at 25 cm. Which of the following is-applicable to this situation?
.: . A. The valve behind the operator must be labeled as a "liot Spot" and the area posted - as a Radiation Area. P B. The operator will exceed his/her 10 CFR 20 dose limits. A> . C.- The area must be posted as a liigh Radiation Area and the operator must have an - , i integrating dose meter, or fiealth Physics coverage.- D. A fiealth Physics technician should be present to monitor the radiation in the rcom with a portable neutron meter while the operator is stationed at the valve.- . . , _. b - - --,. * - ._ . - - - ,. .. ..- . - -....-....... -... - , e: .-:BVP 1- ..
- NRC Exam: ILOT3 Rev 1: ! -Question 12-97-98? ' DC Bus #1 has been de-energized due t'o a fire in the switchboard cabinet. A Reactor trip then -
' - occursi Recovery from this transient is complicated because... - ! , _ -... . , A? control of emergency loads on the 1 AE bus are not possible from the Control Room. ' B.; the #1 Diesel Generator will start and load, but not be able to be stopped from the Control ~ j
- 1 Room. C. Ch'arging and Letdown will not be available due to the loss of the Station Air Compressors. ' ' D.: temperature control will be en the Residual ifcat Release valve because the Condenser and - ' _
- Atmospheric Steam Dumps will not be available.
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- Turbine Driven AFW pump secured.
- 2 Motor Driven AFW pumps running.
- Tavg is 529 F and lowering slowly.
- All narrow range S/G levels are 4%.
. A partial Feedwater isolation has occurred, and all Condenser and Atmospheric Steam Dump valves are shut. Assume all trends continue. The correct operator response to these conditions is to.. A. reduce AFW Gow and initiate Main Steamline Isolation. B. reduce AFW Gow and commence emergency boration. C. initiate Main Steamline isolation and commence emergency boration. D. Establish SGFW bypass Dow, secure AFW How, and initiate Main Steamline Isolation. ~ m ~
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' 3. Question 12-97 100- ~* ' Nith the plant operating at 100% power, Intermediate Range detector N35 fails highifollowedi- -- - i immediatelp byLN36 failing low, Wh'ich of the following is the correct isperator response 7[ Lj , -
_ . , Ad initiate a Reactor Trip and enter E-Oi" Reactor Trip or Safety injection," ! . _ i , - B. = Restore at least one Intermediate Range channel in the next hour or be in Hot Shutdown
- .within the following six hours.
l C.= Since the number of channel operable is less than the number required by the Minimum- Operability Channel requirement, Technical Specification 3.0.3 applies. _;
- t D.1 Power operation may continue in accordance with Technical Specifications.
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- Questio'n 12 97101 - 1. - (D-
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- A' 51, -C 76. A-
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'A' ' 27.'-- -D 52. D-77. - C: 3. - C-28. iB-53. B-78; 'B-
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i 29. B? 54. A- D 5. D - - - 30. D: 55. : 'D 80. - A 6. C. 31. -A 56. B 81. .A 7. - B-- 32. D - 57. D 82. C R. C 33. -A 58. C. 83. D 9. D _ ' 3 4. B 59. D-84. -B 5. -I)rd-10. B 60.- C 85. A-11. B1-36. C 61. A 86. C 12, D 37. B 62. D 87. D 13.- , C-38. A 63. D 88. D B-39. -B 64. C-89. B 15.. A-40. C 65. B 90. A ~ 16. B 41. -A 66. A 91. D-17. - D' 42. A 67. C 92. C - 18.- B 43. C 68. A 93. 'D 19. D-44. B 69. D 94. C t 20. C 45. C 70. A 95. D ^21- 'A 46.
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B 96. A - 22. C 4 7.- C~ 72. C 97. C - 23. .D. - 48. A 73. B 98. A 1B- A -- 74. B 99. C a - 251 D . 50. B. 75. C: 100. D - 101 Y, .. .. - . .. .,..,,. . . ' '_ _ _ _ _ _ _ _ _ _ _ _ _ _, _ _, _ _, _ _ _ _ _ _ _ _ _ _ _ _ _ _ _, ' _ _ _ _ _, _ _ _ _ _ _ _ _ _ _ - .- .. -- -
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i-Attachment 2 SIMULATION FACILITY REPORT
Facility Licensee: Beaver Vallev Unit 1 Facility Docket No: = 50-334 , . Operating Tests Administered from: December 16 17,1997 This form is used only to report simulator observations. These observations do not constitute audit or inspection findings and are not, without further v6.'ification and reviewi indicative of noncompliance with 10 CFR 55.45(b). These observations do not affect NRC certification or approval of the simulation facility other than to provide information that may be used in future evaluations. No licensee action is required in response to these observations.. , No simulator deficiencies, that affected the scenario examinations or JPMs, were identified - during the conduct of the examinations. , i _