IR 05000346/1997308
| ML20211C569 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 09/23/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20211C515 | List: |
| References | |
| 50-346-97-308OL, NUDOCS 9709260244 | |
| Download: ML20211C569 (82) | |
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U. S. NUCLEAR REGULATORY COMMISSION REGION til Docket No:
50 346
License No:
NPF 3 Report No:
50 346/97308(OL)
Licensee:
Toledo Edison Company
. Facility:
Davis Besse Nuclear Power Station Location:
5503 N. State Route 2 Oak Harbor, OH 43449 i
Dates:
July 8 10,1997 Examiners:
D. McNeil, Chief Examiner, Rlll E. Plettner, Examiner, Rlli l
Approved by:
M. N. Leach, Chief Operator Licensing Branch i
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9709260244 970923 i
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EXECUTIVE SUMMARY Davis Besse Nuclear Power Station NRC Examination Report 50 340/97308 A licensee developed and NRC approved initial operator licensing examination was administered to three Senior Reactor Operator (SRO) license applicants. The examination was completed in one week, including dynamic simulator scenario and job performance measures validation on the Davis Besse simulator by NRC examiners.
Results:
Alllicense applicants passed all portions of their respective examinations and
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were issued SRO licenses.
Overall operator performance during the examination was determined to be
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satisf actory. (Sections 05.2,05.3,05.4,05.5)
Examination Summarv:
The examiners were concerned with administrative procedure, DB OP-0000,
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Conduct of Operations, which stated that the shift supervisor was authorized to approve certain deviations from operating procedures without a second review. (Section 03.1)
The submitted examination material was of high quality. With a few
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exceptions in the area of administrative Job Performance Measures (JPMs),
the examination materials required little or no modification prior to administration. (Section 05.1)
One occurrence of poor examination material control (examination security)
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was identified during the examination and resulted in a non-cited violation.
(Section 05.1)
The licensee prepared written examination required a minimum amount of
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revision. (Section 05.2)
Preparation and validation of administrative JPMs was considered
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inadequate. (Section 05.3)
Licensee prepared walkthrough JPMs were considered good. The JPM
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follow up questions were determined to be memory level questions.
(Section 05.4)
Licensee prepared dynamic scenarios were considered good. (Section 05.5)
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Simulatnr deficiencies were noted during the examination. (Section 05.7)
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Reoort Details
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1. Operations
Conduct of Operations
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01.1 General Comments l
During the initiallicense examination validation, examiners observed control room operations. The observed period included a shift turnover and the calibration of an automatic controller. Control room operators were noted to be observing control
toom instrumentation at acceptable time intervals and were attentive to the
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controller calibration in progress. During an additional period of observation, a
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pressurizer pressure Instrument f ailed giving a reactor protection channel trip.
Response to the instrument failure was prompt and accurate. During both visits to the control room, operator demeanor was professional.
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Operations Procedures and Documentation 03.1 General Comments While reviewing DB OP 00000, Conduct of Operations, Section 6.8.3.c, Revision 3, NRC examiners determined that the procedure specified actions that appeared to be in violation of Davls Besse Technical Specification (TS) 6.5.3.1.b. Specifically, TS 6.5.3.1.b permitted, in part, temporary apnroval of changes to non administrative procedures that clearly did not change the Intent of the procedures if reviewed and approved by two members of plant management staff, at least one of whom holds a senior reactor operator's license. However, DB OP-00000, Step 6.8.3.c,
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i specified that steps in operating procedures could be performed in other than the as written sequence if approved by the Shift Supervisor (without obtaining the approval of a second member of management staff). This information was subsequently forwarded to the Davis-Besse NRC Resident Office for further evaluation and is considered an unresolved item. (URI 50 346/97308-01)
Operator Training and Qualification 05.1 General CommtD11 Operator initiallicense examinations were administered at the Davis Besse Nuclear Power Station (DBNPS) to three Senior Reactor Operator (SRO) applicants during the week of July 7,1997. All applicants were previously licensed at the facility as Reactor Operators.
The licensee participated in a pilot process in which the license examination was developed by the licensee and approved by the NRC in accordance with guidance prescribed by NUREG 1021, " Operator Licensing Examiner Standards," Revision 7,
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and superseded in part by Interim Pilot Examination Guldance provided in Generic d
Letter (GL) 95-06, " Changes in the Operator Licensing Program." As part of the
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pilot program, the NRC administered the operating test and the licensee administered the written test.
All materials developed by the licensee for the examination were submitted to thg NRC on schedule. The submitted material was of high quality and with only a few exceptions, was used as written by the f acility instructors. Examination validation by NRC examiners was conducted during the examination week.
Applicants, licensees, and f acility licensees shall not engage in any activity that compromises the integrity of any application, test, or examination required by 10 CFR 55. (10 CFR 55.49) Contrary to this requirement, an operating Job Performance Measure (JPM) was inadvertently lef t unattended at the simulator l
Instructor console for a short period of time during instructor validation of the
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JPMs. The facility Instructor assigned to develop the examination immediately informed the NRC of the loss of examination security. A suitable JPM replacement was suggested by facility instructors and accepted by the NRC examiners.
Compensatory measures were implemented to prevent recurrence of the loss of
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I examination control. This event was of minor significance and is being treated as a Non Cited Violation, consistent with Section IV of the NRC Enforcement Policy (NCV 50 346/97308 02).
05.2 Written Examination a.
Examination Scone using NUREG 1021, " Operator Licensing Examiner Standards," GL95 06, * Changes in the Operator Licensing Program," and NUREG/BR 0122, " Examiner's Handbook for Developing Operation Licensing Written Examinations," examiners reviewed each written examination question. Each question was reviewed for comprehension, validity and level of difficulty, b.
Observations and Findinas NRC examiners made minor comments on twenty of the one hundred questions submitted by the licensee. This was substantially below the number and depth of comments returned to other facilities involved in the pilot examination process. The questions were comprehensive, valid questions and examined at the correct difficulty level.
Nine of the written examination comments concerned a question writing technique that is considered undesirable by NRC examiners, This question technique is sometimes referred to as a True/ False multiple choice question. The technique allows an applicant to easily eliminate two of the four possible answers to a multiple choice question, thus increasing the applicant's probability of guessing the correct answer when the answer is not known. Several of these questions were reworked to eliminate the True/ False aspect of the question.
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- Examiners observed the start of the written examination. All examination rules were read to the applicants and enforced by the training staff. The training staff conformed to all requirements for administering the written examination. All applicants passed the written examination with scores ranging from 90% to 93%.
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Conclusions The written examination provided by the training department provod to be a good evaluation tool for determining applicant competence. Applicants appeared well prepared to take the written examination. Use of the True/Falso multiple choice questions was considerad a weakness in constructing the written examination.
05.3 Administrative Job Performance Measureg a.
Examination Scop _g Using NUREG 1021, Operator Licensing Examiner Standards, examiners reviewed each administrative JPM. Each administrative JPM was reviewed for applicability, importance, and safety significance. The aggregate of the administrative JPMs was reviewed to ensure all required areas of the administrative JPM examination were represented, b.
Observations and Findinas
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The office review of the submitted administrative JPMs indicated they met all NRC guidelines. During validation of the administrative JPMs it appeared that the administrative JPMs were well planned and prepared. However, difficulties were encountered during administration of the administrative JPMs to the applicants, indicating a lack of preparation and validation by facility instructors. The following problems were encountered by NRC examiners while administering the examination:
During validation of one administrative JPM involving temporary procedure
changes, NRC examiners determined one of the JPM cues appeared to reveal a critical task associated with the JPM and requested that this cue be modified to ensure the critical task would not be revealed by the cue. The requested change was inadvertently overlooked and the cue was left in the JPM. One examiner read the cue to an applicant and subsequently realized the change had not been made. The examiner believed the JPM was compromised and did not grade the JPM. Two temporary procedure change questions were written and administered to the applicant in place of the compromised JPM. The other examiner noted the failure to replace the cue and changed the cue while administering the examination to prevent compromise of the administrative JPM for his applicants.
The applicants were given a faulted surveillance test and directed to evaluate
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the test paper work. The intended result was for the applicants to declare a high pressure inject!on pump inoperable due to the data taken during the surveillance. As the applicants reviewed the surveillance they noted several
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admir.istrative deficiencies in the surveillance package. These deficiencies
!ncluded: taking equipment readings at 0800 before the surveillance was issued at 0840, improperly initialing steps, leaving data spaces blank, and incorrect instrument readings. Each applicant noted several administrative deficiencies. When the applicants reached the matrix in the surveklance where the high pressure injection pump readings were to be recorded, they indicated that an insufficient number of readings was taken. They stated 20 minutes of readings were taken, but 25 minutes of readings were required.
Each indicated the test was invalid and would have to be perfortned again.
The administrative JPM did not meet the desired end point because of errors in the surveillance. This was an example of poor validation and preparation of the administrative JPMs. The examiners directed the facility trainers to rewrite this administrative JPM to reflect the actual performance by the applicants. Examiners believed applicant performance demonstrated competence in reviewing surveillance tests and accepted the results of the administrative JPM.
Applicants were given a set of initial conditions and told they were the
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emergency director. They were to classify the event based on the initial conditions and make any protective action recommendations associated with the event. Insufficient data was provided in the initial conditions to allow the applicants to complete the off site dose calculations. During administration of this administrative JPM, examiners had to request the current Xenon concentration to allow completion of the JPM. The Xenon concentration should have been made available in the initiating cue and was an additional example of poor validation of tha administrative JPMs.
During the validation of the examination material, it was noted that the
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administrative JPM materials were ready to administer to the applicants; however, a correct answer key for the administrative JPMs was not provided until after it was requested by NRC examiners. In one case NRC examinors had to request the grading key several times.
One common applicant weakness was noted during the administrative JPMs.
Applicants were given a watch list that met the minimum fire brigade requirements and told an equipment operator had to leave for a medical emergency. They were directed to make a call out to restore the shift to the minimum required manning levels. The applicants indicated the provided shif t watch list did not meet minimum fire brigade requirements before the equipment operator left and would have called out two operators to complete the shift. The applicants f ailed to count the non-technical specification required SRO in the control room as a fire brigade member, c,
Conclusions The list of administrative JPMs provided to the NRC indicated creative thought and good preparation. The administration of the administrative JPMs revealed significant weaknesses in the preparation and validation of the administrative JPMs
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and the preparation of correct answer keys. Applicants appeared well prepared for this portion of the examination.
05.4 Ooeratina Job Performance Measutu a.
Examination Scong Using NUREG 1021, Operator Licensing Examiner Standards, examiners reviewed each JPM. F.ach JPM was reviewed for applicability, importance, and safety significance. The aggregate of the JPMs was reviewed to ensure all required areas of the JPM examination were represented. JPM follow up questions were reviewed for applicability and to determine if they were considered direct look up questions, b.
Observations and Findinos The licensee submitted JPM: that met standards prescribed in NUREG 1021 and GL 95 06 for addressing the various types and numbers of functional safety systems, alternate path and shutdown plant requirements. Critical tasks were correctly identified within the JPMs.
The JPM questions were considered memory level questions by the NRC examiners.
Facility trainers correctly stated that the questions were not direct look up questions; however, the questions were of the type that are required knowledge of an applicant to be able to properly execute the responsibilities of their job assignrnents. An example of one question was, " explain how to take mar.ual control of decay heat flow through DH 14A or DH 14B from the control room after SA 3 has been actuated." The answer to this question was difficult to locate in any reference material and was marked as open reference by the facility instructors.
A competent operator needs to know this information without reference in order to complete required actions in a timely manner during a plant event. Since this type of knowledge is required to be memorized, NRC examiners disallowed the use of references to this and several comparable questions.
One JPM required the use of a special tool to work in the Engineered Safety Features Actuation System (ESFAS) cabinets. The simulator did not have a duplicate of the control room tool and expected appliccnts and licensees enrolled in requalification training to use a straightened paper clip. In order to provide effective training the simulator should duplicate the control room as closely as possible, c.
Conclusions
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Facility instructors presented a good package of JPMs that met all NRC requirements for use to examine SRO upgrada candidates except for the JPM follow up questions. Most JPM follow up questions did not meet the criteria for open reference questions. Operators were well prepared for the operating JPM portion of the examination.
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05.5 Dynamic Simulator Scenarios a.
Examination Sconjt Using NUREG 1021, Operator Licensing Examiner Standards, excminers reviewed each dynamic simulator scenario. Each scenario was reviewed foi content, applicability, and safety significanca, b.
Observations and Findinas The Integrated dynamic scenarios submitted by the licensee had the minimum number of required malfunctions for each applicant. The initial conditions were different for each scenario and included a low power scenario.
The effice review of the scenarios showed that the submitted scenarios were acceptable as written; however, examiners requested one minor modification to one scenarlo to ensure compliance with NUREG 1021 guidelines. During validation of the scenarios, examiners directed facility instrxtors to modify the sequence of one scenario. The change was necessary to increase the probability that the required number of evaluated events would be completed prior to a reactor scram or safety system initiation, c.
Conclusions The dynamic simulator scenarios submitted by the facility instructors rnet all requirements of NUREG 1021. The as submitted scenarios did not include equipment or instrumont failures after the major transient had occurred. The applicants were well prepared for the dynamic simulator scenario portion of the examination. Communications between applicants during dynamic almulator scenarios was improved when compared to previous initiallicensa classes.
05.6 Post Examination Activities The licensee informed the NRC Chief Examiner by telephone that there were no post examination comments for the written examination. There were two clarifications issued by the facility monitor while the written examination was being administered.
Question #47 Clarification given to all three applicants that all
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parameters in choices a d were primary side parameters.
Question #62 Clarification given to all three applicants that choice c
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should be changed from 130*F to 1030*F.
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05.7 Simulator FidgHiy
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Examiners observed some simulator modeling deficiencies during the examination administration. Examiners and f acility instructors were able to provide cues to the
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applicants to disregard the erroneous indications where applicable. The examiners concluded the identified deficiencies did not preclude completion of valid
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evaluatione of license applicant performance. Simulator deficiencies are
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documented in Enclosure 2, Simulation Facility Report.
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V. Manaaement Meetinas X1 Exit Meetina Summarv The chief examiner presented the examination team'n observations and findings to members of the licensee's management on July 10,1997. The licensee acknowledged the findings presented. No proprietary information was identified during the examination or at the exit meeting.
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PARTIAL LIST OF PERSONS CONTACTED Licensee D. Esholman, Manager, Operations J. Freets, Manager, Regulatory Affairs D. Imlay, Superintendent, Operations J. Lash, Plant Manager R. Simpkins, Superintendent, Operations Training NEG
- S. Stasek, Senior Resident Inspector
'L. Vick, Examiner (HOLB)
j K. Zellers, Resident inspector
(*) Personnel not in attendance at the exit meeting on July 10,1997.
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ITEMS OPENED, CLOSED, AND DISCUSSED Opened 50-346/97308 01 URI deviations from procedures may be approved by Shift Supervisor without additional review per DB OP-00000, Conduct of Operations 50 346/97308 02 NCV examination security compromise
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Enclosure 2 SIMULATION FACILITY REPORT Facility Licensee: Davis Besse Nuclear Power Station
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Facility Licensee Docket No: 50 346 Operating Tests Administered: July 08 09,1997 The following documents observations made by the NRC examination team during the July 1997,initiallicense examination. These observations do not constitute audit or inspection findings and are not, without further verification and review, indicative of non compliance with 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 During the conduct of the simulator portion of the operating tests, the following item was observed:
ITEM DESCRIPTION
During administration of a JPM involving placing spent fuel pool purification in service while using decay heat removal, reactor coolant pressure suddenly and inexplicably went to 2500 psig.
At the end of the third scenario with a reactor coolant leak, secondary plant parameters started to give operators erroneous readings. Simulator operators were not sure of the actual cause.
Subsequent scenario events caused a rar;d shutdown of the plant and prevented the erroneous readings from affecting tha outcome of the scenario.
While parforming a JPM to take ittims out of ESFAS Shutdown bypass, the caro on ESFAS channels 1 and 3 moved whan the tool was inserted into the hole that remover, the component from shutdown bypasa, making it difficult for the applicant to seceive the correct indicatio _
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ES 401 Site specific Written Examination Form ES 401 1 Cover Sheet
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l U. S. NUCLEAR REGULATORY COMMISSION
SITE SPECIFIC
WRITTEN EXAMINATlON l
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APPLICANT INFORMATION
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i Name: MASTER Region:
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J Date: 07/10/97 Facility / Unit: Davis Besse License Level:
SRO Reactor Type:
BW INSTRUCTIONS
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Use the answer sheets provided to document your answers. Staple this cover cheet on
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top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires a final grade of at least 80 percent.
Examination papers will be picked up 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after the examination starts.
All work done on this examination is my own. I have neither given not received aid.
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Applicant's Signature RESULTS Exa.nination Value 100 Points
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Applicant's Score Points Applicant's Grade Percent
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i SENIOR REACTOR OPERATOR Page 2 ANSWER SHEET Multiple Choice (Circle or X your choice)
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If you change your answer, write your selection in the blank, MULTIPLE CHolCE 023 a b c d i
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008 a b c d 031 a b c d 009 a b c d.
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012 a bc d 036 a b c d 013 a b c d 036 a b c d 014 a b c d
'037 a b c d 015 a b c d 038 a b c d 016 a b c d 039 a b c d 017 a b c d 040 a b c d
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018 a b c d 041 - a b c d-
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SENIOR REACTOR OPERATOR Page 3 ANSWER SHEET Multiple Choice (Circle or X your choice)
If you change your answer, write your selection in the blank.
MULTIPLE CHOICE 068 a b c d_
046 a b cd 069 a b c d 047 a b cd 070 a b c d 048 a bc d 071 a b c d 049 a bc d 072 a b c d 050 a bc d 073 a b c d 051 a b c d 074 a b c d 052 a b c d 075 a bc d
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053 a bc d 076 a b c d 054 a b c d 077 a b c d 055 a bc d 078 a b c d
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056 a b c d 079 a b c d 057 a b c d 080 a b c d 058 a b c d-081 a b c d 059 a b c d 082 a b c d 060 a bc d 083 a b c d 061 a b c d 084 a b c d 062 a b c d 085 a b c d 063 a bc d 086 a b c d
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i SENIOR REACTOR OPERATOR Page4 ANSWER SHEET Multiple Choice (Circle or X your choice)
J If you change your answer, writo your selection in the blank MULTIPLE CHOICE 091 a b c d
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(' ' ' ' ' ' ' ' ' ' END OF EXAMIN ATION * * * * * * * * * * )
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i SENIOR REACTOR OPERATOR Page 5 I
i ES 402 Policies and Guidelines Attachment 2 l
for Taking NRC Written Examinations i
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j 1.
Cheating on the examination will result in a denial of your application and could result l
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in more severe penalties, i
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Af ter you complete the examination, sign the statement on the cover sheet indicating j
that the work is your own and you have not received or given assistance in completing j
the examination.
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3.
To pass the examination, you must achieve a grade of 80 percent or greater.
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The point value for each question is indicated in parentheses after the question
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There is a time limit of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for completing the examination.
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Use only black ink or dark poncil to ensure legible copies.
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7.
Print your name in the blank provided on th; examination cover sheet and the answer i
sheet, i
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Mark your answers on the answer sheet provided and do not leave any question blank.
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9.
If the intent of a question is unclear, ask questions of the examiner only.
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10.
Restroom trips are permitted, but only one applicant at a time will be allowed to leave.
Avoid all contact with anyone outside the examination oom to eliminate even the
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appearance or possibility of cheating
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11.
When you complete the examination, assemble a package including the examination questions, examination aids, and answer sheets and give it to the examiner or proctor.
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9emember to sign the statement on the examination cover sheet.
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After you have turned in your examination, Isave the examination area as defined by he examiner.
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Minimum RCS P/T to provide NPSU vith two RCPs operating in a loop S.
Saturation Curve 6.
Abnohnal transient envelope-Hote 1 -
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Vide Range Pressure 'Indlcoor PI-RC2A4 or PRS-RC2A1 or v/32 or PI-RC2B4 or P724
,
Curves 3 and 4 Vith RCS pressure greater than 500 psig Use RCS Vide Range Pressure Indicators PI-RC2A4 or PRS-RC2A1 or P732 or PI-RC2B4 or P724 Vith RCS Pressure less that 500 psig use RCS Hot Leg Lov Range Pressure indication from PI-RC2&6
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.
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s 242 DD-OP.02000 Ravision 05 FIGURE 2 Incore T\\C Temperature vs RCS Pressure for ICC 2600
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Figure 3 Sheet 1 of 1
.
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_________ ______ _ _ _
SENIOR REACTOR OPERATOR Page 6 QUESTION: 001 (1.00)
A licensed operator actively performed shift duties in the control room per the following:
Day 1:
0600 1300 Day 2:
0800 1600 Day 3:
0600 1800 Day 4:
1700 2300 Day 5:
0600 1400 Assuming Day 1 is the first day of the quarter, what is the MINIMUM number of additional hours required for the operator to maintain an active license?
a.
b.
c.
d.
..
QUESTION: 002 (1.00)
You are the Shift Supervisor. You were just informed th.it the P9 P was selected to participate in an unannounced drug ocreening test. You direct the RC nom the relief shif t to relieve the RO P so he can go to the drug screening test. Which ONE of the following statements identifies the requirements for the watch relief?
a.
if the relief did NOT attend the shiit turnover, then he must perform a complete shift turnover per DB-OP-00100, Shif t Turnover and sign into the Unit Log as a RO.
b.
If the relief attended shift turnover and has been in the control room continuously since shif t turnover, he may relieve the RO P temporarily without any additional actions.
c.
If the relief attended shif t turnover and is currently signed into the Unit Log as an RO, only an additional entry in the Unit Log documenting the transfer of the RO function is required, d.
If the relief did NOT attend shif t turnover, he must, as a MINIMUM, walkdown the CTRM panels with the off-going RO, discuss evolutions in progress, notify the CTRM SRO and sign into the Unit Lo _ _ _ _ _ _ _ _ -
..
.
-
.
+
i SENIOR REACTOR OPERATOR Page 7
'
OUESTION: 003 (1.00)
l When a task listed on the Month!) Activity Log (MAL) is NOT performed satisfactorily, the l
places an X in the initial block and prepares and signs comments on that l
zone's MAL Comment Sheet, a.
Assistant Shift Supervisor or responsible Zone Operator b.
Shift Manager or Reactor Operator c.
Shif t Supervisor or Reactor Operator d.
Shift Supervisor or SRO designee QUESTION: 004 (1.00)
An estimated critical rod position (ECP) has been calculated for a reactor startup that is to be performed 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after a trio from a 60 day full power run. Which ONE of the following events or conditions will result in the actual critical rod position being lower than the ECP?
..
a.
The startup is delayed for approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, b.
The main steam header pressures are decreased by 35 psig just prior to criticality.
Steam generator feedwater addition rate is reduced by 5 percent just prior to c.
criticality, d.
A new boron sample shows a current boron concentration 20 ppm higher than that used in the ECP calculation.
_-
.__
_
._.- _
_
_
_
. _...
SENIOR REACTOR OPERATOR Page 8 QUESTION: 005 (1.00)
The Shift Supervisor has declared an Alert. In the absence of the Emergency Director, the Shift Supervisor will turn over the role of Emergency Director to the who is stationed in the which is located in the
,
a.
Emergency Plant Manager; Emergency Control Center; DBAB b.
Emergency Plant Manager; Technical Support Center; DBAB c.
Emergency Assistant Plant Manager; Technical Support Center; DBAB d.
Emergency Plant Manager; Technical Support Center; PSF 2nd floor lunch room
.
QUESTION: 000 (1.00)
According to the general isolation guidelines outlined in DB OP 00016, Removal and Restoration of Station Equipment, in order to isolate a Main Fend Pump the sequsnce should be
..
a.
1.
Isolate the steam supply.
2.
Isolate the suction line.
3.
Vent and drain the pump.
4.
Isolate the auxiliary systems, b.
1.
Isolate the discharge line.
2.
Isolate the steam supply.
3.
Isolate the suction line.
4.
Isolate the auxiliary systems, c.
1.
Isolate the discharge line.
2.
Isolate the recirculation line.
3.
Isolate the suction line.
l 4.
Vent and drain the pump, d.
1.
Isolate the steam supply.
2.
Isolate the recirculation line.
3.
Isolate the auxiliary systems.
4.
Isolate the discharge lin }
SENIOR REACTOR OPERATOR Page 9 QUESTION: 007 (1.00)
Which one of the following activities would require a change to clearance versus using a temporary lift of a safety red tag?
a.
Fan belt running checks.
b.
A leak check of a piping system, c.
Installing a megger test unit into a load breaker cubicle.
d.
Racking a circuit breaker into the test position while the breaker is required for personnel protection.
QUESTION: 008 (1.00)
The dose rate in Mechanical Penetration Room (MPR) 3 is at the limit for an area posted as a
- i Radiation Area. Which ONE of the following is the exposure an operator will receive if he works in MPR 3 for 30 minutes?
a.
2.5 mrem
-
b.
50 mrem c.
100 mrem
d.
500 mrem QUESTION: 009 (1.00)
'
You have received.410 R TEDE this year at DBNPS. While you were on a temporary assignment to Oconee Nuclear Station earlier this year, you received.220 R TEDE. You have had two chest x-rays during the year giving an additional.050 R TEDE. Which of the following is the maximum amount of whole body radiation you are allowed to receive a? Davis Besse Station during the remainder of the year without prior approval of the Manager -
Radiation Protection?
a.
0.590 R TEDE b.
1.320 R TEDE c.
1.370 R TEDE d.
3.590 R TEDE
-
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,
i SENIOR REACTOR OPERATOR Page 10
i
'
j QUESTION: 010 (1.00)
if The Emergency Key Locker, located in the CTRM, may be utilized as plant conditions dictate
and as approved by the Shift Supervisor, Assistant Shift Supervisor, or
.
[
a.
Shift Supervisor Administrative Assistant
b.
Nuclear Security Shift Supervisor c.
Reactor Operator j
d.
Shift Manager
!~
i OUESTION: 011 (1.00)
f
A surveillance test has just been completed and restoration includes the racking out of a 4160
VAC breaker and the removal of control power fuses. The individual assigned to perform the
,
racking evolution shall be qualified as at least an and should remove any
j objects hanging from his belt due to the potential shock hazard from
.
_
..
a.
Auxiliary Operator; metal; 125 VDC
b.
Equipment Operator I; metal; 120 VAC -
c.
Equipment Operator I; metal; 125 VDC d.
Equipment Operator I; plastic; 125 VDC
i
!
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.
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"
SENIOR REACTOR OPERATOR Page 11
- QUESTION: 012 (1.00)
Limit and Precaution 2.2.6 of DB-OP-06210, CO2 and Hydrogen Procedure, states that "...
if Hydrogen purity drops to below 90%, the Generator must be shut down and the Hydrogen purged out with CO2." The reason for this requirement is
.
a.
the reduced purity createsthe possibility of an explosive mixture of hydrogen in-air b.
the Seal Oil System pressure regulator will not function properly at reduced purities c.
the stator cooling water intet pressure regulator will not function properly at reduced puritiet t-d.
reduced purity will result in inefficient operation of the main generator due to I
increased friction I
!
QUESTION: 013 (1.00)
i
..
l An HOURLY Operating Specification Fire Watch (OSFW)in EDG Room #2 has been ongoing i
for 3 days when it is discovered that the individual responsible for performing the fire watch
_ has NOT checked the firewatch post for 82 minutes. Which ONE of the following describes the IMMEDIATE action that should be performed per DB-FP-00009, Fire Protection impairment j
and Firewatch?
j a.
Notify the Manager, Plant Operations and process a PCAOR l
b.
Notify the Shift Supervisor, proceed to the firewatch post and perform the fire
'
watch.
[
_ c.
No immediato action is required. Send the fire watch to perform the inspection l:
within the next eight minutes.
d.
Contact Security to have a card reader check of the area performed. Anyone -
entering / leaving the area can sign for the fire watch surveillance.
L
4
1
.
.
SENIOR REACTOR OPERATOR Page 12 QUESTION: 014-(1.00)
OPERATOR AIDS, as described in DB-OP-00004, Operator _ Aids Control,
,
a.
should always be considered as requirements and, therefore, should be strictly adhered to b.
are always considered CONTROLLE') documents' and, therefore, can take the place of approved procedures c.-
can be used as a temporary alteration to an approved procedure until the procedure has undergone the alteration process d.
are usually considered INFORMATION ONLY documents and, therefore, can NOT take the place of approved procedures, unless otherwise specified QUESTION: 015 (1.00)
The specific activity of the primary coolant has exceeded Technical Specification limits and a reactor shutdown and cooldown to less than 530 deg F Tavg has been performed. Which "
ONE of the following was the basis for cooling the Reactor Coolant System to less than 530 deg F?
a.
Prevents the release of activity should a steam generator tube rupture occur.
b-Seals the fuel cladding failure thereby reducing the amount of activity released in the event of a LOCA.
c.
Allows the lodine to react with the metal in the Reactor Coolant System, minimizing the release of iodine activity in the event of a LOCA.
d.
Minimizes the amount of corrosion and erosion of the Reactor Coolant System and the fuel elements caused by the increase of lodine activity and particulate in the coolan. _ _
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j SENIOR REACTOR OPERATOR Page 13
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QUESTION: 016-(1.00)
-
During a plant startup a heat balance was performed at 55% power. The results of this l
calculation agreed with indicated NI power. After power was increased to 95%, a NEW heat
balance calculation was performed. It does not agree with indicated Ni power. Which ONE
of the following describes the EXPECTED relationship between indicated Ni power and
calculated power, and the reason for the difference?
a.
Indicated Ni power will be less than calculated power due to the decreasing steam generator superheat region.
'
b.
Indicated NI power will be less than calculated power due to the decrease in
<
Tcold water temperature.
,
c.
Calculated power will be less than indicated NI power due to increased ambient losses.
d.
Calculated power will be less than indicated NI power due to an increase in Tcold water temperature.
.
QUESTION: 017 (1.00)
i Upon declaration of a Tornado WARNING, which ONE of the following describes the required status of the Emergency Diesel Generators (EDGs)?
a.
Idle speed operation b.
C1 or D1 parallel operation
.-
c.
C1 and D1 parallel operaticn i
'
d.
Isolated bus operation on C1 end D1 (AC110 and AD110 open)
!
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i f
%
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SENIOR REACTOR OPERATOR Page 14
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QUESTION: 018 (1.00)
,
!
Given the following conditions:
.
The plant is operating at 80% power
-
All plant systems are operating normally
-
l A loss of power to 120 VAC Essential Power Panel Y2 has just occurred
-
j What is the expected effect on the RPS system?
a.
RPS Channel 1 will trip and CRD Breaker B will open
,
{
b.
RPS Channel 1 will trip and all CRD Breakers will open c.
RPS Channel 2 will trip and CRD Breaker A will open
,
d.
RPS Channel 2 will trip and all CRD Breakers will open
i QUESTION: 019 (1.00)
,
,
..
!
Which ONE of the following automatic responses will result from the loss of RCP 1-1 (Loop 1) while at 85% power? (All systems are in automatic control with FOUR RCPs initially operating.)
a.
Tave input to ICS from Loop 1 selected and delta Tcold near zero.
b.
A 2.4:1 ratio of SG feed flow and the ICS in a runback to 75% at 20% per
'
mir.ute.
c.
A runback to 75% ULD at 20% per minute and the affected loop SG being on low level limits, d.
Reactor power reduced to less than 75% and the ICS upper load 16mit automatically reset to 80%.
,
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i h
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SENIOR REACTOR OPERATOR Page 15 i
QUESTION: 020 (1.00)
-
l The following plant conditions exist:
72% RTP
-
!
RCP 2-2 secured due to oil cooler leak
-
Due to Makeup System operating problems the'following Reactor Coolant Pump parameters are observed with the exception of pump 2-2 which is NOT running:
'
,
j 1-1 1-2
!-
Seal Rtrn Temp 180 deg F 185 deg F -
175 deg F
Seal Inj Flow 2.4 gpm 3.5 gpm 2.7 gpm -
1:
. Seal Rtrn Flow 1.8 opm.
2.0 gpm 1.5 gpm
-
Seal Diff Press 600 psid 550 psid 650 pcid Which ONE of the following actions is required because of the above Reactor Coolant Pump
!
indications?
-
,
!
a.
TRIP the reactor, STOP the 2-1 Reactor Coolant Pump, GO TO DB-OP-02000,
]
.RPS, SFAS, SFRCS Trip or SG Tube Rupture.
!-
b.
TRIP the reactor, STOP the 1-2 Reactor Coolant Pump,'GO TO DB-OP-02OOO, RPS,'SFAS, SFRCS Trip or SG Tube Rupture.
,
i c.
TRIP the reactor, STOP the 1-2 Reactor Coolant Pump, verify proper feedwater
,
flow ratios and transfer of Tave control.-
i d,
- REDUCE reactor power to 45%, STOP the 1-2 Reactor Coolant Pump, GO TO l
DB-OP-02000, RPS, SFAS, SFRCS Trip or SG Tube Rupture, i
l 4-
i
$
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.
...
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_
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_
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___.
_ _.... - _ _.. - _ _.. _. ~ _.. _..
. _ _ _.,. _ _ __ _,
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SENIOR REACTOR OPERATOR Page 16
!
QUESTION: 021 (1.00)
.
!
Which ONE of the following describes a design aspect of the letdown portion of the Makeup and Purification Systam?
'
a.
The letdown conlers provide TWO primary functions, cooling of the letdown l
fluid and N-16 decay.
e b.
The letdown line is sized to limit a LOCA outside containment to within the capacity of ONE makeup pump.
l c.
The letdown HIGH temperature isolation (160 deg F) isolates allletdown flow except 3 gpm to RE1998 (Failed Fuel Monitor),
d.
The RCS inlet valve (MU1 A or 1B) to EACH letdown cooler OPENS AFTER and
!
CLOSES BEFORE the coolers CCW supply valves (CC1409 or 1410).
,
QUESTION: 022 (1.00)
A plant heatup from COLD SHUTDOWN is in progress. Which ONE of the following describes "
the consequences of the operators failing to MANUALLY RESET any of the SFAS RC
"
PRESSURE LO LO TRIP bistables as the plant heats up?
l a.
An SFAS trip willinitiate at approximately 450 psig.
'
i b.
An SFAS trip willinitiate at approximately 600 psig.
c.
An SFAS trip willinitiate at approximately 1650 psig,
.
d.
The plant will be without low pressure SFAS protection until the bistables are
'
manually reset.
!
.
4
T l
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.-
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--.
-
.--
-...
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,
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SENIOR REACTOR OPERATOR Page 17
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{
QUESTION: 023 (1.00)
At DBNPS, the Emergency Core Cooling Systems are designed such that
,
a.
CFTs will provide a sufficient volume of borated water to be forced into the Rx vessel at less than 600 psig in the RCS b.
above 200 deg F, two independent ECCS subsystems are required to ensure sufficient core cooling even in the event of a single failure c.
Trisodium Phosphate loaded into baskets in CTMT will ensure a post-LOCA pH of less than 3.0 d.
the BWST water volume and boron concentration are sufficient to maintain post LOCA CTMT pH between 5.0 and 13.0, which will minimize the evolution of Xenon gas QUESTION: 024 (1.00)
The following plant conditions exist:
The plant is operating at 100% power.
-
Reactor Engineering reports that the results of yesterday's incore flux map
-
indicate that Control Rod 4-1 is fully inserted into the core.
CTRM API for Rod 4-1 Indicates 100% withdrawn.
-
CTRM RPl for Rod 4-1 indicates 100% withdrawn.
-
Which ONE of the following actions should be performed?
a.
Trip the reactor and GO TO DB-OP-02000, b.
Declare the rod inoperable and remain at 100% power while evaluating, c.
Declare the rod inoperable and reduce power to less than 60% while evaluating.
d.
Commence a rapid shutdown to HOT STANDBY and GO TO DB-OP-02504, Rapid Shutdown.
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SENIOR REACTOR OPERATOR Page 18 OUESTION: 025 (1.00)
,
Which ONE of the following is an indication that Intermediate Range NI 3 is OVER-I compensated, during a reactor startup?
l l
a.
NI 3 reads 5 E 11 amps NI-4 is not on scale l
b.
NI 3 reads 3 E-6 amps NI-4 reads 8 E 7 amps I
i c.
NI 1 and 2 read 8 E + 5 cps NI-3 reads 2 E 11 amps l
l-d.
NI 1 and 2 read 3 E+4 cps NI 3 reads 8 E 11 amps
QUESTION: 026 (1.00)
.
l The following plant conditions exist:
I Reactor power is 50%.
-
BOTH Intermediate Range (IR) Ni detectors have failed HIGH.
-
The IR detector repairs will take 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to complete.
-
..
Which ONE of the following actions must be performed?
a.
POWER OPERATION may continue with no limitations.
!
'.
b.
POWER OPERATION may continue, but power must be maintained greater than
'
5%.
.
c.
Within ONE hour actions must be taken to place the unit in HOT STANDBY with l
Rx trip breakers open within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
l d.
Within ONE hour action must be taken to place the unit in COLD SHUTDOWN with Rx trip breakers open within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
l
f f
.. -.. -
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_ _. ~..
- - -.
.. -. -.. _ - -.. - -
_ - - - - -
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-
- -.. -,
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SENIOR REACTOR OPERATOR Page 19 -
QUESTION: 027 (1.00)
The plant has experienced a LOCA. Plant conditions are such that the upper portion of the core may be uncovered. Which ONE of the following indications could be used to confirm this condition?
a.
Both Source Range Nis reading 4000 counts and slowly decreasing b.
Average incore thermocouple temperatures = 586 deg F and decreasing rapidly, c.
Three incore thermocouples showing negative Subcooling Margin with two Post Accident Monitoring channels operable, d.
Incore Self Powered Neutron Detectors (SPND) showing large positive currents in the upper portion of the core, with SPND output in lower portion of the core remaining constant.
QUESTION: 028 (1.00)
.
The CTMT Air Cooling System, combined with the CTMT Spray system comprise the CTMT Heat Removal system. The purposes of this system are to keep CTMT temperature less than 120 deg F during and removal of post LOCA airborne
,
a.
a Design Basis LOCA: Xenon b.
a Design-Basis LOCA; todine c.
normal power operations; Xenon d.
normal power operations; lodine
.
.
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SENIOR REACTOR OPERATOR Page 20 QUESTION: 029 (1.00)
Following a primary LOCA, BWST LO LO LVL XFER TO EMER SUMP (5-3 A) is received.
Which ONE of the following describes the required operator actions?
a.
BLOCK cnd SHUT the suction c.' the DH and CS pumps to the emergency sump (DH-9A/B) and BLOCK and CvEN the BWST outlet valves (DH 7A/B).
b.
BLOCK and OPEN the suction of the DH and CS pumps to the emergency sump (DH 9A/B) and then BLOCK and SHUT the BWST outlet valves (DH 7A/B).
,
c, BLOCK and OPEN the suction of the DH and CS pumps to the emergency sump (DH 9A/B) which automatically CLOSES the BWST outlet valves (DH 7A/B).
d.
BLOCK and OPEN the BWST outlet valves (DH 7A/B) which allows automatic opening of the suction of the DH and CS pumps to :he emergency sump (DH 9A/B).
QUESTION: 030 (1.00)
-
Which ONE of the following describes the interlocks associated with STARTING and RUNNING -
a condensate pump?
a.
The pump will TRIP or be prevented from starting if hotwelllevel goes below 24 inches, b.
The pump will TRIP or be prevented from starting if vacuum becomes less than 12 inches of mercury, c.
The pump will TRIP if hotwell level goes below 36 inches. This interlock is BYPASSED for 60 seconds to allow pump START.
d.
The pump will TRIPif vacuum becomes less than 24 inches of mercury. This interlock is BYPASSED for 60 seconds to allow pump START.
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SF.NIOR REACTOR OPERATOR Page 21
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-A QUESTION: 031 (1.00)
,
The following plant conditions exist:
Containment pressure is 20 psia and increasing.
-
OTSG #1 pressure is 600 psig decreasing rapidly.
-
OTSG #2 pressure is 655 psig decreasing slowly.
-
OTSG #1 levelis 5 inches on the SU range, decreasing rapidly.
-
OTSG #2 levelis 50 inches on the SU range, increasing slowly.
-
Which ONE of the following describes the operation of the Steam Feed Rupture Control
-
System (SFRCS)?
a.
Initiates AFW and level setpoint for both OTSGs will be 49 inches.
b.
Initiates AFW and level setpoint for both OTSGs will be 124 inches.
I c.
Isolates MFW to OTSG #1, initiates AFW and level setpoint for #2 OTSG will be 55 inches.
d.
Isolates MFW to OTSG #1, initiates AFW and level setpoint for #2 OTSG will
i be 130 inches.
-
t.
QUESTION: 032 (1.00)
Which ONE of the fonowing would requirc immediate initiation of MU/HPl cooling in accordance with DB OP-02000, RPS, SFAS, SFRCS TRIP or SG Tube Rupture, Specific Rule 2, MU/HPl Flow initiation, Throttling, and Termination?
a.
RCS Subcooling margin is 25 deg F.
b.
RCS pressure is less than 1650 psig.
c.
Loop 1 hot leg temperature is 605 oeg F, AND both AFW pumps are abgned to SG 2.
d.
Hot leg temperatures are 560 deg F, all feedwater is lost AND D1 Bus is locked out.
..
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l SENIOR REACTOR OPERATOR Page 22 i
OUESTION: 033 (1.00)
Auxiliary Feedwater has automatically started and each train is supplying its own OTSG.
Which CNE of the following describes the response of AF 6451 and AF 6452 (SG Level Control Valves) controllers when the automatic initiation signal clears?
l e.
Shifts to MANUAL as soon as the AUTO initiation signal clears.
!
b.
Shifts to MANUAL after the AFPT teaches the HIGH Speed Stop (HSS).
c.
Continues to operate on level control until the MANUAL pushbutton is depressed.
l d.
Continues to operate on speed control until the AUTO initiation signal is reset.
.
p
!
QUESTION: 034 (1.00)
,
The cavitating venturis in the Auxiliary Feedwater (AFW) System are installed to
--
.
l a,
limit AFW flow to 800 gpm to EACH steam generator
!
- b.
limit AFW flow to 1050 gpm to EACH steam generator
c.
ensure AFW flowis maintained greater than 800 apm to EACH steam generator
,
d.
ensure AFW flow is maintained greater than 1050 gpm to EACH steam generator f
i
&
i f
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l SENIOR REACTOR OPERATOR Page 23 OUESTION: 035 (1.00)
Which ONE of the following sets of DC buses, when lost, will result in an imtrediate reactor TRIP, if operating at 100% power?
a.
D1N and DAN b.
D2N and DBN c.
D1P and DAP d.
D2P and DBP QUESTION: 036 (1.00)
When a HIGH alarm comes in on an AREA radiation monitor, the local alarm and indicating panel (if so equipped) will alarm and
.
a.
there will be no other alarms associated with the AREA monitor, b.
the CTRM module's red light willlight on the Radiation Monitoring Panel but its amber light will be off.
-
c.
the CTRM module's red lignt willlight on the Radiation Monitoring Panel only if it is a Tech Spec required mauitor, d.
the CTRM module's amber and red lights willlight on the Radiation Monitoring Panel _ and the alarming monitor will be displayed on the CTRM Fire and Radiation CRT.
QUESTION: 037 (1.00)
Which ONE of the following sets of conditions requires tripping of all RCP's during a smali break LOCA?
Temperature RCS pressure Ave T/C Temp (Tl-RC4A2)
(P732)
a.
579 deg F 1800 psig 590 deg F b.
535 deg F 1300 psig 540 deg F c.
525 deg F 1000 psig 540 deg F d.
460 deg F 700 psig 445 deg F
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.-
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SENIOR REACTOR OPERATOR Page 24 QUESTION: 038 (1.00)
l A LOCA has occurred and HPl has AUTOMATICALLY initiated. Which ONE of the following conditions will allow the operator to throttle HPl flow?
a.
HPl flow is 35 gpm with the minimum recirc valves closed, b.
LPI flow has been above 1000 gpm per line for 10 minutes, c.
Subcooling Margin is 25 deg F on Channel 1 and 32 deg F on Channel 2 using Thot, d,
HPl flow is greater than 250 gpm per HPI pump except while LPI piggybacked to the BWST.
QUESTION: 039 (1,00)
Which ONE of the following describes a PORV indication that is available to the operator?
a.
Lighted LEDs ON the PAM panel indicating.25 means that the PORV is FULL --
open.
b.
A red light ON the control switch indicates that tho acoustical detectors have detected flow through the PORV.
c, An extinguished blue light ABOVE the control switch indicates that the PORV solenoid has de-energized to open the PORV.
d.
An illuminated blue light ABOVE the control switch indicates that the PORV pilot solenoid valve has 125 VDC available to open the FOR..
.
,
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.
<
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I SENIOR REACTOR OPERATOR Page 25
,
QUESTION: 040 (1.00)
!
W,'.ich ONE of the following describes the normal position and the purpose of RC 49, PZR MINIMUM FLOW SPRAY VALVE?
~
Normally OPEN to minimize the temperature differential across the PZR spray a.
line.
J b.
Normally CLOSED and is opened to provide emergency spray if RC 2 fails
,
closed.
c.
Normally OPEN to maintain a constant boron concentration between the PZR j-surge line and spray line.
!
d.
Normally CLOSED and is opened to reduce temperature differentialin the spray and surge lines during startup.
QUESTION: 041 (1.00)
Which ONE of the following plant conditions would require the operator to initiate a MANUAL reacic, rRIP?
..
a.
Pressurizer level is 245" and INCREASING in MODE 3.
,
b.
An SFAS Level 1 actuation (HIGH CTMT RADIATION) in MODE 1.
'
c.
An OTSG Tube Leak larger than MAXIMUM MU flow with letdown in service in MODE 1.
.
d.
An overcooling event in excess of Tech, Spec. cooldown rate limits occurs in MODE 3.
!
,
l QUESTION: 042 (1.00)
i When a selected feedwater temperature detector in the ICS fails low, it will be indicated by l
-
!
a.
a decrease in Tave i'
b.
an increase in reactor power c.
a decrease in feedwater flow demand
,
d.
an increase in OTSG Operate Range levelindication
.--
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SENIOR REACTOR OPERATOR Page 26 i
QUESTION: 043 (1.00)
Which ONE of the following conditions will cause an AUTOMATIC TRIP of the S1 and S2
!
switches in the NNI Y cabinet?
i
a.
Loss of voltage from panel YAU.
b.
Loss of the + 24 VDC bus or the 24 VDC bus.
c.
Fuses from YAU and YBU feeding the NNI ABT blow.
d.
Loss of ONE + 24 VDC and ONE -24 VDC power supply.
QUESTION: 044 (1.00)
A Large Break LOCA has occurred, and CTMT conditions are being monitored per DB OP-02000, Table 3. CTMT Hydrogen concentration has reached 3%. Which ONE of the following combinations represents the PREFERRED order of CTMT Hydrogen cornrol systems to be placed in service, per Table 3? (Assume all equipment is on site).
--
1.
Hydrogen Dilution Blowers 2.
Hydrogen Purge System 3.
Hydrogen Recombiner a.
2 then 1 then 3 b.
3 then 2 then 1 c.
3 then 1 then 2 d.
1 then 2 then 3
-.
SENIOR REACTOR OPERATOR Page 27 OUESTION: 045 (1.00)
The following plant conditions exist:
The Reactor has been de-fueled for one week.
-
DH Pump 1 is aligned to the SFP
-
'
SFP temperature is increasing
-
Erratic flow indication exists on FYI DH28, DH Pump 1 Outlet flow
-
The Assistant Shif t Supervisor should initially GO TO which ONE of the following procedures?
a.
DB-OP-06021, Spent Fuel Pool Operating Procedure, b.
DB OP-02527, Loss of Decay Heat Removal Procedure, c.
DB-OP-02003, ECCS Alarm Panel 3 Annunciators, Attachment 3-1 B, SFP LVL, d.
DB OP-06012, Decay Heat and Low Pressure injection System Operating Procedure.
QUESTION: 046 (1.00)
-
The following plant conditions exist:
The reactor is TRIPPED
-
Subcooling Margin is O deg F
-
RCPs are TRIPPED
-
AFW is supplying BOTH SGs at 49 inches
-
RCS pressure is 1700 psig
-
T-cold and SG T-sat are 40 deg F apart
-
Which ONE of the following actions will be consistent with the expected DB-OP-02000 routing for this scenario?
a.
Once SCM is regained, start 2 RCPs in the loop with the highest SG level, b.
Raise SG pressure until secondary T-sat is greater than incore T/C temperature, c.
Open the PORV and reduce RCS pressure until it is 40 to 60 deg F below SG T-sat.
d.
Fully open the AVVs and cooldown at a maximum of 235 deg F/ hour until heat transfer is regained.
+
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SENIOR REACTOR OPERATOR Page 28
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_ QUESTION: 047 (1.00)
i
A normal plant startup is in progress with the reactor CRITICAL at 1.0 E-8 amps power when
,
an atmospheric vent valvo fails OPEN. Which ONE of the following is the expected plant
response? (No operator action.)
,
I
a.
Reactor power willlNCREASE, temperature will DECREASE, and pressure will j
DECREASE.
b.
Reactor power will DECREASE, temperature will DECREASE, and pressure will DECREASE.
'
.
c.
Reactor power willINCREASE, temperature willINCREASE, and pressure will
-
DECREASE.
d.
Reactor power willlNCREASE, temperature will DECREASE, and pressure will
INCREASE.
i QUESTION: 048 (1.00)
l
..
A is provided between_ generator output ground switch 34645 and the to prevent shorting the
,
,
i 4-a.
kirk key interlock; exciter field breaker; main generator output to ground i
b.
kirk key interlock; generator field breaker; main generator rotor to ground
,
c.
kirk key interlock; generator. field breaker; main generator output to ground l-d.
mechanical slide link; generator field breaker; mein generator output to ground
.
i
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. _ _ _. _ _..,
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,
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. SENIOR REACTOR OPERATOR Page 29 i
4
]
QUESTION: 049 (1.00)
The following plant conditions exist:
A complete loss of offsite power
-
BOTH EDGs running
-
1-3-H, BUS D1 LOCKOUT is lit
-
j EDG 1 output breaker AC101 is closed
-
Which ONE of the following is an IMMEDIATE operator action?
,
,
a.
Align Bus D2 to be supplied from EDG 1.
b.
Align bus D2 to be supplied from the SBODG.
c.
Direct an equipment operator to reset the lockout on D1 bus.
"
d.
Shutdown EDG 2 by depressing the local EMERGENCY SilUTDOWN
pushbutton.
.
..
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QUESTION: 050 (1.00)
-
!
The following conditions exist:
The green RESET pushbutton on radioactive monitor RE 600 (MS line RE) was
-
incorrectly installed following a light bulb replacement.
This caused it to bind in the " depressed" position.
-
Which ONE of the following describes how this condition affects the operation of the RE?
<
!'
a.
No effect, the RE is still OPERABLE.
I b.
The ALERT and HIGH lights will be ON.
c.
The ALERT and HIGH lights will NOT function
,
d.
The ALERT light will be ON, the HIGH light will be OFF.
-
.
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SENIOR REACTOR OPERATOR Page 30 QUESTION: 051 (1.00)
Which ONE of the following describes the relationship between SW 1395/1399 (TPCW Heat Exchanger inlet Header Isolation Valves) and CT 2955 (TPCW Heat Exchanger Supply frorn Circ. Water System Stop Valve) operating sequence?
At 30 psig INCREASING SW header pressure, SW 1395/1399 will a.
-
OPEN to supply SW to the TPCW Hxs At 50 psig INCREASING Cooling Water Heat Exchanger SW header
-
pressure, CT 2955 will CLOSE to isolate cire, water to the TPCW Hxs
b.
At 30 psig INCREASING Cooling Water Heat Exchanger SW header
-
pressure, CT 2955 will CLOSE to isolate water to the TPCW Hxs At 50 psig INCREASING SW header pressure, SW 1395/1399 will OPEN
-
to supply to the TPCW Hxs f
At 50 psig DECREASING Cooling Water Heat Exchanger SW header c.
-
.
pressure, CT 2955 will OPEN to supply cire, water to the TPCW Hxs i
At 30 psig DECREASING SW header pressure, SW 1395/1399 will
-
CLOSE to isolate SW to the TPCW Hxs
i d.
At 50 psig DECREASING SW header pressure, SW 1395/1399 will s.
-
CLOSE to isolate SW to the TPCW Hxs At 30 psig DECREASING Cooling Water Heat Exchanger SW header
-
}
pressure, CT 2955 will OPEN to supply circ. water to the TPCW Hxs i
'
.
)
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SENIOR REACTOR OPERATOR Page 31 QUESTION: 052 (1.00)
The following plant conditions exist:
SAC 1-1 running
-
SAC 1-1 receiver pressure 110#
-
EIAC runnin0
-
EIAC receiver pressure 93# and decreasing
-
SA 6445 lA to SA crosstie valve closed
-
These symptoms are a result of which ONE of the following conditions?
a.
Station Air Receiver 1-1 outlet moisture trap failed open.
b.
SAC 1-2 intercooler solenoid operated drain failed open, c.
CC 1495 NON ESSENTIAL CCW TO AUX BLDG ISOLATION instrument air line has ruptured.
d.
A pneumatic gnnder connected to a Turbine Bldg. service connection has a ruptured air hose.
..
QUESTION: 053 (1.00)
The following plant conditions exist:
,
The plant is operating at 100% RTP.
-
The Jockey Fire Pump trips.
-
Annunciator Alarm 9-5-G, FIRE WTR TURB BLDG PRESS LO, is illuminated.
-
Which ONE of the following describes the control room indications if the Fire Protection System pressure continues to DECREASE to 90 psig?
a.
FIRE WTR DSL PMP ON annunciator followed by DSL OIL STRG TK LVL annunciator, b.
FIRE WTR ELEC PMP ON annunciator followed by FIRE WTR DSL PMP ON annunciator, c.
FIRE WTR STRG TK LVL annunciator followed by FIRE WTR ELEC PMP ON annunciator, d.
FIRE WTR ELEC PMP ON annunciator followed by FIRE WTR ELEC PMP SYS TRBL annunciato SENIOR REACTOR OPERATOR Page 32 QUESTION: 054 (1.00)
Which one of the follow!ng complies with Specific Rule 2 of DB-OP-02000, RPS, SFAS, SFRCS Trip or SG Tube Rupture, concerning piggyback operation of the MU/HPl System?
a.
MU flow shall NOT exceed 250 gpm during piggyback operation from LPI with
'
suction on the BWST b.
If LPI pump suction is from the CTMT Emergency sump, MU pumps shall be operated ONLY in piggyback mode If there is LPI flow into the RCS dur:ng an LOCA and LPI suctian is aligned to c.
the BWST, piggyback operation is NOT allowed d.
MU/HPl piggyback operation is ONLY to be utilized when directed to be initiated by DB-OP-02000 QUESTION: 055 (1.00)
DB-OP-02523, CCW Malfunctions, directs the operator to MANUALLY TRIP certain SFAS "
modules in the event of a totalloss of CCW pumps to defeat the HIGH temperature and LOW flow CCW pump TRIPS. The results of tripping ONE side's SFAS modules is that CTMT vecuum reliefs for that side will and SG level setpoint will automatically
'
a.
OPEN; INCREASE from 49" to 124".
b.
CLOSE: DECREASE from 124" to 49".
c.
OPEN; DECREASE from 124" to 49" d.
CLOSE; INCREASE from 49" to 124".
.
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SENIOR REACTOR OPERATOR Page 33
- QUESTION: 056 (1.00)
Given the following conditions:
The plant is at 48% power
-
The Main Turbine has received a valid trip signal
-
The Turbine Stop Valves and Turbine Control Valves have NOT closed
-
The Reactor Operator has tripped the reactor and has depressed "SFRCS
-
Manual Actuation : witches" (Start AFPT and Isolate SG pushbuttons), per DB-OP-02000, Step 4.2 The operator is directed to initiate SFRCS for these conditions because SFRCS actuation will....
a.
prevent the Main Turbine from overspeeding.
b.
prevent a reactor coolant system overcooling transient, c.
provide Auxiliary Feedwater flow to maintain steam generator levels at 124 inches, d.
ensure total steam flow stays within the capacity of the Turbine Bypass Valves (TBV).
"
QUESTION: 057 (1.00)
The following plant conditions exist:
The reactor is operating at 100% rated power
-
Instrument Air pressure is rapidly DECREASING
-
NO SFAS actuation signals have been received
-
. Whk:h ONE of the following describes the effect of a COMPLETE LOSS of instrument Air on the Makeup and Pressurizer Level Control Systems? (Assume NO operator actions are taken.)
a.
MU Tank level willINCREASE off scale HIGH causing the lifting of the letdown line relief valve, b.
MU Tank level will DECREASE and the running MU Pump will f ail when the MU Tank is drained, c.
MU Tank level will DECREASE and INCREASED injection flow will INCREASE pressurizer level.
d.
MU Tank level willlNCREASE causing MU 11 to automatically align to the Clean Waste Receiver Tan _ - _ _ _ _ - _ _ _ _
,
.
SENIOR REACTOR OPERATOR-Page 34 QUESTION: 058 (1.00)
At 1105, while operating at 70% reactor power, a regulating control rod becomes misaligned l
by 7% from the group average due to a malfunction. At 1125 (the sama day), the rod is i
ready to be recovered. Which ONE of the following power levels is the maximum at which the rod may be realigned?
a.
60 %
b.
70 %
c.
80 %
d.
100 %
QUESTION: 059 (1.00)
- Given the following conditions:
The plant is operating at 55% power
"
-
A!) systems are operating as designed
-
Which rod position indicating (PI) system is selected on the Control Room Rod Position-Indication Panel for normal monitoring and why is this one selected?
ABSOLUTE - allows immediate verification that all control rods are fully a.
inserted on a reactor trip b.
RELATIVE-- allows continuous monitoring of sequence fault conditions during control rod motion ABSOLUTE-allows contlauous monitoring of sequence fault conditions during c.
control rod motion d.
RELATIVL -- allows immediate verification that all control rods are fully inserted on a reactor trip
.
.
,
s
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l SENIOR REACTOR OPERATOR Page 35 QUESTION: 060 (1.00)
Initial conditions:
100% RTP
-
- Groups 1-6 100% withdrawn
-
Group 7
' 92% withdrawn
-
-APSRs 30% withdrawn
-
An event occurs and the following alarms / conditions are observed:
(5 2 El CRD ASYMMETRIC ROD
-
(51-3) CRD LCO
-
(5 3 E) CRD SAFETY RODS NOT WITHDRAWN
-
Control Rod Drive Mechanism Position Indication Panel: ASYM ALARM iight LIT
-
NO zero percent lights LIT
-
Rod Control Indication Panel: AUTO INHIBIT
-
Which ONE of the following sections of DB-OP 02516,CRD Malfunctions, should be entered?
a, Dropped rod
.
b.
Misaligned rod c.
Undesired rod motion d.
Rod Position Indication malfunction
!
QUESTION: 061 (1.00)
,
Which ONE of the following incore conditions meets the requirements for entry into DB-OP-02OOO, Section 9, inadequate Core Cooling? '
a.
With BOTH channels available, four (4) working incore detectors display a NEG
' MARGIN light, b.
-With BOTH channels available, three (3) working incore detectors indicate a
- NEG MARGIN light.
c.
With only ONE channel available, two (2) working _incore detectors display a NEG MARGIN light.
d.
With only ONE channel available, five (5) working incore detectors indicate a NEG MARGIN ligh.
.
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SENIOR REACTOR OPERATOR Page 36
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,
'
OUESTION: 062 (1.00)
Which ONE of the following sets of conditions would require an attempt to bump Alt four (4)
RCPs by jumpering interlockst RCS pressure Incore Thermocouples
,
(PIRC2A4)
(ave)
i a.
400 psig 440 deg F
2 b..
1000 psig 680 deg F c.
1600 psig 130 deg F d.
2200 psig 1280 deg F
r QUESTION: 063 (1,00)
i-Which ONE of the following describes WHY prolonged RCP operation is to be MINIMlZED with
RCS pressures below 400 psig?
-
)_
a.
To minimize excessive seat wear b.
To prevent improper seal staging
,
,
c.
To prevent excess seat bleed-eff flow
d.
To minimize excessive thrust bearing wear
.
N
-
,
SENIOR REACTOR OPERATOR Pa e 37 w
i OUESTION: 004 (1.00)
The plant is at 2% power during a plant shutdown whers the reactor operator reports that the SHUTDOWN MARGIN has been calculated to be 0.98%. The proper operator action that must be taken is to initiate
,
n.
boration with the Makeup pumps taking a suction on the BWST at a MINIMUM of 25 gpm b.
boration with the Clean Waste Booster pumps taking a suction on a Concentrates Storage Tank at a MINIMUM of a 95 opm c.
a batch addition directly to the RCS from the BorM Acid Mix Tank (BAMT) with a MINIMUM concentration of 7875 ppm at a MINIMUM of 95 gpm d.
a batch addition with the Boric Acid pumps taking a suction on the Boric Acid Addition Tanks (BAATs) with a MINIMUM concentration of 7875 ppm at a MINIMUM of 25 gpm i
QUESTION: 005 (1.00)
Following a reactor trip, six control rods are stuck out. The Reactor Operator started boration per DB OP-02000, "RPS, SFAS, SFRCS Trip or SG Tube Rupture." Termination of the boration is allowed when it has been verified that:
,
a.
SHUTDOWN MARGIN hac been calculated to be 1.15%.
b.
Source Range startup rate reaches negative 1/3 dpm.
c.
Source Range counts on all channels are less than 100 cps.
d.
Reactor Coolant Systern boron concentration reaches 2000 pp.__.
__.
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_ _
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--
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SENIOR REACTOR OPERATOR Page 38 OUESTION: 000 (1.00)
The following plant conditions exist:
The plant is at 100% power, normal system alignment
-
Annunciator 11.'bA, CCW GURGE TK LVL LO is in alarm
-
.
The RO S has opened DW 2643, DEMIN WATER TO CCW SURGE TANK
.
LI 1402 and 1403 {CCW SURGE TANK LEVEL SIDE 1 AND 2) are on a
-
downward trend if this trend continues, tnen the:
a.
CTMT non essential header will be isolated at a surge tank level of 35 inches.
b.
Aux. Oldg non-essential header will be isolated at a surge tank level of 35 inches.
c.
Train 2 essential header will be isolated at a surge tank le'<el of 45 inches, d.
makeup pump header will be isolated and the makeup pumps will trip at a surge tank level of 45 inches.
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l
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_
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-
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SENIOR REACTOR OPERATOR Page 39 QUESTION: 067 (1.00)
Af ter attempting to TRIP the reactor in the Control Room, the Nls still read 50%. Which ONE of the following combinations represents the preferred order in which the control rod drives should be MANUALLYdeenergized, according to DB OP 02000,lmmediate Operator Actions?
MANUALLY TRIP the three (3) reactor trip breakers in the low voltage a.
-
i switchgear rooms, MANUALLY DEENERGlZE the CRD System by tripping BE 211 and
-
s BF 211.
MOMENTARlLY DEENERGl2E 480 VAC Unit Substations E 2 and F 2
-
simultaneously.
b.
MANUALLY TRIP the three (3) reactor trip breakers in the low voltage
-
switchgear rooms, j
MOMENTARILY DEENERGlZE 480 VAC Unit S abstations E 2 and F 2
-
simultaneously.
MANUALLY DEENERGlZE the CRD System by tripping BE 211 and"
-
BF 211, MOMENTARILY DEENERGlZE 480 VAC Unit Substations E 2 and F 2 c.
-
simultaneously.
MANUALLY TRIP the three (3) reactor trip breakers in the low voltage
-
switchgear rooms.
MANUALLY DEENERGlZE the CRD System by tripping BE 211 and
-
BF 211.
d.
MOMENTARILY DEENERGlZE 480 VAC Unit Substations E 2 and F 2
-
simultaneously.
MANUALLY DEENERGlZE the CRD System by tripping BE 211 and
-
BF 211.
MANUALLY TRIP the three (3) reactor trip breakers in the low voltage
-
switchgear rooms.
_
_
.
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SENIOR REACTOR OPERATOR Page 40 l
i OUESTION: 008 (1.00)
{
The plant is In the process of a power increase to 100% power following repairs to #1 MFP.
At 78% power, the fellowing indications are observed:
SP68, SG 1 MAIN FW CONTROL VALVE is 0% open
-
Tave 586 deg F increasing
-
RCS pressure 2375 psig increasing
-
OTSG #1 level decreasing
-
OTSG #2 levelincreasing slightly
-
Ni power decreasing slightly
-
Based upon the above indications, the crew should:
a.
trip the reactor using either manual pushbutton, then enter Tech. Spec. 3.0.3.
b.
take both FW Loop Demands to HAND along with SP79, SG 1 STARTUP FW CONTROL VALVE.
c.
open the PZR spray valve to mitigate the transient, then perform a normal shutdown.
..
d.
take both FW Loop Demands to HAND along with SP6A, SG 2 MAIN FW CONTROL VALVE and SP7A, SG 2 STARTUP FW CONTROL VALVE.
QUESTION: 069 (1.00)
After a LBLOCA, HPl cooling may be FIRST terminated when flow has been established at >
gpm per line for >
minutes, a.
LPI,1500, 30 b.
MU, 250, 30 c.
LPI,1000, 20 d.-
MU, 250, 20
_ _ _ - - - - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ _
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f SENIOR REACTOR OPERATOR Page 41
i l<
QUESTION: 070 (1.00)
j The following plant conditions exist:
Reactor TRIPPED
'
Containment Pressure 15.0 psia Pressurizer level 90 in, and DECREASING i
RCS pressure 1750 psig and DECREASING RCS temperature 533 deg F j
NI 6, 6, and 7 0%
NI 8 126%
OTSG Pressure Level
850 psig
-
245in.
800 psig 240in.
Which ONE of the following events caused the above conditions?
a.
Steam Line Break b.
Steam Generator Overfeed
"
c.
Steam Generator Tube Rupture d.
Stuck Open Main Steam Safety Valve
..
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-
SENIOR REACTOR OPERATOR Page 42 OUESTION: 071 (1.00)
The plant is experiencing a loss of condenser vacuum. According to DB OP 02518. High Condenser Pressure, at 10 inches HgA, the operator is required to perform the following actions:
1.
TRIP the Reactor.
?
Initiate AFW flow AND isolation of BOTH SGs.-
Theat eft lpe are taken:
in articipation of a loss of the RCPs when the turbine is TRIPPED manually by a.
the operator.
l b. - -
to ensure adequate heat removal from the reactor coolant system when the-condenser is lost, c. -
to MINIMlZE the potential for condenser tube leaks due to the HIGH differential
'
pressure, d.
due to the potential loss of the interconnect water seal between the condenser l
sections.
"
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SENIOR REACTOR OPERATOR Page 43 OUESTION: 072 (1.00)
The following plant conditions exist:
Reactor-TRIPPED from full power Bus A voltage OVAC Bus B voltage OVAC Both EDG's Fall TO WART l
MU pump #2 breaker CLOSED SBODG FAIL-TO START Which ONE of the following describes the correct sequence that should be followed when EDG
I a._
Immediately start EDG #2, energize Bus D1 then immediately energize Bus D2.
b.
TRIP MU Pump #2 breaker and then immediately start EDG #2 to energize Bus D1.
c.
Place MU 19 controller in MANUAL, CLOSE MU 19 and then start EDG #2 and energize Bus D1.
.
d.
Immediately start EDG #2 and energize Bus D1 to restore essential bus loads.
Energize Bus D2 after stripping non essential bus load. -. -
_ -
_. -
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_ -
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SENIOR REACTOR OPERATOR Page 44 i
OUESTION: 073 (1.00)
The following conditions exist:
l Bus lockout has occurred on 4160 VAC Bus C1.
-
l All other plant systems are in a normal lineup.
-
!
Which ONE of the following correctly describes the Instrument AC system essential power alignment?
a.
Y1, NORMALLY powered from inverter YV1,is on ALTERNATE power supplied
'
to YV1 AUTOMAT!CALLY from Voltage Regulator XY1.
,
j b.
Y3, NORMALLY powered from inverter YV3,is on ALTERNATE power supplied
~
i to YV3 AUTOMATICALLY from YAR.
i; c.
Y1 A, NORMALLY powered from Inverter YV1, is on ALTERNATE power supplied to Y1 A MANUALLY from Uninterruptable Bus YAR.
)
d.
Y1 A, NORMALLY powered from inverter YV1, is on ALTERNATE power
!
supplied to YVI AUTOMATICALLY from DC MCC 2.
..
l QUESTION: 074 (1.00)
A fire is in progress at Davis Besse, and the Fire Brigade is manned and on the scene. The Fire Brigade Captain should request Carrall Township Fire Department's assistance if:
a.
the fire is located in a Radiologically Restricted Area and is NOT out.
b.
fire suppression systems have been in operation for 10 minutes and the fire is NOT out, c.
the Fire Brigade was unable to respond to the fire location within the first 10 minutos, d.
the fire is affecting more than one system, and a hose has been laid, even though no water has flowed.
.ipdw ye w * in.i.-
9r t <m y
n.aiwyg sy
=m'ygrg--m-t-= - -
Ap=--v.--++-M=w,-e w,,-
iiv-
u-vvvp.wmry-4*
sMvr
. - _ _ _ _ _ _.. _ _ _ _..... _ _ _ _ _ _. _.. _ _
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S SENIOR REACTOR OPERATOR Page 46 OUESTION: 076 (1.00)
The following plant conditions exist:
The plant is in MODE 3, preparing to startup the reactor.
-
A LARGE fire is in progress in Mechanical Penetration Room 1.
-
Fire fighting operations are in progress.
-
Off site fire department assistance has been called and is enroute to the plant,
-
if a loss of ALL off site power occurs while efforts to fight the fire are in progress, the water for the Station Fire Protection System will be supplied from the
- a.
Jockey Fire Pump with suction on the FWST b.
Diesel Fire Pump (DFP) with suction on the intake structure
.
c.
Electric Fire Pump (EFP) with suction on the intake structure d.
Fire department pumper connection with backup lines taking suction on the FWST
..
QUESTION: 076 (1.00)
In the event that the control room must be evacuated, which ONE of the following actions has the LOWEST priority to be accomplished?
a.
initiate SFRCS b.
Isolate letdown c.
Start the standby Makeup Pump d.
Sound the INITIATE EMERGENCY PROCEDURES alarm and announce control room evacuation i
-
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l SENIOR REACTOR OPERATOR Page 40 OUESTION: 077 (1.00)
A serious CTRM fire has occurred requiring evacuation of the CTRM. Local plant control was established at the Auxiliary Shutdown Panel (ASP). The SS is attempting to determine the presence of primary to secondary heat transfer from the ASP. Which ONE of the following actions should be performed to allow him to make this determination?
a.
Raise the setting on AFP 11 governor from the ASP and observe Thot Indications for a decrease, b.
Raise the setting on AFP 11 governor from the ASP and observe Tcold indications for a decrease.
c.
Direct the RO S to locally throttle ICS118, OTSG 11 ATMOSPHERIC VENT VALVE and observe Tcold indications for a decrease, d.
Direct the RO S to locally throttle open ICS118, OTSG 11 ATMOSPHERIC VENT VALVE and observe Thot indications for a decrease.
QUESTION: 078 (1.00)
"
Which ONE of the following conditions would affect CTMT INTEGRITY?
a.
MS 100, #2 SG MS lsolation Valve is CLOSED.
b.
A CTMT isolation Velve f ailed its Locai Leak Rate Test (LLRT).
c.
A #2 side MS Safety Valve is removed with NO secondary side openings on #2 OTSG.
d.
The door to the CTMT Vessel annulus is removed in #4 Mechanical Penetration Room.
.
-._
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SENIOR REACTOR OPERATOR Page 47
!
QUESTION: 079 (1.00)
Given the following conditions:
i A station blackout has occurred
-
l High Pressure injection and Auxiliary Feedwater have failed to operate
-
i The reactor core has become partially uncovered
-
Core damage may be unavoidable if the crew identifies that:
'
'
a.
T sat monitors indicate O deg F
'
i b.
there is no indication of natural circulation
,
c.
RCS pressure / temperature is in Region 4 of Figure 2 for enadequate Core Cooling l
d.
the self powered neutron detectors indicate 100% reactor power 3 minutes
after the above transient i
..
QUESTION: 080 (1.00)
Plant conditions are as follows:
A small break LOCA has occurred.
-
HPl Pump 12 AUTOMATICALLY started.
-
HPl Pump 1 1 CANNOT be started.
-
Which ONE of the following is the reason that the operator must balance the injection line flows in HPl Train 2?
a.
Ta prevent a break in1 rain 1 from robbing most of the injection flow from Train 2.
b.
To prevent excessive thermal stress to the reactor vessel wall near the injection line with the higher flow, c.
To prevent excessive thermal stress to the injection nozzle in the Train 2 injection line with the higher flow, d._
To prevent a break in one of the Train 2 injection lines from robbing r.iost of the injection flow from the intact injection lin.
-
_
-
.
- _ _
SENIOR REACTOR OPERATOR Page 48
,
QUESTION: 081 (1.00)
Chemistry reported that a high activity condition existed in the RCS based upon their latest sample. You directed the control room operator to check the failed fuel datector RI 1998 for proper response. Which ONE of the following describes how the operator should perform the check on RI 1998 in this situation?
'
a.
Depress the check source button on the digital radiation rnonitor for RI 1998.
b.
Perform DB SC-04187, Daily Check of the Radiation Monitoring System, for RI 1998, c.
Perform DB SC-03200, Shift Channel Check of the Radiation Monitoring System, for RI 1998.
d.
VERIFY power available to RI 1998 and that the scale reading corresponds to chemistry sample results.
QUESTION: 082 (1.00)
.
In an attempt to trip the turbine from 24% power, the CTRM operator depressed the EHC - EMERGENCY TRIP pusnbutton. The turbine TRIP did NOT occur and the reactor is still producing 24% power. The CTRM operator should immediately:
a.
stop the Main Turbine Motor Suction Lube Oil Pump by placing His 2400 in LOCKOUT.
b, send an operator to TURN and PULL the local turbine trip lever at the front standard.
c.
initiate AFW flow to BOTH SGs by depressing SFRCS manual actuation switches HIS 6401 and HIS 6402.
d.
open BOTH generator output breakers (ACB 34560 and 34561) to prevent reverse motorization of the main generato SENIOR REACTOR OPERATOR Page 49 QUESTION: 083 (1.00)
Reactor power is at 86% when the controlling RCS pressure channel fails HIGH. With NO operator action, which ONE of the following statements describes plant response?
a.
SFAS will be actuated b.
Pressudzer heaters will energize c.
Actual RCS pressure willINCREASE d.
Pressurizer spray will go full CLOSED QUESTION: 084 (1.00)
During a severe plant transient where Region 2 inadequate core cooling exists, the crew is directed to depressurize the SGs while maintaining secondary side SG levels. SGs are depressurized to obtain a drop in secondary Tsat below existing primary Tsat which is done to
.
..
a.
40 to 60 deg F; increase the primary to secondary DT to promote natural circulation heat transfer b.
40 to 60 deg F; enhance the strength of the heat source to promote natural circulation heat transfer c.
90 to 110 deg F; increase the primary to secondary DT to promote natural circulation heat transfer d.
90 to 110 deg F; enhance the strength of the heat source to promote natural circulation heat transfer
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i SENIOR REACTOR OPERATOR Page 50 i
'
OUESTION: 085 (1.00)
!
The following plant conditions exist:
Reactor Power 95% steady
-
RCS pressure 2150 psig steady
-
.
PRZR level 220 inches and slowly DECREASING
-
j MU TK LEVEL LO, 2 2 C ALARM
-
i MU FLOW HI TRN 2, 2 4 C ALARM
-
,
i RCP seal flow NORMAL
-
No operator actions have been taken i
-
Which ONE of the following abnormal conditions would explain the above conditions?
i l.
a.-
MU 32 failed open b.
MU pumps are tripped
j c.
Normal MU injection line leak
,
j d.
Loss of suction to running MU pump
..
d j
QUESTION: 086 (1.00)
The plant is in MODE 5 with Decay Heat Removal (DHR) Pump 1 providing RCS flow. The
'
control room operator noticed erratic flow on DHR Pump 1 flow indicator FYl DH 28. If DHR
Pump 1 is cavitating, this could also be indicated by:
.
a.
motor amps erratic, pressure erratic, and pump noise decreasing
'
b.
motor amps erratic, pressure erratic, and pump noise increasing
,
c.
motor amps decreasing, pressure erratic, and pump noise increasing, d.
motor amps erratic, pressure increasing, and pump noise increasing
-.
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SENIOR REACTOR OPERATOR Page 51 QUESTION: 087 (1.00)
A plant transient at 70% power has occurred, which resulted in a drop in Tave of 7 deg F.
Tave has been recovered over the past 10 minutes and is again at setpoint. At the point-where the heatup stops, RCS pressure will:
a.
remain constant as the heaters compensate for PZR spray flow.
b.
decrease because of mixing of the relatively cold RCS liquid in the PZR increase because the RCS pressure instruNnt will see a higher DP with Thot c.
increasing after the heatup stops d.
decrease because the RCS pressure instrument is temperature compensated and Thot will be decreasing after the heatup stops GUESTION: 088 (1.00)
Given the following conditions:
A plant startup is in progress.
-
"
Safety rod Groups 14 are at their OUT LIMIT.
-
The Reactor Operator (RO) commenced withdrawing Group 5 rods and has now
-
stopped.
What will be the indication that sub-critical multiplication is occurring?
a.
A constant, positive start up rate with increasing source range counts, b.
A constant, negative start up rate with source range counts decreasing to a previous level, c.
Start up rate leveling off at zero DPM with source range counts increasing and leveling off at a higher value than previously observed, d.
Start up rate leveling off at zero DPM with intermediate range power level increasing and leveling off at 5.0 E 5 amps.
_
SENIOR REACTOR OPERATOR Page 52 QUESTION: 089 (1.00)
A plant startup is in progress at 6% powe,. Intermediate Range Nuclear Instrument NI 3 has failed DOWNSCALE. Which ONE of the following describes when the Source Range nuclear instrument detector high voltage power supply will be AUTOMATICALLY cut off as reactor power is INCREASED?
a.
When NI 3 fell below 5 X 1010 amps b.
Exceeding 5% power indication on Nl.5 and 7.
c.
Exceeding 10% power indication on NI 6 and 7.
d.
Source Range nuclear instruments wili not automatically deenergize when Ni 3 is reading downscale.
QUESTION: 090 (1.00)
The following plant conditions exist:
..
100% power
-
RE 600 (MS Line 2 RE) in alarm
-
PZR level decreasing (currently 210")
-
Both makeup pumps are running
-
MU 2B, Letdown Isolation, is closed
-
MU 32, PZR Level Control, is fully open
-
f Which ONE of the following describes the correct course of action and reason for those actions under these conditions?
a.
Reactor TRIP; MINIMlZES the duration of offsite releases, b.
CONTROLLED shutdown; avoids INCREASING the RCS leakage rate, c.
CONTROLLED shutdown; avoids lifting the main steam safety valves, d.
Reactor TRIP; MINIMlZES the amount of secondary side contaminatio _ _ _ - _ - _ - _ _ _
_
..
.
SENIOR REACTOR OPERATOR Page 53 OVESTION: 091 (1.00)
An SG Tube Rupture is in progress on OTSG #1. All RCPs are running. The RO S has reported that OTSG #1 level has reached 225 inches and is stillincreasing. The CTRM SRO should direct the RO S to:
Initiate AFW flow and Isolation of both SGs to prevent a high level SFRCS trip.
a.
b, trip all four RCPs to minimize the driving head for flow into the ruptured OTSG.
c.
Increase his steaming rate to achieve up to a 235 deg F/hr. rate down to a Tave of 500 deg F.
d.
Increase his steaming rate to achieve up to a 235 deg F hr. rate down to a Tave of 520 deg F.
QUESTION: 092 (1.00)
The following plant conditions exist:
..
The plant is at 28% power.
-
Primary to secondary leakage has been diagnosed.
-
Both Makeup Pumps are running.
-
A decision has been made to shutdown the plant.
-
Which ONE of the following is consistent with the indications of a SG TUBE RUPTURE?
Prior to the power reduction, letdown is isolated.
a.
-
Makeup Tank levelis dropping at 5 inches / minute,
-
b.
Prior to the power reduction, letdown is isolated.
-
PZR level is dropping at 5 inches / minute.
-
During the power reduction, letdown is isolated.
c.
-
Makeup Tank level is dropping and the rate of drop slows when the
-
power ramp rate slows, d.
During the power reduction, letdown is isolated.
-
PZR level la dropping and pressurizer level starts to recover when the
-
power ramp rate slows,
_
__
__ ___
_.
__
__ _ _ _ _ _. _ _ -.
.
_._
.
SENIOR REACTOR OPERATOR Page 54 OUESTION: 003 (1.00)
The plant has experienced a steam generator tube leak of GREATER than 50 gpm. N0 reactor coolant pumps are available. A natural circulation cooldown is in progress. The RO P was directed to REDUCE RCS pressure to maintain a MINIMUM subcooling margin during the cooldown. A minimum subcooling margin:
a.
prevents a pressurized thermal shock condition when HPlis required for RCS makeup.
b.
will reduce the differential pressure between the RCS and the leaking steam
'
generator, c.
allows a greater cooldown rate by reducing the steam generator tube to shell therrnal stresses.
d.
ensures normal makeup will be capable of maintaining pressurizer level greater than 100 inches.
i QUESTION: 094 (1.00)
'
i The following plant conditions exist:
- 1 AFP is tagged out for maintenance on its governor.
-
The reactor has been tripped because of a large fire in the West Condenser Pit
-
565' elevation.
The fire has been put out,
-
The Fire Captain reports that the fire was limited to the Condensate Pumps and
-
- 2 MFP.
- 2 OTSG, which has an indicated SU level of 10 inches, should be fed using:
a.
the MDFP in its normal 100% power lineup b.
the SUFP, provided flowrote is maintained < 300 gpm.
c.
either MFP, provided flowrate is maintained < 1000 gpm d.
the MDFP in its MFW lineup, provided flowrate is maintained > 800 gpm.
_
,
..
. _. _. _.
.. _. _ _ _...... _ _ -
- _ _
_ _ _ _ _ _ _.. _ _ _ _ _
-
. _ _ _
-
i t
,
-
i SENIOR REACTOR OPERATOR Page55 i
!
OUESTION: 095 (1.00)
}
After operating at full power for three months you note the following:
i
'
!
1.
>
l
- DC BUS 1 TRBL,16 E i
I 2.
VOLTAGE INDICATORS:
'
!
j
+ 125 VDC PANEL D1P, El 6271 reads 0 VDC i
All other indicators read NORMAL values
!j'
3.-
BREAKER STATUS LIGHTS:
l
' A bus load breaker Indicators are OFF
'
j
+ C1 bus breaker _ indicators are OFF
Which ONE of the following represents a correct AUTOMATIC system response to these i
conditions? (assume NO operator actions are taken.)
.
.
a.
Annunciator 141 C, ICS 24 VDC BUS TRIP will alarm.
b.
RCP's 1 1 and 2 2 will automatically TRIP on loss of DC control power.
!
'
c.
The ICS will RUNBACK the unit to 55% power due to LOW level in the Demerators.
.
d.
The #1 AFP will AUTOMATICALLY. START and feed the OTSGs to 49" on the i
!
!
i k
!
!
'
SENIOR REACTOR OPERATOR Page 56 OUESTION: 096 (1.00)
The following plant conditions exist:
The plant is in MODE 5 with the CTMT Equipment Hatch removed and
-
fuel shuffling in progress in the SFP.
CTMT Purge is running on the Penetration Rooms.
-
Area Monitors RE 8440 and 8447 have just gone into alarm (ALERT and HIGH).
-
SFAS is NOT in Sb':'iown Dypass.
-
RP personnel have increased dose rates on their dose rate monitors.
-
The correct CTRM reaponse to these indications is to:
seund the CTMT Evacuation alarm and evacuate personnel from the Auxiliary l
a.
Building and CTMT b.
realign CTMT Purge to CTMT to aid St. tion EVS Trains 1 and 2 in removing any harmful atmosphere
,
.
c.
sound the initiate Emergency Procedures alarm and evacuate personnel from the
!
Auxiliary Building and CTMT
[
.
d.
manually initiate Station EVS Trains 1 and 2 since they are prevented from running under current plant conditions i'
QUESTION: 097 (1.00)
!
After receiving the INSTR AIR HDR PhESS LO alarm at 100% power, the Reactor Operator reports that instrument air pressure (using Pl810) reads 72 psig and the secondary plant appears STABLE. Which ONE of the following sets of actions is required to be performed?
a.
Manually TRIP the reactor and initiate AFW flow and isolation of both SGs.
b.
Rapidly DECREASE power per DB-OP-02504 untilinstrument air INCREASES to approximately 90 psig, c.
Dispatch operators to locate the cause of excessive air demand and maintain reactor power at the present level, d.
START the standby Station Air Compressor and thu..nergency Instrument Air i
.
Compressor, and perform a Rapid Shutdown per DB OP-02504, Rapid Shutdown.
...
.
.
..
.
_
... -
- -.-. _ _ _
... -. -
_ - -.
.-
-
- - -. - - -
l l
l l
,
SENIOR REACTOR OPERATOR Page 57
.
QUESTION: 098 (1.00)
l The following plant conditions exist:
'
\\
-
!
Annunciator Alarms:
,
(41 El PZR LO LVL HTR TRIP
-
(4 2 El PZR LVL LO i
l
-
l Other Symptoms:
l Pressurizer Heaters are deenergized.
-
l L-20" indicated on Pressurizer levelindication LRS RC14 and DECREASING
-
l RCS pressure is INCREASING
-
Which ONE of the following immediate operator actions is required by DB OP 02513, Pressurizer System Abnormal Operation for the above plant symptoms?
a.
Close RC 11, PORV BLOCK.
b.
CLOSE RC 2, PRESSURIZER SPRAY
..
c.
Place MU 32 in HAND, adjust demand to obtain desired makeup flow.
d.
START the standby Makeup Pump.
l
!
. _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _
SENIOR REACTOR OPERATOR Page 58 l
j l
QUESTION: 099 (1.00)
l The following plant conditions exist:
l The plant is in MODE 6.
-
'
Core off load is in progress.
'
-
Refueling Canallevelis noticed to be dropping at the rate of 1 foot / minute.
-
The Fuel Handling Director should:
a.
direct installation of the gate betwoon the SFP and the Cask Fill Pit.
b.
direct all personnel to abandon the bridges and evacuate the area before canal level drops to 19 feet (597' elevation).
c.
order the Fuel Storage Bridge Operator (FSBO) to place his spent fuel assembly in the nearest transfer basket and leave the basket vertical.
d.
order the Main Bridge Operator (MBO) to place his spent fuel assembly in the deep end of the canal 19 feet below the normal ZZ tape reading.
..
-. ~... _ -..
-. -. _
-
. - - -
_ _ - -
. _ -.
- -
..
,
!
!
,
+
SENIOR REACTOR OPERATOR Page 59 i
,
'
j QUESTION: 100 (1.00)
.l The following plant conditions exist:
SU Transformer 01 is tagged out.
-
An SFAS Level 2 Actuation has occurred.
-
The Emergency Diesel Generators (EDGs) have been BLOCKED and SHUT
-
DOWN.
The EDG Lockout Relays have been reset.
j
-
Which ONE of the following describes the response of the EDGs to a subsequent overpressure trip of SU Transformer 02?
,
j.
j a.
Fall to AUTOMATICALLY start but can be MANUALLY started and tied to
associated buses, i
b.
AUTOMATICALLY start and tie to associated buses due to an undervoltage condition on associated buses.
!
c.
Fall to AUTOMATICALLY start but can be MANUALLY started; however BLOCKED from tying to associated buses.
..
.
d.
AUTOMATICALLY start but #2 EDG la BLOCKED from tying to its associated
!
bus due to th" condition of SU Transformer 02.
!,
i
I
!
.
W l
( * * * * * * ' ' ' ' END OF EXAMINATION * * * * * * * * * * )
,
.,
J e
.v,r-
, -,, -.
---r.--
,...... - - -
ym-.-
,,,. -
w-,
r-%-4
..,,+ -- - - -e--,e
--.,-,
-rw,,-.
-.,
.m-
--
-
-.www'--
t--m>--.
SENIOR REACTOR OPERATOP Page 60 ANSWER: 001 (1.00)
ANSWER: 007 (1.00)
ANSWER: 013 (1.00)
c.
d b
REFERENCE:
REFERENCE:
REFERENCE:
10CFR55 DB OP 00015 (6.13) R4 DB FP 00009 (0.5.1) R5 NT-OT-07001 194001 K102..(KA's)
194001 K116..(KA's)
194001 A102..(KA's)
ANSWER: 008 (1.00)
ANSWER: 014 (1.00)
ANSWER: 002 (1.00)
b d
a REFERENCE:
REFERENCE:
REFERENCE:
DB HP 01100 (6.3) R2 DB OP 00004 (6.1) R2 DB-OP 00000 (6.3.2) R3 194001 K103..(KA's)
194001 A107..(KA's)
194001 A103..(KA's)
ANSWER: 009 (1.00)
ANSWER: 015 (1.00)
ANSWER: 003 (1.00)
b a
d REFERENCE:
REFERf.NCE:
REFERENCE:
DB HP 01201 (6.2) R3 T.S. 3.4.8, Bases DB OP-00003 (6.2.5) R2 194001 K104..(KA's)
194001 A114..(KA's)
194001 A106..(KA's)
ANSWER: 010 (1.00)
ANSWER: 016 (1.00)
ANSWER: 004 (1.00)
d b
b RECERENCE:
REFERENCE:
"
REFERENCE:
DB-OP 00000 (6.20.1) R3 DB.OP 02504 (4.1.10) R2 DB NE-06201 R3 DB-PF 06703(C.C.6.9)R3 DB-NE-06202 (5.2) R1 194001 K105..(KA's)
194001 A115..(KA's)
194001 A108..(KA's)
ANSWER: 011 (1.00)
c ANSWER: 017 (1.00)
ANSWER: 005 (1.00)
REFERENCE:
a b
D B - O P 010 0 0 REFERENCE:
REFERENCE:
(2.1.2/2.1.4) R1 RA-EP-02810 (6.3.1) RO E PLAN, 5.2.1.c 194001 K107..(KA's)
194001 A116..(KA's)
194001 A116..(KA's)
ANSWER: 012 (1.00)
ANSWER: 018 (1.00)
ANSWER: 006 (1.00)
a e
c REFERENCE:
REFERENCE:
REFERENCE:
DB-OP-06210 (2.2.6) RO DB-OP 06403 (3.4.1) R1 DB-OP-00016 (6.1.1) R2 194001 K115..(KA's)
001000K105..(KA's)
194001 K101..(KA's)
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ -
....
SENIOR REACTOR OPERATOR Page 61 ANSWER: 019 (1.00)
ANSWER: 025 (1.00)
ANSWER: 031 (1.00)
b c
d REFERENCE:
REFERENCE:
REFERENCE:
DB OP-06401 (4.9.4) R2 DB OP-06912 (4.21.3) R1 DB OP 02000(Table 1)R4 DB OP 02515 (4.1.2) R1 015000A202..(K.A's)
059000A201..(KA's)
003000K305..(KA's)
ANSWER: 026 (1.00)
ANSWER: 032 (1.00)
ANSWER: 020 (1.00)
c d
b REFERENCE:
REFERENCE:
REFERENCE:
T.S. Table 3.31 DB OP 02000(Spec Rule DB OP-02515 (4.1.3) R1 T.S. 3.0.3 2.1.2) R4 004000A205..(KA's)
015000G005..(KA's)
059000G014..(KA's)
ANSWER: 021 (1.00)
ANSWER: 027 (1.00)
ANSWER: 033 (1.00)
d d
c REFERENCE:
REFERENCE:
REFERENCE:
OS 002, Sh.1 (CL 2) R16 B&DD (9.9), R7 DB OP-06233 (3.3.31 R4 004010K403..(KA's)
017020A202..(KA's)
061000A101..(KA's)
ANSWER: 022 (1.00)
ANSWER: 028 (1.00)
ANSWER: 034 (1.00)
b d
a
-
REFERENCE:
REFERENCE:
REFERENCE:
DB OP 06900 (4.62) R4 USAR 6.2.2.2.1 USAR 9.2.7.3 USAR 6.2.2.2,2 013000K101..(KA's)
T.S. 3.6.2.1 001000K101..(KA's)
T.S. 3.6.2.2 ANSWER: 023 (1.00)
ANSWER: 035 (1.00)
a 022000G007..(KA's)
d REFERENCE:
REFERENCE:
T.S. 3/4.5 Bases ANSWER: 029 (1.00)
DB-OP-02538 (4.1.1) RO c
REFERENCE:
9B OP-02000 (Att. 7) R4 063000K302,,(KA's)
ANSWER: 024 (1.00)
c 026020K403..(KA's)
ANSWER: 036 (1.00)
REFERENCE:
d DB OP-02516 (4.1.4) R3 ANSWER: 030 (1.00)
REFERENCE:
a DB OP 06412 (4.12.2) R3 014000A104..(KA's)
REFERENCE:
DB OP 06221 (2.2.5) R1 072000G012..(KA's)
056000A204..(KA's)
_
SENIOR REACTOR OPERATOR Page 62 ANSWER: 037 (1.00)
ANSWER: 042 (1.00)
ANSWER: 048 (1.00)
c c
c REFERENCE:
REFERENCE:
REFERENCE:
-
l DB OP-02000 (Fig.1) R4 M5331761 T1 DB OP 06311 (2.1.14)RO
!
002020K506..(KA's)
016000A201..(KA's)
062000G010..(KA's)
.
!
ANSWER: 038 (1.00)
ANSWER: 043 (1.00)
ANSWER: 049 (1.00)
c b
d
,
REFERENCE:
REFERENCE:
REFERENCE:
DB-OP-02000 (Spec. Rule DB OP-02014 (Att.14 2-DB OP-02521 (3.2.2) R1 f
,
'
2.4.2) R4 D) R1
064000G008..(KA's)
l 006000A106..(KA's)
016000K201..(KA's)
ANSWER: 050 (1.00)
,
ANSWER: 039 (1.00)
ANSWER: 044 (1.00)
c
'
d c
REFERENCE:
!
REFERENCE:
REFERENCE:
ICM DSM 87 25083
DB-OP 02004 (Att. 41 D)
DB-OP 02000 (Table 3)
( 0 P S S Y S-l E O 8 R3 R4 BACKGROUND FOLDER)
'
OS 001 A, Sh. 2 (CL 2)
PCAOR 87 0071 i
R10 028000A202..(KA's)
DB OP 02513 (2.1) R2 073000K301..(KA's)
DB-OP 02513 Operator ANSWER: 045 (1.00)
"
!
Aid b
ANSWER: 051 (1.00)
!
REFERENCE:
d I
010000A203..(KA's)
DB OP-02527 (2.4) R2 REFERENCE:
!
OS-020, Sh. 2 (CL 9) R17 j
ANSWER: 040 (1.00)
033000G015..(KA's)
j a
075000A401..(KA's)
REFERENCE:
ANSWER: 046 (1.00)
'
DBOP-06003 a
ANSWER: 052 (1.00)
l (2.2.3/4.5.9) R3 REFERENCE:
c DB OP-02000 (6.13) R4 REFERENCE:
011000G010..(K A's)
OS-0198, Sh.1 R10 j
035010K301..(KA's)
ANSWER: 041 (1.00)
j d
ANSWER: 047 (1.00)
079000K401..(KA's)
REFERENCE:
- REFERENCE:
ANSWER: 053 (1.00)
.
DB-OP 02525 (2.2) R1 b
i 012000G014..(KA's)
REFERENCE:
039000K508..(KA's)
DB-OP 02529 (2.1.2) R2 086000G008..(KA's)
i
.
.
_
.
.
-.
_.__ _ _ _ _ _, _ _ _ _ _... _ - -
_, -
.
SENIOR REACTOR OPERATOR Page 63 ANSWER: 054 (1.00)
ANSWER: 060 (1.00)
ANSWER: 066 (1.00)
c b
a REFERENCE:
REFERENCE:
REFERENCE:
DB OP-02000 (Spec. Rule DB-OP-02516 (2.2) R3 DB OP-02523 (4.1.6) R1 2.2/2.3) R4 000005G011..(KA's)
000026A105..(KA's)
005000K408..(KA's)
ANSWER: 061 (1.00)
ANSWER: 067 (1.00)
ANSWER: 055 (1.00)
d c
d REFERENCE:
REFERENCE:
REFERENCE:
DB OP 02000 (5.9) R4 DB-OP 02000 (3.2.2) R4 DB OP 02523 (4.6.13) R1 000011 A115..(KA's)
000029A209..(KA's)
008030A304..(KA's)
ANSWER: 062 (1.00)
ANSWER: 068 (1.00)
ANSWER: 056 (1.00)
d a
b REFERENCE:
REFERENCE:
'
REFERENCE:
DB-OP 02000 (9.16) R4 DB-OP-02000 (1,1) R4 DB-OP-02000 (4.2) R4 000011 A201..(KA's)
B&DD (Step 4.2) R7 000029G011..(KA's)
ANSWER: 063 (1.00)
045050K101..(KA's)
a ANSWER: 069 (1.00)
>
REFERENCE:
c ANSWER: 057 (1.00)
DB OP 06900 (4.1) R4 REFERENCE:
-
c DB OP-06005 (2.2.1.e) R2 DB OP 02000 (Spec. Rule REFERENCE 2)R4 B.OP-C2000 (4.8.3) R4 000015G007..(KA's)
000040A205..(KA's)
078000K302..(KA's)
ANSWER: 064 (1.00)
d ANSWER: 070 (1.00)
ANSWER: 058 (1.00)
REFERENCE:
b b
DB-OP 02510 (4.1.2) R0 REFERENCE:
REFERENCE:
Tech. Spec. 3.1.1.1 DB-OP02000 DB OP-02516 (4.2.6) R3 (2.1.3/2.1.4) R4 000024 A205..(KA's)
000001 K301..(KA's)
000040C011..(KA's)
ANSWER: 065 (1.00)
ANSWER: 059 (1.00)
a ANSWER: 071 (1.00)
a REFERENCE:
b REFERENCE:
DB OP-02000 (4.1.2) R4 REFERENCE:
DB OP-06402 (3.9.18) R1 DB OP-02518 (4.1.1.d)
000024 K302..(KA's)
000051 A202..(KA's)
.-
-
--.
-
. -
-
- - -
-.
. - -
.
.
- - -.
-
'
.
SENIOR REACTOR OPERATOR Page 64 ANSWER: 072 (1.00)
ANSWER: 078 (1.00)
ANSWER: 084 (1.00)
b b
c REFERENCE:
REFERENCE:
REFERENCE:
DB OP 02000 (4.5.4) R4 USAR Ch.
6.2, Table DB OP 02000 (9.4.1) R4 6.2 23 000056K302..(KA's)
Tech Spec Defn 1.8 000009A201..(KA's)
ANSWER: 073 (1.00)
000069G003..(KA's)
ANSWER: 085 (1.00)
c c
REFERENCE:
ANSWER: 079 (1.00)
HEFERENCE:
DB OP 02522 (4.1.6) R1 14) R2 REFERENCE:
DB OP 02000 (9.16) R4 000022A201..(KA's)
000057A219..(KA's)
BDD (Step 9.9) R7 ANSWER: 086 (1.00)
ANSWER: 074 (1.00)
000074K102..(KA's)
b b
REFERENCE:
REFERENCE:
ANSWER: 080 (1.00)
DB OP 02527 (4.3.1) R2 DB OP-02529 (4.1.16) R2 d
REFERENCE:
000025 A207..(KA's)
000067G002..(KA's)
DB OP-02000 (Spec. Rule 2)R4 ANSWER: 087 (1.00)
ANSWER: 075 (1.00)
BDD (Spec. Rule 2) R7 b
b REFERENCE:
-
REFERENCE:
000074K204..(KA's)
CE Steam Tables DB OP 06610 (2.2.14) R1 DB OP 02000 (7.10) R4 OPS SYS 1601 ANSWER: 081 (1.00)
a 000027 K304..(KA's)
000067K102..(KA's)
REFERENCE:
DB OP-02535 (4.1.1) R3 ANSWER: 088 (1.00)
ANSWER: 076 (1.00)
c d
000076G008..(KA's)
REFERENCE:
REFERENCE:
DB-OP-06912 (4.14) R1 DB OP 02508 (4.1.1) R0 ANSWER: 082 (1.00)
b 000032A202..(KA's)
000068K318..(KA's)
REFERENCE:
DB-OP 02500 (3.1.1) R1 ANSWER: 089 (1.00)
ANSWER: 077 (1.00)
c d
000007A202..(KA's)
REFERENCE:
REFERENCE:
DB OP 02505 (4.2.1) R1 DB-OP-02519 (Att.1) R3 ANSWER: 083 (1.00)
a 000033 A209..(KA's)
000068A205..(KA's)
REFERENCE:
DB OP 02513 (2.6.3) R2 000008 A201..(KA's)
,
_ _ -
_-. _ _ _ _ _ _ _ _ _ - _
_
-
_
.
l
-
SENIOR REACTOR OPERATOR Page 65 l
ANSWER: 090 (1.00)
ANSWER: 096 (1.00)
c a
REFERENCE:
REFERENCE:
DB-OP-02000 (8.6.3) R4 DB-OP-02530 (Step 3.2.1)
R2 000037K305..(KA's)
R2 ANSWER: 091 (1.00)
c 000061 A205..(KA's)
REFERENCE:
DB-OP-02000 (8.18.6) R4 ANSWER: 097 (1.00)
a 000037K307..(KA's)
REFERENCE:
DB-OP-02528 (3.1.1) R2 ANSWER: 092 (1.00)
b 000065 A206..(KA's)
REFERENCE:
DB-OP-02000 (1.4) R4 ANSWER: 098 (1.00)
DB-OP-02531 (4.3.1) R1 c
REFERENCE:
000038A202..(KA's)
DB-OP-02513 (3.5.2) R2 ANSWER: 093 (1,00)-
OOOO28K305,.(KA's)
b
-
REFERENCE:
ANSWER: 099 (1.00)
DB-OP-02531 (4.5.6) R1 b
REFERENCE:
000038A215..(KA's)
DB-OP-00030 (Att.1) RO ANSWER: 094 (1.00)
OOOO36K303..(KA's)
a 1FERENCE:
ANSWER: 100 (1.00)
De-OP-02000 (Att.1) R4 b
REFERENCE:
000054K102..(KA's)
DB-OP-06316 (5.8.6) R1
- ANSWER: 095 (1.00)
000056A214..(KA's)
c REFERENCE:
DB-OP 02537 (4.1.1) R1 000058A203..(KA's)
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SENIOR REACTOR OPERATOR Page 66
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J ANSWER KEY l
l --
MULTIPLE CHOICE 023 a
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l 001 c 024 c
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002 a 025 c 003 d 026 c
004 - b 027 d 005 ' b 028 d
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006 c 029 c 007 d 030 a 008 b 031 d 009 b 032 d 010 'd 033 e
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011 c'
034 a 012 a 035 d 013 b 036 d 014 d 037 c
~ 015 ~ a 038 c 016 b 039 d
' 017 a 040 a
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018 c 041 d 019 b 042-c 020 b 043 b 021 d:
044 c 022 b 045 b
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- SENIOR REACTOR OPERATOR Page 67-A' N S W E R_
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KEY MULTIPLE CHolCE 068 a-046 a 069-c 047 a 070 b 048 - c 071 - b 049 d 072 b l
l 050 c 073 c
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051 d 074 b
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052 c 075 b 053 b 076 d-054 c 077 d 055 d 078 b
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056 b 079 c 057 c 080 d 058 b 081 a 059 - a 082 b 060 b 083 a 061 d -
084 c 062 d-085 c 063 a-086 b 064 d-087 b 065 a 088 c l066 a 089 c 067 c 090 c
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- SENIOR REACTOR OPERATOR
- age 68 ANSWER KEY i
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I MULTIPLE CHOICE
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091 c
>
092 b 093 b.
094 ' a 095 c-096 a.
097 a 098 n 099-b 100 b
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