IR 05000266/1998301
| ML20236V586 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 07/29/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20236V575 | List: |
| References | |
| 50-266-98-301, 50-301-98-301, NUDOCS 9808040062 | |
| Download: ML20236V586 (75) | |
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U.S. NUCLEAR REGULATORY COMMISSION
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REGIONlli Docket Nos:
50-266; 50-301 License Nos:
50 266/98301(ORS); 50-301/98301(DRS)
Licensee:
Wisconsin Electric Power Cornpany Facility:
Point Beach Nuclear Plant, Units 1 & 2 Location:
6612 Nuclear Road Two Rivers, WI 54241-9516 Dates:
June 22 - 25,1998 Inspectors:
R. Bailey, Lead Examiner / Inspector J. Hansen, Examiner / Resident inspector J. Lariz73, Examiner-in-Training Approved by:
Melvyn Leach, Chief, Operator Licensing Branch Division of Reactor Safety 9808040062 980729
PDR ADOCK 05000266 V
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EXECUTIVE SUMMARY
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l Point Beach Nuclear Plant, Units 1 & 2 f
NRC Inspection Report 50-266/98301; 50-301/98301
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I An NRC developed initial operator licensing examination was administered to three senior i
reactor operator (SRO) applicants at the Point Beach Nuclear Plant facility. The examination process incorporated a one week period for validation of the examination material and a one week period for administration.
Results:
One SRO license applicant passed all portions of the examination and was issued an
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operating license.
Two SRO license applicants passed all portions of their respective examinations but
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were not issued operating licenses. Licenses will be issued upon completion of eligibility requirements for a Point Beach Nuclear Plant operator.
Observations:
An appropriate level of attention to detail was observed during a man control room l
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walkdown and parameter logging evolution. Control operator actions were consistent with licensee expectations as outlined in their administrative procedures. (Section 01.1)
Operating procedures were generally accurate and complete for performing the task
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assigned. However, the examiners identified a procedural deficiency attributed to inadequate verification during the technical accuracy review process. The licensee was unable to determine whether the procedural error had been a technical or clerical mistake. (Section O3.1)
The licensee provided a competent technical review and verification of the examination
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validity and provided positive feedback to the examiners when modification was required. The licensee maintained a high level of examination security. (Section 05.7)
The plant specific simulator accurately mimicked changing plant parameters during
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varied equipment malfunctions and plant conditions. However, a number of long standing software and hardware problems have the potential of impacting future simulator performance and licensed operator training, if not resolved. (Section 05.7)
While the licensee was meeting the requirements of the licensed operator requalification
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program, the licensee's decision to consider a scheduled refueling outage an " unforseen circumstance" for deletion of a scheduled simulator training session may not have been appropriate. As a result, two operating crews received 30% less simulator training than the other four operating crews and selected operators from the two crews had documented performance weaknesses during training which included one failure of the annual operating evaluation. (Section 08.1)
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d.
Report Details Summarv of Plant Status During the examination period, Unit 1 was conducting a plant startup following the Cycle 24 refueling (U1R24) outage and Unit 2 was operated at full capacity.
1._Qperations
Conduct of Operations 01.1 General Comments (71707)
The examiners utilized various operations and administrative procedures during the validation and administration phases of the examination process (see Section O3 for additional comments). Additionally, the examiners observed licensed operator performance in the main control room during normal plant operations on Units 1 and 2.
The examiners observed an appropriate level of attention to detail during a control board walkdown and parameter logging evolution. Operator actions were consistent with licensee expectations as outlined in its administrative procedures.
O3 Operations Procedures and Documentation 03.1 Emergencv Ooerating Procedure Deficiency a.
Insoection Scoce (71707)
The NRC examination utilized various normal, abnormal, and emergency operating procedures during the development process. During the validation phase, the examiners and the licensee verified the reliability and validity of each procedure required by the applicants during the administration phase.
b.
Observations and Findings The examiners identified a procedural deficiency in ememency operating procedure EOP-2, Unit 2, " Faulted Steam Generator Isolation," during the validation of a plant job performance task on the walkthrough portion of the test. The auxiliary feedwater pump
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l P-388 discharge valve for the steam generator 'B' was listed as "AF-44" in the
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procedure but labeled as 2AF-45 locally. A review of the facility prints and diagrams l
revealed that the valve's marking was consistent with the plant design configuration. A I
licensee representative noted the deficiency for corrective action implementation.
The examiners discussed the procedural deficiency with the licensee's management.
l Due to examination security concerns, the licensee delayed prompt action to generate a
temporary procedure chango request. The procedural deficiency affected the j
applicants' performance during that task in that their response under simulated
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emergency conditions was delayed due to an acknowledged error which was not i
expected. However, no adverse consequences were noted in the applicants'
performance.
c.
Conclusions Operating procedures were generally accurate and complete for performing the task assigned. However, the examiners identified a procedural deficiency attributed to j
inadequate verification during the technical accuracy review process. The licensee was I
unable to determine whether the procedural error had been a technical or clerical mistake. The licensee's prompt action to implement a procedural change following completion of the examination process minimized the chance of an operator error during an actual emergency.
Operator Training and Qualification 05.1 General Comments (NUREG-1021)
The examination was NRC developed and administered using the guidance provided in NUREG 1021, " Operator Licensing Examination Standards for Power Reactors," Interim Revision 8. The examination contained-l A written test consisting of multiple-choice questions focusing on a broad
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e spectrum of various plant systems' design and operation.
A walkthrough test consisting of an administrative topics section and a job
performance tasks (JPMs) section focusing on control room and inplant systems'
operation.
A performance-based section focusing on normal and abnormal plant conditions
using the plant specific simulator during a dynamic scenario set.
Each applicant participated in all portions of the examination process with an NRC
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l examiner evaluating that performance.
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05.2 Written Test a.
Scoce (NUREG-1021)
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l The written test was NRC developed and administered using the guidance contained in ES-401 and ES-402, respectfully, of NUREG-1021. Each applicant was administered a 100 question (multiple-choice) examination. The grading of the written examination was conducted using the guidance contained in ES-403 of NUREG-1021.
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b.
Observations and Findings (
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The licensee provided a thorough technical review of each multiple-choice question to verify the validity of the question's stem and distractors such that only one answer was correct. The accuracy of that review was evident by a lack of post-examination i
comments needed to clarify a discrepancy. All three applicants scored greater than 80.0 on a possible 100 points written test. However, four questions were commonly missed by a majority of the applicants (see Enclosure 2 for more detail).
05.3 Facility Walkthrough Test a.
Scooe (NUREG-1021)
The walkthrough test was NRC developed and administered using the guidance contained in ES-301 and ES-302, respectfully, of NUREG-1021. Each applicant was given ten job performance tasks, with seven being administered on the plant specific simulator and three being administered inside the plant facility complex. Additionally, four administrative tasks and two questions were given in a control room setting. The grading of the walkthrough examination was conducted using the guidance contained in ES-303 of NUREG-1021.
i b.
Observations and Findings The licensee provided a thorough technical review of each administrative and job performance task to verify the validity of the expected operator actions and supporting procedural direction. There was no need for clarification during the administration process. An appropriate level of personnel support was afforded the examiners to accomplish the examination process with minimal delay or distraction to the applicants.
All three applicants scored satisfactory during the performance of the tasks assigned
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and no commonly missed performance items were identified.
05.4 Intearated Plant Ooerations Examination a.
Ifyamination Scooe (NUREG-1021)
The integrated plant operations test was NRC developed and administered using the guidance contained in ES-301 and ES-302, respectfully, of NUREG-1021. Each applicant participated in two different dynamic scenarios using the plant specific simulator. Each scenario consisted of selected instrument and component failures with a major transient leading to implementation of the emergency procedures. The grading of the integrated plant operations examination was conducted using the guidance contained in ES-303 of NUREG-1021.
b.
Observations and Findings The licensee provided a thorough technical review of each dynamic scenario malfunction to verify the validity of the expected operator actions and supporting
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l procedural direction. The accuracy of that review was evident by a lack of operator l
performance issues or concerns. An appropriate level of personnel support was afforded the examiners to accomplish the examination process with minimal delay or distraction to the applicants. All three applicants scored satisfactory during the performance of the scenarios and no commonly missed performance items were identified.
05.5 Examination Administration and Security a.
Scoce (NUREG-1021)
l The examiners administered the examination over a four day period. Examination
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administration process and security measures prescribed in ES-201 of NUREG-1021 l
were discussed with the licensee and verified during the administration phase.
t b.
Observations and Findinas The licensee made available to the examiners a locked cabinet located inside of a locked room to store examination material during the validation and administration process. Only personnel signed on the NRC security agreement were allowed access to the room. Additionally, the licensee sequestered the applicants and kept them segregated during the administration period. No security compromise concerns were identified.
l 05.6 Simulator Fidelity
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Examination Scooe (NUREG-1021)
The examiners observed the performance of the plant specific simulator during job performance (JPM) tasks and dynamic scenario sets. The simulator fidelity report (see Enclosure 3) findings were included to reflect simulator performance concems identified during the conduct of the operating test, as prescribed in ES-501 of NUREG-1021. The examiners also reviewed the licensee's list of documented hardware and software simulator discrepancies, b.
Observations and Findinas While most of the plant / simulator discrepancies had been identified and were being evaluated for resolution, the examin.ers determined that a number of these discrepancies had existed for an extended period of time. Specifically, seventeen licensee identified software deficiencies were awaiting resolution since 1996. The examiners were not able to determine if any negative training or slgnificant simulator performance problems had occurred based upon these discrepancies. The examiners also identified that one long standing discrepancy regarding the modeling of DC power sliders for evaluating reactor coolant pump operations was not noted on the discrepancy lis._
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The examiners were informed early that the Unit 1 plant specific simulator modeling had i
not been updated following a recent plant modification to replace both steam generators. The licensee had postponed the modification until completion of the initial
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examination process. The Unit 2 plant specific simulator modeling had been updated to reflect a recent steam generator replacement modification on Unit 2, but was not utilized since the' po: tion of the simulator had not been NRC certified.
OS.7 Conclusions en Ooerator Trainina and Qualifications The applicants appeared to be highly trained as evident by their satisfactory I
performance. The licensee provided a competent technical review and verification of the examination validity and provided positive feedback to the examiners when modification was required. The licensee maintained a high level of security.
The plant specific simulator accurately mimicked changing plant parameters during varied equipment malfunctions and plant conditions. However, a number of long standing software and hardware problems have the potential of impacting future l
simulator performance and licensed operator training, if not resolved.
Miscellaneous Operations issues 08.1 Review of the Simulator Trainina Time Provided Durina Reaualification Trainina a.
Insoection Scoce (NUREG 1021)
The examiners reviewed several licensed operator requalification training records, and interviewed operations and training personnel regarding simulator training provided as part of the requalification program.
b.
Observations and Findinas The examiners were informed that an operating crew would normally receive about 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> of simulator training in a year. However, the examiners determined that two of l
the six operating crews had received only 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br /> of simulator training during the past year which included the most recent annual operating examination. A portion of both crew's scheduled simulator training had been postponed due to a Unit 1 restart. A Training Advisory Committee review made the decision not to require a makeup of the i
previously missed simulator training. Section 5.2.2 of the Licensed Operator Requalification Training Program allowed the committee to drop a scheduled training session based upon an " unforseen circumstance" which was defined as the restart of Unit 1 following a refueling outage. The requalification training program also required factors, such as past performance, to be included in the decision to reschedule or cancel training.
The examiners reviewed the individual requalification training records for both crewt i
The examiners noted that two of the senior reactor operators had documented
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significant performance weaknesses during the most recent simu ator evaluations and l
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A one had failed the annual operating examination. The examiners expressed a concern that reduced training and evaluation time on the simulator might allow a less competent operator's performance to go undetected. The licensee acknowledged the findings and were evaluating the scheduling of simulator training time during the next requalification cycle and intended to increase the number of hours on the simulator during training.
During the review of requalification training records, the examiners identified a concern that individual training records ready for microfilming were not being stored properly.
The examiners noted that cardboard boxes containing individual records were not stored in a fireproof cabinet as required by procedure TI-6.0, " Nuclear Power Business Unit Training Instructions," Revision 1. The licensee had removed the 1907 training records with the intent to send them directly to central files. However, a work backlog had delayed the delivery. The licensee promptly initiated a Condition Report and placed the records in a permanent records storage vault.
c.
Conclusions While the licensee was meeting the requirements of the licensed operator requalification program, the licensee's decision to consider a scheduled refueling outage an " unforseen circumstance" for deletion of a scheduled simulator training session may not have been appropriate. An important aspect of the systematic approach to training concept deals with the evaluation of the operator's mastery of the objectives presented during training which was not consistently applied to all crews.
08.2 (Ocen) IFI 50-266/97019-02: 50-301/97019-02: No Process in Place to Ensure Continued Validation of Expected Operator Response Times During a Design Basis
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Accident.
In October 1997, NRC inspectors determined that the licensee had not been validating l
the 30 minute operator response to isolate the ruptured generator during a steam i
generator tube rupture (SGTR) accident. The timed operator response was an I
assumption stated in the Final Safety Analysis Report and was utilized in several plant design basis documents. The licensee had implemented several correct actions regarding the timeliness of operator response which included:
Researching the design basis justification for the 30 minute operator response
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time to isolate a ruptured steam generator.
Evaluating each operating crew's timeliness during a SGTR event in training
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cycle 98-3 (4 crews completed).
Training all crews on EOP-3, SGTR, implementation during training cycle 98-3
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(4 crews completed).
Reviewing design basis assumptions regarding operator timed performance
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The crew evaluation results indicated that the 30 minute assumption could not be met l
during an actual event. The crews' isolation time ranged from 36 to 50 minutes. The licer.see noted that the isolation times may actually be higher during an accident based upon extensive training just prior to the simulator evaluation phase.
The licensee was continuing an evaluation of non-conservative operator responses as
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l compared to design basis commitments and intends on completing the investigation by October 9,1998. The licensee determined that, in addition to the 30 minute STGR operator response, nine other instances of assumed operator actions existed in design basis documents. The licensee was implementing corrective actions to ensure that the assumptions stated were accurate and that operations personnel could implement the actions within the required times.
The inspectors determined that the licensee investigative and corrective actions were still on going and this inspector follow up item will remain open.
V. Management Meetings l
.X1 Exit Meetina Summarv i
Examiners presented the examination team's observations and findings to members of the
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licensee's management on June 25,1998. The licensee acknowledged the findings presented
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and the need to enroll the applicants in the continuing operator license program. No proprietary information was identified during the examination or at the exit meeting.
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Enclosure 2 Post Examination Comments and NRC Review Reoort
A.
The licensee's post-examination review identified no technical or administrative concerns with the written examination as administered.
B.
The examiners noted a few common deficiencies during the review process based upon 2 or more of the applicants having e.siected an incorrect answer:
1.
Question No. 33 Understanding of the conditions required to secure a reactor coolant pump during a response to inadequate core cooling condition.
2.
Question No. 77 Understanding of the conditions and actions required to secure the only running. residual heat removal pump during l
a degraded RHR system capability condition.
3.
Question No. 90 Understanding of the technical specification requirements l
during a degraded incore thermocouple indication
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condition.
4.
Question No.100 Understanding of the conditions and actions required to a l
failed surveillance test of a radiation monitor following a liquid waste discharge conditio _. _... _.
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Enclosure 3 l
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Simulation Facility Reoort
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Facility Licensee: Point Beach Nuclear Plant, Units 1 & 2 l
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Facility Licensee Docket No: 50-266; 50-301 l
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Operating Tests Administered: June 22 - 25,1998 The following documents observations made by the NRC examination team during the June 1998, initial license examination. These observations do not constitute audit or inspection findings and are not, without further verification and review, indicative of non-compliance with i
10 CFR 55.45(b). These observations do not affect NRC certification or approval of the simulation facility other than to provide information which may be used in future evaluations.
No licensee action is required in response to these observations, i
j During the conduct of the simulator portion of the operating tests, the following item was observed:
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ITEM DESCRIPTION
Simulator failure during The simulator validation had established three setup JPM Performance snaps. During the examination week, one simulated JPM setup crashed following the initialization with no cause determined.
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Control Bank D Group The control bank D rod group indication (LED) had to be Counter manually reset following each setup re-initialization due to
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faulty indication (off by as much as 100 steps).
Main Generator The main generator synchronous scope does not contain a Synch Scope
" Green Band" marking consistent with the main control l
room indication and approved procedure for synchronizing l
with the grid.
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DC Power Sliders The simulator control room does not model DC power l
sliders which are required to be operated upon receiving a i
reactor coolant pump sump oil alarm to determine actual i
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location of the problem, as noted in Alarm Response Book j
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U.S. Nuclear Regulatory Commission Site-Specific Written Examination Appi! cant information Name:
MASTER EXAMINATION Region:
lli Date:
JUNE 22, 1998 Facility / Unit: POINT BEACH l
License Level:
SRO Reactor Type:
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Start Time:
Finish Time:
Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. The passing grade requires a final grade of at least 80.00 percent. Examination papers will be collected four hours after
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the examination starts.
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Applicant Certification
All work done on this examination is my own. I have neither given nor received aid.
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Applicant's Signature i
Results
Examination Value Points
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Applicant's Score Points Applicant's Grade Percent
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SENIOR REACTOR OPERATOR Page 2 ANSWER SHEET Multiple Choice (Circle or X your choice)
If you change your answer, write your selection in the blank.
001 abcd 021 abcd 041 abcd 002 a b c d 022 a b c d 042 a b c d 003 a b c d 023 a b c d 043 a b c d 004 abcd 024 a b c d 044 a b c d 005 a b c d 025 a b c d 045 a b c d 006 a b c d 026 a b c d 046 a b c d 007 a b c d 027 a b c d 047 a b c d 008 a b c d 028 a b c d 048 a b c d l
C09 a b c d 029 a b c d 049 a b c d 010 a b c d 030 a b c d 050 a b c d
011 abcd 031 abcd 051 abcd 012 a b c d 032 a b c d 052 a b c d 013 a b c d 033 a b c d 053 a b c d 014 abcd 034 a b c d 054 a b c d 015 a b c d 035 a b c d 055 a b c d 016 a b c d 036 a b c d 056 a b c d l
l 017 a b c d 037 a b c d 057 a b c d 018 a b c d 038 a b c d 058 a bcd 019 a b c d 039 a b c d 059 a b c d 020 a b c d 040 a b c d 060 a bcd
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i SENIOR REACTOR OPERATOR Page 3
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ANSWER SHEET Multiple Choice (Circle or X your choice)
If you change your answer, write your selection in the blank.
061 abcd 081 abcd j
062 a b c d 082 a b c d 063 a b c d 083 a b c d 064 a b c d 084 a b c d 065 a b c d 085 a b c d
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066 a b c. d 086 a b c d 067 a b c d 087 a b c d 068 a b c d 088 a b c d 069 a b c d 069 a b c d -
070 a b c d 090 a b c d
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071 abcd 091 abcd 072 a b c d 092 a b c d 073 a b c d 093 a b c d 074 a b c d 094 a b c d 075 a b c d 095 a b c d i
076 a b c d 096 a b c d 077 a b c d 097 a b c d 078 a b ' c d 098 a b c d 079 a b c d 099 a b c d 080 a b. c d 100 a b c d (""""" END OF EXAMINATION "**""")
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l SENIOR REACTOR OPERATOR Page 4 WRITTEN EXAMINATION GUIDELINES 1.
[ Read Verbat/mJ 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.
2.
To pass the examination, you must achieve a grade of 80.00 percent or greater. Every question is worth one point.
3.
For an initial examination, the time limit for completing the examination is four hours.
For a requalification examination, the time limit for completing both sections of the examination is three hours. If both sections are administered in the simulator during a single three-hour period, you may retum to a section of the examination that was already completed or retain both sections of the examination until the allotted time has expired.
4.
You may bring pens, pencils, and calculators into the examination room. Use only black ink or pencils to ensure legible copies.
5.
Print your name in the blank provided on the examination cover sheet and the answer sheet. You may be asked to provide the examiner with some form of positive identification.
6.
Mark your answers on the answer sheet provided and do not leave any question blank.
Use only the paper provided and do not write on the back side of the pages. If you decide to change your original answer enter the desired answer in the spaces at the end of the answer row.
7.
If the intent of a question is unclear, ask questions of the NRC examiner or the designated facility instructor only.
8.
Restroom trips are permitted, but only one applicant at a time will be allowed to leave.
Avoid all contact with anyone outside the examination room to eliminate even the appearance or possibility of cheating.
9.
When you complete the examination, give the examination answer sheets to the NRC examiner. Remember to sign the statement on the examination cover sheet indicating that the work is your own and that you have neither given nor received assistance in completing the examination. Leave all other materials at your desk. The scrap paper will be disposed of immediately after the examination.
10.
After you have turned in your examination, leave the examination area as defined by the proctor or NRC examiner. If you are found in this area while the examination is still in progress, your license may be denied or revoked.
11.
Do you have any questions?
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SENIOR REACTOR OPERATOR Page5 i
j QUESTION: 001 (1.00)
What function do the CRDM vent fans perform during Natural Circulation Cooldown?
a.
Attempt to prevent voiding in reactor vessel head region.
b.
Promote natural circulation flow through the RCS.
l-c.
Aid increased flow through the reactor vessel head region.
d.
Reduce stresses on the reactor vessel head due to uneven cooldown.
l OUESTION: 002 (1.00)
l The following conditions exist on Unit 1:
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Reactor trip occurred 1 minute ago Main feedwater FAILED to isolate y
Containment humidity is increasing i
1R "A" SUR: 0.00 dpm IR "B" SUR: - 0.15 dpm Which ONE of the following actions will be taken FIRST following a manual safety injection actuation?
(ASSUME: All other equipment performed as designed.)
a.
Transition to EOP-2, " Faulted Steam Generator Isolation."
b.
Transition to CSP-S.1, " Response to Nuclear Power Generation /ATWS."
c.
Shift FW controllers to manual then close regulating control and bypass valves.
d.
Manually open containment ventilation cooler outlet emergency FCVs.
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SENIOR REACTOR OPERATOR Page 6 i
QUESTION: 003 (1.00)
With the RCS at normal operating pressure and temperature, what is the condition of the steam entering the PRT if a PORV opens? (ASSUME: PRT is at 100*F,5 psig; an ideal thermodynamic process)
a.
Superheated steam at 635'F.
b.
Superheated steam at 313*F.
c.
Saturated steam-water mixture at 213*F.
d.
Saturated steam-water mixture at 228'F.
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l QUESTION: 004 (1.00)
The following plant conditions exist:
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PZR Level Channel select switch is in the NORMAL position RCS Temperature is 550*F j
. Plant startup is in progress Which ONE of the following statements describes the plant response to a failure of pressurizer level channel L427 in the LOW direction? (ASSUME: No operator action)
a.
Letdown isolation occurs; pressurizer level will slowly increase.
b.
Charging pump speed will decrease to minimum; pressurizer level will slowly decrease.
c.
Charging pump speed will remain the same; pressurizer heaters will turn off; pressurizer level will remain the same, d.
No system response occurs since this is a non-controlling channel failure.
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QUESTlON: 005 (1.00)
Plant startup is its progress. If main condenser vacuum were to inadvertently decrease, in accordance with AO?-5A, " Loss of Condenser Vacuum," which ONE of the following conditions would require the operator to trip the reactor?
a.
AT POWER TRIP BLOCKED P-7 status light - LIT Condenser pressure 2.5" Hg Absolute.
b.
AT POWER TRIP BLOCKED P-7 status light - LIT Condenser pressure 3" Hg Absolute.
c.
AT POWER TRIP BLOCKED P-7 status light - CLEAR Condenser pressure 3" Hg Absolute.
d.
AT POWER TRIP BLOCKED P-7 status light - CLEAR Condenser pressure 4"
)
Hg Absolute.
<
i QUESTION: 006 (1.00)
The following conditions exist on Unit 1:
Pressurizer levelis 0 Pressurizer pressure is 1200 psig Containment pressure is ~10 psig Tcold is ~ 380*F SGs levels ~ 39%
SGs pressures ~ 300 psig.
,
Where is the location of the leak?
a.
On a RCS cold leg.
b.
In a Steam Generator tube.
c.
On a Main Steam Line inside containment.
d.
On a feedwater line inside containment upstream of the feedwater check valve.
!
i
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SENIOR REACTOR OPERATOR Page 8
QUESTION: 007 (1.00)
The following conditions exist on Unit 1:
Reactor power 20%
PZR pressure 1985 psig PZR level 25%
8 A' Train of Secondary equipment is aligned for operation
' An 86 relay lockout condition occurs on Bus 1A02,4.16 KVAC. Which ONE of the following statements describes the affect on continued plant operations?
a.
The reactor can be maintained in the current condition with power limited to 50%.
b.
The reactor will automatically trip due to the opening of a RCP breaker, c.
The reactor must be placed in hot shutdown by the operator due to the loss of a RCP.
d.
The reactor will automatically trip due to a low SG Level from a loss of feedwater j
flow.
QUESTION: 008 (1.00)
Which ONE of the following does NOT actuate the 86-TG01 or 86-X01 lockout relays?
!
a.
Bus under frequency.
b.
Main generator ground.
c.
d.
Main transformer ground overcurrent.
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SENIOR REACTOR OPERATOR Page 9
,
QUESTION: 009 (1.00)
{
\\
The following cordions exist on Unit 1:
A LOCA has occurred Transfer to containment sump recirculation is required RCS pressure is approximately 50 psig What is the approximate SI pump TOTAL flow indicated on the main control board and how will this value change following transfer of BOTH trains of ECCS to containment sump recirculation?
Total Flow Flow Change a.
700 gpm Decrease b.
700 gpm increase c.
1800 gpm Decrease d.
1800 gpm increase QUESTION: 010 (1.00)
When would the Containment Spray pumps start in relation to a Containment Spray actuation signal that was received two seconds after an AUTO Si signal?
a.
Immediately after the Si pumps.
l b.
Coincident with the starting of the Containment Accident Fans.
c.
Between the start of the AFW pumps and the Service Water pumps.
d.
Between the start of the RHR pumps and the AFW pumps.
_ - - - _ _ _ _ _. -.. _ _ _
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SENIOR REACTOR OPERATOR Page 10 QUESTION: 011 (1.00)
Unit 1 is at 100% power with no other equipment OOS. Diesel Generator G02 surveillance test commenced on June 15 at 1000 hrs.
The following time line of events occurred:
June 151030 hrs DG started for a one hour run.
June 151100 hrs DG tripped due to failure of fuelinjector.
June 171450 hrs Spare fuelinjector located.
.
June 210200 hrs Post maintenance test of DG revealed a crack in the governor casing. No replacement is available for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
,
When must the Unit 1 shutdown commence due to required Tech Spec LCO Actions?
l a.
June 18 at 0200 hrs l
b.
June 22 at 1000 hrs c.
June 22 at 1100 hrs l
d.
June 28 at 0200 hrs
i l
i l
'
QUESTION: 012 (1.00)
According to 10 CFR 20, what is the definition of Total Organ Dose Equivalent (TODE)?
!
i a.
It is the sum of the Shallow Dose Equivalent, Whole Body (SDE, WB) and the
~
Committed Effective Dose Equivalent (CEDE).
b.
It is the sum of the Deep Dose Equivalent (DDE) and the Committed Dose Equivalent (CDE).
c.
It is the sum of the Deep Dose Equivalent (DDE) and the Committed Effective Dose Equivalent (CEDE).
j i
d.
It is the sum of the Shallow Dose Equivalent, Max Extremity (SDE, ME) and the Deep Dose Equivalent (DDE).
_ _________________
A
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SENIOR REACTOR OPERATOR Page 11 QUESTION: 013 (1.00)
Previously, an operator had the following annual exposure history:
210 mrem Committed Effective Dose Equivalent (CEDE)
45 mrem Shallow Dose Equivalent (SDE)
33 mrem Committed Dose Equivalent (CDE)
28 mrem Today, the operator was required to make two containment entries which resulted in the following:
Entry 1:
Gamma dose 52 mrem; Neutron dose 24 mrem Entry 2:
Gamma dose 124 mrem After the second containment entry, how much radiation exposure is administratively available for the year for this operator?
a.
1545 mrem b.
1569 mrem c.
1614 mrem d.
1745 mrem QUESTION: 014 (1.00)
Which ONE of the following is the MAXIMUM annual Total Exposure Dose Equivalent (TEDE)
limit an individual can receive from PLANNED SPECIAL EXPOSURE?
a.
2 rem b.
5 rem c.
10 rem d.
25 rem
i
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SENIOR REACTOR OPERATOR Page 12 QUESTION: 015 (1.00)
An event occurred at 0800 that was classified as a Site Emergency. The plant evacuation alarm was actuated at the time of the classification.
Which ONE of the following is the MAXIMUM time by which accountability must be completed?
a.
0815 b.
0830 c.
0845 d.
0900 QUESTION: 016 (1.00)
Given the following Unit 1 plant conditions:
A Control Bank D rod was dropped and recovered.
The Pulse to Analog Converter was NOT reset as required by AOP-6A, " Dropped Rod."
,
Which ONE of the following will occur on the next rod movement?
a.
If control rods are inserted, the Rod Insertion Limit Alarm will be received at a lower rod position than required.
b.
If control rods are withdrawn, Overtemperature AT will NOT stop Control Bank D j
withdrawal when required.
c.
If control rods are inserted, Bank C control rods will begin insertion at a lower value of Control Bank D position.
d.
If control rods are inserted, Bank C control rods will begin insertion at a higher value of Control Bank D position.
l
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t SENIOR REACTOR OPERATOR Page 13
.
QUESTION: 017 (1.00)
Which ONE of the following is the MAXIMUM allowable time 'o MANUALLY actuate AFW
'
following an ATWS with a loss of feedwater? (ASSUME: NO AMSAC aduation)
a.
30 seconds b.
60 seconds c.
120 seconds j
d.
180 seconds
'
i QUESTION: 018 (1.00)
,
l Which ONE of the fo lowing is the reason the Containment Purge Supply and Exhaust Valves
,
!
are required to be locked closed during operations at power?
\\
'
a.
The related piping systems outside containment are NOT seismically qualified.
.
b.
The valves are NOT seismically qualified to operate during a design basis earthquake.
!
c.
The valve actuators do NOT have class 1E penetration conductor overcurrent protection devices.
d.
The valves' capability to close during a design basis loss-of-coolant accident has NOT been demonstrated.
i l
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SENIOR REACTOR OPERATOR Page 14 QUESTION: 019 (1.00)
Given the following Unit 2 plant conditions:
The plant is at 100% power.
Pressurizer pressure is in AUTOMATIC control.
The Pressurizer Pressure Channel Defeat switch is in the normal position.
i No operator action is taken.
]
Which ONE of the following actions occurs when pressurizer pressure transmitter PT-431 fails LOW?
a.
Pressurizer pressure INCREASES, resulting in a HIGH pressure reactor trip.
l b.
Pressurizer pressure DECREASES, resulting in a LOW pressure reactor trip.
c.
Pressurizer PORV (RC-430) cycles to maintain pressure below the reactor trip
!
setpoint.
d.
Pressurizer heaters and spray valves operate normally to maintain pressurizer pressure.
j QUESTION: 020 (1.00)
During a normal reduction in power, which ONE of the following is the reason that additional pressurizer heaters should be energized?
a.
. Allow an increased ramp rate for the power change.
b.
. Equalize the reactor coolant system and pressurizer boron concentrations.
c.
Maintain pressurizer pressure in normal operating ranne during the power change.
d.
Ensure positive pressurizer pressure control is established prior to starting the power change.
j
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SENIOR REACTOR OPERATOR Page 15 QUESTION: 021 (1.00)
Which ONE of the following is the MINIMUM number of shifts, per 10 CFR55, on which you must actively perform operator or senior operator duties to maintain your license in an ACTIVE status? (ASSUME 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> shifts.)
a.
5 shifts per calendar quarter b.
5 shifts per calendar year c.
7 shifts per calendar quarter i
d.
7 shifts per calendar yeat QUESTION: 022 (1.00)
i l _-
In procedure OP-5A, " Reactor Coolant Volume Control," there is a PRECAUTION that states:
"Do not secure letdown flow without also securing charging flow..."
i Which ONE of the following statements describes why charging flow should also be isolated?
(ASSUME: All systems are in a normal at power lineup.)
l a.
VCT level will decrease until charging pump suction shifts to the RWST.
L b.
Reduce thermal shock on the Non-Regenerative Heat Exchanger.
c.
VCT level will decrease causing possible damage to the charging pumps.
d.
Reduce thermal shock on the charging penetration into the RCS.
l~
I l
L
,
.)
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SENIOR REACTOR OPERATOR Page 16 I
QUESTION:023 (1.00)
Given the following Unit 1 plant conditions:
A loss of all AC power has occurred.
ECA-0.0, " Loss of All AC Power,"is in effect.
Per ECA-0.0, certain Engineered Safeguards equipment control switches are placed in Pull-out.
Which ONE of the following events is prevented by this switch alignment?
a.
An uncontrolled depressurization of the RCS.
l b.
An uncontrolled start of large loads on safeguards AC buser.
c.
An uncontrolled cooldown of the RCS and possible reactor restart.
d.
An uncontrolled use of watar that may be needed for long term cooldown.
!
QUESTION: 024 (1.00)
Given the following Unit 1 plant conditions:
Unit is operating at 75% steady state power.
All systems are in automatic control.
The "A" SG atmospheric steam dumps fails open.
Which ONE of the following describes the plant response to this condition? (ASSUME: No operator action is taken.)
a.
Turbine load decreases by 5%, reactor power remains stable at 75%.
b.
Control rods withdraw and reactor power increasing to 80% where it stabilizes.
c.
Control rods initially insert then withdraw to maintain reactor power at 75%.
d.
Turbine governor valves open in response to lower steam header pressure to increase turbine load to 80%.
>
l
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SENIOR REACTOR OPERATOR Page 17 QUESTION: 025 (1.00)
What will cause the "A" SG Feedwater Regulating Valve (CS-466) to OPEN when in Auto?
a.
Low level (17%) in the "B" SG.
b.
Auto start of P-38A auxiliary feedwater pump.
c.
Turbine trip with Tavg greater than 554*F.
d.
High-High steam flow coincident with a low steamline pressure of 500 psig.
QUESTION: 026 (1.00)
Unit 1 is shutting down and at 6% power when intermediate range channel N36 fails high.
Which of the following statements best describes how this failure affects the reactor shutdown and subsequent operation of the nuclear instrumentation system?
a.
The reactor will NOT trip, and source range NI's will have to be manually re-energized.
b.
The reactor will trip on high IR flux, and source range NI'.s will have to be manually re-energized.
c.
The reactor will NOT trip, and source range N!'s will re-energize when N35 reaches the proper setpoint.
d.
The reactor will trip on high IR flux, and source range NI's will re-energize when N35 reaches the proper setpoin l l
l SENIOR REACTOR OPERATOR Page 18 CUESTION: 027 (1.00)
Given the following Unit 1 plant conditions:
Unit is at 50% power.
Control rods are in AUTOMATIC.
Which ONE of the following instrument malfunctions would result in a CONTINUOUS rod withdrawal?
a.
PT-485 fails HIGH.
b.
Loop "A" Thot RTD fails HIGH.
c.
d.
Power range channel N-42 fails LOW.
QUESTION: 028 (1.00)
Which ONE of the following describes the response of Feedwater Regulating Valves following loss of a RED or BLUE instrument power to FRV controller?
a.
Fail open; manual control will NOT be available to regain control.
b.
Fail open; but manual control will be available to regain control.
c.
Fail closed; manual control will NOT be aval:Ble to regain control.
d.
Fail closed; but manual control will be available to regain control.
i
i- '
l SENIOR REACTOR OPERATOR Page 19 QUESTION: 029 (1.00)
Which ONE of the following is the reason or basis for establishing SI Core Deluge within 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> after a LOCA, in accordance with EOP-1.2,"Small Break LOCA Cooldown and Depressurization?"
i i
a.
Aids in preventing boron precipitation.
l'
l b.
Entrains additional hydrogen in solution.
I l
c.
Enhances reflux cooling in the steam generators.
i d.
Eliminates non-condensable gases in the reactor head area.
QUESTION: 030 (1.00)
l Which ONE of the following is an action required by AOP-6E, "Altemate Boration/ Loss of Shutdown Margin" when performing an EMERGENCY BORATION?
a.
Stop letdown flow.
b.
Start a boric acid transfer pump.
c.
Maximize c.harging flow to the RCP seals.
d.
Set the required amount of boric acid on the blender.
i QUESTION: 031 (1.00)
Which ONE of the following conditions would require an IMMEDIATE reactor trip during a Loss of CCW, according to AOP-9B, " Component Cooling System Malfunction?"
i a.
No CC pumps running with surge tank level at 8% and decreasing.
l b.
One CC pump running with surge tank level at 22% and decreasing.
c.
" Component Cooling HX Outlet Temp High" alarm is received.
- d.
" Component Cooling Pump Discharge Pressure Low" alarm is received.
l
]
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i SENIOR REACTOR OPERATOR Page 20
i-(
QUESTION: 032 (1.00)
!
Which ONE of the following locations / equipment is protected by a Halon System?
,
a.
Diesel Generator rooms i
b.
Hydrogen Seal Oil package
,
c.
Service Water Pump area d.
Auxiliary Feed Pump room QUESTION: 033 (1.00)
A LOCA has occurred and both RCPs have been re-started per CSP-C.1, " Response to Inadequate Core Cooling."
)
.
Given the below list of criteria:
1. Core cooling provided by low or high head Si
. II. Narrow range reactor vessel level greater than 25 feet i
111. Any RCS hot leg loop less than 350*F j
IV. Core exit thermocouple less than 1200*F-Which ONE of the following gives the three conditions from the above list that allow securing of the RCPs per CSP-C.17 a.
I,11 and lll b.
I,11 and IV c.-
1, Ill and IV.
d.
II, ll1 and IV.
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SENIOR REACTOR OPERATOR Page 21
QUESTION: 034 (1.00)
j l
Which ONE of the following conditions will result in the AUTOMATIC closure of Waste Condensate Discharge valve WL-0187 i
a.
Waste Condensate Pumps trip.
b.
Loss of Circulating Water flow.
c.
High level alarm on the Waste Holdup Tank.
d.
An alarm condition on Radiation Monitor Channel RE-218.
- QUESTION: 035 (1.00)
Given the following Unit 1 plant conditions:
Unit is in Cold Shutdown with RHR Cooling in progress.
The RCS is solid.
RHR flow is lost and CANNOT be restored.
All other systems and components are available.
Which ONE of the following methods of cooling will be utilized to remove the core decay heat?
a.
Feed the RCS with Safety injection; use letdown to remove decay heat.
b.
Start a charging pump with flow through an RHR heat exchanger, and initiate Hot Leg injection.
!
c.
Start a charging pump with flow through an RHR heat exchanger, and initiate i
Cold Leg injection.
,
d.
Feed a SG using an AFW pump, and bleed steam through the respective SG atmospheric t'eam dump.
l.
l
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l SENIOR REACTOR OPERATOR Page 22 l
QUESTION: 036 (1.00)
Which ONE of the following criteria / conditions would require shutdown of the reactor during the performance cf AOP-1 A, " Reactor Coolant Leak", Attachment A, SG Tube Leakage Actions?"
a.
PZR pressure drops below 1925 psig.
b.
Steam Generator tube leakage exceeds 50 gallons per day.
c.
Steam generator activity exceeds 0.2 microcurie per gram Dose Equivalent l-131.
d.
RCS leakage to the secondary side exceeds 500 gallons per day.
)
QUESTION: 037 (1.00)
Which ONE of the following statements explains why AFW flowrate is procedurally restricted to 100 gpm when recovering steam generator (SG) level if the level has fallen below 55 inches on the wide range indication?
a.
To maintain adequate water inventory to +h: SG downcomer.
b.
To prevent reactor restart from an excessive cooldown.
c.
To minimize thermal stresses to S3 components.
d.
To prevent exceeding reactor vessel cooldown rate limit.
_
,......
.
.
l'
SENIOR REACTOR OPERATOR Page 23 -
j.'
QUESTION: 038 (1.00)
Which ONE of the following actions should be taken if a loss of DC control power occurs on a l
single 4160 V breaker due to a blown fuse, according to AOP-0.0, " Vital DC System Malfunction?"
!
a.
Verify no grounds are present and transfer control power to backup.
b.
Request support from maintenance to restore power from alternatb.
c.
Replace both fuses of affected set; a one-time changeout is allowed.
..
d.
Determine the cause for the blown fuse prior to replacing the affected fuse.
QUESTION: 039 (1.00)
,
Which ONE of the following will eventually result in a loss of one or both units due to a j
!
i
- decrease in instrument air header pressure according to AOP-5B, " Loss of Instrument Air?"
a.
Pressurizer PORV's opening.
b.
Main steam stop valves closing.
c.
Pressurizer Spray valves fail open.
d.
The air header cross connects remain open.
I l
,
.
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SENIOR REACTOR OPERATOR Page 24 QUESTION: 040 (1.00)
The following conditions exist on Unit 1:
RCS Pressure 1700 psig Containment Pressure 22 psig Loss of Offsite Power
, Occurred Equipment list:
1. Charging pumps 6. RHR pump 2. CCW pump 7. Motor Driven AFP 3. SI pump 8. Service water pump 4. Containment fans 9. Service air compressor 5. Containment spray pump Wh!ch ONE of the following is the sequential order for loading Emergency Diesel Generator given the above conditions?
l a.
3-6-7-8-4 b.
2-4-8-7-9 c.
2-6-4-8-5 d.
3-2-7-6-1 QUESTION: 041 (1.00)
Unit 1 is operating at 40% power with normal system lineups. Which ONE of the following describes the response of charging flow to a T hot RTD failing HIGH, with NO operator action?
a.
Flow remains the same.
I b.
Flow decreases to its minimum value.
l c.
Initially flow increases, then decreases back to its original value, d.
Initially flow decreases, then increases back to its original value.
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SENIOR REACTOR OPERATOR Page 25 QUESTION: 042 (1.00)
Which ONE of the following conditions on pressurizer level transmitter, LT-428, would result in an increased levelindication?
a.
A leak in the reference leg.
b.
A decrease in containment temperature.
c.
Pressurizer liquid temperature increases.
d.
An increase in containment pressure to 0.3 psig; temperature remains constant.
QUESTION: 043 (1.00)
Which ONE of the following conditions would be indicated by an illuminated status light on the
"Sl/ Spray Ready Panel" with the Unit at FULL POWER 7 A component...
a.
has lost DC control power.
b.
has lost AC control power.
c.
is in its normal alignment condition.
d.
is in an abnormal alignment condiion.
I QUESTION: 044 (1.00)
l A spent fuel element has been damaged during fuel handling operations in the Spent Fuel Pool.
. Which ONE of the following is an initial action required by AOP-8C, " Fuel handling Accident in Primary Auxiliary Building?"
a.
Maximize SFP purification flow.
I'
b.
Verify SFP leak detection system normal.
c.
Ensure normal SFP cooling is available.
d.
Initiate a limited evacuation of the primary auxiliary building.
l
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_ _ _. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _. _ _ _ _ _ _ _ _ _ _ _ _ _ _.. _
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SENIOR REACTOR OPERATOR Page 26 QUESTION: 045 (1.00)
The reactor is at 60% power when N42 begins to SLOWLY drift HIGH and eventually reaches the upper peg. All systems are in automatic, no operator action is taken, and turbine load remains constant.
Which ONE of the following describes the response of the rod control system to this failure?
a.
Rods will withdraw to raise power since N42 is driving the auctioneered nuclear power signal high.
b.
Rods will withdraw restoring Tavg in accordance with programmed Tavg consistent with the N42 reading.
c.
Rods will not move because Tavg and turbine power are remaining constant with no power rate mismatch.
d.
Rods will insert from the turbine runback which occurs when the mismatch between N42 and the other power range channels is 2.5%.
QUESTION: 046 (1.00)
Which ONE of the following would indicate that a rod is "misal;gned", with the associated bank demand position greater than 215 steps, according to AOP-68, " Stuck Rod or Malfunctioning Position Indication?"
A misalignment of:
a.
5 steps b.
10 steps c.
12 steps d.
24 steps i
I i
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SENIOR REACTOR OPERATOR Page 27 QUESTION: 047 (1.00)
>
Which ONE of the following describes the purpose of the back draft dampers installed in the Containment. Air Recirculation System?
a.
Prevent backflow in a cooling unit in the event of fire in containment.
b.
Serve as explosion dampers preventing duct work collapse during an accident.
c.
Prevent unit air backflow when the accident fan is running and the cooling fan is not.
,
d.
Serve as a system air backflow damper in idle cooling units (both accident and cooling fans secured).
'
l QUESTION: 048 (1.00)
Which ONE of the following must approve the raising of an area monitor's Alert alarm to a value j
that is still below the High alarm setpoint, in order to clear a " nuisance" alert alarm, according to l
OM 4.1.7, "RMS Alarm Setpoint & Response Book?"
a.
Control operator, b.
D. cy Shift Superintendent.
,
!
c.
Superintendent - l & C.
!
d.
Health physics technician.
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SENIOR REACTOR OPERATOR Page 28 QUESTION: 049 (1.00)
The following conditions exist on Unit 1:
Reactor power 80%
l The lower detector for N-43 is open circuited (due to a broken cable)
Which ONE of the following is acceptable for determining core quadrant power tilt under the above conditions?
'
a.
Plant Process Computer.
b.
Incore Movable Detectors.
c.
Manual calculations using operable excore detectors.
d.
Manual calculations using estimated current for N-43 lower detector.
QUESTION: 050 (1.00)
Performance of EOP-1.2,"Small Break LOCA Cooldown and Depressurization," is in progress.
What is the PRIMARY reason for starting both control rod shroud fans after depressurizing the RCS7 a.
To provide adequate cooling for the NIS detectors.
b.
To provide adequate cooling for the CRDMs.
c.
To cool down the upper head region of the reactor vessel.
d.
To reduce containment pressure and humidity.
QUESTION: 051 (1.00)
Which ONE of the following is used as the reactor power input to the Rod Insertion Limit (RIL)
computer?
Average AT a.
b.
Auctioneered High Tavg
l I
c.
Turbine impulse Pressure d.
Auctioneered High Power Range N1
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SENIOR REACTOR OPERATOR Page 29 QUESTION: 052 (1.00)
]
Which ONE of the following describes the reason why the start of the containment spray pumps is delayed for 10 seconds upon containment spray actuation?
a.
To allow for containment spray pump discharge valve positioning.
b.
To permit the spray additive tank discharge valves sufficient tims to open, c.
To prevent a brief containment pressure transient from actua*ing containment spray.
d.
To allow the operator time to override the pumps if signal is spurious (prevent inadvertent spray discharge to containment).
l QUESTION: 053 (1.00)
Which ONE of the following conditions will result in an automatic trip of all operating Condensate Pumps?
a.
Steam generator level at 73%
b.
Containment pressure at 5.2 psig.
c.
Condenser hotwell level at 10 inches.
d.
Steam generator feed pump suction pressure at 179 psig.
QUESTION: 054 (1.00)
Which ONE of the following conditions / signals will result in an automatic start of the Turbine -
~ Driven Auxiliary Feedwater Pump?
,
!
a.
Safety injection signal.
b.
Loss of 4.16 KV buses 1 A01 and 1 AO2.
c.
Trip of both Motor Driven AFW pumps, d.
Low-low steam generator (SG) level (15%) in "A" S __ __ _ _____
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SENIOR REACTOR OPERATOR Page 30
QUESTION: 055 (1.00)
The following plant conditions exist on Unit 1:
j i
RCS drained down Reactor vessel level (Ll-447/447A) 50% and decreasing
Which ONE of the following combinations are available to refill the reactor coolant system?
a.
Boric acid transfer pump; chemical drain pump.
b.
Spent fuel pool pump; spent fuel pool skimmer pump.
c.
Reactor coolant drain tank pump; spent fuel p sol skimmer pump.
d.
Charging pumps; refueling water circulating water pump from RWST.
i QUESTION: 056 (1.00)
Which ONE of the following statements describes how plant operations are affected if Loop A RCS Wide Range Pressure instrument, PT-420, fails HIGH during Low Temperature Overpressure Mitigation System operation?
a.
Pressurizer PORV RC-430 opens.
i b.
Pressurizer PORV RC-431C opens.
c.
Both pressurizer relief valves open.
d.
No effect will be seen.
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SENIOR REACTOR OPERATOR Page 31 l
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QUESTION: 057 (1.00)
A 100 gpm breach has occurred in the SFP. Which ONE d the following states the preferred l
makeup source and the reason why it is preferred, per AOP-8F, " Loss of Spent Fuel Pool l
Cooling?"
!
a.
Service Water, most abundant source.
!
b.
Unit 2 RWST, higher concentration of boric acid.
'
!
c.
Fire water, higher capacity than other sources.
d.
DI water, less impurities than other sources.
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QUESTION: 058 (1.00)
-]
All of the following actions are DIRECTLY associated with an anticipated transient without scram mitigating system actuation circuit (AMSAC) trip, EXCEPT:
a.
b.
c.
opening steam supplies to turbine-driven AFP (P-29).
,
I d.
auto start motor-driven AFPs (P-38A & B).
- QUESTION: 059 (1.00)
,
Which ONE of the following fuel handling components does NOT require air / nitrogen for operation?
a.
Fuel transfer cart.
i b.
RCCA change fixture carriage.
c.
Manipulator crane gripper assembly.
I d.
Control rod drive shaft unlatching tool.
'
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SENIOR REACTOR OPERATOR Page 32
QUESTION: 060 (1.00)
The following plant conditions exist:
Steam generator A steam flow:
0.0005 E6 lbm/hr Steam generator B steam flow; 0.6 E6 lbm/hr Steam generator A level:
70 %
Steam generator B level:
10%
Turbine driven AFW pump:
running Motor driven AFW pumps:
running
,
Tavg:
516*F
'
Containment pressure:
5 psig l
SI:
actuated l
If no operator action has been taken, which ONE of the following indicates the status of the main steam isolation valves?
l MSIVA MSIV B l
a.
.open open b.
open-shut c.
shut open l
d.
shut shut
.
QUESTION: 061 (1.00)
Which ONE of the following would be used to verify a RMS channel's high radiation alarm is operable (i.e.' has NOT been disabled)?
a.
The channel's " data" file.
b.
The channel's " parameter" file.
c.
The " normal" light is illuminated.
d.
The " alarm disabled" light is deenergized.
..
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SENIOR REACTOR OPERATOR Page 33 QUESTION: 062 (1.00)
Which ONE of the following describes the interlock associated with the RHR discharge to Containment Spray pump suction MOVs (RH-871 A&B) in order to OPEN the valves?
a.
RHR pressure must be above 225 psig.
b.
RHR temperature must be below 200*F.
c.
The respective suction MOVs from the RWST (RH-870A&B) must be shut and RHR pressure less than 210 psig.
d.
The respective suction hydraulic valves from the containment sump (RH-850A&B) must be shut.
QUESTION:-063 (1.00)
l During operation at 100% power, impulse pressure channel (PT-486) fails LOW due to a loss of instrument power.
Which ONE of the following describes the response of the condenser steam dump control system to this failure and why?
a.
The dump valves modulate open due to a Tavg/ Tref deviation generated by the loss of impulse pressure.
b.
The loss of impulse pressure would only have an affect on the steam dump if it was operating in the " pressure" mode.
c.
The steam dump valves trip open on a turbine trip signal being generated by the loss of impulse pressure.
d.
The steam dump valves remain closed but now are " armed" due to a loss of load condition being sensed.
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SENIOR REACTOR OPERATOR Page 34 l
QUESTION: 064 (1.00)
.
. While reducing power, the operator observes a single control rod that appears to be misaligned l
. from its bank. How can the operator determine if the RCCA is actually misaligned or if the position indication system has a malfunction?
'
a.
Check if any other rods appear to be misaligned. If not, an actual rod
,
'
misalignment is indicated.
!
b.,
Check the " Quadrant Power Tilt" within allowable limits. If not within limits, an actual rod misalignment is indicated.
c.
Move the affected bank and see if the indicated misalignment worsens. If it does not _ worsen, a position indication malfunction is indicated.
d.
Move the affected bank and compare Tavg response to the equivalent reactivity what would be expected from the differential bank worth. If it does not respond as expected, a position indication malfunction is indicated.
QUESTION: 065 (1.00)
A steam line break has occurred outside containment. The reactor has tripped, Si has initiated, CSP-P.1 has been entered and both RCPs are tripped. Conditions for restart of a RCP have been met.
Starting one RCP will result in...
a.
a pressure spike that will worsen a PTS condition.
b.
an increase in the RCS cooldown rate, thus increasing the potential of PTS.
c.
providing mixing of the Si flow, thus alleviating the potential for PTS.
d.
providing mixing of the Si flow, thus alleviating the potential for boron precipitation.
,
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SENIOR REACTOR OPERATOR Page 35 QUESTION: 066 (1.00)
Operations personnel have entered AOP-1B " Reactor Coolant Pump Malfunction." Which ONE of the following conditions would require a RCP to be secured?
a.
Shaft vibration is 12 mils and stable.
b.
Seal leakoff outlet temperature is 195'F and rising.
c.
No.1 sealleakage is 4 gpm and stable.
d.
Motor bearing temperature is 85'F and lowering.
QUESTION: 067 (1.00)
The chemistry department has determined that high coolant activity exists in the Unit 1 RCS.
Unit 1 startup is in progress with temperature being maintained at - 540'F by condenser steam dumps.
Which ONE of the following statements describes the effect of placing the cation bed demineralized in service?
l
The placernent in service of the cation bed dernineralizer will...
a.
cause an increase in the RCS pH.
b.
NOT have any affect on the RCS coolant.
c.
cause a decrease in RCS coolant activity.
d.
cause a decrease in the RCS boron concentration.
!
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SENIOR REACTOR OPERATOR Page 36 QUESTION: 068 (1.00)
Adverse containment conditions are determined by...
a.
calculating the containment saturation pressure..
b.
the operator, taking in5 account the containment temperature.
c.
the saturation temperature in RCS versus containment pressure.
d.
the operator, through the use of containment pressure and/or radiation condition.
QUESTION: 069 (1.00)
The following conditions exist on Unit 1:
l Reactor trip and Safety injection occurred.
SG high level caused a FWisolation.
Tavg has returned to no-load condition.
SG levels have returned to within NORMAL band.
The reactor operator desires to restore feed to the SGs using the FW Regulating Bypass valves.
I Which ONE of the following conditions have to be met to OPEN the FW Regulating Bypass q
valves?
I a.
FWisolation would have to be reset.
b.
Reactor Trip breakers would have to be re-closed and MFP discharge MOV i-re-opened.
I j
c.
Reactor Trip breakers would have to be re-closed and FW isolation would have l
to be reset.
I d.
The Safety injection signal would have to be reset followed by the resetting of the FW isolation.
j
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SENIOR REACTOR OPERATOR Page 37 QUESTION: 070 (1.00)
The actions contained in EOP-3 " Steam Generator Tube Rupture" are being adhered to following a steam generator tube rupture in the "B" SG.
Operations personnel have just completed the RCS cooldown portion of the EOP. Due to an erroneous temperature reading, the actual RCS temperature is higher than the required temperature in the EOP.
What consequences could result from this condition?
a.
The pressurizer will go solid following a subsequent depressurization, b.
The resulting increase in pressure in the ruptured SG will result in lifting the steam line safety valve (s).
c.
Loss of RCS subcooling margin before the RCS pressure is decreased to a value equal to the ruptured SG pressure.
d.
The intact SG pressure will be greater than the ruptured SG pressure and will result in an adequate subcooling margin in the RCS.
.
QUESTION: 071 (1.00)
An automatic Safety injection actuation signal was received on Unit 1. Alarm annunciator REACTOR COOLANT DRAIN TANK HI-LO LEVEL just came in. Assuming all systems are functioning properly and that both RCDT pumps are in auto, what is the status of the RCDT system?
a.
One RCDT pump is running with flow directed to the RWST, b.
One RCDT pump is running with flow directed to the Holdup Tanks.
c.
Both RCDT pumps are running with flow directed to the Holdup Tanks.
d.
Neither RCDT pump is running and NO flow is being directed out of the RCDT.
!
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SENIOR REACTOR OPERATOR Page 38 QUESTION: 072 (1.00)
The following plant conditions exist:
Unit 1 is in refueling sinutdown.
Core geometry changes are in progress.
N-31 is indicating 35 cps.
N-32 is indicating 36 cps ut.d is providing audible count rate.
N-40 is out of service.
Which ONE of the following actions is immediately required if vital instrument power bus Y-03 is lost to source range detector N-327
. a.
Bypass the N-32 Level Trip.
b.
Evacuate the Containment Building.
c.
Suspend all core geometry changes.
d.
Switch the audible ceunt rate selector to N-31.
QUESTION: 073 (1.00)
.
In accordance with Technical Specifications, which ONE of the following situations constitutes a LOSS of Containment Integrity requirements?
a.
A containment forced vent is performed with the reactor operating at 100%
power.
b.
The plant is in refueling shutdown with fuel movement in progress when a containment purge is initiated.
c.
With the reactor at 30% power, inspection of the Containment Equipment Access Hatch reveals a cocked seal.
d.
While performing the operability test of two normally open, redundant containment isolation valves at 100% power, one of the valves fails to CLOSE.
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SENIOR REACTOR OPERATOR Page 39 QUESTION: 074 (1.00)
A reactor trip and Si have occurred on Unit 1. Unit 2 subsequently trips, and you notice that the RCPs for Unit 2 have tripped.
Diagnose why the RCPs tripped on Unit 2.
a.
A malfunction in the fast bus transfer circuitry has occurred.
I b.
The fast bus transfer is blocked on Unit 2 due to the Unit 1 trip with St.
c.
Unit 2 does not have the fast bus transfer option.
d.
There was no Si on Unit 2; SI generates the fast bus transfer.
QUESTION: 075 (1.00)
According to Technical Specifications, which ONE of the following is the additional requirement between a High Radiation Area with levels of 700 mrem /hr and one with levels of 1100 mrem /hr?
a.
The 1100 mrem /hr area requires posting as an EXTREMELY HIGH RADIATION AREA.
b.
The 1100 mrem /hr area requires that an individual entering the area be j
accompanied by an individual qualified in radiation protection procedures.
c.
The 1100 mrem /hr area requires locked doors to prevent unauthorized entry, and the keys shall be maintained under administrative control of the DSS.
'
d.
The 1100 mrem /hr area requires that an individual entering the area take along a radiation monitoring device which continuously indicates the radiation dose rate.
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SENIOR REACTOR OPERATOR Page 40 QUESTION: 076 (1.00)
During refueling operations inside containment, the control room receives an RE-211/212 alarm.
Which ONE of the following actions should be performed FIRST?
a.
Verify Containment Ventilation isolation.
b.
Notify DCS, DTA, and implement emergency plan.
c.
Suspend all refueling operations inside containment.
d.
Notify plant personnel with Gai-Tronics announcements to evacuate.
QUESTION: 077 (1.00)
RHR ccoling was in progress when one RHR pump had to be secured due to mechanical failure. The remaining RHR pump is indicating erratic flow characteristics. According to SEP-1,
" Degraded RHR System Capability," which ONE of the following actions is required?
a.
Check completion of RHR suction line reflood, b.
Stop the RHR pump and isolate RCS drain path.
c.
Check operating RHR pump and vent as necessary.
,
!
d.
Check RHR and adjust flow to between 1100 gpm and 1500 gpm.
,
!
!
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SENIOR REACTOR OPERATOR Page 41 OUESTION: 078 (1.00)
Which ONE of the following is an acceptance criteria for ECCS and a basis for operator actions required in EOPs during a Large Break LOCA7 a.
The loss of core coolable geometry shall NOT exceed 17%.
b.
The maximum fuel element clad temperature shall NOT exceed 2000*F.
c.
The local fuel element clad oxidation shall be LESS ths.10% of the total clad thickness.
d.
The total amount of hydrogen generation shall be LESS than 1% of all the potential Zirconium cladding interaction with water or steam.
QUESTION: 079 (1.00)
The following conditions exist at Unit 1:
The reactor is shutdown.
RHR is in service.
RCS temperature is 280*F.
RCS pressure is 110 psig.
"1T-12 CC Surge Tank Level High or Low" Alarm is in.
The CC Surge Tank level is decreasing.
A leak in which ONE of the following locations would produce the above conditions?
a.
!
b.
Seal Return Heat Exchanger.
c.
Non-Regenerative Heat Exchanger.
d.
RCP Thermal Barrier Heat Exchanger.
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SENIOR REACTOR OPERATOR Page 42 QUESTION: 080 (1.00)
The following conditions exist at Unit 1:
A reactor trip has occurred due to loss of offsite power.
None of the EDGs have started.
The DTA reports the following Critical Safety Function Status Trees:
- Heat Sink Red
- Suberiticality Orange
- Containment Green
- Inventory Yellow
- Core Cooling Red
- Integrity Green Which ONE of the following procedures should be used to mitigate this event?
a.
ECA-0.0," Loss of All AC Power" b.
CSP-S.2, " Response to Loss of Core Shutdown" l
l c.
CSP-C.1, " Response to inadequate Core Cooling" d.
CSP-H.1, " Response to Loss of Secondary Heat Sink"
QUESTION: 081 (1.00)
l While Unit 1 is operating at 100% power, Service Water is lost to Component Cooling Heat Exchangers.
Which ONE of the following components is NOT DIRECTLY affected?
a.
Reactor Coolant Pumps b.
Safety injection Pumps c.
Spent Fuel Pool Heat Exchanger d.
Non-Regenerative Heat Exchanger
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' SENIOR REAC TOR OPERATOR Page 43 QUESTION: 082 (1.00)
Which ONE of the following is the basis for isolating the feedwater to the faulted SG?
a.
To maximize the energy release from the faulted SG.
b.
To maximize the cooldown capability from the non-faulted SG and minimizing RCS cooldown.
c.
To prevent overflowing the faulted SG with feedwater and thus minimizing the cooldown for PTS concerns.
d.
To prevent thermal shock to the faulted SG "U" tubes and thus minimizing the potential of rupturing a SG tube with subsequent off-site release.
QUESTION: 083 (1.00)
Which ONE of the following conditions require ALTERNATE BORATION per AOP-6E,
" Alternate Boration/ Loss of Shutdown Margin?" (ASSUME: Refueling operation in progress)
a.
Keff is determined to be 0.93 b.
1900 ppm boron concentration in RCS.
'
i c.
The reactor is 5.8% shutdown per calculation.
d.
Neutron counts have unexpectedly doubled in a short period of time.
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SENIOR REACTOR OPERATOR Page 44 QUESTION: 084 (1.00)
Unit 1 is performing power ascension in accordance with OP-1C, " Low Power Operation to Normal Power Operations."
With control rods in Auto, the following conditions are observed:
Turbine load Stable.
Reactor Power At approximately 75% and decreasing slowly.
Control Rods Stepping into the core slowly.
RCS Tavg 564*F and decreasing slowly.
Pressurizer Level 42% and decreasing slowly.
Pressurizer Pressure 1935 psig and decreasing slowly.
What actions MUST be performed?
a.
Manually trip the reactor, b.
Switch the controlling Tavg channel to alternate.
c.
Place the control bank selector switch to MANUAL.-
d.
Decrease turbine load to restore Tref - Tavg deviation.
QUESTION: 085 (1.00)
Which ONE of the following conditions will automatically trip the Instrument Air Compressors?
a.
Low lube oil pressure, b.
High lube oil temperature.
c.
High-High discharge temperature.
d.
High-High Intercooler Condenser level.
l
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.
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SENIOR REACTOR OPERATOR Page 45 QUESTION: 086 (1.00)
A LOCA has occurred at Unit 1. Which ONE of the following is accurate conceming the usage of Hydrogen Recombiners in post-DBA Hydrogen control?
a.
The Hydrogen Recombiners are not capable of removing Hydrogen concentrations in excess of 6% by volume.
b.
The Hydrogen Recombiner would not be required within the first 30 days following a LOCA event.
c.
Hydrogen recombination cannot occur until the containment Hydrogen concentration is between 8% and 10% by volume.
d.
When the containment Hydrogen concentration reaches 0.5% by volume, both Hydrogen Recombiners should be placed in service in conjunction with containment ptce.
QUESTION: 087 (1.00)
Which ONE of the following willisolate due to a Containment isolation signal?
a.
The Excess Letdown Heat Exchanger (via CC-769).
b.
The Non-Regenerative Heat Exchanger (via CV-130).
c.
Auxiliary charging, normal letdown, and seal water supply.
d.
Normal charging, excess letdown, and seal water supply.
QUESTION: 088 (1.00)
A HIGH radiation alarm received by which ONE of the following radiation roonitors will result in an automatic control function?
a.
Chemistry Lab monitor RE-103.
b.
SGBD Tank Area monitor 1RE-222.
c.
Condenser Air Ejector Gas monitor 1RE-215.
d.
Containment Fan Cooler SW Return monitor 1RE-216.
I
l SENICR REACTOR OPERATOR Page 46 QUESTION: 089 (1.00)
Fire pump P35B is controlled at C61, which is wired such that it will start the diesel engine automatically when the system pressure is reduced to 80 psig or on loss of AC power.
Which ONE of the following is NOT true concerning the diesel engine Ore pump?
a.
Crankir.,, ; tarts upon receipt of any start signal after a 10 second time celay, b.
The diesel engine can be started manually using the solenoid joy-stick mounted on the engine.
c.
If, during operation, the oil pressure becomes too high, a trouble signal will be genera 9d at C61 and C01, and the engine will stop running.
d.
The battery sets are electrically switched after each time the engine is cranked.
If the engine fails to start, the controller automatically switches to the second set of batteries.
QUESTION: 090 (1.00)
While examining the incore Thermocouple Map on the PPCS, it was noted that only ONE thermocouple in Core Quadrant 1 was responding. Per Technical Specification, what actions should be taken? (ASSUME: Unit at full power)
a.
Restore two (2) additional Thermocouple to OPERABLE within seven (7) days or be in at least HOT SHUTDOWN within the next twelve (12) hours.
b.
Restore two (2) additional Thermocouple to OPERABLE within forty-eight (48)
hours or be in at least HOT SHUTDOWN within the next twelve (12) hours.
c.
Restore one (1) additional Thermocouple to OPERABLE within seven (?) days or be in at least HOT SHUTDOWN within the next twelve (12) hours.
d.
Restere one (1) additional Thermocouple to OPERABLE within forty-eight (48)
hours or be in at least HOT SHUTDOWN within the next twelve (12) hours.
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l SENIOR REACTOR OPERATOR Page 47 QIJF.5 flON: 091 (1.00)
The following conditions exist at Unit 1:
Peactor power 50 %
Pressurizer pressure 1985 psig
!
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Pressurizer level 33 %
RCS Tavg 558'F RCS Thot 572'F (A); 571*F (B)
RCS Tcold 545*F (A); 544*F (B)
If loop B Thot output channel fails LOW, what is the response of the Pressurizer Level Control.
System?
a.
Pressurizer level will remain the same.
b.
Pressurizer level will increase to 39%.
c.
Pressunzer level will decrease to 28%.
d.
Pressurizer level will decrease to the point of Letdown Isolation.
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l SENIOR REACTOR OPERATOR Page 48 l
QUESTION: 092 (1.00)
The following condition exist in Unit 1:
Time 1000 Hrs 1100 Hrs i
PRT Level 70 %
76 %
PRT Temp 120*F 100*F j
PZR Level 45%
43%
)
l Tavg 570*F 569'F RMW System Pumps Running Running Which of the following is the MOST LIKELY source of the PRT levelincrease?
a.
Pressurized PORV (1-RC 430)
b.
Reactor Makeup Water valve (1CV-508)
c.
Letdown relief valve (1CV-203)
d.
Sest return relief valve (1CV-314).
QUESTION: 093 (1.00)
Power escalation is in progress at Unit 1. The P-10 permissive has just illuminated. In the process of controlling Tav0 with main turbine condenser steam dumps, a slight transient condition occurs which results in a reduction of reactor power to 8% on 2 channels and 9% on the other 2 channels.
In relations to power range High Flux-Low Range Reactor Trip, which ONE of the following is the system response?
a.
The nuclear instrument reactor trip bistable de-energize.
b.
The nuclear instrument reactor trip bistables re-energize.
!
c.
The nuclear instrument reactor trip will automatically unblock.
'
d.
The nuclear instrument reactor trip will NOT automatically unbloch I
,
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SENIOR REACTOR OPERATOR Page 49 QUESTION: 094 (1.00)
[
You are assuming the DSS duty after one week of absence.
Which ONE of the following is required by procedure OM 1.1, Attachment 4, during the shift tumover process?
a.
Each on-coming operator shall review the narrative log, rounds sheets, and checklists for all stations.
b.
The on-coming AO watchstander shall perform a radio check with the Control R0om, identifying which watchstation has been relieved and by whom.
c.
Immediately after the shift tumover, completed individual position checklists are forwarded to Document Control througn the Assistant Operations Manager, d.
Each on-coming operator shall review the Control Room narrative log. As a minimum, the review shall include narrative logs since the last time on shift.
OUESTION: 095 (1.00)
The following conditions exist at Unit 1:
"A" RHR loop is in operation for a refueling outage.
Operations is in the process of lowering the RCS level to mid-loop.
RCS temperature is 110*F Containment temperature 87*F RCS localleve(indication is 60" RCS electronic level indication (LT-447/447A) is 55" Given the above conditions, which ONE of the following actions shall be taken?
a.
Stop the rurning RHR pump to prevent cavitation, b.
Continue lowering the RCS level at the current rate.
c.
Align the RHR pump to the RWST to restore P,CS level.
d.
Terminate the lowering of the RCS leve! until a reason for level discrepancy has been determined.
_ _ _ - _
.
_.
- -
- _ _ -
_ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _
_ _ _ _ _ _-_ - _ - _ - _ - __ ___ - __ - ___
~ SENIOR REACTOR OPERATOR Page 50 QUESTION: 096 (1.00)
The WE Emergency Plan, EP 5.0, " Organizational Control of Emergencies," specifies the authority and responsibilities of key personnel during an emergency at PBNP.
- Which ONE of the following is NOT a responsibility of the Shift Superintendent under the WE Emergency Plan, EP 5.0?
a.
Assume the responsibility of the Plant Operation Manager until relieved of this responsibility.
b.
Direct the plant response to assure and control the emergency and initiate the required plant and offsite notifications, c.
Detennine when it is warranted to obtain the assistance of county, state or federal emergency support organizations.
d.
Determines that an event has potential of or is actually exceeding predetermined EALs, then shifts the organization into an emergency mode as necessary.
QUESTION: 097 (1.00)
Unit 1 is operating at 100% power during mid-cycle. All systems are functioning normally.
What is the affect on plant operations if instrument air supplied to the CVCS letdown heat.
exchanger component cooling water outlet valve, CV-130 is lost?
. CV-130 geos...
a.
shut and reactor power decreases due to boron release in the CVCS demineralizers.
b.
shut and the CVCS demineralizers are automatically bypassed on high temperature signal.
c.
open and reactor power increases due to boron retention in the CVCS demineralizers.
L d.
open and the CVCS demineralizers are automatically bypassed on low temperature signal.
t
. _ _ _ _. _ _ _ _
_ _. _ _ _ _ _ _ _ _ _ _ _
. _ - _ _ _.
_ -
--___ _-_________ _______-____ - _ - __ _ _ _ _
!
SENIOR REACTOR OPERATOR Page 51
'
!
. QUESTION: 098 (1.00)
{
i The following conditions exist:
Reactor power:
Tripped on Unit 1; Cold Shutdown on Unit 2 Control Room:
FIRE in progress Offsite Power.
AVAILABLE A control room evacuation has been ordered and AOP-10A, " Safe Shutdcwn-Local Control" has been entered. One of the INITIAL operator actions prior to evacuation is to...
a.
initiate Safety injection on both units.
b.
shut both the MSIVs on Unit 1.
c.
Initiate alternate boration on both units.
i d.
establish feed to the "A" SG on Unit 1.
i
l QUESTION: 099 (1.00)
Which ONE of the following control functions will be initiated if a high alarm is actuated on its associated radiation monitor?
a.
El. 66' Containment Low Range Monitor (1RE-102)- Isolates SG blowdown.
b.
Auxiliary Building Exhaust High Range Gas (RE-319)- Shifts Auxiliary Building Ventilation to the Carbon Filters. Shuts ve. gas release valve RCV-014, if open.
c.
Auxiliary Building Vent Exhaust Gas Monitor (RE-214) - Shifts Auxiliary Building Ventilation to the Carben Filters. Shuts vent gas release valve RCV-014, if open.
d.
Steam Line 1 A Monitor (1RE-231)- Shuts Blowdown Valves (1MS-5958, 1MS-5959), Blowdown Tank Outlet Valve (1 MS-2040), and Blowdown Sample Isolations (1MS-2083,1MS-2084).
t-E__._______
. _ _. _ _. _. _... _. _. _. _.
_ _ _ _
.._._j
_ _ _ - _ _ _ _ _ _ - _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ - _ _ _ -_
___
SENIOR REACTOR OPERATOR Page 52 QUESTION: 100 (1.00)
During a surveillance test, RE-218,' Waste Disposal System Liquid Monitor," has been determined to have its alarm setpoint set to a non-conservative value. (ASSUME: A liquid waste discharge was completed one hour ago.)
Which ONE of the following is the immediate action the DSS should take?
a.
Perform 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> grab sample.
b.
Verify automatic isolation has occurred.
c.
Change the radiation monitor setpoint to the appropriate value.
d.
Notify Chemistry to obtain a confirmatory analysis of the effluent.
I j
r"*""" FND OF EXAMINATION ***"*"")
l l
l
!
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
_ _ _ _ _. _ ______-______ ________ _____ ________ _____-_-_________ - __
____-___ _ - _---._- _ - _ _ _
_ _ - _ _
SENIOR REACTOR OPERATOR Page 53 ANSWER: 001 (1.00)
ANSWER: 005 (1.00)
ANSWER: 009 (1.00)
a.
d.
d.
REFERENCE:
REFERENCE:
REFERENCE:
AOP-5A, " Loss of TRHB 10.8, Rev 5, "
EOP-0.2, Rev.14, "
Condenser Vacuum," pg.
Safety injection System."
Natural Circulation
Cooldown" 060000K603
..(KA's)
BG EOP-0.2, Rev 14, step 000051A202
..(KA's)
9.
ANSWER: 010 (1.00)
00WE09K301
..(KA's)
ANSWER: 006 (1.00)
c.
a.
REFERENCE:
REFERENCE:
TRHB 10.16 " Engineered ANSWER: 002 (1.00)
EOP-1," Loss Of Reactor Safety System" c.
or Secondary Ct.olant,"
TRHB 10.12 " Containment REFERENCE:
Rev 24 Spray System" Logic EOP-0, Rev 6, " Reactor Sheet 8 Trip or Safety injection,"
000040K106
..(KA's)
pg 11 064000A307
..(KA's)
000007A201
..(KA's)
ANSWER: 007 (1.00)
c.
ANSWER: 011 (1.00)
REFERENCE:
b.
ANSWER: 003 (1.00)
AOP-17A U1, * Rapid REFERENCE:
d.
Power Reduction" Tech Specs 15.3.10.B.1.g REFERENCE:
Tech Specs 15.3.1.A.1
// TST-8.2 Steam Tables 003000K201
..(KA's)
064050G009
..(KA's)
000008K101
..(KA's)
ANSWER: 003 (1.00)
ANSWER: 012 (1.00)
ANSWER: 004 (1.00)
a.
b.
a.
REFERENCE:
REFERENCE:
REFERENCE:
TRHB 12.1, " Main 10 CFR 20.1003, Pressurizer Level Control Generator, Exciter and Definitions.
System Lesson Plan Transformer," Table TRHB 13.6, Rev 1 12.1.3. A.
194001K103
..(KA's)
AOP-1A, Reactor Coolant AOP-0.1, Step 2 3, Rev 3 /
Leak, Rev 10 Logic Sheel 3 000028K305
..(KA's)
045000A304
..(KA's)
('
l
_ - - - - - _ _ - - _ _
__ _ - _ _
-
_ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -
SENIOR REACTOR OPERATOR Page 54 ANSWER: 013 (1.00)
ANSWER: 017 (1.00)
ANSWER: 021 (1.00)
a.
b.
c.
REFERENCE:
REFERENCE:
REFERENCE-I NP 4.2.14 " Administrative BG CSP-S.1, Rev.14, 10 CFR 55.53 Dose Levels / Dose Level
" Response to Nuclear OM 3.10, Rev.10, Extension," Rev 1 Power Generation /ATWS."
" Operations Personnel I
Assignment and l
194001K103
..(KA's)
Scheduling."
l
'
000029A115
..(KA's)
194001A103
..(KA's)
ANSWER: 014 (1.00)
b.
ANSWER: 018 (1.00)
'
REFERENCE:
d.
ANSWER: 022 (1.00)
l NP 4.2.18, Rev.
REFERENCE:
d.
0," Planned Special Tech. Spec. Bases REFERENCE:
Exposure" 15.3.6.A.1.c.
OP-5A, Rev. 29, " Reactor Coolant Volume Control."
194001K104
..(KA's)
029000K104
..(KA's)
THRB 10.6, Rev. 4,
" Chemical and Volume l
Control System."
ANSWER: 015 (1.00)
ANSWER: 019 (1.00)
b.
a.
000022K307
..(KA's)
REFERENCE:
REFERENCE:
EP 6.0, Rev. 37 TRHB 10.3, Rev. 3, Figure
" Emergency Measures" 10.3.7 and LP 2438 LC ANSWER: 023 (1.00)
2.1, Logic Sheet 18 and b.
194001A116
..(KA's)
the Setpoint Document.
REFERENCE:
BG ECA-0.0, Rev.19, 000027A101
..(KA's)
" Loss of All Power."
ANSWER: 016 (1.00)
a.
000055A106
..(KA's)
REFERENCE:
ANSWER: 020 (1.00)
AOP-6A, Rev. 8, " Dropped b.
Rod" REFERENCE:
ANSWER: 024 (1.00)
TRHB 13.13, Rev.1, " Rod OP-3A, Rev. 40, " Normal b.
insertion and AT Operation to Low Power REFERENCE:
Deviation Alarms."
Operation."
TRHB 13.8,Rev.1, " Rod Speed and Direct Control."
000003A101
..(KA's)
004000K601
..(KA's)
035010K501
..(KA's)
[
l t _ _ _______ _ _ _-_ ___
_ _ _ _ _ _ -
. - _ _ _ _ _ _ _ _ _ _ _ _.
- _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - - - - -
SENIOR REACTOR OPERATOR Page 55 ANSWER: 025 (1.00)
ANSWER: 029 (1.00)
ANSWER: 034 (1.00)
c.
a.
d.
REFERENCE:
REFERENCE:
REFERENCE:
l TRHB 11.3, Rev. 3, Background Document for TRHB 13.12, Rev. 3,
" Secondary System EOP-1.2, Rev 16
" Process and Area j
Description: Main LPO435, Rev 4, EO 2.1.24 Radiation Monitoring
'
Feedwater System."
System" TRHB 13.7, Rev.1, 000011K313
..(KA's)
RMS Alarm Setpoint and
"Feedwater Control Response Book System" pg 5.
ANSWER: 030 (1.00)
000059K201
..(KA's)
059000K419
..(KA's)
b.
REFERENCE:
AOP-6E, Rev 7,"Altemate ANSWER: 035 (1.00)
ANSWER: 026 (1.00)
Boration/ Loss of Shutdown d.
b.
Margin" REFERENCE:
REFERENCE:
SEP-1.1, Rev. 2,"Altemate TRHB 13.1, Rev. 3, 000024K302
..(KA's)
Core Cooling"
" Nuclear Instrumentation System (Excore 000025K101
..(KA's)
l instrumentation)" Logic ANSWER: 031 (1.00)
Sheets 11 and 12 a..
REFERENCE:
ANSWER: 036 (1.00)
000033A208
..(KA's)
AOP-9B, Rev 12, d.
" Component Cooling REFERENCE:
a System Malfunction" AOP-1 A, " Reactor Coolant l
ANSWER: 027 (1.00)
Leak, Attachment A SG a.
000026K303
..(KA's)
Tube Leakage Actions'"
REFERENCE:
Tech Specs 15.3.1.D TRHB 13.8, Rev.1," Rod Speed and Direct Control" ANSWER: 032 (1.00)
000037A206
..(KA's)
pg.10 d.
LP 2441, EO 2.4 and REFERENCE:
Logic Sheet 16 TRHB 11.14, Rev. 8, " Fire ANSWER: 037 (1.00)
Prok.ation System" c.
000001K105
..(KA's)
REFERENCE:
000067K102
..(KA's)
BG CSP-H.5, Rev.65,
"Respnse to Steam ANSWER: 028 (1.00)
Generator Low Level" d.
ANSWER: 033 (1.00)
LP 1998, F.O 2.6.
REFERENCE:
c.
TRHB 12.9, Rev. 2, REFERENCE:
0000"AK102
..(KA's)
" Electrical System CSP-C.1, Rev.14, Description: Instrument
" Response to inadequate Bus Power," pg.11.
Core Cooling" 000057A218
..(KA's)
000074K201
..(KA's)
,
-.
_ _ _ _ _ - - _ - _ _ _ _ - -
- _ _ - _ _ _ - _ _
. _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _
_ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - - - - _ _ - _ _ _ _ _
SENIOR REACTOR OPERATOR Page 56 ANSWER: 038 (1.00)
ANSWER: 043 (1.00)
ANSWER: 047 (1.00)
c.
d.
c.
REFERENCE:
REFERENCE:
REFERENCE:
AOP-0.0, Rev 12, Step TRHB 13.4, Rev 4, "ESF TRHB 10.1, Rev.5, 4.3, p 5.
Actuation Instrumentation"
" Reactor Containment" C1805 COT and TRHB 000058K302
..(KA's)
013000A401
..(KA's)
Figure 10.1.24 022000G007
..(KA's)
ANSWER: 039 (1.00)
ANSWER: 044 (1.00)
b.
d.
REFERENCE:
REFERENCE:
ANSWER: 048 (1.00)
AOP-58,.Rev 11, " Loss of AOP-8C, Rev 6, " Fuel b.
Instrument Air" handling Accident in REFERENCE:
Primary Auxiliary Building" OM 4.1.7, Rev 0, "RMS 000065A208
..(KA's)
Alarm Setpoint &
000036G010
..(KA's)
Response Book" ANSWER: 040 (1.00)
000061G005
..(KA's)
a.
ANSWER: 045 (1.00)
REFERENCE:
c.
TRHB 10.16, Rev 3, REFERENCE:
ANSWER: 049 (1.00)
" Primary System LP 1547, Rev 3, p 51, EO b.
i Description: ESF" 1.2.40 REFERENCE:
I l
TRHB 13.8, Rev.1, " Rod REl 13.0, Rev.14, 000056K301
..(KA's)
Speed and Direct Control"
" Quadrant Power Tilt" TS 15.3.10.E.3.c 001000K105
..(KA's)
ANSWER: 041 (1.00)
015000K501
..(KA's)
c.
REFERENCE:
ANSWER: 046 (1.00)
TRHB 13.6, Rev 1, d.
ANSWER: 050 (1.00)
" Pressurizer Level Control REFERENCE:
c.
System" AOP-6B, Rev 9, " Stuck REFERENCE:
Rod or Malfunction BG EOP-1.2, Rev 16, 000028A107
..(KA's)
Position Indication"
"Small break LOCA
,
TS 15.3.10.B.1 Cooldown and
!
Depressurization" ANSWER: 042 (1.00)
000005G003
..(KA's)
.
a.
022000K405
..(KA's)
I REFERENCE:
WHB 13.6, Rev.1,
": ressurizer Level Cc strol
.
System" 011000A101
..(KA's)
l L________-_________-_---
_ - - - - - - _ - - - - _ -
J
_ ____ ______ __ __________ _ __ __ ____ _ __.
_ _ _ _ _ _ _ _ _ _ - _ - _ - _
l SENIOR REACTOR OPERATOR Page 57 ANSWER: 051 (1.00)
ANSWER: 055 (1.00)
ANSWER: 059 (1.00)
a.
d.
b.
l REFERENCE:
REFERENCE:
REFERENCE:
'
TRHB 13.13, Rev.1, " Rod OP-4F p.10 TRHB 10.13, Rev. 2, " Fuel insertion Limit and DT handling System" Deviation Alarm" 002000K401
..(KA's)
034000G009
..(KA's)
014000A103
..(KA's)
ANSWER: 056 (1.00)
a.
ANSWER: 060 (1.00)
<
ANSWER: 052 (1.00)
REFERENCE:
b.
I a.
TRHB 13.5, Rev.
REFERENCE:
'
REFERENCE:
2," Pressurizer Pressure TRHB 13.4, Rev. 4, "ESF TRHB 13.4, Rev 4, "ESF Control System" Actuation Instrumentation
Actuation Instrumentation C0905C01, Logic Sheet (Safeguard System)
l (Safeguard System)
18 and P&lD 541F091
'
Section 2.2.6, p 23; Sheet 1 039000K408
..(KA's)
LP0515, Rev 2, Step 2(b)(2), p 9 010000K301
..(KA's)
EO 1.2.8 ANSWER: 061 (1.00)
a.
026000A301
..(KA's)
ANSWER: 057 (1.00)
"EFERENCE:
b.
tt 7739, Rev 4, p 4, EO REFERENCE:
-
ANSWER: 053 (1.00)
AOP-8F, Rev 5, " Loss of FRHB 13.12, Rev 3, b.
Spent Fuel Pool Cooling" Process and Area REFERENCE:
LP0065. Rev 3, Step 3.3, Monitoring System"
,
TRHB 11.2, Rev. 5, p 3 EO. 4.2 l
" Condensate System" pg.
073000A101
..(KA's)
033000A203
..(KA's)
C3912 COT and Logic
'
Sheet 24 ANSWER: 062 (1.00)
ANSWER: 058 (1.00)
c.
056000A204
..(KA's)
a.
REFERENCE:
REFERENCE:
LP0069, Rev 12, p 19; TRHB 13.3, Rev 4, LP0069, Rev 9, p 14 EO l
ANSWER: 054 (1.00)
" Reactor Protection 2.3 b.
(Reactor Trip) Figure Setpoint Document 8.1 REFERENCE:
13.3.19 (Aux Cooling)
LPO169, Rev 9, p 12, EO LP0273, Rev 2, EO 2.6 2.8 C1104 COT 005000K412
..(KA's)
'
061000K402
..(KA's)
012000A402
..(KA's)
i
-_
_ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - _ _ _
-..
.
__
. _ _ _ _ _ _ _ _ _ _ - _ _.
_ _ _ _ _ _ _ _
SENIOR REACTOR OPERATOR Page 58 ANSWER: 063 (1.00)
ANSWER: 068 (1.00)
ANSWER: 073 (1.00)
d.
d c
REFERENCE:
REFERENCE:
REFERENCE:
LPO253, Rev 3, p 8; TP-8.
WOG EOP Guidelines Technical Specifications EO 2.3,2.4 Generic Instructions EOP-0 " Reactor Trip or 000069A201
..(KA's)
041020K603
..(KA's)
Safety injection" 000009K316
..(KA's)
ANSWER: 074 (1.00)
ANSWER: 064 (1.00)
b b
REFERENCE:
REFERENCE:
ANSWER: 069 (1.00)
LP0121 "DC Electrical AOP-68, Rev.9, step 8 d
Distribution" Logic Sheet 4 pg.5, step 10 pg.6.
REFERENCE:
ESFAS 062000G007
..(KA's)
000005A201
..(KA's)
S/G Water Level Control 00WE05K201
..(K i's)
ANSWER: 075 (1.00)
ANSWER: 065 (1.00)
c c
REFERENCE:
REFERENCE:
ANSWER: 070 (1.00)
Technical Specifications CSP-P.1 " Response to c
15.6.11 Pressurized Thermal REFERENCE:
Shock Condition" EOP-3 " Steam Generator 000061G005
..(KA's)
Tube Rupture" 00WE08A202
..(KA's)
000038K306
..(KA's)
ANSWER: 076 (1.00)
c ANSWER: 066 (1.00)
REFERENUE:
b ANSWER: 071 (1.00)
AOP-88, " Irradiated Fuel REFERENCE:
d Handling Accidentin AOP-1B " Reactor Coolant REFERENCE:
Containment."
Pump Malfunction,"
ESF Actuation Rev.10, pgs.4-11 Liquid Rad Waste System 000036K303
..(KA's)
000015K207
..(KA's)
068000A404
..(KA's)
ANSW,ER 077 (1.00)
b ANSWER: 067 (1.00)
ANSWER: 072 (1.00)
REFERENCE:
c c
SEP-1 " Degraded RHR REFERENCE:
REFERENCE:
System Capability" AOP-8A-U1 "High Reactor TRHB 13.1, Nuclear Coolant Activity," Rev.5 Instrumentation System,"
000025A103
..(KA's)
Rev 3, pg 3.
000076A202
..(KA's)
Technical Specifications 000032G003
..(KA's)
. - _ _ _ _ _ _ _ _ _ _ _ _
_ _ _ _ _.
_ _ _. _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ - _ ___-_-
_
_ _ _ _ _ - _
_ _ _ _ _
_ _ _ _ _ _
SENIOR REACTOR OPERATOR Page 59 ANSWER: 078 (1.00)
ANSWER: 083 (1.00)
ANSWER: 087 (1.00)
d d
a REFERENCE:
REFERENCE:
REFERENCE:
10CFR50.46 AOP-6E, Rev. 7, TRHB 10.1, " Reactor
" Alternate Boration/ Loss of Containment" 000011K312
..(KA's)
Shutdown Margin" TRHB 11.8, " Service RP-1C Water System" Logic /P&lDs ANSWER: 079 (1.00)
000024K301
..(KA's)
b 10300A301
..(KA's)
REFERENCE:
AOP-98 " Component ANSWER: 084 (1.00)
,
Cooling System c
ANSWER: 088 (1.00)
'
Malfunction" REFERENCE:
b AOP-6C, " Uncontrolled REFERENCE:
000026A105
..(KA's)
Motion of RCAA(s)"
RMSASRB Cl 00001AK107
..(KA's)
072000K102
..(KA's)
ANSWER: 080 (1.00)
a REFERENCE:
ANSWER: 085 (1.00)
ANSWER: 089 (1.00)
ECA-0.0, Loss of All AC a
c Power" REFERENCE:
REFERENCE:
TRHB 11.11, " Secondary TRHB 11.14, "Secondsry 000055G011
..(KA's)
System Description:
System Descriptions: Fire instrument and Service Air Protection System" System" ANSWER: 081 (1.00)
086000K401
..(KA's)
c 007800K401
..(KA's)
REFERENCE:
TRHB 10.9, "CCW ANSWER: 090 (1.00)
System" ANSWER: 086 (1.00)
d TRHB 11.8, " Service b
REFERENCE:
Water System" REFERENCE:
Tech Specs Table
TRHB 10.11, " Primary 15.3.5-5, No.13.
i 000062A102
..(KA's)
Systems Descriptions:
l Post DBA Hydrogen 017000G005
..(KA's)
i Control ANSWER: 082 (1.00)
System (Post Accident b
Containment Ventilation REFERENCE:
System)
<
BG EOP-2, " Faulted Steam Generator 028000K502
..(KA's)
Isolation" 000040K304
..(KA's)
<
u__
_ _ _ _ _ _ _. _.. _ _.. _ _.. _ _ _ _. _ _ _ _ _ _. _ _ _ _.
_
_
_
- - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ - _ _
_ _ _- ____ - ___-_ _________-___ _ - ____-__ _
_ _ _ _ _ _ _ _.
SENIOR REACTOR OPERATOR Page 60 ANSWER:. 091 (1.00)
ANSWER: 094 (1.00)
a b
REFERENCE:
REFERENCE:
TRHB 13.6, OM 1.1, " Conduct of
" Instrumentation and Operations," Attachment 4 Control Systems
" Standards and Descriptions:
.
Expectations for the Pressurizer Level Control Watchstanding and Shift System" Turnover."
TRHB 13.8, " Rod Speed and Direction Control" 194001A103
..(KA's)
016000K302
..(KA's)
ANSWER: 095 (1.00)
d ANSWER: 092 (1.00j REFERENCE:
b TRCR 55.0 IU-02, REFERENCE:
LP-1308, "RCS Draindown P&lD RCS 541F091 SH2 Reduced inventory OPS" OP-4D " Draining the 007000A301
..(KA's)
Reactor Coolant System" 005000G014
..(KA's)
l ANSWER: 093 (1.00)
)
d
.
' REFERENCE:
ANSWER: 096 (1.00)
TRHR 13.1, " Instrument c
and Control System REFERENCE:
Description: Nuclear WE Emergency Plan, EP instrumentation 5.0, " Organizational System /Excore Control of Emergencies" Instrumentation" Technical Specification 194001A116
..(KA's)
15.2.3, " Limiting Safety System Settings, Protective ANSWER: 097 (1.00)
Instrumentation" c.
STPT 3.1 REFERENCE:
AOP-58, " Loss of l
012000K604
..(KA's)
Instrument Air" l
008000A205
..(KA's)
I
,
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l SENIOR REACTOR OPERATOR Page 61 ANSWER: 098 (1.00)
b.
REFERENCE:
AOP-10A, Rev. 22, " Safe Shutdown-Local Control" 000068G010
..(KA's)
ANSWER: 099 (1.00)
c.
REFERENCE:
RMSASRB, Radiation Monitoring System Alarm Setpoint and Response Book.
072000A301
..(KA's)
ANSWER: 100 (1.00)
c.
REFERENCE:
OM-4.1.7, RMS Alarm Setpoint and Response Book l
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068000A402
..(KA's)
l c""*"" END OF EXAMINATION "*"""*)
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SENIOR REACTOR OPERATOR Page 62 I
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SENIOR REACTOR OPERATOR Page 63 ANSWER KEY 001 a 021 c 041 c 002 c 022 d 042 a 003 d 023 b 043 d 004 a 024 b 044 d 005 d 025 c 045 c 006 a 026 b 046 d
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012 b 032 d 052 a 013 a 033 c 053 b l
014 b 034 d 054 b 015 b
. 035 d 055 d i
016 a 036 d 056 a
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017 b 037 c 057 b 018 d 038 c 058 a 019 ' a 039 b 059 b 020 h 040 a 060 b l
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SENIOR REACTOR OPERATOR Page 64 061 a 081 c 062 c
- 082 b 063 d 083 d 064 b 084 c 065 c 085 a 066 b 086 b 067 c 087 a 068 d 088 b
-. 069 d 089 c 070 c 090 d 071 d 091 a 072 c 092 b 073 c 093 d 074 b 094 b 075 c 095 d 076 c 096 c 077 6 097 c 078 d 098 b
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079 b 099 c 080 a 100 c
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(**"****" END OF EXAMINATION *""*"")
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